ML043070354
| ML043070354 | |
| Person / Time | |
|---|---|
| Site: | Susquehanna  |
| Issue date: | 10/28/2004 |
| From: | Susquehanna |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| Download: ML043070354 (19) | |
Text
Oct. 28, 2004 Page 1 of 1 MANUAL HARD COPY DISTRIBUTION DOCUMENT TRANSMITTAL 2004-44150 USE INFORMATION /
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THE FOLLOWING CHANGES HAVE OCCURRED TO THE HARDCOPY OR ELECTRONIC MANUAL ASSIGNED TO YOU:
TSB2 -
TECHNICAL SPECIFICATIONS BASES UNIT 2 MANUAL REMOVE MANUAL TABLE OF CONTENTS DATE: 10/08/2004
')D MANUAL TABLE OF CONTENTS DATE: 10/27/2004
\\'
TEGORY:
DOCUMENTS TYPE: TSB2 ID:
TEXT 2.1.1 REMOVE:
REV:0 ADD:
REV: 1 CATEGORY:
DOCUMENTS TYPE: TSB2 ID:
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SSES MANUAL
, Manual Name: TSB2 manual
Title:
TECHNICAL SPECIFICATIONS BASES UNIT 2 MANUAL Table Of Contents Issue Date:
10/27/2004 Procedure Name Rev TEXT LOES 49
Title:
LIST OF EFFECTIVE SECTIONS Zssie Date' 10/27/2004 1
Change ID Change Number 3
09/02/2004 TEXT TOC
Title:
TABLE OF CONTENTS TEXT 2.1.1 1
10/27/2004
Title:
SAFETY LIMITS (SLS) REACTOR CORE SLS
/
TEXT 2.1.2 0
11/18/2002
Title:
SAFETY LIMITS (SLS)
REACTOR COOLANT SYSTEM (RCS)NPRESSURE SL TEXT 3.0 0
11/18/2002 A
Title:
LIMITING CONDITION FOR OPERATION (LCO)
APPLICABILITY TEXT 3.1.1 0
/
11/18/2002
Title:
REACTIVITY CONTROL SYSTEMSSHUTDOWN MARGIN (SDM)
TEXT 3.1.2
'11/18/2002
Title:
REACTIVITY CONTROL; SYSTEMS REACTIVITY ANOMALIES TEXT 3.1.3 by 0<
m{
0 11/18/2002
Title:
REACTIVITY 'CONTROL SYSTEMS CONTROL ROD OPERABILITY
- .,,,.4),,,,.
m m
11 A4 tI-s t ltQ tn TI;xT i. 1.4
Title:
REACTIVITY CONTROL U
J.J.X J.0 c
uu SYSTEMS CONTROL ROD SCRAM TIMES
.~~.I TEXT 3.1.5 0
11/18/2002
Title:
REACTIVITY CONTROL SYSTEMS CONTROL ROD SCRAM ACCUMULATORS TEXT 3.1.6 0
11/18/2002
Title:
REACTIVITY CONTROL SYSTEMS ROD PATTERN CONTROL Pagel of 8
Report Date: 10/28/04 Page I of 8 Report Date: 10/28/04
SSES MANUAL Manual Name:
TSB2 Manual
Title:
TECHNICAL SPECIFICATIONS BASES UNIT 2 MANUAL Ii TEXT 3.1.7 0
11/18/2002
Title:
REACTIVITY CONTROL SYSTEMS STANDBY LIQUID CONTROL (SLC)
SYSTEM TEXT 3.1.8 0
11/18/2002
Title:
REACTIVITY CONTROL SYSTEMS SCRAM DISCHARGE VOLUME (SDV) VENT AND DRAIN VALVES TEXT 3.2.1 0
11/18/2002
Title:
POWER DISTRIBUTION LIMITS AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)
TEXT 3.2.2 0
11/18/2002
Title:
POWER DISTRIBUTION LIMITS MINIMUM CRITICAL POWER RATIO (MCPR)
TEXT 3.2.3
Title:
POWER DISTRIBUTION 0
11/18/2002 LIMITS LINEAR HEAT GENERATION RATE (LHGR)
TEXT 3.2.4 0
11/18/2002
Title:
POWER DISTRIBUTION. LIMITS AVERAGE POWER RANGE MONITOR (APRM)
GAIN AND SETPOINTSK--
TEXT 3.3.1.1 0
11/18/2002
Title:
INSTRUMENTATION REACTOR PROTECTION SYSTEM (RPS) INSTRUMENTATION TEXT 3.3.1.2 0
11/18/2002
Title:
