ML20100P062
| ML20100P062 | |
| Person / Time | |
|---|---|
| Site: | Fermi  |
| Issue date: | 12/31/1995 |
| From: | DETROIT EDISON CO. |
| To: | |
| Shared Package | |
| ML20100P050 | List: |
| References | |
| NUDOCS 9603110024 | |
| Download: ML20100P062 (14) | |
Text
, Enclosure to
. NRC-96-0017 Page1' i
FERMI 2 ANNUAL OPERATING REPORT JANUARY I - DECEMBER 31,1995 l
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' DETROIT EDISON COMPANY NRC DOCKET NO. 50-341 FACILITY OPERATING LICENSE NO. NPF-43 4
9603110024 960229 PDR ADOCK 05000341 R PDR
, Enclosure to
. NRC-96-0017 Page 2' Table of Contents Page 1.0 Introduction 3 2.0 Summary of Operations 3 [
' 2.1 Summary of Operating Experience 3 2.2 Summary of Outages and Forced Reductions 3 l Greater than 20 Percent ofFull Power ,
2.3 Fuel Performance 5 i 2.4 Shore Barrier Survey 6 2.5 Safety Relief Valve Challenges 6 l 2.6 Personnel Monitoring and Exposure 6 2.7 Service Life of Main Steam Bypass Line 8 ;
2.8 Specific Activity Analysis of the Primary Coolant 8 l 2.9 ECCS Cooling Performance Evaluation Model 8 Changes or Errors 1 2.10 ECCS Outages 9 I
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i 1.0 : Introduction The Fermi 2 Nuclear Power Plant site is located on the western shore of Lake Erie in j Frenchtown Township, Monroe County, Michigan. The Nuclear Steam Supply System is a e General Electric BWR 4, with a Mark I pressure-suppressian containment. The plant is fully owned by the Detroit Edison Company. ;
2.0 Summan of Operations 2.1 Summary of Operating Experience i Fermi 2 was synchronized to the grid in January 1995 aAer a prolonged shutdown following recovery from a turbine failure that occurred on December 25,1993. During l 1995, with the main generator paralleled to the Detroit Edison electrical grid, the l reactor automatically scrammed two times; once on April 25,1995, due to a pressure .
regulator failure, and once on June 2,1995, due to a turbine trip during mechanical overspeed testing due to an inadequate mechanical overspeed test procedure. The first part of the year was focused on startup and power ascension testing. For the second <
half of the year, starting after the June 2,1995, scram, the reactor maintained power operation with four power reductions, three for testing and preventive maintenance, and one for main generator hydrogen cooler repairs. Core Thermal Power (CTP) is ' l administratively limited to 96 percent due to turbine limitations.
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2.2 Summan of Outages and Forced Reductions Greater than 20 Percent of Full Power January 27,1995 - 127.5 Hours Shutdown Turbine taken off-line to perform scheduled post-outage balancing.
February 1,1995 - 259 Hours Shutdown Forced continuation of outage that began on January 27,1995. Turbine remained off-line to repair the Number 4 jacking oil pump discharge piping. ,
l' February 13,1995 - 663. 2 Hours Shutdown Turbine taken off-line to repair turbine oil system structural concerns. ]
l March 14,1995 - 18.1 Hours Shutdown '
Turbine taken off-line to perform scheduled post-outage balancing.
i March 16,1995 - 20.6 Hours Shutdown Turbine taken off-line to obtain scheduled turbine coastdown bearing vibration data.
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March 26,1995 - 47.8 Hours Shutdown ,
i Turbine taken off-line (forced) to repair a stator cooling water vent line leak. !
April 9,- 1995 - 38.1 Hours Shutdown !
Scheduled manual reactor trip to obtain hot turbine coastdown vibration data at l approximately 80% power. This event was reported ir. LER 95-004.
