Text

Changes Made to the Technical Requirements Manual
	ML11325A322
Person / Time
Site:	Watts Bar
Issue date:	11/16/2011
From:	James Shea; Tennessee Valley Authority
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	Download: ML11325A322 (54)
	v • d • e

Tennessee Valley Authority, 1101 Market Street, Chattanooga, Tennessee 37402 November 16, 2011 10 CFR 50.4 ATTN: Document Control Desk U. S. Nuclear Regulatory Commission Washington, D.C. 20555-0001 Watts Bar Nuclear Plant, Unit 1 Facility Operating License No. NPF-90 NRC Docket No. 50-390

Subject:

CHANGES MADE TO THE TECHNICAL REQUIREMENTS MANUAL The purpose of this letter is to provide the NRC with copies of the changes that have been made to the Watts Bar Nuclear Plant (WBN), Unit 1 Technical Requirements Manual (TRM) through the implementation of Revisions 46, 47 and 49. This information is provided in accordance with WBN TRM Section 5.1, "Technical Requirements Control Program," on a frequency consistent with 10 CFR 50.71(e). These changes have been implemented at WBN during the period since the last update of the TRM on April 20, 2010 and meet the criteria described within the TRM control program for which prior NRC approval is not required. The updates to the TRM are provided in the enclosures listed below. Please note that a revision was initiated as Revision 48 but cancelled prior to being implemented.

Since WBN's TRM is incorporated by reference into the Updated Final Safety Analysis Report (UFSAR), this letter certifies that the content of this update accurately presents the changes made since the previous submittal.

There are no regulatory commitments associated with this submittal. Please direct any questions concerning this matter to Kara Stacy, Program Manager at (423) 751-3489.

I declare under penalty of perjury that the foregoing is true and correct. Executed on this 16th day of November 2011.

Respec

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nea Do0 o Printed on recycled paper

U.S. Nuclear Regulatory Commission Page 2 November 16, 2011

Enclosures:

1.

2.

3.

WBN, Unit 1 Technical Requirements Manual - Table of Contents WBN, Unit 1 Technical Requirements Manual - Changed Pages WBN, Unit 1 Technical Requirements Manual Bases - Changed Pages Enclosures cc (Enclosures):

NRC Regional Administrator - Region II NRC Senior Resident Inspector - Watts Bar Nuclear Plant, Unit 1 NRC Senior Resident Inspector - Watts Bar Nuclear Plant, Unit 2

U.S. Nuclear Regulatory Commission Page 3 November 16, 2011 DKG:JLB Enclosures

bcc (Enclosures):

NRC Project Manager - Watts Bar Nuclear Plant, Unit 1 NRC Project Manager - Watts Bar Nuclear Plant, Unit 2 EDMS, WT 3B-K

Enclosure I WBN, Unit I Technical Requirements Manual - Table of Contents

TABLE OF CONTENTS TECHNICAL REQUIREMENTS TABLE OF CONTENTS LIST O F TABLES.........................................................................................................................................

v LIST O F FIG URES.......................................................................................................................................

vi LIST O F AC RO NYM S...................................................................................................................................

vii LIST O F EFFECTIVE PAG ES.....................................................................................................................

viii 1.0 USE A ND APPLICATIO N................................................................................................

1.1-1 1.1 Definitions...........................................................................................................

1.1-1 1.2 Logical Connectors.............................................................................................

1.2-1 1.3 Com pletion Tim es...............................................................................................

1.3-1 1.4 Frequency...........................................................................................................

1.4-1 TR 3.0 APPLICABILITY...............................................................................................................

3.0-1 TR 3.1 REACTIVITY CO NTRO L SYSTEM S..............................................................................

3.1-1 TR 3.1.1 Boration System s Flow Paths, Shutdow n.........................................................

3.1-1 TR 3.1.2 Boration System s Flow Paths, O perating.........................................................

3.1-3 TR 3.1.3 Charging Pum p, Shutdow n................................................................................

3.1-5 TR 3.1.4 Charging Pum ps, O perating..............................................................................

3.1-6 TR 3.1.5 Borated W ater Sources, Shutdow n...................................................................

3.1-8 TR 3.1.6 Borated W ater Sources, O perating...................................................................

3.1-10 TR 3.1.7 Position Indication System, Shutdow n..............................................................

3.1-13 TR 3.3 INSTRUM ENTATIO N......................................................................................................

3.3-1 TR 3.3.1 Reactor Trip System (RTS) Instrum entation.....................................................

3.3-1 TR 3.3.2 Engineered Safety Features Actuation System (ESFAS) Instrum entation.....................................................

3.3-5 TR 3.3.3 M ovable Incore Detectors..................................................................................

3.3-12 TR 3.3.4 Seism ic Instrum entation.....................................................................................

3.3-14 TR 3.3.5 Turbine Overspeed Protection...........................................................................

3.3-18 TR 3.3.6 Loose-Part Detection System............................................................................

3.3-20 TR 3.3.7 Plant Calorim etric M easurem ent.......................................................................

3.3-22 TR 3.3.8 Hydrogen M onitors.............................................................................................

3.3-24 TR 3.3.9 Power Distribution M onitoring System (PDM S)................................................

3.3-26 TR 3.4 REACTO R CO O LANT SYSTEM (RCS).........................................................................

3.4-1 TR 3.4.1 Safety Valves, Shutdow n...................................................................................

3.4-1 TR 3.4.2 Pressurizer Tem perature Lim its.........................................................................

3.4-3 TR 3.4.3 RCS Vents..........................................................................................................

3.4-5 TR 3.4.4 Chem istry............................................................................................................

3.4-7 TR 3.4.5 Piping System Structural Integrity......................................................................

3.4-10 TR 3.6 CO NTAINM ENT SYSTEM S............................................................................................

3.6-1 TR 3.6.1 Ice Bed Tem perature M onitoring System..........................................................

3.6-1 TR 3.6.2 Inlet Door Position M onitoring System..............................................................

3.6-4 TR 3.6.3 Lower Com partm ent Cooling (LCC) System....................................................

3.6-6 (continued)

Watts Bar-Unit 1 Technical Requirements Last Updated Revision 46

TABLE OF CONTENTS (continued)

TR 3.7 TR 3.7.1 TR 3.7.2 TR 3.7.3 TR 3.7.4 TR 3.7.5 TR 3.8 TR 3.8.1 TR 3.8.2 TR 3.8.3 TR 3.8.4 TR 3.9 TR 3.9.1 TR 3.9.2 TR 3.9.3 TR 3.9.4 5.0 5.1 PLANT SYSTEM S............................................................................................................

3.7-1 Steam Generator Pressure/

Tem perature Lim itations.......................................................................

3.7-1 Flood Protection Plan.........................................................................................

3.7-3 Snubbers.............................................................................................................

3.7-10 Sealed Source Contam ination...........................................................................

3.7-22 Area Tem perature M onitoring............................................................................

3.7-26 ELECTRICAL POW ER SYSTEM S.................................................................................

3.8-1 Isolation Devices.................................................................................................

3.8-1 Containment Penetration Conductor Overcurrent Protection Devices................................................................................

3.8-5 Motor-Operated Valves Thermal Overload Bypass Devices....................................................................................

3.8-10 Subm erged Com ponent Circuit Protection.......................................................

3.8-17 REFUELING O PERATIO NS...........................................................................................

3.9-1 Decay Tim e.........................................................................................................

3.9-1 Com m unications.................................................................................................

3.9-2 Refueling M achine..............................................................................................

3.9-3 Crane Travel - Spent Fuel Storage Pool Building.............................................

3.9-5 ADM INISTRATIVE CO NTRO LS.....................................................................................

5.0-1 Technical Requirem ents (TR) Control Program...............................................

5.0-1 (continued)

Watts Bar-Unit 1 Technical Requirements ii

TABLE OF CONTENTS (continued)

BASES B 3.0 TECHNICAL REQUIREMENTS (TR) AND TECHNICAL SURVEILLANCE REQUIREMENTS (TSR)

A P P LIC A B ILITY.................................................................................................

B 3.0-1 B 3.1 REACTIVITY CONTROL SYSTEMS..............................................................................

B 3.1-1 B 3.1.1 Boration Systems Flow Paths, Shutdown.......................................................................