INSTRUMENTATION SOURCE RANGE MONITOR (SRM) INSTRUMENTATION TEXT 3.3.2.1
Title:
INSTRUMENTATION 0
11/i8/2002 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
Title:
INSTRUMENTATION 0
11/18/2002 POST ACCIDENT MONITORING (PAM)
INSTRUMENTATION LDCN 3710 TEXT 3.3.3.2 0
11/18/2002
Title:
INSTRUMENTATION REMOTE SHUTDOWN SYSTEM Page2 of 8
Report Date: 10/28/04 Page 2 of 8 Report Date: 10/28/04
SSES MANUAL Manual Name:
TSB2 Manual
Title:
TECHNICAL SPECIFICATIONS BASES UNIT 2 MANUAL 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 0
11/18/2002,e
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 0
11/18/2002 PRIMARY CONTAINMENT ISOLATION INSTRUMENTATION 0
11/18/2002 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
Title:
INSTRUMENTATION 0
11/18/2002 REACTOR PROTECTION SYSTEM (RPS)
ELECTRIC POWER MONITORING TEXT 3.4.1 1
11/06/2003
Title:
RECIRCULATION LOOPS OPERATING TEXT 3.4.2 0
11/18/2002
Title:
REACTOR COOLANT SYSTEM (RCS) JET PUMPS TEXT 3.4.3 0
11/18/2002
Title:
SAFETY/RELIEF VALVES (S/RVS)
Page3 of 8
Report Date: 10/28/04 Page 3.
of 8 Report Date: 10/28/04
I SSES MANUAL Manual Name:
TSB2 Manual
Title:
TECHNICAL SPECIFICATIONS BASES UNIT 2 MANUAL I
TEXT 3.4.4 0
11/18/2002
Title:
REACTOR COOLANT SYSTEM (RCS) RCS OPERATIONAL LEAKAGE TEXT 3.4.5
Title:
REACTOR TEXT 3.4.6
Title:
REACTOR 0
11/18/2002 COOLANT SYSTEM (RCS)
RCS PRESSURE ISOLATION VALVE (PIV) LEAKAGE 0
11/18/2002 COOLANT SYSTEM (RCS) RCS LEAKAGE DETECTION INSTRUMENTATION TEXT 3.4.7 0
Title:
TEXT 3.4.8 0
Title:
HOT SHUTDOWN TEXT 3.4.9 0
Title:
COLD SHUTDOWN TEXT 3.4.10 0
Title:
REACTOR COOLANT SYSTEM (RCS) 11/18/2002 RCS SPECIFIC ACTIVITY 11/18/2002 RESIDUAL HEAT REMOVAL (RHR)
SHUTDOWN COOLING SYSTEM 11/18/2002 RESIDUAL HEAT REMOVAL (RHR)
SHUTDOWN COOLING SYSTEM--_-
11/18/2002 RCS PRESSURE AND TEMPERATURE (P/T) LIMITS TEXT 3.4.11 0
11/18/2002
Title:
REACTOR STEAM DOME PRESSURE TEXT 3.5.1 0O 11/18/2002
Title:
EMERGENCY CORE COOLING SYSTEMS (ECCS)
AND REACTOR CORE ISOLATION SYSTEM ECCS -
OPERATING TEXT 3.5.2 0
11/18/2002
Title:
EMERGENCY CORE COOLING SYSTEMS (ECCS) AND REACTOR CORE ISOLATION SYSTEM ECCS -
SHUTDOWN TEXT 3.5.3 0
11/18/2002
Title:
EMERGENCY CORE COOLING SYSTEMS (ECCS) AND REACTOR CORE ISOLATION SYSTEM RCIC SYSTEM COOLING (RCIC)
COOLING (RCIC)
COOLING (RCIC)
TEXT 3.6.1.1 0
11/18/2002
Title:
CONTAINMENT SYSTEMS PRIMARY CONTAINMENT Page4 of 8 Report Date: 10/28/04 Page 4 of 8 Report Date: 10/28/04
v SSES MANUEAL Manual Name:
TSB2 Manual
Title:
TECHNICAL SPECIFICATIONS BASES UNIT,2 MANUAL' 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
Title:
CONTAINMENT T'EXT 3.6.2.1 K-
Title:
CONTAINMENT TEXT 3.6.2.2
Title:
CONTAINMENT TEXT 3.6.2.3
Title:
CONTAINMENT TEXT 3.6.2.4
Title:
CONTAINMENT TEXT 3.6.3.1
Title:
CONTAINMENT TEXT 3.6.3.2
Title:
CONTAINMENT TEXT 3.6.3.3
Title:
CONTAINMENT o
11/18/2002.
