ML20029B642

LER 91-004-00:on 910210,potential Transformer Drawer Opened at Bottom of Switchgear Bus 1A3 Auxiliary Cubicle,Causing Load Shed Signal & Reactor Trip.Caused by Personnel Error. Labels Attached to Switchgear bus.W/910312 Ltr

ML20029B642
Person / Time
Site:	Comanche Peak
Issue date:	03/12/1991
From:	William Cahill, Hope T TEXAS UTILITIES ELECTRIC CO. (TU ELECTRIC)
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
LER-91-004, LER-91-4, TXX-91077, NUDOCS 9103130246
Download: ML20029B642 (8)

Text

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E Log # TXX-91077

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g File i 10200 910.4 7UELECTRIC Ref. # 50.73(a)(2)(iv)

March 12, 1991 2,DOIU,r,,,am, U. S. Nuclear Regulatory Comission Attn: Document Control Desk Washington, D. C. 20555

SUBJECT:

COMANCHE PEAK STEAM ELECTRIC STATION (CPSEd)

DOCKET NO. 50-445 MANUAL OR ALTTOMATIC REACTOR PROTECTION SYSTEM ACTUATION LICENSEE EVENT REPORT 91-004-00 Gentlemen:

Enclosed is Licensee Event Report 91-004-00 for Comanche Peak Steam Electric Station Unit 1 " Reactor Trip Caused by Personnel Error and Insufficient Labeling of Sensitive Equipment."

Sincerely, d A.

William J. Cahill, Jr. '

JAA/daj e - Mr. R. )). Martin, Region IV Resident Inspectors, CPSES (3)

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On February 10,1991, Comanche Peak Steam Electric Station Unit 1 was in Mode 1, Power Operation, at 80 percent reactor power. Whi!e conducting on-the-job training, an Auxiliary Operator opened the 6.9 KV switchgear bus "1 A3" auxiliary cabinet potential transformer drawer resulting in a load shed signal for the bus loads, which included Reactor Coolant Pump 103, followed by a reactor trip due to a loss of flow in reactor coolant loop 3. The root causes were determined to be inadequate labeling on the potential transformer drawer and poor judgement in opening a drawer without knowing the effect it would have on the plant.

Corrective actions include labeling the potential transformer drawers, identifying equipment or panels t at have interlocks or special sensitivity for equipment actuation, and issuing guidance to personnel on opening switchgear doors.

. .'5nclosure to TXX 91077 NRC FORM 3a0A , u.6. Nucd AR ne otAA roRY COWW888CN APPROWD OMt NO. 31t00104 ESTIMATED DURDEN PER RES SE O COMPLY WTH TH18 PFonMATioN LICENSEE EVENT REPORT (LER)- MC'%,g','; ,'f"geg,",*,S

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l. DESCRIPTION.0F_THE REPO'1 TABLE EVENT A. REPORTABLE EVENT CLASSIFICATION ,

Any event or condition that resulted in manual or automatic actuation of any Engineered Safety Feature (ESF), including the Reactor Protection System.

B. PLANT OPERATING COND!TIONS PRIOR TO THE EVENT On February 10,1991, just prior to the event, Comanche Peak Steam Electric Station (CPSES) Unit 1 was in Mode 1, Power Operation, with reactor power at 80 percent.

C. STATUS OF STRUCTURES, SYSTEMS, OR COMPONENTS THAT WERE INOPERABLE AT THE START OF THE EVENT AND THAT CONTRIBUTED TO THE EVENT There were no inoperable structures, systems or components that contributed to the event.

D. NARRATIVE

SUMMARY

OF THE EVENT, INCLUDING DATES AND APPROXIMATE TIMES At 0530 on February 10,1991, an Auxiliary Operator (AO) (utility, non licensed) was conducting on the-job training with a trainee (utility, non licensed). The AO and trainee were trying to locate the reset switches for the Circulating Water Pumps 103 and 1-04 (Ells:(KE)(P)), While looking for the reset switch for Circulating Water -

l Pump 103, the AO opened the potential transformer drawer (Ells:(EA)(FD)) at the L bottom of the switchgear bus "1 A3" auxiliary cubicle (Ells:(EA)(BU)). The AO did not i realize that the drawer was position interlocked with fuses for the bus undervoltage load shed circuitry (Ells:(EA)(RLY 94)). There was no warning or caution label on

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the drawer. Opening the drawer resulted in a load shed signal for the bus loads,

which included Reactor Coolant Pump 103 (Ells

(AB)(P)), followed by a reactor trip due to loss of flow in reactor coolant loop 3. Also lost when the drawer was opened were Condensate Pump 1-01 (Ells:(SG)(P)), Turb!ne Plant Cooling Water Pump 1-01 (Ells:(KB)(P)) and Service Air Compressor 1-01 (Ells:(LF)(CMP)). Circulating Water Pump 1-03 (Ells:(KE)(P)) would also have tripped if it had been in service.

