05000277/LER-1987-016, Corrected LER 87-016-00:on 870820,isolations Occurred When Four of Eight 4 Kv Emergency Transformers Momentarily de-energized.Caused by Inadequate Concern for Potential Electrical Hazard by Maint Craft Supervision
| ML20235J101 | |
| Person / Time | |
|---|---|
| Site: | Peach Bottom  |
| Issue date: | 09/28/1987 |
| From: | Alden W, Birely W PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC |
| To: | NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM) |
| References | |
| LER-87-016-02, LER-87-16-2, NUDOCS 8710010305 | |
| Download: ML20235J101 (9) | |
| Event date: | |
|---|---|
| Report date: | |
| 2772087016R00 - NRC Website | |
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Abstract: 2-87-16 On August 20, 1987 at 1305 hours0.0151 days <br />0.363 hours <br />0.00216 weeks <br />4.965525e-4 months <br /> containment isolations occurred on both units when the Unit 2 Startup Source tripped. Mobile crane movement on-site drew an arc from the startup source line i to the crane causing the trip. The affected emergency buses automatically switched to the Unit 3 Startup Source. The de-energization of relays during the switching caused isolations and !
pump trips on both units involving various systems. The Unit 2
'A' Reactor Protection System M-G set output breakers tripped on undervoltage following the emergency bus switching, causing a half-scram signal. No control rods moved. No one was injured, however the potential existed for severe personnel injury. There were no adverse safety consequences to the plant. Decay heat removal was restored within 30 minutes to both units. The decay j heat load on both units was very low because the reactors had ^
been shutdown for approximately five months. The root cause of this event was inadequate concern for a potential electrical hazard by maintenance craft supervision. The supervisors involved were counseled, all maintenance personnel and station group supervisors were notified of the event, additional training is planned for maintenance foremen to discuss the event, and all maintenance riggers will attend a one hour training class covering the OSHA regulations for mobile crane operation in the vicinity of electrical equipment.
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LICENSEE EVENT REPORT (LER) TEXT CONTINUATION ExPIA ES. 8/31/88 FACILITY NAME ti; DOCKET NUMfin m ggg gyggg n gg, pagg m
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0 l0 012 OF 0 l6 rwix- . w.m_ unc ,.,,, asa ,s u n Unit Conditions Prior to the Event Units 2 and 3 in the Cold Shutdown Condition Description of the Event:
On August 20, 1987 at 1305 hours0.0151 days <br />0.363 hours <br />0.00216 weeks <br />4.965525e-4 months <br />, containment isolations occurred on Unit 2 and Unit 3 when four of the eight 4 kv emergency transformers were momentarily de-energized. The event was-immediately identified by the actuation of control room alarms.
l The taole attached to this LER provides a summary of the safety recponses. A mobile crane was being operated behind the Diesel Generator Building (west side). As it was moved below the 220kV offsite source lines from the South Substation, it drew an are to the boom of the crane. This caused the outdoor SU-25 breaker to trip, isolating the station from the Unit 2 Startup Source. The supply breakers for the 4kv emergency buses that are normally fed l frcm the Unit 2 Startup Source (E-12, E-23, E-32, and E-43) i automatically fast-transferred to re-energize the buses from the Unit 3 Startup Source. The de-energized 13.8 kv unit auxiliary owitchgear were manually switched to the Unit 3 Startup Source one minute af.er the trip.
l Duting the fast transfer, Primary Containment Isolation System relays fed from the affected buses were de-energized, causing the isolations listed in the attached table. Also, the Unit 2 "B" l
and Unit 3 "C" Reridual Heat Removal (RHR) pumps (the only ones operating ia the shutdown cooling mode) and the Unit 2 "C" and Unit 3 "A" Reactor Water Clean-up (RWCU) pumps tripped because their section valves closed.
The Unit 2 A " Reactor Protection System (RPS) motor generator (M-G) set output breakers tripped on undervoltage. Tliis resulted in a half-scram signal, which does not result in control rod movement. Typically, a RPS M-G set trip does not occur during a fast transfer because the M-G sets are equipped with a high
- inertia flywheel to minimize coastdown during switching or raomentary de-energization, and the M-G sets are also equipped with a voltage regulator.
