ML19325D237

LER 89-028-00:on 890912,extra Contact Found in Control Circuits for Svc Water Pumps P4A & P4C Which Was Not on Schematics.Caused by Inadequate Configuration Control During Design Phases.Wiring Insps Program initiated.W/891012 Ltr

ML19325D237
Person / Time
Site:	Arkansas Nuclear
Issue date:	10/12/1989
From:	Ewing E, Jacks J ARKANSAS POWER & LIGHT CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
1CAN108909, LER-89-028, LER-89-28, NUDOCS 8910190259
Download: ML19325D237 (6)

Text

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Arkanees Power & Upht Company O 42 est Captal

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Octouer 12, 1989 ICAN198989 U. S. Nuclear Regulatory Commission Document Control Desk Mail Station P1-137 Washington, D. C. 20555

SUBJECT:

Arkansas Nuclear One - Unit 1 Docket No. 50-313 License No. DPR-51 Licensee Event Report No. 50-313/89-028-00 Gentlemen:

In accordance with 10CFR50.73(a)(2)(1)(B), (a)(2)(ii)(B), (a)(2)(v) and (a)(2)(vii), attache (t is the subject report concerning a wiring error due to inadequate configuratioa controls which resulted in two inoperable Service Water pumps.

Very truly yours, O -

E. . Ewing l

General Manager, l Technical Support and Assessment 1

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Regional Administrator l Region IV U. S. Nuclear Regulatory Commission l

611 Ryan Plaza Drive, Suite 1000 Arlington, TX 76011 INPO Record *, Center 1500 Circle 75 Parkway r

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NRC Fore 366 U.$. Nuclear Regulatory Commission (9 83) Approved DMB No. 3150 01D4 Empires: 8/31/8b l L1CEN$EE EYENT REP 0R1 (L E R)

FACILITY NAME (1) Arkansas Nuclear One. Unit One IDOCAET NJMBIR (2) IPAGE (3) 10161010101 31 11 31110F1016 TITLE (4) Wiring Error Due to Inspequate Configuration Control Results in Two Inoperable Service water Pumps EVENT DATE (5) .ER NUMBER (6) 1 REPORT DATE (7) OTHER F AC]LITIf 51NVOLVED (8) l i i I Lequentiell lRevisioni i i i Month! Day lYear lYear l l Nuatier i i Number IMonthi Day lYear i Facility Names IDocket Nunber(s) 1 1 1 1 I i 1 i i i i 101 >10 010 1 1 Of 91 Il 21 81 91 81 91*-I 01 21 BI--I 01 01 11 01 11 21 81 91 101 ilo 000 t I

' 0PERATING l ITM11 REPORT 15 SUBMITTED PUR5 DANT TO THE REQUIREMENTS Of 10 CFR $;

MODE (9) i Ni (C 4ck one or more of the followinn) (11)

POWERI l__l 20.402(b) l__l 20.406(c) l__l 50.73(a)(2)(iv) l__l 73.71(b)

LEVELI l__l 20.415(a)(1)(1) l__l 50.36(c)(1) l_31 50.73(a)(2)(v) l__l 73.71(c)

(10) 1017141 1 20.4;5(a)(1)(11) l 60.36(c)(2) l_31 50.73(a)(2)(vit) l__l Other-(Specify in l__l 20.405(a)(1)(111) l[]ll50.73(a)(2)(1) l__l 50.73(a)(2)(viii)(A)I Abstract below and l__l 20.406(a)(1)(iv) l_31 60,73(a)(2)(11) l__l 60.73(a)(2)(v111)(B)l in Text. NRC Form I I 20.406(a)(1)(v) i 1 60.72fe)(2)(111) i 1 0.7)(a)(2)(x) I 366A)

L]CfM5ft CONTACT FOR THIS L R (12) l Telephone Number

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Name lArea l Julie D. Jacks. Nuclear safety and Licensing Specialist (Code i

!$10111916181-13111010 COMPLlT[ DN( LINf FOR ( ACH COMPONfNT FAILURE DESCR]p[D IN THIS RIPORT (13) i i i IReportablel I i i i IReportabiel Causel5vsteel Coenonent IManufactureel to N?RDS I ICausel5ysteel Component IManuf acterert to NPRDS I I i i l 1 i l i l i i l i l l l l I I l i i 1 I i l l I I i 1 I l l l 1 1 i i i i l i l i l i