B 3.1-1 B 3.1.2 Boration Systems Flow Paths, Operating.......................................................................

B 3.1-5 B 3.1.3 Charging Pump, Shutdown..............................................................................................

B 3.1-9 B 3.1.4 Charging Pumps, Operating............................................................................................

B 3.1-11 B 3.1.5 Borated Water Sources, Shutdown.................................................................................

B 3.1-14 B 3.1.6 Borated Water Sources, Operating.................................................................................

B 3.1-18 B 3.1.7 Position Indication System, Shutdown............................................................................

B 3.1-23 B 3.3 INSTRUMENTATION......................................................................................................

B 3.3-1 B 3.3.1 Reactor Trip System (RTS) Instrumentation...................................................................

B 3.3-1 B 3.3.2 Engineered Safety Features Actuation System (ESFAS) Instrumentation.....................................................

B 3.3-4 B 3.3.3 Movable Incore Detectors................................................................................................

B 3.3-7 B 3.3.4 Seismic Instrumentation...................................................................................................

B 3.3-10 B 3.3.5 Turbine Overspeed Protection.........................................................................................

B 3.3-14 B 3.3.6 Loose-Part Detection System..........................................................................................

B 3.3-18 B.3.3.7 Plant Calorimetric Measurement.....................................................................................

B 3.3-21 B 3.3.8 Hydrogen Monitors...........................................................................................................

B 3.3-25 B 3.3.9 Power Distribution Monitoring System (PDMS)..............................................................

B 3.3-30 B 3.4 REACTOR COOLANT SYSTEM (RCS).........................................................................

B 3.4-1 B 3.4.1 Safety Valves, Shutdown.................................................................................................

B 3.4-1 B 3.4.2 Pressurizer Temperature Limits.......................................................................................

B 3.4-4 B 3.4.3 R C S V e nts........................................................................................................................

B 3.4-7 B 3.4.4 C he m istry..........................................................................................................................

B 3.4 -10 B 3.4.5 Piping System Structural Integrity....................................................................................

B 3.4-14 B 3.6 CONTAINMENT SYSTEMS............................................................................................

B 3.6-1 B 3.6.1 Ice Bed Temperature Monitoring System........................................................................

B 3.6-1 B 3.6.2 Inlet Door Position Monitoring System............................................................................

B 3.6-6 B 3.6.3 Lower Compartment Cooling (LCC) System..................................................................

B 3.6-10 B 3.7 PLANT SYSTEMS............................................................................................................

B 3.7-1 B 3.7.1 Steam Generator Pressure/Temperature Limitations.....................................................

B 3.7-1 B 3.7.2 Flood Protection Plan.......................................................................................................

B 3.7-4 B 3.7.3 S n ub be rs...........................................................................................................................

B 3.7-12 B 3.7.4 Sealed Source Contamination.........................................................................................

B 3.7-18 B 3.7.5 Area Temperature Monitoring..........................................................................................

B 3.7-22 (continued)

Watts Bar-Unit 1 iii Technical Requirements Last Updated Revision 46

TABLE OF CONTENTS (continued)

B 3.8 B 3.8.1 B 3.8.2 B 3.8.3 B 3.8.4 B 3.9 B 3.9.1 B 3.9.2 B 3.9.3 B 3.9.4 ELECTRICAL POW ER SYSTEM S................................................................................

B 3.8-1 Isolation Devices...............................................................................................................

B 3.8-1 Containment Penetration Conductor O vercurrent Protection Devices.........................................................................

B 3.8-7 Motor-Operated Valves Thermal O verload Bypass Devices..................................................................................

B 3.8-15 Subm erged Com ponent Circuit Protection.....................................................................

B 3.8-19 REFUELING O PERATIO NS...........................................................................................

B 3.9-1 Decay Tim e.......................................................................................................................

B 3.9-1 Com m unications...............................................................................................................

B 3.9-3 Refueling M achine............................................................................................................

B 3.9-5 Crane Travel - Spent Fuel Storage Pool Building.......................................................................................................

B 3.9-8 Watts Bar-Unit 1 Technical Requirements iv

LIST OF TABLES Table No.

Title PaQe 1.1-1 M O D E S...............................................................................................................

1.1-6 3.3.1-1 Reactor Trip System Instrumentation Response Times...................................

3.3-3 3.3.2-1 Engineered Safety Features Actuation System Response Times....................................................

3.3-7 3.3.4-1 Seismic Monitoring Information.........................................................................

3.3-17 3.7.3-1 Snubber Visual Inspection Acceptance Criteria................................................

3.7-14 3.7.3-2 Snubber Visual Inspection Surveillance Frequency.........................................

3.7-15 3.7.3-3 Snubber Transient Event Inspection.................................................................

3.7-17 3.7.3-4 Snubber Functional Testing Plan.......................................................................

3.7-18 3.7.3-5 Snubber Functional Testing Acceptance Criteria.............................................

3.7-20 3.7.5-1 Area Temperature Monitoring............................................................................

3.7-29 3.8.3-1 Motor-Operated Valves Thermal Overload Devices Which Are Bypassed Under Accident Conditions..........................................

3.8-12 3.8.4-1 Submerged Components With Automatic De-energization Under Accident Conditions..................................................................

3.8-19 Watts Bar-Unit 1 v

Technical Requirements

LIST OF FIGURES Figqure No.

3.1.6 3.7.3-1 Title Page Boric Acid Tank Limits Based on RWST Boron Concentration....................... 3.1-12a Sample Plan B for Snubber Functional Test.....................................................

3.7-21 LIST OF MISCELLANEOUS REPORTS AND PROGRAMS Core Operating Limits Report Watts Bar-Unit 1 Technical Requirements vi Last Updated Revision 21

LIST OF ACRONYMS Acronym Title ABGTS Auxiliary Building Gas Treatment System ACRP Auxiliary Control Room Panel ASME American Society of Mechanical Engineers AFD Axial Flux Difference AFW Auxiliary Feedwater System ARO All Rods Out ARFS Air Return Fan System ARV Atmospheric Relief Valve BOC Beginning of Cycle CCS Component Cooling Water System CFR Code of Federal Regulations COLR Core Operating Limits Report CREVS Control Room Emergency Ventilation System CSS Containment Spray System CST Condensate Storage Tank DNB Departure from Nucleate Boiling ECCS Emergency Core Cooling System EFPD Effective Full-Power Days EGTS Emergency Gas Treatment System EOC End of Cycle ERCW Essential Raw Cooling Water ESF Engineered Safety Feature ESFAS Engineered Safety Features Actuation System HEPA High Efficiency Particulate Air HVAC Heating, Ventilating, and Air-Conditioning LCC Lower Compartment Cooler LCO Limiting Condition For Operation MFIV Main Feedwater Isolation Valve MFRV Main Feedwater Regulation Valve MSIV Main Steam Line Isolation Valve MSSV Main Steam Safety Valve MTC Moderator Temperature Coefficient NMS Neutron Monitoring System ODCM Offsite Dose Calculation Manual PCP Process Control Program PDMS Power Distribution Monitoring System PIV Pressure Isolation Valve PORV Power-Operated Relief Valve PTLR Pressure and Temperature Limits Report QPTR Quadrant Power Tilt Ratio RAOC Relaxed Axial Offset Control RCCA Rod Cluster Control Assembly RCP Reactor Coolant Pump RCS Reactor Coolant System RHR Residual Heat Removal RTP Rated Thermal Power RTS Reactor Trip System RWST Refueling Water Storage Tank SG Steam Generator SI Safety Injection SL Safety Limit SR Surveillance Requirement UHS Ultimate Heat Sink Watts Bar-Unit 1 vii Last Updated Revision 46 Technical Requirements

