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TABLE OF CONTENTS
 
B 2.0                                SAFETY LIMITS (SLs) ......................................... B      2.0-1 B 2.1.1                                                                          Reactor Core SLs .................................... B      2.0-1 B 2.1.2                                                                          Reactor Coolant System (RCS) Pressure SL  ........... B      2.0-7
 
B 3.0                                LIMITING CONDITION FOR OPERATION (LCO) APPLICABILITY ........ B      3.0-1 B 3.0                                SURVEILLANCE REQUIREMENT (SR) APPLICABILITY ................. B      3.0-10
 
B 3.1                                                              REACTIVITY CONTROL SYSTEMS .............................. B      3.1-1 B 3.1.1                                                                          SHUTDOWN MARGIN (SDM) ............................... B      3.1-1 B 3.1.2  Reactivity Anomalies ................................ B      3.1-8 B 3.1.3                                                                          Control Rod OPERABILITY ............................. B      3.1-13 B 3.1.4                                                                          Control Rod Scram Times ............................. B      3.1-22 B 3.1.5                                                                          Control Rod Scram Accumulators ...................... B      3.1-29 B 3.1.6  Rod Pattern Control ................................. B      3.1-34 B 3.1.7                                                                          Standby Liquid Control (SLC) System ................. B      3.1-39 B 3.1.8                                                                          Scram Discharge Volume (SDV) Vent and Drain Valves .. B      3.1-48
 
B 3.2                                                              POWER DISTRIBUTION LIMITS ............................... B      3.2-1 B 3.2.1                                                                          AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR) ......................................... B      3.2-1 B 3.2.2                                                                          MINIMUM CRITICAL POWER RATIO (MCPR) ................. B      3.2-6 B 3.2.3                                                                          LINEAR HEAT GENERATION RATE (LHGR)  ................. B      3.2-11
 
B 3.3                                                              INSTRUMENTATION ......................................... B      3.3-1 B 3.3.1.1                                Reactor Protection System (RPS) Instrumentation ......... B      3.3-1 B 3.3.1.2                                Wide Range Neutron Monitor (WRNM) Instrumentation ....... B      3.3-36 B 3.3.2.1                                Control Rod Block Instrumentation ....................... B      3.3-45 B 3.3.2.2                                Feedwater and Main Turbine High Water Level Trip Instrumentation .................................. B      3.3-58 B 3.3.3.1                                Post Accident Monitoring (PAM) Instrumentation .......... B      3.3-65 B 3.3.3.2  Remote Shutdown System .................................. B      3.3-76 B 3.3.4.1                                Anticipated Transient Without Scram Recirculation Pump Trip (ATWS-RPT) Instrumentation ............. B      3.3-83 B 3.3.4.2                                End of Cycle Recirculation Pump Trip (EOC-RPT) Instrumentation . . .                              B 3.3-91a thru B 3.3-91j B 3.3.5.1                                Emergency Core Cooling System (ECCS)
Instrumentation .................................. B      3.3-92 B 3.3.5.2                                Reactor Core Isolation Cooling (RCIC) System Instrumentation .................................. B      3.3-130 B 3.3.5.3  Not Used ................................................ B-3.3-140a B 3.3.5.4                                Reactor Pressure Vessel (RPV) Water Inventory Control Instrumentation .................................. B-3.3.140b B 3.3.6.1                                Primary Containment Isolation Instrumentation ........... B      3.3-141 B 3.3.6.2                                Secondary Containment Isolation Instrumentation ......... B      3.3-169 B 3.3.7.1                                Main Control Room Emergency Ventilation (MCREV)
System Instrumentation ........................... B      3.3-180 B 3.3.8.1                                Loss of Power (LOP) Instrumentation ..................... B      3.3-187 B 3.3.8.2                                Reactor Protection System (RPS) Electric Power Monitoring ....................................... B      3.3-199
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                                                                      i                                                                                                                                                                                        Revision No. 145 TABLE OF CONTENTS (continued)
 
B 3.4                                                              REACTOR COOLANT SYSTEM (RCS) ............................ B      3.4-1 B 3.4.1  Recirculation Loops Operating ....................... B      3.4-1 B 3.4.2  Jet Pumps ........................................... B      3.4-11 B 3.4.3                                                                          Safety Relief Valves (SRVs) and Safety Valves (SVs) . B      3.4-15 B 3.4.4  RCS Operational LEAKAGE ............................. B      3.4-19 B 3.4.5                                                                          RCS Leakage Detection Instrumentation ............... B      3.4-24 B 3.4.6  RCS Specific Activity ............................... B      3.4-29 B 3.4.7                                                                          Residual Heat Removal (RHR) Shutdown Cooling SystemHot Shutdown ............................. B 3.4-33 B 3.4.8                                                                          Residual Heat Removal (RHR) Shutdown Cooling SystemCold Shutdown ............................ B 3.4-38 B 3.4.9                                                                          RCS Pressure and Temperature (P/T) Limits ........... B      3.4-43 B 3.4.10                                                                Reactor Steam Dome Pressure ......................... B      3.4-52
 
B 3.5                                                              EMERGENCY CORE COOLING SYSTEMS (ECCS), RPV WATER INVENTORY CONTROL (WIC), AND REACTOR CORE ISOLATION COOLING (RCIC) SYSTEM ................................... B      3.5-1 B 3.5.1  ECCS ................................................ B      3.5-1 B 3.5.2  Deleted ............................................. B      3.5-18 B 3.5.3  RCIC System ......................................... B      3.5-24 B 3.5.4                                                                          RPV Water Inventory Control ......................... B      3.5-25
 
B 3.6                                                              CONTAINMENT SYSTEMS ..................................... B      3.6-1 B 3.6.1.1  Primary Containment ..................................... B      3.6-1 B 3.6.1.2                                Primary Containment Air Lock ............................ B      3.6-6 B 3.6.1.3                                Primary Containment Isolation Valves (PCIVs) ............ B      3.6-14 B 3.6.1.4                                Drywell Air Temperature ................................. B      3.6-31 B 3.6.1.5                                Reactor Building-to-Suppression Chamber Vacuum Breakers ......................................... B      3.6-34 B 3.6.1.6                                Suppression Chamber-to-Drywell Vacuum Breakers .......... B      3.6-42 B 3.6.2.1                                Suppression Pool Average Temperature .................... B      3.6-48 B 3.6.2.2                                Suppression Pool Water Level ............................ B      3.6-53 B 3.6.2.3                                Residual Heat Removal (RHR) Suppression Pool Cooling .......................................... B      3.6-56 B 3.6.2.4                                Residual Heat Removal (RHR) Suppression Pool Spray ...... B      3.6-60 B 3.6.2.5                                Residual Heat Removal (RHR) Drywell Spray ............... B      3.6-63a B 3.6.3.1  Deleted ................................................. B      3.6-64 B 3.6.3.2                                Primary Containment Oxygen Concentration ................ B      3.6-70 B 3.6.4.1  Secondary Containment ................................... B      3.6-73 B 3.6.4.2                                Secondary Containment Isolation Valves (SCIVs) .......... B      3.6-78 B 3.6.4.3                                Standby Gas Treatment (SGT) System ...................... B      3.6-85
 
B 3.7                                                              PLANT SYSTEMS ........................................... B      3.7-1 B 3.7.1                                                                          High Pressure Service Water (HPSW) System ........... B      3.7-1 B 3.7.2                                                                          Emergency Service Water (ESW) System and Normal Heat Sink ....................................... B      3.7-6 B 3.7.3  Emergency Heat Sink ................................. B      3.7-11 B 3.7.4                                                                          Main Control Room Emergency Ventilation (MCREV)
System ........................................... B      3.7-15 B 3.7.5                                                                          Main Condenser Offgas ............................... B 3.7-22
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                                                                  ii                                                                                                                                                                                    Revision No. 145 TABLE OF CONTENTS (continued)
 
B 3.7                                                              PLANT SYSTEMS  (continued)
B 3.7.6                                                                          Main Turbine Bypass System .......................... B      3.7-25 B 3.7.7                                                                          Spent Fuel Storage Pool Water Level ................. B      3.7-29
 
B 3.8                                                              ELECTRICAL POWER SYSTEMS ................................ B      3.8-1 B 3.8.1                                                                          AC Sources      Operating ............................... B 3.8-1 B 3.8.2                                                                          AC Sources      Shutdown ................................ B 3.8-40 B 3.8.3                                                                          Diesel Fuel Oil, Lube Oil, and Starting Air ......... B      3.8-48 B 3.8.4                                                                          DC Sources      Operating ............................... B 3.8-58 B 3.8.5                                                                          DC Sources      Shutdown ................................ B 3.8-72 B 3.8.6  Battery Parameters .................................. B      3.8-77 B 3.8.7                                                                          Distribution Systems      Operating ..................... B 3.8-83 B 3.8.8                                                                          Distribution Systems      Shutdown ...................... B 3.8-94
 
B 3.9                                                              REFUELING OPERATIONS .................................... B      3.9-1 B 3.9.1  Refueling Equipment Interlocks ...................... B      3.9-1 B 3.9.2                                                                          Refuel Position One-Rod-Out Interlock ............... B      3.9-5 B 3.9.3                                                                          Control Rod Position ................................ B      3.9-8 B 3.9.4                                                                          Control Rod Position Indication ..................... B      3.9-10 B 3.9.5                                                                          Control Rod OPERABILITY      Refueling .................. B 3.9-14 B 3.9.6                                                                          Reactor Pressure Vessel (RPV) Water Level ........... B      3.9-17 B 3.9.7                                                                          Residual Heat Removal (RHR)      High Water Level ....... B 3.9-20 B 3.9.8                                                                          Residual Heat Removal (RHR)      Low Water Level ........ B 3.9-24
 
B 3.10                                                    SPECIAL OPERATIONS ...................................... B      3.10-1 B 3.10.1                                                                Inservice Leak and Hydrostatic Testing Operation .... B      3.10-1 B 3.10.2                                                                Reactor Mode Switch Interlock Testing ............... B      3.10-5 B 3.10.3                                                                Single Control Rod Withdrawal      Hot Shutdown ......... B 3.10-10 B 3.10.4                                                                Single Control Rod Withdrawal      Cold Shutdown ........ B 3.10-14 B 3.10.5                                                                Single Control Rod Drive (CRD)
RemovalRefueling ............................... B 3.10-19 B 3.10.6                                                                Multiple Control Rod WithdrawalRefueling ........... B      3.10-24 B 3.10.7                                                                Control Rod Testing      Operating ...................... B 3.10-27 B 3.10.8                                                                SHUTDOWN MARGIN (SDM) Test      Refueling ............... B 3.10-31
 
PBAPS UNIT 2                                                                                                                                                                                                                                                            iii                                                                                                                                                                                                                    Revision No. 150 PBAPS                                                                                                                                                                        UNIT 2 - LICENSE NO. DPR-44 TECHNICAL SPECIFICATIONS BASES PAGE REVISION LISTING
 
B                                                                                                                                                                TABLE OF CONTENTS
 
page(s)        i .................................................................................................................... Rev 145 ii ................................................................................................................... Rev 145 iii .................................................................................................................. Rev 150
 
B 2.0                                                                                                    SAFETY LIMITS (SLs)
 
page(s)        2.0-1 ........................................................................................................... Rev 1    57 2.0-2 ........................................................................................................... Rev 169 2.0-3 ........................................................................................................... Rev 169 2.0-4 ........................................................................................................... Rev 157 2.0-5 ............................................................................................................. Rev 75 2.0-6 ........................................................................................................... Rev 169 2.0-7 ............................................................................................................. Rev 75 2.0-8 ........................................................................................................... Rev 148 2.0-9 ............................................................................................................. Rev 75 2.0-10 ......................................................................................................... Rev 148
 
B 3.0                                                                                                    LIMITING CONDITION FOR OPERATION (LCO) APPLICABILITY
 
page(s)        3.0-1 ........................................................................................................... Rev 1    56 3.0-2 ........................................................................................................... Rev 152 3.0-3 ........................................................................................................... Rev 152 3.0-4 ........................................................................................................... Rev 141 3.0-5 ........................................................................................................... Rev 141 3.0-5a ......................................................................................................... Rev 141 3.0-5b ......................................................................................................... Rev 141 3.0-6 ............................................................................................................. Rev 52 3.0-7 ........................................................................................................... Rev 141 3.0-7a ......................................................................................................... Rev 141 3.0-8 ........................................................................................................... Rev 173 3.0-9 ........................................................................................................... Rev 145 3.0-9a  ..................................................................................................... Rev 145 3.0-9b ......................................................................................................... Rev 156 3.0-9c ......................................................................................................... Rev 156 3.0-9d ......................................................................................................... Rev 156 3.0-9e ......................................................................................................... Rev 156 3.0-10 ......................................................................................................... Rev 140 3.0-12 ......................................................................................................... Rev 141 3.0-13 ......................................................................................................... Rev 141 3.0-13a ....................................................................................................... Rev 141 3.0-14 ........................................................................................................... Rev 52 3.0-15 ........................................................................................................... Rev 52
 
B 3.1                                                                                                      REACTIVITY CONTROL SYSTEMS
 
page(s)        3.1-5 ............................................................................................................. Rev 72 3.1-6 ............................................................................................................. Rev 72 3.1-7 ............................................................................................................. Rev 72 3.1-8 ........................................................................................................... Rev 113 3.1-9 ........................................................................................................... Rev 113 3.1-10 ........................................................................................................... Rev 94 3.1-11 ......................................................................................................... Rev 113 3.1-14 ........................................................................................................... Rev 49 3.1-15 ............................................................................................................. Rev 2 3.1-16 ........................................................................................................... Rev 79 3.1-17 ........................................................................................................... Rev 63
 
PBAPS UNIT 2                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                      i                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                            Revision No. 175 PBAPS                                                                                                                                                                        UNIT 2 - LICENSE NO. DPR-44 TECHNICAL SPECIFICATIONS BASES PAGE REVISION LISTING
 
B 3.1                                                                                                      REACTIVITY CONTROL SYSTEMS (continued)
 
page(s)        3.1-18 ........................................................................................................... Rev      63 3.1-19 ........................................................................................................... Rev 86 3.1-20 ........................................................................................................... Rev 79 3.1-21 ........................................................................................................... Rev 63 3.1-23 ........................................................................................................... Rev 49 3.1-25 ........................................................................................................... Rev 57 3.1-26 ........................................................................................................... Rev 86 3.1-27 ........................................................................................................... Rev 57 3.1-28 ........................................................................................................... Rev 72 3.1-29 ........................................................................................................... Rev 49 3.1-31  ............................................................................................................ Rev 2 3.1-32  ............................................................................................................ Rev 2 3.1-33  .......................................................................................................... Rev 86 3.1-34 ........................................................................................................... Rev 75 3.1-35 ......................................................................................................... Rev 114 3.1-35a ......................................................................................................... Rev 63 3.1-36 ........................................................................................................... Rev 63 3.1-37 ........................................................................................................... Rev 86 3.1-38 ........................................................................................................... Rev 61 3.1-39 ......................................................................................................... Rev 114 3.1-40 ......................................................................................................... Rev 114 3.1-41 ......................................................................................................... Rev 114 3.1-42 ......................................................................................................... Rev 114 3.1-43 ......................................................................................................... Rev 159 3.1-44 ......................................................................................................... Rev 114 3.1-45 ......................................................................................................... Rev 114 3.1-46 ......................................................................................................... Rev 140 3.1-47 ......................................................................................................... Rev 130 3.1-48 ........................................................................................................... Rev 75 3.1-49 ........................................................................................................... Rev 57 3.1-50 ........................................................................................................... Rev 57 3.1-51 ........................................................................................................... Rev 86 3.1-52 ........................................................................................................... Rev 86
 
B 3.2                                                                                                    POWER DISTRIBUTION LIMITS
 
page(s)                                                                                            3.2-1 ............................................................................................................. Rev 49 3.2-2 ............................................................................................................. Rev 49 3.2-3 ........................................................................................................... Rev 143 3.2-4 ........................................................................................................... Rev 143 3.2-5 ........................................................................................................... Rev 143 3.2-6 ........................................................................................................... Rev 157 3.2-7 ........................................................................................................... Rev 157 3.2-8 ........................................................................................................... Rev 143 3.2-9 ........................................................................................................... Rev 143 3.2-10 ......................................................................................................... Rev 143 3.2-11 ......................................................................................................... Rev 101 3.2-12 ......................................................................................................... Rev 143 3.2-12a ....................................................................................................... Rev 143 3.2-13 ......................................................................................................... Rev 169 3.2-13a ...................................................................................................... Rev. 143
 
PBAPS UNIT 2                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                      ii                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                        Revision No. 175 PBAPS                                                                                                                                                                        UNIT 2 - LICENSE NO. DPR-44 TECHNICAL SPECIFICATIONS BASES PAGE REVISION LISTING
 
B 3.3                                                                                                    INSTRUMENTATION
 
page(s)        3.3-1 ........................................................................................................... Rev 1    34 3.3 6 (inc lusive) ..................................................................................... Rev 24 3.3-7 ........................................................................................................... Rev 124 3.3-8 ........................................................................................................... Rev 143 3.3-9 ........................................................................................................... Rev 143 3.3-10 ........................................................................................................... Rev 36 3.3-11 ........................................................................................................... Rev 36 3.3-12 ........................................................................................................... Rev 50 3.3-12a ....................................................................................................... Rev 143 3.3-12b ....................................................................................................... Rev 143 3.3-12c ....................................................................................................... Rev 123 3.3-17 ........................................................................................................... Rev 87 3.3-18 ......................................................................................................... Rev 143 3.3-19 ......................................................................................................... Rev 143 3.3-20 ......................................................................................................... Rev 134 3.3-21 ......................................................................................................... Rev 134 3.3-23 ......................................................................................................... Rev 149 3.3-23a ....................................................................................................... Rev 149 3.3-24 ......................................................................................................... Rev 159 3.3-25 ......................................................................................................... Rev 159 3.3-26 ........................................................................................................... Rev 36 3.3-27 ......................................................................................................... Rev 123 3.3-27a ....................................................................................................... Rev 123 3.3-27b ....................................................................................................... Rev 143 3.3-28 ......................................................................................................... Rev 149 3.3-28a ....................................................................................................... Rev 149 3.3-29 ......................................................................................................... Rev 143 3.3-30 ......................................................................................................... Rev 114 3.3-31 ......................................................................................................... Rev 114 3.3-32 ......................................................................................................... Rev 114 3.3-33 ......................................................................................................... Rev 152 3.3-34 ......................................................................................................... Rev 143 3.3-35 ......................................................................................................... Rev 123 3.3-35a ....................................................................................................... Rev 123 3.3-35b ....................................................................................................... Rev 143 3.3 40 (inc lusive) ................................................................................. Rev 24 3.3-41 ........................................................................................................... Rev 86 3.3-42 ........................................................................................................... Rev 86 3.3-43 ........................................................................................................... Rev 86 3.3-44 ........................................................................................................... Rev 86 3.3-45 ........................................................................................................... Rev 36 3.3-46 ........................................................................................................... Rev 36 3.3-48 ......................................................................................................... Rev 143 3.3-49 ........................................................................................................... Rev 63 3.3-52 ........................................................................................................... Rev 86 3.3-53 ........................................................................................................... Rev 86 3.3-54 ........................................................................................................... Rev 86 3.3-55 ........................................................................................................... Rev 86 3.3-56 ........................................................................................................... Rev 86 3.3-57 ........................................................................................................... Rev 61
 
PBAPS UNIT 2                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                iii                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                  Revision No. 175 PBAPS                                                                                                                                                                        UNIT 2 - LICENSE NO. DPR-44 TECHNICAL SPECIFICATIONS BASES PAGE REVISION LISTING
 
B 3.3                                                                                                    INSTRUMENTATION (continued) page(s)  3.3-58 ......................................................................................................... Rev 1    46 3.3-59 ......................................................................................................... Rev 143 3.3-60 ......................................................................................................... Rev 143 3.3-61 ......................................................................................................... Rev 159 3.3-62 ......................................................................................................... Rev 143 3.3-63 ........................................................................................................... Rev 86 3.3-64 ......................................................................................................... Rev 143 3.3-67 ............................................................................................................. Rev 7 3.3-68 ............................................................................................................. Rev 3 3.3-69 ........................................................................................................... Rev 57 3.3-70 ........................................................................................................... Rev 55 3.3-71 ........................................................................................................... Rev 52 3.3-72 ............................................................................................................. Rev 3 3.3-73 ............................................................................................................. Rev 3 3.3-74 ........................................................................................................... Rev 86 3.3-75 ........................................................................................................... Rev 86 3.3-76 ......................................................................................................... Rev 132 3.3-77 ......................................................................................................... Rev 132 3.3-78 ........................................................................................................... Rev 52 3.3-79 ......................................................................................................... Rev 132 3.3-80 ......................................................................................................... Rev 132 3.3-81 ......................................................................................................... Rev 132 3.3-82 ......................................................................................................... Rev 132 3.3-83 ......................................................................................................... Rev 167 3.3-87 ......................................................................................................... Rev 159 3.3-89 ........................................................................................................... Rev 86 3.3-90 ........................................................................................................... Rev 86 3.3-91 ........................................................................................................... Rev 86 3.3-91a ....................................................................................................... Rev 137 3.3-91b ....................................................................................................... Rev 143 3.3-91c ......................................................................................................... Rev 49 3.3-91d ....................................................................................................... Rev 143 3.3-91e ....................................................................................................... Rev 143 3.3-91f ........................................................................................................ Rev 159 3.3-91g ....................................................................................................... Rev 143 3.3-91h ......................................................................................................... Rev 86 3.3-91i  ....................................................................................................... Rev 143 3.3-91j  ....................................................................................................... Rev 143 3.3-98 ........................................................................................................... Rev 21 3.3-99 ......................................................................................................... Rev 145 3.3-100 ....................................................................................................... Rev 145 3.3-101 ....................................................................................................... Rev 145 3.3-102 ....................................................................................................... Rev 145 3.3-103 ....................................................................................................... Rev 145 3.3-104 ....................................................................................................... Rev 145 3.3-106 ....................................................................................................... Rev 145 3.3-111 ......................................................................................................... Rev 78 3.3-117 ....................................................................................................... Rev 145 3.3-118 ....................................................................................................... Rev 159 3.3-120 ....................................................................................................... Rev 159 3.3-121 ....................................................................................................... Rev 159 3.3-122 ....................................................................................................... Rev 145 3.3-123 ....................................................................................................... Rev 159 3.3-124 ....................................................................................................... Rev 159 3.3-125 ......................................................................................................... Rev 83 3.3-126 ....................................................................................................... Rev 159 3.3-127 ......................................................................................................... Rev 86
 
PBAPS UNIT 2                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                iv                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                  Revision No. 175 PBAPS                                                                                                                                                                        UNIT 2 - LICENSE NO. DPR-44 TECHNICAL SPECIFICATIONS BASES PAGE REVISION LISTING
 
B 3.3                                                                                                    INSTRUMENTATION (continued)
 
page(s)        3.3-128 ......................................................................................................... Rev 8    6 3.3-129 ......................................................................................................... Rev 86 3.3-135 ....................................................................................................... Rev 159 3.3-136 ....................................................................................................... Rev 159 3.3-137 ....................................................................................................... Rev 159 3.3-138 ......................................................................................................... Rev 86 3.3-139 ......................................................................................................... Rev 86 3.3-140 ......................................................................................................... Rev 86 3.3-140a ..................................................................................................... Rev 145 3.3-140b ..................................................................................................... Rev 145 3.3-140c ..................................................................................................... Rev 145 3.3-140d ..................................................................................................... Rev 145 3.3-140e ..................................................................................................... Rev 145 3.3-140f ...................................................................................................... Rev 145 3.3-140g ..................................................................................................... Rev 145 3.3-140h ..................................................................................................... Rev 145 3.3-140i ...................................................................................................... Rev 145 3.3-140j ...................................................................................................... Rev 145 3.3-141 ....................................................................................................... Rev 134 3.3-142 ......................................................................................................... Rev 48 3.3-143 ......................................................................................................... Rev 48 3.3-144 ......................................................................................................... Rev 57 3.3-145 ......................................................................................................... Rev 57 3.3-147 ....................................................................................................... Rev 143 3.3-148 ....................................................................................................... Rev 134 3.3-149 ....................................................................................................... Rev 134 3.3-149a ....................................................................................................... Rev 75 3.3-150 ......................................................................................................... Rev 75 3.3-151 ......................................................................................................... Rev 20 3.3-155 ......................................................................................................... Rev 32 3.3-156 ......................................................................................................... Rev 75 3.3-157 ....................................................................................................... Rev 135 3.3-158 ....................................................................................................... Rev 145 3.3-159 ....................................................................................................... Rev 145 3.3-159a ....................................................................................................... Rev 57 3.3-160 ....................................................................................................... Rev 159 3.3-161 ......................................................................................................... Rev 48 3.3-162 ......................................................................................................... Rev 45 3.3-165 ......................................................................................................... Rev 86 3.3-166 ....................................................................................................... Rev 134 3.3-167 ....................................................................................................... Rev 114 3.3-168 ....................................................................................................... Rev 143 3.3-169 - 171 (inclusive) .............................................................................. Rev 1 3.3-172 ....................................................................................................... Rev 145 3.3-173 ........................................................................................................... Rev 1 3.3-174 ....................................................................................................... Rev 145 3.3-175 ........................................................................................................... Rev 1 3.3-176 .......................................................................................................... Rev 1 3.3-177 ......................................................................................................... Rev 86 3.3-178 ......................................................................................................... Rev 86 3.3-179 - 181 (inclusive) ............................................................................... Rev 1 3.3-182 ....................................................................................................... Rev 145 3.3-183 ........................................................................................................... Rev 1 3.3-184 ......................................................................................................... Rev 86 3.3-185 ......................................................................................................... Rev 86
 
PBAPS UNIT 2                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                  v                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                        Revision No. 175 PBAPS                                                                                                                                                                        UNIT 2 - LICENSE NO. DPR-44 TECHNICAL SPECIFICATIONS BASES PAGE REVISION LISTING
 
B 3.3                                                                                                    INSTRUMENTATION (continued)
 
page(s)        3.3-186 ......................................................................................................... Rev 8    6 3.3-187 ........................................................................................................... Rev 5 3.3-188 ........................................................................................................ Rev 88 3.3-189 ........................................................................................................ Rev 88 3.3-190 ........................................................................................................ Rev 88 3.3-191 - 192 .................................................................................................. Rev 5 3.3-193 ....................................................................................................... Rev 159 3.3-194 ....................................................................................................... Rev 159 3.3-195 ....................................................................................................... Rev 159 3.3-196 ........................................................................................................... Rev 5 3.3-197 ......................................................................................................... Rev 86 3.3-198 ......................................................................................................... Rev 86 3.3-199 ........................................................................................................... Rev 1 3.3-200 ........................................................................................................... Rev 1 3.3-201 ........................................................................................................... Rev 1 3.3-202 ........................................................................................................... Rev 1 3.3-203 ......................................................................................................... Rev 66 3.3-204 ......................................................................................................... Rev 86 3.3-205 ......................................................................................................... Rev 86
 
B 3.4                                                                                                    REACTOR COOLANT SYSTEM (RCS)
 
page(s)        3.4-1 ........................................................................................................... Rev 1    37 3.4-2 ........................................................................................................... Rev 137 3.4-3 ........................................................................................................... Rev 123 3.4-4 ............................................................................................................. Rev 50 3.4-5 ........................................................................................................... Rev 123 3.4-6 ............................................................................................................. Rev 50 3.4-7 ........................................................................................................... Rev 123 3.4-8 ........................................................................................................... Rev 123 3.4-9 ........................................................................................................... Rev 123 3.4-10 ......................................................................................................... Rev 169 3.4-14 ......................................................................................................... Rev 143 3.4-15 ......................................................................................................... Rev 114 3.4-16 ......................................................................................................... Rev 148 3.4-16a ....................................................................................................... Rev 142 3.4-17 ......................................................................................................... Rev 140 3.4-18 ......................................................................................................... Rev 148 3.4-20 ......................................................................................................... Rev 164 3.4-21 ......................................................................................................... Rev 164 3.4-22 ......................................................................................................... Rev 164 3.4-23 ......................................................................................................... Rev 164 3.4-24 ........................................................................................................... Rev 93 3.4-25 ........................................................................................................... Rev 93 3.4-26 ........................................................................................................... Rev 93 3.4-26a ....................................................................................................... Rev 152 3.4-27 ........................................................................................................... Rev 86 3.4-28 ........................................................................................................... Rev 93 3.4-29 ........................................................................................................... Rev 75 3.4-30 ........................................................................................................... Rev 75 3.4-31 ........................................................................................................... Rev 75 3.4-32 ........................................................................................................... Rev 86 3.4-34 ......................................................................................................... Rev 126 3.4-35 ......................................................................................................... Rev 160 3.4-36 ......................................................................................................... Rev 175 3.4-36a ....................................................................................................... Rev 175
 
PBAPS UNIT 2                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                          vi                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                  Revision No. 175 PBAPS                                                                                                                                                                        UNIT 2 - LICENSE NO. DPR-44 TECHNICAL SPECIFICATIONS BASES PAGE REVISION LISTING
 
B 3.4                                                                                                    REACTOR COOLANT SYSTEM (RCS) (continued)
 
3.4-37 ......................................................................................................... Rev 175 3.4-37a ....................................................................................................... Rev 127 3.4-37b ....................................................................................................... Rev 126 3.4-39 ......................................................................................................... Rev 126 3.4-41 ......................................................................................................... Rev 160 3.4-42 ......................................................................................................... Rev 127 3.4-42a ....................................................................................................... Rev 126 3.4-43 ......................................................................................................... Rev 102 3.4-44 ......................................................................................................... Rev 102 3.4-45 ......................................................................................................... Rev 102 3.4-46 ......................................................................................................... Rev 102 3.4-47 ......................................................................................................... Rev 102 3.4-48 ......................................................................................................... Rev 102 3.4-49 ......................................................................................................... Rev 102 3.4-50 ......................................................................................................... Rev 102 3.4-51 ......................................................................................................... Rev 102 3.4-52 ........................................................................................................... Rev 49 3.4-53 ......................................................................................................... Rev 114 B 3.5                                                                                                    EMERGENCY CORE COOLING SYSTEMS (ECCS) AND REACTOR CORE ISOLATION COOLING (RCIC) SYSTEM page(s)        3.5-1 ........................................................................................................... Rev 1    45 3.5-3 ........................................................................................................... Rev 110 3.5-4 ........................................................................................................... Rev 147 3.5-5 ........................................................................................................... Rev 126 3.5-6 ........................................................................................................... Rev 145 3.5-6a ......................................................................................................... Rev 159 3.5-7 ........................................................................................................... Rev 159 3.5-8 ........................................................................................................... Rev 172 3.5-9 ........................................................................................................... Rev 126 3.5-10 ......................................................................................................... Rev 127 3.5-10a ....................................................................................................... Rev 126 3.5-11 ........................................................................................................... Rev 86 3.5-12 ......................................................................................................... Rev 140 3.5-13 ........................................................................................................... Rev 99 3.5-14 ......................................................................................................... Rev 143 3.5-15 ......................................................................................................... Rev 166 3.5-16 ........................................................................................................... Rev 86 3.5-17 ......................................................................................................... Rev 172 3.5-18 ......................................................................................................... Rev 145 3.5-19 ......................................................................................................... Rev 145 3.5-19a - 23 (Deleted) .............................................................................. Rev 145  3.5-24 ......................................................................................................... Rev 145 3.5-25 ......................................................................................................... Rev 145 3.5-26 ......................................................................................................... Rev 159 3.5-27 ......................................................................................................... Rev 127 3.5-27a ....................................................................................................... Rev 126 3.5-28 ......................................................................................................... Rev 143 3.5-29 ......................................................................................................... Rev 166 3.5-30 ........................................................................................................... Rev 66 3.5-31 ......................................................................................................... Rev 145 3.5-32 ......................................................................................................... Rev 145 3.5-32a ....................................................................................................... Rev 145 3.5-33 ......................................................................................................... Rev 145 3.5-34 ......................................................................................................... Rev 145 3.5-35 ......................................................................................................... Rev 145 3.5-36 ......................................................................................................... Rev 145
 
PBAPS UNIT 2                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                          vii                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                              Revision No. 175 PBAPS                                                                                                                                                                        UNIT 2 - LICENSE NO. DPR-44 TECHNICAL SPECIFICATIONS BASES PAGE REVISION LISTING
 
B 3.5                                                                                                    EMERGENCY CORE COOLING SYSTEMS (ECCS) AND REACTOR CORE ISOLATION COOLING (RCIC) SYSTEM (continued)
 
page(s)        3.5-37 ......................................................................................................... Rev 14    5 3.5-38 ......................................................................................................... Rev 145 3.5-39 ......................................................................................................... Rev 166
 
B 3.6                                                                                                    CONTAINMENT SYSTEMS
 
page(s)        3.6-1 ............................................................................................................. Rev 27 3.6-2 ........................................................................................................... Rev 114 3.6-3 ............................................................................................................. Rev 66 3.6-4 ........................................................................................................... Rev 161 3.6-5 ........................................................................................................... Rev 161 3.6-7 ........................................................................................................... Rev 114 3.6-11 ......................................................................................................... Rev 159 3.6-12 ........................................................................................................... Rev 86 3.6-13 ......................................................................................................... Rev 114 3.6-16 ......................................................................................................... Rev 163 3.6-17 ......................................................................................................... Rev 163 3.6-18 ......................................................................................................... Rev 145 3.6-19 ......................................................................................................... Rev 159 3.6-20 ......................................................................................................... Rev 159 3.6-21 ......................................................................................................... Rev 163 3.6-22 ......................................................................................................... Rev 163 3.6-23 ......................................................................................................... Rev 159 3.6-23a ....................................................................................................... Rev 145 3.6-24 ........................................................................................................... Rev 91 3.6-25 ........................................................................................................... Rev 86 3.6-26 ........................................................................................................... Rev 86 3.6-27 ......................................................................................................... Rev 161 3.6-28 ........................................................................................................... Rev 86 3.6-29 ......................................................................................................... Rev 144 3.6-30  ........................................................................................................ Rev 161 3.6-31  .......................................................................................................... Rev 18 3.6-33  ........................................................................................................ Rev 114 3.6-35  .......................................................................................................... Rev 91 3.6-38  ........................................................................................................ Rev 159 3.6-39  .......................................................................................................... Rev 91 3.6-40 ........................................................................................................... Rev 86 3.6-41 ........................................................................................................... Rev 86 3.6-43 ........................................................................................................... Rev 44 3.6-45 ......................................................................................................... Rev 159 3.6-46 ........................................................................................................... Rev 86 3.6-47 ........................................................................................................... Rev 86 3.6 51 (inc lusive) ................................................................................. Rev 24 3.6-52 ......................................................................................................... Rev 165 3.6-54 ......................................................................................................... Rev 145 3.6-55 ........................................................................................................... Rev 86 3.6-56 ......................................................................................................... Rev 114 3.6-57 ......................................................................................................... Rev 159 3.6-58 ......................................................................................................... Rev 159 3.6-59 ......................................................................................................... Rev 140
 
PBAPS UNIT 2                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                      viii                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                        Revision No. 175 PBAPS                                                                                                                                                                        UNIT 2 - LICENSE NO. DPR-44 TECHNICAL SPECIFICATIONS BASES PAGE REVISION LISTING
 
B 3.6                                                                                                CONTAINMENT SYSTEMS (continued)
 
page(s)        3.6-59a ....................................................................................................... Rev 127 3.6-59b ....................................................................................................... Rev 126 3.6-60 ......................................................................................................... Rev 114 3.6-61 ......................................................................................................... Rev 126 3.6-62 ......................................................................................................... Rev 159 3.6-63 ......................................................................................................... Rev 130 3.6-63a ....................................................................................................... Rev 127 3.6-63b ....................................................................................................... Rev 126 3.6-63c ....................................................................................................... Rev 126 3.6-63d ....................................................................................................... Rev 126 3.6-63e ....................................................................................................... Rev 159 3.6-63f ........................................................................................................ Rev 130 3.6-63g ....................................................................................................... Rev 127 3.6-63h ....................................................................................................... Rev 126 3.6-64 ........................................................................................................... Rev 80 3.6-70 ........................................................................................................... Rev 80 3.6-71 ......................................................................................................... Rev 158 3.6-72 ......................................................................................................... Rev 158 3.6-73 ........................................................................................................... Rev 75 3.6-74 ......................................................................................................... Rev 145 3.6-75 ......................................................................................................... Rev 145 3.6-76 ......................................................................................................... Rev 120 3.6-77 ........................................................................................................... Rev 97 3.6-78 ........................................................................................................... Rev 75 3.6-79 ......................................................................................................... Rev 145 3.6-81 ........................................................................................................... Rev 57 3.6-82 ......................................................................................................... Rev 145 3.6-83 ......................................................................................................... Rev 140 3.6-84 ........................................................................................................... Rev 86 3.6-85 ......................................................................................................... Rev 162 3.6-87 ......................................................................................................... Rev 145 3.6-88 ......................................................................................................... Rev 145 3.6-89 ......................................................................................................... Rev 145 3.6-90 ......................................................................................................... Rev 166
 
B 3.7                                                                                                    PLANT SYSTEMS
 
page(s)        3.7-1 ........................................................................................................... Rev 1    14 3.7-2 ........................................................................................................... Rev 171 3.7-2a ......................................................................................................... Rev 171 3.7-3 ........................................................................................................... Rev 144 3.7-4 ........................................................................................................... Rev 159 3.7-5 ........................................................................................................... Rev 151 3.7-5a ......................................................................................................... Rev 114 3.7-5b ......................................................................................................... Rev 114 3.7-6 ........................................................................................................... Rev 168 3.7-7 ........................................................................................................... Rev 109 3.7-8 ........................................................................................................... Rev 159 3.7-9 ........................................................................................................... Rev 109 3.7-10 ......................................................................................................... Rev 166 3.7-11 ........................................................................................................... Rev 67 3.7-12 ........................................................................................................... Rev 92
 
PBAPS UNIT 2                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                ix                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                  Revision No. 175 PBAPS                                                                                                                                                                        UNIT 2 - LICENSE NO. DPR-44 TECHNICAL SPECIFICATIONS BASES PAGE REVISION LISTING
 
B 3.7                                                                                                    PLANT SYSTEMS (continued)
 
page(s)        3.7-13 ......................................................................................................... Rev 15    9 3.7-14 ........................................................................................................... Rev 86 3.7-15 ......................................................................................................... Rev 116 3.7-16 ......................................................................................................... Rev 116 3.7-16a ....................................................................................................... Rev 116 3.7-16b ....................................................................................................... Rev 121 3.7-17 ......................................................................................................... Rev 145 3.7-18 ......................................................................................................... Rev 116 3.7-19 ......................................................................................................... Rev 145 3.7-20 ......................................................................................................... Rev 145 3.7-20a ....................................................................................................... Rev 166 3.7-21 ......................................................................................................... Rev 121 3.7-23 ........................................................................................................... Rev 66 3.7-24 ........................................................................................................... Rev 86 3.7-25 ......................................................................................................... Rev 143 3.7-26 ......................................................................................................... Rev 143 3.7-27 ......................................................................................................... Rev 143 3.7-28 ......................................................................................................... Rev 143 3.7-29 ........................................................................................................... Rev 75 3.7-30 ........................................................................................................... Rev 86
 
B 3.8                                                                                                    ELECTRICAL POWER SYSTEMS
 
page(s)        3.8-1 ............................................................................................................. Rev 82 3.8-2 ............................................................................................................. Rev 90 3.8-2a ........................................................................................................... Rev 90 3.8-3 ........................................................................................................... Rev 114 3.8-5 ............................................................................................................. Rev 73 3.8-6 ............................................................................................................. Rev 52 3.8-7 ............................................................................................................... Rev 5 3.8-8 ........................................................................................................... Rev 159 3.8-9 ............................................................................................................. Rev 85 3.8-10  ............................................................................................................. Rev 5 3.8-11 ........................................................................................................... Rev 60 3.8-12 ......................................................................................................... Rev 159 3.8-13 ......................................................................................................... Rev 159 3.8-16 ......................................................................................................... Rev 159 3.8-17 ........................................................................................................... Rev 66 3.8-18 ......................................................................................................... Rev 174 3.8-19 ........................................................................................................... Rev 86 3.8-20 ........................................................................................................... Rev 86 3.8-21 ........................................................................................................... Rev 86 3.8-22 ........................................................................................................... Rev 86 3.8-23 ........................................................................................................... Rev 86 3.8-24 ......................................................................................................... Rev 139 3.8-25 ............................................................................................................. Rev 1 3.8-26 ........................................................................................................... Rev 86 3.8-27 ........................................................................................................... Rev 86 3.8-27a ......................................................................................................... Rev 57 3.8-28 ........................................................................................................... Rev 86 3.8-29 ........................................................................................................... Rev 71 3.8-30 ........................................................................................................... Rev 86 3.8-31 ........................................................................................................... Rev 57
 
PBAPS UNIT 2                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                  x                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                        Revision No. 175 PBAPS                                                                                                                                                                        UNIT 2 - LICENSE NO. DPR-44 TECHNICAL SPECIFICATIONS BASES PAGE REVISION LISTING
 
B 3.8                                                                                                    ELECTRICAL POWER SYSTEMS (continued)
 
page(s)        3.8-32 ........................................................................................................... Rev 86 3.8-33 ........................................................................................................... Rev 86 3.8-34 ........................................................................................................... Rev 86 3.8-35 ......................................................................................................... Rev 117 3.8-36 ........................................................................................................... Rev 86 3.8-37 ........................................................................................................... Rev 71 3.8-38 ........................................................................................................... Rev 86 3.8-39 ........................................................................................................... Rev 95 3.8-40 ......................................................................................................... Rev 145 3.8-42 ......................................................................................................... Rev 145 3.8-43 ......................................................................................................... Rev 145 3.8-44 ......................................................................................................... Rev 145 3.8-45 ......................................................................................................... Rev 145 3.8-46 ......................................................................................................... Rev 145 3.8-47 ........................................................................................................... Rev 16 3.8-48 ......................................................................................................... Rev 105 3.8-49 ......................................................................................................... Rev 122 3.8-51 ......................................................................................................... Rev 138 3.8-53 ......................................................................................................... Rev 138 3.8-54 ........................................................................................................... Rev 86 3.8-55 ......................................................................................................... Rev 122 3.8-56 ......................................................................................................... Rev 122 3.8-57 ......................................................................................................... Rev 122 3.8-59 ......................................................................................................... Rev 150 3.8-60 ......................................................................................................... Rev 150 3.8-60a ....................................................................................................... Rev 150 3.8-62 ......................................................................................................... Rev 159 3.8-62a ....................................................................................................... Rev 153 3.8-62b ....................................................................................................... Rev 159 3.8-63 ......................................................................................................... Rev 159 3.8-63a ....................................................................................................... Rev 159 3.8-63b ....................................................................................................... Rev 159 3.8-64 ......................................................................................................... Rev 153 3.8-65 ......................................................................................................... Rev 155 3.8-66 ......................................................................................................... Rev 150 3.8-67 ......................................................................................................... Rev 150 3.8-68 ......................................................................................................... Rev 150 3.8-69 ......................................................................................................... Rev 150 3.8-70 ......................................................................................................... Rev 150 3.8-71 ......................................................................................................... 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Rev 150 3.8-78 ......................................................................................................... Rev 150 3.8-78a ....................................................................................................... Rev 150 3.8-78b ....................................................................................................... Rev 150 3.8-78c ....................................................................................................... Rev 150
 
PBAPS UNIT 2                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                          xi                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                  Revision No. 175 PBAPS                                                                                                                                                                        UNIT 2 - LICENSE NO. DPR-44 TECHNICAL SPECIFICATIONS BASES PAGE REVISION LISTING
 
B 3.8                                                                                                    ELECTRICAL POWER SYSTEMS (continued)
 
pages(s)        3.8-79 ......................................................................................................... Rev 1    55 3.8-80 ......................................................................................................... Rev 150 3.8-81 ......................................................................................................... Rev 150 3.8-82 ......................................................................................................... Rev 150 3.8-85 ......................................................................................................... Rev 150 3.8-87 ......................................................................................................... Rev 159 3.8-88 ......................................................................................................... Rev 159 3.8-89 ......................................................................................................... Rev 159 3.8-90 ......................................................................................................... Rev 159 3.8-91 ........................................................................................................... Rev 85 3.8-92 ........................................................................................................... Rev 86 3.8-94 ......................................................................................................... Rev 145 3.8-95 ......................................................................................................... Rev 145 3.8-96 ......................................................................................................... Rev 145 3.8-97 ......................................................................................................... Rev 145
 
B 3.9                                                                                                    REFUELING OPERATIONS
 
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B 3.10                                                                              SPECIAL OPERATIONS
 
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PBAPS UNIT 2                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                          xii                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                              Revision No. 175 PBAPS                                                                                                                                                                        UNIT 2 - LICENSE NO. DPR-44 TECHNICAL SPECIFICATIONS BASES PAGE REVISION LISTING
 
B 3.10                                                                              SPECIAL OPERATIONS (continued)
 
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All remaining pages are Rev 0 dated 1/18/96.
 
PBAPS UNIT 2                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                      xiii                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                        Revision No. 175 Reactor Core SLs B 2.1.1
 
B 2.0  SAFETY LIMITS (SLs)
 
B 2.1.1  Reactor Core SLs
 
BASES
 
BACKGROUND                                                                      SLs ensure that specified acceptable fuel design limits are not exceeded during steady state operation, normal operational transients, and abnormal operational transients.
 
The fuel cladding integrity SL is set such that no fuel damage is calculated to occur if the limit is not violated.
Because fuel damage is not directly observable, a stepback approach is used to establish an SL, such that 99.9% of the fuel rods avoid transition boiling. Meeting the SL can be demonstrated by analysis that confirms no more than 0.1% of fuel rods in the core are susceptible to transition boiling or by demonstrating that the MCPR is not less than the limit specified in Specification 2.1.1.2 for General Electric (GE)
Company fuel. MCPR greater than the specified limit represents a conservative margin relative to the conditions required to maintain fuel cladding integrity.
 
The fuel cladding is one of the physical barriers that separate the radioactive materials from the environs. The integrity of this cladding barrier is related to its relative freedom from perforations or cracking. Although some corrosion or use related cracking may occur during the life of the cladding, fission product migration from this source is incrementally cumulative and continuously measurable. Fuel cladding perforations, however, can result from thermal stresses, which occur from reactor operation significantly above design conditions.
 
While fission product migration from cladding perforation is just as measurable as that from use related cracking, the thermally caused cladding perforations signal a threshold beyond which still greater thermal stresses may cause gross, rather than incremental, cladding deterioration. Therefore, the fuel cladding SL is defined with a margin to the conditions that would produce onset of transition boiling (i.e., MCPR = 1.00). These conditions represent a significant departure from the condition intended by design for planned operation. This is accomplished by having a Safety Limit Minimum Critical Power Ratio (SLMCPR) design basis, referred to as SLMCPR95/95, which corresponds to a 95%
probability at a 95% confidence level (the 95/95 MCPR criterion) that transition boiling will not occur.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                                  B 2.0-1                                                                                                                                                                                          Revision No. 157 Reactor Core SLs B 2.1.1
 
BASES
 
BACKGROUND                                                                      Operation above the boundary of the nucleate boiling regime (continued)                                        could result in excessive cladding temperature because of the onset of transition boiling and the resultant sharp reduction in heat transfer coefficient. Inside the steam film, high cladding temperatures are reached, and a cladding water (zirconium water) reaction may take place. This chemical reaction results in oxidation of the fuel cladding to a structurally weaker form. This weaker form may lose its integrity, resulting in an uncontrolled release of activity to the reactor coolant.
 
The reactor vessel water level SL ensures that adequate core cooling capability is maintained during all MODES of reactor operation. Establishment of Emergency Core Cooling System initiation setpoints higher than this safety limit provides margin such that the safety limit will not be reached or exceeded.
 
APPLICABLE                                                                      The fuel cladding must not sustain damage as a result of SAFETY ANALYSES                    normal operation and abnormal operational transients. The Tech Spec SL is set generically on a fuel product MCPR correlation basis as the MCPR which corresponds to a 95%
probability at a 95% confidence level that transition boiling will not occur, referred to as SLMCPR95/95.
 
The Reactor Protection System setpoints (LCO 3.3.1.1, "Reactor Protection System (RPS) Instrumentation"), in combination with other LCOs, are designed to prevent any anticipated combination of transient conditions for Reactor Coolant System water level, pressure, and THERMAL POWER level that would result in reaching the MCPR limit.
 
2.1.1.1                    Fuel Cladding Integrity
 
GE critical power correlations are applicable for all critical power calculations at pressures                                                                                                                                                                                                                                                                                                                                                                                                                            700 psia for GNF2 fuel (Ref. 5) and            600 psia for GNF3 fuel (Ref. 6), and core flows                                                                                                                10% of rated flow. For operation at low pressures or low flows, another basis is used, as follows:
 
The pressure drop in the bypass region is essentially all elevation head with a value > 4.5 psi; therefore, the core pressure drop at low power and flows will always be > 4.5 psi. At power, the static head inside
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                                        B 2.0-2                                                                                                                                                                                          Revision No. 169 Reactor Core SLs B 2.1.1
 
BASES
 
APPLICABLE                                                                      2.1.1.1                    Fuel Cladding Integrity  (continued)
SAFETY ANALYSES the bundle is less than the static head in the bypass region because the addition of heat reduces the density of the water. At the same time, dynamic head loss in the bundle will be greater than in the bypass region because of two phase flow effects. Analyses show that this combination of effects causes bundle pressure drop to be nearly independent of bundle power when bundle flow is 28 X 103 lb/hr and bundle pressure drop is 3.5 psi. Because core pressure drop at low power and flows will always be > 4.5 psi, the bundle flow will be > 28 X 10 3 lb/hr.
 
Full scale ATLAS test data taken at pressures from 14.7 psia (0 psig) to 800 psia (785 psig) indicate that the fuel assembly critical power with bundle flow at 28 X 103 lb/hr is approximately 3.35 MWt. This is equivalent to a THERMAL POWER > 50% RTP even when design peaking factors are considered. Therefore, a THERMAL POWER limit of 22.6% RTP for reactor pressure
                                    < 700 psia is conservative. Additional information on low flow conditions is available in Reference 4.
 
2.1.1.2 MCPR
 
The fuel cladding integrity SL is set such that no fuel damage is calculated to occur if the limit is not violated.
Since the parameters that result in fuel damage are not directly observable during reactor operation, the thermal and hydraulic conditions that result in the onset of transition boiling have been used to mark the beginning of the region in which fuel damage could occur. Although it is recognized that the onset of transition boiling would not result in damage to BWR fuel rods, the critical power at which boiling transition is calculated to occur has been adopted as a convenient limit. The Technical Specification SL value is dependent on the fuel product line and the corresponding MCPR correlation, which is cycle independent.
The value is based on the Critical Power Ration (CPR) data statistics and a 95% probability with 95% confidence that rods are not susceptible to boiling transition, referred to as MCPR95/95.
 
The SL is based on GNF2 and/or GNF3 fuel. The 700 psia threshold in the Safety Limit bounds the absolute pressure value for GNF2 fuel (700 psia per Ref. 1) and for GNF3 fuel (600 psia per Ref. 6). For cores with a single fuel product line, the SLMCPR95/95 is the MCPR95/95 for the fuel type. For cores loaded with a mix of applicable fuel types, the SLMCPR95/95 is based on the largest (i.e., most limiting) of the MCPR values for the fuel product lines that are fresh or once-burnt at the start of the cycle.
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                                  B 2.0-3                                                                                                                                                                                          Revision No. 169 Reactor Core SLs B 2.1.1
 
BASES
 
APPLICABLE SAFETY ANALYSES
 
2.1.1.3                    Reactor Vessel Water Level
 
During MODES 1 and 2 the reactor vessel water level is required to be above the top of the active fuel to provide core cooling capability. With fuel in the reactor vessel during periods when the reactor is shut down, consideration must be given to water level requirements due to the effect of decay heat. If the water level should drop below the top of the active irradiated fuel during this period, the ability to remove decay heat is reduced. This reduction in cooling capability could lead to elevated cladding temperatures and clad perforation. The core can be adequately cooled as long as water level is above 2/3 of the core height. The reactor vessel water level SL has been established at the top of the active irradiated fuel to provide a point that can be monitored and to also provide adequate margin for effective action.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                                  B 2.0-4                                                                                                                                                                                          Revision No. 157 Reactor Core SLs B 2.1.1
 
BASES  (continued)
 
SAFETY LIMITS                                        The reactor core SLs are established to protect the integrity of the fuel clad barrier to the release of radioactive materials to the environs. SL 2.1.1.1 and SL 2.1.1.2 ensure that the core operates within the fuel design criteria. SL 2.1.1.3 ensures that the reactor vessel water level is greater than the top of the active irradiated fuel in order to prevent elevated clad temperatures and resultant clad perforations.
 
APPLICABILITY                                        SLs 2.1.1.1, 2.1.1.2, and 2.1.1.3 are applicable in all MODES.
 
SAFETY LIMIT                                          Exceeding an SL may cause fuel damage and create a potential VIOLATIONS                                                              for radioactive releases in excess of 10 CFR 100, "Reactor Site Criteria," limits (Ref. 2) and 10 CFR 50.67, Accident Source Term, for accidents analyzed using AST (Ref 3).
Therefore, it is required to insert all insertable control rods and restore compliance with the SLs within 2 hours. The 2 hour Completion Time ensures that the operators take prompt remedial action and also ensures that the probability of an accident occurring during this period is minimal.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                                  B 2.0-5                                                                                                                                                                                                    Revision No. 75 Reactor Core SLs B 2.1.1
 
BASES
 
REFERENCES                                                                      1.                    NEDC-24011-P-A, General Electric Standard Application for Reactor Fuel, latest approved revision.
: 2.                    1-=0 CFR 100
: 3.                    10 CFR 50.67.
: 4.                    SIL No. 516 Supplement 2, January 19, 1996.
: 5.                    NEDC-33292P, GEXL17 Correlation for GNF2 Fuel, Rev. 3, April 2009
: 6.                    NEDC-33880P, GEXL21 Correlation for GNF3 Fuel, Rev. 1, November 2017
 
PBAPS UNIT 2                                                                                                                                                                                                                                  B 2.0-6                                                                                                                                                                                          Revision No. 169 RCS Pressure SL B 2.1.2
 
B 2.0  SAFETY LIMITS (SLs)
 
B 2.1.2  Reactor Coolant System (RCS) Pressure SL
 
BASES
 
BACKGROUND                                                                      The SL on reactor steam dome pressure protects the RCS against overpressurization. In the event of fuel cladding failure, fission products are released into the reactor coolant. The RCS then serves as the primary barrier in preventing the release of fission products into the atmosphere. Establishing an upper limit on reactor steam dome pressure ensures continued RCS integrity with regard to pressure excursions. Per the UFSAR (Ref. 1), the reactor coolant pressure boundary (RCPB) shall be designed with sufficient margin to ensure that the design conditions are not exceeded during normal operation and abnormal operational transients.
 
During normal operation and abnormal operational transients, RCS pressure is limited from exceeding the design pressure by more than 10%, in accordance with Section III of the ASME Code (Ref. 2). To ensure system integrity, all RCS components are hydrostatically tested at 125% of design pressure, in accordance with ASME Code requirements, prior to initial operation when there is no fuel in the core. Any further hydrostatic testing with fuel in the core may be done under LCO 3.10.1, "Inservice Leak and Hydrostatic Testing Operation."  Following inception of unit operation, RCS components shall be pressure tested in accordance with the requirements of ASME Code, Section XI (Ref. 3).
 
Overpressurization of the RCS could result in a breach of the RCPB reducing the number of protective barriers designed to prevent radioactive releases from exceeding the limits specified in 10 CFR 50.67, "Accident Source Term,"
(Ref. 4). If this occurred in conjunction with a fuel cladding failure, fission products could enter the containment atmosphere.
 
APPLICABLE                                                                      The RCS safety/relief valves and the Reactor Protection SAFETY ANALYSES                    System Reactor Pressure      High Function have settings established to ensure that the RCS pressure SL will not be exceeded.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                                  B 2.0-7                                                                                                                                                                                                    Revision No. 75 RCS Pressure SL B 2.1.2
 
BASES
 
APPLICABLE                                                                      The RCS pressure SL has been selected such that it is at a SAFETY ANALYSES                    pressure below which it can be shown that the integrity of (continued)                                        the system is not endangered. The reactor pressure vessel is designed to Section III, 1965 Edition of the ASME, Boiler and Pressure Vessel Code, including Addenda through the winter of 1965 (Ref. 5), which permits a maximum pressure transient of 110%, 1375 psig, of design pressure 1250 psig.
The SL of 1340 psig is measured in the reactor steam dome.
The SL has been determined to be adequate to ensure the RCS pressure does not exceed the 1375 psig RCS pressure limit (Ref. 7 and 8). The RCS is designed to the ASME Section III, 1980 Edition, including Addenda through winter of 1981 (Ref. 6), for the reactor recirculation piping, which permits a maximum pressure transient of 110% of design pressures of 1250 psig for suction piping and 1500 psig for discharge piping. The RCS pressure SL is selected to be the lowest transient overpressure allowed by the applicable codes.
 
SAFETY LIMITS                                        The maximum transient pressure allowable in the RCS pressure vessel under the ASME Code, Section III, is 110% of design pressure. The maximum transient pressure allowable in the RCS piping, valves, and fittings is 110% of design pressures of 1250 psig for suction piping and 1500 psig for discharge piping. The most limiting of these allowances is the 110%
of design pressures of 1250 psig; therefore, the SL on maximum allowable RCS pressure is established at 1340 psig, as measured at the reactor steam dome.
 
APPLICABILITY                                        SL 2.1.2 applies in all MODES.
 
SAFETY LIMIT VIOLATIONS
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                                  B 2.0-8                                                                                                                                                                                          Revision No. 148 RCS Pressure SL B 2.1.2
 
BASES
 
SAFETY LIMIT VIOLATIONS (continued)                                        Exceeding the RCS pressure SL may cause immediate RCS failure and create a potential for radioactive releases in excess of 10 CFR 50.67, "Accident Source Term," limits (Ref. 4). Therefore, it is required to insert all insertable control rods and restore compliance with the SL within 2 hours. The 2 hour Completion Time ensures that the operators take prompt remedial action and also assures that the probability of an accident occurring during the period is minimal.
 
REFERENCES                                                                      1.                    UFSAR, Section 1.5.2.2.
: 2.                    ASME, Boiler and Pressure Vessel Code, Section III, Article NB-7000.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                                  B 2.0-9                                                                                                                                                                                                    Revision No. 75 RCS Pressure SL B 2.1.2
 
BASES
 
REFERENCES                                                                      3.                    ASME, Boiler and Pressure Vessel Code, Section XI, (continued)                                                                                  Article IW-5000.
: 4.                    10 CFR 50.67.
: 5.                    ASME, Boiler and Pressure Vessel Code, Section III, 1965 Edition, including Addenda to winter of 1965.
: 6.                    ASME, Boiler and Pressure Vessel Code, Section III, 1980 Edition, Addenda to winter of 1981.
: 7.                    G-080-VC-413, "Reactor Vessel Overpressure Protection,"
GE Hitachi Nuclear Energy, 26A8321, Revision 1.
: 8.                    G-080-VC-468,"Peach Bottom Units 2 & 3 Two Safety Relief Valves Out-of-Service Evaluation," GE Hitachi Nuclear Energy, 004N6240-P, Revision 1.
 
PBAPS UNIT 2                                                                                                                                                                                                                            B 2.0-10                                                                                                                                                                                    Revision No. 148 LCO Applicability B 3.0
 
B 3.0  LIMITING CONDITION FOR OPERATION (LCO) APPLICABILITY
 
BASES
 
LCOs                                                                                                                                  LCO 3.0.1 through LCO 3.0.9 establish the general requirements applicable to all Specifications in Sections 3.1 through 3.10 and apply at all times, unless otherwise stated.
 
LCO  3.0.1                                                                      LCO 3.0.1 establishes the Applicability statement within each individual Specification as the requirement for when the LCO is required to be met (i.e., when the unit is in the MODES or other specified conditions of the Applicability statement of each Specification).
 
LCO  3.0.2                                                                      LCO 3.0.2 establishes that upon discovery of a failure to meet an LCO, the associated ACTIONS shall be met. The Completion Time of each Required Action for an ACTIONS Condition is applicable from the point in time that an ACTIONS Condition is entered, unless otherwise specified.
The Required Actions establish those remedial measures that must be taken within specified Completion Times when the requirements of an LCO are not met. This Specification establishes that:
: a.                    Completion of the Required Actions within the specified Completion Times constitutes compliance with a Specification; and
: b.                    Completion of the Required Actions is not required when an LCO is met within the specified Completion Time, unless otherwise specified.
 
There are two basic types of Required Actions. The first type of Required Action specifies a time limit in which the LCO must be met. This time limit is the Completion Time to restore an inoperable system or component to OPERABLE status or to restore variables to within specified limits. If this type of Required Action is not completed within the specified Completion Time, a shutdown may be required to place the unit in a MODE or condition in which the Specification is not applicable.  (Whether stated as a Required Action or not, correction of the entered Condition is an action that may always be considered upon entering ACTIONS.)  The second type of Required Action specifies the
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                                  B 3.0-1                                                                                                                                                                                          Revision No. 156 LCO Applicability B 3.0
 
BASES
 
LCO  3.0.2                                                                      remedial measures that permit continued operation of the (continued)                                        unit that is not further restricted by the Completion Time.
In this case, compliance with the Required Actions provides an acceptable level of safety for continued operation.
 
Completing the Required Actions is not required when an LCO is met or is no longer applicable, unless otherwise stated in the individual Specifications.
 
The nature of some Required Actions of some Conditions necessitates that, once the Condition is entered, the Required Actions must be completed even though the associated Condition no longer exists. The individual LCO's ACTIONS specify the Required Actions where this is the case.
An example of this is in LCO 3.4.9, "RCS Pressure and Temperature Limits."
 
The Completion Times of the Required Actions are also applicable when a system or component is removed from service intentionally. The ACTIONS for not meeting a single LCO adequately manage any increase in plant risk, provided any unusual external conditions (e.g., severe weather, offsite power instability) are considered. In addition, the increased risk associated with simultaneous removal of multiple structures, systems, trains, or components from service is assessed and managed in accordance with 10 CFR 50.65(a)(4). Individual Specifications may specify a time limit for performing an SR when equipment is removed from service or bypassed for testing. In this case, the Completion Times of the Required Actions are applicable when this time limit expires, if the equipment remains removed from service or bypassed.
 
When a change in MODE or other specified condition is required to comply with Required Actions, the unit may enter a MODE or other specified condition in which another Specification becomes applicable. In this case, the Completion Times of the associated Required Actions would apply from the point in time that the new Specification becomes applicable and the ACTIONS Condition(s) are entered.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                                  B 3.0-2                                                                                                                                                                                          Revision No. 152 LCO Applicability B 3.0
 
BASES  (continued)
 
LCO  3.0.3                                                                      LCO 3.0.3 establishes the actions that must be implemented when an LCO is not met and:
: a.                    An associated Required Action and Completion Time is not met and no other Condition applies; or
: b.                    The condition of the unit is not specifically addressed by the associated ACTIONS. This means that no combination of Conditions stated in the ACTIONS can be made that exactly corresponds to the actual condition of the unit. Sometimes, possible combinations of Conditions are such that entering LCO 3.0.3 is warranted; in such cases, the ACTIONS specifically state a Condition corresponding to such combinations and also that LCO 3.0.3 be entered immediately.
 
This Specification delineates the time limits for placing the unit in a safe MODE or other specified condition when operation cannot be maintained within the limits for safe operation as defined by the LCO and its ACTIONS. Planned entry into LCO 3.0.3 should be avoided. If it is not practicable to avoid planned entry into LCO 3.0.3, plant risk should be assesses and managed in accordance with 10 CFR 50.65 (a)(4), and the planned entry into LCO 3.0.3 should have less effect on plant safety than other practicable alternatives.
 
Upon entering LCO 3.0.3, 1 hour is allowed to prepare for an orderly shutdown before initiating a change in unit operation. This includes time to permit the operator to coordinate the reduction in electrical generation with the load dispatcher to ensure the stability and availability of the electrical grid. The time limits specified to enter lower MODES of operation permit the shutdown to proceed in a controlled and orderly manner that is well within the specified maximum cooldown rate and within the capabilities of the unit, assuming that only the minimum required equipment is OPERABLE. This reduces thermal stresses on components of the Reactor Coolant System and the potential for a plant upset that could challenge safety systems under conditions to which this Specification applies. The use and interpretation of specified times to complete the actions of LCO 3.0.3 are consistent with the discussion of Section 1.3, Completion Times.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                                  B 3.0-3                                                                                                                                                                                          Revision No. 152 LCO Applicability B 3.0
 
BASES
 
LCO  3.0.3                                                                      A unit shutdown required in accordance with LCO 3.0.3 may be (continued)                                        terminated and LCO 3.0.3 exited if any of the following occurs:
: a.                                              The LCO is now met.
: b.                                              The LCO is no longer applicable.
: c.                    A Condition exists for which the Required Actions have now been performed.
: d.                    ACTIONS exist that do not have expired Completion Times. These Completion Times are applicable from the point in time that the Condition is initially entered and not from the time LCO 3.0.3 is exited.
 
The time limits of Specification 3.0.3 allow 37 hours for the unit to be in MODE 4 when a shutdown is required during MODE 1 operation. If the unit is in a lower MODE of operation when a shutdown is required, the time limit for entering the next lower MODE applies. If a lower MODE is entered in less time than allowed, however, the total allowable time to enter MODE 4, or other applicable MODE, is not reduced. For example, if MODE 2 is entered in 2 hours, then the time allowed for entering MODE 3 is the next 11 hours, because the total time for entering MODE 3 is not reduced from the allowable limit of 13 hours. Therefore, if remedial measures are completed that would permit a return to MODE 1, a penalty is not incurred by having to enter a lower MODE of operation in less than the total time allowed.
 
In MODES 1, 2, and 3, LCO 3.0.3 provides actions for Conditions not covered in other Specifications. The requirements of LCO 3.0.3 do not apply in MODES 4 and 5 because the unit is already in the most restrictive Condition required by LCO 3.0.3. The requirements of LCO 3.0.3 do not apply in other specified conditions of the Applicability (unless in MODE 1, 2, or 3) because the ACTIONS of individual Specifications sufficiently define the remedial measures to be taken.
 
Exceptions to LCO 3.0.3 are provided in instances where requiring a unit shutdown, in accordance with LCO 3.0.3, would not provide appropriate remedial measures for the associated condition of the unit. An example of this is in LCO 3.7.7, "Spent Fuel Storage Pool Water Level."  LCO 3.7.7 has an Applicability of "During movement of fuel assemblies
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                                  B 3.0-4                                                                                                                                                                                          Revision No. 141 LCO Applicability B 3.0
 
BASES
 
LCO  3.0.3                                                                      in the spent fuel storage pool."  Therefore, this LCO can be (continued)                                        applicable in any or all MODES. If the LCO and the Required Actions of LCO 3.7.7 are not met while in MODE 1, 2, or 3, there is no safety benefit to be gained by placing the unit in a shutdown condition. The Required Action of LCO 3.7.7 of "Suspend movement of fuel assemblies in the spent fuel storage pool" is the appropriate Required Action to complete in lieu of the actions of LCO 3.0.3. These exceptions are addressed in the individual Specifications.
 
LCO  3.0.4                                                                      LCO 3.0.4 establishes limitations on changes in MODES or other specified conditions in the Applicability when an LCO
 
is not met. It allows placing the unit in a MODE or other specified condition stated in that Applicability (e.g., the Applicability desired to be entered) when unit conditions are such that the requirements of the LCO would not be met, in accordance with either LCO 3.0.4.a, LCO 3.0.4.b, or LCO 3.0.4.c.
 
LCO 3.0.4.a allows entry into a MODE or other specified condition in the Applicability with the LCO not met when the associated ACTIONS to be entered following entry into the MODE or other specified condition in the Applicability will permit continued operation within the MODE or other specified condition for an unlimited period of time.
Compliance with ACTIONS that permit continued operation of the unit for an unlimited period of time in a MODE or other specified condition provides an acceptable level of safety for continued operation. This is without regard to the status of the unit before or after the MODE change.
Therefore, in such cases, entry into a MODE or other specified condition in the Applicability may be made and the Required Actions followed after entry into the Applicability.
 
For example, LCO 3.0.4.a may be used when the Required Action to be entered states that an inoperable instrument channel must be placed in the trip condition within the Completion Time. Transition into a MODE or other specified condition in the Applicability may be made in accordance with LCO 3.0.4 and the channel is subsequently placed in the tripped condition within the Completion Time, which begins when the Applicability is entered. If the instrument channel cannot be placed in the tripped condition and the subsequent default ACTION ("Required Action and associated Completion Time not met") allows the OPERABLE train to be placed in operation, use of LCO 3.0.4.a is acceptable because the subsequent ACTIONS to be entered following entry into the MODE include ACTIONS (place the OPERABLE train in operation) that permit safe plant operation for an unlimited period of time in the MODE or other specified condition to be entered.
 
LCO 3.0.4.b allows entry into a MODE or other specified condition in the Applicability with the LCO not met after performance of a risk assessment addressing inoperable systems and components, consideration of the results, determination of the acceptability of entering the MODE or other specified condition in the Applicability, and establishment of risk management actions, if appropriate.
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                                  B 3.0-5                                                                                                                                                                                          Revision No. 141 LCO Applicability B 3.0
 
BASES
 
LCO  3.0.4                                                                      The risk assessment may use quantitative, qualitative, or (continued)                                        blended approaches, and the risk assessment will be conducted using the plant program, procedures, and criteria in place to implement 10 CFR 50.65(a)(4), which requires that risk impacts of maintenance activities be assessed and managed. The risk assessment, for the purposes of LCO 3.0.4.b, must take into account all inoperable Technical Specification equipment regardless of whether the equipment is included in the normal 10 CFR 50.65(a)(4) risk assessment scope. The risk assessments will be conducted using the procedures and guidance endorsed by Regulatory Guide 1.182, Assessing and Managing Risk Before Maintenance Activities at Nuclear Power Plants.
Regulatory Guide 1.182 endorses the guidance in Section 11 of NUMARC 93-01, Industry Guideline for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants.
These documents address general guidance for conduct of the risk assessment, quantitative and qualitative guidelines for establishing risk management actions, and example risk management actions. These include actions to plan and conduct other activities in a manner that controls overall risk, increased risk awareness by shift and management personnel, actions to reduce the duration of the condition, actions to minimize the magnitude of risk increases (establishment of backup success paths or compensatory measures), and determination that the proposed MODE change is acceptable. Consideration should also be given to the probability of completing restoration such that the requirements of the LCO would be met prior to the expiration of ACTIONS Completion Times that would require exiting the Applicability.
 
LCO 3.0.4.b may be used with single, or multiple systems and components unavailable. NUMARC 93-01 provides guidance relative to consideration of simultaneous unavailability of multiple systems and components.
 
The results of the risk assessment shall be considered in determining the acceptability of entering the MODE or other
 
specified condition in the Applicability, and any corresponding risk management actions. The LCO 3.0.4.b risk assessments do not have to be documented.
 
The Technical Specifications allow continued operation with equipment unavailable in MODE 1 for the duration of the
 
Completion Time. Since this is allowable, and since in general the risk impact in that particular MODE bounds the risk of transitioning into and through the applicable MODES or other specified conditions in the Applicability of the generally acceptable, as long as the risk is assessed and LCO, the use of the LCO 3.0.4.b allowance should be
 
managed as stated above. However, there is a small subset
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                            B 3.0-5a                                                                                                                                                                                    Revision No. 141 LCO Applicability B 3.0
 
BASES
 
LCO  3.0.4                                                                      of systems and components that have been determined to be (continued)                                        more important to risk and use of the LCO 3.0.4.b allowance is prohibited. The LCOs governing these system and components contain Notes prohibiting the use of LCO 3.0.4.b by stating that LCO 3.0.4.b is not applicable.
 
LCO 3.0.4.c allows entry into a MODE or other specified condition in the Applicability with the LCO not met based on
 
a Note in the Specification which states LCO 3.0.4.c is applicable. These specific allowances permit entry into MODES or other specified conditions in the Applicability when the associated ACTIONS to be entered do not provide for continued operation for an unlimited period of time and a risk assessment has not been performed. This allowance may apply to all the ACTIONS or to a specific Required Action of a Specification. The risk assessments performed to justify the use of LCO 3.0.4.b usually only consider systems and components. For this reason, LCO 3.0.4.c is typically applied to Specifications which describe values and
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                            B 3.0-5b                                                                                                                                                                                    Revision No. 141 LCO Applicability B 3.0
 
BASES
 
LCO  3.0.4                                                                      parameters (e.g., Reactor Coolant System specific activity),
(continued)                                        and may be applied to other Specifications based on NRC plant-specific approval.
 
The provisions of this Specification should not be interpreted as endorsing the failure to exercise the good practice of restoring systems or components to OPERABLE status before entering an associated MODE or other specified condition in the Applicability.
 
The provisions of LCO 3.0.4 shall not prevent changes in MODES or other specified conditions in the Applicability that are required to comply with ACTIONS. In addition, the provisions of LCO 3.0.4 shall not prevent changes in MODES or other specified conditions in the Applicability that result from any unit shutdown. In this context, a unit shutdown is defined as a change in MODE or other specified condition in the Applicability associated with transitioning from MODE 1 to MODE 2, MODE 2 to MODE 3, and MODE 3 to MODE 4.
 
Upon entry into a MODE or other specified condition in the Applicability with the LCO not met, LCO 3.0.1 and LCO 3.0.2
 
require entry into the applicable Conditions and Required Actions until the Condition is resolved, until the LCO is met, or until the unit is not within the Applicability of the Technical Specification.
 
Surveillances do not have to be performed on the associated inoperable equipment (or on variables outside the specified limits), as permitted by SR 3.0.1. Therefore, utilizing LCO 3.0.4 is not a violation of SR 3.0.1 or SR 3.0.4 for any Surveillances that have not been performed on inoperable equipment. However, SRs must be met to ensure OPERABILITY prior to declaring the associated equipment OPERABLE (or variable within limits) and restoring compliance with the affected LCO.
 
LCO  3.0.5                                                                      LCO 3.0.5 establishes the allowance for restoring equipment to service under administrative controls when it has been removed from service or declared inoperable to comply with ACTIONS. The sole purpose of this Specification is to provide an exception to LCO 3.0.2 (e.g., to not comply with the applicable Required Action(s)) to allow the performance of SRs to demonstrate:
: a.                    The OPERABILITY of the equipment being returned to service; or
: b.                    The OPERABILITY of other equipment.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                                  B 3.0-6                                                                                                                                                                                                    Revision No. 52 LCO Applicability B 3.0
 
BASES
 
LCO  3.0.5                                                                      The administrative controls ensure the time the equipment is (continued)                                        returned to service in conflict with the requirements of the ACTIONS is limited to the time absolutely necessary to perform the allowed SRs. This Specification does not provide time to perform any other preventive or corrective maintenance. LCO 3.0.5 should not be used in lieu of other practicable alternatives that comply with Required Actions and that do not require changing the MODE or other specified conditions in the Applicability in order to demonstrate equipment is OPERABLE. LCO 3.0.5 is not intended to be used repeatedly.
 
An example of demonstrating equipment is OPERABLE with the Required Actions not met would be returning a Control Rod Drive (CRD) Hydraulic Control Unit (HCU) to service in order to perform testing to demonstrate that the CRD is now OPERABLE following HCU maintenance.
 
Examples of demonstrating equipment OPERABILITY include instances in which it is necessary to take an inoperable channel or trip system out of a tripped condition that was directed by a Required Action, if there is no Required Action Note for this purpose. An example of verifying OPERABILITY of equipment removed from service is taking a tripped channel out of the tripped condition to permit the logic to function and indicate the appropriate response during performance of required testing on the inoperable channel.
 
Examples of demonstrating the OPERABILITY of other equipment are taking an inoperable channel or trip system out of the tripped condition 1) to prevent the trip function from occurring during the performance of an SR on another channel in the other trip system, or 2) to permit the logic to function and indicate the appropriate response during the performance of an SR on another channel in the same trip system.
 
The administrative controls in LCO 3.0.5 apply in all cases to systems or components in Chapter 3 of the Technical Specifications, as long as the testing could not be conducted while complying with the Required Actions. This includes the realignment or repositioning of redundant or alternate equipment or trains previously manipulated to comply with ACTIONS, as well as equipment removed from service or declared inoperable to comply with ACTIONS.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                                  B 3.0-7                                                                                                                                                                                          Revision No. 141 LCO Applicability B 3.0
 
BASES
 
LCO  3.0.6                                                                      LCO 3.0.6 establishes an exception to LCO 3.0.2 for support systems that have an LCO specified in the Technical Specifications (TS). This exception is provided because LCO 3.0.2 would require that the Conditions and Required Actions of the associated inoperable supported system LCO be entered solely due to the inoperability of the support system. This exception is justified because the actions that are required to ensure the plant is maintained in a safe condition are specified in the support systems' LCO's Required Actions. These Required Actions may include entering the supported system's Conditions and Required Actions or may specify other Required Actions.
 
