ML043370132
| ML043370132 | |
| Person / Time | |
|---|---|
| Site: | Susquehanna  |
| Issue date: | 11/23/2004 |
| From: | Susquehanna |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| Download: ML043370132 (35) | |
Text
Nov. 23, 2004 Page 1 of 1 MANUAL HARD COPY DISTRIBUTION DOCUMENT TRANSMITTAL 2004-48982 USER INFORMATI Name RLAC OSE M EMPL#:028401 CA#:0363 Address:
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TSB1 -
TECHNICAL SPECIFICATION BASES UNIT 1 MANUAL REMOVE MANUAL TABLE OF CONTENTS DATE: 11/09/2004
)D MANUAL TABLE OF CONTENTS DATE: 11/22/2004 "1/2ATEGORY: DOCUMENTS TYPE: TSB1 ID:
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REV:1 ADD:
REV: 2 CATEGORY:
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IL SSES MANUAL
- a> Manual Name:
TSB1 manual
Title:
TECHNICAL SPECIFICATION BASES UNIT 1 MANUAL Table Of Contents Issue Date:
11/22/2004 Procedure Name Rev TEXT LOES 53
Title:
LIST OF EFFECTIVE SECTIONS Issue Date 11/22/2004 Change ID Change Number 5
11/22/2004 TEXT TOC
Title:
TABLE OF CONTENTS TEXT 2.1.1
Title:
SAFETY LIMITS (SLS)
TEXT 2.1.2
Title:
SAFETY LIMITS (SLS) 1 REACTOR 0
REACTOR 04/27/2004 CORE SLS 11/15/2002 COOLANT SYSTEM SL nTEXVm la n
A
Title:
LIMITING CONDITION FOR OPERATION (LCOL-APPLICABILITY TEXT 3.1.1 0/< ;l11/,5(2o02
Title:
REACTIVITY CONTROL SYSTEMSZSHUTDOWN MARGIN (SDM)
TEXT 3.1.2 11/15/2002
Title:
REACTIVITY CONTROL SYSTEMS REACTIVITY ANOMALIES TEXT 3.1.3 Or 0
11/15/2002
Title:
REACTIVITY CONTROL SYSTEMS CONTROL ROD OPERABILITY TEXT 3.1.4 0
11/15/2002
Title:
REACTIVITY CONTROL SYSTEMS CONTROL ROD SCRAM TIMES TEXT 3.1.5 0
11/15/2002
Title:
REACTIVITY CONTROL SYSTEMS CONTROL ROD SCRAM ACCUMULATORS TEXT 3.1.6 0
11/15/2002
Title:
REACTIVITY CONTROL SYSTEMS ROD PATTERN CONTROL Pagel of 8
Report Date: 11/22/04 Page 1 of 8 Report Date: 11/22/04
SSES MANUAL Manual Name: TSB1 Manual
Title:
TECHNICAL SPECIFICATION BASES UNIT 1 MANUAL TEXT 3.1.7 0
11/15/2002
Title:
REACTIVITY CONTROL SYSTEMS STANDBY LIQUID CONTROL {SLC) SYSTEM TEXT 3.1.8 0
11/15/2002
Title:
REACTIVITY CONTROL SYSTEMS SCRAM DISCHARGE VOLUME (SDV)
VENT AND DRAIN VALVES TEXT 3.2.1 0
11/15/2002
Title:
POWER DISTRIBUTION LIMITS AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)
TEXT 3.2.2 0
11/15/2002
Title:
POWER DISTRIBUTION LIMITS MINIMUM CRITICAL POWER RATIO (MCPR)
TEXT 3.2.3 0
11/15/2002
Title:
POWER DISTRIBUTION LIMITS LINEAR HEAT GENERATION RATE (LHGR)
TEXT 3.2.4 0
11/15/2002
Title:
POWER DISTRIBUTION LIMITS AVERAGE POWER RANGE MONITOR (APRM)
GAIN AND SETPOINTS TEXT 3.3.1.1 0
11/15/2002
Title:
INSTRUMENTATION REACTOR PROTECTION SYSTEM (RPS)
INSTRUMENTATION TEXT 3.3.1.2 0
11/15/2002
Title:
INSTRUMENTATION SOURCE RANGE MONITOR (SRM)
INSTRUMENTATION TEXT 3.3.2.1
Title:
INSTRUMENTATION TEXT 3.3.2.2
Title:
INSTRUMENTATION TEXT 3.3.3.1
Title:
INSTRUMENTATION TEXT 3.3.3.2
Title:
INSTRUMENTATION 0
11/15/2002 CONTROL ROD BLOCK INSTRUMENTATION 0
11/15/2002 FEEDWATER MAIN TURBINE HIGH WATER LEVEL TRIP INSTRUMENTATION 0
11/15/2002 POST ACCIDENT MONITORING (PAM)
INSTRUMENTATION LDCN 3702 0
11/15/2002 REMOTE SHUTDOWN SYSTEM Page2 of 8
Report Date: 11/22/04 Page 2 of 8 Report Date: 11/22/04
SSES MANUAL Manual Name: TSB1 Manual
Title:
TECHNICAL SPECIFICATION BASES UNIT 1 MANUAL TEXT 3.3.4.1 0
11/15/2002
Title:
INSTRUMENTATION END OF CYCLE RECIRCULATION PUMP TRIP (EOC-RPT) INSTRUMENTATION TEXT 3.3.4.2 0
11/15/2002
Title:
INSTRUMENTATION ANTICIPATED TRANSIENT WITHOUT SCRAM RECIRCULATION PUMP TRIP (ATWS-RPT) INSTRUMENTATION TEXT 3.3.5.1 0
11/15/2002
Title:
INSTRUMENTATION EMERGENCY CORE COOLING SYSTEM (ECCS) INSTRUMENTATION TEXT 3.3.5.2
