W3P89-2161, 1989 Rept of Facility Changes,Tests & Experiments Per 10CFR50.59
| ML20005D984 | |
| Person / Time | |
|---|---|
| Site: | Waterford  |
| Issue date: | 12/31/1989 |
| From: | Burski R LOUISIANA POWER & LIGHT CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| W3P89-2161, NUDOCS 9001030061 | |
| Download: ML20005D984 (89) | |
Text
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LPeL n=r Loutenona Power O Ught Company
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l' Now orleans, LA 70160 0340
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R. F. Burski Nuclear Saloty & Regulatory Affairs-Manager
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f December 18, 1989-L 3
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U.S. Nuclear Regulatory Commission j
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ATTN:
Document Control Desk I
Washington, D.C.
20555
Subject:
Waterford 3 SES Docket No.:50-382 License No. NPF-38 1989 Repor,t of Facility Changes, Tests and Experiments I
Gentlemen:
g Enclosed "is the 1989 Report of Facility Changes Tests and e
Experiments-for Waterford 3 which is submitted pursuant to 10CFR50.59.
This annual report covers the period from June 19, 1988 through June 18, 1989.
If you have any questions regarding this report, please contact j-L.W. Laughlin, Site Licensing. Support, at (504) 464-3499.
Very truly yours, f
~E RFB/ OPP /pi Enclosure
-cc:
-R.D.
Martin, NRC Region IV
. F.J. Hebdon, NRC-NRR D.L. Wigginton, NRC-NRR NRC Resident Inspectors Office E.L. Blake W.M. Stevenson i
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i 9001030061 891231 An Entergy Company PDR ADOCK 05000302
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i' Louisiana > Power & Light' Company i
Waterford 3 SES
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Docket No. 50-382 License No. NPF-38
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-REPORT-OF FACILITY CHANGES, TESTS AND EXPERIMENTS t
r FOR 1989 PER 10CFR50.59 1
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Louisiana Power & Light Company Waterford 3 SES 10CFR50.59 Annual Report for 1989 TABLE OF CONTENTS I.
FACILITY CHANGES A.
STATION MODIFICATIONS (SMPs)
Report No.
Title 1.
0065 Steam Generator Feedpump Turbine (SGFPT) Lube Oil Transfer Pump Suction Piping Tie-in Modification 2.
0227 Floor Drains / Fire Protection Testing 3.
0446 Access to Top of the Boric Acid Condensate Tanks 4.
0532 Radioactive Filter Conversions 5.
0661 Switchgear Area Ventilation Modification 6.
0813 Boric Acid & Waste Condensate Level Gauge 7.
0876 Main Turbine Vibration Monitoring System 8.
0920 Addition of a Booster Pump to Blowdown 9.
0983 Containment H2 Analyzer Replacement 10.
1004 Liquid Post Accident Sampling System (PASS) Modifications 11.
1009 Electro-Hydraulic (EH) High Pressure l
Filter Differential Pressure Switch i
Impulsing 12.
1087 Replace Various Tank Indicators / Switches 13.
1092 Gland Steam Pressure Taps l
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- 14..1135 Controlled Bulk Chemical Storage Area f
Building i
15.
1186 Loop Seal on Exhaust Header Inlet to-Vapor Extractors 16.
1219 Condenser Outlet Water' Box Level _ Taps
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17.
1251-Installation of Low Pressure Turbine i
LP1_and LP2 Heavy Disc / Key Plate Rotors
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18.
1259 Modify Overflow Pipe on Bulk Chemical Storage Tanks 19.
1299 Gas Surge Tank (GST) Safety Valve Set Pressure 20.
1304 Seal Oil coolers Temperature Control
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Valves-5 21.
1310 Addition of Lube Oil Centrifuge System-to Feedwater Pump / Turbine j
22.
1327 Block Caustic and - Acid Supply Lines to Blowdown Mixed Bed Demineralisers t
23.
1328 Gaseous Waste Management Rad Monitor i
Replacement 24.
1389 Permanent Installation of the Second H2 Gas Dryer 1
25, 1402 Boze Filter Flushwater Flowmeter
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26, 1472 CP-2 Startup Recorder t
27.
1475 Instrument Air Containment Isolation Check Valve Repair i
28.
1480 Security Camera System Upgrade j
f 29.
1494 Annunciator G0710 Alarm Modification 30.
1565 Steam Generator Feed Pump Turbine (SGFPT)
Isolation on High Hotwell Level 31.
1628 Valve Packing Enhancement Program 32.
1654 RCP Seal Replacement -,
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1 33.
1696 Boron Management to Circulation Water Isolation e
34.
1743 Boron Management and Chemical Volume Control Drain Leakage Prevention 35.
1817 Huclear Measurement Corporation Radiation Monitor Replacement 36.
1835 Orbisphere Dissolved Hydrogen Analyzer on Sample Line P6 37.
1979 Fuel Pool Purification Pump Low Discharge l
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Araunciator 38.
1994 Turbine Building Closed Cooling Water Filtration System 39.
2041 Ratio / Relay Panels for EDG Control Panel (A&B) Pressure 40.
2066 Setpoint Change to NG-IPIC-7630 B.
DESIGN CHANGE PACKAGES (DCPs) 41.
3000 Waste Gas Compressor Replacement (Tie-ins) 42.
3003 Fuel Pool Hx CCW Flow Low Alarm Removal 43.
3018 Charging Pump Seal Cooling System Modifications 44.
3021 Turbine Generator Stator Coil Water " Tee" Installation for conductivity and oxygen Checks 45.
3056 Installation of Pressure Bleodoff Valve for Air Accumulator Check Valve Test 46.
3060 Relocation of Seismic Monitor 47.
3062 CVC-109 Replacement 48.
3070 Provided SUPS Feed for Emergency i
L Notification System l
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49.
3094 Fuel Handling Building Shipladder l
R Handrails, at EL +l.00 L
50.
3115 MS Drain Valves (2")
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= PROCEDURE CHANGES S1.
PE-005-001 (Change 1,2,3 Rev. 1)
Integrated Leak Rate 52.
STP-NE-TEM-003 Variation of Feedwater Suction-Pressure
Fi.-
UNT-5-013 (Rev. 1)
SPECIAL EVALUATIONS 55.
Special Issue DCN-HV-124 and Radioactive Release Path Categories 5 6 '.
CI 257780 Add Drain Valves in Low Points of the Main Steam Headers to the Emergency Feedwater Pump Turbine 57.
CI 260568 Restore Emergency Diesel Generator Fuel Oil' Storage Tank Mechanical Float to Service 58.
CI 2C1111 Replace the Local, Audible High-High Level Alarm on the Emergency Diesel Generator Jacket Water Standpipe, with Plant Monitor Computer Alarms 59.
DRN M 8800750 Delete the Locked-Open Requirement from Valves MS-403A and MS-403B 60.
DRN M 8800918 Add a Drain to Sample Line r-Downstream from Blowdown Domineralizer Conductivity Cells 61.
DRN M 8801398 Make Normal Position for Valve BAM-141 Consistent 62.
DRN M 8801624 Show Condensate Pumps' Gland Leakage Drains l
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DRN M 8801290 Update the Schematic Diagram for the Component Cooling Water Surge Tank Level Instruments
- 64. ~PEIR 70923 Certifying Carbon Steel Y-Strainers in the Chilled Water Pumps' Suction Lines 65.
TAR 88-25 Add Waste Gas Compressor Isolation Valve at its Suction and Discharge Headers, Instead of at the Compressor Inlet and Outlet 66.
TAR 88-26 & 88-38 Defeat the Automatic Closure-Interlock 67.
TAR 88-37 Cap Drain Line (7SIl-250)
Downstream of Valve SI-231A 68.
TAR 88-44 Remove One Defective Heated Junction Thermocouple Heater String from Service 69.
TAR 88-45, Adding Proximity Probe Brackets
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to Reactor Coolant Pump Seal Flange Bolts
Special Procedure Installing Westinghouse Mechanical Steam Generator Tube Plugs
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-.Waterford 3 SES
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1989 Report of Facility Changes, Tests and< Experiments.
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SUMMARY
zy:
(This1 report provides.the'Waterford 3 Facility Changes mades 2
- pursuant-to 10CFR50.59(a)(1)'. - The-report covers : the period =
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=from-June: 19,31988'through; June 18, :- 19 89. - ' None of,the' items in1
'- this: report--represent an=.unreviewed. safety. question.
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!The report identifies 71' Facility Changes:(40, changes under the-y Station Modification program,'10 under, Design Change packages 4 1
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- procedure. changes'andL17.special' safety evaluations).
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FACILITY CHANGES A.
' STATION MODIFICATIONS (SMPs) 1.
Station Modification Packagei SMP-0065 SMP-0065, Steam Generator Feedp_ ump Turbine (SGFPT) Lube Oil Transfer Pump Suction Piping Tie-in Modification Description of change The modification provided two tie-in lines between the SGFPT Reservoir drain connection and the existing suction lines to the feed pump.
A level gage (sight glass) off
.the drain line was also installed.
Reason for Change The.SGFPT manufacturer provided connections in the Lube Oil Reservoir for continuous lube oil purification and a
1/2" vent hole in the standpipe to the outlet connection to prevent inadvertent siphoning when the lube-oil purifiers are not operating.
-In adapting 'these connections to the present batch conditioning system both outlet and " return" connections are piped to the transfer pump suction header.
The arrangement did not-
. allow complete draining of the. reservoir for ' batch processing of the lube oil.
The~ sight glass was requested by Operations to provide visual indication of oil level in the reservoirs.
Safety Evaluation The modification was classified as non-safety and non-seismic.
The addition of the subject tie-in lines and sight' glass to the SGFPT Lube Oil System had no impact on the nuclear safety aspects of the Plant.
No changes to the Plant Technical Specifications were required. ' The modifications did not increase or create the possibility, probability and/or consequences of an accident, neither new nor previously considered.
No unreviewed safety question was involved with the modification.
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' Station Modification Package, SMP-227 SMP-227, -Floor-Drains / Fire Protection Testing Description of Change SMP-227 rerouted the main drain of sprinkler systems FPM-24 and FPM-27.
FPM-24 was routed to the storm drainage system.
FPM-27 was-routed' directly to the sewage lift station.
The open drip cup drain from each system was routed separately to the floor drain system to ensure a monitored pathway.
Reason for Change The new routing of drains for the two systems was provided to eliminate the flooding condition which occurred during draindown after inadvertent system actuation or flow testing.
The modification will minimize flow to the Radwaste drains.
Safety Evaluation The Fire Water System is a closed system dedicated exclusively to fire protection.
The possibility - of radioactive contamination of the system is highly unlikely.
Therefore, discharge of the riser flow test drain line to the storm water drainage system does not pose a risk to the health and safety of the public.
The drainage systems do not serve a safety function since they are not required to achieve safe Plant shutdown.-
Furthermore, the systems are not required to mitigate or monitor the consequences of an accident.
The work completed by this SMP is considered non-safety related.
The change did not affect the Plant Technical Specifications.
No safety equipment is affected by the modification.
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Station Modification Package, SMP-446 SMP-446, ' Access to Top of the Boric Acid Condensate Tanks Description of Change SMP-446 installed four, twenty-five f t. high ladders with safety devices on the four Boric Acid Condensate Tanks (one ladder-per tank).
It also installed a handrail with too plate around the top of each of the subject tanks.
Reason for Change Valves 7BM-F219,- 7BM-F220, 7BM-F252 and 7BM-F253 located at the top of the Boric Acid Condensate Tanks were not readily accessible for operation and maintenance.
Safety Evaluation The access ladders
.and. handrails added by-this modification were non-safety related and non-quality-related.
The modification does not affect the operation of Boron Management - System or any other system.
No Technical ~ Specification was affected.
The installation is not Seismic. Category-I.
The modification does'not involve an unreviewed safety question.
The installed equipment is-not required for safe shutdown of the Plant or for monitoring the consequences of or mitigation of an accident.
No unreviewed safety question was created by the modification.
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Station Modification Package, SMP-532 SMP-532, Radioactive Filter Conversions Description of Change SMP-532 installed Pall Trinity Filter conversion _. tube
-sheets into Carborundum Filter Vessels -for the Waste
- Filter, Laundry
- Filter, CVC Purification
- Filter, Preconcentrator Filter A and B, and Fuel Pool Filtar.
.The' installations allow - use of Pall Trinity cartridge filter elements for. filtration of all radioactive fluids passing through the stated vessels.
The modification-required installation of filter cartridges rated for less than 20 microns, ' except in the case of the waste and laundry filters which require larger but-less than 100 micron filters.
