ML18038B630

Annual Operating Rept Jan-Dec 1995. W/960228 Ltr

ML18038B630
Person / Time
Site:	Browns Ferry
Issue date:	12/31/1995
From:	Salas P TENNESSEE VALLEY AUTHORITY
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 9603040199
Download: ML18038B630 (124)

Text

CATEGORY 1 REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)

DOCKET ¹ FACIL:50-259 Browns Ferry Nuclear Power Station, Unit 1, Tennessee 05000259 50-260 Browns Ferry Nuclear Power Station, Unit 2, Tennessee 05000260 50-296 Browns Ferry Nuclear Power Station, Unit 3, Tennessee 05000296 AUTH. NAME AUTHOR AFFIZIATION SALASPP. Tennessee Valley Authority 'I (

RECIP.NAME RECIPIENT'AFFILIATION "Browns uclear Plant Annual Operating Rept Jan-Dec, j

SUBJECT:

F 1995." /960228 r.

DISTRIBUTION CODE: IE47D COPIES RECEIVED:LTR I ENCL SIZE: SO T TITLE: 50.59 Annual Report of Changes, Tests or Experiments Made W out E

NOTES:

RECIPIENT COPIES RECIPIENT COPIES ID CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL PD2-3-.PD 1 0 WILLIAMSPJ. 1 1 INTERNAL: ACRS 6 6 FILE CENTER 1 1 RGN2 FILE 01, 1 1

'I EXTERNAL: NOAC 1 1 NRC PDR 1 1 D

0 U

N NOTE TO ALL "RIDS" RECIPIENTS:

PLEASE HELP US TO REDUCE WASTE! CONTACT THE DOCUMENT CONTROL DESK, ROOM OWFN SD-5(EXT. 415-2083) TO ELIMINATE YOUR NAME FROM DISTRIBUTION LISTS FOR DOCUMENTS YOU DON'T NEED/

TOTAL NUMBER OF COPIES REQUIRED: LTTR 12 ENCL 'l

Tennessee Valley Authority. Post Office Box 2000. Decatur, Alabama 35609 February 28, 1996 10 CFR 50. 59 (b) (2)

U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555 Gentlemen:

In the Matter of Docket Nos. 50-259 Tennessee Valley Authority 50-260 50-296 BROWNS FERRY NUCLEAR PLANT (BFN) - UNITS 1, 2, AND 3 ANNUAL OPERATING REPORT FOR JANUARY 1, 1995 THROUGH DECEMBER 31, 1995 The purpose of this letter is to transmit, in the enclosure, the BFN Annual Operating Report from January 1, 1995 through December 31, 1995. TVA submits this report according to the BFN Technical Specifications (TS) 6.9.1.2, 6.9.2.1, and 10 CFR 50.59(b)(2).

The enclosed report contains a summary of plant conditions, occupational exposure data, challenges to or failures of main steam relief valves, and reactor vessel fatigue usage. The report also details gaseous and liquid releases. This information is required by TS 6.9.1.2 and 6.9.2.1. Further, this report contains safety evaluations for revisions to the Updated Final Safety Analysis Report, procedures (new or revisions), special operating conditions, special tests, temporary alteration; and, plant modifications that were fi'eld completed during this reporting period. This information is required by 10 CFR 50.59(b)(2).

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U. S. Nuclear Regulatory Commission Page 2 February 28, 1996 There are no commitments contained in this letter. If you have any questions, please contact Pedro Salas at (205) 729-2636.

Sincerel Ped o Salas Manager of Site Licensing Enclosure cc (Enclosure):

Mr. Mark S. Lesser, Branch Chief U. S. Nuclear Regulatory Commission Region II 101 Marietta Street, NW, Suite 2900 Atlanta, Georgia 30323 NRC Resident Inspector Browns Ferry Nuclear Plant 10833 Shaw Road Athens, Alabama 35611 Mr. J. F. Williams, Project Manager U.S. Nuclear Regulatory Commission One White Flint, North 11555 Rockville Pike Rockville, Maryland 20852

ENCLOSURE TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT (BFN)

UNITS 1, 2, AND 3 ANNUAL OPERATING REPORT FOR JANUARY 1, 1995 THROUGH DECEMBER 31, 1995 See Attached

TENNESSEE VALLEYAUTHORITY BROWNS FERRY NUCLEAR PLANT ANNUALOPERATING REPORT January 1, 1995- December 31, 1995

'ocket Number 50-259, 50-260, and 50-296 License Number DPR-33, DPR-52, and DPR-68

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Brmets Ferry Nuclear Plant 1995 Annual Operating Report WSS~JRSSS WSSMtNCf!M~

Acronyms Listing.

Regulatory Guide 1.16,Section I.b. (I) and (2)

Operational Summary...

IOCFR50.59(b)(2) - Summary ofSafety Evaluations Core Component and Operating Limits 10 Field Completed Plant Modifications.. 14 Fire Protection Report Revisions.. 45 New Instructions/Procedure Revisions. 48 Special Operating Conditions. 50 Special Tests. 56 Temporary Alterations. 58 Technical Specification Changes... 60 Updated Final Safety Analysis Report Revisions. 62 Regulatory Guide 1.16,Section I.b. (3) 1995 Release Summary 64 Technical Specification 6.9.1.2 1995 Occupational Exposure Data. 66 Challenges to or Failures of Main Steam Relief Valves.. 71 Technical Specification 6.9.2.1 Reactor Vessel Fatigue Usage Evaluation. 74

Tennessee Valley A uth ority Brogans Ferry Nuclear Plant 1995 Annual Operating Report ~

ACRONFMS LISTING This is a list of acronyms and abbreviations used throughout the 1995 Annual Operating Report.

AC Alternating Current ACU Air Conditioning Unit ADS Automatic Depressurization System; Atmospheric Dilution System AFFF Aqueous Film Forming Foam ALARA As Low As Reasonably Achievable APRM Average Power Range Monitor ASME 'merican Society of Mechanical Engineers ATWS Anticipated Transient Without SCRAM BFN Browns Ferry Nuclear Plant BPWS Banked Position Withdrawal Sequence BTU British Thermal Units BWR -Boiling Water Reactor CAQR Condition Adverse To Quality Report CFR Code of Federal Regulations CISS Containment Isolation Status System CKV Check Valve COLR Core Operating Limits Report CRD Control Rod Drive

'RDR Control Room Design Review CRLD Change Request to a Licensing Document DBA Design Basis Accident DBE Design Basis Earthquake DC Direct Current DCN Design Change Notice DCR 'G Design Change Request Diesel Generator ECCS Emergency Core Cooling System ECN Engineering Change Notice EECW Emergency Equipment Cooling Water EFPD EQective Full Power Days ELLLA Extended Load Line Limit Analysis EMS Equipment Management System EOC End of Cycle F Fahrenheit FCR Field Change Request FCV Flow Control Valve Floor Drain Collector

Tennessee Valley Authority Browns Ferry Nuclear Plant I995AnnualOperaling Report . ACRONYMSLISTING FDCN Field Design Change Notice FI Flow Indicator FIC Flow Indicating Controller FPC Fuel Pool Cooling FSV Flow Solenoid Valve FT Flow Transmitter

.foot GE General Electric GE SIL GE Service Information Letter GEMAC General Electric Measurement and Control gpm Gallons per Minute HELB High-Energy Line Break Hg Mercury HPCI High Pressure Coolant Injection HPFP High Pressure Fire Protection HS Hand switch HWC Hydrogen Water Chemistry ICS Integrated Computer System ILRT Integrated Leak Rate Test ISI Inservice Inspection 1<V Kilovolt lbs Pounds LLRT Local Leak Rate Test LOCA Loss of Coolant Accident LPRM Local Power Range Monitor LS Level Switch LT Level Transmitter .

MCPR Minimum Critical Power Ration Moisture Element MG Motor Generator MOV Motor Operated Valve MSIV Main Steam Isolation Valve MSLRM Main Steam Line Radiation Monitor MSRV Main Steam Relief Valve MWD/ST Megawatt Days per Short Ton MWe Megawatt Electrical MWt Megawatt Thermal NESSD Nuclear Engineering Setpoint and Scaling Document NFPA National Fire Protection Association NMS Neutron Monitoring System NRC Nuclear Regulatory Commission NUMAC Nuclear Measurement Analysis and Control

Tennessee Valley Authority I"'~

Brains Ferry Nuclear Plant 1995 Annual Operating Report ACRONYMS LlSTING

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"%:.""""""UMARC Nuclear Utilities Management and Human Resources Committee NUREG Nuclear Regulatory Commission Regulation PCIOMR Preconditioning Interim Operating Management Recommendations PCIS Primary Containment Isolation System PCV Pressure Control Valve PER Problem Evaluation Report PI Pressure Indicator ppb Parts per Billion ppm Parts per Million PS Pressure Switch psi Pounds per Square Inch PT Pressure Transmitter QA Quality Assurance RBCCW Reactor Building Closed Cooling Water RbNO3 Rubidium Nitrate RCI Radiological Control Instruction RCIC Reactor Core Isolation Cooling RCW Raw Cooling Water RHR Residual Heat Removal RHRSW Residual Heat Removal Service Water RM Radiation Modifier RMOV Reactor Motor Operated Valve RMS Radiation Monitoring System RPS Reactor Protection System RPV Reactor Pressure Vessel RSW Raw Service Water RVLIS Reactor Vessel Level Instrumentation System RWCU Reactor Water Cleanup SBO Station Blackout SCFM Standard Cubic Feet per Minute SER Sequential Events Recorder; Significant Events Report (INPO)

SGTS Standby Gas Treatment System SI Surveillance Instruction SLC Standby Liquid Control SPDS Safety Parameter Display System SSP Site Standard Practice ST Special Test TA Tantalum (Symbol)

TACF Temporary Alteration Control Form TI Technical Instruction Tennessee Valley A uthori ty Brains Ferry Ãuclear Plant 1995 Annual Operating Report ACRONYMS LISTING tAI:

TIP Traversing Incore Probe TPM Thermal Power Monitor TRS Temperature Recorder Switch TVA Tennessee Valley Authority UFSAR Updated Final Safety Analysis Report UPS Uninterruptible Power Supply USQD Unreviewed Safety Question Determination V Volt; Vanadium VAC Volts Alternating Current VDC Volts Direct Current VHF Very High Frequency

Tennessee Valley Authority Brains Ferry Nuclear Plant 1995Annual Operating Report OPERA TIOlVAI.

SUMMARY

1995 OPERATIONAL

SUMMARY

Tennessee Valley Authority Brogans Ferry Nudear Plant

'PERATIONAL kM'-'~ '4" 1995 Annual Operating Report

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SUMMARY

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Unit 1 remains on administrative hold to resolve various Tennessee Valley Authority (TVA) and Nuclear Regulatory Commission (NRC) concerns.

UNIT 2 On January 1, 1995, the unit's power level was at full power (3293 MWt and 1115 MWe).

On February 9, 1995, at 0027 hours3.125e-4 days <br />0.0075 hours <br />4.464286e-5 weeks <br />1.02735e-5 months <br />, the reactor scrammed due to a turbine trip. The turbine trip was caused by a ground to the main generator. The reactor was again critical on February 10 at 0135 hours0.00156 days <br />0.0375 hours <br />2.232143e-4 weeks <br />5.13675e-5 months <br /> and at 100% power on February 23 at 2300 hours0.0266 days <br />0.639 hours <br />0.0038 weeks <br />8.7515e-4 months <br />.

On March 30, 1995, at 1800 hours0.0208 days <br />0.5 hours <br />0.00298 weeks <br />6.849e-4 months <br />, the reactor scrammed due to low scram pilot air header pressure caused by the operation of a handswitch out of sequence during the performance of an SI. The reactor was again critical on April 2 at 0139 hours0.00161 days <br />0.0386 hours <br />2.29828e-4 weeks <br />5.28895e-5 months <br /> and at 100% power on L April 5 at 0001 hours1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

On August 19, 1995, at 0200 hours0.00231 days <br />0.0556 hours <br />3.306878e-4 weeks <br />7.61e-5 months <br />, the reactor scrammed due to a turbine trip on low condenser vacuum. The reactor was again critical on August 20 at 0514 hours0.00595 days <br />0.143 hours <br />8.498677e-4 weeks <br />1.95577e-4 months <br /> and at 100%

power on August 26 at 0700 hours0.0081 days <br />0.194 hours <br />0.00116 weeks <br />2.6635e-4 months <br />.

On December 31, 1995,,Unit 2 was at full power.

UNIT 3 On November 19, 1995, at 0922 hours0.0107 days <br />0.256 hours <br />0.00152 weeks <br />3.50821e-4 months <br />, the reactor startup commenced. The reactor was declared critical at 1203 hours0.0139 days <br />0.334 hours <br />0.00199 weeks <br />4.577415e-4 months <br /> and continued the power ascension test program. On November 25, 1995, the plant closed breaker. The reactor reached 100% power on December 9, at 0133 hours0.00154 days <br />0.0369 hours <br />2.199074e-4 weeks <br />5.06065e-5 months <br />.

On December 31, 1995, Unit 3 was at full power.

Tennessee Valley Authority Brogans Ferry Nuclear Plant l995 Annual Operating Report OPERATIONAL

SUMMARY

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Docket No.: 50-259 OPERATING STATUS

1. Unit Name: Browns Ferry Unit One

2. Reporting Period: Calendar Year 1995

3. Licensed Thermal Power (MWt): 3293

4. Nameplate Rating (Gross MWe): 1152

5. Design Electrical Rating (Net MWe): 1065 6.- Maximum Dependable Capacity (Gross MWe): 0

7. Maximum Dependable Capacity (Net MWe): 0

8. IfChanges Occur in Capacity Ratings (Items Number 3 Through 7)

Since Last Report, Give Reason: N/A

9. Power Level to Which Restricted, ifany (Net MWe): 0

10. Reason for Restrictions, ifany: Administrative Hold 1995 Cumulative*

HoursinRe ortin Period 0 95743 12 Hours Reactor Was Critical ,0 59521

13. Reactor Reserve Shutdown Hours 0 6997 14 Hours Generator On Line 0 58267

15. Unit Reserve Shutdown Hours 0 0 16 Gross Thermal Generation 0 168066787

17. Gross Electrical Generation h 0 55398130 18 Net Electrical Generation h 0 53796427

19. Unit Service Factor 0 60.9 20 Unit Availabilit Factor 0 60.9

21. Unit Ca acit Factor CNet 0 52.8 22 Unit Ca acit Factor ERNet 0 52.8

23. Unit Forced Outa e Rate 0 25.6

24. Shutdowns Scheduled Over Next 6 Months (Type, Date, and Duration of Each):

N/A

25. IfShutdown at End of Reporting Period, Estimated Date of Startup:

To Be Determined

• Excludes hours under administrative hold (June 1, 1985 to present)

Tennessee Valley Authority Brogans Ferry Nuclear Plant l995 Annual Operating Report OPERA TIOiVAL

SUMMARY

Docket No.: 50-260 OPERATING STATUS

l. Unit Name: Browns Ferry Unit Two

2. Reporting Period: Calendar Year 1995

3. Licensed Thermal Power (MWt): 3293

4. Nameplate Rating (Gross MWe): 1152

5. Design Electrical Rating (Net MWe): 1065

6. Maximum Dependable Capacity (Gross MWe): 1098.4

7. Maximum Dependable Capacity (Net MWe): 1065

8. IfChanges Occur in Capacity Ratings (Items Number 3 Through 7)

Since Last Report, Give Reason: N/A

9. Power Level to Which Restricted, ifany (Net MWe): N/A

10. Reason for Restrictions, ifany: N/A 1995 Cumulative*

11. Hours in Re ortin Period 8760 130831

12. Hours Reactor Was Critical 8652 90779

13. Reactor Reserve Shutdown Hours 0 14200

14. Hours Generator On Line 8628 88488

15. Unit Reserve Shutdown Hours 0 0

16. Gross Thermal Generation 28171417 259340332

17. Gross Electrical Generation 9422140 86165318

18. Net Electrical Generation h 9197035 83786871

19. Unit Service Factor 98.5 67.6

20. Unit Availabilit Factor 98.5 67.6

21. Unit Ca acit Factor C Net 98.6 60.1

22. Unit Ca acit Factor ER Net 98.6 60.1

23. Unit Forced Outa e Rate 1.5 16.0 r

24. Shutdowns Scheduled Over Next 6 Months (Type, Date, and Duration of Each):

Refueling Outage, March 23, 1996, 35 days

25. IfShutdown at End of Reporting Period, Estimated Date of Startup:

N/A

• Excludes hours under administrative hold (June 1, 1985 to May 24, 1991)

Tennessee Valley A uthori ty Browns Ferry Nuclear Plant Docket No.: 50-296 OPERATING STATUS

1. Unit Name: Browns.Ferry Unit Three

2. Reporting Period: Calendar Year 1995

3. Licensed Thermal Power (MWt): 3293

4. Nameplate Rating (Gross MWe): 1152

5. Design Electrical Rating (Net MWe): 1065

6. Maximum Dependable Capacity (Gross MWe): 0

7. Maximum Dependable Capacity (Net MWe): 0

8. IfChanges Occur in Capacity Ratings (Items Number 3 Through 7)

Since Last Report, Give Reason: N/A

9. Power Level to Which Restricted, ifany (Net MWe): N/A

10. Reason for Restrictions, ifany: N/A 1995 Cumulative~

Hours in Re ortin Period 1020 74075 12 Hours Reactor Was Critical 989 46295

13. Reactor Reserve Shutdown Hours 0 5150 14 Hours Generator On Line 811 45005

15. Unit Reserve Shutdown Hours 0 0 16 Gross Thermal Generation h 2402014 134270281

17. Gross Electrical Generation h 790430 44264190 18 Net Electrical Generation h 764618 42878627

19. Unit Service Factor 79.5 60.8 20 Unit Availabilit Factor 79.5 60.8

21. Unit Ca acit Factor CNet 70.4 54.4

22. Unit Ca acit Factor ERNet 70.4 54.4

23. Unit Forced Outa e Rate 0.0 21.3

24. Shutdowns Scheduled Over Next 6 Months (Type, Date, and Duration of Each):

N/A

25. IfShutdown at End of Reporting Period, Estimated Date of Startup:

To Be Determined

• Excludes hours under administrative hold (June 1, 1985 to November 19, 1995)

Tennessee Valley Authority . SI9fkfARYOF Brogans Feny Nuclear Plant SAFETY EVALUA TIONS FOR 1995 Annual Operating Report CORE COMPONENTS AND OPERA TING LIMITS

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SUMMARY

OF SAFETY EVALUATIONS FOR CORE COMPONENTS AND-OPKRATING LIMITS

Tennessee Valley Authority

SUMMARY

OF Broivns Ferry Nuclear Plant SAFETY EVALUATIONSFOR

/995AnnualOperatingReport 'ORECOMPONENTSANDOPERATINGLIMITS Cycle 6 operated for approximately 4 months prior to being prematurely shutdown on March 9, 1985. This cycle experienced a total burnup of only 1658 MWD/MT(1504 MWD/ST) or 70 EFPDs. Consequently, the GE7B fuel loaded fresh into Cycle 6 can still be considered as essentially mesh fuel for Cycle 7.

