ML17329A690

LER 91-007-01:on 910802,flowpath Identified Which Could Result in Diversion of Water Away from ECCS & Containment Bldg.Caused by Erroneous Steps in Eop.Eop Revised W/ Assistance of Westinghouse.W/921124 Ltr

ML17329A690
Person / Time
Site:	Cook
Issue date:	11/24/1992
From:	Blind A, Sampson J INDIANA MICHIGAN POWER CO. (FORMERLY INDIANA & MICHIG
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
LER-91-007-01, LER-91-7-1, NUDOCS 9212010289
Download: ML17329A690 (9)

Text

ACCeCE<<'i'SU 0 'TMOUTION DEMON TRACTION SYSTEM l

REGULATO INFORMATION DISTRIBUTION TEM (RIDS)

. ACCESSION NBR:9212010289 DOC.DATE: 92/11/24 NOTARIZED: NO DOCKET FACIL:5fI-316 Donald C. Cook Nuclear Power Plant, Unit 2, Indiana M 05000316 AUTH. NAME AUTHOR AFFILIATION SAMPSON,J.R. Indiana Michigan Power Co. (formerly Indiana 8 Michigan Ele BLIND<A.A. Indiana Michigan Power Co. (formerly Indiana 6 Michigan Ele RECIP.NAME RECIPIENT AFFILIATION

SUBJECT:

LER 91-007-01:on 910802,identified flowpath that diverted ECCS flow caused by plant design/emergency response guidelines. Reviewed plant specific EOPs to ensure that procedures are provided to plant operators.W/921124 ltr.

DISTRIBUTION CODE: ZE22T COPIES RECEIVED:LTR ( ENCL TITLE: 50.73/50.9 Licensee Event Report (LER), Incident Rpt, etc. U SIZE:

NOTES:

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PLEASE HELP US TO REDUCE WASTE! CONTACT THE DOCUMENT CONTROL DESK.

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Indiana Michigan Power Company Cook Nuclear Piani One Cook Place Bridgman, Mi 49106 616 465 5901 lNOlANA NlCHlGAN November 24, 1992 POWM United States Nuclear Regulatory Commission Document Control Desk Rockville, Maryland 20852 Operating Licenses DPR-74 Docket No. 50-316 Document Control Manager:

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

91-007-01 Sincerely, A. A. Blind Plant Manager

/sb Attachment c: D. H. Williams, Jr.

A. B. Davis, Region E. E. Fitzpatrick III P. A. Barrett R. F. Kroeger B. Walters Ft. Wayne NRC Resident Inspector W. M. Dean NRC J. G. Keppler M. R. Padgett G. Charnoff, Esq.

D. Hahn INPO S. J. Brewer B. A. Svensson j

9212010289 921124 PDR ADOCK 05000316 PDR iong i

1' 8

NRC FORM 355 (54)9) APPROVED OMB NO. 3(504104 EXPIRES: 4/30/92 ESTIMATED BURDEN PER RESPONSE TO COMPLY WTH THIS INFORMATION COLLECTION REOUEST! 50.0 HRS. FORWARD LICENSEE EVENT REPORT (LER) COMMENTS REGARDING BURDEN ESTIMATE TO THE RECORDS AND REPORTS MANAGEMENT BRANCH IP.530), U.S. NUCLEAR REGULATORY COMMISSION. WASHINGTON, DC 20555, AND TO 1

THE PAPERWORK REDUCTION PROJECT (31500104I. OFFICE OF MANAGEMENTAND BUDGET, WASHINGTON, DC 20503.

FACILITY NAME Ill DOCKET NVMBER (2) PA E 3 D. C. Cook Nuclear Plant Unit 2 o s o o o31 6ioFO Simulator Scenario Identified Flowpath that Diverted ECCS Flow Caused b Plant Design/Emergency Response Guidelines EVENT DATE (51 LER NUMBKR IB) REPORT DATE (7) OTHER FACILITIES INVOLVED ( ~ I MONTH DAY YEAR YEAR SNOUSNTIAL RKVrSION FACILITYNAMES NUMSER ?err NUMB5 tl MONTH DAY YEAR DOCKET NUMBER(SI D. C. Cook Unit 1 o s o o o3 15 0 8 0 2 9 1 9 1 0 0 7 0 1 112 49? 0 5 0 0 0 THIS REPORT IS SUBMITTED PURSUANT T 0 THE RNOUIREMENTB oF 10 cF R (It (choco ono or moro oi tno fortowins) (11 OPERATING MODE (4) 20A02(b) 20.405(cl 50.73lol CI)(ir) 73.71(III POWER 20AOB ( ~ I (I I 0) 50.35(cl(II 50,73( ~ )(2)lrl 73.71(c)

