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Category:LICENSEE EVENT REPORT (SEE ALSO AO
[Table view] Category:RO)
[Table view] Category:TEXT-SAFETY REPORT
MONTHYEAR05000278/LER-1999-005-03, :on 990920,uplanned Esfas During Planned Mod Activitives in Main CR Were Noted.Caused by Inattention to Detail by Individuals Performing Work.All CR Mods Were Ceased to Allow Review of Mod Work Packages.With1999-10-20020 October 1999
- on 990920,uplanned Esfas During Planned Mod Activitives in Main CR Were Noted.Caused by Inattention to Detail by Individuals Performing Work.All CR Mods Were Ceased to Allow Review of Mod Work Packages.With
ML20217K9931999-10-14014 October 1999 Safety Evaluation Supporting Amend 234 to License DPR-56 ML20217B4331999-10-0505 October 1999 Safety Evaluation Supporting Amend 233 to License DPR-56 05000278/LER-1999-004-03, :on 990901,3A RPS Bus Was Inadvertently Deenergized,During Planned Mod Activities on Main CR Panel. Caused by Electrician Failing to Self Check Work.All CR Work Was Ceased Immediately & Shutdown Meeting Held1999-10-0101 October 1999
- on 990901,3A RPS Bus Was Inadvertently Deenergized,During Planned Mod Activities on Main CR Panel. Caused by Electrician Failing to Self Check Work.All CR Work Was Ceased Immediately & Shutdown Meeting Held
ML20217G3541999-09-30030 September 1999 Monthly Operating Repts for Sept 1999 for Pbaps,Units 2 & 3. with ML20216H7091999-09-24024 September 1999 Safety Evaluation Supporting Amends 229 & 232 to Licenses DPR-44 & DPR-56,respectively ML15112A7681999-09-20020 September 1999 SER Accepting Revision 25 of Pump & Valve Inservice Testing Program,Third 10-year Interval for Plant,Units 1,2 & 3 ML20212D1281999-09-17017 September 1999 Safety Evaluation Supporting Proposed Alternatives CRR-03, 05,08,09,10 & 11 05000278/LER-1999-003-03, :on 990814,HPCIS Was Declared Inoperable Due to Erratic Behavior Resulting in Loss of Single High Train Safety Sys.Caused by Weakness in Procedural Guidance. Readjusted Hydraulic Governor Needle Valve.With1999-09-13013 September 1999
- on 990814,HPCIS Was Declared Inoperable Due to Erratic Behavior Resulting in Loss of Single High Train Safety Sys.Caused by Weakness in Procedural Guidance. Readjusted Hydraulic Governor Needle Valve.With
ML20212A5871999-08-31031 August 1999 Monthly Operating Repts for Aug 1999 for Peach Bottom,Units 2 & 3.With ML20211D5501999-08-23023 August 1999 Safety Evaluation Supporting Amends 228 & 231 to Licenses DPR-44 & DPR-56,respectively ML20212H6311999-08-19019 August 1999 Rev 2 to PECO-COLR-P2C13, COLR for Pbaps,Unit 2,Reload 12 Cycle 13 ML20210N7641999-07-31031 July 1999 Monthly Operating Repts for Jul 1999 for PBAPS Units 2 & 3. with 05000277/LER-1999-005-01, :on 990616,failure to Maintain Provisions of FP Program Occurred.Caused by Less than Adequate Engineering Rigor in Both Development & Review Analysis.Fire Watch Immediately Established.With1999-07-16016 July 1999
- on 990616,failure to Maintain Provisions of FP Program Occurred.Caused by Less than Adequate Engineering Rigor in Both Development & Review Analysis.Fire Watch Immediately Established.With
ML20209H1121999-06-30030 June 1999 Monthly Operating Repts for June 1999 for Pbaps,Units 2 & 3. with ML20195H8841999-05-31031 May 1999 Monthly Operating Repts for May 1999 for Pbaps,Units 2 & 3. with 05000278/LER-1999-002-02, :on 990406,safeguard Sys to Unrelated Door Was Inadvertently Disabled by Security Alarm Station Operator. Caused by Noncompliance with Procedures & Less than Adequate Shift Turnover.Briefed Personnel on Event.With1999-05-0606 May 1999
- on 990406,safeguard Sys to Unrelated Door Was Inadvertently Disabled by Security Alarm Station Operator. Caused by Noncompliance with Procedures & Less than Adequate Shift Turnover.Briefed Personnel on Event.With
