ML060260367
| ML060260367 | |
| Person / Time | |
|---|---|
| Site: | Susquehanna  |
| Issue date: | 01/17/2006 |
| From: | Susquehanna |
| To: | Gerlach R Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| 028401 | |
| Download: ML060260367 (99) | |
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MANUAL HARD COPY DISTRIBUTION DOCUMENT TRANSMITTAL 2006-2897 USER INFORMATION:
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ID:
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- USERS, ELECTRONICALLY REVIEW THE APPROPRIATE DOCUMENTS AND ACKNOWLEDGE COMPLETE IN YOUR NIMS INBOX.
SSES MANUAL Manual Name:
TSB1 Manual
Title:
TECHNICAL SPECIFICATION BASES UNIT 1 MANUAL Table Of Contents Issue Date:
01/16/2006 Procedure Name Rev TEXT LOES 68
Title:
LIST OF EFFECTIVE SECTIONS Issue Date 01/16/2006 Change ID Change Number TEXT TOC
Title:
TABLE OF CONTENTS 7
04/18/2005 TEXT 2.1.1 1
Title:
SAFETY LIMITS (SLS) REACTOR 04/27/2004 CORE SLS TEXT 2.1.2 0
Title:
SAFETY LIMITS (SLS) REACTOR 11/15/2002 COOLANT SYSTEM (RCS) PRESSURE SL TEXT 3.0
Title:
LIMITING CONDITION 1
04/18/2005 FOR OPERATION (LCO) APPLICABILITY TEXT 3.1.1
Title:
REACTIVITY TEXT 3.1.2
Title:
REACTIVITY 0
7 CONTROL SYSTEMS C 0S CONTROL SYSTEMS 11/1512002 SHUTDOWN MARGIN (SDM)
- 11/15/2002 REACTIVITY ANOMALIES TEXT 3.1.3
Title:
REACTIVITY 1
07/06/2005 CONTROL SYSTEMS CONTROL ROD OPERABILITY TEXT 3.1.4
Title:
REACTIVITY TEXT 3.1.5
Title:
REACTIVITY TEXT 3.1.6
Title:
REACTIVITY 1
CONTROL SYSTEMS 1
CONTROL SYSTEMS 1
CONTROL SYSTEMS 07/06/2005 CONTROL ROD SCRAM TIMES 07/06/2005 CONTROL ROD SCRAM ACCUMULATORS 02/17/2005 ROD PATTERN CONTROL Pagel of 8 Report Date: 01/16/06 Page 1 of 8 Report Date: 01/16/06
SSES MANUAL Manual Name: TSB1 Manual
Title:
TECHNICAL SPECIFICATION BASES UNIT 1 MANUAL TEXT 3.1.7
Title:
REACTIVITY CONTROL TEXT 3.1.8
Title:
REACTIVITY CONTROL TEXT 3.2.1
Title:
POWER DISTRIBUTION TEXT 3.2.2
Title:
POWER DISTRIBUTION 1
08/30/2005 SYSTEMS STANDBY LIQUID CONTROL (SLC) SYSTEM 1
10/19/2005 SYSTEMS SCRAM DISCHARGE VOLUME (SDV)
VENT AND DRAIN VALVES 0
11/15/2002 LIMITS AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR) 0 11/15/2002 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 1
07/06/2005
Title:
POWER DISTRIBUTION LIMITS AVERAGE POWER RANGE MONITOR (APRM)
GAIN AND SETPOINTS TEXT 3.3..1.1 2
07/06/2005
Title:
INSTRUMENTATION REACTOR PROTECTION SYSTEM (RPS) INSTRUMENTATION TEXT 3.3.1.2 0
11/15/2002
Title:
INSTRUMENTATION SOURCE RANGE MONITOR (SRM) INSTRUMENTATION TEXT 3.3.1.3
Title:
OPRM INSTRUMENTATION 0
11/22/2004 TEXT 3.3.2.1 1
02/17/2005
Title:
INSTRUMENTATION CONTROL ROD BLOCK INSTRUMENTATION TEXT 3.3.2.2 0
11/15/2002
Title:
INSTRUMENTATION FEEDWATER -
MAIN TURBINE HIGH WATER LEVEL TRIP INSTRUMENTATION TEXT 3.3.3.1 2
07/06/2005
Title:
INSTRUMENTATION POST ACCIDENT MONITORING (PAM) INSTRUMENTATION Page2 of 8
Report Date: 01/16/06 Page 2 of 8 Report Date: 01/16/06
SSES MANUAL Manual Name:
TSB1 Manual
Title:
TECHNICAL SPECIFICATION BASES UNIT 1 MANUAL V
TEXT 3.3.3.2 1
04/18/2005
Title:
INSTRUMENTATION REMOTE SHUTDOWN SYSTEM TEXT 3.3.4.1 0
11/1.5/2002
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
Title:
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 2
07/06/2005 EMERGENCY CORE COOLING SYSTEM (ECCS) INSTRUMENTATION 0
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 TEXT 3.3.8.1 1
09/02/2004
Title:
INSTRUMENTATION LOSS OF POWER (LOP) INSTRUMENTATION TEXT 3.3.8.2 0
11/15/2002
Title:
INSTRUMENTATION REACTOR PROTECTION SYSTEM (RPS) ELECTRIC POWER MONITORING TEXT 3.4.1 2
11/22/2004
Title:
REACTOR COOLANT SYSTEM (RCS) RECIRCULATION LOOPS OPERATING TEXT 3.4.2 0
11/15/2002
Title:
REACTOR COOLANT SYSTEM (RCS) JET PUMPS Page3 of 8
Report Date: 01/16/06 Page 3 of 8 Report Date: 01116106
SSES MANUAL Manual Name: TSB1 Manual
Title:
TECHNICAL SPECIFICATION BASES UNIT 1 MANUAL TEXT 3.4.3
Title:
REACTOR TEXT 3.4.4
Title:
REACTOR TEXT 3.4.5
Title:
REACTOR TEXT 3.4.6
Title:
REACTOR TEXT 3.4.7
Title:
REACTOR 1
01/16/2006 COOLANT SYSTEM (RCS) SAFETY/RELIEF VALVES (S/RVS) 0 11/15/2002 COOLANT SYSTEM (RCS) RCS OPERATIONAL LEAKAGE 1
01/16/2006 COOLANT SYSTEM (RCS) RCS PRESSURE ISOLATION VALVE (PIV) LEAKAGE 1
04/18/2005 COOLANT SYSTEM (RCS) RCS LEAKAGE DETECTION INSTRUMENTATION 1
04/18/2005 COOLANT SYSTEM (RCS) RCS SPECIFIC ACTIVITY 1
04/18/2005 COOLANT SYSTEM (RCS) RESIDUAL HEAT REMOVAL (RHR) SHUTDOWN COOLING SYSTEIo_
UTDOWN 0
11/15/2002 COOLANT SYSTEM (RCS) RESIDUAL HEAT REMOVAL (RHR) SHUTDOWN COOLING SYSTEM HUTDOWN 0
11/15/2002 COOLANT SYSTEM (RCS) RCS PRESSURE AND TEMPERATURE (P/T) LIMITS TEXT 3.4.8
Title:
REACTOR HOT SH TEXT 3.4.9
Title:
REACTOR COLD S TEXT 3.4.10
Title:
REACTOR TEXT 3.4.11
Title:
REACTOR TEXT 3.5.1 COOLANT 0
11/15/2002 SYSTEM (RCS) REACTOR STEAM DOME PRESSURE 2
01/16/2006
Title:
EMERGENCY CORE COOLING SYSTEMS (ECCS) AND REACTOR CORE ISOLATION COOLING (RCIC)
SYSTEM ECCS -
OPERATING TEXT 3.5.2 0
11/15/2002
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 Page4 of 8
Report Date: 01/16/06 Page 4 of 8 Report.Date: 01/16/06
SSES MANUAL
-* Manual Name:
TSBl Manual
Title:
TECHNICAL SPECIFICATION BASES UNIT 1 MANUAL TEXT 3.6.1.1
Title:
CONTAINMENT TEXT 3.6.1.2
Title:
CONTAINMENT TEXT 3.6.1.3
Title:
CONTAINMENT TEXT 3.6.1.4
Title:
CONTAINMENT TEXT 3.6.1.5
Title:
CONTAINMENT TEXT 3.6.1.6 Cp,,'
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 0
11/15/2002 SYSTEMS PRIMARY CONTAINMENT 0
11/15/2002 SYSTEMS PRIMARY CONTAINMENT AIR LOCK 3
12/08/2005 SYSTEMS PRIMARY CONTAINMENT ISOLATION VALVES (PCIVS) 0 11/15/2002 SYSTEMS CONTAINMENT PRESSURE 1
10/05/2005 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 1
01/16/2006 SYSTEMS RESIDUAL HEAT REMOVAL (RHR)
SUPPRESSION POOL COOLING 0
11/15/2002 SYSTEMS RESIDUAL HEAT REMOVAL (RHR)
SUPPRESSION POOL SPRAY 1
04/18/2005 SYSTEMS PRIMARY CONTAINMENT HYDROGEN RECOMBINERS 1
04/18/2005 SYSTEMS DRYWELL AIR FLOW SYSTEM PageS of 8
Report Date: 01/16/06 Page 5 of 8
Report Date: 01/16/06
SSES MANUAL Manual Name:
TSB1 Manual
Title:
TECHNICAL SPECIFICATION BASES UNIT 1 MANUAL TEXT 3.6.3.3 0
11/15/2002
Title:
CONTAINMENT SYSTEMS PRIMARY CONTAINMENT OXYGEN CONCENTRATION TEXT 3.6.4.1 4
10/24/2005
Title:
CONTAINMENT SYSTEMS SECONDARY CONTAINMENT TEXT 3.6.4.2 2
01/03/2005
Title:
CONTAINMENT SYSTEMS SECONDARY CONTAINMENT ISOLATION VALVES (SCIVS)
TEXT 3.6.4.3 3
10/24/2005
Title:
CONTAINMENT SYSTEMS STANDBY GAS TREATMENT (SGT) SYSTEM TEXT 3.7.1 0
11/15/2002
Title:
PLANT SYSTEMS RESIDUAL HEAT REMOVAL SERVICE WATER (RHRSW) SYSTEM AND THE ULTIMATE HEAT SINK (UHS)
TEXT 3.7.2
Title:
PLANT TEXT 3.7.3
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 1
11/09/2004 SYSTEMS EMERGENCY SERVICE WATER (ESW)
SYSTEM 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 1
01/17/2005 SYSTEMS MAIN TURBINE BYPASS SYSTEM 0
11/15/2002 SYSTEMS SPENT FUEL STORAGE POOL WATER LEVEL 3
10/05/2005 111
Title:
ELECTRICAL POWER SYSTEMS AC SOURCES -
OPERATING Page 6 of 8 3
Report Date: 01/16/06
SSES MANUAL Manual Name:
TSB1 Manual
Title:
TECHNICAL SPECIFICATION BASES UNIT 1 MANUAL TEXT 3.8.2
Title:
ELECTRICAL 0
11/1.5/2002 POWER SYSTEMS AC SOURCES -
SHUTDOWN TEXT 3.8.3
Title:
ELECTRICAL TEXT 3.8.4
Title:
ELECTRICAL TEXT 3.8.5
Title:
ELECTRICAL TEXT 3.8.6
Title:
ELECTRICAL TEXT 3.8.7
Title:
ELECTRICAL TEXT 3.8.8
Title:
ELECTRICAL TEXT 3.9.1
Title:
REFUELING C TEXT 3.9.2
Title:
REFUELING C TEXT 3.9.3
Title:
REFUELING C TEXT 3.9.4
Title:
REFUELING C TEXT 3.9.5
Title:
REFUELING C POWER SYSI POWER SYSI POWER SYSI POWER SYS' POWER SYSI POWER SYST
)PERATIONS
)PERATIONS
)PERATIONS OPERATIONS
)PERATIONS 0
11/1.5/2002
'EMS DIESEL FUEL OIL, LUBE OIL, AND STARTING AIR 0
11/15/2002
'EMS DC SOURCES -
OPERATING 0
11/1.5/2002
'EMS DC SOURCES -
SHUTDOWN 0
11/15/2002
'EMS BATTERY CELL PARAMETERS 1
10/05/2005
'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 Pae f8RprtDt:0/60 Page 7 of 8
Report Date: 01/16/06
SSES MANUAL Manual Name: TSB1 Manual
Title:
TECHNICAL SPECIFICATION BASES UNIT 1 MANUAL TEXT 3.9.6 0
11/15/2002
Title:
REFUELING OPERATIONS REACTOR PRESSURE VESSEL (RPV)
TEXT 3.9.7 0
11/15/2002
Title:
REFUELING OPERATIONS RESIDUAL HEAT REMOVAL (RHR) -
TEXT 3.9.8 0
11/15/2002
Title:
REFUELING OPERATIONS RESIDUAL HEAT REMOVAL (RHR) -
WATER LEVEL HIGH WATER LEVEL LOW WATER LEVEL TEXT 3.10.1
Title:
SPECIAL TEXT 3.10.2
Title:
SPECIAL TEXT 3.10.3
Title:
SPECIAL TEXT 3.1.0.4
Title:
SPECIAL TEXT 3.10.5
Title:
SPECIAL TEXT 3.10.6
Title:
SPECIAL TEXT 3.10.7
Title:
SPECIAL TEXT 3.10.8
Title:
SPECIAL OPERATIONS OPERATIONS OPERATIONS OPERATIONS OPERATIONS OPERATIONS OPERATIONS OPERATIONS 0
11/15/2002 INSERVICE LEAK AND HYDROSTATIC TESTING OPERATION 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 0
11/15/2002 SHUTDOWN MARGIN (SDM) TEST -
REFUELING Page8 of 8 Report Date: 01/16/06 Page 8 of 8 Report Date: 01/16/06
SUSQUEHANNA STEAM ELECTRIC STATION LIST OF EFFECTIVE SECTIONS (TECHNICAL SPECIFICATIONS BASES)
Section Title Revision TOC Table of Contents 7
B 2.0 SAFETY LIMITS BASES Page B 2.0-1 0
Page TS / B 2.0-2 2
Page TS / B 2.0-3 3
Pages TS/ B 2.0-4 and TS / B 2.0-5 2
Page TS/B2.0-6 1
Pages B 2.0-7 through B 2.0-9 0
B 3.0 LCO AND SR APPLICABILITY 13ASES Pages B 3.0-1 through B 3.0-4 0
Pages TS / B 3.0-5 through TS / B 3.0-7 i1 Pages TS / B 3.0-8 through TS / B 3.0-9 2
Pages TS I B 3.0-10 through 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 TSi B 3.0-17 0
B 3.1 REACTIVITY CONTROL BASES Pages B 3.1-1 through B'3.1-5 0
Pages TS / B 3.1-6 and TS / B 3.1-7 1
Pages B 3.1-8 through B 3.1-13 0
Page TS / B 3.1-14 1
Pages B 3.1-15 through i 3.-22 0
Page TS / B 3.1 1 i i 1
Pages B 3.1-24 throuih;B 3.1-27 0
Page TS / B 3.1-28 1
PageTS/B3.1-29 I
Pages B 3.1-0 through B 3.1-33 0
Pages TSI B 3.3-34 through TS I B 3.3-36 1
PageTS [B,311-7,,
2 Page TS I B 1-38 1
1 Page'sl 3.1-39 through B 3.1-44 0
Page'rTI B 3.1-45 1
Paes B 3.1-46 and B 3.1-47 0
PaD TS / B 3,1-48 andTSI/ B 3.1-49 1
Page B 3.1-50 I
0 Page TS B 3.1-51 1
B 3.2 POWER DISTRIBUTION LIMITS BASES Page TS / B 3.2-1, 1
Page TS B 3.2-2 2
Page TS B 3.2-3 1
Page TS B 3.2-4 I
2 Pages TS / B 3.2-5 and TS I B 3.2-6 1
Page B 3.2-7 0
Pages TS / B 3.2-8 through TS / B 3.2-10 1
SUEAN
. UNI 1 TS.~ B _OS-Reiso 68 SUSQUEHANNA - UNIT 1 TS I B LOES-1 Revision 68
SUSQUEHANNA STEAM ELECTRIC STATION LIST OF EFFECTIVE SECTIONS (TECHNICAL SPECIFICATIONS BASES)
Section Title Revision Page TS / B 3.2-11 2
Page B 3.2-12 0
Page TS / B 3.2-13 2
Pages B 3.2-14 and B 3.2-15 0
Page TS / B 3.2-16 2
Pages B 3.2-17 and B 3.2-18 0
Page TS / B 3.2-19 3
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/B3.3-6 1
Pages TS / B 3.3-7 through TS / B 3.3-11 2
Page TS/B 3.3-12 2
Page TS / B 3.3-13 1
Page TS / B 3.3-14 2
Pages TS/B 3.3-15 and TS /83.3-16 1
Pages TS/B 3.3-17 and TS/ B3.3-18 2
Pages TS / B 3.3-19 through TS / B 3.3-27 1
Pages TS / B 3.3-28 through TS / B 3.3-31 2
Pages TS / B 3.3-32 and TS / B 3.3-33 4
Pages TS / B 3.3-34 through TS / B 3.3-43 1
Pages TS / B 3.3-43a through TS / B 3.3-43i 0
Pages TS / B 3.3-44 through TS / B 3.3-50 2
Pages TS / B 3.3-51 through TS / B 3.3-53 1
Page TS/B 3.3-54 2
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
Pages TS / B 3.3-69 and TS / B 3.3-70 3
Page TS / B 3.3-71 3
Pages TS 1/3.3-72 through TS 3.3-75 2
Page TS / B 3.3-75a 4
Pages TS / B 3.3-75b and TS / B 3.3-75c 4
Pages B 3.3-76 through 3.3-77 0
Page TS / B 3.3-78 1
Pages B 3.3-79 through B 3.3-89 0
Page TS / B 3.3-90 1
Page B 3.3-91 0
Page TS / B 3.3-92 through TS / B 3.3-100 1
Pages B 3.3-101 through B 3.3-103 0
Page TS /B 3.3-104 1
Pages B 3.3-105 and B 3.3-106 0
Page TS / B 3.3-107 1
Page B 3.3-108 0
SUSQUEHANNA - UNIT I IS / B LOES-2 Revision 68 SUSQUEHANNA - UNIT 1 TS I B LOES-2 Revision 68
SUSQUEHANNA STEAM ELECTRIC STATION LIST OF EFFECTIVE SECTIONS (TECHNICAL SPECIFICATIONS BASES)
Section Title Revision Page TS / B 3.3-109 1
Pages B 3.3-110 and B 3.3-111 0
Pages TS / B 3.3-112 and TS; / B 3.3-112a 1
Pages TS / B 3.3-113 and TS / B 3.3-114 1
Page TS / B 3.3-115 1
Page TS / B 3.3-116 2
Page TS / B 3.3-117 1
Pages B 3.3-118 through B 3.3-122 0
Pages TS / B 3.3-123 and TS / B 3.3-124 1
Page TS / B 3.3-124a 0
Page B 3.3-125 0
Page TS / B 3.3-126 1
Page TS / B 3.3-127 1
Pages B 3.3-128 through B 3.3-130 0
Page TS / B 3.3-131 1
Pages B 3.3-132 through B 3.3-137 0
Page TS / B 3.3-138 1
Pages B 3.3-139 through B 3.3-149 0
Page TS / B 3.3-150 through TS / B 3.3-162 1
Page TS /B3.3-163 2
Pages TS / B 3.3-164 through TS / B 3.3-177 1
Pages TS / B 3.3-178 and TS / B 3.3-179 2
Page TS / B 3.3-179a 2
Page TS / B 3.3-179b 0
Page TS / B 3.3-179c 0
Page TS /B3.3-180 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 3.3-204 0
Page TS / B 3.3-205 1
Pages B 3.3-206 through B 3,.3-219 0
B 3.4 REACTOR COOLANT SYSTEMI 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
Page TS /B3.4-15 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 and B 3.4-31 0
Page TS / B 3.4-32 1
SUSQUEHANNA - UNIT I TS/B LOES-3 Revision 68 SUSQUEHANNA - UNIT 1 TS / B LOES-3 Revision 68
SUSQUEHANNA STEAM ELECTRIC STATION UST OF EFFECTIVE SECTIONS (TECHNICAL SPECIFICATIONS BASES)
Section Title Revision 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 2
Page TS / B 3.4-50 1
Page TS / B 3.4-51 2
Pages TS / B 3.4-52 and TS / B 3.4-53 1
Page TS / B 3.4-54 2
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
Page TS/B3.5-4 1
Page TS/B3.5-5 2
Page TS/B3.5-6 1
Pages B 3.5-7 through B 3.5-10 0
Page TS / B 3.5-11 1
Page TS / B 3.5-12 0
Page TS / B 3.5-13 1
Page TS / B 3.5-14 and TS / B 3.5-15 0
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 TS / B 3.5-26 and TS / B 3.5-27 1
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B 3.6 CONTAINMENT SYSTEMS BASES Page TS/B3.6-1 2
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SUSQUEHANNA STEAM ELECTRIC STATION LIST OF EFFECTIVE SECTIONS (TECHNICAL SPECIFICATIONS BASES)
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TSB1 text LOES 1/4/06
.UNIT1 BRe
.vision 68 SUSQUEHANNA - UNIT 1 TS I B LOES-7 Revision 68
PPL Rev. 1 S/RVs B 3.4.3 B 3.4 REACTOR COOLANT SYSTEM (RCS)
B 3.4.3 Safety/Relief Valves (S/R Vs)
BASES BACKGROUND The ASME Boiler and Pressure Vessel Code requires the reactor pressure vessel be protected from overpressure during upset conditions by self-actuated safety valves. As part of the nuclear pressure relief system, the size and number of S/RVs are selected such that peak pressure in the nuclear system will not exceed the ASME Code limits for the reactor coolant pressure boundary (RCPB).