The Cycle 7 GE11 bundles, Cycle 6 GE7B bundles, and a single once4urnt (Cycle 5) Lead Test Assembly (LTA) bundle are all of the barrier cladding design with no Preconditioning Interim Operating Management Recommendation (PCIOMR) restrictions. The remaining fuel in the Cycle 7 core does not contain barrier cladding and all PCIOMR constraints remain in effect for these bundles.

The effects of the extended shutdown were analyzed by General Electric (GE) and corrections for both the cold and hot design eigenvalues were provided. These corrected eigenvalues were utilized in the appropriate cycle analyses.

BFN Unit 3 Cycle 7 is analyzed for Extended Load Line LimitAnalysis, Increased Core Flow, Final Feedwater Temperature Reduction, and Feedwater Heaters Out of Service. The cycle is also analyzed for Banked Position Withdrawal Sequence (BPWS) rod movement. The BPWS procedure must be followed in order to stay within the licensed Rod Drop Accident design basis.

Cycle 7. is designed for moderate spectral shiA operation. Spectral shift can extend full power operation by increasing the void content (spectrum hardening) during the first part of the cycle which increases plutonium production in the upper part of the core. Spectiuin hardening is enhanced with operation at lower Qow rates and by using rod patterns to obtain more bottom peaked power distributions.

A total of 37 control blades are being replaced this outage. These are in the central A2 sequence locations and make up the primary control cells used in CCC operation. The replacement blades are of the BWR/6 design modified with new low cobalt pins and rollers sized to fit BWR/4 lattice dimensions. This modification replaces both the top and bottom pins and rollers in the control, blade. This modified BWR/6 blade design was first installed and used at BFN in the Unit 2 Cycle 7 reactor core and has been previously evaluated and found to be safe for use in all BFN units.

An additional 124 exposed original equipment blades have also been modified for radiation as low as reasonably achievable (ALARA)considerations. These blades have been modified to use low~balt content spacer pads in lieu of the original rollers at the top of each control blade. The rollers at the bottom are unaffected by this modification. This is the first application of this modification to a BFN unit, but a separate evaluation was

~

previously performed to allow use in all BFN units.

'I A total of 24 local power range monitor (LPRM) strings are being replaced this outage with new LPRMs of the BE NA-300 design. This design was first installed at BFN for use in Unit 2 Cycle 7. The NA-300 LPRM has been evaluated and found to be safe for use in all BFN units.

The reactivity basis for the core requires that the core shall be capable of being made subcritical at any time or at any core condition with the highest worth control rod fully withdrawn. Shutdown margin analyses were performed by GE as part of the reload core design analysis for the Cycle 7 loading pattern. A shutdown margin demonstration test willalso be performed during Cycle 7 startup as required by Technical Specifications to verify that adequate shutdown margin exists. The reactivity effects associated with the extended shutdown have been included in the shutdown and reload analyses.

Tennessee Valley Authority SU8SlfARY OF Browns Feny Nuclear Plant SAFETY EVALUA TIONS FOR

' 'PREP"' '.

1995 Annual Operating Report P~ ' '"'w' ~' '" ~' "'" 's ~" '"'

FIELD COMPLETED PLANTMODIFICATIONS

'" " 5" "~~k' P2'~~ui3 1995 SUM Y OF SAFETY EVALUATIONS FOR FIELD COMPLETED PLANT MODIFICATIONS Tennessee Valley Authority 'I

SUMMARY

OF Brogans Feny Nuclear Plant SAFETYEVALUATIONSFOR l995 Annual Operating Report ~

FIELD COMPLETED PLANTMODIFICATIONS ~

Safety Evaluations or Unreviewed Safety Question Determinations (USQDs) for the following plant modifications, which were field completed during 1995 were summarized in previous Annual Operating

. Reports. Therefore, they are not included in this report.

See Annual ECN/DCN Operating Report No. Descri tion for Year G18384 -

Unit 3 Modification of Valve Packin Confi ration 1992 L01791 Unit 3 - Modification to Provide Adequate Steam to Steam 1989 Jet Air Ejectors to Utilize Low Pressure Steam Supplied b the Auxili Boiler L01845 Unit 3 - Low Pressure Coolant Injection Mode of the RHR 1988 S stem.

L02107 Unit 3 - RHR System Valve Modification (Disc 1988 Re lacement in Pressure Seal An le Valves .

P 00085 Unit 3 - Upgrade of Drywell Temperature and Pressure 1989 Instrumentation and Annunciators Used for Manual Initiation of Containment Spray. A safety evaluation for DCN F30446 was approved on 09/15/94 which downgraded components (safety related to non-safety related) that perform no safety function and that are isolated from other safety equipment. This safety evaluation also addressed UFSAR chan es re uired.

P00126 Unit 3 - Replacement of Pressure, Level, and Temperature 1984 and 1988 Switches with Analog Transmitter/Trip Unit System for Engineered Safeguard Sensor Trip Inputs. A revised safety evaluation was approved 01/28/94 to remove W21920 as a prerequisite. DCN W21920 was canceled and cables originally scheduled to be replaced remain and are acceptable per calculation ED-Q3999-920183. Fuses supplying power &om panels 3-9-15/17 to RPS panels 3-9-83/84/85/86 were replaced with fuses of a diFerent manufacturer and size.

P00422 Unit 3 - Replacement of Voltage Sensing Relays In RPS 1984 and 1988 Circuit Protectors 3A1/3A2, 3B1/3B2, and 3C1/3C2.

P00612 3 - Installation of Stainless Steel Flex Hose from the

'nit 1984 Control Air Permanent Line to the Pilot Stage Assembly of the MSRVs.

P00614 Unit 3 - Installation of Lar er 2" Stem on MSIVs. 1984

I.

Tennessee Valley Authori ty, SQWlfARYOF Browns Feny 1Vuelear Plant SAFETY EVALUATIONSFOR 1995 Annual Operating Report FIELD COMPLETED PLANTMODIFICA .gpss)

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'See Annual ECN/DCN Operating Report No. Descri tion for Year P00621 -

Unit 3 MSIV Modification - New Locking Tabs, Limit 1984 Switch Mountin Plate, and Gate Valves.

P05246 Unit 3 - Replacement of HPCI Booster Pump Suction 1989 Relief Valve.

W11179 Unit 3 - Replacement of Existing Reactor and Refueling 1992 Zone Ventilation Exhaust Radiation Monitors W15365 Unit 3 - Unit 3 Integrated Computer System Upgrade. 1992 This safety evaluation was revised 11/18/93 to delete Attachments 1 and 2 to the safety assessment. The safety evaluation referenced these attachments only. The attachments initially provided design output information that currently exists in the body of the DCN. Maintaining this design information in the safety assessment/safety evaluation rovided no value and thus was deleted.

W1 5756 Unit 0 - Upgrade Evacuation Alarm System, Code Call, 1992 and Paging System. This safety evaluation was revised 04/28/93 to provide a stage for work associated with the new West Portal Access Building. This stage allowed for the closure of work related to the Unit 2 Cycle 6 outage.

The other sta e included all other work.

W 16707 Unit 0 - Security Upgrade. This safety evaluation was 1992 revised 08/26/93 and 02/23/95 for the design and installation of a new Secondary Alarm Station Building as art of the Ph sical Securit U rade.

W16708 Unit 0 - Security Upgrade (Fencing, Gates, and 1992 Sallyports). This safety evaluation was revised 12/23/93 to allow a change in fence post spacing due to underground interference s.

W16726 Unit 3 - Panel 3-9-4 Control Room Design Review 1992 Modifications. This safety evaluation was revised 08//05/93 to remove any work performed on 3-21-68-33A and 35A. DCN W17545 removed the valves and controls.

W17041 Unit 3 - Panel 3-9-8 Control Room Design Review 1992 Modifications.

Tennessee Valley Authority

SUMMARY

OF Brains Ferry Nuclear Plant SAFETYEVALUATIONSFOR 1995 Annual Operating Report FIELD COMPLETED PLANTMODIFICATIONS See Annual ECN/DCN Operating Report No. Descri tion for Year W1 7044 Unit 3 - Panel 3-9-3 Control Room Design Review 1992 Modifications. This safety evaluation was revised 06/02/94 per F30423 to remove words from the safety assessment/safety evaluation reflecting that equipment will be made inoperable. New wording'reflects that equipment "ma "be made ino erable.

W17051 Unit 0 - Re lacement of Level Switches 1992 W17215 Unit 3 - Panels 3-9-54 and 3-9-55 Control Room Design 1992 Review Modifications.

W1 7273 Unit 3 - Shutdown Board Battery Replacement. This 1992 safety evaluation was revised 05/27/93 per DCN F20476 to allow staged DCN raceway/cable installation at any time for all stages. The safety evaluation was revised again 09/01/94 per DCN F31855 to allow the shutdown board battery replacement to occur at reactor power in accordance with Technical Specification Submittal 347T.

This submittal requested extension of the Unit 2 shutdown board battery limiting condition for operation (LCO) from 5 days to 45 days, allowing the installation of each battery, charger, distribution panel, and oiher related components at reactor power for Unit 2 (provided Units 1 and 3 are maintained defueled .

W17434 Unit 0 - Radwaste Flow Transmitter Re lacement 1992 W17463 Unit 3 - Modification of Reactor Vessel Level 1992 Instrumentation System Sensing Lines. A general revision to the safety evaluation was approved 05/05/93. No safety concerns were identified and no unreviewed safety uestion exists.

W17538 Unit 3 - Modifications to D ell Platform. 1992 W17619 Unit 3 - Modifications of Piping, Components, and Pipe 1992 Su orts.

W17720 Unit 3 - Replacement of Class 1E Oil Filled Transformers 1992 with Dry Type Transformers. This safety evaluation was revised 02/03/93 to add stages 5 and 7 which electrically connect the new 480V HVAC board and moves the 480V Control Bay Vent Board B to this new HVAC board. The safety evaluation was revised again 07/14/94 for removal of all RMS-9 tri devices associated with HVAC Board B.

Tennessee Valley A uthori ty

SUMMARY

OF Brains Ferry iVuolear Plant SAFETY EVALUATIONSFOR (995Annual Operating Report FIELD COMPLETED PLANTMODIFICATIONS See Annual ECN/DCN Operating Report No. Descri tion for Year W1 8062 -

Unit 3 Replacement of Forebay Level Instrumentation. 1992 This safety evaluation was revised 08/05/93 per DCN F25807 to delete stage 4. Stage 4 was to remove the existing submerged transducers at the forebay of the intake pumping station. DCN F25807 deleted stage 4 and allowed the existin transducers to be abandoned in lace.

W1 8213 Unit 0 - Decommissioning, Deenergizing, 1992 Abandonment/Removal of Existing Fire Alarm and Detection S stem.

W1 8620 Unit 3 - Electrical Cable Reroute 1992 W18685 Units 1, 2, 3 - Reroute/Re lace Cables 1992 W33408 Unit 3 - Replacement of Sample Tubing to the OFgas 1992 Hydrogen Analyzer Panels 25-254A and 25-254B. The safety evaluation performed for DCN W18209 is a licable to Unit 3 as well as Unit 2.

Tennessee Valley Authority

SUMMARY

OF Browns Feny Nuclear Plant SAFETY EVALUA TIONS FOR 1995 Annual Operating Report . FIELD COMPLETED PLAiVT MODIFICATIONS a4~~" '&xc4cv~avm4

~

w'~" wM" "" -' " "'-'" p~:": ' '+VS' DCN D29636- Removal of Pressure Switches from Applicable Drawings Unit: 2 DCN D29636 addressed the removal of pressure switches 2-PS-IA, 2-PS-1-5, 2-PS-1-18, 2-PS-1-19, 2-PS-1-22, 2-PS-1-23, 2-PS-1-30, 2-PS-1-31, 2-PS-141, 2-PS-142, 2-PS-1-179, and 2-PS-1-180 associated with main steam relief valve three-stage valves and their related cable and conduit from applicable drawings.

This DCN did not cause physical changes to the plant, it only documented as-installed plant configuration.

This modification is safe from a nuclear safety standpoint. No unreviewed safety question exists.

DCN D29637- Revision of Flow Diagram and Physical Piping Drawings Unit: 3 DCN D29637 revised the flow diagram and physical piping drawings. Drawings showed only one gate valve when two gate valves existed on the 6" decant line to the Unit 3 hotwell. Also, a drawing was revised to remove valve 0-77-961 and correct the continuation flag on the hotwell decant line. This valve was already shown on the continued drawing.

This DCN was a documentation change only which corrected drawings to match the physical plant configuration as well as to correct continuations and references to other drawings (i.e., no physical work required). These drawings are, however.

radwaste system drawings.

The operation of the radwaste system is controlled by Technical Specification 3.8/4.8. Documenting this valve on the design drawings has no affect on the operation, maintenance, or testing of the radwaste or condensate systems as this line is normally isolated.

This modification is safe from a nuclear safety standpoint. No unreviewed safety question exists.

DCN D32127- Revision of Flow Diagram Unit: 0 DCN D32127 revises the flow diagram 047E830-3 to show valve 0-ISV-77-1116 as normally closed, valves 0-VLV-77-884 and -902 as normally open, and laundry tank B connected to the floor drain system via 0-DRV-77-882B. This DCN is a documentation change only which will correct the aforementioned drawing to match the physical plant configuration (i.e., no physical work required). This drawing is, however, a UFSAR figure and a Radwaste system drawing.

The operation of the Radwaste system is controlled by Technical Specifications 3.8/4.8. Documenting the "normal" manual valve positions on the design drawings has no affect on the operation, maintenance, or testing of the Radwaste or Condensate systems as this line is normally isolated. Also, the drain line connection on laundry tank B identical to the laundry tank A design.

This modification is safe from a nuclear safety standpoint. No unreviewed safety question exists.

Tennessee Valley Authority SCGNKIRY OF Browns Feny 1Vuclear Plant SAFETY EVALUATIONS FOR Annual Operating<Report FIELD COMPLETED PLANT MODIFICATIONS

"" 'aY"'99$

DCN S36661 - Pressure Rating Change on Standby Liquid Control (SLC)

Unit: 2 This DCN affected documentation only on portions of the SLC System. Problem Evaluation Report BFPER940940 describes a condition in the SLC System in which the pressure rating for the SLC pump and safety relief valve do not agree with the design pressure rating of the SLC pump discharge piping. Flow diagram 247E854-I indicated the positive displacement-type SLC pumps have a capacity of 50 gpm at 1500 psig. The discharge relief valves (RFV-512 and 513) are set at 1425 psig + 75 psi. However, the SLC pump discharge piping is rated at a design pressure of 1400 psig up to valve 63-524 and 1150 psig dowr%eam of valve 63-524. DCN S36661 changed the pressure rating on the SLC pump discharge piping and revised calculations which are affected by changing the SLC pump discharge piping pressure rating.

This modification is safe from a nuclear safety standpoint. No unreviewed safety question exists.

DCN S37684- Change of Seismic QualiTication Unit: 2,3 This safety evaluation supports the proposed changes to the UFSAR by change request BFEP-MN-95057RO. The proposed changes consists of revising Figures 5.2-22 Sheets 1, 2, and 3 to remove BFN-2-FSVA3-28A, BFN-2-FSV43-28B, BFN-2-FSV-29A, BFN-2-FSV-29B, BFN-3-FSV43-28A, BFN-3-FSV43-28B, BFN-3-FSV-29A, and BFN-3-FSV-29B and to add Residual Heat Removal (RHR) valves BFN-2-SMV-74-226, BFN-2-SMV-74-227, BFN-3-SMV-74-226, and BFN-3-SMV-74-227 as primary containment isolation valves.

