ML17334B508

Transmits Preliminary Accident Sequence Precursor Program Analysis for Licensee Peer Review

ML17334B508
Person / Time
Site:	Cook
Issue date:	05/19/1994
From:	John Hickman Office of Nuclear Reactor Regulation
To:	Fitzpatrick E AMERICAN ELECTRIC POWER SERVICE CORP.
References
NUDOCS 9405310331
Download: ML17334B508 (35)

Text

FORD 2 REGULAT ZNFORMATZON DZSTRZBUTZOtZSTEM (RZDS)

ACCESSION NBR:9405310331 DOC.DATE: 94/05/19 NOTARIZED:

NO DOCKET FACIL:50-316 Donald C.

Cook Nuclear Power Plant, Unit 2, Indiana M

05000316 AUTH.NAME AUTHOR AFFILIATION HICKMANFJ.B.

Project Directorate III-3 RECIP.NAME RECIPIENT AFFILIATION FITZPATRICK,E.

American Electric Power Service Corp.

SUBJECT:

Transmits preliminary accident sequence presursor analysis of plant, Unit 2 for licensee peer review.

DISTRIBUTION CODE:

DF01D COPIES RECEIVED:LTR ENCL SIZE:

TITLE: Direct Flow Distribution:

50 Docket (PDR Avail)

NOTES:

D RECIPIENT COPIES ID CODE/NAME LTTR ENCL RECIPIENT COPIES ID CODE/NAME LTTR ENCL F

INTERNAL: NUDOCS-ABSTRACT EXTERNAL: NRC PDR 1

1 1

1 REG FILE 01 NSIC 1

1 1

1 0

D 2

D 0

NOTE TO ALL"RIDS" RECIPIENTS:

PLEASE HELP US TO REDUCE WASTE! CONTACT THE DOCUMENT CONTROL DESK, ROOM Pl-37 (EXT. 504-2065) TO ELIMINATEYOUR NAME FROM DISTRIBUTION LISTS FOR DOCUMENTS YOU DON'T NEED!

N TOTAL NUMBER OF COPIES REQUIRED:

LTTR 4

ENCL 4

Docket No. 50-316 Hay 19, 1994 Hr.

E.

E. Fitzpatrick, Vice President Indiana Michigan Power Company c/o American Electric Power Service Corporation 1 Riverside Plaza

Columbus, Ohio 43215 DISTRIBUTION Docket File NRC

& LPDRs PD3-1 Rdg JRoe JZwolinski

GHolahan, 6/H/3 JMinns 10/D/4 EJordan HNBB-3701 WKropp, RIII JHickman CJamerson ACRS (10)

OGC LHarsh

SUBJECT:

TRANSMITTAL OF PRELIMINARY ACCIDENT SEQUENCE PRECURSOR ANALYSIS OF COOK UNIT 2 EVENT FOR LICENSEE PEER REVIEW Enclosed is a copy of the preliminary Accident Sequence Precursor (ASP)

Program analysis of an operational event which occurred at the D.C.

Cook, Unit 2, plant on August 2, 1993 (Enclosure 1).

The preliminary results of our contractor's (ORNL) analysis of this event indicate that it may be a precursor event for 1993.

The purpose of this letter is to request your review of this analysis.

You are requested to review and comment on the technical adequacy of the

analyses, including the depiction of the plant equipment and equipment capabilities.

We will then evaluate the comments received during this Peer Review for reasonableness and pertinence to the ASP analysis in an attempt to use best-estimate values.

Upon completion of this evaluation, we will revise the conditional core damage probability calculations where necessary to consider information provided, during this review.

The object of the Peer Review processi's to provide as realistic an analysis of the significance of the event as possible.

In order to maintain our schedule for issuance of the 1993 Precursor

Report, you are requested to complete your review and provide your comments within 30 days from the date that you receive this letter.

In order to facilitate your review we have enclosed several items for guidance.

Enclosure 2:

(1) contains specific guidance for the Peer

Review, (2) identifies,-the criteria which we will apply to determine whether any credit should be given in the analysis for the use of licensee-identified additional equipment or specific actions in recovering from the event, and (3) describes the specific information that should be provided by the licensee to support such a claim.

Enclosure 3 is the licensee event report (LER) documenting the subject event.

Enclosures 4 and 5

contain background information regarding the ASP methodology which may be useful in reviewing the analysis.

Enclosure 4, which is Section 2.0 from the 1992 ASP Annual Precursor

Report, describes the precursor event identification and quantification process.

Enclosure 5, which is Appendix A from the same report, describes the ASP modes used in precursor analyses.

