ML19332F617

From kanterella
Jump to navigation Jump to search
LER 89-037-00:on 891110,reactor Trip Occurred as Result of Inadvertent Grounding of Reactor Protection Sys Power Supply During Surveillance Testing.Caused by Inadequate Procedure. Procedures revised.W/891211 Ltr
ML19332F617
Person / Time
Site: Arkansas Nuclear Entergy icon.png
Issue date: 12/11/1989
From: Ewing E, Taylor L
ARKANSAS POWER & LIGHT CO.
To:
NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
1CAN128907, LER-89-037, LER-89-37, NUDOCS 8912180061
Download: ML19332F617 (4)
v • d • e


Text

r- ,

4  : C

, . [ .

Arkansas Power & Upht C1mpany

.. / Arkansas Naciear One

  • N- 3 Raate 3. Box 137 G i i

nasse!ioM. IsR 70B01 Tel 501 !$4 3100 i

L i

December 11, 1989 i- ICAN128967 I U. S. Nuclear Regulatory Commission i 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. 50-313/89-037-00  :

Gentlemen:

In accordance with 10CFR50.73(a)(2)(iv), attached is the subject report r concerning a reactor trip which was caused by the inadvertent grounding of a Reactor Protection System power supply during surveillance testing due to ,

an inadequate procedure.

Very truly yours, E. C. Ewing General Manager,  ;

Technical Support ,

and Assessment 1 ECE/RHS/sgw attachment cc: Regional Administrator i Region IV l U. 5. Nuclear Regulatory Commission 1 611 Ryan Plaza Drive, Suite 1000 Arlington, TX 76011 INPO Records Center l 1500 Circle 75 Parkway  !

Atlanta, GA 30339-3064 l l

l 8912180061 891211 PDR ADOCK 05000313

%N i S PDC II An [ rag Cym

r-

)

. .. v

. 1 Form 1D62.01A i

i NkC Fore 366 U.S. Nuclear Repu1 story Commission (9 63) Approved DMB No. 3160-0104 Empires: 8/31/86 L1CEN5tt tVENT REP 0RT (L E R)

FACILITY NAME (1) Arkanswa Nuclear One, Unit One IDOCAf f NUMBER (2) IPAct (3) t 10161010101 31 11 31110F l013 l TITLE (4) Reactor Trip Caused by the Inadvertent Grounding of a Reactor Protection System Power  ;

Supply During surveil.ance Testing Due to an Inseequate Procedure

.[VtW" DATE (6) I utR NUMBER (6) i REPORT DATI (7) 1 OTHth F AtiLITIES INVOLV[D ($)

I i i i tioquentiell laevisioni i i l l Month Day lYear lYear l l Number l l Number IMonthi Day l Year i Facility Names IDo:ket Nunber(s) i i i l l l l l l l 101 i101010 Il 11 11 0 $1 91 81 91--I 01 31 71**l 01 0! 11 ft 11 11 81 91 101 v101010. .

OPERATING ITH15 REPORT 15 SuSMITitD PukbuANT TO THE REQUIREMENTS OF 10 CFR 5:

MODE (9) l Ni (Check one or more of the followinn) (11)

PDWERI l,,,1 20.a02(b) 1._l 20.406(c) 1,kt 60.73(a)(2)(iv) l_,1 73.71(b) i LEVELI l I 20.406(a)(1)(1) l__l 60.36(c)(1) 1._1 60.73(a)(2)(v) l_,1 73.71(c)

(10) 1017141 l 20.405(a)(1)(11) l__l E0.36(c)(2) 1,_l bO.73(a)(2)(v11) 1,_l Other (Specify in l_,1 20.406(a)(1)(111) l I b0.73(a)(2)(1) 1.,1 60.73(a)(2)(v111)(A)! Abstract below and -

l l 20.406(a)(1)(iv) l ! 60.73(a)(2)(11) l__l 50.73(a)(2)(v111)(B)l in Text, NRC Form i 1 20,406fa)(1)(v)  ! I 60,73 :a)(2)(iii) I l 60.73:a)(2)fr) l 366A)  ;

LICEN5tL CDNTACT FOR THl$ LfR (13)

