ML071080232

Revisions to Technical Specifications Bases Unit 2 Manual

ML071080232
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Issue date:	04/10/2007
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SSES MANUALJ Manual Name: TSB2 Manual

Title:

TECHNICAL SPECIFICATIONS BASES UNIT 2 MANUAL Table Of Contents Issue Date: 04/09/2007 Procedure Name Rev Issue Date Change ID Change Number TEXT LOES 82 04/09/2007

.Title: LIST OF EFFECTIVE SECTIONS TEXT TOC Ii 04/09/2007 'V

Title:

TABLE OF CONTENTS

/\ >-)

TEXT 2.1.1 2 04/03/2007

Title:

SAFETY LIMITS (SLS) REACTOR CORE SLS TEXT 2.1.2 0 11/18/2002Z

Title:

SAFETY LIMITS (SLS) REACTOR COOLANT SYSTEM, '(RCS) PRESSURE SL IY',

TEXT 3.0 2 10/1,2,200,6i

Title:

LIMITING CONDITION FOR OPERATION (LCO).'APPLICABILITY TEXT 3.1.1 1 03/24/2005

Title:

REACTIVITY CONTROLj SYSTEMS SHUTDOWN MARGIN (SDM)

TEXT 3.1.2 0: 11/18/2002

Title:

REACTIVITY> CONTROL, SYSTEMS REACTIVITY ANOMALIES TEXT 3.1.3 / '> / 1 07/06/2005

Title:

REACTIVITY ',CONTROL SYSTEMS CONTROL ROD OPERABILITY TEXT 3.1.4 3 09/29/2006

Title:

REACTIVITY CONTROL SYSTEMS CONTROL ROD SCRAM TIMES TEXT 3.1.5 1* 07/06/2005

Title:

REACTIVITY CONTROL SYSTEMS CONTROL ROD SCRAM ACCUMULATORS TEXT 3.1.6 2 03/24/2005

Title:

REACTIVITY CONTROL SYSTEMS ROD PATTERN CONTROL Report Date: 04/09/07 Pagel1 Page of of ~8 Report Date: 04/09/07

SSES MANUJAL Manual Name: TSB2 Manual

Title:

TECHNICAL SPECIFICATIONS BASES UNIT 2 MANUAL TEXT 3.1.7 2 04/03/2007

Title:

REACTIVITY CONTROL SYSTEMS STANDBY LIQUID CONTROL (SLC) SYSTEM TEXT 3.1.8 1 10/19/2005

Title:

REACTIVITY CONTROL SYSTEMS SCRAM DISCHARGE VOLUME (SDV) VENT AND DRAIN VALVES TEXT 3.2.1 3 04/09/2007

Title:

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

TEXT 3.2.2 2 04/03/2007

Title:

POWER DISTRIBUTION LIMITS MINIMUM CRITICAL POWER RATIO (MCPR)

TEXT 3.2.3 0 11/18/2002

Title:

POWER DISTRIBUTION LIMITS LINEAR HEAT GENERATION RATE LHGR TEXT 3.3.1.1 3 04/09/2007

Title:

INSTRUMENTATION REACTOR PROTECTION SYSTEM (RPS) INSTRUMENTATION TEXT 3.3.1.2 1 04/09/2007

Title:

INSTRUMENTATION SOURCE RANGE MONITOR (SRM) INSTRUMENTATION TEXT 3.3.2.1 2 04/09/2007

Title:

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

Title:

INSTRUMENTATION FEEDWATER - MAIN TURBINE HIGH WATER LEVEL TRIP INSTRUMENTATION TEXT 3.3.3.1 4 04/09/2007

Title:

INSTRUMENTATION POST ACCIDENT MONITORING (PAM) INSTRUMENTATION TEXT 3.3.3.2 1 04/18/2005

Title:

INSTRUMENTATION REMOTE SHUTDOWN SYSTEM TEXT 3.3.4.1 0 11/18/2002

Title:

INSTRUMENTATION END OF CYCLE RECIRCULATION PUMP TRIP (EOC-RPT) INSTRUMENTATION Page 2 . of 8 Report Date: 04/09/07

SSES MANUAL SManual Manual Name: TSB2

Title:

TECHNICAL SPECIFICATIONS BASES UNIT 2 MANUAL TEXT 3.3.4.2 0 11/18/2002

Title:

INSTRUMENTATION ANTICIPATED TRANSIENT WITHOUT SCRAM RECIRCULATION PUMP TRIP (ATWS-RPT) INSTRUMENTATION TEXT 3.3.5.1 3 07/06/2005

Title:

INSTRUMENTATION EMERGENCY CORECOOLING SYSTEM (ECCS) INSTRUMENTATION TEXT 3.3.5.2 0 11/18/2002

Title:

INSTRUMENTATION REACTOR CORE ISOLATION COOLING (RCIC) SYSTEM INSTRUMENTATION TEXT 3.3.6.1 2 04/03/2007

Title:

INSTRUMENTATION PRIMARY CONTAINMENT ISOLATION INSTRUMENTATION TEXT 3.3.6.2 1 11/09/2004

Title:

INSTRUMENTATION SECONDARY CONTAINMENT ISOLATION INSTRUMENTATION TEXT 3.3.7.1 0 11/18/2002

Title:

INSTRUMENTATION CONTROL ROOM EMERGENCY OUTSIDE AIR SUPPLY (CREOAS) SYSTEM INSTRUMENTATION TEXT 3.3.8.1 2 03/31/2006

Title:

INSTRUMENTATION LOSS OF POWER (LOP) INSTRUMENTATION TEXT 3.3.8.2 0 11/18/2002

Title:

INSTRUMENTATION REACTOR PROTECTION SYSTEM (RPS) ELECTRIC POWER MONITORING TEXT 3.4.1 3 04/09/2007

Title:

REACTOR COOLANT SYSTEM (RCS) RECIRCULATION LOOPS OPERATING TEXT 3.4.2 0 11/18/2002

Title:

REACTOR COOLANT SYSTEM (RCS) JET PUMPS TEXT 3.4.3 1 01/16/2006

Title:

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

TEXT 3.4.4 0 11/18/2002

Title:

REACTOR COOLANT SYSTEM (RCS) RCS OPERATIONAL LEAKAGE Report Date: 04/09/07 Page Page~3 . of of 8

.8 Report Date: 04/09/07

SSES MANUAL SManual Manual Name: TSB2

Title:

TECHNICAL SPECIFICATIONS BASES UNIT 2 MANUAL TEXT 3..4.5 1 01/16/2006

Title:

REACTOR COOLANT SYSTEM (RCS) RCS PRESSURE ISOLATION VALVE (PIV) LEAKAGE TEXT 3.4.6 1 04/18/2005

Title:

REACTOR COOLANT SYSTEM (RCS), RCS LEAKAGE DETECTION INSTRUMENTATION TEXT 3.4.7 1 04/18/2005

Title:

REACTOR COOLANT SYSTEM (RCS) RCS SPECIFIC ACTIVITY TEXT 3.4. 8 1 04/18/2005

Title:

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

- 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 2 05/10/2006

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 STEAM DOME PRESSURE TEXT'3.5.1 3 01/16/2006

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 1 04/18/2005

Title:

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

SYSTEM RCIC SYSTEM TEXT 3.6.1.1 1 04/03/2007

Title:

CONTAINMENT SYSTEMS PRIMARY CONTAINMENT TEXT 3.6.1.2 0 11/18/2002

Title:

CONTAINMENT SYSTEMS PRIMARY CONTAINMENT AIR LOCK Report Date: 04/09/07 Page Page4A of8 ~

of Report Date: 04/09/07

SSES MANUAL Manual Name: TSB2 Manual

Title:

TECHNICAL SPECIFICATIONS BASES UNIT 2 MANUAL TEXT 3.6.1.3 5 "04/03/2007

Title:

CONTAINMENT SYSTEMS PRIMARY CONTAINMENT ISOLATION VALVES (PCIVS)

TEXT 3.6.1.4 0 11/18/2002

Title:

CONTAINMENT SYSTEMS CONTAINMENT PRESSURE TEXT 3.6.1.5 1 10/05/2005

Title:

CONTAINMENT SYSTEMS DRYWELL AIR TEMPERATURE TEXT 3.6.1.6 0 11/18/2002

Title:

CONTAINMENT SYSTEMS SUPPRESSION CHAMBER-TO-DRYWELL VACUUM BREAKERS TEXT 3.6.2.1 0 11/18/2002

Title:

CONTAINMENT SYSTEMS SUPPRESSION POOL AVERAGE TEMPERATURE TEXT 3.6.2.2 0 11/18/2002

Title:

CONTAINMENT SYSTEMS SUPPRESSION POOL WATER LEVEL TEXT 3.6.2.3 1 01/16/2006

Title:

CONTAINMENT SYSTEMS RESIDUAL HEAT REMOVAL (RHR)-'SUPPRESSION POOL COOLING TEXT 3.6.2.4 0 11/18/2002

Title:

CONTAINMENT SYSTEMS RESIDUAL HEAT REMOVAL (RHR) SUPPRESSION POOL SPRAY TEXT 3.6.3.1 2 06/13/2006

Title:

CONTAINMENT SYSTEMS PRIMARY CONTAINMENT HYDROGEN RECOMBINERS TEXT 3.6.3.2 1 04/18/2005

Title:

CONTAINMENT SYSTEMS DRYWELL AIR FLOW SYSTEM TEXT 3.6.3.3 0 11/18/2002

Title:

CONTAINMENT SYSTEMS PRIMARY CONTAINMENT OXYGEN CONCENTRATION TEXT 3.6.4.1 6 08/08/2006

Title:

CONTAINMENT SYSTEMS SECONDARY CONTAINMENT Report Date: 04/09/07 PageS5 Page . of of 88 Report Date: 04/09/07

SSES MANUAL Manual Name: TSB2 Manual

Title:

TECHNICAL SPECIFICATIONS BASES UNIT 2 MANUAL TEXT 3.6.4.2 2 01/03/2005

Title:

CONTAINMENT SYSTEMS SECONDARY CONTAINMENT ISOLATION VALVES (SCIVS)

TEXT 3.6.4.3 4 09/21/2006

Title:

CONTAINMENT SYSTEMS STANDBY GAS TREATMENT .(SGT) SYSTEM TEXT 3.7.1 0 11/18/2002

Title:

PLANT SYSTEMS RESIDUAL HEAT REMOVAL SERVICE WATER (RHRSW) 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 BYPASq SYSTEM TEXT. 3.7.7 0 11/18/2002

Title:

PLANT SYSTEMS SPENT FUEL STORAGE POOL WATER LEVEL TEXT 3.8.1 4 04/18/2006

Title:

ELECTRICAL POWER SYSTE MS AC SOURCES - OPERATING TEXT 3.8.2 0 11/18/2002

Title:

ELECTRICAL POWER SYSTEMS. AC SOURCES - SHUTDOWN TEXT 3.8.3 0 11/18/2002

Title:

ELECTRICAL POWER SYSTEMS DIESEL FUEL OIL, LUBE OIL, AND STARTING AIR Page 6 of .8 Report Date: 04/09/07

SSES MANUAL Manual Name: TSB2 Manual

Title:

TECHNICAL SPECIFICATIONS BASES UNIT 2 MANUAL TEXT 3.8.4 2 12/14/2006

Title:

ELECTRICAL POWER SYSTEMS DC SOURCES - OPERATING TEXT 3.8.5 1 12/14/2006

Title:

ELECTRICAL POWER SYSTEMS DC SOURCES - SHUTDOWN TEXT 3.8.6 1 12/14/2006

Title:

ELECTRICAL POWER SYSTEMS BATTERY CELL PARAMETERS TEXT 3.8.7 3 03/31/2006

Title:

ELECTRICAL POWER SYSTEMS DISTRIBUTION SYSTEMS - OPERATING TEXT 3.8.8 0 11/18/2002

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 ROD POSITION 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 TEXT 3.9.6 0 11/18/2002

Title:

REFUELING OPERATIONS REACTOR PRESSURE VESSEL (RPV) WATER LEVEL TEXT 3.9.7 0 .11/18/2002

Title:

REFUELING OPERATIONS RESIDUAL HEAT REMOVAL (RHR) - HIGH WATER LEVEL Report Date: 04/09/07 Page 2 Page 7 of 8~ Report Date: 04/09/07

SSES MANUAL Manual Name: TSB2 Manual

Title:

TECHNICAL SPECIFICATIONS BASES UNIT 2 MANUAL 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 LEAK 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 11/18/2002

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 1 03/24/2005

Title:

SPECIAL OPERATIONS CONTROL ROD TESTING - OPERATING TEXT 3.10.8 2 04/09/2007

Title:

SPECIAL OPERATIONS SHUTDOWN MARGIN (SDM) TEST - REFUELING Report Date: 04/09/07 Page~

Page a of of ~

8 Report Date: 04/09/07

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

Section Title Revision TOC Table of Contents 11 B 2.0 SAFETY LIMITS BASES Page TS / B 2.0-1 1 Page TS / B 2.0-2 3 Page TS / B 2.0-3 4 Page TS / B 2.0-4 5 Page TS /B 2.0-5 1 Pages B 2.0-6 through B 2.0-8 0 B 3.0 LCO AND SR APPLICABILITY BASES Page TS/B 3.0-1 1 Pages TS / B 3.0-2 through TS / B 3.0-4 0 Pages TS / B 3.0-5 through TS / B 3.0-7 1 Pages TS / B 3.0-8 through TS / B 3.0-9 2 Page TS / B 3.0-10 1 Page TS / B 3.0-11 2 Page TS / B 3.0-1 la 0 Page TS / B 3.0-12 1 Pages TS / B 3.0-13 through TS B 3.0-15 2 Pages TS / B 3.0-16 and TS / B 3'.017 0 B 3.1 REACTIVITY CONTROL BASES" Pages B 3.1-1 through'B 3.1-4., 0 Page TS / B 3.1-5 , 1 Pages TS / B 3.1-6 and\T.S/-,B"3.1-7 2 Pages B 3.1-8 thr-ou/gh B 3.1-13 -. 0 Page TS / B 3.1l-4' \ \/ 1 Pages B 3.1-15 through B 3.1-21 0 PageTS / B\3.1-22 0 Page TS ,B 3.1,723 1 Page TS I B 3.1-24 0 I B3.1-25

&3PageTS 1 Page Ts I B 3.1-26 0

\Page TS / B 3.1-27 1 PageTS / B 3.1-28. 2 Page TS / 3.1-29 1 Pages B 3.1-30 through B 3.1-33 0 Pages TS / B 3.1.34 through TS / B 3.1-36 1 Pages TS / B 3.1-37 and TS / B 3.1-38 2 Pages TS / B 3.1-39 through TS / B 3.1-42 1 Pages TS / B 3.1-43 and TS / B 3.1-44 0 Page TS / B 3.1-45 2 Page TS / B 3.1-46 0 Page B 3.1-47 0 Revision 82 TSIB LOES-1 SUSQUEHANNA - UNIT SUSQUEHANNA -

UNIT 22 TS / B LOES-1 Revision 82

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

Section Title Revision Pages TS / B 3.1-48 and TS / B 3.1-49 1 Page B 3.1-50 0 Page TS / B 3.1-51 1 B 3.2 POWER DISTRIBUTION LIMITS BASES Page TS / B 3.2-1 2 Page TS / B 3.2-2 1 Page TS-/ B 3.2-3 3 Page TS / B 3.2-4 1 Pages TS / B 3.2-5 and TS / B 3.2-6 3 Page TS / B 3.2-7 2 Pages TS / B 3.2-8 and TS / B 3.2-9 4 Pages TS / B 3.2-10 through TS / B 3.2-13 1 B 3.3 INSTRUMENTATION Pages TS / B 3.3-1 through TS / B 3.3-4 1 Page TS / B 3.3-5 2 Page TS / B 3.3-6 1 Pages TS / B 3.3-7 through TS I B 3.3-12 3 Page TS / B 3.3-13 2 Page TS / B 3.3-14 3 Pages TS / B 3.3-15 and TS / B 3.3-16 2 Pages TS / B 3.3-17 and TS / B 3.3-18 3 Pages TS / B 3.3-19 through TS / B 3.3-27 2 Pages TS / B 3.3-28 through TS / B 3.3-30 3 Page TS / B 3.3 - 2 Page TS / B 3.3-32 -4..

Page TS / B 3.3-33 3 Page TS / B 3.3-34 2 Pages TS / B 3.3-34a though TS / B 3.3-34i 0 Pages TS / B 3.3-35 and TS / B 3.3-36 2 Pages TS / B 3.3-37 through TS / B 3.3-43 1 Pages TS / B 3.3-44 through TS / B 3.3-54 3 Pages TS / B 3.3-54a through TS / B 3.3-54e 0 Pages B 3.3-55 through B 3.3-63 0 Pages TS / B 3.3-64 and TS / B 3.3-65 2 Page TS / B 3.3-66 4 Page TS / B 3.3-67 3 Page TS / B 3.3-68 4 Page TS / B 3.3.69 5 Pages TS / B 3.3-70 and TS / B 3.3-71 3 Pages TS / B 3.3-72 and TS / B 3.3-73 2 Page TS / B 3.3-74 3 Page TS / B 3.3-75 2 Pages B 3.3-75a and TS / B 3.3-75b 5 Page B 3.3-75c 4 Pages B 3.3-76 and TS / B 3.3-77 0 SUSQUEHANNA - UNIT 2 TS / B LOES-2 Revision 82

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

Section Title Revision Page TS / B 3.3-78 1 Pages B 3.3-79 through B 3.3-91 0 Pages TS / B 3.3-92 through TS / B 3.3-103 1 Page TS / B 3.3-104 2 Pages TS / B 3.3-105 and TS / B 3.3-106 1 Page TS / B 3.3-107 2 Page TS / B 3.3-108 1 Page TS/ B 3.3-109 2 Pages TS / B 3.3-110 through TS / B 3.3-112 1 Page TS / B 3.3-113 2 Page TS / B 3.3-114 1 Page TS / B 3.3-115 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 / 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 Page TS / B 3.3-128 2 Pages TS / B 3.3-129 through TS / 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 1 Pages B 3.3-139 through B 3.3-149 .0 Pages TS/ B 3.3-150 through TS / B 3.3-162 1 Page TS / B 3.3-163 2 Pages TS / B 3.3-164 and TS / B 3.3-165 1 Page TS / B 3.3-166 c 2 Pages TS / B 3.3-167 through TS / B 3.3-177 1 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 1 Page TS / B 3.3-181 2 Pages TS / B 3.3-182 through TS / 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 3.3-209 0 Pages TS / B 3.3-210 through TS / B 3.3-213 1 Pages B 3.3-214 through B 3.3-220 0 SUSQUEHANNA - UNIT 2 TS I B LOES-3 Revision 82

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

Section Title Revision B 3.4 REACTOR COOLANT SYSTEM BASES Pages TS / B 3.4-1 and TS / B 3.4-2 1 Pages TS / B 3.4-3 and TS / B 3.4-4 4 Pages TS / B 3.4-5 and TS / B 3.4-9 3 Pages B 3.4-10 through B 3.4-14 0 Page TS / B 3.4-15 1 Pages TS / B 3.4-16 through TS / B 3.4-18 2 Pages B 3.4-19 through B 3.4-27 0 Pages TS / B 3.4-28 and TS / B 3.4-29 1 Pages B 3.4-30 through B 3.3-31 0 Page TS / B 3.4-32 1 Pages B 3.4-33 through B 3.4-36 0 Page TS / B 3.4-37 1 Pages B 3.4-38 through B 3.4-40 0 Page TS / B 3.4-41 1 Pages B 3.4-42 through B 3.4-48 0 Page TS / B 3.4-49 3 Pages TS / B 3.4-50 through TS B B 3.4-52 2 Page TS / B 3.4-53 1 Pages TS / B 3.4-54 and TS / B 3.4-55 2 Page TS / B 3.4-56 1 Page TS / B 3.4-57. 2 Pages TS / B 3.4-58 through TS / B 3.4-60 1 B 3.5 ECCS AND RCIC BASES -

Pages TS / B 3.5-1 and TS B 3.5-2 1 Pages TS / B 3.5-3 through TS / B 3.5 2 Pages TS / B 3.5-7 through TS / B 3.5-10 1 Pages TS / B 3.5-11 and TS / B 3.5-12 2 Pages TS / B 3.6-13 and TS / B 3.5-14 1 Pages TS / B 3.5-15 and TS / B.3.5-16 2 Page TS / B 3.5-17 3 Page TS / B 3.5-18 1 Pages B 3.5-19 through B 3.5-24 0 Pages TS / B 3.5-25 through TS / B 3.5-27 1 Pages B 3.5-28 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 Page TS / B 3.6-2 2 Pages TS / B 3.6-3 through TS / B 3.6-5 3 Page TS / B 3.6-6 4 Pages TS / B 3.6-6a and TS / B 3.6-6b 2 Page TS / B 3.6-6c 0 SUSQUEHANNA - UNIT 2 TS / B LOES-4 Revision 82

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

Section Title Revision Pages B 3.6-7 through B 3.6-14 0 Page TS / B 3.6-15 3 Page TS / B 3.6-15a 0 Page TS / B 3.6-15b 2 Page TS / B 3.6-16 1 Page TS / B 3.6-17 2 Page TS / B 3.6-17a 0 Pages TS / B 3.6-18 and TS / B 3.6-19 1 Page TS / B 3.6-20 2 Page TS / B 3.6-21 3 Pages TS / B 3.6-21a and TS / B 3.6-21b 0 Pages TS / B 3.6-22 and TS / B 3.6-23 2

.Pages TS / B 3.6-24 and TS / B 3.6-25 1 Page TS / B 3.6-26 2 Page TS / B 3.6-27 3 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 Page TS / B 3.6-32 1 Page TS / B 3.6-33 2 Page TS / B 3.6 1 Pages TS / B 3,6-35 through TS / B 3.6-37 2 Page TS / B 3.6-38 3 Page TS / B 3.6-39 - 7 Pages B 3.6-40 through B 3.6-42 0 Pages TS / B 3.6-43 and TS / B 3.6-44 -- 1 Page TS / B 3.6-45 2 Pages TS / B 3.6-46 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 Pages TS / B 3.6-63 and TS / B 3.6-64 1 Pages B 3.6-65 through B 3.6-68 0 Pages B 3.6-69 through B 3.6-71 1 Page TS / B 3.6-72 2 Pages TS / B 3.6-73 through TS / B 3.6-74 1 Pages B 3.6-75 and B 3.6-76 0 Page TS / B 3.6-77 1 Pages B 3.6-78 through B 3.6-82 0 Page TS / B 3.6-83 3 Page TS / B 3.6-84 2 Page TS / B 3.6-85 4 Page TS / B 3.6-86 through TS / B 3.6-87a 2 Revision 82 TS/BLOES-5 SUSQUEHANNA - UNIT SUSQUEHANNA -

UNIT 22 TS / B LOES-5 Revision 82

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

Section Title Revision Page TS / B 3.6-88 3 Page TS / B 3.6-89 2 Page TS / B 3.6-90 3 Pages TS / B 3.6-91 through TS / B 3.6-95 1 Page TS / B 3.6-96 2 Pages TS / B 3.6-97 and TS / B 3.6-98 1 Page TS / B 3.6-99 2 Page TS / B 3.6-99a 0 Pages TS / B 3.6-100 and TS I B 3.6-101 1 Pages TS / B 3.6-102 and TS I B 3.6-103 2 Page TS / B 3.6-104 3 Page TS / B 3.6-105 2 Page TS / B 3.6-106 3 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 / B 3.7-6c 0 Page TS / B 3.7-7 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--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 B 3.8 ELECTRICAL POWER SYSTEMS BASES Pages B 3.8-1 through B 3.8-3 0 Page TS / B 3.8-4 1 Pages TS / B 3.8-4a and TS / B 3.8-4b 0 Pages TS / B 3.8-5 and Page TS / B 3.8-6 1 Pages B 3.8-7 and B 3.8-8 0 Page TS / B 3.8-9 2 Pages TS / B 3.8-10 and TS / B 3.8-11 1 Pages B 3.8-12 through B 3.8-18 0 Page TS / B 3.8-19 1 Pages B 3.8-20 through B 3.8-22 0 Page TS / B 3.8-23 1 Page B 3.8-24 0 Revision 82 TSIBLOES-6 SUSQUEHANNA - UNIT SUSQUEHANNA -

UNIT 22 TS / B LOES-6 Revision 82

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

Section Title Revision Pages TS / B 3.8-25 and TS / B 3.8-26 1 Pages B 3.8-27 through B 3.8-37 0 Page TS / B 3.8-38 1 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 2 Page TS / B 3.8-65 3 Page TS-/ B 3.8-66 4 Pages TS / B 3.8-67 and TS / B 3.8-68 3 Page TS / B 3.8-69 4 Pages TS / B 3.8-70 through TS / B 3.8-83 1 Pages TS / B 3.8-83A through TS / B 3.8-83D 0 Pages B 3.8-84 through B 3.8-85 0 Page TS / B 3.8-86 1 Page TS / B 3.8-87 2 Pages TS / B 3.8-88 through TS / B 3.8-93 1 Pages B 3.8-94 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 1 Pages TS / B 3.9-3 and TS / B 3.9-4 1 Pages B 3.9-5 through B 3.9-30 0 B 3.10 SPECIAL OPERATIONS BASES Page TS / B 3.10 - 1 Pages B 3.10-2 through B 3.10-32 .- 0 Page TS / B 3.10-33 2 Page B 3.10-34 0 Page B 3.10-35 1 Pages B 3.10-36 and B 3.10-37 0 Page B 3.10-38 1 Page TS / B 3.10-39 2 TSB2 Text LOES.doc 3/27/07 SUSQUEHANNA - UNIT 2 TS / B LOES-7 Revision 82

TABLE OF CONTENTS (TECHNICAL SPECIFICATIONS BASES)

B2.0 SA FETY LIM ITS (SLs)............................................................................ TS/B2.0-1 B2.1.1 Reactor C ore 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 ........ TS/B3.0-1 B3.0 SURVEILLANCE REQUIREMENT (SR) APPLICABILITY ...................... TS/B3.0-10 B3.1 REACTIVITY CONTROL SYSTEMS ................................................... B3.1-1 B3.1.1 Shutdow n Margin (SDM) ................................................................ B3.1-1 B3.1.2 Reactivity A nom alies ...................................................................... B3.1-8 B3.1.3 Control Rod OPERABILITY ............................................................ B3.1-13 B3.1.4 Control Rod Scram Times ........................................................ TS/B3.1-22 B3.1.5 Control Rod Scram Accumulators ............................................ TS/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 Volume (SDV) Vent and Drain Valves ................ B3.1-47 B3.2 POWER DISTRIBUTION LIMITS ................................................... TS/B3.2-1 B3.2.1 Average Planar Linear Heat Generation Rate (APLHGR) ........ TS/B3.2-1 B3.2.2 Minimum Critical Power Ratio (MCPR) ..................................... TS/B3.2-5 B3.2.3 Linear Heat Generation Rate (LHGR) ....................................... TS/B3.2-10 B3.3 INSTRUMENTATION ..................................................................... TS/B3.3-1 B3.3.1.1 Reactor Protection System (RPS) Instrumentation ................... TS/B3.3-1 B3.3.1.2 Source Range Monitor (SRM) Instrumentation ......................... TS/B3.3-35 B3.3.2.1 Control Rod Block Instrumentation .................... TS/B3.3-44 B3.3.2.2 Feedwater - Main Turbine High Water Level Trip .. ..