Reason for Change SMP-532 allows quicker and easier change out of filter elements on a less frequent basis, in accordance with j
the principles of ALARA.
q Safety Evaluation 1
' Changing the waste and laundry filters did not affect any equipment important to safety (i.e.,
not required for l-safe shutdown of the Plant-and not required for j
monitoring or mitigating the consequences of an j
accident).
No Technical Specifications required changes I
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There 1s no decrease to the margin of safety for the Plant.
-The Station i
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Modification did not impact accidents evaluated in the j
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No new accidents were created as a result of the j
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An unreviewed safety question was -not l
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Station Modification Package, SMP-661 SMP-661, Switchgear Area Ventilation Modification Description of Change SMP-661 lowered Switchgear/ Cable Vault area "A"
temperature-setpoint from the range 90 to 104 degree F-to the= range 75 to-80 degree F.
This area is served by AH-30 (3A-SA).
The modification = was accomplished by recalibrating VLL _ and ~ NAC cards of AH-30 temperature loop.
Reason'for Change A problem 'of high temperature around inverters in Switchgear/ Cable Vault Area dA" had been identified.
Safety Evaluation.
Reducing the temperature setpoint improves the environment for equipment in the area and reduces the probability of any failure due to higher temperatures.
The modification assures _that equipment environment in-the affected areas does not. approach worst-case conditions.
The implementation of the SMP did not require: a change to-Technical Specifications.
No impact was made on the basis or margin.to safety of any Technical Specification.
The modification did not result in an unreviewed safety-question.
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Station Modification Package, SMP-813' SMP-813, Boric Acid and Waste Condensate Level Gauge Description of Change SMP-813 replaced the level instrumentation for the Boric Acid and Waste Condensate Tanks.
The new level
' instrumentation' provides the same alarm and control functions ~which are annunciation under high/ low level conditions and stopping the pump on low-low level.
The modification added Dixson indicators with four setpoints.
Foxboro level -indicators were; removed.
Cables were abandoned in place.
Reason for Change The modification was implemented to improve reliability of alarms' and pump shutdown.
The original-level indicator / controllers were experiencing problems with' t
calibration and-setpoint repeatability.
Level indications had been unreliable due to pressure variances caused by the vent gas header.
Safety Evaluation The modifications performed are in the non-safety-related -
Boron and Waste Management Systems.
Neither system.is required for safe shutdown of the Plant..
The modification will not have any impact on the accidents previously evaluated in the FSAR nor create the possibility of new unevaluated accidents.
. The implementation of this SMP does not require a change to the Plant Technical Specifications.
No reduction in margin of safety in the bases of Technical Specifications resulted.
The SMP does not affect safety-related equipment.
No unreviewed safety question is involved with the modification.
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' Station Modification Package, SMP-876 SMp-876,-Main Turbine vibration Monitoring System D_escription of Change SMP-876 L replaced the We'stinghouse Powertronics Turbine Supervisory System. With a
Bently Nevada Turbine Supervisory Instrumentation (TSI) System.
Reason for Change The Westinghouse Turbine Supervisory System had experienced problems and was considered to be an outdated system.
' Safety Evaluation The affected instrumentation is used primarily for diagnostic purposes in balancing the turbine generator and monitoring ' of operating parameters.
Two of the monitored' parameters for. vibration and-differential oxpansion also serve an equipment protection function by tripping the turbine.
All functions of this system, however, are non-safety-related, and their failure would not adversely affect safe Plant shutdown.
The modification will not have impact on the accidents-previously evaluated in the FSAR nor create the possibility of new unevaluated accidents.
The implementation of this SMP does not require a change to the Plant Technical Specifications.
The SMP does not affect safety-related equipment.
No unreviewed safety question is involved with the modification.
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Station Modification Package, SMP-920 p
SMP-920,fAddition of a Booster Pump to Blowdown Description of Change SMP-920 installed a booster pump on:the -35 elevation of the Reactor Auxiliary Building in the Steam Generator blowdown flow path.
Reason for' Change The modification provides increased not positive suction head (NPSH) to the existing blowdown-pumps.
The objective was to increase the capacity of the Blowdown-System.
Safety Evaluation-The modification is non-safety, non-seismic and not necessary to shut down the Plant.
It does not introduce a new path of releases of radioactive-materials to unrestricted areas because it is a loop emanating and terminating at the suction of the existing pumps.
The modification does not interact with safety-related systems and thus its f ailure will not impair the function of any safety-related systems.
The only accident that the Steam. Generator Blowdown System (SGBS) is associated with is the Steam Generator Tube Rupture accident evaluated in Chapter 15 of the FSAR.
The modified portion of the SGBS is a closed loop with respect to the SGBS and does not' have any significant impact on any potential radiological considerations of the system.
The implementation of this SMP does not require a change to the Plant Technical Specifications.
No unreviewed safety question is involved with the modification, 9
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-Station Modification Package, SMP-983 4
- SMP-983, Containment H2 Analyzer' Replacement Description of Change SMP-983 provided replacement of the two Hydrogen Gas Analyzers.
Isolation valves were added to each sample line to f acilitate replacing the analyzers with the Plant on - line.
The isolation valves were locked open or.
closed, as appropriate, in preparation for the new panels that'were later installed.
-Reason for Change The performance of the original Hydrogen. Gas Analyzers was inadequate.
The ' moisture removal devices were inefficient.
Moisture caused by condensation in the sample lines had been drawn through the analyzer.
This happened f requently when the analyzer units were started up.
After several options were investigated to resolve the moisture problems, it was decided to replace the H2 Gas Analyzers.
Safety Evaluation The replacement Containment Hydrogen Analyzer System has been seismically and environmentally reviewed prior to the system-being declared operational.
Control Room Panel CP-33 has been reanalyzed to verify that original seismic qualification was not lost after modifications for new system controls.
The new components comply with IEE 323-74 (more stringent than 323-71) -and IEEE 344-75.
The Analyzer range, accuracy, and alarm actuation did not change as a result of the SMP.
There is no increase in probability and/or consequences of an accident previously evaluated in the FSAR.
No new failure modes were introduced to create an accident different than any already evaluated in the FSAR.
There is no change in the minimal impact that the Hydrogen Analyzer System will have on other equipment important to safety.
No unreviewed safety question resulted from this SMP.
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Station Modification Package, SMP-1004-l
~SMP-1004, Liquid Post Accident Sampling System (PASS)
Modifications s.
Description of Change SMP-1004 modified the existing Reactor Coolant and Safety Injection Sump (SIS) PASS.
The gas analysis section of-the system was changed from_ a " flow through" to a
" batching" system.
Reason for Change The modification enables the operator to have better control of the system and to more easily quantify
' dissolved gas concentration in the reactor coolant'or the SIS sump.
Safety Evaluation The modified system provides sampling capability and-radiological and chemical analysis to meet the original design. basis described in FSAR Section 9.'3.8.
The installation ~ of the modified sampling' system maintained isolation <from the reactor coolant pressure boundary and is therefore considered a non-safety-related
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system.
No'new unmonitored release paths were introduced since discharge of liquid sample flow will be directed to either the Radwaste Tanks or the Containment Sump-and-gaseous samples / residues will be routed either to the Gas Surge Tank System or returned to the Containment atmosphere.
The SMP implementation meets the requirements of Technical Specification 6.8.4 (d).
No changes are required to the Technical Specifications.
Margin to safety is not reduced.
Failure of the PASS would not affect safety-related equipment or the capability to safely shutdown the Plant.
In addition, the system failure will not increase the probability or consequences of an accident.
No unresolved safety question resulted from the SMP.
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j 11... Station Modification Packaoc, SMP-1009-SMP-1009, Electro-Hydraulic (EH) High Pressure-
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Differential Pressure Switch Impulsing t
Description of Change s
SMP-1009 installed 1/4-inch needle valves EH 1031 and 1041 (A and B)- between the isolation valves EH 103 and 104 ( A and B) and the respective Electro-Hydraulic Fluid High Differential Pressure Switches DPS-TA-4301'(A and y
B). - Adjusted needle valves for proper operation.
Reason for Change The-modification was performed to prevent switch housing failures'and falso alarms due to high pressure impulses.
A switch housing failure could result in the loss of a large quantity of EH Fluid.
The potential problem was brought to LP&L's attention by Westinghouse Operation &
Maintenance Memorandum #009.
The initial recommended modification involved installation of four snubbers.
However, the snubbers did not eliminate high differential-pressure alarms when the by-pass valves closed on low accumulator pressure. Westinghouse then recommended the installation of the needle valves.
Safety Evaluation The modifications performed are in the non-safety-related Turbine Electro-Hydraulic Fluid System.
The system is not required for safe shutdown of the Plant and does not affect any safety-related equipment.
The modification will not have any impact on the accidents previously evaluated in the FSAR nor create the-possibility of new unevaluated accidents.
The implementation of this SMP does not require a change to the Plant Technical Specifications.
No reduction in l
margin of safety in the basis of Technical Specifications l
resulted.
No unreviewed safety question is-involved with the modification.
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Station Modification Package, SMP-1087 SMP-1087,. Replace Various Tank Level Indicators / Switches Description of Change n,
SMP-1087 replaced the Shand & Jurs type level indicating switch with level transmitters for various large storage tanks..
The Condensate and Primary Water Storage Tanks level measurement was modified by wiring signal.
conditioners and electrical indicators in series. with-the existing level transmitters (LT-CD9002 and-i
'LT-CD9041).
The signal conditioners were-_then used-to generate the-same control functions as.the Shand & Jurs level indicating switches. For.the remaining seven -tanks U
modified, new level transmitters, electrical indicators and signal conditioners were used.
Now, instrument connections wero-provided near the bottom of each tank-for the transmitter.
Reason for Change The original Shand &.Jurs level instrumentation on the various large. storage tanks were unreliable and required a great-deal of maintenance.
Safety Evaluation a
Of the nine' tanks modified, the Emergency Diesel Oil i
Storage Tanks are considered safety-related/ quality-related, and the Fire Water Tanks are considered quality-related/special. The remaining tanks were non-safety and non-quality related, and their modification had no effect on safe Plant shutdown.
Seismic considerations were 1
incorporated for the safety-related modification.
The probability and consequences of an accident previously evaluated in the FSAR will not be increased.
Loss of one Emergency Diesel Fuel Oil Storage Tank would not adversely affect the safe shutdown of the Plant since a redundant tank would bc available.
1 Potential accidents via this SMP are enveloped by existing FSAR analyses.
Failure of the modification would affect the same equipment that a failure of the original equipment would affect, with no change in the consequences. The function remains the same.
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' No change'-to Technical Specifications-was required'due toi the.
-modification.
Equivalent--
or-better-instrumentation now installed will not reduceLthe margin of safety-' defined.in the basis to any' Technical specification.'.
No unreviewed:: safety question resulted from -the-
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- 13. ' Station Modification-Package, SMP-1092 SMP-1092, Gland Steam Pressure Taps Description of Change SMP-1092 permanently installed a previous-temporary installation of vacuum sensing tubing _ and valves for Gland vacuum on High Pressure and Low Pressure Turbine Rotor Glands.
-Reason for Change The above mentioned tubing and valves, though shown on drawings and on the instrument list as test connections, had never been made a permanent installation with proper supports, workmanlike bends, etc.
Since outer Gland
. vacuum must be monitored during each startup, the subject modification made the tubing and valves suitable for repetitive long-term use.
Safety Evaluation The equipment involved in the modification is non-safety-related~ Balance of Plant (BOP) equipment and is not required for safe shutdown of the Plant.
Failure of _ the installed equipment would not affect equipment important to safety - nor would it adversely affect accidents evaluated in the FSAR.
No Technical Specifications or bases thereof were affected, thus there was no reduction to the margin of safety.
No unreviewed safety question was created as a result of this modification.
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I 14. - Station Modification Package, SMP-1135 SMP-ll35, Controlled Bulk Chemical Storage Area Building Description of Change SMP-1135 constructed 34' x 13' x 12' high steel-framed metal building, controlled storage enclosure for storing diesel generator fuel oil additives, corrosion inhibitors and other bulk chemicals. The f acility is located in the Plant Protected Area, adjacent to the Water Treatment
~Bui'lding southwest corner.
Reason for Change
.The items mentioned above were stored in an open area in the Water Treatment Building that was readily accessible to all personnel, which invalidated quality and safety controls.
Also a-centralized-storage area was required-for bulk barrel chemicals such as corrosion-inhibitors hydrazine, ammonia and biocides.
Safety Evaluation Construction of the above described facility limits unrestricted access to the stored products and helps improve quality and safety controls.
The construction did not require a
change to the Plant -Technical Specifications and had no impact'on the Nuclear Safety aspects of the Plant.