SYSTEMS PRIMARY CONTAINMENT AIR LOCK 0
11/18/2002'>-
SYSTEMS PRIMARY CONTAINMENT ISOLATION VALVES (PCIVS) o 11/18/2002' SYSTEMS CONTAINMENT PRESSURE 0
11/18/2002-SYSTEMS DRYWELL AIR TEMPERATURE 0
11/18/2002 SYSTEMS SUPPRESSION CHAMBER-TO-DRYWELL VACUUM BREAKERS 0
11/18/2002 SYSTEMS SUPPRESSION POOL AVERAGE TEMPERATURE 0
11/18/2002,'
SYSTEMS SUPPRESSION POOL WATER'LEVEL o
11/18/2002 SYSTEMS RESIDUAL HEAT REMOVAL (RHR)'SUPPRESSION POOL COOLING 0
11/18/2002 SYSTEMS RESIDUAL HEAT REMOVAL-(RHR).SUPPRESSION POOL SPRAY 0
11/18/2002 SYSTEMS PRIMARY CONTAINMENT HYDROGEN RECOMBINERS 0
11/18/2002' SYSTEMS DRYWELL AIR FLOW SYSTEM 0
11/18/2002 SYSTEMS PRIMARY CONTAINMENT OXYGEN CONCENTRATION Page5 of 8
Report Date: 10/28/04 Page 5 of 8 Report Date: 10 /28/04
SSES MANUAL Manual Name:
TSB2 Manual
Title:
TECHNICAL SPECIFICATIONS BASES UNIT 2 MANUAL TEXT 3.6.4.1 0
11/18/2002
Title:
CONTAINMENT SYSTEMS SECONDARY CONTAINMENT TEXT 3.6.4.2 0
11/18/2002
Title:
CONTAINMENT SYSTEMS SECONDARY CONTAINMENT TEXT 3.6.4.3 1
10/08/2004
Title:
CONTAINMENT SYSTEMS STANDBY GAS TREATMENT ISOLATION VALVES (SCIVS)
(SGT)
SYSTEM TEXT 3.7.1
Title:
PLANT SYSTEMS ULTIMATE HEAT 0
11/18/2002 RESIDUAL HEAT REMOVAL SERVICE WATER (RHRSW)
SYSTEM AND THE 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 0
11/18/2002 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 0
11/18/2002 SYSTEMS MAIN TURBINE BYPASS SYSTEM 0
11/18/2002 SYSTEMS SPENT FUEL STORAGE POOL WATER LEVEL 1
10/17/2003
Title:
ELECTRICAL POWER SYSTEMS AC SOURCES -
OPERATING TEXT 3.8.2 0
11/18/2002
Title:
ELECTRICAL POWER SYSTEMS AC SOURCES -
SHUTDOWN
'x Page6 of 8
Report Date: 10/28/04 Page 6.
of 8 Report Date: 10/28/04'
SSES MANUAL Manual Name:
TSB2 Manual
Title:
TECHNICAL SPECIFICATIONS BASES UNIT 2 MANUAL TEXT 3.8.3 0
11/18/2002
Title:
ELECTRICAL POWER SYSTEMS DIESEL FUEL OIL, LUBE OIL, AND STARTING AIR TEXT 3.8.4 0
11/18/2002
Title:
ELECTRICAL POWER SYSTEMS DC SOURCES - OPERATING TEXT 3.8.5 0
11/18/2002
Title:
ELECTRICAL POWER SYSTEMS DC SOURCES -
SHUTDOWN TEXT 3.8.6 0
11/18/2002
Title:
ELECTRICAL POWER SYSTEMS BATTERY CELL PARAMETERS TEXT 3.8.7 0
11/18/2002
Title:
ELECTRICAL POWER SYSTEMS DISTRIBUTION SYSTEMS OPERATING TEXT 3.8.8 0
11/18/2002
Title:
ELECTRICAL POWER SYSTEMS DISTRIBUTION SYSTEMS SHUTDOWN TEXT 3.9.1 0
11/18/2002
Title:
REFUELING OPERATIONS REFUELING EQUIPMENT INTERLOCKS TEXT 3.9.2 0
11/18/2002
Title:
REFUELING OPERATIONS. REFUEL POSITION ONE-ROD-OUT INTERLOCK TEXT 3.9.3
Title:
REFUELING TEXT 3.9.4
Title:
REFUELING TEXT 3.9.5
Title:
REFUELING TEXT 3.9.7
Title:
REFUELING 0
11/18/2002 OPERATIONS CONTROL ROD POSITION 0
11/18/2002 OPERATIONS CONTROL ROD POSITION INDICATION 0
11/18/2002 OPERATIONS CONTROL ROD OPERABILITY -
REFUELING 0
11/18/2002 OPERATIONS RESIDUAL HEAT REMOVAL (RHR) - HIGH WATER LEVEL Page7 of 8
Report Date: 10/28/04 Page 7 of 8 Report Date: 10/28/04
SSES MANUAL Manual Name:
TSB2 Manual
Title:
TECHNICAL SPECIFICATIONS BASES UNIT 2 MANUAL TEXT 3.9.8 0
11/18/2002
Title:
REFUELING OPERATIONS RESIDUAL HEAT REMOVAL (RHR) -
LOW WATER LEVEL TEXT 3.10.1 0
11/18/2002
Title:
SPECIAL OPERATIONS INSERVICE LEAK AND HYDROSTATIC TESTING OPERATION TEXT 3.10.2 0
11/18/2002
Title:
SPECIAL OPERATIONS REACTOR MODE SWITCH INTERLOCK TESTING TEXT 3.10.3 0
11/18/2002
Title:
SPECIAL OPERATIONS SINGLE CONTROL ROD WITHDRAWAL - HOT SHUTDOWN TEXT 3.10.4
Title:
SPECIAL TEXT 3.10.5
Title:
SPECIAL 0
.11/18/2002 OPERATIONS SINGLE CONTROL ROD WITHDRAWAL -
COLD SHUTDOWN 0
11/18/2002 OPERATIONS. SINGLE CONTROL ROD DRIVE (CRD)
REMOVAL - REFUELING TEXT 3.10.6
Title:
SPECIAL 0
11/18/2002 MULTIPLE CONTROL ROD WITHDRAWAL - REFUELING OPERATIONS TEXT 3.10.7
Title:
SPECIAL TEXT 3.10.8
Title:
SPECIAL OPERATIONS OPERATIONS 0
11/18/2002 CONTROL ROD TESTING -
0 11/i8/2002 SHUTDOWN MARGIN (SDM)
OPERATING TEST -
REFUELING Page 8 of 8
Report Date: 10/28/04 Page 8 of 8 Report Date: 10/28/04
SUSQUEHANNA STEAM ELECTRIC STATION JIST OF EFFECTIVE SECTIONS (TECHNICAL SPECIFICATIONS BASES)
Section Title Revision TOC Table of Contents 3
B 2.0 SAFETY LIMITS BASES Page TS / B 2.0-1 1
Page TS /IB 2.0-2 2
Page TS /B2.0-3 3
Page TS / B 2.0-4 4
PageTS/B2.0-5 1
Pages B 2.0-6 through B 2.0-8 0
B 3.0 LCO AND SR APPLICABILITY BASES Pages B3.0-1 through B 3.0-7 0
PagesTS/B3.0-8andTS/B3.0-9 1 1 Pages B 3.0-10 through B 3.0-12
\\
0 Pages TS / B 3.0-13 through TS / B 3.01,5 1
B 3.1 REACTIVITY CONTROL BASES Pages B 3.1-1 through-B 3.150
>7 0
Pages TS / B 3.1-6 and TS /
I Pages B 3.1-8 through B 3.1-27 j 0
Page TS/B3.1-28 1
Pages B 3;1-29 through B-3.1-36 0
Page TS / B 3.1-37/ '\\
1 Pages B 3.1-38 tIrouB3.1-51 0
B 3.2 POWER DISTRIBUTION LIMITS BASES Pages TS6AB 3:2-1 through TS / B 3.2-4 1
PagesdTS I B 3.2-5 and.TS / B 3.2-6 2
Page TS PB 3.2-7 1
Page'srTSl B 3.2-8 and TS / B 3.2-9 2
(PSr B 3.2-10 through TS / B 3.2-19 1
B 3.3 INSTRUMENTATION Pages TS I B 3.3-1 through TS / B 3.3-10 1
Page TS / B 3.3-11 2
Pages TS I B 3.3-12 through TS I B 3.3-27 I..