When a support system is inoperable and there is an LCO specified for it in the TS, the supported system(s) are required to be declared inoperable if determined to be inoperable as a result of the support system inoperability.
However, it is not necessary to enter into the supported systems' Conditions and Required Actions unless directed to do so by the support system's Required Actions. The potential confusion and inconsistency of requirements related to the entry into multiple support and supported
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                            B 3.0-7a                                                                                                                                                                                    Revision No. 141 LCO Applicability B 3.0 BASES
 
LCO  3.0.6                                                                      systems' LCOs' Conditions and Required Actions are (continued)                                        eliminated by providing all the actions that are necessary to ensure the plant is maintained in a safe condition in the support system's Required Actions.
However, there are instances where a support system's Required Action may either direct a supported system to be declared inoperable or direct entry into Conditions and Required Actions for the supported system. This may occur immediately or after some specified delay to perform some other Required Action. Regardless of whether it is immediate or after some delay, when a support system's Required Action directs a supported system to be declared inoperable or directs entry into Conditions and Required Actions for a supported system, the applicable Conditions and Required Actions shall be entered in accordance with LCO 3.0.2.
Specification 5.5.11, "Safety Function Determination Program (SFDP)," ensures loss of safety function is detected and appropriate actions are taken. Upon entry into LCO 3.0.6, an evaluation shall be made to determine if loss of safety function exists. Additionally, other limitations, remedial actions, or compensatory actions may be identified as a result of the support system inoperability and corresponding exception to entering supported system Conditions and Required Actions. The SFDP implements the requirements of LCO 3.0.6.
Cross division checks to identify a loss of safety function for those support systems that support safety systems are required. The cross division check verifies that the supported systems of the redundant OPERABLE support system are OPERABLE, thereby ensuring safety function is retained.
If this evaluation determines that a loss of safety function exists, the appropriate Conditions and Required Actions of the LCO in which the loss of safety function exists are required to be entered. This loss of safety function does not require the assumption of additional single failures or loss of offsite power. Since operation is being restricted in accordance with the ACTIONS of the support system, any resulting temporary loss of redundancy or single failure protection is taken into account. Similarly, the ACTIONS for inoperable offsite circuit(s) and inoperable diesel generator(s) provide the necessary restriction for cross train inoperabilities. This explicit cross train verification for inoperable AC electrical power sources also acknowledges that supported system(s) are not declared inoperable solely as a result of inoperability of a normal or emergency electrical power source (refer to the definition of OPERABILITY).
When a loss of safety function is determined to exist, and the SFPD requires entry into the appropriate Conditions and Required Actions of the LCO in which the loss of safety function exists, consideration must be given to the specific type of function affected. Where a loss of function is solely due to a single Technical Specification support system (e.g., loss of automatic start due to inoperable instrumentation, or loss of pump suction source due to low tank level) the appropriate LCO is the LCO for the support system. The ACTIONS for a support system LCO adequately addresses the inoperabilities of that system without reliance on entering its supported system LCO. When the loss of function is the result of multiple support systems, the appropriate LCO is the LCO for the supported system.
 
LCO  3.0.7                                                                      There are certain special tests and operations required to be performed at various times over the life of the unit.
These special tests and operations are necessary to demonstrate select unit performance characteristics, to perform special maintenance activities, and to perform (continued)
PBAPS UNIT 2                                                                                                                                                                                                                                  B 3.0-8                                                                                                                                                                                          Revision No. 173 LCO Applicability B 3.0
 
BASES
 
LCO  3.0.7                                                                      special evolutions. Special Operations LCOs in Section 3.10 (continued)                                        allow specified TS requirements to be changed to permit performances of these special tests and operations, which otherwise could not be performed if required to comply with the requirements of these TS. Unless otherwise specified, all the other TS requirements remain unchanged. This will ensure all appropriate requirements of the MODE or other specified condition not directly associated with or required to be changed to perform the special test or operation will remain in effect.
 
The Applicability of a Special Operations LCO represents a condition not necessarily in compliance with the normal requirements of the TS. Compliance with Special Operations LCOs is optional. A special operation may be performed either under the provisions of the appropriate Special Operations LCO or under the other applicable TS requirements. If it is desired to perform the special operation under the provisions of the Special Operations LCO, the requirements of the Special Operations LCO shall be followed. When a Special Operations LCO requires another LCO to be met, only the requirements of the LCO statement are required to be met regardless of that LCO's Applicability (i.e., should the requirements of this other LCO not be met, the ACTIONS of the Special Operations LCO apply, not the ACTIONS of the other LCO). However, there are instances where the Special Operations LCO's ACTIONS may direct the other LCO's ACTIONS be met. The Surveillances of the other LCO are not required to be met, unless specified in the Special Operations LCO. If conditions exist such that the Applicability of any other LCO is met, all the other LCO's requirements (ACTIONS and SRs) are required to be met concurrent with the requirements of the Special Operations LCO.
 
LCO 3.0.8                                                                        LCO 3.0.8 establishes conditions under which systems are considered to remain capable of performing their intended safety function when associated snubbers are not capable of providing their associated support function(s). This LCO states that the supported system is not considered to be inoperable solely due to one or more snubbers not capable of performing their associated support function(s). LCO 3.0.8 may also be applied to exclude penetration flow paths with nonfunctional snubbers from LCO 3.5.4 RPV WIC Drain Time Calculation.
 
This is appropriate because a limited length of time is allowed for maintenance, testing, or repair of one or more snubbers not capable of performing their associated support function(s) and appropriate compensatory measures are specified in the snubber requirements, which are located
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                                  B 3.0-9                                                                                                                                                                                          Revision No. 145 LCO Applicability B 3.0
 
BASES
 
LCO 3.0.8                                                                                outside of the Technical Specifications (TS) under licensee (continued)                                        control. The snubber requirements do not meet the criteria in 10 CFR 50.36(c)(2)(ii}, and, as such, are appropriate for control by the licensee.
 
If the allowed time expires and the snubber(s) are unable to perform their associated support function(s), the affected supported system's and DRAIN TIME LCO(s) must be declared not met and the Conditions and Required Actions entered in accordance with LCO 3.0.2.
 
The optional allowance of TS 3.0.8 to not declare the supported (sub)system(s) LCO(s) not met for inoperable snubbers may be used at PBAPS for snubbers that have a seismic-only function in addition to other required loading functions such as a hydro-dynamic function during the applicable operating condition. Prior to using LCO 3.0.8, it must be  confirmed that the requirements of TRM 3.16 SNUBBERS are met.
 
LCO 3.0.8.a applies when one or more snubbers are not capable of providing their associated support function(s) to a single train or subsystem of a multiple train or subsystem supported system or to a single train or subsystem supported system.
LCO 3.0.8.a allows 72 hours to restore the snubber(s) before declaring the supported system inoperable or calculating the associated DRAIN TIME. The 72 hour Completion Time is reasonable based on the low probability of a seismic event concurrent with an event that would require operation of the supported system occurring while the snubber(s) are not capable of performing their associated support function and due to the availability of the redundant train of the supported system.
 
LCO 3.0.8.b applies when one or more snubbers are not capable of providing their associated support function(s) to more than one train or subsystem of a multiple train or subsystem supported system. LCO 3.0.8.b allows 12 hours to restore the snubber(s) before declaring the supported system inoperable or calculating the associated DRAIN TIME. The 12 hour Completion Time is reasonable based on the low probability of a seismic event concurrent with an event that would require operation of the supported system occurring while the snubber(s) is (are) not capable of performing their associated support function(s).
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                            B 3.0-9a                                                                                                                                                                                    Revision No. 145 LCO Applicability B 3.0
 
BASES
 
LCO 3.0.8                                                                        Reference TRM 3.16 SNUBBERS Bases for requirements and (continued)                                          commitments for maintaining minimum supporting equipment not associated with the inoperable snubber(s) when entering LCO 3.0.8.a or LCO 3.0.8.b.
When applying LCO 3.0.8.a or LCO 3.0.8.b one of the following two means of heat removal must be available 1) at least one high pressure makeup path (i.e., using high pressure coolant injection (HPCI) or reactor core isolation cooling (RCIC)) and heat removal capability (i.e.,
suppression pool cooling), including a minimum set of supporting equipment required for success, not associated with the inoperable snubber(s), or 2) at least one low pressure makeup path (i.e., low pressure coolant injection (LPCI) or core spray (CS)) and heat removal capability (i.e., suppression pool cooling or shutdown cooling),
including a minimum set of supporting equipment, not associated with the inoperable snubber(s).
 
LCO 3.0.8 requires that risk be assessed and managed.
Industry and NRC guidance on the implementation of 10 CFR 50.65(a)(4) (the Maintenance Rule) does not address seismic risk. However, use of LCO 3.0.8 should be considered with respect to other plant maintenance activities, and integrated into the existing Maintenance Rule process to the extent possible so that maintenance on any unaffected train or subsystem is properly controlled, and emergent issues are properly addressed. The risk assessment need not be quantified, but may be a qualitative awareness of the vulnerability of systems and components when one or more snubbers are not able to perform their associated support function. Reference TRM 3.16 Bases for risk management actions used to satisfy commitments T04781 and T04782.
 
LCO 3.0.8 does not apply to non-seismic functions of snubbers. Prior to using LCO 3.0.8.a for seismic snubbers that may also have non-seismic functions, it must be confirmed that at least one train of each system that is supported by the inoperable snubber(s) would remain capable of performing the systems required safety or support functions for postulated design loads other than seismic loads. LCO 3.0.8.b is not to be applied to seismic snubbers that also have non-seismic functions.
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                            B 3.0-9b                                                                                                                                                                                    Revision No. 145 LCO Applicability B 3.0
 
BASES
 
LCO 3.0.9                                                                        LCO 3.0.9 establishes conditions under which systems described in the Technical Specifications are considered to remain OPERABLE when required barriers are not capable of providing their related support function(s).
 
Barriers are doors, walls, floor plugs, curbs, hatches, installed structures or components, or other devices, not explicitly described in Technical Specifications, that support the performance of the safety function of systems described in the Technical Specifications. This LCO states that the supported system is not considered to be inoperable solely due to required barriers not capable of performing their related support function(s) under the described conditions. LCO 3.0.9 allows 30 days before declaring the supported system(s) inoperable and the LCO(s) associated with the supported system(s) not met. A maximum time based on the risk assessment.
 
If the allowed time expires and the barriers are unable to perform their related support function(s), the supported systems LCO(s) must be declared not met and the Conditions and Required Actions entered in accordance with LCO 3.0.2.
 
This provision does not apply to barriers which support ventilation systems or to fire barriers. The technical Specifications for ventilation systems provide specific Conditions for inoperable barriers. Fire Barriers are addressed by other regulatory requirements and associated plant programs. This provision does not apply to barriers which are not required to support system OPERABILITY (see NRC Regulatory Issue Summary 2001-09, Control of Hazard Barriers, dated April 2, 2001).
 
The provisions of LCO 3.0.9 are justified because of the low risk associated with required barriers not being capable of performing their related support function. This provision is based on consideration of the following initiating event categories:
 
Loss of coolant accidents High energy line breaks; Feedwater line breaks; Internal flooding; External flooding; Turbine missile ejection; and Tornado or high wind.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                            B 3.0-9c                                                                                                                                                                                    Revision No. 156 LCO Applicability B 3.0
 
BASES
 
LCO 3.0.9                                                                        The risk impact of the barriers which cannot perform their (continued)                      related support function(s) must be addressed pursuant to the risk assessment and management provision of the Maintenance Rule, 10 CFR 50.65(a)(4), and the associated implementation guidance, Regulatory Guide 1.160, Monitoring the Effectiveness of Maintenance at Nuclear Power Plants.
Regulatory Guide 1.160 endorses the guidance in Section 11 of NUMARC 93-01, Industry Guideline for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants. This guidance provides for the consideration of dynamic plant configuration issues, emergent conditions, and other aspects pertinent to plant operation with the barriers unable to perform their related support function(s). These considerations may result in risk management and other compensatory actions being required during the period that barriers are unable to perform their related support function(s).
 
LCO 3.0.9 may be applied to one or more trains or subsystems of a system supported by barriers that cannot provide their related support function(s), provided that risk is assessed and managed (including consideration of the effects on Large Early Release and from external events). If applied concurrently to more than one train or subsystem of a multiple train or subsystem supported system, the barriers supporting each of these trains or subsystems must provide their related support function(s) for different categories of initiating events. For example, LCO 3.0.9 may be applied for up to 30 days for more than one train of a multiple train supported system if the affected barrier for one train protects against internal flooding and the affected barrier for the other train protects against tornado missiles. In this example, the affected barrier may be the same physical barrier but serve different protection functions for each train.
 
The HPCI (High Pressure Coolant Injection) and RCIC (Reactor Core Isolation Cooling) systems are single train systems for injecting makeup water into the reactor during an accident or transient event. The HPCI system provides backup in case of a RCIC system failure. The ADS (Automatic Depressurization System) and low pressure ECCS coolant injection provide the core cooling function in the event of a failure of the HPCI system during an accident.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                            B 3.0-9d                                                                                                                                                                                    Revision No. 156 LCO Applicability B 3.0
 
BASES
 
LCO 3.0.9                                                                        Thus, for the purposes of LCO 3.0.9, the HPCI system, the (continued)                      RCIC system, and the ADS are considered independent subsystems of a single system and LCO 3.0.9 can be used on these single train systems in a manner similar to multiple train or subsystem systems.
 
If during the time that LCO 3.0.9 is being used, the required OPERABLE train or subsystem becomes inoperable, it must be restored to OPERABLE status within 24 hours.
Otherwise, the train(s) or subsystem(s) supported by barriers that cannot perform their related support function(s) must be declared inoperable and the associated LCOs declared not met. This 24 hour period provides time to respond to emergent conditions that would otherwise likely lead to entry into LCO 3.0.3 and a rapid plant shutdown, which is not justified given the low probability of an initiating event which would require the barrier(s) not capable of performing their related support function(s).
During this 24 hour period, the plant risk associated with the existing conditions is assessed and managed in accordance with 10 CFR 50.65(a)(4).
 
PBAPS UNIT 2                                                                                                                                                                                                                            B 3.0-9e                                                                                                                                                                                    Revision No. 156 SR Applicability B 3.0
 
B 3.0  SURVEILLANCE REQUIREMENT (SR) APPLICABILITY
 
BASES
 
SRs                                                                                                                                            SR 3.0.1 through SR 3.0.4 establish the general requirements applicable to all Specifications in Sections 3.1 through 3.10 and apply at all times, unless otherwise stated. SR 3.0.2 and SR 3.0.3 apply in Chapter 5 only when invoked by a Chapter 5 Specification.
 
SR  3.0.1                                                                                SR 3.0.1 establishes the requirement that SRs must be met during the MODES or other specified conditions in the Applicability for which the requirements of the LCO apply, unless otherwise specified in the individual SRs. This Specification is to ensure that Surveillances are performed to verify the OPERABILITY of systems and components, and that variables are within specified limits. Failure to meet a Surveillance within the specified Frequency, in accordance with SR 3.0.2, constitutes a failure to meet an LCO.
 
Systems and components are assumed to be OPERABLE when the associated SRs have been met. Nothing in this Specification, however, is to be construed as implying that systems or components are OPERABLE when:
: a.                    The systems or components are known to be inoperable, although still meeting the SRs; or
: b.                    The requirements of the Surveillance(s) are known to be not met between required Surveillance performances.
 
Surveillances do not have to be performed when the unit is in a MODE or other specified condition for which the requirements of the associated LCO are not applicable, unless otherwise specified. The SRs associated with a Special Operations LCO are only applicable when the Special Operations LCO is used as an allowable exception to the requirements of a Specification.
 
Surveillances, including Surveillances invoked by Required Actions, do not have to be performed on inoperable equipment because the ACTIONS define the remedial measures that apply.
Surveillances have to be met and performed in accordance with SR 3.0.2, prior to returning equipment to OPERABLE status.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                            B 3.0-10                                                                                                                                                                                    Revision No. 140 SR Applicability B 3.0
 
BASES
 
SR  3.0.1                                                                                Upon completion of maintenance, appropriate post maintenance (continued)                                        testing is required to declare equipment OPERABLE. This includes ensuring applicable Surveillances are not failed and their most recent performance is in accordance with SR 3.0.2. Post maintenance testing may not be possible in the current MODE or other specified conditions in the Applicability due to the necessary unit parameters not having been established. In these situations, the equipment may be considered OPERABLE provided testing has been satisfactorily completed to the extent possible and the equipment is not otherwise believed to be incapable of performing its function. This will allow operation to proceed to a MODE or other specified condition where other necessary post maintenance tests can be completed.
 
Some examples of this process are:
: a.                    Control Rod Drive maintenance during refueling that requires scram testing at > 800 psi. However, if other appropriate testing is satisfactorily completed and the scram time testing of SR 3.1.4.3 is satisfied, the control rod can be considered OPERABLE. This allows startup to proceed to reach 800 psi to perform other necessary testing.
: b.                    High pressure coolant injection (HPCI) maintenance during shutdown that requires system functional tests at a specified pressure. Provided other appropriate testing is satisfactorily completed, startup can proceed with HPCI considered OPERABLE. This allows operation to reach the specified pressure to complete the necessary post maintenance testing.
 
SR  3.0.2                                                                                SR 3.0.2 establishes the requirements for meeting the specified Frequency for Surveillances and any Required Action with a Completion Time that requires the periodic performance of the Required Action on a "once per..."
interval.
 
SR 3.0.2 permits a 25% extension of the interval specified in the Frequency. This extension facilitates Surveillance scheduling and considers plant operating conditions that may not be suitable for conducting the Surveillance (e.g.,
transient conditions or other ongoing Surveillance or maintenance activities).
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                            B 3.0-11                                                                                                                                                                                                        Revision No. 0 SR Applicability B 3.0
 
BASES
 
SR  3.0.2                                                                                The 25% extension does not significantly degrade the (continued)                                        reliability that results from performing the Surveillance at its specified Frequency. This is based on the recognition that the most probable result of any particular Surveillance being performed is the verification of conformance with the SRs. The exceptions to SR 3.0.2 are those Surveillances for which the 25% extension of the interval specified in the Frequency does not apply. These exceptions are stated in the individual Specifications. The requirements of regulations take precedence over the TS. Therefore, when a test interval is specified in the regulations, the test interval cannot be extended by the TS, and the SR include a Note in the Frequency stating, "SR 3.0.2 is not applicable."
An example of an exception when the test interval is not specified in the regulations is the Note in the Primary Containment Leakage Rate Testing Program, "SR 3.0.2 is not applicable."  This exception is provided because the program already includes extension of test intervals.
 
As stated in SR 3.0.2, the 25% extension also does not apply to the initial portion of a periodic Completion Time that requires performance on a "once per..." basis. The 25%
extension applies to each performance after the initial performance. The initial performance of the Required Action, whether it is a particular Surveillance or some other remedial action, is considered a single action with a single Completion Time. One reason for not allowing the 25%
extension to this Completion Time is that such an action usually verifies that no loss of function has occurred by checking the status of redundant or diverse components or accomplishes the function of the inoperable equipment in an alternative manner.
 
The provisions of SR 3.0.2 are not intended to be used repeatedly to extend Surveillance intervals (other than those consistent with refueling intervals) or periodic Completion Time intervals beyond those specified.
 
SR  3.0.3                                                                                SR 3.0.3 establishes the flexibility to defer declaring affected equipment inoperable or an affected variable outside the specified limits when a Surveillance has not been performed within the specified Frequency. A delay period of up to 24 hours or up to the limit of the specified
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                            B 3.0-12                                                                                                                                                                                    Revision No. 141 SR Applicability B 3.0
 
BASES
 
SR  3.0.3                                                                                Frequency, whichever is greater, applies from the point in (continued)                                        time that it is discovered that the Surveillance has not been performed in accordance with SR 3.0.2, and not at the time that the specified Frequency was not met.
 
This delay period provides adequate time to perform Surveillances that have been missed. This delay period permits the performance of a Surveillance before complying with Required Actions or other remedial measures that might preclude performance of the Surveillance.
 
The basis for this delay period includes consideration of unit conditions, adequate planning, availability of personnel, the time required to perform the Surveillance, the safety significance of the delay in completing the required Surveillance, and the recognition that the most probable result of any particular Surveillance being performed is the verification of conformance with the requirements.
 
When a Surveillance with a Frequency based not on time intervals, but upon specified unit conditions, operating situations, or requirements of regulations (e.g., prior to entering MODE 1 after each fuel loading, or in accordance with 10 CFR 50, Appendix J, as modified by approved exemptions, etc.) is discovered to not have been performed when specified, SR 3.0.3 allows for the full delay period of up to the specified Frequency to perform the Surveillance.
However, since there is not a time interval specified, the missed Surveillance should be performed at the first reasonable opportunity.
 
SR 3.0.3 provides a time limit for, and allowances for the performance of, Surveillances that become applicable as a consequence of MODE changes imposed by Required Actions.
 
SR 3.0.3 is only applicable if there is a reasonable expectation the associated equipment is OPERABLE or that variables are within limits, and it is expected that the Surveillance will be met when performed. Many factors should be considered, such as the period of time since the Surveillance was last performed, or whether the Surveillance, or a portion thereof, has ever been performed, and any other indications, tests, or activities that might support the expectation that the Surveillance will be met when performed.
An example of the use of SR 3.0.3 would be a relay contact that was not tested as required in accordance with a particular SR, but previous successful performances of the SR included the relay contact; the adjacent, physically connected relay contacts were tested during the SR performance; the subject relay contact has been tested by another SR; or historical operation of the subject relay contact has been successful. It is not sufficient to infer the behavior of the associated equipment from the performance
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                            B 3.0-13                                                                                                                                                                                    Revision No. 141 SR Applicability B 3.0
 
BASES
 
SR  3.0.3                                                                        of similar equipment. The rigor of determining whether there (continued)                                        is a reasonable expectation a Surveillance will be met when performed should increase based on the length of time since the last performance of the Surveillance. If the Surveillance has been performed recently, a review of the Surveillance history and equipment performance may be sufficient to support a reasonable expectation that the Surveillance will be met when performed. For Surveillances that have not been performed for a long period or that have never been performed, a rigorous evaluation based on objective evidence should provide a high degree of confidence that the equipment is OPERABLE. The evaluation should be documented in sufficient detail to allow a knowledgeable individual to understand the basis for the determination.
 
Failure to comply with specified Frequencies for SRs is expected to be an infrequent occurrence. Use of the delay period established by SR 3.0.3 is a flexibility which is not intended to be used repeatedly to extend Surveillance intervals. While up to 24 hours or the limit of the specified Frequency is provided to perform the missed Surveillance, it is expected that the missed Surveillance will be performed at the first reasonable opportunity. The determination of the first reasonable opportunity should include consideration of the impact on plant risk (from delaying the Surveillance as well as any plant configuration changes required or shutting the plant down to perform the Surveillance) and impact on any analysis assumptions, in addition to unit conditions, planning, availability of personnel, and the time required to perform the Surveillance.
This risk impact should be managed through
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                        B 3.0-13a                                                                                                                                                                                Revision No. 141 SR Applicability B 3.0
 
BASES
 
SR  3.0.3                                                                                the program in place to implement 10 CFR 50.65(a)(4) and its (continued)                                        implementation guidance, NRC Regulatory Guide 1.182, Assessing and Managing Risk Before Maintenance Activities at Nuclear Power Plants. This Regulatory Guide addresses consideration of temporary and aggregate risk impacts, determination of risk management action thresholds, and risk management action up to and including plant shutdown. The missed Surveillance should be treated as an emergent condition as discussed in the Regulatory Guide. The risk evaluation may use quantitative, qualitative, or blended methods. The degree of depth and rigor of the evaluation should be commensurate with the importance of the component.
Missed Surveillances for important components should be analyzed quantitatively. If the results of the risk evaluation determine the risk increase is significant, this evaluation should be used to determine the safest course of action. All missed Surveillances will be placed in the licensees Corrective Action Program.
 
If a Surveillance is not completed within the allowed delay period, then the equipment is considered inoperable or the variable is considered outside the specified limits and the Completion Times of the Required Actions for the applicable LCO Conditions begin immediately upon expiration of the delay period. If a Surveillance is failed within the delay period, then the equipment is inoperable, or the variable is outside the specified limits and the Completion Times of the Required Actions for the applicable LCO Conditions begin immediately upon the failure of the Surveillance.
 
Completion of the Surveillance within the delay period allowed by this Specification, or within the Completion Time of the ACTIONS, restores compliance with SR 3.0.1.
 
SR  3.0.4                                                                                SR 3.0.4 establishes the requirement that all applicable SRs must be met before entry into a MODE or other specified condition in the Applicability.
 
This Specification ensures that system and component OPERABILITY requirements and variable limits are met before
 
entry into MODES or other specified conditions in the Applicability for which these systems and components ensure safe operation of the unit. The provisions of this Specification should not be interpreted as endorsing the failure to exercise the good practice of restoring systems or components to OPERABLE status before entering an associated MODE or other specified condition in the Applicability.
 
A provision is included to allow entry into a MODE or other specified condition in the Applicability when an LCO is not met due to Surveillance not being met in accordance with LCO 3.0.4.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                            B 3.0-14                                                                                                                                                                                              Revision No. 52 SR Applicability B 3.0
 
BASES
 
SR  3.0.4                                                                                However, in certain circumstances, failing to meet an SR (continued)                                        will not result in SR 3.0.4 restricting a MODE change or other specified condition change. When a system, subsystem, division, component, device, or variable is inoperable or outside its specified limits, the associated SR(s) are not required to be performed, per SR 3.0.1, which states that surveillances do not have to be performed on inoperable equipment. When equipment is inoperable, SR 3.0.4 does not apply to the associated SR(s) since the requirement for the SR(s) to be performed is removed. Therefore, failing to perform the Surveillance(s) within the specified Frequency does not result in an SR 3.0.4 restriction to changing MODES or other specified conditions of the Applicability.
However, since the LCO is not met in this instance, LCO 3.0.4 will govern any restrictions that may (or may not) apply to MODE or other specified condition changes. SR 3.0.4 does not restrict changing MODES or other specified conditions of the Applicability when a Surveillance has not been performed within the specified Frequency, provided the requirement to declare the LCO not met has been delayed in accordance with SR 3.0.3.
 
The provisions of SR 3.0.4 shall not prevent entry into MODES or other specified conditions in the Applicability that are required to comply with ACTIONS. In addition, the provisions of SR 3.0.4 shall not prevent changes in MODES or other specified conditions in the Applicability that result from any unit shutdown. In this context, a unit shutdown is defined as a change in MODE or other specified condition in the Applicability associated with transitioning from MODE 1 to MODE 2, MODE 2 to MODE 3, and MODE 3 to MODE 4.
 
The precise requirements for performance of SRs are specified such that exceptions to SR 3.0.4 are not necessary. The specific time frames and conditions necessary for meeting the SRs are specified in the Frequency, in the Surveillance, or both. This allows performance of Surveillances when the prerequisite condition(s) specified in a Surveillance procedure require entry into the MODE or other specified condition in the Applicability of the associated LCO prior to the performance or completion of a Surveillance. A Surveillance that could not be performed until after entering the LCOs Applicability, would have its Frequency specified such that it is not "due" until the specific conditions needed are met. Alternately, the Surveillance may be stated in the form of a Note, as not required (to be met or performed) until a particular event, condition, or time has been reached. Further discussion of the specific formats of SRs' annotation is found in Section 1.4, Frequency.
 
PBAPS UNIT 2                                                                                                                                                                                                                            B 3.0-15                                                                                                                                                                                              Revision No. 52 SDM B 3.1.1
 
B 3.1  REACTIVITY CONTROL SYSTEMS
 
B 3.1.1  SHUTDOWN MARGIN (SDM)
 
BASES
 
BACKGROUND                                                                                                                                                                          SDM requirements are specified to ensure:
: a.                                        The reactor can be made subcritical from all operating conditions and transients and Design Basis Events;
: b.                                        The reactivity transients associated with postulated accident conditions are controllable within acceptable limits; and
: c.                                        The reactor will be maintained sufficiently subcritical                                                                                                              to preclude inadvertent criticality in the shutdown condition.
 
These requirements are satisfied by the control rods, as described in the UFSAR Section 1.5 (Ref.                                                                                                                                                                                                                                                                                                                                                                                                              1), which can compensate for the reactivity effects of the fuel and water temperature changes experienced during all operating conditions.
 
APPLICABLE                                                                                                                                                                          The control rod drop accident (CRDA) analysis (Refs. 2 SAFETY ANALYSES                                                                                                                                                                          and                              3) assumes the core is subcritical                                                                                                                                                                                                                                                                                                                                                  with the highest worth control rod withdrawn. Typically, the first control rod withdrawn has a very high reactivity worth and, should the core be critical during the withdrawal of the first control rod, the consequences of a CRDA could exceed the fuel damage limits for a CRDA (see Bases for LCO                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                3.1.6, "Rod Pattern Control"). Also, SDM is assumed as an initial condition for the control rod removal error during refueling (Ref.                                                                                                                                                      4) and fuel assembly insertion error during refueling (Ref.                                                                                                                                                      5) accidents. The analysis of these reactivity insertion events assumes the refueling interlocks are OPERABLE when the reactor is in the refueling mode of operation. These interlocks prevent the withdrawal of more than one control rod from the core during refueling.  (Special consideration and requirements for multiple control rod withdrawal during refueling are covered in Special Operations LCO                                                                                                                                                                                                                                                                                                                                        3.10.6, "Multiple Control Rod      Withdrawal      Refueling.")  The analysis assumes this condition is acceptable since the core will be shut down with the highest worth control rod withdrawn, if adequate
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                                                                                                                                                        B 3.1-1                                                                                                                                                                                                                                                              Revision No. 0 SDM B 3.1.1
 
BASES
 
APPLICABLE                                                                                                                                                                          SDM has been demonstrated. Prevention or mitigation of SAFETY ANALYSES                                                                                                                                                                          reactivity insertion events is necessary to limit energy (continued)                                                                                                                                                                          deposition in the fuel to prevent significant fuel damage, which could result in undue release of radioactivity.
Adequate SDM ensures inadvertent criticalities and potential CRDAs involving high worth control rods (namely the first control rod withdrawn) will not cause significant fuel damage.
 
SDM satisfies Criterion                                                                                                                                                                                                                                      2 of the NRC Policy Statement.
 
LCO                                                                                                                                                                          The specified SDM limit accounts for the uncertainty in the demonstration of SDM by testing. Separate SDM limits are provided for testing where the highest worth control rod is determined analytically or by measurement. This is due to the reduced uncertainty in the SDM test when the highest worth control rod is determined by measurement. When SDM is demonstrated by calculations not associated with a test (e.g., to confirm SDM during the fuel loading sequence),
additional margin is included to account for uncertainties in the calculation. To ensure adequate SDM during the design process, a design margin is included to account for uncertainties in the design calculations (Ref.                                                                                                                                                                                                                                                                                                                                                                                                                                                                            6).
 
APPLICABILITY                                                                                                                                                                          In MODES                                                                                1 and                                                  2, SDM must be provided because subcriticality with the highest worth                                                                                                                                                                                                                                                                                                                                                                                control rod withdrawn is assumed in the CRDA analysis (Ref.                                                                                                                                                                                                                                                                                                                                                                                2). In MODES                                                                                                                                  3 and                              4, SDM is required to ensure the reactor will be held subcritical with margin for a single withdrawn control rod.
SDM is required in MODE                                                                                                                                                                                                                                      5 to prevent an open vessel, inadvertent criticality during the withdrawal of a single control rod from a core cell containing one or more fuel assemblies (Ref. 4) or a fuel assembly insertion error (Ref.                                                  5).
 
ACTIONS                                                  A.1
 
With SDM not within the limits of the LCO in MODE                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                        1 or                                        2, SDM must be restored within 6                                                                                                                                                                                                                                                                                                  hours. Failure to meet the specified SDM may be caused by a control rod that cannot be inserted. The allowed Completion Time of 6                                                                                                                                                                                                                                                                                                                                                                                                                                            hours is
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                                                                                                                                                        B 3.1-2                                                                                                                                                                                                                                                              Revision No. 0 SDM B 3.1.1
 
BASES
 
ACTIONS                                                                                                                                                                          A.1          (continued)
 
acceptable, considering that the reactor can still be shut down, assuming no failures of additional control rods to insert, and the low probability of an event occurring during this interval.
 
B.1
 
If the SDM cannot be restored, the plant must be brought to MODE                                        3 in 12                                                                      hours, to prevent the potential for further reductions in available SDM (e.g., additional stuck control rods). The allowed Completion Time of 12                                                                                                                                                                                                                                                                                                                                                                                                                        hours is reasonable, based on operating experience, to reach MODE                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                3 from full power conditions in an orderly manner and without challenging plant systems.
 
C.1
 
With SDM not within limits in MODE                                                                                                                                                                                                                                                                                                                                                  3, the operator must immediately initiate action to fully insert all insertable control rods. Action must continue until all insertable control rods are fully inserted.This action results in the least reactive condition for the core.
 
D.1, D.2, D.3, and D.4
 
With SDM not within limits in MODE                                                                                                                                                                                                                                                                                                                                                  4, the operator must immediately initiate action to fully insert all insertable control rods. Action must continue until all insertable control rods are fully inserted. This action results in the least reactive condition for the core. Action must also be initiated within 1                                                                                                                                                                                    hour to provide means for control of potential radioactive releases. This includes ensuring secondary                                                                                          containment is OPERABLE; at least one Standby Gas Treatment (SGT) subsystem for Unit 2 is OPERABLE; and secondary containment isolation capability (i.e., at least one secondary containment isolation valve and associated instrumentation are OPERABLE, or other acceptable administrative controls to assure isolation capability), in each associated secondary containment penetration flow path not isolated that is assumed to be isolated to mitigate radioactivity releases. This may be performed as
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                                                                                                                                                        B 3.1-3                                                                                                                                                                                                                                                              Revision No. 0 SDM B 3.1.1
 
BASES
 
ACTIONS                                                                                                                                                                          D.1, D.2, D.3, and D.4(continued)
 
an administrative check, by examining logs or other information, to determine if the components are out of service for maintenance or other reasons. It is not necessary to perform the surveillances needed to demonstrate the OPERABILITY of the components. If, however, any required component is inoperable, then it must be restored to OPERABLE status. In this case, SRs may need to be performed to restore the component to OPERABLE status.
Actions must continue until all required components are OPERABLE.
 
E.1, E.2, E.3, E.4, and E.5
 
With SDM not within limits in MODE                                                                                                                                                                                                                                                                                                                                                  5, the operator must immediately suspend CORE ALTERATIONS that could reduce SDM, e.g., insertion of fuel in the core or the withdrawal of control rods. Suspension of these activities shall not preclude completion of movement of a component to a safe condition. Inserting control rods or removing fuel from the core will reduce the total reactivity and are therefore excluded from the suspended actions.
 
Action must also be immediately initiated                                                                                                                                                                                                                                                                                                                                                                                                                        to fully insert all insertable control rods in core cells containing one or more fuel assemblies. Action must continue until all insertable control rods in core cells containing one or more fuel assemblies have been fully inserted. Control rods in core cells containing no fuel assemblies do not affect the reactivity of the core and therefore do not have to be inserted.
 
Action must also be initiated within 1                                                                                                                                                                                                                                                                                                                                                                                          hour to provide means for control of potential radioactive releases. This includes ensuring secondary containment is OPERABLE; at least one SGT subsystem for Unit 2 is OPERABLE; and secondary containment isolation capability (i.e., at least one secondary containment isolation valve and associated instrumentation are OPERABLE, or other acceptable administrative controls to assure isolation capability), in each associated secondary containment penetration flow path not isolated that is assumed to be isolated to mitigate radioactive releases. This may be performed as an administrative check, by examining logs or other
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                                                                                                                                                        B 3.1-4                                                                                                                                                                                                                                                              Revision No. 0 SDM B 3.1.1
 
BASES
 
ACTIONS                                                                                                                                                                          E.1, E.2, E.3, E.4, and E.5(continued)
 
information, to determine if the components are out of service for maintenance or other reasons. It is not necessary to perform the SRs needed to demonstrate the OPERABILITY of the components. If, however,  any required component is inoperable, then it must be restored to OPERABLE status. In this case, SRs may need to be performed to restore the component to OPERABLE status. Action must continue until all required components are OPERABLE.
 
SURVEILLANCE                                  SR3.1.1.1 REQUIREMENTS Adequate SDM must be verified to ensure that the reactor can be made subcritical from any initial operating condition.
This can be accomplished by a test, an evaluation, or a combination of the two. Adequate SDM is demonstrated before or during the first startup after fuel movement or shuffling within the reactor pressure vessel, or control rod replacement. Control rod replacement refers to the decoupling and removal of a control rod from a core location, and subsequent replacement with a new control rod or a control rod from another core location. Since core reactivity will vary during the cycle as a function of fuel depletion and poison burnup, the beginning of cycle (BOC) test must also account for changes in core reactivity during the cycle. Therefore, to obtain the SDM, the initial measured value must be increased by an adder, "R", which is the difference between the calculated value of maximum core reactivity                                                                                                    during the operating cycle and the calculated BOC core reactivity. If the value of                                                                                                                                                                                                                                                                                                                                        R is negative (that is, BOC is the most reactive point in the cycle), no correction to the BOC measured value is required (Ref. 3                                                                                                                                                                                                                                                                                                                                                                                                                                            ). For the SDM demonstrations that rely solely on calculation of the highest worth control rod, additional margin (0.10%                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                    k/k) must be added to the SDM limit of 0.28%                                                                                                                                                                                                                                                                                                                                                                                                                                                                                    k/k to account for uncertainties in the calculation.
 
The SDM may be demonstrated during an in sequence control rod                              withdrawal,                                                                                                              in which the highest worth control rod is analytically determined, or during local criticals, where the                              highest worth control rod is determined by testing.
Local critical tests require the withdrawal of out of
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                                                                                                                                                        B 3.1-5                                                                                                                                                                                                                                                    Revision No. 72 SDM B 3.1.1
 
BASES
 
SURVEILLANCE                                                                                                                                                                          SR  3.1.1.1(continued)
REQUIREMENTS sequence control rods.This testing would therefore require bypassing of the Rod Worth Minimizer to allow the out of sequence withdrawal, and therefore additional requirements must be met (see LCO                                                                                                                                                                                                        3.10.7, "Control Rod Testing Operating").
 
The Frequency of 4                                                                                                                                                                                    hours after reaching criticality is allowed to provide a reasonable amount of time to perform the required calculations and have appropriate verification.
 
During MODES 3 and 4, analytical calculation of SDM may be used to assure the requirements of SR 3.1.1.1 are met.
During MODE                                                                                                              5, adequate SDM is required to ensure that the reactor does not reach criticality during control rod withdrawals. An evaluation of each in vessel fuel movement during fuel loading (including shuffling fuel within the core) is required to ensure adequate SDM is maintained during refueling. This evaluation ensures that the intermediate loading patterns are bounded by the safety analyses for the final core loading pattern. For example, bounding analyses that demonstrate adequate SDM for the most reactive configurations during the refueling may be performed to demonstrate acceptability of the entire fuel movement sequence. These bounding analyses include additional margins to the associated uncertainties. Spiral offload/reload sequences, including modified quadrant spiral offload/reload sequences, inherently satisfy the SR, provided the fuel assemblies are reloaded in the same configuration analyzed for the new cycle. Removing fuel from the core will always result in an increase in SDM.
 
REFERENCES                                                                                                                                                                          1.                                        UFSAR, Sections 1.5.1.8 and 1.5.2.2.7.
: 2.                                        UFSAR, Section                                                                                                                                            14.6.2.
: 3.                                        NEDE-                                        24011-P-A                                                , "General Electric Standard Application for Reactor Fuel," latest approved revision.
: 4.                                        UFSAR, Section                                                                                                                                            14.5.3.3.
: 5.                                        UFSAR, Section                                                                                                                                            14.5.3.4.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                                                                                                                                                        B 3.1-6                                                                                                                                                                                                                                                    Revision No. 72 SDM B 3.1.1
 
BASES
 
REFERENCES                                                                                                                                                                          6.                                        UFSAR, Section                                                                                                                                            3.6.5.4.
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                                                                                                                                                        B 3.1-7                                                                                                                                                                                                                                                    Revision No. 72 Reactivity Anomalies B 3.1.2
 
B 3.1  REACTIVITY CONTROL SYSTEMS
 
B 3.1.2  Reactivity Anomalies
 
BASES
 
BACKGROUND                                                                                                                                                                          In accordance with the UFSAR (Ref.                                                                                                                                                                                                                                                                                                                                                  1), reactivity shall be controllable such that subcriticality is maintained under cold conditions and acceptable fuel design limits are not exceeded during normal operation and abnormal operational transients. Therefore, reactivity anomaly is used as a measure of the predicted versus measured (i.e., monitored) core reactivity during power operation.                                                                                                                                                                                                                                                                                                              A large reactivity anomaly could be the result of unanticipated changes in fuel reactivity or control rod worth or operation at conditions not consistent with those assumed in the predictions of core reactivity, and could potentially result in a loss of SDM or violation of acceptable fuel design limits. Comparing predicted versus measured core r                                                                                                                                                                                                                                                                                                                      eactivity                                                                                          supports the SDM demonstrations (LCO                                                                                                                                                                                              3.1.1, "SHUTDOWN MARGIN (SDM)") in assuring the reactor can be brought safely to cold, subcritical conditions.
 
When the reactor core is critical or in normal power operation, a reactivity balance exists and the net reactivity is zero. A comparison of predicted and measured reactivity is convenient under such a balance, since parameters                                                                                                    are being maintained relatively stable under steady state power conditions. The positive reactivity inherent in the core design is balanced by the negative reactivity of the control components, thermal feedback, neutron leakage, and materials in the core                                                                                                                                                                                                                                                                                                                                                                                                                                  that absorb neutrons, such as burnable absorbers, producing zero net reactivity.
 
In order to achieve the required fuel cycle energy output, the uranium enrichment in the new fuel loading and the fuel loaded in the previous cycles provide excess positive reactivity beyond that required to sustain steady state operation at the beginning of cycle (BOC). When the reactor is critical at RTP and operating moderator temperature, the excess positive reactivity is compensated by burnable absorbers (e.g., gadolinia), control rods, and whatever neutron poisons (mainly xenon and samarium) are present in the fuel. The predicted core reactivity, as represented by
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                                                                                                                                                        B 3.1-8                                                                                                                                                                                                                                          Revision No. 113 Reactivity Anomalies B 3.1.2 BASES
 
BACKGROUND                                                                                                                                                                          core k                                        effective  (k eff ), is calculated by a 3D                                                                                                                                                                                                                  core simulator (continued)                                                                                                                                                                          code as a function of cycle exposure. This calculation is performed for projected operating states and conditions throughout the cycle. The monitored core k                                                                                                                                                                                                                                                                                                                      eff  is calculated by the core monitoring system for actual plant conditions and is then compared to the predicted value for the cycle exposure.
 
APPLICABLE                                                                                                                                                                          Accurate prediction of core reactivity is either an explicit SAFETY ANALYSES                                                                                                                                                                          or implicit assumption in the accident analysis evaluations (Ref.                                                  2). In particular, SDM and reactivity transients, such as control rod withdrawal accidents or rod drop accidents, are very sensitive to accurate prediction of core reactivity. These accident analysis evaluations rely on computer codes that have been qualified against available test data, operating plant data, and analytical benchmarks.
Monitoring reactivity anomaly provides additional assurance that the nuclear methods provide an accurate representation of the core reactivity.
 
The comparison between measured and predicted initial core reactivity provides a normalization for the calculational models used to predict core reactivity. If the measured and predicted core k                                        eff(s)  for identical core conditions at BOC do not reasonably agree, then the assumptions used                                                                                                                                                                                                                                                                                                                                                                                                                                                                                    in the reload cycle design analysis or the calculation models used to predict core k eff  may not be accurate. If reasonable agreement between measured and predicted core reactivity exists at BOC, then the prediction may be normalized to the measured value. Thereafter, any significant deviations in the measured core k                                                                                                                                                                          eff  from the predicted core k                                                                                                                                                                                                                                      eff  that develop during fuel depletion may be an indication that the assumptions of the DBA and transient analyses are no longer valid, or that an unexpected change in core conditions has occurred.
 
Reactivity anomalies satisfy Criterion                                                                                                                                                                                                                                                                                                                                                                                          2 of the NRC Policy Statement.
 
LCO                                                                                                                                                                          Large differences between monitored and                                                                                                                                                                                                                                                                                                                                                                                                    predicted core reactivity may indicate that the assumptions of the DBA and transient analyses are no longer valid, or that the
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                                                                                                                                                        B 3.1-9                                                                                                                                                                                                                                          Revision No. 113 Reactivity Anomalies B 3.1.2
 
BASES
 
LCO                                                                                                                                                                          uncertainties in the "Nuclear Design Methodology" are larger (continued)                                                                                                                                                                          than expected. A limit on the difference between the monitored and the predicted core                                                                                                                                                                                                                                                                                        k eff  of +/-                                        1%                        k/k has been established based on engineering judgment. A >                                                                                                                                                                                                                                                                                                                                                                                                                                                                                    1% deviation in reactivity from that predicted is larger than expected for normal operation and should therefore be evaluated. A deviation as large as 1% would not                                                                                                                                                                                                                                                                                                                                                  exceed the design conditions of the reactor and is on the safe side of the postulated transients.
 
APPLICABILITY                                                                                                                                                                          In MODE                                                                      1, most of the control rods are withdrawn and steady state operation is typically achieved. Under these conditions, the comparison between predicted and monitored core reactivity provides an effective measure of the reactivity anomaly. In MODE                                                                                                                                                                                                                                                                                        2, control rods are typically being withdrawn during a startup.                                                                                                                                                                                                                                                                                                                                                    In MODES                                                                                3 and                                                  4, all control rods are fully inserted and therefore the reactor is in the least reactive state, where monitoring core reactivity is not necessary. In MODE                                                                                                                                                                                                                                                                                                                                                                                5, fuel loading results in a continually changing core reactivity. SDM requirements (LCO                                                                                                                                                                          3.1.1) ensure that fuel movements are performed within the bounds of the safety analysis, and an SDM demonstration is required during the first startup following operations that could have altered core reactivity (e.g., fuel movement, control rod replacement, shuffling).
The SDM test, required by LCO                                                                                                                                                                                                                                                                                                  3.1.1, provides a direct comparison of the predicted and monitored core reactivity at cold conditions; therefore, reactivity anomaly is not required during these conditions.
 
ACTIONS                                        A.1
 
Should an anomaly develop between measured and predicted core reactivity, the core reactivity difference must be restored to within the limit to ensure continued operation is within the core design assumptions. Restoration to within the limit could be performed by an evaluation of the core design and safety analysis to determine the reason for the anomaly. This evaluation normally reviews the core conditions to determine their consistency with input to design calculations. Measured core and process parameters are also normally evaluated to determine that they are within the bounds of the safety analysis, and safety analysis calculational models may be reviewed to verify that they are adequate for representation of the core conditions.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                                                                                                                                                  B 3.1-                                                  10                                                                                                                                                                                                                  Revision No. 94 Reactivity Anomalies B 3.1.2
 
BASES
 
ACTIONS                                                                                                                                                                          A.1(continued)
 
The required Completion Time of 72                                                                                                                                                                                                                                                                                                                                                  hours is based on the low probability of a DBA occurring during this period, and allows sufficient time to assess the physical condition of the reactor and complete the evaluation of the core design and safety analysis.
 
B.1
 
If the core reactivity cannot be restored to within the 1%                        k/k limit, 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                                                                                                              hours. The allowed Completion Time of 12                                                                                                                                                                                                                                                                                                  hours is reasonable, based on operating experience, to reach MODE                                                                                                                                                                                                                                                                                                                                                            3 from full power conditions in an orderly manner and without challenging plant                                                  systems.
 
SURVEILLANCE                                          SR3.1.2.1 REQUIREMENTS The core monitoring system calculates the core                                                                                                                                                                                                                                                                                                                                                                                                                                    k eff  for the reactor conditions obtained from plant instrumentation.                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                      A comparison of the monitored core k                                                                                                                                                                                                                                                                                                                                eff  to the predicted core k eff  at the same cycle exposure is used to calculate the reactivity difference. The comparison is required when the core reactivity has potentially changed by a significant amount. This may occur following a refueling in which                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                            new fuel assemblies are loaded, fuel assemblies are shuffled within the core, or control rods are replaced or shuffled.
Control rod replacement refers to the decoupling and removal of a control rod from a core location, and subsequent replacement with a new control rod or a control rod from another core location. Also, core reactivity changes during the cycle. The 24                                                                                                                                                                                    hour interval after reaching equilibrium conditions following a startup is based on the need for equilibrium xenon concentrations in the core, such that an accurate comparison between the monitored and predicted core k eff  can be made. For the purposes of this SR, the reactor is assumed to be at equilibrium conditions when steady state operations (no control rod movement or core
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                                                                                                                                                  B 3.1-                                                  11                                                                                                                                                                                                        Revision No. 113 Reactivity Anomalies B 3.1.2
 
BASES
 
SURVEILLANCE                                                                                                                                                                          SR3                                        .1.2.1(continued)
REQUIREMENTS flow changes) at                                                                                                                                                                            75%                              RTP have been obtained. The 1000                                        MWD/T Frequency was developed, considering the relatively slow change in core reactivity with exposure and operating experience related to variations in core reactivity. The comparison requires the core to be operating at power levels which minimize the uncertainties and measurement errors, in order to obtain meaningful results. Therefore, the comparison is only done when in MODE 1.
 
REFERENCES                                                                                                                                                                          1.                                        UFSAR, Section 1.5.
: 2.                                        UFSAR, Chapter                                                                                                                                            14.
 
PBAPS UNIT 2                                                                                                                                                                                                                                                                                                                                                  B 3.1-                                                  12                                                                                                                                                                                                                            Revision No. 0 Control Rod OPERABILITY B 3.1.3
 
B 3.1  REACTIVITY CONTROL SYSTEMS
 
B 3.1.3  Control Rod OPERABILITY
 
BASES
 
BACKGROUND                                                                                                                                                                          Control rods are components of the Control Rod Drive (CRD)
System, which is the primary reactivity control system for the reactor. In conjunction with the Reactor Protection System, the CRD System provides the means for the reliable control of reactivity changes to ensure under conditions of normal operation, including abnormal operational transients, that specified acceptable fuel design limits are not exceeded. In addition, the control rods provide the capability to hold the reactor core subcritical under all conditions and to limit the potential amount and rate of reactivity increase caused by a malfunction in the CRD System. The CRD System is designed to satisfy the requirements specified in Reference 1.
 
The CRD System consists of 185                                                                                                                                                                                                                                                                                                            locking piston control rod drive mechanisms (CRDMs) and a hydraulic control unit for each drive mechanism. The locking piston type CRDM is a double acting hydraulic piston, which uses condensate water as the operating fluid. Accumulators provide additional energy for scram. An index tube and piston, coupled to the control rod, are locked at fixed increments by a collet mechanism. The collet fingers engage notches in the index tube to prevent unintentional withdrawal of the control rod, but without restricting insertion.
 
This Specification, along with LCO                                                                                                                                                                                                                                                                                                                                                    3.1.4, "Control Rod Scram Times," and LCO                                                                                                                                                      3.1.5, "Control Rod Scram Accumulators,"
ensure that the performance of the control rods in the event of a Design Basis Accident (DBA) or transient meets the assumptions used in the safety analyses of References                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                2, 3, and                              4.
 
APPLICABLE                                                                                                                                                                          The analytical methods and assumptions used in the SAFETY ANALYSES                                                                                                                                                                          evaluations involving control rods are presented in References                                                                                                    2, 3, and                                                                                          4. The control rods                                                                                                                                                                                                        provide the primary means for rapid reactivity control (reactor scram),
for maintaining the reactor subcritical and for limiting the potential effects of reactivity insertion events caused by malfunctions in the CRD System.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                                                                                                                                                  B 3.1-                                                  13                                                                                                                                                                                                                            Revision No. 0 Control Rod OPERABILITY B 3.1.3
 
BASES
 
APPLICABLE                                                                                                                                                                          The capability to insert the control rods provides assurance SAFETY ANALYSES                                                                                                                                                                          that the assumptions for scram reactivity in the DBA and (continued)                                                                                                                                                                          transient analyses are not violated. Since the SDM ensures the reactor will be subcritical with the highest worth control rod withdrawn                                                                                                                                                                                                                  (assumed single failure), the additional failure of a second control rod to insert, if required, could invalidate the demonstrated SDM and potentially limit the ability of the CRD System to hold the reactor subcritical. If the control rod is stuck at an inserted position and becomes decoupled from the CRD, a control rod drop accident (CRDA) can possibly occur.
Therefore, the requirement that all control rods be OPERABLE ensures the CRD System can perform its intended function.
 
The control rods also protect the fuel from damage which could result in release of radioactivity. The limits protected are the MCPR Safety Limit (SL) (see Bases for SL                    2.1.1, "Reactor Core SLs" and LCO                                                                                                                                                                                                                                                                                                                                        3.2.2, "MINIMUM CRITICAL POWER RATIO (MCPR)"), the 1% cladding plastic strain                                                            fuel design limit (see Bases for LCO                                                                                                                                                                                                                                                                                                                                                                      3.2.3, "LINEAR HEAT GENERATION RATE (LHGR)"), and the fuel damage limit (see Bases for LCO                                                                                                                                                                                    3.1.6, "Rod Pattern Control") during reactivity insertion                                                                                                                                                                                                        events.
 
The negative reactivity insertion (scram) provided by the CRD System provides the analytical basis for determination of plant thermal limits and provides protection against fuel damage limits during a CRDA. The Bases for LCO                                                                                                                                                                                                                                                                                                                                                                                                                                                                                      3.1.4, LCO                              3.1.5, and LCO                                                                                                                                            3.1.6 discuss in more detail how the SLs are protected by the CRD System.
 
Control rod OPERABILITY satisfies Criterion                                                                                                                                                                                                                                                                                                                                                                                                                                            3 of the NRC Policy Statement.
 
LCO                                                                                                                                                                          The OPERABILITY of an individual control rod is based on a combination of factors, primarily, the scram insertion times, the control rod coupling integrity, and the ability to determine the control rod position. Accumulator OPERABILITY is addressed by LCO                                                                                                                                                                                                                                                                                                                      3.1.5. The associated scram accumulator status for a control rod only affects the scram insertion times; therefore, an inoperable accumulator does not immediately require declaring a control rod inoperable.
Although not all control rods are required to be OPERABLE to satisfy the intended reactivity control requirements, strict
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                                                                                                                                                  B 3.1-                                                  14                                                                                                                                                                                                                  Revision No. 49 Control Rod OPERABILITY B 3.1.3
 
BASES
 
LCO                                                                                                                                                                          control over the number and distribution of inoperable (continued)                                                                                                                                                                          control rods is required to satisfy the assumptions of the DBA and transient analyses.
 
APPLICABILITY                                                                                                                                                                          In MODES                                                                                1 and                                                  2, the control rods are assumed to function during a DBA or transient and are therefore required to be OPERABLE in these MODES. In MODES                                                                                                                                                                                                                                                                                                                                                  3 and                                                  4, control rods are not able to be withdrawn since the reactor mode switch is in shutdown and a control rod block is applied. This provides adequate requirements for control rod OPERABILITY during these conditions. Control rod requirements in MODE                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                            5 are located in LCO                                                                                                                                            3.9.5, "Control Rod OPERABILITY                                                                                                                                                                                                                                                                                                                  Refueling."
 
ACTIONS                                                                                                                                                                          The ACTIONS Table is modified by a Note indicating that a separate Condition entry is allowed for each control rod.
This is acceptable, since the Required Actions for each Condition provide appropriate compensatory actions for each inoperable control rod. Complying with the Required Actions may allow for continued operation, and subsequent inoperable control rods are governed by subsequent Condition entry and application of associated Required Actions.
 
A.1, A.2, A.3, and A.4
 
A control rod is considered stuck if it will not insert by either CRD drive water or scram pressure (i.e., the control rod cannot be inserted by CRD drive water and cannot be inserted by scram pressure.)With a fully inserted control rod stuck, only those actions specified in Condition C                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                            are required as long as the control rod remains fully inserted.
The Required Actions are modified by a Note, which allows the rod worth minimizer (RWM) to be bypassed if required to allow continued operation. LCO                                                                                                                                                                                                                                                                                                                      3.3.2.1, "Control Rod Block Instrumentation," provides additional requirements when the RWM is bypassed to ensure compliance with the CRDA analysis.
With                                        one withdrawn control rod stuck, the local scram reactivity rate assumptions may not be met if the stuck control rod separation criteria are not met. Therefore, a verification that the separation criteria are met must be performed immediately. The separation criteria are not met if a) the stuck control rod occupies a location adjacent to two "slow" control rods, b) the stuck control rod occupies a location adjacent to one "slow" control rod, and the one "slow" control rod is also adjacent to another "slow" control rod, or c) if the stuck control rod occupies a
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                                                                                                                                                  B 3.1-                                                  15                                                                                                                                                                                                                            Revision No. 2 Control Rod OPERABILITY B 3.1.3
 
BASES
 
ACTIONS                                                                                                                                                                          A.1, A.2, A.3, and A.4(continued)
 
location adjacent to one "slow" control rod when there is another pair of "slow" control rods adjacent to one another.
The description of "slow" control rods is provided in LCO                              3.1.4, "Control Rod Scram Times."  In addition, the associated control rod drive must be disarmed in 2                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                  hours.
The allowed Completion Time of 2                                                                                                                                                                                                                                                                                                                              hours is acceptable, considering the reactor can still be shut down, assuming no additional control rods fail to insert, and provides a reasonable time to perform the Required Action in an orderly manner. The control rod must be isolated from both scram and normal insert and withdraw pressure. Isolating the control rod from scram and normal insert and withdraw pressure prevents damage to the CRDM. The control rod should be isolated from scram and normal insert and withdraw pressure, while maintaining cooling water to the CRD.
 
Monitoring of the insertion capability of each withdrawn control rod must                                                                                                                                                                also be performed within 24                                                                                                                                                                                                                                                                              hours from discovery of Condition A concurrent with THERMAL POWER greater than the low power setpoint (LPSP) of the RWM.
SR                    3.1.3.3 performs                                                                                                                                                      periodic tests of the control rod insertion capability of withdrawn control rods. Testing each withdrawn control rod ensures that a generic problem does not exist. This Completion Time also allows for an exception to the normal "time zero" for beginning the allowed outage time "clock."  The Required Action A.3 Completion Time only begins upon discovery of Condition A concurrent with THERMAL POWER greater than the actual LPSP of the RWM, since the notch insertions may not be compatible with the requirements of rod pattern control (LCO                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                        3.1.6) and the RWM (LCO                                                                                                                        3.3.2.1). The allowed Completion Time of 24                    hours from discovery of Condition A concurrent with THERMAL POWER greater than the LPSP of the RWM provides a reasonable time to test the control rods, considering the potential for a need to reduce power to perform the tests.
 
To allow continued operation with a withdrawn control rod stuck, an evaluation of adequate SDM is also required within 72                    hours. Should a DBA or transient require a shutdown, to preserve the single failure criterion, an additional control rod would have to be assumed to fail to insert when required. Therefore, the original SDM demonstration may not be valid. The SDM must therefore be evaluated (by measurement or analysis) with the stuck control rod at its
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                                                                                                                                                  B 3.1-                                                  16                                                                                                                                                                                                                  Revision No. 79 Control Rod OPERABILITY B 3.1.3
 
BASES
 
ACTIONS                                                                                                                                                                          A.1, A.2, A.3, and A.4(continued)
 
stuck position and the highest worth OPERABLE control rod assumed to be fully withdrawn.
 
The allowed Completion Time of 72                                                                                                                                                                                                                                                                                                                                        hours to verify SDM is adequate, considering that with a single control rod stuck in a withdrawn position, the remaining OPERABLE control rods are capable of providing the required scram and shutdown reactivity. Failure to reach MODE                                                                                                                                                                                                                                                                                                                                                                                                                                            4 is only likely if an additional control rod adjacent to the stuck control rod also fails to insert during a required scram.
Even with the postulated additional single failure of an adjacent control rod to insert, sufficient reactivity control remains to reach and maintain MODE                                                                                                                                                                                                                                                                                                                                                                                                                                  3 conditions (Ref. 5                                                  and 6).
 
B.1
 
With two or more withdrawn control rods stuck, the plant must be brought to MODE                                                                                                                                                                                                                                      3 within 12                                                                                                              hours. The occurrence of more than one control rod stuck at a withdrawn position increases the probability that the reactor cannot be shut down if required. Insertion of all insertable control rods eliminates the possibility of an additional failure of a control rod to insert. The allowed Completion Time of 12                    hours is reasonable, based on operating experience, to reach MODE                                                                                                    3 from full power conditions in an orderly manner and without challenging plant systems.
 
C.1 and C.2
 
With one or more control rods inoperable for reasons other than being stuck in the withdrawn position, (including a control rod which is stuck in the fully inserted position) operation may continue, provided the control rods are fully inserted within 3                                                                                                                                                                          hours and disarmed (electrically or hydraulically) within 4                                                                                                                                                                                                                                      hours. Inserting a control rod ensures the shutdown and scram capabilities are not adversely affected. The control rod is disarmed to prevent inadvertent withdrawal during subsequent operations. The control rods can be hydraulically disarmed by closing the drive water and exhaust water isolation valves. The control rods can be electrically disarmed by disconnecting power from all four directional control valve solenoids. Required Action                                                            C.1 is modified by a Note, which allows the RWM to be bypassed if required to allow insertion of the inoperable
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                                                                                                                                                  B 3.1-                                                  17                                                                                                                                                                                                                  Revision No. 63 Control Rod OPERABILITY B 3.1.3
 
BASES
 
ACTIONS                                                                                                                                                                          C.1 and C.2(continued)
 
control rods and continued operation. LCO                                                                                                                                                                                                                                                                                                                                                                                                                                  3.3.2.1 provides additional requirements when the RWM is bypassed to ensure compliance with the CRDA analysis.
 
The allowed Completion Times are reasonable, considering the small number of allowed inoperable control rods, and provide time to insert and disarm the control rods in an orderly manner and without challenging plant systems.
 
D.1 and D.2
 
Out of sequence control rods may increase the potential reactivity worth of a dropped control rod during a CRDA. At 10%                              RTP, the analyzed rod position sequence                                                                                                                                                                                                                                                                                                                                                    (Ref. 5                                                  and 6) requires inserted control rods not in compliance with the analyzed rod position sequence to be separated by at least two OPERABLE control rods in all directions, including the diagonal. Therefore, if two or more inoperable control rods are not in compliance with the analyzed rod position sequence                                                                                and not separated by at least two OPERABLE control rods, action must be taken to restore compliance with the analyzed rod position sequence or restore the control rods to OPERABLE status. Condition                                                                                                                                                                                                                                                                                                            D is modified by a Note indicating that the Condition is not applicable when
                                                  > 10%                              RTP, since the analyzed rod position sequence                                                                                                                                                                                                                                                                                                                                                                                                                                                                is not required to be followed under these conditions, as described in the Bases for LCO                                                                                                                                                                                                        3.1.6. The allowed Completion Time of 4 hours is acceptable, considering the low probability of a CRDA occurring.
 
E.1
 
If any Required Action and associated Completion Time of Condition                                                                                          A, C, or                                                                                D are not met, or there are nine or more inoperable control rods, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to MODE                                                                                                                                                                                                                                                                                                                                        3 within 12                                                                                                              hours. This ensures all insertable control rods are inserted and places the reactor in a condition that does not require the active function (i.e., scram) of the control rods. The number of control rods permitted to be inoperable when operating above 10%                              RTP (e.g., no CRDA considerations) could be more than the value specified, but the occurrence of a large number of
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                                                                                                                                                  B 3.1-                                                  18                                                                                                                                                                                                                  Revision No. 63 Control Rod OPERABILITY B 3.1.3
 
BASES
 
ACTIONS                                                                                                                                                                          E.1(continued)
 
inoperable control rods could be indicative                                                                                                                                                                                                                                                                                                                                                                                                                                            of a generic problem, and investigation and resolution of the potential problem should be undertaken. The allowed Completion Time of 12                                                  hours is reasonable, based on operating experience, to reach MODE                                                                                                    3 from full power in an orderly manner and without challenging plant systems.
 
SURVEILLANCE                                    SR3.1.3.1 REQUIREMENTS The position of each control rod must be determined to ensure adequate information on control rod position is available to the operator for determining control rod OPERABILITY and controlling rod patterns. Control rod position may be determined by the use of OPERABLE position indicators, by moving control rods to a position with an OPERABLE indicator, or by the use of other appropriate methods. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
 
SR3.1.3.2 DELETED
 
SR3.1.3.3
 
Control rod insertion capability is demonstrated by inserting each partially or fully withdrawn control rod at least one notch and observing that the control rod moves.
The control rod may then be returned to its original position. This ensures the control rod is not stuck and is free to insert on a scram signal. This Surveillance is not required when THERMAL POWER is less than or equal to the actual LPSP of the RWM, since the notch insertions may not be compatible with the requirements of the analyzed rod position sequence (LCO                                                                                                                                                                                                                            3.1.6) and the RWM (LCO                                                                                                                                                                                                                                      3.3.2.1).
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.                                                                                                                                                                                                                                                                                                                                                                                                                At any time, if a control rod is immovable, a
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                                                                                                                                                  B 3.1-                                                  19                                                                                                                                                                                                                  Revision No. 86 Control Rod OPERABILITY B 3.1.3
 
BASES
 
SURVEILLANCE                                                                                                                                                                          SR3.1.3.3(continued)
REQUIREMENTS determination of that control rod's trippability (OPERABILITY) must be made and appropriate action taken.
For example, the unavailability of the Reactor Manual Control System does not affect the OPERABILITY of the control rods, provided SR 3.1.3.3 is                                                                                                                                                                                                                                                                                                                                                  current in accordance with SR                                                                      3.0.2.
 
SR  3.1.3.4
 
Verifying that the scram time for each control rod to notch position                                                                                06 is                                                              7 seconds provides reasonable assurance that the control rod will insert when required during a DBA or transient, thereby completing its shutdown function.
This SR is performed in conjunction with the control rod scram time testing of SR                                                                                                                                                                                                                                                3.1.4.1, SR                                                                                                              3.1.4.2, SR                                                                                                              3.1.4.3, and SR                                                            3.1.4.4. The LOGIC SYSTEM FUNCTIONAL TEST in LCO                              3.3.1.1, "Reactor Protection System (RPS)
Instrumentation," and the functional                                                                                                                                                                                                                                                                                                                                                                      testing of SDV vent and drain valves in LCO                                                                                                                                                                                              3.1.8, "Scram Discharge Volume (SDV)
Vent and Drain Valves," overlap this Surveillance to provide complete testing of the assumed safety function. The associated Frequencies are acceptable, considering the more frequent testing performed to demonstrate other aspects of control rod OPERABILITY and operating experience, which shows scram times do not significantly change over an operating cycle.
 
SR3.1.3.5
 
Coupling verification is performed to ensure the control rod is connected to the CRDM and will perform its intended function when necessary. The Surveillance requires verifying a control rod does not go to the withdrawn overtravel position. The overtravel position feature provides a positive check on the coupling integrity since only an uncoupled CRD can reach the overtravel position.
The verification is required to be performed any time a control rod is withdrawn to the "full out" position (notch position                                                                                48) or prior to declaring the control rod OPERABLE after work on the control rod or CRD System that could affect coupling (CRD changeout and blade replacement or complete cell disassembly, i.e., guide tube removal). This includes control rods inserted one notch and then returned
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                                                                                                                                                  B 3.1-                                                  20                                                                                                                                                                                                                  Revision No. 79 Control Rod OPERABILITY B 3.1.3
 
BASES
 
SURVEILLANCE                                                                                                                                                                          SR3.1.3.5(continued)
REQUIREMENTS to the "full out" position during the performance of SR                    3.1.3.2. This Frequency is acceptable, considering the low probability that a control rod will become uncoupled when it is not being moved and operating experience related to uncoupling events.
 
REFERENCES                                                                                                                                                                          1.                                        UFSAR, Sections 1.5.1.1 and 1.5.2.2.
: 2.                                        UFSAR, Section                                                                                                                                            14.6.2.
: 3.                                        UFSAR, Appendix K, Section VI.
: 4.                                        UFSAR, Chapter 14.
: 5.                                        NEDO-                                      21231, "Banked Position Withdrawal Sequence,"
Section                                                                      7.2, January                                                                                                                        1977.
: 6.                                        NEDE-                                        24011-P-                                                A, General Electric Standard Application for Reactor Fuel, latest approved revision.
 
PBAPS UNIT 2                                                                                                                                                                                                                                                                                                                                                  B 3.1-                                                  21                                                                                                                                                                                                                  Revision No. 63 Control Rod Scram Times B 3.1.4
 
B 3.1  REACTIVITY CONTROL SYSTEMS
 
B 3.1.4  Control Rod Scram Times
 
BASES
 
BACKGROUND                                                                                                                                                                          The scram function of the Control Rod Drive (CRD) System controls reactivity changes during abnormal operational transients to ensure that specified acceptable fuel design limits are not exceeded (Ref.                                                                                                                                                                                                                                                                                                  1). The control rods are scrammed by positive means using hydraulic pressure exerted on the CRD piston.
 
When a scram signal is initiated, control air is vented from the scram valves, allowing them to open by spring action.
Opening the exhaust valve reduces the pressure above the main drive piston to atmospheric pressure, and opening the inlet valve applies the accumulator or reactor pressure to the bottom of the piston. Since the notches in the index tube are tapered on the lower edge, the collet fingers are forced open by cam action, allowing the index tube to move upward without restriction because of the high differential pressure across the piston. As the drive moves upward and the accumulator pressure reduces below the reactor pressure, a ball check valve opens, letting the reactor pressure complete the scram action. If the reactor pressure is low, such as during startup, the accumulator will fully insert the control rod in the required time without assistance from reactor pressure.
 
APPLICABLE                                                                                                                                                                          The analytical methods and assumptions used in evaluating SAFETY ANALYSES                                                                                                                                                                          the control rod scram function are presented in References                                                                                                    2, 3, and                                                                                          4. The Design Basis Accident (DBA) and transient analyses assume that all of the control rods scram at a specified insertion rate. The resulting negative scram reactivity forms the basis for the determination of plant thermal limits (e.g., the MCPR). Other distributions of scram times (e.g., several control rods scramming slower than the                                                                                average time with several control rods scramming faster than the average time) can also provide sufficient scram reactivity. Surveillance of each individual control rod's scram time ensures the scram reactivity assumed in the DBA and transient analyses can be met.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                                                                                                                                                  B 3.1-                                                  22                                                                                                                                                                                                                            Revision No. 0 Control Rod Scram Times B 3.1.4
 
BASES
 
APPLICABLE                                                                                                                                                                          The scram function of the CRD System protects the MCPR SAFETY ANALYSES                                                                                                                                                                          Safety Limit (SL) (see Bases for SL 2.1.1, "Reactor Core (continued)                                                                                                                                                                            SLs" and LCO                                                                                                                        3.2.2, "MINIMUM CRITICAL POWER RATIO (MCPR)")
and the 1% cladding plastic strain fuel design limit (see Bases for LCO                                                                                                                                  3.2.3                                      , "LINEAR HEAT GENERATION RATE (LHGR)"),
which ensure that no fuel damage will occur if these limits are not exceeded. Above 800                                                                                                                                                                                                                                                                                        psig, the scram function is designed to insert negative reactivity at a                                                                                                                                                                                                                                                                                                                                                                                                                                            rate fast enough to prevent the actual MCPR from becoming less than the MCPR SL, during the analyzed limiting power transient. Below 800                              psig, the scram function is assumed to perform during the control rod drop accident (Ref.                                                                                                                                                                                                                                                                                                                                                            5) and, therefore, also provides protection against violating fuel damage limits during reactivity insertion accidents (see Bases for LCO                              3.1.6, "Rod Pattern Control"). For the reactor vessel overpressure protection analysis, the scram function, along with the safety/relief valves,                                                                                                                                                                                                                                                                                                            ensure that the peak vessel pressure is maintained within the applicable ASME Code limits.
 
Control rod scram times satisfy Criterion                                                                                                                                                                                                                                                                                                                                                                                                                        3 of the NRC Policy Statement.
 
LCO                                                                                                                                                                          The scram times specified in Table                                                                                                                                                                                                                                                                                                            3.1.4-                                                1 (in the accompanying LCO) are required to ensure that the scram reactivity assumed in the DBA and transient analysis is met (Ref. 6).
 
To account for single failures and "slow" scramming control rods, the scram times specified in Table                                                                                                                                                                                                                                                                                                                                                                                                                3.1.4-                                                1 are faster than those assumed in the design basis analysis. The scram times have a margin that allows up to approximately 7% of the control rods (e.g., 185                                                                                                                                                                                                                                                                              x 7%                                                    13) to have scram times exceeding the specified limits (i.e., "slow" control rods) assuming a single stuck control rod (as allowed by LCO                              3.1.3, "Control Rod OPERABILITY") and an additional control rod failing to scram per the single failure criterion. The scram times are specified as a function of reactor steam dome pressure to account for the pressure dependence of the scram times. The scram times are specified relative to measurements based on reed switch positions, which provide the control rod position indication. The reed switch closes ("pickup") when the
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                                                                                                                                                  B 3.1-                                                  23                                                                                                                                                                                                                  Revision No. 49 Control Rod Scram Times B 3.1.4
 
BASES
 
LCO                                                                                                                                                                          index tube passes a specific location and then opens (continued)                                                                                                                                                                          ("dropout") as the index tube travels upward. Verification of the specified scram times in Table                                                                                                                                                                                                                                                                                                                                                                                  3.1.4-                                                1 is accomplished                                                                                                                        through measurement of the "dropout" times.
 
To ensure that local scram reactivity rates are maintained within acceptable limits, no more than two of the allowed "slow" control rods may occupy adjacent locations.
 
Table                                                  3.1.4-                                                1 is modified by two Notes, which state that control rods with scram times not within the limits of the table are considered "slow" and that control rods with scram times > 7                                                                      seconds are considered inoperable as required by SR                    3.1.3.4.
 
This LCO applies only to OPERABLE control rods since inoperable control rods will be inserted and disarmed (LCO                                        3.1.3). Slow scramming control rods may be conservatively declared inoperable and not accounted for as "slow" control rods.
 
APPLICABILITY                                                                                                                                                                          In MODES                                                                                1 and                                                  2, a scram is assumed to function during transients and accidents analyzed for these plant conditions. These events are assumed to occur during startup and power operation; therefore, the scram function of the control rods is required during these MODES. In MODES                                                  3 and                                                  4, the control rods are not able to be withdrawn since the reactor mode switch is in shutdown and a control rod block is applied. This provides adequate requirements for control rod scram capability during these conditions.
Scram requirements in MODE                                                                                                                                                                                                                                                                    5 are contained in LCO                                                                                                                                                                                                                            3.9.5, "Control Rod OPERABILITY      Refueling."
 
ACTIONS                                            A.1
 
When the requirements of this LCO are not met, the rate of negative reactivity insertion during a scram may not be within the assumptions of the safety analyses. Therefore, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to MODE                                                                                                                                                      3 within 12                                                                                                              hours. The allowed Completion Time of 12                                                                                                    hours is reasonable, based on operating experience, to reach MODE                                                                                                                                                                                                                                                          3 from full power conditions in an orderly manner and without challenging plant systems.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                                                                                                                                                  B 3.1-                                                  24                                                                                                                                                                                                                            Revision No. 0 Control Rod Scram Times B 3.1.4
 
BASES  (continued)
 
SURVEILLANCE                                                                                                                                                                          The four SRs of this LCO are modified by a Note stating that REQUIREMENTS                                                                                                                                                                          during a single control rod scram time surveillance, the CRD pumps shall be isolated from the associated scram accumulator. With the CRD pump isolated, (i.e., charging valve closed) the influence of the CRD pump head does not affect the single control rod scram times. During a full core scram, the CRD pump head would be seen by all control rods and would have a negligible effect on the scram insertion times.
 
SR3.1.4.1
 
The scram reactivity used in DBA and transient analyses is based on an assumed control rod scram time. Measurement of the scram times with reactor steam dome pressure                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                            800                              psig demonstrates acceptable scram times for the transients analyzed in References                                                                                                                                                                                                                            3 and                                                  4.
 
Maximum scram insertion times occur at a reactor steam dome pressure of approximately 800                                                                                                                                                                                                                                                                                                  psig because of the competing effects of reactor steam dome pressure and stored accumulator energy. Therefore, demonstration of adequate scram times at reactor steam dome pressure                                                                                                                                                                                                                                                                                                                                                                                                                                                800                              psig ensures that the measured scram times will be within the specified limits at higher pressures. Limits are specified as a function of reactor pressure to account for the sensitivity of the scram insertion times with pressure and to allow a range of pressures over which scram time testing can be performed. To ensure that scram time testing is performed within a reasonable time after a shutdown 120                              days or longer, all control rods are required to be tested before exceeding 40%                                                                                                                                                                                                                                                                              RTP. This Frequency is acceptable considering the additional surveillances performed for control rod OPERABILITY, the frequent verification of adequate accumulator pressure, and the required testing of control rods affected by fuel movement within the associate core cell and by work on control rods or the CRD System.
 
SR3.1.4.2
 
Additional testing of a sample of control rods is required to verify the continued performance of the scram function during the cycle. A representative sample contains at least 10% of the control rods. The sample remains representative
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                                                                                                                                                  B 3.1-                                                  25                                                                                                                                                                                                                  Revision No. 57 Control Rod Scram Times B 3.1.4
 
BASES
 
SURVEILLANCE                                                                                                                                                                          SR3.1.4.2(continued)
REQUIREMENTS if no more than 7.5% of the control rods in the sample tested are determined to be "slow". With more than 7.5% of the sample declared to be "slow" per the criteria in Table                                                  3.1.4-                                                1, additional control rods are tested until this 7.5%                                        criterion (i.e., 7.5% of the active sample size) is satisfied, or until the total number of "slow" control rods (throughout the core, from all Surveillances) exceeds the LCO                              limit. For planned testing, the control rods selected for the sample should be different for each test. Data from inadvertent scrams should be used whenever possible to avoid unnecessary testing at power, even if the control rods with data may have been previously tested in a sample. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
 
SR3.1.4.3
 
When work that could affect the scram insertion time is performed on a control rod or the CRD System, testing must be done to demonstrate that each affected control rod retains adequate scram performance over the range of applicable reactor pressures from zero to the maximum permissible pressure. This surveillance can be met by performance of either scram time testing or Diaphragm Alternative Response Time (DART) testing, when it is concluded that DART testing monitors the performance of all affected components. The testing must be performed once before declaring the control rod OPERABLE. The required testing must demonstrate the affected control rod is still within acceptable limits. The limits for reactor pressures
                                            < 800 psig are established based on a high probability of meeting the acceptance criteria at reactor pressures                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                    800 psig. Limits for                                                                                                                                                                                      800 psig are found in Table 3.1.4-                                                                                                                                                                                                                                                                                                                                        1. If testing demonstrates the affected control rod does not meet these limits, but is within the 7                                                                                                                                                                                                                                                                                                                                        second limit of Table 3.1.4-                                                1, Note 2, the control rod can be declared OPERABLE and "slow."
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                                                                                                                                                  B 3.1-                                                  26                                                                                                                                                                                                                  Revision No. 86 Control Rod Scram Times B 3.1.4
 
BASES
 
SURVEILLANCE                                                                                                                                                                          SR3.1.4.3(continued)
REQUIREMENTS Specific examples of work that could affect the scram times are (but are not limited to) the following:  removal of any CRD for maintenance or modification; replacement of a control rod; and maintenance or modification of a scram solenoid pilot valve, scram valve, accumulator, isolation valve or check valve in the piping required for scram.
 