26K CBi.

9'405310331 9405i9 PDR ADOCN 05000316 P

'DR

1

)

E

1 I ~

~

~,

Mr. E.

E. Fitzpatrick I May 19, 1994 No new OMB clearance is needed for the ASP Peer Review process, since the process is already covered by the existing OMB clearance addressing followup review of events documented in LERs.

Should you have any questions, please contact John B. Hickman at (301) 504-3017.

Sincerely, Original Signed By:

John B. Hickman, Project Manager Project Directorate III-1 Division of Reactor Projects III/IV Office of Nuclear Reactor Regulation 3.

cc w/enclosures:

See next page

Enclosures:

1.

Preliminary ASP Analysis of LER 316/93-007.

2.

Guidance for Performing Peer Review of Preliminary Analysis and Criteria for Recovery Credit.

Licensee Event Report No. 316/93-007.

"Accident Sequence Precursor Identification and guantification,"

Section 2.0 of "Precursors to Potential Severe Core Damage Accidents:

1992 A Status

Report, NUREG/CR-4674, Volume 17.

5.

"ASP Models," Appendix A to NUREG/CR-4674, Volume 17.

LA:PD31 CJamerson P

3 J

'ckman: ll l

PD31 BMarsh 05/[ l/94 05//5 94 05 l094 OFFICIAL RECORD COPY

FILENAME: G: iWPDOCSiDCCOOKiCO-ASP. LTR

'i 4(

, ~

l d

)f I

i

'I II

\\ p I

UNITED STATES NUCLEAR REGULATORY COMIVIISSION WASHINGTON, D.C. 20555-0001 Docket No. 50-316 May 19; 1994 Hr.

E.

E. Fitzpatrick, Vice President Indiana Hichigan Power Company c/o American Electric Power Service Corporation 1 Riverside Plaza

Columbus, Ohio 43215

SUBJECT:

TRANSMITTAL OF PRELIMINARY ACCIDENT SE(UENCE PRECURSOR ANALYSIS OF COOK UNIT 2 EVENT FOR LICENSEE PEER REVIEW Enclosed is a copy of the preliminary Accident Sequence Precursor (ASP) Program analysis of an operational event which occurred at the D.C.

Cook, Unit 2, plant on August 2, 1993 (Enclosure 1).

The preliminary results of our contractor's (ORNL) analysis of this event indicate that it may be a precursor event for 1993.

The purpose of this letter is to request your review of this analysis.

You are requested to review and comment on the technical adequacy of the

analyses, including the depiction of the plant equipment and equipment capabilities.

We will then evaluate the comments received during this Peer Review for reasonableness and pertinence to the ASP analysis in an attempt to use best-estimate values.

Upon completion of this evaluation, we will revise the conditional core damage probability calculations where necessary to consider information provided during this review.

The object of the Peer Review process is to provide as realistic an analysis of the significance of the event as possible.

In order to maintain our schedule for issuance of the 1993 Precursor

Report, you are requested to complete your review and provide your comments within 30 days from the date that you receive this letter, In order to facilitate your review we have enclosed several items for guidance.

Enclosure 2:

(1) contains specific guidance for the Peer

Review, (2) identifies the criteria which we will apply to determine whether any credit should be given in the analysis for the use of licensee-identified additional equipment or specific actions in recovering from the event, and (3) describes the specific information that should be provided by the licensee to support such a claim.

Enclosure 3 is the licensee event report (LER) documenting the subject event.

Enclosures 4 and 5

contain background information regarding the ASP methodology which may be useful in reviewing the analysis.

Enclosure 4, which is Section 2.0 from the 1992 ASP Annual Precursor

Report, describes the precuI'sor event identification and quantification process.

Enclosure 5, which is Appendix A from the same report, describes the ASP iII'odes used in precursor analyses.

g

Hr.

E.

E. Fitzpatrick Hay 19, 1994 No new OHB clearance is needed for the ASP Peer Review process, since the process is already covered by the existing OHB clearance addressing followup review of events documented in LERs.

Should you have any questions, please contact John B. Hickman at (301) 504-3017.

Sincerely, John B. Hickman, Project Hanager Project Directorate III-1 Division of Reactor Projects III/IV Office of Nuclear Reactor Regulation 3.

4, cc w/enclosures:

See next page

Enclosures:

1.

Preliminary ASP Analysis of LER 316/93-007.

2.

Guidance for Performing Peer Review of Preliminary Analysis and Criteria for Recovery Credit.

Licensee Event Report No. 316/93-007.