! l I l l l l l l l 1 1 1 I I I I I l I I I 1 1 I SUPPLEM[NT REPORT EkPECTED (le) i EXPEC1ED l Month Day lYear

~ i SUBN!5510N 1 I l l Yes (if yes. conolete Expected Suboission Date) III No 1 DATf (15) l I i l i 1 A6? TRACT (Limit to 1400 spaces i.e.. approximately fif teen single space typewritten lines) (16)

On 09/12 13/89, an extra contact was found in the control circuits for Service Water pumps P4A and P4C which was not shown on the pumps' control circuit schematics. Under certain required starting conditions (e.g. , an Engineeren 5mfeguards actuation signal prior to or without a main ponerator lockout, causing a " slow" transfer to off-site power), the extra contacts would have caused the *enti-pump" circuit in the power supply breakers for the pumps to lockout the close signal to the breakers. This would prevent an automatic (as designed) or manual control room start of the pumps. Both P4A and P4C were declared inoperable until the circuit was modified and testen on 09/15/89. The event is considered safety significant due to the potential for loss of the Service Water System. The wiring error was detereined to be the result of inadequate configuration control during the design, construction, and startup phases of A40-1. A program of wiring inspections was initiated on both ANO-1 and AN0*2 to provide assurance that no additional safety significant wiring discrepancies existed. Although additional discrepancies have been identified, none have been found to be safety sigr.ificant. Further long-tere actions will resolve the identified discrepancies by revising drawings and/cr equipment as necessary.

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Form 1062.018 I NRC Fem M6A U.S. Nuclear Regulatory Coemission I (9-83) Approved One No. 31bO-0104  !

Empirest 4/31/85 LICENSEE TVENT REPORT (LER) TEXT CONTINUATION ]

FACILITY hapE (1) IDOCKET NVISER (2) l 1

.ER NLpMR ($1 1 PAGE (3) l l l lSeque3ttall luvisioni A konses Nuclear One, Unit One l l Vearl I Number I i n gggy,,1  ;

10161010101 31 11 31 el 91- I 01 21 81 -t 03 0101210F10lf> j YM(If more space is required, use asettional NRC fers 3MA's) (17) j

>4 A. Plant Status a

At the time of the discovery of this condition, Arkansas Nuclear One, Unit One (AND 1) was I operating at 74 percent of full power. l S. Event Description j During the recent NRC 01 agnostic Evaluation Team Inspection (August 21 - September 16, 1989), an issue relating to wiring discrepancies was brought to the attention of Arkansas Power & Light (AP6L) by the NRC evaluators. In response to this issue, AP&L performed selected inspections '

of the electrical schematic diagrams against the as built wirini' configuration of UtD-1 control room cabinets which contain Engineered Safeguards Actuation Systems control circuits. On .

i Septesener 12, 1989, a wiring disempancy was discovered relating to the ANO-1 Service Water pump  !

j' [ti-P) control circuit. A relay contact not shown on the schematic diagram was found to be wired i in the control circuit located in the control room for the ' A' Service Water pump motor breaker.

• An evaluation of the discrepancy, completed September 13, 1999, determined that the exwa contact  ;

would prevent the closure of the breaker for Service Water pump P4A if an Engineered Safeguards -

(ES) actuation signal occurred prior to or without a main generator lockout. The same wiring l trror was also found in the control circuit for Service Water pump P4C, ,

The Service Water pumps are powered from vital 4160V AC busses A3 and A4, which in turn are supplied by non-vital 4160V AC busses Al and A2. Busses Al and A2 are fed by either one of the two l transfomers which supply off site oower or by the unit sumiliary transformer (UAT), supplied by the  ;

main generator. During nomal power operations, busses Al and A2 are supplied by the UAT. If a main generator lockout occurs (e.g.. a reactor trip), the UAT feeder breakers trip and busses A1 and A2 will " fast" transfer to en off site power source, provided that no out-of-synchronisation condition i saists between the UAT and the of f site source. When a " fast" transfer occurs, the busses do not .

see an undervoltage condition. An E5 signal will also trip the UAT feeder breakers to busses A1  !

and A21 however, without the generator lockout there is no "f ast" transfer enabled. Instead,  !

busses Al and A2 will " slow" transfer to an off* site source on undervoltage on the busses. A i

" slow" transfer and its associated undervoltage condition causes a trip of the running Service Water pumps. The " slow" transfer is nomally completed in less than two seconfs and the running Service Water pumps are designed to automatically restart. An ses.rgency diesel generator for each vital bus, A3 and A4, would also start on the undervoltage condition but would not tie on to supply  ;

power to the vital busses as a " dead bus" condition would not exist.

