ML15034A311
| ML15034A311 | |
| Person / Time | |
|---|---|
| Site: | Susquehanna  |
| Issue date: | 01/23/2015 |
| From: | Susquehanna |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| 028401, 2015-3019 | |
| Download: ML15034A311 (25) | |
Text
Jan.
23, 2015 Page 1
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MANUAL HARD COPY DISTRIBUTION DOCUMENT TRANSMITTAL 2015-3019 USER INFORMATION:
GERLACH*ROSEY M EMPL#:028401 CA#:
0363 Address: NUCSA2 Phone#:
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TO:
GERLACH*ROSEY M 01/23/2015 LOCATION:
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TSB2 -
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DOCUMENTS TYPE: TSB2 AoO(
Jan. 23, 2015 Page 2
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REV:4 ADD:
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- USERS, ELECTRONICALLY REVIEW THE APPROPRIATE DOCUMENTS AND ACKNOWLEDGE COMPLETE IN YOUR NIMS INBOX.
SSES MANUJAL Manual Name:
TSB2 Manual
Title:
TECHNICAL SPECIFICATIONS BASES UNIT 2 MANUAL Table Of Contents Issue Date:
01/22/2015 Procedure Name Rev TEXT LOES 124
Title:
LIST OF EFFECTIVE SECTIONS Issue Date 01/22/2015 Change ID Change Number TEXT TOC
Title:
TABLE OF CONTENTS 22 07/02/2014 TEXT 2.1.1 5
01/22/2015
Title:
SAFETY LIMITS (SLS)
REACTOR CORE SLS TEXT 2.1.2 1
Title:
SAFETY LIMITS (SLS)
REACTOR 10/04/2007 COOLANT SYSTEM (RCS)
PRESSURE SL TEXT 3. 0 3
08/20/2009
Title:
LIMITING CONDITION FOR OPERATION (LCO)
APPLICABILITY TEXT 3. 1.1
Title:
REACTIVITY TEXT 3.1.2
Title:
REACTIVITY TEXT 3.1.3
Title:
REACTIVITY TEXT 3.1.4
Title:
REACTIVITY TEXT 3.1.5
Title:
REACTIVITY TEXT 3.1.6
Title:
REACTIVITY 1
CONTROL SYSTEMS 03/24/2005 SHUTDOWN MARGIN (SDM) 0 11/18/2002 CONTROL SYSTEMS REACTIVITY ANOMALIES 2
01/19/2009 CONTROL SYSTEMS CONTROL ROD OPERABILITY 4
01/30/2009 CONTROL SYSTEMS CONTROL ROD SCRAM TIMES 1
07/06/2005 CONTROL SYSTEMS CONTROL ROD SCRAM ACCUMULATORS 3
02/24/2014 CONTROL SYSTEMS ROD PATTERN CONTROL Pagel of 8
Report Date: 01/22/15 Page 1 of
-8 Report Date: 01/22/15
SSES MANUAL Manual Name:
TSB2 Manual
Title:
TECHNICAL SPECIFICATIONS BASES UNIT 2 MANUAL TEXT 3.1.7 3
10/04/2007
Title:
REACTIVITY CONTROL SYSTEMS STANDBY LIQUID CONTROL (SLC)
SYSTEM TEXT 3.1.8 3
05/06/2009
Title:
REACTIVITY CONTROL SYSTEMS SCRAM DISCHARGE VOLUME (SDV)
VENT AND DRAIN VALVES TEXT 3.2.1 4
05/06/2009
Title:
POWER DISTRIBUTION LIMITS AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)
TEXT 3.2.2 3
05/06/2009
Title:
POWER DISTRIBUTION LIMITS MINIMUM CRITICAL POWER RATIO (MCPR)
TEXT 3.2.3 2
05/06/2009
Title:
POWER DISTRIBUTION LIMITS LINEAR HEAT GENERATION RATE LHGR TEXT 3.3.1.1 5
02/24/2014
Title:
INSTRUMENTATION REACTOR PROTECTION SYSTEM (RPS)
INSTRUMENTATION TEXT 3.3.1.2 2
01/19/2009
Title:
INSTRUMENTATION SOURCE RANGE MONITOR (SRM)
INSTRUMENTATION TEXT 3.3.2.1 3
02/24/2014
Title:
INSTRUMENTATION CONTROL ROD BLOCK INSTRUMENTATION TEXT 3.3.2.2 2
02/22/2012
Title:
INSTRUMENTATION FEEDWATER -
MAIN TURBINE HIGH WATER LEVEL TRIP INSTRUMENTATION TEXT 3.3.3.1 8
02/28/2013
Title:
INSTRUMENTATION POST ACCIDENT MONITORING (PAM)
INSTRUMENTATION TEXT 3.3.3.2 1
04/18/2005
Title:
INSTRUMENTATION REMOTE SHUTDOWN SYSTEM TEXT 3.3.4.1 1
05/06/2009
Title:
INSTRUMENTATION END OF CYCLE RECIRCULATION PUMP TRIP (EOC-RPT)
INSTRUMENTATION Page 2 of 8
Report Date: 01/22/15
SSES MANUJAL Manual Name:
TSB2 Manual
Title:
TECHNICAL SPECIFICATIONS BASES UNIT 2 MANUAL TEXT 3.3.4.2 0
11/18/2002
Title:
INSTRUMENTATION ANTICIPATED TRANSIENT WITHOUT SCRAM RECIRCULATION PUMP TRIP (ATWS-RPT)
INSTRUMENTATION TEXT 3.3.5.1
Title:
INSTRUMENTATION TEXT 3.3.5.2
Title:
INSTRUMENTATION 5
02/24/2014 EMERGENCY CORE COOLING SYSTEM (ECCS)
INSTRUMENTATION 0
11/18/2002 REACTOR CORE ISOLATION COOLING (RCIC)
SYSTEM INSTRUMENTATION TEXT 3.3.6.1 7
03/31/2014
Title:
INSTRUMENTATION PRIMARY CONTAINMENT ISOLATION INSTRUMENTATION TEXT 3.3.6.2 4
09/01/2010
Title:
INSTRUMENTATION SECONDARY CONTAINMENT ISOLATION INSTRUMENTATION TEXT 3.3.7.1
Title:
INSTRUMENTATION INSTRUMENTATION 2
10/27/2008 CONTROL ROOM EMERGENCY OUTSIDE AIR SUPPLY (CREOAS)
SYSTEM TEXT 3.3.8.1 3
12/17/2007
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 4
07/20/2010
Title:
RECIRCULATION LOOPS OPERATING TEXT 3.4.2 3
10/23/2013
Title:
JET PUMPS TEXT 3.4.3 3
01/13/2012
Title:
REACTOR COOLANT SYSTEM (RCS) SAFETY/RELIEF VALVES (S/RVS)
TEXT 3.4.4 0
11/18/2002
Title:
RCS OPERATIONAL LEAKAGE Page3 of.8 Report Date: 01/22/15
SSES MANUAL Manual Name:
TSB2 Manual
Title:
TECHNICAL SPECIFICATIONS BASES UNIT 2 MANUAL TEXT 3. 4.5
Title:
SYSTEM (RCS) 03/10/2010 RCS PRESSURE ISOLATION VALVE (PIV)
LEAKAGE TEXT 3.4.6 4
02/19/2014
Title:
RCS LEAKAGE DETECTION INSTRUMENTATION TEXT 3.4.7 2
10/04/2007
Title:
RCS SPECIFIC ACTIVITY TEXT 3.4.8
Title:
REACTOR COOLANT HOT SHUTDOWN TEXT 3.4.9
