ML18037A772

Annual Operating Rept Jan-Dec 1993

ML18037A772
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Site:	Browns Ferry
Issue date:	12/31/1993
From:	TENNESSEE VALLEY AUTHORITY
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TENNESSEE VALLEYAUTHORITY BROWNS. FERRY NUCLEARPLANT ANNUALOPERATING REPORT January 1, 1993 - December 31, 1993 Docket Number 50-259, 50-260, and 50-296 License Number DPR-33, DPR-52, and DPR-68 94030800S7

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Tennessee ValleyAulhoriiy Browns Ferry Nuclear Plant l993 Annual Operating Reporl TABLEOF CONTENTS

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Acronyms Listing.

Regulatory Guirle 1.16,Section I.b. (I) anrl (2)

Operational Summary.

IOCFR59. 59(b) (2) - Suni nmry ofSafety Evaluations Core Component and Operating Limits.

10 Field Completed Plant Modifications.

15 New Instructions/Procedure Revisions..

Special Operating Conditions.

53 58 Special Tests.

4 Updated Final Safety Analysis Report Revisions.

Regulatory Guirie 1.16,Section I.b. (3) 62 1993 Release Summary 81 Technical Specification 6.9.1.2 1993 Occupational Exposure Data.

Challenges to or Failures ofMain Steam Relief Valves.

Technical Specification 6.9.2.1

~ 83 88 Reactor Vessel Fatigue Usage Evaluation.

90

Tennessee ValleyAuthori ty Brains Ferry Nuclear Plant 1993Annnal 0 eralin Re orl

. ACROIVYMSI.ISTING L":::*:-:."*:::--'::::-::::--:::-::i:-':: :o: -:':-:::::::-':-::::::::-" -::-'--::::::::-':i:":-:":":-:-'::-'":::rl::'--::-:::-'::":.--::::-"-::.:-:.':.:-*-'::-:-":-::"-:-"::-:::

'::::--i This is a list ofacronyms and abbreviations used throughout the 1993 Annual Operating Report.

AC AFFF ANS ANSI AP ASME BFN BPWS BWR CAD CAQR CBHZ CCDCR CCW CFM CFR CIV CO2 Cr CRD CRDR CREV DBA DBE DC DCD DCN DG DRD ECN EDG EECW EMI F

FCO FCV FDCN Fe FPC Alternating Current Aqueous Film Forming Foam American Nuclear Society American National Standards Institute Alkaline Per manganate American Society ofMechanical Engineers Browns Ferry Nuclear Plant Banked Position Withdrawal Sequence BoilingWater Reactor Containment Atmosphere Dilution Condition Adverse To Quality Report Control Bay Habitability Zone Core Component Design Change Request Condenser Circulating Water Cubic Feet per Minute Code ofFederal Regulations Combined Intercept Valve Carbon Dioxide Chromium Control Rod Drive Control Room Design Review Control Room Emergency Ventilation Design Basis Accident Design Basis Earthquake Direct Current Dilute Chemical Decontamination Design Change Notice Diesel Generator Direct Reading Dosimeter Engineering Change Notice Emergency Diesel Generator Emergency Equipment Cooling Water Electromagnetic Interference Fahrenheit Flow Control Operator Flow Control Valve Field Design Change Notice Iron (Chemical)

Fuel Pool Cooling

Tennessee ValleyAurhoriry Brogans Ferry Nuclear Planr 1993Annual 0 cratingReport, ACRONYhISLISTING FQ FR-FSV FT GE GEMAC GSC HELB

'HEPA HPCI HPFP HS HVAC HWWV ICS IEEE IRM IST JB kv KW lbs LCO LIC LOMI LOP LPRM

,LT LTTIP MG MOV MSIV MSRV MWD/ST N2 NFPA NG NQA NRC NUREG PCIOMR PCIS Flow Totalizer Flow Recorder Flow Solenoid Valve Flow Transmitter General Electric General Electric Measurement and Control Gland Seal Condenser High-Energy Line Break High-EQiciency Particulate Air(Filter)

High Pressure Coolant Injection High Pressure Fire Protection Hand switch Heating, Ventilation, and AirConditioning Hardened Wetwell Vent Integrated Computer System Institute ofElectrical and Electronic Engineers Intermediate Range Monitor Image Sensing Technology Junction Box Kilovolt Kilowatt Pounds LimitingCondition for Operation Level Indicating Controller Low Oxidation State Metal Ion Loss ofOffsite Power Local Power Range Monitor Level Transmitter Long Term Torus Improvement Plan Motor Generator Motor Operated Valve Main Steam Isolation Valve Main Steam Relief Valve Megawatt Days per Short Ton Nitrogen National Fire Protection Association Nuclear Grade Nuclear Quality Assurance Nuclear Regulatory Commission Nuclear Regulatory Commission Regulation Preconditioning Interim Operating Management Recommendations Primary Containment Isolation System Tennessee ValleyAuthority Brogans Ferry Nuclear Plant l993 Annual Operating Report ACRONYMSLISTING 1$$%K::-:'::-:::::-:"::::::

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'Rx Y~4q ':%:..'( gl~Q; PCV PMT PSIG RBCCW RCIC RCW'FW RHRSW RMOV RMS RPM RPS RPT RPV RSCS RWCU RWM SCFM SDSP SGT SGTS SHV SJAE SLMCPR SOI SOS SPDS SRD SSP TI TLD TS TVA UFSAR UNID UPS USQD V

VAC VDC VTV Pressure Control Valve Post Modification,Test Pounds per Square Inch Gauge Reactor Building Closed Cooling Water Reactor Core Isolation Cooling Raw Cooling Water Reactor Feedwater Residual Heat Removal Service Water Reactor Motor Operated Valve Radiation Monitoring System Revolutions per Minute Reactor Protection System Recirculation Pump Trip Reactor Pressure Vessel Rod Sequence Control System Reactor Water Cleanup Rod Worth Minimizer Standard Cubic Feet per Minute Site Director Standard Practice Standby Gas Treatment Standby Gas Treatment System Shutoff Valve Steam Jet AirEjector Safety LimitMinimum Critical Power Ratio Special Operating Instruction Shift Operations Supervisor Safety Parameter Display System Self Reading Dosimeter Site Standard Practice Technical Instruction Thermoluminescent Dosimeter Technical Specification Tennessee Valley Authority Updated Final Safety Analysis Report Unique Identification Uninterruptible Power Supply Unreviewed Safety Question Determination Volt; Vanadium Volts Alternating Current Volts Direct Current Vent Valve

Tennessee ValleyAuthority Browns Ferry Nuclear Plant 1993 Annual Operating~Re ort OPERA TIONAL

SUMMARY

1993 ll OPERATIONAL

SUMMARY

A

Tennessee ValleyAuthori ty Brains Ferry Nuclear Plant l993 Annual Operating Report OPERA TIONAL