DCN S37684 implemented the corrective action proposed in BFPER950761RI. This Problem Evaluation Report identified the following condition adverse to quality for Units 1, 2, and 3: "FSV-43-28A, FSV<3-38B, FSV<3-29A, and FSV43-29B, Units 1, 2, and 3 of the Water Sampling System at elevation 519'n the RHR corner rooms are listed in the Design Criteria Document (BFN-50-7307), UFSAR and other engineering/plant documentation as primary containment isolation valves. As a consequence, these valves would require seismic Class I qualification. There is no documentation demonstrating that the tubing/piping/supports associated with these valves meet seismic Class I support criteria. In addition, the seismic Class I boundary calculations for Units 2 and 3 do not include these valves within the seismic Class I boundaries."

DCN S37684 revised engineering and plant documentation to be consistent with the RHR valves74-226 and 74-227 (Uriits 2 and 3) being designated as primary containment isolation valves. Leak rate testing of valves 43-28A/B and 43-29A/8 (Units 2 and 3) willno longer be required. The designated RHR valves74-226 and 74-227 (Units 2 and 3) will not require any local leak rate testing. However, these valves willbe within the integrated leak rate testing boundaries.

This change will not require any changes as to how the water sampling system is operated. Designating the existing RHR valves74-226 and 74-227 which are normally closed and manually operated willnot change the procedure by which the system is operated.

Tennessee Valley Authority

SUMMARY

OF Brains Feny Nuclear Plant SAFETY EVALUA TIONS FOR

/995 Annual Operating Report -

FIELD COMPLETED PLANTMODIFICATIONS DCN S38892- Addition of Note on Flow Diagram Identifying Operator Actions Unit: 3 Problem Evaluation Report BFPER951542 identified a single failure which could result in the loss of air conditioning and subsequent loss of two Unit 3 4kV shutdown boards. A single failure of 480V reactor motor operated valve board 3A or 3B could result in loss of cooling to both rooms 3EA and 3EB or rooms 3EC and 3ED, respectively. A gradual temperature rise in the rooms could result in exceeding the shutdown boards component qualification temperature of 104'F and subsequent spurious failures of the electrical boards. However, as justified by calculation ND~031-950029 Revision 0, a single train of air~nditioning equipment is capable of providing adequate cooling provided inanual actions are taken within 37 hours4.282407e-4 days <br />0.0103 hours <br />6.117725e-5 weeks <br />1.40785e-5 months <br />. DCN S38892 added a note to Qow diagram 347ES65-8 to provide design output which delineates the limiting condition identified by calculation ND~031-950029. The note shall identify the operator actions (open fire doors 810, 811, and 824 and turn offthe normal lighting) which ensure the 4kV shutdown board room and bus tie room remain below 104'F.

This modification is safe from a nuclear safety standpoint. No unreviewed safety question exists.

DCN T20662 - Replacement of Thermo-Lag Fire Barrier Material Unit: 2 This DCN replaced the Thermo-Lag 330-1 fire barrier material located on the Division I Residual Heat Removal Service Water/Emergency Equipment Cooling Water circuits at the intake pumping station with new Thermo-Lag 330-1 material. This new installation willresult in an approved one-hour fire barrier installation.

This modification is safe from a nuclear safety standpoint. No unreviewed safety question exists.

DCN T27900- Modiifiications to Improve Functional Reliability of Wide Range Gaseous Emuent Radiation Monitoring System (WRGERMS)

Unit: 0 Temporary Alteration Control Form 0-93-3-90 installed several modifications to improve functional reliability of the WRGERMS. DCN T27900 was issued to make these changes permanent plant features and update the design basis.

The WRGERMS vendor manual and associated vendor and TVA drawings were revised to show these changes. A UFSAR revision was made to clarify descriptive language for the WRGERMS.

The functional requirements for the normal range particulate and iodine channels are deleted from the WRGERMS design. Continuous monitoring for noble gas and grab samples for particulate and iodine are specifie. This change does not alter the function or operation of the noble gas monitoring equipment for which Technical Specification 3.2.F/4.2.F applies.

This modification is safe from a nuclear safety standpoint. No unreviewed safety question exists.

Tennessee Valley A uthori ty SUM1lfARYOF Brains Feny Nuclear Plant SAFETY EVALUATIONS FOR l995 Annual Operating Report FIELD COMPLETED PLANT MODIFICATIONS DCN T29436 - Raise Alarm Setpoint for 2-TA46-108A Unit: 2 DCN T29436 issued a Design Change Authorization against Instrument Tabulation drawing 047B601M4 to indicate the new alarm setpoint value of 55'F for 2-TA46-108A, a Nuclear Engineering Setpoint and Scaling Document for describing calibration information, and an Equipment Management System update request. The alarm setpoint was raised per 2-TRS46-108 Accuracy Evaluation and Setpoint Determination.

The change will not afiect the Offgas System's function to control the release of gaseous radwaste to the environment.

This modification is safe from a nuclear safety standpoint. No unreviewed safety question exists.

DCN T30199 - Issues Unit 3 Nuclear Engineering Setpolnt and Scaling Documents for Baseline Setpoint Calculations Unit: 3 DCN T30199 issues Unit 3 Nuclear Engineering Setpoint and Scaling Documents (NESSDs) for baseline setpoint calculations not associated with previously issued DCNs. This DCN also issues NESSDs for calculations updated to include the latest Significant Operating Report (SOR) switch test report accuracies and for a calculation updated to resolve Quality Assurance audit findings. This DCN (as revised by DCN F34985A) also issues NESSD 3P47340294143, which was updated to incorporate new environmental data being issued by DCN S33300A. DCN F34985 also issues NESSDs 3U-268403D/1D4041 and NESSD 3U-2684003E/1D4041 to correct the lower allowable value from 18.1 to 18.3 seconds.

The setpoint identified in NESSD 3P47340294143 for 3-PS-73-29-1 differs from that shown in UFSAR Table 7.4-1.

Additionally, the instrument unique identification is different for Unit 3. The information shown in Table 7.4-1 is shown elsewhere in Section 7.4, in the control diagram and logic diagram, figures and therefore does not need to be repeated. Information regarding the basis for selection of the setpoints is shown in Section 7.4.3.2.2. The actual trip setting or analytical limit in Table 7.4-1 is too much detail for the UFSAR and results in UFSAR changes being required whenever these changes are made. Therefore, on the basis that information is already provided in Section 7.4 for the instrumentation used and the basis for the setpoints, Table 7.4-1 and the references to it are being deleted by change request BFEP-EEB-95031.

Setpoint and scaling changes issued with this DCN also impact information presented in UFSAR Table 7.8-2. This table has been revised by change request BFEP-BNA-94003 issued with Unit 2 DCN T30200 to refiect the current

'instrumentation and to rcmove the trip setpoint values.

DCN T32638- Update Radwaste Drawings Unit: 0 This DCN corrected various discrepancies between Radwaste System flow diagrams, control diagrams, and physical drawings to accurately depict the configuration installed in the plant. These discrepancies were administrative in nature and do not afiect the design basis, operation or function of the Radwaste System. Discrepancies corrected 0

Tennessee Valley Authority

SUMMARY

OF Browns Ferry Nuclear Plant 1995 Annual Operatin Report 'IELD SAFETYEVALUATIONSFOR COMPLETED PLANTMODIFICATIONS are as follows: 1) Showed correct function code of sump level switches as "LS" instead of "LIS" on the control diagram and physical drawing. 2) Corrected and labeled continuation flags on Qow diagrams. 3) Deleted various valves which were duplicated on separate Qow diagrams. Valve positions are in accordance with Radwaste System operating instructions. 4) Corrected flow diagram to show valves in the condensate recycle line as indicated on physical drawings and as installed in the plant. 5) Corrected minor configuration discrepancies on flow and control diagrams to match physical drawings and plant conditions.

Physical changes to the plant were not required except for component labeling.

This modification is safe from a nuclear safety standpoint. No unreviewed safety question exists.

DCN T33279- ModiTied the Recirculation Flow Control System from a Speed Controller to a Scoop Tube Position Controller Unit: 3 DCN T33279 was issued to delete the speed feedback signal from the Motor Generator (MG) set tachometer to the error limiter circuit in the Recirculation Flow Control System. The previous system controlled generator speed using the feedback signal and had exhibited unstable operation at higher pump speeds. This DCN duplicates, on Unit 3, the modifications performed by DCN T30397 on Unit 2.

This DCN modified the Recirculation Flow Control System from a speed controller to a scoop tube position controller. This eliminates self sustaining oscillations initiated by an electrically or mechanically induced perturbation.

This modification required disconnecting one wire from each of the coils of relay 2A-K60A for the 3A MG set and 2A-K60B for the 3B MG set. This prevents the speed signal from being input into the Error Limiter network and allows the controller output to be used as the feedback signal instead. Recirculation pump speed willstill be indicated in the control room at the pump's M/A station.

This change revises the description of the Recirculation Flow Control System circuit function and operation discus in the UFSAR.

Removing the speed regulation feature from the Recirculation Flow Control System circuitry does not alter the relationship of the recirculation pumps with respect to core flow/reactor power operation characteristics. Only the manner in which the operator makes manual pump speed changes is afiected. Recircttiation pumps willstill be operated within the limits imposed by the Power/Flow Curve (Technical Specification Figure 2.1-2).

DCN T35068- Addition of Check Valves and Service AirPiping Unit: 3 DCN T35068 provided the design to install a check valve in each of the 1" service air purge lines which serve Oifgas System preheaters A and B. The design change also revised applicable design drawings to correctly depict the existing Service/Offgas piping for Unit 3. The Service Air System is physically connected to the Offgas System through normally closed solenoid valves and is used to provide purging air to the OQgas System preheaters prior to performing maintenance on the equipment. The service air purge piping is also used in conjunction with auxiliary boiler steam to warm the Oflgas recombiners from a cold condition prior to placing the steam jet air ejectors in service. NRC IE Information Notice No. 7948 documents the potential for contamination of plant

Tennessee Valley Authority SUMK4RYOF Broils Ferry Nuclear Plant SAFETY EVALUATIONSFOR l995 Annual Operating Report FIELD COMPLETED PLANTMODIFICATIONS P.'--'ervice air systems (which are used as a source for breathing air) that are physically connected to gaseous radwaste systems. The installation of a check valve in each of the service air purge lines ensures adequate separation between the. Service Air System and the OQgas System and reduces the possibility of a Service Air System contamination and Offgas release to the environment. A similar design change has been implemented for the corresponding Unit 2 Service Air/Offgas piping by DCN W3548B.

Additionally, the design change provided the necessary design drawing changes to correctly document the valve manufacturer information for Offgas System valves 3-FCV466-0011 and 3-FCV4664015. The aflected valves are located in the suction piping of steam jet air ejectors A and B.

This modification did not alter the system design or functional requirements of the affected systems nor did the modification affect any safety related equipment.

DCN T35069 - Setpoint Change Unit: 3 DCN T35069 modified Unit 3 OQgas System equipment and setpoints to match the corresponding Unit 2 equipment and setpoints. Specifically, this DCN performed the following: 1) The setpoints for 3-PIS46-21C and -21D were revised to 3.5 psig increasing. These instruments provide automatic valve closure for Offgas System drains on high Offgas System pressure. 2) The setpoints for 3-TA46-108A and -108B were revised from 48'nd 42'F to 55'nd 39'F, respectively. TA46-108A provides an'alarm for high moisture separator reheater inlet temperature.

TA<6-108B provides an alarm for low charcoal bed moisture separator outlet temperature. During low Offgas Qow conditions, the temperature recorded'tends to reflect ambient temperature, which is higher than Offgas eflluent temperatures. Nuisance alarms from these inputs have been recorded on Unit 2 which were subsequently eliminated by revising the setpoints..3) The setpoints for 3-TS46-55 and -56 were revised Rom 110'o 120'F. The setpoint for 3-PS'-37 and 41 were revised from 27" Hg to 26" Hg. High room temperatures in Unit 2 resulted in nuisance alarms associated with mechanical vacuum pump high gland seal water temperature. The nuisance alarms were eliminated by raising the alarm setpoint. Pressure switch setpoints were also revised which trip the mechanical vacuum pump on increasing vacuum. The water provides both a seal and compressing Quid for the vacuum pumping action. The vacuum setpoint was modified to provide adequate margin between the operating pressure and the saturation pressure associated with the increased temperature setpoint. 4) Flow switch 3-FS4649 was removed to match its removal from Unit 2. This flow switch was originally installed to shut offthe Offgas dehumidifier chiller unit upon loss of a flow to prevent damage to the chiller. The chiller is interlocked with the chilled water circulating pump and is supplied with a temperature controller, internal breeze protection and other protective devices which make this flow switch redundant and unneamary.

This DCN modified alarm setpoints without aflecting the existing system operating ranges. The equipment affected by the DCN provide neither nuclear nor protective safety functions.

DCN T35110 - Setpoint Change Unit: 2 This DCN was issued to increase the setpoint on the Unit 2 ~veil equipment drain sump high temperature switch 2-TIS-77-14, which was previously set at 130'F. This setpoint was causing the drywell sump pump to cycle too frequently in the recirculation cooling mode. The switch also produces a control room annunciator alarm at this setpoint, consequently creating a nuisance alarm.

Cl Tennessee Valley Authority

SUMMARY

OF Browns Ferry Nuclear Plant SAFETY EVALUATIONSFOR l995 Annual Operating Report FIELD COMPLETED PLANTMODIFICATIONS 4

The drywell equipment drain sump is a subsystem of the Liquid Radwaste System. The diywell equipment drain sump collects leakage from identified components inside the drywell and, when full, is pumped from the drywell to the waste collector tank located in the Radwaste Building. Temperature switch 2-TIS-77-14 senses sump temperature and controls the sump pumps and discharge valves to recirculate the water through a cooler ifthe water is too hot before discharging to the waste collector tank. The sump water is circulated through the cooler until it cools down to the low temperature setpoint at which time the pumps willstop, or ifthe sump water level is at the high level setpoint, discharge to the waste collector tank. Ifthe sump water level reaches the high-high setpoint, the pumps discharge to the waste collector tank whether or not the water temperature is above the high temperature setpoint. This feature prevents discharging extremely hot water to the Radwaste System except under unusual conditions of high inleakage into the sump that are beyond the capacity of one pump.

This change increased the high temperature setpoint on 2-TIS-77-14 from 130'F to 150'F (INC) and the low temperature reset point from 115'F to 125'F (DEC). This reduces the cycling frequency on the diywell.equipment drain sump pumps in the recirculation cooling mode and reduces nuisance alarms in the control room. This setpoint change did not eliminate the alarm, it will still occur when the sump water reaches 150'F, but this should occur much less oflen than at the 130'F setpoint. The low temperature reset point on this switch was raised to maintain approximately the same recirculation mode cooldown duration.

This setpoint change does not aQ'ect the safety-related function of the Radwaste System to provide primary and secondary containment boundary integrity.

DCN T35528- Removal of Abandoned Fire Protection Components Unit: 2 DCN T35528 was issued to remove High Pressure Fire Protection (HPFP) System water supply tie-ins to abandoned sprinkler systems located in the Unit 3 Reactor Building. Corrosion caused leakage in these deaden'ded branch lines contributed to increased maintenance eQorts on the unused piping. This change disconnected abandoned sprinkler systems from the HPFP main water distribution header.

Fire Protection Report Change Notice FPR-95007 was submitted with the DCN package to update Figures 10.11< Sheet 2, 10.11-7 Sheet 2, and 10.11-8 Sheet 2 in the Fire Protection Report. These piping changes constituted a change to the plant facility as described in the Fire Protection Report, therefore, a 10CFR50.59 safety evaluation was required.

This modification is safe from a nuclear safety standpoint. No unreviewed safety question exists.

DCN T35568 - Change Offgas System Bypass Dampers 0-DMP46-953A and 0-DMP-953B to Open Position Unit: 0 This DCN changed the Offgas System bypass dampers 0-DMP46-953A and 0-DMP46-953B to the open position.: The dampers were in the closed position and water from condensation was accumulating on the top of the backdraQ dampers 0-DMP46-976A and 0-DMP46-976B. Throttling any dampers (O-DMP46-929A, O-DMP&-953B, O-DMP46953A, 0-DMP46-929B) to force flow through both the normal dampers and the bypass dampers will remove condensation.

Tennessee Valley Authority SUMllfARYOF Browns Ferry Nuclear Plant SAFETYEVALUA TIONS FOR 1995 Annual Operating Report k4j 4v.:V4ACAM"'~4NV"" ' ' '"' """ "'""""""""'

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FIELD COMPLETED PLANT MODIFICATIONS

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Dampers 0-DMP46-953A and 0-DMS-953B are isolation dampers located in the parallel stack discharge duct section which is used to allow maintenance on associated backdrafi dampers without disrupting OQ'gas flow. Opening these isolation dampers does not reduce the OQgas fiow and willprevent moisture accumulation on top of the backdraQ dampers.

The basis for radiological effluent monitoring in the Technical Specifications is established by federal regulatory requirements which specify environmental radiological release limits to protect the health and safety of the public. The Offgas System functions to control and limit the amount of radioactivity released &om the plant's main stack in order to help prevent exceeding these established limits. This damper position change has no relationship to establishing or maintaining any margin of safety as defined by the regulations (i.e., the basis) which govern control of radiological releases to the environment. Therefore, this activity does not affect any margin of safety defined in the Technical Specifications.

