ML080360497
| ML080360497 | |
| Person / Time | |
|---|---|
| Site: | Susquehanna  |
| Issue date: | 01/24/2008 |
| From: | Susquehanna |
| To: | Gerlach R Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| 2008-4299 | |
| Download: ML080360497 (54) | |
Text
Jan. 24, 2008 Page 1 of 2 MANUAL HARD COPY DISTRIBUTION DOCUMENT TRANSMITTAL 2008-4299 USER INFORMATION:
GERLACH*ROSE M EMPL#:028401 CA#: 0363 Address: NUCSA2 Phone#: 254-3194 rAT\TMqMTTTAT, T1\NTfrMATT\N TO: GERLACH*ROSE M 01/24/2008 LOCATION: USNRC FROM: NUCLEAR RECORDS DOCUMENT CONTROL CENTER (NUCSA-2)
THE FOLLOWING CHANGES 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 TOC will be issued with the updated material.
TSB2 - TECHNICAL SPECIFICATIONS BASES UNIT 2 MANUAL REMOVE MANUAL TABLE OF CONTENTS DATE: 12/17/2007 ADD MANUAL TABLE OF CONTENTS DATE: 01/23/2008 CATEGORY: DOCUMENTS TYPE: TSB2
,&-D (
ugj'
Jan. 24, 2008 Page 2 of 2 ID: TEXT 3.10.1 REMOVE: REV:0 ADD: REV: 1 CATEGORY: DOCUMENTS TYPE:TSB2 ID: TEXT 3.6.1.3 ADD: REV: 8 REMOVE: REV:7 CATEGORY: DOCUMENTS TYPE: TSB2 ID: TEXT LOES REMOVE: REV:87 ADD: REV: 88 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: TSB2 Manual
Title:
TECHNICAL SPECIFICATIONS BASES UNIT 2 MANUAL Table Of Contents Issue Date: 01/23/2008 Procedure Name Rev Issue Date Change ID Change Number TEXT LOES 88 01/23/2008
Title:
LIST OF EFFECTIVE SECTIONS TEXT TOC 12 10/04/2007
Title:
TABLE OF CONTENTS TEXT 2. 1.1 3 10/04/2007
Title:
SAFETY LIMITS. (SLS) REACTOR CORE SLS TEXT 2.1.2 1 10/04/2007
Title:
SAFETY LIMITS (SLS) REACTOR COOLANT SYSTEM '(RCS) PRESSURE SL TEXT 3.0 2 ý,10/12/2006 -
Title:
LIMITING CONDITION FOR OPERATION,(LCO) APPLICABILITY TEXT 3.1.1 ",03/24/2005
Title:
REACTIVITY CONTROL SYSTEMS SHUTDOWN MARGIN (SDM)
TEXT 3.1.2 0 11/18/2002
Title:
REACTIVITY CONTROL SYSTEMS REACTIVITY ANOMALIES TEXT 3.1.3 1 07/06/2005
Title:
REACTIVITY ,CONTROL SYSTEMS CONTROL ROD OPERABILITY TEXT 3.1.4 3 09/29/2006
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 2 03/24/2005
Title:
REACTIVITY CONTROL SYSTEMS ROD PATTERN CONTROL Report Date: 01/23/08 Page I Pagel of of .8 Report Date: 01/23/08
.SSES MANUAL Manual Name: TSB2 Manual
Title:
TECHNICAL SPECIFICATIONS BASES UNIT 2 MANUAL 0 TEXT 3.1.7 3 10/04/2007
Title:
REACTIVITY CONTROL SYSTEMS STANDBY LIQUID CONTROL (SLC) SYSTEM TEXT 3.1.8 2 10/04/2007
Title:
REACTIVITY CONTROL SYSTEMS SCRAM DISCHARGE VOLUME (SDV) VENT AND DRAIN VALVES TEXT 3.2.1 3 04/09/2007
Title:
POWER DISTRIBUTION LIMITS AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)
TEXT 3.2.2 2 04/03/2007
Title:
POWER DISTRIBUTION LIMITS MINIMUM CRITICAL POWER RATIO (MCPR)
TEXT 3.2.3 1 10/04/2007
Title:
POWER DISTRIBUTION LIMITS LINEAR HEAT GENERATION RATE LHGR TEXT 3.3.1.1 3 04/09/2007
Title:
INSTRUMENTATION REACTOR PROTECTION .SYSTEM (RPS) INSTRUMENTATION TEXT 3.3.1.2 1 04/09/2007
Title:
INSTRUMENTATION SOURCE RANGE MONITOR (SRM) INSTRUMENTATION TEXT 3.3.2.1 2 04/09/2007
Title:
INSTRUMENTATION CONTROL ROD BLOCK INSTRUMENTATION TEXT 3.3.2.2 0 11/18/2002
Title:
INSTRUMENTATION FEEDWATER - MAIN TURBINE HIGH WATER LEVEL TRIP INSTRUMENTATION TEXT 3.3.3.1 6 12/17/2007
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 0 11/18/2002
Title:
INSTRUMENTATION END OF CYCLE RECIRCULATION PUMP TRIP (EOC-RPT) INSTRUMENTATION Page 2 of . Report Date: 01/23/08
SSES MANUAL 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 3 07/06/2005
Title:
INSTRUMENTATION EMERGENCY CORE COOLING SYSTEM (ECCS) INSTRUMENTATION TEXT 3.3.5.2 0 11/18/2002
Title:
INSTRUMENTATION REACTOR CORE ISOLATION COOLING (RCIC) SYSTEM INSTRUMENTATION TEXT 3.3.6.1 3 10/04/2007
Title:
INSTRUMENTATION PRIMARY CONTAINMENT ISOLATION INSTRUMENTATION TEXT 3.3.6.2 2 10/04/2007
Title:
INSTRUMENTATION SECONDARY CONTAINMENT ISOLATION INSTRUMENTATION TEXT 3.3.7.1 1 10/04/2007
Title:
INSTRUMENTATION CONTROL ROOM EMERGENCY OUTSIDE AIR SUPPLY (CREOAS) SYSTEM INSTRUMENTATION 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 3 04/09/2007
Title:
REACTOR COOLANT SYSTEM (RCS) RECIRCULATION LOOPS OPERATING TEXT 3.4.2 0 11/18/2002
Title:
REACTOR COOLANT SYSTEM (RCS) JET PUMPS TEXT 3.4.3 1 01/16/2006
Title:
REACTOR COOLANT SYSTEM (RCS) SAFETY/RELIEF VALVES (S/RVS)
TEXT 3.4,4 0 11/18/2002
Title:
REACTOR COOLANT SYSTEM (RCS) RCS OPERATIONAL LEAKAGE Page 3 of .8 Report Date: 01/23/08
S SES. MANUAL Manual Name: TSB2 Manual
Title:
TECHNICAL SPECIFICATIONS BASES UNIT 2 MANUAL TEXT 3.4.5 2 04/19/2007
Title:
REACTOR COOLANT SYSTEM (RCS) RCS PRESSURE ISOLATION VALVE (PIV) LEAKAGE TEXT 3.4.6 1 04/18/2005
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 1 04/18/2005
Title:
REACTOR COOLANT SYSTEM (RCS) RESIDUAL HEAT REMOVAL (RHR) SHUTDOWN COOLING SYSTEM
- HOT SHUTDOWN TEXT 3.4.9 0 11/18/2002
Title:
REACTOR COOLANT SYSTEM (RCS) RESIDUAL HEAT REMOVAL (RHR) SHUTDOWN COOLING SYSTEM
- COLD SHUTDOWN TEXT 3.4.10 2 05/1-0/2006 1
Title:
REACTOR COOLANT SYSTEM (RCS) RCS PRESSURE AND TEMPERATURE (P/T) LIMITS TEXT 3.4.11 0 11/18/2002
Title:
REACTOR COOLANT SYSTEM (RCS) REACTOR STEAM DOME PRESSURE TEXT 3. 5.1 3 01/16/2006
Title:
EMERGENCY CORE COOLING SYSTEMS (ECCS) AND REACTOR CORE ISOLATION COOLING (RCIC)
SYSTEM ECCS - OPERATING TEXT 3.5.2 0 11/18/2002
Title:
EMERGENCY CORE COOLING SYSTEMS (ECCS) AND REACTOR CORE ISOLATION COOLING (RCIC)
SYSTEM ECCS - SHUTDOWN TEXT 3.5.3 1 04/18/2005
Title:
EMERGENCY CORE COOLING SYSTEMS (ECCS) AND REACTOR CORE ISOLATION COOLING (RCIC)
SYSTEM RCIC SYSTEM TEXT 3.6.1.1 2 10/04/2007
Title:
CONTAINMENT SYSTEMS PRIMARY CONTAINMENT TEXT 3.6.1.2 0 11/18/2002
Title:
CONTAINMENT SYSTEMS PRIMARY CONTAINMENT AIR LOCK Report Date: 01/23/08 Page4 4 Page of 88 Report Date: 01/23/08
SSES MANUAL Manual Name: TSB2 Manual
Title:
TECHNICAL SPECIFICATIONS BASES UNIT 2 MANUAL TEXT 3.6.1.3 8 01/23/2008
Title:
CONTAINMENT SYSTEMS PRIMARY CONTAINMENT ISOLATION VALVES (PCIVS)
TEXT 3.6.1.4 0 11/18/2002
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 11/18/2002
Title:
CONTAINMENT SYSTEMS SUPPRESSION CHAMBER-TO-DRYWELL VACUUM BREAKERS TEXT 3.6.2.1 2 12/17/2007
Title:
CONTAINMENT SYSTEMS SUPPRESSION POOL AVERAGE TEMPERATURE TEXT 3.6.2.2 0 11/18/2002
Title:
CONTAINMENT SYSTEMS SUPPRESSION POOI WATER LEVEL TEXT 3.6.2.3 1 01/16/2006
Title:
CONTAINMENT SYSTEMS RESIDUAL HEAT REMOVAL (RHR) SUPPRESSION POOL COOLING TEXT 3.6.2.4 0 11/18/2002
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 0 11/18/2002
Title:
CONTAINMENT SYSTEMS PRIMARY CONTAINMENT OXYGEN CONCENTRATION TEXT 3.6.4.1 7 10/04/2007
Title:
CONTAINMENT SYSTEMS SECONDARY CONTAINMENT Report Date: 01/23/08 Page 55 Page of of .8 Report Date: 01/23/08
SSES MANUAL Manual Name: TSB2 Manual
Title:
TECHNICAL SPECIFICATIONS BASES UNIT 2 MANUAL TEXT 3.6.4.2 2 01/03/2005
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 1 12/17/2007
Title:
PLANT SYSTEMS RESIDUAL HEAT REMOVAL SERVICE WATER (RHRSW) SYSTEM AND THE ULTIMATE HEAT SINK (UHS)
TEXT 3.7.2 1 11/09/2004
Title:
PLANT SYSTEMS EMERGENCY SERVICE WATER (ESW) SYSTEM TEXT 3.7.3 0 11/18/2002
Title:
PLANT SYSTEMS CONTROL ROOM EMERGENCY OUTSIDE AIR SUPPLY (CREOAS) SYSTEM TEXT 3.7.4 0 11/18/2002
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 1 01/17/2005
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. 8.1 5 12/17/2007
Title:
ELECTRICAL POWER SYSTEMS AC SOURCES - OPERATING TEXT 3.8.2 0 11/18/2002
Title:
ELECTRICAL POWER SYSTEMS AC SOURCES - SHUTDOWN TEXT 3.8.3 0 11/18/2002
Title:
ELECTRICAL POWER SYSTEMS DIESEL FUEL OIL, LUBE OIL, AND STARTING AIR Page 6 of . Report Date: 01/23/08
SSES MANUAL Manual Name: TSB2 Manual
Title:
TECHNICAL SPECIFICATIONS BASES UNIT 2 MANUAL TEXT 3. 8.4 2 12/14/2006
Title:
ELECTRICAL POWER SYSTEMS DC SOURCES - OPERATING TEXT 3.8.5 1 12/14/2006
Title:
ELECTRICAL POWER SYSTEMS DC SOURCES - SHUTDOWN TEXT 3.8.6 1 12/14/2006
Title:
ELECTRICAL POWER SYSTEMS BATTERY CELL PARAMETERS TEXT 3.8.7 3 03/31/2006
Title:
ELECTRICAL POWER SYSTEMS DISTRIBUTION SYSTEMS - OPERATING TEXT 3.8.8 0 11/18/2002
Title:
ELECTRICAL POWER SYSTEMS DISTRIBUTION SYSTEMS - SHUTDOWN TEXT 3. 9.1 0 11/18/2002
Title:
REFUELING OPERATIONS REFUELING EQUIPMENT INTERLOCKS TEXT 3.9.2 0 11/18/2002
Title:
REFUELING OPERATIONS REFUEL POSITION ONE-ROD-OUT INTERLOCK TEXT 3.9.3 0 11/18/2002
Title:
REFUELING OPERATIONS CONTROL ROD POSITION TEXT 3.9.4 0 11/18/2002
Title:
REFUELING OPERATIONS CONTROL ROD POSITION INDICATION TEXT 3.9.5 0 11/18/2002
Title:
REFUELING OPERATIONS CONTROL ROD OPERABILITY - REFUELING TEXT 3.9.6 1 10/04/2007
Title:
REFUELING OPERATIONS REACTOR PRESSURE VESSEL (RPV) WATER LEVEL TEXT 3.9.7 0 11/18/2002
Title:
REFUELING OPERATIONS RESIDUAL HEAT REMOVAL (RHR) - HIGH WATER LEVEL Report Date: 01/23/08 Page Pagel7 of8 8 of Report Date: 01/23/08
SSES MANUAL Manual Name: TSB2 Manual
Title:
TECHNICAL SPECIFICATIONS BASES UNIT 2 MANUAL TEXT 3.9.8 0 11/18/2002
Title:
REFUELING OPERATIONS RESIDUAL HEAT REMOVAL (RHR) - LOW WATER LEVEL TEXT 3.10.1 1 01/23/2008
Title:
SPECIAL OPERATIONS INSERVICE LEAK AND HYDROSTATIC TESTING OPERATION TEXT 3.10.2 0 11/18/2002
Title:
SPECIAL OPERATIONS REACTOR MODE SWITCH INTERLOCK TESTING TEXT 3.10.3 0 11/18/2002
Title:
SPECIAL OPERATIONS SINGLE CONTROL ROD WITHDRAWAL - HOT SHUTDOWN TEXT 3.10.4 0 11/18/2002
Title:
SPECIAL OPERATIONS SINGLE CONTROL ROD WITHDRAWAL COLD SHUTDOWN TEXT 3.10.5 0 11/1-8/2002
Title:
SPECIAL OPERATIONS SINGLE CONTROL ROD DRIVE (CRD) REMOVAL - REFUELING TEXT 3.10.6 0 11/18/2002
Title:
SPECIAL OPERATIONS MULTIPLE CONTROL ROD WITHDRAWAL - REFUELING TEXT 3.10.7 1 03/24/2005
Title:
SPECIAL OPERATIONS CONTROL ROD TESTING - OPERATING TEXT 3.10.8 2 04/09/2007
Title:
SPECIAL OPERATIONS SHUTDOWN MARGIN (SDM) TEST - REFUELING Report Date:. 01/23/08 Page Page88 of 8 of3 Report Date:. 01/23/08
SUSQUEHANNA STEAM ELECTRIC STATION LIST OF EFFECTIVE SECTIONS (TECHNICAL SPECIFICATIONS BASES)
Section Title Revision TOC Table of Contents 12 B 2.0 SAFETY LIMITS BASES Page TS / B 2.0-1 1 Page TS / B 2.0-2 3 Page TS / B 2.0-3 4 Page TS / B 2.0-4 6 Page TS / B 2.0-5 1 Pages TS / B 2.0-6 through TS I B 2.0-8 1 B 3.0 LCO AND SR APPLICABILITY BASES Page TS / B 3.0-1 1 Pages TS I B 3.0-2 through TS / B 3.0-4 0 0'
Pages TS I B 3.0-5 through TS / B 3.0-7 11' Pages TS / B 3.0-8 through TS / B 3.0-9 2 Page TS / B 3.0-10 1 Page TS / B 3.0-11 , 2 Page TS / B 3.0-1 la 0 Page TS / B 3.0-12 . 1 Pages TS / B 3.0-13 through TS"IB 3.0-15" 2 Pages TS / B 3.0-16 and TS_/ B .0-j7 0 B 3.1 REACTIVITY CONTROL<BASES Pages B 3.1-1 throughB3.1,-.4 0 Page TS / B .3.1-5 ,-,1 Pages TS / B<3A1>6andTS I B 3.1-7 2 Pages B 3.1-8"through\3B/3.1-13 0 Page TS 3-1-14' 1 Pages-B3-,1-15\thrbdugh B 3.1-21 0 Page TS" B'3:1-22 0 Page TS IB 3.1-23 1 TS7B/3.BPa1-24 0 (pKageaT4/fB 3.1-25 1
'age TS / B 3.1-26 0
"¢Page/TS / B 3.1-27 1 Page TS / B 3.1-28 2 Page TS / 3.1-29 1 Pages B 3.1-30 through B 3.1-33 0 Pages TS / B 3.1.34 through TS / B 3.1-36 1 Pages TS / B 3.1-37 and TS / B 3.1-38 2 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 Pages TS / B 3.1-43 1 Page TS / B 3.1-44 0 SUSQUEHANNA -UNIT 2 , TS / B LOES-1 Revision 88
SUSQUEHANNA STEAM ELECTRIC STATION LIST OF EFFECTIVE SECTIONS (TECHNICAL SPECIFICATIONS BASES)
Section Title Revision Page TS / B 3.1-45 3 Page TS / B 3.1-46 0 Page TS / B 3.1-47 1 Pages TS / B 3.1-48 and TS / B 3.1-49 1 Page B 3.1-50 0 Page TS / B 3.1-51 2 B 3.2 POWER DISTRIBUTION LIMITS BASES Page TS / B 3.2-1 2 Page TS / B 3.2-2 1 Page TS / B 3.2-3 3 Page TS / B 3.2-4 1 Pages TS / B 3.2-5 and TS / B 3.2-6 3 Page TS / B 3.2-7 2 Pages TS / B 3.2-8 and TS / B 3.2-9 4 Page TS / B 3.2-10 2 Pages TS / B 3.2-11 through 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 Pages TS / B 3.3-7 through TS-/ B 3.3-12 3 Page TS / B 3.3-13 -* 2 PageTS / B 3.3-14 3 Pages TS / B 3.3-15 and TS / B 3.3-16-- 2 Pages TS / B 3.3-17 and TS / B 3.3-18 3 Pages TS / B 3.3-19 through TS / B 3.3-27 2 Pages TS / B 3.3-28 through TS / B 3.3-30 3 Page TS / B 3.3-31 2 Page TS / B 3.3-32 4 Page TS / B 3.3-33 3 Page TS / B 3.3-34 2 Pages TS / B 3.3-34a though 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 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 3.3-54e 0 Pages B 3.3-55 through B 3.3-63 0 Pages TS / B 3.3-64 and TS / B 3.3-65 2 Page TS / B 3.3-66 4 Page TS / B 3.3-67 3 Page TS / B 3.3-68 4 Page TS / B 3.3.69 5 Page TS / B 3.3-70 4 SUSQUEHANNA - UNIT 2 TS / B LOES-2 Revision 88
SUSQUEHANNA STEAM ELECTRIC STATION LIST OF EFFECTIVE SECTIONS (TECHNICAL SPECIFICATIONS BASES)
Section Title Revision 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 Page TS / B 3.3-75a 6 Pages TS / B 3.3-75 b and TS / B 3.3-75c 5 Pages B 3.3-76 and TS / B 3.3-77 0 Page TS / B 3.3-78 1 Pages B 3.3-79 through B 3.3-91 0 Pages TS / B 3.3-92 through TS / B 3.3-103 1 Page TS / B 3.3-104 2 Pages TS / B 3.3-105 and TS / B 3.3-106 1 Page TS / 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 Page TS / B 3.3-114 1 Page TS / B 3.3-115 through TS / B 3.3-118 2 Pages TS / B 3.3-119 through TS / B 3.3-120 1 Pages TS / B 3.3-121 and TS / B 3-.3-122 2 Page TS / B 3.3-123 1 Page TS / B 3.3-124 2 Page TS / B 3.3-124a 0 PageTS / 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 B 3.3-135 through B 3.3-137 0 Page TS / B 3.3-138 1 Pages B 3.3-139 through B 3.3-149 0 Pages TS/ B 3.3-150 and TS / B 3.3-151 1 Page TS / B 3.3-152 2 Page TS / B 3.3-153 1 Page 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 3.3-161 1 Page TS / B 3.3-162 1 Page TS / B 3.3-163 2 Page TS / B 3.3-164 1 Pages TS / B 3.3-165 and TS / B 3.3-166 2 Pages TS / B 3.3-167 and TS / B 3.3-168 1 SUSQUEHANNA - UNIT 2 TS / B LOES-3 Revision 88
SUSQUEHANNA STEAM ELECTRIC STATION LIST OF EFFECTIVE SECTIONS (TECHNICAL SPECIFICATIONS BASES)
Section Title Revision Pages TS / B 3.3-169 and TS / B 3.3-170 2 Pages TS / B 3.3-171 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 2 Page TS / B 3.3-186 1 Pages TS B 3.3-187 and TS/ B 3.3-188 2 Pages TS I B 3.3-189 through TS / B 3.3-191 1 Page TS / B 3.3-192 0 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 1 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 1 Pages TS / B 3.4-3 and TS I B 3.4-4 4 Pages TS / B 3.4-5 and TS / B 3.4-9 3 Pages B 3.4-10 through B 3.4-14 0 Page TS / B 3.4-15 1 Pages TS / B 3.4-16 through TS / B 3.4-18 2 Pages B 3.4-19 through B 3.4-27 0 Pages TS / B 3.4-28 and TS / B 3.4-29 1 Pages B 3.4.-30 through B 3.3-31 0 Page TS / B 3.4-32 1 Page B 3.4-33 and B 3.4-34 0 Pages TS / B 3.4-35 and TS / B 3.4-36 1 Page TS / B 3.4-37 2 Page B 3.4-38 1 Pages B 3.4-39 and B 3.4-40 0 Page TS / B 3.4-41 1 Pages B 3.4-42 through B 3.4-48 0 Revision 88 UNIT 2 TS I B LOES-4 SUSQUEHANNA SUSQUEHANNA -UNIT-2 TS / B LOES-4 Revision 88
SUSQUEHANNA STEAM ELECTRIC STATION LIST OFEFFECTIVE SECTIONS (TECHNICAL SPECIFICATIONS BASES)
Section Title Revision Page TS / B 3.4-49 3 Pages TS / B 3.4-50 through TS / B 3.4-52 2 Page TS / B 3.4-53 1 Pages TS / B 3.4-54 and TS / B 3.4-55 2 Page TS / B 3.4-56 1 Page TS / B 3.4-57 2 Pages TS / B 3.4-58 through TS / B 3.4-60 1 B 3.5 ECCS AND RCIC BASES Pages TS / B 3.5-1 and TS / B 3.5-2 1 Pages TS / B 3.5-3 through TS / B 3.5-6 2 Pages TS / B 3.5-7 through TS / B 3.5-10 1 Pages TS / B 3.5-11 and TS / B 3.5-12 2 Pages TS / B 3.6-13 and TS / B 3.5-14 1 Pages TS / B 3.5-15 and TS / B.3.5-16 2 Page TS / B 3.5-17 3 Page TS / B 3.5-18 1 Pages B 3.5-19 through B 3.5-24 0 Pages TS / B 3.5-25 through TS /B 3.5-27 1 Pages B 3.5-28 through B 3.5-31 0 B 3.6 CONTAINMENT SYSTEMS BASES Page TS / B 3.6-1 2 Page TS / B 3.6-1a 3 Page-TS / B 3.6-2 and TS / B 3.6-3 3 Page TS / B 3.6-4 4 Page TS / B 3.6-5 3 Page TS / B 3.6-6 4 Pages TS / B 3.6-6a and TS / B 3.6-6b 2 Page TS / B 3.6-6c 0 Pages B 3.6-7 through B 3.6-14 0 Page TS / B 3.6-15 3 Page'TS / B 3.6-15a 0 Page TS / B 3.6-15b 2 Pages TS / B 3.6-16 and TS / B 3.6-17 2 Page TS / B 3.6-17a 0 Pages TS / B 3.6-18 and TS / B 3.6-19 1 Page TS / B 3.6-20 2 Page TS / B 3.6-21 3 Pages TS / B 3.6-21a and TS / B 3.6-21b 0 Pages TS / B 3.6-22 and TS / B 3.6-23 2 Pages TS / B 3.6-24 and TS / B 3.6-25 1 Page TS / B 3.6-26 3 Page TS / B 3.6-27 3 Page TS / B 3.6-28 6 SUSQUEHANNA - UNIT 2 TS / B LOES-5 Revision 88
SUSQUEHANNA STEAM ELECTRIC STATION LIST OF EFFECTIVE SECTIONS (TECHNICAL SPECIFICATIONS BASES)
Section Title Revision Page TS / B 3.6-29 3 Page TS / B 3.6-29a 0 Page TS / B 3.6-30 2 Page TS / B 3.6-31 3 Page TS / B 3.6-32 1 Page TS / B 3.6-33 2 Page TS / B 3.6-34 1 Page TS / B 3.6-35 3 Pages TS / B 3.6-36 and TS I B 3.6-37 2 Page TS / B 3.6-38 3 Page TS / B 3.6-39 7 Pages B 3.6-40 through B 3.6-42 0 Pages TS / B 3.6-43 and TS / B 3.6-44 1 Page TS / B 3.6-45 2 Pages TS / B 3.6-46 through TS / B 3.6-50 1 Page TS / B 3.6-51 2 Pages B 3.6-52 through B 3.6-55 0 Pages TS / B 3.6-56 and TS / B 3.6-57 2 Pages B 3.6-58 through B 3.6-62 0 Pages TS / B 3.6-63 and TS / B 3.6-64 1 Pages B 3.6-65 through B 3.6 0 Pages B 3.6-69 through B 3.6-71 1 Page TS / B 3.6-72 2 Pages TS / B 3.6-73 and TS / B 3.6-74 1 Pages B 3.6-75 and B 3.6-76 0 Page TS / B 3.6-77 1 Pages B 3.6-78 through B 3.6-82 0 Page TS / B 3.6-83 3 Page TS / B 3.6-84 2 Page TS / B 3.6-85 4 Page TS / B 3.6-86 through TS / B 3.6-87a 2 Page TS / B 3.6-88 4 Page TS / B 3.6-89 2 Page TS / B 3.6-90 3 Pages TS / B 3.6-91 through TS / B 3.6-95 1 Page TS / B 3.6-96 2 Pages TS / B 3.6-97 and TS / B 3.6-98 1 Page TS / B 3.6-99 2 Page TS / B 3.6-99a 0 Pages TS / B 3.6-100 and TS / B 3.6-101 1 Pages TS / B 3.6-102 and TS / B 3.6-103 2 Page TS / B 3.6-104 3 Page TS / B 3.6-105 2 Page TS / B 3.6-106 3 SUSQUEHANNA - UNIT 2 TS / B LOES-6 Revision 88
SUSQUEHANNA STEAM ELECTRIC STATION LIST OF EFFECTIVE SECTIONS (TECHNICAL SPECIFICATIONS BASES)
Section Title Revision B 3.7 PLANT SYSTEMS BASES Page TS / B 3.7-1 2 Page TS / B 3.7-2 3 Pages TS / B 3.7-3 through TS / B 3.7-6 2 Page TS / B 3.7-6a 2 Pages TS / B 3.7-6b and TS / B 3.7-6c 0 Page TS / B 3.7-7 2 Page TS / B 3.7-8 1 Pages B 3.7-9 through B 3.7-11 0 Pages TS / B 3.7-12 and TS / B 3.7-13 1 Pages TS / B 3.7-14 through TS / B 3.7-18 2 Page TS / B 3.7-18a 0 Pages TS / B 3.7-19 through TS / B 3.7-26 1 Page TS / B 3.7-24 1 Pages TS / B 3.7-25 and TS / B 3.7-26 0 Pages TS / B 3.7-27 through TS I B 3.7-29 2 Page TS / B 3.7-30 1 Page TS / B 3.7-31 1 Page TS / B 3.7-32 0 Page TS / B 3.7-33 1 B 3.8 ELECTRICAL POWER SYSTEMS BASES Pages B 3.8-1 through B 3.8-3 0 Page TS / B 3.8-4 1 Pages-TS / B 3.8-4a and TS I B 3.8-4b 0 Page TS / B 3.8-5 2 Page TS / B 3.8-6 1 Pages B 3.8-7 and B 3.8-8 0 Page TS / B 3.8-9 2 Pages TS / B 3.8-10 and TS / B 3.8-11 1 Pages B 3.8-12 through B 3.8-18 0 Page TS / B 3.8-19 1 Pages B 3.8-20 through B 3.8-22 0 Page TS / B 3.8-23 1 Page B 3.8-24 0 Pages TS / B 3.8-25 and TS / B 3.8-26 1 Pages B 3.8-27 through B 3.8-37 0 Page TS / B 3.8-38 1 Pages TS / B 3.8-39 through TS / B 3.8-55 0 Pages TS / B 3.8-56 through TS / B 3.8-64 2 Page TS / B 3.8-65 3 Page TS / B 3.8-66 4 Pages TS / B 3.8-67 and TS / B 3.8-68 3 Page TS / B 3.8-69 4 SUSQUEHANNA - UNIT 2 TS / B LOES-7 Revision 88
SUSQUEHANNA STEAM ELECTRIC STATION LIST OF EFFECTIVE SECTIONS (TECHNICAL SPECIFICATIONS BASES.)