ML20206N1661999-04-30030 April 1999 Monthly Operating Repts for Apr 1999 for Pbaps,Units 2 & 3. with ML20206A2921999-04-20020 April 1999 Safety Evaluation Concluding That Proposed Changes to EALs for PBAPS Are Consistent with Guidance in NUMARC/NESP-007 & Identified Deviations Meet Requirements of 10CFR50.47(b)(4) & App E to 10CFR50 05000278/LER-1999-001-03, :on 990312,ESF Actuation of Rcics Occurred Due to High Steam Flow Signal During Sys Restoration.Temporary Change to Restoration Procedure Was Initiated to Open RCIC Outboard Steam Isolation Valve in Smaller Increments1999-04-0808 April 1999
- on 990312,ESF Actuation of Rcics Occurred Due to High Steam Flow Signal During Sys Restoration.Temporary Change to Restoration Procedure Was Initiated to Open RCIC Outboard Steam Isolation Valve in Smaller Increments
ML20205K7411999-04-0707 April 1999 Safety Evaluation Supporting Amends 227 & 230 to Licenses DPR-44 & DPR-56,respectively ML20205P5851999-03-31031 March 1999 Monthly Operating Repts for Mar 1999 for Peach Bottom Units 2 & 3.With ML20207G9971999-02-28028 February 1999 Monthly Operating Repts for Feb 1999 for Peach Bottom Units 2 & 3.With 05000278/LER-1998-009-01, :on 981227,unplanned Esfa Were Noted.Caused by Transformer Insulator Failure.Replaced Failed Insulator. with1999-01-20020 January 1999
- on 981227,unplanned Esfa Were Noted.Caused by Transformer Insulator Failure.Replaced Failed Insulator. with
ML20199E3471998-12-31031 December 1998 Monthly Operating Repts for Dec 1998 for Peach Bottom,Units 1 & 2.With ML20206P1651998-12-31031 December 1998 Fire Protection for Operating Nuclear Power Plants, Section Iii.F, Automatic Fire Detection ML20205K0381998-12-31031 December 1998 PECO Energy 1998 Annual Rept. with ML20206D3651998-12-31031 December 1998 1998 PBAPS Annual 10CFR50.59 & Commitment Rev Rept. with ML20206D3591998-12-31031 December 1998 1998 PBAPS Annual 10CFR72.48 Rept. with 05000277/LER-1998-008-01, :on 981130,circuit Breaker SU-25 Tripped.Caused by Less than Adequate Procedural Guidance.Operators Verified Sys Integrity & Successfully Returned Sys to Svc.With1998-12-30030 December 1998
- on 981130,circuit Breaker SU-25 Tripped.Caused by Less than Adequate Procedural Guidance.Operators Verified Sys Integrity & Successfully Returned Sys to Svc.With
05000277/LER-1998-007-02, :on 981107,failure to Meet TS & Associated LCO Requirments of Absolute Difference in APRM & Calculated Power of Less than 2% Was Noted.Caused by Substitute Valves Being Used.Removed Substitute Valves.With1998-12-0404 December 1998
- on 981107,failure to Meet TS & Associated LCO Requirments of Absolute Difference in APRM & Calculated Power of Less than 2% Was Noted.Caused by Substitute Valves Being Used.Removed Substitute Valves.With
ML20196G7021998-12-0202 December 1998 SER Authorizing Proposed Alternative to Delay Exam of Reactor Pressure Vessel Shell Circumferential Welds by Two Operating Cycles ML20196E8261998-11-30030 November 1998 Response to NRC RAI Re Reactor Pressure Vessel Structural Integrity at Peach Bottom Units 2 & 3 ML20198B8591998-11-30030 November 1998 Monthly Operating Repts for Nov 1998 for Pbaps,Units 2 & 3. with 05000278/LER-1998-005-03, :on 981025,inadvertent Unit 3 Electrical Bus E33 Trip (Esfa) During Performance of Unit 2 Electrical Bus E32 Surveillance Test Was Noted.Caused by Personnel Error. Sp S12M-54-E32-XXF4 Was Completed.With1998-11-20020 November 1998
- on 981025,inadvertent Unit 3 Electrical Bus E33 Trip (Esfa) During Performance of Unit 2 Electrical Bus E32 Surveillance Test Was Noted.Caused by Personnel Error. Sp S12M-54-E32-XXF4 Was Completed.With