DCN T35738 - Internal Panel Electrical Separation Modification Unit: 3 This DCN modified circuitry associated with the Reactor Protection System and Residual Heat Removal System and the Primary Containment Isolation System function of the following systems: Reactor Core Isolation Cooling, High Pressure: Coolant Injection, Radwaste. Fuses were provided and replaced; cables were reterminated or relocated; separation barriers and flex conduit were provided for internal panel wiring, and relays were provided to ensure compliance to electrical and physical separation requirements.'his DCN added and replaced fuses and added relays without affecting the existing system operating characteristics.

This modiTication is safe from a nuclear safety standpoint. No unreviewed safety question exists.

.DCN T35933 - Increase Setpoints for Unit 3 Drywell Equipment Drain Sump Temperature Switch Unit: 3 DCN T35933 increases the setpoints for the Unit 3 diywell equipment drain sump high temperature switch (3-TIS-77-14). The existing setpoint causes the drywell sump pump to cycle frequently in the recirculation cooling mode and produces an alarm in the main control room on panel 3-94 (3-TA-77-14). The change in setpoint reduces the cycling frequency of the drywell equipment drain system recirculation cooling mode and consequently, reduces the, frequency of receiving alarm 3-TA-77-14 in the main control room.

This modification is safe from a nuclear safety standpoint. No unreviewed safety question exists.

DCN T36027 - Replacement of Obsolete Recombiner Temperature Controllers, 3-TC-66-76 and -90 Unit: 3 This DCN replaced obsolete Offgas recombiner temperature controllers 3-TC46-76 and 3-TC46-90 with Yokogawa Model UT37 single input controllers. The 0.1 ampere power fuses protecting the controllers were replaced with 1 ampere fuses per Yokogawa recommendations. The new controllers are configured to control temperature in the same manner as the existing controllers, by cycling a power relay on and off to control heating elements. The new equipment provides "proportional" control similar to the previous controllers by varying the duty cycle of the heaters as the Tennessee Valley Authority SMEARY OF Brains Ferry Nuclear Plant SAFETY EVALUATIONS FOR 1995 Annual Operating Report FIELD COMPLETED PLANTMODIFICATIONS CKXwC ' 'C4WZ 'CRT i' ' 'ww' 'v " '~

' ' ' 'kly A&4, temperature setpoint is approached. The ope'rational temperature limits for the Offgas recombiner are not affected by this DCN. This modification has no impact on the operation of the Offgas System.

This modification is safe from a nuclear safety standpoint. No unreviewed safety question exists.

DCN T36111 - Change Low Temperature Alarm Setpolnt on Offgas Charcoal Vault Unit: 2 This DCN changed the Unit 2 Offgas System charcoal bed vault temperature low alarin setpoint, 2-TA46-120B, Rom 73'F to 70'F. The vault experiences temperature excursions as low as 71'F during the winter months resulting in nuisance alarms.

The basis for radiological eflluent monitoring in the Technical Specifications is established by federal regulatory requirements which specify environmental radiological release limits to protect the health and safety of the public. The OQgas System functions to control and limit the amount of radioactivity released &om the plant's main stack in order to help prevent exceeding these established limits. This temperature alarm setpoint change has no relationship to establishing or maintaining any margin of safety as defined by the regulations (i.e., the basis) which govern control of radiological releases to the environment. Therefore, this activity does not aQ'ect any margin of safety defined in the Technical Specifications.

This modification is safe from a nuclear safety standpoint. No unreviewed safety question exists.

DCN T37074- Replacement of Offgas System Components Unit: 3 DCN T37074 provided for the following upgrades to the Unit 3 Offgas System: 1) Replaced obsolete controller 3-TIC46-102 with an Allen-Bradley progratmnable logic controller and digital indicator. 2) Replaced'obsolete GEMAC transmitters 3-LT46-93 and 3-LT46-94 with comparable Rosemount transmitters. 3) Replaced pressure switches 3-PIS-66-21A, 3-PIS46-21B, 3-PIS46-31C, and 3-PIS46-21D. 4) Provided setpoints for 3-TAN-120A and 3-TA-120B (formerly 3-TA46-115A and 3-TA46-115B, respectively) which are consistent with those in Unit 2.

This DCN replaced a temperature controller and level transmitters without affecting the existing system operating ranges. This DCN replaced pressure switches with expanded indicated ranges while the setpoints remain unchanged.

The equipment affected by this DCN provide neither nuclear nor protective safety functions.

This modification is safe from a nuclear safety standpoint. No unreviewed safety question exists.

DCN T37461- Dewpoint Device Replacement Unit: 3 This DCN completed the work identified on DCN T37074, to replace obsolete equipment which included the dewpoint instrument 3-ME<6-110. The dewpoint device replacement was removed from the scope of DCN T37074 to improve work control. The inline dewpoint hygrometer 3-ME46-110 was replaced with an ofiline system that may be isolated from the OQ'gas System for maintenance without isolating the Offgas flow stream which it monitors.

Tennessee Valley Authority SCSIMARY OF Browns Feny Nuclear Plant SAFETY EVALUATIONSFOR l995 Annual Operating Report FIELD COMPLETED PLANTMODIFICATIONS This DCN installed a new panel, 3-25481, in the OQgas Building, to house the ofiline sensor 3-ME%6-110, vacuum pump 3-SPMP&-110, and inline flow indicating controller 3-HC46-110. The new dewpoint analyzer (3-MT4i-110) was installed in the same location as the original equipment, panel 3-25-97. The output of this device willbe a 4-20 mA signal equal to the dewpoint temperature, a range of 0-100'F and provide input to the recorder 3-TRS46-108. This recorder was replaced as it only accepts RTD inputs. The replacement recorder willaccept either RTD or voltage signal inputs. The Offgas high/low temperature alarms Rom this recorder has been disabled.

During long periods of low OQgas flow rates these alarms do not provide useful system operating status and are a source of nuisance alarms in the main control room. Therefore, they were removed. Additionally, the transfer mvitch 3-XS46-108 was removed and the two cooler condenser resistance temperature device inputs, 3-TE<6-108A and B were connected directly to the recorder.

To allow Operations the flexibilityof running both Cooler Condensers 3A and 3B at the same time, a note was added to the mechanical flow drawing to allow dual cooler condenser operation.

DCN T37517- Changed/Removed/Added Breaker and Load Limit Restrictions Unit: 3 This DCN changed/removed/added breaker and load limit restrictions to allow operation of multiple units, Units 2 and 3. This DCN also revised electrical load limits for loss of coolant accident loads, normal loads, and addressed load changes for Appendix R high impedance fault concerns. Revision B issued a load matrix, via Engineering sketches, which willbe used as a guide in determining load restrictions applicability for various scenarios of outwf-service transformers.

This DCN was primarily for documentation purposes only. It documents the results of calculations which define the loading requirements and restrictions for multiple unit operation. No new equipment was introduced by this DCN and no equipment required for safe operation of Units 2 and/or 3 was modified by this DCN.

This DCN also made changes to breaker alignments for Appendix R purposes to resolve high impedance fault concerns.

As a result of these changes, annunciator circuits were modified by DCN F38634. These circuits represent the only physical work completed by this DCN and do not impact any safety related components.

This modification is safe from a nuclear safety standpoint. No unreviewed safety question exists.

DCN T37949- Offgas System Glycol Storage Tank Temperature Low Alarm Setpoint Unit: 2 The DCN changed the Unit 2 Offgas System glycol storage tank temperature low alarm setpoint, 2-TAM-IOOB, Rom 33'F. This change was implemented to reduce nuisance alarms caused by the glycol tank temperature controller alarm setpoint and the tank bulk temperature alarm setpoint having the same value.

This modification is safe from a nuclear safety standpoint. No unreviewed safety question exists.

Tennessee Valley Authority

SUMMARY

OF Brogans Ferry Nuclear Plant SAFETYEVALUATIONSFOR 1995 Annual Operating Report FIELD COMPLETED I'LANTMODIFICATIONS

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.<g<P; '><'j~j'<>a applicable electrical elementary diagrams, electrical connection diagrams, cable and conduit schedules, mechanical flow diagrams, and the Equipment Management System database.

The temperature alarm loop does not provide any control functions. The replacement of the temperature indicating switch and its associated element with components that have a greater range permits a higher setpoint value for alarm. The higher alarm setpoint willremove a nuisance alarm to allow the operators to perform their function more effectively.

This modification is safe from a nuclear safety standpoint. No unreviewed safety question exists.

DCN WI5367- Integrated Computer System PCS) Upgrade Unit: 3 This DCN provided equipment mounting details, cable routing, and cable terminations to complete the BFN Unit 3 ICS upgrade modification. Also included in this DCN was the removal of the Unit 3 GE4020 Plant Process Computer. This modification was required to support TVA's commitment to the NRC to implement NUREG 0696 reqiiirements.

Revisions to the UFSAR were required as a result of this modification.

This modification is safe from a nuclear safety standpoint. No unreviewed safety question exists.

DCN W16710- New Uninterruptable Power Supply (UPS) Building Unit: 0 This DCN provided the modification for a new UPS building along with the equipment and utilities to upgrade the BFN Security System. This Security System upgrade involved the erection of a Class II UPS building and the installation of associated non-safety related equipment outside of Class I buildings. The power sources and other utilities required to feed the newly installed equipment is not part of the power sources used in the event of an ab'normal operational transient. The equipment installed in the new UPS building has no impact on any safety related equipment.

UFSAR figures required revision as a result of this modification.

This modification is safe from a nuclear safety standpoint. No unreviewed safety question exists.

I DCN W16711- New West Access Control Portal Unit: 0 This DCN provided the modification for a new west access control portal building along with the equipment and utilities to upgrade the BFN Security System. The equipment installed in the new west access portal has not impact on any safety related equipment.

The UFSAR required revision in order to depict the location of the new buildings and revise the description of the portal radiation monitors.

This modification is safe from a nuclear safety standpoint. No unreviewed safety question exists.

Tennessee Valley Authority Brogans SUMKQtYOF Feny Nuclear Plant SAFETYEVALUATIONSFOR 1995 Annual O~erati~n Report FIELD COMPLETED PLANTMODIFICATIONS DCN W17185 - Installation of Containment Isolation Status System (CISS)

Unit: 3 This DCN installed and interconnected the CISS for BFN Unit 3. The CISS uses prograrninable logic controllers to monitor the position of Primary Containment Isolation System (PCIS) valves and the status of PCIS isolation initiations. This information is processed to provide a mnmary of PCIS isolation completions on Control Room panel 3-9A.

In addition, this modification relocated two PCIS logic reset handswitches (16A-S32 and 16A-S33) and four PCIS Group 1 isolation logic status indicating lights (16A-DS250, 16A-DS251, 16A-DS252, and 16A-DS253) Rom panel 3-9-5 to panel 3-9A. To make room for the CISS status indications, the drywell floor drain sump flow totalizer (3-FQ-774) was relocated on panel 3-9-4.

This DCN also modified the PCIS Group 8 (Traversing Incore Probe (TIP)) isolation circuitry and installed a relay and Group 8 reset push4utton on panel 3-9-13. This modified the circuitry to remove the existing auto-reset circuitry on the TIP valves and requires an operator to manually reset a Group 8 isolation before the TIP ball valves can be opened. This modification has been performed on Unit 2 via ECN P0469 and satisfies NRC commitments.

Changes to UFSAR figures were required as a result of this DCN.

Installation and operation of CISS does not affect the functional and performance characteristics of the monitored PCIS equipment.

DCN W17344 - Drywell Blower ModiTications Unit: 3 This DCN relabled, rerouted, and/or replaced cables to drywell blowers 3A-I, 3B-l, 3A-2, 3B-2, 3A-3, 3AM, 3BP, 3A-5, and 3B-5, as required, to correct electrical separation concerns. Also, this DCN switches power sources for drywell blowers 3A-2 and 3B-1, 3AM and 3B-3 in order to correct separation concerns and to match the configuration of the Unit 2 diywell blowers. In addition, documentation was provided in this DCN to retag the power and control cables, conduits and cable splices associated with the Unit 3 drywell blowers to remove the incorrect and reflect new unique identifier numbers. '1E'esignation DCN F20401 corrected inconsistencies between the calculated minimum and maximum cable lengths of various cables,

'orrected the construction notes reflecting those lengths, clarified the Appendix R (Qre protection) requirements, and corrected the setroute device to device cable lengths as identified in plant modifications audit BFA92302.

DCN F39080 replaced the thermal overload heaters for drywell blowers 3A-3 and 3B<. The existing thermal overload heaters were sized'smaller than the full load current of the blowers and have resulted in spurious tripping of the units. The replacement thermal overload heaters are adequately sized.

UFSAR figures required a change due to the modification.

The cables rerouted/replaced by this modification have no safety related function and their replacement willhave no aQect on the function, operation, or qualification of any component or system required to ensure nuclear safety.

Tennessee Valley Authority SQKlfARYOF Browns Feny Nuclear Plant SAFETYEVALUA TIONS FOR 1995 Annual Operating Report FIELD COMPLETED PLANT MODIFICATIONS K::""'-~"'""""""':'::::: .""":

DCN W17424- Repair/Replacement of Drywell Environmentally Qualified Cables Unit: 3 The Unit Restart Equipment List (REL) indentified components which must be replaced/repaired to support Unit 3 operation. This DCN addressed REL electrical breakages (cable, conduit seals, limit switches, pressure switches, flow transmitter loops, SIS wiring, thermal overload heaters, splices, etc.) associated with various systems in the Unit 3 diywell.

This change involved the modification of safety-related systems. However, no change was made to the function, operation, or performance of these systems.

The replacement of cables with Class 1E, 10CFR50.49 qualified cables provides enhanced equipment operability, thus assuring that the capability of safety-related equipment is not diminished as a result of inadequate protection of.

less reliable, obsolete cabling.

This modification is safe from a nuclear safety standpoint. No unreviewed safety question exists.

DCN W17627 - Reconnection of Unit 3 Containment Atmosphere Dilution (CAD) Lines and Seismic Qualification to the Unit 3 Reactor Building Wall Unit: 3 This DCN was implemented to reconnect the CAD nitrogen supply lines to Unit 3. The piping was disconnected in the Residual Heat Removal Service Water tunnel because it lacked adequate seismic qualification to support Unit 2 operation. The reconnection restored the system to original configuration and modified existing supports.

However, unlike the original design, the supply lines were reconnected without fiexible ball joints in each line.

Additionally, this DCN provided the neamary analyses and modifications to qualify the aQected portion of the system to seismic Class I requirements and extends the seismic Class I qualification of the two Unit 3 CAD System supply lines to the first anchor inside the Unit 3 Reactor Building. The modification included performing rigorous stress analyses and pipe support calculations and pipe support modifications as a result of these analyses. The stress problems associated with this modification are Nl-384-25T for CAD Line A and Nl-384-26T for Cad Line B.

There were 14 existing supports on each line and every support was modified by this DCN. Additionally, an anchor was added to each of the CAD lines inside the Reactor Building. The supply lines were supported Rom the 16" RHRSW lines. The design change modified the fiame portion supporting the CAD line on 26 of the supports. Two supports were completely redesigned and two supports were designed and installed as anchors. Additionally, the bolts of the support fianges were replaced with high strength bolts. Also, the bolts of the fianges in the piping system analyzed by the stress problems Nl-384-25T and -26T were replaced with high strength bolts.

This design change docs not change the function or operation of the CAD System. This modification was implemented to support the return of the Unit 3 CAD System to service and restore the Unit 3 CAD System to its original design requirements. This design change only seismically qualified the CAD System Rom the reconnection through the first anchor inside the Reactor Building. The design basis in the UFSAR states that the CAD System is capable of supplying nitrogen to Unit 2. Following the completion of this modification, the CAD System will only be capable of supplying nitrogen to Units 2 and 3. The UFSAR was updated to reflect actual plant configuration.

Tennessee Valley Authority

SUMMARY

OF Brains Ferry Nuclear Plant SAFETY EVALUATIONS FOR l995 Annual Operation Report . FIELD COMPLETED PLANT MODIFICATIONS vA DCN W17675 - Check Valve Installation Unit: 3 DCN W17675 installed seismic Class I check valves 3<KV477-1759 and 3ZKV477-1760 in the 4" radwaste line fiom the floor drain sump pump discharge, upstream of isolation valve 3-ISIA77412. By this installation, the 3" drywell floor drain sump piping Rom 3-FCV477-2B to 3-CKV477-1759 and the 4" discharge piping of the Reactor Building floor drain piping from sumps 3A and 3B to 3-CKV477-1759 willonly need to be qualified to seismic Class II criteria as determined by the analysis provided in DCN Q17145. This DCN also qualifies the piping fiom penetration X-19 through flow control valve 3-FCV477-15B to seismic Class I.

This DCN also installs a 3/4" socket welded union in the 3/4" bypass line on the 3A Reactor Building sump discharge. Installation of the union facilitates pump maintenance.

This DCN modified pipe supports to meet design criteria requirements to meet applicable seismic categorizations This modification does not change the system design or functional requirements. UFSAR figures are affected by this modification.

