WBL-22-026, Periodic Submission for Changes Made to the Technical Specification Bases and Technical Requirements Manual: Difference between revisions

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{{#Wiki_filter:Post Office Box 2000, Spring City, Tennessee 37381-2000 WBL-22-026 May 11, 2022 10 CFR 50.4 10 CFR 50.71(e)
{{#Wiki_filter:}}
ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001 Watts Bar Nuclear Plant, Units 1 and 2 Facility Operating License Nos. NPF-90 and NPF-96 NRC Docket Nos. 50-390 and 50-391
 
==Subject:==
Watts Bar Nuclear Plants Unit 1 and 2 - Periodic submission for changes made to the Technical Specification Bases and Technical Requirements Manual
 
==Reference:==
TVA Letter to NRC, WBL-20-047 Watts Bar Nuclear Plant Units 1 and 2 - Periodic Submission for Changes Made to the Technical Specification Bases and Technical Requirements Manual dated September 29, 2020 (ML20273A043)
The purpose of this letter is to provide the Nuclear Regulatory Commission (NRC) with copies of changes to the Watts Bar Nuclear Plant (WBN) Units 1 and 2 Technical Specification (TS) Bases and Technical Requirement Manual (TRM). These changes have been implemented at WBN since the last periodic submission in the referenced letter.
Copies of changes to the TS Bases, revisions 167 through 182 for Unit 1 and revisions 41 through 59 for Unit 2, are provided in accordance with WBN Units 1 and 2 TS section 5.6, Technical Specifications (TS) Bases Control Program.
Copies of changes to the TRM, revisions 70 through 72 for Unit 1 and revisions 13 through 16 for Unit 2, are provided in accordance with WBN Units 1 and 2 TRM section 5.1 Technical Requirements (TR) Control Program.
The changes meet the criteria described within the above control programs for which prior NRC approval is not required.
 
ENCLOSURE 1 WBN UNIT 1 TECHNICAL SPECIFICATION BASES TABLE OF CONTENTS (32 pages)
 
TABLE OF CONTENTS TABLE OF CONTENTS ................................................................................................................................. i LIST OF TABLES        .................................................................................................................................. iv LIST OF FIGURES      .................................................................................................................................. v LIST OF ACRONYMS .................................................................................................................................. vi LIST OF EFFECTIVE PAGES ....................................................................................................................... viii 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 2.0-8 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) Tavg > 200°F ................................................ B 3.1-1 B 3.1.2              SHUTDOWN MARGIN (SDM) Tavg 200°F ................................................ B 3.1-7 B 3.1.3              Core Reactivity............................................................................................... B 3.1-11 B 3.1.4              Moderator Temperature Coefficient (MTC) .................................................. B 3.1-17 B 3.1.5              Rod Group Alignment Limits ......................................................................... B 3.1-23 B 3.1.6              Shutdown Bank Insertion Limits.................................................................... B 3.1-32 B 3.1.7              Control Bank Insertion Limits ........................................................................ B 3.1-37 B 3.1.8              Rod Position Indication .................................................................................. B 3.1-45 B 3.1.9              PHYSICS TESTS Exceptions MODE 1 ....................................................... B 3.1-52 B 3.1.10            PHYSICS TESTS Exceptions MODE 2 ....................................................... B 3.1-59 B 3.2            POWER DISTRIBUTION LIMITS .......................................................................... B 3.2-1 B 3.2.1              Heat Flux Hot Channel Factor (FQ(Z)) ......................................................... B 3.2-1 B 3.2.2              Nuclear Enthalpy Rise Hot Channel N
Factor (F_          H).......................................................................................... B 3.2-11 B 3.2.3              AXIAL FLUX DIFFERENCE (AFD) .............................................................. B 3.2-18 B 3.2.4              QUADRANT POWER TILT RATIO (QPTR) ................................................ B 3.2-23 B 3.3            INSTRUMENTATION ............................................................................................. B 3.3-1 B 3.3.1              Reactor Trip System (RTS) Instrumentation ................................................ B 3.3-1 B 3.3.2              Engineered Safety Feature Actuation System (ESFAS) Instrumentation .......................................................... B 3.3-53 B 3.3.3              Post Accident Monitoring (PAM) Instrumentation ........................................ B 3.3-101 B 3.3.4              Remote Shutdown System ........................................................................... B 3.3-118 B 3.3.5              Loss of Power (LOP) Diesel Generator (DG)
Start Instrumentation .............................................................................. B 3.3-126 B 3.3.6              Containment Vent Isolation Instrumentation ................................................ B 3.3-131 B 3.3.7              Control Room Emergency Ventilation System (CREVS) Actuation Instrumentation ...................................................... B 3.3-138 B 3.3.8              Auxiliary Building Gas Treatment System (ABGTS)
Actuation Instrumentation....................................................................... B 3.3-145 (continued)
Watts Bar-Unit 1                                                        i                                                                    Revision 162
 
TABLE OF CONTENTS (continued)
B 3.4            REACTOR COOLANT SYSTEM (RCS) ............................................................... B 3.4-1 B 3.4.1              RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits ....................................................... B 3.4-1 B 3.4.2              RCS Minimum Temperature for Criticality .................................................... B 3.4-6 B 3.4.3              RCS Pressure and Temperature (P/T) Limits .............................................. B 3.4-9 B 3.4.4              RCS Loops  MODES 1 and 2 .................................................................... B 3.4-16 B 3.4.5              RCS Loops  MODE 3 ................................................................................. B 3.4-19 B 3.4.6              RCS Loops  MODE 4 ................................................................................. B 3.4-24 B 3.4.7              RCS Loops  MODE 5, Loops Filled........................................................... B 3.4-30 B 3.4.8              RCS Loops  MODE 5, Loops Not Filled .................................................... B 3.4-35 B 3.4.9              Pressurizer ..................................................................................................... B 3.4-38 B 3.4.10            Pressurizer Safety Valves ............................................................................. B 3.4-44 B 3.4.11            Pressurizer Power Operated Relief Valves (PORVs)...................................................................................... B 3.4-46 B 3.4.12            Cold Overpressure Mitigation System (COMS) ........................................... B 3.4-52 B 3.4.13            RCS Operational LEAKAGE ......................................................................... B 3.4-65 B 3.4.14            RCS Pressure Isolation Valve (PIV) Leakage.............................................. B 3.4-70 B 3.4.15            RCS Leakage Detection Instrumentation ..................................................... B 3.4-75 B 3.4.16            RCS Specific Activity ..................................................................................... B 3.4-80 B 3.4.17            Steam Generator (SG) Tube Integrity .......................................................... B 3.4-85 B 3.5            EMERGENCY CORE COOLING SYSTEMS (ECCS) ......................................... B 3.5-1 B 3.5.1              Accumulators ................................................................................................. B 3.5-1 B 3.5.2              ECCS  Operating........................................................................................ B 3.5-9 B 3.5.3              ECCS  Shutdown ....................................................................................... B 3.5-19 B 3.5.4              Refueling Water Storage Tank (RWST) ....................................................... B 3.5-23 B 3.5.5              Seal Injection Flow......................................................................................... B 3.5-29 B 3.6            CONTAINMENT SYSTEMS .................................................................................. B 3.6-1 B 3.6.1              Containment ................................................................................................... B 3.6-1 B 3.6.2              Containment Air Locks .................................................................................. B 3.6-5 B 3.6.3              Containment Isolation Valves ....................................................................... B 3.6-11 B 3.6.4              Containment Pressure................................................................................... B 3.6-22 B 3.6.5              Containment Air Temperature....................................................................... B 3.6-25 B 3.6.6              Containment Spray Systems ........................................................................ B 3.6-28 B 3.6.7              Deleted ........................................................................................................... B 3.6-34 B 3.6.8              Hydrogen Mitigation System (HMS) ............................................................. B 3.6-35 B 3.6.9              Emergency Gas Treatment System (EGTS) ............................................... B 3.6-40 B 3.6.10            Air Return System (ARS) .............................................................................. B 3.6-44 B 3.6.11            Ice Bed ........................................................................................................... B 3.6-48 B 3.6.12            Ice Condenser Doors..................................................................................... B 3.6-56 B 3.6.13            Divider Barrier Integrity .................................................................................. B 3.6-64 B 3.6.14            Containment Recirculation Drains ................................................................ B 3.6-69 B 3.6.15            Shield Building ............................................................................................... B 3.6-73 (continued)
Watts Bar-Unit 1                                                  ii                                                      Revision 82, 94, 162
 
TABLE OF CONTENTS (continued)
B 3.7            PLANT SYSTEMS .................................................................................................. B 3.7-1 B 3.7.1              Main Steam Safety Valves (MSSVs) ............................................................ B 3.7-1 B 3.7.2              Main Steam Isolation Valves (MSIVs) .......................................................... B 3.7-7 B 3.7.3              Main Feedwater Isolation Valves (MFIVs) and Main Feedwater Regulation Valves (MFRVs) and Associated Bypass Valves .............................................................. B 3.7-12 B 3.7.4              Atmospheric Dump Valves (ADVs)............................................................... B 3.7-17 B 3.7.5              Auxiliary Feedwater (AFW) System.............................................................. B 3.7-21 B 3.7.6              Condensate Storage Tank (CST) ................................................................. B 3.7-29 B 3.7.7              Component Cooling System (CCS) .............................................................. B 3.7-33 B 3.7.8              Essential Raw Cooling Water (ERCW) System ........................................... B 3.7-38 B 3.7.9              Ultimate Heat Sink (UHS).............................................................................. B 3.7-42 B 3.7.10            Control Room Emergency Ventilation System (CREVS) ............................ B 3.7-44 B 3.7.11            Control Room Emergency Air Temperature Control System (CREATCS) .................................................................. B 3.7-51 B 3.7.12            Auxiliary Building Gas Treatment System (ABGTS).................................... B 3.7-55 B 3.7.13            Fuel Storage Pool Water Level ..................................................................... B 3.7-60 B 3.7.14            Secondary Specific Activity ........................................................................... B 3.7-63 B 3.7-15            Spent Fuel Pool Assembly Storage .............................................................. B 3.7-66 B 3.7-16            Component Cooling System (CCS) - Shutdown.......................................... B 3.7-69 B 3.7-17            Essential Raw Cooling Water (ERCW) System Shutdown ......................... B 3.7-75 B 3.7-18            Fuel Storage Pool Boron Concentration                                                                            . B 3.7-81 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 3.8-39 B 3.8.3              Diesel Fuel Oil, Lube Oil, and Starting Air .................................................... B 3.8-44 B 3.8.4              DC Sources  Operating .............................................................................. B 3.8-54 B 3.8.5              DC Sources  Shutdown .............................................................................. B 3.8-66 B 3.8.6              Battery Parameters ........................................................................................ B 3.8-71 B 3.8.7              Inverters  Operating.................................................................................... B 3.8-80 B 3.8.8              Inverters  Shutdown ................................................................................... B 3.8-84 B 3.8.9              Distribution Systems  Operating ................................................................ B 3.8-87 B 3.8.10            Distribution Systems  Shutdown ................................................................ B 3.8-97 B 3.9            REFUELING OPERATIONS .................................................................................. B 3.9-1 B 3.9.1              Boron Concentration...................................................................................... B 3.9-1 B 3.9.2              Unborated Water Source Isolation Valves ................................................... B 3.9-4 B 3.9.3              Nuclear Instrumentation ................................................................................ B 3.9-7 B 3.9.4              Deleted ........................................................................................................... B 3.9-10 B 3.9.5              Residual Heat Removal (RHR) and Coolant Circulation  High Water Level ............................................................. B 3.9-11 B 3.9.6              Residual Heat Removal (RHR) and Coolant Circulation  Low Water Level .............................................................. B 3.9-15 B 3.9.7              Refueling Cavity Water Level ........................................................................ B 3.9-18 B 3.9.8              Deleted ........................................................................................................... B 3.9-21 B 3.9.9              Spent Fuel Pool Boron Concentration .......................................................... B 3.9-22 B 3.9.10            Decay Time .................................................................................................... B 3.9-24 Watts Bar-Unit 1                                                iii                                                  Revision 150, 162, 167
 
LIST OF TABLES Table No.        Title Page                                                                                              Page B 3.3.4-1        Remote Shutdown System Instrumentation and Controls                                                .. B 3.3-124 B 3.8.1-2        TS Action or Surveillance Requirement (SR)
Contingency Actions ....................................................................... B 3.8-37 B 3.8.9-1        AC and DC Electrical Power Distribution Systems............................................................................................ B 3.8-96 Watts Bar-Unit 1                                        iv                                                            Revision 150, 162
 
LIST OF FIGURES Figure No.      Title                                                                                                      Page B 2.1.1-1        Reactor Core Safety Limits vs Boundary of Protection ....................................................................................................... B 2.0-7 B 3.1.7-1        Control Bank Insertion vs Percent RTP ................................................................. B 3.1-44 B 3.2.1-1        K(z) - Normalized FQ(z) as a Function of Core Height ............................................................................................................. B 3.2-10 B 3.2.3-1        AXIAL FLUX DIFFERENCE Acceptable Operation Limits as a Function of RATED THERMAL POWER ............................................. B 3.2-22 Watts Bar-Unit 1                                                    v                                                                    Revision 162
 
LIST OF ACRONYMS (Page 1 of 2)
Acronym          Title ABGTS            Auxiliary Building Gas Treatment System ACRP            Auxiliary Control Room Panel ASME            American Society of Mechanical Engineers AFD              Axial Flux Difference AFW              Auxiliary Feedwater System ARO              All Rods Out ARFS            Air Return Fan System ADV              Atmospheric Dump Valve BOC              Beginning of Cycle CAOC            Constant Axial Offset Control CCS              Component Cooling System CFR              Code of Federal Regulations COLR            Core Operating Limits Report CREVS            Control Room Emergency Ventilation System CSS              Containment Spray System CST              Condensate Storage Tank DNB              Departure from Nucleate Boiling ECCS            Emergency Core Cooling System EFPD            Effective Full-Power Days EGTS            Emergency Gas Treatment System EOC              End of Cycle ERCW            Essential Raw Cooling Water ESF              Engineered Safety Feature ESFAS            Engineered Safety Features Actuation System HEPA            High Efficiency Particulate Air HVAC            Heating, Ventilating, and Air-Conditioning LCO              Limiting Condition For Operation MFIV            Main Feedwater Isolation Valve MFRV            Main Feedwater Regulation Valve MSIV            Main Steam Line Isolation Valve MSSV            Main Steam Safety Valve MTC              Moderator Temperature Coefficient NMS              Neutron Monitoring System ODCM            Offsite Dose Calculation Manual PCP              Process Control Program PDMS            Power Distribution Monitoring System PIV              Pressure Isolation Valve PORV            Power-Operated Relief Valve PTLR            Pressure and Temperature Limits Report QPTR            Quadrant Power Tilt Ratio RAOC            Relaxed Axial Offset Control RCCA            Rod Cluster Control Assembly RCP              Reactor Coolant Pump RCS              Reactor Coolant System RHR              Residual Heat Removal RTP              Rated Thermal Power Watts Bar-Unit 1                            vi              Revision 104
 
LIST OF ACRONYMS (Page 2 of 2)
Acronym          Title RTS              Reactor Trip System RWST            Refueling Water Storage Tank SG              Steam Generator SI              Safety Injection SL              Safety Limit SR              Surveillance Requirement UHS              Ultimate Heat Sink Watts Bar-Unit 1                          vii
 
TECHNICAL SPECIFICATIONS BASES LIST OF EFFECTIVE PAGES Page    Revision              Page  Revision Number:    Number:              Number:  Number:
i      162              B 2.0-1      0 ii      162                B 2.0-2  140 iii      167                B 2.0-3    0 iv      162                B 2.0-4    59 v      162                B 2.0-5  108 vi      104              B 2.0-6    140 vii        0                B 2.0-7    0 viii      182                B 2.0-8    0 ix      176                B 2.0-9    0 x      177              B 2.0-10  108 xi      171              B 2.0-11    0 xii      178                B 3.0-1  133 xiii      179                B 3.0-2  159 xiv      182                B 3.0-3  159 xv      181                B 3.0-4    68 xvi        19              B 3.0-5    68 xvii      32                B 3.0-6    68 xviii      46              B 3.0-7    0 xix        60              B 3.0-8    141 xx        68              B 3.0-9    133 xxi      75                B 3.0-10  133 xxii      85                B 3.0-11  133 xxiii      94              B 3.0-12    0 xxiv      102              B 3.0-13  165 xxv      110              B 3.0-14  159 xxvi      119              B 3.0-15    68 xxvii      127              B 3.0-16    68 xxviii      141                B 3.1-1    0 xxix      154                B 3.1-2    0 xxx      166                B 3.1-3    0 xxxi      177                B 3.1-4    68 xxxii      182                B 3.1-5    0 B 3.1-6  162 B 3.1-7    0 B 3.1-8    0 B 3.1-9    68 B 3.1-10  162 B 3.1-11    0 B 3.1-12    32 B 3.1-13    0 B 3.1-14    0 B 3.1-15    0 Watts Bar-Unit 1                  viii                    Revision 182
 
TECHNICAL SPECIFICATIONS BASES LIST OF EFFECTIVE PAGES Page    Revision              Page  Revision Number:    Number:              Number:  Number:
B 3.1-16      162              B 3.1-61    39 B 3.1-17      32              B 3.1-62    0 B 3.1-18      32              B 3.1-63  162 B 3.1-19      32              B 3.1-64    0 B 3.1-20      32                B 3.2-1  175 B 3.1-21      32                B 3.2-2  176 B 3.1-22        0                B 3.2-3  175 B 3.1-23      143                B 3.2-4  175 B 3.1-24      143                B 3.2-5  175 B 3.1-25        0                B 3.2-6  175 B 3.1-26      143                B 3.2-7  175 B 3.1-27      143                B 3.2-8  175 B 3.1-28      143                B 3.2-9  175 B 3.1-29      143              B3.2-9a    175 B 3.1-30      162              B3.2-9b    175 B 3.1-31      162              B3.2-9c    162 B 3.1-32      51              B3.2-9d    175 B 3.1-33      143              B3.2-9e    175 B 3.1-34      143              B 3.2-10  175 B 3.1-35      143              B 3.2-11  162 B 3.1-36      162              B 3.2-12  176 B 3.1-37      51              B 3.2-13  176 B 3.1-38      143              B 3.2-14    0 B 3.1-39      143              B 3.2-15  104 B 3.1-40      143              B 3.2-16  104 B 3.1-41      143              B 3.2-17  162 B 3.1-42      162              B 3.2-18    0 B 3.1-43      162              B 3.2-19    0 B 3.1-44      0                B 3.2-20  162 B 3.1-45      143              B 3.2-21    0 B 3.1-46      143              B 3.2-22    0 B 3.1-47      143              B 3.2-23  176 B 3.1-48      143              B 3.2-24    0 B 3.1-49      143              B 3.2-25  104 B 3.1-50      143              B 3.2-26    0 B 3.1-51      104              B 3.2-27  162 B 3.1-52      0                B 3.2-28  145 B 3.1-53      40              B 3.2-29  104 B 3.1-54      40                B 3.3-1    0 B 3.1-55      0                B 3.3-2    0 B 3.1-56      0                B 3.3-3    0 B 3.1-57      162                B 3.3-4    60 B 3.1-58      162                B 3.3-5    60 B 3.1-59      40                B 3.3-6    0 B 3.1-60      40                B 3.3-7    0 B 3.3-8    0 B 3.3-9    0 Watts Bar-Unit 1                    ix                    Revision 176
 
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B 3.3-10      27              B 3.3-59    0 B 3.3-11    149              B 3.3-60    0 B 3.3-12      0                B 3.3-61    0 B 3.3-13    149              B 3.3-62    0 B 3.3-14      13              B 3.3-63    0 B 3.3-15      0                B 3.3-64    0 B 3.3-16      13              B 3.3-65    9 B 3.3-17      0                B 3.3-66    9 B 3.3-18      0                B 3.3-67    0 B 3.3-19      90              B 3.3-68    0 B 3.3-20      90              B 3.3-69    0 B 3.3-21      90              B 3.3-70    0 B 3.3-22      13              B 3.3-71    0 B 3.3-23    146              B 3.3-72    0 B 3.3-24      0                B 3.3-73    0 B 3.3-25      0                B 3.3-74    0 B 3.3-26      90              B 3.3-75    13 B 3.3-27      90              B 3.3-76  177 B 3.3-28      13              B 3.3-77  177 B 3.3-29      0                B 3.3-78  177 B 3.3-30      0                B 3.3-79    0 B 3.3-31      0                B 3.3-80    0 B 3.3-32      0                B 3.3-81    0 B 3.3-33    145              B 3.3-82    0 B 3.3-34      0                B 3.3-83    0 B 3.3-35      0                B 3.3-84  123 B 3.3-36      0                B 3.3-85    90 B 3.3-37      90              B 3.3-86    90 B 3.3-38      90              B 3.3-87  123 B 3.3-39      90              B 3.3-88  123 B 3.3-40      90              B 3.3-89    96 B 3.3-41      90              B 3.3-90    96 B 3.3-42      90              B 3.3-91    90 B 3.3-43      90              B 3.3-92    90 B 3.3-44    162              B 3.3-93  177 B 3.3-45    162              B 3.3-94  162 B 3.3-46    162              B 3.3-95  162 B 3.3-47    162              B 3.3-96  162 B 3.3-48    162              B 3.3-97  34 B 3.3-49    162              B 3.3-98  177 B 3.3-50      90              B 3.3-99    34 B 3.3-51    170              B 3.3-100  176 B 3.3-52    170              B 3.3-101    0 B 3.3-53      0              B 3.3-102    0 B 3.3-54      0              B 3.3-103    0 B 3.3-55      0              B 3.3-104    0 B 3.3-56      0              B 3.3-105    0 B 3.3-57    176 B 3.3-58      0 Watts Bar-Unit 1                    x                    Revision 177
 
TECHNICAL SPECIFICATIONS BASES LIST OF EFFECTIVE PAGES Page      Revision              Page  Revision Number:    Number:              Number:  Number:
B 3.3-106        0                B 3.4-1    0 B 3.3-107        0                B 3.4-2    60 B 3.3-108      171                B 3.4-3    60 B 3.3-109        0                B 3.4-4  162 B 3.3-110      94                B 3.4-5    60 B 3.3-111      171                B 3.4-6    0 B 3.3-112      171                B 3.4-7    55 B 3.3-113      94                B 3.4-8  162 B 3.3-114      94                B 3.4-9    0 B 3.3-115      162              B 3.4-10    0 B 3.3-116      162              B 3.4-11    0 B 3.3-117        0              B 3.4-12    0 B 3.3-118        0              B 3.4-13    0 B 3.3-119      150              B 3.4-14  162 B 3.3-120      150              B 3.4-15    0 B 3.3-121      162              B 3.4-16    0 B 3.3-122      162              B 3.4-17    82 B 3.3-123      162              B 3.4-18  162 B 3.3-124      150              B 3.4-19    0 B 3.3-125      156              B 3.4-20    0 B 3.3-126        0              B 3.4-21    79 B 3.3-127      156              B 3.4-22    0 B 3.3-128      156              B 3.4-23  162 B 3.3-129      164              B 3.4-24  123 B 3.3-130      162              B 3.4-25  123 B 3.3-131      119              B 3.4-26  123 B 3.3-132        9              B 3.4-27  123 B 3.3-133      119              B 3.4-28  162 B 3.3-134      119              B 3.4-29  162 B 3.3-135      162              B 3.4-30    79 B 3.3-136      162              B 3.4-31    79 B 3.3-137      162              B 3.4-32    82 B 3.3-138        0              B 3.4-33  162 B 3.3-139      45                B 3.4-34    29 B 3.3-140        0              B 3.4-35    0 B 3.3-141        0              B 3.4-36    68 B 3.3-142      45                B 3.4-37  162 B 3.3-143      162              B 3.4-38    0 B 3.3-144      162              B 3.4-39    0 B 3.3-145      119              B 3.4-40    0 B 3.3-146      119              B 3.4-41  162 B 3.3-147      119              B 3.4-42    0 B 3.3-148      162              B 3.4-43    0 B 3.4-44    0 Watts Bar-Unit 1                    xi                    Revision 171
 
TECHNICAL SPECIFICATIONS BASES LIST OF EFFECTIVE PAGES Page    Revision              Page  Revision Number:    Number:              Number:  Number:
B 3.4-45      89              B 3.4-86    82 B.3.4-46        0              B 3.4-87  174 B 3.4-47      42              B 3.4-88    82 B 3.4-48      68              B 3.4-89  174 B 3.4-49      42              B 3.4-90  174 B 3.4-50        0              B 3.4-91  174 B 3.4-51      162                B 3.5-1    0 B 3.4-52        0                B 3.5-2  176 B 3.4-53        0                B 3.5-3  176 B 3.4-54        0                B 3.5-4  176 B 3.4-55        0                B 3.5-5    98 B 3.4-56        0                B 3.5-6  162 B 3.4-57        0                B 3.5-7  162 B 3.4-58      68                B 3.5-8    98 B 3.4-59        0                B 3.5-9    61 B 3.4-60      68              B 3.5-10    0 B 3.4-61        0              B 3.5-11  176 B 3.4-62      162              B 3.5-12    39 B 3.4-63      162              B 3.5-13    68 B 3.4-64      89              B 3.5-14    68 B 3.4-65      82              B 3.5-15    0 B 3.4-66      82              B 3.5-16  165 B 3.4-67      82              B 3.5-17  162 B 3.4-68      82              B 3.5-18  165 B 3.4-69      162              B 3.5-19    0 B.3.4-70        0              B 3.5-20    68 B 3.4-71        0              B 3.5-21    0 B 3.4-72        0              B 3.5-22    0 B 3.4-73      162              B 3.5-23    0 B 3.4-74      162              B 3.5-24    0 B 3.4-75      92              B 3.5-25  176 B 3.4-76      12              B 3.5-26    0 B 3.4-77      92              B 3.5-27  162 B 3.4-78      92              B 3.5-28  176 B 3.4-79      162              B 3.5-29    0 B 3.4-80      173              B 3.5-30    0 B 3.4-81      173              B 3.5-31  162 B 3.4-82      173                B 3.6-1    10 B 3.4-83      173                B 3.6-2    10 B 3.4-84      173                B 3.6-3    10 B 3.4-85      173                B 3.6-4  178 B 3.6-5    10 Watts Bar-Unit 1                  xii                    Revision 178
 
TECHNICAL SPECIFICATIONS BASES LIST OF EFFECTIVE PAGES Page    Revision              Page  Revision Number:    Number:              Number:  Number:
B 3.6-6      178              B 3.6-51  162 B 3.6-7      0                B 3.6-52  162 B 3.6-8      0                B 3.6-53  162 B 3.6-9      0                B 3.6-54  162 B 3.6-10      162              B 3.6-55  176 B 3.6-11      0                B 3.6-56    0 B 3.6-12      0                B 3.6-57    0 B 3.6-13      130              B 3.6-58  176 B 3.6-14      98              B 3.6-59    0 B 3.6-15      98              B 3.6-60    36 B 3.6-16      98              B 3.6-61  165 B 3.6-17      0                B 3.6-62  165 B 3.6-18      162              B 3.6-63  176 B 3.6-19      162              B 3.6-64    0 B 3.6-20      178              B 3.6-65    0 B 3.6-21      178              B 3.6-66    0 B 3.6-22      178              B 3.6-67  162 B 3.6-23      176              B 3.6-68  162 B 3.6-24      176              B 3.6-69    0 B 3.6-25      0                B 3.6-70    0 B 3.6-26      29              B 3.6-71    0 B 3.6-27      162              B 3.6-72  162 B 3.6-28      0                B 3.6-73  129 B 3.6-29      0                B 3.6-74  129 B 3.6-30      176              B 3.6-75  169 B 3.6-31      0                B 3.6-76  179 B 3.6-32      162                B 3.7-1    31 B 3.6-33      176                B 3.7-2    31 B 3.6-34      94                B 3.7-3    41 B 3.6-35      0                B 3.7-4  121 B 3.6-36      128                B 3.7-5  121 B 3.6-37      0                B 3.7-6    89 B 3.6-38      162                B 3.7-7    0 B 3.6-39      165                B 3.7-8    0 B 3.6-40      138                B 3.7-9    0 B 3.6-41      71              B 3.7-10    0 B 3.6-42      162              B 3.7-11  162 B 3.6-43      168              B 3.7-12    76 B3.6-43a      101              B 3.7-13    0 B 3.6-44      0                B 3.7-14    0 B 3.6-45      176              B 3.7-15    89 B 3.6-46      0                B 3.7-16  162 B 3.6-47      176              B 3.7-17    0 B 3.6-48      81              B 3.7-18    0 B 3.6-49      0                B.3.7-19    68 B 3.6-50      176              B 3.7-20  162 Watts Bar-Unit 1                  xiii                    Revision 179
 
TECHNICAL SPECIFICATIONS BASES LIST OF EFFECTIVE PAGES Page      Revision              Page  Revision Number:    Number:              Number:  Number:
B 3.7-21        0              B 3.7-67  167 B 3.7-22      147              B 3.7-68  167 B 3.7-23        0              B 3.7-69  123 B 3.7-24        68              B 3.7-70  123 B 3.7-25        0              B 3.7-71  123 B 3.7-26      162              B 3.7-72  123 B 3.7-27      162              B 3.7-73  123 B 3.7-28        89              B 3.7-74  162 B 3.7-29        0              B 3.7-75  123 B 3.7-30        41              B 3.7-76  123 B 3.7-31      162              B 3.7-77  123 B 3.7-32        0              B 3.7-78  123 B 3.7-33      136              B 3.7-79  162 B 3.7-34        0              B 3.7-80  123 B 3.7-35      136              B 3.7-81  167 B 3.7-36      162              B 3.7-82  167 B 3.7-37      162                B 3.8-1  166 B 3.7-38        0                B 3.8-2  166 B 3.7-39        0                B 3.8-3  180 B 3.7-40      162                B 3.8-4  166 B 3.7-41      162                B 3.8-5  166 B 3.7-42        0                B 3.8-6  180 B 3.7-43      162                B 3.8-7  158 B 3.7-44        91                B 3.8-8    0 B 3.7-45        91                B 3.8-9  132 B 3.7-46        91              B 3.8-10  180 B 3.7-47        91              B 3.8-11  132 B 3.7-48      122              B 3.8-12  132 B 3.7-49      168              B 3.8-13  132 B 3.7-50        91              B 3.8-14  180 B 3.7-51        64              B 3.8-15  158 B 3.7-52        64              B 3.8-16  158 B 3.7-53      172              B 3.8-17  180 B 3.7-54      172              B 3.8-18  158 B 3.7-54a      172              B 3.8-19  162 B 3.7-55      119              B 3.8-20  162 B 3.7-56      182              B 3.8-21  162 B 3.7-57      139              B 3.8-22  162 B 3.7-58      168              B 3.8-23  166 B 3.7-59      162              B 3.8-24  162 B 3.7-60      119              B 3.8-25  162 B 3.7-61      162              B 3.8-26  162 B 3.7-62      119              B 3.8-27  162 B 3.7-63        47              B 3.8-28  162 B 3.7-64        0              B 3.8-29  162 B 3.7-65      162              B 3.8-30  162 B 3.7-66      167 Watts Bar-Unit 1                    xiv                    Revision 182
 
TECHNICAL SPECIFICATIONS BASES LIST OF EFFECTIVE PAGES Page    Revision              Page    Revision Number:    Number:              Number:  Number:
B 3.8-31      162              B 3.8-79  162 B 3.8-32      162              B 3.8-80    97 B 3.8-33      162              B 3.8-81    97 B 3.8-34      162              B 3.8-82    97 B 3.8-35      162              B 3.8-83  162 B 3.8-36      166              B 3.8-84    0 B 3.8-37      180              B 3.8-85    97 B 3.8-38      132              B 3.8-86  162 B 3.8-39        0              B 3.8-87  152 B 3.8-40        0              B 3.8-87a  163 B 3.8-41        0              B 3.8-88    0 B 3.8-42        0              B 3.8-89  158 B 3.8-43        0              B 3.8-90  152 B 3.8-44        0              B 3.8-91    0 B 3.8-45        0              B 3.8-92  152 B 3.8-46        0              B 3.8-93  152 B 3.8-47      55              B 3.8-94  158 B 3.8-48      55              B 3.8-95  158 B 3.8-49      162              B 3.8-96  162 B 3.8-50        0              B 3.8-97  181 B 3.8-51      162              B 3.8-98    0 B 3.8-52      162              B 3.8-99    0 B 3.8-53      29              B 3.8-100    0 B 3.8-54      160              B 3.8-101  162 B 3.8-55      160              B 3.9-1    0 B 3.8-56      160              B 3.9-2    68 B 3.8-57      160              B 3.9-3    162 B 3.8-58      160              B 3.9-4    68 B 3.8-59      160              B 3.9-5    0 B 3.8-60      160              B 3.9-6    162 B 3.8-61      160              B 3.9-7    0 B 3.8-62      162              B 3.9-8    0 B 3.8-63      162              B 3.9-9    162 B 3.8-64      162              B 3.9-10  119 B 3.8-65      160              B 3.9-11    0 B 3.8-66      0                B 3.9-12    23 B 3.8-67      160              B 3.9-13    0 B 3.8-68      160              B 3.9-14  162 B 3.8-69      160              B 3.9-15    0 B 3.8-70        0              B 3.9-16    68 B 3.8-71      160              B 3.9-17  162 B 3.8-72      160              B 3.9-18  119 B 3.8-73      160              B 3.9-19    45 B 3.8-74      160              B 3.9-20  162 B 3.8-75      160              B 3.9-21  119 B 3.8-76      160              B 3.9-22  167 B 3.8-77      162              B 3.9-23  167 B 3.8-78      162              B 3.9-24  119 B 3.9-25  119 Watts Bar-Unit 1                  xv                      Revision 181
 
TECHNICAL SPECIFICATION BASES - REVISION LISTING (This listing is an administrative tool maintained by WBN Licensing and may be updated without formally revising the Technical Specification Bases Table-of-Contents)
REVISIONS                                  ISSUED                                SUBJECT NPF-20                                                  11-09-95            Low Power Operating License Revision 1                                              12-08-95            Slave Relay Testing NPF-90                                                  02-07-96            Full Power Operating License Revision 2 (Amendment 1)                                12-08-95            Turbine Driven AFW Pump Suction Requirement Revision 3                                              03-27-96            Remove Cold Leg Accumulator Alarm Setpoints Revision 4 (Amendment 2)                                06-13-96            Ice Bed Surveillance Frequency And Weight Revision 5                                              07-03-96            Containment Airlock Door Indication Revision 6 (Amendment 3)                                09-09-96            Ice Condenser Lower Inlet Door Surveillance Revision 7                                              09-28-96            Clarification of COT Frequency for COMS Revision 8                                              11-21-96            Admin Control of Containment Isol. Valves Revision 9                                              04-29-97            Switch Controls For Manual CI-Phase A Revision 10 (Amendment 5)                              05-27-97            Appendix-J, Option B Revision 11 (Amendment 6)                              07-28-97            Spent Fuel Pool Rerack Revision 12                                            09-10-97            Heat Trace for Radiation Monitors Revision 13 (Amendment 7)                              09-11-97            Cycle 2 Core Reload Revision 14                                            10-10-97            Hot Leg Recirculation Timeframe Revision 15                                            02-12-98            EGTS Logic Testing Revision 16 (Amendment 10)                              06-09-98            Hydrogen Mitigation System Temporary Specification Revision 17                                              07-31-98          SR Detectors (Visual/audible indication)
Revision 18 (Amendment 11)                              09-09-98          Relocation of F(Q) Penalty to COLR Revision 19 (Amendment 12)                              10-19-98            Online Testing of the Diesel Batteries and Performance of the 24 Hour Diesel Endurance Run Watts Bar-Unit 1                                              xvi                                          Revision 19
 
TECHNICAL SPECIFICATION BASES - REVISION LISTING (This listing is an administrative tool maintained by WBN Licensing and may be updated without formally revising the Technical Specification Bases Table-of-Contents)
REVISIONS                                  ISSUED                                SUBJECT Revision 20 (Amendment 13)                              10-26-98            Clarification of Surveillance Testing Requirements for TDAFW Pump Revision 21                                            11-30-98            Clarification to Ice Condenser Door ACTIONS and door lift tests, and Ice Bed sampling and flow blockage SRs Revision 22 (Amendment 14)                              11-10-98            COMS - Four Hour Allowance to Make RHR Suction Relief Valve Operable Revision 23                                            01-05-99            RHR Pump Alignment for Refueling Operations Revision 24 (Amendment 16)                              12-17-98            New action for Steam Generator ADVs due to Inoperable ACAS.
Revision 25                                            02-08-99            Delete Reference to PORV Testing Not Performed in Lower Modes Revision 26 (Amendment 17)                              12-30-98            Slave Relay Surveillance Frequency Extension to 18 Months Revision 27 (Amendment 18)                              01-15-99            Deletion of Power Range Neutron Flux High Negative Rate Reactor Trip Function Revision 28                                            04-02-99            P2500 replacement with Integrated Computer System (ICS). Delete Reference to ERFDS as a redundant input signal.
Revision 29                                            03-13-00            Added notes to address instrument error in various parameters shown in the Bases.
Also corrected the applicable modes for TS 3.6.5 from 3 and 4 to 2, 3 and 4.
Revision 30 (Amendment 23)                              03-22-00            For SR 3.3.2.10, Table 3.3.2-1, one time relief from turbine trip response time testing. Also added Reference 14 to the Bases for LCO 3.3.2.
Revision 31 (Amendment 19)                              03-07-00            Reset Power Range High Flux Reactor Trip Setpoints for Multiple Inoperable MSSVs.
Revision 32                                            04-13-00            Clarification to Reflect Core Reactivity and MTC Behavior.
Watts Bar-Unit 1                                            xvii                                            Revision 32
 
TECHNICAL SPECIFICATION BASES - REVISION LISTING (This listing is an administrative tool maintained by WBN Licensing and may be updated without formally revising the Technical Specification Bases Table-of-Contents)
REVISIONS                                  ISSUED                                SUBJECT Revision 33                                            05-02-00            Clarification identifying four distribution boards primarily used for operational convenience.
Revision 34 (Amendment 24)                              07-07-00            Elimination of Response Time Testing Revision 35                                            08-14-00            Clarification of ABGTS Surveillance Testing Revision 36 (Amendments 22 and 25)                      08-23-00            Revision of Ice Condenser sampling and flow channel surveillance requirements Revision 37 (Amendment 26)                              09-08-00            Administrative Controls for Open Penetrations During Refueling Operations Revision 38                                            09-17-00            SR 3.2.1.2 was revised to reflect the area of the core that will be flux mapped.
Revision 39 (Amendments 21and 28)                      09-13-00            Amendment 21 - Implementation of Best Estimate LOCA analysis.
Amendment 28 - Revision of LCO 3.1.10, Physics Tests Exceptions - Mode 2.
Revision 40                                            09-28-00            Clarifies WBNs compliance with ANSI/ANS-19.6.1 and deletes the detailed descriptions of Physics Tests.
Revision 41 (Amendment 31)                              01-22-01            Power Uprate from 3411 MWt to 3459 MWt Using Leading Edge Flow Meter (LEFM)
Revision 42                                            03-07-01            Clarify Operability Requirements for Pressurizer PORVs Revision 43                                            05-29-01            Change CVI Response Time from 5 to 6 Seconds Revision 44 (Amendment 33)                              01-31-02            Ice weight reduction from 1236 to 1110 lbs per basket and peak containment pressure revision from 11.21 to 10.46 psig.
Revision 45 (Amendment 35)                              02-12-02            Relaxation of CORE ALTERATIONS Restrictions Revision 46                                            02-25-02            Clarify Equivalent Isolation Requirements in LCO 3.9.4 Watts Bar-Unit 1                                            xviii                                            Revision 46
 
TECHNICAL SPECIFICATION BASES - REVISION LISTING (This listing is an administrative tool maintained by WBN Licensing and may be updated without formally revising the Technical Specification Bases Table-of-Contents)
REVISIONS                                  ISSUED                                SUBJECT Revision 47 (Amendment 38)                              03-01-02            RCS operational LEAKAGE and SG Alternate Repair Criteria for Axial Outside Diameter Stress Corrosion Cracking (ODSCC)
Revision 48 (Amendment 36)                              03-06-02            Increase Degraded Voltage Time Delay from 6 to 10 seconds.
Revision 49 (Amendment 34)                              03-08-02            Deletion of the Post-Accident Sampling System (PASS) requirements from Section 5.7.2.6 of the Technical Specifications.
Revision 50 (Amendment 39)                              08-30-02            Extension of the allowed outage time (AOT) for a single diesel generator from 72 hours to 14 days.
Revision 51                                            11-14-02            Clarify that Shutdown Banks C and D have only One Rod Group Revision 52 (Amendment 41)                              12-20-02            RCS Specific Activity Level reduction from
                                                                                <1.0 Ci/gm to <0.265 Ci/gm.
Revision 53 (Amendment 42)                              01-24-03            Revise SR 3.0.3 for Missed Surveillances Revision 54 (Amendment 43)                              05-01-03            Exigent TS SR 3.5.2.3 to delete SI Hot Leg Injection lines from SR until U1C5 outage.
Revision 55                                            05-22-03            Editorial corrections (PER 02-015499),
correct peak containment pressure, and revise I-131 gap inventory for an FHA.
Revision 56                                            07-10-03            TS Bases for SRs 3.8.4.8 through SR 3.8.4.10 clarification of inter-tier connection resistance test.
Revision 57                                            08-11-03            TS Bases for B 3.5.2 Background information provides clarification when the 9 hrs for hot leg recirculation is initiated.
Revision 58 (Amendment 45)                              09-26-03            The Bases for LCO 3.8.7 and 3.8.8 were revised to delete the Unit 2 Inverters.
Revision 59 (Amendment 46)                              09-30-03            Address new DNB Correlation in B2.1.1 and B3.2.12 for Robust Fuel Assembly (RFA)-2.
Revision 60 (Amendment 47)                              10-06-03            RCS Flow Measurement Using Elbow Tap Flow Meters (Revise Table 3.3.1-1(10) &
SR 3.4.1.4).
Watts Bar-Unit 1                                              xix                                              Revision 60
 
TECHNICAL SPECIFICATION BASES - REVISION LISTING (This listing is an administrative tool maintained by WBN Licensing and may be updated without formally revising the Technical Specification Bases Table-of-Contents)
REVISIONS                                  ISSUED                                SUBJECT Revision 61 (Amendments 40 and 48)                      10-14-03            Incorporated changes required to implement the Tritium Program (Amendment 40) and Stepped Boron Concentration increases for RWST and CLAs (Amendment 48) depending on the number of TPBARS installed into the reactor core.
Revision 62                                            10-15-03            Clarified ECCS venting in Bases Section B 3.5.2 (WBN-TS-03-19)
Revision 63                                            12-08-03            The contingency actions listed in Bases Table 3.8.1-2 were reworded to be consistent with the NRC Safety Evaluation that approved Tech Spec Amendment 39.
Revision 64 (Amendment 50)                              03-23-04            Incorporated Amendment 50 for the seismic qualification of the Main Control Room duct work. Amendment 50 revised the Bases for LCO 3.7.10, CREVS, and LCO 3.7.11, CREATCS. An editorial correction was made on Page B 3.7-61.
Revision 65                                            04-01-04            Revised the Bases for Action B.3.1 of LCO 3.8.1 to clarify that a common cause assessment is not required when a diesel generator is made inoperable due to the performance of a surveillance.
Revision 66                                            05-21-04            Revised Page B 3.8-64 (Bases for LCO 3.8.4) to add a reference to SR 3.8.4.13 that was inadvertently deleted by the changes made for Amendment 12.
Revision 67 (Amendment 45)                              03-05-05            Revised the Bases for LCOs 3.8.7, 3.8.8 and 3.8.9 to incorporate changes to the Vital Inverters (DCN 51370). Refer to the changes made for Bases Revision 58 (Amendment 45)
Revision 68 (Amendment 55)                              03-22-05            Amendment 55 modified the requirements for mode change limitations in LCO 3.0.4 and SR 3.0.4 by incorporating TSTF-359, Revision 9.
Watts Bar-Unit 1                                              xx                                            Revision 68
 
TECHNICAL SPECIFICATION BASES - REVISION LISTING (This listing is an administrative tool maintained by WBN Licensing and may be updated without formally revising the Technical Specification Bases Table-of-Contents)
REVISIONS                                  ISSUED                                SUBJECT Revision 68 (Amendment 55 and 56)                      03-22-05            Change MSLB primary to secondary leakage from 1 gpm to 3 gpm (WBN-TS-03-14).
Revision 69 (Amendment 54)                              04-04-05            Revised the use of the terms inter-tier and inter-rack in the Bases for SR 3.8.4.10.
Revision 70 (Amendment 58)                              10-17-05            Alternate monitoring process for a failed Rod Position Indicator (RPI) (TS-03-12).
Revision 71 (Amendment 59)                              02-01-06            Temporary Use of Penetrations in Shield Building Dome During Modes 1-4 (WBN-TS-04-17)
Revision 72                                            08-31-06            Minor Revision (Corrects Typographical Error)  Changed LCO Bases Section 3.4.6 which incorrectly referred to Surveillance Requirement 3.4.6.2 rather than correctly identifying Surveillance Requirement 3.4.6.3.
Revision 73                                            09-11-06            Updated the Bases for LCO 3.9.4 to clarify that penetration flow paths through containment to the outside atmosphere must be limited to less than the ABSCE breach allowance. Also administratively removed from the Bases for LCO 3.9.4 a statement on core alterations that should have been removed as part of Amendment 35.
Revision 74                                              09-16-06          For the LCO section of the Bases for LCO 3.9.4, administratively removed the change made by Revision 73 to the discussion of an LCO note and placed the change in another area of the LCO section.
Revision 75 (Amendment 45)                              09-18-06          Revised the Bases for LCOs 3.8.7, 3.8.8 and 3.8.9 to incorporate a spare inverter for Channel 1-II of the Vital Inverters (DCN 51370).
Watts Bar-Unit 1                                              xxi                                            Revision 75
 
TECHNICAL SPECIFICATION BASES - REVISION LISTING (This listing is an administrative tool maintained by WBN Licensing and may be updated without formally revising the Technical Specification Bases Table-of-Contents)
REVISIONS                                ISSUED                                SUBJECT Revision 76 (Amendment 45)                              09-22-06          Revised the Bases for LCOs 3.8.7, 3.8.8 and 3.8.9 to incorporate a spare inverter for Channel 1-IV of the Vital Inverters (DCN 51370).
Revision 77 (Amendment 45)                              10-10-06          Revised the Bases for LCOs 3.8.7, 3.8.8 and 3.8.9 to incorporate a spare inverter for Channel 1-I of the Vital Inverters (DCN 51370).
Revision 78 (Amendment 45)                              10-13-06          Revised the Bases for LCOs 3.8.7, 3.8.8 and 3.8.9 to incorporate a spare inverter for each of the Vital Inverters (DCN 51370).
Revision 79 (Amendment 60, 61 and                      11-03-06            Steam Generator Narrow Range Level
: 64)                                                                        Indication Increased from 6% to 32% (WBN-TS-05-06) Bases Sections 3.4.5, 3.4.6, and 3.4.7.
Revision 80                                            11-08-06            Revised the Bases for SR 3.5.2.8 to clarify that inspection of the containment sump strainer constitutes inspection of the trash rack and the screen functions.
Revision 81 (Amendment 62)                            11-15-06            Revised the Bases for SR 3.6.11.2, 3.6.11.3, and 3.6.11.4 to address the Increase Ice Weight in Ice Condenser to Support Replacement Steam Generators (WBN-TS-05-09) [SGRP]
Revision 82 (Amendment 65)                            11-17-06            Steam Generator (SG) Tube Integrity (WBN-TS-05-10) [SGRP]
Revision 83                                            11-20-06            Updated Surveillance Requirement (SR) 3.6.6.5 to clarify that the number of unobstructed spray nozzles is defined in the design bases.
Revision 84                                            11-30-06            Revised Bases 3.6.9 and 3.6.15 to show the operation of the EGTS when annulus pressure is not within limits.
Revision 85                                            03-22-07            Revised Bases 3.6.9 and 3.6.15 in accordance with TACF 1-07-0002-065 to clarify the operation of the EGTS.
Watts Bar-Unit 1                                              xxii                                            Revision 85
 
TECHNICAL SPECIFICATION BASES - REVISION LISTING (This listing is an administrative tool maintained by WBN Licensing and may be updated without formally revising the Technical Specification Bases Table-of-Contents)
REVISIONS                                  ISSUED                                SUBJECT Revision 86                                              01-31-08          Figure 3.7.15-1 was deleted as part of Amendment 40. A reference to the figure in the Bases for LCO 3.9.9 was not deleted at the time Amendment 40 was incorporated into the Technical Specifications. Bases Revision 86 corrected this error (refer to PER 130944).
Revision 87                                              02-12-08            Implemented Bases change package TS                                                                                13 for DCN 52220-A. This DCN ties the ABI and CVI signals together so that either signal initiates the other signal.
Revision 88 (Amendment 67)                                03-06-08            Technical Specification Amendment 67 increased the number of TPBARs from 240 to 400.
Revision 89 (Amendment 66)                                05-01-08          Update of Bases to be consistent with the changes made to Section 5.7.2.11 of the Technical Specifications to reference the ASME Operation and Maintenance Code Revision 90 (Amendment 68)                                10-02-08          Issuance of amendment regarding Reactor Trip System and Engineered Safety Features Actuation System completion times, bypass test times, and surveillance test intervals Revision 91 (Amendment 70)                              11-25-2008          The Bases for TS 3.7.10, Control Room Emergency Ventilation System (CREVS) were revised to address control room envelope habitability.
Revision 92 (Amendment 71)                              11-26-2008          The Bases for TS 3.4.15, RCS Leakage Detection Instrumentation were revised to remove the requirement for the atmospheric gaseous radiation monitor as one of the means for detecting a one gpm leak within one hour.
Revision 93 (Amendment 74)                              02-09-2009          Updates the discussion of the Allowable Values associated with the Containment Purge Radiation Monitors in the LCO section of the Bases for LCO 3.3.6.
Revision 94 (Amendment 72)                              02-23-2009          Bases Revision 94 [Technical Specification (TS)] Amendment 72 deleted the Hydrogen Recombiners (LCO 3.6.7) from the TS and moved the requirements to the Technical Requirements Manual.
Watts Bar-Unit 1                                            xxiii                                          Revision 94
 
TECHNICAL SPECIFICATION BASES - REVISION LISTING (This listing is an administrative tool maintained by WBN Licensing and may be updated without formally revising the Technical Specification Bases Table-of-Contents)
REVISIONS                                  ISSUED                                SUBJECT Revision 95                                          03-05-2009            Corrected an error in SR 3.3.2.6 which referenced Function 6.g of TS Table 3.3.2-1.
This function was deleted from the TS by Amendment 1.
Revision 96 (Amendment 75)                            06-19-2009            Modified Mode 1 and 2 applicability for Function 6.e of TS Table 3.3.2-1, "Engineered Safety Feature Actuation System Instrumentation." This is associated with AFW automatic start on trip of all main feedwater pumps. In addition, revised LCO 3.3.2, Condition J, to be consistent with WBN Unit 1 design bases.
Revision 97 (Amendment 76)                            09-23-2009            Amendment 76 updates LCO 3.8.7, Inverters - Operating to reflect the installation of the Unit 2 inverters.
Revision 98 (Amendments 77, 79, &                    10-05-2009            Amendment 77 revised the number of
: 81)                                                                        TPBARs that may be loaded in the core from 400 to 704.
Amendment 79 revised LCO 3.6.3 to allow verification by administrative means isolation devices that are locked, sealed, or otherwise secured.
Amendment 81 revised the allowed outage time of Action B of LCO 3.5.1 from 1 hour to 24 hours.
Revision 99                                          10-09-2009            Bases Revision 99 incorporated Westinghouse Technical Bulletin (TB) 08-04.
Revision 100                                          11-17-2009            Bases Revision 100 revises the LCO description of the Containment Spray System to clarify that transfer to the containment sump is accomplished by manual actions.
Revision 101                                          02-09-2010            Bases Revision 101 implemented DCN 52216-A that will place both trains of the EGTS pressure control valves hand switches in A-AUTO and will result in the valves opening upon initiation of the Containment Isolation phase A (CIA) signal.
They will remain open independent of the annulus pressure and reset of the CIA.
Revision 102                                          03-01-2010            Bases Revision 102 implemented EDC 52564-A which addresses a new single failure scenario relative to operation of the EGTS post LOCA.
Watts Bar-Unit 1                                            xxiv                                            Revision 102
 
TECHNICAL SPECIFICATION BASES - REVISION LISTING (This listing is an administrative tool maintained by WBN Licensing and may be updated without formally revising the Technical Specification Bases Table-of-Contents)
REVISIONS                                  ISSUED                                SUBJECT Revision 103                                          04-05-2010          Bases Revision 103 implemented NRC guidance Application of Generic Letter 80-30 which allows a departure from the single failure criterion where a non-TS support system has two 100% capacity subsystems, each capable of supporting the design heat load of the area containing the TS equipment.
Revision 104 (Amendment 82)                            09-20-2010          Bases Revision 104 implemented License Amendment No. 82, which approved the BEACON-TSM application of the Power Distributing System. The PDMS requirements reside in the TRM.
Revision 105                                          10-28-2010          DCN 53437 added spare chargers 8-S and 9-S which increased the total of 125 VDC Vital Battery Chargers to eight (8).
Revision 106                                          01-20-2011          Revised SR 3.8.3.6 to clarify that identified fuel oil leakage does not constitute failure of the surveillance.
Revision 107 (Amendment 85)                            02-24-2011          Amendment 85 revises TS 3.7.11, "Control Room Emergency Air Temperature Control System (CREATCS). Specifically, the proposed change will only be applicable during plant modifications to upgrade the CREATCS chillers. This "one-time" TS change will be implemented during Watts Bar Nuclear Plant, Unit 1 Cycles 10 and 11 beginning March 1, 2011, and ending April 30, 2012.
Revision 108                                          03-07-2011          Bases Revision 108 deletes reference to NSRB to be notified of violation of a safety limit within 24 hours in TSB 2.2.4. Also, corrected error in SR 3.3.2.4 in the reference to Table 3.3.1-1. It should be Table 3.3.2-1.
Revision 109                                          04-06-2011          Bases Revision 109 clarifies that during plant startup in Mode 2 the AFW anticipatory auto-start signal need not be OPERABLE if the AFW system is in service. PER 287712 was identified by NRC to provide clarification to TS Bases 3.3.2, Function 6.e, Trip of All Turbine Driven Main Feedwater Pumps.
Revision 110                                          04-19-2011          Clarified the text associated with the interconnection of the ABI and CVI functions in the bases for LCO 3.3.6, 3.3.8, 3.7.12 and 3.9.8.
Watts Bar-Unit 1                                            xxv                                          Revision 110
 
TECHNICAL SPECIFICATION BASES - REVISION LISTING (This listing is an administrative tool maintained by WBN Licensing and may be updated without formally revising the Technical Specification Bases Table-of-Contents)
REVISIONS                                  ISSUED                                  SUBJECT Revision 111                                          05-05-2011            Added text to several sections of the Bases for LCO 3.4.16 to clarify that the actual transient limit for I-131 is 14 Ci/gm and refers to the controls being placed in AOI-28.
Revision 112                                          05-24-2011            DCN 55076 replaces the existing four 125-Vdc DG Battery Chargers with four sets of redundant new battery charger assemblies.
Revision 113                                          06-24-2011            Final stage implementation of DCN 55076 which replaced the existing four 125-Vdc DG Battery Chargers with four sets of redundant new battery charger assemblies.
Revision 114                                          12-12-2011            Clarifies the acceptability of periodically using a portion of the 25% grace period in SR 3.0.2 to facilitate 13 week maintenance work schedules.
Revision 115                                          12-21-2011            Revises several surveillance requirements notes in TS 3.8.1 to allow performance of surveillances on WBN Unit 2 6.9 kV shutdown boards and associated diesel generators while WBN Unit 1 is operating in MODES 1, 2, 3, or 4 Revision 116                                          06-27-2012            Revises TS Bases 3.8.1, AC Sources -
Operating, to make the TS Bases consistent with TS 3.8.1, Condition D Revision 117                                          07-27-2012            Revises TS Bases 3.7.10, Control Room Emergency Ventilation System (CREVS), to make the TS Bases consistent with TS 3.7.10, Condition E Revision 118                                          01-30-2013            Revises TS Bases 3.4.16, Reactor Coolant System (RCS) to change the dose equivalent I-131 spike limit and the allowable value for control room air intake radiation monitors.
Revision 119                                          08-17-2013            Revises TS Bases 3.3.6, 3.3.8, 3.7.12, 3.7.13, 3.9.4, 3.9.7, 3.9.8, and adds TS Bases 3.9.10 to reflect selective implementation of the Alternate Source Term methodology for the analysis of Fuel Handling Accidents (FHAs) and make TS Bases consistent with the revised FHA dose analysis.
Watts Bar-Unit 1                                            xxvi                                            Revision 119
 
TECHNICAL SPECIFICATION BASES - REVISION LISTING (This listing is an administrative tool maintained by WBN Licensing and may be updated without formally revising the Technical Specification Bases Table-of-Contents)
REVISIONS                                  ISSUED                                SUBJECT Revision 120                                          01-23-2014            Revised the References to TS Bases 3.1.9, PHYSICS TESTS Exceptions - Mode1, to document NRC approval of WCAP 12472-P-A. Addendum 1-A and 4-A., Addendum 1-A approved the use of the Advance Nodal Code (ANC-Phoenix_ in the BEACON system as the neutronic code for measuring core power distribution. Is also approved the use of fixed incore self-powered neutron detectors (SPD) to calibrate the BEACON system in lieu of incore and excore neutron detectors and core exit thermocouples (CET). For plants that do not have SPDs Addendum 4-A approved Westinghouse methodology that allow the BEACON system to calculate CET uncertainty as a function of reactor power on a plant cycle basis during power ascension following a refueling outage.
Revision 121                                          08-04-2014            Revises references in TS Bases 3.7.1 for consistency with changes to the TS Bases 3.7.1 references approved in Revision 89.
Revision 122 (Amendment 94)                          01-14-2014            Revises TS Bases 3.7.10, Control Room Emergency Ventilation System (CREVS) to make the TS Bases consistent with TS 3.7.10, Actions E, F, G, and H.
Revision 123 (Amendment 104)                          03-16-2016            Amendment 104, TSB Revision 123 adds TS B3.7.16, Component Cooling System (CCS) - Shutdown and adds TS B3.7.17, Essential Raw Cooling Water (ERCW)
System - Shutdown.
Revision 124                                          02-12-2016            Revises TS Bases Table B3.8.9-1, AC and DC Electrical Power Distribution Systems, the second Note.
Revision 125 (Amendment 84, 102,                      03-16-2016            Revises TS Bases Section B3.8-1, AC 103)                                                                        Sources-Operating.
Revision 126                                          03-18-2016            Revises TS Bases Section B3.7.7, Component Cooling System the 1B and 2B surge tank sections.
Revision 127                                          04-18-2016            Revises TS Bases Section B 3.6.4, Containment Pressure and B3.6.6, Containment Spray System to change the maximum peak pressure from a LOCA of 9.36 psig.
Watts Bar-Unit 1                                            xxvii                                        Revision 127
 
TECHNICAL SPECIFICATION BASES - REVISION LISTING (This listing is an administrative tool maintained by WBN Licensing and may be updated without formally revising the Technical Specification Bases Table-of-Contents)
REVISIONS                                  ISSUED                                SUBJECT Revision 128                                            06-27-16            Revises TS Bases Section B3.6.8, Hydrogen Mitigation System (HMS), to delete sentence regarding Hydrogen Recombiners that are abandoned.
Revision 129                                            08-19-16            Revises TS Bases Section 3.6.15, Shield Building, to clarify the use of the Condition B note.
Revision 130                                            12-22-16            Revises TS Bases Sections 3.6.1, 3.6.2, and 3.6.3 to reflect the deletion of TS 3.9.4 in WBN Unit 1 TS Amendment 92.
Revision 131 (Amendment 107)                            01-13-17            Revises TS Bases Section 3.5.4,  Refueling Water Storage Tank (RWST), Applicable Safety Analyses Revision 132 (Amendment 110)                            01-17-17            Revises TS Bases Section 3.8.1, AC Sources -Operating Revision 133 (Amendment 111)                            03-13-17            Adds TS Bases Section 3.0.8 for Inoperability of Snubbers.
Revision 134 (Amendment 112)                            04-25-17            Revise TS Bases Section 3.7.11 Action A.1 regarding CREATCS.
Revision 135                                            05-17-17            Revises TS Bases Section B3.3.3, PAM Instrumentation Revision 136 (Amendment 113)                            05-17-17            Revises TS Bases Section B3.7.7 CCS Revision 137 (Amendment 114)                            07-14-17            Revises TS Bases Section B SR 3.0.2 to add a one-time extension for the surveillance interval.
Revision 138 (Amendment 115)                            11-2-17            Revises TS Bases to adopt the TSTF-522 to revise ventilation system surveillance requirements to operate for 10 hours per month.
Revision 139 (Amendment 116)                            11-2-17            Revises TS Bases Auxiliary Building Gas Treatment System.
Revision 140                                            12-12-17            Revises TS Bases to include the ABB-NV and WLOP secondary CHF correlations.
Revision 141                                            03-08-18            Revises TS Bases 3.0.6 to remove non-standard guidance added by Bases Rev.103 that applied LCO 3.0.6 to non-TS support equipment when the equipment consisted of two 100% capacity subsystems, each capable of supporting both trains of TS equipment.
Watts Bar-Unit 1                                            xxviii                                          Revision 141
 
TECHNICAL SPECIFICATION BASES - REVISION LISTING (This listing is an administrative tool maintained by WBN Licensing and may be updated without formally revising the Technical Specification Bases Table-of-Contents)
REVISIONS                                  ISSUED                                SUBJECT Revision 142                                            04-06-18            Add clarifying information of ECCS gas that some gas is acceptable based on output of DCP 66453.
Revision 143, Amendment 120                            08-20-18            Revises TS Bases and adopts the TSTF-547, Clarification of Rod position requirements.
Revision 144, Amendment 121                            08-16-18            Revises TS Bases 3.0 to extend surveillance requirements and specify intervals.
Revision 145, Amendment 122                            09-21-18            Revises TS Bases 3.2.4 and Bases 3.3.1 related to the reactor trip system instrumentation.
Revision 146, Amendment 119                            10-11-18            Revises TS Bases 3.3.1 Reactor Trip System Instrumentation, to reflect plant modifications to the reactor protection system instrumentation associated with the turbine trip on low fluid oil pressure.
Revision 147                                            11-14-18            Revises TS Bases 3.7.5, AFW System, to increase margin on the AFW MDAFW pumps.
Revision 148                                            11-14-18            Revises TS Bases 3.4.12, References section to update Reference 4 with an updated FSAR Section.
Revision 149                                            2-13-19            Revises TS Bases 3.3.1, Reactor Trip System Instrumentation Revision 150, Amendment 124                              3-19-19            Revises TS Bases 3.3.4, Remote Shutdown System Revision 151, Amendment 123                              6-13-19            Revises TS Bases 3.6.3, Containment Isolation Valves, Surveillance Requirement 3.6.3.5 for the containment purge valves to revise the frequency from "184 days AND Within 92 days after opening the valve" to "In accordance with the Containment Leakage Rate Testing Program."
Revision 152, Amendment 126                              8-1-19            Revises TS Bases 3.8.9, Distribution Systems  Operating, to add a new Condition C.
Revision 153                                              8-1-19            Revises TS Bases 3.2.1 and 3.2.2.
Revision 154                                              8-7-19            Revises TS Bases 3.8.6, surveillance requirements.
Watts Bar-Unit 1                                            xxix                                            Revision 154
 
TECHNICAL SPECIFICATION BASES - REVISION LISTING (This listing is an administrative tool maintained by WBN Licensing and may be updated without formally revising the Technical Specification Bases Table-of-Contents)
REVISIONS                                  ISSUED                                SUBJECT Revision 155                                            8-22-19            Revises TS Bases 3.8.4, DC Sources Operating.
Revision 156 (Amendment 128)                          12-17-19            Revises TS Bases 3.3.5, LOP DG Start Instrumentation, to implement Class 1E unbalanced voltage relays.
Revision 157                                            9-26-19            Removed Reference 5 from B 3.3-162.
CR 1514672 Revision 158 (Amendment 129)                          12-17-19            Revises Tech Spec Bases 3.8.1, 3.8.7, 3.8.8, and 3.8.9 to support performance of the 6.9kV and 480V shutdown board maintenance.
Revision 159                                            1-13-20            Revises Tech Spec Bases 3.0.2 and 3.0.3 to remove the term operational convenience.
Revision 160 (Amendment 130)                            1-29-20            TSTF-500  DC Electrical Rewrite Update to TSTF-360 Revision 161                                            2-20-20            Revises CSST A and B to qualify to GDC-17 requirements in order to be considered as a TS offsite power source substitute for CSST D or C when out of service.
Revision 162 (Amendment 132)                            3-25-20            TSTF- 425  Surveillance Frequency Testing Program Revision 163                                            3-17-20            Revises Tech Spec Bases 3.8.9, Distribution Systems  Operating, regarding the Diesel Auxiliary Building Boards.
Revision 164 (Amendment 133)                              4-8-20            Revises Tech Spec Bases 3.3.5, LOP DG Start Instrumentation for Condition C.
Revision 165 (Amendment 135)                            7-20-20            Revises miscellaneous administrative changes to the Tech Specs Bases.
Revision 166 (Amendment 136)                            9-15-20            Revises Tech Spec Bases 3.8.1 operability of the automatic transfer from a Unit Service Station Transformer to a Common Station Service Transformer A or B at the associated unit board.
Watts Bar-Unit 1                                            xxx                                        Revision 166
 
TECHNICAL SPECIFICATION BASES - REVISION LISTING (This listing is an administrative tool maintained by WBN Licensing and may be updated without formally revising the Technical Specification Bases Table-of-Contents)
REVISIONS                                  ISSUED                              SUBJECT Revision 167 (Amendment 125)                            11-4-20            Revises Tech Spec Bases 3.7.15, Spent Fuel Pool Assembly Storage, and adding the Bases 3.7.18 Fuel Storage Pool Boron Concentration.
Revision 168 (Amendment 137)                            12-8-20            Adopts TSTF-541, Revision 2, Add exceptions to Surveillance Requirements for valves and dampers locked in the actuated position.
Revision 169 (Amendment 139)                            01-05-21            Revises Tech Spec Bases 3.6.15 by deleting existing Condition B and revise the acceptance criteria for annulus pressure.
Revision 170 (Amendment 141)                            02-02-21            Adopts TSTF-569, Revision 2, Revise Response Time Testing Definition.
Revision 171 (Amendment 144)                            03-31-21            Revises the Post Accident Monitoring (PAM) Instrumentation to remove plasma from the text.
Revision 172 (Amendment 145)                            05-19-21            One-Time Change to Tech Spec Bases 3.7.11 to Extend the Completion Time for Main Control Room Chiller Modifications.
Revision 173 (Amendment 146)                            06-22-21            TSTF-490 Deletion of E Bar and revision of the RCS specific activity.
Revision 174 (Amendment 147)                            08-11-21            TSTF-510 Steam Generator Tube Inspection frequency.
Revision 175 (Amendment 142)                            11-04-21            Implement WCAP-17661-P-A, Improved RAOC and CAOC FQ Surveillance Tech Specs.
Revision 176 (Amendment 143)                            11-10-21            Full SpectrumTM Loss of Coolant Accident Analysis (LOCA) and New LOCA - Specific Tritium Producing Burnable Absorber Rod Stress Analysis Methodology.
Revision 177 (Amendment 148)                            11-17-21            Revises Technical Specification Bases B3.3.2 for Function 6.E.
Watts Bar-Unit 1                                            xxxi                                        Revision 177
 
TECHNICAL SPECIFICATION BASES - REVISION LISTING (This listing is an administrative tool maintained by WBN Licensing and may be updated without formally revising the Technical Specification Bases Table-of-Contents)
REVISIONS                                                ISSUED            SUBJECT Revision 178 (Amendment 149)                            11-30-21            Revise Technical Specification 5.7.2.19, Containment Leakage Rate Testing Program to extend containment integrated and local leak rate test intervals.
Revision 179 (Amendment 150)                            1-12-22            Revise Technical Specification SR 3.6.15.4.
Revision 180                                            1-25-22            Revises Technical Specification Bases 3.8.1 AC Sources  Operating to support the upcoming shutdown board cleaning.
Revision 181                                              2-9-22            Revises Technical Specification Bases Table 3.8.9-1 to remove C&A vent Boards 1A2-A and 1B2-B.
Revision 182 (Amendment 151)                              3-1-22            Revises Technical Specification Bases 3.7.12 to add note for one-time exception.
Watts Bar-Unit 1                                            xxxii                                      Revision 182
 
ENCLOSURE 2 WBN UNIT 1 TECHNICAL SPECIFICATION BASES CHANGED PAGES (95 pages)
 
TABLE OF CONTENTS (continued)
B 3.7            PLANT SYSTEMS .................................................................................................. B 3.7-1 B 3.7.1              Main Steam Safety Valves (MSSVs) ............................................................ B 3.7-1 B 3.7.2              Main Steam Isolation Valves (MSIVs) .......................................................... B 3.7-7 B 3.7.3              Main Feedwater Isolation Valves (MFIVs) and Main Feedwater Regulation Valves (MFRVs) and Associated Bypass Valves .............................................................. B 3.7-12 B 3.7.4              Atmospheric Dump Valves (ADVs)............................................................... B 3.7-17 B 3.7.5              Auxiliary Feedwater (AFW) System.............................................................. B 3.7-21 B 3.7.6              Condensate Storage Tank (CST) ................................................................. B 3.7-29 B 3.7.7              Component Cooling System (CCS) .............................................................. B 3.7-33 B 3.7.8              Essential Raw Cooling Water (ERCW) System ........................................... B 3.7-38 B 3.7.9              Ultimate Heat Sink (UHS).............................................................................. B 3.7-42 B 3.7.10            Control Room Emergency Ventilation System (CREVS) ............................ B 3.7-44 B 3.7.11            Control Room Emergency Air Temperature Control System (CREATCS) .................................................................. B 3.7-51 B 3.7.12            Auxiliary Building Gas Treatment System (ABGTS).................................... B 3.7-55 B 3.7.13            Fuel Storage Pool Water Level ..................................................................... B 3.7-60 B 3.7.14            Secondary Specific Activity ........................................................................... B 3.7-63 B 3.7-15            Spent Fuel Pool Assembly Storage .............................................................. B 3.7-66 B 3.7-16            Component Cooling System (CCS) - Shutdown.......................................... B 3.7-69 B 3.7-17            Essential Raw Cooling Water (ERCW) System Shutdown ......................... B 3.7-75 B 3.7-18            Fuel Storage Pool Boron Concentration. B 3.7-81 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 3.8-39 B 3.8.3              Diesel Fuel Oil, Lube Oil, and Starting Air .................................................... B 3.8-44 B 3.8.4              DC Sources  Operating .............................................................................. B 3.8-54 B 3.8.5              DC Sources  Shutdown.............................................................................. B 3.8-66 B 3.8.6              Battery Parameters ........................................................................................ B 3.8-71 B 3.8.7              Inverters  Operating .................................................................................... B 3.8-80 B 3.8.8              Inverters  Shutdown ................................................................................... B 3.8-84 B 3.8.9              Distribution Systems  Operating ................................................................ B 3.8-87 B 3.8.10            Distribution Systems  Shutdown ................................................................ B 3.8-97 B 3.9            REFUELING OPERATIONS .................................................................................. B 3.9-1 B 3.9.1              Boron Concentration...................................................................................... B 3.9-1 B 3.9.2              Unborated Water Source Isolation Valves ................................................... B 3.9-4 B 3.9.3              Nuclear Instrumentation ................................................................................ B 3.9-7 B 3.9.4              Deleted ........................................................................................................... B 3.9-10 B 3.9.5              Residual Heat Removal (RHR) and Coolant Circulation  High Water Level ............................................................. B 3.9-11 B 3.9.6              Residual Heat Removal (RHR) and Coolant Circulation  Low Water Level .............................................................. B 3.9-15 B 3.9.7              Refueling Cavity Water Level ........................................................................ B 3.9-18 B 3.9.8              Deleted ........................................................................................................... B 3.9-21 B 3.9.9              Spent Fuel Pool Boron Concentration .......................................................... B 3.9-22 B 3.9.10            Decay Time .................................................................................................... B 3.9-24 Watts Bar-Unit 1                                                iii                                                  Revision 150, 162, 167
 