"Accident Sequence Precursor Identification and guantification,"

Section 2.0 of "Precursors to Potential Severe Core Damage Accidents:

1992 - A Status

Report, NUREG/CR-4674, Volume 17.

5.

"ASP Hodels," Appendix A to NUREG/CR-4674, Volume 17.

Mr.

E.

E. Fitzpatrick Indiana Michigan Power Company CC:

Regional Administrator, Region III U.S. Nuclear Regulatory Commission 801 Warrenville Road Lisle, Illinois. 60532-4351 Attorney General Department of Attorney General 525 West Ottawa Street

Lansing, Michigan 48913 Township Supervisor Lake Township Hall

,Post Office Box 818

Bridgman, Michigan 49106 Al Blind, Plant Manager Donald C.

Cook Nuclear Plant Post Office Box 458

Bridgman, Michigan 49106 U.S. Nuclear Regulatory Commission Resident Inspector Office 7700 Red Arrow Highway Stevensville, Michigan 49127 Gerald Charnoff, Esquire

Shaw, Pittman, Potts and Trowbridge 2300 N Street, N.

W.

Washington, DC 20037 Mayor, City of Bridgman Post Office Box 366

Bridgman, Michigan 49106 Special Assistant to the Governor Room 1 - State Capitol

Lansing, Michigan 48909 Nuclear Facilities and Environmental Monitoring Section Office Division of Radiological Health Department of Public Health 3423 N.

Logan Street P. 0.

Box 30195

Lansing, Michigan 48909 Donald C.

Cook Nuclear Plant Hr. S.

Brewer American Electric Power Service Corporation 1 Riverside Plaza

Columbus, Ohio 43215 Dcccmber 1993

ENCLOSURE 1

9gp531p>31

I

-W I

PRELIMINARY 0.1 LER Number 316/93-007 Event

Description:

'eactor Trip with Degraded AFW Date ofEvent:

August 2, 1993 Plant:

Cook 2 0.1.1 Summary Cook 2 tripped &om 70% power because ofa spurious main turbine exhaust hood high-temperature signal.

The auxiliary feedwater (AFW)control valves for the east motor-driven AFW pump throttled further than expected, requiring operator action to restore proper flow&om that pump. Operator action was also required to reopen the main steam isolation valves (MSIVs) that drifted closed following the trip. The conditional core damage probability estimated for the event is 6.4

>< 10~.

0.1.2 Event Description On August 2, 1993, Cook 2 tripped from 70% power following a turbine trip caused by spurious actuation of the turbine exhaust hood high-temperature switches.

Eight ofthe nine switch actuation setpoints were found to be significantly lower than the as-left condition recorded during the last calibration one year earlier.

Following the reactor trip, the auxiliary feedwater pumps started and provided flowto the steam generators.

The feedwater control valves from the east motor-driven AFW pump throttled further than expected after receiving a flowretention signal, and operator action was required to maintain correct flowrates. AFWflow switches were subsequently recalibrated and flowretention valve intermediate positions were reset to correct the problem.

The MSIVs, which started to driftclosed followingthe reactor trip, v ere reopened by the operators.

The utilitystated that this driftwas expected followinga trip because ofthe valve actuator design.

0.1.3 Additional Event-Related Information Cook 2 has three AFW pumps; two are motor-driven, and one is turbine-driven. The turbine-driven pump provides flonto all four steam generators (SGs), and each motor-driven pump provides flowto two SGs.

,. Flow retention valves coiitrol the fiowfrom each pump to each SG; these valves can be controlled from the control room. A cross-connect exists that can provide flowfrom one motor-driven pump to the other unit.

The cross-connect valves are manual and normally locked closed.

The main feedwater (MFW) pumps are turbine driven.

PRELIMINARY LER NO: 316/93-007

E I;~

I r ~~)

0

PRELIMINARY 0.1.4 Modeling Assumptions

~

~

~

This event was modeled as a reactor trip with degraded AFWand MFW. To reflect the reduced flowfrom the east motor-driven AFW pump, one ofthe three AFW trains was assumed to be failed in the analysis.

Consistent with other precursor analyses, the probability of not recovering the potentially failed A'FW system was not revised since failures were not observed in the other two trains.

The ASP model for MFWassumes MFW is isolated followinga trip but is potentially available in the event ofa failure ofAFW. Ifthe operators had not promptly responded and reopened the drifting-closed MSIVs, steam to the MFWpumps would have been lost, rendering MFWunavailable.