Title:
REACTOR COOLANT COLD SHUTDOWN 2
SYSTEM (RCS) 1 SYSTEM (RCS) 03/28/2013 RESIDUAL HEAT REMOVAL (RHR)
SHUTDOWN COOLING SYSTEM 03/28/2013 RESIDUAL HEAT REMOVAL (RHR)
SHUTDOWN COOLING SYSTEM TEXT 3.4.10 3
05/06/2009
Title:
RCS PRESSURE AND TEMPERATURE (P/T) LIMITS TEXT 3.4.11 0
11/18/2002
Title:
REACTOR STEAM DOME PRESSURE TEXT 3.5.1 4
07/16/2014
Title:
EMERGENCY CORE COOLING SYSTEMS (ECCS)
AND REACTOR SYSTEM ECCS -
OPERATING TEXT 3.5.2 1
02/24/2014
Title:
EMERGENCY CORE COOLING SYSTEMS (ECCS)
AND REACTOR SYSTEM ECCS -
SHUTDOWN TEXT 3.5.3 3
02/24/2014
Title:
EMERGENCY CORE COOLING SYSTEMS (ECCS)
AND REACTOR SYSTEM RCIC SYSTEM CORE ISOLATION COOLING (RCIC)
CORE ISOLATION COOLING (RCIC)
CORE ISOLATION COOLING (RCIC)
TEXT 3.6.1.1
Title:
02/24/2014 TEXT 3.6.1.2 1
05/06/2009
Title:
CONTAINMENT SYSTEMS PRIMARY CONTAINMENT AIR LOCK
~age4 of 8
Report Date: 01/22/15 Page 4 of 8
Report Date: 01/22/15
SSES MANUAL Manual Name:
TSB2 Manual
Title:
TECHNICAL SPECIFICATIONS BASES UNIT 2 MANUAL 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 TEXT 3.6.2.1
Title:
CONTAINMENT TEXT 3.6.2.2
Title:
CONTAINMENT TEXT 3.6.2.3
Title:
CONTAINMENT TEXT 3.6.2.4
Title:
CONTAINMENT TEXT 3.6.3.1
Title:
CONTAINMENT TEXT 3.6.3.2
Title:
CONTAINMENT TEXT 3.6.3.3
Title:
CONTAINMENT TEXT 3.6.4.1
Title:
CONTAINMENT 14 07/02/2014 SYSTEMS PRIMARY CONTAINMENT ISOLATION VALVES (PCIVS) 1 05/06/2009 SYSTEMS CONTAINMENT PRESSURE 1
10/05/2005 DRYWELL AIR TEMPERATURE SYSTEMS I
0 11/18/2002 SYSTEMS SUPPRESSION CHAMBER-TO-DRYWELL VACUUM BREAKERS 2
12/17/2007 SYSTEMS SUPPRESSION POOL AVERAGE TEMPERATURE 0
11/18/2002 SYSTEMS SUPPRESSION POOL WATER LEVEL 1
01/16/2006 SYSTEMS RESIDUAL HEAT REMOVAL (RHR)
SUPPRESSION POOL COOLING 0
11/18/2002 SYSTEMS RESIDUAL HEAT REMOVAL (RHR)
SUPPRESSION POOL SPRAY 2
06/13/2006 SYSTEMS PRIMARY CONTAINMENT HYDROGEN RECOMBINERS 1
04/18/2005 SYSTEMS DRYWELL AIR FLOW SYSTEM 1
02/28/2013 SYSTEMS PRIMARY CONTAINMENT OXYGEN CONCENTRATION 11 11/06/2014 SYSTEMS SECONDARY CONTAINMENT Page5 of 8
Report Date: 01/22/15 Page 5 of
.8 Report Date: 01/22/15
SSES MANUJAL Manual Name:
TSB2 Manual
Title:
TECHNICAL SPECIFICATIONS BASES UNIT 2 MANUAL TEXT 3.6.4.2 9
04/25/2014 Title; CONTAINMENT SYSTEMS SECONDARY CONTAINMENT ISOLATION VALVES (SCIVS)
TEXT 3.6,4.3 4
09/21/2006
Title:
CONTAINMENT SYSTEMS STANDBY GAS TREATMENT (SGT)
SYSTEM TEXT 3.7,1
Title:
PLANT SYSTEMS ULTIMATE HEAT
- 5.