The extra contact in the P4A and P4C control circuit would have affected the ability of the pumps [

to restart if an E5 signal was present and busses Al and At completed a " slow" transfer to an  ;

off-site source, e.g., a spurious E5 signal for both trains of E5 during power operations, or an ,

E5 signe) initiated prior to the ponerator lockout as could occur on a large break LOCA with a  !

rapid depressurization of the Reactor Coolant System. an E5 signal alone (with no ponerator '

lockout) would result in a " slow" transfer of busses Al and A2, and the extra contact in the P4A

, and P4C control circuit would have caused the " anti pap" circuit in the breakers to lock out the l close signal to the breakers, effectively disabling P4A and P4C. Although the breakers could j still have been mechanically closed locally at the breakers, the extraneous contact in the circuit I would have prevented the breakers from being electrically closed, either automatically, manually ,

l. from the handswitches in the Control Room, or locally at the breakers. l Seth pumps were declared inoperable at 2245 hours0.026 days <br />0.624 hours <br />0.00371 weeks <br />8.542225e-4 months <br /> on September 13, 1989, placing ANO 1 in a l 36 hour4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> time clock in accordance with the Lietting Condition for Operation (LC0) associated with i.

Technical Specification 3.3.6. Service Water pump P48 was not affected by the wiring error and I was operable. On September 15, 1989, t oth pumps were declared operable af ter performance of a j modification to remove the contacts and functional testing of the breakers. Pump P4C was declared operable at 0255 hours0.00295 days <br />0.0708 hours <br />4.21627e-4 weeks <br />9.70275e-5 months <br /> and P4A at 2100 hours0.0243 days <br />0.583 hours <br />0.00347 weeks <br />7.9905e-4 months <br />.

l l Further investigations of the wiring error indicated that the error occurred during the initial l construction of ANO 1. Therefore, Service Water pumps P4A and P4C had been inoperable since that time.

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Form 1062.018 NRC Ft.m 366A U.S. Nuclear Regulatory Commission (9 83) Approved ONB No. 3150 0104 ,

Expires: 8/31/85  ;

LICENSEE EVENT REPORT (LER) TEXT CONT]NUATION <

I FACILITY NAME (1) l DOCKET NdNBER (2) l tfR Nt#SER (6) l PACL (3) l l l 15equentiall (Revision]

Arkansas Nuclear One. Unit One l . l Yearl i Number l Number l '

10151010101 31 11 31 81 91 -I 01 21 ti-- 01 Ol01310F1015 ,

i. TEXT (If more space is required, use additionel NRC Fom 366A's) (17) i

't C. Safety Significance The Technical Speciffcations for ANO-1 require two out of three Service Water pumps to be operable whenuver containment integrity is required. Only one pump is necessary to mitigate the consequences of an accident. During the majority of the time that P4A and P4C were inoperable, P48 was operable and should have been available for accident mitigation had P4A and P4C faileo to start. However, there are several factors which make this event safety significant. Two Service Water pumps were effected by the wiring errors. Under the conditions described, the pumps would not have restarted automatically as designed nor could the operators have manually started the pumps from the control room handswitches. Also, the Service Water Sysr.em is a necessary support system for other E5 cosponents, such as the emergency diesel penarators, high pressure injection, and low pressure injection. A loss of Service Water could result in the f ailure of these systems to perfore their functions, which are necessary for safe shutdown of the unit and for accident mitigation.

D. Root Cruse A review of previous revisions of the applicable electrical schematic was conducted to determine when the extraneous contact was deleted from the drawing. Revision 10, issued in May 1974, revised the scheme to show the contact deleted. (Revisions 1 through 10 were issued during the design, construction, and startup phases on AND-1.) However, for reasons which could not be determined, the contact was not removed from the control circuits for P4A ..nd P4C, even though the schematic was revised to delete it. The related vendor connection drawings were not revised until approximately two years later. Apparently the vendor prints were not actually *as built" when turned over to AP&L by tie Architect /f ngineer (Sechtel) since the drawings show the extra contact in various states of being removed from the circuit.