Instrum entation ........................................................................ B-3.3-55 B3.3.3.1 Post Accident Monitoring (PAM) Instrumentation ..................... TS/B3.3-64 B3.3.3.2 Remote Shutdown System ............................................................. B3.3-76 B3.3.4.1 End of Cycle Recirculalion Pump Trip (EOC-RPT)

Instrum entation ......................................................................... B 3.3-8 1 B3.3.4.2 Anticipated Transient Without Scram Recirculation Pump Trip (ATWS-RPT) Instrumentation .................................. B3.3-92 B3.3.5.1 Emergency Core Cooling System (ECCS)

Instrum entation ................................................................... TS/B3.3-1 01 B3.3.5.2 Reactor Core Isolation Cooling (RCIC) System Instrum entation ......................................................................... 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 - I Revision 11

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 Mo nito ring ................................................................................ B 3 .3-2 14 B3.4 REACTOR COOLANT SYSTEM (RCS) ......................................... TS/B3.4-1 B3.4.1 Recirculation Loops Operating ................................................. TS/B3.4-1 B 3 .4 .2 Jet P um ps ...................................................................................... B 3.4-10 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 (PIT) Limits ........................... TS/B3.4-49 B3.4.1 1 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 ......................................................................... 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 ...........................................................-. TS/B36-1 B3.6.1.2 Primary Containment Air Lock ....... ........... B3.6-7 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 B3.6.2.1 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 C o o ling ..................................................................................... B 3 .6 -6 1 B3.6.2.4 Residual Heat Removal (RHR) Suppression Pool Spray ................ B3.6-65 B3.6.3.1 Not Used ................................................................................. TS/B 3.6-69 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 ........................................................... TS/B3.6-83 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)

SUSQUEHANNA - UNIT 2 TS / B TOC - 2 Revision 11

TABLE OF CONTENTS (TECHNICAL SPECIFICATIONS BASES)

B3.7 PLANT SYSTEM S ......................................................................... 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 (C REOAS) System ............................................................. TS/B 3.7-12 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 - O perating ................................................................ 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 DC Sources - Operating .......................................................... TS/B3.8-56 B3.8.5 DC Sources - Shutdown ................................................................ B3.8-70 B3.8.6 Battery Cell Param eters ................................................................. 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 ............................. B3.10-11 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 3/23/07 SUSQUEHANNA - UNIT 2 TS / B TOC - 3 Revision 11

PPL Rev. 3 APLHGR B 3.2.1 B 3.2 POWER DISTRIBUTION LIMITS B 3.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)

BASES BACKGROUND The APLHGR is a measure of the average LHGR of all the fuel rods in a fuel assembly at any axial location. Limits on the APLHGR are specified to ensure that limits specified in 10 CFR 50.46 are not exceeded during the postulated design basis loss of coolant accident (LOCA).

APPLICABLE SPC performed LOCA calculations for the SPC ATRIUMT-10 fuel SAFETY ANALYSES design. The analytical methods and assumptions used in evaluating the fuel design limits from 10 CFR 50.46 are presented in References 3, 4, 5, and 6 for the SPC analysis. The analytical methods and assumptions used in evaluating Design Basis Accidents (DBAs) that determine the APLHGR Limits are presented in References 3 through 9.

LOCA analyses are performed to ensure that the APLHGR limits are adequate to meet the Peak Cladding Temperature (PCT), maximum cladding oxidation, and maximum hydrogen generation limits of 10 CFR 50.46. The analyses are performed using calculational models that are consistent with the requirements of 10 CFR 50, Appendix K.

A complete discussion of the analysis codes ariprb-oided-in --.

References 3, 4, 5, and 6 for the SPC analysis. The PCT following a postulated LOCA is a function of the average heat generation rate of all the rods of a fuel assembly at any axial location and is not strongly influenced by the rod to rod power distribution within the assembly.

APLHGR limits are developed as a function of fuel type and exposure.

The SPC analysis is valid for full cores of ATRIUMTM -10 fuel. The SPC LOCA analyses also consider several alternate operating modes in the development of the APLHGR limits (e.g., Maximum Extended Load Line Limit Analysis (MELLLA), Suppression Pool Cooling Mode, and Single Loop Operation (SLO)). LOCA analyses were performed for the regions of the power/ flow map bounded by the rod line that runs through 100% RTP and maximum core flow and the upper boundary of the MELLLA region. The MELLLA region is analyzed to determine whether an APLHGR multiplier as a function of core flow is required.

The results of the analysis demonstrate the PCTs are within the 10 CFR 50.46 limit, and that APLHGR multipliers as a function of core flow are not required.

(continued)

SUSQUEHANNA - UNIT 2 TS / B 3.2-1 Revision 2

PPL Rev. 3 APLHGR B 3.2.1 BASES APPLICABLE The SPC LOCA analyses consider the delay in Low Pressure Coolant SAFETY ANALYSES Injection (LPCI) availability when the unit is operating in the (continued) Suppression Pool Cooling Mode. The delay in LPCI availability is due to the time required to realign valves from the Suppression Pool Cooling Mode to the LPCI mode. The results of the analyses demonstrate that the PCTs are within the 10 CFR 50.46 limit.

Finally, the SPC LOCA analyses were performed for Single-Loop Operation. The results of the SPC analysis for ATRIUMTM-10 fuel shows that an APLHGR limit which is 0.8 times the two-loop APLHGR limit meets the 10 CFR 50.46 acceptance criteria, and that the PCT is less than the limiting two-loop PCT.

The APLHGR satisfies Criterion 2 of the NRC Policy Statement (Ref.

10).

LCO The APLHGR limits specified in the COLR are the result of the DBA analyses.

APPLICABILITY The APLHGR limits are primarily derived from LOCA analyses that are assumed-to occur at high power levels. Design calculati6ns and operating experience have shown that as power is-reduced, t56-margin to the required APLHGR limits increases. At THERMAL POWER levels < 25% RTP, the reactor is operating with substantial margin to the APLHGR limits; thus, this LCO is not required.

ACTIONS A.1 If any APLHGR exceeds the required limits, an assumption regarding an initial condition of the DBA may not be met. Therefore, prompt action should be taken to restore the APLHGR(s) to within the required limits such that the plant operates within analyzed conditions. The 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> Completion Time is sufficient to restore the APLHGR(s) to within its limits and is acceptable based on the low probability of a DBA occurring simultaneously with the APLHGR out of specification.

(continued)

SUSQUEHANNA - UNIT 2 TS / B 3.2-2 Revision 1

PPL Rev. 3 APLHGR B 3.2.1 BASES ACTIONS B. 1 (continued)

If the APLHGR cannot be restored to within its required limits within the associated Completion Time, the plant must be brought to a MODE or other specified condition in which the LCO does not apply. To achieve this status, THERMAL POWER must be reduced to < 25% RTP within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. The allowed Completion Time is reasonable, based on operating experience, to reduce THERMAL POWER to < 25% RTP in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.2.1.1 REQUIREMENTS APLHGRs are required to be initially calculated within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after THERMAL POWER is > 25% RTP and then every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> thereafter.

Additionally, APLHGRs must be calculated prior to exceeding 50%

RTP unless performed in the previous 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. APLHGRs are compared to the specified limits in the COLR to ensure that the reactor is operating within the assumptions of the safety analysis. The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Frequency is based on both engineering judgment and recognition of the slowness of changes in power distribution during normal operation. The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> allowance after THERMAL POWER

> 25% RTP is achieved is acceptable given the large inherent margin to operating limits at low power levels and because the APLHGRs must be calculated prior to exceeding 50 01 RTP.

REFERENCES 1. Not Used

2. Not Used

3. EMF-2361(P)(A), "EXEM BWR-2000 ECCS Evaluation Model,"

Framatome ANP.

4. EMF-2292(P)(A) Revision 0, "ATRIUM TM-10:

Appendix K Spray Heat Transfer Coefficients."

5. XN-CC-33(P)(A) Revision 1, "HUXY: A Generalized Multirod Heatup Code with 10CFR50 Appendix K Heatup Option Users Manual," November 1975.

(continued)

SUSQUEHANNA - UNIT 2 TS / B 3.2-3 Revision 3

PPL Rev. 3 APLHGR B 3.2.1 BASES REFERENCES 6. XN-NF-80-19(P)(A), Volumes 2, 2A, 2B, and 2C "Exxon (continued) Nuclear Methodology for Boiling Water Reactors: EXEM BWR ECCS Evaluation Model," September 1982.

7. FSAR, Chapter 4.

8. FSAR, Chapter 6.

9. FSAR, Chapter 15.

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

SUSQUEHANNA - UNIT 2 TS / B 3.2-4 Revision 1

PPL Rev. 3 RPS Instrumentation B 3.3.1.1 B 3.3 INSTRUMENTATION B 3.3.1.1 Reactor Protection System (RPS) Instrumentation BASES BACKGROUND The RPS initiates a reactor scram when one or more monitored parameters exceed their specified limits, to preserve the integrity of the fuel cladding and the Reactor Coolant System (RCS) and minimize the energy that must be absorbed following a loss of coolant accident (LOCA). This can be accomplished either automatically or manually.

The protection and monitoring functions of the RPS have been designed to ensure safe operation of the reactor. This is achieved by specifying limiting safety system settings (LSSS) in terms of parameters directly monitored by the RPS, as well as LCOs on other reactor system parameters and equipment performance. The LSSS are defined in this Specification as the Allowable Values, which, in conjunction with the LCOs, establish the threshold for protective system action to prevent exceeding acceptable limits, including Safety Limits (SLs) during Design Basis Accidents (DBAs).

The RPS, as shown in the FSAR, Figure 7.2-1 (Ref. 1), includes sensors, relays, bypass circuits, and switches that are necessary to cause-initiation of a reactor scram. Functional diversity is provided by monitoring a wide range of dependent and-independent P5arameters.

The input parameters to the scram logic are from-iri-stfrumenta-t6n that monitors reactor vessel water level, reactor vessel pressurel neutron flux, main steam line isolation valve position, turbine control valve (TCV) fast closure trip oil pressure, turbine stop valve (TSV) position, drywell pressure, and scram discharge volume (SDV) water level, as well as reactor mode switch in shutdown position and manual scram signals. There are at least four redundant sensor input signals from each of these parameters (with the exception of the reactor mode switch in shutdown scram signal). When the setpoint is reached, the channel sensor actuates, which then outputs an RPS trip signal to the trip logic. Table B 3.3.1.1-1 summarizes the diversity of sensors capable of initiating scrams during anticipated operating transients typically analyzed.

The RPS is comprised of two independent trip systems (A and B) with two logic channels in each trip system (logic (continued)

SUSQUEHANNA - UNIT 2 TS / B 3.3-1 Revision 1

PPL Rev. 3 RPS Instrumentation B 3.3.1.1 BASES BACKGROUND channels Al and A2, B1 and B2) as shown in Reference 1. The (continued) outputs of the logic channels in a trip system are combined in a one-out-of-two logic so that either channel can trip the associated trip system. The tripping of both trip systems will produce a reactor scram.

This logic arrangement is referred to as a one-out-of-two taken twice logic. Each trip system can be reset by use of a reset switch. If a full scram occurs (both trip systems trip), a relay prevents reset of the trip systems for 10 seconds after the full scram signal is received. This 10 second delay on reset ensures that the scram function will be completed.

Two AC powered scram pilot solenoids are located in the hydraulic control unit for each control rod drive (CRD). Each scram pilot valve is operated with the solenoids normally energized. The scram pilot valves control the air supply to the scram inlet and outlet valves for the associated CRD. When either scram pilot valve solenoid is energized, air pressure holds the scram valves closed and, therefore, both scram pilot valve solenoids must be de-energized to cause a control rod to scram. The scram valves control the supply and discharge paths for the CRD water during a scram. One of the scram pilot valve solenoids for each CRD is controlled by trip system A, and the other solenoid is controlled by trip system B. Any trip of trip system A in conjunction with- any trip in trip system B results in de-energizing both solenoids, air bleedirg._off, scram valves opening, and control rod scram.

The DC powered backup scram valves, which energize on a scram signal to depressurize the scrari, air header, are also controlled by the RPS. Additionally, the RPS System controls the SDV vent and drain valves such that when both trip systems trip, the SDV vent and drain valves close to isolate the SDV.

APPLICABLE The actions of the RPS are assumed in the safety analyses of SAFETY References 3, 4, 5 and 6. The RPS initiates a reactor scram before the ANALYSES, LCO, monitored parameter values reach the Allowable Values, specified by the and APPLICABILITY setpoint methodology and listed in Table 3.3.1.1-1 to preserve the integrity of the fuel cladding, the reactor coolant pressure boundary (RCPB), and (continued)

SUSQUEHANNA - UNIT 2 TS / B 3.3-2 Revision 1

PPL Rev. 3 RPS Instrumentation B 3.3.1.1 BASES APPLICABLE the containment by minimizing the energy that must be absorbed SAFETY following a LOCA.

ANALYSES, LCO, and RPS instrumentation satisfies Criterion 3 of the NRC Policy Statement.

APPLICABILITY (Ref. 2)

(continued)

Functions not specifically credited in the accident analysis are retained for the overall redundancy and diversity of the RPS as required by the NRC approved licensing basis.

The OPERABILITY of the RPS is dependent on the OPERABILITY of the individual instrumentation channel Functions specified in Table 3.3.1.1-1. Each Function must have a required number of OPERABLE channels per RPS trip system, with their setpoints within the specified Allowable Value, where appropriate. The actual setpoint is calibrated consistent with applicable setpoint methodology assumptions. Each channel must also respond within its assumed response time.

Allowable Values are specified for each RPS Function specified in the Table. Nominal trip setpoints are specified in the setpoint calculations.

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

A channel is inoperable if its actual trip setpoint is not within its required Allowable Value.

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 vessel water level), and when the measured output value of the process parameter reaches the setpoint, the associated device changes state. The analytic limits are derived from the limiting values of the process parameters obtained from the safety analysis. 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, (continued)

SUSQUEHANNA - UNIT 2 TS / B 3.3-3 Revision 1

PPL Rev. 3 RPS Instrumentation B 3.3.1.1 BASES APPLICABLE instrument drift, and severe environment errors (for channels that must SAFETY function in harsh environments as defined by 10 CFR 50.49) are ANALYSES, LCO, accounted for.

and APPLICABILITY (continued) The OPERABILITY of scram pilot valves and associated solenoids, backup scram valves, and SDV valves, described in the Background section, are not addressed by this LCO.

The individual Functions are required to be OPERABLE in the MODES specified in the table, which may require an RPS trip to mitigate the consequences of a design basis accident or transient. To ensure a reliable scram function, a combination of Functions are required in each MODE to provide primary and diverse initiation signals.

The RPS is required to be OPERABLE in MODE 5 with any control rod withdrawn from a core cell containing one or more fuel assemblies.

Control rods withdrawn from a core cell containing no fuel assemblies do not affect the reactivity of the core and, therefore, are not required to have the capability to scram. Provided all other control rods remain inserted, the RPS function is not required. In this condition, the required SDM (LCO 3.1.1) and refuel position one-rod-out interlock (LCO 3.9.2) ensure that no event requiring RPS will occur. During normal operation in MODES 3 and 4, all control rods are fully inserted and the Reactor Mode Switch Shutdown Position control rod withdrawal block (LCO 3.3.2.1)- does not allow any control rod to be withdrawn-Under these conditions, the RPS function is not required to be OPERABLE.

The exception to this is SpeciaF Operations (LCO 3.10.3 and LCO 3.10.4) which ensure compliance with appropriate requirements.

The specific Applicable Safety Analyses, LCO, and Applicability discussions are listed below on a Function by Function basis.

Intermediate Range Monitor (IRM) 1.a. Intermediate Range Monitor Neutron Flux-High The IRMs monitor neutron flux levels from the upper range of the source range monitor (SRM) to the lower range of the average power range monitors (APRMs). The IRMs are capable of generating trip signals that can be used to prevent fuel (continued)

SUSQUEHANNA-UNIT2 TS / B 3.3-4 Revision 1

PPL Rev. 3 RPS Instrumentation B 3.3.1.1 BASES APPLICABLE l.a. Intermediate Range Monitor Neutron Flux-High (continued)

SAFETY ANALYSES, LCO, damage resulting from abnormal operating transients in the and APPLICABILITY intermediate power range. In this power range, the most significant source of reactivity change is due to control rod withdrawal. The IRM provides diverse protection for the rod worth minimizer (RWM), which monitors and controls the movement of control rods at low power. The RWM prevents the withdrawal of an out of sequence control rod during startup that could result in an unacceptable neutron flux excursion (Ref. 5). The IRM provides mitigation of the neutron flux excursion.

To demonstrate the capability of the IRM System to mitigate control rod withdrawal events, generic analyses have been performed (Ref. 3) to evaluate the consequences of control rod withdrawal events during startup that are mitigated only by the IRM. This analysis, which assumes that one IRM channel in each trip system is bypassed, demonstrates that the IRMs provide protection against local control rod withdrawal errors and results in peak fuel energy depositions below the 170 cal/gm fuel failure threshold criterion.

The IRMs are also capable of limiting other reactivity excursions during startup, -such as cold water injection events, although no credit is specifically assumed.

The IRM System-is divided into two trip systems-, with-four IRM_-

channels inputting to each trip system. The analysis of Reference 3 assumes that one channel in each trip system is bypassed. Therefore, six channels with three channels in each trip system are required for IRM OPERABILITY to ensure that no single instrument failure will preclude a scram from this Function on a valid signal. This trip is active in each of the 10 ranges of the IRM, which must be selected by the operator to maintain the neutron flux within the monitored level of an IRM range.

The analysis of Reference 3 has adequate conservatism to permit an IRM Allowable Value of 122 divisions of a 125 division scale.

The Intermediate Range Monitor Neutron Flux-High Function must be OPERABLE during MODE 2 when control rods may be withdrawn and the potential for criticality exists. In (continued)

SUSQUEHANNA - UNIT 2 TS / B 3.3-5 Revision 2

PPL Rev. 3 RPS Instrumentation B 3.3.1.1 BASES APPLICABLE l.a. Intermediate Range Monitor Neutron Flux-High (continued)

SAFETY ANALYSES, LCO, MODE 5, when a cell with fuel has its control rod withdrawn, the IRMs and APPLICABILITY provide monitoring for and protection against unexpected reactivity excursions. In MODE 1, the APRM System and the RWM provide protection against control rod withdrawal error events and the IRMs are not required. In addition, the Function is automatically bypassed when the Reactor Mode Switch is in the Run position.

l.b. Intermediate Range Monitor-Inop This trip signal provides assurance that a minimum number of IRMs are OPERABLE. Anytime an IRM mode switch is moved to any position other than "Operate," the detector voltage drops below a preset level, or when a module is not plugged in, an inoperative trip signal will be received by the RPS unless the IRM is bypassed. Since only one IRM in each trip system may be bypassed, only one IRM in each RPS trip system may be inoperable without resulting in an RPS trip signal.

This Function was not specifically credited in the accident analysis but it is'retained for the overall redundancy and diversity of the RPS as required by the NRC approved licensing basis.

Six channels of Intermediate Range Monitor-Inop with three-channels in each trip system are required to be OPERABLE to ensure that no single instrument failure will preclude a scram from this Function on a valid signal.

Since this Function is not assumed in the safety analysis, there is no Allowable Value for this Function.

This Function is required to be OPERABLE when the Intermediate Range Monitor Neutron Flux-High Function is required.

(continued)

SUSQUEHANNA - UNIT 2 TS / B 3.3-6 Revision 1

PPL Rev. 3 RPS Instrumentation B 3.3.1.1 BASES APPLICABLE Average Power Range Monitor (APRM)

SAFETY

ANALYSES, LCO, and The APRM channels provide the primary indication of neutron flux APPLICABILITY within the core and respond almost instantaneously to neutron flux (continued) increases. The APRM channels receive input signals from the local power range monitors (LPRMs) within the reactor core to provide an indication of the power distribution and local power changes. The APRM channels average these LPRM signals to provide a continuous indication of average reactor power from a few percent to greater than RTP. Each APRM channel also includes an Oscillation Power Range Monitor (OPRM) Upscale Function, which monitors small groups of LPRM signals to detect thermal-hydraulic instabilities.

The APRM trip System is divided into four APRM channels and four 2-out-of-4 Voter channels. Each APRM channel provides inputs to each of the four voter channels. The four voter channels are divided into two groups of two each with each group of two providing inputs to one RPS trip system. The system is designed to allow one APRM channel, but no voter channels, to be bypassed. A trip from any one unbypassed APRM will result in a "half-trip" in all four of the voter channels, but no trip inputs to either RPS trip system.

APRM-trip Functions 2.a, 2.b, 2.c, and 2.d are voted independently from OPRM Trip Function 2.f. Therefore, any Function 2.a, 2b.,-2.c, or 2.d trip from any two unbypassed APRM channels will result in a full trip in each of the four voter channels, which in turn results in two trip inputs into each RPS trip system logic channel (Al, A2, B1, and B2),

thus resulting in a full scram signal. Similarly, a Function 2.f trip from any two unbypassed APRM channels will result in a full trip from each of the four voter channels.