04/27/2012 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 3.7.8
Title:
MAINE TEXT 3.8.1 2
05/02/2008 SYSTEMS EMERGENCY SERVICE WATER (ESW)
SYSTEM 1
01/08/2010 SYSTEMS CONTROL ROOM EMERGENCY OUTSIDE AIR SUPPLY (CREOAS)
SYSTEM 0
11/18/2002 SYSTEMS CONTROL ROOM FLOOR COOLING SYSTEM 1
10/04/2007 SYSTEMS MAIN CONDENSER OFFGAS 3
01/25/2011 SYSTEMS MAIN TURBINE BYPASS SYSTEM 1
10/04/2007 SYSTEMS SPENT FUEL STORAGE POOL WATER LEVEL 0
05/06/2009 TURBINE PRESSURE REGULATION SYSTEM 9
02/24/2014
Title:
ELECTRICAL POWER SYSTEMS AC SOURCES -
OPERATING TEXT 3.8.2 0
11/18/2002
Title:
ELECTRICAL POWER SYSTEMS AC SOURCES -
SHUTDOWN Page6 of 8
Report Date: 01/22/15 Page 6 of 8
Report Date: 01/22/15
SSES MANUAL Manual Name: TSB2 WManual
Title:
TECHNICAL SPECIFICATIONS BASES UNIT 2 MANUAL TEXT 3. 8.3 4
10/23/2013
Title:
ELECTRICAL POWER SYSTEMS DIESEL FUEL OIL LUBE OIL AND STARTING AIR 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 TEXT 3.9.2
Title:
REFUELING TEXT 3.9.3
Title:
REFUELING TEXT 3.9.4
Title:
REFUELING TEXT 3.9.5
Title:
REFUELING TEXT 3.9.6
Title:
REFUELING 3
01/19/2009 POWER SYSTEMS DC SOURCES -
OPERATING 1
12/14/2006 POWER SYSTEMS DC SOURCES -
SHUTDOWN 1
12/14/2006 POWER SYSTEMS BATTERY CELL PARAMETERS 4
10/05/2005 POWER SYSTEMS DISTRIBUTION SYSTEMS -
OPERATING 0
11/18/2002 POWER SYSTEMS DISTRIBUTION SYSTEMS -
SHUTDOWN 0
11/18/2002 OPERATIONS REFUELING EQUIPMENT INTERLOCKS 1
09/01/2010 OPERATIONS REFUEL POSITION ONE-ROD-OUT INTERLOCK 0
11/18/2002 OPERATIONS CONTROL ROD POSITION 0
11/18/2002 OPERATIONS CONTROL ROD POSITION INDICATION 0
11/18/2002 CONTROL ROD OPERABILITY OPERATIONS REFUELING 1
10/04/2007 REACTOR PRESSURE VESSEL (RPV)
WATER LEVEL OPERATIONS Page 7 of
.8 Report Date: 01/22/15
SSES MANUAL Manual Name:
TSB2 Manual
Title:
TECHNICAL SPECIFICATIONS BASES UNIT 2 MANUAL 0
TEXT 3.9.7 0
11/18/2002
Title:
REFUELING OPERATIONS RESIDUAL HEAT REMOVAL (RHR)
TEXT 3.9.8 0
11/18/2002
Title:
REFUELING OPERATIONS RESIDUAL HEAT REMOVAL (RHR)
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 3.10.4
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 1
01/23/2008 OPERATIONS INSERVICE LEAK AND HYDROSTATIC TESTING OPERATION 0
11/18/2002 OPERATIONS REACTOR MODE SWITCH INTERLOCK TESTING 0
11/18/2002 OPERATIONS SINGLE CONTROL ROD WITHDRAWAL -
HOT SHUTDOWN 0
11/18/2002 OPERATIONS SINGLE CONTROL ROD WITHDRAWAL -
COLD SHUTDOWN 0
11/18/2002 OPERATIONS SINGLE CONTROL ROD DRIVE (CRD)
REMOVAL -
REFUELING 0
11/18/2002 OPERATIONS MULTIPLE CONTROL ROD WITHDRAWAL -
REFUELING 1
03/24/2005 OPERATIONS CONTROL ROD TESTING -
OPERATING 2
04/09/2007 OPERATIONS SHUTDOWN MARGIN (SDM)
TEST -
REFUELING Page 8 of 8
Report Date: 01/22/15 Page -8 of
_8 Report Date: 01/22/15
SUSQUEHANNA STEAM ELECTRIC STATION LIST OF EFFECTIVE SECTIONS (TECHNICAL SPECIFICATIONS BASES)
Section Title Revision TOC Table of Contents 22 B 2.0 SAFETY LIMITS BASES Page TS / B 2.0-1 2
Pages TS / B 2.0-2 and TS / B 2.0-3 5
Page TS / B 2.0-4 7