Approximately fourteen seconds after the scram insertion, reactor water level approached Level 2, causing partial actuation of Level 2 safety functions. The i reactor recirculation pumps tripped, the high pressure coolant injection system !
started but did not inject, and the reactor core isolation cooling system initiated. I The corrective action was to review the post scram fcadwater logic to determine ifimprovements are possible. This review is ongoing and any necessary improvements are planned for the fifth refueling outage.
1 April 12,1995 - 41.8 Hours Shutdown !
i Turbine taken off-line (forced) to repair main steam to moisture separator reheater isolation valve.
-l April 25,1995 - 249.3 Hours Shutdown l
l Automatic reactor scram on average power range monitor neutron upscale trip j resulting from reactor pressure regulator transient. This event was reported m >
LER 95-005. Replacement of pressure regulator circuit boards and ' !
potentiometers, additional testing to locate potential sources of voltage signal introduction or current loading, and removal of the pressure regulator monitoring system interconnection to the pressure regulator control circuitry provide a high ;
level of confidence that the cause of the event has been corrected. ')
May 13,1995 - 27.5 Hour Power Reduction Reduced reactor power to 50% to repair tube leaks in main condenser.
June 2,1995 - 333.7 Hours Shutdown Automatic main turbine trip on mechanical overspeed while performing main turbine generator overspeed trip test. This event was reported in LER 95-006.
The main turbine tripped because the test on the number 1 trip ring sufficiently vibrated the number 2 trip lever when the "On Load Overspeed Test" push-button was depressed. This occurred because the number 2 trip lever had not fully latched into its reset state during the performance of the overspeed test (27.112.01) on May 4,1995. Procedure 27.112.01 was revised to incorporate holding the reset overspeed push-button depressed for at least 15 seconds to preclude this event from recurring.
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. NRC-96-0017-Page5'
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j- October 13,1995 - Power Reduction l
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i Reactor power was reduced to 65% power for control rod scram time testing and i main turbine valve surveillance testing.
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October 20,1995 - Power Reduction t
Reactor Power was reduced to 70% power for power suppression testing and '
control rod deep / shallow position swapping. -l l
l November 3,1995 - Power Redu: tion Reactor power was reduced to 65% power for control rod scram time testing and scram solenoid pilot valve work.
t ' December 10,1995 - Power Reduction v
$ Reactor power was reduced to 25% power to isolate the main generator southeast
- hydrogen cooler for leak rcpairs and to perform main turbine valve surveillance testing.
4 2.3 Fuel Performance The reactor was operating in Mode 2 at the start of 1995, On January 7,1995, reactor
- operation in Mode I commenced.'
, During 1995 the plant produced about 242 effective fuel power days based on a 3430 )
MWt rating. Following return to Mode 1 operation, the offgas activity at 3293 MWt .
settled at about 100 uCi/sec, confirming the success of the RF04 fuel sipping campaign.
However, during October a leaking fuel rod was confirmed. At the end of the year, the reactor was operating at 96 percent power. Core Thermal Power (CTP) is administratively limited to 96 percent due to turbine limitations.
- On October 9,1995, the offgas radiation monitor rose from 6.5 mR/hr to 10 mR/hr, and decreased back to 8 mR/hr within 20 minutes, and eventually came back to 7.
mR/hr. On October 13th a slight drop in the offgas radiation monitor reading occurred j in conjunction with the movement of control blade 14-47. During an October 14th j
. downpower Xe", a leaking fuel " tracer" isotope, more than doubled. i i
On October 21,1995, a power suppression test was performed which identified the ;
i suspected region of the core containing the leaking fuel bundle. Control blade 10-47 was fully inserted, and control blade 06-47 was left fully inserted to suppress power
. ' levels in this suspect region. Also, as a conservative measure, ramp rate restrictions ,
were applied to the bundles in cell 10-47. A supplemental power suppression test was .!
performed on November 4,1995, in conjunction with control blade operability testing. )
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This test provided more evidence that the suspect region of the core contained the leaking bundle, and a third control blade (10-43) was fully inserted to further suppress power in the suspect region. These efforts are intended to reduce the degradation rate r
. of the leaking fuel rod.