SUMMARY

t

~~~~"

. UNIT1 Unit 1 remains on administrative hold to resolve various Tennessee Valley Authority (TVA) and Nuclear Regulatory Commission (NRC) concerns.

UNIT2 On January 1, 1993, the unit's power level was 72% ofrated power. Reactor coast down began on October 1, 1992, with reactor shutdown starting January 29, 1993, for completion offuel Cycle 6 and the beginning ofthe refueling outage.

Fuel loading for Cycle 7 began on April25, 1993, and was completed on May 3, 1993. Initial criticalitywas achieved at 1342 on May 25, 1993. A problem with overlap between intermediate range monitor (IRM) ranges 6 and 7 (combined with other complicating factors) resulted in the reactor being manually scrammed at 1936 hours0.0224 days <br />0.538 hours <br />0.0032 weeks <br />7.36648e-4 months <br /> on the same day.

Control rod withdrawal for scram recovery began at 1044 hours0.0121 days <br />0.29 hours <br />0.00173 weeks <br />3.97242e-4 months <br /> on May 26 and the reactor was critical by 1248 hours0.0144 days <br />0.347 hours <br />0.00206 weeks <br />4.74864e-4 months <br />.

On May 27, with reactor in run mode, a cold shutdown was initiated due to recirculating pump 2A high vibration and seal leakage.

(A notification ofan unusual event was made to NRC.)

AAer replacement ofrecirculation pump seals, plant restart commenced and the reactor was again critical at 1606 hours0.0186 days <br />0.446 hours <br />0.00266 weeks <br />6.11083e-4 months <br /> on May 31.

Reactor power was decreased and the main generator breaker was opened at 2244 hours0.026 days <br />0.623 hours <br />0.00371 weeks <br />8.53842e-4 months <br /> on June 2 to allow turbine overspeed testing and to allow turbine balancing work.'n support of this work and after reactor power was decreased to 6% ofrated output, the reactor was manually scrammed at 0236 hours0.00273 days <br />0.0656 hours <br />3.902116e-4 weeks <br />8.9798e-5 months <br /> on June 3. Rod pulls for reactor restart began at 0455 hours0.00527 days <br />0.126 hours <br />7.523148e-4 weeks <br />1.731275e-4 months <br /> the same day and the reactor was again critical by 0930 hours0.0108 days <br />0.258 hours <br />0.00154 weeks <br />3.53865e-4 months <br />.

On June 16, power was reduced to 69% (2281 MWt and 720MWe) for power ascension testing. Power was again reduced on August 29 and September 17 for performance of scheduled maintenance (condenser water box, reactor feed pump, high pressure heaters) and rod adjustment.

There also was a power reduction on November 19 to replace 2C hotwell pump oil cooler.

Following the refueling outage, Unit 2 operated for 211 days continuously with no major problems.

Tennessee ValleyAuthority Browns Ferry Nuclear P/ant 1993Annual Operating Report OPERATIONAL

SUMMARY

UNIT3 Unit 3 remains on administrative hold to resolve various TVAand NRC concerns.

P.

Tennessee ValleyAuthority Brogans Ferry Nuclear Plant 1993 Annual Operating Report OPERATIONAL

SUMMARY

Docket No.: 50-259 OPERATING STATUS l.

UnitName: Browns Ferry Unit One 2.

Reporting Period:

Calendar Year 1993 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 12.

13 14.

Hours in Re ortin Period Hours Reactor Was Critical Reactor Reserve Shutdown Hours Hours Generator On Line December 1993 0

0 0

0 Year to Date 0

0 0

0 Cumulative*

95743 59521 6997 58267 15.

'16.

17 18.

Unit Reserve Shutdown Hours Gross Thermal Generation h

Gross Electrical Generation h

Net Electrical Generation Wh

-0 0

0 0

0 0

0 168066787 0

55398130 0

53796427 19 20.

21 22 23.

Unit Service Factor Unit Availabilit Factor Unit Ca acit Factor MDCNet Unit Ca acit Factor ERNet Unit Forced Outa e Rate 0

0 0

0 0

0 0

0 0

0 60.9 60.9 52.8 52.8 25.6 24.

Shutdowns Scheduled Over Next 6 Months (Type, Date, and Duration ofEach):

N/A

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

TO BE DETERMINED

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

Tennessee ValleyAuthority Browns Ferry Nuclear Plant 3993AnnnnlOpernnnERepnre OPERATiONALSO33MARY "i%i':"""" ""'~. "'-""-"'":re'n:. '":none'

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ere Docket No.: 50-260 OPERATING STATUS 1.

Unit Name: Browns Ferry Unit Two 2.

Reporting Period:

Calendar Year 1993 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 Dccembcr 1993 Year to Date Cumulative*

11.

Hours in Re ortin Period 744 8760 113311 12.

Hours Reactor Was Critical 13.

Reactor Reserve Shutdown Hours 14.

Hours Generator On Line 744 0

744 5854 0

5754 74817 14200 72625 15.

Unit Reserve Shutdown Hours 16.

Gross Thermal Generation h

17.

Gross Electrical Generation h

18.

Net Electrical Generation h

0 2445505 830240 811267 0

0 17818949 208547601 5930270 69207918 5776842 67244662 19.

Unit Service Factor 20.

Unit Availabilit Factor 21.

Unit Ca acit Factor C Net 22.

Unit Ca acit Factor ERNet 23.

Unit Forced Outa e Rate 100.0 100.0 102.4 102.4 0

65.7 65.7 61.9 61.9 0

64.1 64.1 55.7 55.7 18.5 24.

Shutdowns Scheduled Over Next 6 Months (Type, Date, and Duration ofEach):

N/A

25. IfShutdown at End ofReporting Period, Estimated Date ofStartup:

N/A

• Excludes hours under administrative hold (June 1, 1985 to May 24, 1991) 0

Tennessee VnlleyAulhorily Browns Ferry Nuclear Plnnl 1993 Annual Opernling Report OPFRATIOJVAL

SUMMARY

¹ '"'~ '""""""""'-".~ '."-'-"". "'=" -':='"""'-"'"'"'""""'"':- "";:.:": '---"':-"" "'

"'ocket No.: 50-296 OPERATING STATUS 1.

Unit Name: Browns Ferry Unit Three 2.

Reporting Period:

Calendar Year 1993 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 12 13.

14 15.

16 17.

18 19.

20 21.

22.

23.

Hours in Re ortin Period Hours Reactor Was Critical Reactor Reserve Shutdown Hours Hours Generator On Line Unit Reserve Shutdown Hours Gross Thermal Generation h

Gross Electrical Generation h

Net Electrical Generation h

Unit Service Factor Unit Availabilit Factor Unit Ca acit Factor MDCNet Unit Ca acit Factor ER Net Unit Forced Outa e Rate December 1993 0

0 0

0 0

0 0

0 0

0 0

0 0

Year to Date 0

0 0

0 0

0 0

0 0

0 0

0 0

Cumulative*

73055 45306 5150 44195 0

131868267 43473760 42114009 60.5 60.5 54.2 54.2 21.6 24.

Shutdowns Scheduled Over Next 6 Months (Type, Date, and Duration ofEach):

N/A

25. IfShutdown at End ofReporting Period, Estimated Date ofStartup:

TO BE DETEIMINED

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

Tennessee ValleyAuthority

SUMMARY

OF Browns Ferry Nuclear Plant SAFETYEVALUATIONSFOR 1993 Annual Operating Report CORE COMPONENTS AND OPERA TINGLIMITS m" ""':::

1993

SUMMARY

OF SAFETY EVALUATIONS FOR CORE COMPONENTS AND OPERATING LIMITS Tennessee ValleyAutlrority

SUMMARY

OF BroNns Ferry Nuclear Plant.

SAFETYEVALUATIONSFOR I993 Annual Operating Report CORE COMPONENTS AND OPERA TINGLIMITS Unit 2 Core Operttting Linrit'sReport (SEBF 00-9303410 R1)

~ Descri tion/Safet Evaluation This safety evaluation supports the BFN Unit 2 Cycle 7 reload core design and the cycle specific updates to the BFN Unit 2 Core Operating Limits Report.

The core design and licensing analyses for this cycle were performed by General Electric (GE) with results documented in the Supplemental Reload Licensing Report.

GE reload core design bases and analysis methods are described in GE Standard Application for Reactor Fuel and the U.S. Supplement.

Operating limits for the cycle (i.e., linear heat generation rate, minimum critical power ratio, and maximum average planar linear heat generation rate) as determined by the licensing analyses are incorporated into the TVA Browns Ferry Nuclear Plant (BFN) Unit 2 Core Operating Limits Report.

Revision 1 ofthis safety evaluation was made to incorporate revision 1 ofthe Supplemental Reload Licensing Report and revision 1 ofthe BFN Unit 2 Core Operating Limits Report.

The Reload Licensing Report was revised to add exposure dependent minimum critical power ratio operating limits. The severity ofany plant pressurization transient event is worst at the end ofthe cycle primarily because the end-of-cycle all-rods-out scram curve gives the worst possible scram response.

Therefore, some limits relief may be obtained by analyzing the transients at other interim points in the cycle and administering the resulting limits on an "exposure dependent" basis. For Unit 2 Cycle 7, GE has added an additional analysis point at end ofcycle (end offull power capability with rated flow) minus 2000 megawatt days per short ton (MWD/ST). The Unit 2 Core Operating Limits Report is revised to incorporate the exposure dependent minimum critical power ratio limits.

The BFN Unit 2 Cycle 7 core is a control cell core design with a predicted fullpower life ofapproximately 8450 MWD/ST. Increased core fiow and coastdown capability increase this to a maximum cycle burnup ofapproximately 9800 MWD/ST. Maximum energy capability is limited by the licensing limit on peak pellet exposure for some ofthe older GE6 product line fuel.

The fuel types and final loading arrangement are specified in GE document 111-02034, Revision 1, "Browns Ferry 2 Cycle 7 Cycle Management Report" and "GE letter

'MQ:

93-073, "Browns Ferry 2 Cycle 7 Full Core Loading Plan Ec CSDM Period Correction Factors".

The fresh fuel types are GE7B and GE9B designs.

The GE7 Tennessee ValleyAulhority

SUMMARY

OF Brains Ferry Nuclear Plant SAFETYEVALUATIONS FOR 1993 Annual Operating Report CORE COMPONENTS AND OPERA TINGLIMITS K=."--":: '-""i':::-::"': " ":"~'":"'""""":"-""'-:-"-::"":""~=."-"":-'"-'-"'""-'-"

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'"'roduct line offuel has been previously loaded.

Use ofthe GE9 product line offuel has been evaluated an approved as Core Component Design Change Request (CCDCR)

No. 52. Both the GE7B and GE9B are barrier cladding designs and have no Preconditioning Interim Operating Management Recommendations (PCIOMR) restrictions.

The existing once-burnt and twice-burnt fuel do not contain barrier cladding and all PCIOMR constraints remain in effect for these bundles.

The core willinclude one twice-burnt reconstituted fuel bundle which willbe used to replace a leaking twice-burnt fuel bundle from Cycle 6. This bundle willbe loaded in a low power peripheral location and requires no special consideration during operation. No other reconstituted bundles remain in the Cycle 7 core. The core willalso include the four Westinghouse QUAD+

demonstration assemblies which were previously loaded in Cycle 6.

The cycle is analyzed for extended load line limitanalysis and increased core flow. 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.

Other hardware changes for Cycle 7 operation include:

~

Allofthe interior A2 control cell control blades (37 total) were replaced this outage.

The new blades are a mix ofmodified boiling water reactor (BWR)/6 control blades and newer design hybrid control blades.

These designs were evaluated and approved as CCDCR No.s 51 and 53.

~

A total of24 local power range monitor (LPRM) strings were replaced this outage.

Allofthe Image Sensing Technology (IST) (oAen referred to as Westinghouse design)

LPRMs that were in Cycle 6 are being replaced.

The new LPRM strings are a mix of GE NA-300 and newer design IST strings.

These designs were evaluated and approved as CCDCR No.s 49 and 50.

~

The Unit 2 process computer was replaced during the reload 6 outage.

New core monitoring software (3D Monicore) from GE willbe utilized in the new computer system.

The core operating limits for linear heat generation rate, maximum average planar linear heat generation rate, and minimum critical power ratio have been removed from the Technical Specifications and placed in the Core Operating Limits Report.

Therefore, incorporation ofnew operating limits for the Cycle 7 core design do not require a Technical Specification change.

The current safety limitminimum critical power ratio (SLMCPR) value specified in BFN Technical Specifications is 1.07. The GE9 fuel design is approved by the NRC for a lower SLMCPR of 1.06. However, for the mixed core which contains both fresh and once-burnt fuel ofthe earlier designs, the more limiting 1.07 value willcontinue to be used and no Technical Specification change is necessary.

l 0

Tennessee ValleyAuthority

SUMMARY

OF Bro>vns Ferry Nuclear Plant SAFETYEVALUATIONS FOR I993 Annual Operating Report CORE COMPONENTS AND OPERATING LIMITS Due to the change from TVAmethods to GE methods for performing Cycle 7 reload licensing analyses and changes in methodology for analyzing certain events, Updated Final Safety Analysis Report (UFSAR) Chapter 14 will,re'quire revising. Also, the BFN Cycle 7 Supplemental Reload Licensing Report needs to be incorporated into Appendix N ofthe UFSAR.

The BFN Cycle 7 reload core design is acceptable from a nuclear safety standpoint and no unreviewed safety question is involved.

Core Contponent Design Change Request - Units 1, 2, 3 (00-9301401 RO)

Descri tion/Safet Evaluation This safety evaluation addresses use ofGE's GE9 (designated GE8X8NB) product line of fuel in BFN Units 1, 2, and 3.

Progressing to advanced fuel assembly design offers potential for improved fuel cycle economics and plant operations.

A fuel assembly consists ofa fuel bundle and a channel which surround it. The rods ofall "the bundle types are spaced and supported by upper and lower tie plates, as well as fuel rod spacers.

The fuel rods are Zircaloy cladding containing uranium fuel pellets. For barrier fuel designs (GE7 and GE9), the fuel rods have a thin liner ofpure zirconium mechanically bonded to the inside ofthe cladding. Later fuel designs also contain water rods which are hollow Zircaloy tubes with fiow holes.

The major design differences between the GE9 fuel assembly being addressed in this safety evaluation and the previously loaded GE7 assembly type are:

~

Fuel rod design changes to enable the fuel bundle to attain higher burnup (e.g.,

changes to plenum volume, helium fillgas pressures, fuel density, fuel/cladding gap, pellet enrichments and gadolinia content).

~

Replacement ofthe four rod positions in the center ofthe 8X8 lattice (2 water rods and 2 fuel rods) with a large central water rod to improve nuclear efficiency.

~

Use ofaxially varying enrichment to optimize pin power distributions.

~

Introduction ofa high-performance ferrule spacer which provides better critical power ratio performance.

Tennessee VnlleyAuthority SUhfhfARYOF Browns Ferry nuclear Plant SAFETYEVALUATIONS FOR 1993 Annual Operating Report CORE COMPONENTS AND OPERATING LIMITS

~

Use ofa new high flowupper tie plate which reduces two-phase bundle delta-P to improve thermal-hydraulic stability margins.

~

Change from Zircaloy-4 to zircaloy-2 for fuel spacer and channel material.

GE fuel is designed and analyzed to meet specific mechanical and nuclear design bases and fuel licensing acceptance criteria. The NRC has reviewed and approved the GE9 fuel design.

With the exception ofthe SLMCPR specification, fuel type dependent information and operating limits have been removed from the Technical Specifications for inclusion in cycle-specific Core Operating Limits Reports.

The current SLMCPR value specified in BFN Technical Specifications is 1.07. The GE9 fuel design is approved by the NRC for a lower SLMCPR of 1.06. However, for mixed cores which contain any fresh or once-burnt fuel ofthe earlier designs, the more limiting 1.07 value willcontinue to be used and no Technical Specification change is necessary.

In the future, ifand 'when the loading criteria is met (i.e., all fresh and once-burnt fuel consisting ofGE9), a Technical Specification change could be implemented to take credit for the lower limit. TVA Nuclear Fuels is involved in establishing reload design bases and reviewing reload licensing reports, and willinitiate the Technical Specification change request to lower the SLMCPR ifand when the opportunity arises.

UFSAR Section 3.2 requires revision since the GE9 fuel assembly channels willbe made from Zircaloy-2 instead ofZircaloy-4. UFSAR Appendix H contains specific designations ofreload bundle types contained in BFN and willhave to be revised.

This CCDCR is acceptable from a nuclear safety standpoint and no unreviewed safety question is involved.

Tennessee ValleyAuthority

SUMMARY

OF Broils Ferry Nuclear Plant SAFETYEVALUATIONS FOR I993 Annual 0 crating Report FIELD COMPLETED PLANTMODIFICATIONS 1993

SUMMARY

OF SAFETY EVALUATIONS FOR

. FIELD COMPLETED PLANT MODIFICATIONS Tennessee ValleyAuthori ty

SUMMARY

OF Brains Ferry Nuclear Plant SAFETYEVALUATIONSFOR 1993 Annual 0 crating Report FIELD COMPLETED PLANT MODIFICATIONS Safety Evaluations or Unreviewed Safety Question Determinations (USQDs) for the following plant modifications, which were field completed during 1993, were summarized in previous Annual Operating Reports.

Therefore, they are not included in this report.

ECN/DCN No.

H05 860 L01656 L01896 L01929 L02003 P00093 P00162 P00360 P00511 P00697 P00732 P 00905 P00974 P02075 P03 106 P05235 P07147 W10618 W1 1179 W14012 W14487 W1 5757 Descri tion Addition ofVibration Monitoring Equipment to Recirculation Pum s 2A and 2B Warm Water Supply from Condenser Circulating Water CCW S stem to Ex erimental Greenhouse Modification to Control AirDistribution System to Provide Hi her AirPressure Addition ofPressure Interlock Switch Downstream of Coolin Su I Pum Dischar e Valve

.Re lacement ofPortions ofCore S ra Pi in Torus Inte rit Modification Removal ofAutomatic Initiation Opening Logic from High Pressure Coolant Injection (HPCI) Steam Line Valves Lon Term Torus Im rovement Plan LTTIP Relocation ofReactor Building Emergency Lighting Transformer Heating, Ventilation, and AirConditioning (HVAC)

Ca acit Replacement ofExisting Sewage Treatment System with a Sewa e Treatment La oon S stem Permanent Power Stora e Facilities Installation/Connection ofAdditional Substations Control Rod Drive CRD Scram Dischar eHeader Re lacement ofFlow Transmitter 19 and -20 Installation of4kV Power Su I Line Common Maintenance Buildin Post Accident Sam lin S stem Replacement ofReactor and Refueling Zone Ventilation Exhaust Radiation Monitors Main Condenser Retubin Pre aration ofthe New Process Com uter Room Evacuation Alarm, Code Call, Paging System Modifications See Annual Operating Report for Year 1991 1988 1987 1992 1988 1988 1989 1988 1988 1988 1988 1988 1987 1988 1992 1992 1991 1992 1992 1992 1992 4

Tennessee ValleyAuthority

SUMMARY

OF Drowns Ferry Nuclear Plant SAFETYEVALUATIONS FOR l993 Annual Operating Report FIELD COMPLETED PLANTMODIFICATIONS 1::: "':-::-

"-""":"-::-:::::-:i-:::-:;::::"'-""ll"

':-.':-"":::-:--:: ':-':i':::-!':::::i:--:::::ii:":::i:::::-:::-':-'::

"-'"::::=:-'::-::---':::: -':::-"

':=:i::::-"'-::::-':-"'CN/DCN No.

W16714 East Access Facilit Descri tion See Annual Operating Report for Year 1992 W16901 W17057 W17083 W17240 W17249 W17317 W17355 W17358 W17360 W17363 W17365 W17367 W17368 W17369 W17370 W17416 W17421 W17429 W17431 W17440 W17465 W17466 W17536 W17666 W17667 Instrumentation to Monitor Temperature ofResidual Heat Removal Service Water (RHRSW) EfHuent from the Residual Heat Removal Heat Exchan ers Control Room Design Review (CRDR) Modifications -.

Panel 3-9-53 CRDR Modifications - Panel 3-9-21 Replacement ofStation Oscillographs with Digital Fault Recorders CRDR Modifications - Panel 2-25-32 Removal ofDamper Motor Operators from Flow Control Operator (FCO) 0-FCO-75-16 and 38 and Lock Dampers in the 0 en Position CRDR Modifications - Panel 2-9-2 CRDR Modifications - Panel 2-9-6 CRDR Modifications - Panel 2-9-8 CRDR Modifications - Panel 2-9-21 CRDR Modifications - Panel 2-9-25 CRDR Modifications - Panel 2-9-4 CRDR Modifications - Panel 2-9-53 CRDR Modifications - Panel 2-9-47 CRDR Modifications - Panel 2-9-54/55 Various Modifications to the Control Air S stem Master Component Electrical List Revision to Add Post-Accident Monitorin Flow Indicators and Transmitters CRDR Modifications - Panel 2-25-32 CRDR Modifications - Panel 2-9-4 Revise Instrumentation Components in 2-FT-74-50/64 Loo s

D ell Floor Steel Qualification/Modification D

ell Floor Steel Qualification/Modification D

ell Platform Installation ofAutomatic Load Tap Changer on Common Station Service Transformer

'A'nstallation ofAutomatic Load Tap Changer on Common Station Service Transformer

'B'992 1992 1992 1992 1992 1992 1992 1992 1992 1992 1992 1992 1992 1992 1992 1992 1992 1992 1992 1992 1992 1992 1992 1992 1992

Tennessee ValleyAuthority

SUMMARY

OF Browns Ferry Nuclear Plant SAFETYEVALUATIONS FOR 1993 Annual Operating Report FIELD COMPLETED PLAiVTMODIFICATIONS

~i54R~~Mv. '~~~<' '""

'"" ~ " '~j~< ~~~':.

~

~<<'CN/DCN No.

W17722 W17723 W17724 W17820 W17821 W17822 W17845 W17907 W17914 W17915 W17928 W18011 W1 8053 W18060 W1 8077 W18092 W1 8470 W18895 W19557 W19558 W19559 W19560 W19562 W19563 W19700 W20321 W20466 W20654 W20657 W20667 Descri tion Retrofit ofUnit Service Station Transformer Unit 2 with Automatic Load Ta Chan er Retrofit ofUnit Service Station Transformer Unit 3 with Automatic Load Ta Chan er Replacement ofUnit Station Service Transformer

'B'oad Ta Chan ers National Fire Protection Association PA U rades NFPA U rades - Cable S readin Room A, NFPAU rades-Cable S readin Room B Re lacement ofValves 2-84-680/617 Unit 2 Reactor Buildin Fire Detection S stem Unit 3 Turbine Buildin Crane Modification Unit 1 Turbine Buildin Crane Modification U

rade ofUnit 2 Turbine Buildin Crane East Access Facilit Seismic Qualification ofUnit 1 RHRSW Header and Pi in Su orts Control Ba Ventilation Intake Ducts Re lace D ell Humidit Sensors and Transmitters Alternate Power Su lies for 2-FCV-74-59 and -73 Reconnection ofRHRSW S stem Pressure Lockin CRDR Modifications - Panel 2-9-3 CRDR Modifications - Panel 2-9-3 CRDR Modifications - Panel 2-9-3 CRDR Modifications - Panel 2-9-3 CRDR Modifications - Panel 2-9-3 CRDR Modifications - Panel 2-9-3 Removal ofPower for Equipment Access Lock Exhaust Fan and Associated Dam ers Standb Gas Treatment SGT Pi in Re lacement CRDR Modifications - Panel 2-9-47 Cable Reroute to Eliminate Need for Thermo-La Thermolag Fire Wrap -'480V Reactor Motor Operated Valve OV Board 2B Installation ofChemical Decontamination Taps on RHR S stem See Annual Operating Report for Year 1992 1992 1992 1992 1992 1992 1992 1992 1992 1992 1992 1992 1992 1992 1992 1992 1992 1992 1992 1992 1992 1992 1992 1992 1992 1992 1992 1992 1992 1992 0,

Tennessee VnlleyAuthority

SUMMARY

OF Brogans Ferry Nuclear Plant SAFETYEVALUATIONS FOR l993 Annual~Oerating Report FIELD COMPLETED PLANTMODIFICATIONS i Xiii;~it'.;.",':~V.@:;Keys';.

'~i Engineering Change Notice (ECN) 80004C - Replacement ofRHRS %Supply Valves-Unit3 Descri tion/Safet Evaluation This USQD addressed the replacement ofcurrent RHRSW supply isolation valves (HCV-23-31,

-37, -43, and -49) with Henry Pratt Nuclear Mark IIbutterfly valves.

The previous valve type has no qualification documentation and had been shown to have a tendency to develop leaks.

The new valves function in the same manner and were designed to meet the process piping design requirements.

A seismic analysis was performed and shows that the replacement valves do not degrade the system's seismic requirements.

The new valve replacements meet the Process Piping Design Conditions and the G-28 Pipe Classification Requirements and have the same failure position.

The replacement valves improve the integrity ofthe RHRSW system by meeting these requirements and have no adverse affects on the system's ability to perform its function in accordance with the UFSAR or Technical Specifications.