Name J Telephone Number

[ Aret i Larry A. Taylor, Nuclear Safety and Licensing Specialist ICode I 1510111916141-13111010 COMPLETE ONT LINE FOR CACH COMPON(NT FAILURE Di$CklB(0 IN TH15 RfPORT (13) i i i IReportabiel i i l I (Reportablel Causel$ysteel Coesonent IManufacturerl to NPRD$ 1 (Causel5ysteel Component lHanufacturert to WPRDS I 1 i l i l l I l I l i i i l I I i l l l i l l l l l l l l l l l l l 1 I i l 1 i i l 1 l 1 i i l I i i i l t i i l I I i i i l i l i i i i i l i SUPPLEMEN" REPORT [kP(CTED (14) l EKPICitD l Monthi Day lYear

~ 1 500MI5510N i l i l l Yes (If yes. comotete Exoected Subsission Date) III No l CATE (1$) l l l 1 l l ABSTRACT (Limit to 1400 spaces, i.e. , approximately fif teen single-space typewritten lines) (16) ,

On November 10, 1989 at approxinately 2266, a reactor trip occurred as a result of the inadvertent [

grounding of a Reactor Protection System (RP5) power supply during surveillance testing. The gfr.unding resulted in the loss of the RPS power supply and oeenergized the reactor power auctioneering circuit for the Integrated Control Systei (ICS) which resulted in a reactor power /feedwater flow mismatch. The ICS automatic response to this mismatch was to reduce feedwater flow to the steam ponerators and to withdraw control rods. These actions resulted in a rise in Reactor Coolant System (RCS) temperature .

and pressure and a reactor trip at 2355 psig. The initial plant response following the trip was as expected, with all post trip parameters being normal. However, due to various steam leakage paths in the secondary systee, the steam ponerator pressures gradually decayed to approximately 860 psig and RC5 temperature decreased to $35 degrees. The cause cf this event was an in40 equate procedure which required connecting a test lead to a soldered connection in the back of the RPS cabinet. This connection was within one eighth inch of the connection which was inadvertently grounded. The RP5 calibration procedure was revised to specify taking the required reading from a more suitable location. In addition, an evaluation was performed to identify and recommend appropriate corrective actions for long standing secondary systee 5eficiencies.

I

1 j t

l

. I f ore 1062.018 NRC Forn 666A U.S. Nuclear Regulatory Commission j (9 63) Approved 048 ho. 3160 01D4 1 Empires: 8/31/95 LICtN$tt EVtKT kiPORT (LER) TEXT CONT 140ATIDN FACILITY NAPE (1) IDDCACT NdMBLR (2) l LFR NJMBfR I6) i PAGt (3) i l l l l$equentiell IRevisioni l Arkansas Nucleer One, Unit One i 1.Yeart I humber i i Numbeg_l 1 10161010101 31 11 31 61 91--I 01 3r 71--I 01 0101210Fl013 i TtXT (If more space is reeutrod. use additional NRC f ore 366A's) (17)  ;

A. Plant $tatus At the time of this event. Arkansas Nuclear One. Unit one ( AND 1) was operating at approximately 74 percent of rated power. Reactor coolant Systeo (RCS) ( AB) pressure was 2150 psig and RC5 ,

everage temperature was $79 degrees. The *0" reactor coolant pump (RCP) was out of service due to en oil leak.

B. Event Description On hovember 10. 1989 at approximatuly 2256, a reactor trip occurred as a result of the +

inadvertint grounding of a Reactor Protection System (RP$) (JC) pnwer supply during surveillance i testing.

A portion of the 18 month RPS channel calibration, which is normally performed during refueling outages, was being performed at power in order to reduce the work load during an upcoming. Short duration mid-cycle outape. During the Performance of the RP5 channe) *B" calibration, an Instrumentation and Control technician inadvertently touched the wrong connection while attempting to connect a test lead to a 3016ered connection located in the back of the 'B" RP5 cabinet. This action caused the circuit breaker for the alb volt direct current (VDC) power supply for RP5 channel *B" to trip. Loss of the RP5 channel "B" power supply doenergized the reactor powtr  !

auctioneering circuit for the Int

  • grated Control system (105) (JA). The loss of the reactor power ,

signal to the ICS resulted in a reactor power /feedwater flow mismatch. The ICS automatic response

  • to this mismatch was to reduce feedwater flow to the Once Through stone Generators (OT5G) and to withdrew control rods to increase power. The reduced feedwater flow caused a secrease in heat '

transfer capability, a rise in RCS temperature and pressure, and a reactor trip on high RC5 pressure at 23b5 psig. The initial plant resporse following the trip was as expected, with all l post trip parameters being normal. However, due to various stems leakape paths in the secondary i system, the OT50 pressures gradually secayed (approximately 30 minutes) to 870 psig for a *A" OTEG and 860 psig for the *B" 075G. The decrease in 015G pressures in turn caused RC5 average temperature ,

to secrease to 535 degrees. These values are slightly below those normally anticipated during post trip conuttions (pressure - 2005 psig, temperature - 545 pogrees).