The Frequency of once prior to declaring the affected control rod OPERABLE is acceptable because of the capability to test the control rod over a range of operating conditions and the more frequent surveillances on other aspects of control rod OPERABILITY.
 
SR3.1.4.4
 
When work that could affect the scram insertion time is performed on a control rod or CRD System, or when fuel movement within the reactor vessel occurs testing must be done to demonstrate each affected control rod is still within the limits of Table                                                                                                                                                                                                                                                                    3.1.4-                                                1 with the reactor steam dome pressure                                                                                                                                              800                              psig. Where work has been performed at high reactor pressure, the requirements of SR                                                                                                                                                                                                                                                                                                                                                                                                                                                                3.1.4.3 and SR                    3.1.4.4 can be satisfied with one test. For a control rod affected by work performed while shut down, however, a zero pressure and high pressure test                                                                                                                                                                                                                                                                                                                                                                      may be required. This testing ensures that, prior to withdrawing the control rod for continued operation, the control rod scram performance is acceptable for operating reactor pressure conditions.
Alternatively, a control rod scram test during hydrostatic pressure testing could also satisfy both criteria. When fuel movement occurs within the reactor pressure vessel, only those control rods associated with the core cells affected by the fuel movement are required to be scram time tested.During a routine refueling outage, it is expected that all control rods will be affected.
 
The Frequency of once prior to exceeding 40%                                                                                                                                                                                                                                                                                                                                                                                                                                                      RTP is acceptable because of the capability to test the control rod over a range of operating conditions and the more frequent surveillances on other aspects of control rod OPERABILITY.
 
REFERENCES                                                                                                                                                                          1.                                        UFSAR, Sections 1.5.1.3 and 1.5.2.2.
: 2.                                        UFSAR, Section                                                                                                                                            14.6.2.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                                                                                                                                                  B 3.1-                                                  27                                                                                                                                                                                                                  Revision No. 57 Control Rod Scram Times B 3.1.4
 
BASES
 
REFERENCES                                                                                                                                                                          3.                                        UFSAR, Appendix K, Section VI.
(continued)
: 4.                                        UFSAR, Chapter 14.
: 5.                                        NEDE-                                        24011-P-A                                                , "General Electric Standard Application for Reactor Fuel," latest approved revision.
: 6.                                        Letter from R. E. Janecek (BWROG) to R. W. Starostecki (NRC), "BWR Owners Group Revised Reactivity Control System Technical Specifications," BWROG-                                                                                                                                                                                                                                                                                                                                                                                                      8754, September 17, 1987.
 
PBAPS UNIT 2                                                                                                                                                                                                                                                                                                                                                  B 3.1-                                                  28                                                                                                                                                                                                                  Revision No. 72 Control Rod Scram Accumulators B 3.1.5
 
B 3.1  REACTIVITY CONTROL SYSTEMS
 
B 3.1.5  Control Rod Scram Accumulators
 
BASES
 
BACKGROUND                                                                                                                                                                          The control rod scram accumulators are part of the Control Rod Drive (CRD) System and are provided to ensure that the control rods scram under varying reactor conditions. The control rod scram accumulators store sufficient energy to fully insert a control rod at any reactor vessel pressure.
The accumulator is a hydraulic cylinder with a free floating piston. The piston separates the water used to scram the control rods from the nitrogen, which provides the required energy. The scram accumulators are necessary to scram the control rods within                                                                                                                                                                                              the required insertion times of LCO                              3.1.4, "Control Rod Scram Times."
 
APPLICABLE                                                                                                                                                                          The analytical methods and assumptions used in evaluating SAFETY ANALYSES                                                                                                                                                                          the control rod scram function are presented in References                                                                                                    1, 2, and                                                                                          3. The Design Basis Accident (DBA) and transient analyses assume that all of the control rods scram at a specified insertion rate. OPERABILITY of each individual control rod scram accumulator, along with LCO                              3.1.3, "Control Rod OPERABILITY," and                                                                                                                                                                                                                                                                                                                                                                                  LCO                              3.1.4, ensures that the scram reactivity assumed in the DBA and transient analyses can be met. The existence of an inoperable accumulator may invalidate prior scram time measurements for the associated control                                                                                                                                                                                                                            rod.
 
The scram function of the CRD System, and therefore the OPERABILITY of the accumulators, protects the MCPR Safety Limit (see Bases for SL 2.1.1, "Reactor Core SLs" and LCO                              3.2.2, "MINIMUM CRITICAL POWER RATIO (MCPR)") and 1%                    cladding plastic strain                                                                                                                                                                                                                                      fuel design limit (see Bases for LCO                              3.2.3, "LINEAR HEAT GENERATION RATE (LHGR)"), which ensure that no fuel damage will occur if these limits are not exceeded (see Bases for LCO                                                                                                                                                                                                                                                                              3.1.4). In addition, the scram function at low reactor vessel pressure (i.e., startup conditions) provides protection against violating fuel design limits during reactivity insertion accidents (see Bases for LCO                              3.1.6, "Rod Pattern Control").
 
Control rod scram accumulators satisfy Criterion                                                                                                                                                                                                                                                                                                                                                                                                                                                                                              3 of the NRC Policy Statement.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                                                                                                                                                  B 3.1-                                                  29                                                                                                                                                                                                                  Revision No. 49 Control Rod Scram Accumulators B 3.1.5
 
BASES  (continued)
 
LCO                                                                                                                                                                          The OPERABILITY of the control rod scram accumulators is required to ensure that adequate scram insertion capability exists when needed over the entire range of reactor pressures. The OPERABILITY of the scram accumulators is based on maintaining adequate accumulator pressure.
 
APPLICABILITY                                                                                                                                                                          In MODES                                                                                1 and                                                  2, the scram function is required for mitigation of DBAs and transients, and therefore the scram accumulators must be OPERABLE to support the scram function. In MODES                                                                                                                                                                                    3 and                                                  4, control rods are not able to be withdrawn since the reactor mode switch is in shutdown and a control rod block is applied. This provides adequate requirements for control rod scram accumulator OPERABILITY during these conditions. Requirements for scram accumulators in MODE                                                                                                                                                                                                        5 are contained in LCO                                                                                                                                                                                                                            3.9.5, "Control Rod OPERABILITY-                                                                                                                                                      Refueling."
 
ACTIONS                                                                                                                                                                          The ACTIONS Table is modified by a Note indicating that a separate Condition entry is allowed for each control rod scram accumulator. This is acceptable since the Required Actions for each Condition provide appropriate compensatory actions for each inoperable accumulator. Complying with the Required Actions may allow for continued operation and subsequent inoperable accumulators governed by subsequent Condition entry and application of associated Required Actions.
 
A.1 and A.2
 
With one control rod scram accumulator inoperable and the reactor steam dome pressure                                                                                                                                                                                                                                                                                          900                              psig, the control rod may be declared "slow," since the control rod will still scram at the reactor operating pressure but may not satisfy the required scram times in Table                                                                                                                                                                                                                                                                                                  3.1.4-                                                1. Required Action                                                                                                                                                                                              A.1 is modified by a Note indicating that declaring the control rod "slow" only applies if the associated control scram time was within the limits of Table                                                                                                                                                                                                                                                                                                            3.1.4-                                                1 during the last scram time test. Otherwise, the control rod would already be considered "slow" and the further degradation of scram performance with an inoperable accumulator could result in excessive scram times. In this event, the associated
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                                                                                                                                                  B 3.1-                                                  30                                                                                                                                                                                                                            Revision No. 0 Control Rod Scram Accumulators B 3.1.5
 
BASES
 
ACTIONS                                                                                                                                                                          A.1 and A.2(continued)
 
control rod is declared inoperable (Required Action                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                            A.2) and LCO                              3.1.3 is entered. This would result in requiring the affected control rod to be fully inserted and disarmed, thereby satisfying its intended function, in accordance with ACTIONS of LCO                                                                                                                                            3.1.3.
 
The allowed Completion Time of 8                                                                                                                                                                                                                                                                                                                              hours is reasonable, based on the large number of control rods available to provide the scram function and the ability of the affected control rod to scram only with reactor pressure at high reactor pressures.
 
B.1, B.2.1, and B.2.2
 
With two or more control rod scram accumulators inoperable and reactor steam dome pressure                                                                                                                                                                                                                                                                                                                                  900                              psig, adequate pressure must be supplied to the charging water header.
With inadequate charging water pressure, all of the accumulators could become inoperable, resulting in a potentially severe degradation of the scram performance.
Therefore, within 20                                                                                                                                                                                                        minutes from discovery of charging water header pressure <                                                                                                                                                                                                                                      940                              psig concurrent with Condition                                                                                          B, adequate charging water header pressure must be restored. The allowed Completion Time of 20                                                                                                                                                                                                                                                                                                                                                                                                                                                      minutes is reasonable, to place a CRD pump into service to restore the charging water header pressure, if required. This Completion Time is based on the ability of the reactor pressure alone to fully insert all control rods.
 
The control rod may be declared "slow," since the control rod will still scram using only reactor pressure, but may not satisfy the times in Table                                                                                                                                                                                                                                                                                                            3.1.4-                                                1. Required Action                                                            B.2.1 is modified by a Note indicating that declaring the control rod "slow" only applies if the associated control scram time is within the limits of Table                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                3.1.4-                                                  1 during the last scram time test. Otherwise, the control rod would already be considered "slow" and the further degradation of scram performance with an inoperable accumulator could result in excessive scram times. In this event, the associated control rod is declared inoperable (Required Action                                                                                                                                                                B.2.2) and LCO                                                                                                                                            3.1.3 entered. This would
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                                                                                                                                                  B 3.1-                                                  31                                                                                                                                                                                                                            Revision No. 2 Control Rod Scram Accumulators B 3.1.5
 
BASES
 
ACTIONS                                                                                                                                                                          B.1, B.2.1, and B.2.2(continued)
 
result in requiring the affected control rod to be fully inserted and disarmed, thereby satisfying its intended function in accordance with ACTIONS of LCO                                                                                                                                                                                                                                                                                                                                                                                                                                  3.1.3.
 
The allowed Completion Time of 1                                                                                                                                                                                                                                                                                                                              hour is reasonable, based on the ability of only the reactor pressure to                                                                                                                                                                                                                                                                                                                                                                                                                                                                          scram the control rods and the low probability of a DBA or transient occurring while the affected accumulators are inoperable.
 
C.1 and C.2
 
With one or more control rod scram accumulators inoperable and the reactor steam dome pressure <                                                                                                                                                                                                                                                                                                                                                                                  900                              psig, the pressure supplied to the charging water header must be adequate to ensure that accumulators remain charged. With the reactor steam dome pressure <                                                                                                                                                                                                                  900                              psig, the function of the accumulators in providing the scram force becomes much more important since the scram function could become severely degraded during a depressurization event or at low reactor pressures. Therefore, immediately upon discovery of charging water header pressure <                                                                                                                                                                                                                                                                                                                                940                              psig, concurrent with Condition                                                                                          C, all control rods associated with inoperable accumulators must be verified to be fully inserted.
Withdrawn control rods with inoperable accumulators may fail to scram under these low pressure conditions. The associated control rods must also be declared inoperable within 1                                                                                hour. The allowed Completion Time of 1                                                                                                                                                                                                                                                                                                                                                                                                    hour is reasonable for Required Action                                                                                                                                                                                                                                                                                                            C.2, considering the low probability of a DBA or transient occurring during the time that the accumulator is inoperable.
 
D.1
 
The reactor mode switch must be immediately placed in the shutdown position if either Required Action and associated Completion Time associated with the loss of the CRD charging pump (Required Actions B.1 and C.1) cannot be met. This ensures that all insertable control rods are inserted and that the reactor is in a                                                                                                                                                                                                                                                condition that does not require the
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                                                                                                                                                  B 3.1-                                                  32                                                                                                                                                                                                                            Revision No. 2 Control Rod Scram Accumulators B 3.1.5
 
BASES
 
ACTIONS                                                                                                                                                                          D.1(continued)
 
active function (i.e., scram) of the control rods. This Required Action is modified by a Note stating that the action is not applicable if all control rods associated with the inoperable scram accumulators are fully inserted, since the function of the control rods has been performed.
 
SURVEILLANCE                                      SR3.1.5.1 REQUIREMENTS SR                    3.1.5.1 requires that the accumulator pressure be periodically checked to ensure adequate accumulator pressure exists to provide sufficient scram force. The primary indicator of accumulator OPERABILITY is the accumulator pressure. A minimum accumulator pressure is specified, below which the capability of the accumulator to perform its intended function becomes degraded and the accumulator is considered inoperable. The minimum accumulator pressure of 940                              psig is well below the expected pressure of approximately 1450                                        psig (Ref.                                                                                                    1).Declaring the accumulator inoperable when the minimum pressure is not maintained ensures that significant degradation in scram times does not occur. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
 
REFERENCES                                                                                                                                                                          1.                                        UFSAR, Section                                                                                                                                            3.4.5.3 and                                                                                                              Figure 3.4.10.
: 2.                                        UFSAR, Appendix K, Section VI.
: 3.                                        UFSAR, Chapter 14.
 
PBAPS UNIT 2                                                                                                                                                                                                                                                                                                                                                  B 3.1-                                                  33                                                                                                                                                                                                                  Revision No. 86 Rod Pattern Control B 3.1.6
 
B 3.1  REACTIVITY CONTROL SYSTEMS
 
B 3.1.6  Rod Pattern Control
 
BASES
 
BACKGROUND                                                                                                                                                                          Control rod patterns during startup conditions are controlled by the operator and the rod worth minimizer (RWM) (LCO                                                                                                    3.3.2.1, "Control Rod Block Instrumentation"),
so that only specified control rod sequences and relative positions are allowed over the operating range of all control rods inserted to 10%                                                                                                                                                                                                                                                                                        RTP. The sequences limit the potential amount of reactivity addition that could occur in the event of a Control Rod Drop Accident (CRDA).
 
This Specification assures that the control rod patterns are consistent with the assumptions of the CRDA analyses of References                                                                                                    1 and                                                  2.
 
APPLICABLE                                                                                                                                                                          The analytical methods and assumptions used in evaluating SAFETY ANALYSES                                                                                                                                                                          the CRDA are summarized in References                                                                                                                                                                                                                                                                                                                                                                                1 and                                                  2. CRDA analyses assume that the reactor operator follows prescribed withdrawal sequences. These sequences define the potential initial conditions for the CRDA analysis. The RWM (LCO                                        3.3.2.1) provides backup to operator control of the withdrawal sequences to ensure that the initial conditions of the CRDA analysis are not violated.
 
Prevention or mitigation of positive reactivity insertion events is necessary to limit the energy deposition in the fuel, thereby preventing significant fuel damage which could result in the undue release of radioactivity. Since the failure consequences for UO  2  have been shown to be insignificant below fuel energy depositions of 300                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                    cal/gm (Ref.                                                  3), the fuel damage limit of 280                                                                                                                                                                                                                                                                                                                              cal/gm provides a margin of safety from significant core damage which would result in release of radioactivity (Refs.                                                                                                                                                                                                                                                                                                                                                                                                                          5). Generic evaluations (Refs.                                                                                                                                                                                    1 and                                                  6) of a design basis CRDA (i.e., a CRDA resulting in a peak fuel energy deposition of 280                              cal/gm) have shown that if the peak fuel enthalpy remains below 280                                                                                                                                                                          cal/gm, then the maximum reactor pressure will be less than the required ASME Code limits (Ref.                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                7) and the calculated offsite doses will be well within the required limits (Ref.                                                                                                                                                                                                                  5).
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                                                                                                                                                  B 3.1-                                                  34                                                                                                                                                                                                                  Revision No. 75 Rod Pattern                                                                                                              Control B 3.1.6
 
BASES
 
APPLICABLE                                                                                                                                                                          Control rod patterns analyzed in Reference                                                                                                                                                                                                                                                                                                                                                                                                                                  1 follow the SAFETY ANALYSES                                                                                                                                                                          analyzed rod position sequence. The analyzed rod position (continued)                                                                                                                                                                          sequence is applicable from the condition of all control rods fully inserted to 10%                                                                                                                                                                                                                                                                    RTP (Ref.                                                                                          2). For the analyzed rod position sequence, the control rods are required to be moved in groups, with all control rods assigned to a specific group required to be within specified banked positions. The banked positions are established to minimize the maximum incremental control rod worth without being overly restrictive during normal plant operation. Generic analysis of the analyzed rod position sequence (Ref.                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                      1) has demonstrated that the 280                                                                                                                                                                                                                                                          cal/gm fuel damage limit will not be violated during a CRDA while following the analyzed rod position sequence mode of operation. The generic                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                          analyzed rod position sequence analysis (Ref.                                                                                                                                                                                                                                                                                                                                                            8) also evaluates the effect of fully inserted, inoperable control rods not in compliance with the sequence, to allow a limited number (i.e., eight) and distribution of fully inserted, inoperable control rods.
 
When performing a shutdown of the plant, an optional rod position sequence (Ref. 9) may be used provided that all withdrawn control rods have been confirmed to be coupled.
The rods may be inserted without the need to stop at intermediate positions since the possibility of a CRDA is eliminated by the confirmation that withdrawn control rods are coupled. When using the (Ref. 9) control rod sequence for shutdown, the RWM may be reprogrammed to enforce the requirements of the improved control rod insertion process, or may be bypassed and the analyzed                                                                                                                                                                                                                                                                                                                                                            rod position sequence implemented under LCO 3.3.2.1, Condition D controls.
 
In order to use the Reference 9 shutdown process, an extra check is required in order to consider a control rod to be confirmed to be coupled. This extra check ensures that no single operator error can result in an incorrect coupling check. For purposes of this shutdown process, the method for confirming that control rods are coupled varies depending on the position of the control rod in the core. Detail on this coupling confirmation requirement are provided in Reference                                                                                          9. If the requirements for use of the control rod insertion process contained in Reference 9 are followed, the plant is considered in compliance with the rod position sequence as required by LCO 3.1.6.
 
Rod pattern control satisfies Criterion                                                                                                                                                                                                                                                                                                                                                            3 of the NRC Policy Statement.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                                                                                                                                                  B 3.1-                                                  35                                                                                                                                                                                                        Revision No.                                                                                                                        114 Rod Pattern Control B 3.1.6
 
BASES  (continued)
 
LCO                                                                                                                                                                          Compliance with the prescribed control rod sequences minimizes the potential consequences of a CRDA by limiting the initial conditions to those consistent with the analyzed rod position sequence. This LCO only applies to OPERABLE control rods. For inoperable control rods required to be inserted, separate requirements are specified in LCO                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                        3.1.3, "Control Rod OPERABILITY," consistent with the allowances for inoperable control rods in the analyzed rod position sequence.
 
APPLICABILITY                                                                                                                                                                          In MODES                                                                                1 and                                                  2, when THERMAL POWER is                                                                                                                                                                                                                                                            10%                              RTP, the CRDA is a Design Basis Accident and, therefore, compliance with the assumptions of the safety analysis is required. When THERMAL POWER is >                                                                                                                                                                                    10%                              RTP, there is no credible control rod configuration that results in a control rod worth that could exceed the 280                                                                                                                                            cal/gm fuel damage limit during a CRDA (Ref.                                        2). In MODES                                                                                                                                  3, 4, and                                                                                          5, since the reactor is shut down and only a single control rod can be withdrawn from a core cell containing fuel assemblies, adequate SDM ensures that the consequences of a CRDA are acceptable, since the reactor will remain subcritical with a single control rod withdrawn.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                                                                                                                                              B 3.1-                                                  35a                                                                                                                                                                                                                        Revision No. 63 Rod Pattern Control B 3.1.6
 
BASES  (continued)
 
ACTIONS                                    A.1 and                                                                      A.2
 
With one or more OPERABLE control rods not in compliance with the analyzed rod position sequence, actions may be taken to either correct the control rod pattern or declare the associated control rods inoperable within 8                                                                                                                                                                                                                                                                                                                                                                                                                                                                                    hours.
Noncompliance with the prescribed sequence may be the result of "double notching," drifting from a control rod drive cooling water transient, leaking scram valves, or a power reduction to                                                                                                                                    10%                              RTP before establishing the correct control rod pattern. The number of OPERABLE control rods not in compliance with the prescribed sequence is limited to eight, to prevent the operator from attempting to correct a control rod pattern that significantly deviates from the prescribed sequence. When the control rod pattern is not in compliance with the prescribed sequence, all control rod movement must be stopped except for moves needed to correct the rod pattern, or scram if warranted.
 
Required Action                                                                                                                                                      A.1 is modified by a Note which allows the RWM to be bypassed to allow the affected control rods to be returned to their correct position. LCO 3                                                                                                                                                                                                                                                                                                                                                                                                              .3.2.1 requires verification of control rod movement by a second licensed operator or a qualified member of the technical staff (i.e.,
personnel trained in accordance with an approved training program). This ensures that the control rods will be moved to the correct position. A control rod not in compliance with the prescribed sequence is not considered inoperable except as required by Required Action                                                                                                                                                                                                                                                                                                                                                                                A.2. The allowed Completion Time of 8                                                                                                                                                                                                        hours is reasonable, considering the restrictions on the number of                                                                                                                                                                                                                                                                                                  allowed out of sequence control rods and the low probability of a CRDA occurring during the time the control rods are out of sequence.
 
B.1 and                                                                      B.2
 
If nine or more OPERABLE control rods are not in compliance with the analyzed rod position sequence, the control rod pattern significantly deviates from the prescribed sequence.
Control rod withdrawal should be suspended immediately to prevent the potential for further deviation from the prescribed sequence. Control rod insertion to correct control rods withdrawn beyond their allowed position is allowed since, in general, insertion of control rods has
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                                                                                                                                                  B 3.1-                                                  36                                                                                                                                                                                                                  Revision No. 63 Rod Pattern Control B 3.1.6
 
BASES
 
ACTIONS                                                                                                                                                                          B.1 and                                                                      B.2(continued)
 
less impact on control rod worth than withdrawals have.
Required Action                                                                                                                                                      B.1 is modified by a Note which allows the RWM to be bypassed to allow the affected control rods to be returned to their correct position.
 
LCO                              3.3.2.1 requires verification of control rod movement by a second licensed operator or a qualified member of the technical staff.
 
When nine or more OPERABLE control rods are not in compliance with the analyzed rod position sequence, the reactor mode switch must be placed in the shutdown position within 1                                                                                hour. With the mode switch in shutdown, the reactor is shut down, and as such, does not meet the applicability requirements of this LCO. The allowed Completion Time of 1                                                                                                                                                                                                        hour is reasonable to allow insertion of control rods to restore compliance, and is appropriate relative to the low probability of a CRDA occurring with the control rods out of sequence.
 
SURVEILLANCE                                              SR3.1.6.1 REQUIREMENTS The control rod pattern is periodically                                                                                                                                                                                                                                                                              verified to be in compliance with the analyzed rod position sequence to ensure the assumptions of the CRDA analyses are met. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. The RWM                                                                                                                                                                                                                                                                                                                                                            provides control rod blocks to enforce the required sequence and is required to be OPERABLE when operating at                                                                                                                                                                                                                                                                                10%                              RTP.
 
REFERENCES                                                                                                                                                                          1.                                        NEDE-                                        24011-P-                                                A, "General Electric Standard Application for Reactor Fuel, latest approved revision.
: 2.                                        Letter (BWROG-                                                                                                                                8644) from T. Pickens (BWROG) to G. C.
Lainas (NRC), "Amendment 17 to General Electric Licensing Topical Report NEDE-                                                                                                                                                                                                                                                                                                  24011-P-                                                  A."
: 3.                                        UFSAR, Section 14.6.2.3.
: 4.                                        Deleted.
: 5.                                        10 CFR                                                            50.67.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                                                                                                                                                  B 3.1-                                                  37                                                                                                                                                                                                                  Revision No. 86 Rod Pattern Control B 3.1.6
 
BASES
 
REFERENCES                                                                                                                                                                          6.                                        NEDO-                                        21778-                                                A, "Transient Pressure Rises Affected (continued)                                                                                                                                                                                                                  Fracture Toughness Requirements for Boiling Water Reactors," December                                                                                                                                                                                              1978.
: 7.                                        ASME, Boiler and Pressure Vessel Code.
: 8.                                        NEDO-                                      21231, "Banked Position Withdrawal Sequence,"
January                                                                      1977.
: 9.                                        NEDO-                                        33091-                                                  A, Improved BPWS Control Rod Insertion Process, Revision 2, July 2004.
 
PBAPS UNIT 2                                                                                                                                                                                                                                                                                                                                                  B 3.1-                                                  38                                                                                                                                                                                                                  Revision No. 61 SLC System B 3.1.7
 
B 3.1  REACTIVITY CONTROL SYSTEMS
 
B 3.1.7  Standby Liquid Control (SLC) System
 
BASES
 
BACKGROUND                                                                                                                                                                          The SLC System is designed to provide the capability of bringing the reactor, at any time in a fuel cycle, from full power and minimum control rod inventory (which is at the peak of the xenon transient) to a subcritical condition with the reactor in the most reactive, xenon free state without taking credit for control rod movement. The SLC System satisfies the requirements of 10                                                                                                                                                                                                                            CFR                              50.62 (Ref.                                                                                                              1) on anticipated transient without scram using highly                                                                                                                                                                                                                                                                                                            enriched boron.Using highly enriched boron                                                                                                                                                                                                                  in the SLC System increases the rate of Boron-                                                10 injection and functions to shutdown the reactor core faster. This limits the heat generated that is transferred to the suppression                                                                                                                                                                                    pool during and ATWS event. Limiting the heat transferred to the suppression pool maintains the pool below design limits, which ensures adequate net positive suction head (NPSH) is available for the emergency core cooling system (ECCS) pumps without credit for containment accident pressure.
 
The SLC System is also used to maintain suppression pool pH at or above 7 following a loss of coolant accident (LOCA) involving significant fission product releases. Maintaining suppression pool pH levels at or above 7 following an accident ensures that sufficient iodine will be retained in the suppression pool water.
 
Reference 1 requires a SLC System with a minimum flow capacity and boron content equivalent in control capacity to 86 gpm of 13 weight percent sodium pentaborate solution.
Natural sodium pentaborate solution is 19.8% atom Boron-                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                      10.
Therefore, the system parameters of concern, boron concentration (C), SLC pump flow rate (Q), and Boron-                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                        10 enrichment (E), may be expressed as a multiple of ratios.
The expression is as follows:
 
C            Q              E
___________ x __________ x ____________
13% weight      86 gpm      19.8% atom
 
If the product of this expression is                                                                                                                                                                                                                                                                                                                                                                                    1, then the SLC System satisfies the criteria of Reference 1. As such, the product of this expression at the minimum acceptance
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                                                                                                                                                  B 3.1-                                                  39                                                                                                                                                                                                        Revision No. 114 SLC System B 3.1.7
 
BASES
 
BACKGROUND                                                                                                                                                                          criteria for the surveillances of concentration, flow rate (continued)                                                                                                                                                                          and boron enrichment is >                                                                                                                                                                                                                                                          1.69, which reflects that the SLC system exceeds the criteria of Reference 1.
 
The SLC System consists of a boron solution storage tank, two positive displacement pumps, two explosive valves that are provided in parallel for redundancy, and associated piping and valves used to transfer borated water from the storage tank to the reactor pressure vessel (RPV). The borated solution is discharged near the bottom of the core shroud, where it then mixes with the cooling water rising through the core. A smaller tank containing demineralized water is provided for testing purposes.
 
APPLICABLE                                                                                                                                                                          The SLC System is manually initiated from the main control SAFETY ANALYSES                                                                                                                                                                  room, as directed by the emergency operating procedures, if the operator believes the reactor cannot be shut down, or kept shut down, with the control rods. The SLC System is used in the event that enough control rods cannot be inserted to accomplish shutdown and cooldown in the normal manner. The SLC System injects borated water into the reactor core to add negative reactivity to compensate for all of the various reactivity effects that could occur during plant operations. To meet this objective, it is necessary to inject a quantity of boron, which produces a concentration of 660                                                                                                                                                                                                        ppm of natural boron, in the reactor coolant at 68&deg;F. To allow for potential leakage and imperfect mixing in the reactor system, an additional amount of boron equal to 25% of the amount cited above is added as a minimum (Ref.                                                                                                                                                      2). The minimum level of sodium pentaborate in solution in the SLC tank (i.e., SR 3.1.7.1,  52%) and the temperature versus concentration limits in Figure                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                  3.1.7-1 are calculated such that the required concentration is achieved, with additional margin associated with using highly enriched boron to increase the rate of Boron-                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                              10 injection, accounting for                                                                                                                                                                                                                                                          dilution in the RPV with normal water level and including the water volume in the residual heat removal shutdown cooling piping and in the recirculation loop piping. This quantity of borated solution is the amount that is above the pump suction shutoff level in the boron solution storage tank. No credit is taken for the portion of the tank volume that cannot be injected. The maximum allowable                                                                                                                                                                                                                                      concentration of sodium pentaborate depicted in Figure                                                                                                                                                                                    3.1.7-                                                1 has been established to ensure that the solution saturation temperature does not exceed 43&deg;F. Using highly enriched boron (i.e., SR 3.1.7.10,                                                                                            92.0%) in the SLC System increases the rate of
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                                                                                                                                                  B 3.1-                                                  40                                                                                                                                                                                                              Revision No.                                                                                                                        114 SLC System B 3.1.7
 
BASES
 
APPLICABLE                                                                                                                                                                          Boron-                                                  10                    injection and functions to shutdown the reactor SAFETY ANALYSES                                                                                                                                                                          core                                        faster. This limits the heat generated that is (continued)                                                                                                                                                                          transferred to t                                                                                                                                          he suppression pool during an ATWS event.
Limiting the heat transferred to the suppression pool maintains the pool below design limits, which ensures adequate NPSH is available for the ECCS pumps without credit for containment accident pressure.
 
The sodium pentaborate solution in the SLC System is also used, post-                                                                                                  LOCA, to maintain suppression pool pH at or above
: 7. The system parameters used in the calculation are the minimum allowable volume, Boron-                                                                                                                                                                                                                                                                                                                    10 enrichment, and concentration of sodium pentaborate in solution in the SLC tank. These minimum allowable values are required to maintain suppression pool pH                                                                                                                                                                                                                                                                                                    7.0 post-                                                                                LOCA.                                                  This prevents radioactive iodine from re-                                                                                                                                                                                                                                                                  evolving, which limits the iodine release to the plant environs and minimizes the radiological consequences to comply with 10 CFR 50.67 l                                                                                                                                                                                                                                                                                                                                                                                                                        imits (Ref. 3).
 
The SLC System satisfies Criteri                                                                                                                                                                                                                                                                                                                    a 3 and                                                                      4 of the NRC Policy Statement.
 
LCO                                                                                                                                                                          The OPERABILITY of the SLC System provides backup capability for reactivity control independent of normal reactivity control provisions provided by the control rods. The OPERABILITY of the SLC System is based on the conditions of the borated solution in the storage tank and the availability of a flow path to the RPV, including the OPERABILITY of the pumps and valves. Two SLC subsystems are required to be OPERABLE; each contains an OPERABLE pump, an explosive valve, and associated piping, valves, and instruments and controls to ensure an OPERABLE flow path.
 
APPLICABILITY                                                                                                                                                                          In MODES                                                                                1 and                                                  2, shutdown capability is required. In MODES 1, 2, and                                                                                          3, SLC System injection capability is required in order to maintain post DBA LOCA suppression pool pH. In MODES                                                  3 and                                                  4, control rods are not able to be withdrawn since the reactor mode switch is in shutdown and a control rod block is applied. This provides adequate controls to ensure that the reactor remains subcritical. In MODE                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                  5, only a single control rod can be withdrawn from a core cell containing fuel assemblies. Demonstration of adequate SDM (LCO                                        3.1.1, "SHUTDOWN MARGIN (SDM)") ensures that the reactor will not become critical. Therefore, the SLC System is not required to be OPERABLE when only a single control rod can be withdrawn.
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                                                                                                                                                  B 3.1-                                                  41                                                                                                                                                                                                        Revision No.                                                                                                                        114 SLC System B 3.1.7
 
BASES
 
APPLICABILITY                                                                                                                                                                          In MODES 1, 2, and 3, the SLC System must be OPERABLE to (continued)                                                                                                                                                                          ensure that offsite doses remain within 10 CFR 50.67 (Ref. 3) limits following a LOCA involving significant fission product releases. The SLC System is designed to maintain suppression pool pH at or above 7 following a LOCA involving significant fission product releases to ensure that                                                                                                                                                                                                                                                                                                                                                                                                    iodine will be retained in the suppression pool water.
 
ACTIONS                                        A.1 and A.2
 
If the boron solution concentration is >                                                                                                                                                                                                                                                                                                                                                                                                              9.82% weight but the concentration and temperature of boron in solution and pump suction piping temperature are within the limits of Figure 3.1.7-                                                                                                                      1, operation is permitted for a limited period since the SLC subsystems are capable of performing the intended function. It is not necessary under these conditions to declare both SLC subsystems inoperable since the SLC subsystems are capable of performing their intended function.
 
The concentration and temperature of boron in solution and pump suction piping temperature must be verified to be within the limits of Figure 3.1.7-                                                                                                                                                                                                                                                                                                                                        1 within 8 hours and once per 12 hours thereafter (Required Action A.1). The temperature versus concentration curve of Figure 3.1.7-1                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                          ,
for concentrations > 9.82% weight,                                                                                                                                                                                                                                                                                                                                                  ensures a 10&deg;F margin will be maintained above the saturation temperature. This verification ensures that boron does not precipitate out of solution in the storage tank or in the pump suction piping due to low boron solution temperature (below the saturation temperature for the given concentration). The Completion Time for performing Required Action A.1 is considered acceptable given the low probability of a Design Basis Accident (DBA) or transient occurring concurrent with the failure of the control rods to shut down the reactor and operating experience which has shown there are relatively slow variations in the measured parameters of concentration and temperature over these time periods.
 
Continued operation is only permitted for 72 hours before boron solution concentration must be restored to                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                            9.82%
weight. Taking into consideration that the SLC System design capability still exists for vessel injection under these conditions and the low probability of the temperature and concentration limits of Figure 3.1.7-                                                                                                                                                                                                                                                                                                                                                                      1 not being met, the allowed Completion Time of 72 hours is acceptable and provides adequate time to restore concentration to within limits.
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                                                                                                                                                  B 3.1-                                                  42                                                                                                                                                                                                        Revision No. 114 SLC System B 3.1.7
 
BASES
 
ACTIONS                                                B.1 (continued)
 
If one SLC subsystem is inoperable for reasons other than Condition A, the inoperable subsystem must be restored to OPERABLE status within 7                                                                                                                                                                                                                                                days o                                      r in accordance with the Risk Informed Completion Time Program.                                                                                                                                                                                                                                                                                                                              In this condition, the remaining OPERABLE subsystem is adequate to perform the shutdown function. However, the overall reliability is reduced because a single failure in the remaining OPERABLE subsystem could result in the loss of SLC System shu                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                              tdown capability. The 7                                                                                                                                                                                    day Completion Time is based on the availability of an OPERABLE subsystem capable of performing the intended SLC System function and the low probability of a DBA or severe transient occurring concurrent with the failure of the Cont                                                                                                  rol Rod Drive (CRD) System to shut down the plant.
Alternatively, a Completion Time can be determined in accordance with                                                                                                                                                      the Risk Informed Completion Time Program.
 
C.1
 
If both SLC subsystems are inoperable for reasons other than Condition A, at least                                                                                                                                                                                                        one subsystem must be restored to OPERABLE status within 8                                                                                                                                                                                                                                                hours. The allowed Completion Time of 8                                        hours is considered acceptable given the low probability of a DBA or transient occurring concurrent with the failure of the control rods to shut down the r                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                          eactor.
 
D.1 and D.2
 
If any Required Action and associated Completion Time is not met, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to MODE                                                                                                                                                      3 within 12                                                                                                              hours and MODE 4 within 3                                                                                                                                                                                                                                    6 hours.
The allowed Completion Times are reasonable, based on operating experience, to reach the required MODES from full power conditions in an orderly manner and without challenging plant                                                  systems.
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                                                                                                                                                  B 3.1-                                                  43                                                                                                                                                                                                        Revision No                                                                                                    . 159 SLC System B 3.1.7
 
BASES
 
SURVEILLANCE                                                                                                                                                                          SR3.1.7.1, SR3.1.7.2, and SR3.1.7.3(continued)
REQUIREMENTS SR                    3.1.7.1 through                                                                                                                                                      SR 3.1.7.3 verify certain characteris                                                                                                                                                                                                                                                                                                                                                                        tics of the SLC System (e.g., the level and temperature of the borated solution in the storage tank), thereby ensuring SLC System OPERABILITY without disturbing normal plant operation. These Surveillances ensure that the proper borated solution level and                                                                                                                                                                                                                                                                    temperature, including the temperature of the pump suction piping, are maintained.
Maintaining a minimum specified borated solution temperature is important in ensuring that                                                                                                                                                                                                                                                                                                  the boron remains in solution and does not precipitate out in the storage tank or                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                    in                    the pump suction piping. The temperature limit specified in SR 3.1.7.2 and SR 3.1.7.3 and the maximum sodium pentaborate concentration specified in Figure 3.1.7-                                                                                                                                                                                                                                                                                                                                1 ensures that a 10&deg;F                                        margin will be maintained above the saturation temperature. Co                                                                                                                                            ntrol room alarms for low SLC storage tank temperature and low SLC System piping temperature are available and are set at 55&deg;F. As such, SR 3.1.7.2 and SR 3.1.7.3 may be satisfied by verifying the absence of low temperature alarms for the SLC storage tank                                                                                                                                                                                                                                                                                                                                                                                                                                              and                              SLC System piping. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
 
SR3.1.7.4 and SR3.1.7.6
 
SR                    3.1.7.4 verifies the continuity of the explosive charges in the injection valves to ensure that proper                                                                                                                                                                                                                                                                                                                                                                                                                                                                  operation will occur if required. Other administrative controls, such as those that limit the shelf life of the explosive charges, must be followed. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
 
SR                    3.1.7.6                                                            verifies that each valve in the system is in its correct position, but does not apply to the squib (i.e.,
explosive) valves. Verifying the correct alignment for manual and power operated valves in the SLC System flow path provides assurance that the prop                                                                                                                                                                                                                                                                                                                      er flow paths will exist for system operation. A valve is also allowed to be in the nonaccident position provided it can be aligned to the accident position from the control                                                                                                                                                                                                                                                                                                                                                  room, or locally by a dedicated operator at the valve control. This is acceptabl                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                    e since the SLC System is a manually initiated system. This Surveillance also does not apply to valves that are locked, sealed, or otherwise secured in position since they are verified to be in the correct position prior to locking, sealing, or securing.                                                                                                                                                                                                                  This verification of valve alignment
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                                                                                                                                                  B 3.1-                                                  44                                                                                                                                                                                                        Revision No. 114 SLC System B 3.1.7
 
BASES
 
SURVEILLANCE                                                                                                                                                                          SR3.1.7.4 and SR3.1.7.6(continue                                                                                                    d)
REQUIREMENTS does not require any testing or valve manipulation; rather, it involves verification that those valves capable of being mispositioned are in the correct position. This SR does                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                not apply to valves that cannot be inadvertently misaligned, such                              as check valves. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
 
SR3.1.7.5
 
This Surveillance requires an examination of the sodium pentaborate solution by using chemical analysis to ensure that the prop                                                                                                                        er concentration of boron exists in the storage tank. Having the proper concentration of boron in the storage tank ensures the SLC subsystems will perform their intended function of injecting no less than the minimum quantity of Boron-                                                                                                                                                                        10 and amount of sod                                                                                                                                                                                              ium                              pentaborate required by ATWS                                                                                                                                                                analyses. The SLC subsystems function to quickly shutdown the reactor in the event of an ATWS. This limits the heat generated that is transferred to the suppression pool during an ATWS event. Limiting the heat transferre                                                                                        d to the suppression pool maintains the pool below design limits, which ensures adequate NPSH is available for the ECCS pumps without                                                                                                                                                                                                                            credit for                                                                                                    containment accident pressure. The SLC subsystems also function to maintain suppression pool pH                                                                                                                                                                                                          7.0 under post-                                                                                                                                          LOCA conditions.
SR                    3.1.7.5 must be performed anytime boron or water is added to the storage tank solution to determine that the boron solution concentration is                                                                                                                                                                                                                                                                                                              8.32% weight and                                                                                                                                                                            9.82% weight.
SR 3.1.7.5 must also be performed anytime the temperature is restored to within limits to ensure that no significant boron precipitation occurred. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
 
SR3.1.7.7Deleted
 
Demonstrating that each SLC System pump develops a flow rate  49.1                                        gpm at a discharge pressure                                                                                                                                                                                                                                                                                          1275                                        psig ensures that pump performance has not degraded below design values during the fuel cycle. This minimum pump flow rate requirement ensures that, when combined with the sodium pentaborate
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                                                                                                                                                  B 3.1-                                                  45                                                                                                                                                                                                        Revision No. 114 SLC System B 3.1.7
 
BASES
 
SURVEILLANCE                                                                                                                                                                          SR  3.1.7.8(continued)
REQUIREMENTS solution concentration requirements, the rate of negative reactivity insertion from the SLC System will adequately compensate for the positive reactivity effects encountered during power reduction, cooldown of the moderator, and xenon decay.                                                                                                                        The rate of negative reactivity insertion is increased by using highly enriched boron in the SLC System solution that increases the rate of Boron-                                                                                                                                                                                                                                                                                                                                                                                                                        10 injection and functions to shutdown the reactor core faster.                                                                                                                                                                                                                                                                                                                                                                                                                                                                                      This limits the heat generated that is transferred to the suppression pool during an ATWS event.                                                                                                                                                                                                                                                                    Limiting the heat transferred to the suppression pool maintains the pool below design limits, which ensures adequate NPSH is available for the ECCS pumps without credit for containment accident pressure.                                                                                          This test confirms one point on the pump design curve and                                                                                          is indicative of overall performance. Such inservice inspections confirm component OPERABILITY, trend performance, and detect incipient failures by indicating abnormal performance. The Frequency of this Surveillance is in accordance with the INSERVICE TESTING PROGRAM.
 
SR3.1.7.9
 
This Surveillance ensures that there is a functioning flow path from the boron solution storage tank to the RPV, including the firing of an explosive valve. The replacement charge for the explosive valve shall be from the same manufactured batch as the one fired or from                                                                                                                                                                                                                                                                                                                                                                                                                                            another batch that has been certified by having one of that batch successfully fired. The Surveillance may be performed in separate steps to prevent injecting boron into the RPV. An acceptable method for verifying flow from the pump to the RPV is to pump demineralized water from a test tank through one SLC subsystem and into the RPV. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                                                                                                                                                  B 3.1-                                                  46                                                                                                                                                                                                        Revision No.                                                                                                                        140 SLC System B 3.1.7
 
BASES
 
SURVEILLANCE                                                                                SR3.1.7.10 REQUIREMENTS (continued)                                                                                                                                                                          Enriched sodium pentaborate solution is made by mixing granular, enriched sodium pentaborate with water. Isotopic tests on the granular sodium pentaborate to verify                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                    the actual B-                                                          10 enrichment must be performed prior to addition to the SLC tank in order to ensure that the proper B-                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                      10 atom percentage is being used. The tests may use vendor certification documents.
 
REFERENCES                                                                                                                                                                          1.                                        10                    CFR                              50.62.
: 2.                                        UFSAR, Section                                                                                                                                            3.8.4.
: 3.                                        10 CFR 50.67.
 
PBAPS UNIT 2                                                                                                                                                                                                                                                                                                                                                  B 3.1-                                                  47                                                                                                                                                                                                        Revision No. 130 SDV Vent and Drain Valves B 3.1.8
 
B 3.1  REACTIVITY CONTROL SYSTEMS
 
B 3.1.8  Scram Discharge Volume (SDV) Vent and Drain Valves
 
BASES
 
BACKGROUND                                                                                                                                                                          The SDV vent and drain valves are normally open and discharge any accumulated water in the SDV to ensure that sufficient volume is available at all times to allow a complete scram. During a scram, the SDV vent and drain valves close to contain reactor water. As discussed in Reference 1, the SDV vent and drain valves need not be considered primary containment isolation valves (PCIVs) for the Scram Discharge System.  (However, at PBAPS, these valves are considered PCIVs.)  The SDV is a volume of header piping that connects to each hydraulic control unit (HCU) and drains into an instrument volume. There are                                                                                                                                                                                                                                                                                                                                                                                                                                                                                              two SDVs (headers) and a common instrument volume that receives all of the control rod drive (CRD) discharges. The instrument volume is connected to a common drain line with two valves in series. Each header is connected to a common vent line with two valves in series for a total of four vent valves.
The header piping is sized to receive and contain all the water discharged by the CRDs during a scram. The design and functions of the SDV are described in Reference                                                                                                                                                                                                                                                                                                                                                                                                                                                                                      2.
 
APPLICABLE                                                                                                                                                                          The Design Basis Accident and transient analyses assume all SAFETY ANALYSES                                                                                                                                                                          of the control rods are capable of scramming. The acceptance criteria for the SDV vent and drain valves are that they operate automatically to close during scram to limit the amount of reactor coolant discharged so that adequate core cooling is maintained and offsite doses remain within the limits of 10                                                                                                                                                                                                                                      CFR                              50.67 (Ref.                                                  3).
 
Isolation of the SDV can also be accomplished by                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                manual closure of the SDV valves. Additionally, the discharge of reactor coolant to the SDV can be terminated by scram reset or closure of the HCU manual isolation valves. For a bounding leakage case, the                                                                                                                                                                                                                                                                    offsite doses are well within the limits of 10 C                                                                                                                      FR                    50.67                                                  (Ref.                                                  3), and adequate core cooling is maintained (Ref.                                                                                                                                                                                              1). The SDV vent and drain valves allow continuous drainage of the SDV during normal plant operation to ensure that the SDV has sufficient                                                                                                                                                                                                                                                                                                                                                                                capacity to contain the reactor coolant discharge during a full core scram. To automatically ensure this capacity, a reactor scram (LCO                                        3.3.1.1, "Reactor Protection System (RPS)
Instrumentation") is initiated if the SDV water level in the
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                                                                                                                                                  B 3.1-                                                  48                                                                                                                                                                                                                  Revision No. 75 SDV Vent and Drain Valves B 3.1.8
 
BASES
 
APPLICABLE                                                                                                                                                                          instrument volume exceeds a specified setpoint. The SAFETY ANALYSES                                                                                                                                                                          setpoint is chosen so that all control rods are inserted (continued)                                                                                                                                                                          before the SDV has insufficient volume to accept a full scram.
 
SDV vent and drain valves satisfy Criterion                                                                                                                                                                                                                                                                                                                                                                                                                                            3 of the NRC Policy                                                            Statement.
 
LCO                                                                                                                                                                          The OPERABILITY of all SDV vent and drain valves ensures that the SDV vent and drain valves will close during a scram to contain reactor water discharged to the SDV piping.
Since the vent and drain lines are                                                                                                                                                                                                                                                                                                                                                  provided with two valves in series, the single failure of one valve in the open position will not impair the isolation function of the system. Additionally, the valves are required to be opened following scram reset to ensure that a path is available for the SDV piping to drain freely at other times.
 
APPLICABILITY                                                                                                                                                                          In MODES                                                                                1 and                                                  2, scram may be required; therefore, the SDV vent                                        and drain valves must be OPERABLE. In MODES                                                                                                                                                                                                                                                                                                                                                                                                                                                      3 and                                                  4, control rods are not able to be withdrawn since the reactor mode switch is in shutdown and a control rod block is applied. This provides adequate controls to ensure that only a single control rod can be withdrawn. Also, during MODE                                        5, only a single                                                                                                                                                                control rod can be withdrawn from a core cell containing fuel assemblies. Therefore, the SDV vent and drain valves are not required to be OPERABLE in these MODES since the reactor is subcritical and only one rod may be withdrawn and subject to scram.
 
ACTIONS                                                                                                                                                                          The ACTIONS Table is modified by Notes                                                                                                                                                                                                                                                                                                                                                                                indicating that a separate Condition entry is allowed for each SDV vent and drain line. This is acceptable, since the Required Actions for each Condition provide appropriate compensatory actions for each inoperable SDV line. Complying with the Required Actions may allow for continued operation, and subsequent inoperable SDV lines are governed by subsequent Condition entry and application of associated Required Actions.
 
When a line is isolated, the potential for an inadvertent scram due to high SDV level is increased. During these periods, the line may be unisolated under administrative control. This allows any accumulated water                                                                                                                                                                                                                                                                                                                                                                                                                                            in the line to be drained, to preclude a reactor scram on SDV high level.
This is acceptable since the administrative controls ensure the valve can be closed quickly, by a dedicated operator, if a scram occurs with the valve open.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                                                                                                                                                  B 3.1-                                                  49                                                                                                                                                                                                                  Revision No. 57 SDV Vent and Drain Valves B 3.1.8
 
BASES
 
ACTIONS                                                A.1 (continued)
When one SDV vent or drain valve is inoperable in one or more lines,                                                                                                              the associated line must be isolated to contain the reactor coolant during a scram. The 7 day Completion Time is reasonable, given the level of redundancy in the lines and the low probability of a scram occurring during the time the valves are inoperable                                                                                                                                                                                                                                                                                                                                                  and the line is not isolated. The SDV is still isolable since the redundant valve in the affected line is OPERABLE. During these periods, the single failure criterion may not be preserved, and a higher risk exists to allow reactor water out of the primary system during a scram.
 
B.1
 
If both valves in a line are inoperable, the line must be isolated to contain the reactor coolant during a scram.
 
The 8                                                  hour Completion Time to isolate the line is based on the low probability of a scram occurring while the line is not isolated and unlikelihood of significant CRD seal leakage.
 
C.1
 
If any Required Action and associated Completion Time is not met, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE                                                                                                                                                                                                                                                3 within 12                                                                                                              hours. The allowed Completion Time of 12                                                                                                                                                                                                        hours is reasonable, based on operating experience, to reach MODE                                                                                                                                                                                                                                                                                                                                                            3 from full power conditions in an orderly manner and without challenging plant systems.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                                                                                                                                                  B 3.1-                                                  50                                                                                                                                                                                                                  Revision No. 57 SDV Vent and Drain Valves B 3.1.8
 
BASES  (continued)
 
SURVEILLANCE                                  SR3.1.8.1 REQUIREMENTS During normal operation, the SDV vent and drain valves should be in the open position (except when performing SR                    3.1.8.2 or SR 3.3.1.1.9 for Function 13, Manual Scram, of Table 3.3.1.1-                                                                                                                                1) to allow for drainage of the SDV piping.
Verifying that each valve is in the open position ensures that the SDV vent and drain valves will perform their intended functions during normal operation. This SR does not require any testing or valve manipulation; rather, it involves verification that the valves are in the correct position.The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
 
SR3.1.8.2
 
During a scram, the SDV vent and drain valves should close to contain the reactor water discharged to the SDV piping.
Cycling each valve through its complete range of motion (closed and open) ensures that the valve will function properly during a scram. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
 
SR3.1.8.3
 
SR                    3.1.8.3 is an integrated test of the SDV vent and drain valves to verify total system performance. After receipt of a simulated or actual scram signal, the closure of the SDV vent and drain valves is verified. The                                                                                                                                                                                                                                                                                                                                                                                                    closure time of 15                    seconds after receipt of a scram signal is based on the bounding leakage case evaluated in the accident analysis (Ref.                                                  2). The LOGIC SYSTEM FUNCTIONAL TEST in LCO                                                                                                                                                                                                                                                                                                                                                                                                                                                        3.3.1.1 and                              the scram time testing of control rods in LCO                                                                                                                                                                                                                                                                                                                                                                                                                                                                  3.1.3 overlap this Surveillance to provide complete testing of the assumed safety function. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                                                                                                                                                  B 3.1-                                                  51                                                                                                                                                                                                                  Revision No. 86 SDV Vent and Drain Valves B 3.1.8
 
BASES  (continued)
 
REFERENCES                                                                                                                                                                          1.                                        NUREG-                                                0803, "Generic Safety Evaluation Report Regarding Integrity of BWR Scram System Piping,"
August                                                            1981.
: 2.                                        UFSAR, Sections                                                                                                                                                      3.4.5.3.1 and 7.2.3.6.
: 3.                                        10                    CFR                              50.67.
 
PBAPS UNIT 2                                                                                                                                                                                                                                                                                                                                                  B 3.1-                                                  52                                                                                                                                                                                                                  Revision No. 86 APLHGR B 3.2.1
 
B 3.2  POWER DISTRIBUTION LIMITS
 
B 3.2.1  AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)
 
BASES
 
BACKGROUND                                                                      The APLHGR is a measure of the average LHGR of all the fuel rods in a fuel assembly at any axial location. Limits on the APLHGR are specified to ensure that the peak cladding temperature (PCT) during the postulated design basis loss of coolant accident (LOCA) does not exceed the limits specified in 10 CFR 50.46.
 
APPLICABLE                                                                      The analytical methods and assumptions used in evaluating SAFETY ANALYSES                    Design Basis Accidents (DBAs) that determine the APLHGR limits are presented in References 1, 2, 3, 4, 5, and 7.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                                  B 3.2-1                                                                                                                                                                                                    Revision No. 49 APLHGR B 3.2.1
 
BASES
 
APPLICABLE SAFETY ANALYSES (continued)
 
LOCA analyses are performed to ensure that the APLHGR limits are adequate to meet the PCT and maximum oxidation limits of 10 CFR 50.46. The analysis is performed using calculational models that are consistent with the requirements of 10 CFR 50, Appendix K. A complete discussion of the analysis code is provided in Reference 11. The PCT following a postulated LOCA is a function of the average heat generation rate of all the rods of a fuel assembly at any axial location and is not strongly influenced by the rod to rod power distribution within an assembly. A conservative multiplier is applied to the LHGR assumed in the LOCA analysis to account for the uncertainty associated with the measurement of the APLHGR.
 
For single recirculation loop operation, a conservative multiplier is applied to the APLHGR as specified in the COLR (Ref. 12). This is due to the conservative analysis assumption of an earlier departure from nucleate boiling with one recirculation loop available, resulting in a more severe cladding heatup during a LOCA.
 
Power-dependent and flow-dependent APLHGR adjustment factors may also be provided per Reference 1 to ensure that fuel design limits are not exceeded due to the occurrence of a postulated transient event during operation at off-rated (less than 100%) reactor power or core flow conditions.
These adjustment factors are applied, if required, per the COLR and decrease the allowable APLHGR value.
 
The APLHGR satisfies Criterion 2 of the NRC Policy Statement.
 
LCO                                                                                                                                            The APLHGR limits specified in the COLR are the result of the fuel design and DBA analyses. The limits are developed as a function of exposure and are applied per the COLR.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                                  B 3.2-2                                                                                                                                                                                                    Revision No. 49 APLHGR B 3.2.1
 
BASES
 
LCO (continued)                                        With only one recirculation loop in operation, in conformance with the requirements of LCO 3.4.1, "Recirculation Loops Operating," the limit is determined by multiplying the exposure dependent APLHGR limit by a conservator factor.
 
APPLICABILITY                                        The APLHGR limits are primarily derived from LOCA analyses that are assumed to occur at high power levels. Design calculations (Ref. 6) and operating experience have shown that as power is reduced, the margin to the required APLHGR limits increases. This trend continues down to the power range of 5% to 15% RTP when entry into MODE 2 occurs. When in MODE 2, the wide range neutron monitor period-short scram function provides prompt scram initiation during any significant transient, thereby effectively removing any APLHGR limit compliance concern in MODE 2. Therefore, at THERMAL POWER levels < 22.6% RTP, the reactor is operating with substantial margin to the APLHGR limits; thus, this LCO is not required.
 
ACTIONS A.1
 
If any APLHGR exceeds the required limits, an assumption regarding an initial condition of the DBA analyses may not be met. Therefore, prompt action should be taken to restore the APLHGR(s) to within the required limits such that the plant operates within analyzed conditions and within design limits of the fuel rods. The 2 hour Completion Time is sufficient to restore the APLHGR(s) to within its limits and is acceptable based on the low probability of a DBA occurring simultaneously with the APLHGR out of specification.
 
B.1
 
If the APLHGR cannot be restored to within its required limits within the associated Completion Time, the plant must be brought to a MODE or other specified condition in which the LCO does not apply. To achieve this status, THERMAL POWER must be reduced to < 22.6% RTP within 4 hours. The
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                                  B 3.2-3                                                                                                                                                                                          Revision No. 143 APLHGR B 3.2.1
 
BASES
 
ACTIONS                                                                                                    B.1  (continued)
 
allowed Completion Time is reasonable, based on operating experience, to reduce THERMAL POWER to < 22.6% RTP in an orderly manner and without challenging plant systems.
 
SURVEILLANCE                                                  SR  3.2.1.1 REQUIREMENTS APLHGRs are required to be initially calculated within 12 hours after THERMAL POWER is                                                                                                                                                                                                                                                                                                                                  22.6% RTP and then periodically thereafter. They are compared to the specified limits in the COLR to ensure that the reactor is operating within the assumptions of the safety analysis. The 12 hour allowance after THERMAL POWER                                                                                                                                                                                                                                                                                                              22.6% RTP is achieved is acceptable given the large inherent margin to operating limits at low power levels. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
 
REFERENCES                                                                      1.                    NEDO-24011-P-A, "General Electric Standard Application for Reactor Fuel," latest approved revision.
: 2.                    UFSAR, Chapter 3.
: 3.                    UFSAR, Chapter 6.
: 4.                    UFSAR, Chapter 14.
: 5.                    NEDO-24229-1, "Peach Bottom Atomic Power Station Units 2 and 3, Single Loop Operation," May 1980.
: 6.                    NEDC-32162P, "Maximum Extended Load Line Limit and ARTS Improvement Program Analyses for Peach Bottom Atomic Power Station Units 2 and 3," Revision 2, March 1995.
: 7.                    NEDC-33566P, "Safety Analysis Report for Exelon Peach Bottom Atomic Power Station, Units 2 and 3, Constant Pressure Power Uprate," Revision 0.
: 8. Deleted
: 9.                    NEDO-30130-A, "Steady State Nuclear Methods,"
April 1985.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                                  B 3.2-4                                                                                                                                                                                                Revision No. 143 APLHGR B 3.2.1
 
BASES
 
REFERENCES 10. Deleted (continued)
: 11.                              NEDC-32163P, "Peach Bottom Atomic Power Station Units 2 and 3 SAFER/GESTR-LOCA Loss-of-Coolant Accident Analysis," January 1993.
: 12.                              Peach Bottom Unit 2 Core Operating Limits Report (COLR).
: 13.                              NEDC-33873P, Safety Analysis Report for Peach Bottom Atomic Power Station, Units 2 and 3, Thermal Power Optimization, Revision 0.
 
PBAPS UNIT 2                                                                                                                                                                                                                                  B 3.2-5                                                                                                                                                                                          Revision No. 143 MCPR B 3.2.2
 
B 3.2  POWER DISTRIBUTION LIMITS
 
B 3.2.2  MINIMUM CRITICAL POWER RATIO (MCPR)
 
BASES
 
BACKGROUND                                                                      MCPR is a ratio of the fuel assembly power that would result in the onset of boiling transition to the actual fuel assembly power. The operating limit MCPR is established to ensure that no fuel damage results during abnormal operational transients, and that 99.9% of the fuel rods are not susceptible to boiling transition if the limit is not violated. Although fuel damage does not necessarily occur if a fuel rod actually experienced boiling transition (Ref. 1), the critical power at which boiling transition is calculated to occur has been adopted as a fuel design criterion.
 
The onset of transition boiling is a phenomenon that is readily detected during the testing of various fuel bundle designs. Based on these experimental data, correlations have been developed to predict critical bundle power (i.e.,
the bundle power level at the onset of transition boiling) for a given set of plant parameters (e.g., reactor vessel pressure, flow, and subcooling). Because plant operating conditions and bundle power levels are monitored and determined relatively easily, monitoring the MCPR is a convenient way of ensuring that fuel failures due to inadequate cooling do not occur.
 
APPLICABLE                                                                      The analytical methods and assumptions used in evaluating SAFETY ANALYSES                    the abnormal operational transients to establish the operating limit MCPR are presented in References 2, 3, 4, 5, 6, 7, 8, and 9. To ensure that the MCPR Safety Limit (SL) is not exceeded during any transient event that occurs with moderate frequency, limiting transients have been analyzed to determine the largest reduction in critical power ratio (CPR). The types of transients evaluated are loss of flow, increase in pressure and power, positive reactivity insertion, and coolant temperature decrease. The limiting transient yields the largest change in CPR (                                                                                                                                                                                                                                                                                                                                                                                                                                                        CPR). When the largest                                                                                CPR (corrected for analytical uncertainties) is combined with the MCPR99.9%, the required operating limit MCPR is obtained.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                                  B 3.2-6                                                                                                                                                                                          Revision No. 157 MCPR B 3.2.2 BASES
 
APPLICABLE                                                                      MCPR99.9% is determined to ensure more than 99.9% of the SAFETY ANALYSES          fuel rods in the core are not susceptible to boiling (continued)    transition using a statistical model that combines all the uncertainties in operating parameters and the procedures used to calculate critical power. The probability of the occurrence of boiling transition is determined using the approved Critical Power correlations. Details of the MCPR99.9% calculation are given in Reference 2. Reference 2 also includes a tabulation of the uncertainties and the nominal values of the parameters used in the MCPR99.9% statistical analysis.
The MCPR operating limits are derived from the MCPR                                                                                                                                                                                                                                                                                                                                                                                                                                                                                    99.9% value and transient analysis, and are dependent on the operating ensure adherence to fuel design limits during the worst core flow and power  state (MCPRf and MCPRp, respectively) to transient that occurs with moderate frequency (Refs. 6, 7, 8, and 9). Flow dependent MCPR limits are determined by steady state thermal hydraulic methods with key physics response inputs benchmarked using the three dimensional BWR simulator code (Ref. 10) to analyze slow flow runout transients. The flow dependent operating limit, MCPRf, is evaluated based on a single recirculation pump flow runout event (Ref. 9).
Power dependent MCPR limits (MCPRp) are determined by approved transient analysis models (Reference 2). Due to the sensitivity of the transient response to initial core flow levels at power levels below those at which the turbine stop valve closure and turbine control valve fast closure scrams are bypassed, high and low flow MCPRp operating limits are provided for operating between 22.6% RTP and the previously mentioned bypass power level.
The MCPR satisfies Criterion 2 of the NRC Policy Statement.
 
LCO                                                                                                                                            The MCPR operating limits specified in the COLR value, MCPRf values, and MCPRp values) (MCPR                                                                                                                                                                                                                                                                                                                                                                                                                                                                                              99.9% are the result of the Design Basis Accident (DBA) and transient analysis. The operating limit MCPR is determined by the larger of the MCPRf specified in the COLR. and MCPRp limits, which are based on the MCPR                                                                                        99.9% limit
 
APPLICABILITY                                        The MCPR operating limits are primarily derived from transient analyses that are assumed to occur at high power levels. Below 22.6% RTP, the reactor is operating at a minimum recirculation pump speed and the moderator void ratio is small. Surveillance of thermal limits below 22.6% RTP is unnecessary due to the large inherent margin that ensures that the MCPR SL is not exceeded even if a limiting transient occurs. Statistical analyses indicate that the nominal value of the initial MCPR expected at 22.6% RTP is > 3.5. Studies of the variation of limiting transient behavior have been performed over the range of power and (continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                                  B 3.2-7                                                                                                                                                                                          Revision No. 157 MCPR B 3.2.2
 
BASES
 
APPLICABILITY                                        flow conditions. These studies encompass the range of key (continued)                                        actual plant parameter values important to typically limiting transients. The results of these studies demonstrate that a margin is expected between performance and the MCPR requirements, and that margins increase as power is reduced to 22.6% RTP. This trend is expected to continue to the 5% to 15% power range when entry into MODE 2 occurs. When in MODE 2, the wide range neutron monitor period-short function provides rapid scram initiation for any significant power increase transient, which effectively eliminates any MCPR compliance concern. Therefore, at THERMAL POWER levels < 22.6% RTP, the reactor is operating with substantial margin to the MCPR limits and this LCO is not required.
 
ACTIONS A.1
 
If any MCPR is outside the required limits, an assumption regarding an initial condition of the design basis transient analyses may not be met. Therefore, prompt action should be taken to restore the MCPR(s) to within the required limits such that the plant remains operating within analyzed conditions. The 2 hour Completion Time is normally sufficient to restore the MCPR(s) to within its limits and is acceptable based on the low probability of a transient or DBA occurring simultaneously with the MCPR out of specification.
 
B.1
 
If the MCPR cannot be restored to within its required limits within the associated Completion Time, the plant must be brought to a MODE or other specified condition in which the LCO does not apply. To achieve this status, THERMAL POWER must be reduced to < 22.6% RTP within 4 hours. The allowed Completion Time is reasonable, based on operating experience, to reduce THERMAL POWER to < 22.6% RTP in an orderly manner and without challenging plant systems.
 
SURVEILLANCE                                                  SR  3.2.2.1 REQUIREMENTS The MCPR is required to be initially calculated within 12 hours after THERMAL POWER is                                                                                                                                                                                                                                                                                                                                  22.6% RTP and periodically thereafter. It is compared to the specified limits
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                                  B 3.2-8                                                                                                                                                                                                Revision No. 143 MCPR B 3.2.2
 
BASES
 
SURVEILLANCE                                                  SR  3.2.2.1  (continued)
REQUIREMENTS in the COLR (Ref. 12) to ensure that the reactor is operating within the assumptions of the safety analysis. The 12 hour allowance after THERMAL POWER                                                                                                                                                                                                                                                                                                              22.6% RTP is achieved is acceptable given the large inherent margin to operating limits at low power levels. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
 
SR  3.2.2.2
 
Because the transient analysis takes credit for conservatism in the scram speed performance, it must be demonstrated that the specific scram speed distribution is consistent with that used in the transient analysis. SR 3.2.2.2 determines the value of                                                                                                                                  , which is a measure of the actual scram speed distribution compared with the assumed distribution. The MCPR operating limit is then determined based on an interpolation between the applicable limits for Option A (scram times of LCO 3.1.4,"Control Rod Scram Times") and Option B (realistic scram times) analyses. The parameter must be determined once within 72 hours after each set of scram time tests required by SR 3.1.4.1, SR 3.1.4.2, and SR 3.1.4.4 because the effective scram speed distribution may change during the cycle or after maintenance that could affect scram times. The 72 hour Completion Time is acceptable due to the relatively minor changes in                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                      expected during the fuel cycle.
 
REFERENCES                                                                      1.                    NUREG-0562, June 1979.
: 2.                    NEDO-24011-P-A, "General Electric Standard Application for Reactor Fuel," latest approved revision.
: 3.                    UFSAR, Chapter 3.
: 4.                    UFSAR, Chapter 6.
: 5.                    UFSAR, Chapter 14.
: 6.                    NEDO-24229-1, "Peach Bottom Atomic Power Station Units 2 and 3, Single Loop Operation," May 1980.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                                  B 3.2-9                                                                                                                                                                                          Revision No. 143 MCPR B 3.2.2
 
BASES
 
REFERENCES                                                                      7.                      NEDC-32162P, "Maximum Extended Load Line Limit and (continued)                                                                                  ARTS Improvement Program Analyses for Peach Bottom Atomic Power Station Units 2 and 3," Revision 2, March 1995.
: 8.                      NEDC-33566P, "Safety Analysis Report for Exelon Peach Bottom Atomic Power Station, Units 2 and 3, Constant Pressure Power Uprate," Revision 0.
: 9.                      NEDC-32428P, "Peach Bottom Atomic Power Station Unit 2 Cycle 11 ARTS Thermal Limits Analyses," December 1994.
: 10.                      NEDO-30130-A, "Steady State Nuclear Methods,"
April 1985.
: 11.                      NEDO-24154, "Qualification of the One-Dimensional Core Transient Model for Boiling Water Reactors,"
October 1978.
: 12.                      Peach Bottom Unit 2 Core Operating Limits Report (COLR).
: 13.                      NEDC-33873P, Safety Analysis Report for Peach Bottom Atomic Power Station, Units 2 and 3, Thermal Power Optimization, Revision 0.
 
PBAPS UNIT 2                                                                                                                                                                                                                            B 3.2-10                                                                                                                                                                                    Revision No. 143 LHGR B 3.2.3
 
B 3.2  POWER DISTRIBUTION LIMITS
 
B 3.2.3  LINEAR HEAT GENERATION RATE (LHGR)
 
BASES
 
BACKGROUND                                                                      The LHGR is a measure of the heat generation rate of a fuel rod in a fuel assembly at any axial location. Limits on LHGR are specified to ensure that fuel design limits are not exceeded anywhere in the core during normal operation, including abnormal operational transients. Exceeding the LHGR limit could potentially result in fuel damage and subsequent release of radioactive materials. Fuel design limits are specified to ensure that fuel system damage, fuel rod failure, or inability to cool the fuel does not occur during the anticipated operating conditions identified in Reference 1.
 
APPLICABLE                                                                      The analytical methods and assumptions used in evaluating SAFETY ANALYSES                    the fuel system design are presented in References 1, 2, 3, 4, 5, 6, 7, 8, 11, and 12. The fuel assembly is designed to ensure (in conjunction with the core nuclear and thermal hydraulic design, plant equipment, instrumentation, and protection system) that fuel damage will not result in the release of radioactive materials in excess of the guidelines of 10 CFR, Parts 20, 50, and 100, as applicable. The mechanisms that could cause fuel damage during operational transients and that are considered in fuel evaluations are:
: a.                    Rupture of the fuel rod cladding caused by strain from the relative expansion of the UO2 pellet; and
: b.                    Severe overheating of the fuel rod cladding caused by inadequate cooling.
 
A value of 1% plastic strain of the fuel cladding has been defined as the limit below which fuel damage caused by overstraining of the fuel cladding is not expected to occur (Ref. 9).
 
Fuel design evaluations have been performed and demonstrate that the 1% fuel cladding plastic strain design limit is not exceeded during continuous operation with LHGRs up to the operating limit specified in the COLR. The analysis also
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                            B 3.2-11                                                                                                                                                                                    Revision No. 101 LHGR B 3.2.3
 
BASES
 
APPLICABLE                                                                      includes allowances for short term transient operation above SAFETY ANALYSES                    the operating limit to account for abnormal operational (continued)                                        transients, plus an allowance for densification power spiking.
 
Power-dependent and flow-dependent LHGR adjustment factors may also be provided per Reference 1 to ensure that fuel design limits are not exceeded due to the occurrence of a postulated transient event during operation at off-rated (less than 100%) reactor power or core flow conditions.
These adjustment factors are applied, if required, per the COLR and decrease the allowable LHGR value.
 
Additionally, for single recirculation loop operation, an LHGR multiplier may be provided per Reference 1. This multiplier is applied per the COLR and decreases the allowable LHGR value. This additional margin may be necessary during SLO to account for the conservative analysis assumption of an earlier departure from nucleate boiling with only one recirculation loop available.
 
The LHGR satisfies Criterion 2 of the NRC Policy Statement.
 
LCO                                                                                                                                            The LHGR is a basic assumption in the fuel design analysis.
The fuel has been designed to operate at rated core power with sufficient design margin to the LHGR calculated to cause a 1% fuel cladding plastic strain. The operating limit to accomplish this objective is specified in the COLR.
 
APPLICABILITY                                        The LHGR limits are derived from fuel design analysis that is limiting at high power level conditions. At core thermal power levels < 22.6% RTP, the reactor is operating with a substantial margin to the LHGR limits and, therefore, the Specification is only required when the reactor is operating at                                22.6% RTP.
 
ACTIONS A.1
 
If any LHGR exceeds its required limit, an assumption regarding an initial condition of the fuel design analysis is not met. Therefore, prompt action should be taken to restore the LHGR(s) to within its required limits such that the plant is operating within analyzed conditions. The
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                            B 3.2-12                                                                                                                                                                                    Revision No. 143 LHGR B 3.2.3
 
BASES
 
ACTIONS A.1 (continued)
 
2 hour Completion Time is normally sufficient to restore the LHGR(s) to within its limits and is acceptable based on the low probability of a transient or Design Basis Accident occurring simultaneously with the LHGR out of specification.
 
B.1
 
If the LHGR cannot be restored to within its required limits within the associated Completion Time, the plant must be brought to a MODE or other specified condition in which the LCO does not apply. To achieve this status, THERMAL POWER is reduced to < 22.6% RTP within 4 hours. The allowed Completion Time is reasonable, based on operating experience, to reduce THERMAL POWER TO < 22.6% RTP in an orderly manner and without challenging plant systems.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                        B 3.2-12a                                                                                                                                                                                Revision No. 143 LHGR B 3.2.3
 
BASES  (continued)
 
SURVEILLANCE                                                  SR  3.2.3.1 REQUIREMENTS The LHGR is required to be initially calculated within 12 hours after THERMAL POWER is                                                                                                                                                                                                                                                                                                                                  22.6% RTP and periodically thereafter. It is compared to the specified limits in the COLR (Ref. 10) to ensure that the reactor is operating within the assumptions of the safety analysis. The 12 hour allowance after THERMAL POWER                                                                                                                                                                                                                                                                                                              22.6% RTP is achieved is acceptable given the large inherent margin to operating limits at lower power levels. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
 
REFERENCES                                                            1.                              NEDO-24011-P-A, "General Electric Standard Application for Reactor Fuel," latest approved revision.
: 2.                              UFSAR, Chapter 3.
: 3.                              UFSAR, Chapter 6.
: 4.                              UFSAR, Chapter 14.
: 5.                              NEDO-24229-1, "Peach Bottom Atomic Power Station Units 2 and 3, Single-Loop Operation," May 1980.
: 6.                              NEDC-32162P, "Maximum Extended Load Line Limit and ARTS Improvements Program Analyses for Peach Bottom Atomic Power Station Units 2 and 3," Revision 2, March 1995.
: 7.                              NEDC-33566P, "Safety Analysis Report for Exelon Peach Bottom Atomic Power Station, Units 2 and 3, Constant Pressure Power Uprate," Revision 0.
: 8.                              NEDC-32163P, "Peach Bottom Atomic Power Station Units 2 and 3 SAFER/GESTR-LOCA Loss-of-Coolant Accident Analysis," January 1993.
: 9.                              G-080-VC-400, Peach Bottom Atomic Power Station Units 2
                                                                  & 3 GNF2 ECCS-LOCA Evaluation, GE Hitachi Nuclear Energy, 0000-0100-8531-R1, March 2011.
: 10.                              NUREG-0800, Section 4.2, Subsection II.A.2(g),
Revision 2, July 1981.
: 11.                              Peach Bottom Unit 2 Core Operating Limits Report (COLR).
: 12.                              005N9630, Peach Bottom Atomic Power Station Units 2 & 3 GNF3 ECCS-LOCA Evaluation, GE Hitachi Nuclear Energy, April 2021.
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                            B 3.2-13                                                                                                                                                                                    Revision No. 169 LHGR B 3.2.3
 
BASES  (continued)
 
REFERENCES                                                  13.                              NEDC-33873P, "Safety Analysis Report for Peach Bottom Atomic Power Station, Units 2 and 3, Thermal Power Optimization, Revision 0.
 
PBAPS UNIT 2                                                                                                                                                                                                                        B 3.2-13a                                                                                                                                                                                Revision No. 143 RPS Instrumentation B 3.3.1.1 a
 
B 3.3  INSTRUMENTATION
 
B 3.3.1.1  Reactor Protection System (RPS) Instrumentation
 
BASES
 
BACKGROUND                                                                      The RPS initiates a reactor scram when one or more monitored parameters exceed their specified limits, to preserve the integrity of the fuel cladding and the Reactor Coolant System (RCS) and minimize the energy that must be absorbed following a loss of coolant accident (LOCA). This can be accomplished either automatically or manually.
 
The protection and monitoring functions of the RPS have been designed to ensure safe operation of the reactor. This is achieved by specifying limiting safety system settings (LSSS) in terms of parameters directly monitored by the RPS, as well as LCOs on other reactor system parameters and equipment performance. The LSSS are defined in this Specification as the Allowable Values, which, in conjunction with the LCOs, establish the threshold for protective system action to prevent exceeding acceptable limits, including Safety Limits (SLs) during Design Basis Accidents (DBAs).
 
The RPS, as shown in the UFSAR Section 7.2, (Ref. 1),
includes sensors, relays, bypass circuits, and switches that are necessary to cause initiation of a reactor scram.
Functional diversity is provided by monitoring a wide range of dependent and independent parameters. The input parameters to the scram logic are from instrumentation that monitors reactor vessel water level, reactor vessel pressure, neutron flux, main steam line isolation valve position, turbine control valve (TCV) fast closure trip oil pressure, turbine stop valve (TSV) position, drywell pressure, scram discharge volume (SDV) water level, condenser vacuum, as well as reactor mode switch in shutdown position, manual scram signals, and RPS test switches.
There are at least four redundant sensor input signals from each of these parameters (with the exception of the manual scram signal and the reactor mode switch in shutdown scram signal). Most channels include electronic equipment (e.g.,
trip units) that compares measured input signals with pre-established setpoints. When the setpoint is exceeded, the channel output relay actuates, which then outputs an RPS trip signal to the trip logic.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                                  B 3.3-1                                                                                                                                                                                          Revision No. 134
 
RPS Instrumentation B 3.3.1.1
 
BASES
 
BACKGROUND                                                                      The RPS is comprised of two independent trip systems (continued)                                        (A and B) with three logic channels in each trip system (logic channels A1, A2, and A3; B1, B2, and B3) as shown in the Reference 1 figures. Logic channels A1, A2, B1, and B2 contain automatic logic for which the above monitored parameters each have at least one input to each of these logic channels. The outputs of the logic channels in a trip system are combined in a one-out-of-two logic so that either channel can trip the associated trip system. The tripping of both trip systems will produce a reactor scram. This logic arrangement is referred to as a one-out-of-two taken twice logic. In addition to the automatic logic channels, logic channels A3 and B3 (one logic channel per trip system) are manual scram channels. Both must be depressed in order to initiate the manual trip function. Each trip system can be reset by use of a reset switch. If a full scram occurs (both trip systems trip), a relay prevents reset of the trip systems for 10 seconds after the full scram signal is received. This 10 second delay on reset ensures that the scram function will be completed.
 