Section Title Revision Pages TS / B 3.8-70 through TS / B 3.8-83 1 Pages TS / B 3.8-83A through TS / B 3.8-83D 0 Pages B 3.8-84 through B 3.8-85 0 Page TS / B 3.8-86 1 Page TS / B 3.8-87 2 Pages TS / B 3.8-88 through TS / B 3.8-93 1 Pages B 3.8-94 through B 3.8-99 0 B 3.9 REFUELING OPERATIONS BASES Pages TS / B 3.9-1 and TS / B 3.9-2 1 Page TS / B 3.9-2a 1 Pages TS / B 3.9-3 and TS / B 3.9-4 1 Pages B 3.9-5 through B 3.9-18 0 Pages TS / B 3.9-19 through TS / B 3.9-21 1 Pages B 3.9-22 through B 3.9-30 0 B 3.10 SPECIAL OPERATIONS BASES Page TS / B 3.10-1 2 Pages TS / B 3.10-2 through TS / B 3.10-5 1 Pages B 3.10-6 through B 3.10-32 0 Page TS / B 3.10-33 2 Page B 3.10-34 0 Page B 3.10-35 1 Pages B 3.10-36 and B 3.10-37 0 Page13 3.10 1 Page TS / B 3.10-39 2 TSB2 Text LOES.doc 1/16/08 Revision 88 TS I B LOES-8 SUSQUEHANNA - UNIT SUSQUEHANNA -
UNIT 22 TS / B LOES-8 Revision 88
PPL Rev. 8 PCIVs B 3.6.1.3 B 3.6 CONTAINMENT SYSTEMS B 3.6.1.3 Primary Containment Isolation Valves (PCIVs)
BASES BACKGROUND The function of the PCIVs, in combination with other accident mitigation systems, is to limit fission product release during and following postulated Design Basis Accidents (DBAs) to within limits. Primary containment isolation within the time limits specified for those isolation valves designed to close automatically ensures that the release of radioactive material to the environment will be consistent with the assumptions used in the analyses for a DBA.
The OPERABILITY requirements for PCIVs help ensure that an adequate primary containment boundary is maintained during and after an accident by minimizing potential paths to the environment. Therefore, the OPERABILITY requirements provide assurance that primary containment function assumed in the safety analyses will be'maintained. For PCIVs, the primary containment isolation function is that the valve must be able to close (automatically or manually)_and/or remain closed, and maintain leakage within that assumed in the DBA LOCA Dose Analysis. These isolation devices are either passive or active (automatic). Manual valves, de-activated automatic valves secured in their closed position (including check valves with flow through the valve secured), blind flanges, and closed systems are considered passive devices. The OPERABILITY requirements for closed systems are discussed in Technical Requirements Manual (TRM) Bases 3.6.4. Check valves, or other automatic valves designed to close without operator action following an accident, are considered active devices. Two barriers in series are provided for each penetration so that no single credible failure or malfunction of an active component can result in a loss of isolation or leakage that exceeds limits assumed in the safety analyses. One of these barriers may be a closed system.
For each division of H20 2 Analyzers, the lines, up to and including the first normally closed valves within the H20 2 (continued)
SUSQUEHANNA - UNIT 2 TS / B 3.6-15 Revision 3
PPL Rev. 8 PCIVs B 3.6.1.3 BASES BACKGROUND Analyzer panels, are extensions of primary containment (continued) (i.e., closed system), and are required to be leak rate tested in accordance with the Leakage Rate Test Program. The H20 2 Analyzer closed system boundary is identified in the Leakage Rate Test Program. The closed system boundary consists of those components, piping, tubing, fittings, and valves, which meet the guidance of Reference 6. The closed system provides a secondary barrier in the event of a single failure of the PCIVs, as described below. The closed system boundary between PASS and the H20 2 Analyzer system ends at the process sampling solenoid operated isolation valves between the systems (SV-22361, SV-22365, SV-22366, SV-22368, and SV-22369). These solenoid operated isolation valves do not fully meet the guidance of Reference 6 for closed system boundary valves in that they are not powered from a Class 1E power source. However, based upon a risk determination, operating these valves as closed system boundary valves is not risk significant. These valves also form the end of the Seismic Category I boundary between the systems. These process sampling solenoid operated isolation valves are normally closed and are required to be leak rate tested in accordance with the Leakage Rate Test Program as part of the closed system for the H20 2 Analyzer system. These valves are "closed system boundary valves!' and may be opened under administrative control, as delineated in Technical Requirements Manual (TRM)
Bases 3.6.4. Opening of these valves to permit testing of PASS in Modes 1, 2, and 3 is permitted in accordance with TRO 3.6.4.
Each H 2 0 2 Analyzer Sampling line penetrating primary containment has two PCIVs, located just outside primary containment. While two PCIVs are provided on each line, a single active failure of a relay in the control circuitry for these valves, could result in both valves failing to close or failing to remain closed. Furthermore, a single failure (a hot short in the common raceway to all the valves) could simultaneously affect all of the PCIVs within a H 2 0 2 Analyzer division. Therefore, the containment isolation barriers for these penetrations consist of two PCIVs and a closed system. For situations where one or both PCIVs are inoperable, the ACTIONS to be taken are similar to the ACTIONS for a single PCIV backed by a closed system.
(continued)
SUSQUEHANNA - UNIT 2 TS / B 3.6-15a Revision 0
PPL Rev. 8 PCIVs B 3.6.1.3 BASES BACKGROUND The drywell vent and purge lines are 24 inches in diameter; the (continued) suppression chamber vent and purge lines are 18 inches in diameter. The containment purge valves are normally maintained closed in MODES 1, 2, and 3 to ensure the primary containment boundary is maintained. The outboard isolation valves have 2 inch bypass lines around them for use during normal reactor operation.
The RHR Shutdown Cooling return line containment penetrations
{X-13A(B)}are provided with a normally closed gate valve
{HV-251 F015A(B)} and a normally open globe valve
{HV-251 F017A(B)} outside containment and a testable check valve {HV-251 F05OA(B)} with a normally closed parallel air operated globe valve {HV-251F122A(B)} inside containment.
The gate valve is manually opened and automatically isolates upon a containment isolation signal from the Nuclear Steam Supply Shutoff System or RPV low level 3 when the RHR System is operated in the Shutdown Cooling Mode only. The LPCI subsystem is an operational mode of the RHR System and uses the same injection lines to the RPV as the Shutdown Cooling Mode.
The design of these containment penetrations is unique in that some valves are containment isolation valves while others perform the function of pressure isolation valves. In order to meet the 10 CFR 50 Appendix J leakage testing requirements, the HV-251 F015A(B) and the closed system outside containment are the only barriers tested in accordance with the Leakage Rate Test Program. Since these containment penetrations {X-13A and X-13B} include a containment isolation valve outside containment that is tested in accordance with 10 CFR 50 Appendix J require-ments and a closed system outside containment that meets the requirements of USNRC Standard Review Plan 6.2.4 (September 1975), paragraph 11.3.e, the containment isolation provisions for these penetrations provide an acceptable alternative to the explicit requirements of 10 CFR 50, Appendix A, GDC 55.
Containment penetrations X-13A(B) are also high/low pressure system interfaces. In order to meet the requirements to have two (2) isolation valves between the high pressure and low pressure systems, the HV-251F050A(B), HV-251F122A(B), and HV-251 F01 5A(B) valves are used to meet this requirement and are tested in accordance with the pressure test program.
(continued)
SUSQUEHANNA - UNIT 2 TS / B 3.6-15b Revision 2
PPL Rev. 8 PCIVs B 3.6.1.3 BASES APPLICABLE The PCIVs LCO was derived from the assumptions related to SAFETY ANALYSES minimizing the loss of reactor coolant inventory, and establishing the primary containment boundary during major accidents. As part of the primary containment boundary, PCIV OPERABILITY supports leak tightness of primary containment.
Therefore, the safety analysis of any event requiring isolation of primary containment is applicable to this LCO.
The DBAs that result in a release of radioactive material within primary containment are a LOCA and a main steam line break (MSLB). In the analysis for each of these accidents, it is assumed that PCIVs are either closed or close within the required isolation times following event initiation. This ensures that potential paths to the environment through PCIVs (including primary containment purge valves) are minimized. The closure time of the main steam isolation valves (MSIVs) for a MSLB outside primary containment is a significant variable from a radiological standpoint. The MSIVs are required to close within 3 to 5 seconds since the 5 second closure time is assumed in the analysis. The safety analyses assume that the purge valves were closed at event initiation. Likewise, it is assumed that the primary containment is isolated such that release of fission products to the environment is controlled.
The DBA analysis assumes that within the required isolation time leakage is terminated, except for the maximum allowable leakage rate, La.
The single failure criterion required to be imposed in the conduct of unit safety analyses was considered in the original design of the primary containment purge valves. Two valves in series on each purge line provide assurance that both the supply'and exhaust lines could be isolated even if a single failure occurred.
The primary containment purge valves may be unable to close in the environment following a LOCA. Therefore, each of the purge valves is required to remain closed during MODES 1, 2, and 3 except as permitted under Note 2 of SR 3.6.1.3.1. In this case, the single failure criterion remains applicable to the primary containment purge valve (continued)
SUSQUEHANNA - UNIT 2 TS / B 3.6-16 Revision 2
PPL Rev. 8 PCIVs B 3.6.1.3 BASES APPLICABLE due to failure in the control circuit associated with each valve.
SAFETY ANALYSIS The primary containment purge valve design precludes a single (continued) failure from compromising the primary containment boundary as long as the system is operated in accordance with this LCO.
Both H20 2 Analyzer PCIVs may not be able to close given a single failure in the control circuitry of the valves. The single failure is caused by a "hot short" in the cables/raceway to the PCIVs that causes both PCIVs for a given penetration to remain open or to open when required to be closed. This failure is required to be considered in accordance with IEEE-279 as discussed in FSAR Section 7.3.2a. However, the single failure criterion for containment isolation of the H20 2 Analyzer penetrations is satisfied by virtue of the combination of the associated PCIVs and the closed system formed by the H20 2 Analyzer piping system as discussed in the BACKGROUND section above.
The closed system boundary between PASS and the H20 2 Analyzer system ends at the process sampling solenoid operated isolation valves between the systems (SV-22361, SV-22365, SV:
22366, SV-22368, and SV-22369). The closed system is not fully qualified to the guidance of Reference 6 in that the closed system boundary valves between the H 2 0 2 system and PASS are not powered from a-Class 1E power source. However, based upon a risk determination, the use of these valves is considered to have no risk significance. This exemption to the requirement of Reference 6 for the closed system boundary is documented in License Amendment No. 170.
PCIVs satisfy Criterion 3 of the NRC Policy Statement. (Ref. 2)
LCO PCIVs form a part of the primary containment boundary. The PCIV safety function is related to minimizing the loss of reactor coolant inventory and establishing the primary containment boundary during a DBA.