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i      162              B 2.0-1    0 ii      162              B 2.0-2    140 iii      167              B 2.0-3    0 iv      162              B 2.0-4    59 v      162              B 2.0-5    108 vi      104              B 2.0-6    140 vii      0                B 2.0-7    0 viii      182              B 2.0-8    0 ix      176              B 2.0-9    0 x      177              B 2.0-10  108 xi      171              B 2.0-11    0 xii      178              B 3.0-1    133 xiii      179              B 3.0-2    159 xiv      182              B 3.0-3    159 xv        181              B 3.0-4    68 xvi        19              B 3.0-5    68 xvii      32              B 3.0-6    68 xviii      46              B 3.0-7    0 xix        60              B 3.0-8    141 xx        68              B 3.0-9    133 xxi        75              B 3.0-10  133 xxii      85              B 3.0-11  133 xxiii      94              B 3.0-12    0 xxiv      102              B 3.0-13  165 xxv      110              B 3.0-14  159 xxvi      119              B 3.0-15    68 xxvii      127              B 3.0-16    68 xxviii      141              B 3.1-1    0 xxix      154              B 3.1-2    0 xxx      166              B 3.1-3    0 xxxi      177              B 3.1-4    68 xxxii      182              B 3.1-5    0 B 3.1-6    162 B 3.1-7    0 B 3.1-8    0 B 3.1-9    68 B 3.1-10  162 B 3.1-11    0 B 3.1-12    32 B 3.1-13    0 B 3.1-14    0 B 3.1-15    0 Watts Bar-Unit 1                  viii                    Revision 182
 
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B 3.3-106      0                B 3.4-1    0 B 3.3-107      0                B 3.4-2    60 B 3.3-108      171              B 3.4-3    60 B 3.3-109      0                B 3.4-4    162 B 3.3-110      94              B 3.4-5    60 B 3.3-111      171              B 3.4-6    0 B 3.3-112      171              B 3.4-7    55 B 3.3-113      94              B 3.4-8    162 B 3.3-114      94              B 3.4-9    0 B 3.3-115      162              B 3.4-10    0 B 3.3-116      162              B 3.4-11    0 B 3.3-117      0                B 3.4-12    0 B 3.3-118      0                B 3.4-13    0 B 3.3-119      150              B 3.4-14  162 B 3.3-120      150              B 3.4-15    0 B 3.3-121      162              B 3.4-16    0 B 3.3-122      162              B 3.4-17    82 B 3.3-123      162              B 3.4-18  162 B 3.3-124      150              B 3.4-19    0 B 3.3-125      156              B 3.4-20    0 B 3.3-126      0                B 3.4-21    79 B 3.3-127      156              B 3.4-22    0 B 3.3-128      156              B 3.4-23  162 B 3.3-129      164              B 3.4-24  123 B 3.3-130      162              B 3.4-25  123 B 3.3-131      119              B 3.4-26  123 B 3.3-132      9                B 3.4-27  123 B 3.3-133      119              B 3.4-28  162 B 3.3-134      119              B 3.4-29  162 B 3.3-135      162              B 3.4-30    79 B 3.3-136      162              B 3.4-31    79 B 3.3-137      162              B 3.4-32    82 B 3.3-138      0                B 3.4-33  162 B 3.3-139      45              B 3.4-34    29 B 3.3-140      0                B 3.4-35    0 B 3.3-141      0                B 3.4-36    68 B 3.3-142      45              B 3.4-37  162 B 3.3-143      162              B 3.4-38    0 B 3.3-144      162              B 3.4-39    0 B 3.3-145      119              B 3.4-40    0 B 3.3-146      119              B 3.4-41  162 B 3.3-147      119              B 3.4-42    0 B 3.3-148      162              B 3.4-43    0 B 3.4-44    0 Watts Bar-Unit 1                    xi                    Revision 171
 
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B 3.4-45      89              B 3.4-86    82 B.3.4-46        0              B 3.4-87  174 B 3.4-47      42              B 3.4-88    82 B 3.4-48      68              B 3.4-89  174 B 3.4-49      42              B 3.4-90  174 B 3.4-50      0                B 3.4-91  174 B 3.4-51      162              B 3.5-1    0 B 3.4-52      0                B 3.5-2    176 B 3.4-53      0                B 3.5-3    176 B 3.4-54      0                B 3.5-4    176 B 3.4-55      0                B 3.5-5    98 B 3.4-56      0                B 3.5-6    162 B 3.4-57      0                B 3.5-7    162 B 3.4-58      68              B 3.5-8    98 B 3.4-59      0                B 3.5-9    61 B 3.4-60      68              B 3.5-10    0 B 3.4-61      0                B 3.5-11  176 B 3.4-62      162              B 3.5-12    39 B 3.4-63      162              B 3.5-13    68 B 3.4-64      89              B 3.5-14    68 B 3.4-65      82              B 3.5-15    0 B 3.4-66      82              B 3.5-16  165 B 3.4-67      82              B 3.5-17  162 B 3.4-68      82              B 3.5-18  165 B 3.4-69      162              B 3.5-19    0 B.3.4-70      0                B 3.5-20    68 B 3.4-71      0                B 3.5-21    0 B 3.4-72      0                B 3.5-22    0 B 3.4-73      162              B 3.5-23    0 B 3.4-74      162              B 3.5-24    0 B 3.4-75      92              B 3.5-25  176 B 3.4-76      12              B 3.5-26    0 B 3.4-77      92              B 3.5-27  162 B 3.4-78      92              B 3.5-28  176 B 3.4-79      162              B 3.5-29    0 B 3.4-80      173              B 3.5-30    0 B 3.4-81      173              B 3.5-31  162 B 3.4-82      173              B 3.6-1    10 B 3.4-83      173              B 3.6-2    10 B 3.4-84      173              B 3.6-3    10 B 3.4-85      173              B 3.6-4    178 B 3.6-5    10 Watts Bar-Unit 1                  xii                    Revision 178
 
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B 3.6-6      178              B 3.6-51  162 B 3.6-7      0                B 3.6-52  162 B 3.6-8      0                B 3.6-53  162 B 3.6-9      0                B 3.6-54  162 B 3.6-10      162              B 3.6-55  176 B 3.6-11      0                B 3.6-56    0 B 3.6-12      0                B 3.6-57    0 B 3.6-13      130              B 3.6-58  176 B 3.6-14      98              B 3.6-59    0 B 3.6-15      98              B 3.6-60    36 B 3.6-16      98              B 3.6-61  165 B 3.6-17      0                B 3.6-62  165 B 3.6-18      162              B 3.6-63  176 B 3.6-19      162              B 3.6-64    0 B 3.6-20      178              B 3.6-65    0 B 3.6-21      178              B 3.6-66    0 B 3.6-22      178              B 3.6-67  162 B 3.6-23      176              B 3.6-68  162 B 3.6-24      176              B 3.6-69    0 B 3.6-25      0                B 3.6-70    0 B 3.6-26      29              B 3.6-71    0 B 3.6-27      162              B 3.6-72  162 B 3.6-28      0                B 3.6-73  129 B 3.6-29      0                B 3.6-74  129 B 3.6-30      176              B 3.6-75  169 B 3.6-31      0                B 3.6-76  179 B 3.6-32      162              B 3.7-1    31 B 3.6-33      176              B 3.7-2    31 B 3.6-34      94              B 3.7-3    41 B 3.6-35      0                B 3.7-4    121 B 3.6-36      128              B 3.7-5    121 B 3.6-37      0                B 3.7-6    89 B 3.6-38      162              B 3.7-7    0 B 3.6-39      165              B 3.7-8    0 B 3.6-40      138              B 3.7-9    0 B 3.6-41      71              B 3.7-10    0 B 3.6-42      162              B 3.7-11  162 B 3.6-43      168              B 3.7-12    76 B3.6-43a      0                B 3.7-13    0 B 3.6-44      0                B 3.7-14    0 B 3.6-45      176              B 3.7-15    89 B 3.6-46      0                B 3.7-16  162 B 3.6-47      176              B 3.7-17    0 B 3.6-48      81              B 3.7-18    0 B 3.6-49      0                B.3.7-19    68 B 3.6-50      176              B 3.7-20  162 Watts Bar-Unit 1                  xiii                    Revision 179
 
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B 3.7-21        0                B 3.7-67  167 B 3.7-22      147              B 3.7-68  167 B 3.7-23        0                B 3.7-69  123 B 3.7-24        68              B 3.7-70  123 B 3.7-25        0                B 3.7-71  123 B 3.7-26      162              B 3.7-72  123 B 3.7-27      162              B 3.7-73  123 B 3.7-28        89              B 3.7-74  162 B 3.7-29        0                B 3.7-75  123 B 3.7-30        41              B 3.7-76  123 B 3.7-31      162              B 3.7-77  123 B 3.7-32        0                B 3.7-78  123 B 3.7-33      136              B 3.7-79  162 B 3.7-34        0                B 3.7-80  123 B 3.7-35      136              B 3.7-81  167 B 3.7-36      162              B 3.7-82  167 B 3.7-37      162              B 3.8-1    166 B 3.7-38        0                B 3.8-2    166 B 3.7-39        0                B 3.8-3    180 B 3.7-40      162              B 3.8-4    166 B 3.7-41      162              B 3.8-5    166 B 3.7-42        0                B 3.8-6    180 B 3.7-43      162              B 3.8-7    158 B 3.7-44        91              B 3.8-8    0 B 3.7-45        91              B 3.8-9    132 B 3.7-46        91              B 3.8-10  180 B 3.7-47        91              B 3.8-11  132 B 3.7-48      122              B 3.8-12  132 B 3.7-49      168              B 3.8-13  132 B 3.7-50        91              B 3.8-14  180 B 3.7-51        64              B 3.8-15  158 B 3.7-52        64              B 3.8-16  158 B 3.7-53      172              B 3.8-17  180 B 3.7-54      172              B 3.8-18  158 B 3.7-54a      172              B 3.8-19  162 B 3.7-55      119              B 3.8-20  162 B 3.7-56      182              B 3.8-21  162 B 3.7-57      139              B 3.8-22  162 B 3.7-58      168              B 3.8-23  166 B 3.7-59      162              B 3.8-24  162 B 3.7-60      119              B 3.8-25  162 B 3.7-61      162              B 3.8-26  162 B 3.7-62      119              B 3.8-27  162 B 3.7-63        47              B 3.8-28  162 B 3.7-64        0                B 3.8-29  162 B 3.7-65      162              B 3.8-30  162 B 3.7-66      167 Watts Bar-Unit 1                    xiv                    Revision 182
 
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B 3.8-31      162              B 3.8-79  162 B 3.8-32      162              B 3.8-80    97 B 3.8-33      162              B 3.8-81    97 B 3.8-34      162              B 3.8-82    97 B 3.8-35      162              B 3.8-83  162 B 3.8-36      166              B 3.8-84    0 B 3.8-37      180              B 3.8-85    97 B 3.8-38      132              B 3.8-86  162 B 3.8-39      0                B 3.8-87  152 B 3.8-40      0              B 3.8-87a  163 B 3.8-41      0                B 3.8-88    0 B 3.8-42      0                B 3.8-89  158 B 3.8-43      0                B 3.8-90  152 B 3.8-44      0                B 3.8-91    0 B 3.8-45      0                B 3.8-92  152 B 3.8-46      0                B 3.8-93  152 B 3.8-47      55              B 3.8-94  158 B 3.8-48      55              B 3.8-95  158 B 3.8-49      162              B 3.8-96  162 B 3.8-50      0                B 3.8-97  181 B 3.8-51      162              B 3.8-98    0 B 3.8-52      162              B 3.8-99    0 B 3.8-53      29              B 3.8-100    0 B 3.8-54      160              B 3.8-101  162 B 3.8-55      160              B 3.9-1    0 B 3.8-56      160              B 3.9-2    68 B 3.8-57      160              B 3.9-3    162 B 3.8-58      160              B 3.9-4    68 B 3.8-59      160              B 3.9-5    0 B 3.8-60      160              B 3.9-6    162 B 3.8-61      160              B 3.9-7    0 B 3.8-62      162              B 3.9-8    0 B 3.8-63      162              B 3.9-9    162 B 3.8-64      162              B 3.9-10  119 B 3.8-65      160              B 3.9-11    0 B 3.8-66      0                B 3.9-12    23 B 3.8-67      160              B 3.9-13    0 B 3.8-68      160              B 3.9-14  162 B 3.8-69      160              B 3.9-15    0 B 3.8-70      0                B 3.9-16    68 B 3.8-71      160              B 3.9-17  162 B 3.8-72      160              B 3.9-18  119 B 3.8-73      160              B 3.9-19    45 B 3.8-74      160              B 3.9-20  162 B 3.8-75      160              B 3.9-21  119 B 3.8-76      160              B 3.9-22  167 B 3.8-77      162              B 3.9-23  167 B 3.8-78      162              B 3.9-24  119 B 3.9-25  119 Watts Bar-Unit 1                  xv                      Revision 181
 
TECHNICAL SPECIFICATION BASES - REVISION LISTING (This listing is an administrative tool maintained by WBN Licensing and may be updated without formally revising the Technical Specification Bases Table-of-Contents)
REVISIONS                                  ISSUED                              SUBJECT Revision 167 (Amendment 125)                            11-4-20            Revises Tech Spec Bases 3.7.15, Spent Fuel Pool Assembly Storage, and adding the Bases 3.7.18 Fuel Storage Pool Boron Concentration.
Revision 168 (Amendment 137)                              12-8-20          Adopts TSTF-541, Revision 2, Add exceptions to Surveillance Requirements for valves and dampers locked in the actuated position.
Revision 169 (Amendment 139)                            01-05-21            Revises Tech Spec Bases 3.6.15 by deleting existing Condition B and revise the acceptance criteria for annulus pressure.
Revision 170 (Amendment 141)                            02-02-21            Adopts TSTF-569, Revision 2, Revise Response Time Testing Definition.
Revision 171 (Amendment 144)                            03-31-21            Revises the Post Accident Monitoring (PAM) Instrumentation to remove plasma from the text.
Revision 172 (Amendment 145)                            05-19-21            One-Time Change to Tech Spec Bases 3.7.11 to Extend the Completion Time for Main Control Room Chiller Modifications.
Revision 173 (Amendment 146)                            06-22-21            TSTF-490 Deletion of E Bar and revision of the RCS specific activity.
Revision 174 (Amendment 147)                            08-11-21            TSTF-510 Steam Generator Tube Inspection frequency.
Revision 175 (Amendment 142)                            11-04-21            Implement WCAP-17661-P-A, Improved RAOC and CAOC FQ Surveillance Tech Specs.
Revision 176 (Amendment 143)                            11-10-21            Full SpectrumTM Loss of Coolant Accident Analysis (LOCA) and New LOCA - Specific Tritium Producing Burnable Absorber Rod Stress Analysis Methodology.
Revision 177 (Amendment 148)                            11-17-21            Revises Technical Specification Bases B3.3.2 for Function 6.E.
Watts Bar-Unit 1                                            xxxi                                        Revision 177
 
TECHNICAL SPECIFICATION BASES - REVISION LISTING (This listing is an administrative tool maintained by WBN Licensing and may be updated without formally revising the Technical Specification Bases Table-of-Contents)
REVISIONS                                                ISSUED            SUBJECT Revision 178 (Amendment 149)                            11-30-21            Revise Technical Specification 5.7.2.19, Containment Leakage Rate Testing Program to extend containment integrated and local leak rate test intervals.
Revision 179 (Amendment 150)                              1-12-22          Revise Technical Specification SR 3.6.15.4.
Revision 180                                              1-25-22          Revises Technical Specification Bases 3.8.1 AC Sources - Operating to support the upcoming shutdown board cleaning.
Revision 181                                              2-9-22            Revises Technical Specification Bases Table 3.8.9-1 to remove C&A vent Boards 1A2-A and 1B2-B.
Revision 182 (Amendment 151)                              3-1-22          Revises Technical Specification Bases 3.7.12 to add note for one-time exception.
Watts Bar-Unit 1                                              xxxii                                      Revision 182
 
FQ(Z)
B 3.2.1 B 3.2  POWER DISTRIBUTION LIMITS B 3.2.1        Heat Flux Hot Channel Factor (FQ(Z))
BASES BACKGROUND              The purpose of the limits on the values of FQ(Z) is to limit the local (i.e., pellet) peak power density. The value of FQ(Z) varies along the axial height (Z) of the core.
FQ(Z) is defined as the maximum local fuel rod linear power density divided by the average fuel rod linear power density, assuming nominal fuel pellet and fuel rod dimensions. Therefore, FQ(Z) is a measure of the peak fuel pellet power within the reactor core.
During power operation, the global power distribution is limited by LCO 3.2.3, "AXIAL FLUX DIFFERENCE (AFD)," and LCO 3.2.4, "QUADRANT POWER TILT RATIO (QPTR)," which are directly and continuously measured process variables. These LCOs, along with LCO 3.1.7, "Control Bank Insertion Limits,"
maintain the core limits on power distributions on a continuous basis.
FQ(Z) varies with fuel loading patterns, control bank insertion, fuel burnup, and changes in axial power distribution.
FQ(Z) is measured periodically using either the Movable Incore Detector System or the Power Distribution Monitoring System (PDMS) (Ref.6).
These measurements are generally taken with the core at or near equilibrium conditions.
Using the measured three dimensional power distributions, it is possible to derive a measured value for FQ(Z). However, because this value represents an equilibrium condition, it does not include the variations in the value of FQ(Z) which are present during non-equilibrium situations, such as load following or power ascension.
To account for these possible variations, the elevation dependent measured planar radial peaking factors, FXY(Z), are increased by an elevation dependent factor, [T(Z)]COLR, that accounts for the expected maximum values of the transient axial power shapes postulated to occur during RAOC operation. Thus, [T(Z))]COLR accounts for the worst case non-equilibrium power shapes that are expected for the assumed RAOC operating space.
The RAOC operating space is defined as the combination of AFD and Control Bank Insertion Limits assumed in the calculation of a particular [T(Z)]COLR function. The [T(Z)]COLR factors are directly dependent on the AFD and Control Bank Insertion Limit assumptions. The COLR may contain different [T(Z)] COLR (continued)
Watts Bar-Unit 1                                  B 3.2-1                                Revision 104, 175 Amendment 82, 142
 
FQ(Z)
B 3.2.1 BASES BACKGROUND      functions that reflect different operating space assumptions. If the limit on FQ(Z)
(continued)      is exceeded, a more restrictive operating space may be implemented to gain margin for future non-equilibrium operation.
Core monitoring and control under non-equilibrium state conditions are accomplished by operating the core within the limits of the appropriate LCOs, including the limits on AFD, QPTR, and control rod insertion.
APPLICABLE      This LCO precludes core power distributions that violate the following SAFETY ANALYSES  fuel design criteria:
: a.        During a loss of coolant accident (LOCA), the 10 CFR 50.46 acceptance criteria must be met (Ref. 1);
: b.        During a loss of forced reactor coolant flow accident, there must be at least 95% probability at the 95% confidence level (the 95/95 DNB criterion) that the hot fuel rod in the core does not experience a departure from nucleate boiling (DNB) condition;
: c.        During an ejected rod accident, the energy deposition to the fuel must not exceed 280 cal/gm (Ref. 2); and
: d.        The control rods must be capable of shutting down the reactor with a minimum required SDM with the highest worth control rod stuck fully withdrawn (Ref. 3).
Limits on FQ(Z) ensure that the value of the initial total peaking factor assumed in the accident analyses remains valid.
FQ(Z) limits assumed in the LOCA analysis are typically limiting relative to (i.e., lower than) the FQ(Z) limit assumed in safety analyses for other postulated accidents. Therefore, this LCO provides conservative limits for other postulated accidents.
FQ(Z) satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).
(continued)
Watts Bar-Unit 1                              B 3.2-2                            Revision 39, 175, 176 Amendment 21, 142, 143
 
FQ(Z)
B 3.2.1 BASES (continued)
LCO              The Heat Flux Hot Channel Factor, FQ(Z), shall be limited by the following relationships:
CFQ FQ Z      KZ for P  0.5 P
CFQ FQ Z      KZ for P  0.5
 
===0.5 where===
CFQ is the FQ(Z) limit at RTP provided in the COLR, K(Z) is the normalized FQ(Z) limit as a function of core height provided in the COLR, and THERMAL POWER P =
RTP The actual values of CFQ and K(Z) are given in the COLR; however, CFQ is normally a number on the order of 2.5, and K(Z) is a function that looks like the one provided in Figure B 3.2.1-1.
For Relaxed Axial Offset Control operation, FQ(Z) is approximated by FQC(Z) and FQW(Z). Thus, both FQC(Z) and FQW(Z) must meet the preceding limits on FQ(Z)
(Ref 6).
An FQC(Z) evaluation requires obtaining an incore power distribution measurement in MODE 1.
The measured value, FMQ(Z), of FQ(Z) is obtained from the incore power distribution measurement and then corrected for fuel manufacturing tolerances and measurement uncertainty.
If the Moveable Incore Detector System (MIDS) is used to obtain the incore power distribution measurement, then:
FQC (Z) = 1.03 FQM (Z) FQMU where 1.03 is the factor that accounts for the fuel manufacturing tolerances and FQMU, which accounts for flux map measurement uncertainty, is 1.05 (Ref. 4).
When the PDMS is used to obtain the incore power distribution measurement, then:
FQC(Z) = 1.03 FQM (Z) (1+UQ/100)
(continued)
Watts Bar-Unit 1                            B 3.2-3                                Revision 104, 175 Amendment 82, 142
 
FQ(Z)
B 3.2.1 BASES LCO              where 1.03 is the factor that accounts for the fuel manufacturing tolerances (continued)    and the factor (1+UQ/100), which accounts for PDMS measurement uncertainty, is calculated and applied automatically by the BEACONTM software (Ref. 5). In order to be consistent with the LOCA analysis and the uncertainty inputs utilized, a minimum uncertainty of 5 should be used for UQ.
FQC(Z) is an excellent approximation for FQ(Z) when the reactor is at the steady state power at which the incore power distribution measurement was obtained.
The expression for FQW(Z) for a MIDS measurement is:
FQW(Z) = 1.03 FXYM(Z) ([T(Z)]COLR / P) AXY(Z) Rj FQMU for P>0.5 FQW(Z) = 1.03 FXYM(Z) ([T(Z)]COLR / 0.5) AXY(Z) Rj FQMU for P 0.5 The expression for FQW(Z) for a PDMS measurement is:
FQW(Z) = 1.03 FXYM(Z) ([T(Z)]COLR / P) AXY(Z) Rj (1 + UQ / 100) for P>0.5 FQW(Z) = 1.03 FXYM(Z) ([T(Z)]COLR / 0.5) AXY(Z) Rj (1 + UQ / 100) for P0.5 The various factors in these expressions are defined below:
FXYM(Z) is the measured radial peaking factor at axial location Z and is equal to the value of FQM(Z)/PM(Z), where PM(Z) is the measured core average axial power shape.
[T(Z)]COLR is the cycle and burnup dependent function, specified in the COLR, which accounts for power distribution transients encountered during non-equilibrium normal operation. [T(Z)]COLR functions are specified for each analyzed RAOC operating space (i.e. each unique combination of AFD limits and Control Bank Insertion Limits). The [T(Z)]COLR functions account for the limiting non-equilibrium axial power shapes postulated to occur during normal operation for each RAOC operating space. Limiting power shapes at both full and reduced power operation are considered in determining the maximum values of [T(Z)]COLR.
The [T(Z)]COLR functions also account for the following effects: (1) the presence of spacer grids in the fuel assembly, (2) the increase in radial peaking in rodded core planes due to the presence of control rods during non-equilibrium normal operation, (3) the increase in radial peaking that occurs during part-power operation due to reduced fuel and moderator temperatures, and (4) the increase in radial peaking due to non-equilibrium xenon effects. The [T(Z)]COLR functions are normally calculated assuming that the Surveillance is performed at nominal RTP conditions with all shutdown and control rods fully withdrawn, i.e., all rods out (ARO). Surveillance specific [T(Z)]COLR values may be generated for a given surveillance core condition.
(continued)
Watts Bar-Unit 1                          B 3.2-4                    Revision 39, 99, 104, 153, 175 Amendment 21, 82, 142
 
FQ(Z)
B 3.2.1 BASES LCO              P is the THERMAL POWER / RTP.
(continued)
AXY(Z) is a function that adjusts the FQW(Z) Surveillance for differences between the reference core condition assumed in generating the [T(Z)]COLR function and the actual core condition that exists when the Surveillance is performed.
Normally, this reference core condition is 100% RTP, all rods out, and equilibrium xenon. For simplicity, AXY(Z) may be assumed to be 1.0 as this will typically result in an accurate FQW(Z) Surveillance result for a Surveillance that is performed at or near the reference core condition, and an underestimation of the available margin to the FQ limit for Surveillances that are performed at core conditions different from the reference condition. Alternatively, the AXY(Z) function may be calculated using the NRC approved methodology in Reference 7.
FQMU and (1 + UQ/100) are factors that account for measurement uncertainty and 1.03 is a factor that accounts for fuel manufacturing tolerances.
Rj is a cycle and burnup dependent analytical factor specified in the COLR that accounts for potential increases in FQW(Z) between Surveillances. Rj values are provided for each RAOC operating space.
The FQ(Z) limits define limiting values for core power peaking and ensure that the 10 CFR 50.46 acceptance criteria are met during a LOCA (Ref. 1).
This LCO requires operation within the bounds assumed in the safety analyses.
Violating the LCO limits for FQ(Z) could result in unacceptable consequences if a design basis event were to occur while FQ(Z) exceeds its specified limits.
Calculations are performed in the core design process to confirm that the core can be controlled in such a manner during operation that it can stay within the FQ(Z) LCO limits. If FQ(Z) cannot be maintained within the LCO limits, reduction of the core power is required, a more restrictive RAOC operating space must be implemented, or core power limits and AFD limits must be reduced.
APPLICABILITY    The FQ(Z) limits must be maintained in MODE 1 to prevent core power distributions from exceeding the limits assumed in the safety analyses.
Applicability in other MODES is not required because there is either insufficient stored energy in the fuel or insufficient energy being transferred to the reactor coolant to require a limit on the distribution of core power.
ACTIONS          A.1 Reducing THERMAL POWER by  1% RTP for each 1% by which FQC(Z) exceeds its limit, maintains an acceptable absolute power density. FQC(Z) is FQM(Z) multiplied by a factor accounting for manufacturing tolerances and measurement uncertainties. FQM(Z) is the measured value of FQ(Z). The (continued)
Watts Bar-Unit 1                            B 3.2-5                                      Revision 175 Amendment 142
 
FQ(Z)
B 3.2.1 BASES ACTIONS          A.1 (continued)
Completion Time of 15 minutes provides an acceptable time to reduce power in an orderly manner and without allowing the plant to remain in an unacceptable condition for an extended period of time.
The maximum allowable power level initially determined by Required Action A.1 may be affected by subsequent determinations of FQ(Z) and would require power reductions within 15 minutes of the FQC(Z) determination, if necessary, to comply with the decreased maximum allowable power level. Decreases in FQC(Z) would allow increasing the maximum allowable power level and increasing power up to this revised limit. If an FQ Surveillance is performed at 100% RTP conditions, and both FQC(Z) and FQW(Z) exceed their limits, the option to reduce the THERMAL POWER limit in accordance with Required Action B.2.1 instead of implementing a new operating space in accordance with Required Action B.1 will result in a further power reduction after Required Action A.1 has been completed. However, this further power reduction would be permitted to occur over the next 4 hours. In the event the evaluated THERMAL POWER reduction in the COLR for Required Action B.2.1 did not result in a further power reduction (for example, if both Condition A and Condition B were entered at less than 100% RTP conditions),
then the THERMAL POWER level established as a result of completing Required Action A.1 will take precedence, and will establish the effective operating power level limit for the unit until both Conditions A and B are exited.
A.2 A reduction of the Power Range Neutron Flux - High trip setpoints by > 1% for each 1% that THERMAL POWER is limited below RATED THERMAL POWER by Required Action A.1 is a conservative action for protection against the consequences of severe transients with unanalyzed power distributions. The Completion Time of 72 hours is sufficient considering the small likelihood of a severe transient in this time period and the preceding prompt reduction in THERMAL POWER in accordance with Required Action A.1. The maximum allowable Power Range Neutron Flux - High trip setpoints initially determined by Required Action A.2 may be affected by subsequent determinations of FQC(Z) and would require Power Range Neutron Flux - High trip setpoint reductions within 72 hours of the FQC(Z) determination, if necessary, to comply with the decreased maximum allowable Power Range Neutron Flux - High trip setpoints.
Decreases in FQC(Z) would allow increasing the maximum allowable Power Range Neutron Flux - High trip setpoints.
A.3 Reduction in the Overpower T trip setpoints (value of K4) by > 1% for each 1%
that THERMAL POWER is limited below RATED THERMAL POWER by Required Action A.1 is a conservative action for (continued)
Watts Bar-Unit 1                              B 3.2-6                                  Revision 175 Amendment 142
 
FQ(Z)
B 3.2.1 BASES ACTIONS          A.3 (continued) protection against the consequences of severe transients with unanalyzed power distributions. The Completion Time of 72 hours is sufficient considering the small likelihood of a severe transient in this time period, and the preceding prompt reduction in THERMAL POWER in accordance with Required Action A.1. The maximum allowable Overpower T trip setpoints initially determined by Required Action A.3 may be affected by subsequent determinations of FQC(Z) and would require Overpower T trip setpoint reductions within 72 hours of the FQC(Z) determination, if necessary, to comply with the decreased maximum allowable Overpower T trip setpoints. Decreases in FQC(Z) would allow increasing the maximum allowable Overpower T trip setpoints.
A.4 Verification that FQC(Z) has been restored to within its limit, by performing SR 3.2.1.1 and SR 3.2.1.2 prior to increasing THERMAL POWER above the limit imposed by Required Action A.1, ensures that core conditions during operation at higher power levels and future operation are consistent with safety analyses assumptions.
Condition A is modified by a NOTE that requires Required Action A.4 to be performed whenever the Condition is entered prior to increasing THERMAL POWER above the limit of Required Action A.1. The Note also states that SR 3.2.1.2 is not required to be performed if this Condition is entered prior to THERMAL POWER exceeding 75% RTP after a refueling. This ensures that SR 3.2.1.1 and SR 3.2.1.2 (if required) will be performed prior to increasing THERMAL POWER above the limit of Required Action A.1, even when Condition A is exited prior to performing Required Action A.4. Performance of SR 3.2.1.1 and SR 3.2.1.2 are necessary to ensure FQ(Z) is properly evaluated prior to increasing THERMAL POWER.
B.1.1 If it is found that the maximum calculated value of FQ(Z) that can occur during normal maneuvers, FQW(Z), exceeds its specified limits, there exists a potential for FQC(Z) to become excessively high if a normal operational transient occurs.
Implementing a more restrictive RAOC operating space, as specified in the COLR, within the allowed Completion Time of 4 hours will restrict the AFD such that peaking factor limits will not be exceeded during non-equilibrium normal operation. Several RAOC operating spaces, representing successively smaller AFD envelopes and, optionally shallower Control Bank Insertion Limits, may be specified in the COLR. The corresponding T(Z) functions for these operating spaces can be used to determine which RAOC operating space will result in acceptable non-equilibrium operation within the FQW(Z) limits.
(continued)
Watts Bar-Unit 1                            B 3.2-7                                      Revision 175 Amendment 142
 
FQ(Z)
B 3.2.1 BASES ACTIONS          B.1.2 (continued)
If it is found that the maximum calculated value of FQ(Z) that can occur during normal maneuvers, FQW(Z), exceeds its specified limits, there exists a potential for FQC(Z) to become excessively high if a normal operational transient occurs.
As discussed above, Required Action B.1.1 requires that a new RAOC operating space be implemented to restore FQW(Z) to within its limits. Required Action B.1.2 requires that SR 3.2.1.1 and SR 3.2.1.2 be performed if control rod motion occurs as a result of implementing the new RAOC operating space in accordance with Required Action B.1.1. The performance of SR 3.2.1.1 and SR 3.2.1.2 is necessary to ensure FQ(Z) is properly evaluated after any rod motion resulting from the implementation of a new RAOC operating space in accordance with Required Action B.1.1.
B.2.1 When FQW(Z) exceeds its limit, Required Action B.2 may be implemented instead of Required Action B.1. Required Action B.2.1 limits THERMAL POWER to less than RATED THERMAL POWER by the amount specified in the COLR. It also requires reductions in the AFD limits by the amount specified in the COLR. This maintains an acceptable absolute power density relative to the maximum power density value assumed in the safety analyses.
If the required FQW(Z) margin improvement exceeds the margin improvement available from the pre-analyzed THERMAL POWER and AFD reductions provided in the COLR, then THERMAL POWER must be further reduced to less than or equal to 50% RTP. In this case, reducing THERMAL POWER to less than or equal to 50% RTP will provide additional margin in the transient FQ by the required change in THERMAL POWER and the increase in the FQ limit. This will ensure that the FQ limit is met during transient operation that may occur at or below 50% RTP.
The Completion Time of 4 hours provides an acceptable time to reduce the THERMAL POWER and AFD limits in an orderly manner to preclude entering an unacceptable condition during future non-equilibrium operation. The limit on THERMAL POWER initially determined by Required Action B.2.1 may be affected by subsequent determinations of FQW(Z) and may require further power reductions, if necessary, to comply with the decreased THERMAL POWER limit.
If the 4 hour Completion Time had expired since the Condition B entry had been made, then Condition C would be required to be entered. Decreases in FQW(Z) during subsequent F QW(Z) determinations, however would allow increasing the THERMAL POWER limit and increasing THERMAL POWER up to this revised limit.
Required Action B.2.1 is modified by a NOTE that states Required Action B.2.4 shall be completed whenever Required Action B.2.1 is performed prior to increasing THERMAL POWER above the limit of Required Action B.2.1.
(continued)
Watts Bar-Unit 1                            B 3.2-8                                    Revision 175 Amendment 142
 
FQ(Z)
B 3.2.1 BASES ACTIONS          B.2.1 (continued)
Required Action B.2.4 requires the performance of SR 3.2.1.1 and SR 3.2.1.2 prior to increasing THERMAL POWER above the limit established by Required Action B.2.1. The Note ensures that the SRs will be performed even if Condition B may be exited prior to performing Required Action B.2.4. The performance of SR 3.2.1.1 and SR 3.2.1.2 is necessary to ensure FQ(Z) is properly evaluated prior to increasing THERMAL POWER.
If an FQ surveillance is performed at 100% RTP conditions, and both FQC(Z) and FQW(Z) exceed their limits, the option to reduce the THERMAL POWER limit in accordance with proposed Required Action B.2.1 instead of implementing a new operating space in accordance with proposed Required Action B.1, will result in a further power reduction after Required Action A.1 has been completed. However, this further power reduction would be permitted to occur over the next 4 hours. In the event the evaluated THERMAL POWER reduction in the COLR for proposed Required Action B.2.1 did not result in a further power reduction (for example, if both Condition A and Condition B were entered at less than 100%
RTP conditions), then the THERMAL POWER level established as a result of completing Required Action A.1 will take precedence and will establish the effective operating power level limit for the unit until both Conditions A and B are exited.
B.2.2 A reduction of the Power Range Neutron Flux - High trip setpoints by  1% for each 1% by which the maximum allowable power is reduced is a conservative action for protection against the consequences of severe transients with unanalyzed power distributions. The Completion Time of 72 hours is sufficient considering the small likelihood of a severe transient in this time period and the preceding prompt reduction in the THERMAL POWER limit and AFD limits in accordance with Required Action B.2.1.
B.2.3 Reduction in the Overpower T trip setpoints value of K4 by  1% for each 1% by which the maximum allowable power is reduced is a conservative action for protection against the consequences of severe transients with unanalyzed power distributions. The Completion Time of 72 hours is sufficient considering the small likelihood of a severe transient in this time period, and the preceding prompt reduction in the THERMAL POWER limit and AFD limits in accordance with the Required Action B.2.1.
(continued)
Watts Bar-Unit 1                          B 3.2-9                                        Revision 175 Amendment 142
 
FQ(Z)
B 3.2.1 BASES ACTIONS          B.2.4 (continued)
Verification that FQC(Z) and FQW(Z) have been restored to within limit, by performing SR 3.2.1.1 and SR 3.2.1.2 prior to increasing THERMAL POWER above the maximum allowable power limit imposed by Required Action B.2.1, ensures that core conditions during operation at higher power levels and future operation are consistent with safety analyses assumptions.
C.1 If Required Actions A.1 through A.4 or B.1.1 through B.2.4 are not met within their associated Completion Times, the plant must be placed in a mode or condition in which the LCO requirements are not applicable. This is done by placing the plant in at least MODE 2 within 6 hours.
This allowed Completion Time is reasonable based on operating experience regarding the amount of time it takes to reach MODE 2 from full power operation in an orderly manner and without challenging plant systems.
SURVEILLANCE    SR 3.2.1.1 REQUIREMENTS
.                Verification that FQC(Z) is within its specified limits involves increasing FQM(Z) to allow for manufacturing tolerance and measurement uncertainties in order to obtain FQC(Z). Specifically, FQM(Z) is the measured value of FQ(Z) obtained from an incore power distribution measurement.
If the Movable Incore Detector System is used to obtain the incore power distribution measurement, then:
FQC(Z) = 1.03 FQM (Z) FQMU where 1.03 is the factor that accounts for the fuel manufacturing tolerances and FQMU, which accounts for flux map measurement uncertainty, is 1.05 (Ref. 4).
When the PDMS is used to obtain the incore power distribution measurement, then:
FQC(Z) = 1.03 FQM (Z) (1+UQ /100) where 1.03 is the factor that accounts for the fuel manufacturing tolerances and the factor (1+UQ/100), which accounts for PDMS measurement uncertainty, is calculated and applied automatically by the BEACON software (Ref.5). In order to be consistent with the LOCA analysis and the uncertainty inputs utilized, a minimum uncertainty of 5 should be used for UQ.
(continued)
Watts Bar-Unit 1                          B 3.2-9a                              Revision 104, 153 175 Amendment 82, 142
 
FQ(Z)
B 3.2.1 BASES SURVEILLANCE    SR 3.2.1.1 (continued)
REQUIREMENTS FQC(Z) is then compared to its specified limits. The limit with which FQC(Z) is compared varies inversely with power above 50% RTP and directly with a function called K(Z) provided in the COLR.
Performing this Surveillance in MODE 1 prior to exceeding 75% RTP following a refueling ensures that some determination of FQC(Z) is made prior to achieving a significant power level where peak linear heat rate could approach the limits assumed in the safety analyses.
If THERMAL POWER has been increased by > 10% RTP since the initial or most recent determination of FQC(Z), another evaluation of this factor is required 24 hours after achieving equilibrium conditions at this higher power level (to ensure that FQC(Z) values are being reduced sufficiently with power increase to stay within the LCO limits). Equilibrium conditions are achieved when the core is sufficiently stable at the intended operating conditions required to perform the surveillance.
The allowance of up to 24 hours after achieving equilibrium conditions at the increased THERMAL POWER level to complete the next FQC(Z) surveillance applies to situations where the FQC(Z) has already been measured at least once at a reduced THERMAL POWER level. The observed margin in the previous surveillance will provide assurance that increasing power up to the next plateau will not exceed the FQ limit, and that the core is behaving as designed.
This Frequency condition is not intended to require verification of these parameters after every 10% increase in RTP above the THERMAL POWER at which the last verification was performed. It only requires verification after a THERMAL POWER is achieved for extended operation that is 10% higher than the THERMAL POWER at which FQC(Z) was last measured.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.2.1.2 The nuclear design process includes calculations performed to determine that the core can be operated within the FQ(Z) limits. Because incore power distribution measurements are taken in steady state conditions, the variations in power distribution resulting from normal operational maneuvers are not present in the measured data. These variations are, however, conservatively calculated by considering a wide range of unit maneuvers in normal operation.
(continued)
Watts Bar-Unit 1                            B 3.2-9b                  Revision 38, 99, 104, 162, 175 Amendment 82, 132, 142
 
FQ(Z)
B 3.2.1 BASES SURVEILLANCE    SR 3.2.1.2 (continued)
REQUIRMENTS The measured FQ (Z) can be determined through a synthesis of the measured planar radial peaking factors, FXYM(Z), and the measured core average axial power shape, PM(Z). Thus, FQC(Z) is given by the following expressions:
FQC(Z) = 1.03 FXYM(Z) PM(Z) FQMU = 1.03 FQM(Z) FQMU                          [MIDS]
or FQC(Z) = 1.03 FXYM(Z) PM(Z) (1 + UQ/100) = 1.03 FQM(Z) (1 + UQ/100)          [PDMS]
For RAOC operation, the analytical [T(Z)]COLR functions, specified in the COLR for each RAOC operating space, are used together with the measured FXY(Z) values to estimate FQ(Z) for non-equilibrium operation within the RAOC operating space. When the FXY(Z) values are measured at HFP ARO conditions (P equals 1.0 and (AXY(Z) equals 1.0), FQW(Z) is given by the following expressions:
FQW(Z) = 1.03 FXYM(Z) [T(Z)]COLR Rj FQMU                                    [MIDS]
or FQW(Z) = 1.03 FXYM(Z) [T(Z)]COLR Rj (1 + UQ/100)                            [PDMS]
Non-equilibrium operation can result in significant changes to the axial power shape. To a lesser extent, non-equilibrium operation can increase the radial peaking factors, FXY(Z), through control rod insertion and through reduced Doppler and moderator feedback at part-power conditions.
The [T(Z)]COLR functions quantify these effects for the range of power shapes, control rod insertion, and power levels characteristic of the operating space.
Multiplying [T(Z)]COLR by the measured full power, un-rodded FXYM(Z) value, and the factors that account for manufacturing and measurement uncertainties gives FQW(Z) , the maximum total peaking factor postulated for non-equilibrium RAOC operation.
The limit with which FQW(Z) is compared varies inversely with power above 50%
RTP and directly with the function K(Z) provided in the COLR.
The [T(Z)]COLR functions are specified in the COLR for discrete core elevations.
Flux map data are typically taken for 30 to 75 core elevations. FQW(Z) evaluations are not applicable for axial core regions, measured in percent of core height:
: a.      Lower core region, from 0 to 10% inclusive,
: b.      Upper core region, from 90 to 100% inclusive, (continued)
Watts Bar-Unit 1                          B 3.2-9c                        Revision 38, 99, 104, 162 Amendment 82, 132
 
FQ(Z)
B 3.2.1 BASES SURVIELLANCE    SR 3.2.1.2 (continued)
REQUIRMENTS
: c.        Grid plane regions, +/-2% inclusive, and
: d.        Core plane regions, within 2% of the bank demand positions of the control banks.
These regions of the core are excluded from the evaluation because of the low probability that they would be more limiting in the safety analysis and because of difficulty of making a precise measurement in these regions. The excluded regions at the top and bottom of the core are specified in the COLR and are defined to ensure that the minimum margin location is adequately surveilled. A slightly smaller exclusion zone may be specified, if necessary, to include the limiting margin location in the surveilled region of the core.
SR 3.2.1.2 requires a Surveillance of FQW(Z) during the initial startup following each refueling within 24 hours after exceeding 75% RTP. THERMAL POWER levels below 75% are typically non-limiting with respect to the limit for FQW(Z).
Furthermore, startup physics testing and flux symmetry measurements, also performed at low power, provide confirmation that the core is operating as expected. This Frequency ensures that verification of FQW(Z) is performed prior to extended operation at power levels where the maximum permitted peak LHR could be challenged and that the first verified performance of SR 3.2.1.2 after a refueling is performed at a power level high enough to provide a high level of confidence in the accuracy of the Surveillance result.
Equilibrium conditions are achieved when the core is sufficiently stable at the intended operating conditions required to perform the Surveillance.
If a previous Surveillance of FQW(Z) was performed at part power conditions, SR 3.2.1.2 also requires that FQW(Z) be verified at power levels  10% RTP above the THERMAL POWER of its last verification within 24 hours after achieving equilibrium conditions. This ensures that FQW(Z) is within its limit using radial peaking factors measured at the higher power level.
The allowance of up to 24 hours after achieving equilibrium conditions will provide a more accurate measurement of FQW(Z) by allowing sufficient time to achieve equilibrium conditions and obtain the power distribution measurement.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
Watts Bar-Unit 1                          B 3.2-9d                                Revision 162, 175 Amendment 132, 142
 
FQ(Z)
B 3.2.1 BASES (continued)
REFERENCES        1. Title 10, Code of Federal Regulations, Part 50.46, "Acceptance Criteria for Emergency Core Cooling Systems for Light Water Nuclear Power Reactors."
: 2. Regulatory Guide 1.77, Rev. 0, "Assumptions Used for Evaluating a Control Rod Ejection Accident for Pressurized water Reactors,"
May1974.
: 3. Title 10, Code of Federal Regulations, Part 50, Appendix A, "General Design Criteria for Nuclear Power Plants," GDC 26, "Reactivity Control System Redundancy and Capability."
: 4. WCAP-7308-L-P-A, "Evaluation of Nuclear Hot Channel Factor Uncertainties," June 1988.
: 5. WCAP-12472-P-A, BEACON Core Monitoring and Operations Support System, August 1994.
: 6. WCAP-10216-P-A, Rev. 1A, Relaxation of Constant Axial Offset Control (and) FQ Surveillance Technical Specification, February 1994.
: 7. WCAP-17661-P-A, Improved RAOC and CAOC FQ Surveillance Technical Specifications, February 2019.
Watts Bar-Unit 1                      B 3.2-9e                              Revision 18, 175 Amendment 11, 142
 
FQ(Z)
B 3.2.1 BASES (continued) 1.1 HIS AREA 1
0.9 0.8 MEN                                OEM Y
Aa  0.7                                                                          MMMOME ENMNM MEE N W 0.6              .......MISS..                    ..      ....
M U1                          'mm 0.5            MEN.                                      MEE, b ILM I              OR      MEN ImmEE US    ION O LY.      MIN          MMME 0.4                  ME                                    ...
w                                    *
* OR                    MIS 06 z
OP                  MEN:'
0.3 I                          EE 0.2      M            M-0.1 IN
                  *C
                    . OMEN YWWY 0 'M"MMM=WY 0  1    2      3      4      5      6    7      8    9    10    11            12 Height (ft)
Figure B 3.2.1-1 (page 1 of 1)
K(Z) - Normalized FQ(Z) as a Function of Core Height Watts Bar-Unit 1                                B 3.2-10                                          Revision 175 Amendment 142
 
F NH B 3.2.2 BASES BACKGROUND      Operation outside the LCO limits may produce unacceptable consequences (continued)    if a DNB limiting event occurs. The DNB design basis ensures that there is no overheating of the fuel that results in possible cladding perforation with the release of fission products to the reactor coolant.
APPLICABLE      Limits on FNH preclude core power distributions that exceed the following fuel SAFETY          design limits:
ANALYSES
: a.        There must be at least 95% probability at the 95% confidence level (the 95/95 DNB criterion) that the hottest fuel rod in the core does not experience a DNB condition;
: b.        During a loss of coolant accident (LOCA), the 10 CFR 50.46 acceptance criteria must be met (Ref. 3);
: c.        During an ejected rod accident, the energy deposition to the fuel must not exceed 280 cal/gm (Ref. 1); and
: d.        Fuel design limits required by GDC 26 (Ref. 2) for the condition when control rods must be capable of shutting down the reactor with a minimum required SDM with the highest worth control rod stuck fully withdrawn.
For transients that may be DNB limited, FNH is a significant core parameter. The limits on FNH ensure that the DNB design basis is met for normal operation, operational transients, and any transients arising from events of moderate frequency. The DNB design basis is met by limiting the minimum local DNB heat flux ratio to a value which satisfies the 95/95 criterion for the DNB correlation used. Refer to the Bases for the Reactor Core Safety Limits, B 2.1.1 for a discussion of the applicable DNBR limits. The W-3 Correlation with a DNBR limit of 1.3, or the ABB-NV correlation with a DNBR limit of 1.13, is applied in the heated region below the first mixing vane grid. In addition, the W-3 or WLOP DNB correlations are applied in the analysis of accident conditions where the system pressure is below the range of the WRB-1 correlation for VANTAGE 5H and VANTAGE+ fuel or the WRB-2M correlation for RFA-2 fuel with IFMs. For system pressures in the range of 500 to 1000 psia, the W-3 correlation DNBR limit is 1.45 instead of 1.3. For system pressures in the range of 185 to 1800 psia, the WLOP correlation DNBR limit is 1.18.
(continued)
Watts Bar-Unit 1                            B 3.2-12                    Revision 13, 39, 59, 140, 176 Amendment 7, 21, 46, 143
 
F NH B 3.2.2 BASES APPLICABLE      Application of these criteria provides assurance that the hottest fuel rod in the SAFETY ANALYSES  core does not experience a DNB.
(continued)
The allowable FNH limit increases with decreasing power level. This functionality in FNH is included in the analyses that provide the Reactor Core Safety Limits (SLs) of SL 2.1.1. Therefore, any DNB events in which the calculation of the core limits is modeled implicitly use this variable value of FNH in the analyses.
Likewise, all transients that may be DNB limited are assumed to begin with an initial FNH as a function of power level defined by the COLR limit equation.
The LOCA safety analyses that verify compliance with the 10 CFR 50.46 acceptance criteria (Ref. 3) model FNH as well as the Nuclear Heat Flux Hot Channel Factor (FQ(Z)).
The fuel is protected in part by Technical Specifications, which ensure that the initial conditions assumed in the safety and accident analyses remain valid. The following LCOs ensure this: LCO 3.2.3, "AXIAL FLUX DIFFERENCE (AFD),"
LCO 3.2.4, "QUADRANT POWER TILT RATIO (QPTR)," LCO 3.1.7, "Control Bank Insertion Limits," LCO 3.2.2, "Nuclear Enthalpy Rise Hot Channel Factor (FNH)," and LCO 3.2.1, "Heat Flux Hot Channel Factor (FQ(Z))."
FNH and FQ(Z) are measured periodically using either the Movable Incore Detector System or the PDMS (Ref. 5). Measurements are generally taken with the core at, or near, steady state conditions. Core monitoring and control under transient conditions (Condition 1 events) are accomplished by operating the core within the limits of the LCOs on AFD, QPTR, and Bank Insertion Limits.
FNH satisfies Criterion 2 of the NRC Policy Statement.
LCO              FNH shall be maintained within the limits of the relationship provided in the COLR.
The FNH limit identifies the coolant flow channel with the maximum enthalpy rise.
This channel has the least heat removal capability and thus the highest probability for a DNB.
The limiting value of FNH, described by the equation contained in the COLR, is the design radial peaking factor used in the unit safety analyses.
A power multiplication factor in this equation includes an additional margin for higher radial peaking from reduced thermal feedback and greater control rod insertion at low power levels. The limiting value of FNH is allowed to increase 0.3% for every 1% RTP reduction in THERMAL POWER.
(continued)
Watts Bar-Unit 1                            B 3.2-13                            Revision 39, 104, 176 Amendment 21, 82, 143
 
QPTR B 3.2.4 B 3.2 POWER DISTRIBUTION LIMITS B 3.2.4 QUADRANT POWER TILT RATIO (QPTR)
BASES BACKGROUND        The QPTR limit ensures that the gross radial power distribution remains consistent with the design values used in the safety analyses. Precise radial power distribution measurements are made during startup testing, after refueling, and periodically during power operation.
The power density at any point in the core must be limited so that the fuel design criteria are maintained. Together, LCO 3.2.3, "AXIAL FLUX DIFFERENCE (AFD)," LCO 3.2.4, and LCO 3.1.7, "Control Rod Insertion Limits," provide limits on process variables that characterize and control the three dimensional power distribution of the reactor core. Control of these variables ensures that the core operates within the fuel design criteria and that the power distribution remains within the bounds used in the safety analyses.
APPLICABLE        This LCO precludes core power distributions that violate the SAFETY ANALYSES    following fuel design criteria:
: a.        During a loss of coolant accident, the 10 CFR 50.46 criteria must be met (Ref. 1);
: b.        During a loss of forced reactor coolant flow accident, there must be at least 95% probability at the 95% confidence level (the 95/95 departure from nucleate boiling (DNB) criterion) that the hot fuel rod in the core does not experience a DNB condition;
: c.        During an ejected rod accident, the energy deposition to the fuel must not exceed 280 cal/gm (Ref. 2); and
: d.        The control rods must be capable of shutting down the reactor with a minimum required SDM with the highest worth control rod stuck fully withdrawn (Ref. 3).
The LCO limits on the AFD, the QPTR, the Heat Flux Hot Channel Factor (FQ(Z)), the Nuclear Enthalpy Rise Hot Channel Factor (FNH), rod group alignment, sequence, overlap, and control bank insertion are established to preclude core power distributions that exceed the safety analyses limits.
(continued)
Watts Bar-Unit 1                              B 3.2-23                                Revision 104, 176 Amendment 82, 143
 
RTS Instrumentation B 3.3.1 Bases SURVEILLANCE    SR 3.3.1.15 (continued)
REQUIREMENTS WCAP-14036-P-A, Revision 1, Elimination of Periodic Protection Channel Response Time Tests (Reference 12), provides the basis and methodology for using allocated signal processing and actuation logic response times in the overall verification of the protection system channel response time. The allocations for sensor, signal conditioning and actuation logic response times must be verified prior to placing the component in operational service and re-verified following maintenance that may adversely affect response time. In general, electrical repair work does not impact response time provided the parts used for repair are of the same type and value. Specific components identified in the WCAP may be replaced without verification testing. One example where response time could be affected is replacing the sensing assembly of a transmitter.
The response time may be verified for components that replace the components that were previously evaluated in Ref. 11 and Ref. 12, provided that the components have been evaluated in accordance with the NRC approved methodology as discussed in Attachment 1 to TSTF-569, Methodology to Eliminate Pressure Sensor and Protection Channel (for Westinghouse Plants only)
Response Time Testing, (Ref. 17).
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.3.1.15 is modified by a Note stating that neutron detectors are excluded from RTS RESPONSE TIME testing. This Note is necessary because of the difficulty in generating an appropriate detector input signal. Excluding the detectors is acceptable because the principles of detector operation ensure a virtually instantaneous response.
REFERENCES      1.        Watts Bar FSAR, Section 6.0, "Engineered Safety Features"
: 2.        Watts Bar FSAR, Section 7.0, "Instrumentation and Controls"
: 3.        Watts Bar FSAR, Section 15.0, "Accident Analysis."
: 4.        Institute of Electrical and Electronic Engineers, IEEE-279-1971, "Criteria for Protection Systems for Nuclear Power Generating Stations,"
April 5, 1972.
: 5.        10 CFR Part 50.49, "Environmental Qualifications of Electric Equipment Important to Safety for Nuclear Power Plants."
: 6.        WCAP-12096, Rev. 7, "Westinghouse Setpoint Methodology for Protection System, Watts Bar 1 and 2," March 1997.
(continued)
Watts Bar-Unit 1                            B 3.3-51                        Revision 34, 90, 162, 170 Amendment 24, 68, 132, 141
 
RTS Instrumentation B 3.3.1 Bases REFERENCES      7. WCAP-10271-P-A, Supplement 1, and Supplement 2, Rev. 1, (continued)        "Evaluation of Surveillance Frequencies and Out of Service Times for the Reactor Protection Instrumentation System," May 1986 and June 1990.
: 8. Watts Bar Technical Requirements Manual, Section 3.3.1, "Reactor Trip System Response Times."
: 9. Evaluation of the applicability of WCAP-10271-P-A, Supplement 1, and Supplement 2, Revision 1, to Watts Bar, Westinghouse Letter WAT-D-10128.
: 10. ISA-DS-67.04, 1982, "Setpoint for Nuclear Safety Related Instrumentation Used in Nuclear Power Plants."
: 11. WCAP-13632-P-A Revision 2, Elimination of Pressure Sensor Response Time Testing Requirements, January 1996
: 12. WCAP-14036-P-A, Revision 1, Elimination of Periodic Protection Channel Response Time Tests, October 1998.
: 13. WCAP-16067-P, Rev. 0, RCS Flow Measurement Using Elbow Tap Methodology at Watts Bar Unit 1, April 2003.
: 14. WCAP-14333 P-A, Revision 1, Probabilistic Risk Analysis of the RPS and ESFAS Test Times and Completion Times, October 1998.
: 15. WCAP-15376-P-A, Revision 1, Risk Informed Assessment of the RTS and ESFAS Surveillance Test Intervals and Reactor Trip Breaker Test and Completion Times, March 2003.
: 16. WCAP-12472-P-A,BEACON Core Monitoring and Operations Support System, August 1994.
: 17. Attachment 1 to TSTF-569, Methodology to Eliminate Pressure Sensor and Protection Channel (for Westinghouse Plants only) Response Time Testing.
Watts Bar-Unit 1                    B 3.3-52                Revision 13, 34, 60, 90, 104, 170 Amendment 24, 47, 68, 82, 141
 
ESFAS Instrumentation B 3.3.2 BASES (continued)
APPLICABLE        Each of the analyzed accidents can be detected by one or more ESFAS SAFETY ANALYSES,  Functions. One of the ESFAS Functions is the primary actuation signal LCO, and          for that accident. An ESFAS Function may be the primary actuation signal for APPLICABILITY    more than one type of accident. An ESFAS Function may also be a secondary, or backup, actuation signal for one or more other accidents. For example, Pressurizer PressureLow is a primary actuation signal for small loss of coolant accidents (LOCAs) and a backup actuation signal for steam line breaks (SLBs) outside containment. Functions such as manual initiation, not specifically credited in the accident safety analysis, are qualitatively credited in the safety analysis and the NRC staff approved licensing basis for the unit. These Functions may provide protection for conditions that do not require dynamic transient analysis to demonstrate Function performance. These Functions may also serve as backups to Functions that were credited in the accident analysis (Ref. 3).
The LCO requires all instrumentation performing an ESFAS Function to be OPERABLE. Failure of any instrument renders the affected channel(s) inoperable and reduces the reliability of the affected Functions.
The LCO generally requires OPERABILITY of four or three channels in each instrumentation function and two channels in each logic and manual initiation function. The two-out-of-three and the two-out-of-four configurations allow one channel to be tripped during maintenance or testing without causing an ESFAS initiation. Two logic or manual initiation channels are required to ensure no single random failure disables the ESFAS.
The required channels of ESFAS instrumentation provide unit protection in the event of any of the analyzed accidents. ESFAS protection functions are as follows:
: 1. Safety Injection Safety Injection (SI) provides two primary functions:
: 1.        Primary side water addition to ensure maintenance or recovery of reactor vessel water level (coverage of the active fuel for heat removal, clad integrity, and compliance with the 10 CFR 50.46 acceptance criteria (Ref. 22); and
: 2.        Boration to ensure recovery and maintenance of SDM (keff < 1.0).
(continued)
Watts Bar-Unit 1                              B 3.3-57                                      Revision 176 Amendment 143
 
ESFAS Instrumentation B 3.3.2 BASES APPLICABLE      c. Auxiliary Feedwater-Safety Injection SAFETY ANALYSES, LCO, and            An SI signal starts the motor driven and turbine driven AFW APPLICABILITY      pumps. The AFW initiation functions are the same as the (continued)          requirements for their SI function. Therefore, the requirements are not repeated in Table 3.3.2-1. Instead, Function 1, SI, is referenced for all initiating functions and requirements.
: d. Auxiliary Feedwater-Loss of Offsite Power A loss of offsite power to the RCP buses will be accompanied by a loss of reactor coolant pumping power and the subsequent need for some method of decay heat removal. The loss of offsite power is detected by a voltage drop on each 6.9 kV shutdown board. Loss of power to either 6.9 kV shutdown board will start the turbine driven AFW pump to ensure that enough water is available to serve as the heat sink for reactor decay heat and sensible heat removal following the reactor trip.
Functions 6.a through 6.d (except the loop T input to the trip time delay) must be OPERABLE in MODES 1, 2, and 3 to ensure that the SGs remain the heat sink for the reactor. SG Water LevelLow Low in any operating SG will cause the motor driven AFW pumps to start. The system is aligned so that upon a start of the pump, water immediately begins to flow to the SGs.
SG Water LevelLow Low in any two operating SGs will cause the turbine driven pumps to start. These Functions do not have to be OPERABLE in MODES 5 and 6 because there is not enough heat being generated in the reactor to require the SGs as a heat sink. In MODE 4, AFW actuation does not need to be OPERABLE because either AFW or residual heat removal (RHR) will already be in operation to remove decay heat or sufficient time is available to manually place either system in operation.
: e. Auxiliary Feedwater-Trip Of All Main Feedwater A trip of all main feed pumps is an indication of a loss of MFW and the subsequent need for some method of decay heat and sensible heat removal to bring the reactor back to no load temperature and pressure. Each turbine driven MFW pump (TDMFWP) is equipped with one pressure switch mounted on the control oil line for the speed control system. A low pressure signal from this pressure switch indicates a trip of that pump.
The electric motor driven standby main feedwater pump (SBMFWP) trip channel is provided by breaker contacts from the supply breaker of the motor driven SBMFWP in the AFW start logic. The breaker contacts monitor the SBMFWP and close (continued)
Watts Bar-Unit 1                B 3.3-76                                      Revision 177 Amendment 148
 
ESFAS Instrumentation B 3.3.2 BASES APPLICABLE      e. Auxiliary Feedwater-Trip Of All Main Feedwater SAFETY ANALYSES,    Pumps (continued)
LCO, and APPLICABILITY      upon the opening of the breaker, indicating that pump has tripped. The trip of both TDMFWPs and the SBMFWP pump will start the motor driven and turbine driven AFW pumps to ensure that enough water is available to act as the heat sink for the reactor.
This Function must be OPERABLE in MODES 1 and 2 in accordance with the applicable Tech Specs Notes to ensure that at least one SG is provided with sufficient water to serve as the heat sink to remove reactor decay heat and sensible heat in the event of an accident.
During startup in MODE 2 the SBMFP will be providing feedwater to the steam generators instead of the TDMFPs as sufficient steam is not yet available. In the unlikely event that the SBMFP trips during this time, the anticipatory AFW auto-start circuitry will actuate starting both the motor driven AFW pumps and the turbine driven AFW pump.
In MODE 1, at approximately 10% RTP, the first TDMFWP is placed in service. Once the first TDMFWP is placed in service, the SBMFWP will be removed from service. Under these conditions, a trip of the sole operating TDMFWP would generate an anticipatory AFW auto-start signal causing all three AFW pumps to start. Once the first TDMFWP is supplying feedwater to the steam generators, the SBMFWP trip channel shall be placed in trip status. This ensures during normal operation, should the TDMFWP(s) in operation trip, the AFW auto start function will actuate starting all of the AFW pumps.
Once the SBMFWP is in service, the SBMFWP trip channel is capable of providing an input signal to the AFW start signal upon the trip of the SBMFWP and the TDMFWP will be removed from service. The SBMFWP will then be the only supply of feedwater to the steam generators going from MODE 1 to MODE 2. In the unlikely event that the SBMFWP trips during this time, the anticipatory AFW auto start circuitry will actuate starting both the motor driven AFW pumps and the turbine driven AFW pump.
(continued)
Watts Bar-Unit 1              B 3.3-77                            Revision 96, 109, 177 Amendment No. 75, 148
 
ESFAS Instrumentation B 3.3.2 BASES APPLICABLE      e. Auxiliary Feedwater-Trip Of All Main Turbine Driven Feedwater SAFETY ANALYSES    Pumps (continued)
LCO, and APPLICABILITY      Mode 1 applicability for the TDMFW pumps allow entry into LCO 3.3.2, Condition J to be suspended for up to 4 hours when placing a TDMFW pump in service or removing a TDMFW pump from service. This provision will reduce administrative burden on the plant. Plant safety is not compromised during this short period because the safety grade AFW auto-start channels associated with steam generator low-low levels are operable.
In Mode 3, decay heat and sensible heat removal is sufficiently low that adequate time is available for the operator to manually activate the AFW system if a loss of MFW were to occur.
In Modes 4 and 5, the RCPs and MFW pumps are normally shut down with decay heat and sensible heat removed by the RHR system. Therefore, neither pump trip is indicative of a condition requiring automatic AFW initiation.
: f. Auxiliary Feedwater-Pump Suction Transfer on Suction Pressure-Low A low pressure signal in the AFW pump suction line protects the AFW pumps against a loss of the normal supply of water for the pumps, the CST. Three pressure switches are located on each motor driven AFW pump suction line from the CST. A low pressure signal sensed by two switches of a set will cause the emergency supply of water for the respective pumps to be aligned. ERCW (safety grade) is then lined up to supply the AFW pumps to ensure an adequate supply of water for the AFW System to maintain at least one of the SGs as the heat sink for reactor decay heat and sensible heat removal. Since the detectors are located in an area not affected by HELBs or high radiation, they will not experience any adverse environmental conditions and the Trip Setpoint reflects only steady state instrument uncertainties. This Function must be OPERABLE in MODES 1, 2, 3, and 4, when the steam generators are relied on to remove decay heat from the reactor, to ensure a safety grade supply of water for the AFW System to maintain the SGs as the heat sink for the reactor. This Function does not have to be OPERABLE in MODES 5 and 6 because there is not enough heat being generated in the reactor to require the SGs as a heat sink. In MODE 4, AFW automatic suction transfer does not need to be OPERABLE because RHR will already be in operation, or sufficient time is available to place RHR in operation, to remove decay heat.
(continued)
Watts Bar-Unit 1            B 3.3-78                                Revision 123, 177 Amendment 104, 148
 
ESFAS Instrumentation B 3.3.2 BASES ACTIONS        O.1 and O.2 (continued)
Failure to restore the inoperable channel to OPERABLE status or place it in the tripped condition within 72 hours requires the plant to be placed in MODE 3 within the following 6 hours. The allowed Completion Time of 6 hours is reasonable, based on operating experience, to reach MODE 3 from full power conditions in an orderly manner and without challenging plant systems. In MODE 3, these functions are no longer required OPERABLE.
The Required Actions have been modified by a Note that allows placing the inoperable channel in the bypassed condition for up to 12 hours while performing routine surveillance testing of the other channels. The 12 hour time limit is justified in Reference 10 and 17.
P.1 and P.2 Condition P applies to the AFW pump start on trip of the SBMFWP pump.
The OPERABILITY of the AFW System must be assured by allowing automatic start of the AFW System pumps. If the SBMFWP trip channel is inoperable, 48 hours are allowed to restore that channel to OPERABLE status or place it in the tripped condition. Failure to restore the inoperable channel to OPERABLE status or place it in the tripped condition within 48 hours requires the plant to be placed in MODE 3 within the following 6 hours. The allowed Completion Time of 6 hours is reasonable, based on operating experience, to reach MODE 3 from full power conditions in an orderly manner and without challenging plant systems. In MODE 3, the plant does not have any analyzed transients or conditions that require the explicit use of the protection function noted above. The allowance of 48 hours to restore the channel to OPERABLE status or place it in the tripped condition is justified in Reference 7.
SURVEILLANCE    The SRs for each ESFAS Function are identified by the SRs column of REQUIREMENTS    Table 3.3.2-1.
A Note has been added to the SR Table to clarify that Table 3.3.2-1 determines which SRs apply to which ESFAS Functions.
Note that each channel of process protection supplies both trains of the ESFAS.
When testing channel I, train A and train B must be examined. Similarly, train A and train B must be examined when testing channel II, channel III, and channel IV. The CHANNEL CALIBRATION and COTs are performed in a manner that is consistent with the assumptions used in analytically calculating the required channel accuracies.
The protection Functions associated with the EAGLE-21TM Process Protection System have an installed bypass capability, and may be tested in either the trip or bypass mode, as approved in Reference 7. When testing is performed in the bypass mode, the SSPS input relays are not operated, as justified in Reference
: 10. The input relays are checked during the CHANNEL CALIBRATION.
(continued)
Watts Bar-Unit 1                            B 3.3-93                      Revision 90, 108, 162, 177 Amendment 68, 132, 148
 
EFAS Instrumentation B 3.3.2 BASES SURVEILLANCE    SR 3.3.2.10 (continued)
REQUIREMENTS The response time may be verified for components that replace the components that were previously evaluated in Ref. 15 and Ref. 16, provided that the components have been evaluated in accordance with the NRC approved methodology as discussed in Attachment 1 to TSTF-569, Methodology to Eliminate Pressure Sensor and Protection Channel (for Westinghouse Plants only) Response Time Testing, (Ref. 21).
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
This SR is modified by a Note indicating that the SR should be deferred until suitable test conditions are established. This deferral is required because there may be insufficient steam pressure to perform the test.
There is an additional note pertaining to this SR on Page 3 of Table 3.3.2-1 of the Technical Specification, which states the following (Ref. 14):
Note h: For the time period between February 23, 2000 and prior to turbine restart (following the next time the turbine is removed from service), the response time test requirement of SR 3.3.2.10 is not applicable for 1-FSV-47-027.
SR 3.3.2.11 SR 3.3.2.11 is the performance of a TADOT as described in SR 3.3.2.8, except that it is performed for the P-4 Reactor Trip Interlock, and the Frequency is once per RTB cycle. This Frequency is based on operating experience demonstrating that undetected failure of the P-4 interlock sometimes occurs when the RTB is cycled.
The SR is modified by a Note that excludes verification of setpoints during the TADOT. The Function tested has no associated setpoint.
SR 3.3.2.12 SR 3.3.2.12 is the verification that the SBMFWP channel is in the trip status when a TDMFWP is supplying feedwater to the steam generators (SGs). The Frequency is once within 4 hours during startup after the first TDMFWP is supplying feedwater to the steam generators and in accordance with the Surveillance Frequency Control Program thereafter. This SR is accomplished by verification of the SBMFWP breaker interlock hand switch (WBN-1-HS-003-0200D, Standby Feedwater Pump Breaker Interlock) is in the trip position. This SR ensures that the SBMFWP breaker interlock hand switch is in the correct position when a TDMFW pump is supplying FW to the SGs.
(continued)
Watts Bar-Unit 1                            B 3.3-98                          Revision 20,30, 162, 177 Amendment 13,23,132, 148
 
EFAS Instrumentation B 3.3.2 BASES REFERENCES      20. Letter from John G. Lamb (NRC) to Mr. Preston D. Swafford (TVA) dated (continued)        March 4, 2009, Includes Enclosures (a) Amendment No. 75 to Facility Operating License No. NPF-90 for Watts Bar Nuclear Plant, Unit 1 and (b) NRC Safety Evaluation (SE) for Amendment No. 75.
: 21. Attachment 1 to TSTF-569, Methodology to Eliminate Pressure Sensor and Protection Channel (Fore Westinghouse Plants only) Response Time Testing.
: 22. Code of Federal Regulations, Title 10, Part 50.46, Acceptance Criteria for Emergency Core Cooling Systems for Light Water Nuclear Power Reactors.
Watts Bar-Unit 1                    B 3.3-100                            Revision 96, 170, 176 Amendment 75, 141, 143
 
PAM Instrumentation B 3.3.3 BASES LCO              6. Reactor Vessel Water Level (continued)
Reactor Vessel Water Level, a non-Type A, Category 1 variable, is provided for verification and long term surveillance of core cooling. It is also used for accident diagnosis and to determine reactor coolant inventory adequacy.
The Reactor Vessel Level Instrumentation System (RVLIS) provides a direct measurement of the liquid level above the bottom of the reactor vessel up to the top of the reactor vessel. Indication is in percent of this distance (i.e., the reactor vessel bottom is 0% and the vessel top is 100%). It also has a dynamic range vessel liquid content (% LIQ) normalized from 20% to 100%. Normalization corrects the transmitted level information for the RCP operational configuration so that the accurate dynamic % LIQ is indicated regardless of the pattern of pumps running or the fluid density. Control room indications are provided through the ICCM display. The ICCM display is the primary indication used by the operator during an accident.
: 7. Containment Sump Water Level (Wide Range)
Containment Sump Water Level is provided for event identification, and verification and long term surveillance of RCS integrity.
Containment Sump Water Level is used to:
Verify water source for recirculation mode of ECCS operation after a LOCA.
Determine whether high energy line rupture has occurred inside or outside containment.
: 8. Containment Lower Compartment Atmospheric Temperature The lower compartment temperature monitors will verify the temperatures in the lower compartment after an accident with display in the main control room. The monitoring system consists of two channels with range 0F to 350F.
(continued)
Watts Bar-Unit 1                      B 3.3-108                                    Revision 171 Amendment 144
 