TECHNICAL REQUIREMENTS LIST OF EFFECTIVE PAGES Page Number iv iii iv v

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Revisions Revision 0 Revision 1 Revision 2 Revision 3 Revision 4 Revision 5 Revision 6 Revision 7 Revision 8 Revision 9 Revision 10 Revision 11 Revision 12 Revision 13 Revision 14 Revision 15 Revision 16 Revision 17 Revision 18 Revision 19 Revision 20 Revision 21 Revision 22 TECHNICAL R LIST OF EFFECTIV Issued 09-30-95 12-06-95 01-04-96 02-28-96 08-18-97 08-29-97 09-08-97 09-12-97 09-22-97 10-10-97 12-17-98 01-08-99 01-15-99 03-30-99 04-07-99 04-07-99 04-13-99 05-25-99 08-03-99 10-12-99 03/13/00 04/13/00 07/07/00 EQUIREMENTS MANUAL E PAGES - REVISION LISTING SUBJECT Initial Issue Submerged Component Circuit Protection Area Temperature Monitoring - Change in MSSV Limit Turbine Driven AFW Pump Suction Requirement Time-frame for Snubber Visual Exams Performance of Snubber Functional Tests at Power Revised Actions for Turbine Overspeed Protection Change OPAT/OTAT Response Time Clarification of Surveillance Frequency for Position Indication System Revised Boron Concentration for Borated Water Sources ICS Inlet Door Position Monitoring - Channel Check Computer-Based Analysis for Loose Parts Monitoring Removal of Process Control Program from TRM Deletion of Power Range Neutron Flux High Negative Rate Reactor Trip Function Submerged Component Circuit Protection Submerged Component Circuit Protection Submerged Component Circuit Protection Flood Protection Plan Submerged Component Circuit Protection Upgrade Seismic Monitoring Instruments Added Notes to Address Instrument Error for Various Parameters COLR, Cycle 3, Rev 2 Elimination of Response Time Testing xii Last Updated Revision 22 Watts Bar-Unit 1 Technical Requirements

Revisions Revision 23 Revision 24 Revision 25 Revision 26 Revision 27 Revision 28 Revision 29 Revision 30 Revision 31 Revision 32 Revision 33 Revision 34 Revision 35 Revision 36 Revision 37 Watts Bar-Unit 1 Technical Requirements TECHNICAL REQUIREMENTS MANUAL LIST OF EFFECTIVE PAGES - REVISION LISTING Issued SUBJECT 01/22/01 Plant Calorimetric (LEFM) 03/19/01 TRM Change Control Program per 50.59 Rule 05/15/01 Change in Preventive Maintenance Frequency for Molded Case Circuit Breakers 05/29/01 Change CVI Response Time from 5 to 6 Seconds 01/31/02 Change pH value in the borated water sources due to TS change for ice weight reduction 02/05/02 Refueling machine upgrade under DCN D-50991-A 02/26/02 Added an additional action to TR 3.7.3 to perform an engineering evaluation of inoperable snubber's impact on the operability of a supported system.

06/05/02 Updated TR 3.3.5.1 to reflect implementation of the TIPTOP program in a Technical Instruction (TI).

10/31/02 Correct RTP to 3459 MWt (PER 02-9519-000) 09/17/03 Editorial correction to Bases for TSR 3.1.5.3.

10/14/03 Updated TRs 3.1.5 and 3.1.6 and their respective bases to incorporate boron concentration changes in accordance with change packages WBN-TS-02-14 and WBN-TS-03-017.

05/14/04 Revised Item 5, "Source Range, Neutron Flux" function of Table 3.3.1-1 to provide an acceptable response time of less than or equal 0.5 seconds. (Reference TS Amendment 52.)

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

09/25/06 Revised the response time for Containment Spray in Table 3.3.2-1 and the RTNDT values in the Bases for TR 3.7.1. Both changes result from the replacement of the steam generators.

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.

xiii Last Updated Revision 37

TECHNICAL REQUIREMENTS MANUAL LIST OF EFFECTIVE PAGES REVISION LISTING Revisions Issued SUBJECT Revision 38 11/29/06 Updated the TRM to be consistent with Tech Spec Amendment

55. TRM Revision 38 modified the requirements for mode change limitations in TR 3.0.4 and TSR 3.0.4 by incorporating changes similar to those outlined in TSTF-359, Revision 9.

(TS-06-24)

Revision 39 04/16/07 Updated the TRM to be consistent with Tech Spec Amendment 42.

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

(TS-07-03)

Revision 40 05/24/07 Updated the TRM and Bases to remove the various requirements for the submittal of reports to the NRC. (TS-07-06)

Revision 41 05/25/07 Revision 41 updates the Bases of TR 3.1.3, 3.1.4 and 3.4.5 to be consistent with Technical Specification Amendment 66. This amendment replaces the references to Section XI of the ASME Boiler and Pressure Vessel Code with the ASME Operation and Maintenance Code for Inservice Testing (IST) activities and removes reference to "applicable supports" from the IST program.

Revision 42 03/20/2008 Revision 42 updates Figure 3.1.6 to remove the 240 TPBAR Limit.

Revision 43 07/17/2008 Revision 43 removes a reporting requirement from TR 3.7.4, "Sealed Source Contamination." The revision also updates the Bases for TR 3.7.4.

Revision 44 10/10/2008 Revision 44 updates Table 3.3.1-1 to be consistent with the changes approved by NRC as Tech Spec Amendment 68.

Revision 45 02/23/2009 Added TR 3.3.8, "Hydrogen Monitors," and the Bases for TR 3.3.8.

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

Revision 46 09/20/2010 Revision 46 implements changes from License Amendment 82 (Technical Specification (TS) Bases Revsion 104) for the approved BEACON-TSM application of the Power Distribution Monitoring System (PDMS).

Revision 47 10/08/2010 Revision 47 changes are in response to PER 215552 which requested clarification be added to the TRM regarding supported system operability when a snubber is declared inoperable or removed from service.

Watts Bar-Unit 1 xiv Technical Requirements Last Updated Revision 47

TECHNICAL REQUIREMENTS MANUAL LIST OF EFFECTIVE PAGES REVISION LISTING Revisions Revision 48 Revision 49 Issued SUBJECT 04/12/2011 CANCELLED 05/24/2011 Revision 49 updated Note 14 of Table 3.3.2-1 to clarify that the referenced time is only for 'partial' transfer of the ECCS pumps from the VCT to the RWST.

Watts Bar-Unit 1 Technical Requirements xv Last Updated Revision 49

THIS PAGE INTENTIONALLY LEFT BLANK WBN, Unit I Technical Requirements Manual - Changed Pages

ESFAS TR 3.3.2 Table 3.3.2-1 (Page 5 of 5)

Engineered Safety Features Actuation System Response Times TABLE NOTATIONS (11)

Containment purge valves only. Containment radiation monitor valves have a response time of 6.5 seconds.

(12)

(13)

Diesel generator start time includes a reactor trip response time of 2 seconds.

Includes diesel generator starting, containment spray pump sequence loading-delay/breaker closure, plus stroke time of 1-FCV-72-39/2.*

The containment integrity analysis of record was performed using 221 seconds for initiation of spray. However, Westinghouse document WAT-D-1 1264 has evaluated the initiation of spray at 234 seconds with the conclusion that the increase will have no effect on the results or conclusions of the Watts Bar LOCA and MSLB containment integrity analysis.

(14)

Diesel generator starting and sequence loading delays included. Response time limit includes the opening of valves to establish flowpath and bring pumps to full speed. The additional partial transfer of ECCS pump suction from the VCT to the RWST (RWST valves open, VCT valves not yet closed) is included.

(15)

Feedwater Isolation Valve (motor) and Feedwater Regulating Valve (air operated) response time includes an ESFAS signal response time of 2 seconds.

Watts Bar-Unit 1 Technical Requirements 3.3-11 Revision 35, 49 05/24/11

Movable Incore Detectors TR 3.3.3 TR 3.3 INSTRUMENTATION TR 3.3.3 Movable Incore Detectors TR 3.3.3 The Movable Incore Detection System shall be OPERABLE with >_ 75% of the detector thimbles, _> 2 detector thimbles per core quadrant, and sufficient movable detectors, drive, and readout equipment to map these thimbles.

NOTES -----------------------------------------------

1.

Either a full core flux map or quarter-core flux maps may be used in calibrations of the Excore Neutron Flux Detection System.

2.

Either a full core flux map or two sets of four thimble locations with quarter core symmetry may be used for monitoring the QUADRANT POWER TILT RATIO.

3.

Only > 50% of the detector thimbles and > 2 detector thimbles per core quadrant are required for Power Distribution Monitoring System (PDMS) calibration after the initial PDMS calibration after each refueling.

APPLICABILITY:

When the Movable Incore Detection System is used for:

a.

Recalibration of the Excore Neutron Flux Detection System, or

b.