Three of the four APRM channels and all four of the voter channels are required to be OPERABLE to ensure that no single failure will preclude a scram on a valid signal. In addition, to provide adequate coverage of the entire core consistent with the design bases for the APRM Functions 2.a, 2.b, and 2.c, at least 20 LPRM inputs with at least three LPRM inputs from each of the four axial levels at which the LPRMs are located must be OPERABLE for each APRM channel, with no more than 9, LPRM detectors declared inoperable since the most recent APRM gain calibration. Per Reference 23, the minimum input (continued)

SUSQUEHANNA - UNIT 2 TS / B 3.3-7 Revision 3

PPL Rev. 3 RPS Instrumentation B 3.3.1.1 BASES APPLICABLE Average Power Range Monitor (APRM) (continued)

SAFETY ANALYSES, LCO, requirement for an APRM channel with 43 LPRM inputs is determined and APPLICABILITY given that the total number of LPRM outputs used as inputs to an APRM channel that may be bypassed shall not exceed twenty-three (23). Hence, 20 LPRM inputs producing a channel trip signal. needed to be operable. For the OPRM Trip Function 2.f, each LPRM in an APRM channel is further associated in a pattern of OPRM "cells," as described in References 17 and 18. Each OPRM cell is capable of producing a channel trip signal.

2.a. Average Power Range Monitor Neutron Flux-High (Setdown)

For operation at low power (i.e., MODE 2), the Average Power Range Monitor Neutron Flux-High (Setdown) Function is capable of generating a trip signal that prevents fuel damage resulting from abnormal operating transients in this power range. For most operation at low power levels, the Average Power Range Monitor Neutron Flux-High (Setdown) Function will provide a secondary scram to the Intermediate Range Monitor Neutron Flux-High Function because of the relative setpoints. With the IRMs at Range 9 or 10, it is possible that the-Average Power Range Monitor Neutron Flux-High (Setdown) I Function will provide the primary trip signal for a corewide increase in power.

The Average Power Range Monitor Neutron Flux-High (Setdown)

Function together with the IRM-High Function provide mitigation for the control rod withdrawal event during startup (Section 15.4.1 of Ref 5).

Also, the Function indirectly ensures that before the reactor mode switch is placed in the run position, reactor power does not exceed 25% RTP (SL 2.1.1.1) when operating at low reactor pressure and low core flow. Therefore, it indirectly prevents fuel damage during significant reactivity increases with THERMAL POWER < 25% RTP.

The Allowable Value is based on preventing significant increases in power when THERMAL POWER is < 25% RTP.

The Average Power Range Monitor Neutron Flux-High (Setdown)

Function must be OPERABLE during MODE 2 when control rods may be withdrawn since the potential for criticality exists. In MODE 1, the Average Power Range Monitor Neutron Flux-High Function provides protection against reactivity transients and the RWM protects against control rod withdrawal error events.

(continued)

SUSQUEHANNA - UNIT 2 TS / B 3.3-8 Revision 3

PPL Rev. 3 RPS Instrumentation B 3.3.1.1 BASES APPLICABLE 2.a. Average Power Range Monitor Neutron Flux-High (Setdown)

SAFETY (continued)

ANALYSES, LCO, and APPLICABILITY There are provisions in the design of the NUMAC PRNM that given certain circumstances, such as loss of one division of RPS power, an individual APRM will default to a "run" mode condition logic. If the plant is in mode 2 when this occurs, the individual APRM will be in a condition where the 'run' mode setpoint (Function 2.c) and not the

'setdown' setpoint (Function 2.a) will be applied. If this condition occurs while in reactor mode 2 condition, the appropriate LCO condition per Table 3.3.1.1-1 needs to be entered.

2.b. Averaae Power Ranae Monitor Simulated Thermal Power-Hiah The Average Power Range Monitor Simulated Thermal Power-High Function monitors neutron flux to approximate the THERMAL POWER being transferred to the reactor coolant. The APRM neutron flux is electronically filtered with a time constant representative of the fuel heat transfer dynamics to generate a signal proportional to the THERMAL POWER in the reactor. The trip level is varied as a function of recirculation drive flow (i.e., at lower core flows, the setpoint is reduced proportional to the reduction in power experienced as core flow i§ reduced with a fixed control rod pattern) but is clamped at an upper limit that is always lower than the Average P-ower-Range-Monitor Neutron Flux-High Function Allowable Value. The Average- Power Range Monitor Simulated Thermal Power-High Function is not credited in any plant Safety Analyses. The Average Power Range Monitor Simulated Thermal Power-High Function Limit is set above the APRM Rod Block to provide defense in depth to the APRM Neutron Flux-High for transients where THERMAL POWER increases slowly (such as loss of feedwater heating event). During these events, the THERMAL POWER increase does not significantly lag the neutron flux response and, because of a lower trip setpoint, will initiate a scram before the high neutron flux scram. For rapid neutron flux increase events, the THERMAL POWER lags the neutron flux and the Average Power Range Monitor Neutron Flux-High Function will provide a scram signal before the Average Power Range Monitor Simulated Thermal Power-High Function setpoint is exceeded.

The Average Power Range Monitor Simulated Thermal Power-High Function uses a trip level generated based on recirculation loop drive flow (W) representative of total core flow. Each APRM channel uses one total recirculation drive flow signal. The total recirculation drive (continued)

SUSQUEHANNA - UNIT 2 TS / B 3.3-9 Revision 3

PPL Rev. 3 RPS Instrumentation B 3.3.1.1 BASES APPLICABLE 2.b. Average Power Range Monitor Simulated Thermal Power-High SAFETY (continued)

ANALYSES, LCO, and APPLICABILITY flow signal is generated by the flow processing logic, part of the APRM channel, by summing the flow calculated from two flow transmitter signal inputs, one from each of the two recirculation drive flow loops.

The flow processing logic OPERABILITY is part of the APRM channel OPERABILITY requirements for this Function.

The adequacy of drive flow as a representation of core flow is ensured through drive flow alignment, accomplished by SR 3.3.1.1.20.

A note is included, applicable when the plant is in single recirculation loop operation per LCO 3.4.1, which requires reducing by AW the recirculation flow value used in the APRM Simulated Thermal Power-High Allowable Value equation. The Average Power Range Monitor Scram Function varies as a function of recirculation loop drive flow (W). AW is defined as the difference in indicated drive flow (in percent of drive flow, which produces rated core flow) between two-loop and single-loop operation at the same core flow. The value of AW is established to conservatively bound the inaccuracy created in the core flow/drive flow correlation due to back flow in the jet pumps associated with the inactive recirculation loop. This adjusted Allowable Value thus maintains thermal margins essentially unchanged from those for two-loop operation.

The THERMAL POWER time.constant of < 7 seconds is based on the fuel heat transfer dynamics and provides a signal proportional to the THERMAL POWER. The simulated thermal time constant is part of filtering logic in the APRM that simulates the relationship between neutron flux and core thermal power.

The Average Power Range Monitor Simulated Thermal Power-High Function is required to be OPERABLE in MODE 1 when there is the possibility of generating excessive THERMAL POWER and potentially exceeding the SL applicable to high pressure and core flow conditions (MCPR SL). During MODES 2 and 5, other IRM and APRM Functions provide protection for fuel cladding integrity.

2.c. Average Power Range Monitor Neutron Flux-High The Average Power Range Monitor Neutron Flux-High Function is capable of generating a trip signal to prevent fuel damage or excessive RCS pressure. For the overpressurization protection analysis of Reference 4, the Average Power Range Monitor Neutron Flux-High (continued)

SUSQUEHANNA - UNIT 2 TS / B 3.3-10 Revision 3

PPL Rev. 3 RPS Instrumentation B 3.3.1.1 BASES APPLICABLE 2.c. Average Power Range Monitor Neutron Flux-High (continued)

SAFETY

ANALYSES, LCO, and Function is assumed to terminate the main steam isolation valve APPLICABILITY (MSIV) closure event and, along with the safety/relief valves (S/RVs), limit the peak reactor pressure vessel (RPV) pressure to less than the ASME Code limits. The control rod drop accident (CRDA) analysis (Ref. 5) takes credit for the Average Power Range Monitor Neutron Flux-High Function to terminate the CRDA.

The CRDA analysis assumes that reactor scram occurs on Average Power Range Monitor Neutron Flux-High Function.

The Average Power Range Monitor Neutron Flux-High Function is required to be OPERABLE in MODE 1 where the potential consequences of the analyzed transients could result in the SLs (e.g., MCPR and RCS pressure) being exceeded. Although the Average Power Range Monitor Neutron Flux-High Function is assumed in the CRDA analysis, which is applicable in MODE 2, the Average Power Range Monitor Neutron Flux-High (Setdown) Function conservatively bounds the assumed trip and, together with the assumed IRM trips, provides adequate protection. Therefore, the Average Power Range Monitor Neutron Flux-High Function is not required in MODE 2.. ...

2.d. Average Power Range M6nitor-lnop Three of the four APRM channels are required to be OPERABLE for each of the APRM Functions. This Function (Inop) provides assurance that the minimum number of APRM channels are OPERABLE.

For any APRM channel, any time its mode switch is not in the "Operate" position, an APRM module required to issue a trip is unplugged, or the automatic self-test system detects a critical fault with the APRM channel, an Inop trip is sent to all four voter channels. Inop trips from two or more unbypassed APRM channels result in a trip output from each of the four voter channels to its associated trip system.

This Function was not specifically credited in the accident analysis, but it is retained for the overall redundancy and diversity of the RPS as required by the NRC approved licensing basis.

(continued)

SUSQUEHANNA - UNIT 2 TS / B 3.3-11 Revision 3

PPL Rev. 3 RPS Instrumentation B 3.3.1.1 BASES APPLICABLE 2.d. Average Power Range Monitor-lnop (continued)

SAFETY

ANALYSES, LCO, and There is no Allowable Value for this Function.

APPLICABILITY This Function is required to be OPERABLE in the MODES where the APRM Functions are required.

2.e. 2-out-of-4 Voter The 2-out-of-4 Voter Function provides the interface between the APRM Functions, including the OPRM Trip Function, and the final RPS trip system logic. As such, it is required to be OPERABLE in the MODES where the APRM Functions are required and is necessary to support the safety analysis applicable to each of those Functions.

Therefore, the 2-out-of-4 Voter Function is required to be OPERABLE in MODES 1 and 2.

All four voter channels are required to be OPERABLE. Each voter channel includes self-diagnostic functions. If any voter channel detects a critical fault in its own processing, a trip is issued from that voter channel to the associated RPS trip system.

The Two-Out-Of-Four Logic Module includes both the 2-out-of-4 Voter hardware and-the APRM Interface hardware. The 2-out-of-4 Voter Function 2.e votes APRM Functions 2.a, 2.b, 2.c, and 2.d independently of Function 2.f. This voting is accomplished by the 2-out-of-4 Voter hardware in the Two-Out-Of-Four Logic Module.

The voter includes separate outputs to RPS for the two independently voted sets of Functions, each of which is redundant (four total outputs). The analysis in Reference 15 took credit for this redundancy in the justification of the 12-hour Completion Time for Condition A, so the voter Function 2.e must be declared inoperable if any of its functionality is inoperable. The voter Function 2.e does not need to be declared inoperable due to any failure affecting only the APRM Interface hardware portion of the Two-Out-Of-Four Logic Module.

There is no Allowable Value for this Function.

(continued)

SUSQUEHANNA-UNIT2 TS / B 3.3-12 Revision 3

PPL Rev. 3 RPS Instrumentation B 3.3.1.1 BASES APPLICABLE 2.f. Oscillation Power Ranqe Monitor (OPRM) Trip SAFETY

ANALYSES, LCO, and The OPRM Trip Function provides compliance with GDC 10, "Reactor APPLICABILITY Design," and GDC 12, "Suppression of Reactor Power Oscillations" (continued) thereby providing protection from exceeding the fuel MCPR safety limit (SL) due to anticipated thermal-hydraulic power oscillations.

References 17, 18 and 19 describe three algorithms for detecting thermal-hydraulic instability related neutron flux oscillations: the period based detection algorithm (confirmation count and cell amplitude), the amplitude based algorithm, and the growth rate algorithm. All three are implemented in the OPRM Trip Function, but the safety analysis takes credit only for the period based detection algorithm. The remaining algorithms provide defense in depth and additional protection against unanticipated oscillations. OPRM Trip Function OPERABILITY for Technical Specification purposes is based only on the period based detection algorithm.

The OPRM Trip Function receives input signals from the local power range monitors (LPRMs) within the reactor core, which are combined into "cells" for evaluation by the OPRM algorithms. Each channel is capable of detecting thermal-hydraulic instabilities, by detecting the related neutron flux oscillations, and issuing a trip signal before the MCPR SL is exceeded. Three of the four channels-are required_-to be OPERABLE.

The OPRM Trip is automatically enabled (bypass removed) when THERMAL POWER is >_30% RTP, as indicated by the APRM Simulated Thermal Power, and reactor core flow is _ the value defined in the COLR, as indicated by APRM measured recirculation drive flow.

This is the operating region where actual thermal-hydraulic instability and related neutron flux oscillations are expected to occur. Reference 21 includes additional discussion of OPRM Trip enable region limits.

These setpoints, which are sometimes referred to as the "auto-bypass" setpoints, establish the boundaries of the OPRM Trip enabled region.

The APRM Simulated Thermal Power auto-enable setpoint has 1%

deadband while the drive flow setpoint has a 2% deadband. The deadband for these setpoints is established so that it increases the enabled region once the region is entered.

(continued)

SUSQUEHANNA - UNIT 2 TS / B 3.3-13 Revision 2

PPL Rev. 3 RPS Instrumentation B 3.3.1.1 BASES APPLICABLE 2.f. Oscillation Power Ranqe Monitor (OPRM) Trip (continued)

SAFETY

ANALYSES, LCO, and The OPRM Trip Function is required to be OPERABLE when the plant APPLICABILITY is at > 25% RTP. The 25% RTP level is selected to provide margin in the unlikely event that a reactor power increase transient occurring without operator action while the plant is operating below 30% RTP causes a power increase to or beyond the 30% APRM Simulated Thermal Power OPRM Trip auto-enable setpoint. This OPERABILITY requirement assures that the OPRM Trip auto-enable function will be OPERABLE when required.

An APRM channel is also required to have a minimum number of OPRM cells OPERABLE for the Upscale Function 2.f to be OPERABLE. The OPRM cell operability requirements are documented in the Technical Requirements Manual, TRO 3.3.9, and are established as necessary to support the trip setpoint calculations performed in accordance with methodologies in Reference 19.

An OPRM Trip is issued from an APRM channel when the period based detection algorithm in that channel detects oscillatory changes in the neutron flux, indicated by the combined signals of the LPRM detectors in a cell, with period confirmations and relative cell amplitude exceeding specified setpoints. One or more cells in a channel exceeding the trip conditions will result in a channel OPRM Tripfrom that channel. An OPRM Trip is also issued from-the channel if either the growth rate or amplitude-based algorithms detect oscillatory changes in the neutron flux for one or more cells in that channel.

(Note: To facilitate placing the OPRM Trip Function 2.f in one APRM channel in a "tripped" state, if necessary to satisfy a Required Action, the APRM equipment is conservatively designed to force an OPRM Trip output from the APRM channel if an APRM Inop condition occurs, such as when the APRM chassis keylock switch is placed in the Inop position.)

There are three "sets" of OPRM related setpoints or adjustment parameters: a) OPRM Trip auto-enable region setpoints for STP and drive flow; b) period based detection algorithm (PBDA) confirmation count and amplitude setpoints; and c) period based detection algorithm tuning parameters.

(continued)

SUSQUEHANNA - UNIT 2 TS / B 3.3-14 Revision 3

PPL Rev. 3 RPS Instrumentation B 3.3.1.1 BASES APPLICABLE 2.f. Oscillation Power Range Monitor (OPRM) Trip (continued)

SAFETY

ANALYSES, LCO, and The first set, the OPRM Trip auto-enable setpoints, as discussed in the APPLICABILITY SR 3.3.1.1.19 Bases, are treated as nominal setpoints with no additional margins added. The settings are defined in the Technical Requirements Manual, TRO 3.3.9, and confirmed by SR 3.3.1.1.19.

The second set, the OPRM PBDA trip setpoints, are established in accordance with methodologies defined in Reference 19, and are documented in the COLR. There are no allowable values for these setpoints. The third set, the OPRM PBDA "tuning" parameters, are established or adjusted in accordance with and controlled by requirements in the Technical Requirements Manual, TRO 3.3.9.

3. Reactor Vessel Steam Dome Pressure-High An increase in the RPV pressure during reactor operation compresses the steam voids and results in a positive reactivity insertion. This causes the neutron flux and THERMAL POWER transferred to the reactor coolant to increase, which could challenge the integrity of the fuel cladding and the RCPB. This trip Function is assumed in the low power generator load rejection without bypass and the recirculation flow controller failure (increasing) event. However, the Reactor Vessel Stearmn Dome Pressure-High Function initiates a scram for transients that results in a pressure increase, counteracting the pressure increase by rapidly reducing core power. For the overpressurization protection analysis of Reference 4, reactor scram (the analyses conservatively assume a scram from either the Average Power Range Monitor Neutron Flux-High signal, or the Reactor Vessel Steam Dome Pressure-High'signal), along with the S/RVs, limits the peak RPV pressure to less than the ASME Section III Code limits.

High reactor pressure signals are initiated from four pressure instruments that sense reactor pressure. The Reactor Vessel Steam Dome Pressure-High Allowable Value is chosen to provide a sufficient margin to the ASME Section III Code limits during the event.

Four channels of Reactor Vessel Steam Dome Pressure-High Function, with two channels in each trip system arranged in a one-out-of-two logic, are required to be OPERABLE to ensure that no single instrument failure will preclude a scram from this Function on a valid signal. The Function is required to be OPERABLE in MODES 1 and 2 when the RCS is pressurized and the potential for pressure increase exists.

(continued)

SUSQUEHANNA - UNIT 2 TS / B 3.3-15 Revision 2

PPL Rev. 3 RPS Instrumentation B 3.3.1.1 BASES APPLICABLE 4. Reactor Vessel Water Level-Low, Level 3 SAFETY

ANALYSES, LCO, and Low RPV water level indicates the capability to cool the fuel may be APPLICABILITY threatened. Should RPV water level decrease too far, fuel damage (continued) could result. Therefore, a reactor scram is initiated at Level 3 to substantially reduce the heat generated in the fuel from fission. The Reactor Vessel Water Level-Low, Level 3 Function is assumed in the analysis of the recirculation line break (Ref. 6). The reactor scram reduces the amount of energy required to be absorbed and, along with the actions of the Emergency Core Cooling Systems (ECCS), ensures that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46.

Reactor Vessel Water Level-Low, Level 3 signals are initiated from four level instruments that sense the difference between the pressure due to a constant column of water (reference leg) and the pressure due to the actual water level (variable leg) in the vessel.

Four channels of Reactor Vessel Water Level-Low, Level 3 Function, with two channels in each trip system arranged in a one-out-of-two logic, are required to be OPERABLE to ensure that no single instrument failure will preclude a scram from this Function on a valid signal.

The Reactor Vessel Water Level-Low, Level 3-Allowable Value is selected to ensure that during'normal operation the separator skirts are not uncovered (this protects available recirculation pump net positive suction head (NPSH) from significant carryunder) and, for transients involving loss of all normal feedwater flow, initiation of the low pressure ECCS subsystems at Reactor Vessel Water-Low Low Low, Level 1 will not be required.

The Function is required in MODES 1 and 2 where considerable energy exists in the RCS resulting in the limiting transients and accidents. ECCS initiations at Reactor Vessel Water Level-Low Low, Level 2 and Low Low Low, Level 1 provide sufficient protection for level transients in all other MODES.

(continued)

SUSQUEHANNA - UNIT 2 TS / B 3.3-16 Revision 2

PPL Rev. 3 RPS Instrumentation B 3.3.1.1 BASES APPLICABLE 5. Main Steam Isolation Valve-Closure SAFETY

ANALYSES, LCO, and MSIV closure results in loss of the main turbine and the condenser as APPLICABILITY a heat sink for the nuclear steam supply system and indicates a need (continued) to shut down the reactor to reduce heat generation. Therefore, a reactor scram is initiated on a Main Steam Isolation Valve-Closure signal before the MSIVs are completely closed in anticipation of the complete loss of the normal heat sink and subsequent overpressurization transient. However, for the overpressurization protection analysis of Reference 4, the Average Power Range Monitor Neutron Flux-High Function, along with the S/RVs, limits the peak RPV pressure to less than the ASME Code limits. That is, the direct scram on position switches for MSIV closure events is not assumed in the overpressurization analysis.

Additionally, MSIV closure is assumed in the transients analyzed in Reference 5 (e.g., low steam line pressure, manual closure of MSIVs, high steam line flow). The reactor scram reduces the amount of energy required to be absorbed and, along with the actions of the ECCS, ensures that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46.

MSIV closure signals are initiated from position switches located on each of the eight MSIVs. Each MSIV has two position switches; one inputs to RPS trip system A while the other inputs to RPS trip system B. Thus, each RPS trip system receives an input-from-eight Main Steam Isolation Valve-Closure channels, each consisting of one position switch. The logic for the Main Steam Isolation Valve-Closure Function is arranged such that either the inboard or outboard valve on three or more of the main steam lines must close in order for a scram to occur.

The Main Steam Isolation Valve-Closure Allowable Value is specified to ensure that a scram occurs prior to a significant reduction in steam flow, thereby reducing the severity of the subsequent pressure transient.

Sixteen channels (arranged in pairs) of the Main Steam Isolation Valve-Closure Function, with eight channels in each trip system, are required to be OPERABLE to ensure that no single instrument failure will preclude the scram from this Function on a valid signal. This Function is only required in MODE 1 since, with the MSIVs open and the heat generation rate high, a pressurization transient can occur if the MSIVs close. In addition, the Function is automatically bypassed when the Reactor Mode Switch is not in the Run position. In MODE 2, the heat generation rate is low enough so that the other diverse RPS functions provide sufficient protection.

(continued)

SUSQUEHANNA - UNIT 2 TS / B 3.3-17 Revision 3

PPL Rev. 3 RPS Instrumentation B 3.3.1.1 BASES APPLICABLE 6. Drvwell Pressure-Hiah SAFETY

ANALYSES, LCO, and High pressure in the drywell could indicate a break in the RCPB. A APPLICABILITY reactor scram is initiated to minimize the possibility of fuel damage and (continued) to reduce the amount of energy being added to the coolant and the drywell. The Drywell Pressure-High Function is assumed in the analysis of the recirculation line break (Ref. 6). The reactor scram reduces the amount of energy required to be absorbed and, along with the actions of Emergency Core Cooling Systems (ECCS), ensures that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46.

High drywell pressure signals are initiated from four pressure instruments that sense drywell pressure. The Allowable Value was selected to be as low as possible and indicative of a LOCA inside primary containment.

Four channels of Drywell Pressure-High Function, with two channels in each trip system arranged in a one-out-of-two logic, are required to be OPERABLE to ensure that no single instrument failure will preclude a scram from this Function on a valid signal. The Function is required in MODES 1 and 2 where considerable energy exists in the RCS, resulting in the limiting transients and accidents.

7.a, 7.b. Scram Discharge Volume Water Level -High -

The SDV receives the water displaced by the motion of the CRD pistons during a reactor scram. Should this volume fill to a point where there is insufficient volume to accept the displaced water, control rod insertion would be hindered. Therefore, a reactor scram is initiated while the remaining free volume is still sufficient to accommodate the water from a full core scram. The two types of Scram Discharge Volume Water Level-High Functions are an input to the RPS logic.

No credit is taken for a scram initiated from these Functions for any of the design basis accidents or transients analyzed in the FSAR.

However, they are retained to ensure the scram function remains OPERABLE.

SDV water level is measured by two diverse methods. The level in each of the two SDVs is measured by two float type level switches and two level transmitters with trip units for a total of eight level signals.

The outputs of these devices are arranged so that there is a signal from a level switch and a level transmitter with trip unit to each RPS logic channel. The level measurement instrumentation satisfies the recommendations of Reference 8.

(continued)

SUSQUEHANNA - UNIT 2 TS / B 3.3-18 Revision 3

PPL Rev. 3 RPS Instrumentation B 3.3.1.1 BASES APPLICABLE 7.a, 7.b. Scram Discharge Volume Water Level - High (continued)

SAFETY ANALYSES, LCO, The Allowable Value is chosen low enough to ensure that there is and APPLICABILITY sufficient volume in the SDV to accommodate the water from a full scram.

Four channels of each type of Scram Discharge Volume Water Level-High Function, with two channels of each type in each trip system, are required to be OPERABLE to ensure that no single instrument failure will preclude a scram from these Functions on a valid signal. These Functions are required in MODES 1 and 2, and in MODE 5 with any control rod withdrawn from a core cell containing one or more fuel assemblies, since these are the MODES and other specified conditions when control rods are withdrawn. At all other times, this Function may be bypassed.