Title:
INSTRUMENTATION TEXT 3.3.6.1
Title:
INSTRUMENTATION TEXT 3.3.6.2
\\
Title:
INSTRUMENTATION TEXT 3.3.7.1
Title:
INSTRUMENTATION INSTRUMENTATION TEXT 3.3.8.1
Title:
INSTRUMENTATION TEXT 3.3.8.2
Title:
INSTRUMENTATION TEXT 3.4.1
Title:
REACTOR COOLANT TEXT 3.4.2
Title:
REACTOR COOLANT TEXT 3.4.3
Title:
11/15/2002 REACTOR CORE ISOLATION COOLING (RCIC) SYSTEM INSTRUMENTATION 1
11/09/2004 PRIMARY CONTAINMENT-ISOLATION INSTRUMENTATION 1
11/09/2004 SECONDARY CONTAINMENT ISOLATION INSTRUMENTATION 0
11/15/2002 CONTROL ROOM EMERGENCY OUTSIDE AIR SUPPLY (CREOAS)
SYSTEM 1
09/02/2004 LOSS OF POWER (LOP) INSTRUMENTATION 0
11/15/2002 REACTOR PROTECTION SYSTEM (RPS) ELECTRIC POWER MONITORING 2
11/22/2004 SYSTEM (RCS)
RECIRCULATION LOOPS OPERATING 0
11/15/2002 SYSTEM (RCS) JET PUMPS 0
11/15/2002 SYSTEM (RCS) SAFETY/RELIEF VALVES (S/RVS)
Page3 of 8 Report Date: 11/22/04 Page 3 of 8 Report Date: 11/22/04
SSES MANUAL Manual Name: TSB1 Manual
Title:
TECHNICAL SPECIFICATION BASES UNIT 1 MANUAL TEXT 3.4.4
Title:
REACTOR TEXT 3.4.5
Title:
REACTOR TEXT 3.4.6
Title:
REACTOR TEXT 3.4.7
Title:
REACTOR TEXT 3.4.8
Title:
REACTOR 0
COOLANT SYSTEM (RCS) 0 COOLANT SYSTEM (RCS) 6 0
COOLANT SYSTEM (RCS) 0 COOLANT SYSTEM (RCS) 0 COOLANT SYSTEM (RCS) 11/15/2002 RCS OPERATIONAL LEAKAGE 11/15/2002 RCS PRESSURE ISOLATION VALVE (PIV) LEAKAGE 11/15/2002 RCS LEAKAGE DETECTION INSTRUMENTATION 11/15/2002 RCS SPECIFIC ACTIVITY 11/15/2002 RESIDUAL HEAT REMOVAL (RHR) SHUTDOWN COOLING SYSTEM HOT SHUTDOWN TEXT 3.4.9
Title:
REACTOR COOLANT COLD SHUTDOWN TEXT 3.4.10
Title:
SYSTEM (RCS) 0 SYSTEM (RCS) 11/15/2002 RESIDUAL HEAT REMOVAL (RHR)
SHUTDOWN COOLING SYSTEM 11/15/2002 RCS PRESSURE AND TEMPERATURE (P/T) LIMITS TEXT 3.4.11 0
11/15/2002
Title:
REACTOR COOLANT SYSTEM (RCS) REACTOR STEAM DOME PRESSURE TEXT 3.5.1 0
11/15/2002
Title:
EMERGENCY CORE COOLING SYSTEMS (ECCS)
AND REACTOR CORE ISOLATION SYSTEM ECCS -
OPERATING TEXT 3.5.2 0
11/15/2002
Title:
EMERGENCY CORE COOLING SYSTEMS (ECCS)
AND REACTOR CORE ISOLATION SYSTEM ECCS -
SHUTDOWN TEXT 3.5.3 0
11/15/2002
Title:
EMERGENCY CORE COOLING SYSTEMS (ECCS)
AND REACTOR CORE ISOLATION SYSTEM RCIC SYSTEM COOLING (RCIC)
COOLING (RCIC)
COOLING (RCIC)
TEXT 3.6.1.1 0
11/15/2002
Title:
CONTAINMENT SYSTEMS PRIMARY CONTAINMENT Page4 of 8
Report Date: 11/22/04 Page 4 of 8 Report Date: 11/22/04
SSES MANUAL Manual Name: TSB1 Manual
Title:
TECHNICAL SPECIFICATION BASES UNIT 1 MANUAL TEXT 3.6.1.2
Title:
CONTAINMENT TEXT 3.6.1.3
Title:
CONTAINMENT TEXT 3.6.1.4
Title:
CONTAINMENT 0
11/15/2002 SYSTEMS PRIMARY CONTAINMENT AIR LOCK 0
11/15/2002 SYSTEMS PRIMARY CONTAINMENT ISOLATION VALVES (PCIVS)
LDCN 3092 0
11/15/2002 SYSTEMS CONTAINMENT PRESSURE TEXT 3.6.1.5
Title:
CONTAINMENT TEXT 3.6.1.6
Title:
CONTAINMENT TEXT 3.6.2.1
Title:
CONTAINMENT TEXT 3.6.2.2
Title:
CONTAINMENT TEXT 3.6.2.3
Title:
CONTAINMENT TEXT 3.6.2.4
Title:
CONTAINMENT TEXT 3.6.3.1
Title:
CONTAINMENT TEXT 3.6.3.2
Title:
CONTAINMENT TEXT 3.6.3.3
Title:
CONTAINMENT 0
11/15/2002 SYSTEMS DRYWELL AIR TEMPERATURE 0
11/15/2002 SYSTEMS SUPPRESSION CHAMBER-TO-DRYWELL VACUUM BREAKERS 0
11/15/2002 SYSTEMS SUPPRESSION POOL AVERAGE TEMPERATURE 0
11/15/2002 SYSTEMS SUPPRESSION POOL WATER LEVEL 0
11/15/2002 SYSTEMS RESIDUAL HEAT REMOVAL (RHR)
SUPPRESSION POOL COOLING 0
11/15/2002 SYSTEMS RESIDUAL HEAT REMOVAL (RHR)
SUPPRESSION POOL SPRAY 0
11/15/2002 SYSTEMS PRIMARY CONTAINMENT HYDROGEN RECOMBINERS 0
11/15/2002 SYSTEMS DRYWELL AIR FLOW SYSTEM 0
11/15/2002 SYSTEMS PRIMARY CONTAINMENT OXYGEN CONCENTRATION PageS of 8
Report Date: 11/22/04 Page 5 of 8 Report Date: 11/22/04
SSES MANUAL Manual Name:
TSB1 Manual
Title:
TECHNICAL SPECIFICATION BASES UNIT 1 MANUAL TEXT 3.6.4.1 0
11/15/2002
Title:
CONTAINMENT SYSTEMS SECONDARY CONTAINMENT TEXT 3.6.4.2 1
11/09/2004
Title:
CONTAINMENT SYSTEMS SECONDARY CONTAINMENT ISOLATION VALVES (SCIVS)
TEXT 3.6.4.3 2
11/09/2004
Title:
CONTAINMENT SYSTEMS STANDBY GAS TREATMENT (SGT)
SYSTEM TEXT 3.7.1
Title:
PLANT SYSTEMS ULTIMATE HEAT TEXT 3.7.2
Title:
PLANT SYSTEMS 0
11/15/2002 RESIDUAL HEAT REMOVAL SERVICE WATER (RHRSW)
SYSTEM AND THE SINK (UHS) 1 11/09/2004 EMERGENCY SERVICE WATER (ESW) SYSTEM TEXT 3.7.3 U~ 'Title:
PLANT TEXT 3.7.4
Title:
PLANT TEXT 3.7.5
Title:
PLANT TEXT 3.7.6
Title:
PLANT TEXT 3.7.7
Title:
PLANT TEXT 3.8.1 0
11/15/2002 SYSTEMS CONTROL ROOM EMERGENCY OUTSIDE AIR SUPPLY (CREOAS)
SYSTEM 0
11/15/2002 SYSTEMS CONTROL ROOM FLOOR COOLING SYSTEM 0
11/15/2002 SYSTEMS MAIN CONDENSER OFFGAS 0
11/15/2002 SYSTEMS MAIN TURBINE BYPASS SYSTEM 0
11/15/2002 SYSTEMS SPENT FUEL STORAGE POOL WATER LEVEL 1
10/17/2003
Title:
ELECTRICAL POWER SYSTEMS AC SOURCES - OPERATING TEXT 3.8.2 0
11/15/2002
Title:
ELECTRICAL POWER SYSTEMS AC SOURCES -
SHUTDOWN Page 6 of 8 Report Date: 11/22/04 Page 6 of 8 Report Date: 11/22/04
SSES MANUAL Manual Nane: TSBl Manual
Title:
TECHNICAL SPECIFICATION BASES UNIT I MANUAL TEXT 3.8.3
Title:
ELECTRICAL TEXT 3.8.4
Title:
ELECTRICAL TEXT 3.8.5
Title:
ELECTRICAL 0
11/15/2002 POWER SYSTEMS DIESEL FUEL OIL, LUBE OIL, AND STARTING AIR 0
11/15/2002 POWER SYSTEMS DC SOURCES -
OPERATING 0
11/15/2002 POWER SYSTEMS DC SOURCES -
SHUTDOWN TEXT 3.8.6
Title:
ELECTRICAL TEXT 3.8.7
Title:
ELECTRICAL TEXT 3.8.8 K>
Title:
ELECTRICAL TEXT 3.9.1
Title:
REFUELING I TEXT 3.9.2
Title:
REFUELING TEXT 3.9.3
Title:
REFUELING TEXT 3.9.4
Title:
REFUELING TEXT 3.9.5
Title:
REFUELING TEXT 3.9.6
Title:
REFUELING POWER SYST POWER SYST POWER SYST OPERATIONS OPERATIONS OPERATIONS OPERATIONS OPERATIONS OPERATIONS 0
11/15/2002
'EMS BATTERY CELL PARAMETERS 0
11/15/2002
'EMS DISTRIBUTION SYSTEMS -
OPERATING 0
11/15/2002
'EMS DISTRIBUTION SYSTEMS -
SHUTDOWN 0
11/15/2002 REFUELING EQUIPMENT INTERLOCKS 0
11/15/2002 REFUEL POSITION ONE-ROD-OUT INTERLOCK 0
11/15/2002 CONTROL ROD POSITION 0
11/15/2002 CONTROL ROD POSITION INDICATION 0
11/15/2002 CONTROL ROD OPERABILITY -
REFUELING 0
11/15/2002 REACTOR PRESSURE VESSEL (RPV)
WATER LEVEL Page7 of 8
Report Date: 11/22/04 Page 7 of 8 Report Date: 11/22/04
SSES MANUAL Manual Name: TSB1 Manual
Title:
TECHNICAL SPECIFICATION BASES UNIT 1 MANUAL TEXT 3.9.7 0
11/15/2002
Title:
REFUELING OPERATIONS RESIDUAL HEAT REMOVAL (RHR) -
HIGH WATER LEVEL TEXT 3.9.8 0
11/15/2002
Title:
REFUELING OPERATIONS RESIDUAL HEAT REMOVAL (RHR) -
LOW WATER LEVEL TEXT 3.10.1
Title:
SPECIAL 0
11/15/2002 OPERATIONS INSERVICE LEAK AND HYDROSTATIC TESTING OPERATION TEXT 3.10.2
Title:
SPECIAL TEXT 3.10.3
Title:
SPECIAL TEXT 3.10.4 U
Title:
SPECIAL TEXT 3.10.5
Title:
SPECIAL TEXT 3.10.6
Title:
SPECIAL TEXT 3.10.7
Title:
SPECIAL OPERATIONS OPERATIONS OPERATIONS OPERATIONS OPERATIONS OPERATIONS 0
11/15/2002 REACTOR MODE SWITCH INTERLOCK TESTING 0
11/15/2002 SINGLE CONTROL ROD WITHDRAWAL HOT SHUTDOWN 0
11/15/2002 SINGLE CONTROL ROD WITHDRAWAL COLD SHUTDOWN 0
11/15/2002 SINGLE CONTROL ROD DRIVE (CRD)
REMOVAL -
REFUELING 0
11/15/2002 MULTIPLE CONTROL ROD WITHDRAWAL -
REFUELING 0
11/15/2002 CONTROL ROD TESTING -
OPERATING TEXT 3.10.8
Title:
SPECIAL 0
11/15/2002 OPERATIONS SHUTDOWN MARGIN (SDM)
TEST -
REFUELING Page 8 of 8
Report Date: 11/22/04 Page 8 of 8 Report Date: 11/22/04
SUSQUEHANNA STEAM ELECTRIC STATION LIST OF EFFECTIVE SECTIONS (TECHNICAL SPECIFICATIONS BASES)
Section TOC Title Table of Contents'.