LEvEL 0 0 0 20AOB(rill)INI 50.35 (cl (2 I 50.73(ol(2)(rN) OTHER ISpocity in Abrtroct below onyin Toot. HRC Form 20.405 le) (1)(IN) 50.734) (2)(I) 50.73(o) (21(riN)(AI SEEAI 20A054) Ill(ir) 5023(ol(2)(NI 50.73(oN1(lriN)(BI 20AOB (o I I I ) lrl 50.'73(o I (2)(NII 50.73( ~ l(2((cl LICENSEE CONTACT FOR THIS LER (12I NAME TELEPHONE NUMBER J. R. Sampson, Operations Superintendent AREA CODE 6 16 465 -59 01 COMPLETK ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT (13)

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NO ABSTRACT ILimit to teOO ooocor, ie., ooprooimMoly !if!ten oinole tpoco typewritten Iinai (15)

This update is being submitted to present the review conducted by the Westinghouse Corporation.

On 8-2-91, a "small break loss of coolant accident" run on the Plant simulator identified a flowpath which may be established by plant conditions and the Emergency Operating Procedures (EOPs) with the potential to divert water away from the emergency core cooling system and the containment building. The flowpath was from the Centrifugal Charging Pump discharge through an leakoff valve, through the seal return line safety valve to the volumeemergency control tank (VCT), and through the VCT safety valve to the chemical and volume control system holdup tanks.

A review of the finding concluded that the amount of water which could be diverted did not significantly affect core cooling. However, the diverted water would be discharged from the containment building. Analysis of the potential dose rate from the diverted water to the whole body at the site boundary was calculated to be insignificant compared to the 10CFR100 accident dose limit and even within the 10CFR20.105 dose limits for unrestricted areas during normal operations. The condition identified has the potential to be a generic issue for Westinghouse-designed plants. The appropriate EOP was revised to address the simulator findings.

NRC Form 345 (54)9)

NRC FORM 366A U.S. NUCLEAR AEGULATOAYCOMMISSION (6 J)9) APPROVED OMB NO. 3)500106 EXPIRES: SI30l92 LICENSEE EVENT REPORT ILER) ESTIMATED BURDEN PER RESPONSE TO COMPLY WTH THIS INFORMATION COLLECTION REQUEST: 50.0 HRS. FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE RECORDS TEXT CONTINUATION AND AEPOATS MANAGEMENT BRANCH IP 530). U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON. DC 20555, AND TO THE PAPERWORK REDUCTION PROJECT (3)50s)loa). OFFICE OF MANAGEMENTAND BUDGET, WASHINGTON. DC 20503.

FACILITY NAME (1) DOCKET NUMBER l21 LER NUMBER (6) I'AGE l3)

YEAR SEOUENTIAL REVISION NUMOER NUMEER D. C. Cook Nuclear Plant Unit 2 o 5 o o o 316 91 007 01 02 oF0 6 TEXT IIImoIe Space IS scow'rNS II>> ed(so'opal HRC FomI 36SA's) (17)

Conditions Prior To Occu rence Unit one (U-1) in mode one at 100X power.

Unit two (U-2) in mode three (hot standby).

Descri tion of Event On 8-2-91, while the Training Department was running a small break loss of coolant accident (LOCA) on the Plant simulator, a flowpath was identified which could result in the diversion of water away from the emergency core cooling system (ECCS) (EIIS/BQ) and the containment building (EIIS/NH).