No Technical Specification change was required.

The margin ofsafety as defined in the basis for any Technical Specification was not reduced.

No unreviewed safety question was involved.

ECN P$130- Refueling Platfortn Roller Mounting Bracket Mollification-Unit 3 Descri tion/Safet Evaluation This USQD addressed modifying the refueling platform by cutting away the outside edge ofthe roller bracket to eliminate interference between the bracket and the limitswitch actuator track.

The refueling platform is a seismic'Class II mechanism; therefore, cutting away the outside edge ofthe roller bracket was required to be done without degrading the seismic Class IIqualification ofthe refueling platform.

The limitswitch is one ofseveral refueling interlocks which are designed to backup procedural core reactivity controls during refueling operations.

During refueling operations, with the mode switch in REFUEL position, the limitswitch (and other interlocks) input into circuitry which prevents (1) the withdrawal ofmore than one control rod and (2) the movement ofthe loaded refueling p atform over the core with any control rod withdrawn. Thus, this limitswitch is a necessary Technical Specification required component.

No Technical Specification change was required and no unreviewed safety question was involved.

Tennessee ValleyAuthority

SUMMARY

OF Broivns Ferry Nuclear Plant SAFETYEVALUATIONSFOR l993 Annnai ~Oeran~nRepnri FIELD COhIPLETED PLANTilIODIFICATIONS ECN P$22I - Revision to Turbine anti React'or BuilrlingControlAirSystem Flow Diagrams-Unit3 Descri tion/Safet Evaluation This USQD addressed a revision to the Unit 3 turbine and reactor building control air system drawings to reflect the as-constructed configuration.

This change did not affect the function or operation ofany Technical Specification required equipment served by the control air system.

No Technical Specification changes were required and no unreviewed safety question was involved.

Design Clrange Notice (DCN) tV12579 - Replacenient ofRatliation MonitorDryivellSample Line Contaitirnent Isolation Valves fVitltSolenoirl Operaterl Valves - Unit 2 (SEBFDCN900104)

Descri tion/Safet Evaluation DCN H7926 replaced electric motor actuators for Class 1E FCV-90-254A and B, -255, and

-257A and B in the Unit 2 radiation monitoring system (RMS) with identical electric motor actuators borrowed from the corresponding Unit 3 valves. However, the qualified life ofthese actuators expired in October 1992. This DCN replaced the above Unit 2 ball valves and electric motor actuators with new Class 1E solenoid operated gate valves.

The FCV prefix for valves 90-254A and B, -255, and 257A and B were changed to FSV. These containment isolation valves close on receipt ofa containment isolation signal. The existing piping and replacement valves were seismically, analyzed and support modifications implemented to assure acceptability.

The existing power and control cables 2PC629-I, 2PC632-I, 2PC635-I, 2-PC644-II, and 2PC647-II, between junction box (JB) 3335 and the subject valves were replaced with environmentally qualified valve pigtail leads which were supplied with the new valves.

'Calculation ED-Q2090-900073 evaluated the adequacy ofthe pigtail leads and the existing cables for voltage drop, ampacity, short circuit, and Appendix R high and low impedance fault considerations and found them acceptable.

The existing Panel 9-9 circuit breakers 203 and 303 were replaced with Class 1E type TED 15 circuit breakers in order to provide adequate circuit protection.

This modification affected the radiation monitoring system, containment isolation system, and 120VAC instrument and control power.

Tennessee ValleyAulhorily

SUMMARY

OF Browns Ferry Nuclear Planl SAFETYEVALUATIONS FOR 1993 Annual Operating Reporl FIELD COMPLETED PLANT MODIFICATIONS K':-":~::""

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Revision 1 ofthis safety evaluation was issued to revise information pertaining to a change to Technical Specification 3.7.A. The table had been recently deleted from the Technical Specifications, therefore, no Technical Specification change was necessary.

This modification required changes to UFSAR figures.

This modification was acceptable from a nuclear safety standpoint and'no unreviewed safety question was involved.

DCN W1$366-Int'egrntell ColiiinlterSysteni (ICS) Upgrade - Unit 2 (SEBFDCN910056 RI)

Descri tion/Safet Evaluation This DCN provided equipment mounting details, cable routing, and cable terminations to complete the Unit 2 ICS upgrade modification. This modification was required to support TVA's commitment to the NRC to implement Nuclear Regulatory Commission Regulation (NUI&G)0696 requirements.

The ICS upgrade modification willultimately provide a separate computer system for each unit. This DCN addressed the Unit 2 upgrade only. Unit 2 had been sharing a GE4020 plant process computer with Unit 1. Because the Unit 1 upgrade is not yet scheduled, this DCN leaves the GE4020 in place for Unit 1. The Unit 3 upgrade is addressed by DCNs W15365 and W15367.

The process computer system provides a quick and accurate determination ofcore thermal performance, improves data reduction, accounting, and logging functions for both the nuclear boiler and balance ofplant equipment, and supplements procedural requirements for control rod manipulation during reactor startup and shutdown.

The new Unit 2 system performs all current nuclear steam supply system and balance ofplant functions provided by the GE4020 computer as well as the following additional functions:

~

Safety parameter display system (SPDS)

~

Sequence ofevents

~

Rod scram time recording

~

Transient recording analysis

~

Rod worth minimizer This safety evaluation was revised to delete wiring for three computer points which are addressed in DCN W20466.

No revision to plant Technical Specifications was required as a result ofthe implementation of this DCN. UFSAR Sections 7.16, F.S, F.6.16, and Appendix 7.7B required revision to reflect

'22-

Tennessee ValleyAuthority

SUMMARY

OF Browns Ferry Nuclear Plant SAFETYEVALUATIONS FOR I993 Annual Operating Report FIELD COMPLETED PLANTMODIFICATIONS C -::-'":::i'~

the ICS installation. In addition, numerous UFSAR figures required revision to reflect changes made to their TVAsource drawings.

This modification did not result in a reduced margin ofsafety as defined in the basis for any Technical Specification. No unreviewed safety question was involved.

DCN WI5723 - Upgratle ofSite Evacttafion, Fire Alarm, Cotle Call, anti Paging Sysfems-Unif' (SEBFDCN920027)

Descri tion/Safet Evaluation This DCN provided for the upgrading ofthe nonsafety-related site evacuation, fire alarm, code call, and paging systems in order to meet the intent ofNRC IE Bulletin 79-18. This DCN addressed only those devices associated with Unit 2.

The DCN provided for the addition ofstrobe lights, alarm bells, electronic sirens, relay boxes, tone generators, uninterruptible power supply gJPS), replacement ofspeakers and associated cabling and conduits.

In addition to this work, the DCN also provided for the fabrication and installation ofequipment and conduit supports and the interface ofthe new equipment with the existing systems.

These additional devices were required to ensure proper notification ofall plant personnel in high noise areas ofa site emergency.

Revision 1 ofthis safety evaluation was prepared in order to address Field Design Change Notice (FDCN) F17991A which corrected the breaker compartment used for the Unit 2 reactor building evacuation alarm UPS system on lighting cabinet LC203.

No Technical Specification change was required.

UFSAR Section 10.18.6 text required revising as a result ofthis modification.

Implementation ofthis modification did not reduce the margin ofsafety as defined in the basis for any Technical Specification. No unreviewed safety question was involved.

Ol

Tennessee ValleyAuthori ty

SUMMARY

OF Brogans Ferry Nuclear Plant SAFETYEVALUATIOjVSFOR 1993 Annnai Operating Eeparl FIELD COhIPLETED PLANTMODIFICATIONS 1'Tn"nn-"""'*"'""'""'4"-""'i'::-."'--"-"w""'""wi"""'"--'-"'""""""*-"'"--"-'-""""-nn"-*"'Q'n'='"a'"""-"

"4-'zc DCN Wl7274 anti F24972 - Replaeetnent ofUnitBatteries 2 anti 3 - Unit 0 (SEBFDCN910185)

Descri tion/Safet Evaluation r

This modification replaced the existing unit batteries 2 and 3 (C and D type LCUN 29) with new.

Class 1E (C and D type LCUN 33) batteries to support continued Unit 2 operation. Unit battery 3 (required for Unit 2 operation) was reaching the end ofits qualified life. Furthermore, BFN

'esign studies concluded that existing Class 1E 250VDC unit batteries 2 and 3 may not,have had adequate capacity to meet expected design loads for multi-unit operation.

The new batteries have a higher one minute rating of2080 amperes, a 30 minute rating of 1472 amperes, and an 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> discharge rating of2320 amper-hours at a terminal voltage of210 volts at 77'F. Although the number ofplates per cell was increased, the size ofeach cell container remains unchanged.

The weight ofeach ofthe 120 cells was increased by approximately 34 pounds, resulting in an overall weight increase of4080 pounds per battery. Due to the increased weight, new, seismic Class I battery racks were installed.

'he operation ofa battery results in the generation ofhydrogen gas which may form an explosive mixture ifexcessive concentrations are allowed to form. The hydrogen evolution from the fully charged replacement batteries operating at r'ated capacity at 77'F is 0.0041 ft3/cell/hour as compared to 0.0036 ft3/celVhour for the batteries replaced.

The present HVACsystem was evaluated and found sufficient to ensure that hydrogen concentration with the new batteries will not reach potentially flammable concentrations.

Electrical calculations were performed to ensure that all associated equipment (i.e., breakers, fuses, and cables) were adequate for the replacement batteries.

Likewise, the existing battery chargers were evaluated and found adequate to charge the fullydischarged replacement batteries, based on actual worst case duty cycle ampere-hour discharge, in approximately 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> under normal load conditions to meet design requirements.

In addition, battery operation required during a postulated station blackout event was evaluated and found adequate.

This DCN retagged the power cables between existing unit battery 3 and battery board 3.

Revision 1 ofthis safety evaluation was prepared in order to address the scope ofFDCN F24972.

This FDCN provided for the downscoping ofW17274 to remove the work associated with relocating the single cell battery charger supports and fuse boxes designed by DCN H4514.

Implementation ofthis modification and system operation followingthe modification are in accordance with Technical Specification requirements.

The battery replacements have no adverse impact on the 250VDC system, unit battery 2, or unit battery 3 performance and therefore have no adverse impact on Technical Specification limitingcondition for Operation (LCO), surveillance intervals, or their bases.

No Technical Specification change was required.

Tennessee VnlleyAutltority

SUMMARY

OF Broils Ferry Nuelenr Plant SAFETYEVALUATIONS FOR 1993 Annual Operntinp Report FIELD COMPLETED PLANTMODIFICATIONS kg,i::,,:",;;",.ppIII'-.;.::.*,$F~i;:,*,';;;i:;";:::;:::::."P,:,,:::,::,:,::::':f6~'~4,::.

A change to UFSAR Section 8.6.3 was initiated to reflect the increased replacement battery disch'arge ratings for unit batteries 2 and 3.

This modification is acceptable from a nuclear safety standpoint and no unreviewed safety question was involved.

'I DCN Wl7337 rtnrl W1 7491 - Hrtrrrenerl Wetivell Vent (HWWVl-Unit 0 (SEBFDCN920109)

Descri tion/Safet Evaluation These DCNs together provided the primary containment ofUnit 2 with an exhaust line from the wetwell vapor space to the stack.

The standby gas treatment system (SGTS), which is a low pressure system, can now be bypassed using the dedicated hardened vent path should torus pressure exceed SGTS operating conditions.

The wetwell vent provides assurance ofpressure reliefthrough'a path with scrubbing offission products.

The vent is sized such that under conditions ofa constant heat input at a rate equal to 1.05% ofrated thermal power and a containment pressure equal to the primary containment pressure limitof56 pounds per square inch gauge (PSIG), the exhaust flow through the vent is sufficient to prevent the containment pressure from jeopardizing the structural integrity ofthe torus.

DCN W17337 provided a common header for all three units to the stack that provides:

~

14" pipe and supports for each unit tie-in to the exhaust point in the stack;

~

Manual butterfly valves 1, 2, 3-64-737;

~

Secondary containment penetration ofUnit 1, including drywell chiller piping stubs (capped for future use);

~

Tie-in to secondary containment penetrations for Unit 2 and Unit 3;

~

Penetration ofthe south air intake louver ofthe stack;

~

Drain line with a steam trap, a bypass valve,'nd two shutoff valves that ties into existing offgas condensate sump drain;

~

Valve pits for valves 1, 2, and 3-64-737, and man hole for the radwaste drain.

DCN W17491 provided for the Unit 2 torus to the common header portion ofthe vent including:

~

14" pipe and supports from the suppression chamber inerting supply to the tie-in to the common header;

~

Secondary containment penetration ofUnit 2 reactor building, including drywell chiller piping stubs (capped for future use);

~

Primary containment isolation valves 2-FCV-64-221 and 2-FCV-64-222 equipped with air operators, pilot solenoid valves, and position indication switches;

Tennessee ValleyAuthority

SUMMARY

OF Browns Ferry Nuclear Plant SAFETYEVALUATIOiVSFOR 1993 Annual Operating Report FIELD COMPLETED PLANTMODIFICATIONS

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Control air piping and valves with containment atmosphere dilution (CAD) (N2) backup to supply the solenoid valves with air/nitrogen; 250VDC power from RMOVboards 2A and 2B to supply the solenoid valves and to provide position indication to the control room; 250VDC power from RMOV board 1A to provide position indication ofvalve 2-64-737 to the control room; Key-lock permissive switches, handswitches, indication lights, and annunciators installed in the control room; Test connections with normally closed isolation valves; Removal ofthe obsolete Group 7 isolation lights to make room for instrumentation associated with this modification; Removal ofabandoned fire protection piping; A 2" drain line that contains two piston check valves for secondary containment, a steam trap, two shutoff valves, and a bypass valve; A 3" portion ofthe 1" diameter spool piece after vent valve 0-67-684 was removed and the cap replaced to eliminate an interference with the 14" HWWVpipe.

Revision 1 ofthis safety evaluation was issued to clarify the "Additional Information" to be provided by BFN Engineering.

A change to UFSAR text, tables, and figures was required due to this modification. The Technical Specifications were reviewed and it was found that valves 2-FCV-64-221 and 2-FCV-64-222 were required to be added to the primary containment isolation valve Table 3.7A.

However, this table was to be deleted from the Technical Specifications. A Technical Specification change had to be implemented to delete this table.

This modification was acceptable from a nuclear safety standpoint and no unreviewed safety question was involved.

DCN Wl7357-CRDR Modifications to Control Room Panel Z-9 Unit 2 (SEBFDCN920179)

Descri tion/Safet Evaluation

/

A This DCN consisted ofmodifications to Panel 2-9-5 for resolving identified human engineering discrepancies between the design ofthe Unit 2 control room and TVA's human factors standards.

These modifications were applications ofhuman engineering principles to improve man-machine interface characteristics and, thus enhance operator response during abnormal and emergency conditions ofthe plant. In general, this DCN performed the following:

~

Rearranged/relocated control switches and instruments;

~

Installed improved component labeling and switch position escutcheons; Tennessee ValleyAuthority

SUMMARY

OF Browns Ferry Nuclear Plant SAFETYEVALUATIONS FOR 1993 Annual Operaling Report FIELD COMPLETED PLANTMODIFICATIONS 4%+4+'

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~

Replaced switch handles with shape and tactilely-coded, black handles;

~

Replaced recorders, indicating meter scales (with colorbanding);

~

Replaced indicating light lenses to confirm with applicable color standards; and

~

Implemented modification while maintaining the seismic integrity ofthe panel.

UFSAR Figures 7.8-1 Sheets 1 and 2, 7.7-1b, 8.7-4c Sheet 2, 3.8-2, 3.8-4, 3.4-8a Sheet 1, 7.5-1, 7.5-3b, and 7.2-3e were affected by this modification. This modification did not change any Technical Specification requirement.

These plant modifications did not reduce nuclear safety at any time. No unreviewed safety question was involved.

DCN Wl7359 - CMMorlifieations to Control Rooni Panel 2-9 Unit 2 (SEBFDCN920186)

Descri tion/Safet Evaluation This DCN consisted ofmodifications to Panel 2-9-8 for resolving identified human engineering discrepancies between the design ofthe Unit 2 control room and TVA's human factors standards.

These modifications were applications ofhuman engineering principles to improve man-machine interface characteristics and, thus, enhance operator response during abnormal and emergency conditions ofthe plant. In general, this DCN performed the following:

~

Rearranged and relocated handswitches, indicating lights, recorders, and meters;

~

Provided improved component labeling;

~

Provided new scales and colorbanding for applicable meters;

~

Provided new switch escutcheons, as necessary;

~

Provided unique identifications (UNIDs) for components which did not have unique identification numbers;

~

Replaced switch handles with shape-coded black handles;

~

Replaced indicating light lenses, as necessary to conform with BFN color conventions;

~

Replaced all recorders; and

~

Implemented panel repairs while maintaining the seismic integrity ofPanel 2-9-8.

Revision 1 o'fthis safety evaluation was issued to remove system 43 (sampling system) from the list ofaffected systems due to the reassignment ofUNID2-CR-47-165 to the "Generator PCB Cooling Water Conductivity" recorder on Panel 2-9-8.

Tennessee VolleyAulhority

SUMMARY

OF Bro>vns Ferry Nuelenr Plant SAFETYEVALUATIONS FOR 3993 Annnni Operating Report FIELD COMPLETED PLANTMODIFICATIONS This DCN relocated and rearranged controls for the offgas system.

UFSAR Figure 7.12-2b was changed to reflect the relocation ofhand switch (HS) 2-HS-90-155. No Technical Specification change was required.

These modifications did not reduce nuclear safety and no unreviewed safety question was involved.

DCN Wl7362 - CRDR Modifications to Control Room Panel 2-9 Unit 2 (SEBFDCN920209)

Descri tion/Safet Evaluation This DCN consisted ofmodifications to Panel 2-9-20 for resolving identified human engineering discrepancies between the design ofthe Unit 2 control room and TVA's human factors standards.

These modifications were applications ofhuman engineering principles to improve man-machine interface characteristics and, thus enhance operator response during abnormal and emergency conditions ofthe plant. In general, this DCN performed the following:

~

Rearranged control switches and instruments;

~

Replaced switch escutcheons and switch handles with black handles;

~

Replaced meter and/or meter scales (with colorbanding, as applicable);

~

Provided hierarchical and component labeling; and

~

Implemented panel repair while maintaining seismic integrity ofPanel 2-9-20.

Revision 1 ofthe safety evaluation addressed the deletion ofmodifications associated with Category 3 and 4 human engineering discrepancy resolutions as performed by FDCN F21492.

Revision 1 included the addition ofSGTS Train A and Train B operating indication and the rerouting of'cables from conduits to existing cable tray.

This modification required UFSAR figures to be revised to reflect changes and component location and identification. A Technical Specification change was not required to implement this DCN. These modifications did not reduce nuclear safety and no unreviewed safety question was involved.

DCN Wl7481 - Generator Backup Relay Logic Modification -.Units2 and 3 (SEBFDCN920228)

Descri tion/Safet Evaluation The modification reconfigured the existing generator 3 backup relay logic to eliminate the possibility that the loss ofthe relay logic potential from a single bus would trip the unit. This modification added a second channel to the generator backup relay logic. The new generator 0

Tennessee ValleyAuthori ty

SUMMARY

OF Browns Ferry Ãuclear Plant SAFETYEVALUATIO1VSFOR 1993 Annual Operating Report FIELD COMPLETED PLAIVTMODIFICATIONS P

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backup impedance relay willreceive its sensing voltage from the existing 500KV bus 2, section 2 potential voltage transformer.

Therefore, failure ofa single potential voltage willtrip one channel ofthe generator backup relay logic but not the other channel.

The two channels ofgenerator backup relay logic were configured in a two-out-of-two logic scheme such that a single channel backup relay operation willnot trip the unit. Ifboth channels ofgenerator backup relay logic operate, the unit willbe tripped by lockout relay 386TF.

Circuit breaker 5264 and 5268 position switches and associated auxiliary relays willdisable and bypass one logic channel ifits associated circuit breaker is open, allowing the remaining channel to trip the unit.

The new equipment was mounted and wiring performed in relay room panels 32, 33, 34, 35, and'7 in the control bay. The change was nonsafety related, nonseismic, and had no Appendix R or 10CFR50.49 environmental requirements.

There were no external cable or raceway modifications involved.

This change also rewired generator 2 backup relay logic circuit auxiliary relays (5244X and 5248X) in relay panel 35. This allows the Unit 1, 2, and 3 auxiliary logic relays to have the proper tier arrangement.

This modification was worked in stages to separate Unit 2 work in order to minimize the impact on Unit 2 during the Cycle 6 outage and to ensure that Unit 2 and Unit 3 main transformer banks were not inoperative at the same time.

This modification afFected equipment required for the operation ofUnit 2. Therefore, this modification was performed in accordance to the requirements of Site Standard Practice (SSP)-12,50.

Compliance with Technical Specifications and standard work control practices during implementation was adequate to assure that operation ofUnit 2 was not'impacted.

This modification did not involve an unreviewed safety question and no Technical Specification changes, radwaste systems, or special tests or experiments were involved. This modification did affect UFSAR Figure 8.3-6 and a change was initiated.

N DCN Wl75Z7-Neiv Control Rooni Entergency Ventt7tttion (CREV) Syste]n - Unit 0 (SEBFDCN920122)

Descri tion/Safet Evaluation This modification provided a new CREV system which replaced the two 500 standard cubic feet per minute (SCFM) emergency pressurization systems A and B.

