ML17325A196

LER 87-008-00:on 870604,ESF Actuation Occurred Due to Extreme High Level in Steam Generator 12.Caused by Mispositioning of Both Level Controller & Feedwater Flow Selector Switch.Level Controller refurbished.W/870702 Ltr

ML17325A196
Person / Time
Site:	Cook
Issue date:	07/02/1987
From:	Blind A, Will Smith INDIANA MICHIGAN POWER CO. (FORMERLY INDIANA & MICHIG
To:	NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM)
References
LER-87-008, LER-87-8, NUDOCS 8707130305
Download: ML17325A196 (6)

Text

REGULATl INFORMATION DIBTRIBUTIONQ'STEM <RIDE)

ACCESSION NBR: 8707130305 DOC. DATE: 87/07/02 NOTARIZED: NO DOCKET 0 FACIL: 50-315 Donald C. Cook Nuclear Power Plant> Unit 1> Indiana 0 05000315 AUTH. NAME AUTHOR AFFILIATION BLINDS A. A. Indiana Zc Michigan Electric Co.

SMITHS'. G. Indiana 8c Michigan Electric Co.

RECIP. NAME RECIP IENT AFFILIATION

SUBJECT:

LER 87-008-00:on 870604iESF actuation occurred due to extreme high level in steam generator 12. Caused bg mispositioning of both level controller 5 feedwater flow selector switch. Level controller refurbished. W/870702 ltr.

DISTRIBUTION CODE: IE22D COPIES RECEIVED: LTR ENCL SIZE:

TITLE: 50. 73 Licensee Event Report (LER)i Incident Rpt> etc.

NOTES:

REC IP IENT COPIES RECIPIENT COPIES ID CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL PD3-3 LA 1 PD3-3 PD 1 1 WIGQINQTONi D 1 INTERNAL: ACRS MICHELSON 1 1 ACRS MOELLER 2 2 AEOD/DOA 1 1 AEOD/DSP/ROAB 2 2 AEOD/DSP/TPAB 1 1 DEDRO 1 NRR/GEST/ADE 0 NRR/DEST/ADS 0 NRR/DEST/CEB 1 1 NRR/DEST/ELB 1 1 NRR/DEST/ICSB 1 1 NRR/DEST/MEB 1 NRR/DEST/MTB 1 1 NRR/DEST/PSB 1 NRR/DEST/RSB 1 1 NRR/DEST/SGB 1 1 NRR/DLPG/HFB 1 1 NRR/DLPG/GAB 1 NRR/DOEA/EAB 1 1 NRR/DREP/RAB ~ 1 1 NRR/DREP/RPB 2. 2 RN~ RB NRR/PMAS/PTSB 1 EQ FILE 02 1 1 RES DEPY GI 1 1 RES E FORDS J 1 1 RES/DE/EIB 1 RQN3 FILE 01 1 1 EXTERNAL: EG~<<G GROHi M 5 5 H ST LOBBY NARD 1 1 LPDR 1 1 NRC PDR 1 1 NSIC HARRIS' 1 NSIC MAYST Q 1 TOTAL NUMBER OF COPIES REQUIRED: LTTR 44 ENCL 42

NRC Form 355 (043) 0 US. NUCLEAR REOULATORY COMMISSION APPROVED OMR NO. 3150410l LICENSEE EVENT REPORT (LER) EXPIRESI SISIISS FACILITYNAMK (I) DOCKET NUMBER (2) PA E TITLE (Cl D.C. Cook Nuclear Plant, Unit 1 0 5 0 0 0 3 1 5 1 OF P 4 ESF Actuation (Reactor Trip) Due to extreme High Steam Generator Level Resultin from Mis ositioned Control Switches EVENT DATE (5) LER NUMBER (5) REPORT OAT K (7) OTHER FACILITIES INVOLVED (SI MONTH OAY YEAR YEAR SEQUENTIAL C.W RNVCUCN MONTH OAY YEAR FACILITYNAMES DOCKET NUMBERISI NUMSNII NUMSNII 0 5 0 0 0 0 6 0 4 8 7 8 7 0 0 0 0 0 2 8 7 0 5 0 0 0 THIS REPORT IS SUBMITTED PURSUANT T0 THE REOUIREMENTs oF 10 cFR (II Icrrocc onr or moro of trrr forloiffnpl (11 OPERATINO MODE (5) 20A02(N) 20A05(c) 50.73(a) (2 l(h) 73.7101)