Pages TS / B 2.0-5 through TS / B 2.0-8 1
B 3.0 LCO AND SR APPLICABILITY BASES Page TS / B 3.0-1 1
Pages TS / B 3.0-2 through TS / B 3.0-4 0
Pages TS / B 3.0-5 through TS / B 3.0-7 1
Page TS / B 3.0-8 3
Pages TS / B 3.0-9 through Page TS / B 3.0-11 2
Page TS / B 3.0-11a 0
Page TS / B 3.0-12 1
Pages TS / B 3.0-13 through TS / B 3.0-15 2
Pages TS / B 3.0-16 and TS / B 3.0-17 0
B 3.1 REACTIVITY CONTROL BASES Pages B 3.1-1 through B 3.1-4' 0
Page TS / B 3.1-5 1
Pages TS / B 3.1-6"and TS /B 2
Pages B 3.1-8 through B 0
Page TS/B 3.1 -13 0
Page TS / B 3 '-
1 Page TS / B 4f'i 0
PageTS/3B. -1 1
Pages TS / B 3. -7 through TS / B 3.1-19 0
Pages TS / B 3.1-20. and TS / B 3.1-21 1
Page TS / B 3.1-22 0
Page TS / B 3.1-23 1
Page TS / B 3.1-24 0
Pages TS / B 3.1-25 through TS / B 3.1-27 1
Page TS / B 3.1-28 2
Page TS / B 3.1-29 1
Pages TS / B 3.1-30 through TS / B 3.1-33 0
Pages TS / B 3.1.34 through TS / B 3.1-36 1
Page TS / B 3.1-37 2
Page TS / B 3.1-38 3
Pages TS / B 3.1-39 and TS / B 3.1-40 2
Page TS / B 3.1-40a 0
Page TS / B 3.1-41 1
Page TS / B 3.1-42 2
SUSQUEHANNA
- UNIT 2 TS / B LOES-1 Revision 124 SUSQUEHANNA - UNIT 2 TS / B LOES-1 Revision 124
SUSQUEHANNA STEAM ELECTRIC STATION LISTOFEFFECTIVE SECTIONS (TECHNICAL SPECIFICATIONS BASES)
Section Title Revision Pages TS / B 3.1-43 1
Page TS / B 3.1-44 0
Page TS / B 3.1-45 3
Page TS / B 3.1-46 through TS / B 3.1-49 1
Page TS / B 3.1-50 0
Page TS / B 3.1-51 3
B 3.2 POWER DISTRIBUTION LIMITS BASES Pages TS / B 3.2-1 and TS / B 3.2-2 2
Page TS / B 3.2-3 4
Page TS / B 3.2-4 1
Page TS / B 3.2-5 3
Page TS / B 3.2-6 4
Page TS / B 3.2-7 3
Pages TS / B 3.2-8 and TS / B 3.2-9 4
Pages TS / B 3.2-10 through TS / B 3.2-12 2
Page TS / B 3.2-13 1
B 3.3 INSTRUMENTATION Pages TS / B 3.3-1 through TS / B 3.3-4 1
Page TS / B 3.3-5 2
Page TS / B 3.3-6
.1 Page TS / B 3.3-7 3
Page TS / B 3.3-8 4
Pages TS / B 3.3-9 through TS / B 3.3-13 3
Page TS / B 3.3-14 4
Pages TS / B 3.3-15 and TS / B 3.3-16 2
Pages TS / B 3.3-17 through TS / B 3.3-21 3
Pages TS / B 3.3-22 through TS / B 3.3-27 2
Page TS / B 3.3-28 3
Page TS / B 3.3-29 4
Pages TS / B 3.3-30 and TS / B 3.3-31 3
Pages TS / B 3.3-32 and TS / B 3.3-33 4
Page TS / B 3.3-34 2
Pages TS / B 3.3-34a and TS / B 3.3-34b 1
Pages TS / B 3.3.34c and TS / B 3.3-34d 0
Page TS / B 3.3-34e 1
Pages TS / B 3.3-34f through TS / B 3.3-34i 0
Pages TS / B 3.3-35 and TS / B 3.3-36 2
Pages TS / B 3.3-37 and TS / B 3.3-38 1
SUSQUEHANNA
- UNIT 2 TS / B LOES-2 Revision 124 SUSQUEHANNA - UNIT 2 TS / B LOES-2 Revision 124
SUSQUEHANNA STEAM ELECTRIC STATION LIST OF EFFECTIVE SECTIONS (TECHNICAL SPECIFICATIONS BASES)
Section Title Revision Page TS / B 3.3-39 2
Pages TS / B 3.3-40 through TS / B 3.3-43 1
Pages TS / B 3.3-44 through TS / B 3.3-54 3
Pages TS / B 3.3-54a through TS / B 3.3-54d 0
Page TS / B 3.3.54e 1
Page TS / B 3.3-55 2
Page TS / B 3.3-56 0
Page TS / B 3.3-57 1
Page TS / B 3.3-58 0
Page TS / B 3.3-59 1
Page TS / B 3.3-60 0