The S/RVs are located on the main steam lines between the reactor vessel and the first isolation valve within the drywell. There are a total of 16 S/RVs of which any 12 are required to be OPERABLE. The S/RVs can actuate by either of two modes: the safety mode or the relief mode. In the safety mode (or spring mode of operation), the valve opens when steam pressure at the valve inlet overcomes the spring force holding the valve closed. This satisfies the Code requirement.
Each S/RV discharges steam through a discharge line to a point below the minimum water level in the suppression pool. Six S/RVs also serve as the Automatic Depressurization System (ADS) valves. The ADS requirements are specified in LCO 3.5.1, "ECCS-Operating."
APPLICABLE SAFETY ANALYSES The overpressure protection system must accommodate the most severe pressurization transient. Evaluations have determined that the most severe transient is the closure of all main steam isolation valves (MSIVs),
followed by reactor scram on high neutron flux (i.e., failure of the direct scram associated with MSIV position) (Ref. 1). For the purpose of the analyses, 12 of the 16 S/RVs are assumed to operate in the safety mode.
The analysis results demonstrate that the design S/RV capacity is capable of maintaining reactor pressure below the ASME Code limit of 110% of vessel design pressure (110% x 1250 psig = 1375 psig). This LCO helps to ensure that the acceptance limit of 1375 psig is met during the Design Basis Event.
(continued)
SUSQUEHANNA-UNIT 1 TS / B3.4-15 Revision 1
PPL Rev. I S/RVs B 3.4.3 BASES APPLICABLE From an overpressure standpoint, the design basis events are bounded SAFETY by the MSIV closure with flux scram event described above. Reference 2 ANALYSES discusses additional events that are expected to actuate the S/RVs.
(continued)
S/RVs satisfy Criterion 3 of the NRC Policy Statement (Ref. 4).
LCO The safety function of 12 of the 16 S/RVs are required to be OPERABLE to satisfy the assumptions of the safety analysis (Refs. 1 and 2). The requirements of this LCO are applicable only to the capability of the S/RVs to mechanically open to relieve excess pressure when the lift setpoint is exceeded (safety function).
The S/RV setpoints are established to ensure that the ASME Code limit on peak reactor pressure is satisfied. The ASME Code specifications require the lowest safety valve setpoint to be at or below vessel design pressure (1250 psig) and the highest safety valve to be set so that the total accumulated pressure does not exceed 110% of the design pressure for overpressurization conditions. The transient evaluations in the FSAR are based on these! setpoints, but also include the additional uncertainty of +/- 3% of the nominal setpoint to provide an added degree of conservatism.
Operation with fewer valves OPERABLE than specified, or with setpoints outside the ASME limits, could result in a more severe reactor response to a transient than predicted, possibly resulting in the ASME Code limit on reactor pressure being exceeded.
APPLICABILITY In MODES 1, 2, and 3, all required S/RVs must be OPERABLE, since considerable energy may be in the reactor core and the limiting design basis transients are assumed to occur in these MODES. The S/RVs may be required to provide pressure relief to discharge energy from the core until such time that the Residual Heat Removal (RHR) System is capable of dissipating the core heat.
In MODE 4 reactor pressure is low enough that the overpressure limit is unlikely to be approached by assumed (continued)
SUSQUEHANNA - UNIT 1
'rs / B 3.4-16 Revision 2
PPL Rev. 1 S/RVs B 3.4.3 BASES APPLICABILITY operational transients or accidents. In MODE 5, the reactor vessel head is (continued) unbolted or removed and the reactor is at atmospheric pressure. The S/RV function is not needed during these conditions.
ACTIONS A.1 and A.2 With less than the minimum number of required S/RVs OPERABLE, a transient may result in the violation of the ASME Code limit on reactor pressure. If the safety function of one or more required S/RVs is inoperable, the plant: must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to MODE 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and to MODE 4 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The allowed Completion Times are reasonable, based on operating experience, to reach required plant conditions from full power conditions in an orderly manner and without challenging plant systems.
SURVEILLANCE SR 3.4.3.1 REQUIREMENT The Surveillance requires that the required S/RVs will open at the pressures assumed in the safety analysis of Reference 1. The demonstration of the S/RV safe lift settings must be performed during shutdown, since this is a bench test, to be done in accordance with the Inservice Testing Program. The lift setting pressure shall correspond to ambient conditions of the valves at nominal operating temperatures and pressures. The S/RV setpoint is +/-3% of the nominal setpoint for OPERABILITY. Requirements for accelerated testing are established in accordance with the Inservice Test Program. Any of the 16 SIRVs, identified in this Surveillance Requirement, with their associated setpoints, can be designated as the 12 required S/RVs. This maintains the assumptions in the overpressure analysis.
A Note is provided to allow up to two of the required 12 S/RVs to be physically replaced with S/RVs with lower setpoints until the next refueling outage. This provides operational flexibility which maintains the assumptions in the over-pressure analysis.
(continued)
SUSQUEHANNA - UNIT 1 TS / B 3.4-17 Revision 2
PPL Rev. 1 SIRVs B 3.4.3 BASES SURVEILLANCE SR 3.4.3.1 (continued)
REQUIREMENTS The Frequency of this Surveillance is established in accordance with the Inservice Testing Program.
REFERENCES
- 1. FSAR, Section 5.2.2.1.4.
- 2. FSAR, Section 15.
- 3. ASME Operation and Maintenance Code.
- 4. Final Policy Statement on Technical Specifications Improvements, July 22, 1993 (58 FR 39132).
SUSQUEHANNA - UNIT 1 T~S / B 3.4-1 8 Revision 2
PPL Rev. 1 RCS PIV Leakage B 3.4.5 B 3.4 B 3.4.5 REACTOR COOLANT SYSTEM (RCS)
RCS Pressure Isolation Valve (PIV) Leakage BASES BACKGROUND The function of RCS PIVs is to separate the high pressure RCS from an attached low pressure system. This protects the RCS pressure boundary described in 10 CFR 50.2, 10 CFR 50.55a(c), and GDC 55 of 10 CFR 50, Appendix A (Refs. 1, 2, and 3). RCS PlVs are defined as any two normally closed valves in series within the reactor coolant pressure boundary (RCPB). F'lVs are designed to meet the requirements of Reference 4. During their lives, these valves can produce varying amounts of reactor coolant leakage through either normal operational wear or mechanical deterioration.
The RCS PIV LCO allows RCS high pressure operation when leakage through these valves exists in amounts that do not compromise safety.
The PIV leakage limit applies to each individual valve. Leakage through these valves is not included in any allowable LEAKAGE specified in LCO 3.4.4, "RCS Operational LEAKAGE."
Although this specification provides a limit on allowable PIV leakage rate, its main purpose is to prevent overpressure failure of the low pressure portions of connecting systems. The leakage limit is an indication that the PIVs between the RCOS and the connecting systems are degraded or degrading. PIV leakage could lead to overpressure of the low pressure piping or components. Failure consequences could be a loss of coolant accident (LOCA) outside of containment, an unanalyzed event that could degrade the ability for low pressure injection.
A study (Ref. 5) evaluated various PIV configurations to determine the probability of intersystem LOCAs. This study concluded that periodic leakage testing of the PIVs can substantially reduce intersystem LOCA probability.
(continued)
SUSQUEHANNA - UNIT 1 B 3.4-24 Revision 0
PPL Rev. 1 RCS PIV Leakage B 3.4.5 BASES BACKGROUND PIVs are provided to isolate the RCS from the following typically (continued) connected systems:
- a. Residual Heat Removal (RHR) System; and
- b. Core Spray System.
The PIVs are listed in Table B 3.4.5-1 "Pressure Isolation Valve".
APPLICABLE Reference 5 evaluated various PIV configurations, leakage testing of the SAFETY valves, and operational changes to determine the effect on the probability ANALYSES of intersystem LOCAs. This study concluded that periodic leakage testing of the PIVs can substantially reduce the probability of an intersystem LOCA.
PIV leakage is not considered in any Design Basis Accident analyses.
This Specification provides for monitoring the condition of the RCPB to detect PIV degradation that has the potential to cause a LOCA outside of containment. RCS PIV leakage satisfies Criterion 2 of the NRC Policy Statement (Ref. 6).
LCO RCS PIV leakage is leakage into closed systems connected to the RCS.
Isolation valve leakage is usually on the order of drops per minute.
Leakage that increases significantly suggests that something is operationally wrong and corrective action must be taken. Violation of this LCO could result in continued degradation of a PIV, which could lead to overpressurization of a low pressure system and the loss of the integrity of a fission product barrier.
The LCO PIV leakage limit is 0.5 gpm per nominal inch of valve size with a maximum limit of 5 gpm (Ref. 4).
APPLICABILITY In MODES 1, 2, and 3, this LCO applies because the PIV leakage potential is greatest when the RCS is pressurized. In MODE 3, valves in the RHR shutdown cooling flow path are not required to meet the requirements of this LCO when in, (continued)
SUSQUEHANNA - UNIT 1 B 3.4-25 Revision 0
PPL Rev. 1 RCS PIV Leakage B 3.4.5 BASESi APPLICABILITY or during transition to or from, the RHR shutdown cooling mode of (continued) operation. This is because RHR shutdown cooling will be placed in operation only below the current pressure permissive setpoint when the high to low pressure interface does not exist.
In MODES 4 and 5, leakage limits are not provided because the lower reactor coolant pressure results in a reduced potential for leakage and for a LOCA outside the containment. Accordingly, the potential for the consequences of reactor coolant leakage is far lower during these MODES.
ACTIONS The ACTIONS are modified by two Notes. Note 1 has been provided to modify the ACTIONS related to RCS PIV flow paths. Section 1.3, Completion Times, specifies 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 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 the Condition of RCS PIV leakage limits exceeded provide appropriate compensatory measures for separate affected RCS PIV flow paths. As such, a Note has been provided that allows separate Condition entry for each affected RCS PIV flow path. Note 2 requires an evaluation of affected systems if a PIV is inoperable. The leakage may have affected system OPERABILITY, or isolation of a leaking flow path with an alternate valve may have degraded the ability of the interconnected system to perform its safety function. As a result, the applicable Conditions and Required Actions for systems made inoperable by PIVs must be entered. This ensures appropriate remedial actions are taken, if necessary, for the affected systems.
A. 1 If leakage from one or more RCS PIVs is not within limit, the flow path must be isolated by at least one closed manual, deactivated automatic, or check valve within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
(continued)
SUSQUEHANNA - UNIT 1 B 3.4-26 Revision 0
.8 3.4.5 BASES ACTIONS A.
(continued)
Required Action A. 1 is modified by a Note stating that the valves used for isolation must meet the same leakage requirements as the PIVs and must be on the RCPB or the high pressure portion of the system.
Four hours provides time to reduce leakage in excess of the allowable limit and to isolate the flow path if leakage cannot be reduced while corrective actions to reseat the leaking PIVs are taken. The 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> allows time for these actions and restricts the time of operation with leaking valves.
B.1 and B.2 If leakage cannot be reduced or the system isolated, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to MODE 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and MODE 4 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. This action may reduce the leakage and also reduces the potential for a LOCA outside the containment. The Completion Times are reasonable, based on operating experience, to achieve the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.
SURVEILLANCE SR 3.4.51 REQUIREMENTS S
345 Performance of leakage testing on each RCS PIV is required to verify that leakage is below the specified limit and to identify each leaking valve.
The leakage limit of 0.5 gpm per inch of nominal valve diameter up to 5 gpm maximum applies to each valve. Leakage testing requires a stable pressure condition. For the two PIVs in series, the leakage requirement applies to each valve individually and not to the combined leakage across both valves. If the PIVs are not individually leakage tested, one valve may have failed completely and not be detected if the other valve in series meets the leakage requirement. In this situation; the protection provided by redundant valves would be lost.
(continued)
SUSQUEHANNA - UNIT 1 B 3.4-27 Revision 0
PPL Rev. 1 RCS PIV Leakage B 3.4.5 BASES SURVEILLANCE S
cniud REQUIREMENTS S
3.45.
(c)
The 24 month Frequency required by the Inservice Testing Program is within the ASME OM Code Frequency requirement and is based on the need to perform this Surveillance during an outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power.
This SR is modified by a Note that states the leakage Surveillance is not required to be performed in MODE 3. Entry into MODE 3 is permitted for leakage testing at high differential pressures with stable conditions not possible in the lower MODES.
REFERENCES
- 1. 10 CFR 50.2.
- 2. 10 CFR 50.55a(c).
- 4. ASME Operation and Maintenance Code.l
- 5. NUREG-0677, May 1980.
- 6. Final Policy Statement on Technical Specifications Improvements, July 22, 1993 (58 FR 39132).
SUSQUEHANNA-UNIT 1 TS / B 3.4-28 Revision 1
PPL Rev. 1 RCS PIV Leakage B 3.4.5 BASES (continued)
TABLE B 3.4.5-1 REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES 1st Isolation 2nd Isolation Valve(s) Number(s)
Valve(s) Number(s)
Service HV-152F006A HV-1 52F005A Core Spray Injection HV-152F037A HV-152F006B HV-152F005B Core Spray Injection HV-152F037B HV-151 FO50A HV-151 FO15A LPCI Injection HV-1511F122A HV-151 F050B HV-151F015B LPCI Injection HV-151 F122B HV-151 F022 HV-151 F023 Head Spray HV-151F009 HV-151 F008 Shutdown Cooling I
I SUSQUEHANNA - UNIT 1 TS / B 3.4-29 Revision 1
PPL Rev. 2 ECCS-Operating
.B3.5.1 B 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) AND REACTOR CORE ISOLATION COOLING (RCIC) SYSTEM B 3.5.1 ECCS-Operating BASES BACKGROUND The ECCS is designed, in conjunction with the primary and secondary containment, to limit the release of radioactive materials to the environment following a loss of coolant accident (LOCA). The ECCS uses two independent methods (flooding and spraying) to cool the core during a LOCA. The ECCS network consists of the High Pressure Coolant Injection (HPCI) System, the Core Spray (CS) System, the low pressure coolant injection (LFPCI) mode of the Residual Heat Removal (RHR)
System, and the Automatic Depressurization System (ADS). The suppression pool provides the required source of water for the ECCS.
Although no credit is taken in the safety analyses for the condensate storage tank (CST), it is capable of providing a source of water for the HPCI and CS systems.
On receipt of an initiation signal, ECCS pumps automatically start; simultaneously, the system aligns and the pumps inject water, taken either from the CST or suppression pool, into the Reactor Coolant System (RCS) as RCS pressure is overcome by the discharge pressure of the ECCS pumps. Although the system is initiated, ADS action is delayed, allowing the operator to interrupt the timed sequence if the system is not needed.
The HPCI pump discharge pressure quickly exceeds that of the RCS, and the pump injects coolant into the vessel to cool the core. If the break is small, the HPCI System will maintain coolant inventory as well as vessel level while the RCS is still pressurized. If HPCI fails, it is backed up by ADS in combination with LPCI and CS. In this event absent operator action, the ADS timed sequence would time out and open the selected safety/relief valves (S/RVs) depressurizing the RCS, thus allowing the LPCI and CS to overcome RCS pressure and inject coolant into the vessel. If the break is large, RCS pressure initially drops rapidly and the LPCI and CS cool the core.