No.unreviewed safety question was generated by the construction.
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- 15.. Station Modification Package, SMP-ll86 SMP-ll86, Loop Seal on Exhaust Header Inlet to Vapor Extractors Description of Change
.SMP-1186 installed an oil mist separator between the main generator-loop seal tank and vapor extraction line 7VT4-37.
The separator was attached to the vent connection atop the loop seal tank.
Air, including the
. entrained oil mist, is now drawn from the loop seal-tank,-
passed through the separator and then enters the vapor extraction line 7VT4-37.
Reason for Change
-Plant Operations had noticed that oil had been collecting in. Vapor Extraction Line 7VT4-37, downstream of the Loop Seal Tank.
This presented a problem since the oil could-impede the air flow in the vapor extraction-lines.
The oil had been coalescing from a-fine oil mist which was mixed with the air drawn from the loop seal tank.
Safety Evaluation The equipment involved in the modification is non-safety-related Balance of Plant (BOP) equipment and is not required for safe-shutdown of the Plant.
Failure of the installed equipment would not affect equipment important to safety nor would it adversely affect accidents evaluated in the FSAR.
No. Technical Specifications or bases thereof were affected, thus there was no reduction to the margin of safety.
No unreviewed safety question was created as a result of this modification.
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Station Modification Package, SMP-1219 SMP-1219, Condenser Outlet Water Box Level _ Taps
. Description of Change SMP-1219 installed isolation valves on each of the six condenser outlet water boxes ( A1, A2, B1, B2, C1, and C2) at.the vendor's connection #74 (upper).
I Reason for Change The tap was needed to obtain an accurate water box level indication. The modification provided for reference legs to be vented to the upper section of the water.' boxes.
Safety Evaluation The equipment involved in~ the modification is non-safety-related Balance of Plant - (BOP) equipment and is not required for safe shutdown of the Plant.
Failure of ~ the installed equipment would not affect l
equipment important to safety nor would it adversely.
I affect accidents evaluated in the FSAR.
No Technical Specifications or bases thereof were-1 affected, thus there was no reduction to the margin of safety.
No unreviewed safety question was created as a result of this modification.
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Station Modification Package, SMP-1251 SMP-1251, Installation of Low Pressure Turbine LP1 and LP2 Heavy Disc / Key Plate Rotors p
Description of Change SMP-1251 replaced two Low Pressure (LP) Turbine Rotors, LPl and LP2, with rotors of heavy. disc / key design.
c Reason for Change Industry experience had indicated that most nuclear low pressure turbines with the keyed shrunk on disc rotor design, similar to Waterford 3 original design, were not usable af ter five to ten years-of service.
This was determined to be due to stress corrosion cracking in the keyway clearances, blade roots and bore areas of the disc, induced by a combination of factors including chemical buildup and corrosion.
Safety Evaluation The equipment involved in the modification is non-safety-related Balance of. Plant (BOP) equipment and is not required for safe shutdown of the Plant.
Failure. of the installed equipment would' not affect 3
equipment. important 'to safety nor would - it. adversely affect accidents evaluated in the FSAR.
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No Technical Specifications or bases thereof were affected, thus there was.no reduction to the margin of j
safety.
No unreviewed safety question was created as a result of this modification.
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- 18. -Station Modification Package, SMP-1259 SMP-1259, Modify Overflow Pipe on Bulk Chemical Storage Tanks Description of Change SMP-1259 rerouted the vent / overflow lines from the tops of the Sulfuric Acid Storage Tank and Caustic' Storage Tank to discharge just above the shell grade.
It also relocated the desiccant breather'on the Acid Tank to a-more accessible location and replaced the existing breather with a model that can be monitored for desiccant level.
Warning signs were installed on both tanks.
Reason for Change The modification was performed to reduce the possibility of personnel injury and equipment-damage.
Safety Evaluation The equipment involved in the modification is non-safety-related. Balance of Plant (BOP) equipment and is not required for safe shutdown of the Plant.
Failure of the installed equipment would not affect equipment important to safety nor would it adversely-affect accidents evaluated in the FSAR.
No Technical Specifications or bases thereof were affected, thus there was no reduction to the margin of safety.
No unroviewed safety question was created as a result of this modification.
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Station Modification Package,-SMP-1299
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SMP-1299, Gas Surge Tank (GST) Safety Valve Set Pressure l
Description of Change SMP-1299 increased the set pressure of the GST Safety
' Valve Eto 55 psig by installing a replacement 41 to 55 t
psig spring in lieu of the existing 30 to 40 psig spring.
At a pressure of 55 psig, the valve will require 58 psig inlet pressure to lift against psig back-pressure.
Also the valve specified design pressure was increased from 50 to 60 psig.
Reason for Change Design pressure for the Gas Surgo Tank had been increased from 40 to 60 psig._ However, the original relief valve set pressure of 35 psig had remained unchanged.
Safety Evaluation The increase in set pressure from 35 to 55 psig for the r
GST safety valve remained 5 psig within the 60 psig design pressure of the Tank.
The margin was.the same relative to the previous 40 psig design pressure'and 35 psig safety valve set pressure.
Therefore, there.was no increase in the probability or consequences.of any accident previously evaluated in the FSAR.
Nor was the possibility of a different accident-created.
The modification did not change the effects of a
malfunction of the modified equipment and other equipment important to safety.
The modification did not require a change to a Technical Specification.
There was no affect on margin to safety as defined in the basis of any Technical Specification.
No unreviewed safety question resulted from the modification.
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- 20. ' Station Modification Package, SMP-1304 s
SMP-1304, Seal 011 Coolers Temperature control Valves Description of Change SMP-1304 installed two temperature control valves (one for the H2 - Scal: 011 Cool'ing and the other for : the ' air sidei seal oil' cooling),
- piping, iso 1ation
- valves, supports, etc.
Also, Instrument Air piping was routed to the Scal oil skid for automatic valve control.
Reason'for Change
-i The Turbine ' Seal oil skid is cooled by the Turbine-Building--Cooling Water System.
There.was no automatic temperature control valves on. this cooling system to maintain the correct range of seal oil temperature during non-steady state conditions (start-up).
Safety Evaluation a
The equipment involved in the modification is non-safety-related Balance of Plant (BOP) equipment and is not required for safe shutdown of the Plant.
Failure of the installed equipment would not affect equipment-important to. safety nor would it adversely affect accidents evaluated in the FSAR.
No Technical Specifications or bases thereof. were affected, thus there was no reduction to the margin of safety.
No unreviewed safety question was created as a-result of this modification.
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Station Modification Package, SMP-1310 SMP-1310, Addition of Lube Oil'Centrifuae System to Feedwater Pump / Turbine Description of Change SMP-1310 installed a permanent Lube Oil-Centrifuge for the Steam _ Generator Feedwater Pump Turbine Lube _ Oil System, replacing the temporary centrifuge system that was--in use.
The installation included all the necessary
- piping, pumps, valves, -heaters,
- controls, and power supplies.
Reason for Change-To provide a permanent system to maintain the quality oil of-the Steam Generator Feedwater Pump Turbine Lube Oil t
System.
Safety Evaluation The equipment. involved in the modification is non-safety-related Balance of Plant (BOP) equipment and is not required for safe shutdown of the Plant.
Failure of the installed equipment would not affect equipment important to safety nor would-it adversely affect accidents evaluated in the FSAR.
No Technical Specifications or bases thereof were affected, thus there was no reduction to the margin of safety.
No unreviewed safety question was created as a result of this modification.
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- 22. - Station Modification Package, SMP-1327 7
SMP-1327, Block Caustic and Acid Supply Lines to-
, Blowdown Mixed Bed Demineralizers Description of' Change SMP-1327 updated drawings to reflect the two blind flanges added and spool pieces removed from dilute caustic 'and. acid lines to the Blowdown Mixed Bed Domineralizers which now utilize non-regenerable resin over near. term and long term for cost effectiveness. The work l had been done prior to the modification package under an immediate work authorization (LCIWA-016701).
The SMP installed two-blind flanges in the concentrated r
acid supply lines to prevent leakage into the dilute acid system..
Reason for Change The Caustic Storage Tank at -4' elevation of the Reactor e
Auxiliary Building (RAB) is no longer needed with the use of non-regenerable resin in the Blowdown Mixed Bed
'Demineralizers.
Since the tank is subject to overflow via leaky valves when its supply lines are pressurized, the supply line was' cut and capped and heat tracing was disabled.'
. Sulfuric acid is needed in; the RAB for waste neutralization purposes.
Two blind flanges were installed in -the concentrated acid supply lines to prevent leakage into the dilute acid system.
Supply of concentrated acid to the Waste Neutralization System will not be affected since dilution water is no longer needed.
.The supply line was blinded.
Safety Evaluation The equipment involved in the modification is non-safety-related equipment that is not required for safe shutdown of the Plant.
Failure of the installed equipment would not affect i
equipment important to safety nor would it adversely affect accidents evaluated in the FSAR.
4 No Technical Specifications or bases thereof were affected, thus no reduction to the margin of safety.
No unreviewed safety question was created as a result of this modification.
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- 23.. Station Modification Package, SMP-1328 SMP-1328, Gaseous Waste Management Rad Monitor l
Replacement
-Description of Change' SMP-1328 replaced the existing. Gaseous Waste Management (GWM) Radiation Monitor PRM-IRE-0648, an on-line monitor, With 'a. General Atomic off-line, extended range gas monitor.
A moisture control unit was also installed.
Process piping changes were made to affected GWM system-piping and valves to increase GWM discharge rate to design flow rate of-50 SCFM.
Reason for Change The original monitor enhibited a high fallure rate and required an inordinat', amount of maintenance to stay in service.
Safety Evaluatior, The non-safety-related equipment was replaced with equipment whicn met or exceeded original requirements.
No process change was involved.
Therefore, no increase-resulted in. the probability or consequences.of an.
accident previously evaluated in the FSAR.
Furthermore, the modification did not create the possibility for accidents different than that evaluated in the FSAR.
The modification did not change the effects of a
malfunction of the modified equipment and other equipment
-important to safety.
i The modification did not require.a change to a Technical Specification.
There was no affect on margin to safety as defined in the basis of any Technical Specification.
No unreviewed safety question resulted from the modification.
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f-24; Station Modification Package, SMP-1389 SMP-1389, Permanent Installation of the Second H2 Gas Dryer Description of Change SNP-1389 permanently installed the second hydrogen: gas-dryer for the Turbine Generator Hydrogen Gas System. The dryer had been-temporarily tied ;in with the - existing 0
permanently installed hydrogen dryer.
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Reason for Change 1
The modification was performed to make the required temporary installation permanent.
Safety Evaluation The equipment involved in the modification is non-safety-related Balance of Plant (BOP) ' equipment and is not.
required for safe' shutdown of the Plant.
t Failure of ' the installed equipment would not affect equipment important to safety.. nor would it adversely affect accidents evaluated in the FSAR.
No Technical Specifications or bases thereof were affected, thus there was no reduction to the margin of safety.
No unreviewed safety question was created as a result of this modification.
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Station Modification Package, SMP-1402 SMP-1402, Boze Filter Flushwater Flowmeter Description of__ Chance j
SMP-1402 replaced and relocated the Boze Water System i
flushwater flow indicator / totalizer.
This device, FIQ-CW-2500 measures and totalizes flushwater disebarge flow from the Boze System to the circulating Wster System.-
The replacement instrument is an electronic ultrasonic device as opposed to the original mechanical unit.
Reason for Chance The original flow indicator / totalizer had proven to be unreliable due to unsuitable application and incorrect i
rounting arrangement.
As a result, the Plant could not ensure accurate flow totalization of effluents from the t
primary water treatment plant.
l Safety Evaluation The equipment involved in the modification is non-safety-related and is not required for safe shutdown of the Plant.
Failure of the installed equipment would not affect equipment important to safety nor could it adversely affect accidents evaluated in the FS?.R.
i No Technical Specifications or - bases thereof were affected, thus there was no reduction to the margin of safety.
No unreviewed safety question was created as a result of this modification.
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Station Modification Package, SMP-1472 SMP-1472, CP-2 Startup Recorder Description of Chance SMP-1472 installed a multi-point, programmable recorder in Control Room Panel CP-2 (Roactor Control Panol).
The recorder is a variable speed and range recorder now occupying the location previously filled by the removed and spared Megawatt Demand Setter (MDS).
Reason for Change This modification was performed to address iuman engineering design considerations. The recorder is being used to monitor critical reactor parameters during startup as well as power operations.
Safety Evaluation The modification of CP-2 did not violate Class IE separation and seismic qualification criteria.