Folbwing the trip, Control Room personnel responded in accordance with J

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. Enclosure to TXX 91077 NRC FORM 366A U.$sNUGli AR R4 GMIORY COMW$$ ION gpFROWD OMS NO 3166 0104

($1edATED BURDEN PER Rt8 NSE County wifw THis p*oRuAttoN LICENSEE EVENT REPORT (LER) Q'ty%g','; y','","*cl[o"*70 E'uEi TEXT CONTINUATION 'g"f7g,y,@,',ll$^70 f,7,"l$3cQ7,Q '

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0IO OI3 OF 017 emergency operating procedures. Plant systems responded as expected, except for the situations discussed below. The plant was stabilized in Mode 3, Hot Standby, at 0632. At 0720 the NRC was notified of the event via the Emer0ency Notification System in accordance with 10CFR50.72.

Following the reactor trip, three momentary Steam Generator level signals (Ells:(JE)(RLY 83)) were received. LO LO Level signals were received for Steam Generators 2 and 3 at approximately 1.69 seconds and a HI HI Level signal was received for Steam Generator 2 at approximately 2.18 seconds after the trip. The Turbine Driven Auxiliary Feedwater Pump (Ells:(BA)(P)) did not start and Steam Generator Blowdown and Sampling (Ells:(KN)(RLY 56)) did not isolate. A Main Feedwater isolation signal was generated as a result of the Hi HI Level signal with Main Feedwater Pump 1 A tripping (Ells:(SJ)(P)). Main Feedwater Pump 1B did not trip. The steam dump turbine load rejection interlock (Ells:(JI)(IEL)) would not reset.

i i The momentary Steam Generator level signals had been previously identified in i another event as the resu!t of steam pressure oscillations following a turbine trip.

l Two channels on all four Stearn Generators are susceptible to low level spikes and two channels on Steam Generators 1 and 2 are susceptible to high level spikes. The l

level transmitters share the same impulse lines as the steam flow transmitters and as l a result, the narrow range level transmitters are susceptible to level spikes whenever l rapid changes in steam flow occur Due to the short duration of the spikes, less than 0.2 seconds, the signals were not present long enough to start Turbine Driven Auxiliary Feedwater Pump, isolate Steam Generator Blowdown and Sampling, or trip l Main Feedwater Pump 18. During the post trip recovery the slave relays for these signals were tested and showed that both trains were satisfactory and would respond as designed to actual Steam Generator LO LO or HI HI signals. A bad Lead / Lag card was found in the Steam Dump Turbine Load Rejection circuitry which l was replaced and calibrated with all alarms functioning and the interlock resetting correctly.

E. THE METHOD OF DISCOVERY OF EACH COMPONENT OR SYSTEM FAILURE OR PROCEDURAL ERROR The reactor trip was annunciated by numerous alarms in the Control Room. The immediate cause of the trip was reported by the AO.

Ehclosure to TXX 91077 hRC FORM 3eAA U,8. NCE AR REOULATORT COWWIS$ON APPROWDOWO NO SMOW E6TIMATED BURDEN PER RES SE T COMPLY WitH THS IN50RM ATION LICENSEE EVENT REPORT (LER) *g,'"","S***l; gl"fcj,o"l,*,*o Og',"'ff,%,

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0IO Ol4lOF 017 n.u, _ . - ~ w n w ..On ll. COMPONENT OR SYSTEM F AILURES A. FAILURE MODE, MECHANISM, AND EFFECT OF EACH FAILED -

COMPONENT No failed components contributed to this event.

B. CAUSE OF EACH COMPONENT OR SYSTEM FAILURE No failed components contributed to this event.

C. SYSTEMS OR SECONDARY FUNCTIONS THAT WERE AFFECTED BY FAILURE OF COMPONENTS WITH MULTIPLE FUNCTIONS No failed components contributed to this event.

D. FAILED COMPONENT INFORMATION No failed components contributed to this event.

Ill. ATLA 1.YSIS OF THE EVENT A. SAFETY SYSTEM RESPONSES THAT OCCURRED The Reactor Protection System (Ells:(JC)) and Auxiliary Feedwater System (Ells:(BA)) actuated during the event; all associated components within these systems functioned as designed. The Steam Generator level signals received immediately after the trip were not valid signals, as explained above, and did not affect the safe shutdown or recovery of the plant.

