ML15317A009
| ML15317A009 | |
| Person / Time | |
|---|---|
| Site: | Susquehanna  |
| Issue date: | 11/05/2015 |
| From: | Gerlach R Susquehanna |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| Download: ML15317A009 (31) | |
Text
Nov. 05, 2015 Page 1 of 2 MANUAL HARD COPY DISTRIBUTION DOCUMENT TRANSMITTAL 2015-44473 USER INFORMATION:
GERLACH*ROSEY N EMPL#:028401 CA#l: 0363 Address: NUCSA2 Phone#: 254-3194 TRANSMITTAL INFORMATION:
TO: GERLACH*ROSEY N 11/05/2015 LOCATION: USNRC FROM: NUCLEAR RECORDS DOCUMENT CONTROL CENTER (NUCSA-2)
THE FOLLOWING CHANJGES HAVE OCCURRED TO THE HARDCOPY OR ELECTRONIC MANUAL ASSIGNED TO YOU. HARDCOPY USERS MUST ENSURE THE DOCUMENTS PROVIDED MATCH THE INFORMATION ON THIS TRANSMITTAL. WHEN REPLACING THIS MATERIAL IN YOUR HARDCOPY MANUAL, ENSURE THE UPDATE DOCUMENT ID IS THE SAME DOCUMENT ID YOU'RE REMOVING FROM YOUR MANUAL. TOOLS FROM THE HUMAN PERFORMANCE TOOL BAG SHOULD BE UTILIZED TO ELIMINATE THE CHANCE OF ERRORS.
ATTENTION: "REPLACE" directions do not affect the Table of Contents, Therefore no T0C will be issued with the updated material.
TSBI - TECHNICAL SPECIFICATION BASES UNIT 1 MANUAL REMOVE MANUAL TABLE OF CONTENTS DATE: 08/05/2015 ADD M*ANUAL TABLE OF CONTENTS DATE: 11/04/2015 CATEGORY: DOCUMENTS TYPE: TSBl S00i
Nov. 05, 2015 Page 2 of 2 ID: TEXT 3.4.10 REMOVE: REV:3 ADD: REV: 4 CATEGORY: DOCUMENTS TYPE: TSB1 ID: TEXT LOES ADD: REV: 121 REMOVE: REV:120 ANY DISCREPANCIES WITH THE MATERIAL PROVIDED, CONTACT DCS @ X3107 OR X3136 FOR ASSISTANCE. UPDATES FOR HARDCOPY MANUALS WILL BE DISTRIBUTED WITHIN 3 DAYS IN ACCORDANCE WITH DEPARTMENT PROCEDURES. PLEASE MAKE ALL CHANGES AND ACKNOWLEDGE COMPLETE IN YOUR NIMS INBOX UPON COMPLETION OF UPDATES. FOR ELECTRONIC MANUAL USERS, ELECTRONICALLY REVIEW THE APPROPRIATE DOCUMENTS AND ACKNOWLEDGE COMPLETE IN YOUR NIMS INBOX.
SSES MANUAL Manual Name: TSBI Manual
Title:
TECHNICAL SPECIFICATION BASES UNIT 1 MANUAL Table Of Contents Issue Date: 11/04/2015 Procedure Name Rev Issue Date Change ID Change Number TEXT LOES 121 11/04/2 015
Title:
LIST OF EFFECTIVE SECTIONS TEXT TOC 23 07/02/2014
Title:
TABLE OF CONTENTS TEXT 2.1.1 6 01/22 /2 015
Title:
SAFETY LIMITS (SLS) REACTOR CORE SLS TEXT 2.1.2 1 10/ 04/2 007
Title:
SAFETY LIMITS (SLS) REACTOR COOLANT SYSTEM (RCS) PRESSURE S TEXT 3.0 3 08/20/2009
Title:
LIMITING CONDITION FOR OPERATION (LCO) APPLICABILITY TEXT 3.1.1 1 04/18/2006
Title:
REACTIVITY CONTROL SYSTEMS SHUTDOWN MARGIN (SDM)
TEXT 3.1.2 0 11/15/2002
Title:
REACTIVITY CONTROL SYSTEMS REACTIVITY ANOMALIES TEXT 3.1.3 2 01/19/2009
Title:
REACTIVITY CONTROL SYSTEMS CONTROL ROD OPERABILITY TEXT 3.1.4 4 01/30/2009
Title:
REACTIVITY CONTROL SYSTEMS CONTROL ROD SCRAM TIMES TEXT 3.1.5 1 07/06/2005
Title:
REACTIVITY CONTROL SYSTEMS CONTROL ROD SCRAM ACCUMULATORS TEXT 3.1.6 3 02/24/2014
Title:
REACTIVITY CONTROL SYSTEMS ROD PATTERN CONTROL Report Date: 11/05/15 Pagel Page ! of of 8 Report Date: 11/05/15
SSES MANUA Manual Name: TSBI Manual
Title:
TECHNICAL SPECIFICATION BASES UNIT 1 MANUAL TEXT 3.1.7 3 04/23/2008
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 2 04/23/2008
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 04/23/2008
Title:
POWER DISTRIBUTION LIMITS LINEAR HEAT GENERATION RATE (LHGR)
TEXT 3.3.1.1 6 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 4 02/24/2014
Title:
INSTRUMENTATION CONTROL ROD BLOCK INSTRUMENTATION TEXT 3.3.2.2 2 04/05/2010
Title:
INSTRUMENTATION FEEDWATER MAIN TURBINE HIGH WATER LEVEL TRIP INSTRUMENTATION TEXT 3.3.3.1 9 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 2 02/24/2014
Title:
INSTRUMENTATION END OF CYCLE RECIRCULATION PUMP TRIP (EOC-RPT) INSTRUMENTATION W Report Date: 11/05/15 Page 2.
Page of 8 of .~
Report Date: 11/05/15
SSES MANTJAL Manual Name: TSB1
- aulTitle: TECHNICA SPECIFICATION BASES UNIT 1 MNA 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 3 08/20/2009
Title:
INSTRUMENTATION EMERGENCY CORE COOLING SYSTEM (ECCS) INSTRUMENTATION TEXT 3.3.5.2 0 11/15/2002
Title:
INSTRUMENTATION REACTOR CORE ISOLATION COOLING CRCIC) 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 2 10/27/2008
Title:
INSTRUMENTATION CONTROL ROOM EMERGENCY OUTSIDE AIR SUPPLY (CREOAS) SYSTEM INSTRUMENTATION TEXT 3.3.8.1 2 12/17/2007
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 4 04/27/2010
Title:
REACTOR COOLANT SYSTEM (RCS) RECIRCULATION LOOPS OPERATING TEXT 3.4.2 3 10/23/2013
Title:
REACTOR COOLANT SYSTEM (RCS) 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 11/15/2002
Title:
REACTOR COOLANT SYSTEM (RCS) RCS OPERATIONAL LEAKAGE Report Date: 11/05/15 Page Page A3 of of 88 Report Date: 11/05/15
SSES MANqUAL Manual Name: TSBI SManual
Title:
TECHNICAL SPECIFICATION BASES UNIT 1 MANUAL TEXT 3.4.5 1 01/16/2006
Title:
REACTOR COOLANT SYSTEM CRCS) RCS PRESSURE ISOLATION VALVE (PIV) LEAKAGE TEXT 3.4.6 4 02/19/2 014
Title:
REACTOR COOLANT SYSTEM (RCS) RCS LEAKAGE DETECTION INSTRUMENTATION TEXT 3.4.7 2 10/04/2007
Title:
REACTOR COOLANT SYSTEM (RCS) RCS SPECIFIC ACTIVITY TEXT 3.4.8 2 03/28/2013
Title:
REACTOR COOLANT SYSTEM (RCS) RESIDUAL HEAT REMOVAL (RHR) SHUTDOWN COOLING SYSTEM
- HOT SHUTDOWN TEXT 3.4.9 1 03/28/2013
Title:
REACTOR COOLANT SYSTEM (RCS) RESIDUAL HEAT REMOVAL (RER) SHUTDOWN COOLING SYSTEM
- COLD SHUTDOWN TEXT 3.4.10
Title:
(RCS) 11/04/2015 RCS PRESSURE AND TEMPERATURE (P/T) LIMITS 0
TEXT 3.4.11 0 11/15/2002
Title:
REACTOR COOLANT SYSTEM (RCS) REACTOR STEAM DOME PRESSURE TEXT 3.5.1 4 07/16/2014
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 CRCIC)
SYSTEM ECCS - SHUTDOWN TEXT 3.5.3 3 02/24/2014
Title:
EMERGENCY CORE COOLING SYSTEMS (ECCS) AND REACTOR CORE ISOLATION COOLING (RCIC)
SYSTEM RCIC SYSTEM TEXT 3.6.1.1 5 02/24/2014
Title:
PRIMARY CONTAINMENT TEXT 3.6.1.2 1 04/23/2008
Title:
CONTAINMENT SYSTEMS PRIMARY CONTAINMENT AIR LOCK Report Date: 11/05/15 Page4 Page
- of of 8 Report Date: 11/05/15
SSES MANUALJM Manual Name: TSB1 Manual
Title:
TECHNICAL SPECIFICATION BASES UNIT 1 MANUAL TEXT 3.6.1.3 11 07/02/2014
Title:
CONTAINMENT SYSTEMS PRIMARY CONTAINMENT ISOLATION VALVES (PCIVS)
TEXT 3.6.1.4 1 04/23/2008
Title:
CONTAINMENT SYSTEMS CONTAINMENT PRESSURE TEXT 3.6.1.5 1 10/05/2005
Title:
CONTAINMENT SYSTEMS DRYWELL AIR TEMPERATURE TEXT 3.6.1.6 0 11/15/2002
Title:
CONTAINMENT SYSTEMS SUPPRESSION CHANBER-TO-DRYWELL VACUUM BREAKERS TEXT 3.6.2.1 2 04/23/2008
Title:
CONTAINMENT SYSTEMS SUPPRESSION POOL AVERAGE TEMPERATURE TEXT 3.6.2.2 0 11/15/2002
Title:
CONTAINMENT SYSTEMS SUPPRESSION POOL WATER. LEVEL TEXT 3.6.2.3 1 01/16/2006
Title:
CONTAINMENT SYSTEMS RESIDUAL HEAT REMOVAL (RER) SUPPRESSION POOL COOLING TEXT 3.6.2.4 0 11/15/2002
Title:
CONTAINMENT SYSTEMS RESIDUAL HEAT REMOVAL (RHR) SUPPRESSION POOL SPRAY TEXT 3.6.3.1 2 06/13/2006
Title:
CONTAINMENT SYSTEMS PRIMARY CONTAINMENT HYDROGEN RECOMBINERS TEXT 3.6.3.2 1 04/18/2005
Title:
CONTAINMENT SYSTEMS DRYWELL AIR FLOW SYSTEM TEXT 3.6.3.3 1 02/28/2013
Title:
CONTAINMENT SYSTEMS PRIMARY CONTAINMENT OXYGEN CONCENTRATION TEXT 3.6.4.1 11 11/06/2014
Title:
CONTAINMENT SYSTEMS SECONDARY CONTAINMENT Report Date: 11/05/15 Page~
Page 5 of of 8 Report Date: 11/05/15
SSES MANUJAL Manual Name:- TSB1
-[!:*Manual
Title:
TECHNICAL SPECIFICATION BASES UNIT 1 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 4 04/05/2010
