Text

Periodic Submission for Changes Made to Technical Specification Bases and Technical Requirement Manual
	ML14307A981
Person / Time
Site:	Watts Bar
Issue date:	11/03/2014
From:	Walsh K; Tennessee Valley Authority
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	Download: ML14307A981 (102)
	v • d • e

Tennessee Valley Authority, Post Office Box 2000 Spring City, Tennessee 37381 November 3, 2014 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, Unit 1 Facility Operating License No. NPF-90 NRC Docket No. 50-390

Subject:

WATTS BAR NUCLEAR PLANT (WBN) UNIT 1 - pERtODtC SUBMISSION FOR CHANGES MADE TO THE WBN TECHNICAL SPECIFICATION BASES AND TECHNICAL REQUIREMENTS MANUAL

References:

Tennessee Valley Authority (TVA) letter to the Nuclear Regulatory Commission (NRC) "Changes Made to the Technical Specifications Bases and Technical Requirements Manual" dated April29, 2013 The purpose of this letter is to provide the Nuclear Regulatory Commission (NRC) with copies of changes to the WBN Technical Specification (TS) Bases, through Revision '121, and WBN Technical Requirements Manual (TRM), through Revision 54, in accordance with WBN TS Section 5.6, "TS Bases Control Program," and WBN TRM Section 5.1, "Technical Requirements Control Program," respectively. These changes have been implemented at WBN during the period since WBN's last update (April 29, 2013) and 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 10 CFR 50.71(e). Per the provisions of 10 CFR 50.71(e), the Updated Final Safety Analysis Report will be provided in a separate letter. The WBN TS Bases and TRM updates for the table of contents and change pages are provided in the enclosures.

There are no new regulatory commitments in this letter. lf you have questions regarding this letter, please call Gordon Arent, Director of Watts Bar Site Licensing, at (423) 365-2004.

U.S. Nuclear Regulatory Commission Page 2 November 3, 2014 I certify the information provided accurately presents changes made since the last TS Bases and TRM update was submitted on April 29, 2013.

Respectfully, Kevin T. Walsh Site Vice President Watts Bar Nuclear Plant

Enclosures:

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

NRC Regional Administrator - Region ll NRC Senior Resident lnspector - Watts Bar Nuclear Plant, Unit 1 NRC Senior Resident lnspector - Watts Bar Nuclear Plant, Unit 2

U.S. Nuclear Regulatory Commission Page 3 November 3, 2014 cc (Enclosures):

U. S. Nuclear Regulatory Commission Region ll Marquis One Tower 245 Peachtree CenterAve., NE Suite 1200 Atlanta, Georgia 30303-1257 NRC Resident Inspector Unit 1 Watts Bar Nuclear Plant 1260 Nuclear Plant Road Spring City, Tennessee 37381 NRC Resident Inspector Unit 2 Watts Bar Nuclear Plant 1260 Nuclear Plant Road Spring City, Tennessee 37381

ENCLOSURE 1 WBN TECHNICAL SPECIFICATION BASES TABLE OF CONTENTS E1-1

TABLE OF CONTENTS B 3.1 REACTMryCONTROLSYSTEMS .....B 3.1-1 B 3.1 .1 SHUTDOWN MARGTN (SDM) Ta,s > 200"F ......B 3.1-1 B 3.1 .2 SHUTDOWN MARGIN (SDM) T"w < 200"F ......B 3.1-7 B 3.1.3 Core Reactivity.................!.!!.!!..ri.r..i:.... ............... B 3.1-12 B 3.1 .4 ModeratorTemperature Coefficient (MTC).......... ...................8 3.1-18 B 3.1.5 Rod Group Alignment 1imits........... ....................B 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..... ..............B 3.1.48 B 3.1 .9 PHYSICS TESTS Exceptions MODE 1 .................. ................8 3.1-55 B 3.1 .10 PHYSICSTESTSExceptionsMODE2.................. ................83.1$2 B 3.2 POWER DISTRIBUTION LIMITS ...........B 3.2-1 B 3.2.1 Heat Flux Hot Channel Factor (FO(Z)........ .......8 3.2-1 B 3.2.2 Nuclear Enthalpy Rise Hot Channel Factor (F*aH)......... ..................B 3.2-12 B 3.2.3 AXlAL FLUXDTFFERENCE (AFD).......... ..........B 3.2-19 B 3.2.4 QUADRANTPOWERTTLTRATTO(OPTR)........ ...................83.2-24 B 3.3 INSTRUMENTATION .........B 3.}1 B 3.3.1 Reac'torTrip System (RTS) lnstrumentation....... ....................8 3.&1 B 3.3.2 Engineered Safety Feafure Ac{uation Sptem (ESFAS) lnstrumentiation................. ....................8 3.3-&t B 3.3.3 PostAccident Monitoring (PAM) lnstrumentiation................ ...B 3.3-121 B 3.3.4 Remote Shutdown System ............8 3.T141 B 3.3.5 Loss of Power (LOP) Diesel Generator (DG)

Start lnstrumentation ...............8 3.3-147 B 3.3.6 Containment Vent lsolation lnsfumentiation ................ ........... B 3.$154 B 3.3.7 Control Room Emergency Ventilation Sptem (CREVS) Actuation lnsbumentation................ ................. B 3.&163 B 3.3.8 Auxiliary Building Gas Treatment System (ABGTS)

Actuation lnstrumentation................. ............B 3.+.171 (continued)

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TABLE OF CONTENTS (continued)

B 3.4 REACTOR COOl-ANT SYSTEM (RCS) B 3.+1 B 3.4.1 RCS Pressure, Tempemfure, and Flow Departure from Nucleate Boiling (DNB) 1imits........... ..8 3.+1 B 3.4.2 RCS Minimum Temperature for Critica1ity................. ..............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.+17 B 3.4.5 RCS Loops-MODE 3................... ....................83.+21 B 3.4.6 RCS Loops:MODE 4................... ....................B 3.+27 B 3.4.7 RCS Loops-.MODE 5, Loops Fi11ed............ .....B 3.4-33 B 3.4.8 RCS Loops-MODE 5, Loops Not Filled..... .....8 3.+38 B 3.4.9 Pressurizer. ................ B 3.441 B 3.4.10 Pressurizer Safety Va1ves......... .....8 3.44 B 3.4.11 Pressurizer Power Operated Relief Valves (PORVS)...... .................B 3.+51 B 3.4.12 ColdOverpressureMitigationSystem(COMS)....... ...............83.+58 B 3.4.13 RCS Operational LEAI(AGE.................. .............8 3.+74 B 3.4.14 RCS Pressure lsolation Valve (PlV) Leakage....... ..................B 3.+81 B 3.4.15 RCS Leakage Detec'tion lnstumentation................ ................B 3.447 B 3.4.16 RCS Spec1ficActivity......... .............B 3.+93 B 3.4.17 Steam Generator (SG) Tube lntegrity ................ B 3.+99 B 3.5 EMERGENCY CORE COOLTNG SYSTEMS (ECCS) ....................B 3.S1 B 3.5.1 Accumulators .............8 3.$1 B 3.5.2 ECCS-Operating............ .............8 3.$10 B 3.5.3 ECCS-Shutdown.......... ..............B 3.*20 B 3.5.4 Retueling WaterStorage Tank (RWST)....... ......83.5-24 B 3.5.5 Seallnjection F1ow............. .............B 3.S31 B 3.6 CoNTAINMENT SYSTEMS .................. .B 3.&1 B 3.6.1 Containment ...............8 3.G1 B 3.6.2 Containment Air Locks ................... B 3.6-6 B 3.6.3 Containment lsolation Valves ........B 3.G14 B 3.6.4 Contiainment Pressure...... ..............B 3.G28 B 3.6.5 Containment Air Temperature.............. ............... B 3.G31 B 3.6.6 B 3.6.7 Hydrogen Recombiners - Deleted .B 3.il3 B 3.6.8 Hydrogen Mitigation System (HMS).......... .........B 3.649 B 3.6.9 Emergency Gas Treatment System (EGTS) ..... B 3.G55 B 3.6.10 Air Retum System (ARS).......... .....B 3.&60 B 3.6.11 lce Bed ..8 3.665 B 3.6.12 lce Condenser Doors....... ...............8 3.674 B 3.6.13 Divider Banier lntegrity........ ...........8 3.S84 B 3.6.14 Containment Recirculation Drains .B 3.&90 B 3.6.15 Shield 8ui1din9.................. ..............B 3.S95 (continued)

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TABLE OF CONTENTS (continued)

B 3.7 PLANT SYSTEMS ..............B 3.7-1 B 3.7.1 Main Steam SafetyValves (MSSVS)... ...............B 3.7-1 B 3.7.2 Main Steam lsolation Valves (MSlVs) ................8 3.7-7 B 3.7.3 Main Feedwater lsolation Valves (MFlVs) and Main Feedwater Regulation Valves (MFRVS) and Associated Bypass Va1ves.......... ..........B 3.7-13 B 3.7.4 Atmospheric Dump Valves (ADVs)......... ............B 3.7-20 B 3.7.5 Auxiliary Feedwater (AFW) System......... ...........B 3.7-24 B 3.7.6 Condensate Storage Tank (CST). .83.7-U B 3.7.7 Component Cooling System (CCS)........... .........B 3.7-38 B 3.7.8 Essential Raw Cooling Water (ERCW) System........ ..............8 3.743 B 3.7.9 Ultimate Heat Sink (UHS)........... ....B 3.748 B 3.7.10 ConholRoomEmeryencyVentilationSystem(CREVS) .......B3.7-51 B 3.7.11 Contrcl Room Emergency Air Temperature B 3.7.12 Auxiliary Building Gas Treatment System (ABGTS)...... .........8 3.7$2 B 3.7.13 Fuel Storage PoolWater Leve1............ ...............B 3.768 B 3.7.14 Secondary Specific Activity................. ................8 3.7-71 B 3.7-15 Spent Fuel Assembly Storage .......8 3.7-75 B 3.8 ELECTRICAL POWER SYSTEMS... ......8 3.8.1 B 3.8.1 AC Sources-Operating ...............8 3.&1 B 3.8.2 AC Sources-Shutdown.... ...........B 3.&37 B 3.8.3 Diesel FuelOil, Lube Oil, and Starting Air................ ...............B 3.&43 B 3.8.4 DC Sources-Operating.... ...........B 3.&54 B 3.8.5 DC Sources-Shutdown.... ...........B 3.&70 B 3.8.6 Battery Cell Parameters................. 83.8-74 B 3.8.7 lnverters-Operating........ .............B 3.&81 B 3.8.8 lnverters-Shutdown ....................8 3.&85 B 3.8.9 Disfibution Systems-Operating .B 3.&89 B 3.8.10 Distibution Systems-Shutdown....... ...............8 3.8-99 B 3.9 REFUELTNG OPERATlONS.................. .B 3.9-1 B 3.9.1 Boron Concenfation......... ..............8 3.9-1 B 3.9.2 Unborated Water Source lsolation Valves .........B 3.9.5 B 3.9.3 Nuclear lnstrumentiation............... ..8 3.9-8 B 3.9.4 Deleted ..B 3.9-12 B 3.9.5 Residual Heat Removal (RHR) and Coolant Circulation-High Water Level............ .......B 3.9-17 B 3.9.6 Residual Heat Removal (RHR) and Coolant Circrulation

- Low Water Level ............ ........ B 3.9-21 B 3.9.7 Retueling Cavity Water Level. ........8 3.9-25 B 3.9.8 Deleted ..B 3.9-29 B 3.9.9 SpentFuelPoolBoronConcenbation................... ..................83.9-33 B 3.9.10 DecayTime............ ....8 3.9-35 Watts Bar-Unit 1 Revision 1 19

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

ContingencyActions........ B 3.&36 B 3.8.9.1 AC and DC Electrical Power Distribution Systems.-..... .B 3.&98 a

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LIST OF FIGURES Figure No. Title B 2.1.1-1 Reactor Core Safety Limib vs Boundary of Prctection ...................82.0-7 83.1.7-1 Confol Bank lnsertion vs Percent RTP............. ..........8 3.147 B 3.2.'l-1 K(z) - Normalized Fq(z) as a Function of Core Hei9ht.......... ...............B 3.2-11 83.2.*'l AXIAL FLUX DIFFERENCE Acceptable Operation Limits as a Func'tion of RATED THERMAL POWER...... ..................8 3.2-23 Watts Bar-Unit 1