Revision 5 -
I B 2.0 SAFETY LIMITS BASES Page B 2.0-1 Page TS/ B 2.0-2 Page TS / B 2.0-3 Pages TS/ B 2.0-4 and TS J B 2.0-5 Page TS/B2.0-6 Pages B 2.0-7 through B 2.0-9 0
2 3
'2
.1 0
'B 3.0 LCO AND SR APPLICABILITY BASES Pages B 3.0-1 through B 3.0-7,'
Pages TS / B 3.0-8 and TS / B 3.0-9 Pages B 3.0-10 through B 3.0-12 Pages TS / B 3.0-13 through TS / B 3.0-15 B 3.1 REACTIVITY CONTROL BASES N
Pages B 3.1-1 through B 3.1-5 1.
Pages TS / B 3.1-6 and TS /.B 3.1-7-N Pages B 3.1-8 through B 3.1-27/7r^-.
Pages TS / B 3.1-28 Pages B 3.1-29 through B 3136 Pages TS I B 3.1-37 Pages B 3.1-38 through B3i. 51-,51 B 3.2 POWER DISTRIBUTION LIMITS BASES PageTS/B3.2;1 Page TS / P'3:2-2 Page TS J.B 3.2-3 -.,
Page7TS lB 3s24' Pages TS B 3.2-5 and TS B 3.2-6 PageB 3.2-7 (Pages 6TS / B 3.2-8 through TS / B 3.2-10 Page)TS / B 3.2-11
~PagdB '3.2-12 Page TS I B3.2-13 Pages B 3.2-14 and B 3.2-15 Page TS I B 3.2-16 Pages B 3.2-17 and B 3.2-18 Page TS / B 3.2-19 0
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B 3.3 INSTRUMENTATION Pages TS I B 3.3-1 through TS I B 3.3-10 Page TS / B 3.3-11 Pages TS / B 3.3-12 through TS / B 3.3-27
- Pages TS I B 3.3-28 through TS l.B 3.3-31
- Pages TS / B 3.3-32 and TS / B 3.3-33 SUSQUEHANNA - UNIT I TS/BLOES-1 Revision 53 SUSQUEHANNA - UNIT 1 TS / B LOES-1
-Revision 53
.__.4._
I SUSQUEHANNA STEAM ELECTRIC STATION LIST OFEFFECTIVE SECTIONS (TECHNICAL SPECIFICATIONS BASES)
Section Title Pages TS / B 3.3-34 through TS / B 3.3-43 Pages TS / B 3.3-43a through TS / B 3.3-43i Pages TS / B 3.3-44 through TS / B 3.3-54 Pages B 3.3-55 through B 3.3-63 Pages TS / B 3.3-64 and TS / B 3.3-65 Page TS / B 3.3-66 Page TS / B 3.3-67 Page TS /B 3.3-68 Pages TS / B 3.3-69 and TS / B 3.3-70 Pages TS I B 3.3-71 through TS I B 3.3-75 Page TS / B 3.3-75a Pages TS / B 3.3-75b through TS I B 3.3-75c Pages B 3.3-76 through B 3.3-89 Page TS / B 3.3-90 Page B 3.3-91 Page TS / B 3.3-92 through TS / B 3.3-1 00 Pages B 3.3-101 through B 3.3-103 Page TS / B 3.3-104 Pages B 3.3-105 and B 3.3-106 Page TS / B 3.3-107 Page B 3.3-108 Page TS / B 3.3-109 Pages B 3.3-110 and B 3.3-111 Pages TS / B 3.3-112 and TS / B 3.3-112a Pages B 3.3-113 and B 3.3-114 Page TS / B 3.3-115 Page TS / B 3.3-116 Page TS / B 3.3-117 Pages B 3.3-118 through B 3.3-122 Pages TS / B 3.3-123 through TS I B 3.3-124 Page TS / B 3.3-124a Pages B 3.3-125 and B 3.3-126 Page TS / B 3.3-127 Pages B 3.3-128 through B 3.3-130 Page TS / B 3.3-131 Pages B 3.3-132 through B 3.3-137 Page TS / B 3.3-138 Pages B 3.3-139 through B 3.3-149 Page TS / B 3.3-150 through TS / B 3.3-162 Page TS / B 3.3-163 Pages TS / B 3.3-164 through TS / B 3.3-177 Pages TS I B 3.3-178 and TS / B 3.3-179 Page TS I B 3.3-179a Page TS / B 3.3-179b Page TS / B 3.3-179c Page TS / B 3.3-1 80 Page TS / B 3.3-181 Revision' 1
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- UNIT I TSIBLOES-2 Revision 53 SUSQUEHANNA - UNIT 1 TS I B, LOES-2 Revision 53
SUSQUEHANNA STEAM ELECTRIC STATION UST OF EFFECTIVE SEC7TONS (TECHNICAL SPECIFICATIONS BASES)
Section Title.