Water from the break returns to the suppression pool where it is used again and again. Water in the suppression pool is circulated through a heat exchanger cooled by the RHR Service Water System. Depending on the location and size of (continued)
SUSQUEHANNA-UNIT 1 B 3.5-1 Revision 0
PPL Rev. 2 ECCS-Operating B 3.5.1 BASES BACKGROUND the break, portions of the ECCS may be ineffective; however the overall (continued) design is effective in cooling the core regardless of the size or location of the piping break. Although no credit is taken in the safety analysis for the RCIC System, it performs a similar function as HPCI, but has reduced makeup capability. Nevertheless, it will maintain inventory and cool the core while the RCS is still pressurized following a reactor pressure vessel (RPV) isolation.
All ECCS subsystems are designed to ensure that no single active component failure will prevent automatic initiation and successful operation of the minimum required ECCS equipment.
The CS System (Ref. 1) is composed of two independent subsystems.
Each subsystem consists of two motor driven pumps, a spray sparger above the core, and piping and valves to transfer water from the suppression pool to the sparger. The CS System is designed to provide cooling to the reactor core when reactor pressure is low. Upon receipt of an initiation signal, the CS pumps in both subsystems are automatically started when AC power is available. When the RPV pressure drops sufficiently, CS Syslem flow to the RPV begins. A full flow test line is provided to route water from and to the suppression pool to allow testing of the CS System without spraying water in the RPV.
LPCI is an independent operating mode of the RHR System. There are two LPCI subsystems (Ref. 2), each consisting of two motor driven pumps and piping and valves to transfer water from the suppression pool to the RPV via the corresponding recirculation loop. The two LPCI subsystems can be interconnected via the RHR System cross tie valves; however, at least one of the two cross tie valves is maintained closed with its power removed to prevent loss of both LPCI subsystems during a LOCA. The LPCI subsystems are designed to provide core cooling at low RPV pressure. Upon receipt of an initiation signal, all four LPCI pumps are automatically started. RHR System valves in the LPCI flow path are automatically positioned to ensure the proper flow path for water from the suppression pool to inject into the recirculation loops. When the RPV pressure drops sufficiently, the LPCI flow to the RPV, via the corresponding recirculation loop, begins. The water then enters the reactor through the jet pumps.
(continued)
SUSQUEHANNA-UNIT 1 B 3.5-2 Revision 0
PPL Rev. 2 ECCS-Operating B 3.5.1 BASES BACKGROUND Full flow test lines are provided for each LPCI subsystem to route water (continued) from the suppression pool, to allow testing of the LPCI pumps without injecting water into the RPV. These test lines also provide suppression pool cooling capability, as described in LCO 3.6.2.3, "RHR Suppression Pool Cooling."
The HPCI System (Ref. 3) consists of a steam driven turbine pump unit, piping, and valves to provide steam to the turbine, as well as piping and valves to transfer water from the suction source to the core via the feedwater system line, where the coolant is distributed within the RPV through the feedwalter sparger. Suction piping for the system is provided from the CST and the suppression pool. Pump suction for HPCI is normally aligned to the CST source to minimize injection of suppression pool water into the RPV. Whenever the CST water supply is low, an automatic transfer to the suppression pool water source ensures a water supply for continuous operation of the HPCI System. The steam supply to the HPCI turbine is piped from a main steam line upstream of the associated inboard main steam isolation valve.
The HPCI System is designed to provide core cooling for a wide range of reactor pressures (165 psia to 1225 psia). Upon receipt of an initiation signal, the HPCI turbine stop valve and turbine control valve open and the turbine accelerates to a specified speed. As the HPCI flow increases, the turbine control valve is automatically adjusted to maintain design flow.
Exhaust steam from the HPCI turbine is discharged to the suppression pool. A full flow test line is provided to route water to the CST to allow testing of the HPCI System during normal operation without injecting water into the RPV, The ECCS pumps are provided with minimum flow bypass lines, which discharge to the suppression pool. The valves in these lines automatically open to prevent pump damage due to overheating when other discharge line valves are closed. To ensure rapid delivery of water to the RPV and to minimize water hammer effects, all ECCS pump discharge lines are filled with water. The HPCI, LPCI and CS System discharge lines are kept full of water using a "keep fill" system that is supplied using the condensate transfer system.
(continued)
SUSQUEHANNA - UNIT 1 TS / B 3.5-3 Revision 2
PPL Rev. 2 ECCS-Operating B 3.5.1 BASES BACKGROUND (continued)
The ADS (Ref. 4) consists of 6 of the 16 S/RVs. It is designed to provide depressurization of the RCS during a small break LOCA if HPCI fails or is unable to maintain required water level in the RPV. ADS operation reduces the RPV pressure to within the operating pressure range of the low pressure ECC'S subsystems (CS and LPCI), so that these subsystems can provide coolant inventory makeup. Each of the S/RVs used for automatic depressurization is equipped with two gas accumulators and associated inlet check valves. The accumulators provide the pneumatic power to actuate the valves.
APPLICABLE SAFETY ANALYSES The ECCS performance is evaluated for the entire spectrum of break sizes for a postulated LOCA. The accidents for which ECCS operation is required are presented in References 5, 6, and 7. The required analyses and assumptions are defined in Reference 8. The results of these analyses are also described in Reference 9.
This LCO helps to ensure that the following acceptance criteria for the ECCS, established by 10 CFR 50.46 (Ref. 10), will be met following a LOCA, assuming the worst case single active component failure in the ECCS:
- a. Maximum fuel element cladding temperature is
- 22000F;
- b. Maximum cladding oxidation is < 0.17 times the total cladding thickness before oxidation;
< 0.01 times the hypothetical amount that would be generated if all of the metal in the cladding surrounding the fuel, excluding the cladding surrounding the plenum volume, were to react;
- d. The core is maintained in a coolable geometry; and
- e. Adequate long term cooling capability is maintained.
(continued)
I SUSQUEHANNA - UNIT 1 TS / B 3.5-4 Revision 1
PPL Rev. 2 ECCS-Operating B 3.5.1 BASES APPLICABLE SPC performed LOCA calculations for the SPC ATRIUM'-10 fuel design.
SAFETY The limiting single failures for the SPC analyses are discussed in ANALYSES Reference 11. For ca large break LOCA, the SPC analyses identify the (continued) recirculation loop suction piping as the limiting break location. The SPC analysis identifies the failure of the LPCI injection valve into the intact recirculation loop as the most limiting single failure.
For a small break LOCA, the SPC analyses identify the recirculation loop discharge piping as the limiting break location, and a battery failure as the most severe single failure. One ADS valve failure is analyzed as a limiting single failure for events requiring ADS operation. The remaining OPERABLE ECCS subsystems provide the capability to adequately cool the core and prevent excessive fuel damage.
The ECCS satisfy Criterion 3 of the NRC Policy Statement (Ref. 15).
LCO Each ECCS injection/spray subsystem and six ADS valves are required to be OPERABLE. The ECCS injection/spray subsystems are defined as the two CS subsystems, the two LPCI subsystems, and one HPCI System.
The low pressure EC"CS injection/spray subsystems are defined as the two CS subsystems and the two LPCI subsystems.
With less than the required number of ECCS subsystems OPERABLE, the potential exists that during a limiting design basis LOCA concurrent with the worst case single failure, the limits specified in Reference 10 could be exceeded. All ECCS subsystems must therefore be OPERABLE to satisfy the single failure criterion required by Reference 10.
LPCI subsystems may be considered OPERABLE during alignment and operation for decay heat removal when below the actual RHR cut in permissive pressure in MODE 3, if capable of being manually realigned (remote or local) to the LPCI mode and not otherwise inoperable. At these low pressures and decay heat levels, a reduced complement of ECCS subsystems should provide the required core cooling, thereby allowing operation of RHR shutdown cooling when necessary.
(continued)
SUSQUEHANNA - UNIT 1 TS / B 3.5-5 Revision 2
PPL Rev. 2 ECCS-Operating B 3.5.1 BASES (continued)
APPLICABILITY All ECCS subsystems are required to be OPERABLE during MODES 1, 2, and 3, when there is considerable energy in the reactor core and core cooling would be required to prevent fuel damage in the event of a break in the primary system piping. In MODES 2 and 3, when reactor steam dome pressure is c 150 psig, ADS and HPCI are not required to be OPERABLE because the low pressure ECCS subsystems can provide sufficient flow below this pressure. ECCS requirements for MODES 4 and 5 are specified in LCO 3.5.2, "ECCS-Shutdown."
ACTIONS A Note prohibits the application of LCO 3.0.4.b to an inoperable HPCI subsystem. There is an increased risk associated with entering a MODE or other specified condition in the Applicability with an inoperable HPCI subsystem and the provisions of LCO 3.0.4.b, which allow entry into a MODE or other specified condition in the Applicability with the LCO not met after performance of a risk assessment addressing inoperable systems and components, should not be applied in this circumstance.
A.1 If any one low pressure ECCS injection/spray subsystem is inoperable for reasons other than Condition B, the inoperable subsystem must be restored to OPERABLE status within 7 days. In this Condition, the remaining OPERABLE subsystems provide adequate core cooling during a LOCA. However, overall ECCS reliability is reduced, because a single failure in one of the remaining OPERABLE subsystems, concurrent with a LOCA, may result in the ECCS not being able to perform its intended safety function. The 7 day Completion Time is based on a reliability study (Ref. 12) that evaluated the impact on ECCS availability, assuming various components and subsystems were taken out of service. The results were used to calculate the average availability of ECCS equipment needed to mitigate the consequences of a LOCA as a function of allowed outage times (i.e., Completion Times).
B. 1 If one LPCI pump in one or both LPCI subsystems is inoperable, the inoperable LPCI pumps must be restored to OPERABLE status within 7 days. In this Condition, the remaining OPERABLE LPCI pumps and at least one CS subsystem (continued)
SUSQUEHANNA - UNIT 1 TS / B 3.5-6 Revision 1
PPL Rev. 2 ECCS-Operating B 3.5.1 BASES ACTIONS B. 1 (continued) provide adequate core cooling during a LOCA. However, overall ECCS reliability is reduced, because a single failure in one of the remaining OPERABLE subsystems, concurrent with a LOCA, may result in the ECCS not being able to perform its intended safety function. A 7 day Completion Time is based on a reliability study cited in Reference 12 and has been found to be acceptable through operating experience.
C.1 and C.2 If the inoperable low pressure ECCS subsystem or LPCI pump(s) cannot be restored to OPERABLE status within the associated Completion Time, the plant must be brought to a MODE in which the LCO does 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 /> and to MODE 4 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.
D.1 and D.2 If the HPCI System is inoperable and the RCIC System is verified to be OPERABLE, the HPFCI System must be restored to OPERABLE status within 14 days. In this Condition, adequate core cooling is ensured by the OPERABILITY of the redundant and diverse low pressure ECCS injection/spray subsystems in conjunction with ADS. Also, the RCIC System will automatically provide makeup water at most reactor operating pressures. Verification of RCIC OPERABILITY is therefore required when HPCI is inoperable. This may be performed as an administrative check by examining logs or other information to determine if RCIC is out of service for maintenance or other reasons. It does not mean to perform the Surveillances needed to demonstrate the OPERABILITY of the RCIC System. If the OPERABILITY of the RCIC System cannot be verified, however, Condition H must be immediately entered. If a single active component fails concurrent with a design basis LOCA, there is a potential, depending on the specific failure, that the minimum required ECCS equipment (continued)
SUSQUEHANNA-UNIT 1/R TS / B 3.5-7 Revision 0
PPL Rev. 2 ECCS-Operating B 3.5.1 BASES ACTIONS D.1 and D.2 (continued) will not be available. A 14 day Completion Time is based on a reliability study cited in Reference 12 and has been found to be acceptable through operating experience.
E.1 and E.2 If Condition A or Condition B exists in addition to an inoperable HPCI System, the inoperable low pressure ECCS injection/spray subsystem or the LPCI pump(s) or the HPCI System must be restored to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. In this Condition, adequate core cooling is ensured by the OPERABILITY of the ADS and the remaining low pressure ECCS subsystems. However, the overall ECCS reliability is significantly reduced because a single failure in one of the remaining OPERABLE subsystems concurrent with a design basis LOCA may result in the ECCS not being able to perform its intended safety function. Since both a high pressure system (HPCI) and a low pressure subsystem are inoperable, a more restrictive Completion Time of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> is required to restore either the HPCI System or the low pressure ECCS injection/spray subsystem to OPERABLE status. This Completion Time is-based on a reliability study cited in Reference 12 and has been found to be acceptable through operating experience.
F. 1 The LCO requires six ADS valves to be OPERABLE in order to provide the ADS function. Reference 11 contains the results of an analysis that evaluated the effect of one ADS valve being out of service. Per this analysis, operation of only five ADS valves will provide the required depressurization. However, overall reliability of the ADS is reduced, because a single failure in the OPERABLE ADS valves could result in a reduction in depressurization capability. Therefore, operation is only allowed for a limited time. The 14 day Completion Time is based on a reliability study cited in Reference 12 and has been found to be acceptable through operating experience.
(continued)
SUSQUEHANNA - UNIT 1 TS / B3.5-8 Revision 0
PPL Rev. 2 ECCS-Operating B 3.5.1 BASES ACTIONS G.1 and G.2 (continued)
If Condition A or Condition B exists in addition to one inoperable ADS valve, adequate core cooling is ensured by the OPERABILITY of HPCI and the remaining low pressure ECCS injection/spray subsystem.
However, overall ECCS reliability is reduced because a single active component failure concurrent with a design basis LOCA could result in the minimum required ECCS equipment not being available. Since both a high pressure system (ADS) and a low pressure subsystem are inoperable, a more restrictive Completion Time of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> is required to restore either the low pressure ECCS subsystem or the ADS valve to OPERABLE status. This Completion Time is based on a reliability study cited in Reference 112 and has been found to be acceptable through operating experience.
H.1 and H.2 If any Required Action and associated Completion Time of Condition D, E, F, or G is not met, or if two or more ADS valves are inoperable, the plant must be brought to a condition in which the LCO does 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 /> and reactor steam dome pressure reduced to < 150 psig within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.
When multiple ECCS subsystems are inoperable, as stated in Condition I, LCO 3.0.3 must be entered immediately.
SURVEILLANCE SR 3.5.1.1 REQUIREMENTS The flow path piping has the potential to develop voids and pockets of entrained air. Maintaining the pump discharge lines of the HPCI System, CS System, and LPCI subsystems (continued)
SUSQUEHANNA - UNIT 1 TS / B 3.5-9 Revision 0
PPL Rev. 2 ECCS-Operating B 3.5.1 BASES SURVEILLANCE SR 3.5.1.1 (continued)
REQUIREMENTS full of water ensures that the ECCS will perform properly, injecting its full capacity into the RC'S upon demand. This will also prevent a water hammer following an ECCS initiation signal. One acceptable method of ensuring that the lines are full is to vent at the high points. The 31 day Frequency is based on the gradual nature of void buildup in the ECCS piping, the procedural controls governing system operation, and operating experience.
SR 3.5.1.2 Verifying the correct alignment for manual, power operated, and automatic valves in the ECCS flow paths provides assurance that the proper flow paths will exist for ECCS operation. This SR does not apply to valves that are locked, sealed, or otherwise secured in position since these were verified to be in the correct position prior to locking, sealing, or securing.
A valve that receives an initiation signal is allowed to be in a nonaccident position provided the valve will automatically reposition in the proper stroke time. This SR does not require any testing or valve manipulation; rather, it involves verification that those valves capable of potentially being mispositioned are in the correct position. This SR does not apply to valves that cannot be inadvertently misaligned, such as check valves. For the HPCI System, this SR also includes the steam flow path for the turbine and the flow controller position.
The 31 day Frequency of this SR was derived from the Inservice Testing Program requirements for performing valve testing at least once every 92 days. The Frequency of 31 days is further justified because the valves are operated under procedural control and because improper valve position would only affect a single subsystem. This Frequency has been shown to be acceptable through operating experience.
This SR is modified by a Note that allows LPCI subsystems to be considered OPERABLE during alignment and operation for decay heat removal with reactor steam dome pressure less than the RHR cut in permissive pressure in MODE 3, if capable of being manually realigned (remote or local) to the (continued)
SUSQUEHANNA - UNIT 1 TS / B 3.5-1 0 Revision 0
PPL Rev. 2 ECCS-Operating B 3.5.1 BASES SURVEILLANCE SR 3.5.1.2 (continued)
REQUIREMENTS LPCI mode and not otherwise inoperable. This allows operation in the RHR shutdown cooling mode during MODE 3, if necessary.
SR 3.5.1.3 Verification every 31 days that ADS gas supply header pressure is 2 135 psig ensures adequate gas pressure for reliable ADS operation.
The accumulator on each ADS valve provides pneumatic pressure for valve actuation. The design pneumatic supply pressure requirements for the accumulator are! such that, following a failure of the pneumatic supply to the accumulator, at least one valve actuations can occur with the drywell at 70% of design pressure.
The ECCS safety analysis assumes only one actuation to achieve the depressurization required for operation of the low pressure ECCS. This minimum required pressure of > 135 psig is provided by the containment instrument gas system. The 31 day Frequency takes into consideration administrative controls over operation of the gas system and alarms associated with the containment instrument gas system.
SR 3.5.1.4 Verification every 31 days that at least one RHR System cross tie valve is closed and power to its operator is disconnected ensures that each LPCI subsystem remains independent and a failure of the flow path in one subsystem will not affect the flow path of the other LPCI subsystem.
Acceptable methods of removing power to the operator include opening the breaker, or racking out the breaker, or removing the breaker. If both RHR System cross tie valves are open or power has not been removed from at least one closed valve operator, both LPCI subsystems must be considered inoperable. The 31 day Frequency has been found acceptable, considering that these valves are under strict administrative controls that will ensure the valves continue to remain closed with motive power removed.
(continued)
SUSQUEHANNA - UNIT 1 TS / B 3.5-11 Revision 1
PPL Rev. 2 ECCS-Operating
.B 3.5.1 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.5.1.5 Verification every 31 days that each 480 volt AC swing bus transfers automatically from the normal source to the alternate source on loss of power while supplying its respective bus demonstrates that electrical power is available to ensure proper operation of the associated LPCI inboard injection and minimum flow valves and the recirculation pump discharge and bypass valves. Therefore, each 480 volt AC swing bus must be OPERABLE for the associated LPCI subsystem to be OPERABLE. The test is performed by actuating the load test switch or by disconnecting the preferred power source to the transfer switch and verifying that swing bus automatic transfer is accomplished. The 31 day Frequency has been found to be acceptable through operating experience.