LIST OF ACRONYMS (Page 1 ot 2)

Acronym Title ABGTS Auxiliary Building Gas Treatment System ACRP Auxiliary Control Room Panel ASME American Society of Mechanical Engineers AFD Axial Flux Difference AFW Auxiliary Feedwater System ARO All Rods Out ARFS Air Retum 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 lsolation Valve MFRV Main Feedwater Regulation Valve MSIV Main Steam Line lsolation Valve MSSV Main Steam Safety Valve MTC Moderator Temperature Coeffi cient NMS Neutron Monitoring System ODCM Offsite Dose Calculation Manual PCP Process Control Program PDMS Power Distribution Monitoring System P!V 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 a

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LIST OF ACRONYMS (Page 2 ot 2)

Acronvm 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 tt Watts Bar-Unit 1 vI

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)ry 115 B 3.1-16 0 xvi 119 B 3.1-17 0 B 2.0-1 0 B 3.1-18 32 B 2.0-2 59 B 3.1-19 32 B 2.0-3 0 B 3.1-20 32 B2.04 59 B 3.1-21 32 B 2.0-5 108 B 3.1-22 32 B 2.0 59 B 3.1-23 0 B 2.0-7 0 B 3.1-24 51 B 2.0-8 0 B 3. 1-25 51 B 2.0-9 0 B 3.1-26 0 B 2.0-1 0 0 B 3.1-27 104 B 2.0-11 108 B 3.1-28 0 B 2.0-12 0 B 3.1-29 0 B 3.0-1 55 B 3.1-30 104 B 3.0-2 0 B 3.1-31 0 B 3.0-3 0 B 3.1-32 0 B 3.04 68 B 3.1-33 0 B 3.0-5 68 B 3.1-y 0 B 3.0 68 B 3.1-35 51 B 3.0-7 0 B 3.1-36 0 B 3.0-8 103 B 3. 1-37 0 B 3.0-9 0 B 3.1-38 0 B 3.0-10 0 B 3.1-39 0 B 3.0-1 1 114 B 3.140 51 B 3.0-12 53 B 3.141 0 B 3.0-13 68 B 3.142 0 B 3.0-14 68 B 3.143 0 B 3.1-1 0 B3. 14 0 Watts Bar-Unit 1 vilt Revision 121

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TECHNICAL SPECIFIGATION BASES . REVISION LISTING (This listing is an administrative tool maintained by WBN Licensing and may be updated without formally revising the Technical Specification Bases Table-of-Contents)

REVISIONS ISSUED SUBJECT NPF-20 1 1-09-95 Low Power Operating License Revision 1 12-08-95 Slave Relay Testing NPF-90 02-07-96 Full Power Operating License Revision 2 (Amendment 1) 12-08-95 Turbine Driven AFW Pump Suction Requirement Revision 3 03-27-96 Remove Cold Leg Accumulator Alarm Setpoints Revision 4 (Amendment 2) 06-13-96 lce Bed Surveillance Frequency And Weight Revision 5 07-03-96 Containment Airlock Door lndication Revision 6 (Amendment 3) 09-09-96 Ice Condenser Lower lnlet Door Surveillance Revision 7 09-28-96 Clarification of COT Frequency for COMS Revision 8 11-21-96 Admin Control of Containment Iso!. Valves Revision 9 04-29-97 Switch Controls For Manual Cl-Phase A Revision 10 (Amendment 5) 05-27-97 Appendix-J, Option B Revision 11 (Amendment 6) 07-28-97 Spent Fuel Pool Rerack Revision 12 09-10-97 Heat Trace for Radiation Monitors Revision 13 (Amendment 7) 09-1 1-97 Cycle 2 Core Reload Revision 14 10-10-97 Hot Leg Recirculation Timeframe Revision 15 02-12-98 EGTS Logic Testing Revision 16 (Amendment 10) 06-09-98 Hydrogen Mitigation System Temporury Specification Revision 17 07-31-98 SR Detectors (Visual/audible indication)

Revision 18 (Amendment 1 1) 09-09-98 Relocation of F(Q) Penalty to COLR Revision 19 (Amendment 12) 10-19-98 Online Testing of the Diesel Batteries and Performance of the 24 Hour Diesel Endurance Run Watts Bar-Unit 1 xvii

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 20 (Amendment 13) 10-26-98 Clarification of Suleillance 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 AlIowance 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 Detete 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 lntegrated Computer System (lCS). Delete Reference to ERFDS as a redundant input signal.

Revision 29 03-1 3-00 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 . REVISPN 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)

REV!SIONS 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 Surueillance Testing Revision 36 (Amendments 22 and 25) 08-23-00 Revision of lce Condenser sampling and flow channel surveillance requirements Revision 37 (Amendment 26) 09-08-00 Administrative Controls for Open Penetrations During Refueling Operations Revision 38 09-17-00 SR 3.2.1 .2 was revised to reflect the area of the core that will be flux mapped.

Revision 39 (Amendments 21and 28) 09-13-00 Amendment 21 - lmplementation of Best Estimate LOCA analysis.

Amendment 28 - Revision of LGO 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 Requirements for Pressunzer PORVs Revisio n 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 xix

TECHNICAL SPECIFICATION BASES . REVISION LISTING (This listlng is an adminislrative too! maintained by WBN Licenslng and may be updated without formally rcvising the Technical Specification Bases Table-of-Contents)

REVISIONS ISSUED SUBJECT Revision 47 (Amendment 38) 03-01-02 RCS operational LEAKAGE and SG Altemate Repair Criteria for Axial Outside Diameter Stress Conosion 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 PoshAccident Sampling System (PASS) requirements from Section 5.7 .2.6 of the Technical Specifi cations.

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 (Amendmenl42) 01-24-03 Revise SR 3.0.3 for Missed Surveillances Revision 54 (Amendment 43) 05-01-03 Exigent TS SR 3.5.2.3 to delete Sl Hot Leg lnjection lines from SR until UlC5 outage.

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

conect 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 clarifi cation 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 Conelation in 82.1.1 and 83.2.12 for Robust FuelAssembly (RFA)-2.

Revision 60 (Amendment 47) 1&06-03 RCS Flow Measurement Using Elbow Tap Flow Meters (Revise Table 3.3.1-1(10) &

sR 3.4.1.4).

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TECHNICAL SPECIFIGATION 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 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-o3-19)

Revision 63 12-08-03 The contingency actions listed in Bases Tabf e 3.8. 1-2 were reworded to be consistent with the NRC Safety Evaluation that approved Tech Spec Amendment 39.

Revision 64 (Amendment 50) 03-23-04 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 clariff 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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TECHNIGAL SPECIFICATION BASES . REVISION LISTING (This listing is an administrative tool maintained by WBN Licensing and may be updated without formally revising the Technical Specification Bases Table-of-Contents)

REVISIONS ISSUED SUBJECT Revision 68 (Amendment 55 and 56) 03-22-05 Change MSLB primary to secondary leakage from 1 gpm to 3 gpm (WBN-TS-03-14).

Revision 69 (Amendment 54) 04-04-05 Revised the use of the terms inter-tier and inter-rack in the Bases for SR 3.8.4.10.

Revision 70 (Amendment 58) 10-17-05 Alternate monitoring process for a failed Rod Position Indicator (RPI) (TS-03-12).

Revision 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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TEGHNICAL SPECIFICATION BASES . REVISION LISTING (This listing is an administrative tool malntalned by WBN Licensing and may be updated without formally revising the Technical Speclflcation Bases Tableof{ontents)

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 s1370).

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 s1370).

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 11-03-06 Steam Generator Nanow Range Level

64) lndication lncreased from 6Yo to 32o/o (WBN-T5-05-06) Bases Sections 3.4.5, 3.4.6, and 3.4.7.

Revision 80 11-08-06 Revised the Bases for SR 3.5.2.8 to clariff that inspection of the containment sump strainer constitutes inspection of the trash rack and the screen functions.

Revision 81 (Amendment 62) 1 1-15-06 Revised the Bases for SR 3.6.11 .2, 3.6.1 1 .3, and 3.6.11.4 to address the lncrease lce Weight in lce Condenser to Support Replacement Steam Generators WBN-TS-0$0e) [SGRP]

Revision 82 (Amendment 65) 11-17-06 Steam Generator (SG) Tube lntegrity (wBN-rs-os-10) [SGRPI Revision 83 11-20-06 Updated Surveillance Requirement (SR) 3.6.6.5 to clariff that the number of unobstructed spray nozzles is defined in the design bases.

Revision 84 11-30-06 Revised Bases 3.6.9 and 3.6.15 to show the operation of the EGTS when annulus pressure is not within limits.

Revision 85 03-22-07 Revised Bases 3.6.9 and 3.6.15 in accordance with TACF 1-07-0002-065 to clariff the operation of the EGTS.

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TECHNICAL SPECIFICATION BASES . REVISPN 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)

REV!SIONS ISSUED SUBJECT Revision 86 01-31-08 Figure 3.7 .15-1 was deleted as part of Amendment 40. A reference to the figure in the Bases for LCO 3.9.9 was not deleted at the time Amendment 40 was incorporated into the Technical Specifications. Bases Revision 86 corrected this error (refer to PER 130e44).

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 signal.

Revision 88 (Amendment 67) 03-06-08 Technical Specification Amendment 67 increased the number of TPBARS from 240 to 400.

Revision 89 (Amendment 66) 05-01-08 Update of Bases to be consistent with the changes made to Section 5.7 .2.11 of the Technical Specifications to reference the ASME Operation and Maintenance Code Revision 90 (Amendment 68) 10-02-08 lssuance of amendment regarding Reactor Trip System and Engineered Safety Features Actuation System completion times, bypass test times, and su rvei lance test intervals I

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 3.4.1 5, "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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I TECHNIGAL SPECIFICATION BASES . REVISIoN LISTING (This listing is an administrative tool maintained by WBN Licensing and may be updated without formally revising the Technical Specification Bases Table-of-Contents)

REVISIONS ISSUED SUBJECT Revision 95 03-05-2009 Corrected an error in SR 3.3.2.6 which referenced Function 6.9 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 appficability 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 - Operating" to reflect the installation of the Unit 2 inverters.

Revision 98 (Amendments 77,79, & 10-05-2009 Amendment 77 revised the number of

81) TPBARS that may be loaded in the core from 400 to 704.

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

Amendment 81 revised the allowed outage time of Action B of LCO 3.5.1 from t hour to 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 11-17-2009 Bases Revision 100 revises the LCO description of the Containment Spray System to clarify that transfer to the containment sump is accomplished by manual actions.

Revision 101 02-09-2414 Bases Revision 101 implemented DCN 52216-A that 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 (CIA) signal.

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

Revision 142 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.

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TECHNICAL SPECIFICATION BASES . REVISPN LISTING (This listing is an administrative tool maintained by WBN Licensing and may be updated without formally revising the Technical Specification Bases Table-of-Contents)

REVISIONS ISSUED SUBJECT Revision 103 04-05-2010 Bases Revision 103 implemented NRC guidance "Application of Generic Letter 80-30" which allows a departure from the single failure criterion where a non-TS support system has two 100o/o capacity subsystems, each capable of supporting the design heat Ioad 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 clariff 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 30,2A12.

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.

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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 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 1 4 ltGilgm and refers to the controls being placed in AOI-28.

Revision 112 05-24-2011 DCN 55076 replaces the existing four 125-Vdc DG Battery Chargers with four sets of redundant new battery charger assemblies.

Revision 1 13 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 1 15 12-21-2011 Revises severa! 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 .10, 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 I-131 spike limit and the allowable value for control room air intake radiation monitors.

Revision 1 19 08-17-2013 Revises TS Bases 3.3.6, 3.3.8, 3.7.12, 3.7 .13, 3.9.4, 3.9.7, 3.9.8, and adds TS Bases 3.9.10 to reflect selective implementation of the Alternate Source Term methodology for the analysis of Fuel Handling Accidents (FHAs) and make TS Bases consistent with the revised FHA dose analysis.

Revision 120 01-23-2014 Revised the References to TS Bases 3.1 .9, PHYSICS TESTS Exceptions - Mode1, to document NRC approval of WCAP 12472-P-Watts Bar-Unit 1 xxvii

TECHNIGAL 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)

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 a references in TS Bases 3.7.1 for consistency with changes to the TS Bases 3.7.1 references approved in Revision 89.