DCN W17702- Replacement of Board Room Air Conditioning Units (ACUs)

Unit: 3 This DCN installed redundant ACUs on elevation 593'nd elevation 604'f the Reactor Building to cool the Unit 3 Reactor Building cicctric board rooms on elevations 593'nd 621.25'. These new ACUs (3A and 3B) are designated 3-ACU431-7205 and -7206, respectively. These ACUs maintain the safety related electric distribution boards in both Unit 3 Reactor Building board rooms below the maximum environmental steady state temperature of 104'F. The new ACUs are sized to maintain temperature in these rooms during normal operating conditions for Units 1 and 2 with a simultaneous loss of coolant accident (LOCA) condition in Unit 3. This is the most severe Heating, Ventilation, and Air Conditioning load requirement for these rooms, bounding the requirements for normal operation for any two of the three units with the third unit in a LOCA condition. These board rooms are important to safety as they feed power to safety related equipment required by all three units for safe shutdown.

This modification removed the Unit 3 shutdown board room emergency cooling units, associated refrigerant piping, ductwork, cables, controls, related supports, and dampers serving the electric board rooms. The existing emergency cooling units for the electric board rooms were standard 5-ton commercial grade systems that are not qualified for operation within the expected Reactor Building post-accident (harsh) environment.

CRLD BFEP-BNA-92152 has previously been written to change Technical Specification Table 3.11.A to remove the duct mounted smoke detectors. No further changes to the Technical Specifications is required.

DCN W17779- Relocation of Power Supply for Control Bay Water Chiller 3B Unit: 3 DCN W17779 relocated the control bay water chiller 3B feeder from 480V shutdown board 3B to the new 480V Heating, Ventilation. and Air-Conditioning (HVAC) board B (installed per W17720). The power supply was relocated to reduce load on 480V shutdown board 3B and to provide a more equitable load distribution between 4kV shutdown boards 3EC and 3ED.

Tennessee Valley Authority

SUMMARY

OF Brains Feny Nuclear Plant SAFETY EVALUATIONS FOR 1995 Annual Operating Report FIELD COMPLETED PLANTMODIFICATIONS P~~'~v~ '.",:~ ~..~~~~a. '"'. ';:."':: . ': ':..e c.. .::.s r: '

This modification did not alter the function, operation, or qualification of equipment within the affected systems or any other systems.

Revisions to the UFSAR were required as a result of this modification.

DCN W17792- Installation of Safety Related Air Conditioning Units Unit: 1 DCN W17792 installed two redundant safety related air conditioning units to cool both the Unit I elevation 593'nd elevation 621.25'eactor Building electric board rooms. This modification involves installing the two 25-ton air conditioning units, drains, condenser water piping, ductwork, dampers, cables and remote and local control stations to provide redundant cooling capacity. The system is capable of maintaining the rooms at or below 104'F during normal and accident conditions to allow for operation of the electrical panels. Cooling water for the air conditioning units is supplied by the Emergency Equipment Cooling Water (EECW) System which was modified to include two supply lines which were routed to a common air conditioning unit header. The power for these units is Class IE and from separate power divisions. In case of loss of offsite power, the reactor motor operator valve boards 1A and 1B are supplied by the diesel generators.

This modification involves adding new connections to the EECW System. However, these connections do not have an adverse impact on the safety related functions of the EECW. Adequate separation, redundancy, and qualification of the added connections insure that EECW is capable of performing its safety related functions.

This modification is safe from a nuclear safety standpoint. No unreviewed safety question exists.

DCN W17846- Control Bay Ductwork Modifications Unit: 0 This DCN provided design modifications to enhance control bay air distribution by installation of balancing dampers and new seismic Class I ductwork to rooms previously supplied by mechanical exhaust only. New ductwork to the Unit 2 shutdown board room provides additional cooling due to battery charger equipment upgrades. This DCN ensures acceptable control by environmental temperatures during post-loss of coolant accident conditions.

Revision A of this modification was based on input to a single calculation (MD~031-920537). Subsequently, the heat loads were measured and the actual heat loads were used as the basis for revising the control by Heating, Ventilation, and Air Conditioning (HVAC) analysis calculation, MD+031-920537. Revision 3 of MD+031-920537, based on an HVAC configuration requiring fewer modifications than W17846A, resulted in satisfactory results. As a result, W17846 was downscoped to delete the safety related cooling to the cable spreading rooms and reduce the scope of the duct revisions (W17846B).

Revision 2 of the safety evaluation reflects the removal of fire dampers 144, 153, and 154 as documented by DCN F33059 and Fire Protection Report Number FPR-94019. The removal of these fire dampers is within the design intent of DCN W17846B but was not documented by DCN W17846B. In addition, Fire Protection Report Volume 1 was revised by FRP-94014 to reflect the deletion of the spreading room exhaust fans from the Safe Shutdown Analysis Section of the Fire Protection Report per DCN W17846B.

This DCN did not delete or modify any system or equipment protective safety features. This DCN willonly improve

Tennessee Valley Authority Brogans

SUMMARY

OF Ferry Nuclear Plant SAFETY EVALUATIONS FOR 1995 Annual Operating Report FIELD COMPLETED PLANTMODIFICATIONS control bay HVAC System support of equipment important to safety. Changes were required to the associated UFSAR flow diagram figures.

DCN W17908 - Reactor Building Fire Protection Upgrade Unit: 3 This modification installed new fire detection equipment in the Unit 3 Reactor Building. This DCN also provided for the completion of the data communications loop, which connects each of the local fire alarm panels in Unit 3 Reactor Building with the previously installed portion of the data communications loop by DCN W17904.

The new system is designed to provide an early detection of fires with a high degree of reliability.

When this safety evaluation was approved, it was determined that portions of Technical Specifications Section 3.11/4.11 pertaining to the fire detection system would be'aQ'ected by implementation of the modification.

However, those sections were to be removed from the Technical Specifications in accordance with the requirements of NRC Generic Letter 88-12. Those portions of the Fire Protection Report affected by this modification are contained in CRLD BFEP-MN-93003.

This modification is safe from a nuclear safety standpoint. No unreviewed safety question exists.

DCN W17933 - Backup AirSupply Unit: 3 DCN W17933 provided for modification of the Control Air System and the Containment Atmosphere Dilution System in order to provide a reliable backup source of control gas to the pneumatic operators on valves 3-FCV4644020 and 3-FCV4021, which are part of the Suppression Chamber Vacuum Relief System. This DCN modified the Control Air System and the CAD System so that nitrogen from the CAD System can be used to operate the butterfly valves in the vacuum relief lines when the, Control Air System cannot provide the necesmy compressed air.

These modifications constitute a plant enhancement that will ensure reliable operation of the butterfly valves in the Suppression Chamber Vacuum Relief System under all design basis conditions. No change to the basic operational function of the butterfly valves or the CAD System is involved.

DCN W18028 - Fire Damper Installation Unit: 3 This safety evaluation was written for DCN W18028 and in support of Fire Protection Report Change Notice FPR-95006, Revision 0, to Fire Protection Report Volume 1. DCN W18028 provided 27 seismic Class I procured fire dampers in the Unit 3 Reactor Building to minimize the potential spread of fire between fire zones and meet the separation requirements of 10CFR50 Appendix R. The fire dampers were installed in the reactor zone and refueling floor supply and exhaust systems duchvork floor penetrations at elevations 593.0', 621.25', and 639.0'f the Reactor Building.

These fire dampers arc three-hour rated which will maintain the existing one-hour fire barrier rating for the floor penetrations under a postulated fire in a specified zone.

Tennessee Valley Authority

SUMMARY

OF Brains Ferry Nuclear Plant 1995 Annual Operating Report, SAFETY EVALUATIONS FOR FIELD COMPLETED PLANTMODIFICATIOJVS The fire dampers are a curtain type, spring loaded, and self actuated (fusible link set at 165'F) meeting the requirements of Underwriters Laboratories 555. Duct access door openings are provided for routine maintenance and inspection. These fire dampers are designed to close against the air flow; therefore, shut offof the reactor zone and refueling floor supply and exhaust fans during a fire is not required.

The changes made to the Fire Protection Report by FPR-95006 are minor technical and editorial in nature. The changes update and clarify text, tables, and drawings, and bring the Fire Protection Report into conformance with the current system configurations and equipment alignment. The changes conform with the Appendix R analysis and supporting evaluations/calculations.

DCN W18049- ModiTication of Automatic Preaction System for Unit 3 Reactor Building Unit: 3 DCN W18049 provided for modification of the automatic preaction system for the Unit 3 Reactor Building elevation 565', 593', and abandonment/removal of the existing water spray systems on floor elevation 593'.

The protection from the existing water spray system is no longer required due to enhanced ceiling and below obstruction coverage. Decommissioning of these water spray systems also removes the possibility of spurious valve actuauons causing water spray damage to plant equipment.

The existing preaction sprinkler system in the Unit 3 Reactor Building was extensively modified in order to bring the system into conformance with design standards of the National Fire Protection Association (NFPA). The deviations from NFPA code requirements that existed were identified to the NRC. A commitment to upgrade the fire suppression system for the Unit 3 Reactor Building elevation 593's contained in the "Fire Protection Upgrade Program" which is part of the Fire Protection Report Volume 3. This modification installed piping, piping supports, sprinkler heads, low point drain valves and inspector test assemblies for the Unit 3 Reactor Building's preaction sprinkler system at elevation 593'. The system provides sprinkler coverage for the majority of the elevation 593'loor area including two water curtains, one for the main equipment hatch and one for the stairwell located in the Southeast corner.

The preaction sprinkler system branch piping added by this modification was connected to an existing feed main header supplied by preaction valve 3-FCV426-77. This existing preaction valve was provided with a new air supervision system under DCN W18048. This modification does not provide coverage for the elevation 593'oom containing the regenerative and non-regenerative heat exchangers. There are no appreciable combustibles in this high radiation area, therefore, coverage is not provided or required in this area. The existing sprinkler heads protecting the clean-up backwash receiving tank room are closed head type and are acceptable for continued use due to low combustibles in the room and radiation as low as reasonably achievable (ALARA)concerns.

DCN W19321- Anticipated Transients Without Scram (ATWS) Modifications Unit: 3 DCN W19321 completed the implementation of the Recirculation Pump Trip (RPT) and Alternate Rod Insertion (ARI) portions of the ATWS modifications for BFN Unit 3. This modification is similar to the change implemented on BFN Unit 2 by ECNs P7045, P7089, P7090, and DCN W17249.

Appropriate annunciation, indication, and computer signals were also provided for the ATWS System by this DCN.

Tennessee Valley Authority SUM8lfARYOF Broils Ferry Nuclear Plant SAFETY EVALUATIONS FOR

/995 Annual Operating Report FIELD COMPLETED PLANTMODIFICATIONS

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'I'he ATWS (AIU/f~')System is not safety related. The changes do not alter the function or method of operation of any safe shutdown system. The ATWS (AIU/I1Iri')modification miniinizes the probability of a spurious trip. This modification enhances the existing reactor protection features and does not reduce the margin of safety.

UFSAR sections were updated by change request BFEP-BNA-92253 to document the changes implemented by this modification.

Unit 3 Technical Specifications, Sections 3.2/4.2, were updated during the implementation of the ATWS modification on BFN Unit 2.

DCN W19785- CORRECTION OF ELECTRICALPROGRAM CONCERNS Unit: 3 This DCN corrected electrical program concerns required for Unit 3 recovery. In general, this DCN provided design to modify the control circuits for various breakers to provide isolation for Appendix R concerns; also, cables were replaced by this DCN with environmentally qualified cables; provided raceway and cable reroutes for separation concerns; and relabeled cables to correctly reflect the divisional grouping as defined in Design Criteria BFN-50-728.

This modification has no adverse impact on the standby diesel generators, the 250V DC Power System, the Diesel 125V DC System, the 4kV. shutdown boards, or any other system. 4kV shutdown boards A, B, C, D, 3EA, 3EB, 3EC, and 3ED operate and function the same as they did before this modification.

Revision B of this DCN added relays and cables to the following diesel breakers: 1818, 1822, 1812, 1816, 1838, 1842, 1832, and 1836. This affected all of the diesel generators associated with Units 1, 2, and 3. The requirements for the Appendix R manual actions resetting 86G lockout relays and resetting/closing diesel field breakers (for Diesels A, B, C, and D) added by corrective action of BFPER930014 RO was removed by this modification. This modification willprevent potential reverse power flow to diesel generators during an Appendix R event, thereby eliminating the requirement for additional manual actions.

No description in the UFSAR requires revision as a result of this modification. However, the handswitches, relays, and fuses installed by the DCN involve circuits depicted on UFSAR Figure 8.4-3 requiring a change to the UFSAR.

DCN W20145 - Curb Modification and High Pressure Coolant Injection (HPCI) Room Floor Drain ModiTication Unit: 3 DCN W20145 was issued to allow for the drainage of flood water from the Unit 3 main steam valve vault at elevation 565'o the torus room at elevation 519'ollowing a postulated feedwater line break in the valve vault and to prevent flood water drained to the torus room from entering the HPCI room. This DCN modified main steam and feedwater valve vault and HPCI room entrance curbs, floor drains, and floor penetrations, as required, to provide flood protection from internally produced flood waters for the Reactor Building general floor area and HPCI room.

An 8" concrete curb was installed in the labyrinth passage leading to the valve vault to prevent flooding of the 565'levation Reactor Building general floor area. Additionally, two existing 24" floor penetrations in the valve vault area, which were previously sealed, were opened to provide a flow path from the valve vault to the torus room.

Tennessee Valley Authority SUN'MARYOF Broi9pns Feny Nuclear Plant SAFETY EVAI.UATIOltVS FOR 1995 Annaai Operating Report PIELD COMPLETEDPIANTMODIPICATIONS Flood protection for the HPCI room was provided by increasing the height of the existing curb at the entrance to the HCPI room from 12" to a new height of 18".

Neither the function of the Reactor Building structure nor the Radwaste System was changed by this modification.

DCN W20217- Redesign of Common Accident Signal Logic Unit: 2,3 DCN W20217 restored the Unit 3 inputs to the auident signal logic that is shared by all three units. This DCN also modified the Unit 2 and Unit 3 diesel breaker trip circuitry so that following an initiation of a common accident signal (which trips the diesel breakers), a second diesel breaker trip on a "unit priority" basis is provided to ensure that the diesel supplied buses are stripped prior to starting the Residual Heat Removal pumps and other Emergency Core Cooling System loads in response to a RHR initiation signal. The other unit's diesel breakers are unaffected by this trip. This new diesel breaker "Unit Priority Re-Trip" willonly occur ifan accident signal (spurious or real) has previously initiated the diesel breaker trip circuitry.

Only BFN Unit 2 and 3 operation and accident signal logic are addressed in this safety evaluation. Unit 1 inputs to the common accident signal and pre-accident signal logic, and the Unit 1 interface with the Unit 2 RHR and Core Spray pump starting logic, remain disabled and have not been analyzed.

This modification does not prevent any systems or components from fulfillingtheir safety related functions.

This modification is safe from a nuclear safety standpoint. No unreviewed safety question exists.

\

DCN W20899 - Motor Operated Valve (MOV) Upgrade Program Unit: 3 DCN W20899 provided information for the modification to various MOVs in the High Pressure Coolant Injection (HPCI)

System. These changes included changes to actuators and other components that are affected by the GL 89-10 program, environmental qualification program, Appendix R program, and miscellaneous non-integnted issues. One of the changes involved changing the gear ratio in valve 3-FCV-73-35. This change in gear ratio changed the stroke time for this valve.

This stroke time is addressed in section 7.4.3.2.5 of the UFSAR.

This safety evaluation focuses on the stroke time change only because of the impact on the UFSAR as a result of this DCN.

The increased stroke time does not involve an unreviewed safety question because the valve is normally closed and closes if it receives the appropriate containment isolation signal. There is no stroke time established for this function. For HPCI the opening stroke time will still allow adequate HPCI flow into the reactor. The closing stroke time for the HPCI is not a concern because the valve does not need to be closed to allow rated HPCI flow into the reactor vessel. The increased stroke time does not impact any other HPCI components or the way the HPCI System responds to any initiation signals.

Tennessee Valley Authority SUM1KCRYOF Brogans Feny Nuclear Plant SAFETY EVAL,UATIONS FOR 1995 Annual Operating Report FIELD COMPLETED PLANTMODIFICAT101VS KAi::-:::;:.::::::,'-:::.i,-:,:;-.,:::i: "::"::::i i.,::::::i:.:i:::i::::.":.':::i:--::::-:::::::-.::-:-: i':::::!

DCN W21284 - 480V Load Shed Logic Unit: 3 This DCN added automatic load shed logic to the Unit 3 480V AuxiliaryPower System. Previously, the 480V Auxiliary Power System had provision for undervoltage load shedding only. Additional load shed logic was added to accommodate a loss of coolant accident event concurrent with loss of offsite power.

The separation and isolation design for the load shed process ensures that the connection to existing equipment will not modify the functional or performance characteristics of the affected loads. Post modification testing was performed to demonstrate that the operation, performance, and function of the affected components remained unchanged.

Addition of automatic load shedding potentially affects Unit 3 Technical Specification Section 3.9.A.S by addition of a requirement that Unit 3 load shed system logic be operable. In addition, Section 4.9.A.3 is potentially affected with a requirement for surveillance activities to ensure load shed system logic operability. A change to the Tcchnical Specifications has been proposed. The proposed Technical Specification change is identical to that included in the Unit 2 Technical Specifications.

This modification is safe from a nuclear safety standpoint. No unreviewed safety question exists.