PAM Instrumentation B 3.3.3 BASES LCO                            14, 15. Steam Generator Water Level (Wide and Narrow Range)
(continued)
Narrow range steam generator level is used to make a determination on the nature of the accident in progress, e.g., verify a steam generator tube rupture. Steam generator level (Narrow Range) is also used to help identify the ruptured steam generator following a tube rupture and verify that the intact steam generators are an adequate heat sink for the reactor. Narrow range steam generator water level is used when verifying plant conditions for termination of SI during secondary plant high energy line breaks outside containment.
: 16. AFW Valve Status The status of each AFW swap over to Essential Raw Cooling Water (ERCW) valve is monitored with non-Type A Category 1 indication in the control room. Indication on each valve for fully open or fully closed position is provided. AFW valve status is monitored to give verification to the operator that automatic transfer to ERCW has taken place.
17, 18, 19, 20. Core Exit Temperature Core Exit Temperature is provided for verification and long term surveillance of core cooling.
Core exit thermocouples, in conjunction with RCS wide range temperatures, are sufficient to provide indication of radial distribution of the coolant enthalpy rise across representative sections of the core.
Core Exit Temperature is used to support determination of whether to terminate SI, if still in progress, or to reinitiate SI if it has been stopped.
Core Exit Temperature is also used for unit stabilization and cooldown control.
The Inadequate Core Cooling Monitor (ICCM) is used to monitor the core exit thermocouples. There are two isolated systems, with each system monitoring at least four thermocouples per quadrant. The display gives the average quadrant value, the high quadrant value, and the low quadrant value for each quadrant.
Two OPERABLE channels are required in each quadrant to provide adequate indication of coolant temperature rise in representative regions 7of the core. Two isolated channels of two thermocouples each ensure a single failure will not disable the ability to identify significant temperature gradients.
The incore thermocouple monitoring system described in Reference 4 supports the plant operating procedures.
(continued)
Watts Bar-Unit 1                                B 3.3-111                                      Revision 135, 171 Amendment 144
 
PAM Instrumentation B 3.3.3 BASES LCO              21. Auxiliary Feedwater Flow (continued)
AFW Flow is provided to monitor operation of decay heat removal via the SGs.
Redundant monitoring capability is provided by two independent trains of instrumentation for each SG. Each differential pressure transmitter provides an input to a control room indicator. Since the primary indication used by the operator during an accident is the control room indicator, the PAM specification deals specifically with this portion of the instrument channel.
AFW flow is used three ways:
to verify AFW flow to the SGs; to determine whether to terminate SI if still in progress, in conjunction with SG water level (narrow range); and to regulate AFW flow so that the SG tubes remain covered.
: 22. Reactor Coolant System Subcooling Margin Monitor The RCS subcooling margin monitor is used to determine the temperature margin to saturation of the primary coolant. Control room indications are provided through the ICCM display and digital panel meters. The ICCM display is the primary indication used by the operator during an accident.
: 23. Refueling Water Storage Tank Level RWST water level is used to verify the water source availability to the ECCS and Containment Spray (CS) Systems. It alerts the operator to manually switch the CS suction from the RWST to the containment sump. It may also provide an indication of time for initiating cold leg recirculation from the sump following a LOCA.
: 24. Steam Generator Pressure Steam pressure is used to determine if a high energy secondary line rupture has occurred and the availability of the steam generators as a heat sink. It is also used to verify that a faulted steam generator is isolated. Steam pressure may be used to ensure proper cooldown rates or to provide a diverse indication for natural circulation cooldown.
(continued)
Watts Bar-Unit 1                      B 3.3-112                                  Revision 135, 171 Amendment 144
 
RCS Specific Activity B 3.4.16 B 3.4 REACTOR COOLANT SYSTEM (RCS)
B 3.4.16 RCS Specific Activity BASES BACKGROUND            The maximum dose that an individual at the exclusion area boundary can receive for 2 hours following an accident, or at the low population zone outer boundary for the radiological release duration, is specified in 10 CFR 100 (Ref. 1). The maximum dose to the whole body and the thyroid that an individual occupying the Main Control Room can receive for the accident duration is specified in 10 CFR 50, Appendix A, GDC 19. The limits on specific activity ensure that the offsite and control room doses are appropriately limited during analyzed transients and accidents.
The RCS specific activity LCO limits the allowable concentration level of radionuclides in the reactor coolant. The LCO limits are established to minimize the offsite and Main Control Room dose consequences in the event of a steam generator tube rupture (SGTR) or main steam line break (MSLB) accident.
The LCO contains specific activity limits for both DOSE EQUIVALENT I-131 and DOSE EQUIVALENT XE-133. The allowable levels are intended to ensure that offsite and control room doses meet the appropriate acceptance criteria in the Standard Review Plan (Ref. 2).
APPLICABLE            The LCO limits on the specific activity of the reactor coolant ensure that the SAFETY ANALYSES        resulting offsite and control room doses meet the appropriate SRP acceptance criteria following a MSLB or SGTR accident. The SGTR and MSLB safety analyses (Refs. 3 and 4) assume the specific activity of the reactor coolant at the LCO limit and an existing reactor coolant steam generator (SG) tube leakage rate of 150 gallons per day (GPD). The safety analyses assume the specific activity of the secondary coolant at its limit of 0.1 Ci/gm DOSE EQUIVALENT I-131 from LCO 3.7.14, "Secondary Specific Activity."
(continued)
Watts Bar-Unit 1                                B 3.4-80                                Revision 52, 173 Amendment 41, 146
 
RCS Specific Activity B 3.4.16 BASES APPLICABLE      The analyses for the SGTR and MSLB accidents establish the acceptance limits SAFETY ANALYSES  for RCS specific activity. Reference to these analyses are used to assess (continued)    changes to the unit that could affect RCS specific activity, as they relate to the acceptance limits.
The analyses are for two cases of reactor coolant specific activity. One case assumes specific activity at 0.265 Ci/gm DOSE EQUIVALENT I-131 with an iodine spike immediately after the accident that increases the iodine activity in the reactor coolant by a factor of 500 times the iodine production rate necessary to maintain a steady state iodine concentration of 0.265 Ci/gm DOSE EQUIVALENT I-131. The second case assumes the initial reactor coolant iodine activity at 14 Ci/gm DOSE EQUIVALENT I-131 due to a pre-accident iodine spike caused by an RCS transient. In both cases, the noble gas activity in the reactor coolant equals the LCO limit of 1200 Ci/gm DOSE EQUIVALENT XE-133.
The analyses also assume a loss of offsite power at the same time as the SGTR and MSLB event. The SGTR causes a reduction in reactor coolant inventory.
The reduction initiates a reactor trip from a low pressurizer pressure signal or an RCS overtemperature T signal. The MSLB results in a reactor trip due to low steam pressure.
For the SGTR, the coincident loss of offsite power causes the steam dump valves to close to protect the condenser. The rise in pressure in the ruptured SG discharges radioactively contaminated steam to the atmosphere through the SG power operated relief valves and the main steam safety valves. The unaffected SGs remove core decay heat by venting steam to the atmosphere until the cooldown ends.
Operation with iodine specific levels greater than the LCO limit is permissible, if the activity levels do not exceed 14 Ci/gm DOSE EQUIVALENT I-131, in the applicable specification, for more than 48 hours. The safety analyses have concurrent and pre-accident iodine spiking levels up to 14 Ci/gm DOSE EQUIVALENT I-131.
The limits on RCS specific activity are also used for establishing standardization in radiation shielding and plant personnel radiation protection practices.
RCS specific activity satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).
(continued)
Watts Bar-Unit 1                          B 3.4-81                      Revision 52, 111, 118, 173 Amendment 41, 91, 146
 
RCS Specific Activity B 3.4.16 BASES LCO              The specific iodine activity is limited to 0.265 Ci/gm DOSE EQUIVALENT I-131, and the noble gas specific activity in the reactor coolant is limited to 1200 Ci/gm DOSE EQUIVALENT XE-133, which ensure that offsite and control room doses will meet the appropriate SRP acceptance criteria (Ref. 2).
The MSLB and SGTR accident analyses (Refs. 3 and 4) show that the calculated doses are within acceptable limits. Violation of the LCO may result in reactor coolant radioactivity levels that could, in the event of a MSLB or SGTR, lead to doses that exceed the SRP acceptance criteria (Ref. 2).
APPLICABILITY    In MODES 1, 2, 3, and 4, operation within the LCO limits for DOSE EQUIVALENT I-131 and DOSE EQUIVALENT XE-133 is necessary to limit the potential consequences of a SGTR or MSLB to within the SRP acceptance criteria (Ref. 2).
In MODES 5 and 6, the steam generators are not being used for decay heat removal, the RCS and steam generators are depressurized, and primary to secondary leakage is minimal. Therefore, the monitoring of RCS specific activity is not required.
(continued)
Watts Bar-Unit 1                          B 3.4-82                                  Revision 52,173 Amendment 41, 146
 
RCS Specific Activity B 3.4.16 BASES (continued)
ACTIONS          A.1 and A.2 With the DOSE EQUIVALENT I-131 greater than the LCO limit, samples at intervals of 4 hours must be taken to demonstrate that the limit of 14 Ci/gm is not exceeded. The Completion Time of 4 hours is required to obtain and analyze a sample. Sampling is done to continue to provide a trend.
The DOSE EQUIVALENT I-131 must be restored to within limits within 48 hours.
The Completion Time of 48 hours is acceptable since it is expected that, if there were an iodine spike, the normal coolant iodine concentration would be restored within this time period. Also, there is a low probability of a MSLB or SGTR occurring during this time period.
A Note permits the use of the provisions of LCO 3.0.4.c. This allowance permits entry into the applicable MODE(S) while relying on the ACTIONS. This allowance is acceptable due to the significant conservatism incorporated into the specific activity limit, the low probability of an event which is limiting due to exceeding this limit, and the ability to restore transient specific activity excursions while the plant remains at, or proceeds to power operation.
B.1 With the DOSE EQUIVALENT XE-133 greater than the LCO limit, DOSE EQUIVALENT XE-133 must be restored to within limit within 48 hours. The allowed Completion Time of 48 hours is acceptable since it is expected that, if there were a noble gas spike, the normal coolant noble gas concentration would be restored within this time period. Also, there is a low probability of a MSLB or SGTR occurring during this time period.
A Note permits the use of the provisions of LCO 3.0.4.c. This allowance permits entry into the applicable MODE(S), relying on Required Action B.1 while the DOSE EQUIVALENT XE-133 LCO limit is not met. This allowance is acceptable due to the significant conservatism incorporated into the specific activity limit, the low probability of an event which is limiting due to exceeding this limit, and the ability to restore transient-specific activity excursions while the plant remains at, or proceeds to, power operation.
C.1 and C.2 If a Required Action and the associated Completion Time of Condition A or B is not met, or if the DOSE EQUIVALENT I-131 is greater than 14 Ci/gm, the reactor must be brought to MODE 3 within 6 hours and MODE 5 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.
(continued)
Watts Bar-Unit 1                              B 3.4-83                    Revision 52, 68, 111, 118, 173 Amendment 41, 55, 91, 146
 
RCS Specific Activity B 3.4.16 BASES SURVEILLANCE    SR 3.4.16.1 REQUIREMENTS SR 3.4.16.1 requires performing a gamma isotopic analysis as a measure of the noble gas specific activity of the reactor coolant. This measurement is the sum of the degassed gamma activities and the gaseous gamma activities in the sample taken. This Surveillance provides an indication of any increase in the noble gas specific activity.
Trending the results of this Surveillance allows proper remedial action to be taken before reaching the LCO limit under normal operating conditions. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
Due to the inherent difficulty in detecting Kr-85 in a reactor coolant sample due to masking from radioisotopes with similar decay energies, such as F-18 and I-134, it is acceptable to include the minimum detectable activity for Kr-85 in the SR 3.4.16.1 calculation. If a specific noble gas nuclide listed in the definition of DOSE EQUIVALENT XE-133 is not detected, it should be assumed to be present at the minimum detectable activity. A Note modifies the SR, which requires the SR to only be performed in MODES 1, 2, and 3 with Tavg 500&deg;F.
SR 3.4.16.2 This Surveillance is performed to ensure iodine remains within limit during normal operation and following rapid power changes when fuel failure is more apt to occur. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. The Frequency, between 2 and 6 hours after a power change  15% RTP within a 1 hour period, is established because the iodine levels peak during this time following fuel failure; samples at other times would provide inaccurate results.
(continued)
Watts Bar-Unit 1                          B 3.4-84                  Revision 52, 111, 118, 162, 173 Amendment 41, 91, 132, 146
 
RCS Specific Activity B 3.4.16 BASES REFERENCES      1. Title 10, Code of Federal Regulations, Part 100.11, Determination of Exclusion Area, Low Population Zone, and Population Center Distance, 1973.
: 2. Standard Review Plan (SRP) Section 15.1.5 Appendix A (SLB) and Section 15.6.3 (SGTR).
: 3. Watts Bar FSAR, Section 15.5.4, Environmental Consequences of a Postulated Main Steam Line Break.
: 4. Watts Bar FSAR, Section 15.5.5, Environmental Consequences of a Postulated Steam Generator Tube Rupture.
Watts Bar-Unit 1                    B 3.4-85                                    Revision 173 Amendment 146
 
SG Tube Integrity B 3.4.17 BASES (continued)
APPLICABLE        The steam generator tube rupture (SGTR) accident is the limiting design SAFETY            basis event for SG tubes and avoiding an SGTR is the basis for this ANALYSES          Specification. The analysis of a SGTR event assumes a bounding primary to secondary LEAKAGE rate equal to the operational LEAKAGE rate limits in LCO 3.4.13, RCS Operational LEAKAGE, plus the leakage rate associated with a double-ended rupture of a single tube. The accident analysis for a SGTR assumes the contaminated secondary fluid is only briefly released to the atmosphere via safety valves and the majority is discharged to the main condenser.
The analysis for design basis accidents and transients other than a SGTR assume the SG tubes retain their structural integrity (i.e., they are assumed not to rupture). In these analyses, the steam discharge to the atmosphere is based on the total primary to secondary LEAKAGE from 150 gallons per day (gpd) per steam generator and 1 gallon per minute (gpm) in the faulted steam generator.
For accidents that do not involve fuel damage, the primary coolant activity level of DOSE EQUIVALENT I-131 is assumed to be equal to the LCO 3.4.16 RCS Specific Activity, limits. For accidents that assume fuel damage, the primary coolant activity is a function of the amount of activity released from the damaged fuel. The dose consequences of these events are within the limits of GDC 19 (Ref. 2), and 10 CFR 100 (Ref. 3) or the NRC approved licensing basis.
Steam generator tube integrity satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).
LCO              The LCO requires that SG tube integrity be maintained. The LCO also requires that all SG tubes that satisfy the plugging criteria be plugged in accordance with the Steam Generator Program.
During an SG inspection, any inspected tube that satisfies the Steam Generator Program plugging criteria is removed from service by plugging. If a tube was determined to satisfy the plugging criteria but was not plugged, the tube may still have tube integrity.
In the context of this Specification, a SG tube is defined as the entire length of the tube, including the tube wall, between the tube-to-tubesheet weld at the tube inlet and the tube-to-tubesheet weld at the tube outlet. The tube-to-tubesheet weld is not considered part of the tube.
A SG tube has tube integrity when it satisfies the SG performance criteria. The SG performance criteria are defined in Specification 5.7.2.12, Steam Generator Program, and describe acceptable SG tube performance. The Steam Generator Program also provides the evaluation process for determining conformance with the SG performance criteria (continued)
Watts Bar-Unit 1                          B 3.4-87                              Revision 82, 174 Amendment 65, 147
 
SG Tube Integrity B 3.4.17 BASES LCO              The operational LEAKAGE performance criterion provides an observable (continued)    indication of SG tube conditions during plant operation. The limit on operational LEAKAGE is contained in LCO 3.4.13, RCS Operational LEAKAGE, and limits primary to secondary LEAKAGE through any one SG to 150 gallons per day.
This limit is based on the assumption that a single crack leaking this amount would not propagate to a SGTR under the stress conditions of a LOCA or a main steam line break. If this amount of LEAKAGE is due to more than one crack, the cracks are very small, and the above assumption is conservative.
APPLICABILITY    Steam generator tube integrity is challenged when the pressure differential across the tubes is large. Large differential pressures across SG tubes can only be experienced in MODE 1, 2, 3, or 4.
RCS conditions are far less challenging in MODES 5 and 6 than during MODES 1, 2, 3, and 4. In MODES 5 and 6, primary to secondary differential pressure is low, resulting in lower stresses and reduced potential for LEAKAGE.
ACTIONS          The ACTIONS are modified by a Note that the Conditions may be entered independently for each SG tube. This is acceptable because the Required Actions provide appropriate compensatory actions for each affected SG tube.
Complying with the Required Actions may allow for continued operation, and subsequent affected SG tubes are governed by subsequent Condition entry, and application of associated Required Actions.
A.1 and A.2 Condition A applies if it is discovered that one or more SG tubes examined in an inservice inspection satisfy the tube plugging criteria but were not plugged in accordance with the Steam Generator Program as required by SR 3.4.17.2. An evaluation of SG tube integrity of the affected tube(s) must be made. Steam generator tube integrity is based on meeting the SG performance criteria described in the Steam Generator Program. The SG plugging criteria define limits on SG tube degradation that allow for flaw growth between inspections while still providing assurance that the SG performance criteria will continue to be met. In order to determine if a SG tube that should have been plugged, has tube integrity, an evaluation must be completed that demonstrates that the SG performance criteria will continue to be met until the next refueling outage or SG tube inspection. The tube integrity determination is based on the estimated condition of the tube at the time the situation is discovered and the estimated growth of the degradation prior to the next SG tube inspection. If it is determined that tube integrity is not being maintained, Condition B applies.
(continued)
Watts Bar-Unit 1                        B 3.4-89                              Revision 82, 174 Amendment 65, 147
 
SG Tube Integrity B 3.4.17 BASES ACTIONS          A.1 and A.2 (continued)
A Completion Time of 7 days is sufficient to complete the evaluation while minimizing the risk of plant operation with a SG tube that may not have tube integrity.
If the evaluation determines that the affected tube(s) have tube integrity, Required Action A.2 allows plant operation to continue until the next refueling outage or SG inspection provided the inspection interval continues to be supported by an operational assessment that reflects the affected tubes.
However, the affected tube(s) must be plugged prior to entering MODE 4 following the next refueling outage or SG inspection. This Completion Time is acceptable since operation until the next inspection is supported by the operational assessment.
B.1 and B.2 If the Required Actions and associated Completion Times of Condition A are not met or if SG tube integrity is not being maintained, the reactor must be brought to MODE 3 within 6 hours and MODE 5 within 36 hours.
The allowed Completion Times are reasonable, based on operating experience, to reach the desired plant conditions from full power conditions in an orderly manner and without challenging plant systems.
SURVEILLANCE    SR 3.4.17.1 REQUIREMENTS During shutdown periods the SGs are inspected as required by this SR and the Steam Generator Program. NEI 97-06, Steam Generator Program Guidelines (Ref. 1), and its referenced EPRI Guidelines, establish the content of the Steam Generator Program. Use of the Steam Generator Program ensures that the inspection is appropriate and consistent with accepted industry practices.
During SG inspections a condition monitoring assessment of the SG tubes is performed. The condition monitoring assessment determines the as found condition of the SG tubes. The purpose of the condition monitoring assessment is to ensure that the SG performance criteria have been met for the previous operating period.
The Steam Generator Program determines the scope of the inspection and the methods used to determine whether the tubes contain flaws satisfying the tube plugging criteria. Inspection scope (i.e., which tubes or areas of tubing within the SG are to be inspected) is a function of existing and potential degradation locations. The Steam Generator Program also specifies the inspection methods to be used to find potential degradation. Inspection methods are a function of degradation morphology, nondestructive examination (NDE) technique capabilities, and inspection locations.
(continued)
Watts Bar-Unit 1                        B 3.4-90                              Revision 82, 174 Amendment 65, 147
 
SG Tube Integrity B 3.4.17 BASES SURVEILLANCE    SR 3.4.17.1 (continued)
REQUIREMENTS The Steam Generator Program defines the Frequency of SR 3.4.17.1. The Frequency is determined by the operational assessment and other limits in the SG examination guidelines (Ref. 6). The Steam Generator Program uses information on existing degradations and growth rates to determine an inspection Frequency that provides reasonable assurance that the tubing will meet the SG performance criteria at the next scheduled inspection. In addition, Specification 5.7.2.12 contains prescriptive requirements concerning inspection intervals to provide added assurance that the SG performance criteria will be met between scheduled inspections. If crack indications are found in any SG tube, the maximum inspection interval for all affected and potentially affected SGs is restricted by Specification 5.7.2.12 until subsequent inspections support extending the inspection interval.
SR 3.4.17.2 During an SG inspection, any inspected tube that satisfies the Steam Generator Program plugging criteria is removed from service by plugging. The tube plugging criteria delineated in Specification 5.7.2.12 are intended to ensure that tubes accepted for continued service satisfy the SG performance criteria with allowance for error in the flaw size measurement and for future flaw growth. In addition, the tube plugging criteria, in conjunction with other elements of the Steam Generator Program, ensure that the SG performance criteria will continue to be met until the next inspection of the subject tube(s). Reference 1 provides guidance for performing operational assessments to verify that the tubes remaining in service will continue to meet the SG performance criteria.
The Frequency of prior to entering MODE 4 following a SG inspection ensures that the Surveillance has been completed and all tubes meeting the plugging criteria are plugged prior to subjecting the SG tubes to significant primary to secondary pressure differential.
REFERENCES      1. NEI 97-06, Steam Generator Program Guidelines.
: 2. 10 CFR 50 Appendix A, GDC 19, Control Room.
: 3. 10 CFR 100, Reactor Site Criteria.
: 4. ASME Boiler and Pressure Vessel Code, Section III, Subsection NB.
: 5. Draft Regulatory Guide 1.121, Basis for Plugging Degraded Steam Generator Tubes, August 1976.
: 6. EPRI, Pressurized Water Reactor Steam Generator Examination Guidelines.
Watts Bar-Unit 1                        B 3.4-91                              Revision 82, 174 Amendment 65, 147
 
Accumulators B 3.5.1 BASES BACKGROUND      This interlock also prevents inadvertent closure of the valves during normal (continued)    operation prior to an accident. Although not required for accident mitigation, the valves will automatically open as a result of an SI signal. These features ensure that the valves meet the requirements of the Institute of Electrical and Electronic Engineers (IEEE) Standard 279-1971 (Ref. 1) for "operating bypasses" and that the accumulators will be available for injection without reliance on operator action.
The accumulator size, water volume, and nitrogen cover pressure are selected so that three of the four accumulators are sufficient to partially cover the core before significant clad melting or zirconium water reaction can occur following a LOCA. The need to ensure that three accumulators are adequate for this function is consistent with the LOCA assumption that the entire contents of one accumulator will be lost via the RCS pipe break during the blowdown phase of the LOCA.
APPLICABLE      The accumulators are assumed OPERABLE in both the large and small break SAFETY ANALYSES  LOCA analyses at full power (Ref. 2). These are the Design Basis Accidents (DBAs) that establish the acceptance limits for the accumulators. Reference to the analyses for these DBAs is used to assess changes in the accumulators as they relate to the acceptance limits.
In performing the LOCA calculations, conservative assumptions are made concerning the availability of ECCS flow. In the early stages of a LOCA, with or without a loss of offsite power, the accumulators provide the sole source of makeup water to the RCS. The assumption of loss of offsite power is also considered to determine if it yields limiting results. The loss of offsite power assumption imposes a delay wherein the ECCS pumps cannot deliver flow until the emergency diesel generators start, come to rated speed, and go through their timed loading sequence. In cold leg break scenarios, the entire contents of one accumulator are assumed to be lost through the break.
The limiting large break LOCA is a break in the cold leg. During this event, the accumulators discharge to the RCS as soon as RCS pressure decreases to below accumulator pressure.
(continued)
Watts Bar-Unit 1                            B 3.5-2                                  Revision 39, 176 Amendment 21, 143
 
Accumulators B 3.5.1 BASES APPLICABLE      As a conservative estimate, no credit is taken for ECCS pump flow until an SAFETY ANALYSES  effective delay has elapsed. This delay accounts for the diesels starting (for loss (continued)    of offsite power assumption) and the pumps being loaded and delivering full flow.
The delay time is conservatively set with an additional 2 seconds to account for SI signal generation. During this time, the accumulators are analyzed as providing the sole source of emergency core cooling. No operator action is assumed during the blowdown stage of a large break LOCA.
The small break LOCA analysis also assumes a time delay before pumped flow is assumed to inject into the reactor coolant system. For intermediate breaks, the rate of blowdown is such that the increase in fuel clad temperature is terminated solely by the accumulators, with pumped flow then providing continued cooling.
As break size decreases, the accumulators and centrifugal charging pumps both play a part in terminating the rise in clad temperature. At very small break sizes, the safety injection pumps are capable of mitigating the inventory loss during the small break LOCA, and the accumulators do not play a significant role in the accident mitigation.
This LCO helps to ensure that the following acceptance criteria established for the ECCS by 10 CFR 50.46, Paragraph b (Ref. 3) will be met with a high level of probability following a LOCA:
: a.        Maximum fuel element cladding temperature is  2200F;
: b.        Maximum cladding oxidation is  0.17 times the total cladding thickness before oxidation;
: c.        Maximum hydrogen generation from a zirconium water reaction is  0.01 times the hypothetical amount that would be generated if all of the metal in the cladding cylinders surrounding the fuel, excluding the cladding surrounding the plenum volume, were to react; and
: d.        Core is maintained in a coolable geometry.
Since the accumulators discharge during the blowdown phase of a LOCA, they do not contribute to the long term cooling requirements of 10 CFR 50.46.
The contained water volume is the same as the deliverable volume for the accumulators, since the accumulators are emptied, once discharged. Both large and small-break analyses use a nominal accumulator line volume from the accumulator to the check valve. The safety analysis assumes accumulator water volumes of 7518 gallons and 8191 gallons. To allow for instrument inaccuracy, values of 7630 gallons and 8000 gallons are specified.
(continued)
Watts Bar-Unit 1                            B 3.5-3                              Revision 3, 39, 176 Amendment 21, 143
 
Accumulators B 3.5.1 BASES APPLICABLE      The minimum boron concentration setpoint is used in the post LOCA boron SAFETY ANALYSES  concentration calculation. The calculation is performed to assure reactor (continued)    subcriticality in a post LOCA environment. Of particular interest is the large break LOCA, since no credit is taken for control rod assembly insertion. A reduction in the accumulator minimum boron concentration would produce a subsequent reduction in the available containment sump concentration for post LOCA shutdown and an increase in the maximum sump pH. The maximum boron concentration is used in determining the cold leg to hot leg recirculation injection switchover time and minimum sump pH.
The maximum nitrogen cover pressure limit prevents accumulator relief valve actuation, and ultimately preserves accumulator integrity. The LOCA analyses support a range of 585 to 690 psig. To account for the accumulator tank design pressure rating, and to allow for instrument accuracy values of  610 psig and 660 psig are specified for the pressure indicator in the main control room.
The effects on containment mass and energy releases from the accumulators are accounted for in the appropriate analyses (Refs. 2 and 4).
The accumulators satisfy Criterion 3 of the NRC Policy Statement.
LCO              The LCO establishes the minimum conditions required to ensure that the accumulators are available to accomplish their core cooling safety function following a LOCA. Four accumulators are required to ensure that 100% of the contents of three of the accumulators will reach the core during a LOCA. This is consistent with the assumption that the contents of one accumulator spill through the break. If less than three accumulators are injected during the blowdown phase of a LOCA, the ECCS acceptance criteria of 10 CFR 50.46 (Ref. 3) could be violated.
For an accumulator to be considered OPERABLE, the isolation valve must be fully open, power removed above 1000 psig, and the limits established in the SRs for contained volume, boron concentration, and nitrogen cover pressure must be met.
(continued)
Watts Bar-Unit 1                            B 3.5-4                            Revision 3, 39, 176 Amendment 21, 143
 
ECCS - Operating B 3.5.2 BASES BACKGROUND      The centrifugal charging subsystem of the ECCS also functions to supply (continued)    borated water to the reactor core following increased heat removal events, such as a main steam line break (MSLB). The limiting design conditions occur when the negative moderator temperature coefficient is highly negative, such as at the end of each cycle.
During low temperature conditions in the RCS, limitations are placed on the maximum number of ECCS pumps that may be OPERABLE. Refer to the Bases for LCO 3.4.12, "Cold Overpressure Mitigation System (COMS)," for the basis of these requirements.
The ECCS subsystems are actuated upon receipt of an SI signal. The actuation of safeguard loads is accomplished in a programmed time sequence for a loss of offsite power. If offsite power is available, the safeguard loads start immediately.
If offsite power is not available, the Engineered Safety Feature (ESF) buses shed normal operating loads and are connected to the emergency diesel generators (EDGs). Safeguard loads are then actuated in the programmed time sequence.
The time delay associated with diesel starting, sequenced loading, and pump starting determines the time required before pumped flow is available to the core following a LOCA.
The active ECCS components, along with the passive accumulators and the RWST covered in LCO 3.5.1, "Accumulators," and LCO 3.5.4, "Refueling Water Storage Tank (RWST)," provide the cooling water necessary to meet GDC 35 (Ref. 1).
APPLICABLE      The LCO helps to ensure that the following acceptance criteria for the ECCS, SAFETY ANALYSES  established by 10 CFR 50.46, Paragraph b (Ref. 2), will be met with a high level of probability following a LOCA:
: a.        Maximum fuel element cladding temperature is  2200&deg;F;
: b.        Maximum cladding oxidation is  0.17 times the total cladding thickness before oxidation; (continued)
Watts Bar-Unit 1                          B 3.5-11                                Revision 39, 176 Amendment 21, 143
 
RWST B 3.5.4 BASES APPLICABLE      in the inadvertent ECCS actuation analysis, although it is typically a nonlimiting SAFETY ANALYSES  event and the results are very insensitive to boron concentrations. The (continued)    maximum temperature ensures that the amount of cooling provided from the RWST during the heatup phase of a feedline break is consistent with safety analysis assumptions; the minimum is an assumption in both the MSLB and inadvertent ECCS actuation analyses, although the inadvertent ECCS actuation event is typically nonlimiting.
The MSLB analysis has considered a delay associated with the interlock between the VCT and RWST isolation valves, and the results show that the departure from nucleate boiling design basis is met. The delay has been established as 27 seconds, with offsite power available, or 37 seconds without offsite power.
For a large break LOCA Analysis, the minimum water volume limit of 370,000 gallons and the minimum boron concentration limit is used to compute the post LOCA sump boron concentration necessary to assure subcriticality.
The large break LOCA is the limiting case since the safety analysis assumes least negative reactivity insertion.
The upper limit on boron concentration of 3300 ppm is used to determine the maximum allowable time to switch to hot leg recirculation following a LOCA. The purpose of switching from cold leg to hot leg injection is to avoid boron precipitation in the core following the accident.
In the ECCS analysis, the containment spray temperature is assumed to be equal to the RWST lower temperature limit of 60&deg;F. If the lower temperature limit is violated, the containment spray further reduces containment pressure, which decreases the rate at which steam can be vented out the break and increases peak clad temperature. The acceptable temperature range of 60&deg;F to 105&deg;F is assumed in the large and small-break LOCA analyses per approved methods (Ref. 3). The upper temperature limit of 105&deg;F is also used in the containment OPERABILITY analysis. Exceeding the upper temperature limit could result in a higher peak clad temperature, because there is less heat transfer from the core to the injected water following a LOCA. For the containment response following an MSLB, the lower limit on boron concentration and the upper limit on RWST water temperature are used to maximize the total energy release to containment.
The RWST satisfies Criterion 3 of the NRC Policy Statement.
(continued)
Watts Bar-Unit 1                          B 3.5-25                Revision 13, 61, 88, 98, 131, 176 Amendment 7, 40, 48, 67, 77, 107, 143
 
RWST B 3.5.4 BASES (continued)
REFERENCES        1. Watts Bar FSAR, Section 6.3, Emergency Core Cooling System, and Section 15.0, Accident Analysis.
: 2. Watts Bar Drawing 1-47W605-243, Electrical Tech Spec Compliance Tables.
: 3. WCAP-16996-P-A, Revision 1, Realistic LOCA Evaluation Methodology Applied to the Full Spectrum of Break Sizes (FULL SPETRUM LOCA Methodology), November 2016.
Watts Bar-Unit 1                    B 3.5-28                                Revision 176 Amendment 143
 
Containment B 3.6.1 BASES SURVEILLANCE    SR 3.6.1.1 REQUIREMENTS Maintaining the containment OPERABLE requires compliance with the visual examinations and leakage rate test requirements of the Containment Leakage Rate Testing Program. Failure to meet air lock, Shield Building containment bypass leakage path, and purge valve with resilient seal leakage limits specified in LCO 3.6.2 and LCO 3.6.3 does not invalidate the acceptability of these overall leakage determinations unless their contribution to overall Type A, B, and C leakage causes that to exceed limits. As left leakage prior to the first startup after performing a required leakage test is required to be < 0.6 La for combined Type B and C leakage and  0.75 La for overall Type A leakage. At all other times between required leakage rate tests, the acceptance criteria is based on an overall Type A leakage limit of  1.0 La. At  1.0 La the offsite dose consequences are bounded by the assumptions of the safety analysis.
SR Frequencies are as required by the Containment Leakage Rate Testing Program. These periodic testing requirements verify that the containment leakage rate does not exceed the leakage rate assumed in the safety analysis.
REFERENCES      1.      Title 10, Code of Federal Regulations, Part 50, Appendix J, Option B, "Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors - Performance-Based Requirements."
: 2.      Watts Bar FSAR, Section 15.0, "Accident Analysis."
: 3.      Watts Bar FSAR, Section 6.2, "Containment Systems."
: 4.      Nuclear Energy Institute (NEI) 94-01, Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J, Revision 3-A, dated July 2012.
Watts Bar-Unit 1                          B 3.6-4                                  Revision 10, 178 Amendment 5, 149
 
Containment Air Locks B 3.6.2 BASES (continued)
APPLICABLE        The DBAs that result in a significant release of radioactive material within SAFETY            containment are a loss of coolant accident and a rod ejection accident ANALYSES          (Ref. 2). In the analysis of each of these accidents, it is assumed that containment is OPERABLE such that release of fission products to the environment is controlled by the rate of containment leakage. The containment was designed with an allowable leakage rate (La) of 0.25% of containment air weight per day (Ref. 2), at Pa = 15.0 psig. This allowable leakage rate forms the basis for the acceptance criteria imposed on the SRs associated with the air locks.
The containment air locks satisfy Criterion 3 of the NRC Policy Statement.
LCO              Each containment air lock forms part of the containment pressure boundary. As part of containment pressure boundary, the air lock safety function is related to control of the containment leakage rate resulting from a DBA. Thus, each air lock's structural integrity and leak tightness are essential to the successful mitigation of such an event.
Each air lock is required to be OPERABLE. For the air lock to be considered OPERABLE, the air lock interlock mechanism must be OPERABLE, the air lock must be in compliance with the Type B air lock leakage test, and both air lock doors must be OPERABLE. The interlock allows only one air lock door of an air lock to be opened at one time. This provision ensures that a gross breach of containment does not exist when containment is required to be OPERABLE.
Closure of a single door in each air lock is sufficient to provide a leak tight barrier following postulated events. Nevertheless, both doors are kept closed when the air lock is not being used for normal entry into and exit from containment.
APPLICABILITY    In MODES 1, 2, 3, and 4, a DBA could cause a release of radioactive material to containment. In MODES 5 and 6, the probability and consequences of these events are reduced due to the pressure and temperature limitations of these MODES. Therefore, the containment air locks are not required in MODES 5 and 6 to prevent leakage of radioactive material from containment.
ACTIONS          The ACTIONS are modified by a Note that allows entry and exit to perform repairs on the affected air lock component. If the outer door is inoperable, then it may be easily accessed for most repairs. It is preferred that the air lock be accessed from inside containment by entering through the other OPERABLE air lock. However, if this is not practicable, or if repairs on either door must be performed from the barrel side of the door then it is permissible to enter the air (continued)
Watts Bar-Unit 1                              B 3.6-6                                Revision 130, 178 Amendment 149
 
Containment Isolation Valves B 3.6.3 BASES SURVEILLANCE    SR 3.6.3.3 (continued)
REQUIREMENTS The Note allows valves and blind flanges located in high radiation areas to be verified closed by use of administrative means. Allowing verification by administrative means is considered acceptable, since access to these areas is typically restricted for ALARA reasons. Therefore, the probability of misalignment of these containment isolation valves, once they have been verified to be in their proper position, is small.
SR 3.6.3.4 Verifying that the isolation time of each power operated and automatic containment isolation valve is within limits is required to demonstrate OPERABILITY. The isolation time test ensures the valve will isolate in a time period less than or equal to that assumed in the safety analyses. The isolation time and Frequency of this SR are in accordance with the Inservice Testing Program or in accordance with the Surveillance Frequency Control Program.
SR 3.6.3.5 For containment purge valves with resilient seals, additional leakage rate testing beyond the test requirements of 10 CFR 50, Appendix J, Option B (Ref. 4), is required to ensure OPERABILITY.
Operating experience has demonstrated that this type of seal has the potential to degrade in a shorter time period than do other seal types. Based on this observation and the importance of maintaining this penetration leak tight (due to the direct path between containment and the environment), these valves will not be placed on the maximum extended test interval. Therefore, these valves will be tested in accordance with NEI 94-01, Revision 3-A, which allows a maximum test interval of 30 months. (Ref.3).
SR 3.6.3.6 Automatic containment isolation valves close on a containment isolation signal to prevent leakage of radioactive material from containment following a DBA. This SR ensures that each automatic containment isolation valve will actuate to its isolation position on a containment isolation signal. This Surveillance is not required for valves that are locked, sealed, or otherwise secured in the required position under administrative control.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
(continued)
Watts Bar-Unit 1                            B 3.6-20                      Revision 10, 151, 162, 178 Amendment 5, 123, 132, 149
 
Containment Isolation Valves B 3.6.3 BASES SURVEILLANCE    SR 3.6.3.7 REQUIREMENTS (continued)      Verifying that each 24 inch containment lower compartment purge valve is blocked to restrict opening to  50&deg;F is required to ensure that the valves can close under DBA conditions within the times assumed in the analyses of References 1 and 2. If a LOCA occurs, the purge valves must close to maintain containment leakage within the values assumed in the accident analysis. At other times when purge valves are required to be capable of closing (e.g., during movement of irradiated fuel assemblies), pressurization concerns are not present, thus the purge valves can be fully open. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.6.3.8 This SR ensures that the combined leakage rate of all Shield Building bypass leakage paths is less than or equal to the specified leakage rate. This provides assurance that the assumptions in the safety analysis are met. The as-left bypass leakage rate prior to the first startup after performing a leakage test, requires calculation using maximum pathway leakage (leakage through the worse of the two isolation valves). If the penetration is isolated by use of one closed and de-activated automatic valve, closed manual valve, or blind flange, then the leakage rate of the isolated bypass leakage path is assumed to be the actual pathway leakage through the isolation device. If both isolation valves in the penetration are closed, the actual leakage rate is the lesser leakage rate of the two valves. At all other times, the leakage rate will be calculated using minimum pathway leakage.
The frequency is required by Containment Leakage Rate Testing Program. This SR simply imposes additional acceptance criteria. Although not a part of La, the Shield Building Bypass leakage path combined leakage rate is determined using the 10 CFR 50, Appendix J, Option B, Type B and C leakage rates for the applicable barriers.
REFERENCES      1.      Watts Bar FSAR, Section 15.0, "Accident Analysis."
: 2.      Watts Bar FSAR, Section 6.2.4.2, "Containment Isolation System Design,"
and Table 6.2.4-1, "Containment Penetrations and Barriers."
: 3. Nuclear Energy Institute (NEI) 94-01, Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J, Revision 3-A, dated July 2012.
: 3.      Title 10, Code of Federal Regulations, Part 50 Appendix J, Option B, "Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors - Performance - Based Requirements."
Watts Bar-Unit 1                          B 3.6-21                      Revision 10, 151, 162, 178 Amendment 5, 123, 132, 149
 
Containment Pressure B 3.6.4 B 3.6 CONTAINMENT SYSTEMS B 3.6.4 Containment Pressure BASES BACKGROUND            The containment pressure is limited during normal operation to preserve the initial conditions assumed in the accident analyses for a loss of coolant accident (LOCA) or steam line break (SLB). These limits also prevent the containment pressure from exceeding the containment design negative pressure differential
(-2.0 psid) with respect to the Shield Building annulus atmosphere in the event of inadvertent actuation of the Containment Spray System or Air Return Fans.
Containment pressure is a process variable that is monitored and controlled.
The containment pressure limits are derived from the input conditions used in the containment functional analyses and the containment structure external pressure analysis. Should operation occur outside these limits coincident with a Design Basis Accident (DBA), post accident containment pressures could exceed calculated values.
APPLICABLE            Containment internal pressure is an initial condition used in the DBA SAFETY ANALYSES      analyses to establish the maximum peak containment internal pressure. The limiting DBAs considered, relative to containment pressure, are the LOCA and SLB, which are analyzed using computer pressure transients. The worst case LOCA generates larger mass and energy release than the worst case SLB.
Thus, the LOCA event bounds the SLB event from the containment peak pressure standpoint (Ref. 1).
The initial pressure condition used in the containment analysis was 15.0 psia.
This resulted in a maximum peak pressure from a LOCA of 9.36 psig. The containment analysis (Ref. 1) shows that the maximum allowable internal containment pressure (15.0 psig) bounds the calculated results from the limiting LOCA. The maximum containment pressure resulting from the worst case LOCA, does not exceed the containment design pressure, 13.5 psig.
(continued)
Watts Bar-Unit 1                                B 3.6-22                  Revision 44, 55, 76, 127, 178 Amendment 33, 149
 
Containment Pressure B 3.6.4 BASES APPLICABLE      The containment was also designed for an external pressure load equivalent SAFETY ANALYSES  to 2.0 psig. The inadvertent actuation of the Containment Spray System was (continued)    analyzed to determine the resulting reduction in containment pressure. The initial pressure condition used in this analysis was -0.1 psig. This resulted in a minimum pressure inside containment of 1.4 psig, which is less than the design load.
For certain aspects of transient accident analyses, maximizing the calculated containment pressure is not conservative. In particular, the cooling effectiveness of the Emergency Core Cooling System during the core reflood phase of a LOCA analysis increases with increasing containment backpressure. Therefore, for the reflood phase, the containment backpressure is calculated in a manner designed to conservatively minimize, rather than maximize, the containment pressure response in accordance with WCAP-16996-P-A, Revision 1 (Ref. 2).
Containment pressure satisfies Criterion 2 of the NRC Policy Statement.
LCO              Maintaining containment pressure at less than or equal to the LCO upper pressure limit ensures that, in the event of a DBA, the resultant peak containment accident pressure will remain below the containment design pressure.
Maintaining containment pressure at greater than or equal to the LCO lower pressure limit ensures that the containment will not exceed the design negative differential pressure following the inadvertent actuation of the Containment Spray System or Air Return Fans.
APPLICABILITY    In MODES 1, 2, 3, and 4, a DBA could cause a release of radioactive material to containment. Since maintaining containment pressure within limits is essential to ensure initial conditions assumed in the accident analyses are maintained, the LCO is applicable in MODES 1, 2, 3 and 4.
In MODES 5 and 6, the probability and consequences of these events are reduced due to the pressure and temperature limitations of these MODES.
Therefore, maintaining containment pressure within the limits of the LCO is not required in MODES 5 or 6.
(continued)
Watts Bar-Unit 1                            B 3.6-23                                  Revision 176 Amendment 143
 
Containment Pressure B 3.6.4 BASES (continued)
ACTIONS          A.1 When containment pressure is not within the limits of the LCO, it must be restored to within these limits within 1 hour. The Required Action is necessary to return operation to within the bounds of the containment analysis. The 1 hour Completion Time is consistent with the ACTIONS of LCO 3.6.1, "Containment,"
which requires that containment be restored to OPERABLE status within 1 hour.
When opening or closing Penetration 1-EQH-271-0010 or 1-EQH-271-0011 in the Shield Building Dome, the differential pressure between the Containment and the Annulus may exceed the equal to or greater than -0.1 and equal to or less than +0.3 psid requirement. During this operation, time is allowed for Containment/Annulus pressure equalization to be re-established.
B.1 and B.2 If containment pressure cannot be restored to within limits within the required Completion Time, 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 6 hours and to MODE 5 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.
SURVEILLANCE      SR 3.6.4.1 REQUIREMENTS Verifying that containment pressure is within limits ( -0.1 and  +0.3 psid relative to the annulus, value does not account for instrument error, Ref. 3) ensures that plant operation remains within the limits assumed in the containment analysis.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
REFERENCES        1.      Watts Bar FSAR, Section 6.2.1, "Containment Functional Design."
: 2.      WCAP-16996-P-A, Revision 1, Realistic LOCA Evaluation Methodology Applied to the Full Spectrum of Break Sizes (FULL SPECTRUM LOCA Methodology), November 2016.
: 3.      Watts Bar Drawing 1-47W605-242, "Electrical Tech Spec Compliance Tables."
Watts Bar-Unit 1                          B 3.6-24                        Revision 29, 71, 162, 176 Amendment 59, 132, 143
 
Containment Spray System B 3.6.6 BASES APPLICABLE      The DBA analyses show that the maximum peak containment pressure of SAFETY ANALYSES  9.36 psig results from the LOCA analysis and is calculated to be less than the (continued)    containment design pressure. The maximum peak containment atmosphere temperature results from the SLB analysis. The calculated transient containment atmosphere temperatures are acceptable for the DBA SLB.
The modeled Containment Spray System actuation from the containment analysis is based on a response time associated with exceeding the containment High-High pressure signal setpoint to achieving full flow through the containment spray nozzles. A delayed response time initiation provides conservative analyses of peak calculated containment temperature and pressure responses. The Containment Spray System total response time of 221 seconds is composed of signal delay, diesel generator startup, and system startup time.
For certain aspects of transient accident analyses, maximizing the calculated containment pressure is not conservative. In particular, the ECCS cooling effectiveness during the core reflood phase of a LOCA analysis increases with increasing containment backpressure. For these calculations, the containment backpressure is calculated in a manner designed to conservatively minimize, rather than maximize, the calculated transient containment pressures in accordance with WCAP-16996-P-A, Revision 1 (Ref. 3).
Inadvertent actuation of the Containment Spray System is evaluated in the analysis, and the resultant reduction in containment pressure is calculated. The maximum calculated steady state pressure differential relative to the Shield Building annulus is 1.4 psid, which is below the containment design external pressure load of 2.0 psid.
The Containment Spray System satisfies Criterion 3 of the NRC Policy Statement.
LCO              During a DBA, one train of Containment Spray System and RHR Spray System is required to provide the heat removal capability assumed in the safety analyses. To ensure that these requirements are met, two containment spray trains and two RHR spray trains must be OPERABLE with power from two safety related, independent power supplies. Therefore, in the event of an accident, at least one train in each system operates.
Each containment spray train typically includes a spray pump, header, valves, a heat exchanger, nozzles, piping, instruments, and controls to ensure an OPERABLE flow path capable of taking suction from the RWST upon an ESF actuation signal and transferring suction to the containment sump. This suction path realignment is accomplished by manual operator action upon receipt of a Low-Low level alarm for the RWST.
(continued)
Watts Bar-Unit 1                          B 3.6-30                  Revision 44, 55, 76, 100, 127, 176 Amendment 33, 143
 
Containment Spray System B 3.6.6 BASES (continued)
SURVEILLANCE      SR 3.6.6.6 REQUIREMENTS (continued)      The Surveillance descriptions from Bases 3.5.2 for SR 3.5.2.2 and 3.5.2.4 apply as applicable to the RHR spray system.
REFERENCES        1.      Title 10, Code of Federal Regulations, Part 50, Appendix A, "General Design Criterion (GDC) 38, "Containment Heat Removal," GDC 39, "Inspection of Containment Heat Removal System," GDC 40, "Testing of Containment Heat Removal Systems, and GDC 50, "Containment Design Basis."
: 2.      Watts Bar FSAR, Section 6.2, "Containment Systems."
: 3.      WCAP-16996-P-A, Revision 1, Realistic LOCA Evaluation Methodology Applied to the Full Spectrum of Break Sizes (FULL SPECTRUM LOCA Methodology), November 2016.
: 4.      American Society of Mechanical Engineers (ASME) OM Code, "Code for Operation and Maintenance of Nuclear Power Plants."
Watts Bar-Unit 1                          B 3.6-33                              Revision 89, 176 Amendment 66, 143
 
EGTS B 3.6.9 BASES SURVEILLANCE    SR 3.6.9.2 REQUIREMENTS (continued)      This SR verifies that the required EGTS filter testing is performed in accordance with the Ventilation Filter Testing Program (VFTP-Technical Specification Section 5.7.2.14). The EGTS filter tests are in accordance with Regulatory Guide 1.52 (Ref. 4). The VFTP includes testing HEPA filter performance, charcoal adsorber efficiency, minimum system flow rate, and the physical properties of the activated charcoal (general use and following specific operations). Specific test frequencies and additional information are discussed in detail in the VFTP. It should be noted that for the EGTS, the VFTP pressure drop value across the entire filtration unit does not account for instrument error (Ref. 5).
SR 3.6.9.3 The automatic startup ensures that each EGTS train responds properly.
The SR excludes automatic dampers and valves that are locked, sealed, or otherwise secured in the actuated position. The SR does not apply to dampers or valves that are locked, sealed, or otherwise secured in the actuated position since the affected dampers or valves were verified to be in the actuated position prior to being locked, sealed, or otherwise secured.
Placing an automatic valve or damper in a locked, sealed, or otherwise secured position requires an assessment of the operability of the system or any supported systems, including whether it is necessary for the valve or damper to be repositioned to the non-actuated position to support the accident analysis. Restoration of an automatic valve or damper to the non-actuated position requires verification that the SR has been met within its required Frequency. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.6.9.4 The proper functioning of the fans, dampers, filters, adsorbers, etc., as a system is verified by the ability of each train to produce the required system flow rate within the specified timeframe. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
(continued)
Watts Bar-Unit 1                            B 3.6-43                            Revision 29, 162, 168 Amendment 132, 137
 
ARS B 3.6.10 BASES BACKGROUND      through the ice condenser doors into the ice condenser compartment where the (continued)    steam portion of the flow is condensed. The air flow returns to the upper compartment through the top deck doors in the upper portion of the ice condenser compartment. The ARS fans operate continuously after actuation, circulating air through the containment volume and purging all potential hydrogen pockets in containment. When the containment pressure falls below a predetermined value, the ARS fans are manually de-energized. Thereafter, the fans are manually cycled on and off if necessary to control any additional containment pressure transients.
The ARS also functions, after all the ice has melted, to circulate any steam still entering the lower compartment to the upper compartment where the Containment Spray System can cool it.
The ARS is an ESF system. It is designed to ensure that the heat removal capability required during the post accident period can be attained. The operation of the ARS, in conjunction with the ice bed, the Containment Spray System, and the Residual Heat Removal (RHR) System spray, provides the required heat removal capability to limit post accident conditions to less than the containment design values.
APPLICABLE      The limiting DBAs considered relative to containment temperature and pressure SAFETY ANALYSES  are the loss of coolant accident (LOCA) and the steam line break (SLB). The LOCA and SLB are analyzed using computer codes designed to predict the resultant containment pressure and temperature transients. DBAs are assumed not to occur simultaneously or consecutively. The postulated DBAs are analyzed, in regard to ESF systems, assuming the loss of one ESF bus, which is the worst case single active failure and results in one train each of the Containment Spray System, RHR System, and ARS being inoperable (Ref. 1).
The DBA analyses show that the maximum peak containment pressure results from the LOCA analysis and is calculated to be less than the containment design pressure.
For certain aspects of transient accident analyses, maximizing the calculated containment pressure is not conservative. In particular, the cooling effectiveness of the Emergency Core Cooling System during the core reflood phase of a LOCA analysis increases with increasing containment backpressure. For these calculations, the containment backpressure is calculated in a manner designed to conservatively minimize, rather than maximize, the calculated transient containment pressures, in accordance with WCAP-16996-P-A, Revision 1 (Ref. 2).
(continued)
Watts Bar-Unit 1                          B 3.6-45                                    Revision 176 Amendment 143
 
ARS B 3.6.10 BASES SURVEILLANCE    SR 3.6.10.1 REQUIREMENTS Verifying that each ARS fan starts on an actual or simulated actuation signal, after a delay of  8.0 minutes and  10.0 minutes and operates for  15 minutes is sufficient to ensure that all fans are OPERABLE and that all associated controls and time delays are functioning properly. It also ensures that blockage, fan and/or motor failure, or excessive vibration can be detected for corrective action. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.6.10.2 Verifying ARS fan motor current with the return air backdraft dampers closed confirms one operating condition of the fan. This test is indicative of overall fan motor performance. Such inservice tests confirm component OPERABILITY, trend performance, and detect incipient failures by indicating abnormal performance. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.6.10.3 Verifying the OPERABILITY of the air return damper to the proper opening torque (Ref. 3) provides assurance that the proper flow path will exist when the fan is started. By applying the correct torque to the damper shaft, the damper operation can be confirmed. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
REFERENCES      1.        Watts Bar FSAR, Section 6.8, "Air Return Fans."
: 2.        WCAP-16996-P-A, Revision 1, Realistic LOCA Evaluation Methodology Applied to the Full Spectrum of Break Sizes (FULL SPECTRUM LOCA Methodology), November 2016.
: 3.        System Description N3-30RB-4002.
Watts Bar-Unit 1                          B 3.6-47                              Revision 162, 176 Amendment 132, 143
 
Ice Bed B 3.6.11 BASES APPLICABLE      The limiting DBAs considered relative to containment temperature and SAFETY ANALYSES  pressure are the loss of coolant accident (LOCA) and the steam line break (SLB).
The LOCA and SLB are analyzed using computer codes designed to predict the resultant containment pressure and temperature transients. DBAs are not assumed to occur simultaneously or consecutively.
Although the ice condenser is a passive system that requires no electrical power to perform its function, the Containment Spray System and the ARS also function to assist the ice bed in limiting pressures and temperatures. Therefore, the postulated DBAs are analyzed in regards to containment Engineered Safety Feature (ESF) systems, assuming the loss of one ESF bus, which is the worst case single active failure and results in one train each of the Containment Spray System and ARS being inoperable.
The limiting DBA analyses (Ref. 1) show that the maximum peak containment pressure results from the LOCA analysis and is calculated to be less than the containment design pressure. For certain aspects of the transient accident analyses, maximizing the calculated containment pressure is not conservative.
In particular, the cooling effectiveness of the ECCS during the core reflood phase of a LOCA analysis increases with increasing containment backpressure.
For these calculations, the containment backpressure is calculated in a manner designed to conservatively minimize, rather than maximize, the calculated transient containment pressures, in accordance with WCAP-16996-P-A, Revision 1 (Ref. 2). The maximum peak containment atmosphere temperature results from the SLB analysis and is discussed in the Bases for LCO 3.6.5, "Containment Air Temperature."
In addition to calculating the overall peak containment pressures, the DBA analyses include calculation of the transient differential pressures that occur across subcompartment walls during the initial blowdown phase of the accident transient. The internal containment walls and structures are designed to withstand these local transient pressure differentials for the limiting DBAs.
The ice bed satisfies Criterion 3 of the NRC Policy Statement.
LCO              The ice bed LCO requires the existence of the required quantity of stored ice, appropriate distribution of the ice and the ice bed, open flow paths through the ice bed, and appropriate chemical content and pH of the stored ice. The stored ice functions to absorb heat during a DBA, thereby limiting containment air temperature and pressure. The chemical content and pH of the ice provide core SDM (boron content) and remove radioactive iodine from the containment atmosphere when the melted ice is recirculated through the ECCS and the Containment Spray System, respectively.
(continued)
Watts Bar-Unit 1                            B 3.6-50                              Revision 176 Amendment 143
 
Ice Bed B 3.6.11 BASES REFERENCES      1. Watts Bar FSAR, Section 6.2, "Containment Systems"
: 2. WCAP-16996-P-A, Revision 1, Realistic LOCA Evaluation Methodology Applied to the Full Spectrum of Break Sizes (FULL SPECTRUM LOCA Methodology), November 2016.
: 3. Watts Bar Drawing 1-47W605-242, "Electrical Tech Spec Compliance Tables"
: 4. Westinghouse Letter, WAT-D-10686, Upper Limit Ice Boron Concentration In Safety Analysis Watts Bar-Unit 1                    B 3.6-55                            Revision 176 Amendment 143
 
Ice Condenser Doors B 3.6.12 BASES APPLICABLE        Although the ice condenser is a passive system that requires no electrical SAFETY ANALYSES  power to perform its function, the Containment Spray System and ARS (continued)      also function to assist the ice bed in limiting pressures and temperatures.
Therefore, the postulated DBAs are analyzed with respect to Engineered Safety Feature (ESF) systems, assuming the loss of one ESF bus, which is the worst case single active failure and results in one train each of the Containment Spray System and the ARS being rendered inoperable.
The limiting DBA analyses (Ref. 1) show that the maximum peak containment pressure results from the LOCA analysis and is calculated to be less than the containment design pressure. For certain aspects of transient accident analyses, maximizing the calculated containment pressure is not conservative. In particular, the cooling effectiveness of the ECCS during the core reflood phase of a LOCA analysis increases with increasing containment backpressure. For these calculations, the containment backpressure is calculated in a manner designed to conservatively minimize, rather than maximize, the calculated transient containment pressures, in accordance with WCAP-16996-P-A, Revision 1 (Ref. 2).
The maximum peak containment atmosphere temperature results from the SLB analysis and is discussed in the Bases for LCO 3.6.5, "Containment Air Temperature."
An additional design requirement was imposed on the ice condenser door design for a small break accident in which the flow of heated air and steam is not sufficient to fully open the doors.
For this situation, the doors are designed so that all of the doors would partially open by approximately the same amount. Thus, the partially opened doors would modulate the flow so that each ice bay would receive an approximately equal fraction of the total flow.
This design feature ensures that the heated air and steam will not flow preferentially to some ice bays and deplete the ice there without utilizing the ice in the other bays.
In addition to calculating the overall peak containment pressures, the DBA analyses include the calculation of the transient differential pressures that would occur across subcompartment walls during the initial blowdown phase of the accident transient. The internal containment walls and structures are designed to withstand the local transient pressure differentials for the limiting DBAs.
The ice condenser doors satisfy Criterion 3 of the NRC Policy Statement.
(continued)
Watts Bar-Unit 1                            B 3.6-58                                Revision 176 Amendment 143
 
Ice Condenser Doors B 3.6.12 BASES SURVEILLANCE      SR 3.6.12.6 (continued)
REQUIREMENTS The above test lifting forces were established based upon test results gathered on newly manufactured Intermediate Deck Doors set up in fixturing to simulate plant installation tolerances. The lifting force values developed were to account for and envelope expected door panel variations in weight and hinge friction and alignments. The intent of the surveillance is to establish a method of detecting abnormalities or deteriorating conditions of the door panels or hinges after completion of refueling outage maintenance activities.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.6.12.7 Verifying, by visual inspection, that the top deck doors are in place, not obstructed, and verifying free movement of the vent assembly provides assurance that the doors are performing their function of keeping warm air out of the ice condenser during normal operation, and would not be obstructed if called upon to open in response to a DBA. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
REFERENCES        1.        Watts Bar FSAR, Section 15.0, "Accident Analysis."
: 2.        WCAP-16996-P-A, Revision 1, Realistic LOCA Evaulation Methodology Applied to the Full Spectrum of Break Sizes (FULL SPECTRUM LOCA Methodology), November 2016.
: 3.        TVA Letter to NRC dated July 31, 1996 - Proposed License Amendment
                            - Containment Systems.
Watts Bar-Unit 1                            B 3.6-63                    Revision 6, 21, 162, 165, 176 Amendment 3, 132, 135, 143
 
Shield Building B 3.6.15 BASES ACTIONS          A.1 In the event shield building OPERABILITY is not maintained, shield building OPERABILITY must be restored within 24 hours. Twenty-four hours is a reasonable Completion Time considering the limited leakage design of containment and the low probability of a Design Basis Accident occurring during this time period.
B.1 and B.2 If the shield building cannot be restored to OPERABLE status within the required Completion Time, 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 6 hours and to MODE 5 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.
SURVEILLANCE    SR 3.6.15.1 REQUIREMENTS Verifying that shield building annulus negative pressure is within limit (equal to or more negative than -1 inches water gauge, value does not account for instrument error, Ref. 2 and 3) ensures that operation remains within the limit assumed in the large break loss of coolant accident dose analysis (Ref. 6). The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.6.15.2 Maintaining shield building OPERABILITY requires maintaining each door in the access opening closed, except when the access opening is being used for normal transient entry and exit. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
(continued)
Watts Bar-Unit 1                          B 3.6-75                Revision 15, 29, 101, 129, 162, 169 Amendment 132, 139
 
Shield Building B 3.6.15 BASES SURVEILLANCE    SR 3.6.15.3 REQUIREMENTS (continued)    This SR would give advance indication of gross deterioration of the concrete structural integrity of the shield building. The Frequency of this SR is the same as that of SR 3.6.1.1. The verification is done during shutdown.
SR 3.6.15.4 The EGTS produces a negative pressure to prevent leakage from the building.
This Surveillance verifies that the shield building can be rapidly drawn down to equal to or more negative than -0.50 inches of water gauge ("wg) in the annulus at an elevation equivalent to the top of the Auxiliary Building. This test is used to ensure shield building boundary integrity. At elevations higher than the Auxiliary Building, the EGTS is required to maintain a pressure equal to or more negative than -0.25 "wg. The low pressure sense line for the pressure controller is located in the annulus at elevation 783. By verifying that the annulus pressure is equal to or more negative than -0.63 "wg at elevation 783, the annulus pressurization requirements stated above are met. The ability of an EGTS train with final flow 3600 and  4400 cfm, within 20 seconds after a start signal, to produce the required negative pressure during the test operation provides assurance that the building is adequately sealed. The time limit ensures that no significant quantity of radioactive material leaks from the Shield Building prior to developing the negative pressure. Upon failure to meet this SR, the leak tightness of the shield building must be immediately assessed to determine the impact on the OPERABILITY of the shield building. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
REFERENCES      1.      DELETED
: 2.      Watts Bar Drawing 1-47W605-242, "Electrical Tech Spec Compliance Tables."
: 3.      TVA Calculation EPMMMA121889, Temperature Induced Differential Pressure Effects in Reactor Building and Aux. Bldg. Secondary Containment, Revision 008.
: 4.      WBN UFSAR Section 6.2.3.2.2, Emergency Gas Treatment System (EGTS).
: 5.      WBN UFSAR Section 9.4.6, Reactor Building Purge Ventilating System (RBPVS).
: 6.      TVA Calculation TIRPS197, Offsite Doses Due to a Regulatory Guide 1.4 Loss of Coolant Accident, Revision 023.
Watts Bar-Unit 1                            B 3.6-76                  Revision 15, 29, 101, 129, 162, 169, 179 Amendment 132, 139, 150
 
CREVS B 3.7.10 BASES SURVEILLANCE    SR 3.7.10.1 REQUIREMENTS Standby systems should be checked periodically to ensure that they function properly. As the environment and normal operating conditions on this system are not too severe, testing each train provides an adequate check of this system.
The systems need only be operated for  15 minutes to demonstrate the function of the system. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.7.10.2 This SR verifies that the required CREVS testing is performed in accordance with the Ventilation Filter Testing Program (VFTP). The CREVS filter tests are in accordance with Regulatory Guide 1.52 (Ref. 6). The VFTP includes testing the performance of the HEPA filter, charcoal adsorber efficiency, minimum flow rate, and the physical properties of the activated charcoal. Specific test Frequencies and additional information are discussed in detail in the VFTP.
SR 3.7.10.3 This SR verifies that each CREVS train starts and operates on an actual or simulated actuation signal. The SR excludes automatic dampers and valves that are locked, sealed, or otherwise secured in the actuated position. The SR does not apply to dampers or valves that are locked, sealed, or otherwise secured in the actuated position since the affected dampers or valves were verified to be in the actuated position prior to being locked, sealed, or otherwise secured. Placing an automatic valve or damper in a locked, sealed, or otherwise secured position requires an assessment of the operability of the system or any supported systems, including whether it is necessary for the valve or damper to be repositioned to the non-actuated position to support the accident analysis.
Restoration of an automatic valve or damper to the non-actuated position requires verification that the SR has been met within its required Frequency. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.7.10.4 This SR verifies the OPERABILITY of the CRE boundary by testing for unfiltered air inleakage past the CRE boundary and into the CRE. The details of the testing are specified in the Control Room Envelope Habitability Program.
The CRE is considered habitable when the radiological dose to CRE occupants calculated in the licensing basis analyses of DBA consequences is no more than 5 rem whole body or its equivalent to any part of the body and the CRE occupants are protected from hazardous chemicals and smoke. This SR verifies that the unfiltered air inleakage into the CRE is no greater than the flow rate assumed in the licensing basis analyses of DBA consequences. When unfiltered air inleakage is greater than the assumed flow rate, Condition B must be entered.
Required Action B.3 allows time to restore the CRE boundary to OPERABLE status provided mitigating actions can ensure that the CRE remains within the (continued)
Watts Bar-Unit 1                            B 3.7-49                    Revision 64, 91, 162, 168 Amendment 50, 70, 132, 137
 
CREATCS B 3.7.11 BASES ACTIONS          A.1 (continued)
The Completion Time is modified by a footnote that states an allowance is permitted for one CREATCS train to be inoperable for 60 days. This TS provision is only authorized for one entry per train during modification activities planned for the upgrade of the main control room chillers beginning no earlier than May 1, 2022, and ending no later than May 1, 2023, provided the following compensatory measures are implemented as described in TVA letter CNL-20-012, dated May 19, 2020.
* A temporary, non-safety related chiller system with a temporary DG to provide power to the temporary chiller system will be installed and operated as described in the LAR.
* Instructions for operation of the temporary cooling equipment will be provided.
* During replacement of the CREATCS chillers, TVA will employ a graded approach to defense-in-depth and protected equipment strategies based on the operating status of the affected unit. The risk of the activity will be assessed and managed, including the use of physical barriers as needed.
Additionally, TVA procedures preclude work on or near protected equipment and limit access to the area to emergency situations and non-intrusive monitoring of running equipment per operator rounds.
* During replacement of the CREATCS chillers, no elective maintenance will be performed on TS related support equipment for the Operable CREATCS chiller except for any required TS SRs.
B.1 and B.2 In MODE 1, 2, 3, or 4, if the inoperable CREATCS train cannot be restored to OPERABLE status within the required Completion Time, the plant must be placed in a MODE that minimizes the risk. To achieve this status, the plant must be placed in at least MODE 3 within 6 hours, and in MODE 5 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.
C.1 and C.2 In MODE 5 or 6, or during movement of irradiated fuel, if the inoperable CREATCS train cannot be restored to OPERABLE status within the required Completion Time, the OPERABLE CREATCS train must be placed in operation immediately. This action ensures that the remaining train is OPERABLE, that no failures preventing automatic actuation will occur, and that active failures will be readily detected.
(continued)
Watts Bar-Unit 1                            B 3.7-53                                Revision 45, 172 Amendment 35, 145
 
CREATCS B 3.7.11 BASES ACTIONS          C.1 and C.2 (continued)
An alternative to Required Action C.1 is to immediately suspend activities that present a potential for releasing radioactivity that might require isolation of the control room. This places the unit in a condition that minimizes accident risk.
This does not preclude the movement of fuel to a safe position.
D.1 In MODE 5 or 6, or during movement of irradiated fuel assemblies, with two CREATCS trains inoperable, action must be taken immediately to suspend activities that could result in a release of radioactivity that might require isolation of the control room. This places the unit in a condition that minimizes risk. This does not preclude the movement of fuel to a safe position.
E.1 If both CREATCS trains are inoperable in MODE 1, 2, 3, or 4 the CREATCS may not be capable of performing its intended function. Therefore, LCO 3.0.3 must be entered immediately. The Completion Time is modified by a footnote that states an allowance to monitor the main control room temperature every hour and verify the main control room temperature is less than or equal to 90&deg;F is permitted for up to four days in lieu of the immediate entry into LCO 3.0.3. If the main control room temperature exceeds 90&deg;F, or the duration without a train of CREATCS being OPERABLE exceeds four days, immediate entry into LCO 3.0.3 is required. This provision is only applicable during modification activities planned for the upgrade of the main control room chillers beginning no earlier than May 1, 2022, and ending no later than May 1, 2023, provided the following compensatory measures are implemented as described in TVA letter CNL-20-012, dated May 19, 2020.
* A temporary, non-safety related chiller system with a temporary DG to provide power to the temporary chiller system will be installed and operated as described in the LAR.
* Instructions for operation of the temporary cooling equipment will be provided.
* During replacement of the CREATCS chillers, TVA will employ a graded approach to defense-in-depth and protected equipment strategies based on the operating status of the affected unit. The risk of the activity will be assessed and managed, including the use of physical barriers as needed.
Additionally, TVA procedures preclude work on or near protected equipment and limit access to the area to emergency situations and non-intrusive monitoring of running equipment per operator rounds.
(continued)
Watts Bar-Unit 1                          B 3.7-54                                  Revision 45, 172 Amendment 35, 145
 
CREATCS B 3.7.11 BASES ACTIONS          E.1 (continued)
* During replacement of the CREATCS chillers, no elective maintenance will be performed on TS related support equipment for the Operable CREATCS chiller except for any required TS SRs.
The purpose of the footnote is to ensure the MCR temperature is being controlled. The specified temperature limit of 90&deg;F is above the normal operating temperature of the MCR (approximately 75&deg;F), providing operational flexibility when implementing the mitigating actions. This temperature does not impact the operability of equipment or habitability of the MCR. The limit of 90&deg;F maintains margin below the lowest specification for the MCR equipment cabinets of 104&deg;F.
Subsequent to immediate MCR temperature verification, the one-hour frequency is adequate given the indications available in the MCR. Main control room temperature data is measured and displayed from readily available equipment in the MCR and operators will have awareness of temperature trending relative to the 90&deg;F limit.
SURVEILLANCE    SR 3.7.11.1 REQUIREMENTS This SR verifies that the heat removal capability of the system is sufficient to remove the heat load assumed in the sizing calculations in the control room.
This SR consists of a combination of testing and calculations. This is accomplished by verifying that the system has not degraded. The only measurable parameters that could degrade undetected during normal operation are the system air flow and chilled water flow rate. Verification of these two flow rates will provide assurance that the heat removal capacity of the system is still adequate. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
REFERENCES      1.        Watts Bar FSAR, Section 9.4.1, "Control Room Area Ventilation System."
: 2.        Watts Bar FSAR, Section 3.7.3.18, Seismic Qualification of Main Control Room Suspended Ceiling and Air Delivery Components.
: 3.        NRC Safety Evaluation dated February 12, 2004, for License Amendment 50.
Watts Bar-Unit 1                          B 3.7-54a                          Revision 64, 162, 172 Amendment 50, 132, 145
 
