ML20011F583

LER 89-022-01:on 891114,normal Offsite Power Feeder Breaker to 4,160-volt Ac ESF Bus Opened,Resulting in Loss of Power to Bus 2A3.Caused by Inadequate post-maint Test Controls. Test Switch Opened & Job Order changed.W/900227 Ltr

ML20011F583
Person / Time
Site:	Arkansas Nuclear
Issue date:	02/27/1990
From:	Ewing E, Millar D ARKANSAS POWER & LIGHT CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
2CAN029011, 2CAN29011, LER-89-022, LER-89-22, NUDOCS 9003060308
Download: ML20011F583 (4)

Text

, - -

f j Arkenees Power & Ught Company

. / 42 West Canitol

, lutie Rock AD 72203

(. Te1501377 4:00 i .

1 F

February 27, 1990 '

i l 2CAN029011 L.

'U. S.. Nuclear Regulatory Commission Docoment Control Desk Mail Station P1-137 Washington, D.-C. 20555 *

SUBJECT:

Arkansas Nuclear One - Unit 2 '

Docket No. 50-368 License No. NPF-6 l Licensee Event Report No. 50-368/89-022-01 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, ,

l

? _

E. C. Ewing  :

General Manager, I. Technical Support l and Assessment ECE/0M/sgw attachment cc: Regional Administrator Region IV ,

U. S. Nuclear Regulatory Commission 611 Ryan Plaza Drive, Suite 1000 Arlington, TX 76011 INPO Records Center 1500 Circle 75 Parkway Atlanta, GA 30539-3064 b # '

9003060308 PDR ADOCK900227 05000368 !A -

A PDC- ~is-

. Form 1062.01A NRC Fors 366 U.S. Nuclear Regulatory Commission (9 83) Approved OMB No. 3150-0104 LICENSEE EVENT REPORT (L E R)

FACILITY NAME (1) Arkansas Nuclear One, Unit Two (DOCAET NUMBER (2) lPAGE (3) 1015l010101 31 61 81110rl013 TITLE *

(4) Inadequate Post Maintenance Test Controls Resulted in Deenergizing a 4160 VAC Engineered Safety Features Electric Bus Unexpectedly While Performing Post Maintenance Testing on an Auxiliary Relay .

EVENT DATE (5) LER HUMBER (6) REPORT DATE (7) OTHER FACILITIES INVOLVED (8) <

l l l 5equential l Revision l l Month Day Year Year Number Number IMonthi Day Year Facility Names 1, Docket Number (s) l l 0 5 0 0 0 4 11 1 11 4 81 9 81 91--I 01 21 2 --

' 01 11 Of 21 21 7 91 01 0 5 0 0 0  !

OPERAING lTH 5 REPORT 15 5UBMITTED PUR5UANT TO THE REQUlitEMENT5 0F 10 CFR 5:

WWE' f9) 5 (Check one or more of the followino) (11) i POWER ,,,,, 20.402(b) 1,,,,,1 20. 405(c ) lJ i 50.73(a)(2)(iv) l l 73.71(b) 1 LEVELI l_I 20.405(a)(1)(1) l_l 50.36(c)(1) i l 50.73(a)(2)(v) l l 73.71(c) l (10) 1 01010 1 20.405(a)(1)(ii) l.,,,1 50.36(c)(2) l l__l 50.73(a)(2)(v11)50.73(a)(2)(v111)(A))_l I

_l 20.405(a)(1)(111) l_I 50.73(a)(2)(1) l Abstract below and Other (SpecJ i~1 50.73(a)(2)(li)- l_l 50.73(a)(2)(viii)(B)l in Text, NRC Form i 1- 1l 20.405(a)(1)(iv) 20.405(a)(1)(v) l $0.73(a)(2)(111) I_ti 50.73(a)(2)(x) '

l' 366A)

LICENSEE CONTACT FOR THIS LER (12) ,

Name l Telephone Number  !

l Area i Dana Millar Nuclear safety and Licensing Specialist ICode 1 510111916141-13!11010 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT (h3) i 1 l l Reportable l l l l l lReportablel Cause System Component Manufacturer to NPRDS Cause System Component Manufacturer to NPROS l 1 i i l l i I I l l l l.I I I I I I I I l '

I I i I i I I i SUPPLEMENT REPORT EXPECTI:D (14) EXPECTED Month Day Year l $UBMIS$10N

_l~l Yes (if yes, complete Expected submission Date) til No l DATE (15) i I I l L AB5 TRACT (Limit to 1400 spaces, i.e., approximately fif teen single-spacs typewritten lines) (L6)

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 285 undervoltage relay contacts. Immediately following this step, the normal offsite 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 doenergized, a Low Pressure Safety injection (LPSI) pump, which was supplying flow for decay heat removal and a Service Water pump doenergized. A standby LPSI pump powered from the redundant 4160 VAC E$F electrical bus was started in approximately one minute and flow reestablished. Since the plant had been shutdown for several days prior to this event, the 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 reevaluated and satisfactorily completed. The root cause of this event was determined to be inadequate post maintenance test controls. An evaluation of the controls that 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).

h

Firm 1062.01B 5 WRC'Fom M6A U.S. Nuclear Regulatory Commission (9 83) .