Revision 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 3.3-204 0
PageTS/B3.3-205 1
Pages B 3.3-206 through B 3.3-219 0
B 3.4 REACTOR COOLANT SYSTEM BASES Pages B 3.4-1 and B 3.4-2 0
Page TS / B 3.4-3 and Page TS / B 3.4-4 3
Pages TS / B 3.4-5 through TS / B 3.4-9 2
Pages B 3.4-10 through B 3.4-14 0
PageTS/B3.4-15 1
Pages TS / B 3.4-16 and TS / B 3.4-17 2
Page TS / B 3.4-18 1
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Page 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 I B 3.4-60 1
B 3.5 ECCS AND RCIC BASES Pages B 3.5-1 and B 3.5-2 0
Page TS / B 3.5-3 2
Pages TS / B 3.5-4 and TS / B 3.5-5 1
Pages B 3.5-6 through B 3.5-10 0
PageTS/B3.5-11 1
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Pages TS / B 3.5-16 through TS / B 3.5-18 1
Pages B 3.5-19 through B 3.5-24 0
Page TS / B 3.5-25 1
Pages B 3.5-26 through B 3.5-31 0
B 3.6 CONTAINMENT SYSTEMS BASES Page TS / B 3.6-1 2
Page TS / B 3.6-1 a 3
Pages TS / B 3.6-2 through TS / B 3.6-5 2
PageTS/B3.6-6 3
Pages TS I B 3.6-6a and TS I B 3.6-6b 2
Page TS / B 3.6-6c 0
SUSQUEHANNA
- UNIT I TSIBLOES-3 Revision 53 SUSQUEHANNA - UNIT 1 TS / B LOES-3 Revision 53
SUSQUEHANNA STEAM ELECTRIC STATION LIST OF EFFECTIVE SECTIONS (TECHNICAL SPECIFICATIONS BASES)
Section Title Pages B 3.6-7 through B 3.6-14 Page TS / B 3.6-15 Pages TS / B 3.6-15a and TS I B 3.6-15b Page B 3.6-16 Page TS I B 3.6-17 Page TS I B 3.6-17a Pages TS / B 3.6-18 and TS B 3.6-19 Page TS I B 3.6-20 Page TS l B 3.6-21 Page TS l B 3.6-22 Page TS I B 3.6-22a Page TS I B 3.6-23 Pages TS / B 3.6-24 through TS I B 3.6-25 Page TS lB 3.6-26 Page TS l B 3.6-27 Page TSlB3.6-28 Page TS lB 3.6-29 Page TS/B 3.6-30 Page TS l B 3.6-31 Pages B 3.6-32 through B 3.6-35 PageTS/B3.6-36 Page B 3.6-37 Page TS/B 3.6-38 Page B 3.6-39 Page TS / B 3.6-40 Pages B 3.641 through B 3.6-43 Pages TS / B 3.6-44 through TS / B 3.6-51 Page TS / B 3.6-52.
Pages B 3.6-53 through B 3.6-63 PageTS /8 3.6-64 Pages B 3.6-65 through B 3.6-83 Page TS / B 3.6-84 Pages TS / B 3.6-85 through TS I B 3.6-88 Page TS / B 3.6-88a Page TS / B 3.6-89 Page TS / B 3.6-90 Page TS / B 3.6-91 Pages TS / B 3.6-92 through TS I B 3.6-96 Page TS / B 3.6-97 Pages TS / B 3.6-98 and TS I B 3.6-99 Page TS / B 3.6-100 Pages TS / B 3.6-101 and TS / B 3.6-102 Pages TS / B 3.6-103 through TS / B 3.6-105 Pages TS / B 3.6-106 and TS / B 3.6-107 Revision 0
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SUSQUEHANNA STEAM ELECTRIC STATION LIST OF EFFECTIVE SECTIONS (TECHNICAL SPECIFICATIONS BASES)
Section Title Revision 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 3.7-6b and TS I B 3.7-6c 0
Page TS / B 3.7-7 2
Pages TS / B 3.7-8 through TS / B 3.7-11 1
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
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B 3.9 REFUELING OPERATIONS BASES Pages TS / B 3.9-1 and TS / B 3.9-1a I
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TSB1 text LOES 1 U9/04 SUSQUEHANNA - UNIT I TSIBLOES-5 Revision 53 SUSQUEHANNA - UNIT 1 ITS / B LOES-5 Revision 53
TABLE OF CONTENTS (TECHNICAL SPECIFICATIONS BASES)
B2.0 SAFETY LIMITS (SLs).................................................
B2.0-1 B2.1.1 Reactor Core SLs................................................
B2.O-1 B2.1.2 Reactor Coolant System (RCS) Pressure SL................................. B2.0-7 B3.0 LIMITING CONDITION FOR OPERATION (LCO) APPLICABILITY.............. B3.0-1 B3.0 SURVEILLANCE REQUIREMENT (SR) APPLICABILITY............................ B3.0-10 B3.1 REACTIVITY CONTROL SYSTEMS................................................
B3.1-1 B3.1.1 Shutdown Margin (SDM)................................................
B3.1-1 B3.1.2 Reactivity Anomalies................................................
B3.1-8 B3.1.3 Control Rod OPERABILITY................................................
B3.1-13 B3.1.4 Control Rod Scram Times........................
........................ B3.1-22 B3.1.5 Control Rod Scram Accumulators................................................ B3.1-29 B3.1.6 Rod Pattern Control..............
.................................. 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 83.2 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).............................................. B3.2-10 B3.2.4 Average Power Range Monitor (APRM) Gain and Setpoints...
8; B3.2-14 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 B3.3.1.3 Oscillation Power Range Monitor (OPRM)..Power.Range.Monitor.(OPR M)TS/B3.3-43a B3.3.2.1 Control Rod Block Instrumentation.;
TS/B3.3-44 B3.3.2.2 Feedwater - Main Turbine High Water Level Trip Instrumentation......
B3.3-55 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 Recirculation Pump Trip (EOC-RPT)
Instrumentation.
B3.3-81 B3.3.4.2 Anticipated Transient Without Scram Recirculation Pump Trip (A7WS-RPT) Instrumentation............................. ;.... B3.3-92 B3.3.5.1 Emergency Core Cooling System (ECCS)
Instrumentation.
3.3-101 B3.3.5.2 Reactor Core Isolation Cooling (RCIC) System Instrumentation.........
8.;
B3.3-135 B3.3.6.1 Primary Containment Isolation Instrumentation............................. B3.3-147 B3.3.6.2 Secondary Containment Isolation Instrumentation.TS/3.3-180 B3.3.7.1 Control Room Emergency Outside Air Supply (CREOAS)
System Instrumentation....
B3.3-192 (continued)
SUSQUEHANNA - UNIT 1 TS/ BTOC - 1 Revision 5
TABLE OF CONTENTS (TECHNICAL SPECIFICATIONS BASES)
B3.3 INSTRUMENTATION (continued)
B3.3.8.1 Loss of Power (LOP) Instrumentation....................................
TS/B3.3-205 B3.3.8.2 Reactor Protection System (RPS) Electric Power Monitoring...................
B3.3-213 B3.4 REACTOR COOLANT SYSTEM (RCS).
.................................. B3.4-1 B3.4.1 Recirculation Loops Operating..................................
B3.4-1 B3.4.2 Jet Pumps.......................
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..............................
8...
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 B3.4.11 Reactor Steam Dome Pressure.....................
TS/B3.4-58 B3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) AND REACTOR CORE ISOLATION COOLING (RCIC) SYSTEM.................................. B3.5-1 B3.5.1 ECCS - Operating............