8. Turbine Stop Valve-Closure Closure of the TSVs results in the loss of a heat sink that produces reactor pressure, neutron flux, and heat flux transients that must be limited. Therefore, a reactor scram is initiated at the start of TSV closure in anticipation of the transients that would result from the closure of these valves. The Turbine Stop Valve-Closure Function is the primnary scram signal for the turbine trip event analyzed in Reference 5. -For this event, the reactor scram reduces the amount of energy required to be absorbed and, along with the actions -of the End of Cycle Recirculation Pump Trip (EOC-RPT) System, ensures that the MCPR SL is not exceeded. Turbine Stop Valve-Closure signals are initiated from position switches located on each of the four TSVs. Two independent position switches are associated with each stop valve.

One of the two switches provides input to RPS trip system A; the other, to RPS trip system B.

Thus, each RPS trip system receives an input from four Turbine Stop Valve-Closure channels, each consisting of one position switch. The logic for the Turbine Stop Valve-Closure Function is such that three or more TSVs must be closed to produce a scram. This Function must be enabled at THERMAL POWER Ž30% RTP. This is accomplished automatically by pressure instruments sensing turbine first stage pressure. Because an increase in the main turbine bypass flow can affect this function non-conservatively, THERMAL POWER is derived from first stage pressure. The main turbine bypass valves must not cause the trip Function to be bypassed when THERMAL POWER is

Ž30% RTP.

(continued)

SUSQUEHANNA - UNIT 2 TS / B 3.3-19 Revision 2

PPL Rev. 3 RPS Instrumentation B 3.3.1.1 BASES APPLICABLE 8. Turbine Stop Valve-Closure (continued)

SAFETY ANALYSES, LCO, The Turbine Stop Valve-Closure Allowable Value is selected to be and APPLICABILITY high enough to detect imminent TSV closure, thereby reducing the severity of the subsequent pressure transient.

Eight channels (arranged in pairs) of Turbine Stop Valve-Closure Function, with four channels in each trip system, are required to be OPERABLE to ensure that no single instrument failure will preclude a scram from this Function if any three TSVs should close. This Function is required, consistent with analysis assumptions, whenever THERMAL POWER is >_30% RTP. This Function is not required when THERMAL POWER is < 30% RTP since the Reactor Vessel Steam Dome Pressure- High and the Average Power Range Monitor Neutron Flux-High Functions are adequate to maintain the necessary safety margins.

9. Turbine Control Valve Fast Closure, Trip Oil Pressure-Low Fast closure of the TCVs results in the loss of a heat sink that produces reactor pressure, neutron flux, and heat flux transients that must be limited. Therefore, a reactor scram is initiated on TCV fast closure in anticipation of the transients that would result from the closure of these valves. The Turbine Control Valve Fast-Closure, Trip Oil Pressure-Low Function is the primary scram signal for the-generator load rejection event analyzed in Reference 5. For this event, the reactor scram reduc[s the amount of energy required to be absorbed and, along with the actions of the EOC-RPT System, ensures that the MCPR SL is not exceeded.

Turbine Control Valve Fast Closure, Trip Oil Pressure-Low signals are initiated by the electrohydraulic control (EHC) fluid pressure at each control valve. One pressure instrument is associated with each control valve, and the signal from each transmitter is assigned to a separate RPS logic channel. This Function must be enabled at THERMAL POWER _>30% RTP. This is accomplished automatically by pressure instruments sensing turbine first stage pressure. Because an increase in the main turbine bypass flow can affect this function non-conservatively, THERMAL POWER is derived from first stage pressure. The main turbine bypass valves must not cause the trip Function to be bypassed when THERMAL POWER is _>30% RTP.

(continued)

SUSQUEHANNA - UNIT 2 TS / B 3.3-20 Revision 2

PPL Rev. 3 RPS Instrumentation B 3.3.1.1 BASES APPLICABLE 9. Turbine Control Valve Fast Closure, Trip Oil Pressure-Low SAFETY (continued)

ANALYSES, LCO, and APPLICABILITY The Turbine Control Valve Fast Closure, Trip Oil Pressure-Low Allowable Value is selected high enough to detect imminent TCV fast closure.

Four channels of Turbine Control Valve Fast Closure, Trip Oil Pressure-Low Function with two channels in each trip system arranged in a one-out-of-two logic are required to be OPERABLE to ensure that no single instrument failure will preclude a scram from this Function on a valid signal. This Function is required, consistent with the analysis assumptions, whenever THERMAL POWER is > 30% RTP. This Function is not required when THERMAL POWER is < 30% RTP, since the Reactor Vessel Steam Dome Pressure-High and the Average Power Range Monitor Neutron Flux-High Functions are adequate to maintain the necessary safety margins.

10. Reactor Mode Switch-Shutdown Position The Reactor Mode Switch-Shutdown Position Function provides signals, -via the manual scram logic channels, to each of the four RPS logic channels, which are redundant to the automatic protective instrumne ntation channels and provide manual reactor trip capability.

This Functionwas not specifically credited in the accident analysis, but it is retained for the overall redundancy and diversity of the .RPS as required by the NRC approved-licensing basis.

The reactor mode switch is a single switch with four channels, each of which provides input into one of the RPS logic channels.

There is no Allowable Value for this Function, since the channels are mechanically actuated based solely on reactor mode switch position.

Four channels of Reactor Mode Switch-Shutdown Position.

Function, with two channels in each trip system, are available and required to be OPERABLE. The Reactor Mode Switch-Shutdown Position Function is required to be OPERABLE in MODES 1 and 2, and MODE 5 with any control rod withdrawn from a core cell containing one or more fuel assemblies, since these are the MODES and other specified conditions when control rods are withdrawn.

(continued)

SUSQUEHANNA - UNIT 2 TS / B 3.3-21 Revision 2

PPL Rev. 3 RPS Instrumentation B 3.3.1.1 BASES APPLICABLE 11. Manual Scram SAFETY ANALYSES, LCO, The Manual Scram push button channels provide signals, via the and APPLICABILITY manual scram logic channels, to each of the four RPS logic channels, (continued) which are redundant to the automatic protective instrumentation channels and provide manual reactor trip capability. This Function was not specifically credited in the accident analysis but it is retained for the overall redundancy and diversity of the RPS as required by the NRC approved licensing basis.

There is one Manual Scram push button channel for each of the four RPS logic channels. In order to cause a scram it is necessary that at least one channel in each trip system be actuated.

There is no Allowable Value for this Function since the channels are mechanically actuated based solely on the position of the push buttons.

Four channels of Manual Scram with two channels in each trip system arranged in a one-out-of-two logic are available and required to be OPERABLE in MODES 1 and 2, and in MODE 5 with any control rod withdrawn from a core cell containing one or more fuel assemblies, since these are the MODES and other specified conditions when control rods are withdrawn.

ACTIONS A Note has been-provided to modify the ACTIONS related to RPS instrumentation channels. Section 1.3, Completorr-Times, specifies that once a Condition has been. entered, subsequent divisions, subsystems, components, or variables expressed in the Condition, discovered to be inoperable or not within limits, will not result in separate entry into the Condition. Section 1.3 also specifies that Required Actions of the Condition continue to apply for each additional failure, with Completion Times based on initial entry into the Condition.

However, the Required Actions for inoperable RPS instrumentation channels provide appropriate compensatory measures for separate inoperable channels. As such, a Note has been provided that allows separate Condition entry for each inoperable RPS instrumentation channel.

A.1 and A.2 Because of the diversity of sensors available to provide trip signals and the redundancy of the RPS design, an allowable out of service time of (continued)

SUSQUEHANNA - UNIT 2 TS / B 3.3-22 Revision 2

PPL Rev. 3 RPS Instrumentation B 3.3.1.1 BASES ACTIONS A.1 and A.2 (continued) 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> has been shown to be acceptable (Refs. 9, 15 and 16) to permit restoration of any inoperable channel to OPERABLE status.

However, this out of service time is only acceptable provided the associated Function's inoperable channel is in one trip system and the Function still maintains RPS trip capability (refer to Required Actions B.1, B.2, and C.1 Bases). If the inoperable channel cannot be restored to OPERABLE status within the allowable out of service time, the channel or the associated trip system must be placed in the tripped condition per Required Actions A.1 and A.2. Placing the inoperable channel in trip (or the associated trip system in trip) would conservatively compensate for the inoperability, restore capability to accommodate a single failure, and allow operation to continue.

Alternatively, if it is not desired to place the channel (or trip system) in trip (e.g., as in the case where placing the inoperable channel in trip would result in a full scram), Condition D must be entered and its Required Action taken.

As noted, Action A.2 is not applicable for APRM Functions 2.a, 2.b, 2.c, 2.d, or 2.f. Inoperability of one required APRM channel affects both trip systems. For that condition, Required Action A.1 must be satisfied, and is the only action (other than restoring OPERABILITY) that will restore capability to accommodate a single failure.

Inoperability of more than one required APFRM channel of-the-same trip function results in loss of trip capability and entry into Condition C, as well as entry into Condition A for each channel.

B.1 and B.2 -

Condition B exists when, for any one or more Functions, at least one required channel is inoperable in each trip system. In this condition, provided at least one channel per trip system is OPERABLE, the RPS still maintains trip capability for that Function, but cannot accommodate a single failure in either trip system.

Required Actions B.1 and B.2 limit the time the RPS scram logic, for any Function, would not accommodate single failure in both trip systems (e.g., one-out-of-one and one-out-of-one arrangement for a typical four channel Function). The reduced reliability of this logic arrangement was not evaluated in References 9, 15 or 16 for the 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Completion Time. Within the 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> allowance, the associated Function will have all required channels OPERABLE or in trip (or any combination) in one trip system.

(continued)

SUSQUEHANNA-UNIT2 TS / B 3.3-23 Revision 2

PPL Rev. 3 RPS Instrumentation B 3.3.1.1 BASES ACTIONS B.1 and B.2 (continued)

Completing one of these Required Actions restores RPS to a reliability level equivalent to that evaluated in References 9, 15 and 16 which justified a 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> allowable out of service time as presented in Condition A. The trip system in the more degraded state should be placed in trip or, alternatively, all the inoperable channels in that trip system should be placed in trip (e.g., a trip system with two inoperable channels could be in a more degraded state than a trip system with four inoperable channels if the two inoperable channels are in the same Function while the four inoperable channels are all in different Functions). The decision of which trip system is in the more degraded state should be based on prudent judgment and take into account current plant conditions (i.e., what MODE the plant is in).

If this action would result in a scram, it is permissible to place the other trip system or its inoperable channels in trip.

The 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> Completion Time is judged acceptable based on the remaining capability to trip, the diversity of the sensors available to provide the trip signals, the low probability of extensive numbers of inoperabilities affecting all diverse Functions, and the low probability of an-event requiring the initiation of a scram.

Alternately, if itris not desired to place the inoperable channels_(or one trip system) in trip (e.g., as in the case where placing the inoperable channel or associated trip sysfem in trip would result in a scram),

Condition D must be entered and its Required Action taken.

As noted, Condition B is not applicable for APRM Functions 2.a, 2.b, 2.c, 2.d, or 2.f. Inoperability of an APRM channel affects both trip systems and is not associated with a specific trip system as are the APRM 2-out-of-4 Voter (Function 2.e) and other non-APRM channels for which Condition B applies. For an inoperable APRM channel, Required Action A.1 must be satisfied, and is the only action (other than restoring OPERABILITY) that will restore capability to accommodate a single failure. Inoperability of a Function in more than one required APRM channel results in loss of trip capability for that Function and entry into Condition C, as well as entry into Condition A for each channel. Because Conditions A and C provide Required Actions that are appropriate for the inoperability of APRM Functions 2.a, 2.b, 2.c, 2.d, or 2.f, and because (continued)

SUSQUEHANNA - UNIT 2 TS / B 3.3-24 Revision 2

PPL Rev. 3 RPS Instrumentation B 3.3.1.1 BASES ACTIONS B.1 and B.2 (continued) these Functions are not associated with specific trip systems as are the APRM 2-out-of-4 Voter and other non-APRM channels, Condition B does not apply.

C.1 Required Action C.1 is intended to ensure that appropriate actions are taken if multiple, inoperable, untripped channels within the same trip system for the same Function result in the Function not maintaining RPS trip capability. A Function is considered to be maintaining RPS trip capability when sufficient channels are OPERABLE or in trip (or the associated trip system is in trip), such that both trip systems will generate a trip signal from the given Function on a valid signal. For the typical Function with one-out-of-two taken twice logic, this would require both trip systems to have one channel OPERABLE or in trip (or the associated trip system in trip). For Function 5 (Main Steam Isolation Valve-Closure), this would require both trip systems to have each channel associated with the MSIVs in three main steam lines (not necessarily the same main steam lines for both trip systems)

OPERABLE or in trip (or the associated trip- system in trip).

For Fdnction 8 (Turbine Stop Valve-Closure), this would require both trip systems to have three channels, each OPERABLE or-in tdp_.(or the associated trip system in trip).

The Completion Time is intended to allow the operator time to evaluate and repair any-discovered inoperabilities. The 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Completion Time is acceptable because it minimizes risk while allowing time for restoration or tripping of channels.

D._1 Required Action D.1 directs entry into the appropriate Condition referenced in Table 3.3.1.1-1. The applicable Condition specified in the Table is Function and MODE or other specified condition dependent and may change as the Required Action of a previous Condition is completed: Each time an inoperable channel has not met any Required Action of Condition A, B, or C and the associated Completion Time has expired, Condition D will be entered for that channel and provides for transfer to the appropriate subsequent Condition.

(continued)

SUSQUEHANNA -- UNIT 2 TS / B 3.3-25 Revision 2

PPL Rev. 3 RPS Instrumentation B 3.3.1.1 BASES ACTIONS E.1, F.1, G.1, and J.1 (continued)

If the channel(s) is not restored to OPERABLE status or placed in trip (or the associated trip system placed in trip) within the allowed Completion Time, the plant must be placed in a MODE or other specified condition in which the LCO does not apply. The allowed Completion Times- are reasonable, based on operating experience, to reach the specified condition from full power conditions in an orderly manner and without challenging plant systems. In addition, the Completion Time of Required Actions E.1 and J.1 are consistent with the Completion Time provided in LCO 3.2.2, "MINIMUM CRITICAL POWER RATIO (MCPR)."

H.1 If the channel(s) is not restored to OPERABLE status or placed in trip (or the associated trip system placed in trip) within the allowed Completion Time, the plant must be placed in a MODE or other specified condition in which the LCO does not apply. This is done by immediately initiating action to fully insert all insertable control rods in core cells containing one or more fuel assemblies. 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.

1.1 and 1.2 Required Actions 1.1 and 1.2 are intended to ensure that appropriate actions are taken if more than two inoperable or bypassed'OPRM channels result in not maintaining OPRM trip capability.

In the 4-OPRM channel configuration, any 'two' of the OPRM channels out of the total of four and one 2-out-of-4 voter channels in each RPS trip system are required to function for the OPRM safety trip function to be accomplished. Therefore, three OPRM channels assures at least two OPRM channels can provide trip inputs to the 2-out-of-4 voter channels even in the event of a single OPRM channel failure, and the minimum of two 2-out-of-4 voter channels per RPS trip system assures at least one voter channel will be operable per RPS trip system even in the event of a single voter channel failure.

References 15 and 16 justified use of alternate methods to detect and suppress oscillations under limited conditions. The alternate methods are consistent with the guidelines identified in Reference 20. The (continued)

SUSQUEHANNA - UNIT 2 TS / B 3.3-26 . Revision 2

PPL Rev. 3 RPS Instrumentation B 3.3.1.1 BASES ACTIONS 1.1 and 1.2 (continued) alternate-methods procedures require increased operator awareness and monitoring for neutron flux oscillations when operating in the region where oscillations are possible. If operator observes indications of oscillation, as described in Reference 20, the operator will take the actions described by procedures, which include manual scram of the reactor. The power/flow map regions where oscillations are possible are developed based on the methodology in Reference 22. The applicable regions are contained in the COLR.

The alternate methods would adequately address detection and mitigation in the event of thermal hydraulic instability oscillations.

Based on industry operating experience with actual instability oscillations, the operator would be able to recognize instabilities during this time and take action to suppress them through a manual scram.

In addition, the OPRM system may still be available to provide alarms to the operator if the onset of oscillations were to occur.

The 12-hour allowed Completion Time for Required Action 1.1 is based on engineering judgment to allow orderly transition to the alternate methods while limiting the period of time during which no automatic or alternate detect and suppress trip capability is formally in place.

Based on the small probability of an instability event occurring at all, the 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is judged to be reasonable. - . .. ..

The 120-day allowed Completion Time, the time that was evaluated in References 15 and 16, is considered adequate because with operation minimized in regions where oscillations may occur and implementation of the alternate methods, the likelihood of an instability event that could not be adequately handled by the alternate methods during this 120-day period was negligibly small.

The primary purpose of Required Actions 1.1 and 1.2 is to allow an orderly completion, without undue impact on plant operation, of design and verification activities required to correct unanticipated equipment design or functional problems that cause OPRM Trip Function INOPERABILITY in all APRM channels that cannot reasonably be corrected by normal maintenance or repair actions. These Required Actions are not intended and were not evaluated as a routine alternative to returning failed or inoperable equipment to OPERABLE status.

(continued)

SUSQUEHANNA - UNIT 2 TS / [] 3.3-27 Revision 2

PPL Rev. 3 RPS Instrumentation B 3.3.1.1 BASES SURVEILLANCE As noted at the beginning of the SRs, the SRs for each RPS REQUIREMENTS instrumentation Function are located in the SRs column of Table 3.3.1.1-1.

The Surveillances are modified by a Note to indicate that when a 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 RPS trip 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 (Refs. 9, 15 and 16) 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 the RPS will trip when necessary.

SR 3.3.1.1.1 and SR 3.3.1.1.2 Performance of the CHANNEL CHECK ensures that a gross failure of instrumentation has not occurred. A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a similar parameter on other channels. It is based on the assumption that instrumnent channels monitoring the same parameter should read approximately the same value. Significani deviations between_-

instrument channels could be an indicatioh of excessive instrument drift in one of the channels or something even more serious. A CHANNEL CHECK will detect gross channel failure; thus, it is key to verifying the instrumentation continues to operate properly between each CHANNEL CALIBRATION.

Agreement criteria, which are determined by the plant staff based on an investigation of a combination of the channel instrument uncertainties, may be used to support this parameter comparison and include indication and readability. If a channel is outside the criteria, it may be an indication that the instrument has drifted outside its limit, and does not necessarily indicate the channel is Inoperable.

(continued)

SUSQUEHANNA-UNIT2 TS / B 3.3-28 Revision 3

PPL Rev. 3 RPS Instrumentation B 3.3.1.1 BASES SURVEILLANCE SR 3.3.1.1.1 and SR 3.3.1.1.2 (continued)

REQUIREMENTS The Frequency of once every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> for SR 3.3.1.1.1 is based upon operating experience that demonstrates that channel failure is rare.

The Frequency of once every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for SR 3.3.1.1.2 is based upon operating experience that demonstrates that channel failure is rare and the evaluation in References 15 and 16. The CHANNEL CHECK supplements less formal checks of channels during normal operational use of the displays associated with the channels required by the LCO.

SR 3.3.1.1.3 To ensure that the APRMs are accurately indicating the true core average power, the APRMs are calibrated to the reactor power calculated from a heat balance. The Frequency of once per 7 days is based on minor changes in LPRM sensitivity, which could affect the APRM reading between performances of SR 3.3.1.1.8.

A restriction to satisfying this SR when < 25% RTP is provided that requires the SR to be met only at _>25% RTP because it is difficult to accurately maintain APRM indication of core THERMAL POWER consistent with a heat balance when < 25% RTP. At low power levels, a high degree of accuracy is unnecessary because of the large, inhereht margin to thermal limits (MCPR, LHGR and APLHGR). At

Ž 25% RTP, the Surveillance is required to have-been satisfactorily performed within the last 7 days, in accordance with SR 3.0,2. A Note is provided which allows an increase in THERMAL POWER above 25% if the 7 day Frequency is not met per SR 3.0.2. In this event, the SR must be performed within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reaching or exceeding 25% RTP. Twelve hours is based on operating experience and in consideration of providing a reasonable time in which to complete the SR.

(continued)

SUSQUEHANNA - UNIT 2 TS / B 3.3-29 Revision 3

PPL Rev. 3 RPS Instrumentation B 3.3.1.1 BASES SURVEILLANCE SR 3.3.1.1.4 REQUIREMENTS (continued) A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the entire channel will perform the intended function.

As noted, SR 3.3.1.1.4 is not required to be performed when entering MODE 2 from MODE 1, since testing of the MODE 2 required IRM Functions cannot be performed in MODE 1 without utilizing jumpers, lifted leads, or movable links. This allows entry into MODE 2 if the 7 day Frequency is not met per SR 3.0.2. In this event, the SR must be performed within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after entering MODE 2 from MODE 1.

Twelve hours is based on operating experience and in consideration of providing a reasonable time in which to complete the SR.

A Frequency of 7 days provides an acceptable level of system average unavailability over the Frequency interval and is based on reliability analysis (Ref. 9).

SR 3.3.1.1.5 A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the entire channel will perform the intended function. A Frequency of 7 days provides an acceptable leveLQf system average availability over the Frequency and is based on the reliability analysis of Reference-9. (The Manual Scram Function's CHANNEL FUNCTIONAL TEST Frequency was credited in the analysis to extend many automatic scram Functions' Frequencies.)

SR 3.3.1.1.6 and SR 3.3.1.1.7 These Surveillances are established to ensure that no gaps in neutron flux indication exist from subcritical to power operation for monitoring core reactivity status.

The overlap between SRMs and IRMs is required to be demonstrated to ensure that reactor power will not be increased into a neutron flux region without adequate indication. The overlap is demonstrated prior to fully withdrawing the SRMs from the core. Demonstrating the overlap prior to fully withdrawing the SRMs from the core is required to ensure the SRMs are on-scale for the overlap demonstration.

(continued)

SUSQUEHANNA - UNIT 2 TS / B 3.3-30 Revision 3

PPL Rev. 3 RPS Instrumentation B 3.3.1.1 BASES SURVEILLANCE SR 3.3.1.1.6 and SR 3.3.1.1.7 (continued)

REQUIREMENTS The overlap between IRMs and APRMs is of concern when reducing power into the IRM range. On power increases, the system design will prevent further increases (by initiating a rod block) if adequate overlap is not maintained. Overlap between IRMs and APRMs exists when sufficient IRMs and APRMs concurrently have onscale readings such that the transition between MODE 1 and MODE 2 can be made without either APRM downscale rod block, or IRM upscale rod block. Overlap between SRMs and IRMs similarly exists when, prior to fully withdrawing the SRMs from the core, IRMs are above mid-scale on range 1 before SRMs have reached the upscale rod block.

As noted, SR 3.3.1.1.7 is only required to be met during entry into MODE 2 from MODE 1. That is, after the overlap requirement has been met and indication has transitioned to the IRMs, maintaining overlap is not required (APRMs may be reading downscale once in MODE 2).

If overlap for a group of channels is not demonstrated (e.g.,

IRM/APRM overlap), the reason for the failure of the Surveillance should be determined and the appropriate channel(s) declared inoperable. Only those appropriate channels that are required in the current MODE or condition should be declared inoperable.

A Frequency of 7 days is reasonable based on engineering judgment and the reliability of the IRMs and APRMs.

SR 3.3.1.1.8 -

LPRM gain settings are determined from the local flux profiles that are either measured by the Traversing Incore Probe (TIP) System at all functional locations or calculated for TIP locations that are not functional. The methodology used to develop the power distribution limits considers the uncertainty for both measured and calculated local flux profiles. This methodology assumes that all the TIP locations are functional for the first LPRM calibration following a refueling outage, and a minimum of 25 functional TIP locations for subsequent LPRM calibrations. The calibrated LPRMs establish the relative local flux profile for appropriate representative input to the APRM System. The 1000 MWD/MT Frequency is based on operating experience with LPRM sensitivity changes.

(continued)

SUSQUEHANNA - UNIT 2 TS / B 3.3-31 Revision 2

PPL Rev. 3 RPS Instrumentation B 3.3.1.1 BASES SURVEILLANCE S R 3.3.1.1.9 and SR 3.3.1.1.14 REQUIREMENTS (continued) A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the entire channel will perform the intended function. The 92 day Frequency of SR 3.3.1.1.9 is based on the reliability analysis of Reference 9.

SR 3.3.1.1.9 is modified by a Note that provides a general exception to the definition of CHANNEL FUNCTIONAL TEST. This exception is necessary because the design of instrumentation does not facilitate functional testing of all required contacts of the relay, which input into the combinational logic. (Reference 10) Performance of such a test could result in a plant transient or place the plant in an undo risk situation.