ML20206R2571998-11-17017 November 1998 PBAPS Graded Exercise Scenario Manual (Sections 1.0 - 5.0) Emergency Preparedness 981117 Scenario P84 ML20198C6751998-11-0505 November 1998 Rev 3 to COLR for PBAPS Unit 3,Reload 11,Cycle 12 ML20195E5341998-10-31031 October 1998 Monthly Operating Repts for Oct 1998 for Pbaps,Units 2 & 3. with ML20155C6071998-10-26026 October 1998 Safety Evaluation Supporting Amend 226 to License DPR-44 ML20155C1681998-10-22022 October 1998 Safety Evaluation Accepting Proposed Alternative Plan for Exam of Reactor Pressure Vessel Shell Longitudinal Welds ML20155H7721998-10-12012 October 1998 Rev 1 to COLR for Peach Bottom Atomic Power Station Unit 2, Reload 12,Cycle 13 05000277/LER-1998-006-02, :on 980915,automatic RWCU Isolation Occurred While Placing RWCU Sys in Svc.Caused by Unexpected Surge of Water.Procedure Change Was Initiated to Open MO-2-12-74 & RWCU Sys Was Successfully Returned to Svc.With1998-10-0909 October 1998
- on 980915,automatic RWCU Isolation Occurred While Placing RWCU Sys in Svc.Caused by Unexpected Surge of Water.Procedure Change Was Initiated to Open MO-2-12-74 & RWCU Sys Was Successfully Returned to Svc.With
ML20154J2401998-10-0505 October 1998 Safety Evaluation Supporting Amends 224 & 228 to Licenses DPR-44 & DPR-56,respectively ML20154H4771998-10-0505 October 1998 Safety Evaluation Supporting Amends 225 & 229 to Licenses DPR-44 & DPR-56,respectively ML20154G6821998-10-0101 October 1998 SER Related to Request for Relief 01A-VRR-1 Re Inservice Testing of Automatic Depressurization Sys Safety Relief Valves at Peach Bottom Atomic Power Station,Units 2 & 3 ML20154G6631998-10-0101 October 1998 Safety Evaluation Supporting Amends 223 & 227 to Licenses DPR-44 & DPR-56,respectively ML20154H5541998-09-30030 September 1998 Monthly Operating Repts for Sept 1998 for Pbaps,Units 2 & 3. with 05000278/LER-1998-004-03, :on 980820,automatic RWCU Isolation Occurred While Placing B RWCU Sys Demineralizer in Svc.Caused by less-than-adequate Control of Equipment.Isolated B Demineralizer & Returned RWCU Sys to Svc1998-09-18018 September 1998
- on 980820,automatic RWCU Isolation Occurred While Placing B RWCU Sys Demineralizer in Svc.Caused by less-than-adequate Control of Equipment.Isolated B Demineralizer & Returned RWCU Sys to Svc
05000277/LER-1998-005-02, :on 980824,noted Failure to Meet TS Actions for Suppression chamber-to-drywell Vacuum Breaker Not Being Fully Seated.Caused by Personnel Failing to Take All TS Required Actions.Temporary Procedure Changes Were Made1998-09-18018 September 1998
- on 980824,noted Failure to Meet TS Actions for Suppression chamber-to-drywell Vacuum Breaker Not Being Fully Seated.Caused by Personnel Failing to Take All TS Required Actions.Temporary Procedure Changes Were Made
ML20153B9651998-09-14014 September 1998 Safety Evaluation Supporting Amend 9 to License DPR-12 1999-09-30
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Text
s CCN-90-14094' PHILADELPHIA ELECTRIC COMPANY
. PEACll BUTIUM A1DMIC POWER STATION R. D.1, Ikix 208 'I Delta, Pennsylvania 17314
, . PEEN banUtl. tnt POWER OF R ECEL11Nca (717)456-7014 ~
. May 16,'1990 Docket No.' 50-277
. Document Control Desk U. S.. Nuclear Regulatory Commission Washington, DC 20555 ;
SUBJECT:
Licensee: Event-Report .
i Peach Bottom Atomic-Power-Station - Unit 2 1
This LER' concerns a violation of Technical Specifications because of a. ~
missed-surveillance'due to a personnel error' i
Reference:
Docket No. 50-277 Report Number: 2-90-008 Revision Number: 00 Event Date:- 04/14/90 Discovery Date: 4/17/90 Report Dath 05/16/90 1
Facilit a {
Peach Bottom Atomic Power Station '
RD 1. Box 208,-Delta,-PA 17314 r This LER is being submitted pursuant to the-requirements of 10 CFR' '
50.73(a)(2)(1)(B). i
. Since' rely.