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/B3.6-1 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.B3.6-41 B3.6.1.5 Drywell Air Temperature.TS/B3.6-44 B3.6.1.6 Suppression Chamber-to-Drywell Vacuum Breakers TS/B3.6-47 B3.6.2.1 Suppression Pool Average Temperature............................
B3.6-53 B3.6.2.2 Suppression Pool Water Level............................
B3.6-59 B3.6.2.3 Residual Heat Removal (RHR) Suppression Pool Cooling........................................
B3.6-62 B3.62.4 Residual Heat Removal (RHR) Suppression Pool Spray................ B3.6-66 B3.6.3.1 Primary Containment Hydrogen Recombiners................................ 83.6-70 B3.6.3.2 Drywell Air Flow System........................................
8 3.6-76 B3.6.3.3 Primary Containment Oxygen Concentration............................... ;.: B3.6-81 B3.6.4.1 Secondary Containment
............. TS/B3.6-84 B3.6.4.2 Secondary Containment Isolation Valves (SCIVs).................... TS/B3.6-91 B3.6.4.3 Standby Gas Treatment (SGT) System................................ :.TS/B3.6-101 (continued)
SUSQUEHANNA-UNIT1 TS/BTOC-2 Revision 5
TABLE OF CONTENTS (TECHNICAL SPECIFICATIONS BASES)
B3.7 PLANT SYSTEMS..................................
TSIB3.7-1 B3.7.1 Residual Heat Removal Service Water (RHRSW) System and the Ultimate Heat Sink (UHS)............................. TS/B3.7-1 B3.7.2 Emergency Service Water (ESW) System............................. TS/B3.7-7 B3.7.3 Control Room Emergency Outside Air Supply (CREOAS) System.TS/B3.7-12 B3.7.4 Control Room Floor Cooling System.TS/B3.7-19 B3.7.5 Main Condenser Offgas.B3.7-24 B3.7.6 Main Turbine Bypass System.TS/B3.7-27 B3.7.7 Spent Fuel Storage Pool Water Level.B3.7-31 B3.8 ELECTRICAL POWER SYSTEM.TS/B3.8-1 B3.8.1 AC Sources - Operating.TS/B3.8-1 B3.8.2 AC Sources - Shutdown.B3.8-38.
B3.8.3 Diesel Fuel Oil, Lube Oil, and Starting Air.B3.8-45 B3.8.4 DC Sources - Operating.TS/B3.8-54 B3.8.5 DC Sources - Shutdown.B3.8-66 B3.8.6 Battery Cell Parameters.
B3.8-71 B3.8.7 Distribution Systems - Operating.B3.8-78 B3.8.8 Distribution Systems - Shutdown.B3.8-86 B3.9
-REFUELING OPERATIONS.
TSIB3.9-1 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......................................... B3.10-29 B3.10.8 SHUTDOWN MARGIN (SDM) Test-Refueling............................. B3.10-33 TSB1 Ted TOC 1119/04 SUSQUEHANNA -
UNIT I TSIBTOC-3 Revision 5 SUSQUEHANNA - UNIT 1 TS/BTOC-3 Revision 5
PPL Rev. 0 OPRM Instrumentation B 3.3.1.3 B 3.3 INSTRUMENTATION B 3.3.1.3 Oscillation Power Range Monitor (OPRM)
BASES I
BACKGROUND General Design Criterion 10 (GDC 10) requires the reactor core and associated coolant, control, and protection systems to be designed with appropriate margin to assure that acceptable fuel design limits are not exceeded during any condition of normal operation including the affects
.of anticipated operational occurrences. Additionally, GDC 12 requires the reactor core and associated coolant control and protection systems to be designed to assure that power oscillations which can result in conditions exceeding acceptable fuel design limits are either not possible or can be reliably and readily detected and suppressed. The OPRM System provides compliance with GDC 10 and GDC 12 thereby providing protection from exceeding the fuel MCPR safety limit.
References 1, 2, and 3 describe three separate algorithms for detecting stability related oscillations: the period based detection algorithm, the amplitude based algorithm, and the growth rate algorithm. The OPRM System hardware implements these algorithms in microprocessor based modules. These modules execute the algorithms based on LPRM inputs and generate alarms and trips based on these calculations. These trips result in tripping the Reactor Protection System (RPS) when the appropriate RPS trip logic is satisfied, as described in the Bases for LCO 3.3.1.1, "RPS Instrumentation." Only the period based detection algorithm is used in the safety analysis (Ref. 1, 2, 6, & 7). The remaining algorithms provide defense-in-depth and additional protection against unanticipated oscillations.
The period based detection algorithm detects a stability-related oscillation based on the occurrence of a fixed number of consecutive LPRM signal period confirmations followed by the LPRM signal amplitude exceeding a specified setpoint. Upon detection of a stability related oscillation a trip is generated for that OPRM channel.
(continued)
SUSQUEHANNA - UNIT 1 ITS / B 3.3-43a Revision 0
PPL Rev. 0 OPRM Instrumentation B 3.3.1.3 BASES BACKGROUND (continued)
The OPRM System consists of 4 OPRM trip channels, each channel consisting of two OPRM modules. Each.OPRM module receives input from LPRMs. Each OPRM module also receives input from the NMS average power range monitor (APRM) power and flow signals to automatically enable the trip function of the OPRM module.
Each OPRM module is continuously tested by a self-test function. On detection of any OPRM module failure, either a Trouble alarm or INOP alarm is activated. The OPRM module provides an INOP alarm when the self-test feature indicates that the OPRM module may not be capable of meeting its functional requirements.
APPLICABLE SAFETY ANALYSES It has been shown that BWR cores may exhibit thermal-hydraulic reactor instabilities in high power and low flow portions of the core power to flow operating domain. GDC 10 requires the reactor core and associated coolant control and protection systems to be designed with appropriate margin to assure that acceptable fuel design limits are not exceeded during any condition of normal operation, including the effects of anticipated operational occurrences. GDC 12 requires assurance that power oscillations which can result in conditions exceeding acceptable fuel design limits are either not possible or can be reliably and readily detected and suppressed. The OPRM System provides compliance with GDC 10 and GDC 12 by detecting the onset of oscillations and suppressing them by initiating a reactor scram. This assures that the MCPR safety limit will not be violated for anticipated oscillations.