At the end of the year offgas activity continued to hover slightly above 100 uCi/sec providing evidence that the power seppession effort on the leaking bundle has been successful. ' Because of the low ofigas activity, the leaking fuel rod is believed to have a very tight leak. l i
2.4 Shore Barrier Survey A survey of the Fermi 2 shore barrier was completed as required by Technical l Specification 4/1.3. The results of the survey indicated no damage, significant I movement, or deterioration of the barrier. All forty-seven survey point elevations were within the tolerance specified in Technical Specification Table 3.7.3-1. Civil l Engineering Drawings 6C721-44 through 49 were revised to incorporate the survey ]
data. No unusual incidents occurred in 1995 that would have required additional l surveillance.
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2.5 Safety Relief Valve Challenges There were no safety relief valve challenges during 1995. l l
2.6 Personnel Monitoring and Exposy.rs !
Table 2.6-1 provides a breakdown of radiation exposure by work and job function as !
required by Technical Specification 6.9,1.5.a. !
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Enclosure to NRC-96-0017 .
Pabe 7 Table 2.5-1 Detrost Edison Fermi 2 1995 Regulatory Guide 1.16 Direct Reading Dosemeter (DRD) Deep Dose Equivalent (DDE) Dose Report -
Personnel Receiving Exposure DRD DDE Monrem Function Department Sistion Utility Contred Station Utigty Contract Employees Employees Warhers Employees Employees Workers -
Rouhne Marilenonce 143 2 65 4.624 0.005 2.196 Opershon & Opershons 109 1 24 5 831 0.000 3.912 Survesence Hoekh Physses 31 0 2 2.446 0 000 0.201 Supervisory 110 15 128 1.759 0 010 0.977 Engpneenna 121 2 6 1.294 0 000 0 007 Reactor Maintenance 5 0 1 0.3S9 0 000 0.132 Mesnienance Opershons 0 0 0 0.000 0.000 0.000 HesRh Physses 0 0 0 0 000 0 000 0.000 Supennsory 0 0 0 0.000 0 000 0 000 Engpneenna 0 2 0 0 000 0 000 0 000 insennce Maintenance 0 0 0 0.171 0 000 0.000 Inspechon Opershons 0 0 0 0 000 0 000 0.000 Health Physses 3 0 0 0.475 0.000 0.000 Supervisory 0 0 1 0 000 0 000 0 026 Engineenna 0 0 0 0 000 0.000 0 000 _,
Special Maintenance 20 0 68 1.725 0 000 1.427 Momtenance Opershons 5 0 4 0 000 0.000 0.109 Health Physses 6 0 0 0.735 0.000 0.000 i Supervisory 9 0 56 0.516 0.000 1.815 Engineenna 2 0 1 0 004 0 000 0 013 Weste Maintenance 0 0 1 0 000 0.000 0.017 Processma Opershons 0 0 0 0 000 0.000 0.150 HesRh Physics 1 0 0 0.000 0.000 0.000 Supervisory 0 0 3 0.295 0.000 0 045 Engineering 0 0 0 0 000 0.000 0 000 Total Maintenance 168 2 135 8.879 0.005 - 3.772 Operations 114 1 28 5.911 0.000 4.171 Heefth Physics 41 0 2 3.951 0.000 0.201 T; ..-1 119 15 188 2.275 0.010 2.883 E. _ _ _ _ _ _ ;.
- 123 4 7 1.238 0.000 0.029 Grand Total Personnel 947 Montem 31.356 NOTE: Thic report was produced using only secondary (D8tD) extemed dosimetry - R does not include any intemel exposure
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. .- ' Enclosure to .
- .NRC-96-Odl7
' Page 8 2.7 Service Life of Main Steam Bvoass Lint !
