ML20245K940
| ML20245K940 | |
| Person / Time | |
|---|---|
| Site: | Three Mile Island  |
| Issue date: | 06/29/1989 |
| From: | Hukill H GENERAL PUBLIC UTILITIES CORP. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| C311-89-2046, NUDOCS 8907050316 | |
| Download: ML20245K940 (31) | |
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GPU Nuclear Corporation Nuclear
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- -l,, oms:r8o Middletown, Pennsylvan'a 17057 0191 l 717 944 7621 1 l
TELEX 84 2386 Writer's Direct Dial Number:
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June 29,1989 J C311-89-2046 !
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l U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555
Dear Sir:
Three Mile Island Nuclear Station, Unit 1 (TMI-1) j Operating License No. DPR-50 ?
Docket No. 50-289 10 CFR 50.59 Report for 1988 4
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In.accordance with the requirements of 10 CFR 50.59, enclosed are summaries of the changes to TMI-1 systems and procedures as described in the Safety Analysis Report. g i
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Sincerely,
. D.
Vice President & Director, TMI-1 i i
HDH/SM0/spb:2046 cc: W. Russell i F. Young I R. Hernan Enclosure ,
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s9070so316 890o29 I PDR ADOCK 05000'289 !
R PDC 5547 GPU Nuclear Corporation is a subsidiary of the General Public Utilities Corporation
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Procedure: AP.1038 Administrative Controls.- Fire Protection Program (PCR 1-EG-88-050)
Description of Change:
This change. deleted the reporting requirements that were contained in T.S. 3.18 and 4.18. These Tech. Specs. were deleted in Amendment 146. Additional Operations Surveillance Program procedures for recently completed fire suppression systems and several maintenance procedures were added to exhibits.
Other administrative changes were made.
Safety Evaluation Summary:
.The' changes do not present any unreviewed safety questions or deviations from the Fire Protection Program. The changes were administrative in nature.
o e s T 4 Procedure: HVA-1-9 Reactor Compartment Air Temp Hi (PCR 1-05-87-0081)
Description of Change:
The setpoint for the Reactor Compartment Air Temperature alarm was changed from 120 F to 210 F to eliminate nuisance alarms.
Safety Evaluation Summary:
Investigation has shown that actual primary shield concrete temperatures are !
maintained at less than 185 F while air temperatures are at 200-205 F. i ACI-ASME 359-1982 states that the temperature limit for concrete in local areas (penetrations) shall be less than 200oF. The new setpoint provides sufficient margin for corrective action if the alarm should be received.
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4 Procedure: OP 1102-4 Power Operation (TCN 1-88-0018) i l
Description of Change:
The OTSG high level Main Feedwater isolation function of the Heat Sink Protection System may be placed in ' DEFEAT' at the direction of the Plant Operations Director.
Safety Evaluation Summary:
0TSG fouling had caused a steady increase in OTSG levels, increasing the potential for an unintended isolation of Main Feedwater. Analyses have shown that a Main Feedwater overfill represents a less severe pressurized thermal shock transient than that analyzed and approved by the NRC, that the Emergency Feedwater System would not be prevented from performing its design function, that Main Steam piping would not be damaged, and that operator action can accomplish the required isolation after determining the validity of the condition.
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-procedure: OP1102-4PowerOperation-(PCRl1-05-88-0237)' i l
' Description of Change:
This change provides additional guidance on T... reduction,:-rod positioning, and power reduction for end-of-cycle coastdown.
Safety Evaluation Summary:
The T..._ reduction is, limited to that previously experienced in moderator
' coefficient testing, which is a decrease of approximately 5*F. It is reasonable to stay within~this. band where plant stability under automatic ICS control has been demonstrated. The rod position recommendation allows for continued full power operation while maintaining a control band for ICS control.
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3 t procedure: ATP 1210-10 Abnormal Transient Rules, Guides and Graphs (PCR 1-0S-88-0509)
Description of-Change:
The reactor coolant pump trip criterion was modified to allow more time (10 minutes versus 2 minutes) before the pumps have to be left running in a voided system.
Safety Evaluation Sunnary:
I The requirement to trip the pumps immediately upon an indicated loss of subcooling margin is unchanged and still provides the assurance that the requirements of 10 CFR 50.46 are met. The procedure revision reduces the chances of running the pump to destruction when it is not necessary to do so.
