ML18038B154
| ML18038B154 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 12/31/1994 |
| From: | TENNESSEE VALLEY AUTHORITY |
| To: | |
| Shared Package | |
| ML18038B153 | List: |
| References | |
| NUDOCS 9503060284 | |
| Download: ML18038B154 (182) | |
Text
TENNESSEE VALLEYAUTHORITY BROWNS FERRY NUCLEARPLA.NT ANNUALOPERATING REPORT January 1, 1994-December 31, 1994 Docket Number 50-259, 50-260, and 50-296 License Number DPR-33, DPR-52, and DPR-68 9503060284 950228 PDR ADOCK 05000259 R
II 0
O~
Tennessee Valley Authority Broivns Ferry Nuclear Plant
'99@Annual Operating Report'
'-'~-""S."""~%M--
--:~..'.!!!".kk.
W!.."NO - 'L
--Nt-'cronyms Listing..
Regulatory Guide 1.16,Section I.b.(I) and (2)
Operational Summary.
IOCFR50.59(b) (2) - Sumnrary ofSafety Fvaiuations Core Component and Operating Limits Field Completed Plant Modifications..
Fire Protection-Report Revisions 13 50 New Instructions/Procedure Revisions 57 Special Operating Conditions.
60 Special Tests 63 Temporary Alterations.
Temporary Shielding-Request.
Updated Final Safety Analysis Report Revisions.
Regulatory Guide 1.16,Section I.b. (3) 1994 Release Summary.
67 70 73 79 Technical Specification 6.9.1.2 1994 Occupational Exposure Data.
81 Challenges to or Failures ofMain Steam Relief Valves.....
86 Technical Specification 6.9.2.1 Reactor Vessel Fatigue Usage Evaluation.
0 ll ll~
0 Tennessee ValleyAuthority Browns Ferry Nuclear Plant 1994 Annual Operating Report ACRONYMSLISTING This.is a list of acronyms and abbreviations used throughout the 1994 Annual Operating Report.
0 ADS AFFF ALARA APRM ASME ATWS BFN BPWS BWR CAQR CFR CISS CKV COLR CRD CRDR CRLD DBA DBE DC DCN DG ECCS ECN EECW
FCV FDC FDCN FI FIC FPC FT ft GE Automatic Depressurization System; Atmosp Aqueous;Film Forming Foam As Low As Reasonably Achievable "Average Power Range Monitor American Society ofMechanical;Engineers Anticipated Transient Without Scram Browns Ferry Nuclear Plant Banked Position Withdrawal Sequence Boiling Water Reactor Condition Adverse To Quality Report Code ofFederal Regulations Containment Isolation Status System Check Valve Core Operating LimitsReport Control Rod. Drive Control Room Design Review Change Request to a Licensing Document Design Basis Accident Design Basis Earthquake Direct Current Design Change. Notice Diesel Generator
'Emergency Core Cooling System Engineering Change Notice Emergency Equipment Cooling Water Effective Full Power Days Extended Load Line LimitAnalysis, Equipment'anagement System End ofCycle Fahrenheit Flow Control Valve Floor Drain Collector Field Design Change Notice Flow Indicator Flow Indicating Controller Fuel Pool Cooling Flow Transmitter
'foot General Electric heric Dilution System
0 4I O~
e Tennessee ValleyAuthority Browns Ferry ¹clear Plant 1994Annual Operating Report ACROiVYMSLISTIJVG 0
GE SIL GEMAC gpm HELB Hg HPCI HPFP HS HWC ICS IL'RT ISI I-Tabs kv lbs LLRT LOCA LPRM LS LT MCPR ME MG MOV MSIV MSL'RM MSRV MWD/ST MWe MWt NESSD NFPA NMS
- NRC NUMAC NUMARC NUIMG PCIOMR PCIS PCV PER P.I ppb ocument esources Committee ent Recommendations GE Service Information Letter General Electric Measurement and Control Gallons per Minute High-Energy Line Break Mercury High Pressure Coolant. Injection High Pressure Fire Protection Handswitch Hydrogen Water Chemistry Integrated Computer System Integrated Leak Rate Test Inservice Inspection Instrument Tabulations Kilovolt Pounds Local Leak Rate Test Loss ofCoolant Accident Local Power Range Monitor Level.Switch Level Transmitter Minimum Critical Power Ration Moisture Element Motor Generator Motor Operated Valve Main Steam Isolation Valve Main Steam Line Radiation. Monitor Main Steam. Relief Valve Megawatt Days per Short Ton Megawatt Electrical Megawatt Thermal Nuclear Engineering Setpoint and Scaling D National Fire Protection Association Neutron Monitoring System Nuclear Regulatory Commission Nuclear Measurement. Analysis and Control Nuclear Utilities Management and Human R Nuclear Regulatory Commission Regulation:
Preconditioning Interim Operating Managem Primary Containment Isolation System Pressure Control Valve Problem Evaluation Report Pressure Indicator, Parts per Billion 0
il O~
Tennessee ValleyAuthori ty Browns Ferry Nuclear Plant 1994 Annual Operating Rcport ACRONYhfS'LISTING ppm PS psi PT QA RBCCW RbNO3 RCIC RCW RHR RHRSW RM RMOV RMS RPS RPV RSW RVLIS RWCU SBO SCFM SER SGTS SI SJAE SLC SPDS SSP ST TACF TI TIP TPM TRS TVA UFSAR UPS V
VAC VDC Parts per Million Pressure Switch Pounds per Square Inch Pressure Transmitter Quality Assurance Reactor Building Closed Cooling Water Rubidium Nitrate Reactor Core Isolation Cooling Raw Cooling Water Residual Heat Removal.
Residual Heat Removal Service Water Radiation Modifier Reactor Motor Operated Valve Radiation Monitoring System Reactor Protection System Reactor Pressure Vessel Raw Service Water Reactor Vessel Level Instrumentation System Reactor Water Cleanup Station Blackout Standard Cubic Feet per Minute Sequential Events Recorder; Significant Events Report (INPO)
Standby Gas Treatment System Surveillance Instruction Steam Jet AirEjector Standby Liquid Control Safety Parameter Display System Site Standard Practice Special Test Temporary&teration Control Form Technical Instruction Traversing Incore Probe Thermal Power Monitor, Temperature Recorder Switch Tennessee Valley Authority Updated Final Safety Analysis Report Uninterruptible Power Supply Volt; Vanadium Volts Alternating Current Volts Direct Current Very High Frequency Ih 0
il~
Tennessee ValleyA'uthority Broivns Ferry Nuclear'Plant 1994 Annual Operating Report OPERA TIOiVAL
SUMMARY
1994 OPERATIONAL
SUMMARY
II O~
iki
Tennessee ValleyAuthori ty Browns Ferry Nuclear Plant 199@Annual Operating Report OPERA TIO1VAL
SUMMARY
UNIT 1 Unit 1 remains on administrative hold to resolve various Tennessee Valley Authority (TVA) and Nuclear Regulatory Commission (NRC) concerns.
UNIT2 On January 1, 1994', the unit's power level was a full power (3291 MWtand 1117 MWe).
On April 15, 1994, at 0218 the reactor scrammed on low scram air header pressure.
The reactor remained shut down while maintenance was being performed on a compressed air system.
The reactor was again critical on April 17 at 2233.
After being placed in the run mode at 0226-on April 18, the reactor auto scrammed at 0355 on main steam isolation valve (MSIV) closure due to low reactor pressure.
All bypass valves failed full open and then reclosed; The reactor was critical once again on April 20 at 2241 and at 100% power by 2015 on April22.
Reactor coastdown began on July 24, 1994, with reactor shutdown starting October 1 for Following the refueling outage, control rod withdrawal began at 0507 on November 21, 1994, and the reactor declared critical at 0620 the-same day.
The reactor was manually scrammed and the turbine tripped at 0101 on November 28 to allow for the turbine to be balanced.
The reactor was critical at 2242 on November 30 but scrammed again on December 2.at 0717 due to stator coolant system instrumentation failure.
Criticality was achieved. again by 2210 the same day.
On'December 31, 1994, Unit 2 was at 3291 MWt and 11'14 MWe.
UNIT3 Unit 3 remains on administrative hold to resolve various TVAand NRC concerns.
il~
4I~
Tennessee ValleyAulhority Browns'Ferry Nuclear Plant 1994 Annual Operating Report OPERA TIDAL
SUMMARY
Docket No.: 50-259 OPERATING STATUS l.
2.
3.
4 5.
6.
7'.
9; 10; Unit Name: Browns Ferry Uriit One Reporting Period:
Calendar Year 1994.
L'icensed Thermal Power'(MWt): 3293 Nameplate Rating (Gross MWe): 1'152 Design:Electrical R'ating,(Net MWe)::1065 Maximum:Dependable Capacity (Gross MWe): 0 Maximum Dependable Capacity (Net MWe): 0 If Changes Occur in Capacity Ratings (Items Number 3 Through 7)
Since'Last Report, Give Reason:
N/A Power Level to Which Restricted, if,any. (Net MWe): 0 Reason-for Restrictions, ifany: Administrative Hold Year December to, 1994 Date Cumulative*
12.
13 14 15 16.
17 18 19 20.
21 22 23.
HoursinRe ortin Period Hours Reactor Was Critical Reactor Reserve Shutdown Hours Hours Generator On Line Unit Reserve Shutdown Hours Gross Thermal Generation'Wh Gross Electrical Generation h
Net Electrical Generation h
Unit Service Factor Unit Availabilit Factor Unit Ca acit Factor C Net Unit Ca acit Factor ERNet Unit Forced Outa e Rate
'0 0
, 0 0
'0 0
i0 0
0 0
0 0
0
, 0 0
0 0
0 0
0 0
'0 0
0 0
0 95743 59521 6997 58267 0
168066787 55398130 53796427 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 ofStartup:
To Be Determined
- Excludes hours under administrative:hold (June 1, 1985 thru end ofreporting period)
0 i
il~
Tennessee ValleyAuthori ty Browns Ferry 1Vuelear Plant l994 Annual Operating Report OPERA TIONAL
SUMMARY
Docket No;: 50-260 OPERATING STATUS l.
Unit Name: Browns Ferry Unit Two 2'.
Reporting Period:
Calendar. Year 1994 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.
If Changes Occur in Capacity Ratings (Items Number 3 Through 7)
Since Last Report, Give Reason:
N/A 9.
Power Level to Which Restrictedif any (Net MWe): N/A 10.
Reason for Restrictions, ifany: N/A 12 13 Hours in Re ortin Period Hours Reactor Was Critical Reactor Reserve Shutdown Hours 744 8760 729 7310' Year December to 1994 Date Cumulative*
122071 82127 14200 14 Hours Generator Gn Line 711'234
'79859 15 16 17 18.
19 Unit Reserve Shutdown Hours Gross Thermal Generation Gross Electrical Generation h
Net Electrical Generation h
Unit Service Factor 0
0 0
21'48854 22621314 231168915 82.6 65.4 95.5 726260 7535260
,'76743178
, 708446 7345174 74589836 20 21.
22 23 Unit Availabilit Factor Unit'Ca acit Factor C Net Unit Ca acit 'Factor ER Net Unit Forced Outa e Rate 95.5 89.4 89.4 2.7 82.6 78.7 78.7 2.2 65.4 57.4 57.4 17.3 24.
Shutdowns Scheduled Over Next 6 Months (Type, Date, and Duration ofEach):
None 25.
IfShutdown at End'of Reporting Period, Estimated Date ofStartup:
N/A
- Excludes hours under administrative hold (June 1, 1985 to May 24, 1991) 0 il~
Tennessee ValleyAuthority Brains Ferry Nuclear Plant 1994 Annual Operating,Report'PERA TIONAL
SUMMARY
Docket No.'. 50-296 OPERATING'STATUS l.
2.
3.
4 5.
6.
7.
8'.
9.
10.
Unit Name: Browns Ferry Unit Three Reporting Period:
Calendar Year 1994 Licensed Thermal Power (MWt): 3293 Nameplate Rating (Gross MWe): 1152 Design Electrical, Rating (Net MWe): 1065 Maximum Dependable Capacity (Gross MWe): 0 Maximum Dependable Capacity (Net MWe): 0 If'Changes Occur in Capacity Ratings (Items Number,3 Through 7)
Since Last Report, Give Reason: N/A Power Level to Which Restricted, ifany (Net MWe): 0 Reason, for Restrictions, ifany: Administrative Hold Year December to 1994 Date Cumulative~
12.
13 14.
16.
17 18 19 20.
21'2.
'3.
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-MWh Net Electrical Generation, h
Unit Service Factor Unit Availabilit Factor Unit Ca acit Factor C Net Unit Ca acit Factor ERNet Unit Forced Outa e Rate 0
0 0
0 0
0 0
0 0
0 0
0 0
0 0
'0' 0
0 0
, 0 0
0.
0.
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.
If.Shutdown at End ofReporting Period, Estimated Date ofStartup:
To Be Determined
- Excludes'.hours under administrative hold (June-1, 1985:thru end'of reporting period)
Cl 4N
(
l Tennessee ValleyAuthority
SUMMARY
OF Browns Ferry Nuclear Plant SAFETYEVALUATIOjVSFOR I994Annual Operating Report CORE COMPOiVENTSAND OPERA TliVGLIMITS 1994
SUMMARY
OF SAFETY EVALUATIONS FOR CORK COMPONENTS AND OPERATING LIMITS 0
~
0 Tennessee ValleyAuthority
SUMMARY
OF Browns Ferry. Nuclear Plant SAFETYEVALUATIONSFOR 1994 Annual Opcratlng.Rcport CORE COMPOlVENTSAND OPERA TINGLIMITS Unit Z:Core. Operati'ng LimitsReport (COLR)
Descri tion/Safet Evaluation This safety evaluation supports Revision 2 of the TVA Browns Ferry Nuclear Plant (BFN)
Unit 2 COLR.
The COLR contains the operating limits for the cycle determined 'from the reload licensing analyses as documented, in the Supplemental Keload Licensing Report.
Revision 2 of the COLR was made to incorporate revised'nd-of-cycle (EOC) minimum critical power ratio (MCPR) operating limits,to allow a more;bottom peaked EOC exposure distribution than assumed'in the original licensing analyses.
The revised MCPR'limits were determined by General Electric (GE) from a reanalysis of the limiting transients; The reanalyses were performed using NRC approved methods and design bases.
Operation within the revised MCPR limits.incorporated in Revision 2 ofthe COLR willensure the MCPR safety limit specified in the Technical Specifications is not v'iolated during any anticipated operational transient.
Shutdown-margins reported in the licensing analyses for Cycle 7,are adequate to accommodate. any effects of the more bottom peaked exposure distribution ensuring Technical Specification limits on shutdown margin are met.
By operating within the established. limits, the more bottom peaked exposure distribution for Cycle
.7 will'ot reduce the margin of safety as defined in the bases for any Technical Specification.
Since the core operating limits:are contained in the COLR, revising the MCPR limits does.not require a Technical Specification change.
These changes will need to be incorporated into Appendix N ofthe Updated Final'Safety Analysis Report (UFSAR).
- No unreviewed safety question is involved.
Unit 2 COLR Descri tion/Safet Evaluation This safety evaluation supports the BFN Unit 2 Cycle'8 reload core design, and the cycle specific updates to the BFN Unit 2 COLR'.
The reload core design and licensing analyses 'for this cycle were performed by 'GE with results documented in the'Supplemental Reload, Licensing Report.
Operating limits for the cycle (i.e., Linear Heat Generation Rate, Minimum Critical Power Ratio, and Maximum I
0 P
4I~
Tennessee ValleyAuthori ty
~
SUMMARY
OF Bro>vns Ferry Nuclear Plant SAFETYEVALUATIONS FOR 1994 Annual Operating Report CORE COMPONEiVTSAND OPERA TINGLIMITS Average Planar Linear Heat Generation Rate) as determined by the licensing analyses are incorporated into the TVABFN Unit 2 COLR.
The BFN 2 Cycle 8 core is a control cell core design with a predicted full power life of approximately 8900 megawatt days per short ton (MWD/ST) (equivalent to about 415 effective full power days (EFPDs)).
Increased core fiow, feedwater temperature reduction, and coastdown capability increase this to a maximum cycle burnup of approximately 9720 MWD/ST or 453 EFPDs assuming a 95% capacity factor is achieved (note:
Cycle 8 is scheduled to operate from November 1994 to March 1996).
The fresh fuel types are GE7B and GE9B designs which are the same types as were loaded in Cycle 7.
both are barrier cladding designs and have no Preconditioning Interim Operating Management Recommendations (PCIOMR) restrictions.
The remaining twice-burnt fuel and once-burnt reinsert fuel from Cycle 6 does not contain barrier cladding and all PCIOMR constraints remain in efFect for these bundles.
The core will'also include the 4 Westinghouse QUAD+ demonstration assemblies which were previously loaded in Cycles 6 and 7.
0 The cycle. is analyzed for Extended Load'Line LimitAnalysis (ELLLA),Increased Core Flow, FFWTR, and Feedwater Heaters Out of Service.
The cycle is also analyzed for Banked Position Withdrawal Sequence (BPWS) rod movement.
The BPWS procedure must be followed in order to stay within the'licensed Rod Drop Accident design basis.
Cycle 8 is designed for aggressive spectral shift operation.
Spectral shift can extend full'ower operation by increasing the void content (spectrum hardening) during the first part of the cycle which increases plutonium production in the upper part of the core.
Spectrum hardening is enhanced with operation at lower flow rates and by using rod patterns to obtain more bottom peaked power distributions.
No control blades or local power range monitor (LPRM) strings were replaced during the Fall 1994 outage.
The BFN Unit 2 Cycle 8 reload core design is acceptable from a nuclear safety standpoint.
Due to necessary revisions to.the UFSAR and the BFN Unit 2 COLR, a safety evaluation was required. No Technical Specification revisions are required.
No unreviewed.safety question is involved.