Monitoring the QUADRANT POWER TILT RATIO, or

c.

Measurement of FNAH, and FQ(Z) or

d.

Calibration of the Power Distribution Monitoring System (PDMS).

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

Movable Incore Detection A.1

NOTE ------------

System inoperable.

TR 3.0.3 is not applicable.

Restore the inoperable system to Prior to using the OPERABLE status.

system for monitoring or calibration functions.

Revision 46 3.3-12 09/20/2010 Watts Bar-Unit 1 Technical Requirements

Hydrogen Monitors TR 3.3.8 TECHNICAL SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY TSR 3.3.8.1 Perform CHANNEL CHECK.

31 days TSR 3.3.8.2 Perform COT.

31 days TSR 3.3.8.3 Perform CHANNEL CALIBRATION.

18 months Watts Bar-Unit 1 3.3-25 Revision 45

PDMS TR 3.3.9 TR 3.3 INSTRUMENTATION TR 3.3.9 Power Distribution Monitoring System (PDMS)

TR 3.3.9 APPLICABILITY:

The PDMS shall be OPERABLE with:

a.

THERMAL POWER > 25% RTP, and

b.

The required channel inputs from the plant computer for each function defined in Table 3.3.9-1.

When the PDMS is used for:

a.

Calibration of the Excore Neutron Flux Detection System, or

b.

Monitoring the QUADRANT POWER TILT RATIO, or

c.

Measurement of FN AH and FQ(Z), or

d.

Verifying the position of a rod with inoperable position indicators.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A, PDMS inoperable.

A.1

NOTE------

TR 3.0.3 is not applicable.

Restore the inoperable Prior to using the system system to OPERABLE for incore power status.

distribution measurement purposes.

Watts Bar-Unit 1 Technical Requirements 3.3-26 Revision 46 09/20/2010

PDMS TR 3.3.9 TECHNICAL SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY TSR 3.3.9.1 Perform CHANNEL CHECK for each required 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months instrumentation channel specified in Table 3.3.9-1.

TSR 3.3.9.2 Verify by administrative means that the 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months surveillance requirements for each required channel specified in Table 3.3.9-1 are satisfied.

TSR 3.3.9.3 Perform PDMS calibration.

Once after each refueling AND NOTE ---------

Not required to be performed until 31 Effective Full Power Days (EFPD) after the Core Exit Thermocouple (CET) chess knight move pattern not satisfied.

31 EFPD thereafter with the CET chess knight move pattern not satisfied AND 180 EFPD thereafter with the CET chess knight move pattern satisfied Watts Bar-Unit 1 Technical Requirements 3.3-27 Revision 46 09/20/2010

PDMS TR 3.3.9 Table 3.3.9-1 (Page 1 of 1)

Power Distribution Monitoring System (PDMS) Instrumentation REQUIRED FUNCTION REQUIRED CHANNELS 3

SURVEILLANCE REQUIREMENTS SR 3.3.1.6 SR 3.3.1.11(4)

1. Power Range Neutron Flux Monitors SURVEILLANCE DESCRIPTION NIS Calibration CHANNEL CALIBRATION CHANNEL CALIBRATION CHANNEL CALIBRATION NIS Calorimetric CHANNEL CALIBRATION LEFM Availability

2.

RCS Cold Leg Temperature

3.

Reactor Power 2(1) 1(2)

SR 3.3.1. 10")

SR 3.3.3.2(6)

SR 3.3.1.2(')

SR 3.3.1.10(8)

TSR 3.3.7.1(7)

SR 3.1.8.1 SR 3.3.3.2(9)

4.

Control Bank Position (per bank)

RPI Calibration

5.

Core Exit Thermocouple Temperature 17 with > 2 per core quadrant CHANNEL CALIBRATION (1)

(2)

(3)

(4)

(5)

(6)

(7)

(8)

(9)

Either Narrow Range or Wide Range RTDs Either secondary calorimetric power, average power range neutron flux power, or average RCS Loop AT power Either the Demand Position Indication or the average of the individual Rod Position Indications Neutron detectors are excluded from CHANNEL CALIBRATION Applies to Narrow Range RTDs only Applies to Wide Range RTDs only Not applicable to average RCS Loop AT power Applies to average RCS Loop AT power only Applies to Core Exit Thermocouple Temperatures only Watts Bar-Unit 1 Technical Requirements 3.3-28 Revision 46 09/20/2010

Flood Protection Plan TR 3.7.2 This Page Intentionally Left Blank Watts Bar-Unit 1 Technical Requirements 3.7-9 Revision 17 05/25/99

Snubbers TR 3.7.3 TR 3.7 PLANT SYSTEMS TR 3.7.3 Snubbers TR 3.7.3 All required* snubbers shall be OPERABLE.

APPLICABILITY:

MODES 1, 2, 3, and 4.

MODES 5 and 6 for snubbers located on systems required OPERABLE in those MODES.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

One or more required*

A.1 Declare the supported system Immediately snubber(s) removed from inoperable.

service.

B.

One or more required*

B.1 Declare the supported system Immediately snubber(s) found inoperable and enter the inoperable, applicable LCO action statement for the supported system.

B.2 Perform an engineering 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months evaluation per Table 3.7.3-5 Note 3 on the attached component.

(continued)

Required means the snubber performs a support function that is necessary for the suported TS system to perform its specified safety function.

Watts Bar-Unit 1 Technical Requirements 3.7-10 Revision 47 10/08/10

Snubbers TR 3.7.3 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C.

Required Action(s) and C.1 Initiate a PER.

Immediately associated Completion Time(s) not met.

Watts Bar-Unit 1 Technical Requirements 3.7-11 Revision 47 10/08/10

Snubbers TR 3.7.3 TECHNICAL SURVEILLANCE REQUIREMENTS

-NOTES-

1.

Each snubber shall be demonstrated OPERABLE by performance of the following augmented inservice inspection program.

2.

Snubbers which fail the visual inspection or the functional test acceptance criteria shall be repaired or replaced. Replacement snubbers and snubbers which have repairs which might affect the functional test results shall be tested to meet the functional test criteria before installation in the unit. Mechanical snubbers shall have met the acceptance criteria subsequent to their most recent service, and the freedom-of-motion test must have been performed within 12 months before being installed in the unit.

3.

As used herein, type of snubber shall mean snubbers of the same design and manufacturer, irrespective of capacity.

SURVEILLANCE FREQUENCY TSR 3.7.3.1 Visually inspect each snubber in accordance with the In accordance with acceptance criteria in Table 3.7.3-1.

Table 3.7.3-2.

TSR 3.7.3.2 Perform a transient event inspection of all hydraulic 6 months following and mechanical snubbers in accordance with transient event.

Table 3.7.3-3.

(continued)

Watts Bar-Unit 1 Technical Requirements 3.7-12 09/30/95

Snubbers TR 3.7.3 TECHNICAL SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY TSR 3.7.3.3 Perform a functional test on a representative sample Each refueling of snubbers in accordance with Table 3.7.3-4 to outage.

determine acceptance with criteria in Table 3.7.3-5.

TSR 3.7.3.4

NOTES

1.

The maximum expected service life for various seals, springs, and other critical parts shall be determined and established based on engineering information and shall be extended or shortened based on monitored test results and failure history.

2.

Critical parts shall be replaced so that the maximum service life will not be exceeded during a period when the snubber is required to be OPERABLE.

3.

The parts replacement shall be documented and the documentation shall be retained for the duration of the unit operating license.

Verify that the service life of hydraulic and Each refueling mechanical snubbers has not been exceeded or will outage.

not be exceeded prior to the next scheduled surveillance inspection.

Watts Bar-Unit 1 Technical Requirements 3.7-13 09/30/95

Snubbers TR 3.7.3 Table 3.7.3-1 (Page 1 of 1)

Snubber Visual Inspection Acceptance Criteria

1.

Visual inspection shall verify that:

a.

There are no visible indications of damage or impaired OPERABILITY (See Note 6).

b.

Attachments to the foundation or supporting structure are functional; and

c.

Fasteners for attachment of the snubber to the component and to the snubber anchorage are functional.

2.

Snubbers which appear inoperable as a result of visual inspections shall be classified as unacceptable and may be reclassified acceptable for the purpose of establishing the next visual inspection interval, provided that:

a.