4
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7 H~}
1 ant Manager cc: J. J. Lyash, USNRC Senior Resident inspector '!
T. T. Martin, USNRC, Region I '
i 900522'0107 900516 PDR S
ADOCK 05000277 PDC
[p J 9
1
NRC Fe,in 364
- U S. NUCLLN LE LULATORT COMMistiON 4.PT00VED DMS NO 3156 0104
'"a'*''*
LICENSEE EVENT REPORT (LER)
DOCKEY NUMSIR (28 P AGI '3' F ACILITY NAME III Peach Bottom Atomic Power Station - Unit 2 0 l5 l 0 [0 l 0121717 1 lOFl 0 l3
'" Technical Specification Violation Caused By Personnel Error Results In Missed Low Preneure core coolino System S'Irveillances EVENT DAf t (St LE R NUMSER (G) REPORT DATE til OTHER F ACILITIEE INVOLVED (8)
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NAME TE LEPHONE NUMBE R A84t A CQDE A. A. Fulvio, Regulatory Encjineer 711 l7 4l 5 l 6 l -l71 of 114 COMPLif t ONE LINE FOR E ACH COMPONE AT F AILURE Of 8CRitfD IN THl3 REPORT 1131 COMPONENT "Yo %,R0" * "'s' ' CAust sy STEM COvPONENT a OnT Astt CAUSE Sv 81 E M X'g*C M(NNC pq $
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On 4/17/90, 0850 hours0.00984 days <br />0.236 hours <br />0.00141 weeks <br />3.23425e-4 months <br />. Operations personnel discovered that testing of Low Pressure Coolant injection (LPCI) pumps and Core Spray subsystems were not performed as I required by Technical Specification 4.5.A.4. This testing was required because the D LPCI pump was inoperable.when the Unit 2 Reactor mode was changed from shutdown to i
' refuel on 4/14/90, 2218 hours0.0257 days <br />0.616 hours <br />0.00367 weeks <br />8.43949e-4 months <br />. On 4/17/90, 0950 hours0.011 days <br />0.264 hours <br />0.00157 weeks <br />3.61475e-4 months <br />, a shift permit applied to the !
D LPCI pump was cleared and the pump declared operable. The cause of this event was personnel error in not recognizing the inoperable status of the D LPCI pump when ;
changing the reactor mode from shutdown to refuel. The D LPCI pump was declared I inoperable due to a unique design condition on Unit 2. No actual safety consequences ..,
l occurred as a result of this event. Appropriate Operations personnel will be ;
I informed of this event. Procedural controls that involve declaring the D LPCI pump l
inoperable for the unique design condition requirement will be revised to require the
! mode switch be included within the blocking permit for the D LPCI pump. There were no previous similar LERs.
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- 4 LICENSEE EVENT REPORT (LER) TEXT CONTINUATION emoveo ove no siso-om ,
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- EXPtRES
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.Esach Bottom Atomic Power Station' va^a :
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Requirements'for the: Report'
\-
y This report =is required per'10 CFR 50.73(a)(2)(1)(B) because of?a condition.
prohibited;by' Technical Specifications-(Tech Specs).
l 'Un'it' Status at Time of Discovery Unit 2 was-in the Refuel Mode '
The-20 Residual-Heat Removal (RHR) (EIIS:B0) pump (EIIS:P) was tagged-out as part of a shift permit (i e.t pump declared. inoperable)-
lThe.2C-RHRpumpwas11nserviceintheshutdowncooling(SDC)modeof_RHR.'
l Description of Event- ,
.On 4/17/90, 0850 hours0.00984 days <br />0.236 hours <br />0.00141 weeks <br />3.23425e-4 months <br />, Operations personnel discovered that testing of Low Pressure.