ABGTS B 3.7.12 BASES (continued)
LCO                Two independent and redundant trains of the ABGTS are required to be OPERABLE to ensure that at least one train is available, assuming a single failure that disables the other train, coincident with a loss of offsite power. Total system failure, such as from a loss of both ventilation trains or from an inoperable ABSCE boundary, could result in exceeding a dose of 5 rem whole body or its equivalent to any part of the body to the main control room occupants in the event of a large radioactive release.
The ABGTS is considered OPERABLE when the individual components necessary to control exposure in the fuel handling building are OPERABLE in both trains. An ABGTS train is considered OPERABLE when its associated:
: a.        Fan is OPERABLE;
: b.        HEPA filter and charcoal adsorber are not excessively restricting flow, and are capable of performing their filtration function; and
: c.        Heater, moisture separator, ductwork, valves, and dampers are OPERABLE, and air circulation can be maintained.
The LCO is modified by a Note allowing the ABSCE boundary to be opened intermittently under administrative controls that ensure the ABSCE can be closed consistent with the safety analysis. For entry and exit through doors the administrative control of the opening is performed by the person(s) entering or exiting the area. For other openings, these controls are proceduralized and consist of stationing a dedicated individual at the opening who is in continuous communication with the control room. This individual will have a method to rapidly close the opening when a need for auxiliary building isolation is indicated. The ABSCE boundary must be able to be restored within four minutes (including the time for restoration of the ABSCE boundary and drawdown) in accordance with UFSAR Section 15.5.3.
As a one-time exception for the Watts Bar Unit 2 Cycle 4 Refueling Outage, scheduled to commence in spring 2022, during which the Unit 2 Replacement Steam Generators (RSGs) will be installed, the breaches of the ABSCE boundary needed to support the Unit 2 RSG project activities (Unit 2 Upper Containment Personnel Air Lock Access, Unit 2 Lower Containment Personnel Air Lock Access, Unit 2 Containment Equipment Hatch, and Auxiliary Building General Supply Fan 737 Elevation Room A12 Access and Backup) may be opened on a continuous basis, under administrative controls that ensure the ABSCE can be closed consistent with the safety analysis.
These controls are the same as for the existing Note as described above.
APPLICABILITY      In MODE 1, 2, 3, or 4, the ABGTS is required to be OPERABLE to provide fission product removal associated with ECCS leaks due to a LOCA and leakage from containment and annulus.
In MODE 5 or 6, the ABGTS is not required to be OPERABLE since the ECCS is not required to be OPERABLE.
(continued)
Watts Bar-Unit 1                              B 3.7-56                    Revision 55, 87, 119, 139, 182 Amendment 92, 116, 151
 
ABGTS B 3.7.12 BASES (continued)
SURVEILLANCE      SR 3.7.12.1 REQUIREMENTS Standby systems should be checked periodically to ensure that they function properly. As the environmental and normal operating conditions on this system are not severe, testing each train provides an adequate check on this system.
Operation with the heaters on for  15 continuous minutes demonstrates OPERABILITY of the system. Periodic operation ensures that heater failure, blockage, fan or motor failure, or excessive vibration can be detected for corrective action. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.7.12.2 This SR verifies that the required ABGTS testing is performed in accordance with the Ventilation Filter Testing Program (VFTP). The ABGTS filter tests are in accordance with Regulatory Guide 1.52 (Ref. 8). The VFTP includes testing HEPA filter performance, charcoal adsorber efficiency, minimum system flow rate, and the physical properties of the activated charcoal (general use and following specific operations). Specific test frequencies and additional information are discussed in detail in the VFTP.
SR 3.7.12.3 This SR verifies that each ABGTS train starts and operates on an actual or simulated actuation signal. The SR excludes automatic dampers and valves that are locked, sealed, or otherwise secured in the actuated position. The SR does not apply to dampers or valves that are locked, sealed, or otherwise secured in the actuated position since the affected dampers or valves were verified to be in the actuated position prior to being locked, sealed, or otherwise secured.
Placing an automatic valve or damper in a locked, sealed, or otherwise secured position requires an assessment of the operability of the system or any supported systems, including whether it is necessary for the valve or damper to be repositioned to the non-actuated position to support the accident analysis.
Restoration of an automatic valve or damper to the non-actuated position requires verification that the SR has been met within its required Frequency. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.7.12.4 This SR verifies the integrity of the ABSCE. The ability of the ABSCE to maintain negative pressure with respect to potentially uncontaminated adjacent areas is periodically tested to verify proper function of the ABGTS. During the post accident mode of operation, the ABGTS is designed to maintain a slight negative pressure in the ABSCE, to prevent unfiltered LEAKAGE. The ABGTS is designed to maintain a negative pressure between -0.25 and -0.5 inches water gauge (continued)
Watts Bar-Unit 1                            B 3.7-58                        Revision 29, 35, 162, 168 Amendment 132, 137
 
Spent Fuel Pool Assembly Storage B 3.7.15 B 3.7 PLANT SYSTEMS B 3.7.15 Spent Fuel Pool Assembly Storage BASES BACKGROUND            The spent fuel pool contains flux trap rack modules with 1386 storage positions that are designed to accommodate fuel with a maximum enrichment of 4.95 +/-
0.05 weight percent U-235 without restrictions.
The water in the spent fuel storage pool normally contains soluble boron, which results in large subcriticality margins under actual operating conditions. The double contingency principle discussed in ANSI N-16.1-1975, and the April 1978 NRC letter (Reference 1) allows credit for soluble boron under other abnormal or accident conditions, since only a single accident need be considered at one time.
To mitigate postulated criticality-related events, boron is dissolved in the pool water.
APPLICABLE            The accident analyses are provided in the FSAR.
SAFETY ANALYSES The initial enrichment of fuel assemblies in the fuel storage pool along with the concentration of dissolved boron in the fuel storage pool satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).
LCO                  The restrictions on the initial enrichment of fuel assemblies within the spent fuel pool in accordance with Specification 4.3.1.1 in the accompanying LCO, ensures the keff will always remain subcritical, assuming the pool to be flooded with unborated water.
APPLICABILITY        This LCO applies whenever any fuel assembly is stored in the spent fuel storage pool.
(continued)
Watts Bar-Unit 1                                B 3.7-66                              Revision 11, 61, 167 Amendment 6, 40, 125
 
Spent Fuel Pool Assembly Storage B 3.7.15 BASES (continued)
ACTIONS          A.1 Required Action A.1 is modified by a Note indicating that LCO 3.0.3 does not apply.
If unable to move irradiated fuel assemblies while in Mode 5 or 6, LCO 3.0.3 would not be applicable. If unable to move irradiated fuel assemblies while in Mode 1, 2, 3, or 4, the action is independent of reactor operation. Therefore, inability to move fuel assemblies is not sufficient reason to require a reactor shutdown.
When the initial enrichment of fuel assemblies stored in the spent fuel storage pool is not in accordance with Specification 4.3.1.1, the immediate action is to initiate action to make the necessary fuel assembly movements to bring the configuration into compliance with Specification 4.3.1.1.
SURVEILLANCE      SR 3.7.15.1 REQUIREMENTS This SR verifies by administrative means that the initial enrichment of the fuel assembly is in accordance with Specification 4.3.1.1 in the accompanying LCO.
REFERENCES        1.        Double contingency principle of ANSI N16.1-1975, as specified in the April 14, 1978, NRC letter (Section 1.2) and implied in the proposed revision to Regulatory Guide 1.13 (Section 1.4, Appendix A).
: 2.        FSAR Section 4.3.2.7.
Watts Bar-Unit 1                            B 3.7-67                            Revision 11, 61, 167 Amendment 6, 40, 125
 
Spent Fuel Pool Assembly Storage B 3.7.15 BASES (continued)
INTENTIONALLY LEFT BLANK Watts Bar-Unit 1          B 3.7-68                    Revision 11, 61, 167 Amendment 6, 40, 125
 
Fuel Storage Pool Boron Concentration B 3.7.18 B 3.7 PLANT SYSTEMS B 3.7.18 Fuel Storage Pool Boron Concentration BASES BACKGROUND            In the BORALTM flux trap rack design, the spent fuel storage pool is designed to accommodate new fuel with a maximum enrichment of 4.95 +/- 0.05 wt % U-235, or spent fuel regardless of the discharge fuel burnup.
The water in the spent fuel storage pool normally contains soluble boron, which results in large subcriticality margins under actual operating conditions. Analysis demonstrates that the effective neutron multiplication factor (keff) of the spent fuel pool loaded with fuel of the highest anticipated reactivity, at a temperature corresponding to the highest reactivity, is less than 1.0 for the pool flooded with unborated water, and does not exceed 0.95 for the pool flooded with borated water with 500 ppm soluble boron (an additional 50 ppm of soluble boron has been added to account for grid growth.) The double contingency principle discussed in ANSI N-16.1-1975 and the April 1978 NRC letter (Ref. 1) allows credit for soluble boron under other abnormal or accident conditions, since only a single accident need be considered at one time. To mitigate postulated criticality related events, boron is dissolved in the pool water.
APPLICABLE            The following accident conditions have been evaluated:
SAFETY                      The effect of SFP temperature exceeding the normal range ANALYSES                    A dropped fuel assembly A misloaded fuel assembly (a fuel assembly in the wrong location within the storage rack)
A mislocated fuel assembly (a fuel assembly in the wrong location outside the storage rack)
Rack movement due to seismic activity The results of the evaluation show that a soluble boron concentration of 500 ppm is sufficient to ensure that the maximum keff is below the regulatory limit of 0.95.
The boron dilution analysis assumes an initial boron concentration of 2,300 ppm for the limiting evaluation. The accident analyses are provided in the FSAR, Section 4.3.2.7 (Ref. 2)
The concentration of dissolved boron in the fuel storage pool satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).
(continued)
Watts Bar-Unit 1                                  B 3.7-81                                    Revision 167 Amendment 125
 
Fuel Storage Pool Boron Concentration B 3.7.18 BASES (continued)
LCO              The fuel storage pool boron concentration is required to be 2300 ppm. The specified concentration of dissolved boron in the fuel storage pool preserves the assumptions used in the analyses. This concentration of dissolved boron is the minimum required concentration for fuel assembly storage and movement within the fuel storage pool.
APPLICABILITY    This LCO applies whenever fuel assemblies are stored in the spent fuel storage pool.
ACTIONS          A.1 The Required Actions are modified by a Note indicating that LCO 3.0.3 does not apply.
When the concentration of boron in the fuel storage pool is less than required, immediate action must be taken to preclude the occurrence of an accident or to mitigate the consequences of an accident in progress. This is most efficiently achieved by immediately suspending the movement of fuel assemblies. The concentration of boron is restored simultaneously with suspending movement of fuel assemblies. However, prior to resuming movement of the fuel assemblies, the concentration of boron must be restored. This does not preclude movement of a fuel assembly to a safe position.
If the LCO is not met while moving irradiated fuel assemblies in MODE 5 or 6, LCO 3.0.3 would not be applicable. If moving irradiated fuel assemblies while in MODE 1, 2, 3, or 4, the fuel movement is independent of reactor operation.
Therefore, inability to suspend movement of fuel assemblies is not sufficient reason to require a reactor shutdown.
SURVEILLANCE      SR 3.7.18.1 REQUIREMENTS This SR verifies that the concentration of boron in the fuel storage pool is within the required limit. As long as this SR is met, the analyzed accidents are fully addressed. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
REFERENCES        1. Double contingency principle of ANSI N16.1-1975, as specified in the April 14, 1978 NRC letter (Section 1.2) and implied in the proposed revision to Regulatory Guide 1.13 (Section 1.4, Appendix A)
: 2. FSAR, Section 4.3.2.7 Watts Bar-Unit 1                            B 3.7-82                                    Revision 167 Amendment 125, 132
 
AC Sources - Operating B 3.8.1 BASES (continued)
BACKGROUND        When credited, the available 6.9 kV FLEX DG must be able to connect to a (continued)      shutdown board that will power the necessary components to facilitate unit shutdown and cooldown. Typically, the FLEX DG would be available to be alighted to the shutdown board associated with the inoperable DG. However, if the DG is inoperable because the associated 6.9 kV Shutdown Board is out of service (e.g., for shutdown board cleaning), then the available FLEX DG must be able to align to an available shutdown board to ensure an entire train of ESF systems remains available to maintain plant safety.
APPLICABLE        The initial conditions of DBA and transient analyses in the SAFETY ANALYSES  FSAR, Section 6 (Ref. 4) and Section 15 (Ref. 5), assume ESF systems are OPERABLE. The AC electrical power sources are designed to provide sufficient capacity, capability, redundancy, and reliability to ensure the availability of necessary power to ESF systems so that the fuel, Reactor Coolant System (RCS), and containment design limits are not exceeded. These limits are discussed in more detail in the Bases for Section 3.2, Power Distribution Limits; Section 3.4, Reactor Coolant System (RCS); and Section 3.6, Containment Systems.
The OPERABILITY of the AC electrical power sources is consistent with the initial assumptions of the Accident analyses and is based upon meeting the design basis of the plant. This results in maintaining at least two DG's associated with one load group or one offsite circuit OPERABLE during Accident conditions in the event of:
: a.        An assumed loss of all offsite power or all onsite AC power; and
: b.        A worst case single failure.
The AC sources satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii).
LCO              Two qualified circuits between the Watts Bar Hydro 161 kV switchyard and the onsite Class 1E Electrical Power System and separate and independent DGs for each train ensure availability of the required power to shut down the reactor and maintain it in a safe shutdown condition after an anticipated operational occurrence (AOO) or a postulated DBA.
Qualified offsite circuits are those that are described in the FSAR and are part of the licensing basis for the plant.
Each offsite circuit must be capable of maintaining acceptable frequency and voltage, and accepting required loads during an accident, while connected to the 6.9 kV shutdown boards.
(continued)
Watts Bar-Unit 1                              B 3.8-3                      Revision 125, 132, 161, 180 Amendment 103, 110
 
AC Sources - Operating B 3.8.1 BASES LCO                      Each DG must be capable of starting, accelerating to rated speed and voltage, (continued)              and connecting to its respective 6.9 kV shutdown board on detection of loss-of-voltage. This will be accomplished within 10 seconds. Each DG must also be capable of accepting required loads within the assumed loading sequence intervals and continue to operate until offsite power can be restored to the 6.9 kV shutdown boards. These capabilities are required to be met from a variety of initial conditions such as DG in standby with the engine hot and DG in standby with the engine at ambient conditions. Additional DG capabilities must be demonstrated to meet required Surveillances, e.g., capability of the DG to revert to standby status on an accident signal while operating in parallel test mode.
Proper sequencing of loads, including tripping of nonessential loads, is a required function for DG OPERABILITY.
The AC sources in one train must be separate and independent (to the extent possible) of the AC sources in the other train. For the DGs, separation and independence are complete. However, CSST A or B can only supply two 6.9 kV shutdown boards in the same load group, to ensure the separation criteria is met.
For the offsite AC sources, separation and independence are to the extent practical. A circuit may be connected to more than one ESF bus, with fast transfer capability to the other circuit OPERABLE, and not violate separation criteria. A circuit that is not connected to an ESF bus is required to have OPERABLE fast transfer interlock mechanisms to at least two ESF buses to support OPERABILITY of that circuit.
Note: The offsite power configurations identified above for normal operation and alternate operation are mutually exclusive with respect to OPERABILITY of offsite circuits. For example, for normal operation, if 6.9 kV Shutdown Board 2A-A is out of service for planned maintenance (e.g., shutdown board cleaning), then the offsite circuit defined as from the 161 kV Watts Bar Hydro Switchyard through CSST C to 6.9 kV Shutdown Board 1A-A and to 6.9 kV Shutdown Board 2A-A would be inoperable because the normally closed circuit breaker connecting 6.9 kV Shutdown Board 2A-A would be open. Additionally, the normally open circuit breaker connecting CSST D to 6.9 kV Shutdown Board 2A-A, which is credited for alternate operation, would also remain open. Alternate operation would not be available for the offsite circuit associate with CSST D. However, for normal operation, the offsite circuit defined as the 161 kV Watts Bar Hydro Switchyard, through CSST D to 6.9 kV Shutdown Board 1B-B and to 6.9 kV Shutdown Board 2B-B would still be operable. Therefore, for Unit 2 shutdown board cleaning, only one offsite circuit would be considered inoperable (and Condition D would be entered).
(continued)
Watts Bar-Unit 1                                      B 3.8-6                              Revision 125, 180 Amendment 84, 103
 
AC Sources - Operating B 3.8.1 BASES ACTIONS          B.1 and C.1 (continued)
To ensure a highly reliable power source remains with one or more DGs inoperable in Train A OR with one or more DGs inoperable in Train B, it is necessary to verify the availability of the required offsite circuits on a more frequent basis. Since the Required Action only specifies perform, a failure of SR 3.8.1.1 acceptance criteria does not result in a Required Action being not met. However, if a circuit fails to pass SR 3.8.1.1, it is inoperable. Upon required offsite circuit inoperability, additional Conditions and Required Actions must then be entered.
B.2 In order to extend the Required Action B.5 Completion Time for an inoperable DG from 72 hours to 10 days, it is necessary to evaluate the availability of the 6.9 kV FLEX DG within 2 hours upon entry into LCO 3.8.1 and every 12 hours thereafter. Since Required Action B.2 only specifies evaluate, discovering the 6.9 kV FLEX DG unavailable does not result in the Required Action being not met (i.e., the evaluation is performed). However, on discovery of an unavailable 6.9 kV FLEX DG, the Completion Time for Required Action B.5 starts the 72 hour and/or 24 hour clock.
6.9 kV FLEX DG availability requires that:
: 1) 6.9 kV FLEX DG fuel tank level is verified locally to be  8-hour supply; and
: 2) 6.9 kV FLEX DG supporting system parameters for starting and operating are verified to be within required limits for functional availability (e.g., batter state of charge).
The 6.9 kV FLEX DG is not used to extend the Completion Time for more than one inoperable DG at any one time.
If the inoperable DG is inoperable because the associated Unit 2 6.9 kV Shutdown Board is out of service for planned maintenance (e.g., shutdown board cleaning), then the available FLEX DG must be ale to align to an available shutdown board to ensure an entire train of ESF systems remains available to maintain plant safety and to support the ability to shutdown and cooldown both units.
Note that entry into both Conditions B and D would be appropriate for planned maintenance such as shutdown board cleaning.
(continued)
Watts Bar-Unit 1                            B 3.8-10                        Revision 50, 125, 132, 180 Amendment 39, 84, 103, 110
 
AC Sources - Operating B 3.8.1 BASES ACTIONS          C.4 (continued)
In Condition C, the remaining OPERABLE DGs and offsite circuits are adequate to supply electrical power to the onsite Class 1E Distribution System. The 72 hour Completion Time takes into account the capacity and capability of the remaining AC sources, a reasonable time for repairs, and the low probability of a DBA occurring during this period. Restoration of at least on DG within 72 hours results in reverting back under Condition B and continuing to track the time zero Completion Time for one DG inoperable.
The second Completion Time for Required Action C.4 establishes a limit on the maximum time allowed for any combination of required AC power sources to be inoperable during any single contiguous occurrence of failing to meet the LCO. If Condition C is entered while, for instance, an offsite circuit is inoperable and that circuit is subsequently restored OPERABLE, the LCO may already have been not met for up to 72 hours. This could lead to a total of 144 hours, since initial failure to meet the LCO, to restore the DGs. At this time, an offsite circuit could again become inoperable, the DGs restored OPERABLE, and an additional 72 hours (for a total of 9 days) allowed prior to complete restoration of the LCO.
The 6 day Completion Time provides a limit on time allowed in a specified condition after discovery of failure to meet the LCO. This limit is considered reasonable for situations in which Conditions A and C are entered concurrently.
The AND connector between the 72 hour and 6 day Completion Times means that both Completion Times apply simultaneously, and the more restrictive Completion Time must be met.
As in Required Action C.2, the Completion Time allows for an exception to the normal time zero for beginning the allowed outage time clock. This will result in establishing the time zero at the time that the LCO was initially not met, instead of at the time Condition C was entered.
D.1, D.2, and D.3 Condition D is modified by two notes that limit the conditions that allow entry into Condition D. The first note states that Condition D is only applicable during planned maintenance. This will allow the plant configuration to be aligned to minimize features being inoperable when the opposite unit shutdown board is made inoperable. The second note limits the applicability of Condition D to the time period when the opposite unit is defueled. This note limits the time period allowing Condition D to be entered, minimizing when the allowance can be utilized.
Note that entry into both Conditions B and D would be appropriate for planned maintenance such as shutdown board cleaning.
(continued)
Watts Bar-Unit 1                          B 3.8-14                        Revision 50, 125, 132, 180 Amendment 39, 84, 103,110
 
AC Sources - Operating B 3.8.1 BASES ACTIONS          E.1 and E.2 (continued)
: b.        The time required to detect and restore an unavailable required offsite power source is generally much less than that required to detect and restore an unavailable onsite AC source.
With both of the required offsite circuits inoperable, sufficient onsite AC sources are available to maintain the plant in a safe shutdown condition in the event of a DBA or transient. In fact, a simultaneous loss of offsite AC sources, a LOCA, and a worst case single failure were postulated as a part of the design basis in the safety analysis. Thus, the 24 hour Completion Time provides a period of time to effect restoration of one of the offsite circuits commensurate with the importance of maintaining an AC electrical power system capable of meeting its design criteria.
According to Reference 6, with the available offsite AC sources, two less than required by the LCO, operation may continue for 24 hours. If two offsite sources are restored within 24 hours, unrestricted operation may continue. If only one offsite source is restored within 24 hours, power operation continues in accordance with Condition A or Condition D, as applicable.
F.1 and F.2 According to Regulatory Guide 1.93 (Ref. 6), operation may continue in Condition F for a period that should not exceed 12 hours.
In Condition F, individual redundancy is lost in both the offsite electrical power system and the onsite AC electrical power system. Since power system redundancy is provided by two diverse sources of power, however, the reliability of the power systems in this Condition may appear higher than that in Condition E (loss of both required offsite circuits). This difference in reliability is offset by the susceptibility of this power system configuration to a single bus or switching failure. The 12 hour Completion Time takes into account the capacity and capability of the remaining AC sources, a reasonable time for repairs, and the low probability of a DBA occurring during this period.
Note that if an offsite circuit and a DG are inoperable because the associated Unit 2 6.9 kV Shutdown Board is out of service for planned maintenance (e.g.,
shutdown board cleaning), then entry into both Conditions B and D is appropriate and a longer Completion Time (i.e., a maximum of 7 days instead of 12 hours) is justified because Unit 2 is defueled and Condition E of LOC 3.8.9 is entered simultaneously, which results in the associated required downstream fed subsystems being declared inoperable immediately so the subsystems will be governed by their own LCOs (Ref. 12).
(continued)
Watts Bar-Unit 1                            B 3.8-17                    Revision 50, 125, 132, 158, 180 Amendment 39, 84, 110, 129
 
AC Sources - Operating B 3.8.1 BASES REFERENCES      1. Title 10, Code of Federal Regulations, Part 50, Appendix A, General Design Criterion (GDC) 17, Electrical Power Systems.
: 2. Watts Bar FSAR, Section 8.2, Offsite Power System, and Tables 8.3-1 to 8.3-3, Safety-Related Standby Power Sources and Distribution Boards, Shutdown Board Loads Automatically Tripped Following a Loss of Nuclear Unit and Preferred Power, and Diesel Generator Load Sequentially Applied Following a Loss of Nuclear Unit and Preferred Power.
: 3. Regulatory Guide 1.9, Rev. 3, Selection, Design, Qualification and Testing of Emergency Diesel Generator Units Used as Class 1E Onsite Electric Power Systems at Nuclear Power Plants, July 1993.
: 4. Watts Bar FSAR Section 6, Engineered Safety Features.
: 5. Watts Bar FSAR, Section 15.4, Condition IV-Limiting Faults.
: 6. Regulatory Guide 1.93, Rev. 0, Availability of Electric Power Sources, December 1974.
: 7. Generic Letter 84-15, Proposed Staff Actions to Improve and Maintain Diesel Generator Reliability, July 2, 1984.
: 8. Title 10, Code of Federal Regulations, Part 50, Appendix A, GDC 18, Inspection and Testing of Electric Power Systems.
: 9. Regulatory Guide 1.137, Rev. 1, Fuel Oil Systems for Standby Diesel Generators, October 1979.
: 10. Watts Bar Drawing 1-47W605-242, Electrical Tech Spec Compliance Tables.
: 11. Generic Letter 84-15, "Proposed Staff Actions to Improve and Maintain Diesel Generator Reliability," dated July 2, 1984.
: 12. Letter from Kimberly J. Green (NRC) to Mr. James Barstow (TVA) dated November 26, 2019, with
 
==Enclosures:==
(1) Amendment No. 129 to Facility Operating License No. NPF-90, (2) Amendment No. 32 to Facility Operating License No. NPF-96, and (3) NRC Safety Evaluation.
(continued)
Watts Bar-Unit 1                      B 3.8-3                            Revision 50, 125, 180 Amendment 39, 84
 
Distribution Systems - Operating B 3.8.9 Table B 3.8.9-1 (page 1 of 1)
AC and DC Electrical Power Distribution Systems TYPE                      VOLTAGE                        TRAIN A*                            TRAIN B*
AC safety buses                  6900 V                        Shdn Bd                            Shdn Bd 1A-A, 2A-A                          1B-B, 2B-B 480 V                        Shdn Bd                            Shdn Bd 1A1-A, 1A2-A                      1B1-B, 1B2-B 2A1-A, 2A2-A                      2B1-B, 2B2-B Rx MOV Bd                          Rx MOV Bd 1A1-A, 1A2-A                      1B1-B, 1B2-B 2A1-A**, 2A2-A                    2B1-B**, 2B2-B C & A Vent Bd                    C & A Vent Bd 1A1-A                              1B1-B 2A1-A                              2B1-B Rx Vent Bd                        Rx Vent Bd 1A-A, 2A-A**                      1B-B, 2B-B**
AC vital buses                  120 V                        Channel I                          Channel II Vital bus 1-I                    Vital bus 1-II Vital bus 2-I                    Vital bus 2-II Channel III                        Channel IV Vital bus 1-III                    Vital bus 1-IV Vital bus 2-III                    Vital bus 2-IV DC buses                      125 V                        Board I                            Board II Board III                          Board IV
* Each train of the AC and DC electrical power distribution systems is a subsystem.
      **        For WBN Unit 1, 480V Reactor MOV Boards 2A1-A and 2B1-B and 480V Reactor Vent Boards 2A-A and 2B-B are available for economic and operational convenience. The boards contain no Unit 1 Technical Specification (TS) Required loads. The boards are considered part of the Unit 1 / Unit 2 Electrical Power Distribution System and meet Unit 1 TS Requirements and testing only while connected. WBN Unit 1 is designed to be operated, shutdown, and maintained in a safe shutdown status without any of these boards or their loads. As such, the boards may be disconnected from service without entering a Unit 1 LCO provided their loads are not substituting for a Unit 1 TS Required load.
Watts Bar-Unit 1                                            B 3.8-97                      Revision 33, 124, 152, 163, 181 Amendment 126
 
Spent Fuel Pool Boron Concentration B 3.9.9 B 3.9 REFUELING OPERATIONS B 3.9.9 Spent Fuel Pool Boron Concentration BASES BACKGROUND            The spent fuel storage rack criticality analysis assumes 2300 ppm soluble boron in the fuel pool when fuel is being stored.
APPLICABLE            This requirement ensures the presence of at least 2300 ppm soluble boron SAFETY ANALYSES        in the spent fuel pool water as assumed in the spent fuel rack criticality analysis for normal storage and a dropped fuel assembly event.
The RCS boron concentration satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii) .
LCO                    The LCO requires that the boron concentration in the spent fuel pool be greater than or equal to 2300 ppm anytime fuel is being stored in the pool.
APPLICABILITY          This LCO is applicable when the spent fuel pool is flooded and fuel is in the pool.
The assembly is verified to comply with the criticality loading criteria specified in Specification 4.3.1.1 before placing it in the Spent Fuel Pool.
ACTIONS        A.1 If the spent fuel pool boron concentration does not meet the above requirements, action must be initiated to restore fuel storage pool boron concentration to within limits.
Suspension of CORE ALTERATIONS and positive reactivity additions shall not preclude moving a component to a safe position.
(continued)
Watts Bar-Unit 1                                B 3.9-22                            Revision 11, 86, 167 Amendment 6, 125
 
Spent Fuel Pool Boron Concentration B 3.9.9 BASES (continued)
SURVEILLANCE      SR 3.9.9.1 REQUIREMENTS This SR requires that the spent fuel pool boron concentration be verified greater than or equal to 2300 ppm. This surveillance is to be performed when fuel is stored in the spent fuel pool and in accordance with the Surveillance Frequency Control Program.
REFERENCES        1. Watts Bar FSAR, Section 4.3.2.7.
Watts Bar-Unit 1                            B 3.9-23                                  Revision 167 Amendment 125, 132
 
ENCLOSURE 3 WBN UNIT 1 TECHNICAL REQUIREMENTS MANUAL TABLE OF CONTENTS (16 pages)
 
TABLE OF CONTENTS TECHNICAL REQUIREMENTS TABLE OF CONTENTS LIST OF TABLES ......................................................................................................................................... v LIST OF FIGURES....................................................................................................................................... vi LIST OF ACRONYMS ................................................................................................................................. vii LIST OF EFFECTIVE PAGES .................................................................................................................... viii 1.0            USE AND APPLICATION ................................................................................................ 1.1-1 1.1                        Definitions ........................................................................................................... 1.1-1 1.2                  Logical Connectors .................................................................................................. 1.2-1 1.3                  Completion Times .................................................................................................... 1.3-1 1.4                  Frequency................................................................................................................. 1.4-1 TR 3.0          APPLICABILITY ............................................................................................................... 3.0-1 TR 3.1          REACTIVITY CONTROL SYSTEMS .............................................................................. 3.1-1 TR 3.1.1                  Boration Systems Flow Paths, Shutdown ......................................................... 3.1-1 TR 3.1.2                  Boration Systems Flow Paths, Operating ......................................................... 3.1-3 TR 3.1.3                  Charging Pump, Shutdown ................................................................................ 3.1-5 TR 3.1.4                  Charging Pumps, Operating .............................................................................. 3.1-6 TR 3.1.5                  Borated Water Sources, Shutdown ................................................................... 3.1-8 TR 3.1.6                  Borated Water Sources, Operating ................................................................... 3.1-10 TR 3.1.7                  Position Indication System, Shutdown .............................................................. 3.1-13 TR 3.3          INSTRUMENTATION ...................................................................................................... 3.3-1 TR 3.3.1                  Reactor Trip System (RTS) Instrumentation..................................................... 3.3-1 TR 3.3.2                  Engineered Safety Features Actuation System (ESFAS) Instrumentation ..................................................... 3.3-5 TR 3.3.3                  Movable Incore Detectors .................................................................................. 3.3-12 TR 3.3.4                  Seismic Instrumentation ..................................................................................... 3.3-14 TR 3.3.5                  Turbine Overspeed Protection........................................................................... 3.3-18 TR 3.3.6                  Loose-Part Detection System ............................................................................ 3.3-20 TR 3.3.7                  Plant Calorimetric Measurement ....................................................................... 3.3-22 TR 3.3.8                  Hydrogen Monitors ............................................................................................. 3.3-24 TR 3.3.9                  Power Distribution Monitoring System (PDMS) ................................................ 3.3-26 TR 3.4          REACTOR COOLANT SYSTEM (RCS) ......................................................................... 3.4-1 TR 3.4.1                  Safety Valves, Shutdown ................................................................................... 3.4-1 TR 3.4.2                  Pressurizer Temperature Limits ........................................................................ 3.4-3 TR 3.4.3                  RCS Vents .......................................................................................................... 3.4-5 TR 3.4.4                  Chemistry ............................................................................................................ 3.4-7 TR 3.4.5                  Piping System Structural Integrity ..................................................................... 3.4-10 TR 3.6          CONTAINMENT SYSTEMS ............................................................................................ 3.6-1 TR 3.6.1                  Ice Bed Temperature Monitoring System ......................................................... 3.6-1 TR 3.6.2                  Inlet Door Position Monitoring System .............................................................. 3.6-4 TR 3.6.3                  Lower Compartment Cooling (LCC) System .................................................... 3.6-6 (continued)
Watts Bar-Unit 1                                                              i Technical Requirements                                                                                                                                Revision 56
 
TABLE OF CONTENTS (continued)
TR 3.7          PLANT SYSTEMS ........................................................................................................... 3.7-1 TR 3.7.1              Steam Generator Pressure/
Temperature Limitations ...................................................................... 3.7-1 TR 3.7.2              Flood Protection Plan ......................................................................................... 3.7-3 TR 3.7.3              DELETED. .......................................................................................................... 3.7-10 TR 3.7.4              Sealed Source Contamination ........................................................................... 3.7-22 TR 3.7.5              Area Temperature Monitoring............................................................................ 3.7-26 TR 3.8          ELECTRICAL POWER SYSTEMS ................................................................................. 3.8-1 TR 3.8.1              Isolation Devices ................................................................................................ 3.8-1 TR 3.8.2              Containment Penetration Conductor Overcurrent Protection Devices................................................................................ 3.8-5 TR 3.8.3              Motor-Operated Valves Thermal Overload Bypass Devices .................................................................................... 3.8-10 TR 3.8.4              Submerged Component Circuit Protection ....................................................... 3.8-17 TR 3.9          REFUELING OPERATIONS ........................................................................................... 3.9-1 TR 3.9.1              Deleted ................................................................................................................ 3.9-1 TR 3.9.2              Communications ................................................................................................. 3.9-2 TR 3.9.3              Refueling Machine.............................................................................................. 3.9-3 TR 3.9.4              Crane Travel - Spent Fuel Storage Pool Building............................................. 3.9-5 5.0            ADMINISTRATIVE CONTROLS ..................................................................................... 5.0-1 5.1                    Technical Requirements (TR) Control Program ............................................... 5.0-1 (continued)
Watts Bar-Unit 1                                                    ii Technical Requirements                                                                                                                        Revision 62
 
TABLE OF CONTENTS (continued)
BASES B 3.0          TECHNICAL REQUIREMENTS (TR) AND TECHNICAL SURVEILLANCE REQUIREMENTS (TSR)
APPLICABILITY ................................................................................................. B 3.0-1 B 3.1          REACTIVITY CONTROL SYSTEMS .............................................................................. B 3.1-1 B 3.1.1        Boration Systems Flow Paths, Shutdown ....................................................................... B 3.1-1 B 3.1.2        Boration Systems Flow Paths, Operating ....................................................................... B 3.1-5 B 3.1.3        Charging Pump, Shutdown. ............................................................................................. B 3.1-9 B 3.1.4        Charging Pumps, Operating ............................................................................................ B 3.1-11 B 3.1.5        Borated Water Sources, Shutdown ................................................................................. B 3.1-14 B 3.1.6        Borated Water Sources, Operating ................................................................................. B 3.1-18 B 3.1.7        Position Indication System, Shutdown ............................................................................ B 3.1-23 B 3.3          INSTRUMENTATION ...................................................................................................... B 3.3-1 B 3.3.1        Reactor Trip System (RTS) Instrumentation................................................................... B 3.3-1 B 3.3.2        Engineered Safety Features Actuation System (ESFAS) Instrumentation ..................................................... B 3.3-4 B 3.3.3        Movable Incore Detectors. ............................................................................................... B 3.3-7 B 3.3.4        Seismic Instrumentation ................................................................................................... B 3.3-10 B 3.3.5        Turbine Overspeed Protection......................................................................................... B 3.3-14 B 3.3.6        Loose-Part Detection System .......................................................................................... B 3.3-18 B.3.3.7        Plant Calorimetric Measurement ..................................................................................... B 3.3-21 B 3.3.8        Hydrogen Monitors ........................................................................................................... B3.3-25 B 3.3.9        Power Distribution Monitoring System (PDMS) .............................................................. B3.3-30 B 3.4          REACTOR COOLANT SYSTEM (RCS)......................... ............................................... B 3.4-1 B 3.4.1        Safety Valves, Shutdown ................................................................................................. B 3.4-1 B 3.4.2        Pressurizer Temperature Limits ...................................................................................... B 3.4-4 B 3.4.3        RCS Vents ........................................................................................................................ B 3.4-7 B 3.4.4        Chemistry .......................................................................................................................... B 3.4-10 B 3.4.5        Piping System Structural Integrity ................................................................................... B 3.4-14 B 3.6          CONTAINMENT SYSTEMS ............................................................................................ B 3.6-1 B 3.6.1        Ice Bed Temperature Monitoring System............. .......................................................... B 3.6-1 B 3.6.2        Inlet Door Position Monitoring System ............................................................................ B 3.6-6 B 3.6.3        Lower Compartment Cooling (LCC) System .................................................................. B 3.6-10 B 3.7          PLANT SYSTEMS ........................................................................................................... B 3.7-1 B 3.7.1        Steam Generator Pressure/Temperature Limitations .................................................... B 3.7-1 B 3.7.2        Flood Protection Plan ....................................................................................................... B 3.7-4 B 3.7.3        DELETED ......................................................................................................................... B 3.7-12 B 3.7.4        Sealed Source Contamination ......................................................................................... B 3.7-18 B 3.7.5        Area Temperature Monitoring.......................................................................................... B 3.7-22 B 3.8          ELECTRICAL POWER SYSTEMS ................................................................................. B 3.8-1 B 3.8.1        Isolation Devices .............................................................................................................. B 3.8-1 B 3.8.2        Containment Penetration Conductor Overcurrent Protection Devices......................................................................... B 3.8-7 B 3.8.3        Motor-Operated Valves Thermal Overload Bypass Devices .................................................................................. B 3.8-15 B 3.8.4        Submerged Component Circuit Protection ..................................................................... B 3.8-19 (continued)
Watts Bar-Unit 1                                                        iii Technical Requirements                                                                                                                            Revision 62
 
TABLE OF CONTENTS (continued)
B 3.9          REFUELING OPERATIONS ........................................................................................... B 3.9-1 B 3.9.1        Deleted .............................................................................................................................. B 3.9-1 B 3.9.2        Communications............................................................................................................... B 3.9-3 B 3.9.3        Refueling Machine............................................................................................................ B 3.9-5 B 3.9.4        Crane Travel - Spent Fuel Storage Pool Building ....................................................................................................... B 3.9-8 (continued)
Watts Bar-Unit 1                                                          iv Technical Requirements                                                                                                                              Revision 53
 
LIST OF TABLES Table No.        Title                                                                                                                          Page 1.1-1            MODES ............................................................................................................... 1.1-6 3.3.1-1          Reactor Trip System Instrumentation Response Times.... .............................. 3.3-3 3.3.2-1          Engineered Safety Features Actuation System Response Times .................................................... 3.3-7 3.3.4-1          Seismic Monitoring Information ......................................................................... 3.3-17 3.7.3 3.7.3-5 ............. ............. ................................................................................................. DELETED 3.7.5-1          Area Temperature Monitoring...... ..................................................................... 3.7-29 3.8.3-1          Motor-Operated Valves Thermal Overload Devices Which Are Bypassed Under Accident Conditions .......................................... 3.8-12 3.8.4-1          Submerged Components With Automatic De-energization Under Accident Conditions .................................................................. 3.8-19 Watts Bar-Unit 1                                                                  v Technical Requirements                                                                                                                              Revision 62
 
LIST OF FIGURES Figure No.      Title                                                                            Page 3.1.6          Boric Acid Tank Limits Based on RWST Boron Concentration ....................... 3.1-12a 3.7.3-1        DELETED LIST OF MISCELLANEOUS REPORTS AND PROGRAMS Core Operating Limits Report Watts Bar-Unit 1                                    vi Technical Requirements                                                                                  Revision 62
 
LIST OF ACRONYMS Acronym        Title ABGTS          Auxiliary Building Gas Treatment System ACRP            Auxiliary Control Room Panel ASME            American Society of Mechanical Engineers AFD            Axial Flux Difference AFW            Auxiliary Feedwater System ARO            All Rods Out ARFS            Air Return Fan System ARV            Atmospheric Relief Valve BOC            Beginning of Cycle CCS            Component Cooling Water System CFR            Code of Federal Regulations COLR            Core Operating Limits Report CREVS          Control Room Emergency Ventilation System CSS            Containment Spray System CST            Condensate Storage Tank DNB            Departure from Nucleate Boiling ECCS            Emergency Core Cooling System EFPD            Effective Full-Power Days EGTS            Emergency Gas Treatment System EOC            End of Cycle ERCW            Essential Raw Cooling Water ESF            Engineered Safety Feature ESFAS          Engineered Safety Features Actuation System HEPA            High Efficiency Particulate Air HVAC            Heating, Ventilating, and Air-Conditioning LCC            Lower Compartment Cooler LCO            Limiting Condition For Operation MFIV            Main Feedwater Isolation Valve MFRV            Main Feedwater Regulation Valve MSIV            Main Steam Line Isolation Valve MSSV            Main Steam Safety Valve MTC            Moderator Temperature Coefficient NMS            Neutron Monitoring System ODCM            Offsite Dose Calculation Manual PCP            Process Control Program PDMS            Power Distribution Monitoring System PIV            Pressure Isolation Valve PORV            Power-Operated Relief Valve PTLR            Pressure and Temperature Limits Report QPTR            Quadrant Power Tilt Ratio RAOC            Relaxed Axial Offset Control RCCA            Rod Cluster Control Assembly RCP            Reactor Coolant Pump RCS            Reactor Coolant System RHR            Residual Heat Removal RTP            Rated Thermal Power RTS            Reactor Trip System RWST            Refueling Water Storage Tank SG              Steam Generator SI              Safety Injection SL              Safety Limit SR              Surveillance Requirement UHS            Ultimate Heat Sink Watts Bar-Unit 1                                      vii Technical Requirements                                    Revision 46
 
TECHNICAL REQUIREMENTS LIST OF EFFECTIVE PAGES Page      Revision            Page  Revision Number      Number              Number  Number i        56                3.1-4  69 ii        62                3.1-5  38 iii      62                3.1-6  51 iv        53                3.1-7    0 v        62                3.1-8    0 vi        62                3.1-9  37 vii        46                3.1-10    0 viii      72                3.1-11  33 ix        69                3.1-12    0 x        72              3.1-12a  42 xi        69                3.1-13  71 xii        22                3.3-1    0 xiii      37                3.3-2    0 xiv        47                3.3-3  34 xv        58                3.3-4  44 xvi        72                3.3-5    0 1.1-1        0                3.3-6    0 1.1-2        22                3.3-7  26 1.1-3        0                3.3-8  36 1.1-4        31                3.3-9  68 1.1-5        0                3.3-10    0 1.1-6        0                3.3-11  49 1.2-1        0                3.3-12  46 1.2-2        0                3.3-13    0 1.2-3        0                3.3-14  40 1.3-1        0                3.3-15  40 1.3-2        0                3.3-16    0 1.3-3        0                3.3-17  19 1.3-4        0                3.3-18  38 1.3-5        0                3.3-19  38 1.3-6        0                3.3-20  63 1.3-7        0                3.3-21    0 1.3-8        0                3.3-22  23 1.3-9        0                3.3-23  23 1.3-10        0                3.3-24  45 1.3-11        0                3.3-25  45 1.3-12        0                3.3-26  72 1.3-13        0                3.3-27  70 1.4-1        0                3.3-28  46 1.4-2        0                3.4-1    0 1.4-3        0                3.4-2    0 1.4-4        0                3.4-3    0 3.0-1        38                3.4-4    0 3.0-2        38                3.4-5    0 3.0-3        39                3.4-6    0 3.0-4        38                3.4-7    0 3.1-1        38                3.4-8    0 3.1-2        0                3.4-9    0 3.1-3        51 Watts Bar-Unit 1                      viii Technical Requirements                                      Revision 72
 
TECHNICAL REQUIREMENTS LIST OF EFFECTIVE PAGES Page      Revision            Page  Revision Number      Number              Number  Number 3.4-10        64                3.8-7    0 3.4-11        0                3.8-8    0 3.4-12        52                3.8-9  25 3.6-1        0                3.8-10    0 3.6-2        0                3.8-11  69 3.6-3        0                3.8-12    0 3.6-4        56                3.8-13    0 3.6-5        56                3.8-14  55 3.6-6        0                3.8-15  60 3.6-7        0                3.8-16  59 3.7-1        0                3.8-17    0 3.7-2        0                3.8-18  69 3.7-3        17                3.8-19  18 3.7-4        17                3.9-1  53 3.7-5        17                3.9-2    0 3.7-6        17                3.9-3  28 3.7-7        17                3.9-4  28 3.7-8        17                3.9-5    0 3.7-9        17                5.0-1  24 3.7-10        62 3.7-11        62 3.7-12        62 3.7-13        62 3.7-14        62 3.7-15        62 3.7-16        62 3.7-17        62 3.7-18        62 3.7-19        62 3.7-20        62 3.7-21        62 3.7-22        43 3.7-23        0 3.7-24        0 3.7-25        0 3.7-26        40 3.7-27        40 3.7-28        40 3.7-29        2 3.7-30        2 3.8-1        0 3.8-2        0 3.8-3        0 3.8-4        25 3.8-5        0 3.8-6        0 Watts Bar-Unit 1                      ix Technical Requirements                                      Revision 69
 
TECHNICAL REQUIREMENTS BASES LIST OF EFFECTIVE PAGES Page        Revision            Page  Revision Number        Number            Number  Number B 3.0-1        0              B 3.3-13  19 B 3.0-2        0              B 3.3-14  66 B 3.0-3        0              B 3.3-15  38 B 3.0-4        38              B 3.3-16    6 B 3.0-5        38              B 3.3-17  38 B 3.0-6        0              B 3.3-18  63 B 3.0-7        0              B 3.3-19  63 B 3.0-8        0              B 3.3-20  63 B 3.0-9        50              B 3.3-21  23 B 3.0-10        39              B 3.3-22  23 B 3.0-11        39              B 3.3-23  23 B 3.0-12        38              B 3.3-24  23 B 3.1-1        0              B 3.3-25  45 B 3.1-2        0              B 3.3-26  45 B 3.1-3        38              B 3.3-27  45 B 3.1-4        0              B 3.3-28  45 B 3.1-5        51              B 3.3-29  45 B 3.1-6        0              B 3.3-30  54 B 3.1-7        69              B 3.3-31  72 B 3.1-8        20              B 3.3-32  46 B 3.1-9        38              B 3.3-33  46 B 3.1-10        41              B 3.3-34  70 B 3.1-11        51                B 3.4-1    0 B 3.1-12        0                B 3.4-2    0 B 3.1-13        41                B 3.4-3    0 B 3.1-14        0                B 3.4-4    0 B 3.1-15        20                B 3.4-5    0 B 3.1-16        37                B 3.4-6    0 B 3.1-17        37                B 3.4-7    0 B 3.1-18        0                B 3.4-8    0 B 3.1-19        0                B 3.4-9    0 B 3.1-20        20              B 3.4-10    0 B 3.1-21        27              B 3.4-11    0 B 3.1-22        37              B 3.4-12    0 B 3.1-23        0              B 3.4-13    0 B 3.1-24        71              B 3.4-14  64 B 3.1-25        71              B 3.4-15  38 B 3.3-1        0              B 3.4-16  52 B 3.3-2        0                B 3.6-1    0 B 3.3-3        0                B 3.6-2  20 B 3.3-4        22                B 3.6-3  20 B 3.3-5        22                B 3.6-4    0 B 3.3-6        0                B 3.6-5    0 B 3.3-7        46                B 3.6-6  65 B 3.3-8        46                B 3.6-7  56 B 3.3-9        46                B 3.6-8  61 B 3.3-10        19                B 3.6-9    0 B 3.3-11        40              B 3.6-10    0 B 3.3-12        40              B 3.6-11    0 Watts Bar-Unit 1                      x Technical Requirements                                        Revision 72
 
TECHNICAL REQUIREMENTS BASES LIST OF EFFECTIVE PAGES Page        Revision            Page  Revision Number        Number            Number  Number B 3.6-12        0              B 3.8-22  69 B 3.7-1        36                B 3.9-1  53 B 3.7-2        38                B 3.9-2  53 B 3.7-3        36                B 3.9-3    0 B 3.7-4        57                B 3.9-4    0 B 3.7-5        17                B 3.9-5  28 B 3.7-6        17                B 3.9-6    0 B 3.7-7        17                B 3.9-7  28 B 3.7-8        17                B 3.9-8    0 B 3.7-9        17                B 3.9-9    0 B 3.7-10        17 B 3.7-11        17 B 3.7-12        62 B 3.7-13        62 B 3.7-14        62 B 3.7-15        62 B 3.7-16        62 B 3.7-17        62 B 3.7-18        0 B 3.7-19        43 B 3.7-20        0 B 3.7-21        0 B 3.7-22        0 B 3.7-23        20 B 3.7-24        40 B 3.7-25        40 B 3.8-1        0 B 3.8-2        0 B 3.8-3        0 B 3.8-4        0 B 3.8-5        0 B 3.8-6        25 B 3.8-7        25 B 3.8-8        0 B 3.8-9        0 B 3.8-10        0 B 3.8-11        0 B 3.8-12        0 B 3.8-13        25 B 3.8-14        25 B 3.8-15        0 B 3.8-16        0 B 3.8-17        69 B 3.8-18        0 B 3.8-19        0 B 3.8-20        0 B 3.8-21        0 Watts Bar-Unit 1                      xi Technical Requirements                                        Revision 69
 
TECHNICAL REQUIREMENTS MANUAL LIST OF EFFECTIVE PAGES - REVISION LISTING Revisions                    Issued                                SUBJECT Revision 0                  09-30-95  Initial Issue Revision 1                  12-06-95  Submerged Component Circuit Protection Revision 2                  01-04-96  Area Temperature Monitoring - Change in MSSV Limit Revision 3                  02-28-96  Turbine Driven AFW Pump Suction Requirement Revision 4                  08-18-97  Time-frame for Snubber Visual Exams Revision 5                  08-29-97  Performance of Snubber Functional Tests at Power Revision 6                  09-08-97  Revised Actions for Turbine Overspeed Protection Revision 7                  09-12-97  Change OPT/OTT Response Time Revision 8                  09-22-97  Clarification of Surveillance Frequency for Position Indication System Revision 9                  10-10-97  Revised Boron Concentration for Borated Water Sources Revision 10                  12-17-98  ICS Inlet Door Position Monitoring - Channel Check Revision 11                  01-08-99  Computer-Based Analysis for Loose Parts Monitoring Revision 12                  01-15-99  Removal of Process Control Program from TRM Revision 13                  03-30-99  Deletion of Power Range Neutron Flux High Negative Rate Reactor Trip Function Revision 14                  04-07-99  Submerged Component Circuit Protection Revision 15                  04-07-99  Submerged Component Circuit Protection Revision 16                  04-13-99  Submerged Component Circuit Protection Revision 17                  05-25-99  Flood Protection Plan Revision 18                  08-03-99  Submerged Component Circuit Protection Revision 19                  10-12-99  Upgrade Seismic Monitoring Instruments Revision 20                  03/13/00  Added Notes to Address Instrument Error for Various Parameters Revision 21                  04/13/00  COLR, Cycle 3, Rev 2 Revision 22                  07/07/00  Elimination of Response Time Testing (continued)
Watts Bar-Unit 1                                xii Technical Requirements                                                                      Revision 22
 
TECHNICAL REQUIREMENTS MANUAL LIST OF EFFECTIVE PAGES - REVISION LISTING Revisions                    Issued                                SUBJECT Revision 23                01/22/01  Plant Calorimetric (LEFM)
Revision 24                03/19/01  TRM Change Control Program per 50.59 Rule Revision 25                05/15/01  Change in Preventive Maintenance Frequency for Molded Case Circuit Breakers Revision 26                05/29/01  Change CVI Response Time from 5 to 6 Seconds Revision 27                01/31/02  Change pH value in the borated water sources due to TS change for ice weight reduction Revision 28                02/05/02  Refueling machine upgrade under DCN D-50991-A Revision 29                02/26/02  Added an additional action to TR 3.7.3 to perform an engineering evaluation of inoperable snubbers impact on the operability of a supported system.
Revision 30                06/05/02  Updated TR 3.3.5.1 to reflect implementation of the TIPTOP program in a Technical Instruction (TI).
Revision 31                10/31/02  Correct RTP to 3459 MWt (PER 02-9519-000)
Revision 32                09/17/03  Editorial correction to Bases for TSR 3.1.5.3.
Revision 33                10/14/03  Updated TRs 3.1.5 and 3.1.6 and their respective bases to incorporate boron concentration changes in accordance with change packages WBN-TS-02-14 and WBN-TS-03-017.
Revision 34                05/14/04  Revised Item 5, Source Range, Neutron Flux function of Table 3.3.1-1 to provide an acceptable response time of less than or equal 0.5 seconds. (Reference TS Amendment 52.)
Revision 35                04/06/05  Revised Table 3.3.2-1, Engineered Safety Features Actuation systems Response Times, to revise containment spray response time and to add an asterisk note to notation 13 of the table via Change Package WBN-TS-04-16.
Revision 36                09/25/06  Revised the response time for Containment Spray in Table 3.3.2-1 and the RTNDT values in the Bases for TR 3.7.1. Both changes result from the replacement of the steam generators.
Revision 37                11/08/06  Revised TR 3.1.5 and 3.1.6 and the Bases for these TRs to update the boron concentration limits of the RWST and the BAT.
(continued)
Watts Bar-Unit 1                              xiii Technical Requirements                                                                    Revision 37
 
TECHNICAL REQUIREMENTS MANUAL LIST OF EFFECTIVE PAGES REVISION LISTING Revisions                Issued                                  SUBJECT Revision 38            11/29/06    Updated the TRM to be consistent with Tech Spec Amendment 55. TRM Revision 38 modified the requirements for mode change limitations in TR 3.0.4 and TSR 3.0.4 by incorporating changes similar to those outlined in TSTF-359, Revision 9. (TS-06-24)
Revision 39            04/16/07  Updated the TRM to be consistent with Tech Spec Amendment 42.
TRM Revision 39 modified the requirements of TSR 3.0.3 by incorporating changes similar to those outlined in TSTF-358.
(TS-07-03)
Revision 40            05/24/07  Updated the TRM and Bases to remove the various requirements for the submittal of reports to the NRC. (TS-07-06)
Revision 41            05/25/07  Revision 41 updates the Bases of TR 3.1.3, 3.1.4 and 3.4.5 to be consistent with Technical Specification Amendment 66. This amendment replaces the references to Section XI of the ASME Boiler and Pressure Vessel Code with the ASME Operation and Maintenance Code for Inservice Testing (IST) activities and removes reference to applicable supports from the IST program.
Revision 42            03/20/2008  Revision 42 updates Figure 3.1.6 to remove the 240 TPBAR Limit.
Revision 43            07/17/2008  Revision 43 removes a reporting requirement from TR 3.7.4, Sealed Source Contamination. The revision also updates the Bases for TR 3.7.4.
Revision 44            10/10/2008  Revision 44 updates Table 3.3.1-1 to be consistent with the changes approved by NRC as Tech Spec Amendment 68.
Revision 45            02/23/2009  Added TR 3.3.8, Hydrogen Monitors, and the Bases for TR 3.3.8.
This change is based on Technical Specification (TS) Amendment 72 which removed the Hydrogen Monitors (Function 13 of LCO 3.3.3) from the TS.
Revision 46            09/20/2010  Revision 46 implements changes from License Amendment 82 (Technical Specification (TS) Bases Revsion 104) for the approved BEACON-TSM application of the Power Distribution Monitoring System (PDMS).
Revision 47            10/08/2010  Revision 47 changes are in response to PER 215552 which requested clarification be added to the TRM regarding supported system operability when a snubber is declared inoperable or removed from service.
(continued)
Watts Bar-Unit 1                            xiv Technical Requirements                                                                    Revision 47
 
TECHNICAL REQUIREMENTS MANUAL LIST OF EFFECTIVE PAGES REVISION LISTING Revisions                Issued                                  SUBJECT Revision 48            04/12/2011  CANCELLED Revision 49            05/24/2011  Revision 49 updated Note 14 of Table 3.3.2-1 to clarify that the referenced time is only for partial transfer of the ECCS pumps from the VCT to the RWST.
Revision 50            12/12/2011  Clarifies the acceptability of periodically using a portion of the 25%
grace period in TSR 3.0.2 to facilitate 13 week maintenance work schedules.
Revision 51            08/09/2013  Adds a note to TR 3.1.2 and TR 3.1.4 to permit securing one charging pump in order to supporting transition into or from the Applicability of Technical Specification 3.4.12 (PER 593365).
Revision 52            08/30/2013  Clarifies that TR 3.4.5, Piping System Structural Integrity, applies to all ASME Code Class 1, 2, and 3 piping systems, and is not limited to reactor coolant system piping.
Revision 53            12/12/2013  Technical Specification Amendment 92 added Limiting Condition for Operation (LCO) 3.9.10, Decay Time, which was redundant to Technical Requirement (TR) 3.9.1, Decay Time. Revision 53 removes TR 3.9.1 from the Technical Requirements Manual (TRM) and the TRM Bases.
Revision 54            01/23/2014  TRM which updates Technical Requirement (TR) 3.3.9, Power Distribution Monitoring System, to reflect the Addendum to WCAP 12472-P-A.
Revision 55            01/14/2015  Provided in the attachment is TRM Revision 55 which revises TRM Table 3.8.3-1 pages 3 and 5, Motor-Operated Valves Thermal Overload Devices which are bypassed under accident conditions.
This revision results in the valves requiring their Thermal Overload Bypasses to be operable.
Revision 56            04/30/2015  This revision restructures TR 3.6.2 CONDITIONS, REQUIRED ACTIONS, and COMPLETION TIME(s) to address two distinct cases of system inoperability. TRM BASES B 3.6.2 was also revised to coincide with the changes described above and to include additional detail regarding use of indirect means for performing channel checks Revision 57            05/07/2015  This revision changes the elevation of the Mean Sea Level by submergence during floods vary from 714.5 ft to 739.2 ft in TRM Bases B 3.7.2, Flood Protection Plan.
Revision 58            05/19/2015  This revision is an administrative change in TRM Bases 3.4.5 background information.
(continued)
Watts Bar-Unit 1                            xv Technical Requirements                                                                      Revision 58
 
TECHNICAL REQUIREMENTS MANUAL LIST OF EFFECTIVE PAGES REVISION LISTING Revisions                  Issued                                  SUBJECT Revision 59            10/13/2015  This revision adds the Unit 1 and Unit 2 FCV-67-0066 and FCV-67-0067 valves to TRM Table 3.8.3-1.
Revision 60            06/01/2016  This revision is to add 2-FCV-70-153 valve to TRM Table 3.8.3-1 Sheet 4 of 5.
Revision 61            02/21/2017  Revises TRM Bases 3.6.2 Inlet Door Position Monitoring System actions.
Revision 62            03/31/2017  This revision deletes TRM and TRM Bases section 3.7.3, Snubbers via the License Amendment 111.
Revision 63              5/17/2017  Revises the obsolete analog system that was limited to monitoring 1 sensor for each RCS collection point.
Revision 64              8/22/17    Clarified ASME Code Class in the section description in Section 3.4.5, Piping System Structural Integrity.
Revision 65                4/6/18    Revised TRM Bases Section 3.6.2, to more closely match information provided in the UFSAR. The Bases as written limits credit for the lower inlet door main panel annunciator as part of the Inlet Door Position Monitoring System.
Revision 66 (Amendment    10/11/18  Revises TRM Bases Section 3.3.5, Turbine Overspeed 119)                                Protection, to change the background information.
Revision 67                8/14/19  Revises TRM Section 3.3.9, PDMS and TRM Bases Section B3.3.9, PDMS to align the EFPD with the NRC SER.
Revision 68 (Amendment    12/17/19  Revises TRM Table 3.3.2-1 Page 3, to add Unbalanced voltage to 131)                                item 14.
Revision 69              4/21/20    Revises TRM to change TSRs 3.1.2.3, 3.8.3.1, and 3.8.4.2 due to the frequency of SR 3.6.3.6 being changed to 36 months.
Revision 70              9/24/20    Revises TRM and Bases to change the frequency of the PDMS calibration with the CET chess knight move pattern not satisfied from 30 to 31 EPFD.
Revision 71              11/4/20    Revises TRM and Bases TR 3.1.7 involving the requirements for the Position Indication System during Shutdown conditions.
Revision 72              1/11/22    Revises TRM and Bases to correct TRM 3.3.9 PDMS regarding the great than or equal too and greater than because a process parameter is never exactly equal to a value.
Watts Bar-Unit 1                              xvi Technical Requirements                                                                      Revision 72
 
ENCLOSURE 4 WBN UNIT 1 TECHNICAL REQUIREMENTS MANUAL CHANGED PAGES (10 pages)
 
TECHNICAL REQUIREMENTS LIST OF EFFECTIVE PAGES Page      Revision            Page  Revision Number      Number              Number  Number i        56                3.1-4  69 ii        62                3.1-5  38 iii        62                3.1-6    51 iv        53                3.1-7    0 v        62                3.1-8    0 vi        62                3.1-9  37 vii        46                3.1-10    0 viii      72                3.1-11  33 ix        69                3.1-12    0 x        72              3.1-12a  42 xi        69                3.1-13  71 xii        22                3.3-1    0 xiii      37                3.3-2    0 xiv        47                3.3-3  34 xv        58                3.3-4  44 xvi        72                3.3-5    0 1.1-1        0                3.3-6    0 1.1-2        22                3.3-7  26 1.1-3        0                3.3-8  36 1.1-4        31                3.3-9  68 1.1-5        0                3.3-10    0 1.1-6        0                3.3-11  49 1.2-1        0                3.3-12  46 1.2-2        0                3.3-13    0 1.2-3        0                3.3-14  40 1.3-1        0                3.3-15  40 1.3-2        0                3.3-16    0 1.3-3        0                3.3-17  19 1.3-4        0                3.3-18  38 1.3-5        0                3.3-19  38 1.3-6        0                3.3-20  63 1.3-7        0                3.3-21    0 1.3-8        0                3.3-22  23 1.3-9        0                3.3-23  23 1.3-10        0                3.3-24  45 1.3-11        0                3.3-25  45 1.3-12        0                3.3-26  72 1.3-13        0                3.3-27  70 1.4-1        0                3.3-28  46 1.4-2        0                3.4-1    0 1.4-3        0                3.4-2    0 1.4-4        0                3.4-3    0 3.0-1        38                3.4-4    0 3.0-2        38                3.4-5    0 3.0-3        39                3.4-6    0 3.0-4        38                3.4-7    0 3.1-1        38                3.4-8    0 3.1-2        0                3.4-9    0 3.1-3        51 Watts Bar-Unit 1                      viii Technical Requirements                                      Revision 72
 
TECHNICAL REQUIREMENTS BASES LIST OF EFFECTIVE PAGES Page        Revision            Page  Revision Number        Number            Number  Number B 3.0-1        0              B 3.3-13  19 B 3.0-2        0              B 3.3-14  66 B 3.0-3        0              B 3.3-15  38 B 3.0-4        38              B 3.3-16    6 B 3.0-5        38              B 3.3-17  38 B 3.0-6        0              B 3.3-18  63 B 3.0-7        0              B 3.3-19  63 B 3.0-8        0              B 3.3-20  63 B 3.0-9        50              B 3.3-21  23 B 3.0-10        39              B 3.3-22  23 B 3.0-11        39              B 3.3-23  23 B 3.0-12        38              B 3.3-24  23 B 3.1-1        0              B 3.3-25  45 B 3.1-2        0              B 3.3-26  45 B 3.1-3        38              B 3.3-27  45 B 3.1-4        0              B 3.3-28  45 B 3.1-5        51              B 3.3-29  45 B 3.1-6        0              B 3.3-30  54 B 3.1-7        69              B 3.3-31  72 B 3.1-8        20              B 3.3-32  46 B 3.1-9        38              B 3.3-33  46 B 3.1-10        41              B 3.3-34  70 B 3.1-11        51                B 3.4-1    0 B 3.1-12        0                B 3.4-2    0 B 3.1-13        41                B 3.4-3    0 B 3.1-14        0                B 3.4-4    0 B 3.1-15        20                B 3.4-5    0 B 3.1-16        37                B 3.4-6    0 B 3.1-17        37                B 3.4-7    0 B 3.1-18        0                B 3.4-8    0 B 3.1-19        0                B 3.4-9    0 B 3.1-20        20              B 3.4-10    0 B 3.1-21        27              B 3.4-11    0 B 3.1-22        37              B 3.4-12    0 B 3.1-23        0              B 3.4-13    0 B 3.1-24        71              B 3.4-14  64 B 3.1-25        71              B 3.4-15  38 B 3.3-1        0              B 3.4-16  52 B 3.3-2        0                B 3.6-1    0 B 3.3-3        0                B 3.6-2  20 B 3.3-4        22                B 3.6-3  20 B 3.3-5        22                B 3.6-4    0 B 3.3-6        0                B 3.6-5    0 B 3.3-7        46                B 3.6-6  65 B 3.3-8        46                B 3.6-7  56 B 3.3-9        46                B 3.6-8  61 B 3.3-10        19                B 3.6-9    0 B 3.3-11        40              B 3.6-10    0 B 3.3-12        40              B 3.6-11    0 Watts Bar-Unit 1                      x Technical Requirements                                        Revision 72
 
TECHNICAL REQUIREMENTS MANUAL LIST OF EFFECTIVE PAGES REVISION LISTING Revisions                  Issued                                  SUBJECT Revision 59            10/13/2015  This revision adds the Unit 1 and Unit 2 FCV-67-0066 and FCV-67-0067 valves to TRM Table 3.8.3-1.
Revision 60            06/01/2016  This revision is to add 2-FCV-70-153 valve to TRM Table 3.8.3-1 Sheet 4 of 5.
Revision 61            02/21/2017  Revises TRM Bases 3.6.2 Inlet Door Position Monitoring System actions.
Revision 62            03/31/2017  This revision deletes TRM and TRM Bases section 3.7.3, Snubbers via the License Amendment 111.
Revision 63              5/17/2017  Revises the obsolete analog system that was limited to monitoring 1 sensor for each RCS collection point.
Revision 64              8/22/17    Clarified ASME Code Class in the section description in Section 3.4.5, Piping System Structural Integrity.
Revision 65                4/6/18    Revised TRM Bases Section 3.6.2, to more closely match information provided in the UFSAR. The Bases as written limits credit for the lower inlet door main panel annunciator as part of the Inlet Door Position Monitoring System.
Revision 66 (Amendment    10/11/18  Revises TRM Bases Section 3.3.5, Turbine Overspeed 119)                                Protection, to change the background information.
Revision 67                8/14/19  Revises TRM Section 3.3.9, PDMS and TRM Bases Section B3.3.9, PDMS to align the EFPD with the NRC SER.
Revision 68 (Amendment    12/17/19  Revises TRM Table 3.3.2-1 Page 3, to add Unbalanced voltage to 131)                                item 14.
Revision 69              4/21/20    Revises TRM to change TSRs 3.1.2.3, 3.8.3.1, and 3.8.4.2 due to the frequency of SR 3.6.3.6 being changed to 36 months.
Revision 70              9/24/20    Revises TRM and Bases to change the frequency of the PDMS calibration with the CET chess knight move pattern not satisfied from 30 to 31 EPFD.
Revision 71              11/4/20    Revises TRM and Bases TR 3.1.7 involving the requirements for the Position Indication System during Shutdown conditions.
Revision 72              1/11/22    Revises TRM and Bases to correct TRM 3.3.9 PDMS regarding the great than or equal too and greater than because a process parameter is never exactly equal to a value.
Watts Bar-Unit 1                              xvi Technical Requirements                                                                      Revision 72
 
Position Indication System, Shutdown TR 3.1.7 TR 3.1 REACTIVITY CONTROL SYSTEMS TR 3.1.7 Position Indication System, Shutdown TR 3.1.7                The group demand position indicators shall be OPERABLE and capable of determining within +/- 12 steps the demand position for each shutdown or control rod that is not fully inserted.
APPLICABILITY:          MODES 3, 4, and 5, when the reactor trip breakers are closed.
ACTIONS CONDITION                              REQUIRED ACTION                      COMPLETION TIME A.      One or more group              A.1      Restore required group                15 minutes demand position                          demand position indicators                                indicator inoperable.                              OPERABILITY.
B. Required Action and              B.1      Initiate action to fully insert all  Immediately associated completion Time                  rods.
of Condition A not met.
OR B.2      Open reactor trip breakers.            Immediately TECHNICAL SURVEILLANCE REQUIREMENTS SURVEILLANCE                                                  FREQUENCY TSR 3.1.7.1            Determine that each group demand                                Within 4 hours after position indicator is OPERABLE by                              closing the reactor movement of the associated                                      trip breakers if not shutdown or control rod 10 steps in                            completed within any one direction.                                              previous 31 days.
AND 31 days thereafter Watts Bar-Unit 1                                    3.1-13                                      Revision 8, 71 Technical Requirements
 
PDMS TR 3.3.9 TR 3.3 INSTRUMENTATION TR 3.3.9 Power Distribution Monitoring System (PDMS)
TR 3.3.9              The PDMS shall be OPERABLE with:
: a. THERMAL POWER > 25% RTP, and
: b. The required channel inputs from the plant computer for each function defined in Table 3.3.9-1.
APPLICABILITY:        When the PDMS is used for:
: a. Calibration of the Excore Neutron Flux Detection System, or
: b. Monitoring the QUADRANT POWER TILT RATIO, or
: c. Measurement of FNH and FQ(Z), or
: d. Verifying the position of a rod with inoperable position indicators.
ACTIONS CONDITION                        REQUIRED ACTION                    COMPLETION TIME A. PDMS inoperable.              A.1    -----------NOTE-------------
TR 3.0.3 is not applicable.
Restore the inoperable            Prior to using the system system to OPERABLE                for incore power status.                          distribution measurement purposes.
Watts Bar-Unit 1                                  3.3-26                                  Revision 46, 72 Technical Requirements
 
PDMS TR 3.3.9 TECHNICAL SURVEILLANCE REQUIREMENTS SURVEILLANCE                                        FREQUENCY TSR 3.3.9.1    Perform CHANNEL CHECK for each required            24 hours instrumentation channel specified in Table 3.3.9-1.
TSR 3.3.9.2    Verify by administrative means that the            24 hours surveillance requirements for each required channel specified in Table 3.3.9-1 are satisfied.
TSR 3.3.9.3    Perform PDMS calibration.                          Once after each refueling AND
                                                                    ---------NOTE---------
Not required to be performed until 31 Effective Full Power Days (EFPD) after the Core Exit Thermocouple (CET) chess knight move pattern not satisfied.
31 EFPD thereafter with the CET chess knight move pattern not satisfied AND 180 EFPD thereafter with the CET chess knight move pattern satisfied Watts Bar-Unit 1                                  3.3-27                            Revision 46, 67, 70 Technical Requirements
 
Position Indication System, Shutdown B 3.1.7 BASES (continued)
APPLICABLE            The rod Position Indication System is a system SAFETY ANALYSES      which provides information to the operator which could be used to initiate operator action. However, no DBA or transient assumes operator action to manually trip the reactor, or to take some alternative action if an automatic reactor trip does not occur (Ref. 2). Hence, the shutdown and control rods, including position indication, are not assumed to be OPERABLE to mitigate the consequences of a DBA or transient during shutdown conditions. Positive reactivity addition due to withdrawal of control rods is compensated for by boron concentration.
TR                    TR 3.1.7 specifies that the group demand position indicators be OPERABLE and capable of determining within +/- 12 steps the demand position for each shutdown or control rod not fully inserted. For the control rod position indicators to be OPERABLE requires meeting the surveillance requirement of the TR. This requirement provides adequate assurance that control rod position indication during shutdown conditions and rod testing is accurate, and that design assumptions are not challenged. OPERABILITY of the required position indicators ensures that inoperable, misaligned, or mispositioned control rods can be detected.
APPLICABILITY        This TR covers only the requirements on Rod Position Indication during MODES 3, 4, and 5 with the reactor trip breakers closed. Rod Position Indication during MODES 1 and 2 are covered by Technical Specification 3.1.8. In MODE 6 and in MODES 3, 4, and 5 with trip breakers open or all rods fully inserted Rod Position Indication is not required to be OPERABLE. Rod Position Indication OPERABILITY is required only when rods are not fully inserted.
(continued)
Watts Bar-Unit 1                                B 3.1-24                                      Revision 71 Technical Requirements
 