No systems used for accident mitigation were impacted by this SMP.
The modification enhances Plant operation, especially during primary system startups. No associated.
increases in probability or consequences-of accidents previously evaluated in the FSAR and no different accident possibilities resulted from the modification.
Since scismic and electrical separation considerations are intact, failure of the recorder should not impact other equipment important to safety.
No Technical Specification changes were necessary as a result of the modification.
No impact on the basis of Technical Specifications or associated margins of safety.
No unreviewed safety question resulted from the modification.
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Station Modification Package, SMP-1475 SMP-1475, Instrument Air Containment Isolation Check Valve Repair Description of Chanac SMP-1475 replaced Instrument Air Inside Containment Isolation Check Valvo 2IAV-60.!A/B (IAMVAAA910) with an identical Volan check valve.
The valve number assigned to the valve on drawings and other documentation was t
changed to 2IA-V614A/B.
Reason for Chanac The valve was replaced with an identical Velan check valve as initiated by a maintenance repair.
EBASCO Services recommended assigning a now tag number to the valvo.
It had been their experience that reassigning existing tag numbers to replaccmont valves loads to confusing documentation-files
_. o r quality-related Components.
Safety Evaluation The valve was replaced with an identical replacement valvo.
Therefore, no change to nuclear safety aspects of the Plant resulted. A safety evaluation was performed since the chango in valvo identification tag number resulted in a change to a figure in the FSAR.
The change had no impact on accident analysis equipment, or margin to safety.
No unroviewed safety question resulted from this modification.
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Station Modification Package, SMP-1480 SMP-1480, Security camera System Upgrade i
Description of Change j
SMP-1480 provided upgrades to the Security camera System.
i The modification package contained Security Safeguard information.
The details of the modification are available under Safeguard Information controls.
j Reason for Change f
r The modifications were
- t. formed in an effort to assure compliance with FSAR requirements / commitments.
Safety Evaluation The equipment involved in the modification is non-safety-related and is not required for safe shutdown of the Plant.
Failure of the _ installed equipment would not affect i
equipment important to safety nor would it adversely l
affect accidents evaluated in the FSAR.
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i No Technical Specifications or bases thereof were affected, thus there was no reduction to the margin of safety.
r No unreviewed safety question was created as a result of.
i.h.ts modification.
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Station Modification Package, SMP-1494 i
SMP-1494, Annunciator G0710 Alarm Modification l
Description of Change The local control panel annunciator was replaced by a new annunciator provided by the original vendor.
This allowed the continued use of the 42 alarm points at the local transmitter / recorder panel and reduced the number of selected alarms to 16.
Reason for Change since the transmitter / recorder is in the main accumulation conter, annunciation in the main Control Room is generic and tells the operator to check the local panol to determine whether a major alarm actually exists.
The Control Room operators were distracted by constant alarming of window G0710 located in CP-19 and they requested that the number of system alarms be reduced.
Operations requested that alarms of 39 points remain in the local panel and 16 specific alarms be generated in the main control Room common annunciator window G0710.
Safety Evaluation This modification involved the replacement of the local control panel annunciator due to chassis circuitry changes for the Primary / Secondary Sampling System.
The number of Primary /Socondary Sampling System alarm inputs to the common annunciator in the main Control Room was reduced from 39 to 16 chemical and/or radiological paramotors. The remaining 23 alarm points no longer have an audible alarm in the control room but are continuously recorded and monitored in the plant computer.
In the event of abnormal chemistry for those sampling points, the computer will flash at the CRT in the Control Room to indicate a problem.
The Primary /Socondary Sampling System local panel previously indicated records and annunciates the chemical / radiological paramotors for the secondary sampling points for systems including steam generator 1
& 2 blowdown, steam generator effluent, combined heater outlet, condencate pump discharge, main steam 1 &
2, makeup domineralizer ef fluent, condenser hotwells lA, 2A, l
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1 IB, 2B, 3C and 2C and troubic alarms for the secondary sampling system itself (i.e., chiller, annunciator and chemical waste tank). This modification provided audible annunciation in the main control Room of those critical i
1 parameters of each system which indicate a potential environment for the formation of corrosion products and subsequent steam generator tube degradation.
I Additionally, an audible and visual alarm for the steam generator blowdown radiation monitor will continue to be provided in the Control Room to monitor reactor coolant leakage to the secondary side as well as to indicate instrument failure.
i-There were no physical changes to the Transmitter /
Recorder Panel (Tag No. 4) whero local annunciator alarm located.
All 39 points still generate local alarms and l
light their respective windows.
Annunciator circuits are considered non-safety-related and their modification did not compromise the ability to safely shutdown the Plant.
The implementation of this modification did not require change to the Plant Technical Specifications.
This change did not involve an unreviewed safety question.
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Station Modification Package, SMP-1565 SMP-1565, Steam Generator Feed Pump Turbine (SGFPT)
Isolation on High Hotwell Level Description of Chance i
Under the previous design, when condensato in the Hotwell rosc~ to a level where there was a danger of water-induction to the SGFPT through the drain linos, the SGFPT was tripped by a level switch in the SGFPT drain tank.
This design led to three unplanned Plant trips.
b This modification resolved the above problem by implementing the following:
Routing the SGFPT drain lines from Hotwell Section i
B1 and B2 to the Al Hotwell.
This took advantage of the lower condensato level in A1.
Adding air operated isolation valves in the SGFPT which will close in the event of a
rise of condensato in the Al Hotwell to a height of 60 inches.
Removing three lovel switches from cach SGFPT drain tank and installing a sight glass on each tank.
Rolocating two of the switches to the Al Hotwell to provide signals to the isolation valves.
Installing higher range sight glasses on cach section of each Hotwell.
Providing Control Room annunciation and computer indication of the isolation valvo position.
Reason for Chanac Sinco commercial operation, there have been several SGFPT trips during periods where a condensor waterbox was removed from service for c1 caning.
The occurrences have led to a loss of generating capacity and other undesired events.
The concern over those events prompted LP&L to deviso a solution to the SGFPT trip problems.
Safety Evaluation This modification provided the design for isolating the SGFPT drain tanks from the condensor Hotwell and removing 33
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. the SGFPT water level trip switches from the drain tanks.
The af fected. equipment was non-safety related, non-quality related and this modification had no effect on-the. plant.'s ability for safe shutdown or mitigation of
. an accident.. Furthermore, no Technical Specification was-
- affected, i.
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Station Modification Package, SMP-1628 SMP-1628, Valve Packing Enhancement Program I
L Description of Chanac This modification upgraded the existing valvo packing i
program.to minimize valve packing leakage.
l Reason for Change This modification minimizes expenditures of time, money, and man-rem associated with unpacking valves, packing
- valves, adjusting packing and verifying valve operability.
Safety Evaluation This modification eliminated leakoff lines on selected valves, revised valve packing instructions and added springloads on packing for valvos throughout the Plant.
The leakoff modifications did not involve changes to any i
scismic, ASME Codo or class requirements or other quality standards.
None of these modifications increased the probability or consequences of an accident beyond that previously analyzed in the FSAR.
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Station Modification Package, SMP-1654 1
SMP-1654, RCP Seal Replacement Description of Change i
This modification replaced the original RCP seal cartridge with a newly designed seal cartridge during Refuci 2.
The new seal design implemented by this modification was subsequently replaced with the original design due to an unforeseen complication that arose i
regarding the now seal after Refuel 2.
However, prior to this happening, the new cartridge consisted of the Byron Jackson Type N-9000 stages and a Type SU-9250 vapor 1
scal.
The new seal operated with a controlled bleedoff of 1.5 gpm instead of 1.0 gpm at which the original seal operated.
This resulted in an increase in the heat load to the Component Cooling Water System.
This change is within the design capacity of the system.
Waterford 3 has operated with controlled bleedoff flows in excess of 1.5 gpm previously with no adverso consequences. The RCP controlled biccdoff flow is indicated on the Plant Monitoring Computer.
Reason for Chanac The Reactor Coolant Pump seals had a history of poor reliability resulting in forced outages due to seal failures.
The original seals were sensitive to temperature and pressure changes resulting in failures.
This modification replaced the original seal design with a new design.
The modif! cation required no chtnges to existing
- piping, instrumentation, or other pump components.
The new seals complied wit.h the original equipment specifications.
This modification was considered a component replaccmont under Section XI of the ASME code.
Safety Evaluation Replaccmont of the Reactor Coolant Pump scals with the new design did not involve an unroviewed safety question.
The now scals are designed to the same or more stringent criteria than the original type SU seals.
The use of the now design did not increase the probability of occurrence of an accident or malfunction of equipment important to safety previously stated in the FSAR.
The new seals did not create the possibility for an accident or malfunction of a different type than any evaluated previously in the FSAR.
Also, the new design did not reduce the margin of safety as described in the bases for any Technical Specification.
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Station Modification Package, SMP-1696 SMP-1696, Boron Manaccment to Circulation Water Isolation i'
Description of Chance Boron Management Valve BM-556 was an administratively controlled valve used to control the combined Boron-Management and Liquid Radioactive Waste System discharges to the Circulating Water system and then to the Mississippi River.
The valve was located 25 feet above the -35 floor elevation and accessible only by climbing on conduit or cable trays.
This modification removed a temporary - cross-connect
- hose, and provided an administrativo1y controlled valve in a more accessible location. The temporary cross-connect hose was no longer needed due to changes in Waterford 3's EPA discharge permit.
The new valve was installed upstream of valve BM-554 and the inlet to Radiation Monitor RT-627 to provide doubic isolation of the Boron Management system.
from circulating water.
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Reason for Change This modification was instituted to relocate an administrative 1y controlled valve used to combine Boron Management and Liquid Radioactive Waste System discharges
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to a more accessible location.
The previous valve, BM-556, was located 25 feet above the -35 floor elevation and accessible only by climbing on conduit or cable trays.
Also, a temporary cross-connect hose was removed due to changes in Waterford 3's EPA discharge permit.
Safety Evaluation This modification relocated an administratively controlled valve used to combine Boron Management and Liquid Radioactive Wasto System discharges to a more accessible location. Implementation of this modification did not increase the probability of a malfunction of equipment important to safety for Plant safe shutdown or mitigation of an accident as previously evaluated in the FSAR.
Also, the margin of safety as defined in the basis to Technical Specifications has not been reduced.
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Station Modification Package, SMP-1743
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SMP-1743, Boron Management and chemical Volume Control prain LeakaQe Prevention Description of Change This modification installed secondary isolation devices e
to prevent unwanted leakage for selected drain lines leading to the equipment drain sump which are located at the -35 level of the RAB.
This modification reduced the amount of liquid radwaste generated and provides greater control of RCS inventory.
Reason for Change This modification was implemented to eliminate leakage of selected drain lines of various systems.
The leakage was allowing noble gases and other gaseous nuclides to escape into the atmosphere of the RAB.
Safety Evaluation The installation of this modification to install spectacle blinds and additional valves for various system drain lines which discharge into the equipment drain sunp has no safety significance on the Plant's safe shutdown or accident mitigation capability.
This change does not alter any previous accident or equipment malfunction analyses as described in the FSAR.
- Also, this modification does not change the Technical Specifications.
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Station Modification Package, SMP-1817 SMP-1817, Nuclear Measurement Corporation Radiation Monitor Replacement Description of Change This - modification replaced four existing Nuclear Measurement Corporation non-safety related radiation monitors in service at Waterford 3 with General Atomic-Sorrento Electronics Division monitors built for the specific application.
The following monitors were replaced:
1.
RE-BD-0100, Steam Generator Blowdown
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RE-CH-0202, CVCS Letdown 3.
RE-BM-0627, Boron Management Effluent Discharge 4.
RE-WM-0647, Liquid Waste Management Effluent Discharge This modification also included the installation of purge and cooling water lines and other actions to support i
monitor operation.
Reason for Change t
This modification replaced radiation monitors manuf actured by Nuclear Measurement Corporation (NMC) and furnished by Combustion Engineering (CE) because they were susceptible to operation and maintenance problems.
Additionally, spare parts from NMC were difficult to procure.
LP&L also replaced these units with ones manufactured by General Atomic to standardize the manufacture of the radiation monitoring equipment in the Plant.
Safety Evaluation l
This modification involved the replacement of four existing non-safety related radiation monitors with new monitors.
The new monitors meet or exceed the original requirements of the system.
The purpose of these radiation monitors is to continuously monitor the liquid in each system and to alarm if radioactivity levels exceed the predetermined setpoints.
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The system must be initiated locally at each monitor.
once initiated, the system draws a continuous sample from the line being monitored.
The liquid is then passed through the monitor and then pumped back into the line.