Two scram pilot valves are located in the hydraulic control unit for each control rod drive (CRD). Each scram pilot valve is solenoid operated, with the solenoids normally energized. The scram pilot valves control the air supply to the scram inlet and outlet valves for the associated CRD.
When either scram pilot valve solenoid is energized, air pressure holds the scram valves closed and, therefore, both scram pilot valve solenoids must be de-energized to cause a control rod to scram. The scram valves control the supply and discharge paths for the CRD water during a scram. One of the scram pilot valve solenoids for each CRD is controlled by trip system A, and the other solenoid is controlled by trip system B. Any trip of trip system A in conjunction with any trip in trip system B results in de-energizing both solenoids, air bleeding off, scram valves opening, and control rod scram.
 
The backup scram valves, which energize on a scram signal to depressurize the scram air header, are also controlled by the RPS. Additionally, the RPS controls the SDV vent and drain valves such that when logic channels A1 and B1 are deenergized or when logic channel A3 is deenergized the
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                                  B 3.3-2                                                                                                                                                                                                              Revision No. 0 RPS Instrumentation B 3.3.1.1
 
BASES
 
BACKGROUND                                                                      inboard SDV vent and drain valves close to isolate the SDV, (continued)                                        and when logic channels A2 and B2 are deenergized or when logic channel B3 is deenergized the outboard SDV vent and drain valves close to isolate the SDV.
 
APPLICABLE                                                                      The actions of the RPS are assumed in the safety analyses of SAFETY ANALYSES,          References 2 and 3. The RPS is required to initiate a LCO, and                                                                                          reactor scram when monitored parameter values exceed the APPLICABILITY                                        Allowable Values, specified by the setpoint methodology and listed in Table 3.3.1.1-1, to maintain OPERABILITY and to preserve the integrity of the fuel cladding, the reactor coolant pressure boundary (RCPB), and the containment, by minimizing the energy that must be absorbed following a LOCA.
 
RPS instrumentation satisfies Criterion 3 of the NRC Policy Statement. Functions not specifically credited in the accident analysis are retained for the overall redundancy and diversity of the RPS as required by the NRC approved licensing basis.
 
The OPERABILITY of the RPS is dependent on the OPERABILITY of the individual instrumentation channel Functions specified in Table 3.3.1.1-1. Each Function must have a required number of OPERABLE channels per RPS trip system, with their setpoints within the specified Allowable Value, where appropriate. The actual setpoint is calibrated consistent with applicable setpoint methodology assumptions.
 
Allowable Values, where applicable, are specified for each RPS Function specified in the Table. Trip setpoints are specified in the setpoint calculations. The trip setpoints are selected to ensure that the actual setpoints do not exceed the Allowable Value between successive CHANNEL CALIBRATIONS. Operation with a trip setting less conservative than the trip setpoint, but within its Allowable Value, is acceptable. A channel is inoperable if its actual trip setting is not within its required Allowable Value.
 
Trip setpoints are those predetermined values of output at which an action should take place. The setpoints are compared to the actual process parameter (e.g., reactor vessel water level), and when the measured output value of
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                                  B 3.3-3                                                                                                                                                                                                              Revision No. 0 RPS Instrumentation B 3.3.1.1
 
BASES
 
APPLICABLE                                                                      the process parameter exceeds the setpoint, the associated SAFETY ANALYSES,          device (e.g., trip unit) changes state. The analytic or LCO, and                                                                                          design limits are derived from the limiting values of the APPLICABILITY                                        process parameters obtained from the safety analysis or (continued)                                        other appropriate documents. The Allowable Values are derived from the analytic or design limits, corrected for calibration, process, and instrument errors. The trip setpoints are determined from analytical or design limits, corrected for calibration, process, and instrument errors, as well as instrument drift. In selected cases, the Allowable Values and trip setpoints are determined by engineering judgement or historically accepted practice relative to the intended function of the trip channel. The trip setpoints determined in this manner provide adequate protection by assuring instrument and process uncertainties expected for the environments during the operating time of the associated trip channels are accounted for.
 
The OPERABILITY of scram pilot valves and associated solenoids, backup scram valves, and SDV valves, described in the Background section, are not addressed by this LCO.
 
The individual Functions are required to be OPERABLE in the MODES or other specified conditions specified in the Table, which may require an RPS trip to mitigate the consequences of a design basis accident or transient. To ensure a reliable scram function, a combination of Functions are required in each MODE to provide primary and diverse initiation signals.
 
The only MODES specified in Table 3.3.1.1-1 are MODES 1 and 2, and MODE 5 with any control rod withdrawn from a core cell containing one or more fuel assemblies. No RPS Function is required in MODES 3 and 4, since all control rods are fully inserted and the Reactor Mode Switch Shutdown Position control rod withdrawal block (LCO 3.3.2.1) does not allow any control rod to be withdrawn. In MODE 5, control rods withdrawn from a core cell containing no fuel assemblies do not affect the reactivity of the core and, therefore, are not required to have the capability to scram.
Provided all other control rods remain inserted, no RPS function is required. In this condition, the required SDM (LCO 3.1.1) and refuel position one-rod-out interlock (LCO 3.9.2) ensure that no event requiring RPS will occur.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                                  B 3.3-4                                                                                                                                                                                                              Revision No. 0 RPS Instrumentation B 3.3.1.1
 
BASES
 
APPLICABLE                                                                      The specific Applicable Safety Analyses, LCO, and SAFETY ANALYSES,          Applicability discussions are listed below on a Function by LCO, and                                                                                          Function basis.
APPLICABILITY (continued)
Wide Range Neutron Monitor (WRNM)
 
1.a. Wide Range Neutron Monitor Period-Short
 
The WRNMs provide signals to facilitate reactor scram in the event that core reactivity increase (shortening period) exceeds a predetermined reference rate. To determine the reactor period, the neutron flux signal is filtered. The period of this filtered neutron flux signal is used to generate trip signals when the respective trip setpoints are exceeded. The time to trip for a particular reactor period is dependent on the filter time constant, actual period of the signal and the trip setpoints. This period based signal is available over the entire operating range from initial control rod withdrawal to full power operation. In the startup range, the most significant source of reactivity change is due to control rod withdrawal. The WRNM provides diverse protection from the rod worth minimizer (RWM), which monitors and controls the movement of control rods at low power. The RWM prevents the withdrawal of an out of sequence control rod during startup that could result in an unacceptable neutron flux excursion (Ref. 2). The WRNM provides mitigation of the neutron flux excursion. To demonstrate the capability of the WRNM System to mitigate control rod withdrawal events, an analysis has been performed (Ref. 3) to evaluate the consequences of control rod withdrawal events during startup that are mitigated only by the WRNM period-short function. The withdrawal of a control rod out of sequence, during startup, analysis (Ref.
: 3) assumes that one WRNM channel in each trip system is bypassed, demonstrates that the WRNMs provide protection against local control rod withdrawal errors and results in peak fuel enthalpy below the 170 cal/gm fuel failure threshold criterion.
 
The WRNMs are also capable of limiting other reactivity excursions during startup, such as cold water injection events, although no credit is specifically assumed.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                                  B 3.3-5                                                                                                                                                                                                    Revision No. 24
 
RPS Instrumentation B 3.3.1.1
 
BASES
 
APPLICABLE                                                                      1.a. Wide Range Neutron Monitor Period-Short SAFETY ANALYSES,          (continued)
LCO, and APPLICABILITY                                        The WRNM System is divided into two groups of WRNM channels, with four channels inputting to each trip system. The analysis of Reference 3 assumes that one channel in each trip system is bypassed. Therefore, six channels with three channels in each trip system are required for WRNM OPERABILITY to ensure that no single instrument failure will preclude a scram from this Function on a valid signal.
 
The analysis of Reference 3 has adequate conservatism to permit an Allowable Value of 13 seconds.
 
The WRNM Period-Short Function must be OPERABLE during MODE 2 when control rods may be withdrawn and the potential for criticality exists. In MODE 5, when a cell with fuel has its control rod withdrawn, the WRNMs provide monitoring for and protection against unexpected reactivity excursions.
In MODE 1, the APRM System and the RWM provide protection against control rod withdrawal error events and the WRNMs are not required. The WRNMs are automatically bypassed when the mode switch is in the Run position.
 
1.b. Wide Range Neutron MonitorInop
 
This trip signal provides assurance that a minimum number of WRNMs are OPERABLE. Anytime a WRNM mode switch is moved to any position other than "Operate," a loss of power occurs, or the self-test system detects a failure which would result in the loss of a safety-related function, an inoperative trip signal will be received by the RPS unless the WRNM is bypassed. Since only one WRNM in each trip system may be bypassed, only one WRNM in each RPS trip system may be inoperable without resulting in an RPS trip signal.
 
This Function was not specifically credited in the accident analysis but it is retained for the overall redundancy and diversity of the RPS as required by the NRC approved licensing basis.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                                  B 3.3-6                                                                                                                                                                                                    Revision No. 24 RPS Instrumentation B 3.3.1.1
 
BASES
 
APPLICABLE                                                                      1.b. Wide Range Neutron MonitorInop  (continued)
SAFETY ANALYSES, LCO, and                                                                                          Six channels of the Wide Range Neutron Monitor      Inop APPLICABILITY                                        Function, with three channels in each trip system, are required to be OPERABLE to ensure that no single instrument failure will preclude a scram from this Function on a valid signal. Since this Function is not assumed in the safety analysis, there is no Allowable Value for this Function.
This Function is required to be OPERABLE when the Wide Range Neutron Monitor Period-Short Function is required.
 
Average Power Range Monitor (APRM)
The APRM channels provide the primary indication of neutron flux within the core and respond almost instantaneously to neutron flux increases. The APRM channels receive input signals from the local power range monitors (LPRMs) within the reactor core to provide an indication of the power distribution and local power changes. The APRM channels average these LPRM signals to provide a continuous indication of average reactor power from a few percent to greater than RTP. Each APRM also includes an Oscillation Power Range Monitor (OPRM) Upscale Function which monitors small groups of LPRM signals to detect thermal-hydraulic instabilities.
The APRM System is divided into four APRM channels and four 2-out-of-4 voter channels. Each APRM channel provides inputs to each of the four voter channels. The four voter channels are divided into two groups of two each, with each group of two providing inputs to one RPS trip system. The system is designed to allow one APRM channel, but no voter channels, to be bypassed. A trip from any one unbypassed APRM will result in a half-trip in all four of the voter channels, but no trip inputs to either RPS trip system.
APRM trip Functions 2.a, 2.b, 2.c, and 2.d are voted independently from OPRM Upscale Function 2.f. Therefore, any Function 2.a, 2.b, 2.c, or 2.d trip from any two unbypassed APRM channels will result in a full trip in each of the four voter channels, which in turn results in two trip inputs into each RPS trip system logic channel (A1, A2, B1, and B2), thus resulting in a full scram signal.
Similarly, a Function 2.f trip from any two unbypassed APRM channels will result in a full trip from each of the four voter channels. Three of the four APRM channels and all four of the voter channels are required to be OPERABLE to ensure that no single failure will preclude a scram on a valid signal. In addition, to provide adequate coverage of the entire core, consistent with the design bases for the APRM Functions 2.a, 2.b, and 2.c, at least 20 LPRM inputs, with at least three LPRM inputs from each of the four axial levels at which the LPRMs are located, must be operable for each APRM channel, and the number of LPRM inputs that have become inoperable (and bypassed) since the last APRM calibration (SR 3.3.1.1.2) must be less than ten for each APRM channel. For the OPRM Upscale, Function 2.f, LPRMs are assigned to cells of 3 or 4 detectors. A minimum of 8 cells per channel, each with a minimum of 2 OPERABLE LPRMs, must be OPERABLE for the OPRM Upscale Function 2.f to be (continued) OPERABLE.
 
PBAPS UNIT 2                                                                                                                                                                                                                                  B 3.3-7                                                                                                                                                                                          Revision No. 124 RPS Instrumentation B 3.3.1.1
 
BASES
 
APPLICABLE                                                                      2.a. Average Power Range Monitor Neutron Flux-High SAFETY ANALYSES,          (Setdown)  (continued)
LCO, and APPLICABILITY                                        For operation at low power (i.e., MODE 2), the Average Power capable of generating a trip signal that prevents fuel Range Monitor Neutron Flux-High (Setdown) Function is damage resulting from abnormal operating transients in this power range. For most operation at low power levels, the Average Power Range Monitor Neutron Flux-High (Setdown)
Function will provide a secondary scram to the Wide Range Neutron Monitor Period-Short Function because of the relative setpoints. At higher power levels, it is possible that the Average Power Range Monitor Neutron Flux-High (Setdown) Function will provide the primary trip signal for a corewide increase in power.
No specific safety analyses take direct credit for the Average Power Range Monitor Neutron Flux-High (Setdown)
Function. However, this Function indirectly ensures that before the reactor mode switch is placed in the run position, reactor power does not exceed 22.6% RTP (SL 2.1.1.1) when operating at low reactor pressure and low core flow. Therefore, it indirectly prevents fuel damage during significant reactivity increases with THERMAL POWER
                                  < 22.6% RTP.
The Allowable Value is based on preventing significant increases in power when THERMAL POWER is < 22.6% RTP.
 
The Average Power Range Monitor Neutron Flux-High (Setdown) Function must be OPERABLE during MODE 2 when control rods may be withdrawn since the potential for criticality exists.
In MODE 1, the Average Power Range Monitor Neutron Flux-High Function provides protection against reactivity transients and the RWM and rod block monitor protect against control rod withdrawal error events.
2.b. Average Power Range Monitor Simulated Thermal Power-High
 
The Average Power Range Monitor Simulated Thermal Power-High Function monitors average neutron flux to approximate the THERMAL POWER being transferred to the reactor coolant. The APRM neutron flux is electronically filtered with a time constant representative of the fuel heat transfer dynamics to generate a signal proportional to the THERMAL POWER in the reactor. The trip level is varied as a function of recirculation drive flow (i.e., at lower core flows, the setpoint is reduced proportional to the reduction in power experienced as core flow is reduced with a fixed control rod pattern) but is clamped at an upper limit that is always lower than  the Average Power Range Monitor Neutron Flux-High Function Allowable Value. A note is included, applicable when the plant is in single recirculation loop operation per LCO 3.4.1, which requires the flow value, used in the Allowable Value equation, be reduced by                                                                                                                                                                                                                                                                                                                                                                                                                W. The value of                                                                                                                                                                W
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                                  B 3.3-8                                                                                                                                                                                          Revision No. 143 RPS Instrumentation B 3.3.1.1
 
BASES
 
APPLICABLE                                                                      2.b. Average Power Range Monitor Simulated Thermal SAFETY ANALYSES,          Power-High  (continued)
LCO, and APPLICABILITY is established to conservatively bound the inaccuracy created in the core flow/drive flow correlation due to back flow in the jet pumps associated with the inactive recirculation loop. The Allowable Value thus maintains thermal margins essentially unchanged from those for two loop operation. The value of                                                                                          W is plant specific and is defined in plant procedures. The Allowable Value equation for single loop operation is only valid for flows down to W =                                                                                                                                                                                                                                                                                                                                                                                                                                                                            W; the Allowable Value does not go below 60.3% RTP. This is acceptable because back flow in the inactive recirculation loop is only evident with drive flows of approximately 35%
or greater (Reference 19). The Nominal Trip Setpoint (NTSP) and the as-found and as-left tolerances (Leave Alone Zone) were determined in accordance with Reference 10.
 
The Average Power Range Monitor Simulated Thermal Power-High Function is not specifically credited in the safety analysis but is intended to provide an additional margin of protection from transient induced fuel damage during operation where recirculation flow is reduced to below the minimum required for rated power operation. The Average Power Range Monitor Simulated Thermal Power-High Function provides protection against transients where THERMAL POWER increases slowly (such as the loss of feedwater heating event) and protects the fuel cladding integrity by ensuring that the MCPR SL is not exceeded. During these events, the THERMAL POWER increase does not significantly lag the neutron flux scram. For rapid neutron flux increase events, the THERMAL POWER lags the neutron flux and the Average Power Range Monitor Neutron Flux-High Function will provide a scram signal before the Average Power Range Monitor Simulated Thermal Power-High Function setpoint is exceeded.
 
Each APRM channel uses one total drive flow signal representative of total core flow. The total drive flow signal is generated by the flow processing logic, part of the APRM channel, by summing up the flow calculated from two flow transmitter signal inputs, one from each of the two recirculation loop flows. The flow processing logic OPERABILITY is part of the APRM channel OPERABILITY requirements for this Function. The APRM flow processing logic is considered inoperable whenever it cannot deliver a flow signal less than or equal to actual Recirculation flow conditions for all steady state and transient reactor conditions while in Mode 1. Reduced or Downscale flow conditions due to planned maintenance or testing activities during derated plant conditions (i.e. end of cycle coast down) will result in conservative setpoints for the APRM Simulated Thermal Power-High function, thus maintaining that function operable.
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                                  B 3.3-9                                                                                                                                                                                          Revision No. 143 RPS Instrumentation B 3.3.1.1
 
BASES
 
APPLICABLE                                                                      2.b. Average Power Range Monitor Simulated Thermal SAFETY ANALYSES,          Power-High  (continued)
LCO, and APPLICABILITY                                        The Allowable Value is based on analyses that take credit for the Average Power Range Monitor Simulated Thermal Power-High Function for the mitigation of non-limiting events.
The THERMAL POWER time constant of                                                                                                                                                                                                                                                                                                                                                                7 seconds is based on the fuel heat transfer dynamics and provides a signal proportional to the THERMAL POWER.
 
The Average Power Range Monitor Simulated Thermal Power-High Function is required to be OPERABLE in MODE 1 when there is the possibility of generating excessive THERMAL POWER and potentially exceeding the SL applicable to high pressure and core flow conditions (MCPR SL). During MODES 2 and 5, other WRNM and APRM Functions provide protection for fuel cladding integrity.
 
2.c. Average Power Range Monitor Neutron Flux-High
 
The Average Power Range Monitor Neutron Flux-High Function is capable of generating a trip signal to prevent fuel damage or excessive RCS pressure. For the overpressurization protection analysis of Reference 4, the Average Power Range Monitor Neutron Flux-High Function is assumed to terminate the main steam isolation valve (MSIV) closure event and, along with the safety/relief valves (S/RVs), limit the peak reactor pressure vessel (RPV) pressure to less than the ASME Code limits. The control rod drop accident (CRDA) analysis (Ref. 5) takes credit for the Average Power Range Monitor Neutron Flux-High Function to terminate the CRDA.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                            B 3.3-10                                                                                                                                                                                              Revision No. 36
 
RPS Instrumentation B 3.3.1.1
 
BASES
 
APPLICABLE                                                                      2.c. Average Power Range Monitor Neutron Flux-High SAFETY ANALYSES,          (continued)
LCO, and APPLICABILITY                                        The Allowable Value is based on the Analytical Limit assumed in the CRDA analysis.
 
The Average Power Range Monitor Neutron Flux-High Function is required to be OPERABLE in MODE 1 where the potential consequences of the analyzed transients could result in the SLs (e.g., MCPR and RCS pressure) being exceeded. Although the Average Power Range Monitor Neutron Flux-High Function is assumed in the CRDA analysis, which is applicable in MODE 2, the Average Power Range Monitor Neutron Flux-High (Setdown) Function conservatively bounds the assumed trip and, together with the assumed WRNM trips, provides adequate protection. Therefore, the Average Power Range Monitor Neutron Flux-High Function is not required in MODE 2.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                            B 3.3-11                                                                                                                                                                                              Revision No. 36 RPS Instrumentation B 3.3.1.1
 
BASES
 
APPLICABLE                                                                                2.d. Average Power Range MonitorInop SAFETY ANALYSES, LCO, and                                                                                                    Three of the four APRM channels are required to be OPERABLE APPLICABILITY                                                  for each of the APRM Functions. This Function (Inop)
(continued)                                                  provides assurance that the minimum number of APRM channels are OPERABLE.
 
For any APRM channel, any time its mode switch is not in the Operate position, an APRM module required to issue a trip is unplugged, or the automatic self-test system detects a critical fault with the APRM channel, an Inop trip is sent to all four voter channels. Inop trips from two or more unbypassed APRM channels result in a trip output from each of the four voter channels to its associated trip system.
This Function was not specifically credited in the accident analysis, but it is retained for the overall redundancy and diversity of the RPS as required by the NRC approved licensing basis.
There is no Allowable Value for this Function.
This Function is required to be OPERABLE in the MODES where the APRM Functions are required.
2.e. 2-Out-Of-4 Voter The 2-Out-Of-4 Voter Function provides the interface between the APRM Functions, including the OPRM Upscale Function, and the final RPS trip system logic. As such, it is required to be OPERABLE in the MODES where the APRM Functions are required and is necessary to support the safety analysis applicable to each of those Functions. Therefore, the 2-Out-Of-4 Voter Function needs to be OPERABLE in MODES 1 and 2.
All four voter channels are required to be OPERABLE. Each voter channel includes self-diagnostic functions. If any voter channel detects a critical fault in its own processing, a trip is issued from that voter channel to the associated trip system.
The 2-Out-Of-4 Logic Module includes 2-Out-Of-4 Voter hardware and the APRM Interface hardware. The 2-Out-Of-4 Voter Function 2.e votes APRM Functions 2.a, 2.b, 2.c and 2.d independently of Function 2.f. This voting is accomplished by the 2-Out-Of-4 Voter hardware in the 2-Out-Of-4 Logic Module.
Each 2-Out-Of-4 Voter includes two redundant sets of outputs to RPS. Each output set contains two independent contacts; one contact for Functions 2.a, 2.b, 2.c and 2.d, and the other contact for Function 2.f. The analysis in Reference 12 took credit for this redundancy in the justification of the 12-hour Completion Time for Condition A, so the voter Function 2.e must be declared inoperable if any of its functionality is inoperable. However, the voter Function 2.e does not need to be declared inoperable due to any failure affecting only the plant interface portions of the 2-Out-Of-4 Logic Module that are not necessary to perform the 2-Out-Of-4 Voter function.
There is no Allowable Value for this Function.
(continued)
PBAPS UNIT 2                                                                                                                                                                                                                                                B 3.3-12                                                                                                                                                                                                                  Revision No. 50 RPS Instrumentation B 3.3.1.1
 
BASES
 
APPLICABLE                                                                                2.f. Oscillation Power Range Monitor (OPRM) Upscale SAFETY ANALYSES, LCO, and                                                                                                    The OPRM Upscale Function provides compliance with 10 CFR APPLICABILITY                                                  50, Appendix A, General Design Criteria (GDC) 10 and 12, (continued)                                                  thereby providing protection from exceeding the fuel MCPR safety limit (SL) due to anticipated thermal-hydraulic power oscillations.
 
Reference 22 describes the Detect and Suppress-Confirmation Density (DSS-CD) long-term stability solution and the licensing basis Confirmation Density Algorithm (CDA).
Reference 22 also describes the DSS-CD Armed Region and the three additional algorithms for detecting thermal-hydraulic instability related neutron flux oscillations: the period based detection algorithm (PBDA), the amplitude based algorithm (ABA), and the growth rate algorithm (GRA). All four algorithms are implemented in the OPRM Upscale Function, but the safety analysis takes credit only for the CDA. The remaining three algorithms provide defense-in-depth and additional protection against unanticipated oscillations.
OPRM Upscale Function OPERABILITY is based only on the CDA.
 
The OPRM Upscale Function receives input signals from the local power range monitors (LPRMs) within the reactor core, which are combined into cells for evaluation by the OPRM algorithms.
 
DSS-CD operability requires at least 8 responsive OPRM cells per channel. The DSS-CD software includes a self-check for the responsive OPRM cells; therefore, no SR is necessary.
 
The OPRM Upscale Function is required to be OPERABLE when the plant is            17.6% RTP, which is established as a power level that is greater than or equal to 5% below the lower boundary of the Armed Region. This requirement is designed to encompass the region of power-flow operation where anticipated events could lead to thermal-hydraulic instability and related neutron flux oscillations. The OPRM Upscale Function is automatically trip-enabled when THERMAL POWER, as indicated by the APRM Simulated Thermal Power, is            22.6% RTP corresponding to the MCPR monitoring threshold and reactor recirculation drive flow, is less than 75% of rated flow. This region is the OPRM Armed Region. Note (h) allows for entry into the DSS-CD Armed Region without automatic arming of DSS-CD prior to completely passing through the DSS-CD Armed Region during the first startup and the first shutdown following DSS-CD implementation.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                                            B 3.3-12a                                                                                                                                                                                                    Revision No. 143 RPS Instrumentation B 3.3.1.1
 
BASES
 
APPLICABLE                                                                                2.f. Oscillation Power Range Monitor (OPRM)
SAFETY ANALYSES,                    Upscale (continued)
LCO, and APPLICABILITY                                                  As described in Reference 22 and 24, the RTP values for the OPRM Upscale Function to be OPERABLE (          17.6% RTP) and for the OPRM Upscale Function to be auto-enabled (          22.6% RTP) are sufficiently conservative for protection of the plant against thermal-hydraulic instabilities. The basis for the 5% RTP difference between the OPRM Upscale OPERABLE (17.6%
RTP) and OPRM Upscale auto-enable value (22.6% RTP) is to ensure that no credible event, e.g., loss of feed water heating, could result in a plant power excursion where an inoperable OPRM channel entered into the OPRM Armed Region.
Peach Bottom plant specific analyses performed at these low power levels (Reference 24) have demonstrated that any power excursion resulting from credible events is bounded by 5%
RTP. In addition, both the core-wide and channel decay ratios at the OPRM Upscale auto-enabled values are extremely low as documented in Reference 22, which demonstrates the low possibility of thermal-hydraulic instabilities at low power and confirms the conservatisms in the OPRM Upscale Function auto-enable RTP value. The conservatisms in the determination of the values for OPRM Upscale Function OPERABLE and the OPRM Upscale Function auto enabled sufficiently compensate for possible inaccuracy of the APRM simulated thermal power signal versus actual core thermal power at power levels < 22.6% RTP. Therefore, there is no need to perform any calibration of the APRM simulated thermal power signal to calculated power with RTP < 22.6% in order to determine the OPRM Upscale Function OPERABLE.
 
If any OPRM auto-enable setpoint is in a non-conservative condition, i.e., the OPRM Upscale is not auto-enabled with RTP            22.6% and reactor recirculation drive flow            75% of rated, the associated channel is considered inoperable for the OPRM Upscale function. Alternatively, the auto-enable setpoint may be adjusted to place the channel in a conservative condition (armed). If placed in the armed condition, the channel is considered OPERABLE.
 
Note (h) reflects the need for plant data collection in order to test the DSS-CD equipment. Testing the DSS-CD equipment ensures its proper operation and prevents spurious reactor trips. Entry into the DSS-CD Armed Region without automatic arming of DSS-CD during this initial testing phase also allows for changes in plant operations to address maintenance or other operational needs. However, during this initial testing period, the OPRM Upscale Function is OPERABLE and DSS-CD operability and capability to automatically arm shall be maintained at recirculation drive flow rates above the DSS-CD Armed Region flow boundary.
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                                            B 3.3-12b                                                                                                                                                                                                    Revision No. 143 RPS Instrumentation B 3.3.1.1
 
BASES
 
APPLICABLE                                                                                2.f. Oscillation Power Range Monitor (OPRM)
SAFETY ANALYSES,                    Upscale (continued)
LCO, and APPLICABILITY                                                  An OPRM Upscale trip is issued from an OPRM channel when the confirmation density algorithm in that channel detects oscillatory changes in the neutron flux, indicated by periodic confirmations and amplitude exceeding specified setpoints for a specified number of OPRM cells in the channel. An OPRM Upscale trip is also issued from the channel if any of the defense-in-depth algorithms (PBDA, ABA, GRA) exceed their trip condition for one or more cells in that channel.
 
Three of the four channels are required to be operable. Each channel is capable of detecting thermal-hydraulic instabilities, by detecting the related neutron flux oscillations, and issuing a trip signal before the SLMCPR is exceeded. There is no Allowable Value for this function.
 
The OPRM Upscale Function is not LSSS SL-related (Ref. 22) and Reference 23 confirms that the OPRM Upscale Function settings based on DSS-CD also do not have traditional instrumentation setpoints determined under an instrument setpoint methodology.
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                                            B 3.3-12c                                                                                                                                                                                                    Revision No. 123 RPS Instrumentation B 3.3.1.1
 
BASES
 
APPLICABLE                                                                      3. Reactor PressureHigh SAFETY ANALYSES, LCO, and                                                                                          An increase in the RPV pressure during reactor operation APPLICABILITY                                        compresses the steam voids and results in a positive (continued)                                        reactivity insertion. This causes the neutron flux and THERMAL POWER transferred to the reactor coolant to increase, which could challenge the integrity of the fuel cladding and the RCPB. No specific safety analysis takes direct credit for this Function. However, the Reactor Pressure      High Function initiates a scram for transients that result in a pressure increase, counteracting the pressure increase by rapidly reducing core power. For the overpressurization protection analysis of Reference 4, the Reactor Pressure      High Function is credited as a backup Scram Function only. The analyses conservatively assume the scram occurs on the Average Power Range Monitor Scram Clamp signal, not the Reactor Pressure      High signal. The reactor scram, along with the S/RVs, limits the peak RPV pressure to less than the ASME Section III Code limits.
 
High reactor pressure signals are initiated from four pressure transmitters that sense reactor pressure. The Reactor Pressure      High Allowable Value is chosen to provide a sufficient margin to the ASME Section III Code limits during the event.
 
Four channels of Reactor Pressure      High Function, with two channels in each trip system arranged in a one-out-of-two logic, are required to be OPERABLE to ensure that no single instrument failure will preclude a scram from this Function on a valid signal. The Function is required to be OPERABLE in MODES 1 and 2 when the RCS is pressurized and the potential for pressure increase exists.
: 4. Reactor Vessel Water LevelLow (Level 3)
 
Low RPV water level indicates the capability to cool the fuel may be threatened. Should RPV water level decrease too far, fuel damage could result. Therefore, a reactor scram is initiated at Level 3 to substantially reduce the heat generated in the fuel from fission. The Reactor Vessel Water Level      Low (Level 3) Function is assumed in the analysis of events resulting in the decrease of reactor coolant inventory (Ref. 6). This is credited as a backup scram function for large and intermediate break LOCAs inside
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                            B 3.3-13                                                                                                                                                                                                        Revision No. 0 RPS Instrumentation B 3.3.1.1
 
BASES
 
APPLICABLE                                                                      4. Reactor Vessel Water LevelLow (Level 3)  (continued)
SAFETY ANALYSES, LCO, and                                                                                          primary containment. The reactor scram reduces the amount APPLICABILITY                                        of energy required to be absorbed and, along with the actions of the Emergency Core Cooling Systems (ECCS),
ensures that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46.
 
Reactor Vessel Water Level      Low (Level 3) signals are initiated from four level transmitters that sense the difference between the pressure due to a constant column of water (reference leg) and the pressure due to the actual water level (variable leg) in the vessel.
 
Four channels of Reactor Vessel Water Level      Low (Level 3)
Function, with two channels in each trip system arranged in a one-out-of-two logic, are required to be OPERABLE to ensure that no single instrument failure will preclude a scram from this Function on a valid signal.
 
The Reactor Vessel Water Level      Low (Level 3) Allowable Value is selected to ensure that during normal operation the separator skirts are not uncovered (this protects available recirculation pump net positive suction head (NPSH) from significant carryunder) and, for transients involving loss of all normal feedwater flow, initiation of the low pressure ECCS subsystems at Reactor Vessel Water      Low Low Low (Level 1) will not be required.
 
The Function is required in MODES 1 and 2 where considerable energy exists in the RCS resulting in the limiting transients and accidents. ECCS initiations at Reactor Vessel Water Level      Low Low (Level 2) and Low Low Low (Level 1) provide sufficient protection for level transients in all other MODES.
: 5. Main Steam Isolation ValveClosure
 
MSIV closure results in loss of the main turbine and the condenser as a heat sink for the nuclear steam supply system and indicates a need to shut down the reactor to reduce heat generation. Therefore, a reactor scram is initiated on a Main Steam Isolation Valve      Closure signal before the MSIVs are completely closed in anticipation of the complete loss of the normal heat sink and subsequent overpressurization
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                            B 3.3-14                                                                                                                                                                                                        Revision No. 0 RPS Instrumentation B 3.3.1.1
 
BASES
 
APPLICABLE                                                                      5. Main Steam Isolation ValveClosure  (continued)
SAFETY ANALYSES, LCO, and                                                                                          transient. However, for the overpressurization protection APPLICABILITY                                        analysis of Reference 4, the Average Power Range Monitor Scram Clamp Function, along with the S/RVs, limits the peak RPV pressure to less than the ASME Section III Code limits.
That is, the direct scram on position switches for MSIV closure events is not assumed in the overpressurization analysis. The reactor scram reduces the amount of energy required to be absorbed and, along with the actions of the ECCS, ensures that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46.
 
MSIV closure signals are initiated from position switches located on each of the eight MSIVs. Each MSIV has two position switches; one inputs to RPS trip system A while the other inputs to RPS trip system B. Thus, each RPS trip system receives an input from eight Main Steam Isolation Valve      Closure channels, each consisting of one position switch. The logic for the Main Steam Isolation Valve      Closure Function is arranged such that either the inboard or outboard valve on three or more of the main steam lines must close in order for a scram to occur. In addition, certain combinations of valves closed in two lines will result in a half-scram.
 
The Main Steam Isolation Valve      Closure Allowable Value is specified to ensure that a scram occurs prior to a significant reduction in steam flow, thereby reducing the severity of the subsequent pressure transient.
 
Eight channels of the Main Steam Isolation Valve      Closure Function, with four channels in each trip system, are required to be OPERABLE to ensure that no single instrument failure will preclude the scram from this Function on a valid signal. This Function is only required in MODE 1 since, with the MSIVs open and the heat generation rate high, a pressurization transient can occur if the MSIVs close. In MODE 2, the heat generation rate is low enough so that the other diverse RPS functions provide sufficient protection.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                            B 3.3-15                                                                                                                                                                                                        Revision No. 0 RPS Instrumentation B 3.3.1.1
 
BASES
 
APPLICABLE                                                                      6. Drywell PressureHigh SAFETY ANALYSES, LCO, and                                                                                          High pressure in the drywell could indicate a break in the APPLICABILITY                                        RCPB. A reactor scram is initiated to minimize the (continued)                                        possibility of fuel damage and to reduce the amount of energy being added to the coolant and the drywell. The Drywell Pressure      High Function is assumed to scram the reactor during large and intermediate break LOCAs inside primary containment. The reactor scram reduces the amount of energy required to be absorbed and, along with the actions of the ECCS, ensures that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46.
 
High drywell pressure signals are initiated from four pressure transmitters that sense drywell pressure. The Allowable Value was selected to be as low as possible and indicative of a LOCA inside primary containment.
 
Four channels of Drywell Pressure      High Function, with two channels in each trip system arranged in a one-out-of-two logic, are required to be OPERABLE to ensure that no single instrument failure will preclude a scram from this Function on a valid signal. The Function is required in MODES 1 and 2 where considerable energy exists in the RCS, resulting in the limiting transients and accidents.
: 7. Scram Discharge Volume Water LevelHigh
 
The SDV receives the water displaced by the motion of the CRD pistons during a reactor scram. Should this volume fill to a point where there is insufficient volume to accept the displaced water, control rod insertion would be hindered.
Therefore, a reactor scram is initiated while the remaining free volume is still sufficient to accommodate the water from a full core scram. No credit is taken for a scram initiated from the Scram Discharge Volume Water Level      High Function for any of the design basis accidents or transients analyzed in the UFSAR. However, this function is retained to ensure the RPS remains OPERABLE.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                            B 3.3-16                                                                                                                                                                                                        Revision No. 0 RPS Instrumentation B 3.3.1.1
 
BASES
 
APPLICABLE                                                                      7. Scram Discharge Volume Water LevelHigh  (continued)
SAFETY ANALYSES, LCO, and                                                                                          SDV water level is measured by two diverse methods. The APPLICABILITY                                        level is measured by two float type level switches and two thermal probes for a total of four level signals. The outputs of these devices are arranged so that one switch provides input to one RPS logic channel. The level measurement instrumentation satisfies the recommendations of Reference 8.
 
The Allowable Value is chosen low enough to ensure that there is sufficient volume in the SDV to accommodate the water from a full scram.
 
Four high water level inputs to the RPS from four switches are required to be OPERABLE, with two switches in each trip system, to ensure that no single instrument failure will preclude a scram from this Function on a valid signal. This Function is required in MODES 1 and 2, and in MODE 5 with any control rod withdrawn from a core cell containing one or more fuel assemblies, since these are the MODES and other specified conditions when control rods are withdrawn. At all other times, this Function may be bypassed.
: 8. Turbine Stop ValveClosure
 
Closure of the TSVs results in the loss of a heat sink that produces reactor pressure, neutron flux, and heat flux transients that must be limited. Therefore, a reactor scram is initiated at the start of TSV closure in anticipation of the transients that would result from the closure of these valves. The Turbine Stop Valve      Closure Function is the primary scram signal for the turbine trip event analyzed in Reference 7 and the feedwater controller failure event. For these events, the reactor scram reduces the amount of energy required to be absorbed and ensures that the MCPR SL is not exceeded.
 
Turbine Stop Valve      Closure signals are initiated from four position switches; one located on each of the four TSVs.
Each switch provides two input signals; one to RPS trip system A and the other, to RPS trip system B. Thus, each RPS trip system receives an input from four Turbine Stop Valve      Closure channels. The logic for the Turbine Stop
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                            B 3.3-17                                                                                                                                                                                              Revision No. 87 RPS Instrumentation B 3.3.1.1
 
BASES
 
APPLICABLE                                                                      8. Turbine Stop ValveClosure  (continued)
SAFETY ANALYSES, LCO, and                                                                                          Valve      Closure Function is such that three or more TSVs must APPLICABILITY                                        be closed to produce a scram. In addition, certain combinations of two valves closed will result in a half-scram. This Function must be enabled at THERMAL POWER 26.3% RTP as measured at the turbine first stage pressure.
This is normally accomplished automatically by pressure switches sensing turbine first stage pressure; therefore, opening of the turbine bypass valves may affect this Function.
 
The Turbine Stop Valve      Closure Allowable Value is selected to be high enough to detect imminent TSV closure, thereby reducing the severity of the subsequent pressure transient.
 
Eight channels of Turbine Stop Valve      Closure Function, with four channels in each trip system, are required to be OPERABLE to ensure that no single instrument failure will preclude a scram from this Function if any three TSVs should close. This Function is required, consistent with analysis assumptions, whenever THERMAL POWER is                                                                                                                                                                                                                                                                                                                                                                                                        26.3% RTP. This Function is not required when THERMAL POWER is < 26.3% RTP since the Reactor Pressure      High and the Average Power Range Monitor Scram Clamp Functions are adequate to maintain the necessary safety margins.
: 9. Turbine Control Valve Fast Closure, Trip Oil PressureLow
 
Fast closure of the TCVs results in the loss of a heat sink that produces reactor pressure, neutron flux, and heat flux transients that must be limited. Therefore, a reactor scram is initiated on TCV fast closure in anticipation of the transients that would result from the closure of these valves. The Turbine Control Valve Fast Closure, Trip Oil Pressure      Low Function is the primary scram signal for the generator load rejection event analyzed in Reference 7 and the generator load rejection with bypass failure event. For these events, the reactor scram reduces the amount of energy required to be absorbed and ensures that the MCPR SL is not exceeded.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                            B 3.3-18                                                                                                                                                                                    Revision No. 143 RPS Instrumentation B 3.3.1.1
 
BASES
 
APPLICABLE                                                                      9. Turbine Control Valve Fast Closure, Trip Oil SAFETY ANALYSES,          PressureLow  (continued)
LCO, and APPLICABILITY                                        Turbine Control Valve Fast Closure, Trip Oil Pressure      Low signals are initiated by the relayed emergency trip supply oil pressure at each control valve. One pressure switch is associated with each control valve, and the signal from each switch is assigned to a separate RPS logic channel. This Function must be enabled at THERMAL POWER                                                                                                                                                                                                                                                                                                                                                                                                                                      26.3% RTP. This is normally accomplished automatically by pressure switches sensing turbine first stage pressure; therefore, opening of the turbine bypass valves may affect this Function.
 
The Turbine Control Valve Fast Closure, Trip Oil Pressure      Low Allowable Value is selected high enough to detect imminent TCV fast closure.
 
Four channels of Turbine Control Valve Fast Closure, Trip Oil Pressure      Low Function with two channels in each trip system arranged in a one-out-of-two logic are required to be OPERABLE to ensure that no single instrument failure will preclude a scram from this Function on a valid signal. This Function is required, consistent with the analysis assumptions, whenever THERMAL POWER is                                                                                                                                                                                                                                                                                                                                                                                                        26.3% RTP. This Function is not required when THERMAL POWER is < 26.3% RTP, since the Reactor Pressure      High and the Average Power Range Monitor Scram Clamp Functions are adequate to maintain the necessary safety margins.
: 10. Turbine CondenserLow Vacuum
 
The Turbine Condenser      Low Vacuum Function protects the integrity of the main condenser by scramming the reactor and thereby decreasing the severity of the low condenser vacuum transient on the condenser. This function also ensures integrity of the reactor due to loss of its normal heat sink. The reactor scram on a Turbine Condenser      Low Vacuum signal will occur prior to a reactor scram from a Turbine Stop Valve      Closure signal. This function is not specifically credited in any accident analysis but is being retained for the overall redundancy and diversity of the RPS as required by the NRC approved licensing basis.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                            B 3.3-19                                                                                                                                                                                    Revision No. 143 RPS Instrumentation B 3.3.1.1
 
BASES
 
APPLICABLE                                                                      10. Turbine CondenserLow Vacuum  (continued)
SAFETY ANALYSES, LCO, and                                                                                          Turbine Condenser      Low Vacuum signals are initiated from APPLICABILITY                                        four vacuum pressure transmitters that provide inputs to associated trip systems. There are two trip systems and two channels per trip system. Each trip system is arranged in a one-out-of-two logic and both trip systems must be tripped in order to scram the reactor.
 
The Turbine Condenser      Low Vacuum Allowable Value is specified to ensure that a scram occurs prior to the integrity of the main condenser being breached, thereby limiting the damage to the normal heat sink of the reactor.
 
Four channels of the Turbine Condenser      Low Vacuum Function with two channels in each trip system, are required to be OPERABLE to ensure that no single instrument failure will preclude a scram from this function on a valid signal. This Function is only required in MODE 1 where considerable energy exists which could challenge the integrity of the main condenser if vacuum is low. In MODE 2, the Turbine Condenser      Low Vacuum Function is not required because at low power levels the transients are less severe.
: 11. Deleted
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                            B 3.3-20                                                                                                                                                                                    Revision No. 134 RPS Instrumentation B 3.3.1.1
 
BASES
 
APPLICABLE                                                                      12. Reactor Mode SwitchShutdown Position SAFETY ANALYSES, LCO, and                                                                                          The Reactor Mode Switch      Shutdown Position Function provides APPLICABILITY                                signals, via the manual scram logic channels, directly to the scram pilot solenoid power circuits. These manual scram logic channels are redundant to the automatic protective instrumentation channels and provide manual reactor trip capability. This Function was not specifically credited in the accident analysis, but it is retained for the overall redundancy and diversity of the RPS as required by the NRC approved licensing basis.
 
The reactor mode switch is a keylock four-position, four-bank switch. The reactor mode switch is capable of scramming the reactor if the mode switch is placed in the shutdown position. Scram signals from the mode switch are input into each of the two RPS manual scram logic channels.
 
There is no Allowable Value for this Function, since the channels are mechanically actuated based solely on reactor mode switch position.
 
Two channels of Reactor Mode Switch      Shutdown Position Function, with one channel in each manual scram trip system, are available and required to be OPERABLE. The Reactor Mode Switch      Shutdown Position Function is required to be OPERABLE in MODES 1 and 2, and MODE 5 with any control rod withdrawn from a core cell containing one or more fuel assemblies, since these are the MODES and other specified conditions when control rods are withdrawn.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                            B 3.3-21                                                                                                                                                                                    Revision No. 134 RPS Instrumentation B 3.3.1.1
 
BASES
 
APPLICABLE                                                                      13. Manual Scram SAFETY ANALYSES, LCO, and                                                                                          The Manual Scram push button channels provide signals, via APPLICABILITY                                        the manual scram logic channels, directly to the scram pilot (continued)                                        solenoid power circuits. These manual scram logic channels are redundant to the automatic protective instrumentation channels and provide manual reactor trip capability. This Function was not specifically credited in the accident analysis but it is retained for the overall redundancy and diversity of the RPS as required by the NRC approved licensing basis.
 
There is one Manual Scram push button channel for each of the two RPS manual scram logic channels. In order to cause a scram it is necessary that each channel in both manual scram trip systems be actuated.
 
There is no Allowable Value for this Function since the channels are mechanically actuated based solely on the position of the push buttons.
 
Two channels of Manual Scram with one channel in each manual scram trip system are available and required to be OPERABLE in MODES 1 and 2, and in MODE 5 with any control rod withdrawn from a core cell containing one or more fuel assemblies, since these are the MODES and other specified conditions when control rods are withdrawn.
: 14. RPS Channel Test Switch
 
There are four RPS Channel Test Switches, one associated with each of the four automatic scram logic channels (A1, A2, B1, and B2). These keylock switches allow the operator to test the OPERABILITY of each individual logic channel without the necessity of using a scram function trip. This is accomplished by placing the RPS Channel Test Switch in test, which will input a trip signal into the associated RPS logic channel. The RPS Channel Test Switches were not specifically credited in the accident analysis. However, because the Manual Scram Functions at Peach Bottom Atomic Power Station, were not configured the same as the generic model in Reference 9, the RPS Channel Test Switches were included in the analysis in Reference 10. Reference 10 concluded that the Surveillance Frequency extensions from
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                            B 3.3-22                                                                                                                                                                                                        Revision No. 0 RPS Instrumentation B 3.3.1.1
 
BASES
 
APPLICABLE                                                                      14. RPS Channel Test Switch  (continued)
SAFETY ANALYSES, LCO, and                                                                                          RPS Functions, described in Reference 9, were not affected APPLICABILITY                                        by the difference in configuration, since each automatic RPS channel has a test switch which is functionally the same as the manual scram switches in the generic model. As such, the RPS Channel Test Switches are retained in the Technical Specifications.
 
There is no Allowable Value for this Function since the channels are mechanically actuated based solely on the position of the switches.
 
Four channels of RPS Channel Test Switch with two channels in each trip system arranged in a one-out-of-two logic are available and required to be OPERABLE in MODES 1 and 2, and in MODE 5 with any control rod withdrawn from a core cell containing one or more fuel assemblies, since these are the MODES and other specified conditions when control rods are withdrawn.
 
ACTIONS                                                                                                    Note 1 has been provided to modify the ACTIONS related to RPS instrumentation channels. 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 inoperable RPS instrumentation channels provide appropriate compensatory measures for separate inoperable channels. As such, a Note has been provided that allows separate Condition entry for each inoperable RPS instrumentation channel.
 
Note 2 has been provided to modify the ACTIONS for the RPS instrumentation functions of APRM Flow Biased Neutron-Flux High (Function 2.b) and APRM Fixed Neutron Flux-High (Function 2.c) when they are inoperable due to failure of SR 3.3.1.1.2 and gain adjustments are necessary. Note 2 allows entry into associated Conditions and Required Actions to be delayed for up to 2 hours if the APRM is indicating a lower power value than the calculated power (i.e., the gain adjustment factor (GAF) is high (non-conservative)). The GAF for any channel is defined as the power value determined
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                            B 3.3-23                                                                                                                                                                                    Revision No. 149
 
RPS Instrumentation B 3.3.1.1
 
BASES
 
ACTIONS (continued)
 
by the heat balance divided by the APRM reading for that channel. Upon completion of the gain adjustment, or expiration of the allowed time, the channel must be returned to OPERABLE status or the applicable Condition entered and the Required Actions taken. This Note is based on the time required to perform gain adjustments on multiple channels.
 
A.1 and A.2
 
Because of the diversity of sensors available to provide trip signals and the redundancy of the RPS design, an allowable out of service time of 12 hours has been shown to be acceptable (Refs. 9, 12 & 13) to permit restoration of any inoperable channel to OPERABLE status. However, this out of service time is only acceptable provided the associated
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                        B 3.3-23a                                                                                                                                                                      Revision No. 149
 
RPS Instrumentation B 3.3.1.1 BASES
 
ACTIONS                                                                                                    A.1 and A.2 (continued)
 
Function's inoperable channel is in one trip system and the Function still maintains RPS trip capability (refer to Required Actions B.1, B.2, and C.1 Bases). Alternatively, a Completion Time can be determined in accordance with the Risk Informed Completion Time (RICT) Program. A Note has been provided to indicate that a RICT is only applicable when a loss of function has not occurred.
If the inoperable channel cannot be restored to OPERABLE status within the allowable out of service time, the channel or the associated trip system must be placed in the tripped condition per Required Actions A.1 and A.2. Placing the inoperable channel in trip (or the associated trip system in trip) would conservatively compensate for the inoperability, restore capability to accommodate a single failure, and allow operation to continue. Alternatively, if it is not desired to place the channel (or trip system) in trip (e.g.,
as in the case where placing the inoperable channel in trip would result in a full scram), Condition D must be entered and its Required Action taken.
 
As noted, Action A.2 is not applicable for APRM Functions 2.a, 2.b, 2.c, 2.d, or 2.f. Inoperability of one required APRM channel affects both trip systems. For that condition, Required Action A.1 must be satisfied, and is the only action (other than restoring operability) that will restore capability to accommodate a single failure. Inoperability of more than one required APRM channel of the same trip function results in loss of trip capability and entry into Condition C, as well as entry into Condition A for each channel.
 
B.1 and B.2
 
Condition B exists when, for any one or more Functions, at least one required channel is inoperable in each trip system. In this condition, provided at least one channel per trip system is OPERABLE, the RPS still maintains trip capability for that Function, but cannot accommodate a single failure in either trip system.
 
Required Actions B.1 and B.2 limit the time the RPS scram logic, for any Function, would not accommodate single failure in both trip systems (e.g., one-out-of-one and one-out-of-one arrangement for a typical four channel Function). The reduced reliability of this logic arrangement was not evaluated in References 9, 12 or 13 for the 12 hour Completion Time. Within the 6 hour allowance, the associated Function will have all required channels OPERABLE or in trip (or any combination) in one trip system.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                            B 3.3-24                                                                                                                                                                                    Revision No. 159 RPS Instrumentation B 3.3.1.1
 
BASES
 
ACTIONS                                                                                                    B.1 and B.2  (continued)
 
Completing one of these Required Actions restores RPS to a reliability level equivalent to that evaluated in References 9, 12 or 13, which justified a 12 hour allowable out of service time as presented in Condition A. The trip system in the more degraded state should be placed in trip or, alternatively, all the inoperable channels in that trip system should be placed in trip (e.g., a trip system with two inoperable channels could be in a more degraded state than a trip system with four inoperable channels if the two inoperable channels are in the same Function while the four inoperable channels are all in different Functions). The decision of which trip system is in the more degraded state should be based on prudent judgment and take into account current plant conditions (i.e., what MODE the plant is in).
If this action would result in a scram or RPT, it is permissible to place the other trip system or its inoperable channels in trip.
 
The 6 hour Completion Time is judged acceptable based on the remaining capability to trip, the diversity of the sensors available to provide the trip signals, the low probability of extensive numbers of inoperabilities affecting all diverse Functions, and the low probability of an event requiring the initiation of a scram. Alternatively, a Completion Time can be determined in accordance with the Risk Informed Completion Time (RICT) Program. A Note has been provided to indicate that a RICT is only applicable when a loss of function has not occurred.
 
Alternately, if it is not desired to place the inoperable channels (or one trip system) in trip (e.g., as in the case where placing the inoperable channel or associated trip system in trip would result in a scram, Condition D must be entered and its Required Action taken.
 
As noted, Condition B is not applicable for APRM Functions 2.a, 2.b, 2.c, 2.d, or 2.f. Inoperability of an APRM channel affects both trip systems and is not associated with a specific trip system as are the APRM 2-Out-Of-4 voter and other non-APRM channels for which Condition B applies. For an inoperable APRM channel, Required Action A.1 must be satisfied, and is the only action (other than restoring operability) that will restore capability to accommodate a single failure. Inoperability of a Function in more than one required APRM channel results in loss of trip capability for that Function and entry into Condition C, as well as entry into Condition A for each channel. Because Condition A and C provide Required Actions that are appropriate for the inoperability of APRM Functions 2.a, 2.b, 2.c, 2.d, or 2.f, and these functions are not associated with specific trip systems as are the APRM 2-Out-Of-4 voter and other non-APRM channels, Condition B does not apply.
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                            B 3.3-25                                                                                                                                                                                    Revision No. 159 RPS Instrumentation B 3.3.1.1
 
BASES
 
ACTIONS C.1
 
Required Action C.1 is intended to ensure that appropriate actions are taken if multiple, inoperable, untripped channels within the same trip system for the same Function result in an automatic Function, or two or more manual Functions, not maintaining RPS trip capability. A Function is considered to be maintaining RPS trip capability when sufficient channels are OPERABLE or in trip  (or the associated trip system is in trip), such that both trip systems will generate a trip signal from the given Function on a valid signal. For the typical Function with one-out-of-two taken twice logic and the WRNM and APRM Functions, this would require both trip systems to have one channel OPERABLE or in trip (or the associated trip system in trip).
For Function 5 (Main Steam Isolation Valve      Closure), this would require both trip systems to have each channel associated with the MSIVs in three main steam lines (not necessarily the same main steam lines for both trip systems)OPERABLE or in trip (or the associated trip system in trip). For Function 8 (Turbine Stop Valve      Closure),
this would require both trip systems to have three channels, each OPERABLE or in trip (or the associated trip system in trip). For Functions 12 (Reactor Mode Switch      Shutdown Position) and 13 (Manual Scram), this would require both trip systems to have one channel, each OPERABLE or in trip (or the associated trip system in trip).
 
The Completion Time is intended to allow the operator time to evaluate and repair any discovered inoperabilities. The 1 hour Completion Time is acceptable because it minimizes risk while allowing time for restoration or tripping of channels.
 
D.1
 
Required Action D.1 directs entry into the appropriate Condition referenced in Table 3.3.1.1-1. The applicable condition specified in the Table is Function and MODE or other specified condition dependent and may change as the Required Action of a previous Condition is completed. Each time an inoperable channel has not met any Required Action of Condition A, B, or C and the associated Completion Time has expired, Condition D will be entered for that channel and provides for transfer to the appropriate subsequent Condition.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                            B 3.3-26                                                                                                                                                                                              Revision No. 36 RPS Instrumentation B 3.3.1.1
 
BASES
 
ACTIONS                                                                                                    E.1, F.1 and G.1 (continued)
If the channel(s) is not restored to OPERABLE status or placed in trip (or the associated trip system placed in trip) within the allowed Completion Time, the plant must be placed in a MODE or other specified condition in which the LCO does not apply. The allowed Completion Times are reasonable, based on operating experience, to reach the specified condition from full power conditions in an orderly manner and without challenging plant systems. In addition, the Completion Time of Required Action E.1 is consistent with the Completion Time provided in LCO 3.2.2, "MINIMUM CRITICAL POWER RATIO (MCPR)."
H.1
 
If the channel(s) is not restored to OPERABLE status or placed in trip (or the associated trip system placed in trip) within the allowed Completion Time, the plant must be placed in a MODE or other specified condition in which the LCO does not apply. This is done by immediately initiating action to fully insert all insertable control rods in core cells containing one or more fuel assemblies. Control rods in core cells containing no fuel assemblies do not affect the reactivity of the core and are, therefore, not required to be inserted. Action must continue until all insertable control rods in core cells containing one or more fuel assemblies are fully inserted.
 
I.1 If OPRM Upscale trip capability is not maintained, Condition I exists and Backup Stability Protection (BSP) is required.
The Manual BSP Regions are described in Reference 22. The Manual BSP Regions are procedurally established consistent with the guidelines identified in Reference 22 and require specified manual operator actions if certain predefined operational conditions occur.
The Completion Time of immediately is based on the importance of limiting the period of time during which no automatic or alternate detect and suppress trip capability is in place.
 
I.2 and I.3 Actions I.2 and I.3 are both required to be taken in conjunction with Action I.1 if OPRM Upscale trip capability is not maintained. As described in Section 7.4 of Reference 22, the Automated BSP Scram Region is designed to avoid reactor instability by automatically preventing entry into (continued)
PBAPS UNIT 2                                                                                                                                                                                                                            B 3.3-27                                                                                                                                                                                    Revision No. 123 RPS Instrumentation B 3.3.1.1
 
BASES
 
ACTIONS                                                                                                    I.2 and I.3  (continued)
 
the region of the power and flow-operating map that is susceptible to reactor instability. The reactor trip would be initiated by the modified APRM Simulated Thermal Power-High scram setpoints for flow reduction events that would have terminated in the Manual BSP Region I. The Automated BSP Scram Region ensures an early scram and SLMCPR protection.
 
The Completion Time of 12 hours to complete the specified actions is reasonable, based on operational experience, and based on the importance of restoring an automatic reactor trip for thermal-hydraulic instability events.
 
BSP is intended as a temporary means to protect against thermal-hydraulic instability events. The action should be initiated immediately to document the situation and prepare the report.
The reporting requirements of Specification 5.6.8 document the corrective actions and schedule to restore the required channels to an OPERABLE status. The Completion Time of 90 days shown in Specification 5.6.8 is adequate to allow time to evaluate the cause of the inoperability and to determine the appropriate corrective actions and schedule to restore the required channels to OPERABLE status.
 
J.1
 
If the Required Action I is not completed within the associated Completion Time, then Action J is required. The Bases for the Manual BSP Regions and associated Completion Time is addressed in the Bases for I.1. The Manual BSP Regions are required in conjunction with the BSP Boundary.
 
J.2
 
The BSP Boundary, as described in Section 7.3 of Reference 22, defines an operating domain where potential instability events can be effectively addressed by the specified BSP manual operator actions. The BSP Boundary is constructed such that a flow reduction event initiated from this boundary and terminated at the core natural circulation line (NCL) would not exceed the Manual BSP Region I stability criterion. Potential instabilities would develop slowly as a result of the feedwater temperature transient (Ref. 22).
 
The Completion Time of 12 hours to complete the specified actions is reasonable, based on operational experience, to reach the specific condition from full power conditions in an orderly manner and without challenging plant systems.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                        B 3.3-27a                                                                                                                                                                                Revision No. 123 RPS Instrumentation B 3.3.1.1
 
BASES
 
ACTIONS J.3 (continued)
BSP is a temporary means for protection against thermal-hydraulic instability events. An extended period of inoperability without automatic trip capability is not justified. Consequently, the required channels are required to be restored to OPERABLE status within 120 days.
 
Based on engineering judgment, the likelihood of an instability event that could not be adequately handled by the use of the BSP Regions (See Action J.1) and the BSP Boundary (See Action J.2) during a 120-day period is negligibly small. The 120-day period is intended to allow for resolution of a variety of equipment problems (e.g.,
design changes, extensive analysis, or other unforeseen circumstances). This action is not intended to be used for operational convenience. Correction of most equipment failures or inoperabilities is expected to normally be accomplished within the completion times allowed for Actions for Conditions A and I.
 
A Note is provided to indicate that LCO 3.0.4 is not applicable. The intent of the note is to allow plant startup while operating within the 120-day Completion Time for Required Action J.3. The primary purpose of this exclusion is to allow an orderly completion of design and verification activities, in the event of a required design change, without undue impact on plant operation.
 
K.1
 
If the required channels are not restored to OPERABLE status and the Required Actions of J are not met within the associated Completion Times, then the plant must be placed in an operating condition in which the LCO does not apply. To achieve this status, the plant must be brought to less than 17.6% RTP within 4 hours. The allowed Completion Time is reasonable, based on operating experience, to reach the specified operating power level from full power conditions in an orderly manner and without challenging plant systems.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                        B 3.3-27b                                                                                                                                                                                Revision No. 143 RPS Instrumentation B 3.3.1.1 BASES  (continued)
 
SURVEILLANCE                                                  As noted at the beginning of the SRs, the SRs for each RPS REQUIREMENTS                                                  instrumentation Function are located in the SRs column of Table 3.3.1.1-1.
 
The Surveillances are modified by a Note to indicate that when a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours, provided the associated Function maintains RPS trip capability. Upon completion of the Surveillance, or expiration of the 6 hour allowance, the channel must be returned to OPERABLE status or the applicable Condition entered and Required Actions taken. This Note is based on the reliability analysis (Refs. 9, 12 & 13) assumption of the average time required to perform channel Surveillance.
That analysis demonstrated that the 6 hour testing allowance does not significantly reduce the probability that the RPS will trip when necessary.
 
SR 3.3.1.1.1 Performance of the CHANNEL CHECK ensures that a gross failure of instrumentation has not occurred. A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a similar parameter on other channels. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value.
Significant deviations between instrument channels could be an indication of excessive instrument drift in one of the channels or something even more serious. A CHANNEL CHECK will detect gross channel failure; thus, it is key to verifying the instrumentation continues to operate properly between each CHANNEL CALIBRATION.
Agreement criteria are determined by the plant staff based on a combination of the channel instrument uncertainties, including indication and readability. If a channel is outside the criteria, it may be an indication that the instrument has drifted outside its limit.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. The CHANNEL CHECK supplements less formal, but more frequent, checks of channels during normal operational use of the displays associated with the channels required by the LCO.
 
SR  3.3.1.1.2 To ensure that the APRMs are accurately indicating the true core average power, the APRMs are adjusted to the reactor power calculated from a heat balance if the heat balance calculated reactor power exceeds the APRM channel output by (continued) more than 2% RTP.
 
PBAPS UNIT 2                                                                                                                                                                                                                            B 3.3-28                                                                                                                                                                                    Revision No. 149
 
RPS Instrumentation B 3.3.1.1
 
BASES
 
SURVEILLANCE                                                  SR  3.3.1.1.2  (continued)
REQUIREMENTS This Surveillance does not preclude making APRM channel adjustments, if desired, when the heat balance calculated reactor power is less than the APRM channel output. To provide close agreement between the APRM indicated power and to preserve operating margin, the APRM channels are normally adjusted to within +/- 2% of the heat balance calculated reactor power. However, this agreement is not required for OPERABILITY when APRM output indicates a higher reactor power than the heat balance calculated reactor power.
 
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                        B 3.3-28a                                                                                                                                                                                Revision No. 149 RPS Instrumentation B 3.3.1.1
 
BASES
 
SURVEILLANCE                                                  SR  3.3.1.1.2  (continued)
REQUIREMENTS A restriction to satisfying this SR when < 22.6% RTP is provided that requires the SR to be met only at                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                      22.6% RTP because it is difficult to accurately maintain APRM indication of core THERMAL POWER consistent with a heat balance when < 22.6% RTP. At low power levels, a high degree of accuracy is unnecessary because of the large, inherent margin to thermal limits (MCPR, LHGR and APLHGR). At                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                                              22.6%
RTP, the Surveillance is required to have been satisfactorily performed in accordance with SR 3.0.2. A Note is provided which allows an increase in THERMAL POWER above 22.6% if the Frequency is not met per SR 3.0.2. In this event, the SR must be performed within 12 hours after reaching or exceeding 22.6% RTP. Twelve hours is based on operating experience and in consideration of providing a reasonable time in which to complete the SR.
 
SR  3.3.1.1.3
 
(Not Used.)
 
SR  3.3.1.1.4
 
A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the entire channel will perform the intended function. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
 
SR  3.3.1.1.5 and SR  3.3.1.1.6
 
A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the entire channel will perform the intended function. Any setpoint adjustment shall be made consistent with the assumptions of the current plant specific setpoint methodology.
 
As noted, SR 3.3.1.1.5 is not required to be performed when entering MODE 2 from MODE 1, since testing of the MODE 2 required WRNM Functions cannot be performed in MODE 1 without utilizing jumpers, lifted leads, or movable links.
This allows entry into MODE 2 if the Frequency is not met per SR 3.0.2. In this event, the SR must be performed within 12 hours after entering MODE 2 from MODE 1. Twelve hours is based on operating experience and in consideration of providing a reasonable time in which to complete the SR.
 
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                            B 3.3-29                                                                                                                                                                                    Revision No. 143 RPS Instrumentation B 3.3.1.1
 
BASES
 
SURVEILLANCE                                                  SR  3.3.1.1.7 REQUIREMENTS (continued)                                        (Not Used.)
 
SR  3.3.1.1.8
 
LPRM gain settings are determined from the local flux profiles measured by the Traversing Incore Probe (TIP)
System. This establishes the relative local flux profile for appropriate representative input to the APRM System.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
 
SR  3.3.1.1.9 and SR  3.3.1.1.14
 
A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the entire channel will perform the intended function. Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology. For Function 5, 7, and 8 channels, verification that the trip settings are less than or equal to the specified Allowable Value during the CHANNEL FUNCTIONAL TEST is not required since the channels consist of mechanical switches and are not subject to drift. An exception to this are two of the Function 7 level switches which are not mechanical. These Scram Discharge Volume (SDV) RPS switches (Fluid Components Inc.) are heat sensitive electronic level detectors which actuate by sensing a difference in temperature. The temperature detectors are permanently affixed within the scram discharge volume piping conservatively below the level (allowable value as measured in gallons) at which an RPS actuation signal will occur. Since there is no drift involved with the physical location of these switches, verifying the trip settings are less than or equal to the specified allowable value during the CHANNEL FUNCTIONAL TEST is not required.
Additionally, historical calibration data has indicated that the FCI level switches have not exceeded their Allowable Value when tested.
 
In addition, Function 5 and 7 instruments are not accessible while the unit is operating at power due to high radiation and the installed indication instrumentation does not provide accurate indication of the trip setting. For the Function 9 channels, verification that the trip settings are less than or equal to the specified Allowable Value during the CHANNEL
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                            B 3.3-30                                                                                                                                                                                    Revision No. 114 RPS Instrumentation B 3.3.1.1
 
BASES
 
SURVEILLANCE                                                  SR  3.3.1.1.9 and SR  3.3.1.1.14  (continued)
REQUIREMENTS FUNCTIONAL TEST is not required since the instruments are not accessible while the unit is operating at power due to high radiation and the installed indication instrumentation does not provided accurate indication of the trip setting. Waiver of these verifications for the above functions is considered acceptable since the magnitude of drift assumed in the setpoint calculation is based on a 24 month calibration interval. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
 
SR  3.3.1.1.10, SR  3.3.1.1.12, SR  3.3.1.1.15, and SR  3.3.1.1.16
 
A CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor. This test verifies that the channel responds to the measured parameter within the necessary range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drifts between successive calibrations, consistent with the current plant specific setpoint methodology.
 
As noted for SR 3.3.1.1.10, radiation detectors are excluded from CHANNEL CALIBRATION due to ALARA reasons (when the plant is operating, the radiation detectors are generally in a high radiation area; the steam tunnel). This exclusion is acceptable because the radiation detectors are passive devices, with minimal drift. To complete the radiation CHANNEL CALIBRATION, SR 3.3.1.1.16 requires that the radiation detectors be calibrated in accordance with the Surveillance Frequency Control Program.
 
SR 3.3.1.1.12 for Function 3.3.1.1-1.2.b is modified by two Notes as identified in Table 3.3.1.1-1. The first Note requires evaluation of channel performance for the condition where the as-found setting for the channel setpoint is outside its as-found tolerance but conservative with respect to the Allowable Value. Evaluation of channel performance will verify that the channel will continue to behave in accordance with safety analysis assumptions and the channel performance assumptions in the setpoint methodology. The purpose of the assessment is to ensure confidence in the channel performance prior to returning the channel to service. For channels determined to be OPERABLE but degraded, after returning the channel to service the performance of these channels will be evaluated under the
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                            B 3.3-31                                                                                                                                                                                    Revision No. 114 RPS Instrumentation B 3.3.1.1
 
BASES
 
SURVEILLANCE                                                  SR  3.3.1.1.10, SR  3.3.1.1.12, SR  3.3.1.1.15, REQUIREMENTS                                                  and SR  3.3.1.1.16  (continued)
 
plant Corrective Action Program. Entry into the Corrective Action Program will ensure required review and documentation of the condition. The second Note requires that the as-left setting for the channel be within the Leave Alone Zone around the NTSP. Where a setpoint more conservative than the NTSP is used in the plant surveillance procedures (ATSP), the Leave Alone Zone and as-found tolerances, as applicable, will be applied to the surveillance procedure setpoint. This will ensure that sufficient margin to the Safety Limit and/or Analytical Limit is maintained. If the as-left channel setting cannot be returned to a setting within the Leave Alone Zone around the NTSP, then the channel shall be declared inoperable. The second Note also requires that NTSP and the methodologies for calculating the Leave Alone Zone and the as-found tolerances be in the Bases for the applicable Technical Specifications.
 
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
 
As noted for SR 3.3.1.1.12, neutron detectors are excluded from CHANNEL CALIBRATION because they are passive devices, with minimal drift, and because of the difficulty of simulating a meaningful signal. Changes in neutron detector sensitivity are compensated for by performing the calorimetric calibration (SR 3.3.1.1.2) and the LPRM calibration against the TIPs (SR 3.3.1.1.8).
 
A second note is provided for SR 3.3.1.1.12 that allows the WRNM SR to be performed within 12 hours of entering MODE 2 from MODE 1. Testing of the MODE 2 WRNM Functions cannot be performed in MODE 1 without utilizing jumpers, lifted leads or movable links. This Note allows entry into MODE 2 from MODE 1, if the Frequency is not met per SR 3.0.2. Twelve hours is based on operating experience and in consideration of providing a reasonable time in which to complete the SR.
 
A third note is provided for SR 3.3.1.1.12 that includes in the SR the recirculation flow (drive flow) transmitters, which supply the flow signal to the APRMs. The APRM Simulated Thermal Power-High Function (Function 2.b) and the OPRM Upscale Function (Function 2.f), both require a valid drive flow signal. The APRM Simulated Thermal Power-High
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                            B 3.3-32                                                                                                                                                                                    Revision No. 114 RPS Instrumentation B 3.3.1.1
 
BASES
 
SURVEILLANCE                                                  SR  3.3.1.1.10, SR 3.3.1.1.12, SR 3.3.1.1.15, REQUIREMENTS                                                  and SR 3.3.1.1.16  (continued)
 
Function uses drive flow to vary the trip setpoint. The OPRM Upscale Function uses drive flow to automatically enable or bypass the OPRM Upscale trip output to RPS. A CHANNEL CALIBRATION of the APRM drive flow signal requires both calibrating the drive flow transmitters and establishing a valid drive flow / core flow relationship. The drive flow
                      /core flow relationship is established once per refuel cycle, while operating at or near rated power and flow conditions.
This method of correlating core flow and drive flow is consistent with GE recommendations. Changes throughout the cycle in the drive flow / core flow relationship due to the changing thermal hydraulic operating conditions of the core are accounted for in the margins included in the bases or analyses used to establish the setpoints for the APRM Simulated Thermal Power-High Function and the OPRM Upscale Function.
 
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
 
SR 3.3.1.1.11
 
A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the entire channel will perform the intended function. For the APRM Functions, this test supplements the automatic self-test functions that operate continuously in the APRM and voter channels. The scope of the APRM CHANNEL FUNCTIONAL TEST is limited to verification of system trip output hardware. Software controlled functions are tested only incidentally. Automatic internal self-test functions check the EPROMs in which the software-controlled logic is defined. Any changes in the EPROMs will be detected by the self-test function resulting in a trip and/or alarm condition. The APRM CHANNEL FUNCTIONAL TEST covers the APRM channels (including recirculation flow processing - applicable to Function 2.b and the auto-enable portion of Function 2.f only), the 2-Out-Of 4 voter channels, and the interface connections into the RPS trip systems from the voter channels. Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. (NOTE: The actual voting logic of the 2-Out-Of-4 Voter Function is tested as part of SR 3.3.1.1.17.)
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                            B 3.3-33                                                                                                                                                                                    Revision No. 152 RPS Instrumentation B 3.3.1.1
 
BASES
 
SURVEILLANCE                                                  SR 3.3.1.1.11 (continued)
REQUIREMENTS A Note is provided for Function 2.a that requires this SR to be performed within 12 hours of entering MODE 2 from MODE 1.
Testing of the MODE 2 APRM Function cannot be performed in MODE 1 without utilizing jumpers or lifted leads. This Note allows entry into MODE 2 from MODE 1 if the associated Frequency is not met per SR 3.0.2.
 
A second Note is provided for Function 2.b that clarifies that the CHANNEL FUNCTIONAL TEST for Function 2.b includes testing of the recirculation flow processing electronics, excluding the flow transmitters.
 
SR 3.3.1.1.13
 
This SR ensures that scrams initiated from the Turbine Stop Valve-Closure and Turbine Control Valve Fast Closure, Trip Oil Pressure-Low Functions will not be inadvertently bypassed when THERMAL POWER is  26.3% RTP. This involves calibration of the bypass channels. Adequate margins for the instrument setpoint methodologies are incorporated into the actual setpoint. Because main turbine bypass flow can affect this setpoint nonconservatively (THERMAL POWER is derived from turbine first stage pressure), the main turbine bypass valves must remain closed during the calibration at THERMAL POWER 26.3% RTP to ensure that the calibration is valid.
 
If any bypass channel's setpoint is nonconservative (i.e.,
the Functions are bypassed at                                                                                                                                                                                                                                                                                                              26.3% RTP, either due to open main turbine bypass valve(s) or other reasons), then the affected Turbine Stop Valve      Closure and Turbine Control Valve Fast Closure, Trip Oil Pressure      Low Functions are considered inoperable. Alternatively, the bypass channel can be placed in the conservative condition (nonbypass). If placed in the nonbypass condition, this SR is met and the channel is considered OPERABLE.
 
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                            B 3.3-34                                                                                                                                                                                    Revision No. 143 RPS Instrumentation B 3.3.1.1
 
BASES
 
SURVEILLANCE                                                  SR  3.3.1.1.17 REQUIREMENTS (continued)                                        The LOGIC SYSTEM FUNCTIONAL TEST demonstrates the OPERABILITY of the required trip logic for a specific channel. The functional testing of control rods (LCO 3.1.3), and SDV vent and drain valves (LCO 3.1.8),
overlaps this Surveillance to provide complete testing of the assumed safety function.
 
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
 
The LOGIC SYSTEM FUNCTIONAL TEST for APRM Function 2.e simulates APRM and OPRM trip conditions at the 2-Out-Of-4 voter channel inputs to check all combinations of two tripped inputs to the 2-Out-Of-4 logic in the voter channels and APRM related redundant RPS relays.
 
SR  3.3.1.1.18
 
This SR ensures that the individual channel response times are maintained less than or equal to the original design value. The RPS RESPONSE TIME acceptance criterion is included in Reference 11.
 
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
 
SR  3.3.1.1.19  Deleted
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                            B 3.3-35                                                                                                                                                                                    Revision No. 123 RPS Instrumentation B 3.3.1.1
 
BASES  (continued)
 
REFERENCES                                                                      1.                    UFSAR, Section 7.2.
: 2.                    UFSAR, Chapter 14.
: 3.                    NEDO-32368, "Nuclear Measurement Analysis and Control Wide Range Neutron Monitoring System Licensing Report for Peach Bottom Atomic Power Station, Units 2 and 3,"
November 1994.
: 4.                    NEDC-33566P, "Safety Analysis Report for Exelon Peach ottom Atomic Power Station, Units 2 and 3, Constant Pressure Power Uprate," Revision 0.
: 5.                    UFSAR, Section 14.6.2.
: 6.                    UFSAR, Section 14.5.4.
: 7.                    UFSAR, Section 14.5.1.
: 8.                    P. Check (NRC) letter to G. Lainas (NRC), "BWR Scram Discharge System Safety Evaluation," December 1, 1980.
: 9.                    NEDO-30851-P-A, "Technical Specification Improvement Analyses for BWR Reactor Protection System," March 1988.
: 10.          MDE-87-0485-1, "Technical Specification Improvement Analysis for the Reactor Protection System for Peach Bottom Atomic Power Station Units 2 and 3," October 1987.
: 11.          UFSAR, Section 7.2.3.9.
: 12.          NEDC-32410P-A, Nuclear Measurement Analysis and Control Power Range Neutron Monitor (NUMAC PRNM)
Retrofit Plus Option III Stability Trip Function, October 1995.
: 13.          NEDC-32410P Supplement 1, Nuclear Measurement Analysis and Control Power Range Neutron Monitor (NUMAC PRNM) Retrofit Plus Option III Stability Trip Function, Supplement 1, November 1997.
: 14. Deleted
: 15.            Deleted (continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                        B 3.3-35a                                                                                                                                                                                Revision No. 123 RPS Instrumentation B 3.3.1.1
 
BASES
 
REFERENCES 16. Deleted (continued)
: 17. Deleted
: 18. Deleted
: 19.            NEDO-24229-1, "Peach Bottom Atomic Power Station Units 2 and 3 Single-Loop Operation," May 1980.
: 20.            Setpoint Methodology for Peach Bottom Atomic Power Station and Limerick Generating Station, CC-MA-103-2001.
: 21.            Backup Stability Protection (BSP) for Inoperable Option III Solutions, OGO2-0119, July 17, 2002.
: 22.            GE Hitachi Nuclear Energy, "GE Hitachi Boiling Water Reactor, Detect and Suppress Solution - Confirmation Density," NEDC-33075P-A, Revision 8, November 2013.
: 23.            GEH letter to NRC, "NEDC-33075P-A, Detect and Suppress Solution - Confirmation Density (DSS-CD) Analytical Limit (TAC No. MD0277)," October 29, 2008. (ADAMS Accession No. ML083040052).
: 24.            000N7936-R0, "Project Task Report - Exelon Generation Company LLC, Peach Bottom Atomic Power Station Unit 2
                                              & 3 MELLLA+, Task T0202: Thermal-Hydraulic Stability,"
April 2014.
: 25.            NEDC-33873P, Safety Analysis Report for Peach Bottom Atomic Power Station, Units 2 and 3, Thermal Power Optimization, Revision 0.
 
PBAPS UNIT 2                                                                                                                                                                                                                        B 3.3-35b                                                                                                                                                                                Revision No. 143 WRNM Instrumentation B 3.3.1.2
 
B 3.3  INSTRUMENTATION
 
B 3.3.1.2  Wide Range Neutron Monitor (WRNM) Instrumentation
 
BASES
 
BACKGROUND                                                                      The WRNMs are capable of providing the operator with information relative to the neutron flux level at very low flux levels in the core. As such, the WRNM indication is used by the operator to monitor the approach to criticality and determine when criticality is achieved.
 