Position Indication System, Shutdown B 3.1.7 BASES (continued)
ACTIONS              A.1 With one or more group demand position indicators inoperable, OPERABILITY of the affected group demand position indicators must be restored promptly.
The 15 minutes provides an acceptable time to evaluate whether the group demand position indicators represent the actual demand position of the affected rods and whether the affected rods are not fully inserted in an orderly manner without allowing the plant to remain in an unacceptable condition for an extended period of time.
The immediate Completion Times are consistent with the required time for actions to be pursued without delay and in a controlled manner.
B.1 and B.2 If OPERABILITY of the group demand position indicators is not restored within 15 minutes, the unit must be placed in a condition where the demand position indicators are not required. This is accomplished by fully inserting all rods or opening the reactor trip breakers immediately.
TECHNICAL            TSR 3.1.7.1 SURVEILLANCE REQUIREMENTS          Exercising rods at a Frequency of 31 days allows the operator to determine that all withdrawn rods, including the group step counter demand position indicator, continue to be OPERABLE. A movement of 10 steps is adequate to demonstrate motion and verify a corresponding step change in the group step counter demand position indicator. Four hours is provided to perform the first surveillance after closing the reactor trip breakers if the surveillance has not been performed within the previous 31 days. The 31-day Frequency takes into consideration other information available to the operator in the control room and the remote likelihood that rods would be withdrawn from fully inserted for extended periods of time during shutdown conditions.
REFERENCES            1.      Watts Bar FSAR, Section 4.2.3 "Reactivity Control System."
: 2.      WCAP-11618, "MERITS Program-Phase II, Task 5, Criteria Application,"
including Addendum 1 dated April, 1989.
Watts Bar-Unit 1                              B 3.1-25 Technical Requirements                                                                        Revision 8, 71
 
PDMS B 3.3.9 Bases (continued)
APPLICABLE            The PDMS is used for periodic measurement of the core power distribution to SAFETY                confirm operation within design limits and periodic calibration of the excore ANALYSES              detectors. This system does not initiate any automatic protection action. The PDMS is not assumed to be OPERABLE to mitigate the consequences of a DBA or transient (References 2 and 3).
TR                    TR 3.3.9 requires the PDMS to be OPERABLE with the specified number of instrument channel inputs from the plant computer for each function listed in Table 3.3.9-1. The PDMS is OPERABLE when the required channel inputs are available, the calibration data set is valid, and reactor power is > 25% RTP.
This TR ensures the OPERABILITY of the PDMS when required to monitor the power distribution within the core. The PDMS is used for periodic surveillance of the incore power distribution and calibration of the excore detectors. The surveillance of incore power distribution verifies that the peaking factors are within their design envelope (Reference 3). The peaking factor limits include measurement uncertainty which bounds the actual measurement uncertainty of an OPERABLE PDMS (Reference 1).
Maintaining the minimum number of instrumentation channel inputs ensures the uncertainty is bounded by the uncertainty methodology. Similarly, when THERMAL POWER is 25% RTP, then the accuracy of the adjustment provided by the CETs to the measured PDMS power distribution may not be bounded by the uncertainties documented in Reference 1.
APPLICABILITY          The PDMS must be OPERABLE when it is used for calibration of the Excore Neutron Flux Detection System, monitoring the QPTR, measurement of FNH and FQ(z), or verifying the position of a rod with inoperable position indicators.
ACTIONS                A.1 The Required Action A.1 has been modified by a Note stating that the provisions of TR 3.0.3 do not apply.
With THERMAL Power  25% RTP or with one or more required channel inputs inoperable or unavailable to the PDMS, the PDMS must not be used to obtain an incore power distribution measurement. Therefore, the Required Action A.1 prohibits the use of the inoperable system for the applicable monitoring or calibration functions.
(continued)
Watts Bar-Unit 1                                B 3.3-31                                    Revision 46, 72 Technical Requirements
 
PDMS B 3.3.9 BASES (continued)
TECHNICAL              The subsequent PDMS calibration frequency is 31 Effective Full Power Days SURVEILLANCE          (EFPD) when the CET chess knight move pattern is not satisfied. The CET REQUIREMENTS          chess knight move pattern is satisfied when every interior core location (fuel (continued)    assemblies not face adjacent to the core baffle) is no further than a chess knights move from an OPERABLE CETC. The 31 EFPD frequency calibration requirement is modified by a note that clarifies that subsequent PDMS calibration is not required to be performed until 31 EFPD after the CET chess knight move pattern is not satisfied.
The subsequent PDMS calibration frequency is 180 EFPD when the CET chess knight move pattern is satisfied. The CET chess knight move pattern provides coverage of all interior fuel assemblies (coverage of fuel assemblies with a face along the baffle is not required). Fuel assemblies which are within a chess knights move of an OPERABLE CET are covered.
REFERENCES            1.      WCAP-12472-P-A, BEACON Core Monitoring and Operations Support System, August 1994.
: 2.      10 CFR 50.46, Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors."
: 3.      WCAP-11618, MERITS Program-Phase II, Task 5, Criteria Application, including Addendum 1 dated April, 1989.
(continued)
Watts Bar-Unit 1                                  B 3.3-34                              Revision 46, 67, 70 Technical Requirements
 
ENCLOSURE 5 WBN UNIT 2 TECHNICAL SPECIFICATION BASES TABLE OF CONTENTS (26 pages)
 
TABLE OF CONTENTS TABLE OF CONTENTS .... i LIST OF TABLES  vi LIST OF FIGURES . vi LIST OF ACRONYMS  vii LIST OF EFFECTIVE PAGES .. x 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 3.0-11 B 3.1      REACTIVITY CONTROL SYSTEMS ...... B 3.1-1 B 3.1.1        SHUTDOWN MARGIN (SDM) - Tavg > 200&deg;F .......... B 3.1-1 B 3.1.2          SHUTDOWN MARGIN (SDM) - Tavg  200&deg;F  B 3.1-8 B 3.1.3        Core Reactivity ... B 3.1-12 B 3.1.4          Moderator Temperature Coefficient (MTC)  B 3.1-18 B 3.1.5        Rod Group Alignment Limits  B 3.1-25 B 3.1.6        Shutdown Bank Insertion Limits .. B 3.1-35 B 3.1.7        Control Bank Insertion Limits ...... B 3.1-40 B 3.1.8        Rod Position Indication . B 3.1-48 B 3.1.9        PHYSICS TESTS Exceptions  MODE 1 . B 3.1-57 B 3.1.10        PHYSICS TESTS Exceptions  MODE 2 . B 3.1-64 B 3.2      POWER DISTRIBUTION LIMITS . B 3.2-1 B 3.2.1          Heat Flux Hot Channel Factor (FQ(Z)) .... B 3.2-1 B 3.2.2          Nuclear Enthalpy Rise Hot Channel Factor (F ) ..... B 3.2-14 B 3.2.3          AXIAL FLUX DIFFERENCE (AFD) ............ B 3.2-21 B 3.2.4        QUADRANT POWER TILT RATIO (QPTR) .. B 3.2-26 Watts Bar - Unit 2                            i                            (continued)
 
TABLE OF CONTENTS B 3.3    INSTRUMENTATION .                                  B 3.3-1 B 3.3.1        Reactor Trip System (RTS) Instrumentation .              B 3.3-1 B 3.3.2        Engineered Safety Feature Actuation System (ESFAS)
Instrumentation .... B 3.3-64 B 3.3.3        Post Accident Monitoring (PAM) Instrumentation.            B 3.3-122 B 3.3.4        Remote Shutdown System ... B 3.3-138 B 3.3.5        Loss of Power (LOP) Diesel Generator (DG) Start Instrumentation ....                        B 3.3-144 B 3.3.6        Containment Vent Isolation Instrumentation ................ B 3.3-151 B 3.3.7        Control Room Emergency Ventilation System (CREVS)
Actuation Instrumentation ................          B 3.3-159 B 3.3.8        Auxiliary Building Gas Treatment System (ABGTS) Actuation Instrumentation .... B 3.3-165 B 3.4      REACTOR COOLANT SYSTEM (RCS) .. B 3.4-1 B 3.4.1          RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits ... B 3.4-1 B 3.4.2          RCS Minimum Temperature for Criticality . B 3.4-6 B 3.4.3          RCS Pressure and Temperature (P/T) Limits  B 3.4-9 B 3.4.4          RCS Loops - MODES 1 and 2 . B 3.4-16 B 3.4.5          RCS Loops - MODE 3 ... B 3.4-20 B 3.4.6          RCS Loops - MODE 4 ... B 3.4-25 B 3.4.7          RCS Loops - MODE 5, Loops Filled .. B 3.4-31 B 3.4.8          RCS Loops - MODE 5, Loops Not Filled  B 3.4-35 B 3.4.9          Pressurizer .. B 3.4-38 B 3.4.10        Pressurizer Safety Valves .... B 3.4-42 B 3.4.11        Pressurizer Power Operated Relief Valves (PORVs) .. B 3.4-46 B 3.4.12        Cold Overpressure Mitigation System (COMS) .... B 3.4-52 B 3.4.13        RCS Operational LEAKAGE .... B 3.4-65 B 3.4.14        RCS Pressure Isolation Valve (PIV) Leakage ... B 3.4-71 B 3.4.15        RCS Leakage Detection Instrumentation .. B 3.4-76 B 3.4.16        RCS Specific Activity ............... B 3.4-82 B 3.4.17        Steam Generator (SG) Tube Integrity  B 3.4-88 Watts Bar - Unit 2                            ii                                (continued)
 
TABLE OF CONTENTS B 3.5      EMERGENCY CORE COOLING SYSTEMS (ECCS) ................ B 3.5-1 B 3.5.1          Accumulators .. B 3.5-1 B 3.5.2          ECCS - Operating .. B 3.5-9 B 3.5.3          ECCS - Shutdown .. B 3.5-20 B 3.5.4          Refueling Water Storage Tank (RWST) .... B 3.5-24 B 3.5.5          Seal Injection Flow .... B 3.5-30 B 3.6      CONTAINMENT SYSTEMS .. B 3.6-1 B 3.6.1          Containment ... B 3.6-1 B 3.6.2          Containment Air Locks .. B 3.6-6 B 3.6.3          Containment Isolation Valves .. B 3.6-13 B 3.6.4          Containment Pressure .. B 3.6-27 B 3.6.5          Containment Air Temperature . B 3.6-30 B 3.6.6          Containment Spray System . B 3.6-34 B 3.6.7          RESERVED FOR FUTURE ADDITION .......... B 3.6-41 B 3.6.8          Hydrogen Mitigation System (HMS)  B 3.6-42 B 3.6.9          Emergency Gas Treatment System (EGTS) . B 3.6-48 B 3.6.10        Air Return System (ARS) .. B 3.6-54 B 3.6.11        Ice Bed . B 3.6-59 B 3.6.12        Ice Condenser Doors .... B 3.6-69 B 3.6.13        Divider Barrier Integrity . B 3.6-78 B 3.6.14        Containment Recirculation Drains .. B 3.6-83 B 3.6.15        Shield Building  B 3.6-87 B 3.7      PLANT SYSTEMS ................ B 3.7-1 B 3.7.1          Main Steam Safety Valves (MSSVs) ............. B 3.7-1 B 3.7.2          Main Steam Isolation Valves (MSIVs) .... B 3.7-8 B 3.7.3          Main Feedwater Isolation Valves (MFIVs) and Main Feedwater Regulation Valves (MFRVs) and Associated Bypass Valves .......................................................... B 3.7-13 B 3.7.4          Atmospheric Dump Valves (ADVs) . B 3.7-19 B 3.7.5          Auxiliary Feedwater (AFW) System  B 3.7-23 B 3.7.6          Condensate Storage Tank (CST)  B 3.7-32 (continued)
Watts Bar - Unit 2                            iii
 
TABLE OF CONTENTS B 3.7      PLANT SYSTEMS (continued)
B 3.7.7          Component Cooling System (CCS) .... B 3.7-36 B 3.7.8          Essential Raw Cooling Water (ERCW) System  B 3.7-42 B 3.7.9        Ultimate Heat Sink (UHS) . B 3.7-47 B 3.7.10        Control Room Emergency Ventilation System (CREVS) .... B 3.7-50 B 3.7.11        Control Room Emergency Air Temperature Control System (CREATCS) .. B 3.7-59 B 3.7.12        Auxiliary Building Gas Treatment System (ABGTS) .... B 3.7-63 B 3.7.13        Fuel Storage Pool Water Level  B 3.7-68 B 3.7.14        Secondary Specific Activity ............. B 3.7-71 B 3.7.15        Spent Fuel Pool Assembly Storage.                      B 3.7-74 B 3.7.16        Component Cooling System (CCS) - Shutdown.............        B 3.7-77 B 3.7.17        Essential Raw Cooling Water (ERCW) - Shutdown.              B 3.7-84 B 3.7.18        Fuel Storage Pool Boron Concentration                  B 3.7-89 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 3.8-38 B 3.8.3          Diesel Fuel Oil, Lube Oil, and Starting Air ................ B 3.8-43 B 3.8.4          DC Sources - Operating .................. B 3.8-53 B 3.8.5          DC Sources - Shutdown .................. B 3.8-68 B 3.8.6          Battery Parameters ...................... B 3.8-72 B 3.8.7          Inverters - Operating .............................. B 3.8-78 B 3.8.8          Inverters - Shutdown .............................. B 3.8-82 B 3.8.9          Distribution Systems - Operating ............... B 3.8-86 B 3.8.10        Distribution Systems - Shutdown .... B 3.8-95 B 3.9      REFUELING OPERATIONS . B 3.9-1 B 3.9.1          Boron Concentration .. B 3.9-1 B 3.9.2          Unborated Water Source Isolation Valves . B 3.9-5 B 3.9.3          Nuclear Instrumentation .... B 3.9-8 B 3.9.4        RESERVED FOR FUTURE ADDITION  B 3.9-11 B 3.9.5          Residual Heat Removal (RHR) and Coolant Circulation - High Water Level ... B 3.9-12 (continued)
Watts Bar - Unit 2                              iv                                Revision 41
 
TABLE OF CONTENTS B 3.9      REFUELING OPERATIONS (continued)
B 3.9.6          Residual Heat Removal (RHR) and Coolant Circulation - Low Water Level .... B 3.9-16 B 3.9.7          Refueling Cavity Water Level ... B 3.9-20 B 3.9.8        RESERVED FOR FUTURE ADDITION  B 3.9-23 B 3.9.9          Spent Fuel Pool Boron Concentration .... B 3.9-24 B 9.10          Decay Time ....................................................................................... B 3.9-26 Watts Bar - Unit 2                                      v
 
TABLE OF CONTENTS LIST OF TABLES TABLE NO                                        TITLE                                          PAGE B 3.3.4-1        Remote Shutdown System Instrumentation and Controls                      B 3.3-143a B 3.8.1-2        TS Action or Surveillance Requirements Contingency                        B3.8-37a Actions.
B 3.8.9-1        AC and DC Electrical Power Distribution Systems                          B 3.8-94 LIST OF FIGURES FIGURE NO.                                        TITLE                                          PAGE B 2.1.1-1        Reactor Core Safety Limits vs. Boundary of Protection ..................... B 2.0-6 B 3.1.7-1        Control Bank Insertion vs. Percent RTP ................... B 3.1-47 B 3.2.1-1        K(Z) - Normalized FQ(Z) as a Function of Core Height . B 3.2-13 B 3.2.3-1        TYPICAL AXIAL FLUX DIFFERENCE Acceptable Operation Limits as a Function of RATED THERMAL POWER ................................................. B 3.2-25 Watts Bar - Unit 2                            vi                                        Revision 19
 
LIST OF ACRONYMS (Page 1 of 3)
ACRONYM            TITLE ABGTS              Auxiliary Building Gas Treatment System ACRP              Auxiliary Control Room Panel AFD                Axial Flux Difference AFW                Auxiliary Feedwater System ARFS              Air Return Fan System ARO                All Rods Out ARV                Atmospheric Relief Valve ASME              American Society of Mechanical Engineers BOC                Beginning of Cycle CAOC              Constant Axial Offset Control CCS                Component Cooling Water System CFR                Code of Federal Regulations COLR              Core Operating Limits Report CREVS              Control Room Emergency Ventilation System CSS                Containment Spray System CST                Condensate Storage Tank DNB                Departure from Nucleate Boiling ECCS              Emergency Core Cooling System EFPD              Effective Full-Power Days EGTS              Emergency Gas Treatment System EOC                End of Cycle (continued)
Watts Bar - Unit 2                            vii
 
LIST OF ACRONYMS (Page 2 of 3)
ACRONYM            TITLE ERCW              Essential Raw Cooling Water ESF                Engineered Safety Feature ESFAS              Engineered Safety Features Actuation System HEPA              High Efficiency Particulate Air HVAC              Heating, Ventilating, and Air-Conditioning LCO                Limiting Condition For Operation MFIV              Main Feedwater Isolation Valve MFRV              Main Feedwater Regulation Valve MSIV              Main Steam Line Isolation Valve MSSV              Main Steam Safety Valve MTC                Moderator Temperature Coefficient NMS                Neutron Monitoring System ODCM              Offsite Dose Calculation Manual PCP                Process Control Program PIV                Pressure Isolation Valve PORV              Power-Operated Relief Valve PTLR              Pressure and Temperature Limits Report QPTR              Quadrant Power Tilt Ratio RAOC              Relaxed Axial Offset Control RCCA              Rod Cluster Control Assembly RCP                Reactor Coolant Pump RCS                Reactor Coolant System (continued)
Watts Bar - Unit 2                            viii
 
LIST OF ACRONYMS (Page 3 of 3)
ACRONYM            TITLE RHR                Residual Heat Removal RTP                Rated Thermal Power RTS                Reactor Trip System RWST              Refueling Water Storage Tank SG                Steam Generator SI                Safety Injection SL                Safety Limit SR                Surveillance Requirement UHS                Ultimate Heat Sink Watts Bar - Unit 2                            ix
 
TECHNICAL SPECIFICATIONS BASES LIST OF EFFECTIVE PAGES (continued)
PAGE              REVISION                    PAGE      REVISION NUMBER            NUMBER                    NUMBER      NUMBER i                    0                      B 3.0-3      30 ii                    0                      B 3.0-4      0 iii                  0                      B 3.0-5      0 iv                    41                      B 3.0-6      0 v                    0                      B 3.0-7      0 vi                    19                      B 3.0-8      0 vii                  0                      B 3.0-9      0 viii                  0                      B 3.0-10    0 ix                    0                      B 3.0-10a    7 x                    59                      B 3.0-10b    7 xi                    59                      B 3.0-10c    7 xii                  59                      B 3.0-11    0 xiii                  58                      B 3.0-12    38 xiv                  59                      B 3.0-13    30 xv                    57                      B 3.0-14    30 xvi                  57                      B 3.0-15    0 xvii                  57                      B 3.0-16    0 xviii                54                      B 3.1-1      0 xix                  55                      B 3.1-2      0 xx                    41                      B 3.1-3      0 xxi                  13                      B 3.1-4      0 xxii                  26                      B 3.1-5      0 xxiii                38                      B 3.1-6      34 xxiv                  46                      B 3.1-7      0 xxv                  57                      B 3.1-8      0 xxvi                  59                      B 3.1-9      0 B 2.0-1              0                      B 3.1-10    0 B 2.0-2              0                      B 3.1-11    34 B 2.0-3              0                      B 3.1-12    0 B 2.0-4              0                      B 3.1-13    0 B 2.0-5              0                      B 3.1-14    0 B 2.0-6              0                      B 3.1-15    0 B 2.0-7              0                      B 3.1-16    0 B 2.0-8              0                      B 3.1-17    34 B 2.0-9              0                      B 3.1-18    0 B 2.0-10              0                      B 3.1-19    0 B 3.0-1              7                      B 3.1-20    0 B 3.0-2              30                      B 3.1-21    0 Watts Bar - Unit 2                    x                      Revision 59
 
TECHNICAL SPECIFICATIONS BASES LIST OF EFFECTIVE PAGES (continued)
PAGE              REVISION                    PAGE      REVISION NUMBER            NUMBER                    NUMBER      NUMBER B 3.1-59    0 B 3.1-22              0                      B 3.1-60    0 B 3.1-23              0                      B 3.1-61    0 B 3.1-24              0                      B 3.1-62    34 B 3.1-25              16                      B 3.1-63    34 B 3.1-26              0                      B 3.1-64    0 B 3.1-27              0                      B 3.1-65    0 B 3.1-28              16                      B 3.1-66    0 B 3.1-29              16                      B 3.1-67    0 B 3.1-30              16                      B 3.1-68    34 B 3.1-31              16                      B 3.1-69    34 B 3.1-32              34                      B 3.1-70    0 B 3.1-33              34                      B 3.2-1      56 B 3.1-34              0                      B 3.2-2      57 B 3.1-35              0                      B 3.2-3      56 B 3.1-36              16                      B 3.2-4      56 B 3.1-37              16                      B 3.2-5      56 B 3.1-38              16                      B 3.2-6      56 B 3.1-39              34                      B 3.2-7      56 B 3.1-40              16                      B 3.2-8      56 B 3.1-41              16                      B 3.2-9      56 B 3.1-42              16                      B 3.2-10    56 B 3.1-43              16                      B 3.2-11    56 B 3.1-44              34                      B 3.2-12    56 B 3.1-45              34                      B 3.2-13    56 B 3.1-46              0                      B 3.2-13a    56 B 3.1-47              16                      B 3.2-13b    56 B 3.1-48              16                      B 3.2-14    34 B 3.1-49              16                      B 3.2-15    57 B 3.1-50              16                      B 3.2-16    57 B 3.1-51              16                      B 3.2-17    0 B 3.1-52              16                      B 3.2-18    0 B 3.1-53              16                      B 3.2-19    34 B 3.1-54              16                      B 3.2-20    0 B 3.1-55              0                      B 3.2-21    0 B 3.1-56              0                      B 3.2-22    0 B 3.1-57              0                      B 3.2-23    0 B 3.1-58              0                      B 3.2-24    34 Watts Bar - Unit 2                    xi                      Revision 59
 
TECHNICAL SPECIFICATIONS BASES LIST OF EFFECTIVE PAGES (continued)
PAGE              REVISION                    PAGE      REVISION NUMBER            NUMBER                    NUMBER      NUMBER B 3.2-25            0                        B 3.3-32    0 B 3.2-26            57                      B 3.3-33    0 B 3.2-27            0                        B 3.3-34    0 B 3.2-28            0                        B 3.3-35    0 B 3.2-29            0                        B 3.3-36    0 B 3.2-30            34                      B 3.3-37    0 B 3.2-31            0                        B 3.3-38    17 B 3.3-1              0                        B 3.3-39    17 B 3.3-2              0                        B 3.3-40    0 B 3.3-3              0                        B 3.3-41    0 B 3.3-4              0                        B 3.3-42    0 B 3.3-5              0                        B 3.3-43    0 B 3.3-6              0                        B 3.3-44    0 B 3.3-7              0                        B 3.3-45    0 B 3.3-8              0                        B 3.3-46    0 B 3.3-9              0                        B 3.3-47    0 B 3.3-10            0                        B 3.3-48    0 B 3.3-11            0                        B 3.3-49    0 B 3.3-12            0                        B 3.3-50    0 B 3.3-13            18                      B 3.3-51    34 B 3.3-14            0                        B 3.3-52    34 B 3.3-15            18                      B 3.3-53    34 B 3.3-16            0                        B 3.3-54    34 B 3.3-17            0                        B 3.3-55    34 B 3.3-18            0                        B 3.3-56    34 B 3.3-19            0                        B 3.3-57    34 B 3.3-20            0                        B 3.3-58    34 B 3.3-21            0                        B 3.3-59    34 B 3.3-22            0                        B 3.3-60    34 B 3.3-23            0                        B 3.3-61    0 B 3.3-24            0                        B 3.3-62    46 B 3.3-25            0                        B 3.3-63    46 B 3.3-26            0                        B 3.3-64    0 B 3.3-27            21                      B 3.3-65    0 B 3.3-28            0                        B 3.3-66    0 B 3.3-29            0                        B 3.3-67    0 B 3.3-30            0                        B 3.3-68    0 B 3.3-31            0                        B 3.3-69    0 Watts Bar - Unit 2                    xii                    Revision 59
 
TECHNICAL SPECIFICATIONS BASES LIST OF EFFECTIVE PAGES (continued)
PAGE              REVISION                    PAGE      REVISION NUMBER            NUMBER                    NUMBER      NUMBER B 3.3-70            0                        B 3.3-108    0 B 3.3-71            0                        B 3.3-109    0 B 3.3-72            57                      B 3.3-110    0 B 3.3-73            0                        B 3.3-111    58 B 3.3-74            0                        B 3.3-112    0 B 3.3-75            0                        B 3.3-113    36 B 3.3-76            0                        B 3.3-114    36 B 3.3-77            0                        B 3.3-115    34 B 3.3-78            0                        B 3.3-116    34 B 3.3-79            0                        B 3.3-117    34 B 3.3-80            0                        B 3.3-118    0 B 3.3-81            0                        B 3.3-119    57 B 3.3-82            0                        B 3.3-120    0 B 3.3-83            0                        B 3.3-121    58 B 3.3-84            0                        B 3.3-122    0 B 3.3-85            0                        B 3.3-123    0 B 3.3-86            0                        B 3.3-124    0 B 3.3-87            0                        B 3.3-125    0 B 3.3-88            0                        B 3.3-126    0 B 3.3-89            0                        B 3.3-127    0 B 3.3-90            0                        B 3.3-128    0 B 3.3-91            0                        B 3.3-129    0 B 3.3-92            0                        B 3.3-130    0 B 3.3-93            58                      B 3.3-131    0 B 3.3-94            58                      B 3.3-132    0 B 3.3-95            0                        B 3.3-133    0 B 3.3-96            0                        B 3.3-134    0 B 3.3-97            0                        B 3.3-135    34 B 3.3-98            0                        B 3.3-136    34 B 3.3-99            0                        B 3.3-137    0 B 3.3-100            0                        B 3.3-138    0 B 3.3-101            0                        B 3.3-139    19 B 3.3-102            0                        B 3.3-140    19 B 3.3-103            0                        B 3.3-141    34 B 3.3-104            0                        B 3.3-142    34 B 3.3-105            0                        B 3.3-143    0 B 3.3-106            0                        B3.3-143a    29 B 3.3-107            0                        B3.3-143b    20 Watts Bar - Unit 2                  xiii                    Revision 59
 
TECHNICAL SPECIFICATIONS BASES LIST OF EFFECTIVE PAGES (continued)
PAGE              REVISION                    PAGE      REVISION NUMBER            NUMBER                    NUMBER      NUMBER B 3.3-144            27                      B 3.4-14    0 B 3.3-145            0                        B 3.4-15    49 B 3.3-146            27                      B 3.4-16    0 B 3.3-147            27                      B 3.4-17    0 B 3.3-148            36                      B 3.4-18    0 B 3.3-149            34                      B 3.4-19    34 B 3.3-150            27                      B 3.4-20    0 B 3.3-151            0                        B 3.4-21    0 B 3.3-152            0                        B 3.4-22    0 B 3.3-153            0                        B 3.4-23    0 B 3.3-154            0                        B 3.4-24    59 B 3.3-155            34                      B 3.4-25    0 B 3.3-156            34                      B 3.4-26    0 B 3.3-157            34                      B 3.4-27    0 B 3.3-158            34                      B 3.4-28    0 B 3.3-159            0                        B 3.4-29    34 B 3.3-160            0                        B 3.4-30    59 B 3.3-161            0                        B 3.4-31    59 B 3.3-162            0                        B 3.4-32    59 B 3.3-163            34                      B 3.4-33    59 B 3.3-164            34                      B 3.4-34    59 B 3.3-165            0                        B 3.4-35    0 B 3.3-166            0                        B 3.4-36    0 B 3.3-167            0                        B 3.4-37    34 B 3.3-168            34                      B 3.4-38    0 B 3.4-1              0                        B 3.4-39    0 B 3.4-2              0                        B 3.4-40    0 B 3.4-3              0                        B 3.4-41    34 B 3.4-4              34                      B 3.4-42    0 B 3.4-5              34                      B 3.4-43    0 B 3.4-6              0                        B 3.4-44    0 B 3.4-7              0                        B 3.4-45    0 B 3.4-8              34                      B 3.4-46    0 B 3.4-9              0                        B 3.4-47    0 B 3.4-10            0                        B 3.4-48    0 B 3.4-11            49                      B 3.4-49    0 B 3.4-12            0                        B 3.4-50    0 B 3.4-13            0                        B 3.4-51    34 Watts Bar - Unit 2                  xiv                      Revision 59
 
TECHNICAL SPECIFICATIONS BASES LIST OF EFFECTIVE PAGES (continued)
PAGE              REVISION                    PAGE      REVISION NUMBER            NUMBER                    NUMBER      NUMBER B 3.4-89    39 B 3.4-52            0                        B 3.4-90    0 B 3.4-53            0                        B 3.4-91    0 B 3.4-54            0                        B 3.4-92    39 B 3.4-55            0                        B 3.4-93    39 B 3.4-56            0                        B 3.4-94    39 B 3.4-57            0                        B 3.5-1      0 B 3.4-58            0                        B 3.5-2      57 B 3.4-59            0                        B 3.5-3      57 B 3.4-60            0                        B 3.5-4      57 B 3.4-61            34                      B 3.5-5      0 B 3.4-62            34                      B 3.5-6      34 B 3.4-63            34                      B 3.5-7      34 B 3.4-64            51                      B 3.5-8      0 B 3.4-65            0                        B 3.5-9      0 B 3.4-66            0                        B 3.5-10    0 B 3.4-67            0                        B 3.5-11    57 B 3.4-68            0                        B 3.5-12    0 B 3.4-69            34                      B 3.5-13    0 B 3.4-70            34                      B 3.5-14    0 B 3.4-71            0                        B 3.5-15    34 B 3.4-72            0                        B 3.5-16    34 B 3.4-73            0                        B 3.5-17    0 B 3.4-74            34                      B 3.5-18    34 B 3.4-75            34                      B 3.5-19    0 B 3.4-76            0                        B 3.5-20    0 B 3.4-77            0                        B 3.5-21    0 B 3.4-78            0                        B 3.5-22    0 B 3.4-79            0                        B 3.5-23    0 B 3.4-80            34                      B 3.5-24    0 B 3.4-81            0                        B 3.5-25    0 B 3.4-82            48                      B 3.5-26    0 B 3.4-83            48                      B 3.5-27    57 B 3.4-84            48                      B 3.5-28    0 B 3.4-85            48                      B 3.5-29    57 B 3.4-86            48                      B 3.5-30    0 B 3.4-87            48                      B 3.5-31    0 B 3.4-88            0                        B 3.5-32    34 Watts Bar - Unit 2                    xv                      Revision 57
 
TECHNICAL SPECIFICATIONS BASES LIST OF EFFECTIVE PAGES (continued)
PAGE              REVISION                    PAGE      REVISION NUMBER            NUMBER                    NUMBER      NUMBER B 3.6-37    0 B 3.5-33            0                        B 3.6-38    34 B 3.6-1              0                        B 3.6-39    34 B 3.6-2              0                        B 3.6-40    57 B 3.6-3              0                        B 3.6-41    0 B 3.6-4              0                        B 3.6-42    0 B 3.6-5              52                      B 3.6-43    0 B 3.6-6              0                        B 3.6-44    0 B 3.6-7              52                      B 3.6-45    0 B 3.6-8              0                        B 3.6-46    34 B 3.6-9              0                        B 3.6-47    0 B 3.6-10            0                        B 3.6-48    0 B 3.6-11            0                        B 3.6-49    12 B 3.6-12            34                      B 3.6-50    0 B 3.6-13            0                        B 3.6-51    34 B 3.6-14            0                        B 3.6-52    43 B 3.6-15            0                        B 3.6-53    0 B 3.6-16            0                        B 3.6-54    0 B 3.6-17            0                        B 3.6-55    57 B 3.6-18            0                        B 3.6-56    0 B 3.6-19            0                        B 3.6-57    34 B 3.6-20            0                        B 3.6-58    57 B 3.6-21            34                      B 3.6-59    11 B 3.6-22            34                      B 3.6-60    0 B 3.6-23            34                      B 3.6-61    0 B 3.6-24            52                      B 3.6-62    57 B 3.6-25            34                      B 3.6-63    34 B 3.6-26            52                      B 3.6-64    34 B 3.6-27            52                      B 3.6-65    0 B 3.6-28            57                      B 3.6-66    34 B 3.6-29            57                      B 3.6-67    34 B 3.6-30            0                        B 3.6-68    57 B 3.6-31            0                        B 3.6-69    0 B 3.6-32            34                      B 3.6-70    0 B 3.6-33            0                        B 3.6-71    57 B 3.6-34            0                        B 3.6-72    50 B 3.6-35            0                        B 3.6-73    50 B 3.6-36            57                      B 3.6-74    34 Watts Bar - Unit 2                  xvi                      Revision 57
 
TECHNICAL SPECIFICATIONS BASES LIST OF EFFECTIVE PAGES (continued)
PAGE              REVISION                    PAGE      REVISION NUMBER            NUMBER                    NUMBER      NUMBER B 3.7-21    0 B 3.6-75            34                      B 3.7-22    34 B 3.6-76            34                      B 3.7-23    0 B 3.6-77            57                      B 3.7-24    44 B 3.6-78            0                        B 3.7-25    0 B 3.6-79            0                        B 3.7-26    0 B 3.6-80            0                        B 3.7-27    0 B 3.6-81            34                      B 3.7-28    0 B 3.6-82            34                      B 3.7-29    34 B 3.6-83            0                        B 3.7-30    34 B 3.6-84            0                        B 3.7-31    0 B 3.6-85            0                        B 3.7-32    0 B 3.6-86            34                      B 3.7-33    42 B 3.6-87            4                        B 3.7-34    34 B 3.6-88            4                        B 3.7-35    0 B 3.6-89            45                      B 3.7-36    6 B 3.6-90            45                      B 3.7-37    0 B 3.6-91            53                      B 3.7-38    0 B 3.7-1              0                        B 3.7-39    34 B 3.7-2              0                        B 3.7-40    34 B 3.7-3              42                      B 3.7-41    34 B 3.7-4              0                        B 3.7-42    0 B 3.7-5              0                        B 3.7-43    0 B 3.7-6              0                        B 3.7-44    33 B 3.7-7              0                        B 3.7-44a    33 B 3.7-8              0                        B 3.7-45    34 B 3.7-9              0                        B 3.7-46    34 B 3.7-10            0                        B 3.7-47    0 B 3.7-11            0                        B 3.7-48    34 B 3.7-12            34                      B 3.7-49    0 B 3.7-13            0                        B 3.7-50    0 B 3.7-14            0                        B 3.7-51    0 B 3.7-15            0                        B 3.7-52    0 B 3.7-16            0                        B 3.7-53    0 B 3.7-17            0                        B 3.7-54    0 B 3.7-18            34                      B 3.7-55    9 B 3.7-19            0                        B 3.7-56    43 B 3.7-20            0                        B 3.7-57    0 Watts Bar - Unit 2                  xvii                    Revision 57
 
TECHNICAL SPECIFICATIONS BASES LIST OF EFFECTIVE PAGES (continued)
PAGE              REVISION                    PAGE      REVISION NUMBER            NUMBER                    NUMBER      NUMBER B 3.8-1      40 B 3.7-58            0                        B 3.8-2      40 B 3.7-59            0                        B 3.8-3      54 B 3.7-60            0                        B 3.8-4      40 B 3.7-61            47                      B 3.8-5      40 B 3.7-62            47                      B 3.8-6      54 B 3.7-62a            47                      B 3.8-7      54 B 3.7-62b            47                      B 3.8-8      28 B 3.7-63            0                        B 3.8-9      5 B 3.7-64            13                      B 3.8-10    5 B 3.7-65            13                      B 3.8-10a    54 B 3.7-66            34                      B 3.8-10b    5 B 3.7-67            43                      B 3.8-11    5 B 3.7-67a            34                      B 3.8-12    5 B 3.7-68            0                        B 3.8-12a    54 B 3.7-69            34                      B 3.8-13    28 B 3.7-70            0                        B 3.8-14    28 B 3.7-71            0                        B 3.8-15    28 B 3.7-72            0                        B 3.8-16    54 B 3.7-73            34                      B 3.8-17    34 B 3.7-74            41                      B 3.8-18    34 B 3.7-75            41                      B 3.8-19    34 B 3.7-76            41                      B 3.8-20    34 B 3.7-77            6                        B 3.8-21    34 B 3.7-78            6                        B 3.8-22    40 B 3.7-79            6                        B 3.8-23    34 B 3.7-80            0                        B 3.8-24    34 B 3.7-81            0                        B 3.8-25    34 B 3.7-82            0                        B 3.8-26    0 B 3.7-83            34                      B 3.8-27    34 B 3.7-84            0                        B 3.8-28    34 B 3.7-85            0                        B 3.8-29    34 B 3.7-86            0                        B 3.8-30    34 B 3.7-87            0                        B 3.8-31    34 B 3.7-88            34                      B 3.8-32    34 B 3.7-89            41                      B 3.8-33    34 B 3.7-90            41                      B 3.8-34    34 B 3.7-91            41                      B 3.8-35    40 Watts Bar - Unit 2                  xviii                    Revision 54
 
TECHNICAL SPECIFICATIONS BASES LIST OF EFFECTIVE PAGES (continued)
PAGE              REVISION                    PAGE      REVISION NUMBER            NUMBER                    NUMBER      NUMBER B3.8-70b    31 B 3.8-36            40                      B 3.8-71    31 B 3.8-37            54                      B 3.8-72    31 B 3.8-37a            40                      B3.8-72a    31 B 3.8-38            0                        B3.8-72b    31 B 3.8-39            0                        B 3.8-73    31 B 3.8-40            0                        B 3.8-74    31 B 3.8-41            0                        B 3.8-75    34 B 3.8-42            0                        B 3.8-76    34 B 3.8-43            0                        B 3.8-77    34 B 3.8-44            0                        B3.8-77a    0 B 3.8-45            0                        B 3.8-78    0 B 3.8-46            0                        B 3.8-79    0 B 3.8-47            34                      B 3.8-80    0 B 3.8-48            34                      B 3.8-81    34 B 3.8-49            0                        B 3.8-82    0 B 3.8-50            34                      B 3.8-83    0 B 3.8-51            34                      B 3.8-84    34 B 3.8-52            0                        B 3.8-85    0 B 3.8-53            31                      B 3.8-86    28 B 3.8-54            31                      B3.8-86a    35 B 3.8-55            31                      B 3.8-87    0 B 3.8-56            31                      B 3.8-88    28 B 3.8-57            31                      B 3.8-89    28 B 3.8-58            31                      B 3.8-90    28 B 3.8-59            31                      B 3.8-91    28 B 3.8-60            31                      B 3.8-92    28 B 3.8-61            31                      B 3.8-93    28 B 3.8-62            31                      B3.8-93a    37 B 3.8-63            34                      B3.8-93b    28 B 3.8-64            34                      B 3.8-94    34 B 3.8-65            34                      B 3.8-94a    55 B 3.8-66            31                      B 3.8-95    0 B 3.8-67            34                      B 3.8-96    0 B 3.8-68            0                        B 3.8-97    0 B 3.8-69            31                      B 3.8-98    34 B 3.8-70            31                      B 3.9-1      0 B 3.8-70a            31                      B 3.9-2      0 Watts Bar - Unit 2                  xix                      Revision 55
 
TECHNICAL SPECIFICATIONS BASES LIST OF EFFECTIVE PAGES (continued)
PAGE              REVISION                    PAGE      REVISION NUMBER            NUMBER                    NUMBER      NUMBER B 3.9-3              34 B 3.9-4              0 B 3.9-5              0 B 3.9-6              0 B 3.9-7              34 B 3.9-8              0 B 3.9-9              0 B 3.9-10            34 B 3.9-11            0 B 3.9-12            0 B 3.9-13            0 B 3.9-14            0 B 3.9-15            34 B 3.9-16            0 B 3.9-17            0 B 3.9-18            0 B 3.9-19            34 B 3.9-20            0 B 3.9-21            34 B 3.9-22            0 B 3.9-23            0 B 3.9-24            41 B 3.9-25            41 B 3.9-26            0 B 3.9-27            0 B 3.9-28            0 Watts Bar - Unit 2                    xx                    Revision 41
 
TECHNICAL SPECIFICATION BASES - REVISION LISTING (This listing is an administrative tool maintained by WBN Licensing and may be updated without formally revising the Technical Specification Bases Table-of-Contents)
REVISIONS                                  ISSUED                              SUBJECT NPF-20                                                10-22-15            Low Power Operating License Revision 1                                              2-12-16          TS Bases Table B 3.8.9-1, AC and DC Electrical Power Distribution Systems Revision 2                                              3-18-16            Revise TS Bases B3.3.7, Component Cooling System (CCS), regarding the 1B and 2B surge tank sections.
Revision 3                                              7-11-16            Revise TS Bases B3.6.4, Containment Pressure, and B3.6.6, Containment Spray System regarding the maximum peak containment pressure from a LOCA of 11.73 psig.
Revision 4                                              8-19-16          Revise TS Bases B3.6.15, Shield Building, to clarify the use of the Condition B note.
Revision 5                                              1-17-17          Revises TS Bases B 3.8.1 AC-Sources Revision 6                                              2-24-17          Revises TS Bases B 3.7.7, Component Cooling System (CCS), and B 3.7.16, Component Cooling System (CCS) -
Shutdown.
Revision 7                                              3-13-17            Adds TS Bases B 3.0.8 for Inoperability of Snubbers.
Revision 8                                              4-7-17            Revises TS Bases B 3.4.6.3 to correct the steam generator minimum narrow range level.
Revision 9                                              4-25-17          Revises TS Bases B3.7-10 CREVS.
Revision 10                                              7-14-17          Revises TS Bases SR B3.0.2 for a one-time extension of the Alternating Current Sources.
Revision 11, Amendment 14                                9-29-17          Revises TS Bases B3.6.11 to change the ice mass weight.
Revision 12, Amendment 15                                11-2-17          Revises TS Bases to adopt the TSTF-522 to revise ventilation system surveillance requirements to operate for 10 hours per month.
Revision 13, Amendment 16                                11-2-17          Revises TS Bases B3.7-12 to provide action when both trains of ABGTS are inoperable. Also, B3.8-37a correction of unit error.
Watts Bar - Unit 2                                        xxi                                              Revision 13
 
TECHNICAL SPECIFICATION BASES - REVISION LISTING (This listing is an administrative tool maintained by WBN Licensing and may be updated without formally revising the Technical Specification Bases Table-of-Contents)
REVISIONS                                  ISSUED                              SUBJECT Revision 14                                              11-9-17          Revises TS Bases B 3.8.1 AC Sources -
Operating LCO to correct a typo 1.a.
Revision 15                                            12-13-17            Revises TS Bases B3.6.4 and B 3.6.6 to change the calculated peak pressure.
Revision 16, Amendment 20                              08-20-18            Revises TS Bases B3.1.5, B3.1.6, B3.1.7, and B3.1.8 which adopts the TSTF-547, Clarification of Rod position requirements.
Revision 17, Amendment 21                              09-21-18            Revises TS Bases 3.2.4 and Bases 3.3.1 related to the reactor trip system instrumentation.
Revision 18                                            02-13-19            Revises TS Bases 3.3.1 related to the reactor trip system instrumentation.
Revision 19, Amendment 25                              03-19-19            Revises TS Bases 3.3.4 which adds Table 3.3.4-1, Remote Shutdown System Instrumentation and Controls Revision 20                                            03-21-19            Revises TS Bases Table 3.3.4-1, Remote Shutdown System Instrumentation and Controls Revision 21                                            05-16-19            Revises TS Bases 3.3.1, RTS Instrumentation, to relocation of the Turbine Trip - Low Fluid Oil Pressure trip function from pressure switches in the low pressure fluid oil header to a new location In the high pressure EHC System trip header.
Revision 22, Amendment 24                              06-13-19            Revises TS Bases 3.6.3,Containment Isolation Valves, to change the frequency to in accordance with the Containment Leakage Rate Testing Program.
Revision 23, Amendment 29                                08-1-19          Revises TS Bases 3.8.9, Distribution Systems - Operating, to add a new Condition C.
Revision 24                                              08-1-19          Revises TS Bases 3.2.1 and 3.2.2.
Revision 25                                              08-7-19          Revises TS Bases 3.8.6, surveillance requirements.
Revision 26                                            08-19-19            Revises TS Bases 3.8.4, DC Sources-Operating Watts Bar - Unit 2                                        xxii                                              Revision 26
 
TECHNICAL SPECIFICATION BASES - REVISION LISTING (This listing is an administrative tool maintained by WBN Licensing and may be updated without formally revising the Technical Specification Bases Table-of-Contents)
REVISIONS                                  ISSUED                              SUBJECT Revision 27, Amendment 31                              12-17-19            Revises TS Bases 3.3.5, LOP DG Start Instrumentation, to implement Class 1E unbalanced voltage relays.
Revision 28, Amendment 32                              12-17-19            Revises Tech Specs 3.8.1, 3.8.7, 3.8.8, and 3.8.9 to support performance of the 6.9kV and 480V shutdown board maintenance.
Revision 29                                              1-09-20          Revises TS Bases Table B 3.3.4-1, Remote Shutdown System Instrumentation and Controls Revision 30                                              1-13-20          Revises TS Bases 3.0.2 and 3.0.3 to remove the Term operatioal convenience.
Revision 31, Amendment 33                                1-29-20          TSTF - 500 - DC Electrical Rewrite -
Update to TSTF-360 Revision 32                                              2-20-20          Revises CSST A and B to qualify to GDC-17 requirements in order to be considered as a TS offsite power source substitute for CSST D or C when out of service.
Revision 33, Amendment 35                                3-03-20          Revises Tech Spec Bases 3.7.8 for a one-time extension of completion time for inoperable ERCW train.
Revision 34, Amendment 36                                3-25-20          TSTF-425 - Surveillance Frequency Testing Program.
Revision 35                                              3-19-20          Revises Tech Spec Bases 3.8.9, Distribution Systems - Operating, regarding the Diesel Auxiliary Building Boards.
Revision 36, Amendment 37                                4-8-20            Revises Tech Spec Bases 3.3.5, LOP DG Start Instrumentation for Condition C.
Revision 37                                              6-1-20            Corrected reference in Tech Spec Bases 3.8.9 Condition E, Action E.1, for the repairs to U1 Shutdown boards.
Revision 38, Amendment 39                                7-20-20          Revises miscellaneous administrative changes.
Watts Bar - Unit 2                                        xxiii                                          Revision 38
 
TECHNICAL SPECIFICATION BASES - REVISION LISTING (This listing is an administrative tool maintained by WBN Licensing and may be updated without formally revising the Technical Specification Bases Table-of-Contents)
REVISIONS                                  ISSUED                              SUBJECT Revision 39, Amendment 40                                8-26-20          Revises Tech Spec Bases to allow the use of Westinghouse leak-limiting non-nickel banded Alloy 800 sleeves to repair degraded steam generator tubes.
Revision 40, Amendment 41                                9-15-20          Revises Tech Spec Bases to delete Surveillance Requirement 3.8.1.22 requirement to verify the operability of the automatic transfer from a Unit 2 Service Station Transformer to a Common Station Service Transformer A or B at the associated unit board.
Revision 41, Amendment 27                                11-4-20          Revises Tech Spec Bases 3.7.15, Spent Fuel Pool Assembly Storage, and adding the Bases 3.7.18 Fuel Storage Pool Boron Concentration.
Revision 42, Amendment 42                              11-10-20            Revises Tech Spec Bases 3.7.1, MSSVs, for measurement uncertainty recapture power uprate.
Revision 43, Amendment 43                                12-8-20          Adopts TSTF-541, Revision 2, Add exceptions to Surveillance Requirements for valves and dampers locked in the actuated position.
Revision 44                                            12-14-20            Revises Tech Spec Bases 3.7.5, AFW System, reducing assumed accumulation for the lowest MSSV setpoint.
Revision 45, Amendment 45                              01-05-21            Revise Tech Spec 3.6.15 by deleting existing Condition B and revise the acceptance criteria for annulus pressure.
Revision 46, Amendment 47                              02-02-21            Adopts TSTF-569, Revision 2, Revise Response Time Testing Definition.
Watts Bar - Unit 2                                        xxiv                                          Revision 46
 
TECHNICAL SPECIFICATION BASES - REVISION LISTING (This listing is an administrative tool maintained by WBN Licensing and may be updated without formally revising the Technical Specification Bases Table-of-Contents)
REVISIONS                                  ISSUED                              SUBJECT Revision 47, Amendment 51                              05-19-21            One-Time Change to Tech Spec Bases 3.7.11 to Extend the Completion Time for Main Control Room Chiller Modifications.
Revision 48, Amendment 52                              06-22-21            TSTF-490 - DELETION OF E BAR DEFINITION AND REVISION TO RCS SPECIFIC ACTIVITY TECH SPEC Revision 49, Amendment 53                              06-30-21            Implement WCAP-18124-NP-A to Tech Spec Bases 3.4.3, RCS P/T Limits Revision 50                                            08-26-21            Revises Tech Spec Bases LCO 3.6.12, Ice Condenser Doors, notes issue.
Revision 51                                              09-2-21          Revises Tech Spec Bases 3.4.12, Reference 8.
Revision 52, Amendment 56                              11-30-21            Revises Tech Spec 5.7.2.19, Containment Leakage Rate Testing Program to extend containment integrated and local leak rate test intervals.
Revision 53, Amendment 58                                1-12-22          Revises Tech Spec SR 3.6.15.4.
Revision 54                                              1-25-22          Revises Tech Spec Bases 3.8.1, AC Sources - Operating to support the future shutdown board cleaning activities.
Revision 55                                              2-9-22            Revises Tech Spec Bases Table 3.8.9-1 to remove C&A 1A2-A, 2A2-A, 1B2-B and 2B2-B.
Revision 56, Amendment 49                                3-9-22            Implement WCAP-17661-P-A, Improved RAOC and CAOC FQ Surveillance Tech Specs.
Revision 57, Amendment 50                                3-16-22          Revises Tech Spec 5.9.5 implementation of Full Spectrum LOCA and New LOCA Specific Tritium Producing Burnable Absorber rod stress analysis methodology Watts Bar - Unit 2                                        xxv                                          Revision 57
 
TECHNICAL SPECIFICATION BASES - REVISION LISTING (This listing is an administrative tool maintained by WBN Licensing and may be updated without formally revising the Technical Specification Bases Table-of-Contents)
REVISIONS                                              ISSUED            SUBJECT Revision 58, Amendment 55                              03-30-22            Revises Tech Spec Bases 3.3.2 Function 6.E.
Revision 59, Amendment 57                              04-06-22            Revises Tech Spec Bases RCS Loops -
Steam Generator Secondary Side Water Level.
Watts Bar - Unit 2                                        xxvi                                        Revision 59
 
ENCLOSURE 6 WBN UNIT 2 TECHNICAL SPECIFICATION BASES CHANGED PAGES (109 pages)
 
TABLE OF CONTENTS B 3.7      PLANT SYSTEMS (continued)
B 3.7.7          Component Cooling System (CCS) .... B 3.7-36 B 3.7.8          Essential Raw Cooling Water (ERCW) System  B 3.7-42 B 3.7.9        Ultimate Heat Sink (UHS) . B 3.7-47 B 3.7.10        Control Room Emergency Ventilation System (CREVS) .... B 3.7-50 B 3.7.11        Control Room Emergency Air Temperature Control System (CREATCS) .. B 3.7-59 B 3.7.12        Auxiliary Building Gas Treatment System (ABGTS) .... B 3.7-63 B 3.7.13        Fuel Storage Pool Water Level  B 3.7-68 B 3.7.14        Secondary Specific Activity ............. B 3.7-71 B 3.7.15        Spent Fuel Pool Assembly Storage.                      B 3.7-74 B 3.7.16        Component Cooling System (CCS) - Shutdown.............        B 3.7-77 B 3.7.17        Essential Raw Cooling Water (ERCW) - Shutdown.              B 3.7-84 B 3.7.18        Fuel Storage Pool Boron Concentration                  B 3.7-89 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 3.8-38 B 3.8.3          Diesel Fuel Oil, Lube Oil, and Starting Air ................ B 3.8-43 B 3.8.4          DC Sources - Operating .................. B 3.8-53 B 3.8.5          DC Sources - Shutdown .................. B 3.8-68 B 3.8.6          Battery Parameters ...................... B 3.8-72 B 3.8.7          Inverters - Operating .............................. B 3.8-78 B 3.8.8          Inverters - Shutdown .............................. B 3.8-82 B 3.8.9          Distribution Systems - Operating ............... B 3.8-86 B 3.8.10        Distribution Systems - Shutdown .... B 3.8-95 B 3.9      REFUELING OPERATIONS . B 3.9-1 B 3.9.1          Boron Concentration .. B 3.9-1 B 3.9.2          Unborated Water Source Isolation Valves . B 3.9-5 B 3.9.3          Nuclear Instrumentation .... B 3.9-8 B 3.9.4        RESERVED FOR FUTURE ADDITION  B 3.9-11 B 3.9.5          Residual Heat Removal (RHR) and Coolant Circulation - High Water Level ... B 3.9-12 (continued)
Watts Bar - Unit 2                              iv                                Revision 41
 
TECHNICAL SPECIFICATIONS BASES LIST OF EFFECTIVE PAGES (continued)
PAGE              REVISION                    PAGE      REVISION NUMBER            NUMBER                    NUMBER      NUMBER i                    0                      B 3.0-3      30 ii                    0                      B 3.0-4      0 iii                  0                      B 3.0-5      0 iv                    41                      B 3.0-6      0 v                    0                      B 3.0-7      0 vi                    19                      B 3.0-8      0 vii                  0                      B 3.0-9      0 viii                  0                      B 3.0-10    0 ix                    0                      B 3.0-10a    7 x                    59                      B 3.0-10b    7 xi                    57                      B 3.0-10c    7 xii                  57                      B 3.0-11    0 xiii                  58                      B 3.0-12    38 xiv                  59                      B 3.0-13    30 xv                    57                      B 3.0-14    30 xvi                  57                      B 3.0-15    0 xvii                  57                      B 3.0-16    0 xviii                54                      B 3.1-1      0 xix                  55                      B 3.1-2      0 xx                    41                      B 3.1-3      0 xxi                  13                      B 3.1-4      0 xxii                  26                      B 3.1-5      0 xxiii                38                      B 3.1-6      34 xxiv                  46                      B 3.1-7      0 xxv                  57                      B 3.1-8      0 xxvi                  59                      B 3.1-9      0 B 2.0-1              0                      B 3.1-10    0 B 2.0-2              0                      B 3.1-11    34 B 2.0-3              0                      B 3.1-12    0 B 2.0-4              0                      B 3.1-13    0 B 2.0-5              0                      B 3.1-14    0 B 2.0-6              0                      B 3.1-15    0 B 2.0-7              0                      B 3.1-16    0 B 2.0-8              0                      B 3.1-17    34 B 2.0-9              0                      B 3.1-18    0 B 2.0-10              0                      B 3.1-19    0 B 3.0-1              7                      B 3.1-20    0 B 3.0-2              30                      B 3.1-21    0 Watts Bar - Unit 2                    x                      Revision 59
 
TECHNICAL SPECIFICATIONS BASES LIST OF EFFECTIVE PAGES (continued)
PAGE              REVISION                    PAGE      REVISION NUMBER            NUMBER                    NUMBER      NUMBER B 3.1-59    0 B 3.1-22              0                      B 3.1-60    0 B 3.1-23              0                      B 3.1-61    0 B 3.1-24              0                      B 3.1-62    34 B 3.1-25              16                      B 3.1-63    34 B 3.1-26              0                      B 3.1-64    0 B 3.1-27              0                      B 3.1-65    0 B 3.1-28              16                      B 3.1-66    0 B 3.1-29              16                      B 3.1-67    0 B 3.1-30              16                      B 3.1-68    34 B 3.1-31              16                      B 3.1-69    34 B 3.1-32              34                      B 3.1-70    0 B 3.1-33              34                      B 3.2-1      56 B 3.1-34              0                      B 3.2-2      57 B 3.1-35              0                      B 3.2-3      56 B 3.1-36              16                      B 3.2-4      56 B 3.1-37              16                      B 3.2-5      56 B 3.1-38              16                      B 3.2-6      56 B 3.1-39              34                      B 3.2-7      56 B 3.1-40              16                      B 3.2-8      56 B 3.1-41              16                      B 3.2-9      56 B 3.1-42              16                      B 3.2-10    56 B 3.1-43              16                      B 3.2-11    56 B 3.1-44              34                      B 3.2-12    56 B 3.1-45              34                      B 3.2-13    56 B 3.1-46              0                      B 3.2-13a    56 B 3.1-47              16                      B 3.2-13b    56 B 3.1-48              16                      B 3.2-14    34 B 3.1-49              16                      B 3.2-15    57 B 3.1-50              16                      B 3.2-16    57 B 3.1-51              16                      B 3.2-17    0 B 3.1-52              16                      B 3.2-18    0 B 3.1-53              16                      B 3.2-19    34 B 3.1-54              16                      B 3.2-20    0 B 3.1-55              0                      B 3.2-21    0 B 3.1-56              0                      B 3.2-22    0 B 3.1-57              0                      B 3.2-23    0 B 3.1-58              0                      B 3.2-24    34 Watts Bar - Unit 2                    xi                      Revision 59
 
TECHNICAL SPECIFICATIONS BASES LIST OF EFFECTIVE PAGES (continued)
PAGE              REVISION                    PAGE      REVISION NUMBER            NUMBER                    NUMBER      NUMBER B 3.2-25            0                        B 3.3-32    0 B 3.2-26            57                      B 3.3-33    0 B 3.2-27            57                      B 3.3-34    0 B 3.2-28            0                        B 3.3-35    0 B 3.2-29            0                        B 3.3-36    0 B 3.2-30            34                      B 3.3-37    0 B 3.2-31            0                        B 3.3-38    17 B 3.3-1              0                        B 3.3-39    17 B 3.3-2              0                        B 3.3-40    0 B 3.3-3              0                        B 3.3-41    0 B 3.3-4              0                        B 3.3-42    0 B 3.3-5              0                        B 3.3-43    0 B 3.3-6              0                        B 3.3-44    0 B 3.3-7              0                        B 3.3-45    0 B 3.3-8              0                        B 3.3-46    0 B 3.3-9              0                        B 3.3-47    0 B 3.3-10            0                        B 3.3-48    0 B 3.3-11            0                        B 3.3-49    0 B 3.3-12            0                        B 3.3-50    0 B 3.3-13            18                      B 3.3-51    34 B 3.3-14            0                        B 3.3-52    34 B 3.3-15            18                      B 3.3-53    34 B 3.3-16            0                        B 3.3-54    34 B 3.3-17            0                        B 3.3-55    34 B 3.3-18            0                        B 3.3-56    34 B 3.3-19            0                        B 3.3-57    34 B 3.3-20            0                        B 3.3-58    34 B 3.3-21            0                        B 3.3-59    34 B 3.3-22            0                        B 3.3-60    34 B 3.3-23            0                        B 3.3-61    0 B 3.3-24            0                        B 3.3-62    46 B 3.3-25            0                        B 3.3-63    46 B 3.3-26            0                        B 3.3-64    0 B 3.3-27            21                      B 3.3-65    0 B 3.3-28            0                        B 3.3-66    0 B 3.3-29            0                        B 3.3-67    0 B 3.3-30            0                        B 3.3-68    0 B 3.3-31            0                        B 3.3-69    0 Watts Bar - Unit 2                    xii                    Revision 59
 
TECHNICAL SPECIFICATIONS BASES LIST OF EFFECTIVE PAGES (continued)
PAGE              REVISION                    PAGE      REVISION NUMBER            NUMBER                    NUMBER      NUMBER B 3.3-70            0                        B 3.3-108    0 B 3.3-71            0                        B 3.3-109    0 B 3.3-72            57                      B 3.3-110    0 B 3.3-73            0                        B 3.3-111    58 B 3.3-74            41                      B 3.3-112    0 B 3.3-75            0                        B 3.3-113    36 B 3.3-76            0                        B 3.3-114    36 B 3.3-77            0                        B 3.3-115    34 B 3.3-78            0                        B 3.3-116    34 B 3.3-79            0                        B 3.3-117    34 B 3.3-80            0                        B 3.3-118    0 B 3.3-81            0                        B 3.3-119    57 B 3.3-82            0                        B 3.3-120    0 B 3.3-83            0                        B 3.3-121    58 B 3.3-84            0                        B 3.3-122    0 B 3.3-85            0                        B 3.3-123    0 B 3.3-86            0                        B 3.3-124    0 B 3.3-87            0                        B 3.3-125    0 B 3.3-88            0                        B 3.3-126    0 B 3.3-89            0                        B 3.3-127    0 B 3.3-90            0                        B 3.3-128    0 B 3.3-91            0                        B 3.3-129    0 B 3.3-92            0                        B 3.3-130    0 B 3.3-93            58                      B 3.3-131    0 B 3.3-94            58                      B 3.3-132    0 B 3.3-95            0                        B 3.3-133    0 B 3.3-96            0                        B 3.3-134    0 B 3.3-97            0                        B 3.3-135    34 B 3.3-98            0                        B 3.3-136    34 B 3.3-99            0                        B 3.3-137    0 B 3.3-100            0                        B 3.3-138    0 B 3.3-101            0                        B 3.3-139    19 B 3.3-102            0                        B 3.3-140    19 B 3.3-103            0                        B 3.3-141    34 B 3.3-104            0                        B 3.3-142    34 B 3.3-105            0                        B 3.3-143    0 B 3.3-106            0                        B3.3-143a    29 B 3.3-107            0                        B3.3-143b    20 Watts Bar - Unit 2                  xiii                    Revision 58
 
TECHNICAL SPECIFICATIONS BASES LIST OF EFFECTIVE PAGES (continued)
PAGE              REVISION                    PAGE      REVISION NUMBER            NUMBER                    NUMBER      NUMBER B 3.3-144            27                      B 3.4-14    0 B 3.3-145            0                        B 3.4-15    49 B 3.3-146            27                      B 3.4-16    0 B 3.3-147            27                      B 3.4-17    0 B 3.3-148            36                      B 3.4-18    0 B 3.3-149            34                      B 3.4-19    34 B 3.3-150            27                      B 3.4-20    0 B 3.3-151            0                        B 3.4-21    0 B 3.3-152            0                        B 3.4-22    0 B 3.3-153            0                        B 3.4-23    0 B 3.3-154            0                        B 3.4-24    59 B 3.3-155            34                      B 3.4-25    0 B 3.3-156            34                      B 3.4-26    0 B 3.3-157            34                      B 3.4-27    0 B 3.3-158            34                      B 3.4-28    0 B 3.3-159            0                        B 3.4-29    34 B 3.3-160            0                        B 3.4-30    59 B 3.3-161            0                        B 3.4-31    59 B 3.3-162            0                        B 3.4-32    59 B 3.3-163            34                      B 3.4-33    59 B 3.3-164            34                      B 3.4-34    59 B 3.3-165            0                        B 3.4-35    0 B 3.3-166            0                        B 3.4-36    0 B 3.3-167            0                        B 3.4-37    34 B 3.3-168            34                      B 3.4-38    0 B 3.4-1              0                        B 3.4-39    0 B 3.4-2              0                        B 3.4-40    0 B 3.4-3              0                        B 3.4-41    34 B 3.4-4              34                      B 3.4-42    0 B 3.4-5              34                      B 3.4-43    0 B 3.4-6              0                        B 3.4-44    0 B 3.4-7              0                        B 3.4-45    0 B 3.4-8              34                      B 3.4-46    0 B 3.4-9              0                        B 3.4-47    0 B 3.4-10            0                        B 3.4-48    0 B 3.4-11            49                      B 3.4-49    0 B 3.4-12            0                        B 3.4-50    0 B 3.4-13            0                        B 3.4-51    34 Watts Bar - Unit 2                  xiv                      Revision 59
 
TECHNICAL SPECIFICATIONS BASES LIST OF EFFECTIVE PAGES (continued)
PAGE              REVISION                    PAGE      REVISION NUMBER            NUMBER                    NUMBER      NUMBER B 3.4-89    39 B 3.4-52            0                        B 3.4-90    0 B 3.4-53            0                        B 3.4-91    0 B 3.4-54            0                        B 3.4-92    39 B 3.4-55            0                        B 3.4-93    39 B 3.4-56            0                        B 3.4-94    39 B 3.4-57            0                        B 3.5-1      0 B 3.4-58            0                        B 3.5-2      57 B 3.4-59            0                        B 3.5-3      57 B 3.4-60            0                        B 3.5-4      57 B 3.4-61            34                      B 3.5-5      0 B 3.4-62            34                      B 3.5-6      34 B 3.4-63            34                      B 3.5-7      34 B 3.4-64            51                      B 3.5-8      0 B 3.4-65            0                        B 3.5-9      0 B 3.4-66            0                        B 3.5-10    0 B 3.4-67            0                        B 3.5-11    57 B 3.4-68            0                        B 3.5-12    0 B 3.4-69            34                      B 3.5-13    0 B 3.4-70            34                      B 3.5-14    0 B 3.4-71            0                        B 3.5-15    34 B 3.4-72            0                        B 3.5-16    34 B 3.4-73            0                        B 3.5-17    0 B 3.4-74            34                      B 3.5-18    34 B 3.4-75            34                      B 3.5-19    0 B 3.4-76            0                        B 3.5-20    0 B 3.4-77            0                        B 3.5-21    0 B 3.4-78            0                        B 3.5-22    0 B 3.4-79            0                        B 3.5-23    0 B 3.4-80            34                      B 3.5-24    0 B 3.4-81            0                        B 3.5-25    0 B 3.4-82            48                      B 3.5-26    0 B 3.4-83            48                      B 3.5-27    57 B 3.4-84            48                      B 3.5-28    0 B 3.4-85            48                      B 3.5-29    57 B 3.4-86            48                      B 3.5-30    0 B 3.4-87            48                      B 3.5-31    0 B 3.4-88            0                        B 3.5-32    34 Watts Bar - Unit 2                    xv                      Revision 57
 
TECHNICAL SPECIFICATIONS BASES LIST OF EFFECTIVE PAGES (continued)
PAGE              REVISION                    PAGE      REVISION NUMBER            NUMBER                    NUMBER      NUMBER B 3.6-37    0 B 3.5-33            0                        B 3.6-38    34 B 3.6-1              0                        B 3.6-39    34 B 3.6-2              0                        B 3.6-40    57 B 3.6-3              0                        B 3.6-41    0 B 3.6-4              0                        B 3.6-42    0 B 3.6-5              52                        B 3.6-43    0 B 3.6-6              0                        B 3.6-44    0 B 3.6-7              52                        B 3.6-45    0 B 3.6-8              0                        B 3.6-46    34 B 3.6-9              0                        B 3.6-47    0 B 3.6-10            0                        B 3.6-48    0 B 3.6-11            0                        B 3.6-49    12 B 3.6-12            34                        B 3.6-50    0 B 3.6-13            0                        B 3.6-51    34 B 3.6-14            0                        B 3.5-52    43 B 3.6-15            0                        B 3.6-53    0 B 3.6-16            0                        B 3.6-54    0 B 3.6-17            0                        B 3.6-55    57 B 3.6-18            0                        B 3.6-56    0 B 3.6-19            0                        B 3.6-57    34 B 3.6-20            0                        B 3.6-58    57 B 3.6-21            34                        B 3.6-59    11 B 3.6-22            34                        B 3.6-60    0 B 3.6-23            34                        B 3.6-61    0 B 3.6-24            52                        B 3.6-62    57 B 3.6-25            34                        B 3.6-63    34 B 3.6-26            52                        B 3.6-64    34 B 3.6-27            52                        B 3.6-65    0 B 3.6-28            57                        B 3.6-66    34 B 3.6-29            57                        B 3.6-67    34 B 3.6-30            0                        B 3.6-68    57 B 3.6-31            0                        B 3.6-69    0 B 3.6-32            34                        B 3.6-70    0 B 3.6-33            0                        B 3.6-71    57 B 3.6-34            0                        B 3.6-72    50 B 3.6-35            0                        B 3.6-73    50 B 3.6-36            57                        B 3.6-74    34 Watts Bar - Unit 2                  xvi                      Revision 57
 
TECHNICAL SPECIFICATIONS BASES LIST OF EFFECTIVE PAGES (continued)
PAGE              REVISION                    PAGE      REVISION NUMBER            NUMBER                    NUMBER      NUMBER B 3.7-21    0 B 3.6-75            34                      B 3.7-22    34 B 3.6-76            34                      B 3.7-23    0 B 3.6-77            57                      B 3.7-24    44 B 3.6-78            0                        B 3.7-25    0 B 3.6-79            0                        B 3.7-26    0 B 3.6-80            0                        B 3.7-27    0 B 3.6-81            34                      B 3.7-28    0 B 3.6-82            34                      B 3.7-29    34 B 3.6-83            0                        B 3.7-30    34 B 3.6-84            0                        B 3.7-31    0 B 3.6-85            0                        B 3.7-32    0 B 3.6-86            34                      B 3.7-33    42 B 3.6-87            4                        B 3.7-34    34 B 3.6-88            4                        B 3.7-35    0 B 3.6-89            45                      B 3.7-36    6 B 3.6-90            45                      B 3.7-37    0 B 3.6-91            53                      B 3.7-38    0 B 3.7-1              0                        B 3.7-39    34 B 3.7-2              0                        B 3.7-40    34 B 3.7-3              42                      B 3.7-41    34 B 3.7-4              0                        B 3.7-42    0 B 3.7-5              0                        B 3.7-43    0 B 3.7-6              0                        B 3.7-44    33 B 3.7-7              0                        B 3.7-44a    33 B 3.7-8              0                        B 3.7-45    34 B 3.7-9              0                        B 3.7-46    34 B 3.7-10            0                        B 3.7-47    0 B 3.7-11            0                        B 3.7-48    34 B 3.7-12            34                      B 3.7-49    0 B 3.7-13            0                        B 3.7-50    0 B 3.7-14            0                        B 3.7-51    0 B 3.7-15            0                        B 3.7-52    0 B 3.7-16            0                        B 3.7-53    0 B 3.7-17            0                        B 3.7-54    0 B 3.7-18            34                      B 3.7-55    9 B 3.7-19            0                        B 3.7-56    43 B 3.7-20            0                        B 3.7-57    0 Watts Bar - Unit 2                  xvii                    Revision 57
 
TECHNICAL SPECIFICATIONS BASES LIST OF EFFECTIVE PAGES (continued)
PAGE              REVISION                    PAGE      REVISION NUMBER            NUMBER                    NUMBER      NUMBER B 3.8-1      40 B 3.7-58            0                        B 3.8-2      40 B 3.7-59            0                        B 3.8-3      54 B 3.7-60            0                        B 3.8-4      40 B 3.7-61            47                      B 3.8-5      40 B 3.7-62            47                      B 3.8-6      54 B 3.7-62a            47                      B 3.8-7      54 B 3.7-62b            47                      B 3.8-8      8 B 3.7-63            0                        B 3.8-9      5 B 3.7-64            13                      B 3.8-10    5 B 3.7-65            13                      B 3.8-10a    54 B 3.7-66            34                      B 3.8-10b    5 B 3.7-67            43                      B 3.8-11    5 B 3.7-67a            34                      B 3.8-12    5 B 3.7-68            0                        B 3.8-12a    54 B 3.7-69            34                      B 3.8-13    28 B 3.7-70            0                        B 3.8-14    28 B 3.7-71            0                        B 3.8-15    28 B 3.7-72            0                        B 3.8-16    54 B 3.7-73            34                      B 3.8-17    34 B 3.7-74            41                      B 3.8-18    34 B 3.7-75            41                      B 3.8-19    34 B 3.7-76            41                      B 3.8-20    34 B 3.7-77            6                        B 3.8-21    34 B 3.7-78            6                        B 3.8-22    40 B 3.7-79            6                        B 3.8-23    34 B 3.7-80            0                        B 3.8-24    34 B 3.7-81            0                        B 3.8-25    34 B 3.7-82            0                        B 3.8-26    0 B 3.7-83            34                      B 3.8-27    34 B 3.7-84            0                        B 3.8-28    34 B 3.7-85            0                        B 3.8-29    34 B 3.7-86            0                        B 3.8-30    34 B 3.7-87            0                        B 3.8-31    34 B 3.7-88            34                      B 3.8-32    34 B 3.7-89            41                      B 3.8-33    34 B 3.7-90            41                      B 3.8-34    34 B 3.7-91            41                      B 3.8-35    40 Watts Bar - Unit 2                  xviii                    Revision 54
 