POWER 20AOS tc l(1)(i) 5035(c)(I ) 50,73(a)(2)(r) 7371(cl LEUEL p 9 p 20AOS(cHI ) (Nl 5024(cl(2) 50.73(a) (2)(rN) DTHER ISprcrfy in Aororct oolow onrf In Toit, IIRC Fomr 20.405(c) (1) (IN) 50.734)(2) (I) 50.734) (2)(rNI)(A) 388AI 20A054) (1)(lr) 50.734)(21(NI 50.73(a I (2) (rid) (5) 20AOS(c)(1)4) 50.734)(2)(NI) 50.73(c) (2)(xl LICENSEE CONTACT FOR THIS LKR (12)

NAME TELEPHONE NUMBER AREA CODE A. A. Blind Assistant, Plant Manager 6 1 64 6 5- 5 9 0 1 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS RKPORT (13) r CAUSE SYSTEM COMPONENT MANUFAC. EPORTABLE MANUFAC. EPORTABLE TUR ER TO NPRDS CAUSE SYSTEM COMPONENT TURER TO NPROS SUPPLKMENTAL REPORT EXPECTED ('ll) MONTH OAY YEAR EXPECTED SUBMISSION DATE (15)

YKS Ilfyrt, ccrnplrtr EXPECTED SIIEMISSION DATEI NO ABSTRACT ILlmlt to lr00 rPrtrr, I A, rPProiimorrly fit ttn rlnPlororto tyorrwltmn ffnrrl (IS)

On June 4, 1987, at 1034 hours0.012 days <br />0.287 hours <br />0.00171 weeks <br />3.93437e-4 months <br />, an ESF actuation (turbine/reactor trip) occurred due to an extreme high level in Steam Generator number 12. The level transient occurred during the performance of a test procedure. During the execution of the test, an operator was directed to place the automatic/manual level controller in the manual position, then transfer the steam flow and feedwater flow selector switches from the Channel I position to the Channel II position. Once transferred, the level controller can be placed in the automatic position and testing can commence on Channel I.

The cause of the event was the mispositioning of both the level controller (due to malfunction) and the feedwater flow selector switch (due to misleading switch position labeling) . This resulted in an increase in feedwater pump speed, the automatic opening of the Steam Generator number 12 feedwater flow control valve and sub-sequent extreme high Steam Generator level.

The level controller was removed, inspected, refurbished and determined to be operable. The escutcheon plate on the feedwater flow selector switch has been relabeled to identify the correct position. In addition, a memo was issued to all operators describing lessons learned from this event.

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NRC Form 355 1543)

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NRC Form 358A U)L NUCLEAR REOULATORY COMMISSION 983 I LICENSEE EVENT REPORT (LER) TEXT CONTINUATION APPROVEO OMS NO. 3IEO-0104 EXPIRES: 8/31/88 FACIUTY NAME ill DOCKET NUMBER (2I LER NUMBER ISI PACE l3I YEAR 54CVENTIAL RtVISION NVM54R NVM 4R D.C. Cook Nuclear Plant, Unit 1 0 6 0 0 0 3 1 5 8 7 0 0 8 0 0 0 2 oF 0 4 TEXT /SIRIW NMCe /I/etvlRS vtt ahWont////IC %%drm SSSI8/ l1TI Conditions Prior to Occurrence Unit 1 in Mode 1 (Power Operation) at 90 percent Reactor Thermal Power.