Since this recovery action could be performed in the control room and the observed MSIVresponse was apparently not unusual at Cook, a nonrecovery probability of0.04 was added to the nominal MFWnonrecovery probability (0.07) to estimate the overall MFWnonrecovery used in the analysis (0.11). The nonrecovery probabilities used in ASP analyses are described in Section A.3.2 ofNUREGICR-4674, Volume 17 (Precursors to Potential Se>'ere Core Damage Accidents: 1992, A Status Report).

The potential use ofthe locked-closed cross-connect between both units'FW systems was not addressed in the analysis.

0.1.5 Analysis Results The conditional core damage probability estimated for this event is 6.4 < 10~. The dominant core damage sequence, highlighted on the event tree in Fig. 1, involves a postulated failure ofAFW and MFW following the trip and subsequent failure offeed and bleed cooling.

PRELIMINARY LER NO: 316/93-007

I

PRELIMINARY PORVI PORVI TRAHS RT AFW MFW SRV SRV HPI HPR CHAL RESEAT OPEN NO NPNPN OK CD 12 CD OK OK 13 CD CD OK OK 15 CD CD CD 15 ATWS

~

'ig.

I. Dominant core damage sequences for LER 316/93-007.

PRELIMINARY LER No: 316/93-007

l

~

PRELIMINARY CONDITIONAL CORE DAHAGE PROBABILITY CALCULATIOHS Event Ident ifier:

316/93-007 Event

Description:

Reactor trip uith degraded AFM Event Date:

08/02/93 Plant:

Cook 2 INITIATINGEVENT NONRECOVERABLE INITIATINGEVENT PROBABILITIES TRANS 1.DE+00 SEOUEHCE CONDITIONAL PROBABILITY SUHS End State/Ini t later Probability TRANS 6.4E.06 Total 6.4E 06 TRANS 3.4E.OS Total SEQVEHCE CONDITIONAL PROBABILITIES (PROBABILITY ORDER) 3.4E.05 Sequence End State Prob H Rec'*

13 trans -ft AFM NFM -hpi(f/b> -hpr/-hpi porv.open 17 trans -rt AFM I(FM hpi(f/b) 16 trans -rt AFM HFM -hpi(f/b) hpr/-hpi CD CD CD 4.6E-06 1.6E-06 1.7E-07 2.9E-02 2.4E-02 2.9E.02 18 trans rt

~~ nonrecovery credit for edited case ATMS 3.4E.OS 1.2E-01 SEQVEHCE CONDITIONAL PROBABILITIES (SEOUEHCE ORDER)

Sequence End State Prob N Rec+*

15 trans -rt AFM HFM -hpi(f/b) -hpr/-hpi porv.open 16 trans -rt AFM NFM -hpi(f/b) hpr/-hpi 17 trans -rt AFM MFM hpi(f/b) 18 trans rt

~ non-recovery credit for edited case CD CD CD AIMS 4.6E 06 1.7E.07 1.6E-06 3.4E.OS 2.9E-02 2.9E-02 2.4E 02 1.2E-01 SEOVENCE teDELt ctiaspi1989ipwrbseai.cmp BRANCH HCOELt c: iaspi1989icook. sl1 PRDBABILITY FII.E:

c:iaspi1989lpwr bsl1.pro Conrecovery Limit BRANCH FREQVENCIES/PROBABILITIES Event Ident ifier: 316/93.007 PRELI XIIXARY LER NO: 316/93-007

~

PRELIMINARY Branch trans loop loca rt rt/loop emerg.power AFII Branch Hodel:

1.OF.3+ser Train 1

Cond Prob:

Train 2 Cond Prob:

Train 3 Cond Prob:

Serial Component Prob:

afw/emerg.power HFW Branch Hodel:

1.0'+opr Train 1

Cond Prob:

porv.or.srv.chall porv.or.srv.reseat porv.or.srv.reseat/emerg.power seat. loca ep.rec<st) ep.rec hpi hpi(f/b) hpr/-hpi porvr open branch codet file

~ ~ forced System 3.4E-04 1.6E-05 2.4E.06 2.8E-04 O.OE+00 2.9E-03 3.8E-04 i 5.3E-03 2.0E.02 i Fatted 1.0E.01 5.0E-02 2.8E-04 5.0E.02 1.0E+00 > 1.0E+00 1.0E+00 4.0E-02 3.0E.02 3.0E 02 2.5E.01 6.9E-01 5.2E 02 1.0E-03 1.0E.03

'1.5E.04 3.0E.O2 HonrRecov 1.0E+00 2.4E.01 4.3E-01 1.2E 01 1.0E+00 8.0E.01 2.6E-01 3.4E 01 7.0E-02

> 1.1E 01 1.0E+00 1.1E.02 1 AL OE+00 1.0E+00 1.0E+00 1.0E+00 8.4E 01 8.4E.01 T.i&00 1.BE+ 00 Opr Fail 1.0E-O3

).QE-02 T.OE-03 4r.OE 04 Event Identifier: 316/93-007 PRELI~IINARY 5

LER No: 316/93-007

'I J

t t

r

ENCLOSURE 2

I

GUIDANCE FOR LICENSEE PEER REVIEM OF PRELIMINARY ASP ANALYSIS Back round The preliminary precursor analysis of an operational event which occurred at your plant has been provided for your review.