Watts Bar-Unit 1 xxvilt

ENCLOSURE 2 WBN TECHNIGAL SPECIFICATION BASES CHANGED PAGES E2-1

PHYSICS TESTS Exceptions - MODE 1 3.1 .9 BASES SURVEILLANCE SR 3.1.9.4 (continued)

REQU!REMENTS

f. Samarium concentration; and

g. Deleted Using the ITC accounts for Doppler reactivity in the calculation because the reactor is subcritical, and the fueltemperature will be changing at the same rate as the RCS.

The Frequency of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is based on the generally slow change in required boron concentration and on the low probability of an accident without the required SDM.

REFERENCES 1. Title 10, Code of Federal Regulations, Part 50, Appendix B, "Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants."

2. Title 10, Code of Federal Regulations, Part 50.59, "Changes, Tests, and Experiments."

3. Regulatory Guide 1 .68, Revision 2, "lnitial Test Programs for Water-Cooled Nuclear Power Plants," August 1978.

4. ANSI/ANS-19.6.1, "Reload Startup PHYSICS TESTS for Pressurized Water Reactors," American National Standards lnstitute.

5. WCAP -9272-P-A, "Westinghouse Reload Safety Evaluation Methodology Report," July 1985.

6. Watts Bar FSAR, Section 14.2, "Test Program."

7. WCAP-11618, "MERITS Program - Phase Il, Task 5, Criteria Application," dated Novembet 1987, including Addendum 1, April 1 989.

8. WCAP-12472-P-A,"BEACON Core Monitoring and Operations Support System," August 1994 (Addendum 1-A, January 2000, Addendum 4-4, September 2012).

Watts Bar-Unit 1 B 3.1 -61 Revision 40, 104, 120 Amendment 82

Containment Vent lsolation lnstrumentation B 3.3.6 B 3.3 INSTRUMENTATION B 3.3.6 Containment Vent lsolation lnstrumentation BASES BACKGROUND Containment Vent lsolation (CVl) lnstrumentation closes the containment isolation valves in the Containment Purge System. This action isolates the containment atmosphere from the environment to minimize releases of radioactivity in the event of an accident. The Reactor Building Purge System may be in use during reactor operation and with the reactor shutdown.

Containment vent isolation is initiated by a safety injection (Sl) signal or by manual actuation. The Bases for LCO 3.3.2, "Engineered Safety Feature Actuation System (ESFAS) lnstrumentation,' discuss initiation of Sl signals.

Redundant and independent gaseous radioactivity monitors measure the radioactivity levels of the containment purge exhaust, each of which will initiate its associated train of automatic Containment Vent lsolation upon detection of high gaseous radioactivity.

The Reactor Building Purge System has inner and outer containment isolation valves in its supply and exhaust ducts. This system is described in the Bases for LCO 3.6.3, "Containment lsolation Valves.'

(continued)

Watts Bar-Unit 1 B 3.3-154 Revision 43, 87 , 1 10, 1 19 Amendment 92

Containment Vent lsolation lnstrumentation B 3.3.6 BASES (continued)

APPLICABLE The containment isolation valves for the Reactor Building Purge System SAFETY ANALYSES close within six seconds following the DBA. The containment vent isolation radiation monitors act as backup to the Sl signal to ensure closing of the purge air system supply and exhaust valves. Containment isolation in tum ensures meeting the containment leakage rate assumptions of the safety analyses, and ensures that the calculated accidental offsite radiological doses are below 10 CFR 100 (Ref. 1) limits.

The Containment Vent lsolation instrumentation satisfies Criterion 3 of the NRC Policy Statement.

(continued)

Watts Bar-Unit 1 B 3.3-1544 Revision 43, 87 , 1 10, J 19 Amendment 92

Containment Vent lsolation lnstrumentation B 3.3.6 BASES LCO 3. Containment Radiation (continued)

The LCO specifies two required channels of radiation monitors to ensure that the radiation monitoring instrumentation necessary to initiate Containment Vent lsolation remains OPERABLE.

For sampling systems, channel OPERABILITY involves more than OPERABILITY of the channelelectronics. OPERABILITY may also require conect valve lineups and sample pump operation, as well as detector OPERABILITY, if these supporting features are necessary for trip to occur under the conditions assumed by the safety analyses.

Table 3.3.6-1 specifies the Allowable Values (AVs) for the Contrainment Purge Exhaust Radiation Monitors. This AV is based on expected concentrations for a small break LOCA, which is more restrictive than the 10 CFR 100limits. The specified AV is more @nservative than the analytical limit assumed in the safety analysis in order to account for instrument uncertainties appropriate to the trip function. The actual nominal Trip Setpoint is normally still more conservative than that required by the AV. lf the setpoint does not exceed the applicable AV, the radiation monitor is considered OPERABLE.

4. Safetv lniection (Sl)

Refer to LCO 3.3.2. Function 1, for all initiating Functions and requirements.

APPLICABILITY The Manual lnitiation, Automatic Actuation Logic and Actuation Relays, Safety lnjection, and Containment Radiation Functions are required OPERABLE in MODES 1 , 2, 3, and 4. Under these conditions, the potential exists for an accident that could release significant fission product radioactivity into containment. Therefore, the Containment Vent lsolation lnstrumentration must be OPERABLE in these MODES. See additionaldiscussion in the Background and Applicable Safety Analysis sections.

(continued)

Watts Bar-Unit 1 B 3.3-156 Revision 45, 87, 93, 1 19 Amendment 35,74,92

Containment Vent lsolation lnstrumentation B 3.3.6 BASES APPLICABILITY While in MODES 5 and 6, the Containment Vent lsolation lnstrumentation need (continued) not be OPERABLE since the potentialfor radioactive releases is minimized and operator action is sufficient to ensure post accident offsite doses are maintained within the limits of Reference '1.

ACTIONS The most common cause of channel inoperability is outright failure or drift sufficient to exceed the tolerance allowed by unit specific calibration procedures.

Typically, the drift is found to be small and results in a delay of actuation rather than a total loss of function. lf the Trip Setpoint is less conservative than the tolerance specified by the calibration procedure, the channel must be declared inoperable immediately and the appropriate Condition entered.

A Note has been added to the ACTIONS to clariff the application of Completion Time rules. The Conditions of this Specification may be entered independently for each Function listed in Table 3.3.6-1. The Completion Time(s) of the inoperable channel(s/train(s) of a Function will be tracked separately for each Function starting from the time the Condition was entered for that Function.

A.1 Condition A applies to the failure of one containment purge isolation radiation monitor channel. Since the two containment radiation monitors are both gaseous detectors, failure of a single channel may result in loss of the redundancy.

Consequently, the failed channel must be restored to OPERABLE status. The 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months allowed to restore the affected channel is justified by the low likelihood of events occurring during this interval, and recognition that one or more of the remaining channels will respond to most events.

(continued)

Watts Bar-Unit 1 B 3.3-157 Revision 1 19 Amendment 92

Containment Vent lsolation lnstrumentation B 3.3.6 BASES ACTIONS 8.1 (continued)

Condition B applies to all Containment Vent lsolation Functions and addresses the train orientation of the Solid State Protection System (SSPS) and the master and slave relays for these Functions. lt also addresses the failure of multiple radiation monitoring channels, or the inability to restore a single failed channel to OPERABLE status in the time allowed for Required Action A.1.

lf a train is inoperable, multiple channels are inoperable, or the Required Action and associated Completion Time of Condition A are not met, operation may continue as long as the Required Action for the applicable Conditions of LCO 3.6.3 is met for each valve made inoperable by failure of isolation instrumentation. A Note has been added above the Required Actions to allow one train of actuation logic to be placed in bypass and to delay entering the Required Actions for up to four hourc to perform surveillance testing provided the other train is OPERABLE. The 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months allowance is consistent with the Required Actions for actuation logic trains in LCO 3.3.2, "Engineered Safety Features Actuation System lnstrumentation" and allows periodic testing to be conducted while at power without causing an actual actuation. The delay for entering the Required Actions relieves the administrative burden of entering the Required Actions for isolation valves inoperable solely due to the performance of surveillance testing on the actuation logic and is acceptable based on the OPERABILIry of the opposite train.

(continued)

Watts Bar-Unit 1 B 3.3-158 Revision 1 19 Amendment 92

Containment Vent lsolation lnstrumentation B 3.3.6 BASES SURVEILLANCE A Note has been added to the SR Table to clarify that Table 3.3.6-1 determines REQUIREMENTS which SRs apply to which Containment Vent lsolation Functions.

sR 3.3.6.1 Performance of the CHANNEL CHECK once every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ensures that a gross failure of instrumentation has not occuned. A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a similar parameter on other channels. lt is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value.

Significant deviations between the two instrument channels could be an indication of excessive instrument drift in one of the channels or of something even more serious. A CHANNEL CHECK will detect gross channel failure; thus, it is key to verifying the instrumentation continues to operate properly between each CHANNEL CALIBRATION.

(continued)

Watts Bar-Unit 1 B 3.3-159 Revision 45, 1 19 Amendment 35, 92

ABGTS Actuation Instrumentation B 3.3.8 B 3.3 INSTRUMENTATION B 3.3.8 Auxiliary Building Gas Treatment (ABGTS) Actuation lnstrumentation BASES BACKGROUND The ABGTS ensures that radioactive materials in the fuel building atmosphere following a loss of coolant accident (LOCA) are filtered and adsorbed prior to exhausting to the environment. The system is described in the Bases for LCO 3.7.12, "Auxiliary Building Gas Treatment System." The system initiates filtered exhaust of air from the fuel handling area, ECCS pump rooms, and penetration rooms automatically following receipt of a Containment Phase A lsolation signal. lnitiation may also be performed manually as needed from the main control room.

There are a total of two channels, one for each train. A Phase A isolation signal from the Engineered Safety Features Actuation System (ESFAS) initiates auxiliary building isolation and starts the ABGTS. These actions function to prevent exfiltration of contaminated air by initiating filtered ventilation, which imposes a negative pressure on the Auxiliary Building Secondary Contrainment Enclosure (ABSCE).

(continued)

Watts Bar-Unit 1 B 3.3-171 Revision 87, 1 10, 1 19 Amendment 92

ABGTS Actuation I nstrumentation B 3.3.8 BASES (continued)

APPLICABLE The ABGTS ensures that radioactive materials in the ABSCE atmosphere SAFETY ANALYSES following a LOCA are filtered and adsorbed prior to being exhausted to the environment. This action reduces the radioactive content in the auxiliary building exhaust following a LOCA so that offsite doses remain within the limits specified in 10 CFR 100 (Ref. 1).

The ABGTS Actuation lnstrumentation satisfies Criterion 3 of the NRC Policy Statement.

(continued)

Watts Bar-Unit 1 B 3.3-172 Revision 87, 1 10, 1 19 Amendment 92

ABGTS Actuation lnstrumentation B 3.3.8 BASES (continued)

LCO The LCO requirements ensure that instrumentiation necessary to initiate the ABGTS is OPEMBLE.

1. Manual lnitiation The LCO requires two channels OPEMBLE. The operator can initiate the ABGTS at any time by using either of two switches in the control room. This action will cause actuation of all components in the same manner as any of the automatic actuation signals.

The LCO for Manual lnitiation ensures the proper amount of redundancy is maintained in the manual actuation circuitry to ensure the operator has manual initiation capability.

Each channel consists of one hand switch and the interconnecting wiring to the actuation logic relays.

2. Deleted (continued)

Watts Bar-Unit 1 B 3.3-173 Revision 1 19 Amendment 92

ABGTS Actuation I nstrumentation B 3.3.9 BASES LCO 3. Containment Phase A lsolation (continued)

Refer to LCO 3.3.2, Function 3.a, for all initiating Functions and requirements.

APPLICABILITY The manual ABGTS initiation must be OPERABLE in MODES 1, 2,3, and 4 to ensure the ABGTS operates to remove fission products associated with leakage after a LOCA. The Phase A ABGTS Actuation is also required in MODES 1,2,3, and 4 to remove fission products caused by post LOCA Emergency Core Cooling Systems leakage.

While in MODES 5 and 6, the ABGTS instrumentation need not be OPERABLE.