The monitors were inserted into the RM-80/RM-11 communication loop to allow process variables to bo monitored in the Control Room.
The implementation of this modification did not affect the Plant's safe shutdown or accident mitigation capability.
Also, no change to the plant Technical specification was required nor did this modification involve an unreviewed safety question.
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Station Modification Package, SMP-1835 SMP-1835, TAR 86-067: orbisphere Dissolved Hydrogen l
Analyzer on Sample Line P6 r
Description of Change I
This modification documents the alterations of TAR 86-067 which installed an on-line instrument on a primary i
sample line (Purification Filter Inlet) to take' grab samples.
The instrument was installed downstream of Sample Cooler PSL-MCLR-0001D and upstream of Pressure Reducing Valve PSL-1361.
Reason for Change TAR 86-067, which is documented by this modification, was completed to eliminate human exposure and reduce decontamination time while taking grab samples.
Safety Evaluation This modification, which documented TAR 86-067, did not increase the probability of an accident previously evaluated in the FSAR.
Furthermore, this modification did not affect the safe shutdown or accident mitigation of Waterford 3.
The margin of safety as defined in the basis to any Technical Specification was not reduced.
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Station Modification Packaoc, SMP-1979 l
SMP-1979, Fuel Pool Purification Pump Low Discharge Annunciator Description of Change This modification doloted the Fuel Pool Purification Pump Low Dischargo Pressure Alarm from. the Control Room Annunciator, operations requested that it be removed as part of the effort to alleviate alarms which were L
constantly illuminated in the Control Room to f acilitato L
rapid identification of incoming alarms. The alarm point was deleted _by determinating the cable from the field.
Reason for Change The Fucl Pool Purification Pump is not used continuously during normal plant operations.
This causes the low discharge pressure alarm point to be locked in.
This modification was part of the effort to alleviato alarms which were constantly illuminated in the Control Room.
This assists in rapid identification of incoming alarms.
Safety Evaluation This morlification deleted the Fuel Pool Purification Pump Low Discharge Pressuro Alarm point from the Control Room Annunciator. The alarm loop is a non-safety related loop and not required for the safety of the Plant or for any safe shutdown function. The purification pump is not run continuously during normal Plant operation.
It is only i
run if water has to be filtered and/or do-ionized prior to a refueling outage.
The performance of the pump is
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measured by the discharge pressure in conjunction with the suction pressure.
Operation personnel frequently monitor the system when it is in operation.
Failure of the purification pump will not impair operation of the Fucl Pool Cooling Pumps or Heat Exchangers.
Hence this modification did not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the FSAR, create the possibility for an accident or malfunction of a different type than evaluated by the FSAR or reduce the margin of safety as defined in the basis for any Technical Specification.
Also, this modification did not require a change to the Plant Technical Specifications.
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Station Modification Package, SMP-1994 SMP-1994, Turbine Building closed cooling Water Filtration System Description of Change The modification installed a permanent filter mounted in the Turbine Building which can be connected with temporary lines to drain valves on the suction and discharge sides of either Turbine cooling Water Pump.
The filter housing can accept filters from 0.5 to 90 micron particulate removal size.
Reason for Change The Turbine cooling Water System, on occasion, had excessive amounts of solids and iron particulates.
As a result, this modification was implemented as an option of filtering the water in lieu of the current " feed and bleed" method.
Safety Evaluation This modification was installed in a non-safety related system and consequently has no affect on previously evaluated accidents, equipment importance to safety or any basis to Plant Technical Specifications.
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Station-Modification Package, SMP-2041 SMP-2041, Ratio / Relay Panels for EDG Control Panel (A&B) Pressure Description of Change This modification provided the details for installing a ratio relay panel on each Emergency Diesel Generator.
Each panel consisted of a seismically designed _ instrument L
stand supporting 4 ratio relays.
Isolation and test connections were provided on the ratio relay panel facilitating maintenance and testing of the relays and corresponding pressure indicators.
The ratio relay panels were mounted as closo as possible to the engine connection block to minimize possible leak points, and were located within the curbing of the engine skid so that any spillage or leakage of process fluid would not present a personnel or fire hazard.
The EDG engine processes affected by this modification include the fuel oil, turbo fuel oil, engine lubo oil, and jacket cooling water.
Impulse tubing to the ratio relay panel was tied into the existing process tubing near the engine connection block. Each process was tubed through isolation / calibration manifold valves to the inlet of the ratio relays.
A supply line was routed to the panel supplying instrument air to the ratio relays.
Instrument air also had block valves to isolate each relay for maintenance without having to remove the others from service.
The ratio relays provide a pneumatic pressure signal to the EDG process pressure being input to that relay.
The pneumatic lines were routed through isolation / calibration manifold valves before leaving the ratio relay panel and returning the existing tubing to the EDG control panol. This arrangement allows servicing and calibration of any one relay without affecting the other three EDG processes.
Reason for Change Maintenanco and Plant Engineering personnel, noticing discrepancies between main Control Room and local indications of certain EDG engine processes, found that these processes were directly tubed to the EDG Control Panel.
This situation posed a potential hazard due to the flammable nature of the process fluids and the exposed contacts within the control panel.
As a result, l
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i PEIR 60746 was. initiated to investigate the situation and present options to rectify the problem. Af ter consulting L
with the manuf acturer of the diesel generators it was decided that ratio relays, which were missing from the t,
instrument tubing between the engine skid and control panel, be installed as closo as possible to the engine connection block. These relays would provide a pneumatic pressure signal equal to the pressure of the process '
fluid.
The pneumatic signal could then be used. to indicato process pressures and eliminate the hazards associated with flammable fluids.
e Safety Evaluation 1
This modification involved changes to the instrument tubing between the EDG engine skid and the EDG control panel.
The tubing was modified to include ratio relays, which were originally intended to be part of the EDG
- system, but were never actually installed during-construction.
In addition to furnishing the missing
- relays, this package provided isolation and test connections at the ratio relay panel to enhance maintenance and testing of the relay components.
The ratio relay panel was located just off the engino connection block.
The panel roccives safety-related impulse tubing from the engine processes and provides a one-to-one (1:1) non-safety pneumatic signal of process pressure.
This pneumatic signal is used to indicate j
process pressure in the EDG control panel.
Each relay panel has four (4) ratio relays with block and bleed manifold valves on either side of each relay.
The previous arrangement had process impulse tubing routed directly to the EDG control pancl.
Engine lubo oil, turbo lube oil, fuel oil, and jacket cooling water are the four (4) EDG cngine processes affected by this modification - all are safety-related.
Tubing on the process side of each relay is safety-related and was procured to Quality Class 1 requirements.
New components on the pneumatic side of the relay are non-safety; however, because of seismic II/I situation, and because the existing impulse tubing to the EDG control panel was installed as Safety Class 3 and Seismic Class I, pneumatic tubing was seismically qualified and seismically supported. Fittings and valves were purchased as Quality Class 3 items from an LP&L approved supplier on the Qualified Suppliers List and l
45
were dedicated by LP&L for use in the Safety Class 3 system.
Structural steel for the instrument stands was procured as Quality Class 1 in order to ensure seismic integrity and supports modified or installed by this SMP were seismically qualified.
Section XI of the ASME Boiler and Pressure Vessel Code, paragraph IWA-7400 states, "The following items are exempt from the requirements of this article (replacements):... (d) piping, valves and fittings 1 inch nominal pipe size and less, except that materials and primary stress levels shall be consistent with the requirements of the applicable Construction Code." (Ref.
6).
Valves specified for installation in this L
modification were 1/2 inch instrument valves.
In this analysis, it was shown that stress levels developed as a result of seismic activity were well within yield stress limits of materials used (0.5% of SS yield stress).
Since process operating pressures (0-80 psig) are within the prossure ratings of ANSI Class 2500 valves 1
(0-4130 psig-@ 450 degree F; Ref. 9), it can be stated that stress levels are maintained within the requirements of ASME Class 3 components.
Per this argument and paragraph IWA-7400 allowing the exemption of valves 1 inch nominal pipe size and less, the use of valves did not pose an open safety concern and did not violate the pressure boundary integrity of the EDG engine processes.
Thus, Class 3 nuclear qualification of the EDG system was maintained.
The relays specified for installation in this modification were seismically qualified to demonstrate component integrity by an independent laboratory prior to being placed in service on the EDG system.
l The implementation of this SMP did not require a change to the plant Technical Specifications.
The foregoing constitutos that this modification did not involve an unroviewed safety question.
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40.
Station Modification Package, SMP-2066 SMP-2066, Setpoint Change to NG-IPIC-7630 Description of Change This modification changed the instrument setpoint of NG-IPIC-7630 from 0.5 psig to 2.0 psig.
This is the setpoint for nitrogen gas to.the reactor drain tank controller.
Reason for Change The controller for the low pressure alarm for the reactor drain tank was set at 0.5 psig.
This controller is needed to maintain pressure above the low pressure alarm.
The previous setpoint of 0.5 psig was too close to O psig for operations to verify positive pressure.
Operations did not have another method to supply nitrogen gas to the reactor drain tank other than by this controller.
This modification provided operations with a more accurate method to supply nitrogen gas to the reactor drain tank.
Safety Evaluation This modification changed the setpoint for the controller of the reactor drain tank to 2.0 psig to enable operations to improve positive prassure capabilities on the reactor drain tank. This modification did not affect the probability or consequences of an accident as previously evaluated.
Also, the margin of safety as defined in the basis to any Technical Specification was not affected or reduced, 47
I B.
DESIGN CHANGE PACKAGES (DCPs) 41.
Design Change Package, DCP-3000 DCP-3000, Waste Gas Compressor Replacement (Tie-ins)
Description of Change DCP-3000 incorporated Temporary Alteration Request (TAR)88-025.
The TAR installed four diaphragm scaled stainless steel Safety class 3 globe valves.
Two valves were installed per Wasto Gas Compressor, one on the suction and one on the discharge at the header Tee's.
Reason for Change The valves were installed to enable the isolation of the compressors from the systems for reconfiguration of the downstream piping for replacement of the Wasto Gas Compressors (DC-3026 Phaso II).
Thesc valves were required in order to keep the GWMs in service during the installation of each new compressor.
Safety Evaluation The installation of the two valves per Wasto-Gas Compressor does not increase the probability of an accident previously evaluated in the FSAR.
The FSAR addresses failure of a Gas Decay Tank (GDT).
The installed valves do not isolate the GDT.
Failure of a valve would not impact the evaluated event.
Therefore, the consequences of the accident evaluated in the FSAR was not increased.
(Radioactive Waste Gas System Leak
~
or Failure Event, FSAR Section 15.7.3.1)
Failure of a system valve is bounded by the limiting fault (Failure of a GDT).
Therefore, the possibility of an accident which is different than any already evaluated in the FSAR was not created.
The valves will be left open until required to replace the Waste Gas Compressors.
Therefore, the probability of malfunction of equipment important to safety previously evaluated in the FSAR will not be increased.
The loss of a compressor due to a stuck closed valve would neither increase or mitigate the consequences of a gaseous release.
Consequences of a malfunction of equipment important to safety previously evaluated in the FSAR was not increased.
48
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. Similar; valves used in the GWMs.
The possibility of a
^
malfunction of equipment important to safety different h,....
than any already evaluated in the FSAR was not created.
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These valves do not-affect the. GDTs (addressed by.
1 Technical Specifications).
Therefore the - margin. to i
safety as defined in the basis 'of any Technical
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Specification will not be reduced.
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Design Change Package, DCP-3003 DCP-3003, Fuel Pool Hx CCW Flow Low Alarm Removal Description of Change DC-3003 (CI-253518) spared annunciator window N-0610,
" Fuel Pool Hx CCW Flow Low Alarm", by determinating cable 30712C at both ends and replacing the window with a blank lens.
Reason for Change I
.The modification was performed to remove a nuisance annunciator alarm point.
The alarm was set at 500 gpm While actual flow under temperature control valve control l
is approximately 100 gpm.
The window was continually activated and had become a
nuisance to operations Personnel.
Safety Evaluation 1
This design change removes an alarm which is not discussed in, nor is credit taken for, any accident analysis.
Therefore, the probability of an accident previously evaluated in the FSAR will not be increased.
Removal of the alarm does not increase the consequences of any accident pre'tiously evaluated in the FSAR.
CCW flow is isolated to the Fuel Pool Heat Exchanger upon receipt of an SIAS signal.
Removal of the alarm does not create the possibility of an' accident which is different than any already evaluated in the FSAR..
The alarm is not required during an accident since CCW flow is isolated to the Fuel Pool Heat Exchanger during an accident.
The Plant FSAR does not take credit for the subject alarm in an accident.