The power operated, automatic isolation valves are required to have isolation times within limits and actuate on an (continued)
SUSQUEHANNA - UNIT 2 TS / B 3.6-17 Revision 2
PPL Rev. 8 PCIVs B 3.6.1.3 BASES LCO automatic isolation signal. The valves covered by this LCO are (continued) listed in Table B 3.6.1.3-1.
The normally closed PCIVs are considered OPERABLE when manual valves are closed or open in accordance with appropriate administrative controls, automatic valves are in their closed position, blind flanges are in place, and closed systems are intact. These passive isolation valves and devices are those listed in Table B 3.6.1.3-1.
Purge valves with resilient seals, secondary containment bypass
.valves, MSIVs, and hydrostatically tested valves must meet additional leakage rate requirements. Other PCIV leakage rates are addressed by LCO 3.6.1.1, "Primary Containment," as Type B or C testing.
This LCO provides assurance that the PCIVs will perform their designed safety functions to minimize the loss of reactor coolant inventory and establish the primary containment boundary during accidents.
APPLICABILITY In MODES 1, 2, and 3, a DBA could cause a release of radioactive material to primary containment. In MODES 4 and 5, the probability and consequences of these events are reduced due to the pressure and temperature limitations of these MODES. Therefore, most PCIVs are not required to be (continued)
SUSQUEHANNA - UNIT 2 TS / B 3.6-17a Revision 0
PPL Rev. 8 PCIVs B 3.6.1.3 BASES APPLICABILITY OPERABLE and the primary containment purge valves are (continued) not required to be closed in MODES 4 and 5. Certain valves, however, are required to be OPERABLE to prevent inadvertent reactor vessel draindown. These valves are those whose associated instrumentation is required to be OPERABLE per LCO 3.3.6.1, "Primary Containment Isolation Instrumentation."
(This does not include the valves that isolate the associated instrumentation.)
ACTIONS The ACTIONS are modified by a Note allowing penetration flow path(s) to be unisolated intermittently under administrative controls. These controls consist of stationing a dedicated operator at the controls of the valve, who is in continuous communication with the control room. In this way, the penetration can be rapidly isolated when a need for primary containment isolation is indicated.
A second Note has been added to provide clarification that, for the purpose of this LCO, separate Condition entry is allowed for each penetration flow path. This is acceptable, since the Required Actionis for each Condition provide appropriate compensatory actions for each inoperable PCIV. Complying with the RequiredActions may allow for continued operation, and subsequent inoperable PCIVs are governed by subsequent Condition entry and application of associated Required Actions.
The ACTIONS are modified by Notes 3 and 4. Note 3 ensures that appropriate remedial actions are taken, if necessary, if the affected system(s) are rendered inoperable by an inoperable PCIV (e.g., an Emergency Core Cooling System subsystem is inoperable due to a failed open test return valve). Note 4 ensures appropriate remedial actions are taken when the primary containment leakage limits are exceeded. Pursuant to LCO 3.0.6, these actions are not required even when the associated LCO is not met. Therefore, Notes 3 and 4 are added to require the proper actions be taken.
A. 1 and A.2 With one or more penetration flow paths with one PCIV inoperable except for purge valve leakage not within limit, (continued)
SUSQUEHANNA - UNIT 2 TS / B 3.6-18 Revision I
PPL Rev. 8 PCIVs B 3.6.1.3 BASES ACTIONS A.1 and A.2 (continued) the affected penetration flow paths must be isolated. The method of isolation must include the use of at least one isolation barrier that cannot be adversely affected by a single active failure. Isolation barriers that meet this criterion are a closed and de-activated automatic valve, a closed manual valve, a blind flange, and a check valve with flow through the valve secured.
For a penetration isolated in accordance with Required Action A.1, the device used to isolate the penetration should be the closest available valve to the primary containment. The Required Action must be completed within the 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> Completion Time (8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> for main steam lines). The Completion Time of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is reasonable considering the time required to isolate the penetration and the relative importance of supporting primary containment OPERABILITY during MODES 1, 2, and 3. For main steam lines, an 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> Completion Time is allowed. The Completion Time of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> for the main steam lines allows a period of time to restore the MSIVs to OPERABLE status given the fact that MSIV closure will result in isolation of the main steam I.ine(s) and a potential for plant shutdown.
For affected penetrations that have been isolated in accordance with Required Action A. 1, the affected penetration flow path(s).
must be verified to be isolated on a periodic basis. This is necessary to ensure that primary containment penetrations required to be isolated following an accident, and no longer capable of being automatically isolated, will be in the isolation position should an event occur. This Required Action does not require any testing or device manipulation. Rather, it involves verification that those devices outside containment and capable of potentially being mispositioned are in the correct position. The Completion Time of "once per 31 days for isolation devices outside primary containment" is appropriate because the devices are operated under administrative controls and the probability of their misalignment is low. For the devices inside primary containment, the time period specified "prior to entering MODE 2 or 3 from MODE 4, if primary containment was de-inerted while in MODE 4, if not performed within the previous 92 days" is based on engineering judgment and is considered reasonable in view of the inaccessibility of the devices and other administrative controls ensuring that device misalignment is an unlikely possibility.
(continued)
SUSQUEHANNA - UNIT 2 TS / B 3.6-19 Revision 1
PPL Rev. 8 PCIVs B 3.6.1.3 BASES ACTIONS A.1 and A.2 (continued)
Condition A is modified by a Note indicating that this Condition is only applicable to those penetration flow paths with two PCIVs except for the H20 2 Analyzer penetrations. For penetration flow paths with one PCIV, Condition C provides the appropriate Required Actions. For the H20 2 Analyzer penetrations, Condition D provides the appropriate Required Actions.
Required Action A.2 is modified by a Note that applies to isolation devices located in high radiation areas, and allows them to be verified by use of administrative means. Allowing verification by administrative means is considered acceptable, since access to these areas is typically restricted. Therefore, theprobability of misalignment of these devices, once they have been verified to be in the proper position, is low.
B. 1 With one or more penetration flow paths with two PCIVs inoperable except for purge valve'leakage not within limit, either the inoperable PCIVs must be restored to OPERABLE status or the affected penetration flow path must be isolated within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
The method of isolation must include the use of at least one isolation barrier that cannot be adversely affected by a single active failure. Isolation barriers that meet this criterion are a closed and de-activated automatic valve, a closed manual valve, and a blind flange. The 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Completion Time is consistent with-the ACTIONS of LCO 3.6.1.1.
Condition B is modified by a Note indicating this Condition is only applicable to penetration flow paths with two PCIVs except for the H2 0 2 Analyzer penetrations. For penetration flow paths with one PCIV, Condition C provides the appropriate Required Actions. For the H20 2 Analyzer penetrations, Condition D provides the appropriate Required Actions.
C.1 and C.2 With one or more penetration flow paths with one PCIV inoperable, the inoperable valve must be restored to (continued)
SUSQUEHANNA - UNIT 2 TS / B 3.6-20 Revision 2
PPL Rev. 8 PCIVs B 3.6.1.3 BASES ACTIONS C.1 and C.2 (continued)
OPERABLE status or the affected penetration flow path must be isolated. The method of isolation must include the use of at least one isolation barrier that cannot be adversely affected by a single active failure. Isolation barriers that meet this criterion are a closed and de-activated automatic valve, a closed manual valve, and a blind flange. A check valve may not be used to isolate the affected penetration. Required Action C.1 must be completed within the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Completion Time. The Completion Time of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> is reasonable considering the relative stability of the closed system (hence, reliability) to act as a penetration isolation boundary and the relative importance of supporting primary containment OPERABILITY during MODES 1, 2, and 3. The closed system must meet the requirements of Reference 6. For conditions where the PCIV and the closed system are inoperable, the Required Actions of TRO 3.6.4, Condition B apply. For the Excess Flow Check Valves (EFCV), the Completion Time of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is reasonable considering the instrument and the small pipe diameter of penetration (hence, reliability) to act as a penetration isolation boundary and the small pipe diameter of the affected penetrations. In the event the affected penetration flow path is isolated inaccordance with Required Action C.1, the affected penetration must be verified to be isolated on a periodic basis. This is necessary to ensure that primary containment penetrations required -to be isolated following an accident are isolated. The Completion Time of once per 31 days for verifying each affected penetration is isolated is appropriate because the valves are operated under administrative controls and the probability of their misalignment is low.
Condition C is modified by a Note indicatingthat this Condition is only applicable to penetration flow paths with only one PCIV. For penetration flow paths with two PCIVs and the H20 2 Analyzer penetration, Conditions A, B, and D provide the appropriate Required Actions.
Required Action C.2 is modified by a Note that applies to valves and blind flanges located in high radiation areas and allows them to be verified by use of administrative means. Allowing verification by administrative means is considered acceptable, since access to these areas is typically (continued)
SUSQUEHANNA - UNIT 2 TS / B 3.6-21 Revision 3
PPL Rev. 8 PCIVs B 3.6.1.3 BASES ACTIONS C.1 and C.2 (continued) restricted. Therefore, the probability of misalignment of these valves, once they have been verified to be in the proper position, is low.
D.1 and D.2 With one or more H20 2 Analyzer penetrations with one or both PCIVs inoperable, the inoperable valve(s) must be restored to OPERABLE status or the affected penetration flow path must be isolated. The method of isolation must include the use of at least one isolation barrier that cannot be adversely affected by a single active failure. Isolation barriers that meet this criterion are a closed and de-activated automatic valve, a closed manual valve, and a blind flange. A check valve may not be used to isolate the affected penetration. Required Action D.1 must be completed within the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Completion Time. The Completion Time of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> is reasonable considering the unique design of the H20 2 Analyzer penetrations. The containment isolation barriers for these penetrations consist of two PCIVs and a closed system. In addition, the Completion Time of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> is reasonable considering the relative stability of the closed system (hence, reliability) to act as a penetration. isolation boundary and the relative importance of supporting primary containment OPERABILITY duringMODES 1, 2, and 3. In the event the affected penetration flow path is isolated in accordance with Required Action D.1, the affected penetration must be verified to be isolated on a periodic basis. This is necessary to ensure that primary containment penetrations required to be isolated following an accident are isolated. The Completion Time of once per 31 days for verifying each affected penetration is isolated is appropriate because the valves are operated under administrative controls and the probability of their misalignment is low.
When an H20 2 Analyzer penetration PCIV is to be closed and deactivated in accordance with Condition D, this must be accomplished by pulling the fuse for the power supply, and either determinating the power cables at the solenoid valve, or jumpering of the power side of the solenoid to ground.
(continued)
SUSQUEHANNA - UNIT 2 TS / B 3.6-21 a Revision 0
PPL.Rev. 8 PCIVs B 3.6.1.3 BASES ACTIONS D.1 and D.2 (continued)
The OPERABILITY requirements for the closed system are discussed in Technical Requirements Manual (TRM)
Bases 3.6.4. In the event that either one or both of the PCIVs and the closed system are inoperable, the Required Actions of TRO 3.6.4, Condition B apply.
Condition D is modified by a Note indicating that this Condition is only applicable to the H20 2 Analyzer penetrations.
(continued)
SUSQUEHANNA - UNIT 2 TS / B 3.6-21 b Revision 0
PPL Rev. 8 PCIVs B 3.6.1.3 BASES ACTIONS (continued) E. 1 With the secondary containment bypass leakage rate not within limit, the assumptions of the safety analysis may not be met.
Therefore, the leakage must be restored to within limit within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. Restoration can be accomplished by isolating the penetration that caused the limit to be exceeded by use of one closed and de-activated automatic valve, closed manual valve, or blind flange. When a penetration is isolated, the leakage rate for the isolated penetration is assumed to be the actual pathway leakage through the isolation device. If two isolation devices are used to isolate the penetration, the leakage rate is assumed to be the lesser actual pathway leakage of the two devices. The 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> Completion Time is reasonable considering the time required to restore the leakage by isolating the penetration and the relative importance of secondary containment bypass leakage to the overall containment function.
F. 1 In the event oneor more containment purge valves are not within the purge valve leakage limits, purge valve leakage must be restored to within limits. The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Completion Time is reasonable, considering that one containment purge valve remains closed, except as controlled by SR 3.6.1.3.1 so that a gross breach of containment does not exist.
G.1 and G.2 If any Required Action and associated Completion Time cannot be met in MODE 1, 2, or 3, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and to MODE 4 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.
(continued)
SUSQUEHANNA - UNIT 2 TS / B 3.6-22 Revision 2
PPL Rev. 8 PCIVs B 3.6.1.3 BASES ACTIONS (continued) H.1 and H.2 If any Required Action and associated Completion Time cannot be met, the unit must be placed in a condition in which the LCO does not apply. If applicable, action must be immediately initiated to suspend operations with a potential for draining the reactor vessel (OPDRVs) to minimize the probability of a vessel draindown and subsequent potential for fission product release.
Actions must continue until OPDRVs are suspended or valve(s) are restored to OPERABLE status. If suspending an OPDRV would result in closing the residual heat removal (RHR). shutdown cooling isolation valves, an alternative Required Action is provided to immediately initiate action to restore the valve(s) to OPERABLE status. This allows RHR to remain in service while actions are being taken to restore the valve.