Description of systems (reference Figure 1):

Upon receipt of a safety injection signal (EIIS/JE-ACO), the safety injection system (SIS) centrifugal charging pump (CCP) (EIIS/BQ-P) suction transfers from the volume control tank (VCT) (EIIS/CB-TK) to the refueling water storage tank (RWST) (EIIS/BP-TK). The CCP emergency leakoffs (ELO) isolation valves (EIIS/BQ-LOV) close and then reopen if the Reactor Coolant System (RCS) pressure increases'o greater than 2000 psi. If the accident has progressed to the point of switchover from the RWST to the containment recirculation sump (EIIS/NH-RVR), the CCP suction is realigned to the discharge of the residual heat removal (RHR) pumps (EIIS/BP-P). With the RHR pump supplying the suction to the CCP, the CCP suction pressure could be as high as 205 psig (RHR pump shutoff head pressure). Since the suction to the CCP is supplied by the discharge of the RHR pump and the CCP ELO (if opened) is returned to the suction of the pump, the pressure in the ELO line could be in excess of the 150 psig pressure setpoint for safety valve SV-54 (EIIS/CB-RV) located upstream of the seal water heat exchanger (EIIS/CB-HX).

54 lifts and one CCP is in operation, then approximately 60 gpm flow If SV-would be diverted from the Emergency Core Cooling System (ECCS) into the VCT. If not terminated, the VCT would valve SV-53 (EIIS/CB-RV). SV-53 is set at 75 psig and will divert flow fill and then lift the VCT safety to the chemical and volume control system (CVCS) holdup tanks (EIIS/WD-TK).

NRC Form 366A (669)

NRC FORM SSBA US. NUCLEAR REGULATORY COMMISSION (6()91 APPROVEO 0MB NO. 31504)OS 5 XP I R ES: S/30/92 ESTIMATED BURDEN PER RESPONSE TO COMPLY WTH THIS LICENSEE EVENT REPORT ILER) INFORMATION COLLECTION REQUEST: 508) HRS. FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE RECORDS TEXT CONTINUATION AND REPORTS MANAGEMENT BRANCH (P-530), U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555. AND TO THE PAPERWORK REDUCTION PROJECT (3150410SI, OFFICE OF MANAGEMENTANO BUDGET, WASHINGTON,OC 20503.

FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER LB) PAGE (3)

YEAR LyL~s SSOUSssTIAL ssUMos4

~ AEYSION o NUMosR D. C. Cook Nuclear Plant Uni.t 2 05000316 9 1 007 01 0 3oF 0 6 TEXT /// moss ssssois rrquirat, oss sddkbrul /t/RC Foms 3555's/ (17)

Cause of Event The potential divert flowpath identified by the simulator scenario was caused by steps in the Emergency Operating Procedure (EOP). The EOP instructs the operators to open the CCP ELO isolation valves in preparation for switchover of the CCP discharge from the ECCS lineup through the boron injection tank (BIT) (EIIC/BQ-TK) to the normal charging discharge lineup.

The EOP steps to reset and open the CCP ELO valves in preparation for the switchover of the pump discharge to the normal lineup are based on the EOP Westinghouse Owners Group (WOG) Emergency Response Guidelines (ERG). The EOP ERG states that the ELO should be reestablished before the BIT is isolated.

The affects of this ELO alignment in the above described Plant condition were not previously recognized by the Cook Nuclear Plant design and safety valve setpoints review.

Although not discussed in the Westinghouse letter, American Electric Power Service Corporation (AEPSC) has been informed by Westinghouse that it is not possible to conclude positively that the RCS will be below 2000 psi by the time of switchover to recirculation. This may be especially true for ice condenser plants, because the lower containment spray initiation setpoint for ice condenser plants makes it more likely that containment spray will be activated resulting in more rapid depletion of the refueling water storage tank.

Anal sis of Event This condition is being reported in accordance with 10CFR50.73(a)2(ii)(B) as a condition outside of the design basis of the Cook Nuclear Plant.

Following a review of the findings of this LER, it has be'en concluded that the amount of water diverted from the ECCS is not considered to be significant, based on the following:

In order to establish the normal CCP discharge lineup per the EOP, the RCS inventory control must be manageable with the conditions described in the investigation section of this report.

2. In addition, under the conditions described, the RCS cooldown would be progressing in preparation for placing the RHRs in service for decay heat removal and cooldown.