Ik gQ>>
Tennessee ValleyAuthority
SUMMARY
OF Browns Ferry Nuclear Plant SAFETYEVALUATIONS FOR I994 Annual Operating Report CORE COMPONEiVTS AND OPERA TINGLIMITS Core Component Design Change Request No. 54 Descri tion/Safet Evaluation This safety evaluation addresses the use of modified original equipment control rods in BFN Units 1, 2, and 3.
The control rods are being modified to replace the rollers with low cobalt spacer.pads.
The original equipment control rod roller and pin materials are cobalt bearing Stellite 3 and Haynes Alloy 25 respectively.
The replacement spacer pad materials are low cobalt Inconel X-750 (spacer pads) and PH13-8Mo (retaining ring). The modification is being made for as low as reasonably achievable (ALARA)purposes to remove a large contributor of cobalt to the reactor coolant system thereby reducing dose rates to site personnel.
The control rod blade contains rollers at the top and bottom to guide the control rod as it is inserted and withdrawn from the core.
Only the top rollers were replaced.
The original rollers together with the portion of the retaining pins inside the rollers were removed and replaced with low cobalt spacer pads.
The spacer pads consist of two halves threaded together and independently locked together by a snap ring. Each halfofthe assembly consists of a round washer conically tapered to a flat.contact surface that interfaces with the fuel channels.
The thickness of the assembled spacer pads is the same as the diameter of the rollers they replace.
It should be noted that the described modifications were performed'n irradiated blades by remote underwater operations.
The spacer pad design affords. easier and faster replacement over the original pin and roller design under these conditions.
GE procedure BNI-SWP-009 describes the modification process.
The modified control rod configuration was interchangeable with existing control rod assemblies and was compatible with existing nuclear steam system hardware.
The use of modified original equipment control rods having the upper rollers replaced with spacer pads is acceptable from-a nuclear safety standpoint.
Changes to UFSAR Section 3.4.5.1.1 and.Figure 3.4-4 are necessary.
The control rod modification does not significantly affect control rod reactivity worth, scram insertion performance, or drop velocity.
There is no impact on shutdown margins, scram times, or operating limits and no changes to the Technical Specifications are required.
No unreviewed safety question is involved.
il O~
Tennessee ValleyAuthority
SUMMARY
OF Browns Ferry Nuclear Plant SAFETYEVALUATIONSFOR L994 Annual'Operating Report FIELD COMPLETED PLANTMODIFICATIONS 1994
SUMMARY
OF SAFETY EVALUATIONS FOR FIELD COMPLETED PLA.NTMODIFICATIONS
!1 il 0
Tennessee ValleyAuthority
SUMMARY
OF Browns Ferry Nuclear Plant SAFETY EVALUATIONS 'FOR 1994 Annual Operating RePort FIELD COMPLETED PLANTMODIFICATIONS Safety Evaluations or Unreviewed Safety Question Determinations (USQDs) for the following plant modifications, which were field completed during 1994 were summarized in previous Annual Operating Reports.
Therefore, they are not included'in this report.
ECN/DCN No.
P0161 P0533 P0596
'P0652 P0706 P0730 L2079 P3023 P3092 P3104 H7054 W13294 W14155
'W15365 Descri tion Remove Automatic Initiation Opening Logic from RCIC Steam Line Valves - Unit 3 Install Torus Tem erature Monitorin S stem - Unit 3 Control Rod Drive (CRD) Flow Control Valve Re lacement - Unit 3 Replacement ofFlow Control Valve (3-FCV-71-40 with Pneumatic 0 crated Soft Seated Check Valve - Unit 3 Replacement ofPower Supplies for Analog Trip System-Unit 3 ModifyResidual Heat Removal (RHR) Head Spray-Unit 3 Re lacmentofOx enandH dro enAnal zer-Unit3 Re lacement ofPressure, Switches -.Unit 3 Replacement ofFlow Indicating Switches 3-FS-74-50 and - UEt 3 Replacement ofFlow Transmitter (FT) 3-FT-73-33 with Environmentall uallified Transmitter - Unit 3 Modification to Packin Confi uration - Unit 2 Re lacement ofDoor Interlocks - Unit 0 Installation ofCarrier Heat Pump for the Instrument Maintenance Shop - Unit 0 Revision 1 (approved 07/21/94):of this safety. evaluation was,prepared to address the changes initiated by F28698.
This Field Design Change Notice (FDCN) removes the requirements to replace the existing circuit breaker trip unit in compartment 2C of480V service building main board with a GE Radiation Monitoring System (RMS)-9 unit.
Calculation ED-N0215-910084 R14. was issued to show that the existin EC-1 tri device is ade uate.
Conduit, Cable, and MultiplierEquipment Setting for Unit.3 Process Computer - Unit 3 Revision 1 ofthis safety evaluation was prepared to delete the revision level for Safety Assessment SABFEDCN910032.
See Annual Operating Report for Year 1989 1988 1988 1988 1988 1989 1988 1988 1988 1989 1989 1990 1992 1992
!I 0
0
Tennessee ValleyAuthority
SUMMARY
OF BroNns Ferry Nuclear Plant SAFETYEVALUATIONS FOR 1994 Annual Operating Report FIELD COMPLETED PLANTMODIFICATIONS ECN/DCN No.
W15724 W16710 W1 6713 W16726 W16960 W17040 W17041 W17044
'.17057 W17082 W17133 W1 7215 W17251 W17252 Descri tion Upgrade Evacuation Alarm System, Code Call, and Paging S stems - Unit 3 Uninterruptible Power Supply (UPS) Building - Unit 0 Revision 1 ofthis safety evaluation was issued to addr'ess changes made by F21845.
The changes to the safety evaluation included a format change, reference change, and minor editorial changes due to a revision to the Change Re uest to a Licensin Document CRLD.
Contractor Facilities - Unit 0 Control Room Design Review (CRDR) Modifications for Panel 3-9 Unit 3-Revision 1 (approved 08/05/93) ofthis safety evaluation was issued to address the removal ofposition indication from Panel 3-9-4 for the recirculation loop equializer valves and to delete reference to.a recently deleted CRLD This was considered a minor chan e.
CRDR Modifications for Panel 3-9 Unit 3 CRDR Modifications for Panel 3-9 Unit 3 CRDR Modifications for Panel'3-9 Unit 3 CRDR Modifications for Panel 3-9 Unit 3 Revision 1 (approved 06/02/94) ofthis safety evaluation clarifies implementation restrictions and interim configuration associated with this DCN. This was considered a minor revision.
CRDR Modifications.for Panel 3-9 Unit 3 CRDR Modifications for Panel 3-9 Unit 3 CRDR Modifications for Panel 3-9 Unit 3 CRDR Modifications for Panel 3-9-54 and 3-9 Unit 3 Installation ofMain Control Room Workstations - Unit 0 Revision 1 ofthis safety evaluation incorporated the changes associated with FDCN F20395 which deleted the very high frequency (VHF) radio console from the Unit 0 workstation but retained the capability to reinstall the console at a future time. Additionally, the FDCN installed conduit, added utilitypower and installed signal cables from the Unit 1 operators desk to the common area workstations.
Installation ofMain Control Room Workstations - Unit 2 See Annual Operating Report for Year 1992 1992 1992 1992 1992 1992 1992 1992 1992 1992 1992 1992 1992 1992 0
Tennessee ValleyAuthorily
SUMMARY
OF Browns Ferry Nuclear Plant SAFETYEVALUATIONS FOR l994 Annual Operating Report FIELD COMPLETED PLANTMODIFICATIONS ECN/DCN No.
W17257 W17310 W17347 W1 7427 W17447 W17514 W17536 W1 7545 W1.7725 W17904 W1 8207 W1 8209 W18554 W18685 W18812 Descri tion Control Ba Elevation 593'ir Conditionin
- Unit 0 Replacement ofObsolete GE Measurement and Control (GEMAC) Transmitters with-Rosemount Transmitters-Unit 2
'otor Operated Valve (MOV)Thrust Requirements-Unit 2 Revision 1 ofthis safety evaluation (approved 10/13/94) was prepared to address the changes initiated by F31688.
F31688 deleted valve FCV 2-FCV-78-68 from W17347 and from the GL 89-10 program.
This valve is normally
'losed, stays closed for all design basis earthquakes (DBEs) and is not required for any DBE. Therefore, this valve does not have an active safety function as defined in GL 89-10 and the GL 89-10 scoping calculation has deleted this valve from the sco e ofGL 89-10.
CRDR Modifications for Panel 3-25 Unit 3 Replacement ofFuel Pool Cooling (FPC) Pump 1/4" Seal Water Line with Stainless Steel Tubing and Add Throttle Valve-Units 1, 2, 3 Replacement ofDrywell Control AirSystem Dewpoint Tem erature Monitorin Loo s - Unit 2 Modifications to D, ell Platform - Unit 3 Recirculation Ringheader, Risers, Safe Ends, and Jet Pump Instrumentation Nozzle Safe Ends Re lacement - Unit 3 Addition ofStation Batte No. 5 - Units 1, 2, 3 Fire Alarm and Detection S stem U rade - Unit 0 Installation ofHydrogen Water Chemistry (HWC) System
- Unit 2 Installation ofHWC S stem - Unit 2 Modification to Transformer TS3E - Unit 3 Reroute/Re lace Cables - Unit 1, 2, 3 Modifications to Condensate Transfer System Piping-Unit 2 See Annual Operating Report for Year 1992 1991 1992 1992 1992 1992 1992 1992 1992 1992 1992 1992 1992 1992 1992
0 0
ili
Tennessee ValleyAuthority
SUMMARY
OF Browns Ferry Nuclear Plant SAFETYEVALUATIOlVSFOR l994 Annual Operating Report FIELD COMPLETED PLANTMODIFICATIONS ECN P0112
- Average Poiuer Range Monit'or (APEM} Sitnulaterl T1rernra1 Trip Modification - Unit3 Descri tion/Safet Evaluation This ECN added a thermal power monitor (TPM) to the APRM subsystem of the neutron monitoring system (NMS). The addition ofthe TPM was in respons'e to the operating history ofBoiling Water Reactors (BWRs) which shows numerous scrams resulting from momentary neutron flux spikes.
These spikes are typically caused by power increase anomalies such as disturbances in the recirculation
- system, disturbances during large flow control load maneuvers, transients during turbine stop valve tests, etc.
During a power increase transient, the neutron flux leads the reactor thermal power because offuel time constants.
This situation can result in neutron flux trip levels before the reactor thermal power has increased significantly, thereby causing an unnecessary reactor scram.
The flux.spikes typically represent no danger to the fuel since, they are only one or two seconds in duration and are less than the 120% fluxtrip limit.
The TPM should'liminate this problem by providing a measurement whi'ch is more representative of the reactor thermal power during a transient than the previous design.
The TPM utilizes the APRM output signal as its input and provides an output signal which closely approximates the average heat flux (thermal power) during a,transient or steady-state condition.
This is accomplished by the use of'a time constant which is representative of the fuel dynamics.
This time constant is sufficiently long so that flux spikes such as those described above are averaged over,a longer time period and hence do not result in, the generation ofa trip signal from the TPM.
The addition of the TPM resulted in the replacement of the flow-referenced APRM trip function with a flow-referenced TPM trip, but preserved the existing trip channel on straight APRM flowutilizing a nonflow-referenced set.point.
The APRM signal which previously was an input to a'flow-referenced trip unit would now provide the input to the fixed.120% trip unit and also to the TPM time constant module.
Total recirculation drive flow is used to provide the flow referencing to the TPM.
The TPM then provides a signal to the flow-referenced APRM thermal power trip unit. A channel trip would result from either a fixed APRM trip unit or the flow-referenced TPM trip unit.
The Technical Specifications already incorporate the necessary changes to reflect this ECN, however, the UI'.SAR requires revision to reflect the design. No unreviewed safety question is involved.
Cl 0
ili
Tennessee ValleyAuthority
SUMMARY
OF Browns Ferry Nuclear Plant SAFETY EVALUATIONS FOR l994Annual Operating'Report FIELD COMPLETED PLANT MODIFICATIONS L '"' ':""':'
"'%P""
ECN P0244-Replacetnent ofDifferentialPressure Transmitters - Unit 2 Descri tion/Safet Evaluation The ECN replaced differential pressure transmitters PDT-68-65 and PDT-68-62 ofthe reactor water recirculation system.
It also removed load resistors that are not needed when Rosemount transmitters are used and changed the. installation requirement from seismic Class I to seismic Class II.
The original Foxboro Model 611DM transmitters were replaced with Rosemount Model 1151 HP 7B22PB transmitters.
The affected components are not safety related and.no function of the reactor water recirculation system is changed.
No other system or equipment is effected.
No Technical Specification changes are required.
The UFSAR is not affected.
This change is acceptable from a nuclear safety standpoint and no unreviewed safety question is involved.
ECN P0$11 - Replacement ofReactor Building Emergency Lighting Transfornter - Unit 2 Descri tion/Safet Evaluation This ECN replaced the reactor building emergency lighting transformer with a newer model due to discontinuance of the existing older model.
This ECN also relocated the transformer from reactor building Elevation 608.3'o Elevation 593'.
The new transformer will perform the same function as the present transformer.
The transformer was seismically qualified to prevent damage to safety-related equipment.
No Technical Specification changes are required.
No unreviewed safety question is involved.
4k O~
ili
1 Tennessee ValleyAuthority
SUMMARY
OF Browns Ferry Nuclear:Plant SAFETYEVALUATIONS FOR 1994 Annual Operating, Report FIELD COMPLETED PLANTMODIFICATIONS 4~ %:9:;:" ' iP,;; ~;:;. yj;,;. ~:,< "W ':x ~"..~.'..Y. w 'X.:;x
. ' >'w',qw, '
ECN P0583 Revision I'-Individual Test Points for3-FCV-71 Unit 3 Descri tion/Safet Evaluation The testable check valve (CKV) on the injection line to the RPV had two test connection points, one on. each side of.the valve.
Originally, the test lines were connected to one common test point. This ECN removed the inter-tie piping to valves74-536 and 71-534 and
,provides"separate lines with individual test points. It also changed the valve number prefixes for 74-536 and 74-535 to 71-536 and 71-535, respectively.
Those modifications will correct the. mislabeled. valve tags and'improve the:testing offeedwater valve 3-568 where the reactor core isolation cooling (RCIC) connects to the feedwater system.
It will also allow a temporary pressure hose to be installed across.FCV 71-40 to remove any differential pressure when cycling the valve for a test.
A revision to the Technical Specification. is not required.
The margin, of safety as defined in the Technical'pecification is not reduced.
UFSAR Figure 4.7-1A requires a revision to indicate that the change is applicable to Unit 3 only.
No unreviewed safety question is involved.
DCN P0597
- Replacement and Modification of Reactor Feedivater Punip Floiv Instrunientation - Unit3 Descri tion/Safet Evaluation This ECN,provided the documentation to replace and modify the. reactor feedwater, pump flow instrumentation (FT-3-6, -13, -20, and FI-3-6,. -13, -20) located in.the reactor, feedwater system.
'The existing nonsafety-related flow transmitters, Bailey Meter Company Type 555 weighing 23 lbs. were replaced with lighter (12 lbs.).-Rosemount Model 1151DP transmitters and their indicating range was increased to enable a greater flow rate indication.
This change allows the operator to:monitor the actual feedwater flow of any two feedwater
.pumps, while the other one pump is out of service.
With this change, the feedwater pumps can provide.a flow ofwater to the reactor, equivalent to approximately 90% of normal flow during fullpower operation and stay within'the operating range ofthe flow instrumentation.
The setpoints for opening the:minimum, recirculation valve and the.low reactor vessel level signal to ramp the pumps to a speed corresponding to 75% offull power are not affected.
il~
il~
Tennessee ValleyAuthorlty
SUMMARY
OF Bra Nns Ferry iVuelear Plant
'SAFETY EVALUATIONS FOR e
1994 Annual Operating Report FIELD COMPLETED PLAIVTMODIFICATIONS The feedwater system is not safety related and's not required for. safe shutdown of the reactor.
The instruments do not impact the seismic qualifications of the panels they are located on,',because the new transmitters are placed in the old-location," occupy approximately the same space, and.are approximately half the weight ofthe replaced old;transmitters.
- Also, they do not degrade any Class 1E system.
No Technical. Specification change is required.
However, Section 1'l.8'.3.1 of the UFSAR is aQected.'o unreviewed safety question is.involved; ECN P0737-Replace AirLock.Electrical Penetration- 'Unit3 Descri tion/Safet Evaluation This ECN, replaced existing penetration with one meeting environmental qualification.
The penetration contains cables for airlock light,. telephone, and door status circuits. A failure of
,this-penetration would cause leakage into secondary containment ifthe inner air lock door was open during a'loss of coolant accident (LOCA) or High-Energy Line Break (HELB) inside primary containment.
The new electrical penetration is qualified to American Society of Mechanical Engineers (ASME) 'Code Section III-,pr'imary containment penetrations requirements.
Also, the new penetration is welded into the existing bulkhead sheaves and are seismically qualified.
No Technical Specification change is required; No unreviewed safety question is involved.
ECN POS$2-Modification to Offgas Gas'Reheater Effluent Moisture Loop - Unit'3
'Descri tion/Safet Evaluation ECN P0852 deals with a modification.to the oQgas system, and involves instruments which measure the moisture content of the eftluent gases from the gas reheater before it enters the prefilters and the charcoal adsorbers.
The original scope of ECN P0852'was to replace the I
I 0
ggl
Tennessee ValleyAuthority
SUMMARY
OF Browns Ferry 'Nuclear Plant SAFETYEVALUATIONS FOR l994Annual Operating Report FIELD COMPLETED PLANT MODIFICATIONS moisture sensor ME-66-110, transmitter MT-66-110, and recorder TRS-66-108 in Units 1, 2, and 3 due to equipment obsolescence and unavailability of spare parts.
- However, the transmitter was later taken out ofthe ECN scope and'the ECN only provided the design for the replacement ofthe recorder and the moisture element.
The ECN has been implemented in Unit 2.