SURVEILLANCE REQUIREMENTS SR 3.6.1.3.1 This SR ensures that the primary containment purge valves are closed as required or, if open, open .for an allowable reason. If a purge valve is open in violation of this SR, the valve is considered inoperable. If the inoperable valve is not otherwise known to have excessive leakage when closed, it is not considered to have leakage outside of limits. The SR is also modified by Note 1, stating that primary containment purge valves are only required to be closed in MODES 1, 2, and 3. If a LOCA inside primary containment occurs in these MODES, the purge valves may not be capable of closing before the pressure pulse affects systems downstream of the purge valves, or the release of radioactive material will exceed limits prior to the purge valves closing. At other times when the purge valves are required to be capable of closing (e.g., during handling of irradiated fuel), pressurization concerns are not present and the purge valves are allowed to be open. The SR is modified by Note 2 stating that the SR is not required to be met when the purge valves are open for the stated reasons. The Note states that these valves may be opened for inerting, de-inerting, pressure control, ALARA or air quality considerations for personnel entry, or Surveillances that require the valves to be open. The vent and purge valves are capable of closing in the environment following (continued)
SUSQUEHANNA - UNIT 2 TS / B 3.6-23 Revision 2
PPL Rev. 8 PCIVs B 3.6.1.3 BASES SURVEILLANCE REQUIREMENTS SR 3.6.1.3.1 (continued) a LOCA. Therefore, these valves are allowed to be open for limited periods of time. The 31 day Frequency is consistent with other PCIV requirements discussed in SR 3.6.1.3.2.
SR 3.6.1.3.2 This SR verifies that each primary containment isolation manual valve and blind flange that is located outside primary containment and not locked, sealed, or otherwise secured and is required to be closed during accident conditions is closed. The SR helps to ensure that post accident leakage of radioactive fluids or gases outside the primary containment boundary is within design limits.
This SR does not require any testing or valve manipulation..
Rather, it involves verification that those PCIVs outside primary containment, and capable of being mispositioned, are in the correct position. Since verification of valve position for PCIVs outside primary containment is relatively easy, the 31 day Frequency was-chosen to provide added assurance that the PCIVs are in the correct positions.
Two Notes have been added to this SR. The firstNote allows valves and blind flanges located in high radiation areas to be verified by use of administrative controls. Allowing verification by administrative controls is considered acceptable since access to these areas is typically restricted during MODES 1, 2, and 3 for ALARA reasons. Therefore, the probability of misalignment of these PCIVs, once they have been verified to be in the proper position, is low. A second Note has been included to clarify that PCIVs that are open under administrative controls are not required to meet the SR during the time that the PCIVs are open.
This SR does not apply to valves that are locked, sealed, or otherwise secured in the closed position, since these were verified to be in the correct position upon locking, sealing, or securing.
SR 3.6.1.3.3 This SR verifies that each primary containment manual isolation valve and blind flange that is located inside (continued)
SUSQUEHANNA - UNIT 2 TS /B 3.6-24 Revision 1
PPL Rev. 8 PCIVs B 3.6.1.3 BASES SURVEILLANCE REQUIREMENTS SR 3.6.1.3.3 (continued) primary containment and not locked, sealed, or otherwise secured and is required to be closed during accident conditions is closed. The SR helps to ensure that post accident leakage of radioactive fluids or gases outside the primary containment boundary is within design limits. For PCIVs inside primary containment, the Frequency defined as "prior to entering MODE 2 or 3 from MODE 4 if primary containment was de-inerted while in MODE 4, if not-performed within the previous 92 days" is appropriate since these PCIVs are operated under administrative controls and the probability of their misalignment is low. This SR does not apply to valves that are locked, sealed, or otherwise secured in the closed position, since these were verified to be in the correct position upon locking, sealing, or securing. Two Notes have been added to this SR. The first Note allows valves and blind flanges located in high radiation areas to be verified by use of administrative controls. Allowing verification by administrative controls is considered acceptable since the primary containment is inerted and access to these areas is typically restricted during MODES 1, 2, and 3 for ALARA reasons. Therefore, the probability of misalignment of these PCIVs, once they have been verified to be in their proper position, is low. A second Note has been included to clarify that PCIVs that are open -under administrative controls are not required to meet the SR during the time that the PCIVs are open.
SR 3.6.1.3.4 The traversing incore probe (TIP) shear isolation valves are actuated by explosive charges. Surveillance of explosive charge continuity provides assurance that TIP valves will actuate when required. Other administrative controls, such as those that limit the shelf life of the explosive charges, must be followed. The 31 day Frequency is based on operating experience that has demonstrated the reliability of the explosive charge continuity.
SR 3.6.1.3.5 Verifying the isolation time of each power operated and each automatic PCIV is within limits is required to demonstrate (continued)
SUSQUEHANNA - UNIT 2 TS / B 3.6-25 Revision 1
PPL Rev. 8 PCIVs B 3.6.1.3 BASES SURVEILLANCE REQUIREMENTS SR 3.6.1.3.5 (continued)
OPERABILITY. MSIVs may be excluded from this SR since MSIV full closure isolation time is demonstrated by SR 3.6.1.3.7.
The isolation time test ensures that the valve will isolate in a time period less than or equal to that assumed in the Final Safety Analyses Report. The isolation time and Frequency of this SR are in accordance with the requirements of the Inservice Testing Program.
SR 3.6.1.3.6 For primary containment purge valves with resilient seals, the Appendix J Leakage Rate Test Interval of 24 months is sufficient.
The acceptance criteria for these valves is defined in the Primary Containment Leakage Rate Testing Program, 5.5.12.
The SR is modified by a Note stating that the primary containment purge valves are only required to meet leakage rate testing requirements in MODES 1, 2, and 3. If a LOCA inside primary containment occurs in these MODES, purge, valve leakage must be minimized to ensure offsite radiological release is within limits. At other times when the purge valves are required to be capable of closing (e.g., during handling of
-irradiated fuel), pressurization concerns are not present and the purge valves are not required to meet any specific leakage criteria.
SR 3.6.1.3.7 Verifying that the isolation time of each MSIV is within the specified limits is required to demonstrate OPERABILITY. The isolation time test ensures that the MSIV will isolate in a time period that does not exceed the times assumed in the DBA analyses. This ensures that the calculated radiological consequences of these events remain within regulatory limits.
(continued)
SUSQUEHANNA - UNIT 2 TS / B 3.6-26 Revision 3
PPL Rev. 8 PCIVs B 3.6.1.3 BASES SURVEILLANCE REQUIREMENTS SR 3.6.1.3.7 (continued)
The Frequency of this SR is in accordance with the requirements of the Inservice Testing Program.
SR 3.6.1.3.8 Automatic PCIVs close on a primary containment isolation signal to prevent leakage of radioactive material from primary containment following a DBA. This SR ensures that each automatic PCIV will actuate to its isolation position on a primary containment isolation signal. The LOGIC SYSTEM FUNCTIONAL TEST in SR 3.3.6.1.5 overlaps this SR to provide complete testing of the safety function.
The 24 month Frequency was developed considering it is prudent that some of these Surveillances be performed only during a unit outage since isolation of penetrations could eliminate cooling water flow and disrupt the normal operation of some critical components.
Operating experience has shown that these components usually pass this Surveillance when performed at the 24 month Frequency.
Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.
SR 3.6.1.3.9 This SR requires a-demonstration that a representative sample of reactor instrumentation-line excess flow check valves (EFCV) are OPERABLE by verifying that the valve actuates to check flow on a simulated instrument line break. As defined in FSAR Section 6.2.4.3.5 (Reference 4), the conditions under which an EFCV will isolate, simulated instrument line breaks are at flow rates which develop a differential pressure of between 3 psid and 10 psid. This SR provides assurance that the instrumentation line EFCVs will perform its design function to check flow. No specific valve leakage limits are specified because no specific leakage limits are defined in the FSAR. The 24 month Frequency is based on the need to perform some of these Surveillances under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. The representative sample consists of an approximate equal number of EFCVs such that each EFCV is tested at least once every 10 years (nominal). The nominal 10 year interval is based on other performance-based testing programs, such as Inservice Testing (snubbers) and Option B to 10 CFR 50, Appendix J. In addition, the EFCVs in the sample are representative of the various plant configurations, models, sizes and operating environments. This ensures that any potential common problem with a specific type or application of EFCV is (continued)
SUSQUEHANNA - UNIT 2 TS / B 3.6-27 Revision 3
PPL Rev. 8 PCIVs B 3.6.1.3 BASES SURVEILLANCE SR 3.6.1.3.9 (continued)
REQUIREMENTS detected at the earliest possible time. EFCV failures will be evaluated to determine if additional testing in that test interval is warranted to ensure overall reliability and that failures to isolate are very infrequent.
Therefore, testing of a representative sample was concluded to be acceptable from a reliability standpoint (Reference 7).
SR 3.6.1.3.10 The TIP shear isolation valves are actuated by explosive charges. An in place functional test is not possible with this design. The explosiYe squib is removed and tested to provide assurance that the valves will actuate when required. The replacement charge for the explosive squib shall be from the same manufactured batch as the one fired or from another batch that has been certified by having one of the batch successfully fired. The Frequency of 24 months on a STAGGERED TEST BASIS is considered adequate given the administrative controls on replacement charges and the frequent checks of circuit continuity (SR 3.6.1.3.4).
SR 3.6.1.3. 11 This SR ensures that the leakage rate of secondary containment bypass leakage paths is less than the specified leakage rate. This provides assurance that the assumptions in the radiological evaluations of Reference 4 are met. The secondary containment leakage pathways and Frequency are defined by the Primary Containment Leakage Rate Testing Program. This SR simply imposes additional acceptance criteria. A note is added to this SR which states that these valves are only required to meet this leakage limit in MODES 1, 2, and 3. In the other MODES, the Reactor Coolant System is not pressurized and specific primary containment leakage limits are not required.
SR 3.6.1.312 The analyses in References 1 and 4 are based on the specified leakage rate. Leakage through each MSIV must be < 100 scfh for anyone MSIV or < 300 scfh for total leakage through the MSIVs combined with the Main Steam Line Drain Isolation Valve, HPCI Steam Supply Isolation Valve and the RCIC Steam Supply Isolation Valve. The MSIVs can be tested at either > Pt (22.5 psig) or Pa (45 psig). Main Steam Line Drain Isolation, HPCI and RCIC Steam Supply Line Isolation Valves, are tested at Pa (45 psig). A note is added to this SR which states that these valves are only required to meet this leakage limit in MODES 1, 2, and 3.
In the other (continued)
SUSQUEHANNA - UNIT 2 TS / B 3.6-28 Revision 6
PPL Rev. 8 PCIVs B 3.6.1.3 BASES SURVEILLANCE REQUIREMENTS SR 3.6.1.3.12 (continued) conditions, the Reactor Coolant System is not pressurized and specific primary containment leakage limits are not required. The Frequency is required by the Primary Containment Leakage Rate Testing Program.
SR 3.6.1.3.13 Surveillance of hydrostatically tested lines provides assurance that the calculation assumptions of Reference 2 are met. The acceptance criteria for the combined leakage of all hydrostatically tested lines is 3.3 gpm when tested at 1.1 Pa, (49.5 psig). The combined leakage rates must be demonstrated in accordance with the leakage rate test Frequency required by the Primary Containment Leakage Testing Program.
As noted in Table B 3.6.1.3-1, PCIVs associated with this SR are not Type C tested. Containment bypass leakage is prevented since the line terminates below the minimum water level in the suppression chamber. These valves are tested in accordance with the IST Program. Therefore, these valves leakage is not included as containment leakage.
This SR has been modified by a Note that states that these valves are only required to meet the combined leakage rate in MODES 1, 2, and 3, since this is when the Reactor Coolant System is pressurized and primary containment is required. In some instances, the valves are required to be capable of automatically closing during MODES other than MODES 1, 2,
,and 3. However, specific leakage limits are not applicable in these other MODES or conditions.
REFERENCES 1. FSAR, Chapter 15.
- 2. Final Policy Statement on Technical Specifications Improvements, July 22, 1993 (58 FR 39132).
- 3. 10 CFR 50, Appendix J, Option B.
- 4. FSAR, Section 6.2.
5.. NEDO-30851-P-A, "Technical Specification Improvement Analyses for BWR Reactor Protection System," March 1988.
(continued)
SUSQUEHANNA - UNIT 2 TS / B 3.6-29 Revision 4
PPL Rev. 8 PCIVs B 3.6.1.3 BASES REFERENCES (continued) 6. Standard Review Plan 6.2.4, Rev. 1, September 1975.
- 7. NEDO-32977-A, "Excess Flow Check Valve Testing Relaxation," June 2000.
SUSQUEHANNA - UNIT 2 TS / B 3.6-29a Revision 0
PPL Rev. 8 PCIVs B 3.6.1.3 Table B 3.6.1.3-1 Primary Containment Isolation Valve (Paae I of 10)
Isolation Signal LCO 3.3.6.1 (MaximumFunction No.