TECHNICAL SPECIFICATIONS BASES LIST OF EFFECTIVE PAGES (continued)
PAGE              REVISION                    PAGE      REVISION NUMBER            NUMBER                    NUMBER      NUMBER B3.8-70b    31 B 3.8-36            40                      B 3.8-71    31 B 3.8-37            54                      B 3.8-72    31 B 3.8-37a            40                      B3.8-72a    31 B 3.8-38            0                        B3.8-72b    31 B 3.8-39            0                        B 3.8-73    31 B 3.8-40            0                        B 3.8-74    31 B 3.8-41            0                        B 3.8-75    34 B 3.8-42            0                        B 3.8-76    34 B 3.8-43            0                        B 3.8-77    34 B 3.8-44            0                        B3.8-77a    0 B 3.8-45            0                        B 3.8-78    0 B 3.8-46            0                        B 3.8-79    0 B 3.8-47            34                      B 3.8-80    0 B 3.8-48            34                      B 3.8-81    34 B 3.8-49            0                        B 3.8-82    0 B 3.8-50            34                      B 3.8-83    0 B 3.8-51            34                      B 3.8-84    34 B 3.8-52            0                        B 3.8-85    0 B 3.8-53            31                      B 3.8-86    28 B 3.8-54            31                      B3.8-86a    35 B 3.8-55            31                      B 3.8-87    0 B 3.8-56            31                      B 3.8-88    28 B 3.8-57            31                      B 3.8-89    28 B 3.8-58            31                      B 3.8-90    28 B 3.8-59            31                      B 3.8-91    28 B 3.8-60            31                      B 3.8-92    28 B 3.8-61            31                      B 3.8-93    28 B 3.8-62            31                      B3.8-93a    37 B 3.8-63            34                      B3.8-93b    28 B 3.8-64            34                      B 3.8-94    34 B 3.8-65            34                      B 3.8-94a    55 B 3.8-66            31                      B 3.8-95    0 B 3.8-67            34                      B 3.8-96    0 B 3.8-68            0                        B 3.8-97    0 B 3.8-69            31                      B 3.8-98    34 B 3.8-70            31                      B 3.9-1      0 B 3.8-70a            31                      B 3.9-2      0 Watts Bar - Unit 2                  xix                      Revision 55
 
TECHNICAL SPECIFICATIONS BASES LIST OF EFFECTIVE PAGES (continued)
PAGE              REVISION                    PAGE      REVISION NUMBER            NUMBER                    NUMBER      NUMBER B 3.9-3              34 B 3.9-4              0 B 3.9-5              0 B 3.9-6              0 B 3.9-7              34 B 3.9-8              0 B 3.9-9              0 B 3.9-10            34 B 3.9-11            0 B 3.9-12            0 B 3.9-13            0 B 3.9-14            0 B 3.9-15            34 B 3.9-16            0 B 3.9-17            0 B 3.9-18            0 B 3.9-19            34 B 3.9-20            0 B 3.9-21            34 B 3.9-22            0 B 3.9-23            0 B 3.9-24            41 B 3.9-25            41 B 3.9-26            0 B 3.9-27            0 B 3.9-28            0 Watts Bar - Unit 2                    xx                    Revision 41
 
TECHNICAL SPECIFICATION BASES - REVISION LISTING (This listing is an administrative tool maintained by WBN Licensing and may be updated without formally revising the Technical Specification Bases Table-of-Contents)
REVISIONS                                  ISSUED                              SUBJECT Revision 39, Amendment 40                                8-26-20          Revises Tech Spec Bases to allow the use of Westinghouse leak-limiting non-nickel banded Alloy 800 sleeves to repair degraded steam generator tubes.
Revision 40, Amendment 41                                9-15-20          Revises Tech Spec Bases to delete Surveillance Requirement 3.8.1.22 requirement to verify the operability of the automatic transfer from a Unit 2 Service Station Transformer to a Common Station Service Transformer A or B at the associated unit board.
Revision 41, Amendment 27                                11-4-20          Revises Tech Spec Bases 3.7.15, Spent Fuel Pool Assembly Storage, and adding the Bases 3.7.18 Fuel Storage Pool Boron Concentration.
Revision 42, Amendment 42                              11-10-20            Revises Tech Spec Bases 3.7.1, MSSVs, for measurement uncertainty recapture power uprate.
Revision 43, Amendment 43                                12-8-20          Adopts TSTF-541, Revision 2, Add exceptions to Surveillance Requirements for valves and dampers locked in the actuated position.
Revision 44                                            12-14-20            Revises Tech Spec Bases 3.7.5, AFW System, reducing assumed accumulation for the lowest MSSV setpoint.
Revision 45, Amendment 45                              01-05-21            Revise Tech Spec 3.6.15 by deleting existing Condition B and revise the acceptance criteria for annulus pressure.
Revision 46, Amendment 47                              02-02-21            Adopts TSTF-569, Revision 2, Revise Response Time Testing Definition.
Watts Bar - Unit 2                                        xxiv                                          Revision 46
 
TECHNICAL SPECIFICATION BASES - REVISION LISTING (This listing is an administrative tool maintained by WBN Licensing and may be updated without formally revising the Technical Specification Bases Table-of-Contents)
REVISIONS                                  ISSUED                              SUBJECT Revision 47, Amendment 51                              05-19-21            One-Time Change to Tech Spec Bases 3.7.11 to Extend the Completion Time for Main Control Room Chiller Modifications.
Revision 48, Amendment 52                              06-22-21            TSTF-490 - DELETION OF E BAR DEFINITION AND REVISION TO RCS SPECIFIC ACTIVITY TECH SPEC Revision 49, Amendment 53                              06-30-21            Implement WCAP-18124-NP-A to Tech Spec Bases 3.4.3, RCS P/T Limits Revision 50                                            08-26-21            Revises Tech Spec Bases LCO 3.6.12, Ice Condenser Doors, notes issue.
Revision 51                                              09-2-21          Revises Tech Spec Bases 3.4.12, Reference 8.
Revision 52, Amendment 56                              11-30-21            Revises Tech Spec 5.7.2.19, Containment Leakage Rate Testing Program to extend containment integrated and local leak rate test intervals.
Revision 53, Amendment 58                                1-12-22          Revises Tech Spec SR 3.6.15.4.
Revision 54                                              1-25-22          Revises Tech Spec Bases 3.8.1, AC Sources - Operating to support the future shutdown board cleaning activities.
Revision 55                                              2-9-22            Revises Tech Spec Bases Table 3.8.9-1 to remove C&A 1A2-A, 2A2-A, 1B2-B and 2B2-B.
Revision 56, Amendment 49                                3-9-22            Implement WCAP-17661-P-A, Improved RAOC and CAOC FQ Surveillance Tech Specs.
Revision 57, Amendment 50                                3-16-22          Revises Tech Spec 5.9.5 implementation of Full Spectrum LOCA and New LOCA Specific Tritium Producing Burnable Absorber rod stress analysis methodology Watts Bar - Unit 2                                        xxv                                          Revision 57
 
TECHNICAL SPECIFICATION BASES - REVISION LISTING (This listing is an administrative tool maintained by WBN Licensing and may be updated without formally revising the Technical Specification Bases Table-of-Contents)
REVISIONS                                              ISSUED            SUBJECT Revision 58, Amendment 55                              03-30-22            Revises Tech Spec Bases 3.3.2 Function 6.E.
Revision 59, Amendment 57                              04-06-22            Revises Tech Spec Bases RCS Loops -
Steam Generator Secondary Side Water Level.
Watts Bar - Unit 2                                        xxvi                                        Revision 59
 
FQ (Z)
B 3.2.1 B 3.2 POWER DISTRIBUTION LIMITS B 3.2.1 Heat Flux Hot Channel Factor (FQ (Z))
BASES BACKGROUND          The purpose of the limits on the values of FQ (Z) is to limit the local (i.e., pellet) peak power density. The value of FQ (Z) varies along the axial height (Z) of the core.
FQ (Z) is defined as the maximum local fuel rod linear power density divided by the average fuel rod linear power density, assuming nominal fuel pellet and fuel rod dimensions. Therefore, FQ (Z) is a measure of the peak fuel pellet power within the reactor core.
During power operation, the global power distribution is limited by LCO 3.2.3, "AXIAL FLUX DIFFERENCE (AFD)," and LCO 3.2.4, "QUADRANT POWER TILT RATIO (QPTR)," which are directly and continuously measured process variables. These LCOs, along with LCO 3.1.7, "Control Bank Insertion Limits," maintain the core limits on power distributions on a continuous basis.
FQ (Z) varies with fuel loading patterns, control bank insertion, fuel burnup, and changes in axial power distribution.
FQ (Z) is measured periodically using the Power Distribution Monitoring System (PDMS). These measurements are generally taken with the core at or near equilibrium conditions.
Using the measured three dimensional power distributions, it is possible to derive a measured value for FQ (Z). However, because this value represents an equilibrium condition, it does not include the variations in the value of FQ (Z) which are present during non-equilibrium situations, such as load following or power ascension.
To account for these possible variations, The elevation dependent measured planar radial peaking factors, FXY(Z), are increased by an elevation dependent factor, [T(Z)]COLR , that accounts for the expected maximum values of the transient axial power shapes postulated to occur during RAOC operation. Thus, [T(Z)]COLR accounts for the worst case non-equilibrium power shapes that are expected for the assumed RAOC operating space.
The RAOC operating space is defined as the combination of AFD and Control Bank Insertion Limits assumed in the calculation of a particular
[T(Z)]COLR function. The [T(Z)]COLR factors are directly dependent on the (continued)
Watts Bar - Unit 2                          B 3.2-1                                    Revision 56 Amendment 49
 
FQ (Z)
B 3.2.1 BASES (continued)
BACKGROUND        AFD and Control Bank Insertion Limit assumptions. The COLR may (continued)      contain different [T(Z)]COLR functions that reflect different operating space limitations. If the limit on FQ(Z) is exceeded, a more restrictive operating space may be implemented to gain margin for future non-equilibrium operation.
Core monitoring and control under non-equilibrium conditions are accomplished by operating the core within the limits of the appropriate LCOs, including the limits on AFD, QPTR, and control rod insertion.
APPLICABLE        This LCO precludes core power distributions that violate the following fuel SAFETY            design criteria:
ANALYSES
: a. During a loss of coolant accident (LOCA), the 10 CFR 50.46 acceptance criteria must be met (Ref. 1);
: b. During a loss of forced reactor coolant flow accident, there must be at least 95% probability at the 95% confidence level (the 95/95 DNB criterion) that the hot fuel rod in the core does not experience a departure from nucleate boiling (DNB) condition;
: c. During an ejected rod accident, the energy deposition to the fuel must not exceed 280 cal/gm (Ref. 2); and
: d. The control rods must be capable of shutting down the reactor with a minimum required SDM with the highest worth control rod stuck fully withdrawn (Ref. 3).
Limits on FQ (Z) ensure that the value of the initial total peaking factor assumed in the accident analyses remains valid.
FQ (Z) limits assumed in the LOCA analysis are typically limiting relative to (i.e., lower than) the FQ (Z) limit assumed in safety analyses for other postulated accidents. Therefore, this LCO provides conservative limits for other postulated accidents.
FQ (Z) satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).
(continued)
Watts Bar - Unit 2                          B 3.2-2                              Revision 56, 57 Amendment 49, 50
 
FQ (Z)
B 3.2.1 BASES (continued)
LCO                The Heat Flux Hot Channel Factor, FQ (Z), shall be limited by the following relationships:
CFQ F Z          K Z    for P  0.5 P
CFQ FQ Z          K Z      for P 0.5
 
===0.5 where===
CFQ is the FQ(Z) limit at RTP provided in the COLR, K(Z) is the normalized FQ(Z) limit as a function of core height provided in the COLR, and THERMAL POWER P
RTP The actual values of CFQ and K(Z) are given in the COLR; however, CFQ is normally a number on the order of 2.5, and K(Z) is a function that looks like the one provided in Figure B 3.2.1-1.
For Relaxed Axial Offset Control operation, FQ(Z) is approximated by FQC (Z) and FQW (Z). Thus, both FQC (Z) and FQW (Z) must meet the preceding limits on FQ (Z) (Ref 5).
An FQC (Z) evaluation requires obtaining an incore power distribution measurement in MODE 1. The measured value, FQM(Z), of FQ(Z) is obtained from the incore power distribution measurement and then corrected for fuel manufacturing tolerances and measurement uncertainty.
(continued)
Watts Bar - Unit 2                        B 3.2-3                                  Revision 56 Amendment 49
 
FQ (Z)
B 3.2.1 BASES LCO                Using the PDMS to obtain the incore power distribution measurement:
(continued)
UQ FQC  1.03 FQM Z  1 100 where 1.03 is the factor that accounts for the fuel manufacturing tolerances and the factor (1+UQ/100), which accounts for measurement uncertainty, is calculated and applied automatically by the BEACON' software (Ref. 4). In order to be consistent with the LOCA analysis and the uncertainty inputs utilized, a minimum uncertainty of 5 should be used for UQ.
FQC (Z) is an excellent approximation for FQ(Z) when the reactor is at the steady state power at which the incore power distribution measurement was obtained.
The expression for FQW (Z) is:
FQW(Z) = 1.03 FXYM (Z) ([T (Z)]COLR /P) AXY(Z) Rj (1 + UQ / 100) for P>0.5 FQW(Z) = 1.03 FXYM(Z) ([T(Z)]COLR / 0.5) AXY(Z) Rj (1 + UQ / 100) for P0.5 The various factors in these expressions are defined below:
FXYM(Z) is the measured radial peaking factor at axial location Z and is equal to the value of FQM(Z)/PM(Z), where PM(Z) is the measured core average axial power shape.
[T(Z)]COLR is the cycle and burnup dependent function, specified in the COLR, which accounts for power distribution transients encountered during non-equilibrium normal operation. [T(Z)]COLR functions are specified for each analyzed RAOC operating space (i.e. each unique combination of AFD limits and Control Bank Insertion Limits). The
[T(Z)]COLR functions account for the limiting non-equilibrium axial power shapes postulated to occur during normal operation for each RAOC operating space. Limiting power shapes at both full and reduced power operation are considered in determining the maximum values of
[T(Z)]COLR. The [T(Z)]COLR functions also account for the following effects:
(1) the presence of spacer grids in the fuel assembly, (2) the increase in radial peaking in rodded core planes due to the presence of control rods during non-equilibrium normal operation, (3) the increase in radial peaking that occurs during part-power operation due to reduced fuel and moderator temperatures, and (4) the increase in radial peaking due to non-equilibrium xenon effects. The [T(Z)]COLR functions are normally calculated assuming that the Surveillance is performed at nominal RTP conditions with all shutdown and control rods fully withdrawn, i.e., all rods out (ARO). Surveillance specific [T(Z)]COLR values may be generated for a given surveillance core condition.
Watts Bar - Unit 2                        B 3.2-4                                  (continued)
Revision 24, 56 Amendment 49
 
FQ (Z)
B 3.2.1 BASES LCO                P is the THERMAL POWER / RTP.
(continued)
AXY(Z) is a function that adjusts the FQW(Z) Surveillance for differences between the reference core condition assumed in generating the
[T(Z)]COLR function and the actual core condition that exists when the Surveillance is performed. Normally, this reference core condition is 100% RTP, all rods out, and equilibrium xenon. For simplicity, AXY(Z) may be assumed to be 1.0 as this will typically result in an accurate FQW(Z) Surveillance result for a Surveillance that is performed at or near the reference core condition, and an underestimation of the available margin to the FQ limit for Surveillances that are performed at core conditions different from the reference condition. Alternatively, the AXY(Z) function may be calculated using the NRC approved methodology in Reference 6.
(1 + UQ/100) is a factor that accounts for measurement uncertainty and 1.03 is a factor that accounts for fuel manufacturing tolerances.
Rj is a cycle and burnup dependent analytical factor specified in the COLR that accounts for potential increases in FQW(Z) between Surveillances. Rj values are provided for each RAOC operating space.
The FQ (Z) limits define limiting values for core power peaking and ensure that the 10 CFR 50.46 acceptance criteria are met during a LOCA (Ref. 1).
This LCO requires operation within the bounds assumed in the safety analyses. Violating the LCO limits for FQ(Z) could result in unacceptable consequences if a design basis event were to occur while FQ(Z) exceeds its specified limits. Calculations are performed in the core design process to confirm that the core can be controlled in such a manner during operation that it can stay within the LOCA FQ (Z) limits. If FQ (Z) cannot be maintained within the LCO limits, reduction of the core power is required, a more restrictive RAOC operating space must be implemented, or core power limits and AFD limits must be reduced.
APPLICABILITY      The FQ (Z) limits must be maintained in MODE 1 to prevent core power distributions from exceeding the limits assumed in the safety analyses.
Applicability in other MODES is not required because there is either insufficient stored energy in the fuel or insufficient energy being transferred to the reactor coolant to require a limit on the distribution of core power.
Watts Bar - Unit 2                        B 3.2-5                                    (continued)
Revision 24, 56 Amendment 49
 
FQ (Z)
B 3.2.1 BASES ACTIONS            A.1 Reducing THERMAL POWER by  1% RTP for each 1% by which FQC (Z) exceeds its limit, maintains an acceptable absolute power density.
FQC (Z) is FQM (Z) multiplied by a factor accounting for manufacturing tolerances and measurement uncertainties. FQM (Z) is the measured value of FQ (Z). The Completion Time of 15 minutes provides an acceptable time to reduce power in an orderly manner and without allowing the plant to remain in an unacceptable condition for an extended period of time.
The maximum allowable power level initially determined by Required Action A.1 may be affected by subsequent determinations of FQC(Z) and would require power reductions within 15 minutes of the FQC(Z) determination, if necessary, to comply with the decreased maximum allowable power level. Decreases in FQC(Z) would allow increasing the maximum allowable power level and increasing power up to the revised limit.
If an FQ surveillance is performed at 100% RTP conditions and both FQC(Z) and FQW(Z) exceed their limits, the option to reduce the THERMAL POWER limit in accordance with Required Action B.2.1, instead of implementing a new operating space in accordance with Required Action B.1, will result in a further power reduction after Required Action A.1 has been completed. However, this further power reduction would be permitted to occur over the next 4 hours. In the event the evaluated THERMAL POWER reduction in the COLR for Required Action B.2.1 did not result in a further power reduction (for example, if both Condition A and Condition B were entered at less than 100% RTP conditions), then the THERMAL POWER level established as a result of completing Required Action A.1 will take precedence, and will establish the effective operating power level limit for the unit until both Conditions A and B are exited.
A.2 A reduction of the Power Range Neutron Flux - High trip setpoints by 1% for each 1% that THERMAL POWER is limited below RATED THERMAL POWER by Required Action A.1, is a conservative action for protection against the consequences of severe transients with unanalyzed power distributions. The Completion Time of 72 hours is sufficient considering the small likelihood of a severe transient in this time period and the preceding prompt reduction in THERMAL POWER in accordance with Required Action A.1. The maximum allowable Power Range Neutron Flux - High trip setpoints initially determined by Required Action A.2 may be affected by subsequent determinations of FQC(Z) and would require Power Range Neutron Flux - High trip setpoint reductions (continued)
Watts Bar - Unit 2                          B 3.2-6                                Revision 56 Amendment 49
 
FQ (Z)
B 3.2.1 BASES ACTIONS            A.2 (continued) within 72 hours of the FQC(Z) determination, if necessary, to comply with the decreased maximum allowable Power Range Neutron Flux - High trip setpoints. Decreases in FQC(Z) would allow increasing the maximum allowable Power Range Neutron Flux - High trip setpoints.
A.3 Reduction in the Overpower T trip setpoints (value of K4) by  1% for each 1% that THERMAL POWER is limited below RATED THERMAL POWER by Required Action A.1, is a conservative action for protection against the consequences of severe transients with unanalyzed power distributions. The Completion Time of 72 hours is sufficient considering the small likelihood of a severe transient in this time period, and the preceding prompt reduction in THERMAL POWER in accordance with Required Action A.1. The maximum allowable Overpower T trip setpoint initially determined by Required Action A.3 may be affected by subsequent determinations of FQC(Z) and would require Overpower T trip setpoint reductions within 72 hours of the FQC(Z) determination, if necessary, to comply with the decreased maximum allowable Overpower T trip setpoints. Decreases in FQC(Z) would allow increasing the maximum allowable Overpower T trip setpoints.
A.4 Verification that FQC (Z) has been restored to within its limit, by performing SR 3.2.1.1 and SR 3.2.1.2 prior to increasing THERMAL POWER above the limit imposed by Required Action A.1, ensures that core conditions during operation at higher power levels and future operation are consistent with safety analyses assumptions.
Condition A is modified by a NOTE that requires Required Action A.4 to be performed whenever the Condition is entered prior to increasing THERMAL POWER above the limit of Required Action A.1. The Note also states that SR 3.2.1.2 is not required to be performed if this Condition is entered prior to THERMAL POWER exceeding 75% RTP after a refueling. This ensures that SR 3.2.1.1 and SR 3.2.1.2 (if required) will be performed prior to increasing THERMAL POWER above the limit of Required Action A.1 even when Condition A is exited prior to performing Required Action A.4. Performance of SR 3.2.1.1 and SR 3.2.1.2 are necessary to ensure FQ(Z) is properly evaluated prior to increasing THERMAL POWER.
(continued)
Watts Bar - Unit 2                        B 3.2-7                                    Revision 56 Amendment 49
 
FQ (Z)
B 3.2.1 BASES ACTIONS            B.1.1 (continued)
If it is found that the maximum calculated value of FQ (Z) that can occur during normal maneuvers, FQW (Z), exceeds its specified limits, there exists a potential for FQC (Z) to become excessively high if a normal operational transient occurs. Implementing a more restrictive RAOC operating space, as specified in the COLR, within the allowed Completion Time of 4 hours will restrict the AFD such that peaking factor limits will not be exceeded during non-equilibrium normal operation. Several RAOC operating spaces, representing successively smaller AFD envelopes and, optionally shallower Control Bank Insertion Limits, may be specified in the COLR. The corresponding T(Z) functions for these operating spaces can be used to determine which RAOC operating space will result in acceptable non-equilibrium operation within the FQW(Z) limit.
B.1.2 If it is found that the maximum calculated value of FQ(Z) that can occur during normal maneuvers, FQW(Z), exceeds its specified limits, there exists a potential for FQC(Z) to become excessively high if a normal operational transient occurs. As discussed above, Required Action B.1.1 requires that a new RAOC operating space be implemented to restore FQW(Z) to within its limits. Required Action B.1.2 requires that SR 3.2.1.1 and SR 3.2.1.2 be performed if control rod motion occurs as a result of implementing the new RAOC operating space in accordance with Required Action B.1.1. The performance of SR 3.2.1.1 and SR 3.2.1.2 is necessary to ensure FQ(Z) is properly evaluated after any rod motion resulting from the implementation of a new RAOC operating space in accordance with Required Action B.1.1.
B.2.1 When FQW(Z) exceeds it limit, Required Action B.2 may be implemented instead of Required Action B.1. Required Action B.2.1 limits THERMAL POWER to less than RATED THERMAL POWER by the amount specified in the COLR. It also requires reductions in the AFD limits by the amount specified in the COLR. This maintains an acceptable absolute power density relative to the maximum power density value assumed in the safety analyses.
If the required FQW(Z) margin improvement exceeds the margin improvement available from the pre-analyzed THERMAL POWER and AFD reductions provided in the COLR, then THERMAL POWER must be further reduced to less than or equal to 50% RTP. In this case, reducing THERMAL POWER to less than or equal to 50% RTP will provide (continued)
Watts Bar - Unit 2                          B 3.2-8                                Revision 56 Amendment 49
 
FQ (Z)
B 3.2.1 BASES ACTIONS            B.2.1 (continued) additional margin in the transient FQ by the required change in THERMAL POWER and the increase in the FQ limit. This will ensure that the FQ limit is met during transient operation that may occur at or below 50% RTP.
The Completion Time of 4 hours provides an acceptable time to reduce the THERMAL POWER and AFD limits in an orderly manner to preclude entering an unacceptable condition during future non-equilibrium operation. The limit on THERMAL POWER initially determined by Required Action B.2.1 may be affected by subsequent determinations of FQW(Z) and may require further power reductions, if necessary, to comply with the decreased THERMAL POWER limit. If the 4 hour completion time had expired since the condition B entry had been made, then Condition C would be required to be entered. Decreases in FQW(Z),
during subsequent FQW(Z) determinations, however, would allow increasing the THERMAL POWER limit and increasing THERMAL POWER up to this revised limit.
Required Action B.2.1 is modified by a NOTE that states Required Action B.2.4 shall be completed whenever Required Action B.2.1 is performed prior to increasing THERMAL POWER above the limit of Required Action B.2.1. Required Action B.2.4 requires the performance of SR 3.2.1.1 and SR 3.2.1.2 prior to increasing THERMAL POWER above the limit established by Required Action B.2.1. The Note ensures that the SRs will be performed even if Condition B may be exited prior to performing Required Action B.2.4. The performance of SR 3.2.1.1 and SR 3.2.1.2 is necessary to ensure FQ(Z) is properly evaluated prior to increasing THERMAL POWER.
If an FQ surveillance is performed at 100% RTP conditions, and both FQC(Z) and FQW(Z) exceed their limits, the option to reduce the THERMAL POWER limit in accordance with proposed Required Action B.2.1 instead of implementing a new operating space in accordance with proposed Required Action B.1, will result in a further power reduction after Required Action A.1 has been completed. However, this further power reduction would be permitted to occur over the next 4 hours. In the event the evaluated THERMAL POWER reduction in the COLR for proposed Required Action B.2.1 did not result in a further power reduction (for example, if both Condition A and Condition B were entered at less than 100% RTP conditions), then the THERMAL POWER level established as a result of completing Required Action A.1 will take precedence, and will establish the effective operating power level limit for the unit until both Conditions A and B are exited.
(continued)
Watts Bar - Unit 2                        B 3.2-9                                  Revision 56 Amendment 49
 
FQ (Z)
B 3.2.1 BASES ACTIONS            B.2.2 (continued)
A reduction of the Power Range Neutron Flux - High trip setpoints by 1% for each 1%, by which the maximum allowable power is reduced is a conservative action for protection against the consequences of severe transients with unanalyzed power distributions. The Completion Time of 72 hours is sufficient considering the small likelihood of a severe transient in this time period and the preceding prompt reduction in the THERMAL POWER limit and AFD limits in accordance with Required Action B.2.1.
B.2.3 Reduction in the Overpower T trip setpoints value of K4 by  1% for each 1% by which the maximum allowable power is reduced is a conservative action for protection against the consequences of severe transients with unanalyzed power distributions. The Completion Time of 72 hours is sufficient considering the small likelihood of a severe transient in this time period, and the preceding prompt reduction in the THERMAL POWER limit and AFD limits in accordance with Required Action B.2.1.
B.2.4 Verification that FQC (Z) and FQW (Z) have been restored to within limit, by performing SR 3.2.1.1 and SR 3.2.1.2 prior to increasing THERMAL POWER above the maximum allowable power limit imposed by Required Action B.2.1, ensures that core conditions during operation at higher power levels and future operation are consistent with safety analyses assumptions.
C.1 If Required Actions A.1 through A.4 or B.1.1 through B.2.4 are not met within their associated Completion Times, the plant must be placed in a mode or condition in which the LCO requirements are not applicable.
This is done by placing the plant in at least MODE 2 within 6 hours.
This allowed Completion Time is reasonable based on operating experience regarding the amount of time it takes to reach MODE 2 from full power operation in an orderly manner and without challenging plant systems.
(continued)
Watts Bar - Unit 2                        B 3.2-10                                  Revision 56 Amendment 49
 
FQ (Z)
B 3.2.1 BASES SURVEILLANCE      SR 3.2.1.1 REQUIREMENTS Verification that FQC (Z) is within its specified limits involves increasing FQM (Z) to allow for manufacturing tolerance and measurement uncertainties in order to obtain FQC (Z). Specifically, FQM (Z) is the measured value of FQ (Z) obtained from the incore power distribution measurement.
Using the PDMS to obtain the incore power distribution measurement:
1.03        1 100 where 1.03 is the factor that accounts for the fuel manufacturing tolerances and the factor (1+UQ/100), which accounts for measurement uncertainty, is calculated and applied automatically by the BEACON' software (Ref. 4). In order to be consistent with the LOCA analysis and the uncertainty inputs utilized, a minimum uncertainty of 5 should be used for UQ.
FQC (Z) is then compared to its specified limits. The limit with which FQC (Z) is compared varies inversely with power above 50% RTP and directly with a function called K(Z) provided in the COLR.
Performing this Surveillance in MODE 1 prior to exceeding 75% RTP following a refueling ensures that some determinations of FQC(Z) is made prior to achieving a significant power level where peak linear heat rate could approach the limits assumed in the safety analysis.
If THERMAL POWER has been increased by  10% RTP since the initial or most recent determination of FQC (Z), another evaluation of this factor is required 24 hours after achieving equilibrium conditions at this higher power level (to ensure that FQC (Z) values are being reduced sufficiently with power increase to stay within the LCO limits). Equilibrium conditions are achieved when the core is sufficiently stable at the intended operating conditions required to perform the surveillance.
The allowance of up to 24 hours after achieving equilibrium conditions at the increased THERMAL POWER level to complete the next FQC(Z) surveillance applies to situations where the FQC(Z) has already been measured at least once at a reduced THERMAL POWER level. The observed margin in the previous surveillance will provide assurance that increasing power up to the next plateau will not exceed the FQ limit, and that the core is behaving as designed.
(continued)
Watts Bar - Unit 2                        B 3.2-11                            Revision 24, 34, 56 Amendment 36, 49
 
FQ (Z)
B 3.2.1 BASES SURVEILLANCE      SR 3.2.1.1 (continued)
REQUIREMENTS This Frequency condition is not intended to require verification of these parameters after every 10% increase in RTP above the THERMAL POWER at which the last verification was performed. It only requires verification after a THERMAL POWER is achieved for extended operation that is 10% higher than the THERMAL POWER at which FQC(Z) was last measured.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.2.1.2 The nuclear design process includes calculations performed to determine that the core can be operated within the FQ (Z) limits. Because incore power distribution measurements are taken in steady state conditions, the variations in power distribution resulting from normal operational maneuvers are not present in the incore power distribution measurement data. These variations are, however, conservatively calculated by considering a wide range of unit maneuvers in normal operation.
The measured FQ(Z) can be determined through a synthesis of the measured planar radial peaking factors, FXYM(Z), and the measured core average axial power shape, PM(Z). Thus, FQC(Z) is given by the following expression:
FQC(Z) = 1.03 FXYM(Z) PM(Z) (1 + UQ/100) = 1.03 FQM(Z) (1 + UQ/100)
For RAOC operation, the analytical [T(Z)]COLR functions, specified in the COLR for each RAOC operating space, are used together with the measured FXY(Z) values to estimate FQ(Z) for non-equilibrium operation within the RAOC operating space. When the FXY(Z) values are measured at HFP ARO conditions (P equals to 1, AXY(Z) equals 1.0), FQW(Z) is given by the following expression:
FQW(Z) = 1.03 FXYM(Z) [T(Z)]COLR Rj (1 + UQ/100)
Non-equilibrium operation can result in significant changes to the axial power shape. To a lesser extent, non-equilibrium operation can increase the radial peaking factors, FXY(Z), through control rod insertion and through reduced Doppler and moderator feedback at part-power conditions.
The [T(Z)]COLR functions quantify these effects for the range of power shapes, control rod insertion, and power levels characteristic of the operating space. Multiplying [T(Z)]COLR by the measured full power, un-(continued)
Watts Bar - Unit 2                        B 3.2-12                                  Revision 56 Amendment 49
 
FQ (Z)
B 3.2.1 BASES SURVEILLANCE      SR 3.2.1.2 (continued)
REQUIREMENTS rodded FXYM(Z) value, and the factor that accounts for manufacturing and measurement uncertainties gives FQW(Z) , the maximum total peaking factor postulated for non-equilibrium RAOC operation.
The limit with which FQW (Z) is compared varies inversely with power above 50% RTP and directly with the function K(Z) provided in the COLR.
The [T(Z)]COLR functions are specified in the COLR for discrete core elevations. Incore power distribution measurement results are typically calculated at 30 to 75 core elevations. FQW (Z) evaluations are not applicable for axial core regions, measured in percent of core height:
: a. Lower core region, from 0 to 10% inclusive,
: b. Upper core region, from 90 to 100% inclusive,
: c. Grid plane regions, +2% inclusive, and
: d. Core Plane regions, within 2% of the bank demand positions of the control banks.
These regions of the core are excluded from the evaluation because of the low probability that they would be more limiting in the safety analysis and because of the difficulty of making a precise measurement in these regions. The excluded regions are specified in the COLR and are defined to ensure that the minimum margin location is adequately surveilled. A slightly smaller exclusion zone may be specified, if necessary, to include the limiting margin location in the surveilled region of the core.
SR 3.2.1.2 requires a Surveillance of FQW(Z) during the initial startup following each refueling within 24 hours after exceeding 75% RTP.
THERMAL POWER levels below 75% are typically non-limiting with respect to the limit for FQW(Z). Furthermore, startup physics testing and flux symmetry measurements, also performed at low power, provide confirmation that the core is operating as expected. This Frequency ensures that verification of FQW(Z) is performed prior to extended operation at power levels where the maximum permitted peak LHR could be challenged and that the first required performance of SR 3.2.1.2 after a refueling is performed at a power level high enough to provide a high level of confidence in the accuracy of the Surveillance result.
(continued)
Watts Bar - Unit 2                        B 3.2-13                                  Revision 56 Amendment 49
 
FQ (Z)
B 3.2.1 BASES (continued)
SURVEILLANCE      SR 3.2.1.2 (continued)
REQUIREMENTS Equilibrium conditions are achieved when the core is sufficiently stable at the intended operating conditions required to perform the Surveillance.
If a previous Surveillance of FQW(Z) was performed at part power conditions, SR 3.2.1.2 also requires that FQW(Z) be verified at power levels > 10% RTP above the THERMAL POWER of its last verification within 24 hours after achieving equilibrium conditions. This ensures that FQW(Z) is within its limit using radial peaking factors measured at the higher power level.
The allowance of up to 24 hours after achieving equilibrium conditions will provide a more accurate measurement of FQW(Z) by allowing sufficient time to achieve equilibrium conditions and obtain the power distribution measurement.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
REFERENCES        1. Title 10, Code of Federal Regulations, Part 50.46, "Acceptance Criteria for Emergency Core Cooling Systems for Light Water Nuclear Power Reactors."
: 2. Regulatory Guide 1.77, Rev. 0, "Assumptions Used for Evaluating a Control Rod Ejection Accident for Pressurized water Reactors,"
May 1974.
: 3. Title 10, Code of Federal Regulations, Part 50, Appendix A, "General Design Criteria for Nuclear Power Plants," GDC 26, "Reactivity Control System Redundancy and Capability."
: 4. WCAP-12472-P-A, BEACON' Core Monitoring and Operations Support System, August 1994, (Addendum 2, April 2002).
: 5. WCAP-10216-P-A, Rev 1A, Relaxation of Constant Axial Offset Control (and) FQ Surveillance Technical Specification, February 1994.
: 6. WCAP-17661-P-A, Improved RAOC and CAOC FQ Surveillance Technical Specifications, February 2019.
Watts Bar - Unit 2                        B 3.2-13a                                Revision 56 Amendment 49
 
FQ (Z)
B 3.2.1 BASES DO NOT GPERATE IN THIS AREA (0,1.0)
(12.0.425)
: 0. -Y TMIS tIGUM 13 !OR ILL'JBTRATIOIN ONI.I _
N 0.4 0                  -                  nO NOT USi roR            ---
zi OPRAAiION.
0.3 0.2      -    ---          --  ---          ---          ---------
U.i 0
0    12          3      0            6      7    e  s    10    11    1_'
maight (ft)
Figure B 3.2.1-1 (page 1 of 1)
K(Z) - Normalized FQ (Z) as a Function of Core Height Watts Bar - Unit 2                          B 3.2-13b                                Revision 56 Amendment 49
 
FNH B 3.2.2 BASES BACKGROUND        Operation outside the LCO limits may produce unacceptable (continued)      consequences if a DNB limiting event occurs. The DNB design basis ensures that there is no overheating of the fuel that results in possible cladding perforation with the release of fission products to the reactor coolant.
APPLICABLE        Limits on F preclude core power distributions that exceed the following SAFETY            fuel design limits:
ANALYSES
: a. There must be at least 95% probability at the 95% confidence level (the 95/95 DNB criterion) that the hottest fuel rod in the core does not experience a DNB condition;
: b. During a loss of coolant accident (LOCA), the 10 CFR 50.46 acceptance criteria must be met (Ref. 3);
: c. During an ejected rod accident, the energy deposition to the fuel must not exceed 280 cal/gm (Ref. 1); and
: d. Fuel design limits required by GDC 26 (Ref. 2) for the condition when control rods must be capable of shutting down the reactor with a minimum required SDM with the highest worth control rod stuck fully withdrawn.
For transients that may be DNB limited, F is a significant core parameter. The limits on F ensure that the DNB design basis is met for normal operation, operational transients, and any transients arising from events of moderate frequency. The DNB design basis is met by limiting the minimum local DNB heat flux ratio to a value which satisfies the 95/95 criterion for the DNB correlation used. Refer to the Bases for the Reactor Core Safety Limits, B 2.1.1, for a discussion of the applicable DNBR limits. The W-3 Correlation with a DNBR limit of 1.3 is applied in the heated region below the first mixing vane grid. In addition, the W-3 DNB correlation is applied in the analysis of accident conditions where the system pressure is below the range of the WRB-2M correlation for RFA-2 fuel with IFMs. For system pressures in the range of 500 to 1000 psia, the W-3 correlation DNBR limit is 1.45 instead of 1.3.
Application of these criteria provides assurance that the hottest fuel rod in the core does not experience a DNB.
(continued)
Watts Bar - Unit 2                        B 3.2-15                                  Revision 57 Amendment 50
 
FNH B 3.2.2 BASES APPLICABLE        The allowable F limit increases with decreasing power level. This SAFETY            functionality in F is included in the analyses that provide the Reactor ANALYSES          Core Safety Limits (SLs) of SL 2.1.1. Therefore, any DNB events in (continued)      which the calculation of the core limits is modeled implicitly use this variable value of F in the analyses. Likewise, all transients that may be DNB limited are assumed to begin with an initial F as a function of power level defined by the COLR limit equation.
The LOCA safety analyses that verify compliance with the 10 CFR 50.46 acceptance criteria (Ref. 3) model F as well as the Nuclear Heat Flux Hot Channel Factor (FQ(Z)).
The fuel is protected in part by Technical Specifications, which ensure that the initial conditions assumed in the safety and accident analyses remain valid. The following LCOs ensure this: LCO 3.2.3, "AXIAL FLUX DIFFERENCE (AFD)," LCO 3.2.4, "QUADRANT POWER TILT RATIO (QPTR)," LCO 3.1.7, "Control Bank Insertion Limits," LCO 3.2.2, "Nuclear Enthalpy Rise Hot Channel Factor (F )," and LCO 3.2.1, "Heat Flux Hot Channel Factor (FQ(Z))."
F and FQ(Z) are measured periodically using the PDMS (Ref. 4).
Measurements are generally taken with the core at, or near, steady state conditions. Core monitoring and control under transient conditions (Condition 1 events) are accomplished by operating the core within the limits of the LCOs on AFD, QPTR, and Bank Insertion Limits.
F satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).
LCO                F shall be maintained within the limits of the relationship provided in the COLR.
The F limit identifies the coolant flow channel with the maximum enthalpy rise. This channel has the least heat removal capability and thus the highest probability for a DNB.
The limiting value of F , described by the equation contained in the COLR, is the design radial peaking factor used in the unit safety analyses.
(continued)
Watts Bar - Unit 2                          B 3.2-16                                Revision 57 Amendment 50
 
QPTR B 3.2.4 B 3.2 POWER DISTRIBUTION LIMITS B 3.2.4 QUADRANT POWER TILT RATIO (QPTR)
BASES BACKGROUND        The QPTR limit ensures that the gross radial power distribution remains consistent with the design values used in the safety analyses. Precise radial power distribution measurements are made during startup testing, after refueling, and periodically during power operation.
The power density at any point in the core must be limited so that the fuel design criteria are maintained. Together, LCO 3.2.3, "AXIAL FLUX DIFFERENCE (AFD)," LCO 3.2.4, and LCO 3.1.7, "Control Rod Insertion Limits," provide limits on process variables that characterize and control the three dimensional power distribution of the reactor core. Control of these variables ensures that the core operates within the fuel design criteria and that the power distribution remains within the bounds used in the safety analyses.
APPLICABLE        This LCO precludes core power distributions that violate the following fuel SAFETY            design criteria:
ANALYSES
: a. During a large break loss of coolant accident, the 10 CFR 50.46 criteria must be met (Ref. 1);
: b. During a loss of forced reactor coolant flow accident, there must be at least 95% probability at the 95% confidence level (the 95/95 departure from nucleate boiling (DNB) criterion) that the hot fuel rod in the core does not experience a DNB condition;
: c. During an ejected rod accident, the energy deposition to the fuel must not exceed 280 cal/gm (Ref. 2); and
: d. The control rods must be capable of shutting down the reactor with a minimum required SDM with the highest worth control rod stuck fully withdrawn (Ref. 3).
The LCO limits on the AFD, the QPTR, the Heat Flux Hot Channel Factor N
(FQ(Z)), the Nuclear Enthalpy Rise Hot Channel Factor (FH    ), rod group alignment, sequence, overlap, and control bank insertion are established to preclude core power distributions that exceed the safety analyses limits.
(continued)
Watts Bar - Unit 2                        B 3.2-26                                  Revision 57 Amendment 50
 
QPTR B 3.2.4 APPLICABLE        The QPTR limits ensure that FN and FQ(Z)remain below their limiting SAFETY            values by preventing an undetected change in the gross radial power ANALYSES          distribution.
(continued)
In MODE 1, the FH and FQ(Z) limits must be maintained to preclude core power distributions from exceeding design limits assumed in the safety analyses.
The QPTR satisfies Criterion 2 of 10 CFR50.35(c)(2)(ii).
LCO                The QPTR limit of 1.02, at which corrective action is required, provides a margin of protection for both the DNB ratio and linear heat generation rate contributing to excessive power peaks resulting from X-Y plane power tilts. A limiting QPTR of 1.02 can be tolerated before the margin for uncertainty in FQ(Z) and (FAHN )is possibly challenged.
APPLICABILITY      The QPTR limit must be maintained in MODE 1 with THERMAL POWER
                  > 50% RTP to prevent core power distributions from exceeding the design limits.
Applicability in MODE 1 < 50% RTP and in other MODES is not required because there is either insufficient stored energy in the fuel or insufficient energy being transferred to the reactor coolant to require the implementation of a QPTR limit on the distribution of core power. The QPTR limit in these conditions is, therefore, not important. Note that the FH and Fa(Z) LCOs still apply, but allow progressively higher peaking factors at 50% RTP or lower.
ACTIONS            AA With the QPTR exceeding its limit, a power level reduction of 3% RTP for each 1% by which the QPTR exceeds 1.00 is a conservative tradeoff of total core power with peak linear power. The Completion Time of 2 hours allows sufficient time to identify the cause and correct the tilt. Note that the power reduction itself may cause a change in the tilted condition.
(continued)
Watts Bar - Unit 2                        B 3.2-27
 
RTS Instrumentation B 3.3.1 BASES SURVEILLANCE      SR 3.3.1.15 (continued)
REQUIREMENTS The response time may be verified for components that replace the components that were previously evaluated in Ref. 11 and Ref. fit, provided that the components have been evaluated in accordance with the NRC approved methodology as discussed in Attachment 1 to TSTF-569,"Methodology to Eliminate Pressure Sensor and Protection Channel (for Westinghouse Plants only) Response Time Testing,'
(Ref. 18).
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.3.1.15 is modified by a Note stating that neutron detectors are excluded from RTS RESPONSE TIME testing. This Note is necessary because of the difficulty in generating an appropriate detector input signal. Excluding the detectors is acceptable because the principles of detector operation ensure a virtually instantaneous response.
REFERENCES          1. Watts Bar FSAR, Section 6.0, "Engineered Safety Features."
2,    Watts Bar FSAR, Section 7.0, "Instrumentation and Controls."
: 3. Watts Bar FSAR, Section 15.0, "Accident Analysis."
: 4. Institute of Electrical and Electronic Engineers, IEEE-279-1971, "Criteria for Protection Systems for Nuclear Power Generating Stations," April 5, 1972,
: 5. 10 CFR Part 50.49, "Environmental Qualifications of Electric Equipment Important to Safety for Nuclear Power Plants."
: 6. Regulatory Guide 1.105, "Setpoints for Safety Related Instrumentation," Revision 3.
: 7. WCAP-10271-P-A, Supplement 1, and Supplement 2, Rev. 1, "Evaluation of Surveillance Frequencies and Out of Service Times for the Reactor Protection Instrumentation System," May 1986 and June 1990.
: 8. Watts Bar Technical Requirements Manual, Section 3.3.1, "Reactor Trip System Response Times."
: 9. Evaluation of the applicability of WCAP-10271-P-A, Supplement 1, and Supplement 2, Revision 1, to Watts Bar, Westinghouse Letter WAT-D-10128.
(continued)
Watts Bar - Unit 2                          B 3.3-62 Revision 34, 46 Amendment 36, 47
 
RTS Instrumentation B 3.3.1 BASES REFERENCES        10. Deleted (continued)
: 11. WCAP-13632-P-A Revision 2, "Elimination of Pressure Sensor Response Time Testing Requirements," January 1996
: 12. WCAP-14036-P-A, Revision 1,"Elimination of Periodic Protection Channel Response Time Tests," October 1998.
: 13. Deleted.
: 14. WCAP-14333 P-A, Revision 1, "Probabilistic Risk Analysis of the RPS and ESFAS Test Times and Completion Times,"
October 1998.
: 15. WCAP-15376-P-A, Revision 1, "Risk Informed Assessment of the RTS and ESFAS Surveillance Test Intervals and Reactor Trip Breaker Test and Completion Times," March 2003
: 16. WCAP-12472-P-A, `BEACONTM Core Monitoring and Operations Support System," August 1994(Addendum 2, April 2002).
: 17. TVA Calculation WBPE0689009007,"Demonstrated Accuracy Calculation for Reactor Coolant Pump Undervoltage."
: 18. Attachment 1 to TSTF-569, "Methodology to Eliminate Pressure Sensor and Protection Channel (for Westinghouse Plants only)
Response Time Testing."
Watts Bar - Unit 2                    B 3.3-63                              Revision 46 Amendment 47
 
ESFAS Instrumentation B 3.3.2 BASES APPLICABLE        The LCO generally requires OPERABILITY of four or three channels in SAFETY            each instrumentation function and two channels in each logic and manual ANALYSES,          initiation function. The two-out-of-three and the two-out-of-four LCO, and          configurations allow one channel to be tripped during maintenance or APPLICABILITY      testing without causing an ESFAS initiation. Two logic or manual (continued)      initiation channels are required to ensure no single random failure disables the ESFAS.
The required channels of ESFAS instrumentation provide unit protection in the event of any of the analyzed accidents.
ESFAS protection functions are as follows:
: 1. Safety Injection Safety Injection (SI) provides two primary functions:
: 1. Primary side water addition to ensure maintenance or recovery of reactor vessel water level (coverage of the active fuel for heat removal, clad integrity, and compliance with the 10 CFR 50.46 acceptance criteria (Ref. 23); and
: 2. Boration to ensure recovery and maintenance of SDM (keff < 1.0).
These functions are necessary to mitigate the effects of high energy line breaks (HELBs) both inside and outside of containment. The SI signal is also used to initiate other Functions such as:
Phase A Isolation; Containment Vent Isolation; Reactor Trip; Turbine Trip; Feedwater Isolation; Start of all auxiliary feedwater (AFW) pumps; Control room ventilation isolation; and Enabling automatic switchover of Emergency Core Cooling Systems (ECCS) suction to containment sump.
(continued)
Watts Bar - Unit 2                          B 3.3-72                                Revision 57 Amendment 50
 
ESFAS Instrumentation B 3.3.2 BASES APPLICABLE        e. Auxiliary Feedwater - Trip Of All Main Turbine Driven SAFETY                Feedwater Pumps
: ANALYSES, LCO, and              A trip of all main feed pumps is an indication of a loss of APPLICABILITY        MFW and the subsequent need for some method of decay (continued)        heat and sensible heat removal to bring the reactor back to no load temperature and pressure. Each turbine driven MFW pump (TDMFWP) is equipped with one pressure switch on the control oil line for the speed control system. A low pressure signal from this pressure switch indicates a trip of that pump. The electric motor driven standby main feedwater pump (SBMFWP) trip channel is provided by breaker contacts from the supply breaker of the motor driven SBMFWP in the AFW start logic. The breaker contacts monitor the SBMFWP and close upon the opening of the breaker, indicating that pump has tripped. The trip of both TDMFWPs and the SBMFWP pump will start the motor driven and turbine driven AFW pumps to ensure that enough water is available to act as the heat sink for the reactor.
This Function must be OPERABLE in MODES 1 and 2 in accordance with the applicable Tech Specs Notes to ensure that at least one SG is provided with water to serve as the heat sink to remove reactor decay heat and sensible heat in the event of an accident.
During unit startup in MODE 2 the SBMFP will be providing feedwater to the steam generators. In the unlikely event that the SBMFP trips during this time, the anticipatory AFW auto-start circuity will actuate starting both the motor driven AFW pumps and the turbine driven AFW pump.
In MODE 1, at approximately 10% RTP, the first TDMFP is placed in service. Once the first TDMFP is placed in service, the SBMFP will be removed from service. Under these conditions, a trip of the sole operating TDMFWP would generate an anticipatory AFW auto-start signal causing all three AFW pumps to start. Once the first TDMFWP is supplying feedwater to the steam generators, the SBMFWP trip channel shall be placed in trip status. This ensures during normal operation, should the TDMFWP(s) in operation trip, the AFW auto start function will actuate starting all of the AFW pumps.
Once the SBMFWP is in service, the SBMFWP trip channel is capable of providing an input signal to the AFW start signal upon the trip of the SBMFWP and the TDMFWP will be removed from service. The SBMFWP will then be the (continued)
Watts Bar - Unit 2              B 3.3-93                                  Revision 58 Amendment 55
 
ESFAS Instrumentation B 3.3.2 BASES APPLICABLE        e. Auxiliary Feedwater - Trip Of All Main Turbine Driven SAFETY                Feedwater Pumps (continued)
: ANALYSES, LCO, and              only supply of feedwater to the steam generators going from APPLICABILITY        MODE 1 to MODE 2. In the unlikely event that the SBMFWP trips during this time, the anticipatory AFE auto-Start circuitry will actuate starting both the motor driven AFW pumps and the turbine driven AFW pump.
MODE 1 applicability for the TDMFW pumps allows entry into LCO 3.3.2, Condition J to be suspended for up to 4 hours when placing a TDMFW pump in service or removing a TDMFW pump from service.
This provision will reduce administrative burden on the plant.
Plant safety is not compromised during this short period because the safety grade AFW auto start channels associated with steam generator low-low levels are operable.
In Mode 3, decay heat and sensible heat removal is sufficiently low that adequate time is available for the operator to manually activate the AFW system if a loss of MFW were to occur.
In Modes 4 and 5, the RCPs and MFW pump are normally shut down with decay heat and sensible heat removed by the RHR system. Therefore, neither pump trip is indicative of a condition requiring automatic AFW initiation.
: f. Auxiliary Feedwater - Pump Suction Transfer on Suction Pressure - Low A low pressure signal in the AFW pump suction line protects the AFW pumps against a loss of the normal supply of water for the pumps, the CST. Three pressure switches are located on each motor driven AFW pump suction line from the CST. A low pressure signal sensed by two switches of a set will cause the emergency supply of water for the respective pumps to be aligned. ERCW (safety grade) is then lined up to supply the AFW pumps to ensure an adequate supply of water for the AFW System to maintain at least one of the SGs as the heat sink for reactor decay heat and sensible heat removal.
(continued)
Watts Bar - Unit 2              B 3.3-94                                  Revision 58 Amendment 55
 
ESFAS Instrumentation B 3.3.2 BASES (continued)
ACTIONS            O.1 and O.2 (continued)
The Required Actions have been modified by a Note that allows placing the inoperable channel in the bypassed condition for up to 12 hours while performing routine surveillance testing of the other channels. The 12 hour time limit is justified in References 10 and 17.
P.1 and P.2 Condition P applies to the AFW pump start on trip of the SBMFWP pump.
The OPERABILITY of the AFW System must be assured by allowing automatic start of the AFW System pumps. If the SBMFWP trip channel is inoperable, 48 hours are allowed to restore that channel to OPERABLE status or place it in the tripped condition. Failure to restore the inoperable channel to OPERABLE status or place it in the tripped condition within 48 hours requires the plant to be placed in MODE 3 within the following 6 hours. The allowed Completion Time of 6 hours is reasonable, based on operating experience, to reach MODE 3 from full power conditions in an orderly manner and without challenging plant systems. In MODE 3, the plant does not have any analyzed transients or conditions that require the explicit use of the protection function noted above. The allowance of 48 hours to restore the channel to OPERABLE status or place it in the tripped condition is justified in Reference 7.
Watts Bar - Unit 2                        B 3.3-111                                  Revision 58 Amendment 55
 
ESFAS Instrumentation B 3.3.2 BASES SURVEILLANCE      SR 3.3.2.10 (continued)
REQUIREMENTS WCAP-14036-P-A, Revision 1, Elimination of Periodic Protection Channel Response Time Tests (Ref. 16), provides the basis and methodology for using allocated signal processing and actuation logic response times in the overall verification of the protection system channel response time. The allocations for sensor, signal conditioning and actuation logic response times must be verified prior to placing the component in operational service and re-verified following maintenance that may adversely affect response time. In general, electrical repair work does not impact response time provided the parts used for repair are of the same type and value. Specific components identified in the WCAP may be replaced without verification testing. One example where response time could be affected is replacing the sensing assembly of a transmitter.
The response time may be verified for components that replace the components that were previously evaluated in Ref. 15 and Ref. 16, provided that the components have been evaluated in accordance with the NRC approved methodology as discussed in Attachment 1 to TSTF-569, Methodology to Eliminate Pressure Sensor and Protection Channel (for Westinghouse Plants only) Response Time Testing, (Ref. 23).
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
This SR is modified by a Note indicating that the SR should be deferred until suitable test conditions are established. This deferral is required because there may be insufficient steam pressure to perform the test.
SR 3.3.2.11 SR 3.3.2.11 is the performance of a TADOT as described in SR 3.3.2.8, except that it is performed for the P-4 Reactor Trip Interlock, and the Frequency is once per RTB cycle. This Frequency is based on operating experience demonstrating that undetected failure of the P-4 interlock sometimes occurs when the RTB is cycled.
The SR is modified by a Note that excludes verification of setpoints during the TADOT. The Function tested has no associated setpoint.
Watts Bar - Unit 2                        B 3.3-119                          Revision 34, 46, 57 Amendment 36, 47, 50
 
ESFAS Instrumentation B 3.3.2 BASES REFERENCES        10. Evaluation of the applicability of WCAP-10271-P-A, Supplement 1, (continued)          and Supplement 2, Revision 1, to Watts Bar, Westinghouse letter to TVA WAT-D-10128.
: 11. Deleted
: 12. Deleted
: 13. Deleted
: 14. Not Applicable for Unit 2
: 15. WCAP-13632-P-A Revision 2, Elimination of Pressure Sensor Response Time Testing Requirements, January 1996.
: 16. WCAP-14036-P-A, Revision 1, Elimination of Periodic Protection Channel Response Time Tests, October 1998.
: 17. WCAP-14333-P-A, Revision 1, Probablistic Risk Analysis of the RPS and ESFAS Test Times and Completion Times, October 1998
: 18. Deleted
: 19. Westinghouse letter to TVA, WAT-D-11248, Revised Justification for Applicability of Instrumentation Technical Specification Improvements to the Automatic Switchover to Containment Sump Signal, June 2004.
: 20. Letter from John G. Lamb (NRC) to Mr. Preston D. Swafford (TVA) dated March 4, 2009, Includes Enclosures (a) Amendment No. 75 to Facility Operating License No. NPF-90 for Watts Bar Nuclear Plant, Unit 1 and (b) NRC Safety Evaluation (SE) for Amendment No. 75.
: 21. Deleted
: 22. WCAP-13878-P-A, Revision 2, Reliability Assessment of Potter &
Brumfield MDR Series Relays.
: 23. Code of Federal Regulations, Title 10, Part 50.46, "Acceptance Criteria for Emergency Core Cooling Systems for Light Water Nuclear Power Reactors."
Watts Bar - Unit 2                      B 3.3-121                                Revision 58 Amendment 55
 
RCS P/T Limits B 3.4.3 BASES [contin APPLICABLE        The PIT limits are not derived from esign Basis Accident(SBA)
SAFETY            analyses. They are prescribed during normal operation to avoid ANALYSES          encountering pressure, temperature, and temperature rate of change conditions that might cause undetected flaws to propagate and cause nonductile failure of the RCPB, an unanalyzed condition. References 8 and 9 establish the methodology for determining the PIT limits. Although the PIT limits are not derived from any BA, the PIT limits are acceptance limits since they preclude operation in an unanalyzed condition.
RCS PIT limits satisfy Criterion 2 of 19 CFR 50.36(c)(2)(ii).
LCO                The two elements of this LC are:
: a. The limit curves for heatup, cooldown, and ISLH testing; and
: b. Limits on the rate of change of temperature.
The LCD limits apply to all components of the RCS, except the pressurizer. These limits define allowable operating regions and permit a large number of operating cycles while providing a wide margin to nonductile failure.
The limits for the rate of change of temperature control and the thermal gradient through the vessel wall are used as inputs for calculating the heatup, cooldown, and ISLH testing P/T limit curves. Thus, the LCO for the rate of change of temperature restricts stresses caused by thermal gradients and also ensures the validity of the PIT limit curves.
Violating the LCO limits places the reactor vessel outside of the bounds of the stress analyses and can increase stresses in other RCPB components. The consequences depend on several factors, as follow:
: a. The severity of the departure from the allowable operating P/T regime r the severity of the rate of change of temperature;
: b. The length of time the limits were violated (longer violations allow the temperature gradient in the thick vessel walls to become more pronounced); and
: c. The existences, sizes, and orientations of flaws in the vessel material.
Watts Bar - Unit 2                        B 3.4-11                                (continued)
Revision 49 Amendment 53
 
RCS P1T Limits B 3.4.3 BASES SURVEILLANCE      SR 3.4.3.1 REQUIREMENTS Verification that operation is within the PTLR limits is when RCS pressure and temperature conditions are undergoing planned changes. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
Surveillance for heatup, cooldown, or ISLH testing may be discontinued when the definition given in the relevant plant procedure for ending the activity is satisfied.
This SR is modified by a Note that only requires this SR to be performed during system heatup, cooldown, and ISLH testing. No SR is given for criticality operations because LCO 3.4.2 contains a more restrictive requirement.
REFERENCES        1.      Appendix "B" to RCS System Description N3-68-4091,"Watts Bar Unit 2 RCS Pressure and Temperature Limits Report."
: 2.      Title 19, Code of Federal Regulations, Part 50, Appendix G, "Fracture Toughness Requirements."
: 3.      ASME Boiler and Pressure Vessel Code, Section XI, Appendix G, "Fracture Toughness Criteria for Protection Against Failure."
: 4.      ASTM E 185-82,"Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels,"
July 1982.
: 5.      Title 10, Code of Federal Regulations, Part 50, Appendix H, "Reactor Vessel Material Surveillance Program Requirements."
: 6.      Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials," May 1988.
: 7.      ASME Boiler and Pressure Vessel Code, Section XI, Appendix E, "Evaluation of Unanticipated Operating Events."
: 8.      WCAP-14040-A, Revision 4,"Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves," May 2004.
: 9.      WCAP-18124-NP-A, Revision 0 "Fluence Determination with RAPTOR-M3G and FERRET," July 2018.
Watts Bar - Unit 2                        B 3.4-15 Revision 34, 49 Amendment 36, 53
 
RCS Loops - MODE 3 B 3.4.5 BASES (continued)
ACTIONS            D.1, D.2, and D.3 (continued)
If all RCS loops are inoperable or no RCS loop is in operation, except as during conditions permitted by the Note in the LCO section, all CRDMs must be de-energized by opening the RTBs or de-energizing the MG sets. All operations involving a reduction of RCS boron concentration must be suspended, and action to restore one of the RCS loops to OPERABLE status and operation must be initiated. Boron dilution requires forced circulation for proper mixing and opening the RTBs or de-energizing the MG sets removes the possibility of an inadvertent rod withdrawal. The immediate Completion Time reflects the importance of maintaining operation for heat removal. The action to restore must be continued until one loop is restored to OPERABLE status and operation.
SURVEILLANCE      SR 3.4.5.1 REQUIREMENTS This SR requires verification that the required loops are in operation.
Verification includes flow rate, temperature, and pump status monitoring, which help ensure that forced flow is providing heat removal. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.4.5.2 SR 3.4.5.2 requires verification of SG OPERABILITY. SG OPERABILITY is verified by ensuring that the secondary side narrow range water level is greater than or equal to 32% (value accounts for instrument error, Ref. 1) for required RCS loops. If the SG secondary side narrow range water level is less than 32%, the tubes may become uncovered and the associated loop may not be capable of providing the heat sink for removal of the decay heat. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.4.5.3 Verification that the required RCPs are OPERABLE ensures that safety analyses limits are met. The requirement also ensures that an additional RCP can be placed in operation, if needed, to maintain decay heat removal and reactor coolant circulation. Verification is performed by verifying proper breaker alignment and power availability to the required RCPs. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
REFERENCES        1.      Watts Bar Drawing 2-47W605-242, Electrical Tech Spec Compliance Tables Watts Bar - Unit 2                        B 3.4-24 Revision 34, 59 Amendment 36, 57
 
RCS Loops - MODE 4 B 3.4.6 BASES SURVEILLANCE      SR 3.4.6.2 REQUIREMENTS (continued)        This SR requires verification that one required RCS or RHR loop is in operation when the rod control system is not capable of rod withdrawal.
Verification includes flow rate, temperature, or pump status monitoring, which help ensure that forced flow is providing heat removal. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.4.6.3 SR 3.4.6.3 requires verification of SG OPERABILITY. SG OPERABILITY is verified by ensuring that the secondary side narrow range water level is greater than or equal to 32% (value accounts for instrument error, Ref. 1).
If the SG secondary side narrow range water level is less than 32%, the tubes may become uncovered and the associated loop may not be capable of providing the heat sink necessary for removal of decay heat.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.4.6.4 Verification that the required pump is OPERABLE ensures that an additional RCS or RHR pump can be placed in operation, if needed, to maintain decay heat removal and reactor coolant circulation. Verification is performed by verifying proper breaker alignment and power available to the required pump. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
REFERENCES        1. Watts Bar Drawing 2-47W605-242, Electrical Tech Spec Compliance Tables Watts Bar - Unit 2                        B 3.4-30                        Revision 8, 34, 59 Amendment 8, 36, 57
 
RCS Loops - MODE 5, Loops Filled B 3.4.7 B 3.4 REACTOR COOLANT SYSTEM (RCS)
B 3.4.7 RCS Loops - MODE 5, Loops Filled BASES BACKGROUND          In MODE 5 with the RCS loops filled, the primary function of the reactor coolant is the removal of decay heat and the transfer of this heat to either the steam generator (SG) secondary side coolant or the component cooling water via the residual heat removal (RHR) heat exchangers.
While the principal means for decay heat removal is via the RHR System, the SGs are specified as a backup means for redundancy. Even though the SGs cannot produce steam in this MODE, they are capable of being a heat sink due to their large contained volume of secondary water. As long as the SG secondary side water is at a lower temperature than the reactor coolant, heat transfer will occur. The rate of heat transfer is directly proportional to the temperature difference. The secondary function of the reactor coolant is to act as a carrier for soluble neutron poison, boric acid.
In MODE 5 with RCS loops filled, the reactor coolant is circulated by means of two RHR loops connected to the RCS, each loop containing an RHR heat exchanger, an RHR pump, and appropriate flow and temperature instrumentation for control, protection, and indication.
One RHR pump circulates the water through the RCS at a sufficient rate to prevent boric acid stratification.
The number of loops in operation can vary to suit the operational needs.
The intent of this LCO is to provide forced flow from at least one RHR loop for decay heat removal and transport. The flow provided by one RHR loop is adequate for decay heat removal. The other intent of this LCO is to require that a second path be available to provide redundancy for heat removal.
The LCO provides for redundant paths of decay heat removal capability.
The first path can be an RHR loop that must be OPERABLE and in operation. The second path can be another OPERABLE RHR loop or maintaining two SGs with secondary side water levels greater than or equal to 32% narrow range to provide an alternate method for decay heat removal.
(continued)
Watts Bar - Unit 2                          B 3.4-31                                  Revision 59 Amendment 57
 
RCS Loops - MODE 5, Loops Filled B 3.4.7 BASES (continued)
APPLICABLE        In MODE 5, RCS circulation is considered in the determination of the time SAFETY            available for mitigation of the accidental boron dilution event. The RHR ANALYSES          loops provide this circulation.
RCS Loops - MODE 5 (Loops Filled) have been identified in 10 CFR 50.36(c)(2)(ii) as important contributors to risk reduction.
LCO                The purpose of this LCO is to require that at least one of the RHR loops be OPERABLE and in operation with an additional RHR loop OPERABLE or two SGs with secondary side water level greater than or equal to 32% narrow range. One RHR loop provides sufficient forced circulation to perform the safety functions of the reactor coolant under these conditions. An additional RHR loop is required to be OPERABLE to meet single failure considerations. However, if the standby RHR loop is not OPERABLE, an acceptable alternate method is two SGs with their secondary side water levels greater than or equal to 32% narrow range.
Should the operating RHR loop fail, the SGs could be used to remove the decay heat.
Note 1 allows one RHR loop to be inoperable for a period of up to 2 hours, provided that the other RHR loop is OPERABLE and in operation. This permits periodic surveillance tests to be performed on the inoperable loop during the only time when such testing is safe and possible.
Note 2 requires that the secondary side water temperature of each SG be 50F above each of the RCS cold leg temperatures before the start of a reactor coolant pump (RCP) with an RCS cold leg temperature less than the COMS arming temperature specified in the PTLR. This restriction is to prevent a low temperature overpressure event due to a thermal transient when an RCP is started.
Note 3 provides for an orderly transition from MODE 5 to MODE 4 during a planned heatup by permitting removal of RHR loops from operation when at least one RCS loop is in operation. This Note provides for the transition to MODE 4 where an RCS loop is permitted to be in operation and replaces the RCS circulation function provided by the RHR loops.
RHR pumps are OPERABLE if they are capable of being powered and are able to provide flow if required. An SG can perform as a heat sink when it has an adequate water level and is OPERABLE.
(continued)
Watts Bar - Unit 2                        B 3.4-32                                Revision 59 Amendment 57
 
RCS Loops - MODE 5, Loops Filled B 3.4.7 BASES (continued)
APPLICABILITY      In MODE 5 with RCS loops filled, this LCO requires forced circulation of the reactor coolant to remove decay heat from the core and to provide proper boron mixing. One loop of RHR provides sufficient circulation for these purposes. However, one additional RHR loop is required to be OPERABLE, or the secondary side water level of at least two SGs is required to be greater than or equal to 32% narrow range.
Operation in other MODES is covered by:
LCO 3.4.4, RCS Loops - MODES 1 and 2; LCO 3.4.5, RCS Loops - MODE 3; LCO 3.4.6, RCS Loops - MODE 4; LCO 3.4.8, RCS Loops - MODE 5, Loops Not Filled; LCO 3.9.5, Residual Heat Removal (RHR) and Coolant Circulation - High Water Level (MODE 6); and LCO 3.9.6, Residual Heat Removal (RHR) and Coolant Circulation -
Low Water Level (MODE 6).
ACTIONS            A.1 and A.2 If one RHR loop is inoperable and the required SGs have secondary side water levels less than 32% narrow range, redundancy for heat removal is lost. Action must be initiated immediately to restore a second RHR loop to OPERABLE status or to restore the required SG secondary side water levels. Either Required Action A.1 or Required Action A.2 will restore redundant heat removal paths. The immediate Completion Time reflects the importance of maintaining the availability of two paths for heat removal.
B.1 and B.2 If no RHR loop is in operation, except during conditions permitted by Note 1, or if no loop is OPERABLE, all operations involving a reduction of RCS boron concentration must be suspended and action to restore one RHR loop to OPERABLE status and operation must be initiated. To prevent boron dilution, forced circulation is required to provide proper mixing and preserve the margin to criticality in this type of operation. The immediate Completion Times reflect the importance of maintaining operation for heat removal.
(continued)
Watts Bar - Unit 2                        B 3.4-33                                  Revision 59 Amendment 57
 
RCS Loops - MODE 5, Loops Filled B 3.4.7 BASES (continued)
SURVEILLANCE      SR 3.4.7.1 REQUIREMENTS This SR requires verification that the required loop is in operation.
Verification includes flow rate, temperature, or pump status monitoring, which help ensure that forced flow is providing heat removal. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.4.7.2 Verifying that at least two SGs are OPERABLE by ensuring their secondary side narrow range water levels are greater than or equal to 32% (value accounts for instrument error, Ref. 1) narrow range ensures an alternate decay heat removal method in the event that the second RHR loop is not OPERABLE. If both RHR loops are OPERABLE, this Surveillance is not needed. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.4.7.3 Verification that a second RHR pump is OPERABLE ensures that an additional pump can be placed in operation, if needed, to maintain decay heat removal and reactor coolant circulation. Verification is performed by verifying proper breaker alignment and power available to the RHR pump.
If secondary side water level is greater than or equal to 32% narrow range in at least two SGs, this Surveillance is not needed. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
REFERENCES        1. Watts Bar Drawing 2-47W605-242, Electrical Tech Spec Compliance Tables Watts Bar - Unit 2                        B 3.4-34                            Revision 34, 59 Amendment 36, 57
 
COMS B 3.4.12 BASES (continued)
REFERENCES        1. Title 10, Code of Federal Regulations, Part 50, Appendix G, Fracture Toughness Requirements.
: 2. Generic Letter 88-11, NRC Position on Radiation Embrittlement of Reactor Vessel Materials and Its Impact on Plant Operation.
: 3. ASME Boiler and Pressure Vessel Code, Section III.
: 4. Watts Bar FSAR, Section 15.2, Condition II - Faults of Moderate Frequency.
: 5. Title 10, Code of Federal Regulations, Part 50.46, Acceptance Criteria for Emergency Core Cooling Systems for Light Water Nuclear Power Reactors.
: 6. Title 10, Code of Federal Regulations, Part 50, Appendix K, ECCS Evaluation Models.
: 7. Generic Letter 90-06, Resolution of Generic Issue 70,
                      'Power-Operated Relief Valve and Block Valve Reliability, and Generic Issue 94, 'Additional Low-Temperature Overpressure Protection for Light Water Reactors,' pursuant to 10 CFR 50.44(f).
: 8. Westinghouse Letter to TVA, LTR-SCS-17-34, Cold Overpressure Mitigation System (COMS) Setpoint Analysis for Tritium Production, August 14, 2017.
Watts Bar - Unit 2                    B 3.4-64                                Revision 51
 