Therefore, for this SR, the CHANNEL FUNCTIONAL TEST verifies acceptable response by verifying the change of state of the relay, which inputs into the combinational logic. The required contacts not tested during the CHANNEL FUNCTIONAL TEST are tested under the LOGIC SYSTEM FUNCTIONAL TEST, SR 3.3.1.1.15. This is acceptable because operating experience shows that the contacts not tested during inputs into the combinational logic. The required contacts not tested during the CHANNEL FUNCTIONAL TEST are tested under the LOGIC SYSTEM FUNCTIONAL TEST, SR 3.3.1.1.15.

This i§ acceptable because operating experience shows.that the contacts not tested during the CHANNEL FUNCTIONAL-TESJTU_

normally pass the LOGIC SYSTEM FUNCTIONAL TEST, and the testing methodology minimizes-the risk of unplanned transients.

The 24 month Frequency of SR 3.3.1.1.14 is based on the need to perform 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 that these components usually pass the Surveillance when performed at the 24 month Frequency.

SR 3.3.1.1.10, SR 3.3.1.1.11, SR 3.3.1.1.13, and SR 3.3.1.1.18 A CHANNEL CALIBRATION verifies that the channel responds to the measured parameter within the necessary range and accuracy.

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

(continued)

SUSQUEHANNA - UNIT 2 TS / B 3.3-32 Revision 4

PPL Rev. 3 RPS Instrumentation B 3.3.1.1 BASES SURVEILLANCE SR 3.3.1.1.10, SR 3.3.1.1.11, SR 3.3.1.1.13, and SR 3.3.1.1.18 REQUIREMENTS (continued)

Note 1 for SR 3.3.1.1.18 states that neutron detectors are excluded from CHANNEL CALIBRATION because they are passive devices, with minimal drift, and because of the difficulty of simulating a meaningful signal. Changes in neutron detector sensitivity are compensated for by performing the 7 day calorimetric calibration (SR 3.3.1.1.3) and the 1000 MWD/MT LPRM calibration against the TIPs (SR 3.3.1.1.8).

A Note is provided for SR 3.3.1.1.11 that requires the IRM SRs to be performed within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> of entering MODE 2 from MODE 1. Testing of the MODE 2 APRM and IRM Functions cannot be performed in MODE 1 without utilizing jumpers, lifted leads, or movable links. This Note allows entry into MODE 2 from MODE 1 if the associated Frequency is not met per SR 3.0.2. Twelve hours is based on operating experience and in consideration of providing a reasonable time in which to complete the SR.

A second note is provided for SR 3.3.1.1.18 that requires that the recirculation flow (drive flow) transmitters, which supply the flow signal to the APRMs, be included in the SR for Functions 2.b and 2.f. The APRM Simulated Thermal Power-High Function (Function 2.b) and the OPRM Trip Function (Function 2.f) both require-a valid drive flow signal. The APRM Simulated Thermal Power-High Function uses drive flow to vary the trip setpoint. The OPRM Trip Function uses drive flow to automatically enable or bypass the OPRM Trip output to the RPS.

A CHANNEL CALIBRATION of the APRM drive flow signal requires both calibrating the drive flow transmitters and the processing hardware in the APRM equipment. SR 3.3.1.1.20 establishes a valid drive flow / core flow relationship. Changes throughout the cycle in the drive flow / core flow relationship due to the changing thermal hydraulic operating conditions of the core are accounted for in the margins included in the bases or analyses used to establish the setpoints for the APRM Simulated Thermal Power-High Function and the OPRM Trip Function.

The Frequency of 184 days for SR 3.3.1.1.11, 92 days for SR 3.3.1.1.12 and 24 months for SR 3.3.1.1.13 and SR 3.3.1.1.18 is based upon the assumptions in the determination of the magnitude of equipment drift in the setpoint analysis.

(continued)

SUSQUEHANNA-UNIT2 TS / B 3.3-33 Revision 3

PPL Rev. 3 RPS Instrumentation B 3.3.1.1 BASES SURVEILLANCE SR 3.3.1.1.12 REQUIREMENTS (continued) A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the entire channel will perform the intended function. For the APRM Functions, this test supplements the automatic self-test functions that operate continuously in the APRM and voter channels. The scope of the APRM CHANNEL FUNCTIONAL TEST is that which is necessary to test the hardware.

Software controlled functions are tested as part of the initial verification and validation and are only incidentally tested as part of the surveillance testing. Automatic self-test functions check the EPROMs in which the software-controlled logic is defined. Changes in the EPROMs will be detected by the self-test function and alarmed via the APRM trouble alarm. SR 3.3.1.1.1 for the APRM functions includes a step to confirm that the automatic self-test function is still operating.

The APRM CHANNEL FUNCTIONAL TEST covers the APRM channels (including recirculation flow processing -- applicable to Function 2.b and the auto-enable portion of Function 2.f only), the 2-out-of-4 Voter channels, and the interface connections into the RPS trip systems from the voter channels.

Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology. The 184-day Frequency of SR 3.3.1.1.12 is based on the reliability analyse_._of References 15 and 16. (NOTE: The actual voting logic of the 2-out-of-4 Voter Function is tested as part of SR 3.3.1.1.15. The auto-enable setpoints for the OPRM Trip are confirmed by SR 3.3.1.1.19.)

A Note is provided for Function 2.a that requires this SR to be performed within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> of entering MODE 2 from MODE 1. Testing of the MODE 2 APRM Function cannot be performed in MODE 1 without utilizing jumpers or lifted leads. This Note allows entry into MODE 2 from MODE 1 if the associated Frequency is not met per SR 3.0.2.

A second Note is provided for Functions 2.b and 2.f that clarifies that the CHANNEL FUNCTIONAL TEST for Functions 2.b and 2.f includes testing of the recirculation flow processing electronics, excluding the flow transmitters.

(continued)

SUSQUEHANNA - UNIT 2 TS / B 3.3-34 Revision 2

PPL Rev. 3 RPS Instrumentation B 3.3.1.1 BASES SURVEILLANCE SR 3.3.1.1.15 REQUIREMENTS (continued) The LOGIC SYSTEM FUNCTIONAL TEST demonstrates the OPERABILITY of the required trip logic for a specific channel. The functional testing of control rods (LCO 3.1.3), and SDV vent and drain valves (LCO 3.1.8), overlaps this Surveillance to provide complete testing of the assumed safety function.

The LOGIC SYSTEM FUNCTIONAL TEST for APRM Function 2.e simulates APRM and OPRM trip conditions at the 2-out-of-4 Voter channel inputs to check all combinations of two tripped inputs to the 2-out-of-4 logic in the voter channels and APRM-related redundant RPS relays.

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 that these components usually pass the Surveillance when performed at the 24 month Frequency.

SR 3.3.1.1.16 This SR ensures that scrams initiated from the Turbine Stop Valve-Closure and Turbine Control Valve Fast Closure, TdpOil-Pressure-Low Functions will not be inadvertently bypassed when THERMAL POWER is 2! 30% RTP. This is performerd by a Functional check that ensures the scram feature is not bypassed at > 30% RTP. Because main turbine bypass-flow can affect this function nonconservatively (THERMAL POWER is derived from turbine first stage pressure), the opening of the main turbine bypass valves must not cause the trip Function to be bypassed when Thermal Power is > 30% RTP.

If any bypass channel's trip function is nonconservative (i.e., the Functions are bypassed at > 30% RTP, either due to open main turbine bypass valve(s) or other reasons), then the affected Turbine Stop Valve-Closure and Turbine Control Valve Fast Closure, Trip Oil Pressure-Low Functions are considered inoperable. Alternatively, the bypass channel can be placed in the conservative condition (nonbypass). If placed in the nonbypass condition, this SR is met and the channel is considered OPERABLE.

(continued)

SUSQUEHANNA-UNIT2 TS / B 3.3-34a Revision 0

PPL Rev. 3.

RPS Instrumentation B 3.3.1.1 BASES SURVEILLANCE SR 3.3.1.1.16 (continued)

REQUIREMENTS The Frequency of 24 months is based on engineering judgment and reliability of the components.

SR 3.3.1.1.17 This SR ensures that the individual channel response times are less than or equal to the maximum values assumed in the accident analysis. This test may be performed in one measurement or in overlapping segments, with verification that all components are tested.

The RPS RESPONSE TIME acceptance criteria are included in Reference 11.

RPS RESPONSE TIME for the APRM 2-out-of-4 Voter Function (2.e) includes the APRM Flux Trip output relays and the OPRM Trip output relays of the voter and the associated RPS relays and contactors.

(Note: The digital portion of the APRM, OPRM and 2-out-of-4 Voter channels are excluded from RPS RESPONSE TIME testing because self-testing and calibration checks the time base of the digital electronics. Confirmation of the time base is adequate to assure.

required- response times are met. Neutron detectors are excluded from RPS RESPONSE TIME testing because the principles of detector operation virtually ensure an instantaneous response time. See References 12 arid 13).. . . ...

As noted, neutron detectors are excluded from RPS RESPONSE TIME testing because the principles of detector operation virtually ensure an instantaneous response time.

RPS RESPONSE TIME tests are conducted on an 24 month STAGGERED TEST BASIS. Note 3 requires STAGGERED TEST BASIS Frequency to be determined based on 4 channels per trip system, in lieu of the 8 channels specified in Table 3.3.1.1-1 for the MSIV Closure Function because channels are arranged in pairs.

This Frequency is based on the logic interrelationships of the various channels required to produce an RPS scram signal. The 24 month Frequency is consistent with the typical industry refueling cycle and is based upon plant operating experience, which shows that random failures of instrumentation components causing serious response time degradation, but not channel failure, are infrequent occurrences.

(continued)

SUSQUEHANNA - UNIT 2 TS / B 3.3-34b Revision 0

PPL Rev. 3 RPS Instrumentation B 3.3.1.1 BASES SURVEILLANCE SR 3.3.1.1.17 (continued)

REQUIREMENTS SR 3.3.1.1.17 for Function 2.e confirms the response time of that function, and also confirms the response time of components common to Function 2.e and other RPS Functions. (Reference 14)

Note 3 allows the STAGGERED TEST BASIS Frequency for Function 2.e to be determined based on 8 channels rather than the 4 actual 2-out-of-4 Voter channels. The redundant outputs from the 2-out-of-4 Voter channel (2 for APRM trips and 2 for OPRM trips) are considered part of the same channel, but the OPRM and APRM outputs are considered to be separate channels for application of SR 3.3.1.1.17, so N = 8. The note further requires that testing of OPRM and APRM outputs from a 2-out-of-4 Voter be alternated. In addition to these commitments, References 15 and 16 require that the testing of inputs to each RPS Trip System alternate.

Combining these frequency requirements, an acceptable test sequence is one that:

a. Tests each RPS Trip System interface every other cycle,

b. -Alternates the testing of APRM and OPRM outputs from any spiecific 2-out-of-4 Voter Channel,

c. Alternates between divisions at least every other test cycle.

(continued)

SUSQUEHANNA-UNIT2 TS / B 3.3-34c Revision 0

PPL Rev. 3 RPS Instrumentation B 3.3.1.1 BASES SURVEILLANCE SR 3.3.1.1.17 (continued)

REQUIREMENTS The testing sequence shown in the table below is one sequence that satisfies these requirements.

Function 2.e Testing Sequence for SR 3.3.1.1.17 Voter "Staggering" 24-Month Output Cycle Tested Voter Al Voter A2 Voter 1 Voter B2 RPS Trip Division Output Output Output Output System 1 St OPRM Al OPRM A 1 2nd APRM B1 APRM B 1 3rd OPRM A2 OPRM A 2 4th APRM B2 APRM B 2 5th APRM A1 APRM A 1 6 OPRM B1 OPRM B 1 7 th APRM A2 APRM A 2 8 OPRM B2 " OPRM 5- 2 After 8 cycles,-the sequence repeats.

Each test of an OPRM or APRM output tests each of the redundant outputs from the 2-out-of-4 Voter channel for that Function and each of the corresponding relays in the RPS. Consequently, each of the RPS relays is tested every fourth cycle. The RPS relay testing frequency is twice the frequency justified by References 15 and 16.

(continued)

SUSQUEHANNA - UNIT 2 TS / B 3.3-34d Revision 0

PPL Rev. 3 RPS Instrumentation B 3.3.1.1 BASES SURVEILLANCE SR 3.3.1.1.19 REQUIREMENTS (continued) This surveillance involves confirming the OPRM Trip auto-enable setpoints. The auto-enable setpoint values are considered to be nominal values as discussed in Reference 21. This surveillance ensures that the OPRM Trip is enabled (not bypassed) for the correct values of APRM Simulated Thermal Power and recirculation drive flow.

Other surveillances ensure that the APRM Simulated Thermal Power and recirculation drive flow properly correlate with THERMAL POWER (SR 3.3.1.1.2) and core flow (SR 3.3.1.1.20), respectively.

If any auto-enable setpoint is nonconservative (i.e., the OPRM Trip is bypassed when APRM Simulated Thermal Power >_30% and recirculation drive flow < value equivalent to the core flow value defined in the COLR, then the affected channel is considered inoperable for the OPRM Trip Function. Alternatively, the OPRM Trip auto-enable setpoint(s) may be adjusted to place the channel in a conservative condition (not bypassed). If the OPRM Trip is placed in the not-bypassed condition, this SR is met, and the channel is considered OPERABLE.

For purposes of this surveillance, consistent with Reference 21, the conversion from core flow values defined in the COLR to drive flow values used for this SR can be conservatively determined by a linear scaling assuming that 100% drive flow corresponds to 100 Mlb/hr core flow, with no adjustment made for expected deviations between core flow and drive flow below 100%.

The Frequency of 24 months is based on engineering judgment and reliability of the components.

SR 3.3.1.1.20 The APRM Simulated Thermal Power-High Function (Function 2.b) uses drive flow to vary the trip setpoint. The OPRM Trip Function (Function 2.f) uses drive flow to automatically enable or bypass the OPRM Trip output to RPS. Both of these Functions use drive flow as a representation of reactor core flow. SR 3.3.1.1.18 ensures that the drive flow transmitters and processing electronics are calibrated. This SR adjusts the recirculation drive flow scaling factors in each APRM channel to provide the appropriate drive flow/core flow alignment.

(continued)

SUSQUEHANNA-UNIT2 TS / B 3.3-34e Revision 0

PPL Rev. 3 RPS Instrumentation B 3.3.1.1 BASES SURVEILLANCE SR 3.3.1.1.20 (continued)

REQUIREMENTS The Frequency of 24 months considers that any change in the core flow to drive flow functional relationship during power operation would be gradual and the maintenance of the Recirculation System and core components that may impact the relationship is expected to be performed during refueling outages. This frequency also considers the period after reaching plant equilibrium conditions necessary to perform the test, engineering judgment of the time required to collect and analyze the necessary flow data, and engineering judgment of the time required to enter and check the applicable scaling factors in each of the APRM channels. This timeframe is acceptable based on the relatively small alignment errors expected, and the margins already included in the APRM Simulated Thermal Power - High and OPRM Trip Function trip - enable setpoints.

REFERENCES 1. FSAR, Figure 7.2-1.

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

3. NEDO-23842, "Continuous Control Rod Withdrawal in the Startup Range," April 18, 1978.

4. FSAR, Section 5.2.2.

5. FSAR, Chapter 15.

6. FSAR, Section 6.3.3.

7. Not used.

8. P. Check (NRC) letter to G. Lainas (NRC), "BWR Scram Discharge System Safety Evaluation," December 1, 1980.

9. NEDO-30851-P-A, "Technical Specification Improvement Analyses for BWR Reactor Protection System," March 1988.

SUSQUEHANNA - UNIT 2 TS / B 3.3-34f Revision 0

PPL Rev. 3 RPS Instrumentation B 3.3.1.1 BASES REFERENCES 10. NRC Inspection and Enforcement Manual, Part 9900: Technical (continued) Guidance, Standard Technical Specification 1.0 Definitions, Issue date 12/08/86.

11. FSAR, Table 7.3-28.

12. NEDO-32291-A "System Analyses for Elimination of Selected Response Time Testing Requirements," October 1995.

13. NRC Safety Evaluation Report related to Amendment No. 171 for License No. NPF 14 and Amendment No. 144 License No. NPF 22.

14. NEDO 32291-A, Supplement 1, "System Analyses for the Elimination of Selected Response Time Testing Requirements,"

October 1999.

15. NEDC-32410P-A, "Nuclear Measurement Analysis and Control Power Range Neutron Monitor (NUMAC PRNM) Retrofit Plus Option III Stability Trip Function," October 1995.

16. NEDC-3241 OP-A Supplement 1, "Nuclear Measurement Analysis and Control Power Range Neutron Monitor (NUMAC PRNM)

Retrofit Plus Option III Stability Trip Function," November 1997.

17. NEDO-31960-A, "BWR Owners' Group Long-Term Stability Solutions Licensing Methodology," November 1995.

18. NEDO-31-960-A, Supplement 1, "BWR Owners' Group Long-Term Stability Solutions Licensing Methodology,"

November 1995.

19. NEDO-32465-A, "BWR Owners' Group Long-Term Stability Detect and Suppress Solutions Licensing Basis Methodology and Reload Applications," August 1996.

20. BWROG Letter BWROG 9479, L. A. England (BWROG) to M. J. Virgilio, "BWR Owners' Group Guidelines for Stability Interim Corrective Action," June 6, 1994.

SUSQUEHANNA - UNIT 2 TS / B 3.3-34g Revision 0

PPL Rev. 3 RPS Instrumentation B 3.3.1.1 BASES REFERENCES 21. BWROG Letter BWROG 96113, K. P. Donovan (BWROG)

(continued) to L. E. Phillips (NRC), "Guidelines for Stability Option III

'Enable Region' (TAC M92882)," September 17, 1996.

22. EMF-CC-074(P)(A), Volume 4, "BWR Stability Analysis:

Assessment of STAIF with Input from MICROBURN-B2."

23. GE Letter to PPL, GE-2005-EMC426, "Susquehanna 1 & 2 Minimum LPRM Input Requirement for NUMAC APRM 4-Channel Design," April 26, 2005.

SUSQUEHANNA - UNIT 2 TS / B 3.3-34h Revision 0

PPL Rev. 3 RPS Instrumentation B 3.3.1.1 Table B 3.3.1.1-1 (page 1 of 1)

RPS Instrumentation Sensor Diversity Scram Sensors for Initiating Events RPV Variables Anticipatory Fuel Initiation Events (a) (b) (c) (d) (e) (f) (g)

MSIV Closure X X X X Turbine Trip (w/bypass) - X X X X Generator Trip (w/bypass) X X X Pressure Regulator Failure X X X X X (primary pressure decrease)

(MSIV closure trip)

Pressure Regulator Failure X X X (primary pressure decrease)

(Level 8 trip)

Pressure Regulator Failure X X (primary pressure increase)

Feedwater Controller Failure X X X X (high reactor water level)

Feedwater Controller Failure X X X (low reactor water level)

Loss of Condenser Vacuum - -X X X X Loss of AC Power (loss of X - X X k transformer)

Loss of AC Power (loss of X X X X X X grid connections)

(a) Reactor Vessel Steam Dome Pressure-High (b) Reactor Vessel Water Level-High, Level 8 (c) Reactor Vessel Water Level-Low, Level 3 (d) Turbine Control Valve Fast Closure (e) Turbine Stop Valve-Closure (f) Main Steam Isolation Valve-Closure (g) Average Power Range Monitor Neutron Flux-High SUSQUEHANNA - UNIT 2 TS / B 3.3-34i Revision 0

PPL Rev. 1 SRM Instrumentation B 3.3.1.2 B 3.3 INSTRUMENTATION B 3.3.1.2 Source Range Monitor (SRM) Instrumentation BASES BACKGROUND The SRMs provide the operator with information relative to the neutron flux level at startup and low flux levels in the core. As such, the SRM indication is used by the operator to monitor the approach to criticality and determine when criticality is achieved. The SRMs are not fully withdrawn from the core until the SRM to intermediate range monitor (IRM) overlap is demonstrated (as required by SR 3.3.1.1.6), when the SRMs are normally fully withdrawn from the core.

The SRM subsystem of the Neutron Monitoring System (NMS) consists of four channels. Each of the SRM channels can be bypassed, but only one at any given time, by the operation of a bypass switch. Each channel includes one detector that can be physically positioned in the core. Each detector assembly consists of a miniature fission chamber with associated cabling, signal conditioning equipment, and electronics associated with the various SRM functions. The signal conditioning equipment converts the current pulses from the fission chamber to analog DC currents that correspond to the count rate. Each channel also includes indication, alarm, and control rod blocks. However, this LCO specifies OPERABILITY requirements only for the monitoring and indication functions of the SRMs.

During refueling, shutdown, and low power operations, the primary indication of neutron flux levels is provided by the SRMs or special movable detectors connected to the normal SRM circuits. The SRMs provide monitoring of reactivity changes during fuel or control rod movement and give the control room operator early indication of unexpected subcritical multiplication that could be indicative of an approach to criticality.

APPLICABLE Prevention and mitigation of prompt reactivity excursions during SAFETY ANALYSES refueling and low power operation is provided by LCO 3.9.1, "Refueling Equipment Interlocks"; LCO 3.1.1, "SHUTDOWN MARGIN (SDM)"; LCO 3.3.1.1, "Reactor Protection System (RPS)

Instrumentation"; IRM Neutron Flux-High and Average Power Range Monitor (APRM) Neutron Flux-High (continued)

SUSQUEHANNA-UNIT2 TS / B 3.3-35 Revision 2

PPL Rev. 1 SRM Instrumentation B 3.3.1.2 BASES APPLICABLE (Setdown) Functions; and LCO 3.3.2.1, "Control Rod Block SAFETY ANALYSES Instrumentation."

(continued)

The SRMs have no safety function and are not assumed to function during any FSAR design basis accident or transient analysis.

However, the SRMs provide the only on-scale monitoring of neutron flux levels during startup and refueling. Therefore, they are being retained in Technical Specifications.

LCO During startup in MODE 2, three of the four SRM channels are required to be OPERABLE to monitor the reactor flux level prior to and during control rod withdrawal, subcritical multiplication and reactor criticality, and neutron flux level and reactor period until the flux level is sufficient to maintain the IRMs on Range 3 or above. All but one of the channels are required in order to provide a representation of the overall core response during those periods when reactivity changes are occurring throughout the core.

In MODES 3 and 4, with the reactor shut down, two SRM channels provide redundant monitoring of flux levels in the core.

In MODE 5, during a spiral offload or reload, an SRM outside the fueled region will no longer be required to be OPERABLE, since it is not capable of monitoring neutron flux in the fueled-region of the core.

Fueled region is a continuous area with fuel. Thus, CORE.-

ALTERATIONS are allowed in a quadrant with no OPERABLE SRM in an adjacent quadrant provided the Table 3.3.1.2-1, footnote (b),

requirement that the bundles being spiral reloaded or spiral offloaded are all in a single fueled region containing at least one OPERABLE SRM is met. Spiral reloading and offloading encompass reloading or offloading a cell on the edge of a continuous fueled region (the cell can be reloaded or offloaded in any sequence).

In nonspiral routine operations, two SRMs are required to be OPERABLE to provide redundant monitoring of reactivity changes occurring in the reactor core. Because of the local nature of reactivity changes during refueling, adequate coverage is provided by requiring one SRM to be OPERABLE in (continued)

SUSQUEHANNA - UNIT 2 TS / B 3.3-36 Revision 2

PPL Rev. 1 SRM Instrumentation B 3.3.1.2 BASES LCO the quadrant of the reactor core where CORE ALTERATIONS are (continued) being performed, and the other SRM to be OPERABLE in an adjacent quadrant containing fuel. These requirements ensure that the reactivity of the core will be continuously monitored during CORE ALTERATIONS.

Special movable detectors, according to footnote (c) of Table 3.3.1.2-1, may be used during CORE ALTERATIONS in place of the normal SRM nuclear detectors. These special detectors must be connected to the normal SRM circuits in the NMS, such that the applicable neutron flux indication can be generated. These special detectors provide more flexibility in monitoring reactivity changes during fuel loading, since they can be positioned anywhere within the core during refueling. They must still meet the location requirements of SR 3.3.1.2.2 and all other required SRs for SRMs.

For an SRM channel to be considered OPERABLE, it must be providing neutron flux monitoring indication.

APPLICABILITY The SRMs are required to be OPERABLE in MODES 2, 3, 4, and 5 prior to the IRMs being on scale on Range 3 to provide for neutron monitoring. In MODE 1, the APRMs provide adequate monitoring of reactivity changes in the core; therefore, the SRMs are not required.