Analyses demonstrate that tripping of the pumps within 10 minutes of reactor trip results in acceptable peak clad temperatures.
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.. Proce' dure: ATP 1210-10. Abnormal' Transient Rules, Guides and Graphs (PCR1-0S-88-0509) 1 Description of Changei TheLEmergency.Feedwater Throttling-Criterion for minimum; total flow was revised from 350 gpm to 250 gpm.
Safety Evaluation Summary:.
Recent analyses performed by B&W using TMI-1-plant specific considerations such as emergency feedwater. nozzle elevation and reactor coolant. pump spillover point have demonstrated that as little as 100 gpm'per loop is required to-maintain the core adequately cooled. The value of 125 gpm was chosen to account for the maximum instrument error associated with the EFW flow transmitter.
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' j Proceduresi ATP 1210-1 ~ Reactor Trip .
. (PCR 1-OS-88-0501) )
ATP 1210-2 Loss of'25*F Subcooled Margin (PCR.1-OS-88-0502) -!
! ATP 1210-5 .OTSG' Tube Leak / Rupture' (PCR 1-0S-88-0505)' i ATP 1210-8 lRCS Superheated . . (PCR 1-05-88-0507) l ATP=1210-10 Abnormal Transient Rules, (PCR 1-OS-88-0509) .I Guides and Graphs
-DescriptionhfChange:
The~ required OTSG level.to sustain boiler condenser made cooling was' revised '
.from 90-9b% to 75-85%.
. Safety Evaluation Summary:
Recent analyses performed by B&W used TMI-1 specific considerations.such as power level, emergency feedwater nozzle elevation, and the' different geometry of a Westinghouse reactor coolant pump spillover point. This analysis established the level required to maintain the core ~ adequately cooled as greater than 70% (excluding instrument error). The range of 75-85% was. chosen j to account for the maximum instrument error associated with the OTSG operating range instruments.
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Procedures: B-3-8 BWST Level Lo (PCR 1-0S-88-0454)
E-1-8 BWST Level Lo-Lo (PCR 1-0S-88-0455) 1210-6 Small Break LOCA Cooldown (PCR 1-05-88-0456) 1210-7 Large Break LOCA Cooldown (PCR 1-0S-88-0457) 1101-1 Plant Limits and Precautions (PCR 1-0S-88-0458) 1101-2 Plant Setpoints (PCR 1-05-88-0459) 1104-5 Reactor Building Spray System (PCR 1-0S-88-0460) 1502-5.19 Borated Water Storage Tank (PCR 1-MT-88-8634)
Level Indicator Description of Change:
Increase the Lo-Lo Level Alarm setpoint in the Borated Water f'orage Tank from 3 feet 0 inches to 6 feet 4 inches, and provide revised instiurcions for throttling building spray flow. j Safety Evaluation Summary:
The revised Lo-Lo Level Alarm setpoint provides allowance for level instrument !
error, operator action time, Reactor Building sump valve opening time, and i ensures the transfer is accomplished above an acceptable minimum level in the l BWST.
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Each of the following potentially affected items was evaluated and determined to be acceptable:
ECCS pump performance
- a. assurance of no air entrainment l
- b. affect on recirculation NPSHA Reactor Building sump pH !
Reactor Building pressure and temperature Offsite dose Boron concentration and reactivity I
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Modification: Installation of Local Flow Measurement instrumentation for-SW-P-2A/.B_and AH-P-3A/B-(BA 412484)
Description of Modification:
This modification provided local flow measurement for the two pumps in the Screerhouse Ventilation Cooling Water System (SW-P-2A/B) and.the two pumps-in the Control . Building Chilled Water System ( AH-P-3A/B).
Summary'of Safety Evaluation:
This modification installed pressure differential instrumentation for-In-Service Testing purposes. The installation of local flow elements and flow indicators does not affect the' performance of either system as it provides indication and has no control function. No-existing Tech. Specs, were affected and only drawing updates were required for the FSAR.
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Modification: installation of Auxiliary Current Transformers for Auxiliary
, Transformer (BA'128967)
Description:
TheLinsta11ation of auxiliary current transformers corrects the mis-coordination between the' auxiliary transformer ground _ relay and a 4.16KV bus incoming relay. One auxiliary current transformer was placed in.the control. cabinets of each auxiliary transformer'and connected in the ground relay current transformer'_ circuit.