Pages TS I B 3.3-28 through TS / B 3.3-30 2
Page TS / B 3.3-31 1
Pages TS I B 3.3-32 and TS l B 3.3-33 2
Pages TS I B 3.3-34 through TS I B 3.3-54 1
Pages B 3.3-55 through B 3.3-63 0
Pages TS I B 3.3-64 and TS I B 3.3-65 2
Page TS I B 3.3-66 4
SUSQUEHANNA - UNIT 2 TS I B LOES-1 Revision 49
SUSQUEHANNA STEAM ELECTRIC STATION LISTOFEFFEC7TVE SECTIONS (TECHNICAL SPECIFICATIONS BASES)
Section Title Revision Page TS / B 3.3-67 3
Page TS / B 3.3-68 4
Pages TS / B 3.3-69 and TS I B 3.3-70 3
Pages TS / B 3.3-71 through TS / B 3.3-75 2
Page TS I B 3.3-75a 4
Pages TS / B 3.3-75b through TS I B 3.3-75c 3
Pages B 3.3-76 through B 3.3-91 0
Pages TS / B 3.3-92 through TS i B 3.3-103 1
Page TS / B 3.3-104 2
Pages TS / B 3.3-105 and TS I B 3.3-106 1
Page TS I B 3.3-107 2
Page TS i B 3.3-108 1
Page TS/ B 3.3-109 2
Pages TS / B 3.3-110 through TS / B 3.3-115 1
Pages TS / B 3.3-116 through TS / B 3.3-118 2
Pages TS / B 3.3-119 through TS / B 3.3-120 1
Pages TS / B 3.3-121 and TS / B 3.3-122 2
Page TS / B 3.3-123 1
Page TS / B 3.3-124 2
Page TS / B 3.3-124a 0
Pages TS I B 3.3-125 and TS / B 3.3-126 1
Page TS / B 3.3-127 2
Pages TS / B 3.3-128 through TS / B 3.3-131 1
Page TS / B 3.3-132 2
Pages TS / B 3.3-133 and TS / B 3.3-134 1
Pages B 3.3-135 through B 3.3-137 0
Page TS / B 3.3-138 1
Pages B 3.3-139 through B 3.3-149 0
Pages TS/ B 3.3-150 through TS / B 3.3-162 1
Page TS / B 3.3-163 2
Pages TS / B 3.3-164 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 1
Pages TS / B 3.3-180 through TS / B 3.3-191 1
Pages B 3.3-192 through B 3.3-205 0
Page TS / B 3.3-206 1.
Pages B 3.3-207 through B 3.3-220 0
B 3.4 REACTOR COOLANT SYSTEM BASES Pages TS / B 3.4-1 and TS I B 3.4-2 1
Pages TS / B 3.4-3 through TS I B 3.4-6 2
Page TS / B 3.4-7.
1 Pages TS / B 3.4-8 and TS i B 3.4-9 2
SUSQUEHANNA - UNIT 2 TS/IB LOES-2 Revision 49
SUSQUEHANNA STEAM ELECTRIC STATION LST OF EFFECTIVE SECTIONS (TECHNICAL SPECIFICATIONS BASES)
Section Title Pages B 3.4-10 through B 3.4-14 Page TS l B 3.4-15 Pages TS I B 3.4-16 and TS I B 3.4-17 Page TS l B 3.4-18 Pages B 3.4-19 through B 3.4-28 Page TS I B 3.4-29 Pages B 3.3-30 through B 3.3-48 Page TS I B 3.4-49 Page TS I B 3.4-50 Page TS I B 3.4-51 Pages TS I B 3.4-52 and TS I B 3.4-53 Pages TS 1 B 3.4-54 and TS I B 3.4-55 Pages TS / B 3.4-56 through TS I B 3.4-60 Revision
- 0 1
2
.1-0 1
0 2
1 2
1 2
1 B 3.5 ECCS AND RCIC BASES' Pages TS l B 3.5-1 and TS I B 3.5-2 Page TS I B 3.5-3 Pages TS I B 3.5-4 through TS I B 3.5-10 Page TS I B 3.5-11 Pages TS I B 3.5-12 through TS I B 3.514 Pages TS I B 3.5-15 through TS I B.3.5-17 PageTS/B 3.18 Pages B 3.5-19 through B 3.5-24 Page TS I B 3.5-25 Pages B 3.5-26 through B 3.5-31 I
2 1
2 I
2
.1 0
.1 0
B 3.6 CONTAINMENT SYSTEMS BASES Page TS I B 3.6-1 Page TS I B 3.6-1 a Pages TS I B 3.6-2 through TS 1 B 3.6-5 Page TS I B 3.6-6 Pages TS I B 3.6-6a and TS I B 3.6-6b Page TS I B 3.6-6c Pages B 3.6-7 through B 3.6-14 Page TS I B 3.6-15 Pages TS I B 3.6-15a and TS I B 3.6-1 5b Page TS I B 3.6-16 Page TS / B 3.6-17 Page TS / B 3.6-17a Pages TS / B 3.6-18 and TS / B 3.6-19 Page TS I B 3.6-20 Page TS I B 3.6-21 Pages TS I B 3.6-21 a and TS I B 3.6-21 b 2
3 2
3 2
0 0
3 0
1 2
0 2
3 0
SUSQUEHANNA
- UNIT 2 TS I B LOES-3 Revision 49 SUSQUEHANNA - UNIT 2 TS I B LOES-3 Revision 49
SUSQUEHANNA STEAM ELECTRIC STATION LIST OF EFFECTIVE SECTIONS (TECHNICAL SPECIFICATIONS BASES)
Section Title Revision.