Descri tion of Event On June 4, 1987, at 1034 hours0.012 days <br />0.287 hours <br />0.00171 weeks <br />3.93437e-4 months <br />, an Engineered Safety Features actuation (turbine trip/reactor trip) occurred due to a extreme high level in Steam Generator (EIIS/SG) number 12. The Steam Generator level transient occurred during the performance of a surveillance test procedure, 1-THP-4030-STP.019 (Steam Generator 1 and 2 mismatch set number 1). During the execution of the test, the control room operator is directed to place the automatic/manual level controller (EIIS/LC) in the manual position, then transfer both the steam flow and feedwater flow selector switches (EIIS/33) from the "Channel ILI position to the "Channel IIN position. Once transferred, the automatic/manual level controller can be placed in the automatic position and testing can commence on channel I.

When the control room operator (Licensed Reactor Operator) acted

'o place the automatic/manual level controller in manual, the switch did not. fully travel to the manual position detent (the switch is spring loaded to the manual position and should have traveled to manual once it was moved, even slightly, from the automatic detent position) . Consequently the controller remained in automatic, unknown to the operator. The operator then rotated steam flow selector switch from channel I to channel II. No adverse effects resulted from this transfer. The operator then acted to transfer the feedwater flow selector switch from channel I to channel II. It was later determined that the selector switch was not fully rotated to the channel II position due to misleading switch position labeling and therefore neither the channel I or channel II output was selected. With the level controller still in automatic, it compared a full power steam flow signal with a zero feedwater flow signal. This resulted in an increase in feed-water pump (EIIS/P) speed and the automatic opening of the steam generator number 12 feedwater flow control valve (EIIS/FCV).

The ensuing transient occurred so rapidly the actions taken by the operator to regain control of steam generator level were ineffective [this was complicated by the fact that there was an erroneous low steam generator level indication available to the operator due to a mechanical interference within the steam generator level recorder (EIIS/LR)] and a turbine trip/reactor trip sequence was initiated [opening of the reactor trip breakers (EIIS/BKR), insertion of reactor control rods (EIIS/ROD), feedwater isolation, automatic starting of the motor-driven and turbine-driven auxiliary feedwater pumps (EIIS/P) ] . There was no automatic or manual actuation of the intermediate head safety injection system.

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NRC Form 388A US. NUCLEAR REGULATORY COMMISSION 883)

LICENSEE EVENT REPORT (LER) TEXT CONTINUATION APPROVEO OMB NO. 3180-0105 EXPIRES: 8/31/88 PACIUTY NAME lll DOCKET NUMBER IE) LER NUMBER (8) PAGE (3) rr 55CVENTIAL 'P RSVr5ION YEAR NVMS ~ R NVM55R D.C. Cook Nuclear Plant, Unit 1 0 s 0 o o 3 1 5 8 7 0 0 8 8 7 0 3 QF 0 4 TE)TT /e/rrroo rpooo/5/orp/8mS Ir ~//RC a 38883/ I)T)

, Operations personnel immediately implemented the special emergency operating procedure 1-0HP-4023.E-O to verify proper response of the automatic protection system (EIIS/JC) and to assess plant conditions for initiating appropriate recovery actions.

The Unit was stabilized in Mode 3 (Hot Standby) at approximately 1100 hours0.0127 days <br />0.306 hours <br />0.00182 weeks <br />4.1855e-4 months <br />, June 4, 1987. The NRC was notified of the event via the ENS at 1120 hours0.013 days <br />0.311 hours <br />0.00185 weeks <br />4.2616e-4 months <br />, June 4, 1987.

With the exception of the malfunction of the level controller switch mechanism and the steam generator level recorder, there were no inoperative structures, components, or systems that contributed to this event.

