05000368/LER-1989-022

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LER 89-022-02:on 891114,4,160-volt Ac ESF Electric Bus Unexpectedly Deenergized During post-maint Testing on Auxiliary Relay.Caused by Inadequate Test Controls.Job Order Changed to Include Test switch.W/900525 Ltr
ML20055C649
Person / Time
Site: Arkansas Nuclear Entergy icon.png
Issue date: 05/25/1990
From: Ewing E, Millar D
ARKANSAS POWER & LIGHT CO.
To:
NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
2CAN059014, 2CAN59014, LER-89-022, LER-89-22, NUDOCS 9005300231
Download: ML20055C649 (4)


LER-2089-022,
Event date:
Report date:
3682089022R00 - NRC Website

text

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I . Arkansas Power & Light Company o%N a 3B 17G RusseHville. AR 72801 l Tel 501964 3100 J I

l May 25, 1990 2CAN059914 U. S. Nuclear Regulatory Commission i Document Control Desk .

Mail-Station P1-137 i Washington, D. C. 20555

SUBJECT:

Arkansas Nuclear One - Unit 2 Docket No. 50-368 License No NPF-6 Licensee Event Report No. 50-368/89-022-02 i

Gentlemen:

In accordance with-10CFR50.73(a)(2)(iv), attached is the subject supplemental

. report concerning inadequate post maintenance test controls which resulted in deenergizing a 4160 VAC Engineered Safety Features electric bus unexpectedly-while performing-post maintenance testing on an auxiliary relay. This report .

is being supplemented to provide a revised schedule for the implementation of the post maintenance testing program. I Very truly yours, i

/

E. C. Ewing i General Manager  ?

Technical Support and Assessment ECC/0M/sgw Attachment cc: Regional Administrctor  !

Region IV U. S. Nuclear Regulatory Commission 611 Ryan Plaza Drive, Suite 1000 '

Arlington, TX 76011 ,

INP0 Records Center Suite 1500 1100 Circle 75 Parkway Atlar,ta, GA 30339-3064 gg y ,

9005300231 900523 //

PDR S

ADOCK 05000368 An Entergy Company - y!

PDC-

'tg Foro 1062.01A NRC Fors 366 U.S. Nuclear Regulatory Commission (9 B3) Approved OMB No. 3150 0104 Expires: 8/3.'/85

LICEN$[E EVENT REPORT (L E R)

TAUDh hAME (1) Arkansas Nuclear One, Unit Two 100CAETNUMBER(2)(PAGE(U 10151010101 31 61 Bill 0Fl013 TITLE (4) Inadequate Post Maintenance Test Controls Resulted in Deenergizing a 4160 VAC Eagineered Safety Features Electric Bus Unexpectedly While Performing Post Maintenance Testing on an Auxiliary Relay EVENT DATE (5) LER NLMELR (6) REPORT DATI;(7)

OTHER FACILITIES INVOLVED (6) levision 15equentia l' I Month Day Year Year Number t

i fNmber Month Day Year Facility Names Docket Number (si 11 1 0 5 0 0 0

1. 11 4 81 9 81 9' -

Of 21 21 - 1 01 2 01 5 21.5 91 01 0 5 0 0 0 OPERA'ING ITH. 5 REPORT 15 5UBMITTED PUR5UANT

O THE REQUllsEMEN15 OF 10 CFR 5:

M30f (9) 5 (Check one or more of the followina) (11)

POWER) 20.402(b) l_ l 20.405(c)

LEVEll

_I 20.405(a)(1)(1)

I l,XI 50.73 a)(2)(v) a)(2)(iv) l_l 73.71(b)

(10) 101010 t 50.36(c)(1) _t 50.73 l l 20.405(a)(1)(11) l_

l_I 50.36(c)(2) l _ l 50.73 a)(2)(vit) l_ I 73.71(c) i

_ l 20.405(a)(1)(iii) l l_ l 50.73(a)(2)(viii)(A)l _ l Othen (Specify in Abstract below and i

_ l 20.405(a)(1)(iv) l_ 1 50.73(a)(2)(t) ii) l_ l 50.73(a)(2)(viii)(B)l in Text, NRC Form l l 20.405(a)(1)(v) I_l150.73(a)(2)(iii) 50.7.l(a)(2)( l I 50.73(a)(2)(x) l 366A)

LICEN5EE CONTACT FOR THIS LER (12)

Ge l Telephong Number Dana Millar, Nuclear Safety and Licensing Specialist tArea l ICode i 1510111916l41-13111010 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DE5CRIBED IN THIS REPORT (13) l l l 4Reportablel

. Causel$ystem i I I IReportablej LNanonent IManufacturert in NPRDS I ICause Systeml Component IManufacturer to NPRDS I I I I I I I I i 1 1 1 I I i i i 11 1 I I I I I i I I i i i l l I I I i i l I I i l l I I SIiPPLEMENT REPORT EXPECTl;D (14)