Tennessee ValleyAulhorily SUNNARYOF.

Browns Ferry PIuelear Plant SAFETYEVALUATIOiVSFOR.

I993 Annual Operaling Reporl FIELD CONPLETED PLANTNODIFICATIONS KS~~'- '-""'~"" ": "":;"> "'"'4';;-'"-':.,'.'":-"~"':"";;::" ""

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The new CREV system filtertrain assemblies are classified as safety related and were constructed to meet American Society ofMechanical Engineers (ASME) N509 requirements.

The new system utilizes the existing Unit 2 high-efficiency particulate air (HEPA) filterbank located in the Unit 2 ventilation tower intake plenum. New ductwork connects the HEPA filter bank to the new charcoal filter assemblies which are located in the relay equipment room on elevation 617'f the control bay. The new ductwork is essentially leakproof, all welded construction.

Electric (15.5kW) duct heaters, sized to main'tain incoming outside air relative humidity below 70%, are located upstream ofcharcoal absorbers to ensure high charcoal absorption efticiency. Discharge ducts are routed within the ceiling space above the Unit 2 hallway and terminate in each control room (Unit I/2 and 3). The'discharge into each room will mix the filtered outside air with control bay return.air returning to the control room air handling units for recirculation. The cooling capacity ofthe existing recirculation units willaccommodate the added heat load ofthe new CREV system units. The two filtertrains are 100% redundant with each having the capability to provide 3000 SCFM offresh filtered outside air for positive pressurization.

The two filterunits are located in the same room on the 617'levation, but are on opposite sides ofthe room to maintain separation between the redundant trains.

The new CREV system is started automatically on a Group 6 accident signal (primary containment isolation system [PCIS]) or high radiation signal in the Unit I or 3 control bay air intake ducts.

Segregated electrical normal and emergency power, controls, and equipment for each ofthe trains are classified 1E. The new CREV system train A equipment is powered from 480V RMOVboard 1A, served by Unit ll2 emergency diesel generators (EDGs) A or B, while the new train B equipment is powered from 480V RMOV board 3B, served by Unit 3 EDGs 3EC or 3EB. Supports for the CREV system components are seismic Class I design and the equipment installed has been seismically qualified.

Technical Specification Amendment Change TS-323 was approved by the NRC. This change removed the LCOs in place until the end ofUnit 2 Cycle 6, identified that the CREV system is to assist other sources ofpressurization in maintaining positive pressure, and revised the required damper action at isolation to the necessary dampers listed in a Technical Specification table instead ofa specific list in the text.

Implementation ofthis modification did aA'ect the UFSAR. The discussion ofthe control bay habitability zone (CBHZ) and ventilation ofUFSAR Section 10.12 and Figures 1.6-3 Sheet 2, 7.12-2a Sheet 2, 7.12-2b Sheet 2, 7.12-2b Sheet 3, 8.5-8b, 8.5-11c, 10.12-2a, 10.12-2b, 10.12-3 required updating as a result ofthis modification.

This modification was determined to be acceptable from a nuclear safety standpoint.

The change enhances the operator's ability to respond to an accident by insuring the dilution ofradioactive contaminated air entering into the main control room. No unreviewed safety question was involved.

Tennessee ValleyAutltorily

SUMMARY

OF Brogans Ferry Nuclear Plant SAFETYEVALUATIONSFOR 1993 Annual Operating Report FIELD COMPLETED PLANT MODIFICATIONS 4". K""'""-""'"::::""'"" "-:." '::

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SPAN?g.$iv:;t(8:,,c;:.'<.p'p'~ i%%';.'.:: <'.:,.<span,., <sx,j(+jxjo'Qgvix i'i~> <<:.",.ix5gc; g)~.,:i.;; <Ã(g':. >$$4gg'.,::.,i~< y'?gt"'r'~%'j"~. 4>$<c<~p,", /~imp~

DCiVGI7693C-Replttcentent ofGE Brettl<er Trip Units - Unit3 (SEBFDCN920031 R1)

Descri tion/Safet Evaluation This DCN provided necessary documentation to allow modification ofvarious GE 480V type AKcircuit breakers, which were modified to replace the existing, noninstantaneous, electronic trip units (EC, power sensor, and SST) with microprocessor based electronic trip units (RMS-9).

The existing GE 480V AK-25 and AK-50 circuit breakers primarily employed either EC trip devices, SST trip devices, or a power sensor trip device.

The EC devices were high maintenance while the SSTs and power sensors were obsolete.

The RMS-9 conversion kit (vendor recommended replacement part) was a pre-engineered unit representing product improvement and was identified by GE as providing improved flexibility,accuracy, reliability, and long life.

This DCN provided engineering support for using the generic equipment replacement process to replace the existing trip devices with RMS-9 conversion kits. BFN's Generic Substitution Engineering Requirements Specification incorporates installation and procurement requirements.

Revision B ofthis DCN incorporated two additional circuit breakers on nonsafety-related 480V lighting board 3 for RMS-9 conversion with noninstantaneous trip applications.

Revision C ofthis DCN was issued to revise specification N1M-002; Data Sheet 3007 (Generic Substitution Data Sheet), to delete all GE RMS-9 micro-versa trip devices with the instantaneous trip function (GE Model Code 02). This revision was performed due to a "Fail'ure to Comply Notification to the NRC" in accordance with 10CFR21.21.

UFSAR figures were affected by this modification and a change was initiated.

The Technical Specifications for BFN do not explicitly address 480V circuit breaker trip devices or their individual adjustments or characteristics.,The generic replacement oftrip devices with RMS-9 units provided enhanced equipment reliability thus assuring that the capability ofsafety-related equipment was not diminished as a result ofinadequate protection ofless reliable devices.

Therefore, this DCN had no adverse impact on Technical Specifications LCO, surveillance

~ intervals, or their bases for any plant system.

No, Technical Specification change was required Nuclear safety was not decreased with the installation ofnoninstantaneous trip RMS-9 trip units and no unreviewed safety question was involved.

Tennessee ValleyAuthori ty

SUMMARY

OF Browns Ferry Nuclear Plant SAFETYEVALUATIONSFOR 1993 Annual Operating Report FIELD COMPLETED PLAiVTMODIFICATIOiVS DCN O'17810- Reactor Water Cleanup(RWCU) Punip Suction antlDisclsarge Piping Replacement/Reroute - Unit3 (SEBFDCN920154 R1)

Descri tion/Safet Evaluation DCNs W17810, W17811, W17701, W18486, and W19297 combined accomplish the reroute of the RWCU piping outside primary containment, the replacement ofselected sections ofpipe both inside and outside primary containment from 304 to 316 nuclear grade (NG) stainless steel material, pipe support modifications, pump enhancements, instrumentation modifications, and temperature detection modifications for the upgrade ofthe RWCU system for Unit 3.

DCN W17810 provided the following:

~

Equipment in the Unit 3 RWCU pump room was replaced or enhanced.

Equipment included pumps, motors, piping, and valves.

Enhancements include new John Crane mechanical pump seals, new pump bearing housing material (i.e., stainless steel), a flange connection to the pump discharge nozzle, and bonnetless globe valves;

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Prefabricated piping assemblies were installed to provide cooling water and purge water to the pump mechanical seals, bearing housing, and stuffing box jacket;

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Relief valves were added in the CRD seal cooling water supply line to prevent overpressurization ofthe RWCU pumps;

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Instrumentation located in the pump room was relocated outside the pump room to Panel 25-5-1;

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Three temperature indicating switches were upgraded.

Two were moved to Panel 25-5-1;

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Three temperature elements were added to each ofthe two pumps;

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Flow indicating switches monitoring the pump discharge pressure were replaced.

These were moved to Panel 25-5-1;

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Two new pressure indicators were added to the existing gauges.

Existing pressure indicators used to measure pump suction and discharge during normal operation were moved to Panel 25-5-1;

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Accelerometers were added to the pump bearing housing to monitor pump vibration. These were added to Panel 25-5-1;

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A level transmitter and pressure transmitter were removed from Panel 25-5-1 to a new seismic Class I designed and mounted panel, located next to Panel 25-5-1; Relays and associated wiring were removed from Panel 25-2 to 25-5-1. New relays were added to Panel 25-5-1;

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Three thermowells were replaced with one being relocated;

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New power and control cables were added for the new accelerometer and thermocouples used to monitor RWCU pumps 3A and 3B performances.

Also, existing equipment from Panel 25-2 was relocated to 25-5-1 and cables spliced to the new location. Power to this new location was added from an existing source;

~

New supports were installed for the new electrical conduits from pump room to panels applicable to this modification;

Tennessee VnlleyAutltority

SUMMARY

OF Bro>vns Ferry Nuelenr Plant SAFETYEVALUATIONSFOR 1993 Annunl Oj erotic Report FIELD COMPLETED PLANTMODIFICATIONS

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The small bore piping and instruments, disconnected by DCN W17811, were redesigned and/or reconnected;

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A 3/4" vent line was installed downstream ofthe nonregenerative heat exchangers;

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The Unit 3 seal heat exchanger was elevated as part ofthe prefabricated seal cooling water assembly;

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Two check valves in the RWCU return lines were installed to satisfy single active component failure criteria in the event ofa high-energy line break (HELB) in the RWCU system;

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Handswitch 3-HS-69-2B in the heat exchanger. room was elevated to move it out ofthe range ofpipe whip and/or jet impingement postulated at the anchor near the X-14 penetration;

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A seismic Class IIladder was installed to reach the handswitch described above.

This modification complied with Technical Specification 3.6.G/4.6.G and did not modify any applicable portion ofthese requirements.

This modification did not require changes to the welds listed in Section 4.6.G.2 as these welds are for inside the primary containment.

Surveillance Requirement 4.6.G.2 lists welds which require additional inspections due to whip concerns.

These welds were affected by DCN W17811.

Technical Specification cha'nge TS-330 was initiated to remove these welds from the Technical Specification.

Temperature switches will be added by DCN W19297 for RWCU system HELB detection and mitigation which willalso require a Technical Specification change to Section 3.2/4.2.

UFSAR figures were impacted by this modification. In addition, UFSAR Table 7.3-2 Sheet 2 is to be updated by DCN W19297 to remove eight temperature switches inside the RWCU pump rooms which detect high temperatures and to add four temperature monitors to the RWCU heat exchanger room near penetration X-14 to detect HELBs.

This change was acceptable from a nuclear safety standpoint and no unreviewed safety question was involved.

DCN Wl7811 - RWCU Pump Suction anti Discitarge Piping Replacentent/Reroute - Unit3 (SEBFDCN920067 R2)

'I Descri tion/Safet Evaluation This DCN included the followingmodifications to the Unit 3 RWCU piping system:

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Removed and replaced the 6" diameter RWCU process piping from the RWCU 69-500 valve inside Unit 3 primary containment to the X-14 penetration.. This section ofprocess pipe utilized the existing piping support locations and followed the existing pipe routing. Existing valves69-500 and FCV-69-1 and all small bore branch connections for this section of process pipe was removed and reinstalled at their present locations;

Tennessee ValleyAuthorily

SUMMARY

OF Broivns Ferry Nuclear Plant SAFETYEVALUATIONS FOR 1993 Annual Opernling Reporl FIELD COMPLETED PLANTMODIFICATIONS

.",,"';;:.'<< ",:"~,.-,;:; -. ",, :;,:.;.:.::;;~,.'.~.:,.::;;',.;,':...k..:: S~g.".;:..%-,:

. ~~s,,';M

~

The RWCU piping outside the drywell was rerouted to establish a cold leg suction for the RWCU pumps.

The flowfrom the recirculation system suction willfirst pass through the regenerative and nonregenerative heat exchangers and then to the suction ofthe RWCU

. pumps which willdischarge through the demineralizer then back'through the regenerative heat exchanger to the feedwater header;

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The rerouted piping was replaced with stainless steel 316 and stainless steel 316 NG. In addition, the hot piping between the regenerative and nonregenerative heat exchangers and

'he RWCU return from the regenerative heat exchanger to the first valve into the feedwater headers was replaced with stainless steel 316 and stainless steel 316 NG.

C The pipe rerouting was performed to alleviate the pump seal mechanical maintenance problems associated with the RWCU pumps which have been mostly attributed to the high temperature water passing through the pump producing thermal stresses.

The new routing provides cold water (approximately 130'F) to the pump suction.

The material replacement was performed because the existing RWCU piping was made ofa material which is susceptible to IGSCC.

Stainless steel 316 NG is an approved material less susceptible to IGSCC.

This modification was a pipe replacement and reroute only. AIIvalves and small bore pipe were reinstalled and reconnected so as not to change the function ofthe system.

The small bore pipe and valves inside primary containment were reinstalled to original configuration. The valve controls were also restored to original configuration.

This DCN was worked in conjunction with DCNs W17810, W17701, and W18486.

These DCNs perform the pipe stress and supports analysis and seismically qualify the RWCU modification and restore the system pressure boundary integrity.

A The breaching ofprimary containment required that the reactor remain shutdown and reactor water temperature less than 212'F with fuel in the vessel or defueled until primary containment integrity was returned to Technical Specifications.

Implementation ofthis design makes the RWCU system inoperable.

The RWCU system is the normal method ofcontrolling coolant chemistry in the vessel.

Technical Specification 3.6.B establishes coolant chemistry limits which were followed. Surveillance requirement 4.6.G.2 lists welds which require additional inspections due to pipe whip concerns.

These welds were affected by this design.

Technical Specification change TS-330 was initiated which willremove these welds from the Technical Specifications.

Also, additional temperature switches were added for RWCU system HELB detection and mitigation which require a Technical Specification change to Section 3.2/4.2.

These items are addressed in DCN W19297.

This modification did not require a special test or experiment nor affect the radwaste system.

This modification did indirectly affect the UFSAR.

This modification was acceptable from a nuclear safety standpoint and no unreviewed safety question was involved.

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i 0;

Tennessee ValleyAuthority

SUMMARY

OF Drowns Ferry Nuclear Plant SAFETYEVALUATIONS FOR J993 Annual Operaling Report FIELD COMPLETED PLANTMODIFICATIONS DCN WI7999 - Offgas Stack Motlifications-Unit 0 (SEBFDCN920194 RO)

Descri tion/Safet Evaluation This modification redesigned the isolation capability ofthe cubicle exhaust, steam packing exhaust, SGTS dilution crossties, and Units 1, 2, and 3 dilution lines. Additionally, this DCN installed a bypass line in the cubicle exhaust and steam packing exhaust line with the identical isolation capability. This modification also restored power and controls to the Unit 1 dilution fan and the associated flow control dampers in addition to increasing the speed ofthe fan to ensure adequate flowfor operation ofthe wide range gaseous efHuent radiation monitoring system.

This change was necessary to alleviate the concern that during operation ofthe SGTS, the stack allows release ofradioactive effluent at ground level via reverse flowthrough common discharge paths.

This modification provided for independent automatic closure ofdampers due to back draft or no flow. As such, potential for ground level releases due to reverse flow in any line will be mitigated..

The implementation ofthis modification affected secondary containment and reactor building ventilation during the interim configuration. The proper control ofthe secondary containment is provided by Technical Specification 3/4.7.C and Technical Specification 3/4.7.B controls the SGTS.

The modification also affected the wide range gaseous effluent radiation monitoring system during implementation.

The gaseous monitoring is properly controlled by Technical Specifications 3/4.2.K and 3/4.8.B. This modification also required the temporary breach offire barriers. No Technical Specification change was required to implement this modification;- No

~

unreviewed safety question was involved.

This modification required revision to UFSAR Section 5.3.3.1, 5.3.3.7, 9.5.4, and 14.6 and to UFSAR Figures 9.5-1, 8.5-12a, 8.5-13a, 8.7-4c, 12.2-61, and 12.2-62.

DCN W18298-RWCU Piping Reroute, Replacenient, Snpports, Break Detection - Unit2 (SEBFD CN920181 R2)

Descri tion/Safet Evaluation This modification included:

1) Reroute ofthe Unit 2 RWCU piping such that the water coming directly from the reactor (recirculation line suction) first passes through the regenerative and nonregenerative heat exchangers through the RWCU pumps, to and from the demineralizer room, and then back through the regenerative heat exchanger and into the feedwater lines; 2)

Installation ofa check valve along seismic Class I pipe outside primary containment penetration X-9B; 3) Installation offour new safety-related temperature detectors in the RWCU heat Tennessee VnlleyAuthority

SUMMARY

OF Browns Ferry Nuclear Plant SAFETYEVALUATIONSFOR l993 Annual Operating Report FIELD COMPLETED PLANTMODIFICATIONS k':~ "'"" ~V

~'~c" "'

'o"" "" ' ~",'Z'"

"j WM" c'j@WP)%jib.;i%i%Pj':)(>ji~~pi:4"j~".'s'"j~j)vY:.'::::j@p)k(p74<F':~s'<"'-%%'~p'<i~(e ~w >++A'<%j.

++><~("i.x':<' <%."'A~q.'i exchanger room for HELB detection; and 4) Installation ofa 2" chemical decontamination connection with double isolation valves to provide a permanent decontamination connection for the Unit 2 RWCU system piping.

Revision 2 ofthe safety evaluation was issued to incorporate FDCN F20801A which added the 2" chemical decontamination connection.

The connection is in the steam vault immediately upstream ofcheck valve 2-69-579

. The connection enhances and expedites the chemical decontamination process since valve disassembly willno longer by required to install the decontamination equipment.

Chemical decontamination ofthe RWCU system has been shown to significantly reduce radiation dose.

The decontamination connection was designed and installed to the design criteria and general engineering specifications which ensures that the integrity ofthe RWCU system is maintained over the expected and unexpected range oftransient and accident conditions which might be experienced, including a design basis earthquake (DBE). Therefore, this change had no adverse efFect on the RWCU system or any other system.

As the result ofthis modification, UFSAR Section 4.9, Figures 4.9-1 and 4.9-3, and Section 7.3, Table 7.3-2 (Sheet 2) and Figure 7.3-2d required updating.

AUnit 2 Technical Specification change is required to Section 3.2/4.2 and 3.7/4.7.

The changes include adding the four temperature monitors to the RWCU heat exchanger room pipe chase area for HELB detection and mitigation.

This change was acceptable from a nuclear safety standpoint.

No unreviewed safety question was involved.

DCN W2002$ - Turbine Floor Valve Hoists - Unit3 (SEBFDCN920212)

Descri tion/Safet Evaluation This DCN installed a pivot beam and hoist above combined intercept valve (CIV)-3A2and relief valve 3-1-574, a pivot beam and hoist above CIV-3B2 (3-FCV-1-111) and reliefvalve 3-1-580, and a pivot beam and hoist above CIV-3C2 (3-FCV-1-114) and reliefvalve 3-1-585 on the east side ofthe turbines located on elevation 617'f the turbine building in Unit 3.

The cranes are bolted to existing rigid frames in the turbine building along the T-16 column line.

The cranes are designed to pick up the'CIVs and the reliefvalves and either place them in a location where they can be serviced locally, or connected to the overhead crane for movement through the building for service.

This modification allows the overhead cranes access to valves without having to modify the safety trips or manually manipulate the hoists to pick up the valves.

0

Tennessee ValleyAuthori ty

SUMMARY

OF Browns Ferry Nuclear Plant SAFETY EVALUATIONS FOR 1993 Annual Operatic Report FIELD COMPLETED PLANTMODIFICATIONS The cranes are vendor supplied and are bolted to the rigid frames in the turbine building (the brackets for the cranes are also vendor supplied).

The cranes are jib cranes, have a rotation of approximately 180', an effective beam length of21', and a rated capacity of 16,000 lbs. The hoists are manually operated units.

The existing rigid frames in the turbine building have been evaluated for 125% ofthe rated load ofthe crane.

Stiffeners were added to the existing rigid frames due to these additional loads.

Due to high radiation levels in the area where the cranes were installed, this modification was

'mplemented while the turbines were not operating.

The installation ofthe pivot beams and hoists have no effect on the operation ofany safety-related system since the pivot beams willonly be used when there is no flowthrough the turbines and then only on nonsafety-related main steam components.

The hoists are manual units, so there is no effect on any electrical or control systems which may have a safety-related function.

Their attachment to the rigid frames in the turbine building (not a safety-related structure) has been evaluated and the rigid frames stiffened accordingly.

No Technical Specification change was required.

UFSAR Figures 1.6-12 and 1.6-16 were affected by this change.

This change was acceptable from a nuclear safety standpoint and no unreviewed safety'question was involved.

DCN W20162 - Disabling ofRoti Sequence Control Systent-(RSCS) - Unit 2 (SEBFDCN920217 RO)

Descri tion/Safet Evaluation This modification disabled the RSCS function for Unit 2. The RSCS is a part ofthe reactor manual control system which is a subsystem ofthe CRD system.