This analysis was performed as a part of the'RC's Accident Sequence Precursor (ASP) Program.

The ASP Program uses probabilistic risk assessment techniques to provide estimates of operating event significance in terms of the potential for core damage.

The types of events evaluated include loss of off-site power (LOOP), Loss-of-Coolant Accident (LOCA), degradation of plant conditions, and safety equipment failures or unavailabilities that could increase the probability of core damage from postulated accident sequences.

This preliminary analysis was conducted using the information contained in the plant-specific final safety analysis report (FSAR), individual plant examination (IPE),

and the licensee event report (LER) for this event.

These sources are identified in the write-up documenting the analysis.

The analysis methodology followed the process described in Section

2. 1 and Appendix A of Volume l7 of NUREG/CR-4674, copies of which have been provided in this package for your use in this review.

Guidance for Peer Review and Criteria for Recover Credit The review of the preliminary analysis should use Section

2. 1 and Appendix A of NUREG/CR-4674 for guidance.

Comments regarding the analysis should address:

~

Characterization of possible plant response,

~

Representation of expected plant response used in the analytical

models,

~

Representation of plant safety equipment configuration and capabilities at the time of the event, and

~

Assumptions regarding equipment recovery probabilities.

If you desire credit for plant features or recovery measures that were not considered in our preliminary analysis of this event (e.g.,

the use of additional

systems, equipment, or specific actions),

your request must be supported by appropriate, documented information that will allow us to reanalyze the event in the light of the information you provide.

The identified plant features or recovery measures must have existed at the time of the event, and should include:

Normal or emergency operating procedures, Piping and instrumentation diagrams (P&IOs),

Electrical one-line diagrams, Results of thermal-hydraulic analysis, bperator-training (both procedures and simulator), etc.

Plant-specific system reliability - supporting information should include the basis for the stated reliability value (method of determining the system's reliability, available data used in determination, etc.)

Also, the documentation should address the impact of the use of the specific recovery measure on:

The sequence of events, The timing of events, The probability of operator error in using the system or equipment, and Other, systems/processes already modeled in the analysis.

For example, Plant A (a PWR) experiences a reactor trip and, during the subsequent recovery, it is discovered that one train of the auxiliary

, feedwater (AFW) system is unavailable.

Absent any further information regrading this event, the ASP Program would analyze it as a reactor trip with one train of AFW unavailable.

The AFW train modeling would be patterned after information gathered either from the plant PSAR or the IPE.

However, if information is received about the use of an additional system (such as a standby steam generator feedwater system) in recovering from this event, the transient would be modeled as a reactor trip with one train of AFW unavailable, but this unavailability would be mitigated by the use of the standby feedwater system.

The mitigation effect for the standby feedwater system would be credited in the analysis provided that the standby feedwater system characteristics are documented in the

FSAR, accounted for in the IPE, procedures for using the system during recovery existed at the time of the event, the plant operators had been trained in the use of the system prior to the event, a clear diagram (one-line diagram or better) of the system is available, previous analyses have indicated that there would be sufficient time available to implement the procedure successfully, and results of an assessment that evaluates the effect that use of the standby feedwater system has on already existing processes of procedures that would normally be used to deal with the event are available.

Materials Provided for Review The following materials have been provided in the package to facilitate your review of the preliminary analysis of the operational event:

The specific licensee event report (LER), augmented inspection team AIT) report, or other pertinent reports as appropriate (separate enclosure).

A calculation summary "sheet indicating the-dominant sequences and pertinent aspects of the modeling details (contained in the analysis writeup).

An event tree with the dominant sequence(s) highlighted (contained in the analysis writeup).