In MODE 2, with IRMs on Range 3 or above, the IRMs provide.-_

adequate monitoring and the SRMs are not required.

ACTIONS A.1 and B.1 In MODE 2, with the IRMs on Range 2 or below, SRMs provide the means of monitoring core reactivity and criticality. With any number of the required SRMs inoperable, the ability to monitor neutron flux is degraded. Therefore, a limited time is allowed to restore the inoperable channels to OPERABLE status.

Provided at least one SRM remains OPERABLE, Required Action A.1 allows 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> to restore the required SRMs to OPERABLE status.

This time is reasonable because there is adequate capability remaining to monitor the core, there is (continued)

SUSQUEHANNA -UNIT 2 TS / B 3.3-37 Revision 1

PPL Rev. 1 SRM Instrumentation B 3.3.1.2 BASES ACTIONS A.1 and B.1 (continued) limited risk of an event during this time, and there is sufficient time to take corrective actions to restore the required SRMs to OPERABLE status or to establish alternate IRM monitoring capability. During this time, control rod withdrawal and power increase is not precluded by this Required Action. Having the ability to monitor the core with at least one SRM, proceeding to IRM Range 3 or greater (with overlap required by SR 3.3.1.1.6), and thereby exiting the Applicability of this LCO, is acceptable for ensuring adequate core monitoring and allowing continued operation.

With three required SRMs inoperable, Required Action B.1 allows no positive changes in reactivity (control rod withdrawal must be immediately suspended) due to inability to monitor the changes.

Required Action A.1 still applies and allows 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> to restore monitoring capability prior to requiring control rod insertion. This allowance is based on the limited risk of an event during this time, provided that no control rod withdrawals are allowed, and the desire to concentrate efforts on repair, rather than to immediately shut down, with no SRMs OPERABLE.

C.1 In MODE 2, if the required number of SRMs is not restored to OPERABLE status within the a/lowed Completion Time, the reactor shall be placed in MODE 3. With all control rods fully inserted, the core is in its least reactive state with the most margin to criticality. The allowed Completion Time of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is reasonable, based on operating experience, to reach MODE 3 from full power conditions in an orderly manner and without challenging plant systems.

D.1 and D.2 With one or more required SRMs inoperable in MODE 3 or 4, the neutron flux monitoring capability is degraded or nonexistent. The requirement to fully insert all insertable control rods ensures that the reactor will be at its minimum reactivity level while no neutron monitoring capability is available. Placing the reactor mode switch in the shutdown position prevents subsequent control rod withdrawal by (continued)

SUSQUEHANNA - UNIT 2 TS / B 3.3-38 Revision 1

PPL Rev. 1 SRM Instrumentation B 3.3.1.2 BASES ACTIONS D.1 and D.2 (continued) maintaining a control rod block. 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 sufficient to accomplish the Required Action, and takes into account the low probability of an event requiring the SRM occurring during this interval.

E.1 and E.2 With one or more required SRM inoperable in MODE 5, the ability to detect local reactivity changes in the core during refueling is degraded.

CORE ALTERATIONS must be immediately suspended and action must be immediately initiated to insert all insertable control rods in core cells containing one or more fuel assemblies. Suspending CORE ALTERATIONS prevents the two most probable causes of reactivity changes, fuel loading and control rod withdrawal, from occurring.

Inserting all insertable control rods ensures that the reactor will be at its minimum reactivity given that fuel is present in the core.

Suspension of CORE ALTERATIONS shall not preclude completion of the movement of a component to a safe, conservative position.

Action (once required to be initiated) to insert control rods must continue until all insertable rods in core cells containing one or more fuel assemblies are inserted.. .. .. .

SURVEILLANCE The SRs for each SRM Applicable MODE or other specified conditions REQUIREMENTS are found in the SRs column of Table 3.3.1.2-1.

SR 3.3.1.2.1 and SR 3.3.1.2.3 Performance of the CHANNEL CHECK ensures that a gross failure of instrumentation has not occurred. A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a similar parameter on another channel. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value. Significant deviations between the instrument channels could be an indication of excessive instrument drift in one of the channels or something even more serious. A CHANNEL CHECK will detect gross channel failure; thus, it (continued)

SUSQUEHANNA - UNIT 2 TS / B 3.3-39 Revision 1

PPL Rev. 1 SRM Instrumentation B 3.3.1.2 BASES SURVEILLANCE SR 3.3.1.2.1 and SR 3.3.1.2.3 (continued)

REQUIREMENTS is key to verifying the instrumentation continues to operate properly between each CHANNEL CALIBRATION.

Agreement criteria which are determined by the plant staff based on an investigation of a combination of the channel instrument uncertainties may be used to support this parameter comparison and include indication and readability. If a channel is outside the criteria, it may be an indication that the instrument has drifted outside its limit, and does not necessarily indicate the channel is Inoperable.

The Frequency of once every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> for SR 3.3.1.2.1 is based on operating experience that demonstrates channel failure is rare. While in MODES 3 and 4, reactivity changes are not expected; therefore, the 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Frequency is relaxed to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for SR 3.3.1.2.3. The CHANNEL CHECK supplements less formal checks of channels during normal operational use of the displays associated with the channels required by the LCO.

SR -3.3.1.2.2 To pro-vide adequate coverage of potential. reactivity changes in the core, a maximum- of two SRMs are required to be OPERABLE.One SRM is required to be OPERABLE in the quadrant where CORE ALTERATIONS are being performed, and the other OPERABLE SRM must be in an adjacent quadrant containing fuel. However, in accordance with Table 3.3.1.2-1, only one SRM is required during a spiral reload until the fueled region is large enough to encompass a second installed SRM. Note 1 states that the SR is required to be met only during CORE ALTERATIONS. It is not required to be met at other times in MODE 5 since core reactivity changes are not occurring. This Surveillance consists of a review of plant logs to ensure that SRMs required to be OPERABLE for given CORE ALTERATIONS are, in fact, OPERABLE. In the event that only one SRM is required to be OPERABLE, per Table 3.3.1.2-1, footnote (b), only the a. portion of this SR is required. Note 2 clarifies that more than one of the three requirements can be met by the same OPERABLE SRM. The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Frequency is based upon operating experience and supplements operational controls over refueling activities (continued)

SUSQUEHANNA - UNIT 2 TS / B 3.3-40 Revision 1

PPL Rev. 1 SRM Instrumentation B 3.3.1.2 BASES SURVEILLANCE SR 3.3.1.2.2 (continued)

REQUIREMENTS that include steps to ensure that the SRMs required by the LCO are in the proper quadrant.

SR 3.3.1.2.4-This Surveillance consists of a verification of the SRM instrument readout to ensure that the SRM reading is greater than a specified minimum count rate, which ensures that the detectors are indicating count rates indicative of neutron flux levels within the core. The signal-to-noise ratio shown in Figure 3.3.1.2-1 is the SRM count rate at which there is a 95% probability that the SRM signal indicates the presence of neutrons and only a 5% probability that the SRM signal is a result of noise (Ref. 1). With few fuel assemblies loaded, the SRMs will not have a high enough count rate to satisfy the SR. Therefore, allowances are made for loading sufficient "source" material, in the form of irradiated fuel assemblies, to establish the minimum count rate.

To accomplish this, the SR is modified by a Note that states that the count rate is not required to be met on an SRM that has less than or equal to four fuel assemblies adjacent to the SRM and no other fuel assemblie s are in the associated core quadrant. With four or less fuel assemblies loaded around each SRM and no other-fuel assemblies in the associated core quadrant, even with a control rod withdrawn, the configuration will not be crtical.- The signal to noise ratio is only required to be determined every 7 or 31 days per the requirements of SR 3.3.1.2.5 OF 3.3.1.2.6.

The Frequency is based upon channel redundancy and other information available in the control room, and ensures that the required channels are frequently monitored while core reactivity changes are occurring. When no reactivity changes are in progress, the Frequency is relaxed from 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

SR 3.3.1.2.5 and SR 3.3.1.2.6 Performance of a CHANNEL FUNCTIONAL TEST demonstrates the associated channel will function properly. SR 3.3.1.2.5 is (continued)

SUSQUEHANNA - UNIT 2 TS / B 3.3-41 Revision 1

PPL Rev. 1 SRM Instrumentation B 3.3.1.2 BASES SURVEILLANCE SR 3.3.1.2.5 and SR 3.3.1.2.6 (continued)

REQUIREMENTS required in MODE 5, and the 7 day Frequency ensures that the channels are OPERABLE while core reactivity changes could be in progress. This Frequency is reasonable, based on operating experience and on other Surveillances (such as a CHANNEL CHECK),

that ensure proper functioning between CHANNEL FUNCTIONAL TESTS.

SR 3.3.1.2.6 is required in MODE 2 with IRMs on Range 2 or below, and in MODES 3 and 4. Since core reactivity changes do not normally take place, the Frequency has been extended from 7 days to 31 days.

The 31 day Frequency is based on operating experience and on other Surveillances (such as CHANNEL CHECK) that ensure proper functioning between CHANNEL FUNCTIONAL TESTS.

Verification of the signal to noise ratio also ensures that the detectors are inserted to an acceptable operating level. In a fully withdrawn condition, the detectors are sufficiently removed from the fueled region of the core to essentially eliminate neutrons from reaching the detector. Any count rate obtained while the detectors are fully withdrawn is assumed to be "noise" only.

The Note to the Surveillance allows the Surveillance to be delayed until entry into the specified condition of the Applicability (THERMAL POWER decreased to IRM Range 2 or below). The SR must be performed within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after IRMs are on Range 2 or below. The allowance to enter the Applicability with the 31 day Frequency not met is reasonable, based on the limited time of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> allowed after entering the Applicability and the inability to perform the Surveillance while at higher power levels. Although the Surveillance could be performed while on IRM Range 3, the plant would not be expected to maintain steady state operation at this power level. In this event, the 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Frequency is reasonable, based on the SRMs being otherwise verified to be OPERABLE (i.e., satisfactorily performing the CHANNEL CHECK) and the time required to perform the Surveillances.

(continued)

SUSQUEHANNA - UNIT 2 TS / B 3.3-42 Revision 1

PPL Rev. 1 SRM Instrumentation B 3.3.1.2 BASES SURVEILLANCE SR 3.3.1.2.7 REQUIREMENTS (continued) Performance of a CHANNEL CALIBRATION at a Frequency of 24 months verifies the performance of the SRM detectors and associated circuitry. The Frequency considers the plant conditions required to perform the test, the ease of performing the test, and the likelihood of a change in the system or component status. The neutron detectors are excluded from the CHANNEL CALIBRATION because they cannot readily be adjusted. The detectors are fission chambers that are designed to have a relatively constant sensitivity over the range and with an accuracy specified for a fixed useful life.

Note 2 to the Surveillance allows the Surveillance to be delayed until entry into the specified condition of the Applicability. The SR must be performed in MODE 2 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> of entering MODE 2 with IRMs on Range 2 or below. The allowance to enter the Applicability with the 24 month Frequency not met is reasonable, based on the limited timeof 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> allowed after entering the Applicability and the inability to perform the Surveillance while at higher power levels. Although the Surveillance could be performed while on IRM Range 3, the plant would not be expected to maintain steady state operation at this power level. In this event, the 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Frequency is reasonable, based on the-SRMs being otherwise verified to be OPERABLE (i.e., satisfactorily performing the CHANNEL CHECK) and the time required to perform the Surveillances.

REFERENCES 1. General Electric Service Information Letter (SIL) 478 "SRM Minimum-Count Rate" dated December 16, 1988.

(continued)

SUSQUEHANNA - UNIT 2 TS / B 3.3-43 Revision 1

PPL Rev. 2 Control Rod Block Instrumentation B 3.3.2.1 B 3.3 INSTRUMENTATION B 3.3.2.1 Control Rod Block Instrumentation 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 Nominal Trip Setpoint (NTSP) is a predetermined setting for a protective device chosen to ensure automatic actuation prior to the process variable reading the Analytical Limit and thus ensuring that the Safety Limit (SL) would not be exceeded. The NTSP accounts for various uncertainties. As such, the NTSP meets the definition of a Limfting Safety System Setting (LSSS), because the protective instrurment channel activates to protect a reactor core or RCS pressure boundary Safety-Limit. Rod Block Monitor'functions la, l1b and-.lc are LSSSs.

Technical Specification contain values related to the OPERABILITY of equipment required for safe operation of the facility. OPERABLE is defined in Technical Specifications as "... being capable of performing its specified safety function(s)." For automatic protective devices related to SLs, the required safety function is to ensure that a SL is not exceeded and therefore the NTSP is the LSSS, as defined by 10 CFR 50.36. However, use of the NTSP to define OPERABILITY in Technical Specifications would be an overly restrictive requirement if it were applied as an OPERABILITY limit for the "as-found" value during a Surveillance. This would result in Technical Specification compliance problems, as well as reports and corrective actions required by the rule, which are not necessary to ensure safety.

(continued)

SUSQUEHANNA - UNIT 2 TS / B 3.3-44 Revision 3

PPL Rev. 2 Control Rod Block Instrumentation B 3.3.2.1 BASES BACKGROUND Use of the NTSP to define "as-found" OPERABILITY under the (continued) expected circumstances described above would result in actions required by both rule and Technical Specifications that are not warranted. However, there is also some point beyond which the device would have not been able to perform its function due, for example, to greater than expected drift. This value needs to be specified in the Technical Specifications in order to define OPERABILITY of the devices and is designated as the Allowable Value, which is the least conservative value of the as-found setpoint that a channel can have during testing.

The Allowable Value specified in SR 3.3.2.1.7 is the least conservative value of the as-found setpoint that a channel can have when tested, such that a channel is OPERABLE if the setpoint is found conservative with respect to the Allowable Value during the CHANNEL CALIBRATION.

The purpose of the RBM is to limit control rod withdrawal if localized neutron flux exceeds a predetermined setpoint during control rod manipulations. It is assumed to function to block further control rod withdrawal to preclude MCPR Safety Limit violation. The RBM supplies a trip signal to the Reactor Manual Control System (RMCS) to appropriately inhibit control rod withdrawal during power operation abdve-the lowpower 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 simulated thermal power signal from one of the four redundant Average Power Range Monitor (APRM) channels supplies a reference signal for one of the RBM channels and a simulated thermal power signal from another of the APRM channels supplies the reference signal to the second RBM channel. This reference signal is used to determine which RBM range setpoint (low, intermediate, or high) is enabled. If the APRM simulated thermal 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).

(continued)

SUSQUEHANNA - UNIT 2 TS / B 3.3-45 Revision 3

PPL Rev. 2 Control Rod Block Instrumentation B 3.3.2.1 BASES BACKGROUND The purpose of the RWM is to control rod patterns during startup, such (continued) 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 insert blocks when the actual sequence deviates beyond allowances 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 -lirminate the possibility of a CRDA, administrative controls require that any partially inserted control rods, which have not been confirmed to be coupled sipce their last withdrawal, be fully irnserted-prior-to reaching the THERMAL POWER of *10% RTP. If a controf-rod has been checked for coupling at notch 48 and the rod has not been moved inward, this rod is in contact with its drive and is not required to be fully inserte-d 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.

(continued)

SUSQUEHANNA - UNIT 2 TS / B 3.3-46 Revision 3

PPL Rev. 2 Control Rod Block Instrumentation B 3.3.2.1 BASES (continued)

BACKGROUND With the reactor mode switch in the shutdown position, a control rod (continued) 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.

Allowable Values are specified for each applicable Rod Block Function listed in Table 3.3.2.1-1. The NTSPs (actual trip setpoints) are selected to ensure that the setpoints are conservative with respect to the Allowable Value. A channel is inoperable if its actual trip setpoint is non-conservative with respect to its required Allowable Value.

NTSPs 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 Analytical Limits are derived from the limiting values of the process parameters obtained from the safety analysis. The Allowable Values are derived from the Analytical Limits, corrected for calibration, process, and some of the instrument errors.

The. NTSPs are then determined, accounting for the remaining channel uncertainties. The trip setpoints derived in this manner provide adequate protection because instrumentation uncertainties, ppcess effects, calibration tolerances, and drift are accounted for.

The specific Applicable Safety Analyses, LCO, and Applicability discussions are listed below on a Function by Function basis.

APPLICABLE 1. Rod Block Monitor SAFETY

ANALYSES, LCO, and The RBM is designed to prevent violation of the MCPR SL and the APPLICABILITY cladding 1% strain fuel design limit that may result from a Single Control Rod Withdrawal (RWE) event.

The analytical methods and assumptions used in evaluating the RWE event are summarized in Reference 14. The fuel thermal performance as a function of RBM Allowable Value is determined from the analysis.

The NTSP and Allowable Values are chosen as a function of power level. NTSP operating limits are established based on the specified Allowable Values.

(continued)

SUSQUEHANNA - UNIT 2 TS / B 3.3-47 Revision 3

PPL Rev. 2 Control Rod Block Instrumentation B 3.3.2.1 BASES (continued)

APPLICABLE 1. Rod Block Monitor (continued)

SAFETY

ANALYSES, LCO, and The RBM function satisfies Criterion 3 of the NRC Policy Statement APPLICABILITY (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.

Nominal setpoints are those predetermined values of output at which an action should take place. The trip setpoints are compared to the actual process parameter, the calculated RBM flux (RBM channel signal). When the normalized RBM flux value exceeds the applicable trip. setpoint, the RBM provides a trip output.

Th6 analytic limits are derived from the limiting values of the process parameters. Using the GE setpoint methodology, based-on ISA RP 67.04, Part II "Methodologies for the Determinatiorr of Setpointsfor Nuclear Safety-Related Instrumentation" setpoint calculation Method 2, the RBM Allowable Values are determined from the analytical limits using the statistical combination of the RBM input signal calibration error, process measurement error, primary element accuracy and instrument accuracy under trip conditions. Accounting for these errors assures that a setpoint found during calibration at the Allowable Value has adequate margin to protect the analytical limit thereby protecting the Safety Limit.

For the digital RBM, there is a normalization process initiated upon rod selection, so that only RBM input signal drift over the interval from the rod selection to rod movement needs to be considered in determining the nominal trip setpoints. The RBM has no drift characteristic with no as-left or as-found tolerances since it only performs digital calculations on the digitized input signals provided by the APRMs.

(continued)

SUSQUEHANNA - UNIT 2 TS / B 3.3-48 Revision 3

PPL Rev. 2 Control Rod Block Instrumentation B 3.3.2.1 BASES APPLICABLE 1. Rod Block Monitor (continued)

SAFETY

ANALYSES, LCO, and The RBM Allowable Value demonstrates that the analytical limit would APPLICABILITY not be exceeded, thereby protecting the safety limit. The nominal trip setpoints and Allowable Values determined in this manner provide adequate protection because instrumentation uncertainties, process effects, calibration tolerances, instrument drift, and the environment errors are accounted for and appropriately applied for the RBM. There are no margins applied to the RBM nominal trip setpoint calculations which could mask RBM degradation.

The RBM will function when operating greater than or equal to 28%

RTP. Below this power level, the RBM is not required to be OPERABLE.

The RBM selects one of three different RBM flux trip setpoints to be applied based on the current value of THERMAL POWER. THERMAL POWER is indicated to each RBM channel by a simulated thermal power (STP) reference signal input from an associated reference APRM channel. The OPERABLE range is divided into three "power ranges," a "low power range," an "intermediate power range," and a "high power range." The RBM flux trip setpoint applied within each of the.*e three power ranges is, respectively, the "low trip setpoint," the "intermediate trip -setpoint," and the "high tip setpoint" (Allowable Values for which are defined in the COLR). To-determine the-cUrrent power range, each RBM channel compares its current STP input value to three power setpoints, the "low power setpoint",(28%), the "intermediate power setpoint" (current value defined in the COLR), and the "high power setpoint" (current value defined in the COLR), which define, respectively, the lower limit of the low power range, the lower limit of the intermediate power range, and the lower limit of the high power range. The trip setpoint applicable for each power range is more restrictive than the corresponding setpoint for the lower power range(s). When STP is below the low power setpoint, the RBM flux trip outputs are automatically bypassed but the low trip setpoint continues to be applied to indicate the RBM flux setpoint on the NUMAC RBM displays.

The calculated setpoints and applicable power ranges are bounding values. In the equipment implementation, it is necessary to apply a "deadband" to each setpoint. The deadband is applied to the RBM trip setpoint selection logic and the RBM trip automatic bypass logic such (continued)

SUSQUEHANNA - UNIT 2 TS / B 3.3-49 Revision 3

PPL Rev. 2 Control Rod Block Instrumentation B 3.3.2.1 BASES APPLICABLE 1. Rod Block Monitor (continued)

SAFETY

ANALYSES, LCO, and that the setpoint being applied is always equal to or more conservative APPLICABILITY than the required setpoint. Since the RBM flux trip setpoint applicable to the higher power ranges are more conservative than the corresponding trip setpoints for lower power ranges, the trip setpoint applicable to the higher power range (high power range or intermediate power range) continues to be applied when STP decreases below the lower limit of that range until STP is below the power range setpoint by a value exceeding the deadband. Similarly, when STP decreases below the low power setpoint, the automatic bypass of RBM flux trip outputs will not be applied until STP decreases below the trip setpoint a value exceeding the deadband.

The RBM channel uses THERMAL POWER, as represented by the STP input value from its reference APRM channel, to automatically enable RBM flux trip outputs (remove the automatic bypass) and to select the RBM flux trip setpoint to be applied. However, the RBM Upscale function is only required to be OPERABLE when the MCPR values are less than the values defined in the COLR, depending on the.THERMAL POWER level. Therefore, even though the RBM Upscale Function is implemented in each RBM channel as a single trip funZction with a selected trip setpoint, it is characterized in Table 3.3.2.1-1-as three Functions, the Low Power Range - Upscale Function, the Intermediate Power Range - UpsCale Function,-and the High Power Range - Upscale.Eunction, to facilitate correct definition of the OPERABILITY requirements for the Functions. Each Function corresponds to one of the RBM power ranges. Due to the deadband effects on the determination of the current power range, the transition between these three Functions will occur at slightly different THERMAL POWER levels for increasing power versus decreasing power.

2. Rod Worth Minimizer 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 Pattern Control."

(continued)

SUSQUEHANNA - UNIT 2 TS / B 3.3-50 Revision 3

PPL Rev. 2 Control Rod Block Instrumentation B 3.3.2.1 BASES APPLICABLE 2. Rod Block Minimizer (continued)

SAFETY

ANALYSES, LCO, and When performing a shutdown of the plant, an optional BPWS control APPLICABILITY 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 acontrol rod pattern not in compliance with the BPWS. The RWM mag be bypassed as required by these conditions, but then it must be considered inoperable and the Required Actions of this LCO followed.

Compliance with the BPWS, and therefore OPERABILITY of the RWM, is required in MODES 1 and 2 when THERMAL POWER is

< 10% RTP. When THERMAL POWER is > 10% RTP, there is no possible control rod configuration that results in a control rod worth that 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.

(continued)

SUSQUEHANNA - UNIT 2 TS / B 3.3-51 Revision 3

PPL Rev. 2 Control Rod Block Instrumentation B 3.3.2.1 BASES (continued)

APPLICABLE 3. Reactor Mode Switch-Shutdown Position SAFETY

ANALYSES, (continued) 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. _

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 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is based on the low probability of an event occurring coincident with a failure in the remaining OPERABLE channel.

(continued)

SUSQUEHANNA-UNIT2 TS / B 3.3-52 Revision 3

PPL Rev. 2 Control Rod Block Instrumentation B 3.3.2.1 BASES ACTIONS B.1 (continued)

If Required Action A.1 is not met and the associated Completion Time has expired, the inoperable channel must be placed in trip within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 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 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 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.

Alternatively, startup may continue if at least 12 control rods have already been withdrawn, or a reactor startup with an inoperable RWM wa§ 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 reyiew of plant logs and control ro-onrrindicatiorns-.-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 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.

(continued)

SUSQUEHANNA - UNIT 2 TS / B 3.3-53 Revision 3

PPL Rev. 2 Control Rod Block Instrumentation B 3.3.2.1 BASES (continued)

ACTIONS D. 1 (continued)

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 enscired 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.

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.

(continued)

SUSQUEHANNA - UNIT 2 TS / B 3.3-54 Revision 3

PPL Rev. 2 Control Rod Block Instrumentation B 3.3.2.1 BASES (continued)

SURVEILLANCE The Surveillances are modified by a Note to indicate that when an REQUIREMENTS RBM channel is placed in an inoperable status solely for performance (continued) 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 Refs. 9, 12, and 13 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 184 days is based on reliability analyses (Refs. 8, 12, and 13).