DCN W21813- Replacement/ModiTication of Electrical Components Unit: 3 DCN W21813 was issued to replace components in the Radwaste System, Reactor Water Cleanup System, 250V DC Power System, and 240V AC Lighting System which had been identified in the Restart Equipment List as breakages. The change involved cable and component replacement and cable rerouting to correct separation anomalies. Implementation of this DCN involved the replacement of cables, fuses and fuseblocks, cable splices, thermal overloads, level switches and limit switches.

No change is made to the function, operation, or performance of the Radwaste System or any other systems associated with this DCN.

This modification is safe from a nuclear safety stand point. No unreviewed safety question exists.

DCN W21917- Replacement/Modiifiication of Electrical Components Unit: 3 DCN W21917 was issued to resolve selected Restart Equipment List breakage items Rom the following systems: Fire Protection System, Qffgas System, Reactor Building Closed Cooling Water System, Recirculation Flow Control System, Fuel Pool Cooling System, Traversing Incore Probe System, Control Air System, and the Reactor Feedwater System.

Implementation of this DCN involved the installation of cables, fuses and fuseblocks, breakers, solenoid valves, and check valves.

Tennessee Valley Authority

SUMMARY

OF Brains Ferry Nuclear Plant SAFETY EVALUATIONSFOR 1995 Annual Operating Report FIELD COMPLETED PLANTMODIFICATIONS DCN W27299 - Deletion of Non-Essential Raw Cooling Water (RCW) Feeds to Emergency Equipment Cooliag Water (EEC%)

Unit: 3  !

This DCN removed the piping, valves, and fittings that make up the RCW and EECW System interface ties to equipment normally supplied by EECW such that only EECW willsupply or have the capability to supply the equipment. The EECW System is the primary source of cooling water to the affected safety related equipment and the RCW System serves a backup function. However, the RCW System is not a safety related system and cannot be relied upon to provide cooling water following a design basis earthquake. In addition, the stagnant Qow piping ia the interface between the RCW and the EECW Systems causes severe corrosion problems in piping and valves.

This modification eliminated an interface that is not required for safe operation or shutdown of the plant. The modification eliminated equipment requiring periodic inspection and maintenance but it did not change how the RCW System or the EECW System functions. The welded caps provide the required pressure boundary and the aQ'ected secondary containment penetrations willprovide isolation in accordance with applicable requirements.

A safety evaluation was required because changes to the UFSAR were required.

DCN W29948- Upgrade of Instrumeat and Coatrols Portion of Condeasate Demineralizer System Unit: 3 This DCN replaced the controls for the Unit 3 Condensate Demineralizer System with a progranunable logic controller based system. Previously existing panel 25-121 was removed and replaced with a new panel. This new panel contains the programmable logic controller system as well as controls and indications necessary for manual control of the demineralizers should both progranunable logic controller subsystems fail. This DCN replaced most of the monitoring instrument loops associated with the control system. This DCN also electrically connected Condensate Demineralizer System valves replaced by DCN T34196.

This modification did not alter the condensate demineralizers in such a way as to deviate from the UFSAR functional description. It resulted in replacement of obsolete equipment with statecf-the-art instrumentation and controls.

This modification is safe from a nuclear safety standpoint. No unreviewed safety question exists:

DCN W29949 - Condensate Demineralizer Vessel Internal Upgrade Unit: 3 DCN W29949 provided for Unit 3 condensate demineralizers vessel internal modifications. This modification involved removal of the existing internal components and provided an upgraded filter lattice structure along with enhanced filter elements.

The Condensate Demineralizer System is not a safety related system and is not required for safe shutdown of the reactor. This modification required a change to the UFSAR in regard to the square footage of the demineralizer vessels.

This modification is safe from a nuclear safety standpoint. No unreviewed safety question exists.

41-

Tennessee Valley Authority

SUMMARY

OF Broivns Feny Nuclear Plant SAFETY EVALUATIONSFOR k '<<

1995 Annual Operating Report

" " r<<'iCY FIELD COMPLETED PLANTMODIFICATIONS DCN W33292 - Control Air Cross Ties Unit: 0 Section F.4.g of Appendix F of the UFSAR, "Unit Sharing and Interactions" states, "Incidents or accidents in one unit shall not directly influence the operation of the other units...". The existing field configuration and the as~nstructed drawings of the Control Air System did not comply with this requirement. Header isolation valves 1-32-586 and 1-32-2378, which separate Unit 1 and Unit 2 Control Air headers were open. This was also true for the Unit 2 and Unit 3 header isolation valve 2-32-976.

The unit cross tie valves were originally required to be closed when ECN L1410 was implemented. ECN L1410 provided one operating standby dryer-filter station for all three units. The addition of the standby dryer-filter stations along with positive unit isolation of the Control Air header was intended to provide more reliability and protection from multi-unit trips. This is consistent with UFSAR Section 10.14.4.1 which states that the

"...Control Air header for each unit is connected to the adjacent unit through manual valves which are normally closed."

Cross-tying the individual units'ontrol Air headers through the open header isolation valves had become necessary because of the unreliability of the Control Air diyers, whose failure can cause unexpected depressuration of the affected header. As a result, several enhancements to the existing system were identified which would significantly improve its reliability.

These enhancements included an automatic unit segregation on decreasing pressure in the Control Air System. This is accomplished by adding pressure controlled valves between Units 1 and 2, and between Units 2 and 3. These valves are in the Control Air header and willautomatically isolate the low pressure side of the header, effectively isolating a Control Air loss to a single unit. This satisfies the UFSAR requirement that incidents in any one unit will not directly influence the operation of the other units.

These enhancements also included upgrading the Control Air compressors control system and increasing the system operating pressure to 115 psig (the Control Air receivers have been pressure tested to 150 psi). The control system was also enhanced by replacing the pressure regulators and pressure diflerential transmitter with two Rosemount pressure transmitters and by replacing the mercury manometer with two Yogogawa solid state profpammable controllers, one normal and one backup. These replacements will not change the function of the circuit in any way but will allow more precise control of the loading and unloading of the compressors, resulting in a narrower operating band for the compressors.

These changes provide improved Control Air supplies to the operating units, increases unit reliability, increases unit availability, reduces the potential for SCRAM due to loss of Control Air in adjacent units, and brings plant configuration into compliance with licensing documents. A safety evaluation was required because the modifications change text and figures in the UFSAR.

DCN W34880- Unit Preferred Transformer Redistribution Unit: 3 DCN W34880 removed the non-safety related alternate power feed for the Unit 3 unit preferred transformer and distribution system from the safety related 480V shutdown board 3B compartment 5B and relocated it to 480V reactor motor operated valve (RMOV) board 3B.

A2-

Tennessee Valley Authority

SUMMARY

OF Brains Ferry Nuclear Plant R4FETYEVALUATIONS FOR 1995 Annual Operating Report FIELD COMPLETED PLANTMODIFICATIOiVS R'~':

The alternate feed for the Unit 3 unit preferred bus was fed from 480V shutdown board 3B compartment SB. The Unit 3 safety related Instrument and Control (INC) Bus B is also fed from the same location, resulting in electrical separation and isolation concerns.

This DCN separated these two loads by relocating the Unit 3 unit preferred transformer feeder from 480V shutdown board 3B compartment SB to 480V RMOV board 3B compartment 17D. Electrical isolation from the safety related 480V RMOV board willbe provided by use of a Class 1E circuit breaker. The Unit 3 safety related INC Bus B remains connected to 480V shutdown board 3B compartment SB.

Removal of the unit preferred load from 480V shutdown board 3B required a breaker setting change for the circuit breaker associated with 480V shutdown board 3B compartment SB.

Moving the Unit 3 unit preferred transformer feeder to 480V RMOV board 3B compartment 17D required the addition of an electrical bucket assembly for insertion to this compartment as well as a new electrical circuit breaker. Additionally, to assure adequate electrical protection of unit preferred transformer, a fused disconnect switch was installed by this DCN.

The fused disconnect switch was located physically near transformer and electrically upstream of TUP3.

The cables from 480V RMOV board 3B to the disconnect switch and from the disconnect switch to the unit preferred transformer are new Class 1E qualified cable and are run in conduit. Reworked ground cable from the Unit 3 unit preferred transformer to a new splice to the Unit 1 unit preferred tiansformer ground is also new Class 1E qualified cable, run with the power feed cable in conduit. These new cables are run in dedicated conduit (some existing), either alone or with these two cables only.

This DCN also reset the unit preferred motor-motor generator set 2 tie to battery board 3 emergency feeder breaker (battery board 3, breaker 1003). This breaker had been identified as not providing adequate coordination in baseline electrical calculation ED+3999-920121.

The separation of safety related and non-safety related loads and the changing of settings enhances equipment reliability.

Analyses have been prepared to assure that circuit breaker settings and cable installation represent the proper selection, application, and adjustments for each specific circuit, and that required margins and equipment protection have been assured.

DCN W36676- Installation of Zinc Injection System Unit: 3 This DCN installed a Zinc Injection (GEZlp) System for Unit 3. The GEZIp System consists of a simple recirculation loop offof the feedwater pumps. Specifically, the zinc solution is obtained by passing a stream of feedwater (4-100 GPM) from the feedwater pumps discharge header, through a dissolution vessel containing pelletized DZO, and back to the feedwater pumps suction header.

The Zinc Injection System is a part of Feedwater System and is a non-safety related installation.

A new section 11.8.3.9 willbe added to the UFSAR to describe the Zinc Injection Skid operation.

The reactor water chemistry limits discussed in Technical Specification 3.6/4.6 willbe maintained.

This modification is safe from a nuclear safety standpoint. No unreviewed safety question exists.

t Tennessee Valley Authority

SUMMARY

OF Browns Feny Nuclear Plant SAFETY EVAL,UATIO1VSFOR

! 995 Annnai Operating Report FIELD COMPLETED PLAiVTMODIFICATIOiVS

.: 'g".,Qit':;", ea;:Mpi injjp" ."eEa sr!::"."-,:!/tvetei::,.j) i.'"'iv "'en,,+r,,;'V':;:ntit ECN P00613 - Modification to Main Steam Isolation Valves to Reduce Valve Leak Rates During Refueling Outage Leak Rate Testing Unit: 3 h

ECN P00613 provided for alteration to the MSIVs to reduce their leak rates during refueling outage leak rate testing and to improve their reliability to maintain low leakage rates.

The operation of the MSIVs is not changed by the modification. The valves'unction is not altered. The valves willactually be capable of providing better isolation due to the reduced leak rates. This change does not alter the MSIVs function or closing time nor does it negate the MSIVs seismic qualification.

This modification is safe from a nuclear safety standpoint. No unreviewed safety question exists.

ECN P07084- Tubing Supports and Routing Unit: 3 This ECN involved the OQ'gas System, Radwaste System, Radiation Monitoring System, and the Traversing Incore Probe System. The changes included additions, deletions, and modifications of existing tubing supports and rerouting of tubing. These modifications were made in response to seismic and thermal design concerns. The change in configuration of tubing is to facilitate proper support of tubing to meet seismic and thermal requirements and not to change the intended function of the tubing.

This modification is safe from a nuclear safety standpoint. No unreviewed safety question exists.

Tennessee Valley Authority SUMllfARYOF Browns Feny Nuclear Plant SAFETY EVAI.UATIONS FOR 1995 Annual Operating Report FIRE PROTECTION REPORT REVISIONS 1995 SUM Y OF SAFETY EVALUATIONS.

FOR FIRE PROTECTION REPORT REVISIONS Tennessee Valley Authority

SUMMARY

OF Browns Ferry Nuclear Plant SAFETY EVALUATIONS FOR l995 Annual Report - FIRE PROTECTION REPORT REVISIONS j::,:::.:~:,:,Operating M',"%~"!,"." .::;...,:.:,:.,"%g,,

Fire Protection Report - Incorporation of Unit 3 Safe Shutdown Program This safety evaluation was performed to evaluate changes to the Fire Protection Report, Volume 1 due to the incorporation of the Unit 3 Safe Shutdown Program.

The previous report reflected the fire protection program for BFN with Unit 2 being the only operational unit. The revised report reflects the Appendix R safe shutdown capability of the plant with both Unit 2 and Unit 3 being operational.

The Unit 2/Unit 3 Fire Protection Report revised the Safe Shutdown Analysis and the Safe Shutdown Program sections to include the Unit 3 fire protection program. The Fire Protection Plan and Fire Hazards Analysis sections were revised as necesmy to reflect the modifications of Unit 3.

\

This Fire Protection Report revision was a documentation change only and did not perform modifications to any systems or conduct any tests or experiments.

The dual unit (Unit 2/Unit 3) Fire Protection Report is a consolidation of the fire protection program at BFN for Unit 2 and Unit 3. The Unit 3 program is similar to the Unit 2 program as provided within the guidelines of 10CFR50 Appendix R. The safe shutdown analysis provided for Unit 3 considered the unit specific components and the components common to both units. Modifications required to assure compliance to the program were performed via design change notices (DCNs). These design changes were evaluated for adverse impact to the plant safety systems individually within the DCN 10CFR50.59 process.

Fire Protection Report Change - FPR-95012 This revision to the Fire Protection Report, Volume 1, encompassed the following areas'of change: a) organizational and title changes within TVANuclear Senior Management, b) added a statement pertaining to the agreement between TVANuclear and TVA Fossil and Hydro relative to the fire brigade incident commander training, c) indicates the transfer of fire protection engineering activities to the Fire Protection Section, d) added a statement indicating that the Fire Protection Section responsibilities including secondary responsibility for the Fire Hazards analysis and Fire Protection Report Plan, implementation of the Fire Protection Program, responding to fire emergencies, and provide system engineering support for all fire protection systems, e) revised Table 9.3.11.I, reflecting the revised fire protection support staff, implementing the site standardization goals within TVANuclear, I) updated Table 9.3.11.E for fire rated doors to match the detection panel previously made to Table 9.3.11.A for fire detection instrumentation, and g) revised the Fire Hazards Analysis Section introduction) to include information on reviews of information notices, generic letters, operating experiences, etc.

Changes a) through e) above were necessary to reflect the current status of those areas. Change I) was made to remove inconsistencies between the two tables and to ensure that the appropriate information is provided in the tables for the purpose of taking compensatory measures. Change g) was made to provide information regarding reviews of NRC Information Notices and Generic Letters, Operating Experience, etc.

S Tennessee Valley Authority

SUMMARY

OF Brains Feny Nuclear Plant SAFETY EVALUATIONS FOR 1995 Annual Operating Report FIRE PROTECTION REPORT REVISIO1VS R~ '~":~ ~'&WAR"~':~4 '~'""4k "'"'::.~ac-"- '": ': '-

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Fire Protection Report Change Notice No. FPR-95023 This safety evaluation was performed to evaluate changes identified in the Fire Protection Report Change Notice No.

FPR-95-23 to the Fire Protection Report, Volume 1. The changes are a result of editorial corrections, changes to combustible loading as reflected in a revision to calculation MD-N0026-910163, and errors identified in the Fire Protection Report, Volume 1 during a review of the report.

The changes in combustible loadings were due to modifications performed under various Unit 3 design change notices (DCNs). Allchanges to combustible loads have been evaluated under the specific DCN safety assessment and Appendix R reviews. The changes were found to be acceptable.

The changes correct or clarify the location of the compensatory actions required by the Appendix R Safe Shutdown Program.

This change is safe from a nuclear safety standpoint. No unreviewed safety question exists.

47-

SUMMARY

OF t

Tennessee Valley Authority Brains Feny Nuclear Plant SAFETY EVALUATIONS FOR 1995 Annual Operating Report NEH'INSTRUCTIONS/PROCEDUREREVISIONS 4jdP:;<<<:<Ã~:<;:.".r'vjgj<'<<>>.,"'-@<':>4)A's 6"8~"" > '~4" ".:<:i<a@@."::;:z,'.>> %<<'.44/>>~~s+o. q.- - .g~~," p >>-s <'vQ:p"" <~:'z'>>.

1995

SUMMARY

OF SAFETY EVALUATIONS FOR NKW INSTRUCTIONS OR PROCEDURE REVISIONS 48-

Tennessee Valley Authority SUNDRY OF Brains Feny Nuclear Plant SAFETF EVAI.UATIONS FOR 1995 Annual Operating Report

,.""',:".!i%~i'iiiiiii'~l';,::;';,':-.'ji.';;.ll':lF.' i...:. ~!~8~!iiiPi"',::::;ll"',::;:;!~%~..".".'ll'iN" "':~!P NEH INSTRUCTIONS/PROCEDURE REVISIONS

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3-SI47.A.2.a-f - Primary Containment Integrated Leak Rate Test This surveillance instruction implements the requirements of 10CFR50, Appendix J, and Technical Specdication 4.7.A.2. The requirements include leak testing primary containment at accident pressure once every 40 + 10 months during cold shutdown condition. This leak rate test involves measuring temperature, pressure, and dewpoint of the primary containment atmosphere with the reactor vessel vented over an 8-hour (minimum) period at a minimum diQerential pressure of 49.6 psi..The actual test pressure is 50.8 +.4 psi. The data is used to correlate the mass leak rate of primary containment atmosphere at accident prcssure over a 24-hour period. This instruction is performed in conjunction with 3-Tl-179 that'installs test instrumentation and 3-Tl-173 that inspects primary containment, leak tests the core spray pump'seals, and bubble tests air leakage paths from priaary containment.

3-SI47.A.2.a-f will have no negative effects on the UFSAR described safety functions. This surveillance instruction willensure the leak tight integrity function of primary containment as required by the Technical Specifications.

This change is safe from a nuclear safety standpoint. No unreviewed safety question exists.