The RSCS is a system that complements and acts as a backup to the rod worth minimizer (RWM) in imposing restrictions on control rod movements.

Such restrictions are imposed to reduce the consequences ofa postulated rod drop accident.

The scope ofthis DCN was to deenergize and remove the RSCS and associated switches from Panel 2-9-27 and 2-9-5, deenergize the Group Notch control system in Panel 2-9-28, and disable to rod select panel dim backlighting.

Tennessee ValleyAulliorily

SUMMARY

OF Browns Ferry Nuclenr Plant SAFETYEVALUATIONSFOR 1993Annonl OpernIIngIIeporl FIELD COMPLETED PLANTMODIFICATIONS This change for BWRs (use ofa single channel rod pattern control system) is described by GE in Amendment 17 ofits NEDE-240-P-A report. NRC has reviewed this amendment and written a safety evaluation.

The evaluation concludes that it is acceptable to remove Technical Specification requiremerit for RSCS for BWR 4 and 5 reactors with either the Group Notch or BPWS RSCS, and to lower the turnoffsetpoint for the RWM to 10% power. It also recommends that rod patterns used for these reactors be at least equivalent to BPWS patterns.

BPWS control is added to Unit 2 RWM by computer replacement DCN W15366A.

'echnical Specification amendment request TVA-BFN-TS-310 eliminated all requirements for-the RSCS and decreased the RWM power level cutofffrom 20% to 10% rated power for Units 1, 2, and 3. This amendment request was incorporated prior to rod withdrawal for startup.

This modification affected UFSAR Figures 3.4-8a, 7.7-1a, 7.7.-1b, and UFSAR text contained in Sections 1.2, 7.0, and Appendices 7.7A and 7.7B.

Compliance with Technical Specifications during implementation ofthis DCN were adequate to assure nuclear safety. No unreviewed safety question is involved.

DCN W20313 - Jet Pang> Riser Brace Clan>p - Unit 3 (SEBFDCN930042 RO)

Descri tion/Safet Evaluation Jet pump riser braces are made ofstainless steel and provide structural support at the upper part.

ofthe jet pump assembly.

The design ofthe riser brace allows for relative vertical movement

'etween the jet pump riser pipe and the reactor pressure vessel (RPV) due to differential thermal expansion ofthe reactor vessel and jet pump riser materials.

The riser brace also provides lateral support to maintain jet pump alignment and structural integrity during operations.

Riser braces consist oftwo layers ofhorizontal riser brace leaves.

At one end these attach to the reactor vessel wall on each side ofthejet pump riser pipe. At the other end they attach to a yoke which facilitates fit-up to the riser pipe. All attachments are welded connections.

This attachment is a two layer design which was selected to eliminate crevices and reduce structural discontinuities.

During the in-vessel visual inspection ofUnit 3, crack indications were identified at the two attachment welds ofthe riser brace to the riser pipe adjacent to jet pump number 5 at reactor vessel 90'zimuth.

A positive locking mechanism was installed between the riser brace yoke and the riser pipe to capture the riser brace.

This locking mechanism performs the same function as the deficient weld attachment.

The locking mechanism uses threaded fasteners,.locked with tack welds, and allows future monitoring ofthe existing cracks.

With the repair clamp installed, the jet pump riser brace

Tennessee ValleyAuthority

SUMMARY

OF Brogans Ferry iVuelear Plant SAFETYEVALUATIOiVSFOR l993 Annual Operating Report FIELD COMPLETED PLANTMODIFICATIONS

"'""~'""'~~K~~""""'"~c'l"'":~"'"""''""".'" """"'""M'"'-'":3 continues to perform the same safety function as the original welded jet pump riser brace. In the event offailure ofthe welded joint, the installed clamp willaccommodate all loads normally carried by the welded joint.

No Technical Specification change was required.

UFSAR figures, graphs, and tables do not include the details which would have been affected by the installation ofthe clamp. Installation ofthe mechanical clamp does not change the operating system or function ofthe system as described in the UFSAR. Although the specific description ofhow the jet pumps are supported is not included in the UFSAR, a safety evaluation was required because the existing indication on the jet pump riser brace willnot be repaired, and a new device which serves as a redundant support for this brace was installed.

This modification was acceptable from a nuclear safety standpoint and no unreviewed safety question was involved.

DCN W20462 - Rttiu CooIing Water (RCg Puntp 3D Control CircuitMotlificntion-Unit 3 (SEBFDCN930015 RO)

Descri tion/Safet Evaluation This modification modified Unit 3 RCW pump 3D control circuit by the inclusion ofauto-start permissive contacts from the normal (offsite power) feeder breaker and from a preaccident signal relay.

This modification corrected a discrepancy between existing installed design and a system design criteria requirement.

The design criteria requires that the RCW pumps not automatically start during either a loss ofoffsite power (LOP) or a design basis accident (DBA) condition.

Similarly, the UFSAR has a requirement to not allow an auto-start ofthe RCW pumps during a LOP. Condition Adverse to Quality Report (CAQR) BFP910008 identified the failure ofthe existing plant configuration to comply with this UFSAR requirement.

The intent ofthis requirement is to ensure successful load sequencing on the EDGs during LOP by inhibiting an auto-start ofthe nonsafety-related RCW pumps.

Additionally, when offsite power is available during a DBA, the safety-related electrical distribution system bus voltage levels willnot be adversely impacted by the auto-start ofthe RCW pumps. Ofthe Unit 3 RCW pumps, only 3D did not satisfy the above design requirement.

The RCW pumps are not safety related.

They provide cooling for main turbine associated equipment and the reactor building closed cooling water (RBCCW) heat exchangers during normal, nonaccident conditions.

Post-accident the emergency equipment cooling water (EECW) system performs these functions and the RCW pumps can be manually restarted as required.

Tennessee ValleyAulhoriiy

SUMMARY

OF Browns Ferry Nuclear Plant SAFETYEVALUATIONS FOR 1993 Annual Operaling Report FIELD COMPLETED PLANTMODIFICATIONS This modification installed new control wiring, qualified in accordance with Institute ofElectrical and Electronic Engineers (IEEE)-383, in 4KV shutdown board 3EC. It did not install new cable nor reroute existing cable.

This change is acceptable from a nuclear safety standpoint.

This modification required a revision to UFSAR Section 10.7.

No Technical Specification change was required and no unreviewed safety question was involved.

DCN$206$6-Design Docnnientation Revision to EliniinateNeerlfor Cable Protection for Appendix R Events in Listerl Fire Zones - Unit 2 (SEBFDCN930006 RO)

Descri tion/Safet Evaluation Thermo-Lag 330 fire barrier material, manufactured by Thermal Science, Inc., which is utilized at BFN as a 10CFR50 Appendix R fire barrier material, has failed to demonstrate that it is capable ofprotecting cables from fire damage and is no longer acceptable (NRC Bulletin 92-01).

For compensation ofthis issue, fire watches have been posted at various locations throughout the plant.

DCN S20656 was issued to remove the requirement for the following cables to be protected for Appendix R events in the listed fire zones.

This change did not involved any physical modifications ofthe plant.

Tennessee ValleyAulhority

SUMMARY

OF Broils Ferry Nuclenr Plont SAFETYEVALUATIONSFOR 1993 Annunl Opernting Report FIEI.D COMPLETED PLANTMODIFICATIONS

~ooduit 2V1742 2ES173-IS 1

2ES 141-I 2ES215-I 2ES141-I 2ES 141-I 2PL5250-II Cable 2V2225 2ES80-I 2ES140-I 2ES214-I 2ES769-I 2ES55-IS1 2PL5260-II Fire Zone 2-3 2-3 2-3 2-3 2-3 2-3 2-4 4

Appendix R Affected Com onent 2-FCV-74-108 normal power supply 2-PCV-1-18 normal power supply 2-PCV-1-42 normal

'ower supply Unit 2 HPCI logic Bus I power supply Core spray logic power for 2-FCV-74-53 and 2-FCV-74-67 2-PCV-1-22 alternate power supply Unit 2 panel 9-9 IAC Bus B normal power supply (2-PNLA-009-0009-3)

No Technical Specification change was required.

The BFN Fire Protection Report, Volume 1, Unit 2 Appendix R Safe Shutdown Program, required revision in support ofthis design documentation change.

UFSAR Section 10.11 states that the Fire Protection Report, Volume 1 is the licensing basis for BFN's Fire Protection Program and changes to it are subject to the requirements of 10CFR50.59.

This change was acceptable from a nuclear safety standpoint and no unreviewed safety question

-was involved.

DCNS21532 - RWMLoiu Poiuer Sefpoint Cltnttge - Uttir2 (SEBFDCN930005 RO)

Descri tion/Safet Evaluation This DCN changed the RWM low power setpoint for Unit 2. The RWM is a part ofthe reactor manual control system which is a subsystem ofthe CRD system.

The RWM system restricts rod movement to a certain predetermined sequence ofwithdrawal and insertion below the RWM low power setpoint.

Movement restriction limits the maximum reactivity worth ofindividual control rods and the rate with which reactivity occurs within the system.

-4 l-

Tennessee ValleyAulhorily

SUMMARY

OF Broivns Ferry Nuelenr Plant SAFETYEVALUATIONSFOR 1993 Annnol Operolinh Eeporl FIELD COhIPLETED PLANT MODIFICATIONS This change for BWRs (reduction in RWM low power setpoint from 20% to 10%) is addressed by GE in Amendment 17 ofits NEDE-24011-P-A. NRC has reviewed this amendment and written a safety evaluation.

The evaluation concludes that it is acceptable to remove Technical

'pecification requirements for RSCS for BWR 4 and 5 reactors with either the Group Notch or BPWS RSCS, and to lower the turnoffsetpoint for the (required) RWMto 10% power.

This change was not safety related. It reduces and facilitates operator's action for startup, shutdown, and emergency conditions.

A Technical Specification change was required for this DCN (TVA-BFN-TS-310). This Technical Specification change eliminated all requirements for the RSCS and decreased the RWM power level cutofffrom 20% to 10% rated power for Units 1, 2, and 3. Elimination of Unit 2 RSCS function was by DCN W20162.

This change did not adversely impact nuclear safety and no unreviewed safety question was involved.

DCN$21 664-HPCI Floiv Diagrani anti Meclsain'cal Control Diagram Revisions - Unit 2 (SEBFDCN930039 RO)

Descri tion/Safet Evaluation This DCN revised HPCI flow diagram 2-47E812-1 and mechanical control diagram 2-47E610-73-1 to show the gland seal condenser (GSC) condensate pump recirculation valve 2-73-621 normally closed.

Past performances ofHPCI surveillance tests have resulted in GSC high level alarms and the HPCI GSC blower tripping due to overflowing condensate.

This blower is not designed to pump water and the thermal overloads ultimately trip power to the blower during overflow conditions.

Troubleshooting activities associated with these occurrences identified misadjustment ofvalve 2-73-621 as being the cause ofthe GSC overflows. The subject valve was positioned either fullyopen or partially throttled during these occurrences.

Steam leakage from the HPCI turbine gland seals is piped to the HPCI GSC.

Condensate from the condenser is drained into the GSC hotwell. A level switch automatically starts the GSC condensate pump when the level in the hotwell rises above the high level setpoint.

When the hotwell level has been reduced to the low level trip setpoint, the GSC condensate pump is automatically shutoff Drainage from the hotwell is pumped through the cooling water return line back to the HPCI booster pump suction line (during standby operation it is pumped into a clean radwaste system drain). The GSC blower is automatically turned on at HPCI initiation.

The blower pumps noncondensible gases extracted from the gland seal leakage to the SGTS.

GSC operability is not required for the HPCI system to perform its post-accident mitigation function.

Tennessee ValleyAuthori ty

SUMMARY

OF Browns Ferry Nuclear Plant SAFETYEVALUATIONSFOR 1993 Annual Operating Report FIEI.D COMPI.ETED PLANTMODIFICATIONS The subject valve is located in the 1.25" condensate recirculation line back to the GSC. A review ofthe design documentation and vendor information for the HPCI GSC indicated that this valve's function is to provide control ofthe operating level ofthe condensate in the GSC hotwell. Control ofthe hotwell level is achieved utilizing this valve to adjust the condensate pumpdown rate. The'required pumpdown rate willvary with the operating conditions.

The BFN Unit 2 Technical Specifications do address HPCI system availability and surveillance requirements in Sections 3.5.E/4.5.E.

However, neither the position ofvalve 2-73-621 nor the HPCI GSC itselfare discussed in the Technical Specifications.

Therefore, no Technical Specification change was required.

This documentation only change did affect UFSAR Figures 6.4-1, 7.4-1a, and 7.4-1b Revising the Unit 2 HPCI flow and mechanical control diagrams to depict valve 2-73-621 normally closed did not have an adverse affect on nuclear safety and did not involve an unreviewed safety question.

DCN W21933 - Replacetttent ofReact'or Feetlivater (RFQ Cont'rol Systent Master Level Controller - Unit 2 (SEBFDCN930011 RO)

Descri tion/Safet Evaluation This modification impacted the RFW control system.

The purpose ofthe RFW control system is to maintain the water level in the reactor vessel within a programmed range during all modes of plant operation to ensure that excessive moisture carryover and steam carryunder are not experienced.

Abnormal operation or failure ofthe RFW control system, while not a nuclear safety issue, has a high probability ofcausing a plant scram, with the subsequent reduction of plant availability and revenue.

During Unit 2 Cycle 6 operation, two reactor scrams were caused by the RFW control system master controller, LIC-46-5. One scram was directly caused by the master controller output failing low, causing a scram on low water level, and the second scram was caused by troubleshooting activities that were investigating abnormal operation ofthe master controller.

The existing master level controller was a GEMAC 540 instrument, which had water level as an input and provided a demand signal output to the three reactor feedpump turbine manual/bias control stations.

The existing water level input was switchable between level transmitter LT-3-53 and LT-3-60 reactor vessel level loops, and could input directly in single element mode or be summed with a steam flow/feed flow mismatch signal in three element mode.

The problems that BFN (and other GE plants ofsimilar age) had experienced with these type of 43-

Tennessee ValleyAuthority

SUMMARY

OF Broivns Ferry Nuclear Plant SAFETYEVALUATIONSFOR 1993 Annual 0 crating RePort FIELD COMPLETED PLANT MODIFICATIONS controllers were due to the age ofthe controllers (approximately 20 years), the vintage ofthe design (mid 1960s), and the general obsolescence ofthe GEMAC line.

To alleviate these problems, this modification replaced the existing GEMAC master level controller with a microprocessor-based single loop controller. To take advantage ofthe additional functionality ofthe Yokogawa controller, and to provide a measure offault tolerance, the LT-3-206 reactor vessel level loop signal was input to the controller, in addition to the existing reactor vessel level input. The controller was programmed to fail over to single element control using the LT-3-206 input upon failure ofthe normal input, and willfail to manual control at the last valid output upon failure ofboth level inputs. Manual action is required to return the controller to the normal input.

Two annunciator windows were added to XA-55-6C in Panel 9-6 to alert the unit operator to the fail over ofthe controller to the L-3-206 reactor vessel level loop and to a catastrophic failure of the controller that causes a fail over to manual control. A+24VDC power supply, a fuse, and two interposing relays were added in Panel 906 to enable the transistor switch output ofthe Yokogawa controller to drive the annunciator windows.

This modification required that UFSAR Figure 7.10-1 and UFSAR Sections 7.10.3.1 and 7.10.3.4 be revised to reflect these changes.

No Technical Specification changes were required.

This modification was acceptable from a nuclear safety standpoint and no unreviewed safety question was involved.

DCN W22$68-Retnoval ofHantlsiviteltes for 2-FCV-01-56 an(1 2-FCV-71 Unit 2 (SEBFDCN930009 RO)

Descri tion/Safet Evaluation This modification removed the local maintenance handswitches for valves 2-FCV-01-56 (main steam line drain outboard isolation valve) and 2-FCV-71-03 (reactor core isolation cooling

[RCIC] steam supply outboard isolation valve). These switches were for maintenance and testing activities, which can be accomplished without local handswitches.

Therefore, since these switches served no function during unit operation or accident mitigation, equipment reliability was enhanced by the removal ofunnecessary circuit components.

The local handswitches were 2-HS-01-56B and 2-HS-71-03B and were located in the Unit 2 steam tunnel.

These handswitches were removed, including their associated indicating lights and cables/conduits from the handswitches to the junction boxes located nearby (approximately 5'). The splice located in the local junction boxes were removed and new splices installed reconfiguring the cabling from 250V RMOV board 2B to the valves.

The junction box openings resulting from the removal of the conduits were sealed and the junction boxes closed, returning them to their original condition. This resulted in each valve's cabling to the 250V RMOVboard having spare A4-

Tennessee ValleyAuthority

SUMMARY

OF Browns Ferry Nuclear Plant SAFETYEVALUATIONSFOR 1993 Annual Opera(ing Report FIELD COMPLETED PLAJVTMODIFICATIOiVS conductors.

This modification was completed by determinating the spare conductors and changing the circuitry at 250V RMOV board 2B. This resulted in removal ofthe valve's maintenance switches, but the valve's operational logic for operator commands and accident mitigation logic was unchanged.

Both ofthe motor operated valves (MOV) affected by this modification are primary containment isolation valves. However, implementation ofthis modification did not change the valves operational logic for operator initiated commands or accident mitigation logic commands.

The removal ofthese handswitches do not affect the valves capability to perform their safety-related functions ofproviding primary containment isolation when required.

Implementation ofthis modification enhanced primary containment integrity by providing circuitry which is more reliable by eliminating unnecessary circuit components.

Thus this modification did not reduce nuclear safety.

This modification required revision ofUFSAR Figures 4.7-1b, 4.7-2a, and 7.3-2f.

No Technical Specification change was required and no unreviewed safety question was involved.

DCN W23238 - Replacenient ofMninSteani Line Drain Valves - Unit 2 (SEBFDCN930014 RO)

Descri tion/Safet Evaluation This modification replaced main steam drain valves 2-FCV-1-168, -169, -170, and -171 with a more reliable valve that is more capable ofwithstanding the extreme conditions.

Specifically, the 600 8 Hancock Model 5500W valves (2" carbon steel globe valves) were replaced with 1700 8 Anchor/Darling Series 700 2", cast carbon steel, double disc, gate valves.

The existing valves were damaged due to excessive vibration in the drain lines. The Limitorque SMB-000 motor operators installed on these valves had also experienced substantial damage, primarily to the limit switches and terminal blocks inside the limitswitch compartments ofthe operators.

This modification provided new Limitorque SMB-000 motor operators which are compatible with the new Anchor/Darling valves. This change also provided thermal overloads sized for use with the new motor operators and repaired or replaced damaged cable and flexible conduit.

This change also revised Unit 2 main steam mechanical control diagram 2-47E610-1-1 to correctly depict the subject drain valves as normally open as shown on main steam fiow diagram

'-47E801-1.

The positions ofthe valves were not physically changed.

Only a drawing error correction was made.

During normal power operation with turbine revolutions per minute (RPMs) > 1700, these drain valves are open.

Schematic diagram 0-45E777-15 indicates that all four ofthese valves willclose when any one ofthe eight main steam isolation valves (MSIV) closes and the turbine speed is > 1700 RPM. This design feature ensures that all four main steam 45-

~,

Tennessee ValleyAutltority ~

SUMMARY

OF Broils Ferry Nuclear Plant SAFETYEVALUATIOlVSFOR l993 Annual Operating Report FIELD COMPLETED PLAlVTMODIFICATIOIVS lines willhave individually orificed drain lines ifone main steam line isolates with the turbine still on line. The individual orifices prevent condensate backflow into the isolated main steam line due to the high'er pressures in the other lines.

UFSAR Section 4.11 describes in general the design ofthe main steam line drainage features.

This change did not affect this text. UFSAR Figure 11.1-1A required changing to depict the subject main steam drain valves as gate valves instead ofglobe valves.

Technical Specifications do not address the affected main steam drain valves or provide requirements for their operability. This change had no effect on the performance ofthe main steam system.

Therefore, no Technical Specification change was required.

This modification did not have an adverse affect on nuclear safety and did not involve an unreviewed safety question.

DCN T23972 - ControI Bay Spreatling Rooni AirFloiv Rates - Units 1, 2, 3 (SEBFDCN930021 RO)

Descri tion/Safet Evaluation Technical Instruction, O-TI-35, Test Deficiency No. 2, identified control bay spreading room air flowrates outside the specified design flow range. DCN T23972 was issued to reduce the spreading room design supply air flow rate from 5900 cubic feet per minute (CFM) to 5000 CFM and the design exhaust air flow rate from 12,100 CFM to 10,000 CFM based on analysis in calculation MD-N0031-930037.

The spreading room supply fan motor pulleys were replaced with variable pitch sheaves in order to adjust fan speeds to achieve the new design air flow rates.

The control bay spreading room ventilation system consists ofone 100% capacity supply fan-serving the Unit 1 and 2 (combined) spreading room and one 100% capacity supply fan serving the Unit 3 spreading room. Two, redundant, 100% capacity exhaust fans serve both spreading rooms through a common exhaust duct. Both supply fans and one exhaust fan are normally operated with the supply fans drawing on outside air and the exhaust fan discharging to the control building ventilation stack.

The spreading room fans are not safety related and do not provide any safety related function. The spreading room air flow must be balanced with the exhaust flow exceeding the total supply flowin order to prevent a positive spreading room pressure in relation to the CBHZ. This action is necessary in order to preclude inleakage of unfiltered, potentially radioactive spreading room air into the CBHZ followingan event involving radioactive release to the environment.

The functional description ofthe spreading room fans in UFSAR Section 10.12.5.3 was revised to describe the air flow'balancing technique used to keep the spreading rooms at a negative

