Text

Periodic Submission for Changes Made to the WBN Technical Specification Bases and Technical Requirements Manual
	ML19078A109
Person / Time
Site:	Watts Bar
Issue date:	03/19/2019
From:	Anthony Williams; Tennessee Valley Authority
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	WBL-19-023
	Download: ML19078A109 (189)
	v • d • e

Tennessee Valley Authority, Post Office Box 2000 Spring City, Tennessee 37381 wBL-19-023 March 19, 2019 10 cFR 50.4 10 CFR 50 71(e)

U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555-0001 Watts Bar Nuclear Plant, Units 1 and2 Facility Operating License Nos. NPF-90 and NPF-96 NRC Docket Nos. 50-390 and 50-391

Subject:

Watts Bar Nuclear Plant Units 1 and 2 - Periodic Submission for Changes Made to the WBN Technical Specification Bases and Technical Requirements Manual

References:

TVA letter to NRC, "Watts Bar Nuclear Plant (WBN) Units I and 2 -

Periodic Submission for Changes Made to the WBN Technical Specification Bases and Technical Requirements Manual," dated November 2, 2017 (ML17306A802)

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 and2 Technical Specification (TS) Bases and to provide copies of changes to the Unit 1 and 2 Technical Requirements Manual (TRM). These changes have been implemented at WBN during the period since the last update (Reference 1). Copies of the TS Bases, Revisions 139 through 149 for Unit 1 and Revisions 12 through 18 for Unit 2, are provided in accordance with WBN Units 1 and2 TS Section 5.6, "Technical Specifications (TS)

Bases Control Program." ln addition, copies of changes to the WBN Units 1 and 2 TRM, Revisions 65 and 66 for Unit 1 and Revisions 8 through 10 for Unit 2, are provided in accordance with WBN 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. Both control programs require such changes to be provided to the NRC on a frequency consistent with Title 10 of the Code of Federal Regulations (10 CFR) 50.71(e). The WBN TS Bases and TRM updates for the table of contents and change pages are provided in the enclosures.

U.S. Nuclear Regulatory Commission Page 2 March 19, 2019 There are no new regulatory commitments in this letter. Should you have questions regarding this submittal, please contact Kim Hulvey, Site Licensing Manager, at (423) 365-7720.

ctfully, Anthony L. Williams lV Site Vice President Watts Bar Nuclear Plant

Enclosures:

1 - WBN Unit 1 Technical Specification Bases - Table of Contents 2 - WBN Unit 1 Technical Specifications Bases - Changed Pages 3 - WBN Unit 1 Technical Requirements Manual - Table of Contents 4 - WBN Unit 1 Technical Requirements Manual - Changed Pages 5 - WBN Unit 2 Technical Specification Bases - Table of Contents 6 - WBN Unit 2 Technical Specifications Bases - Changed Pages 7 - WBN Unit 2 Technical Requirements Manual - Table of Contents 8 - WBN Unit 2 Technical Requirements Manual - Changed Pages cc (Enclosures):

NRC RegionalAdministrator - Region ll NRC Senior Resident lnspector NRR Project Manager

ENCLOSURE 1 WBN UNIT { TECHNICAL SPECIFICATION BASES TABLE OF CONTENTS

TABLE OF CONTENTS B 2.0 SAFEW LIMITS (SLs)............ ..32.0-1 B 2.1 .1 Reactor Core SLs ....8 2.0-1 B 2.1 .2 Reactor Coolant System (RCS) Pressure SL ...8 2.0-g B 3.0 LTMTTING CONDlTION FOR OPERATION (LCO) App1rCABt11Ty............................8 3.0_1 B 3.0 SURVElLIANCE REQUTREMENT (SR) AppLtCABlLlTY......... .............B 3.0-10 B 3.1 REACTIVIWCONTROLSYSTEMS ....8 3.1-1 B 3.1.1 SHUTDOWN MARGTN (SDM) T"w > 200"F .....8 3.1_1 B 3.1 .2 SHUTDOWN MARGIN (SDM) T"w < 200"F .....8 3.1_7 B 3.1 .3 Core Reactivity................. .............B 9.1-12 B 3.1 .4 ModeratorTemperature Coefficient (MTC).......... ..................8 3.1-1g B 3.1.5 Rod Group Alignment Limits........... ...................8 3.1-24 B 3.1.6 Shutdown Bank lnsertion Limits ...8 3.1-35 B 3.1 .7 Control Bank lnsertion Limits........ B 3.1.40 B 3.1 .8 Rod Position lndication .................8 3.1.48 B 3.1 .9 PHYSICS TESTS Exceptions MODE 1 .................. ...............8 3.j-5S B 3.1 .10 PHYSICS TESTS Exceptions MODE 2 .................. ...............8 9.1fi2 B 3.2 POWER DISTRIBUTION LIMITS.. ...,....83,2.1 B 3.2.1 Heat Flux Hot Channel Factor (FO(Z))........ ......89.2-1 B 3.2.2 Nuclear Enthalpy Rise Hot Channel Factor(FNU)......... ................B 9.2-12 B 3.2.3 AXTAL FLUXDTFFERENCE (AFD).......... .........8 s.2-1s B 3.2.4 QUADRANT POWER TlLT RATlO (OPTR) .....83.2_24 B 3.3 INSTRUMENTATION .......8 3.3-1 B 3.3.1 ReactorTrip System (RTS) lnstrumentation .....8 3.3-1 B 3.3.2 Engineered Safety Feature Actuation System (ESFAS) lnstrumentation..........,..... ....................8 3.3+t B 3.3.3 PostAccident Monitoring (PAM) lnstrumentation............... ...8 3.3-121 B 3.3.4 Remote Shutdoivn System ...........8 g.g-141 B 3.3.5 Loss of Porer (LOP) Diesel Generator (DG)

Start lnstrumentation........ ......89.3-147 B 3.3.6 ContainmentVent lsolation lnstrumentation............ ..............8 3.3-154 B 3.3.7 Control Room Emergency Ventilation System (CREVS) Actuation lnstrumentation................ ................B 3.3-163 B 3.3.8 Auxiliary Building Gas Treatment System (ABGTS)

Actuation lnstrumentation............... ............B 3.9-111 (continued)

Watts Bar-Unit 1 Revision 90

TABLE OF CONTENTS (continued)

B 3.4 REACTOR COOl-ANT SYSTEM (RCS).......... ..........8 3.4-1 B 3.4.1 RCS Pressure, Temperature, and Floru Departure from Nucleate Boiling (DNB) 1imits........... .B 3.4-1 B 3.4.2 RCS Minimum TemperatureforCritical'ty................. .............8 3.4 B 3.4.3 RCS Pressure and Temperature (P/T) Limits........... .............8 3.4-9 B 3.4.4 RCS Loops-MODES 1and2......... ................8 3.4-17 B 3.4.5 RCS Loops-MODE 3.................. ...................83.+21 B 3.4.6 RCS Loops-MODE4.................. ...................8 3.4-27 B 3.4.7 RCS Loops-MODE 5, Loops Fi11ed............ ....8 3.4-33 B 3.4.8 RCS Loops-MODE 5, Loops Not Filled..... ....B 3.4-38 B 3.4.9 Pressurizer.. .............8 3.44'l B 3.4.10 Pressurizer SafetyVa1ves................... ...............8 3A4 B 3.4.11 Pressurizer Power Operated Relief Valves (PORVS) B 3.4-51 B 3.4.12 Cold Overpressure Mitigation System (COMS) B 3.4-58 B 3.4.13 RCS OperationalLEAI(AGE................... ...........8 3.4-74 B 3.4.14 RCS Pressure lsolation Valve (PlV) Leakage....... .................B 3.4-81 B 3.4.15 RCS Leakage Detection lnstrumentation................ ...............B 3.4-87 B 3.4.16 RCS SpecificActivity ....................8 3.4-93 B 3.4.17 Steam Generator (SG) Tube lntegrity ...............8 3.4-99 B 3.5 EMERGENCY CORE COOLTNG SYSTEMS (ECCS)........ ............8 3.5-1 B 3.5.1 Accumulators ...........8 3.5-1 B 3.5.2 ECCS-Operating......... ..............B 3.5-10 B 3.5.3 ECCS-Shutdown........... ............B 3.5-20 B 3.5.4 Refueling Water Storage Tank (RWST) ............8 3.5-24 B 3.5.5 Seallnjection F1ow............. ...........8 3.5-31 B 3.6 CoNTAlNMENT SYSTEMS.................. B 3.6-1 B 3.6.1 Contiainment .............B 3.6-1 B 3.6.2 ContainmentAir Locks..... .............8 3.6 B 3.6.3 Contiainment lsolation Valves .......8 3.6-14 B 3.6.4 Containment Pressure...... ............8 3.6-28 B 3.6.5 Containment Air Temperature .............. .............8 3.6-31 B 3.6.6 Containment Spray Systems.................. ...........8 3.6-35 B 3.6.7 Hydrogen Recombiners - Deleted B 3.643 B 3.6.8 Hydrogen Mitigation System (HMS) ..................8 3.649 B 3.6.9 Emergency Gas Treatment System (EGTS)........ ..................8 3.6-55 B 3.6.10 Air Retum Sptem (ARS)........... ...8 3.660 B 3.6.11 lce Bed B 3.65 B 3.6.12 lce Condenser Doors...... ..............8 3.6-74 B 3.6.13 Divider Barrier lntegrity .................8 3.6-84 B 3.6.14 Containment Recirculation Drains.......... ...........8 3.6-90 B 3.6.15 Shield 8ui1din9.................. ............8 3.6-95 (continued)

Watts Bar-Unit 1 Revision 82,94

TABLE OF CONTENTS (continued)

B 3.7 Pr-ANT SYSTEMS... .........8 3.7-1 B 3.7.1 Main Steam SafetyValves (MSSVs)... ..............8 3.7-1 B 3.7.2 Main Steam lsolation Valves (MSlVs)....... ........89.2-l B 3.7.3 Main Feedwater lsolation Valves (MFlVs) and Main Feedwater Regulation Valves (MFRVs) and Associated Bypass Va1ves.......... .........8 g.Z-19 B 3.7.4 Atmospheric Dump Valves (ADVs) ...................8 3.1-20 B 3.7.5 Auxiliary Feedwater (AFW System......... ..........8 3.7-24 B 3.7.6 Condensate Storage Tank (CST). B3.t-y B 3.7.7 Component Cooling Sptem (CCS)........... ........8 3.7-38 B 3.7.8 Essential Raiv Cooling Water (ERCW System B3.l4g B 3.7.9 Ultimate Heat Sink (UHS).......... ...83.14A B 3.7.10 ControlRoomEmergencyVentilationSystem(CREVS)...... .83.7-51 B 3.7.11 Control Room Emergency Air Temperature ControlSystem (CREATCS). .B 3.7-58 B 3.7.12 Auxiliary Building Gas Treatment System (ABGTS)..... .........8 g.T$2 B 3.7.13 Fuel Storage PoolWater Leve1............ ..............8 3.78 B 3.7.14 Secondary Specific Ac1ivity................. ...............8 3.1-11 B 3.7-15 Spent FuelAssembly Storage....... ....................B g.t-75 B 3.7-16 Component Cooling Sptem (CCS) - Shutdorn.... ................8 g1-Za B 3.7-17 Essential Raw Cooling Water (ERCW System Shutdown.... B 3.7-83 B 3.8 ELECTRICAL POWER SYSTEMS .......8 3.8-1 B 3.8.1 AC Sources-Operating..... .........B 3.9-1 B 3.8.2 AC Sources-Shutdown .............B 3.8-37 B 3.8.3 Diesel Fuel Oil, Lube Oil, and Starting Air................ ..............8 3.8-43 B 3.8.4 DC Sources-Operating .............8 3.8-54 B 3.8.5 DC Sources-Shutdown .............8 3.9-70 B 3.8.6 Battery Cell Parameters.................. ...................8 g.g-74 B 3.8.7 lnverters-Operating....... ............8 3.8-81 B 3.8.8 lnverters-Shutdown....... ............8 3.8-85 B 3.8.9 Distribution Systems-Operating....... ..............8 3.8-89 B 3.8.10 Distribution Systems-Shutdown....... ..............B 3.8-99 B 3.9 REFUELTNG OPERATlONS.................. ....................B 3.9-1 B 3.9.1 Boron Concentration........ .............B 3.9-1 B 3.9.2 Unborated Water Source lsolation Valves ........B 3.9-5 B 3.9.3 Nuclear lnstrumentation................ B 3.9-8 B 3.9.4 Deleted........ .............8 3.9-12 B 3.9.5 Residual Heat Removal (RHR) and Coolant B 3.9.6 Circulation

- High Water Level............

Residual Heat Removal (RHR) and Coolant

......8 3.9-17 B 3.9.7 Circulation

- Low Water Leve1............

Refueling Cavity Water Level ...........

.......8 3.9-21

.................8 3.9-25 B 3.9.8 Deleted........ .............8 3.9-29 B 3.9.9 SpentFuelPoolBoronConcentration................... .................83.9-33 B 3.9.10 DecayTime. .............B 3.9-35 Watts Bar-Unit 1 Revision 123

LIST OF TABLES Table No. Title Paoe paoe B 3.8.1-2 TS Action or Surveillance Requirement (SR)

Contingency Actions ......8 3.8-36 B 3.8.9-1 AC and DC ElectricalPorver Distribution Systems....... ....................B 3.8-98 Watts Bar-Unit 1 iv

LIST OF FIGURES Fioure No. Title 82.1.1-1 Reactor Core Safety Limits vs Boundary of Protection .................82.0-l 83.1.7-1 ControlBank lnsertion vs Percent RTP............. .........8 gjl47 83.2.1-'l K(z) - Normalized Fq(z) as a Function of Core Hei9ht.......... .............8 9.2-11 B 3.2.3-1 AXIAL FLUX DIFFERENCE Acceptabte Operation Limits as a Function of RATED THERMAL POWER ,8 3.2-29 Watts Bar-Unit 1

LIST OF ACRONYMS (Page 1 of 2)

Acronlrm Title ABGTS Auxiliary BuiHing 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 lsolation Valve MSSV Main Steam Safety Valve MTC Moderator Tem peratu re Coefficient NMS Neutron Monitoring System ODCM Offsite Dose Calculation Manual PCP Process Control Program PDMS Power Distribution Monitoring System PIV Pressure lsolation 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 lnjection SL Safety Limit SR Surveillance Requirement UHS Ultimate Heat Sink Watts Bar-Unit 1 vil

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i 90 B 2.0-1 0 ii 94 B 2.0-2 140 iii 123 B 2.0-3 0 iv 0 82.04 59 V 0 B 2.0-5 108 vi 104 B 2.0-6 140 vii 0 B 2.0-7 0 viii 149 B 2.0-8 0 ix 145 B 2.0-9 0 x 149 B 2.0-10 0 xi 135 B 2.0-11 108 xii 148 B 2.0-12 0 xiii 142 B 3.0-1 133 xiv 147 B 3.0-2 0

)(V 139 B 3.0-3 0 xvi 125 B 3.0-4 68

><vii 19 B 3.0-5 68 xviii 32 B 3.0-6 68 xix 46 B 3.0-7 0

)o( 60 B 3.0-8 141 xxi 68 B 3.0-9 133

>uii 75 B 3.0-9a 133 rc<iii 85 B 3.0-9b 133 xxiv 94 B 3.0-10 0

)ou 102 B 3.0-1 1 144 ncvi 110 B 3.0-12 53

>o<vii 120 B 3.0-13 68

>o<viii 129 B 3.0-14 68 xxix 141 B 3.1-1 0 149 B 3.1-2 0 B 3.1-3 0 B 3.14 68 B 3.1-5 0 B 3.1-6 0 B 3.1-7 0 B 3.1-8 0 B 3.1-9 68 B 3.1-10 0 B 3. 1-11 0 B 3. 1-12 0 B 3.1-13 32 B 3. 1-14 0 B 3.1-15 0 Watts Bar-Unit 1 viii Revision 149

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REVISIONS ISSUED SUBJECT NPF-20 11-09-95 Low power Operating License Revision 1 12-08-95 Slave Relay Testing NPF-90 02-07-gO Fuil power Operating License Revision 2 (Amendment'1) 12-08-95 Turbine Driven AFW Pump Suction Requirement Revision 3 03-27-96 Remove Cold Leg AccumulatorAlarm Setpoints Revision 4 (Amendment 2) 06-13-96 lce Bed Surveillance Frequency And Weight Revision 5 07-03-96 Containment Airlock Door lndication Revision 6 (Amendment 3) 09-09-96 lce Condenser Lower Inlet Door Surveillance Revision 7 09-28-96 Clarification of COT Frequency for COMS Revision I 11-21-96 Admin Control of Containment lsol. Valves Revision I 04-29-97 Switch Controls For Manual Cl-Phase A Revision 10 (Amendment 5) 05-27-92 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 Cycle2 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 (VisuaUaudible indication)

Revision 18 (Amendment '11) 09-09-98 Relocation of F(O) Penalty to COLR Revision 19 (Amendment 12) 10-19-98 Online Testing of the DieselBatteries and Performance of the 24 Hour Diesel Endurance Run Watts Bar-Unit 1 xvii 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 1 1-30-98 Clarification to lce Condenser Door ACTIONS and door lift tests, and lce 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 lnoperable 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 (lCS). 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.

AIso 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 lnoperable MSSVS.

Revision 32 04-13-00 Clarification to Reflect Core Reactivity and MTC Behavior.

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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 lce Condenser sampling and flow channel surveillance requ irements 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 - lmplementation of Best Estimate LOCA analysis.

Amendment 28 - Revision of LCO 3.1 .10, "Physics Tests Exceptions - Mode 2."

Revision 40 09-28-00 Clarifies WBN's 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 FIow Meter (LEFM)

Revision 42 03-07-01 Clarify Operability Requ irements for Pressurizer PORVs Revision 43 05-29-01 Change CVI Response Time from 5 to 6 Seconds Revision 44 (Amendment 33) 01-31 -02 lce 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 lsolation Requirements in LCO 3.9.4 Watts Bar-Unit 1 xrx 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 operationat LEAKAGE and SG Alternate Repair Criteria forAxial Outside Diameter Stress Corrosion Cracking (oDSCC)

Revision 48 (Amendment 36) 03-06-02 lncrease 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 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months 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 pCi/gm to <0.265 pCi/gm.

Revision 53 (Amendment42) 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 detete St Hot Leg lnjection lines from SR until U1C5 outage.

Revision 55 05-22-03 Editorialcorrections (PER 02-015499),

correct peak containment pressure, and revise l-131 gap inventory for an FHA.

Revision 56 07-10-03 TS Bases for SRs 3.8.4.8 through SR 3.8.4. 1 0 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 lnverters.

Revision 59 (Amendment 46) 09-30-03 Address new DNB Correlation in 82.1.1 and B,3.2.12 for Robust FuetAssembty (RFA)-2.

Revision 60 (Amendment 47) 10-06-03 RCS Flow Measurement Using Elbow Tap Flow Meters (Revise Tabte 3.3.1-i(10) &

sR 3.4.1.4).

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REVISIONS ISSUED SUBJECT Revision 61 (Amendments 40 and 48) 10-14-03 lncorporated 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-03-1e)

Revision 63 12-08-03 The contingency actions listed in Bases Table 3.8.1-2were reworded to be consistent with the NRC Safety Evaluation that approved Tech Spec Amendment 39.

Revision 64 (Amendment 50) 03-23-04 lncorporated 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 3.7.1 1, "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 lnverters (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.

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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 prim ary 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 lndicator (RPl) (TS-03-12).

Revision 7 1 (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 09-1 1-06 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-ll of the Vital lnverters (DCN 51370).

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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-Gontents)

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-lV of the Vital lnverters (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-l of the Vital lnverters (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 lnverters (DCN 51370).

Revision 79 (Amendment 60, 61 and 1 1-03-06 Steam Generator Narrow Range Level

64) lndication lncreased from 60/o to 32o/o WBN-T5-05-06) Bases Sections 3.4.5, 3.4.6, and 3.4.7.

Revision 80 1 1-08-06 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 lncrease lce Weight in lce Condenser to Support Replacement Steam Generators (WBN-TS-05-0e) [SGRP]

Revision 82 (Amendment 65) 11-17-06 Steam Generator (SG) Tube lntegrity (wBN-rs-05-10) [SGRP]

Revision 83 1 1-20-06 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 1 1-30-06 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.

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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 !SSUED SUBJECT Revision 86 01-31-08 Figure 3.7 .1 5-1 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 lmplemented 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 signa!.

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 g0 (Amendment 68) 10-02-08 lssuance 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 lnstrumentation" 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.

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REVISIONS ISSUED SUBJECT Revision 95 03-05-2009 Corrected an error in SR 3.3.2.6 which referenced Function 6.9 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 lnstrumentation." This is associated with AFW automatic start on trip of all main feedwater pumps. ln 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, "lnverters - Operding'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 7O4.

Amendment 79 revised LCO 3.6.3 to allow verification by administrative means isolation devices that are locked, sealed, or othenrise secured.

Amendment 81 revised the allowed outage time of Action B of LCO 3.5.1 from t hourto 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Revision 99 10-09-2009 Bases Revision 99 incorporated Westinghouse Technical Bulletin (TB) 08-04.

Revision 100 1'l-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 manualactions.

Revision 101 02-09-2010 Bases Revision 101 implemented DCN 52216-Athat will place both trains of the EGTS pressure control valve's hand switches in A-AUTO and will result in the valves opening upon initiation of the Containment lsolation phase A (ClA) signal.

They will remain open independent of the annulus pressure and reset of the ClA.

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 Revision 102

TECHNICAL SPEGIFICATION BASES . REVISION LISTING (This listing is an administrative tool maintained by WBN Licensing and may be updated without formally revising the Technical Specification Bases Tabte-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 100o/o capacity subsysteffis, 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 1 1 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 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months in TSB 2.2.4. Also, corrected error in SR 3.3.2.4 in the reference to Table 3.3.1-1 . lt 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 1 10 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 xxvt Revision 1 10

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 pCi/gm and refers to the controls being placed in AOI-28.

Revision 112 05-24-2011 DCN 55076 replaces the existing fou r 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 25o/o 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 1 16 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 .1 0, Control Room Emergency Ventilation System (CREVS), to make the TS Bases consistent with TS 3.7.10, Condition E Revision 1 18 01-30-2013 Revises TS Bases 3.4.16, Reactor Coolant System (RCS) to change the dose equivalent l-1 31 spike limit and the allowable value for control room air intake radiation monitors.

Revision 1 19 08-17-2A13 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 xxvii Revision 1 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-Gontents)

REVISIONS ISSUED SUBJECT Revision 120 01-23-2014 Revised the References to TS Bases 3.1.9, PHYSICS TESTS Exceptions - Model , 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. ls 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) 01-1 4-2014 Revises TS Bases 3.7.1 0, Control Room Emergency Ventilation System (CREVS) to make the TS Bases consistent with TS 3.7.1 0, Actions E, F, G, and H.

Revision 123 (Amendment 104) 03-16-2016 Amendment 104, TSB Revision 123 adds TS 83 .7.16, "Component Cooling System (CCS) - Shutdown" and adds TS 83 .7.17 ,

"Essential Raw Cooling Water (ERCW)

System - Shutdown.

Revision 124 02-12-2016 Revises TS Bases Table 83.8.9-1 , "AC and DC Electrical Power Distribution Systems,"

the second Note.

Revision 125 (Amendment 84, 102, 03-16-2016 Revises TS Bases Section 83.8-1, "AC 1 03) Sources-Operating. "

Revision 126 03-18-2016 Revises TS Bases Section B3.7.7, "Component Cooling System" the 1B and 28 surge tank sections.

Revision 127 04-18-2016 Revises TS Bases Section B 3.6.4, "Containment Pressure" and 83.6.6, "Containment Spray System to change the maximum peak pressure from a LOCA of 9.36 psig.

Watts Bar-Unit 1 xxviii 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 83.6.8, "Hydrogen Mitigation System (HMS)", to delete sentence regarding Hydrogen Recombiners that are abandoned.

Revision 129 08-19-16 Revises TS Bases Section 3.6.1 5, .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) 01-1 3-17 Revises TS Bases Section 3.5.4, " Refueling Water Storage Tank (RWST), Applicable Safety Analyses" Revision 132 (Amendment 1 10) 01-17-17 Revises TS Bases Section 3.8.1, ,,AC Sources -Operating" Revision 133 (Amendment 111) 03-1 3-17 Adds TS Bases Section 3.0.8 for I noperability of Snubbers.

Revision 134 (Amendment 112) 04-25-17 Revise TS Bases Section 3.7.1 1 Action A.1 regarding CREATCS.

Revision 135 05-17-17 Revises TS Bases Section 83.3.3, 'PAM lnstrumentation" Revision 136 (Amendment 1 13) 05-17-17 Revises TS Bases Section 83.7.7 "CCS" Revision 137 (Amendment 114) 07 17 Revises TS Bases Section B SR 3.0.2 to add a one-time extension for the surveillance interval.

Revision 138 (Amendment 1 15) 11-2-17 Revises TS Bases to adopt the TSTF-522 to revise ventilation system surveillance requirements to operate for 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months 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 03-08-1 I 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 100o/o capacity subsystems, each capable of supporting both trains of TS equipment.

Watts Bar-Unit 1 xxtx 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 04-06-1 I 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 08-16-1 8 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 10-1 1-18 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 Iow 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 2-06-19 Revises TS Bases 3.3.1, "Reactor Trip System I nstrumentation" Watts Bar-Unit 1 Revision 149

ENCLOSURE 2 WBN UNIT 1 TECHNICAL SPEGIFICATION BASES CHANGED PAGES

Reactor Core SLs B 2.1 .1 BASES BACKGROUND DNB is not a directly measurable parameter during operation; therefore, (continued) THERMAL POWER, reactor coolant temperature, and pressure are related to DNB through critical heat flux (cHfl_correlations. The primary DNB correlations are the WRB-1 correlation (Ref. 7) foTVANTAGE 5H and VANTAGE+ fueland the WRB-2M correlation (Ref.8) for RFA-2 fuelwith lFMs. These DNB correlations take credit for significant improvement in the accuracy of the CHF predictions. The W-3, ABB-NV, or WLOP CHF correlations (Refs. 9, 10, and '1'1) are used for conditions outside the range of the WRB-I correlation for VANTAGE 5H and VANTAGE+ fuel or the WRB-2M correlation for RFA-2 fuelwith lFMs.

The proper functioning of the Reactor Protection System (RPS) and steam generator safe$ valves prevents violation of the reactor core SLs.

APPLICABLE The fuel cladding must not sustain damage as a result of normal operation SAFETY ANALYSES and AOOs. The reactor core SLs are established to preclude violation of the following fuel design criteria:

a. There must be at least 95% probability at a 95% confidence level (the 95/95 DNB criterion) that the hot fuel rod in the core does not experience DNB; and

b. The hot fuel pellet in the core must not experience centerline fuel melting.

The Reactor Trip System setpoints (Ref. 2), in combination wlth all the LCOs, are designed to prevent any anticipated combination of transient conditions for Reactor Coolant System (RCS) temperature, pressure, and THERMAL POWER levelthat would result in a departure from nucleate boiling ratio (DNBR) of less than the DNBR limit and preclude the existence of flow instabilities.

Automatic enforcement of these reactor core SLs is provided by the following functions:

a. High pressurizer pressure trip;

b. Low pressurizer pressure trip;

c. Overtemperature AT trip (continued)

Watts Bar-Unit 1 B 2.0-2 Revision 59, 140 Amendment 46

Reactor Core SLs B 2.1 .1 BASES References 4. WCAP -9272-P-A, "Westinghouse Reload Safety Evaluation (continued) Methodology," July 1 985.

5. Title 10, Code of Federal Regulations, Part 50.72, "lmmediate Notification Requirements for operating Nuclear Power Reactors."

6. Title 10, Code of Federal Regulations, Part 50.73, "Licensee Event Report System."

7. WCAP-8762-P-A, "New Westinghouse Correlation WRB-1 for Predicting Critical Heat Flux in Rod Bundles with Mixing Vane Grids," July 1984.

8. WCAP-1 5025-P-A, "Modified WRB-2 Correlation, WRB-2M, for Predicting Critical Heat Flux in 17 x 17 Rod Bundles with Modified LPD Mixing Vane Grids," April 1999.

9. Tong, L. S., "Boiling Crisis and Critical Heat Flux," AEC Critical Review Series, TID-25887, 1972.

10. Tong, L. S., "Critical Heat Fluxes on Rod Bundles," in "Two-Phase FIow and Heat Transfer in Rod Bundles," Engineers, New pages 31 through 41 ,

American Society of Mechanical York, 1969.

11. WCAP-14565-P-A Addendum z-P-A (Proprietary) / WCAP-1 5306-NP-A Addendum 2-NP-A (Non-Proprietary), "Addendum 2 to WCAP-14565-P-A, Extended Application of ABB-NV Correlation WLOP for PWR Low Pressure Applications, " April 2008.

Watts Bar-Unit 1 B 2.0-6 Revision 13, 59, 140 Amendment 7, 46

LCO Applicability B 3.0 BASES LCO 3.0.6 the entry into multiple support and supported systems' LCOs' Conditions and (continued) Required Actions are eliminated by providing all the actions that are necessary to ensure the unit is maintained in a safe condition in the support system's Required Actions.

However, there are instances where a support system's Required Action may either direct a supported system to be declared inoperable or direct entry into Conditions and Required Actions for the supported system. This may occur immediately or after some specified delay to perform some other Required Action. Regardless of whether it is immediate or afrer some delay, when a support system's Required Action directs a supported system to be declared inoperable or directs entry into Conditions and Required Actions for a supported system, the applicable Conditions and Required Actions shall be entered in accordance with LCO 3.0.2.

Specifi cation 5.7 .2.1 8, "Safety Function Determination Program (SFDP),'

ensures loss of safety function is detected and appropriate actions are taken.

Upon entry into LGO 3.0.6, an evaluation shall be made to determine if loss of safety function exists. Additionally, other limitations, remedial actions, or compensatory actions may be identified as a result of the support system inoperability and conesponding exception to entering supported system Conditions and Required Actions. The SFDP implements the requirements of LCO 3.0.6.

Cross train checks to identifiT a loss of safety function for those support systems that support multiple and redundant safety systems are required. The cross train check verifies that the supported systems of the redundant OPEMBLE support system are OPEMBLE, thereby ensuring safety function is retained. lf this evaluation determines that a loss of safety function exists, the appropriate conditions and Required Actions of the LCo in which the loss of safety function exists are required to be entered.

(continued)

Watts Bar-Unit 1 B 3.0-8 Revision 141

SR Applicability B 3.0 BASES (continued) sR 3.0.2 SR 3.0.2 establishes the requirements for meeting the specified Frequency for Surveillances and any Required Action with a Completion Time that requires the periodic performance of the Required Action on a "once per . . ." interval.