The replacement ofthe recorder 3-TRS-66-108.is addressed by CRDR DCN W17057. DCN F18188 was generated to remove the recorder from the scope of ECN P0852 for Unit 3.
Hence, for Unit 3, the scope ofECN P0852 involves changes associated with the replacement ofthe existing moisture element, 3-ME-066-0110, with a-moisture element with an expanded range.
DCN F21442 was generated to address the document deficiencies and the downscoping of ECN.P0852 for Unit 3.
This change involves a modificati'on to the offgas system which is considered part of the radwaste
- system, i.e., gaseous radwaste.
The modification does not impact offgas system functional or operational characteristics, therefore nuclear safety is not decreased.
The modification is in a section of.the Unit 3 offgas system that is fully isolable from the other units and not required for the function ofeither ofthe other units.
No Technical Specification change is required.and the UFSAR is not affected.'his modification is acceptable from a nuclear safety standpoint.
No unreviewed safety question is involved.
ECN L1937 - Installnlion ofBnckwasIs Connections on Che Core Sprny nnrl AFAR Room Coolers - Vnit' Descri tion/Safet Evaluation This ECN provided 1" flush connections along with associated isolation valves for the RHR room coolers, core spray room coolers, and the core spray motor bearing coolers.
The flushing lines were installed between the cooler and their respective isolation valves.
These modifications willfacilitate, from a maintenance perspective, the connection of the backwash lines and the isolation and flushing ofthe coolers.
The new connections willprovide a means of accomplishing the cooler cleaning program without cutting and welding pipe each time a cooler is cleaned.
!I O~
't
~i
Tennessee ValleyAuthori ty
SUMMARY
OF Browns Ferry'Nuclear Plant SAFETYEVALUATIONSFOR 1994 Annual, Operating Report FIELD COMPLETED PLANT MODIFICATIONS K': '
"'-" ':-""""'-'""-'":"'"""'"'""'i"""
The new flushing connections along with associated
'isolation valves are safety related and designed,to TVA Class P'requirements.
The manual'lushing connection isolation valve on the inlet and outlet line of each cooler are normally closed, thus maintaining the integrity of the emergency equipment cooling water (EECW).
Since the new connections'are used only for maintenance, and on an annual basis, the required flow ofEECW cooling water to each respective cooler to achieve safe shutdown ofthe plant is not compromised.
No Technical Specification revision is required.
No unreviewed safety question is involved.
ECN L2050'- Modifications to Automatic and Manual ControlsforSteam Jet AirEj ectors (SJAEs).- Unit3 Descri tion/Safet Evaluation The automatic controls on, the steam supply to'the,SJAEs are subject to instability during startup transfers from auxiliary boiler steam to nuclear process steam and automatic SJAE switchovers.
As a result, gross excess dilution,steam is often being provided by the SJAEs to the offgas system during these transients.
Therefore, this ECN was'written as a short-term fix to allow manual, operator action to minimize the control instability problem.
(The long term.
fix to be covered by another ECN will involve alleviating the problems with the existing automatic control system.)
The modifications covered'by this ECN constitute an interim fix to the SJAE steam supply pressure controller instability and alleviate the o6gas system operational problems caused by the controller instability.
No safety-related system or function is affected by these modifications.
No Technical Specification changes are required.
No unreviewed safety question is involved.
il~
~5~
Tennessee ValleyAuthori ty
SUMMARY
OF Browns Ferry Nuclear Plant SAFETYEVALUATIONSFOR 1994 Annual Operating Report FIELD COMPLETED PLANTMODIFICATIONS
..;~a",.:.(~~ >~ex"'~a
>~q'
+<<.',v<, x.g~.".'>
.><'w':<'. c..~x;"
ECN P3025 - Replacement ofHigh Pressure Coolant Inj ection (HPCI) Pressure Switches
-.Unit3 Descri ti n/ afet Evaluation This ECN. replaced HPCI pressure switches (PS) PS-73-22A and PS-72-.22B with new pressure switches to meet environmental qualifications.
The new etIuipment meets the same requirements and performs the same function as the original equipment.
No Technical Specification change is required.
No unreviewed safety question is involved.
DCN II4277 - Removal of Thermal Overloads on RHR and Residual Heat Removal Service Wafer (RHRSQ Valves - Unit2 Descri tion/Safet Evaluation DCN H4277 implements a removal of thermal overload protection for selected MOVs by
'bypassing the, thermal overload relays and removing the thermal'verload heater elements from the starters ofthe followingvalves:
~
2-FCV-23-34
~
2-FCV-23-40
~
2-FCV-23-46
~
2-FCV-23-52
~
2-FCV-74-59
~
2-FCV-74-73 This safety evaluation was,revised to address DCN-F33607.
This FDCN removes the special requirements from the safety evaluation and documents the test requirements on the single line drawings for 480V reactor motor operated valve (RMOV) Boards 2A and 2B.
Additionally, the frequency of the testing was changed to coincide with BFN's valve diagnostic testing commitments for NRC Generic Letter 89-10.
Thermal overload protection is intended to provide motor protection.from the harmful consequences of operational overloads and motor stall conditions.
This protection is not intended to,provide, nor will it provide, valve system boundary integrity or flow integrity protection.
The valves involved in this DCN had experienced tripping of their overload 0
iki
<Ii
Tennessee'alley Aulhority
SUMMARY
OF Brains Ferry Nuclear Plant SAFETYEVALUATIONSFOR 1994Annual Operating Report FIELD COMPLETED PLANT MODIFICATIONS
,protective devices as a result of,the valves duty requirement for providing throttling of their respective system flows.
The RHRSW, valves are the throttle valves on. the discharge side of the Unit 2 RHR.heat
.exchangers.
These valves are positioned, by the operator as needed to establish and control cooling water for.the RHR heat exchangers as needed for the applicable RHR operating mode.
The RHR'alves are the test return valves and are normally closed.
These valves require closure. ifthey are not in their normal position at the onset ofa:DBE LOCA. These valves are also required to open in support ofthe RHR system torus cooling mode ofoperation, and they must maihtain their position integrity for long term decay heat removal; The Technical'Specification does not directly discuss~thermal overload protection for MOVs.
- However, assuring the integrity of required MOV. actuations is a necessary element, of assuring the basis for Technical Specifications.
Therefore; assuring the capability of plant operators to position valves when necessary to obtain required system configurations and flows will enhance. the, existing margins of safety..
Thus, this modification did. not reduce the margins ofsafety as defined in the basis for any Technical'Specification.
No Technical Specification revisions are required as a result of this modification.
.figures have been revised to depict'he, removal of the thermal overload, protection as implemented:by this DCN and the text and figures require revision to reflect the testing
,requirement note being added by DCN F33607.
No unreviewed'safety question is involved.
DCNH55S6-Rewiring ofOutput Switch Contacts - Unit 2 Descri tion/Safet Evaluation DCN'H5586 provided wiring changes'in auxiliary instrument room Panel 2-9-19 to allow for surveillance testing".of the drywell differential pressure alarm and drywell-compressor control'oops without violating;the intent ofTechriical;Specification Sections 3.2.F and'3.7.A.6.
The existing circuitry was rewired.to allow the control.and alarm function to-remain intact for one channel while the other channel is being calibrated utilizing a lifted lead to isolate the circuit function under test.
il Oi 4>
Tennessee. ValleyAuthori ty
SUMMARY
OF Browns Ferry Nuclear Plant SAFETYEVALUATIONSFOR l994 Annual Operating Report FIELD COMPLETED PLANTMODIFICATIONS The function of the differential pressure alarm and control circuitry modified by this DCN is not degraded in any manner.
No Technical Specification change is required.
No unreviewed safety question is involved.
DCNH5614-Replacement ancb'or Rerouting of1Vormal LigittingCircuits -. Unit 1 Descri tion/Safet Evaluation This DCN replaced and rerouted two normal'lighting circuits to provide corrective action for CAQR BFP890261P.
This DCN did not'involve modifications to the circuits'-associated end devices.
The cables required modification as a result of cable damage due to settling of the Unit 1 air intake structure.
The cables replaced by this modification have,no safety related function and their replacement t
I has no acct.on the function, operation, nor.qualification of any component,
- system, or structure required to ensure nuclear safety.
No Technical Specification changes are required.
No unreviewed safety question is, involved.
DCN W6846-Reroute Cable to Appropriate Fire Zones - Unit 3 Descri tion/Safet Evaluation This safety. evaluation was revised to incorporate DCN F20203A which provided design to reroute cable 3PP733-I3B out of fire zones 3-3 and 3-4.
This modification only involved cable and, raceway in the Unit 3 reactor building and diesel generator (DG) building.
This DCN did not involve modifications to the sources nor the end devices associated with these cables.
t Cable 3PP733-I was replaced with environmentally qualified cable end'to end and routed in fire area 22, fire zone 3-2, and fire zone 3-1 in a new 4" conduit.
The reroute ofthe cable in ili 4>
l Tennessee ValleyAuthority
SUMMARY
OF Bra wns Ferry Nuclear Plant'AFETYEVALUATIONS,FOR l994 Annual Operating Report FIELD COMPLETED PLANTMODIFICATIONS fire, areas was required to assure Appendix R separation.
Old conduit was removed and existing penetrations spared.
Existing cable 3PP733-I was,retagged as 3ABN15, capped, and abandoned in tray'CE-I.
The replacement of cable 3PP733'-I required'he:opening of breaker number 9 in 4kV shutdown board 3EB.
This disabled, alternate feed to 480V shutdown boards 3A and 3B.
- However, the normal feed to shutdown boards 3A and 3B were not
- aQected, by the implementation ofthis modification and were available to comply with Technical Specification 3:9.C/4.9.C.
During the conduit installation, the breaching of.fire barriers (walls, floors, and seal assemblies) complied with Technical Specification 3.11.G/4.11;G.
The new seals meet the requirements for secondary containment,,flood protection, and fire resistance, as applicable.
There are no impacts on nor potential'changes to the Technical Specifications resulting from implementation ofthis modification. No unreviewed'safety question is involved.
ECNI'7019 - Upgrarle ofReactor.Wnter Cleanup- (RWCU) Sampling Station - Unit 2 Descri tion/Safet Evaluation 4
This modification:was made to upgrade the inline coolant chemistry instrumentation in order to better control coolant chemistry.
This modification replaced or rerouted sample lines, replaced. RWCU sampling panel and chiller, installed an online ion chromatograph and associated computer with a UPS in the RWCU sampling subsystem ofthe sampling and water quality system.
No. Technical Specification change is required.'FSAR,figures will.require revision.
No unreviewed safety question. is involved.
0 Oi
Tennessee ValleyAuthori ty
SUMMARY
OF Bro>vns'Ferry Nuclear Plant SAFETYEVALUATIONS FOR'994 Annual Operating Report FIELD COMPLETED PLANTMODIFICATIONS DCN W1$367- Integrated Contputer System (ICS) Upgrade - Unit3 Descri tion/Safet Evaluation This'DCN provided equipment mounting details, cable routing, and cable terminations,to complete the Unit 3 ICS.upgrade modification. Also included in this DCN was the. removal of the Unit 3 GE4020 Plant Process Computer.
This modification Was. required to support TVA's commitment to the NRC to implement Nuclear Regulatory Commission Regulation (NUREG) 0696 requirements.
The ICS'pgrade modification will ultimately,provide a separate computer system for each unit. This DCN addressed the Unit 3 upgrade only.
The process computer system provides a quick and accurate d'etermination of core thermal performance, improves data reduction, accounting, and logging functions for, both the. nuclear boiler and balance of plant equipment,. and supplements procedural requirements for control rod,manipulation:during reactor startup and shutdown.
The new Unit 3 system performs all current nuclear steam supply system and'balance of plant 'functions provided by the-GE4020
,computer as well as the following additional functions:
~
Safety parameter display system (SPDS)
~
Sequence ofevents Rod scram'time.recording
~
Transient recording analysis
~
'Rod'worth-minimizer No revision to plant Technical Specifications is required as a result of.the-implementation of this DCN. UFSAR Sections 7.16, 7.16.5.3, and Appendix 7;7B require revision to reflect the ICS installation.
In addition, numerous UFSAR figures require revision to.refiect changes made to their TVAsource drawings..
This modification does not result in,a reduced margin of safety as defined in the basis for any Technical'Specification.
No unreviewed'safety question is involved.
4li
Tennessee ValleyAuthori ty
SUMMARY
OF Browns Ferry Nuclear Plant SAFETYEVALUATIONSFOR 1994 Annual Operating Report FIELD COMPLETED PLANTMODIFICATIONS DCN.WI643$ - Reactor Vessel Level Instrumentation System (RVLIS) Instrumentation-Unit 2 Descri tion/Safet Evaluation DCN W16435B was originally issued providing for a periodic backfill of the RVLIS instrument reference leg condensing pots at local Panels 25-5 and 25'-'6. The reference. legs of the RVLIS pressure/level transmitters were to be arranged such that similar reactor pressure vessel (RPS)/emergency core cooling system (ECCS) channels shared the same condensing pot, thus 'channelizing'he condensing pots.
NRC Bulletin 93-'03, "Resolution of Issues Related to Reactor Vessel Instrumentation In BWRs", required a modification to ensure the level instrumentation was of a high reliability for,long-term operation, to be in place following the first cold shutdown after July 30, 1993.
To meet this commitment, the scope of DCN W16435B was reduced to address installing a continuous backfill to each RVLIS instrument reference line.
DCN W16435C was issued as Phase I to add continuous backfill to each condensing pot by injecting backfill from the CRD system into the reference leg headers at Panels 25-5 and 25-6.
Also, only the reference leg for level transmitter (LT) 2-LT-3-53 was moved from condensing pot 3-820 to 3-821, via internal tubing reroute inside Panel 25-5, to ensure that feedwater level control is maintained and high water level trip of feedwater and main turbine is not disabled.
A transient or perturbation of the reference leg could potentially cause a scram or ECCS initiation.
DCN F29466 was written against DCN W16435C to reinstate the arrangement ofthe RVLIS instrument reference legs as originally designed per DCN W16435B.
The scope ofthis work was considered as Phase IIimplementation.
The 'channelization'f the instrumentation to the condensing pots will reduce a perturbation or transient in a reference leg to'actuation of half the RPS/ECCS logic.
2-LT-3-53 was returned to condensing port 3-820 and transmitter 2-LT-3-207 and 2-PT-3-207 reference legs were moved to 3-821.
Calculation and the Electrical Calculation Checklist have concluded that this activity does not afFect the accuracy of the RVLIS instrumentation.
No permanent change to the Technical Specification is required.
A temporary change to the Technical Specification (Temporary Technical Specification No. 343T) was required to allow implementing the modification with reactor head on and fuel.in the vessel.
UFSAR figures willrequire updating as a result ofthis modification.
-2S-
Cl ili i5i
Tennessee ValleyAuthori ty
SUMMARY
OF Browns Ferry Nuclear Plant SAFETYEVALUATIONSFOR 1994 Annual Operating Report FIELD COMPLETED PLANTMODIFICATIONS This change is acceptable from a nuclear safety standpoint and no.unreviewed safety question is involved.
DCN W16809-CRDR Modifications to Pane1 3-9-20'- Vnit3 Descri tion/Safet Evaluation This DCN consisted of modifications to Panel 3-9-20 for resolving identified human engineering discrepancies between. the design of the Unit 3 control room and TVA's human factors standards.
These modifications are applications of human engineering principles to improve man-machine interface characteristics and, thus, enhance operator response during abnormal and emergency conditions ofthe plant.
DCN F25936 deletes the addition of redundant EECW sectionalizing valve position indications originally intended to be installed in Panel 3-9-20.
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 color banding, as applicable;
~
Provided hierarchical and component labeling; and
~
Implemented panel repair maintaining seismic integrity ofPanel 3-9-20.
In addition to the general modifications listed above, this DCN performed the following:
~
Duplicated standby gas treatment system (SGTS) outlet flow indication, 0-FI-65-50B/3 and 0-FI-65-71B/3, from Panel 3-9-25 to Panel 3-9-20;
~
Provided SGTS train operability indication, 0-XI-65-18B/3, 0-XI-65-40B/3, and 0-XI-65-69B/3 at Panel 3-90-20;
~
Duplicated control, indication, and annunciation associated with the service air crosstie valve (Handswitch (HS) 0-HS-33-1A/1 and 0-PA-33-lA/1) from Panel 1-9-20 to Panel 3-9-20;
~
Modified control air header pressure instrument loop 3-P-32-88 to provide a wider range;
~
Added an instrument loop (0-P-33-3) for, indication ofservice air header pressure, pressure indicator (PI),O-PI-33-3A/3, at Panel 3-9-20;
~
Replaced existing indication ofDG cooler high discharge temperature with annunciation at 3-XA-55-20, Panel 3-9-20; l
0 4li
<5>
\\
Tennessee ValleyAuthority
SUMMARY
OF Browns Ferry Nuclear Plant SAFETYEVALUATIONSFOR "1994Annual Operating Report FIELD COMPLETED PLANTMODIFICATIONS
~
Added an instrument loop (0-P-26-44) for indication of high pressure fire protection header pressure, 0-PI-26-44A/3, at Panel 3-9-20;
~
Removed the CO2 system capability for cable, spreading rooms A and B.
Revision 1 ofthis safety evaluation was issued to incorporate changes made by DCN F27244 to the CO2 storage, fire protection, and purging system.
DCN F27244 removes the CO2 system capability for cable spreading rooms A.and B.
The National Fire Protection Association (NFPA) sprinkler modifications installed by DCNs W17821 and W17822 provide adequate fire protection for cable spreading rooms A and B, and the existing CO2 system is no longer required.
These modifications:do not change any Technical Specification requirements.
UFSAR figures and Fire Protection Report figures and text willrequire updating.
These changes do not reduce nuclear safety and no unreviewed safety question is involved.
Descri tion/Safet Evaluation This DCN'onsisted'f modifications to Panel 3-9-6 for resolving identified human engineering discrepancies between the design of the Unit 3 control room and TVA's human factors standards.