Tennessee ValleyAuthority

SUMMARY

OF Browns Ferry Nuclear Plan!

SAFETYEVALUATIONS FOR l993 Annual Operating Report FIELD COMPLETED PLANT MODIFICATIONS pressure relative to the CBHZ. UFSAR Figure 10.12-2b was revised to show the new spreading room design air flow rates.

The cable spreading room HVACis a nonsafety-related system and is not addressed in the Technical Specifications or its bases.

Therefore, no change was required to the Technical Specifications.

This change was acceptable from a nuclear safety standpoint and no unreviewed safety question was involved.

DCN$24344-Unique Itlentifier(UNII))Nuntber Reassigntnent - Unit 0 (SEBFDCN930025 RO)

Descri tion/Safet Evaluation DCN W19165 installed a 3" plug valve in the radwaste building floor drain piping to comply with maximum permissible flood calculation requirements.

The UNID number shutoff valve 0-SHV-77-613 was assigned to this valve, however, this UNID number conflicted with existing 3/4" vent valves (VTV)that were already assigned number 1,2,3-VTV-77-613. DCN S24344 was issued to change the floor drain valve UNID number to 0-SHV-77-680. The design drawing was revised to show the new valve number assignment.

This is a documentation change only.

The plant design basis was not altered and no physical work was performed on the plant.

No Technical Specification change was required.

UFSAR Figures 9.2-3a and 9.3-1b required updating as a result ofthis change.

This change was an administrative drawing change only. However, the radwaste system and UFSAR figures were involved, therefore, a safety evaluation was required.

Nuclear safety was not affected and no unreviewed safety question was involved.

DCN T24470- Renioval ofSGTS Ahantlonetl Duet - UnitI (SEBFDCN930026 RO)

Descri tion/Safet Evaluation This modification removed abandoned duct from the SGTS header to the Unit 1 refuel floor zone. Allequipment associated with this duct was removed and the floor penetrations sealed.

In accordance with design criteria documents, licensing commitments, and the Nuclear Performance Plan, the duct had to be removed before startup following the Unit 2 Cycle 6 outage.

The SGTS safety actions required for safe shutdown due to transients, accidents, and special events do not Tennessee ValleyAuthority

SUMMARY

OF Browns Ferry Nuclear Plant SAFETYEVALUATIONS FOR 1993 Annual Operating Report FIELD COMPLETED PLANTMODIFICATIONS

, gg~~~~'"':" " ':::-": '""'"':"',"':"" ""'""'-'" "-"4""'::: "":i"~"""~:-'""'"""."'. i~@":':' M "~'~a~KC require the operation ofthis duct. The duct was no longer required since a dedicated path through dampers FCO-66-44 and FCO-64-45 was available to direct the refuel zone exhaust to the SGTS.

The safety function ofthe duct was to maintain pressure boundary integrity during system operation.

Based on the fact that the duct was not required to provide a flowpath before, during, or following an accident, the duct was permanently removed and the header blanked off Technical Specifications do not list or discuss the duct. Therefore, no Technical Specification change was required.

A description ofthe duct removed is not presented within the UFSAR's text, tables, or graphs.

However, the existence ofthe operators was reflected on Figures 5.3-3a and a revision to this figure was required.

The removal ofthe duct did not acct the ability ofthe SGTS to perform its function.

Therefore, the margin ofsafety as defined in the basis for any Technical Specification was not reduced.

No unreviewed safety question was involved.

DCN$24484-RCWSystent Pressttre Relief Valve Setting - Unit 0 (SEBFD CN93027 RO)

Descri tion/Safet Evaluation This change revised drawing 1-47E844-1 to add a new pressure reliefvalve setting for valves 0-24-1058 and -1059 for the RCW system.

The original setting was 60 PSIG which was very close to the operating pressure causing the valve to open.

The new setting is 72 PSIG which allows enough margin between the pressure reliefvalve setting and normal operating pressure to prevent unnecessary pressure reliefvalve discharge.

This change involved the nonsafety-related portion ofthe RCW system.

No Technical Specification change was required.

UFSAR Figure 10.7-1a required updating to reflect the new setpoint change.

This change was acceptable from a nuclear safety standpoint and no unreviewed safety question was involved.

II Tennessee ValleyAuthority

SUMMARY

OF Browns Ferry Nucleor Plnnt SAFETYEVALUATIONSFOR I993 Annunl Opernting Report FIELD COMPLETED PLANTMODIFICATIONS

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DCN T24611 - Relocation ofControls/Inrlications Associaterl ivitltPCIS Circuitry-Unit 2 (SEBFDCN930031 RO)

Descri tion/Safet Evaluation This modification added contacts from PCIS isolation relay 86-76-91B (92B) to the isolation logic forH202 analyzer A (B) isolation valves which are part ofPCIS Group 6. In addition, this DCN added a dropping resistor to Unit 2 offgas flow instrument 66-20 to provide a voltage input to flowrecorder 2-FR-66-20.

This DCN reconfigured flow controller 2-FC-2-29 to provide a meter output consistent with the function ofmeters for other controllers.

Repairs were made to a panel opening created by previous modifications. Indicating lights which are part of the PCIS valve mimic mounted on Panel 2-9-3 were rearranged.

This modification also allowed minor wiring termination detail changes which did not result in functional changes to a schematic.

Allmodifications performed by this DCN were implemented in the control bay which is not a radiological control area.

The affected control room panels are seismic Class I. Allrepairs performed by this DCN utilized standard repair techniques implemented throughout the control room to ensure that the modifications met seismic qualification requirements.

A resistor was added to Unit 2 ofrgas flow instrument 66-20 to convert the current input to a voltage which may be monitored by recorder 2-FR-66-20. No other radwaste system changes were performed by this modification.

No Technical Specification changes were required as a result ofthis modification. No UFSAR text or figures required updating.

This modification does not reduce nuclear safety and no unreviewed safety question was involved.

DCN T24667-SGT BuiklingDrain Sutnps Motlification-Unit 0 (SEBFD CN930032 RO)

Descri tion/Safet Evaluation This DCN provided minor modifications to the SGT building drain sumps in order to close and seal them air tight. A weld neck flange and cover plate was welded to the 8" float well standpipe on the SGT building No.

1 sump.

(The No. 2 building sump float well was already equipped with a flange and cover plate.) The 3/4" makeup water fillline penetration on each sump cover plate was sealed with a bolted, split ring cover plate.

Unused holes in the sumps were plugged and sealed.

All sump manhole covers, float well covers, and pipe seals were caulked with RTV adhesive sealant to make the sumps air tight.

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Tennessee ValleyAuthority

SUMMARY

OF Browns Ferry Nuclear Plant SAFETYEVALUATIONSFOR I993Annual Operating Report FIEID COMPLETED PLANTMODIFICATIONS KU~kekw.

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'v'"'his action was taken to eliminate secondary containment bypass air leakage paths through these sumps into the SGTS main air supply duct. This arrangement allows these radioactive drain sumps to be purged and vented through the SGT filtertrains thereby preventing possible airborne contamination conditions in the SGT building. The SGT building is not located within the secondary containment boundary, so airborne contamination conditions are to be avoided in this building as this could result in unmonitored and unfiltered radioactive releases to the environment.

These changes brought the sumps into better compliance with UFSAR Section 10.16.4.2 which describes radioactive drainage sumps as being "closed".

This modification did not alter the function, operation, or test requirements ofthe SGTS as described in the Technical Specifications or its bases.

Sealing the SGT building drain sumps reduces secondary containment air bypass leakage thereby improving the flow margin of Technical Specification related SGTS air flow testing.

Also, the function and operation ofthe radwaste system's SGT building drain sumps were not affected by this change.

Thus, no change was required to the Technical Specifications.

Sealing the SGT building drain sumps had no adverse impact on nuclear safety and no unreviewed safety question was involved.

DC1V$249ZZ - E/iniijiation ofbfanaal Opertttor ActionforHSWVOperation - Unit 0 (SEBFDCN930036 RO)

Descri tion/Safet Evaluation During the conduct ofUnit 2 restart test procedure 2-BFN-RTP-065, flowfrom the offgas and SGTS was found to produce a positive pressure in the stack plenum which could cause backflow through certain components and result in ground level radioactive releases.

This invalidated the original assumption ofzero ground level release as was stated in UFSAR Chapter 14 analysis.

This test demonstrated that the drain line from the offgas stack plenum and from above the stack roofdeck would be pressurized by system fans with radioactive gases which could backflow through these lines. These gases were postulated to escape either from-a break in the drain piping in the stack, yard, or radwaste building or by failure ofthe nonsafety-related sumps (i.e.;,

loss or water seal) to which the drain lines exit. Consequently it was postulated that ground level releases could occur which would result in post-accident control room doses in excess oflimits.

In order to alleviate this problem, DCN W14099 installed a locked closed radwaste valve (0-77-2284) in the 3" stack drain line above the 599'-6" elevation in the stack building. The DCN also seismically qualified (seismic Class I) the 3" stack drain from the stack plenum and stack roofdeck through the locked closed valve to eliminate potential ground level radioactive releases from the stack through this line.

DCNs W17491 and W17337 and associated system design criteria designed and installed the HWWVfor Unit 2. A similar HWWVmodification willbe required due to a commitment to the Tennessee VolleyAuthority

SUMMARY

OF Broils Ferry Nuclear Plnnl SAFETYEVALUATIONSFOR l993 Annual 0 erotic Report FIELD COMPLETED PLANTMODIFICATIONS NRC for Units 1 and 3 prior to their operation.

The modification as installed for Unit 2 required a manual operator action to open the stack drain valve (0-77-2284) prior to HWWVoperation to prevent the HWWVcondensation from collapsing the stack roofdeck. This manual operator action is not preferred since HWWVoperation willoccur in accordance with the BFN symptom based emergency operating instructions in which no allowable time frame for this manual action can be specified nor can the stack environment for operator access be guaranteed in the event of design basis or beyond design basis accidents.

Therefore, DCN S24922 eliminates the manual operator action for HWWVoperation for any unit by opening the stack drain valve. Potential ground level releases through the 3" stack drain are prevented by (1) qualifying the exposed piping in the stack and the radwaste building as seismic Class IIpressure boundary retention, (2) evaluating the potential for release from the piping embedded in seismic Class I concrete in the stack, (3) demonstrating that the water seal where the stack drain terminates underwater in the offgas condensate sump in the radwaste building and in the stack sump in the basement ofthe stack remain, and (4) evaluating the potential for radioactive gases to be released from the buried portion ofthe stack drain.

Technical Specification Section 3.7 describes the primary and secondary containment systems but places no requirements or restrictions on operation ofthe stack drain. Section 3.8 describes the radioactive waste systems but places no requirements or restrictions on operation ofthe stack drain. No change to the Technical Specification were required.

This change was acceptable from a nuclear safety standpoint and no unreviewed safety question was involved.

DCN W25957-Connnttnications Systettt Upgratje - Unit 0 (SEBFDCN930049 RO)

Descri tion/Safet Evaluation This modification provided additional desk consoles for the plant radio communication system and added encryption capability for portions ofthe radio communication system.

This modification installed one expansion chassis on each oftwo existing Zetron radio controllers which are located in the communications room. New desktop consoles (Zetron Model 4016) located in the Operations Support Center, plant simulator, and shift operations supervisor (SOS) station in the simulator room were connected to the new expansion chassis.

The radio communications system is not addressed in any Technical Specification, nor is the system addressed in the basis ofany Technical Specification for any BFN unit. Therefore, a Technical Specification change was not required.

Tennessee VnlleyAutlrority

SUMMARY

OF Browns Ferry Nucleor Plant SAFETYEVALUATIONS FOR 1993 Annunl Opernting Report FIELD COMPLETED PLANT MODIFICATIONS AP@A@ghgglll heal(VIP hVgW)$$)pggg(VVg+)WPWPgVAgYAV@g,, AP,A +g@VgP$ $

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Section 10.18 ofthe UFSAR provides details ofthe plant communications system.

Specifically, Section 10.18.3.1.3 describes those areas from which the emergency radio system may be accessed.

This DCN expands the number offacilities from which this system may be accessed, thus requiring a change to the UFSAR.

This change was acceptable from a nuclear safety standpoint and no unreviewed safety question was involved.

Tennessee ValleyAuthority

SUMMARY

OF Browns Ferry Nuclear Plant SAFETYEVALUATIONS FOR l993 Annual Operating Report NEfVINSTRUCTIONS/PROCEDURE REVISIONS K:-'"""'":"':::""'"':-"': """-"""""'"""'"'"':""."':;".": i'-":.""."-"'""""..""-'":- '-'-"""""'""-"""""'-

""-'UMMARY OF SAFETY EVALUATIONS FOR NEW INSTRUCTIONS OR PROCEDURE REVISIONS

Tennessee ValleyAuthority

SUMMARY

OF Brains Ferry Nuclear Plant SAFETYEVALUATIONSFOR 1993Annual Operating Report NEfVINSTRUCTIONS/PROCEDUREREVISIONS Pacific Nuclear Procerlure DP-2337 BFNLoiu Oxitlation State Metal Ion (LOMI)Decontaniination ofthe RWCV System - Operating Procetlure Pacific Nuclear Procetlure DP-2337 BFNLOMIDecontamination ofthe, Reactor Recirculation System - RHR Operating Procerlure (68-9301403-RO) escri ti n/Safet Evaluation Pacific Nuclear Procedures DP-2337-1 and DP-2337-2 provide detailed guidance for the use ofdilute chemical decontamination (DCD). BFN willuse DCD to reduce radiation levels in portions ofthe reactor recirculation, RWCU, and RHR systems.

This safety evaluation is limited to concerns with in-process and post-process material compatibility and potential system/component degradation.

Additional aspects ofplant equipment configuration and DCD implementation are covered in a separate evaluation.

The DCD ofthe reactor recirculation, RWCU, and RHR piping systems willbe accomplished using a three step LOMI-alkaline permanganate (AP)-LOMIprocess.

However, only LOMIwill be allowed to circulate through the'RPV annulus and the reactor recirculation loop piping upstream (RPV side) ofvalves 2-FCV-68-1 and 2-FCV-68-77.

The LOMInonregenerative dilute process reduces Fe+3 to Fe+2 in the corrosion film through the use ofthe vanadous ion, V+, as the reducing agent.

AP is an oxidizing step which enhances LOMI(or other) decontaminations ifthe corrosion film contains >10%

chromium. AP consists ofpotassium permanganate and sodium hydroxide to oxidize Cr+ to Cr+6 in the filmfor further removal.

AP is removed by oxalic acid. Destruction ofAP with oxalic acid is usually part ofthe AP step.

No change to the Technical Specifications is required for performance ofthis activity.

During performance ofthis activity, the plant willbe maintained in a configuration which is consistent with Technical Specification requirements.

Although there is no direct impact on system operational characteristics as described in the UFSAR, this process does have some limited potential for indirectly changing system operational characteristics by changing corrosion rates and increasing the likelihood of corrosion induced failure. Therefore, a safety evaluation was prepared.

No unreviewed safety question was involved.

Tennessee ValleyAuthori ty

SUMMARY

OF Browns Ferry Miclear Plant SAFETYEVALUATIONSFOR l993 Annual Operating Report NEfVINSTRUCTIONS/PROCEDURE REVISIONS 5':

'-:::: -"':-:::':::::::--'::-::-:':'i--'"-:-:'i""::-"':-"'-"-'"':::::":--'--"'i- "::- -:-:;-::- i"::::::::-::""':-:-:":::::"=::-:::::i'--::"'l':::-i':::-::-':!ll': "':

Special Operating Instrucii'on 3-SOI-31A-Special Operating Instruction O-SOI Pacific Nuclear Proceilure DP-231$ Pacific Nuclear Proceilure DP-Z31 $-Z-Pacific Nuclear Proceilure DP-231$ (00-9111427)

Descri tion/Safet Evaluation Draindoivn ofUnit3 Reactor Vessel and Cavity to Condensate Storage System Clienrical Decontanunation ofFuel Pool Cooling (FPg Pumps and Heat Zrchangers BFNLOMIDecontamination ofthe R8'CU Systeni - Operating Procedure BFNLOMIDecontamination ofthe Reactor Recirculation System - Operating Procedure Dilute CITROXDecontamination ofthe BFN FPC System This safety evaluation and the associated safety assessment addressed the DCD ofportions ofthe reactor recirculation system, RWCU system, RHR system, and FPC system to reduce radiation levels.

Concerns addressed in this evaluation/assessment were limited to issues ofin-process or post-process material compatibility and potential system/component degradation.

Additional aspects ofplant activities and equipment configuration required to support decontamination were covered in a separate assessment and/or evaluation.

For the reactor recirculation, RWCU, and RHR systems, only Unit 3 piping and components were aA'ected.

For the FPC system, not only portions ofthe Unit 3 system were aAected, but also similar portions ofthe Unit 1 and Unit 2 systems.

The DCD ofthe reactor recirculation, RWCU, and RHR systems willbe accomplished using a three step LOMI-AP-LOMIprocess.

However, only LOMIwillbe allowed to circulate throughout the RPV annulus and the reactor recirculation loop piping upstream (RPV side) ofvalves 3-FCV-6S-1 and 3-FCV-68-77. The DCD ofthe FPC piping willbe accomplished using a CITROX-AP-CITROXor a CITROX (only) process.

The LOMInonregenerative dilute process reduces Fe+3 to Fe+

in the corrosion film through the use ofthe vanadous ion, V+2, as the reducing agent.

AP is an oxidizing step which enhances LOMI(or other) decontaminations ifthe corrosion filmcontains >10%

chromium. AP consists ofpotassium permanganate and sodium hydroxide to oxidize Cr+3 to Cr+6 in the filmfor further removal.

AP is removed by oxalic acid. CITROX is a regenerative dilute decontamination process consisting ofa 0.25% solution with 2:1 ratio ofcitric acid to oxalic acid. CITROX A has no inhibitor while CITROXB has an inhibitor.

Tennessee ValleyAuthority

SUMMARY

OF Broivns Ferry Nuclear Plant SAFETYEVAI.UATIONS FOR 1993Annual Operating Report NElVINSTRUCTIONS/PROCEDUREREVISIONS G -" '"".': ":: -"': " ":-:"-"'::"'.-:- """"'::-'"':

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No change to the Technical Specifications is required for performance ofthis activity.

During performance ofthis activity, the plant willbe maintained in a configuration which is consistent with Technical Specification requirements.

Although there is no direct impact on system operational characteristics as described in the UFSAR, this process does have some limited potential for indirectly changing system operational characteristics by changing corrosion rates and increasing the likelihood of corrosion induced failure. Therefore, a safety evaluation was prepared.

No unreviewed safety question was involved.

SSP-6.10 - ASMESection XIAugnientetl Nontlestracti ve Exaininations 1/3-$I-4.6. G - Inser vice Inspection Program - Units 1 and 3

,2-$I-4.6, G - Inservice Inspection Prograni - Unit 2 (00-9309415 RO)

Descri tion/Safet Evaluation Nuclear Quality Assurance (NQA) Plan 89A, Revision 3, Section 4.1.3.c.7.b.11 specifies that the Site Nuclear Assurance and Licensing Manager is responsible for planning and implementing the ASME Section XI nondestructive examination inspection activities. The revision ofSSP-6.10, 1/3-SI-4.6.G, and 2-SI-4.6.G transfers that responsibility to the Site Engineering and Modifications Manager.

IOCFR50.54.a(3) states that changes may be made to the Quality Assurance Program provided the change does not reduce commitments and that the changes be submitted to the NRC at least annually. This revision was initiated by the consolidation ofall nuclear corporate functions in the new Technical Support organization.

This consolidation was evaluated and determined not to be a reduction in commitment.

Quality Methods Procedure 105.6 defines "Reduction in Commitment" as a change in the nuclear quality assurance program previously accepted by the NRC that diminishes the intent or scope ofthe program potentially permitting deficiencies to arise in the design, fabrication, construction, or operation ofthe facilityresulting in increased risk to the public health and safety.

The NQA Plan specifies the training and certification of inspectors, requires inspection points and acceptance criteria be included in instructions, and requires sufficient authority and organizational freedom ofpersonnel performing quality verifications. Only the organization performing nondestructive examination was changed by this revision. Neither the Site Nuclear Assurance and Licensing Manager nor the Site Engineering and Modifications Manager has direct responsibility or authority for the activity being verified (structural integrity ofASME code classes I, 2, and 3 equivalent

Tennessee ValleyAuthori ty

SUMMARY

OF Browns Ferry Nuclear Plant SAFETYEVALUATIONS FOR 1993 Annual Operating Report SPECIAL OPERA TING CONDITIONS 1993

SUMMARY

OF SAFETY EVALUATIONS FOR SPECIAL OPERATING CONDITIONS

Tennessee ValleyAuthority

SUMMARY

OF Broivns Ferry Nuclear Plant SAFETYEVALUATIONS FOR I993 Annual Operating Report SPECIAL OPERATING CONDITIONS R@k& Y'~~':

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and disperse low-level radioactive liquid waste from the radwaste treatment facilities is not specified in Technical Specifications.

Plant procedures ensure that the Technical Specification requirements for radioactivity concentration after dilution are met.

Therefore, no Technical Specification change is required.

The UFSAR states that interlocks are provided which prevent the discharge ofliquid radwaste into a condenser cooling water discharge conduit when fewer than two ofthe associated circulating water pumps are in operation.

This activity simulated a portion of that interlock. Additionally, the normal source ofwater to dilute and disperse low-level radioactive liquid waste from the radwaste treatment facilities was changed from two CCW pumps to one CCW pump while two CCW pumps were out ofservice.

Standard work control practices commensurate with existing Technical Specifications and special requirements which were specified were adequate to ensure an unreviewed safety question did not exist during this activity.

Operation ofCCEY Cooling Toiuersin Closed Motle ofOperation (SEBFSA930050 RI)