SR 3.0.2 permits a25o/o extension of the intervalspecified in the Frequency.

This extension facilitates Surveillance scheduling and considers plant operating conditions that may not be suitable for conducting the Surveillance (e.9.,

transient conditions or other ongoing Surveillance or maintenance activities). For each of the SRs listed in Table SR 3.0.2-1 the specified Frequency is met if the Surveillance is performed on or before the date listed on Table SR 3.0.2-1. The Surveillance Frequency extension limits expire on the dates listed in Table SR 3.0.2-1 or when the unit enters MODE 5, whichever occurs first.

The25% extension does not significantly degrade the reliability that results from performing the Surveillance at its specified Frequency. This is based on the recognition that the most probable result of any particular Surveillance being performed is the verification of conformance with the SRs. The exceptions to SR 3.0.2 are those Surveillances for which the 25o/o extension of the interval specified in the Frequency does not apply. These exceptions are stated in the individual Specifications. The requirements of regulations take precedence over the TS. Therefore, when a test interval is specified in the regulations, the test interval cannot be extended by the TS, and the surveillance requirement will include a note in the frequency stating, 'SR 3.0.2 does not apply." An example of an exception when the test interval is not specified in the regulations, is the discussion in the Containment Leakage Rate Testing Program, that SR 3.0.2 does not apply. This exception is provided because the program already includes extension of test intervals.

As stated in SR 3.0.2, the25o/o extension also does not apply to the initial portion of a periodic Completion Time that requires performance on a "once per . . ."

basis. The 25olo extension applies to each performance after the initial performance. The initial performance of the Required Action, whether it is a particular Surveillance or some other remedial action, is considered a single action with a single Completion Time. One reason for not allowing lhe25%

extension to this Completion Time is that such an action usually verifies that no loss of function has occuned by checking the status of redundant or diverse components or accomplishes the function of the inoperable equipment in an alternative manner.

The provisions of SR 3.0.2 are not intended to be used repeatedly merely as an operational convenience to extend Surveillance intervals (other than those consistent with refueling intervals) or periodic Completion Time intervals beyond those specified, with the exception of surveillances required to be performed on a 31-day frequency. For surveillances performed on a 31-day frequency, the normal surveillance interval may be extended in accordance with Specification 3.0.2 cyclically as required to remain synchronized to the 13-week maintenance work schedules. This practice is acceptable based on the results of an evaluation of 31-day frequency surveillance test histories that demonstrate that no adverse failure rate changes have occurred nor would be expected to develop as a result of cyclical use of surveillance interval extensions and the fact that the total number of 31{ay frequency surveillances performed in any one-year period remains unchanged.

(continued)

Watts Bar-Unit 1 B 3.0-1 1 Revision 10, 114, 137, 144 Amendment 5, 114, 121

Rod Group Alignment Limits B 3.1 .5 B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.5 Rod Group Alignment Limits BASES BACKGROUND The OPEMBILITY (i.e., trippability) of the shutdown and control rods is an initial I assumption in all safety analyses that assume rod insertion upon reactor trip.

Maximum rod misalignment is an initial assumption in the safeg analysis that directly affects core power distributions and assumptions of available SDM.

The applicable criteria for these reactivity and power distribution design requirements are 10 CFR 50, Appendix A, GDC 10, "Reactor Design," and GDC 26, "Reactivity Control System Redundancy and Capability," (Ref. 1), and 10 CFR 50.46,'Acceptance Criteria for Emergency Core Cooling Systems for Light Water Nuclear Power Reactors" (Ref. 2).

Mechanical or electrical failures may cause a control or shutdown rod to become inoperable or to become misaligned from its group. Rod inoperability or misalignment may cause increased power peaking, due to the asymmetric reactivity distribution and a reduction in the total available rod worth for reactor shutdown. Therefore, rod alignment and OPERABILIry are related to core operation in design power peaking limits and the core design requirement of a minimum SDM.

Limits on rod alignment have been established, and all rod positions are monitored and controlled during power operation to ensure that the power distribution and reactivity limits defined by the design power peaking and SDM limits are preserved.

Rod cluster control assemblies (RCCAs), or rods, are moved by their control rod drive mechanisms (CRDMS). Each CRDM moves its RCCA one step (approximately 5/8 inch) at a time, but at varying rates (steps per minute) depending on the signal output from the Rod Control System.

The RCCAs are divided among control banks and shutdown banks. Each bank may be further subdivided into two groups to provide for precise reactivity control (Shutdown Banks C and D have only one group each). A group consists (continued)

Watts Bar-Unit 1 B 3. 1-24 Revision 51 , 143 Amendment 120

Rod Group Alignment Limits B 3.1 .5 BASES BACKGROUND of two or more RCCAs that are electrically paralleled to step simultaneously.

(continued) Except for Shutdown Banks C and D, a bank of RCCAs consists of two groups that are moved in a staggered fashion, but ahrays within one step of each other.

There are four control banks and four shutdown banks.

The shutdown banks are maintained either in the fully inserted or fully withdrawn position. The control banks are moved in an overlap pattern, using the following withdrawal sequence: When control bank A reaches a predetermined height in the core, control bank B begins to move out with control bank A. Control bank A stops at the position of maximum withdrawal, and control bank B continues to move out. When control bank B reaches a predetermined height, control bank C begins to move out with control bank B. This sequence continues until control banks A, B, and C are at the fully withdrawn position, and control bank D is approximately halfrvay withdrawn. The insertion sequence is the opposite of the withdrawal sequence. The control rods are arranged in a radially symmetric pattern, so that control bank motion does not introduce radial asymmetries in the core power distributions.

The axial position of shutdown rods and control rods is indicated by two separate and independent systems, which are the Bank Demand Position lndication System (commonly called group step counters) and the Rod Position Indication (RPl) System.

The Bank Demand Position Indication System counts the pulses from the rod control system that moves the rods. There is one step counter for each group of rods. lndividual rods in a group all receive the same signalto move and should, therefore, all be at the same position indicated by the group step counter for that group. The Bank Demand Position lndication System is considered highly precise (t '1 step or t 5/8 inch). lf a rod does not move one step for each demand pulse, the step counter will still count the pulse and incorrectly reflect the position of the rod.

The RPI System provides an accurate indication of actual control rod position, but at a lower precision than the step counters. This system is based on inductive analog signals from a series of coils spaced along a hollow tube with a center to center distance of 3.75 inches, which is six steps. The normal (continued)

Watts Bar-Unit 1 B 3. 1-25 Revision 51 ,143 Amendment 120

Rod Group Alignment Limits B 3.1 .5 BASES BACKGROUND indication accuracy of the RPI System is t 6 steps (r 3.75 inches),

(continued) and the maximum uncertainty is t12 steps (t 7.5 inches). With an indicated deviation of 12 steps between the group step counter and RPl, the maximum deviation between actual rod position and the demand position could be 24 steps, or'15 inches.

APPLICABLE Control rod misalignment accidents are analyzed in the safety analysis SAFETY ANALYSES (Ref. 3). The acceptance criteria for addressing control rod inoperability or misalignment are that:

a. There be no violations of:

1. Specified acceptable fuel design limits, or

2. Reactor Coolant System (RCS) pressure boundary integrity; and

b. The core remains subcritical after accident transients other than a main steam line break (MSLB).

Two types of misalignment are distinguished. During movement of a control rod group, one rod may stop moving, while the other rods in the group continue. This condition may cause excessive power peaking. The second type of misalignment occurs if one rod fails to insert upon a reactor trip and remains stuck fully withdrawn. This condition requires an evaluation to determine that sufficient reactivity worth is held in the control rods to meet the SDM requirement, with the maximum worth rod stuck fully withdrawn.

Three types of analysis are performed in regard to static rod misalignment (Ref. ). The first type of analysis considers the case where any one rod is completely inserted into the core with all other rods completely withdrawn. With control banks at their insertion limits, the second type of analysis considers the case when any one rod is completely inserted into the core. The third type of analysis considers the case of a completely withdrawn single rod from a bank inserted to its insertion limit. Satisfying limits on departure from nucleate boiling ratio in each of these cases bounds the situation when a rod is misaligned from its group by 12 steps.

(continued)

Watts Bar-Unit 1 B 3. 1-26 Revision 143 Amendment 120

Rod Group Alignment Limits B 3.1.5 BASES APPLICABLE Another $pe of misalignment occurs if one RCCA fails to insert upon a SAFETY ANALYSES reactor trip in response to a main steam pipe rupture and remains stuck (continued) fully withdrawn. This condition is assumed in the evaluation to determine that the required sDM is met with the maximum worth RCCA also fully withdrawn (Ref. 5). The reactor is shutdown by the boric acid injection delivered by the ECCS.

The Required Actions in this LCO ensure that either deviations from the alignment limits will be corrected orthat THERMAL POWER wiil be adjusted so that excessive local linear heat rates (LHRs) will not occur, and that the requirements on SDM and ejected rod worth are preserved.

Continued operation of the reactor with a misaligned control rod is allowed if the heat flux hot channel factor (Fq(Z)) and the nuclear enthalpy hot channel facto(FN611) are verified to ne wltnin their limits in the COLiiand the safety analysis is verified to remain valid. When a control rod is misaligned, the assumptions that are used to determine the rod insertion limits, AFD limits, and quadrant power tilt limits are not preserved. Therefore, the limits may not preserve the design peaking factors, and Fq(Z) and FNa11 must be veiified direcfly using incore power distribution measurements. Bases Section 3.2 (Power Distribution Limits) contains more complete discussions of the relation of Fq(Z) and Fa11to the operating limits.

Shutdown and controlrod OPEMBILITY and alignment are direcfly related to power distributions and SDM, which are initialconditions assumed in safety analyses. Therefore they satisf, Criterion 2 ot '10 CFR 50.36(cx2xii).

LCO The limits on shutdown or control rod alignments ensure that the assumptions in the safe$ analysis will remain valid. The requirements on oPEFlABlLlrY ensure that upon reactor trip, the assumed reactivity will be available and will be inserted.

The control rod OPEMBILITY requirements (i.e., trippabitity) are acceptable from the alignment requirements, which ensure that the RCCAs and banks maintain the correct power distribution and rod alignment. The rod OPERABILIW requirement is satisfied provided the rod willfully insert in the required rod drop time assumed in the safety analysis. Rod control malfunctions that result in the inability to move a rod (e.9., rod lift coil failures), but that do not impact trippability, do not result in rod inoperability.

The requirement to maintain the rod alignment to within plus or minus 12 steps is conservative. The minimum misalignment assumed in safety analysis is 24 steps (15 inches), and in some cases a total misalignment from fully withdrawn to fully inserted is assumed.

(continued)

Watts Bar-Unit 1 B 3.1-27 Revision 104, 143 Amendment 82, 120

Rod Group Alignment Limits B 3.1.5 BASES LCO Failure to meet the requirements of this LCO may produce unacceptable power (continued) peaking factors and LHRs, or unacceptable SDMs, all of which may constitute initial conditions inconsistent with the safety analysis.

APPLICABILITY The requirements on RCCA OPEMBILITY and alignment are applicable in MODES 1 and2 because these are the only MODES in which neutron (or fission) power is generated, and the OPERABILITY (i.e., trippability) and alignment of rods have the potential to affect the safety of the plant. ln MODES 3, 4, 5, and 6, the alignment limits do not apply because the control rods are bottomed and the reactor is shut down and not producing fission power. ln the shutdown MODES, the OPERABILITY of the shutdown and control rods has the potential to affect the required SDM, but this effect can be compensated for by an increase in the boron concentration of the RCS. See LCO 3.1.1, "SHUTDOWN MARGIN (SDM) - T",s

> 200oF," for SDM in MODES 3 and 4, LCO 3.1.2, "Shutdown Margin (SDM)-Ta,s s 200'F" for SDM in MODE 5, and LCO 3.9.1, "Boron Concentration," for boron concentration requirements during refueling.

ACTIONS A.1.1 and A.1.2 When one or more rods are inoperable (i.e., untrippable), there is a possibility that the required SDM may be adversely affected. Under these conditions, it is important to determine the SDM, and if it is less than the required value, initiate boration untilthe required SDM is recovered. The Completion Time of t hour is adequate for determining SDM and, if necessary, for initiating boration to restore SDM.

ln this situation, SDM verification must include the worth of the untrippable rod, as well as a rod of maximum worth.

(continued)

Watts Bar-Unit 1 B 3.1-28 Revision 143 Amendment 120

Rod Group Alignment Limits B 3.1 .5 BASES ACTIONS A.2 (continued) lf the inoperable rod(s) cannot be restored to OPEMBLE status, the plant must I be brought to a MODE or condition in which the LCO requirements are not I applicable. To achieve this status, the unit must be brought to at least MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

The allowed Completion Time is reasonable, based on operating experience, for reaching MODE 3 from full power conditions in an orderly manner and wlthout challenging plant systems.

811and8.12 ,

When a rod becomes misaligned, it can usually be moved and is stilltrippable. j An alternative to realigning a single misaligned RCCA to the group average position is to align the remainder of the group to the position of the misaligned RCCA. However, this must be done without violating the bank sequence, overlap, and insertion limits specified in LCO 3.1.6, "Shutdown Bank lnsertion Limits," and LCO 3.1.7, "ControlBank lnsertion Limits." I ln many cases, realigning the remainder of the group to the misaligned rod may I not be desirable. For example, realigning control bank B to a rod that is misaligned 15 steps from the top of the core would require a significant power reduction, since controlbank D must be moved fully in and controlbank C must be moved in to approximately 100 to 115 steps.

(continued)

Watts Bar-Unit 1 B 3.1-29 Revision 143 Amendment 120

Rod Group Alignment Limits B 3.1.5 BASES ACTIONS B.1.1 and 8.1.2 (continued)

Power operation may continue with one RCCA trippable but misaligned, provided that SDM is verified within t hour. The Completion Time of t hour represents the time necessary lor determining the actual unit SDM and, if necessary, aligning and starting the necessary systems and components to initiate boration.

8.2. B.3. B.4. and B.5 For continued operation with a misaligned rod, RTP must be reduced, SDM must periodically be verified within limits, hot channelfactors (Fo(Z) and FNo6) must be verified within limits, and the safety analyses must be re-evaluated to confirm continued operation is permissible.

Reduction of power to7io/o RTP ensures that local LHR increases due to a misaligned RCCA will not cause the core design criteria to be exceeded (Ref. 6).

The Completion Time of 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months gives the operator sufficient time to accomplish an orderly power reduction without challenging the Reactor Protection System.

When a rod is known to be misaligned, there is a potentialto impact the SDM.

Since the core conditions can change with time, periodic verification of SDM is required. A Frequency of '12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is sufficient to ensure this requirement continues to be met.

Verifying that Fq(Z), as approximated by Fco(Z) and Fwo1Z1, and FN66 are within the required limits ensures that current operation al71o/o RTP with a rod misaligned is not resulting in power distributions that may invalidate safeg analysis assumptions at full power. The Completion Time of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months allows sufficient time to obtain an incore power distribution measurement and to calculate Fq(Z) and FN611.

Once current conditions have been verified acceptable, time is available to perform evaluations of accident analysis to determine that core limits will not be exceeded during a Design Basis Event for the duration of operation under these conditions. The accident analyses presented in UFSAR Change 15 (Ref. 3) that may be adversely affected will be evaluated to ensure that the analyses remain valid for the duration of continued operation under these conditions. A Completion Time of 5 days is sufficient time to obtain the required input data and to perform the analysis.

(continued)

Watts Bar-Unit 1 B 3.1-30 Revision 104, 143 Amendment 82, 120

Rod Group Alignment Limits B 3.1 .5 BASES ACTIONS c.1 (continued) when Required Actions cannot be completed within their completion Time, the unit must be brought to a MoDE or condition in which the LCo requirements are not applicable. To achieve this status, the unit must be brought to at least MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , which obviates concerns about the development of undesirable xenon or power distributions. The allowed Completion Time of 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months is reasonable, based on operating experience, for reaching MODE 3 from full power conditions in an orderly manner and without challenging the plant systems.

D.1.1 and D.'1.2 More than one control rod becoming misaligned from its group average position is not expected, and has the potential to reduce SDM. Therefore, SDM must be evaluated. One hour allows the operator adequate time to determine SDM.

Restoration of the required sDM, if necessary, requires increasing the RCS boron concentration to provide negative reactivity, as described in the Bases of LCO 3.1.1. The required Completion Time of t hour for initiating boration is reasonable, based on the time required for potentialxenon redistribution, the low probability of an accident occurring, and the steps required to complete the action. This allows the operator sufficient time to align the required valves and start the boric acid pumps. Boration will continue until the required SDM is restored.

o.2 lf more than one rod is found to be misaligned or becomes misaligned because of bank movement, the unit conditions fall outside of the accident analysis assumptions. Since automatic bank sequencing would continue to cause misalignment, the unit must be brought to a MODE or condition in which the LCo requirements are not applicable.

To achieve this status, the unit must be brought to at least MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . The allowed Completion Time is reasonable, based on operating experience, for reaching MODE 3 from full power conditions in an orderly manner and without challenging plant systems.

(continued)

Watts Bar-Unit 1 B 3.1 -31 Revision 143 Amendment 120

Rod Group Alignment Limits B 3.1.5 BASES SURVEILI-.ANCE sR 3.1.5.1 REQUIREMENTS Verification that the position of individual rod is within alignment limits at a Frequency of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months provides a history that allows the operator to detect a rod that is beginning to deviate from its expected position. lf the rod position deviation monitor is inoperable, a Frequency of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months accomplishes the same goal. The specified Frequency takes into account other rod position information that is continuously available to the operator in the control room, so that during actual rod motion, deviations can immediately be detected.

The SR is modified by a Note that permits it to not be performed for rods associated with an inoperable demand position indicator or an inoperable rod position indicator. The alignment limit is based on the demand position indicator which is not available if the indicator is inoperable. LCO 3.1.8, 'Rod Position lndication,' provides Actions to veriff the rods are in alignment when one or more rod position indicators are inoperable.

The Surveillance is modified by a Note which states that the SR is not required to be performed until t hour after associated rod motion. Control rod temperature affects the accuracy of the rod position indication system. Due to changes in the magnetic permeability of the drive shaft as a function of temperature, the indicated position is expected to change with time as the drive shaft temperature changes. The one hour period allows control rod temperature to stabilize following rod movement in order to ensure the indicated rod position is accurate.

sR 3.1.5.2 Verifying each control rod is OPERABLE would require that each rod be hipped.

However, in MODES 1 and2, hipping each controlrod would result in radialor axial power tilts, or oscillations. Exercising each individual control rod every 92 days provides increased confidence that all rods continue to be OPERABLE without exceeding the alignment limit, even if they are not regularly tripped.

Moving each control rod by 10 steps will not cause radial or axial power tilts, or oscillations, to occur. The 92 day Frequency takes into consideration other information available to the operator in the control room and SR 3.1.5.1, which is performed more frequently and adds to the determination of OPEMBILITY of the rods. Between required performances of SR 3.1.5.2 (determination of control rod OPEMBILITY by movement), if a controlrod(s) is discovered to be immovable, but remains trippable and aligned, the control rod(s) is considered to be OPERABLE. At any time, if a control rod(s) is immovable, a determination of the trippability (OPERABILITY) of the control rod(s) must be made, and appropriate action taken.

(continued)

Watts Bar-Unit 1 B 3. 1-32 Revision 143 Amendment 12A

Rod Group Alignment Limits B 3.1.5 BASES SURVEILI-ANCE sR 3.1.5.3 REQUIREMENTS (continued) Verification of rod drop times allows the operator to determine that the maximum rod drop time permitted is consistent with the assumed rod drop time used in the safety analysis. Measuring rod drop times prior to reactor criticalig, after each reactor vessel head removal, ensures that the reactor internals and rod drive mechanism will not interfere with rod motion or rod drop time, and that no degradation in these systems has occurred that would adversely affect control rod motion or drop time. This testing is performed with all RCPs operating and the average moderator temperature > 551'F to simulate a reactor trip under actual conditions.

This Surveillance is performed during a plant outage, due to the plant conditions needed to perform the SR and the potential for an unplanned plant transient if the Surveillance were performed with the reactor at power.

Watts Bar-Unit 1 B 3.1 -33 Revision 143 Amendment 120

Shutdown Bank lnsertion Limits B 3.1 .6 BASES BACKGROUND Hence, they are not capable of adding a large amount of positive reactivity.

(continued) Boration or dilution of the Reactor Coolant System (RCS) compensates for the reactivity changes associated with large changes in RCS temperature. The design calculations are performed with the assumption that the shutdown banks are withdrawn first. The shutdown banks are controlled manually by the control room operator. During normal unit operation, the shutdown banks are either fully withdrawn or fully inserted. The shutdown banks must be completely withdrawn from the core, prior to withdrawing any control banks during an approach to criticality. The shutdown banks can be fully withdrawn without the core going critical. This provides available negative reactivity in the event of boration errors.

The shutdown banks are then left in this position untilthe reactor is shut down.

They add negative reactivity to shut down the reactor upon receipt of a reactor trip signal.

APPLICABLE On a reactor trip, all RCCAs (shutdown banks and control banks), except SAFEWANALYSES the most reactive RCCA, are assumed to insert into the core. The shutdown banks shall be at or above their insertion limits and available to insert the maximum amount of negative reactivi$ on a reactor trip signal. The control banks may be partially inserted in the core, as allowed by LCO 3.1.7, "Control Bank lnsertion Limits." The shutdown bank and control bank insertion limits are established to ensure that a sufficient amount of negative reactivity is available to shut down the reactor and maintain the required SDM (see LCO 3.1.1, "SHUTDOWN MARGIN (SDM) - T",s ) 200oF," and LCO 3.1.2, "SHUTDOWN MARGIN (SDM) - T",s 3 200'F) following a reactortrip from full power. The combination of control banks and shutdown banks (less the most reactive RCCA, which is assumed to be fully withdrawn) is sufficient to take the reactor from full power conditions at rated temperature to zero power, and to maintain the required SDM at rated no load temperature (Ref. 3). The shutdown bank insertion limit also limits the reactivity worth of an ejected shutdown rod.

The acceptance criteria for addressing shutdown and control rod bank insertion limits and inoperability or misalignment is that:

a. There be no violations of:

1. Specified acceptable fuel design limits, or

2. RCS pressure boundary integrity; and (continued)

Watts Bar-Unit 1 B 3.1-36 Revision 143 Amendment 120

Shutdown Bank lnsertion Limits B 3.1 .6 BASES APPLICABLE b. The core remains subcritical after accident transients other than a main SAFETY ANALYSES steam line break (MSLB).

(continued)

As such, the shutdown bank insertion limits affect safety analysis involving core reactivity and SDM (Ref. 3).

The shutdown bank insertion limits preserve an initial condition assumed in the safety analyses and, as such, satisff Criterion 2 ot '10 CFR 50.36(c)(2xii).

LCO The shutdown banks must be within their insertion limits any time the reactor is critical or approaching criticality. This ensures that a sufficient amount of negative reactivity is available to shut down the reactor and maintain the required SDM following a reactor trip.

The shutdown bank insertion limits are defined in the COLR.

The LCO is modified by a Note indicating the LCO requirement is not applicable to shutdown banks being inserted while performing SR 3.1.3.2. This SR verifies the freedom of the rods to move, and may require the shutdown bank to move below the LCO limits, which would normally violate the LCO. This NOTE applies to each shutdown bank as it is moved below the insertion limit to perform the SR.

This Note is not applicable should a malfunction stop performance of the SR.

APPLICABILITY The shutdown banks must be within their insertion limits, with the reactor in MODES 1 and2. This ensures that a sufficient amount of negative reactivity is available to shut down the reactor and maintain the required SDM following a reactor trip. The shutdown banks do not have to be within their insertion limits in MODE 3, unless an approach to criticality is being made. Refer to LCO 3.1.1 and LCO 3.1.2 for SDM requirements in MODES 3, 4, and 5. LCO 3.9.1, "Boron Concentration," ensures adequate SDM in MODE 6.

(continued)

Watts Bar-Unit 1 B 3.1-37 Revision 143 Amendment 120

Shutdown Bank lnsertion Limits B 3.1.6 BASES ACTIONS 4.1.4.2.1. A.2.2. and A.3 lf one shutdown bank is inserted less than or equal to 10 steps below the insertion limit, 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is allowed to restore the shutdown bank to within the limit.

This is necessary because the available sDM may be reduced with a shutdown bank not within its insertion limit. Also, verification of SDM or initiation of boration within'1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months is required, since the SDM in MODES 1 and2 is ensured by adhering to the control and shutdown bank insertion limits (see LCo 3.1.1). lf a shutdown bank is not within its insertion limit, sDM will be verified by performing a reactivity balance calculation, considering the effects listed in the BASES for SR 3.1.',l.1.

While the shutdown bank is outside the insertion limit, all control banks must be within their insertion limits to ensure sufficient shutdown margin is available. The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months completion Time is sufficient to repair most rod control failures that would prevent movement of a shutdown bank.

8.1.1.8.1.2 and B.2 When one or more shutdown banks is not within insertion limits for reasons other than condition A, 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months is allowed to restore the shutdown banks to within the insertion limits. This is necessary because the available SDM may be significantly reduced, with one or more of the shutdown banks not within their insertion limits. Also, verification of SDM or initiation of boration within t hour is required, since the sDM in MODES 1 and2 is ensured by adhering to the control and shutdown bank insertion limits (see Lco 3.1 .1 ). lf shutdown banks are not within their insertion limits, then sDM will be verified by performing a reactivity balance calculation, considering the effects listed in the Bases for sR 9.1.'1.1.

The allowed completion Time of 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months provides an acceptable time for evaluating and repairing minor problems without allowing the plant to remain in an unacceptable condition for an extended period of time.

c.1 lf the Required Actions and associated completion Times are not met, the unit must be brought to a MODE where the LCO is not applicable. The allowed completion Time of 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months is reasonable, based on operating experience, for reaching the required MODE from full power conditions in an orderly manner and without challenging plant systems.

Watts Bar-Unit 1 B 3.1-38 Revision 143 Amendment 120

Shutdown Bank lnsertion Limits B 3.1 .6 BASES SURVEILLANCE sR 3.1.6.1 REQUIREMENTS Verification that the shutdown banks are within their insertion limits prior to an approach to criticality ensures that when the reactor is critical, or being taken critical, the shutdown banks will be available to shut down the reactor, and the required sDM will be maintained following a reactor trip. This sR and Frequency ensure that the shutdown banks are withdrawn before the control banks are withdrawn during a unit startup.

The surveillance is modified by a Note which states that the sR is not required to be performed for shutdown banks until t hour after motion of rods in those banks.

Rod temperature affects the accuracy of the rod position indication system. Due to changes in the magnetic permeabilig of the drive shaft as a function of temperature, the indicated position is expected to change with time as the drive shaft temperature changes. The one hour period allows rod temperature to stabilize following rod movement in order to ensure the indicated position is accurate.

since the shutdown banks are positioned manually by the control room operator, a verification of shutdown bank position at a Frequency of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , after the reactor is taken critical, is adequate to ensure that they are within their insertion limits. Also, the 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Frequency takes into account other information available in the control room for the purpose of monitoring the status of shutdown rods.

REFERENCES 1. Title 10, Code of Federal Regulations, Part 50, Appendix A, General Design Criterion 10, "Reactor Design," General Design CriterionZO, "Reactivity Control System Redundancy and Capability," and General Design Criterion 28, "Reactivity Limits."

2. Title 10, Code of Federal Regulations, Part 50.46, 'Acceptance Criteria for Emergency Core Cooling Systems for Light Water Nuclear Power Reactors."

3. watts Bar FSAR, section 15.0, "Accident Analyses."

Watts Bar-Unit 1 B 3.1-39 Revision 143 Amendment 120

Control Bank lnsertion Limits B 3.1 .7 BASES BACKGROUND to move with bank C on a withdrawal, as an example may be at 116 steps.

(continued) Therefore, in this example, control bank C overlaps control bank D from 116 steps to the fully withdrawn position for control bank C. The fully withdrawn position and predetermined overlap positions are defined in the COLR.

The control banks are used for precise reactivity control of the reactor. The positions of the control banks are normally controlled automatically by the Rod Control System, but can also be manually controlled. They are capable of adding reactivity very quickly (compared to borating or diluting).

The power density at any point in the core must be limited, so that the fuel design criteria are maintained. Together, LCO 3.1.5, "Rod Group Alignment Limits,'

LCO 3.1.6, "Shutdown Bank lnsertion Limits," LCO 3.1.7, "ControlBank lnsertion Limits, LCO 3.2.3, "AXIAL FLUX DIFFERENCE (AFD)," and LCO 3.2.4, "QUADRANT POWER TILT MTIO (QPTR)," provide limits on control component operation and on monitored process variables, which ensure that the core operates within the fuel design criteria.

The shutdown and controlbank insertion and alignment limits, AFD, and QPTR are process variables that together characterize and controlthe three dimensional power distribution of the reactor core. Additionally, the control bank insertion limits control the reactivity that could be added in the event of a rod ejection accident, and the shutdown and control bank insertion limits ensure the required SDM is maintained.

Operation within the subject LCO limits will prevent fuel ctadding failures that would breach the primary fission product barrier and release fission products to the reactor coolant in the event of a loss of coolant accident (LOCA), loss of flow, ejected rod, or other accident requiring termination by a Reactor Trip System (RTS) trip function.

(continued)

Watts Bar-Unit 1 B 3. 141 Revision 143 Amendment 120

Control Bank lnsertion Limits B 3.1 .7 BASES APPLICABLE The shutdown and controlbank insertion limits, AFD, and QPTR LCOs are SAFEW ANALYSES required to prevent power distributions that could result in fuel cladding failures in the event of a LOCA, loss of flow, ejected rod, or other accident requiring termination by an RTS trip function.

The acceptance criteria for addressing shutdown and control bank insertion limits and inoperability or misalignment are that:

a. There be no violations of:

1. Specified acceptable fuel design limits, or

2. Reactor Coolant System pressure boundary integrity; and

b. The core remains subcritical after accident transients other than a main steam line break (MSLB).

As such, the shutdown and control bank insertion limits affect safety analysis involving core reactivity and power distributions (Ref. 3 through 13).

The SDM requirement is ensured by limiting the control and shutdown bank insertion limits so that allowable inserted worth of the RCCAs is such that sufficient reactivig is available in the rods to shut down the reactor to hot zero power with a reactivity margin that assumes the maximum worth RccA remains fullywithdrawn upon trip (Ref. 5,6,8 and 11).

Operation at the insertion limits orAFD limits may approach the maximum allowable linear heat generation rate or peaking factor with the allowed eprR present. operation at the insertion limit may also indicate the maximum ejected RCCA worth could be equal to the limiting value in fuel cycles that have sufficiently high ejected RCCA worths.

The control and shutdown bank insertion limits ensure that safety analyses assumptions for SDM, ejected rod worth, and power distribution peaking factors are preserved (Ref. 3 through 13).