SR 3.5.1.6 Cycling the recirculation pump discharge and bypass valves through one complete cycle of full travel demonstrates that the valves are mechanically OPERABLE and provides assurance that the valves will close when required to ensure the proper LPCI flow path is established. Upon initiation of an automatic LPCI subsystem injection signal, these valves are required to be closed to ensure full LPCI subsystem flow injection in the reactor via the recirculation jet pumps. De-energizing the valve in the closed position will also ensure the proper flow path for the LPCI subsystem. Acceptable methods of de-energizing the valve include opening the breaker, or racking out the breaker, or removing the breaker.
The specified Frequency is once during reactor startup before THERMAL POWER is > 25%
0RTP. However, this SR is modified by a Note that states the Surveillaince is only required to be performed if the last performance was rrjore than 31 days ago. Therefore, implementation of this Note requires this test to be performed during reactor startup before exceeding 25% RTP. Verification during reactor startup prior to reaching
> 25% RTP is an exception to the normal Inservice Testing Program generic valve cycling Frequency of 92 days, but is considered acceptable due to (continued)
SUSQUEHANNA - UNIT 1 TS / B 3.5-12 Revision 0
PPL Rev. 2 ECCS-Operating B 3.5.1 BASES SURVEILLANCE SR 3.5.1.6 (continued)
REQUIREMENTS the demonstrated reliability of these valves. If the valve is inoperable and in the open position, the associated LPCI subsystem must be declared inoperable.
SR 3.5.1.7. SR 3.5.1.8. and SR 3.5.1.9 The performance requirements of the low pressure ECCS pumps are determined through application of the 10 CFR 50, Appendix K criteria (Ref. 8). This periodic Surveillance is performed (in accordance with the ASME OM Code requirements for the ECCS pumps) to verify that the ECCS pumps will develop the flow rates required by the respective analyses. The low pressure ECCS pump flow rates ensure that adequate core cooling is provided to satisfy the acceptance criteria of Reference 10.
The pump flow rates are verified against a system head equivalent to the RPV pressure expected during a LOCA. The total system pump outlet pressure is adequate to overcome the elevation head pressure between the pump suction and the vessel discharge, the piping friction losses, and RPV pressure present during a LOCA. These values may be established during preoperational testing.
The flow tests for the HPCI System are performed at two different pressure ranges such that system capability to provide rated flow is tested at both the higher and lower operating ranges of the system. Additionally, adequate steam flowv must be passing through the main turbine or turbine bypass valves to continue to control reactor pressure when the HPCI System diverts steam flow. Reactor steam pressure is considered adequate when Ž 920 psig to perform SR 3.5.1.8 and Ž 150 psig to perform SR 3.5.1.9. However, the requirements of SR 3.5.1.9 are met by a successful performance at any pressure < 165 psig. Adequate steam flow is represented by at least 1.25 turbine bypass valves open.
Therefore, sufficient time is allowed after adequate pressure and flow are achieved to perform these tests. Reactor startup is allowed prior to performing the low pressure Surveillance test because the reactor pressure is low and the time allowed to satisfactorily perform the Surveillance test is short. The reactor pressure is allowed to be increased to normal operating pressure since it is assumed that the low pressure test has been satisfactorily (continued)
SUSQUEHANNA - UNIT 1 TS / B 3.5-13 Revision 1
PPL Rev. 2 ECCS-Operating B 3.5.1 BASES SURVEILLANCE SR 3.5.1.7, SR 3.5.1.8, and SR 3.5.1.9 (continued)
REQUIREMENTS completed and there is no indication or reason to believe that HPCI is inoperable.
Therefore, SR 3.5.1.8 and SR 3.5.1.9 are modified by Notes that state the Surveillances are not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after the reactor steam pressure and flow are adequate to perform the test.
The Frequency for SR 3.5.1.7 and SR 3.5.1.8 is in accordance with the Inservice Testing Plrogram requirements. The 24 month Frequency for SR 3.5.1.9 is based on the need to perform the Surveillance under the conditions that apply just prior to or during a startup from a plant outage.
Operating experience has shown that these components usually pass the SR when performed at the 24 month Frequency, which is based on the refueling cycle. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.
SR 3.5.1.10 The ECCS subsystems are required to actuate automatically to perform their design functions. This Surveillance verifies that, with a required system initiation signal (actual or simulated), the automatic initiation logic of HPCI, CS, and LPCI will cause the systems or subsystems to operate as designed, including actuation of the system throughout its emergency operating sequence, automatic pump startup and actuation of all automatic valves to their required positions. This functional test includes the LPCI and CS interlocks between Unit 1 and Unit 2 and specifically requires the following:
A functional test of the interlocks associated with the LPCI and CS pump starts in response to an automatic initiation signal in Unit 1 followed by a false automatic initiation signal in Unit 2; A functional test of the interlocks associated with the LPCI and CS pump starts in response to an automatic initiation signal in Unit 2 followed by a false automatic initiation signal in Unit 1; and (continued)
SUSQUEHANNA - UNIT 1 TS / B 3.5-14 Revision 0
PPL Rev. 2 ECCS-Operating
.B 3.5.1 BASES SURVEILLANCE SR 3.5.1.10 (continued)
REQUIREMENTS A functional test of the interlocks associated with the LPCI and CS pump starts in response to simultaneous occurrences of an automatic initiation signal in both Unit 1 and Unit 2 and a loss of Offsite power condition affecting both Unit 1 and Unit 2.
The purpose of this functional test (preferred pump logic) is to assure that if a false LOCA signal were to be received on one Unit simultaneously with an actual LOCA signal on the second Unit, the preferred LPCI and CS pumps are started and the non-preferred LPCI and CS pumps are tripped for each Unit. This functional test is performed by verifying that the non-preferred LPCI and CS pumps are tripped. The verification that preferred LPCI and CS pumps start is performed under a separate surveillance test. Only one division of LPCI preferred pump logic is required to be OPERABLE for each Unit, because no additional failures needs to be postulated with a false LOCA signal. If the preferred or non-preferred pump logic for CS is inoperable, the associated CS pumps shall be declared inoperable and the pumps should not be operated to ensure that the opposite Unit's CS pumps or 4.16 kV ESS Buses are protected.
This SR also ensures that the HPCI System will automatically restart on an RPV low water level (Level 2) signal received subsequent to an RPV high water level (Level 8) trip and that the suction is automatically transferred from the CST to the suppression pool. The LOGIC SYSTEM FUNCTIONAL TEST performed in LCO 3.3.5.1 overlaps this Surveillance.
This SR can be accomplished by any series of sequential overlapping or total steps such that the entire channel is tested.
The 24 month Frequency is acceptable because operating experience has shown that these components usually pass the SR when performed at the 24 month Frequency, which is based on the refueling cycle. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.
This SR is modified by a Note that excludes vessel injection/spray during the Surveillance. Since all active components are testable and full flow can be demonstrated by recirculation through the test line, coolant injection into the RPV is not required during the Surveillance.
(continued)
SUSQUEHANNA - UNIT 1 TS / B3.5-15 Revision 0
PPL Rev. 2 ECCS-Operating B 3.5.1 BASES SURVEILLANCE SR 3.5.1.11 REQUIREMENTS (continued)
The ADS designated S/RVs are required to actuate automatically upon receipt of specific initiation signals. A system functional test is performed to demonstrate that the mechanical portions of the ADS function (i.e.,
solenoids) operate as designed when initiated either by an actual or simulated initiation signal, causing proper actuation of all the required components. SR 3.5.1.12 and the LOGIC SYSTEM FUNCTIONAL TEST performed in LCO '3.3.5.1 overlap this Surveillance to provide complete testing of the assumed safety function.
The 24 month Frequency is based on the need to perform portions of the 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 SR when performed at the 24 month Frequency, which is based on the refueling cycle. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.
This SR is modified by a Note that excludes valve actuation. This prevents an RPV pressure blowdown.
SR 3.5.1.12 A rmanual actuation of each ADS valve is performed to verify that the valve and solenoid are functioning properly. This is demonstrated by one of the two methods described below. Proper operation of the valve tailpipes is ensured through the use of foreign material exclusion during maintenance.
One method is by manual actuation of the ADS valve under hot conditions. Proper functioning of the valve and solenoid is demonstrated by the response of the turbine control or bypass valve or by a change in the measured flow or by any other method suitable to verify steam flow.
Adequate reactor steam dome pressure must be available to perform this test to avoid damaging the valve due to seat impact during closure. Also, adequate steam flow must be passing through the main turbine or turbine bypass valves to continue to control reactor pressure when the ADS valves divert steam flow upon opening. Sufficient time is therefore allowed after the required pressure and flow are achieved to perform this SR. Adequate pressure at which this SR is to be performed is 150 psig.
Howeyer, the requirements of SR 3.5.1.12 are met by a successful performance at any pressure. Adequate steam flow is represented by at least 1.25 turbine bypass valves open. Reactor startup is allowed prior to performing this SR by this method because valve OPERABILITY and the setpoints for (continued)
SUSQUEHANNA - UNIT 1 TS / B 3.5-16 Revision 1
PPL Rev. 2 ECCS-Operating B 3.5.1 BASES SURVEILLANCE SR 3.5.1.12 (continued)
REQUIREMENTS overpressure protection are verified, per ASME requirements, prior to valve installation. Therefore, this SR is modified by a Note that states the Surveillance is not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reactor steam pressure and flow are adequate to perform the test. The 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> allowed for manual actuation after the required pressure is reached is sufficient to achieve stable conditions and provides adequate time to complete the Surveillance.
Another method is by manual actuation of the ADS valve at atmospheric temperature and pressure during cold shutdown. When using this method, proper functioning of the valve and solenoid is demonstrated by visual observation of actuator movement. Actual disc travel is measured during valve refurbishment and testing per ASME requirements. Lifting the valve at atmospheric pressure is the preferred method because lifting the valves with steam flow increases the likelihood that the valve will leak.
The Note that modifies this SR is not needed when this method is used because the SR is performed during cold shutdown.
SR 3.5.1.11 and the LOGIC SYSTEM FUNCTIONAL TEST performed in LCO 3.3.5.1 overlap, this Surveillance to provide complete testing of the assumed safety function. The Frequency of 24 months on a STAGGERED TEST BASIS ensures that both solenoids for each ADS valve are alternately tested. The Frequency is based on the need to perform the Surveillance under the conditions that apply just prior to or during a startup from a plant outage. Operating experience has shown that these components usually pass the SR when performed at the 24 month Frequency, which is based on the refueling cycle. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.
SR 3.5.1.13 This SR ensures that the ECCS RESPONSE TIME for each ECCS injection/spray subsystem is less than or equal to the maximum value assumed in the accident analysis. Response Time testing acceptance criteria are included in Reference 13. This SR is modified by a Note that allows the instrumentation portion of the response time to be assumed to be based on historical response time data and therefore, is excluded from the ECCS RESPONSE TIME testing. This is allowed since the instrumentation response time is a small part of the ECCS RESPONSE TIME (e.g., sufficient margin exists in the diesel generator start time when compared to the instrumentation response time) (Ref. 14).
(continued)
SUSQUEHANNA - UNIT 1 TS / B 3.5-17 Revision 1
PPL Rev. 2 ECCS-Operating B 3.5.1 BASES SURVEILLANCE REQUIREMENTS SR 3.5.1.13 (continued)
The 24-month Frequency is consistent with the typical industry refueling cycle and is acceptable based upon plant operating experience.
REFERENCES
- 1.
FSAR, Section 6.3.2.2.3.
- 2.
FSAR, Section 6.3.2.2.4.
- 3.
FSAR, Section 6.3.2.2.1.
- 4.
FSAR, Section 6.3.2.2.2.
- 5.
FSAR, Section 15.2.4.
- 6.
FSAR, Section 15.2.5.
- 7.
FSAR, Section 15.2.6.
- 8.
- 9.
FSAR, Section 6.3.3.
- 10.
- 11.
FSAR, Section 6.3.3.
- 12. Memorandum from R.L. Baer (NRC) to V. Stello, Jr. (NRC),
"Recommended Interim Revisions to LCOs for ECCS Components,"
December 1, 1975.
- 13. FSAR, Section 6.3.3.3.
- 14. NEDO 32291-A, "System Analysis for the Elimination of Selected Response Time Testing Requirements, October 1995.
- 15.
Final Policy Statement on Technical Specifications Improvements, July 22, 1993 (58 FR 39132).
SUSQUEHANNA-UNIT 1 TS / B 3.5-18 Revision 1
PPL Rev. 1 RHR Suppression Pool Cooling B 3.6.2.3 B 3.6 CONTAINMENT SYSTEMS B 3.6.2.3 Residual Heat Removal (RHR) Suppression Pool Cooling BASES BACKGROUND Following a Design Basis Accident (DBA), the RHR Suppression Pool Cooling System removes heat from the suppression pool. The suppression pool is designed to absorb the sudden input of heat from the primary system. In the long term, the pool continues to absorb residual heat generated by fuel in the reactor core. Some means must be provided to remove heat from the suppression pool so that the temperature inside the primary containment remains within design limits.
This function is provided by two redundant RHR suppression pool cooling subsystems. The purpose of this LCO is to ensure that both subsystems are OPERABLE in applicable MODES.
Each RHR subsystem contains either one of the two RHR pumps and a flow path capable of recirculating water from the suppression chamber through an RHR heat exchanger and is manually initiated and independently controlled. The two subsystems perform the suppression pool cooling function by circulating water from the suppression pool through the RHR heat exchangers and returning it to the suppression pool. RHR service water, circulating through the tube side of the heat exchangers, exchanges heat with the suppression pool water and discharges this heat to the external heat sink.
The heat removal capability of one RHR pump in one subsystem is sufficient to meet the overall DBA pool cooling requirement for loss of coolant accidents (LOCAs) and transient events such as a turbine trip or stuck open safety/relief valve (SIRV). S/RV leakage and High Pressure Coolant Injection and Reactor Core Isolation Cooling System testing increase suppression pool temperature more slowly. The RHR Suppression Pool Cooling System is also used to lower the suppression pool water bulk temperature following such events.
APPLICABLE Reference 1 contains the results of analyses used to predict primary SAFETY containment pressure and temperature following large and small break ANALYSES LOCAs. The intent of the analyses is to demonstrate that the heat removal capacity of the RHR (continued)
SUSQUEHANNA - UNIT 1 B 3.6-62 Revision 0
PPL Rev. 1 RHR Suppression Pool Cooling B 3.6.2.3 BASES APPLICABLE Suppression Pool Cooling System is adequate to maintain the primary SAFETY containment conditions within design limits. The suppression pool ANALYSES temperature is calculated to remain below the design limit.
(continued)
The RHR Suppression Pool Cooling System satisfies Criterion 3 of the NRC Policy Statement. (Ref. 3)
LCO During a DBA, a minimum of one RHR suppression pool cooling subsystem is required to maintain the primary containment peak pressure and temperature below design limits (Ref. 1). To ensure that these requirements are met, two RHR suppression pool cooling subsystems must be OPERABLE. Therefore, in the event of an accident, at least one subsystem is OPERABLE assuming the worst case single active failure.
An RHR suppression pool cooling subsystem is OPERABLE when one of the pumps, the heat exchanger, and associated piping, valves, instrumentation, and controls are OPERABLE.
APPLICABILITY In MODES 1, 2, and 3, a DBA could cause a release of radioactive material to primary containment and cause a.heatup and pressurization of primary containment. In MODES 4 and 5, the probability and consequences of these events are reduced due to the pressure and temperature limitations in these MODES. Therefore, the RHR Suppression Pool Cooling System is not required to be OPERABLE in MODE 4 or 5.
ACTIONS A.1 With one RHR suppression pool cooling subsystem inoperable, the inoperable subsystem must be restored to OPERABLE status within 7 days. In this Condition, the remaining RHR suppression pool cooling subsystem is adequate to perform the primary containment cooling function. However, the overall reliability is reduced because a single failure in the OPERABLE subsystem could result in loss of primary containment cooling capability. The 7 day Completion Time is acceptable in light of the redundant RHR suppression (continued)
SUSQUEHANNA - UNIT 1 B 3.6-63 Revision 0
PPL Rev. 1 RHR Suppression Pool Cooling B 3.6.2.3 BASES ACTIONS A._
(continued) pool cooling capabilities afforded by the OPERABLE subsystem and the low probability of a DBA occurring during this period.
B. 1 With two RHR suppression pool cooling subsystems inoperable, one subsystem must be restored to OPERABLE status within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. In this condition, there is a substantial loss the of primary containment pressure and temperature mitigation function. The 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> Completion Time is based on this loss of function and is considered acceptable due to the low probability of a DBA and the potential avoidance of a plant shutdown transient that could result in the need for the RHR suppression pool cooling subsystems to operate.
C.1 and C.2 If the Required Action and associated Completion Time cannot be met the plant must be brought to a MODE in which the LCO does 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 /> and to MODE 4 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.
SURVEILLANCE REQUIREMENTS SR 3.6.2.3.1 Verifying the correct alignment for manual, power operated, and automatic valves in the RHR suppression pool cooling mode flow path provides assurance that the proper flow path exists for system operation. This SR does not apply to valves that are locked, sealed, or otherwise secured in position since these valves were verified to be in the correct position prior to locking, sealing, or securing. A valve is also allowed to be in the nonaccident position provided it can be aligned to the accident position within the time assumed in the accident analysis. This is acceptable since the RHR suppression pool cooling mode is manually initiated. This SR does not require any testing or valve manipulation; rather, it involves verification that those valves capable of being mispositioned are in the correct position. This SR does not apply to valves that cannot be inadvertently misaligned, such as check valves.
(continued)
SUSQUEHANNA - UNIT 1 TS / B 3.6-64 Revision 1
PPL Rev. 1 RHR Suppression Pool Cooling B 3.6.2.3 BASES SURVEILLANCE S
cniud REQUIREMENTS SR 3.6.2.3.1 (continued)
The Frequency of 31 days is justified because the valves are operated under procedural control, improper valve position would affect only a single subsystem, the probability of an event requiring initiation of the system is low, and the subsystem is a manually initiated system. This Frequency has been shown to be acceptable based on operating experience.