The OPRM Instrumentation satisfies Criteria 3 of the NRC Policy Statement.
LCO Four channels of the OPRM System are required to be OPERABLE to ensure that stability related oscillations are detected and suppressed prior to exceeding the MCPR safety limit. Only one of the two OPRM modules is required for.OPRM channel OPERABILITY. The minimum number of LPRMs required OPERABLE to maintain an OPRM channel OPERABLE is consistent with the minimum number of LPRMs required to maintain the APRM system OPERABLE per LCO 3;3.1.1.
(continued)
SUSQUEHANNA - UNIT 1 TS / B3.3-43b Revision 0
PPL Rev. 0 OPRM Instrumentation B 3.3.1.3 BASES LCO (continued)
The OPRM setpoints are determined based on the NRC approved methodology described in NEDO-32465-A (Ref 6). The Allowable Value for the OPRM Period Based Algorithm setpoint (SP) is derived from the analytic limit corrected for instrument and calibration errors as contained in the COLR.
The OPRM bypass flow setpoint (SR 3.3.1.3.5) is conservatively established based on the greater of 60 MLb/Hr. (NEDO-32465-A) and the value obtained based on the NRC approved methodology described in EMF-CC-074(P)(A), Volume 4, (Ref. 11).
APPLICABILITY The OPRM instrumentation is required to be OPERABLE in order to detect and suppress neutron flux oscillations in the event of thermal-hydraulic instability. As described in References 1, 2, and 3, the power/core flow region protected against anticipated oscillations is defined by THERMAL POWER 2 30% RTP and core flow ' 65 MLb/Hr.
The OPRM trip is required to be enabled in this region, and the OPRM must be capable of enabling the trip function as a result of anticipated transients that place the core in that power/flow condition. Therefore, the OPRM is required to be OPERABLE with THERMAL POWER > 25% RTP and at all core flows while above that THERMAL POWER. It is not necessary for the OPRM to be operable with THERMAL POWER < 25%
RTP because transients from below this THERMAL POWER are not anticipated to result in power that exceeds 30% RTP.
ACTIONS A Note has been provided to modify the ACTIONS related to the OPRM 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 OPRM 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 OPRM instrumentation channel.
(continued)
SUSQUEHANNA - UNIT I TS I B 3.3-43c Revision 0
PPL Rev. 0 OPRM Instrumentation B 3.3.1.3 BASES ACTIONS A.1. A.2 and A.3 (continued)
Because of the reliability and on-line self-testing of the OPRM instrumentation and the redundancy of the RPS design, an allowable out of service time of 30 days has been shown to be acceptable (Ref. 7) to permit restoration of any inoperable channel to OPERABLE status.
However, this out of service time is only acceptable provided the OPRM instrumentation still maintains OPRM trip capability (refer to Required Actions B.1 and B.2). The remaining OPERABLE OPRM channels continue to provide trip capability (see Condition B) and provide operator information relative to stability activity. The remaining OPRM modules have high reliability. With this high reliability, there is a low probability of a subsequent channel failure within the allowable out of service time. In addition, the OPRM modules continue to perform on-line self-testing and alert the operator if any further system degradation occurs.
If the inoperable channel cannot be restored to OPERABLE status within the allowable out of service time, the OPRM channel or associated RPS trip system must be placed in the tripped condition per Required actions A.1 and A.2. Placing the inoperable OPRM channel in trip (or the associated RPS trip system in trip) would conservatively compensate for the inoperability, provide the capability to accommodate a single failure, and allow operation to continue. Alternately, if it is not desired to place the OPRM channel (or RPS trip system) in trip (e.g., as in the case where placing the inoperable channel in trip would result in a full scram), the alternate method of detecting and suppressing thermal hydraulic instability oscillations is required (Required Action A.3). This alternate method is described in Reference 5. It consists of increased operator awareness and monitoring for neutron flux oscillations when operating in the region where oscillations are possible. If indications of oscillation, as described in Reference 5 are observed by the operator, the operator will take the actions described by procedures which include initiating a manual scram of the reactor. The power/flow map regions are developed based on methodology in Reference 11. The applicable regions are contained in the COLR.
-(continued)
SUSQUEHANNA - UNIT 1 TS / B3.3-43d Revision 0
PPL Rev. 0 OPRM Instrumentation B 3.3.1.3 BASES ACTIONS (continued)
B.1 and B.2 Required action B.1 is intended to ensure that-appropriate actions are taken if multiple, inoperable, untripped OPRM channels within the same RPS trip system result in not maintaining OPRM trip capability. OPRM trip capability is considered to be maintained when sufficient OPRM channels are OPERABLE or in trip (or the associated RPS trip system is in trip), such that a valid OPRM signal will generate a trip signal in both RPS trip systems (this would require both RPS trip systems to have at least one OPRM channel OPERABLE or the associated RPS trip system in trip).
Because of the low probability of the occurrence of an instability, 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is an acceptable time to initiate the alternate method of detecting and suppressing thermal hydraulic instability oscillations described in Action A.3 above. The alternate method of detecting and suppressing thermal hydraulic instability oscillations would adequately address detection and mitigation in the event of instability oscillations. Based on industry operating experience with actual instability oscillation, 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. Since plant operation is minimized in areas where oscillations may occur, operation for 120 days without OPRM trip capability is considered acceptable with implementation of the alternate method of detecting and suppressing thermal hydraulic instability oscillations.
C.1 With any Required Action and associated Completion Time not met, 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 />.
Reducing THERMAL POWER to < 25% RTP places the plant in a condition where instabilities are not likely to occur. The 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is reasonable, based on operating experience, to reduce THERMAL POWER < 25% RTP from full power conditions in an orderly manner and without challenging plant systems.
(continued)
I SUSQUEHANNA-UNIT I
-. TS / B 3.343e Revision 0
PPL Rev. 0 OPRM Instrumentation B 3.3.1.3
- 3.
BASES (continued)
SURVEILLANCE SR 3.3.1.3.1 REQUIREMENTS A CHANNEL FUNCTIONAL TEST is performed to ensure that the entire channel will perform the intended function. A Frequency of 184 days provides an acceptable level of system average availability over the Frequency and is based on the reliability of the channel (Ref. 7).