The insertion limits satisfy Criterion 2 ot 10 CFR 50.36(cx2xii), in that they are I initial conditions assumed in the safety analysis. I (continued)

Watts Bar-Unit 1 B 3. 142 Revision 143 Amendment 120

Control Bank Insertion Limits B 3.1 .7 BASES LCO The limits on control banks sequence, overlap, and physical insertion, as defined in the COLR, must be maintained because they serve the function of preserving power distribution, ensuring that the SDM is maintained, ensuring that ejected rod worth is maintained, and ensuring adequate negative reactivity insertion is available on trip. The overlap between control banks provides more uniform rates of reactivity insertion and withdrawal and is imposed to maintain acceptable power peaking during control bank motion.

The LCO is modified by a Note indicating the LCO requirement is not applicable to control banks being inserted while performing SR 3.1.5.2. This SR verifies the freedom of the rods to move, and may require the control bank to move below the LCO limits, which would normally violate the LCO. This Note applies to each control bank as it is moved below the insertion limit to perform the SR. This Note is not applicable should a malfunction stop performance of the SR.

APPLICABILITY The controlbank sequence, overlap, and physical insertion limits shallbe maintained with the reactor in MODES 1 and 2 with k"r 2 1.0. These limits must be maintained, since they preserve the assumed power distribution, ejected rod worth, SDM, and reactivity rate insertion assumptions. Applicability in MODES 3, 4, and 5 is not required, since neither the power distribution nor ejected rod worth assumptions would be exceeded in these MODES.

ACTIONS 4.1. 4.2.1. A.2.2. and A.3 lf Control Bank A, B, or C is inserted less than or equal to 10 steps below the insertion, sequence, or overlap limits, 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is allowed to restore the control bank to within the limits. Verification of SDM or initiation of boration within t hour is required, since the SDM in MODES 'l and2 is ensured by adhering to the control and shutdown bank insertion limits (see LCO 3.1.1). lf a control bank is not within its insertion limit, SDM will be verified by performing a reactivig balance calculation, considering the effects listed in the BASES for SR 3.1.1.1.

While the control bank is outside the insertion, sequence, or overlap limits, all shutdown banks must be within their insertion limits to ensure sufficient shutdown margin is available and that power distribution is controlled. The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Completion Time is sufficient to repair most rod control failures that would prevent movement of a shutdown bank.

Condition A is limited to Control Banks A, B, or C. The allowance is not required for Control Bank D because the full power bank insertion limit can be met during performance of the SR 3.'1.5.2 control rod freedom of movement (trippability) testing.

Watts Bar-Unit 1 B 3.143 Revision 143 Amendment 120

Control Bank lnsertion Limits B 3.1 .7 BASES ACTIONS 8.1.1. 8.1.2. 8.2. C.'1.1. C.1.2. and C.2 (continued)

When the control banks are outside the acceptable insertion limits for reasons other than Condition A, they must be restored to within those limits. This restoration can occur in two ways:

a. Reducing power to be consistent with rod position; or

b. Moving rods to be consistent with power.

Also, verification of sDM or initiation of boration to regain sDM is required within t hour, since the SDM in MODES 1 and2 normally ensured by adhering to the controland shutdown bank insertion limits (see LCO 3.1.1, "SHUTDOWN MARGIN (SDM) - Ta,g ) 200'F") has been upset. lf control banks are not within their insertion limits, then sDM will be verified by performing a reactivity balance calculation, considering the effects listed in the Bases for SR 3.1.1.1.

similarly, if the control banks are found to be out of sequence or in the wrong overlap configuration for reasons other than Condition A, they must be restored to meet the limits.

Operation beyond the LCO limits is allowed for a short time period in order to take conservative action because the simultaneous occurrence of either a LoCA, loss of flow accident, ejected rod accident, or other accident during this short time period, together with an inadequate power distribution or reactivity capability, has an acceptably low probability.

The allowed completion Time of 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for restoring the banks to within the insertion, sequence, and overlap limits provides an acceptable time for evaluating and repairing minor problems without allowing the plant to remain in an unacceptable condition for an extended period of time.

D.1 lf the Required Actions cannot be completed within the associated completion Times, the plant must be brought to MODE 2 with k"6 < 1.0, where the LCO is not applicable. The allowed Completion Time of 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months is reasonable, based on operating experience, for reaching the required MODE from full power conditions in an orderly manner and without challenging plant systems.

(continued)

Watts Bar-Unit 1 B 3.144 Revision 143 Amendment 120

Control Bank lnsertion Limits B 3.1 .7 BASES (continued)

SURVEIL!.ANCE SR 3.1.7.1 REQUlREMENTS This Surveillance is required to ensure that the reactor does not achieve criticality with the control banks below their insertion limits.

The estimated critical position (ECP) depends upon a number of factors, one of which is xenon concentration. lf the ECP was calculated long before criticality, xenon concentration could change to make the ECP substantially in error.

Conversely, determining the ECP immediately before criticality could be an unnecessary burden. There are a number of unit parameters requiring operator attention at that point. Performing the ECP calculation within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months prior to criticality avoids a large error from changes in xenon concentration, but allows the operator some flexibility to schedule the ECP calculation with other startup activities.

sR 3.1.7.2 With an OPERABLE bank insertion limit monitor, verification of the controlbank insertion limits at a Frequency of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is sufficient to ensure OPERABILITY of the bank insertion limit monitor and to detect control banks that may be approaching the insertion limits since, normally, very little rod motion occurs in 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . lf the insertion limit monitor becomes inoperable, verification of the control bank position at a Frequency of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months is sufficient to detect control banks that may be approaching the insertion limits.

The Surveillance is modified by a Note stating that the SR is not required to be performed for control banks until t hour after motion of rods in those banks.

Control rod temperature affects the accuracy of the rod position indication system. Due to changes in the magnetic permeability of the drive shaft as a function of temperature, the indicated position is expected to change with time as the drive shaft temperature changes. The one hour period allows control rod temperature to stabilize following rod movement in order to ensure the indicated rod position is accurate.

sR 3.1.7.3 When control banks are maintained within their insertion limits as checked by SR 3.1.7.2 above, it is unlikely that their sequence and overlap will not be in accordance with requirements provided in the COLR.

The Surveillance is modified by a Note stating that the SR is not required to be performed for control banks until t hour after motion of rods in those banks.

Control rod temperature affects the accuracy of the rod position indication system. Due to changes in the magnetic permeability of the drive shaft as a function of temperature, the indicated position is expected to change with time as the drive shaft temperature changes. The one hour period allows control rod temperature to stabilize following rod movement in order to ensure the indicated rod position is accurate.

(continued)

Watts Bar-Unit 1 B 3. 145 Revision 143 Amendment 120

Control Bank Insertion Limits B 3.1 .7 BASES SURVEILLANCE SR 3.1.7.3 (continued)

REQUIREMENTS A Frequency of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is consistent with the insertion limit check above in sR 3.1.7.2.

REFERENCES 1. Title 10, Code of Federal Regulations, Part 50, Appendix A, General Design Criterion 10, "Reactor Design," General Design Criterion26, "Reactivity Control System Redundancy and Capability," and General Design Criterion 28, "Reactivity Limits."

2. Title 10, Code of Federal Regulations, Part 50.46, 'Acceptance Criteria for Emergency Core Cooling Systems for Light Water Nuctear Power Reactors."

3. Watts Bar FSAR, Section 15.2.1, "Uncontrolled Rod Cluster Control Assembly Bank Withdrawal From a Subcritical Condition."

4. Watts Bar FSAR, Section 15.2.2, "Uncontrolled Rod Cluster Control Assembly Bank Withdrawal At Power."

5. Watts Bar FSAR, Section 15.2.3, "Rod Cluster Control Assembly Misalignment."

6. Watts Bar FSAR, Section 15.2.4, "Uncontrolled Boron Dilution."

7. Watts Bar FSAR, Section 15.2.5, "Partial Loss of Forced Reactor Cootant Flow."

8. Watts Bar FSAR, Section 15.2.13,'Accidental Depressurization of the Main Steam System."

9. Watts Bar FSAR, Section 15.3.4, "Complete Loss of Forced Reactor Coolant Flow."

10. Watts Bar FSAR, Section 15.3.6, "single Rod Cluster Control Assembly Withdrawal At Full Power."

11. Watts Bar FSAR, Section 15.4 .2.1, "Major Rupture of Main Steam Line."

12. Watts Bar FSAR, Section 15.4.4, "single Reactor Coolant Pump Locked Rotor."

13. Watts Bar FSAR, Section 15.4.6, "Rupture of a Control Rod Drive Mechanism Housing (Rod Cluster Control Assembly Ejection)."

Watts Bar-Unit 1 B 3.146 Revision 143 Amendment 120

Control Bank lnsertion Limits B 3.1 .7 BASES (ALL POWERS, 225)

(0.197 ,225) 0.703, 225 Lts 0.0, 180

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c, (0.0, 64l, 0.227 , O o,4 0,6 FRACTION OF RATED THERMAL POWER FIGURE B 3.1 .7-1 CONTROL BANK INSERTION vs. RTP Watts Bar-Unit 1 B 3. 147 Revision 143 Amendment 120

Rod Position lndication B 3.1.8 B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.8 Rod Position lndication BASES BACKGROUND According to GDC 13 (Ref. 1), instrumentation to monitor variables and systems over their operating ranges during normal operation, anticipated operational occurrences, and accident conditions must be OPEMBLE. LCO 3.'1.8 is required to ensure OPERABILITY of the control rod position indicators to determine control rod positions and thereby ensure compliance with the control rod alignment and insertion limits.

The OPEMBILITY, including position indication, of the shutdown and control rods is an initial assumption in all safety analyses that assume rod insertion upon reactor trip. Maximum rod misalignment is an initial assumption in the safety analysis that directly affects core power distributions and assumptions of available SDM. Rod position indication is required to assess OPEMBILITY and misalignment.

Mechanical or electricalfailures may cause a control rod to become inoperable or to become misaligned from its group. Control rod inoperability or misalignment may cause increased power peaking, due to the asymmetric reactivity distribution and a reduction in the total available rod worth for reactor shutdown. Therefore, control rod alignment and OPERABILIry are related to core operation in design power peaking limits and the core design requirement of a minimum SDM.

Limits on control rod alignment and OPERABILITY have been established, and all rod positions are monitored and controlled during power operation to ensure that the power distribution and reactivity limits defined by the design power peaking and SDM limits are preserved.

Rod cluster control assemblies (RCCAs), or rods, are moved out of the core (up or withdrawn) or into the core (down or inserted) by their control rod drive mechanisms. The RCCAs are divided among control banks and shutdown banks. Each bank may be further subdivided into two groups to provide for precise reactivity control (Shutdown Banks C and D have only one group each).

The axial position of shutdown rods and control rods are determined by two separate and independent systems: the Bank Demand Position lndication System (commonly called group step counters) and the analog Rod Position lndication (RPl) System.

(continued)

Watts Bar-Unit 1 B3148 Revision 51 , 143 Amendment 120

Rod Position lndication B 3.1.8 BASES BACKGROUND The Bank Demand Position lndication System counts the pulses from the Rod (continued) control system that move the rods. There is one step counter for each group of rods. lndividual rods in a group all receive the same signalto move and should, therefore, all be at the same position indicated by the group step counter for that group. The Bank Demand Position lndication System is considered highly precise (t 1 step or t 5/8 inch). lf a rod does not move one step for each demand pulse, the step counter will still count the pulse and incorrectly reflect the position of the rod.

The RPI system provides an accurate indication of actual control rod position, but at a lower precision than the step counters. This system is based on inductive analog signals from a series of alternating primary and secondary coils spaced along a hollow tube. The normal indication accuracy of the Rpl system is t 6 steps (t 3.75 inches), and the maximum uncertainty is r 12 steps (t 7.5 inches). With an indicated deviation of 12 steps between the group step counter and RPl, the maximum deviation between actual rod position and the demand position could be 24 steps, or 15 inches.

APPLICABLE control and shutdown rod position accuracy is essential during power operation.

SAFETY ANALYSES Power peaking, ejected rod worth, or SDM limits may be violated in the event of a Design Basis Accident (Ref. 2 through 12), with control or shutdown rods operating outside their limits undetected. Therefore, the acceptance criteria for rod position indication is that rod positions must be known with sufficient accuracy in order to verify the core is operating within the group sequence, overlap, design peaking limits, ejected rod worth, and with minimum SDM (LCO 3.1.6, "Shutdown Bank lnsertion Limits," and LCO 3.1.7, "ControlBank lnsertion Limits"). The rod positions must also be known in order to verify the alignment limits are preserved (LCO 3.1.5, "Rod Group Alignment Limits,').

Control rod positions are continuously monitored to provide operators with information that ensures the plant is operating within the bounds of the accident analysis assumptions.

The control rod position indicator channels satisfy Criterion 2 of 10 CFR 50.36(c)(2)(ii). The control rod position indicators monitor control rod position, which is an initial condition of the accident.

LCO LCO 3.1.8 specifies that the RPI System and the Bank Demand position lndication System be OPERABLE for all control rods. For the control rod position indicators to be OPEMBLE requires meeting the sR of the LCo (when required) and the following:

(continued)

Watts Bar-Unit 1 B 3. 149 Revision 143 Amendment 120

Rod Position lndication B 3.1 .8 BASES LCO a. The RPI System indicates within 12 steps of the group step counter (continued) demand position when LCO 3.1.5, "Rod Group Alignment Limits;" met.

b. For the RPI System there are no failed coils; and

c. The Bank Demand lndication System has been calibrated either in the fully inserted position or to the RPI System.

The't2 step agreement limit between the Bank Demand Position lndication System and the RPI System indicates that the Bank Demand Position lndication System is adequately calibrated, and can be used for indication of the measurement of control rod bank position.

A deviation of less than the allowable limit, given in LCO 3.1.5, in position indication for a single control rod, ensures high confidence that the position uncertainty of the corresponding control rod group is within the assumed values used in the analysis (that specified control rod group insertion limits).

These requirements ensure that control rod position indication during power operation and PHYSICS TESTS is accurate, and that design assumptions are not challenged. OPERABILIw of the position indicator channels ensures that inoperable, misaligned, or mispositioned control rods can be detected.

Therefore, power peaking, ejected rod worth, and SDM can be controlled within acceptable limits.

The LCO is modified by a Note stating that the RPI system is not required to be OPEMBLE for '1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months following movement of the associated rods. Control and shutdown rod temperature affects the accuracy of the RPI System. Due to changes in the magnetic permeability of the drive shaft as a function of temperature, the indicated position is expected to change with time as the drive shaft temperature changes. The one hour period allows temperature to stabilize following rod movement in order to ensure the indicated position is accurate.

APPLICABILITY The requirements on the RPI and step counters are only applicable in MODES 1 and 2 (consistent with LCO 3.1.5, LCO 3.1.6, and LCO 3.1.7), because these are the only MODES in which power is generated, and the OPERABTLITY and alignment of rods have the potential to affect the safety of the plant. ln the shutdown MODES, the OPEMBILITY of the shutdown and controlbanks has the potentialto affect the required SDM, but this effect can be compensated for by an increase in the boron concentration of the Reactor Coolant System.

(continued)

Watts Bar-Unit 1 B 3.1-50 Revision 70, 104, 143 Amendment 58, 82, 120

Rod Position Indication B 3.1.8 BASES ACTIONS The ACTIONS table is modified by a Note indicating that a separate Condition entry is allowed for each inoperable rod position indicator and each demand position indicator. This is acceptable because the Required Actions for each Condition provide appropriate compensatory actions for each inoperable position indicator.

A.1 and A.2 When one RPI channel per group in one or more groups fails, the position of the rod can still be determined indirectly by use of incore power distribution measurement information. lncore power distribution measurement information can be obtained from flux traces using the Movable lncore Detector System or from an OPEMBLE Power Distribution Monitoring System (PDMS) (Ref. 15).

The Required Action may also be satisfied by ensuring at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months that Fq satisfies LCO 3.2.1, F*o11 satisfies LCO 3.2.2, and SHUTDOWN MARGIN is within the limits provided in the COLR, provided the non-indicating rods have not been moved. Based on experience, normal power operation does not require excessive movement of banks. lf a bank has been significantly moved, the Required Action of C.1 or C.2 below is required. Therefore, verification of RCCA position within the Completion Time of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months is adequate for allowing continued full power operation, since the probability of simultaneously having a rod significantly out of position and an event sensitive to that rod position is small.

Required Action A.1 requires verification of the position of a rod with an inoperable RPI once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months which may put excessive wear and tear on the moveable incore detector system. Required Action A.2 provides an alternative.

Required Action A.2 requires verification of rod position using the incore power distribution measurement information every 31 EFPD, which coincides with the normal measurements to verify core power distribution.

Required Action A.2 includes six distinct requirements for verification of the position of rods associated with an inoperable RPI using incore power distribution measurements information:

a. lnitial verification within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months of the inoperability of the RPI;

b. Re-verification once every 31 Effective Full Power Days (EFPD) thereafter;

c. Verification within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months after discovery of each unintended rod movement.

An unintended rod movement is defined as the release of the rod's stationary gripper when no action was demanded either manually or automatically from the rod control system, or a rod motion in a direction other than the direction demanded by the rod control system. Verifying that no unintended rod movement has occurred is performed by monitoring the rod control system stationary gripper coil current for indications of rod movement; (continued)

Watts Bar-Unit 1 B 3.1-51 Revision 70, 104, 143 Amendment 58, 82, 120

Rod Position lndication B 3.1.8 BASES ACTIONS d. Verification within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months if the rod with an inoperable RPI is intentionally (continued) moved greaterthan 12 steps;

e. Verification prior to exceeding 50% RTP if power is reduced below 50% RTP; and

f. Verification within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months of reaching 100o/o RTP if power is reduced to less than 100% power RTP.

Should the rod with the inoperable RPI be moved more than 12 steps, or if reactor power is changed, the position of the rod with the inoperable RPI must be verified.

4.3 Reduction of THERMAL POWER to s 50% RTP puts the core into a condition where rod position is not significantly affecting core peaking factors (Ref. 4). The allowed Completion Time of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months is reasonable, based on operating experience, for reducing power to s 50% RTP from full power conditions without challenging plant systems and allowing for rod position determination by Required Actions A.1 and A.2 above.

8.1 and B.2 When more than one RPI per group in one or more groups fail, additional actions are necessary. Placing the Rod ControlSystem in manualassures unplanned rod motion will not occur. The immediate Completion Time for placing the Rod Control System in manual reflects the urgency with which unplanned rod motion must be prevented while in this Condition.

The inoperable RPls must be restored, such that a maximum of one RPI per group is inoperable, within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . The 24hour Completion Time provides sufficient time to troubleshoot and restore the RPI system to operation while avoiding the plant challenges associated with the shutdown without full rod position indication.

Based on operating experience, normal power operation does not require excessive rod movement. lf one or more rods has been significantly moved, the Required Action of C.1 or C.2 below is required.

C.1 and C.2 With one or more RPI inoperable in one or more groups and the affected groups have moved greater than24 steps in one direction since the last determination of rod position, additional actions are needed to verify the position of rods with inoperable RPl. Within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , the position of the rods with inoperable position indication must be determined using either the moveable incore detectors or PDMS to verify these rods are still properly positioned, relative to their group positions.

(continued)

Watts Bar-Unit 1 B 3. 1-52 Revision 70, 104, 143 Amendment 58, 82, 120

Rod Position Indication B 3.1.8 BASES ACTIONS C.1 and C.2 (continued) lf, within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , the rod positions have not been determined, THERMAL POWER must be reduced to s 50% RTP within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months to avoid undesirable power distributions that could resuJt from continued operation at > 50% RTP, if one or more rods are misaligned by more lhan24 steps. The allowed Completion Time of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months provides an acceptable period of time to veriff the rod positions.

D.1.1 and D.1.2 With one or more demand position indicators per bank inoperable in one or more banks, the rod positions can be determined by the RPI System. Since normal power operation does not require excessive movement of rods, verification by administrative means that the rod position indicators are OPEMBLE and the most withdrawn rod and the least wtthdrawn rod are s 12 steps apart within the allowed Completion Time of once every 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months is adequate.

D.2 Reduction of THERMAL POWER to < 50% RTP puts the core into a condition where rod position is not significantly affecting core peaking factor limits (Ref. 13). The allowed Completion Time of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months provides an acceptable period of time to verify the rod positions per Required Actions D.1.1 and D.1.2 or reduce powerto s 50% RTP.

E.'l lf the Required Actions cannot be completed within the associated Completion Time, the plant must be brought to a MODE in which the requirement does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . The allowed Completion Time is reasonable, based on operating experience, for reaching the required MODE from full power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE sR 3.1.8.1 REQUIREMENTS Verification that the RPI agrees with the demand position within 12 steps ensures that the RPI is operating correctly.

This Surveillance is performed prior to reactor criticality after each removal of the reactor head, as there is the potential for unnecessary plant transients if the SR were performed with the reactor at power.

The Surveillance is modified by a Note which states it is not required to be met for RPls associated with rods that do not meet LCO 3.1.5. lf a rod is known to not be within 12 steps of the group demand position, the ACTIONS of LCO 3.1.5 provide the appropriate Actions.

(continued)

Watts Bar-Unit 1 B 3.1-53 Revision 70, 104, 143 Amendment 58, 82, 120

rlN tan 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 FN66 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 peak cladding temperature (PCT) must not exceed 2200F for small breaks, and there must be a high level of probability that the peak cladding temperature does not exceed 2200'F for large brealls (Ref. 3);

c. During an ejected rod accident, the energy deposition to the fuel must not exceed 280 caUgm (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, FN6x is a significant core parameter. The limits on FN6x BnSUr 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 localDNB 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 lor a discussion of the applicable DNBR limits. The w-3 corretation with a DNBR timit of 1 .3, or the ABB-NV correlation with a DNBR limit of 1 .1 3, is applied in the heated region below the first mixing vane grid. ln 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 orthe WRB-2M correlation for RFA-2 fuelwith lFMs. 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 3.2-1 3 Revision 13, 39, 59 , 140 Amendment 7,21 , 46

QPTR B 3.2.4 BASES ACTIONS 8.1 (continued) lf Required Actions A.1 through A.6 are not completed within their associated completion Times, the unit must be brought to a MoDE or condilion in which the requirements do not apply. To achieve this status, THERMAL povVER must be reduced to < 50% RTP within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . The allorred comptetion Time of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months is reasonable, based on operating experience regarding the amount of time required to reach the reduced power levelwithout challenging plant systems.

SURVEILLANCE sR 3.2.4.1 REQUIREMENTS sR 3.2.4.1 is modified by two Notes. Note 1 allows eprR to be catcutated with three power range channels if THERMAL pot ,ER is < 75% RTp and the input from one power range neutron flux channel is inoperable. Note 2 allows performance of sR 3.2.4.2 in lieu of sR 3.2.4.1 if more than one input from power range neutron flux channels are inoperable.

This surveillance verifies that the QPTR, as indicated by the Nuclear lnstrumentation system (Nls) excore channets, is within its limits. The Frequency of 7 days when the QPTR alarm is OPERABLE is acceptable because of the low probability that this alarm can remain inoperable without detection.

when the QPTR alarm is inoperable, the Frequency is increased to .12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

This Frequency is adequate to detect any relatively slow changes in eprR, because for those changes of QPTR that occur quickly (e.g., a dropped rod),

there $pically are other indications of abnormality that prompt a verification of core porertilt.

sR 3.2.4.2 This surveillance is modified by a Note, which states the surveillance is only required to be performed if input to QPTR from one or more power Range Neutron Flux channels are inoperable when THERMAL povvER is > 75% RTp.

with an NIS power range channel inoperable, tilt monitoring for a portion of the reactor core becomes degraded. Large tiltrs are Iikely detected with the remaining channels, but the capability for detection of smail power tilts in some quadrants is decreased. Performing sR 3.2.4.2 at a Frequency of 12 hourc provides an accurate altemative means for ensuring that any tih remains within its limits.

(continued)

Watts Bar-Unit 1 B 3.2-29 Revision 145 Amendment 122

QPTR B 3.2.4 BASES SURVEILLANCE SR 3.2.4.2 (continued)

REQUIREMENTS (continued) For the purpose of monitoring the QPTR when the input from one or more power range neutron flux channels are inoperable, incore power distribution measurement information is used to confirm that the indicated QPTR is consistent with the reference normalized symmetric power distribution. The incore power distribution information can be used to generate an incore "tilt."

This tilt can be compared to the reference incore tilt to generate and incore QPTR. Therefore, incore QPTR can be used to confirm that excore QPTR is within limits.

The incore power distribution measurement information can be obtained from either the movable incore detectors or from an OPEMBLE Power Distribution Monitoring System (PDMS) (Ref. ). !f the movable incore detectors are used, then the incore detector monitoring is performed with a full core flux map or two sets of four thimble locations with quarter core symmetry. The two sets of four symmetric thimbles is a set of eight unique detector locations. These locations are C-8, E-5, E-1 1, H-3, H-13, L-1 1, and N-8.

The reference normalized symmetric power distribution is available from the last incore power distribution measurement information used to calibrate the excore axial offset. The reference incore power distribution measurement information may have been obtained from either a full core flux map using the Movable lncore Detector System orfrom an OPERABLE PDMS. The fullcore flux map information may be reduced to the information from only the two sets of four symmetric thimbles with quarter core symmetry for like comparisons, if practical.

With the input from one or more power range neutron flux channels inoperable, the indicated QPTR may be changed from the value indicated with all four channels OPERABLE. To confirm that no change in tilt has actually occuned, which might causes the QPTR limit to be exceeded, the normalized quadrant tilt is compared against the reference normalized quadrant tilt. Nominally, quadrant tilt from the surveillance should be within 2o/o of lhe tilt shown by the reference incore power distribution measurement information.

(continued)

Watts Bar-Unit 1 B 3.2-29 Revision 104, 145 Amendment 82, 122

RTS lnstrumentation B 3.3.1 Bases APPLICABLE 4. Intermediate Ranoe Neutron Flux SAFEWANALYSES, LCO, and The lntermediate Range Neutron Flux trip Function ensures that APPLICABILITY protection is provided against an uncontrolled RCCA bank (continued) rod withdrawal accident from a subcritical condition during startup. This trip Function provides backup protection to the Power Range Neutron Flux-Low Setpoint trip Function. The NIS intermediate range detectors are located external to the reactor vessel and measure neutrons leaking from the core. This Function also provides a signal to prevent automatic and manua! rod withdrawal prior to initiating a reactor trip.

The LCO requires two channels of lntermediate Range Neutron Flux to be OPERABLE. Two OPERABLE channels are sufficient to ensure no single random failure will disable this trip Function.

Because this trip Function is important only during startup, there is generally no need to disable channels for testing while the Function is required to be OPEMBLE. Therefore, a third channel is unnecessary.

ln MODE 1 below the P-10 setpoint, and in MODE 2, when there is a potential for an uncontrolled RCCA bank rod withdrawa! accident during reactor startup, the lntermediate Range Neutron FIux trip must be OPERABLE. Above the P-1 0 setpoint, the Power Range Neutron Flux-High Setpoint trip provides core protection for a rod withdrawal accident.

ln MODE 3, 4, or 5, the lntermediate Range Neutron Flux trip does not have to be OPERABLE because the control rods must be fully inserted and only the shutdown rods may be withdrawn. The reactor cannot be started up in this condition. The core also has the required SDM to mitigate the consequences of a positive reactivity addition accident. ln MODE 6, all rods are fully inserted and the core has a required increased SDM.

(continued)

Watts Bar-Unit 1 B 3.3-14 Revision 149

RTS Instrumentation B 3.3.1 Bases APPLICABLE 5. Source Range Neutron Flux SAFEWANALYSES, LCO, and The LCO requirement for the Source Range Neutron Flux trip APPUCABILITY Function ensures that protection is provided against an uncontrolled (continued) RCCA rod bank withdrawal accident from a subcritical condition during startup. This trip Function provides redundant protection to the Power Range Neutron Flux-Low Setpoint and Intermediate Range Neutron Flux trip Functiolls. ln MODES 3, 4, and 5, administrative controls also prevent the uncontrolled withdrawal of rods. The NIS source range detectors are located external to the reactor vessel and measure neutrons leaking from the core. The NIS source range detectors do not provide any inputs to control systems. The source range trip is the only RTS automatic protection function required in MODES 3, 4, and 5.

Therefore, the functional capability at the specified Trip Setpoint is assumed to be available.

The LCO requires two channels of Source Range Neutron Flux to be OPERABLE. Two OPERABLE channels are sufficient to ensure no single random failure will disable this trip Function. The LCO also requires one channel of the Source Range Neutron FIux to be OPERABLE in MODE 3, 4, or 5 with RTBs open.

The Source Range Neutron Flux Function provides protection for control rod withdrawal from subcritical, boron dilution and control rod ejection events. The Function also provides visual neutron flux indication in the control room.

In MODE 2 when below the P setpoint during a reactor startup, the Source Range Neutron FIux trip must be OPERABLE. Above the P-6 setpoint, the lntermediate Range Neutron Flux trip and the Power Range Neutron Flux-Low Setpoint trip will provide core protection for reactivity accidents. Above the P-6 setpoint, the NIS Source Range Neutron Flux trip Function may be manually blocked. Above the P-10 setpoint, the NIS Source Range Neutron Flux trip function is automatically blocked.

(continued)

Watts Bar-Unit 1 B 3.3-15 Revision 149

RTS lnstrumentation B 3.3.1 Bases APPLICABLE a. Turbine Trip-Low Fluid Oil Pressure (continued)

SAFETY ANALYSES LCO, and power, will not actuate a reactor trip. Three APPLICABILITY pressure switches monitor the Emergency Trip Header pressure in the Turbine Electrohydraulic Control System high pressure header. A low pressure condition sensed by two-out-of-three pressure switches will actuate a reactor trip. These pressure switches do not provide any input to the control system. The unit is designed to withstand a complete loss of load and not sustain core damage or challenge the RCS pressure limitations. Core protection is provided by the Pressurizer Pressure-High trip Function and RCS integrity is ensured by the pressunzer safety valves.

The LCO requires three channels of Turbine Trip Low Fluid Oil Pressure to be OPERABLE in MODE 1 above-P-9.

Below the P-9 setpoint, a turbine trip does not actuate a reactor trip. ln MODE 2, 3, 4,5, or 6, there is no potential for a turbine trip, and the Turbine Trip-Low Fluid Oil Pressure trip Function does not need to be OPERABLE.

b. Turbine Trip-Turbine Stop Valve Closure The Turbine Trip-Turbine Stop Valve Closure trip Function anticipates the loss of heat removal capabilities of the second ary system following a turbine trip from a power level below the P-9 setpoint, approximately 50o/o power. This action will not actuate a reactor trip. The trip Function anticipates the loss of secondary heat remova! capability that occurs when the stop valves close.