Safety Evaluation Summary:
This modification' corrects an existing 4.16KV ground' relay mis-coordination therefore ensuring that a fault on a 4160V bus will be cleared by the bus '
incoming breaker instead of potentially separating the offsite sources to the unit electrical distribution system. The reliability of the power sources to the balance of the plant auxiliary transformers is enhanced.
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Modification: Addition of Nuclear Services Closed Cycle Cooling Water Check-Valve (BA 413930).
Description of Modification:
This modification' installed.a Seismic Class 1 check valve (NS-V-205) between the Demineralized Water System (DW) and the Nuclear Services Closed Cycle Cooling Water System.(NSCCW). Prior to the modification, there was a normally open globe valve (NS-V-102) which was a boundary between a Seismic Class 1 system (NSCCW) and a non-seismic system (DW). In the event of-a seismic, failure of the DW line, NS-V-102 would not have precluded the loss of. water inventory and tank pressure. The new check valve prevents these losses.
Safety Evaluation Summary:
This modification provides a passive seismic boundary for the NSCCW/DW' system interface and increases the reliability of the NS-TI surge tank. This modification does not alter the operation or functional design requirements of the NSCCW' system.
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Modification: Emergency Feedwater Steam Trap Drain Modification (BA 123219).
Description of Modification:
This modifiention provided an-improved drain ~ syst'em so that the Emergency Feedwater. Turbine will be protected from exposure to potential damage resulting from' condensate. carry-over to the turbine.during startup of the turbine. The modification involved rerouting the one inch steam trap discharge to a nearby floor drain and carping the existing discharge to the main condenser.
Safety Evaluation Summary:
This modification ensures that no condensate will be carried over to the EFW L turbine and therefore increases the reliability that the turbine driven EFW pump will perform its safety function.
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Modification: Valva Addition for Supplying Cooling Water for Conductivity and Ammonia Analyzers (BA 418624)
Description of Modification:
As part of the Secondary Plant Chemistry upgrade, the conductivity and hydrazine analyzers for the condensate booster pump discharge were relocated from the Secondary Plant Sample Room to the Turbine Building. Valves SC-V188 and SC-V189 were added in secondary services cooling water piping to supply cooling water to the analyzers,. q Safety Evaluation Summary:
The installation of these valves does not alter the operation of any safety-related equipment. The secondary closed cooling water system is not in the basis of any Technical Specification. '
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Modification: Coadenser. Pit Sprinkler System (BA 413925)
Description of Modification:
< This modification' installed a wet' pipe sprinkler system in the Condenser Pit area of the TMI-1 Turbine Building. The new system provides automatic fire.
suppression water' coverage for this area. The modification was in response to-
'American Nuclear. Insurers (ANI) Recommendation:(81.5a) to protect the area against the ignition of combustibles such as turbine lube oil which could accumulate under the condenser on elevation 292, feet of the Turbine Building.
The modification involved the connection of a supply line from the Fire Service Water header to.a piping. distribution system containing' closed-head-fusible 2
sprinkler nozzles. The supply line contains an isolation valve. .The existing Alarm PLA-7-9 provides control room indication of system actuation.
Safety Evaluation Summary:
The modification improves system performance by providing automatic water spray coverage for the Condenser Pit' area, thereby improving the fire suppression capability for the area' . The existing Turbine Building sprinkler systems will.
not be adversely affected since a fire is postulated to occur in only one area at a time; only an individual sprinkler system is considered to place a demand on the Fire Service Water pumps. The demand of the Condenser Pit System will not have any adverse effectsoon the Main Fire Service Water Supply since it will be designed as a pipe schedule system in accordance with NFPA 13, 1987. ,
The pressure and. flow available at the inlet of the system is equal to that '
supplied off of the main header-to the other Turbine Building sprinkler and deluge systems.
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Modification:' Fire Service Tie-In for Outage Equipment Storage and Outage Support Buildings (BA_418638)
' Description of Modification:
.This modification installed a water service.line from an existing hydrant'in the Fire: Service Water System to provide fire service water to the.0utage
'_ Equipment Storage and Outage Support Buildings.
Safety Evaluation Summary:
'This modification does not affect the system performance requirements of the Fire Service Water System. The new line does not prevent the yard main'from
' supplying any existing systems.