These modifications are applications of human engineering principles to improve man-machine interface characteristics and, thus, enhance operator response during abnormal and emergency conditions ofthe plant.
DCN F25936 deletes the addition of redundant EECW sectionalizing valve position indications originally intended to be installed in Panel 3-9-20.
In general, this DCN performed the following:
~
Rearranged and/or replaced control switches, indicating lights, and meters;
~
Provided new labels for all components with improved functional descriptions;
~
Provided hierarchical labels for identification ofsystems and their associated components;
~
Replaced switch handles with tactile and shape coded black handles, as applicable;
~
Replaced switch position escutcheons;
~
Replaced indicating light lenses to conform with BFN standards;
~
Provided color banding for specific scales;
~
Replaced meter scales to conform with BFN human factors design standards; 0
~i
Tennessee ValleyAuthority
SUMMARY
OF Browns Ferry Nuclear Plant SAFETYEvALUATIOlVSFOR l994Annual.Operating Report FIELD COMPLETED PLANTMODIFICATIOlVS
~
Replaced existing analog recorders with functionally identical recorders; and
~
Implemented modifications while maintaining the seismic integrity ofthe panel; These modifications do not change any Technical Specification requirements.
UFSAR figures willrequire updating.
These;changes do not reduce. nuclear safety and no unreviewed safety question is involved.
DCN 8'1 718$ - Installation ofContainntent Isolation Status System (CISS) - Unit'3 Descri tion/Safet
'Evaluation, This DCN installed and interconnected the CISS for Unit 3.
The CISS uses, programmable logic controllers to monitor the position of.primary containment 'isolation system (PCIS) valves.and the status-of PCIS isolation.initiations.
This information is processed to provide a summary ofPCIS isolation completions on Control Room Panel 3-9-4.
In addition, this modification relocated two PCIS logic reset handswitches (16A-S32 and 16A-S33),and four PCIS Group 1 isolation logic status indicating'lights'(16A-DS250, 16A-DS251,. 16A-DS252, and 16A-DS253) from Panel 3-9-5 to Panel 3-9-4.
To make room for the CISS status indications, the drywell floor drain sump flow totalizer (3-FQ-77-6) was relocated'on Panel 3-9-4.
This DCN also modified the PCIS Group 8 (traversing incore probe (TIP)) isolation circuitry and installed a relay and Group 8 reset.pushbutton on Panel 3-9-13.
This modifies the circuitry to remove, the existing auto-reset circuitry on the TIP valves and requires an operator
.to manually reset,a Group 8 isolation before. the TIP ball valves can be opened.
No Technical Specification changes were required.
UFSAR figures willrequire updating as a result. ofthis modification.
By. maintaining, the current design, function, and performance of PCIS control and indication components, it is assured that the installation of CISS will not reduce the margin of safety as defined inthe basis for any Technical Specification.
No unreviewed'afety question is involved.
!I
~i ill
Tennessee ValleyAuthority
SUMMARY
OF Browns Ferry Nuclear Plant SAFETYEVALUATIONS FOR l994Annual Operating Report FIELD COMPLETED PLANTMODIFICATIONS DCN Wl7531 - ADSMorlifictttion-Unit3 Descri tion/Safet Evaluation DCN W17531 made the following changes to the design ofthe ADS:
~
Replaced the existing ADS timer (120 seconds) with a seismically qualified'ime delay relay ofsimilar manufacture;
~
Revised the setpoint of the 120 second timer from 120 seconds + 5 seconds; to 95 + 7 seconds in order to comply with BFN UFSAR Appendix N Section N.6.5.10 and BFN Unit 3 Technical Specification Section 3.2.B;
~
Added an ADS inhibit switch into both trains ofthe ADS initiation logic;
~
Added a time delay relay to each train of ADS logic to bypass the high drywell pressure initiation signal after a low-low-low(Level I) RPV water level signal occurs and the timer times out;
~
Established the setpoint of the time delay relay at 265 seconds in order to maintain an analytical limitofsix minutes.
The modification changed setpoints and electrical design features ofthe ADS function.
The setpoints presently stipulated in BFN Unit 3 Technical Specification Table 3.2;B require revision as a result ofthis DCN. This change is in a conservative direction and supported by calculation and setpoint and scaling documents.
This Technical Specification revision was a requirement for return to operation ofDCN W17531 and its associated FDCNs.
The hardware changes did not remove any safety functions described or inferred in the bases for Technical Specification Sections 3.2.B and 3.6 without proper administrative control.
Additionally, the modifications ensure that a valid design basis event (i.e., main steam line break outside primary containment with loss of high pressure makeup) applicable to BFN Unit 3 can be successfully mitigated without operator intervention.
Therefore, the margin of safety associated with Technical'Specification Sections 3.2.B and 3.6 is in no way reduced.
This modification is safe from a nuclear safety standpoint and no unreviewed safety question is involved.
Cl ili E
il>
Tennessee ValleyAuthority
SUMMARY
OF Brains Ferry Nuclear Plant SAFETYEVALUATIOjVSFOR 1994 Annual Operating Report FIELD COMPLETED PLANTMODIFICATIONS DCN W17803-Change Valve Operating Metliumfrom Water to Air-Unit'3 Descri tion/Safet Evaluation This modification changed the valve actuation medium for 3-FCV-67-50 and 3-FCV-67-51 from hydraulics supplied from the EECW.header to pneumatics from the control air system.
Additionally, the operating logic.for the valves was slightly modified'uch that the valves will no longer cycle on EECW header pressure with a demand from raw cooling water (RCW),
but upon shutting from low EECW header pressure will remain shut and require a manual reset to be reopened.
This change was necessary to reconcile-Condition Adverse to Quality Report (CAQR) BFP900232 which was written to document the failure of these valves to perform their intended safety function of closing in a timely manner upon low EECW header pressure to ensure adequate flow to essential EECW loads.
These valves were operated with raw water from the EECW header as the actuating medium.
The failure of these valves to operate correctly was attributed to silt blockage in the actuator lines'due to the use of raw water for hydraulic actuation.
These control valves are normally closed, backup supply valves to the portion of the RCW system which supplies the Unit 3 reactor building closed cooling water (RBCCW) heat exchanges.
No change to Technical Specifications is required as a result of the implementation of this modification. This modification does require revisions to the UFSAR to properly depict plant configuration.
This modification is safety from a nuclear safety standpoint and no unreviewed safety question is involved.
DCN W18052 - 1VFPA Upgratles To Unit3 React'or BuiklingElevation 639 - Unit3 Descri tion/Safet Evaluation W18052 modified the Aqueous Film Forming Foam (AFFF) system which supplies the fire suppression for the lube oil system of the recirculation motor generator (MG) sets located on the 639'levation ofthe Unit 3 reactor building. All existing pipe, pipe supports, valves, and sprinkler heads from the branch off from the 4" raw service water (RSW) supply header between valves 3-26-1279 and 3-26-1280 was remov'ed and replaced by this modification.
The size of the main header of this AFFF system was increased to 4".
An air supervision system was installed with the modification.
This design change only installed the air supply hardware and pressure switch.
This DCN did not provide the signals and associated alarms for the air supervision system.
The power,
- signals, and associated alarms for the air 0
~I>
Tennessee ValleyAull>ority
SUMMARY
OF Brains Ferry Nuclear Plant SAFETYEVALUATIONS FOR l994Annual Operating RePort FIELD COMPLETED PLANTMODIFICATIONS supervision pressure sw'itch are provided in DCNs W17905 and W17908.
Also, the, smoke and heat detection system which is associated with:the AFFF'preaction system is installed by DCNs W1,7905,and W17908.
In addition to the fire protection hardware modifications, the curb design on the 639'levation was modified to improve the containment of lube oil in case of a spill. The containment of the;lube oil to the curbed areas limits the area required to be covered by the fire suppression system.
This design change was provided'o upgrade the fire suppression system to meet the design requirements ofNFPA codes 16A and 13 and design criteria; No Technical Specification change is required.
This change does not have any impact on the Unit 2 Appendix R analysis nor does it affect the capability, performance, or function of any component important to safety.
This modification is safe from a nuclear safety standpoint.
No unreviewed safety question is involved.
DCN W19260 - Anticipated Transient Witliout Scram (ATWS) Standby Liquid Control (SLC) Modification - Unit 3
'Descri tion/Safet Evaluation DCN W19260 modified the Unit 3 SLC system to comply with the Code of Federal Regulations (CFR),10CFR50;62 ATWS equivalency requirements.
The modification ensures that the SLC system has the capability to inject a borated water solution into the reactor vessel at a flowrate, level ofboron concentration, and Boron-10 isotope. enrichment that willcontrol reactivit'y to-at least the equivalent ofthat resulting from the injection of86 gallons per minute (gpm) of 13 weight percent sodium pentaborate solution with a natural boron concentration within the reactor core ofat least 660 parts per million (ppm).
The modification enriches the SLC boric acid to 92 atom percent Boron-10 and decreases the sodium pentaborate solution concentration to s -9.2 weight percent.
The modification also replaced and rescaled temperature switches 3'-TS-63-3 and -4, rescaled temperature control loop 3-T-63-2, and rescaled:level loops 3-L-63-1 and -1B.
Enriching the SLC boric acid and decreasing the. sodium pentaborate solution concentration improves the ability of the SLC system to bring,the reactor from full power to a cold shutdown,and.brings the system into compliance with the 10CFR50.62 ATWS equivalency requirements as specified, in the Technical. Specifications.
0
~i
~li
Tennessee ValleyAuthoritv
SUMMARY
OF Browns Ferry Nuclear Plant
'SAFETYEVALUATIONS FOR 1994 Annual Operating, Report FIELD COMPLETED PLANTMODIFICATIONS The temperature and level alarm setpoint changes, as well as the heater control setpoint
- changes, are commensurate with the chemical composition changes.and continue to alert the operator of solution temperature and volume changes that might indicate a possible solution concentration change.
Nuclear Engineering Setpoint and Scaling Documents (NESSDs) are issued to ensure the instrument setpoints, scaling, and accuracy requirements are implemented and,controlled.
Increasing the minimum amount of required Boron-10 in the Unit 3.SLC tank adds to the margin ofsafety as defined in the basis ofthe Technical Specifications.
This modification. involved a change to the Unit 3'echnical Specifications.
A design calculation for Unit 3 has determined that 186 lbs. is the minimum amount of Boron-10 needed for injection into the reactor coolant to achieve cold shutdown.
To maintain consistency between all units, the minimum amount of Boron-10 needed, for each unit was calculated-and the most limitingvalue calculated was used for all units. Thus, a change willbe performed for the Units 1, 2, and 3 Technical Specifications.
Also,. revision to the text ofthe UI'SAR is required.
The modifications to the Unit 3 SLC system do not reduce nuclear safety and involves a t
change to the plant as described in the UFSAR.
No unreviewed safety question is involved.
DCN W20206-Removal ofPCIS and RPS Trip Signals from Main Steam Line Radiation Monitors (MSLRMs) - Unit 2 Descri tion/Safet Evaluation DCN W20206 deleted the following safety-related reactor scram and PCIS functions associated with the,Unit 2 MSLRMs:
~
Reactor scram
~
Closure ofthe MSIVs
~
Closure of-the main steam line drain valves
~
Closure ofthe reactor water recirculation loop sample valves Nonsafety-related automatic trips to deenergize the main condenser mechanical vacuum pumps and to initiate closure of the vacuum pump suction line isolation valves will remain active.
ii il~
i
Tennessee ValleyAuthority
SUMMARY
OF Browns Ferry Nuclear Plant SAFETYEVALUATIONS FOR 1994Annual Operating Report FIELD COMPLETED PLANTMODIFICATIOiVS Calculations performed in support of this DCN demonstrate that with these changes, offsite radiation doses will remain well within (<25%) the limits of 10CFR100 and the ability to safely shutdown the plant is not degraded.
The operation and. ability of the MSLRM system to perform its required function (as changed by this DCN) are not adversely affected by this modification. This modification willnot prevent any associated systems from performing their safety-related functions and is therefore acceptable from a nuclear safety standpoint.
Technical Specification changes required to support this modification were approved (Technical Specification Amendment 322) and no additional Technical Specification changes are required.
This DCN requires changes to the UFSAR for Unit 2.
This modification is acceptable from a nuclear safety standpoint and no unreviewed safety question is involved.
DCN W22478 - MSLRMRepltteetnent - Unit3 Descri tion/Safet Evaluation This DCN replaced the MSLRMs with more accurate models.
DCN W22478 also deleted the following safety-related reactor scram and primary containment isolations initiated by the MSLRMs. The replacement of:the MSLRMs is similar to modifications performed on Unit 2 by DCN H1263.
MSLRMs (GE Model 194X629), located in main control room'Panel 3-9-10, were replaced with GE Nuclear Measurement Analysis and-Control (NUMAC) digital radiation monitors.
The followingMSLRMs are affected:
~
3-RM-90-136 MSLRM Channel A
~
3-RM-90-137 MSLRMChannel C
~
3-RM-90-138 MSLRM Channel B
~
3-RM-90-139 MSLRMChannel D MSLRM system operation and protective functions required by the existing design basis were changed by this DCN.
Calculations performed by this DCN demonstrate that the changes made by this DCN will not increase offsite radiation doses above the limits of 10CFR100.
The operation and ability ofthe MSLRM system to perform its required functions (as changed by this DCN) are not adversely affected by this modification.
This modification will not prevent any associated systems from performing their safety-related. functions and is therefore acceptable fr'om a nuclear safety standpoint.
<Ii
1 Tennessee'alley Authority
SUMMARY
OF Browns Ferry Nuclear Plant SAFETYEVALUATIONSFOR f994 Annual Operating Report FIELD COMPLETED PLANTMODIFICATIONS e
This modification did require a Technical Specification change and the system design and functional requirements as described'in the.UFSAR were significantly affected by this DCN.
No unreviewed safety question is, involved since NRC approval was obtained prior to implementation and return to operation ofthis modification.
DCN W22$00 -.Addition of Filter Capacitors to Signal Circuit for Rosemount Transmitters - Unit 3, Descri tion/Safet Evaluation This DCN installed capacitors across GEMAC flow.modifiers (square root converters) in flow loops to filter process noise fluctuations.
The new capacitors were installed in Panels 9-19, 9-29, and 9-38 in the auxiliary instrument room for the feedwater, recirculation, RHR, and radwaste system flow loops.
A similar modification was performed on BFN Unit 2 as an addition to the scope, ofECNP0381 by Field Change Request 86-204. ECN P0381 had been worked and closed for all three units without the filtercapacitors being installed on Unit 3.
The components and associated circuits affected by this modification provide non-safety related indication to the main. control room and signal inputs to the reactor feedwater and recirculation pump control circuitry.
This.modification had no affect on any safety related components or system operability and functions.
Therefore, the change is acceptable from a nuclear safety standpoint.
The components and associated circuits affected by this modification are not listed or described in the Technical Specifications.
Therefore, no Technical Specification change was required. No unreviewed safety question is involved.
DCÃ W22767 - Installation of Helium Leak Test Connections at Conrlenser Vncuum Pumps-Unit3 Descri tion/Safet Evaluation DCN W22767 provided the design to install offgas system air inleakage test connections on the suction and discharge piping ofmain condenser vacuum pumps 3A and'3B.
The new test connections were used to install a portable'helium leak detector on the condenser vacuum I
I 0
ll~
Tennessee ValleyAuthority
SUMMARY
OF Browns Ferry Nuclear Plant SAFETYEVALUATIOJVSFOR I994'Annual OperatingReport'IELD COMPLETED PLANTMODIFICATIOiVS pump piping to conduct offgas system air inleakage testing.
The test connections consist. of a welded pipe nipple with;an isolation valve and a 1/4 turn quick connect'hose coupling and are located':on the suction and;discharge piping of each vacuum-pump between the pump and the inlet and.outlet check valves.
Offgas system air inleakage testing. is performed in accordance with Technical Instruction,(TI) 2-TI-55 to identify sources of air inleakage in systems penetrating
.the main condenser, including fianges, valves, penetrations, and other components exposed'to condenser vacuum.
Installation of the test connections affords an, alternate means of conducting air inleakage testing by allowing use ofthe condenser vacuum pumps to maintain condenser vacuum during testing instead of.the steam jet air ejectors.
Similar test connections have been installed for
,the Unit 2 condenser vacuum pumps by Temporary Alteration Control Form (TACF) 2 87-66 and have, been documented by ECN P5332.
This modificati'on has no affect on the normal operating characteristics of the offgas system nor.does it affect, any safety related equipment.
No Technical Specification change is required No unreviewed'safety question is involved.
DCN T25331 - Installation ofChart Paper and Pen. Supply Cabinets-,Unit 2 Descri tion/Safet Evaluation This DCN permanently mounted two:recorder chart paper and pen:supply cabinets in.a former janitor's closet above the stairwell at. the P-line wall in.the.Unit'2 control room.
By moving the supply cabinets into the room, the total paper quantity was increased to 16 ft.
Combustible load calculations have been revised.'he. fire,severity increases from 57 minutes to 182 minutes.
However, the paper is enclosed in steel cabinets.
Therefore the, fire severity willbe significantly less than the estimated values.
The room is. also protected with a photo-electric smoke detector.
- Hence, a fire would be promptly detected and'nnunciated in the main control room and manual fire extinguishing measures can be taken.
Also, alternative shutdown capability is available for a fire in this area in accordance with A'ppendix R safe shutdown procedures.
The safe shutdown. capability,ofthe plant. is not affected.
II II
Tennessee ValleyAuthority
SUMMARY
OF Bro>vns Ferry Nuclear Plant SAFETYEVALUATIONS FOR 199@Annual Operating Repor't FIELD COMPL'ETED PLAiVTMODIFICATIONS No Technical'Specification change is required.
The Fire Protection Report, Volume 1, willbe revised to reflect the combustible loading change in the area.
No unreviewed safety question is involved.