In accordance with Detroit Edison letter VP-86-0154, dated November 7, !
1986, the cumulative time the main steam bypass lines are operated with the !
bypass valves between 30 percent and 45 percent opened will be reponed annually. A cumulative value of 100 days is not to be exceeded without prior i NRC notification.
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Evaluations performed by Stone and Webster and by Hopper and Associates 1 concluded that the bypass lines are acceptable for safe operation when !
, operated within the 100 day constraint. Based on these evaluations, the new i
! main steam bypass piping that was installed in 1985 has a service life which j will allow it to function for the life of the plant under anticipated operating l conditions. The main steam bypass lines usage was 36.10 days as of j December 31,1995.
'l 2.8 Specific Activity Analysis of the Primary Coolant Exceeding ,
the Limits of Technical Specification 3.4.5 During 1995, the specific activity of the primary coolant did not exceed the :
' limits of Technical Specification 3.4.5. l l
2.9 ECCS Cooling Performance Evaluation Model Changes or Errors 1
During 1995, there were no changes or errors discovered in an evaluation model for calculating ECCS cooling performance or in the application of a ECCS cooling performance evaluation model that affects the temperature ;
calculation of the fuel cladding by Detroit Edison. j However, on December 15,1995, General Electric (GE) reponed to the NRC,
' in letter RJR:95-118, that a concern had been raised by a domestic utility which could affect the GE LOCA analysis. GE believes that this concern will !
impact peak clad temperature less that 10 degrees Fahrenheit based on !
engineeringjudgement and extrapolation of previous LOCA analysis. An !
event is considered by the NRC, pursuant to 10CFR50.46(a)(3)(II), to be l significant if the peak clad temperature increase is more than 50 degrees Fahrenheit. Therefore, the amount of change described in GE letter RJR:95- )
118 is considered insignificant and well within the margins of safety analysis. I i
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. NRC-96-Od17 Page 9 2.10 ECCS Outages Pursuant to Fermi 2 Technical Specification 6.9.1.5.c, a summary of the ECCS system outages which occurred between January 1,1995 and December 31, 1995 is provided. The tabulation of ECCS outage hours (Table 2.10-1) includes both forced and planned outages for the Low Pressure Coolant Injection (LPCI), Core Spray, High Pressure Coolant Injection (HPCI), and Automatic Depressurization Systems (ADS). An' outage was considered to be whenever one of the ECCS systems was out-of-service at a time it was !
required to be operable per Technical Specifications. !
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. . Enclosure to NRC-96-Odl7 Page 10 ECCS Outanes Table 2.10-1 ECCS Outage Hours !
January 1,1995 to December 31,1995
-l ECCS SYSTE_M FORCED HOURS PLANNED HOURS l LPCI Division I 0.0 40.36 ;
LPCI Division II 120.58 22.25 i Core Spray Division I. 0.0 16.27 i Core Spray Division II 0.0 40.10 I ADS Division I 0.0 - 62.25 j
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ADS Division II 9.5 0.0 HPCI 95.52 101.02 l DIVISION I LOW PRESSURE COOLANT INJECTION ECCS System Outage: Division I Low Pressure Coolant Iniection Out of Service from 1754 02/16/95 - to 2120 02/16/95 '
Duration: 3,43 hours4.976852e-4 days <br />0.0119 hours <br />7.109788e-5 weeks <br />1.63615e-5 months <br /> Planned Outage
,i Outage Summary; Surveillance Performance ECCS System Outage: Division I Low Pressure Coolant Iniection Out of Service from 0049 05/17/95 to 0640 05/17/95 Duration: 5.85 hours9.837963e-4 days <br />0.0236 hours <br />1.405423e-4 weeks <br />3.23425e-5 months <br /> Planned Outage l
Outage Summary: Surveillance Performance j ECCS System Outage: Division I Low Pressure Coolant Iniection Out of Service from 1045 08/14/95 to 1500 08/14/95 Duration: 4.25 hours2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br /> Planned Outage Outage Summary: RHR Pump A taken out of service to replace discharge pressure switch.