Page TS / B 3.3-61 1
Pages TS / B 3.3-62 and TS / B 3.3-63 0
Pages TS / B 3.3-64 and TS / B 3.3-65 2
Page TS / B 3.3-66 4
Page TS / B 3.3-67 3
Page TS / B 3.3-68 4
Page TS I B 3.3.69 5
Page TS / B 3.3-70 4
Page TS / B 3.3-71 3
Pages TS / B 3.3-72 and TS / B 3.3-73 2
Page TS / B 3.3-74 3
Page TS / B 3.3-75 2
Pages TS / B 3.3-75a through TS / B 3.3-75c 6
Pages TS / B 3.3-76 and TS / B 3.3-77 0
Page TS / B 3.3-78 1
Pages TS / B 3.3-79 through TS / B 3.3-81 0
Page TS / B 3.3-82 1
Page TS / B 3.3-83 0
Pages TS / B 3.3-84 and TS / B 3.3-85 1
Page TS / B 3.3-86 0
Page TS / B 3.3-87 1
Page TS / B 3.3-88 0
Page TS / B 3.3-89 1
Pages TS / B 3.3-90 and TS / B 3.3-91 0
Pages TS / B 3.3-92 through TS / B 3.3-103 1
Page TS / B 3.3-104 3
Pages TS / B 3.3-105 and TS / B 3.3-106 1
Page TS / B 3.3-107 2
Page TS / B 3.3-108 1
Page TS / B 3.3-109 2
Pages TS / B 3.3-110 through TS / B 3.3-112 1
Page TS / B 3.3-113 2
SUSQUEHANNA - UNIT 2 TS / B LOES-3 Revision 124
SUSQUEHANNA STEAM ELECTRIC STATION LIST OF EFFECTIVE SECTIONS (TECHNICAL SPECIFICATIONS BASES)
Section Title Revision Page TS / B 3.3-114 1
Page TS / B 3.3-115 2
Page TS / B 3.3-116 3
Pages TS / B 3.3-117 and TS/ B 3.3-118 2
Page TS / B 3.3-119 1
Page TS / B 3.3-120 2
Pages TS / B 3.3-121 and TS / B 3.3-122 3
Page TS / B 3.3-123 1
Page TS / B 3.3-124 2
Page TS / B 3.3-124a 0
Page TS / B 3.3-125 1
Page TS / B 3.3-126 2
Page TS / B 3.3-127 3
Page TS / B 3.3-128 2
Pages TS / B 3.3-129 through TS / B 3.3-131 1
Page TS / B 3.3-132 2
Pages TS / B 3.3-133 and TS/ B 3.3-134 1
Pages TS / B 3.3-135 through TS / B 3.3-137 0
Page TS / B 3.3-138 1
Pages TS / B 3.3-139 through TS / B 3.3-146 0
Pages B 3.3-147 through B 3.3-149 0
Page TS / B 3.3-150 1
Pages TS / B 3.3-151 through TS / B 3.3-154 2
Page TS / B 3.3-155 1
Pages TS / B 3.3-156 through TS / B 3.3-158 2
Pages TS / B 3.3-159 through TS / B 3.3-162 1
Rages TS / B 3.3-163 through TS / B 3.3-166 2
Pages TS / B 3.3-167 and TS / B 3.3-168 1
Pages TS / B 3.3-169 and TS / B 3.3-170 3
Pages TS / B 3.3-171 through TS /B 3.3-174 1
Page TS./ B 3.3-174a 1
Pages TS/ B 3.3-175 through TS / B 3.3-177 1
Page TS / B 3.3-178 2
Page TS / B 3.3-179 3
Page TS / B 3.3-179a 2
Page TS / B 3.3-180 1
Page TS / B 3.3-181 3
Page TS / B 3.3-182 1
Page TS / B 3.3-183 2
Page TS / B 3.3-184 1
Page TS / B 3.3-185 4
Page 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
Page TS / B 3.3-192 0
SUSQUEHANNA
- UNIT 2 TS / B LOES-4 Revision 124 SUSQUEHANNA - UNIT 2 TS / B LOES-4 Revision 124
SUSQUEHANNA STEAM ELECTRIC STATION LIST OF EFFECTIVE SECTIONS (TECHNICAL SPECIFICATIONS BASES)
Section Title Revision Page TS / B 3.3-193 1
Pages TS / B 3.3-194 and TS / B 3.3-195 0
Page TS / B 3.3-196 2
Pages TS / B 3.3-197 through TS / B 3.3-205 0
Page TS / B 3.3-206 1
Pages B 3.3-207 through B 3.3-209 0
Page TS / B 3.3-210 1
Page TS / B 3.3-211 2
Pages TS / B 3.3-212 and TS / B 3.3-213 1
Pages B 3.3-214 through B 3.3-220 0