The probability or consequences of malfunction of equipment important to safety previously evaluated in the FSAR will not be increased.
It is known that flow will be low to the Fuel Pool Heat Exchanger during an accident since flow is isolated upon receipt of a SIAS.
The alarm removal will not create the possibility of a malfunction of equipment important to safety different than any already evaluated in the FSAR.
It is not required during an accident.
50
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The FSAR ' does' not take credit for the subject alarm during an accident. Therefore, removal of the alarm does l
not reduce.the margin to safety defined in the basis of any Technical Specification.
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- Removal of. nuisance-alarmswhere allowable-frees
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- immediate ' or continued attention, therefore enhancing L
-safety aspects of the Plant.
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Desion Chanac Package, DCP-3018 DCP-3018, Charoing Pump Seal cooling System Modifications Description of Change r
The level sensing probes on the Charging Pump Seal Cooling System reservoir tank were replaced with float switches.
An isolation valve was added to the
[
domincralized water supply.
The existing isolation and solenoid valves were replaced with leak-tight valves.
L An indicating light was added to indicate when the makeup l
solenoid valve is energized.
j Reason for Chango The modification was performed to improve the operational reliability of the Charging Pump Seal Cooling System.
The unreliabic system had resulted in inadequate seal cooling and premature seal packing wear.
The isolation valvo added to the domineralized water supply was provided to allow work on the system without isolating domincralized water to all three Charging Pumps.
The modification f acilitates preventing contamination of the room due to tank overflowing.
It prevents unnecessary load on the waste management system by reducing or climinating tank overflow.
Safety Evaluation The reliability enhancements and the addition of an isolation valve on the non-safety water supply to the non-safety scal cooling system does not increase the probability or consequences of an accident previously evaluated in the FSAR.
Nor do they create the possibility of an accident different than any already cvaluated in the FSAR.
The changos do not adversely affect the operation of nor the consequence of malfunction of any equipment important to safety different than that already evaluated in the FSAR.
No impact is made upon the Technical Specification basis and l
therefore no impact on margin to safety.
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Design Chance Package, DCP-3021 DCP-3021, Turbine Generator Stator Coil Water " Tee" Installation For Conductivity and Oxygen Checks Description of Change DCP-3021 installed isolation valves and quick-disconnect sample points located near permanently installed I
conductivity cells, Reason for Change v
The modification was performed to provide stator coil l
Water samplo points, to allow Chemistry personnel to test 1
water conductivity and oxygen content. This provides for verification of the permanent conductivity cells.
Safety Evaluation The change does not impact nuclear safety aspects of the plant nor any accident probabilities or consequences evaluated in the FSAR.
The system affected (Main Generator Stator Cooling Water) performs no safety function nor does it support a system that performs a safety function.
It is not required to achieve safe shutdown or to mitigate the consequences of an accident.
No Technical Specification was affected, nor potentially affected, thus no margin to safety defined in the basis to any Technical Specification was reduced.
The 50.59 nuclear safety evaluation was performed due to the change to drawing LOU-1564-G-152, sheet 2 which affected FSAR Figure 9.3-1, shoot 2.
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Design Change Package, DCP-3056 DCP-3056, Installation of Pressure Bleedoff Valve For Air Accumulator Check Valve Test Description of Change DCP-3056 installed bleedoff needle valves immediately upstream of the check valves for valves SI-MVAAA-602A and SI-MVAAA-602B, Reason for Change The modification providos for testing the air accumulator check valves for safety-related valves SI-MVAAA-602A and SI-MVAAA-602B.
Each accumulator has an associated check valve designed to seat and sustain accumulator pressure.
e The accumulators are sized to provide enough air for one stroke open - one stroko closed in the event that the instrument air supply is lost.
Safety Evaluation Failure of the installed bloodoff needle valves would have no affect on the accumulator check valves performing their designed function.
The probability and consequences of an accident previousy evaluated in the FSAR would not be increased by the f ailure. The bleedoff needle valves do not perform a safety-related function.
No nuclear safety significant accident different than those evaluated in the FSAR would be created.
No increase to the probability or consequences of malfunction of equipment important to safety evaluated in the FSAR or different than that ovaluated in the FSAR l
were croated since failure of the bicedoff valves have no affect on the pressure retention capabilities of the safety-rclated accumulators.
No reduction to margin of safety resulted from the subject modification since there was.no affect on the basis for any Technical Specification.
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Design' Change Package, DCP-3060 DCP-3060, Relocation of Seismic Monitor a
Description of' Change The modification relocated Seismic Monitor SM-IYR-6021 from intermediate leg 2A cf the Reactor Coolant System piping to the. Safety Injection System piping.(Line No.
IRC14-44RL1) between Valve ISI-V1502B and the "D" Ring.
Reason for change
'. h e temperature, vibration and radiation IcVels experienced by the seismic monitor are greatly reduced jn the new location.
Scismic Monitor SM-IYR-6021 has repeatedly
- failed, suspected due to heat from the 4
associated piping and vibration due to the monitor's
-close proximity to RCP-2A pump.
Safety Evaluation There is no accident evaluation in Chapter 15 of-the FSAR for the seismic Monitoring System.
Seismic Monitor.
SM-IyR-6021 is a self-contained passive device requiring-no internal. or external power or control connections.
3ccause the monitor performs no active function, it cannot increase or decrease the probability. of an accident previously evaluated in Chapter 15 of the FSAR.
s The solo purpose of the seismic monitor per Regulatory Guide 1.12 is to record triaxial peak accelerations of Reactor' Coolant System piping during a seismic event.
.This monitor as well as the entire seismic monitoring system plays.no part in mitigating-the consequences of an accident, Because the relocation / installation of seismic moni n r
SM-IYR-6021 meets or exceeds seismic Category 1
requirements, the possibility of an accident different than any already evaluated in the FSAR will not be
- created, The monitor does not affect consequences or probability l.
of malfunction of equipment important to safety.
Both L.
the monitor and mounting support meet scismic Category f
1 requirements which keeps them in place during a seismic
?
event.
Due to whip restraints, the monitor will not l
become a missile in the event of a pipe break.
Since it is passive, its failure does not affect other equipment.
Design reviews and calculations verified that the monitor will not create a seismic II/I hazard er other type hazard to safety-related components.
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The relevant: margin of safety covers the seismic stresses
.for the pipe and the function'of components around the new monitor -location. The new location does not increase.
g.
.or. decrease the-integrity of IRC14-44RL1.
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Design Change Package, DCP-3062 DCP-3062, CVC-109 Replacement Description of' Change The modification replaced valve CVC-109, a. Safety Class 2 gate -valve with a Masoneilan valve.-
A similar replacement was done with valve CVC-103 in SMP-934.
The modification was performed - under ASME Section XI 1980 Edition, Winter 1981 Addendum.
Reason for Change During past outages, CVC-109 required extensive work to pass LLRT and operability tests.
Safety Evaluation The replacement valve meets the same requirements as the original valve.
Therefore, the probability and/or consequences.of an accident previously evaluated in the FSAR were not increased.
The valve replacement did not involve a change in valve i
function.
Therefore, the possibility of an accident which is different than any already evaluated in the FSAR will not be created.
Changing the valve type was evaluated as not increasing the probability or consequences-of malfunction of equipment important to safety previously evaluated in the FSAR since the new valve had to meet the same requirements and functionality.
The replacement valve does not impact any equipment different than the original valve.
Since the functional design was unchanged and there was no reduction in requirements imposed on the replacement equipment, the margin of safety as defined in the basis of Technical Specifications were not reduced.
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Design Change Package, DCP-30701 DCP-3070, Provided SUPS Feed for Emergency Notification System Description of Change t.
'The modification made the wiring and conduit revisions-
-o necessary to provide power from Power Distribution Panel PDP-3014AB (SUPS AB)~to the NRC and OPS Hotline Control Packages.
Reason for Change The modification was performed to increase the reliability of the ENS communication equipment.
It assures that loss of offsite power would not result'in loss of communications between the Waterford 3 facility
-and the NRC Operations Center via the Emergency Notification System.
The OPS Hotline portion of the modification was necessary to be in agreement with FSAR-Section 9.5.2.1.
' Safety Evaluation As stated in FSAR Section 8.3.1.1.1.c, the SUPS is i
arranged such that any type of single failure or fault 3
will not prevent-proper protective action of the safety-related systems.
Therefore, the probability _ and/or consequences;of'an. accident previously evaluated in the FSAR was not increased by this design change.
Neither the ENS nor the OPS Hotline are Safety Related
]
systems.
They are not-required for Safe Shutdown of the Plant.
The failure of the systems will not create the l
possibility of an accident different than any already 1
evaluated in the FSAR.
Since the systems involved (ENS and OPS Hotline) are not required for safe shutdown and do not support systems required for safe shutdown, the margin of safety. as defined in the basis of any Technical Specification is not affected.
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4 9.. Design Change Package, DCP-3094 DCP-3094, Fuel Handling' Building Shipladder Handrails, at EL +1.00 i
Description of Change The major work items associated with the modification included:
- 1) permanent removal of the existing access
.l hatch.at EL +1.00; 2) installation of handrails around
-the hatch opening; and = 3) extension of the shipladder y
handrails up to the new handrails around the opening.
t Reason for-Change The design change was implemented to improve personnel safety near the shipladder access hatch-located in the Fuel Handling Building at EL +1.00.
Safety Evaluation a
There is no accident evaluation in FSAR-Chapter 15 for the. Fuel Handling Building access hatch on EL
+1.00.
Therefore, removal of the hatch and the associated handrail work will not increase or decrease the-probability or consequences of an accident previously evaluated in the FSAR.
The modification is non-safety /non-nuclear in scope and only involves new and existing handrails and removal of i
an access hatch.
The hatch was not a fire door and its
. removal did-not adversely affect the capacity for the Fuel Handling Building HVAC to operate in the normal or accident mode.
Therefore, implementation did not create the possibility of an accident which is different than L
any already evaluated in the FSAR.
L DC-3094 did not involve installation or modification of
^
safety-related equipment, nor did it create a seismic II/I.
hazard.
Therefore, the probability and/or consequences of malfunction of equipment to safety l.
previously evaluated in the FSAR was not increased.
L Furthermore, it did not create the possibility of a E
malfunction of equipment important to safety different than any already evaluated in the FSAR.
Since the modification did not adversely affect the capacity for the Fuel Handling Building HVAC to operate in the normal or accident modo, the margin of safety as defined in the basis of Technical Specifications were not l-reduced.
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50.
Design-Change Package, DCP-3115 l
DCP-3115, MS Drain Valves (2")
Description-of Change The modification installed an: additional 4"
- long, 2"
Schedule 80 carbon steel pipe 'and 2"
carbon _ steel threaded' cap to the outlet of valve MS 1052B.
Reason for Change Valves MS 1052B, MS 1051B and the pipe 3n between had
-i been leak repaired.
DCP-3115 was performed to allow pressure testing the. rework / repair without _ having to
' hydro the MS Header.
The cap will also contain further leakage if both valves MS 1052B and MS-1051B were to f ail again.
Safety Evaluation seismic Loads were not af fected by the additional weight of the cap and pipe addition.
Tho' design intent and function of the valves were not affected by the change.
No additional pipe or valve stresses were induced.
Therefore, no increase was caused to the probability and/or consequences of an accident previously evaluated in the FSAR.
Since additional loads were. acceptable, the possibility of an. accident different than any already evaluated in the FSAR was not created.
Since seismic requirements are not violated; pressure boundary is stil acceptable; and the MSIV Drain Line is not affected, the probability and/or consequences-of a malfunction of equipment important to safety previously evaluated in the FSAR will not be increased.
Drain lines will function per design.
MSIVs are not affected by the modification.
The possibility of - a malfunction of equipment important to safety different than any already evaluated in the FSAR will not be created.
Since the function and operation of the MSIVs will not be affected, the margin of safety as defined in the basis to Technical Specification will not be reduced.
60
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Special-Issue-h[-
PE-005-001,2 Revision 1, Change 1, 2, & 3,-
Integrated Leak Rate Description of Change s
This procedure assures LP&L that - leakage through the
. primary reactor containment,. including systems and components penetrating ' primary containment, -does not exceed the allowable leakage rate..
The change _ (1) incorporated new sensor volume fractions and locations, t
(2) allowed the test to continue after isolating leaks, and (3) changed a valve line-up for hydrogen analyzers.
A statement discussing item (2) appears in FSAR 6.2.6.1.
Reason for Chance W3E88-0065 provided new sensor locations and volume fraction calculations for the integrated leak rate test (ILRT).
Various heat loads now-in the containment were not present during the pre-operational'ILRT. This change in test method countered.the likely ~ increased top to bottom temperature differential.