B. DURATION OF SAFETY SYSTEM TRAIN INOPERABILITY No safety system trains were inoperable as a result of this event.

I

. Enclosure to TXX 91077 NRC FORW 3e6A U 8. WCd AA MGt.A.AIORY COMW.610N yp;pygg gyg gg g3gyggg4 EXPLRES 6%92 atinAATIO DUROEN PER RES80NSE TO COWPLY WlIH THtB MORMATION LICENSEE EVENT REPORT (LER) 7,y,],"*,'; ,'f 7cg%" C ",o"[',RE R nc ,

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0l0 015 OF 017 u.i s - . - ~-A ., o n C. SAFETY CONSEQUENCES AND IMPLICATIONS OF THE EVENT The reactor trip was the resuu of the loss of Roactor Coolant Pump 1-03. The reactor trip event is discussed in Section 15.3.1 of the CPSES Final Safety Analysis Report under " Partial Loss of Forced Reactor Coolant Flow." The analysis uses conservative assumptions to demonstrate that Departure from Nucleate Boiling Ratio (DNBR) will never decrease below the limiting value of 1.30 during the event. The event of February 10,1991, occurred at 80 percent reactor power, and all protective functions responded as designed. The event is completely bounded by the FSAR accident analysis which assumes an initial power level of 102 percent and makes conservative assumptions which reduce the capability of safety systems to mitigate the consequences of the transient. The event of February 10 did not adversely affect the safe operation of CPSES Unit 1 or the health and safety of the public.

IV. C AUSE OF THE EVENT A. ROOT CAUSE

1. The 6.9 KV switchgear bus "1 A3" auxi'iary cabinet potential transformer drawer did not have an adequate label. This drawer contains a vital position interlock with bus undervoltage load shedding circuitry, but was not labeled as such.

2. The AO used poor judgement in opening a drawer without knowing the effect it would have on the plant. Operators are expected to know the result of any action they take in the plant.

B. GENERIC CONSIDERATIONS Similar equipment needs to be identified and labeled to prevent similar events from occurring.

1 L _ _ _ . _ _ _ _ _ _ _ _ _ _ ___ _______ _ __________ _ _ _ _ ______ _ ___ _ ____ _ _______ ______ _

Ehclosure to TXX 91077 NT4 FORM 3e$4 U.S. NUCd AR R6GULATOM COMuem APPROWD OMB NO. 3@0104 EXPIRE S. 410Dif l ESTIMATE 0 8URDEN PER RESONSE TO COMPLY WTTH TH41NFORMAf 0N i

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00, 20666. AND TO THC PAPERWORM REDUCTION PROJECT (3%0104 0FFICE Or MANAGEMENT AND SUDGET WASHiNOTON, DC. 20603. _j r.a., w w Nn m a n un a, wm yen j.g Egn s.mni ai * *p COMANCHE PEAK - UNIT 1 OI5101010141415 911 - 0IOl4 - 0i0 016 OF 0l7 u o - - ~ ~w- =A .ni n V. CORRECTIVE ACTIONS CORRECBVE ACBONS TO PREVENT RECURRENCE A. ROOT CAUSE i

1. The 6.9 KV switchgear bus "1 A3" auxiliary cabinet potential Ensformer drawer did not heve an adequate label.

CORRECBVE ACBON Labels have been affixed to the 6.9 KV switchgear bus "1 A1", "1 A2", "1 A3", and "1 A4" auxiliary cabinet potential transformer drawers.

2. The AO used poorjudgement in opening a drawer without knowing the effect it would have on the plant. Operators are expected to know the result of any action they take in tha plant.

CORRECTIVE ACTION Operations Management has issued a memorandum that provides additional guidance on opening switchgear doors. This memorandum has been included in the Control Room " Lessons Learned" notebook. B. GENERIC CONSIDERATIONS Similar equipment needs to be identified and labeled to prevent similar events from occurring. CORRECTIVE ACTION A survey will be conducted to identify equipment that could cause similar

                              - problems. The equipment will be labeled appropriately.                 .

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   '.   ' hnclosure to TXX 91077 NROFORW$;na                          U 5. NUCLE AR RE0VLATORY COWWSSON APPROVEDOMS N0 31660104 E XPIRE 8:4'I M E ESTNATED BUROEN PER RESPONSE TO COMPLY W!TH T>nt PFORMA10N "C'"                   "' "E "' **""* ' " * ^ " C ""E"'8 * ***

LICENSEE EVENT REPORT (LER) GURDEN ESTNATE TO THE RCCORDS ##D REPORTS MANAGEWENT TEXT CONTINUATION "*""i"*'"uC'5^""'***'"'C""***"***"""" M:,30666. AND TO THE PAPERWORK REDUCTON PROJECT 13t1001040 0FFICE OF WANAGEMENT AND BUDGET. WASH.NOTON,00. 70603. Facdov kane (1) Docw Nunte (2) Li R NwSt* fel PaJe (% vn w w a w COMANCHE PEAK UNIT 1 0151010101414 l 5 911 - Ol0l4 - 010 017 OF 0I7 a.i e - . - . NRc F- m .m n . VI. PREVIOUS SIMILAR EVENTS There have been no previous reactor trips attributed to the causes identified during this event investigation. No previous reactor trips have been caused by personnel taking action in the plant that was not covered by a procedure or work control process, nor have any previous reactor trips been caused by insufficient labeling of plant equipment. Vll. ARQlIl0NAklNEQBMAIl0B , The times listed in the report are approximate and Central Standard Time, L

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