SR 3.6.2.3.2 Verifying that each RHR pump develops a flow rate > 9750 gpm while operating in the suppression pool cooling mode with flow through the associated heat exchanger ensures that pump performance has not degraded during the cycle. Flow is a normal test of centrifugal pump performance required by ASME OM Code (Ref. 2). This test confirms one point on the pump design curve, and the results are indicative of overall performance. Such inservice inspections confirm component OPERABILITY, trend performance, and detect incipient failures by indicating abnormal performance. The Frequency of this SR is in accordance with the Inservice Testing Program.
REFERENCES
- 1. FSAR, Section 6.2.
- 2. ASME Operation and Maintenance Code.
- 3. Final Policy Statement on Technical Specifications Improvements, July 22, 1993 (58 FR 39132).
I SUSQUEHANNA - UNIT 1 T S / B 3.6-65 Revision 1
Jan. 17, 2006 Page 1
of 3
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4, SSES MANUAL Manual Name: TSB2 a1 Manual
Title:
TECHNICAL SPECIFICATIONS BASES UNIT 2 MANUAL Table Of Contents Issue Date:
01/16/2006 Procedure Name Rev TEXT LOES 67
Title:
LIST OF EFFECTIVE SECTIONS Issue Date 01/1.6/2006 Change ID Change Number TEXT TOC
Title:
TABLE OF CONTENTS 7
04/18/2005 TEXT 2.1.1 1
Title:
SAFETY LIMITS (SLS) REACTOR 10/27/2004 CORE SLS TEXT 2.1.2 0
Title:
SAFETY LIMITS (SLS) REACTOR 11/18/2002 COOLANT SYSTEM (RCS) PRESSURE SL TEXT 3.0
Title:
LIMITING CONDITION 1
04/18/2005 I
FOR OPERATION (LiCO) 0APPLICABILITY TEXT 3.1.1
Title:
REACTIVITY TEXT 3.1.2
Title:
REACTIVITY 1
O 03/24/2005 CONTROL SYSTEMS SHUTDOWN MARGIN (SDM) 7
- 0 :1 11/18/2002 CONTROL SYSTEMS REACTIVITY ANOMALIES TEXT 3.1.3
Title:
REACTIVITY TEXT 3.1.4
Title:
REACTIVITY TEXT 3.1.5
Title:
REACTIVITY TEXT 3.1.6
Title:
REACTIVITY ll:
- ~ ;)
1 CONTROL SYSTEMS 2
CONTROL SYSTEMS 1
CONTROL SYSTEMS 2
CONTROL SYSTEMS 07/06/2005 CONTROL ROD OPERABILITY 07/06/2005 CONTROL ROD SCRAM TIMES 07/06/2005 CONTROL ROD SCRAM ACCUMULATORS 03/24/2005 ROD PATTERN CONTROL Page 1 of 8 Report Date: 01/16/06 Page 1 of 8 Report Date: 01/16/06
SSES MANUAL Manual Name: TSB2 Manual
Title:
TECHNICAL SPECIFICATIONS BASES UNIT 2 MANUAL TEXT 3.1.7 1
08/30/2005
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 2
10/05/2005
Title:
POWER DISTRIBUTION LIMITS AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)
TEXT 3.2.2 1
03/24/2005
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 3.2.4 1
07/06/2005
Title:
POWER DISTRIBUTION LIMITS AVERAGE POWER RANGE MONITOR (APRM) GAIN AND SETPOINTS <
TEXT 3.3.1.1 2
07/06/2005
Title:
INSTRUMENTATION REACTOR PROTECTION SYSTEM (RPS) INSTRUMENTATION TEXT 3.3.1.2 0
11/18/2002
Title:
INSTRUMENTATION SOURCE RANGE MONITOR (SRM) INSTRUMENTATION TEXT 3.3.1.3
Title:
OPRM INSTRUMENTATION 0
11/22/2004 TEXT 3.3.2.1 1
02/17/2005
Title:
INSTRUMENTATION CONTROL ROD BLOCK INSTRUMENTATION TEXT 3.3.2.2 0
11/18/2002
Title:
INSTRUMENTATION FEEDWATER -
MAIN TURBINE HIGH WATER LEVEL TRIP INSTRUMENTATION TEXT 3.3.3.1 2
07/06/2005
Title:
INSTRUMENTATION POST ACCIDENT MONITORING (PAM) INSTRUMENTATION Page2 of 8 Report Date: 01/16/06 Page 2 of 8 Report Date: 01/16/06
SSES MANUALI 1,
Manual Name:
TSB2 Manual
Title:
TECHNICAL SPECIFICATIONS BASES UNIT 2 MANUAL 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 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 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 0
11/18/2002 REACTOR CORE ISOLATION COOLING (RCIC) SYSTEM INSTRUMENTATION 1
11/09/2004 PRIMARY CONTAINMENT ISOLATION INSTRUMENTATION 1
11/09/2004 SECONDARY CONTAINMENT ISOLATION INSTRUMENTATION 0
11/18/2002 CONTROL ROOM EMERGENCY OUTSIDE AIR SUPPLY (CREOAS)
SYSTEM TEXT 3.3.8.1 1
09/02/2004
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 2
11/22/2004
Title:
REACTOR COOLANT SYSTEM (RCS) RECIRCULATION LOOPS OPERATING TEXT 3.4.2 0
11/18/2002
Title:
JET PUMPS Page3 of 8
Report Date: 01/16/06 Page 3 of 8 Report Date: 01/16/06
SSES MANUAL Manual Name:
TSB2 Manual
Title:
TECHNICAL SPECIFICATIONS BASES UNIT 2 MANUAL TEXT 3.4.3
Title:
REACTOR COOLANT TEXT 3.4.4
Title:
REACTOR COOLANT TEXT 3.4.5
Title:
REACTOR COOLANT TEXT 3.4.6
Title:
REACTOR COOLANT TEXT 3.4.7
Title:
REACTOR COOLANT TEXT 3.4.8
Title:
REACTOR COOLANT HOT SHUTDOWN TEXT 3.4.9
Title:
REACTOR COOLANT COLD SHUTDOWN 1
SYSTEM (RCS) 0 SYSTEM (RCS) 1 SYSTEM (RCS) 1 SYSTEM (RCS) 1 SYSTEM (RCS) 1 SYSTEM (RCS) 0 SYSTEM (RCS) 01/16/2006 SAFETY/RELIEF VALVES (S/RVS) 11/18/2002 RCS OPERATIONAL LEAKAGE 01/16/2006 RCS PRESSURE ISOLATION VALVE (PIV) LEAKAGE 04/18/2005 RCS LEAKAGE DETECTION INSTRUMENTATION 04/18/2005 RCS SPECIFIC ACTIVITY 04/18/2005 RESIDUAL HEAT REMOVAL (RHR) SHUTDOWN COOLING SYSTEMI.-
11/18/2002 RESIDUAL HEAT REMOVAL (RHR) SHUTDOWN COOLING SYSTEM TEXT 3.4.10
Title:
REACTOR COOLANT TEXT 3.4.11
Title:
REACTOR COOLANT TEXT 3.5.1 0
11/18/2002 SYSTEM (RCS) RCS PRESSURE AND TEMPERATURE (P/T) LIMITS 0
11/18/2002 SYSTEM (RCS) REACTOR STEAM DOME PRESSURE 3
01/16/2006
Title:
EMERGENCY CORE COOLING SYSTEMS (ECCS) AND REACTOR CORE ISOLATION COOLING SYSTEM ECCS -
OPERATING TEXT 3.5.2 0
11/18/2002
Title:
EMERGENCY CORE COOLING SYSTEMS (ECCS) AND REACTOR CORE ISOLATION COOLING SYSTEM ECCS -
SHUTDOWN TEXT 3.5.3 1
04/18/2005
Title:
EMERGENCY CORE COOLING SYSTEMS (ECCS) AND REACTOR CORE ISOLATION COOLING SYSTEM RCIC SYSTEM (RCIC)
(RCIC)
(RCIC)
Pa e
of 8Rr at : 0 / 6 0 Page 4 of 8
Report Date: 01/16/06
SSES MANUAL
, Manual Name:
TSB2 Manual
Title:
TECHNICAL SPECIFICATIONS BASES UNIT 2 MANUAL TEXT 3.6.1.1 0
11/18/2002
Title:
CONTAINMENT SYSTEMS PRIMARY CONTAINMENT TEXT 3.6.1.2
Title:
CONTAINMENT 0
11/1.8/2002 SYSTEMS PRIMARY CONTAINMENT AIR LOCK TEXT 3.6.1.3
Title:
CONTAINMENT 3
SYSTEMS PRIMARY 12/) 8/2005 CONTAINMENT ISOLATION VALVES (PCIVS)
TEXT 3.6.1.4
Title:
CONTAINMENT 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 0
11/1.8/2002 SYSTEMS CONTAINMENT PRESSURE 1
10/05/2005 SYSTEMS DRYWELL AIR TEMPERATURE 0
11/1.8/2002 SYSTEMS SUPPRESSION CHAMBER-TO-DRYWELL VACUUM BREAKERS 0
11/1.8/2002 SYSTEMS SUPPRESSION POOL AVERAGE TEMPERATURE 0
11/18/2002 SYSTEMS SUPPRESSION POOL WATER LEVEL 1
01/1.6/2006 SYSTEMS RESIDUAL HEAT REMOVAL (RHR) SUPPRESSION POOL COOLING 0
11/18/2002 SYSTEMS RESIDUAL HEAT REMOVAL (RHR) SUPPRESSION POOL SPRAY 1
04/18/2005 SYSTEMS PRIMARY CONTAINMENT HYDROGEN RECOMBINERS 1
04/18/2005 SYSTEMS DRYWELL AIR FLOW SYSTEM PageS of 8 Report Date: 01/16/06 Page 5 of 8 Report Date: 01/16/06
SSES MANUAL Manual Name:
TSB2 Manual
Title:
TECHNICAL SPECIFICATIONS BASES UNIT 2 MANUAL TEXT 3.6.3.3 0
11/18/2002
Title:
CONTAINMENT SYSTEMS PRIMARY CONTAINMENT OXYGEN CONCENTRATION TEXT 3.6.4.1 4
10/24/2005
Title:
CONTAINMENT SYSTEMS SECONDARY CONTAINMENT TEXT 3.6.4.2 2
01/03/2005
Title:
CONTAINMENT SYSTEMS SECONDARY CONTAINMENT ISOLATION VALVES (SCIVS)
TEXT 3.6.4.3 3
10/24/2005
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
Title:
PLANT TEXT 3.7.3
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 1
11/09/2004 SYSTEMS EMERGENCY SERVICE WATER (ESW)
SYSTEM 0
11/18/2002 SYSTEMS CONTROL ROOM EMERGENCY OUTSIDE AIR SUPPLY (CREOAS)
SYSTEM 0
11/18/2002 SYSTEMS CONTROL ROOM FLOOR COOLING SYSTEM 0
11/18/2002 SYSTEMS MAIN CONDENSER OFFGAS 1
01/17/2005 SYSTEMS MAIN TURBINE BYPASS SYSTEM 0
11/18/2002 SYSTEMS SPENT FUEL STORAGE POOL WATER LEVEL 3
10/05/2005
Title:
ELECTRICAL POWER SYSTEMS AC SOURCES -
OPERATING Page6 of 8
Report Date: 01/16/06 Page 6 of 8 Report Date: 01/16/06
SSES MANUAL
, Manual Name:
TSB2 Manual
Title:
TECHNICAL SPECIFICATIONS BASES UNIT 2 MANUAL TEXT 3.8.2
Title:
ELECTRICAL 0
POWER SYSTEMS 11/18/2002 AC SOURCES -
SHUTDOWN TEXT 3.8.3
Title:
ELECTRICAL 0
11/18/2002 POWER SYSTEMS DIESEL FUEL OIL, LUBE OIL, AND STARTING AIR TEXT 3.8.4
Title:
ELECTRICAL 0
POWER SYSTEMS 11/18/2002 DC SOURCES -
OPERATING TEXT 3.8.5
Title:
ELECTRICAL 0
11/18/2002 POWER SYSTEMS DC SOURCES -
SHUTDOWN TEXT 3.8.6
Title:
ELECTRICAL 0
POWER SYSTEMS 11/18/2002 BATTERY CELL PARAMETERS TEXT 3.8.7 i
-O
Title:
ELECTRICAL TEXT 3.8.8
Title:
ELECTRICAL 1
10/05/2005 POWER SYSTEMS DISTRIBUTION SYSTEMS -
OPERATING 0
11/18/2002 POWER SYSTEMS DISTRIBUTION SYSTEMS -
SHUTDOWN TEXT 3.9.1
Title:
REFUELING TEXT 3.9.2
Title:
REFUELING TEXT 3.9.3
Title:
REFUELING TEXT 3.9.4
Title:
REFUELING TEXT 3.9.5
Title:
REFUELING OPERATIONS OPERATIONS OPERATIONS OPERATIONS OPERATIONS 0
11/18/2002 REFUELING EQUIPMENT INTERLOCKS 0
11/18/2002 REFUEL POSITION ONE-ROD-OUT INTERLOCK 0
11/18/2002 CONTROL ROD POSITION 0
11/18/2002 CONTROL ROD POSITION INDICATION 0
11/18/2002 CONTROL ROD OPERABILITY -
REFUELING Page7 of 8
Report Date: 01/16/06 Page 7 of 8
Report Date: 01/16/06
SSES MANUAL Manual Name: TSB2 Manual
Title:
TECHNICAL SPECIFICATIONS BASES UNIT 2 MANUAL TEXT 3.9.7 0
11/18/2002
Title:
REFUELING OPERATIONS RESIDUAL HEAT REMOVAL (RHR) -
HIGH WATER LEVEL TEXT 3.9.8 0
11/18/2002
Title:
REFUELING OPERATIONS RESIDUAL HEAT REMOVAL (RHR) -
LOW WATER LEVEL TEXT 3.10.1
Title:
SPECIAL TEXT 3.10.2
Title:
SPECIAL TEXT 3.10.3
Title:
SPECIAL TEXT 3.10.4
Title:
SPECIAL TEXT 3.10.5
Title:
SPECIAL TEXT 3.10.6
Title:
SPECIAL TEXT 3.10.7
Title:
SPECIAL TEXT 3.10.8
Title:
SPECIAL OPERATIONS OPERATIONS OPERATIONS OPERATIONS OPERATIONS OPERATIONS OPERATIONS OPERATIONS 0
11/18/2002 INSERVICE LEAK AND HYDROSTATIC TESTING OPERATION 0
11/18/2002 REACTOR MODE SWITCH INTERLOCK TESTING 0
11/18/2002 SINGLE CONTROL ROD WITHDRAWAL HOT SHUTDOWN 0
11/18/2002 SINGLE CONTROL ROD WITHDRAWAL COLD SHUTDOWN 0
11/18/2002 SINGLE CONTROL ROD DRIVE (CRD)
REMOVAL REFUELING 0
11/18/2002 MULTIPLE CONTROL ROD WITHDRAWAL -
REFUELING 1
03/24/2005 CONTROL ROD TESTING -
OPERATING 1
03/24/2005 SHUTDOWN MARGIN (SDM)
TEST -
REFUELING Page8 of 8
Report Date: 01/16/06 Page 8 of 8
Report Date: 01/16/06
SUSQUEHANNA STEAM ELECTRIC STATION LIST OF EFFECTIVE SECTIONS (TECHNICAL SPECIFICATIONS BASES)
Section
-Title Revision TOC Table of Contents 7
B 2.0 SAFETY LIMITS BASES Page TS/B2.0-1 1
Page TS / B 2.0-2 2
Page TS / B 2.0-3 3
Page TS/B2.0-4 4
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 Pages B 3.0-1 through B 3.0-4 0
Pages TS / B 3.0-5 through rs / B 3.0-7 1
Pages TS / B 3.0-8 through Ts / B 3.0-9 2
Pages TS I B 3.0-10 through 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 3.0-17 0
B 3.1 REACTIVITY CONTROL BASES 4j Pages B 3.1-1 through B 3.1-4 0
Page TS / B 3.1-5 i
1 Pages TS / B 3.1-6 and TSI /B 3.1 -
7 2
Pages B 3.1-8 through 1-13 0
Page TS /B 3.1-14 1
Pages B 3.1-15 througI B 3.1-22 0
Page TS / B 3.1-2it3 1
Pages B 3.1-2 ug3.1-27 0
Page TS / 3.1-8 2
Page TS / f 29 1
Pages '3.'1-3`trough 8 3.1-33 0
Page ST, /
3.T1.34 through TS / B 3.1-36 1
Page S/ B-31-37 2
rage T /83.1-38 2
Pages 3.1-39 through B 3.1-44 0
PageS/ B 3.1-45 1
Pages 83.1-46 and B 3.1-47 0
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 Pages TS / B 3.2-1 and TS / B 3.2-2 1
Page TS /B3.2-3 3
Page TS/B3.2-4 1
Pages TS / B 3.2-5 and TS / B 3.2-6 3
Page TS /B3.2-7 2
Pages TS / B 3.2-8 and TS / B 3.2-9 3
SUSQEHANA -UNIT2 T / BLOESl Rvisin 6 SUSQUEHANNA - UNIT 2 TS I B LOES-1 Revision 67
SUSQUEHANNA STEAM ELECTRIC STATION LIST OF EFFECTIVE SECTIONS (TECHNICAL SPECIFICATIONS BASES)
Section Title Revision Pages TS / B 3.2-10 through TS I B 3.2-17 1
Page TS / 3.2-18 2
Page TS / 3.2-19 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
Page TS / B 3.3-7 2
Pages TS / B 3.3-8 through TS / B 3.3-11 2
Page TS / B 3.3-12 2
Page TS / B3.3-13 I
Page TS / B 3.3-14 2
Pages TS / B 3.3-15 and TS / B 3.3-16 1
Pages TS / B 3.3-17 and TS / B 3.3-18 2
Pages TS / B 3.3-19 through TS / B 3.3-27 1
Pages TS / B 3.3-28 through TS / B 3.3-30 2
Page TS /B3.3-31 1
Page TS /B3.3-32 3
Page Ts / B 3.3-33 2
Pages TS / B 3.3-34 through TS / B 3.3-43 1
Pages TS / B 3.3-43a though TS / B 3.3-43i 0
Pages TS / B 3.3-44 through TS / B 3.3-54 2
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
Pages TS / B 3.3-69 and TS / B 3.3-70 3
Pages TS / B 3.3-71 3
Pages TS / B 3.3-72 through TS / B 3.3-75 2
Page TS / B 3.3-75a 4
Pages TS / B 3.3-75b and TS / B 3.3-75c 4
Pages B 3.3-76 through TS / 13 3.3-77 0
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 /B3.3-107 2
Page TS / B3.3-108 1
Page TS / B 3.3-109 2
Pages TS / B 3.3-110 through TSI B 3.3-112 1
Page TS /B3.3-113 2
Page TS/B3.3-114 1
Page TS /B3.3-115 2
SUSQUEHANNA
- UNIT 2 TSIB LOES-2 Revision 67 SUSQUEHANNA - UNIT 2 TS IB LOES-2 Revision 67
SUSQUEHANNA STEAM ELECTRIC STATION LIST OF EFFECTIVE SECTIONS (TECHNICAL SPECIFICATIONS BASES)
Section Title Revision Pages TS / B 3.3-116 through TS / B 3.3-118 2
Pages TS / B 3.3-119 through TS / B 3.3-120 1
Pages TS / B 3.3-121 and TS; / B 3.3-122 2
Page TS / B 3.3-123 1
Page TS /B3.3-124 2
Page TS / B 3.3-124a 0
Page TS / B 3.3-125 1
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SUSQUEHANNA - UNIT 2 TS 113 LOES-3 Revision 67
SUSQUEHANNA STEAM ELECTRIC STATION LIST OF EFFECTIVE SECTIONS (TECHNICAL SPECIFICATIONS BASES)
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SUSQUEHANNA
- UNIT 2 TSiBLOES-4 Revision 67 SUSQUEHANNA - UNIT 2 TS If B LOES-4 Revision 67
SUSQUEHANNA STEAM ELECTRIC STATION LIST OF EFFECTIVE SECTIONS (TECHNICAL SPECIFICATIONS BASES)
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SUSQUEHANNA - UNIT 2 TS I B LOES-5 Revision 67
SUSQUEHANNA STEAM ELECTRIC STATION LIST OF EFFECTIVE SECTIONS (TECHNICAL SPECIFICATIONS BASES)
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TSB2 LOES 1/3/06 SUSQUEHANNA - UNIT 2 TSI B LOES-6 Revision 67
PPL Rev. 1 SIRVs B 3.4.3 B 3.4 REACTOR COOLANT SYSTEM (RCS)
B 3.4.3 Safety/Relief Valves (S/RVs)
BASES BACKGROUND The ASME Boiler and Pressure Vessel Code requires the reactor pressure vessel be protected from overpressure during upset conditions by self-actuated safety valves. As part of the nuclear pressure relief system, the size and number of S/RVs are selected such that peak pressure in the nuclear system will not exceed the ASME Code limits for the reactor coolant pressure boundary (RCPB).