See additional discussion in the Background and Applicable Safety Analysis sections.

ACTIONS The most common cause of channel inoperability is outright failure or drift sufficient to exceed the tolerance allowed by unit specific calibration procedures.

Typically, the drift is found to be small and results in a delay of actuation rather than a total loss of function. lf the Trip Setpoint is less conservative than the tolerance specified by the calibration procedure, the channel must be declared inoperable immediately and the appropriate Condition entered.

A Note has been added to the ACTIONS to clarifu the application of Completion Time rules. The Conditions of this Specification may be entered independently for each Function listed in Table 3.3.8-1 in the accompanying LCO. The Completion Time(s) of the inoperable channel(s/train(s) of a Function will be tracked separately for each Function starting from the time the Condition was entered for that Function.

(continued)

Watts Bar-Unit 1 B 3.3-174 Revision 87, 1 19 Amendment 92

ABGTS Actuation lnstrumentation ACTIONS (continued)

Condition A applies to the actuation logic train function from the Phase A lsolation and the manual function. Condition A applies to the failure of a single actuation logic train or manual channel. lf one channel or train is inoperable, a period of 7 days is allowed to restore it to OPEMBLE status. lf the train cannot be restored to OPERABLE stiatus, one ABGTS train must be placed in operation.

This accomplishes the actuation instrumentation function and places the unit in a conservative mode of operation. The 7 day Completion Time is the same as is allowed if one train of the mechanical portion of the system is inoperable. The basis for this time is the same as that provided in LCO 3.7.12.

8.1.1.8.1.2.8.2 Condition B applies to the failure of two ABGTS actuation logic signals from the Phase A lsolation or two manual channels. The Required Action is to place one ABGTS train in operation immediately. This accomplishes the actuation instrumentation function that may have been lost and places the unit in a conservative mode of operation. The applicable Conditions and Required Actions of LCO 3.7.12 must also be entered for the ABGTS train made inoperable by the inoperable actuation instrumentation. This ensures appropriate limits are placed on train inoperability as discussed in the Bases for lco 3.7.12.

Altematively, both trains may be placed in the emergency radiation protection mode. This ensures the ABGTS Function is performed even in the presence of a single failure.

Watts Bar-Unit 1 B 3.3-175 Revision 1 19 Amendment 92

ABGTS Actuation I nstrumentation B 3.3.8 BASES ACTIONS C.1 and C.2 Condition C applies when the Required Action and associated Completion Time for Condition A or B have not been met and the plant is in MODE 1,2,3, ot 4.

The plant must be brought to a MODE in which the LCO requirements are not applicable. To achieve this status, the plant must be brought to MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and 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 A Note has been added to the SR Table to clariff that Table 3.3.8-1 determines REQUIREMENTS which SRs apply to which ABGTS Actuation Functions.

sR 3.3.8.1 SR 3.3.8.1 is the performance of a TADOT. This test is a check of the manual actuation functions and is performed every 18 months. Each manualactuation function is tested up to, and including, the relay coils. ln some instances, the test includes actuation of the end device (e.9., pump stiarts, valve cycles, etc.). The Frequency is based on operating experience and is consistent with the typical industry refueling cycle.

The SR is modified by a Note that excludes verification of setpoints during the TADOT. The Functions tested have no setpoints associated with them.

REFERENCES 1. Title 10, Code of Federal Regulations, Part 100.1 1 , "Determination of Exclusion Area, Low Population Zone, and Population Center Distance."

Watts Bar-Unit 1 B 3.3-176 Revision 1 19 Amendment 92

Page lntentionally Left Blank Watts Bar-Unit 1 B 3.3- 177 Revision 1 19 Amendment 92

Page lntentionally Left Blank Watts Bar-Unit 1 B 3.3-178 Revision 1 19 Amendment 92

t MSSVs B 3.7.1 BASES ACTIONS A.1 (continued) ln the case of only a single inoperable MSSV on one or more steam generators a reactor power reduction alone is sufficient to limit primary side heat generation such that overpressurization of the secondary side is precluded for any RCS heatup event. Furthermore, for this case there is sufficient total steam flow capacity provided by the turbine and remaining OPERABLE MSSVs to preclude overpressuration in the event of an increased reactor power due to reactivity insertion, such as in the event of an uncontrolled RCCA bank withdrawal at power. Therefore, Required Action A.1, requires an appropriate reduction in reactor power within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

The maximum THERMAL POWER corresponding to the heat removalcapacity of the remaining OPERABLE MSSVs is determined using a conservative heat balance between the reactor coolant system heat generation and the steam relief through the OPEMBLE MSSVS, as shown below and described in the attachment to Reference 5:

Allowable THEUMAL POWER Levet (%o) : 100W where: w" = Minimum total steam relief capacity of the OPERABLE MSSVs on any one steam generator, in lbm/sec.

hr = heat of vaporization at the highest MSSV full-open pressure, in Btu/lbm.

O= NSSS power rating of the plant (includes reactor coolant pump heat) in MlA/t.

J( = Unitconversionfactor: g47.82Btu/sec/MW.

Note: The values in Specification 3.7.1 include an allowance for instrument and channel uncertainties to the allowable RTP obtained with this algorithm.

(continued)

Watts Bar-Unit 1 B 3.7-4 Revision 31 , 121 Amendment 19

BASES ACTIONS B.1 and 8.2 (continued) ln the case of multiple inoperable MSSVs on one or more steam generators, with a reactor power reduction alone there may be insufficient total steam flow capacity provided by the turbine and remaining OPERABLE MSSVs to preclude overpressurization in the event of an increased reactor power due to reactivity insertion, such as in the event of an uncontrolled RCCA bank withdrawal at power. The 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Completion Time for Required Action 8.1 is consistent with A.1. An additional 32 hours3.703704e-4 days 0.00889 hours 5.291005e-5 weeks 1.2176e-5 months is allowed in Required Action B.2 to reduce the setpoints. The Cgmpletion Time of 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months is based on a reasonable time to conect the MSSV inoperability, the time required to perform the power reduction, operating experiehce in resetting all channels of a protective function, and on the low probability of the occurence of a transient that could result in steam generator overpressure during this period.

The maximum T ERMAL POWER corresponding to the heat remova! capacity of the remaining PERABLE MSSVs is determined using a conservative heat balance calculati as described above (Action A.1) and in the attachment to Reference 5. Th values in Specification 3.7.1 include an allowance for instrument and c annel uncertainties to the allowable RTP obtained with this algorithm.

Required Action .2 is modified by a Note, indicating that the Power Range Neutron Flux-Hig reactor trip setpoint reduction is only required in MODE 1. ln MODES 2 and 3 e reactor protection system trips specified in LCO 3.3.1, "Reactor Trip Sy m lnstrumentation," provide sufficient protection.

C.1 and C.2 If the Required A ons are not completed within the associated Completion Time, or if one or ore steam generators have > 4 inoperable MSSVS, the plant must be placed i a MODE in which the LCO does not apply. To achieve this status, the plant ust be placed in at least MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and in MODE 4 within 1 hours. The allowed Completion Times are reasonable, based on operating e ience, to reach the required plant conditions from full power conditions in an erly manner and without challenging plant systems.

SURVEILI.ANCE sR 3.7.1 .1 REQUIREMENTS This SR verifies e OPERABIL!ry of the MSSVs by the verification of each MSSV Iift setpoi in accordance with the lnservice Testing Program. The ASME OM Code (Ref. 4 requires that safety and relief valve tests be (continued)

Watts Bar-Unit 1 B 3.7-5 Revision 31, 89, 121 Amendment 19, 66

ABGTS B 3.7.12 B 3.7 PLANT SYSTEMS B 3.7.12 Auxiliary Building Gas Treatment System (ABGTS)

BASES BACKGROUND The ABGTS filters airbome radioactive particulates from the area of active Unit 1 ECCS components and Unit 1 penetration rooms following a loss of coolant accident (LOCA)"

The ABGTS consists of two independent and redundant trains. Each train consists of a heater, a prefilter, moisture separator, a high efficiency particulate air (HEPA) filter, two activated charcoal adsorber sections for removal of gaseous activity (principally iodines), and a fan. Ductwork, valves or dampers, and instrumentation also form part of the system. A second bank of HEPA filters follows the adsorber section to collect carbon fines and provide backup in case the main HEPA filter bank fails. The downstream HEPA filter is not credited in the analysis. The system initiates filtered ventilation of the Auxiliary Building Secondary Containment Enclosure (ABSCE) exhaust air following receipt of a Phase A containment isolation signal.

The ABGTS is a standby system, not used during normal plant operations.

During emergency operations, the ABSCE dampers are realigned and ABGTS fans are started to begin filtration. Air is exhausted from the Unit 1 ECCS pump rooms, Unit 1 penetration rooms, and fuel handling area through the filter trains.

The prefilters or moisture separators remove any large particles in the air, and any entrained water droplets present, to prevent excessive loading of the HEPA filters and charcoal adsorbers.

The ABGTS is discussed in the FSAR, Sections 6.5.1,9.4.2,15.0, and 6.2.3 (Refs. 1, 2, 3, and 4, respectively).

(continued)

Watts Bar-Unit 1 B 3.7-62 Revision 87, 1 10, 1 19 Amendment 92

ABGTS B 3.7.12 BASES (continued)

APPLICABLE The ABGTS design basis is established by the oonsequences of the limiting SAFETY ANALYSES Design Basis Accident (DBA), which is a LOCA. The analysis of 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 airbome radioactive material provided by the one remaining train of this filtration system. The amount of ftssion 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.a (Ref. 6).

The ABGTS satisfies Criterion 3 of the NRC Policy Statement.

(continued)

Watts Bar-Unit 1 B 3.7-62A Revision 87, 1 10, 1 19 Amendment 92

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 could result in the atmospheric release from the ABSCE exceeding the 10 CFR 100 (Ref. 7) limits in the event of a LOCA.

The ABGTS is considered OPERABLE 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 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.

APPLICABILIry 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.

(continued)

Watts Bar-Unit 1 B 3.7-63 Revision 55, 87 , 1 19 Amendment 92

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 OPEMBLE train is adequate to perform the ABGTS function. The 7 day Completion Time is based on the risk from an event occuning requiring the inoperable ABGTS train, and the remaining ABGTS train providing the required protection.

8.1 and B.2 When Required Action A.1 cannot be completed within the associated Completion Time, or when both ABGTS trains are inoperable, 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 3.7-64 Revision 1 19 Amendment 92

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 environmental and normal operating conditions on this system are not severe, testing each train once every month provides an adequate check on this system.

Monthly heater operation dries out any moisture accumulated in the charcoal from humidity in the ambient air. The system must be operated for > 10 continuous hours with the heaters energized. 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 ftltertests are in accordance with Regulatory Guide 1.52 (Ref. 8). The VFTP includes testing HEPA filter performance, charcoal adsorber efftciency, minimum system flow rate, and the physical properties of the activated charcoal (general use and following speciflc 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 Amendment 92

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 Federal Regulations, Part 1 00.1 1 , "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 147W605-242,'ElectricalTech Spec Compliance Tables."

1'1. Deleted.

Watts Bar-Unit 1 B 3.7-67 Revision 29, 55, 1 19 Amendment 92

Fuel Storage Pool Water Leve!

B 3.7.1 3 B 3.7 PLANT SYSTEMS B 3.7.1 3 Fuel Storage Pool Water Level BASES BACKGROUND The minimum water level in the fuel storage pool meets the assumptions of iodine decontamination factors following a fuel handling accident. The specified water level shields and minimizes the general area dose when the storage racks are filled to their maximum capacity. The water also provides shielding during the movement of spent fuel.

A general description of the fuel storage pool design is given in the FSAR, Section 9.1.2 (Ref. 1). A description of the Spent Fuel Pool Cooling and Cleanup System is given in the FSAR, Section 9.1.3 (Ref. 2). The assumptions of the fuel handling accident are given in the FSAR, Section 15.4.5 (Ref. 3).

APPLICABLE The minimum water level in the fuel storage pool meets the assumptions of the SAFETY fuel handling accident described in Regulatory Guide 1.183 (Ref. 6). The Total ANALYSES Effective Dose Equivalent (TEDE) for control room occupants, individuals at the exclusion area boundary, and individuals within the low population zone will remain within 10 CFR 50.67 (Ref. 7) and Regulatory Position C.4.4 ot Regulatory Guide 1.183 (Ref. 6) for a fuel handling accident.