The cause of the rejection is clearly established and remedied for that particular snubber and for other snubbers irrespective of type that may be generically susceptible; and

b.

The affected snubber is functionally tested in the as-found condition and determined OPERABLE per Table 3.7.3-5, Snubber Functional Test Acceptance Criteria.

3.

All snubbers connected to an inoperable common hydraulic fluid reservoir shall be counted as unacceptable for determining the next inspection interval.

4.

Snubbers which have been made inoperable as the result of unexpected transients, isolated damage or other such random events, when the provisions of Table 3.7.3-3 have been met and any other appropriate corrective action implemented, shall not be counted in determining the next visual inspection interval.

5.

A review and evaluation shall be performed and documented to justify continued operation with an unacceptable snubber. If continued operation cannot be justified, the snubber shall be declared inoperable and the ACTION requirements shall be met.

6.

Mechanical and hydraulic snubbers are considered OPERABLE (until proven otherwise by functional testing), unless they are disconnected at either end, experienced gross deformation of the snubber or structural support, or the hydraulic fluid level is empty for hydraulic snubbers.

Watts Bar-Unit 1 Technical Requirements 3.7-14 Revision 47 10/08/10

Snubbers TR 3.7.3 Table 3.7.3-3 (Page 1 of 1)

Snubber Transient Event Inspection

1.

An inspection shall be performed of all hydraulic and mechanical snubbers attached to sections of systems that have experienced unexpected, potentially damaging transients as determined from a review of operational data and a visual inspection of the systems within six months following such an event.

2.

In addition to satisfying the visual inspection acceptance criteria, freedom-of-motion of mechanical snubbers shall be verified using one of the following:

a. Manually induced snubber movement.

b. Evaluation of in-place snubber piston setting.

c. Stroking the mechanical snubber through its full range of travel.

Watts Bar-Unit I Technical Requirements 3.7-17 09/30/95

Snubbers TR 3.7.3 Table 3.7.3-4 (Page 1 of 2)

Snubber Functional Testing Plan

1.

The representative sample of snubbers shall include each type and shall be tested using sample plan A for hydraulic snubbers and sample plan B for mechanical snubbers.

2.

The NRC Regional Administrator shall be notified in writing of any changes to the sample plan prior to the test period.

SAMPLE PLAN A

1.

At least 10% of the total hydraulic snubber population shall be functionally tested either in-place or in a bench test.

2.

For each hydraulic snubber of a type that does not meet the functional test acceptance criteria of Table 3.7.3-5, an additional 10% of hydraulic snubbers shall be functionally tested until no more failures are found or until all hydraulic snubbers have been functionally tested.

SAMPLE PLAN B

1.

An initial representative sample of 37 mechanical snubbers shall be functionally tested in accordance with Figure 3.7.3-1. For each mechanical snubber type which does not meet the functional test acceptance criteria of Table 3.7.3-5, another sample of at least 19 snubbers shall be tested. The results from this sample plan shall be plotted using an "Accept" line which follows the equation N = 36.49 + 18.18C where "C" is the number of snubbers which do not meet functional test acceptance criteria. If the point plotted falls on or below the "Accept" line, testing of that type of snubber may be terminated. If the point plotted falls above the "Accept" line, testing must continue until the point falls in the "Accept" region or all mechanical snubbers have been tested.

(continued)

Watts Bar-Unit 1 Technical Requirements 3.7-18 Revision 47 10/08/10 WBN, Unit I Technical Requirements Manual Bases - Changed Pages

Movable Incore Detectors B 3.3.3 B 3.3 INSTRUMENTATION B 3.3.3 Movable Incore Detectors BASES BACKGROUND The Movable Incore Detection System uses six miniature fission chamber neutron detectors to measure fuel assembly reaction rates. The miniature fission chambers are positioned by a Detector Drive System which pushes and pulls the detectors in and out of the reactor core through one of 58 thimbles, which are open at one end. The thimbles have Reactor Coolant System (RCS) pressure on the outside and atmospheric pressure inside. Each thimble is inserted into the center position of the fuel assembly, all the way to the top of the fuel assembly. The drive system slowly pushes the detector up through the fuel assembly, inside the thimble, to the top of the core. An x-y plot (position verses flux level) is initiated with the slow withdrawal of the detectors through the core from top to a point below the bottom. As the detector traverses the thimble tube it obtains the raw currents for 61 axial levels. At each level, the computer takes three rapid looks, averages the readings and uses this as the base reading for that axial point. In a similar manner, other core locations are selected and plotted. Each detector provides axial flux distribution data along the center of a fuel assembly.

Each of the six miniature neutron detectors has its own drive unit. A series of five-and ten-path transfer devices are then used to direct a detector into one of several possible fuel assemblies. In this manner, the six detectors and drive units are used to monitor 58 fuel assemblies in the core.

The operability of the Movable Incore Detectors with the specified minimum complement of equipment ensures that the measurements obtained from use of this system accurately represent the spatial neutron flux distribution of the core.

The OPERABILITY of this system is demonstrated by irradiating each detector used and determining the acceptability of the data obtained. The detectors are normalized to one another through cross-calibration comparison (looking for relative readings) of each detector's output. This effectively makes the data from all the detectors approximately the same as the reference detector.

Either the Movable Incore Detector System or the Power Distribution Monitoring System (PDMS) may be used for calibration of the Excore Neutron Flux Detection System, monitoring the QUADRANT POWER TILT RATIO, or measurement of FN LH and FQ(z). Since the Movable Incore Detector System is utilized by the PDMS, it is required to be OPERABLE before it is used to calibrate the PDMS. The initial calibration of PDMS after a refueling requires a full incore flux map (> 75% of the detector thimbles) but subsequent calibrations require only > 50% of the detector thimbles.

(continued)

Watts Bar-Unit 1 Technical Requirements B 3.3-7 Revision 46 09/20/2010

Movable Incore Detectors B 3.3.3 BASES BACKGROUND (continued)

When the Movable Incore Detectors are used directly for measuring FN AH, the Nuclear Enthalpy Rise Hot Channel Factor (Technical Specification 3.2.2) and FQ(z), the Heat Flux Hot Channel Factor (Technical Specification 3.2.1), a full incore flux map is required. Quarter-core flux maps, as designed in Reference 1, may be used to calibrate the excore axial offset. Either full incore flux maps or symmetric incore thimbles may be used for monitoring the QPTR.

APPLICABLE SAFETY ANALYSES The Movable Incore Detector System is used for periodic surveillance of the power distribution and calibration of the excore detectors.

Surveillance of the power distribution verifies that the peaking factors are within the design envelope. The system is not used continuously and does not initiate any automatic protection action. The Movable Incore Detector System is not assumed to be OPERABLE to mitigate the consequences of a DBA or transient (Ref. 2).

TR TR 3.3.3 specifies that the Movable Incore Detection System shall be OPERABLE with at least 75% of the detector thimbles, and a minimum of two detector thimbles per core quadrant. Also, sufficient movable detectors, drive, and readout equipment to map these thimbles.

This TR ensures the OPERABILITY of the Movable Incore Detector Instrumentation when required to monitor the flux distribution within the core. The Movable Incore Detector System is used for periodic surveillance of the power distribution, and calibration of the excore detectors. The surveillance of power distribution verifies that the peaking factors are within their design envelope (Ref. 2).

Three notes modify TR 3.3.3. Note 1 clarifies that quarter-core flux maps may be used to calibrate the Excore Neutron Flux Detection System in lieu of a full core flux map with > 75% of the detector thimbles in accordance with Reference 1.

Note 2 clarifies that symmetric incore thimbles may be used for monitoring the QPTR in lieu of a full incore flux map with > 75% of the detector thimbles in accordance with Technical Specification 3.2.4. Note 3 clarifies that a flux map with > 50% of the detector thimbles and > 2 detector thimbles per core quadrant may be used for subsequent calibration of the PDMS after the initial calibration after a refueling in lieu of a full incore flux map with > 75% of the detector thimbles and > 2 detector thimbles per core quadrant in accordance with Reference 3.

APPLICABILITY The Movable Incore Detection System must be OPERABLE when it is used for recalibration of the Excore Neutron Flux Detection System, or monitoring the QPTR, measurement of F NAH and FQ (z), or calibration of the PDMS.