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. .. - 1 it ' Modification: FuellHandling Bridge Equipment Upgrade (BA 412232)
-Description of Modification:
The main objective of the fuel handling bridge upgrades was to imp' rove the control rod mast performance on the main fuel handling bridge. This was accomplished by replacing the existing hydraulic rod drive system with a hoist driven mechanical rod drive system that provides greater reliability, speed of operation and Mark B-4 or B-5 control rod handling capability. A_new load i sensing system, new fuel assembly grap;,les, new control console, a TV i positioning system, an Inching Modification system, and programmable geared {
limit switches'were installed. 1 Safety Evaluation Summary:
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The fuel handling bridge upgrades increase fuel handling equipment reliability while decreasing equipment maintenance. The operating modes and system control have remained the same. These modifications do not affect the basis for any Technical. Specification.
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, j Modification: Closed Circuit Television for the Reactor Building (BA 412537) i Description of Modification:
This modification provided a closed circuit television system (CCTV) inside the Reactor Building at TMI-1 for observation / surveillance of various areas inside containment. The CCTV system consists of four (4) permanent television cameras l inside the "D" ring area and associated equipment. This associated equipment has the capability of utilizing up to a maximum of eight (8) cameras. The cameras and associated equipment are left in place inside containment and operational during plant operation.
Safety Evaluation Summary:
This modification added approximately 44 lbs. of aluminum material to the !
500 lbs of aluminum already inside containment (per FSAR 6.5.3) bringing the total amount of alumin. n to 544 lbs. As a result of this addition, the hydrogen concentration will reach three percent of the total containment volume in a shorter time period - 9.8 days in lieu of the 11 days as indicated in the FSAR. This addition of aluminum was acceptable.
Nuclear safety or safe plant operations are not adversely affected by this modification because components added with materials prohibited inside containment have been evaluated as having no impact on safety systems function.
In addition, cabling and equipment installed inside and outside containment and 1 conduit installed inside containment does not constitute a hazard to safe plant operations because the design and installation in accordance with existing plant electrical separation and seismic (anti-falldown) criteria. Cabling is protected by conduit in areas where the cabling is susceptible to mechanical damage (walkway areas, etc.).
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Y -Modification:' Two Hour Backup Instrument Air Charging _ Compressor.
(BA 412525)
Description of Modification: o 1 1 This modification' installed a permanent source of charging air for the Two Hour Backup Instrument Air System (2HBUIA) storage bottles. This modification added an air compressor, power supply, air filter, air. dryer and supply tubing.to.
replace a truck mounted breathing air. storage tank.
Safety Evaluation Summary:
Installation of the charging air compressor does not alter the function, design requirements'or quality standards of the 2HBU1A system or any other safety related system.
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Modification: FW-V5A/B Reactor Trip Interlock (BA 128968) 1 Description of Modification:
This modification changed the interlocks for opening and closing the Main i Feedwater Block Valves FW-VSA/B in order to improve plant control during startups and incorporate B&W Owners Group Safety and Performance Improvement Program (SPIP) Recommendations. The interlock to open FW-V5A/B when the startup valve (FW-V16A/B) is at 90% open was removed. The close interlock was changed from Startup valve less than 70% open to reactor trip. I Safety Evaluation Summary:
The margin of safety is not reduced as well as nuclear safety and safe plant operations are not adversely affected. By closing the MFW block valves on a Reactor Trip the probability of overfeed or overfill events occurring is reduced and the block valve will still close prior to an overcooling event due to the reactor trip on either high or low pressure. Therefore, Known-Safe-State for ICS/NNI power failures is not adversely affected by this .
modification. By removing the interlock to automatically open the MFW Block I Valves smooth MFW control can be achieved during plant startups, by eliminating MFW transients induced by the block valves automatically opening.
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Modification: Load Reduction on.USS 1P and 1S (BA 412520)
Description of Modification:
This' modification reduced the loading on USS IP and 1S'(DG 1A and 1B loading) n by transferring 230KV Substation breaker loads and count room loads to B0p-busses. Additionally the obsolete Motor-Generator Set (supplied. count room equipment).was replaced with a regulating. transformer. This modification was worked in conjunction with a Metropolitan Edison modification to install backup emergency diesel generators for the substation breaker loads.