Pages TS / B 3.6-22 and TS I B 3.6-23 2
Pages TS / B 3.6-24 through TS / B 3.6-26 1
Page TS / B 3.6-27 3
Page TS / B 3.6 6
Page TS / B 3.6-29 3
Page TS / B 3.6-29a 0
Page TS / B 3.6-30 2
Page TS / B 3.6-31 3
Pages TS I B 3.6-32 through TS I B 3.6-34 1
Pages TS / B 3.6-35 through TS /B 3.6-37 2
Page TS / B 3.6-38 1
Page TS / B 3.6-39 4
Pages B 3.6-40 through B 3.6-42 0
Pages TS / B 3.6-43 through TS / B 3.6-50 1
Page TS / B 3.6-51
-2 Pages B 3.6-52 through B 3.6-62 0
Page TS / B 3.6-63 1
Pages B 3.6-64 through B 3.6-82 0
Page TS / B 3.6-83 2
Pages TS I B 3.6-84 through TS I B 3.6-87 1
Page TS / B 3.6-87a 1
Page TS / B 3.6-88 2
PagesTS / B 3.6-89 through TS./ B 3.6-99 1
Page B 3.6-100 0
Pages TS / B 3.6-101 through TS I B 3.6-106 1
B 3.7 PLANT SYSTEMS BASES Pages TS I B 3.7-1 through TS / B 3.7-6 2
Page TS / B 3.7-6a 2
Pages TS I B 3.7-6b and TS / B 3.7-6c 0
Pages TS I B 3.7-7 and TS / B 3.7-8 1
Pages B 3.7-9 through B 3.7-11 0
Pages TS / B 3.7-12 and TS / B 3.7-13 1
Pages TS / B 3.7-14 through TS / B 3.7-18 2
Page TS / B 3.7-18a 0
Pages TS / B 3.7-19 through TS / B 3.7-26 1
Pages B 3.7-24 through B 3.7-26 0
Pages TS / 3.7-27 through TS / B 3.7-29 1
Pages B 3.7-30 through B 3.7-33 0
B 3:8 ELECTRICAL POWER SYSTEMS BASES Pages B 3.8-1 through B 3.84 0
Page TS / B 3.8-5 1
SUSQUEHANNA-UNIT2 TSIB LOES-4 Revision 49 SUSQUEHANNA - UNIT 2 TS /B LOES-4 Revision 49
SUSQUEHANNA STEAM ELECTRIC STATION LIST OF EFFECTIVE SECTIONS (TECHNICAL SPECIFICATIONS BASES)
Section Title Revision Pages B 3.8-6 through B 3.8-8 0
Pages TS / B 3.8-9 through TS I B 3.8-11 1
Pages B 3.8-12 through B 3.8-18 0
Page TS I B 3.8-19 1
Pages B 3.8-20 through B 3.8-22 0
Page TS / B 3.8-23 1
Page B 3.8-24 0
Pages TS I B 3.8-25 and TS / B 3.8-26 1
Pages B 3.8-27 through B 3.8-37 0
Page TS / B 3.8-38 Pages TS / B 3.8-39 through TS / B 3.8-55 0
Pages TS / B 3.8-56 through TS / B 3.8-64 1
Page TS / B 3.8-65 2
Page TS / B 3.8-66 2
Pages TS I B 3.8-67 through TS i B 3.8-68 1
Page TS / B 3.8-69 2
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B 3.9 REFUELING OPERATIONS BASES Pages TS / B 3.9-1 and TS I B 3.9-2 1
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TSB2 text LOES 10114M04 SUSQUEHANNA - UNIT 2 TS / B LOES-5 Revision 49
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PPL Rev. I Reactor Core SLs B 2.1.1 B 2.0 SAFETY LIMITS (SLs)
B 2.1.1 Reactor Core SLs BASES BACKGROUND GDC 10 (Ref. 1) requires, and SLs ensure, that specified acceptable fuel design limits are not exceeded during steady state operation, normal operational transients, and anticipated operational occurrences (AOOs).