Title:
PLANT SYSTEMS RESIDUAL HEAT REMOVAL SERVICEWATER_ CRHRSW) SYSTEM AND THE ULTIMATE HEAT SINK (UES)
TEXT 3.7.2 2 02/11/2009
Title:
PLANT SYSTEMS EMERGENCY SERVICE WATER (ESW) SYSTEM TEXT 3.7.3 1 01/08/2010
Title:
PLANT SYSTEMS CONTROL ROOM EMERGENCY OUTSIDE AIR SUPPLY (CREOAS) SYSTEM TEXT 3.7.4 0 11/15/2002
Title:
PLANT SYSTEMS CONTROL ROOM FLOOR COOLING SYSTEM TEXT 3.7.5 1 10/04/2007
Title:
PLANT SYSTEMS MAIN CONDENSER OFFGAS TEXT 3.7.6 2 04/23/2008
Title:
PLANT SYSTEMS MAIN TURBINE BYPASS SYSTEM TEXT 3.7.7 1 10/04/2007
Title:
PLANT SYSTEMS SPENT FUEL STORAGE POOL WATER LEVEL TEXT 3.7.8 0 04/23/2008
Title:
PLANT SYSTEMS TEXT 3.8.1 7 02/24/2014
Title:
ELECTRICAL POWER SYSTEMS AC SOURCES - OPERATING
- TEXT 3.8.2 0 11/15/2002 S-
Title:
ELECTRICAL POWER SYSTEMS AC SOURCES -SHUTDOWN Report Date: 11/05/15 Page6 of 8 Page 6 of Report Date: 11/05/15
SSE~S MANUAL
. Manual Name : TSBI Manual
Title:
TECHNICAL SPECIFICATION BASES UNIT 1 MANUAL TEXT 3.8.3 4 10/23/2013
Title:
ELECTRICAL POWER SYSTIJ *EMS DIESEL FUEL OIL, LUBE OIL, AND STARTING AIR TEXT 3.8.4 3 01/19/2009 Title : ELECTRICAL POWER SYST 'EMS DC SOURCES - OPERATING TEXT 3.8.5 1 12/14/2006
__Title: ELECTRICAL POWER SYST ?EMS DC SOURCES - SHUTDOWN TEXT 3.8.6 1 12/14/2006
Title:
ELECTRICAL POWER SYST PEMS BATTERY CELL PARAMVETERS TEXT 3.8.7 1 10/05/2005
Title:
- ELECTRICAL POWER SYST ?EMS DISTRIBUTION SYSTEMS - OPERATING TEXT 3.8.8 0 11/15/2002
Title:
ELECTRICAL POWER SYSI ?ENS DISTRIBUTION SYSTEMS - SHUTDOWN TEXT 3.9.1 0 11/15/2002
Title:
REFUELING C)PERATIONS REFUELING EQUIPMENT INTERLOCKS TEXT 3.9.2 1 09/01/2010
Title:
REFUELING C)PERATIONS REFUEL POSITION ONE-ROD-OUT INTERLOCK TEXT 3.9.3 0 11/15/2002
Title:
REFUELING C)PERATIONS CONTROL ROD POSITION TEXT 3.9.4 0 11/15/2002
Title:
REFUELING C)PERATIONS CONTROL ROD POSITION INDICATION TEXT 3.9.5 0 11/15/2002
Title:
REFUELING C)PERATIONS CONTROL ROD OPERABILITY - REFUELING TEXT 3.9.6 1 10/04/2007
Title:
REFUELING C)PERATIONS REACTOR PRESSURE VESSEL (RPV) WATER LEVEL Report Date: 11/05/15 Page Page 2 of 88 of Report Date: 11/05/15
- SEg MANUITAL Manual Name:s TSBI
?***Manual
Title:
TECHNICAL SPECIFICATION BASES UNIT 1 MANUAL TEXT 3.9.7 0 11/15/2002
Title:
REFUELING OPERATIONS RESIDUAL HEAT REMOVAL (RHR) - HIGH WATER LEVEL TEXT 3.9.8 0 11/15/2002
Title:
REFUELING OPERATIONS RESIDUAL HEAT REMOVAL (RER) - LOW WATER LEVEL TEXT 3.10.1 1 01/23/2008
Title:
SPECIAL OPERATIONS INSERVICE LEAK AND HYDROSTATIC TESTING OPERATION TEXT 3.10.2 0 11/15/2002
Title:
SPECIAL OPERATIONS REACTOR MODE SWITCH INTERLOCK TESTING TEXT 3.10.3 0 11/15/2002
Title:
SPECIAL OPERATIONS SINGLE CONTROL ROD WITHDRAWAL - HOT SHUTDOWN TEXT 3.10.4 0 11/15/2002
Title:
SPECIAL OPERATIONS SINGLE CONTROL ROD WITHDRAWAL - COLD SHUTDOWN S
TEXT 3.10.5 0 11/15/2002
Title:
SPECIAL OPERATIONS SINGLE CONTROL ROD DRIVE (CRD) REMOVAL - REFUELING TEXT 3.10.6 0 11/-15/2002
Title:
SPECIAL OPERATIONS MULTIPLE CONTROL ROD WITHDRAWAL - REFUELING TEXT 3.10.7 1 04/18/2006
Title:
SPECIAL OPERATIONS CONTROL ROD TESTING - OPERATING TEXT 3.10.8 1 04/12/2006
Title:
SPECIAL OPERATIONS SHUTDOWN MARGIN (SDM) TEST - REFUELING Report Date: 11/05/15 Page8 Page 8 of of 88 Report Date: 11/05/15
SUSQUEHANNA STEAM ELECTRIC STATION LIST OF EFFECTIVE SECTIONS (TECHNICAL SPECIFICATIONS BASES)
Section Title Revision TOC Table of Contents 23 B 2.0 SAFETY LIMITS BASES Page TS / B2.0-1 1 Page TS /B 2.0-2 3 Page TS /8B2.0-3 6 Page TS /B 2.0-4 4 Page"S / B 2.0-5 6 Page TS / B 2.0-6 1 Pages TS / B 2.0-7 through TS / B 2.0-9 1 B 3.0 LCO AND SR APPLICABILITY BASES Page TS / B3.0-1 I*
Pages TS / B 3.0-2 through TS/IB 3.0-4 0 .....
Pages TS / B 3.0-5 through TS / B 3.0-7 *,1 Page TS / B3.0-8 3 Pages TS / B 3.0-9 through TS I B 3.0-11 2 Page TS / B 3.0-11a 0 Page TS /B 3.0-12I Pages TS / B 3.0-13 through TS I2 Pages TS /B3.0-16 and TS 0 B 3.1 REACTIVITY CONTROL BAy Pages B83.1-1 trug 0 Page TS / B 3.1-5A 1 PagesTS /Ba3.1 d . 72 Pages B3.1-8 t 3.
Page 3. 20ghB 3.1-19 0 20 and TS / B 3.1-21 1 Pa
- 0
- 3.1-23I 3.1-240 Pa TS / B 3.1-25 through TS /8B3.1-27 1
- TS B 31-282
,= TS/ B 31-29I Pages B 3.1-30 through B 3.1-33 0 Pages TS I B 3.3-34 through TS I B 3.3-36 1 Page TS /8B3.1-37 2 Page TS I B 3.1-38 3 Pages TS /8B3.1-39 and TS /8B3.1-40 2 Page TS I B 3.1-40a 0 Pages TS /B 3.1-41 and TS / B3.1-42 2 Revision 121 SUSQUEHANNA SUSQUEHANNA - UNIT
- UNIT 11 TS / B LOES-1 TSIB LOES-1 Revision 121
SUSQUEHANNA STEAM ELECTRIC STATION LIST OF EFFECTIVE SECTIONS (TECHNICAL SPECIFICATIONS BASES)
Section Title Revision Page TS I B 3.1.43 1 Page TS / B 3.1-44 0 Page TS I B 3.1-45 3 Pages TS I B 3.1-46 through TS I B 3.1-49 1 Page TS I B 3.1-50 0 Page TS / B 3.1-51 3 B 3.2 POWER DISTRIBUTION LIMITS BASES Page TS I B 3.2-1 2 Pages TS I B 3.2-2 and TS / B 3.2-33 Pages TS / B 3.2-4 and TS I B 3.2-5 2
- Page TSI/B 3.2-6 3
_ ageB3.2-7 .. ...... _1__
Pages TS / B 3.2-8 and TS I B 3.2-9 3 Page TS / B 3.2.10 2 Page TS / B 3.2-11 3 Page TS I B 3.2-12 1 Page TS / B 3.2-13 2 B 3.3 INSTRUMENTATION Pages TS / B 3.3-1 through TS I B 3.3-4 1 Page TS I B 3.3-5 2 Page TS / B 3.3-6 I Page TS / B 3.3-7 3 Page TS I B 3.3-7a I Page TS / B 3.3-8 5 Pages TS I B 3.3-9 through TS / B 3.3-12 3 Pages TS I B 3.3-12a I Pages TS / B 3.3-12b and TS I B 3.3-12c 0 Page TS I B 3.3-13 1 Page TS / B 3.3-14 3 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 4 Page TS / B 3.3-19 1 Pages TS I B 3.3-20 through TS I B 3.3-22 2 Page TS / B 3.3-22a 0 Pages TS / B 3.3-23 and TS I B 3.3-24 2 Pages TS I B 3.3-24a and TS / B 3.3-24b 0 Page TS / B 3.3-25 3 Page TS / B 3.3-26 2 Page TS / B 3.3-27 1 Page TS / B 3.3-28 3 Page TSI/B 3.3-29 4 Page TS I B 3.3-30 3 Page TS / B 3.3-30a 0 SUSQUEHANNA-UNIT1 TS / B LOES-2 Revision 121
SUSQUEHANNA STEAM ELECTRIC STATION LIST OF EFFECTIVE SECTIONS (TECHNICAL SPECIFICATIONS BASES)
Section Title Revision Page TS / B 3.3-31 4 Page TS / B 3.3-32 5 Pages TS / B 3.3-32a 0 Page TS / B 3.3-32b 1 Page TS / B 3.3-33 5 Page TS I B 3.3-33a 0 Page TS / B 3.3-34 1 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 Page TSI/B 3.3-39 2 Pages TS I B 3.3-40 through TS I B 3.3-43 1 Page TSI/B 3.3-44 4 Pages TS I B 3.3_-4a and TS/_ B 3.3-44b -0 Page TS / B 3.3-45 3 Pages TS / B 3.3-45a and TS I B 3.3-45b 0 Page TS / B3.3-46 3 Pages TS / B 3.3-47 2 Pages TS I B 3.3-48 through TS / B 3.3-51 3 Pages TS I B 3.3-52 and TS / B 3.3-53 2 Page TS / B 3-3-53a 0 Page TS / B 3.3-54 5 Page TS / B 3.3-55 2 Pages TS I B 3.3-56 and TS / B 3.3-57 1 Page TS I B 3.3-58 0 Page TS I 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 / B 3.3-69 5 Pages TS / B 3.3-70 4 Page TS / B3.3-71 3 Pages TS I B 3.3-72 and TS I B 3.3-73 2 Page TS / B 3.3-74 3 Page TS I B 3.3-75 2 Page TS / B 3.3-75a 6 Page TS / B 3.3-75b 7 Page TS / B 3.3-75c 6 Pages B 3.3-76 through B 3.3-77 0 Page TS I B 3.3-78 1 UNIT 1I TSIBLOES-3 Revision 121 SUSQUEHANNA UNIT SUSQUEHANNA- - TS / B LOES-3 Revision 121
SUSQUEHANNA STEAM ELECTRIC STATION LIST OF EFFECTIVE SECTIONS (TECHNICAL SPECIFICATIONS BASES)
Section Title Revision Pages B 3.3-79 through B 3.3-81 0 Page TS I B 3.3-82 2 Page B 3.3-83 0 Pages B 3.3-84 and B 3.3-85 1 Page B 3.3-86 0 Page B 3.3-87 1 Page B 3.3-88 0 Page B 3.3-89 1 Page TS I B 3.3-90 1 Page B 3.3-91 0.