According to Reference 6, there is 23 ft of water between the top of the damaged fuel bundle and the fuel pool surface during a fuel handling accident. With 23 ft of water, the assumptions of Reference 6 can be used directly. ln practice, this LCO preserves this assumption for the bulk of the fuel in the storage racks. ln the case of a single bundle dropped and lying horizontally on top of the spent fuel racks, however, there may be < 23 ft of water above the top of the fuel bundle and the surface, indicated by the width of the bundle. To offset this small nonconservatism, the analysis assumes that all fuel rods fail, although analysis shows that only the first few rows fail fom a hypothetical maximum drop.

The fuel storage poolwater level satisfies Criterion 2 of the NRC Policy Statement.

(continued)

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Fuel Storage Pool Water Level B 3.7.13 BASES SURVEILLANCE SR 3.7.13.1 (continued)

REQUIREMENTS pool is normally stable. Water level changes are controlled by plant procedures and are acceptable based on operating experience.

During refueling operations, the level in the fuel storage pool is in equilibrium with the refueling canal, and the level in the refueling canal is checked daily in accordance with SR 3.9.7.1.

REFERENCES 1. Watts Bar FSAR, Section 9.1.2,'Spent Fuel Storage."

2. Watts Bar FSAR, Section 9.1.3, "Spent Fuel Pool Cooling and Cleanup System."

3. Watts Bar FSAR, Section 15.4.5, "Fuel Handling Accident."

4. Deleted

5. Deleted

6. Regulatory Guide 1.183, ?ltemate RadiologicalSource Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors," July 2000.

7. Title 10, Code of Federal Regulations, 10 CFR 50.67, "Accident Source Term."

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Refueling Cavity Water Level B 3.9.7 B 3.9 REFUELING OPERATIONS B 3.9.7 Refueling Cavity Water Level BASES BACKGROUND The movement of inadiated fuel assemblies within containment requires a minimum water level of 23 ft above the top of the reactor vessel flange. During refueling, this maintains sufficient water level in the containment, refueling canal, fuel transfer canal, refueling cavity, and spent fuel pool. Sufficient water is necessary to retain iodine fission product activity in the water in the event of a fuel handling accident (Refs. 2 and 8). Sufficient iodine activity would be retained to limit offsite doses from the accident to the limits defined in 10 CFR 50.67 (Ref. 7) and Regulatory Position C.4.4. ot Regulatory Guide 1.183 (Ref. 8).

APPLICABLE During movement of inadiated fuel assemblies, the water level in the refueling SAFETY ANALYSES canal and the refueling cavity is an initial condition design parameter in the analysis of a fuel handling accident in containment (Refs. 2 and 8). A minimum water level of 23 tl (Regulatory Position 2 of Appendix B to Regulatory Guide 1.183 (Ref. 8)) allows an overall iodine decontamination factor of 200 to be used in the accident analysis for iodine. This relates to the assumption that 99.5% of the total iodine released from the pellet to cladding gap of all the dropped fuel assembly rods is retained by the refueling cavity water. The fuel pellet to cladding gap is assumed to contain 8% of the l-131 , 1oo/o of the Kr-85, and 5% of the other noble gases and iodides from the total fission product inventory in accordance with Regulatory Position 3.1 of Regulatory Guide 1.183 (Ref. 8).

The fuel handling accident analysis inside containment is described in Reference 2. With a minimum water level of 23 ft in conjunction with a minimum decay time of 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months prior to fuel handling without containment closure or Auxiliary Building isolation, the analysis and test programs demonstrate that the iodine release due to a postulated fuel handling accident is adequately captured by the water and offsite doses are maintained within allowable limits (Refs. 7 and 8).

Refueling cavity water level satisfies Criterion 2 of the NRC Policy Statement.

(Continued)

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Refueling Cavity Water Level B 3.9.7 BASES (continued)

SURVEILLANCE sR 3.9.7.1 REQUIREMENTS Verification of a minimum water level of 23 ft above the top of the reactor vessel flange ensures that the design basis for the analysis of the postulated fuel handling accident during refueling operations is met. Water at the required level above the top of the reactor vessel flange limits the consequenoes of damaged fuel rods that are postulated to result from a fuel handling accident inside containment (Ref. 2).

The Frequency of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is based on engineering judgment and is considered adequate in view of the large volume of water and the normal procedural controls of valve positions, which make significant unplanned level changes unlikely.

REFERENCES 1. Deleted.

2. Watts Bar FSAR, Section 15.4.5, "Fuel Handling Accident."

3. NUREG-0800, "Standard Review Plan," Section 15.7.4, "Radiological Consequences of Fuel-Handling Accidents," U.S. Nuclear Regulatory Commission.

4. Title 10, Code of Federal Regulations, Part 20.1201 (a), (aX1 ), and (2)(2),

"Occupational Dose Limits for Adults."

5. Deleted

6. Deleted.

7. Title 10, Code of Federal Regulations, 10 CFR 50.67, "Accident Source Term."

8. Regulatory Guide 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors," July 2000.

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Decay Time B 3.9.1 0 B 3.9 REFUELING OPERATIONS B 3.9.10 Decay Time BASES BACKGROUND Section 15.5.6 of the WBN, Unit 1 UFSAR (Ref. 1) defines the assumptions of the fuel handling accident radiological analysis, including a minimum decay time for inadiated fuel assemblies prior to movement. This assumption ensures that the inventory of radioactive isotopes is at a level that supports the safety analysis assumptions.

To ensure that irradiated fuel assemblies have decayed for the appropriate period of time, a limitation is established to require the reactor core to be subcritical for a time period at least equivalent to the minimum decay time assumption in the fuel handling analysis prior to allowing inadiated fuel to be moved.

Given that no inadiated fuel assembly will be moved outside of the containment until the minimum decay time requirement is met, this requirement also ensures that any irradiated fuel assemblies that are moved outside of the containment meet the decay time assumption in the radiological analysis of the fuel handling accident.

APPLICABLE The radiological analysis of the fuel handling accident (Ref. 1) assumes a SAFETY ANALYSES minimum decay time prior to movement of inadiated fuel assemblies. The requirements of LCO 3.3.7, "Control Room Emergency Ventilation System (CREVS) Actuation lnstrumentation,' LCO 3.7.10, "Control Room Emergency Ventilation System (CREVS),' LCO 3.7.11, 'Control Room Emergency Air Temperature Control System (CREATCS)," and LCO 3.9.7, "Refueling Cavity Water Level,' in conjunction with a minimum decay time of 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months prior to irradiated fuel movement ensures that the release of fission product radioactivity, subsequent to a fuel handling accident, results in doses that are within the requirements of 10 CFR 50.67 (Ref. 2) and Regulatory Position C.4.4 ot Regulatory Guide 1.183 (Ref. 3).

The decay time satisfies Criterion 2ot 10 CFR 50.36(cx2xii).

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Decay Time B 3.9.10 BASES (continued)

LCO A minimum decay time of 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months is required prior to moving inadiated fuel assemblies within containment. This preserves an assumption in the fuel handling accident analysis (Ref. 1), and ensures that the radiological consequences of a postulated fuel handling accident inside containment are within acceptable limits.

APPLICABILITY This LCO applies during movement of irradiated fuel assemblies within the containment, since the potentialfor a release of fission products exists.

ACTIONS 4.1 When the initial conditions for prevention of an accident cannot be met, steps should be taken to preclude the accident from occurring. When the reactor is subcritical for < 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months , movement of irradiated fuelassemblies within containment must be suspended. This action precludes the possibility of a fuel handling accident in containment. This action does not preclude moving a fuel assembly to a safe position.

The immediate Completion Time is consistent with the required times for actions to be performed without delay and in a controlled manner.

SURVEILLANCE sR 3.9.10.1 REQUIREMENTS This SR verifies that the reactor has been subcritical for at least 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months prior to moving inadiated fuel assemblies by confirming the date and time of subcriticality. This ensures that any inadiated fuel assemblies have decayed for at least 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months prior to movement. The Frequency of "Prior to movement of irradiated fuel in the containmenf is appropriate, be@use it ensures that the decay time requirement has been met just prior to moving the inadiated fuel.

REFERENCES 1. Watts Bar UFSAR, Section 15.5.6, "Environmental Consequences of a Postulated Fuel Handling Accident."

2. Title 10, Code of Federal Regulations, 10 CFR 50.67, "Accident Source Term."

3. Regulatory Guide 1 .183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors," July 2000.

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ENCLOSURE 3 WBN TECHNICAL REQUIREMENTS MANUAL TABLE OF CONTENTS E3-1

TABLE OF CONTENTS TECHNICAL REQUIREMENTS TABLE OF CONTENTS LIST OF ACRONYMS ........................vii LIST OF EFFECTIVE PAGES ..........viii 1.0 usE AND APP1lCATlON.................. ...............1.1-1 1.1 Definitions .......................1.1-1 1.2 Logical Connec'tors ..............1.2-1 1.3 Completion Times.......... ......1.&1 1.4 Frequency.... ....1.+1 TR 3.0 APPLlCABlLlTY............... ............3.0-1 TR 3.1 REACTMry CONTROL SYSTEMS... ............3.1.1 TR 3.1.1 Boration Systems Flow Paths, Shutdown.... ...........3.1-1 TR 3.1.2 Boration Systems Flow Paths, Operating ...............3.1-3 TR 3.1.3 Charging Pump, Shutdown.... .............3.1-5 TR 3.1.4 Charging Pumps, Operating ...............3.1-6 TR 3.1.5 Borated Water Sources, Shutdown.... .....................3.1-8 TR 3.1.6 Borated Water Sour@s, Operating ....3.1-10 TR 3.1.7 Position lndication System, Shutdown ....................3.1-13 TR 3.3 TNSTRUMENTATTON ..................3.&1 TR 3.3.1 ReactorTrip System (RTS) lnstrumentation........ ...3.3-1 TR 3.3.2 Engineered Safety Features Actuation System (ESFAS) lnsfumentation................ ................3.&5 TR 3.3.3 Movable lncore Detectors.................. .3.3-12 TR 3.3.4 Seismic lnstrumentation................ ......3.&14 TR 3.3.5 Turbine Overspeed Protection.... ........3.&.18 TR 3.3.6 Loose-Part Detection System........ .....3.&20 TR 3.3.7 Plant Calorimehic Measurement............ .................3.122 TR 3.3.8 Hydrogen Monitors...... ...3.3-24 TR 3.3.9 Power Distribution Monitoring System (PDMS)........ ...................3.&26 TR 3.4 REACTOR COOT.ANT SYSTEM (RCS)........... ....................3.+1 TR 3.4.1 SafetyValves, Shutdown.... ................3.+1 TR 3.4.2 Pressurizer Temperature Limits ........... ...................3.4-3 TR 3.4.3 RCS Vents ......................3.+5 TR 3.4.4 Chemistry.... ....................3.+7 TR 3.4.5 Piping System Structural lntegrity ......3.+10 TR 3.6 CoNTAlNMENT SYSTEMS.................. ...........3.G1 TR 3.6.1 TR 3.6.2 lnlet Door Position Monitoring System........ ............3.64 TR 3.6.3 Lower Compartment Cooling (LCC) System ..........3.ffi (continued)

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TABLE OF CONTENTS (continued)

TR 3.7 PLANT SYSTEMS .......................3.7-1 TR 3.7.1 Steam Generator Pressure/

Temperafure Limitations .......3.7-1 TR 3.7.2 Flood Protection P1an....... ...................3.7-3 TR 3.7.3 Snubbers..... ....................3.7-10 TR 3.7.4 Sealed Source Contamination................... ..............3.7-22 TR 3.7.5 Area Temperafure Monitorin9............... ...................3.7-26 TR 3.8 ELECTRTCAL POWER SYSTEMS... ...............3.&1 TR 3.8.1 lsolation Devices ............3.&1 TR 3.8.2 Containment Penetration Conductor Overcunent Protection Devices........ .........3.8-5 TR 3.8.3 Motor-Operated Valves Thermal Overload Bypass Devices....... ..............3.&10 TR 3.8.4 Submerged Component Circuit Protection .............3.&17 TR 3.9 REFUELTNG OPERATTONS.................. ..........3.9,1 TR 3.9.1 Deleted........ ....................3.9.1 TR 3.9.2 Communications.............. ....................3.9-2 TR 3.9.3 Retueling Machine....... ...3.9.3 TR 3.9.4 Crane Travel- Spent Fuel Storage Pool Building ...3.9'5 5.0 ADMlNlSTRATIVE CONTRO1S................ ......5.G,1 5.1 Technical Requirements (TR) Conhol Program .....5.&1 (continued)