A copy of Section

2. I and Appendix A of NUREG/CR-4674, Volume I7 (separate enclosures).

ENCLOSURE 3

~ 4'

~

ac v

HRCFORM366 ieebi UA. NUCLfAilREGULATORY COMMISSION LICENSEE EVENT REPORT (LER)

APPROVED OMS NO. l150010e EXPIRf5'(30/52 fSTIMATED SUROEN PER

RESPONSE

TO COMPLY WTH THIS INFQAMATIQH COLLECTION REQUEST 500 IIRS, FORwARD COMMENTS REGARDING SVRDEH ESTIMATE TO THE AECORDS AHD REPORTS MANAGEMENTSRANCH IP030). U.S NUCLfAR RfGULATORY COMMISSION, WASHINGTON. DC 20555. ANO TO THE PAPERWORK REDUCTION PROJECT 13150010e).

OFFICE OF MANAGEMENTAND SUDGET,WASHINGTON. OC 20503.

DOtNALD C.

COOK NUCLEAR PLANT - UNIT 2 OOCKKT NVMSER 12) 0 5

0 0

0 3

~A 5

I 1

QF 0

4 TITLE I~ I REACTOR TRIP FROM SPURIOUS TURBINE EXHAUST HOOD HIGH TEMPERATURE TRIP EVENT DATE 151 MQHTtt OAY YEAR YEAR LER NUMbER 16)

SLQVKHTtAL HVMSER A E PORT OATE 17)

OAY YEAR ACVS~ MQHTH

~tuttOCA OOCKKT HVMSERISI 0

5 0

0 0

FACILITYNAMES OTHER FACILITIES IHVOI.VEDISI 802 93 93 007 0 009 01 9

3 0

5 0

0 0

OPKAATIHC MOOS I~ I POWER LEYEL 0

7 0 20.e02(b) 20 e051e)(II(il 20.e05 (e )(I)(ii) lb.e05I ~ )II)itii) 20 4)5( ~l(lIlit) 20.e05( ~ II1 IIrI 20.4)5(c) 50M(ci())

50.$ 6 (c) 121 60.7liellllli) 60.7$ (elllllil lb.734 I(2 I(riil LICENSEE CONTACT FOR THIS LER (121 50.7 34)(2 l lit) 50,7 34) 12)4) 50.7$ (ellll(rii1 lb.7341(2) Irii i)

(AI 50.7$ (e I(2)lriiiI I~ I 50,73(el 12 )(~ I THIS REPOAT 1$ SUSMITTED PURSUANT 70 THE RLQUIREAIENTs oF '10 cFR II; Ice@co oor ot torte ot tor totto~iotl (11 7$.7)(~ I 7$.7) Ic)

OTHEII ISorcittt ie Aotttrct ortow rtrt to Fret, iYACForte 3FFAI G. A. VEBER PLANT ENGINEERING SUPERINTENDENT TELEPHONE HVMSER AREA CODE 6

164 65-590 1

COMPLKTE OHE LINE FOR EACH COMPONENT FAILURE DESCAISED IN THIS AEPORT I'Ill CAUSE SY$ 7EV COVPOHEttT

MAHUFAC, TVREA EPOATASLE TQ HPRQ$

,e CAVSE SYSTIKM COMPOHEHT MAHVFAC TVREA EPORTASLK TO NPRDS X

T A T

S M

2 3

5 I

r

"~t 5~~'t:gv~crJ'x$

etQt+0 '+Sg H

SUPPLEMENTAL REPORT EXPECTED Ilei

>ES tH rt co o ~ rr tttrfctED Sttdet'$$ tott DATft X

HO ASSTA*CT

~ ~

t tc 't$0 tottrt

~ r rooter ~rtr ~ Afttre tiet ~ toter trot~otter Itrtt 11 ~ I EX~ECTED SVSMISSIOPI DATE (151 erOtrTtt DAY YEAR On A gust 2,

1993, at 1226 hours0.0142 days <br />0.341 hours <br />0.00203 weeks <br />4.66493e-4 months <br />, the Unit 2 reactor tripped as a result of a main turbine trip, caused by a spurious actuation of the Exhaust Hood High Temperature Trip Switches.

Investigation revealed that eight of nine Exhaust Hood High Temperature Trip Switch setpoints were found to be significantly below the normal trip setpoint.

Investigation of the event determined that the method used to calibrate the switches may have caused the setpoint to be misadjusted, and that vibration can cause a downward shift in the setpoint.

These factors, combined with a slight increase in hood temperatures and vib ation levels which resulted from the removal ~f a main condenser half from

service, are believed to have caused the spurious trip.

To prevent recurrence, the Main Turbine Exhaust Hood high tern eratu trip was disabled.