Tripping the reactor in anticipation of loss of secondary heat removal acts to minimize the pressure and temperature transient on the reactor. This trip Function will not and is not required to operate in the presence of a single channel failure. The unit is designed to withstand a complete loss of load and not sustain core damage or challenge the RCS pressure limitations. Core protection is provided by the Pressu nzer Pressure-High trip Function, and RCS integrity is ensured by the (continued)

Watts Bar-Unit 1 B 3.3-30 Revision 13, 146 Amendment 7, 1 19

RTS lnstrumentation B 3.3.1 Bases ACTIONS D.1 and D.2 (continued)

Condition D applies to the Power Range Neutron Flux-High Function.

The NIS power range detectors provide input to the CRD System and the SG Water Level Control System and, therefore, have a two-out-of-four trip logic. A known inoperable channel must be placed in the tripped condition. This results in a partial trip condition requiring only one-out-of-three logic for actuation. The 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months allowed by Required Action D.1 to place the inoperable channel in the tripped condition is justified in Reference 14.

The Required Actions have been modified by two Notes. Note 1 allows the inoperable channel to be placed in the bypassed condition for up to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months while performing routine surveillance testing of other channels. With one channel inoperable, the Note also allows routine surveillance testing of another channel with the inoperable channel in bypass. The Note also allows placing the inoperable channel in the bypass condition to allow setpoint adjustments of other channels when required to reduce the Power Range Neutron Flux-High setpoint in accordance with other Technical Specifications. The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months time limit is justified in Reference 14.

Note 2 states to perform SR 3.2.4.2 if input to QPTR from one or more Power Range Neutron Flux channels are inoperable with THERMAL POWER > 75o/o RTP.

lf Required Action D.1 cannot be met within the specified Completion Time, the plant must be placed in a MODE where this Function is no longer required OPERABLE. Seventy-eight hours are allowed to place the plant in MODE 3.

The 78-hour Completion Time includes 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months for the MODE reduction as required by Required Action D.2. This is a reasonable time, based on operating experience, to reach MODE 3 from full power in an orderly manner and without challenging plant systems.

(continued)

Watts Bar-Unit 1 B 3.341 Revision 90, 145 Amendment 68, 122

RTS lnstrumentation B 3.3.1 Bases ACTIONS E.1 and E.2 (continued)

Condition E applies to the following reactor trip Functions:

o Power Range Neutron Flux-Low; and o Power Range Neutron Flux-High Positive Rate A known inoperable channel must be placed in the tripped condition within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . Placing the channel in the tripped condition results in a partial trip condition requiring only one-out-of-two logic for actuation of the two-out-of-three trips and one-out-of-three logic for actuation of the two-out-of-four trips. The 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months allowed to place the inoperable channel in the tripped condition is justified in Reference 14.

lf the inoperable channel cannot be placed in the trip condition within the specifled Completion Time, the plant must be placed in a MODE where these Functions are not required OPERABLE. An additional6 hours is allowed to place the plant in MODE 3. Six hours is a reasonable time, based on operating experience, to place the plant in MODE 3 from full power in an orderly manner and without challenging plant systems.

The Required Actions have been modified by a Note that allows placing the inoperable channel in the bypassed condition for up to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months while performing routine surveillance testing of the other channels. The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months time limit is justified in Reference 14.

(continued)

Watts Bar-Unit 1 B 3.3-42 Revision 27,90, 104, 145 Amendment 18, 68 , 82, 122

COMS B 3.4.12 BASES SURVEILISNCE SR 3.4.12.7 (continued)

REOUIREMENTS The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months allowance to meet the requirement considers the unlikelihood of a low temperature overpressure event during this time.

A Note has been added indicating that this SR is required to be met within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after decreasing RCS cold leg temperature to < 350EF.

sR 3.4.12.8 Performance of a CHANNEL CALIBRATION on each required PORV actuation channel is required every 18 months to adjust the whole channel so that it responds and the valve opens within the required range and accuracy to known input.

REFERENCES 1. Title 10, Code of Federal Regulations, Part 50, Appendix G, "Fracture Toughness Requirements."

2. Generic Letter 88-11, "NRC Position on Radiation Embriftlement of Reactor Vessel Materials and lts lmpact on Plant Operation."

3. ASME Boiler and Pressure Vessel Code, Section lll.

4. Watts Bar FSAR, Section 5.2.2.4, 'RCS Pressure ControlDuring Low I Temperature Operation.' I

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."

(continued)

Watts Bar-Unit 1 B 3.4-72 Revision 7, 148

ECCS - Operating B 3.5.2 BASES (continued)

SURVEILLANCE sR 3.5.2.1 REQUIREMENTS Verification of proper valve position ensures that the flow path from the ECCS pumps to the RCS is maintained. Misalignment of these valves could render both ECCS trains inoperable. Securing these valves in position by removal of power or by key locking the control in the correct position ensures that they cannot change position as a result of an active failure or be inadvertently misaligned. These valves are of the type, described in Reference 6, that can disable the function of both ECCS trains and invalidate the accident analyses. A 12-hour Frequency is considered reasonable in view of other administrative controls that will ensure a mispositioned valve is unlikely.

sR 3.5.2.2 Veriffing the correct alignment for manual, power operated, and automatic valves in the ECCS flow paths provides assurance that the proper flow paths exist for ECCS operation. This SR does not apply to valves that are locked, sealed, or othenrise secured in position, since these were verified to be in the correct position prior to locking, sealing, or securing. A valve that receives an actuation signal is allowed to be in a nonaccident position provided the valve will automatically reposition within the proper stroke time. This Surveillance does not require any testing or valve manipulation. Rather, it involves verification that those valves capable of being mispositioned are in the correct position. The 31 day Frequency is appropriate because the valves are operated under administrative control, and an improper valve position would only affect a single train. This Frequency has been shown to be acceptable through operating experience.

sR 3.5.2.3 With the exception of the operating centrifugal charging pump, the ECCS pumps are normally in a standby, nonoperating mode. As such, flow path piping has the potential to develop voids and pockets of entrained gases. Maintaining the piping from the ECCS pumps to the RCS full of water by venting the ECCS pump casings and accessible suction and discharge piping high points ensures that the system will perform properly, injecting its full capacig into the RCS upon demand.* This will also prevent water hammer, pump cavitation, and pumping of noncondensible gas (e.9., air, nitrogen, or hydrogen) into the reactor vessel following an Sl signal or during shutdown cooling.** The 31 day Frequency takes into consideration the gradual nature of gas accumulation in the ECCS piping and the procedural controls governing system operation. A note is added to the FREQUENCY that surveillance performance is not required for safety injection hot leg injection lines until startup from the Fall2003 Refueling Outage. (Ref. 7)

(continued)

Watts Bar-Unit 1 B 3.5-17 Revision 54,62, 142 Amendment 43

ECCS - Operating B 3.5.2 BASES (continued)

SURVEILLANCE SR 3.5.2.3 (continued)

REQUIREMENTS

For the accessible locations, UT may be substituted to demonstrate the piping is full of water. An accessible ECCS high point is defined as one that:

1) Has a vent connection installed.

2) The high point can be vented with the dose received remaining within ALARA expectations. ALARA for venting ECCS high point vents is considered to not be within A1ARA expectations when the planned, intended collective dose for the activity is unjustifiably higher than industry norm, or the licensee's past experience, for this (or similar) work activity.

3) The high point can be vented with industrial safety expectations remaining within the industry norm.

t*While lower levels of gas are ideal, due to the robust design of the ECCS system, it can still perform its design functions despite gas volumes below a specifically defined value being present in the system. An evaluation was performed on the effects of the gas on system performance which included transient effects on piping, components, and supports. The gas was also evaluated for the potential to delay the ECCS flow delivery to ensure the analyzed delivery times remained valid. An allowable gas volume for ECCS piping was calculated based on this analysis. (Ref. 8) sR 3.5.2.4 Periodic surveillance testing of ECCS pumps to detect gross degradation caused by impeller structural damage or other hydraulic component problems is required by the American Society of Mechanical Engineers (ASME) OM Code. This type of testing may be accomplished by measuring the pump developed head at only one point of the pump characteristic curve. This verifies both that the measured performance is within an acceptable tolerance of the original pumps baseline performance and that the performance at the test flow is greater than or equal to the performance assumed in the plant safety analysis. SRs are specified in the lnservice Testing Program, which encompasses the ASME OM Code. The ASME OM Code provides the activities and Frequencies necessary to satisfu the requirements.

SR 3.5.2.5 and 3.5.2.6 These surveillances demonstrate that each automatic ECCS valve actuates to the required position on an actual or simulated Sl signal and that each ECCS pump starts on receipt of an actual or simulated S! signal. This Surveillance is not required for valves that are locked, sealed, or otherwise secured in the required position under administrative control. The 18 month Frequency is based on the need to perform these Surveillances under the conditions that apply during a plant outage and the potential for unplanned plant transients if the Surveillances were performed with the reactor at power. The 18 month Frequency is also acceptable based on consideration of the design reliability Watts Bar-Unit 1 B 3.5-1 8 Revision 54, 80, 142 Amendment 43

ECCS - Operating B 3.5.2 BASES (continued)

(and confirming operating experience) of the equipment. The actuation logic is tested as part of ESF Actuation System testing, and equipment performance is monitored as part of the lnservice Testing Program.

SURVEILLANCE sR 3.5.2.7 REQUIREMENTS (continued) Realignment of valves in the flow path on an Sl signal is necessary for proper ECCS performance. These valves are secured in a throttled position for restricted flow to a ruptured cold leg, ensuring that the other cold legs receive at least the required minimum flow. The 18 month Frequency is based on the same reasons as those stated in SR 3.5.2.5 and SR 3.5.2.6.

sR 3.5.2.8 Periodic inspections of the containment sump suction inlet ensure that it is unrestricted and stays in proper operating condition. The advanced sump strainer design installed at \AIBN incorporates both the trash rack function and the screen function. lnspection of the advanced strainer constitutes fulfillment of the trash racUscreen inspection. The 18 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage, on the need to have access to the location, and because of the potential for an unplanned transient if the Surveillance were performed with the reactor at power. This Frequency has been found to be sufficient to detect abnormal degradation and is confirmed by operating experience.

REFERENCES 1. Title 10, Code of Federal Regulations, Part 50, Appendix A, General Design Criterion 35, "Emergency Core Cooling System."

2. Title 10, Code of Federal Regulations, Part 50.46, "Acceptance Criteria for Emergency Core Cooling Systems for Light Water Nuclear Power Plant."

3. Watts Bar FSAR, Section 6.3, "Emergency Core Cooling System."

4. FSAR Bar FSAR, Section 15.0, 'Accident Analysis."

5. NRC Memorandum to V. Stello, Jr., from R.L. Baer, "Recommended lnterim Revisions to LCOs for ECCS Components," December 1 , 1975.

6. lE lnformation Notice No. 87-01 , "RHR Valve Misalignment Causes Degradation of ECCS in PWRs," January 6, 1987.

7. WBN License Amendment Request WBN-TS-03-1 1 dated Apri! 8, 2003.

8. NEI 09-1 0, Revision 1a-A " Guidelines for Effective Prevention and Management of System Gas Accumulation," dated April ,2013.

Watts Bar-Unit 1 B 3.5-19 Revision 54, 80, 142 Amendment 43

EGTS B 3.6.9 BASES BACKGROUND The prefilters remove large particles in the air, and the moisture separators (continued) remove entrained water droplets present, to prevent excessive loading of the HEPA filters and charcoal absorbers. Heaters are included to reduce the relative humidity of the airstream on systems that operate in high humidity.

Operation with the heaters on for 2 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 conective actions. Cross-over flow ducts are provided between the two trains to allow the active train to draw air through the inaciive train and cool the air to keep the charcoal beds on the inactive train from becoming too hot due to absorption of fission products.

The containment annulus vacuum fans maintain the annulus at - 5 inches water gauge vacuum during normal operations. During accident conditions, the containment annulus vacuum fans are isolated from the air cleanup portion of the system.

The EGTS reduces the radioactive content in the shield building atmosphere following a DBA. Loss of the EGTS could cause site boundary doses, in the event of a DBA, to exceed the values given in the licensing basis.

APPLICABLE The EGTS design basis is established by the consequences of the limiting SAFETY ANALYSES DBA, which is a LOCA. The accident analysis (Ref. 3) considers two different single failure scenarios. The first one assumes that only one train of the EGTS is functional due to a postulated single failure that disables the other train. An alternate scenario assumes a single failure of the pressure control loop associated with one train of PCOs. The first scenario is bounding for thyroid dose while the alternate scenario is bounding for beta and gamma doses. The accident analysis accounts for the reduction in airborne radioactive material provided by the number of filter trains in operation for each failure scenario. The amount of fission products available for release from containment is determined fora LOCA.

The safety analysis conservatively assumes the annulus is at atmospheric pressure prior to the LOCA. The analysis further assumes that upon receipt of a Containment lsolation Phase A (ClA) signal from the RPS, the EGTS fans automatically start and achieve a minimum flow of 3600 cfm per train within 18 seconds (20 seconds from the initiating event.) This does not include 10 seconds for diesel generator startup. The analysis shows that the annulus pressure will rise to a positive value and then decrease to the EGTS control point for a single failure of one EGTS train, or slightly more negative for a single failure of a pressure control loop associated with one train of PCOs. The normalalignment for both EGTS control loops is the A-Auto position. With both EGTS con-trol loops in A-Auto, both trains will function upon initiation of a CIA signal. ln the event of a LOCA, the annulus vacuum control system isolates and both trains of the EGTS pressure control loops will be placed in service to maintain the required negative pressure. lf annulus vacuum is lost during normal operations, the A-Auto position is unaffected by the loss of vacuum. This operational configuration is acceptable because the accident dose analysis conservatively assumes the annulus is at atmospheric pressure at event initiation. (Ref. 6)

The EGTS satisfies Criterion 3 of the NRC Policy Statement.

(continued)

Watts Bar-Unit 1 B 3.6-56 Revision 84, 85, 101 , 102, 138 Amendment 1 15

EGTS B 3.6.9 BASES ACTIONS B.1 and 8.2 (continued) most repairs. lf the EGTS train 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 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and to MODE 5 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . 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 3.6.9.1 REQUIREMENTS Operating each EGTS train for > 15 minutes with heaters on ensures that all trains are OPERABLE and that all associated controls are functioning properly.

It also ensures that blockage, fan or motor failure, or excessive vibration can be detected for corrective action. The 31 day Frequency was developed in consideration of the known reliability of fan motors and controls, the two train redundancy available.

sR 3.6.9.2 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. lt should be noted that for the EGTS, the VFTP pressure drop value across the entire filtration unit does not account for instrument enor (Ref.5).

(continued)

Watts Bar-Unit 1 B 3.6-s8 Revision 29, 138 Amendment 1 15

AFW System B 3.7.5 BASES BACKGROUND The AFW Systeh 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 psifor accumulation and pressure drop between the SG and MSSV). 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 generatorwater level -

low-low by the ESFAS (LCO 3.3.2). The motor driven pumps start on a two-out-ofthree low-low level signal in any steam generator and the turbine driven pump starts on a two-out-of-three low{ow level signal in any two steam generators.

The system also actuates on loss of offsite power, safety injection, and trip of both turbinedriven 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 normal SAFETYANALYSES feedwater.

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 psifor accumulation and pressure drop between the SG and MSSV.

ln 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.

(continued)

Watts Bar-Unit 1 B 3.7-25 Revision 147

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 OPEMBLE when the individual components necessary to control exposure in the fuel handling building are OPEMBLE in both trains. An ABGTS train is considered OPERABLE when its associated:

a. Fan is OPEMBLE;

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 individualwill 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.

APPLICABILITY ln MODE 1,2,3, or 4, the ABGTS is required to be OPEMBLE to provide fission product removal associated with ECCS leaks due to a LOCA and leakage from containment and annulus.

ln MODE 5 or 6, the ABGTS is not required to be OPERABLE since the ECCS is not required to be OPEMBLE.

(continued)

Watts Bar-Unat 1 B 3.73 Revision 55, 87, 1 19, 139 Amendment 92, 1 16

ABGTS B 3.7.12 BASES (continued)

ACTIONS 4.1 With one ABGTS train inoperable, action must be taken to restore OPERABLE status within 7 days. During this period, the remaining OPERABLE train is adequate to perform the ABGTS function. The 7 day Completion Time is based on the risk from an event occurring requiring the inoperable ABGTS train, and the remaining ABGTS train providing the required protection.

B.1. B.2 and B.3 lf the ABSCE boundary is inoperable, the ABGTS trains cannot perform their intended functions. Actions must be taken to restore an OPEMBLE ABSCE boundary within seven days. During the period that the ABSCE boundary is inoperable, action must be inltiated to implement mitigating actions consistent with the intent, as applicable, of GDC 19, 60, 61, 63, &t and 10 CFR Part 100 (Ref. 7) to protect plant personnelfrom potential hazards such as radioactive contamination, temperature and relative humidity, and physical security. Actions must be taken within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to verify that, in the event of a DBA, main control room occupant radiological exposures will not exceed 10 CFR 50 Appendix A GDC 19 limits. These mitigating actions (i.e., actions that are taken to offset the consequences of the inoperable ABSCE boundary) should be preplanned to address these concerns for intentional and unintentional entry into the condition.

The 24-hour Completion Time is reasonable based on the low probability of a DBA occurring during this time period, and the use of mitigating actions. The seven-day Completion Time is reasonable based on the determination that the mitigating actions will ensure protection of main control room occupants within analped limits (Ref. 11) while limiting the probability that main control room occupants will have to implement protective measures that may adversely affect their ability to controlthe reactor and maintain it in a safe shutdown condition in the event of a DBA. ln addition, the seven-day Completion Time is a reasonable time to diagnose, plan and possibly repair, and test most problems with the ABSCE boundary.

C.1 and C.2 When Required Actions A.1 or Required Actions 8.1,8.2, and B.3 cannot be completed within the associated Completion Time, orwhen both ABGTS trains are inoperable for reasons other than an inoperable ABSCE boundary (i.e.,

Condition B), the plant must be placed in a MODE in which the LCO does not apply. To achieve this status, the plant must be placed in MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and in MODE 5 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . The Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

(continued)

Watts Bar-Unit 1 B 3.7$4 Revision 1 19, 139 Amendment 92, 116

ABGTS B 3.7.12 BASES SURVEILLANCE SR 3.7.12.1 REQUIREMENTS Standby systems should be checked periodically to ensure that they function properly. As the environmentaland normal operating conditions on this system are not severe, testing each train once every month provides an adequate check on this system.

Operation with the heaters on for > 15 continuous minutes demonstrates OPEMBILITY of the system. Periodic operation ensures that heaterfailure, blockage, fan or motor failure, or excessive vibration can be detected for conective action. The 31 day Frequency is based on the known reliability of the equipment and the two train redundancy available.

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 fitter 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.

(continued)

Watts Bar-Unit 1 B 3.7-65 Revision 1 19, 138 Amendment 92, 115

ABGTS B 3.7.12 BASES REFERENCES 5. Deleted (continued)

6. Regulatory Guide 1.4, 'Assumptions Used for Evaluating the Potential Radiological Consequences of a Loss of Coolant Accident for Pressurized Water Reactors."

7. Title'10, Code of FederalRegulations, Part 100.11, "Determination of Exclusion Area, Low Population Zone, and Population Center Distance."

8. Regulatory Guide 1.52 (Rev. 2), "Design, Testing and Maintenance Criteria for Post Accident Engineered-Safety-Feature Atmospheric Cleanup System Air Filtration and Adsorption Units of Light-Water Cooled Nuclear Power Plants."

9. NUREG-0800, Section 6.5.1, "Standard Review Plan," Rev.2,'ESF Atmosphere Cleanup System," July 1981.

10. Watts Bar Drawing 1-47W605-242, "ElectricalTech Spec Compliance Tables."

11. TVA Calculation MDQ0000302014000618, 'Offsite and ControlRoom Doses without the Auxiliary Building Secondary Containment Enclosure (ABSCE) during a LOCA."

Watts Bar-Unit 1 B 3.7-67 Revision 29, 55, 119, 139 Amendment 92, 1 16

ENCLOSURE 3 WBN UNIT 1 TECHNICAL REQUIREMENTS MANUAL TABLE OF CONTENTS

TABLE OF CONTENTS TECHNICAL REQUIREMENTS TABLE OF CONTENTS LIST OF TABLES ............................ v LIST OF F!GURES.... ...................... vi LIST OF ACRONYMS ..................... vii LIST OF EFFECTIVE PAGES ........ viii 1.0 1.1 1.2 1.3 1.4 TR 3.0 TR 3.1 REACTTVTTY CONTROL SYSTEMS ............ 3.1_1 TR 3.1 .1 Boration Systems Floirr Paths, Shutdorvn ............. 3.1-1 TR 3.1.2 Boration Systems Flor Paths, Operating .............3.1-3 TR 3.1.3 Charging Pump, Shutdown ..............3.1-s TR 3.1.4 Charging Pumps, Operating..... ........3.1 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 lndication System, Shutdown 3.1-13 TR 3.3 INSTRUMENTATION ............... 3.3_1 TR 3.3.1 ReactorTrip System (RTS) lnstrumentation........ ......................3.3-1 TR 3.3.2 Engineered Safety Features Actuation System (ESFAS) lnstrumentation................ ..............3.3-5 TR 3.3.3 Movable lncore Detectors................ ..................... .9.3-12 TR 3.3.4 Seismic lnstrumentation ............... --9.9-14 TR 3.3.5 Turbine Overspeed Protection ......... 3.3-18 TR 3.3.6 Loos+Part Detection System......... ..3.3-20 TR 3.3.7 Plant Calorimetric Measurement............ S.g-22 TR 3.3.8 Hydrogen Monitors ......9.9-24 TR 3.3.9 Poirrer Distribution Monitoring System (PDMS)....... 3.3-26 TR 3.4 RFACTOR COOl-ANT SYSTEM (RCS) .......... ..................3.4_1 TR 3.4.1 SafetyValves, Shutdown ..................9.4-1 TR 3.4.2 Pressurizer Temperature Limits........... ................. 3.4-3 TR 3.4.3 RCS Vents ................... 3.4-5 TR 3.4.4 Chemistry.... .................2.4-T TR 3.4.5 Piping System Structurallntegrity ....3.4-10 TR 3.6 CoNTATNMENT SYSTEMS.................. ........ 3.6_1 TR 3.6.1 lce Bed Temperature Monitoring System .............3.6-1 TR 3.6.2 lnlet Door Position Monitoring System .................. 3.6-4 TR 3.6.3 Lorer Compartment Cooling (LCC) System ........ 3.66 Watts Bar-Unit 1 Technical Req u irements Revision 56

TABLE OF CONTENTS (continued)

TR 3.7 pt-ANT SYSTEMS ....................3.7_1 TR 3.7.1 Steam Generator Pressure/

Temperature Limitations... ..3.7-1 TR 3.7.2 Flood Protection Plan ..9.1_g TR 3.7.3 DELETED. 3.7-10 TR 3.7.4 Sealed Source Contamination................... g.T_22 TR 3.7.5 Area Temperature Monitoring .............. 3.1_26 TR 3.8 ELECTRTCAL POWER SYSTEMS ...............3.8_1 TR 3.8.1 lsolation Devices........ ..3.g_1 TR 3.8.2 Containment Penetration Conductor Overcurrent Protection Devices ..............3.9-s TR 3.8.3 Motor-Operated Valves Thermal Overload Bypass Devices........ 3.9_10 TR 3.8.4 Submerged Component Circuit protection 3.g_11 TR 3.9 REFUELTNG OPERATlONS.................. ........ 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.g_s 5.0 ADMTNlSTMTlVE CONTRO1S................ ...5.0_1 5.1 TechnicalRequirements (TR) Controlprogram ... S.0_1 Watts Bar-Unit 1 Techn ical Req uirements Revision 62

TABLE OF CONTENTS (continued)

BASES B 3.0 TECHNTCAL REQUTREMENTS CrR) AND TECHNTCAL SURVETLLANCE REQUTREMENTS CrSR)

APPL!CABlLlW.............. 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 Sptems Flom Paths, Operating.... .. B 3.1-5 B 3.1.3 Charging Pump, Shutdown. ...... B 3.1-g B 3.1.4 Charging Pumps, Operating ..... B 3.1-11 B 3.1 .5 Borated Water Sources, Shutdown B 3.j-14 B 3.1.6 Borated Water Sources, Operating..... B 3.1-1g B 3.1 .7 Position lndication System, Shutdown.... ....... B 9.1-23 B 3.3 INSTRUMENTATION B 3.3-1 B 3.3.1 ReactorTrip System (RTS) lnstrumentation . B 3.3-1 B 3.3.2 Engineered Safety Features Actuation System (ESFAS) lnstrumentation................ B 3.3-4 B 3.3.3 Movable lncore Detectors................. B 3.3-7 B 3.3.4 Seismic lnstrumentation................ B 3.3-10 B 3.3.5 Turbine Overspeed Protection.... B 3.3-14 B 3.3.6 Loose-Part Detection System ... B 3.3-1g B.3.3.7 Plant Calorimetric Measurement............. ....... B 3.3-21 B 3.3.8 Hydrogen Monitors...... 83.3-25 B 3.3.9 Porer Distribution Monitoring Sptem (PDMS)........ .......... 83.3-30 B 3.4 REACTOR COOl3NT SYSTEM (RCS)........... B 3.4_1 B 3.4.1 SafetyValves, Shutdown .......... B 3.4-1 B 3.4.2 PressurizerTemperature Limits........... .......... B 3.4-4 B 3.4.3 RCSVents... .........89.4-7 B 3.4.4 Chemistry B 3.4-10 B 3.4.5 Piping System Structural lntegnty........ .......... B 9.4-14 B 3.6 CoNTATNMENT SYSTEMS.................. ........ B 3.6_1 B 3.6.1 lce Bed Temperature Monitoring Syrstem......... B 3.6-1 B 3.6.2 lnlet Door Position Monitoring System........ ... B 3.66 B 3.6.3 Lorer Compartment Cooling (LCC) System . B 3.6-10 B 3.7 PI.ANT SYSTEMS 837-1 B 3.7.1 Steam Generator Pressure/Temperature Limitations.... ..... B 3.2-1 B 3.7.2 Flood Protection Plan....... .........83.74 B 3.7.3 DELETED.... ......... B g1-j2 B 3.7.4 Sealed Source Contamination.................. ..... B 3.7-19 B 3.7.5 Area Temperature Monitoring.............. .......... B g.Z-22 B 3.8 ELECTRICAL POWER SYSTEMS B 3.8-1 B 3.8.1 lsolation Devices .. B 3.g-1 B 3.8.2 Containment Penetration Conductor Overcunent Protection Devices ....... B 3.9-7 B 3.8.3 Motor-Operated Valves Thermal Overload Bypass Devices B 3.9-15 B 3.8.4 Submerged Component Circuit Protection.... B 3.9-19 continued Watts Bar-Unit 1 Technical Req uirements Revision 62

TABLE OF CONTENTS (continued)

B 3.9 REFUELTNG OPEMTlONS.................. ........ 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 PoolBuilding B 3.9-B Watts Bar-Unit 1 Technical Req uirements Revision 53

LIST OF TABLES Table No. Title paoe 1.1-1 MoDES .........:........... ...1.r.

3.3.1-1 ReactorTrip System lnstrumentiation ResponseTimes........ .....3.3-3 3.3.2-1 Engineered Safety Features Actuation System Response Times.......... ....................3.3-7 3.3.4-1 Seismic Monitoring lnformation... ......9.3-11 g.7.3-',t - 3.7.3-5....... ...........DE1ETED I 3.7.5-1 Area Temperature Monitoring...... ,..9]-29 3.8.3-1 Motor-OperatedValvesThermalOverloadDevicesWhich Are Bypassed UnderAccident Condilions ....................3.8-12 3.8.4-1 Submerged Components Wilh Automatic De-energization UnderAccident Conditions ..3.8-19 Watts Bar-Unit 1 Technical Req uirements Revision 62

LISTOF FIGURES Fioure No. Title paqe 3.1.6 BoricAcid rank Limits Based on RWST Boron concentration.......................3.1-12a 3.7.3-1 DELETED LIST OF MISCELTANEOUS REPORTS AND PROGMMS Core Operating Limits Report Watts Bar-Unit 1 vt Technical Req uirements Revision 62

LIST OF ACRONYMS Acronvm 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 Particutate Air HVAC Heating, ventilating, and Air-conditioning LCC Lower Compartment Cooler LCO Limiting Condition For Operation MFIV Main Feedwater lsolation Valve MFRV Main Feedwater Regulation Valve MSIV Main Steam Line lsolation 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 PM Pressure lsolation Valve PORV Power-Operated Relief Valve PTLR Pressure and remperature 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 Sl Safety Injection SL Safety Limit SR Surveillance Requirement UHS Ultimate Heat Sink Watts Bar-Unit 1 vI Technical Requirements Revision 46

TECHNICAL REQUIREMENTS LIST OF EFFECTIVE PAGES Page Revision Page Revision Number Number Number Number i 56 3.1-4 0 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 6s 3.1-1 1 33 ix 64 3.1-12 0 x 66 3.1-12a 42 xi 62 3.1-13 8 xii 22 3.3-1 0 xiii 37 3.3-2 0 xiv 47 3.3-3 34

)(V 58 3.3-4 44

>cvi 66 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 3 1 .1-5 0 3.3-10 0 1.1 -6 0 3.3-1 1 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 1.34 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 46 1.3-13 0 3.3-27 46 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 3.44 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 vlil Techn ical Req u irements Revision 66

TECHNICAL REQU I REMENTS 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 3.8-1 1 0 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 18 3.7-3 17 3.8-19 18 3.7-4 17 3.9-1 53 3.7-s 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 3.84 25 3.8-5 0 3.8-6 0 Watts Bar-Unit 1 tx Technical Req uirements Revision 64

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 3.04 38 B 3.3-16 6 B 3.0-5 38 B 3.3-17 38 B 3.0 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 3.0-1 1 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 3.14 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 20 B 3.3-31 54 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 il 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 0 B 3.4-14 u B 3.1-25 8 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 3.34 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 3.3-1 1 40 B 3.6-10 0 B 3.3-12 40 B 3.6-1 1 0 Watts Bar-Unit 1 Technical Req uirements Revision 66

TECHNICAL REQUIREMENTS BASES LIST OF EFFECTIVE PAGES Page Revision Page Revision Number Number Number Number B 3.6-12 0 B 3.8-22 18 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 3.74 57 B 3.9-4 0 B 3.7-5 17 B 3.9-5 28 B 3.7 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 2s B 3.8-7 25 B 3.8-8 0 B 3.8-9 0 B 3.8-10 0 B 3.8-1 1 0 B 3.8-12 0 B 3.8-13 25 B 3.8-14 25 B 3.8-1s 0 B 3.8-16 0 B 3.8-17 0 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 Req uirements Revision 62

TECHNICAL REQUIREMENTS MANUAL LIST OF EFFECTIVE PAGES - REVISION LISTING Revisions lssued SUBJECT Revision 0 09-30-95 lnitial lssue 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 09-1 2-97 Change OPAT/OTAT Response Time Revision 8 Og-22-g7 Clarification of Surveillance Frequency for Position lndication System Revision 9 10-10-97 Revised Boron Concentration for Borated Water Sources Revision 10 12-17-98 ICS lnlet 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 UpgradeSeismicMonitoring lnstruments Revision 20 03/13/00 Added Notes to Address lnstrument Error for Various Parameters Revision 21 04113/00 COLR, Cycle 3, Rev 2 Revision 22 07107100 Elimination of Response Time Testing Watts Bar-Unit 1 xii Technical Requirements Revision 22

TECHNICAL REQUIREMENTS MANUAL LIST OF EFFECTIVE PAGES - REVISION LISTING Revisions lssued SUBJECT Revision 23 UDAU Phnt Calorimetric (LEFM)

Revision 24 03119101 TRM Change Controt Program per 50.59 Rute Revision 25 05115101 change in Preventive Maintenance Frequency for Molded case Circuit Breakers Revision 26 05129101 Change CVI Response Time from 5 to 6 Seconds Revision 27 01131102 Change pH value in the borated water sources due to TS change for ice weight reduction Revision 28 02105102 Refueling machine upgrade under DCN D-50991-A Revision 29 02126102 Added an additionat action to TR 3.7.3 to perform an engineering evaluation of inoperable snubbe/s impact on the operability of a supported system.