I ECCS System Outage: Division I Low Pressure Coolant Iniection Out of Service from 0520 08/16/95 to 0810 08/17/95 Duration: 26.83 hours9.606481e-4 days <br />0.0231 hours <br />1.372354e-4 weeks <br />3.15815e-5 months <br /> Planned Outage Outage Summary: ' RHR and RHRSW Systems Outage to pr,rform various PM activities. Following completion of the activities and required surveillances, the Division I LPCI System was returned to service.
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Enclosure to NRC-96-00'17 Page11 DIVISION 11 LOW PRESSURE COOLANT INJECTION ECCS System Outage: Division II Low Pressure Coolant Iniection Out of Service from 2010 02/21/95 to 2045 02/26/95 Duration: 120.58 hours6.712963e-4 days <br />0.0161 hours <br />9.589947e-5 weeks <br />2.2069e-5 months <br /> Forced Outage Outage Summary: The Division II LPCI System, RHR Pump B, was removed from service !
to repair check valve El1-F03IB. .
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ECCS System Outage; Division II Low Pressure Coolant Iniection Out of Service from 2200 03/29/95 to 1500 03/29/95 l Duration: 0.53 hours6.134259e-4 days <br />0.0147 hours <br />8.763227e-5 weeks <br />2.01665e-5 months <br /> Planned Outage j Outane Summaw: Suweillance Performance ECCS System Outage: Division II Low Pressure Coolant Iniection t L Out of Sewice from 1339 04/27/95 to 1413 04/27/95 ,
Duration: 0.57 hours6.597222e-4 days <br />0.0158 hours <br />9.424603e-5 weeks <br />2.16885e-5 months <br /> Planned Outage l Outage Summary: Isolated EECW first to RHR Pump B, then to RHR Pump D, to . l check for EECW blockage. EECW isolation causes inoperability of RHR pumps. 1 ECCS System Outage: Division II Low Pressure Coolant Iniection 1 Out of Service from 0900 07/18/95 to 1353 07/18/95 )
Duration: 4.88 hours0.00102 days <br />0.0244 hours <br />1.455026e-4 weeks <br />3.3484e-5 months <br /> Planned Outage )
l Outage Summary: The Division II LPCI System, RHR Pump B, was removed from service
- to perform an oil change.
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L ECCS System Outage: Division II Low Pressure Coolant Iniection Out of Service from 0142 09/30/95 to 062109/30/95 Duration: 4.65 hours7.523148e-4 days <br />0.0181 hours <br />1.074735e-4 weeks <br />2.47325e-5 months <br /> Planned Outage Outane Summary: Surveillance Performance i
l ECCS System Outage: Division II Low Pressure Coolant Iniection Out of Service from 1923 12/27/95 to 070012/28/95
! Duration: 11.62 hours7.175926e-4 days <br />0.0172 hours <br />1.025132e-4 weeks <br />2.3591e-5 months <br /> Planned Outage Outane Summan: Surveillance Performance I I
Enclosure to NRC-96-00'17 Page 12 DIVISION I CORE SPRAY ECCS System Outage: Division I Core Spray Out of Service from 0544 07/05/95 to 2200 07/05/95 Duration: 16.27 hours3.125e-4 days <br />0.0075 hours <br />4.464286e-5 weeks <br />1.02735e-5 months <br /> Planned Outage Outage S_ummary: The Division I Core Spray System was removed from service to perform various PM activities. Follow'mg completion of the activities and required surveillances, the Division I Core Spray System was returned to service.
DIVISION II CORE SPRAY ECCS System Outage: Division II Core Sprav Out of Service from 0524 08/03/95 to 2130 08/04/95 Duration: 40.10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> Planned Outage Outage Summary: The Division II Core Spray System was removed from service to perform various PM activities. Following completion of the activities and required surveillances, the Division II Core Spray System was returned to service.