DCN T2$4$$ - Replacement of Unit 1/Unit 2 Battery and Battery Board Room F1ow Switches -.Unit 0 Descri tion/Safet Evaluation" An exception to the system design criteria down-graded the ventilation system associated with the Units 1'nd '2 250VDC battery rooms to nonsafety related..System evaluations and administrative controls have. been provided to assure the 250VDC batteries can still achieve their safety-related function.
Since, the outdoor. air supply and exhaust ventilation system associated with the Unit 1 and.
Unit 2 250VDC battery rooms,no longer have the protective safety function ofmaintaining a negative pressure with respect to surrounding areas, air.fiow switches associated with that ventilation system was removed by this DCN.
DCN T25455 rewired the control circuits associated with the Units 1 and 2 250VDC battery ventilation system blowers to remove the flow,switches and:the time delay relays and place an electrical interlock between each set of blowers (i.e., between 1A and 1B exhaust blowers, between 1A and 1B supply blowers, between.2A and 2B exhaust blowers, and between 2A and 2B supply.'blowers).
The new electrical interlock willstart the opposite train blower upon the deenergization of the, selected train blower.
The blower train is selected via the existing handswitches.
Airflow difficulties willactivate the existing alarm as before.
DCN T25455 revised the System 31 Equipment Management System '(EMS), mechanical control'iagram 0-47E931-6, mechanical flow diagram 0-47E865-4, electrical.schematic diagrams 0-45E779-18 and 2-45E779-18, and associated connection diagrams.
The subject iristrumentation or. associated'entilation equipment is not described in the Technical Specifications.
Consequently, a change to the Technical Specifications is not required.
II O~
'1 Tennessee'alley Authori ty
SUMMARY
OF Browns Ferry Nuclear Plant SAFETYEVALUATIONS FOR (994 Annual Operating Report FIELD COMPLETED PLANTMODIFICATIONS UFSAR Figures 8.5-7b, 8.5-8a, and 10.12-2a require a revision as
.a result of this modification.
This change had.no adverse aQect on nuclear safety.
No unreviewed safety question is involved.
DCN W2$87$ - Replacement ofDemineralizer Vessel Internals - Unit.2 This DCN involved modifications of the condensate, filter-demineralizer system related to the internal parts and elements ofeach condensate.demineralizer vessels.
The condensate filter-demineralizer system for each reactor unit consists of nine filter-demineralizer:units, a backwash
- system, a precoat
- system, and' body feed system; The condensate filter-demineralizer system is used,to remove ionic and particulate material from feedwater so as to maintain a high reactor feedwater quality. The system-minimizes corrosion products entering the reactor which could afFect fuel,performance and accessibility to primary system-,components, and:reduce the capacity required ofthe RWCU system.
The equipment is also used to protect the primary system against, intrusion of foreign materials, especially chlorides, which could occur due to condenser leakage.
The condensate water requires
- a. high degree of purity in order to meet fuel warranty requirements.
These requirements by GE specify a low.level of. iron content <5 parts per billion (ppb) and a'low level of conductivity <.1 micromho/cm.
A high level of iron in the condensate water may cause plate out on the fuel rods with.subsequent hot spots.
The existing condensate filter-demineralizer system at BFN had become very inefficient (short run time between precoats) w'ith consequent excessive generation ofresin.
Replacement resin and its disposal is expensive.
By implementing this modification, the efficiency, ofthe system willincrease and the run length between each precoat will be extended resulting.'in significant savings in precoat material and resin disposal cost and an overall'reduction.in the handling=of radioactive waste.
This modification installed bigger filter elements but this did not change the function of the system.
40-
C
!I ll
Tennessee ValleyAuthority
SUMMARY
OF Browns Ferry Nuclear Plant SAFETYEVALUATIONS FOR J994Annual Operallng Report FIELD COMPLETED PLANT MODIFICATIONS This change is acceptable from a nuclear safety standpoint.
No Technical Specification changes are required.
A change is required to the UFSAR in regard to the filtration area of the new elements in the demineralizer vessels.
No unreviewed safety. question is involved.
DC'26769-Floor Drain Collector Pump Impeller Changeout - Unit 0 Descri tion/Safet Evaluation This change completed the design work needed to document the installation of the 8 1/4" impellers in the floor drain collector (FDC) pumps in association with Special Test (ST) 89-06.
The original design of these pumps call'ed for 7 7/8" impellers.
The larger impellers were installed as part ofthe ST to determine ifthey would result in improved efficiency ofthe floor drain filters.
This change also corrected the applicable drawings to indicate that the Cation Floe addition line.to the FDC tank is actually l/2" O.D. tubing instead of3/8" as was currently shown.
Implementation of this change does not adversely affect the function or operation of the radwaste system.
This change is intended to increase the capacity ofthe'FDC pumps and also improve the efficiency ofthe floor drain filters. The functions and flow paths ofthe radwaste system remain unchanged.
The portions of the radwaste system aQected by this change ar' nonsafety related and cannot cause an accident.
These portions of the radwaste system are adequately designed for the increased pressures that will'esult from the larger diameter impellers.
This change does not aQect any. information presented in the Technical Specifications.
This change does however aQect radwaste flow diagrams 0-47E830-2 and -3 which are the parent drawings for UFSAR Figures 9.2-3b and 9.2-3c.
This modification does not have an adverse affect on nuclear safety and does not involve an unreviewed safety question.
~
i ili
\\
Tennessee ValleyAuthority
SUMMARY
OF Browns Ferry Nuclear Plant SAFETYEVALUATIOiVSFOR 1994 Annual Operating Report FIELD COMPLETED PLAiVTMODIFICATIONS DCN TZ7897-MollificationofCircuits aml Setpoints on Offgas FloivInstrument Loops-Unit Z Descri tion/Safet Evaluation This safety evaluation addresses the non-class 1E 2-F-66-111'A/B instrument loops that measure and record the Unit 2 offgas system flow rates at the 6-hour'holdup pipe.
This modification removed the low flow alarm associated with 2-FS-66-111B (20 standard cubic feet per, minute (SCFM) decreasing on range 0 to 300 SCFM) -to prevent nuisance alarms from this instrument.
The low setpoint for 2-FS-66-111A was changed from 6 SCFM to 8 SCFM for greater margin at that low flow.
To eliminate process noises displayed on indicator 2-FI-66-111A, a vendor supplied capacitor was replaced with a higher capacitive value.
These instruments do not provide any safety function or any control functions.
The above loops provide only visual indication (local and main control room), recorder output, and alarm annunciation in the main control room.
These instruments do not affect the operation of the offgas system.
No Technical Specification change is required.
No unreviewed safety question is involved; DCN TZ7975 - Settings forLimitSivitch (LS), LS Unit2 Descri tion/Safet Evaluation This DCN revised the settings for the LS-5 limit switches for the Unit 2 MSIVs from 90%
open to 85% open as recommended by GE Service Information Letter (SIL), GE SIL 568, and further discussed in NRC Information Notice 94-08.
In effect, the limit switches will be actuated.with,the valve 1/2" further closed than was currently the case (BFN's MSIVs have total stem travel of 10").
This change does not affect the settings for the MSIV 90% open limit switches (LS-3 and LS-4) that initiate a reactor scram via the RPS.
This change also removes the limit switch setting for the affected switches from the main steam instrument tabulations (I-Tabs) (0-47B601-001 Series) and mechanical control diagram 2-47E610-1-1.
This information willcontinue to be presented on 2-730E927RF-6 and 2-45N2631-3 and -4.
4l ili
'L Tennessee ValleyAuthority
SUMMARY
OF Browns Ferry nuclear Plant SAFETYEVALUATIOjVS FOR'994 Annual Operating Report FIELD COMPLETED PLAiVTMODIFICATIONS r
The LS-5 limit switches affected by this change are provided only, to turn on and offgreen position indication lights.
The settings for these limit:switches are not discussed in,the Technical Specifications.
Therefore, this change does not affect any information presented in the Technical Specifications.
UFSAR Section 4.6 describes-the design ofthe MSIVs including a discussion ofthe function of the'.affected. limit switches and their setting at 90% open.
Therefore, this change does affect this text.
This change is acceptable from a nuclear safety standpoint and no unreviewed safety question is involved.
DCN W2950$ - Install Lube'Oil Purifierfor the Unit3 Main'Turbine Oil Tanks - Unit 3 Descri tion/Safet Evaluation DCN W29505 installed:a permanent turbine lubricating oil purifier for the.Unit 3 main turbine and,reactor feed pump turbine oil'tanks.
Revision A ofDCN W29505 implemented'tage l
of the design change which installed a welded pipe connection and isolation valve on the 4" drain piping for the Unit 3 main turbine oil'tanks.
Revision B implemented Stage 2 of the design change which installed the remainder ofthe modifications including the. lube oil purifier, skid, associated piping and fire detection and suppression features.
This design change met the design, material,.and construction standards applicable to the affected systems and'tructures.
The modification does not affect any safety-related equipment.
This modification is safe from a nuclear safety standpoint.
No Technical Specification change is required. A change to the Fire Protection Report. is required to reflect minor changes to the high pressure fire protection (HPFP).piping serving the main turbine oil tanks and the surrounding area.
No unreviewed safety, question is involved.
0 ik il~
Tennessee ValleyAuthori ty
SUMMARY
OF Bro>vns Ferry Nuclear Plant SAFETYEVALUATIONS FOR 1994 Annual Operating Report FIELD COMPLETED PLANTMODIFICATIONS 5 "-': '--'
'::"':;.""""-"-'-".;-"':-"":--:-"""""':".:: -',:-';:::"'-::=""""'-"",'="':.:"-"" "::.":."=:-!
DCN T30397-.Cltange Feetlback on Recirculation Fioiv Control Loops - Unit 2 Descri tion/Safet Evaluation DCN T30397 was issued,to.delete the speed'feedback signal:for the MG set tachometer to the Error Limiter circuit in,.the recirculationflow control system.
The existing system controlled generator speed using the feedback signal and had exhibited unstable operation at higher pump speeds.
This change eliminates the: speed feedback loop and replaces it with scoop tube position demand feedback.
This w'ill convert the recirculation control system from a speed controller to a scoop tube position controller.
This removes a feedback loop which facilitates self sustaining oscillations initiated'by an electrically or mechanically induced perturbation.
The recirculation flow control system is not discussed in the Technical Specifications or its Bases, therefore, no change is required to the Technical Specifications.
This change revises the description of the recirculation flow control system circuit function and operation discussed in the UFSAR.
This modification does not decrease nuclear safety and no unreviewed safety question is involved.
DCNS30691 - Relocation ofOutboartl Containment Isolation Bountlary - Unit 2 Descri tion/Safet Evaluation DCN S30691 was issued for Unit 2 to change the outboard containment isolation boundary in feedwater
.'line B
(penetration X-9B) which includes RWCU return check valve 2-CKV-69-'579 and CRD return check valve 2-'CKV-85-576 to a single boundary valve, RWCU'return check valve 2-CKV-69-630.
0 This change willimprove 10CFR50 Appendix J Local Leak Rate Test (LLRT) on penetration X-9B.
Leakage on the current containment isolation boundary valve, RWCU return check valve 2-CKV-69-579, usually requires extensive maintenance in order to get it to pass LLRT leakage criteria.
A redundant check. valve, 2-CKV-69-630,. was installed: in this line under DCN W18298 to preclude;single-check valve failure blowdown. of the feedwater system, an environmental analysis concern.
This valve is correctly configured (including installed LLRT test vent/drain connections) to function as a containment isolation valve and replaces two parallel containment boundary leakage
.paths, check valves 2-CKV-69-579 and A4-
k 0
Ik
Tennessee ValleyAulhorny
SUMMARY
OF Browns Ferry Pluelear Plant SAFETYEVALUATIOjVSFOR l994Annual Opernling Reporr FIELD COMPLETED PLANTMODIFICATIONS 2-CKV-85-576. This modification required no changes to any piping system design pressure or temperature ratings nor did it change seismic qualifications of any systems or components.
Check valve 2-69-630 was procured and installed to ASME III, Class I and Seismic I safety-related criteria, thus, this valve meets all design and Quality Assurance (QA) requirements to serve as.a containment isolation valve.
This DCN changed the piping classification downstream of check valves 2-CKV-69-579 and 2-CKV-85-576 to the upstream side of valve 2-CKV-69-630 from TVA Piping Class B to Class C.
The ASME Section XI Inservice Inspection (ISI) boundary was changed to reflect the same boundary.,Containment isolation valve drawings were revised to delete check valves 2-CKV-69-579 and 2-CKV-85-576 from the containment isolation function and add check valve 2-CKV-69-630 as the new isolation valve. These are design document changes only, no field work was required.
No Technical Specification changes are required.
The UFSAR.required revision to document the change in the containment isolation boundary.
0 This change is acceptable from a nuclear safety standpoint and no unreviewed safety question is involved.
DCN T31546-Removal ofLoiv Condenser Vacuum Scram - Unit 2 Descri tion/Safet Evaluation This modification eliminates the possibility of an inadvertent scram from "Low Condenser Vacuum" initiated by pressure switches 2-PS-2-1A/1B/5A/8A. The pressure switch setpoints were reduced by DCN M00074A to 0.8 in Hg (Mercury) vacuum eliminating any sensing capabilities ofthe switches to scram the reactor, but the switch circuits were left intact which could cause inadvertent scrams.
The basis for the turbine condenser low vacuum scram was to provide an anticipatory scram to reduce the reactor vessel pressure
- increase, caused by a turbine trip on low condenser vacuum.
The low vacuum setpoint of 2-PS-2-1A/1B/5A/8A (25 in Hg vacuum decreasing) was selected to initiate a scram before closure of the turbine stop valves on low condenser vacuum, initiated by pressure switches 2-PS-47-72A/72B/73A/73B/74A/74B (setpoint 21.8 in Hg vacuum decreasing).
In the'BFN, Unit 1, 2, 3 Safe Shutdown Analysis, ND-Q0999-910033 R9, no credit is taken for the low vacuum scram anticipatory signal provided by pressure switches 45-
il
n Tennessee ValleyAuthority
SUMMARY
OF Drowns Ferry Nuclear Planl SAFETYEVALUATIOiVSFOR I99IAnnnoi OporotinRRcport FIELD COMPLETED PLAIITMODIFICATIONS I9 o
In 9
n n E 9 9
n T n E
9 n
ER 9
2-PS-1A/1B/5A/8A. The NRC issued Amendments 89, 113, and 118 to BFN Units 1,.2, and 3 Facility Operating. License Nos. DPR-33, DPR-52, and DPR-68.
'These amendments revised;the Technical Specifications requirements, tables,. and bases to eliminate the main condenser low vacuum scram, initiated by pressure switches 2-PS-2-1A/1B/5A/8A;, therefore, no 'Technical. Specification changes are required.
The UFSAR text and. figures require revision to delete these switches from Unit 2.
This change does'not decrease:nuclear safety and no unreviewed safety question is involved.
DCN T31916-Setpoint Change for2-TA-66-1088-Unit 2 Descri tion/Safet Evaluation The offgas system is part of the gaseous radwaste
- system, which. collects and processes gaseous radioactive wastes from the-main condenser air ejectors, the startup vacuum, pumps, and,the. gland seal condensers, and'controls their release-to the atmosphere through. the plant stack so that the total radiation exposure to persons outside the controlled area is ALARA and does not exceed applicable regulations.
Troubleshooting on Work Order 94-10708-00 and Problem Evaluation Report (PER)
BFPER940343 addressed low temperature nuisance alarms that affect the low temperature alarm for loop 2-'TRS-66-108.
GE made the recommendation. that the low temperature alarm (TRS-66-108) be set at 39'F.
This new setpoint allows for better. than design performance of the cooler condenser component upstream of the moisture separator.
This DCN lowers the existing setpoint value of 42'F (2-TA-66-108B), annunciated in 2-XA-55-53, Window 17.
The system'still.functions as, designed but the nuisance alarms are reduced.
Recorder '2-TRS-66-108.and,its associated'oop components are not described in the Technical Specifications.
Therefore, a change is not required to the Technical Specifications.
The UFSAR is not affected.
This change, is acceptable from a nuclear safety standpoint and no unreviewed,safety question is involved.
0 0
0 O~
Tennessee ValleyAuihorlly
SUMMARY
OF Drowns Ferry Nuclear Plant SAFETYEVALUATIONS FOR l994 Annual Operallng Report FIELD COMPLETED PLANTMODIFICATIONS r
p ".:::,'.;,;,,':-.':::,.':,.::;,;,:;:: ',
',.::" '.:,:;,.;:;;::..::;:;.,:;,",";;, ',.;.;:;::;::;:.;,:.".,'.:;..';i,",!:;,,'i; '::;...:.;,;,";::;;,"';;:;:,";,',:,,;;;;:.,';:,;,,;,". '-;,":.,
DCN$32301 - Relocation ofOutboard Containment Isolation Boumlary - Unit3 Descri tion/Safet Evaluation DCN S32301 was issued to change the outboard containment isolation boundary for Penetrations X-9A and X-9B.
For Penetration X-9A, the boundary was relocated from RWCU check valve 3-CKV-69-624 to check valve 3-CKV-69-628: For Penetration X-9B, the. boundary was relocated from RWCU check valve 3-CKV-69-579 and CRD system check valve 3-CKV-85-576 to RWCU check valve 3-CKV-69-629.
Valves 3-CKV.-69-628 and 3-CKV-69-629 were originally installed to preclude a single check valve failure blowdown of the feedwater line, an environmental equipment qualification concern.
However, they are suited for the Appendix J containment'boundary function.
These valves are correctly configured (including installed LLRT,test connections) to function as containment isolation valves.
Additionally, the relocation of the LLRT boundary for Penetration X9-B to check valve 3-CKV-69-629 will eliminate a potential leak path through the CRD system.
This modification did not require changes to any piping system design pressure or temperature ratings nor did it change seismic qualifications of any systems or components.
Check valves 3-CKV-69-628 and 3-CKV-69-629 were procured and installed to ASME, Class I and seismic safety-related criteria, thus they meet all design and QA requirements to serve as containment isolation valves.