Calculations attached 1
to W3E88-0065 allowed LP&L to reassign sensor' positions and the volume fraction assigned to each sensor.
The temperature and humidity results data allow LP&L to accurately account for the partial' pressure caused ' by' water vapor.-
Safety Evaluatio_n_
PE-005-001 requires the plant to be in COLD SHUTDOWN with containment isolation valves closed and equipment inside containment vented to containment atmosphere.
It also
~
requires shutdown cooling to be in-service.
These test-features are as described in the FSAR (re Section 6.2.6.1).
However, the specific test steps and sensor locations are not described.
Therefore, the procedure change described above cannot affect either probabilities H
or consequences of either accidents or malfunctions described in the FSAR.
Further, staging and de-staging l
the test cannot cause a new accident because of the l
administrative controls provided in PE-005-001 (especially post-test valve line-ups).
Other administrative controls preclude violating Technical Specification
- LCOs, safety limits and surveillance requirements.
Thus, the change to PE-005-001 described above does not reduce any margin of safety.
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Special Issue STP-NE-TEM-003, Variation of Feedwater Suction Pressure N
Description of Test The test reduced ~the setting on the condensate polisher
^
2 master. controller, thus increasing the pressure -drop across the polisher discharge valves.
The increased pressure-drop lowered the force-against which heater drain-pumps act.
Reason for Test This test determined whether the pumping capacity of the Li!
heater drain pumps could be increased by reducing the suction-pressure to the main feedwater pump.
Safety Evaluation
,J:
A test prerequisite required the plant to have all.three heater drain pumps on-line and plant power to exceed 95 percent. Operating at least 95 percent assured dampening any small perturbations caused by the test.
The test perturbated non-safety-related equipment in
-the condensate system. Therefore, tra potential consequences of an accident did not change.
The accident analysis assumes non-safety-related equipment does not operate at all.
.Therefore, the FSAR-analyses bound reduced
'feedwater suction pressure.
The test administratively ensured that feedwater pumps' suction pressure remained above 350 psig during the test.
The feedwater pumps automatically adjust to (within design) perturbations and maintain steam generator levels;- so the assumed probability of either (1) a feedwater malfunction, or (2) feedwater accident, stayed intact during this test.
The. test did not affect a Technical Specification LCO, safety limit, or surveillance requirement.
Thus the margin of safety remained intact.
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Special Issue STP-WA-01027046, Lithium Tracer Test (Foodwater Flow Noztle Description of Test This test determined the condition of the feedwater flow nozzles by accurately measuring feedwater flow to each steam generator using a chemical tracer.
The-associated procedure provided (1) instructions. for installing sampling equipmont, (2) necessary plant conditions,-and (3) test steps.
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Safety Evaluation The test did not change the operation of the Plant, other Lthan requiring operators to place the feedwater control; in manual.-
Steam generator controls were not affected.
The concentration of tracer chemical was well below the corrosive threshold. This low concentration also ensured ao that'the feedwater system pH was not affected, The FSAR analysis for a feedwater line break bounds the possible consequences of a hypothetical break in the-sampling equipment used, j
The test did not affect a Technical Specification LCO, 1
safety limit, or surveillance requirement.
Thus the
.j margin of safety remained intact.
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- 54. ' Plant Procedure, UNT-5-013 (Revision 1)_
UNT-5-013 (Revision 1), Fire Protection Program-Description of Change i
Revision 1 to UNT-5-013,
provided the following changes:
Inclusion of Reference to Plant Technical Specification 3.0.3 and 3.0.4 Addition of a fire water tank Provision for fire pump testing tolerance and Reactor Containment Building (RCB) fire-hose consideration Reason for Change To establish better functional test criteria for fire pumps.
Safety Evaluation The level of fire protection was not decreased.
This is based on the fact that the revision was mostly editorial / administrative in nature. Certain enhancements were added at the request of the NRC.
A provision to allow measures for RCB hose stations being inoperable was added to provide better ALARA consideratons.
The flow test requirements for fire pumps was clarified to be -
consistent with ASME Section XI.
This procedure revision did not affect Plant safe shutdown or' accident mitigation capability.
Also the-margin of safety as defined in the basis for Technical Specifications was not reduced.
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- 55. Special Issue' DCN-HV-124 and Radioactive Release Path Categories Description of Change.
~
LP&L established a new offluent release path category.
Reason for Chango LP&L documentation, before the analysis of the now paths (re PEIR 71098), did not justify secondary ventilation systems-discharging directly to the atmosphere.
Safety Evaluation e
Waterford ventilation exhausts to the atmosphere fall into one out of three possible categories.
Each exhaust path is either a " principal cff3uent release pathway,"
or a "non-potential release pathway," or a " secondary" release pathway.
First, Waterford Technical Specifications explicitly list and provide administrative. controls over " principal offluent-release pathways."
Under accidents discussed in Chapter 15, these pathways make the largest contribution to site boundary dosage estimates..
Second, some Waterford ventilation systems serve areas with small or nonexistent source terms, i.e.,. a "non-potential rolcase pathway." Releases from these exhausts cannot occur.
This ventilation system group has no source term in the served room (s).
The ventilation systems arc either (1) separato from others, or (2) have multiple barriers separating them f rom any potential 7
source term.
Examples include HVAC for the Administration Building and the control room envelope.
Third, Waterford's design includes a ventilation system group that forms a " secondary" release pathway.
A secondary pathway is not normally radioactive; however, a
remote possibility exists that it can become radioactive during an unanticipated event.
Those pathways differ from the principal pathways listed on FSAR Table ll.3-9a.
Any potential radioactivo release for them results in insignificant (not discernible) dose consequences.
l 65
Secondary pathways have the following characteristics.
There is only a remote-possibility for attaining ten' percent of_the concentration limits in 10 CFR Part 20, Appendix <B, Table 2, Column 1 cither at or beyond the site boundary.
Technical Specifications define the site boundary. When considering all hypothetical " worst case"
-unanticipated' _ occurrences, the_
potential activity released corresponds to less than or equal to one percent of the annual air and organ dose limits from an equivalent single release pathway.
Lastly, all of-the estimated
" worst case" scenarios for all' secondary release pathways combined do not exceed ten percent.of the design basis annual air and organ dose limits.
The following lists the secondary pathways:
Emergency diesel generator ventilation system Battery. rooms exhaust Computer battery room exhaust Switchgear area smoke purge
)
Electrical penetration and cable vault and -relay room smoke purge Electrical penetraton area smoke purge Hot machine shop and decontamination shop
-ventilation system Radwaste Solidification Building (no ventilation system)
Radwa'ste Compactor Building air (no ventilation system)
While a containment equipment hatch is open After examining the potential sources in the area serviced by the ventilation systems for each secondary pathway, LP&L concludes that enough radioactivity could not gather to affect off-site receptors.
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- 56. Condition Identification 257780 5
J CI 257780, Add Drain Valves in Low Points of the Main Steam Headers to'the Emergency Feedwater
. Pump Turbine-Description of Change
.LP&L installed a short drain line, drain isolation' valve tail piece and cap to sockets welded'to-several (re LCIWA-24477) low points of the main steam headers to the emergency feedwater pump turbine.
Reason for Change
~
7 The drains allow personnel to remove condensed water from the steam header, thus preventing water hammer.
Safety Evaluation The new drains were installed under the same requirements used for the steam header, i.e, ASME Section III, Class 3.
A single drain assembly weighed approximately twelve pounds. This weight at each specified location was found
+
acceptable with respect to steam header seismic loading.
-t Thus, the drain assemblies are no more likely to break than any other part of the steam header.
The FSAR main steam line break analysis conservatively
. bounds any failure resulting from a break in these new drains.
The analysis assumes a much larger break area than the drain line cross-section.
These extra drains do not affect any-Technical Specification
- LCo, safety
- limits, or surveillance requirements.
Therefore, the Plant's margin of safety remained unchanged.
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'57. Condition Identification 260568 t
I CI 260568, WA 01029982, Restore Emergency Diesel Generator Fuel Oil Storage Tank Mechanical Float to Service Description of Change LP&L restored a " spared instrument" to service.
SM-1087 determinated. low. level alarm circuits from mechanical floats on the emergency diesel generator fuel oil storage tanks.
SM-1087 added remote leve; indication via delta-pressure transmitters.
Although SM-1087 directed abandoning the float, personnel found the float to be an integral part of the tanks.
Therefore, the float itself was only ' spared administratively.
This CI undid the administrative sparing action.
Reason for Change PEIR 60963 recommended against using the remote indication installed by SM-1087 for checking compliance with Technical Specification fuel oil storage-requirements.
The remote indication transmitters could not account for normal specific gravity variances of the stored fuel oil.
Safety Evaluation
-The floats give operators a ' mechanically coupled tank level indication. The mechanical coupling makes specific gravity variances immaterial.
With tank level versus volume curves, personnel can easily check for Technical Specification compliance regarding the. minimum amount of' fuel oil on hand (similar numbers allow personnel to calibrate the remote indication).
This change did not alter the as-built condition of the Plant.
Therefore, FSAR assumptions and the margin of safety were all unaffected.
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- 58. Condition Identification 261111 CI 261111, WA 01031151,. Replace the Local, Audible High-Iligh= Level Alarm on the Emergency Diesel Generator Jacket Water-Standpipe With Plant Monitor Computer Alarms Description of Change Jacket water cools operating-omergency diesel generator engines.
MP-1731 replaced local alarms for high-high '
jacket water standpipe level with Plant Monitor Computer (PMC) alarms.
Reason for Change This change eliminated a nuisance alarm from a misapplied instrument model.
The instrument connected to_the PMC could accommodate the normal standpipe level fivetuations during diesel operation without locking in the composite emergency diesel generator." trouble" annunciator in the Control Room.
Safety Evaluation The level alarm performs a passive measuring function.-
No automatic actions are-associated with the old or new alarm circuits.
Separate level switches control the amount of water in the standpipe.
Both the PMC alarm and the local alarm require plant operators to physically acknowledge them.
Thus, the old and new alarms are equivalent.
Because the alarm circuits have no automatic control or trip function, the FSAR assumptions concerning design l_
basis accidents remain unchanged.
Because LP&L retained l
the standpipe level alarm in the PMC, the margin of safety was also unaffected by this change.
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- 59. Document Revision Notice, M 8800750
'DRN.M 8800750, Delete the Locked-Open Requirement'from-Valves MS-403A and MS-403B Description of Change The :DRN withdrew the requirement imposed by SM-740 to
' keep _ main steam line valves MS-403A and MS-403B locked i
open.-
Those one-inch valves drain condensate from the emergency feedwater pump turbine steam supply headers.
Reason for Change
.m Physically locking open these valves was an administrative burden not justified by code or regulatory requirements.(re PEIR 60811).
Safety Evaluation
-3 a
Prudence dictated that these two valves should stay open to minimize the potential for water hammer.
Whether LP&L administratively keeps these valves _open, or physically locks them open does-not affect the ability of emergency feedwater to mitigate the consequences of an accident.
Other sections of this steam supply, header can accumulate water.
However, ' LP&L. manually drains '
those other sections.-
Therefore, the water hammer potential was unchanged by converting to administrative controls keeping MS-403A and MS-403B open.
This change upholds LP&L offorts to minimize water hammer events.
It does not affect the probability or consequences of an accident because the normal valve position remains open.
The revised valve line-up L-requirements do not affect _ the emergency feedwater Technical Specification; therefore, the margin of safety romains constant.
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9 60.. Document Revision Notice, M 8800918 DRN M 8800918, DCN-MP-796, CIWA 834332, Add a Drain to Sample Line Downstream From Blowdown Domineralizer Conductivity Cells Description of Change DCN-MP-796 administratively controlled adding a. pipe to route discharge from the steam generator blowdown domineralizer conductivity cells in the sampling system to.the sample recovery tank.
DCN-MP-796 was completed in 1983, before commercial operation.
DRN M-8800918 -
. clarified that this new pipe also included a ' drain
-assembly not previously shown on FSAR Figure 9.3-2, Sheet 4.
Reason for Change This design allows LP&L to recycle blowdown water rather than discharge it.
The drain facilitates system.
maintenance, j
i Safety Evaluation The new drain. assembly on the recycling line conforms to the same construction codes as the recycling line itself.
Therefore, its failure-probability is no different than l
the recycling line itself.
The drain valve is normally closed and the tail piece is normally capped.
i Because the new drain design conformed to the original licensed plant and it included double protection against
.j inadvertent diversion, no analyzed accidents. were affected by this change.
Further, the consequences of any accident caused by a single failure involving the new drain would be bounded by the accident consequences already analyzed in the FSAR.