The root cause of the event was inadequate configuration control during the design, construction, and startup phases of AND-1. A contributing cause was the failure to ptoperly implement a change to the circuit.

E. Basis for Reportability As two of the three Service Water pumps were inoperable for a period of time longer than allowed by the associated Technical Specification LCO, this condition resulted in operation prohibited by the Technical Specifications and is therefore reportable in accordance with 2f.CFR50.73(a)(2)(1)(B).

, Also, since the same condition was found to af fect both P4A and P4C, a loss of the Service Water 1 System could have occurred if P48 had been inoperable or had f ailed during the postulated sequence l

of events. As the Service Water System did not meet single failure criteria during the time P4A and P4C were inoperable, this condition is considered to be outside the design basis of the plant and is therefore reportable in accordance with 10CFR50.73(a)(2)(11)(B). This condition is also reportable in accordance with 10CFR50.73(a)(2)(v), a condition that alone could have prevented the fulfillment of the safety function of a system needed to maintain safe shutdown of the reactor.

l remove residual heat, or r,itigate the consequences of an accident. 51m11erty, in this event a j single cause or condition caused two independent trains to become inoperable in a system designed to maintain safe shutdown of the reactor, remove residual heat, or mitigate the consequences of an

( accident, and is therefore reportable in accordance with 20CFR50.73(a)(2)(vii).

1 l This condition was reported to the NRC Operations Center on September 13,1989, at 2330 hours0.027 days <br />0.647 hours <br />0.00385 weeks <br />8.86565e-4 months <br />, in I

accordance with 10CFR50.72(a)(2)(111).

F. Corrective Actions Due to the potential generic implications of this wirin0 discrepancy, AP&L promptly initiated an action plan to inspect selected additional safety related electrical equipment in the control rooms on both units at ANO. The objective of this action plan was to provide assurance through a sampling program that no additional safety significant wiring discrepancies existed between the as built configuration and the electrical schematic diagrams, connections diagrams, or vendor drawings, l

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Form 1062.018 ~j

' NRC F6m 36&A U.S. Nuclear Regulatory Commission i

"(9 83) Approved DNR No. 3160 0104 Iapires: 4/3) '% )

LIC(N$(( EV[NT REPORT (LER) T1X1 CONTIN 0AT10N ]

i FACit!TY NAML (1). 100ChEl NONDER (2) i .tk NueEtt (6) l PAGE (3) J l l l lioquentiall IRevisioni i Arkansas Nuclear One, Unit One l l Yearl I humber '

Number 1 )

10l$1010101 31 11 al el 91--I 01 21 8 --' 01 Ol014 M 10lb ,

TEKT (If more space is required, use additional NRC fore 3 %A's) (17) ,

i This inspection plan consisted of the following:

1. Identification of wiring by scheme numbers;

2. Field verification of schemes by visual inspection of control panels, teminal blocks and I internal component connections; I

3. Analysis of schematic wiring discrepancies for functional impact; and J

4 Initiation of additional actions as required based on the results of the analyses perfomed.

A three pttle approach was developed to perform the inspections. Phase 1 consisted of inspentons l and analyses of 4160V safety-related pump schemes in control panels for both units' control rooms.  ;

Phase 2 consisted of inspections and analysis of additional teminal block connections in the same ,

l panels. Phase 3 currently in progress, consists of edditional inspections and verifications of a  !

larger variety of electrical control cabinets and motor control centers both inside and outside  !

the ANO 1 control room. This is being done to provide additional assurance that no further safety issues exist.

Phase 1 was completed on September 18, 1989, and Phase 2 was completed on September 19, 1989.