The S/RVs are located on the main steam lines between the reactor vessel and the first isolation valve within the drywell. There are a total of 16 S/RVs of which any 12 are required to be OPERABLE. The S/RVs can actuate by either of two modes: the safety mode or the relief mode. In the safety mode (or spring mode of operation), the valve opens when steam pressure at the valve inlet overcomes the spring force holding the valve closed. This satisfies the Code requirement.
Each S/RV discharges steam through a discharge line to a point below the minimum water level in the suppression pool. Six S/RVs also serve as the Automatic Depressurization System (ADS) valves. The ADS requirements are specified in LCO 3.5.1, "ECCS-Operating."
APPLICABLE SAFETY ANALYSES The overpressure protection system must accommodate the most severe pressurization transient. Evaluations have determined that the most severe transient is the closure of all main steam isolation valves (MSIVs),
followed by reactor scram on high neutron flux (i.e., failure of the direct scram associated with MSIV position) (Ref. 1). For the purpose of the analyses, 12 of the 16 S/RVs are assumed to operate in the safety mode.
The analysis results demonstrate that the design S/RV capacity is capable of maintaining reactor pressure below the ASME Code limit of 110% of vessel design pressure (110% x 1250 psig = 1375 psig). This LCO helps to ensure that the acceptance limit of 1375 psig is met during the Design Basis Event.
(continued)
SUSQUEHANNA - UNIT 2 TS / B3.4-15 Revision 1
PPL Rev. 1 S/RVs B 3.4.3 BASES APPLICABLE From an overpressure standpoint, the design basis events are bounded SAFETY by the MSIV closure with flux scram event described above. Reference 2 ANALYSES discusses additional events that are expected to actuate the S/RVs.
(continued)
S/RVs satisfy Criterion 3 of the NRC Policy Statement (Ref. 4).
LCO The safety function of 12 of the 16 S/RVs are required to be OPERABLE to satisfy the assumptions of the safety analysis (Refs. 1 and 2). The requirements of this LCO are applicable only to the capability of the S/RVs to mechanically open to relieve excess pressure when the lift setpoint is exceeded (safety function).
The S/RV setpoints are established to ensure that the ASME Code limit on peak reactor pressure is satisfied. The ASME Code specifications require the lowest safety valve setpoint to be at or below vessel design pressure (1250 psig) and the highest safety valve to be set so that the total accumulated pressure does not exceed 110% of the design pressure for overpressurization conditions. The transient evaluations in the FSAR are based on these! setpoints, but also include the additional uncertainty of +/- 3% of the nominal setpoint to provide an added degree of conservatism.
Operation with fewer valves OPERABLE than specified, or with setpoints outside the ASME limits, could result in a more severe reactor response to a transient than predicted, possibly resulting in the ASME Code limit on reactor pressure being exceeded.
APPLICABILITY In MODES 1, 2, and 3, all required S/RVs must be OPERABLE, since considerable energy may be in the reactor core and the limiting design basis transients are assumed to occur in these MODES. The S/RVs may be required to provide pressure relief to discharge energy from the core until such time that the Residual Heat Removal (RHR) System is capable of dissipating the core heat.
In MODE 4 reactor pressure is low enough that the overpressure limit is unlikely to be approached by assumed (continued)
SUSQUEHANNA - UNIT 2 TS / B3.4-16 Revision 2
PPL Rev. 1 SIRVs B 3.4.3 BASES APPLICABILITY operational transients or accidents. In MODE 5, the reactor vessel head is (continued) unbolted or removed and the reactor is at atmospheric pressure. The S/RV function is not needed during these conditions.
ACTIONS A.1 and A.2 With less than the minimum number of required S/RVs OPERABLE, a transient may result in the violation of the ASME Code limit on reactor pressure. If the safety function of one or more required S/RVs is inoperable, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to MODE 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and to MODE 4 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The allowed Completion Times are reasonable, based on operating experience, to reach required plant conditions from full power conditions in an orderly manner and without challenging plant systems.
SURVEILLANCE SR 3.4.3.1 REQUIREMENT The Surveillance requires that the required S/RVs will open at the pressures assumed in the safety analysis of Reference 1. The demonstration of the S/RV safe lift settings must be performed during shutdown, since this is a bench test, to be done in accordance with the Inservice Testing Program. The lift setting pressure shall correspond to ambient conditions of the valves at nominal operating temperatures and pressures. The S/RV setpoint is +/-3% of the nominal setpoint for OPERABILITY. Requirements for accelerated testing are established in accordance with the Inservice Test Program. Any of the 16 S/RVs, identified in this Surveillance Requirement, with their associated setpoints, can be designated as the 12 required S/RVs. This maintains the assumptions in the overpressure analysis.
A Note is provided to allow up to two of the required 12 S/RVs to be physically replaced with S/RVs with lower setpoints until the next refueling outage. This provides operational flexibility which maintains the assumptions in the over-pressure analysis.
(continued)
SUSQUEHANNA - UNIT 2 TiS I B 3.4-17 Revision 2
PPL Rev. I S/RVs B 3.4.3 BASES SURVEILLANCE S
cniud REQUIREMENTS SR 3.4.3.1 (continued)
The Frequency of this Surveillance is established in accordance with the Inservice Testing Program.
REFERENCES
- 1. FSAR, Section 5.2.2.1.4.
- 2. FSAR, Section 15.
- 3. ASME Operation and Maintenance Code.
- 4. Final Policy Statement on Technical Specifications Improvements, July 22, 1993 (58 FR 39132).
I SUSQUEHANNA - UNIT 2 TS / B 3.4-18 Revision 2
PPL Rev. 1 RCS PIV Leakage B 3.4.5 B 3.4 REACTOR COOLANT SYSTEM (RCS)
B 3.4.5 RCS Pressure Isolation Valve (PIV) Leakage BASES BACKGROUND The function of RCS PIVs is to separate the high pressure RCS from an attached low pressure system. This protects the RCS pressure boundary described in 10 CFR 50.2, 10 CFR 50.55a(c), and GDC 55 of 10 CFR 50, Appendix A (Refs. 1, 2, and 3). RCS PlVs are defined as any two normally closed valves in series within the reactor coolant pressure boundary (RCPB). PIlVs are designed to meet the requirements of Reference 4. During their lives, these valves can produce varying amounts of reactor coolant leakage through either normal operational wear or mechanical deterioration.
The RCS PIV LCO allows RCS high pressure operation when leakage through these valves exists in amounts that do not compromise safety.
The PIV leakage limit applies to each individual valve. Leakage through these valves is not included in any allowable LEAKAGE specified in LCO 3.4.4, "RCS Operational LEAKAGE."
Although this specification provides a limit on allowable PIV leakage rate, its main purpose is to prevent overpressure failure of the low pressure portions of connecting systems. The leakage limit is an indication that the PIVs between the RCS and the connecting systems are degraded or degrading. PIV leakage could lead to overpressure of the low pressure piping or components. Failure consequences could be a loss of coolant accident (LOCA) outside of containment, an unanalyzed event that could degrade the ability for low pressure injection.
A study (Ref. 5) evaluated various PIV configurations to determine the probability of intersystem LOCAs. This study concluded that periodic leakage testing of the PIVs can substantially reduce intersystem LOCA probability.
(continued)
SUSQUEHANNA - UNIT 2 B 3.4-24 Revision 0
PPL Rev. 1 RCS PIV Leakage B 3.4.5 BASES BACKGROUND PIVs are provided to isolate the RCS from the following typically (continued) connected systems:
- a. Residual Heat Removal (RHR) System; and
- b. Core Spray System.
The PlVs are listed in Table B 3.4.5-1 "Pressure Isolation Valve".
APPLICABLE Reference 5 evaluated various PIV configurations, leakage testing of the SAFETY valves, and operational changes to determine the effect on the probability ANALYSES of intersystem LOCAs. This study concluded that periodic leakage testing of the PIVs can substantially reduce the probability of an intersystem LOCA.
PIV leakage is not considered in any Design Basis Accident analyses.
This Specification provides for monitoring the condition of the RCPB to detect PIV degradation that has the potential to cause a LOCA outside of containment. RCS 1PIV leakage satisfies Criterion 2 of the NRC Policy Statement (Ref. 6).
LCO RCS PIV leakage is leakage into closed systems connected to the RCS.
Isolation valve leakage is usually on the order of drops per minute.
Leakage that increases significantly suggests that something is operationally wrong and corrective action must be taken. Violation of this LCO could result in continued degradation of a PIV, which could lead to overpressurization of a low pressure system and the loss of the integrity of a fission product barrier.
The LCO PIV leakage limit is 0.5 gpm per nominal inch of valve size with a maximum limit of 5 gpm (Ref. 4).
APPLICABILITY In MODES 1, 2, and 3, this LCO applies because the PIV leakage potential is greatest when the RCS is pressurized. In MODE 3, valves in the RHR shutdown cooling flow path are not required to meet the requirements of this LCO when in, (continued)
SUSQUEHANNA - UNIT 2 B 3.4-25 Revision 0
PPL Rev. 1 RCS PIV Leakage B 3.4.5 BASES APPLICABILITY or during transition to or from, the RHR shutdown cooling mode of (continued) operation. This is because RHR shutdown cooling will be placed in operation only below the current pressure permissive setpoint when the high to low pressure interface does not exist.
In MODES 4 and 5, leakage limits are not provided because the lower reactor coolant pressure results in a reduced potential for leakage and for a LOCA outside the containment. Accordingly, the potential for the consequences of reactor coolant leakage is far lower during these MODES.
ACTIONS The ACTIONS are modified by two Notes. Note 1 has been provided to modify the ACTIONS related to RCS PIV flow paths. Section 1.3, Completion Times, specifies 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 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 the Condition of RCS PIV leakage limits exceeded provide appropriate compensatory measures for separate affected RCS PIV flow paths. As such, a Note has been provided that allows separate Condition entry for each affected RCS PIV flow path. Note 2 requires an evaluation of affected systems if a PIV is inoperable. The leakage may have affected system OPERABILITY, or isolation of a leaking flow path with an alternate valve may have degraded the ability of the interconnected system to perform its safety function. As a result, the applicable Conditions and Required Actions for systems made inoperable by PIVs must be entered. This ensures appropriate remedial actions are taken, if necessary, for the affected systems.
A.1 If leakage from one or more RCS PIVs is not within limit, the flow path must be isolated by at least one closed manual, deactivated automatic, or check valve within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
(continued)
SUSQUEHANNA - UNIT 2 B 3.4-26 Revision 0
PPL Rev. 1 RCS PIV Leakage B 3.4.5 BASES ACTIONS A.1 (continued)
Required Action A.i is modified by a Note stating that the valves used for isolation must meet the same leakage requirements as the PIVs and must be on the RCPB or the high pressure portion of the system.
Four hours provides time to reduce leakage in excess of the allowable limit and to isolate the flow path if leakage cannot be reduced while corrective actions to reseat the leaking PIVs are taken. The 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> allows time for thesis actions and restricts the time of operation with leaking valves.
B.1 and B.2 If leakage cannot be reduced or the system isolated, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to MODE 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and MODE 4 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. This action may reduce the leakage and also reduces the potential for a LOCA outside the containment. The Completion Times are reasonable, based on operating experience, to achieve the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.
SURVEILLANCE SR 3.4.5.1 REQUIREMENTS Performance of leakage testing on each RCS PIV is required to verify that leakage is below the specified limit and to identify each leaking valve.
The leakage limit of 0.5 gpm per inch of nominal valve diameter up to 5 gpm maximum applies to each valve. Leakage testing requires a stable pressure condition. For the two PIVs in series, the leakage requirement applies to each valve individually and not to the combined leakage across both valves. If the PIVs are not individually leakage tested, one valve may have failed completely and not be detected if the other valve in series meets the leakage requirement. In this situation, the protection provided by redundant valves would be lost.
(continued)
SUSQUEHANNA - UNIT 2 B 3.4-27 Revision 0
PPL Rev. I RCS PIV Leakage B 3.4.5 BASES SURVEILLANCE SR 3.4.5.1 (continued)
REQUIREMENTS The 24 month Frequency required by the Inservice Testing Program is within the ASME OM Code Frequency requirement and is based on the need to perform this Surveillance during an outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power.
This SR is modified by a Note that states the leakage Surveillance is not required to be performed in MODE 3. Entry into MODE 3 is permitted for leakage testing at high differential pressures with stable conditions not possible in the lower MODES.
REFERENCES
- 1. 10 CFR 50.2.
- 2. 10 CFR 50.55a(c).
- 4. ASME Operation and Maintenance Code.
- 5. NUREG-0677, May 1980.
- 6. Final Policy Statement on Technical Specifications Improvements, July 22, 1993 (58 FR 39132).
SUSQUEHANNA - UNIT 2 TS / B 3.4-28 Revision 1
.B 3.4.5 BASES (continued)
TABLE B 3.4.5-1 REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES 1st Isolation 2nd Isolation Valve(s) Number(s)
Valve(s) Number(s)
Service HV-252F006A HV-252F005A Core Spray Injection HV-252F037A HV-252F006B HV-252F005B Core Spray Injection HV-252F037B HV-251 F050A HV-251 F01 5A LPCI Injection HV-251 F122A HV-251 F050B HV-251 FO15B LPCI Injection HV-251 F122B HV-251 F022 HV-251 F023 Head Spray HV-251 F009 HV-251 F008 Shutdown Cooling SUSQUEHANNA - UNIT 2 T'S / B 3.4 Revision 1
PPL Rev. 3 ECCS-Operating B 3.5.1 B 3.5 B 3.5.1 EMERGENCY CORE COOLING SYSTEMS (ECCS) AND REACTOR CORE ISOLATION COOLING (RCIC) SYSTEM ECCS-Operating BASES BACKGROUND The ECCS is designed, in conjunction with the primary and secondary containment, to limit the release of radioactive materials to the environment following a loss of coolant accident (LOCA). The ECCS uses two independent methods (flooding and spraying) to cool the core during a LOCA. The ECCS network consists of the High Pressure Coolant Injection (HPCI) System, the Core Spray (CS) System, the low pressure coolant injection (LFPCI) mode of the Residual Heat Removal (RHR)
System, and the Automatic Depressurization System (ADS). The suppression pool provides the required source of water for the ECCS.
Although no credit is taken in the safety analyses for the condensate storage tank (CST), it is capable of providing a source of water for the HPCI and CS systems.
On receipt of an initiation signal, ECCS pumps automatically start; simultaneously, the system aligns and the pumps inject water, taken either from the CST or suppression pool, into the Reactor Coolant System (RCS) as RCS pressure is overcome by the discharge pressure of the ECCS pumps. Although the system is initiated, ADS action is delayed, allowing the operator to interrupt the timed sequence if the system is not needed.
The HPCI pump discharge pressure quickly exceeds that of the RCS, and the pump injects coolant into the vessel to cool the core. If the break is small, the HPCI System will maintain coolant inventory as well as vessel level while the RCS is still pressurized. If HPCI fails, it is backed up by ADS in combination with LPCI and CS. In this event absent operator action, the ADS timed sequence would time out and open the selected safety/relief valves ( S/RVs) depressurizing the RCS, thus allowing the LPCI and CS to overcome RCS pressure and inject coolant into the vessel. If the break is large, RCS pressure initially drops rapidly and the LPCI and CS cool the core.
Water from the break returns to the suppression pool where it is used again and again. Water in the suppression pool is circulated through a heat exchanger cocled by the RHR Service Water System. Depending on the location and size of the break, portions of the ECCS may be ineffective; (continued)
SUSQUEHANNA - UNIT 2 0I T" / B 3.5-1 Revision 1
PPL Rev. 3 ECCS-Operating B 3.5.1 BASES BACKGROUND however, the overall design is effective in cooling the core regardless of (continued) the size or location of the piping break. Although no credit is taken in the safety analysis for the RCIC System, it performs a similar function as HPCI, but has reduced makeup capability. Nevertheless, it will maintain inventory and cool the core while the RCS is still pressurized following a reactor pressure vessel (RPV) isolation.