Isolation Plant System Valve Number Valve Description Type of Valve Time (Seconds))
Containment 2-57-199 (d) ILRT Manual N/A Atmospheric 2-57-200 (d) ILRT Manual N/A Control HV-25703 Containment Purge Automatic Valve 2-b, 2.d, 2.e (15)
HV-25704 Containment Purge Automatic Valve 2.b, 2.d, 2.e (15)
HV-25705 Containment Purge Automatic Valve 2.b, 2.d, 2.e (15)
HV-25711 Containment Purge Automatic Valve 2.b, 2.d, 2.e (15)
HV-25713 Containment Purge Automatic Valve 2.b, 2.d, 2.e (15)
HV-25714 Containment Purge Automatic Valve 2.b, 2.d, 2.e (15)
HV-25721 Containment Purge Automatic Valve 2.b, 2.d, 2.e (15)
HV-25722 Containment Purge Automatic Valve 2.b, 2.d, 2.e (15)
HV-25723 Containment Purge Automatic Valve 2-b, 2.d, 2.e (15)
HV-25724 Containment Purge Automatic Valve 2.b, 2.d, 2.e (15)
HV-25725 Containment Purge Automatic Valve 2.b, 2.d, 2.e (15)
HV-25766 (a) Suppression Pool Cleanup Automatic Valve 2.b, 2.d (35)
HV-25768 (a) Suppression Pool Cleanup Automatic Valve 2.b, 2.d (30)
SV-2571 00 A Containment Radiation Detection Automatic Valve 2.b, 2.d, 2.f Syst SV-257100 B Containment Radiation Detection Automatic Valve 2.b, 2.d, 2.f Syst SV-2571 01 A Containment Radiation Detection Automatic Valve 2.b, 2.d, 2.f Syst SV-2571 01 B Containment Radiation Detection -Automatic Valve- 2.b, 2.d, 2.f Syst SV-257102 A Containment Radiation Detection -Automatic Valve 2.b, 2.d, 2.f Syst SV-257102 B - Containment Radiation Detection Automatic Valve 2.b, 2.d, 2.f Syst SV-257103 A Containment Radiation Detection Automatic Valve 2.b, 2.d, 2.f Syst SV-257103 B Containment Radiation Detection Automatic Valve 2.b, 2.d, 2.f Syst SV-257104 Containment Radiation Detection Automatic Valve 2.b, 2.d, 2.f Syst SV-257105 Containment Radiation Detection Autormatic Valve 2.b, 2.d, 2.f Syst SV-257106 Containment Radiation Detection Automatic Valve 2.b, 2.d, 2.f Syst SV-257107 Containment Radiation Detection Automatic Valve 2.b, 2.d, 2.f Syst SV-25734 A (e) Containment Atmosphere Sample Automatic Valve 2.b, 2.d SV-25734 B (e) Containment Atmosphere Sample Automatic Valve 2.b, 2.d SV-25736 A (e) Containment Atmosphere Sample Automatic Valve 2.b, 2.d SV-25736 B (e) Containment Atmosphere Sample Automatic Valve 2.b, 2.d SV-25737 Nitrogen Makeup Automatic Valve 2.b, 2.d, 2.e SUSQUEHANNA - UNIT 2 TS / B 3.6-30 Revision 2
PPL Rev. 8 PCIVs B 3.6.1.3 Table B 3.6.1.3-1 Primary Containment Isolation Valve (Pace 2 of 10)
Isolation Signal LCO 3.3.6.1 Function No.
Plant System Valve Number Valve Description Type of Valve (Maximum Isolation Time (Seconds))
Containment SV-25738 Nitrogen Makeup Automatic Valve 2.b, 2.d, 2.e Atmospheric SV-25740 A (e) Containment Atmosphere Sample Automatic Valve 2.b, 2.d Control SV-25740 B (e) Containment Atmosphere Sample Automatic Valve 2.b, 2.d (continued) SV-25742 A (e) Containment Atmosphere Sample Automatic Valve 2.b, 2.d SV-25742 B (e) Containment Atmosphere Sample Automatic Valve 2.b, 2.d SV-25750 A (e) Containment Atmosphere Sample Automatic Valve 2.b, 2.d SV-25750 B (e) Containment Atmosphere Sample Automatic Valve 2.b, 2.d SV-25752 A (e) Containment Atmosphere Sample Automatic Valve 2.b, 2.d SV-25752 B (e) Containment Atmosphere Sample Automatic Valve 2.b, 2.d SV-25767 Nitrogen Makeup Automatic Valve 2.b, 2.d, 2.e SV-25774 A (e) Containment Atmosphere Sample Automatic Valve 2.b, 2.d SV-25774 B (e) Containment Atmosphere Sample Automatic Valve 2.b, 2.d SV-25776 A (e) Containment Atmosphere Sample Automatic Valve 2.b, 2.d SV-25776 B (e) Containment Atmosphere Sample Automatic Valve 2.b, 2.d SV-25780 A (e) Containment Atmosphere Sample Automatic Valve 2.b, 2.d SV-25780 B (e) Containment Atmosphere Sample Automatic Valve 2.b, 2.d SV-25782 A (e) Containment Atmosphere Sample Automatic Valve 2.b, 2.d SV-25782 B (e) Containment Atmosphere Sample Automatic Valve 2.b, 2.d SV-25789 Nitrogen Makeup Automatic Valve 2.b, 2.d, 2.e Containment 2-26-072 (d) Containment Instrument Gas Manual Check N/A Instrument Gas 2-26-074 (d) Containment Instrument Gas Manual Check N/A 2-26-152 (d) Containment Instrument Gas Manual Check N/A 2-26-154 (d) Containment Instrument Gas Manual Check N/A 2-26-164 (d) Containment Instrument Gas Manual Check N/A HV-22603 Containment Instrument Gas Automatic Valve 2.c, 2.d (20)
SV-22605 Containment Instrument Gas Automatic Valve 2.c, 2.d SV-22651 Containment Instrument Gas Automatic Valve 2.c, 2.d SV-22654 A Containment Instrument Gas Power Operated N/A SV-22654 B Containment Instrument Gas Power Operated N/A SV-22661 Containment Instrument Gas Automatic Valve 2.b, 2.d SV-22671 Containment Instrument Gas Automatic Valve 2.b, 2.d Core Spray HV-252F001 A (b)(c) CS Suction Power Operated N/A HV-252F001 B (b)(c) CS Suction Power Operated N/A HV-252F005 A CS Injection Power Operated N/A HV-252F005 B CS Injection Power Operated N/A HV-252F006 A CS Injection Air Operated N/A
_Check Valve HV-252F006 B CS Injection Air Operated N/A
-__V-252F006_B CS__InjectionCheck Valve HV-252F01 5 A (b)(c) CS Test Automatic Valve 2.c, 2.d (80)
HV-252F015 B (b)(c) CS Test Automatic Valve 2.c, 2.d (80)
HV-252F031 A (b)(c) CS Minimum Recirculation Flow Power Operated N/A HV-252F031 B (b)(c) CS Minimum Recirculation Flow Power Operated N/A SUSQUEHANNA- UNIT 2 TS / B 3.6-31 Revision 3
PPL Rev. 8 PCIVs B 3.6.1.3 Table B 3.6.1.3-1 Primary Containment Isolation Valve (Page3 of 10)
Isolation Signal LCO 3.3.6.1 Function No.
Valve Number Valve Description Type of Valve (Maximum stion Plant System (Maximum Isolation Time (Seconds))
Core Spray HV-252F037 A CS Injection Power Operated N/A (continued) (Air)
HV-252F037 B CS Injection Power Operated N/A (Air)
XV-252F01 8 A Core Spray Excess Flow N/A Check Valve XV-252F018 B Core Spray Excess Flow N/A Check Valve Demin Water 2-41-017 (d) Demineralized Water Manual N/A 2-41-018 (d) Demineralized Water Manual N/A HPCI 2-55-038 (d) HPCI Injection. Manual N/A 255F046 (b) (c) (d) HPCI Minimum Recirculation Flow Manual Check N/A 255F049 (a) (d) HPCI Manual Check N/A HV-255F002 HPCI Steam Supply Automatic Valve 3.a, 3.b, 3.c, 3.e, 3f, 3.g, (50)
HV-255F003 HPCI Steam Supply Automatic Valve 3.a, 3.b, 3.c, 3.e, 3.f, 3.g, (50)
HV-255F006 HPCI Injection Power Operated N/A HV-255F012 (b) (c) HPCI Minimum Recirculation Flow Power Operated N/A HV-255F042 (b) (c) HPCI Suction - - Automatic Valve 3.a, 3.b, 3.c, 3.e, Sf,
- - 3.g, (90)
HV-255F066 (a) HPCI Turbine Exhaust Power Operated N/A HV-255F075 HPCI Vacuunm Breaker Automatic Valve 3.b, 3.d, (15)
HV-255F079 HPCI Vacuum Breaker Automatic Valve 3.b, 3.d, (15)
HV-255F100 HPCI Steam Supply Automatic Valve 3.a, 3.b, 3.c, 3.e, 3f, 3.g, (6)
XV-255F024 A HPCI Excess Flow N/A Check Valve XV-255F024 B HPCI Excess Flow N/A Check Valve XV-255F024 C HPCI Excess Flow N/A Check Valve XV-255F024 D HPCI Excess Flow N/A Check Valve Liquid Radwaste HV-261 08 Al Liquid Radwaste Automatic Valve 2.b, 2.d (15)
Collection HV-26108 A2 Liquid Radwaste Automatic Valve 2.b, 2.d (15)
HV-26116 Al Liquid Radwaste Automatic Valve 2.b, 2.d (15)
HV-26116 A2 Liquid Radwaste Automatic Valve 2.b, 2.d (15)
Nuclear Boiler 241 F01 0 A (d) Feedwater Manual Check N/A 241 F010 B (d) Feedwater Manual Check N/A 241 F039 A (d) Feedwater Isolation Valve Manual Check N/A 241 F039 B (d) Feedwater Isolation Valve Manual Check N/A 241818 A (d) Feedwater Isolation Valve Manual Check N/A 241818 B (d) Feedwater Isolation Valve Manual Check N/A SUSQUEHANNA - UNIT 2 TS / B 3.6-32 Revision 1
PPL Rev. 8 PCIVs B 3.6.1.3 Table B 3.6.1.3-1 Primary Containment Isolation Valve (Page 4 of 10)
Isolation Signal LCO 3.3.6.1 Function No.
Plant System Valve Number Valve Description Type of Valve (Maximum Isolation Time (Seconds))
Nuclear Boiler HV-241iF01i6 MSL Drain Automatic Valve 1.a, 1.b, 1.c, 1.d, 1.e (continued) (10)
HV-241 F019 MSL Drain Automatic Valve 1.a, 1.b, 1.c, 1.d, 1.e (15)
HV-241 F022 A MSIV Automatic Valve 1.a, 1.b, 1.c, 1.d, 1.e (5)
HV-241 F022 B MSIV Automatic Valve 1. a, 1.b, 1.c, 1.d, 1.e (5)
HV-241 F022 C MSIV Automatic Valve 1 .a, 1.b, 1.c, 1.d, 1.e (5)
HV-241 F022 D MSIV Automatic Valve 1.a, 1.b, 1.c, 1.d, 1.e (5)
HV-241 F028 A MSIV Automatic Valve 1.a, 1b, 1.c, 1.d, 1.e (5)
HV-241 F028 B MSlV Automatic Valve 1.a, 1.b, 1.c, 1id, 1.e (5)
HV-241 F028 C MSIV Automatic Valve 1 .a, *.b, 1.c, 1.d, 1.e (5)
HV-241 F028 D MSIV Automatic Valve 1.a, 1.b, 1.c, 1.d, 1.e
_ _ _-_(5)
HV-241 F032 A Feedwater Isolation Valve -Power Operated- N/A Check Valves HV-241 F032 B Feedwater Isolation Valve Power Operated N/A
-'Check Valves XV-241 F009 Nuclear Boiler EFCV Excess Flow N/A
-__-_Check Valve XV-241 F070 A Nuclear Boiler EFCV Excess Flow N/A Check Valve XV-241 F070 B Nuclear Boiler EFCV Excess Flow N/A Check Valve XV-241 F070 C Nuclear Boiler EFCV Excess Flow N/A Check Valve XV-241 F070 D Nuclear Boiler EFCV Excess Flow N/A Check Valve XV-241 F071 A Nuclear Boiler EFCV Excess Flow N/A Check Valve XV-241 F071 B Nuclear Boiler EFCV Excess Flow N/A Check Valve XV-241 F071 C Nuclear Boiler EFCV Excess Flow N/A Check Valve XV-241 F071 D Nuclear Boiler EFCV Excess Flow N/A Check Valve XV-241 F072 A Nuclear Boiler EFCV Excess Flow N/A Check Valve XV-241 F072 B Nuclear Boiler EFCV Excess Flow N/A Check Valve XV-241 F072 C Nuclear Boiler EFCV Excess Flow N/A Check Valve XV-241 F072 D Nuclear Boiler EFCV Excess Flow N/A Check Valve XV-241 F073 A Nuclear Boiler EFCV Excess Flow N/A Check Valve SUSQUEHANNA - UNIT 2 TS / B 3.6-33 Revision 2
PPL Rev. 8 PCIVs B 3.6.1.3 Table B 3.6.1.3-1 Primary Containment Isolation Valve (Page 5 of 10)
Isolation Signal LCO 3.3.6.1 Function No.