Although numerous drawing discrepancies were discovered, the results of the Phase 1 and 2 inspections ,

and evaluations indicate that significant safety concerns do not exist in the wiring configurations  :

of safety related electrical equipment at ANO. Based on the results of these inspections, the  ;

schemes have been detemined to accurately reflect the required logic, and equipment operation can be anticipated 6s designed. Other than the relay contact circuit found in the ' A' and 'C' Service  ;

Water pumps, no ' sneak circuits or unknown failure modes of equipment were discovered during '

i Phases 1 and 2. The discr6pancies identified were primertly associated with the interna 3 connection diagrams (commonly referred to as vendor prints) and consisted mainly of errors such as deficient or incorrect labeitng and indications of spare wiring shown on internal connection diagrams that t does not exist physically in the plant.

Phase 3 of the inspection plan, which applied to AND*1 only, expanded the sample site to include additional safety-related cabinets and components both inside and outside the ANO-1 control room.

These inspections have been completed. A review of the identified discrepancies found none to be safety significant. The formal documentation and resolutions of the identified discrepancies are still pending. i Actions for ANO 2 beyond Phase 2 have been developed into e verification project plan which has  !

been initiated during the current AND-2 refueling outage 2R7. Based on the results of the Phase I 1 and Phase 2 inspections, the following criteria were developed to determine priorities for f further wiring inspections: t

1. Nature of the cabinet construction Cabinets with many discrete components and single conductor wiring with torsinal blocks for external cables have many more field accessible teminations and therefore, the potential '

for more errors. ,

2. Date of installation Equipment installed in the last few years was installed with enhanceo configuration control (e.g., better labeling, more thorough testing) and is less Itkely to have had a significant number of modifications since installation. +

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3. Number of modifications performed .

Equipment with few modifications has had fewer deswing revisions, hence less chance of errors.

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Form 10f.t.018 i NRC Fem SMA U.S. Nuclear Regulatory Consission l (9-83) Approved DM8 No. 3160 0104 r Empires 8/11/85 L!rtfistt tel REPORT (LER) TEXT CONTINUATION  !

~

s FACILITY NAME (1) lDOCKEl NMER (2) l LEW W M IR (6) l PAGL (3) l l l 15eguentiell lRevisioni Arkansas Nuclear One. Unit One- l 1. Year i N.ater i Nueuer i 10lbl010101 31 Il 31 el 9 1 Oi 71 St.- Oi 0101610Fl016 j TEXT (If more space is required, use eeditiona) NRC Fom 3MA's) (17)

4. Function of equipment ,

' Safety-related equipeer:t with the greatest potential fo4 plant impact w111 be verified first.

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l. Amount of internal single conductor wiring  !

Equipment with a high number of teminations has the poteritial for more disempancies.

t Based on visual inspections *ngineering judgement, the known history of each cabinet, ano the I criteria listed above, the AND 2 cabinets which are designated 'Q' and '$' were evaluated and i the cabinets to be inspected during the 2R7 outage were selected. The 'Q' 480V actor contro)  ;

centers were already scheduled to be inspected during this outaget the inspection criteria was enhanced due to this event. For the equipment selected to be inspected, scheme drawings will be marked up and evaluated for discrepancies in accordance with guidelines developed for the project.

For both AND 1 and AND 2, drawings with identified wiring discrepancies will be revised with a

Penting" note until the discrepancy 16 resolved and the drawing is updated appropriately. Also, t long tere plans for resolving the identified wiring discrepancies are being developed. Drawings  !

and/or equipment will be corrected as appropriate. This effort will also include actions such as s'nsuring adequate labeling of components, verifying equipment meets housekeeping requirements, and verifying that fiberglass sleeving is adequate for separation criteria.

G. Additfonal Infomation The following Licensee Event Reports have been subettted concerning wiring errors which affected >

or potentially affected equipment operability

• LER '50-368/86 020-00 Inoperable Emergency Diesel Generator caused by use of Incorrect Design Drawings to Perform Plant Modification and inadequate Post Modification ,

Testing  :

t LER 50-313/88-004-00 Reactor Building Hydrogen Concentration Instrument Inoperable due to >

Inadequate Post-Modification Testing  !

LER 50-313/88-013 01 Potential Failure of a High Pressure Injection Pump ts $ tart on Engineered Safeguards signal due to a treaker Wiring Error j Additionally, LER 50 313/89-002-00 discusses a reactor trip event which was coq 11cated by a wiring l error in the Integrated Control tystem af fecting the startup and low load feedwat6r control valves.

Energy Industry identification System (EIIS) codes are identified in the text as (KX). .

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