Approved OMB No. 3150-U104 Expires: 8/31/85 LICENSEE EVENT REPORT (LER) TEXT CONTINUATION

• FACILITY N M (1) l DOCKET NUMBER (2) l LER NUMBER (6) l PAGE (3) l l l l5equentiall , Revision l Arkansas Nuclear One, Unit Two l 1 Year Number Number l a

10151010101 31 61 81 81 9 --

01 Fi 2 --

01 It0f210F1013 TEXT (If more space is requimd, use additional NRC Form 366A's) (17)

A. Plant Status" At the time of occurrence of this event Arkansas Nuclear One, Unit Two (ANO-2) was in Mode 5 (Cold $hutdown). Reactor Coolant System (RCS) (AB) temperature was approximately 178 degrees ,

Fahrenheit and RCs pressure was about 250 psia. The seventh refueling outage (2R7) for ANO-2 commenced September 25, 1989 and ended November 22, 1989.

B. Event Description '

During refueling outage 2R7, detailed as-built wiring verification inspections of several Control Room cabinets were performed by engineering personnel. On November 10, 1989, it was discovered that an aux 111ery relay (271X/2B5), actuated by undervoltage relays which monitor voltage on a 480 VAC Engineered Safety Features (ESF) Motor Control Center (NCC) (285), was not correctly '

wired, r

MCC 285 receives power from a 4160 VAC E$F electrical bus (2A3), which is normally energized by the stations offsite power system. In the unlikely event of a loss of offsite electrical power or

• degraded offsite power voltage conditions, the offsite power source is automatically disconnected and an Emergency Diesel Generator (EDG) (EK) (2K4A) is started to supply 2A3 and 285 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 condition actuate to energite -

auxiliary relay 27-1X/285. When the auxiliary relay is energized it functions to close contacts in each of two redundant automatic starting circuits for the associated EDG and to provide an open signal to the nomal offsite electrical feeder breaker to 2A3 (2A309). During the wiring inspection, it was found that one of the contacts on 27-1X/2B5 which provides one of the redundant 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 job 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 starting of the EDG and verifying both of the redundant EDG amergency start relays would energize when an undervoltage condition was simulated on 285.

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 285 bus undervoltage relay contacts to simuiste 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. Since the handswitch '

for 2KaA was in a pull-to-lock position, the EDG did not automatically start to provide electrical power for 2A3 cnd its associated loads. When power was lost to 2A3, a Low Pressure Safety Injection (LPSI) pump (BP P), which was supplying the shutdown cooling system (SDC) for decay heat removal, and a operating Service Water (SW) pump (B1 P) deenergized as designed.

Operations personnel restored SDC flow in approximately one minute by starting a standby LPSI pump powered from the redundant 4160 VAC E5F electrical bus. Bus 2A3 was reenergized in approximately three minutes from offsite power by reclosing breaker 2A309 and the SW pump restarted.

An evaluation of the event was condacted and it was determined that personnel developing and

\ reviewing the test instructions failed to recognize that auxiliary relay 27-1X/2B5 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 deenergizing 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, e'

i

.' I 2

e'* Form 1062.01B NRC Fom 366A U.S. Nuclear Regulatory Commission (9 83) Approved OMB No. 3150-0104

- LICENSEE lYENT REPORT (LER) TEXT CONTINUATION FACILITY MAME (2) (DOCKET NUMBER (2) l .ER NUMBER (6) l PAGE (3) l l l liequentiall I Revision) 1 Arkansas Nuclear One, Unit Two l l Year Number Number l 1 10151010101 31 61 81 86 9 --

01 21 2 --

06 Il0l310F1013 i TEXT (If more space is required, use socitional NRC fors 366A's) (17) i The loss of the Service Water pump resulted in the associated EDG being inoperable untti the pump was returned to service. However, the EDG was already prevented from automstically starting by placing the EDG start handswitch 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 .

l The root cause of this event was determined to be inadequate post maintenance test controls. The steps necestery to perform 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 formal 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 10CFR50.73(a)(2)(iv), as an unplanned actuation of an ESF system.

A 10CFR50.72(b)(2)(ii) notification to the NRC 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 i 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 interin memorandum has been prepared to require an impact statement on significant post maintenance 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 perfomed 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 procedures. The affected procedures will be revised by April 30, 1990. The second phase will be "

to develop a post maintenance testing guideline. The purpose of the guideline is to aid a i planner in developing work packages and to provide consistency in post maintenance testing. The ,

guideline will be issued by June 30, 1990.

G. Additional Information A simliar 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 Power and Light Company did not consider that the actuation of this auxiliary relay should be g considered an 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 reportability, additional guidance has been provided by Plant Licensing to Operations personnel regarding 10CFR50.72 notifications. Additionally, training will be provided for Operations personnel on 10CFR50.72 reporting criteria. This training will be completed by June 22, 1990.

Energy Industry Identification system (E!!S) codes are included in the text as [XX).