SR 3.3.1.3.2 LPRM gain settings are determined from the local flux profiles measured by the Traversing Incore Probe (TIP) System. This establishes the relative local flux profile for appropriate representative input to the OPRM System. The 1000 MWD/MT Frequency is based on operating experience with LPRM sensitivity changes.
SR 3.3.1.3.3 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. Calibration of the channel provides a check of the internal reference voltage and the internal processor clock frequency. It also compares the desired trip setpoints with those in processor memory. Since the OPRM is a digital system, the internal reference voltage and processor clock frequency are, in turn, used to automatically calibrate the internal analog to digital converters. As noted, 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 1000 MWD/MT LPRM calibration using the TlPs (SR 3.3.1.3.2).
(continued)
SUSQUEHANNA - UNIT 1 TS / B 3.3-43f Revision 0
PPL Rev. 0 OPRM Instrumentation B 3.3.1.3 BASES SURVEILLANCE The Frequency of 24 months is based upon the assumption of the REQUIREMENTS magnitude of equipment drift provided by the equipment supplier. (Ref. 7)
(continued)
SR 3.3.1.3.4 The LOGIC SYSTEM FUNCTIONAL TEST demonstrates the OPERABILITY of the required trip logic for a specific channel. The functional testing of control rods, in LCO 3.1.3, "Control Rod OPERABILITY," and scram discharge volume (SDV) vent and drain valves, in LCO 3.1.8, Scram Discharge Volume (SDV) Vent and Drain Valves," overlaps this Surveillance to provide complete testing of the assumed safety function. The OPRM self-test function may be utilized to perform this testing for those components that it is designed to monitor.
The 24 month Frequency is based on engineering judgement, reliability of the components and operating experience.
SR 3.3:1.3.5 The SR ensures that trips initiated from the OPRM System will not be inadvertently bypassed when THERMAL POWER is ? 30% RTP and core flow is 5 65 MLb/Hr. This normally involves calibration of the bypass channels. Adequate margins for the instrument setpoint methodology are incorporated into the actual setpoints (Ref. 7).
If any bypass channel setpoint is nonconservative (i.e., the OPRM module is bypassed at 2 30% RTP and core flow is m 65 MLb/Hr), then the affected OPRM module is considered inoperable. Alternatively, the bypassed channel can be manually placed in the conservative position (Manual Enable). If placed in the MANUAL ENABLE condition, this SR is met and the module is considered OPERABLE.
The 24 month Frequency is based on engineering judgement and reliability of the components.
(continued)
SUSQUEHANNA - UNIT I TS / B3.3-43g Revision 0
PPL Rev. 0 OPRM Instrumentation B 3.3.1.3 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.3.1.3.6 This SR ensures that the individual channel response times are less than or equal to the maximum values assumed in the safety analysis (Ref. 6).
The OPRM self-test function may be utilized to perform this testing for those components it is designed to monitor. The LPRM amplifier cards inputting to the OPRM are excluded from the OPRM RESPONSE TIME testing. The RPS RESPONSE TIME acceptance criteria are included in Reference 8.
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 a 24 month STAGGERED TEST BASIS. This Frequency is based upon operating experience, which shows that random failures of instrumentation components causing serious time degradation, but not channel failure, are infrequent occurrences.
(continued)
SUSQUEHANNA - UNIT 1 TS I B 3.3-43h Revision 0
PPL Rev. 0 OPRM Instrumentation B 3.3.1.3 BASES REFERENCES
- 1. NEDO 31960-A, "BWR Owners Group Long-Term Stability Solutions Licensing Methodology". November 1995.-
- 2. NEDO 31960-A, Supplement 1 "BWR Owners Group Long-Term Stability Solutions Licensing Methodology", November 1995.
- 3. NRC Letter, A. Thadani to L.A. England, "Acceptance for Referencing of Topical Reports NEDO-31960, Supplement 1, 'BWR Owners Group Long-Term Stability Solutions Licensing Methodology".
July 12, 1994.
- 4. Generic Letter 94-02, uLong-Term Solutions and Upgrade of Interim Operating Recommendations for Thermal-Hydraulic Instabilities in Boiling Water Reactors", July 11, 1994.
- 5. BWROG Letter BWROG-9479, "Guidelines for Stability Interim Corrective Action", June 6, 1994.
- 6. NEDO-32465-A, "BWR Owners Group Reactor Stability Detect and Suppress Solutions Licensing Basis Methodology and Reload Applications", August 1996.
- 7. CENPD-400-P-A, Rev 01, "Generic Topical Report for the ABB Option IlIl Oscillation Power Range Monitor (OPRM)", May 1995.
- 8. FSAR Table 7.3-28.
- 9. FSAR Section 4.4.4.6.
- 10. FSAR Section 7.2.
- 11. EMF-CC-074(P)(A), Volume 4, uBWR Stability Analysis: Assessment of STAIF with Input from MICROBURN-B2."
SUSQUEHANNA UNIT I TS I B 3.3-43i Revision 0 SUSQUEHANNA - UNIT 1 TS / B 3.3-43i Revision 0
PPL Rev. 2 Recirculation Loops Operating B 3.4.1 B3.4 B 3.4.1 REACTOR COOLANT SYSTEM (RCS)
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 intemals.
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 from the vessel, through 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 cladding, where heat (continued)
SUSQUEHANNA - UNIT 1 B 3.4-1 Revision 0
R PPL Rev. 2 Recirculaton Loops Operating B 3.4.1 BASES BACKGROUND (continued) is transferred to the coolant. As it rises, the coolant 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 SAFETY ANALYSES The operation of the Reactor Coolant Recirculation System is an initial condition assumed in the design basis loss of coolant accident (LOCA)
(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 the core flow response 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)
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PPL Rev. 2 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 ATRIUM'-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 flow biased simulated THERMAL POWER 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 Policy 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 Flow Biased 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)
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PPL Rev. 2 Recirculation Loops Operating B 3.4.1 BASES LCO (continued)
The LCO is modified by a Note that allows up tp 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> to establish the 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 charactdristics 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 3.3.1.3 (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)
I SUSQUEHANNA - UNIT 1 TS / B3.4-4 Revision 3
PPL Rev. 2 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 g
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 without 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)
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-TS/ B3.4-5 Revision 2
PPL Rev. 2 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,u 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).
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