(continued)

Revision 46 09/20/2010 Watts Bar-Unit 1 Technical Requirements B 3.3-8

Movable Incore Detectors B 3.3.3 BASES (continued)

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

An inoperable Movable Incore Detection Systems cannot be used for recalibration of the Excore Neutron Flux Detection System, or monitoring the QPTR or measurement of F NAH and FQ(z), or calibration of the PDMS. Therefore, the Required action A.1 prohibits the use of the inoperable system for the above applicable monitoring or calibration functions.

TECHNICAL TSR 3.3.3.1 SURVEILLANCE REQUIREMENTS The Movable Incore Detector System must be demonstrated OPERABLE at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by the setting of each detector's operating voltage. The operating voltage is set by determining the operating region for each detector after inserting it into a high flux region of the core. The acceptability of each detector is verified by the performance of a detector drift check. The operating voltage must be determined prior to using the Movable Incore Detector System for recalibration of the Excore Neutron Flux Detection System, or monitoring the QPTR, measurement of FNAH and FQ(z), or calibration of the PDMS. This surveillance ensures that the measurements obtained from use of this system accurately represents the spatial neutron flux distribution of the core. The Frequency of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months has been established, based on engineering judgment, and has been shown to be acceptable through operating experience.

REFERENCES

1.

WCAP-8648, "Excore Detector Recalibration Using Quarter-core Flux Maps," June 1976.

2.

WCAP-11618, "MERITS Program-Phase II, Task 5, Criteria Application,"

including Addendum 1 dated April, 1989.

3.

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

Watts Bar-Unit 1 Technical Requirements B 3.3-9 Revision 46 09/20/2010

Seismic Instrumentation B 3.3.4 B 3.3 INSTRUMENTATION B 3.3.4 Seismic Instrumentation BASES BACKGROUND The seismic instrumentation is made up of several instruments such as accelerometers, an accelerograph, recorders, etc. These instruments are placed in several appropriate locations throughout the plant in order to provide data on the seismic input to containment, data on the frequency, amplitude and phase relationship of the seismic response of the containment structure, and data on the seismic input to other Seismic Category I structures (Ref. 1).

The seismic instrumentation is used to promptly determine the nature and severity of a seismic event and to predict the impact (i.e., potential for damage) on nuclear power plant features which are important to safety. This is required to permit comparison of the measured response to that used in the design basis for the unit to determine if plant shutdown is required pursuant to Appendix A of 10 CFR Part 100. The instrumentation is consistent with the recommendations of Reference 1.

The original seismic instrumentation was replaced with state of the art digital instrumentation in order to permit application of EPRI OBE exceedance criteria delineated in References 4 and 5. Use of these criteria is permitted by Reference 6 provided that upgraded instrumentation is used. The replacement instrumentation is capable of recording a seismic event and performing appropriate analyses of the recorded data to provide an immediate basis for determining whether an OBE exceedance has occurred. Reference 6 directs that this information must be evaluated within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months after an event and a walkdown of critical plant features must be accomplished within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months after an event in order to make a determination as to whether a plant shutdown in warranted.

APPLICABLE SAFETY ANALYSES The OPERABILITY of the seismic instrumentation ensures that sufficient capability is available to promptly determine the magnitude of a seismic event and to determine the impact on those features important to safety. This capability is required to permit comparison of the measured response to that used in the design basis for the unit to determine if plant equipment inspection is required pursuant to Appendix A of 10 CFR part 100 prior to restart. Seismic risks which appear as dominant sequences in PRAs occur for very severe earthquakes with magnitudes which are a factor of two or three above the Safe Shutdown Earthquake and Design Basis Earthquake. The Seismic Instrumentation System was not designed to function or to provide comparative information for such severe earthquakes. This instrumentation is more pertinent to determining the need to shut down following a seismic event and the ability to restart the plant after seismic events which are not risk contributors, and is therefore not of prime importance in risk dominant sequences (Ref. 2).

(continued)

Revision 19 10/12/99 Watts Bar-Unit 1 Technical Requirements B 3.3-10

Hydrogen Monitors B 3.3.8 BASES REFERENCES

1.

10 CFR 50.44, "Standards for Combustible Gas Control System in Light-Water-Cooled Power Reactors," October 16, 2003.

2.

10 CFR 50.2, "Definition of Safety Related Structures, Systems, and Components."

3.

Regulatory Guide 1.97, Revision 2, December 1980, "Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident."

4.

TSTF-447, Revision 1, "Elimination of Hydrogen Recombiners and Change to Hydrogen and Oxygen Monitors."

5.

Commitment (NCO080031 001) made in TVA's letter dated September 4, 2008, to maintain a hydrogen monitoring system capable of diagnosing beyond design basis accidents.

Watts Bar-Unit I Technical Requirements B 3.3-29 Revision 45

PDMS B 3.3.9 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 Reference 1. 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). In 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 Incore Detector System (Technical Requirement 3.3.3).

The PDMS incore power distribution measurement can be used to determine the most limiting core peaking factors, F NAH, the Nuclear Enthalpy Rise Hot Channel Factor (Technical Specification 3.2.2) and FQ(z), the Heat Flux Hot Channel 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 Specification 3.3.1), monitoring the QUADRANT POWER TILT RATIO (QPTR) (Technical Specification 3.2.4), and verifying the position of a rod with inoperable position indicators (Technical Specification 3.1.8).

The PDMS requires information on current plant and core conditions in order to determine the core power distribution using the core peaking factor measurement and measurement uncertainty methodology described in Reference 1. The OPERABILITY 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 Incore 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 F NAH. Similarly, either the PDMS or the Movable Incore Detector System may be used for verifying the position of a rod with inoperable position indicators, but only the PDMS must satisfy OPERABILITY requirements prior to this function.

(continued)

Revision 46 09/20/2010 Watts Bar-Unit 1 Technical Requirements B 3.3-30

PDMS B 3.3.9 Bases (continued)

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

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

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

Maintaining the minimum number of instrumentation channel inputs ensures the uncertainty is bounded by the uncertainty methodology. Similarly, when THERMAL POWER is less than 25% RTP, then the accuracy of the adjustment provided by the CETs to the measured PDMS power distribution may not be bounded by the uncertainties documented in Reference 1.

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

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

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

(continued)

Revision 46 09/20/2010 Watts Bar-Unit I Technical Requirements B 3.3-31

PDMS B 3.3.9 BASES (continued)

TECHNICAL TSR 3.3.9.1 SURVEILLANCE REQUIREMENTS Performance of the CHANNEL CHECK ensures that a gross instrumentation failure has not occurred. A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a similar parameter on other channels. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value. Significant deviations between the two instrument channels could be an indication of excessive instrument drift in one of the channels.

A CHANNEL CHECK will detect gross channel failure, thus it is a key to verifying the instrumentation continues to operate properly between each CHANNEL CALIBRATION. Agreement criteria are determined by the unit staff, based on a combination of the channel instrument uncertainties, including isolation, indication, and readability. If a channel is outside the criteria, it may be an indication that the sensor or the signal processing equipment has drifted outside its limit. If the channels are within the criteria, it is an indication that the channels are OPERABLE.

The Frequency of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is sufficient considering the PDMS provides automatic validation of the channel inputs and either discards the inoperable channel input or declares itself inoperable, but at the same time ensures that the required channel inputs to the PDMS are manually verified to be valid within a reasonable time frame prior to using the PDMS to obtain an incore power distribution measurement.

TSR 3.3.9.2 Verification by administrative means of the surveillance requirements required elsewhere ensures the instrumentation channels satisfy nominal accurancy and reliability for power operation. Many of these surveillance requirements are CHANNEL CALIBRATIONS.

CHANNEL CALIBRATIONS are typically performed every 18 months, or approximately at every refueling. CHANNEL CALIBRATION is a complete check of the instrument loop, including the sensor. The test verifies that the channel responds to a measured parameter within the necessary range and accuracy.

CHANNEL CALIBRATION must be performed consistent with the assumptions of the Watts Bar setpoint methodology. The difference between the current "as found" values and the previous test "as left" values must be consistent with the drift allowance used in the setpoint methodology.

(continued)

Watts Bar-Unit 1 B 3.3-32 Revision 46 Technical Requirements 09/20/2010

PDMS B 3.3.9 BASES (continued)

TECHNICAL Nine notes modify the instrumentation channels specified in Table 3.3.9-1.