DIVISION I AUTOMATIC DEPRESSURIZATION SYSTEM ECCS System Outage: Division I ADS Out of Service from 1045 08/14/95 to 1500 08/14/95 Duration: 4.25 hours2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br /> Planned Outage Outage Summary: Division I ADS was removed from service to replace ElIN021 A, flow indicating switch; RHR pump C002A discharge flow permissive for min-flow line, per EDP 27275 (replacement of D/P switch with new " blind" D/P switch). However, due to the critical load forecast at this time due to hot weather, the work order was deactivated to prevent any unneccessary chances of plant shutdown. The work was completed on September 11,1995 (see below).
ECCS System Outage: Division I ADS Out of Service from 0900 09/11/95 to 1900 09/13/95 Duration: 58 hours6.712963e-4 days <br />0.0161 hours <br />9.589947e-5 weeks <br />2.2069e-5 months <br /> Planned Outage Outage Summary: Division I ADS was removed from service to replace ElIN021 A, flow indicating switch; RHR pump C002A discharge flow permissive for min-flow line, per EDP 27275 (replacement of D/P switch with new " blind" D/P switch).
4 - o Enclosure to NRC-96-0017 Page 13 DIVISION II AUTOMATIC DEPRESSURIZATION SYSTEM ECCS System Outage: Division II ADS Out of Service from 1424 06/02/95 to 2355 06/02/95 Duration: 9.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> Forced Outage Outage Summary: Division II ADS was removed from service to replace B21N694B, Drywell Pressure High Trip Unit, due to calibration problems while performing a surveillance. The trip unit, which also provides input to ADS, was replaced and ADS was returned to service.
HIGH PRESSURE CORE INJECTION ECCS System Outage: HPCI Out of Service from 150101/03/91 to 1200 01/07/95 Duration: 93.02 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> Forced Outage Outage Summaiv: HPCI was determined inoperable due to the inability to complete required surveillance within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> of adequate system pressure. The surveillance performed was unsatisfactory due to the test return line control valve failure to meet the acceptance criteria (E41-F011). HPCI was returned to operable status following the closure and de-energizing of the upstream valve (E4150-F008). During extended periods of time, HPCI was in standby and capable ofinjection.
ECCS System Outage: HPCI Out of Service from 0229 01/25/95 to 0120 01/27/95 Duration: 46.85 hours9.837963e-4 days <br />0.0236 hours <br />1.405423e-4 weeks <br />3.23425e-5 months <br /> Planned Outage .
Outage Summary: The HPCI system was removed from service to perform various CM and PM activities. Following completion of the activities and required surveillances, the HPCI system was returned to service.
Out of Service from 1230 02/20/95 M 1500 02/20/95 Duration: 2.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> Forced Outage l
Q_utage Summary: The HPCI system was denared inoperable due to the E41-F076 vacuum breaker being inoperable. Following evaluation, the failure of this valve was determined to not affect HPCI operability. HPCI was then returned to service. HPCI remained functional during this entire time period. l l
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NRCr96-0017 Page 14 ECCS System Outage: HPCI Out of Service from 0412 09/19/95 to 0448 09/21/95 Duration: 48.60 hours6.944444e-4 days <br />0.0167 hours <br />9.920635e-5 weeks <br />2.283e-5 months <br /> Planned Outage Outage Summary: The HPCI system was removed from service to perform various CM and PM activities. Following completion of the activities and required surveillances, the HPCI system was returned to service.
ECCS System Outage: HPCI Out of Service from 165812/18/95 to 223212/18/95 Duration: 5.57 hours6.597222e-4 days <br />0.0158 hours <br />9.424603e-5 weeks <br />2.16885e-5 months <br /> Planned Outage Outage Summary: The HPCI system was removed from service to tighten a loose !
deflector collar on the HPCI pump. HPCI was returned to service following the completion of this work. .
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