Plant System Valve Number Valve Description Type of Valve (Maximum Isolation Time (Seconds))
Nuclear Boiler XV-241 F073 B Nuclear Boiler EFCV Excess Flow N/A (continued) Check Valve XV-241 F073 C Nuclear Boiler EFCV Excess Flow N/A Check Valve XV-241 F073 D Nuclear Boiler EFCV Excess Flow N/A Check Valve Nuclear Boiler XV-24201 Nuclear Boiler Vessel Instrument Excess Flow N/A Vessel Check Valve Instrumentation XV-24202 Nuclear Boiler Vessel Instrument Excess Flow N/A Check Valve XV-242F041 Nuclear Boiler Vessel Instrument Excess Flow N/A Check Valve XV-242F043 A Nuclear Boiler Vessel Instrument Excess Flow N/A Check Valve XV-242F043 B Nuclear Boiler Vessel Instrument Excess Flow N/A Check Valve XV-242F045 A Nuclear Boiler Vessel Instrument Excess Flow N/A Check Valve XV-242F045 B Nuclear Boiler Vessel Instrument Excess Flow N/A Check Valve XV-242F047 A Nuclear Boiler Vessel Instrument -Excess Flow N/A Check Valve XV-242F047B Nuclear Boiler Vessel Instrument Excess Flow N/A Check Valve XV-242F051 A Nuclear Boiler Vessel Instrument Excess Flow N/A Check Valve XV-242F051 B Nuclear Boiler Vessel Instrument Excess Flow N/A Check Valve XV-242F051 C Nuclear Boiler Vessel Instrument Excess Flow N/A Check Valve XV-242F051 D Nuclear Boiler Vessel Instrument Excess Flow N/A Check Valve XV-242F053 A Nuclear Boiler Vessel Instrument Excess Flow N/A Check Valve XV-242F053 B Nuclear Boiler Vessel Instrument Excess Flow N/A Check Valve XV-242F053 C Nuclear Boiler Vessel Instrument Excess Flow N/A Check Valve XV-242F053 D Nuclear Boiler Vessel Instrument Excess Flow N/A Check Valve XV-242F055 Nuclear Boiler Vessel Instrument Excess Flow N/A I Check Valve XV-242F057 Nuclear Boiler Vessel Instrument Excess Flow N/A Check Valve XV-242F059 A Nuclear Boiler Vessel Instrument Excess Flow N/A Check Valve XV-242F059 B Nuclear Boiler Vessel Instrument Excess Flow N/A Check Valve SUSQUEHANNA - UNIT 2 TS / B 3.6-34 Revision 1
PPL Rev. 8 PCIVs B 3.6.1.3 Table B 3.6.1.3-1 Primary Containment Isolation Valve (Paqe 6 of 10)
Isolation Signal LCO Plant System Valve Number Valve Description Type of Valve 3.3.6.1 Function No.
(Maximum Isolation Time (Seconds))
Nuclear Boiler XV-242F059 C Nuclear Boiler Vessel Instrument Excess Flow NIA Vessel Check Valve Instrumentation XV-242F059 D Nuclear Boiler Vessel Instrument Excess Flow N/A (continued) Check Valve XV-242F059 E Nuclear Boiler Vessel Instrument Excess Flow N/A Check Valve XV-242F059 F Nuclear Boiler Vessel Instrument Excess Flow N/A Check Valve XV-242F059 G Nuclear Boiler Vessel Instrument Excess Flow N/A Check Valve XV-242F059 H Nuclear Boiler Vessel Instrument Excess Flow N/A Check Valve XV-242F059 L Nuclear Boiler Vessel Instrument Excess Flow N/A Check Valve XV-242F059 M Nuclear Boiler Vessel Instrument Excess Flow N/A Check Valve XV-242F059 N Nuclear Boiler Vessel Instrument Excess Flow N/A Check Valve XV-242F059 P Nuclear Boiler Vessel Instrument Excess Flow N/A Check Valve XV-242F059 R Nuclear Boiler Vessel Instrument -Excess Flow N/A Check Valve XV-242F059 S Nuclear Boiler Vessel Instrument Excess Flow N/A
-Check Valve XV-_242F059 T Nuclear Boiler Vessel Instrument Excess Flow N/A Check Valve XV-242F059 U Nuclear Boiler Vessel Instrument Excess Flow N/A Check Valve XV-242F061 Nuclear Boiler Vessel Instrument Excess Flow N/A Check Valve RB Chilled Water HV-28781 Al RB Chilled Water Automatic Valve 2.c, 2.d (40)
System HV-28781 A2 RB Chilled Water Automatic Valve 2.c, 2.d (40)
HV-28781 B1 RB Chilled Water Automatic Valve 2.c, 2.d (40)
HV-28781 B2 RB Chilled Water Automatic Valve 2.c, 2.d (40)
HV-28782 Al RB Chilled Water Automatic Valve 2.c, 2.d (12)
HV-28782 A2 RB Chilled Water Automatic Valve 2.c, 2.d (12)
HV-28782 B1 RB Chilled Water Automatic Valve 2.c, 2.d (12)
HV-28782 B2 RB Chilled Water Automatic Valve 2.c, 2.d (12)
HV-28791 Al RB Chilled Water Automatic Valve 2.b, 2.d (15)
HV-28791 A2 RB Chilled Water Automatic Valve 2.b, 2.d (15)
HV-28791 B1 RB Chilled Water Automatic Valve 2.b, 2.d (15)
HV-28791 B2 RB Chilled Water Automatic Valve 2.b, 2.d (15)
HV-28792 Al RB Chilled Water Automatic Valve 2.b, 2.d (8)
HV-28792 A2 RB Chilled Water Automatic Valve 2.b, 2.d (8)
HV-28792 B1 RB Chilled Water Automatic Valve 2.b, 2.d (8)
SUSQUEHANNA - UNIT 2 TS / B 3.6-35 Revision 3
PPL Rev. 8 PCIVs B 3.6.1.3 Table B 3.6.1.3-1 Primary Containment Isolation Valve (Page 7 of 10)
Isolation Signal LCO 3.3.6.1 Function No.
Valve Number Valve Description Type of Valve (Maximum Isolation Plant System Time (Seconds))
RB Chilled Water HV-28792 B2 RB Chilled Water Automatic Valve 2.b, 2.d (8)
System (continued)
RBCCW HV-21313 RBCCW Automatic Valve 2.c, 2.d (30)
HV-21314 RBCCW Automatic Valve 2.c, 2.d (30)
HV-21345 RBCCW Automatic Valve 2.c, 2.d (30)
HV-21346 RBCCW Automatic Valve 2.c, 2.d (30)
RCIC 2-49-020 (d) RCIC Injection Manual N/A 249F021 (b) (c) (d) RCIC Minimum Recirculation Flow Manual Check N/A 249F028 (a) (d) RCIC Vacuum Pump Discharge Manual N/A 249F040 (a) (d) RCIC Turbine Exhaust Manual N/A FV-249F019 (b) (c) RCIC Minimum Recirculation Flow Power Operated N/A HV-249F007 RCIC Steam Supply Automatic Valve 4.a, 4.b, 4.c, 4.e, 4.f, 4.g (20)
HV-249F008 RCIC Steam Supply Automatic Valve 4.a, 4.b, 4.c, 4.e, 4.f, 4.g (20)
HV-249F013 RCIC Injection Power Operated N/A HV-249F031 (b) (c) RCIC Suction Power Operated N/A HV-249F059 (a) RCIC Turbine Exhaust Power Operated N/A HV-249F060 (a) RCIC Vacuum Puhlp Discharge Power Operated N/A HV-249F062 RCIC Vacuum Breaker -Automatic Valve- 4.b, 4.d (10)
HV-249F084 RCIC Vacuum Breaker Automatic Valve 4.b, 4.d (10)
HV-249F088 RCIC Steam Supply Automatic Valve 4.a, 4.b, 4.c, 4.e, 4.f, 4.g (12)
XV-249F044 A - RCIC Excess Flow N/A Check Valve XV-249F044 B RCIC Excess Flow N/A Check Valve XV-249F044 C RCIC Excess Flow N/A Check Valve XV-249F044 D RCIC Excess Flow N/A Check Valve Reactor 243F01 3 A (d) Recirculation Pump Seal Water Manual Check N/A Recirculation 243F01 3 B (d) Recirculation Pump Seal Water Manual Check N/A HV-243F01 9 Reactor Coolant Sample Automatic Valve 2.b (9)
HV-243F020 Reactor Coolant Sample Automatic Valve 2.b (2)
XV-243F003 A Reactor Recirculation Excess Flow N/A Check Valve XV-243F003 B Reactor Recirculation Excess Flow N/A Check Valve XV-243F004 A Reactor Recirculation Excess Flow N/A Check Valve XV-243F004 B Reactor Recirculation Excess Flow N/A Check Valve SUSQUEHANNA - UNIT 2 TS / B 3.6-36 Revision 2
PPL Rev. 8 PCIVs B 3.6.1.3 Table B 3.6.1.3-1 Primary Containment Isolation Valve (Page 8 of 10)
Isolation Signal LCO 3.3.6.1 Function No.
Plant System Valve Number Valve Description Type of Valve (Maximumstion (Maximum Isolation Time (Seconds))
Reactor XV-243F009 A Reactor Recirculation Excess Flow N/A Recirculation Check Valve (continued) XV-243F009 B Reactor Recirculation Excess Flow N/A Check Valve XV-243F009 C Reactor Recirculation Excess Flow N/A Check Valve XV-243F009 D Reactor Recirculation Excess Flow N/A Check Valve XV-243F01 0 A Reactor Recirculation Excess Flow N/A Check Valve XV-243F010 B Reactor Recirculation Excess Flow N/A Check Valve XV-243F01 0 C Reactor Recirculation Excess Flow N/A Check Valve XV-243F010 D Reactor Recirculation Excess Flow N/A Check Valve XV-243F01 1 A Reactor Recirculation Excess Flow N/A Check Valve XV-243F011 B Reactor Recirculation Excess Flow N/A Check Valve XV-243F01 1 C Reactor Recirculation Excess Flow N/A Check Valve XV-243F01 1 D Reactor Recirculation Excess Flow N/A Check Valve XV-243F01 2 A Reactor Recirculation Excess Flow N/A Check Valve XV-243F01 2 B Reactor Recirculation Excess Flow N/A Check Valve XV-243F01 2 C Reactor Recirculation Excess Flow N/A Check Valve XV-243F012 D Reactor Recirculation Excess Flow N/A Check Valve XV-243F017 A Recirculation Pump Seal Water Excess Flow N/A Check Valve XV-243F017 B Recirculation Pump Seal Water Excess Flow N/A Check Valve XV-243F040 A Reactor Recirculation Excess Flow N/A Check Valve XV-243F040 B Reactor Recirculation Excess Flow N/A Check Valve XV-243F040 C Reactor Recirculation Excess Flow N/A Check Valve XV-243F040 D Reactor Recirculation Excess Flow N/A Check Valve XV-243F057 A Reactor Recirculation Excess Flow N/A Check Valve XV-243F057 B Reactor Recirculation Excess Flow N/A Check Valve Residual Heat HV-251 F004 A (b) (c) RHR - Suppression Pool Suction Power Operated N/A Removal HV-251 F004 B (b) (c) RHR - Suppression Pool Suction Power Operated N/A HV-251 F004 C (b) (c) RHR - Suppression Pool Suction Power Operated N/A SUSQUEHANNA - UNIT 2 TS / B 3.6-37 Revision 2
PPL Rev. 8 PCIVs B 3.6.1.3 Table B 3.6.1.3-1 Primary Containment Isolation Valve (Page 9 of 10)
Isolation Signal LCO 3.3.6.1 Function No.
Valve Number Valve Description Type of Valve (Maximum stion Plant System (Maximum Isolation Time (Seconds))
Residual Heat HV-251 F004 D(b) (c) RHR - Suppression Pool Suction Power Operated N/A Removal HV-251 F007 A (b) (c) RHR - Minimum Recirculation Power Operated N/A (continued) HV-251 F007 B (b) (c) RHR - Minimum Recirculation Power Operated N/A HV-251 F008 RHR - Shutdown Cooling Suction Automatic Valve 6.a, 6.b, 6.c (52)
HV-251 F009 RHR - Shutdown Cooling Suction Automatic Valve 6.a, 6.b, 6.c (52)
HV-251 F01 1 A (b) (d) RHR - Suppression Pool Cooling Manual N/A HV-251 F01 1 B (b) (d) RHR - Suppression Pool Cooling Manual N/A HV-251 F01 5 A (f) RHR - Shutdown Cooling Power Operated N/A Return/LPCI Injection HV-251F015 B (f) RHR - Shutdown Cooling Power Operated N/A Return/LPCI Injection HV-251 F016 A (b) RHR - Drywell Spray Automatic Valve 2.c, 2.d (90)
HV-251 F016 B (b) RHR - Drywell Spray Automatic Valve 2.c, 2.d (90)
HV-251 F022 RH R - Reactor Vessel Head Spray Automatic Valve 2.d, 6.a, 6.b, 6.c (30)
HV-251 F023 RHR - Reactor Vessel Head Spray Automatic Valve 2.d, 6.a, 6.b, 6.c (20)
HV-251 F028 A (b) RHR - Suppression Pool Automatic Valve 2.c, 2.d (90)
Cooling/Spray HV-251 F028 B (b) RHR - Suppression Pool Automatic Valve 2.c, 2.d (90)
Cooling/Spray HV-251 F050 A (g) RHR - Shutdown-Cooling Air Operated N/A Return/LPCI Injection - -Check Valve -
HV-251 F050 B (g) RHR - Shutdown Cooling Air Operated N/A Return/LPCI Fnjection Check Valve HV-251 F103 A (b) RHR Heat Exchanger Vent Power Operated N/A HV-2-1 F103 B (b) - RHR Heat Exchanger Vent Power Operated N/A HV-251 F122 A (g) RHR - Shutdown Cooling Power Operated N/A Return/LPCI Injection (Air)
HV-251 F122 B (g) RHR - Shutdown Cooling Power Operated N/A Return/LPCI Injection (Air)
PSV-25106 A (b) (d) RHR- Relief Valve Discharge Relief Valve N/A PSV-25106 B (b) (d) RHR- Relief Valve Discharge Relief Valve N/A PSV-251 F126 (d) RHR- Shutdown Cooling Suction Relief Valve N/A XV-25109 A RHR Excess Flow N/A Check Valve XV-25109 B RHR Excess Flow N/A Check Valve XV-25109 C RHR Excess Flow N/A Check Valve XV-25109 D RHR Excess Flow N/A Check Valve RWCU HV-244F001 (a) RWCU Suction Automatic Valve 5.a, 5.b, 5.c, 5.d, 5.f, 5.g (30)
HV-244F004 (a) RWCU Suction Automatic Valve 5.a, 5.b, 5.c, 5.d, 5.e, 5.f, 5.g (30)
XV-24411 A RWCU Excess Flow N/A Check Valve XV-24411 B RWCU Excess Flow N/A Check Valve SUSQUEHANNA - UNIT 2 TS / B 3.6-38 Revision 3
PPL Rev. 8 PCIVs B 3.6.1.3 Table B 3.6.1.3-1 Primary Containment Isolation Valve (Page 10 of 10)
Isolation Signal LCO 3.3.6.1 Function No.