SURVEILLANCE REQUIREMENTS Note 1 allows the RCS Cold Leg Temperature to come from either the Narrow (continued) or Wide Range RTDs.

Note 2 allows up to three parameters to be used for reactor power input into PDMS, but BEACONTM will only accept two options at any one time.

Note 3 allows the control bank position input to come from either the Demand Position Indication or the average of the individual Pod Position Indications.

Note 4 clarifies that neutron detectors are excluded from the CHANNEL CALIBRATION. The CHANNEL CALIBRATION for the power range neutron detectors consists of a normalization of the detectors based on a power calorimetric performed above 15% RTP in accordance with SR 3.3.1.2.

Note 5 clarifies that the SR 3.3.1.10 requirements specified in this table apply only to the Narrow Range RTDs.

Note 6 clarifies that the SR 3.3.3.2 requirements specified in this table apply only to the Wide Range RTDs.

Note 7 clarifies that the calorimetric heat balance adjustment is not applicable to the average RCS Loop &T power input.

Note 8 clarifies that the CHANNEL CALIBRATION requirements specified in this table apply only to the RCS Loop AT power.

Note 9 clarifies that the CHANNEL CALIBRATION requirements specified in this table apply only to the CET temperatures.

TSR 3.3.9.3 The PDMS must be calibrated to a flux map obtained above 25% RTP to ensure the accuracy of the calibration data set which is generated from the incore flux map, CETs, and other input channels. The initial calibration in each fuel cycle must utilize incore flux measurements from at least 75% of the detector thimbles, with at least two incore thimbles in each core quadrant. The incore flux measurements in combination with at least the minimum channel inputs from Table 3.3.9-1 are used to generate the calibration data set, including nodal calibration factors and the thermocouple mixing factors. Subsequent PDMS calibrations require only > 50% of the detector thimbles, with at least two incore thimbles in each core quadrant.

Watts Bar-Unit 1 B 3.3-33 Revision 46 Technical Requirements 09/20/2010

PDMS B 3.3.9 BASES (continued)

TECHNICAL SURVEILLANCE REQUIREMENTS (continued)

The subsequent PDMS calibration frequency is 31 Effective Full Power Days (EFPD) when the CET chess knight move pattern is not satisfied. The CET chess knight move pattern is satisfied when every interior core location (fuel 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 calibration is not required to be performed until 31 EFPD after the CET chess knight move pattern is not satisfied.

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

REFERENCES

1.

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

2.

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

3.

WCAP-1 1618, "MERITS Program-Phase II, Task 5, Criteria Application,"

including Addendum I dated April, 1989.

(continued)

Watts Bar-Unit 1 Technical Requirements B 3.3-34 Revision 46 09/20/2010

Flood Protection Plan B 3.7.2 This Page Intentionally Left Blank Watts Bar-Unit 1 Technical Requirements B 3.7-11 Revision 17 5/25/99

Snubbers B 3.7.3 B 3.7 PLANT SYSTEMS B 3.7.3 Snubbers BASES BACKGROUND Component standard supports, are those metal supports which are designed to transmit loads from the pressure-retaining boundary of the component to the building structure. Although classified as component standard supports, snubbers require special consideration due to their unique function. Snubbers are either operated hydraulically or mechanically, depending on the nature of the support needed. They are designed to provide no transmission of force during normal plant operations, but function as a rigid support when subjected to dynamic transient loadings. Therefore, snubbers are chosen in lieu of rigid supports where restricting thermal grow during normal operation would induce excessive stresses in the piping nozzles or other equipment. The location and size of the snubbers are determined by stress analysis. Depending on the design classification of the particular piping, different combinations of load conditions are established. These conditions combine loading during normal operation, seismic loading and loading due to plant accidents/transients to four different loading sets. These loading sets are designated: normal, upset, emergency, and faulted condition. The actual loading included in each of the four conditions, depends on the design classification of the piping. The calculated stresses in the piping and other equipment, for each of the four conditions, must be in conformance with established design limits.

Supports for pressure-retaining components are designed in accordance with the rules of the ASME Boiler and Pressure Vessel Code,Section III, Division 1 (Ref. 1). The combination of loadings for each support, including the appropriate stress levels, meet the criteria of Regulatory Guide 1.124, "Design Limits and Loading Combinations for Class 1 Linear-Type Component Supports" (Ref. 2),

and Regulatory Guide 1.130, "Design Limits and Loading Combinations for Class 1 Plate-and-Shell-Type Component Supports" (Ref. 3).

(continued)

Watts Bar-Unit 1 Technical Requirements B 3.7-12 Revision 47 10/08/10

Snubbers B 3.7.3 BASES (continued)

APPLICABLE SAFETY ANALYSIS Pipe and equipment supports, in general, are not directly considered in designing the accident sequences for theoretical hazard evaluations. Further, various Probabilistic Risk Assessment (PRA) studies have indicated that snubbers are not of prime importance in a risk significant sequence (Ref. 4 and 5). Therefore, the function of the snubbers is not essential in mitigating the consequences of a DBA or transient (Ref. 6).

TR TR 3.7.3 requires that all snubbers shall be OPERABLE. Individual snubbers may be removed from service for functional testing within the limits established herein without violating these requirements, although Required Actions and Completion Times still apply.

APPLICABILITY The OPERABILITY of the snubbers is required in MODES 1, 2, 3, and 4. For MODES 5 and 6, the OPERABILITY is limited to the snubbers located on those systems which need to be OPERABLE in MODES 5 and 6.

ACTIONS A.1 If one or more required snubbers are removed from service, the supported system must be declared inoperable immediately and the appropriate LCO entered for that system.

B.1 and B.2 If one or more required snubbers are discovered inoperable, then the supported system must be declared inoperable immediately and the appropriate LCO entered for that system.

Discovery of any snubber as inoperable requries the performance of an engineering evaluation per Table 3.7.3-5 during the 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months to:

a)

Determine the cause of the failure As a result of this evaluation, the need for testing other snubbers will be considered. The results from the testing will be used to consider expanded functional testing and cause examination with consideration of manufacturing and design deficiency. It should be noted that the testing must be independent and not combined with TSR 3.7.3.3.

(continued)

Watts Bar-Unit 1 Technical Requirements B 3.7-13 Revision 47 10/08/10

Snubbers B 3.7.3 BASES ACTIONS B.1 and B.2 (continued) b)

Determine the impact on the supported component This evaluation shall determine if the inoperable snubber has adversely affected the attached component.

The 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months is based on engineering experiences and is reasonable, considering the time it will take to identify the problem and take the proper corrective actions. This requirement is considered met for those snubbers rendered inoperable by removal for functional testing by the generic engineering evaluation included in Reference 9.

Another alternative is to perform an engineering evaluation to demonstrate inoperable snubber(s) do not impact the OPERABILITY of the supported system for the existing plant condition. (Reference 10)

C.1 If Required Actions under Condition A or Condition B are not met within the required Completion Time, a Service Request shall be written to generate a Problem Evaluation Report for the occurrence.

TECHNICAL SURVEILLANCE REQUIREMENTS The TSRs are preceded by three Notes. Note 1 states that the snubber inservice inspection program shall be carried out in accordance with the requirements in Tables 3.7.3-1, 2 and 3. This represents an enhanced snubber inservice inspection program compared to the Inservice Inspection Program which stipulates inservice inspection in accordance with ASME section Xl. The snubber inservice inspection program includes the requirements of Generic Letter 90-09, "Alternative Requirements for Snubber Visual Inspection Intervals and Corrective Actions." ASME Section XI, 1989 Edition, Subsection IWF-5300(a) and (b) require that inservice examinations of snubbers, using the VT-3 visual examination method described in IWA-2213, and inservice tests of snubbers be performed in accordance with the first Addenda to ASME/ANSI OM-1987, Part 4. Note 2 requires repair or replacement of snubbers which fail inspection, and testing of repaired snubbers before installation. Note 3 indicates that a "snubber type," as used in this TR, is determined by the design and manufacturer, but not by size.