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TABLE OF CONTENTS (continued)

BASES B 3.0 TECHNICAL REQUIREMENTS (TR) AND TECHNTCAL SURVETLT-ANCE REQUIREMENTS rrSR)

ApPLtCABtLtTY............... ...................8 3.0-1 B 3.1 REACTMry CONTROL SYSTEMS... ............B 3.1.1 B 3.1.1 Boration Systems Flow Paths, Shutdown.... ....8 3.1-1 B 3.1.2 Boration Systems Flow Paths, Operating ........B 3.1-5 B 3.1.3 Charging Pump, Shutdown.... ......B 3.1-9 B 3.1.4 Charging Pumps, Operating ........B 3.1-11 B 3.1.5 Borated Water Sources, Shutdown ..................B 3.1-14 B 3.1.6 Borated Water Sources, Operating ..................B 3.1-18 B 3.1 .7 Position lndication Syrstem, Shutdown.... .........B 3.1-23 B 3.3 TNSTRUMENTATTON ..................8 3.&1 B 3.3.1 ReactorTrip System (RTS) lnstrumentation........ .................B 3.$1 B 3.3.2 Engineered Safety Feafu res Actuation Syatem (ESFAS) lnstrumentation................ ................B 3.34 B 3.3.3 Movable lncore Detecto8.................. ...............8 3.&7 B 3.3.4 Seismic lnstrumentation................ ....................8 3.&10 B 3.3.5 Turbine Overspeed Protection.... .B 3.&14 B 3.3.6 Loose-Part Detection System........ ...................8 3.&18 8.3.3.7 Plant Calorimetric Measurement............ ..........B 3.3-21 B 3.3.8 Hydrogen Monitors....... ................B 3.&25 B 3.3.9 Power Distribution Monitoring System (PDMS)........ ............B 3.&30 B 3.4 REACTOR COOLANT SYSTEM (RCS)........... .....B 3.+1 B 3.4.1 SafetyValves, Shutdown .............B 3.+1 B 3.4.2 Pressurizer Temperafure Limib ........... ............8 3.4 B 3.4.3 RCS Vents.. .............83.+7 B 3.4.4 Chemistry.... .............B 3.+1O B 3.4.5 Piping System Structural lntegrity ....................B 3.+14 B 3.6 CoNTAlNMENT SYSTEMS.................. ...........B 3.G1 B 3.6.1 lce Bed Temperature Monitoring System......... ....................B 3.&1 B 3.6.2 lnlet Door Position Monitoring System........ .....8 3.66 B 3.6.3 Lower Compartment Cooling (LCC) System ...B 3.G10 B 3.7 PI.ANT SYSTEMS ..83.7.1 B 3.7.1 Steam Generator Pressureffemperature Limitations ..........B 3.7-1 B 3.7.2 Flood Protection Plan....... ............B 3.74 B 3.7.3 Snubbers..... .............B 3.7-12 B 3.7.4 Sealed Source Contamination................... .......B 3.7-18 B 3.7.5 Area Temperafure Monitorin9............... ............B 3.7-22 (continued)

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TABLE OF CONTENTS (continued)

B 3.8 ELECTRICAL POWER SYSTEMS... ...............B 3.&1 B 3.8.1 lsolation Devices .....B 3.&1 B 3.8.2 Containment Penetration Conductor Overcunent Protection Devices........ ..B 3.&7 B 3.8.3 Motor-Operated Valves Thermal Overload Bypass DeMces........ ...........8 3.8-15 B 3.8.4 Submerged Component Circuit Prctection ......8 3.&19 B 3.9 REFUELTNG OPERATTONS.................. ..........8 3.9,1 B 3.9.1 Deleted........ .............B 3.$1 B 3.9.2 Communications............... ............B 3.9.3 B 3.9.3 Retueling Machine....... .................B 3.9-5 B 3.9.4 Crane Travel - Spent Fuel Storage PoolBuilding ...................B 3.9 Watts Bar-Unit 1 IV Technical Requi rements Last Updated Revision 53

LIST OF TABLES Table No. Title Paqe 1.1-1 MODES....... ....................1.1-6 3.3.1-1 ReactorTripSystemlnstrumentiationResponseT'imes.... .........3.&3 3.3.2-1 Engineered Safety Features Actuation System Response Times.......... .....................3.&7 3.3.4-1 Seismic Monitoring lnformation ..........3.T17 3.7.3-1 Snubber Visual lnspection Acceptance Criteria .....3.7-14 3.7.3-2 Snubber Visual lnspection Surveillance Frequency... .................3.7-1 5 3.7.3-3 Snubber Transient Event lnspection ..3.7-17 3.7.34 Snubber Functional Testing Plan ............. ...............3.7-18 3.7.3-5 Snubber FunctionalTesting Acceptrance Criteria ...3.7-20 3.7.5-1 Area Temperature Monitorin9............... ...................3.7-29 3.8.3-1 Motor-Operated Valves Thermal Overload Devices \{hich Are Bypassed Under Accident Conditions................. ....3.&12 3.8.4-1 Submerged Componenb With Automatic De-energization UnderAccident Conditions ...3.&19 Watts Bar-Unit 1 Technical Requirements

LIST OF FIGURES Fioure No. Tite Paoe 3.1.6 Boric Acid Tank Limib Based on RWST Boron Concentration ..3.1-12a 3.7.3-1 Sample Plan B for Snubber Functional Test.............. ..................3.7-21 LIST OF MISCELLANEOUS REPORTS AND PROGRAMS Core Operating Limits Report Watts Bar-Unit 1 VI Techn ical Requi rements Last Updated Revision 21

LIST OF ACRONYMS Acronvm Title ABGTS Auxiliary Building Gas Treatment System ACRP Auxiliary Contro! Room Panel ASME American Society of Mechanical Engineers AFD Axial Flux Difference AFW Auxiliary Feedwater System ARO AII Rods Out ARFS Air Return Fan System ARV Atmospheric Relief Valve BOC Beginning of Cycle CCS Component Gooling 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 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 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 Therma! 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 1 vil Last Updated Revision 46 Technical Requirements

TECHN ICAL REQUIREMENTS LIST OF EFFECTIVE PAGES Page Revision Page Revision Number Number Number Number i 46 3.1-4 0 ii 53 3.1-5 38 iii 46 3.1-6 51 iv 53 3.1-7 0 V 0 3.1-8 0 vi 21 3.1-9 37 vii 46 3.1-10 0 viii 54 3.1-1 1 33 ix 53 3.1-12 0 x il 3.1-12a 42 xi 53 3.1-13 I xii 22 3.3-1 0 xiii 37 3.3-2 0 xiv 47 3.3-3 34 XV il 3.34 44 1.1-1 0 3.3-5 0 1.1-2 22 3.3-6 0 1 .1-3 0 3.3-7 26 1.14 31 3.3-8 36 1 .1-5 0 3.3-9 3 1 .1-6 0 3.3-10 0 1.2-1 0 3.3-1 1 49 1.2-2 0 3.3-12 46 1.2-3 0 3.3-13 0 1.3-1 0 3.3-14 40 1.3-2 0 3.3-15 40 1.3-3 . 0 3.3-16 0 1.34 0 3.3-17 19 1.3-5 0 3.3-18 38 1.3-6 0 3.3-19 38 1.3-7 0 3.3-20 40 1.3-8 0 3.3-21 0 1.3-g 0 3.3-22 23 1 .3-10 0 3.3-23 23 1.3-11 0 3.3-24 45 1.3-12 0 3.3-25 45 1 .3-13 0 3.3-26 46 1.4-1 0 3.3-27 46 1.4-2 0 3.3-28 46 1.4-3 0 3.4-1 0 1.44 0 3.4-2 0 3.0-1 38 3.4-3 0 3.0-2 38 3.44 0 3.0-3 39 3.4-5 0 3.0-4 38 3.4-6 0 3.1-1 38 3.4-7 0 3.1-2 0 3.4-8 0 3.1-3 51 3.4-9 0 Watts Bar-Unit 1 vilt Technical Requirements Last Updated Revision 54

TECHN ICAL REQU IREMENTS LIST OF EFFECTIVE PAGES Page Revision Page Revision Number Number Number Number 3.4-10 52 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 3.64 0 3.8-13 0 3.6-5 0 3.8-14 0 3.6-6 0 3.8-15 0 3.6-7 0 3.8-16 0 3.7-1 0 3.8-17 0 3.7-2 0 3.8-18 18 3.7-3 17 3.8-19 18 3.74 17 3.9-1 53 3.7-5 17 3.9-2 0 3.7-6 17 3.9-3 28 3.7-7 17 3.94 28 3.7-8 17 3.9-5 0 3.7-9 17 5.0-1 24 3.7-14 47 3.7-11 47 3.7-12 0 3.7-13 0 3.7-14 47 3.7-15 0 3.7-16 0 3.7-17 0 3.7-18 47 3.7-19 5 3.7-20 0 3.7-21 0 3.7-22 43 3.7-23 0 3.7-24 0 3.7-25 0 3.7-26 40 3.7-27 40 3.7-28 40 3.7-29 2 3.7-30 2 3.8-1 0 3.8-2 0 3.8-3 0 3.8-4 25 3.8-5 0 3.8-6 0 Watts Bar-Unit 1 tx Technical Requi rements Last Updated Revision 53

TECHN ICAL 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 0 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 3.0-6 0 B 3.3-18 0 B 3.0-7 0 B 3.3-19 40 B 3.0-8 0 B 3.3-20 11 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 3.1-1 0 41 B 3.3-34 54 B 3.1-11 51 B 3.4-1 0 B 3.1-12 0 B 3.4-2 0 B 3.1-1 3 41 B 3.4-3 0 B 3.1-14 0 B3.44 0 B 3.1-1 5 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 52 B 3.1-25 I B 3.4-15 38 B 3.3-1 0 B 3.4-16 52 B 3.3-2 0 B 3.&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 10 B 3.3-8 46 B 3.6-7 20 B 3.3-9 46 B 3.6-8 10 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 Requirements Last Updated Revision 54

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 B3.74 17 B 3.94 0 B 3.7-5 17 B 3.9-5 28 B 3.7-6 17 B 3.9-6 0 B 3.7-7 17 B 3.9-7 28 B 3.7-8 17 B 3.9-8 0 B 3.7-9 17 B 3.9-9 0 B 3.7-10 17 B 3.7-11 17 B 3.7-12 47 B 3.7-13 47 B 3.7-14 47 B 3.7-15 47 B 3.7-16 47 B 3.7-17 47 B 3.7-18 0 B 3.7-19 43 B 3.7-20 0 B 3.7-21 0 B 3.7-22 0 B 3.7-23 20 B 3.7-24 40 B 3.7-25 40 B 3.8-1 0 B 3.8-2 0 B 3.8-3 0 B 3.8-4 0 B 3.8-5 0 B 3.8-6 25 B 3.8-7 25 B 3.8-8 0 B 3.8-9 0 B 3.8-10 0 B 3.8-1 1 0 B 3.8-12 0 B 3.8-13 25 B 3.8-14 25 B 3.8-15 0 B 3.8-16 0 B 3.8-17 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 xt Technical Requi rements Last Updated Revision 53

TECHNICAL REQUIREMENTS MANUAL LIST OF EFFECTIVE PAGES - REVISION LISTING Revisions Issued SUBJECT Revision 0 09-30-95 Initial 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 Performan@ of Snubber Functional Tests at Power Revision 6 09-08-97 Revised Actions for Turbine Overspeed Protection Revision 7 09-12-97 Change OPAT/OTAT Response Time Revision 8 09-22-97 Clarification of Surveillance Frequency for Position lndication System Revision I 1G'10-97 Revised Boron Concenfation for Borated Water Sources Revision 10 12-17-98 lCS lnlet Door Position Monitoring - ChannelCheck Revision 11 01-0&99 Computer-Based Analysis for Loose Parts Monitoring Revision 12 01-1$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 0+07-99 Submerged Component Circuit Protec'tion Revision 15 04-07-99 Submeryed Component Circuit Protection Revision 16 M-1&99 Submeryed Component Circuit Protection Revision 17 0$2S99 Flood Protec'tion Plan Revision 18 0&0&99 Submerged Component Circuit Protection Revision 19 10-12-99 Upgrade Seismic Monitoring lnstruments Revision 20 03/13/00 Added Notes to Address lnstrument Enor for Various Parameters Revision 21 04113/00 COLR, Cycle 3, Rev 2 Revision 22 07107100 Elimination of Response Time Testing Watts Bar-Unit 1 xil Technica! Requi rements Last Updated Revision 22

TECHNICAL REQUIREMENTS MANUAL LIST OF EFFECTIVE PAGES - REVISION LISTING Revisions lssued SUBJECT Revision 23 O1l22lO1 Plant Calorimetric (LEFM)

Revision 24 03/19/01 TRM Change ControlProgram per50.59 Rule Revision25 05/15/01 Change in Preventive Maintenance Frequency for Molded Case Circuit Breakers Revision 26 05129101 Change CVI Response Time fom 5 to 6 Seconds Revision 27 01131102 Change pH value in the borated water sour@s due to TS change for ice weight reduction Revision 28 02105102 Refueling machine upgrade under DCN D-50991-A Revision 29 02126102 Added an additional 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 OOlO5l02 Updated TR 3.3.5.1 to reflect implementation of the TIPTOP program in a Technical lnshuction (Tl).