The WRNM subsystem of the Neutron Monitoring System (NMS) consists of eight channels. Each of the WRNM channels can be bypassed, but only one at any given time per RPS trip system, by the operation of a bypass switch. Each channel includes one detector that is permanently positioned in the core. Each detector assembly consists of a miniature fission chamber with associated cabling, signal conditioning equipment, and electronics associated with the various WRNM functions. The signal conditioning equipment converts the current pulses from the fission chamber to analog DC currents that correspond to the count rate. Each channel also includes indication, alarm, and control rod blocks.
However, this LCO specifies OPERABILITY requirements only for the monitoring and indication functions of the WRNMs.
 
During refueling, shutdown, and low power operations, the primary indication of neutron flux levels is provided by the WRNMs or special movable detectors connected to the normal WRNM circuits. The WRNMs provide monitoring of reactivity changes during fuel or control rod movement and give the control room operator early indication of unexpected subcritical multiplication that could be indicative of an approach to criticality.
 
APPLICABLE                                                                      Prevention and mitigation of prompt reactivity excursions SAFETY ANALYSES                    during refueling and low power operation is provided by LCO 3.9.1, "Refueling Equipment Interlocks"; LCO 3.1.1, "SHUTDOWN MARGIN (SDM)"; LCO 3.3.1.1, "Reactor Protection System (RPS) Instrumentation"; WRNM Period-Short and
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                            B 3.3-36                                                                                                                                                                                              Revision No. 24 WRNM Instrumentation B 3.3.1.2
 
BASES
 
APPLICABLE                                                                      Average Power Range Monitor (APRM) Startup High Flux Scram SAFETY ANALYSES                    Functions; and LCO 3.3.2.1, "Control Rod Block (continued)                                        Instrumentation."
 
The WRNMs have no safety function associated with monitoring neutron flux at very low levels and are not assumed to function during any UFSAR design basis accident or transient analysis which would occur at very low neutron flux levels.
However, the WRNMs provide the only on-scale monitoring of neutron flux levels during startup and refueling.
Therefore, they are being retained in Technical Specifications.
 
LCO                                                                                                                                            During startup in MODE 2, three of the eight WRNM channels are required to be OPERABLE to monitor the reactor flux level and reactor period prior to and during control rod withdrawal, subcritical multiplication and reactor criticality. These three required channels must be located in different core quadrants in order to provide a representation of the overall core response during those periods when reactivity changes are occurring throughout the core.
 
In MODES 3 and 4, with the reactor shut down, two WRNM channels provide redundant monitoring of flux levels in the core.
 
In MODE 5, during a spiral offload or reload, a WRNM outside the fueled region will no longer be required to be OPERABLE, since it is not capable of monitoring neutron flux in the fueled region of the core. Thus, CORE ALTERATIONS are allowed in a quadrant with no OPERABLE WRNM in an adjacent quadrant provided the Table 3.3.1.2-1, footnote (b),
requirement that the bundles being spiral reloaded or spiral offloaded are all in a single fueled region containing at least one OPERABLE WRNM is met. Spiral reloading and offloading encompass reloading or offloading a cell on the edge of a continuous fueled region (the cell can be reloaded or offloaded in any sequence).
 
In nonspiral routine operations, two WRNMs are required to be OPERABLE to provide redundant monitoring of reactivity changes in the reactor core. Because of the local nature of reactivity changes during refueling, adequate coverage is provided by requiring one WRNM to be OPERABLE for the connected fuel in the quadrant of the reactor core where
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                            B 3.3-37                                                                                                                                                                                              Revision No. 24 WRNM Instrumentation B 3.3.1.2
 
BASES
 
LCO                                                                                                                                            CORE ALTERATIONS are being performed. There are two WRNMs (continued)                                        in each quadrant. Any CORE ALTERATIONS must be performed in a region of fuel that is connected to an OPERABLE WRNM to ensure that the reactivity changes are monitored within the fueled region(s) of the quadrant. The other WRNM that is required to be OPERABLE must be in an adjacent quadrant containing fuel. These requirements ensure that the reactivity of the core will be continuously monitored during CORE ALTERATIONS.
Special movable detectors, according to footnote (c) of Table 3.3.1.2-1, may be used in place of the normal WRNM nuclear detectors. These special detectors must be connected to the normal WRNM circuits in the NMS, such that the applicable neutron flux indication can be generated.
These special detectors provide more flexibility in monitoring reactivity changes during fuel loading, since they can be positioned anywhere within the core during refueling. They must still meet the location requirements of SR 3.3.1.2.2 and all other required SRs for WRNMs.
The Table 3.3.1.2-1, footnote (d), requirement provides for conservative spatial core coverage.
 
For a WRNM channel to be considered OPERABLE, it must be providing neutron flux monitoring indication.
 
APPLICABILITY                                        The WRNMs are required to be OPERABLE in MODES 2, 3, 4, and 5 prior to the WRNMs reading 125E-5 % power to provide for neutron monitoring. In MODE 1, the APRMs provide adequate monitoring of reactivity changes in the core; therefore, the WRNMs are not required. In MODE 2, with WRNMs reading greater than 125E-5 % power, the WRNM Period-Short function provides adequate monitoring and the WRNMs monitoring indication is not required.
 
ACTIONS                                                                                                    A.1 and B.1 In MODE 2, the WRNM channels provide the means of monitoring core reactivity and criticality. With any number of the required WRNMs inoperable, the ability to monitor neutron flux is degraded. Therefore, a limited time is allowed to restore the inoperable channels to OPERABLE status.
Provided at least one WRNM remains OPERABLE, Required Action A.1 allows 4 hours to restore the required WRNMs to OPERABLE status. or in accordance with the Risk Informed Completion Time (RICT) Program. A Note has been provided to indicate that a RICT is not applicable if a loss of function exists under the given Condition. This time is reasonable because there is adequate capability remaining to monitor the core, there is limited risk of an event during this time, and there is sufficient time to take corrective actions to restore the required WRNMs to OPERABLE status.
During this time, control rod withdrawal and power increase (continued) is not precluded
 
PBAPS UNIT 2                                                                                                                                                                                                                            B 3.3-38                                                                                                                                                                                              Revision No. 24 WRNM Instrumentation B 3.3.1.2
 
BASES
 
ACTIONS                                                                                                    A.1 and B.1  (continued)
 
by this Required Action. Having the ability to monitor the core with at least one WRNM, proceeding to WRNM indication greater than 125E-5 % power, and thereby exiting the Applicability of this LCO, is acceptable for ensuring adequate core monitoring and allowing continued operation.
 
With three required WRNMs inoperable, Required Action B.1 allows no positive changes in reactivity (control rod withdrawal must be immediately suspended) due to inability to monitor the changes. Required Action A.1 still applies and allows 4 hours to restore monitoring capability prior to requiring control rod insertion. This allowance is based on the limited risk of an event during this time, provided that no control rod withdrawals are allowed, and the desire to concentrate efforts on repair, rather than to immediately shut down, with no WRNMs OPERABLE.
 
C.1
 
In MODE 2, if the required number of WRNMs is not restored to OPERABLE status within the allowed Completion Time, the reactor shall be placed in MODE 3. With all control rods fully inserted, the core is in its least reactive state with the most margin to criticality. The allowed Completion Time of 12 hours is reasonable, based on operating experience, to reach MODE 3 from full power conditions in an orderly manner and without challenging plant systems.
 
D.1 and D.2
 
With one or more required WRNMs inoperable in MODE 3 or 4, the neutron flux monitoring capability is degraded or nonexistent. The requirement to fully insert all insertable control rods ensures that the reactor will be at its minimum reactivity level while no neutron monitoring capability is available. Placing the reactor mode switch in the shutdown position prevents subsequent control rod withdrawal by maintaining a control rod block. The allowed Completion Time of 1 hour is sufficient to accomplish the Required Action, and takes into account the low probability of an event requiring the WRNM occurring during this interval.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                            B 3.3-39                                                                                                                                                                                              Revision No. 24 WRNM Instrumentation B 3.3.1.2
 
BASES
 
ACTIONS                                                                                                    E.1 and E.2 (continued)
With one or more required WRNMs inoperable in MODE 5, the ability to detect local reactivity changes in the core during refueling is degraded. CORE ALTERATIONS must be immediately suspended and action must be immediately initiated to fully insert all insertable control rods in core cells containing one or more fuel assemblies.
Suspending CORE ALTERATIONS prevents the two most probable causes of reactivity changes, fuel loading and control rod withdrawal, from occurring. Inserting all insertable control rods ensures that the reactor will be at its minimum reactivity given that fuel is present in the core.
Suspension of CORE ALTERATIONS shall not preclude completion of the movement of a component to a safe, conservative position.
 
Action (once required to be initiated) to insert control rods must continue until all insertable rods in core cells containing one or more fuel assemblies are inserted.
 
SURVEILLANCE                                                  As noted at the beginning of the SRs, the SRs for each WRNM REQUIREMENTS                                                  Applicable MODE or other specified conditions are found in the SRs column of Table 3.3.1.2-1.
 
SR  3.3.1.2.1 and SR  3.3.1.2.3
 
Performance of the CHANNEL CHECK ensures that a gross failure of instrumentation has not occurred. A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a similar parameter on another channel. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value. Significant deviations between the instrument channels could be an indication of excessive instrument drift in one of the channels or something even more serious.
A CHANNEL CHECK will detect gross channel failure; thus, it is key to verifying the instrumentation continues to operate properly between each CHANNEL CALIBRATION.
 
Agreement criteria are determined by the plant staff based on a combination of the channel instrument uncertainties, including indication and readability. If a channel is outside the criteria, it may be an indication that the instrument has drifted outside its limit.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                            B 3.3-40                                                                                                                                                                                              Revision No. 24 WRNM Instrumentation B 3.3.1.2
 
BASES
 
SURVEILLANCE                                                  SR  3.3.1.2.1 and SR  3.3.1.2.3  (continued)
REQUIREMENTS The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. The CHANNEL CHECK supplements less formal, but more frequent, checks of channels during normal operational use of the displays associated with the channels required by the LCO.
 
SR  3.3.1.2.2
 
To provide adequate coverage of potential reactivity changes in the core, one WRNM is required to be OPERABLE for the connected fuel in the quadrant where CORE ALTERATIONS are being performed, and the other OPERABLE WRNM must be in an adjacent quadrant containing fuel. Note 1 states that the SR is required to be met only during CORE ALTERATIONS. It is not required to be met at other times in MODE 5 since core reactivity changes are not occurring. This Surveillance consists of a review of plant logs to ensure that WRNMs required to be OPERABLE for given CORE ALTERATIONS are, in fact, OPERABLE. In the event that only one WRNM is required to be OPERABLE, per Table 3.3.1.2-1, footnote (b), only the
: a. portion of this SR is required. Note 2 clarifies that more than one of the three requirements can be met by the same OPERABLE WRNM. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
 
SR  3.3.1.2.4
 
This Surveillance consists of a verification of the WRNM instrument readout to ensure that the WRNM reading is greater than a specified minimum count rate, which ensures that the detectors are indicating count rates indicative of neutron flux levels within the core. The signal-to-noise ratio shown in Figure 3.3.1.2-1 is the WRNM count rate at which there is a 95% probability that the WRNM signal indicates the presence of neutrons and only a 5% probability that the WRNM signal is the result of noise (Ref. 1). With few fuel assemblies loaded, the WRNMs will not have a high enough count rate to satisfy the SR. Therefore, allowances are made for loading sufficient "source" material, in the form of irradiated fuel assemblies, to establish the minimum count rate.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                            B 3.3-41                                                                                                                                                                                              Revision No. 86 WRNM Instrumentation B 3.3.1.2
 
BASES
 
SURVEILLANCE                                                  SR  3.3.1.2.4  (continued)
REQUIREMENTS To accomplish this, the SR is modified by Note 1 that states that the count rate is not required to be met on a WRNM that has less than or equal to four fuel assemblies adjacent to the WRNM and no other fuel assemblies are in the associated core quadrant. With four or less fuel assemblies loaded around each WRNM and no other fuel assemblies in the associated core quadrant, even with a control rod withdrawn, the configuration will not be critical. In addition, Note 2 states that this requirement does not have to be met during spiral unloading. If the core is being unloaded in this manner, the various core configurations encountered will not be critical.
 
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
 
SR  3.3.1.2.5
 
Performance of a CHANNEL FUNCTIONAL TEST demonstrates the associated channel will function properly. SR 3.3.1.2.5 is required in MODES 2, 3, 4 and 5 and ensures that the channels are OPERABLE while core reactivity changes could be in progress. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                            B 3.3-42                                                                                                                                                                                              Revision No. 86 WRNM Instrumentation B 3.3.1.2
 
BASES
 
SURVEILLANCE                                                  SR  3.3.1.2.5  (continued)
REQUIREMENTS Verification of the signal to noise ratio also ensures that the detectors are correctly monitoring the neutron flux.
 
The Note to the Surveillance allows the Surveillance to be delayed until entry into the specified condition of the Applicability (THERMAL POWER decreased to WRNM reading of 125E-5 % power or below). The SR must be performed within 12 hours after WRNMs are reading 125E-5 % power or below.
The allowance to enter the Applicability with the Frequency not met is reasonable, based on the limited time of 12 hours allowed after entering the Applicability. Although the Surveillance could be performed while at higher power, the plant would not be expected to maintain steady state operation at this power level. In this event, the 12 hour Frequency is reasonable, based on the WRNMs being otherwise verified to be OPERABLE (i.e., satisfactorily performing the CHANNEL CHECK) and the time required to perform the Surveillances.
 
SR  3.3.1.2.6
 
Performance of a CHANNEL CALIBRATION verifies the performance of the WRNM detectors and associated circuitry. The Frequency considers the plant conditions required to perform the test, the ease of performing the test, and the likelihood of a change in the system or component status. Note 1 excludes the neutron detectors from the CHANNEL CALIBRATION because they cannot readily be adjusted. The detectors are fission chambers that are designed to have a relatively constant sensitivity over the range and with an accuracy specified for a fixed useful life.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                            B 3.3-43                                                                                                                                                                                              Revision No. 86 WRNM Instrumentation B 3.3.1.2
 
BASES
 
SURVEILLANCE                                                  SR  3.3.1.2.6  (continued)
REQUIREMENTS Note 2 to the Surveillance allows the Surveillance to be delayed until entry into the specified condition of the Applicability. The SR must be performed in MODE 2 within 12 hours of entering MODE 2 with WRNMs reading 125E-5 % power or below. The allowance to enter the Applicability with the Frequency not met is reasonable, based on the limited time of 12 hours allowed after entering the Applicability. Although the Surveillance could be performed while at higher power, the plant would not be expected to maintain steady state operation at this power level. In this event, the 12 hour Frequency is reasonable, based on the WRNMs being otherwise verified to be OPERABLE (i.e., satisfactorily performing the CHANNEL CHECK) and the time required to perform the Surveillance.
 
REFERENCES                                                                      1.                    NRC Safety Evaluation Report for Amendment Numbers 147 and 149 to Facility Operating License Numbers DPR-44 and DPR-56, Peach Bottom Atomic Power Station, Unit Nos. 2 and 3, August 28, 1989.
 
PBAPS UNIT 2                                                                                                                                                                                                                            B 3.3-44                                                                                                                                                                                              Revision No. 86 Control Rod Block Instrumentation B 3.3.2.1
 
B 3.3  INSTRUMENTATION
 
B 3.3.2.1  Control Rod Block Instrumentation
 
BASES
 
BACKGROUND                                                                      Control rods provide the primary means for control of reactivity changes. Control rod block instrumentation includes channel sensors, logic circuitry, switches, and relays that are designed to ensure that specified fuel design limits are not exceeded for postulated transients and accidents. During high power operation, the rod block monitor (RBM) provides protection for control rod withdrawal error events. During low power operations, control rod blocks from the rod worth minimizer (RWM) enforce specific control rod sequences designed to mitigate the consequences of the control rod drop accident (CRDA). During shutdown conditions, control rod blocks from the Reactor Mode Switch      Shutdown Position Function ensure that all control rods remain inserted to prevent inadvertent criticalities.
 
The purpose of the RBM is to limit control rod withdrawal if localized neutron flux exceeds a predetermined setpoint during control rod manipulations. It is assumed to function to block further control rod withdrawal to preclude a MCPR Safety Limit (SL) violation. The RBM supplies a trip signal to the Reactor Manual Control System (RMCS) to appropriately inhibit control rod withdrawal during power operation above the low power range setpoint. The RBM has two channels, either of which can initiate a control rod block when the channel output exceeds the control rod block setpoint. One RBM channel inputs into one RMCS rod block circuit and the other RBM channel inputs into the second RMCS rod block circuit. The RBM channel signal is generated by averaging a set of local power range monitor (LPRM) signals at various core heights surrounding the control rod being withdrawn. A signal from one of the four redundant average power range monitor (APRM) channels supplies a reference signal for one of the RBM channels and a signal from another of the APRM channels supplies the reference signal to the second RBM channel. This reference signal is used to determine which RBM range setpoint (low, intermediate, or high) is enabled.
If the APRM is indicating less than the low power range setpoint, the RBM is automatically bypassed. The RBM is also automatically bypassed if a peripheral control rod is selected (Ref. 1). A rod block signal is also generated if an RBM inoperable trip occurs, since this could indicate a problem with the RBM channel.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                            B 3.3-45                                                                                                                                                                                              Revision No. 36 Control Rod Block Instrumentation B 3.3.2.1
 
BASES
 
BACKGROUND                                                                      The inoperable trip will occur if, during the nulling (continued)                                        (normalization) sequence, the RBM channel fails to null or too few LPRM inputs are available, if a critical self-test fault has been detected, or the RBM instrument mode switch is moved to any position other than "Operate".
 
The purpose of the RWM is to control rod patterns during startup and shutdown, such that only specified control rod sequences and relative positions are allowed over the operating range from all control rods inserted to 10% RTP.
The sequences effectively limit the potential amount and rate of reactivity increase during a CRDA. Prescribed control rod sequences are stored in the RWM, which will initiate control rod withdrawal and insert blocks when the actual sequence deviates beyond allowances from the stored sequence. The RWM determines the actual sequence based position indication for each control rod. The RWM also uses feedwater flow and steam flow signals to determine when the reactor power is above the preset power level at which the RWM is automatically bypassed (Ref. 2). The RWM is a single channel system that provides input into both RMCS rod block circuits.
 
With the reactor mode switch in the shutdown position, a control rod withdrawal block is applied to all control rods to ensure that the shutdown condition is maintained. This Function prevents inadvertent criticality as the result of a control rod withdrawal during MODE 3 or 4, or during MODE 5 when the reactor mode switch is required to be in the shutdown position. The reactor mode switch has two channels, each inputting into a separate RMCS rod block circuit. A rod block in either RMCS circuit will provide a control rod block to all control rods.
 
APPLICABLE                                                                      1. Rod Block Monitor SAFETY ANALYSES, LCO, and                                                                                          The RBM is designed to prevent violation of the MCPR APPLICABILITY                                        SL and the cladding 1% plastic strain fuel design limit that may result from a single control rod withdrawal error (RWE) event. The analytical methods and assumptions used in evaluating the RWE event are summarized in Reference 1. A
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                            B 3.3-46                                                                                                                                                                                              Revision No. 36 Control Rod Block Instrumentation B 3.3.2.1
 
BASES
 
APPLICABLE                                                                      1. Rod Block Monitor  (continued)
SAFETY ANALYSES, LCO, and                                                                                          statistical analysis of RWE events was performed to APPLICABILITY                                        determine the RBM response for both channels for each event.
From these responses, the fuel thermal performance as a function of RBM Allowable Value was determined. The Allowable Values are chosen as a function of power level.
The Allowable Values are specified in the CORE OPERATING LIMITS REPORT (COLR). Based on the specified Allowable Values, operating limits are established.
 
The RBM Function satisfies Criterion 3 of the NRC Policy Statement.
 
Two channels of the RBM are required to be OPERABLE, with their setpoints within the appropriate Allowable Values to ensure that no single instrument failure can preclude a rod block from this Function. The actual setpoints are calibrated consistent with applicable setpoint methodology.
 
Trip setpoints are specified in the setpoint calculations.
The trip setpoints are selected to ensure that the setpoints do not exceed the Allowable Values between successive CHANNEL CALIBRATIONS. Operation with a trip setting less conservative than the trip setpoint, but within its Allowable Value, is acceptable. Trip setpoints are those predetermined values of output at which an action should take place. The setpoints are compared to the actual process parameter (e.g., reactor power), and when the measured output value of the process parameter exceeds the setpoint, the associated device (e.g., trip unit) changes state. The analytic or design limits are derived from the limiting values of the process parameters obtained from the safety analysis or other appropriate documents. The Allowable Values are derived from the analytic or design limits, corrected for calibration, process, and instrument errors. The trip setpoints are determined from analytical or design limits, corrected for calibration, process, and instrument errors, as well as, instrument drift. In selected cases, the Allowable Values and trip setpoints are determined by engineering judgement or historically accepted practice relative to the intended function of the channel.
The trip setpoints determined in this manner provide adequate protection by assuring instrument and process uncertainties expected for the environments during the operating time of the channels are accounted for.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                            B 3.3-47                                                                                                                                                                                                        Revision No. 0 Control Rod Block Instrumentation B 3.3.2.1
 
BASES
 
APPLICABLE                                                                      1. Rod Block Monitor  (continued)
SAFETY ANALYSES, LCO, and                                                                                          The RBM is assumed to mitigate the consequences of an RWE APPLICABILITY                                        event when operating                                                                                                                                                                                                                    28.4% RTP. Below this power level, the consequences of an RWE event will not exceed the MCPR SL and, therefore, the RBM is not required to be OPERABLE.
Analyses (Ref. 1) have shown that with an intital MCPR greater than or equal to the limit specified in the COLR, no RWE event will result in exceeding the MCPR SL. Therefore, under these conditions, the RBM is also not required to be OPERABLE.
: 2. Rod Worth Minimizer
 
The RWM enforces the analyzed rod position sequence to ensure that the initial conditions of the CRDA analysis are not violated. The analytical methods and assumptions used in evaluating the CRDA are summarized in References 3, 4, 5, and 11. The analyzed rod position sequence requires that control rods be moved in groups, with all control rods assigned to a specific group required to be within specified banked positions. Requirements that the control rod sequence is in compliance with the analyzed rod position sequence are specified in LCO 3.1.6, "Rod Pattern Control."
 
When performing a shutdown of the plant, an optional control rod sequence (Ref. 11) may be used if the coupling of each withdrawn control rod has been confirmed. The rods may be inserted without the need to stop at intermediate positions.
When using the Reference 11 control rod insertion sequence for shutdown, the RWM may be reprogrammed to enforce the requirements of the improved control rod insertion process, or may be bypassed and the improved control rod shutdown sequence implemented under the controls in Condition D.
 
The RWM Function satisfies Criterion 3 of the NRC Policy Statement.
 
Since the RWM is a hardwired system designed to act as a backup to operator control of the rod sequences, only one channel of the RWM is available and required to be OPERABLE (Ref. 6). Special circumstances provided for in the Required Action of LCO 3.1.3, "Control Rod OPERABILITY," and LCO 3.1.6 may necessitate bypassing the RWM to allow continued operation with inoperable control rods, or to allow correction of a control rod pattern not in compliance with the analyzed rod position sequence. The RWM may be bypassed as required by these conditions, but then it must be considered inoperable and the Required Actions of this LCO followed.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                            B 3.3-48                                                                                                                                                                                    Revision No. 143 Control Rod Block Instrumentation B 3.3.2.1
 
BASES
 
APPLICABLE                                                                      2. Rod Worth Minimizer  (continued)
SAFETY ANALYSES, LCO, and                                                                                          Compliance with the analyzed rod position sequence, and APPLICABILITY                                        therefore OPERABILITY of the RWM, is required in MODES 1 and 2 when THERMAL POWER is < 10% RTP. When THERMAL POWER is > 10% RTP, there is no possible control rod configuration that results in a control rod worth that could exceed the 280 cal/gm fuel damage limit during a CRDA (Refs. 4 and 6).
In MODES 3 and 4, all control rods are required to be inserted into the core; therefore, a CRDA cannot occur. In MODE 5, since only a single control rod can be withdrawn from a core cell containing fuel assemblies, adequate SDM ensures that the consequences of a CRDA are acceptable, since the reactor will be subcritical.
: 3. Reactor Mode SwitchShutdown Position
 
During MODES 3 and 4, and during MODE 5 when the reactor mode switch is required to be in the shutdown position, the core is assumed to be subcritical; therefore, no positive reactivity insertion events are analyzed. The Reactor Mode Switch      Shutdown Position control rod withdrawal block ensures that the reactor remains subcritical by blocking control rod withdrawal, thereby preserving the assumptions of the safety analysis.
 
The Reactor Mode Switch      Shutdown Position Function satisfies Criterion 3 of the NRC Policy Statement.
 
Two channels are required to be OPERABLE to ensure that no single channel failure will preclude a rod block when required. There is no Allowable Value for this Function since the channels are mechanically actuated based solely on reactor mode switch position.
 
During shutdown conditions (MODE 3, 4, or 5), no positive reactivity insertion events are analyzed because assumptions are that control rod withdrawal blocks are provided to prevent criticality. Therefore, when the reactor mode switch is in the shutdown position, the control rod withdrawal block is required to be OPERABLE. During MODE 5 with the reactor mode switch in the refueling position, the refuel position one-rod-out interlock (LCO 3.9.2, "Refuel Position One-Rod-Out Interlock") provides the required control rod withdrawal blocks.
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                            B 3.3-49                                                                                                                                                                                              Revision No. 63 Control Rod Block Instrumentation B 3.3.2.1
 
BASES  (continued)
 
ACTIONS A.1
 
With one RBM channel inoperable, the remaining OPERABLE channel is adequate to perform the control rod block function; however, overall reliability is reduced because a single failure in the remaining OPERABLE channel can result in no control rod block capability for the RBM. For this reason, Required Action A.1 requires restoration of the inoperable channel to OPERABLE status. The Completion Time of 24 hours is based on the low probability of an event occurring coincident with a failure in the remaining OPERABLE channel.
 
B.1
 
If Required Action A.1 is not met and the associated Completion Time has expired, the inoperable channel must be placed in trip within 1 hour. If both RBM channels are inoperable, the RBM is not capable of performing its intended function; thus, one channel must also be placed in trip. This initiates a control rod withdrawal block, thereby ensuring that the RBM function is met.
 
The 1 hour Completion Time is intended to allow the operator time to evaluate and repair any discovered inoperabilities and is acceptable because it minimizes risk while allowing time for restoration or tripping of inoperable channels.
 
C.1, C.2.1.1, C.2.1.2, and C.2.2
 
With the RWM inoperable during a reactor startup, the operator is still capable of enforcing the prescribed control rod sequence. However, the overall reliability is reduced because a single operator error can result in violating the control rod sequence. Therefore, control rod movement must be immediately suspended except by scram.
Alternatively, startup may continue if at least 12 control rods have already been withdrawn, or a reactor startup with an inoperable RWM was not performed in the last 12 months.
These requirements minimize the number of reactor startups initiated with the RWM inoperable. Required Actions C.2.1.1 and C.2.1.2 require verification of these conditions by review of plant logs and control room indications. Once Required Action C.2.1.1 or C.2.1.2 is satisfactorily
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                            B 3.3-50                                                                                                                                                                                                        Revision No. 0 Control Rod Block Instrumentation B 3.3.2.1
 
BASES
 
ACTIONS                                                                                                    C.1, C.2.1.1, C.2.1.2, and C.2.2  (continued)
 
completed, control rod withdrawal may proceed in accordance with the restrictions imposed by Required Action C.2.2.
Required Action C.2.2 allows for the RWM Function to be performed manually and requires a double check of compliance with the prescribed rod sequence by a second licensed operator (Reactor Operator or Senior Reactor Operator) or other qualified member of the technical staff. The RWM may be bypassed under these conditions to allow continued operations. In addition, Required Actions of LCO 3.1.3 and LCO 3.1.6 may require bypassing the RWM, during which time the RWM must be considered inoperable with Condition C entered and its Required Actions taken.
 
D.1
 
With the RWM inoperable during a reactor shutdown, the operator is still capable of enforcing the prescribed control rod sequence. Required Action D.1 allows for the RWM Function to be performed manually and requires a double check of compliance with the prescribed rod sequence by a second licensed operator (Reactor Operator or Senior Reactor Operator) or other qualified member of the technical staff.
The RWM may be bypassed under these conditions to allow the reactor shutdown to continue.
 
E.1 and E.2
 
With one Reactor Mode Switch      Shutdown Position control rod withdrawal block channel inoperable, the remaining OPERABLE channel is adequate to perform the control rod withdrawal block function. However, since the Required Actions are consistent with the normal action of an OPERABLE Reactor Mode Switch      Shutdown Position Function (i.e., maintaining all control rods inserted), there is no distinction between having one or two channels inoperable.
 
In both cases (one or both channels inoperable), suspending all control rod withdrawal and initiating action to fully insert all insertable control rods in core cells containing one or more fuel assemblies will ensure that the core is subcritical with adequate SDM ensured by LCO 3.1.1. Control rods in core cells containing no fuel assemblies do not
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                            B 3.3-51                                                                                                                                                                                                        Revision No. 0 Control Rod Block Instrumentation B 3.3.2.1
 
BASES
 
ACTIONS                                                                                                    E.1 and E.2  (continued)
 
affect the reactivity of the core and are therefore not required to be inserted. Action must continue until all insertable control rods in core cells containing one or more fuel assemblies are fully inserted.
 
SURVEILLANCE                                                  As noted at the beginning of the SRs, the SRs for each REQUIREMENTS                                                  Control Rod Block instrumentation Function are found in the SRs column of Table 3.3.2.1-1.
 
The Surveillances are modified by a Note to indicate that when an RBM channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours provided the associated Function maintains control rod block capability. Upon completion of the Surveillance, or expiration of the 6 hour allowance, the channel must be returned to OPERABLE status or the applicable Condition entered and Required Actions taken.
This Note is based on the reliability analysis (Refs. 8, 9,
                                  & 10) assumptions of the average time required to perform channel surveillances. That analysis demonstrated that the 6 hour testing allowance does not significantly reduce the probability that a control rod block will be initiated when necessary.
 
SR  3.3.2.1.1
 
A CHANNEL FUNCTIONAL TEST is performed for each RBM channel to ensure that the entire channel will perform the intended function. Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                            B 3.3-52                                                                                                                                                                                              Revision No. 86 Control Rod Block Instrumentation B 3.3.2.1
 
BASES
 
SURVEILLANCE                                                  SR  3.3.2.1.2 and SR  3.3.2.1.3 REQUIREMENTS (continued)                                        A CHANNEL FUNCTIONAL TEST is performed for the RWM to ensure that the entire system will perform the intended function.
The CHANNEL FUNCTIONAL TEST for the RWM is performed by withdrawing a control rod not in compliance with the prescribed sequence and verifying a control rod block occurs. It is permissible to simulate the withdrawn control rod condition into the RWM in order to verify a control rod block occurs. SR 3.3.2.1.2 is performed during a startup and SR 3.3.2.1.3 is performed during a shutdown (or power reduction to                                                                                                                                    10% RTP). As noted in the SRs, SR 3.3.2.1.2 is not required to be performed until 1 hour after any control rod is withdrawn at                                                                                                                                                                                                                                                                                          10% RTP in MODE 2. As noted, SR 3.3.2.1.3 is not required to be performed until 1 hour after THERMAL POWER is                                                                                                                                                                                                                                        10% RTP in MODE 1. This allows entry at                                                                                            10% RTP in MODE 2 for SR 3.3.2.1.2 and entry into MODE 1 when THERMAL POWER is                                                                                                                                                                                                                                                                                                    10% RTP for SR 3.3.2.1.3 to perform the required Surveillance if the Frequency is not met per SR 3.0.2. The 1 hour allowance is based on operating experience and in consideration of providing a reasonable time in which to complete the SRs. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
 
SR  3.3.2.1.4
 
The RBM setpoints are automatically varied as a function of power. Three Allowable Values are specified in the COLR, each within a specific power range. The power at which the control rod block Allowable Values automatically change are based on the APRM signal's input to each RBM channel. Below the minimum power setpoint, the RBM is automatically bypassed. These power Allowable Values must be verified using a simulated or actual signal periodically to be less than or equal to the specified values. If any power range setpoint is nonconservative, then the affected RBM channel is considered inoperable. Alternatively, the power range
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                    B 3.3-53                                                                                                                                                                                              Revision No. 86 Control Rod Block Instrumentation B 3.3.2.1
 
BASES
 
SURVEILLANCE                                                  SR  3.3.2.1.4  (continued)
REQUIREMENTS channel can be placed in the conservative condition (i.e.,
enabling the proper RBM setpoint). If placed in this condition, the SR is met and the RBM channel is not considered inoperable. As noted, neutron detectors are excluded from the Surveillance because they are passive devices, with minimal drift, and because of the difficulty of simulating a meaningful signal.
 
Neutron detectors are adequately tested in SR 3.3.1.1.2 and SR 3.3.1.1.8. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
 
SR  3.3.2.1.5
 
A CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor. This test verifies the channel responds to the measured parameter within the necessary range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drifts between successive calibrations consistent with the plant specific setpoint methodology.
 
As noted, neutron detectors are excluded from the CHANNEL CALIBRATION because they are passive devices, with minimal drift, and because of the difficulty of simulating a meaningful signal. Neutron detectors are adequately tested in SR 3.3.1.1.2 and SR 3.3.1.1.8. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                            B 3.3-54                                                                                                                                                                                              Revision No. 86 Control Rod Block Instrumentation B 3.3.2.1
 
BASES
 
SURVEILLANCE                                                  SR  3.3.2.1.6 REQUIREMENTS (continued)                                        The RWM is automatically bypassed when power is above a specified value. This automatic action can itself be bypassed to allow for control rod sequence enforcement up to 100% RTP. The power level is determined from feedwater flow and steam flow signals. The automatic bypass setpoint must be verified periodically to be > 10% RTP. If the RWM low power setpoint is nonconservative, then the RWM is considered inoperable. Alternately, the low power setpoint channel can be placed in the conservative condition (nonbypass). If placed in the nonbypassed condition, the SR is met and the RWM is not considered inoperable. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
 
SR  3.3.2.1.7
 
A CHANNEL FUNCTIONAL TEST is performed for the Reactor Mode Switch      Shutdown Position Function to ensure that the entire channel will perform the intended function. The CHANNEL FUNCTIONAL TEST for the Reactor Mode Switch      Shutdown Position Function is performed by attempting to withdraw any control rod with the reactor mode switch in the shutdown position and verifying a control rod block occurs.
 
As noted in the SR, the Surveillance is not required to be performed until 1 hour after the reactor mode switch is in the shutdown position, since testing of this interlock with the reactor mode switch in any other position cannot be performed without using jumpers, lifted leads, or movable links. This allows entry into MODES 3 and 4 if the Frequency is not met per SR 3.0.2. The 1 hour allowance is based on operating experience and in consideration of providing a reasonable time in which to complete the SR.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                            B 3.3-55                                                                                                                                                                                              Revision No. 86 Control Rod Block Instrumentation B 3.3.2.1
 
BASES
 
SURVEILLANCE                                                  SR  3.3.2.1.7  (continued)
REQUIREMENTS The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
 
SR  3.3.2.1.8
 
The RWM will only enforce the proper control rod sequence if the rod sequence is properly input into the RWM computer.
This SR ensures that the proper sequence is loaded into the RWM so that it can perform its intended function. The Surveillance is performed once prior to declaring RWM OPERABLE following loading of sequence into RWM, since this is when rod sequence input errors are possible.
 
REFERENCES                                                                      1.                    NEDC-32162-P, "Maximum Extended Load Line Limit and ARTS Improvement Program Analysis for Peach Bottom Atomic Power Station, Units 2 and 3," Revision 1, February 1993.
: 2.                    UFSAR, Sections 7.10.3.4.8 and 7.16.3.
: 3.                    NEDE-24011-P-A, "General Electric Standard Application for Reactor Fuel," latest approved revision.
: 4.                    "Modifications to the Requirements for Control Rod Drop Accident Mitigating Systems," BWR Owners' Group, July 1986.
: 5.                    NEDO-21231, "Banked Position Withdrawal Sequence,"
January 1977.
: 6.                    NRC SER, "Acceptance of Referencing of Licensing Topical Report NEDE-24011-P-A," "General Electric Standard Application for Reactor Fuel, Revision 8, Amendment 17," December 27, 1987.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                            B 3.3-56                                                                                                                                                                                              Revision No. 86 Control Rod Block Instrumentation B 3.3.2.1
 
BASES
 
REFERENCES                                                                      7.                    NEDC-30851-P-A, "Technical Specification Improvement (continued)                                                                                  Analysis for BWR Control Rod Block Instrumentation,"
October 1988.
: 8.                    GENE-770-06-1, "Addendum to Bases for Changes to Surveillance Test Intervals and Allowed Out-of-Service Times for Selected Instrumentation Technical Specifications," February 1991.
: 9.                    NEDC-32410P-A, Nuclear Measurement Analysis and Control Power Range Neutron Monitor (NUMAC PRNM)
Retrofit Plus Option III Stability Trip Function, March 1995.
: 10.          NEDC-32410P Supplement 1, Nuclear Measurement Analysis and Control Power Range Neutron Monitor (NUMAC PRNM) Retrofit Plus Option III Stability Trip Function, Supplement 1, November 1997.
: 11.          NEDO-33091-A, Improved BPWS Control Rod Insertion Process, Revision 2, July 2004
 
PBAPS UNIT 2                                                                                                                                                                                                                            B 3.3-57                                                                                                                                                                                              Revision No. 61 Feedwater and Main Turbine High Water Level Trip Instrumentation B 3.3.2.2
 
B 3.3  INSTRUMENTATION
 
B 3.3.2.2  Feedwater and Main Turbine High Water Level Trip Instrumentation
 
BASES
 
BACKGROUND                                                                      The feedwater and main turbine high water level trip instrumentation is designed to detect a potential failure of the Feedwater Level Control System that causes excessive feedwater flow.
 
With excessive feedwater flow, the water level in the reactor vessel rises toward the high water level setpoint, causing the trip of the three feedwater pump turbines and the main turbine.
 
Digital Feedwater Control System (DFCS) high water level signals are provided by six level sensors, three narrow range and three wide range. The three narrow range level transmitters are used to satisfy the TS requirement. The three level sensors sense the difference between the pressure due to a constant column of water (reference leg) and the pressure due to the actual water level in the reactor vessel (variable leg). The three level signals are input into two independent and redundant digital control systems within the DFCS. Each control system includes redundant controllers capable of performing the high level trip function. All three level signals are used by the digital control systems to produce a validated level signal for use for the high level trip function.
 
Each independent digital control system has two redundant digital outputs (channels) to provide redundant signals to an associated trip system. Each independent digital control system processes input signals and compares them to pre-established setpoints. When the setpoint is exceeded, the two digital outputs actuate two contacts arranged in parallel so that either digital output can trip the associated trip system. The tripping of both digital trip systems will initiate a trip of the feedwater pump turbines and the main turbine.
 
A trip of the feedwater pump turbines limits further increase in reactor vessel water level by limiting further addition of feedwater to the reactor vessel. A trip of the main turbine and closure of the stop valves protects the turbine from damage due to water entering the turbine.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                            B 3.3-58                                                                                                                                                                                    Revision No. 146 Feedwater and Main Turbine High Water Level Trip Instrumentation B 3.3.2.2
 
BASES  (continued)
 
APPLICABLE                                                                      The feedwater and main turbine high water level trip SAFETY ANALYSES                    instrumentation is assumed to be capable of providing a turbine trip in the design basis transient analysis for a feedwater controller failure, maximum demand event (Ref. 1).
The high water level trip indirectly initiates a reactor scram from the main turbine trip (above 26.3% RTP) and trips the feedwater pumps, thereby terminating the event.
The reactor scram mitigates the reduction in MCPR.
 
Feedwater and main turbine high water level trip instrumentation satisfies Criterion 3 of the NRC Policy Statement.
 
LCO                                                                                                                                            The LCO requires two DFCS channels per trip system of high water level trip instrumentation to be OPERABLE to ensure the feedwater pump turbines and main turbine will trip on a valid reactor vessel high water level signal. Two DFCS channels (one per trip system) are needed to provide trip signals in order for the feedwater and main turbine trips to occur.
 
Two level signals are also required to ensure a single sensor failure will not prevent the trips of the feedwater pump turbines and main turbine when reactor vessel water level is at the high water level reference point.
 
Each channel must have its setpoint set within the specified Allowable Value of SR 3.3.2.2.3. The Allowable Value is set to ensure that the thermal limits are not exceeded during the event. The actual setpoint is calibrated to be consistent with the applicable setpoint methodology assumptions. Trip setpoints are specified in the setpoint calculations. The trip setpoints are selected to ensure that the setpoints do not exceed the Allowable Value between successive CHANNEL CALIBRATIONS. Operation with a trip setting less conservative than the trip setpoint, but within its Allowable Value, is acceptable.
 
Trip setpoints are those predetermined values of output at which an action should take place. The setpoints are compared to the actual process parameter (e.g., reactor vessel water level), and when the measured output value of the process parameter exceeds the setpoint, the associated device (e.g., trip unit) changes state. The analytic or design limits are derived from the limiting values of the process parameters obtained from the safety analysis or
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                            B 3.3-59                                                                                                                                                                                    Revision No. 143 Feedwater and Main Turbine High Water Level Trip Instrumentation B 3.3.2.2
 
BASES
 
LCO                                                                                                                                                      other appropriate documents. The Allowable Values are (continued)                                                  derived from the analytic or design limits, corrected for calibration, process, and instrument errors. A channel is inoperable if its actual trip setting is not within its required Allowable Value. The trip setpoints are determined from analytical or design limits, corrected for calibration, process and instrument errors, as well as, instrument drift.
The trip setpoints determined in this manner provide adequate protection by assuring instrument and process uncertainties expected for the environment during the operating time for the associated channels are accounted for.
 
APPLICABILITY                                                  The feedwater and main turbine high water level trip instrumentation is required to be OPERABLE at  22.6% RTP to ensure that the fuel cladding integrity Safety Limit and the cladding 1% plastic strain limit are not violated during the feedwater controller failure, maximum demand event. As discussed in the Bases for LCO 3.2.3, "LINEAR HEAT GENERATION RATE (LHGR)," and LCO 3.2.2, "MINIMUM CRITICAL POWER RATIO (MCPR)," sufficient margin to these limits exists below 22.6% RTP; therefore, these requirements are only necessary when operating at or above this power level.
 
ACTIONS                                                                                                              A Note has been provided to modify the ACTIONS related to feedwater and main turbine high water level trip instrumentation channels. 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 inoperable feedwater and main turbine high water level trip instrumentation channels provide appropriate compensatory measures for separate inoperable channels. As such, a Note has been provided that allows separate Condition entry for each inoperable feedwater and main turbine high water level trip instrumentation channel.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                            B 3.3-60                                                                                                                                                                                    Revision No. 143 Feedwater and Main Turbine High Water Level Trip Instrumentation B 3.3.2.2
 
BASES
 
ACTIONS                                                                                                    A.1 (continued)
With one or more feedwater and main turbine high water level trip channels inoperable, but with feedwater and main turbine high water level trip capability maintained (refer to Required Action B.1 Bases), the remaining OPERABLE channels can provide the required trip signal. However, overall instrumentation reliability is reduced because a single active instrument failure in one of the remaining channels may result in the instrumentation not being able to perform its intended function. Therefore, continued operation is only allowed for a limited time with one or more channels inoperable. If the inoperable channels cannot be restored to OPERABLE status within the Completion Time, the channels must be placed in the tripped condition per Required Action A.1. Placing the inoperable channel in trip would conservatively compensate for the inoperability, restore capability to accommodate a single active instrument failure, and allow operation to continue with no further restrictions. Alternately, if it is not desired to place the channel in trip (e.g., as in the case where placing the inoperable channel in trip would result in the feedwater and main turbine trip), Condition C must be entered and its Required Action taken.
 
The Completion Time of 72 hours is based on the low probability of the event occurring coincident with a single failure in a remaining OPERABLE channel. Alternatively, a Completion Time can be determined in accordance with the Risk Informed Completion Time (RICT) Program. A Note has been provided to indicate that a RICT is only applicable when a loss of function has not occurred.
 
B.1
 
Required Action B.1 is intended to ensure that appropriate actions are taken if multiple, inoperable, untripped channels result in the High Water Level Function of DFCS not maintaining feedwater and main turbine trip capability. In this condition, the feedwater and main turbine high water level trip instrumentation cannot perform its design function. Therefore, continued operation is only permitted for a 2 hour period, during which feedwater and main turbine high water level trip capability must be restored. The trip capability is considered maintained when sufficient channels are OPERABLE or in trip such that the feedwater and main turbine high water level trip logic will generate a trip
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                            B 3.3-61                                                                                                                                                                                    Revision No. 159 Feedwater and Main Turbine High Water Level Trip Instrumentation B 3.3.2.2
 
BASES
 
ACTIONS                                                                                                    B.1  (continued) signal on a valid signal. This requires one channel per trip system to be OPERABLE or in trip. If the required channels cannot be restored to OPERABLE status or placed in trip, Condition C must be entered and its Required Action taken.
The 2 hour Completion Time is sufficient for the operator to take corrective action, and takes into account the likelihood of an event requiring actuation of feedwater and main turbine high water level trip instrumentation occurring during this period. It is also consistent with the 2 hour Completion Time provided in LCO 3.2.2 for Required Action A.1, since this instrumentation's purpose is to preclude a MCPR violation.
 
C.1 and C.2 With any Required Action and associated Completion Time not met, the plant must be brought to a MODE or other specified condition in which the LCO does not apply. To achieve this status, THERMAL POWER must be reduced to < 22.6% RTP within 4 hours. Alternatively, the affected feedwater pump(s) and affected main turbine valve(s) may be removed from service since this performs the intended function of the instrumentation. As discussed in the Applicability section of the Bases, operation below 22.6% RTP results in sufficient margin to the required limits, and the feedwater and main turbine high water level trip instrumentation is not required to protect fuel integrity during the feedwater controller failure, maximum demand event. The allowed Completion Time of 4 hours is based on operating experience to reduce THERMAL POWER to < 22.6% RTP from full power conditions in an orderly manner and without challenging plant systems.
Required Action C.1 is modified by a Note which states that the Required Action is only applicable if the inoperable channel is the result of an inoperable feedwater pump turbine or main turbine stop valve. The Note clarifies the situations under which the associated Required Action would be the appropriate Required Action.
 
SURVEILLANCE                                                  The Surveillances are modified by a Note to indicate that REQUIREMENTS                                                  when a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours provided the associated Function maintains feedwater and main turbine high water level trip capability. Upon completion of the Surveillance, or expiration of the 6 hour allowance, the channel must be returned to OPERABLE status or the applicable Condition entered and Required Actions taken. This Note is based on the reliability analysis (Ref. 2) assumption of the average (continued) time required to perform
 
PBAPS UNIT 2                                                                                                                                                                                                                            B 3.3-62                                                                                                                                                                                    Revision No. 143 Feedwater and Main Turbine High Water Level Trip Instrumentation B 3.3.2.2
 
BASES
 
SURVEILLANCE                                                  channel Surveillance. That analysis demonstrated that the REQUIREMENTS                                                  6 hour testing allowance does not significantly reduce the (continued)                                        probability that the feedwater pump turbines and main turbine will trip when necessary.
 
SR  3.3.2.2.1
 
Performance of the CHANNEL CHECK once every 24 hours ensures that a gross failure of instrumentation has not occurred. A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a similar parameter on other channels. The CHANNEL CHECK may be performed by comparing indication or by verifying the absence of the DFCS "TROUBLE" alarm in the control room. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value. Significant deviations between instrument channels could be an indication of excessive instrument drift in one of the channels, or something even more serious. A CHANNEL CHECK will detect gross channel failure; thus, it is key to verifying the instrumentation continues to operate properly between each CHANNEL CALIBRATION.
 
Agreement criteria are determined by the plant staff based on a combination of the channel instrument uncertainties, including indication and readability. If a channel is outside the criteria, it may be an indication that the instrument has drifted outside its limits.
 
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. The CHANNEL CHECK supplements less formal, but more frequent, checks of channel status during normal operational use of the displays associated with the channels required by the LCO.
 
SR  3.3.2.2.2
 
A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the entire channel will perform the intended function. Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                            B 3.3-63                                                                                                                                                                                              Revision No. 86 Feedwater and Main Turbine High Water Level Trip Instrumentation B 3.3.2.2
 
BASES
 
SURVEILLANCE                                                  SR  3.3.2.2.3 REQUIREMENTS (continued)                                        CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor. This test verifies the channel responds to the measured parameter within the necessary range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drifts between successive calibrations, consistent with the assumptions of the current plant specific setpoint methodology.
 
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
 
SR  3.3.2.2.4
 
The LOGIC SYSTEM FUNCTIONAL TEST demonstrates the OPERABILITY of the required trip logic for a specific channel. The system functional test of the feedwater and main turbine stop valves is included as part of this Surveillance and overlaps the LOGIC SYSTEM FUNCTIONAL TEST to provide complete testing of the assumed safety function.
Therefore, if a stop valve is incapable of operating, the associated instrumentation channels would be inoperable.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
 
REFERENCES                                                                      1.                    UFSAR, Section 14.5.2.2.
: 2.                    GENE-770-06-1, "Bases for Changes to Surveillance Test Intervals and Allowed Out-Of-Service Times for Selected Instrumentation Technical Specifications,"
February 1991.
: 3.                    NEDC-33873P, Safety Analysis Report for Peach Bottom Atomic Power Station, Units 2 and 3, Thermal Power Optimization, Revision 0.
 
PBAPS UNIT 2                                                                                                                                                                                                                            B 3.3-64                                                                                                                                                                                    Revision No. 143 PAM Instrumentation B 3.3.3.1
 
B 3.3  INSTRUMENTATION
 
B 3.3.3.1  Post Accident Monitoring (PAM) Instrumentation
 
BASES
 
BACKGROUND                                                                      The primary purpose of the PAM instrumentation is to display plant variables that provide information required by the control room operators during accident situations. This information provides the necessary support for the operator to take the manual actions for which no automatic control is provided and that are required for safety systems to accomplish their safety functions for Design Basis Events.
The instruments that monitor these variables are designated as Type A, Category I, and non-Type A, Category I, in accordance with Regulatory Guide 1.97 (Ref. 1).
 
The OPERABILITY of the accident monitoring instrumentation ensures that there is sufficient information available on selected plant parameters to monitor and assess plant status and behavior following an accident. This capability is consistent with the recommendations of Reference 1.
 
APPLICABLE                                                                      The PAM instrumentation LCO ensures the OPERABILITY of SAFETY ANALYSES                    Regulatory Guide 1.97, Type A variables so that the control room operating staff can:
 
Perform the diagnosis specified in the Emergency Operating Procedures (EOPs). These variables are restricted to preplanned actions for the primary success path of Design Basis Accidents (DBAs), (e.g.,
loss of coolant accident (LOCA)), and
 
Take the specified, preplanned, manually controlled actions for which no automatic control is provided, which are required for safety systems to accomplish their safety function.
 
The PAM instrumentation LCO also ensures OPERABILITY of Category I, non-Type A, variables so that the control room operating staff can:
 
Determine whether systems important to safety are performing their intended functions;
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                            B 3.3-65                                                                                                                                                                                                        Revision No. 0 PAM Instrumentation B 3.3.3.1
 
BASES
 
APPLICABLE                                                                                                                    Determine the potential for causing a gross breach of SAFETY ANALYSES                                                              the barriers to radioactivity release; (continued)
Determine whether a gross breach of a barrier has occurred; and
 
Initiate action necessary to protect the public and for an estimate of the magnitude of any impending threat.
 
The plant specific Regulatory Guide 1.97 Analysis (Refs. 2, 3, and 4) documents the process that identified Type A and Category I, non-Type A, variables.
 
Accident monitoring instrumentation that satisfies the definition of Type A in Regulatory Guide 1.97 meets Criterion 3 of the NRC Policy Statement. Category I, non-Type A, instrumentation is retained in Technical Specifications (TS) because they are intended to assist operators in minimizing the consequences of accidents.
Therefore, these Category I variables are important for reducing public risk.
 
LCO                                                                                                                                                LCO 3.3.3.1 requires two OPERABLE channels for all but one Function to ensure that no single failure prevents the operators from being presented with the information necessary to determine the status of the plant and to bring the plant to, and maintain it in, a safe condition following that accident. Furthermore, provision of two channels allows a CHANNEL CHECK during the post accident phase to confirm the validity of displayed information.
 
The exception to the two channel requirement is primary containment isolation valve (PCIV) position. In this case, the important information is the status of the primary containment penetrations. The LCO requires one position indicator for each active PCIV. This is sufficient to redundantly verify the isolation status of each isolable penetration either via indicated status of the active valve and prior knowledge of passive valve or via system boundary status. If a normally active PCIV is known to be closed and deactivated, position indication is not needed to determine status. Therefore, the position indication for valves in this state is not required to be OPERABLE.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                            B 3.3-66                                                                                                                                                                                                        Revision No. 0
 
PAM Instrumentation B 3.3.3.1
 
BASES
 
LCO                                                                                                                                            The following list is a discussion of the specified (continued)                                                  instrument Functions listed in Table 3.3.3.1-1 in the accompanying LCO.
: 1. Reactor Pressure
 
Instruments:                              PR-2-2-3-404 A, B
 
Reactor pressure is a Category I variable provided to support monitoring of Reactor Coolant System (RCS) integrity and to verify operation of the Emergency Core Cooling Systems (ECCS). Two independent pressure transmitters with a range of 0 psig to 1500 psig monitor pressure and associated independent wide range recorders are the primary indication used by the operator during an accident.
Therefore, the PAM Specification deals specifically with this portion of the instrument channel.
 
2, 3. Reactor Vessel Water Level (Wide Range and Fuel Zone)
 
Instruments:                              Wide Range: LR-2-2-3-110 A, B (Green Pen)
Fuel Zone:  LR-2-2-3-110 A, B (Blue Pen)
 
Reactor vessel water level is a Category I variable provided to support monitoring of core cooling and to verify operation of the ECCS. The wide range and fuel zone water level channels provide the PAM Reactor Vessel Water Level Functions. The ranges of the wide range water level channels and the fuel zone water level channels overlap to cover a range of -325 inches (just below the bottom of the active fuel) to +50 inches (above the normal water level).
Reactor vessel water level is measured by separate differential pressure transmitters. The output from these channels is recorded on two independent pen recorders, which is the primary indication used by the operator during an accident. Each recorder has two channels, one for wide range reactor vessel water level and one for fuel zone reactor vessel water level. Therefore, the PAM Specification deals specifically with these portions of the instrument channels.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                            B 3.3-67                                                                                                                                                                                                        Revision No. 7 PAM Instrumentation B 3.3.3.1
 
BASES
 
LCO                                                                                                                                            4. Suppression Chamber Water Level (Wide Range)
(continued)
Instruments:                              LR-8123 A, B
 
Suppression chamber water level is a Category I variable provided to detect a breach in the reactor coolant pressure boundary (RCPB). This variable is also used to verify and provide long term surveillance of ECCS function. The wide range suppression chamber water level measurement provides the operator with sufficient information to assess the status of both the RCPB and the water supply to the ECCS.
The wide range water level recorders monitor the suppression chamber water level from the bottom of the ECCS suction lines to five feet above normal water level. Two wide range suppression chamber water level signals are transmitted from separate differential pressure transmitters and are continuously recorded on two recorders in the control room.
These recorders are the primary indication used by the operator during an accident. Therefore, the PAM Specification deals specifically with this portion of the instrument channel.
 
5, 6. Drywell Pressure (Wide Range and Subatmospheric Range)
 
Instruments: Wide Range:          PR-8102 A, B (Red Pen)
Subatmospheric Range: PR-8102 A, B (Green Pen)
 
Drywell pressure is a Category I variable provided to detect breach of the RCPB and to verify ECCS functions that operate to maintain RCS integrity. The wide range and subatmospheric range drywell pressure channels provide the PAM Drywell Pressure Functions. The wide range and subatmospheric range drywell pressure channels overlap to cover a range of 5 psia to 225 psig (in excess of four times the design pressure of the drywell). Drywell pressure signals are transmitted from separate pressure transmitters and are continuously recorded and displayed on two independent control room recorders. Each recorder has two channels, one for wide range drywell pressure and one for subatmospheric range drywell pressure. These recorders are the primary indication used by the operator during an accident. Therefore, the PAM Specification deals specifically with these portions of the instrument channels.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                            B 3.3-68                                                                                                                                                                                                        Revision No. 3 PAM Instrumentation B 3.3.3.1
 
BASES
 
LCO                                                                                                                                            7. Drywell High Range Radiation (continued)
Instruments:                              RR-8103 A, B
 
Drywell high range radiation is a Category I variable provided to monitor the potential of significant radiation releases and to provide release assessment for use by operators in determining the need to invoke site emergency plans. Post accident drywell radiation levels are monitored by four instrument channels each with a range of 1 to 1x108 R/hr. These radiation monitors drive two dual channel recorders located in the control room. Each recorder and the two associated channels are in a separate division. As such, two recorders and two channels of radiation monitoring instrumentation (one per recorder) are required to be OPERABLE for compliance with this LCO. Therefore, the PAM Specification deals specifically with these portions of the instrument channels.
: 8. Primary Containment Isolation Valve (PCIV) Position
 
PCIV position is a Category I variable provided for verification of containment integrity. In the case of PCIV position, the important information is the isolation status of the containment penetration. The LCO requires one channel of valve position indication in the control room to be OPERABLE for each active PCIV in a containment penetration flow path, i.e., two total channels of PCIV position indication for a penetration flow path with two active valves. For containment penetrations with only one active PCIV having control room indication, Note (b) requires a single channel of valve position indication to be OPERABLE. This is sufficient to redundantly verify the isolation status of each isolable penetration via indicated status of the active valve, as applicable, and prior knowledge of passive valve or system boundary status. If a penetration flow path is isolated, position indication for the PCIV(s) in the associated penetration flow path is not needed to determine status. Therefore, the position indication for valves in an isolated penetration flow path is not required to be OPERABLE. The PCIV position PAM instrumentation consists of position switches, associated wiring and control room indicating lamps for active PCIVs (check valves and manual valves are not required to have position indication). Therefore, the PAM Specification deals specifically with these instrument channels.
 
Each penetration is treated separately and each penetration flow path is considered a separate function. Therefore, separate condition entry is allowed for each inoperable penetration flow path.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                            B 3.3-69                                                                                                                                                                                              Revision No. 57 PAM Instrumentation B 3.3.3.1
 
BASES
 
LCO                                                                                                                                            9, 10. Deleted (continued)
: 11. Suppression Chamber Water Temperature
 
Instruments:                              TR-8123 A, B TIS-2-2-71 A, B Recorders
 
Suppression chamber water temperature is a Category I variable provided to detect a condition that could potentially lead to containment breach and to verify the effectiveness of ECCS actions taken to prevent containment breach. The suppression chamber water temperature instrumentation allows operators to detect trends in suppression chamber water temperature in sufficient time to take action to prevent steam quenching vibrations in the suppression pool. Suppression chamber water temperature is monitored by two redundant channels. Each channel is assigned to a separate safeguard power division. Each channel consists of 13 resistance temperature detectors (RTDs) mounted in thermowells installed in the suppression chamber shell below the minimum water level, a processor, and control room recorders. The RTDs are mounted in each of 13 of the 16 segments of the suppression chamber. The RTD
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                            B 3.3-70                                                                                                                                                                                              Revision No. 55 PAM Instrumentation B 3.3.3.1
 
BASES  (continued)
 
LCO                                                                                                                                    11. Suppression Chamber Water Temperature (continued)
 
inputs are averaged by the processor to provide a bulk average temperature output to the associated control room recorder. The allowance that only 10 RTDs are required to be OPERABLE for a channel to be considered OPERABLE provided no 2 adjacent RTDs are inoperable is acceptable based on engineering judgement considering the temperature response profile of the suppression chamber water volume for previously analyzed events and the most challenging RTDs inoperable. These recorders are the primary indication used by the operator during an accident. Therefore, the PAM Specification deals specifically with this portion of the instrument channels. Four recorders are provided. A recorder in each division is required to be OPERABLE to satisfy the LCO.
 
APPLICABILITY                                        The PAM instrumentation LCO is applicable in MODES 1 and 2.
These variables are related to the diagnosis and preplanned actions required to mitigate DBAs. The applicable DBAs are assumed to occur in MODES 1 and 2. In MODES 3, 4, and 5, plant conditions are such that the likelihood of an event that would require PAM instrumentation is extremely low; therefore, PAM instrumentation is not required to be OPERABLE in these MODES.
 
ACTIONS
 
A Note has been provided to modify the ACTIONS related to PAM instrumentation channels. 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
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                            B 3.3-71                                                                                                                                                                                              Revision No. 52 PAM Instrumentation B 3.3.3.1
 
BASES
 
ACTIONS (continued)                                                  inoperable PAM instrumentation channels provide appropriate compensatory measures for separate Functions. As such, a Note has been provided that allows separate Condition entry for each inoperable PAM Function.
 
A.1
 
When one or more Functions have one required channel that is inoperable, the required inoperable channel must be restored to OPERABLE status within 30 days. The 30 day Completion Time is based on operating experience and takes into account the remaining OPERABLE channels (or, in the case of a Function that has only one required channel, other non-Regulatory Guide 1.97 instrument channels to monitor the Function), the passive nature of the instrument (no critical automatic action is assumed to occur from these instruments), and the low probability of an event requiring PAM instrumentation during this interval.
 
B.1
 
If a channel has not been restored to OPERABLE status in 30 days, this Required Action specifies initiation of action in accordance with Specification 5.6.6, which requires a written report to be submitted to the NRC. This report discusses the results of the root cause evaluation of the inoperability and identifies proposed restorative actions.
This action is appropriate in lieu of a shutdown requirement, since alternative actions are identified before loss of functional capability, and given the likelihood of plant conditions that would require information provided by this instrumentation.
 
C.1
 
When one or more Functions have two required channels that are inoperable (i.e., two channels inoperable in the same Function), one channel in the Function should be restored to OPERABLE status within 7 days. The Completion Time of 7 days is based on the relatively low probability of an event requiring PAM instrument operation and the availability of alternate means to obtain the required information. Continuous operation with two required
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                            B 3.3-72                                                                                                                                                                                                        Revision No. 3 PAM Instrumentation B 3.3.3.1
 
BASES
 
ACTIONS C.1 (continued)
 
channels inoperable in a Function is not acceptable because the alternate indications may not fully meet all performance qualification requirements applied to the PAM instrumentation. Therefore, requiring restoration of one inoperable channel of the Function limits the risk that the PAM Function will be in a degraded condition should an accident occur.
 
D.1
 
This Required Action directs entry into the appropriate Condition referenced in Table 3.3.3.1-1. The applicable Condition referenced in the Table is Function dependent.
Each time an inoperable channel has not met the Required Action of Condition C and the associated Completion Time has expired, Condition D is entered for that channel and provides for transfer to the appropriate subsequent Condition.
 
E.1
 
For the majority of Functions in Table 3.3.3.1-1, if the Required Action and associated Completion Time of Condition C is not met, the plant must be brought to a MODE in which the LCO not apply. To achieve this status, the plant must be brought to at least MODE 3 within 12 hours.
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.
 
F.1
 
Since alternate means of monitoring drywell high range radiation have been developed and tested, the Required Action is not to shut down the plant, but rather to follow the directions of Specification 5.6.6. These alternate means may be temporarily installed if the normal PAM channel cannot be restored to OPERABLE status within the allotted time. The report provided to the NRC should discuss the alternate means used, describe the degree to which the alternate means are equivalent to the installed PAM channels, justify the areas in which they are not equivalent, and provide a schedule for restoring the normal PAM channels.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                            B 3.3-73                                                                                                                                                                                                        Revision No. 3 PAM Instrumentation B 3.3.3.1
 
BASES  (continued)
 
SURVEILLANCE                                                  SR  3.3.3.1.1 REQUIREMENTS Performance of the CHANNEL CHECK once every 31 days ensures that a gross failure of instrumentation has not occurred. A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel against a similar parameter on other channels. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value. Significant deviations between instrument channels could be an indication of excessive instrument drift in one of the channels or something even more serious. A CHANNEL CHECK will detect gross channel failure; thus, it is key to verifying the instrumentation continues to operate properly between each CHANNEL CALIBRATION. The high radiation instrumentation should be compared to similar plant instruments located throughout the plant.
 
Agreement criteria are determined by the plant staff, based on a combination of the channel instrument uncertainties, including isolation, indication, and readability. If a channel is outside the criteria, it may be an indication that the sensor or the signal processing equipment has drifted outside its limit.
 
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. The CHANNEL CHECK supplements less formal, but more frequent, checks of channels during normal operational use of those displays associated with the channels required by the LCO.
 
SR  3.3.3.1.2 Deleted
 
SR  3.3.3.1.3
 
These SRs require CHANNEL CALIBRATIONs to be performed. A CHANNEL CALIBRATION is a complete check of the instrument loop, including the sensor. The test verifies the channel responds to measured parameter with the necessary range and accuracy. For the PCIV Position Function, the CHANNEL CALIBRATION consists of verifying the remote indication conforms to actual valve position.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                            B 3.3-74                                                                                                                                                                                              Revision No. 86 PAM Instrumentation B 3.3.3.1
 
BASES
 
SURVEILLANCE                                                  SR  3.3.3.1.3  (continued)
REQUIREMENTS The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
 
REFERENCES                                                                      1.                    Regulatory Guide 1.97, "Instrumentation for Light Water Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident,"
Revision 3, May 1983.
: 2.                    NRC Safety Evaluation Report, "Peach Bottom Atomic Power Station, Unit Nos. 2 and 3, Conformance to Regulatory Guide 1.97," January 15, 1988.
: 3.                    Letter from G. Y. Suh (NRC) to G. J. Beck (PECo) dated February 13, 1991 concerning "Conformance to Regulatory Guide 1.97 for Peach Bottom Atomic Power Station, Units 2 and 3".
: 4.                    Letter from S. Dembek (NRC) to G. A. Hunger (PECO Energy) dated March 7, 1994 concerning "Regulatory Guide 1.97 - Boiling Water Reactor Neutron Flux Monitoring, Peach Bottom Atomic Power Station (PBAPS),
Units 2 and 3".
 
PBAPS UNIT 2                                                                                                                                                                                                                            B 3.3-75                                                                                                                                                                                              Revision No. 86
 
Remote Shutdown System B 3.3.3.2
 
B 3.3  INSTRUMENTATION
 
B 3.3.3.2  Remote Shutdown System
 
BASES
 
BACKGROUND                                                                      The Remote Shutdown System provides the control room operator with sufficient instrumentation and controls to maintain the plant in a safe shutdown condition from a location other than the control room for at least one hour.
This capability is necessary to protect against the possibility of the control room becoming inaccessible. A safe shutdown condition is defined as MODE 3. With the plant in MODE 3, the Reactor Core Isolation Cooling (RCIC)
System and the safety/relief valves can be used to remove core decay heat and meet all safety requirements. The long term supply of water for the RCIC and the ability to control reactor pressure and level from outside the control room allow extended operation in MODE 3.
 
In the event that the control room must be abandoned, a reactor trip and MSIV closure is assumed to have been initiated from the control room prior to abandonment. For the design event, it is assumed the loss of feedwater (as a result of MSIV closure) causes an automatic start of RCIC due to low reactor level. Although HPCI also typically initiates on low reactor level, it is conservatively assumed that it does not start for the design event due to damage in the control room. No LOOP, accident condition or other failures are assumed. At the remote shutdown panel, reactor level and pressure is maintained with RCIC and operation of SRVs H, E and L. SRV operation maintains pressure below the SRV lift setpoint and transfers decay heat to the suppression pool. This can be maintained for at least one hour without suppression pool cooling. If control room access cannot be regained in one hour, procedures provide direction to bring the plant to cold shutdown.
 
The OPERABILITY of the Remote Shutdown System ensures there are sufficient controls and information available for those plant parameters necessary to maintain the plant in MODE 3 for at least one hour. Other controls and indication on the remote shutdown panel are provided, but they are not required for OPERABILITY.
 
APPLICABLE                                                                      The Remote Shutdown System is required to provide SAFETY ANALYSES                    instrumentation and controls at appropriate locations outside the control room with a design capability to control reactor pressure and level, including the necessary instrumentation and controls, to maintain the plant in a safe condition in MODE 3.
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                            B 3.3-76                                                                                                                                                                                    Revision No. 132
 
Remote Shutdown System B 3.3.3.2
 
BASES
 
APPLICABLE                                                                      The criteria governing the design and the specific system SAFETY ANALYSES                    requirements of the Remote Shutdown System are located in (continued)                                        the UFSAR (Refs. 1 and 2).
 
The Remote Shutdown System is considered an important contributor to reducing the risk of accidents; as such, it meets Criterion 4 of the NRC Policy Statement.
 
LCO                                                                                                                                            The Remote Shutdown System LCO provides the requirements for the OPERABILITY of the instrumentation and controls necessary to maintain the plant in MODE 3 from a location other than the control room. The instrumentation and controls required are listed in Table B 3.3.3.2-1.
 
The controls, instrumentation, and transfer switches are those required for:
 
Reactor pressure vessel (RPV) pressure control;
 
Decay heat removal; and
 
RPV inventory control
 
The Remote Shutdown System is OPERABLE if all instrument and control channels needed to support the remote shutdown function are OPERABLE.
 
The Remote Shutdown System instruments and control circuits covered by this LCO do not need to be energized to be considered OPERABLE. This LCO is intended to ensure that the instruments and control circuits will be OPERABLE if plant conditions require that the Remote Shutdown System be placed in operation.
 
APPLICABILITY                                        The Remote Shutdown System LCO is applicable in MODES 1 and 2. This is required so that the plant can be maintained in MODE 3 for an extended period of time from a location other than the control room.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                            B 3.3-77                                                                                                                                                                                    Revision No. 132
 
Remote Shutdown System B 3.3.3.2
 
BASES
 
APPLICABILITY                                        This LCO is not applicable in MODES 3, 4, and 5. In these (continued)                                        MODES, the plant is already subcritical and in a condition of reduced Reactor Coolant System energy. Under these conditions, considerable time is available to restore necessary instrument control Functions if control room instruments or control becomes unavailable. Consequently, the TS do not require OPERABILITY in MODES 3, 4, and 5.
 
ACTIONS
 
A Note has been provided to modify the ACTIONS related to Remote Shutdown System Functions. 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 inoperable Remote Shutdown System Functions provide appropriate compensatory measures for separate Functions.
As such, a Note has been provided that allows separate Condition entry for each inoperable Remote Shutdown System Function.
 
A.1
 
Condition A addresses the situation where one or more required Functions of the Remote Shutdown System is inoperable. This includes the control and transfer switches for any required function.
 
The Required Action is to restore the Function (all required channels) to OPERABLE status within 30 days. The Completion Time is based on operating experience and the low probability of an event that would require evacuation of the control room.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                            B 3.3-78                                                                                                                                                                                              Revision No. 52 Remote Shutdown System B 3.3.3.2
 
BASES
 
ACTIONS B.1 (continued)
If the Required Action and associated Completion Time of Condition A are not met, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 12 hours. The allowed Completion Time is reasonable, based on operating experience, to reach the required MODE from full power conditions in an orderly manner and without challenging plant systems.
 
SURVEILLANCE                                                  SR  3.3.3.2.1 REQUIREMENTS SR 3.3.3.2.1 verifies that each instrument and control circuit in Table B 3.3.3.2-1 performs the intended function.
This verification is performed from the remote shutdown panel and locally, if necessary. Operation of equipment from the remote shutdown panel is not necessary. The Surveillance can be satisfied by performance of a continuity check of the circuitry. This will ensure that if the control room becomes inaccessible, the plant can be maintained in MODE 3 from the remote shutdown panel. Each required transfer switch and circuit is limited to those that are necessary to maintain reactor level and pressure from the remote shutdown panel during operation in Mode 3.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
 
SR  3.3.3.2.2
 
CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor. The test verifies the channel responds to measured parameter values with the necessary range and accuracy. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
 
REFERENCES                                                                      1.                    UFSAR, Section 1.5.1.
: 2.                    UFSAR, Section 7.18.
: 3. Drawing E-540-13.
: 4.            IR 2556042.
 
PBAPS UNIT 2                                                                                                                                                                                                                            B 3.3-79                                                                                                                                                                                    Revision No. 132 Remote Shutdown System B 3.3.3.2
 
Table B 3.3.3.2-1 (page 1 of 3)
Remote Shutdown System Instrumentation
 
FUNCTION                                                                                                                                                                                                                                                                                                                            REQUIRED NUMBER OF CHANNELS
 
Instrument Parameter
: 1. Reactor Pressure                                                                                                                      2
: 2.                    Reactor Level (Wide Range)                                                                                                                                                                                                                                                                                    2
: 3. Torus Temperature                                                                                                                    2
: 4. Torus Level                                                                                                                                        1
: 5.                    Condensate Storage Tank Level                                                                                                                                                                                                                                                      1
: 6. RCIC Flow                                                                                                                                              1
: 7.                    RCIC Turbine Speed                                                                                                                                                                                                                                                                                                                                                                    1
: 8.                    RCIC Pump Suction Pressure                                                                                                                                                                                                                                                                                    1
: 9.                    RCIC Pump Discharge Pressure                                                                                                                                                                                                                                                                1
: 10.          RCIC Turbine Supply Pressure                                                                                                                                                                                                                                                                1
: 11.          RCIC Turbine Exhaust Pressure                                                                                                                                                                                                                                                      1
: 12. Drywell Pressure                                                                                                                                                                                      1
 
Transfer/Control Parameter
: 13.          RCIC Pump Flow                                                                                                                                                                                                                                                                                                                                                                                                            1
: 14.          RCIC Drain Isolation to Radwaste                                                                                                                                                                                                                        1
: 15.          RCIC Steam Pot Drain Steam Trap Bypass                                                                                                                                                            1
: 16.          RCIC Drain Isolation to Main Condenser                                                                                                                                                            1
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                            B 3.3-80                                                                                                                                                                                    Revision No. 132 Remote Shutdown System B 3.3.3.2
 
Table B 3.3.3.2-1 (page 2 of 3)
Remote Shutdown System Instrumentation
 
FUNCTION                                                                                                                                                                                                                                                                                                                            REQUIRED NUMBER OF CHANNELS
 
Transfer/Control Parameter  (continued)
: 17.          RCIC Exhaust Line Drain Isolation                                                                                                                                                                                                              2 (1/valve)
: 18.          RCIC Steam Isolation                                                                                                                                                                                                                                                                                                                                                2 (1/valve)
: 19.          RCIC Suction from Condensate Storage Tank                                                                                                                              1
: 20.          RCIC Pump Discharge                                                                                                                                                                                                                                                                                                                                                          2 (1/valve)
: 21.          RCIC Minimum Flow                                                                                                                                                                                                                                                                                                                                                                              1
: 22.          RCIC Pump Discharge to Full Flow Test Line                                                                                                                    1
: 23.          RCIC Suction from Torus                                                                                                                                                                                                                                                                                                                  2 (1/valve)
: 24.          RCIC Steam Supply                                                                                                                                                                                                                                                                                                                                                                              1
: 25.          RCIC Lube Oil Cooler Valve                                                                                                                                                                                                                                                                                    1
: 26.          RCIC Trip Throttle Valve Operator Position                                                                                                                    1
: 27.          RCIC Trip Throttle Valve Position                                                                                                                                                                                                              1
: 28.          RCIC Vacuum Breaker                                                                                                                                                                                                                                                                                                                                                          1
: 29.          RCIC Condensate Pump                                                                                                                                                                                                                                                                                                                                                1
: 30.          RCIC Vacuum Pump                                                                                                                                                                                                                                                                                                                                                                                        1
: 31.          Safety/Relief Valves (S/RVs)                                                                                                                                                                                                                                                                3 (1/valve)
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                            B 3.3-81                                                                                                                                                                                    Revision No. 132 Remote Shutdown System B 3.3.3.2
 
Table B 3.3.3.2-1 (page 3 of 3)
Remote Shutdown System Instrumentation
 
FUNCTION                                                                                                                                                                                                                                                                                                                            REQUIRED NUMBER OF CHANNELS
 
Transfer/Control Parameter  (continued)
: 32.          Auto Isolation Reset                                                                                                                                                                                                                                                                                                                                                2 (1/division)
: 33. Instrument Transfer                                                                                                                                                                        5 (1/transfer switch)
 
PBAPS UNIT 2                                                                                                                                                                                                                            B 3.3-82                                                                                                                                                                                    Revision No. 132 ATWS-RPT Instrumentation B 3.3.4.1
 
B 3.3  INSTRUMENTATION
 
B 3.3.4.1  Anticipated Transient Without Scram Recirculation Pump Trip (ATWS-RPT) Instrumentation
 
BASES
 
BACKGROUND                                                                      The ATWS-RPT System initiates an RPT, adding negative reactivity, following events in which a scram does not (but should) occur, to lessen the effects of an ATWS event.
Tripping the recirculation pumps adds negative reactivity from the increase in steam voiding in the core area as core flow decreases. When Reactor Vessel Water Level      Low Low (Level 2) or Reactor Pressure      High setpoint is reached, the recirculation pump motor breakers trip.
 
The ATWS-RPT System includes sensors, relays, and switches that are necessary to cause initiation of an RPT. The channels include electronic equipment that compares measured input signals with pre-established setpoints. When the setpoint is exceeded, the channel output relay actuates, which then outputs an ATWS-RPT signal to the trip logic.
 
The ATWS-RPT consists of two trip systems. There are two ATWS-RPT Functions:  Reactor Pressure      High and Reactor Vessel Water Level      Low Low (Level 2). Each trip system has two channels of Reactor Pressure      High and two channels of Reactor Vessel Water Level      Low Low (Level 2). Each ATWS-RPT trip system is a one-out-of-two logic for each Function. Thus, one Reactor Water Level      Low Low (Level 2) or one Reactor Pressure      High signal is needed to trip a trip system. Both trip systems must be in a tripped condition to initiate the trip of both recirculation pumps (by tripping the respective recirculation pump motor breakers). There is one recirculation pump drive motor breaker provided for each of the two recirculation pumps for a total of two breakers.
 