This DCN changed the piping classification downstream of check valves 3-CKV-69-579 and 3-CKV-85-576 to the upstream side of3-CKV-69-629 from TVApiping Class B to Class C.
Also, the piping classification downstream ofcheck valve 3-CKV-69-624 to the upstream side of 3-CKV-69-628 changed from Class B to Class C.
The ASME.Section XI ISI boundary was changed to reflect the same boundaries.
This DCN updated containment isolation valve drawings.
These are design document changes only, no field work was required.
No Technical Specification change is required.
This change does acct UFSAR Figure 5.2-22'heet 3 which identifies the Unit 3 containment isolation valves.
This change is acceptable from a nuclear safety standpoint and no unreviewed safety question is involved.
il 0
Oi
Tennessee VnlleyAuthori ty
SUMMARY
OF Drowns Ferry Nuclear Plant SAFETYEVALUATIONS FOR 1994 Annual Operating Report FIELD COMPL'ETED PLANTMODIFICATIOiVS i
DCN T33096-Appentlix R Manual Actions - Unit 2 Descri tion/Safet Evaluation This safety evaluation. was written in support of DCN T33096 which was generated to provide. corrective action. for PER BFPER940763 and in support of an associated revision to Volume 1 of the Fire Protection Report.
The purpose of the DCN was to transmit. revised Appendix R manual action:requirements to the plant which willbe incorporated into the. Safe Shutdown Instructions which are used to safety shut down the, plant in the event of an Appendix. R fire. These, revised, manual actions, simply ensure that the equipment and power alignments relied upon for the. safe shutdown of the plant are available for a fire in any given area of the plant.
The manual actions;consist of aligning normal or alternate power supplies (depending on which. is available for a given area) and manually starting/stopping equipment for certain areas.
These manual actions did not result in any physical changes being made to
.the plant.
The;transfer ofplant equipment to'its alternate power supply is within the original design basis ofthe plant and does not affect the function or operability. ofany plant systems.
No Technical Specification change is required.
Volume I of the Fire Protection Report will require revision to update the Appendix R Safe Shutdown, Analysis and the Appendix R Safe Shutdown:Program.
No unreviewed safety question is involved.
DCN $33326 - Draiving Rei>ision to AlloivExisting Coating on Stainless Steel Torus to Remain in Place-Unit 2 Descri tion/Safet Evaltiation This DCN.revised design drawings to allow the existing coating on stainless steel in the Unit.2 torus to remain in place.
48-
0 0
il~
Tennessee ValleyAuthori ty
SUMMARY
OF Broivns Ferry'Nuclear Plant SAFETY EVALUATIONS, FOR 1994 Annual Operating Report FIELD COMPLETED PLANTMODIFICATIONS, Coating has been found applied to the following stainless steel. structures within the Unit 2 suppression chamber:
~
T-quenchers
~
Main vent bellows
~
Miscellaneous support steel on the catwalk
~
Electrical conduits
~
Electrical junction boxes
~
Small'bore piping and valve bodies The coating, applied to the stainless steel components inside the suppression chamber is Valspar 78, which is. qualified for. design basis accident (DBA) conditions when applied to carbon steel'surfaces.
'Stainless steel does not require any. coating.
The coating that was applied (Valspar 78) is a DBA qualified and approved coating system for Browns Ferry Coating Service Level I use over carbon steel.
It'has not been tested and qualified over stainless steel.
The items. listed above were initially coated with Valspar 78 in the early 1980's time frame and were subsequently, recoated during followup coating activities in the torus. 'However, the stainless steel components were sandblasted.
and coated by qualified individuals using approved procedures for the surface preparation and application ofthis coating on carbon steel.
Due:to the adhesion ofthe. coating to the stainless steel T-quenchers and the other structures and components, the coating will be left in place and DCN S33326 was issued to show this condition. on applicable drawings.
TVA has performed tests and technically justified the existing, coating on stainless steel structure/components within the suppression chamber.
It will not dislodge during any anticipated accident or transient.
Any, disbonding of the coating that might occur will be in the form of small particles that cannot impair the flow of recirculation water through the ECCS strainers.
Consequently, leaving the coating on the. structures/components is, safe from a nuclear safety standpoint and does not require any changes to Technical Specifications.
The UFSAR will require a revision to identify.the coating situation within the suppression. chamber.
No unreviewed safety question is involved.
II ib 0
Tennessee ValleyAuthori ty
SUMMARY
,OF
,BroN ns'Ferry hluclear Plant
,SAFETYEVALUATIONS FOR 1994 Annual Operating Report FIRE PROTECTION REPORT REVISIOJVS wv' ~'~ 'www'~@wow '
a "wit " '
~WwM~~
'~%vs 1994
SUMMARY
OF SAFETY EVALUATIONS FOR FIRE PROTECTION REPORT REVISIONS 0
0 il~
l Tennessee ValleyAuthority
SUMMARY
OF Browns Ferry Nuclear Plant SAFETYEVALUATIONSFOR 1994 Annual Operating Report FIRE PROTECTION REPORT REVISIONS Fire Protection Report Volume 1 Descri tion/Safet Evaluation This safety evaluation addresses a revision'to the SSDP in the Fire Protection Volume 1. This change was made to incorporate the corrective actions as identified in PER BFPER930143.
The corrective actions required review of Section III(Required Safe Shutdown Equipment) and Section V (Testing and Monitoring) of the SSDP.
The review identified safe shutdown components for which testing requirements are either not provided or incorrectly stated.
Appropriate changes to the SSDP were made based on these reviews.
The system descriptions provided in Section VI of SSDP were also revised to reflect the current system configurations.
Section VIIof the SSDP (Revision/Control of the Program) was revised to reflect the current method of controlling changes to the Fire Protection Report.
Changes to the Fire Protection Report are now controlled in accordance with SSP-12.15.
These changes are acceptable from a nuclear safety standpoint.
Fire Protection requirements are not described in the Technical Specifications.
Therefore, no Technical Specification change was required.
UFSAR text and tables are affected by these changes.
No unreviewed safety question is involved.
Fire Protection Report Volume 1 Descri tion/Safet Evaluation This safety evaluation addresses a revision to the SSDP in the Fire Protection Volume 1.
The SSDP portion ofthe Fire Protection Report identifies equipment required to ensure a safe shutdown ofUnit 2 following an Appendix R fire. Section IIIofthe.SSDP provides a listing ofthe equipment required for safe shutdown following and Appendix R fire. This section also identifies the required compensatory measures to be taken ifa required. piece of equipment is not able to,perform its Appendix R function. The technical basis for the compensatory actions described by Section IIIare justified in Section V ofthe program.
0
I Tennessee ValleyAuthorlty
SUMMARY
OF BroNns'Ferry Nuclear Plant SAFETYEVAIUATIONS FOR I994 Annual Operating Report FIRE PROTECTION:REPORT REVISIONS This change was made to clarify the compensatory actions.to,be taken in the event that a component identified by Section IIIof the SSDP is not considered capable, of performing.its Appendix R function, while still. Technical Specification operable.
The.,change will require that when a piece of Appendix R equipment does not meet spacial separation requirements (i;e;, failure to meet 10CFR50 Appendix R Section III.G.2.b or where safe shutdown components are identified outside their designated area),
compensatory measures. shall be established in accordance with Paragraph 9.3.11.G.l.a ofthe Fire Pro'tection Plan.
These changes are acceptable from a nuclear. safety,, standpoint.
Fire Protection requirements are not described in the Technical Specifications.
Therefore, no Technical Specification change was required.
UFSAR text and tables are afFected by these changes.
No unreviewed safety question is involved.
Descri tion/Safet Evaluation This safety evaluation addressed a change to the Fire Protection Report Volume 1.
The change will.update, clarify, erihance, remove discrepancies'and errors in text or tables.
These changes are acceptable from a nuclear safety standpoint.
Fire-Protection:requirements are not described in the Technical Specifications.
Therefore, no Technical, Specification change was required.
UFSAR text and tables are afFected by these changes.
No unreviewed'safety question is'involved.
0 0
0 ll~
0 Tennessee lralleyAuthority SVWDIARYOF Browns Ferry Nuclear Plant SAFETYEVALUATIONS FOR J 994 Annual Operating Report FIRE PROTECTION REPORT REVISIONS m~~,"
~.'.~ "~~ ',".'.',~!'.: r<..",."? ':.'<'<<?'4;., >x'-ii~x':g~,",-',,? < ~.'i~% '::.>:."<':<.
'x<&<":.~~'.x;::.zv:,",.,~ '"x. x 'wpg,..'<w<.':v.
<?0j<"'ire Protection Report Voluine 1 Descri tion/Safet Evaluation This safety evaluation addressed a
change to the firewatch/compensatory measure requirements associated with the following preaction'ystem areas of protection (i.e., those areas containing redundant safe shutdown equipment):
~
1,2,3-FCV-26-77 (Reactor Building General Area Suppression System, Elevations 541'Water Curtain], 565'93', and 621')
~
1,2,3-FCV-26-88 (Recirculation MG Set AFFF'Suppression System, Elevation 639')
Presently, when one or more of the required sprinkler systems is partially or completely impaired (i.e., part or all of the area of protection is declared inoperable),
a continuous/area fire watch is established within one hour for each of the above listed impaired systems (2
systems per unit, 3 units, therefore, possibly 6 individual fire watches).
The responsibilities of these fire watches are to move throughout the area ofimpairment (within the same preaction protection area) once each hour (he/she is not to be stationed in one specific area but is to travel throughout the entire area of impairment, and he/she is to not,leave the area of impairment without proper relief). This connotates that the firewatch is to be "dedicated" to this particular impairment and have no other assigned function, but the firewatch is not restricted in his/her deployment.
For clarification, under the present requirements. of the Fire Protection Report, this firewatch can only act as.a compensatory measure for specified areas under one ofthe a6ected preaction coverage areas (he/she under no circumstances, can act as the required compensatory measure for an area outside the preaction area ofprotection he/she is presently monitoring). This change is to document (in Table 9.3.1.1.B ofthe Fire Protection Report Volume 1) the ability of this firewatch to perform the functions of compensatory measures
.as a continuous/area fire watch for these preaction areas of impairment (as mentioned above) in areas protected by one or more. of the associated preaction sprinkler systems.
This will provide (i.e., crossing. from one area of protection to another in the same unit and/or crossing unit boundaries, iffeasible) to act as compensatory measures and still satisfy the requirement of the Fire Protection Report to monitor each area of impairment at least once per hour.
This change will also address detection systems serving the same areas of protection as indicated above (and will be documented in Table 9.3.11.A of the Fire Protection Report Volume 1).
i
Tennessee ValleyAuthori ty
SUMMARY
OF Brains Ferry Nuclear Plant SAFETYEVAI.UATIONSFOR 1994Annual.OperatingRePort
'FIREPROTECTIONREPORTREVISIONS i
i
'I This change-is acceptable from a nuclear safety standpoint.
This change is administrative in nature and has no impact on the defined margins of'safety in the Technical Specifications.
The existing Technical'pecifications and the UFSAR (referenced over to the Fire Protection.
Report) are adequate in permitting safe implementation of:this change.
The function of the firewatches as defined in this'change are not modified.
No unreviewed safety question is involved.
Fire Protection Report Volume 1 Descri tion/Safet Evaluation This safety evaluation-addressed an increase'in.total combustible loading for the main control room areas.
The increase in combustible loading resulted.due to installation of control room work stations, and carpeting.
The increase in combustible loading, in the main control rooms due to installation of work stations and carpets has,minimal impact on the fire hazards ofthe area.
The change does not a6ect.the, analyzed fire event (Unit 2 Appendix R safe shutdown analysis).
The alternate safe shutdown, capability is;not compromised in.the event ofa control;building fire.
The modifications to the control rooms were accomplished by, DCN W17256.
The existing Technical Specifications are adequate to permit safe implementation of this change.
No unreviewed safety question is involved.'54-
0 0
Tennessee ValleyAulhorily
SUMMARY
OF BroN ns Ferry Nuclear Plant SAFETYEVALUATIONSFOR I994 Annual Operating Report FIRE PROTECTION REPORT REVISIONS Fire. Protection Report Volunie 1 Descri tion/Safet Evaluation The purpose of this safety evaluation was to determine the safety implications of changes described in Fire Protection Change Notice FPR-94003 and affect the description of the sprinkler system.
The revision reflects upgrades to the Unit 3 reactor building sprinkler systems to meet the NFPA code requirements.
The changes have no impact on the Unit 2 Appendix R analysis nor does it affect the capability, performance, or function of any component important to safety.
The modifications to the automatic preaction sprinkler system. in the Unit 3.reactor building were accomplished by DCNs W18048, W,18049, and W18050; The existing Technical Specifications are adequate to permit safe implementation of the change.
No unreviewed'-safety question is involved.
Fire:-Protection Report Volume 1 Descri tion/Safet
'Evaluation The purpose of this safety evaluation was to determine. the safety implications of changes described in Fire Protection Change Notice FPR-94004.
Changes are being made in Section 5:2 ofthe:Fire,Hazard Analysis to reflect the method in which. the adequacy ofthe fire pumps
'is determined.
Flow tests are being performed in lieu of pressure drop calculations to determine the adequacy of,fire pumps.
The Appendix R Safe Shutdown Program (Section III) Compensatory Measures are being revised for.the shutdown board rooms and the cable spreading rooms.
The continuous fire watch requirement is being changed to an hourly roving fire watch since NFPA code compliant fire detection systems have been installed in the shutdown board'ooms and the cable spreading rooms.
This change does-not affect the Appendix R required equipment or other Technical Specification systems.
The existing Technical Specifications are adequate to permit safe implementation ofthis change.
4l
!l~
Tennessee VnlleyAuthority
SUMMARY
OF Browns Ferry Nuclenr Plnnt SAFETYEVALUATIO1VS FOR 1994 Annual Opernti ng Report FIRE PROTECTION REPORT REVISIONS The changes to be implemented by this revision do not involve, influence, or change any system operational characteristics as described in the UFSAR.
No unreviewed safety question is involved.
Fire Protection Report Volume 1 Descri tion/Safet Evaluation This safety evaluation evaluates a
change to revise the combustible loading (heat of combustion per unit area - BTU/ft2), area square footage, and fire severity information for Fire Areas 16, 18, 19, 21, and 25 in the Fire Protection Report.
These changes resulted from a review as part ofthe corrective action to BFPER940949.
The combustible fire loading information in the Fire Protection Report is utilized to assess the magnitude of a potential fire in the
- area, and the corresponding capability of the fire suppression systems to extinguish such fires. The safety evaluation shows that the changes in combustible loading are either insignificant or. present no significant challenge to the fire barriers.
The potential fire is still contained within the fire barriers and does not affect the redundant safe shutdown capability.
The existing Technical Specifications are adequate to permit safe implementation of this change.
No Technical Specifications change is required.
No unreviewed safety question is involved.
0 Oi
Tennessee ValleyAuthori ty
SUMMARY
OF Browns Ferry Nuclear Plant SAFETYEVALUATIONSFOR 1994Annual Operating Report NEW'INSTRUCTIONS/PROCEDUREREVISIONS 1994
SUMMARY
OF SAFETY EVALUATIONS FOR NKWINSTRUCTIONS OR PROCKDURK REVISIONS 0
0 il
Tennessee ValleyAuthority
SUMMARY
OF
,Bro~vns Ferry Nuclear Plant SAFETYEVAIUATIONS FOR 1994 Annual Operating Report NE/VINSTRUCTIONS/PROCEDURE REVISIONS Surveillance Instruction (Sl) 2-SI-4. 7.A.2.n-f - Primary Containment Integrated Leak Rate Test PLRTj Descri tion/Safet Evaluation 2-SI-4.7;A.2.a-f implements the requirements of 10CFR50, Appendix J,
and Technical Specification 4.7.A.2.
The requirements include leak testing primary containment at accident pressure once ever 40 +:10 months during cold'hutdown: condition.
This. leak rate test involves 'measuring temperature,
.pressure, and dewpoint of the:primary containment atmosphere with the reactor vessel vented'ver an 8-hour (minimum) period at a minimum difFerential pressure of 49.6 pounds per squareiinch (psi).
The actual test pressure is 50.8 +
- 4 psi. The data is used to correlate the mass leak rate ofprimary containment atmosphere at accident pressure over a 24-hour period.
This instruction is performed, in conjunction with 2-TI-179 that inspects primary containment, leak tests the core spray pump seals, and bubble tests air leakage paths from primary containment.
The revision to 2-SI-4.7.A.2.a-f willhave no negative efFects on the UFSAR described safety functions.
This SI will ensure the leak tight integrity function of primary containment. as required by the Technical Specifications.
No unreviewed safety question is involved.
2-TI-275D Revision 4-DryivellLeak Investigation.- Chemistry Descri tion/Safet Evaluation This;procedure provides information to assist personnel in investigating unidentified drywell leakage.
The procedure outlines sampling and analyses designed to yield information as.to the sources(s) of drywell'eakage.
The, procedure was revised'o include the use of a tracer chemical" which when injected. at the appropriate point(s) willassist in identifying the source(s) ofdrywell leakage when the usual methods fail.
0 The specified tracer is Rubidium Nitrate (RbNO3).
RbNO3 was selected because of its relatively low impact on plant systems and its prior use as a tracer at, another utility. After injection, samples of drywell sumps are to be analyzed for the rubidium cation by ion chromatography.
The presence of rubidium cation in the drywell floor drain sump will indicate that leakage originated'from.the tested source.
The quantity ofrubidium cation. in the drywell,floor drain sump may allow the quantity of leakage from the tested source to be estimated.
II il 0
1 Tennessee YnlleyAuthori ty
SUMMARY
OF Browns Ferry Nuclear Plant SAFETYEVALUATIOJVSFOR 1994Annual Operating Report JVE0 INSTRUCTIONS/PROCEDURE REVISIONS When the tracer is used, samples ofthe drywell floor drain and drywell equipment drain sumps will be obtained 'by Chemistry (with assistance from Operations) as the sumps are pumped over to radwaste.