Pages TS/IB 3.3-92 through TS IB 3.3-100 1
- Pages TSI/B 3.3-101 through TS/B 3.3-103 .0 PageTSI/B3.3-_104 .....-
Pages TS I B 3.3-1 05 and TS I B 3.3-106 0 Page TS / B3.3-107 I Page TS I B 3.3-108 0 Page TS I B 3.3-109 1 Pages TS I B 3.3-110 and TS/B83.3-111 0 Pages TS I B 3.3-112 and TS I B 3.3-1 12a I Pages TS / B 3.3-113 through TS / B 3.3-115 1 Page TS / B 3.3-116 3 Page TS / B 3.3-117 1 Pages TS I B 3.3-118 through TS I B 3.3-122 0 Pages TS I B 3.3-123 and TS I B 3.3-124 1 Page TS I B 3.3-124a 0 Page TS I B 3.3-125 0 Pages TS I B 3.3-126 and TS I B 3.3-127 1 Pages TS I B 3.3-1 28 through TS/ B 3.3-130 0 Page TS / B 3.3-131 1 Pages TS I B 3.3-1 32 through TS I B 3.3-1 34 0 Pages B 3.3-135 through B 3.3-137 0 Page TS / B 3.3-138 1 Pages B 3.3-1 39 through B 3.3-149 0 Pages TS I B 3.3-150 and TS I B 3.3-151 1 Pages TS I B 3.3-1 52 through TS/IB 3.3-154 2 Page TS I B 3.3-155 1 Pages TS I B 3.3-1 56 through TS I B 3.3-158 2 Pages TS / B 3.3-159 and TS I B 3.3-160 1 Page TS / B 3.3-161 2 Page TS I B 3.3-162 1 Page TS I B 3.3-163 2 Page TS I B 3.3-164 1 Pages TS I B 3.3-165 through TS I B 3.3-167 2 TSIBLOES-4 Revision 121 SUSQUEHANNA SUSQUEHANNA-UNIT1- UNIT I TS / B LOES-4 Revision 121
SUSQUEHANNA STEAM ELECTRIC STATION LIST OF EFFECTIVE SECTIONS (TECHNICAL SPECIFICATIONS BASES)
Section Title Revision Pages TS/IB 3.3-168 and TS / B 3.3-169 1 Page TS I B 3.3-170 3 Page TS I B 3.3-171 2 Pages TS I B 3.3-172 through TS / B 3.3-1 77 1 Pages TS / B 3.3-178 and TS I B 3.3-179 2 Page TS / B 3.3-179a 2 Pages TS I B 3.3-179b and TS / B 3.3-179c 0 Page TS / B 3.3-180 1 Page TIS / B 3.3-181 3 Page TS / B 3.3-182 1 Page TS / B 3.3-183 2 Page TS /8B3.3-184 1
_P~ageTS /B 3.3-_185 - _
Page TS / B 3.3-186 1 Pages TS I B 3.3-187 and TS I B 3.3-188 2 Pages TS / B 3.3-189 through TS I B 3.3-191 1 Page TS / B 3.3-192 0 Page TS / B 3.3-193 1 Pages TS I B 3.3-194 and TS I B 3.3-195 0 Page TS I B 3.3-196 2 Pages TS I B 3.3-197 through TS I B 3.3-204 0 Page TS / B 3.3-205 1 Pages B 3.3-206 through B 3.3-209 0 Page TS / B 3.3-210 1 Pages B 3.3-211 through B 3.3-219 0 B 3.4 REACTOR COOLANT SYSTEM BASES Pages B 3.4-1 and B 3.4-2 0 Pages TS / B 3.4-3 and Page TS I B 3.4-4 4 Page TS / B 3.4-5 3 Pages TS / B 3.4-6 through TS / B 3.4-9 2 Page TS / B 3.4-10 1 Pages TS / 3.4-11 and TS I B 3.4-12 0 Page TS / B 3.4-13 2 Page TS / B 3.4-14 1 Page TS /8B3.4-15 2 Pages TS /8B3.4-16 and TS 18B3.4-17 4 Page TS / B 3.4-18 2 Pages B 3.4-19 through B 3.4-27 0 Pages TS /8B3.4-28 and TS /8B3.4-29 1 Page TS / B 3.4-30 2 Page TS / B 3.4-31 1 Pages TS / B 3.4-32 and TS / B 3.4-33 2 Page TS /8B3.4-34 1 Page TS I B 3.4-34a 0 TS/BLOES-5 Revision 121 SUSQUEHANNA - UNIT SUSQUEHANNA -
UNIT I 1 TS / B LOES-5 Revision 121
SUSQUEHANNA STEAM ELECTRIC STATION LIST OF EFFECTIVE SECTIONS (TECHNICAL SPECIFICATIONS BASES)
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/ B ILOES-7 UNIT TS/BLOES-.7
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S USQUEHANNA STEAM ELECTRIC STATION LIST OF EFFECTIVE SECTIONS (TECHNICAL SPECIFICATIONS BASES)
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UNIT I1 TS / B LOES-9 TS/B LOES-9 Revision 121
SUSQUEHANNA STEAM ELECTRIC STATION LIST OF EFFECTIVE SECTIONS (TECHNICAL SPECIFICATIONS BASES)
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Rev. 4 RCS P/T" Limits B 3.4.10 B 3.4 REACTOR COOLAkNT SYSTEM (RCS)
B 3.4.10 RCS Pressure and Temperature (PIT) Limits BASES BACKGROUND All components of the RCS are designed to withstand effects of cyclic loads due to system pressure and temperature changes. These loads are introduced by startup (heatup) and shutdown (cooldown) operations, power transients, and reactor trips. This LCO limits the pressure and temperature changes during RCS heatup and cooldown, within the design assumptions and the stress limits for cyclic operation.
This Specification contains PIT limit curves for heatup, cooldown, and
....... in~servi~ce_ lea~kage and_ hyd~rosta~ti~c testing, and limits for the maximum rate of change of reactor coolant temperature. The heatup curve provides limits for both heatup and criticality.
Each PIT limit curve defines an acceptable region for normal operation. The usual use of the curves is operational guidance during heatup or cooldown maneuvering, when pressure and temperature indications are monitored and compared to the applicable curve to determine that operation is within the allowable region.
The LCO establishes operating limits that provide a margin to brittle failure of the reactor vessel and piping of the reactor coolant pressure boundary (RCPB). The vessel is the component most subject to brittle failure.
Therefore, the LCO limits apply mainly to the vessel.
10 CFR 50, Appendix G (Ref. 1), requires the establishment of P/T limits for material fracture toughness requirements of the RCPB materials.
Reference I requires an adequate margin to brittle failure during normal operation, anticipated operational occurrences, and system hydrostatic tests.
It mandates the use of the ASME Code,Section XI, Appendix G (Ref. 2).
The actual shift in the RTNDT of the vessel material will be established periodically by removing and evaluating the irradiated reactor vessel material specimens, in accordance with ASTM E 185 (Ref. 3) and Appendix H of 10 CFR 50 (Ref. 4). The operating PIT limit curves will be adjusted, as necessary, based on the evaluation findings and the recommendations of RG 1.99, "Radiation Embrittlement of Reactor Vessel Materials" (Ref. 5).
The calculations to determine neutron fluence will be developed using the BWRVIP RAMA code methodology, which is NRC approved and meets the intent of RG 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence" (Ref. 11). See FSAR Section 4.1.4.5 for determining fluence (Ref. 12).
(continued)
SUSQUEHANNA - UNIT I1SIB3449Rvso TS / B 3.4-49 Revision 3
Rev. 4 RCS P/T"Limits B 3.4.10 BASES BACKGROUND The P/T limit curves are composite curves established by superimposing (continued) limits derived from stress analyses of those portions of the reactor vessel and head that are the most restrictive. At any specific pressure, temperature, and temperature rate of change, one location within the reactor vessel will dictate the most restrictive limit. Across the span of the P/T limit curves, different locations are more restrictive, and, thus, the curves are composites of the most restrictive regions.
The heatup curve used to develop the P/T limit curve composite represents a different set of restrictions than the cooldown curve used to develop the P/T limit curve composite because the directions of the thermal gradients through the vessel-wall are reversed. The thermal gradient-reversal-alters the location of the tensile stress between the outer and inner walls.
The criticality limits include the Reference I requirement that they be at least 40°F above the heatup curve or the cooldown curve and not lower than the minimum permissible temperature for the inservice leakage and hydrostatic testing.
The consequence of violating the LCO limits is that the RCS has been operated under conditions that can result in brittle failure of the RCPB, possibly leading to a nonisolable leak or loss of coolant accident. In the event these limits are exceeded, an evaluation must be performed to determine the effect on the structural integrity of the RCPB components.
ASME Code,Section XI, Appendix E (Ref. 6), provides a recommended methodology for evaluating an operating event that causes an excursion outside the limits.
APPLICABLE The P/l" limits are not derived from Design Basis Accident (DBA) analyses.
SAFETY They are prescribed during normal operation to avoid encountering pressure, ANALYSES temperature, and temperature rate of change conditions that might cause undetected flaws to propagate and cause nonductile failure of the RCPB, a condition that is unanalyzed. Reference 7 establishes the methodology for determining the P/l" limits. Since the P/T limits are not derived from any DBA, there are no acceptance limits related to the P/T limits. Rather, the P/T limits are acceptance limits themselves since they preclude operation in an unanalyzed condition.
RCS P/T limits satisfy Criterion 2 of the NRC Policy Statement (Ref. 8).