Safety Evaluation Summary:
This modification was implemented to improve the performance and increase the margin of available Diesel Generator power. The modification increased the-c safety margin of the 4B0V system by transferring non-essential, non-safety.
. loads to a B0P bus.
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r Modification: -WDL-V535 Low Level Interlock (BA 413916)
L Description of Modification:
o This modification supported lowering the Reactor Building sump minimum level to L -allow more effective use of the Miscellaneous Waste Evaporator by eliminating unnecessary cycling. The modification involved the replacement of a setpoint module. Loweri_ng the level was accomplished by changing the setpoints for the Reactor Building sump minimum level and the interlock for closing the sump drain valve WDL-V-535.
Safety Evaluation Summary:
This modification provided the hardware' changes necessary to. lower.the Reactor
. Building sump minimum level. Lowering the low level interlock setpoint ensures l
. that WDL-V535 closes automatically prior to uncovering the gravity drain line between the RB sump and the Aux. Bldg. sump. This ensures the RB atmosphere and Aux. Bldg. atmosphere will not be tied together. A calculation was performed to verify that the post-LOCA RB sump level will be sufficient to satisfy Decay Heat and Building Spray pump NPSH requirements.
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Modification: Penetration Pressurization Purge Interspace Automatic Actuation-Elimination and Valve Upgrade (BA.128124)
Description of Modification:
This modification converted check valves to normally closed globe valves in the Penetration Pressurization supply lines to the purge ~ valve interspaces and eliminated the auto initiation. valves. The globe valves are now considered containment isolation valves.
Safety Evaluation Summary:
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malfunction than the check valves to assure containment integrity. The safety
- l. analysis did not rely on the automatic purge interspace pressurization l- provision, therefore, there is no increase in the probability or consequence.of.
an accident previously evaluated.
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Modification:- Replacement of Core Flood Valve 1A and IB Motors (BA 418668)-
Description ~of Modification:
This modification replaced the two pole high speed motors on the Limitorque valve operators for CF-V-1A and CF-V-1B with lower speed four pole motors.
3 Safety Evaluation' Summary:
Installing lower speed motors on core flood discharge valves does not adversely-if > affect the safety function or the response time of the Core Flood System.
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- Modification: Removal of RM-G16 through RM-G21 Local Readouts (CMR #88-161)
Description of Modification:
The RM-G16, G17, G18, G19, G20, and G21 local meters and alarms were removed.
Safety Evaluation Summary:
The subject local readouts were considered maintenance burdens and, in some instances, ALARA concerns. The design intent of these monitors.is to provide containment isolation capability on the particular process lines being monitored. RM-G16 through RM-G21 are not required for' personnel protection; other equipment is installed for this purpose. Active components required to )
meet the containment isolation function are unaffected by this change. '
Therefore, removal of local readouts has no adverse impact on nuclear safety.
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Modification: -Replacement of the RM-L12 Ratemeter-(CMR 88-121)
Description of Modification:
l Nuclear..Research Corporation'DRM-200 digital ratemeter was installed for RM-L12 as a replacement for the obsolete AR-25 ratemeter and MQ234 signal analyzer.
Safety Evaluation Summary:
Installation of the replacement ratemeter does not change RM-L12 interlock or l performance capabilities. The efficiency of the RM-L12 channel is unchanged {
and the I-131 and Cs-137 sensitivity are unaffected. However, while reviewing d the FSAR in preparation of the above referenced S.E., it was discovered that the-FSAR referenced I-131 sensitivity-for RM-L12 is incorrect. The original system design requirements are'not affected by this. change.
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Modification: OTSG Blowdown System'(BA 412081)
Description of Modification:
This modification added a' cross-connection between the existing Liquid Sampling System and the Hot Drain System to allow the Once Through Steam Generators (OTSG) to be blown down to the Main Condenser to assist in controlling secondary side water chemistry. This assistance is desirable at power levels below 30% as in this regime the OTSGs are not operating in the true once-through mode but rather are recirculating the secondary coolant, thereby concentrating water impurities. Additionally, a modification was made to existing drain piping to increase the flexibility of use for portable recirculation / wet lay-up subsystems.
Safety Evaluation Summary:
This modification leaves the Liquid Sampling System's ability to permit remote sampling of OTSG secondary side coolant undisturbed during normal operations, 1 unit cooldown and post-accident conditions. No aspect of containment integrity is challenged.