FSAR 9.3.2.2.2 does not 1
refer to specific pipe design features such as the subject drain.
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- 61. Document Revision Notice, M 8801398 p
1 DRN M 8801398 (PEIR 10458),
Make Normal Position For Valve BAM-141 Consistent Description of Change Make drawing (re FSAR Figure 9.3-6, sheet 2) consistent with the requirements determined-while compiling a corresponding " component functional evaluation form" (re PEIR 10458).
1.
Reason for Change The drawing conflicts with the analysis below.
Safety Evaluation A basic function-for the chemical and volume control system is to provide control of the boron concentration in the RCS.
It is also used to inject concentrated boric acid - into the RCS upon a safety injection actuation signal.
A makeup system.is provided for changes -in reactor coolant boron concentration.
Four modes of operation by the makeup controller in the makeup subsystem are provided (dilute mode, borate mode, manual blend mode, automatic mode).
L Valve BAM-141 receives a
signal from the ' makeup controller.
This subsystem is used during normal-l operation.
However, during an accident, the SIAS causes the charging pump suction to be switched from the volume control tank to the boric acid pump discharge.
If the pumped boric acid supply were unavailable, the charging pumps are also lined up for gravity feed from.the boric acid makeup tank.
SIAS opens valve BAM-133 to allow boric acid to flow from the boric acid pumps-to the charging pump suction directly.
Valve CVC-510 receives a close signal upon SIAS..This prevents the mode which blends primary water with boric acid and sends it to the volume control tank.
In conclusion, valve BAM-141, which controls the boric acid flow in the blending system arrangement, is not required to be " active".
However, during an accident, the valve body and operator cannot fail such that boric acid is able to escape to the environment.
Therefore, it takes a " passive" role during accident mitigation.
During normal operation, the system is in the automatic mode which requires the BAM-141 to be in the open position.
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- 62. Document Revision Notice, M 8801624 DRN M 8801624, SMP-1591, TAR 86-56, show condensate Pumps' Gland Leakage Drains
' Description of Change LP&LLinstalled 3/4" drain lines on each condensate pump.
to send water to the floor drain in the condensate-pipe chase.
This was a new feature added to FSAR Figure 10.4-2, Sheet-1.-
Reason for Change The new drain line prevents water in holding pans
'.c connected to condensate pumps from spilling on the floor in - the pump cavity.
The drains were connected to existing nozzles on the drain pans.
Safety Evaluation r
The drain line performs a passive function on the non-safety-related condensate pumps.
Therefore, accidents analyzed in the FSAR cannot be affected.
Further, any single f ailure involving these drains cannot-introduce a new accident or malfunction because of their passive role.
No Technical Specification LCOs, safety limits, or surveillance requirenents needed adjustment to accommodate this design.
Thus the designed' margin of safety was unaffected by installing these.three drains.
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- 63. Document Revision Notice, M 88022,90 i i-DRN M 8801290 (PEIR 55036),
Update the Schematic Diagram for the Compon3nt Cooling Water Surge Tank Level Instruments Description of Change C
LP&L deleted the design feature allowing some safety 4
' grade level, instruments to.be isolated from non-safety level instruments attached to the component cooling water l
(CCW) surge tank.-
The surge tank passively maintains "c
minimum CCW pump suction pressure..
Safety Evaluation The as-Lailt. configuration climinated a low-level tap on each side of the surge tank.
Thus, all differential pressure cells (for each independent train) installed to q
measure surge tank level rely.on the same high head tap.
However, that does not affect'the dual train design of' F
the surge tank described in FSAR 9.2.2.2.1.
Operating pumps and the emergency diesel generators rely i
on an OPERABLE CCW system-during any postulated accident.
The revised design can detect all possible surge tank
.i failures.
No surge tank failure can directly cause a release of radiation to the general public.
This change integrated piping used for both Class 1E and non-safety-related instruments, but with safety-related ASME Category 3 pipe.
All CCW pipe is ASME Category 3.
The new design did not delete any of the ASME Category 3 requirements applicable to the subject pipe.
The pipe i
failure probability was thus unchanged.
The bases for CCW Technical Specifications generally requires redundant cooling capacity.
This change maintains that redundancy without affectng cooling capacity.
Thus no margin of safety was reduced.
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- 64. Special Issue I
PEIR 70923, Certifyino Carbon Steel Y-Strainers in the Chilled Water Pumps' Suction Lines Description =of Change This document justifies extended use of carbon steel Y-strainers in the chilled water pumps' suction lines.
They were found in place of removable safety-class spool
. pieces.
The Y-strainer material was not traceable._
Safety Evaluation
~
According to the PEIR response, it is unlikely that these
- 18 3/4" long, low pressure, low temperature pipe segments would fall under-any postulated transient.
a The PEIR response considered the worst case loading on
~these strainers.
The response found typical in-line Y-strainers could accommodate the expected stresses. The typical Y-strainer would have been as strong or stronger i
than the spool pieces called for by the design.
A minimum wall calculation showed that for a six inch diameter pipe, designed for-120 psig and 104 degree F, a thickness 'of less than 0.03 inches is required.
The body of the strainer significantly exceeded that minimum wall thickness.-
WA #01013121 et al replaced the Y-strainers with the designed spool pieces.
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- 65. Special Issue TAR 88-25, Add Waste Gas Compressor Isolation Valve at its Suction and Discharge Headers, Instead of at the Compressor Inlet and: Outlet l
i Description of Change LP&L added one valve on each waste gas compressor header tee to facilitate replacing waste gas _ compressors (re station modification DC-3000).
Reason for Change The valves allowed the gaseous waste management system (GWMS) to Eremain in-service while replacing the-l compressors.
Safety Evaluation The new valves were similar to others in GWMS; therefore,
]
their failure probability was no greater than the other
{
valves in GWMS.
Further, the analyzed GWMS failure (see i
i below) was no more likely during this TAR because of the valves selected.
The presence of the new valves.had no effect on the consequences of a GWMS failure.
While installed, the valves were -" locked-open" unless l
they were actually needed during the replacement.
Closing either-valve would not affect the GWMS accident 1
analysis because that accident focuses on a ruptured gaseous decay tank.
The source term for the
. corresponding GWMS failure dose analyses greatly _ exceeds the source term possible from a simple valve or pipe failure.
The new valves did not affect a Technical Specification LCO, safety setting, or surveillance requirement.
- Thus, the TAR did not affect the margin of safety.
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o 66.
Special Issue y
TAR 88-26 and 88-38, Defeat the Automatic Closure Interlock Description of Change
~
TAR 88-38 de-terminated the automatic closure interlock (ACI) signal to valve SI-405B while LP&L replaced-RC-IPT-0106.
TAR 88-26 installed gaging. devices on SI-401A and SI-401B to work on their motor operators.
.l A "401" and "405" pair form the double isolation between the'RCS'and shutdown cooling.
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Valves-SI-401A and B are interlocked'to prevent. opening
'I at greater than or equal to 392 psig and automatically close at greater than or equal to 700 psig RCS pressure.
The RCS was depressurized and vented when TAR 88-26 was l
installed, thus overpressure protection was not required.
Valve SI-405B was administratively controlled to assure that the TAR was removed before -primary pressure was increased above 350 psig.
The= interlock involved-
~
provides overpressure protection during startup : L and shutdown.
SI-405B was administratively closed while the
-l interlock was inoperative, i
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Special Issue
' TAR 88-37, Cap Drain Line (7SIl-250) Downstream of Valve SI-231A Description of Change Valve - SI-231A leaked by its r,c at - and' caused safety injection tank (SIT) _2A to lose inventory through a a
non-safety related drain.
LP&L cut the drain line'and capped it.to block the leak.
The drain serves a maintenance role, allowing LP&L to remove water from the injection header.
Safety Evaluation PEIR 70901 verified that cutting and capping drain line 7sIl-250 conformed to structural and seismic requirements.
Further, the materials used complied with ANSI B31.1 pipe. requirements.
By complying with all code requirements, this modification did not affect either the probability or the consequences of either an accident or ' a malfunction
-described in the FSAR.
Because the design intends for both the cap and the valve to isolate flow paths, no new-malfunction was created by this TAR.
Lastly,- the TAR did not affect compliance with any Technical Specification LCo, safety limit, or surveil-lance' requirement. Thus the margin of safety remained unchanged while using this TAR.
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T 68.
Special Issue TAR 88-44, Remove One Defective-Heated Junction Thermocouple Heater String From Service
- Description of Change LP&L substituted a resistor for defective heater RC-IHTR-0131B to allow the proper operation of the three-associated heated junction thermocouple (HJTC).
This action only removed one sensor in the probe from service.
The other sensors could then be considered " Technical.
Specification OPERABLE".
Safety Evaluation The remaining configuration fully implemented the assumptions of the FSAR and Technical Specifications.
Thus, this TAR neither adversely affected the probability or consequences of a failure or malfunction, nor reduced the safety margin.
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- 69. Special Ir; sue TAR 88-45, Adding Proximity Probe Brackets to Reactor Coolant Pump Seal Flange Bolts Description of Change LP&L,added proximity and deflection sensors for each reactor coolant pump shaft.
This TAR installed four
=two-pound aluminum brackets to hold the sensors.
LP&L connected the brackets to reactor coolant pump seal flange holddown bolts'.
Those bolts extend several threads'above the holddown nut. LP&L fabricated a second nut to fit above'the holddown nut.
The nut had a plate connected.
The bracket attached to the plate.
Reason for Change LP&L sought to gather data on reactor coolant pump shaf t eccentricity. This data was important in determining the root cause of reactor coolant pump seal failures.
To.
trend the eccentricity, LP&L collected data for.
approximately one year.
The brackets held the sensors and minimized' radiation-exposure to personnel who.
ci otherwise would have had to take the readings.-
Safety Evaluation LP&L. considered (1) the maximum seismic stress on the bracket, (2) the extra hydrogen generation potentia 1'of this TAR, and (3) the effect of the bracket on reactor coolant pump operation.
Calculation EC-088-019 confirmed that the brackets themselves would not fail'during a design basis seismic event.
The addition of four two-pound aluminum brackets to the containment inventory was accommodated by the hydrogen generation margin reported in FSAR 6.2.
L The mass of a bracket was not sufficient enough to affect reactor coolant pump operation.
The probability and consequences of an accident did not change upon this bracket installation because LP&L demonstrated that the bracket would not fail and the p
extra mass of aluminum was inconsequential.
Because the i
bracket was shown strong enough not to break, it could not cause any new malfunction or accident.
- Further, installing these brackets did not conflict with any Technical Specification
- LCO, safety
- limit, or surveillance requirement. Therefore, safety margins were not reduced by this modification.
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- 70. Special Issue TAR 89-15, simulate a PPS Temperature Compensation Loop with a Resistor Reason for Change The compensation loop on resistance temperature detector
~
(RTD) TE-112CB had an intermittent ground that affected a
the temperature signal to ' RCS _ Loop 1 Tc,c,r.n of Core Protection Calculator (CPC) "B"._
This TAR replaced the.
ccmpensation loop with a fixed resistor.
Safety Evaluation-The fixed resistance value chosen (4.7 ohms) was based
~
3 on-field measurement of the compensation loop.
Corresponding manufacturer _ data showed an RTD lead has a 1.3 ohm resistance that changes very little with-temperature.
The balance of the nominal resistance came from field leads which w a s' also a weak function of temperature'.
The resistor introduced a 10.25 degree F error-into CPC channel "B".
- However, Combustion Engineering = confirmed that error would not affect the CPC safety function.
i CPC "B" still operated as assumed in the FSAR; therefore, the consequences of a malfunction would have been' no greater than stated in the-FSAR.
The resistor installation met quality and seismic requirements; so the malfunction probability was not affected.
This TAR has been removed-and the intermittent ground eliminated.
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- 71. - Special Procedure for Installing Westinghouse Mechanical Steam Generator Tube Plugs Description of Procedure This procedure,. adapted from a Westinghouse document, describes the steps and equipment necessary to install and remove mechanical plugs in steam generator tube ends.
Safety Evaluation Plugging tubes that have thinned walls reduces the.
probability of these same tubes causing an accident. : The Westinghouse plug design has many operating-years behind it. substantiating its reliability.
In deference - - to conceins raised in Bulletin 89-01, the plug material used at Waterford is less susceptible-to primary water stress corrosion cracking (PWSCC) than mechanical plugs at North Anna.
Possible accidents from a loose or broken plug include steam generator tube rupture and a reactor coolant pump
-shaft seizure.
A plug _ f ailure could not-cause an accident different from those in Waterford's FSAR.
Overall, plugging enhances nuclear safety.
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