APPLICABLE                                                                      The ATWS-RPT is not assumed in the safety analysis. The SAFETY ANALYSES,          ATWS-RPT initiates an RPT to aid in preserving the integrity LCO, and                                                                                          of the fuel cladding following events in which a scram does APPLICABILITY                                        not, but should, occur. Based on its contribution to the reduction of overall plant risk, however, the instrumentation meets Criterion 4 of the NRC Policy Statement.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                            B 3.3-83                                                                                                                                                                                    Revision No. 167 ATWS-RPT Instrumentation B 3.3.4.1
 
BASES
 
APPLICABLE                                                                      The OPERABILITY of the ATWS-RPT is dependent on the SAFETY ANALYSES,          OPERABILITY of the individual instrumentation channel LCO, and                                                                                          Functions. Each Function must have a required number of APPLICABILITY                                        OPERABLE channels in each trip system, with their (continued)                                        setpoints within the specified Allowable Value of SR 3.3.4.1.3. The actual setpoint is calibrated consistent with applicable setpoint methodology assumptions. Channel OPERABILITY also includes the associated recirculation pump drive motor breakers. A channel is inoperable if its actual trip setting is not within its required Allowable Value.
 
Allowable Values are specified for each ATWS-RPT Function specified in the LCO. Trip setpoints are specified in the setpoint calculations. The trip setpoints are selected to ensure that the setpoints do not exceed the Allowable Value between CHANNEL CALIBRATIONS. Operation with a trip setting less conservative than the trip setpoint, but within its Allowable Value, is acceptable. Trip setpoints are those predetermined values of output at which an action should take place. The setpoints are compared to the actual process parameter (e.g., reactor vessel water level), and when the measured output value of the process parameter exceeds the setpoint, the associated device changes state.
The analytic or design limits are derived from the limiting values of the process parameters obtained from the safety analysis. The Allowable Values are derived from the analytic or design limits, corrected for calibration, process, and instrument errors as well as instrument drift.
In selected cases, the Allowable Values and trip setpoints are determined by engineering judgement or historically accepted practice relative to the intended function of the channel. The trip setpoints determined in this manner provide adequate protection by assuring instrument and process uncertainties expected for the environments during the operating time of the associated channels are accounted for.
 
The individual Functions are required to be OPERABLE in MODE 1 to protect against common mode failures of the Reactor Protection System by providing a diverse trip to mitigate the consequences of a postulated ATWS event. The Reactor Pressure      High and Reactor Vessel Water Level      Low Low (Level 2) Functions are required to be OPERABLE in MODE 1 since the reactor is producing significant power and
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                            B 3.3-84                                                                                                                                                                                                        Revision No. 0 ATWS-RPT Instrumentation B 3.3.4.1
 
BASES
 
APPLICABLE                                                                      the recirculation system could be at high flow. During this SAFETY ANALYSES,          MODE, the potential exists for pressure increases or low LCO, and                                                                                          water level, assuming an ATWS event. In MODE 2, the reactor APPLICABILITY                                        is at low power and the recirculation system is at low flow; (continued)                                        thus, the potential is low for a pressure increase or low water level, assuming an ATWS event. Therefore, the ATWS-RPT is not necessary. In MODES 3 and 4, the reactor is shut down with all control rods inserted; thus, an ATWS event is not significant and the possibility of a significant pressure increase or low water level is negligible. In MODE 5, the one rod out interlock ensures that the reactor remains subcritical; thus, an ATWS event is not significant. In addition, the reactor pressure vessel (RPV) head is not fully tensioned and no pressure transient threat to the reactor coolant pressure boundary (RCPB) exists.
 
The specific Applicable Safety Analyses and LCO discussions are listed below on a Function by Function basis.
: a.                    Reactor Vessel Water LevelLow Low (Level 2)
 
Low RPV water level indicates that a reactor scram should have occurred and the capability to cool the fuel may be threatened. Should RPV water level decrease too far, fuel damage could result. The ATWS-RPT System is initiated at Level 2 to assist in the mitigation of the ATWS event. The resultant reduction of core flow reduces the neutron flux and THERMAL POWER and, therefore, the rate of coolant boiloff.
 
Reactor vessel water level signals are initiated from four level transmitters that sense the difference between the pressure due to a constant column of water (reference leg) and the pressure due to the actual water level (variable leg) in the vessel.
 
Four channels of Reactor Vessel Water Level      Low Low (Level 2), with two channels in each trip system, are available and required to be OPERABLE to ensure that no single instrument failure can preclude an ATWS-RPT from this Function on a valid signal. The Reactor Vessel Water Level      Low Low (Level 2) Allowable Value
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                            B 3.3-85                                                                                                                                                                                                        Revision No. 0 ATWS-RPT Instrumentation B 3.3.4.1
 
BASES
 
APPLICABLE                                                                      a.                    Reactor Vessel Water LevelLow Low (Level 2)
SAFETY ANALYSES,                                                    (continued)
LCO, and APPLICABILITY                                                                                  is chosen so that the system will not be initiated after a Level 3 scram with feedwater still available, and for convenience with the reactor core isolation cooling initiation.
: b. Reactor PressureHigh
 
Excessively high RPV pressure may rupture the RCPB.
An increase in the RPV pressure during reactor operation compresses the steam voids and results in a positive reactivity insertion. This increases neutron flux and THERMAL POWER, which could potentially result in fuel failure and overpressurization. The Reactor Pressure      High Function initiates an RPT for transients that result in a pressure increase, counteracting the pressure increase by rapidly reducing core power generation. For the overpressurization event, the RPT aids in the termination of the ATWS event and, along with the safety/relief valves, limits the peak RPV pressure to less than the ASME Section III Code limits.
 
The Reactor Pressure      High signals are initiated from four pressure transmitters that monitor reactor steam dome pressure. Four channels of Reactor Pressure High, with two channels in each trip system, are available and are required to be OPERABLE to ensure that no single instrument failure can preclude an ATWS-RPT from this Function on a valid signal. The Reactor Pressure      High Allowable Value is chosen to provide an adequate margin to the ASME Section III Code limits.
 
ACTIONS                                                                                                    A Note has been provided to modify the ACTIONS related to ATWS-RPT instrumentation channels. 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
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                            B 3.3-86                                                                                                                                                                                                        Revision No. 0 ATWS-RPT Instrumentation B 3.3.4.1
 
BASES
 
ACTIONS                                                                                                    additional failure, with Completion Times based on initial (continued)                                        entry into the Condition. However, the Required Actions for inoperable ATWS-RPT instrumentation channels provide appropriate compensatory measures for separate inoperable channels. As such, a Note has been provided that allows separate Condition entry for each inoperable ATWS-RPT instrumentation channel.
 
A.1 and A.2
 
With one or more channels inoperable, but with ATWS-RPT trip capability for each Function maintained (refer to Required Actions B.1 and C.1 Bases), the ATWS-RPT System is capable of performing the intended function. However, the reliability and redundancy of the ATWS-RPT instrumentation is reduced, such that a single failure in the remaining trip system could result in the inability of the ATWS-RPT System to perform the intended function. Therefore, only a limited time is allowed to restore the inoperable channels to OPERABLE status. Because of the diversity of sensors available to provide trip signals, the low probability of extensive numbers of inoperabilities affecting all diverse Functions, and the low probability of an event requiring the initiation of ATWS-RPT, 14 days is provided to restore the inoperable channel (Required Action A.1). Alternately, the inoperable channel may be placed in trip (Required Action A.2), since this would conservatively compensate for the inoperability, restore capability to accommodate a single failure, and allow operation to continue. As noted, placing the channel in trip with no further restrictions is not allowed if the inoperable channel is the result of an inoperable breaker, since this may not adequately compensate for the inoperable breaker (e.g., the breaker may be inoperable such that it will not open). Alternatively, a Completion Time can be determined in accordance with the Risk Informed Completion Time (RICT) Program. A Note has been provided to indicate that a RICT is only applicable when a loss of function has not occurred. If it is not desired to place the channel in trip (e.g., as in the case where placing the inoperable channel would result in an RPT), or if the inoperable channel is the result of an inoperable breaker, Condition D must be entered and its Required Actions taken.
 
B.1
 
Required Action B.1 is intended to ensure that appropriate actions are taken if multiple, inoperable, untripped channels within the same Function result in the Function not
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                            B 3.3-87                                                                                                                                                                                    Revision No. 159 ATWS-RPT Instrumentation B 3.3.4.1
 
BASES
 
ACTIONS                                                                                                    B.1  (continued)
 
maintaining ATWS-RPT trip capability. A Function is considered to be maintaining ATWS-RPT trip capability when sufficient channels are OPERABLE or in trip such that the ATWS-RPT System will generate a trip signal from the given Function on a valid signal, and both recirculation pumps can be tripped. This requires one channel of the Function in each trip system to be OPERABLE or in trip, and the recirculation pump drive motor breakers to be OPERABLE or in trip.
 
The 72 hour Completion Time is sufficient for the operator to take corrective action (e.g., restoration or tripping of channels) and takes into account the likelihood of an event requiring actuation of the ATWS-RPT instrumentation during this period and that one Function is still maintaining ATWS-RPT trip capability.
 
C.1
 
Required Action C.1 is intended to ensure that appropriate Actions are taken if multiple, inoperable, untripped channels within both Functions result in both Functions not maintaining ATWS-RPT trip capability. The description of a Function maintaining ATWS-RPT trip capability is discussed in the Bases for Required Action B.1 above.
 
The 1 hour Completion Time is sufficient for the operator to take corrective action and takes into account the likelihood of an event requiring actuation of the ATWS-RPT instrumentation during this period.
 
D.1 and D.2
 
With any Required Action and associated Completion Time not met, the plant must be brought to a MODE or other specified condition in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 2 within 6 hours (Required Action D.2). Alternately, the associated recirculation pump may be removed from service since this performs the intended function of the instrumentation (Required Action D.1). The allowed Completion Time of
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                            B 3.3-88                                                                                                                                                                                                        Revision No. 0 ATWS-RPT Instrumentation B 3.3.4.1
 
BASES
 
ACTIONS                                                                                                    D.1 and D.2  (continued)
 
6 hours is reasonable, based on operating experience, both to reach MODE 2 from full power conditions and to remove a recirculation pump from service in an orderly manner and without challenging plant systems.
 
Required Action D.1 is modified by a Note which states that the Required Action is only applicable if the inoperable channel is the result of an inoperable RPT breaker. The Note clarifies the situations under which the associated Required Action would be the appropriate Required Action.
 
SURVEILLANCE                                                  The Surveillances are modified by a Note to indicate that REQUIREMENTS                                                  when a channel is placed in an inoperable status solely for performance of required Surveillances, entry into the associated Conditions and Required Actions may be delayed for up to 6 hours provided the associated Function maintains ATWS-RPT trip capability. Upon completion of the Surveillance, or expiration of the 6 hour allowance, the channel must be returned to OPERABLE status or the applicable Condition entered and Required Actions taken.
This Note is based on the reliability analysis (Ref. 1) assumption of the average time required to perform channel Surveillance. That analysis demonstrated that the 6 hour testing allowance does not significantly reduce the probability that the recirculation pumps will trip when necessary.
 
SR  3.3.4.1.1
 
Performance of the CHANNEL CHECK ensures that a gross failure of instrumentation has not occurred. A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a similar parameter on other channels. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value.
Significant deviations between the instrument channels could be an indication of excessive instrument drift in one of the channels or something even more serious. A CHANNEL CHECK will detect gross channel failure; thus, it is key to verifying the instrumentation continues to operate properly between each CHANNEL CALIBRATION.
 
Agreement criteria are determined by the plant staff based on a combination of the channel instrument uncertainties, including indication and readability. If a channel is outside the criteria, it may be an indication that the instrument has drifted outside its limit.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                            B 3.3-89                                                                                                                                                                                              Revision No. 86 ATWS-RPT Instrumentation B 3.3.4.1
 
BASES
 
SURVEILLANCE                                                  SR  3.3.4.1.1  (continued)
REQUIREMENTS The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. The CHANNEL CHECK supplements less formal, but more frequent, checks of channels during normal operational use of the displays associated with the required channels of this LCO.
 
SR  3.3.4.1.2
 
A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the entire channel will perform the intended function. Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology.
 
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
 
SR  3.3.4.1.3
 
A CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor. This test verifies the channel responds to the measured parameter within the necessary range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drifts between successive calibrations, consistent with the assumptions of the current plant specific setpoint methodology.
 
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
 
SR  3.3.4.1.4
 
The LOGIC SYSTEM FUNCTIONAL TEST demonstrates the OPERABILITY of the required trip logic for a specific channel. The system functional test of the pump breakers is included as part of this Surveillance and overlaps the LOGIC SYSTEM FUNCTIONAL TEST to provide complete testing of the assumed safety function. Therefore, if a breaker is incapable of operating, the associated instrument channel(s) would be inoperable.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                            B 3.3-90                                                                                                                                                                                              Revision No. 86 ATWS-RPT Instrumentation B 3.3.4.1
 
BASES
 
SURVEILLANCE                                                  SR  3.3.4.1.4  (continued)
REQUIREMENTS The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
 
REFERENCES                                                                      1.                    GENE-770-06-1, "Bases for Changes To Surveillance Test Intervals and Allowed Out-of-Service Times For Selected Instrumentation Technical Specifications,"
February 1991.
 
PBAPS UNIT 2                                                                                                                                                                                                                            B 3.3-91                                                                                                                                                                                              Revision No. 86 EOC-RPT Instrumentation B 3.3.4.2
 
B 3.3  INSTRUMENTATION
 
B 3.3.4.2  End of Cycle Recirculation Pump Trip (EOC-RPT) Instrumentation
 
BASES
 
BACKGROUND                                                                      The EOC-RPT instrumentation initiates a recirculation pump trip (RPT) to reduce the peak reactor pressure and power resulting from turbine trip or generator load rejection transients and to minimize the decrease in core MCPR during these transients.
 
The benefit of the additional negative reactivity in excess of that normally inserted on a scram reflects end of cycle reactivity considerations. Flux shapes at the end of cycle are such that the control rods insert only a small amount of negative reactivity during the first few feet of rod travel upon a scram caused by Turbine Control Valve (TCV) Fast Closure, Trip Oil Pressure      Low or Turbine Stop Valve (TSV)      Closure. The physical phenomenon involved is that the void reactivity feedback due to a pressurization transient can add positive reactivity at a faster rate than the control rods can add negative reactivity.
 
The EOC-RPT instrumentation, as shown in Reference 1, is composed of sensors that detect initiation of closure of the TSVs or fast closure of the TCVs, combined with relays, logic circuits, and fast acting circuit breakers that interrupt power from the recirculation pump adjustable speed drives (ASDs) to each of the recirculation pump motors. When the setpoint is exceeded, the channel output relay actuates, which then outputs an EOC-RPT signal to the trip logic. When the RPT breakers trip open, the recirculation pumps coast down under their own inertia. The EOC-RPT has two identical trip systems, either of which can actuate an RPT.
 
Each EOC-RPT trip system is a two-out-of-two logic for each Function; thus, either two TSV      Closure or two TCV Fast Closure, Trip Oil Pressure      Low signals are required for a trip system to actuate. If either trip system actuates, both recirculation pumps will trip. There are two EOC-RPT breakers in series per recirculation pump. One trip system trips one of the two EOC-RPT breakers for each recirculation
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                        B 3.3-91a                                                                                                                                                                                Revision No. 137 EOC-RPT Instrumentation B 3.3.4.2
 
BASES
 
BACKGROUND                                                                      pump, and the second trip system trips the other EOC-RPT (continued)                                        breaker for each recirculation pump.
 
APPLICABLE                                                                      The TSV      Closure and the TCV Fast Closure, Trip Oil SAFETY ANALYSES,          Pressure      Low Functions are designed to trip the LCO, and                                                                                  recirculation pumps in the event of a turbine trip or APPLICABILITY                                        generator load rejection to mitigate the neutron flux, heat flux, and pressurization transients, and to minimize the decrease in MCPR. The analytical methods and assumptions used in evaluating the turbine trip and generator load rejection, as well as other safety analyses that utilize EOC-RPT, are summarized in References 2, 3, and 4.
 
To mitigate pressurization transient effects, the EOC-RPT must trip the recirculation pumps after initiation of closure movement of either the TSVs or the TCVs. The combined effects of this trip and a scram reduce fuel bundle power more rapidly than a scram alone so that the Safety Limit MCPR is not exceeded. Alternatively, APLHGR operating limits (LCO 3.2.1, "AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)"), the MCPR operating limits (LCO 3.2.2, "MINIMUM CRITICAL POWER RATIO (MCPR)"), and the LHGR operating limits (LCO 3.2.3, LINEAR HEAT GENERATION RATE (LHGR)) for an inoperable EOC-RPT, as specified in the COLR, are sufficient to allow this LCO to be met. The EOC-RPT function is automatically disabled when turbine first stage pressure is < 26.3% RTP.
 
EOC-RPT instrumentation satisfies Criterion 3 of the NRC Policy Statement.
 
The OPERABILITY of the EOC-RPT is dependent on the OPERABILITY of the individual instrumentation channel Functions, i.e., the TSV-Closure and the TCV Fast Closure, Trip Oil Pressure-Low Functions. Each Function must have a required number of OPERABLE channels in each trip system, with their setpoints within the specified Allowable Value of SR 3.3.4.2.3. Channel OPERABILITY also includes the associated EOC-RPT breakers. Each channel (including the associated EOC-RPT breakers) must also respond within its assumed response time.
 
Allowable Values are specified for each EOC-RPT Function specified in the LCO. Trip setpoints are specified in the plant design documentation. The trip setpoints are selected
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                        B 3.3-91b                                                                                                                                                                                Revision No. 143 EOC-RPT Instrumentation B 3.3.4.2
 
BASES
 
APPLICABLE                                                                      to ensure that the actual setpoints do not exceed the SAFETY ANALYSES,          Allowable Value between successive CHANNEL CALIBRATIONS.
LCO, and                                                                                          Operation with a trip setpoint less conservative than the APPLICABILITY                                        trip setpoint, but within its Allowable Value, is (continued)                                        acceptable. A channel is inoperable if its actual trip setting is not within its required Allowable Value. Trip setpoints are those predetermined values of output at which an action should take place. The setpoints are compared to the actual process parameters (e.g. TSV position), and when the measured output value of the process parameter exceeds the setpoint, the associated device (e.g., limit switch) changes state. The analytic limit for the TCV Fast Closure, Trip Oil Pressure-Low Function was determined based on the TCV hydraulic oil circuit design. The Allowable Value is derived from the analytic limit, corrected for calibration, process, and instrument errors. The trip setpoint is determined from the analytical limit corrected for calibration, process, and instrumentation errors, as well as instrument drift, as applicable. The Allowable Value and trip setpoint for the TSV-Closure Function was determined by engineering judgment and historically accepted practice for similar trip functions.
 
The specific Applicable Safety Analysis, LCO, and Applicability discussions are listed below on a Function by Function basis.
 
Alternatively, since the instrumentation protects against a MCPR SL violation, with the instrumentation inoperable, modifications to the APLHGR operating limits (LCO 3.2.1, "AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)"), the MCPR operating limits (LCO 3.2.2, "MINIMUM CRITICAL POWER RATIO (MCPR)"), and the LHGR operating limits (LCO 3.2.3, LINEAR HEAT GENERATION RATE (LHGR)) may be applied to allow this LCO to be met. The appropriate MCPR operating limits and power-dependent thermal limit adjustments for the EOC-RPT inoperable condition are specified in the COLR.
 
Turbine Stop ValveClosure
 
Closure of the TSVs and a main turbine trip result in the loss of a heat sink that produces reactor pressure, neutron flux, and heat flux transients that must be limited.
Therefore, an RPT is initiated on TSV      Closure in anticipation of the transients that would result from closure of these valves. EOC-RPT decreases peak reactor power and aids the reactor scram in ensuring that the MCPR SL is not exceeded during the worst case transient.
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                        B 3.3-91c                                                                                                                                                                                          Revision No. 49 EOC-RPT Instrumentation B 3.3.4.2
 
BASES
 
APPLICABLE                                                                      Turbine Stop ValveClosure  (continued)
SAFETY ANALYSIS, LCO, and                                                                                          Closure of the TSVs is determined by measuring the position APPLICABILITY                                        of each valve. There are position switches associated with each stop valve, the signal from each switch being assigned to a separate trip channel. The logic for the TSV      Closure Function is such that two or more TSVs must be closed to produce an EOC-RPT. This Function must be enabled at THERMAL POWER                                                                                                                                              26.3% RTP as measured at the turbine first stage pressure. This is normally accomplished automatically by pressure switches sensing turbine first stage pressure; therefore, opening of the turbine bypass valves may affect this Function. Four channels of TSV      Closure, with two channels in each trip system, are available and required to be OPERABLE to ensure that no single instrument failure will preclude an EOC-RPT from this Function on a valid signal.
The TSV      Closure Allowable Value is selected to detect imminent TSV closure.
 
This EOC-RPT Function is required, consistent with the safety analysis assumptions, whenever THERMAL POWER is 26.3% RTP. Below 26.3% RTP, the Reactor Pressure      High and the Average Power Range Monitor (APRM) Scram Clamp Functions of the Reactor Protection System (RPS) are adequate to maintain the necessary safety margins.
 
Turbine Control Valve Fast Closure, Trip Oil Pressure - Low
 
Fast closure of the TCVs during a generator load rejection results in the loss of a heat sink that produces reactor pressure, neutron flux, and heat flux transients that must be limited. Therefore, an RPT is initiated on TCV Fast Closure, Trip Oil Pressure      Low in anticipation of the transients that would result from the closure of these valves. The EOC-RPT decreases peak reactor power and aids the reactor scram in ensuring that the MCPR SL is not exceeded during the worst case transient.
 
Fast closure of the TCVs is determined by measuring the electrohydraulic control fluid pressure at each control valve. There is one pressure switch associated with each control valve, and the signal from each switch is assigned to a separate trip channel. The logic for the TCV Fast Closure, Trip Oil Pressure      Low Function is such that two or more TCVs must be closed (pressure switch trips)
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                        B 3.3-91d                                                                                                                                                                                Revision No. 143 EOC-RPT Instrumentation B 3.3.4.2
 
BASES
 
APPLICABLE                                                                      Turbine Control Valve Fast Closure, Trip Oil PressureLow SAFETY ANALYSIS,          (continued)
LCO, and APPLICABILITY                                        to produce an EOC-RPT. This Function must be enabled at THERMAL POWER                                                                                                                                              26.3% RTP as measured at the turbine first stage pressure. This is normally accomplished automatically by pressure switches sensing turbine first stage pressure; therefore, opening of the turbine bypass valves may affect this Function. Four channels of TCV Fast Closure, Trip Oil Pressure      Low, with two channels in each trip system, are available and required to be OPERABLE to ensure that no single instrument failure will preclude an EOC-RPT from this Function on a valid signal. The TCV Fast Closure, Trip Oil Pressure      Low Allowable Value is selected high enough to detect imminent TCV fast closure.
 
This protection is required consistent with the safety analysis whenever THERMAL POWER is                                                                                                                                                                                                                                                                                                                                                                26.3% RTP. Below 26.3% RTP, the Reactor Pressure      High and the APRM Scram Clamp Functions of the RPS are adequate to maintain the necessary safety margins.
 
ACTIONS                                                                                                    A Note has been provided to modify the ACTIONS related to EOC-RPT instrumentation channels. 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 inoperable EOC-RPT instrumentation channels provide appropriate compensatory measures for separate inoperable channels. As such, a Note has been provided that allows separate Condition entry for each inoperable EOC-RPT instrumentation channel.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                        B 3.3-91e                                                                                                                                                                                Revision No. 143 EOC-RPT Instrumentation B 3.3.4.2
 
BASES
 
ACTIONS                                                                                                    A.1 and A.2 (continued)
With one or more required channels inoperable, but with EOC-RPT trip capability maintained (refer to Required Action B.1 Bases), the EOC-RPT System is capable of performing the intended function. However, the reliability and redundancy of the EOC-RPT instrumentation is reduced such that a single failure in the remaining trip system could result in the inability of the EOC-RPT System to perform the intended function. Therefore, only a limited time is allowed to restore compliance with the LCO. Because of the diversity of sensors available to provide trip signals, the low probability of extensive numbers of inoperabilities affecting all diverse Functions, and the low probability of an event requiring the initiation of an EOC-RPT, 72 hours is provided to restore the inoperable channels (Required Action A.1). Alternately, the inoperable channels may be placed in trip (Required Action A.2) since this would conservatively compensate for the inoperability, restore capability to accommodate a single failure, and allow operation to continue. As noted in Required Action A.2, placing the channel in trip with no further restrictions is not allowed if the inoperable channel is the result of an inoperable breaker, since this may not adequately compensate for the inoperable breaker (e.g., the breaker may be inoperable such that it will not open).
Alternatively, a Completion Time can be determined in accordance with the Risk Informed Completion  Time (RICT)
Program. A note has been provided to indicate that a RICT is only applicable when a loss of function has not occurred.
If it is not desired to place the channel in trip (e.g., as in the case where placing the inoperable channel in trip would result in an RPT, or if the inoperable channel is the result of an inoperable breaker), Condition C must be entered and its Required Actions taken.
 
B.1
 
Required Action B.1 is intended to ensure that appropriate actions are taken if multiple, inoperable, untripped channels within the same Function result in the Function not maintaining EOC-RPT trip capability. A Function is considered to be maintaining EOC-RPT trip capability when sufficient channels are OPERABLE or in trip, such that the EOC-RPT System will generate a trip signal from the given Function on a valid signal and both recirculation pumps can be tripped. This requires two channels of the Function in the same trip system, to each be OPERABLE or in trip, and the associated EOC-RPT breakers to be OPERABLE.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                        B 3.3-91f                                                                                                                                                                                Revision No. 159 EOC-RPT Instrumentation B 3.3.4.2
 
BASES
 
ACTIONS                                                                                                    B.1  (continued)
 
The 2 hour Completion Time is sufficient time for the operator to take corrective action, and takes into account the likelihood of an event requiring actuation of the EOC-RPT instrumentation during this period. It is also consistent with the 2 hour Completion Time provided in LCO 3.2.1 and 3.2.2 for Required Action A.1, since this instrumentation's purpose is to preclude a thermal limit violation.
 
C.1 and C.2
 
With any Required Action and associated Completion Time not met, THERMAL POWER must be reduced to < 26.3% RTP within 4 hours. Alternately, for an inoperable breaker (e.g., the breaker may be inoperable such that it will not open) the associated recirculation pump may be removed from service, since this performs the intended function of the instrumentation. The allowed Completion Time of 4 hours is reasonable, based on operating experience, to reduce THERMAL POWER to < 26.3% RTP from full power conditions in an orderly manner and without challenging plant systems.
 
Required Action C.1 is modified by a Note which states that the Required Action is only applicable if the inoperable channel is the result of an inoperable RPT breaker. The Note clarifies the situations under which the associated Required Action would be the appropriate Required Action.
 
SURVEILLANCE                                                  The Surveillances are modified by a Note to indicate that REQUIREMENTS                                                  when a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours provided the associated Function maintains EOC-RPT trip capability. Upon completion of the Surveillance, or expiration of the 6 hour allowance, the channel must be returned to OPERABLE status or the applicable Condition entered and Required Actions taken. This Note is based on the reliability analysis (Ref. 5) assumption of the average time required to perform channel Surveillance. That analysis demonstrated that the 6 hour testing allowance does not significantly reduce the probability that the recirculation pumps will trip when necessary.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                        B 3.3-91g                                                                                                                                                                                Revision No. 143 EOC-RPT Instrumentation B 3.3.4.2
 
BASES
 
SURVEILLANCE                                                  SR  3.3.4.2.1 REQUIREMENTS (continued)                                        A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the entire channel will perform the intended function.
 
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
 
SR  3.3.4.2.2
 
CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor. This test verifies the channel responds to the measured parameter within the necessary range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drifts between successive calibrations consistent with the plant specific setpoint methodology.
 
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
 
SR  3.3.4.2.3
 
The LOGIC SYSTEM FUNCTIONAL TEST demonstrates the OPERABILITY of the required trip logic for a specific channel. The system functional test of the pump breakers is included as a part of this test, overlapping the LOGIC SYSTEM FUNCTIONAL TEST, to provide complete testing of the associated safety function. Therefore, if a breaker is incapable of operating, the associated instrument channel(s) would also be inoperable.
 
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                          B 3.3-91h                                                                                                                                                                                                                        Revision No. 86 EOC-RPT Instrumentation B 3.3.4.2
 
BASES
 
SURVEILLANCE                                                  SR  3.3.4.2.4 REQUIREMENTS (continued)                                        This SR ensures that an EOC-RPT initiated from the TSV      Closure and TCV Fast Closure, Trip Oil Pressure      Low Functions will not be inadvertently bypassed when THERMAL POWER is  26.3% RTP. This involves calibration of the bypass channels. Adequate margins for the instrument setpoint methodologies are incorporated into the actual setpoint. Because main turbine bypass flow can affect this setpoint nonconservatively (THERMAL POWER is derived from first stage pressure) the main turbine bypass valves must remain closed during the calibration at THERMAL POWER 26.3% RTP to ensure that the calibration remains valid. If any bypass channel's setpoint is nonconservative (i.e., the Functions are bypassed at                                                                                                                                                                                                                                                                      26.3% RTP, either due to open main turbine bypass valves or other reasons), the affected TSV      Closure and TCV Fast Closure, Trip Oil Pressure      Low Functions are considered inoperable. Alternatively, the bypass channel can be placed in the conservative condition (nonbypass). If placed in the nonbypass condition, this SR is met with the channel considered OPERABLE.
 
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
 
SR  3.3.4.2.5
 
This SR ensures that the individual channel response times are less than or equal to the maximum values assumed in the accident analysis. The EOC-RPT SYSTEM RESPONSE TIME acceptance criterion is included in Reference 6.
 
A Note to the Surveillance states that breaker interruption time may be assumed from the most recent performance of SR 3.3.4.2.6. This is allowed since the time to open the contacts after energization of the trip coil and the arc suppression time are short and do not appreciably change, due to the design of the breaker opening device and the fact that the breaker is not routinely cycled.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                        B 3.3-91i                                                                                                                                                                                Revision No. 143 EOC-RPT Instrumentation B 3.3.4.2
 
BASES
 
SURVEILLANCE                                                  SR  3.3.4.2.5  (continued)
REQUIREMENTS Response times cannot be determined at power because operation of final actuated devices is required. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
 
SR  3.3.4.2.6
 
This SR ensures that the RPT breaker interruption time (arc suppression time plus time to open the contacts) is provided to the EOC-RPT SYSTEM RESPONSE TIME test. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
 
REFERENCES                                                                      1.                    UFSAR, Figure 7.9.4A, Sheet 3 of 3 (EOC-RPT logic diagram).
: 2.                    UFSAR, Section 7.9.4.4.3.
: 3.                    UFSAR, Section 14.5.1.2.4.
: 4.                    NEDE-24011-P-A, "General Electric Standard Application for Reactor Fuel," latest approved version.
: 5.                    GENE-770-06-1-A, "Bases for Changes to Surveillance Test Intervals and Allowed Out-Of-Service Times for Selected Instrumentation Technical Specifications,"
December 1992.
: 6.                    Core Operating Limits Report.
: 7.                    NEDC-33873P, Safety Analysis Report for Peach Bottom Atomic Power Sation, Units 2 and 3, Thermal Power Opitimization, Revision 0.
 
PBAPS UNIT 2                                                                                                                                                                                                                        B 3.3-91j                                                                                                                                                                                Revision No. 143
 
ECCS Instrumentation B 3.3.5.1
 
B 3.3  INSTRUMENTATION
 
B 3.3.5.1  Emergency Core Cooling System (ECCS) Instrumentation
 
BASES
 
BACKGROUND                                                                      The purpose of the ECCS instrumentation is to initiate appropriate responses from the systems to ensure that the fuel is adequately cooled in the event of a design basis accident or transient.
 
For most abnormal operational transients and Design Basis Accidents (DBAs), a wide range of dependent and independent parameters are monitored.
 
The ECCS instrumentation actuates core spray (CS), low pressure coolant injection (LPCI), high pressure coolant injection (HPCI), Automatic Depressurization System (ADS),
and the diesel generators (DGs). The equipment involved with each of these systems is described in the Bases for LCO 3.5.1, "ECCS      Operating."
 
Core Spray System
 
The CS System may be initiated by automatic means.
Automatic initiation occurs for conditions of Reactor Vessel Water Level      Low Low Low (Level 1) or Drywell Pressure      High with a Reactor Pressure      Low permissive. The reactor vessel water level and the reactor pressure variables are monitored by four redundant transmitters, which are, in turn, connected to four pressure compensation instruments. The drywell pressure variable is monitored by four redundant transmitters, which are, in turn, connected to four trip units. The outputs of the pressure compensation instruments and the trip units are connected to relays which send signals to two trip systems, with each trip system arranged in a one-out-of-two taken twice logic (each trip unit sends a signal to both trip systems.)  Each trip system initiates two of the four CS pumps.
 
Upon receipt of an initiation signal, if normal AC power is available, CS pumps A and C start after a time delay of approximately 13 seconds and CS pumps B and D start after a time delay of approximately 23 seconds. If normal AC power is not available, the four CS pumps start simultaneously after a time delay of approximately 6 seconds after the respective DG is ready to load.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                            B 3.3-92                                                                                                                                                                                                        Revision No. 0 ECCS Instrumentation B 3.3.5.1
 
BASES
 
BACKGROUND                                                                      Core Spray System  (continued)
 
The CS test line isolation valve, which is also a primary containment isolation valve (PCIV), is closed on a CS initiation signal to allow full system flow assumed in the accident analyses and maintain primary containment isolated in the event CS is not operating.
 
The CS pump discharge flow is monitored by a differential pressure indicating switch. When the pump is running and discharge flow is low enough so that pump overheating may occur, the minimum flow return line valve is opened. The valve is automatically closed if flow is above the minimum flow setpoint to allow the full system flow assumed in the accident analysis.
 
The CS System also monitors the pressure in the reactor to ensure that, before the injection valves open, the reactor pressure has fallen to a value below the CS System's maximum design pressure. The variable is monitored by four redundant transmitters, which are, in turn, connected to four pressure compensation instruments. The outputs of the pressure compensation instruments are connected to relays whose contacts are arranged in a one-out-of-two taken twice logic.
 
Low Pressure Coolant Injection System
 
The LPCI is an operating mode of the Residual Heat Removal (RHR) System, with two LPCI subsystems. The LPCI subsystems may be initiated by automatic means. Automatic initiation occurs for conditions of Reactor Vessel Water Level      Low Low Low (Level 1); Drywell Pressure      High with a Reactor Pressure      Low (Injection Permissive). The drywell pressure variable is monitored by four redundant transmitters, which, in turn, are connected to four trip units. The reactor vessel water level and the reactor pressure variables are monitored by four redundant transmitters, which are, in turn, connected to four pressure compensation instruments.
The outputs of the trip units and pressure compensation instruments are connected to relays which send signals to two trip systems, with each trip system arranged in a one-out-of-two taken twice logic (each trip unit sends a signal to both trip systems). Each trip system can initiate all four LPCI pumps.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                            B 3.3-93                                                                                                                                                                                                        Revision No. 0 ECCS Instrumentation B 3.3.5.1
 
BASES
 
BACKGROUND                                                                      Low Pressure Coolant Injection System  (continued)
 
Upon receipt of an initiation signal if normal AC power is available, the LPCI A and B pumps start after a delay of approximately 2 seconds. The LPCI C and D pumps are started after a delay of approximately 8 seconds. If normal AC power is not available, the four LPCI pumps start simultaneously with no delay as soon as the standby power source is available.
 
Each LPCI subsystem's discharge flow is monitored by a differential pressure indicating switch. When a pump is running and discharge flow is low enough so that pump overheating may occur, the respective minimum flow return line valve is opened. If flow is above the minimum flow setpoint, the valve is automatically closed to allow the full system flow assumed in the analyses.
 
The RHR test line suppression pool cooling isolation valve, suppression pool spray isolation valves, and containment spray isolation valves (which are also PCIVs) are also closed on a LPCI initiation signal to allow the full system flow assumed in the accident analyses and maintain primary containment isolated in the event LPCI is not operating.
 
The LPCI System monitors the pressure in the reactor to ensure that, before an injection valve opens, the reactor pressure has fallen to a value below the LPCI System's maximum design pressure. The variable is monitored by four redundant transmitters, which are, in turn, connected to four pressure compensation instruments. The outputs of the pressure compensation instruments are connected to relays whose contacts are arranged in a one-out-of-two taken twice logic. Additionally, instruments are provided to close the recirculation pump discharge valves to ensure that LPCI flow does not bypass the core when it injects into the recirculation lines. The variable is monitored by four redundant transmitters, which are, in turn, connected to four pressure compensation instruments. The outputs of the pressure compensation instruments are connected to relays whose contacts are arranged in a one-out-of-two taken twice logic.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                            B 3.3-94                                                                                                                                                                                                        Revision No. 0 ECCS Instrumentation B 3.3.5.1
 
BASES
 
BACKGROUND                                                                      Low Pressure Coolant Injection System  (continued)
 
Low reactor water level in the shroud is detected by two additional instruments. When the level is greater than the low level setpoint LPCI may no longer be required, therefore other modes of RHR (e.g., suppression pool cooling) are allowed. Manual overrides for the isolations below the low level setpoint are provided.
 
High Pressure Coolant Injection System
 
The HPCI System may be initiated by automatic means.
Automatic initiation occurs for conditions of Reactor Vessel Water Level      Low Low (Level 2) or Drywell Pressure      High.
The reactor vessel water level variable is monitored by four redundant transmitters, which are, in turn, connected to four pressure compensation instruments. The drywell pressure variable is monitored by four redundant transmitters, which are, in turn, connected to four trip units. The outputs of the pressure compensation instruments and the trip units are connected to relays whose contacts are arranged in a one-out-of-two taken twice logic for each Function.
 
The HPCI pump discharge flow is monitored by a flow switch.
When the pump is running and discharge flow is low enough so that pump overheating may occur, the minimum flow return line valve is opened. The valve is automatically closed if flow is above the minimum flow setpoint to allow the full system flow assumed in the safety analysis.
 
The HPCI test line isolation valve (which is also a PCIV) is closed upon receipt of a HPCI initiation signal to allow the full system flow assumed in the accident analysis and maintain primary containment isolated in the event HPCI is not operating.
 
The HPCI System also monitors the water levels in the condensate storage tank (CST) and the suppression pool because these are the two sources of water for HPCI operation. Reactor grade water in the CST is the normal source. Upon receipt of a HPCI initiation signal, the CST
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                            B 3.3-95                                                                                                                                                                                                        Revision No. 0 ECCS Instrumentation B 3.3.5.1
 
BASES
 
BACKGROUND                                                                      High Pressure Coolant Injection System  (continued)
 
suction valve is automatically signaled to open (it is normally in the open position) unless both suppression pool suction valves are open. If the water level in the CST falls below a preselected level, first the suppression pool suction valves automatically open, and then the CST suction valve automatically closes. Two level switches are used to detect low water level in the CST. Either switch can cause the suppression pool suction valves to open and the CST suction valve to close. The suppression pool suction valves also automatically open and the CST suction valve closes if high water level is detected in the suppression pool. To prevent losing suction to the pump, the suction valves are interlocked so that one suction path must be open before the other automatically closes.
 
The HPCI provides makeup water to the reactor until the reactor vessel water level reaches the Reactor Vessel Water Level      High (Level 8) trip, at which time the HPCI turbine trips, which causes the turbine's stop valve and the control valves to close. The logic is two-out-of-two to provide high reliability of the HPCI System. The HPCI System automatically restarts if a Reactor Vessel Water Level      Low Low (Level 2) signal is subsequently received.
 
Automatic Depressurization System
 
The ADS may be initiated by automatic means. Automatic initiation occurs when signals indicating Reactor Vessel Water Level      Low Low Low (Level 1); Drywell Pressure      High or ADS Bypass Low Water Level Actuation Timer; Reactor Vessel Water Confirmatory Level      Low (Level 4); and CS or LPCI Pump Discharge Pressure      High are all present and the ADS Initiation Timer has timed out. There are two transmitters each for Reactor Vessel Water Level      Low Low Low (Level 1) and Drywell Pressure      High, and one transmitter for Reactor Vessel Water Confirmatory Level      Low (Level 4) in each of the two ADS trip systems. Each of these transmitters connects to a trip unit, which then drives a relay whose contacts form the initiation logic.
 
Each ADS trip system includes a time delay between satisfying the initiation logic and the actuation of the ADS valves. The ADS Initiation Timer time delay setpoint chosen is long enough that the HPCI has sufficient operating time
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                            B 3.3-96                                                                                                                                                                                                        Revision No. 0 ECCS Instrumentation B 3.3.5.1
 
BASES
 
BACKGROUND                                                                      Automatic Depressurization System  (continued)
 
to recover to a level above Level 1, yet not so long that the LPCI and CS Systems are unable to adequately cool the fuel if the HPCI fails to maintain that level. An alarm in the control room is annunciated when either of the timers is timing. Resetting the ADS initiation signals resets the ADS Initiation Timers.
 
The ADS also monitors the discharge pressures of the four LPCI pumps and the four CS pumps. Each ADS trip system includes two discharge pressure permissive switches from all four LPCI pumps and one discharge pressure permissive switch from all four CS pumps. The signals are used as a permissive for ADS actuation, indicating that there is a source of core coolant available once the ADS has depressurized the vessel. Two CS pumps in proper combination (C or D and A or B) or any one of the four LPCI pumps is sufficient to permit automatic depressurization.
 
The ADS logic in each trip system is arranged in two strings. Each string has a contact from each of the following variables:  Reactor Vessel Water Level      Low Low Low (Level 1); Drywell Pressure      High; Low Water Level Actuation Timer; and Reactor Vessel Water Level      Low Low Low (Level 1) Permissive. One of the two strings in each trip system must also have a Reactor Vessel Water Confirmatory Level      Low (Level 4). After the contacts for the initiation signal from either drywell pressure or reactor vessel level (and the timer for reactor vessel level timing out) close, the following must be present to initiate an ADS trip system:  all other contacts in both logic strings must close, the ADS initiation timer must time out, and a CS or LPCI pump discharge pressure signal must be present. Either the A or B trip system will cause all the ADS relief valves to open. Once the Drywell Pressure      High signal, the ADS Low Water Level Actuation Timer, or the ADS initiation signal is present, it is individually sealed in until manually reset.
 
Manual inhibit switches are provided in the control room for the ADS; however, their function is not required for ADS OPERABILITY (provided ADS is not inhibited when required to be OPERABLE).
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                            B 3.3-97                                                                                                                                                                                                        Revision No. 0 ECCS Instrumentation B 3.3.5.1
 
BASES
 
BACKGROUND                                                                      Diesel Generators (continued)
The DGs may be initiated by automatic means. Automatic initiation occurs for conditions of Reactor Vessel Water Level      Low Low Low (Level 1) or Drywell Pressure      High. The DGs are also initiated upon loss of voltage signals.  (Refer to the Bases for LCO 3.3.8.1, "Loss of Power (LOP)
Instrumentation," for a discussion of these  signals.)  The reactor vessel water level variable is  monitored by four redundant transmitters, which are, in turn, connected to four pressure compensation instruments. The drywell pressure variable is monitored by four redundant transmitters, which are, in turn, connected to four trip units. The outputs of the four pressure compensation instruments and the trip units are connected to relays which send signals to two trip systems, with each trip system arranged in a one-out-of-two taken twice logic (each trip unit sends a signal to both trip systems). The A trip system initiates all four DGs and the B trip system initiates all four DGs. The DGs receive their initiation signals from the CS System initiation logic. The DGs can also be started manually from the control room and locally from the associated DG room. Upon  receipt of a loss of coolant accident (LOCA) initiation signal, each DG is automatically started, is ready to load in approximately 10 seconds, and will run in standby conditions (rated voltage and speed, with the DG output breaker open). The DGs will only energize their respective Engineered Safety Feature buses if a loss of offsite power occurs.  (Refer to Bases for LCO 3.3.8.1.)
 
APPLICABLE                                                                      The actions of the ECCS are explicitly assumed in the safety SAFETY ANALYSES,          analyses of References 1, 2, and 3. The ECCS is initiated LCO, and                                                                                          to preserve the integrity of the fuel cladding by limiting APPLICABILITY                                        the post LOCA peak cladding temperature to less than the 10 CFR 50.46 limits.
 
ECCS instrumentation satisfies Criterion 3 of the NRC Policy Statement. Certain instrumentation Functions are retained for other reasons and are described below in the individual Functions discussion.
 
The OPERABILITY of the ECCS instrumentation is dependent upon the OPERABILITY of the individual instrumentation channel Functions specified in Table 3.3.5.1-1. Each Function must have a required number of OPERABLE channels,
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                            B 3.3-98                                                                                                                                                                                              Revision No. 21 ECCS Instrumentation B 3.3.5.1
 
BASES
 
APPLICABLE                                                                      with their setpoints within the specified Allowable Values, SAFETY ANALYSES,          where appropriate. The actual setpoint is calibrated LCO, and                                                                                  consistent with applicable setpoint methodology assumptions.
APPLICABILITY                                        Table 3.3.5.1-1 is modified by a footnote which is added to (continued)                                                            show that certain ECCS instrumentation Functions also perform DG initiation.
 
Allowable Values are specified for each ECCS Function specified in the Table. Trip setpoints are specified in the setpoint calculations. The trip setpoints are selected to ensure that the settings do not exceed the Allowable Value between CHANNEL CALIBRATIONS. Operation with a trip setting less conservative than the trip setpoint, but within its Allowable Value, is acceptable. A channel is inoperable if its actual trip setpoint is not within its required Allowable Value. Trip setpoints are those predetermined values of output at which an action should take place. The setpoints are compared to the actual process parameter (e.g., reactor vessel water level), and when the measured output value of the process parameter exceeds the setpoint, the associated device (e.g., trip unit) changes state. The analytic or design limits are derived from the limiting values of the process parameters obtained from the safety analysis or other appropriate documents. The Allowable Values are derived from the analytic or design limits, corrected for calibration, process, and instrument errors.
The trip setpoints are determined from analytical or design limits, corrected for calibration, process, and instrument errors, as well as, instrument drift. In selected cases, the Allowable Values and trip setpoints are determined from engineering judgement or historically accepted practice relative to the intended functions of the channel. The trip setpoints determined in this manner provide adequate protection by assuming instrument and process uncertainties expected for the environments during the operating time of the associated channels are accounted for. For the Core Spray and LPCI Pump Start      Time Delay Relays, adequate margins for applicable setpoint methodologies are incorporated into the Allowable Values and actual setpoints.
 
In general, the individual Functions are required to be OPERABLE in the MODES or other specified conditions that may require ECCS (or DG) initiation to mitigate the consequences of a design basis transient or accident. To ensure reliable ECCS and DG function, a combination of Functions is required to provide primary and secondary initiation signals.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                            B 3.3-99                                                                                                                                                                                    Revision No. 145 ECCS Instrumentation B 3.3.5.1
 
BASES
 
APPLICABLE                                                                      The specific Applicable Safety Analyses, LCO, and SAFETY ANALYSES,          Applicability discussions are listed below on a Function by LCO, and                                                                                          Function basis.
APPLICABILITY (continued)
Core Spray and Low Pressure Coolant Injection Systems
 
1.a, 2.a. Reactor Vessel Water LevelLow Low Low (Level 1)
 
Low reactor pressure vessel (RPV) water level indicates that the capability to cool the fuel may be threatened. Should RPV water level decrease too far, fuel damage could result.
The low pressure ECCS and associated DGs are initiated at Reactor Vessel Water Level      Low Low Low (Level 1) to ensure that core spray and flooding functions are available to prevent or minimize fuel damage. The DGs are initiated from Function 1.a signals. This Function, in conjunction with a Reactor Pressure      Low (Injection Permissive) signal, also initiates the closure of the Recirculation Discharge Valves to ensure the LPCI subsystems inject into the proper RPV location. The Reactor Vessel Water Level      Low Low Low (Level 1) is one of the Functions assumed to be OPERABLE and capable of initiating the ECCS during the transients analyzed in References 1 and 3. In addition, the Reactor Vessel Water Level      Low Low Low (Level 1) Function is directly assumed in the analysis of the recirculation line break (Ref. 4) and the control rod drop accident (CRDA) analysis. The core cooling function of the ECCS, along with the scram action of the Reactor Protection System (RPS),
ensures that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46.
 
Reactor Vessel Water Level      Low Low Low (Level 1) signals are initiated from four level transmitters that sense the difference between the pressure due to a constant column of water (reference leg) and the pressure due to the actual water level (variable leg) in the vessel.
 
The Reactor Vessel Water Level      Low Low Low (Level 1)
Allowable Value is chosen to allow time for the low pressure core flooding systems to activate and provide adequate cooling.
 
Four channels of Reactor Vessel Water Level      Low Low Low (Level 1) Function are only required to be OPERABLE when the ECCS are required to be OPERABLE to ensure that no single instrument failure can preclude ECCS initiation.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                        B 3.3-100                                                                                                                                                                                Revision No. 145 ECCS Instrumentation B 3.3.5.1
 
BASES
 
APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY
 
1.b, 2.b. Drywell PressureHigh
 
High pressure in the drywell could indicate a break in the reactor coolant pressure boundary (RCPB). The low pressure ECCS and associated DGs are initiated upon receipt of the Drywell Pressure      High Function with a Reactor Pressure      Low (Injection Permissive) in order to minimize the possibility of fuel damage. The DGs are initiated from Function 1.b signals. This Function also initiates the closure of the recirculation discharge valves to ensure the LPCI subsystems inject into the proper RPV location. The Drywell Pressure      High Function with a Reactor Pressure      Low (Injection Permissive), along with the Reactor Water Level      Low Low Low (Level 1) Function, is directly assumed in the analysis of the recirculation line break (Ref. 4).
The core cooling function of the ECCS, along with the scram action of the RPS, ensures that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46.
 
High drywell pressure signals are initiated from four pressure transmitters that sense drywell pressure. The Allowable Value was selected to be as low as possible and be indicative of a LOCA inside primary containment.
 
The Drywell Pressure      High Function is required to be OPERABLE when the ECCS or DG is required to be OPERABLE in conjunction with times when the primary containment is required to be OPERABLE. Thus, four channels of the CS and LPCI Drywell Pressure      High Function are required to be OPERABLE in MODES 1, 2, and 3 to ensure that no single instrument failure can preclude ECCS and DG initiation. In MODES 4 and 5, the Drywell Pressure      High Function is not required, since there is insufficient energy in the reactor to pressurize the primary containment to Drywell Pressure High setpoint. Refer to LCO 3.5.1 for Applicability Bases for the low pressure ECCS subsystems and to LCO 3.8.1 for Applicability Bases for the DGs.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                        B 3.3-101                                                                                                                                                                                Revision No. 145 ECCS Instrumentation B 3.3.5.1
 
BASES
 
APPLICABLE                                                                      1.c, 2.c. Reactor PressureLow (Injection Permissive)
SAFETY ANALYSES, LCO, and                                                                                          Low reactor pressure signals are used as permissives for the APPLICABILITY                                        low pressure ECCS subsystems. This ensures that, prior to (continued)                                        opening the injection valves of the low pressure ECCS subsystems or initiating the low pressure ECCS subsystems on a Drywell Pressure      High signal, the reactor pressure has fallen to a value below these subsystems' maximum design pressure and a break inside the RCPB has occurred respectively. This Function also provides permissive for the closure of the recirculation discharge valves to ensure the LPCI subsystems inject into the proper RPV location.
The Reactor Pressure      Low is one of the Functions assumed to be OPERABLE and capable of permitting initiation of the ECCS during the transients analyzed in References 1 and 3. In addition, the Reactor Pressure      Low Function is directly assumed in the analysis of the recirculation line break (Ref. 4). The core cooling function of the ECCS, along with the scram action of the RPS, ensures that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46.
 
The Reactor Pressure      Low signals are initiated from four pressure transmitters that sense the reactor dome pressure.
 
The Allowable Value is low enough to prevent overpressuring the equipment in the low pressure ECCS, but high enough to ensure that the ECCS injection prevents the fuel peak cladding temperature from exceeding the limits of 10 CFR 50.46.
 
Four channels of Reactor Pressure      Low Function are only required to be OPERABLE when the ECCS is required to be OPERABLE to ensure that no single instrument failure can preclude ECCS initiation.
 
1.d, 2.g. Core Spray and Low Pressure Coolant Injection Pump Discharge FlowLow (Bypass)
 
The minimum flow instruments are provided to protect the associated low pressure ECCS pump from overheating when the pump is operating and the associated injection valve is not fully open. The minimum flow line valve is opened when low flow is sensed, and the valve is automatically closed when the flow rate is adequate to protect the pump. The LPCI and
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                        B 3.3-102                                                                                                                                                                                Revision No. 145 ECCS Instrumentation B 3.3.5.1
 
BASES
 
APPLICABLE                                                                      1.d, 2.g. Core Spray and Low Pressure Coolant Injection SAFETY ANALYSES,          Pump Discharge FlowLow (Bypass)  (continued)
LCO, and APPLICABILITY                                        CS Pump Discharge Flow      Low Functions are assumed to be OPERABLE and capable of closing the minimum flow valves to ensure that the low pressure ECCS flows assumed during the transients and accidents analyzed in References 1, 2, and 3 are met. The core cooling function of the ECCS, along with the scram action of the RPS, ensures that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46.
 
One differential pressure switch per ECCS pump is used to detect the associated subsystems' flow rates. The logic is arranged such that each switch causes its associated minimum flow valve to open. The logic will close the minimum flow valve once the closure setpoint is exceeded. The LPCI minimum flow valves are time delayed such that the valves will not open for 10 seconds after the switches detect low flow. The time delay is provided to limit reactor vessel inventory loss during the startup of the RHR shutdown cooling mode. The Pump Discharge Flow      Low Allowable Values are high enough to ensure that the pump flow rate is sufficient to protect the pump, yet low enough to ensure that the closure of the minimum flow valve is initiated to allow full flow into the core.
 
Each channel of Pump Discharge Flow      Low Function (four CS channels and four LPCI channels) is only required to be OPERABLE when the associated ECCS is required to be OPERABLE to ensure that no single instrument failure can preclude the ECCS function.
 
1.e, 1.f. Core Spray Pump StartTime Delay Relay
 
The purpose of this time delay is to stagger the start of the CS pumps that are in each of Divisions I and II to prevent overloading the power source. This Function is necessary when power is being supplied from the offsite sources or the standby power sources (DG). The CS Pump Start      Time Delay Relays are assumed to be OPERABLE in the accident and transient analyses requiring ECCS initiation.
That is, the analyses assume that the pumps will initiate when required and excess loading will not cause failure of the power sources.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                        B 3.3-103                                                                                                                                                                                Revision No. 145 ECCS Instrumentation B 3.3.5.1
 
BASES
 
APPLICABLE                                                                      1.e, 1.f. Core Spray Pump StartTime Delay Relay SAFETY ANALYSES,          (continued)
LCO, and APPLICABILITY                                        There are eight Core Spray Pump Start      Time Delay Relays, two in each of the CS pump start logic circuits (one for when offsite power is available and one for when offsite power is not available). One of each type of time delay relay is dedicated to a single pump start logic, such that a single failure of a Core Spray Pump Start      Time Delay Relay will not result in the failure of more than one CS pump. In this condition, three of the four CS pumps will remain OPERABLE; thus, the single failure criterion is met (i.e.,
loss of one instrument does not preclude ECCS initiation).
The Allowable Value for the Core Spray Pump Start      Time Delay Relays is chosen to be long enough so that the power source will not be overloaded and short enough so that ECCS operation is not degraded.
 
Each channel of Core Spray Pump Start      Time Delay Relay Function is required to be OPERABLE only when the associated CS subsystem is required to be OPERABLE.
 
2.d. Reactor PressureLow Low (Recirculation Discharge Valve Permissive)
 
Low reactor pressure signals are used as permissives for recirculation discharge valve closure. This ensures that the LPCI subsystems inject into the proper RPV location assumed in the safety analysis. The Reactor Pressure      Low Low is one of the Functions assumed to be OPERABLE and capable of closing the valve during the transients analyzed in References 1 and 3. The core cooling function of the ECCS, along with the scram action of the RPS, ensures that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46. The Reactor Pressure      Low Low Function is directly assumed in the analysis of the recirculation line break (Ref. 4).
 
The Reactor Pressure      Low Low signals are initiated from four pressure transmitters that sense the reactor pressure.
 
The Allowable Value is chosen to ensure that the valves close prior to commencement of LPCI injection flow into the core, as assumed in the safety analysis.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                        B 3.3-104                                                                                                                                                                                Revision No. 145 ECCS Instrumentation B 3.3.5.1
 
BASES
 
APPLICABLE                                                                      2.d. Reactor PressureLow Low (Recirculation Discharge SAFETY ANALYSES,          Valve Permissive)  (continued)
LCO, and APPLICABILITY                                        Four channels of the Reactor Pressure      Low Low Function are only required to be OPERABLE in MODES 1, 2, and 3 with the associated recirculation pump discharge valve open. With the valve(s) closed, the function of the instrumentation has been performed; thus, the Function is not required. In MODES 4 and 5, the loop injection location is not critical since LPCI injection through the recirculation loop in either direction will still ensure that LPCI flow reaches the core (i.e., there is no significant reactor back pressure).
 
2.e. Reactor Vessel Shroud LevelLevel 0
 
The Reactor Vessel Shroud Level      Level 0 Function is provided as a permissive to allow the RHR System to be manually aligned from the LPCI mode to the suppression pool cooling/spray or drywell spray modes. The reactor vessel shroud level permissive ensures that water in the vessel is approximately two thirds core height before the manual transfer is allowed. This ensures that LPCI is available to prevent or minimize fuel damage. This function may be overridden during accident conditions as allowed by plant procedures. Reactor Vessel Shroud Level      Level 0 Function is implicitly assumed in the analysis of the recirculation line break (Ref. 4) since the analysis assumes that no LPCI flow diversion occurs when reactor water level is below Level 0.
 
Reactor Vessel Shroud Level      Level 0 signals are initiated from two level transmitters that sense the difference between the pressure due to a constant column of water (reference leg) and the pressure due to the actual water level (variable leg) in the vessel. The Reactor Vessel Shroud Level      Level 0 Allowable Value is chosen to allow the low pressure core flooding systems to activate and provide adequate cooling before allowing a manual transfer.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                        B 3.3-105                                                                                                                                                                                                    Revision No. 0 ECCS Instrumentation B 3.3.5.1
 
BASES
 
APPLICABLE                                                                      2.e. Reactor Vessel Shroud LevelLevel 0  (continued)
SAFETY ANALYSES, LCO, and                                                                                          Two channels of the Reactor Vessel Shroud Level      Level 0 APPLICABILITY                                        Function are only required to be OPERABLE in MODES 1, 2, and 3. In MODES 4 and 5, the specified initiation time of the LPCI subsystems is not assumed, and other administrative controls are adequate to control the valves associated with this Function (since the systems that the valves are opened for are not required to be OPERABLE in MODES 4 and 5 and are normally not used).
 
2.f. Low Pressure Coolant Injection Pump StartTime Delay Relay
 
The purpose of this time delay is to stagger the start of the LPCI pumps that are in each of Divisions I and II, to prevent overloading the power source. This Function is only necessary when power is being supplied from offsite sources.
The LPCI pumps start simultaneously with no time delay as soon as the standby source is available. The LPCI Pump Start      Time Delay Relays are assumed to be OPERABLE in the accident and transient analyses requiring ECCS initiation.
That is, the analyses assume that the pumps will initiate when required and excess loading will not cause failure of the power sources.
 
There are eight LPCI Pump Start      Time Delay Relays, two in each of the RHR pump start logic circuits. Two time delay relays are dedicated to a single pump start logic. Both timers in the RHR pump start logic would have to fail to prevent an RHR pump from starting within the required time; therefore, the low pressure ECCS pumps will remain OPERABLE; thus, the single failure criterion is met (i.e., loss of one instrument does not preclude ECCS initiation). The Allowable Values for the LPCI Pump Start      Time Delay Relays are chosen to be long enough so that most of the starting transient of the first pump is complete before starting the second pump on the same 4 kV emergency bus and short enough so that ECCS operation is not degraded.
 
Each channel of LPCI Pump Start      Time Delay Relay Function is required to be OPERABLE only when the associated LPCI subsystem is required to be OPERABLE.
 
continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                        B 3.3-106                                                                                                                                                                                Revision No. 145 ECCS Instrumentation B 3.3.5.1
 
BASES
 
APPLICABLE                                                                      High Pressure Coolant Injection (HPCI) System SAFETY ANALYSES, LCO, and                                                                                          3.a. Reactor Vessel Water LevelLow Low (Level 2)
APPLICABILITY (continued)                                        Low RPV water level indicates that the capability to cool the fuel may be threatened. Should RPV water level decrease too far, fuel damage could result. Therefore, the HPCI System is initiated at Level 2 to maintain level above the top of the active fuel. The Reactor Vessel Water Level      Low Low (Level 2) is one of the Functions assumed to be OPERABLE and capable of initiating HPCI during the transients analyzed in References 1 and 3. Additionally, the Reactor Vessel Water Level      Low Low (Level 2) Function associated with HPCI is credited as a backup to the Drywell Pressure      High Function for initiating HPCI in the analysis of the recirculation line break. The core cooling function of the ECCS, along with the scram action of the RPS, ensures that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46.
 
Reactor Vessel Water Level      Low Low (Level 2) signals are initiated from four level transmitters that sense the difference between the pressure due to a constant column of water (reference leg) and the pressure due to the actual water level (variable leg) in the vessel.
 
The Reactor Vessel Water Level      Low Low (Level 2) Allowable Value is high enough such that for complete loss of feedwater flow, the Reactor Core Isolation Cooling (RCIC)
System flow with HPCI assumed to fail will be sufficient to avoid initiation of low pressure ECCS at Reactor Vessel Water Level      Low Low Low (Level 1).
 
Four channels of Reactor Vessel Water Level      Low Low (Level 2) Function are required to be OPERABLE only when HPCI is required to be OPERABLE to ensure that no single instrument failure can preclude HPCI initiation. Refer to LCO 3.5.1 for HPCI Applicability Bases.
 
3.b. Drywell PressureHigh
 
High pressure in the drywell could indicate a break in the RCPB. The HPCI System is initiated upon receipt of the Drywell Pressure      High Function in order to minimize the possibility of fuel damage. The Drywell Pressure      High Function is directly assumed in the analysis of the
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                        B 3.3-107                                                                                                                                                                                                    Revision No. 0 ECCS Instrumentation B 3.3.5.1
 
BASES
 
APPLICABLE                                                                      3.b. Drywell PressureHigh  (continued)
SAFETY ANALYSES, LCO, and                                                                                          recirculation line break (Ref. 4). The core cooling APPLICABILITY                                        function of the ECCS, along with the scram action of the RPS, ensures that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46.
 
High drywell pressure signals are initiated from four pressure transmitters that sense drywell pressure. The Allowable Value was selected to be as low as possible to be indicative of a LOCA inside primary containment.
 
Four channels of the Drywell Pressure      High Function are required to be OPERABLE when HPCI is required to be OPERABLE to ensure that no single instrument failure can preclude HPCI initiation. Refer to LCO 3.5.1 for the Applicability Bases for the HPCI System.
 
3.c. Reactor Vessel Water LevelHigh (Level 8)
 
High RPV water level indicates that sufficient cooling water inventory exists in the reactor vessel such that there is no danger to the fuel. Therefore, the Level 8 signal is used to trip the HPCI turbine to prevent overflow into the main steam lines (MSLs). The Reactor Vessel Water Level      High (Level 8) Function is assumed to trip the HPCI turbine in the feedwater controller failure transient analysis if HPCI is initiated.
 
Reactor Vessel Water Level      High (Level 8) signals for HPCI are initiated from two level transmitters from the wide range water level measurement instrumentation. Both Level 8 signals are required in order to trip the HPCI turbine.
This ensures that no single instrument failure can preclude HPCI initiation. The Reactor Vessel Water Level      High (Level 8) Allowable Value is chosen to prevent flow from the HPCI System from overflowing into the MSLs.
 
Two channels of Reactor Vessel Water Level      High (Level 8)
Function are required to be OPERABLE only when HPCI is required to be OPERABLE. Refer to LCO 3.5.1 and LCO 3.5.2 for HPCI Applicability Bases.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                        B 3.3-108                                                                                                                                                                                                    Revision No. 0 ECCS Instrumentation B 3.3.5.1
 
BASES
 
APPLICABLE                                                                      3.d. Condensate Storage Tank LevelLow SAFETY ANALYSES, LCO, and                                                                                          Low level in the CST indicates the unavailability of an APPLICABILITY                                        adequate supply of makeup water from this normal source.
(continued)                                        Normally the suction valves between HPCI and the CST are open and, upon receiving a HPCI initiation signal, water for HPCI injection would be taken from the CST. However, if the water level in the CST falls below a preselected level, first the suppression pool suction valves automatically open, and then the CST suction valve automatically closes.
This ensures that an adequate supply of makeup water is available to the HPCI pump. To prevent losing suction to the pump, the suction valves are interlocked so that the suppression pool suction valves must be open before the CST suction valve automatically closes. The Function is implicitly assumed in the accident and transient analyses (which take credit for HPCI) since the analyses assume that the HPCI suction source is the suppression pool.
 
Condensate Storage Tank Level      Low signals are initiated from two level switches. The logic is arranged such that either level switch can cause the suppression pool suction valves to open and the CST suction valve to close. The Condensate Storage Tank Level      Low Function Allowable Value is high enough to ensure adequate pump suction head while water is being taken from the CST.
 
Two channels of the Condensate Storage Tank Level      Low Function are required to be OPERABLE only when HPCI is required to be OPERABLE to ensure that no single instrument failure can preclude HPCI swap to suppression pool source.
Refer to LCO 3.5.1 for HPCI Applicability Bases.
 
3.e. Suppression Pool Water LevelHigh
 
Excessively high suppression pool water could result in the loads on the suppression pool exceeding design values should there be a blowdown of the reactor vessel pressure through the safety/relief valves. Therefore, signals indicating high suppression pool water level are used to transfer the suction source of HPCI from the CST to the suppression pool to eliminate the possibility of HPCI continuing to provide additional water from a source outside containment. To prevent losing suction to the pump, the suction valves are interlocked so that the suppression pool suction valves must be open before the CST suction valve automatically closes.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                        B 3.3-109                                                                                                                                                                                                    Revision No. 0 ECCS Instrumentation B 3.3.5.1
 
BASES
 
APPLICABLE                                                                      3.e. Suppression Pool Water LevelHigh  (continued)
SAFETY ANALYSES, LCO, and                                                                                          This Function is implicitly assumed in the accident and APPLICABILITY                                        transient analyses (which take credit for HPCI) since the analyses assume that the HPCI suction source is the suppression pool.
 
Suppression Pool Water Level      High signals are initiated from two level switches. The logic is arranged such that either switch can cause the suppression pool suction valves to open and the CST suction valve to close. The Allowable Value for the Suppression Pool Water Level      High Function is chosen to ensure that HPCI will be aligned for suction from the suppression pool to prevent HPCI from contributing to any further increase in the suppression pool level.
 
Two channels of Suppression Pool Water Level      High Function are required to be OPERABLE only when HPCI is required to be OPERABLE to ensure that no single instrument failure can preclude HPCI swap to suppression pool source. Refer to LCO 3.5.1 for HPCI Applicability Bases.
 
3.f. High Pressure Coolant Injection Pump Discharge FlowLow (Bypass)
 
The minimum flow instrument is provided to protect the HPCI pump from overheating when the pump is operating at reduced flow. The minimum flow line valve is opened when low flow is sensed, and the valve is automatically closed when the flow rate is adequate to protect the pump. The High Pressure Coolant Injection Pump Discharge Flow      Low Function is assumed to be OPERABLE and capable of closing the minimum flow valve to ensure that the ECCS flow assumed during the transients analyzed in Reference 4 is met. The core cooling function of the ECCS, along with the scram action of the RPS, ensures that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46.
 
One flow switch is used to detect the HPCI System's flow rate. The logic is arranged such that the transmitter causes the minimum flow valve to open. The logic will close the minimum flow valve once the closure setpoint is exceeded.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                        B 3.3-110                                                                                                                                                                                                    Revision No. 0 ECCS Instrumentation B 3.3.5.1
 
BASES
 
APPLICABLE                                                                      3.f. High Pressure Coolant Injection Pump Discharge SAFETY ANALYSES,          FlowLow (Bypass)  (continued)
LCO, and APPLICABILITY                                        The High Pressure Coolant Injection Pump Discharge Flow      Low Allowable Value is high enough to ensure that pump flow rate is sufficient to protect the pump, yet low enough to ensure that the closure of the minimum flow valve is initiated to allow full flow into the core.
 
One channel is required to be OPERABLE when the HPCI is required to be OPERABLE. Refer to LCO 3.5.1 for HPCI Applicability Bases.
 
Automatic Depressurization System
 
4.a, 5.a. Reactor Vessel Water LevelLow Low Low (Level 1)
 
Low RPV water level indicates that the capability to cool the fuel may be threatened. Should RPV water level decrease too far, fuel damage could result. Therefore, ADS receives one of the signals necessary for initiation from this Function. This signal actuates the Function 4.h, 5.h timer.
The Reactor Vessel Water Level      Low Low Low (Level 1) is one of the Functions assumed to be OPERABLE and capable of initiating the ADS during the accident analyzed in Reference 4. The core cooling function of the ECCS, along with the scram action of the RPS, ensures that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46.
 
Reactor Vessel Water Level      Low Low Low (Level 1) signals are initiated from four level transmitters that sense the difference between the pressure due to a constant column of water (reference leg) and the pressure due to the actual water level (variable leg) in the vessel. Four channels of Reactor Vessel Water Level      Low Low Low (Level 1) Function are required to be OPERABLE only when ADS is required to be OPERABLE to ensure that no single instrument failure can preclude ADS initiation. Two channels input to ADS trip system A, while the other two channels input to ADS trip system B. Refer to LCO 3.5.1 for ADS Applicability Bases.
 
The Reactor Vessel Water Level      Low Low Low (Level 1)
Allowable Value is chosen to allow time for the low pressure core flooding systems to initiate and provide adequate cooling.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                        B 3.3-111                                                                                                                                                                                          Revision No. 78 ECCS Instrumentation B 3.3.5.1
 
BASES
 
APPLICABLE                                                                      4.b, 5.b. Drywell PressureHigh SAFETY ANALYSES, LCO, and                                                                                          High pressure in the drywell could indicate a break in the APPLICABILITY                                        RCPB. Therefore, ADS receives one of the signals necessary (continued)                                        for initiation from this Function in order to minimize the possibility of fuel damage. The Drywell Pressure      High is assumed to be OPERABLE and capable of initiating the ADS during the accidents analyzed in Reference 4. The core cooling function of the ECCS, along with the scram action of the RPS, ensures that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46.
 
Drywell Pressure      High signals are initiated from four pressure transmitters that sense drywell pressure. The Allowable Value was selected to be as low as possible and be indicative of a LOCA inside primary containment.
 
Four channels of Drywell Pressure      High Function are only required to be OPERABLE when ADS is required to be OPERABLE to ensure that no single instrument failure can preclude ADS initiation. Two channels input to ADS trip system A, while the other two channels input to ADS trip system B. Refer to LCO 3.5.1 for ADS Applicability Bases.
 
4.c, 5.c. Automatic Depressurization System Initiation Timer
 
The purpose of the Automatic Depressurization System Initiation Timer is to delay depressurization of the reactor vessel to allow the HPCI System time to maintain reactor vessel water level. Since the rapid depressurization caused by ADS operation is one of the most severe transients on the reactor vessel, its occurrence should be limited. By delaying initiation of the ADS Function, the operator is given the chance to monitor the success or failure of the HPCI System to maintain water level, and then to decide whether or not to allow ADS to initiate, to delay initiation further by recycling the timer, or to inhibit initiation permanently. The Automatic Depressurization System Initiation Timer Function is assumed to be OPERABLE for the accident analysis of Reference 4 that requires ECCS initiation and assumes failure of the HPCI System.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                        B 3.3-112                                                                                                                                                                                                    Revision No. 0 ECCS Instrumentation B 3.3.5.1
 
BASES
 
APPLICABLE                                                                      4.c, 5.c. Automatic Depressurization System Initiation SAFETY ANALYSES,          Timer  (continued)
LCO, and APPLICABILITY                                        There are two Automatic Depressurization System Initiation Timer relays, one in each of the two ADS trip systems. The Allowable Value for the Automatic Depressurization System Initiation Timer is chosen so that there is still time after depressurization for the low pressure ECCS subsystems to provide adequate core cooling.
 
Two channels of the Automatic Depressurization System Initiation Timer Function are only required to be OPERABLE when the ADS is required to be OPERABLE to ensure that no single instrument failure can preclude ADS initiation.  (One channel inputs to ADS trip system A, while the other channel inputs to ADS trip system B. Refer to LCO 3.5.1 for ADS Applicability Bases.
 
4.d, 5.d. Reactor Vessel Water Level Low Low Low (Level 1) (Permissive)
 
Low reactor water level signals are used as permissives in the ADS trip systems. This ensures after a high drywell pressure signal or a low reactor water level signal (Level 1) is received and the timer times out that a low reactor water level (Level 1), signal is present to allow the ADS initiation (after a confirmatory Level 4 signal, see Bases for Functions 4.e, 5.e, Reactor Vessel Water Confirmatory Level      Low (Level 4).
 
Reactor Vessel Water Level      Low Low Low (Level 1), signals are initiated from four level transmitters that sense the difference between the pressure due to a constant column of water (reference leg) and the pressure doe to the actual water level (variable leg) in the vessel. The Reactor Vessel Water Level      Low Low Low (Level 1) Allowable Value is chosen to allow time for the low pressure core flooding system to initiate and provide adequate cooling.
 
Four channels of the Reactor Vessel Water Level      Low Low Low (Level 1) Function are required to be OPERABLE to ensure that no single instrument failure can preclude ADS initiation. Two channels input to ADS trip system A while the other two channels input to ADS trip system B. Refer to LCO 3.5.1 for ADS Applicability Bases.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                        B 3.3-113                                                                                                                                                                                                    Revision No. 0 ECCS Instrumentation B 3.3.5.1
 
BASES
 
APPLICABLE                                                                      4.e, 5.e. Reactor Vessel Water Confirmatory LevelLow SAFETY ANALYSES,          (Level 4)
LCO, and APPLICABILITY                                        The Reactor Vessel Water Confirmatory Level      Low (Level 4)
(continued)                                        Function is used by the ADS only as a confirmatory low water level signal. ADS receives one of the signals necessary for initiation from Reactor Vessel Water Level      Low Low Low (Level 1) signals. In order to prevent spurious initiation of the ADS due to spurious Level 1 signals, a Level 4 signal must also be received before ADS initiation commences.
 
Reactor Vessel Water Confirmatory Level      Low (Level 4) signals are initiated from two level transmitters that sense the difference between the pressure due to a constant column of water (reference leg) and the pressure due to the actual water level (variable leg) in the vessel. The Allowable Value for Reactor Vessel Water Confirmatory Level      Low (Level 4) is selected to be above the RPS Level 3 scram Allowable Value for convenience.
 
Two channels of Reactor Vessel Water Confirmatory Level      Low (Level 4) Function are only required to be OPERABLE when the ADS is required to be OPERABLE to ensure that no single instrument failure can preclude ADS initiation. One channel inputs to ADS trip system A, while the other channel inputs to ADS trip system B. Refer to LCO 3.5.1 for ADS Applicability Bases.
 
4.f, 4.g, 5.f, 5.g. Core Spray and Low Pressure Coolant Injection Pump Discharge PressureHigh
 
The Pump Discharge Pressure      High signals from the CS and LPCI pumps are used as permissives for ADS initiation, indicating that there is a source of low pressure cooling water available once the ADS has depressurized the vessel.
Pump Discharge Pressure      High is one of the Functions assumed to be OPERABLE and capable of permitting ADS initiation during the events analyzed in Reference 4 with an assumed HPCI failure. For these events the ADS depressurizes the reactor vessel so that the low pressure ECCS can perform the core cooling functions. This core cooling function of the ECCS, along with the scram action of the RPS, ensures that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                        B 3.3-114                                                                                                                                                                                                    Revision No. 0 ECCS Instrumentation B 3.3.5.1
 
BASES
 
APPLICABLE                                                                      4.f, 4.g, 5.f, 5.g. Core Spray and Low Pressure Coolant SAFETY ANALYSES,          Injection Pump Discharge PressureHigh  (continued)
LCO, and APPLICABILITY                                        Pump discharge pressure signals are initiated from twelve pressure transmitters, two on the discharge side of each of the four LPCI pumps and one on the discharge side of each CS pump. There are two ADS low pressure ECCS pump permissives in each trip system. Each of the permissives receives inputs from all four LPCI pumps (different signals for each permissive) and two CS pumps, one from each subsystem (different pumps for each permissive). In order to generate an ADS permissive in one trip system, it is necessary that only one LPCI pump or two CS pumps in proper combination (C or D and A or B) indicate the high discharge pressure condition in each of the two permissives. The Pump Discharge Pressure      High Allowable Value is less than the pump discharge pressure when the pump is operating in a full flow mode and high enough to avoid any condition that results in a discharge pressure permissive when the CS and LPCI pumps are aligned for injection and the pumps are not running. The actual operating point of this function is not assumed in any transient or accident analysis. However, this Function is indirectly assumed to operate (in Reference
: 4) to provide the ADS permissive to depressurize the RCS to allow the ECCS low pressure systems to operate.
 
Twelve channels of Core Spray and Low Pressure Coolant Injection Pump Discharge Pressure      High Function are only required to be OPERABLE when the ADS is required to be OPERABLE to ensure that no single instrument failure can preclude ADS initiation. Four CS channels associated with CS pumps A through D and eight LPCI channels associated with LPCI pumps A through D are required for both trip systems.
Refer to LCO 3.5.1 for ADS Applicability Bases.
 
4.h, 5.h. Automatic Depressurization System Low Water Level Actuation Timer
 
One of the signals required for ADS initiation is Drywell Pressure      High. However, if the event requiring ADS initiation occurs outside the drywell (e.g., main steam line break outside containment), a high drywell pressure signal may never be present. Therefore, the Automatic Depressurization System Low Water Level Actuation Timer is used to bypass the Drywell Pressure      High Function after a
 
continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                        B 3.3-115                                                                                                                                                                                                    Revision No. 0 ECCS Instrumentation B 3.3.5.1
 
BASES
 
APPLICABLE                                                                      4.h, 5.h. Automatic Depressurization System Low Water Level SAFETY ANALYSES,          Actuation Timer  (continued)
LCO, and APPLICABILITY                                        certain time period has elapsed. Operation of the Automatic Depressurization System Low Water Level Actuation Timer Function is assumed in the accident analysis of Reference 4 that requires ECCS initiation and assumes failure of the HPCI system.
 