- Coolant Injection (LPCI) pumps.and Core Spray (EIIS:BM)csubsystens.were _not performed as required by Tech Spec 4.5.A.4. : Tech Spec 4.5.A.4. requires that.with one LPCI pump
~ inoperable, the remaining 1LPCI pumps an6 associated flowpaths and both . core spray-subsystems be demonstrated to be operable by' testing within 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />sTand at;least once every 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />'.thereaf ter until. the LPCI pump is restored'to an operable status.-
-If this! testing cannot be performed, Tech. Spec 3.5 A.7-requires the initiation'of an orderly shutdown and to be:in the cold shutdown condition within 48' hours' This .
testing was required because the 20 RHR pump (.i.e..lLPCI' pump) was inoperable when ,
the Unit 2' Reactor (EIIS:RCT) mode-was changed from shutdown to refuel on'4/14/90, 2218. hours. Previous to this time, Unit 2 had been in the cold shutdown coridition and LPCI was not required to be operable. The 20 RHR pump was inoperable.because of' a unique design condition on Unit 2Lwhich necessitates the.2D RHR pump to be declared:
inoperable and blocked from service whenever the-2A,-2B or 2C RHR pumps are used;for SDC. .This unique design condition exists because the minimum flow valve'(Ells:V) for-the 2D RHR pump is.a normally open. valve to meet the requirements ofL10 CFR 50 ~
Appendix:R. As a result, to prevent inadvertent Reactor drain down events during SDC' ?
- operation, the minimum flow valveihas to be closed-(thereby rendering the 20.RHR pump.
1 On 4/17/90, 0950 hours0.011 days <br />0.264 hours <br />0.00157 weeks <br />3.61475e-4 months <br />, a shift' permit applied to the 20 RHR pump was-clearedLand the pump was declared operable. The 2C RHR pump was secured from the SDC mode of operation prior to clearing the permit on the 2D RHR: pump.
Analysis of the Event No actua'l safety consequences occurred as a result of this event.
Subsequent testing proved that the other RHR pumps and the Core Spray System were 'i l operable 'during the time the mode switch was placed in refuel until the 20 RHR pump -
was declared operable. Because the other RHR pumps and' Core Spray System were ;
operable through this time, this equipment would be available to provide low pressure core cooling if,needed.
I -
5
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1 NRC Feem 3944 U.S. NUCLt A1781ULATORY COMMISSION )
- LICENSEE EVENT REPORT (LER) TEXT CONTINUATl3N - Aeovto ous No sino-oio. J EXPIRFS 8/31/W F ACl487Y NAME (1) . DOCKli NUMBER (2) LER NUMeth (Si P A04 (3) -
P2ach Bottom Atomic Power station "^" :
"0%W' '
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' Unit 21 OF 0 l5 l0 l0 l0 l 2]7 l 7 91 o o lo l8 -
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Th. ', W_ Is m . mc w.m.nm <
Additionally, lech Specs allow' 7 days of continued reactor operation with 1 RHR. pump inoperable. Ine 2D RHR pump was made-operable within this 7 day: Limiting Condition i
for Operation (i.e ~2D RHR pump was inoperable in the refuel mode for approximately 2 1 1/2' days).
~
Cause of the Event
- The cause of this event was personnel error in not recognizing S e-inoperable, status a of the 20 RHR pump when-changing from the shutdown mode to the refuel mode of i operation. General-Plant' Procedure GP-11C, " Reactor Protection System Refuel Mode
. Operation'.' requires that prior to changing the plant mode tearefuel, LPCI be operable.
as specified in Tech Specs.. This step was signed:off-as satisfactory by Operations !
personnel (Utility, Licensed) '
Corrective Actions .J 1
The 20 RHR pump was returned to an operable status on.4/17/90, 0950 hours0.011 days <br />0.264 hours <br />0.00157 weeks <br />3.61475e-4 months <br />, thereby.
complyingw'ith Tech Specs. Appropriate Operations personnel will be informed of this event. To provide an additional. barrier to prevent' recurrence of this event, .
procedural controls that involve declaring the 20 RHR pump inoperable when using the L2A, 28 or 2C RHR pumps for SDC will be revised to require the mode switch (EIIS:JS) be included within the blocking permit for the 20 pump.
Previous Similar Events ,
There have been no previous similar LERs caused by personnel error involving the non-- '
performance of surveillances that are required as a result of a mode switch change. '
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