(continued)
SUSQUEHANNA - UNIT 1 SIB3450Rvso TS / B 3.4-50 Revision 1
Rev. 4 RCS P/T Limits 8 3.4.10 BASES APPLICABLE The Effective Full Power Years (EFPY) shown on the curves are SAFETY approximations of the ratio of the energy that has been and is anticipated to ANALYSES be generated in a year to the energy that could have been generated ifthe (continued) unit ran at original thermal power rating of 3293 MWT for the entire year.
These values are based on fluence limits that are not to be exceeded.
LCO The elements of this LCO are:
- a. RCS pressure and temperature are to the right of the applicable curves specified in Figures 3.4.10-1 through 3.4.10-3 and within the applicable heat-up or cool down rate specified in SR 3.4.10.1 during RCS heatup, cooldown, and inservice leak and hYdrostatic testing; b._ The temperature difference between the reactorivessel bottom~he~ad coolant and the reactor pressure vessel (RPV) coolant is _<1450 F during recirculation pump startup, and during increases in THERMAL POWER or loop flow while operating at low THERMAL POWER or loop flow;
- c. The temperature difference between the reactor coolant in the respective recirculation loop and in the reactor vessel is < 50°F during recirculation pump startup, and during increases in THERMAL POWER or loop flow while operating at low THERMAL POWER or loop flow;
- d. RCS pressure and temperature are to the right of the criticality limits specified in Figure 3.4.10-3 prior to achieving criticality; and
- e. The reactor vessel flange and the head flange temperatures are > 70°F when tensioning the reactor vessel head bolting studs.
These limits define allowable operating regions and permit a large number of operating cycles while also providing a wide margin to nonductile failure.
The P/T limit composite curves are calculated using the worst case of material properties, stresses, and temperature change rates anticipated under all heatup and cooldown conditions. The design calculations account for the reactor coolant fluid temperature impact on the inner wall of the vessel and the temperature gradients through the vessel wall. Because these fluid temperatures drive the vessel wall temperature gradient, monitoring reactor coolant temperature provides a conservative method of ensuring the P/T limits are not exceeded. Proper monitoring of vessel temperatures to assure compliance with brittle fracture temperature limits and vessel thermal stress limits during normal heatup and cooldown, and.
during inservice leakage and hydrostatic testing, is established in PPL Calculation EC-062-0573 (Ref. 9). For P/T curves A, B, and C, the bottom head drain line coolant temperature should be monitored and maintained to the right of the most limiting curve.
(continued)
SUSQUEHANNA - UNIT 1 S/B3451Rvso TS / B 3.4-51 Revision 3
Rev. 4 RCS P/T Limits B 3.4.10 BASES LCO Curve A must be used for any ASME Section III Design Hydrostatic Tests (continued) performed at unsaturated reactor conditions. Curve A may also be used for ASME Section XI inservice leakage and hydrostatic testing when heatup and cooldown rates can be limited to 20°F in a one-hour period.
Curve A is based on pressure stresses only. Thermal stresses are assumed to be insignificant. Therefore, heatup and cooldown rates are limited to 20°F in a one-hour period when using Curve A to ensure minimal thermal stresses. The recirculation loop suction line temperatures should be monitored to determine the temperature change rate.
Curves B and C are to be used for non-nuclear and nuclear heatup and cooldown, respectively. In addition, Curve B may be used for ASME Section Xl in*service leakage and hydrostatic_ testing, but not for ASME Section III Design Hydrostatic Tests performed at unsaturated reactor conditions. Heatup and cooldown rates are limited to 100°F in a one-hour period when using Curves B and C. This limits the thermal gradient through the vessel wall, which is used to calculate the thermal stresses in the vessel wall. Thus, the LCO for the rate of coolant temperature change limits the thermal stresses and ensures the validity of the P/T curves. The vessel belt-line fracture analysis assumes a l0 0°F/hr coolant heatup or cooldown rate in the beitline area. The 100°F limit in a one-hour period applies to the coolant in the beltline region, and takes into account the thermal inertia of the vessel wall. Steam dome saturation temperature (TsAT), as derived from steam dome pressure, should be monitored to determine the beltline temperature change rate at temperatures above 212°F. At temperatures below 212°F, the recirculation loop suction line temperatures should be monitored.
During heatups and cooldowns, the reactor vessel could experience a vacuum (negative pressure) at low temperatures (unsaturated conditions) and low rates of temperature change. Under a vacuum, the vessel wall would experience a uniform compressive loading, which would counteract the tensile stress due to any thermal gradients through the vessel wall. To ensure the margin to brittle fracture is no less than at any other pressure, Curves A, B, and C require a minimum vessel metal temperature of 70°F when the reactor vessel is at a negative pressure.
(continued)
SUSQUEHANNA - UNIT I1Sl TS / B 3.4-52
.- 2Rvso Revision 3
Rev. 4 RCS PIT Limits B] 3.4.10 BASES LCO Violation of the limits places the reactor vessel outside of the bounds of the (continued) stress analyses and can increase stresses in other RCS components. The consequences depend on several factors, as follows:
- a. The severity of the departure from the allowable operating pressure temperature regime or the severity of the rate of change of temperature;
- b. The length of time the limits were violated (longer violations allow the temperature gradient in the thick vessel walls to become more pronounced); and
- c. The existences, sizes, and orientations of flaws in the vessel material.
APPLICABILITY The potential for violating a P/T limit exists at all times. For example, PIT limit violations could result from ambient temperature conditions that result in the reactor vessel metal temperature being less than the minimum allowed temperature for boltup. Therefore, this LCO is applicable even when fuel is not loaded in the core.
ACTIONS A.1 and A.2 Operation outside the PIT limits while in MODES 1, 2, and 3 must be corrected so that the RCPB is retumed to a condition that has been verified
- 1 by stress analyses.
The 30 minute Completion Time reflects the urgency of restoring the parameters to within the analyzed range. Most violations will not be severe, and the activity can be accomplished in this time in a controlled manner.
B]esides restoring operation within limits, an evaluation is required to determine if RCS operation can continue. The evaluation must verify the RCPB integrity remains acceptable and must be completed if continued operation is desired. Several methods may be used, including comparison with pre-analyzed transients in the stress analyses, new analyses, or inspection of the components.
(continued)
SUSQUEHANNA - UNIT I TS / B 3.4-52a Rvso 0 Revision
Rev. 4 RCS P/T- Limits B 3.4.10 BASES ACTIONS A.1I and A.2. (continued)
ASME Code, Section Xl, Appendix E (Ref. 6), may be used to support the evaluation. However, its use is restricted to evaluation of the vessel beltline.
The 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Completion Time is reasonable to accomplish the evaluation of a mild violation. More severe violations may require special, event specific stress analyses or inspections. A favorable evaluation must be completed if continued operation is desired.
Condition A is modified by a Note requiring Required Action A.2 be completed whenever the Condition is entered. The Note emphasizes the need to perforrm the evaluation of the effects of the excursion Outside the
-- --- -- - allowable limits*-Restoration alone per Required Action A. 1 is-insufficient because higher than analyzed stresses may have occurred and may have affected the RCPB integrity.
B.1I and B.2 If a Required Action and associated Completion Time of Condition A are not met, the plant must be placed in a lower MODE because either the RCS remained in an unacceptable P/T region for an extended period of increased stress, or a sufficiently severe event caused entry into an unacceptable region. Either possibility indicates a need for more careful examination of the event, best accomplished with the RCS at reduced pressure and temperature. With the reduced pressure and temperature conditions, the possibility of propagation of undetected flaws is decreased.
(continued)
SUSQUEHANNA -UNIT1IT .- 3Rvso TS ! B 3.4-53 Revision 2
Rev. 4 RCS P/T Limits B 3.4.10 BASES ACTIONS B.1 and B.2 (continued)
Pressure and temperature are reduced by placing the plant in 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 in 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.
C.1 and C.2 Operation outside the P/T limits in other than MODES 1, 2, and 3 (including defueled conditions) must be corrected so that the RCPB is returned to a condition that-has-been verified by stress analyses. The Required Action must be initiated without delay and continued until the limits are restored.
Besides restoring the P/T limit parameters to within limits, an evaluation is required to determine if RCS operation is allowed. This evaluation must verify that the RCPB integrity is acceptable and must be completed before approaching criticality or heating up to > 200°F. Several methods may be used, including comparison with pre-.analyzed transients, new analyses, or inspection of the components. ASME Code,Section XI, Appendix E (Ref. 6),
may be used to support the evaluation; however, its use is restricted to evaluation of the beltiine.
SURVEILLANCE SR 3.4.10.1 REQUIREMENTS Verification that operation is within limits (i.e., to the right of the applicable curves in Figures 3.4.10-1 through 3.4.10-3) is required every 30 minutes when RCS pressure and temperature conditions are undergoing planned changes. This Frequency is considered reasonable in view of the control room indication available to monitor RCS status. Also, since temperature rate of change limits are specified in hourly increments, 30 minutes permits a reasonable time for assessment and correction of minor deviations.
Surveillance for heatup, cooldown, or inservice leakage and hydrostatic testing may be discontinued when the criteria given in the relevant plant procedure for ending the activity are satisfied.
This SR has been modified with a Note that requires this Surveillance to be performed only during system heatup and cooldown operations and inservice leakage and hydrostatic testing.
(continued)
SUSQUEHANNA - UNIT 1 TS SIB3454Rvso
/ B 3.4-54 Revision 2
Rev. 4 RCS P/T Limits B 3.4.10 BASES SURVEILLANCE SR 3.4.10.1 (continued)
REQUIREMENTS Notes to the acceptance criteria for heatup and cooldown rates ensure that more restrictive limits are applicable when the P/T limits associated with hydrostatic and inservice testing are being applied.
SR 3.4.10.2 A separate limit is used when the reactor is approaching criticality.
Consequently, the RCS pressure and temperature must be verified within the appropriate limits (i.e., to the right of the criticality curve in
- -Figure 3.4:10-3)-before withdrawing~ control rods that will make the-reactor critical.
Performing the Surveillance within 15 minutes before control rod withdrawal for the purpose of achieving criticality provides adequate assurance that the limits will not be exceeded between the time of the Surveillance and the time of reactor criticality. Although no Surveillance Frequency is specified, the requirements of SR 3.4.10.2 must be met at all times when the reactor is critical.
SR 3.4.10.3 and SR 3.4.10.4 Differential temperatures within the applicable limits ensure that thermal stresses resulting from the startup of an idle recirculation pump will not exceed design allowances. In addition, compliance with these limits ensures that the assumptions of the analysis for the startup of an idle recirculation loop (Ref. 10) are satisfied.
Performing the Surveillance within 15 minutes before starting the idle recirculation pump provides adequate assurance that the limits will not be exceeded between the time of the Surveillance and the time of the idle pump start.
An acceptable means of demonstrating compliance with the temperature differential requirement in SR 3.4.10.4 is to compare the temperatures of the operating recirculation loop and the idle loop. If both loops are idle, compare the temperature difference between the reactor coolant within the idle loop to be started and coolant in the reactor vessel.