Cause of Event Several unrelated items (both human factors related) combined to result in this event: 1) the automatic/manual level controller transfer switch malfunctioned which prevented typical (spring assisted) travel to the manual position, and 2) the feedwater flow selector switch was not fully rotated to the channel II position [The reason this did not occur is twofold. First, the escutcheon plate on the subject selector switch incorrectly identified the channel II position at 45 degrees from channel I (in reality the channel II position is 90 degrees from channel I). Secondly, the subject selector switch is the only switch with the channel II position at 90 degrees, the channel II position on all corresponding selector switches is at the 45 degree position.

Anal sis of Event This Engineered Safety Features actuation which resulted in a reactor trip, is reportable pursuant to 10 CFR 50.73 (a)(2)(iv) .

The automatic protection system responses; turbi'ne trip/reactor trip, and resultant actuations, were all verified to have functioned properly as a result of the Engineered Safety Features actuation.

Based on the above, public were not affected.

it is concluded that the health and safety of the It was noted, however, that the control room temperature indication, the chart drive for the wide range reactor coolant system temperature recorders, and the Unit Megawatt Recorder malfunctioned. Further investigation revealed that these abnormal responses were due to a broken wire on a critical control room power circuit breaker. This broken wire was the result of a cable pull

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LICENSEE EVENT REPORT (LER) TEXT CONTINUATION APPROVEO OMS NO. 3180M)04 EXPIRES: 8/31/88 SACILITY NAME 11) OOCKET NUMSER )31 LER NUMSER (8) PAGE (3)

YEAR SEOVSNTIAL REVISION NVM SR NUMSSR D.C. Cook Nuclear Plant, Unit TEXT (4 IIRNS rpecr lr /PEUksIL INS ~ HRC Form CSEA'rl ))7) 1 0 s 0 0 0 3 1 5 8 7 0 0 8 0 0 4 oF 0 4 operation which took place approximately 15 minutes prior to the reactor trip actuation. In addition, the Operation Sequence Monitor (OSM) failed.to print the correct or accurate equipment actuation times, consequently a complete time study was not possible. Through secondary means, it was determined that the reactor trip breakers actuated in less than 100 ms which is consistent with previous data.

All other applicable actuations were also verified.

Corrective Actions Immediate corrective action involved Operations personnel implementing plant procedures to verify proper response of the automatic protection system and to assess plant conditions for initiating appropriate recovery actions. The automatic/manual level controller was removed, inspected, refurbished, and determined to be operable. The steam generator number 12 level recorder was repaired and returned to service. The escutcheon plate for the number 12 Steam Generator feedwater flow selector switch has been relabeled to identify the required 90 degree throw. In addition,,

a memo was issued to all operators describing lessons learned from this event. Corrective action has been taken on all attendent equipment malfunctions noted in the event analysis section.

Failed Com onent Identification No component failures were directly related to the cause of this event.

Previous Similar Events None.

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~ PIONl looerla INDIANA 8r MICHIGANELECTRIC COMPANY OONALO C. COOK NUCLEAR PLANT P.O. Box 456, Brldgemen, MI 49106 Telephone (616) 465.5901 July 2, 1987 United States Nuclear Regulatory Commission Document Control Desk Washington, D.C. 20555 Operating License DPR-58 Docket No. 50-315 Document Control Manager:

Zn accordance with the criteria established by 10 CFR 50.73 entitled Licensee Event Re ortin S stem, the following report is being submitted:

87-008-00 Sincerely, W. G. Smith, r.

Plant Manager WGS:afh Attachment cc: A. B. Davis, Region ZZZ J. E. Dolan M. P. Alexich R. F. Kroeger H. B. Brugger R. W. Jurgensen NRC Resident, Znspector D. L. Wigginton R. C. Callen, MPSC D. Hahn, MDPH G. Charnoff, Esq.

Dottie Sherman, ANZ Library ZNPO J. G. Feinstein/B.P. Lauzau PNSRC A. A. Blind File