R_estoration of Equipment to Normal:
Unit 3 - The shutdown cooling isolations were reset and the "C"-
l RHR pump was returned to service by 1315 hours0.0152 days <br />0.365 hours <br />0.00217 weeks <br />5.003575e-4 months <br /> (within 10 minutes) in the shutdown cooling mode. The remaining oc .e .. u.. ,
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010 0l3 OF 0-l6 rnn w o, . <. .ae a nva ,, sawas nn isolations'were reset at 1327 hours0.0154 days <br />0.369 hours <br />0.00219 weeks <br />5.049235e-4 months <br /> and the "A" RWCU pump was returned to service at 2000 hours0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br />.
Unit 2 - The isolations'were; reset at 1320 hour0.0153 days <br />0.367 hours <br />0.00218 weeks <br />5.0226e-4 months <br />s-(within 15.
minutes) and the "C" RWCU, pump was returned.to service at 1342 hours0.0155 days <br />0.373 hours <br />0.00222 weeks <br />5.10631e-4 months <br />.. The "A" RPS M-G set output breaker. trip
.was reset,-the M-G set' output breakers were reclosed and the half-scram was reset at 1320 hours0.0153 days <br />0.367 hours <br />0.00218 weeks <br />5.0226e-4 months <br />..
In accordance with system operating procedure S.8.3.D.2,
" Unscheduled Tripping Of 2'Off-Site'Startup Source," the E-2 Diesel Generator was started at 1650 hours0.0191 days <br />0.458 hours <br />0.00273 weeks <br />6.27825e-4 months <br />, and aligned to,the.E-22 (Unit 2) and E-23 (Unit 3) emergency buses.' - This measure was taken to ensure reliable power to the RPS system on each unit and to permit operation of the Unit 2 "B" RHR Pump without l potentially overloading the Unit 3 Startup Source. The Unit 2 "B" RHR pump is powered from the E-22 bus. The Unit 2 "B" RHR pump was returned to service in the shutdown cooling mode at 1655 hours0.0192 days <br />0.46 hours <br />0.00274 weeks <br />6.297275e-4 months <br />.
The Unit 2 Startup Source line was not immediately re-energized l
so that the crane operation could commence safely. After completion of the crane operation, the Startup Source line was re-energized and the 13.8 kv unit auxiliary switchgear normally fed from the Unit 2 Startup' Source were realigned back to the Unit 2 Startup Source at 1802 hours0.0209 days <br />0.501 hours <br />0.00298 weeks <br />6.85661e-4 months <br />. The E-22 and.E-23 buses l were again aligned to the. Unit 3 Startup Source and the E-2 Diesel Generator was shutdown at 0950 hours0.011 days <br />0.264 hours <br />0.00157 weeks <br />3.61475e-4 months <br /> on August 21, 1987 l since the second off-site source had been restored to the unit auxiliary loads and there was no longer a concern for overloading the Unit 3 Startup Source. The 4.16.kv emergency buses remained on the Unit 3 Startup' Source for several days to allow previously planned work in the area of the Unit 2 Emergency Transformer (through which the emergency buses are.normally powered). On August 24, 1987 the E-12, E-32 and E-43 buses were realigned to the Unit 2 Startup Source at 0045,.0100 and 0115 hours0.00133 days <br />0.0319 hours <br />1.901455e-4 weeks <br />4.37575e-5 months <br />, respectively. These buses were powered from their alternate
-source for'approximately 3 1/2 days, with the exception of the E-22 bus which was powered by the E-2 Die.mel Generator for approximately 17 hours1.967593e-4 days <br />0.00472 hours <br />2.810847e-5 weeks <br />6.4685e-6 months <br /> during that perit The E-23 bus was realigned on August 28, 1987 at 2022 hout.s at the completion of a scheduled bus outage to permit relay calibrations. This bus was powered from its alternate source for approximately 7 days, it was powered by the E-2 Diesel Generator for approximately 17 hours1.967593e-4 days <br />0.00472 hours <br />2.810847e-5 weeks <br />6.4685e-6 months <br /> and its outage lasted approximately 14 hours1.62037e-4 days <br />0.00389 hours <br />2.314815e-5 weeks <br />5.327e-6 months <br />.