Revision 30 06,105102 Updated TR 3.3.5.1 to reflect implementation of the TtpTOp program in a Technical lnstruction (Tl).

Revision 31 10131102 Conect RTP to 3459 trl\ /t (PER 02-9519000)

Revision 32 09117103 Editorialcorrection to Bases forTSR 3.1.S.3.

Revision 33 10114103 Updated TRs 3.1.5 and 3.1.6 and their respective bases to incorporate boron concentration changes in accordance with change packages WBN-TS42- 1 4 and WBN-TS-0&0.1 7.

Revision 34 05l14lV Revised ltem 5, 'Source Range, Neutron Ftux'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 rable 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 09/25106 Revised the response time for Containment Spray in Table 3.3.2-1and the RTxpl values in the Bases for TR 3.7.1. Both changes result from the replacement of the steam genenators.

Revision 37 11108106 Revised TR 3.1.5 and 3.1.6 and the Bases for these TRs to update the boron concentration limits of the RV\IST and the BAT.

Watts Bar-Unit 1 xilt Technical Req uirements Revision 37

TECHNICAL REQUIREMENTS MANUAL LIST OF EFFECTIVE PAGES REVISION LISTING Revisions lssued SUBJECT Revision 38 '11129106 Updated the TRM to be consistentwith Tech Spec Amendment 55. TRM Revision 38 modified the requirements formode change limitations in TR 3.0.4 and TSR 3-0.4 by incorporating changes similar to those ouflined in TSTF-359, Revision 9. (T5-06-24)

Revision 39 ul16l07 Updated the TRM to be consistent with rech spec Amendment 42.

TRM ReMsion 39 modified the requirements of TSR 3.0.3 by incorporating changes similarto those ouflined in TSTF-3SS.

os-07-03)

Revision 40 05124107 Updated the TRM and Bases to remove the various requirements for the submiftat of reports to the NRC. (TS-07-06)

Revision 41 05125107 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 Xl of theASME Boiler and Pressure Vessel Code with the ASME Operation and Maintenance Code for lnservice Testing (lST) activities and removes reference to'applicable supports, from the IST program.

Revision 42 $120n008 Revision 42 updates Figure 3.1.6 to remove the 240 TpBAR Limit.

Revision 43 0711712008 Revision 43 removes a reporting requirementfrom TR 3.7.4, "Sealed Source Contamination.' The revision also updates the Bases for TR 3.7.4.

Revision 44 1011012008 Revision 4,4 ufiates Table 3.3.1-1 to be consistent with the changes approved by NRC as Tech SpecAmendment6g.

Revision 45 0?/2312009 Added rR 3.3.8, "Hydrogen Monitors,' and the Bases for TR 3.3.g.

This change is based on Technical Specification (l-S) Amendment 72 which removed the Hydrogen Monitors (Function 13 of LCO 3.3.3) from the TS.

Revision 46 091202,010 Revision 46 implements changes from License Amendment g2 (Technical Specification (TS) Bases Revsion 1O4) for the approved BEACON-TSM application of the power Distribution Monitoring System (PDMS).

Revision 47 1010A12010 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.

Watts Bar-Unit 1 Techn ical Req uirements Revision 47

TECHNICAL REQUIREMENTS MANUAL LIST OF EFFECTIVE PAGES REVISION LISTING Revisions lssued SUBJECT Revision 48 UnA2U1 CANCELLED Revision 49 05124120'11 Revision 49 updated Note 14 of rable 9.9.2-1 to clariff that the referenced time is onlyfor'partial'transfer of the ECCS pumps from theVCTtothe RWST.

Revision 50 Pl2r2ul clarifies the acceptability of periodically using a portion of lhe 25o/o grace period in TSR 3.0.2 to facilitate 13 week maintenance work schedules.

Revision 51 08/091201 3 Adds a note to TR 3.1 .2 and TR 3. 1 .4 to permit securing one charging pump in orderto supporting transition into orfrom the Applicability of Technical Specification 3.4.12 (pER 593365).

Revision 52 0813012013 Clarifies that TR 3.4.5, "Piping System Structural lntegrity,, applies to all ASME Code Class '1,2, and 3 piping systems, and is not limited to reactor coolant system piping.

Revision 53 12/1212013 Technical specification Amendment 92 added Limiting condition for Operation (LCO) 3.9.10, 'Decay Time," which was redundant to Technical Requirement CfR) 3.9.1, "Decay Time.' Revisbn 53 removes TR 3.9.1 from the Technical Requirements Manual (TRM) and the TRM Bases.

Revision 54 0112312014 TRM which updates Technical Requirement (TR) 3.3.9, .power Distribution Monitoring System," to reflect the Addendum to WCAp 12472-P-A.

Revision 55 0111412015 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 o4.1302015 This revision restructures TR 3.6.2 CONDITlONS, REQUlRED 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 0510712015 This revision changes the elevation of the Mean Sea Level by submergence during floods vary from714.5 ft to 739.2 ft in TRM Bases 83.7.2, Flood Protection Plan.

Revision 58 05l19f201i This revision is an administrative change in TRM Bases 3.4.5 background information.

Watts Bar-Unit 1 Technical Req uirements Revision 58

TECHNICAL REQUIREMENTS MANUAL LIST OF EFFECTIVE PAGES REVISION LISTING ReMsions tssued SUBJECT Revision 59 1011312015 This revision adds the Unit 1 and Unit 2 FCV7-0066 and FCV-67-0067 valves to TRM Table 3.8.$1.

Revision 60 CF,10112016 This revision is to add 2-FCV-70-153 vatve to TRM Table 3.8.3-1 Sheet 4 of 5.

Revision 61 0A212:017 Revises TRM Bases 3.6.2 "lnlet Door Position Monitoring System' actions.

Revision 62 03131D0'17 This revision deletes TRM and TRM Bases section 3.7.3, Snubbers" via the License Amendment 11 1.

Revision 63 511712017 Revises the obsolete analog system that was limited to monitoring 1 sensorfor each RCS collection point.

Revision 64 81221'17 Clarified ASME Code Class in the section description in Section 3.4.5, Piping System Structurat tntegrity.

Revision 65 416118 Revised TRM Bases Section 3.6.2, to more closely match information provided in the UFSAR. The Bases as written limits credit for the loruer inlet door main panel annunciator as part of the lnlet Door Position Monitoring System.

Revision 66 (Amendment 10111118 Revises TRM Bases Section 3.3.5, "Turbine Overspeed 119) Protection", to change the background information.

I I

Watts Bar-Unit 1 xvi Technical Req uirements Revision 66

ENGLOSURE 4 WBN UNIT 1 TECHNICAL REQUIREMENTS MANUAL CHANGED PAGES

Tu rbine Overspeed Protection B 3.3.5 B 3.3 INSTRUMENTATION B 3.3.5 Turbine Overspeed Protection BASES BACKGROUND The Digital Electro Hydraulic (DEH) system provides for redundant and diverse overspeed protection to isolate main steam to the turbo-generator when the rated operating speed of 1800 rpm is exceeded. The DEH overspeed trip function which is set at 1854 rpm (103 percent of rated speed) will initiate a turbine trip by closing all steam valves (throttle, governor, reheat, stop and interceptor valves).

This trip function will trip (open) the emergency trip header valves, which will result in hydraulic closure of the aforementioned steam valves. \Mth this arrangement, the DEH provides diverse trip functions to prevent the turbine speed from exceeding 120 percent of rated speed. DEH is a fault tolerant system, such that the loss of a single speed probe or module will not result in an unnecessary trip, but will not impede the system from performing its intended function.

lf for some reason the DEH turbine trip does not function and the turbine speed increases to 1980 rpm (110 percent of rated speed), the lndependent Overspeed Protection System (IOPS) will initiate a turbine trip by opening the emergency trip header valves which will then hydraulically close all steam valves (throttle, governor, reheat, stop, and interceptor valves) and prevent the turbine speed from exceeding 120 percent of rated speed. The untilwillthen coast down to turning gear operation. The IOPS has separate speed probes and control modules, independent of any speed instrumentation from the DEH system. The IOPS is a fault tolerant system such that a failure of a single speed probe or module will not result in an unnecessary trip, but will not impede the system from performing its intended function.

(continued)

Watts Bar-Unit 1 B 3.3-14 Revision 66 Tech nical Requ irements

lnlet Door Position Monitoring System B 3.6.2 B 3.6 CONTAINMENT SYSTEMS B 3.6.2 Inlet Door Position Monitoring System BASES BACKGROUND Ninety-six limit switches monitor the position of the lower inlet doors. Two switches are mounted on the door frame for each door panel. The position and movement of the switches are such that the doors must be effectively sealed before the switches are actuated. A single annunciator window in the control room gives a common alarm signalwhen any door is open. open/shut indication is also provided at the lower inlet door position display panel located in the Main Control Room. For door monitoring purposes, the ice condenser is divided into six zones, each containing four inlet door assemblies, or a total of eight door panels. The limit switches on the doors in any single zone are wired to a single iignt on the inlet door position display panel such t6at a closed light indicateslhat all the doors in that zone are shut and an open light indicates that one or more doors in that zone are open (Ref. 1). Additional information on the design of the lower inlet door monitoring instrumentation is provided in UFSAR Section 6.7.1S.

Monitoring of inlet door position is necessary because the inlet doors form the barrier to air flow through the inlet ports of the ice condenser for normal unit operation. Failure of the lnlet Door Position Monitoring System requires an alternate OPERABLE monitoring system to be used to ensure that the ice condenser is not degraded.

APPLICABLE Proper operation of the inlet doors is necessary to mitigate SAFETY ANALYSES the consequences of a loss of coolant accident or a main steam line break inside containment. The lnlet Door Position Monitoring System, however, is not required for proper operation of the inlet doors, nor is it considered OPERABLE as an initial condition for a DBA. Hence, the lnlet Door Position Monitoring System is not a consideration in the analyses of DBAs. Based on the PRA Summary Report in Reference 2, the lnlet Door Position Monitoring System has not been identified as a significant risk contributor.

(continued)

Watts Bar-Unit 1 B 3.6-6 4t6t18 Tech nical Requirements Revision 10, 65

ENGLOSURE 5 WBN UNIT 2 TECHNIGAL SPECIFICATION BASES TABLE OF CONTENTS

TABLE OF CONTENTS TABLE OF CONTENTS ...

LIST OF TABLES ..

VI LIST OF FIGURES Vi LIST OF ACRONYMS ......

vii LIST OF EFFECTIVE PAGES x

B 2.0 SAFETY LIMITS (SLs) ... B2.O_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_l B 3.0 LlMITING CONDtTtoN FoR opERATtON (LCO)

APPLICABILITY ... B 3.0-1 B 3.0 SURVEILLANCE REQUIREMENT (SR) APPLICABIL]TY ... B 3.0-11 B 3.1 REACTIVITY CONTROL SYSTEMS ...... B 3.1-1 B 3.1 .1 SHUTDOWN MARGTN (SDM) _ T"un r 2OO"F B 3.1_1 B 3. 1.2 SHUTDOWN MARGIN (SDM) - T",s ( 2oo"F

... B 3.1-8 B 3.1 .3 Core Reactivity ...

... Bg.1_12 B 3 .1.4 Moderator Temperature Coefficient (MTC)

...... B 3.1_1g B 3.1 .5 Rod Group Alignment Limits B g.1_20 B 3.1 .6 Shutdown Bank lnsertion Limits B 3.1_35 B 3.1 .7 Control Bank lnsertion Limits B 3.1_40 B 3.1 .g Rod Position lndication B 3.1-4g B 3.1 .g PHYSICS TESTS Exceptions_MODE 1 ......

B 3.1_Sz B 3.1 .10 PHYSICS TESTS Exceptions_MODE 2 ......

B g.1_64 B 3.2 POWER DISTRIBUTION LIMITS B 3.2-1 B 3.2,1 Heat Flux Hot Channet Factor (FoG))

B 3.2-1 B 3.2.2 Nuclear Enthalpy Rise Hot Channel Factor (fXH) . B 3.2-14 B 3.2.3 AXlAL FLUX DTFFERENCE (AFD)

B 3.2-21 B 3.2.4 QUADMNT POWER TILT RATlO (OPTR)

B 3.2-26 Watts Bar - Unit 2 (continued)

TABLE OF CONTENTS B 3.3 INSTRUMENTATION B 3.3-1 B 3.3.1 Reactor Trip System (RTS) lnstrumentation ... B 3.3-1 B 3.3.2 Engineered Safety Feature Actuation System (ESFAS) lnstrumentation ... B 3.3-64 B 3.3.3 Post Accident Monitoring (PAM) lnstrumentation... B 3.3-122 B 3.3.4 Remote Shutdown System B 3.3-139 B 3.3.5 Loss of Power (LOP) Diesel Generator (DG) Start lnstrumentation ... B 3.3-144 B 3.3.6 Containment Vent lsolation lnstrumentation ... B 3.3-151 B 3.3.7 Control Room Emergency Ventilation System (CREVS)

Actuation lnstrumentation ... B 3.3-159 B 3.3.8 Auxiliary Building Gas Treatment System (ABGTS) Actuation lnstrumentation ... 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, B 3.4-31 B 3.4.8 RCS Loops - MODE 5, B 3.4-35 B 3.4.9 Prgssurizgr .. ! .. .. .,. . ... . . ... . ... .. B 3.4-38 B 3.4.10 Prgssurizer 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 lsolation Valve (PlV) Leakage ... ... ... ... ... ... .. .. ... B 3.4-71 B 3.4.15 RCS Leakage Detection lnstrumentation ... ... ... . .. ... B 3.4-76 B 3.4.16 RCSSpecificActivity .i... .......... ..... ............... B 3.4-82 B 3.4.17 Steam Generator (SG) Tube Integrity ... . ... .. .. ... ... ... ...

. B 3.4-88 Watts Bar - Unit 2 (continued)

TABLE OF CONTENTS B 3.5 EMERGENCY CORE COOLTNG 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 lsolation 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 Containmgnt Spray System ... ... ... ... ... . ... ... ... ... ... . B 3.6-34 B 3.6.7 RESERVED FOR FUTURE ADDITION tI.It ..... B 3.6-41 B 3.6.9 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 Rgturn System (ARS) .....,,,....... ..... ........... B 3.6-54 B 3.6.1 1 B 3.6-59 B 3.6.12 lce Condenser Doors B 3.6-69 B 3.6.13 Divider Barrier lntegrity 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 lsolation Valves (MSlVs) B 3.7-8 B 3.7.3 Main Feedwater lsolation Valves (MFlVs) 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 ilt

TABLE OF CONTENTS B 3.7 PLANT SYSTEMS (continued)

B 3 .7.7 Component Cooling System (CCS) B 3.7-36 B 3.7.9 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 PoolWater Level ... B 3.7-68 B 3 .7.14 Secondary Specific Activity B 3.7-71 B 3.7.15 Spent Fuel Assembly Storage . ... . ... ... ... . ... ... .. B 3.7-74 B 3 .7.16 Component Cooling System (CCS) - Shutdown .. B 3.7-ll B 3.7.17 Essential Raw Cooling Water (ERCW) - Shutdown B 9.7-U 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, ,ro;;;-;;;;; . 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 lnverters - Operating B 3.8-78 B 3.8.8 lnverters - Shutdown ... .. . ... . B 3.8-82 B 3.8.9 Distribution Systems - Operating ..... .......... ..... B 3.9-96 B 3.8.10 DistributionSystems-Shutdown..... ...... B 3.9-95 B 3.9 REFUELING OPERATIONS B 3.9-1 B 3.9.1 B 3.9-1 B 3.9.2 B 3.9-5 B 3.9.3 Nuclear lnstrumentation ... ... ... . B 3.9-8 B 3.9.4 RESERVED FOR FUTURE ADDITION ..t.. 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 - Unat2

TABLE OF CONTENTS B 3.9 REFUELING OPERATIONS (continued)

B 3.9.6 Residual Heat Removal(RHR) and Coolant Circulation - LowWater Level ... B 3.9-16 B 3.9.7 Refueling CavityWater 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

TABLE OF CONTENTS LIST OF TABLES TABLE NO TITLE PAGE B 3.8.1-2 TS Action or Surveillance Requirements Contingency Actions.......... B3.g-37a B 3.8.9-1 AC and DC Electrical Power Distribution Systems ....... B 3.g-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 lnsertion vs. Percent RTP .... . . ......... ..... B 3.1-47 B 3.2.1-1 K(Z) - Normalized Fo(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 .....t...I B 3.2-25 Watts Bar - Unit 2 VI Amendment 5

LIST OF ACRONYMS ACRONYM TITLE ABGTS Auxiliary Building Gas Treatment System ACRP Auxiliary Contro! Room Panel AFD Axial Flux Difference AF1ru 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 - Unat2

LIST OF ACRONYMS 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 lsolation Valve MFRV Main Feedwater Regulation Valve MSIV Main Steam Line lsolation 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 lsolation Valve PORV Power-Operated Relief Valve PTLR Pressure and Temperature Limits Report QPTR Quadrant Power Tilt Ratio RAOC Relaxed Axial Offset Contro!

RCCA Rod Cluster Control Assembly RCP Reactor Coolant Pump RCS Reactor Coolant System continued Watts Bar - Unit 2

LIST OF ACRONYMS ACRONYM TITLE RHR Residual Heat Removal RTP Rated Thermal Power RTS Reactor Trip System RWST Refueling Water Storage Tank SG Steam Generator Sl Safety lnjection SL Safety Limit SR Surveillance Requirement UHS Ultimate Heat Sink Watts Bar - Unit 2 ix

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PAGE AMENDMENT PAGE AMENDMENT NUMBER NUMBER NUMBER NUMBER 0 B 3.0-6 0 0 B 3.0-7 0 0 B 3.0-8 0 0 B 3.0-9 0 0 B 3.0-10 0 vi 5 B 3.0-10a 7 vii 0 B 3.0-10b 7 viii 0 B 3.0-10c 7 ix 0 B 3.0-11 0 x 18 B 3.0-12 10 xi 16 B 3.0-13 0 xii 18 B 3.0-14 0 xiii 0 B 3.0-15 0 xiv 0 B 3.0-16 0 xiv I B 3.0-17 0 XV 0 B 3.0-18 0 xvi 15 B 3.1-1 0 xvii 9 B 3.1-2 0 xviii 14 B 3.1-3 0 xix 1 B 3.1-4 0 xx 0 B 3.1-5 0 xxi 13 B 3.1-6 0 xxii 18 B 3.1-7 0 B 2.0-1 0 B 3.1-8 0 B 2.0-2 0 B 3.1-9 0 B 2.0-3 0 B 3.1-10 0 B 2.0-4 0 B 3. 1-11 0 B 2.0-5 0 B 3.1-12 0 B 2.0-6 0 B 3.1-13 0 B 2.0-7 0 B 3.1-14 0' B 2.0-8 0 B 3.1-15 0 B 2.0-9 0 B 3.1-16 0 B 2.0-10 0 B 3.1-17 0 B 3.0-1 7 B 3.1-19 0 B 3.0-2 0 B 3.1-19 0 B 3.0-3 0 B 3.1-20 0 B 3.0-4 0 B 3.1-21 0 B 3.0-5 0 B 3.1-22 0 Watts Bar - Unat2 Revision 18

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PAGE AMENDMENT PAGE AMENDMENT NUMBER NUMBER NUMBER NUMBER B 3.1-23 0 B 3.1-61 0 B 3.1-24 0 B 3.1-62 0 B 3.1-25 16 B 3.1-63 0 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 0 B 3.1-31 16 B 3.1-69 0 B 3.1-32 16 B 3. 1-70 0 B 3.1-33 16 B 3.2-1 0 B 3.1-34 0 B 3.2-2 0 B 3.1-35 0 B 3.2-3 0 B 3.1-36 16 B 3.2-4 0 B 3.1-37 16 B 3.2-5 0 B 3.1-38 16 B 3.2-6 0 B 3.1-39 16 B 3.2-7 0 B 3.1-40 16 B 3.2-8 0 B 3.1-41 16 B 3.2-9 0 B 3.1-42 16 B 3.2-10 0 B 3.1-43 16 B 3 .2-11 0 B 3.1-44 16 B 3.2-12 0 B 3.1-45 16 B 3.2-13 0 B 3.1-46 0 B 3.2-14 0 B 3.1-47 16 B 3.2-15 0 B 3.1-48 16 B 3.2-16 0 B 3.1-49 16 B 3.2-17 0 B 3.1-50 16 B 3.2-18 0 B 3.1-51 16 B 3.2-19 0 B 3.1-52 16 B 3.2-20 0 B 3.1-53 16 B 3.2-21 0 B 3.1-54 16 B 3.2-22 0 B 3.1-55 0 B 3.2-23 0 B 3.1-56 16 B 3.2-24 0 B 3.1-57 0 B 3.2-25 0 B 3.1-58 0 B 3.2-26 0 B 3.1-59 0 B 3.2-27 0 B 3.1-60 0 B 3.2-28 0 Watts Bar - Unit2 xt Revision 16

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PAGE AMENDMENT PAGE AMENDMENT NUMBER NUMBER NUMBER NUMBER B 3.2-29 0 B 3.3-36 0 B 3.2-30 17 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-g 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 0 B 3.3-14 0 B 3.3-52 0 B 3.3-15 18 B 3.3-53 0 B 3.3-16 0 B 3.3-54 0 B 3.3-17 0 B 3.3-55 0 B 3.3-18 0 B 3.3-56 0 B 3.3-19 0 B 3.3-57 0 B 3.3-20 0 B 3.3-58 0 B 3.3-21 0 B 3.3-59 0 B 3.3-22 0 B 3.3-60 0 B 3.3-23 0 B 3.3-61 0 B 3.3-24 0 B 3.3-62 0 B 3.3-25 0 B 3.3-63 0 B 3.3-26 0 B 3.3-64 0 B 3.3-27 0 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 B 3.3-32 0 B 3.3-70 0 B 3.3-33 0 B 3,3-71 0 B 3.3-34 0 B 3.3-72 0 B 3.3-35 0 B 3.3-73 0 Watts Bar - Unit 2 xii Revision 18

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PAGE AMENDMENT PAGE AMENDMENT NUMBER NUMBER NUMBER NUMBER B 3.3-74 B 3.3-112 0 B 3.3-75 B 3.3-1 13 0 B 3.3-76 B 3.3-114 0 B 3.3-77 B 3.3-1 15 0 B 3.3-78 B 3.3-1 16 0 B 3.3-79 B 3.3-117 0 B 3.3-80 B 3.3-1 18 -0 B 3.3-81 B 3.3-1 19 0 B 3.3-82 B 3.3-120 0 B 3.3-83 B 3.3-121 0 B 3.3-84 B 3.3-122 0 B 3.3-85 B 3.3-123 0 B 3.3-86 B 3.3-124 0 B 3.3-87 B 3.3-125 0 B 3.3-gg B 3.3-126 0 B 3.3-89 B 3.3-127 0 B 3.3-90 B 3.3-128 0 B 3.3-91 B 3.3-129 0 B 3.3-92 B 3.3-130 0 B 3.3-93 B 3.3-131 0 B 3.3-94 B 3.3-132 0 B 3.3-95 B 3.3-133 0 B 3.3-96 B 3.3-134 0 B 3.3-97 B 3.3-135 0 B 3.3-98 B 3.3-136 0 B 3.3-99 B 3.3-137 0 B 3.3-100 B 3.3-139 0 B 3.3-101 B 3.3-139 0 B 3.3-102 B 3.3-140 0 B 3.3-103 B 3.3-141 0 B 3.3-104 B 3.3-142 0 B 3.3-105 B 3.3-143 0 B 3.3-106 B 3.3-144 0 B 3.3-107 B 3.3-145 0 B 3.3-108 B 3.3-146 0 B 3.3-109 B 3.3-147 0 B 3.3-1 10 B 3.3-148 0 B 3.3-111 Watts Bar - Unit 2 xiii

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PAGE AMENDMENT PAGE AMENDMENT NUMBER NUMBER NUMBER NUMBER B 3.3-149 0 B 3.4-19 0 B 3.3-150 0 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 0 B 3.3-155 0 B 3.4-25 0 B 3.3-156 0 B 3.4-26 0 B 3.3-157 0 B 3.4-27 0 B 3.3-158 0 B 3.4-28 0 B 3.3-159 0 B 3.4-29 0 B 3.3-160 0 B 3.4-30 I B 3.3-161 0 B 3.4-31 0 B 3.3-162 0 B 3.4-32 0 B 3.3-163 0 B 3.4-33 0 B 3.3-164 0 B 3.4-34 0 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 0 B 3.3-169 0 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 0 B 3.4-4 0 B 3.4-42 0 B 3.4-5 0 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 0 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 0 B 3.4-49 0 B 3.4-12 0 B 3.4-50 0 B 3.4-13 0 B 3.4-51 0 B 3.4-14 0 B 3.4-52 0 B 3.4-15 0 B 3.4-53 0 B 3.4-16 0 B 3.4-54 0 B 3.4-17 0 B 3.4-55 0 B 3.4-19 0 Watts Bar - Unit 2 xiv Revision 8

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PAGE AMENDMENT PAGE AMENDMENT NUMBER NUMBER NUMBER NUMBER B 3.4-56 0 B 3.4-93 0 B 3.4-57 0 B 3.4-94 0 B 3.4-58 0 B 3.5-1 0 B 3.4-59 0 B 3.5-2 0 B 3.4-60 0 B 3.5-3 0 B 3.4-61 0 B 3.5-4 0 B 3.4-62 0 B 3.5-5 0 B 3.4-63 0 B 3.5-6 0 B 3.4-64 0 B 3.5-7 0 B 3.4-65 0 B 3.5-8 0 B 3.4-66 0 B 3.5-9 0 B 3.4-67 0 B 3.5-10 0 B 3.4-68 0 B 3.5-11 0 B 3.4-69 0 B 3.5-12 0 B 3.4-70 0 B 3.5-13 0 B 3.4-71 0 B 3.5-14 0 B 3.4-72 0 B 3.5-15 0 B 3.4-72 0 B 3.5-16 0 B 3.4-73 0 B 3.5-17 0 B 3.4-74 0 B 3.5-18 0 B 3.4-75 0 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 3.4-79 0 B 3.5-22 0 B 3.4-79 0 B 3.5-23 0 B 3.4-80 0 B 3.5-24 0 B 3.4-81 0 B 3.5-25 0 B 3.4-82 0 B 3.5-26 0 B 3.4-83 0 B 3.5-27 0 B 3.4-84 0 B 3.5-28 0 B 3.4-85 0 B 3.5-29 0 B 3.4-96 0 B 3.5-30 0 B 3.4-97 0 B 3.5-31 0 B 3.4-89 0 B 3.5-32 0 B 3.4-89 0 B 3.5-33 0 B 3.4-90 0 B 3.6-1 0 B 3.4-91 0 B 3.6-2 0 B 3.4-92 0 Watts Bar - Unit 2 XV

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PAGE AMENDMENT PAGE AMENDMENT NUMBER NUMBER NUMBER NUMBER B 3.6-3 0 B 3.6-41 0 B 3.6-4 0 B 3.6-42 0 B 3.6-5 0 B 3.6-43 0 B 3.6-6 0 B 3.6-44 0 B 3.6-7 0 B 3.6-45 0 B 3.6-8 0 B 3.6-46 0 B 3.6-9 0 B 3.6-47 0 B 3.6-10 0 B 3.6-48 0 B 3.6-1 1 0 B 3.6-49 12 B 3.6-12 0 B 3.6-50 0 B 3.6-13 0 B 3.6-51 12 B 3.6-14 0 B 3.5-52 0 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 0 B 3.6-18 0 B 3.6-56 0 B 3.6-19 0 B 3.6-57 0 B 3.6-20 0 B 3.6-58 0 B 3.6-21 0 B 3.6-59 11 B 3.6-22 0 B 3.6-60 0 B 3.6-23 0 B 3.6-61 0 B 3.6-24 0 B 3.6-62 0 B 3.6-25 0 B 3.6-63 0 B 3.6-26 0 B 3.6-64 11 B 3.6-27 15 B 3.6-65 0 B 3.6-28 0 B 3.6-66 0 B 3.6-29 0 B 3.6-67 0 B 3.6-30 0 B 3.6-68 0 B 3.6-31 0 B 3.6-69 0 B 3.6-32 0 B 3.6-70 0 B 3.6-33 0 B 3.6-71 0 B 3.6-34 0 B 3.6-72 0 B 3.6-35 0 B 3.6-73 0 B 3.6-36 15 B 3.6-74 0 B 3.6-37 0 B 3.6-75 0 B 3.6-38 0 B 3.6-76 0 B 3.6-39 0 B 3.6-77 0 B 3.6-40 0 Watts Bar - Unit 2 xvi Revision 15