Work requests will be submitted to connect temporary sample lines for sampling the sumps, as required.
Injection of the tracer will be performed by Technical Support.
Details concerning the injection apparatus are outside the scope of this procedure.
Work requests willbe submitted to make connection for injection.
The amount of tracer which is to be added'ill result in only very slight increases in conductivity and activity of the reactor coolant.
These increases will be insignificant in comparison with the limits of Technical Specification 3.6.B/4.6.B.
No increase in chloride concentration or change in pH of the reactor coolant, are anticipated.
No damage to 304 stainless steel or Zir'caloy cladding are anticipated.
Therefore, the margin of safety as defined in the basis for any Technical Specification is not reduced.
No Technical Specification changes are required. No unreviewed safety question is involved.
0 Cl il Oi
Tennessee ValleyAuthority
SUMMARY
OF Brois ns Ferry Nuclear Plant SAFETYEVALUATIONFOR
'1994 Annual Operating Report SPECIAL OPERATING CONDITION 1994
SUMMARY
OF SAFETY EVALUATION FOR SPECIAL OPERATING CONDITION 40-
0 0
O~
Tennessee ValleyAuthori ty
SUMMARY
OF Browns Ferry Nuclear Plant SAFETYEVALUATIOlVFOR l994 Annual Operating Report SPECIAL OPERA TING CONDITION Final Feetlwater Temperature Reduction (FFWTR) Operation Descri tion/ afet Evaluation This safety evaluation addresses the use of a mode of operation with FFWTR.
The primary use ofFFWTR is intended to be continuous operation with'FFWTR'at end of cycle; however, FFWTR operation can be used as a contingency mode ofoperation during the cycle.
With normal feedwater heating, a reactor reaches the end offull power capability and begins coasting down in power.
The feedwater temperature for BFN is calculated to be reduced by approximately 45'F at rated power conditions by intentionally valving out the last stage high pressure feedwater heaters.
By reducing the feedwater temperature, the reactor can be kept at full thermal power for approximately 2 weeks after normal end offull power capability.
This mode ofoperation for cycle extension purposes is called FFWTR operation.
The use ofFFWTR operation for cycle extension is described in GE Standard Application for Reactor
In addition, a safety evaluation of the impact of FFWTR operation on cycle dependent reload analysis is required.
The FFWTR operation option has existed for several years and has been utilized by many BWRs.
The preferred method of temperature reduction is to close the turbine extraction steam flow path to high pressure heater string Pl (heaters A-l,B-l, and C-1). However, the analyses in GE Report NEDC-32356P show that the use of other methods, including removal of any single heater or combinations oflow pressure and high pressure heaters, is acceptable as long as the final feedwater temperature is within the nominal 47'F reduction assumed in the analyses.
In addition, the heaters can be removed from service by isolating extraction steam only or by isolating extraction steam, feedwater and condensate.
In all cases, operation with feedwater heaters out ofservice is subject to the requirements ofthe vendor turbine manual; The effects of FFWTR for the BFN units on accident events such LOCA, anticipated transients without scram, containment LOCA loads, the mechanical integrity of the reactor internal components and the feedwater nozzle/sparger fatigue usage have been analyzed by GE.
The impact of FFWTR on plant operating limits and fuel thermal-mechanical performance is cycle dependent and has been analyzed by GE for the current Unit 2 Cycle 7.
Future cycle dependent reload analyses will evaluate the impact ofFFWTR on plant operating limits and fuel thermal mechanical performance.
The impact oFFFWTR operation on transient and accident analyses is most limiting at end of cycle (all control rods fullywithdrawn) and bound the impact ofFFWTR operation during the cycle.
Use of partially inserted control rods decreases the severity of transients.
Since I
4l il il~
Tennessee ValleyAuthority
SUMMARY
OF Broivns Ferry Nuclear Plant SAFETYEVALUATIONFOR 1994 Annual Operating Report SPECIAL OPERA TING CONDITION
- c;,', '"
+'." ",,:'.;':;,
",',",;.;,": i
':~~ i.': ';;:: ';;."+~@:,c'"
.,';,,~~'; ',.~;;:,cm FFWTR operation is intended primarily for cycle extension at end of cycle, end of cycle will be determined by Reactor Engineering and may occur before all control rods are fully withdrawn.
Any use of FFWTR operation before the declared end of cycle is only for contingencies (such as maintenance) and may reduce the value ofFFWTR operation at end of cycle.
Operation above the rated power rod line,. e.g., in the Extended Load Line Limit Analysis (ELLLA),regionofthe power flow map, with FFWTR increases the containment loads during a design basis LOCA. The analyses by GE justify only short term operation in the ELLLA region that should not exceed 14 continuous days per occurrence.
Contingency FFWTR operation may be used more than once'in a cycle ifneeded as the time in the ELLLAregion is not cumulative between separate occurrences.
The impact of FFWTR operation on fatigue of feedwater nozzles and sparger. is cumulative and the time in FFWTR operation must be tracked; Operation is considered to be in FFWTR mode ifthe feedwater temperature is more than 10'F below the normal (all heaters in service) feedwater temperature.
Approximately 600 days ofFFWTR operation can be accommodated in the 16 year refurbishment cycle.
The GE analyses assumed a bounding reduction in feedwater temperature of 47'F at rated power conditions with an allowance of 10'or uncertainty in calculations and monitoring.
Thus, the safety assessment is valid for an observed FFWTR of up to 57'F at rated power conditions.
However, the intentional continuous operation with a temperature reduction greater than 47'F is not justified by the analysis and the feedwater temperature should be brought within the analyzed band.
The use of FFWTR mode of operation is acceptable from a nuclear safety standpoint.
No Technical Specification revisions are required. A modification to 'Section 11.8.3.2 (Feedwater Heaters) ofthe UFSAR willbe. made to reflect the maximum feedwater temperature reduction justified.
No unreviewed safety question is involved.
42-
li 4l 0
Tennessee ValleyAuthority Browns Ferry Ãuclear Plant
SUMMARY
OF SAFETYEVALUATIONS FOR 1994Annual Operating Report SPECIAL TESTS 1994
SUMMARY
OF SAFETY EVALUATIONS FOR SPECIAL TESTS 43-
0 il Oi
Tennessee;e ValleyAulhorny Browni Ferrv Nuclear Plant
SUMMARY
OF SAFETYEVALUATIONSFOR l994 Annual Operating Report SPECEAL TESTS Post Modification Test 2-PMT-BF-066.004-Unit 2 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 of offgas 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 of jumpers 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 addresse'd any necessary lifting of certain 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 liAs 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 of components 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'afety question was involved.
2-PMT-BF-066. 006-Unit 2 Descri tion/Safet Evaluation This PMT was,performed to verify the design functions of the affected components were unchanged and electrical faults were not introduced into the associated component's circuitry after, 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" indicating lights; This procedure 44-
0 il~
Tennessee ValleyAuthority
SUMMARY
OF Drowns Ferry iVuelear Plant SAFETYEVALUATIONSFOR l994 Annual Operating Report SPECIAL TESTS 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 ofthe 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 1, 2, and 3 Descri tion/Safet Evaluation The purpose of this 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 ofadditional 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 of data.
Current probes are clamped over power and signal cables to monitor the levels of EMI emanating from them while the plant equipment they interface with is in operation.
Thes'e 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'oints and data recorded.
Plant handheld radio transceivers and repeaters are keyed on and off'n 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 gath'ered 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 45-
0 0
0>
l Tennessee ValleyAulhorily
SUMMARY
OF Browns'Ferry Nuelenr Plnnl SAFETYEVALUATIONSFOR e
I994'Annuol 0~crating Repori SPECIAL TESTS plant equipment was required for,this data-to be representative ofthe actual EMI environment by plant instrumentation.
This special test was acceptable from'a.nuclear safety standpoint and no unreviewed safety question. was involved..
0 0
~'
Tennessee ValleyAuthority
SUMMARY
OF Broiuns Ferry Nuclear Plant
.SAFETYEVALUATIONSFOR l994 Annual'Operating Report TACFs 1994
SUMMARY
OF SAPKTYEVALUATIONS FOR TACFs 47-
0 0
Tennessee VnlleyAuthority
SUMMARY
OF Browns Ferry Nuclear Plant SAFETYEVALUATIONSFOR 1994Annual Operating Report TACFs I
TACF, 0-94-Z-67 Descri tion/Safet Evaluation This safety evaluation addressed the removal ofthe valve disk from 0-FCV-67-88. Tliis valve aligns/isolates the Al'HRSW pump to/from the EECW system.
This valve also has a limit switch that will align the Al'ump logic to start on. an EECW puriip initiation signal.
This valve had been identified as having a broken disk.
The valve had been removed and all the disk had:been recovered from the system.
This safety evaluation allowed the valve to 'be reinstalled without the disk and'allowed the Al pump to be usedfor EECW but'inoperable for RHRSW,.
This change did. not reduce the margin of safety as defined in the basis for any Technical Specification.
The alignment maintained while this. alteration's in place is an alignment consistent with plant procedures.
'The.alteration:will neither affect the ability ofthe Al or A3 pumps to supply water to the EECW system; nor willit impede'the ability ofthe A2 pump to supply, the RHRSW system.
No Technical. Specification change is required.
No unreviewed safety question;is involved.
TACF 0-94-04-77 Descri tion/Safet:Evaluation As part of the Radwaste Improvement Program, ST 89-06:was performed to optimize water processing of'the radwaste system.
Included in this test was the installation of the following equipment:
A. Pall regimesh elements into the floor drain and waste collector filters B. Addition of accumulators on the discharge and suction of the floor drain and waste collector filter, aid.pumps C. Addition of a cation polymer injection system.
This system includes a skid mounted tank and pump for the introduction of polymer (Betz polymer 1175) into the floor drain filter vessel.
This polymer increases the run:time and the effectiveness of the resin:
This equipment affects only the floor drain filter system.
48-
, II, il 0
~i
Tennessee ValleyAuthority
SUMMARY
OF Broivns Ferry Nuclear Plant SAFETYEVALUATIONSFOR 1994 Annual Operating Report TACFs Initial plant design has filter aid systems and polymer injection systems installed.
- However, the filter aid system was not functional until the pulsation dampers were added to each pump by Item 2 above.
The permanently installed polymer system has never been used due to incomplete wiring on.the pumps.
Addition ofItem 3 above allowed polymer injection from a different injection path.
Based upon the discussion,.injection of filter aid and polymer are covered by existing UFSAR.
This test was successful in optimizing radwaste filter performances.
Based upon. results, Plant Management requested that ST 89-06 remain open to document the physical and operational changes while DCRs were being processed.
The Pall regimesh elements (Item I) were documented under DCN. H7889.
DCR 3579, submitted to document Items B and C, has never been implemented.
It should be noted that plant operating procedures associated with this special'est document operation ofthis equipment.
TACF 0-94-04-77 was written to document Items B and C.
This equipment is not required to support, any margin of safety and is not considered able to impact any margin of safety as defined in the basis ofTechnical Specifications.
No Technical Specification change is required.
This TACF does not have an adverse affect on nuclear safety and does not involve an unreviewed safety question.
Ii il Oi
Tennessee ValleyAuthority
SUMMARY
OF Browns Ferry ¹clear Plant SAFETYEVALUATIONFOR e
I994 Annual Operating Report TEMPORARYSHIELDINGREQUEST 1994 TEMPO
SUMMARY
OF SAFETY EVALUATION FOR Y SHIELDING REQUEST
,I Cb 0
ik~
Tennessee VnlleyAuthori ty
SUMMARY
OF Browns Ferry Nuclear Plant SAFETYEVALUATIONFOR 1994 Annual Operating Report TEMPORARYSHIELDINGREQUEST A safety evaluation was performed since acceptance of shielding packages is in part based on the use of NRC accepted operability criteria. that were intended for evaluation of nonconforming and'egraded conditions.
No unreviewed safety question is involved.
0 0
O~
Tennessee ValleyAuthori ty
SUMMARY
OF Browns Ferry Nuclear 'Plant SAFETYEVALUATIONSFOR I994Annua'1'Operating Report UFSAR REVISIOiVS t
1994
SUMMARY
OF SAFETY EVALUATIONS FOR UFSAR REVISIONS
~,
0 0
O~
Tennessee ValleyAuthority
SUMMARY
OF Browns Ferry Nuclear Plant SAFETY EVALUATIOiVS.'FOR e l994 Annual Operating Report UFSAR REVISIOJVS'FSAR Sections 6;4.1, HPCI System, aml 7.4.3.Z, HPCI System Control and Instrumentation Descri tion/Safet Evaluati'on This safety evaluation was written. in support of corrective action promulgated'y PER BFPER940570.
This PER identified inaccuracies in the, functional descriptions given. for the HPCI system, flow indicating controller '(FIC), 2-FIC-73-33, in.the,UFSAR and the system design criteria.
The HPCI system is designed to automatically start upon receipt'f an initiating signal (low-low reactor vessel level or high drywell pressure) and'inject water. into the vessel at full design flow rate (5000 gpm).
The controls are arranged such that the system can automatically realign to inject into the. reactor vessel to fulfillits safety function, even though the system may initially be operating in the test mode.
When testing the HPCI system, the flow controller could be in either MAN or AUTO with the flow demand signal adjusted to less than full design flow.rate.
If.a HPCI initiation.signal is received while in this mode, the HPCI system would automatically realign to its safety related injection position,
- however, the flow controller. would remain in the mode (AUTO or MAN) and at the flow demand setting, in use at the time the initiating signal was received.
Operator action would be required to place the flow. controller back in AUTO and readjust the flow demand signal for full rated design flow.
Functional descriptions given in both the'UFSAR and, the. design criteria currently state that when in the test mode and,a HPCI 'initiate signal is received, the flow controller would automatically return to the AUTO mode, with flow demand set at full design flowrate.
This change will revise the UFSAR to correctly describe the HPCI flow controller function and the design. criteria,to clarify the design functional requirements as reflected by the current as-built configuration of the HPCI control system.
An additional change to.UFSAR Section 7.4.3.2.4 is being made to correct a typographical error. that describes the Unit 2 level switch configuration which trips the HPCI turbine.,
These are documentation changes only, no physical work willbe done on the plant.
This change does not decrease nuclear safety.
No change is required to the Technical Specifications and no radwaste system(s) are involved.
No unreviewed safety question i's involved.
4i 4$
i iki
Tennessee ValleyAuthority SUhNIARYOF Browns Ferry Nuclear Plant SAFETYEVALUATIONS FOR J994Annual Operating Report UFSAR REVISIONS UFSAR Section 8.3, Transmission System Descri tion/Safet Evaluation This safety evaluation was written in support of CRLD BFEP-EEB-94003 RO.
This CRLD brings UFSAR Section 8.3 up to date with current. transmission line/substation configuration and bulk plant loading conditions. It also updates voltage,and frequency graphs for various events, utilizing data collected during peak conditions in the summer of 1993 as baseline data.
Specifically, it revised the text to reflect Unit 2 only operation rather than 3 unit.operation. It reflects the current configuration of transmission lines, including transmission line crossings, connections to offsite substations, and connections to onsite transformers.
It recognizes the presence of capacitor banks in the 161KV switchyard and describes the revised Section 8.3 figures.
The figure revisions show peak electrical system conditions, electrical system response during a Unit 2 LOCA combined with various electrical equipment
- failures, electrical system response for loss ofa Cumberland Steam Plant unit (one ofTVA's largest 2 units), and for trip of or a fault on the BFN Unit 2 generator, and for a fault on the 500KV bus or transmission t
Ime.
These changes do not reduce the margin of safety as defined in the basis for any Technical Specification.
The UFSAR changes affect the transmission line system, which is non-safety related and is located entirely outside ofthe nuclear plant buildings.
No Technical Specification change is required.
No unreviewed safety question is involved.
UFSAR Chapter S Section 8.10, Station Blackout (SBO)
Descri tion/Safet Evaluation This safety evaluation was written in support ofBFN Unit 2.CRLD BFEP-'EEB-94001 RO.
This CRLD added Section 8.10, Station Blackout, to Chapter 8, Electrical Power Systems, of the UFSAR.
This section summarizes SBO background and the electrical power system's 0
~
~
~
Tennessee ValleyAuthori ty.
SUMMARY
OF Browns.Ferry Pluclear:Plant
'SAFETYEVALUATALONS FOR 1994 Annual Operaling'Report UFSAR'REV1$1ONS capability to provide electrical power in support of coping with a 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> SBO event, at BFN Unit 2.
A SBG Technical Specification is currently an open item. The NRC will.notify BFN,ifa SBO Technical Specification is required.
Therefore, there is no Technical Specification change required'.at the. time ofthis report.
This change to the UFSAR is acceptable from a nuclear safety standpoint.
No unreviewed safety question is involved; UFSAR Section 10. 6, RBCCWSystem Descri tion/Safet Evaluation This safety evaluation addressed a change to the UFSAR:to remove mention of specific parameters/additives and associated limits (listed in Table 10.6-2,, RBCCW System Heat Exchanger Operating Conditions) and replace with a statement that there will be additives to.
minimize corrosion.
The parameters/additives and associated limits are listed in Site Standard Practice SSP-13.1, Chemistry Program.
The UFSAR change-will eliminate the need to revise the UFSAR every time SSP-13.1 is revised to incorporate the most current industry information for corrosion control.
SSP-13.1 reflects the"most current information/guidance for the operation of the chemistry program based on regulatory requirements,. industry, practices, vendor recommendation, and
-technological advancements.
The SSP'is consistent with the TVANuclear Power Chemistry Manual which superseded the April 1985 DPM.N79E2 as the most up to date.current'source for water chemistry information for BFN.
The information contained in the SSP is more restrictive than that in the section ofthe UFSAR proposed to change.
'This change is acceptable from a nuclear safety standpoint.
No Technical Specification changes are required.
No unrevi'ewed:safety question is involved.