RCS Specific Activity B 3.4.16 BASES B 3.4 REACTOR COOLANT SYSTEM {RCS}
B 3.4.16 RCS Specific Activity BASES BACKGROUND          The maximum dose that an individual at the exclusion area boundary can receive for 2 hours following an accident, or at the low population zone outer boundary for the radiological release duration, is specified in 10 CFR 160 (Ref. 1). The maximum dose to the whole body and the thyroid that an individual occupying the Main Control Room can receive for the accident duration is specified in 10 CFR 50, Appendix A, GDC 19.
The limits an specific activity ensure that the offsite and control room doses are appropriately limited during analyzed transients and accidents.
The RCS specific activity LCO limits the allowable concentration level of radionuclides in the reactor coolant. The LCO limits are established to minimize the offsite and Main Control Room dose consequences in the event of a steam generator tube rupture (SGTR)or main steam line break (MSLB)accident.
The LCO contains specific activity limits for both DOSE EQUIVALENT 1-131 and DOSE EQUIVALENT XE-133. The allowable levels are intended to ensure that offsite and control room doses meet the appropriate acceptance criteria in the Standard Review Plan (Ref. 2).
(continued)
Wafts Bar - Unit 2                        B 3.4-82                                    Revision 48 Amendment 52
 
RCS Specific Activity B 3.4.16 BASES APPLICABLE        The LCD limits on the specific activity of the reactor coolant ensure that SAFETY            the resulting offsite and control room doses meet the appropriate SRP ANALYSES          acceptance criteria following a MSLB or SGTR accident. The SGTR and MSLB safety analyses (Refs. 3 and 4)assume the specific activity of the reactor coolant at the LC limit and an existing reactor coolant steam generator(SG)tube leakage rate of 150 gallons per day (GP). The safety analyses assume the specific activity of the secondary coolant at its limit of 0.1 &#xb5;0/gm DOSE EQUIVALENT 1-131 from LCO 3.7.14, "Secondary Specific Activity."
The analyses for the SGTR and MSLB accidents establish the acceptance limits for RCS specific activity. Reference to these analyses is used to assess changes to the unit that could affect RCS specific activity, as they relate to the acceptance limits.
The analyses are for two cases of reactor coolant specific activity. One case assumes specific activity at 0.265 FtCilgm DOSE EQUIVALENT 1-131 with an iodine spike immediately after the accident that increases the iodine activity in the reactor coolant by a factor of 500 times the iodine production rate necessary to maintain a steady state iodine concentration of 0.265 ~LCilgm DOSE EQUIVALENT 1-131. The second case assumes the initial reactor coolant iodine activity at 14 &#xb5;Cilgm DOSE EQUIVALENT 1-131 due to a pre-accident iodine spike caused by an RCS transient. In both cases, the noble gas activity in the reactor coolant equals the LCO limit of 1200/&#xb5;0/gm DOSE EQU IVALENT XE-133.
The analyses also assume a loss of offsite power at the same time as the SGTR and MSLB event. The SGTR causes a reduction in reactor coolant inventory. The reduction initiates a reactor trip from a low pressurizer pressure signal or an RCS overtemperature AT signal. The MSLB results in a reactor trip due to low steam pressure.
For the SGTR, the coincident loss of offsite power causes the steam dump valves to close to protect the condenser. The rise in pressure in the ruptured SG discharges radioactively contaminated steam to the atmosphere through the SG power operated relief valves and the main steam safety valves. The unaffected SGs remove core decay heat by venting steam to the atmosphere until the cooldown ends.
(continued)
Wafts Bar - Unit 2                        B 3.4-83                                  Revision 48 Amendment 52
 
RCS Specific Activity B 3.4.16 BASES APPLICABLE        Operation with iodine specific activity levels greater than the LCO limit is SAFETY            permissible, if the activity levels do not exceed 14 pCilgm DOSE ANALYSES          EQUIVALENT 1-131, in the applicable specification, for more than (continued)        48 hours. The safety analyses have concurrent and pre-accident iodine spiking levels up to 14 tLCilgm DOSE EQUIVALENT 1-131.
The limits on RCS specific activity are also used for establishing standardization in radiation shielding and plant personnel radiation protection practices.
RCS specific activity satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).
LCO                The specific iodine activity is limited to 0.265 ~tCilgm DOSE EQUIVALENT 1-131, and the noble gas specific activity in the reactor coolant is limited to 1200 EtCilgm DOSE EQUIVALENT XE-133, which ensure that offsite and control room doses will meet the appropriate SRP acceptance criteria (Ref. 2).
The MSLB and SGTR accident analyses (Refs. 3 and 4)show that the calculated doses are within acceptable limits. Violation of the LCO may result in reactor coolant radioactivity levels that could, in the event of a MSLB or SGTR, lead to doses that exceed the SRP acceptance criteria (Ref. 2).
APPLICABILITY      In MODES 1, 2, 3, and 4, operation within the LCO limits for DOSE EQUIVALENT 1-131 and DOSE EQUIVALENT XE-133 is necessary to limit the potential consequences of a SGTR or MSLB to within the SRP acceptance criteria (Ref. 2).
In MODES 5 and 6, the steam generators are not being used for decay heat removal, the RCS and steam generators are depressurized, and primary to secondary leakage is minimal. Therefore, the monitoring of RCS specific activity is not required.
(continued)
Watts Bar - Unit 2                        B 3.4-84                                    Revision 48 1_iiii'FTsi~*'f~a
 
RCS Specific Activity B 3.4.16 BASES ACTIONS            A.1 and A.2 With the DOSE EQUIVALENT 1-131 greater than the LCO limit, samples at intervals of 4 hours must be taken to demonstrate that the limit of 14 ~tCilgm is not exceeded. The Completion Time of 4 hours is required to obtain and analyze a sample. Sampling is done to continue to provide a trend.
The DOSE EQUIVALENT 1-131 must be restored to within limits within 48 hours. The Completion Time of 48 hours is acceptable since it is expected that, if there were an iodine spike, the normal coolant iodine concentration would be restored within this time period. Also, there is a low probability of a MSLB or SGTR occurring during this time period.
A Note permits the use of the provisions of LCO 3.6.4.c. This allowance permits entry into the applicable MODE(S) while relying on the ACTIONS.
This allowance is acceptable due to the significant conservatism incorporated into the specific activity limit, the low probability of an event which is limiting due to exceeding this limit, and the ability to restore transient specific activity excursions while the plant remains at, or proceeds to power operation.
B.1 With the DOSE EQUIVALENT XE-133 greater than the LCO limit, DOSE EQUIVALENT XE-133 must be restored to within limit within 48 hours.
The allowed Completion Time of 48 hours is acceptable since it is expected that, if there were a noble gas spike, the normal coolant noble gas concentration would be restored within this time period. Also, there is a low probability of a MSLB or SGTR occurring during this time period.
A Note permits the use of the provisions of LCO 3.6.4.c. This allowance permits entry into the applicable MODES), relying on Required Action B.1 while the DOSE EQUIVALENT XE-133 LCO limit is not met. This allowance is acceptable due to the significant conservatism incorporated into the specific activity limit, the low probability of an event which is limiting due to exceeding this limit, and the ability to restore transient-specific activity excursions while the plant remains at, or proceeds to, power operation.
(continued)
Watts Bar - Unit 2                        B 3.4-85                                    Revision 48 Amendment 52
 
RCS Specific Activity B 3.4.16 BASES ACTIONS            C.1 and C.2 (continued)
If a Required Action and the associated Completion Time of Condition A or B is not met, or if the DSE EQUIVALENT 1-131 is greater than 14 l.tCilgm, the reactor must be brought to MODE 3 within 6 hours and MODE 5 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.
SURVEILLANCE      SR 3.4.16.1 REQUIREMENTS SR 3.4.16.1 requires performing a gamma isotopic analysis as a measure of the noble gas specific activity of the reactor coolant. This measurement is the sum of the degassed gamma activities and the gaseous gamma activities in the sample taken_ This Surveillance provides an indication of any increase in the noble gas specific activity.
Trending the results of this Surveillance allows proper remedial action to be taken before reaching the LCO limit under normal operating conditions. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
Due to the inherent difficulty in detecting Kr-85 in a reactor coolant sample due to masking from radioisotopes with similar decay energies, such as F-18 and 1-134, it is acceptable to include the minimum detectable activity for Kr-85 in the SR 3.4.16.1 calculation. If a specific noble gas nuclide listed in the definition of DOSE EQUIVALENT XE-133 is not detected, it should be assumed to be present at the minimum detectable activity. A Note modifies the SR, which requires the SR to only be performed in MODES 1, 2, and 3 with Ta,9?5000F.
(continued)
Wafts Bar - Unit 2                        B 3.4-86                              Revision 34, 48 Amendment 36, 52
 
RCS Specific Activity B 3.4.16 BASES SURVEILLANCE      SR 3.4.16.2 REQUIREMENTS (continued)      This Surveillance is performed to ensure iodine remains within limit during normal operation and following rapid power changes when fuel failure is more apt to occur. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. The Frequency, between 2 hours and 6 hours after a power change ~: 15% RTP within a 1 hour period, is established because the iodine levels peak during this time following fuel failure; samples at other times would provide inaccurate results.
REFERENCES        1.      Title 16, Code of Federal Regulations, Part 100.11, "Determination of Exclusion Area, Low Population Zone, and Population Center Distance," 1973.
: 2. Standard Review Plan (SRP)Section 15.1.5 Appendix A (SLB)and Section 15.6.3(SGTR).
: 3.      Watts Bar FSAR, Section 15.5.4, "Environmental Consequences of a Postulated Main Steam Line Break."
: 4.      Wats Bar FSAR, Section 15.5.5, "Environmental Consequences of a Postulated Steam Generator Tube Rupture."
Watts Bar - Unit 2                          B 3.4-87                            Revision 34, 48 Amendment 36, 52
 
Accumulators B 3.5.1 BASES BACKGROUND        This interlock also prevents inadvertent closure of the valves during (continued)      normal operation prior to an accident. Although not required for accident mitigation, the valves will automatically open as a result of an SI signal.
These features ensure that the valves meet the requirements of the Institute of Electrical and Electronic Engineers (IEEE) Standard 279-1971 (Ref. 1) for "operating bypasses" and that the accumulators will be available for injection without reliance on operator action.
The accumulator size, water volume, and nitrogen cover pressure are selected so that three of the four accumulators are sufficient to partially cover the core before significant clad melting or zirconium water reaction can occur following a LOCA. The need to ensure that three accumulators are adequate for this function is consistent with the LOCA assumption that the entire contents of one accumulator will be lost via the RCS pipe break during the blowdown phase of the LOCA.
APPLICABLE        The accumulators are assumed OPERABLE in both the large and small SAFETY            break LOCA analyses at full power (Ref. 2). These are the Design Basis ANALYSES          Accidents (DBAs) that establish the acceptance limits for the accumulators. Reference to the analyses for these DBAs is used to assess changes in the accumulators as they relate to the acceptance limits.
In performing the LOCA calculations, conservative assumptions are made concerning the availability of ECCS flow. In the early stages of a LOCA, with or without a loss of offsite power, the accumulators provide the sole source of makeup water to the RCS. The assumption of loss of offsite power is also considered to determine if it yields limiting results. The loss of offsite power assumption imposes a delay wherein the ECCS pumps cannot deliver flow until the diesel generators start, come to rated speed, and go through their timed loading sequence. In cold leg break scenarios, the entire contents of one accumulator are assumed to be lost through the break.
The limiting large break LOCA is a break in the cold leg. During this event, the accumulators discharge to the RCS as soon as RCS pressure decreases to below accumulator pressure.
(continued)
Watts Bar - Unit 2                          B 3.5-2                                  Revision 57 Amendment 50
 
Accumulators B 3.5.1 BASES APPLICABLE        As a conservative estimate, no credit is taken for ECCS pump flow until SAFETY            an effective delay has elapsed. This delay accounts for the diesels ANALYSES          starting (for loss of offsite power assumption) and the pumps being (continued)      loaded and delivering full flow. The delay time is conservatively set to account for SI signal generation. During this time, the accumulators are analyzed as providing the sole source of emergency core cooling. No operator action is assumed during the blowdown phase of a large break LOCA.
The small break LOCA analysis also assumes a time delay before pumped flow is assumed to inject into the reactor coolant system. For intermediate breaks, the rate of blowdown is such that the increase in fuel clad temperature is terminated solely by the accumulators, with pumped flow then providing continued cooling. As break size decreases, the accumulators and centrifugal charging pumps both play a part in terminating the rise in clad temperature. At very small break sizes, the safety injection pumps are capable of mitigating the inventory loss during the small-break LOCA, and the accumulators do not play a significant role in the accident mitigation.
This LCO helps to ensure that the following acceptance criteria established for the ECCS by 10 CFR 50.46, Paragraph b (Ref. 3) will be met with a high level of probability following a LOCA:
: a. Maximum fuel element cladding temperature is  2200F;
: b. Maximum cladding oxidation is  0.17 times the total cladding thickness before oxidation;
: c. Maximum hydrogen generation from a zirconium water reaction is 0.01 times the hypothetical amount that would be generated if all of the metal in the cladding cylinders surrounding the fuel, excluding the cladding surrounding the plenum volume, were to react; and
: d. Core is maintained in a coolable geometry.
Since the accumulators discharge during the blowdown phase of a LOCA, they do not contribute to the long term cooling requirements of 10 CFR 50.46.
The contained water volume is the same as the deliverable volume for the accumulators, since the accumulators are emptied, once discharged.
Both large and small-break analyses use a nominal accumulator line volume from the accumulator to the check valve. The safety analysis assumes accumulator water volumes of 7518 gallons and 8191 gallons.
To allow for instrument inaccuracy, values of 7630 gallons and 8000 gallons are specified.
(continued)
Watts Bar - Unit 2                          B 3.5-3                                Revision 57 Amendment 50
 
Accumulators B 3.5.1 BASES APPLICABLE        The minimum boron concentration setpoint is used in the post LOCA SAFETY            boron concentration calculation. The calculation is performed to assure ANALYSES          reactor subcriticality in a post LOCA environment. Of particular interest is (continued)      the large break LOCA, since no credit is taken for control rod assembly insertion. A reduction in the accumulator minimum boron concentration would produce a subsequent reduction in the available containment sump concentration for post LOCA shutdown and an increase in the maximum sump pH. The maximum boron concentration is used in determining the cold leg to hot leg recirculation injection switchover time and minimum sump pH.
The maximum nitrogen cover pressure analysis limit of 690 psig prevents accumulator relief valve actuation, and ultimately preserves accumulator integrity. The LOCA analyses support a range of 585 psig to 690 psig.
To account for the accumulator tank design pressure rating, and to allow for instrument accuracy values of  610 psig and  660 psig are specified for the pressure indicator in the main control room.
The effects on containment mass and energy releases from the accumulators are accounted for in the appropriate analyses (Refs. 2 and 4).
The accumulators satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii).
LCO                The LCO establishes the minimum conditions required to ensure that the accumulators are available to accomplish their core cooling safety function following a LOCA. Four accumulators are required to ensure that 100% of the contents of three of the accumulators will reach the core during a LOCA. This is consistent with the assumption that the contents of one accumulator spill through the break. If less than three accumulators are injected during the blowdown phase of a LOCA, the ECCS acceptance criteria of 10 CFR 50.46 (Ref. 3) could be violated.
For an accumulator to be considered OPERABLE, the isolation valve must be fully open, power removed above 1000 psig, and the limits established in the SRs for contained volume, boron concentration, and nitrogen cover pressure must be met.
Watts Bar - Unit 2                          B 3.5-4                                (continued)
Revision 57 Amendment 50
 
ECCS - Operating B 3.5.2 BASES BACKGROUND        The ECCS subsystems are actuated upon receipt of an SI signal. The (continued)        actuation of safeguard loads is accomplished in a programmed time sequence for a loss of offsite power. If offsite power is available, the safeguard loads start immediately. If offsite power is not available, the Engineered Safety Feature (ESF) buses shed normal operating loads and are connected to the diesel generators (DGs). Safeguard loads are then actuated in the programmed time sequence. The time delay associated with diesel starting, sequenced loading, and pump starting determines the time required before pumped flow is available to the core following a LOCA.
The active ECCS components, along with the passive accumulators and the RWST covered in LCO 3.5.1, Accumulators, and LCO 3.5.4, Refueling Water Storage Tank (RWST), provide the cooling water necessary to meet GDC 35 (Ref. 1).
APPLICABLE        The LCO helps to ensure that the following acceptance criteria for the SAFETY            ECCS, established by 10 CFR 50.46, Paragraph b (Ref. 2), will be met ANALYSES          with a high level of probability following a LOCA:
: a. Maximum fuel element cladding temperature is  2200&deg;F;
: b. Maximum cladding oxidation is  0.17 times the total cladding thickness before oxidation;
: c. Maximum hydrogen generation from a zirconium water reaction is 0.01 times the hypothetical amount generated if all of the metal in the cladding cylinders surrounding the fuel, excluding the cladding surrounding the plenum volume, were to react;
: d. Core is maintained in a coolable geometry; and
: e. Adequate long term core cooling capability is maintained.
The LCO also limits the potential for a post trip return to power following an MSLB event and ensures that containment temperature limits are met.
(continued)
Watts Bar - Unit 2                        B 3.5-11                                  Revision 57 Amendment 50
 
RWST B 3.5.4 BASES APPLICABLE        In the ECCS analysis, the containment spray temperature is assumed to SAFETY            be equal to the RWST lower temperature limit of 60&deg;F. If the lower ANALYSES          temperature limit is violated, the containment spray further reduces (continued)      containment pressure, which decreases the rate at which steam can be vented out the break and increases peak clad temperature. The acceptable temperature range of 60&deg;F to 105&deg;F is assumed in the large break and small-break LOCA analyses per approved methods (Ref. 2).
The upper temperature limit of 105&deg;F is also used in the containment OPERABILITY analysis. Exceeding the upper temperature limit could result in a higher peak clad temperature, because there is less heat transfer from the core to the injected water following a LOCA. For the containment response following an MSLB, the lower limit on boron concentration and the upper limit on RWST water temperature are used to maximize the total energy release to containment.
The RWST satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii).
LCO                The RWST ensures that an adequate supply of borated water is available to cool and depressurize the containment in the event of a Design Basis Accident (DBA), to cool and cover the core in the event of a LOCA, to maintain the reactor subcritical following a DBA, and to ensure adequate level in the containment sump to support ECCS and Containment Spray System pump operation in the recirculation mode.
To be considered OPERABLE, the RWST must meet the water volume, boron concentration, and temperature limits established in the SRs.
APPLICABILITY      In MODES 1, 2, 3, and 4, RWST OPERABILITY requirements are dictated by ECCS and Containment Spray System OPERABILITY requirements. Since both the ECCS and the Containment Spray System must be OPERABLE in MODES 1, 2, 3, and 4, the RWST must also be OPERABLE to support their operation. Core cooling requirements in MODE 5 are addressed by LCO 3.4.7, RCS Loops - MODE 5, Loops Filled, and LCO 3.4.8, RCS Loops - MODE 5, Loops Not Filled.
MODE 6 core cooling requirements are addressed by LCO 3.9.5, Residual Heat Removal (RHR) and Coolant Circulation - High Water Level, and LCO 3.9.6, Residual Heat Removal (RHR) and Coolant Circulation - Low Water Level.
Watts Bar - Unit 2                        B 3.5-27                                (continued)
Revision 57 Amendment 50
 
RWST B 3.5.4 BASES (continued)
SURVEILLANCE      SR 3.5.4.1 REQUIREMENTS The RWST borated water temperature should be verified to be within the limits assumed in the accident analyses band. The specified temperature range is  60 &deg;F and  105 &deg;F and does not account for instrument error.
The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
The SR is modified by a Note that eliminates the requirement to perform this Surveillance when ambient air temperatures are within the operating limits of the RWST. With ambient air temperatures within the band, the RWST temperature should not exceed the limits.
SR 3.5.4.2 The required minimum RWST water level is  370,000 gallons (value does not account for instrument error). Verification of the presence of this water volume ensures that a sufficient initial supply of water is available for injection and to support continued ECCS and Containment Spray System pump operation on recirculation. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.5.4.3 The boron concentration of the RWST should be verified to be within the required limits. This SR ensures that the reactor will remain subcritical following a LOCA. Further, it assures that the resulting sump pH will be maintained in an acceptable range so that boron precipitation in the core will not occur and the effect of chloride and caustic stress corrosion on mechanical systems and components will be minimized. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
REFERENCES        1. Watts Bar FSAR, Section 6.3, Emergency Core Cooling System, and Section 15.0, Accident Analysis.
: 2. WCAP-16996-P-A, Revision 1, "Realistic LOCA Evaluation Methodology Applied to the Full Spectrum of Break Sizes (FULL SPECTRUM LOCA Methodology)," November 2016.
Watts Bar - Unit 2                        B 3.5-29 Revision 34, 57 Amendment 36, 50
 
Containment B 3.6.1 BASES (continued)
REFERENCES        1. Title 10, Code of Federal Regulations, Part 50, Appendix J, Option B, Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors - Performance-Based Requirements.
: 2. Watts Bar FSAR, Section 15.0, Accident Analysis.
: 3. Watts Bar FSAR, Section 6.2, Containment Systems.
: 4. Nuclear Energy Institute (NEI) 94-01, Industry Guideline for Implementing Performance-Based Option of 10 CFR 50, Appendix J, Revision 3-A, dated July 2012.
Watts Bar - Unit 2                    B 3.6-5                                  Revision 52 Amendment 56
 
Containment Air Locks B 3.6.2 BASES (continued)
APPLICABLE        The DBAs that result in a significant release of radioactive material within SAFETY            containment are a loss of coolant accident and a rod ejection accident ANALYSES          (Ref. 2). In the analysis of each of these accidents, it is assumed that containment is OPERABLE such that release of fission products to the environment is controlled by the rate of containment leakage. The containment was designed with an allowable leakage rate (La) of 0.25%
of containment air weight per day (Ref. 2), at Pa = 15.0 psig. This allowable leakage rate forms the basis for the acceptance criteria imposed on the SRs associated with the air locks.
The containment air locks satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii).
LCO                Each containment air lock forms part of the containment pressure boundary. As part of containment pressure boundary, the air lock safety function is related to control of the containment leakage rate resulting from a DBA. Thus, each air lock's structural integrity and leak tightness are essential to the successful mitigation of such an event.
Each air lock is required to be OPERABLE. For the air lock to be considered OPERABLE, the air lock interlock mechanism must be OPERABLE, the air lock must be in compliance with the Type B air lock leakage test, and both air lock doors must be OPERABLE. The interlock allows only one air lock door of an air lock to be opened at one time. This provision ensures that a gross breach of containment does not exist when containment is required to be OPERABLE. Closure of a single door in each air lock is sufficient to provide a leak tight barrier following postulated events. Nevertheless, both doors are kept closed when the air lock is not being used for normal entry into and exit from containment.
APPLICABILITY      In MODES 1, 2, 3, and 4, a DBA could cause a release of radioactive material to containment. In MODES 5 and 6, the probability and consequences of these events are reduced due to the pressure and temperature limitations of these MODES. Therefore, the containment air locks are not required in MODES 5 and 6 to prevent leakage of radioactive material from containment.
Watts Bar - Unit 2                          B 3.6-7                                    (continued)
Revision 52 Amendment 56
 
Containment Isolation Valves B 3.6.3 BASES SURVEILLANCE      SR 3.6.3.5 REQUIREMENTS (continued)      For containment purge valves with resilient seals, additional leakage rate testing beyond the test requirements of 10 CFR 50, Appendix J, Option B (Ref. 4), is required to ensure OPERABILITY.
Operating experience has demonstrated that this type of seal has the potential to degrade in a shorter time period than do other seal types.
Based on this observation and the importance of maintaining this penetration leak tight (due to the direct path between containment and the environment), these valves will not be placed on the maximum extended test interval. Therefore, these valves will be tested in accordance with NEI 94-01, Revision 3-A, which allows a maximum test interval of 30 months (Ref. 3).
SR 3.6.3.6 Automatic containment isolation valves close on a containment isolation signal to prevent leakage of radioactive material from containment following a DBA. This SR ensures that each automatic containment isolation valve will actuate to its isolation position on a containment isolation signal. This Surveillance is not required for valves that are locked, sealed, or otherwise secured in the required position under administrative control. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
(continued)
Watts Bar - Unit 2                        B 3.6-24                            Revision 22, 34, 52 Amendment 24, 36, 56
 
Containment Isolation Valves B 3.6.3 BASES REFERENCES        1. Watts Bar FSAR, Section 15.0, Accident Analysis.
: 2. Watts Bar FSAR, Section 6.2.4.2, Containment Isolation System Design, and Table 6.2.4-1, Containment Penetrations and Barriers.
: 3. Nuclear Energy Institute (NEI) 94-01, Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J, Revision 3-A, dated July 2012.
: 4. Title 10, Code of Federal Regulations, Part 50 Appendix J, Option B, Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors - Performance - Based Requirements.
Watts Bar - Unit 2                    B 3.6-26                            Revision 22, 52 Amendment 24, 56
 
Containment Pressure B 3.6.4 B 3.6 CONTAINMENT SYSTEMS B 3.6.4 Containment Pressure BASES BACKGROUND          The containment pressure is limited during normal operation to preserve the initial conditions assumed in the accident analyses for a loss of coolant accident (LOCA) or steam line break (SLB). These limits also prevent the containment pressure from exceeding the containment design negative pressure differential (-2.0 psid) with respect to the shield building annulus atmosphere in the event of inadvertent actuation of the Containment Spray System or Air Return Fans.
Containment pressure is a process variable that is monitored and controlled. The containment pressure limits are derived from the input conditions used in the containment functional analyses and the containment structure external pressure analysis. Should operation occur outside these limits coincident with a Design Basis Accident (DBA), post accident containment pressures could exceed calculated values.
APPLICABLE          Containment internal pressure is an initial condition used in the DBA SAFETY              analyses to establish the maximum peak containment internal pressure.
ANALYSES            The limiting DBAs considered, relative to containment pressure, are the LOCA and SLB, which are analyzed using computer pressure transients.
The worst case LOCA generates larger mass and energy release than the worst case SLB. Thus, the LOCA event bounds the SLB event from the containment peak pressure standpoint (Ref. 1).
The initial pressure condition used in the containment analysis was 15.0 psia. This resulted in a maximum peak containment pressure from a LOCA of 9.36 psig. The containment analysis (Ref. 1) shows that the maximum allowable internal containment pressure (15.0 psig) bounds the calculated results from the limiting LOCA. The maximum containment pressure resulting from the worst case LOCA does not exceed the maximum allowable calculated containment pressure of 15.0 psig.
(continued)
Watts Bar - Unit 2                          B 3.6-27                          Revision 3, 15, 52 Amendment 56
 
Containment Pressure B 3.6.4 BASES APPLICABLE        The containment was also designed for an external pressure load SAFETY            equivalent to 2.0 psig. The inadvertent actuation of the Containment ANALYSES          Spray System was analyzed to determine the resulting reduction in (continued)      containment pressure. The initial pressure condition used in this analysis was -0.1 psig. This resulted in a minimum pressure inside containment of 1.4 psig, which is less than the design load.
For certain aspects of transient accident analyses, maximizing the calculated containment pressure is not conservative. In particular, the cooling effectiveness of the Emergency Core Cooling System during the core reflood phase of a LOCA analysis increases with increasing containment backpressure. Therefore, for the reflood phase, the containment backpressure is calculated in a manner designed to conservatively minimize, rather than maximize, the containment pressure response in accordance with WCAP-16996-P-A, Revision 1 (Ref. 2).
Containment pressure satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).
LCO                Maintaining containment pressure at less than or equal to the LCO upper pressure limit ensures that, in the event of a DBA, the resultant peak containment accident pressure will remain below the containment design pressure. Maintaining containment pressure at greater than or equal to the LCO lower pressure limit ensures that the containment will not exceed the design negative differential pressure following the inadvertent actuation of the Containment Spray System or Air Return Fans.
APPLICABILITY      In MODES 1, 2, 3, and 4, a DBA could cause a release of radioactive material to containment. Since maintaining containment pressure within limits is essential to ensure initial conditions assumed in the accident analyses are maintained, the LCO is applicable in MODES 1, 2, 3 and 4.
In MODES 5 and 6, the probability and consequences of these events are reduced due to the pressure and temperature limitations of these MODES. Therefore, maintaining containment pressure within the limits of the LCO is not required in MODES 5 or 6.
Watts Bar - Unit 2                        B 3.6-28                                  (continued)
Revision 57 Amendment 50
 
Containment Pressure B 3.6.4 BASES ACTIONS            A.1 When containment pressure is not within the limits of the LCO, it must be restored to within these limits within 1 hour. The Required Action is necessary to return operation to within the bounds of the containment analysis. The 1 hour Completion Time is consistent with the ACTIONS of LCO 3.6.1, Containment, which requires that containment be restored to OPERABLE status within 1 hour.
B.1 and B.2 If containment pressure cannot be restored to within limits within the required Completion Time, 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 6 hours and to MODE 5 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.
SURVEILLANCE      SR 3.6.4.1 REQUIREMENTS Verifying that containment pressure is within limits ( -0.1 and  +0.3 psid relative to the annulus, value does not account for instrument error) ensures that plant operation remains within the limits assumed in the containment analysis. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
REFERENCES        1. Watts Bar FSAR, Section 6.2.2, Containment Heat Removal Systems.
: 2. WCAP-16996-P-A, Revision 1, "Realistic LOCA Evaluation Methodology Applied to the Full Spectrum of Break Sizes (FULL SPECTRUM LOCA Methodology)," November 2016.
Watts Bar - Unit 2                        B 3.6-29 Revision 34, 57 Amendment 36, 50
 
Containment Spray System B 3.6.6 BASES (continued)
APPLICABLE        The limiting DBAs considered relative to containment are the loss of SAFETY            coolant accident (LOCA) and the steam line break (SLB). The DBA ANALYSES          LOCA and SLB are analyzed using computer codes designed to predict the resultant containment pressure and temperature transients. No two DBAs are assumed to occur simultaneously or consecutively. The postulated DBAs are analyzed, in regard to containment ESF systems, assuming the loss of one ESF bus, which is the worst case single active failure, resulting in one train of the Containment Spray System, the RHR System, and the ARS being rendered inoperable (Ref. 2).
The DBA analyses show that the maximum peak containment pressure of 9.36 psig results from the LOCA analysis and is calculated to be less than the containment maximum allowable pressure of 15 psig. The maximum peak containment atmosphere temperature results from the SLB analysis.
The calculated transient containment atmosphere temperatures are acceptable for the DBA SLB.
The modeled Containment Spray System actuation from the containment analysis is based on a response time associated with exceeding the containment High-High pressure signal setpoint to achieving full flow through the containment spray nozzles. A delayed response time initiation provides conservative analyses of peak calculated containment temperature and pressure responses. The Containment Spray System total response time of 234 seconds is composed of signal delay, diesel generator startup, and system startup time.
For certain aspects of transient accident analyses, maximizing the calculated containment pressure is not conservative. In particular, the ECCS cooling effectiveness during the core reflood phase of a LOCA analysis increases with increasing containment backpressure. For these calculations, the containment backpressure is calculated in a manner designed to conservatively minimize, rather than maximize, the calculated transient containment pressures in accordance with WCAP-16996-P-A, Revision 1 (Ref. 3).
Inadvertent actuation of the Containment Spray System is evaluated in the analysis, and the resultant reduction in containment pressure is calculated. The maximum calculated steady state pressure differential relative to the Shield Building annulus is 1.4 psid, which is below the containment design external pressure load of 2.0 psid.
The Containment Spray System satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii).
Watts Bar - Unit 2                        B 3.6-36                                  (continued)
Revision 3, 15, 57 Amendment 50
 
Containment Spray System B 3.6.6 BASES (continued)
REFERENCES        1. Title 10, Code of Federal Regulations, Part 50, Appendix A, General Design Criterion (GDC) 38, Containment Heat Removal, GDC 39, Inspection of Containment Heat Removal System, GDC 40, Testing of Containment Heat Removal Systems, and GDC 50, Containment Design Basis.
: 2. NPG-SDD-WBN2-72-4001, Containment Heat Removal Spray System.
: 3. WCAP-16996-P-A, Revision 1, Realistic LOCA Evaluation Methodology Applied to the Full Spectrum of Break Sizes (FULL SPECTRUM LOCA Methodology), November 2016.
: 4. American Society of Mechanical Engineers (ASME) OM Code, Code for Operation and Maintenance of Nuclear Power Plants.
Watts Bar - Unit 2                    B 3.6-40                                Revision 57 Amendment 50
 
EGTS B 3.6.9 BASES SURVEILLANCE      SR 3.6.9.2 (continued)
REQUIREMENTS Specific test frequencies and additional information are discussed in detail in the VFTP. It should be noted that for the EGTS, the VFTP pressure drop value across the entire filtration unit does not account for instrument error.
SR 3693 The automatic startup ensures that each EGTS train responds properly.
The SR excludes automatic dampers and valves that are locked, sealed, or otherwise secured in the actuated position. The SR does not apply to dampers or valves that are locked, sealed, or otherwise secured in the actuated position since the affected dampers or valves were verified to be in the actuated position prior to being locked, sealed, or otherwise secured. Placing an automatic valve or damper in a locked, sealed, or otherwise secured position requires an assessment of the operability of the system or any supported systems, including whether it is necessary for the valve or damper to be repositioned to the non-actuated position to support the accident analysis. Restoration of an automatic valve or damper to the non-actuated position requires verification that the SR has been met within its required Frequency. This testing includes the automatic swapping logic of the EGTS pressure control isolation valves in response to the actuation signal. Performance of this swapping logic test will ensure the availability of EGTS functions in the event of an initial single failure of one of the pressure control loops. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3694 The proper functioning of the fans, dampers, filters, adsorbers, etc., as a system is verified by the ability of each train to produce the required system flow rate. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
Watts Bar - Unit 2                        B 3.6-52                                  (continued)
Revision 34,43 Amendment 36,43
 
ARS B 3.6.10 BASES BACKGROUND        purging all potential hydrogen pockets in containment. When the (continued)      containment pressure falls below a predetermined value, the ARS fans are manually de-energized. Thereafter, the fans are manually cycled on and off if necessary to control any additional containment pressure transients.
The ARS also functions, after all the ice has melted, to circulate any steam still entering the lower compartment to the upper compartment where the Containment Spray System can cool it.
The ARS is an ESF system. It is designed to ensure that the heat removal capability required during the post accident period can be attained. The operation of the ARS, in conjunction with the ice bed, the Containment Spray System, and the Residual Heat Removal (RHR)
System spray, provides the required heat removal capability to limit post accident conditions to less than the containment design values.
APPLICABLE        The limiting DBAs considered relative to containment temperature and SAFETY            pressure are the loss of coolant accident (LOCA) and the steam line ANALYSES          break (SLB). The LOCA and SLB are analyzed using computer codes designed to predict the resultant containment pressure and temperature transients. DBAs are assumed not to occur simultaneously or consecutively. The postulated DBAs are analyzed, in regard to ESF systems, assuming the loss of one ESF bus, which is the worst case single active failure and results in one train each of the Containment Spray System, RHR System, and ARS being inoperable (Ref. 1). The DBA analyses show that the maximum peak containment pressure results from the LOCA analysis and is calculated to be less than the containment design pressure.
For certain aspects of transient accident analyses, maximizing the calculated containment pressure is not conservative. In particular, the cooling effectiveness of the Emergency Core Cooling System during the core reflood phase of a LOCA analysis increases with increasing containment backpressure. For these calculations, the containment backpressure is calculated in a manner designed to conservatively minimize, rather than maximize, the calculated transient containment pressures, in accordance with WCAP-16996-P-A, Revision 1 (Ref. 2).
(continued)
Watts Bar - Unit 2                        B 3.6-55                                  Revision 57 Amendment 50
 
ARS B 3.6.10 BASES (continued)
REFERENCES        1. Watts Bar FSAR, Section 6.8, Air Return Fans.
: 2. WCAP-16996-P-A, Revision 1, Realistic LOCA Evaluation Methodology Applied to the Full Spectrum of Break Sizes (FULL SPECTRUM LOCA Methodology), November 2016.
: 3. System Description N3-30RB-4002.
Watts Bar - Unit 2                  B 3.6-58                                Revision 57 Amendment 50
 
Ice Bed B 3.6.11 BASES APPLICABLE        For these calculations, the containment backpressure is calculated in a SAFETY            manner designed to conservatively minimize, rather than maximize, the ANALYSES          calculated transient containment pressures, in accordance with WCAP-(continued)      16996-P-A, Revision 1 (Ref. 2). The maximum peak containment atmosphere temperature results from the SLB analysis and is discussed in the Bases for LCO 3.6.5, Containment Air Temperature.
In addition to calculating the overall peak containment pressures, the DBA analyses include calculation of the transient differential pressures that occur across subcompartment walls during the initial blowdown phase of the accident transient. The internal containment walls and structures are designed to withstand these local transient pressure differentials for the limiting DBAs.
The ice bed satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii).
LCO                The ice bed LCO requires the existence of the required quantity of stored ice, appropriate distribution of the ice and the ice bed, open flow paths through the ice bed, and appropriate chemical content and pH of the stored ice. The stored ice functions to absorb heat during a DBA, thereby limiting containment air temperature and pressure. The chemical content and pH of the ice provide core SDM (boron content) and remove radioactive iodine from the containment atmosphere when the melted ice is recirculated through the ECCS and the Containment Spray System, respectively.
APPLICABILITY      In MODES 1, 2, 3, and 4, a DBA could cause an increase in containment pressure and temperature requiring the operation of the ice bed.
Therefore, the LCO is applicable in MODES 1, 2, 3, and 4.
In MODES 5 and 6, the probability and consequences of these events are reduced due to the pressure and temperature limitations of these MODES. Therefore, the ice bed is not required to be OPERABLE in these MODES.
Watts Bar - Unit 2                          B 3.6-62                                (continued)
Revision 57 Amendment 50
 
Ice Bed B 3.6.11 BASES SURVEILLANCE      SR 3.6.11.7 REQUIREMENTS (continued)      This SR ensures that initial ice fill and any subsequent ice additions meet the boron concentration and pH requirements of SR 3.6.11.5. The SR is modified by a NOTE that allows the chemical analysis to be performed on either the liquid or resulting ice of each sodium tetraborate solution prepared. If ice is obtained from offsite sources, then chemical analysis data must be obtained for the ice supplied.
REFERENCES        1. Watts Bar FSAR, Section 6.2, Containment Systems and Section 6.7, Ice Condenser System.
: 2. WCAP-16996-P-A, Revision 1, Realistic LOCA Evaluation Methodology Applied to the Full Spectrum of Break Sizes (FULL SPECTRUM LOCA Methodology), November 2016.
: 3. Westinghouse Letter, WAT-D-10686, Upper Limit Ice Boron Concentration In Safety Analysis.
Watts Bar - Unit 2                        B 3.6-68                                Revision 57 Amendment 50
 
Ice Condenser Doors B 3.6.12 BASES APPLICABLE        Although the ice condenser is a passive system that requires no electrical SAFETY            power to perform its function, the Containment Spray System and ARS ANALYSES          also function to assist the ice bed in limiting pressures and temperatures.
(continued)      Therefore, the postulated DBAs are analyzed with respect to Engineered Safety Feature (ESF) systems, assuming the loss of one ESF bus, which is the worst case single active failure and results in one train each of the Containment Spray System and the ARS being rendered inoperable.
The limiting DBA analyses (Ref. 1) show that the maximum peak containment pressure results from the LOCA analysis and is calculated to be less than the containment design pressure. For certain aspects of transient accident analyses, maximizing the calculated containment pressure is not conservative. In particular, the cooling effectiveness of the ECCS during the core reflood phase of a LOCA analysis increases with increasing containment backpressure. For these calculations, the containment backpressure is calculated in a manner designed to conservatively minimize, rather than maximize, the calculated transient containment pressures, in accordance with WCAP-16996-P-A, Revision 1 (Ref. 2).
The maximum peak containment atmosphere temperature results from the SLB analysis and is discussed in the Bases for LCO 3.6.5, Containment Air Temperature.
An additional design requirement was imposed on the ice condenser door design for a small break accident in which the flow of heated air and steam is not sufficient to fully open the doors.
For this situation, the doors are designed so that all of the doors would partially open by approximately the same amount. Thus, the partially opened doors would modulate the flow so that each ice bay would receive an approximately equal fraction of the total flow.
This design feature ensures that the heated air and steam will not flow preferentially to some ice bays and deplete the ice there without utilizing the ice in the other bays.
In addition to calculating the overall peak containment pressures, the DBA analyses include the calculation of the transient differential pressures that would occur across subcompartment walls during the initial blowdown phase of the accident transient. The internal containment walls and structures are designed to withstand the local transient pressure differentials for the limiting DBAs.
The ice condenser doors satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii).
Watts Bar - Unit 2                          B 3.6-71                                (continued)
Revision 57 Amendment 50
 
Ice Condenser Doors B 3.6.12 BASES (continued)
LCO                This LCO establishes the minimum equipment requirements to assure that the ice condenser doors perform their safety function. The ice condenser inlet doors, intermediate deck doors, and top deck doors must be closed to minimize air leakage into and out of the ice condenser, with its attendant leakage of heat into the ice condenser and loss of ice through melting and sublimation. The doors must be OPERABLE to ensure the proper opening of the ice condenser in the event of a DBA.
OPERABILITY includes being free of any obstructions that would limit their opening, and for the inlet doors, being adjusted such that the opening and closing torques are within limits. The ice condenser doors function with the ice condenser to limit the pressure and temperature that could be expected following a DBA.
APPLICABILITY      In MODES 1, 2, 3, and 4, a DBA could cause an increase in containment pressure and temperature requiring the operation of the ice condenser doors. Therefore, the LCO is applicable in MODES 1, 2, 3, and 4.
The probability and consequences of these events in MODES 5 and 6 are reduced due to the pressure and temperature limitations of these MODES. Therefore, the ice condenser doors are not required to be OPERABLE in these MODES.
ACTIONS            A Note provides clarification that, for this LCO, separate Condition entry is allowed for each ice condenser door.
A.1 If one or more ice condenser inlet doors are inoperable due to being physically restrained from opening, the door(s) must be restored to OPERABLE status within 1 hour. The Required Action is necessary to return operation to within the bounds of the containment analysis. The 1 hour Completion Time is consistent with the ACTIONS of LCO 3.6.1, Containment, which requires containment to be restored to OPERABLE status within 1 hour.
(continued)
Watts Bar - Unit 2                        B 3.6-72                                Revision 50
 
Ice Condenser Doors B 3.6.12 BASES (continued)
ACTIONS            B.1 and B.2 (continued)
If one or more ice condenser doors are determined to be partially open or otherwise inoperable for reasons other than Condition A or if a door is found that is not closed, it is acceptable to continue plant operation for up to 14 days, provided the ice bed temperature instrumentation is monitored once per 4 hours to ensure that the open or inoperable door is not allowing enough air leakage to cause the maximum ice bed temperature to approach the melting point. The Frequency of 4 hours is based on the fact that temperature changes cannot occur rapidly in the ice bed because of the large mass of ice involved. The 14-day Completion Time is based on long term ice storage tests that indicate that if the temperature is maintained below 27F, there would not be a significant loss of ice from sublimation. If the maximum ice bed temperature is > 27F at any time, or ice bed temperature is not verified to be within the specified Frequency as augmented by the provisions of SR 3.0.2, the situation reverts to Condition C and a Completion Time of 48 hours is allowed to restore the inoperable door to OPERABLE status or enter into Required Actions D.1 and D.2. [NOTE: Entry into Condition B is not required due to personnel standing on or opening an intermediate deck or top deck door for short durations to perform required surveillances, minor maintenance such as ice removal, or routine tasks such as system walkdowns.]
C.1 If Required Actions or Completion Times of B.1 or B.2 is not met, the doors must be restored to OPERABLE status and closed positions within 48 hours. The 48-hour Completion Time is based on the fact that, with the very large mass of ice involved, it would not be possible for the temperature to increase to the melting point and a significant amount of ice to melt in a 48-hour period.
D.1 and D.2 If the ice condenser doors cannot be restored to OPERABLE status within the required Completion Time, 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 6 hours and to MODE 5 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.
(continued)
Watts Bar - Unit 2                        B 3.6-73                                  Revision 50
 
Ice Condenser Doors B 3.6.12 BASES SURVEILLANCE      SR 3.6.12.7 REQUIREMENTS (continued)      Verifying, by visual inspection, that the top deck doors are in place, not obstructed, and verifying free movement of the vent assembly provides assurance that the doors are performing their function of keeping warm air out of the ice condenser during normal operation, and would not be obstructed if called upon to open in response to a DBA. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
REFERENCES        1. Watts Bar FSAR, Section 6.2.1, Containment Functional Design and Section 6.7, Ice Condenser System.
: 2. WCAP-16996-P-A, Revision 1, Realistic LOCA Evaluation Methodology Applied to the Full Spectrum of Break Sizes (FULL SPECTRUM LOCA Methodology), November 2016.
Watts Bar - Unit 2                        B 3.6-77                              Revision 34, 57 Amendment 36, 50
 
Shield Building B 3.6.15 BASES APPLICABILITY      In MODES 5 and 6, the probability and consequences of these events are (continued)      low due to the Reactor Coolant System temperature and pressure limitations in these MODES. Therefore, shield building OPERABILITY is not required in MODE 5 or 6.
ACTIONS            A.1 In the event shield building OPERABILITY is not maintained, shield building OPERABILITY must be restored within 24 hours. 24 hours is a reasonable Completion Time considering the limited leakage design of containment and the low probability of a Design Basis Accident occurring during this time period.
B.1 and B.2 If the shield building cannot be restored to OPERABLE status within the required Completion Time, 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 6 hours and to MODE 5 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.
(continued)
Watts Bar - Unit 2                        B 3.6-89                            Revision 4, 45 Amendment 45
 
Shield Building B 3.6.15 BASES SURVEILLANCE      SR 3.6.15.1 REQUIREMENTS Verifying that shield building annulus negative pressure is within limit (equal to or more negative than -1 inches water gauge; value does not account for instrument error, Ref. 4) ensures that operation remains within the limit assumed in the large break loss of coolant accident dose analysis (Ref. 5). The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.6.15.2 Maintaining shield building OPERABILITY requires maintaining each door in the access opening closed, except when the access opening is being used for normal transient entry and exit. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.6.15.3 This SR would give advance indication of gross deterioration of the concrete structural integrity of the shield building. The Frequency of this SR is the same as that of SR 3.6.1.1. The verification is done during shutdown.
(continued)
Watts Bar - Unit 2                        B 3.6-90 Revision 34, 45 Amendment 36, 45
 
Shield Building B 3.6.15 BASES SURVEILLANCE      SR 3.6.15.4 REQUIREMENTS (continued)      The EGTS produces a negative pressure to prevent leakage from the building. This Surveillance verifies that the shield building can be rapidly drawn down to equal to or more negative than -0.50 inches water gauge
(" wg) in the annulus at an elevation equivalent to the top of the Auxiliary Building. This test is used to ensure shield building boundary integrity. At elevations higher than the Auxiliary Building, the EGTS is required to maintain a pressure equal to or more negative than -0.25" wg. The low pressure sense line for the pressure controller is located in the annulus at elevation 783. By verifying that the annulus pressure is equal to or more negative than -0.63" wg at elevation 783, the annulus pressurization requirements stated above are met. The ability of an EGTS train with final flow 3600 cfm and 4400 cfm, within 20 seconds after a start signal, to produce the required negative pressure during the test operation provides assurance that the building is adequately sealed. The time limit ensures that no significant quantity of radioactive material leaks from the Shield Building prior to developing the negative pressure. Upon failure to meet this SR, the leak tightness of the shield building must be immediately assessed to determine the impact on the OPERABILITY of the shield building. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
REFERENCES        1.      Deleted.
: 2.      WBN UFSAR Section 6.2.3.2.2, Emergency Gas Treatment System (EGTS).
: 3.      WBN UFSAR Section 9.4.6, Reactor Building Purge Ventilating System (RBPVS).
: 4.      TVA Calculation EPMMMA121889, Temperature Induced Differential Pressure Effects in Reactor Building and Aux. Bldg.
Secondary Containment, Revision 008.
: 5.      TVA Calculation TIRPS197, Offsite Doses Due to a Regulatory Guide 1.4 Loss of Coolant Accident, Revision 023.
Watts Bar - Unit 2                          B 3.6-91                      Revision 4, 34, 45, 53 Amendment 36, 45, 58
 
MSSVs B 3.7.1 BASES APPLICABLE        The MSSVs are assumed to have two active failure modes. The active SAFETY            failure modes are spurious opening, and failure to reclose once opened.
ANALYSES (continued)      The MSSVs satisfy Criterion 3 of 10 CFR 50.36(c){2){ii}.
LCO                The accident analysis requires that five MSSVs per steam generator be OPERABLE to provide overpressure protection for design basis transients occurring at 100.6% RTP. The LCO requires that five MSSVs per steam generator be OPERABLE in compliance with Reference 2 and the DBA analysis.
The OPERABILITY of the MSSVs is defined as the ability to open upon demand within the setpoint tolerances to relieve steam generator overpressure, and reseat when pressure has been reduced. The OPERABILITY of the MSSVs is determined by periodic surveillance testing in accordance with the Inservice Testing Program.
This LCO provides assurance that the MSSVs will perform their designed safety functions to mitigate the consequences of accidents that could result in a challenge to the RCPB, or Main Steam System integrity.
APPLICABILITY      In MODES 1, 2, and 3, five MSSVs per steam generator are required to be OPERABLE to prevent Main Steam System overpressurization.
In MODES 4 and 5, there are no credible transients requiring the MSSVs.
The steam generators are not normally used for heat removal in MODES 5 and 6, and thus cannot be overpressurized; there is no requirement for the MSSVs to be OPERABLE in these MODES.
ACTIONS            The ACTIONS table is modified by a Note indicating that separate Condition entry is allowed for each MSSV.
With one or more MSSVs inoperable, action must be taken so that the available MSSV relieving capacity meets Reference 2 requirements.
(continued)
Watts Bar - Unit 2                        B 3.7-3                              Revision 42 Amendment 42
 
AFW System B 3.7.5 BASES BACKGROUND        The AFW System is designed to supply sufficient water to the steam (continued)      generator(s) to remove decay heat with steam generator pressure at the lowest MSSV setpoint (plus 3% tolerance plus 7 psi for accumulation and pressure drop between the SG and MSSV)of the MSSVs. Subsequently, the AFW System supplies sufficient water to cool the unit to RHR entry conditions, with steam released through the ADVs.
The AFW System actuates automatically on steam generator water level low-low by the ESFAS(LCO 3.3.2). The motor driven pumps start on a two-out-of-three low-low level signal in any steam generator and the turbine driven pump starts on a two-out-of-three low-low level signal in any two steam generators. The system also actuates on loss of offsite power, safety injection, and trip of both turbine-driven MFW pumps.
The AFW System is discussed in the FSAR, Section 10.4.9 (Ref. 1).
APPLICABLE        The AFW System mitigates the consequences of any event with loss of SAFETY            normal feedwater.
ANALYSES The design basis of the AFW System is to supply water to the steam generator to remove decay heat and other residual heat by delivering at least the minimum required flow rate to the steam generators at pressures corresponding to the lowest steam generator safety valve set pressure plus 3% tolerance plus 7 psi for accumulation and pressure drop between the SG and MSSV.
In addition, the AFW System must supply enough makeup water to replace steam generator secondary inventory lost as the unit cools to MODE 4 conditions. Sufficient AFW flow must also be available to account for flow losses such as pump recirculation and line breaks.
The limiting Design Basis Accidents (DBAs)and transients for the AFW System are as follows:
: a. Feedwater Line Break (FWLB); and
: b. Loss of MFW.
In addition, the minimum available AFW flow and system characteristics are serious considerations in the analysis of a small break loss of coolant accident(LOCA).
(continued)
Watts Bar - Unit 2                      B 3.7-24                                  Revision 44
 
CST B 3.7.6 BASES APPLICABLE        The limiting event for the condensate volume is the large feedwater line SAFETY            break coincident with a loss of offsite power. Single failures that also ANALYSES          affect this event include the following:
(continued)
: a. Failure of the diesel generator powering the motor driven AFW pump to the unaffected steam generators (requiring additional steam to drive the remaining AFW pump turbine); and
: b. Failure of the steam driven AFW pump (requiring a longer time for cooldown using only one motor driven AFW pump).
These are not usually the limiting failures in terms of consequences for these events.
A non-limiting event considered in CST inventory determinations is a break in either the main feedwater bypass line or AFW line near where the two join. This break has the potential for dumping condensate until terminated by operator action. This loss of condensate inventory is partially compensated for by the retention of steam generator inventory.
Because the CST is the preferred source of feedwater and is relied on almost exclusively for accidents and transients, the CST satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii).
LCO                As the preferred water source to satisfy accident analysis assumptions, the CST must contain sufficient cooling water to remove decay heat for 2 hours following a reactor trip from 100.6% RTP, and then to cool down the RCS to RHR entry conditions, assuming a coincident loss of offsite power and the most adverse single failure. In doing this, it must retain sufficient water to ensure adequate net positive suction head for the AFW pumps during cooldown, as well as account for any losses from the steam driven AFW pump turbine, or before isolating AFW to a broken line.
The CST level required is equivalent to a usable volume of 200,000 gallons, which is based on holding the unit in MOUE 3 for 2 hours, followed by a cooldown to RHR entry conditions at 50&deg;F/hour.
This basis is established in Reference 4 and exceeds the volume required by the accident analysis.
The OPERABILITY of the CST is determined by maintaining the tank level at or above the minimum required level.
Watts Bar - Unit 2                        B 3.7-33                                  (continued)
Revision 42 Amendment 42
 
CREVS B 3.7.10 BASES ACTIONS            H_1 (continued)
If both CREVS trains are inoperable in MODE 1, 2, 3, or 4, for reasons other than Condition B or Condition E the CREVS may not be capable of performing the intended function and the plant is in a condition outside the accident analyses. Therefore, LCO 3.0.3 must be entered immediately.
SURVEILLANCE      SR 3.7.10.1 REQUIREMENTS Standby systems should be checked periodically to ensure that they function properly. As the environment and normal operating conditions on this system are not too severe, testing each train provides an adequate check of this system. The systems need only be operated for
                  > 15 minutes to demonstrate the function of the system. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.7.10.2 This SR verifies that the required CREVS testing is performed in accordance with the Ventilation Filter Testing Program (VFTP). The CREVS filter tests are in accordance with Regulatory Guide 1.52 (Ref. 6).
The VFTP includes testing the performance of the HEPA filter, charcoal adsorber efficiency, minimum flow rate, and the physical properties of the activated charcoal. Specific test Frequencies and additional information are discussed in detail in the VFTP.
SR 3.7.10.3 This SR verifies that each CREVS train starts and operates on an actual or simulated actuation signal. The SR excludes automatic dampers and valves that are locked, sealed, or otherwise secured in the actuated position. The SR does not apply to dampers or valves that are locked, sealed, or otherwise secured in the actuated position since the affected dampers or valves were verified to be in the actuated position prior to being locked, sealed, or otherwise secured. Placing an automatic valve or damper in a locked, sealed, or otherwise secured position requires an assessment of the operability of the system or any supported systems, including whether it is necessary for the valve or damper to be repositioned to the non-actuated position to support the accident analysis.
Restoration of an automatic valve or damper to the non-actuated position requires verification that the SR has been met within its required Frequency. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
continued Watts Bar - Unit 2                        B 3.7-56                          Revision 9, 34,43 Amendment 9, 36,43
 
CREATCS B 3.7.11 ACTIONS            A_1 (continued)
The Completion Time is modified by a footnote that states an allowance is permitted for one CREATCS train to be inoperable for 60 days. This TS provision is only authorized for one entry per train during modification activities planned for the upgrade of the main control room chillers beginning no earlier than May 1, 2022, and ending no later than May 1, 2023, provided the following compensatory measures are implemented as described in TVA letter CNL-20-012, dater{ May 19, 2020.
A temporary, non-safety related chiller system with a temporary G to provide power to the temporary chiller system will be installed and operated as described in the LAR.
Instructions for operation of the temporary cooling equipment will be provided.
During replacement of the CREATCS chillers, TVA will employ a graded approach to defense-in-depth and protected equipment strategies based on the operating status of the affected unit. The risk of the activity will be assessed and managed, including the use of physical barriers as needed. Additionally, TVA procedures preclude work on or near protected equipment and limit access to the area to emergency situations and non-intrusive monitoring of running equipment per operator rounds.
* During replacement of the CREATCS chillers, no elective maintenance will be performed on TS related support equipment for the Operable CREATCS chiller except for any required TS S Rs.
B.1 and B.2 In MODE 1, 2, 3, or 4, if the inoperable CREATCS train cannot be restored to OPERABLE status within the required Completion Time, the plant must be placed in a MODE that minimizes the risk. To achieve this status, the plant must be placed in at least MODE 3 within 6 hours, and in MODE 5 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.
continued Watts Bar - Unit 2                          B 3.7-61                                Revision 47 Amendment 51
 
CREATCS B 3.7.11 ACTIONS            C.1 and C.2 (conti nued)
In MODE 5 or 6, or during movement of irradiated fuel, if the inoperable CREATCS train cannot be restored to OPERABLE status within the required Completion Time, the OPERABLE CREATCS train must be placed in operation immediately. This action ensures that the remaining train is OPERABLE, that no failures preventing automatic actuation will occur, and that active failures will be readily detected.
An alternative to Required Action C.1 is to immediately suspend activities that present a potential for releasing radioactivity that might require isolation of the control room. This places the unit in a condition that minimizes accident risk. This does not preclude the movement of fuel to a safe position.
D.1 In MODE 5 or 6, or during movement of irradiated fuel assemblies, with two CREATCS trains inoperable, action must be taken immediately to suspend activities that could result in a release of radioactivity that might require isolation of the control room. This places the unit in a condition that minimizes risk_ This does not preclude the movement of fuel to a safe position.
E.1 If both CREATCS trains are inoperable in MODE 1, 2, 3, or 4, the CREATCS may not be capable of performing its intended function.
Therefore, LCO 3.0.3 must be entered immediately. The Completion Time is modified by a footnote that states an allowance to monitor the main control room temperature every hour and verify the main control room temperature is less than or equal to 90OF is permitted for up to four days in lieu of the immediate entry into LCO 3.0.3. If the main control room temperature exceeds 90&deg;F, or the duration without a train of CREATCS being OPERABLE exceeds four days, immediate entry into LCO 3.0.3 is required. This provision is only applicable during modification activities planned for the upgrade of the main control room chillers beginning no earlier than May 1, 2022, and ending no later than May 1, 2023, provided the following compensatory measures are implemented as described in TVA letter CNL-20-012, dated May 19, 2020.
continued Watts Bar - Unit 2                        B 3.7-62                                  Revision 47 Amendment 51
 
CREAMS B 3.7.11 BASES (continued)
ACTIONS            E_1 (continued)
A temporary, non-safety related chiller system with a temporary G to provide power to the temporary chiller system will be installed and operated as described in the LAR.
Instructions for operation of the temporary cooling equipment will be provided.
During replacement of the CREAMS chillers, TVA will employ a graded approach to defense-in-depth and protected equipment strategies based on the operating status of the affected unit. The risk of the activity will be assessed and managed, including the use of physical barriers as needed. Additionally, TVA procedures preclude work on or near protected equipment and limit access to the area to emergency situations and non-intrusive monitoring of running equipment per operator rounds.
During replacement of the CREAMS chillers, no elective maintenance will be performed on TS related support equipment for the Operable CREATCS chiller except for any required TS SRs.
The purpose of the footnote is to ensure the MCR temperature is being controlled. The specified temperature limit of 90OF is above the normal operating temperature of the MCR (approximately 75&deg;F), providing operational flexibility when implementing the mitigating actions. This temperature does not impact the operability of equipment or habitability of the MCR. The limit of 90&deg;F maintains margin below the lowest specification for the MCR equipment cabinets of 1040F. Subsequent to immediate MCR temperature verification, the one-hour frequency is adequate given the indications available in the MCR. Main control room temperature data is measured and displayed from readily available equipment in the MCR and operators will have awareness of temperature trending relative to the 90OF limit.
SURVEILLANCE      SR 3.7.11.1 REQUIREMENTS This SR verifies that the heat removal capability of the system is sufficient to remove the heat load assumed in the sizing calculations in the control room. This SR consists of a combination of testing and calculations. This is accomplished by verifying that the system has not degraded. The only measurable parameters that could degrade undetected during normal operation are the system air flow and chilled water flow rate. Verification of these two flow rates will provide assurance that the heat removal capacity of the system is still adequate. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
Watts Bar - Unit 2                        B 3.7-62a                            Revision 34, 47 Amendment 36, 51
 
CREAMS B 3.7.11 BASES (continued)
REFERENCES            Watts Bar FSAR, Section 9.4.1, "Control Room Area Ventilation System."
: 2. Watts Bar FSAR, Section 3.7.3.18, "Seismic Qualification of Main Control Room Suspended Ceiling and Air Delivery Components."
: 3. NRC Safety Evaluation dated February 12, 2004, for License Amendment 50.
Watts Bar - Unit 2                B 3.7-62b                                Revision 47 Amendment 51
 
ABGTS B 3.7.12 BASES SURVEILLANCE      SR 3.7.12.3 REQUIREMENTS (continued)      This SR verifies that each ABGTS train starts and operates on an actual or simulated actuation signal. The SR excludes automatic dampers and valves that are locked, sealed, or otherwise secured in the actuated position. The SR does not apply to dampers or valves that are locked, sealed, or otherwise secured in the actuated position since the affected dampers or valves were verified to be in the actuated position prior to being locked, sealed, or otherwise secured. Placing an automatic valve or damper in a locked, sealed, or otherwise secured position requires an assessment of the operability of the system or any supported systems, including whether it is necessary for the valve or damper to be repositioned to the non-actuated position to support the accident analysis.
Restoration of an automatic valve or damper to the non-actuated position requires verification that the SR has been met within its required Frequency. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
SR 3.7.12.4 This SR verifies the integrity of the ABSCE. The ability of the ABSCE to maintain negative pressure with respect to potentially uncontaminated adjacent areas is periodically tested to verify proper function of the ABGTS. During the post accident mode of operation, the ABGTS is designed to maintain a slight negative pressure in the ABSCE, to prevent unfiltered LEAKAGE. The ABGTS is designed to maintain a negative pressure between -0.25 inches water gauge and -0.5 inches water gauge (value does not account for instrument error) with respect to atmospheric pressure at a nominal flow rate ? 9300 cfm and < 9900 cfm. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
REFERENCES          1. Watts Bar FSAR, Section 6.5.1, "Engineered Safety Feature (ESF)
Filter Systems."
: 2. Watts Bar FSAR, Section 9.4.2,"Fuel Handling Area Ventilation System."
: 3. Watts Bar FSAR, Section 15.0, "Accident Analysis."
: 4. Watts Bar FSAR, Section 6-2.3, "Secondary Containment Functional Design."
(continued)
Watts Bar - Unit 2                        B 3.7-67 Revision 34,43 Amendment 36,43
 
Spent Fuel Assembly Storage B 3.7.15 BASES (continued)
B 3.7 PLANT SYSTEMS B 3.7.15 Spent Fuel Pool Assembly Storage BASES BACKGROUND          The spent fuel pool contains flux trap rack modules with 1386 storage positions that are designed to accommodate fuel with a maximum enrichment of 4.95 +/- 0.05 weight percent U-235 without restrictions.
The water in the spent fuel storage pool normally contains soluble boron, which results in large subcriticality margins under actual operating conditions. The double contingency principle discussed in ANSI N-16.1-1975, and the April 1978 NRC letter (Reference 1) allows credit for soluble boron under other abnormal or accident conditions, since only a single accident need be considered at one time. To mitigate postulated criticality-related events, boron is dissolved in the pool water.
APPLICABLE          The accident analyses are provided in the FSAR.
SAFETY ANALYSES            The initial enrichment of fuel assemblies in the fuel storage pool along with the concentration of dissolved boron in the fuel storage pool satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).
LCO                  The restrictions on the initial enrichment of fuel assemblies within the spent fuel pool in accordance with Specification 4.3.1.1 in the accompanying LCO, ensures the kerf will always remain subcritical, assuming the pool to be flooded with unborated water.
APPLICABILITY        This LCO applies whenever any fuel assembly is stored in the spent fuel storage pool.
(continued)
Watts Bar - Unit 2                          B 3.7-74                                  Revision 41 Amendment 27
 
Spent Fuel Assembly Storage B 3.7.15 BASES (continued)
ACTIONS            A_1 Required Action A.1 is modified by a Note indicating that LCO 3.0.3 does not apply.
If unable to move irradiated fuel assemblies while in MODE 5 or 6, LCO 3.0.3 would not be applicable. If unable to move irradiated fuel assemblies while in MODE 1, 2, 3, or 4, the action is independent of reactor operation. Therefore, inability to move fuel assemblies is not sufficient reason to require a reactor shutdown.
When the initial enrichment of fuel assemblies stored in the spent fuel storage pool is not in accordance with Specification 4.3.1.1, the immediate action is to initiate action to make the necessary fuel assembly movements to bring the configuration into compliance with Specification 4.3.1.1.
SURVEILLANCE      SR 3.7.15.1 REQUIREMENTS This SR verifies by administrative means that the initial enrichment of the fuel assembly is in accordance with Specification 4.3.1.1 in the accompanying LCO.
REFERENCES        1. Double contingency principle of ANSI N16.1-1975, as specified in the April 14, 1978, NRC letter {Section 1.2} and implied in the proposed revision to Regulatory Guide 1.13 (Section 1.4, Appendix A).
: 2. FSAR Section 4.3.2.7.
Watts Bar - Unit 2                        B 3.7-75                                  Revision 41 Amendment 27
 
Spent Fuel Assembly Storage B 3.7.15 BASES (continued)
INTENTIONALLY LEFT BLANK Wafts Bar - Unit 2          B 3.7-76                        Revision 41 Amendment 27
 
Fuel Storage Pool Boron Concentration B 3.7.18 B 3.7 PLANT SYSTEMS B 3.7.18 Fuel Storage Pool Boron Concentration BASES BACKGROUND          In the BORALTm flux trap rack design, the spent fuel storage pool is designed to accommodate new fuel with a maximum enrichment of 4.95 +/- 0.05 wt % U-235, or spent fuel regardless of the discharge fuel burnup.
The water in the spent fuel storage pool normally contains soluble boron, which results in large subcriticality margins under actual operating conditions. Analysis demonstrates that the effective neutron multiplication factor (keff) of the spent fuel pool loaded with fuel of the highest anticipated reactivity, at a temperature corresponding to the highest reactivity, is less than 1.9 for the pool flooded with unborated water, and does not exceed 0.95 for the pool flooded with borated water with 500 ppm soluble boron (an additional 50 ppm of soluble boron has been added to account for grid growth.) The double contingency principle discussed in ANSI N-15.1-1975 and the April 1978 NRC letter (Ref. 1) allows credit for soluble boron under other abnormal or accident conditions, since only a single accident need be considered at one time.
To mitigate postulated criticality related events, boron is dissolved in the pool water.
APPLICABLE          The following accident conditions have been evaluated:
SAFETY                  . The effect of SFP temperature exceeding the normal range ANALYSES                . A dropped fuel assembly A misloaded fuel assembly (a fuel assembly in the wrong location within the storage rack)
A mislocated fuel assembly (a fuel assembly in the wrong location outside the storage rack)
* Rack movement due to seismic activity The results of the evaluation show that a soluble boron concentration of 500 ppm is sufficient to ensure that the maximum ker, is below the regulatory limit of 0.95. The boron dilution analysis assumes an initial boron concentration of 2,300 ppm for the limiting evaluation. The accident analyses are provided in the FSAR, Section 4.3.2.7 (Ref. 2)
The concentration of dissolved boron in the fuel storage pool satisfies Criterion 2 of 10 CFR 59.36(c)(2)(ii).
(continued)
Watts Bar-Unit 2                              B 3.7-89                                  Revision 41 Amendment 27
 
Fuel Storage Pool Boron Concentration B 3.7.18 BASES (continued)
LCO              The fuel storage pool boron concentration is required to be X2300 ppm.
The specified concentration of dissolved boron in the fuel storage pool preserves the assumptions used in the analyses. This concentration of dissolved boron is the minimum required concentration for fuel assembly storage and movement within the fuel storage pool.
APPLICABILITY    This LCO applies whenever fuel assemblies are stored in the spent fuel storage pool.
ACTIONS          A_1 The Required Actions are modified by a Note indicating that LCO 3.0.3 does not apply.
When the concentration of boron in the fuel storage pool is less than required, immediate action must be taken to preclude the occurrence of an accident or to mitigate the consequences of an accident in progress.
This is most efficiently achieved by immediately suspending the movement of fuel assemblies. The concentration of boron is restored simultaneously with suspending movement of fuel assemblies. However, prior to resuming movement of the fuel assemblies, the concentration of boron must be restored. This does not preclude movement of a fuel assembly to a safe position.
If the LCO is not met while moving irradiated fuel assemblies in MODE 5 or B, LCO 3.0.3 would not be applicable. If moving irradiated fuel assemblies while in MODE 1, 2, 3, or 4, the fuel movement is independent of reactor operation. Therefore, inability to suspend movement of fuel assemblies is not sufficient reason to require a reactor shutdown.
(continued)
Watts Bar-Unit 2                          B 3.7-90                                Revision 41 Amendment 27
 
Fuel Storage Pool Boron Concentration B 3.7.18 BASES (continued)
SURVEILLANCE      SR 3.7.18.1 REQUIREMENTS This SR verifies that the concentration of boron in the fuel storage pool is within the required limit. As long as this SR is met, the analyzed accidents are fully addressed. The 72 hour Frequency is appropriate because no major replenishment of pool water is expected to take place over such a short period of time.
REFERENCES        1. Double contingency principle of ANSI N15.1-1975, as specified in the April 14, 1978 NRC letter (Section 1.2) and implied in the proposed revision to Regulatory Guide 1.13(Section 1.4, Appendix A).
: 2. FSAR, Section 4.3.2.7 Watts Bar-Unit 2                          B 3.7-91                                Revision 41 Amendment 27
 
AC Sources - Operating B 3.8.1 BASES APPLICABLE        When credited, the available 6.9 kV FLEX DG must be able to connect to SAFETY            a shutdown board that will power the necessary components to facilitate ANALYSES          unit shutdown and cooldown. Typically, the FLEX DG would be available to be aligned to the shutdown board associated with the inoperable DG.
However, if the DG is inoperable because the associated 6.9 kV Shutdown Board is out of service (e.g., for shutdown board cleaning),
then the available FLEX DG must be able to align to an available shutdown board to ensure an entire train of ESF systems remains available to maintain plant safety.
The initial conditions of DBA and transient analyses in the FSAR, Section 6 (Ref. 4) and Section 15 (Ref. 5), assume ESF systems are OPERABLE. The AC electrical power sources are designed to provide sufficient capacity, capability, redundancy, and reliability to ensure the availability of necessary power to ESF systems so that the fuel, Reactor Coolant System (RCS), and containment design limits are not exceeded.
These limits are discussed in more detail in the Bases for Section 3.2, Power Distribution Limits; Section 3.4, Reactor Coolant System (RCS);
and Section 3.6, Containment Systems.
The OPERABILITY of the AC electrical power sources is consistent with the initial assumptions of the Accident analyses and is based upon meeting the design basis of the plant. This results in maintaining at least two DGs associated with one load group or one offsite circuit OPERABLE during Accident conditions in the event of:
: a. An assumed loss of all offsite power or all onsite AC power; and
: b. A worst case single failure.
The AC sources satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii).
LCO                Two qualified circuits between the Watts Bar Hydro 161 kV switchyard and the onsite Class 1E Electrical Power System and separate and independent DGs for each train ensure availability of the required power to shut down the reactor and maintain it in a safe shutdown condition after an anticipated operational occurrence (AOO) or a postulated DBA.
Qualified offsite circuits are those that are described in the FSAR and are part of the licensing basis for the plant.
Each offsite circuit must be capable of maintaining acceptable frequency and voltage, and accepting required loads during an accident, while connected to the 6.9 kV shutdown boards.
(continued)
Watts Bar - Unit 2                          B 3.8-3                                  Revision 54
 