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PAGE AMENDMENT PAGE AMENDMENT NUMBER NUMBER NUMBER NUMBER B 3.6-78 0 B 3.7-25 B 3.6-79 0 B 3.7-26 B 3.6-80 0 B 3.7-27 B 3.6-81 0 B 3.7-28 B 3.6-82 0 B 3.7-29 B 3.6-83 0 B 3.7-30 B 3.6-84 0 B 3.7-31 B 3.6-85 0 B 3.7-32 B 3.6-86 0 B 3.7-33 B 3.6-87 4 B 3.7-34 B 3.6-88 4 B 3.7-35 B 3.6-89 4 B 3.7-36 B 3.6-90 0 B 3.7-37 83.6-91 4 B 3.7-38 B 3.7-1 0 B 3.7-39 B 3.7-2 0 B 3.7-40 B 3.7-3 0 B 3.7-41 B 3.7-4 0 B 3.7-42 B 3.7-5 0 B 3.7-43 B 3.7-6 0 B 3.7-44 B 3.7-7 0 B 3.7-45 B 3.7-8 0 B 3.7-46 B 3.7-9 0 B 3.7-47 B 3.7-10 0 B 3.7-48 B 3.7-11 0 B 3.7-49 B 3.7-12 0 B 3.7-50 B 3.7-13 0 B 3.7-51 B 3.7-14 0 B 3.7-52 B 3.7-15 0 B 3.7-53 B 3.7-16 0 B 3.7-54 B 3.7-17 0 B 3.7-55 B 3.7-18 0 B 3.7-56 B 3.7-19 0 B 3.7-57 B 3.7-20 0 B 3.7-58 B 3.7-21 0 B 3.7-59 B 3.7-22 0 B 3.7-60 B 3.7-23 0 B 3.7-61 B 3.7-24 0 Watts Bar - Unit 2 xvil Revision 9

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PAGE AMENDMENT PAGE AMENDMENT NUMBER NUMBER NUMBER NUMBER B 3.7-62 0 B 3.8-10a 5 B 3.7-63 0 B 3.8-1 1 5 B 3.7-64 13 B 3.8-12 5 B 3.7-65 13 B 3.8-12a 5 B 3.7-66 13 B 3.9-12b 5 B 3.7-67 0 B 3.8-13 5 B 3.7-67a 13 B 3.8-14 5 B 3.7-68 0 B 3.8-15 5 B 3.7-69 0 B 3.8-16 5 B 3.7-7A 0 B 3.8-17 0 B 3.7-71 0 B 3.8-18 0 B 3.7-72 0 B 3.8-19 0 B 3.7-73 0 B 3.8-20 0 B 3.7-74 0 B 3.8-21 0 B 3.7-75 0 B 3.8-22 0 B 3.7-76 0 B 3.8-23 0 B 3.7-77 6 B 3.8-24 0 B 3.7-78 6 B 3.8-25 0 B 3.7-79 6 B 3.8-26 0 B 3.7-80 0 B 3.8-27 0 B 3.7-81 0 B 3.8-28 0 B 3.7-82 0 B 3.8-28a 5 B 3.7-83 0 B 3.8-29 0 B 3.7-84 0 B 3.8-30 0 B 3.7-85 0 B 3.9-31 0 B 3.7-86 0 B 3.8-32 0 B 3.7-87 0 B 3.8-33 0 B 3.7-88 0 B 3.8-34 0 B 3.8-1 0 B 3.8-35 0 B 3.8-2 5 B 3.8-36 0 B 3.8-3 0 B 3.8-37 0 B 3.8-4 14 B 3.8-37a 13 B 3.8-5 0 B 3.8-38 0 B 3.9-6 0 B 3.9-39 0 B 3.8-7 0 B 3.8-40 0 B 3.8-8 0 B 3.8-41 0 B 3.8-9 5 B 3.8-42 0 B 3.8-10 5 Watts Bar - Unit 2 xviii Revision 14

TECHN ICAL SPECI FICATIONS BASES LIST OF EFFECTIVE PAGES (continued)

PAGE AMENDMENT PAGE AMENDMENT NUMBER NUMBER NUMBER NUMBER B 3.8-43 0 B 3.8-81 0 B 3.8-44 0 B 3.8-82 0 B 3.8-45 0 B 3.8-83 0 B 3.8-46 0 B 3.8-84 0 B 3.8-47 0 B 3.8-85 0 B 3.8-48 0 B 3.8-86 0 B 3.8-49 0 B 3.8-87 0 B 3.8-50 0 B 3.8-88 0 B 3.8-51 0 B 3.8-89 0 B 3.9-52 0 B 3.8-90 0 B 3.8-53 0 B 3.8-91 0 B 3.8-54 0 B 3.9-92 0 B 3.9-55 0 B 3.8-93 0 B 3.8-56 0 B 3.8-94 1 B 3.8-57 0 B 3.9-95 0 B 3.8-58 0 B 3.8-96 0 B 3.8-59 0 B 3.9-97 0 B 3.8-60 0 B 3.8-98 0 B 3.8-61 0 B 3.9-1 0 B 3.8-62 0 B 3.9-2 0 B 3.8-63 0 B 3.9-3 0 B 3.8-64 0 B 3.9-4 0 B 3.8-65 0 B 3.9-5 0 B 3.8-66 0 B 3.9-6 0 B 3.8-67 0 B 3.9-7 0 B 3.8-69 0 B 3.9-8 0 B 3.8-69 0 B 3.9-10 0 B 3.9-70 0 B 3.9-11 0 B 3.8-71 0 B 3.9-12 0 B 3.8-72 0 B 3.9-13 0 B 3.8-73 0 B 3.9-14 0 B 3.8-74 0 B 3.9-15 0 B 3.8-75 0 B 3.9-16 0 B 3.8-76 0 B 3.9-17 0 B 3.8-77 0 B 3.9-18 0 B 3.8-78 0 B 3.9-19 0 B 3.8-79 0 B 3.9-20 0 B 3.8-80 0 B 3.9-21 0 Watts Bar - Unit 2 xrx Revision 1

TECHNICAL SPECI FICATIONS BASES LIST OF EFFECTIVE PAGES (continued)

PAGE AMENDMENT PAGE AMENDMENT NUMBER NUMBER NUMBER NUMBER B 3.9-22 0 B 3.9-23 0 B 3.9-24 0 B 3.9-25 0 B 3.9-26 0 B 3.9-27 0 B 3.9-29 0 Watts Bar - Unat 2 xx

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 tiOte-of-Contents)

REVISIONS ISSUED SUBJECT NPF.2O 10-22-1s 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 83.3.7, "Component Cooling System (CCS)," regarding the 1B and 28 surge tank sections.

Revision 3 7-11-16 Revise TS Bases 83.6.4, "Containment Pressur," and 83.6.6, "Containment Spray System" regarding the maximum peak containment pressure from a LOCA of 11.73 psig.

Revision 4 8-19-16 Revise TS Bases 83.6.15, "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 83.7-10 CREVS.

Revision 10 7-14-17 Revises TS Bases SR 83.0.2for a one-time extension of the Alternating Current Sources.

Revision 11, Amendment 14 9-29-17 Revises TS Bases 83.6.11 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 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months per month.

Revision 13, Amendment 16 11-2-17 Revises TS Bases 83.7-12to provide action when both trains of ABGTS are inoperable. Also, 83.8-37a correction of unit error.

Watts Bar - Unit 2 xxt Revision 13

TECHNICAL SPECIFICATION BASES . REVISION LISTING (Thls llstlng ls an admlnlstratlve tool maintaaned by WBt{ Llcenstng and may be updated wlthout formally revlslng the Technlcal Speclflcatlon Bases Table-of-Gontents)

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 83.6.4 and B 3.6.G to change the calculated peak pressure.

Revision 16, Amendment 20 08-20-1 I Revises TS Bases B3.1 .5, B3.1 .6, 83. 1.7 ,

and B3.1.8 which adopts the TSTF-547 ,

Clarification of Rod. position requirements.

Revision 17, Amendment 21 09-21-1 8 Revises TS Bases 3.2.4 and Bases 3.3.1 related to the reactor trip system instrumentation.

Revision 18 02-13-1e

[::,'""i,irTm#J,fr1L",1,:i:^" I Watts Bar - Unat 2 xxil Revision 18

ENCLOSURE 6 WBN UNIT 2 TECHNICAL SPECIFICATION BASES CHANGED PAGES

Rod Group Alignment Limits B 3.1 .5 B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.5 Rod Group Alignment Limits BASES BACKGROUND The OPERABILITY (e.9., trippability) of the shutdown and control rods is an initial assumption in all safety analyses that assume rod insertion upon reactor trip. Maximum rod misalignment is an initial assumption in the safety analysis that directly affects core power distributions and assumptions of available SDM.

The applicable criteria for these reactivity and power distribution design requirements are 10 CFR 50, Appendix A, GDC 10, "Reactor Design,"

and GDC 26, "Reactivity Control System Redundancy and Capability,"

(Ref. 1), and 10 CFR 50.46, "Acceptance Criteria for Emergency Core Cooling Systems for Light Water Nuclear Power Reactors" (Ref. 2).

Mechanical or electricalfailures may cause a control or shutdown rod to become inoperable or to become misaligned from its group. Rod inoperability or misalignment may cause increased power peaking, due to the asymmetric reactivity distribution and a reduction in the total available rod worth for reactor shutdown. Therefore, rod alignment and OPERABILITY are related to core operation in design power peaking limits and the core design requirement of a minimum SDM.

Limits on rod alignment have been established, and all rod positions are monitored and controlled during power operation to ensure that the power distribution and reactivity limits defined by the design power peaking and SDM limits are preserved.

Rod cluster control assemblies (RCCAS), or rods, are moved by their control rod drive mechanisms (CRDMS). Each CRDM moves its RCCA one step (approximately 5/8 inch) at a time, but at varying rates (steps per minute) depending on the signal output from the Rod Control System.

The RCCAs are divided among control banks and shutdown banks. Each bank may be further subdivided into two groups to provide for precise reactivity control (Shutdown Banks C and D have only one group each).

A group consists of two or more RCCAs that are electrically paralleled to step simultaneously. Except for Shutdown Banks C and D, (continued)

Watts Bar - Unit 2 B 3.1-25 Revision 16 Amendment 20

Rod Group Alignment Limits B 3.1 .5 BASES APPLICABLE Continued operation of the reactor with a misaligned control rod is SAFETY allowed if the heat flux hot channel factor (Fo(Z)) and the nuclear enthalpy ANALYSES hot channelfactor (FN*) are verified to be within their limits in the COLR (continued) and the safety analysis is verified to remain valid. When a control rod is misaligned, the assumptions that are used to determine the rod insertion limits, AFD limits, and quadrant power tilt limits are not preserved.

Therefore, the limits may not preserve the design peaking factors, and Fq(Z) and FNal ffiUSt be verified directly using incore power distribution measurements. Bases Section 3.2 (Power Distribution Limits) contains more complete discussions of the relation of Fq(Z) and FN66 to the operating limits.

Shutdown and control rod OPERABILITY and alignment are directly related to power distributions and SDM, which are initial conditions assumed in safety analyses. Therefore they satisfy Criterion 2 of 10 CFR 50.36(c)(2Xii).

LCO The limits on shutdown or control rod alignments ensure that the assumptions in the safety analysis will remain valid. The requirements on OPERABILITY ensure that upon reactor trip, the assumed reactivity will be available and will be inserted. The control rod OPERABILITY requirements (i.e., trippability) are separate from the alignment requirements, which ensure that the RCCAs and banks maintain the correct power distribution and rod alignment. The rod OPERABILITY requirement is satisfied provided the rod willfully insert in the required rod drop time assumed in the safety analysis. Rod control malfunctions that result in the inability to move a rod (e.9., rod lift coilfailures), but that do not impact trippability, do not result in rod inoperability.

The requirement to maintain the rod alignment to within plus or minus 12 steps is conservative. The minimum misalignment assumed in safety analysis is 24 steps (15 inches), and in some cases a total misalignment from fully withdrawn to fully inserted is assumed.

Failure to meet the requirements of this LCO may produce unacceptable power peaking factors and LHRs, or unacceptable SDMs, all of which may constitute initial conditions inconsistent with the safety analysis.

Watts Bar - Unit 2 B 3. 1-28 (continued)

Revision 16 Amendment 20

Rod Group Alignment Limits B 3.1 .5 BASES (continued)

APPLICABILITY The requirements on RCCA OPERABILITY and alignment are applicable in MODES 1 and2 because these are the only MODES in which neutron (or fission) power is generated, and the OPEMBILITY (i.e., trippability) and alignment of rods have the potentialto affect the safety of the plant.

ln MODES 3, 4, 5, and 6, the alignment limits do not apply because the control rods are bottomed and the reactor is shut down and not producing fission power. ln the shutdown MODES, the OPERABILITY of the shutdown and control rods has the potential to affect the required SDM, but this effect can be compensated for by an increase in the boron concentration of the RCS. See LCO 3.1.1, "SHUTDOWN MARGIN (SDM) - T"un > 200'F," for SDM in MODES 3 and 4, LCO 3.1.2, "Shutdown Margin (SDM) - T",s s 200"F" for SDM in MODE 5, and LCO 3.9.1, "Boron Concentration," for boron concentration requirements during refueling.

ACTIONS A.1.1 and A.1.2 When one or more rods are inoperable (i.e., untrippable), there is a possibility that the required SDM may be adversely affected. Under these conditions, it is important to determine the SDM, and if it is less than the required value, initiate boration untilthe required SDM is recovered. The Completion Time of t hour is adequate for determining SDM and, if necessary, for initiating boration to restore SDM.

ln this situation, SDM verification must include the worth of the untrippable rod, as well as a rod of maximum worth.

4.2 lf the inoperable rod(s) cannot be restored to OPERABLE status, the plant must be brought to a MODE or condition in which the LCO requirements are not applicable. To achieve this status, the unit must be brought to at least MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

The allowed Completion Time is reasonable, based on operating experience, for reaching MODE 3 from full power conditions in an orderly manner and without challenging plant systems.

(continued)

Watts Bar - Unat2 B 3.1-29 Revision 16 Amendment 20

Rod Group Alignment Limits B 3.1 .5 BASES ACTIONS B.1.1 and B.1.2 (continued)

When a rod becomes misaligned, it can usually be moved and is still trippable.

An alternative to realigning a single misaligned RCCA to the group average position is to align the remainder of the group to the position of the misaligned RCCA. However, this must be done without violating the bank sequence, overlap, and insertion limits specified in LCO 3.1.6, "Shutdown Bank lnsertion Limits," and LCO 3.1.7, "Control Bank lnsertion Limits."

ln many cases, realigning the remainder of the group to the misaligned rod may not be desirable. For example, realigning control bank B to a rod that is misaligned 15 steps from the top of the core would require a significant power reduction, since control bank D must be moved fully in and control bank C must be moved in to approximately 100 to 115 steps.

Power operation may continue with one RCCA trippable but misaligned, provided that SDM is verified within t hour. The Completion Time of t hour represents the time necessary for determining the actual unit SDM and, if necessary, aligning and starting the necessary systems and components to initiate boration.

continued Watts Bar - Un,t2 B 3.1-30 Revision 16 Amendment 20

Rod Group Alignment Limits B 3.1 .5 BASES ACTIONS B.2.8.3. B.4. and B.5 (continued)

For continued operation with a misaligned rod, RTP must be reduced, SDM must periodically be verified within limits, hot channelfactors (Fo(Z) and FNo6) must be verified within limits, and the safety analyses must be re-evaluated to confirm continued operation is permissible.

Reduction of power to 75o/o RTP ensures that local LHR increases due to a misaligned RCCA will not cause the core design criteria to be exceeded (Ref. 6). The Completion Time of 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months gives the operator sufficient time to accomplish an orderly power reduction without challenging the Reactor Protection System.

When a rod is known to be misaligned, there is a potentialto impact the SDM. Since the core conditions can change with time, periodic verification of SDM is required. A Frequency of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is sufficient to ensure this requirement continues to be met.

Verifying that Fq(Z), as approximated by Fco(Z) and Fwqlz; and FN66 are within the required limits ensures that current operation at7lo/o RTp with a rod misaligned is not resulting in power distributions that may invalidate safety analysis assumptions at full power. The Completion Time of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months allows sufficient time to obtain an incore power distribution measurement and to calculate Fq(Z) and FNa1.

Once current conditions have been verified acceptable, time is available to perform evaluations of accident analysis to determine that core limits will not be exceeded during a Design Basis Event for the duration of operation under these conditions. The accident analyses presented in UFSAR Chapter 15 (Ref. 3) that may be adversely affected wifl be evaluated to ensure that the analyses remain valid for the duration of continued operation under these conditions. A Completion Time of 5 days is sufficient time to obtain the required input data and to perform the analysis.

c.1 When Required Actions cannot be completed within their Completion Time, the unit must be brought to a MODE or Condition in which the LCO requirements are not applicable. To achieve this status, the unit must be brought to at least MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , which obviates concerns about the development of undesirable xenon or power distributions. The allowed Completion Time of 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months is reasonable, based on operating experience, for reaching MODE 3 from full power conditions in an orderly manner and without challenging the plant systems.

continued B 3.1-31 Revision 16 Amendment 20

Rod Group Alignment Limits B 3.1 .5 BASES ACTIONS D.'1.1 and D.1.2 (continued)

More than one control rod becoming misaligned from its group average position is not expected, and has the potentialto reduce SDM. Therefore, SDM must be evaluated. One hour allows the operator adequate time to determine SDM. Restoration of the required SDM, if necessary, requires increasing the RCS boron concentration to provide negative reactivity, as described in the Bases of LCO 3.1.1. The required Completion Time of t hour for initiating boration is reasonable, based on the time required for potentialxenon redistribution, the low probability of an accident occurring, and the steps required to complete the action. This allows the operator sufficient time to align the required valves and start the boric acid pumps.

Boration will continue untilthe required SDM is restored.

D.2 lf more than one rod is found to be misaligned or becomes misaligned because of bank movement, the unit conditions fall outside of the accident analysis assumptions. Since automatic bank sequencing would continue to cause misalignment, the unit must be brought to a MODE or Condition in which the LCO requirements are not applicable.

To achieve this status, the unit must be brought to at least MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . The allowed Completion Time is reasonable, based on operating experience, for reaching MODE 3 from full power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE sR 3.1.5.1 REQUIREMENTS verification that the position of individual rods is within alignment limits at a Frequency of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months provides a history that allows the operator to detect a rod that is beginning to deviate from its expected position. lf the rod position deviation monitor is inoperable, a Frequency of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months accomplishes the same goal. The specified Frequency takes into account other rod position information that is continuously available to the operator in the control room, so that during actual rod motion, deviations can immediately be detected.

The sR is modified by a NorE that permits it to not be performed for rods associated with an inoperable demand position indicator or an inoperable rod position indicator. The alignment limit is based on the demand position indicator which is not available if the indicator is inoperable. LCo 3.1.8, "Rod Position lndication," provides Actions to verify the rods are in alignment when one or more rod position indicators are inoperable.

continued B 3.1-32 Revision 16 Amendment 20

Rod Group Alignment Limits B 3.1 .5 BASES SURVEILLANCE SR 3.1.5.1 (continued)

REQUIREMENTS (continued) The Surveillance is modified by a NOTE which states that the SR is not required to be performed until t hour after associated rod motion. Control rod temperature affects the accuracy of the rod position indication system. Due to changes in the magnetic permeability of the drive shaft as a function of temperature, the indicated position is expected to change with time as the drive shaft temperature changes. The one hour period allows control rod temperature to stabilize following rod movement in order to ensure the indicated rod position is accurate.

sR 3.1.5.2 Verifying each control rod is OPERABLE would require that each rod be tripped. However, in MODES 1 and2, tripping each control rod would result in radial or axial power tilts, or oscillations. Exercising each individual control rod every 92 days provides increased confidence that all rods continue to be OPERABLE without exceeding the alignment limit, even if they are not regularly tripped. Moving each control rod by 10 steps will not cause radial or axial power tilts, or oscillations, to occur.

The 92 day Frequency takes into consideration other information available to the operator in the control room and SR 3.1.5.1, which is performed more frequently and adds to the determination of OPERABILITY of the rods. Between required performances of SR 3.1.5.2 (determination of controlrod OPERABILIry by movement), if a control rod(s) is discovered to be immovable, but remains trippable and aligned, the control rod(s) is considered to be OPERABLE. At any time, if a control rod(s) is immovable, a determination of the trippability (OPERABILITY) of the control rod(s) must be made, and appropriate action taken.

sR 3.1.5.3 Verification of rod drop times allows the operator to determine that the maximum rod drop time permitted is consistent with the assumed rod drop time used in the safety analysis. Measuring rod drop times prior to reactor criticality after each reactor vessel head removal ensures that the reactor internals and rod drive mechanism will not interfere with rod motion or rod drop time, and that no degradation in these systems has occurred that would adversely affect control rod motion or drop time. This testing is performed with all RCPs operating and the average moderator temperature 2 551'F to simulate a reactor trip under actual conditions.

This Surveillance is performed during a plant outage, due to the plant conditions needed to perform the SR and the potentialfor an unplanned plant transient if the Surveillance were performed with the reactor at power.

Watts Bar - Unit 2 B 3.1-33 (continued)

Revision 16 Amendment 20

Shutdown Bank lnsertion Limits B 3.1 .6 BASES BACKGROUND Hence, they are not capable of adding a large amount of positive (continued) reactivity. Boration or dilution of the Reactor Coolant System (RCS) compensates for the reactivity changes associated with large changes in RCS temperature. The design calculations are performed with the assumption that the shutdown banks are withdrawn first. The shutdown banks are controlled manually by the control room operator. During normal unit operation, the shutdown banks are either fully withdrawn or fully inserted. The shutdown banks must be completely withdrawn from the core, prior to withdrawing any control banks during an approach to criticality. The shutdown banks can be fully withdrawn without the core going critical. This provides available negative reactivity in the event of boration errors. The shutdown banks are then left in this position untilthe reactor is shut down. They add negative reactivity to shut down the reactor upon receipt of a reactor trip signal.

APPLICABLE On a reactor trip, all RCCAs (shutdown banks and control banks), except SAFETY the most reactive RCCA, are assumed to insert into the core. The ANALYSES shutdown banks shall be at or above their insertion limits and available to insert the maximum amount of negative reactivity on a reactor trip signal.

The control banks may be partially inserted in the core, as allowed by LCO 3.1.7, "Control Bank lnsertion Limits." The shutdown bank and control bank insertion limits are established to ensure that a sufficient amount of negative reactivity is available to shut down the reactor and maintain the required SDM (see LCO 3.1.1, "SHUTDOWN MARGIN (SDM) - T",e > 200"F," and LCO 3.1.2, "SHUTDOWN MARGIN (SDM) -

T"w < 200"F') following a reactor trip from full power. The combination of control banks and shutdown banks (less the most reactive RCCA, which is assumed to be fully withdrawn) is sufficient to take the reactor from full power conditions at rated temperature to zero power, and to maintain the required SDM at rated no load temperature (Ref. 3). The shutdown bank insertion limit also limits the reactivity worth of an ejected shutdown rod.

The acceptance criteria for addressing shutdown and control rod bank insertion limits and inoperability or misalignment is that:

a. There be no violations of:

1. Specified acceptable fuel design limits, or

2. RCS pressure boundary integrity; and

b. The core remains subcritical after accident transients other than a main steam line break (MSLB).

(continued)

Watts Bar - Un',t 2 B 3.1-36 Revision 16 Amendment 20

Shutdown Bank lnsertion Limits B 3.1 .6 BASES APPLICABLE As such, the shutdown bank insertion limits affect safety analysis SAFETY involving core reactivity and SDM (Ref. 3).

ANALYSES (continued) The shutdown bank insertion limits preserve an initialcondition assumed in the safety analyses and, as such, satisfo Criterion 2 of 10 CFR 50.36(cX2Xii).

LCO The shutdown banks must be within their insertion limits any time the reactor is critical or approaching criticality. This ensures that a sufficient amount of negative reactivity is available to shut down the reactor and maintain the required SDM following a reactor trip.

The shutdown bank insertion limits are defined in the COLR.

The LCO is modified by a Note indicating the LCO requirement is not applicable to shutdown banks being inserted while performing SR 3.1.5.2.

This SR verifies the freedom of the rods to move, and may require the shutdown bank to move below the LCO limits, which would normally violate the LCO. This Note applies to each shutdown bank as it is moved below the insertion limit to perform the SR. This Note is not applicable should a malfunction stop performance of the SR.

APPLICABILITY The shutdown banks must be within their insertion limits, with the reactor in MODES 1 and 2. This ensures that a sufficient amount of negative reactivity is available to shut down the reactor and maintain the required SDM following a reactor trip. The shutdown banks do not have to be within their insertion limits in MODE 3, unless an approach to criticality is being made. Refer to LCO 3.1.1 and LCO 3.1.2 for SDM requirements in MODES 3, 4, and 5. LCO 3.9.1, "Boron Concentration," ensures adequate SDM in MODE 6.

Watts Bar - Unit 2 B 3.1-37 (continued)

Revision 16 Amendment 20

Shutdown Bank lnsertion Limits B 3.1 .6 BASES ACTIONS A.1. A.2.1. A.2.2 and A.3 lf one shutdown bank is inserted less than or equal to 10 steps below the insertion limit,24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is allowed to restore the shutdown bank to within the limit. This is necessary because the available SDM may be reduced with a shutdown bank not within its insertion limit. Also, verification of SDM or initiation of boration within t hour is required, since the SDM in MODES 1 and2 is ensured by adhering to the control and shutdown bank insertion limits (see LCO 3.1.1). lf a shutdown bank is not within its insertion limit, SDM will be verified by performing a reactivity balance calculation, considering the effects listed in the BASES for SR 9.1.1.1.

while the shutdown bank is outside the insertion limit, all controt banks must be within their insertion limits to ensure sufficient shutdown margin is available. The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months completion Time is sufficient to repair most rod control failures that would prevent movement of a shutdown bank.

8.1 .1 . 8.1.2 and 8.2 When one or more shutdown banks is not within insertion limits for reasons other than Condition A, 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months is allowed to restore the shutdown banks to within the insertion limits. This is necessary because the available SDM may be significantly reduced, with one or more of the shutdown banks not within their insertion limits. Also, verification of sDM or initiation of boration within t hour is required, since the SDM in MODES 1 and2 is ensured by adhering to the controland shutdown bank insertion limits (see Lco 3.1.1.). lf shutdown banks are not within their insertion limits, then SDM will be verified by performing a reactivity balance calculation, considering the effects listed in the Bases for sR 3.1.1.1.

The allowed completion Time of 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months provides an acceptable time for evaluating and repairing minor problems without allowing the plant to remain in an unacceptable condition for an extended period of time.

c.1 lf the Required Actions and associated Completion Times are not met, the unit must be brought to a MODE where the LCO is not applicable.

The allowed Completion Time of 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months is reasonable, based on operating experience, for reaching the required MODE from full power conditions in an orderly manner and without challenging plant systems.

Watts Bar - Unat2 B 3.1-38 Revision 16 Amendment 20

Shutdown Bank lnsertion Limits B 3.1 .6 BASES SURVEILLANCE sR 3.1.6.1 REQUIREMENTS Verification that the shutdown banks are within their insertion limits prior to an approach to criticality ensures that when the reactor is critical, or being taken critical, the shutdown banks will be available to shut down the reactor, and the required SDM will be maintained following a reactor trip.

This SR and Frequency ensure that the shutdown banks are withdrawn before the contro! banks are withdrawn during a unit startup.

The Surveillance is modified by a Note which states that the SR is not required to be performed for shutdown banks until t hour after motion of rods in those banks. Rod temperature affects the accuracy of the rod position indication system. Due to changes in the magnetic permeability of the drive shaft as a function of temperature, the indicated position is expected to change with time as the drive shaft temperature changes.

The one hour period allows rod temperature to stabilize following rod movement in order to ensure the indicated position is accurate.

Since the shutdown banks are positioned manually by the control room operator, a verification of shutdown bank position at a Frequency of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , after the reactor is taken critical, is adequate to ensure that they are within their insertion limits. Also, the 12-hour Frequency takes into account other information available in the control room for the purpose of monitoring the status of shutdown rods.

REFERENCES 1. Title 10, Code of Federal Regulations, Part 50, Appendix A, General Design Criterion 10, "Reactor Design," General Design Criterion 26, "Reactivity Contro! System Redundancy and Capability," and General Design Criterion 2S, "Reactivity Limits."

2. Title 10, Code of Federal Regulations, Part 50.46, "Acceptance Criteria for Emergency Core Cooling Systems for Light Water Nuclear Power Reactors."

3. Watts Bar FSAR, Section 15.0, "Accident Analyses."

Watts Bar - Unit 2 B 3.1-39 Revision 16 Amendment 20

Control Bank lnsertion Limits B3.1 .7 B 3.1 REACTIVITY CONTROL SYSTEMS B 3 .1 .7 Control Bank lnsertion Limits BASES BACKGROUND The insertion limits of the shutdown and control rods are initial assumptions in all safety analyses that assume rod insertion upon reactor trip. The insertion limits directly affect core power and fuel burnup distributions and assumptions of available ejected rod worth, SDM, and initial reactivity insertion rate.

The applicable criteria for these reactivity and power distribution design requirements are 10 CFR 50, Appendix A, GDC 10, "Reactor Design,"

GDC 26, "Reactivity Control System Redundancy and Capability," and GDC 28, "Reactivity Limits" (Ref. 1), and 10 CFR 50.46, 'Acceptance Criteria for Emergency Core Cooling Systems for Light Water Nuclear Power Reactors" (Ref. 2). Limits on control rod insertion have been established, and all rod positions are monitored and controlled during power operation to ensure that the power distribution and reactivity limits defined by the design power peaking and SDM limits are preserved.

The rod cluster control assemblies (RCCAs) are divided among control banks and shutdown banks. Each bank may be further subdivided into two groups to provide for precise reactivity control (Shutdown Banks C and D have only one group each). A group consists of two or more RCCAs that are electrically paralleled to step simultaneously. Except for Shutdown Banks C and D, a bank of RCCAs consists of two groups that are moved in a staggered fashion, but always within one step of each other. There are four control banks and four shutdown banks. See LCO 3.1.5, "Rod Group Alignment Limits," for control and shutdown rod OPERABILITY and alignment requirements, and LCO 3.1.8, "Rod Position lndication," for position indication requirements.

The control bank insertion limits are specified in the COLR. An example is provided for information only in Figure B 3.1.7-1. The control banks are required to be at or above the insertion limit lines.

Figure B 3.1.7-1 also indicates how the control banks are moved in an overlap pattern. overlap is the distance traveled together by two control banks. The predetermined position of control bank C, at which control bank D will begin to move with bank C on a withdrawal, as an example may be at '116 steps. Therefore, in this example, control bank C overlaps contro! bank D from 116 steps to the fully withdrawn position for control bank C. The fully withdrawn position and predetermined overlap positions are defined in the COLR.

(continued)

Watts Bar - Unit 2 B 3.1-40 Revision 16 Amendment 20

Control Bank lnsertion Limits B 3.1 .7 BASES BACKGROUND The control banks are used for precise reactivity control of the reactor.

(continued) The positions of the control banks are normally controlled automatically by the Rod Control System, but can also be manually controlled. They are capable of adding reactivity very quickly (compared to borating or diluting).