Wri i

uc ions or a manua ur ne trip on receipt')~

ie Main Turbine Exhaust Hood Extreme High Temperature Alarm have replaced the defeated automatic trip.

ppt 4 g g//

~i HRC Forte 366 (6ebi

MRC fORM 056A 1645 I ILS, NUCLEAR REGULATORYCOMMISSION LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION AppRQYED QMs No.s150410e EXPIRES'. Alt0ISt ESTIMATED SURDEN PER

RESPONSE

TO COMPLY WTN THIS IHfORMATION COLLECTION REQUEST: SOAl NRS. fORWARD COMMENTS REGARDIHG SURDEN ESTIMATE TO TNE RECORDS AND REPORTS MANAGEMENTSRANCN IP4501. U.S. NUCLEAR REGUI ATORY COMMISSION. WASHINGTON. DC 20555, AHD TO TNE PAPERWORK REDUCTION PROJECT ISI50010ll, OFFICE OF MANAGEMENTAND SUDGET,WASHINGTON, DC t050 CILITY NAME 111 DOCKET NUMSER Itl V5 AII LER NUMSER ISI

>eIe SEQUENTIAL NUMPTA SSI ASYI$IQH

A

~ HUM f A

~AGE IS)

D.

C.

COOK NUCLEAR PLhhT - UNIT 2 TEXT W mme epece r eeeeeed, we~ AFC'nn SSSA'el st 1 0

5 0

0.

0 3

]

6 9

3 0 0 7

0 0 0 20F 0

Conditions Prior to Occurre ce Unit 2 was operating in Mode 1 at 70.5 percent of Rated Thermal Power.

Condenser 'B'ash water box had been removed from service within the previous 15 minutes.

Descri tion of Event On August 2, 1993, at 1226 hours0.0142 days <br />0.341 hours <br />0.00203 weeks <br />4.66493e-4 months <br />, the Unit 2 reactor (EIIS/JE) tripped as a

result of a Main Turbine (EIIS/TA) Exhaust Hood high temperature trip.

Following the turbine/reactor trip sequence, (turbine (EIIS/TA-TRB) trip, opening of the reactor trip breakers (EIIS/JE-BKR), insertion of reactor control rods (EIIS/BA-P), and automatic start of the auxiliary feedwater pumps (EIIS/BA-P)), Operations personnel immediately implemented Emergency Operating Procedure 2

OHP 4023.E-O to verify proper response of the automatic protection systems and to assess plant conditions for appropriate recovery actions.

Abnormalities noted during the event

'ncludeds Feedwater valves (EIIS/BA-FCV) from the East Motor Driven Auxiliary Feed Pump (EIIS/BA) throttled further than expected after receiving a flow retention signal, requiring operator action to maintain correct flow rates.

The Auxiliary Feedwater (AFM) flows from the other motor-driven and turbine-driven pumps were not affected and delivered flow in excess of that required for safety analysis concerns.

Flow switches were subsequently recalibrated and flow retention intermediate valve positions were reset.

Main Steam Isolation Valves (EIIS/SB-ISV) started drifting closed following the reactor trip.

The valves were promptly reopened.

A review of several past trip reports indicates that this is not unusual and is an expected consequence following a trip due to actuator design.

No corrective actions are planned.

Cause of Event The turbine trip was initiated by a spurious actuation of the Turbine Exhaust Hood High Temperature Switches.

Eight of nine switch actuation setpoints were found to be significantly lower than as-left condition recorded in August-.

1992 when they SRE"e last calibrated.

The investigaticn of this event found that the calibration accuracy is affected by the ability to F.sition the switch for bench calibration precisely as it will be positioned in the field.

Any difference will affect the accuracy of the calibration; Calibration accuracy is also susceptible'o the method by which heat is applied to the switch sensing element.

The method used in the previous calibration (app)ication of heat using

a. heat gun) may

. have allowed a difference to exist between the temperature sensed by the HRC fann 566 A (665 1

d

~4 W

NOMIC FORM 255A (6JIQ)

ILITYNAME 111 US. NUCLEAR REGULATORYCOMMISSION LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION DOCKET NUMSER ul APPROVED OMS NO. 21500104 EXPIRES: AJ20/02 ESTIMATED EURDEN PEll RESPONSE TO COMPLY WTH THIS INFORMATION COLLECTION REOUESTI 600 NRS. FORWARD COMMENTS REGARDIHG 4URDEN ESTIMATE TO THE RECORDS AND REPORTS MANAGEMENTERAHCH (F4201. U.S, NUCLEAR REGULATORY COMMISSION, WASHIHQTOH, DC 20555, ANDTO 11IE PAPERWORK REDUCTION PROJECT 1215001041, OFFICE OF MANAGEMENTAHD SUDGET,WASHINGTON, DC 20502.