(continued)
SUSQUEHANNA - UNIT I1SIB3455Rvso TS / B 3.4-55 Revision 2
Rev. 4 RCS P/T Limits B 3.4.10 BASES SURVEILLANCE SR 3.4.10.3 and SR 3.4.10.4 (continued)
REQUIREMENTS SR 3.4.10.3 has been modified by a Note that requires the Surveillance to be performed only in MODES 1, 2, 3, and 4. In MODE 5, the overall stress on limiting components is lower. Therefore, AT limits are not required. The Note also states the SR is only required to be met during a recirculation pump start-up, because this is when the stresses occur.
SR 3.4.10.5 and SR 3.4.10.6 Differential temperatures within the applicable limits ensure that thermal
--stres-es resultingg from increases irn-THERMAL POWER or recirc~lation loop flow during single recirculation loop operation will not exceed design allowances. Performing the Surveillance within 15 minutes before beginning such an increase in power or flow rate provides adequate assurance that the limits will not be exceeded between the time of the Surveillance and the time of the change in operation.
An acceptable means of demonstrating compliance with the temperature differential requirement in SR 3.4.10.6 is to compare the temperatures of the operating recirculation loop and the idle loop.
Plant specific startup test data has determined that the bottom head is not subject to temperature stratification at power levels > 27% of RTP and with single loop flow rate _>21,320 gpm (50% of rated loop flow). Therefore, SR 3.4.10.5 and SR 3.4.10.6 have been modified by a Note that requires the Surveillance to be met only under these conditions. The Note for SR 3.4.10.6 further limits the requirement for this Surveillance to exclude comparison of the idle loop temperature if the idle loop is isolated from the RPV since the water in the loop cannot be introduced into the remainder of the. Reactor Coolant System.
SR 3.4.10.7. SR 3.4.10.8, and SR 3.4.10.9 Limits on the reactor vessel flange and head flange temperatures are generally bounded by the other P/T limits during system heatup and cooldown. However, operations approaching MODE 4 from MODE 5 and in MODE 4 with RCS temperature less than or equal to certain specified values require assurance that these temperatures meet the LCO limits.
(continued)
SUSQUEHANNA - UNIT 1 S/B3456Rvso TS / B 3.4-56 Revision 2
Rev. 4 RCS P/T Limits B 3.4.10 BASES SURVEILLANCE SR 3.4.10.7, SR 3.4.10.8, and SR 3.4.10.9 (continued)
REQUIREMENTS The flange temperatures must be verified to be above the limits 30 minutes before and while tensioning the vessel head bolting studs to ensure that once the head is tensioned the limits are satisfied. When in MODE 4 with RCS temperature < 80°F, 30 minute checks of the flange temperatures are required because of the reduced margin to the limits. When in MODE 4 with RCS temperature < 100°F, monitoring of the flange temperature is required every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> to ensure the temperature is within the specified limits.
The 30 minute Frequency reflects the urgency of maintaining the temperatures-within limits, and also limits the time that the temperature limits could be exceeded. The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Frequency is reasonable based on the rate of temperature change possible at these temperatures.
REFERENCES 1. 10 CFR 50, Appendix G.
- 2. ASME, Boiler and Pressure Vessel Code,Section XI, Appendix G.
- 3. ASTM E 185-73.
- 5. Regulatory Guide 1.99, Revision 2, May 1988.
- 6. ASME, Boiler and Pressure Vessel Code, Section Xl, Appendix E.
(continued)
SUSQUEHANNA - UNIT 1 T / B 3.4-57 TS .- 7Rvso Revision 4
Rev. 4 RCS P/T Limits B 3.4.10 BASES REFERENCES 7. Licensed Topical Reports:
(continued)
- a. Structural Integrity Associates Report No. SlR-05-044, Revision 1-A, "Pressure-Temperature Limits Report Methodology for Boiling Water Reactors," June 2013.
- b. Structural Integrity Associates Report No. 0900876.401, Revision 0-A, "Linear Elastic Fracture Mechanics Evaluation of GE BWR Water Level Instrument Nozzles for Pressure-Temperature Curve Evaluations," May 2013.
- 8. Final Policy Statement on Technical Specifications Improvements,
- - ----- - July-22, -1993 (58 FR 39132). ---.......-
- 9. PPL Calculation EC-062-0573, "Study to Support the Bases Section of Technical Specification 3.4.10."
- 10. FSAR, Section 15.4.4.
- 11. Regulatory Guide 1.190, March 2001.
- 12. ESAR, Section 4.1.4.5.
SUSQUEHANNA - UNIT I1SIB345aRvso TS / B 3.4-57a Revision 0
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SSES MANUAL Manual Name: TSBI Manual
Title:
TECHNICAL SPECIFICATION BASES UNIT 1 MANUAL Table Of Contents Issue Date: 11/04/2015 Procedure Name Rev Issue Date Change ID Change Number TEXT LOES 121 11/04/2 015
Title:
LIST OF EFFECTIVE SECTIONS TEXT TOC 23 07/02/2014
Title:
TABLE OF CONTENTS TEXT 2.1.1 6 01/22 /2 015
Title:
SAFETY LIMITS (SLS) REACTOR CORE SLS TEXT 2.1.2 1 10/ 04/2 007
Title:
SAFETY LIMITS (SLS) REACTOR COOLANT SYSTEM (RCS) PRESSURE S TEXT 3.0 3 08/20/2009
Title:
LIMITING CONDITION FOR OPERATION (LCO) APPLICABILITY TEXT 3.1.1 1 04/18/2006
Title:
REACTIVITY CONTROL SYSTEMS SHUTDOWN MARGIN (SDM)
TEXT 3.1.2 0 11/15/2002
Title:
REACTIVITY CONTROL SYSTEMS REACTIVITY ANOMALIES TEXT 3.1.3 2 01/19/2009
Title:
REACTIVITY CONTROL SYSTEMS CONTROL ROD OPERABILITY TEXT 3.1.4 4 01/30/2009
Title:
REACTIVITY CONTROL SYSTEMS CONTROL ROD SCRAM TIMES TEXT 3.1.5 1 07/06/2005
Title:
REACTIVITY CONTROL SYSTEMS CONTROL ROD SCRAM ACCUMULATORS TEXT 3.1.6 3 02/24/2014
Title:
REACTIVITY CONTROL SYSTEMS ROD PATTERN CONTROL Report Date: 11/05/15 Pagel Page ! of of 8 Report Date: 11/05/15
SSES MANUA Manual Name: TSBI Manual
Title:
TECHNICAL SPECIFICATION BASES UNIT 1 MANUAL TEXT 3.1.7 3 04/23/2008
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 2 04/23/2008
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 04/23/2008
Title:
POWER DISTRIBUTION LIMITS LINEAR HEAT GENERATION RATE (LHGR)
TEXT 3.3.1.1 6 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 4 02/24/2014
Title:
INSTRUMENTATION CONTROL ROD BLOCK INSTRUMENTATION TEXT 3.3.2.2 2 04/05/2010
Title:
INSTRUMENTATION FEEDWATER MAIN TURBINE HIGH WATER LEVEL TRIP INSTRUMENTATION TEXT 3.3.3.1 9 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 2 02/24/2014
Title:
INSTRUMENTATION END OF CYCLE RECIRCULATION PUMP TRIP (EOC-RPT) INSTRUMENTATION W Report Date: 11/05/15 Page 2.
Page of 8 of .~
Report Date: 11/05/15
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INSTRUMENTATION ANTICIPATED TRANSIENT WITHOUT SCRAM RECIRCULATION PUMP TRIP (ATWS-RPT) INSTRUMENTATION TEXT 3.3.5.1 3 08/20/2009
Title:
INSTRUMENTATION EMERGENCY CORE COOLING SYSTEM (ECCS) INSTRUMENTATION TEXT 3.3.5.2 0 11/15/2002
Title:
INSTRUMENTATION REACTOR CORE ISOLATION COOLING CRCIC) 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 2 10/27/2008
Title:
INSTRUMENTATION CONTROL ROOM EMERGENCY OUTSIDE AIR SUPPLY (CREOAS) SYSTEM INSTRUMENTATION TEXT 3.3.8.1 2 12/17/2007
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 4 04/27/2010
Title:
REACTOR COOLANT SYSTEM (RCS) RECIRCULATION LOOPS OPERATING TEXT 3.4.2 3 10/23/2013
Title:
REACTOR COOLANT SYSTEM (RCS) 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 11/15/2002
Title:
REACTOR COOLANT SYSTEM (RCS) RCS OPERATIONAL LEAKAGE Report Date: 11/05/15 Page Page A3 of of 88 Report Date: 11/05/15
SSES MANqUAL Manual Name: TSBI SManual
Title:
TECHNICAL SPECIFICATION BASES UNIT 1 MANUAL TEXT 3.4.5 1 01/16/2006
Title:
REACTOR COOLANT SYSTEM CRCS) RCS PRESSURE ISOLATION VALVE (PIV) LEAKAGE TEXT 3.4.6 4 02/19/2 014
Title:
REACTOR COOLANT SYSTEM (RCS) RCS LEAKAGE DETECTION INSTRUMENTATION TEXT 3.4.7 2 10/04/2007
Title:
REACTOR COOLANT SYSTEM (RCS) RCS SPECIFIC ACTIVITY TEXT 3.4.8 2 03/28/2013
Title:
REACTOR COOLANT SYSTEM (RCS) RESIDUAL HEAT REMOVAL (RHR) SHUTDOWN COOLING SYSTEM
- HOT SHUTDOWN TEXT 3.4.9 1 03/28/2013
Title:
REACTOR COOLANT SYSTEM (RCS) RESIDUAL HEAT REMOVAL (RER) SHUTDOWN COOLING SYSTEM
- COLD SHUTDOWN TEXT 3.4.10
Title:
(RCS) 11/04/2015 RCS PRESSURE AND TEMPERATURE (P/T) LIMITS 0
TEXT 3.4.11 0 11/15/2002
Title:
REACTOR COOLANT SYSTEM (RCS) REACTOR STEAM DOME PRESSURE TEXT 3.5.1 4 07/16/2014
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 CRCIC)
SYSTEM ECCS - SHUTDOWN TEXT 3.5.3 3 02/24/2014
Title:
EMERGENCY CORE COOLING SYSTEMS (ECCS) AND REACTOR CORE ISOLATION COOLING (RCIC)
SYSTEM RCIC SYSTEM TEXT 3.6.1.1 5 02/24/2014
Title:
PRIMARY CONTAINMENT TEXT 3.6.1.2 1 04/23/2008
Title:
CONTAINMENT SYSTEMS PRIMARY CONTAINMENT AIR LOCK Report Date: 11/05/15 Page4 Page
- of of 8 Report Date: 11/05/15
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Title:
TECHNICAL SPECIFICATION BASES UNIT 1 MANUAL TEXT 3.6.1.3 11 07/02/2014
Title:
CONTAINMENT SYSTEMS PRIMARY CONTAINMENT ISOLATION VALVES (PCIVS)
TEXT 3.6.1.4 1 04/23/2008
Title:
CONTAINMENT SYSTEMS CONTAINMENT PRESSURE TEXT 3.6.1.5 1 10/05/2005
Title:
CONTAINMENT SYSTEMS DRYWELL AIR TEMPERATURE TEXT 3.6.1.6 0 11/15/2002
Title:
CONTAINMENT SYSTEMS SUPPRESSION CHANBER-TO-DRYWELL VACUUM BREAKERS TEXT 3.6.2.1 2 04/23/2008
Title:
CONTAINMENT SYSTEMS SUPPRESSION POOL AVERAGE TEMPERATURE TEXT 3.6.2.2 0 11/15/2002
Title:
CONTAINMENT SYSTEMS SUPPRESSION POOL WATER. LEVEL TEXT 3.6.2.3 1 01/16/2006
Title:
CONTAINMENT SYSTEMS RESIDUAL HEAT REMOVAL (RER) SUPPRESSION POOL COOLING TEXT 3.6.2.4 0 11/15/2002
Title:
CONTAINMENT SYSTEMS RESIDUAL HEAT REMOVAL (RHR) SUPPRESSION POOL SPRAY TEXT 3.6.3.1 2 06/13/2006
Title:
CONTAINMENT SYSTEMS PRIMARY CONTAINMENT HYDROGEN RECOMBINERS TEXT 3.6.3.2 1 04/18/2005
Title:
CONTAINMENT SYSTEMS DRYWELL AIR FLOW SYSTEM TEXT 3.6.3.3 1 02/28/2013
Title:
CONTAINMENT SYSTEMS PRIMARY CONTAINMENT OXYGEN CONCENTRATION TEXT 3.6.4.1 11 11/06/2014
Title:
CONTAINMENT SYSTEMS SECONDARY CONTAINMENT Report Date: 11/05/15 Page~
Page 5 of of 8 Report Date: 11/05/15
SSES MANUJAL Manual Name:- TSB1
-[!:*Manual
Title:
TECHNICAL SPECIFICATION BASES UNIT 1 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 4 04/05/2010
Title:
PLANT SYSTEMS RESIDUAL HEAT REMOVAL SERVICEWATER_ CRHRSW) SYSTEM AND THE ULTIMATE HEAT SINK (UES)
TEXT 3.7.2 2 02/11/2009
Title:
PLANT SYSTEMS EMERGENCY SERVICE WATER (ESW) SYSTEM TEXT 3.7.3 1 01/08/2010
Title:
PLANT SYSTEMS CONTROL ROOM EMERGENCY OUTSIDE AIR SUPPLY (CREOAS) SYSTEM TEXT 3.7.4 0 11/15/2002
Title:
PLANT SYSTEMS CONTROL ROOM FLOOR COOLING SYSTEM TEXT 3.7.5 1 10/04/2007
Title:
PLANT SYSTEMS MAIN CONDENSER OFFGAS TEXT 3.7.6 2 04/23/2008
Title:
PLANT SYSTEMS MAIN TURBINE BYPASS SYSTEM TEXT 3.7.7 1 10/04/2007
Title:
PLANT SYSTEMS SPENT FUEL STORAGE POOL WATER LEVEL TEXT 3.7.8 0 04/23/2008
Title:
PLANT SYSTEMS TEXT 3.8.1 7 02/24/2014
Title:
ELECTRICAL POWER SYSTEMS AC SOURCES - OPERATING
- TEXT 3.8.2 0 11/15/2002 S-
Title:
ELECTRICAL POWER SYSTEMS AC SOURCES -SHUTDOWN Report Date: 11/05/15 Page6 of 8 Page 6 of Report Date: 11/05/15
SSE~S MANUAL
. Manual Name : TSBI Manual
Title:
TECHNICAL SPECIFICATION BASES UNIT 1 MANUAL TEXT 3.8.3 4 10/23/2013
Title:
ELECTRICAL POWER SYSTIJ *EMS DIESEL FUEL OIL, LUBE OIL, AND STARTING AIR TEXT 3.8.4 3 01/19/2009 Title : ELECTRICAL POWER SYST 'EMS DC SOURCES - OPERATING TEXT 3.8.5 1 12/14/2006
__Title: ELECTRICAL POWER SYST ?EMS DC SOURCES - SHUTDOWN TEXT 3.8.6 1 12/14/2006
Title:
ELECTRICAL POWER SYST PEMS BATTERY CELL PARAMVETERS TEXT 3.8.7 1 10/05/2005
Title:
- ELECTRICAL POWER SYST ?EMS DISTRIBUTION SYSTEMS - OPERATING TEXT 3.8.8 0 11/15/2002
Title:
ELECTRICAL POWER SYSI ?ENS DISTRIBUTION SYSTEMS - SHUTDOWN TEXT 3.9.1 0 11/15/2002
Title:
REFUELING C)PERATIONS REFUELING EQUIPMENT INTERLOCKS TEXT 3.9.2 1 09/01/2010
Title:
REFUELING C)PERATIONS REFUEL POSITION ONE-ROD-OUT INTERLOCK TEXT 3.9.3 0 11/15/2002
Title:
REFUELING C)PERATIONS CONTROL ROD POSITION TEXT 3.9.4 0 11/15/2002
Title:
REFUELING C)PERATIONS CONTROL ROD POSITION INDICATION TEXT 3.9.5 0 11/15/2002
Title:
REFUELING C)PERATIONS CONTROL ROD OPERABILITY - REFUELING TEXT 3.9.6 1 10/04/2007
Title:
REFUELING C)PERATIONS REACTOR PRESSURE VESSEL (RPV) WATER LEVEL Report Date: 11/05/15 Page Page 2 of 88 of Report Date: 11/05/15
- SEg MANUITAL Manual Name:s TSBI
?***Manual
Title:
TECHNICAL SPECIFICATION BASES UNIT 1 MANUAL TEXT 3.9.7 0 11/15/2002
Title:
REFUELING OPERATIONS RESIDUAL HEAT REMOVAL (RHR) - HIGH WATER LEVEL TEXT 3.9.8 0 11/15/2002
Title:
REFUELING OPERATIONS RESIDUAL HEAT REMOVAL (RER) - LOW WATER LEVEL TEXT 3.10.1 1 01/23/2008
Title:
SPECIAL OPERATIONS INSERVICE LEAK AND HYDROSTATIC TESTING OPERATION TEXT 3.10.2 0 11/15/2002
Title:
SPECIAL OPERATIONS REACTOR MODE SWITCH INTERLOCK TESTING TEXT 3.10.3 0 11/15/2002
Title:
SPECIAL OPERATIONS SINGLE CONTROL ROD WITHDRAWAL - HOT SHUTDOWN TEXT 3.10.4 0 11/15/2002
Title:
SPECIAL OPERATIONS SINGLE CONTROL ROD WITHDRAWAL - COLD SHUTDOWN S
TEXT 3.10.5 0 11/15/2002
Title:
SPECIAL OPERATIONS SINGLE CONTROL ROD DRIVE (CRD) REMOVAL - REFUELING TEXT 3.10.6 0 11/-15/2002
Title:
SPECIAL OPERATIONS MULTIPLE CONTROL ROD WITHDRAWAL - REFUELING TEXT 3.10.7 1 04/18/2006
Title:
SPECIAL OPERATIONS CONTROL ROD TESTING - OPERATING TEXT 3.10.8 1 04/12/2006
Title:
SPECIAL OPERATIONS SHUTDOWN MARGIN (SDM) TEST - REFUELING Report Date: 11/05/15 Page8 Page 8 of of 88 Report Date: 11/05/15
SUSQUEHANNA STEAM ELECTRIC STATION LIST OF EFFECTIVE SECTIONS (TECHNICAL SPECIFICATIONS BASES)
Section Title Revision TOC Table of Contents 23 B 2.0 SAFETY LIMITS BASES Page TS / B2.0-1 1 Page TS /B 2.0-2 3 Page TS /8B2.0-3 6 Page TS /B 2.0-4 4 Page"S / B 2.0-5 6 Page TS / B 2.0-6 1 Pages TS / B 2.0-7 through TS / B 2.0-9 1 B 3.0 LCO AND SR APPLICABILITY BASES Page TS / B3.0-1 I*
Pages TS / B 3.0-2 through TS/IB 3.0-4 0 .....
Pages TS / B 3.0-5 through TS / B 3.0-7 *,1 Page TS / B3.0-8 3 Pages TS / B 3.0-9 through TS I B 3.0-11 2 Page TS / B 3.0-11a 0 Page TS /B 3.0-12I Pages TS / B 3.0-13 through TS I2 Pages TS /B3.0-16 and TS 0 B 3.1 REACTIVITY CONTROL BAy Pages B83.1-1 trug 0 Page TS / B 3.1-5A 1 PagesTS /Ba3.1 d . 72 Pages B3.1-8 t 3.
Page 3. 20ghB 3.1-19 0 20 and TS / B 3.1-21 1 Pa
- 0
- 3.1-23I 3.1-240 Pa TS / B 3.1-25 through TS /8B3.1-27 1
- TS B 31-282
,= TS/ B 31-29I Pages B 3.1-30 through B 3.1-33 0 Pages TS I B 3.3-34 through TS I B 3.3-36 1 Page TS /8B3.1-37 2 Page TS I B 3.1-38 3 Pages TS /8B3.1-39 and TS /8B3.1-40 2 Page TS I B 3.1-40a 0 Pages TS /B 3.1-41 and TS / B3.1-42 2 Revision 121 SUSQUEHANNA SUSQUEHANNA - UNIT
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- Page TSI/B 3.2-6 3
_ ageB3.2-7 .. ...... _1__
Pages TS / B 3.2-8 and TS I B 3.2-9 3 Page TS / B 3.2.10 2 Page TS / B 3.2-11 3 Page TS I B 3.2-12 1 Page TS / B 3.2-13 2 B 3.3 INSTRUMENTATION Pages TS / B 3.3-1 through TS I B 3.3-4 1 Page TS I B 3.3-5 2 Page TS / B 3.3-6 I Page TS / B 3.3-7 3 Page TS I B 3.3-7a I Page TS / B 3.3-8 5 Pages TS I B 3.3-9 through TS / B 3.3-12 3 Pages TS I B 3.3-12a I Pages TS / B 3.3-12b and TS I B 3.3-12c 0 Page TS I B 3.3-13 1 Page TS / B 3.3-14 3 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 4 Page TS / B 3.3-19 1 Pages TS I B 3.3-20 through TS I B 3.3-22 2 Page TS / B 3.3-22a 0 Pages TS / B 3.3-23 and TS I B 3.3-24 2 Pages TS I B 3.3-24a and TS / B 3.3-24b 0 Page TS / B 3.3-25 3 Page TS / B 3.3-26 2 Page TS / B 3.3-27 1 Page TS / B 3.3-28 3 Page TSI/B 3.3-29 4 Page TS I B 3.3-30 3 Page TS / B 3.3-30a 0 SUSQUEHANNA-UNIT1 TS / B LOES-2 Revision 121
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Section Title Revision Pages B 3.3-79 through B 3.3-81 0 Page TS I B 3.3-82 2 Page B 3.3-83 0 Pages B 3.3-84 and B 3.3-85 1 Page B 3.3-86 0 Page B 3.3-87 1 Page B 3.3-88 0 Page B 3.3-89 1 Page TS I B 3.3-90 1 Page B 3.3-91 0.