(continued)

Watts Bar-Unit 1 Technical Requirements B 3.7-14 Revision 47 10/08/10

Snubbers B 3.7.3 BASES TECHNICAL SURVEILLANCE REQUIREMENTS (continued)

TSR 3.7.3.1 TSR 3.7.3.1 comprises a visual inspection of the snubbers. A pre-fuel load visual inspection and functional test has been performed on each snubber using the acceptance criteria listed in this TSR. The baseline considers that the snubbers have experienced thermal cycling and normal operating service as a result of previous hot functional testing. The initial inservice inspection must be performed on the snubbers prior to completion of the first refueling outage. The frequency of subsequent surveillances depends on the number of snubbers found inoperable from each previous inspection as provided in Table 3.7.3-2 and the Inservice Inspection Program. The acceptance criteria and the remedial ACTIONS are listed in Table 3.7.3-1.

The visual inspections are designed to detect obvious indications of inoperability of the snubbers. Removal of insulation or direct contact with the snubbers is not required initially. However, suspected causes of inoperability are to be investigated and all snubbers of the same type and all snubbers subjected to the same failure mode are to be inspected more frequently.

Until proven otherwise by functional testing, mechanical and hydraulic snubbers are considered OPERABLE, unless they are disconnected at either end, experienced gross deformation of the snubber or structural support, or the hydraulic fluid level is empty for hydraulic snubbers.

The visual inspection frequency is based upon the number of unacceptable snubbers found during the previous inspection. Therefore, the required inspection intervals vary inversely with the number of inoperable snubbers found during an inspection. If a snubber fails the visual acceptance criteria, the snubber is declared unacceptable and cannot be declared OPERABLE via functional testing. However, if the cause of rejection is understood and remedied for that type of snubber and for any other type of snubbers, that may be generically susceptible, and OPERABILITY verified by testing, that snubber may be reclassified acceptable for the purpose of establishing the next surveillance interval.

Snubbers maybe categorized, according to accessibility, as noted in the Note to Table 3.7.3-2. The accessibility of each snubber is determined based on radiation level as well as other factors such as temperature, atmosphere, location, etc. The recommendations of Regulatory Guide 8.8, "Information Relevant to Maintaining Occupational Radiation Exposure as Low as Practicable" (Ref. 7) and Regulatory Guide 8.10, "Operation Philosophy for Maintaining Occupational Radiation Exposure as Low as Practicable" (Ref. 8),

are considered in planning and implementing the visual inspection program.

(continued)

Watts Bar-Unit 1 Technical Requirements B 3.7-15 Revision 47 10/08/10

Snubbers B 3.7.3 BASES TECHNICAL TSR 3.7.3.2 SURVEILLANCE REQUIREMENTS TSR 3.7.3.2 comprises the inspection of all snubbers attached to systems that (continued) have experienced unexpected, potentially damaging transients. The potential impact of the transients is assessed by reviewing operating data and by visually inspecting the associated systems. The review and the inspection must be performed within six months of the event. In addition to the inspection, the freedom-of-motion of the mechanical snubber(s) is verified in accordance with Table 3.7.3-3.

TSR 3.7.3.3 TSR 3.7.3.3 comprises the functional testing of hydraulic and mechanical snubbers. The testing for these snubbers have been separated into two sample plans. Sample plan A (10%) is used for the steam generator hydraulic snubbers based on the small population. Sample plan B (37) is used for mechanical snubbers based on the large population. The plans, when used in combination, are a conservative approach versus using only the sample plan B for the entire population.

Snubber functional testing is performed prior to completion of each refueling outage. This frequency is based on engineering experience and is reasonable for testing of a representative sample of snubbers. Credit may be taken toward meeting minimum outage testing requirements for mechanical snubbers functionally tested within the refueling cycle. Snubbers may be removed from service for functional testing in Modes 1 through 4 provided that the following administrative controls are implemented:

1.

Required Actions and Completion Times must be met.

2.

No more than one snubber may be removed from service at a time on any line and attached piping which is analyzed as a seismic subsystem.

Multiple snubbers may be removed for testing simultaneously only if separated by two or more seismic anchors.

3.

Snubbers on trained systems or portions of systems may be removed only on the train which is undergoing maintenance in that work week.

Snubbers on non-trained systems or portions of systems may only be removed following a documented risk assessment. Snubbers may not be removed from service for testing on one train of a system while the other train has been declared inoperable for any reason.

4.

Snubbers adjacent to equipment nozzles may not be removed for testing except in Modes 5 and 6. In determining the applicability of this limitation, engineering judgment must be used regarding the placement of the snubber relative to the nozzle, the routing of the affected piping, and any other supports available to protect equipment function.

(continued)

Watts Bar-Unit 1 B 3.7-16 Revision 47 Technical Requirements 10/08/10

Snubbers B 3.7.3 BASES TECHNICAL SURVEILLANCE REQUIREMENTS (continued)

TSR 3.7.3.4 The TSR is preceded by three Notes which underline the need for considering service life of sub-components and to replace these sub-components before the end of the respective service lives. The replacement of sub-components must be documented and the documentation retained for further reference. TSR 3.7.3.4 addresses the monitoring of the service life of the snubbers. The requirement to monitor the snubber service life is included to ensure that the snubbers periodically undergo a performance evaluation in view of their age and operating conditions. The expected service life is established by the manufacturer and is based on operating experience with critical snubber parts such as seals and springs in a radiation environment. The every refueling outage Frequency is based on engineering experience and is reasonable for the verification service life.

REFERENCES

1.

ASME Boiler and Pressure Vessel Code,Section III and XI.

2.

Regulatory Guide 1.124, "Design Limits and Loading Combinations for Class 1 Linear-Type Component Supports".

3.

Regulatory Guide 1.130, "Design Limits and Loading Combinations for Class 1 Plate-and-Shell-Type Component Supports".

4.

"Zion Probabilistic Safety Study", Commonwealth Edison Company, September 1981.

5.

"Millstone Unit 3 Probabilistic Safety Study," North-east Utilities Company, August 1983.

6.

NRC Staff Review of Nuclear Steam Supply System Vendor Owners Groups' Application of The Commission's Interim Policy Statement Criteria to Standard Technical Specifications, Attachment to letter dated May, 1988 from T.E. Murley, NRC to W.S. Wilgus, Chairman The B&W Owners Group.

7.

Regulatory Guide 8.8, "Information Relevant to Maintaining Occupational Radiation Exposure as Low as Practicable."

8.

Regulatory Guide 8.10, "Operating Philosophy for Maintaining Occupational Radiation Exposure as Low as Practicable."

9.

SA/SE WBPLCE-97-028-0, RIMS T28970829803.

10.

Screening Review WBPLCE-02-003-0.

11.

Screening Review WBPLCE-1 0-009-0.

Watts Bar-Unit 1 Technical Requirements B 3.7-17 Revision 47 10/08/10

Sealed Source Contamination B 3.7.4 B 3.7 PLANT SYSTEMS B 3.7.4 Sealed Source Contamination BASES BACKGROUND A sealed source is any byproduct, source, or special nuclear material that is encased in a capsule designed to prevent leakage or escape of the material (Ref. 1). Sealed sources are classified into three groups according to their use (sources in use, not in use, and startup sources and fission detectors) and may contain beta, gamma, or alpha emitting material. The limitations on removable contamination for sources requiring leak testing, including alpha emitters, is based on Reference 2. Those sources that are frequently handled are required to be tested more often than those which are not. Sealed sources which are continuously enclosed within a shielded mechanism (i.e. sealed sources within radiation monitoring, excore fission detector assemblies or boron measuring devices) are considered to be stored and need not be tested unless they are removed from the shielded mechanism.

APPLICABLE SAFETY ANALYSES The sealed source contamination requirement ensures that leakage from sealed sources will not exceed allowable intake values. This TR is important to the safety of plant personnel, however it is not required to mitigate the consequences of a DBA or transient (Ref. 3).

TR TR 3.7.4 requires that the removable contamination shall be less than 0.005 microcuries for each sealed source containing the following radioactive material:

a.

Greater than 100 microcuries of beta and/or gamma emitting material; or

b.

Greater than 5 microcuries of alpha emitting material.

APPLICABILITY Since the limits on the removable contamination for each sealed source containing radioactive material are not MODE dependent, this TR is applicable at all times.

(continued)

Watts Bar-Unit 1 Technical Requirements B 3.7-18 09/30/95

ML11325A322

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Subject:

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