All ECCS subsystems are designed to ensure that no single active component failure will prevent automatic initiation and successful operation of the minimum required ECCS equipment.
The CS System (Ref. 1) is composed of two independent subsystems.
Each subsystem consists of two motor driven pumps, a spray sparger above the core, and piping and valves to transfer water from the suppression pool to the sparger. The CS System is designed to provide cooling to the reactor core when reactor pressure is low. Upon receipt of an initiation signal, the CS pumps in both subsystems are automatically started when AC power is available. When the RPV pressure drops sufficiently, CS System flow to the RPV begins. A full flow test line is provided to route water from and to the suppression pool to allow testing of the CS System without spraying water in the RPV.
LPCI is an independent operating mode of the RHR System. There are two LPCI subsystems (Ref. 2), each consisting of two motor driven pumps and piping and valves to transfer water from the suppression pool to the RPV via the corresponding recirculation loop. The two LPCI subsystems can be interconnected via the RHR System cross tie valves; however, at least one of the two cross tie valves is maintained closed with its power removed to prevent loss of both LPCI subsystems during a LOCA. The LPCI subsystems are designed to provide core cooling at low RPV pressure. Upon receipt of an initiation signal, all four LPCI pumps are automatically started. RHR System valves in the LPCI flow path are automatically positioned to ensure the proper flow path for water from the suppression pool to inject into the recirculation loops. When the RPV pressure drops sufficiently, the LPCI flow to the RPV, via the corresponding recirculation loop, begins. The water then enters the reactor through the jet pumps.
(continued)
SUSQUEHANNA - UNIT 2 TS / B 3.5-2 Revision 1
PPL Rev. 3 ECCS-Operating B 3.5.1 BASES BACKGROUND Full flow test lines are provided for each LPCI subsystem to route water (continued) from the suppression pool, to allow testing of the LPCI pumps without injecting water into the RPV. These test lines also provide suppression pool cooling capability, as described in LCO 3.6.2.3, "RHR Suppression Pool Cooling."
The HPCI System (Ref. 3) consists of a steam driven turbine pump unit, piping, and valves to provide steam to the turbine, as well as piping and valves to transfer water from the suction source to the core via the feedwater system line, where the coolant is distributed within the RPV through the feedwater sparger. Suction piping for the system is provided from the CST and the suppression pool. Pump suction for HPCI is normally aligned to the CST source to minimize injection of suppression pool water into the RPV. Whenever the CST water supply is low, an automatic transfer to the suppression pool water source ensures an adequate suction head for the pump and an uninterrupted water supply for continuous operation of the HPCI System. The steam supply to the HPCI turbine is piped from a main steam line upstream of the associated inboard main steam isolation valve.
The HPCI System is designed to provide core cooling for a wide range of reactor pressures (165 psia to 1225 psia). Upon receipt of an initiation signal, the HPCI turbine stop valve and turbine control valve open and the turbine accelerates to a specified speed. As the HPCI flow increases, the turbine control valve is automatically adjusted to maintain design flow.
Exhaust steam from the HPCI turbine is discharged to the suppression pool. A full flow test line is provided to route water to the CST to allow testing of the HPCI System during normal operation without injecting water into the RPV.
The ECCS pumps are provided with minimum flow bypass lines, which discharge to the suppression pool. The valves in these lines automatically open to prevent pump damage due to overheating when other discharge line valves are closed. To ensure rapid delivery of water to the RPV and to minimize water hammer effects, all ECCS pump discharge lines are filled with water. The HPCI, LPCI and CS System discharge lines are kept full of water using a "keep fill" system that is supplied using the condensate transfer system.
(continued)
SUSQUEHANNA - UNIT 2 730 / B 3.5-3 Revision 2
PPL Rev. 3 ECCS-Operating B 3.5.1 BASES BACKGROUND (continued)
The ADS (Ref. 4) consists of 6 of the 16 S/RVs. It is designed to provide depressurization of the RC") during a small break LOCA if HPCI fails or is unable to maintain required water level in the RPV. ADS operation reduces the RPV pressure to within the operating pressure range of the low pressure ECCS subsystems (CS and LPCI), so that these subsystems can provide coolant inventory makeup. Each of the S/RVs used for automatic depressurization is equipped with two gas accumulators and associated inlet check valves. The accumulators provide the pneumatic power to actuate the valves.
APPLICABLE SAFETY ANALYSES The ECCS performance is evaluated for the entire spectrum of break sizes for a postulated LOCA. The accidents for which ECCS operation is required are presented in References 5, 6, and 7. The required analyses and assumptions are defined in Reference 8. The results of these analyses are also described in Reference 9.
This LCO helps to ensure that the following acceptance criteria for the ECCS, established by 10 CFR 50.46 (Ref. 10), will be met following a LOCA, assuming the worst case single active component failure in the ECCS:
- a. Maximum fuel element cladding temperature is < 22000F;
- b. Maximum cladding oxidation is < 0.17 times the total cladding thickness before oxidation;
< 0.01 times the hypothetical amount that would be generated if all of the metal in the cladding surrounding the fuel, excluding the cladding surrounding the plenum volume, were to react;
- d. The core is maintained in a coolable geometry; and
- e. Adequate long term cooling capability is maintained.
SPC performed LOCA calculations for the SPC ATRIUM m-10 fuel design.
The limiting single failures for the SPC analyses are discussed in Reference 11. The LOCA calculations examine both recirculation pipe and non-recirculation pipe breaks. For the recirculation pipe breaks, breaks on both the discharge and suction side of the recirculation pump are performed for two geometries; double-ended guillotine break and split break. The LOCA calculations demonstrate that the most limiting (highest PCT) break is a double-ended guillotine break in the recirculation pump suction piping. The limiting single failure is the failure of the LPCI injection valve in the intact recirculation loop to open.
(continued)
SUSQUEHANNA - UNIT 2 TS / B 3.5-4 Revision 2
PPL Rev. 3 ECCS-Operating B 3.5.1 BASES -
APPLICABLE SAFETY One ADS valve failure is analyzed as a limiting single failure for events ANALYSES requiring ADS operation. The remaining OPERABLE ECCS subsystems (continued) provide the capability to adequately cool the core and prevent excessive fuel damage.
The ECCS satisfy Criterion 3 of the NRC Policy Statement (Ref. 15).
I LCO Each ECCS injection/spray subsystem and six ADS valves are required to be OPERABLE. The ECCS injection/spray subsystems are defined as the two CS subsystems, the two LPCI subsystems, and one HPCI System.
The low pressure ECCS injection/spray subsystems are defined as the two CS subsystems; and the two LPCI subsystems.
With less than the required number of ECCS subsystems OPERABLE, the potential exists that during a limiting design basis LOCA concurrent with the worst case single failure, the limits specified in Reference 10 could be exceeded. All ECCS subsystems must therefore be OPERABLE to satisfy the single failure criterion required by Reference 10.
LPCI subsystems may be considered OPERABLE during alignment and operation for decay heat removal when below the actual RHR cut in permissive pressure in MODE 3, if capable of being manually realigned (remote or local) to the LPCI mode and not otherwise inoperable. At these low pressures and decay heat levels, a reduced complement of ECCS subsystems should provide the required core cooling, thereby allowing operation of RHR shutdown cooling when necessary.
(continued)
SUSQUEHANNA - UNIT 2 TS / B 3.5-5 Revision 3
PPL Rev. 3 ECCS-Operating B 3.5.1 BASES APPLICABILITY All ECCS subsystems are required to be OPERABLE during MODES 1, 2, and 3, when there is considerable energy in the reactor core and core cooling would be required to prevent fuel damage in the event of a break in the primary system piping. In MODES 2 and 3, when reactor steam dome pressure is < 150 psig, ADS and HPCI are not required to be OPERABLE because the low pressure ECCS subsystems can provide sufficient flow below this pressure. ECCS requirements for MODES 4 and 5 are specified in LCO 3.5.2, ECCS-Shutdown."
ACTIONS A Note prohibits the application of LCO 3.0.4.b to an inoperable HPCI subsystem. There is an increased risk associated with entering a MODE or other specified condition in the Applicability with an inoperable HPCI subsystem and the provisions of LCO 3.0.4.b, which allow entry into a MODE or other specified condition in the Applicability with the LCO not met after performance of a risk assessment addressing inoperable systems and components, should not be applied in this circumstance.
A.1 If any one low pressure ECCS injection/spray subsystem is inoperable for reasons other than Condition B, the inoperable subsystem must be restored to OPERABLE status within 7 days. In this Condition, the remaining OPERABLE subsystems provide adequate core cooling during a LOCA. However, overall ECCS reliability is reduced, because a single failure in one of the remaining OPERABLE subsystems, concurrent with a LOCA, may result in the ECCS not being able to perform its intended safety function. The 7 day Completion Time is based on a reliability study (Ref. 12) that evaluated the impact on ECCS availability, assuming various components and subsystems were taken out of service. The results were used to calculate the average availability of ECCS equipment needed to mitigate the consequences of a LOCA as a function of allowed outage times (i.e., Completion Times).
B.1 If one LPCI pump in one or both LPCI subsystems is inoperable, the inoperable LPCI pumps must be restored to OPERABLE status within 7 days.
In this Condition, the remaining OPERABLE LPCI pumps and at least one CS subsystem provide adequate core cooling during a LOCA.
However, overall ECCS reliability is reduced, because a single failure in one of the remaining OPERABLE subsystems, concurrent with a LOCA, may result in the ECCS not being able to perform its intended safety function. A 7 day Completion Time is based on a reliability study cited in Reference 12 and has been found to be acceptable through operating experience.
(continued)
SUSQUEHANNA - UNIT 2 TS / B3.5-6 Revision 2
PPL Rev. 3 ECCS-Operating B 3.5.1 BASES ACTIONS C.1 and C.2 (continued)
If the inoperable low pressure ECCS subsystem or LPCI pump(s) cannot be restored to OPERABLE status within the associated Completion Time, the plant must be brought to a MODE in which the LCO does 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 /> and to MODE 4 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.
D.1 and D.2 If the HPCI System is inoperable and the RCIC System is verified to be OPERABLE, the HPCI System must be restored to OPERABLE status within 14 days. In this Condition, adequate core cooling is ensured by the OPERABILITY of the redundant and diverse low pressure ECCS injection/spray subsystems in conjunction with ADS. Also, the RCIC System will automatically provide makeup water at most reactor operating pressures. Verification of RCIC OPERABILITY is therefore required when HPCI is inoperable. This may be performed as an administrative check by examining logs or other information to determine if RCIC is out of service for maintenance or other reasons. It does not mean to perform the Surveillances needed to demonstrate the OPERABILITY of the RCIC System. If the OPERABILITY of the RCIC System cannot be verified, however, Condition H must be immediately entered. If a single active component fails concurrent with a design basis LOCA, there is a potential, depending on the specific failure, that the minimum required ECCS equipment will not be available. A 14 day Completion Time is based on a reliability study cited in Reference 12 and has been found to be acceptable through operating experience.
E.1 and E.2 If Condition A or Condition B exists in addition to an inoperable HPCI System, the inoperable low pressure ECCS injection/spray subsystem or the LPCI pump(s) or the HPCI System must be restored to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. In this Condition, adequate core cooling is ensured by the OPERABILITY of the ADS and the remaining low pressure ECCS subsystems. However, the overall ECCS reliability is significantly reduced because a single failure in one of the remaining OPERABLE subsystems (continued)
SUSQUEHANNA - UNIT 2 TS, / B 3.5-7 Revision 1
PPL Rev. 3 ECCS-Operating B 3.5.1 BASES ACTIONS E.1 and E.2 (continued) concurrent with a design basis LOCA may result in the ECCS not being able to perform its intended safety function. Since both a high pressure system (HPCI) and a low pressure subsystem are inoperable, a more restrictive Completion Time of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> is required to restore either the HPCI System or the low pressure ECCS injection/spray subsystem to OPERABLE status. This Completion Time is based on a reliability study cited in Reference 12 and has been found to be acceptable through operating experience.
F. 1 The LCO requires six ADS valves to be OPERABLE in order to provide the ADS function. Reference 11 contains the results of an analysis that evaluated the effect of one ADS valve being out of service. Per this analysis, operation of only five ADS valves will provide the required depressurization. However, overall reliability of the ADS is reduced, because a single failure in the OPERABLE ADS valves could result in a reduction in depressurization capability. Therefore, operation is only allowed for a limited time. The 14 day Completion Time is based on a reliability study cited in Reference 12 and has been found to be acceptable through operating experience.
G.1 and G.2 If Condition A or Condition B exists in addition to one inoperable ADS valve, adequate core cooling is ensured by the OPERABILITY of HPCI and the remaining low pressure ECCS injection/spray subsystem.
However, overall ECCS reliability is reduced because a single active component failure concurrent with a design basis LOCA could result in the minimum required ECCS equipment not being available. Since both a high pressure system (ADS) and a low pressure subsystem are inoperable, a more restrictive Completion Time of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> is required to restore either the low pressure ECCS subsystem or the ADS valve to OPERABLE status. This Completion Time is based on a reliability study cited in Reference 12 and has been found to be acceptable through operating experience.
(continued)
SUSQUEHANNA - UNIT 2 TS / B 3.5,-8 Revision 1
PPL Rev. 3 ECCS-Operating B 3.5.1 BASES ACTIONS H.1 and H.2 (continued)
If any Required Action and associated Completion Time of Condition D, E, F, or G is not met, or if two or more ADS valves are inoperable, the plant must be brought to a condition in which the LCO does 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 /> and reactor steam dome pressure reduced to < 150 psig within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.
1.1 When multiple ECCS subsystems are inoperable, as stated in Condition I, LCO 3.0.3 must be entered immediately.
SURVEILLANCE SR 351.1 REQUIREMENTS The flow path piping has the potential to develop voids and pockets of entrained air. Maintaining the pump discharge lines of the HPCI System, CS System, and LPCI subsystems full of water ensures that the ECCS will perform properly, injecting its full capacity into the RCS upon demand.
This will also prevent a water hammer following an ECCS initiation signal.
One acceptable method of ensuring that the lines are full is to vent at the high points. The 31 day Frequency is based on the gradual nature of void buildup in the ECCS piping, the procedural controls governing system operation, and operating experience.
SR 3.5.1.2 Verifying the correct alignment for manual, power operated, and automatic valves in the ECCS flow paths provides assurance that the proper flow paths will exist for ECCS operation. This SR does not apply to valves that are locked, sealed, or otherwise secured in position since these were verified to be in the correct position prior locking, sealing, or securing. A valve that receives an initiation signal is allowed to be in a nonaccident position provided the valve will automatically reposition in the proper stroke time. This SR does not require any testing or valve manipulation; rather, it involves verification that those valves capable of potentially being mispositioned are in (continued)
SUSQUEHANNA - UNIT 2 Tr_- / B 3.5-9 Revision 1
PPL Rev. 3 ECCS-Operating B 3.5.1 BASES SURVEILLANCE S
cniud REQUIREMENTS SR 3-512 (continued) the correct position. This SR does not apply to valves that cannot be inadvertently misaligned, such as check valves. For the HPCI System, this SR also includes the steam flow path for the turbine and the flow controller position.
The 31 day Frequency of this SR was derived from the Inservice Testing Program requirements for performing valve testing at least once every 92 days. The Frequency of 31 days is further justified because the valves are operated under procedural control and because improper valve position would only affect a single subsystem. This Frequency has been shown to be acceptable through operating experience.
This SR is modified by a Note that allows LPCI subsystems to be considered OPERABLE during alignment and operation for decay heat removal with reactor steam dome pressure less than the RHR cut in permissive pressure in MODE 3, if capable of being manually realigned (remote or local) to the LPCI mode and not otherwise inoperable. This allows operation in the RHR shutdown cooling mode during MODE 3, if necessary.
SR 3.5.1.3 Verification every 31 days that ADS gas supply header pressure is 2 135 psig ensures adequate gas pressure for reliable ADS operation.
The accumulator on each ADS valve provides pneumatic pressure for valve actuation. The design pneumatic supply pressure requirements for the accumulator are such that, following a failure of the pneumatic supply to the accumulator, at least one valve actuations can occur with the drywell at 70% of design pressure.
The ECCS safety analysis assumes only one actuation to achieve the depressurization required for operation of the low pressure ECCS. This minimum required pressure of 2 135 psig is provided by the containment instrument gas system. The 31 day Frequency takes into consideration administrative controls over operation of the gas system and alarms associated with the containment instrument gas system.
(continued)
SUSQUEHANNA - UNIT 2 TS/ B 3.5-10 Revision 1
PPL Rev. 3 ECCS-Operating B 3.5.1 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.5.1.4 Verification every 31 days that at least one RHR System cross tie valve is closed and power to its operator is disconnected ensures that each LPCI subsystem remains independent and a failure of the flow path in one subsystem will not affect the flow path of the other LPCI subsystem.
Acceptable methods of removing power to the operator include opening the breaker, or racking out the breaker, or removing the breaker. If both RHR System cross tie valves are open or power has not been removed from at least one closed valve operator, both LPCI subsystems must be considered inoperable. The 31 day Frequency has been found acceptable, considering that these valves are under strict administrative controls that will ensure the valves continue to remain closed with motive power removed.
SR 3.5.1.5 Verification every 31 days that each 480 volt AC swing bus transfers automatically from the normal source to the alternate source on loss of power while supplying its respective bus demonstrates that electrical power is available to ensure proper operation of the associated LPCI inboard injection and minimum flow valves and the recirculation pump discharge and bypass valves. Therefore, each 480 volt AC swing bus must be OPERABLE for the associated LPCI subsystem to be OPERABLE. The test is performed by actuating the load test switch or by disconnecting the preferred power source to the transfer switch and verifying that swing bus automatic transfer is accomplished. The 31 day Frequency has been found to be acceptable through operating experience.
SR 3.5.1.6 Cycling the recirculation pump discharge and bypass valves through one complete cycle of full travel demonstrates that the valves are mechanically OPERABLE and provides assurance that the valves will close when required to ensure the proper LPCI flow path is established. Upon initiation of an automatic LPCI subsystem injection signal, these valves are required to be closed to ensure full LPCI subsystem flow injection in the reactor via the recirculation jet pumps. De-energizing the valve in the closed position will also ensure the proper flow path for the LPCI subsystem. Acceptable methods of de-energizing the valve include opening the breaker, or racking out the breaker, or removing the breaker.