stion Plant System Valve Number Valve Description Type of Valve (Maximum (Maximum Isolation Time (Seconds))
RWCU XV-24411 C RWCU Excess Flow N/A (continued) Check Valve XV-24411 D RWCU Excess Flow N/A Check Valve XV-244F046 RWCU Excess Flow N/A Check Valve HV-24182 A RWCU Return Power Operated N/A HV-24182 B RWCU Return Power Operated N/A SLCS 248F007 (a) (d) SLCS Manual Check N/A HV-248F006 (a) SLCS Power Operated N/A Check Valve TIP System C51-J004 A (Ball TIP Ball Valves Automatic Valve 7.a, 7.b (5)
Valve)
C51-J004 B (Ball TIP Ball Valves Automatic Valve 7.a, 7.b (5)
Valve)
C51-J004 C (Ball TIP Ball Valves Automatic Valve 7.a, 7.b (5)
Valve)
C51-J004 D (Ball TIP Ball Valves Automatic Valve 7.a, 7.b (5)
Valve)
C51-J004 E (Ball TIP Ball Valves Automatic Valve 7.a, 7.b (5)
Valve)
TIP System C51-J004 A (Shear TIP Shear Valves S-quib Valve N/A (continued) Valve)
C51-J004 B (Shear TIP Shear Valves Squib Valve N/A Valve)
C51-J004 C (Shear TIP Shear Valves Squib Valve N/A VaIve)
C51-J004 D (Shear TIP Shear Valves Squib Valve N/A Valve)
C51-J004 E (Shear TIP Shear Valves Squib Valve N/A Valve)
(a) Isolation barrier remains filled or a water seal remains in the line post-LOCA, isolation valve is tested with water.
Isolation valve leakage is not included in 0.60 La total Type B and C tests.
(b) Redundant isolation boundary for this valve is provided by the closed system whose integrity is verified by the Leakage Rate Test Program. This footnote does not apply to valve 255F046 (HPCI) when the associated PCIV, HV255F012 is closed and deactivated. Similarly, this footnote does not apply to valve 249F021 (RCIC) when its associated PCIV, FV249F019 is closed and deactivated.
(c) Containment Isolation Valves are not Type C tested. Containment bypass leakage is prevented since the line terminates below the minimum water level in the Suppression Chamber. Refer to the IST Program.
(d) LCO 3.3.3.1, "PAM Instrumentation," Table 3.3.3.1-1, Function 6, (PCIV Position) does not apply since these are relief valves, check valves, manual valves or deactivated and closed.
(e) The containment isolation barriers for the penetration associated with this valve consists of two PCIVs and a closed system. The closed system provides a redundant isolation boundary for both PCIVs, and its integrity is required to be verified by the Leakage Rate Test Program.
(f) Redundant isolation boundary for this valve is provided by the closed system whose integrity is verified by the Leakage Rate Test Program.
(g) These valves are not required to be 10 CFR 50, Appendix J tested since the HV-251 F015A(B) valves and a closed system form the 10 CFR 50, Appendix J boundary. These valves form a high/low pressure interface and are pressure tested in accordance with the pressure test program.
SUSQUEHANNA - UNIT 2 TS / B 3.6-39 Revision 7
PPL Rev. 1 Inservice Leak and Hydrostatic Testing Operation B 3.10.1 B 3.10 SPECIAL OPERATIONS B 3.10.1 Inservice Leak and Hydrostatic Testing Operation BASES BACKGROUND The purpose of this Special Operations LCO is to allow certain reactor coolant pressure tests to be performed in MODE 4 with temperatures as high as 212°F when operational conditions or the metallurgical characteristics of the reactor pressure vessel (RPV) require the pressure testing at temperatures > 200°F (normally corresponding to MODE 3) or to allow completing these reactor coolant pressure tests when the initial conditions do not require temperatures > 200 0 F.
Furthermore, the purpose is to allow continued performance of control rod scram time testing required by SR 3.1.4.1, SR 3.1.4.3 or SR 3.1.4.4 if reactor coolant temperatures exceed 200°F when the control rod scram tirfe testing is initiated in conjunction with an inservice leak or hydrostatic test. These control rod scram time.tests would be performed in accordance with LCO 3.10.4, "Single Control Rod Withdrawal - Cold Shutdown," during MODE 4 operation.
Inservice hydrostatic testing and system leakage pressure tests required by SectionXI of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Ref. 1) are performed prior to the reactor going critical after a refueling outage.
Recirculation pump operation and a water solid RPV (except for an air bubble for pressure control) are used to achieve the necessary temperatures and pressures required for these tests. The minimum temperatures (at the required pressures) allowed for these tests are determined from the RPV pressure and temperature (P/T) limits required by LCO 3.4.10, "Reactor Coolant System (RCS) Pressure and Temperature (P/T) Limits." These limits are conservatively based on the fracture toughness of the reactor vessel, taking into account anticipated vessel neutron fluence.
With increased reactor vessel fluence over time, the minimum allowable vessel temperature increases at a given pressure. Periodic updates to the RPV P/T limit curves are performed as necessary, based upon the results of analyses of irradiated surveillance specimens removed from the vessel. Hydrostatic and leak testing may eventually be required with minimum reactor coolant temperatures > 200 0 F. However, even with required minimum reactor (continued)
SUSQUEHANNA - UNIT 2 TS / B 3.10-1 Revision 2
PPL Rev. 1 Inservice Leak and Hydrostatic Testing Operation B 3.10.1 BASES BACKGROUND coolant temperatures < 2000 F, maintaining RCS temperatures within (continued) a small band during the test can be impractical. Removal of heat addition from recirculation pump operation and reactor core decay heat is coarsely controlled by control rod drive hydraulic system flow and reactor water cleanup system non-regenerative heat exchanger operation. Test conditions are focused on maintaining a steady state pressure, and tightly limited temperature control poses an unnecessary burden on the operator and may not be achievable in certain instances.
The hydrostatic and RCS system leakage tests require increasing pressure to 1035 (+10, -0) psig. Scram time testing required by SR 3.1.4.1 and SR 3.1.4.4 requires reactor pressures > 800 psig.
Other testing may be performed in conjunction with the allowances for inservice leak or hydrostatic tests and control rod scram time tests.
APPLICABLE Allowing the reactor to be considered in MODE 4 when the reactor SAFETY ANALYSES coolant temperature is > 200'F but < 212'F, during, or as a consequence of hydrostatic or leak testing, or as a consequence of control rod scram time testing initiated in conjunction with an inservice leak or hydrostatic test, effectively provides an exception to MODE 3 requirements, including OPERABILITY of primary containment and the full complement of redundant Emergency Core Cooling Systems.
Since the tests are performed nearly water solid, at low decay heat values, and near MODE 4 conditions, the stored energy in the reactor core will be very low. Under these conditions, the potential for failed fuel and a subsequent increase in coolant activity above the LCO 3.4.7, "RCS Specific Activity," limits are minimized. In addition, the secondary containment will be OPERABLE, in accordance with this Special Operations LCO, and will be capable of handling any airborne radioactivity or steam leaks that could occur during the performance of hydrostatic or leak testing. The required pressure testing conditions provide adequate assurance that the consequences of a steam leak will be conservatively bounded by the consequences of the postulated main steam line break outside of primary containment described in Reference 2. Therefore, these requirements will conservatively limit radiation releases to the environment.
(continued)
SUSQUEHANNA - UNIT 2 TS / B 3.10-2 Revision 1
PPL Rev. 1 Inservice Leak and Hydrostatic Testing Operation B 3.10.1 BASES APPLICABLE In the event of a large primary system leak, the reactor vessel would SAFETY ANALYSES rapidly depressurize, allowing the low pressure core cooling systems (continued) to operate. The capability of the low pressure coolant injection and core spray subsystems, as required in MODE 4 by LCO 3.5.2, "ECCS-Shutdown," would be more than adequate to keep the core flooded under this low decay heat load condition. Small system leaks would be detected by leakage inspections before significant inventory loss occurred.
For the purposes of this test, the protection provided by normally required MODE 4 applicable LCOs, in addition to the secondary containment requirements required to be met by this Special Operations LCO, will ensure acceptable consequences during normal hydrostatic test conditions and during postulated accident conditions.
As described in LCO 3.0.7, compliance with Special Operations LCOs is optional, and therefore, no criteria of the NRC Policy Statement apply. Special Operations LCOs provide flexibility to perform certain operations by appropriately modifying requirements of other LCOs. A discussion of the criteria satisfied for the other LCOs is provided in their respective Bases.
LCO As described in LCO 3.0.7, compliance with this Special Operations LCO is optional. Operation at reactor coolant temperatures > 2000F but < 212°F can be in accordance with Table 1.1-1 for MODE 3 operation without meeting this Special Operations LCO or its ACTIONS. This option may be required due to plant conditions or P/T limits, however, which require testing at temperatures > 200'F, while the ASME inservice test itself requires the safety/relief valves to be gagged, preventing their OPERABILITY. Additionally, even with required minimum reactor coolant temperatures < 2000 F, RCS temperatures may drift above 200°F during the performance of inservice leak and hydrostatic testing or during subsequent control rod scram time testing, which is typically performed in conjunction with inservice leak and hydrostatic testing. While this Special Operations LCO is provided for inservice leak and hydrostatic testing, and for scram time testing initiated in conjunction with an inservice leak or hydrostatic test, parallel performance of other tests and inspections is not precluded.
(continued)
SUSQUEHANNA - UNIT 2 TS / B 3.10-3 Revision I
PPL Rev. 1 Inservice Leak and Hydrostatic Testing Operation B 3.10.1 BASES LCO If it is desired to perform these tests while complying with this Special (continued) Operations LCO, then the MODE 4 applicable LCOs and specified LCOs must be met. This Special Operations LCO allows changing Table 1.1-1 temperature limits for MODE 4 to "< 212" and suspending the requirements of LCO 3.4.9, "Residual Heat Removal (RHR)
Shutdown Cooling System-Cold Shutdown." The additional requirements for secondary containment LCOs to be met will provide sufficient protection for operations at reactor coolant temperatures
> 200°F for the purpose of performing an inservice leak or hydrostatic test, and for control rod scram time testing initiated in conjunction with an inservice leak or hydrostatic test.
This LCO allows primary containment to be open for frequent unobstructed access to perform inspections, and for outage activities on various systems to continue consistent with the MODE 4 applicable requirements.
APPLICABILITY The MODE 4 requirements may only be modified for the performance of, or as a consequence of inservice leak or hydrostatic tests, or as a consequence of control rod scram time testing initiated in conjunction with an inservice leak or hydrostatic test, so that these operations can be considered as-in MODE 4, even though the reactor coolant temperature is > 2000-F. The additional requirement for secondary containment OPERABILITY according to the imposed MODE 3 requirements provides conservatism in the response of the unit to any event that may occur. Operations in all other MODES are unaffected by this LCO.
ACTIONS A Note has been provided to modify the ACTIONS related to inservice leak and hydrostatic testing operation. Section 1.3, Completion Times, specifies that once a Condition has been entered, subsequent divisions, subsystems, components, or variables expressed in the Condition discovered to be inoperable or not within limits, will not result in separate entry into the Condition. Section 1.3 also specifies that Required Actions of the Condition continue to apply for each additional failure, with Completion Times based on initial entry into the Condition. However, the Required Actions for each requirement of the LCO not met provide appropriate compensatory measures for separate requirements that are not met. As such, a Note has been provided that allows separate Condition entry for each requirement of the LCO.
(continued)
SUSQUEHANNA - UNIT 2 TS / B 3.10-4 Revision 1
PPL Rev. 1 Inservice Leak and Hydrostatic Testing Operation B 3.10.1 BASES ACTIONS A. 1 (continued)
If an LCO specified in LCO 3.10.1 is not met, the ACTIONS applicable to the stated requirements are entered immediately and complied with. Required Action A.1 has been modified by a Note that clarifies the intent of another LCO's Required Action to be in MODE 4 includes reducing the average reactor coolant temperature to < 2000 F.
A.2.1 and A.2.2 Required Action A.2.1 and Required Action A.2.2 are alternate Required Actions that can be taken instead of Required Action A. 1 to restore compliance with the normal MODE 4 requirements, and thereby exit this Special Operation LCO's Applicability. Activities that could further increase reactor coolant temperature or pressure are suspended immediately, in accordance with Required Action A.2.1, and the reactor coolant temperature is reduced to establish normal MODE 4 requirements. The allowed Completion Time of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for Required Action A.2.2 is based on engineering judgment and provides sufficient tinme to reduce the average reactor coolant temperature from the highest expected value to <200°F with normal cooldown procedures. The Completion Time is also consistent with the time provided in LCO 3.0.3 to reach MODE 4 from MODE 3.
SURVEILLANCE SR 3.10.1.1 REQUIREMENTS The LCOs made applicable are required to have their Surveillances met to establish that this LCO is being met. A discussion of the applicable SRs is provided in their respective Bases.
REFERENCES 1. American Society of Mechanical Engineers, Boiler and Pressure Vessel Code,Section XI.
- 2. FSAR, Section 15.6.4 SUSQUEHANNA - UNIT 2 TS / B 3.10-5 Revision 1