Tennessee Valley Authority SUMIISARYOF Brains Feny Nuclear Plant SAFETY EVALUATIONFOR 1995 Annual Operating Report SPECIAL OPERATING CONDITIONS 1995 SUM Y OF SAFETY EVALUATION FOR SPECIAL OPERATING CONDITION 0

Tennessee Valley Authority

SUMMARY

OF Broils Feny Nuclear Plant SAFETYEVALUATIONFOR 1995 Annual Operating Re ort . SPECIAL OPERATING CONDITIONS Lost Article Report 3-95-034 This safety evaluation addresses BFN Unit 3 Lost Article Report Number 3-95434. The lost article is a 12" cable tie. The cable tie is primarily composed of a nylon 66 material made by DuPont and has a small stainless steel chip at the cable tie head. The cable tie was spotted in the Unit 3 reactor cavity on the upper core Qange at azimuth 270'on August 10, 1995, but could not be recoveraL At the time this piece was discovered, the Unit 3 core was oK-loaded.

A cable tie may be strong enough to possibly interfere with control rod drive operation prior to vessel heatup ifit were to migrate into the internals of a control rod drive. The lost article could migrate into a control rod guide tube and into the annulus opening, but it would be trapped in the control rod drive screen Qlters and would not be able to move into the drive internals. AQer heatup to rated temperatures, the cable tie is expected to decompose. Similarly, the stainless steel chip in the cable tie head would also be trapped in the screen Qlters ifit were to migrate to the drive.

This change is safe from a nuclear safety standpoint. No unreviewed safety question exists.

Lost Article Report 3-95435 This safety evaluation addresses BFN Unit 3 Lost Article Report Number 3-95435. The lost article is a 1" by 3" piece of duct tape. The duct tape was spotted in the Unit 3 reactor core at core location 53-22 on August 10, 1995, but could not be recovered. At the time this item was identified as a lost article, the Unit 3 core was off-loaded.

l A piece of duct tape may be strong enough to possibly interfere with control rod drive, operation prior to vessel heatup ifit were to migrate into the internals of a control rod drive. The lost article could migrate into a control rod guide tube and into the annulus opening, but it would be t'rapped in the control rod drive screen Qlters and would not be able to move into the drive internals. After heatup to rated temperatures, the lost duct tape is expected to decompose.

This change is safe from a nuclear safety standpoint. No unreviewed safety question exists.

Lost Article Report 3-95450 This safety evaluation addresses BFN Unit 3 Lost Article Report Number 3-95450. The lost article is a cigarette butt. The cigarette butt was spotted on the Unit 3 reactor core plate at core location 1649 on August 24, 1995, but could not be recovered. At the time this item was identified as a lost article, the Unit 3 core was off-loaded.

A cigarette butt may be strong enough to possibly interfere with control rod drive operation prior to vessel heatup ifit were to migrate into the internals of a control rod drive. The lost article could migrate into a control rod guide tube and into the annulus opening, but it would be trapped in the control rod drive screen Qlters and would not be able to move into the drive internals. After heatup to rated temperatures, the cigarette butt is expected to decompose.

This change is safe from a nuclear safety standpoint. No unreviewed safety question exists.

Tennessee Valley Authority

SUMMARY

OF Brains Ferry Nuclear Plant ~

SAFETY EVALUATIONFOR 1995 Annual Operating Report SPECIAL OPERATING CO1VDITIONS P""-""::::""'"-:"""'""""" -""""'" """'"'"'""':

Lost Article Report 3-95-2I This safety evaluation addresses BFN Unit 3 Lost Article Report Number 3-95-21. The lost article is a socket cap screw Rom the Unit 3 refueling bridge. The screw was about 1/2" long with a screw head diameter of 1/4" and was made of carbon steel. The cap screw was determined to be missing over the open Unit 3 reactor vessel on July 13, 1995. At the time the cap screw was lost, the Unit 3 core was off-loaded. The exact location where the screw fell is not known; however, there are four locations where the screw could have fallen: 1) Into a control rod guide tube; 2)

Onto the core support plate; 3)'nto an empty fuel support casting and resting inside the fuel support casting cavity;

4) Into an empty fuel support casting, through a fuel support casting inlet orifice, and into the vessel bottom.

The missing cap screw is of insufficient mechanical strength to cause a recirculation pump seizure. For a rod drop accident to occur, the control blade must become decoupled from the control rod drive. However, there is no credible mechanism by which the cap screw could cause the blade to uncouple Rom the driv.

The lost article is strong enough to possibly'nterfere with control rod drive operation ifit were to migrate into the internals of a control rod drive. The lost article could migrate into a control rod guide tube and into the annulus opening, but it would be trapped in the control rod drive screen filters and would not be able to move into the drive internals.

Lost Article Report 3-95-58 This safety evaluation addresses BFN Unit 3 lost articles described on Lost Article Report Number 3-95-58. Several of the lost articles documented under report 3-95-58 were subsequently recovered from the reactor vessel and are not addressed in this safety evaluation. The remaining lost articles are as follows: 1) A knotted, frayed string hanging Rom jet pump ¹1 just below the wedge. 2) A very small piece of white tape between jet pump ¹7 diffuser and the vessel wall. 3) A small pine of twisted, threaded metal laying on the diffuser plate at the base ofjet pump ¹20.

These lost articles" were identified during an inspection of the Unit 3 annulus region conducted on October 15, 1995, but could not be recovered. Lost articles 1 and 2 are lightweight, as evidenced by the fact that they were swept into the annulus region from the vessel lower plenum by the relatively low flow velocities generated within the vessel by the Residual Heat Removal (RHR) System in the shutdown cooling mode. Article 3 is believed to have been in the annulus region for some time before RHR operation in the shutdown cooling mode (it is small enough that it could have been overlooked in previous inspections of internal components in the vessel). At the time these articles were identified, the Unit 3 core was oQ'-loaded.

The missing string/tape/metal are of insuificient mechanical strength to cause a recirculation pump seizure. For a rod drop accident to occur, the control blade must become decoupled &om the control rod drive. However, there is no credible mechanism by which the lost articles described in this safety evaluation could cause the blade to uncouple Rom the drive.

Lost Article Report 3-95-59 This safety evaluation addresses BFN Unit 3 lost arucles described on Lost Article Report Number 3-95-59. Several of the lost articles documented under report 3-95-59 were subsequently recovered from the reactor vessel and are not addresses in this safety evaluation. The remaining lost articles are as follows: 1) A small piece of plastic or paint chip laying on the diffuser plate between jet pumps ¹4 and ¹5. 2) A small piece of plastic or paint chip laying on the difiuser plate between jet pumps ¹1 and ¹20. 3) A dark chip of unknown material lying on the diffuser plate near jet pump ¹6. 4) A 1" by 2" piece of unknown material (plastic or paper) under instrumentation lines near jet pump ¹16.

5) A ball of tape lying on the diffuser plate between the base ofjet pump ¹19 and the vessel wall.

Tennessee Valley Authority SUN8IIARJr OF Browns Ferry Nuclear Plant SAFET7/EVALUA TION FOR (995 Annual Operating Report SPECIAL OPERATING CONDITIONS, These lost articles were identifie during an inspection of the Unit 3 annulus region conducted on October 16, 1995, but could not be recovered. The lost articles are all lightweight, as evidenced by the fact that they were swept into the annulus region from the vessel lower plenum by the relatively low flow velocities generated within the vessel by the Residual Heat Removal System in the shutdown cooling mode. At the time these articles were identified, the Unit 3 core was off-loaded.

The missing plastic/tape pieces are of insuflicient mechanical strength to cause a recirculation pump seizure. For a rod drop accident to occur, the control blade must become decoupled Rom the control rod drive. However, there is no credible mechanism by which the plastic pieces described in this safety evaluation could cause the blade to uncouple from the drive.

Lost Article Report 3-9542 This safety evaluation addresses BFN Unit 3 lost article described on Lost Article Report Number 3-9542. The lost article is a 1" by 2" piece of paper which was last spotted floating in the annulus of the Unit 3 reactor vessel. This lost article was identified during the Unit 3 core loading on October 23, 1995, but could not be recovered. At the time this article was identified, the Unit 3 core was partially loaded.

This piece of paper is not expected to be strong enough to possibly interfere with control rod drive operation prior to if vessel heatup, even it were to migrate into the internals of a control rod drive. If this lost article did migrate into a control rod guide tube and into the annulus opening, it would be trapped in the control rod drive screen filters and

~ would not be able to move into the drive internals. Afier heatup to rated temperatures, the lost piece of paper is expected to decompose. No other equipment important to safety is affected by this lost article.

This change is safe from a nuclear safety standpoint. No unreviewed safety question exists.

Lost Article Report 3-95A6 This safety evaluation addresses BFN Unit 3 lost article described on Lost Article Report Number 3-95M. The lost article is a 1" by 1/2" piece of tape which disappeared from sight at the upper edge of the annulus in the northeast quadrant of the Unit 3 reactor vessel. This lost article was identified during the Unit 3 core loading on October 23, 1995, but could not be recovered. At the time this article was identified, the Unit 3 core was partially loaded.

A piece of tape may be strong enough to possibly interfere with control rod drive operation prior to vessel heatup ifit were to migrate into the internals of a control rod drive. A lost article could migrate into a control rod guide tube and into the annulus opening, but it would be trapped in the control rod drive screen filters and would not be able to move into the drive internals. After heatup to rated temperatures, the lost tape piece is expected to decompose. No other equipment important to safety is affected by this lost article.

This change is safe from a nuclear safety standpoint. No unreviewed safety question exists.

Lost Article Reports 3-95-12 and 3-95-15 This safety evaluation addresses BFN Unit 3 Lost Article Report Numbers 3-95412 and 3-95415. The lost article for report number 3-95412 is two fragments from a plastic diy tube capture device on the GE dry tube removal tool.

The fragments are about 1/4" wide and shaped in a half circle that is about 0.125" thick and about 1" outside diameter. The plastic parts broke offand were lost in the Unit 3 reactor cavity while clamping onto intermediate range monitor A, and were lost at core location IS-52 on May 15, 1995.

Tennessee Valley A uthori ty

SUMMARY

OF Brains Ferry Nuclear Plant SAFETY EVALUATION FOR 1995 Annual Operating Re ort

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SPECIAL OPERATING CONDITIONS

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'he lost article described in report number 3-95415 is the top portion of another plastic capture device Rom the GE dry tube removal tool of roughly the same dimensions. This lost article was spotted on the fuel support casting at core location 33-28 on May 16, 1995, but could not be recovered.

All lost pieces described are made of TC-806 A/B urethane. At the time these pieces were lost, the Unit 3 core was oQ-'loaded.

The missing urethane pieces are of insuQicient mechanical strength to cause a recirculation pump seizure. For a rod drop accident to occur, the control blade must become decoupled from the control rod drive. However, here is no credible mechanism by which the plastic pieces described in this safety evaluation could cause the blade to uncouple from the drive.

A plastic piece may be strong enough to possibly interfere with control rod drive operation prior to vessel heatup ifit were to migrate into the internals of a control rod drive. The lost article could migrate into a control rod guide tube and into the annulus opening, but it would be trapped in the control rod drive screen filters and would not be able to move into the drive internals. After heatup to rated temperatures, the lost plastic pieces are expected to decompose.

Removal of Alpha Subtraction Channel for the Radiation Control (RADCON) Continuous AirMonitors (CAMs)

This safety evaluation was written to address non-use of Channel 2 as a background subtraction channel for the RADCON CAMs. Channel 2 is intended to subtract out any alpha contribution to the readings of Channel 1 (beta particulate). This function of the CAMs is not beneficial at low levels of alpha detection. The scope of work required to remove the subtraction channel was simply to edit the CAN parameters at the control console in the applicable control room. This was accomplished via sofbvare driven menus and a keylock switch that is controlled by Operations. Overall function of the CAMs and the desired information provided to the plant did not change as a result of not using the alpha background subtraction.

The alpha channel had been a source of numerous problems. Not using the alpha channel greatly improved the RADCON CAMs'reliability. There was no adverse impact on equipment reliability from not using alpha subtraction.

The RACDON CAMs are non-safety related and perform no safety function.

Temporary Configuration for Unit 3 Residual Heat Removal Service Water (fu&SW) Pipe Replacement This safety evaluation evaluated the temporary configuration which was entered for the purpose of replacing the Unit 3 RHRSW, System 23, piping inside the Reactor Building. The temporary configuration aQected the RHRSW piping as well as the secondary containment, System 64C, boundary. The temporary configuration was accomplished through the use of blanks installed in the RHRSW supply and discharge lines in the tunnels on the south side of the Reactor Building. At the completion of the activity, the system was returned to its as4esigned configuration.

The replacement activity was performed in general as a "like for like" replacement performed under work requests/work orders. Replacement activities did not require a 10CFR50.59 evaluation. However, since the plant was temporarily in a configuration difierent than as described in the UFSAR, a safety evaluation was prepared in order to ensure the plant was maintained in a safe configuration and that no unreviewed safety questions were created.

Tennessee Valley Authority Brogans 'UMlfARYOF Feny Nuclear Plant =

SAFETY EVALUATIONFOR

'= SPECIAL OPERATINGCONDITIONS l995AnnualOperatingReport

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Both the RHRSW and secondary containment continued to function as designed for Unit 2 operation. The accident mitigation capability of Unit 2 was unaffected by the piping replacement process; therefore, there was no effect on the margin of safety for Unit 2.

Tennessee Valley Authority

SUMMARY

OF Browns Ferry Nuclear Plant SAFETY EVALUATIONSFOR 1995 Annual Operating Report SPECIAL TESTS 1995 SUM Y OF SAFETY EVALUATIONS FOR SPECIAL TESTS

Tennessee Valley Authority SUMK4RYOF Browns Ferry Nuclear Plant SAFETY EVALUATIONS FOR l995 Annual Operating Report SPECIAL TESTS 6'~x'&%~@" '5%8": "" ' '%C" '@'"::55%'5 "s'":+"': """ 3$ 'u' " '<<'k@kPNCC~S PkwSS>

'ost Modification Test (PMT) 3-PMT-BF%67.042 This test established the minimum Emergency Equipment Cooling Water header pressure required'to provide adequate cooling water flow to Unit 2/3 safe shutdown equipment. This test also verified that adequate cooling was provided to the Unit 3 Reactor Building Closed Cooling Water heat exchangers during an Appendix R Qre, while maintaining the minimum required flow to safe shutdown equipment.

This test was performed without adverse impact on nuclear safety and without a change to the Fire Protection Report or Technical Specifications. However, since the AUTO CLOSE function of 347-50/51 was removed from service during the PMT, a safety evaluation was required to address this change in the facility as described in the UFSAR.

No unreviewed safety question exists.

Tennessee Valley Authority SUNMARY OF Browns Ferry nuclear Plant SAFETY EVALUATIONSFOR 1995 Annual Operating Re ort TEMPORARYAI.TERATIONS

'v~~~W 1995 SUM Y OF SAFETY EVALUATIONS FOR TEMPO Y ALTERATIONS Tennessee Valley Authority SMKIIARYOF Brogans Feny Nuclear Plant SAFETY'EVALUA T101VS FOR 1995 Annual Operating Report TEMPORARYALTERATIONS Temporary Alteration Control Form 2-9541-242 This temporary alteration removed the generator and turbine trip initiations which are generated by the operation of the 64GF generator field ground relay-and added an additional alarm to the switchyard annunciation alarm panel, 0-XA%554010. This was accomplished by performing wirelifis at the 64GF auxiliary relays, 64GFX and 64GFXX, and installing a jumper from Bay 91 to Bay 92 of the communications input terminals.

I The worst case failure associated with this alteration is a generator failure which is bounded by the generator trip transient.

This change is safe from a nuclear safety standpoint. No unreviewed safety question exists.

Tennessee Valley Authority SUMllfARFOF .

Brains Feny Nuclear Plant ~ SAFETF EVALUA TIONS FOR 1995 Annual Operating Rep~ort " TECHNICALSPECIFICATION CHANGES 1995 SUM Y OF SAFETY EVALUATIONS FOR TECHNICAL SPECIFICATION C GKS 40-

Tennessee Valley Authority SUMKARYOF Browns Feny Nuclear Plant SAFETY EVALUATIONS FOR l995 Annual Operating P:-:= ':=-""'-'": '::::- .<--'@:""Ressort

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TECHNICALSPECIFICATION CHANGES TS 9&44 This change involved changes to Technical Speci6cation Bases 3.7/4.7 to allow an alternative method to be used to, demonstrate drywell to suppression chamber vacuum breaker closure. This change eliminated entering a 24-hour cold shutdown limiting condition for operation solely on the basis of faBed indicating lights when no evidence of failure of any vacuum breaker to fullyclose exists.

This change does not affect the manner in which the vacuum breakers are verified to perform their safety function (OPEN) or change the frequency for performing cycling of the'vacuum breakers.

This change to Technical Specification Bases is in compliance with all applicable design criteria to ensure that the function and operation of primary containment and associated safety related systems are not adversely affected. The changes affect only the method used to demonstrate closure of the vacuum breakers by allowing an alternative method in the case when a green check light became inoperable. The method for determining vacuum breaker opening remains unchanged..The proposed alter'nate method of demonstrating vacuum breaker closure is similar to and consistent with the methodology used to verify drywell to suppression chamber leakage is within allowable limits as specified by Technical Specification 4.7.A.4.d and is conservative in that not only vacuum breaker leakage, but also bypass leakage willbe monitored using the alternative method.