I I I I i l I i l l EXPECTED Mont h Day Year

.I'l Yes (If yes complete Expected Submission Date) til No i SUBMISSION l AB5 TRACT (Limit to 1400 spaces, i.e., approximately fifteen single-space typewritten I DALE (15) i I 1 I i 1 lines) (16)

On November 14, 1989, maintenance personnel initiated a post maintenance test, using instructions in a maintenance job order, to simulate an undervoltage on a 480 VAC Engineered Safety Features (ESF) Motor Control Center (2B5) by placing a jumper across the 2B5 undervoltage relay contacts. Immediately following this step, the normal of fsite power feeder breaker to the associated 4160 VAC ESF bus (2A3) unexpectedly opened resulting in the loss of power, to 2A3. The electrical bus deenergized as designed, The test steps provided in the job order did not identify that 2A3 would deenergize as part of the test. When 2A3 was deenergized, a low Pressure Safety Injection (LPSI) pump, which was supplying flow for decay heat removal and a Service Water pump deenergized. A standby LPSI pump powered from the redundant 4160 VAC ESF electrical bus was started in approximately one minute and flow reestablished. Since the plant had been shutdown for several days prior to this event 'he reactor decay heat levels were low and the momentary interruption of flow did not result in any significant Reactor Coolant System temperature or pressure increases. The test was teenluated and satisfactorily completed. The root cause of this event was determined to be inadequate post maintenance test controls.' An evaluation of the controls tLat are in place has been performed and the appropriate station procedures will be revised to reflect the results of the evaluation. This event is reportable pursuant to 10CFR50.73(a)(2)(iv).

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4

g. %e Form 2062 # B NRC F:ro 366A U.S. Nuclear Regulatory Commission (9-83) Approved DMB No. 31'20-0104 LICENSEE EVENT REPORT (LER) TEXT CONTINVATION FACILin hAME (1) (DOCAET NUMBER (2) l LER NUMBE R (6) l FAGE (3) l l l j5equentiali l Revision!

Arkansas Nuclear One, Unit Two l l Year Numbea Number l 101510!0l01 31 61 81 81 9 -- 01 21 2 -- 01 2101210F10I3 TEM (If more space is required, use additional NRC form 366A's) (17)

A. Plant Status l

.' At the tinie of occurrence of this event Arkansas Nuclear One, Unit two (AND 2) was im Mode $

(Cold Shutdown). teactor Coolant System (RCS) (AB) temperature was approximately 178 degrees Fahrenheit and RCS pressure was about 250 psia. The seventh refueling outt.ge (2R7) for AND-?

commenced September 25, 1989 and ended November 22, 1989. .

t B. Event Description ,

During refueling outage 2R7, detailed as-built wiring verification inspections of several Control Rone cabinets were performed by engineering personnel. On November 10, 1989, it was discovered that an amiliary relay (27-1X/2B5), actuated by undervoltage relays which molitor voltage on a 480 VAC Engineered Safety Features (ESf) Motor Control Center (MCC) (285), was not correctly wired.

MCC 2B5 receives power from a 4160 VAC ESF electrical bus (2A3), which is normally energized by the stations offsite power system, in the unlikely event of a loss of offsite electrical poner or degraded offsite power voltage conditions, the offsite power scurce is automatically disconnected and an Emergency Diesel Generator (EDG) (EK) (2K4A) is started to supply 2A3 and 2B5 with power.

These functions (opening of offsite power feeder breaker and EDG start) are initiated by the 285 bus undervoltage relays which upon detection of a low voltage conditioit actuate to energize auxiliary relay 27 1X/2B5. When the auxiliary relay is erargized it functions to close contacts f in each of two redundant automatic starting circuits for the associated EDG and to provide an open signal to the normal offsite electrical feeder breaker to 2A3 (2A309). During the wiring intpection, it was found that one of the contacts on 27-1X/2B5 which provides one of the tedundant EDG automatic starting circuits was not wired into the circuit as indicated on the applicable design drawings. ,

Upon a request by engineering, a job order was issued and the wiring discrepancy corrected. To ensure the wiring was correctly performed, engineering personnel were contacted to provide recommendations for testing the circuit.

The testing method, decided upon between engineering, maintenance and operations was included in the maintenance $b order, which was used to correct the wiring error, to provide the necessary steps to perform the test. The proposed test included placing the start handswitch for 2K4A in the pull-to-lock position to prevent actual ttarting of the EDG and verifying both of the redundant EDG emergency start relays would energize when an undervoltage condition was simulated on 2B5.

At approximately 0245 hours0.00284 days <br />0.0681 hours <br />4.050926e-4 weeks <br />9.32225e-5 months <br /> on November 14, 1989, maintenance personnel initiated the test by placing a jumper across the 2B5 bus undervoltage relay contacts to simulate undervoltage on the bus, immediately following this step, the normal offsite power feeder breaker (2A309) to the >

4160 VAC ESF bus unexpectedly opened resulting in loss of power to bus 2A3. Sinct the handswitch for 2K4A was in a pull i o-lock position, the EDG did not autcoatically start to provide electrical power for 2A3 and its associated loads. When power was lost to 2A3, a Low Pressure Safety injection-(LPSI) pump (BP P), which was supplying the shutdown cocling system (SDC) for decay heat removal,

, anti a operating Service Water (SW) pump (Bl*P) deenergized as designed.