Descri tion/Safet Evaluation This safety evaluation addressed the ability to operate the CCW cooling towers in the closed mode without having the high temperature liApump trip feature operational.

Due to increasing upstream river water temperatures, the ability to operate a single unit near maximum load and not exceed the maximum sustained (24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />) downstream river water temperature had become difficult. The elevated river temperatures were due to a combination oflow river water flow and extremely high () 90'F) air temperatures for an exterided period oftime. Certain river water temperatures willrequire the closed mode of CCW. The maximum inlet water temperature limitfor plant cooling water is 95'F. This maximum temperature intake limitis based on the heat removal capacity ofsafety-related equipment.

This mode recirculates the water discharged from the cooling towers to the CCW intake structure and allows the plant to operate without discharging water into the river. This is possible on a long term basis only ifthe cooling towers can dissipate all the heat discharged from the turbine condenser.

Otherwise, the heat in the forebay would approach the 95'F allowable design limitfor maximum intake water temperature.

This'afety evaluation addressed putting into place appropriate compensatory measures to monito'r and manually trip the cooling tower liftpumps, or return the system to other

Tennessee VolleyAuthority

SUMMARY

OF Bro>vns Ferry Nuclenr Plonl SAFETYEVALUATIOiVSFOR 1993 Annunl Operating Report SPECIAL TESTS

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MS 1993

SUMMARY

OF SAFETY EVALUATIONS FOR SPECIAL TESTS

Tennessee ValleyAuthori ty

SUMMARY

OF Browns Ferry Nuclear Plant SAFETYEVALUATIONSFOR 1993 Annual OperatingReport, SPECIAL TESTS K~X~~%%4m&"~~'+~4~'~~'M4A~~~ '

'ost Modifi etta on Test 2-PMT-BF-066. 004 - Unit2 (66-9303412 RO)

Descri tion/Safet Evaluation This procedure functionally tested modifications (relocation and/or rewiring ofcontrols) made to control room Panel 2-9-8 by DCN W170360 Stage

1. This safety evaluation specifically addressed the testing ofoffgas system related components in the main steam system, offgas system, and radiation monitoring system which are associated with the steam jet air ejector (SJAE) operation and inlet/outlet/drain valves auto-isolation logic.

Functional testing ofthese components with the plant in shutdown or refuel mode required the installation ofjumpers in auxiliary instrument room Panel 2-9-36 to simulate condenser vacuum and steam pressure adequate so that the auto-isolation logic for offgas channels A and B could be reset, thereby enabling the controls for these valves.

The safety evaluation also addressed any necessary liAingofcertain internal wires in Panel 2-9-36 to allow proper verification ofcontact configuration ofcontrol switches 2-HS-1-150 and 2-HS-1-152.

These jumpers and wire lifts constituted a temporary alteration to a radwaste system, thus requiring the safety evaluation.

Appropriate administrative controls were used to assure that the as-designed configuration was maintained.

The function and performance characteristics ofcomponents affected by this test were unchanged.

The temporary alterations installed and removed by this test did not affect normal operational parameters, setpoints, calibration intervals, or functional test intervals nor did they affect any Technical Specifications or their bases.

No Technical Specification change was required.

This test was acceptable from a nuclear safety standpoint and no unreviewed safety question was involved.

2-PMT-BF-066. 006-Unit 2 (66-9301402)

Descri tion/Safet Evaluation This PMT was performed to verify the design functions ofthe affected components were unchanged and electrical faults were not introduced into the associated component's circuitry aAer the installation ofDCN W17368 Stage

1. DCN W17368 consisted of modifications to control room offgas Panel 2-9-53. In general, the modification rearranged and/or replaced control switches, instruments, temperature recorders, and Tennessee VolleyAulhorily

SUMMARY

OF Bro>vns Ferry Nuoleor Plont SAFETYEVALUATIONS FOR 1993 Annuol Operoting Report SPEC/AL TESTS indicating lights. This procedure was performed during the time period when the reactor was in cold shutdown and the offgas system was not required to be operable.

The components modified and tested were nonsafety related and were not required for safe shutdown ofthe plant. The offgas components tested were outside the boundary of the offgas stack and its associated ducting. Therefore, the offgas nuclear safety functions were not compromised or affected.

No Technical Specification changes were required and no unreviewed safety question was involved.

Special Test 0-ST-93 Units I, 2, anil 3 Descri tion/Safet Evaluation The purpose ofthis special test was to perform electromagnetic Interference (EMI) mapping at BFN. Phase one performed specific mapping ofthe Unit 1, 2, and 3 refuel

'floor and control room locations in proximity to the reactor and refuel zone radiation monitors'etectors and drawers.

Subsequent phases willperform EMI mapping of additional areas in all three units to obtain an overall plant EMI profile. Each phase will be detailed in an appendix to this special test with additional appendices added as testing scope and requirements are expanded and defined.

There was no impact on plant systems.

EMI mapping is not intrusive in that no cables or equipment is rendered inoperable during the collection ofdata.

Current probes are clamped over power and signal cables to monitor the levels ofEMI emanating from them while the plant equipment they interface with is in operation.

These probes connect to receiving and recording instruments and do not inject signals into the cables they connect onto. Likewise, oscilloscope probes are attached to selected terminal points and data recorded.

Plant handheld radio transceivers and repeaters are keyed on and offin permissible locations to simulate normal use and determine their effect on monitored plant equipment.

Antennae connected to receiving and recording instruments are positioned at various locations and rotated to map the EMI profile present in each area surveyed while the plant is in operation.

These receiving antennae do not transmit outgoing signals.

This was a special test that gathered data only. It in no way'affected system operational characteristics.

It did not affect compliance with any Technical Specification nor did it conflict with or alter anything contained in the UFSAR. Normal operational alignment of all plant equipment was required for this data to be representative ofthe actual EMI environment by plant instrumentation.

Tennessee VnlleyAuthority

SUMMARY

OF Browns Ferry Nuclenr Plnnt SAFETYEVALUATIONSFOR 1993 Annunl Opernting Report SPECIAL TESTS This special test was acceptable from a nuclear safety standpoint and no unreviewed safety question was involved.

Tennessee ValleyAuthority

SUMMARY

OF Browns Ferry Nuclear Plant SAFETYEVALUATIONS FOR I993 Annual Operating Report UFSAR REVISIONS t

1993

SUMMARY

OF SAFETY EVALUATIONS FOR UFSAR REVISIONS 46-

Tennessee ValleyAuthori ty

SUMMARY

OF Browns Ferry Nuclear Plant SAFETYEVALUATIONSFOR 1993 Annual 0 crating Re ort UFSAR REVISIONS CXCZ:::'-::-::-'::-':-:::::""'::-'--:::::::-:i'-'-'-":'::::"-'::."-::i':::-:i'l:::-:i:"':::--'::;:-"""::"::-'-

'::-:-i':--: "::-":-"::::>:'::-:::.":--

0'"":"'":":>""ig@&i"-'""""~Y'.""""'<"~i~<i~<~O." "$'j': ""s::><x<P~4<'.+if'%~>'i"j<; ~%@%~4~

".@:.@j'<~4+4ÃbFA~< Fire Protection Report (39-9309-016 RO) Descri tion/Safet Evaluation This revision to Fire Protection Report, Volume 1, deletes Sections 9.3.11.D.l.e and 9.3.11.D.l.f which require spreading room A and B carbon dioxide (CO ) systems to be operable whenever equipment protected by the CO2 system is required to be operable. DCNs W17821 and DCN W17822 installed automatic fire suppression systems in cable spreading rooms A and B on elevation'606'n the control bay. These modifications corrected all deviations from NFPA code requirements that existed for the spreader room sprinkler systems. Allwork was completed during the Unit 2 Cycle 6 outage. As a result, the new sprinkler systems provide adequate fire protection for spreader room A and B and the existing CO2 systems are no longer required. Also, testing requirements and compensatory fire protection actions associated with these CO2 systems are no longer required. This revision to the Fire Protection Report was safe from a nuclear safety standpoint, did not require any changes to Technical Specifications, did not require any special tests or experiments, and did not involve any changes to radwaste system. However, this revision did a6'ect procedures described in the UFSAR, and therefore required a safety evaluation. No unreviewed safety question was involved. Fire Protection Report (SEBFSR930063 RO) Descri tion/Safet Evaluation The BFN Fire Protection Report identifies the equipment required to ensure a safe shutdown following an Appendix R event. The Fire Protection Report also identifies the required compensatory measure ifa required piece ofequipment is not able to perform its Appendix R function. Several components and systems identified in the Fire Protection Report are addressed by an existing Technical Specification. Ifthe Technical Specification LCO is more stringent than the Appendix R requirement, then the LCO provides adequate compensatory measures. Therefore, the Technical Specification LCO has been listed as the compensatory measure when applicable. The definition ofcompensatory measures in the Fire Protection Report requires the action associated with the Technical Specification LCO to be taken. This safety evaluation evaluated changing the compensatory measures ofthe Fire Protection Report equipment which have associated Technical Specification Tennessee ValleyAuthority

SUMMARY

OF Browns Ferry Nuclear P/ant SAFETYEVALUATIONS FOR 1993 Annual Operating Report UFSAR REVISIONS LCOs. This change would permit either compensatory measure A (Technical Specification LCO action) or compensatory measure B (a fire watch) for Fire Protection Report equipment which is Technical Specification operable but is not capable of performing its Appendix R function. This change willprevent the Fire Protection Report from requiring entrance into a Technical Specification LCO when a piece ofequipment or system is Technical Specification operable but has been identified as not being capable of performing its Appendix R function.

This compensatory measure change is consistent with Appendix R equipment for which no Technical Specification is applicable.

For that equipment where Technical Specification

- LCOs'do not provide adequate measures to meet the Appendix R analysis, a separate compensatory measure has been established.

The specified measure is intended to assure safe shutdown capability is restored within seven days by either restoring the equipment function or by taking temporary measures to assure equivalent shutdown capability exists.

Equivalent shutdown capability is defined as 1) providing temporary equipment or procedures which willensure the out ofservice equipment function does not affect safe shutdown capability or 2) providing adequate fire watch capability to ensure fires are prevented and/or discovered in a time frame which willassure the out ofservice equipment is not needed to support reactor safe shutdown in case ofa fire. Based on this, this Fire Protection Report change does not adversely affect nuclear safety.

No Technical Specification change is required.

This Fire Protection Report change does affect the process regarding compensatory measures for Appendix R equipment as described in the Fire Protection Report which is part ofthe UFSAR.

This change is acceptable from a nuclear safety standpoint and no unreviewed safety question is involved.

Fire Proteclion Rep ort (SEBFMSAR930024 RO)

Descri tion/Safet Evaluation Fire Protection Report, Volume 1 (Revision 1) was submitted to the NRC in support of Technical Specification Change 306 (Removal ofFire Protection from Technical Specifications).

This Fire Protection Report, Volume 1, since NRC approval, is the plant document containing the Fire Protection Operational Requirements (i.e., LCO and surveillance requirements) originally contained in the Technical Specifications.

Tennessee ValleyAuChority

SUMMARY

OF Browns Ferry Nuclear Plant SAFETYEVALUATIONSFOR l993 Annual Operating Report UFSAR REVISIONS Since submittal to the NRC ofour proposed Fire Protection - Volume 1 (Revision 1),

several design documentation changes/documents have been issued which require incorporation into the Fire Protection Report, Volume 1 prior to plant approval.

The requested changes addressed the implementation ofthe compensatory actions hose stations and the Site Fire Protection Organizational Chart (added tables to reflect these requirements), addition ofair supervision to the system design description ofthe Fire Protection Plan for the preaction sprinkler system supplying the intake pumping station, revision ofUnit 2 Appendix R Safe Shutdown Program to reflect the inability to take credit for the Thermo-Lag 330 fire wrap on specific cables in fire zone 2-4.

With NRC approval ofthe BFN Fire Protection Report, Volume 1 and Technical Specification Change 306, there is no potential Technical Specification impact (with Fire Protection fullyremoved from the Technical Specifications).

The BFN Fire Protection Report, Volume 1, Fire Protection Plan (Section 4) and Unit 2 Appendix R Safe Shutdown Program both require revision in support ofthe design and documentation changes discussed above.

UFSAR Section 10.11 states that the Fire Protection Report, Volume 1 is the licensing basis for BFN's fire protection program and changes to it are subject to the requirements of 10CFR50.59.

This change was acceptable from a nuclear safety standpoint and no unreviewed safety question was involved.

Fire Protection Report (26-9310-018)

Descri tion/Safet Evaluation This safety evaluation addressed revisions to the Fire Protection Report, Volume 1 test frequencies as shown on the following page:

Tennessee ValleyAulhorily

SUMMARY

OF BroNns Ferry Nuclear Plant SAFETYEVALUATIONSFOR l993 Annual Operating Reporl UFSAR REVISIONS Test Re uirement 9.4.11.B.l.a 9.4.11.B.1.b 9.4.11.C.1.a 9.4.11.D.1 9.4.11.A.1 D~eecri tion High pressure fire protection (HPFP) electric driven fire pump operability HPFP valve position verification (inside loop)

Sprinkler system valve position verification CO2 system valve position.

verification Fire detection instrument functional Monthly Monthly Monthly Semi-annual None None None Annual Monthly Quarterly The excellent performance ofthe electric fire pumps (1 failure during the past 52 tests) along with the existing BFN hold order, lay-up, and status control programs used to control valve positions, and the change in the NFPA code test frequency for fire detectors serve as justification for relaxatioh.

This change also addressed a clarification note to Table 9.3.11.E as shown below:

NOTE: In accordance with the Fire Hazard Analysis for Fire Areas 1 8'c 3, Fire Zone doors 490 2 635 are not required within Fire Area 1 when Unit 2 and/or Unit 1 is the only operating unit. Fire Zone doors 506 8. 651 are not required within Fire Area 3 when Unit 2 and/or Unit 3 is the only operating unit.

This change was for clarification only and simply restates information under "Fire Protection Evaluation" on page 23 and 42 ofthe Fire Hazards Analysis in Fire Protection Report, Volume 1.

This revision to the Fire Protection Report, Volume 1 was safe from a nuclear safety standpoint, did not require any changes to Technical Specifications, did not require any special tests or experiments, did not involve any changes to radwaste systems, and did not adversely affect the ability to achieve and maintain safe shutdown in the event ofa fire.

However, this revision did changes text and affect procedures referenced in the UFSAR, therefore a safety evaluation was required.

No unreviewed safety question is involved.

Tennessee VolleyAulhoriiy SUNNA RY OF Browns Ferry Pluelenr Plant SAFETYEVALUATIONS FOR I993 Annunl Opernli ng Reporl UFSAR REVISIONS

"".")c<4j>')) jj c<q%>ij gk~i)c))'..)~@cjkpp) ))"jN)))'."sp::p )'jcj:.<)').::?p?g'<:>::))'j>)'))')~)~%(p)))~)R )g~g&~) '>')%(>~i )pygmy)'.

Fire Protection Report (SEBFDCN930037 RO)

Descri tion/Safet Evaluation This safety evaluation addressed CRLD BFEP-BNA-93021 RO revision to the Appendix R Safe Shutdown Section ofVolume 1 ofthe Fire Protection'Report.

This revision changed the required selection ofa fire pump for a fire in the control bay (fire area 16) from fire pump Ato fire pump C. Fire pump A has its power cable routed in fire area 16 thus its availability cannot be assured for a fire in this area.

Fire pump C has been verified to be available for a fire in fire area 16.

Since only one ofthese equal capacity fire pumps is required to confine and extinguish a fire in fire area 16, this was an acceptable change.

There were no physical modifications associated with this change.

A review ofthe Technical Specifications identified no changes as a result ofthis revision to the Fire Protection Report.

UFSAR Section 10.11 states that the Fire Protection Report, Volume 1 is the licensing basis for BFN's fire protection program and changes to it are subject to the requirements of 10CFR50.59.

This change was acceptable from a nuclear safety standpoint and no unreviewed safety question was involved.

Fire Protection Report (SEBFSAR930040 RI)

Descri tion/Safet Evaluation The Fire Protection Report, Volume 1, Section 9.4.11.E.l.a requires a monthly visual inspection ofthe fire hose stations to assure all required equipment is at the station.

This equipment is needed by the fire fighters to manually fight any potential fire. This monthly inspection was revised to be performed on a once per year frequency.

The Fire Protection Report, Volume 1, Section 9.4.11.E.l.b requires at least once per 18 months that the hose be removed for inspection and re-racking, and inspection ofall gaskets and replacement ofdegraded gaskets in the couplings.

This 18 month inspection frequency was increased to a yearly inspection and was combined with the yearly inspection ofall equipment in the fire hose stations.

Tennessee ValleyAuthority

SUMMARY

OF Bro>vns Ferry Nuelenr Plnnt SAFETYEVALUATIOiVSFOR 1993 Annunl Opernling Report UFSAR REVISIONS a~~'~ a@~wM~ ': 'w~k"""5"'i'i'M"~~'w". '4~~"~~~4AM The Fire Protection Report, Volume 1, Sections 9.3.11.F.1 and 9.4.11.F.1 and Table 9.3.11.D contain the LCO and the surveillance requirements for yard fire hydrants and hose houses.

The hose houses contain the fire hoses and associated equipment needed to manually fight a potential fire in the yard area.

These two sections and Table 9.3.11.D were revised to delete the requirements associated with the hose houses.

A This fire hose stations are in a protected/controlled area ofBFN and are not accessible to the general public and as such, there is a very low probability ofthe equipment being removed, tampered with, or damaged.

Based on a review ofthe pertinent NFPA standards, the limited/controlled access ofBFN, and the positive results from the last two years ofmonthly inspections, the decrease in inspection frequency from once per month to yearly was acceptable.

The increase ofinspection frequency from 18 months to yearly is consistent with NFPA-1962 recommendations.

The deletion ofthe requirements for the hose houses from the Fire Protection Report, Volume 1, Section 9.3.11.F. I and 9.4.11F.1 and Table 9.3.11.D was acceptable since it is replaced with the equivalent coverage provided by a fire truck (pumper) or its backup fire truck (pumper).

This requirement was deleted because BFN maintains a fullyequipped fire truck (pumper) and trained operators.

The fire truck (pumper) and backup fire truck (pumper) contain sufficient fire hoses and fire fighting equipment that they provides an equivalent capability.

No Technical Specification change was required.

4 This change was acceptable from a nuclear safety standpoint and no unreviewed safety question was involved.

UFSAR Appentl~ C, Structural {)ualificationofSubsystents and Contponents, Sections C4 Through C10 (SEBFSAR930018 RO)

Descri tion/Safet Evaluation This safety evaluation addressed changes made by CRLD BFEP-CEB-93002 RO.

The information incorporated into Appendix C which is to be included in Amendment 10 ofthe UFSAR has been extensively'eviewed by the NRC. Their review and approval was issued in the NRC Safety Evaluation Report on TVANUREG-1232 Supplements 1 and 2 (BFN plant s'pecific). The torus attached piping and equipment was reviewed and approved in U.S. NRC, "Safety Evaluation Report Mark I Containment Long-Term Tennessee VolleyAull>ority

SUMMARY

OF Browns Ferry Nuclenr Plnnt SAFETYEVALUATIONS FOR I993 Annual Operating Report UFSAR REVISIOlVS

%45'Q954a4%'w>'MM~~NRNPsc@44>h~'M "0'k"%k8~k Program", NUREG 0661 and the U.S. NRC, "MarkI Containment Long-Term Program" for BFN, Units 1, 2, and 3, Safety Evaluation - Pool Dynamics Load, Safety Evaluation-Structural Review. The HVACsupports and ducts were reviewed and approved in a NRC Safety Evaluation Letter to TVA, "Evaluation ofSeismic Design Criteria for HVAC

- BFN". This safety evaluation ofAppendix C incorporated the above referenced and issued NRC safety evaluations.

The items identified as update changes to the UFSAR in this safety evaluation did not afFect the safety function ofany system, structure, or component.

This change incorporated text which was inadvertently removed during the previous UFSAR update, made clarifications ofafFected sections, made editorial changes, and made changes that represent corrections as the means ofobtaining consistency with other related documents.

No physical plant modifications or revisions to existing procedures were required.

No Technical Specification changes were required.

This UFSAR update was acceptable from a nuclear safety standpoint and no unreviewed safety question was involved.

UFSAR AppentlU-C, St'ractnral Qttalification ofSttbsg>stettts anti Cotnponents, Sections C 0, Cl, C2, anti C3 (SEBFSAR930017 RO)

Descri tion/Safet Evaluation This safety evaluation addressed changes made by CRLD BFEP-CEB-93001 RO.

The information incorporated into Appendix C which is to be included in Amendment 10 ofthe UFSAR has been extensively reviewed by the NRC. Their review and approval was issued in the NRC Safety Evaluation Report on TVANUREG-1232 Supplements 1 and 2 (BFN plant specific). The torus attached piping and equipment was reviewed and approved in U.S. NRC, "Safety Evaluation Report Mark I Containment Long-Term Program", NUREG 0661 and the U.S. NRC, "MarkI Containment Long-Term Program" for BFN, Units 1, 2, and 3, Safety Evaluation - Pool Dynamics Load, Safety Evaluation-Structural Review. The HVACsupports and ducts were reviewed and approved in a NRC Safety Evaluation Letter to TVA, "Evaluation ofSeismic Design Criteria forHVAC

- BFN". This safety evaluation ofAppendix G incorporated the above referenced and issued NRC safety evaluations.

Tennessee ValleyAulhoriIy S UMIMARYOF Browns Ferry Nuclear Planl SAFETYEVAI.UATIOJVSFOR I993 Annual Operating Reporl UFSAR REVISIOJYS