41-

Tennessee Valley Authority SUMllfARYOF Browns Feny Nuclear Plant SAFETY EVAI.UATIONS FOR Annual Operating Report UFSAR REVISIONS '995 1995 SUM Y OF SAFETY EVALUATIONS FOR UFSAR REVISIONS 42-

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Tennessee Valley Authority SUML44RY OF Browns Ferry Nuclear Plant 1995 Annual Operating Report

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SAFETY EVALUATIONS FOR LUNARREVISIONS Section 10.10, Emergency Equipment Cooling Water (EECW) System This update of the UFSAR (CRLD-BFEP-MN-94065RO) involved clarification of Section 10.10. The statement in regard to maximum required design flow was not consistent with current calculations. The following statement in the UFSAR was revised: "The maximum required design EECW flow for the three-unit plant is approximately 9800 gpm including 2900 gpm for cooling non~ntial components (i.e., the RBCCW heat exchangers and Control Air compressor intercoolers and aftercoolers)." The statement was revised to read: "The maximum required design EECW flow for the three unit plant is satisfied by EECW pumps rated at 400-hp with a capacity of 4500 gpm at a 275-foot head. This includes cooling non~ntial components (i.e., the RBCCW heat exchangers and Control Air compressor intercoolers and aAercoolers)."

No equipment was added, modified, or removed from BFN as a result of this UFSAR change., Revising the UFSAR section will not result in any new or different design input. However, the change willprovide clarifications and make the UFSAR consistent with other design documents and site procedures.

This change is safe from a nuclear safety'standpoint. No unreviewed safety question exists.

Section 7.3.4.8.4, Primary Containment Isolation System Instrumentation This safety evaluation was issued to evaluate changes to UFSAR Section 7.3.4.8.4 to correct the UFSAR text which addresses main steam line high flow. UFSAR Section 7.3.4.8.4 states that high flow in each main steam line is sensed by four indicating type differential pressure transmitters. These trattsmitters do not have indication. The UFSAR also states that an alarm is actuated whenever the difference between the steam flow in any two steam lines exceeds a preset amount. In actuality, a trip occurs (main steam isolation valve (MSIV) and main steam line drain valve closure) whenever the steam flow in any main steam line exceeds a preset amount. This UFSAR text change is consistent with the Main Steam System Design Criteria, the Safe Shutdown Analysis Calculation, and the Setpoint and Scaling Calculation for the differential pressure transmitters.

The safety related function of the main steam line instrumentation is to provide signals for MSIV and main steam drain valve closure. The documentation change does not change the safety related function of these loops. This change does not change any setpoints associated with these loops.

This change is safe from a nuclear safety standpoint. No unreviewed safety question exists.

Subsection 4.4 (Nuclear Steam Pressure Relief System), Subsection 7.4.3.3 (ADS Instrumentation and Controls), Subsection 14.0 (Plant Safety Analysis), Table 7.4-2 (ADS Instrumentation)

This UFSAR change corrects erroneous information with respect to purpose, function, and description of the main steam relief valve (MSRV) Discharge Tailpipe Temperature Recorder, TR-1-1, and the thermocouples installed in the MSRV discharge tailpipes. Allof the changes were documentation changes only. Incorporation of the changes-did not require field work to implement.

This change is safe from a nuclear safety standpoint. No unreviewed safety question exists.

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Tennessee Valley A uthori ty Brains Feny Nuclear Plant 1995 Annual Operating Report RELEASE

SUMMARY

1995 RELEASE SUM

'8 1995 RELEASE

SUMMARY

ANNUALOPERATING REPORT GASEOUS RELEASES LIQUID RELEASES FISSIONS 8 IODINES PARTICULATES TRITIUM FISSIONS 8 TRITIUM DISSOLVED GROSS ACTIVATION >8 DAY HALF- ACTIVATION NOBLE GASES ALPHA MONTH PRODUCTS (CI) (CI) LIVES (CI) (CI) PRODUCTS (Cl) (CI) (CI) (CI)

JANUARY 1.07E+00 1.33E-05 3:32E-05 6.69E-01 2.86E-02 6.06E-01 1.09E-04 ND FEBRUARY 1.28E+00. 1.98E-05 2.58E-05 6.76E-01 1.95E-02 6.73E-01, 1.59E-04 ND MARCH 9.66E-01 3.01E-05 6.41E-05 5.81E-01 1.85E-02 8.04E-01 9.26E-06 ND APRIL 6.19E-01 2.39E-05 3.60E-05 9.34E-01 3.98E-02 8.92E-01 7.95E-05 ND MAY 2.84E-01 2.16E-05 2.57E-05 7.19E-01 2.95E-02 6.29E-01 1.72E-05 ND JUNE 8.89E-02 3.24E-05 5.32E-06 7.87E-01 6.90E-02 8.41E-01 5.60E-04 ND JULY 3.24E-01 4.72E-05 6.50E-06 8.18E-01 7.41E-02 1.86E+00 6.10E-'04 ND AUGUST 1.37E+00 7.88E-05 5.45E-06 1.19E+00 1.18E-01 1.68E+00 7.17E-04 ND SEPTEMBER 2.92E+00 3.95E-05 9.02E-06 1.16E+00 1.61E-01 2.38E+00 1.86E-03 ND OCTOBER 5.08E-01 4.19E-05 5.18E-05 1.09E+00 1.41E-01 1.60E+00 9.84E-04 ND NOVEMBER 8.28E+00 3.60E-05 5.71E-05 6.58E-01 9.80E-02 2.97E+00 2.86E-03 ND DECEMBER 6.33E+00 2.68E-04 1.23E-04 1.09E+00 1.46E-01 1.42E+00 1.11E-03 ~

ND ND is for non-detectable Variation in the data for gaseous releases have been correlated with the numbers of operating fans. There were no excursion of interest nor releases which exceeded Tech Spec limits.

65

Tennessee Valley Authority Browns Ferry1Vuclear Plant 1995 Annual Operating Report OCCUPATIONALEXPOSURE DATA 1995 OCCUPATIONAL KXPOSURK-DATA

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REXPR219~ TENNESSEE VALL AUTHORITY PAGE: 5 RUfl DATE: 01-10-96 BFN RADIATION EXPOSURE SYSTEM "UN II'>E: 14:04:46 IIUMBER OF PERSONNEL AND tiIAN-REtil BY WORK JOB FUNCTION TOTAL NUIIBER OF INDIVIDUALS NUtiIBER OF PERSONNEL (> 100 MILLIREII) TOTAL MAN-REM f10=REACTOR OPS SURVEILLANCE GROUP STATION UTILITY CONTRACT TOTAL STATION UTILITY CONTRACT TOTAL EMPLOYEES EIIPLOYEES AND OTHERS PERSONS EMPLOYEES EMPLOYEES AND OTHERS - tilAN- REI.I MAINTENANCE PERSONNEL 133 21 304 458 11.843 2.532 6.528 20.903 OPERATING PERSONNEL 108 3 2 113 21.349 0.480 0.000 21.829 HEALTH PHYSICS PERSONNEL 54 6 1 61 10 '49 0.579 0.000 11.228 SUPERVISORY PERSONNEL 30 0 51 81 4.277 0.000 2.037 6.314 ENGINEERING PERSONNEL 23 1 40 64 1.989 0.008 1.443 3.440 lilO T~ml 63.7%4 I

MD= ROUT NE f IAI NTE NANCE GROUP STATION UTILITY CONIRACT TOTAL STATION UTILITY CONTRACT TOTAL EMPLOYEES EflPLOYEES AND OIIIERS PERSONS EllPLOYEES Efilf'LOYEES AND -OI HERS f,lhll- RFt I IllINTENANCE PERSONNEL 159 25 680 864 27 '83 2.502 111 579 141 . 764

'99 '33

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OPERATIIIG PERSONNEL 96 3 9 108 6 0.028 2 9.660 HEALTH PHYSICS PERSONNEL 56 6 1 63 6.316 0.646 0 '01 7.163 SUPERVISORY PERSONNEL 22 2 69 93 1,. 285 0 219

~ 8.723 10.227 EtfGINEERING PERSONNEL 24 4 44 72 1. 'I94

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0,294 2.945 4 '33 I,'10 HU3 TN~U 73.777 ~tiH9 T26.08T f10=IN-SERVICE INSPECTION GROUP STATION UTILITY CONTRACT TOTAL STATION UTILITY CONTRACT TOTAL EMPLOYEES EMPLOYEES AND OTHERS PERSONS EIIPLOYEES El'IPLOYEES AND OTHERS IIAtl-REII flhItfTENANCE PERSONNEL 0.000 0 000 F 0. 041 0.041 OPERATING PERSONNEL HEALTH PHYSICS PERSONtfEL SUPERVISORY PFRSONNEL EtfGINF ERI tfG PERSONtfEL MO f10=SPECIAL IIAINTENANCE GROUP UTILITY CONTRACT TOTAL STATION , UTILITY CONTRACT TOTAL EIIPLOYEES EtilPLOYEES AND O'IHERS PERSONS EMPLOYEES EMPLOYEES AND OTHERS LIAN-REfl PERSONNEL'TATION flAINTENANCE 136 24 738 898 11.464 3.311 127 '39 142.014 67

8 REXPR219

.'LUtl DATE: i- 0-96 TEN tJES SEE VALL BFN RADIATION EXPOSUR AUTHOR I TY PAGE:

SYSTELI

".! fW TI!.IE: 14:04:46 NULIBER OF PERSONNEL AND MAN-REM BY )YORK JOB FUNCTION TOTAL NULIBER OF INDIVIDUALS t!UflBER OF PERSONNEL (> 100 LIILLIRELI) TOTAL L'IAN-RELI OPERATING PERSONNEL 56 2 3 61 1.255 0.104 0.668 2.027 IIEALTH PHYSICS PERSONNEL 53 6 1 60 5.360 1.043 0.021 6.424 SUPFRVISORY PERSONtJEL 9 ~ 2 71 82 0.242 0.058 8.092 8 392

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'JGIWEERIWG PERSONNEL 15 1 46 62 0.950 0.000 5.568 6.518 LIO 8 TB5.375

!.IO=IVASTE PROCES I WG GROUP STATION UTILITY CONTRACT TOTAL STATION UTILITY CONTRACT TOTAL EMPLOYEES ELIPLOYEES AND OTHERS PERSONS ELIPLOYEES EMPLOYEES AWD OTHERS fiIAN-REt.l 1lAItJTENANCE PERSONtlEL OFERATING PERSONNEL 18 10 13 1.

32 0.279 0.754 0.026 0.000 0.057 0 '62 HEALTH PHYSICS PERSONNEL 0' 11 0.282 1.036 10 10 0.136 0.000 0.000 0. 136 SUPERVISORY PERSOtJNEL 3 3 0.060 0.000 0.000 0.060 i.fJGIifEERIWG PERSO"ft<EL 0 0 0 0.000 0.000 0.000 0.000 LIO tlO=REFUEL GROUP STATION UTILITY CONTRACT.: TOTAL STATION UTILITY CONTRACT TOTAL ELIPLOYEES EflPLOYEES AND OTHERS PERSONS EMPLOYEES ELIPLOYEES AND OTHERS LIAN-REt'I LIAINTENANCE PERSONNEL 0 7 0.000 0.000 0.011 0.011 OPERATING PERSONNEL 13 13 0.074 0.000 0.000 0.074 HEALTH PHYSICS PERSONNEL 0.002 0.000 0.000 SUPERVISORY PERSONNEL 1 1 0.002 ENGIWEERItJG PERSONNEL 1 2 0.001 0.000 0.015, 0.016 0 0 0.000 0.000 0.000 0.000 MO T5 0. 077 0. 000 1030 107 2085 3222 114.461 11.830 278 '83 404 .374 68

REXPR219 i- 0-96 TENNESSEE VALL AUTHORITY PAGE.

RUN DATE: BFN RADIATION EXPOSUR SYSTEM ilUN TIIIE'4:04'46 NUMBER OF PERSONNEL AND MAN-REM BY IYORK JOB FUNCTION TOTAL NUIIBER OF INDIVIDUALS NUMBER OF PERSONNEL (> 100 IJIILLIREII) TOTAL MAN-REMI GROJP STATION UTILITY CONTRACT TOTAL STATION UTILITY CONTRACT TOTAL EMPLOYEES EIIIPLOYEES AND OTHERS PERSONS EMPLOYEES EMPLOYEES AND OTHERS I IAN - RE I I IIAINTENANCE PERSONNEL 446 71 1745 2262 51. 269 8.371 245.455 305.095 OPFRATING PERSONNEL 283 8 15 306 30.431 0.612 3.583 34 .626 HEALTH PHYSICS PERSONNEL 174 18 3 195 22.463 2.268 0.222 24.953 SUPERVISORY PERSONNEL 65 4 192 261 5;865 0.277 18.867 25.009 ENGINEERING PERSONNEL 62 6 130 198 4.433 0.302 9.956 14.691 1030 107 2085 3222 114.461 11.830 278.083 404.374 69

REXPR219 RUN DATE: 01-10-96 TENNESSEE VALL>AUTHORITY BFN RADIATION EXPOSURE SYSTEM PAGE:

RUN TIf,'IE: 14:04:46 NUMBER OF PERSONNEL AND LIAN-REM BY WORK JOB FUNCTION TOTAL NUMBER OF INDIVIDUALS STATION UTILITY I'62 GROUP CONTRACT TOTAL MAINTENANCE PERSONNEL 26 728 916 OPERATING PERSONNEL 113 2 11 126 HEALTH PHYSICS PERSONNEL 54 6 1 61 SUPERVISORY PERSONNEL 30 0 72 102 ENGIt<EERItlG PERSONNEL 24 4 47 75 383 38 859 1280 70

Tennessee Valley Authority Brains Feny Nuclear Plant CHALLENGES TO OR FAILURES OF 1995 Annual Operating Report MAINSTEAM RELIEF VALVES K"" ",'""'"""::'""""-""-'- " ".""':".":-""";::.""""'""""

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1995 CHALLENGES TO OR FAILURES OF MAINSTEAM RELIEF VALVES L

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Tennessee Valley Authority Brogans Feny Nuclear Plant CHALLEiVGES TO OR FAILURES OF 1995 Annual Operating Report K:.'4R-:"".~Y::: ';-".'- m": '::"."-'."":

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MAINSTEAM RELIEF VALVES UNIT 1 None UNIT 2 All-thirteen Unit 2 MSRVs were operable, during the entire reporting period.

Based upon discharge tailpipe temperatures, six MSRVs were identified as having leaking pilot valves during the reporting period. Three of the leaking pilot valves were replaced during the late-March reactor scram and recovery. The replaced valves were manually cycled during the reactor restart and power ascension. In addition to the pilot valve leakage, one MSRV was identified as having main seat leakage. Based upon Suppression Pool bulk temperature trends, the proximity of the tailpipe thermocouples to the MSRV discharge Qange, and the absences of any acoustic monitor indications, all MSRV leakage has been estimated to be minor with no impact to valve operability.

During the reporting period, two challenges to the MSRVs occurred. The challenges occurred during the reactor scrams on February 9, 1995 and August 19, 1995. Prior to reactor restart, plant process computer data were reviewed and MSRV performance was evaluated. During both scrams, all nine Main Steam Turbine Bypass valves cycled open and the Main Steam Isolation Valves stayed open. The logic for a Primary Containment Isolation System (PCIS) - Group I was never satisfied. Therefore, reactor vessel over pressurization protection was provided by both the MSRVs and the Turbine Bypass Valves. The availability of the Turbine Bypass Valves helped to soften the reactor vessel pressure transient and limit the duration.

During the February 1995 scram, peak reactor vessel pressure indication was about 1130 psig which is 5 psig above the highest MSRV setpoint. A review of the plant process computer data indicated that all thirteen MSRVs lifted within allowable specifications of the lifting setpoint.

Before the event, two MSRVs were exhibiting pilot valve leakage. Post-scram discharge tailpipe temperatures indicated three leaking pilot valves. Later in the operating cycle, the three leaking pilot valves were replaced.

During the August 1995 scram, peak reactor vessel pressure indication was about 1110 psig which is about 5 psig above the lowest nominal setpoint. During review of the plant process data, there no indication of any MSRV openings. Wyle MSRV setpoint certifications, reactor vessel pressure instrumentation calibration records, and plant process computer data were reviewed.

The review concluded that lack of MSRV actuation was reasonable and expected based upon reactor vessel pressure instrument accuracies, line losses, actual MSRV setpoints (pre-startup certifications), and potential MSRV setpoint drift (reference BFN PER 951076). This position was based upon the observed peak reactor pressure (1110 psig) did not exceed the allowable Technical Specification limit of 1116 psig (1105+ 11 psig) when all of the uncertainties were reviewed.

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Tennessee Valley Authority Browns Feny ¹clear Plant I99$ Annaal 0 eraerng Revere REACTOR VESSEL FATIGUE USAGE EVALUATION

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1995 REACTORVKSSKL FATIGUE USAGE EVALUATION

Tennessee Valley Authority Browns Ferry Nuclear Plant 1995 Annual Operating Report REACTOR VESSEL FATIGUE USAGE EVALUATION The cumulative usage factors for the reactor vessels are as follows:

Location Unit 1 Unit 2 Unit 3 Shell at water line 0.00620 0.00582 0.00438 Feedwater nozzles 0.29782 0.23065 0.16268 Closure studs 0.24204 0.22966 0.15708

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