Revision 31 10131102 Conect RTP to 3459 MWt (PER 02-9519400)

Revision 32 Ogl17l03 Editorial conection to Bases for TSR 3.1.5.3.

Revision 33 1ol14l0g Updated TRs 3.1.5 and 3.1.6 and their respective bases to incorporate boron concentration changes in accordance with change packages WBN-TS-O2- 1 4 and WBN-TS-O$O 1 7.

Revision 34 OSll4t(X Revised ltem 5, "Source Range, Neutron Flurf function of Table 3.3.1-1 to provide an acceptable response time of less than or equal0.5 seconds. (Reference TS Amendment 52.)

Revision 35 @/06/05 Revised Table 3.3.2-1, "Engineered Safety Features Actuation systems Response Times," to revise containment spray response time and to add an asterisk note to notiation 13 of the table via Change Package WBN-TS-04-16.

Revision 36 09125106 Revised the response time for Containment Spray in Table 3.3.2-'l and the RTHor values in the Bases for TR 3.7.1. Both changes result fom the replacement of the steam generators.

Revision 37 11/08/06 Revised TR 3.1.5 and 3.1.6 and the Bases for these TRs to update the boron concentration limits of the RWST and the BAT.

Watts Bar-Unit 1 xilt Tech nical Req ui rements Last Updated Revision 37

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

Revision 39 04116107 Updated the TRM to be consistent with Tech Spec Amendment 42.

TRM Revision 39 modified the requirements of TSR 3.0.3 by incorporating changes similar to those outlined in TSTF-358.

(TS-07-03)

Revision 40 05t24t07 Updated the TRM and Bases to remove the various requirements for the submittal of reports to the NRC. (T5-07-06)

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 XI of the ASME 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 a3t2012008 Revision 42 updates Figure 3.1 .6 to remove the 240 TPBAR Limit.

Revision 43 0711712008 Revision 43 removes a reporting requirement from TR 3.7.4, "Sealed Source Contiamination." The revision also updates the Bases for TR 3.7.4.

Revision 44 10110/2008 Revision 44 updates Table 3.3.1-1 to be consistent with the changes approved by NRC as Tech Spec Amendment 68.

Revision 45 0212312009 Added TR 3.3.8, "Hydrogen Monitors," and the Bases for TR 3.3.8.

This change is based on Technical Specification (TS) Amendment T2which removed the Hydrogen Monitors (Function 13 of LCO 3.3.3) from the TS.

Revision 46 0912012010 Revision 46 implements changes from License AmendmentSz (Technical Specification (TS) Bases Revsion 104) for the approved BEACON-TSM application of the Power Distribution Monitoring System (PDMS).

Revision 47 1010812010 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 xiv Technical Requi rements Last Updated Revision 47

TECHNICAL REQUIREMENTS MANUAL LIST OF EFFECTIVE PAGES REVISION LISTING Revisions lssued SUBJECT Revision 48 04t12t2011 CANCELLED Revision 49 051241201 1 Revision 49 updated Note 14 of Table 3.3.2-1 to clarify that the referenced time is only for 'partial' transfer of the ECCS pumps from the VCT to the RWST.

Revision 50 1211212011 Clarifies the acceptability of periodically using a portion of the 25%

grace period in TSR 3.0.2 to facilitate 13 week maintenance work schedules.

Revision 51 08ta9t2013 Adds a note to TR 3.1 .2 and TR 3.1.4 to permit securing one charging pump in order to supporting transition into or from the Applicability of Technical Specification 3.4.12 (PER 593365).

Revision 52 0813012013 Clarifies that TR 3.4.5, "Piping System Structura! lntegdty," applies to all ASME Code Class 1,2, and 3 piping systeffis, and is not limited to reactor coolant system piping.

Revision 53 1211212013 Technical Specification Amendment 92 added Limiting Condition for Operation (LCO) 3.9.10, "Decay Time," which was redundant to Technical Requirement (TR) 3.9.1, "Decay Time." Revision 53 removes TR 3.9.1 from the Technical Requirements Manual (TRM) and the TRM Bases.

Revision 54 0111412014 Update the references for B 3.3.9 to incorporate NRC approved WCAP-12472-P-A Addendum 1-A and 4-A. The WCAPs provide the bases to allow the core exit thermocouple uncertainty to be calculated on a cycle specific basis. The new method obtiains thermocouple data during power ascension following a refueling outage. This information is prccessed bythe PDMS to determine the power dependent thermocouple uncertainty.

Watts Bar-Unit 1 Technical Requirements Last Updated Revision 54

ENCLOSURE 4 WBN TECHNICAL REQUIREMENTS MANUAL CHANGED PAGES E4-1

Boration Systems FIow Paths, Operating TR 3.1.2 TR 3.1 REACTIVITY CONTROL SYSTEMS TR 3.1 .2 Boration Systems FIow Paths, Operating TR 3.1.2 Two of the following three boron injection flow paths shall be OPERABLE:

a. One flow path from the boric acid tanks, through a boric acid transfer pump, through a charging pump to the Reactor Coolant System (RCS).

b. Two flow paths from the Refueling Water Storage Tank (RWST), through charging pumps to the RCS.

ln MODE 3, a charging pump may be made incapable of injecting to support transition into or fom the Applicability of the TS ICO 3.4.12, "Cold Overpressure Mitigation System (COMS),' for up to four hours or until the temperature of all the RCS cold legs exceeds 375'F, whichever occurs first.

APPLICABILITY: MODES 1, 2, and 3.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME One required flow path A.1 Restore required flow path to inoperable. OPERABLE status.

OR A.2.1 Be in MODE 3.

AND 4.2.2 Borate to a SDM equivalent to :

1o/o LWk at 200oF.

AND 4.2.3 Restore required path to 246 hours0.00285 days 0.0683 hours 4.06746e-4 weeks 9.3603e-5 months OPERABLE status.

B. Required Action and 8.1 Be in MODE 4.

associated Completion Time of Condition A not met.

Watts Bar-Unit 1 3.1-3 Revision 51 Technical Requirements 08/09/1 3

Charging Pumps, Operating TR 3.1.4 TR 3.1 REACTIVITY CONTROL SYSTEMS TR 3.1 .4 Charging Pumps, Operating TR 3.1.4 Two charging pumps shall be OPERABLE.

NOTE ln MODE 3, a charging pump may be made incapable of injecting to support transition into or from the Applicability of the TS LCO 3.4.12, 'Cold Overpressure Mitigation System (COMS)," for up to four hours or until the temperature of all the RCS cold legs exceeds 375'F, whichever occurs first.

APPLICABIL!TY: MODES 1, 2, and 3.

ACTIONS CONDlTION REQUIRED ACTION COMPLETION TIME A. One required charging Restore required 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months pump inoperable. charging pump to OPERABLE status.

OR 4.2.1 Be in MODE 3. 78 hours9.027778e-4 days 0.0217 hours 1.289683e-4 weeks 2.9679e-5 months AND A.2.2 Borate to a SDM 78 hours9.027778e-4 days 0.0217 hours 1.289683e-4 weeks 2.9679e-5 months equivalent to > 1o/o Aldk at 200"F.

AND 4.2.3 Restore required 246 hours0.00285 days 0.0683 hours 4.06746e-4 weeks 9.3603e-5 months charging pump to OPERABLE status.

B. Required Action and B.1 Be in MODE 4.

associated Completion Time of Condition A not met.

Watts Bar-Unit 1 3.1-6 Revision 51 Technical Requirements 08/09/1 3

Piping System Structural lntegrity TR 3.4.5 TR 3.4 REACTOR COOLANT SYSTEM (RCS TR 3.4.5 Piping System Structural Integrity TR 3.4.5 The structural integrity of ASME Code Class 1,2, and 3 components in all systems shall be maintained in accordance with TSR 3.4.5.1 and TSR 3.4.5.2.

APPLICABILITY: AIIMODES.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME Structural integrity of any ASME Restore structural integrity of Prior to increasing Code Class 1 component(s) not affected component(s) to within Reactor Coolant within limits. limit. System temperature

> 50oF above the minimum temperature required by NDT considerations.

lsolate affected component(s). Prior to increasing Reactor Coolant System temperature

> 50oF above the minimum temperature required by NDT considerations.

(continued)

Watts Bar-Unit 1 3.4-10 Revision 38, 52 Technical Requirements 08/30/2013

Piping System Structural lntegrity TR 3.4.5 TECHN ICAL SU RVEILLANCE REQU IREMENTS SURVEILLANCE FREQUENCY TSR 3.4.5.1 Inspect each reactor coolant pump flywheel according to According to the the recommendations of Regulatory Position C.4.b of recommend-actions Regulatory Guide 1.14, Revision 1, August 1975. of Regulatory Positien C.4.b of Regulatory Guide 1.14, Revision 1.

TSR 3.4.5.2 Veriff the structural integrity of ASME Code Class 1,2, ln accordance with and 3 components in al! systems are in accordance with the lnseruice the lnservice lnspection Program. lnspection Program.

Watts Bar-Unit 1 3.4-12 Revision 52 Technical Requirements 0813012013

Deleted TR 3.9.1 TR 3.9 REFUELING OPERATIONS TR 3.9.1 Deleted Watts Bar-Unit 1 3.9-1 Revision 53 Technical Requirements 12t12t2013

Boration Systems Flow Paths, Operating B 3.1 .2 B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1 .2 Boration Systems Flow Paths, Operating BASES BACKGROUND A description of the Boration Systems Flow Paths is provided in the Bases for Technical Requirement 3.1.1, "Boration Systems Flow Paths, Shutdown."

APPLICABLE The boration subsystem is not assumed to be OPERABLE to mitigate the SAFETY ANALYSES consequences of a DBA or Transient. ln the case of a malfunction of the Chemical and Volume Control System, which causes a boron dilution event, the response required by the operator is to close the appropriate valves in the reactor makeup system and/or stop the primary water pumps. This action is required before the shutdown margin is lost. Operation of the boration subsystem is not assumed to mitigate this event (Ref. 1). OPEMBILITY of the charging pumps, the Refueling Water Storage Tank (RWST), and the appropriate flow paths is required as part of the Emergency Core Cooling System (ECCS).

The Technical Specifications for the ECCS address the requirements of these components.

TR TR 3.1.2 requires at least two boron injection flow paths to be OPERABLE during MODES 1,2, and 3, in order to provide two redundant paths to accomplish (1) normal makeup, (2) chemical shim reactivity control, and (3) miscellaneous fill and transfer operations. This requirement may be achieved by having two of the following three flow paths OPEMBLE:

a. One flow path from the boric acid storage tanks, through a boric acid transfer pump, through a charging pump to the Reactor Coolant System (Rcs).

b. Two flow paths from the RWST, through a charging pump to the RCS.