There are four Automatic Depressurization System Low Water Level Actuation Timer relays, two in each of the two ADS trip systems. The Allowable Value for the Automatic Depressurization System Low Water Level Actuation Timer is chosen to ensure that there is still time after depressurization for the low pressure ECCS subsystems to provide adequate core cooling.
 
Four channels of the Automatic Depressurization System Low Water Level Actuation Timer Function are only required to be OPERABLE when the ADS is required to be OPERABLE to ensure that no single instrument failure can preclude ADS initiation. Refer to LCO 3.5.1 for ADS Applicability Bases.
 
ACTIONS                                                                                                    A Note has been provided to modify the ACTIONS related to ECCS instrumentation channels. 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 inoperable ECCS instrumentation channels provide appropriate compensatory measures for separate inoperable Condition entry for each inoperable ECCS instrumentation channel.
 
A.1
 
Required Action A.1 directs entry into the appropriate Condition referenced in Table 3.3.5.1-1. The applicable Condition referenced in the table is Function dependent.
Each time a channel is discovered inoperable, Condition A is entered for that channel and provides for transfer to the appropriate subsequent Condition.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                        B 3.3-116                                                                                                                                                                                                    Revision No. 0 ECCS Instrumentation B 3.3.5.1
 
BASES
 
ACTIONS                                                                                                    B.1, B.2, and B.3 (continued)
Required Actions B.1 and B.2 are intended to ensure that appropriate actions are taken if multiple, inoperable, untripped channels within the same Function result in redundant automatic initiation capability being lost for the feature(s). Required Action B.1 features would be those that are initiated by Functions 1.a, 1.b, 2.a, and 2.b (e.g., low pressure ECCS). The Required Action B.2 system would be HPCI. For Required Action B.1, redundant automatic initiation capability is lost if (a) two or more Function 1.a channels are inoperable and untripped such that both trip systems lose initiation capability, (b) two or more Function 2.a channels are inoperable and untripped such that both trip systems lose initiation capability, (c) two or more Function 1.b channels are inoperable and untripped such that both trip systems lose initiation capability, or (d) two or more Function 2.b channels are inoperable and untripped such that both trip systems lose initiation capability. For low pressure ECCS, since each inoperable channel would have Required Action B.1 applied separately (refer to ACTIONS Note), each inoperable channel would only require the affected portion of the associated system of low pressure ECCS and DGs to be declared inoperable. However, since channels in both associated low pressure ECCS subsystems (e.g., both CS subsystems) are inoperable and untripped, and the Completion Times started concurrently for the channels in both subsystems, this results in the affected portions in the associated low pressure ECCS and DGs being concurrently declared inoperable.
 
For Required Action B.2, redundant automatic HPCI initiation capability is lost if two or more Function 3.a or two Function 3.b channels are inoperable and untripped such that the trip system loses initiation capability. In this situation (loss of redundant automatic initiation capability), the 24 hour allowance of Required Action B.3 is not appropriate and the HPCI System must be declared inoperable within 1 hour.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                        B 3.3-117                                                                                                                                                                                Revision No. 145 ECCS Instrumentation B 3.3.5.1
 
BASES
 
ACTIONS                                                                                                    B.1, B.2, and B.3  (continued)
 
Notes are also provided (the Note to Required Action B.1 and the Note to Required Action B.2) to delineate which Required Action is applicable for each Function that requires entry into Condition B if an associated channel is inoperable.
This ensures that the proper loss of initiation capability check is performed. Required Action B.1 (the Required Action for certain inoperable channels in the low pressure ECCS subsystems) is not applicable to Function 2.e, since this Function provides backup to administrative controls ensuring that operators do not divert LPCI flow from injecting into the core when needed. Thus, a total loss of Function 2.e capability for 24 hours is allowed, since the LPCI subsystems remain capable of performing their intended function.
 
The Completion Time is intended to allow the operator time to evaluate and repair any discovered inoperabilities. This Completion Time also allows for an exception to the normal "time zero" for beginning the allowed outage time "clock."
For Required Action B.1, the Completion Time only begins upon discovery that a redundant feature in the same system (e.g.,
both CS subsystems) cannot be automatically initiated due to inoperable, untripped channels within the same  Function as described in the paragraph above. For Required Action B.2, the Completion Time only begins upon discovery that the HPCI System cannot be automatically initiated due to two inoperable, untripped channels for the associated Function in the same trip system. The 1 hour Completion Time from discovery of loss of initiation capability is acceptable because it minimizes risk while allowing time for restoration or tripping of channels.
 
Because of the diversity of sensors available to provide initiation signals and the redundancy of the ECCS design, an allowable out of service time of 24 hours has been shown to be acceptable (Ref. 5) to permit restoration of any inoperable channel to OPERABLE status. Alternatively, a Completion Time can be determined in accordance with the Risk Informed Completion Time (RICT) Program. A Note has been provided to indicate that a RICT is only applicable when a loss of function has not occurred. If the inoperable channel cannot be restored to OPERABLE status within the
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                        B 3.3-118                                                                                                                                                                                Revision No. 159 ECCS Instrumentation B 3.3.5.1
 
BASES
 
ACTIONS                                                                                                    B.1, B.2, and B.3  (continued)
 
allowable out of service time, the channel must be placed in the tripped condition per Required Action B.3. Placing the inoperable channel in trip would conservatively compensate for the inoperability, restore capability to accommodate a single failure, and allow operation to continue.
Alternately, if it is not desired to place the channel in trip (e.g., as in the case where placing the inoperable channel in trip would result in an initiation), Condition H must be entered and its Required Action taken.
 
C.1 and C.2
 
Required Action C.1 is intended to ensure that appropriate actions are taken if multiple, inoperable channels within the same Function result in redundant automatic initiation capability being lost for the feature(s). Required Action C.1 features would be those that are initiated by Functions 1.c, 1.e, 1.f, 2.c, 2.d, and 2.f (i.e., low pressure ECCS). Redundant automatic initiation capability is lost if either (a) two or more Function 1.c channels are inoperable in the same trip system such that the trip system loses initiation capability, (b) two or more Function 1.e channels are inoperable affecting CS pumps in different subsystems, (c) two or more Function 1.f channels are inoperable affecting CS pumps in different subsystems, (d) two or more Function 2.c channels are inoperable in the same trip system such that the trip system loses initiation capability, (e) two or more Function 2.d channels are inoperable in the same trip system such that the trip system loses initiation capability, or (f) three or more Function 2.f channels are inoperable. In this situation (loss of redundant automatic initiation capability), the 24 hour allowance of Required Action C.2 is not appropriate and the feature(s) associated with the inoperable channels must be declared inoperable within 1 hour. Since each inoperable channel would have Required Action C.1 applied separately (refer to ACTIONS Note), each inoperable channel would only require the affected portion of the associated system to be declared inoperable. However, since channels for both low pressure ECCS subsystems are inoperable (e.g.,
both CS subsystems), and the Completion Times started concurrently for the channels in both subsystems, this results in the affected portions in both subsystems being
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                        B 3.3-119                                                                                                                                                                                                    Revision No. 0 ECCS Instrumentation B 3.3.5.1
 
BASES
 
ACTIONS                                                                                                    C.1 and C.2  (continued)
 
concurrently declared inoperable. For Functions 1.c, 1.e, 1.f, 2.c, 2.d, and 2.f, the affected portions are the associated low pressure ECCS pumps.
 
The Note states that Required Action C.1 is only applicable for Functions 1.c, 1.e, 1.f, 2.c, 2.d, and 2.f. Required Action C.1 is not applicable to Function 3.c (which also requires entry into this Condition if a channel in this Function is inoperable), since the loss of one channel results in a loss of the Function (two-out-of-two logic).
This loss was considered during the development of Reference 5 and considered acceptable for the 24 hours allowed by Required Action C.2.
 
The Completion Time is intended to allow the operator time to evaluate and repair any discovered inoperabilities. This Completion Time also allows for an exception to the normal "time zero" for beginning the allowed outage time "clock."
For Required Action C.1, the Completion Time only begins upon discovery that the same feature in both subsystems (e.g.,
both CS subsystems) cannot be automatically initiated due to inoperable channels within the same Function as described in the paragraph above. The 1 hour Completion Time from discovery of loss of initiation capability is acceptable because it minimizes risk while allowing time for restoration of channels.
 
Because of the diversity of sensors available to provide initiation signals and the redundancy of the ECCS design, an allowable out of service time of 24 hours has been shown to be acceptable (Ref. 5) to permit restoration of any inoperable channel to OPERABLE status. Alternatively, a Completion Time can be determined in accordance with the Risk Informed Completion Time (RICT) Program. A Note has been provided to indicate that a RICT is only applicable when a loss of function has not occurred.If the inoperable channel cannot be restored to OPERABLE status within the allowable out of service time, Condition H must be entered and its Required Action taken. The Required Actions do not allow placing the channel in trip since this action would either cause the initiation or it would not necessarily result in a safe state for the channel in all events.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                        B 3.3-120                                                                                                                                                                                Revision No. 159 ECCS Instrumentation B 3.3.5.1
 
BASES
 
ACTIONS                                                                                                    D.1, D.2.1, and D.2.2 (continued)
Required Action D.1 is intended to ensure that appropriate actions are taken if multiple, inoperable, untripped channels within the same Function result in a complete loss of automatic component initiation capability for the HPCI System. Automatic component initiation capability is lost if two Function 3.d channels or two Function 3.e channels are inoperable and untripped. In this situation (loss of automatic suction swap), the 24 hour allowance of Required Actions D.2.1 and D.2.2 is not appropriate and the HPCI System must be declared inoperable within 1 hour after discovery of loss of HPCI initiation capability. As noted, Required Action D.1 is only applicable if the HPCI pump suction is not aligned to the suppression pool, since, if aligned, the Function is already performed.
 
The Completion Time is intended to allow the operator time to evaluate and repair any discovered inoperabilities. This Completion Time also allows for an exception to the normal "time zero" for beginning the allowed outage time "clock."
For Required Action D.1, the Completion Time only begins upon discovery that the HPCI System cannot be automatically aligned to the suppression pool due to two inoperable, untripped channels in the same Function. The 1 hour Completion Time from discovery of loss of initiation capability is acceptable because it minimizes risk while allowing time for restoration or tripping of channels.
 
Because of the diversity of sensors available to provide initiation signals and the redundancy of the ECCS design, an allowable out of service time of 24 hours has been shown to be acceptable (Ref. 5) to permit restoration of any inoperable channel to OPERABLE status. Alternatively, a Completion Time can be determined in accordance with the Risk Informed Completion Time (RICT) Program. A Note has been provided to indicate that a RICT is only applicable when a loss of function has not occurred. If the inoperable channel cannot be restored to OPERABLE status within the allowable out of service time, the channel must be placed in the tripped condition per Required Action D.2.1 or the suction source must be aligned to the suppression pool per Required Action D.2.2. Placing the inoperable channel in trip performs the intended function of the channel (shifting the suction source to the suppression pool). Performance of either of these two Required Actions will allow operation to continue. If Required Action D.2.1 or D.2.2 is performed, measures should be taken to ensure that the HPCI System
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                        B 3.3-121                                                                                                                                                                                Revision No. 159 ECCS Instrumentation B 3.3.5.1
 
BASES
 
ACTIONS                                                                                                    D.1, D.2.1, and D.2.2  (continued)
 
piping remains filled with water. Alternately, if it is not desired to perform Required Actions D.2.1 and D.2.2 (e.g.,
as in the case where shifting the suction source could drain down the HPCI suction piping), Condition H must be entered and its Required Action taken.
 
E.1 and E.2
 
Required Action E.1 is intended to ensure that appropriate actions are taken if multiple, inoperable channels within the Core Spray and Low Pressure Coolant Injection Pump, Discharge Flow - Low (Bypass) Functions result in redundant automatic initiation capability being lost for the feature(s). For Required Action E.1, the features would be those that are initiated by Functions 1.d and 2.g (e.g., low pressure ECCS). Redundant automatic initiation capability is lost if (a) two or more Function 1.d channels are inoperable affecting CS pumps in different subsystems or (b) three or more Function 2.g channels are inoperable.
Since each inoperable channel would have Required Action E.1 applied separately (refer to ACTIONS Note), each inoperable channel would only require the affected low pressure ECCS pump to be declared inoperable. However, since channels for more than one low pressure ECCS pump are inoperable, and the Completion Times started concurrently for the channels of the low pressure ECCS pumps, this results in the affected low pressure ECCS pumps being concurrently declared inoperable.
 
In this situation (loss of redundant automatic initiation capability), the 7 day allowance of Required Action E.2 is not appropriate and the subsystem associated with each inoperable channel must be declared inoperable within 1 hour. A Note is also provided (Note 2 to Required Action E.1) to delineate that Required Action E.1 is only applicable to low
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                        B 3.3-122                                                                                                                                                                                Revision No. 145 ECCS Instrumentation B 3.3.5.1
 
BASES
 
ACTIONS                                                                                                    E.1 and E.2  (continued)
 
pressure ECCS Functions. Required Action E.1 is not applicable to HPCI Function 3.f since the loss of one channel results in a loss of function (one-out-of-one logic). This loss was considered during the development of Reference 5 and considered acceptable for the 7 days allowed by Required Action E.2. Alternatively, a Completion Time can be determined in accordance with the Risk Informed Completion Time (RICT) Program. A Note has been provided to indicate that a RICT is only applicable when a loss of function has not occurred.
 
The Completion Time is intended to allow the operator time to evaluate and repair any discovered inoperabilities. This Completion Time also allows for an exception to the normal "time zero" for beginning the allowed outage time "clock."
 
For Required Action E.1, the Completion Time only begins upon discovery that a redundant feature in the same system (e.g., both CS subsystems) cannot be automatically initiated due to inoperable channels within the same Function as described in the paragraph above. The 1 hour Completion Time from discovery of loss of initiation capability is acceptable because it minimizes risk while allowing time for restoration of channels.
 
If the instrumentation that controls the pump minimum flow valve is inoperable, such that the valve will not automatically open, extended pump operation with no injection path available could lead to pump overheating and failure. If there were a failure of the instrumentation, such that the valve would not automatically close, a portion of the pump flow could be diverted from the reactor vessel injection path, causing insufficient core cooling. These consequences can be averted by the operator's manual control of the valve, which would be adequate to maintain ECCS pump protection and required flow. Furthermore, other ECCS pumps would be sufficient to complete the assumed safety function if no additional single failure were to occur. The 7 day Completion Time of Required Action E.2 to restore the inoperable channel to OPERABLE status is reasonable based on the remaining capability of the associated ECCS subsystems, the redundancy available in the ECCS design, and the low probability of a DBA occurring during the allowed out of service time. If the inoperable channel cannot be restored to OPERABLE status within the allowable out of service time, Condition H must be entered and its Required Action taken.
The Required Actions do not allow placing the channel in trip since this action would not necessarily result in a safe state for the channel in all events.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                        B 3.3-123                                                                                                                                                                                Revision No. 159 ECCS Instrumentation B 3.3.5.1 BASES
 
ACTIONS                                                                                                    F.1 and F.2 (continued)
Required Action F.1 is intended to ensure that appropriate actions are taken if multiple, inoperable, untripped channels within similar ADS trip system A and B Functions result in redundant automatic initiation capability being lost for the ADS. For example, redundant automatic initiation capability is lost if either (a) one or more Function 4.a channel and one or more Function 5.a channel are inoperable and untripped, (b) one or more Function 4.b channel and one or more Function 5.b channel are inoperable and untripped, (c) one or more Function 4.d channel and one or more Function 5.d channel are inoperable and untripped, or (d) one Function 4.e channel and one Function 5.e channel are inoperable and untripped.
In this situation (loss of automatic initiation capability), the 96 hour or 8 day allowance, as applicable, of Required Action F.2 is not appropriate and all ADS valves must be declared inoperable within 1 hour after discovery of loss of ADS initiation capability.
The Completion Time is intended to allow the operator time to evaluate and repair any discovered inoperabilities. This Completion Time also allows for an exception to the normal "time zero" for beginning the allowed outage time "clock."
For Required Action F.1, the Completion Time only begins upon discovery that the ADS cannot be automatically initiated due to inoperable, untripped channels within similar ADS trip system Functions as described in the paragraph above. The 1 hour Completion Time from discovery of loss of initiation capability is acceptable because it minimizes risk while allowing time for restoration or Because of the diversity of sensors available to provide tripping of channels.
 
initiation signals and the redundancy of the ECCS design, an allowable out of service time of 8 days has been shown to be acceptable (Ref. 5) to permit restoration of any inoperable channel to OPERABLE status if both HPCI and RCIC are OPERABLE. Alternatively, a Completion Time can be determined in accordance with the Risk Informed Completion Time (RICT)
Program. A Note has been provided to indicate that a RICT is only applicable when a loss of function has not occurred.
If either HPCI or RCIC is inoperable, the time is shortened to 96 hours or in accordance with the Risk Informed Completion Time (RICT) Program. A Note has been provided to indicate that a RICT is only applicable when a loss of function has not occurred. If the status of HPCI or RCIC changes such that the Completion Time changes from 8 days to 96 hours or in accordance with the Risk Informed Completion Time Program., the time begins upon discovery of HPCI or RCIC inoperability. However, the total time for an inoperable, untripped channel cannot exceed 8 days or in accordance with the Risk Informed Completion Time Program.
If the status of (continued)
PBAPS UNIT 2                                                                                                                                                                                                                        B 3.3-124                                                                                                                                                                                Revision No. 159 ECCS Instrumentation B 3.3.5.1
 
BASES
 
ACTIONS                                                                                                    F.1 and F.2  (continued)
 
HPCI or RCIC changes such that the Completion Time changes from 96 hours to 8 days, the "time zero" for beginning the 8 day "clock" begins upon discovery of the inoperable, untripped channel. If the inoperable channel cannot be restored to OPERABLE status within the allowable out of service time, the channel must be placed in the tripped condition per Required Action F.2. Placing the inoperable channel in trip would conservatively compensate for the inoperability, restore capability to accommodate a single failure, and allow operation to continue. Alternately, if it is not desired to place the channel in trip (e.g., as in the case where placing the inoperable channel in trip would result in an initiation), Condition H must be entered and its Required Action taken.
 
G.1 and G.2
 
Required Action G.1 is intended to ensure that appropriate actions are taken if multiple, inoperable channels within similar ADS trip system Functions result in automatic initiation capability being lost for the ADS. For example, automatic initiation capability is lost if either (a) one Function 4.c channel and one Function 5.c channel are inoperable, (b) a combination of Function 4.f, 4.g, 5.f, and 5.g channels are inoperable such that channels associated with five or more low pressure ECCS pumps are inoperable, or (c) one or more Function 4.h channels and one or more Function 5.h channels are inoperable.
 
In this situation (loss of automatic initiation capability),
the 96 hour or 8 day allowance, as applicable, of Required Action G.2 is not appropriate, and all ADS valves must be declared inoperable within 1 hour after discovery of loss of ADS initiation capability.
 
The Completion Time is intended to allow the operator time to evaluate and repair any discovered inoperabilities. This Completion Time also allows for an exception to the normal "time zero" for beginning the allowed outage time "clock."
For Required Action G.1, the Completion Time only begins
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                        B 3.3-125                                                                                                                                                                                          Revision No. 83 ECCS Instrumentation B 3.3.5.1
 
BASES
 
ACTIONS                                                                                                    G.1 and G.2  (continued)
 
upon discovery that the ADS cannot be automatically initiated due to inoperable channels within similar ADS trip system Functions as described in the paragraph above. The 1 hour Completion Time from discovery of loss of initiation capability is acceptable because it minimizes risk while allowing time for restoration or tripping of channels.
 
Because of the diversity of sensors available to provide initiation signals and the redundancy of the ECCS design, an allowable out of service time of 8 days has been shown to be acceptable (Ref. 5) to permit restoration of any inoperable channel to OPERABLE status if both HPCI and RCIC are OPERABLE (Required Action G.2). Alternatively, a Completion Time can be determined in accordance with the Risk Informed Completion Time (RICT) Program. A Note has been provided to indicate that a RICT is only applicable when a loss of function has not occurred. If either HPCI or RCIC is inoperable, the time shortens to 96 hours or in accordance with the Risk Informed Completion Time (RICT) Program. A Note has been provided to indicate that a RICT is only applicable when a loss of function has not occurred. If the status of HPCI or RCIC changes such that the Completion Time changes from 8 days to 96 hours or in accordance with the Risk Informed Completion Time Program, the time begins upon discovery of HPCI or RCIC inoperability. However, the total time for an inoperable channel cannot exceed 8 days or in accordance with the Risk Informed Completion Time Program.
If the status of HPCI or RCIC changes such that the Completion Time changes from 96 hours to 8 days, the "time zero" for beginning the 8 day "clock" begins upon discovery of the inoperable channel. If the inoperable channel cannot be restored to OPERABLE status within the allowable out of service time, Condition H must be entered and its Required Action taken. The Required Actions do not allow placing the channel in trip since this action would not necessarily result in a safe state for the channel in all events.
 
H.1
 
With any Required Action and associated Completion Time not met, the associated feature(s) may be incapable of performing the intended function, and the supported feature(s) associated with inoperable untripped channels must be declared inoperable immediately.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                        B 3.3-126                                                                                                                                                                                Revision No. 159 ECCS Instrumentation B 3.3.5.1
 
BASES  (continued)
 
SURVEILLANCE                                                  As noted in the beginning of the SRs, the SRs for each ECCS REQUIREMENTS                                                  instrumentation Function are found in the SRs column of Table 3.3.5.1-1.
 
The Surveillances are modified by a Note to indicate that when a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours as follows:  (a) for Functions 3.c and 3.f; and (b) for Functions other than 3.c and 3.f provided the associated Function or the redundant Function maintains ECCS initiation capability. Upon completion of the Surveillance, or expiration of the 6 hour allowance, the channel must be returned to OPERABLE status or the applicable Condition entered and Required Actions taken. This Note is based on the reliability analysis (Ref. 5) assumption of the average time required to perform channel surveillance. That analysis demonstrated that the 6 hour testing allowance does not significantly reduce the probability that the ECCS will initiate when necessary.
 
SR  3.3.5.1.1
 
Performance of the CHANNEL CHECK ensures that a gross failure of instrumentation has not occurred. A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a similar parameter on other channels. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value.
Significant deviations between the instrument channels could be an indication of excessive instrument drift in one of the channels or something even more serious. A CHANNEL CHECK guarantees that undetected outright channel failure is limited; thus, it is key to verifying the instrumentation continues to operate properly between each CHANNEL CALIBRATION.
 
Agreement criteria are determined by the plant staff, based on a combination of the channel instrument uncertainties, including indication and readability. If a channel is outside the criteria, it may be an indication that the instrument has drifted outside its limit.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                        B 3.3-127                                                                                                                                                                                          Revision No. 86 ECCS Instrumentation B 3.3.5.1
 
BASES
 
SURVEILLANCE                                                  SR  3.3.5.1.1  (continued)
REQUIREMENTS The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. The CHANNEL CHECK supplements less formal, but more frequent, checks of channels during normal operational use of the displays associated with the channels required by the LCO.
 
SR  3.3.5.1.2
 
A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the entire channel will perform the intended function. Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology.
 
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
 
SR  3.3.5.1.3 and SR  3.3.5.1.4
 
A CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor. This test verifies the channel responds to the measured parameter within the necessary range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drifts between successive calibrations, consistent with the assumptions of the current plant specific setpoint methodology.
 
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                        B 3.3-128                                                                                                                                                                                          Revision No. 86 ECCS Instrumentation B 3.3.5.1
 
BASES
 
SURVEILLANCE                                                  SR  3.3.5.1.5 REQUIREMENTS (continued)                                        The LOGIC SYSTEM FUNCTIONAL TEST demonstrates the OPERABILITY of the required initiation logic for a specific channel. The system functional testing performed in LCO 3.5.1, LCO 3.5.2, LCO 3.8.1, and LCO 3.8.2 overlaps this Surveillance to complete testing of the assumed safety function.
 
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
 
REFERENCES                                                                      1.                    UFSAR, Section 6.5.
: 2.                    UFSAR, Section 7.4.
: 3.                    UFSAR, Chapter 14.
: 4.                    NEDC-32163-P, "Peach Bottom Atomic Power Station Units 2 and 3, SAFER/GESTR-LOCA, Loss-of-Coolant Accident Analysis," January 1993.
: 5.                    NEDC-30936-P-A, "BWR Owners' Group Technical Specification Improvement Analyses for ECCS Actuation Instrumentation, Part 2," December 1988.
 
PBAPS UNIT 2                                                                                                                                                                                                                        B 3.3-129                                                                                                                                                                                          Revision No. 86 RCIC System Instrumentation B 3.3.5.2
 
B 3.3  INSTRUMENTATION
 
B 3.3.5.2  Reactor Core Isolation Cooling (RCIC) System Instrumentation
 
BASES
 
BACKGROUND                                                                      The purpose of the RCIC System instrumentation is to initiate actions to ensure adequate core cooling when the reactor vessel is isolated from its primary heat sink (the main condenser) and normal coolant makeup flow from the Reactor Feedwater System is insufficient or unavailable, such that RCIC System initiation occurs and maintains sufficient reactor water level such that an initiation of the low pressure Emergency Core Cooling Systems (ECCS) pumps does not occur. A more complete discussion of RCIC System operation is provided in the Bases of LCO 3.5.3, "RCIC System."
 
The RCIC System may be initiated by automatic means.
Automatic initiation occurs for conditions of Reactor Vessel Water Level      Low Low (Level 2). The variable is monitored by four transmitters that are connected to four pressure compensation instruments. The outputs of the pressure compensation instruments are connected to relays whose contacts are arranged in a one-out-of-two taken twice logic arrangement. Once initiated, the RCIC logic seals in and can be reset by the operator only when the reactor vessel water level signals have cleared.
 
The RCIC test line isolation valve is closed on a RCIC initiation signal to allow full system flow and maintain primary containment isolated in the event RCIC is not operating.
 
The RCIC System also monitors the water level in the condensate storage tank (CST) since this is the initial source of water for RCIC operation. Reactor grade water in the CST is the normal source. Upon receipt of a RCIC initiation signal, the CST suction valve is automatically signaled to open (it is normally in the open position) unless the pump suction from the suppression pool valves is open. If the water level in the CST falls below a preselected level, first the suppression pool suction valves automatically open, and then the CST suction valve automatically closes. Two level switches are used to detect low water level in the CST. Either switch can cause the suppression pool suction valves to open. The opening of the
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                        B 3.3-130                                                                                                                                                                                                    Revision No. 0 RCIC System Instrumentation B 3.3.5.2
 
BASES
 
BACKGROUND                                                                      suppression pool suction valves causes the CST suction valve (continued)                                        to close. This prevents losing suction to the pump when automatically transferring suction from the CST to the suppression pool on low CST level.
 
The RCIC System provides makeup water to the reactor until the reactor vessel water level reaches the high water level (Level 8) setting (two-out-of-two logic), at which time the RCIC steam supply valve closes. The RCIC System restarts if vessel level again drops to the low level initiation point (Level 2).
 
APPLICABLE                                                                      The function of the RCIC System is to respond to transient SAFETY ANALYSES,          events by producing makeup coolant to the reactor. The RCIC LCO, and                                                                                  System is not an Engineered Safeguard System and no credit APPLICABILITY                                        is taken in the safety analyses for RCIC System operation.
Based on its contribution to the reduction of overall plant risk, however, the system, and therefore its instrumentation meets Criterion 4 of NRC Policy Statement.
 
The OPERABILITY of the RCIC System instrumentation is dependent upon the OPERABILITY of the individual instrumentation channel Functions specified in Table 3.3.5.2-1. Each Function must have a required number of OPERABLE channels with their setpoints within the specified Allowable Values, where appropriate. A channel is inoperable if its actual trip setting is not within its required Allowable Value. The actual setpoint is calibrated consistent with applicable setpoint methodology assumptions.
 
Allowable Values are specified for each RCIC System instrumentation Function specified in the Table. Trip setpoints are specified in the setpoint calculations. The setpoints are selected to ensure that the settings do not exceed the Allowable Value between CHANNEL CALIBRATIONS.
Operation with a trip setting less conservative than the trip setpoint, but within its Allowable Value, is acceptable. Each Allowable Value specified accounts for instrument uncertainties appropriate to the Function. These uncertainties are described in the setpoint methodology.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                        B 3.3-131                                                                                                                                                                                                    Revision No. 0 RCIC System Instrumentation B 3.3.5.2
 
BASES
 
APPLICABLE                                                                      The individual Functions are required to be OPERABLE in SAFETY ANALYSES,          MODE 1, and in MODES 2 and 3 with reactor steam dome LCO, and                                                                                          pressure > 150 psig since this is when RCIC is required to APPLICABILITY                                        be OPERABLE.  (Refer to LCO 3.5.3 for Applicability Bases (continued)                                        for the RCIC System.)
 
The specific Applicable Safety Analyses, LCO, and Applicability discussions are listed below on a Function by Function basis.
: 1. Reactor Vessel Water LevelLow Low (Level 2)
 
Low reactor pressure vessel (RPV) water level indicates that normal feedwater flow is insufficient to maintain reactor vessel water level and that the capability to cool the fuel may be threatened. Should RPV water level decrease too far, fuel damage could result. Therefore, the RCIC System is initiated at Level 2 to assist in maintaining water level above the top of the active fuel.
 
Reactor Vessel Water Level      Low Low (Level 2) signals are initiated from four level transmitters that sense the difference between the pressure due to a constant column of water (reference leg) and the pressure due to the actual water level (variable leg) in the vessel.
 
The Reactor Vessel Water Level      Low Low (Level 2) Allowable Value is set high enough such that for complete loss of feedwater flow, the RCIC System flow with high pressure coolant injection assumed to fail will be sufficient to avoid initiation of low pressure ECCS at Level 1.
 
Four channels of Reactor Vessel Water Level      Low Low (Level 2) Function are available and are required to be OPERABLE when RCIC is required to be OPERABLE to ensure that no single instrument failure can preclude RCIC initiation.
Refer to LCO 3.5.3 for RCIC Applicability Bases.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                        B 3.3-132                                                                                                                                                                                                    Revision No. 0 RCIC System Instrumentation B 3.3.5.2
 
BASES
 
APPLICABLE                                                                      2. Reactor Vessel Water LevelHigh (Level 8)
SAFETY ANALYSES, LCO, and                                                                                          High RPV water level indicates that sufficient cooling water APPLICABILITY                                        inventory exists in the reactor vessel such that there is no (continued)                                        danger to the fuel. Therefore, the Level 8 signal is used to close the RCIC steam supply valve to prevent overflow into the main steam lines (MSLs).
 
Reactor Vessel Water Level      High (Level 8) signals for RCIC are initiated from four level transmitters, which sense the difference between the pressure due to a constant column of water (reference leg) and the pressure due to the actual water level (variable leg) in the vessel. These four level transmitters are connected to two pressure compensation instruments (channels).
 
The Reactor Vessel Water Level      High (Level 8) Allowable Value is high enough to preclude isolating the injection valve of the RCIC during normal operation, yet low enough to trip the RCIC System prior to water overflowing into the MSLs.
 
Two channels of Reactor Vessel Water Level      High (Level 8)
Function are available and are required to be OPERABLE when RCIC is required to be OPERABLE to ensure that no single instrument failure can preclude RCIC initiation. Refer to LCO 3.5.3 for RCIC Applicability Bases.
: 3. Condensate Storage Tank LevelLow
 
Low level in the CST indicates the unavailability of an adequate supply of makeup water from this normal source.
Normally, the suction valve between the RCIC pump and the CST is open and, upon receiving a RCIC initiation signal, water for RCIC injection would be taken from the CST.
However, if the water level in the CST falls below a preselected level, first the suppression pool suction valves automatically open, and then the CST suction valve automatically closes. This ensures that an adequate supply of makeup water is available to the RCIC pump. To prevent losing suction to the pump, the suction valves are interlocked so that the suppression pool suction valves must be open before the CST suction valve automatically closes.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                        B 3.3-133                                                                                                                                                                                                    Revision No. 0 RCIC System Instrumentation B 3.3.5.2
 
BASES
 
APPLICABLE                                                                      3. Condensate Storage Tank LevelLow  (continued)
SAFETY ANALYSES, LCO, and                                                                                          Two level switches are used to detect low water level in the APPLICABILITY                                        CST. The Condensate Storage Tank Level      Low Function Allowable Value is set high enough to ensure adequate pump suction head while water is being taken from the CST.
 
Two channels of the CST Level      Low Function are available and are required to be OPERABLE when RCIC is required to be OPERABLE to ensure that no single instrument failure can preclude RCIC swap to suppression pool source. Refer to LCO 3.5.3 for RCIC Applicability Bases.
 
ACTIONS                                                                                                    A Note has been provided to modify the ACTIONS related to RCIC System instrumentation channels. 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 inoperable RCIC System instrumentation channels provide appropriate compensatory measures for separate inoperable channels. As such, a Note has been provided that allows separate Condition entry for each inoperable RCIC System instrumentation channel.
 
A.1
 
Required Action A.1 directs entry into the appropriate Condition referenced in Table 3.3.5.2-1. The applicable Condition referenced in the Table is Function dependent.
Each time a channel is discovered to be inoperable, Condition A is entered for that channel and provides for transfer to the appropriate subsequent Condition.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                        B 3.3-134                                                                                                                                                                                                    Revision No. 0 RCIC System Instrumentation B 3.3.5.2
 
BASES
 
ACTIONS                                                                                                    B.1 and B.2 (continued)
Required Action B.1 is intended to ensure that appropriate actions are taken if multiple, inoperable, untripped channels within the same Function result in a complete loss of automatic initiation capability for the RCIC System. In this case, automatic initiation capability is lost if two Function 1 channels in the same trip system are inoperable and untripped. In this situation (loss of automatic initiation capability), the 24 hour allowance of Required Action B.2 is not appropriate, and the RCIC System must be declared inoperable within 1 hour after discovery of loss of RCIC initiation capability.
 
The Completion Time is intended to allow the operator time to evaluate and repair any discovered inoperabilities. This Completion Time also allows for an exception to the normal "time zero" for beginning the allowed outage time "clock."
For Required Action B.1, the Completion Time only begins upon discovery that the RCIC System cannot be automatically initiated due to two or more inoperable, untripped Reactor Vessel Water Level      Low Low (Level 2) channels such that the trip system loses initiation capability. The 1 hour Completion Time from discovery of loss of initiation capability is acceptable because it minimizes risk while allowing time for restoration or tripping of channels.
 
Because of the redundancy of sensors available to provide initiation signals and the fact that the RCIC System is not assumed in any accident or transient analysis, an allowable out of service time of 24 hours has been shown to be acceptable (Ref. 1) to permit restoration of any inoperable channel to OPERABLE status. Alternatively, a Completion Time can be determined in accordance with the Risk Informed Completion Time (RICT) Program. A Note has been provided to indicate that a RICT is only applicable when a loss of function has not occurred. If the inoperable channel cannot be restored to OPERABLE status within the allowable out of service time, the channel must be placed in the tripped condition per Required Action B.2. Placing the inoperable channel in trip would conservatively compensate for the inoperability, restore capability to accommodate a single failure, and allow operation to continue. Alternately, if it is not desired to place the channel in trip (e.g., as in the case where placing the inoperable channel in trip would result in an initiation), Condition E must be entered and its Required Action taken.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                        B 3.3-135                                                                                                                                                                                Revision No. 159 RCIC System Instrumentation B 3.3.5.2
 
BASES
 
ACTIONS C.1 (continued)
A risk based analysis was performed and determined that an allowable out of service time of 24 hours (Ref. 1) is acceptable to permit restoration of any inoperable channel to OPERABLE status (Required Action C.1). Alternatively, a Completion Time can be determined in accordance with the Risk Informed Completion Time (RICT) Program. A Note has been provided to indicate that a RICT is only applicable when a loss of function has not occurred. A Required Action (similar to Required Action B.1) limiting the allowable out of service time, if a loss of automatic RCIC initiation capability exists, is not required. This Condition applies to the Reactor Vessel Water Level      High (Level 8) Function whose logic is arranged such that any inoperable channel will result in a loss of automatic RCIC initiation capability (closure of the RCIC steam supply valve). As stated above, this loss of automatic RCIC initiation capability was analyzed and determined to be acceptable.
The Required Action does not allow placing a channel in trip since this action would not necessarily result in a safe state for the channel in all events.
 
D.1, D.2.1, and D.2.2
 
Required Action D.1 is intended to ensure that appropriate actions are taken if multiple, inoperable, untripped channels within the same Function result in automatic component initiation capability being lost for the feature(s). For Required Action D.1, the RCIC System is the only associated feature. In this case, automatic initiation capability is lost if two Function 3 channels are inoperable and untripped. In this situation (loss of automatic suction swap), the 24 hour allowance of Required Actions D.2.1 and D.2.2 is only appropriate after Action D.1 has been performed. Action D.1 requires that the RCIC System be declared inoperable within 1 hour from discovery of loss of RCIC initiation capability. As noted, Required Action D.1 is only applicable if the RCIC pump suction is not aligned to the suppression pool since, if aligned, the Function is already performed.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                        B 3.3-136                                                                                                                                                                                Revision No. 159 RCIC System Instrumentation B 3.3.5.2
 
BASES
 
ACTIONS                                                                                                    D.1, D.2.1, and D.2.2  (continued)
 
The Completion Time is intended to allow the operator time to evaluate and repair any discovered inoperabilities. This Completion Time also allows for an exception to the normal "time zero" for beginning the allowed outage time "clock."
For Required Action D.1, the Completion Time only begins upon discovery that the RCIC System cannot be automatically aligned to the suppression pool due to two inoperable, untripped channels in the same Function. The 1 hour Completion Time from discovery of loss of initiation capability is acceptable because it minimizes risk while allowing time for restoration or tripping of channels.
 
Because the RCIC System is not assumed in any accident or transient analysis, an allowable out of service time of 24 hours has been shown to be acceptable (Ref. 1) to permit restoration of any inoperable channel to OPERABLE status.
Alternatively, a Completion Time can be determined in accordance with the Risk Informed Completion Time (RICT)
Program. A Note has been provided to indicate that a RICT is only applicable when a loss of function has not occurred. If the inoperable channel cannot be restored to OPERABLE status within the allowable out of service time, the channel must be placed in the tripped condition per Required Action D.2.1, which performs the intended function of the channel. Alternatively, Required Action D.2.2 allows the manual alignment of the RCIC suction to the suppression pool, which also performs the intended function. If Required Action D.2.1 or D.2.2 is performed, measures should be taken to ensure that the RCIC System piping remains filled with water. If it is not desired to perform Required Actions D.2.1 and D.2.2 (e.g., as in the case where shifting the suction source could drain down the RCIC suction piping), Condition E must be entered and its Required Action taken.
 
E.1
 
With any Required Action and associated Completion Time not met, the RCIC System may be incapable of performing the intended function, and the RCIC System must be declared inoperable immediately.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                        B 3.3-137                                                                                                                                                                                Revision No. 159 RCIC System Instrumentation B 3.3.5.2
 
BASES  (continued)
 
SURVEILLANCE                                                  As noted in the beginning of the SRs, the SRs for each RCIC REQUIREMENTS                                                  System instrumentation Function are found in the SRs column of Table 3.3.5.2-1.
 
The Surveillances are modified by a Note to indicate that when a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed as follows:
(a) for up to 6 hours for Function 2 and (b) for up to 6 hours for Functions 1 and 3, provided the associated Function maintains trip capability. Upon completion of the Surveillance, or expiration of the 6 hour allowance, the channel must be returned to OPERABLE status or the applicable Condition entered and Required Actions taken.
This Note is based on the reliability analysis (Ref. 1) assumption of the average time required to perform channel surveillance. That analysis demonstrated that the 6 hour testing allowance does not significantly reduce the probability that the RCIC will initiate when necessary.
 
SR  3.3.5.2.1
 
Performance of the CHANNEL CHECK ensures that a gross failure of instrumentation has not occurred. A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a parameter on other similar channels. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value.
Significant deviations between the instrument channels could be an indication of excessive instrument drift in one of the channels or something even more serious. A CHANNEL CHECK will detect gross channel failure; thus, it is key to verifying the instrumentation continues to operate properly between each CHANNEL CALIBRATION.
 
Agreement criteria are determined by the plant staff based on a combination of the channel instrument uncertainties, including indication and readability. If a channel is outside the criteria, it may be an indication that the instrument has drifted outside its limit.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                        B 3.3-138                                                                                                                                                                                          Revision No. 86 RCIC System Instrumentation B 3.3.5.2
 
BASES
 
SURVEILLANCE                                                  SR  3.3.5.2.1  (continued)
REQUIREMENTS The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. The CHANNEL CHECK supplements less formal, but more frequent, checks of channels during normal operational use of the displays associated with the channels required by the LCO.
 
SR  3.3.5.2.2
 
A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the entire channel will perform the intended function. Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology.
 
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
 
SR  3.3.5.2.3
 
A CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor. This test verifies the channel responds to the measured parameter within the necessary range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drifts between successive calibrations, consistent with the plant specific setpoint methodology.
 
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
 
SR  3.3.5.2.4
 
The LOGIC SYSTEM FUNCTIONAL TEST demonstrates the OPERABILITY of the required initiation logic for a specific channel. The system functional testing performed in LCO 3.5.3 overlaps this Surveillance to provide complete testing of the safety function.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                        B 3.3-139                                                                                                                                                                                          Revision No. 86 RCIC System Instrumentation B 3.3.5.2
 
BASES
 
SURVEILLANCE                                                  SR  3.3.5.2.4  (continued)
REQUIREMENTS The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
 
REFERENCES                                                                      1.                    GENE-770-06-2, "Addendum to Bases for Changes to Surveillance Test Intervals and Allowed Out-of-Service Times for Selected Instrumentation Technical Specifications," February 1991.
 
PBAPS UNIT 2                                                                                                                                                                                                                        B 3.3-140                                                                                                                                                                                          Revision No. 86 Not Used B 3.3.5.3 B 3.3  INSTRUMENTATION
 
B 3.3.5.3  Not Used
 
PBAPS UNIT 2                                                                                                                                                                                                                                                                                                                                          B 3.3-140a                                                                                                                                                                                                                                                                    Amendment No. 145 RPV Water Inventory Control Instrumentation B 3.3.5.4
 
B 3.3  INSTRUMENTATION
 
B 3.3.5.4  Reactor Pressure Vessel (RPV) Water Inventory Control Instrumentation
 
BASES
 
BACKGROUND                                                                      The RPV contains penetrations below the top of the active fuel (TAF) that have the potential to drain the reactor coolant inventory to below the TAF. If the water level should drop below the TAF, the ability to remove decay heat is reduced, which could lead to elevated cladding temperatures and clad perforation. Safety Limit 2.1.1.3 requires the RPV water level to be above the top of the active irradiated fuel at all times to prevent such elevated cladding temperatures.
 
Technical Specifications are required by 10 CFR 50.36 to include limiting safety system settings (LSSS) for variables that have significant safety functions. LSSS are defined by the regulation as "Where a LSSS is specified for a variable on which a safety limit has been placed, the setting must be chosen so that automatic protective actions will correct the abnormal situation before a Safety                                                                                                                                                                                                                                                                                                                                                    Limit (SL) is exceeded."
The Analytical Limit is the limit of the process variable at which a safety action is initiated to ensure that a SL is not exceeded. Any automatic protection action that occurs on reaching the Analytical Limit therefore ensures that the SL is not exceeded. However, in practice, the actual settings for automatic protection channels must be chosen to be more conservative than the Analytical Limit to account for instrument loop uncertainties related to the setting at which the automatic protective action would actually occur. The actual settings for the automatic isolation channels are the same as those established for the same functions in MODES 1, 2, and 3 in LCO 3.3.5.1, "Emergency Core Cooling System (ECCS) Instrumentation," or LCO 3.3.6.1, "Primary Containment Isolation instrumentation".
 
With the unit in MODE 4 or 5, RPV water inventory control is not required to mitigate any events or accidents evaluated in the safety analyses. RPV water inventory control is required in MODES 4 and 5 to protect Safety Limit 2.1.1.3 and the fuel cladding barrier to prevent the release of radioactive material should a draining event occur. Under the definition of DRAIN TIME, some penetration flow paths may be excluded from the DRAIN TIME calculation if they will be isolated by valves that will close automatically without offsite power prior to the RPV water level being equal to the TAF when actuated by RPV water level isolation instrumentation.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                  B 3.3-140b                                                                                                                                                                          Revision No. 145 RPV Water Inventory Control Instrumentation B 3.3.5.4
 
BASES (continued)
 
BACKGROUND                                                                                                                                                                            The purpose of the RPV Water Inventory Control (continued)                                        Instrumentation is to support the requirements of LCO 3.5.4, Reactor Pressure Vessel (RPV) Water Inventory Control, and the definition of DRAIN TIME. There are functions that are required for manual initiation or operation of the ECCS injection/spray subsystem required to be OPERABLE by LCO 3.5.4 and other functions that support automatic isolation of Residual Heat Removal subsystem and Reactor Water Cleanup system penetration flow path(s) on low RPV water level.
 
The RPV Water Inventory Control Instrumentation supports operation of core spray (CS) and low pressure coolant injection (LPCI). The equipment involved with each of these systems is described in the Bases for LCO 3.5.4.
 
APPLICABLE                                                                                                                                                                  With the unit in MODE 4 or 5, RPV water inventory control is SAFETY ANALYSIS                    not required to mitigate any events or accidents evaluated in the safety analyses. RPV water inventory control is required in MODES 4 and 5 to protect Safety Limit 2.1.1.3 and the fuel cladding barrier to prevent the release of radioactive material should a draining event occur.
 
A double-ended guillotine break of the Reactor Coolant System (RCS) is not postulated in MODES 4 and 5 due to the reduced RCS pressure, reduced piping stresses, and ductile piping systems. Instead, an event is postulated in which a single operator error or initiating event allows draining of the RPV water inventory through a single penetration flow path with the highest flow rate, or the sum of the drain rates through multiple penetration flow paths susceptible to a common mode failure (e.g., seismic event (except when the risk is assessed and managed in accordance with TS LCO 3.0.8), loss of normal power, single human error). It is assumed, based on engineering judgment, that while in MODES 4 and 5, one low pressure ECCS injection/spray subsystem can be manually initiated to maintain adequate reactor vessel water level.
 
As discussed in References 1, 2, 3, 4, and 5, operating experience has shown RPV water inventory to be significant to public health and safety. Therefore, RPV Water Inventory Control satisfies Criterion 4 of 10 CFR 50.36(c)(2)(ii).
 
Permissive and interlock setpoints are generally considered as nominal values without regard to measurement accuracy.
 
The specific Applicable Safety Analyses, LCO, and Applicability discussions are listed below on a Function by
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                                                                                                                                          B 3.3-140c                                                                                                                                                                                                                                                                              Revision No. 145 RPV Water Inventory Control Instrumentation B 3.3.5.4
 
BASES (continued)
 
APPLICABLE                                                                                                                                                                            Function basis.
SAFETY ANALYSES (continued)                                                                                                                                                                          Core Spray and Low Pressure Coolant Injection Systems
 
1.a, 2.a. Reactor Pressure - Low (Injection Permissive)
 
Low reactor pressure signals are used as permissives for the low pressure ECCS injection/spray subsystem manual injection functions. This function ensures that, prior to opening the injection valves of the low pressure ECCS subsystems, the reactor pressure has fallen to a value below these subsystems' maximum design pressure. While it is assured during MODES 4 and 5 that the reactor pressure will be below the ECCS maximum design pressure, the Reactor Pressure - Low signals are assumed to be OPERABLE and capable of permitting initiation of the ECCS.
 
The Reactor Pressure - Low signals are initiated from four pressure transmitters that sense the reactor dome pressure.
 
The Allowable Value is low enough to prevent overpressuring the equipment in the low pressure ECCS.
 
The four channels of Reactor Pressure - Low Function are required to be OPERABLE in MODES 4 and 5 when ECCS manual initiation is required to be OPERABLE by LCO 3.5.4.
 
1.b, 2.b. Core Spray and Low Pressure Coolant Injection Pump Discharge Flow - Low (Bypass)
 
The minimum flow instruments are provided to protect the associated low pressure ECCS pump from overheating when the pump is operating and the associated injection valve is not fully open. The minimum flow line valve is opened when low flow is sensed, and the valve is automatically closed when the flow rate is adequate to protect the pump.
 
One differential pressure switch per ECCS pump is used to detect the associated subsystems' flow rates. The logic is arranged such that each transmitter causes its associated minimum flow valve to open. The logic will close the minimum flow valve once the closure setpoint is exceeded.
The LPCI minimum flow valves are time delayed such that the valves will not open for 10 seconds after the switches detect low flow. The time delay is provided to limit reactor vessel inventory loss during the startup of the Residual Heat Removal (RHR) shutdown cooling mode.
 
__________________________________________________(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                  B 3.3-140d                                                                                                                                                                          Revision No. 145 RPV Water Inventory Control Instrumentation B 3.3.5.4
 
BASES (continued)
 
APPLICABLE                                                              The Pump Discharge Flow - Low Allowable Values are high SAFETY ANALYSES                    enough to ensure that the pump flow rate is sufficient to (continued)                                        protect the pump, yet low enough to ensure that the closure of the minimum flow valve is initiated to allow full flow into the core.
 
One channel of the Pump Discharge Flow - Low Function is required to be OPERABLE in MODES 4 and 5 when the associated Core Spray or LPCI pump is required to be OPERABLE by LCO 3.5.4 to ensure the pumps are capable of injecting into the Reactor Pressure Vessel when manually initiated.
 
A note is added to TS Table 3.3.5.4-1 for Function 2.b to clarify the intent of allowing credit for an OPERABLE Low Pressure Coolant Injection subsystem when it is aligned and operating in the decay heat removal mode of RHR. This note is appropriate since the associated RHR pump minimum flow valve (while operating in the decay heat removal mode) is closed and deactivated to prevent inadvertent vessel drain down events.
 
1.c, 2.c. Manual Initiation
 
The Manual Initiation hand switch channels introduce signals into the appropriate ECCS logic to provide manual initiation capability. There is one hand switch for each CS and LPCI pump (four for CS and four for LPCI).
 
RHR System Isolation
 
3.a Reactor Vessel Water Level - Low, Level 3
 
The definition of DRAIN TIME allows crediting the closing of penetration flow paths that are capable of being isolated by valves that will close automatically without offsite power prior to the RPV water level being equal to the TAF when actuated by RPV water level isolation instrumentation.
 
The Reactor Vessel Water Level - Low, Level 3 Function associated with RHR System isolation may be credited for automatic isolation of penetration flow paths associated with the RHR System.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                  B 3.3-140e                                                                                                                                                                          Revision No. 145 RPV Water Inventory Control Instrumentation B 3.3.5.4
 
BASES (continued)
 
APPLICABLE                                                                                                                                                                  Reactor Vessel Water Level - Low, Level 3 signals are Reactor SAFETY ANALYSES            initiated from four level transmitters that sense the (continued)                                        difference between the pressure due to a constant column of water (reference leg) and the pressure due to the actual water level (variable leg) in the vessel. While four channels (two channels per trip system) of the Reactor Vessel Water Level - Low, Level 3 Function are available, only two channels (all in the same trip system) are required to be OPERABLE.
 
The Reactor Vessel Water Level - Low, Level 3 Allowable Value was chosen to be the same as the Primary Containment Isolation Instrumentation Reactor Vessel Water Level - Low, Level 3 Allowable Value (LCO 3.3.6.1), since the capability to cool the fuel may be threatened.
 
The Reactor Vessel Water Level - Low, Level 3 Function is only required to be OPERABLE when automatic isolation of the associated penetration flow path is credited in calculating DRAIN TIME.
 
Reactor Water Cleanup (RWCU) System Isolation
 
4.a Reactor Vessel Water Level - Low, Level 3
 
The definition of DRAIN TIME allows crediting the closing of penetration flow paths that are capable of being isolated by valves that will close automatically without offsite power prior to the RPV water level being equal to the TAF when actuated by RPV water level isolation instrumentation. The Reactor Vessel Water Level - Low, Level 3 Function associated with RWCU System isolation may be credited for automatic isolation of penetration flow paths associated with the RWCU System.
 
Reactor Vessel Water Level - Low, Level 3 signals are initiated from four level transmitters that sense the difference between the pressure due to a constant column of water (reference leg) and the pressure due to the actual water level (variable leg) in the vessel. While four channels (two channels per trip system) of the Reactor Vessel Water Level - Low, Level 3 Function are available, only two channels (all in the same trip system) are required to be OPERABLE.
 
The Reactor Vessel Water Level - Low, Level 3 Allowable Value was chosen to be the same as the RPS Reactor Vessel Water Level - Low, Level 3 Allowable Value (LCO 3.3.1.1), since the capability to cool the fuel may be threatened.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                  B 3.3-140f                                                                                                                                                                          Revision No. 145 RPV Water Inventory Control Instrumentation B 3.3.5.4
 
BASES (continued)
 
APPLICABLE                                                    This Function isolates the inboard and outboard RWCU pump SAFETY ANALYSES                    suction penetration.
(continued)
 
The Reactor Vessel Water Level - Low, Level 3 Function is only required to be OPERABLE when automatic isolation of the associated penetration flow path is credited in calculating DRAIN TIME.
 
ACTIONS                                                                                                    A Note has been provided to modify the ACTIONS related to RPV Water Inventory Control instrumentation channels. 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 continue to apply for each additional failure, with Completion Times based on initial entry into the Condition. However, the Required Actions for inoperable RPV Water Inventory Control instrumentation channels provide appropriate compensatory measures for separate inoperable Condition entry for each inoperable RPV Water Inventory Control instrumentation channel.
 
A.1
 
Required Action A.1 directs entry into the appropriate Condition referenced in Table 3.3.5.4-1. The applicable Condition referenced in the Table is Function dependent.
Each time a channel is discovered inoperable, Condition A is entered for that channel and provides for transfer to the appropriate subsequent Condition.
 
B.1 and B.2
 
RHR System Isolation, Reactor Vessel Water Level - Low, Level 3, and Reactor Water Cleanup System Isolation, Reactor Vessel Water Level - Low, Level 3 functions are applicable when automatic isolation of the associated penetration flow path is credited in calculating DRAIN TIME. If the instrumentation is inoperable, Required Action B.1 directs an immediate declaration that the associated penetration flow path(s) are incapable of automatic isolation. Required Action B.2 directs calculation of DRAIN TIME. The calculation cannot credit automatic isolation of the affected penetration flow paths.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                  B 3.3-140g                                                                                                                                                                          Revision No. 145 RPV Water Inventory Control Instrumentation B 3.3.5.4
 
BASES (continued)
 
ACTIONS                                                                                  C.1 (continued)
Low reactor steam dome pressure signals are used as permissives for the low pressure ECCS injection/spray subsystem manual injection functions. If the permissive is inoperable, manual initiation of ECCS is prohibited.
Therefore, the permissive must be placed in the trip condition within 1 hour. With the permissive in the trip condition, manual initiation may be performed. Prior to placing the permissive in the tripped condition, the operator can take manual control of the pump and the injection valve to inject water into the RPV.
 
The Completion Time of 1 hour is intended to allow the operator time to evaluate any discovered inoperabilities and to place the channel in trip.
 
D.1
 
If a Core Spray or Low Pressure Coolant Injection Pump Discharge Flow - Low bypass function is inoperable, there is a risk that the associated low pressure ECCS pump could overheat when the pump is operating and the associated injection valve is not fully open. In this condition, the operator can take manual control of the pump and the injection valve to ensure the pump does not overheat. If a manual initiation function is inoperable, the ECCS subsystem pumps can be started manually and the valves can be opened manually, but this is not the preferred condition.
 
The 24 hour Completion Time was chosen to allow time for the operator to evaluate and repair any discovered inoperabilities. The Completion Time is appropriate given the ability to manually start the ECCS pumps and open the injection valves and to manually ensure the pump does not overheat.
 
E.1
 
With the Required Action and associated Completion Time of Condition C or D not met, the associated low pressure ECCS injection/spray subsystem may be incapable of performing the intended function, and must be declared inoperable immediately.
 
As noted in the beginning of the SRs, the SRs for each RPV Water Inventory Control instrument Function are found in the SRs column of Table 3.3.5.4-1.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                  B 3.3-140h                                                                                                                                                                          Revision No. 145 RPV Water Inventory Control Instrumentation B 3.3.5.4
 
BASES (continued)
 
SURVEILLANCE SR 3.3.5.4.1 REQUIREMENTS Performance of the CHANNEL CHECK ensures that a gross failure of instrumentation has not occurred. A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a similar parameter on other channels. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value.
Significant deviations between the instrument channels could be an indication of excessive instrument drift in one of the channels or something even more serious. A CHANNEL CHECK guarantees that undetected outright channel failure is limited; thus, it is key to verifying the instrumentation continues to operate properly between each CHANNEL FUNCTIONAL TEST.
 
Agreement criteria are determined by the plant staff, based on a combination of the channel instrument uncertainties, including indication and readability. If a channel is outside the criteria, it may be an indication that the instrument has drifted outside its limit.
 
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
 
The CHANNEL CHECK supplements less formal, but more frequent, checks of channels during normal operational use of the displays associated with the channels required by the LCO.
 
SR 3.3.5.4.2
 
A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the entire channel will perform the intended function. A successful test of the required contact(s) of a channel relay may be performed by the verification of the change of state of a single contact of the relay. This clarifies what is an acceptable CHANNEL FUNCTIONAL TEST of a relay. This is acceptable because all of the other required contacts of the relay are verified by other Technical Specifications and non-Technical Specifications tests.
 
Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology.
 
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                  B 3.3-140i                                                                                                                                                                          Revision No. 145 RPV Water Inventory Control Instrumentation B 3.3.5.4
 
BASES (continued)
 
SURVEILLANCE SR 3.3.5.4.3 REQUIREMENTS (continued)                                        The LOGIC SYSTEM FUNCTIONAL TEST demonstrates the OPERABILITY of the required initiation logic for a specific channel. The system functional testing performed in LCO 3.5.4 overlaps this Surveillance to complete testing of the assumed safety function.
 
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
 
REFERENCES        1. Information Notice 84-81 "Inadvertent Reduction in Primary Coolant Inventory in Boiling Water Reactors During Shutdown and Startup," November 1984.
: 2. Information Notice 86-74, "Reduction of Reactor Coolant Inventory Because of Misalignment of RHR Valves," August 1986.
: 3. Generic Letter 92-04, "Resolution of the Issues Related to Reactor Vessel Water Level Instrumentation in BWRs Pursuant to 10 CFR 50.54(F), " August 1992.
: 4. NRC Bulletin 93-03, "Resolution of Issues Related to Reactor Vessel Water Level Instrumentation in BWRs," May 1993.
: 5. Information Notice 94-52, "Inadvertent Containment Spray and Reactor Vessel Draindown at Millstone 1," July 1994.
 
PBAPS UNIT 2                                                                                                                                                                                                                  B 3.3-140j                                                                                                                                                                                Revision No. 145 Primary Containment Isolation Instrumentation B 3.3.6.1
 
B 3.3  INSTRUMENTATION
 
B 3.3.6.1  Primary Containment Isolation Instrumentation
 
BASES
 
BACKGROUND                                                                      The primary containment isolation instrumentation automatically initiates closure of appropriate primary containment isolation valves (PCIVs). 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). 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 isolation instrumentation includes the sensors, relays, and switches that are necessary to cause initiation of primary containment and reactor coolant pressure boundary (RCPB) isolation. Most channels include electronic equipment (e.g., trip units) that compares measured input signals with pre-established setpoints. When the setpoint is exceeded, the channel output relay actuates, which then outputs a primary containment isolation signal to the isolation logic. Functional diversity is provided by monitoring a wide range of independent parameters. The input parameters to the isolation logics are (a) reactor vessel water level, (b) reactor pressure, (c) main steam line (MSL) flow measurement, (d) (deleted), (e) main steam line pressure, (f) drywell pressure, (g) high pressure coolant injection (HPCI) and reactor core isolation cooling (RCIC) steam line flow, (h) HPCI and RCIC steam line pressure, (i) reactor water cleanup (RWCU) flow, (j) Standby Liquid Control (SLC) System initiation, (k) area ambient temperatures, (l) reactor building ventilation and refueling floor ventilation exhaust radiation, and (m) main stack radiation. Redundant sensor input signals from each parameter are provided for initiation of isolation.
 
Primary containment isolation instrumentation has inputs to the trip logic of the isolation functions listed below.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                        B 3.3-141                                                                                                                                                                                Revision No. 134 Primary Containment Isolation Instrumentation B 3.3.6.1
 
BASES
 
BACKGROUND                                                                      1. Main Steam Line Isolation (continued)
Most MSL Isolation Functions receive inputs from four channels. The outputs from these channels are combined in a one-out-of-two taken twice logic to initiate isolation of the Group I isolation valves (MSIVs and MSL drains, MSL sample lines, and recirculation loop sample line valves).
To initiate a Group I isolation, both trip systems must be tripped.
 
The exceptions to this arrangement are the Main Steam Line Flow      High Function and Turbine Building Main Steam Tunnel Temperature      High Functions. The Main Steam Line Flow      High Function uses 16 flow channels, four for each steam line.
One channel from each steam line inputs to one of the four trip strings. Two trip strings make up each trip system and both trip systems must trip to cause an MSL isolation. Each trip string has four inputs (one per MSL), any one of which will trip the trip string. The trip systems are arranged in a one-out-of-two taken twice logic. This is effectively a one-out-of-eight taken twice logic arrangement to initiate a Group I isolation. The Turbine Building Main Steam Tunnel Temperature-High Function receives inputs from twelve channels, four channels at each of the three different locations along the steam line. High temperature on any channel is not related to a specific MSL. The channels are arranged in a one-out-of-two taken twice logic for each location.
: 2. Primary Containment Isolation
 
Most Primary Containment Isolation Functions receive inputs from four channels. The outputs from these channels are arranged in a one-out-of-two taken twice logic. Isolation of inboard and outboard primary containment isolation valves occurs when both trip systems are in trip.
 
The exception to this arrangement is the Main Stack Monitor Radiation      High Function. This Function has two channels, whose outputs are arranged in two trip systems which use a one-out-of-one logic. Each trip system isolates one valve per associated penetration. The Main Stack Monitor Radiation      High Function will isolate vent and purge valves greater than two inches in diameter during containment purging (Ref. 2).
 
The valves isolated by each of the Primary Containment Isolation Functions are listed in Reference 1.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                        B 3.3-142                                                                                                                                                                                          Revision No. 48 Primary Containment Isolation Instrumentation B 3.3.6.1
 
BASES
 
BACKGROUND                                                                      3., 4. High Pressure Coolant Injection System Isolation and (continued)                                        Reactor Core Isolation Cooling System Isolation
 
The Steam Line Flow      High Functions that isolate HPCI and RCIC receive input from two channels, with each channel comprising one trip system using a one-out-of-one logic.
Each of the two trip systems in each isolation group (HPCI and RCIC) is connected to the two valves on each associated penetration. Each HPCI and RCIC Steam Line Flow      High channel has a time delay relay to prevent isolation due to flow transients during startup.
 
The HPCI and RCIC Isolation Functions for Drywell Pressure      High and Steam Supply Line Pressure      Low receive inputs from four channels. The outputs from these channels are combined in a one-out-of-two taken twice logic to initiate isolation of the associated valves.
 
The HPCI and RCIC Compartment and Steam Line Area Temperature      High Functions receive input from 16 channels, four channels at each of four different locations. The channels are arranged in a one-out-of-two taken twice logic for each location.
 
The HPCI and RCIC Steam Line Flow      High Functions, Steam Supply Line Pressure      Low Functions, and Compartment and Steam Line Area Temperature      High Functions isolate the associated steam supply and turbine exhaust valves and pump suction valves. The HPCI and RCIC Drywell Pressure      High Functions isolate the HPCI and RCIC test return line valves.
The HPCI and RCIC Drywell Pressure      High Functions, in conjunction with the Steam Supply Line Pressure      Low Functions, isolate the HPCI and RCIC turbine exhaust vacuum relief valves.
: 5. Reactor Water Cleanup System Isolation
 
The Reactor Vessel Water Level      Low (Level 3) Isolation Function receives input from four reactor vessel water level channels. The outputs from the reactor vessel water level channels are connected into a one-out-of-two taken twice logic which isolates both the inboard and outboard isolation valves. The RWCU Flow      High Function receives input from two channels, with each channel in one trip system using a one-out-of-one logic, with one channel tripping the inboard valve and one channel tripping the outboard valves. The SLC
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                        B 3.3-143                                                                                                                                                                                          Revision No. 48 Primary Containment Isolation Instrumentation B 3.3.6.1
 
BASES
 
BACKGROUND                                                                      5. Reactor Water Cleanup System Isolation  (continued)
 
System Isolation Function receives input from two channels with each channel in one trip system using a one-out-of-one logic. When either SLC pump is started remotely, one channel trips the inboard isolation valve and one channel isolates the outboard isolation valves.
 
The RWCU Isolation Function isolates the inboard and outboard RWCU pump suction penetration and the outboard valve at the RWCU connection to reactor feedwater.
: 6. Shutdown Cooling System Isolation
 
The Reactor Vessel Water Level      Low (Level 3) Function receives input from four reactor vessel water level channels. The outputs from the channels are connected to a one-out-of-two taken twice logic, which isolates both valves on the RHR shutdown cooling pump suction penetration. The Reactor Pressure      High Function receives input from two channels, with each channel in one trip system using a one-out-of-one logic. Each trip system is connected to both valves on the RHR shutdown cooling pump suction penetration.
: 7. Feedwater Recirculation Isolation
 
The Reactor Pressure      High Function receives inputs from four channels. The outputs from the four channels are connected into a one-out-of-two taken twice logic which isolates the feedwater recirculation valves.
: 8.                Traversing Incore Probe System Isolation
 
The Reactor Vessel Water Level-Low, Level 3 Isolation Function receives input from two reactor vessel water level channels. The outputs from the reactor vessel water level channels are connected into one two-out-of-two logic trip system. The Drywell Pressure-High Isolation function receives input from two drywell pressure channels. The outputs from the drywell pressure channels are connected into one two-out-of-two logic trip system.
 
When either Isolation Function actuates, the TIP drive mechanisms will withdraw the TIPs, if inserted, and close the TIP system isolation ball valves when the TIPs are fully withdrawn. The redundant TIP system isolation valves are manual shear valves.
 
TIP System Isolation Functions isolate the Group II(D) TIP valves (isolation ball valves).
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                        B 3.3-144                                                                                                                                                                                          Revision No. 57 Primary Containment Isolation Instrumentation B 3.3.6.1
 
BASES
 
APPLICABLE                                                                      The isolation signals generated by the primary containment SAFETY ANALYSES,          isolation instrumentation are implicitly assumed in the LCO, and                                                                                          safety analyses of References 1 and 3 to initiate closure APPLICABILITY                                of valves to limit offsite doses. Refer to LCO 3.6.1.3, "Primary Containment Isolation Valves (PCIVs)," Applicable Safety Analyses Bases for more detail of the safety analyses.
Primary containment isolation instrumentation satisfies Criterion 3 of the NRC Policy Statement. Certain instrumentation Functions are retained for other reasons and are described below in the individual Functions discussion.
The OPERABILITY of the primary containment instrumentation is dependent on the OPERABILITY of the individual instrumentation channel Functions specified in Table 3.3.6.1-1. Each Function must have a required number of OPERABLE channels, with their setpoints within the specified Allowable Values, where appropriate. A channel is inoperable if its actual trip setting is not within its required Allowable Value. The actual setpoint is calibrated consistent with applicable setpoint methodology assumptions.
Allowable Values, where applicable, are specified for each Primary Containment Isolation Function specified in the Table. Trip setpoints are specified in the setpoint calculations. The trip setpoints are selected to ensure that the setpoints do not exceed the Allowable Value between CHANNEL CALIBRATIONS. Operation with a trip setting less conservative than the trip setpoint, but within its Allowable Value, is acceptable. Trip setpoints are those predetermined values of output at which an action should take place. The setpoints are compared to the actual process parameter (e.g., reactor vessel water level), and when the measured output value of the process parameter exceeds the setpoint, the associated device (e.g., trip unit) changes state. The analytic or design limits are derived from the limiting values of the process parameters obtained from the safety analysis or other appropriate documents. The Allowable Values are derived from the analytic or design limits, corrected for calibration, process, and instrument errors. The trip setpoints are determined from analytical or design limits, corrected for calibration, process, and instrument errors, as well as, instrument drift. In selected cases, the Allowable Values and trip setpoints are determined by engineering judgement or historically accepted practice relative to the intended function of the channel. The trip setpoints determined in this manner provide adequate protection by assuring instrument and process uncertainties expected for the environments during the operating time of the associated channels are accounted for.
Certain Emergency Core Cooling Systems (ECCS) and RCIC valves (e.g., minimum flow) also serve the dual function of automatic PCIVs. The signals that isolate these valves are also associated with the automatic initiation of the ECCS (continued)
PBAPS UNIT 2                                                                                                                                                                                                                        B 3.3-145                                                                                                                                                                                          Revision No. 57 Primary Containment Isolation Instrumentation B 3.3.6.1
 
BASES
 
APPLICABLE                                                                      and RCIC. The instrumentation requirements and ACTIONS SAFETY ANALYSES,          associated with these signals are addressed in LCO 3.3.5.1, LCO, and                                                                                          "Emergency Core Cooling Systems (ECCS) Instrumentation," and APPLICABILITY                                        LCO 3.3.5.2, "Reactor Core Isolation Cooling (RCIC) System (continued)                                        Instrumentation," and are not included in this LCO.
 
In general, the individual Functions are required to be OPERABLE in MODES 1, 2, and 3 consistent with the Applicability for LCO 3.6.1.1, "Primary Containment."
Functions that have different Applicabilities are discussed below in the individual Functions discussion.
 
The specific Applicable Safety Analyses, LCO, and Applicability discussions are listed below on a Function by Function basis.
 
Main Steam Line Isolation
 
1.a. Reactor Vessel Water LevelLow Low Low (Level 1)
 
Low reactor pressure vessel (RPV) water level indicates that the capability to cool the fuel may be threatened. Should RPV water level decrease too far, fuel damage could result.
Therefore, isolation of the MSIVs and other interfaces with the reactor vessel occurs to prevent offsite dose limits from being exceeded. The Reactor Vessel Water Level      Low Low Low (Level 1) Function is one of the many Functions assumed to be OPERABLE and capable of providing isolation signals.
The Reactor Vessel Water Level      Low Low Low (Level 1)
Function associated with isolation is assumed in the analysis of the recirculation line break (Ref. 1). The isolation of the MSLs on Level 1 supports actions to ensure that offsite dose limits are not exceeded for a DBA.
 
Reactor vessel water level signals are initiated from four level transmitters that sense the difference between the pressure due to a constant column of water (reference leg) and the pressure due to the actual water level (variable leg) in the vessel. Four channels of Reactor Vessel Water Level      Low Low Low (Level 1) Function are available and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                        B 3.3-146                                                                                                                                                                                                    Revision No. 0 Primary Containment Isolation Instrumentation B 3.3.6.1
 
BASES
 
APPLICABLE                                                                      1.a. Reactor Vessel Water LevelLow Low Low (Level 1)
SAFETY ANALYSES,          (continued)
LCO, and APPLICABILITY                                        The Reactor Vessel Water Level      Low Low Low (Level 1)
Allowable Value is chosen to be the same as the ECCS Level 1 Allowable Value (LCO 3.3.5.1) to ensure that the MSLs isolate on a potential loss of coolant accident (LOCA) to prevent offsite doses from exceeding 10 CFR 50.67 limits.
 
This Function isolates MSIVs, MSL drains, MSL sample lines and recirculation loop sample line valves.
 
1.b. Main Steam Line PressureLow
 
Low MSL pressure indicates that there may be a problem with the turbine pressure regulation, which could result in a low reactor vessel water level condition and the RPV cooling down more than 100&deg;F/hr if the pressure loss is allowed to continue. The Main Steam Line Pressure      Low Function is directly assumed in the analysis of the pressure regulator failure (Ref. 3). For this event, the closure of the MSIVs ensures that the RPV temperature change limit (100&deg;F/hr) is not reached. In addition, this Function supports actions to ensure that Safety Limit 2.1.1.1 is not exceeded.  (This Function closes the MSIVs during the depressurization transient in order to maintain reactor steam dome pressue
                                  > 700 psia. The MSIV closure results in a scram, thus reducing reactor power to < 22.6% RTP.)
 
The MSL low pressure signals are initiated from four transmitters that are connected to the MSL header. The transmitters are arranged such that, even though physically separated from each other, each transmitter is able to detect low MSL pressure. Four channels of Main Steam Line Pressure      Low Function are available and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function.
 
The Allowable Value was selected to be high enough to prevent excessive RPV depressurization.
 
The Main Steam Line Pressure      Low Function is only required to be OPERABLE in MODE 1 since this is when the assumed transient can occur (Ref. 1).
 
This Function isolates MSIVs, MSL drains, MSL sample lines and recirculation loop sample line valves.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                        B 3.3-147                                                                                                                                                                                Revision No. 143 Primary Containment Isolation Instrumentation B 3.3.6.1
 
BASES
 
APPLICABLE                                                                      1.c. Main Steam Line FlowHigh SAFETY ANALYSES, LCO, and                                                                                          Main Steam Line Flow      High is provided to detect a break of APPLICABILITY                                        the MSL and to initiate closure of the MSIVs. If the steam (continued)                                        were allowed to continue flowing out of the break, the reactor would depressurize and the core could uncover. If the RPV water level decreases too far, fuel damage could occur. Therefore, the isolation is initiated on high flow to prevent or minimize core damage. The Main Steam Line Flow      High Function is directly assumed in the analysis of the main steam line break (MSLB) (Ref. 3). The isolation action, along with the scram function of the Reactor Protection System (RPS), ensures that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46 and offsite doses do not exceed the 10 CFR 50.67 limits.
 
The MSL flow signals are initiated from 16 transmitters that are connected to the four MSLs. The transmitters are arranged such that, even though physically separated from each other, all four connected to one MSL would be able to detect the high flow. Four channels of Main Steam Line Flow      High Function for each MSL (two channels per trip system) are available and are required to be OPERABLE so that no single instrument failure will preclude detecting a break in any individual MSL.
 
The Allowable Value is chosen to ensure that offsite dose limits are not exceeded due to the break.
 
This Function isolates MSIVs, MSL drains, MSL sample lines and recirculation loop sample line valves.
 
1.d. Deleted
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                        B 3.3-148                                                                                                                                                                                Revision No. 134 Primary Containment Isolation Instrumentation B 3.3.6.1
 
BASES
 
APPLICABLE                                                                      1.e  Turbine Building Main Steam Tunnel Temperature-High SAFETY ANALYSES, LCO, and                                                                                          The Turbine Building Main Steam Tunnel Temperature Function APPLICABILITY                                        is provided to detect a break in a main steam line and provides diversity to the high flow instrumentation.
 
Turbine Building Main Steam Tunnel Temperature signals are initiated from resistance temperature detectors (RTDs) located along the main steam line between the Reactor Building and the turbine. Twelve channels of Turbine Building Main Steam Tunnel Temperature-High Function are available and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function.
 
The Allowable Value is chosen to detect a leak equivalent to between 1% and 10% rated steam flow.
 
This Function isolates MSIVs, MSL drains, MSL sample lines and recirculation loop sample line valves.
 
1.f. Reactor Building Main Steam Tunnel Temperature-High
 
The Reactor Building Main Steam Tunnel Temperature Function is provided to detect a break in a main steam line and provides diversity to the high flow instrumentation.
 
Reactor Building Main Steam Tunnel Temperature signals are initiated from resistance temperature detectors (RTDs) located in the Main Steam Line Tunnel ventilation exhaust duct. Four channels of Reactor Building Main Steam Tunnel Temperature      High Function are available and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                        B 3.3-149                                                                                                                                                                                Revision No. 134 Primary Containment Isolation Instrumentation B 3.3.6.1
 
BASES
 
APPLICABLE                                                                      1.f  Reactor Building Main Steam Tunnel Temperature-High SAFETY ANALYSES,            (continued)
LCO, and APPLICABILITY                                        The Allowable Value is chosen to detect a leak equivalent to between 1% and 10% rated steam flow.
 
This Function isolates MSIVs, MSL drains, MSL sample lines and recirculation loop sample line valves.
 
Primary Containment Isolation
 
2.a. Reactor Vessel Water LevelLow (Level 3)
 
Low RPV water level indicates that the capability to cool the fuel may be threatened. The valves whose penetrations communicate with the primary containment are isolated to limit the release of fission products. The isolation of the primary containment on Level 3 supports actions to ensure that offsite dose limits of 10 CFR 50.67 are not exceeded.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                  B 3.3-149a                                                                                                                                                                                    Revision No. 75 Primary Containment Isolation Instrumentation B 3.3.6.1
 
BASES
 
APPLICABLE                                                                      2.a. Reactor Vessel Water LevelLow (Level 3)  (continued)
SAFETY ANALYSES, LCO, and                                                                                          The Reactor Vessel Water Level      Low (Level 3) Function APPLICABILITY                                        associated with isolation is implicitly assumed in the UFSAR analysis as these leakage paths are assumed to be isolated post LOCA.
 
Reactor Vessel Water Level      Low (Level 3) signals are initiated from level transmitters that sense the difference between the pressure due to a constant column of water (reference leg) and the pressure due to the actual water level (variable leg) in the vessel. Four channels of Reactor Vessel Water Level      Low (Level 3) Function are available and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function.
 
The Reactor Vessel Water Level      Low (Level 3) Allowable Value was chosen to be the same as the RPS Level 3 scram Allowable Value (LCO 3.3.1.1), since isolation of these valves is not critical to orderly plant shutdown.
 
This Function isolates the Group II(A) valves listed in Reference 1 with the exception of RWCU isolation valves and RHR shutdown cooling pump suction valves which are addressed in Functions 5.c and 6.b, respectively.
 
2.b. Drywell PressureHigh
 
High drywell pressure can indicate a break in the RCPB inside the primary containment. The isolation of some of the primary containment isolation valves on high drywell pressure supports actions to ensure that offsite dose limits of 10 CFR 50.67 are not exceeded. The Drywell Pressure      High Function, associated with isolation of the primary containment, is implicitly assumed in the UFSAR accident analysis as these leakage paths are assumed to be isolated post LOCA.
 