(continued)
SUSQUEHANNA - UNIT 2 TS; / B 3.5-11 Revision 2
PPL Rev. 3 ECCS-Operating B 3.5.1 BASES SURVEILLANCE SR 3.5.1.6 (continued)
REQUIREMENTS The specified Frequency is once during reactor startup before THERMAL POWER is > 25% RTP. However, this SR is modified by a Note that states the Surveillance is only required to be performed if the last performance was more than 31 days ago. Therefore, implementation of this Note requires this test to be performed during reactor startup before exceeding 25% RTP. Verification during reactor startup prior to reaching
> 25% RTP is an exception to the normal Inservice Testing Program generic valve cycling Frequency of 92 days, but is considered acceptable due to the demonstrated reliability of these valves. If the valve is inoperable and in the open position, the associated LPCI subsystem must be declared inoperable.
SR 3.5.1.7. SR 3.5.1.8. and SR 3.5.1.9 The performance requirements of the low pressure ECCS pumps are determined through application of the 10 CFR 50, Appendix K criteria (Ref. 8). This periodic Surveillance is performed (in accordance with the ASME OM Code requirements for the ECCS pumps) to verify that the ECCS pumps will develop the flow rates required by the respective analyses. The low pressure ECCS pump flow rates ensure that adequate core cooling is provided to satisfy the acceptance criteria of Reference 10.
The pump flow rates are verified against a system head equivalent to the RPV pressure expected during a LOCA. The total system pump outlet pressure is adequate to overcome the elevation head pressure between the pump suction and the vessel discharge, the piping friction losses, and RPV pressure present during a LOCA. These values may be established during preoperational testing.
The flow tests for the HPCI System are performed at two different pressure ranges such that system capability to provide rated flow is tested at both the higher and lower operating ranges of the system. Additionally, adequate steam flow must be passing through the main turbine or turbine bypass valves to continue to control reactor pressure when the HPCI System diverts steam flow. Reactor steam pressure is considered adequate when 2 920 psig to perform SR 3.5.1.8 and 2 150 psig to perform SR 3.5.1.9. However, the requirements of SR 3.5.1.9 are met by a successful performance at any pressure < 165 psig. Adequate steam flow is represented by at least 1.25 turbine bypass valves open.
Therefore, sufficient time is allowed after adequate pressure and flow are achieved to perform these tests. Reactor startup is allowed prior to performing the low pressure Surveillance test because the reactor pressure is low and the time allowed to satisfactorily (continued)
SUSQUEHANNA - UNIT 2 TS / B 3.5-12 Revision 2
PPL Rev. 3 ECCS-Operating B 3.5.1 BASES SURVEILLANCE SR 3.5.1.7. SR 3.5.1.8, and SR 3.5.1.9 (continued)
REQUIREMENTS perform the Surveillance test is short. The reactor pressure is allowed to be increased to normal operating pressure since it is assumed that the low pressure test has been satisfactorily completed and there is no indication or reason to believe that HPCI is inoperable.
Therefore, SR 3.5.1.8 and SR 3.5.1.9 are modified by Notes that state the Surveillances are not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after the reactor steam pressure and flow are adequate to perform the test.
The Frequency for SR 3.5.1.7 and SR 3.5.1.8 is in accordance with the Inservice Testing Program requirements. The 24 month Frequency for SR 3.5.1.9 is based on the need to perform the Surveillance under the conditions that apply just prior to or during a startup from a plant outage.
Operating experience has shown that these components usually pass the SR when performed at the 24 month Frequency, which is based on the refueling cycle. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.
SR 3.5.1.10 The ECCS subsystems are required to actuate automatically to perform their design functions. This Surveillance verifies that, with a required system initiation signal (actual or simulated), the automatic initiation logic of HPCI, CS, and LPCI will cause the systems or subsystems to operate as designed, including actuation of the system throughout its emergency operating sequence, automatic pump startup and actuation of all automatic valves to their required positions. This functional test includes the LPCI and CS interlocks between Unit 1 and Unit 2 and specifically requires the following:
A functional test of the interlocks associated with the LPCI and CS pump starts in response to an automatic initiation signal in Unit 1 followed by a false automatic initiation signal in Unit 2; A functional test of the interlocks associated with the LPCI and CS pump starts in response to an automatic initiation signal in Unil 2 followed by a false automatic initiation signal in Unit 1; and (continued)
SUSQUEHANNA - UNIT 2 TS / B 3.5-13 Revision I
PPL Rev. 3 ECCS-Operating B 3.5.1.
BASES SREQVUILRELMENTS SR 3.5.1.10 (continued)
A functional test of the interlocks associated with the LPCI and CS pump starts in response to simultaneous occurrences of an automatic initiation signal in both Unit 1 and Unit 2 and a loss of Offsite power condition affecting both Unit 1 and Unit 2.
The purpose of this functional test (preferred pump logic) is to assure that if a false LOCA signal were to be received on one Unit simultaneously with an actual LOCA signal on the second Unit, the preferred LPCI and CS pumps are started and the non-preferred LPCI and CS pumps are tripped for each Unit. This functional test is performed by verifying that the non-preferred LPCI and CS pumps are tripped. The verification that preferred LPCI and CS pumps start is performed under a separate surveillance test. Only one division of LPCI preferred pump logic is required to be OPERABLE for each Unit, because no additional failures needs to be postulated with a false LOCA signal. If the preferred or non-preferred pump logic for CS is inoperable, the associated CS pumps shall be declared inoperable and the pumps should not be operated to ensure that the opposite Unit's CS pumps or 4.16 kV ESS Buses are protected.
This SR also ensures that the HPCI System will automatically restart on an RPV low water level (Level 2) signal received subsequent to an RPV high water level (Level 8) trip and that the suction is automatically transferred from the CST to the suppression pool. The LOGIC SYSTEM FUNCTIONAL TEST perfonmed in LCO 3.3.5.1 overlaps this Surveillance.
This SR can be accomplished by any series of sequential overlapping or total steps such that the entire channel is tested.
The 24 month Frequency is acceptable because operating experience has shown that these components usually pass the SR when performed at the 24 month Frequency, which is based on the refueling cycle. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.
This SR is modified by a Note that excludes vessel injection/spray during the Surveillance. Since all active components are testable and full flow can be demonstrated by recirculation through the test line, coolant injection into the RPV is not required during the Surveillance.
(continued)
SUSQUEHANNA - UNIT 2 TS / B 3.5-14 Revision 1
PPL Rev. 3 ECCS-Operating B 3.5.1 BASES SURVEILLANCE SR 351.11 REQUIREMENTS (continued)
The ADS designated S/RVs are required to actuate automatically upon receipt of specific initiation signals. A system functional test is performed to demonstrate that the mechanical portions of the ADS function (i.e.,
solenoids) operate as designed when initiated either by an actual or simulated initiation signal, causing proper actuation of all the required components. SR 3.5.1.12 and the LOGIC SYSTEM FUNCTIONAL TEST performed in LCO 3.3.5.1 overlap this Surveillance to provide complete testing of the assumed safety function.
The 24 month Frequency is based on the need to perform portions of the 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 SR when performed at the 24 month Frequency, which is based on the refueling cycle. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.
This SR is modified by a Note that excludes valve actuation. This prevents an RPV pressure blowdown.
SR 3.5.1.12 A manual actuation of each ADS valve is performed to verify that the valve and solenoid are functioning properly. This is demonstrated by one of the two methods described below. Proper operation of the valve tailpipes is ensured through the use of foreign material exclusion during maintenance.
One method is by manual actuation.of ADS valve under hot conditions.
Proper functioning cf the valve and solenoid is demonstrated by the response of turbine control or bypass valve or by a change in the measured flow or by any other method suitable to verify steam flow.
Adequate reactor steam dome pressure must be available to perform this test to avoid damagiing the valve due to seat impact during closure. Also, adequate steam flow must be passing through the main turbine or turbine bypass valves to continue to control reactor pressure when the ADS valves divert steam flow upon opening. Sufficient time is therefore allowed after the required pressure and flow are achieved to perform this SR. Adequate pressure at which this SR is to be performed is 150 psig.
However, the requirements of SR 3.5.1.12 are met by a successful performance at any pressure. Adequate steam flow is represented by at least 1.25 turbine bypass valves open. Reactor startup is allowed prior to performing this SR by this method because valve OPERABILITY and the setpoints for (continued)
SUSQUEHANNA - UNIT 2 TS / B 3.5-15 Revision 2
PPL Rev. 3 ECCS-Operating B 3.5.1 BASES SURVEILLANCE SR3511(cniud REQUIREMENTS S
.5.
(c) overpressure protection are! verified, per ASME requirements, prior to valve installation. Therefore, this SR is modified by a Note that states the Surveillance is not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reactor steam pressure and flow are adequate to perform the test. The 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> allowed for manual actuation after the required pressure is reached is sufficient to achieve stable conditions and provides adequate time to complete the Surveillance.
Another method is by manual actuation of the ADS valve at atmospheric temperature and pressure during cold shutdown. When using this method, proper functioning of the valve and solenoid is demonstrated by visual observation of actuator movement. Actual disc travel is measured during valve refurbishment and testing per ASME requirements. Lifting the valve at atmospheric pressure requires controlling the actuator to set the valve disc softly on its seat to prevent valve damage. Lifting of the valves at atmospheric pressure is the preferred method because lifting the valves with steam flow increases the likelihood that the valve will leak.
The Note that modified this SR is not needed when this method is used because the SR is performed during cold shutdown.
SR 3.5.1.11 and the LOGIC SYSTEM FUNCTIONAL TEST performed in LCO 3.3.5.1 overlap this Surveillance to provide complete testing of the assumed safety function. The Frequency of 24 months on a STAGGERED TEST BASIS ensures that both solenoids for each ADS valve are alternately tested. The Frequency is based on the need to perform the Surveillance under the conditions that apply just prior to or during a startup from a plant outage. Operating experience has shown that these components usually pass the SR when performed at the 24 month Frequency, which is based on the refueling cycle. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.
SR 3.5.1.13 This SR ensures that the E'CCS RESPONSE TIME for each ECCS injection/spray subsystem is less than or equal to the maximum value assumed in the accident analysis. Response Time testing acceptance criteria are included in Reference 13. This SR is modified by a Note that allows the instrumentation portion of the response time to be assumed to be based on historical response time data and therefore, is excluded from the ECCS RESPONSE TIME testing. This is allowed since the instrumentation response time is a small part of the ECCS RESPONSE TIME (e.g., sufficient margin exists in the diesel generator start time when compared to the instrumentation response time) (Ref. 14).
(continued)
SUSQUEHANNA - UNIT 2 TS / B 3.5-16 Revision 2
PPL Rev. 3 ECCS-Operating
.B 3.5.1
- BASES SURVEILLANCE REQUIREMENTS SR 3.5.1.13 (continued)
The 24 month Frequency is consistent with the typical industry refueling cycle and is acceptable based upon plant operating experience.
REFERENCES
- 1.
FSAR, Section 6.3.2.2.3.
- 2.
FSAR, Section 6.3.2.2.4.
- 3.
FSAR, Section 6.3.2.2.1.
- 4.
FSAR, Section 6.3.2.2.2.
- 5.
FSAR, Section 15.2.8.
- 6.
FSAR, Section 15.6.4.
- 7.
FSAR, Section 15.6.5.
- 8.
- 9.
FSAR, Section 6.3.3.
- 10.
- 11.
FSAR, Section 6.3.3.
- 12. Memorandum from R.L. Baer (NRC) to V. Stello, Jr. (NRC),
"Recommended Interim Revisions to LCOs for ECCS Components,"
December 1, '1975.
- 13. FSAR, Section 6.3.3.3.
- 14. NEDO 32291-A, 'System Analysis for the Elimination of Selected Response Timie Testing Requirements, October 1995.
- 15. Final Policy Statement on Technical Specifications Improvements, July 22,1993 (58 FR 39132).
I I
I SUSQUEHANNA - UNIT 2 TS / B 3.5-17 Revision 3
PPL Rev. 3 ECCS-Operating B 3.5.1 THIS PAGE INTENTIONALLY LEFT BLANK SUSQUEHANNA - UNIT 2 TS / B 3.5-18 Revision 1
PPL Rev. 1 RHR Suppression Pool Cooling B 3.6.2.3 B 3.6 CONTAINMENT SYSTEMS B 3.6.2.3 Residual Heat Removal (RHR) Suppression Pool Cooling BASES BACKGROUND Following a Design Basis Accident (DBA), the RHR Suppression Pool Cooling System removes heat from the suppression pool. The suppression pool is designed to absorb the sudden input of heat from the primary system. In the long term, the pool continues to absorb residual heat generated by fuel in the reactor core. Some means must be provided to remove heat from the suppression pool so that the temperature inside the primary containment remains within design limits.
This function is provided by two redundant RHR suppression pool cooling subsystems. The purpose of this LCO is to ensure that both subsystems are OPERABLE in applicable MODES.
Each RHR subsystem contains either one of the two RHR pumps and a flow path capable of recirculating water from the suppression chamber through an RHR heat exchanger and is manually initiated and independently controlled. The two subsystems perform the suppression pool cooling function by circulating water from the suppression pool through the RHR heat exchangers and returning it to the suppression pool. RHR service water, circulating through the tube side of the heat exchangers, exchanges heat with the suppression pool water and discharges this heat to the external heat sink.
The heat removal capability of one RHR pump in one subsystem is sufficient to meet the overall DBA pool cooling requirement for loss of coolant accidents (LOCAs) and transient events such as a turbine trip or stuck open safety/relief valve (S/RV). S/RV leakage and High Pressure Coolant Injection and Reactor Core Isolation Cooling System testing increase suppression pool temperature more slowly. The RHR Suppression Pool Cooling System is also used to lower the suppression pool water bulk temperature following such events.
APPLICABLE Reference 1 contains the results of analyses used to predict primary SAFETY containment pressure and temperature following large and small break ANALYSES LOCAs. The intent of the analyses is to demonstrate that the heat removal capacity of the RHR (continued)
SUSQUEHANNA - UNIT 2 B 3.6-61 Revision 0
PPL Rev. 1 RHR Suppression Pool Cooling B 3.6.2.3 BASES APPLICABLE Suppression Pool Cooling System is adequate to maintain the primary SAFETY containment conditions within design limits. The suppression pool ANALYSES temperature is calculated to remain below the design limit.
(continued)
The RHR Suppression Pool Cooling System satisfies Criterion 3 of the NRC Policy Statement. (Ref. 3)
LCO During a DBA, a minimum of one RHR suppression pool cooling subsystem is required to maintain the primary containment peak pressure and temperature below design limits (Ref. 1). To ensure that these requirements are met, two RHR suppression pool cooling subsystems must be OPERABLE. Therefore, in the event of an accident, at least one subsystem is OPERABLE assuming the worst case single active failure.
An RHR suppression pool cooling subsystem is OPERABLE when one of the pumps, the heat exchanger, and associated piping, valves, instrumentation, and controls are OPERABLE.
APPLICABILITY In MODES 1, 2, and 3, a DBA could cause a release of radioactive material to primary containment and cause a heatup and pressurization of primary containment. In MODES 4 and 5, the probability and consequences of these events are reduced due to the pressure and temperature limitations in these MODES. Therefore, the RHR Suppression Pool Cooling System is not required to be OPERABLE in MODE 4 or 5.
ACTIONS A._
With one RHR suppression pool cooling subsystem inoperable, the inoperable subsystem must be restored to OPERABLE status within 7 days. In this Condition, the remaining RHR suppression pool cooling subsystem is adequate to perform the primary containment cooling function. However, the overall reliability is reduced because a single failure in the OPERABLE subsystem could result in loss of primary containment cooling capability. The 7 day Completion Time is acceptable in light of the redundant RHR suppression (continued)
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PPL Rev. 1 RHR Suppression Pool Cooling B 3.6.2.3 BASES ACTIONS A.1 (continued) pool cooling capabilities afforded by the OPERABLE subsystem and the low probability of a DBA occurring during this period.
B.1 With two RHR suppression pool cooling subsystems inoperable, one subsystem must be restored to OPERABLE status within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. In this condition, there is a substantial loss the of primary containment pressure and temperature mitigation function. The 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> Completion Time is based on this loss of function and is considered acceptable due to the low probability of a DBA and the potential avoidance of a plant shutdown transient that could result in the need for the RHR suppression pool cooling subsystems to operate.
C.1 and C.2 If the Required Action and associated Completion Time cannot be met, the plant must be brought to a MODE in which the LCO does 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 /> and to MODE 4 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.
SURVEILLANCE R 3.6.2.3.1 REQUIREMENTS Verifying the correct alignment for manual, power operated, and automatic valves in the RHR suppression pool cooling mode flow path provides assurance that the proper flow path exists for system operation. This SR does not apply to valves that are locked, sealed, or otherwise secured in position since these valves were verified to be in the correct position prior to locking, sealing, or securing. A valve is also allowed to be in the nonaccident position provided it can be aligned to the accident position within the time assumed in the accident analysis. This is acceptable since the RHR suppression pool cooling mode is manually initiated. This SR does not require any testing or valve manipulation; rather, it involves verification that those valves capable of being mispositioned are in the correct position. This SR does not apply to valves that cannot be inadvertently misaligned, such as check valves.
(continued)
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PPL Rev. 1 RHR Suppression Pool Cooling B 3.6.2.3 BASES SUP' 1fELLANCE
>; RENA64tTS SR 3.6.2.3.1 (continued)
The Frequency of 31 days is justified because the valves are operated under procedural control, improper valve position would affect only a single subsystem, the probability of an event requiring initiation of the system is low, and the subsystem is a manually initiated system. This Frequency has been shown to be acceptable based on operating experience.
SR 3.6.2.3.2 Verifying that each RHR pump develops a flow rate 2 9750 gpm while operating in the suppression pool cooling mode with flow through the associated heat exchanger ensures that pump performance has not degraded during the cycle. Flow is a normal test of centrifugal pump performance required by ASME OM Code (Ref. 2). This test confirms one point on the pump design curve, and the results are indicative of overall performance. Such inservice inspections confirm component OPERABILITY, trend performance, and detect incipient failures by indicating abnormal performance. The Frequency of this SR is in accordance with the Inservice Testing Program.
I REFERENCES
- 1. FSAR, Section 6.2.
- 2. ASME Operation and Maintenance Code.
- 3. Final Policy Statement on Technical Specifications Improvements, July 22, 1993 (513 FR 39132).
I SUSQUEHANNA - UNIT 2 TS / B 3.6-64 Revision 1