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EXPtRE4 8/31/95 PACILtYv Names et: DOCK 57 NutdSER (21 . (SR NUaABER ($) PA04138 j Peach' Bottom' Atomic Power Station "'" -
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waeson . e .m.wnncoemum4 m EIIS Codes:
The EIIS Codes for the systems' involved are CE,-RWCU, BO, RHR; -
SB, Main Steam; LD,, Instrument Nitrogen (Air) Supply; JM, Primary
' Containment Isolation; WD, Liquid Radwaste'(sumps); IG, Incore ~
Monitoring (TIP); SH, Condenser Vacuum (mechanical vacuum pump);
EK, Emergency Onsite Power Supply; FK, Switchyard'(substation) : l and JC, RPS-. JThe~EIIS Codes for the components involved are BKR, ~
I breaker; 30, bus; P, pump; RLY, relay; ISV, isolation valve; DG, 4 diesel generator; CBL, cable (power line) and MG, motor generator set. ,
1 l
Consequences of the Event:
No one was injured; however, the potential existed for' serious. )
injury to the crane operator and other personnel nearby'or who j could have been nearby'the crane.
There were no adverse plant safety consequences. Becau'se Shutdown Cooling and RWCU isolated, the normal means of decay heat removal were temporarily unavailable. However, on Unit 3 the Shutdown Cooling isolation was reset and RHR was returned.to service within ten minutes. On Unit 2, the RWCU system was returned to service within 37 minutes and RHR was returned to service within four hours. The decay heat load on-both units was very low because both the reactors had been shutdown for nearly five months and the Unit 2 core had been reloaded during that period. Reactor coolant temperature on both units did'not- l significantly increase.
If these isolations had occurred under the same circumstances but ;
the decay heat load and coolant temperature were greater, it 1 would have been more significant. However, even'if for some. 1 reason the isolations could not be reset, several alternative methods of coolant makeup and energy removal could have been used. As long as coolant inventory was maintained at normal' levels, adequate core cooling would be provided. Station-Procedure GP-12, " Core Cooling Procedure",. outlines alternative' ;
sources of makeup, such as Condensate Transfer from stay-full j lines, Core Spray from Condensate Storage Tank or Torus, Control '
Rod Drives from Condensate Storage Tank and RHR from Torus. The i procedure also outlines several methods of energy removal j including draining to the Torus through RHR piping. Each of the !
alternative methods of makeup / energy removal listed above were !
available during this event. There are several other methods une renan nu Is sv g_g
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LUCENSEE EVENT REPORT (LER) TEXT CONTINUATION -
'ERPIRt3 8/31/96 PLCabity Naasa (11 00CR8T NUteSER L3) Ltn NUMSER tot ,404 (31-Peach Bottom Atomic Power Station naa "2%Po n ",'#en Unit 2 o is l0 l0 l0 l 2l7l7 81 7 0l1l6 -
0l0 0l5 OF Q l6 TERT IM more aseee e mounet une eennenamnC Fenn Wei (111 outlined in GP-12.which, depending on reactor pressure:and- ,
equipment status,'could be'used'under other circumstances; -j If this event hadLoccurred while at power operation _with unit l auxiliary loads being powered by..the main generator the.
consequences would have been minimal. Shutdown Cooling,would not_ ;
have been.needed and the' isolation of RWCU, RHR Head Spray and- 3 Instrument Nitrogen, for a'brief period,-would not have'resulted )
in adverse consequences. Head Spray is'not required to mi.tigate ,
the consequences.of an accident. The Instrument Nitrogen Supply. j header to the Torus remains pressurized'without' makeup from'the l compressors for' considerably longer than this period of: .)
isolation. Only-one. loop-of Instrument; Nitrogen: Supply.to-the l
Drywell isolated; the 'A' loop was still available. Reactor coolant chemistry would not significantly be affected by.
Isolation of RWCU for several hours unless an abnormal chemistry-transient occurred. If this event'had occur _ red at' low power operation with unit auxiliary loads being supplied'by the offsite H sources, a low level scram may have occurred due to the. loss of .l power to operating condensate pump (s).