High drywell pressure signals are initiated from pressure transmitters that sense the pressure in the drywell. Four channels of Drywell Pressure      High are available and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                        B 3.3-150                                                                                                                                                                                          Revision No. 75 Primary Containment Isolation Instrumentation B 3.3.6.1
 
BASES
 
APPLICABLE                                                                      2.b. Drywell PressureHigh  (continued)
SAFETY ANALYSES, LCO, and                                                                                          The Allowable Value was selected to be the same as the ECCS APPLICABILITY                                        Drywell Pressure      High Allowable Value (LCO 3.3.5.1), since this may be indicative of a LOCA inside primary containment.
 
This Function isolates the Group II(B) valves listed in Reference 1.
 
2.c. Main Stack Monitor RadiationHigh
 
Main stack monitor radiation is an indication that the release of radioactive material may exceed established limits. Therefore, when Main Stack Monitor Radiation      High is detected when there is flow through the Standby Gas Treatment System, an isolation of primary containment purge supply and exhaust penetrations is initiated to limit the release of fission products. However, this Function is not assumed in any accident or transient analysis in the UFSAR because other leakage paths (e.g., MSIVs) are more limiting.
 
The drywell radiation signals are initiated from radiation detectors that isokinetically sample the main stack utilizing sample pumps. Two channels of Main Stack Radiation      High Function are available and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function.
 
The Allowable Value is set below the maximum allowable release limit in accordance with the Offsite Dose Calculation Manual (ODCM).
 
This Function isolates the containment vent and purge valves and other Group III(E) valves listed in Reference 1.
 
2.d., 2.e. Reactor Building Ventilation and Refueling Floor Ventilation Exhaust RadiationHigh
 
High secondary containment exhaust radiation is an indication of possible gross failure of the fuel cladding.
The release may have originated from the primary containment due to a break in the RCPB. When Reactor Building or Refueling Floor Ventilation Exhaust Radiation      High is detected, the affected ventilation pathway and primary
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                        B 3.3-151                                                                                                                                                                                          Revision No. 20 Primary Containment Isolation Instrumentation B 3.3.6.1
 
BASES
 
APPLICABLE                                                                      2.d., 2.e. Reactor Building Ventilation and Refueling Floor SAFETY ANALYSES,          Ventilation Exhaust RadiationHigh  (continued)
LCO, and APPLICABILITY                                        containment purge supply and exhaust valves are isolated to limit the release of fission products. Additionally, Ventilation Exhaust Radiation      High Function initiates Standby Gas Treatment System.
 
The Ventilation Exhaust Radiation      High signals are initiated from radiation detectors that are located on the ventilation exhaust piping coming from the reactor building and the refueling floor zones, respectively. The signal from each detector is input to an individual monitor whose trip outputs are assigned to an isolation channel. Four channels of Reactor Building Ventilation Exhaust      High Function and four channels of Refueling Floor Ventilation Exhaust      High Function are available and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function.
 
The Allowable Values are chosen to promptly detect gross failure of the fuel cladding during a refueling accident.
 
These Functions isolate the Group III(C) and III(D) valves listed in Reference 1.
 
High Pressure Coolant Injection and Reactor Core Isolation Cooling Systems Isolation
 
3.a., 3.b., 4.a., 4.b. HPCI and RCIC Steam Line FlowHigh and Time Delay Relays
 
Steam Line Flow      High Functions are provided to detect a break of the RCIC or HPCI steam lines and initiate closure of the steam line isolation valves of the appropriate system. If the steam is allowed to continue flowing out of the break, the reactor will depressurize and the core can uncover. Therefore, the isolations are initiated on high flow to prevent or minimize core damage. The isolation action, along with the scram function of the RPS, ensures that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46. Specific credit for these Functions is not assumed in any UFSAR accident analyses since the
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                        B 3.3-152                                                                                                                                                                                                    Revision No. 0 Primary Containment Isolation Instrumentation B 3.3.6.1
 
BASES
 
APPLICABLE                                                                      3.a., 3.b., 4.a., 4.b. HPCI and RCIC Steam Line Flow-High SAFETY ANALYSES,          and Time Delay Relays  (continued)
LCO, and APPLICABILITY                                        bounding analysis is performed for large breaks such as recirculation and MSL breaks. However, these instruments prevent the RCIC or HPCI steam line breaks from becoming bounding.
 
The HPCI and RCIC Steam Line Flow      High signals are initiated from transmitters (two for HPCI and two for RCIC) that are connected to the system steam lines. A time delay is provided to prevent isolation due to high flow transients during startup with one Time Delay Relay channel associated with each Steam Line Flow      High channel. Two channels of both HPCI and RCIC Steam Line Flow      High Functions and the associated Time Delay Relays are available and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function.
 
The Allowable Values for Steam Line Flow      High Function and associated Time Delay Relay Function are chosen to be low enough to ensure that the trip occurs to maintain the MSLB event as the bounding event.
 
These Functions isolate the associated HPCI and RCIC steam supply and turbine exhaust valves and pump suction valves.
 
3.c., 4.c. HPCI and RCIC Steam Supply Line PressureLow
 
Low MSL pressure indicates that the pressure of the steam in the HPCI or RCIC turbine may be too low to continue operation of the associated system's turbine. These isolations prevent radioactive gases and steam from escaping through the pump shaft seals into the reactor building but are primarily for equipment protection and are also assumed for long term containment isolation. However, they also provide a diverse signal to indicate a possible system break. These instruments are included in Technical Specifications (TS) because of the potential for risk due to possible failure of the instruments preventing HPCI and RCIC initiations (Ref. 4).
 
The HPCI and RCIC Steam Supply Line Pressure      Low signals are initiated from transmitters (four for HPCI and four for RCIC) that are connected to the system steam line. Four
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                        B 3.3-153                                                                                                                                                                                                    Revision No. 0 Primary Containment Isolation Instrumentation B 3.3.6.1
 
BASES
 
APPLICABLE                                                                      3.c., 4.c. HPCI and RCIC Steam Supply Line PressureLow SAFETY ANALYSES,          (continued)
LCO, and APPLICABILITY                                        channels of both HPCI and RCIC Steam Supply Line Pressure      Low Functions are available and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function.
 
The Allowable Values are selected to be high enough to prevent damage to the system's turbine.
 
These Functions isolate the associated HPCI and RCIC steam supply and turbine exhaust valves and pump suction valves.
 
3.d., 4.d. Drywell PressureHigh (Vacuum Breakers)
 
High drywell pressure can indicate a break in the RCPB. The HPCI and RCIC isolation of the turbine exhaust vacuum breakers is provided to prevent communication with the drywell when high drywell pressure exists. The HPCI and RCIC turbine exhaust vacuum breaker isolation occurs following a permissive from the associated Steam Supply Line Pressure      Low Function which indicates that the system is no longer required or capable of performing coolant injection.
The isolation of the HPCI and RCIC turbine exhaust vacuum breakers by Drywell Pressure      High is indirectly assumed in the UFSAR accident analysis because the turbine exhaust leakage path is not assumed to contribute to offsite doses.
 
High drywell pressure signals are initiated from pressure transmitters that sense the pressure in the drywell. Four channels for both HPCI and RCIC Drywell Pressure      High (Vacuum Breakers) Functions are available and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function.
 
The Allowable Value was selected to be the same as the ECCS Drywell Pressure      High Allowable Value (LCO 3.3.5.1), since this is indicative of a LOCA inside primary containment.
 
This Function isolates the associated HPCI and RCIC vacuum relief valves and test return line valves.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                        B 3.3-154                                                                                                                                                                                                    Revision No. 0 Primary Containment Isolation Instrumentation B 3.3.6.1
 
BASES
 
APPLICABLE                                                                      3.e., 4.e. HPCI and RCIC Compartment and Steam Line Area SAFETY ANALYSES,          TemperatureHigh LCO, and APPLICABILITY                                        HPCI and RCIC Compartment and Steam Line Area temperatures (continued)                                        are provided to detect a leak from the associated system steam piping. The isolation occurs when a very small leak has occurred and is diverse to the high flow instrumentation. If the small leak is allowed to continue without isolation, offsite dose limits may be reached.
 
These Functions are not assumed in any UFSAR transient or accident analysis, since bounding analyses are performed for large breaks such as recirculation or MSL breaks.
 
HPCI and RCIC Compartment and Steam Line Area Temperature      High signals are initiated from resistance temperature detectors (RTDs) that are appropriately located to protect the system that is being monitored. The HPCI and RCIC Compartment and Steam Line Area Temperature      High Functions each use 16 temperature channels. Sixteen channels for each HPCI and RCIC Compartment and Steam Line Area Temperature      High Function are available and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function.
 
The Allowable Values are set low enough to detect a leak.
 
These Functions isolate the associated HPCI and RCIC steam supply and turbine exhaust valves and pump suction valves.
 
Reactor Water Cleanup (RWCU) System Isolation
 
5.a. RWCU FlowHigh
 
The high flow signal is provided to detect a break in the RWCU System. Should the reactor coolant continue to flow out of the break, offsite dose limits may be exceeded.
Therefore, isolation of the RWCU System is initiated when high RWCU flow is sensed to prevent exceeding offsite doses.
This Function is not assumed in any UFSAR transient or accident analysis, since bounding analyses are performed for large breaks such as MSLBs.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                        B 3.3-155                                                                                                                                                                                          Revision No. 32 Primary Containment Isolation Instrumentation B 3.3.6.1
 
BASES
 
APPLICABLE                                                                      5.a. RWCU FlowHigh  (continued)
SAFETY ANALYSES, LCO, and                                                                                          The high RWCU flow signals are initiated from transmitters APPLICABILITY                                        that are connected to the pump suction line of the RWCU System. Two channels of RWCU Flow      High Function are available and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function.
 
The RWCU Flow      High Allowable Value ensures that a break of the RWCU piping is detected.
 
This Function isolates the inboard and outboard RWCU pump suction penetration and the outboard valve at the RWCU connection to reactor feedwater.
 
5.b. Standby Liquid Control (SLC) System Initiation
 
The isolation of the RWCU System is required when the SLC System has been initiated to prevent dilution and removal of the boron solution by the RWCU System (Ref. 5). SLC System initiation signals are initiated from the remote SLC System start switch.
 
There is no Allowable Value associated with this Function since the channels are mechanically actuated based solely on the position of the SLC System initiation switch.
 
For reactivity insertion accidents, two channels of the SLC System Initiation Function are available and are required to be OPERABLE in MODES 1 and 2, since these are the only MODES where the reactor can be critical. In addition, for accidents involving significant fission product releases, both channels are required to be OPERABLE in MODES 1, 2, and 3. The SLC System is designed to maintain suppression pool pH at or above 7 following a LOCA to ensure that sufficient iodine will be retained in the suppression pool water. These MODES are consistent with the Applicability for the SLC System (LCO 3.1.7).
 
This Function isolates the inboard and outboard RWCU pump suction penetration and the outboard valve at the RWCU connection to reactor feedwater.
 
5.c. Reactor Vessel Water LevelLow (Level 3)
 
Low RPV water level indicates that the capability to cool the fuel may be threatened. Should RPV water level decrease too far, fuel damage could result. Therefore, isolation of some interfaces with the reactor vessel occurs to isolate the potential sources of a break. The isolation of the RWCU System on Level 3 supports actions to ensure that the fuel
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                        B 3.3-156                                                                                                                                                                                          Revision No. 75 Primary Containment Isolation Instrumentation B 3.3.6.1
 
BASES
 
APPLICABLE                                                                      5.c. Reactor Vessel Water LevelLow (Level 3)  (continued)
SAFETY ANALYSES, LCO, and                                                                                          peak cladding temperature remains below the limits of APPLICABILITY                                        10 CFR 50.46. The Reactor Vessel Water Level      Low (Level 3)
Function associated with RWCU isolation is not directly assumed in the UFSAR safety analyses because the RWCU System line break is bounded by breaks of larger systems (recirculation and MSL breaks are more limiting).
 
Reactor Vessel Water Level      Low (Level 3) signals are initiated from four level transmitters that sense the difference between the pressure due to a constant column of water (reference leg) and the pressure due to the actual water level (variable leg) in the vessel. Four channels of Reactor Vessel Water Level      Low (Level 3) Function are available and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function.
 
The Reactor Vessel Water Level      Low (Level 3) Allowable Value was chosen to be the same as the RPS Reactor Vessel Water Level      Low (Level 3) Allowable Value (LCO 3.3.1.1),
since the capability to cool the fuel may be threatened.
 
This Function isolates the inboard and outboard RWCU suction penetration and the outboard valve at the RWCU connection to reactor feedwater.
 
Shutdown Cooling System Isolation
 
6.a. Reactor PressureHigh
 
The Reactor Pressure      High Function is provided to isolate the shutdown cooling portion of the Residual Heat Removal (RHR) System. This Function is provided only for equipment protection to prevent an intersystem LOCA scenario, and credit for the Function is not assumed in the accident or transient analysis in the UFSAR.
 
The Reactor Pressure      High signals are initiated from two relays driven by trip units associated with pressure transmitters that sense RPV pressure at different taps on the RPV. Two channels of Reactor Pressure      High Function are available and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function. The Function is only required to be OPERABLE in
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                        B 3.3-157                                                                                                                                                                                Revision No. 135 Primary Containment Isolation Instrumentation B 3.3.6.1
 
BASES
 
APPLICABLE                                                                      6.a. Reactor PressureHigh  (continued)
SAFETY ANALYSES, LCO, and                                                                                          MODES 1, 2, and 3, since these are the only MODES in which APPLICABILITY                                        the reactor can be pressurized; thus, equipment protection is needed. The Allowable Value was chosen to be low enough to protect the system equipment from overpressurization.
 
This Function isolates both RHR shutdown cooling pump suction valves.
 
6.b. Reactor Vessel Water LevelLow (Level 3)
 
Low RPV water level indicates that the capability to cool the fuel may be threatened. Should RPV water level decrease too far, fuel damage could result. Therefore, isolation of some reactor vessel interfaces occurs to begin isolating the potential sources of a break. The Reactor Vessel Water Level      Low (Level 3) Function associated with RHR Shutdown Cooling System isolation is not directly assumed in safety analyses because a break of the RHR Shutdown Cooling System is bounded by breaks of the recirculation and MSL. The RHR Shutdown Cooling System isolation on Level 3 supports actions to ensure that the RPV water level does not drop below the top of the active fuel during a vessel draindown event caused by a leak (e.g., pipe break or inadvertent valve opening) in the RHR Shutdown Cooling System.
 
Reactor Vessel Water Level      Low (Level 3) signals are initiated from four level transmitters that sense the difference between the pressure due to a constant column of water (reference leg) and the pressure due to the actual water level (variable leg) in the vessel. Four channels (two channels per trip system) of the Reactor Vessel Water Level      Low (Level 3) Function are available and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                        B 3.3-158                                                                                                                                                                                      Revision No. 145 Primary Containment Isolation Instrumentation B 3.3.6.1
 
BASES
 
APPLICABLE                                                                      6.b. Reactor Vessel Water LevelLow (Level 3)  (continued)
SAFETY ANALYSES, LCO, and                                                                                          The Reactor Vessel Water Level      Low (Level 3) Allowable APPLICABILITY                                        Value was chosen to be the same as the RPS Reactor Vessel Water Level      Low (Level 3) Allowable Value (LCO 3.3.1.1),
since the capability to cool the fuel may be threatened.
 
The Reactor Vessel Water Level      Low (Level 3) Function is only required to be OPERABLE in MODE 3, to prevent this potential flow path from lowering the reactor vessel level to the top of the fuel. In MODES 1 and 2, another isolation (i.e., Reactor Pressure      High) and administrative controls ensure that this flow path remains isolated to prevent unexpected loss of inventory via this flow path.
 
This Function isolates both RHR shutdown cooling pump suction valves.
 
Feedwater Recirculation Isolation
 
7.a. Reactor PressureHigh
 
The Reactor Pressure      High Function is provided to isolate the feedwater recirculation line. This interlock is provided only for equipment protection to prevent an intersystem LOCA scenario, and credit for the interlock is not assumed in the accident or transient analysis in the UFSAR.
 
The Reactor Pressure      High signals are initiated from four transmitters that are connected to different taps on the RPV. Four channels of Reactor Pressure      High Function are available and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function. The Function is only required to be OPERABLE in MODES 1, 2, and 3, since these are the only MODES in which the reactor can be pressurized; thus, equipment protection is needed. The Allowable Value was chosen to be low enough to protect the system equipment from overpressurization.
 
This Function isolates the feedwater recirculation valves.
 
Traversing Incore Probe System Isolation
 
8.a. Reactor Vessel Water Level-Low, Level 3
 
Low RPV water level indicates that the capability to cool the fuel may be threatened. The valves whose penetrations communicate with the primary containment are isolated to
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                        B 3.3-159                                                                                                                                                                                Revision No. 145 Primary Containment Isolation Instrumentation B 3.3.6.1
 
BASES
 
APPLICABLE                                                                                8.a. Reactor Vessel Water Level-Low, Level 3 (continued)
SAFETY ANALYSES, LCO, and                                                                                          limit the release of fission products. The isolation of the APPLICABILITY                                                  primary containment on Level 3 supports actions to ensure that (continued)                                        offsite dose limits of 10 CFR 100 are not exceeded. The Reactor Vessel Water Level-Low, Level 3 Function associated with isolation is implicitly assumed in the FSAR analysis as these leakage paths are assumed to be isolated post LOCA.
 
Reactor Vessel Water Level-Low, Level 3 signals are initiated from level transmitters that sense the difference between the pressure due to a constant column of water (reference leg) and the pressure due to the actual water level (variable leg) in the vessel. Two channels of Reactor Vessel Water Level-Low, Level 3 Function are available and are required to be OPERABLE to ensure that no single instrument failure can initiate an inadvertent isolation actuation. The isolation function is ensured by the manual shear valve in each penetration.
 
The Reactor Vessel Water Level-Low, Level 3 Allowable Value was chosen to be the same as the RPS Level 3 scram Allowable Value (LCO 3.3.1.1), since isolation of these valves is not critical to orderly plant shutdown.
 
This Function isolates the Group II(D) TIP valves.
 
8.b. Drywell Pressure-High
 
High drywell pressure can indicate a break in the RCPB inside the primary containment. The isolation of some of the primary containment isolation valves on high drywell pressure supports actions to ensure that offsite dose limits of 10 CFR 100 are not exceeded. The Drywell Pressure-High Function, associated with isolation of the primary containment, is implicitly assumed in the FSAR accident analysis as these leakage paths are assumed to be isolated post LOCA.
 
High drywell pressure signals are initiated from pressure transmitters that sense the pressure in the drywell. Two channels of Drywell Pressure-High per Function are available and are required to be OPERABLE to ensure that no single instrument failure can initiate an inadvertent actuation. The isolation function is ensured by the manual shear valve in each penetration.
 
The allowable Value was selected to be the same as the ECCS Drywell Pressure-High Allowable Value (LCO 3.3.5.1), since this may be indicative of a LOCA inside primary containment.
 
This Function isolates the Group II(D) TIP valves.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                                      B 3.3-159a                                                                                                                                                                                                        Revision No. 57 Primary Containment Isolation Instrumentation B 3.3.6.1
 
BASES  (continued)
 
ACTIONS                                                                                                              The ACTIONS are modified by two Notes. Note 1 allows 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. Note 2 has been provided to modify the ACTIONS related to primary containment isolation instrumentation channels. 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 inoperable primary containment isolation instrumentation channels provide appropriate compensatory measures for separate inoperable channels. As such, a Note has been provided that allows separate Condition entry for each inoperable primary containment isolation instrumentation channel.
 
A.1
 
Because of the diversity of sensors available to provide isolation signals and the redundancy of the isolation design, an allowable out of service time of 12 hours for Functions 1.d, 2.a, and 2.b and 24 hours for Functions other than Functions 1.d, 2.a, and 2.b has been shown to be acceptable (Refs. 6 and 7) to permit restoration of any inoperable channel to OPERABLE status. This out of service time is only acceptable provided the associated Function is still maintaining isolation capability (refer to Required Action B.1 Bases). Alternatively, a Completion Time can be determined in accordance with the Risk Informed Completion Time (RICT) Program. A Note has been provided to indicate that a RICT is only applicable when a loss of function has not occurred. If the inoperable channel cannot be restored to OPERABLE status within the allowable out of service time, the channel must be placed in the tripped condition per Required Action A.1. Placing the inoperable channel in trip would conservatively compensate for the inoperability, restore capability to accommodate a single failure, and allow operation to continue with no further restrictions.
Alternately, if it is not desired to place the channel in trip (e.g., as in the case where placing the inoperable channel in trip would result in an isolation), Condition C must be entered and its Required Action taken.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                                            B 3.3-160                                                                                                                                                                                                    Revision No. 159 Primary Containment Isolation Instrumentation B 3.3.6.1
 
BASES
 
ACTIONS B.1 (continued)
Required Action B.1 is intended to ensure that appropriate actions are taken if multiple, inoperable, untripped channels within the same Function result in redundant isolation capability being lost for the associated penetration flow path(s). For those MSL, Primary Containment, HPCI, RCIC, RWCU, SDC, and Feedwater Recirculation Isolation Functions, where actuation of both trip systems is needed to isolate a penetration, the Functions are considered to be maintaining isolation capability when sufficient channels are OPERABLE or in trip (or the associated trip system in trip), such that both trip systems will generate a trip signal from the given Function on a valid signal. For those Primary Containment, HPCI, RCIC, RWCU, and SDC isolation functions, where actuation of one trip system is needed to isolate a penetration, the Functions are considered to be maintaining isolation capability when sufficient channels are OPERABLE or in trip, such that one trip system will generate a trip signal from the given function on a valid signal. This ensures that at least one of the PCIVs in the associated penetration flow path can receive an isolation signal from the given Function. For all Functions except 1.c, 1.e, 2.c, 3.a, 3.b, 3.e, 4.a, 4.b, 4.e, 5.a, 5.b, and 6.a, this would require both trip systems to have one channel OPERABLE or in trip.
For Function 1.c, this would require both trip systems to have one channel, associated with each MSL, OPERABLE or in trip. For Functions 1.e, 3.e and 4.e, each Function consists of channels that monitor several locations within a given area (e.g., different locations within the Turbine Building main steam tunnel area). Therefore, this would require both trip systems to have one channel per location OPERABLE or in trip. For Functions 2.c, 3.a, 3.b, 4.a, 4.b, 5.a, and 6.a, this would require one trip system to have one channel OPERABLE or in trip.
 
The Completion Time is intended to allow the operator time to evaluate and repair any discovered inoperabilities. The 1 hour Completion Time is acceptable because it minimizes risk while allowing time for restoration or tripping of channels.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                        B 3.3-161                                                                                                                                                                                          Revision No. 48 Primary Containment Isolation Instrumentation B 3.3.6.1
 
BASES
 
ACTIONS                                                                                                    B.1  (continued)
 
Entry into Condition B and Required Action B.1 may be necessary to avoid an MSL isolation transient resulting from a temporary loss of ventilation in the main steam line tunnel area. As allowed by LCO 3.0.2 (and discussed in the Bases of LCO 3.0.2), the plant may intentionally enter this Condition to avoid an MSL isolation transient following the loss of ventilation flow, and then raise the setpoints for the Main Steam Tunnel Temperature      High Function to 250&deg;F causing all channels of Main Steam Tunnel Temperature      High Function to be inoperable.
However, during the period that multiple Main Steam Tunnel Temperature      High Function channels are inoperable due to this intentional action, an additional compensatory measure is deemed necessary and shall be taken: an operator shall observe control room  indications of the duct temperature so the main steam line isolation valves may be promptly closed in the event of a rapid increase in MSL tunnel temperature indicative of a steam line break.
 
C.1
 
Required Action C.1 directs entry into the appropriate Condition referenced in Table 3.3.6.1-1. The applicable Condition specified in Table 3.3.6.1-1 is Function and MODE or other specified condition dependent and may change as the Required Action of a previous Condition is completed. Each time an inoperable channel has not met any Required Action of Condition A or B and the associated Completion Time has expired, Condition C will be entered for that channel and provides for transfer to the appropriate subsequent Condition.
 
D.1, D.2.1, and D.2.2
 
If the channel is not restored to OPERABLE status or placed in trip within the allowed Completion Time, the plant must be placed in a MODE or other specified condition in which the LCO does not apply. This is done by placing the plant in at least MODE 3 within 12 hours and in MODE 4 within 36 hours (Required Actions D.2.1 and D.2.2). Alternately, the associated MSLs may be isolated (Required Action D.1),
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                        B 3.3-162                                                                                                                                                                                          Revision No. 45 Primary Containment Isolation Instrumentation B 3.3.6.1
 
BASES
 
ACTIONS                                                                                                    D.1, D.2.1, and D.2.2  (continued)
 
and, if allowed (i.e., plant safety analysis allows operation with an MSL isolated), operation with that MSL isolated may continue. Isolating the affected MSL accomplishes the safety function of the inoperable channel.
The 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.
 
E.1
 
If the channel is not restored to OPERABLE status or placed in trip within the allowed Completion Time, the plant must be placed in a MODE or other specified condition in which the LCO does not apply. This is done by placing the plant in at least MODE 2 within 6 hours.
 
The allowed Completion Time of 6 hours is reasonable, based on operating experience, to reach MODE 2 from full power conditions in an orderly manner and without challenging plant systems.
 
F.1
 
If the channel is not restored to OPERABLE status or placed in trip within the allowed Completion Time, plant operations may continue if the affected penetration flow path(s) is isolated. Isolating the affected penetration flow path(s) accomplishes the safety function of the inoperable channels.
Alternately, if it is not desired to isolate the affected penetration flow path(s) (e.g., as in the case where isolating the penetration flow path(s) could result in a reactor scram), Condition G must be entered and its Required Actions taken. The 1 hour Completion Time is acceptable because it minimizes risk while allowing sufficient time for plant operations personnel to isolate the affected penetration flow path(s).
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                        B 3.3-163                                                                                                                                                                                                    Revision No. 0 Primary Containment Isolation Instrumentation B 3.3.6.1
 
BASES
 
ACTIONS                                                                                                    G.1 and G.2 (continued)
If the channel is not restored to OPERABLE status or placed in trip within the allowed Completion Time, or the Required Action of Condition F is not met and the associated Completion Time has expired, the plant must be placed in a MODE or other specified condition in which the LCO does not apply. This is done by placing the plant in at least MODE 3 within 12 hours and in MODE 4 within 36 hours. 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.
 
H.1 and H.2
 
If the channel is not restored to OPERABLE status or placed in trip within the allowed Completion Time, the associated SLC subsystem(s) is declared inoperable or the RWCU System is isolated. Since this Function is required to ensure that the SLC System performs its intended function, sufficient remedial measures are provided by declaring the associated SLC subsystems inoperable or isolating the RWCU System.
 
The 1 hour Completion Time is acceptable because it minimizes risk while allowing sufficient time for personnel to isolate the RWCU System.
 
I.1 and I.2
 
If the channel is not restored to OPERABLE status or placed in trip within the allowed Completion Time, the associated penetration flow path should be closed. However, if the shutdown cooling function is needed to provide core cooling, these Required Actions allow the penetration flow path to remain unisolated provided action is immediately initiated to restore the channel to OPERABLE status or to isolate the RHR Shutdown Cooling System (i.e., provide alternate decay heat removal capabilities so the penetration flow path can be isolated). Actions must continue until the channel is restored to OPERABLE status or the RHR Shutdown Cooling System is isolated.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                        B 3.3-164                                                                                                                                                                                                    Revision No. 0 Primary Containment Isolation Instrumentation B 3.3.6.1
 
BASES  (continued)
 
SURVEILLANCE                                                  As noted at the beginning of the SRs, the SRs for each REQUIREMENTS                                                  Primary Containment Isolation instrumentation Function are found in the SRs column of Table 3.3.6.1-1.
 
The Surveillances are modified by a Note to indicate that when a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours provided the associated Function maintains trip capability. Upon completion of the Surveillance, or expiration of the 6 hour allowance, the channel must be returned to OPERABLE status or the applicable Condition entered and Required Actions taken. This Note is based on the reliability analysis (Refs. 6 and 7) assumption of the average time required to perform channel surveillance. That analysis demonstrated that the 6 hour testing allowance does not significantly reduce the probability that the PCIVs will isolate the penetration flow path(s) when necessary.
 
SR  3.3.6.1.1
 
Performance of the CHANNEL CHECK ensures that a gross failure of instrumentation has not occurred. A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a similar parameter on other channels. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value.
Significant deviations between the instrument channels could be an indication of excessive instrument drift in one of the channels or of something even more serious. A CHANNEL CHECK will detect gross channel failure; thus, it is key to verifying the instrumentation continues to operate properly between each CHANNEL CALIBRATION.
 
Agreement criteria are determined by the plant staff based on a combination of the channel instrument uncertainties, including indication and readability. If a channel is outside the criteria, it may be an indication that the instrument has drifted outside its limit.
 
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. The CHANNEL CHECK supplements less formal, but more frequent, checks of channels during normal operational use of the displays associated with the channels required by the LCO.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                        B 3.3-165                                                                                                                                                                                          Revision No. 86 Primary Containment Isolation Instrumentation B 3.3.6.1
 
BASES
 
SURVEILLANCE                                                  SR  3.3.6.1.2 REQUIREMENTS (continued)                      A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the entire channel will perform the intended function. Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology. For Function 1.e, 1.f, 3.e, and 4.e channels, verification that trip settings are less than or equal to the specified Allowable Value during the CHANNEL FUNCTIONAL TEST is not required since the installed indication instrumentation does not provide accurate indication of the trip setting. This is considered acceptable since the magnitude of drift assumed in the setpoint calculation is based on a 24 month calibration interval.
 
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
 
SR  3.3.6.1.3, SR  3.3.6.1.4, and SR  3.3.6.1.5 (SR  3.3.6.1.6 Deleted)
 
A CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor. This test verifies the channel responds to the measured parameter within the necessary range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drifts between successive calibrations, consistent with the assumptions of the current setpoint methodology.
 
Specific to Main Steam Line Pressure-Low (Technical Specification Table 3.3.6.1-1, Function 1.b) and the Main Steam Line Flow-High (Technical Specification Table 3.3.6.1-1, Function 1.c), there is a plant specific program which verifies that this instrument channel functions as required by verifying the as-left and as-found settings are consistent with those established by the setpoint methodology.
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                        B 3.3-166                                                                                                                                                                                Revision No. 134 Primary Containment Isolation Instrumentation B 3.3.6.1
 
BASES
 
SURVEILLANCE                                                  SR  3.3.6.1.3, SR  3.3.6.1.4, SR  3.3.6.1.5, and REQUIREMENTS                                                  SR  3.3.6.1.6  (continued)
 
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
 
SR  3.3.6.1.7
 
The LOGIC SYSTEM FUNCTIONAL TEST demonstrates the OPERABILITY of the required isolation logic for a specific channel. The system functional testing performed on PCIVs in LCO 3.6.1.3 overlaps this Surveillance to provide complete testing of the assumed safety function.
 
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
 
REFERENCES                                                                      1.                    UFSAR, Section 7.3.
: 2.                    NRC Safety Evaluation Report for Amendment Numbers 156 and 158 to Facility Operating License Numbers DPR-44 and DPR-56, Peach Bottom Atomic Power Station, Unit Nos. 2 and 3, September 7, 1990.
: 3.                    UFSAR, Chapter 14.
: 4.                    NEDO-31466, "Technical Specification Screening Criteria Application and Risk Assessment,"
November 1987.
: 5.                    UFSAR, Section 4.9.3.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                        B 3.3-167                                                                                                                                                                                Revision No. 114 Primary Containment Isolation Instrumentation B 3.3.6.1
 
BASES
: 6.                    NEDC-31677P-A, "Technical Specification Improvement Analysis for BWR Isolation Actuation Instrumentation,"
July 1990.
: 7.                    NEDC-30851P-A Supplement 2, "Technical Specifications Improvement Analysis for BWR Isolation Instrumentation Common to RPS and ECCS Instrumentation," March 1989.
: 8.                    NEDC-33873P, Safety Analysis Report for Peach Bottom Atomic Power Station, Units 2 and 3, Thermal Power Optimization, Revision 0.
 
PBAPS UNIT 2                                                                                                                                                                                                                        B 3.3-168                                                                                                                                                                                Revision No. 143 Secondary Containment Isolation Instrumentation B 3.3.6.2
 
B 3.3  INSTRUMENTATION
 
B 3.3.6.2  Secondary Containment Isolation Instrumentation
 
BASES
 
BACKGROUND                                                                      The secondary containment isolation instrumentation automatically initiates closure of appropriate secondary containment isolation valves (SCIVs) and starts the Standby Gas Treatment (SGT) System. The function of these systems, in combination with other accident mitigation systems, is to limit fission product release during and following postulated Design Basis Accidents (DBAs) (Ref. 1).
Secondary containment isolation and establishment of vacuum with the SGT System within the required time limits ensures that fission products that leak from primary containment following a DBA, or are released outside primary containment, or are released during certain operations when primary containment is not required to be OPERABLE are maintained within applicable limits.
 
The isolation instrumentation includes the sensors, relays, and switches that are necessary to cause initiation of secondary containment isolation. Most channels include electronic equipment (e.g., trip units) that compares measured input signals with pre-established setpoints. When the setpoint is exceeded, the channel output relay actuates, which then outputs a secondary containment isolation signal to the isolation logic. Functional diversity is provided by monitoring a wide range of independent parameters. The input parameters to the isolation logic are (1) reactor vessel water level, (2) drywell pressure, (3) reactor building ventilation exhaust high radiation, and (4) refueling floor ventilation exhaust high radiation.
Redundant sensor input signals from each parameter are provided for initiation of isolation.
 
The outputs of the channels are arranged in a one-out-of-two taken twice logic. Automatic isolation valves (dampers) isolate and SGT subsystems start when both trip systems are in trip. Operation of both trip systems is required to isolate the secondary containment and provide for the necessary filtration of fission products.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                        B 3.3-169                                                                                                                                                                                                    Revision No. 1 Secondary Containment Isolation Instrumentation B 3.3.6.2
 
BASES  (continued)
 
APPLICABLE                                                                      The isolation signals generated by the secondary containment SAFETY ANALYSES,          isolation instrumentation are implicitly assumed in the LCO, and                                                                                          safety analyses of References 1 and 2 to initiate closure APPLICABILITY                                        of valves and start the SGT System to limit offsite doses.
 
Refer to LCO 3.6.4.2, "Secondary Containment Isolation Valves (SCIVs)," and LCO 3.6.4.3, "Standby Gas Treatment (SGT) System," Applicable Safety Analyses Bases for more detail of the safety analyses.
 
The secondary containment isolation instrumentation satisfies Criterion 3 of the NRC Policy Statement. Certain instrumentation Functions are retained for other reasons and are described below in the individual Functions discussion.
 
The OPERABILITY of the secondary containment isolation instrumentation is dependent on the OPERABILITY of the individual instrumentation channel Functions. Each Function must have the required number of OPERABLE channels with their setpoints set within the specified Allowable Values, as shown in Table 3.3.6.2-1. The actual setpoint is calibrated consistent with applicable setpoint methodology assumptions. A channel is inoperable if its actual trip setting is not within its required Allowable Value.
 
Allowable Values are specified for each Function specified in the Table. Trip setpoints are specified in the setpoint calculations. The trip setpoints are selected to ensure that the setpoints do not exceed the Allowable Value between CHANNEL CALIBRATIONS. Operation with a trip setting less conservative than the trip setpoint, but within its Allowable Value, is acceptable.
 
Trip setpoints are those predetermined values of output at which an action should take place. The setpoints are compared to the actual process parameter (e.g., reactor vessel water level), and when the measured output value of the process parameter exceeds the setpoint, the associated device (e.g., trip unit) changes state. The analytic or design limits are derived from the limiting values of the process parameters obtained from the safety analysis or other appropriate documents. The Allowable Values are derived from the analytic or design limits, corrected for calibration, process, and instrument errors. The trip setpoints are then determined from analytical or design limits, corrected for calibration, process, and instrument
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                        B 3.3-170                                                                                                                                                                                                    Revision No. 1 Secondary Containment Isolation Instrumentation B 3.3.6.2
 
BASES
 
APPLICABLE                                                                      errors, as well as, instrument drift. In selected cases, SAFETY ANALYSES,          the Allowable Values and trip setpoints are determined by LCO, and                                                                                          engineering judgement or historically accepted practice APPLICABILITY                                        relative to the intended function of the channel. The (continued)                                        trip setpoints determined in this manner provide adequate protection by assuring instrument and process uncertainties expected for the environments during the operating time of the associated channels are accounted for.
 
In general, the individual Functions are required to be OPERABLE in the MODES or other specified conditions when SCIVs and the SGT System are required.
 
The specific Applicable Safety Analyses, LCO, and Applicability discussions are listed below on a Function by Function basis.
: 1. Reactor Vessel Water LevelLow (Level 3)
 
Low reactor pressure vessel (RPV) water level indicates that the capability to cool the fuel may be threatened. Should RPV water level decrease too far, fuel damage could result.
An isolation of the secondary containment and actuation of the SGT System are initiated in order to minimize the potential of an offsite dose release. The Reactor Vessel Water Level      Low (Level 3) Function is one of the Functions assumed to be OPERABLE and capable of providing isolation and initiation signals. The isolation and initiation systems on Reactor Vessel Water Level      Low (Level 3) support actions to ensure that any offsite releases are within the limits calculated in the safety analysis.
 
Reactor Vessel Water Level      Low (Level 3) signals are initiated from level transmitters that sense the difference between the pressure due to a constant column of water (reference leg) and the pressure due to the actual water level (variable leg) in the vessel. Four channels of Reactor Vessel Water Level      Low (Level 3) Function are available and are required to be OPERABLE in MODES 1, 2, and 3 to ensure that no single instrument failure can preclude the isolation function.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                        B 3.3-171                                                                                                                                                                                                    Revision No. 1 Secondary Containment Isolation Instrumentation B 3.3.6.2
 
BASES
 
APPLICABLE                                                                      1. Reactor Vessel Water LevelLow (Level 3)  (continued)
SAFETY ANALYSES, LCO, and                                                                                          The Reactor Vessel Water Level      Low (Level 3) Allowable APPLICABILITY                                        Value was chosen to be the same as the RPS Level 3 scram Allowable Value (LCO 3.3.1.1), since isolation of these valves and SGT System start are not critical to orderly plant shutdown.
 
The Reactor Vessel Water Level      Low (Level 3) Function is required to be OPERABLE in MODES 1, 2, and 3 where considerable energy exists in the Reactor Coolant System (RCS); thus, there is a probability of pipe breaks resulting in significant releases of radioactive steam and gas. In MODES 4 and 5, the probability and consequences of these events are low due to the RCS pressure and temperature limitations of these MODES; thus, this Function is not required.
: 2. Drywell PressureHigh
 
High drywell pressure can indicate a break in the reactor coolant pressure boundary (RCPB). An isolation of the secondary containment and actuation of the SGT System are initiated in order to minimize the potential of an offsite dose release. The isolation on high drywell pressure supports actions to ensure that any offsite releases are within the limits calculated in the safety analysis. The Drywell Pressure      High Function associated with isolation is not assumed in any UFSAR accident or transient analyses but will provide an isolation and initiation signal. It is retained for the overall redundancy and diversity of the secondary containment isolation instrumentation as required by the NRC approved licensing basis.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                        B 3.3-172                                                                                                                                                                                Revision No. 145 Secondary Containment Isolation Instrumentation B 3.3.6.2
 
BASES
 
APPLICABLE                                                                      2. Drywell PressureHigh  (continued)
SAFETY ANALYSES, LCO, and                                                                                          High drywell pressure signals are initiated from pressure APPLICABILITY                                        transmitters that sense the pressure in the drywell. Four channels of Drywell Pressure      High Functions are available and are required to be OPERABLE to ensure that no single instrument failure can preclude performance of the isolation function.
 
The Allowable Value was chosen to be the same as the ECCS Drywell Pressure      High Function Allowable Value (LCO 3.3.5.1) since this is indicative of a loss of coolant accident (LOCA).
 
The Drywell Pressure      High Function is required to be OPERABLE in MODES 1, 2, and 3 where considerable energy exists in the RCS; thus, there is a probability of pipe breaks resulting in significant releases of radioactive steam and gas. This Function is not required in MODES 4 and 5 because the probability and consequences of these events are low due to the RCS pressure and temperature limitations of these MODES.
 
3., 4. Reactor Building Ventilation and Refueling Floor Ventilation Exhaust RadiationHigh
 
High secondary containment exhaust radiation is an indication of possible gross failure of the fuel cladding.
The release may have originated from the primary containment due to a break in the RCPB or during refueling due to a fuel handling accident. When Ventilation Exhaust Radiation      High is detected, secondary containment isolation and actuation of the SGT System are initiated to limit the release of fission products as assumed in the UFSAR safety analyses (Ref. 4).
 
The Ventilation Exhaust Radiation      High signals are initiated from radiation detectors that are located on the ventilation exhaust piping coming from the reactor building and the refueling floor zones, respectively. The signal from each detector is input to an individual monitor whose trip outputs are assigned to an isolation channel. Four
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                        B 3.3-173                                                                                                                                                                                                    Revision No. 1 Secondary Containment Isolation Instrumentation B 3.3.6.2
 
BASES
 
APPLICABLE                                                                      3, 4. Reactor Building Ventilation and Refueling Floor SAFETY ANALYSES,          Ventilation Exhaust RadiationHigh  (continued)
LCO, and APPLICABILITY                                        channels of Reactor Building Ventilation Exhaust Radiation      High Function and four channels of Refueling Floor Ventilation Exhaust Radiation      High Function are available and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function.
 
The Allowable Values are chosen to promptly detect gross failure of the fuel cladding.
 
The Reactor Building Ventilation and Refueling Floor Ventilation Exhaust Radiation      High Functions are required to be OPERABLE in MODES 1, 2, and 3 where considerable energy exists; thus, there is a probability of pipe breaks resulting in significant releases of radioactive steam and gas. In MODES 4 and 5, the probability and consequences of these events are low due to the RCS pressure and temperature limitations of these MODES; thus, these Functions are not required. In addition, the Functions are also required to be OPERABLE during movement of RECENTLY IRRADIATED FUEL assemblies in the secondary containment, because the capability of detecting radiation releases due to fuel failures (due to fuel uncovery or dropped fuel assemblies) must be provided to ensure that offsite dose limits are not exceeded.
 
ACTIONS                                                                                                      A Note has been provided to modify the ACTIONS related to secondary containment isolation instrumentation channels.
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 inoperable secondary containment isolation instrumentation channels provide appropriate compensatory measures for separate inoperable channels. As such, a Note has been provided that allows separate Condition entry for each inoperable secondary containment isolation instrumentation channel.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                        B 3.3-174                                                                                                                                                                                Revision No. 145 Secondary Containment Isolation Instrumentation B 3.3.6.2
 
BASES
 
ACTIONS A.1 (continued)
Because of the diversity of sensors available to provide isolation signals and the redundancy of the isolation design, an allowable out of service time of 12 hours for Functions 1 and 2, and 24 hours for Functions other than Functions 1 and 2, has been shown to be acceptable (Refs. 5 and 6) to permit restoration of any inoperable channel to OPERABLE status. This out of service time is only acceptable provided the associated Function is still maintaining isolation capability (refer to Required Action B.1 Bases). If the inoperable channel cannot be restored to OPERABLE status within the allowable out of service time, the channel must be placed in the tripped condition per Required Action A.1. Placing the inoperable channel in trip would conservatively compensate for the inoperability, restore capability to accommodate a single failure, and allow operation to continue. Alternately, if it is not desired to place the channel in trip (e.g., as in the case where placing the inoperable channel in trip would result in an isolation), Condition C must be entered and its Required Actions taken.
 
B.1
 
Required Action B.1 is intended to ensure that appropriate actions are taken if multiple, inoperable, untripped channels within the same Function result in a complete loss of isolation capability for the associated penetration flow path(s) or a complete loss of automatic initiation capability for the SGT System. A Function is considered to be maintaining secondary containment isolation capability when sufficient channels are OPERABLE or in trip, such that both trip systems will generate a trip signal from the given Function on a valid signal. This ensures that at least one of the two SCIVs in the associated penetration flow path and at least one SGT subsystem can be initiated on an isolation signal from the given Function. For Functions 1, 2, 3, and 4, this would require both trip systems to have one channel OPERABLE or in trip.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                        B 3.3-175                                                                                                                                                                                                    Revision No. 1 Secondary Containment Isolation Instrumentation B 3.3.6.2
 
BASES
 
ACTIONS                                                                                                    B.1  (continued)
 
The Completion Time is intended to allow the operator time to evaluate and repair any discovered inoperabilities. The 1 hour Completion Time is acceptable because it minimizes risk while allowing time for restoration or tripping of channels.
 
C.1.1, C.1.2, C.2.1, and C.2.2
 
If any Required Action and associated Completion Time of Condition A or B are not met, the ability to isolate the secondary containment and start the SGT System cannot be ensured. Therefore, further actions must be performed to ensure the ability to maintain the secondary containment function. Isolating the associated secondary containment penetration flow path(s) and starting the associated SGT subsystem (Required Actions C.1.1 and C.2.1) performs the intended function of the instrumentation and allows operation to continue.
 
Alternately, declaring the associated SCIVs or SGT subsystem(s) inoperable (Required Actions C.1.2 and C.2.2) is also acceptable since the Required Actions of the respective LCOs (LCO 3.6.4.2 and LCO 3.6.4.3) provide appropriate actions for the inoperable components.
 
One hour is sufficient for plant operations personnel to establish required plant conditions or to declare the associated components inoperable without unnecessarily challenging plant systems.
 
SURVEILLANCE                                                  As noted at the beginning of the SRs, the SRs for each REQUIREMENTS                                                  Secondary Containment Isolation instrumentation Function are located in the SRs column of Table 3.3.6.2-1.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                        B 3.3-176                                                                                                                                                                                                    Revision No. 1 Secondary Containment Isolation Instrumentation B 3.3.6.2
 
BASES
 
SURVEILLANCE                                                  The Surveillances are modified by a Note to indicate that REQUIREMENTS                                                  when a channel is placed in an inoperable status solely for (continued)                                        performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours provided the associated Function maintains secondary containment isolation capability. Upon completion of the Surveillance, or expiration of the 6 hour allowance, the channel must be returned to OPERABLE status or the applicable Condition entered and Required Actions taken.
This Note is based on the reliability analysis (Refs. 5 and 6) assumption that of the average time required to perform channel surveillance. That analysis demonstrated the 6 hour testing allowance does not significantly reduce the probability that the SCIVs will isolate the associated penetration flow paths and that the SGT System will initiate when necessary.
 
SR  3.3.6.2.1
 
Performance of the CHANNEL CHECK ensures that a gross failure of instrumentation has not occurred. A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a similar parameter on other channels. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value.
Significant deviations between the instrument channels could be an indication of excessive instrument drift in one of the channels or something even more serious. A CHANNEL CHECK will detect gross channel failure; thus, it is key to verifying the instrumentation continues to operate properly between each CHANNEL CALIBRATION.
 
Agreement criteria are determined by the plant staff based on a combination of the channel instrument uncertainties, including indication and readability. If a channel is outside the criteria, it may be an indication that the instrument has drifted outside its limit.
 
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. The CHANNEL CHECK supplements less formal, but more frequent, checks of channel status during normal operational use of the displays associated with channels required by the LCO.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                        B 3.3-177                                                                                                                                                                                          Revision No. 86 Secondary Containment Isolation Instrumentation B 3.3.6.2
 
BASES
 
SURVEILLANCE                                                  SR  3.3.6.2.2 REQUIREMENTS (continued)                                        A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the entire channel will perform the intended function. Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
 
SR  3.3.6.2.3 and SR  3.3.6.2.4T
 
A CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor. This test verifies the channel responds to the measured parameter within the necessary range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drifts between successive calibrations, consistent with the current plant specific setpoint methodology.
 
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
 
SR  3.3.6.2.5
 
The LOGIC SYSTEM FUNCTIONAL TEST demonstrates the OPERABILITY of the required isolation logic for a specific channel. The system functional testing performed on SCIVs and the SGT System in LCO 3.6.4.2 and LCO 3.6.4.3, respectively, overlaps this Surveillance to provide complete testing of the assumed safety function.
 
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                        B 3.3-178                                                                                                                                                                                          Revision No. 86 Secondary Containment Isolation Instrumentation B 3.3.6.2
 
BASES  (continued)
 
REFERENCES                                                                      1.                    UFSAR, Section 14.6.
: 2.                    UFSAR, Chapter 14.
: 3.                    UFSAR, Section 14.6.5.
: 4.                    UFSAR, Sections 14.6.3 and 14.6.4.
: 5.                    NEDC-31677P-A, "Technical Specification Improvement Analysis for BWR Isolation Actuation Instrumentation,"
July 1990.
: 6.                    NEDC-30851P-A Supplement 2, "Technical Specifications Improvement Analysis for BWR Isolation Instrumentation Common to RPS and ECCS Instrumentation," March 1989.
 
PBAPS UNIT 2                                                                                                                                                                                                                        B 3.3-179                                                                                                                                                                                                    Revision No. 1 MCREV System Instrumentation B 3.3.7.1
 
B 3.3  INSTRUMENTATION
 
B 3.3.7.1  Main Control Room Emergency Ventilation (MCREV) System Instrumentation
 
BASES
 
BACKGROUND                                                                      The MCREV System is designed to provide a radiologically controlled environment to ensure the habitability of the control room for the safety of control room operators under all plant conditions. Two independent MCREV subsystems are each capable of fulfilling the stated safety function. The instrumentation and controls for the MCREV System automatically initiate action to pressurize the main control room (MCR) to minimize the consequences of radioactive material in the control room environment.
 
In the event of a Control Room Air Intake Radiation      High signal, the MCREV System is automatically started in the pressurization mode. The outside air from the normal ventilation intake is then passed through one of the charcoal filter subsystems. Sufficient outside air is drawn in through the normal ventilation intake to maintain the MCR slightly pressurized with respect to the turbine building.
 
The MCREV System instrumentation has two trip systems with two Control Room Air Intake Radiation      High channels in each trip system. The outputs of the Control Room Air Intake Radiation      High channels are arranged in two trip systems, which use a one-out-of-two logic. The tripping of both trip systems will initiate both MCREV subsystems. The channels include electronic equipment (e.g., trip units) that compares measured input signals with pre-established setpoints. When the setpoint is exceeded, the channel output relay actuates, which then outputs a MCREV System initiation signal to the initiation logic.
 
APPLICABLE                                                                      The ability of the MCREV System to maintain the habitability SAFETY ANALYSES,          of the MCR is explicitly assumed for certain accidents as LCO, and                                                                                          discussed in the UFSAR safety analyses (Refs. 1, 2, and 3).
APPLICABILITY                                        MCREV System operation ensures that the radiation exposure of control room personnel, through the duration of any one of the postulated accidents, does not exceed acceptable limits.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                        B 3.3-180                                                                                                                                                                                                    Revision No. 1 MCREV System Instrumentation B 3.3.7.1
 
BASES
 
APPLICABLE                                                                      MCREV System instrumentation satisfies Criterion 3 of the SAFETY ANALYSES,          NRC Policy Statement.
LCO, and APPLICABILITY                                        The OPERABILITY of the MCREV System instrumentation is (continued)                                        dependent upon the OPERABILITY of the Control Room Air Intake Radiation      High instrumentation channel Function.
The Function must have a required number of OPERABLE channels, with their setpoints within the specified Allowable Values, where appropriate. A channel is inoperable if its actual trip setting is not within its required Allowable Value. The actual setpoint is calibrated consistent with applicable setpoint methodology assumptions.
 
Allowable Values are specified for the MCREV System Control Room Air Intake Radiation      High Function. Trip setpoints are specified in the setpoint calculations. The trip setpoints are selected to ensure that the setpoints do not exceed the Allowable Value between successive CHANNEL CALIBRATIONS. Operation with a trip setting less conservative than the trip setpoint, but within its Allowable Value, is acceptable. Trip setpoints are those predetermined values of output at which an action should take place. The setpoints are compared to the actual process parameter (e.g., control room air intake radiation),
and when the measured output value of the process parameter exceeds the setpoint, the associated device changes state.
The analytic limits are derived from the limiting values of the process parameters obtained from the safety analysis.
The Allowable Values are derived from the analytic limits, corrected for calibration, process, and instrument errors.
The trip setpoints are determined from analytical or design limits, corrected for calibration, process, and instrument errors, as well as, instrument drift. The trip setpoints derived in this manner provide adequate protection by ensuring instrument and process uncertainties expected for the environments during the operating time of the associated channels are accounted for.
 
The control room air intake radiation monitors measure radiation levels in the fresh air supply plenum. A high radiation level may pose a threat to MCR personnel; thus, automatically initiating the MCREV System.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                        B 3.3-181                                                                                                                                                                                                    Revision No. 1 MCREV System Instrumentation B 3.3.7.1
 
BASES
 
APPLICABLE                                                                      The Control Room Air Intake Radiation      High Function SAFETY ANALYSES,          consists of four independent monitors. Two channels of LCO, and                                                                                          Control Room Air Intake Radiation      High per trip system are APPLICABILITY                                        available and are required to be OPERABLE to ensure that no (continued)                                        single instrument failure can preclude MCREV System initiation. The Allowable Value was selected to ensure protection of the control room personnel.
 
The Control Room Air Intake Radiation      High Function is required to be OPERABLE in MODES 1, 2, and 3 and during CORE ALTERATIONS, and movement of irradiated fuel assemblies in the secondary containment, to ensure that control room personnel are protected during a LOCA, or fuel handling event. During MODES 4 and 5, when these specified conditions are not in progress (e.g., CORE ALTERATIONS), the probability of a LOCA or fuel damage is low; thus, the Function is not required.
 
ACTIONS                                                                                                    A Note has been provided to modify the ACTIONS related to MCREV System instrumentation channels. 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 inoperable MCREV System instrumentation channels provide appropriate compensatory measures for separate inoperable channels. As such, a Note has been provided that allows separate Condition entry for each inoperable MCREV System instrumentation channel.
 
A.1 and A.2
 
Because of the redundancy of sensors available to provide initiation signals and the redundancy of the MCREV System design, an allowable out of service time of 6 hours has been shown to be acceptable (Ref. 4), to permit restoration of any inoperable channel to OPERABLE status. However, this out of service time is only acceptable provided the Control Room Air Intake Radiation      High Function is still maintaining MCREV System initiation capability. The Function is considered to be maintaining MCREV System
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                        B 3.3-182                                                                                                                                                                                Revision No. 145 MCREV System Instrumentation B 3.3.7.1
 
BASES
 
ACTIONS                                                                                                    A.1 and A.2  (continued)
 
initiation capability when sufficient channels are OPERABLE or in trip such that the two trip systems will generate an initiation signal from the given Function on a valid signal.
For the Control Room Air Intake Radiation      High Function, this would require the two trip systems to have one channel per trip system OPERABLE or in trip. In this situation (loss of MCREV System initiation capability), the 6 hour allowance of Required Action A.2 is not appropriate. If the Function is not maintaining MCREV System initiation capability, the MCREV System must be declared inoperable within 1 hour of discovery of the loss of MCREV System initiation capability in both trip systems.
 
The 1 hour Completion Time (A.1) is acceptable because it minimizes risk while allowing time for restoring or tripping of channels.
 
If the inoperable channel cannot be restored to OPERABLE status within the allowable out of service time, the channel must be placed in the tripped condition per Required Action A.2. Placing the inoperable channel in trip would conservatively compensate for the inoperability, restore capability to accommodate a single failure, and allow operation to continue. Alternately, if it is not desired to place the channel in trip (e.g., as in the case where placing the inoperable channel in trip would result in an initiation), Condition B must be entered and its Required Action taken.
 
B.1 and B.2
 
With any Required Action and associated Completion Time not met, the associated MCREV subsystem(s) must be placed in operation per Required Action B.1 to ensure that control room personnel will be protected in the event of a Design Basis Accident. The method used to place the MCREV subsystem(s) in operation must provide for automatically re-initiating the subsystem(s) upon restoration of power following a loss of power to the MCREV subsystem(s).
Alternately, if it is not desired to start the subsystem(s),
the MCREV subsystem(s) associated with inoperable, untripped
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                        B 3.3-183                                                                                                                                                                                                    Revision No. 1 MCREV System Instrumentation B 3.3.7.1
 
BASES
 
ACTIONS                                                                                                    B.1 and B.2  (continued)
 
channels must be declared inoperable within 1 hour. Since each trip system can affect both MCREV subsystems, Required Actions B.1 and B.2 can be performed independently on each MCREV subsystem. That is, one MCREV subsystem can be placed in operation (Required Action B.1) while the other MCREV subsystem can be declared inoperable (Required Action B.2).
 
The 1 hour Completion Time is intended to allow the operator time to place the MCREV subsystem(s) in operation. The 1 hour Completion Time is acceptable because it minimizes risk while allowing time for placing the associated MCREV subsystem(s) in operation, or for entering the applicable Conditions and Required Actions for the inoperable MCREV subsystem(s).
 
SURVEILLANCE                                                  The Surveillances are modified by a Note to indicate that REQUIREMENTS                                                  when a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours, provided the associated Function maintains MCREV System initiation capability. Upon completion of the Surveillance, or expiration of the 6 hour allowance, the channel must be returned to OPERABLE status or the applicable Condition entered and Required Actions taken.
This Note is based on the reliability analysis (Ref. 4) assumption of the average time required to perform channel surveillance. That analysis demonstrated that the 6 hour testing allowance does not significantly reduce the probability that the MCREV System will initiate when necessary.
 
SR  3.3.7.1.1
 
Performance of the CHANNEL CHECK ensures that a gross failure of instrumentation has not occurred. A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a similar parameter on other channels. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value.
Significant deviations between the instrument channels could be an indication of excessive instrument drift in one of the channels or something even more serious. A CHANNEL CHECK will detect
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                        B 3.3-184                                                                                                                                                                                          Revision No. 86 MCREV System Instrumentation B 3.3.7.1
 
BASES
 
SURVEILLANCE                                                  SR  3.3.7.1.1  (continued)
REQUIREMENTS gross channel failure; thus, it is key to verifying the instrumentation continues to operate properly between each CHANNEL CALIBRATION.
 
Agreement criteria are determined by the plant staff, based on a combination of the channel instrument uncertainties, including indication and readability. If a channel is outside the criteria, it may be an indication that the instrument has drifted outside its limit.
 
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. The CHANNEL CHECK supplements less formal, but more frequent, checks of channel status during normal operational use of the displays associated with channels required by the LCO.
 
SR  3.3.7.1.2
 
A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the entire channel will perform the intended function. Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology.
 
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
 
SR  3.3.7.1.3
 
A CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor. This test verifies the channel responds to the measured parameter within the necessary range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drifts between successive calibrations, consistent with the assumptions of the plant specific setpoint methodology.
 
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                        B 3.3-185                                                                                                                                                                                          Revision No. 86 MCREV System Instrumentation B 3.3.7.1
 
BASES
 
SURVEILLANCE                                                  SR  3.3.7.1.4 REQUIREMENTS (continued)                                        The LOGIC SYSTEM FUNCTIONAL TEST demonstrates the OPERABILITY of the required initiation logic for a specific channel. The system functional testing performed in LCO 3.7.4, "Main Control Room Emergency Ventilation (MCREV)
System," overlaps this Surveillance to provide complete testing of the assumed safety function.
 
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
 
REFERENCES                                                                      1.                    UFSAR, Section 10.13.
: 2.                    UFSAR, Section 12.3.4.
: 3.                    UFSAR, Section 14.9.1.5.
: 4.                    GENE-770-06-1, "Bases for Changes to Surveillance Test Intervals and Allowed Out-of-Service Times for Selected Instrumentation Technical Specifications,"
February 1991.
 
PBAPS UNIT 2                                                                                                                                                                                                                        B 3.3-186                                                                                                                                                                                          Revision No. 86 LOP Instrumentation B 3.3.8.1
 
B 3.3  INSTRUMENTATION
 
B 3.3.8.1  Loss of Power (LOP) Instrumentation
 
BASES
 
BACKGROUND                                                                      Successful operation of the required safety functions of the Emergency Core Cooling Systems (ECCS) is dependent upon the availability of adequate power for energizing various components such as pump motors, motor operated valves, and the associated control components. The LOP instrumentation monitors the 4 kV emergency buses voltage. Offsite power is the preferred source of power for the 4 kV emergency buses.
If the LOP instrumentation detects that voltage levels are too low, the buses are disconnected from the offsite power sources and connected to the onsite diesel generator (DG) power sources.
 
Each Unit 2 4 kV emergency bus has its own independent LOP instrumentation and associated trip logic. The voltage for each bus is monitored at five levels, which can be considered as two different undervoltage Functions: one level of loss of voltage and four levels of degraded voltage. The Functions cause various bus transfers and disconnects. The degraded voltage Function is monitored by four undervoltage relays per source and the loss of voltage Function is monitored by one undervoltage relay for each emergency bus. The degraded voltage outputs and the loss of voltage outputs are arranged in a one-out-of-one trip logic configuration. Each channel consists of four protective relays that compare offsite source voltages with pre-established setpoints. When the sensed voltage is below the setpoint for a degraded voltage channel, the preferred offsite source breaker to the 4 kV emergency bus is tripped and autotransfer to the alternate offsite source is initiated. If the alternate source does not provide adequate voltage to the bus as sensed by its degraded grid relays, a diesel generator start signal is initiated.
 
A description of the Unit 3 LOP instrumentation is provided in the Bases for Unit 3 LCO 3.3.8.1.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                        B 3.3-187                                                                                                                                                                                                    Revision No. 5 LOP Instrumentation B 3.3.8.1
 
BASES  (continued)
 
APPLICABLE                                                                      The LOP instrumentation is required for Engineered Safety SAFETY ANALYSES,          Features to function in any accident with a loss of offsite LCO, and                                                                                          power. The required channels of LOP instrumentation ensure APPLICABILITY                                        that the ECCS and other assumed systems powered from the DGs, provide plant protection in the event of any of the Reference 1 (UFSAR) analyzed accidents in which a loss of offsite power is assumed. The first level is loss of voltage. This loss of voltage level detects and disconnects the Class 1E buses from the offsite power source upon a total loss of voltage. The second level of undervoltage protection is provided by the four levels of degraded grid voltage relays which are set to detect a sustained low voltage condition. These degraded grid relays disconnect the Class 1E buses from the offsite power source if the degraded voltage condition exists for a time interval which could prevent the Class 1E equipment from achieving its safety function. The degraded grid relays also prevent the Class 1E equipment from sustaining damage from prolonged operation at reduced voltage. The combination of the loss of voltage relaying and the degraded grid relaying provides protection to the Class 1E distribution system for all credible conditions of voltage collapse or sustained voltage degradation. The initiation of the DGs on loss of offsite power, and subsequent initiation of the ECCS, ensure that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46.
 
Accident analyses credit the loading of the DG based on the loss of offsite power during a loss of coolant accident.
The diesel starting and loading times have been included in the delay time associated with each safety system component requiring DG supplied power following a loss of offsite power.
 
The LOP instrumentation satisfies Criterion 3 of the NRC Policy Statement.
 
The OPERABILITY of the LOP instrumentation is dependent upon the OPERABILITY of the individual instrumentation relay channel Functions specified in Table 3.3.8.1-1. Each Function must have a required number of OPERABLE channels per 4 kV emergency bus, with their setpoints within the specified Allowable Values except the bus undervoltage relay which does not have an Allowable Value. A degraded voltage channel is inoperable if its actual trip setpoint is not within its required Allowable Value. Setpoints are calibrated consistent with the Improved Instrument Setpoint Control Program (IISCP) methodology assumptions.
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                        B 3.3-188                                                                                                                                                                                          Revision No. 88 LOP Instrumentation B 3.3.8.1
 
BASES
 
APPLICABLE                                                                      The loss of voltage channel is inoperable if it will not SAFETY ANALYSES,          start the diesel on a loss of power to a 4 kV emergency bus.
LCO, and APPLICABILITY                                        The Allowable Values are specified for each applicable (continued)                                        Function in the Table 3.3.8.1-1. The nominal setpoints are selected to ensure that the setpoints do not exceed the Allowable Value between CHANNEL CALIBRATIONS. Operation with a trip setpoint within the Allowable Value, is acceptable. Trip setpoints are those predetermined values of output at which an action should take place. The setpoints are compared to the actual process parameter (e.g., voltage), and when the measured output value of the process parameter exceeds the setpoint, the protective relay output changes state. The Allowable Values were set equal to the limiting values determined by the voltage regulation calculation. The setpoints were corrected using IISCP methodology to account for relay drift, relay accuracy, potential transformer accuracy, measuring and test equipment accuracy margin, and includes a calibration leave alone zone. IISCP methodology utilizes the square root of the sum of the squares to combine random non-directional accuracy values. IISCP then includes relay drift, calibration leave alone zones, and margins. The setpoint assumes a nominal 35/1 potential transformer ratio.
 
The specific Applicable Safety Analyses, LCO, and Applicability discussions for Unit 2 LOP instrumentation are listed below on a Function by Function basis.
 
In addition, since some equipment required by Unit 2 is powered from Unit 3 sources, the Unit 3 LOP instrumentation supporting the required sources must also be OPERABLE. The OPERABILITY requirements for the Unit 3 LOP instrumentation is the same as described in this section, except Function 4 (4 kV Emergency Bus Undervoltage, Degraded Voltage LOCA) is not required to be OPERABLE, since this Function is related to a LOCA on Unit 3 only. The Unit 3 instrumentation is listed in Unit 3 Table 3.3.8.1-1.
: 1. 4 kV Emergency Bus Undervoltage (Loss of Voltage)
 
When both offsite sources are lost, a loss of voltage condition on a 4 kV emergency bus indicates that the respective emergency bus is unable to supply sufficient power for proper operation of the applicable equipment.
Therefore, the power supply to the bus is transferred from offsite power to DG power. This ensures that adequate power will be available to the required equipment.
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                        B 3.3-189                                                                                                                                                                                          Revision No. 88 LOP Instrumentation B 3.3.8.1
 
BASES
 
APPLICABLE                                                                      1. 4 kV Emergency Bus Undervoltage (Loss of Voltage)
SAFETY ANALYSIS,  (continued)
LCO, and APPLICABILITY                                        The single channel of 4 kV Emergency Bus Undervoltage (Loss of Voltage) Function per associated emergency bus is only required to be OPERABLE when the associated DG and offsite circuit are required to be OPERABLE. This ensures no single instrument failure can preclude the start of three of four DGs.  (One channel inputs to each of the four DGs.)  Refer to LCO 3.8.1, "AC Sources      Operating," and 3.8.2, "AC Sources      Shutdown," for Applicability Bases for the DGs.
 
2., 3., 4., 5. 4kV Emergency Bus Undervoltage (Degraded Voltage)
 
A degraded voltage condition on a 4 kV emergency bus indicates that, while offsite power may not be completely lost to the respective emergency bus, available power may be insufficient for starting large ECCS motors without risking damage to the motors that could disable the ECCS function.
 
Therefore, power to the bus is transferred from offsite power to onsite DG power when there is insufficient offsite power to the bus. This transfer will occur only if the voltage of the preferred and alternate power sources drop below the Degraded Voltage Function Allowable Values (degraded voltage with a time delay) and the source breakers trip which causes the bus undervoltage relay to initiate the DG. This ensures that adequate power will be available to the required equipment.
 
Four Functions are provided to monitor degraded voltage at four different levels. These Functions are the Degraded Voltage Non-LOCA, Degraded Voltage LOCA, Degraded Voltage High Setting, and Degraded Voltage Low Setting. These relays monitor the following voltage levels with the following time delays:  the Function 2 relay, 2286 - 2706 volts in approximately 2 seconds when source voltage is reduced abruptly to zero volts (inverse time delay); the Function 3 relay, 3409 - 3829 volts in approximately 30 seconds when source voltage is reduced abruptly to 2940 volts (inverse time delay); the Function 4 relay, 3766 -
3836 volts in approximately 10 seconds; and the Function 5 relay, 4116 - 4186 volts in approximately 60 seconds. The Function 2 and 3 relays are inverse time delay relays.
These relays operate along a repeatable characteristic curve. With relay operation being inverse with time, for
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                        B 3.3-190                                                                                                                                                                                          Revision No. 88 LOP Instrumentation B 3.3.8.1
 
BASES
 
APPLICABLE                                                                      2., 3., 4., 5. 4 kV Emergency Bus Undervoltage (Degraded SAFETY ANALYSES,          Voltage)  (continued)
LCO, and APPLICABILITY                                an abrupt reduction in voltage the relay operating time will be short; conversely, for a slight reduction in voltage, the operating time delay will be long.
 
The Degraded Voltage LOCA Function preserves the assumptions of the LOCA analysis and the combined Functions of the other relays preserves the assumptions of the accident sequence analysis in the UFSAR. The Degraded Voltage Non-LOCA Function provides assurance that equipment powered from the 4kV emergency buses is not damaged by degraded voltage that might occur under other than LOCA conditions. This degraded grid non-LOCA relay has an associated 60 second timer. This timer allows for offsite source transformer load tap changer operation. Degraded voltage conditions can be mitigated by tap changer operations and other manual actions. The 60 second timer provides the time for these actions to take place.
 
The degraded grid voltage Allowable Values are low enough to prevent inadvertent power supply transfer, but high enough to ensure that sufficient power is available to the required equipment. The Time Delay Allowable Values are long enough to provide time for the offsite power supply to recover to normal voltages, but short enough to ensure that sufficient power is available to the required equipment.
 
Two channels (one channel per source) of 4 kV Emergency Bus Degraded Voltage (Functions 2, 3, 4, and 5) per associated bus are required to be OPERABLE when the associated DG and offsite circuit are required to be OPERABLE. This ensures no single instrument failure can preclude the start of three of four DGs (each logic inputs to each of the four DGs). Refer to LCO 3.8.1 and LCO 3.8.2 for Applicability Bases for the DGs.
 
ACTIONS                                                                                                    A Note has been provided (Note 1) to modify the ACTIONS related to LOP instrumentation channels. 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 (continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                        B 3.3-191                                                                                                                                                                                                    Revision No. 5 LOP Instrumentation B 3.3.8.1
 
BASES
 
ACTIONS                                                                                                    entry into the Condition. However, the Required Actions for (continued)                                                  inoperable LOP instrumentation channels provide appropriate compensatory measures for separate inoperable channels. As such, a Note has been provided that allows separate Condition entry for each inoperable LOP instrumentation channel.
 
A.1
 
Pursuant to LCO 3.0.6, the AC Sources      Operating ACTIONS would not have to be entered even if the LOP instrumentation inoperability resulted in an inoperable offsite circuit.
Therefore, the Required Action of Condition A is modified by a Note to indicate that when performance of a Required Action results in the inoperability of an offsite circuit, Actions for LCO 3.8.1, "AC Sources      Operating," must be immediately entered. A Unit 2 offsite circuit is considered to be inoperable if it is not supplying or not capable of supplying (due to loss of autotransfer capability) at least three Unit 2 4 kV emergency buses when the other offsite circuit is providing power or capable of supplying power to all four Unit 2 4 kV emergency buses. A Unit 2 offsite circuit is also considered to be inoperable if the Unit 2 4 kV emergency buses being powered or capable of being powered from the two offsite circuits are all the same when at least one of the two circuits does not provide power or is not capable of supplying power to all four Unit 2 4 kV emergency buses. Inoperability of a Unit 3 offsite circuit is the same as described for a Unit 2 offsite circuit, except that the circuit path is to the Unit 3 4 kV emergency buses required to be OPERABLE by LCO 3.8.7, "Distribution Systems      Operating."  The Note allows Condition A to provide requirements for the loss of a LOP instrumentation channel without regard to whether an offsite circuit is rendered inoperable. LCO 3.8.1 provides appropriate restriction for an inoperable offsite circuit.
 
Required Action A.1 is applicable when one 4 kV emergency bus has one or two required Function 3 (Degraded Voltage High Setting) channels inoperable or when one 4 kV emergency bus has one or two required Function 5 (Degraded Voltage Non-LOCA) channels inoperable. In this Condition, the affected Function may not be capable of performing its intended function automatically for these buses. However, the operators would still receive indication in the control room of a degraded voltage condition on the unaffected buses and a manual transfer of the affected bus power supply to
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                        B 3.3-192                                                                                                                                                                                                    Revision No. 5 LOP Instrumentation B 3.3.8.1 BASES
 
ACTIONS                                                                                                    A.1  (continued)
 
the alternate source could be made without damaging plant equipment. Therefore, Required Action A.1 allows 14 days to restore the inoperable channel(s) to OPERABLE status or place the inoperable channel(s) in trip. Alternatively, a Completion Time can be determined in accordance with the Risk Informed Completion Time (RICT) Program. A Note has been provided to indicate that a RICT is only applicable when a loss of function has not occurred. Placing the inoperable channel in trip would conservatively compensate for the inoperability, restore design trip capability to the LOP instrumentation, and allow operation to continue.
Alternatively, if it is not desired to place the channel in trip (e.g., as in the case where placing the channel in trip would result in DG initiation), Condition D must be entered and its Required Action taken.
 
The 14 day Completion Time is intended to allow time to restore the channel(s) to OPERABLE status. The Completion Time takes into consideration the diversity of the Degraded Voltage Functions, the capabilities of the remaining OPERABLE LOP Instrumentation Functions on the affected 4 kV emergency bus and on the other 4 kV emergency buses (only one 4 kV emergency bus is affected by the inoperable channels), the fact that the Degraded Voltage High Setting and Degraded Voltage Non-LOCA Functions provide only a marginal increase in the protection provided by the voltage monitoring scheme, the low probability of the grid operating in the voltage band protected by these Functions, and the ability of the operators to perform the Functions manually.
Alternatively, a Completion Time can be determined in accordance with the Risk Informed Completion Time (RICT)
Program. A Note has been provided to indicate that a RICT is only applicable when a loss of function has not occurred.
 
B.1
 
Pursuant to LCO 3.0.6, the AC Sources      Operating ACTIONS would not have to be entered even if the LOP instrumentation inoperability resulted in an inoperable offsite circuit.
Therefore, the Required Action of Condition B is modified by a Note to indicate that when performance of a Required Action results in the inoperability of an offsite circuit, Actions for LCO 3.8.1, "AC Sources      Operating," must be immediately entered. A Unit 2 offsite circuit is considered to be inoperable if it is not supplying or not capable of supplying (due to loss of autotransfer capability) at least three Unit 2 4 kV emergency buses when the other offsite circuit is providing power or capable of supplying power to all four Unit 2 4 kV emergency buses. A Unit 2 offsite circuit is also considered to be inoperable if the Unit 2 4 kV emergency buses being powered or capable of being powered from the two offsite circuits are all the same when at least one of the two circuits does not provide power or
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                        B 3.3-193                                                                                                                                                                                Revision No. 159 LOP Instrumentation B 3.3.8.1
 
BASES
 
ACTIONS                                                                                                    B.1  (continued)
 
is not capable of supplying power to all four Unit 2 4 kV emergency buses. Inoperability of a Unit 3 offsite circuit is the same as described for a Unit 2 offsite circuit, except that the circuit path is to the Unit 3 4 kV emergency buses required to be OPERABLE by LCO 3.8.7, "Distribution Systems - Operating."  This allows Condition B to provide requirements for the loss of a LOP instrumentation channel without regard to whether an offsite circuit is rendered inoperable. LCO 3.8.1 provides appropriate restriction for an inoperable offsite circuit.
 
Required Action B.1 is applicable when two 4 kV emergency buses have one required Function 3 (Degraded Voltage High Setting) channel inoperable, or when two 4 kV emergency buses have one required Function 5 (Degraded Voltage Non-LOCA) channel inoperable, or when one 4 kV emergency bus has one required Function 3 channel inoperable and a different 4 kV emergency bus has one required Function 5 channel inoperable. In this Condition, the affected Function may not be capable of performing its intended function automatically for these buses. However, the operators would still receive indication in the control room of a degraded voltage condition on the unaffected buses and a manual transfer of the affected bus power supply to the alternate source could be made without damaging plant equipment.
Therefore, Required Action B.1 allows 24 hours to restore the inoperable channels to OPERABLE status or place the inoperable channels in trip. Alternatively, a Completion Time can be determined in accordance with the Risk Informed Completion Time (RICT) Program. A Note has been provided to indicate that a RICT is only applicable when a loss of function has not occurred. Placing the inoperable channel in trip would conservatively compensate for the inoperability, restore design trip capability to the LOP instrumentation, and allow operation to continue. Alternatively, if it is not desired to place the channel in trip (e.g., as in the case where placing the channel in trip would result in DG initiation), Condition D must be entered and its Required Action taken.
 
The 24 hour Completion Time is intended to allow time to restore the channel(s) to OPERABLE status. The Completion Time takes into consideration the diversity of the Degraded Voltage Functions, the capabilities of the remaining OPERABLE LOP Instrumentation Functions on the affected 4 kV emergency buses and on the other 4 kV emergency buses (only two 4 kV emergency buses are affected by the inoperable channels), the fact that the Degraded Voltage High Setting and Degraded Voltage Non-LOCA Functions provide only a
 
(continued)
 
PBAPS UNIT 2                                                                                                                                                                                                                        B 3.3-194                                                                                                                                                                                Revision No. 159 LOP Instrumentation B 3.3.8.1
 
BASES
 
ACTIONS                                                                                                    B.1  (continued)
 
marginal increase in the protection provided by the voltage monitoring scheme, the low probability of the grid operating in the voltage band protected by these Functions, and the ability of the operators to perform the Functions manually.
Alternatively, a Completion Time can be determined in accordance with the Risk Informed Completion Time (RICT)
Program. A Note has been provided to indicate that a RICT is only applicable when a loss of function has not occurred
 
C.1
 
Pursuant to LCO 3.0.6, the AC Sources      Operating ACTIONS would not have to be entered even if the LOP Instrumentation inoperability resulted in an inoperable offsite circuit.
Therefore, the Required Action of Condition C is modified by a Note to indicate that when performance of the Required Action results in the inoperability of an offsite circuit, Actions for LCO 3.8.1, "AC Sources      Operating," must be immediately entered. A Unit 2 offsite circuit is considered to be inoperable if it is not supplying or not capable of supplying (due to loss of autotransfer capability) at least three Unit 2 4 kV emergency buses when the other offsite circuit is providing power or capable of supplying power to all four Unit 2 4 kV emergency buses. A Unit 2 offsite circuit is also considered to be inoperable if the Unit 2 4 kV emergency buses being powered or capable}}

Latest revision as of 17:56, 4 October 2024

Submittal of Changes to Technical Specifications Bases
ML24107A246
Person / Time
Site: Peach Bottom  Constellation icon.png
Issue date: 04/15/2024
From: David Helker
Constellation Energy Generation
To:
Office of Nuclear Reactor Regulation, Document Control Desk
References
Download: ML24107A246 (1)
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