YEAR LER NUMSER 161

$50VEHTIAL gg.

Ivvvpfr:z~ AEVI5KIH NVV FA PAGE IS)

D. C.

COOK hVCLEAR PLANT - UNIT 2 TEXT Illmon Herlr~ VPP~AtRC %%dnn 2554211171 o

5 o

o o

31 693 0

0 7

00 03 OF 0 4

Cause of Event Co t'ensing element of the switch and that sensed by the calibration standard.

It was also demonstrated by test that vibration can cause the switch to actuate below its setpoint.

Just 15 minutes prior to the event, cooling water to the "B" North Low Pressure Turbine (LPT) Condenser half was isolated to permit inspection for tube leakage.

Removal of a condenser half from service has the effect of increasing hood temperature on the associated LPT and can also increase vibration levels.

Although slightly elevated, both vibration and hood temperatures remained well within operating limits.

However, the slight increase in these parameters, combined with the lower than normal as-found switch setpoints and the tendency of the switches to actuate prematurely when subjected to vibration, is believed to have caused the spurious trip.

Anal sis of Event This report is being submitted in accordance with 10 CFR 50.73, paragraph (a) (2) (i'v), as an event that resulted in an unplanned automatic actuation of the Engineered Safety Features, including the Reactor Protection System.

The automatic protection responses, including reactor trip and its associated actuations were verified to have functioned properly as a result of the reactor trip signal.

Feedwater valves from the East Motor Driven Auxiliary Feed

Pump, which throttled further than expected, were under the control of the reactor operator, and readjusted as required in accordance with the reactor trip response procedure (E-0).

The main steam isolation valves, which started drifting closed, were reopened promptly.

Based on the above, it is concluded that the event did not involve an unreviewed safety question as defined in 10 CFR 50.59(a)(2) nor did it adversely impact the health and safety of the public.

Corrective Actions A review by AEPSC and ABB personnel determined that the Turbine Exhaust Hood

(

high temperature trip served no safety-related function.

The trip had been originally installed as a means of tripping the turbine in the event of generator motoring, to prevent damage to the turbine generator.

Following this review, the automatic main turbine trip from high exhaust hood temperature was disabled and replaced with instructions for a manual main turbine trip.

On receipt of a Main Turbine Exhaust Hood Extreme High Temperature Alarm, and after verifying the extreme high temperature condition per the revised annunciator response procedure, the operator will trip the turbine.

The calibration method for the Turbine Exhaust Hood high temperature trip has been modified to use a water bath for heat application to provide assurance of a uniform heat medium.

HRC F ena 266A ISEE I

NRC F~ORM S66A I6891 ILS NUCLEAR REGVLATORYCOMMISSION LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION AFPROV EO OM6 NO. SI500104 EXPIRES: >>ISOI92 ESTIMATED SVRDEN PER hESPONSE TO COMPI.Y WTH THIS INPORMATION COLLECTION REOVESTI 500 HRS. FORWARD COMMENTS REGARDING SURDEN ESTIMATE TO%HE RECORDS

• ND REPORTS MANAGEMENTSRANCH IP820I, U.S. NUCLEAR REGULATORY COMMISSION. WASHINGTON. OC 20555, ANDTO 1HE PAPERWORK REDUCTION PROJECT 12150010II. OFFICE OF MANAGEMENTAND SVDGET.WASHINGTON,DC20502.

ACILITYNAME III DOCKET NVMSER Ql YEAR LER NUMSER (61 p~ SEOVENT>AL HUM TA

~.ip AiV<<klN re,'QQM

~AGE ul D.

C.

COOK NUCLEAR PLANT - UNIT 2 I

TEKT IN>>H>>>> <<<<C>> << I>>>>>>ied. I>>>>>>>>A>>>>n>>RC hem %562I I Ill 0

s 0

0 0

31 69 3 0 0

7 0 0 0 4 oF 0 4 Corrective Actions Cont'd Balance weights were added to the main turbine during unit startup to reduce vibration levels.

The East Motor Driven Auxiliary Feed Pump flow switches were recalibrated and flow retention intermediate valve positions were reset.

a led Com onen Zd t'cat'o Plant Designation:

Manufacturer1 ModelT EZZS Code:

Low Pressure Turbine Exhaust Hood Temperature Switch Thermal Sensors Mercoid Corp.

DA-37-804-6 EZZS/TA-TS Previous Silllilar Events None

<<RC F>>>>>> 266A I689l

0 r J