TR 3.'1.2 is modified by a Note. As indicated in this note, operation in MODE 3 with a charging pump made incapable of injecting in order to facilitate entry into or exit from the Applicability of TS LCO 3.4.12, "Cold Overpressure Mitigation Systems (COMSI is necessary with a COMS arming temperature at or near the Mode 4 boundary temperature of 350'F. TS LCO 3.4.12 requires that certain pumps be rendered incapable of injecting at and below the COMS arming temperature. When this temperature is at or near the MODE 3 boundary temperature, time is needed to make pumps incapable of injecting prior to entering the COMS Applicability, and provide time to restore the inoperable pumps to OPEMBLE status on exiting the COMS Applicability.

(continued)

Watts Bar-Unit 1 B 3.1-5 Revision 51 Technical Requirements 08/09/1 3

Charging Puffip, Operating B 3.1 .4 B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1 .4 Charging Pumps, Operating BASES BACKGROUND A description of the Boration Systems Flow Paths is provided in the Bases for Technical Requirement 3. 1 . 1,'Boration Systems Flow Paths, Shutdown."

APPLICABLE The boration subsystem is not assumed to be OPERABLE to mitigate the SAFETY ANALYSES consequenoes of a DBA or transient. ln the case of a malfunction of the Chemicaland Volume ControlSystem (CVCS), which causes a boron dilution event, the response required by the operator is to close the appropriate valves in the reactor makeup system and/or stop the primary water pumps. This action is required before the shutdown margin is lost. Operation of the boration subsystem is not assumed to mitigate this event (Ref. 1). OPERABILITY of the charging pumps, the refueling water storage tank, and the appropriate flow paths is required as part of the Emergency Core Cooling System (ECCS). The Technical Specifications for the ECCS address the requirements of these components.

TR TR 3.1.4 requires at least two charging pumps to be OPERABLE during MODES 1,2, and 3 in order to assure redundant pumps to the two redundant flow paths to accomplish (1) normalmakeup, (2) chemicalshim reactivity control, and (3) miscellaneous fill and transfer operations.

TR 3.1.2 is modified by a Note. As indicated in this note, operation in MODE 3 with a charging pump made incapable of injecting in order to facilitate entry into or exit from the Applicability of TS LCO 3.4.'12, "Cold Overpressure Mitigation Systems (COMSI is necessary with a COMS arming temperature at or near the Mode 4 boundary temperature of 350"F. TS LCO 3.4.12 requires that certain pumps be rendered incapable of injecting at and below the COMS arming temperature. When this temperature is at or near the MODE 3 boundary temperature, time is needed to make pumps incapable of injecting prior to entering the COMS Applicability, and provide time to restore the inoperable pumps to OPERABLE status on exiting the COMS Applicability.

(continued)

Watts Bar-Unit 1 B 3.1-11 Revision 51 Technical Requirements 08/09/1 3

BASES B 3.3 INSTRUMENTATION B 3.3.9 Power Distribution Monitoring System (PDMS)

BASES BACKGROUND The Power Distribution Monitoring System (PDMS) generates a continuous measurement of the incore power distribution using the methodology documented in References 1, 4, and 5. The PDMS employs an advanced three-dimensional nodal code to calculate the incore power distribution. The reference incore power distribution is periodically normalized to the incore flux measurements from the movable incore detectors. On a nominal once-per-minute basis, the incore power distribution is updated with plant instrumentation, most notably from the Core Exit Thermocouples (CETs). ln this way, the information from the up-to-the-minute PDMS incore power distribution is equivalent to a full incore flux map using the Movable lncore Detector System (Technical Requirement 3.3.3).

The PDMS incore power distribution measurement can be used to determine the most limiting core peaking factors, FN66, the Nuclear Enthalpy Rise Hot Channel Factor (Technical Specification 3.2.21and Fq(z), the Heat Flux Hot Ghannel Factor (Technical Specification 3.2.1). The incore power distribution measurement can also be used in the calibration of the excore neutron flux detection system (Technical Specifi cation 3.3. 1 ), monitoring the QUADMNT POWER TILT RATIO (OPTR) (TechnicalSpecification 3.2.4), and veriffing the position of a rod with inoperable position indicators (Technical Specification 3.1.8).

The PDMS requires information on cunent plant and core conditions in order to determine the core power distribution using the core peaking factor measurement and measurement uncertainty methodology described in References 1, 4, and 5.

The OPERABILIW of the PDMS with the specified minimum complement of instrumentation channel inputs ensures that the measurements obtained from use of this system accurately represent the spatial neutron flux distribution of the core. The PDMS requires input for power range detector calibrated voltage values, average reactor vessel inlet temperature, reactor power level, control bank positions, and temperatures from the CETs.

Either the PDMS or the Movable lncore Detector System may be used for calibration of the Excore Neutron Flux Detection System, monitoring the QUADRANT POWER TILT RATIO, or measurement of Fq(z) or FN66. Similarly, either the PDMS or the Movable lncore Detector System may be used for veriffing the position of a rod with inoperable position indicatorc, but only the PDMS must satisff OPERABILITY requirements priorto this function.

(continued)

Watts Bar-Unit 1 B 3.3-30 Revision 46, 54 Technical Requirements 0112312014

BASES Bases (continued)

APPLICABLE The PDMS is used for periodic measurement of the core power distribution to SAFETY confirm operation within design limits and periodic calibration of the excore ANALYSES detectorc. This system does not initiate any automatic protection action. The PDMS is not assumed to be OPERABLE to mitigate the consequences of a DBA or transient (References 2 and 3).

TR TR 3.3.9 requires the PDMS to be OPEMBLE with the specified number of instrument channel inputs from the plant computer for each function listed in Table 3.3.9-1. The PDMS is OPERABLE when the required channel inputs are available, the calibration data set is valid, and reactor power is 2 25% RTP.

This TR ensures the OPERABILITY of the PDMS when required to monitorthe power distribution within the core. The PDMS is used for periodic surveillance of the incore power distribution and calibration of the excore detectors. The surveillance of incore power distribution verifies that the peaking factors are within their design envelope (Reference 3). The peaking factor limits include measurement uncertainty which bounds the actual measurement uncertainty of an OPERABLE PDMS (References 1, 4, and 5).

Maintaining the minimum number of instrumentation channel inputs ensures the uncertainty is bounded by the uncertainty methodology. Similarly, when THERMAL POWER is less than2So/o RTP, then the accuracy of the adjustment provided by the CETs to the measured PDMS power distribution may not be bounded by the uncertainties documented in .References 1, 4, and 5.

APPLICABILITY The PDMS must be OPERABLE when it is used for calibration of the Excore Neutron Flux Detection System, monitoring the QPTR, measurement of FN611 and Fq(z), or veriffing the position of a rod with inoperable position indicators.

ACTIONS 4.1 The Required Action A.1 has been modified by a Note stating that the provisions of TR 3.0.3 do not apply.

With THERMAL Power less than 25% RTP orwith one or more required channel inputs inoperable or unavailable to the PDMS, the PDMS must not be used to obtain an incore power distribution measurement. Therefore, the Required Action A.1 prohibits the use of the inoperable system for the applicable monitoring or calibration functions.

(continued)

Watts Bar-Unit 1 B 3.3-31 Revision 46, 54 Technical Requirements 01t23t2014

PDMS B 3.3.9 BASES BASES (continued)

TECHNICAL The subsequent PDMS calibration frequency is 31 Effective Full Power Days SURVEILI.ANCE (EFPD) when the CET chess knight move pattem is not satisfied. The CET REQUIREMENTS chess knight move pattern is satisfied when every interior core location (fuel (continued) assemblies not face adjacent to the core baffle) is no further than a chess knight's move from an OPERABLE CETC. The 31 EFPD frequency calibration requirement is modified by a note that clarifies that subsequent PDMS qalibration is not required to be performed until 31 EFPD after the CET chess knight move pattem is not satisfied.

The subsequent PDMS calibration frequency is 180 EFPD when the CET chess knight move paftern is satisfied. The CET chess knight move pattem provides coverage of all interior fuel assemblies (coverage of fuel assemblies with a face along the baffle is not required). Fuel assemblies which are within a chess knight's move of an OPEMBLE CET are covered.

REFERENCES 1. WCAP-12472-P-A, "BEACON Core Monitoring and Operations Support System," August 1994.

2. 10 CFR 50.46, "Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors."

3. WCAP-1 1618, "MERITS Program-Phase ll, Task 5, Criteria Application,"

including Addendum 1 dated April, 1989.

4. WCAP-12472-P-A, Addendum 1-A, "BEACON Core Monitoring and Operations Support Systeffi," January 2000.

5. WCAP-12472-P-A, Addendum 4-A, "BEACON Core Monitoring and Operations Support Systffi," September 2A12.

(continued)

Watts Bar-Unit 1 B 3.3-34 Revision 46, 54 Technical Requirements 0112312014

Piping System Structural Integrity B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.5 Piping System Structural Integrity BASES BACKGROUND lnservice inspection of ASME Code Class 1,2, and 3 components and pressure testing of ASME Code Class 1,2, and 3 pumps and valves in all systems are performed in accordance with Section Xl of the ASME Boiler and Pressure VesselCode (Ref. 1) and applicable Addenda, as required by 10 CFR 50.55a(g)

(Ref. 2). Exception to these requirements apply where relief has been granted by the Commission pursuant to 10 CFR 50.55a(gx6)(i) and (a)(3). !n general, the surveillance intervals specified in Section Xl of the ASME Code apply.

However, the lnservice lnspection Program includes a clarification of the frequencies for performing the inservice inspection and testing activities required by Section Xl of the ASME Code. This clarification is provided to ensure consistency in surveillance intervals throughout the Technical Specifications.

Each reactor coolant pump flyrheel is, in addition, inspected as recommended in Regulatory Position C.4.b of Regulatory Guide 1.14, Revision 1, August 1975 (Ref.3).

Additionally, programmatic information on lnservice lnspection is provided in Technical Specifications, Chapter 5.0, Administrative Controls, Section 5.7 .2.1 1, lnservice lnspection Program.

APPLICABLE Certain components which are designed and manufactured to the requirements SAFETY ANALYSES of specific sections of the ASME Boiler and Pressure Vessel Code are part of the primary success path and function to mitigate DBAs and transients. However, the operability of these components is addressed in the relevant specifications that cover individual components. In addition, this particular Requirement covers only structural integrity inspection/testing requirements for these components, which is not a consideration in designing the accident sequences for theoretical hazard evaluation (Ref.4).

TR TR 3.4.5 requires that the structural integrity of the ASME Code Class 1,2, and3 components in all systems be maintained in accordance with TSR 3.4.5.1 and TSR 3.4.5.2. ln those areas where conflic{ may exist between the Technical Specifications and the ASME Boiler and Pressure Vessel Code, the Technical Specifi cations take precedence.

(continued)

Watts Bar-Unit 1 B 3.4-14 Revision 41 , 52 Technical Requirements a813012013

Piping System Structural lntegrity B 3.4.5 BASES (continued)

TECHNICAL TSR 3.4.5.1 SURVEILLANCE REQUIREMENTS This surveillance stipulates inspection of the coolant pump flyrheel in accordance with Regulatory Position C.4.b of Regulatory Guide 1.14, Revision 1.

This inspection verifies the structural integrity of the flyrheel.

TSR 3.4.5.2 TSR 3.4.5.2 requires the verification of structural integrity of ASME Code Class 1,2, and 3 components in all systems are in accordance with the lnservice lnspection Program.

REFERENCES 1. ASME Boiler and Pressure Vessel Code, Section Xl.

2. 10 CFR 50.55a, "Codes and Standards."

3. Regulatory Guide 1.14, Revisior 1, 1975.

4. WCAP-1 1618, "MERITS Program-Phase ll, Task 5, Criteria Application,"

including Addendum 1 dated April, 1989.

Watts Bar-Unit 1 B 3.4-16 Revision 52 Technical Requirements 0813a12013

Deleted B 3.9.1 B 3.9 REFUELING OPERATIONS B 3.9.1 Deleted (continued)

Watts Bar-Unit 1 B 3.9-1 Revision 53 Technical Requirements 12112t2013

Deleted B 3.9.1 (continued)

THIS PAGE INTENTIONALLY LEFT BIANK Watts Bar-Unit 1 B 3.9-2 Revision 53 Technical Requirements 1211212013

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