Pages TS/IB 3.3-92 through TS IB 3.3-100 1
- Pages TSI/B 3.3-101 through TS/B 3.3-103 .0 PageTSI/B3.3-_104 .....-
Pages TS I B 3.3-1 05 and TS I B 3.3-106 0 Page TS / B3.3-107 I Page TS I B 3.3-108 0 Page TS I B 3.3-109 1 Pages TS I B 3.3-110 and TS/B83.3-111 0 Pages TS I B 3.3-112 and TS I B 3.3-1 12a I Pages TS / B 3.3-113 through TS / B 3.3-115 1 Page TS / B 3.3-116 3 Page TS / B 3.3-117 1 Pages TS I B 3.3-118 through TS I B 3.3-122 0 Pages TS I B 3.3-123 and TS I B 3.3-124 1 Page TS I B 3.3-124a 0 Page TS I B 3.3-125 0 Pages TS I B 3.3-126 and TS I B 3.3-127 1 Pages TS I B 3.3-1 28 through TS/ B 3.3-130 0 Page TS / B 3.3-131 1 Pages TS I B 3.3-1 32 through TS I B 3.3-1 34 0 Pages B 3.3-135 through B 3.3-137 0 Page TS / B 3.3-138 1 Pages B 3.3-1 39 through B 3.3-149 0 Pages TS I B 3.3-150 and TS I B 3.3-151 1 Pages TS I B 3.3-1 52 through TS/IB 3.3-154 2 Page TS I B 3.3-155 1 Pages TS I B 3.3-1 56 through TS I B 3.3-158 2 Pages TS / B 3.3-159 and TS I B 3.3-160 1 Page TS / B 3.3-161 2 Page TS I B 3.3-162 1 Page TS I B 3.3-163 2 Page TS I B 3.3-164 1 Pages TS I B 3.3-165 through TS I B 3.3-167 2 TSIBLOES-4 Revision 121 SUSQUEHANNA SUSQUEHANNA-UNIT1- UNIT I TS / B LOES-4 Revision 121
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SUSQUEHANNA STEAM ELECTRIC STATION LIST OF EFFECTIVE SECTIONS (TECHNICAL SPECIFICATIONS BASES)
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Rev. 4 RCS P/T" Limits B 3.4.10 B 3.4 REACTOR COOLAkNT SYSTEM (RCS)
B 3.4.10 RCS Pressure and Temperature (PIT) Limits BASES BACKGROUND All components of the RCS are designed to withstand effects of cyclic loads due to system pressure and temperature changes. These loads are introduced by startup (heatup) and shutdown (cooldown) operations, power transients, and reactor trips. This LCO limits the pressure and temperature changes during RCS heatup and cooldown, within the design assumptions and the stress limits for cyclic operation.
This Specification contains PIT limit curves for heatup, cooldown, and
....... in~servi~ce_ lea~kage and_ hyd~rosta~ti~c testing, and limits for the maximum rate of change of reactor coolant temperature. The heatup curve provides limits for both heatup and criticality.
Each PIT limit curve defines an acceptable region for normal operation. The usual use of the curves is operational guidance during heatup or cooldown maneuvering, when pressure and temperature indications are monitored and compared to the applicable curve to determine that operation is within the allowable region.
The LCO establishes operating limits that provide a margin to brittle failure of the reactor vessel and piping of the reactor coolant pressure boundary (RCPB). The vessel is the component most subject to brittle failure.
Therefore, the LCO limits apply mainly to the vessel.
10 CFR 50, Appendix G (Ref. 1), requires the establishment of P/T limits for material fracture toughness requirements of the RCPB materials.
Reference I requires an adequate margin to brittle failure during normal operation, anticipated operational occurrences, and system hydrostatic tests.
It mandates the use of the ASME Code,Section XI, Appendix G (Ref. 2).
The actual shift in the RTNDT of the vessel material will be established periodically by removing and evaluating the irradiated reactor vessel material specimens, in accordance with ASTM E 185 (Ref. 3) and Appendix H of 10 CFR 50 (Ref. 4). The operating PIT limit curves will be adjusted, as necessary, based on the evaluation findings and the recommendations of RG 1.99, "Radiation Embrittlement of Reactor Vessel Materials" (Ref. 5).
The calculations to determine neutron fluence will be developed using the BWRVIP RAMA code methodology, which is NRC approved and meets the intent of RG 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence" (Ref. 11). See FSAR Section 4.1.4.5 for determining fluence (Ref. 12).
(continued)
SUSQUEHANNA - UNIT I1SIB3449Rvso TS / B 3.4-49 Revision 3
Rev. 4 RCS P/T"Limits B 3.4.10 BASES BACKGROUND The P/T limit curves are composite curves established by superimposing (continued) limits derived from stress analyses of those portions of the reactor vessel and head that are the most restrictive. At any specific pressure, temperature, and temperature rate of change, one location within the reactor vessel will dictate the most restrictive limit. Across the span of the P/T limit curves, different locations are more restrictive, and, thus, the curves are composites of the most restrictive regions.
The heatup curve used to develop the P/T limit curve composite represents a different set of restrictions than the cooldown curve used to develop the P/T limit curve composite because the directions of the thermal gradients through the vessel-wall are reversed. The thermal gradient-reversal-alters the location of the tensile stress between the outer and inner walls.
The criticality limits include the Reference I requirement that they be at least 40°F above the heatup curve or the cooldown curve and not lower than the minimum permissible temperature for the inservice leakage and hydrostatic testing.
The consequence of violating the LCO limits is that the RCS has been operated under conditions that can result in brittle failure of the RCPB, possibly leading to a nonisolable leak or loss of coolant accident. In the event these limits are exceeded, an evaluation must be performed to determine the effect on the structural integrity of the RCPB components.
ASME Code,Section XI, Appendix E (Ref. 6), provides a recommended methodology for evaluating an operating event that causes an excursion outside the limits.
APPLICABLE The P/l" limits are not derived from Design Basis Accident (DBA) analyses.
SAFETY They are prescribed during normal operation to avoid encountering pressure, ANALYSES temperature, and temperature rate of change conditions that might cause undetected flaws to propagate and cause nonductile failure of the RCPB, a condition that is unanalyzed. Reference 7 establishes the methodology for determining the P/l" limits. Since the P/T limits are not derived from any DBA, there are no acceptance limits related to the P/T limits. Rather, the P/T limits are acceptance limits themselves since they preclude operation in an unanalyzed condition.
RCS P/T limits satisfy Criterion 2 of the NRC Policy Statement (Ref. 8).
(continued)
SUSQUEHANNA - UNIT 1 SIB3450Rvso TS / B 3.4-50 Revision 1
Rev. 4 RCS P/T Limits 8 3.4.10 BASES APPLICABLE The Effective Full Power Years (EFPY) shown on the curves are SAFETY approximations of the ratio of the energy that has been and is anticipated to ANALYSES be generated in a year to the energy that could have been generated ifthe (continued) unit ran at original thermal power rating of 3293 MWT for the entire year.
These values are based on fluence limits that are not to be exceeded.
LCO The elements of this LCO are:
- a. RCS pressure and temperature are to the right of the applicable curves specified in Figures 3.4.10-1 through 3.4.10-3 and within the applicable heat-up or cool down rate specified in SR 3.4.10.1 during RCS heatup, cooldown, and inservice leak and hYdrostatic testing; b._ The temperature difference between the reactorivessel bottom~he~ad coolant and the reactor pressure vessel (RPV) coolant is _<1450 F during recirculation pump startup, and during increases in THERMAL POWER or loop flow while operating at low THERMAL POWER or loop flow;
- c. The temperature difference between the reactor coolant in the respective recirculation loop and in the reactor vessel is < 50°F during recirculation pump startup, and during increases in THERMAL POWER or loop flow while operating at low THERMAL POWER or loop flow;
- d. RCS pressure and temperature are to the right of the criticality limits specified in Figure 3.4.10-3 prior to achieving criticality; and
- e. The reactor vessel flange and the head flange temperatures are > 70°F when tensioning the reactor vessel head bolting studs.
These limits define allowable operating regions and permit a large number of operating cycles while also providing a wide margin to nonductile failure.
The P/T limit composite curves are calculated using the worst case of material properties, stresses, and temperature change rates anticipated under all heatup and cooldown conditions. The design calculations account for the reactor coolant fluid temperature impact on the inner wall of the vessel and the temperature gradients through the vessel wall. Because these fluid temperatures drive the vessel wall temperature gradient, monitoring reactor coolant temperature provides a conservative method of ensuring the P/T limits are not exceeded. Proper monitoring of vessel temperatures to assure compliance with brittle fracture temperature limits and vessel thermal stress limits during normal heatup and cooldown, and.
during inservice leakage and hydrostatic testing, is established in PPL Calculation EC-062-0573 (Ref. 9). For P/T curves A, B, and C, the bottom head drain line coolant temperature should be monitored and maintained to the right of the most limiting curve.
(continued)
SUSQUEHANNA - UNIT 1 S/B3451Rvso TS / B 3.4-51 Revision 3
Rev. 4 RCS P/T Limits B 3.4.10 BASES LCO Curve A must be used for any ASME Section III Design Hydrostatic Tests (continued) performed at unsaturated reactor conditions. Curve A may also be used for ASME Section XI inservice leakage and hydrostatic testing when heatup and cooldown rates can be limited to 20°F in a one-hour period.
Curve A is based on pressure stresses only. Thermal stresses are assumed to be insignificant. Therefore, heatup and cooldown rates are limited to 20°F in a one-hour period when using Curve A to ensure minimal thermal stresses. The recirculation loop suction line temperatures should be monitored to determine the temperature change rate.
Curves B and C are to be used for non-nuclear and nuclear heatup and cooldown, respectively. In addition, Curve B may be used for ASME Section Xl in*service leakage and hydrostatic_ testing, but not for ASME Section III Design Hydrostatic Tests performed at unsaturated reactor conditions. Heatup and cooldown rates are limited to 100°F in a one-hour period when using Curves B and C. This limits the thermal gradient through the vessel wall, which is used to calculate the thermal stresses in the vessel wall. Thus, the LCO for the rate of coolant temperature change limits the thermal stresses and ensures the validity of the P/T curves. The vessel belt-line fracture analysis assumes a l0 0°F/hr coolant heatup or cooldown rate in the beitline area. The 100°F limit in a one-hour period applies to the coolant in the beltline region, and takes into account the thermal inertia of the vessel wall. Steam dome saturation temperature (TsAT), as derived from steam dome pressure, should be monitored to determine the beltline temperature change rate at temperatures above 212°F. At temperatures below 212°F, the recirculation loop suction line temperatures should be monitored.