The power density at any point in the core must be limited, so that the fuel design criteria are maintained. Together, LCO 3.'1.5, "Rod Group Alignment Limits," LCO 3.1.6, "Shutdown Bank lnsertion Limits,"

LCO 3.1.7, "Control Bank lnsertion Limits," LCO 3.2.3,'MIAL FLUX DIFFERENCE (AFD)," and LCO 3.2.4, "QUADRANT POWER TtLT RATIO (QPTR),'provide limits on control component operation and on monitored process variables, which ensure that the core operates within the fuel design criteria.

The shutdown and control bank insertion and alignment limits, AFD, and QPTR are process variables that together characterize and control the three dimensional power distribution of the reactor core. Additionally, the control bank insertion limits controlthe reactivity that could be added in the event of a rod ejection accident, and the shutdown and control bank insertion limits ensure the required SDM is maintained.

Operation within the subject LCO limits will prevent fuel cladding failures that would breach the primary fission product barrier and release fission products to the reactor coolant in the event of a loss of coolant accident (LOCA), loss of flow, ejected rod, or other accident requiring termination by a Reactor Trip Svstem (RTS) trip function.

APPLICABLE The shutdown and control bank insertion limits, AFD, and QPTR LCOs SAFETY are required to prevent power distributions that could result in fuel ANALYSES cladding failures in the event of a LOCA, loss of flow, ejected rod, or other accident requiring termination by an RTS trip function.

The acceptance criteria for addressing shutdown and control bank insertion limits and inoperability or misalignment are that:

a. There be no violations ot

1. Specified acceptable fuel design limits, or

2. Reactor Coolant System pressure boundary integrity; and

b. The core remains subcritical after accident transients other than a main steam line break (MSLB).

As such, the shutdown and control bank insertion limits affect safety analysis involving core reactivity and power distributions (Ref. 3 through 13).

continued Watts Bar - Unit 2 B 3.1-41 Revision 16 Amendment 20

Control Bank lnsertion Limits 83.1 .7 BASES APPLICABLE The SDM requirement is ensured by limiting the control and shutdown SAFETY bank insertion limits so that allowable inserted worth of the RCCAs is ANALYSES such that sufficient reactivity is available in the rods to shut down the (continued) reactor to hot zero power with a reactivity margin that assumes the maximum worth RCCA remains fully withdrawn upon trip (Ref. 5, 6, 8 and 11).

Operation at the insertion limits or AFD limits may approach the maximum allowable linear heat generation rate or peaking factor with the allowed QPTR present. Operation at the insertion limit may also indicate the maximum ejected RCCA worth could be equalto the limiting value in fuel cycles that have sufficiently high ejected RCCA worths.

The control and shutdown bank insertion limits ensure that safety analyses assumptions for SDM, ejected rod worth, and power distribution peaking factors are preserved (Ref. 3 through 13).

The insertion limits satisff Criterion 2 of 10 CFR 50.36(cX2)(ii), in that they are initial conditions assumed in the safety analysis.

LCO The limits on control banks sequence, overlap, and physical insertion, as defined in the COLR, must be maintained because they serve the function of preserving power distribution, ensuring that the SDM is maintained, ensuring that ejected rod worth is maintained, and ensuring adequate negative reactivity insertion is available on trip. The overlap between control banks provides more uniform rates of reactivity insertion and withdrawaland is imposed to maintain acceptable power peaking during control bank motion.

The LCO is modified by a Note indicating the LCO requirement is not applicable to control banks being inserted while performing SR 3.1.5.2.

This SR verifies the freedom of the rods to move, and may require the control bank to move below the LCO limits, which would normally violate the LCO. This Note applies to each control bank as it is moved below the insertion limit to perform the SR. This Note is not applicable should a malfunction stop performance of the SR.

APPLICABILITY The control bank sequence, overlap, and physical insertion limits shall be maintained with the reactor in MODES 1 and2 with k"r 2 1.0. These limits must be maintained, since they preserve the assumed power distribution, ejected rod worth, SDM, and reactivity rate insertion assumptions. Applicability in MODES 3, 4, and 5 is not required, since neither the power distribution nor ejected rod worth assumptions would be exceeded in these MODES.

(continued)

Watts Bar - Unit 2 B 3.1-42 Revision 16 Amendment 20

Control Bank lnsertion Limits 83.1 .7 BASES ACTIONS 4.1. 4.2.1. 4.2.2. and 4.3 lf Control Bank A, B, or C is inserted less than or equalto 10 steps below the insertion, sequence, or overlap limits, 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is allowed to restore the control bank to within the limits. Verification of SDM or initiation of boration within t hour is required, since the SDM in MODES 1 and 2 is ensured by adhering to the control and shutdown bank insertion limits (See LCO 3.1.1). lf a control bank is not within its insertion limit, SDM will be verified by performing a reactivity balance calculation, considering the effects listed in the BASES for SR 3.1.1.1.

While the control bank is outside the insertion, sequence, or overlap limits, all shutdown banks must be within their insertion limits to ensure sufficient shutdown margin is available and that power distribution is controlled. The24 hour Completion Time is sufficient to repair most rod control failures that would prevent movement of a shutdown bank.

Condition A is limited to Control Banks A, B, or C. The allowance is not required for Control Bank D because the full power bank insertion limit can be met during performance of the SR 3.1.5.2 controlrod freedom of movement (trippability) testing.

8.1.1.8.1.2.8.2. C1.1. C.l.2 and C.2 When the control banks are outside the acceptable insertion limits for reasons other than Condition A, they must be restored to within those limits. This restoration can occur in two ways:

a. Reducing power to be consistent with rod position; or

b. Moving rods to be consistent with power.

Also, verification of SDM or initiation of boration to regain SDM is required within '1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , since the SDM in MODES I and 2 normally ensured by adhering to the control and shutdown bank insertion limits (see LCO 3.1.'1, "SHUTDOWN MARGIN (SDM) - T",n , 200"F") has been upset. lf control banks are not within their insertion limits, then SDM will be verified by performing a reactivity balance calculation, considering the effects listed in the Bases for SR 3.1.1 .1.

Similarly, if the control banks are found to be out of sequence or in the wrong overlap configuration for reasons other than Condition A, they must be restored to meet the limits.

(continued)

Watts Bar - Unit 2 B 3.1-43 Revision 16 Amendment 20

Control Bank lnsertion Limits 83.1 .7 BASES ACTIONS 8.1.1.8.1.2. 8.2. C.1.1. C.l.2 and C.2 (continued)

Operation beyond the LCO limits is allowed for a short time period in order to take conservative action because the simultaneous occurrence of either a LOCA, loss of flow accident, ejected rod accident, or other accident during this short time period, together with an inadequate power distribution or reactivity capability, has an acceptably low probability.

The allowed Completion Time of 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for restoring the banks to within the insertion, sequence, and overlap limits provides an acceptable time for evaluating and repairing minor problems without allowing the plant to remain in an unacceptable condition for an extended period of time.

D.1 lf the Required Actions cannot be completed within the associated Completion Times, the plant must be brought to MODE 2 with k"n < 1.0, where the LCO is not applicable. The allowed Completion Time of 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months is reasonable, based on operating experience, for reaching the required MODE from full power conditions in an orderly manner and without challenginq plant svstems.

SURVEILLANCE sR 3.1.7.1 REQUIREMENTS This Surveillance is required to ensure that the reactor does not achieve criticality with the control banks below their insertion limits.

The estimated critical position (ECP) depends upon a number of factors, one of which is xenon concentration. lf the ECP was calculated long before criticality, xenon concentration could change to make the ECp substantially in error. Conversely, determining the ECP immediately before criticality could be an unnecessary burden. There are a number of unit parameters requiring operator attention at that point. Performing the ECP calculation within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months prior to criticality avoids a large error from changes in xenon concentration, but allows the operator some flexibility to schedule the ECP calculation with other startup activities.

sR 3.1.7.2 with an OPERABLE bank insertion limit monitor, verification of the control bank insertion limits at a Frequency of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is sufficient to ensure OPERABILITY of the bank insertion limit monitor and to detect control banks that may be approaching the insertion limits since, normally, very little rod motion occurs in 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . lf the insertion limit monitor becomes inoperable, verification of the control bank position at a Frequency of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months is sufficient to detect control banks that may be approaching the insertion limits.

(continued)

Watts Bar - Unat 2 B 3.1-44 Revision 16 Amendment 20

Control Bank lnsertion Limits 83.1 .7 BASES SURVEILLANCE The Surveillance is modified by a Note stating that the SR is not required REQUIREMENTS to be performed for control banks until t hour after motion of rods in those (continued) banks. Control rod temperature affects the accuracy of the rod position indication system. Due to changes in the magnetic permeability of the drive shaft as a function of temperature, the indicated position is expected to change with time as the drive shaft temperature changes. The one hour period allows control rod temperature to stabilize following rod movement in order to ensure the indicated rod position is accurate.

sR 3.1.7.3 When control banks are maintained within their insertion limits as checked by SR 3.1.7.2 above, it is unlikely that their sequence and overlap will not be in accordance with requirements provided in the COLR.

The Surveillance is modified by a Note stating that the SR is not required to be performed for control banks until t hour after motion of rods in those banks. Control rod temperature affects the accuracy of the rod position indication system. Due to changes in the magnetic permeability of the drive shaft as a function of temperature, the indicated position is expected to change with time as the drive shaft temperature changes. The one hour period allows control rod temperature to stabilize following rod movement in order to ensure the indicated rod position is accurate.

A Frequency of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is consistent with the insertion limit check above in SR 3.1.7.2 (continued)

Watts Bar - Unit 2 B 3.1-45 Revision 16 Amendment 20

Control Bank lnsertion Limits 83.1 .7 BASES OE o

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E, FRACTION OF RATED THERMAT POWER Figure B 3. 1.7-1 CONTROL BANK INSERTION VS. RTP Watts Bar - Unl,t 2 B 3.1-47 Revision 16 Amendment20

Rod Position lndication B 3.1 .8 B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.8 Rod Position Indication BASES BACKGROUND According to GDC 13 (Ref. 1), instrumentation to monitor variables and systems over their operating ranges during normal operation, anticipated operational occurrences, and accident conditions must be OPERABLE.

LCO 3.1.8 is required to ensure OPERABILITY of the control rod position indicators to determine control rod positions and thereby ensure compliance with the control rod alignment and insertion limits.

The OPERABILITY, including position indication, of the shutdown and control rods is an initial assumption in all safety analyses that assume rod insertion upon reactor trip. Maximum rod misalignment is an initial assumption in the safety analysis that directly affects core power distributions and assumptions of available SDM. Rod position indication is required to assess OPERABILITY and misalignment.

Mechanical or electricalfailures may cause a control rod to become inoperable or to become misaligned from its group. Control rod inoperability or misalignment may cause increased power peaking, due to the asymmetric reactivity distribution and a reduction in the total available rod worth for reactor shutdown. Therefore, control rod alignment and OPERABILITY are related to core operation in design power peaking limits and the core design requirement of a minimum SDM.

Limits on control rod alignment and OPERABILITY have been established, and all rod positions are monitored and controlled during power operation to ensure that the power distribution and reactivity limits defined by the design power peaking and SDM limits are preserved.

Rod cluster control assemblies (RCCAs), or rods, are moved out of the core (up or withdrawn) or into the core (down or inserted) by their control rod drive mechanisms. The RCCAs are divided among control banks and shutdown banks. Each bank may be further subdivided into two groups to provide for precise reactivity control(Shutdown Banks C and D have only one group each).

The axial position of shutdown rods and control rods are determined by two separate and independent systems: the Bank Demand Position lndication System (commonly called group step counters) and the Rod Position lndication (RPl) System.

(continued)

Watts Bar - Unit 2 B 3.1-48 Revision 16 Amendment 20

Rod Position lndication B 3.1 .8 BASES BACKGROUND The Bank Demand Position lndication System counts the pulses from the (continued) Rod Control System that move the rods. There is one step counter for each group of rods. lndividual rods in a group all receive the same signal to move and should, therefore, all be at the same position indicated by the group step counter for that group. The Bank Demand Position lndication System is considered highly precise (t 1 step or t 5/8 inch). lf a rod does not move one step for each demand pulse, the step counter will still count the pulse and incorrectly reflect the position of the rod.

The RPI System provides an accurate indication of actual control rod position, but at a lower precision than the step counters. This system is based on inductive analog signals from a series of alternating primary and secondary coils spaced along a hollow tube. The normal indication accuracy of the RPI System is t 6 steps (t 3.75 inches), and the t

maximum uncertainty is 12 steps (t 7.5 inches). With an indicated deviation of 12 steps between the group step counter and RPl, the maximum deviation between actual rod position and the demand position could be 24 steps, or 15 inches.

The Power Distribution Monitoring System (PDMS) as controlled by Technical Requirements Manual Section 3.3.9 develops a detailed three dimensional power distribution via its nodal code coupled with updates from plant instrumentation, including the fixed incore detectors. The monitored power distribution is compared to the reference power distribution corresponding to all control rods properly aligned. Agreement between the two power distributions can be used to indirectly verify the control rod is aligned.

APPLICABLE Controland shutdown rod position accuracy is essentialduring power SAFETY operation. Power peaking, ejected rod worth, or SDM limits may be ANALYSES violated in the event of a Design Basis Accident (Ref. 2 through 12), with control or shutdown rods operating outside their limits undetected.

Therefore, the acceptance criteria for rod position indication is that rod positions must be known with sufficient accuracy in order to verify the core is operating within the group sequence, overlap, design peaking limits, ejected rod worth, and with minimum SDM (LCO 3.1.6, "Shutdown Bank lnsertion Limits," and LCO 3.1.7, "ControlBank lnsertion Limits").

The rod positions must also be known in order to verify the alignment limits are preserved (LCO 3.1.5, "Rod Group Alignment Limits"). Control rod positions are continuously monitored to provide operators with information that ensures the plant is operating within the bounds of the accident analysis assumptions.

The control rod position indicator channels satisfy Criterion 2 of 10 CFR 50.36(cX2)(ii). The control rod position indicators monitor control rod position, which is an initial condition of the accident.

(continued)

Watts Bar - Unit 2 B 3.1-49 Revision 16 Amendment 20

Rod Position lndication B 3.1 .8 BASES (continued)

LCO LCO 3.1.8 specifies that the RPI System and the Bank Demand Position lndication System be OPEMBLE for all control rods. For the control rod position indicators to be OPERABLE requires meeting the SR of the LCO (when required) and the following:

a. The RPI System indicates within 12 steps of the group step counter demand position when LCO 3.1.5, "Rod Group Alignment Limits;"

met.

b. For the RPI System there are no failed coils; and

c. The Bank Demand lndication System has been calibrated either in the fully inserted position or to the RPI System.

The 12 step agreement limit between the Bank Demand Position lndication System and the RPI System indicates that the Bank Demand Position lndication System is adequately calibrated, and can be used for indication of the measurement of control rod bank position.

A deviation of less than the allowable limit, given in LCO 3.1.5, in position indication for a single control rod, ensures high confidence that the position uncertainty of the corresponding control rod group is within the assumed values used in the analysis (that specified control rod group insertion limits).

These requirements ensure that control rod position indication during power operation and PHYSICS TESTS is accurate, and that design assumptions are not challenged. OPEMBILITY of the position indicator channels ensures that inoperable, misaligned, or mispositioned control rods can be detected. Therefore, power peaking, ejected rod worth, and SDM can be controlled within acceptable limits.

The LCO is modified by a NOTE stating that the RPI system is not required to be OEPRABLE for t hour following movement of the associated rods. Control and shutdown rod temperature affects the accuracy of the RPI System. Due to changes in the magnetic permeability of the drive shaft as a function of temperature, the indicated position is expected to change with time as the drive shaft temperature changes. The one hour period allows temperature to stabilize following rod movement in order to ensure the indicated position is accurate.

(continued)

Watts Bar - Unit 2 B 3.1-50 Revision 16 Amendment 20

Rod Position lndication B 3.1 .8 BASES (continued)

APPLICABILITY The requirements on the RPI and step counters are only applicable in MODES 1and2 (consistentwith LCO 3.1.5, LCO 3.1.6, and LCO 3.1.7),

because these are the only MODES in which power is generated, and the OPERABILITY and alignment of rods have the potentialto affect the safety of the plant. ln the shutdown MODES, the OPERABILITY of the shutdown and control banks has the potentialto affect the required SDM, but this effect can be compensated for by an increase in the boron concentration of the Reactor Coolant System.

ACTIONS The ACTIONS table is modified by a Note indicating that a separate Condition entry is allowed for each inoperable rod position indicator and each demand position indicator. This is acceptable because the Required Actions for each Condition provide appropriate compensatory actions for each inoperable position indicator.

A.1 and A.2 When one RPI channel per group in one or more groups fails, the position of the rod can still be determined indirectly by use of incore power distribution measurement information. I ncore power distribution measurement information is obtained from an OPERABLE Power Distribution Monitoring System (PDMS) (Ref. 15). The Required Action may also be satisfied by ensuring at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months that Fo satisfies LCO 3.2.1, FNs satisfies LCO 3.2.2, and SHUTDOWN MARGIN is within the limits provided in the COLR, provided the non-indicating rods have not been moved. Based on experience, normal power operation does not require excessive movement of banks. lf a bank has been significantly moved, the Required Action of C.1 or C.2 below is required.

Therefore, verification of rod position within the Completion Time of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months is adequate for allowing continued full power operation, since the probability of simultaneously having a rod significantly out of position and an event sensitive to that rod position is small.

Required Action A.1 requires verification of a rod with an inoperable RPI once per I hours. Required Action A.2 provides an alternative. Required Action A.2 requires verification of rod position using incore power distribution measurement information every 31 EFPD, which coincides with the normal measurements to verifo core power distribution.

Required Action A.2 includes six distinct requirements for verification of the position of rods associated with an inoperable RPI using incore power distribution measurement information:

a. lnitialverification within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months of the inoperability of the RPI; (continued)

Watts Bar - Unatz B 3.1-51 Revision 16 Amendment20

Rod Position lndication B 3.1 .g BASES (continued)

ACTIONS A.1 and A.2 (continued)

b. Re-verification once every 31 Effective Full Power Days (EFpD) thereafter;

c. Verification within I hours after discovery of each unintended rod movement. An unintended rod movement is defined as the release of the rod's stationary gripper when no action was demanded either manually or automatically from the rod controlsystem, or a rod motion in a direction other than the direction demanded by the rod control system. Verifying that no unintended rod movement has occurred is performed by monitoring the rod control system stationary gripper coil current for indications of rod movement;

d. Verification within I hours if the rod with an inoperable RPI is intentionally moved greater than 12 steps;

e. Verification prior to exceeding 50% RTP if power is reduced below 50% RTP; and

f. Verification within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months of reaching 100% RTP if power is reduced to less than 100% power RTP.

Should the rod with the inoperable RPI be moved more than 12 steps, or if reactor power is changed, the position of the rod with the inoperable RPI must be verified.

AJ Reduction of THERMAL PowER to s 50% RTP puts core into a condition where rod position is not significantly affecting core peaking factors (Ref.

4). The allowed Completion Time of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months is reasonable, based on operating experience, for reducing power to s 50% RTP from full power conditions without challenging plant systems and allowing for rod position determination by Required Actions A.1 and A.2 above.

(continued)

Watts Bar - Unit2 B 3.1-52 Revision 16 Amendment 20

Rod Position Indication B 3.1 .g BASES (continued)

ACTIONS B.1 and B.2 (continued)

When more than one RPI per group in one or more groups fail, additional actions are necessary. Placing the Rod Control System in manual assures unplanned rod motion will not occur. The immediate completion Time for placing the Rod Control System in manual reflects the urgency with which unplanned rod motion must be prevented while in this Condition.

The inoperable RPls must be restored, such that a maximum of one Rpl per group is inoperable, within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Completion Time provides sufficient time to troubleshoot and restore the Rpl system to operation while avoiding the plant challenges associated with the shutdown without full rod position indication.

Based on operating experience, normal power operation does not require excessive rod movement. lf one or more rods has been significanfly moved, the Required Action of C.1 or C.2 below is required.

C.l and C.2 with one or more RPI inoperable in one or more groups and the affected groups have moved greater than 24 steps in one direction since the last determination of rod position, additional actions are needed to verify the position of rods with inoperable RPl. Within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , the position of the rods with inoperable position indication must be determined using the PDMS to verify these rods are still properly positioned, relative to their group positions.

lf, within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , the rod positions have not been verified, THERMAL POWER must be reduced to s 50% RTP within I hours to avoid undesirable power distributions that could result from continued operation at> 50o/o RTP, if one or more rods are misaligned by more than 24 steps.

The allowed completion Time of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months provides an acceptable period of time to verify the rod positions.

D.l.1 andD.l.2 With one or more demand position indicators per bank inoperable in one or more banks, the rod positions can be determined by the Rpl System.

Since normal power operation does not require excessive movement of rods, verification by administrative means that the rod position indicators are OPERABLE and the most withdrawn rod and the least withdrawn rod are 312 steps apart within the allowed Completion Time of once every 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months is adequate.

(continued)

Watts Bar - Untz B 3.1-53 Revision 16 Amendment 20

Rod Position lndication B 3.1 .g BASES ACTIONS D.2 (continued)

Reduction of THERMAL POWER to < 50% RTP puts the core into a condition where rod position is not significantly affecting core peaking factor limits (Ref. 13). The allowed Completion Time of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months provides an acceptable period of time to verifo the rod positions per Required Actions D.1.1 and D.1.2or reduce powers 50% RTP.

E.1 lf the Required Actions cannot be completed within the associated Completion Time, the plant must be brought to a MODE in which the requirement does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . The allowed Completion Time is reasonable, based on operating experience, for reaching the required MODE from full power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE sR 3.1.8.1 REQUIREMENTS Verification that the RPI agrees with the demand position within 12 steps ensures that the RPI is operating correctly.

This Surveillance is performed prior to reactor criticality after each removal of the reactor head, as there is the potential for unnecessary plant transients if the SR were performed with the reactor at power.

The Surveillance is modified by a NOTE which states it is not required to be met for RPls associated with rods that do not meet LCO 3.1.5. lf a rod is known to not be within 12 steps of the group demand position, ACTIONS of LCO 3.1.5 provide the appropriate Actions.

Watts Bar - Unit 2 B 3.1-54 (continued)

Revision 16 Amendment20

Rod Position lndication B 3.1.8 BASES (continued)

This Page lnitially Left Blank Watts Bar - Unit 2 B 3.1-56

QPTR B 3.2.4 BASES ACTIONS B.1 (continued) lf Required Actions A.1 through A.6 are not completed within their associated Completion Times, the unit must be brought to a MODE or condition in which the requirements do not apply. To achieve this status, THERMAL POWER must be reduced to < 50% RTP within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . The allowed Completion Time of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months is reasonable, based on operating experience regarding the amount of time required to reach the reduced power level without challenging plant systems.

SURVEILLANCE sR 3.2.4.1 REQUIREMENTS SR 3.2.4.1 is modified by two Notes. Note 1 allows QPTR to be calculated with three power range channels if THERMAL POWER is 375o/o RTP and the input from one power range neutron flux channel is inoperable. Note 2 allows performance of SR 3.2.4.2 in lieu of SR 3.2.4.1 if more than one input from power range neutron flux channels are inoperable.

This Surveillance verifies that the QPTR, as indicated by the Nuclear lnstrumentation System (NlS) excore channels, is within its limits. The Frequency of 7 days when the QPTR alarm is OPEMBLE is acceptable because of the low probability that this alarm can remain inoperable without detection.

When the QPTR alarm is inoperable, the Frequency is increased to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . This Frequency is adequate to detect any relatively slow changes in QPTR, because for those changes of QPTR that occur quickly (e.9., a dropped rod), there typically are other indications of abnormality that prompt a verification of core power tilt.

sR 3.2.4.2 This Surveillance is modified by a Note, which states the surveillance is only required to be performed if input to QPTR from one or more Power Range Neutron Flux channels are inoperable with THERMAL POWER

>75% RTP.

With an NIS power range channel inoperable, tilt monitoring for a portion of the reactor core becomes degraded. Large tilts are likely detected with the remaining channels, but the capability for detection of smal! power tilts in some quadrants is decreased. Performing SR 3.2.4.2 at a Frequency of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months provides an accurate alternative means for ensuring that any tilt remains within its limits.

(continued)

Watts Bar - Unit 2 B 3.2-30 Revision 17 Amendment 21

RTS Instrumentation B 3.3.1 BASES APPL!CABLE a. Power Ranqe Neutron FIux - High Positive Rate (continued)

SAFETY ANALYSES, The LCO requires all four of the Power Range Neutron Flux -

LCO, and High Positive Rate channels to be OPERABLE.

APPLICABILITY (continued) ln MODE 1 or 2, when there is a potentia! to add a large amount of positive reactivity from a rod ejection accident (REA), the Power Range Neutron FIux - High Positive Rate trip must be OPERABLE. In MODE 3, 4, 5, or 6, the power Range Neutron Flux - High Positive Rate trip Function does not have to be OPERABLE because other RTS trip Functions and administrative controls will provide protection against positive reactivity additions. Also, since only the shutdown banks may be withdrawn in MODE 3,4, or S, the remaining complement of control bank worth ensures a sufficient degree of SDM in the event of an REA. In MODE 6, no rods are withdrawn and the SDM is increased during refueling operations. The reactor vessel head is atso removed or the closure bolts are detensioned preventing any pressure buildup. ln addition, the NIS power range detectors cannot detect neutron levels present in this MODE.

b. Power Ranoe Neutron Flux - Hiqh Negative Rate Deleted

4. lntermediate Ranoe Neutron Flux The lntermediate Range Neutron Flux trip Function ensures that protection is provided against an uncontrolled RCCA bank rod withdrawal accident from a subcritical condition during startup.

This trip Function provides backup protection to the power Range Neutron Flux - Low Setpoint trip Function. The NIS intermediate range detectors are located extemal to the reactor vessel and measure neutrons leaking ftom the core. This Function also provides a signalto prevent automatic and manual rod withdrawal prior to initiating a reactor trip.

The LCO requires two channels of lntermediate Range Neutron Flux to be OPEMBLE. Two OPERABLE channels are sufficient to ensure no single random failure willdisable this trip Function.

(continued)

Watts Bar - Unit 2 B 3.3-13 Revision 18

RTS lnstrumentation B 3.3.1 BASES APPLICABLE 5. Source Ranoe Neutron Flux (continued)

SAFEry ANALYSES, The Source Range Neutron Flux Function provides protection for LCO, and control rod withdrawalfrom subcritical, boron dilution and control APPLICABILITY rod ejection events. The Function also provides visual neutron flux (continued) indication in the controlroom.

ln MODE 2 when below the P-6 setpoint during a reactor startup, the Source Range Neutron Flux trip must be OPEMBLE. Above the P-6 setpoint, the lntermediate Range Neutron Flux trip and the Power Range Neutron Flux - Low Setpoint trip will provide core protection for reactivity accidents. Above the P-6 setpoint, the NtS Source Range Neutron Flux trip Function may be manually blocked. Above the P-10 setpoint, the NIS Source Range Neutron Flux trip function is automatically blocked.

ln MODE 3, 4, or 5 with the reactor shut down, the Source Range Neutron Flux trip Function must also be OPERABLE. lf the CRD System is capable of rod withdrawal, the Source Range Neutron Flux trip must be OPERABLE to provide core protection against a rod withdrawal accident. lf the GRD System is not capable of rod withdrawal, the source range detectors are not required to trip the reactor. However, their monitoring Function must be OPEMBLE to monitor core neutron levels and provide visual indication and audible alarm of reactivity changes that may occur as a result of events like a boron dilution. The requirements for the NIS source range detectors in MODE 6 are addressed in LCO 3.9.3, "Nuclear lnstrumentation."

6. Overtemoerature AT The Overtemperature AT trip Function is provided to ensure that the design limit DNBR is met. This trip Function also limits the range overwhich the Overpower AT trip Function must provide protection. The inputs to the Overtemperature AT trip include pressurizer pressure, coolant temperature, axial power distribution, and reactor power as indicated by loop AT assuming full reactor coolant flow. Protection from violating the DNBR limit is assured for those transients that are slow with respect to delays from the core to the measurement system. The Function monitors both variation in power and flow since a decrease in flow has the same effect on AT as a power increase. The Overtemperature AT trip (continued)

Watts Bar - Unit 2 B 3.3-15 Revision 18

RTS Instrumentation B 3.3.1 BASES ACTIONS C.1 and C.2 (continued) must be opened within the next hour. The additional hour provides sufficient time to accomplish the action in an orderly manner. With the RTBs open, these Functions are no longer required. The Completion Time is reasonable considering that in this Condition, the remaining OPERABLE channel or train is adequate to perform the safety function, and given the low probability of an event occurring during this interval.

D.1 and D.2 Condition D applies to the Power Range Neutron Flux - High Function.

The NIS power range detectors provide input to the CRD System and the SG Water Level Control System and, therefore, have a two-out-of-four trip logic. A known inoperable channel must be placed in the tripped condition. This results in a partial trip condition requiring only one-out-of-three logic for actuation. The 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months allowed by Required Action D.1 to place the inoperable channel in the tripped condition is justified in Reference 14.

The Required Actions have been modified by two Notes. Note 1 allows the inoperable channel to be placed in the bypassed condition for up to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months while performing routine surveillance testing of other channels.

With one channel inoperable, the Note also allows routine surveillance testing of another channel with the inoperable channel in bypass. The Note also allows placing the inoperable channel in the bypass condition to allow setpoint adjustments of other channels when required to reduce the Power Range Neutron Flux-High setpoint in accordance with other Technical Specifications. The 12hour time limit is justified in Reference 14.

Note 2 states to perform SR 3.2.4.2 if input to QPTR from one or more Power Range Neutron Flux channels are inoperable with THERMAL POWER > 75% RTP.

lf Required Action D.1 cannot be met within the specified Completion Time, the plant must be placed in a MODE where this Function is no longer required OPERABLE. Seventy-eight hours are allowed to place the plant in MODE 3. The 78 hour9.027778e-4 days 0.0217 hours 1.289683e-4 weeks 2.9679e-5 months Completion Time includes 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months for the MODE reduction as required by Required Action D.2. This is a reasonable time, based on operating experience, to reach MODE 3 from full power in an orderly manner and without challenging plant systems.

(continued)

Watts Bar - Unit 2 B 3.3-38 Revision 17 Amendment 21

RTS lnstrumentation B 3.3.1 BASES ACTIONS E.1 and E.2 (continued)

Condition E applies to the following reactor trip Functions:

o Power Range Neutron Flux - Low; and o Power Range Neutron Flux - High Positive Rate.

A known inoperable channel must be placed in the tripped condition within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . Placing the channel in the tripped condition results in a partial trip condition requiring only one-out-of-two logic for actuation of the two-out-of-three trips and one-out-ofthree logic for actuation of the two-out-of-four trips. The72 hours allowed to place the inoperable channel in the tripped condition is justified in Reference 14.

lf the inoperable channel cannot be placed in the trip condition within the specified Completion Time, the plant must be placed in a MODE where these Functions are not required OPEMBLE. An additional6 hours is allowed to place the plant in MODE 3. Six hours is a reasonable time, based on operating experience, to place the plant in MODE 3 from full power in an orderly manner and without challenging plant systems.