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- Modification: L Use' of Morpholine- FeedwaterChemistry at TMI-1 '(BA 128130);
n Description of Modification:
' Morpholine.is now used'as a corrosion' inhibitor in the feedwater/ steam /
condensate system asla supplement to the all-volatile treatment _normally.
employed:in the _ secondary side' of _ a pWR.
Safety Evaluation Summary: ,
There'are no adverse; effects that res' ult from the use of morpholine and long term improvement in secondary system corrosion is achieved. The use of morpholine at TMI-1 was qualified for the materials of construction. It was-
. determined-that there was no increase in the probability or. consequences of an accidentior' malfunction of eguipment evaluated in the FSAR.
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.,t,=. -i l Modification: Letdown Reactor Building Isolation Valve (MU-V3) Circuit !
Modification (BA 128973)
Description of Modification:
This modification affects the Reactor Building isolation signal for MU-V3.
MU-V3 will now isolate the Reactor Building on 1600 psig or 500 psig RC pressure or 4 psig RB pressure ES signals versus the present RB isolation on i either a reactor trip or a 4 psig RB pressure. This change will reduce the thermal cycling on letdown coolers due to reactor trips.
Safety Evaluation Summary:
The pressure and temperature conditions experienced by the letdown system i before isolaticn occurs (on 1600 psig RCS pressure or 4 psig Reactor Building j pressure) are within the design limits of the system. Those accidents for '
which a reactor trip signal is generated but not the 1600 psig isolation signal were evaluated. It was determined that there was no increase in the probability or consequences of an accident or malfunction of eq'sipment i evaluated in the FSAR. No new type of accident or transient was introduced into the plant design.
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Modification: . Main Generator Protection Improvements (BA 412444)
Description of Modification:
This modification included the following activities:
- 1. Addition of a reverse power. relay for generator anti-motoring protection.
- 2. Addition of a second reverse power relay to prevent accidental overspending of the generator by converting the existing simultaneous tripping scheme to a sequential tripping scheme for normal shutdown.
- 3. Addition of a pyrolysate collector for generator hydrogen gas sampling.
Tagging compounds were added to the generator insulation paints'to enable i the area of local overheating to be identified by gas chromatography. !
I The existing Main Turbine / Generator protection automatic tripping functions (e.g. turbine overspeed trip) were not changed. l l
Safety Evaluation Summary:
These main generator protection improvements provide increased protection of the main generator in accordance with the GPUN evaluation of the manufacturer's i recommendations. This modification performs no nuclear safety function and has been designed so that the equipment it adds will not become a missle that could 4 damage safety related equipment. This modification does not adversely affect nuclear safety and does not involve an unreviewed safety question.
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. Modification: ICS/NNI Enhanced Reliability.(BA 412534) , )
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Description of Modification: i i
This' modification included the following changes: j
- 1. Supply " Auto" and " Hand" power from separate external power supplies, j
- 2. AddLauto transformer to inverter 1E so as to maintain normal voltage to the l
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- 3. Isolate." Auto"Tand " Hand" power sources including neutral.
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.4. Incorporate " Smart Automatic Signal Selector".(SASS) so as to prevent a trip due to failure of one of'the redundant input to the ICS/NNI.
i The-ICS/NNI.. functional operation during nor:nal operation remains the same.
' Safety Evaluation Summary: q The above changes improved the reliability of'.the ICS/NNI such that a' failure. 1
-of.a. single external power supply would not cause the plant to trip. If a
. redundant input to the ICS/NNI failed, SASS would automatically switch to the z 1 operating input. This modification does not adversely affect nuclear safety and .does not . involve an unreviewed safety question.
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- fjodification: Fire Pump Control Circuit Modification (BA 412539)
Description of Modification:
The purpose of this modification is to modify the control circuitry of the fire !
pump such that it will not automatically start during LOCA conditions.
Safety Evaluation Summary:
This modification prevents the fire pump from being loaded onto the emergency diesel generators in the event of a Loss of Coolant Accident (LOCA). This prevents an overload (above 3000 KW) of the operating emergency diesel generator if the redundant emergency diesel generator fails.
Fire fighting capability is not degraded because this pump is backed up by diesel driven fire pumps. This modification does not adversely effect nuclear safety and does not involve En unreviewed safety question.
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