The fuel cladding integrity SL is set such that no significant fuel damage is calculated to occur if the limit is not violated. Because fuel damage is not directly observable, a stepback approach is used to establish an SL, such that the MCPR is not less than the limit specified in Specification 2.1.1;2 for Siemens Power Corporation fuel. MCPR greater than the specified limit represents a conservative margin relative to the conditions required to maintain fuel cladding integrity.
The fuel cladding is one of the physical barriers that separate the radioactive materials from the environs. The integrity of this cladding barrier is related to its relative freedom from perforations or cracking.
Although some corrosion or use related cracking may occur during the life of the cladding, fission product migration from this source is incrementally cumulative and continuously measurable. Fuel cladding perforations, however, can result from thermal stresses, which occur from reactor operation significantly above design conditions.
While fission product migration from cladding perforation is just as measurable as that from use related cracking, the thermally caused cladding perforations signal a threshold beyond which still greater thermal stresses may cause gross, rather than incremental; cladding deterioration.
Therefore, the fuel cladding SL is defined with a margin to the conditions that would produce onset of transition boiling (i.e., MCPR = 1.00). These conditions represent a significant departure from the condition intended by design for planned operation. The MCPR fuel cladding integrity SL ensures that during normal operation and during A0Os, at least 99.9% of' the fuel rods in the core do not experience transition boiling.
Operation above the boundary of the nucleate boiling regime could result in excessive cladding temperature because of the onset of transition boiling
-and the resultant sharp reduction in heat transfer coefficient. Inside the steam film, high cladding temperatures are reached, and a cladding water (zirconium water) reaction may take place. This chemical reaction results in oxidation of the fuel cladding to a structurally weakerform. This weaker form may lose its integrity, resulting in an uncontrolled release of activity to the reactor coolant (continued)
SUSQUEHANNA - UNIT 2 TS / B2.0-1 Revision I
PPL Rev. 1 Reactor Core SLs B 2.1.1 BASES APPLICABLE The fuel cladding must not sustain damage as a result of normal operation SAFETY and AOOs. The reactor core SLs are established to preclude violation of ANALYSES the fuel design criterion that an MCPR limit is to be established, such that at least 99.9% of the fuel rods in the core would not be expected to experience the onset of transition boiling.
The Reactor Protection System setpoints (LCO 3.3.1.1, "Reactor Protection System (RPS) Instrumentation"), in combination with the other LCOs, are designed to prevent any anticipated combination of transient conditions for Reactor Coolant System water level, pressure, and THERMAL POWER level that would result in reaching the MCPR limit 2.1.1.1 Fuel Cladding Integrity The use of the ANFB-10 (Reference 4) correlation is valid for critical power calculations at pressures > 571 psia and bundle mass fluxes > 0.115 x 106 lb/hr-ft2 for ANFB-10. For operation at low pressures or low flows, the fuel cladding integrity SL is established by a limiting condition on core THERMAL POWER, with the following basis:
Provided that the water level in the vessel downcomer is maintained above the top of the active fuel, natural circulation is sufficient to ensure a minimum bundle flow for all fuel assemblies that have a relatively high power and potentially can approach a critical heat flux condition. For the SPC Atrium 10 design, the minimum bundle flow is >28 x 10 Ib/hr. For Atrium-10 fuel design, the coolant minimum
-bundle flow and maximum area are such that the mass flux is always>.25 x 10 Ib/hr-ft. Full scale critical power test data taken from various SPC and GE fuel designs at pressures from 14.7 psia to 1400 psia indicate the fuel assembly critical power at 0.25 x 10 lb/hr-ft is approximately 3.35 MWt. At 25% RTP, a bundle power of approximately 3.35 MWt corresponds to a bundle radial peaking factor of approximately 3.0, which is significantly higher than the expected peaking factor. Thus, a THERMAL POWER limit of 25% RTP for reactor pressures < 785 psig is conservative.