B 3.4 REACTOR COOLANT SYSTEM BASES Pages TS / B 3.4-1 and TS / B 3.4-2 2
Pages TS / B 3.4-3 through TS / B 3.4-5 4
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Section Title Revision B 3.9 REFUELING OPERATIONS BASES Pages TS / B 3.9-1 and TS / B 3.9-2 1
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PPL Rev. 5 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 AREVA NP 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 AOOs, 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 weaker form. This weaker form may lose its integrity, resulting in an uncontrolled release of activity to the reactor coolant.
(continued)
SUSQUEHANNA - UNIT 2 TS / B 2.0-1 Revision 2
PPL Rev. 5 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 SPCB (Reference 4) correlation is valid for critical power calculations at pressures > 571.4 psia and bundle mass fluxes
> 0.087 x 106 lb/hr-ft2 for SPCB. 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 AREVA NP ATRIUM-10 design, the minimum bundle flow is > 28 x 10 3
lb/hr. For AREVA NP ATRIUM-10 fuel design, the coolant minimum bundle flow and maximum area are such that the mass flux is always > 0.25 x 106 lb/hr-ft2. Full scale critical power test data taken from various AREVA NP and GE fuel designs at pressures from 14.7 psia to 1400 psia indicate the fuel assembly critical power at 0.25 x 106 Ib/hr-ft2 is approximately 3.35 MWt. At 23% RTP, a bundle power of approximately 3.35 MWt corresponds to a bundle radial peaking factor of approximately 2.8, which is significantly higher than the expected peaking factor. Thus, a THERMAL POWER limit of 23% RTP for reactor pressures < 557 psig is conservative and for conditions of lesser power would remain the same.
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)
SUSQUEHANNA - UNIT 2 TS / B 2.0-2 Revision 5
PPL Rev. 5 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 SPCB 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 SPCB 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.
AREVA NP ATRIUM-10 fuel is monitored using the SPCB 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. 90-02 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 shut 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 / B 2.0-3 Revision 5
PPL Rev. 5 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 be 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 regulatory limits. 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.
ANF-524 (P)(A), Revision 2, "Critical Power Methodology for Boiling Water Reactors," Supplement 1 Revision 2 and Supplement 2, November 1990.
- 3.
Deleted.
- 4.
EMF-2209(P)(A), "SPCB Critical Power Correlation," AREVA NP,
[See Core Operating Limits Report for Revision Level].
- 5.
EMF-2158(P)(A), Rev. 0, "Siemens Power Corporation Methodology for Boiling Water Reactors: Evaluation and Validation of CASMO-4 / MICROBURN-B2," October 1999.
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