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 / B 3.3-54a Revision 0

PPL Rev. 2 Control Rod Block Instrumentation B 3.3.2.1 BAS ES SURVEILLANCE SR 3.3.2.1.4 REQUIREMENTS (continued) The RBM setpoints are automatically varied as a function of Simulated Thermal Power. Three control rod block Allowable Values are specified in Table 3.3.2.1-1, each within a specific power range. The power at which the control rod block Allowable Values automatically change are based on the APRM signal's input to each RMB channel.

Below the minimum power setpoint, the RBM is automatically bypassed. These control rod block NTSPs must be verified periodically to be less than or equal to the specified Allowable Values.

If any power range setpoint is non-conservative, then the affected RBM channel is 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 3.3.1.1.3 and SR 3.3.1.1.8. The 24 month Frequency is based on the actual trip setpoint methodology utilized for these channels.

SR 3.3.2.1.5 The- RWM is automatically bypassed when power is above a specified valje.. 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 / B 3.3-54b Revision 0

PPL Rev. 2 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 detecfors 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 / B 3.3-54c Revision 0

PPL Rev. 2 Control Rod Block Instrumentation B 3.3.2.1 BASES SURVEILLANCE SR 3.3.2.1.7 (continued)

REQUIREMENTS SR 3.3.2.1.7 for the RBM Functions is modified by two Notes as identified in Table 3.3.2.1-1. The RBM Functions are Functions that are LSSSs for reactor core Safety Limits. The first Note requires evaluation of channel performance for the condition where the as-found setting for the channel setpoint is not the NTSP but is conservative with respect to the Allowable Value. For digital channel components, no as-found tolerance or as-left tolerance can be specified. Evaluation of instrument performance will verify that the instrument will continue to behave in accordance with design-basis assumptions. The purpose of the assessment is to ensure confidence in the instrument performance prior to returning the instrument to service. These channels will also be identified in the Corrective Action Program. Entry into the Corrective Action Program will ensure required review and documentation of the condition for continued OPERABILITY. The second Note requires that the as-left setting for the instrument be returned to the NTSP. If the as-left instrument setting cannot be returned to the NTSP, then the instrument channel shall be declared inoperable. The second note also requires that the NTSP and NTSP methodology are to be contained in a document controlled by 10 CFR 50.59.

SR-3.3.2.1.8 The RWM will only enforce the proper control rod sequence ifthe 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.

(continued)

SUSQUEHANNA - UNIT 2 TS / B 3.3-54d Revision 0

PPL Rev. 2 Control Rod Block Instrumentation B 3.3.2.1 BASES REFERENCES 4. "Modifications to the Requirements for Control Rod Drop Accident (continued) 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 3309-1-A, Revision 2, "Improved BPWS Control Rod Insertion Process," April 2003. - .

12. NEDC 3241OP-A, "Nuclear Measurement Analysis and Control Power Range Neutron Monitor (NUMAC PRNM) Retrofit Plus Option III Stability Trip Function," October 1995.

13. NEDC 32410P-A, Supplement 1, "Nuclear Measurement Analysis and Control Power Range Neutron Monitor (NUMAC PRNM)

Retrofit Plus Option III Stability Trip Function," November 1997.

14. XN-NF-90-10(P)(A) Volume 4, Revision 1, "Exxon Nuclear Methodology for Boiling Water Reactors: Application of the ENC Methodology to BWR Reloads," Exxon Nuclear Company, June 1986.

SUSQUEHANNA - UNIT 2 TS / B 3.3-54e Revision 0

PPL Rev. 4 PAM Instrumentation B 3.3.3.1 B 3.3 INSTRUMENTATION B 3.3.3.1 Post Accident Monitoring (PAM) Instrumentation BASES BACKGROUND The primary purpose of the PAM instrumentation is to display plant variables that provide information required by the control room operators during accident situations. This information provides the necessary support for the operator to take the manual actions for which no automatic control is provided and that are required for safety systems to accomplish their safety functions for Design Basis Events.

The instruments that monitor these variables are designated as Type A, Category I, and non-Type A, Category I, in accordance with Regulatory Guide 1.97 (Ref. 1).

The OPERABILITY of the accident monitoring instrumentation ensures that there is sufficient information available on selected plant parameters to monitor and assess plant status and behavior following an accident. This capability is consistent with the recommendations of Reference 1.

-APPLICABLE The- PAM instrumentation LCO ensures the OPERABILITY of SAFETY ANALYSES Regulatory Guide 1.97, Type A variables so that the control room operating staff can: . . .. ...

• Perform the diagnosis specified in the Emergency Operating Procedures (EOPs). These variables are restricted to preplanned actions for the primary success path of Design Basis Accidents (DBAs), (e.g., loss of coolant accident (LOCA)), and

• Take the specified, preplanned, manually controlled actions for which no automatic control is provided, which are required for safety systems to accomplish their safety function.

The PAM instrumentation LCO also ensures OPERABILITY of Category I, non-Type A, variables so that the control room operating staff can:

Determine whether systems important to safety are performing their intended functions; (continued)

SUSQUEHANNA - UNIT 2 TS / B 3.3-64 Revision 2

PPL Rev. 4 PAM Instrumentation B 3.3.3.1 BASES APPLICABLE Determine the potential for causing a gross breach of the barriers SAFETY ANALYSES to radioactivity release; (continued)

Determine whether a gross breach of a barrier has occurred; and

• Initiate action necessary to protect the public and for an estimate of the magnitude of any impending threat.

The plant specific Regulatory Guide 1.97 Analysis (Ref. 2 and 3) documents the process that identified Type A and Category I, non-Type A, variables.

Accident monitoring instrumentation that satisfies the definition of Type A in Regulatory Guide 1.97 meets Criterion 3 of the NRC Policy Statement. (Ref. 4) Category I, non-Type A, instrumentation is retained in Technical Specifications (TS) because they are intended to assist operators in minimizing the consequences of accidents. Therefore, these Category I variables are important for reducing public risk.

LCO LCO 3.3.3.1 requires two OPERABLE channels for all but one Function to ensure that no single failure prevents the operators from being presented with the information necessary to determine the status of the plant and to bring the plant to, and maintain-it in, a safe condition following that accident. - .. .. ...

Furthermore, provision of two'channels allows a CHANNEL CHECK during the post accident phase to confirm the validity of displayed information. -

The exception to the two channel requirement is primary containment isolation valve (PCIV) position. In this case, the important information is the status of the primary containment penetrations. The LCO requires one position indicator for each active PCIV. This is sufficient to redundantly verify the isolation status of each isolable penetration either via indicated status of the active valve and prior knowledge of passive Valve or via system boundary (continued)

SUSQUEHANNA - UNIT 2 TS / B 3.3-65 Revision 2

PPL Rev. 4 PAM Instrumentation B 3.3.3.1 BASES LCO status. If a normally active PCIV is known to be closed and (continued) deactivated, position indication is not needed to determine status.

Therefore, the position indication for valves in this state is not required to be OPERABLE.

The following list is a discussion of the specified instrument Functions listed in Table 3.3.3.1-1 in the accompanying LCO. Table B3.3.3.1-1 provides a listing of the instruments that are used to meet the operability requirements for the specific functions.

1. Reactor Steam Dome Pressure Reactor steam dome pressure is a Type A, Category 1, variable provided to support monitoring of Reactor Coolant System (RCS) integrity and to verify operation of the Emergency Core Cooling Systems (ECCS). Two independent pressure channels, consisting of three wide range control room indicators and one wide range control room recorder per channel with a range of 0 psig to 1500 psig, monitor pressure. The wide range recorders are the primary method of indication available for use by the operators during an accident, therefore, the PAM Specification deals specifically with this portion of the instrument channel.

2. Reactor Vessel Water Level Reactor vessel water level is a Type A, Category 1, variable provided to support monitoring of core Cooling and to verify operation of the ECCS. A combination of three different level instrument ranges, with two independent channels each, monitor Reactor Vessel Water Level.

The extended range instrumentation measures from -150 inches to 180 inches and outputs to three control room level indicators per channel. The wide range instrumentation measures from -150 inches to 60 inches and outputs to one control room recorder and three control room indicators per channel. The fuel zone range instrumentation measures from -310 inches to -110 inches and outputs to a control room recorder (one channel) and a control room indicator (one channel). These three ranges of instruments combine to provide level indication from the bottom of the Core to above the main steam line. The wide range level recorders, the fuel zone level indicator and level recorder, and one inner ring extended range level indicator per channel are the primary method of indication available for use by the operator during an accident, therefore the PAM (continued)

SUSQUEHANNA - UNIT 2 TS / B 3.3-66 Revision 4

PPL Rev. 4 PAM Instrumentation B 3.3.3.1 BASES LCO 2. Reactor Vessel Water Level (continued)

Specification deals specifically with this portion of the instrument channel.

3. Suppression Chamber Water Level Suppression chamber water level is a Type A, Category 1, variable provided to detect a breach in the reactor coolant pressure boundary (RCPB). This variable is also used to verify and provide long term surveillance of ECCS function. A combination of two different level instrument ranges, with two independent channels each, monitor Suppression chamber water level. The wide range instrumentation measures from the ECCS suction lines to approximately the top of the chamber and outputs to one control room recorder per channel. The wide range recorders are the primary method of indication available for use by the operator during an accident, therefore the PAM Specification deals specifically with this portion of the instrument channel.

4. Primary Containment Pressure Primary Containment pressure is a Type A, Category 1, variable provided to detect a breach of the RCPB and to verify ECCS functions that operate to maintain RCS integrity. A combination of two-different pressure instrument ranges, with two independent channels.each, monitor primary containment pr-essure. The LOCA range measures from -15 psig to 65 psig and outputs to one control room recorder per channel. The accident range measures from 0 psig to 250 psig and outputs to one control room recorder per channel (same recorders as the LOCA range). The recorders (both ranges) are the primary method of indication available for use by the operator during an accident, therefore the PAM Specification deals specifically with this portion of the instrument channel.

5. Primary Containment High Radiation Primary containment area radiation (high range) is provided to monitor the potential of significant radiation releases (continued)

SUSQUEHANNA - UNIT 2 TS / B 3.3-67 Revision 3

PPL Rev. 4 PAM Instrumentation B 3.3.3.1 BASES LCO 5. Primary Containment Hiqh Radiation (continued) and to provide release assessment for use by operators in determining the need to invoke site emergency plans. Two independent channels, which output to one control room recorder per channel with a range of 100 to 1X1 08 R/hr, monitor radiation. The PAM Specification deals specifically with this portion of the instrument channel.

6. Primary Containment Isolation Valve (PCIV) Position PCIV position is provided for verification of containment integrity. In the case of PCIV position, the important information is the isolation status of the containment penetration. The LCO requires a channel of valve position indication in the control room to be OPERABLE for an active PCIV in a containment penetration flow path, i.e., two total channels of PCIV position indication for a penetration flow path with two active valves. For containment penetrations with only one active PCIV having control room indication, Note (b) requires a single channel of valve position indication to be OPERABLE. This is sufficient to redundantly verify the isolation status of each isolable penetration via indicated status of the active valve, as applicable, and prior knowledge of passive valve or system boundary status. If a penetration flow path is isolated, position indication for the PCIV(s) in the associated penetfation flow path is not needed to determine status. -Therefore, the position indication for valves in an isolated penetration flow-path is not required to be OPERABLE. Those valves which require position indication are specified in Table B 3.6.1.3-1. Furthermore, the loss of position indication does not necessarily result in the PCIV being inoperable.

The PCIV position PAM instrumentation consists of position switches unique to PCIVs, associated wiring and control room indicating lamps (not necessarily unique to a PCIV) for active PCIVs (check valves and manual valves are not required to have position indication). Therefore, the PAM Specification deals specifically with these instrument channels.

(continued)

SUSQUEHANNA - UNIT 2 TS / B 3.3-68 Revision 4

PPL Rev. 4 PAM Instrumentation B 3.3.3.1 BASES LCO 7. Neutron Flux (continued)

Wide range neutron flux is a Category I variable provided to verify reactor shutdown. The Neutron Monitoring System Average Power Range Monitors (APRM) provides reliable neutron flux measurement from 0% to 125% of full power. The APRM consists of four channels each with their own chassis powered with redundant power supplies.

The APRM sends signals to the analog isolator module which in turn sends individual APRM signals to the recorders used for post accident monitoring. The PAM function for neutron flux is satisfied by having two channels of APRM provided for post accident monitoring. The PAM Specification deals specifically with this portion of the instrument channel.

The Neutron Monitoring System (NMS) was evaluated against the criteria established in General Electric NEDO-31558A to ensure its acceptability for post-accident monitoring. NEDO-31558A provides alternate criteria for the NMS to meet the post-accident monitoring guidance of Regulatory Guide 1.97. Based on the evaluation, the NMS was found to meet the criteria established in NEDO-31558A.

The APRM sub-function of the NMS is used to provide the Neutron Flux monitoring identified in TS 3.3.3.1 (Ref- 5 and 6).

8. Not Used.

(continued)

SUSQUEHANNA - UNIT 2 TS / B 3.3-69 Revision 5

PPL Rev. 4 PAM Instrumentation B 3.3.3.1 BASES LCO 9. Drywell Atmosphere Temperature (continued)

Drywell atmosphere temperature is a Category I variable provided to verify RCS and containment integrity and to verify the effectiveness of ECCS actions taken to prevent containment breach. Two independent temperature channels, consisting of two control room recorders per channel with a range of 40 to 440 degrees F, monitor temperature.

The PAM Specification deals specifically with the inner ring temperature recorder portion of the instrument channel.

10. Suppression Chamber Water Temperature Suppression Chamber water temperature is a Type A, Category 1, variable provided to detect a condition that could potentially lead to containment breach and to verify the effectiveness of ECCS actions taken to prevent containment breach. The suppression chamber water temperature instrumentation allows operators to detect trends in suppression chamber water temperature in sufficient time to take action to prevent steam quenching vibrations in the suppression pool.

Two channels are required to be OPERABLE. Each channel consists of eight sensors of which a minimum of four sensors (one sensor in each quadrant) must be OPERABLE to consider a channel OPERABLE. The outputs for the temperature sensors are displayed on two independent indicators in the control room and recorded on the monitoring units located in the control room on a back panel.The temperature indicators are the primary method of indication-available for use by the operator during'n accident, therefore the PAM Specification deals specifically with this portion of the instrument channel.

APPLICABILITY The PAM instrumentation LCO is applicable in MODES 1 and 2.

These variables are related to the diagnosis and preplanned actions required to mitigate DBAs. The applicable DBAs are assumed to occur in MODES 1 and 2. In MODES 3, 4, and 5, plant conditions are such that the likelihood of an event that would require PAM instrumentation is extremely low; therefore, PAM instrumentation is not required to be OPERABLE in these MODES.

(continued)

SUSQUEHANNA-UNIT2 TS / B 3.3-70 Revision 3

PPL Rev. 4 PAM Instrumentation B 3.3.3.1 BASES (continued)

ACTIONS A Note has been provided to modify the ACTIONS related to PAM instrumentation channels. Section 1.3, Completion Times, specifies that once a Condition has been entered, subsequent divisions, subsystems, components, or variables expressed in the Condition discovered to be inoperable or not within limits, will not result in separate entry into the Condition. Section 1.3 also specifies that Required Actions of the Condition continue to apply for each additional failure, with Completion Times based on initial entry into the Condition.

However, the Required Actions for inoperable PAM instrumentation channels provide appropriate compensatory measures for separate Functions. As such, a Note has been provided that allows separate Condition entry for each inoperable PAM Function.

A.1 When one or more Functions have one required channel that is inoperable, the required inoperable channel must be restored to OPERABLE status within 30 days. The 30 day Completion Time is based on operating experience and takes into account the remaining OPERABLE channels, the passive nature of the instrument (no critical automatic action is assumed to occur from these instruments), and the low probability of an event requiring PAM instrumentation during this interval.

B.1 ...

If a channel has not been restoied to OPERABLE status in 30 days, this Required Action specifies initiation of action in accordance with Specification 5.6.7, which requires a written report to be submitted to the NRC. This report discusses the results of the root cause evaluation of the inoperability and identifies proposed restorative actions.

(continued)

SUSQUEHANNA - UNIT 2 TS / B 3.3-71 Revision 3

PPL Rev. 4 PAM Instrumentation B 3.3.3.1 BASES ACTIONS B.1 (continued)

This action is appropriate in lieu of a shutdown requirement because alternative actions are identified before the written report is submitted to the NRC, and given the likelihood of plant conditions that would require information provided by this instrumentation.

C.1 When one or more Functions have two required channels that are inoperable (i.e., two channels inoperable in the same Function), one channel in the Function should be restored to OPERABLE status within 7 days. The Completion Time of 7 days is based on the relatively low probability of an event requiring PAM instrument operation and the availability of alternate means to obtain the required information. Continuous operation with two required channels inoperable in a Function is not acceptable because the alternate indications may not fully meet all performance qualification requirements applied to the PAM instrumentation. Therefore, requiring restoration of one inoperable channel of the Function limits the risk that-the PAM Function will be in a degraded condition should an accident occur.

D. 1 This Required Action directs entry into the appropriate Condition referenced in Table 3.3.3.1-1. The applicable Condition referenced in the Table is Function dependent. Each time an inoperable channel has not met any Required Action of Condition C, as applicable, and the associated Completion Time has expired, Condition D is entered for that channel and provides for transfer to the appropriate subsequent Condition.

E. 1 For the majority of Functions in Table 3.3.3.1-1, if any Required Action and associated Completion Time of Condition C are not met, the plant must be brought to a MODE in which the LCO not apply. To achieve this status, the plant must be brought to at least MODE 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions (continued)

SUSQUEHANNA- UNIT 2 TS / B 3.3-72 Revision 2

PPL Rev. 4 PAM Instrumentation B 3.3.3.1 BASES ACTIONS E.1 (continued) from full power conditions in an orderly manner and without challenging plant systems.

F.1 Since alternate means of monitoring primary containment area radiation have been developed and tested, the Required Action is not to shut down the plant, but rather to follow the directions of Specification 5.6.7. These alternate means will be temporarily installed if the normal PAM channel cannot be restored to OPERABLE status within the allotted time. The report provided to the NRC should discuss the alternate means used, describe the degree to which the alternate means are equivalent to the installed PAM channels, justify the areas in which they are not equivalent, and provide a schedule for restoring the normal PAM channels.

SURVEILLANCE The following SRs apply to each PAM instrumentation Function in REQUIREMENTS Table 3:3.3.1-1.

SR 3.3.3.1.1 -

Performance of the CHANNEL-CHECK once every 31 days ensures that a gross failure of instrumentation has not occurred. A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel against a similar parameter on other channels. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value. Significant deviations between instrument channels could be an indication of excessive instrument drift in one of the channels or something even more serious. A CHANNEL CHECK will detect gross channel failure; thus, it is key to verifying the instrumentation continues to operate properly between each CHANNEL CALIBRATION.

Agreement criteria which are determined by the plant staff based on an investigation of a combination of the channel instrument uncertainties, may be used to support this (continued)

SUSQUEHANNA - UNIT 2 TS / B 3.3-73 Revision 2

PPL Rev. 4 PAM Instrumentation B 3.3.3.1 BASES SURVEILLANCE SR 3.3.3.1.1 (continued)

REQUIREMENTS parameter comparison and include indication and readability. If a channel is outside the criteria, it may be an indication that the instrument has drifted outside its limit, and does not necessarily indicate the channel is Inoperable.

The Frequency of 31 days is based upon plant operating experience, with regard to channel OPERABILITY and drift, which demonstrates that failure of more than one channel of a given Function in any 31 day interval is rare. The CHANNEL CHECK supplements less formal checks of channels during normal operational use of those displays associated with the required channels of this LCO.

SR 3.3.3.1.2 and SR 3.3.3.1.3 A CHANNEL CALIBRATION is performed every 24 months except for the PCIV Position Function. The PCIV Position Function is adequately demonstrated by the Remote Position Indication performed in accordance with 5.5.6, "Inservice Testing Program". CHANNEL CALIBRATION verifies that the channel responds to measured parameter with the necessary range and accuracy, and does not include alarms.

The CHANNEL CALIBRATION for the Containment High Radiation instruments shall consist of an-electronic calibration of the-channel, not including the detector, for range decades above 10 R/hr and a one point calibration check of the detector below 10 R/hr with an installed or portable gamma source.

The Frequency is based on operating experience and for the 24 month Frequency consistency with the industry refueling cycles.

(continued)

SUSQUEHANNA - UNIT 2 TS / B 3.3-74 Revision 3

PPL Rev. 4 PAM Instrumentation B 3.3.3.1 BASES (continued)

REFERENCES 1. Regulatory Guide 1.97 Rev. 2, "Instrumentation for Light Water Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident," February 6, 1985

2. Nuclear Regulatory Commission Letter A. Schwencer to N. Curtis, Emergency Response Capability, Conformance to R.G. 1.97, Rev. 2, dated February 6, 1985.

3. PP&L Letter (PLA-2222), N. Curtis to A. Schwencer, dated May 31, 1984.

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

5. NEDO-31558A, BWROG Topical Report, Position on NRC Reg.

Guide 1.97, Revision 3 Requirements for Post Accident Neutron Monitoring System (NMS).

6. Nuclear Regulatory Commission Letter from C. Poslusny to R.G.

Byram dated July 3, 1996.

SUSQUEHANNA - UNIT 2 TS / B 3.3-75 Revision 2

PPL Rev. 4 PAM Instrumentation B 3.3.3.1 TABLE B 3.3.3.1-1 Post Accident Instruments (Page 1 of 3)

Instrument/Variable Element Transmitter Recorder Indicator

1. Reactor Steam N/A PT-24201A UR-24201A PI-24202A Dome Pressure (Red)* PI-24202A1 PI-24204A N/A PT-24201 B UR-24201 B PI-24202B (left side)

(Red)* PI-24202B1 (left side)

PI-24204B (left side)

2. Reactor Vessel N/A LT-24201A UR-24201A LI-24201A (left side)

Water Level (Wide Range) (Blue)* LI-24201A1 (left side)

LI-24203A (left side)

N/A LT-24201 B UR-24201 A LI-24201 B (left side)

(Wide Range) (Blue)* LI-24201 B1 (left side)

LI-24203B (left side)

N/A LT-24203A N/A LI-24201A (right side)- 1 )

(Extended Range) LI-24201A1 (right side)

LI-24203A (right side) 1 N/9 LT-24203B N/A LI-24201 B (right side)*( )

(Extended Range) LI-24201 BI (right side)

__ __ LI-24203B (right side)

N/A LT-24202A U P,-24201 A N/A (Fuel Zone Range) (Brown)*

N/A LT-24202B UR-24201 B (Fuel Zone Range) (Brown)*

3. Suppression N/A LT-25776A UR-25776A N/A Chamber Water Level (Wide Range) (Red)*

N/A LT-25776B UR-25776B N/A (Wide Range) (Red)*

N/A LT-25775A UR-25776A LI-25775A (Narrow Range) (Blue) I N/A LT-25775B UR-25776B LI-25775B (Narrow Range) (Blue) I _ _ _ _ I SUSQUEHANNA - UNIT 2 TS / B 3.3-75a Revision 5

PPL Rev. 4 PAM Instrumentation B 3.3.3.1 TABLE B 3.3.3.1-1 Post Accident Instruments (Page 2 of 3)

Instrument/Variable { Element Transmitter Recorder Indicator

4. Primary N/A PT-25709A UR-25701A (Dark Blue)* N/A Containment Pressure (0 to 250 psig)

N/A PT-25709B UR-25701 B (Dark Blue)* N/A (0 to 250 psig)

N/A PT-25710A UR-25701A (Red)* N/A

(-15 to 65 psig)

N/A PT-25710B UR-25701 B (Red)* N/A

(-15 to 65 psig)

5. Primary Containment RE-25720A RITS-25720A UR-25776A N/A High Radiation (Green)*

RE-25720B RITS-25720B UR-25776B N/A (Green)*

6. PCIV Position See Technical Specification Bases Table B 3.6.1.3-1 for PCIV that require position indication to be OPERABLE

7. Neutron Flux N/h APRM-1 NR-C51-2R603A N/A

_(red pen)*.