'R~g)'pK@(~j'Qx'x4~.,$~4.>.>"'xk~cg;.g~<... x..~c:.',.c:v.:.'x~$V~)c'%, <:.'.c':;;xiP.

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'~MY'xc.'c'he items identified as update changes to the UFSAR in this safety evaluation did not affect the safety function ofany system, structure, or component.

This change incorporated text which was inadvertently 'removed during the previous UFSAR update,,

made clarifications ofaffected sections, made editorial changes, and made changes that represent corrections as the means ofobtaining consistency with other related documents.

No physical plant modifications or revisions to existing procedures were required.

No Technical Specification change was required.

These changes were acceptable from a nuclear safety standpoint and no unreviewed. safety question was involved.

UFSAR Secli'ori 4.1Z - lnservice lnspeclion anti Testing (00-9302407)

Descri tion/Safet Evaluation UFSAR Section 4.12 documents the requirements for inservice inspections, pump and valve tests, and pressure tests.

This change updates Section 4.12 to remove references to specific editions of Section XIofthe ASME Boiler and Pressure Vessel Code except for historical purposes.

This willensure that the edition ofthe code that is currently in effect is utilized to determine all testing requirements.

The effective code editions willbe documented in plant procedures.

Deletion ofthe specific ASME Section XI Code edition requirements willbring UFSAR Section 4.12 into agreement with Technical Specifications.

Therefore, no change to the Technical Specifications is required.

J This change is acceptable from a nuclear safety standpoint and no unreviewed safety question is involved.

UFSAR Secti'on 7.9.4.5-Recirculation Panip Trip (RPT) Conf'rol System (SEBFSAR930010)

Descri tion/Safet Evaluation This safety evaluation supported a change to the description ofthe RPT control system presented in the UFSAR.

Tennessee VnlleyAuthority

SUMMARY

OF Bro>vns Ferry Pluelenr Plnnt SAFETYEVALUATIONS FOR 1993 Annunl Operating RePort UFSAR REV1$1ONS 9aM " ~~m4%'Md4M~ '

~sec" '" '"~~ '~ '

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The change

1) deletes the words "Safety Class 3" and states that the system is not classified as safety related; 2) states that the equipment is "expected" (versus "required")

to remain functional during a DBE; and 3) provides a more accurate description ofthe reactor protection system (RPS)/RPT trip logic interface.

There is no hardware problem with the installed end ofcycle RPT equipment.

There is a UFSAR "language" problem. The UFSAR wording over-stated requirements on the equipment and presented a misleading description ofthe RPS/RPT interface.

"Safety Class 3" is a safety-related category defined in American National Standards Institute (ANSI)/American Nuclear Society (ANS)-52.1. It generally applies to electrical equipment and to some piping systems.

It is the "least important" ofthe safety-related categories in that standard.

TVABFN design criteria and implementing procedure do not use this terminology. Also this terminology is not used in proposed ANSVANS-58.14, upon which the TVABFN design criteria and procedure are modeled.

Per the TVABFN design criteria and procedure and per the proposed ANSI/ANS standard, the end ofcycle RPT function is not classified as safety related.

The end of cycle RPT equipment is classified as "quality related" per the TVABFN criteria and procedure.

Per the proposed ANSI/ANS standard the end ofcycle RPT equipment is classified as nonsafety related, "supplemented grade".

The "supplemented grade" classification is identical to the TVABFN "quality related" classification. It was inconsistent with the TVABFN criteria and procedure for the UFSAR to continue to say that the end ofcycle RPT equipment was "Safety Class 3".

The end ofcycle RPT equipment was designed to applicable seismic Category I, Class lE standards.

The design has not been changed.

Therefore, it remains "designed to seismic Category 1, Class lE standards".

Because it is designed to these standards, it is "expected to remain functional during a DBE". However, there is no requirement documented in the Safe Shutdown Analysis or elsewhere that the system must remain functional during the design basis seismic event.

These changes were made to correctly state the requirements and describe the existing hardware configuration.

The changes did not result in any modification to plant systems or components nor reflect any deficiency in currently installed plant equipment.

/

No Technical Specification changes were required.

These changes were acceptable from a nuclear safety standpoint and no unreviewed safety question was involved.

Tennessee ValleyAutborily

SUMMARY

OF Broivns Ferry Nuclear Plant SAFETYEVALUATIONSFOR 1993 Annual Opera!in'eport UFSAR REVISIONS

~p~>,,':,<,:. "<~",%,",i'."..'~. ""P:"4': > '.."".~".

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<q'; c~'">~.;,g~i~p;.;.,~.. pi~,,.>.",",..',.i~,i">~>;:;:,.:;" ';;>~:"~~~~ a UFSAR Section 7.13 - Area Rarliation Monitoring System UriSARSection 7.15 - Health Physics Laboratory Ratliation Monitoring Equipment (SEBFSAR930045 RO)

Descri tion/Safet Evaluation In order to provide consistency between equipment operation, procedures, and the UFSAR revisions to Sections 7.13 and 7.15 ofthe UFSAR were requested.

These revisions provided the following:

~

Corrected typographical errors;

~

Removed mounting details for friskers;

~

Provided suEicient level ofdetail to describe the personnel contamination monitoring equipment used at BFN;

~

Provided location ofportal monitor;

~

Removed Section 7.13.5.3.5 to provide additional operational flexibilitypertaining to the monitoring oflaundry ifoutside monitoring services can perform the activity in a more cost efl'ective manner;

~

Provided a description ofthe physical features ofthe door access control system;

~

Removed detailed calibration intervals for instruments used for training and provided a statement that the plant willmaintain equipment necessary to properly assess radioactive hazards which may be encountered;

~

Deleted contamination monitoring equipment description from section 7.15.3 because the previous section on radiation monitoring can be interpreted to apply to equipment for measuring radiations from radioactive contamination;

~

Reflects the use ofself-reading dosimeters and, ifradiation exposure conditions warrant, the removal ofthermoluminescent dosimeters (TLDs) as a requirement to enter the plant radiological controlled area;

~

Removed details ofdose receptors in Section 7.15.4 to provide flexibilityin providing only the monitoring which is required;

~

Changed Section 7.15.4 to be consistent with the change from the use ofdirect reading dosimeters (DRD). It is anticipated that self reading dosimeters (SRD) may be used in the future as the oflicial dose record device.

The equipment affected by this activity is employed to satisfy certain Technical Specifications.

Revising the descriptions ofequipment used in the radiation control program in the UFSAR does not decrease the capability required by the Technical Specifications.

No Technical Specification change is required.,

Tennessee ValleyAulboriiy

SUMMARY

OF Brogans Ferry Nuclear Plant SAFETYEVALUATIONS FOR 1993 Annual Operating Report, UFSAR REVISIONS

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These changes were acceptable from a nuclear safety standpoint and no unreviewed safety question was involved.

UFSAR Section 8.8-AuxiliaryDC Poiver Siipp1y anil Distribution (SEBFSAR900036 R t)

Descri tion/Safet Evaluation UFSAR Section 8.8 states that the batteries are capable ofsupplying the connected loads for a period ofapproximately 3 1/2 hours and that the load requirement calculation that confirms this is shown in Table 8.8-2. Calculation ED-Q1000-87049 R1 does not agree with the information provided in Table 8.8-2 and does not provide confirmation that the 24VDC batteries can supply the connected load for 3 1/2 hours. UFSAR Section 8.8.2.2, Subparagraph 2 states that the 24VDC batteries shall be capable ofsupplying the connected loads for a period of3 hours.

This is substantiated by calculation ED-Q2000'-87049 R l. Therefore to provide consistency and accuracy within the UFSAR, Table 8.8-2 and any reference to it were requested to be deleted.

The batteries continue to becapable ofsupplying their connected loads for the required 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> and meet all other regulatory commitments.

The effect ofthis change is to remove incorrect and unverifiable information from the UFSAR. There is not effect on battery or 24VDC power system design capabilities or functional performance.

The only time that this change has significance is when a battery charger fails and a battery is called upon to furnish backup power until charger power can be restored.

The UFSAR text reduction in battery capability from 3 1/2 to 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> is inconsequential when compared to operator response time to mitigate the consequences ofthis event.

Specifically, operator response time may be conservatively estimated to be less than 1

hour based on previous operating data.

Since the 24VDC batteries perform no safety-related functions and are not explicitly described in any BFN Technical Specification and operator reaction time is well within the calculated 3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> limitwhich the batteries are postulated to be able to carry their attached loads, there willbe no decrease to any safety margin defined in any BFN Technical Specification bases.

This change was acceptable from a nuclear safety standpoint and no unreviewed safety question was involved.

Tennessee ValleyAulhority

SUMMARY

OF Brogans Ferry Nuclear Plant SAFETYEVALUATIONSFOR 1993 Annual Operating Report

=

UFSAR REVISIONS UFSAR Section 12.2 - Principal Strnctnres anti Foundations (SEBFSAR930013) t Descri tion/Safet Evaluation This change to Section 12.2 incorporates text which was inadvertently removed during the previous UFSAR update, clarifications ofaffected sections, editorial changes, and changes that represent corrections as the means ofobtaining consistency with other related documents.

No physical plant modifications or revisions to existing procedures are required for these changes.

The description ofthe changes are:

Section 12.2.4.1, Page 12.2-51, Paragraph 1: Insert "Case 3 - Earthquake" in the heading at the top ofthe page; Section 12.2.7.8.3, Page 12.2-71:

Add a new paragraph between the third and fourth paragraphs to reflect the current seismic analysis performed for the vacuum pipe building; Table 12.2-27:

Revise the calculated safety factors to reflect the results ofthe dynamic earthquake analysis performed on the intake pumping station; Table 12.2-27, Principal Design Case Ib: Revise Principal Design Case Ib to reflect proper loading for 0.1g earthquake; Table 12.2-27, Principal Design Case III: Change the word "unwanted" to "unwatered";

Section 12.2.7.2, Page 12.2-65, Paragraph 1: Replace "greater than" with "equal to";

Section 12.2.7.2, Pager 12.2-65, Paragraph 3: Replace "7.3" with "8.5";

Section 12.2.7.2, Page 12.2-65, Paragraph 4: Add "post earthquake" between "the" and "static". Delete "and seismic";

Section 12.2.7.2, Page 12.2-65, Paragraph 4: Change "analyses" to "analysis";

Section 12.2.10.1, Concrete Structure, Page 12.2-83:

Add a first paragraph to describe the two SGT buildings and their location relative to the reactor building and each other. The description ofbuilding no. 2 is an addition to Section 12.2.10.1; Section 12.2.10.2, Dynamic Earthquake analysis, Page 12.2-84: Delete the entire section and replace with a writeup to reflect the current dynamic earthquake analysis; Table 12.2-35:

Revise Table 12.2-35 to reflect the changes in the mathematical models used in the seismic analysis ofthe SGT building no.

1 and add the corresponding values for building no. 2; Section 12.2.14, Concrete Structures, Page 12.2-88: Delete the last paragraph and replace with a revised writeup to reflect the current seismic design basis ofthe offgas treatment building; Table 12.2-27, Principal Design Case V: Add a principal design case for the high walls surrounding the RHRSW pumps; Tennessee ValleyAuthority

SUMMARY

OF Brogans Ferry Nuclear Plant SAFETY EVALUATIONS FOR 1993 Annual Operating Reporl UFSAR REVISIONS M'~'d~~""""-':-~":8 '-"':'"'"'"'ll'"'"-': " "~"'"'""~"'"".""'-"'M2"~A~l~"';~"l'~

~

Section 12.2.16, Page 12.2-90, Paragraph 2: Add "or tornado winds" between "Design Basis Earthquake (0.20g)" and "to";

~

Section 12.2.2.7.2, Page 12.2-33, Paragraph 5: Delete the fiAhparagraph and replace with a new writeup.

This'UFSAR update to Section 12.2 does not change the function or operability ofany safety-related system or component.

Review ofthe BFN Technical Specifications indicates that there are no impacts on or potential changes to the Technical Specifications, therefore this change does not require a change to the Technical Specifications.

This change is acceptable from a nuclear safety standpoint and no unreviewed safety question is involved.

UFSAR Section 13.3 - Training Progranis (00-9304-014 RO)

Descri tion/Safet Evaluation This UFSAR revision deleted the detailed descriptions ofthe Nuclear (Nonlicensed)

Operator, Nuclear Systems Operator Training - Cold License, Nuclear Power Plant Fundamentals Course, Hot License Training Programs, License Requalification Program, and Replacement Training for Operators contained in Sections 13.3.3, 13.3.4, 13.3.5, 13.3.6, 13.3.7, and 13.3.8. New Sections 13.3.3, 13.3.4, and 13.3.5 were added which contain brief program summaries and identify the programs as accredited by the National Academy ofNuclear Training.

Existing Sections 13.3.9, 13.3.9.1, 13.3.9.2, 13.3.10, 13.3.10.1, 13.3.10.2, and 13.3.10.3 were renumbered.

Additionally, the reference to Site Director Standard Practices (SDSPs) contained in the existing 13.3.9 were changed to Site Standard Practices (SSP) to reflect current nomenclature.

These changes did not reduce the scope ofthe programs in any way. They simply reduced the amount ofadministrative detail contained in the descriptions.

Training program structure and content guidance continues to be governed by the corporate level training program procedures contained in the Nuclear Power Training Manual. These program procedures are subject to review in accordance with 10CFR50.59.

Additionally, these procedures are the procedures which describe the programs accredited by the Academy.

Accreditation is n'ot granted or maintained unless Tennessee ValleyAuthority,

SUMMARY

OF Brogans Ferry Nuclear Plant SAFETYEVAI.UATIONS FOR 1993 Annual Operating Reporl UFSAR REVISIONS

""':='"<<"'"".'m""5""'~"'",:~4 -""'."%m!"%~'""':-'~'"FZRl compliance with the Academy published guidelines is established.

These guidelines are consistent with, and in many areas, exceed the regulatory requirements previously required.

This change was acceptable from a nuclear safety standpoint and no unreviewed safety question was involved.

Tennessee ValleyAuthority Brogans Ferry Nuclear Plant l993Annual Operating Report 1993 RELEASESUhNSARY SE:-:::

-'-:"'-"'-::"-:::-::::--::--i-:::::-::::::::::"::- ":::::-::-:-':i-::-::::::.:::,:: "'---::;:::::::::::::::;i:: '--::"-::::i 1993 RELEASE

SUMMARY

~

i

1993 RELEAGE GUMlfARY ANNUAL OPERATING REPORT GASEOUS RELEASES LIQUIDRELEASES MONTH JANUARY FEBRUARY MARCH APRIL JULY AUGUST OCIOBER NOVEMBER DECEMBER FISSIONS &

ACTIVATION PRODUCTS CI 2.66E+03 6.67E+02

<1.42E+01

<9.26E+00

<5.01E+01 2.18E+02 1.25E+02 8.26E+01 1.03E+02 8.86E+01 5.65E+01 1.36E+01 IODINES CI 5.65E-04 2.06E-03 3.70E-04 2.28E-05

<1.75 E-04 1.48-04 3.02-04 3.64E-04 4.95E-04 2.52E-04 2.34E-04 1.67E-04 PARTICULATES

)8 DAY HALF-LIVES CI 6.12E-03 5.00E-03 5.07E-03 2.46E-04 8.95E-04 2.58E-03 1.17E-04 9.11E-05 2.77E-04 2.00E-04 5.61E-04 4.70E-04 TRmUM CI 9.55E-01 2.24E-01 2.72E-01 1.79E-01 2.03E-01 5.56E-01 1.39E+00 1.35E+00 1.57E+00 8.93E-01 8.12E-01 9.50E-01 FISSION &

ACTIVATION PRODUCTS CI 4.17E-01 3.59E-01 2.93E+00 8.92E-02 1.57E-01 1.06-01 9.57E-02 8.45E-02 6.97E-02 1.08E-01 3.17E-01 7A4E-02 TRITIUM CI 2.72E+00 1.80E+00 7.40E-01 3.02E-01 7.76E-01 5.05E-01 5.30E-01 1.01E+00 8.45E-01 8.31E-01 1.28E+00 1.07E+00 DISSOLVED NOBLE GASES CI 1.50E-01 9.93E-03 ND 2.86E-05 1.78E-04 4.68E-04 6.01E-04 1.64E-03 2.59E-03 8.41E-03 1.43E-03 GROSS ALPHA CI ND ND ND 8.92E-05 1.83E-05 ND 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.

Tennessee ValleyAuthority Bro>vns Ferry Nuclear Plant 1993 Annual Operating Report 1993 OCCUPATIONALEXPOSUREDATA

~'%YR

' m~M~"

" '"'cÃ'~'" '

""~'" 5~4v'

"~~mm~

'~

' 'M

'- "-'%YKC~M~~"' '"""~%P"~~""'"-'-'~%R%4Ya ""MYMSF~%R";~5 1993 OCCUPATIONAL EXPOSURE DATA'83-

REXPR21 RUN DATE.

1-20-94 RUN TIME: 18:00:29 TENNESSEE VAL Y

AUTHORITY BFN RADIATION EXPOSURE SYSTEM NUMBER OF PERSONNEL AND MAN-REM BY WORK JOB FUNCTION TOTAL NUMBER OF INDIVIDUALS NUMBER OF PERSONNEL

(> 100 M-REM)

TOTAL MAN-REM MD=REACTOR OPS SURVEILLANCE GROUP STATION UTILITY EMPLOYEES EMPLOYEES CONTRACT TOTAL AND OTHERS PERSONS STATION EMPLOYEES UTILITY EMPLOYEES CONTRACT AND OTHERS TOTAL M-REMS MAINTENANCE PERSONNEL OPERATING PERSONNEL HEALTH PHYSICS PERSONNEL SUPERVISORY PERSONNEL ENGINEERING PERSONNEL MO 229 161 78 54 64 11 2

0 3

9 654 19 68 78 80 894 182 146 135 153 3.725 33.752 7.503 2.677 3.621

0. 139 0.511 0.000 0.053 0.330 14.239 0.082 7.074 2.509 4.895

18. 103 34.345 14.577 5 '39 8.846 MD=ROUTINE MAINTENANCE GROUP STATION UTILITY EMPLOYEES EMPLOYEES CONTRACT TOTAL AND OTHERS PERSONS STATION EMPLOYEES UTILITY EMPLOYEES CONTRACT AND OTHERS TOTAL M-REMS MAINTENANCE PERSONNEL OPERATING PERSONNEL'EALTH PHYSICS PERSONNEL SUPERVISORY PERSONNEL ENGINEERING PERSONNEL MO 427 154 89 59 62 26 2

2 3

9 1232 1685 21 177 68 159 119 181 97 168 125.721 5.601 12.643 4.214 9.468 4.666 0.080 0.660 1.211 1.466 295.634 1.881 11.156 9.324 11.819 426.021 7.562 24.459 14.749 22 '53 MD=IN-SE.VICE INSPECTION GROUP STATION EMPLOYEES UTILITY EMPLOYEES CONTRACT TOTAL AND OTHERS PERSONS STATION EMPLOYEES UTILITY EMPLOYEES

CONTRACT, TOTAL AND OTHERS M-REMS MAINTENANCE PERSONNEL OPERATING PERSONNEL HEALTH PHYSICS PERSONNEL SUPERVISORY PERSONNEL ENGINEERING PERSONNEL MO 31 2

19 3

7 0

0 0

1 3

200 1

36 2

23 231 3

55 6

33

1. 974 0 ~ 001 1.337 0.048 1.200 0.000 0.000 0.000 0.039 1.275

22. 219

0. 001 1.355 0.029 6.800

24. 193 0.002 2.692 0.116 9.355 MD=SPECIAL MAINTENANCE GROUP MAINTENANCE PERSONNEL STATION UTILITY EMPLOYEES EMPLOYEES CONTRACT TOTAL AND OTHERS PERSONS 296 4

1129 1429 STATION UTILITY EMPLOYEES EMPLOYEES 39.258 0.146 CONTRACT TOTAL AND OTHERS M-REMS 182.039 221.443

REXPR219 RUN DATE: 01-20-94 RUN TIME: 18:00:29 TENNESSEE VALL Y

AUTHORITY

, BFN RADIATION EXPOSURE SYSTEhl NUMBER OF PERSONNEL AND MAN-REM BY WORK JOB FUNCTION TOTAL NUMBER OF INDIVIDUALS NUhtBER OF PERSONNEL

(> 100 M-REhl)

TOTAL MAN-REM OPERATING PERSONNEL HEALTH PHYSICS PERSONNEL SUPERVISORY PERSONNEL ENGINEERING PERSONNEL MO 67 70 17 30 19 63 107 97 86 133 126 133

2. 616 4.723 0.524 1.232 0.000 0.000 0.093 0.402 0 '42 6 '80

6. 169 14.464 2 '58 11.503 6.786 16.098 MD=WASTE PROCESING GROUP STATIO'I UTILITY EMPLOYEES EMPLOYEES CONTRACT TOTAL AND OTHERS PERSONS STATION EMPLOYEES UTILITY EMPLOYEES CONTRACT ~

TOTAL AND OTHERS M-REMS MAINTENANCE PERSONNEL OPERATING PERSONNEL HEALTH PHYSICS PERSONNEL SUPERVISORY PERSONNEL ENGINEERING PERSONNEL MO 32 10 28 2

1 4

0 0

0 0

79 2

7 1

1 115 12 35 3

2 1.024 0.642 0;770

'0.011 0.006 0.042 0.000 0.000 0.000 0 F 000 3.051 0.752 0.659 0.001 0.335 4.117 1.394 1.429 0.012

0. 341 MO=REFUEL GROUP STATION UTILITY EMPLOYEES EMPLOYEFS CONTRACT TOTAL AND OTHERS PERSONS STATION EMPLOYEES UTILITY EMPLOYEES CONTRACT TOTAL AND OTHERS M-REMS MAINTENANCE PERSONNEL OPERATING PERSONNEL HEALTH PHYSICS PERSONNEL SUPERVISORY PERSONNEL ENGINEERING PERSONNEL MO 57 29 13 2

2 131 1

14 1

33 189 30 27 4

38 2.987 1.736 1.847 0.082

0. 015 M%7 0.012 0.000 0.000 0.000 0.074 17.717 0.010 1.248 0.316 2.724 20.716 1.746 3.095 0.398 2.813 Me)3

,2095 92 4383 6570 270.958 11.199 625 '04 907.861

REXPR219 RUN DATE: 01-20-94 RUN TIME: 18:00:29 TENNESSEE VALLEY AUTHORITY BFN RADIATION EXPOSURE SYSTEM NUMBER OF PERSONNEL AND MAN-REM BY WORK JOB FUNCTION TOTAL NUMBER OF INDIVIDUALS'UMBER OF PERSONNEL () 100 M-REM)

TOTAL MAN-REM GROUP STATION UTILITY EMPLOYEES EMPLOYEES CONTRACT TOTAL AND OTHERS PERSONS STATION EMPLOYEES UTILITY EMPLOYEES CONTRACT AND OTHERS TOTAL M-REMS MAINTENANCE PERSONNEL OPERATING PERSONNEL HEALTH PHYSICS PERSONNEL SUPERVISORY PERSONNEL ENGINEERING PERSONNEL 1072 423 297 137 166 2095 46 4

2 10 30 92 3425 63 256 308 331 4383 4543 490 555 455 527

.6570 174. 689 44.348 28.823 7.556 15.542 270.958 5.005 0.591 0.660 1.396 3.547 11.199 534.899 3.068 28.272 18.348 41.117 714

~ 593 48.007 57.755 27.300 60.206 625.704 907.861 r

REXPR21 RUN DATE. 94 RUN TIME: 18'00'29 TENNESSEE VAL~

Y AUTHORITY BFN RADIATION EXPOSURE SYSTEM NUMBER OF PERSONNEL AND MAN-REM BY WORK JOB FUNCTION TOTAL NUMBER OF INDIVIDUALS GROUP STATION UTILITY CONTRACT TOTAL MAINTENANCE PERSONNEL OPERATING PERSONNEL HEALTH PHYSICS PERSONNEL SUPERVISORY PERSONNEL ENGINEERING PERSONNEL 419 133 85 26 59 722 19 1

1 2

10 33 1217 2

73 37 105 1434 1655 136 159 65 174 2189

0

Tennessee ValleyAulhority Brains Ferry Nuclear Plant CHALLENGES TO OR FAILURESOF 1993 Annual Op~erating Report MAINSTEAMRELIEF VALVES K~~"'-'."".'-~"':". " """": ':"::::- :":: '="":"'::." "':."-:-:""":::""--':"".:.".:.'

::v"":: -""'-~

1993 CHALLENl ES TO OR FAILURES OF MAINSTEAM RELIEF VALVES

Tennessee ValleyAuthori ty Browns Ferry Nuclear Plant CHALLENGES TO OR FAILURESOF I993 Annual Operating Report MAINSTEAMRELIEF VALVES UNITI None During the reporting period, Unit 2 operated for approximately eight months. During the operating period, the main steam safety reliefvalves (MSRVs) operated without challenges or failures.

During a four month refuel outage, thirteen MSRV pilot cartridges were removed and sent to Wyle Laboratories, Huntsville, Alabama for post-cycle setpoint verification. Test results indicated that eleven out ofthirteen pilot cartridges failed to meet the acceptance criteria ofsetpoint+ 1%.

This failure was reported under Licensee Event Report 50-260/93003.

Prior to reactor restart, a fullycertified compliment ofpilot cartridges were installed.

UNIT3 None Tennessee ValleyAuthority Brains Ferry Nuclear Plant I993 Annual Operating Report REACTOR VESSEI. FATIGUE USAGE EVALUATION K:.: ':-'.-"":""""-':'""'::

<.. y< kj@Vgi"c.,~g,":x,, k.,>g:., N<~iiijs,.)@~M<'qN+gi'@3jgz+~<,.";::~~<.'.>pwca?~(<N;<'>.'.Nag':4;.<.".'.j~;.; ~(z::aP?Zy@c'<'<?.'.%~+~)(g@';:g¹ivgg<j?'<<,

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1993 REACTOR VESSEL FATIGUE USAGE EVALUATION Tennessee ValleyAuthori ty Browns Ferry Nuclear Plant 1993 Annual Operating Report REACTOR VESSEL FATIGUE USAGE EVALUATION

"'""'6'o'""'""'" "'" """':.'.">>>"'j:~ '"'"'

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,.;:.n.MN~~r,. '.c ~x;:@:sg>.:.:~>>~>>~%>>>>>>5~>V~~>::

'<.'>>~~~>>',p.~%PE.,:,:~y~a;s-~~s':>>"

~~:,'@~>,.~p.,'.;.C~>~@$4,"~%~~~~s:.:,:y>>;<>>>>g!ggy>>~s;,>(y.s>>XN".'P~a@

The cumulative usage factors for the reactor vessels are as follows:

Location Shell at water line Feedwater nozzle Closure studs UniiI 0.00620 0.29782 0.24204 Unit 2 0.00565 0.22274 0.22045 Unii3 0.00431 0.16139 0.14360