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Cause of the Event:
The root cause of this event was insufficient' concern for i electrical safety by job supervision and the consequential inadequate pre-job planning. The supervisors were-adequately j trained and qualified to assure a safe operation, but ;
unintentionally failed to do so. The supervisors and crane operator wrongly assumed that the power lines above the area were ;
de-energ! zed since the station was not at power. operation. There was no procedure being used. .i An 80-ton mobile crane was being used to move cargo units. !
containing tools and equipment to support'the annual diesel ,
generator inspections. The moving of the cargo units was under the direction.of maintenance craft supervision (company employed, I non-degreed, non-licensed foremen). Normally a smaller crane is
.used for'this job; however,.on this day the 80-ton crane was the only suitable-one available. When the boom of the crane was moved in close proximity to the 220 kv offsite power source line
- 220-08, an arc was drawn to the boom, and the fault traveled ;
through the crane to the ground. This caused the SU-25 breaker to trip open, initiating the event.
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LICENSEE EVENT REPORT (LER) TEXT CONTINUATION ExPint$ 8/31/e6 F ACILITY NAME na DOCKST NUMSER til Lem NUMesR les PA04 838 )
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0l0 o l6 or 0. l6 rextu,, ., w. .m. - :, anw,nn Corrective Actions:
1 On' September 3, 1987 at the conclusion of-the Maintenance Department's l' investigation, maintenance personnel at all work.
locations (fossil, hydro-electric and nuclear) were apprised of .)
this. incident and instructed in the importance of including i electrical safety in the job planning process. ,]
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l On September 11, 1987 the Peach Bottom Superintendent-Operations )
l issued a letter to station group supervisors discussing this j event and requiring that work supervision contact the Electrical 1 Supervisor (or Shift Supervision in his absence) prior to i beginning work under or near transmission lines.
The supervisors responsible for the crane activity were counseled regarding their inattention to a potential electrical hazard. It '
was stressed to these individuals tnat electrical equipment must be treated as though it is energized unless it is known to be blocked out cf service.
An additional training class will be held on October 12, 1987 for l maintenance foremen and assistant safety representatives ,
including a discussion of this event and stressing electrical. 1 safety. Assistant safety representatives are individuals within the Maintenance Department assigned to a station to support i' maintenance activities from a safety standpoint. All maintenance riggers will attend a one hour training class in the near future to educate them in the Occupational Safety & Health ;
Administration (OSHA) regulations for mobile crane operation in j the vicinity of electrical equipment. Additionally, the electrical hazards portion of formal craft training for riggers will be expanded.
The Engineering & Research Department will be requested to investigate the capability of the RPS M-G sets to avoid a breaker i trip during a 4.16 kv bus fast transfer and determine if corrective actions are warranted to reduce the possibility of unnecessary scrams, i
i Previous Similar Occurrences:
1 i
There have been other events involving loss of a startup source, but none were caused by a similar incident.
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PHILADELPHIA ELECTRIC COMPANY 2301 M ARKET STREET P.O. BOX 8699 PHILADELPHIA A, PA.19101 (215)841 4000 September 28, 1987 Docket No. 50-277 50-278 Document Control Desk i U.S. Nuclear Regulatory Commission Washington, DC 20555
SUBJECT:
Licensee Event Report Peach Bottom Atomic Power Station - Units 2 and 3 This LER was originally submitted by letter dated September 21, 1997 with the wrong docket number in coded field 2 on the first page of the LER. We regret this error and have enclosed a corrected copy.
Reference:
Docket Nos. 50-277 and 50-278 Report Number: 2-87-16 Revision Number: 00 Event Date: August 20, 1987 Report Date: September 21, 1987 Facility: Peach Bottom Atomic Power Station RD 1, Box 208, Delta, PA 17314 This LER is being submitted pursuant to the requirements of 10 CFR 50.73(a)(2)(iv).
Very ruly oure W. M. Alden Engineer-In-Charge Licensing Section cc: W. T. Russell, Administrator, Region I, USNRC T. P. Johnson, NRC Resident Inspector W
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