2.1.1.2 MCPR The MCPR SL ensures sufficient conservatism in the operating MCPR limit that, in the event of an AOO from the limiting condition of operation, at least 99.9% of the fuel rods in the core would be expected to avoid boiling transition. The margin between calculated boiling transition (i.e.,
MCPR = 1.00) and the MCPR SL is based on a detailed statistical procedure (continued)
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PPL Rev. 1 Reactor Core SLs B 2.1.1 BASES APPLICABLE 2.1.1.2 MCPR (continued)
SAFETY ANALYSES that considers the uncertainties in monitoring the core operating state. One specific uncertainty included in the SL is the uncertainty in the critical power correlation. References 2, 4 and 5 describe the methodology used in determining the MCPR SL The ANFB-10 critical power correlation is based on a significant body of practical test data. As long as the core pressure and flow are within the range of validity of the correlation (refer to Section B 2.1.1.1), the assumed reactor conditions used in defining the SL introduce conservatism into the limit because bounding high radial power factors and bounding flat local peaking distributions are used to estimate the number of rods in boiling transition. These conservatisms and the inherent accuracy of the ANFB-10 correlation provide a reasonable degree of assurance that during sustained operation at the MCPR SL there would be no transition boiling in the core.
If boiling transition were to occur, there is reason to believe that the integrity of the fuel would not be compromised.
Significant test data accumulated by the NRC and private organizations indicate that the use of a boiling transition limitation to protect against cladding failure is a very conservative approach. Much of the data indicate that BWR fuel can survive for an extended period of time in an environment of boiling transition.
SPC ATRIUM-10 fuel is monitored using the ANFB-10 Critical Power Correlation. The effects of channel bow on MCPR are explicitly included in the calculation of the MCPR SL. Explicit treatment of channel bow in the MCPR SL addresses the concerns of the NRC Bulletin No. 9D-02 entitled "Loss of Thermal Margin Caused by Channel Box Bow."
Monitoring required for compliance with the MCPR SL is specified in LCO 3.2.2, Minimum Critical Power Ratio.
2.1.1.3 Reactor Vessel Water Level During MODES 1 and 2 the reactor vessel water level is required to be above the top of the active fuel to provide core cooling capability. With fuel in the reactor vessel during periods when the reactor is §hut down, consideration must be given to water level requirements due to the effect of decay heat. If the water level should drop below the top of the active irradiated fuel during this period, the ability to remove decay heat is reduced. This reduction in cooling capability could lead to elevated cladding temperatures and clad perforation in the event that the water level becomes < 2/3 of the core height.
(continued)
SUSQUEHANNA - UNIT.2 TS/ B2.0 Revision 3
PPL Rev. 1 Reactor Core SLs B 2.1.1 BASES APPLICABLE 2.1.1.3 Reactor Vessel Water Level (continued)
SAFETY ANALYSES The reactor vessel water level SL has been established at the top of the active irradiated fuel to provide a point that can'66 monitored and to also provide adequate margin for effective action.
SAFETY LIMITS The reactor core SLs are established to protect the integrity of the fuel clad barrier to the release of radioactive materials to the environs. SL 2.1.1.1 and SL 2.1.1.2 ensure that the core operates within the fuel design criteria.
SL 2.1.1.3 ensures that the reactor vessel water level is greater than the top of the active irradiated fuel in order to prevent elevated clad temperatures and resultant clad perforations.
APPLICABILITY SLs 2.1.1.1, 2.1.1.2, and 2.1.1.3 are applicable in all MODES.
SAFETY LIMIT Exceeding an SL may cause fuel damage and create a potential for.
VIOLATIONS radioactive releases in excess of 10 CFR 100, "Reactor Site Criteria," limits (Ref. 3). Therefore, it is required to insert all insertable control rods and restore compliance with the SLs within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. The 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> Completion Time ensures that the operators take prompt remedial action and also ensures that the probability of an accident occurring during this period is minimal.
REFERENCES
- 1.
10 CFR 50, Appendix A, GDC 10.
- 2.
ANFB 524 (P)(A), Revision 2, "Critical Power Methodology for Boiling Water Reactors," Supplement I Revision 2 and Supplement 2, November 1990.
- 3.
- 4.
EMF-1997(P)(A), Revision 0, "ANFB-10 Critical Power Correlation,"
July 1998 and EMF-1997(P)(A) Supplement 1 Revision 0," ANFB-10 Critical Power Correlation: High Local Peaking Results," July 1998.
- 5.
EMF-2158(P)(A), Rev. 0, "Siemens Power Corporaton Methodology for Boiling Water Reactors: Evaluation and Validation of CASMO-4 / MICROBURN-B2," October 1999..
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