The major contributor to the OTSG pressure segradatiol and associated RCS cooldown was stone ,

leakage through the moisture separator reheater (MSR) isolation valves, then through manually positioned M5R distiller level control valves to high pressure heaters (*1A and B. The relief valves on the shall side of the high pressure heaters lif ted, relieving steem to the atmosphere.

Other sources of stone leakage included the feedwater pump turbine stops and governnr valves, and the main turbine bypass valves.

The operators recognited the abnormal post trip response and took timely and appropriate corrective cctions to isolate the variour stone leakage paths. The plant was stabilised in the hot shutdown condition at approximately 2340.

At 0335 on November 11, 1989 reactor startup was commenced. The main turbine was tied to the line on November 12 at 1605.

C. Safety $10nificance During this event, a reactor trip was initiated at an RC$ pressure of 2355 psig, as required, and all control rods inserted. All plant parameters remained within normal bands with the exception of CT5G pressures and RCS average temperature, which were slightly lower than normally anticipated during post trip conditions. The operators took timely and appropriate corrective actions to stabilize the plant in the hot shutdown condition. Although the malfunctioning of secondary systes components complicated the operator's post trip responses, they did not create any significant safety problems. Therefore, the safety significance of this event is considered einimal, e

_ _ . . . _ _ _ , _ . , . _ _ . . ~ .

~

r .

l Fors 1062.018 l

.. WRC Fors 3M A U.S. Nuclear Repuistory Coamission j (9-83) Approved 088 No. 3150-0104 .

l fxpires 8/31/85 l LICEN$(( [YENT REPORT (LER) TEXT CONTINUATION FACILITY NRML (1) IDOCA[1 NVIG(R (2) l t[R IkDSIR (6) j PAGL (3) i l I l$equentiell IRevisioni Arkansas Nuclear One. Unit One I l Yearl I humber I i Number i 10161010101 31 11 31 61 91--I Of 31 7t--t O! 0101310F1013 '

M (If more space is riquired, use aeditional NRC Fcts 3%A's) (17)

O. Root Cause The root cause of this evert was determined to be en inadequate procedure. The RPS calibration procedure called for attaching a test lead to a soldereo connection in the back of the RPS cabinet which is in close proximity to other connections. The connection which was inadvertently touched by the technician initiating the trip was approximately 1/8 inch from the specified test point.

E. Basis for Reportability This event is repertable pursvent to 3DCFR50.73(a)(2)(iv) as an autematic attuation of the RP$. t This event was also reported in accordance with 10CFR50.72 on November 10, 1989 at 2400.

F. Corrective Actions The appropriate procedures were vsvised to specify a more suitable location in the RP$ cabinets to get the required measurements.

Additionally, af ter a subsequent reactor trip on November 14. 1989 (LER 50-313/89-038 00), during which secondary system problems similar to those experienced during this trip were observed, a Secondary Systems Evaluation Team was formed to address long standing problems associated with the operation of AND-1 secondary plant systems.

The objectives of the tese were to identify the long standing material problems on the AND-1 secondary plart that created the need for operator compensatory actions durih 2 transients. The problems were addressed individually and on an integrated system operation basis to determine which deficiencies required correction prior to restart from the old cycle outage currently in progress.

i As a result of these assessments, six itees were identified as significant enough to require action prior to restart. These six deficiencies were:

e Feedwater pump turbine high pressure stop and governor valves leak.

  • Meistun 5erstator reheatee distiller level controller is inoperable, e Hester drain tank T40 high level dump valves leak excessively.
  • Main feedwater pump recirculation valves leak excessively.

Appropriate corrective actions will be implemented with respect to each of these deficiencies prior to restart from the outage currently in progress.

C. Additional Information There have been no previous events in which reactor trips occurred during surveillance testing due to an inadequate procedure.

Energy Industry Identification Systee (E!!$) codes are indicated in the text as [XX).

--

Retrieved from "https://kanterella.com/w/index.php?title=ML19332F617&oldid=2606063"