N/A APRM-2 NR-C51-2R603B N/A (red pen)*

N/A APRM-3 NR-C51-2R603C N/A (red pen)*

N/A APRM-4 NR-C51-2R603D N/A (red pen)*

8. Containment AE-25745A AT-25745A UR-25701 A (Blue Violet)* N/A Oxygen and (Hydrogen) UR-25701A (Violet)*

Hydrogen Analyzer AE-25745B AT-25745B UR-25701 B (Blue Violet)*

N/A (Hydrogen) UR-25701 B (Violet)*

AE-25746A AT-25746A UR-25701 A (Orange)* N/A (Oxygen)

AE-25746B AT-25746B UR-25701 B (Orange)* N/A (Oxygen) I _I__ _ _

SUSQUEHANNA - UNIT 2 TS / B 3.3-75b Revision 5

PPL Rev. 4 PAM Instrumentation B 3.3.3.1 TABLE B 3.3.3.1-1 Post Accident Instruments (Page 3 of 3)

Instrument/Variable Element Transmitter Recorder Indicator

9. Drywell Atmosphere TE-25790A TT-25790A UR-25701A (Brown)* N/A Temperature TR-25790A (point # 1)

TE-257908 TT-25790B UR-25701 B (Brown)* N/A TR-25790B (point # 1)

10. Suppression Chamber TE-25753 TX-25751 N/A TIAH-25751*

Water Temperature TE-25755 TI-25751 TE-25757 TE-25759 TE-25763 TE-25765 TE-25767 TE-25769 TE-25752 TX-25752 N/A TIAH-25752*

TE-25754 TI-25752 TE-25758 TE-25760 TE-25762 -

TE-25766 TE-25768 TE-25770 Indicates that the instrument (and associated components in the instrument channel) is considered as instrument channel surveillance acceptance criteria.

(1) In the case of the inner ring indicators for extended range level, it is recommended that LI-24201 A and LI-24201 B be used as acceptance criteria, however LI-24201 Al, LI-24201 B1, LI-24203A. or LI-24203B may be used in their place provided that surveillance requirements are satisfied. Only one set of these instruments needs to be OPERABLE.

SUSQUEHANNA-UNIT2 TS I B 3.3-75c Revision 4

PPL Rev. 3 Recirculation Loops Operating B 3.4.1 B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.1 Recirculation Loops Operating BASES BACKGROUND The Reactor Coolant Recirculation System is designed to provide a forced coolant flow through the core to remove heat from the fuel. The forced coolant flow removes more heat from the fuel than would be possible with just natural circulation. The forced flow, therefore, allows operation at significantly higher power than would otherwise be possible. The recirculation system also controls reactivity over a wide span of reactor power by varying the recirculation flow rate to control the void content of the moderator. The Reactor Coolant Recirculation System consists of two recirculation pump loops external to the reactor vessel. These loops provide the piping path for the driving flow of water to the reactor vessel jet pumps. Each external loop contains one variable speed motor driven recirculation pump, a motor generator (MG) set to control pump speed and associated piping, jet pumps, valves, and instrumentation. The recirculation pump, piping, and valves are part of the reactor coolant pressure boundary and are located inside the drywell structure. The jet pumps are reactor vessel internals.

The recirculated coolant consists of saturated water from the steam separators and dryers that has been subcooled by incoming feedwater.

This water passes down the annulus between-the reactor vessel wall and the core shroud. A portion of the coolant flows frorm the Vessel, tfir6ugh the two external recirculation loops-, and becomes the driving flow for the jet pumps. Each of the two external recirculation loops discharges high pressure flow into an external manifold, from which individual recirculation inlet lines are routed to the jet pump risers within the reactor vessel. The remaining portion of the coolant mixture in the annulus becomes the suction flow for the jet pumps. This flow enters the jet pump at suction inlets and is accelerated by the driving flow. The drive flow and suction flow are mixed in the jet pump throat section. The total flow then passes through the jet pump diffuser section into the area below the core (lower plenum), gaining sufficient head in the process to drive the required flow upward through the core. The subcooled water enters the bottom of the fuel channels and contacts the fuel (continued)

SUSQUEHANNA -UNIT 2 TS / B 3.4-1 Revision 2

PPL Rev. 3 Recirculation Loops Operating B 3.4.1 BASES BACKGROUND cladding, where heat is transferred to the coolant. As it rises, the coolant (continued) begins to boil, creating steam voids within the fuel channel that continue until the coolant exits the core. Because of reduced moderation, the steam voiding introduces negative reactivity that must be compensated for to maintain or to increase reactor power. The recirculation flow control allows operators to increase recirculation flow and sweep some of the voids from the fuel channel, overcoming the negative reactivity void effect.

Thus, the reason for having variable recirculation flow is to compensate for reactivity effects of boiling over a wide range of power generation without having to move control rods and disturb desirable flux patterns.

Each recirculation loop is manually started from the control room. The MG set provides regulation of individual recirculation loop drive flows. The flow in each loop is manually controlled.

APPLICABLE The operation of the Reactor Coolant Recirculation System is an initial SAFETY condition assumed in the design basis loss of coolant accident (LOCA)

ANALYSES (Ref. 1). During a LOCA caused by a recirculation loop pipe break, the intact loop is assumed to provide coolant flow during the first few seconds of the accident. The initial core flow decrease is rapid because the recirculation pump in the broken loop ceases to pump reactor coolant to the vessel almost immediately. The pump in the intact loop coasts down relatively slowly. This pump coastdown governs th6 core flow resp-onse for the next several seconds until.the jet pump suction is uncovered (Ref. 1). The analyses assume that both loops are operating at the same flow prior to the accident. However, the LOCA analysis was reviewed for the case with a flow mismatch between the two loops, with the pipe break assumed to be in the loop with the higher flow. While the flow coastdown and core response are potentially more severe in this assumed case (since the intact loop starts at a lower flow rate and the core response is the same as if both loops were operating at a lower flow rate), a small mismatch has been determined to be acceptable based on engineering judgement. The recirculation system is also assumed to have sufficient flow coastdown characteristics to maintain fuel thermal margins during abnormal operational transients (Ref. 2), which are analyzed in Chapter 15 of the FSAR.

(continued)

SUSQUEHANNA -UNIT 2 TS / B 3.4-2 Revision 2

PPL Rev. 3 Recirculation Loops Operating B 3.4.1 BASES APPLICABLE Plant specific LOCA analyses have been performed assuming only one SAFETY operating recirculation loop. These analyses have demonstrated that, in ANALYSES the event of a LOCA caused by a pipe break in the operating recirculation (continued) loop, the Emergency Core Cooling System response will provide adequate core cooling, provided that the APLHGR limit for SPC ATRIUMTM-10 fuel is modified.

The transient analyses of Chapter 15 of the FSAR have also been performed for single recirculation loop operation and demonstrate sufficient flow coastdown characteristics to maintain fuel thermal margins during the abnormal operational transients analyzed provided the MCPR requirements are modified. During single recirculation loop operation, modification to the Reactor Protection System (RPS) average power range monitor (APRM) instrument setpoints is also required to account for the different relationships between recirculation drive flow and reactor core flow. The APLHGR, LHGR, and MCPR limits for single loop operation are specified in the COLR. The APRM Simulated Thermal Power-High setpoint is in LCO 3.3.1.1, "Reactor Protection System (RPS)

Instrumentation." In addition, a restriction on recirculation pump speed is incorporated to address reactor vessel internals vibration concerns and assumptions in the event analysis.

Recirculation loops operating satisfies Criterion 2 of-the NRC-Potrcy Statement (Ref. 5). .

LCO Two recirculation loops are required to be in operation with their flows matched within the limits specified in SR 3.4.1.1 to ensure that during a LOCA caused by a break of the piping of one recirculation loop the assumptions of the LOCA analysis are satisfied. With the limits specified in SR 3.4.1.1 not met, the recirculation loop with the lower flow must be considered not in operation. With only one recirculation loop in operation, modifications to the required APLGHR limits (LCO 3.2.1, "AVERAGE PLANAR LINEAR HEAT GENERATION RATE"), LHGR limits (LCO 3.2.3, "LINEAR HEAT GENERATION RATE (LHGR)"), MCPR limits (LCO 3.2.2, "MINIMUM CRITICAL POWER RATIO (MCPR)"), and APRM Simulated Thermal Power-High setpoint (LCO 3.3.1.1) may be applied to allow continued operation consistent with the safety analysis assumptions.

Furthermore, restrictions are placed on recirculation pump speed to ensure the initial assumption of the event analysis are maintained.

(continued)

SUSQUEHANNA - UNIT 2 TS / B 3.4-3 Revision 4

PPL Rev. 3 Recirculation Loops Operating B 3.4.1 BASES LCO The LCO is modified by a Note that allows up to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> to establish the (continued) required limits and setpoints after a change from two recirculation loops operation to single recirculation loop operation. If the limits and setpoints are not in compliance with the applicable requirements at the end of the this period, the ACTIONS required by the applicable specifications must be implemented. This time is provided to stabilize operation with one recirculation loop by: limiting flow in the operating loop, limiting total THERMAL POWER, monitor APRM and local power range monitor (LPRM) neutron flux noise levels; and, fully implementing and confirming the required limit and setpoint modifications.

APPLICABILITY In MODES 1 and 2, requirements for operation of the Reactor Coolant Recirculation System are necessary since there is considerable energy in the reactor core and the limiting design basis transients and accidents are assumed to occur.

In MODES 3, 4, and 5, the consequences of an accident are reduced and the coastdown characteristics of the recirculation loops are not important.

ACTIONS A.1 When operating with no recirculation loops operating in MODE 1, the potential for thermal-hydraulic oscillations is greatly-increased. Although this transient is protected for expected modes of oscillation by the OPRM system, when OPERABLE per LCO 3.3.1.1, function 2.f (Reference 3, 4),

the prudent response to the natural circulation condition is to preclude potential thermal-hydraulic oscillations by immediately placing the mode switch in the shutdown position.

B.1 Recirculation loop flow must match within required limits when both recirculation loops are in operation. If flow mismatch is not within required limits, matched flow must be restored within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. If matched flows are not restored, the recirculation loop with lower flow must be declared "not in operation." Should a LOCA occur with recirculation loop flow not matched, the core flow coastdown and resultant core response may not be bounded by the LOCA analyses. Therefore, only a limited time is allowed prior to imposing restrictions associated with single loop operation. Operation with only one recirculation loop satisfies the requirements of the LCO and the initial conditions of the accident sequence.

(continued)

SUSQUEHANNA - UNIT 2 TS / B 3.4-4 Revision 4

PPL Rev. 3 Recirculation Loops Operating B 3.4.1 BASES ACTIONS The 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> Completion Time is based on the low probability of an accident (continued) occurring during this time period, providing a reasonable time to complete the Required Action, and considering that frequent core monitoring by operators allows abrupt changes in core flow conditions to be quickly detected.

These Required Actions do not require tripping the recirculation pump in the lowest flow loop when the mismatch between total jet pump flows of the two loops is greater than the required limits. However, in cases where large flow mismatches occur, low flow or reverse flow can occur in the low flow loop jet pumps, causing vibration of the jet pumps. If zero or reverse flow is detected, the condition should be alleviated by changing recirculation pump speed to re-establish forward flow or by tripping the pump.

C.1 With no recirculation loops in operation while in MODE 2 or if after going to single loop operations the required limits and setpoints cannot be established, the plant must be brought to MODE 3, where the LCO does not apply within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. In this condition, the recirculation loops are not required to be operating because of the reduced severity of DBAs and minimal dependence on the recirculation loop.coastdown characteristics.

The allowed Completion Time of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is reasonable-to reach-MODE 3 from full power conditions in an orderly manner withbut challenging plant systems.

SURVEILLANCE SR 3.4.1.1 REQUIREMENTS This SR ensures the recirculation loops are within the allowable limits for mismatch. At low core flow (i.e., < 75 million Ibm/hr), the MCPR requirements provide larger margins to the fuel cladding integrity Safety Limit such that the potential adverse effect of early boiling transition during a LOCA is reduced. A larger flow mismatch can therefore be allowed when core flow is < 75 million Ibm/hr. The recirculation loop jet pump flow, as used in this Surveillance, is the summation of the flows from all of the jet pumps associated with a single recirculation loop.

(continued)

SUSQUEHANNA - UNIT 2 TS / B 3.4-5 Revision 3

PPL Rev. 3 Recirculation Loops Operating B 3.4.1 BASES SURVEILLANCE SR 3.4.1.1 (continued)

REQUIREMENTS The mismatch is measured in terms of core flow. If the flow mismatch exceeds the specified limits, the loop with the lower flow is considered inoperable. The SR is not required when both loops are not in operation since the mismatch limits are meaningless during single loop or natural circulation operation. The Surveillance must be performed within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after both loops are in operation. The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Frequency is consistent with the Surveillance Frequency for jet pump OPERABILITY verification and has been shown by operating experience to be adequate to detect off normal jet pump loop flows in a timely manner.

SR 3.4.1.2 As noted, this SR is only applicable when in single loop operation. This SR ensures the recirculation pump limit is maintained. The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Frequency is based on operating experience and the operators inherent knowledge of the current reactor status.

REFERENCES 1. FSAR, Section 6.3.3.7.

2. FSAR, Section 5.4.1.4.

3. GE NEDO-31960-A "BWROG Long Term Stability Solutions Licensing Methodology," November, 1995.

4. GE NEDO-31960-A "BWROG Long Term Stability Solutions Licensing Methodology," Supplement 1, November 1995.

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

SUSQUEHANNA - UNIT 2 TS / B 3.4-6 Revision 3

PPL Rev. 3 Recirculation Loops Operating B 3.4.1 THIS PAGE INTENTIONALLY LEFT BLANK SUSQUEHANNA - UNIT 2 TS / B 3.4-7 Revision 3

PPL Rev. 3 Recirculation Loops Operating B 3.4.1 THIS PAGE INTENTIONALLY LEFT BLANK SUSQUEHANNA - UNIT 2 TS / B 3.4-8 Revision 3

PPL Rev. 3 Recirculation Loops Operating B 3.4.1 THIS PAGE INTENTIONALLY LEFT BLANK SUSQUEHANNA-UNIT2 TS / B 3.4-9 Revision 3

PPL Rev. 2 SDM Test-Refueling B 3.10.8 B 3.10 SPECIAL OPERATIONS B 3.10.8 SHUTDOWN MARGIN (SDM) Test-Refueling BASES BACKGROUND The purpose of this MODE 5 Special Operations LCO is to permit SDM testing to be performed for those plant configurations in which the reactor pressure vessel (RPV) head is either not in place or the head bolts are not fully tensioned.

LCO 3.1.1, "SHUTDOWN MARGIN (SDM)," requires that adequate SDM be demonstrated following fuel movements or control rod replacement within the RPV. The demonstration must be performed prior to or within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after criticality is reached. This SDM test may be performed prior to or during the first startup following the refueling.

Performing the SDM test prior to startup requires the test to be performed while in MODE 5, with the vessel head bolts less than fully tensioned (and possibly with the vessel head removed). While in MODE 5, the reactor mode switch is required to be in the shutdown or refuel position, where the applicable control rod blocks ensure that the reactor will not become critical. The SDM test requires the reactor mode switch to be in the startup/hot standby position, since more than one control rod will be withdrawn for the purpose of demonstrating adequate SDM. This Special Operations LCO provides the appropriate additional controls to allow withdrawing more than one control rod from a core cell containing one or more fuel assemblies when the reactor vessel head bolts are less than fully tensioned.

APPLICABLE Prevention and mitigation of unacceptable reactivity excursions during SAFETY ANALYSES control rod withdrawal, with the reactor mode switch in the startup/hot standby position while in MODE 5, is provided by the intermediate range monitor (IRM) neutron flux scram (LCO 3.3.1.1, "Reactor Protection System (RPS) Instrumentation"), and control rod block instrumentation (LCO 3.3.2.1, "Control Rod Block Instrumentation").

The limiting reactivity excursion during startup conditions while in MODE 5 is the control rod drop accident (CRDA).

(continued)

SUSQUEHANNA - UNIT 2 B 3.10-34 Revision 0

PPL Rev. 2 SDM Test-Refueling B 3.10.8 BASES APPLICABLE CRDA analyses assume that the reactor operator follows prescribed SAFETY ANALYSES withdrawal sequences. For SDM tests performed within these defined (continued) sequences, the analyses of Reference 1 is applicable. However, for some sequences developed for the SDM testing, the control rod patterns assumed in the safety analyses of Reference 1 may not be met. Therefore, special CRDA analyses, performed in accordance with an NRC approved methodology, are required to demonstrate the SDM test sequence will not result in unacceptable consequences should a CRDA occur during the testing. For the purpose of this test, the protection provided by the normally required MODE 5 applicable LCOs, in addition to the requirements of this LCO, will maintain normal test operations as well as postulated accidents within the bounds of the appropriate safety analyses (Ref. 1). In addition to the added requirements for the RWM, APRM, and control rod coupling, the notch out mode is specified for control rod withdrawals that are not in conformance with the BPWS. Requiring the notch out mode limits withdrawal steps to a single notch, which limits inserted reactivity, and allows adequate monitoring of changes in neutron flux, which may occur during the test.

As described in LCO 3.0.7, compliance with Special Operations LCOs is optional, and therefore, no criteria of the NRC Policy Statement apply. Special Operations LCOs provide flexibility to perform certain operations by appropriately modifying requirements of other LCOs. A discussion of the-criteria satisfied for the other LCOs is provided in their respective Bases.

LCO As described in LCO 3.0.7, compliance with this Special Operations LCO is optional. SDM tests may be performed while in MODE 2, in accordance with Table 1.1-1, without meeting this Special Operations LCO or its ACTIONS. For SDM tests performed while in MODE 5, additional requirements must be met to ensure that adequate protection against potential reactivity excursions is available. To provide additional scram protection, beyond the normally required IRMs, the APRMs are also required to be OPERABLE (LCO 3.3.1.1, Functions 2.a, 2.d and 2.e) as though the reactor were in MODE 2. Because multiple control rods will be withdrawn and the reactor will potentially become critical, RPS MODE 2 requirements for Functions 2.a, 2.d and 2.e of Table 3.3.1.1-1 (continued)

SUSQUEHANNA - UNIT 2 TS / B 3.10-35 Revision 1

PPL Rev. 2 SDM Test-Refueling B 3.10.8 BASES LCO must be enforced and the approved control rod withdrawal sequence (continued) must be enforced by the RWM (LCO 3.3.2.1, Function 2, MODE 2), or must be verified by a second licensed operator or other qualified member of the technical staff. The SDM may be demonstrated during an in sequence control rod withdrawal, in which the highest worth control rod is analytically determined, or during local criticals, where the highest worth control rod is determined by analysis or testing.

Local critical tests require the withdrawal of control rods in a sequence that is not in conformance with the BPWS. This testing would therefore require bypassing or reprogramming of the rod worth minimizer to allow the withdrawal of rods not in conformance with BPWS, and therefore additional requirements must be met (see LCO 3.10.7, "Control Rod Testing - Operating").

Control rod withdrawals that do not conform to the banked position withdrawal sequence specified in LCO 3.1.6, "Rod Pattern Control,"

(i.e., out of sequence control rod withdrawals) must be made in the individual notched withdrawal mode to minimize the potential reactivity insertion associated with each movement.

Coupling integrity of withdrawn control rods is required to minimize the probability of a CRDA and ensure proper functioning of the withdrawn control rods, if they are required to scram. Because the reactor vessel head may be removed during these tests, no other CORE ALTERATIONS may be in progress. Furthermore, -,inice -the&-hntrol rod scram function with the RCS at-atmospheric pressure relies solely on the CRD accumulator, it is essential that the CRD charging-water header remain pressurized. This Special Operations LCO then allows changing the Table 1.1-1 reactor mode switch position requirements to include the startup/hot standby position, such that the SDM tests may be performed while in MODE 5.

APPLICABILITY These SDM test Special Operations requirements are only applicable if the SDM tests performed in accordance with LCO 3.1.1, "SDM" are to be performed while in MODE 5 with the reactor vessel head removed or the head bolts not fully tensioned.

Additional requirements during these tests to (continued)

SUSQUEHANNA - UNIT 2 B 3.10-36 Revision 0

PPL Rev. 2 SDM Test-Refueling B 3.10.8 BASES APPLICABILITY enforce control rod withdrawal sequences and restrict other CORE (continued) ALTERATIONS provide protection against potential reactivity excursions. Operations in all other MODES are unaffected by this LCO.

ACTIONS A.1 With one or more control rods discovered uncoupled during this Special Operation, a controlled insertion of each uncoupled control rod is required; either to attempt recoupling, or to preclude a control rod drop.

This controlled insertion is preferred since, if the control rod fails to follow the drive as it is withdrawn (i.e., is "stuck" in an inserted position),

placing the reactor mode switch in the shutdown position per Required Action B.1 could cause substantial secondary damage. If recoupling is not accomplished, operation may continue, provided the control rods are fully inserted within 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> and disarmed (electrically or hydraulically) within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. Inserting a control rod ensures the shutdown and scram capabilities are not adversely affected. The control rod is disarmed to prevent inadvertent withdrawal during subsequent operations. The control rods can be hydraulically disarmed by closing the drive water and exhaust water isolation valves.

Electrically the control rods can be disarmed by disconnecting power frorm all four directional control valve solenoids. Required Action A.1 is modified by a Note that allows the RWM to be bypassed if required to allow insertion of -the inoperable control rods and continued operation.

LCO 3.3.2.1, "Control Rod Block Instrumentation," Acions-pro-vvcde additional requirements when the RWM is bypassed to ensure compliance with the CRDA analysis.

The allowed Completion Times are reasonable, considering the small number of allowed inoperable control rods, and provide time to insert and disarm the control rods in an orderly manner and without challenging plant systems.

Condition A is modified by a Note allowing separate Condition entry for each uncoupled control rod. This is acceptable since the Required Actions for this Condition provide appropriate compensatory actions for each uncoupled control rod. Complying with the Required Actions may allow for continued operation. Subsequent uncoupled control rods (continued)

SUSQUEHANNA - UNIT 2 B 3.10-37 Revision 0

PPL Rev. 2 SDM Test-Refueling B 3.10.8 BASES ACTIONS A.1 (continued) are governed by subsequent entry into the Condition and application of the Required Actions.

B. 1 With one or more of the requirements of this LCO not met for reasons other than an uncoupled control rod, the testing should be immediately stopped by placing the reactor mode switch in the shutdown or refuel position. This results in a condition that is consistent with the requirements for MODE 5 where the provisions of this Special Operations LCO are no longer required.

SURVEILLANCE SR 3.10.8.1 REQUIREMENTS Performance of the applicable SRs for LCO 3.3.1.1, Functions 2.a and 2.d will ensure that the reactor is operated within the bounds of the safety analysis.

SR 3.10.8.1, SR 3.10.8.2, and SR 3.10.8.3 LCO 3.3.1.1, Functions 2.a, 2.d and 2.e, made applicable in this Special Operations LCO,-are required to have applicable Surveillances met to establish that this Special Operations LCO is being met. Iowever, the control rod withdrawal sequences during the SDM tests maybe enforced by the RWM (LCO 3.3.2.1, Function 2, MODE 2 requirements) or by a second licensed operator or other qualified member of the technical staff. As noted, either the applicable SRs for the RWM (LCO 3.3.2.1) must be satisfied according to the applicable Frequencies (SR 3.10.8.2), or the proper movement of control rods must be verified (SR 3.10.8.3). This latter verification (i.e., SR 3.10.8.3) must be performed during control rod movement to prevent deviations from the specified sequence. These surveillances provide adequate assurance that the specified test sequence is being followed.

(continued)

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PPL Rev. 2 SDM Test-Refueling B 3.10.8 BASES SURVEILLANCE SR 3.10.8.4 REQUIREMENTS (continued) Periodic verification of the administrative controls established by this LCO will ensure that the reactor is operated within the bounds of the safety analysis. The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Frequency is intended to provide appropriate assurance that each operating shift is aware of and verfies compliance with these Special Operations LCO requirements.

SR 3.10.8.5 Coupling verification is performed to ensure the control rod is connected to the control rod drive mechanism and will perform its intended function when necessary. The verification is required to be performed any time a control rod is withdrawn to the "full out" notch position, or prior to declaring the control rod OPERABLE after work on the control rod or CRD System that could affect coupling. This Frequency is acceptable, considering the low probability that a control rod will become uncoupled when it is not being moved as well as operating experience related to uncoupling events.

SR 3.10.8.6 CRD charging water header pressure verification is performed to ensure the motive force is available to scram the control rods in the event of a scram-signal. A minimum accumulator pressure is specified, below which the capability of the accumulator to peifdrm its in-ti~ded function becomes degraded and the accumulator is considered inoperable. The minimum accumulator pressure of 940 psig is well below the expected pressure of 1100 psig. The 7 day Frequency has been shown to be acceptable through operating experience and takes into account indications available in the control room.

REFERENCE 1. XN-NF-80-19(P)(A) Volume 1 and Supplements 1 and 2, "Exxon Nuclear Methodology for Boiling Water Reactors," Exxon Nuclear Company, March 1983.

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