[ Operations personnel restored SDC flow in approximately one minute by starting a standby LPS.1 pump powered from the redundant 4160 VAC ESF electrical bus. Bus 2A3 was reenergized in approximately three minutes from offsite power by reclosing breaker 2A309 and ths SW pump restarted.

An evaluation of the event was conducted and it was determined that personnel developing and reviewing the test instructions failed to recognize that auxiliary relay 27-1X/285 was also used to automatically open the offsite power feeder breaker (2A309) to bus 2A3 and therefore, performance

' of the test as written would result in d*energi2ing the electrical bus.

C. Safety Significance Electrical bus 2A3 deenergized as designed when the undervoltage condition was simulated on 2B5.

Since the plant had been shutdown for several days prior to this event, core decay heat levels were low. The unexpected loss of decay heat removal flow did not result in any significant increase in RCS temperature or pressure.

4  %

Ftre 3062.01B NRC F: m 360A U.S. Nuclear R:gulatory Commission j9-83)' Approved OMB No. 3150-0104

, Expires: 8/31/85 LICENSEE EVENT REPORT (LER) TEXT CONTINUATION

~

FACILITY KAME (1) IDDCAE1 NUMBER (2) ! LfR NUMEER (6) l FAGE (3) l l l 5eouentiell (Revision l Arkantes Nuclear One Unit Two I i Year Number Number i 101510101013161 Bill 9 --

01 ?l 2 --

06 Pl01310F1013 TEXT (If more space is required, use additional hRC Form 366A's) (17)

The loss of the Service Weter pump resulted in the associated EDG being inoperable until the pump '

was returned to service. However, the EDG was already prevented from automatically starting by placing the EDG start handswitet in pull to-lock as directed by the test in the job order.

As a result of this event, there were no significant safety concerns.

D. Root Cause The root cause of this e'. e r wcs determined to be inadequate post maintenance test controls. The steps necestery to perfors, the post maintenance testing were determined by engineering personnel and it was decided to implement the test by including the test instructions in the maintenance job order used to correct the wiring problem. A fomal technical review (e.g. , independent review) of the testing steps was not performed prior to performing the activity.

E. Basis for Reportability This event is being reported pursuant to 10CFp.50.73(a)(2)(iv), as an ursplanned actuation of an ESF system.

A 10CFR50.72(b)(2)(11) notification to the NkC was made on November 17, 1989 at 0050 hours5.787037e-4 days <br />0.0139 hours <br />8.267196e-5 weeks <br />1.9025e-5 months <br />.

F. Corrective Actions A test switch located in the electrical trip circuit for 2A309 should have been opened to preclude the inadvertent tripping of the breaker and loss of power to 2A3. The job order instructions were changed to include this test switch and the test was performed satisfactorily with no further unexpected events.

An interim memorandum has been prepared to require an impact statement on significant post main-tenance tests implemented by job orders. The memorandum was issued on December 19, 1989. A procedure revision and training on the revision was completed by January 15, 1990.

An evaluation of the controls of the post maintenance testing that are currently in place has been performed and has resulted in the identification of more extensive procedure revisions than originally anticipated. A two phase plan to implement enhancements to plant procedures has been developed. The first stage will identify and revise the applicable station administrative pro-cedures. The second phase will be to develop a post maintenance testieg guideline. The purpose of the guideline is to aid a planner in developing work packages and to provide consistency in post maintenance testing.

As actions have been taken to implement the two phase plan, it has been identified that additional procedures and training will be necessary to ensure a satisfactory program is established. The pro-cedures, guidelines and training for the post maintenance testing program will be completed and the program implemented by July 31, 1990. After implementation of the program, each corrective mainten-ance job order which is issued for 'Q', 'S' or 'F' components will be evaluated for post maintenance testing guidelines using the approved plant procedure " Control of post-Maintenance Testing".

Additionally an evaluation of the current preventative maintenance procedures will be performed to assess the need for testing enhancements. This evaluation will be completed by October 1, 1990, G. Additional Information A similar event due to inadequate work controls was reported in LER 50-313/88 023 00.

The 10CFR50.72 notification was not made in a timely manner following occurrence of the event due to a lack of understanding that the unplanned opening of the offsite power feeder breaker to the 4160 VAC ESF electrical bus should be considered an ESF actuation. Prior to this event, Arkansas Fower and Light Company did not consider that the actuation of this auxiliary relay should be considered en ESF actuation and was, therefore, not reported under 10CFR50.72 or 10CFR50.73 reporting criteria.

To enhance the current process used in evaluating events or plant conditions for r9 portability, additional guidance has been provided by Plant Licensing to Operations personnel regarding 10CFR50.72 notifications. Additionally, training will be provided for Operations personnel on 10CFR$0.72 reporting criteria. This training will be ccepleted by June 22, 1990.

Energy Industry Identification System (Ells) codes are included in the text as [XX).

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