During heatups and cooldowns, the reactor vessel could experience a vacuum (negative pressure) at low temperatures (unsaturated conditions) and low rates of temperature change. Under a vacuum, the vessel wall would experience a uniform compressive loading, which would counteract the tensile stress due to any thermal gradients through the vessel wall. To ensure the margin to brittle fracture is no less than at any other pressure, Curves A, B, and C require a minimum vessel metal temperature of 70°F when the reactor vessel is at a negative pressure.
(continued)
SUSQUEHANNA - UNIT I1Sl TS / B 3.4-52
.- 2Rvso Revision 3
Rev. 4 RCS PIT Limits B] 3.4.10 BASES LCO Violation of the limits places the reactor vessel outside of the bounds of the (continued) stress analyses and can increase stresses in other RCS components. The consequences depend on several factors, as follows:
- a. The severity of the departure from the allowable operating pressure temperature regime or the severity of the rate of change of temperature;
- b. The length of time the limits were violated (longer violations allow the temperature gradient in the thick vessel walls to become more pronounced); and
- c. The existences, sizes, and orientations of flaws in the vessel material.
APPLICABILITY The potential for violating a P/T limit exists at all times. For example, PIT limit violations could result from ambient temperature conditions that result in the reactor vessel metal temperature being less than the minimum allowed temperature for boltup. Therefore, this LCO is applicable even when fuel is not loaded in the core.
ACTIONS A.1 and A.2 Operation outside the PIT limits while in MODES 1, 2, and 3 must be corrected so that the RCPB is retumed to a condition that has been verified
- 1 by stress analyses.
The 30 minute Completion Time reflects the urgency of restoring the parameters to within the analyzed range. Most violations will not be severe, and the activity can be accomplished in this time in a controlled manner.
B]esides restoring operation within limits, an evaluation is required to determine if RCS operation can continue. The evaluation must verify the RCPB integrity remains acceptable and must be completed if continued operation is desired. Several methods may be used, including comparison with pre-analyzed transients in the stress analyses, new analyses, or inspection of the components.
(continued)
SUSQUEHANNA - UNIT I TS / B 3.4-52a Rvso 0 Revision
Rev. 4 RCS P/T- Limits B 3.4.10 BASES ACTIONS A.1I and A.2. (continued)
ASME Code, Section Xl, Appendix E (Ref. 6), may be used to support the evaluation. However, its use is restricted to evaluation of the vessel beltline.
The 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Completion Time is reasonable to accomplish the evaluation of a mild violation. More severe violations may require special, event specific stress analyses or inspections. A favorable evaluation must be completed if continued operation is desired.
Condition A is modified by a Note requiring Required Action A.2 be completed whenever the Condition is entered. The Note emphasizes the need to perforrm the evaluation of the effects of the excursion Outside the
-- --- -- - allowable limits*-Restoration alone per Required Action A. 1 is-insufficient because higher than analyzed stresses may have occurred and may have affected the RCPB integrity.
B.1I and B.2 If a Required Action and associated Completion Time of Condition A are not met, the plant must be placed in a lower MODE because either the RCS remained in an unacceptable P/T region for an extended period of increased stress, or a sufficiently severe event caused entry into an unacceptable region. Either possibility indicates a need for more careful examination of the event, best accomplished with the RCS at reduced pressure and temperature. With the reduced pressure and temperature conditions, the possibility of propagation of undetected flaws is decreased.
(continued)
SUSQUEHANNA -UNIT1IT .- 3Rvso TS ! B 3.4-53 Revision 2
Rev. 4 RCS P/T Limits B 3.4.10 BASES ACTIONS B.1 and B.2 (continued)
Pressure and temperature are reduced by placing the plant in 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 in 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.
C.1 and C.2 Operation outside the P/T limits in other than MODES 1, 2, and 3 (including defueled conditions) must be corrected so that the RCPB is returned to a condition that-has-been verified by stress analyses. The Required Action must be initiated without delay and continued until the limits are restored.
Besides restoring the P/T limit parameters to within limits, an evaluation is required to determine if RCS operation is allowed. This evaluation must verify that the RCPB integrity is acceptable and must be completed before approaching criticality or heating up to > 200°F. Several methods may be used, including comparison with pre-.analyzed transients, new analyses, or inspection of the components. ASME Code,Section XI, Appendix E (Ref. 6),
may be used to support the evaluation; however, its use is restricted to evaluation of the beltiine.
SURVEILLANCE SR 3.4.10.1 REQUIREMENTS Verification that operation is within limits (i.e., to the right of the applicable curves in Figures 3.4.10-1 through 3.4.10-3) is required every 30 minutes when RCS pressure and temperature conditions are undergoing planned changes. This Frequency is considered reasonable in view of the control room indication available to monitor RCS status. Also, since temperature rate of change limits are specified in hourly increments, 30 minutes permits a reasonable time for assessment and correction of minor deviations.
Surveillance for heatup, cooldown, or inservice leakage and hydrostatic testing may be discontinued when the criteria given in the relevant plant procedure for ending the activity are satisfied.
This SR has been modified with a Note that requires this Surveillance to be performed only during system heatup and cooldown operations and inservice leakage and hydrostatic testing.
(continued)
SUSQUEHANNA - UNIT 1 TS SIB3454Rvso
/ B 3.4-54 Revision 2
Rev. 4 RCS P/T Limits B 3.4.10 BASES SURVEILLANCE SR 3.4.10.1 (continued)
REQUIREMENTS Notes to the acceptance criteria for heatup and cooldown rates ensure that more restrictive limits are applicable when the P/T limits associated with hydrostatic and inservice testing are being applied.
SR 3.4.10.2 A separate limit is used when the reactor is approaching criticality.
Consequently, the RCS pressure and temperature must be verified within the appropriate limits (i.e., to the right of the criticality curve in
- -Figure 3.4:10-3)-before withdrawing~ control rods that will make the-reactor critical.
Performing the Surveillance within 15 minutes before control rod withdrawal for the purpose of achieving criticality provides adequate assurance that the limits will not be exceeded between the time of the Surveillance and the time of reactor criticality. Although no Surveillance Frequency is specified, the requirements of SR 3.4.10.2 must be met at all times when the reactor is critical.
SR 3.4.10.3 and SR 3.4.10.4 Differential temperatures within the applicable limits ensure that thermal stresses resulting from the startup of an idle recirculation pump will not exceed design allowances. In addition, compliance with these limits ensures that the assumptions of the analysis for the startup of an idle recirculation loop (Ref. 10) are satisfied.
Performing the Surveillance within 15 minutes before starting the idle recirculation pump provides adequate assurance that the limits will not be exceeded between the time of the Surveillance and the time of the idle pump start.
An acceptable means of demonstrating compliance with the temperature differential requirement in SR 3.4.10.4 is to compare the temperatures of the operating recirculation loop and the idle loop. If both loops are idle, compare the temperature difference between the reactor coolant within the idle loop to be started and coolant in the reactor vessel.
(continued)
SUSQUEHANNA - UNIT I1SIB3455Rvso TS / B 3.4-55 Revision 2
Rev. 4 RCS P/T Limits B 3.4.10 BASES SURVEILLANCE SR 3.4.10.3 and SR 3.4.10.4 (continued)
REQUIREMENTS SR 3.4.10.3 has been modified by a Note that requires the Surveillance to be performed only in MODES 1, 2, 3, and 4. In MODE 5, the overall stress on limiting components is lower. Therefore, AT limits are not required. The Note also states the SR is only required to be met during a recirculation pump start-up, because this is when the stresses occur.
SR 3.4.10.5 and SR 3.4.10.6 Differential temperatures within the applicable limits ensure that thermal
--stres-es resultingg from increases irn-THERMAL POWER or recirc~lation loop flow during single recirculation loop operation will not exceed design allowances. Performing the Surveillance within 15 minutes before beginning such an increase in power or flow rate provides adequate assurance that the limits will not be exceeded between the time of the Surveillance and the time of the change in operation.
An acceptable means of demonstrating compliance with the temperature differential requirement in SR 3.4.10.6 is to compare the temperatures of the operating recirculation loop and the idle loop.
Plant specific startup test data has determined that the bottom head is not subject to temperature stratification at power levels > 27% of RTP and with single loop flow rate _>21,320 gpm (50% of rated loop flow). Therefore, SR 3.4.10.5 and SR 3.4.10.6 have been modified by a Note that requires the Surveillance to be met only under these conditions. The Note for SR 3.4.10.6 further limits the requirement for this Surveillance to exclude comparison of the idle loop temperature if the idle loop is isolated from the RPV since the water in the loop cannot be introduced into the remainder of the. Reactor Coolant System.
SR 3.4.10.7. SR 3.4.10.8, and SR 3.4.10.9 Limits on the reactor vessel flange and head flange temperatures are generally bounded by the other P/T limits during system heatup and cooldown. However, operations approaching MODE 4 from MODE 5 and in MODE 4 with RCS temperature less than or equal to certain specified values require assurance that these temperatures meet the LCO limits.
(continued)
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Rev. 4 RCS P/T Limits B 3.4.10 BASES SURVEILLANCE SR 3.4.10.7, SR 3.4.10.8, and SR 3.4.10.9 (continued)
REQUIREMENTS The flange temperatures must be verified to be above the limits 30 minutes before and while tensioning the vessel head bolting studs to ensure that once the head is tensioned the limits are satisfied. When in MODE 4 with RCS temperature < 80°F, 30 minute checks of the flange temperatures are required because of the reduced margin to the limits. When in MODE 4 with RCS temperature < 100°F, monitoring of the flange temperature is required every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> to ensure the temperature is within the specified limits.
The 30 minute Frequency reflects the urgency of maintaining the temperatures-within limits, and also limits the time that the temperature limits could be exceeded. The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Frequency is reasonable based on the rate of temperature change possible at these temperatures.
REFERENCES 1. 10 CFR 50, Appendix G.
- 2. ASME, Boiler and Pressure Vessel Code,Section XI, Appendix G.
- 3. ASTM E 185-73.
- 5. Regulatory Guide 1.99, Revision 2, May 1988.
- 6. ASME, Boiler and Pressure Vessel Code, Section Xl, Appendix E.
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Rev. 4 RCS P/T Limits B 3.4.10 BASES REFERENCES 7. Licensed Topical Reports:
(continued)
- a. Structural Integrity Associates Report No. SlR-05-044, Revision 1-A, "Pressure-Temperature Limits Report Methodology for Boiling Water Reactors," June 2013.
- b. Structural Integrity Associates Report No. 0900876.401, Revision 0-A, "Linear Elastic Fracture Mechanics Evaluation of GE BWR Water Level Instrument Nozzles for Pressure-Temperature Curve Evaluations," May 2013.
- 8. Final Policy Statement on Technical Specifications Improvements,
- - ----- - July-22, -1993 (58 FR 39132). ---.......-
- 9. PPL Calculation EC-062-0573, "Study to Support the Bases Section of Technical Specification 3.4.10."
- 10. FSAR, Section 15.4.4.
- 11. Regulatory Guide 1.190, March 2001.
- 12. ESAR, Section 4.1.4.5.
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