AC Sources - Operating B 3.8.1 BASES LCO                      2) From the 161 kV Watts Bar Hydro Switchyard (Bay 4), through (continued)                  CSST A (winding Y) to 6.9 kV Unit Board 1C and 6.9 kV Unit Board 2C (Breakers 1524 and 1534 closed) to 6.9 kV Shutdown Board 1B-B and 6.9 kV Shutdown Board 2B-B (Breakers 1726 and 1826 closed), respectively.
Note: When using either CSST A or B as a qualified offsite circuit, the CSST (A or B) not in use as a qualified circuit must be available.
Each DG must be capable of starting, accelerating to rated speed and voltage, and connecting to its respective 6.9 kV shutdown board on detection of loss-of-voltage. This will be accomplished within 10 seconds.
Each DG must also be capable of accepting required loads within the assumed loading sequence intervals, and continue to operate until offsite power can be restored to the 6.9 kV shutdown boards. These capabilities are required to be met from a variety of initial conditions such as DG in standby with the engine hot and DG in standby with the engine at ambient conditions. Additional DG capabilities must be demonstrated to meet required Surveillances, e.g., capability of the DG to revert to standby status on an accident signal while operating in parallel test mode.
Proper sequencing of loads, including tripping of non-essential loads, is a required function for DG OPERABILITY.
The AC sources in one train must be separate and independent (to the extent possible) of the AC sources in the other train. For the DGs, separation and independence are complete. However, CSST A or B can only supply two 6.9 kV shutdown boards in the same load group to ensure the separation criteria is met.
For the offsite AC sources, separation and independence are to the extent practical. A circuit may be connected to more than one ESF bus, with fast transfer capability to the other circuit OPERABLE, and not violate separation criteria. A circuit that is not connected to an ESF bus is required to have OPERABLE fast transfer interlock mechanisms to at least two ESF buses to support OPERABILITY of that circuit.
Note: The offsite power configurations identified above for normal operation and alternate operation are mutually exclusive with respect to OPERABILITY of offsite circuits. For example, for normal operation, if 6.9 kV Shutdown Board 1A-A is out of service for planned maintenance (e.g.,
shutdown board cleaning), then the offsite circuit defined as from the 161 kV Watts Bar Hydro Switchyard through CSST C to 6.9 kV Shutdown Board 1A-A and to 6.9 kV Shutdown Board 2A-A would be inoperable because the normally closed circuit breaker connecting 6.9 kV Shutdown Board 1A-A would be open. Additionally, the normally open circuit Watts Bar - Unit 2                        B 3.8-6                                    (continued)
Revision 54
 
AC Sources - Operating B 3.8.1 BASES LCO                breaker connecting CSST D to 6.9 kV Shutdown Board 1A-A, which is (continued)      credited for alternate operation, would also remain open. Alternate operation would not be available for the offsite circuit associated with CSST D. However, for normal operation, the offsite circuit defined as the 161 kV Watts Bar Hydro Switchyard, through CSST D to 6.9 kV Shutdown Board 1B-B and to 6.9 kV Shutdown Board 2B-B would still be operatable. Therefore, for Unit 1 shutdown board cleaning, only one offsite circuit would be considered inoperable (and Condition D would be entered).
APPLICABILITY      The AC sources are required to be OPERABLE in MODES 1, 2, 3, and 4 to ensure that:
: a. Acceptable fuel design limits and reactor coolant pressure boundary limits are not exceeded as a result of AOOs or abnormal transients; and
: b. Adequate core cooling is provided, and containment OPERABILITY and other vital functions are maintained in the event of a postulated DBA.
The AC power requirements for MODES 5 and 6 are covered in LCO 3.8.2, AC Sources - Shutdown.
Watts Bar - Unit 2                        B 3.8-7                                  (continued)
Revision 54
 
AC Sources - Operating B 3.8.1 BASES ACTIONS            B.2 (continued) 6.9 kV FLEX DG availability requires that:
: 1) 6.9 kV FLEX DG fuel tank level is verified locally to be  8-hour supply; and
: 2) 6.9 kV FLEX DG supporting system parameters for starting and operating are verified to be within required limits for functional availability (e.g., batter state of charge).
The 6.9 kV FLEX DG is not used to extend the Completion Time for more than one inoperable DG at any one time.
If the inoperable DG is inoperable because the associated Unit 1 6.9 kV Shutdown Board is out of service for planned maintenance (e.g.,
shutdown board cleaning), then the available FLEX DG must be able to align to an available shutdown board to ensure an entire train of ESF systems remains available to maintain plant safety and to support the ability to shutdown and cooldown both units.
Note that entry into both Conditions B and D would be appropriate for planned maintenance such as shutdown board cleaning.
B.3 and C.2 Required Actions B.3 and C.2 are intended to provide assurance that a loss of offsite power, during the period that a DG is inoperable, does not result in a complete loss of safety function of critical systems. These features are designed with redundant safety related trains. This includes motor driven auxiliary feedwater pumps. Single train systems, such as the turbine driven auxiliary feedwater pump, are not included. Redundant required feature failures consist of inoperable features associated with a train, redundant to the train that has inoperable DG(s).
The Completion Time for Required Actions B.3 and C.2 are 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. In this Required Action, the Completion Time only begins on discovery that both:
: a.      An inoperable DG exists; and
: b.      A required feature on the other train (Train A or Train B) is inoperable.
(continued)
Watts Bar - Unit 2                          B 3.8-10a                              Revision 5, 54 Amendment 5
 
AC Sources - Operating B 3.8.1 BASES ACTIONS            C.4 (continued)
The second Completion Time for Required Action C.4 establishes a limit on the maximum time allowed for any combination of required AC power sources to be inoperable during any single contiguous occurrence of failing to meet the LCO. If Condition C is entered while, for instance, an offsite circuit is inoperable and that circuit is subsequently restored OPERABLE, the LCO may already have been not met for up to 72 hours.
This could lead to a total of 144 hours, since initial failure to meet the LCO, to restore the DGs. At this time, an offsite circuit could again become inoperable, the DGs restored OPERABLE, and an additional 72 hours (for a total of 9 days) allowed prior to complete restoration of the LCO. The 6 day Completion Time provides a limit on time allowed in a specified condition after discovery of failure to meet the LCO. This limit is considered reasonable for situations in which Conditions A and C are entered concurrently. The AND connector between the 72 hour and 6 day Completion Times means that both Completion Times apply simultaneously, and the more restrictive Completion Time must be met.
As in Required Action C.2, the Completion Time allows for an exception to the normal time zero for beginning the allowed outage time clock.
This will result in establishing the time zero at the time that the LCO was initially not met, instead of at the time Condition C was entered.
D.1, D.2, and D.3 Condition D is modified by two notes that limit the conditions that allow entry into Condition D. The first note states that Condition D is only applicable during planned maintenance. This will allow the plant configuration to be aligned to minimize features being inoperable when the opposite unit shutdown board is made inoperable. The second note limits the applicability of Condition D to the time period when the opposite unit is defueled. This note limits the time period allowing Condition D to be entered, minimizing when the allowance can be utilized.
Note that entry into both Conditions B and D would be appropriate for planned maintenance such as shutdown board cleaning.
Condition D is entered for an offsite circuit inoperable solely due to an inoperable power source to 6.9 kV Shutdown Board 1A-A or 1B-B.
Required Action D.1 verifies the OPERABILITY of the remaining offsite circuit within an hour of the inoperability and every 8 hours thereafter.
Since the Required Action only specifies "perform," a failure of the SR 3.8.1.1 acceptance criteria does not result in a Required Action not met.
(continued)
Watts Bar - Unit 2                        B 3.8-12a                                Revision 5, 54 Amendment 5
 
AC Sources - Operating B 3.8.1 BASES ACTIONS            F.1 and F.2 (continued)
Note that if an offsite circuit and a DG are inoperable because the associated Unit 1 6.9 kV Shutdown board is out of service for planned maintenance (e.g., shutdown board cleaning), then entry into both Conditions B and D is appropriate and a longer Completion Time (i.e., a maximum of 7 days instead of 12 hours) is justified because Unit 1 is defueled and Condition E of LCO 3.8.9 is entered simultaneously, which results in the associated required down stream fed subsystems being declared inoperable immediately so that the subsystems will be governed by their own LCOs (Ref. 11).
G.1 and G.2 With one or more DG(s) in Train A inoperable simultaneous with one or more DG(s) in Train B inoperable, there are no remaining standby AC sources. Thus, with an assumed loss of offsite electrical power, insufficient standby AC sources are available to power the minimum required ESF functions. Since the offsite electrical power system is the only source of AC power for this level of degradation, the risk associated with continued operation for a very short time could be less than that associated with an immediate controlled shutdown (the immediate shutdown could cause grid instability, which could result in a total loss of AC power). Since any inadvertent generator trip could also result in a total loss of offsite AC power, however, the time allowed for continued operation is severely restricted. The intent here is to avoid the risk associated with an immediate controlled shutdown and to minimize the risk associated with this level of degradation.
According to Reference 6, with one or more DG(s) in Train A inoperable simultaneous with one or more DG(s) in Train B inoperable, operation may continue for a period that should not exceed 2 hours.
H.1 and H.2 If the inoperable AC electric power sources cannot be restored to OPERABLE status within the required Completion Time, 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 6 hours and to MODE 5 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.
(continued)
Watts Bar - Unit 2                        B 3.8-16                          Revision 5, 28, 54 Amendment 5, 32
 
AC Sources - Operating B 3.8.1 BASES REFERENCES          3. Regulatory Guide 1.9, Rev. 3, Selection, Design, Qualification and (continued)          Testing of Emergency Diesel Generator Units Used as Class 1E Onsite Electric Power Systems at Nuclear Power Plants, July 1993.
: 4. Watts Bar FSAR Section 6, Engineered Safety Features.
: 5. Watts Bar FSAR, Section 15.4, Condition IV-Limiting Faults.
: 6. Regulatory Guide 1.93, Rev. 0, Availability of Electric Power Sources, December 1974.
: 7. Generic Letter 84-15, Proposed Staff Actions to Improve and Maintain Diesel Generator Reliability, July 2, 1984.
: 8. Title 10, Code of Federal Regulations, Part 50, Appendix A, GDC 18, Inspection and Testing of Electric Power Systems.
: 9. Regulatory Guide 1.137, Rev. 1, Fuel Oil Systems for Standby Diesel Generators, October 1979.
: 10. IEEE-308-1971, IEEE Standard Criteria for Class 1E Power Systems for Nuclear Power Generating Stations, Institute of Electrical and Electronic Engineers.
: 11. Letter from Kimberly J. Green (NRC) to Mr. James Barstow (TVA) dated November 26, 2019, with
 
==Enclosures:==
(1) Amendment No.
129 to Facility Operating License No. NPF-90, (2) Amendment No.
32 to Facility Operating License No. NPF-96, and (3) NRC Safety Evaluation.
Watts Bar - Unit 2                      B 3.8-37                                Revision 54
 
Distribution Systems - Operating B 3.8.9 Table B 3.8.9-1 (page 1 of 1)
AC and DC Electrical Power Distribution Systems TYPE        VOLTAGE                  TRAIN A*                              TRAIN B*
AC safety        6900 V        Shutdown Board 1A-A, 2A-A            Shutdown Board 1B-B, 2B-B buses 480 V      Shutdown Board 1A1-A, 1A2-A          Shutdown Board 1B1-B, 1B2-B 2A1-A, 2A2-A                          2B1-B, 2B2-B Rx MOV Board 1A1-A**, 1A2-A          Rx MOV Board 1B1-B**, 1B2-B 2A1-A, 2A2-A                          2B1-B, 2B2-B C & A Vent Board 1A1-A, 2A1-A          C & A Vent Board 1B1-B, 2B1-B Rx Vent Board 1B-B**, 2B-B Rx Vent Board 1A-A**, 2A-A AC vital        120 V                  Channel I                            Channel II buses Vital bus 1-I                          Vital bus 1-II Vital bus 2-I                          Vital bus 2-II Channel III                          Channel IV Vital bus 1-III                      Vital bus 1-IV Vital bus 2-III                      Vital bus 2-IV DC buses        125 V                    Board I                              Board II Board III                            Board IV
* Each train of the AC and DC electrical power distribution systems is a subsystem.
**  For WBN Unit 2, the 480V Reactor MOV Boards 1A1-A and 1B1-B and 480V Reactor Vent Boards 1A-A and 1B-B are available for economic and operational convenience. The boards contain no Unit 2 Technical Specification (TS) Required loads. The boards are considered part of the Unit 1 / Unit 2 Electrical Power Distribution System and meet Unit 2 TS Requirements and testing only while connected. WBN Unit 2 is designed to be operated, shutdown, and maintained in a safe shutdown status without any of these boards or their loads. As such, the boards may be disconnected from service without entering a Unit 2 LCO provided their loads are not substituting for a Unit 2 TS required load.
Watts Bar - Unit 2                          B 3.8-94a                        Revision 1, 23, 35, 55 Amendment 29
 
Spent Fuel Pool Boron Concentration B 3.9.9 BASES (continued)
B 3.9 REFUELING OPERATIONS B 3.9.9 Spent Fuel Pool Boron Concentration BASES BACKGROUN          The spent fuel storage rack criticality analysis assumes 2390 ppm soluble boron in the fuel pool when fuel is being stored.
APPLICABLE          This requirement ensures the presence of at least 2300 ppm soluble SAFETY              boron in the spent fuel pool water as assumed in the spent fuel rack ANALYSES            criticality analysis for normal storage and a dropped fuel assembly event.
The RCS boron concentration satisfies Criterion 2 of 10 CFR 50.36(c){2}(ii).
LCO                  The LCD requires that the boron concentration in the spent fuel pool be greater than or equal to 2300 ppm anytime fuel is being stored in the pool.
APPLICABILITY        This LCO is applicable when the spent fuel pool is flooded and fuel is in the pool. The assembly is verified to comply with the criticality loading criteria specified in Specification 4.3.1.1 before placing it in the Spent Fuel Pool.
ACTIONS              A_1 If the spent fuel pool boron concentration does not meet the above requirements, action must be initiated to restore fuel storage pool boron concentration to within limits.
Suspension of CORE ALTERATIONS and positive reactivity additions shall not preclude moving a component to a safe position.
(continued)
Watts Bar - Unit 2                            B 3.9-24 Revision 41 Amendment 27
 
Spent Fuel Pool Boron Concentration B 3.9.9 BASES (continued)
SURVEILLANCE      SR 3.9.9.1 REQUIREMENTS This SR requires that the spent fuel pool boron concentration be verified greater than or equal to 2300 ppm. This surveillance is to be performed when fuel is stored in the spent fuel pool and in accordance with the Surveillance Frequency Control Program.
REFERENCES          1. Watts Bar FSAR, Section 4.3.2.7.
Watts Bar - Unit 2                        B 3.9-25 Revision 34, 41 Amendment 27, 36
 
ENCLOSURE 7 WBN UNIT 2 TECHNICAL REQUIREMENTS MANUAL TABLE OF CONTENTS (11 pages)
 
LIST OF TABLES TABLE NO.                                      TITLE                              PAGE 1.1-1          MODES . 1.1-6 3.0.2-1        Technical Surveillance Requirement.. 3.0-5 3.3.1-1        Reactor Trip System Instrumentation Response Times ....... 3.3-2 3.3.2-1        Engineered Safety Features Actuation System Response Times .. 3.3-5 3.3.4-1        Seismic Monitoring Instrumentation  3.3-15 3.3.9-1        Power Distribution Monitoring (PDMS) Instrumentation  3.3-23 3.7.3-1        Deleted . 3.7-8 3.7.3-2        Deleted ... 3.7-9 3.7.2-3        Deleted  3.7-11 3.7.3-4        Deleted . 3.7-12 3.7.3-5        Deleted .. 3.7-14 3.7.5-1        Area Temperature Monitoring ....... 3.7-22 3.8.3-1        Motor-Operated Valves Thermal Overload Devices Which Are Bypassed Under Accident Conditions  3.8-9 3.8.4-1        Submerged Components With Automatic De-energization Under Accident Conditions ....... 3.8-17 Watts Bar - Unit 2                            v Technical Requirements
 
LIST OF FIGURES FIGURE NO.                                      TITLE                                                    PAGE 3.1.6          Boric Acid Tank Limits Based on RWST Boron Concentration Level 1 RWST Concentration .................................................. 3.1-13 DELETED                                                                        3.7-15 3.7.3-1 LIST OF MISCELLANEOUS REPORTS AND PROGRAMS Core Operating Limits Report Watts Bar - Unit 2                            vi Technical Requirements
 
LIST OF ACRONYMS (Page 1 of 2)
ACRONYM              TITLE ABGTS                Auxiliary Building Gas Treatment System ACRP                  Auxiliary Control Room Panel AFD                  Axial Flux Difference AFW                  Auxiliary Feedwater System ARFS                  Air Return Fan System ARO                  All Rods Out ARV                  Atmospheric Relief Valve ASME                  American Society of Mechanical Engineers BOC                  Beginning of Cycle CCS                  Component Cooling Water System CFR                  Code of Federal Regulations COLR                  Core Operating Limits Report CREVS                Control Room Emergency Ventilation System CSS                  Containment Spray System CST                  Condensate Storage Tank DNB                  Departure from Nucleate Boiling ECCS                  Emergency Core Cooling System EFPD                  Effective Full-Power Days EGTS                  Emergency Gas Treatment System EOC                  End of Cycle ERCW                  Essential Raw Cooling Water ESF                  Engineered Safety Feature ESFAS                Engineered Safety Features Actuation System HEPA                  High Efficiency Particulate Air HVAC                  Heating, Ventilating, and Air-Conditioning LCC                  Lower Compartment Cooler LCO                  Limiting Condition For Operation MFIV                  Main Feedwater Isolation Valve MFRV                  Main Feedwater Regulation Valve MSIV                  Main Steam Line Isolation Valve MSSV                  Main Steam Safety Valve (continued)
Watts Bar - Unit 2                            vii Technical Requirements
 
LIST OF ACRONYMS (Page 2 of 2)
ACRONYM              TITLE MTC                  Moderator Temperature Coefficient N/A                  Not Applicable NMS                  Neutron Monitoring System ODCM                  Offsite Dose Calculation Manual PCP                  Process Control Program PDMS                  Power Distribution Monitoring System PIV                  Pressure Isolation Valve PORV                  Power-Operated Relief Valve PTLR                  Pressure and Temperature Limits Report QPTR                  Quadrant Power Tilt Ratio RAOC                  Relaxed Axial Offset Control RCCA                  Rod Cluster Control Assembly RCP                  Reactor Coolant Pump RCS                  Reactor Coolant System RHR                  Residual Heat Removal RTP                  Rated Thermal Power RTS                  Reactor Trip System RWST                  Refueling Water Storage Tank SG                    Steam Generator SI                    Safety Injection SL                    Safety Limit SR                    Surveillance Requirement TSR                  Technical Surveillance Requirement UHS                  Ultimate Heat Sink Watts Bar - Unit 2                          viii Technical Requirements
 
TECHNICAL REQUIREMENTS - LIST OF EFFECTIVE PAGES PAGE            REVISION                        PAGE        REVISION NUMBER            NUMBER                        NUMBER        NUMBER i                    14                        1.4-1          0 ii                  5                          1.4-2          0 iii                  14                          1.4-3          0 iv                  5                          1.4-4          0 v                    0                          3.0-1          0 vi                  0                          3.0-2          0 vii                  0                          3.0-3          6 viii                0                          3.0-4          0 ix                  16                          3.0-5          6 x                    16                          3.0-6          7 xi                  13                          3.1-1          0 xii                  16                          3.1-2          0 xiii                0                          3.1-3          0 xiv                  12                          3.1-4          12 xv                  16                          3.1-5          0 1.1-1                0                          3.1-6          0 1.1-2                0                          3.1-7          0 1.1-3                0                          3.1-8          0 1.1-4                14                          3.1-9          0 1.1-5                0                          3.1-10        0 1.1-6                0                          3.1-11        0 1.2-1                0                          3.1-12        0 1.2-2                0                          3.1-13        0 1.2-3                0                          3.1-14        13 1.3-1                0                          3.3-1          0 1.3-2                0                          3.3-2          2 1.3-3                0                          3.3-3          0 1.3-4                0                          3.3-4          0 1.3-5                0                          3.3-5          0 1.3-6                0                          3.3-6          0 1.3-7                0                          3.3-7          0 1.3-8                0                          3.3-8          11 1.3-9                0                          3.3-9          0 1.3-10              0                          3.3-10        0 Watts Bar - Unit 2                  ix Technical Requirements                                            Revision 16
 
TECHNICAL REQUIREMENTS - LIST OF EFFECTIVE PAGES PAGE            REVISION                      PAGE        REVISION NUMBER            NUMBER                        NUMBER        NUMBER 3.3-11              0                        3.7-3          0 3.3-12              0                        3.7-4          0 3.3-13              0                        3.7-5          5 3.3-14              0                        3.7-6          5 3.3-15              0                        3.7-7          5 3.3-16              0                        3.7-8          5 3.3-17              15                        3.7-9          5 3.3-18              14                        3.7-10        5 3.3-18a              14                        3.7-11        5 3.3-19              0                        3.7-12        5 3.3-20              0                        3.7-13        5 3.3-21              16                        3.7-14        5 3.3-22              0                        3.7-15        5 3.3-23              16                        3.7-16        0 3.4-1                0                        3.7-17        0 3.4-2                0                        3.7-18        0 3.4-3                0                        3.7-19        0 3.4-4                0                        3.7-20        0 3.4-5                0                        3.7-21        0 3.4-6                0                        3.7-22        10 3.4-7                0                        3.7-23        0 3.4-8                0                        3.8-1          0 3.4-9                0                        3.8-2          0 3.4-10              7                        3.8-3          0 3.4-11              0                        3.8-4          0 3.4-12              0                        3.8-5          0 3.6-1                0                        3.8-6          0 3.6-2                0                        3.8-7          0 3.6-3                0                        3.8-8          12 3.6-4                0                        3.8-9          0 3.6-5                0                        3.8-10        0 3.6-6                0                        3.8-11        0 3.7-1                0                        3.8-12        0 3.7-2                0                        3.8-13        0 Watts Bar - Unit 2                  x Technical Requirements                                          Revision 16
 
TECHNICAL REQUIREMENTS - LIST OF EFFECTIVE PAGES PAGE            REVISION                      PAGE        REVISION NUMBER            NUMBER                        NUMBER        NUMBER 3.8-14              3                        B 3.1-6        0 3.8-15              0                        B 3.1-7        12 3.8-16              12                        B 3.1-8        0 3.8-17              8                        B 3.1-9        0 3.8-18              0                        B 3.1-10      0 3.8-19              8                        B 3.1-11      0 3.9-1                0                        B 3.1-12      0 3.9-2                0                        B 3.1-13      0 3.9-3                0                        B 3.1-14      0 3.9-4                0                        B 3.1-15      0 3.9-5                0                        B 3.1-16      0 5.0-1                0                        B 3.1-17      0 B 3.1-18      0 B 3.1-19      0 B 3.0-1              0                        B 3.1-20      0 B 3.0-2              0                        B 3.1-21      0 B 3.0-3              0                        B 3.1-22      0 B 3.0-4              0                        B 3.1-23      0 B 3.0-5              0                        B 3.1-24      0 B 3.0-6              0                        B 3.1-25      13 B 3.0-7              0                        B 3.1-26      0 B 3.0-8              0                        B 3.3-1        0 B 3.0-9              0                        B 3.3-2        0 B 3.0-10            0                        B 3.3-3        0 B 3.0-11            0                        B 3.3-4        0 B 3.0-12            0                        B 3.3-5        0 B 3.0-13            0                        B 3.3-6        0 B 3.0-14            0                        B 3.3-7        0 B 3.0-15            0                        B 3.3-8        0 B 3.1-1              0                        B 3.3-9        0 B 3.1-2              0                        B 3.3-10      0 B 3.1-3              0                        B 3.3-11      0 B 3.1-4              0                        B 3.3-12      0 B 3.1-5              0                        B 3.3-13      0 Watts Bar - Unit 2                  xi Technical Requirements                                          Revision 13
 
TECHNICAL REQUIREMENTS - LIST OF EFFECTIVE PAGES PAGE            REVISION                      PAGE        REVISION NUMBER            NUMBER                        NUMBER        NUMBER B 3.3-14            15                        B 3.6-2        0 B 3.3-15            15                        B 3.6-3        0 B 3.3-16            15                        B 3.6-4        0 B 3.3-17            14                        B 3.6-5        0 B 3.3-18            14                        B 3.6-6        9 B 3.3-19            14                        B 3.6-7        0 B 3.3-20            14                        B 3.6-8        4 B 3.3-21            14                        B 3.6-9        0 B 3.3-22            14                        B 3.6-10      0 B 3.3-23            0                        B 3.6-11      0 B 3.3-24            0                        B 3.6-12      0 B 3.3-25            0                        B 3.7-1        0 B 3.3-26            0                        B 3.7-2        0 B 3.3-27            0                        B 3.7-3        0 B 3.3-28            16                        B 3.7-4        0 B 3.3-29            16                        B 3.7-5        0 B 3.3-30            0                        B 3.7-6        0 B 3.3-31            0                        B 3.7-7        0 B 3.4-1              0                        B 3.7-8        5 B 3.4-2              0                        B 3.7-9        5 B 3.4-3              0                        B 3.7-10      5 B 3.4-4              0                        B 3.7-11      5 B 3.4-5              0                        B 3.7-12      5 B 3.4-6              0                        B 3.7-13      5 B 3.4-7              0                        B 3.7-14      5 B 3.4-8              0                        B 3.7-15      0 B 3.4-9              0                        B 3.7-16      0 B 3.4-10            0                        B 3.7-17      0 B 3.4-11            0                        B 3.7-18      0 B 3.4-12            0                        B 3.7-19      0 B 3.4-13            7                        B 3.7-20      0 B 3.4-14            0                        B 3.7-21      0 B 3.4-15            0                        B 3.7-22      0 B 3.6-1              0                        B 3.8-1        0 Watts Bar - Unit 2                  xii Technical Requirements                                          Revision 16
 
TECHNICAL REQUIREMENTS - LIST OF EFFECTIVE PAGES PAGE            REVISION                      PAGE        REVISION NUMBER            NUMBER                        NUMBER        NUMBER B 3.8-2              0 B 3.8-3              0 B 3.8-4              0 B 3.8-5              0 B 3.8-6              0 B 3.8-7              0 B 3.8-8              0 B 3.8-9              0 B 3.8-10            0 B 3.8-11            0 B 3.8-12            0 B 3.8-13            0 B 3.8-14            0 B 3.8-15            12 B 3.8-16            0 B 3.8-17            0 B 3.8-18            0 B 3.8-19            12 B 3.9-1              0 B 3.9-2              0 B 3.9-3              0 B 3.9-4              0 B 3.9-5              0 B 3.9-6              0 B 3.9-7              0 B 3.9-8              0 Watts Bar - Unit 2                  xiii Technical Requirements                                          Revision 12
 
TECHNICAL REQUIREMENTS MANUAL LIST OF EFFECTIVE PAGES - REVISION LISTING Revisions            Issued                          SUBJECT Revision 01        11/25/15 Revises TRM and TRM Bases section 3.7.3, Snubbers.
Revision 02        05/22/16 TR Table 3.3.1-1, Reactor Trip System Instrumentation Response Times , to change the overtemperature and over power times.
Revision 03        06/27/16 TR Table 3.8.3-1, Motor-Operated Valves Thermal Overload Devices which are Bypassed under Accident Conditions, add valve 2-FCV-70-133 and delete 4 obsolete valves.
Revision 04        02/21/17 Revises TRM Bases 3.6.2, Inlet Door Position Monitoring System, Actions.
Revision 05        03/31/17 Revises TRM and TRM Bases to delete section 3.7.3 Snubbers.
Revision 06        07/08/17 Revises TRM section 3.0, Technical Surveillance Requirements (TSR) Applicability and adds Table 3.0.2-1.
Revision 07        08/22/17 Revises the TR 3.4.5 Title to add ASME Class 1, 2, and 3 in the TRM and Bases. Also revised TSR Table 3.0.2-1 to add two addition TSRs.
Revision 08        03/08/18 Revises TR Table 3.8.4-1 to revise the dual fan motors which were replaced with single fan motors.
Revision 09        04/06/18 Revises TRM Bases B3.6.2 to more closely match information provided in the UFSAR. The Bases as written limits credit for the lower inlet door main panel annunciator as part of the Inlet Door Position Monitoring system.
Revision 10        04/27/18 Revises TRM Table 3.7.5-1, Item 9 to correct the unit identifier on the Mechanical Equipment Room.
Revision 11        12/17/19 Revises TRM Table 3.3.2-1 Item 14 to add unbalanced voltage relay Revision 12        04/21/20 Revises TRM to change TSRs 3.1.2.3, 3.8.3.1, and 3.8.4.2 due to the frequency of SR 3.6.3.6 being changed to 36 months.
Watts Bar - Unit 2                        xiv Technical Requirements                                                            Revision 12
 
TECHNICAL REQUIREMENTS MANUAL LIST OF EFFECTIVE PAGES - REVISION LISTING Revisions            Issued                        SUBJECT Revision 13          11/4/20 Revises TRM and Bases TR 3.1.7 involving the requirements for the Position Indication System during Shutdown conditions.
Revision 14        11/10/20 Revises TRM and Bases to allow execution of the measurement uncertainty recovery (MUR) uprate.
Revision 15        06/14/21 Revises TRM and Bases to revise TR 3.3.6 to include both channels of the collection regions of Loose-Part Detection System.
Revision 16        01/11/22 Revises TRM and Bases to correct TRM 3.3.9 PDMS regarding the great than or equal too and greater than because a process parameter is never exactly equal to a value.
Watts Bar - Unit 2                        xv Technical Requirements                                                          Revision 16
 
ENCLOSURE 8 WBN UNIT 2 TECHNICAL REQUIREMENTS MANUAL CHANGED PAGES (26 pages)
 
TABLE OF CONTENTS TECHNICAL REQUIREMENTS TABLE OF CONTENTS .                                      i LIST OF TABLES                                        v LIST OF FIGURES .                                      vi LIST OF MISCELLANEOUS REPORTS AND PROGRAMS ..                            vi LIST OF ACRONYMS                                        vii LIST OF EFFECTIVE PAGES ..                                  ix TR 1.0    USE AND APPLICATION ..                              1.1-1 TR 1.1          Definitions                              1.1-1 TR 1.2          Logical Connectors ..............                1.2-1 TR 1.3          Completion Times ..                            1.3-1 TR 1.4          Frequency                                1.4-1 TR 3.0    APPLICABILITY ......... 3.0-1 TR 3.1    REACTIVITY CONTROL SYSTEMS ...                            3.1-1 TR 3.1.1        Boration Systems Flow Paths, Shutdown ................      3.1-1 TR 3.1.2        Boration Systems Flow Paths, Operating .................... 3.1-3 TR 3.1.3        Charging Pump, Shutdown ..                        3.1-5 TR 3.1.4        Charging Pumps, Operating .                        3.1-6 TR 3.1.5        Borated Water Sources, Shutdown .                    3.1-8 TR 3.1.6        Borated Water Sources, Operating .                    3.1-10 TR 3.1.7        Position Indication System, Shutdown                  3.1-14 TR 3.3    INSTRUMENTATION . 3.3-1 TR 3.3.1        Reactor Trip System (RTS) Instrumentation .              3.3-1 TR 3.3.2        Engineered Safety Features Actuation System .            3.3-4 TR 3.3.3        RESERVED FOR FUTURE ADDITION .                          3.3-11 TR 3.3.4        Seismic Instrumentation                          3.3-12 TR 3.3.5        RESERVED FOR FUTURE ADDITION .                          3.3-16 TR 3.3.6        Loose-Part Detection System ..                    3.3-17 TR 3.3.7        Plant Calorimetric Measurement .                    3.3-18 TR 3.3.8        Hydrogen Monitor ..                          3.3-19 TR 3.3.9        Power Distribution Monitoring System (PDMS) .              3.3-21 Watts Bar - Unit 2                              i                                  (continued)
Technical Requirements                                                            Revision 14
 
TABLE OF CONTENTS (continued)
TECHNICAL REQUIREMENTS BASES B 3.0      TECHNICAL REQUIREMENT (TR) AND TECHNICAL SURVEILLANCE REQUIREMENT (TSR) APPLICABILITY .............. B 3.0-1 B 3.1      REACTIVITY CONTROL SYSTEMS .. B 3.1-1 B 3.1.1          Boration Systems Flow Paths, Shutdown ................... B 3.1-1 B 3.1.2          Boration Systems Flow Paths, Operating ................... B 3.1-5 B 3.1.3          Charging Pump, Shutdown ...... B 3.1-9 B 3.1.4          Charging Pumps, Operating .... B 3.1-12 B 3.1.5          Borated Water Sources, Shutdown .... B 3.1-15 B 3.1.6          Borated Water Sources, Operating .... B 3.1-19 B 3.1.7          Position Indication System, Shutdown .......... B 3.1-24 B 3.3      INSTRUMENTATION . B 3.3-1 B 3.3.1          Reactor Trip System (RTS) Instrumentation . B 3.3-1 B 3.3.2          Engineered Safety Features Actuation System (ESFAS) Instrumentation  B 3.3-4 B 3.3.3          RESERVED FOR FUTRE ADDITION . B 3.3-7 B 3.3.4          Seismic Instrumentation  B 3.3-8 B 3.3.5          RESERVED FOR FUTURE ADDITION . B 3.3-13 B 3.3.6          Loose-Part Detection System .. B 3.3-14 B 3.3.7          Plant Calorimetric Measurement .                    B 3.3-17 B 3.3.8          Hydrogen Monitor .. B 3.3-18 B 3.3.9          Power Distribution Monitoring System (PDMS)  B 3.3-22 B 3.4      REACTOR COOLANT SYSTEM (RCS) .. B 3.4-1 B 3.4.1          Safety Valves, Shutdown .. B 3.4-1 B 3.4.2          Pressurizer Temperature Limits .. B 3.4-4 B 3.4.3          Reactor Vessel Head Vent System.. B 3.4-7 B 3.4.4          Chemistry  B 3.4-10 B 3.4.5          Piping System Structural Integrity ... B 3.4-13 B 3.6      CONTAINMENT SYSTEMS .. B 3.6-1 B 3.6.1          Ice Bed Temperature Monitoring System ......... B 3.6-1 B 3.6.2          Inlet Door Position Monitoring System  B 3.6-6 B 3.6.3          Lower Compartment Cooling (LCC) System . B 3.6-10 Watts Bar - Unit 2                              iii Technical Requirements                                                          Revision 14
 
TECHNICAL REQUIREMENTS - LIST OF EFFECTIVE PAGES PAGE            REVISION                        PAGE        REVISION NUMBER            NUMBER                        NUMBER        NUMBER i                    14                        1.4-1          0 ii                  5                          1.4-2          0 iii                  14                          1.4-3          0 iv                  5                          1.4-4          0 v                    0                          3.0-1          0 vi                  0                          3.0-2          0 vii                  0                          3.0-3          6 viii                0                          3.0-4          0 ix                  16                          3.0-5          6 x                    16                          3.0-6          7 xi                  13                          3.1-1          0 xii                  16                          3.1-2          0 xiii                0                          3.1-3          0 xiv                  12                          3.1-4          12 xv                  16                          3.1-5          0 1.1-1                0                          3.1-6          0 1.1-2                0                          3.1-7          0 1.1-3                0                          3.1-8          0 1.1-4                14                          3.1-9          0 1.1-5                0                          3.1-10        0 1.1-6                0                          3.1-11        0 1.2-1                0                          3.1-12        0 1.2-2                0                          3.1-13        0 1.2-3                0                          3.1-14        13 1.3-1                0                          3.3-1          0 1.3-2                0                          3.3-2          2 1.3-3                0                          3.3-3          0 1.3-4                0                          3.3-4          0 1.3-5                0                          3.3-5          0 1.3-6                0                          3.3-6          0 1.3-7                0                          3.3-7          0 1.3-8                0                          3.3-8          11 1.3-9                0                          3.3-9          0 1.3-10              0                          3.3-10        0 Watts Bar - Unit 2                  ix Technical Requirements                                            Revision 16
 
TECHNICAL REQUIREMENTS - LIST OF EFFECTIVE PAGES PAGE            REVISION                      PAGE        REVISION NUMBER            NUMBER                        NUMBER        NUMBER 3.3-11              0                        3.7-3          0 3.3-12              0                        3.7-4          0 3.3-13              0                        3.7-5          5 3.3-14              0                        3.7-6          5 3.3-15              0                        3.7-7          5 3.3-16              0                        3.7-8          5 3.3-17              15                        3.7-9          5 3.3-18              14                        3.7-10        5 3.3-18a              14                        3.7-11        5 3.3-19              0                        3.7-12        5 3.3-20              0                        3.7-13        5 3.3-21              16                        3.7-14        5 3.3-22              0                        3.7-15        5 3.3-23              16                        3.7-16        0 3.4-1                0                        3.7-17        0 3.4-2                0                        3.7-18        0 3.4-3                0                        3.7-19        0 3.4-4                0                        3.7-20        0 3.4-5                0                        3.7-21        0 3.4-6                0                        3.7-22        10 3.4-7                0                        3.7-23        0 3.4-8                0                        3.8-1          0 3.4-9                0                        3.8-2          0 3.4-10              7                        3.8-3          0 3.4-11              0                        3.8-4          0 3.4-12              0                        3.8-5          0 3.6-1                0                        3.8-6          0 3.6-2                0                        3.8-7          0 3.6-3                0                        3.8-8          12 3.6-4                0                        3.8-9          0 3.6-5                0                        3.8-10        0 3.6-6                0                        3.8-11        0 3.7-1                0                        3.8-12        0 3.7-2                0                        3.8-13        0 Watts Bar - Unit 2                  x Technical Requirements                                          Revision 16
 
TECHNICAL REQUIREMENTS - LIST OF EFFECTIVE PAGES PAGE            REVISION                      PAGE        REVISION NUMBER            NUMBER                        NUMBER        NUMBER 3.8-14              3                        B 3.1-6        0 3.8-15              0                        B 3.1-7        12 3.8-16              12                        B 3.1-8        0 3.8-17              8                        B 3.1-9        0 3.8-18              0                        B 3.1-10      0 3.8-19              8                        B 3.1-11      0 3.9-1                0                        B 3.1-12      0 3.9-2                0                        B 3.1-13      0 3.9-3                0                        B 3.1-14      0 3.9-4                0                        B 3.1-15      0 3.9-5                0                        B 3.1-16      0 5.0-1                0                        B 3.1-17      0 B 3.1-18      0 B 3.1-19      0 B 3.0-1              0                        B 3.1-20      0 B 3.0-2              0                        B 3.1-21      0 B 3.0-3              0                        B 3.1-22      0 B 3.0-4              0                        B 3.1-23      0 B 3.0-5              0                        B 3.1-24      0 B 3.0-6              0                        B 3.1-25      13 B 3.0-7              0                        B 3.1-26      0 B 3.0-8              0                        B 3.3-1        0 B 3.0-9              0                        B 3.3-2        0 B 3.0-10            0                        B 3.3-3        0 B 3.0-11            0                        B 3.3-4        0 B 3.0-12            0                        B 3.3-5        0 B 3.0-13            0                        B 3.3-6        0 B 3.0-14            0                        B 3.3-7        0 B 3.0-15            0                        B 3.3-8        0 B 3.1-1              0                        B 3.3-9        0 B 3.1-2              0                        B 3.3-10      0 B 3.1-3              0                        B 3.3-11      0 B 3.1-4              0                        B 3.3-12      0 B 3.1-5              0                        B 3.3-13      0 Watts Bar - Unit 2                  xi Technical Requirements                                          Revision 13
 
TECHNICAL REQUIREMENTS - LIST OF EFFECTIVE PAGES PAGE            REVISION                      PAGE        REVISION NUMBER            NUMBER                        NUMBER        NUMBER B 3.3-14            15                        B 3.6-2        0 B 3.3-15            15                        B 3.6-3        0 B 3.3-16            15                        B 3.6-4        0 B 3.3-17            14                        B 3.6-5        0 B 3.3-18            14                        B 3.6-6        9 B 3.3-19            14                        B 3.6-7        0 B 3.3-20            14                        B 3.6-8        4 B 3.3-21            14                        B 3.6-9        0 B 3.3-22            14                        B 3.6-10      0 B 3.3-23            0                        B 3.6-11      0 B 3.3-24            0                        B 3.6-12      0 B 3.3-25            0                        B 3.7-1        0 B 3.3-26            0                        B 3.7-2        0 B 3.3-27            0                        B 3.7-3        0 B 3.3-28            16                        B 3.7-4        0 B 3.3-29            16                        B 3.7-5        0 B 3.3-30            0                        B 3.7-6        0 B 3.3-31            0                        B 3.7-7        0 B 3.4-1              0                        B 3.7-8        5 B 3.4-2              0                        B 3.7-9        5 B 3.4-3              0                        B 3.7-10      5 B 3.4-4              0                        B 3.7-11      5 B 3.4-5              0                        B 3.7-12      5 B 3.4-6              0                        B 3.7-13      5 B 3.4-7              0                        B 3.7-14      5 B 3.4-8              0                        B 3.7-15      0 B 3.4-9              0                        B 3.7-16      0 B 3.4-10            0                        B 3.7-17      0 B 3.4-11            0                        B 3.7-18      0 B 3.4-12            0                        B 3.7-19      0 B 3.4-13            7                        B 3.7-20      0 B 3.4-14            0                        B 3.7-21      0 B 3.4-15            0                        B 3.7-22      0 B 3.6-1              0                        B 3.8-1        0 Watts Bar - Unit 2                  xii Technical Requirements                                          Revision 16
 
TECHNICAL REQUIREMENTS MANUAL LIST OF EFFECTIVE PAGES - REVISION LISTING Revisions            Issued                        SUBJECT Revision 13          11/4/20 Revises TRM and Bases TR 3.1.7 involving the requirements for the Position Indication System during Shutdown conditions.
Revision 14        11/10/20 Revises TRM and Bases to allow execution of the measurement uncertainty recovery (MUR) uprate.
Revision 15        06/14/21 Revises TRM and Bases to revise TR 3.3.6 to include both channels of the collection regions of Loose-Part Detection System.
Revision 16        01/11/22 Revises TRM and Bases to correct TRM 3.3.9 PDMS regarding the great than or equal too and greater than because a process parameter is never exactly equal to a value.
Watts Bar - Unit 2                        xv Technical Requirements                                                          Revision 16
 
Definitions 1.1 1.1 Definitions (continued)
Term                        Definition RATED THERMAL POWER        RTP shall be a total reactor core heat transfer rate to the (RTP)                      reactor coolant of 3459 MWt.
REACTOR TRIP SYSTEM        The RTS RESPONSE TIME shall be that time interval from (RTS) RESPONSE TIME        when the monitored parameter exceeds its RTS trip setpoint at the channel sensor until loss of stationary gripper coil voltage. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured. In lieu of measurement, response time may be verified for selected components provided that the components and the methodology for verification have been previously reviewed and approved by the NRC.
SHUTDOWN MARGIN (SDM)      SDM shall be the instantaneous amount of reactivity by which the reactor is subcritical or would be subcritical from its present condition assuming:
: a. All rod cluster control assemblies (RCCAs) are fully inserted except for the single RCCA of highest reactivity worth, which is assumed to be fully withdrawn. With any RCCA not capable of being fully inserted, the reactivity worth of the RCCA must be accounted for in the determination of SDM; and
: b. In MODES 1 and 2, the fuel and moderator temperatures are changed to the nominal zero power design level.
STAGGERED TEST BASIS        A STAGGERED TEST BASIS shall consist of the testing of one of the systems, subsystems, channels, or other designated components during the interval specified by the Surveillance Frequency so that all systems, subsystems, channels, or other designated components are tested during n Surveillance Frequency intervals, where n is the total number of systems, subsystems, channels, or other designated components in the associated function.
THERMAL POWER              THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant.
Watts Bar - Unit 2                      1.1-4                                    (continued)
Technical Requirements Revision 14
 
Position Indication System, Shutdown TR 3.1.7 TR 3.1 REACTIVITY CONTROL SYSTEMS TR 3.1.7 Position Indication System, Shutdown TR 3.1.7              The group demand position indicators shall be OPERABLE and capable of determining within +/- 12 steps the demand position for each shutdown or control rod that is not fully inserted.
APPLICABILITY:        MODES 3, 4, and 5, when the reactor trip breakers are closed.
ACTIONS CONDITION                          REQUIRED ACTION                      COMPLETION TIME A. One or more                  A.1    Restore required group              15 minutes group demand                        demand position position                            indicator indicators                          OPERABILITY.
inoperable.
B. Required Action and          B.1    Initiate action to fully insert all Immediately associated completion                rods.
Time of Condition A not met.                                  OR B.2      Open reactor trip breakers.        Immediately TECHNICAL SURVEILLANCE REQUIREMENTS SURVEILLANCE                                            FREQUENCY TSR 3.1.7.1        Determine that each group demand position indicator          Within 4 hours after is OPERABLE by movement of the associated                    closing the reactor shutdown or control rod 10 steps in any one                  trip breakers if not direction.                                                    completed within previous 31 days.
AND 31 days thereafter Watts Bar - Unit 2                              3.1-14 Technical Requirements                                                                    Revision 13
 
Loose-Part Detection System TR 3.3.6 TR 3.3 INSTRUMENTATION TR 3.3.6 Loose-Part Detection System TR 3.3.6                    The Loose-Part Detection System shall be OPERABLE.
APPLICABILITY:              MODES 1 and 2.
--------------------------------------------------------NOTE-------------------------------------------------------------
TR 3.0.3 is not applicable.
ACTIONS CONDITION                                  REQUIRED ACTION                          COMPLETION TIME A.      Both channels of one or                A.1          Document in accordance              In accordance with more collection regions of                            with the Corrective                  the Corrective Action Loose-Part Detection                                  Action Program.                      Program.
System inoperable
        > 30 days.
TECHNICAL SURVEILLANCE REQUIREMENTS SURVEILLANCE                                                      FREQUENCY TSR 3.3.6.1              Perform CHANNEL CHECK.                                                    24 hours TSR 3.3.6.2              Perform CHANNEL OPERATIONAL TEST.                                        31 days TSR 3.3.6.3              Perform CHANNEL CALIBRATION.                                              18 months Watts Bar - Unit 2                                        3.3-17                                              Revision 15 Technical Requirements
 
Plant Calorimetric Measurement TR 3.3.7 TR 3.3 INSTRUMENTATION TR 3.3.7 Plant Calorimetric Measurement TR 3.3.7                      The Leading Edge Flow Meter (LEFM) shall be used for the completion of SR 3.3.1.2.
APPLICABILITY:              MODE 1 > 15% RTP.
--------------------------------------------------------NOTES-----------------------------------------------------------
See Bases for Definitions of LEFM flow meter NORMAL, MAINTENANCE, and FAIL status.
ACTIONS CONDITION                                  REQUIRED ACTION                          COMPLETION TIME A.      LEFM in MAINTENANCE                    A.1          Restore LEFM to                      72 hours mode.                                                NORMAL mode.
B.      Required Action and                    B.1          Ensure THERMAL                      Immediately associated Completion                                POWER  98.6% RTP Time of Condition A not                              (3411 MWt).
met.
AND B.2          Perform SR 3.3.1.2                  As required by SR based on feedwater                  3.3.1.2 venturis calorimetric.
AND B.3          Maintain THERMAL                    Until LEFM is POWER  98.6% RTP                    restored to NORMAL (3411 MWt).                          status and SR 3.3.1.2 is performed using LEFM.
C.      LEFM in FAIL mode.                      C.1          Restore LEFM to                      Prior to next NORMAL mode.                        performance of SR 3.3.1.2 Watts Bar - Unit 2                                        3.3-18                                              Revision 14 Technical Requirements
 
Plant Calorimetric Measurement TR 3.3.7 ACTIONS (continued)
D. Required Action and              D.1      Ensure THERMAL              Prior to performance associated Completion                      POWER  98.6% RTP            of SR 3.3.1.2 Time of Condition C not                    (3411 MWt).
met.
AND D.2      Perform SR 3.3.1.2          As required by SR based on feedwater          3.3.1.2 venturis calorimetric.
AND D.3      Maintain THERMAL            Until LEFM is POWER  98.6% RTP            restored to NORMAL (3411 MWt).                  status and SR 3.3.1.2 is performed using LEFM.
TECHNICAL SURVEILLANCE REQUIREMENTS SURVEILLANCE                                      FREQUENCY TSR 3.3.7.1        Verify status of the LEFM, using the self-diagnostics    Prior to performance feature indicated by the LEFM                            of SR 3.3.1.2 NORMAL/MAINTENANCE/FAIL mode status indication as displayed on the plant computer system, is not in FAIL or MAINTENANCE status.
Watts Bar - Unit 2                            3.3-18a                                  Revision 14 Technical Requirements
 
Power Distribution Monitoring System (PDMS)
TR 3.3.9 TR 3.3 INSTRUMENTATION TR 3.3.9 Power Distribution Monitoring System (PDMS)
TR 3.3.9            The PDMS shall be OPERABLE with:
: a. THERMAL POWER >25% RTP, and
: b. The required channel inputs from the plant computer for each function OPERABLE as defined in Table 3.3.9-1 APPLICABILITY:      When the PDMS is used for:
: a. Calibration of the Excore Neutron Flux Detection System, or
: b. Monitoring the QUADRANT POWER TILT RATIO, or
: c. Measurement of F and FQ(Z), or
: d. Verifying the position of a rod with inoperable position indicators.
ACTIONS CONDITION                        REQUIRED ACTION                        COMPLETION TIME A. PDMS inoperable.                A.1        -------------NOTE-------------
TR 3.0.3 is not applicable.
Restore the inoperable            Prior to using the system to OPERABLE                system for incore status.                            power distribution measurement purposes.
Watts Bar - Unit 2                            3.3-21                                          Revision 16 Technical Requirements
 
Power Distribution Monitoring System (PDMS)
TR 3.3.9 Table 3.3.9-1 (Page 1 of 1)
Power Distribution Monitoring System (PDMS) Instrumentation REQUIRED        SURVEILLANCE              SURVEILLANCE FUNCTION              CHANNELS          REQUIREMENTS              DESCRIPTION
: 1. RCS Cold Leg                    2(7)        SR 3.3.1.10(5)      CHANNEL CALIBRATION Temperature                                  SR 3.3.3.2(6)        CHANNEL CALIBRATION
: 2. Reactor Power                    1(1)        SR 3.3.1.2(4)        NIS Adjustment SR 3.3.1.10(3)      CHANNEL CALIBRATION TSR 3.3.7.1(4)      LEFM Availability
: 3. Control Bank Position            1(2)        SR 3.1.8.1          RPI Calibration (per bank)
: 4. Self-Powered Detector        218 total(8)
Segments                        and 10 per core quadrant and 4 per top core quadrant and 4 per bottom core quadrant (1) Either secondary calorimetric power, average power range neutron flux power, or average RCS Loop T power (2) Either the Demand Position Indication or the average of the individual Rod Position Indications (3) Applies to average RCS Loop T power only (4) Not applicable to average RCS Loop T power (5) Applies to Narrow Range RTDs only.
(6) Applies to Wide Range RTDs only.
(7) Either Narrow Range or Wide Range RTDs.
(8)  145 Self-Powered Detector Segments are sufficient for incore power distribution measurements generated by the PDMS subsequent to the initial incore power distribution measurement in each operating cycle.
Watts Bar - Unit 2                            3.3-23                                  Revision 16 Technical Requirements
 
Position Indication System, Shutdown B 3.1.7 BASES (continued)
TR                  TR 3.1.7 specifies that the group demand position indicators be OPERABLE and capable of determining within +/- 12 steps the demand position for each shutdown or control rod not fully inserted. For the control rod position indicators to be OPERABLE requires meeting the surveillance requirement of the TR. This requirement provides adequate assurance that control rod position indication during shutdown conditions and rod testing is accurate, and that design assumptions are not challenged. OPERABILITY of the required position indicators ensures that inoperable, misaligned, or mispositioned control rods can be detected.
APPLICABILITY      This TR covers only the requirements on Rod Position Indication during MODES 3, 4, and 5 with the reactor trip breakers closed. Rod Position Indication during MODES 1 and 2 are covered by Technical Specification 3.1.8. In MODE 6 and in MODES 3, 4, and 5 with trip breakers open or all rods fully inserted, Rod Position Indication is not required to be OPERABLE. Rod Position Indication OPERABILITY is required only when rods are not fully inserted.
ACTIONS            A.1 With one or more group demand position indicators inoperable, OPERABILITY of the affected group demand position indicators must be restored promptly.
The 15 minutes provides an acceptable time to evaluate whether the group demand position indicators represent the actual demand position of the affected rods and whether the affected rods are not fully inserted in an orderly manner without allowing the plant to remain in an unacceptable condition for an extended period of time.
The immediate Completion Times are consistent with the required time for actions to be pursued without delay and in a controlled manner.
B.1 and B.2 If OPERABILITY of the group demand position indicators is not restored within 15 minutes, the unit must be placed in a condition where the demand position indicators are not required. This is accomplished by fully inserting all rods or opening the reactor trip breakers immediately.
Watts Bar - Unit 2                          B 3.1-25                                (continued)
Technical Requirements                                                              Revision 13
 
Loose-Part Detection System B 3.3.6 B 3.3 INSTRUMENTATION B 3.3.6 Loose-Part Detection System BASES BACKGROUND          The Loose-Part Detection System consists of twelve sensors with associated pre-amplifiers, signal conditioners and digital signal processor units, and a CPU with its supporting equipment. The sensors are located in the six natural collection regions. These regions consist of the top and bottom plenums of the reactor vessel and the primary coolant inlet plenum to each steam generator. The entire system is described in Reference 1.
The Loose-Part Detection System provides the capability to detect acoustic disturbances indicative of loose parts within the Reactor Coolant System (RCS) pressure boundary. This system is provided to avoid or mitigate damage to RCS components that could occur from these loose parts. The Loose-Part Detection System Technical Requirement is consistent with the recommendations of Reference 2.
APPLICABLE          The presence of a loose part in the RCS can be indicative of degraded SAFETY              reactor safety resulting from failure or weakening of a safety-related ANALYSES            component. A loose part, whether it is from a failed or weakened component, or from an item inadvertently left in the primary system during construction, refueling, or maintenance, can contribute to component damage and material wear by frequent impacting with other parts in the system. Also, a loose part increases the potential for control-rod jamming and for accumulation of increased levels of radioactive crud in the primary system (Ref. 2).
The Loose-Part Detection System provides the capability to detect loose parts in the RCS which could cause damage to some component in the RCS. Loose parts are not assumed to initiate any DBA, and the detection of a loose part is not required for mitigation of any DBA (Ref. 3).
TR                  TR 3.3.6 requires the Loose-Part Detection System to be OPERABLE.
This is necessary to ensure that sufficient capability is available to detect loose metallic parts in the RCS and avoid or mitigate damage to the RCS components. This requirement is provided in Reference 2.
(continued)
Watts Bar - Unit 2                          B 3.3-14                                  Revision 15 Technical Requirements
 
Loose-Part Detection System B 3.3.6 BASES (continued)
APPLICABILITY      TR 3.3.6 is required to be met in MODES 1 and 2 as stated in Reference 2. These MODES of applicability are provided in Reference 2.
The Applicability has been modified by a Note stating that the provisions of TR 3.0.3 do not apply.
ACTIONS            A.1 If both channels of one or more collection regions of the Loose-Part Detection System are inoperable for more than 30 days, document the inoperability of the channels in accordance with the Corrective Action Program.
TECHNICAL          TSR 3.3.6.1 SURVEILLANCE REQUIREMENTS        Performance of a CHANNEL CHECK for the Loose-Part Detection System once every 24 hours ensures that a gross failure of instrumentation has not occurred. In addition, the Loose-Part Detection System performs an automatic system self-test each day which provides a printable daily report and displays any faults discovered during the test.
The CHANNEL CHECK activity will review the daily report, observe the display to determine if any faults were discovered during the system self-test, verify the system is in an operable condition and verify there are no alarms. The CHANNEL CHECK will detect gross channel failure; thus, it is key to verifying that the instrumentation continues to operate properly between each CHANNEL CALIBRATION.
The Surveillance and the Surveillance Frequency are provided in Reference 2.
TSR 3.3.6.2 A CHANNEL OPERATIONAL TEST is to be performed every 31 days on each required channel to ensure the entire channel will perform the intended function. This test verifies the capability of the Loose-Part Detection System to detect impact signals which would indicate a loose part in the RCS. The Surveillance and the Surveillance Frequency are provided in Reference 2.
(continued)
Watts Bar - Unit 2                          B 3.3-15                                Revision 15 Technical Requirements
 
Loose-Part Detection System B 3.3.6 BASES TECHNICAL          TSR 3.3.6.3 SURVEILLANCE REQUIREMENTS        CHANNEL CALIBRATION is a complete check of the instrument loop and (continued)      the sensor. The 18 month Surveillance Frequency is based upon operating experience and is consistent with the typical industry refueling cycle. The Surveillance and the Surveillance Frequency are provided in Reference 2. Reference 1 describes the built-in capabilities of the system to verify proper channel calibration. This is an acceptable option to using a mechanical impact device for sensors located in plant areas where plant personnel radiation exposure is considered by Plant Management to be excessive.
REFERENCES        1.      Watts Bar FSAR, Section 7.6.7, Loose Part Monitoring System (LPMS) System Description.
: 2.      Regulatory Guide 1.133, Loose-Part Detection Program for the Primary System of Light-Water-Cooled Reactors.
: 3.      WCAP-11618, MERITS Program-Phase II, Task 5, Criteria Application, including Addendum 1 dated April, 1989.
Watts Bar - Unit 2                          B 3.3-16                                Revision 15 Technical Requirements
 
Plant Calorimetric Measurement B 3.3.7 B 3.3 INSTRUMENTATION B 3.3.7 Plant Calorimetric Measurement BASES BACKGROUND          The predominant contribution to the secondary plant calorimetric measurement uncertainty is the uncertainty associated with the feedwater flow measurement. Traditionally, a differential pressure (P) transmitter across a venturi in each main feedwater line has been used to provide the feedwater flow. However, the venturis are subject to fouling and the uncertainty associated with the flow derived from the P indication can be large and increases as the flow deviates from the optimum conditions for which the P transmitter was calibrated.
More recently, Leading Edge Flow Meters (LEFMs) have been used to provide the feedwater flow input to the secondary plant calorimetric measurement. The uncertainty associated with the LEFM is relatively small and is independent of the actual feedwater flow.
Most of the original safety analyses supporting plant operation (including LOCA analyses required by 10CFR50 Appendix K) were performed at a maximum power level of 3411 MWt plus an allowance for the secondary plant calorimetric uncertainty assumed to be 2% above rated thermal power. Hence, it is possible to support operation at a higher power level while remaining within the original analyses.
The RATED THERMAL POWER for Unit 2 is 3459 MWt which represents an increase of 1.4% RTP from the originally licensed value of 3411 MWt.
This uprate is based on reduced uncertainties associated with the secondary plant calorimetric measurement that is attained through the use of the LEFM CheckPlus supplied by Caldon, Inc. Many of the accident analyses are performed at 102% of 3411 MWt, or 3479 MWt, where the 2% RTP is an allowance for the uncertainty associated with the power calorimetric measurement. With the LEFMs, the power calorimetric measurement uncertainty is less than 0.6% RTP. Without performing new Appendix K accident analyses, the LEFM CheckPlus can be used to allow the plant to be operated at a redefined 100% RTP of 3459 MWt.
(continued)
Watts Bar - Unit 2                          B 3.3-17                                Revision 14 Technical Requirements
 
Plant Calorimetric Measurement B 3.3.7 BASES BACKGROUND          However, this allowance is predicated on the availability of the LEFM (continued)        CheckPlus for performance of the calorimetric measurement. There are three possible statuses (modes) for a LEFM CheckPlus flow meter; NORMAL, MAINTENANCE, and FAIL.
A LEFM Flow Meter status is considered NORMAL if:
The LEFM System Computer indicates that flow meter status (mode) to be Normal.
A LEFM flow meter status is considered MAINTENANCE if:
The LEFM System Computer indicates that flow meter status (mode) to be Maintenance.
A LEFM flow meter status is considered FAIL if:
The LEFM System Computer indicates that flow meter status (mode) to be Fail.
The LEFM CheckPlus Flow Meter status is automatically determined and reported by the LEFM System computer based upon the number of functional planes in the LEFM and upon its data quality. If an LEFM flow meter is in a status other than NORMAL, the uncertainty for that meter is increased. When an LEFM CheckPlus system has only one of its two LEFM Check subsystems fully operational, resulting in that meter computing flow from just the remaining fully operational LEFM Check subsystem, that LEFM flow meter is considered to be in the MAINTENANCE mode.
If the WBN Unit 2 LEFM is in MAINTENANCE mode then a 72 hour Completion Time will be used prior to reducing power to 3411 MWt. This is conservative because with the LEFM in MAINTENANCE mode the power calorimetric measurement uncertainty is still less than 0.6% RTP. If the LEFM is in FAIL mode, then the LEFM CheckPlus is unavailable, and the uncertainties associated with the feedwater venturi-based measurement (2% RTP) must be used to ensure compliance with the safety analysis value of the core power of 3479 MWt.
Surveillance Requirement SR 3.3.1.2 requires the performance of a comparison of the results of the calorimetric heat balance calculation to Nuclear Instrumentation System (NIS) channel output. SR 3.3.1.2 note 1 requires that the NIS channels be adjusted if the absolute difference is
                    > 2% RTP.
(continued)
Watts Bar - Unit 2                        B 3.3-18                                Revision 14 Technical Requirements
 
Plant Calorimetric Measurement B 3.3.7 BASES BACKGROUND          SR 3.3.1.2 is required to be performed every 24 hours (daily). At that (continued)        time, the NIS indication must be normalized to indicate within at least +/-
2% RTP of the calorimetric measurement. The plant may then be run for the next 24 hour period using this normalized NIS indication, such that the calorimetric power does not exceed 100% RTP. Although the calorimetric power indication may be monitored continuously for control of the unit power, the calorimetric power indication is not required to be consulted again until the daily calorimetric comparisons of the NIS indication are performed.
The following general guidance is provided for operation of WBN Unit 2:
: 1)      When the LEFM CheckPlus is in NORMAL status, the plant should be operated in a manner consistent with the LEFM CheckPlus based calorimetric measurement and at 3459 MWt (100% RTP).
: 2)      If the LEFM CheckPlus is in MAINTENANCE status (mode), the plant may be operated at 3459 MWt (100% RTP) in a manner consistent with the LEFM CheckPlus based calorimetric measurement for up to 72 hours.
: 3)      If the LEFM CheckPlus is unavailable (i.e., in FAIL mode), the plant may be operated at 3459 MWt (100% RTP) using the NIS indication until the next performance of SR 3.3.1.2 is due.
: 4)      If the LEFM CheckPlus based calorimetric measurement is unavailable (i.e., in FAIL mode) at the time SR 3.3.1.2 is due, or if the LEFM CheckPlus has been in MAINTENANCE mode for 72 hours, the normalized feedwater venturi-based calorimetric measurement should be used for the performance of SR 3.3.1.2.
However, to maintain consistency with the uncertainty analysis, the maximum allowable power should be reduced to 3411 MWt or 98.6% RTP. Either the NIS indication or the normalized feedwater venturi-based calorimetric power indication may be used to control the unit power.
(continued)
Watts Bar - Unit 2                          B 3.3-19                                Revision 14 Technical Requirements
 
Plant Calorimetric Measurement B 3.3.7 BASES APPLICABLE          Each of the analyzed accidents are evaluated for the range of power SAFETY              levels over which the reactor is allowed to be operated. Typically, the ANALYSES            analyses are most limiting when initiated from a higher power level. For Unit 2, the majority of the original analyses were performed for a core power of at least 3411 MWt, plus an allowance for the secondary power calorimetric measurement of 2% RTP. In general, these same analyses are used to support the revised RATED THERMAL POWER definition of a core power of 3459 MW t. With the application of a 0.6% RTP uncertainty (based on the use of the LEFM CheckPlus feedwater flow input into the secondary calorimetric measurement), the analyses are evaluated at a power level of 3479 MWt. Analyses that use statistical methods, such as the analysis of the dropped RCCA event, are explicitly evaluated for operation at 3411 MWt with a 2% RTP uncertainty allowance and for operation at 3459 MWt with a 0.6% RTP uncertainty allowance.
The setpoints for those functions of the Reactor Protection System that are based on percentage of power (i.e., the NIS) have been calculated based on analytical margins available at the 3459 MWt definition of 100%
RTP. Operation back at 3411 MWt does not require these setpoints to be adjusted.
TR                  The TR requires the LEFM CheckPlus to be used for the completion of the daily secondary plant calorimetric measurement required in SR 3.3.1.2. The use of the LEFM CheckPlus ensures that the basis for operation at the RATED THERMAL POWER of 3459 MWt is maintained.
APPLICABILITY      The requirement to use the LEFM CheckPlus for the performance of the secondary plant calorimetric measurement required by SR 3.3.1.2 is applicable to Unit 2 in mode 1 above 15% RTP, consistent with the applicability of SR 3.3.1.2.
ACTIONS            A.1 If the LEFM is in MAINTENANCE mode plant operation may continue using the power indications from the LEFM system for up to 72 hours.
This is considered a reasonable amount of time to restore the LEFM to NORMAL status.
(continued)
Watts Bar - Unit 2                        B 3.3-20                                Revision 14 Technical Requirements
 
Plant Calorimetric Measurement B 3.3.7 BASES ACTIONS            B.1, B.2, and B.3 (continued)
If the Required Action and associated Completion Time of Condition A is not met (i.e., the LEFM has not been returned to NORMAL status mode within 72 hours), Condition B is entered. Required Action B.1 requires that the reactor power be reduced to, or maintained at, a power level less than or equal to 98.6% RTP (3411 MWt). This power reduction is performed immediately, and prior to performing SR 3.3.1.2.
Required Action B.2 directs the performance of SR 3.3.1.2 using the feedwater venturi indications of feedwater flow. Once SR 3.3.1.2 is performed using the feedwater venturi indications of feedwater flow, the required power uncertainty is 2% RTP. In order to maintain compliance with the safety analyses, it is necessary to operate the plant at a maximum core thermal power of 3411 MWt.
Required Action B.3 serves as a reminder that the core power is to be maintained at a value less than or equal to 3411 MWt until the LEFM is returned to service (i.e., NORMAL status mode) and SR 3.3.1.2 has been performed using the LEFM indication of feedwater flow. Once SR 3.3.1.2 has been performed using the LEFM, the plant can again be operated at 3459 MWt.
C.1 If the LEFM becomes unavailable (i.e., is in FAIL mode) during the intervals between performance of SR 3.3.1.2, plant operation may continue using the power indications from the NIS system. However, in order to remain in compliance with the bases for operation at a RATED THERMAL POWER of 3459 MWt, the LEFM must be returned to service (i.e., NORMAL status mode) prior to performance of SR 3.3.1.2.
D.1, D.2, and D.3 If the Required Action and associated Completion Time of Condition C is not met (i.e., the LEFM has not been returned to NORMAL status prior to the performance of SR 3.3.1.2), Condition D is entered. Required Action D.1 requires that the reactor power be reduced to, or maintained at, a power level less than or equal to 98.6% RTP (3411 MWt). This power reduction is performed prior to performing SR 3.3.1.2, in order to remain within the plant's design bases immediately upon performance of SR 3.3.1.2.
(continued)
Watts Bar - Unit 2                        B 3.3-21                                  Revision 14 Technical Requirements
 
Plant Calorimetric Measurement B 3.3.7 BASES ACTIONS            Required Action D.2 directs the performance of SR 3.3.1.2 using the (continued)        feedwater venturi indications of feedwater flow. Once SR 3.3.1.2 is performed using the feedwater venturi indications of feedwater flow, the required power uncertainty is 2% RTP. In order to maintain compliance with the safety analyses, it is necessary to operate the plant at a maximum core thermal power of 3411 MWt.
Required Action D.3 serves as a reminder that the core power is to be maintained at a value less than or equal to 3411 MWt until the LEFM is returned to service and SR 3.3.1.2 has been performed using the LEFM indication of feedwater flow. Once SR 3.3.1.2 has been performed using the LEFM, the plant can again be operated at 3459 MWt.
TECHNICAL          TSR 3.3.7.1 SURVEILLANCE REQUIREMENTS        TSR 3.3.7.1 requires that the availability of the LEFM be verified prior to its use for the performance of SR 3.3.1.2. The self diagnostic features of the LEFM CheckPlus are used for this surveillance. If the LEFM NORMAL/MAINTENANCE/FAIL mode status indication as displayed on the plant computer system is not in FAIL or MAINTENANCE status, it is considered operable (i.e., in NORMAL status mode).
REFERENCES          1. License Amendment Request TVA-WBN-TS-19-06, increases the licensed power for operation of WBN Unit 2 to 3459 MWt, Docket No. 50-391.
Watts Bar - Unit 2                        B 3.3-22                                  Revision 14 Technical Requirements
 
Power Distribution Monitoring System (PDMS)
B 3.3.9 BASES (continued)
APPLICABLE          The PDMS is used for periodic measurement of the core power SAFETY              distribution to confirm operation within design limits and periodic ANALYSES            calibration of the excore detectors. This system does not initiate any automatic protection action. The PDMS is not assumed to be OPERABLE to mitigate the consequences of a DBA or transient (References 2, 3, and 4).
TR                  TR 3.3.9 requires the PDMS to be OPERABLE with the specified number of instrument channel inputs from the plant computer for each function listed in Table 3.3.9-1. The PDMS is OPERABLE when the required channel inputs are available, the calibration data set is valid, and reactor power is > 25% RTP.
This TR ensures the OPERABILITY of the PDMS when required to monitor the power distribution within the core. The PDMS is used for periodic surveillance of the incore power distribution and calibration of the excore detectors. The surveillance of incore power distribution verifies that the peaking factors are within their design envelope (References 3 and 4). The peaking factor limits include measurement uncertainty which bounds the actual measurement uncertainty of an OPERABLE PDMS (Reference 1).
Maintaining the minimum number of instrumentation channel inputs ensures the uncertainty is bounded by the uncertainty methodology.
Similarly, when THERMAL POWER is  25% RTP, then the accuracy of the adjustment provided by the Self-Powered Detector (SPD) elements to the measured PDMS power distribution may not be bounded by the uncertainties documented in Reference 1.
APPLICABILITY      The PDMS must be OPERABLE when it is used for calibration of the Excore Neutron Flux Detection System, monitoring the QPTR, measurement of FN  H and FQ(Z), or verifying the position of a rod with inoperable position indicators.
Watts Bar - Unit 2                        B 3.3-28                                    (continued)
Technical Requirements                                                                Revision 16
 
Power Distribution Monitoring System (PDMS)
B 3.3.9 BASES (continued)
ACTIONS            A.1 The Required Action A.1 has been modified by a Note stating that the provisions of TR 3.0.3 do not apply.
With THERMAL Power  25% RTP or with one or more required channel inputs inoperable or unavailable to the PDMS, the PDMS must not be used to obtain an incore power distribution measurement. Therefore, the Required Action A.1 prohibits the use of the inoperable system for the applicable monitoring or calibration functions.
TECHNICAL          TSR 3.3.9.1 SURVEILLANCE REQUIREMENTS        Performance of the CHANNEL CHECK ensures that a gross instrumentation failure 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 two instrument channels could be an indication of excessive instrument drift in one of the channels.
A CHANNEL CHECK will detect gross channel failure, thus it is a key to verifying the instrumentation continues to operate properly between each CHANNEL CALIBRATION. Agreement criteria are determined by the unit 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. If the channels are within the criteria, it is an indication that the channels are OPERABLE.
The Frequency of 24 hours is sufficient considering the PDMS provides automatic validation of the channel inputs and either discards the inoperable channel input or declares itself inoperable, but at the same time ensures that the required channel inputs to the PDMS are manually verified to be valid within a reasonable time frame prior to using the PDMS to obtain an incore power distribution measurement.
(continued)
Watts Bar - Unit 2                            B 3.3-29                                Revision 16 Technical Requirements}}

Latest revision as of 01:50, 18 November 2024

Periodic Submission for Changes Made to the Technical Specification Bases and Technical Requirements Manual
ML22131A168
Person / Time
Site: Watts Bar  Tennessee Valley Authority icon.png
Issue date: 05/11/2022
From: Anthony Williams
Tennessee Valley Authority
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
WBL-22-026
Download: ML22131A168 (335)
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