The Required Actions have been modified by a Note that allows placing the inoperable channel in the bypassed condition for up to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months while performing routine surveillance testing of the other channels. The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months time limit is justified in Reference 14.

(continued)

Watts Bar - Unit 2 B 3.3-39 Revision 17 Amendment 21

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 normaloperation 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, Pa (15.0 psig), bounds the calculated results from the limiting LOCA. The rnaximum 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

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 I 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 fullflow 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 2U 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. ln 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 10 CFR 50, Appendix K (Ref. 3).

lnadvertent 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)(2Xii).

Watts Bar - Unit 2 B 3.6-36 (continued)

Revision 3, 15

EGTS B 3.6.9 BASES BACKGROUND The prefilters remove large particles in the air, and the moisture (continued) separators remove entrained water droplets present, to prevent excessive loading of the HEPA filters and charcoal absorbers. Heaters are included to reduce the relative humidity of the airstream on systems that operate in high humidity. Operation with the heaters on for 215 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. Cross-over flow ducts are provided between the two trains to allow the active train to draw air through the inactive train and cool the air to keep the charcoal beds on the inactive train from becoming too hot due to absorption of fission products.

The containment annulus vacuum fans maintain the annulus at -5 inches water gauge vacuum during norma! operations. During accident conditions, the containment annulus vacuum fans are isolated from the air cleanup portion of the system.

The EGTS reduces the radioactive content in the shield building atmosphere following a DBA. Loss of the EGTS could cause site boundary doses, in the event of a DBA, to exceed the values given in the licensing basis.

APPLICABLE The EGTS design basis is established by the consequences of the SAFETY limiting DBA, which is a LOCA. The accident analysis (Ref. 3) considers ANALYSES two different single failure scenarios. The first one assumes that only one train of the EGTS is functional due to a postulated single failure that disables the other train. An alternate scenario assumes a single failure of the pressure control loop associated with one train of pressure control operators (PCO). The first scenario is bounding for thyroid dose while the alternate scenario is bounding for beta and gamma doses. The accident analysis accounts for the reduction in airborne radioactive material provided by the number of filter trains in operation for each failure scenario. The amount of fission products available for release from containment is determined for a LOCA.

The safety analysis conservatively assumes the annulus is at -0.5 inches water gauge pressure prior to the LOCA. The analysis further assumes that upon receipt of a Containment lsolation Phase A (ClA) signalfrom the RPS, the EGTS fans automatically start and achieve a minimum flow of 3600 cfm (per train) within 18 seconds (20 seconds from the initiating event.) This does not include 10 seconds for diesel generator startup.

(continued)

Watts Bar - Unit 2 B 3.6-49 Revision 12 Amendment 15

EGTS B 3.6.9 BASES ACTIONS A.1 (continued)

With one EGTS train inoperable, the inoperable train must be restored to OPERABLE status within 7 days. The components in this degraded condition are capable of providing looo/o of the iodine removal needs after a DBA. The 7-day Completion Time is based on consideration of such factors as the availability of the OPERABLE redundant EGTS train and the low probability of a DBA occurring during this period. The Completion Time is adequate to make most repairs.

B.1 and B.2 lf the EGTS train 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 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and to MODE S within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . 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 3.6.9.1 REQUIREMENTS Operating each EGTS train for ) 15 minutes with heaters on ensures that all trains are OPERABLE and that all associated controls are functioning properly. lt also ensures that blockage, fan or motor failure, or excessive vibration can be detected for corrective action. The 31-day Frequency was developed in consideration of the known reliability of fan motors and controls, the two train redundancy available.

sR 3.6.9.2 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 inctudes 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).

continued Watts Bar - Unit 2 B 3.6-51 Revision 12 Amendment 15

ABGTS 83.7.12 APPLICABLE The ABGTS design basis is established by the consequences of the SAFETY limiting Design Basis Accident (DBA), which is a LOCA. The analysis of ANALYSES the LOCA assumes that radioactive materials leaked from the Emergency Core Cooling System (ECCS) are filtered and adsorbed by the ABGTS.

The DBA analysis assumes that only one train of the ABGTS is functional due to a single failure that disables the other train. The accident analysis accounts for the reduction in airborne radioactive material provided by the one remaining train of this filtration system. The amount of fission products available for release from the ABSCE is determined for a LOCA.

The assumptions and analysis for a LOCA follow the guidance provided in Regulatory Guide 1.4 (Ref. 5).

The ABGTS satisfies Criterion 3 of 10 CFR 50.36(cX2Xii).

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. Totalsystem 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 individualcomponents necessary to control exposure in the Auxiliary 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 individualwill 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.

continued Watts Bar - Unit 2 B 3.7-64 Revision 13 Amendment 16

ABGTS B 3.7.12 APPLICABILITY ln 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.

ln MODE 5 or 6, the ABGTS is not required to be OPERABLE since the ECCS is not required to be OPERABLE.

ACTIONS 4.1 With one ABGTS train inoperable, action must be taken to restore OPERABLE status within 7 days. During this period, the remaining OPERABLE train is adequate to perform the ABGTS function. The 7-day Completion Time is based on the risk from an event occurring requiring the inoperable ABGTS train, and the remaining ABGTS train providing the required protection.

B.1. B.2. and B.3 lf the ABSCE boundary is inoperable, the ABGTS trains cannot perform their intended functions. Actions must be taken to restore an OEpRABLE ABSCE boundary within seven days. During the period that the ABSCE boundary is inoperable, action must be initiated to implement mitigating actions consistent with the intent, as applicable, of GDC 19, 60, 61, 63, 64 and 10 CFR Part 100 (Ref. 6) to protect plant personnelfrom potential hazards such as radioactive contamination, temperature and relative humidity, and physica! security. Actions must be taken within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to verify that, in the event of a DBA, main control room occupant radiological exposures will not exceed 10 CFR 50 Appendix A GDC 19 limits. These mitigating actions (i.e., actions that are taken to offset the consequences of the inoperable ABSCE boundary) should be preplanned to address these concerns for intentional and unintentional entry into the condition.

The 24-hour Completion Time is reasonable based on the low probability of a DBA occurring during this time period, and the use of mitigating actions. The seven-day Completion Time is reasonable based on the determination that the mitigating actions will ensure protection of main control room occupants within analyzed limits (Ref. 9) while limiting the probability that main control room occupants will have to implement protective measures that may adversely affect their ability to controlthe reactor and maintain it in a safe shutdown condition in the event of a DBA. ln addition, the seven-day Completion Time is a reasonable time to diagnose, plan and possibly repair, and test most problems with the ABSCE boundary.

continued Watts Bar - Unit 2 B 3.7-65 Revision 13 Amendment 16

ABGTS B 3 .7.12 BASES (continued)

ACTIONS C.1 and C.2 (continued)

When Required Action A.1 or Required Actions 8.1,8.2, and B.3 cannot be completed within the associated Completion Time, or when both ABGTS trains are inoperable for reasons other than an inoperable ABSCE boundary (i.e., Condition B), the plant must be placed in a MODE in which the LCO does not apply. To achieve this status, the plant must be placed in MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and in MODE 5 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

The Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

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 once every month 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 31-day Frequency is based on the known reliability of the equipment and the two train redundancy available.

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. 7).

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 detai! in the VFTP.

continued Watts Bar - Unit 2 B 3.7-66 Revision 13 Amendment 16

ABGTS B 3 .7.12 BASES REFERENCES 7. Regulatory Guide 1.52 (Rev. 2), "Design, Testing and Maintenance (continued) Criteria for Post Accident Eng ineered-Safety-Feature Atmospheric Cleanup System Air Filtration and Adsorption Units of Light-Water Cooled Nuclear Power Plants."

8. NUREG-0800, Section 6.5.1 , "Standard Review Plan," Rev. 2, "ESF Atmosphere Cleanup System," July 1981.

TVA Calculation MDQ000030201400061 8, "Offsite and Control

9. Room Doses without the Auxiliary Building Secondary Containment Enclosure (ABSCE) during a LOCA."

Watts Bar - Unit 2 B 3.7-67a Revision 13 Amendment 16

AC Sources - Operating B 3.8.1 BASES LCO to power two shutdown boards in the same load group through either (continued) CSST A or CSST B and its associated Unit Boards, either directly from the CSST through the Unit Board or by automatic transfer from the Unit Station Service Transformer (USST) to the CSST. Use of CSST A or B as an offsite source requires that CSST A and B both be available and that the associated power and control feeders be in their normal positions to ensure independence. Due to independence limitations, CSST A and B cannot be credited for supply of both the offsite power sources simultaneously. The medium voltage power system starts at the low-side of the common station servicertransformers.

Each required offsite circuit is that combination of power sources described below that are either connected to the Class 1E AC Electrical Power Distribution System, or is available to be connected to the Class 1E AC Electrical Power Distribution System through automatic transfer at the 6.9 kV Shutdown or Unit Boards within a few seconds as required.

The following offsite power configurations meet the requirements of the LCO:

1. Normal Operation (i.e., all 6.9 kV shutdown boards aligned to their normal offsite circuit) - Two offsite circuits consisting of (a) AND (b)

(no board transfers required; a loss of either circuit will not prevent the minimum safety functions from being performed);

a. From the 161 kV Watts Bar Hydro Switchyard (Bay 13), through CSST C (winding Y) to 6.9 kV Shutdown Board 1A-A and (winding X) to 6.9kV Shutdown Board 2A-A:AND

b. From the 161 kV Watts Bar Hydro Switchyard (Bay 4), through CSST D (winding X) to 6.9 kV Shutdown Board 1B-B and (winding Y) to 6.9 kV Shutdown Board 2B-B.

2. Alternate Operation (i.e., one or more 6.9 kV shutdown boards aligned to their alternate offsite circuit) - Two offsite circuits consisting of (a)

AND (b) AND (c) (as needed) (Note: 6.9 kV shutdown board(s) aligned to normal circuit require an OPERABLE automatic transfer; a loss of either circuit will not prevent the minimum safety functions from being performed);

a. From the 161 kV Watts Bar Hydro Switchyard (Bay 13), through CSST C (winding Y) to 6.9 kV Shutdown Board 1A-A (normal)

AND/OR Shutdown Board 2B-B (alternate) and (winding X) to 6.9 kV Shutdown Board 2A-A (normal) AND/OR Shutdown Board 1&B (alternate);

(continued)

Watts Bar - Unat 2 B 3.8-4 Revision 14

AC Sources - Operating B 3.8.1 BASES Bases Table 3.8.1-2 TS Action or Surveillance Requirement (SR) Gontingency Actions Contingency Actions Applicable TS Applicable to be Implemented Action or SR Modes

1. Verify that the offsite power system is stable. This sR 3.8 .1 .14 1,2 action will establish that the offsite power system is Action B.5 1,2,3,4 within single-contingency limits and will remain stable upon the loss of any single component supporting the system. lf a grid stability problem exists, the planned DG outage will not be scheduled.

2. Verify that no adverse weather conditions are sR 3.8 .1 .14 1,2 expected during the outage period. The planned DG Action 8.5 1,2, 3, 4 outage will be postponed if inclement weather (such as severe thunderstorms or heavy snowfall) is projected.

3. Do not remove from service the ventilation systems Action 8.5 1,2,3,4 for the 6.9 kV shutdown boardrooms, the elevation 772 transformer rooms, or the 480-volt shutdown board rooms, concurrently with the DG, or implement appropriate compensatory measures.

4. Do not remove the reactor trip beakers from service Action 8.5 1,2, 3, 4 concurrently with planned DG outage maintenance.

5. Do not remove the turbine-driven auxiliary feedwater Action B.5 1,2,3,4 (AFW) pump from service concurrently with a Un;,t 2 DG outage.

6. Do not remove the AFW level control valves to the Action 8.5 1,2,3,4 steam generators from service concurrently with a Unit 2 DG outage

7. Do not remove the opposite train residual heat Action 8.5 1,2, 3, 4 remove (RHR) pump from service concurrently with a Unit 2 DG outage.

Watts Eiar - Un'ltz B 3.8-37a Revision 5, 13 Amendment 5

ENCLOSURE 7 WBN UNIT 2 TECHNICAL REQUIREMENTS MANUAL TABLE OF CONTENTS

TABLE OF CONTENTS TECHNICAL REQUI REMENTS TABLE OF CONTENTS LIST OF TABLES LIST OF FIGURES LIST OF MISCELLANEOUS REPORTS AND PROGRAMS LIST OF ACRONYMS ......

LIST OF EFFECTIVE PAGES TR 1.0 1 .1-1 TR 1.1 1 .1-1 TR 1.2 Logical Connectors 1.2-1 TR 1.3 Completion Times 1.3-1 TR 1.4 1.4-1 TR 3.0 APPLICABILITY 3.0-1 TR 3.1 REACTIVTTYCONTROLSYSTEMS ...... ...... 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 ... .. .. ... ... .. .. .. r 3.1-10 TR 3.1 .7 Position lndication System, Shutdown ... ... ... . ... .. .. ... ... ... ... 3.1-14 TR 3.3 INSTRUMENTATION 3.3-1 TR 3.3.1 Reactor Trip System (RTS) lnstrumentation ...... ....... 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 lnstrumentation ... 9.3-12 TR 3.3.5 RESERVED FOR FUTURE ADDITION ... 3.3-16 TR 3.3.6 Loose-Part Detection System ..... g.g-17 TR 3.3.7 RESERVED FOR FUTURE ADDITION ... ....... 3.3-18 TR 3.3.8 Hydrogen Monitor .. 3.3-19 TR 3.3.9 Power Distribution Monitoring System (PDMS) . 9.3-21 Watts Bar - Unit 2 (continued)

Technical Req uirements

TABLE OF CONTENTS (continued)

TECHNICAL REQUI REMENTS 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 Tem peratu re Limits 3.4-3 TR 3.4.3 Reactor Vessel Head Vent System 3.4-5 TR 3.4.4 3.4-7 TR 3.4.5 Piping System Structural lntegrity .. ... ... ... .. 3.4-10 TR 3.6 CoNTAINMENTSYSTEMS ... ........... 3.6_1 TR 3.6.1 lce Bed Temperature Monitoring System ......... 3.6-1 TR 3.6.2 lnlet Door Position Monitoring System ... 3.6-4 TR 3.6.3 Lower Compartment Cooling (LCC) System .... 3.6-6 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 3.7-5 TR 3.7.4 Sealed Source Contamination 3.7-16 TR 3.7.5 Area Temperature Monitoring 3.7-19 TR 3.8 ELECTRICAL POWER SYSTEMS 3.9-1 TR 3.9.1 lsolation Devices . ... . ... .. .. ... ... ... ... . 3.8-1 TR 3.9.2 Containment Penetration Conductor Overcurrent Protection Devices 3.8-4 TR 3.8.3 Motor-Operated Valves Thermal Overload Bypass Devices ............ 3.9-g TR 3.8.4 Submerged Component Circuit Protection 3.9-15 TR 3.9 REFUELING OPEMTIONS ... . 3.9-1 TR 3.9.1 RESERVED FOR FUTURE ADDITION ... 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-s TR 5.0 ADMlNlSTRATlVE CONTROLS ... ................ 5.0_1 TR 5.1 Technical Requirements Control Program ... .... S.O-1 Watts Bar - Unit 2 Technical Requirements Revision 5

TABLE OF CONTENTS (continued)

TECHNICAL REQUIREMENTS BASES B 3.0 TECHNTCAL REQUIREMENT (TR) AND TECHNTCAL SURVElLLANCE REQUTREMENT (TSR) AppLtCABtLtTy .............. 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 lndication System, Shutdown B 3.1-24 B 3.3 INSTRUMENTATION B 3.3-1 B 3.3.1 Reactor Trip System (RTS) lnstrumentation ... B 3.3-1 B 3.3.2 Engineered Safety Features Actuation System (ESFAS) lnstrumentation ... B 3.34 B 3.3.3 RESERVED FOR FUTRE ADDITION ...... B 3.3-7 B 3.3.4 Seismic lnstrumentation ... 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 RESERVED FOR FUTURE ADDITION ... 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 Tem peratu re 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 lntegrity B 3.4-13 B 3.6 CONTAINMENT SYSTEMS ... B 3.6-1 B 3.6.1 lce Bed Temperature Monitoring System B 3.6-1 B 3.6.2 lnlet 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

TABLE OF CONTENTS (continued)

TECHNICAL REQUIREMENTS BASES 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 3.7-8 B 3.7.4 Sealed Source Contamination ... B 3.7-15 B 3.7.5 Area Temperature Monitoring B 3.7-19 B 3.8 ELECTRICAL POWER SYSTEMS B 3.8-1 B 3.9.1 !solation Devices B 3.9-1 B 3.8.2 Contai nment Penetration Cond uctor Overcu nent Protection Devices B 3.8-7 B 3.9.3 Motor Operated Valves Thermal Overload Bypass Devices B 3.8-13 B 3.8.4 Submerged Component Circuit Protection B 3.8-16 B 3.9 REFUELING OPERATIONS ... B 3.9-1 B 3.9.1 RESERVED FOR FUTURE ADDITION ... ....... B 3.9-1 B 3.9.2 Communications ... B 3.9-2 B 3.9.3 Refueling Machine B 3.9-4 B 3.9.4 Crane Travel - Spent Fuel Storage Pool Building ... ... B 3.9-7 Watts Bar - Unit 2 iv Technical Req u irements Revision 5

LIST OF TABLES TABLE NO. TITLE PAGE 1 .1-1 1 .1-6 3.0.2-1 3.0-5 3.3 .1-1 3.3-2 3.3.2-1 3.3-5 3.3.4-1 3.3-15 3.3.9-1 3.3-23 3.7.3-1 3.7-8 3.7.3-2 3.7-g 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-g 3.8.4-1 Submerged Components With Automatic De-energization Under Accident Conditions Watts Bar - Unit 2 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 3.7.t1 DELETED 3.7-15 LIST OF MISCELLANEOUS REPORTS AND PROGRAMS Core Operating Limits Report Watts Bar - Unit 2 VI Tech n ical Req uirements

LIST OF ACRONYMS (Page 1 ot 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 lsolation Valve MSSV Main Steam Safety Valve (continued)

Watts Bar - Unit 2 vii Technical Req uirements

LIST OF ACRONYMS (Page 2 of 2)

ACRONYM TITLE MTC Moderator Tem perature 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 lnjection SL Safety Limit SR Surveillance Req uirement TSR Technical Surveillance Requirement UHS Ultimate Heat Sink Watts Bar - Unit2 vilt Technical Requirements

TECHN ICAL REQUI REMENTS LIST OF EFFECTIVE PAGES PAGE REVISION PAGE REVISION NUMBER NUMBER NUMBER NUMBER 0 1.4-2 0 5 1.4-3 0 0 1.4-4 0 5 3.0-1 0 0 3.0-2 0 0 3.0-3 6 0 3.0-4 0 0 3.0-5 6 10 3.0-6 7 x 10 3.1-1 0 xi 8 3.1-2 0 xii 9 3.1-3 0 xiii 0 3.1-4 0 xiv 10 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 0 3.1-9 0 1.1-5 0 3.1-10 0 1 .1-6 0 3.1-1 1 0 1.2-1 0 3.1-12 0 1.2-2 0 3.1-13 0 1.2-3 0 3.1-14 0 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 0 1.3-9 0 3.3-9 0 1 .3-10 0 3.3-10 0 1.4-1 0 Watts Bar - Un,t2 tx Technical Requirements

TECHN ICAL REQ UI REMENTS LIST OF EFFECTIVE PAGES PAGE REVISION PAGE REVISION NUMBER NUMBER NUMBER 3.3-1 1 3.7-4 0 3.3-12 0 3.7-5 5 3.3-13 0 3.7-6 5 3.3-14 0 3.7-7 5 3.3-15 0 3.7-8 5 3.3-16 0 3.7-9 5 3.3-17 0 3.7-10 5 3.3-18 0 3.7-11 5 3.3-19 0 3.7-12 5 3.3-20 0 3.7-13 5 3.3-21 0 3.7-14 5 3.3-22 0 3.7-15 5 3.3-23 0 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 0 3.6-4 0 3.8-9 0 3.6-5 0 3.8-10 0 3.6-6 0 3.8-1 1 0 3.7-1 0 3.8-12 0 3.7-2 0 3.8-13 0 3.7-3 0 3.8-14 3 Watts Bar - Unit 2 Technical Requirements

TECHN ICAL REQ UI REMENTS LIST OF EFFECTIVE PAGES PAGE REVISION PAGE REVISION NUMBER NUMBER NUMBER NUMBER 3.8-15 0 B 3.1-7 0 3.8-16 0 B 3.1-8 0 3.8-17 8 B 3.1-g 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 0 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 B 3.1-6 0 B 3.3-14 0 Watts Bar - Unit 2 xi Technical Requirements

TECHN ICAL REQUI REMENTS LIST OF EFFECTIVE PAGES PAGE REVISION PAGE REVISION NUMBER NUMBER NUMBER NUMBER B 3.3-15 0 B 3.6-8 4 B 3.3-16 0 B 3.6-9 0 B 3.3-17 0 B 3.6-10 0 B 3.3-18 0 B 3.6-11 0 B 3.3-19 0 B 3.6-12 0 B 3.3-20 0 B 3.7-1 0 B 3.3-21 0 B 3.7-2 0 B 3.3-22 0 B 3.7-3 0 B 3.3-23 0 B 3.7-4 0 B 3.3-24 0 B 3.7-5 0 B 3.3-25 0 B 3.7-6 0 B 3.3-26 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 o 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 B 3.6-2 0 B 3.8-2 0 B 3.6-3 0 B 3.8-3 0 B 3.6-4 0 B 3.8-4 0 B 3.6-s 0 B 3.8-5 0 B 3.6-6 I B 3.8-6 0 B 3.6-7 0 B 3.8-7 0 Watts Bar - Unit2 xt!

Technical Requirements

TECHNICAL REQUI REMENTS LIST OF EFFECTIVE PAGES PAGE REVISION PAGE REVISION NUMBER NUMBER NUMBER NUMBER 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 0 B 3.8-16 0 B 3.8-17 0 B 3.8-18 0 B 3.8-19 0 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 - Unat 2 xilt Technical Req uirements

TECHNICAL REQUIREMENTS MANUAL LIST OF EFFECTIVE PAGES - REVISION LISTING Revisions lssued SUBJECT Revision 01 11125115 Revises TRM and rRM Bases section 3.7.3, "Snubbers".

Revision 02 05122116 TR Table 3.3.1-1, "Reactor Trip System lnstrumentation Response Times" , to change the overtemperature and over power times.

Revision 03 06127116 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 02121117 Revises TRM Bases 3.6.2, "lnlet Door position Monitoring System," Actions.

Revision 05 Ogl31l17 Revises TRM and TRM Bases to delete section 3.7.3 "Snubbers."

Revision 06 07108117 Revises TRM section 3.0, "Technical Surveillance Requirements (ISR) Applicability" and adds Table 3.0.2-1.

Revision 07 OBl22l17 Revises the TR 3.4.5 Title to add ASME Ctass 1,2, and 3 in the TRM and Bases. Also revised TSR Table 3.0.2-1to add two addition TSRs.

Revision 08 03/08/18 Revises TR Table 3.8.4-1to revise the dual fan motors which were replaced with single fan motors.

Revision 09 04106118 Revises TRM Bases 83.6.2 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 lnlet Door Position Monitoring system.

Revision 10 04127118 Revises TRM Table 3.7.5-1, ltem g to correct the unit identifier on the Mechanical Equipment Room.

Watts Bar - Unatz xiv Technical Requirements

ENCLOSURE 8 WBN UNIT 2 TECHNICAL REQUIREIUIENTS MANUAL CHANGED PAGES

Area Temperature Monitoring 3.7.5 Table 3.7.5-1 (Page 1 of 2)

Area Temperature Monitoring NORMAL ABNORMAL LIMIT LIMIT AREA ("F) ('F) 1 . Aux Bldg el 772 next to 480V Sd Bd transformer 1 A2-A. < 104 < 110

2. Aux BIdg el 772 next to 480V Sd Bd transformer 1 81-8. < 104 < 110

3. Aux Bldg el 772 next to 480V Sd Bd transform er 2A2-A. < 104 < 110

4. Aux BIdg el 772 next to 480V Sd Bd transform er 2B,2-B.. < 104 < 110

5. Aux BIdg el 772 next to 480V Rx MOV Bd 1A2-A. <83 < 104

6. Aux Bldg el 772 next to 480V Rx MOV Bd 2A2-A. <83 < 104

7. Aux Bldg el 772 next to 480V Rx MOV Bd 2B,2-B.. <93 < 104

8. Aux BIdg el 772 across from spare 125V vital battery <83 < 104 charger 1-S.

9. Aux Bldg el 772 U2 Mech Equip Room. <91

10. Aux Bldg el 757 U1 Sd Bd room behind stairs S-A3. <85 < 104 11 . Aux Bldg el 757 U2 Sd Bd room behind stairs S-A13. <85 < 104

12. Aux Bldg el 757 U1 Refueling beside Aux boration < 104 < 115 makeup tk.

13. Aux BIdg el 737 U2 outside supply fan room. < 104 < 110

14. Aux Bldg el 713 U2 across from AFW pumps. < 104 < 110

15. Aux Bldg el 692 U2 outside AFW pump room door. < 104 < 110

16. Aux BIdg el 692 U2 near boric acid concentrate filter vault. < 104 < 110

17. Aux Bldg el 676 next to O-L-629. < 104 < 110

18. North steam valve vault room U2. (at affected MSSVs) >50 >50

19. South steam valve vault room U2. (at affected MSSVs) >50 >50 (continued)

Watts Bar - Unit 2 3.7-22 Revision 10 Technical Requirements

Submerged Component Circuit Protection TR 3.8.4 Table 3.8.4-1 (Page 1 of 3)

Submerged Components \Mth Automatic De-energization Under Accident Conditions BOARD COMPT LOAD FCTN 6.9KV SHUTDOWN BOARD 2A-A 20 2-DPL-68-341A-A SI 21 2-DPL-68-341F SI 6.9KV SHUTDOWN BOARD 2B-B 20 2-DPL-68-341D-B SI 21 2-DPL-68-341H" SI 48OV SHUTDOWN BOARD 2A1.A 7B 2-MTR-30-83-A CIB 7C 2-MTR-30-74-A CIB 48OV SHUTDOWN BOARD 281-8 7C 2-MTR-30-92-B CIB 7D 2-MTR-30-75-B crB 48OV SHUTDOWN BOARD 2A2-A 7A 2-MTR-30-88-A CIB 7D 2-MTR-30-77-A crB 48OV SHUTDOWN BOARD 2B,2-B 7B 2-MTR-30-80-B CIB 7D 2-MTR-30-78-B CIB 48OV REACTOR MOV BOARD 164 2-MTR-31-265 CIA 2p.1-A 17E 2-PO-213-A1l(1-5) SI 18F2 2-PO-213-A1t(6-1 0) SI 48OV REACTOR MOV BOARD 16A 2-MTR-31-266 CIA 2B1-B 16E 2-PO-213-81(1-5) SI 17E 2-PO-213-81t(6-1 0) SI (continued)

Watts Bar - Unit 2 3.8-17 Revision 8 Technical Requirements

Submerged Component Circuit Protection rR 3.8.4 Table 3.8.4-1 (Page 3 of 3)

Submerged Components With Automatic De-energization Under Accident Conditions BOARD COMPT LOAD FCTN 125VDC VITAL BATTERY 420 2-FCV-43-2-8 CIA BOARD IV 421 2-FCV-43-1 1-B CIA 422 2-FCV-43-22-8 CIA p.23 2-FCV-43-34-B CIA 424 2-FCV-77-1 6-8 CIA A43 2-FCV-77-127-B C!A 444 a-FCV-77-9-B crA 445 2-FCV-77-1 8-B CIA 826 2-FCV-30-8/50-B CVI B.27 2-FCV-90-108-B CVI 828 2-FCV-90-1 10-B CVI B,32 2-FCV 15t57-B cvr 833 2-FCV-90-114-B CVI 834 2-FCV-30-58-B CVI 836 2-FCV-90-1 16-8 CVI C5 2-FCV-31-327-8 CIA C6 2-FCV-30-329-B CIA c26 2-FCV-61 -122-8 crA c34 2-FCV-90-109-B CVI c35 2-FCV-90-1 15-B CVI c41 2-FCV-43-54D-B CIA c42 2-FCV-43-56D-B CIA c43 2-FCV-43-59D-B CIA c44 2-FCV-43-63D-B CIA CIA: CONTAINMENT ISOLATION PHASE A CIB: CONTAINMENT ISOLATION PHASE B CU: CONTAINMENT VENT ISOLATION SI:

SAFETY INJECTION No adverse impact on power supplies if energized after accident signal reset.

Watts Bar - Unit 2 3.8-19 Revision 8 Technical Requirements

lnlet Door Position Monitoring System B 3.6.2 B 3.6 CONTAINMENT SYSTEMS B 3.6.2 lnlet Door Position Monitoring System BASES BACKGROUND Ninety-six limit switches monitor the position of the lower inlet doors.

Two switches are mounted on the door frame for each door panel.

The position and movement of the switches are such that the doors must be effectively sealed before the switches are actuated. A single annunciator window in the control room gives a common alarm signal when any door is open. Open/shut indication is also provided at the lower inlet door position display panel located in the Main Control Room. For door monitoring purposes, the ice condenser is divided into six zones, each containing four inlet door assemblies, or a total of eight door panels.

The limit switches on the doors in any single zone are wired to a single light on the inlet door position display panel such that a closed light indicates that allthe doors in that zone are shut and an open light indicates that one or more doors in that zone are open (Ref. 1).

Additional information on the design of the lower inlet door monitoring instrumentation is provided in UFSAR Section 6.7.15.

Monitoring of inlet door position is necessary because the inlet doors form the barrier to air flow through the inlet ports of the ice condenser for normal unit operation. Failure of the lnlet Door Position Monitoring System requires an alternate OPERABLE monitoring system to be used to ensure that the ice condenser is not degraded.

APPLICABLE Proper operation of the inlet doors is necessary to mitigate the SAFETY consequences of a loss of coolant accident or a main steam line break ANALYSES inside containment. The lnlet Door Position Monitoring System, however, is not required for proper operation of the inlet doors, nor is it considered OPERABLE as an initial condition for a DBA. Hence, the lnlet Door Position Monitoring System is not a consideration in the analyses of DBAs. Based on the PRA Summary Report in References 2 and 3, the lnlet Door Position Monitoring System has not been identified as a significant risk contributor.

(continued)

Watts Bar - Unatz B 3.6-6 Revision I Technical Req uirements

L-19-023, Periodic Submission for Changes Made to the WBN Technical Specification Bases and Technical Requirements Manual

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