II ili
Tennessee ValleyAulhority
SUMMARY
OF Browns Ferry Nuclear Plant SAFETYEVAIUATIONS FOR l994 Annual Operating Report UFSAR REVISIOiVS UFSAR Section 13;I, Organizational Structure for tite Contluct ofOperations and Section 13.Z, Organization and Responsibility Descri tion/Safet Evaluation This safety evaluation addressed a total revision to UFSAR Sections 13.1 and 13.2 due to major organizational changes and as part of the annual update.
Section 13.1 remains essentially the same except for a new statement which is added as a reference to the Nuclear Power responsibility for preoperational and startup testing programs as discussed in UFSAR Sections 13.4 and 13;5 and BFN Operations has been deleted from Amendment 10, Section 13.1.2 (now discussed in the TVATopical Report).
Section 13.2 organization description has been deleted and has been formatted to reference TVATopical Report, TVANuclear Quality Assurance Plan, and BFN SSP-1.4.
These changes are administrative requirements associated with qualification and training of personnel.
These changes are administrative in nature.
The NRC SER allows TVA to reference the TVA Topical Report in lieu of the organization description normally found in the UFSAR. This change is acceptable from a nuclear safety standpoint.
No Technical Specification changes are required.
No unreviewed safety question is involved.
UFSAR Section 13.6.5.1, Radiological Emergency Plan Descri tion/Safet Evaluation The UFSAR currently states "The plan contains information for control of emergency conditions modeled after those contained in NUREG-0654, Revision 1." In 1992, the NRC issued Regulatory Guide 1.101 Revision 3, Emergency Planning and Preparedness for Nuclear Power Reactors.
The issuance of Regulatory Guide 1.101 allows nuclear power plants to utilize the methodology described within the guide to develop emergency action levels as an alternate method to that described within NUREG-0654 Revision 1.
Following a review of the Nuclear Utilities Management and Human'Resources Committee.(NUMARC)/NESP-007 Revision 2 methodology it was determined that it would be utilized at BFN. Based upon this
- decision, new emergency action levels were engineered and prepared.
A review of the UFSAR revealed that a change would be necessary to allow for the use ofNUMARC/NESP-007 Revision 2 guideline.
This revision provides for the allowance ofNUMARC/NESP-007 Revision 2 as described in Regulatory Guide 1.101 Revision 3 dated 1992.
0 0
iS~
\\
Tennessee ValleyAuthority
SUMMARY
OF Browns'FerryNuclear Plant SAFETYEVALUATIONS FOR 1994 Annual OperatingReport'FSAR REVISIONS This change is acceptable from a nuclear safety standpoint.
A review ofthe Technical"Specifications applicable to emergency planning were:reviewed and no sections were noted effecting this change to the UFSAR.
Therefore, no changes to the Technical Specifications are required.
No unreviewed safety question is involved.
'UFSAR Appendix M, Report on Pipe Failures.outsiile Containment Descri tion/Safet Evaluation This safety evaluation addressed changes to UFSAR Appendix M. These changes willupdate this section of the UFSAR with current information on the"methods used to analyze outside containment pipe failures.
The current method ofanalyzing high energy line breaks outside of containment meets the requirements. of NM&G.0588.
This change is not a change to the facility since no.physical changes to the plant were made.
This change does not affect any existing. plant procedures nor requires any revision to any plant procedures.
'This change cannot reduce the margin of safety as defined in.the, basis. for any Technical Specification since. the analysis in Appendix M uses the existing system design parameters as input and evaluates the effects of pipe failures outside ofprimary containment on the reactor building environment.
This change to the UFSAR does not impact any Technical. Specifications.
Therefore, no changes to the Technical Specifications. are required; No unreviewed safety question-is involved.
41 il
Tennessee ValleyAuthori ty Bro>ms Ferry Nuclear, Plant 1994 Annual Operating Report RELEASE
SUMMARY
1994 REI EASK
SUMMARY
0 ll II
1994 RELEASE
SUMMARY
ANNUALOPERATING REPORT GASEOUS RELEASES LIQUIDRELEASES MONTH JANUARY FEBRUARY MARCH FISSIONS &
ACTIVATION PRODUCTS (CI) 8.98E+00 6.58E+00 7.85E+QQ IODINES (CI)
'.74E-04 6.19E-04 3.17E-04 PARTICULATES
>8 DAYHALF-LIVES (CI) 4.02E-04 7.52K-04 6.97E-04 TRITIUM (CI) 9.21E-01 6.97E-01 8.38E-01 FISSIONS 4 ACTIVATION PRODUCTS (Ci) 2.16E-02 2.14E-Q2 1.60E-02 TRITIUM (CI) 1.17E+Q0 1.28E+00 1.20E&0 DISSOLVED NOBLE GASES (CI) 1.03E-03 6.80E-05 1.19E-03 GROSS ALPHA (CI ND ND APRIL 1.45E+Ql.
1.73E+01 4.07E-04 4.30E-04 5.30E-04 9.12E-04 9.85E-01 1.01E+00 2.01E-02 2,32E-02 2.11K+00 1.31E+00 1.40E-03 ND 9.44E-Q5 ND JULY 7.32E+Ql 2.8QE+01 4.56E-04 3.76E-04 2.86E-04 2.26K-04 1.05E+00 1.31K+00 1.80E-02 2.14E-02 6.93K-01 7.07E-01 4.86E-05 8.71K-QS ND AUGUST SEPTEMBER OCTOBER 3.28E+01 2.37E+01 5.14E-02 3.96E-04 3.92E-04 4.33E-04 4.61E-04 3.29E-04 7.08E-04 1.34E+QO 1.39E+00 9.50E-01 1.91E-02 2.14E-02 5.91E-02 6.22E-01 1.17E+QO 1.42K+00 4.51E-05 ND 4.09E-03 ND 1.03E-03 ND NOVEMBER DECEMBER 1.64E-01 1.31' 6.27E-06 1.49E-Q5 5.42E-04 1.97E-05
~
S.67E-01 3.Q7E-01 9.09E-02 3.05E-02 1;76E+00 1.20E+00 ND 2.3QE-03 ND ND is for non-detectable.
Variation in the data for gaseous releases have been correlated ivith the numbers ofoperating fans. There were no excursion ofinterest nor releases which exceeded ODCM limits.
41 0
il~
Tennessee ValleyAuthority Brains Ferry Nuclear Plant l994 Annual Operating Report OCCUPA TIONALEXPOSURE DATA 1994 OCCUPATIONAL EXPOSURE BATA il ik~
AEXPB2 AUN DATE: 01 - 'I 1 - 95 IIUN 'I IhIE: 11:36:45 T
E N
N E
S S
E E
V A Y
A U f II 0 fl I
'1 Y
BFN RADIATION EXf'OSUfIE SYSfEM NUMBER OF PERSONNEL AND MAN-AEM BY WORK JOB FUNCTION TOTAL NUMBEA OF INDIVIDUALS NUMBER OF PERSONNEL
(>
100 MILLIREH) 10TAL hlAN-AEM 82 MD=REACTOR OPS SUAVEILLANCE GAOUP STATION UTILITY EMPLOYEES EMPLOYEES CONTRACT TOTAL AND OTHERS PERSONS STATION EMPLOYEES U'TILITY EMPLOYEES CONTRACT 101AI AND 01IIEflS HAN-RFH MAINTENANCE PERSONNEL OPERATING PERSONNEL HEALTH PHYSICS PERSONNEL SUPERVISOAY PEASONNEL ENGINEERING PERSONNEL 204 118 55 45 51 29 9
6 3
6 423 9
24 69 76 656 136 85 117 133
- 10. 184 23.881 7.004 3.732 2.381 1.876 0.797 0.217 0.019 0.038 3.378 0.140 0.728 3
~ 810 1.928 15.438 24. 818 7.949 7.561 4.347 MO
&7.782 2.947 thO=AOU1 INE MAINTENANCE GROUP STATION EMPLOYEES UTILITY EMPLOYEES CONTRACT TOTAL AND OTHERS PERSONS STATION EMPLOYEES UTILIfY EMPLOYEES COtl IAACf 10 fbi.
AND 01IIEAS gAN RPH MAINTENANCE.PERSONNEL OPERATING PERSONNEL HEALTH PIIYSICS PERSONNEL SUPERVISORY PEASOtJNEL ENGINEERING PEASOtJNEL 220 94 56 36 52 29 3
4 3
13 889 13 11 79 89 1138 110 71 11e 15n 21.605 7.220 4.767 1.387 2.659 1.472 0.024 0.037 0.034 0.098 115. 814 0.243 0.086
- 9. 128 3.301 138.89I 7.487 n.'e90 10.5n9 6 058 hlo 1.665 120.572 167.675 MD=IN~ SERVICE INSPECTION GROUP STATION UTILITY EMPLOYEES EMPLOYEES CONTRACT TOTAL AND OTHERS PERSOtJS STATION EhlPLOYEES UTILITY EMPLOYEES COtl IAACI IOIAL AND 0 1 IIL'flS MAN-REl!
MAINTENANCE PEASONtJEL OPERATING PEASONNEL HEALTH PHYSICS PERSONNEL SUPERVISORY PERSONNEL ENGINEERING PERSONNEL MO 21 0
3 2
8 189 6
1 8
69 73 210 6
4 11 86
- 9. 705 0.000 0.015 0.382 4.688 T4.790 0.000 0.000 0.000 0.302
- 8. 161 B.463 80.869 5.830 0.065
- 4. 130
- 76. 117 90.57n 5.830 0.080 4.814 Oe.orr>
lr7 Oll
-IOO Vrz MD=SPECIAL MAINTENANCE GROUP MAINTENANCE PERSONNEL 138 931 1078 STATION UTILITY CONTRACT TOTAL EMPLOYEES EMPLOYEES AND OTHERS PEASOtJS STATION EhlPLOYEES 10.576 UTILI TY COtl IAACI 10 I AL EMPLOYEES AtlD 01IIFAS HPN-REH 0.042 202 'in I 213 959
II 0
0
BF.XPB219 RUN DATE! 01-11-95 BUN TIME: 11:36:45 OPERATING PERSONNEL HEALTH PHYSICS PERSONNEL SUPERVISORY PERSONNEL FNGINEERING PERSONNEL MO TENNESSEE VALLEY AUTHORITY BFN RADIATION EXPOSURE SYSTEM NUMBER OF PERSONNEL AND MAN-REM BY WORK JOB FUNCTION TOTAL NUMBER OF INDIVIDUALS NUMBER OF PERSONNEL
(>.100 HILLIREM) 37 6
5 48 0.532 51 1
4 56 1.347 13 0
75 88 0 '73 20 6
88 114 0.611
~59 22
%TUB TMZ T3&M TOTAL MAN-REM 0.076 0.003 0.000
-0.015 0.067 0.029 8.443 8.080 PAGE:
83 0.675 1.379 8.516 8.706 MD=WASTE PROCESING GROUP STATION UTILITY EMPLOYEES EMPLOYEES CONTRACT TOTAL AND OTHERS PERSONS STATION EMPLOYEES UTILITY EMPLOYEES COtt TRACT 101AL Atto OTIIEBS MAN-REH MAINTENANCE PERSONNEL OPERATING PERSONNEL HEALTH PHYSICS PERSONNEL SUPERVISORY PERSONNEL ENGINEERING PERSONNEL 20 17 8
6 0
3 0
0 0
0 29 1
0 0
0 52 18 8
6 0
0.300 0.452
- 0. 062 0
~ 067 0.000 0.020 0.000 0.000 0.000 0.000 0.450 0.180 0.000 0.000 0.000 0.770 0.632 0.062 0.067 0.000 MD=REFUEL GROUP STATION UTILITY EMPLOYEES EMPLOYEES CONTRACT TOTAL AND OTHERS PERSONS STATION EMPLOYEES UTILITY EMPLOYEES CON tRAC1 T01AL AND 011 tERS HAN-REH MAINTENANCE.PERSONNEL OPERATING PERSONNEL HEALTH PHYSICS. PERSONNEL SUPERVISORY PERSONNEL ENGINEERING PERSONNEL 216 83 47 37 44 30 6
5 3
10 582 12 24 29 90 828 101 76 69 144 59 '37 9 '83 9.178 5.896 5.370 8.819 0.456 0.683 0.195 1.009 165.915
- 3. 104
- 11. 189 2.393 6.792 233.771 12.943 21.050 8.484
- 13. 171 MO
~2T 5Z T89.393 289."4 19 1702 194 3825 5721 202.494 24.393 114.850 9nt.737
I!
~
~
AEXPR219 IlUN DAfE: 01-11-95 llUN T IME:
1 1: 36: 45 T
E N
N E
S S.E,E V A.L L E Y
A U T ll 0 fl I T
Y BFN RADIATION EXPOSURE SYSTEM NUMBEA OF PERSONNEL AND MAN-REM BY WORK JOB FUNCTION TOTAL NUMBER OF INDIVIDUALS PAGE:
84 GROUP NUMBEA OF PERSONNEL STATION UTILITY EMPLOYEES EMPLOYEES
(%
100 MILLIREM)
CONTRACT TOTAL AND OTHEAS PERSONS STATION EMPLOYEES TOTAL MAN-AEM UTILITY EMPLOYFES CONTRACT TOlAL.
AtJO OIIIEAS MAN Rl H MAINTENANCE PERSONNEL OPERATING PERSONNEL HEALTH PHYSICS, PERSONNEL SUPERVISOAY PERSONNEL ENGINEERING PERSONNEL 819 349 220 139 175 1702 100 24 16 10 44 194 3043 46
,64 260 412 3825 3962 419 300 409 631 5721 111. 407 41;468 22.373 11.537 15.709 202.494 12.229 1.353 0.940 0.550 9 '21 24.393 569;067 9.564 12.097 27;904 96.218 714.850 692; 703 52.385 35.410
- 39. 991.
121.248 941.737
II Cl
REXPR219 RUN DATE: 01-11.95 RUN TIME: 11:36:45 TENNESSEE VALLEY AUTHOR I TY BFN RADIATION EXPOSURE SYSTEM NUMBER OF PERSONNEL AND MAN-REM BY iYORK JOB FUNCTION TOTAL NUMBER OF INDIVIDUALS PAGE:
85 GROUP STATION UTILITY CONTRACT TOTAL MAINTENANCE PERSONNEL OPERATING PERSONNEL HEALTH PHYSICS PERSONNEL SUPERVISORY PERSONNEL ENGINEERING PERSONNEL 265 128 56 48 56 553 15 7
1 1
17 41 1055 13 24 85 127 1304 1335 148 81 134 200 1898
ll il~
Tennessee ValleyAuthori ty Broi~nsFerry Nuclear Plant CHALLENGES TO OR FAILURES OF J994 Annual Operating RePort MAINSTEAMRELIEF VALVES 1994 CHALLENGKSTO OR FAILURES OF MAINSTEAM RELIEF VALVES L'
il~
Oi
Tennessee ValleyAuthoriiy Browns Ferry Pluclear Plani CFIALLENGES TO OR FAILURESOF
/994 Annual Operating Report MAINSTEAMRELIEF VALVES UNIT 1 None UNIT2 During the operating cycle, increasing MSRV discharge tailpipe temperature trends were noted.
The tailpipe temperature trends were indications that the MSRV pilot valves were leaking steam.
During the same period; the. safety-related acoustic monitor system indicated essentially no change.
Therefore, the leakage rate was believed,to be minimal.
Due to the proximity. of the discharge temperature thermowells to. the MSRV pilot valve discharge port, minimal.leakage rates will be detected by,the thermocouples.
The existina plant equipment does not.provide sufficient process parameters,to support a detailed calculation to quantify minimal mass flow rates through the MSRV, pilot valves.
Suppression Pool temperature (bulk and:,indiv'idual bays) can be used to determine
@ross leakage rates due to the increased heat load. During the operating period, Suppression Pool temperatures were routine~i monitored.
Unaccounted for heat loads were not observed in either the vicinityofthe'MSRV discharge point or in the bulk. temperature.
In, accordance with Target Rock Corporation Test Report 3892'(dated:
August 5, 1993)
MSRVpilot leakage rates ofabout 200 pounds per hour (ibm/hr) will'result in a deviation of'he MS'ilot valve setpoint by approximately +
1 percent (BFN Technical Specification limit); Calculation of the theoretic bulk Suppression Pool temperature rise can be completed under the following assumptions and conditions:
~
Assume a MSRV pilot valve leakage rate of 100. ibm/hr (one-half of 200 ibm/hr required for +1 percent setpoint change) ofsteam.
~
'Thermodynamic properties based upon saturated steam at 1005 psig with an enthalpv of 1192.BTus/ibm is condensed in the. Suppression Pool at a bulk temperature ofabout 74'F with an enthalpy.42 "BTUs/Ibm.
~
.Assume a Suppression Pool capacity of about 127,000 cubic feet (950,000 gallons) with a specific volume of0.01605 cubic feet/ibm.
~
Assume perfect mixing occurs.
Under these, conditions, the'bulk Suppression Temperature would increase about 1'F every three days:
Localized thermal effects due to non-p'erfect mixing would have intensified'the local bay, temperatures and would have resulted much larger local temperature, change rates.
Therefore, using the bounding condition, of 100 Ibm/hr with no Suppression Pool heating, and no acoustic, monitor. response, there appears to no basis'for classifying the observed pilot valve leakage, as-a pilot valve ailure.
UNIT3 None Ig I
1 1
il O~
Tennessee ValleyAuthority Bra wns Ferry Ãuclear Plant 1994Annual Operating RePort
'REACTOR VESSEL FATIGUE USAGE EVALUATIOiV 1994 REACTOR VKSSKL FATIGUE USAGE EVALUATION 4i c:a 0
0
Tennessee ValleyAuthority Browns Ferry nuclear Plant l99e Anneal OperallnS R'eparl REACTOR VESSEL FATIGUE USAGEEVALUATIOIV The cumulative usage factors for the reactor vessels are as follows:
Location Shell at water line Feedwater nozzles Closure studs Unit 1 0.00620 0.29782 0.24204 Unit 2 0.00572 0.22754 0.22766 Unit 3 0.00431 0.16139 0.14360 A~
0 0
ili