NRC Fosso 366A (689)

NAC FORM 368A U.S. NUCLEAR REGULATORY COMMISSION (689) APPROVEO 0MB NO.31504104 EXPIRES; 4)30)92 ESTIMATED BURDEN PER RESPONSE TO COMPLY WTH THIS LICENSEE EVENT REPORT (LER) INFORMATION COLLECTION REQUEST: 600 HRS. FORWARD

'EXT CONTINUATION COMMENTS REGARDING BUADEN ESTIMATE TO THE RECORDS ANO REPORTS MANAGEMENT BRANCH (F430). U.S. NUCLEAR AEGULATORY COMMISSION. WASHINGTON, DC 20555, AND TO THE PAPERWOAK REDUCTION PROJECT (31500104), OFFICE OF MANAGEMENTAND BUDGET,WASHINGTON, OC 20503.

FACILITY NAME 111 DOCKET NUMBER LTI LER NUMBER (6) PAGE (3)

YEAR SEDVENTIAL II8 V IS IO N 9% NVMSEII IIUMFSII D. C. Cook Nuclear Plant Unit 2 o s o o o 316 9 1 007 0 1 0 4 OF 0 6 TEXT /IImoro z>>co is n Jrrr)orL oso orssisior>>JHRC Farm 3IJSA'4) (IT)

3. Also, considering that the flowrate diverted from the sump is estimated to be about 60 gpm, the flow diverted is considered small in relation to

,the total volume of water expected to be in the sump (well in excess of 255,000 gallons). This volume in the containment sump is considering the stable condition expected for the ECCS and RCS at the time of the CCP discharge realignment.

Although the amount of water diverted from the ECCS is not considered to be significant from a core cooling perspective, the diverted water does represent a condition whereby coolant is being discharged from the containment building.

Therefore, the diverted water is considered to be a condition outside of the Cook Nuclear Plant design basis.

An analysis of the potential dose consequences of this condition was performed. The analysis assumed a 10 gpm leak from the VCT or CVCS holdup tanks and 1X failed fuel recirculation coolant activity. With the assumed stated conditions, the whole body dose rate at the site boundary was determined to be only 0.293 mR/hr. This is insignificant compared to the 10CFR100 accident dose limit and is even below the 10CFR20.105 dose limits for unrestricted areas during normal operation.

r Westinghouse's review of this issue resulted in a determination that for a small break (2 ' inches) that may allow the RCS to remain at an elevated pressure, the flow out the break is expected to be sufficient to preclude repressurizing the RCS above 2,000 psig. For smaller breaks where RCS repressurizing may occur, the operator is expected to terminate safety injection before charging pump deadheading could occur. Additionally, Westinghouse reviewed a concern for adequate charging flow with the 'miniflow isolated. Break sizes needed to result in charging flow below that required for adequate cooling were found to be less than 3/8 inch in diameter, a break size that is not considered to be a LOCA. Therefore, Westinghouse believes that adequate cooling of the charging pumps during operation in the recirculation modes is assured under LOCA conditions with miniflow isolated.

Based on the above analysis, it is concluded that the condition identified by this simulator scenario would not significantly impact public health and safety.

NRC Form 368A (689)

NRC FORM 366A US. NUCLEAR REGULATORY COMMISSION (689 I APPROV ED 0MB NO. 3'I 500104 EXPIRE5: 4/30/92 ESTIMATED BURDEN PER RESPONSE TO COMPLY WTH THIS LICENSEE EVENT REPORT ILER) INFORMATION COLLECTION REOUESTI 500 HRS. FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE RECORDS TEXT CONTINUATION AND REPORTS MANAGEMENT BRANCH (F430), U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON. OC 20555. AND TO THE PAPERWORK REDUCTION PROJECT (31500104I. OFFICE OF MANAGEMENTAND BUDGET, WASHINGTON, DC 20503.

FACILITY NAME (I) DOCKET NUMBER (2) LER NUMBER (6) PAGE (3)

YEAR SEOUENTIAL REVISION NUMOER NUMBER D. C. Cook Nuclear Plant Unit 2 0 s o 0 o 316 91 007 01 05 QF0 6 TEXT ll!moro spsco is /squi/od, uso odddlo/N/ill/C Fom/ 3664'4/ ((7)

Corrective Actions By letter dated November 3, 1992, Westinghouse notified potentially affected licensees of the results of their review of the issues discussed in our original LER submittal. Although licensees were notified of the issue, Westinghouse's evaluation concluded that the situation does not represent a Substantial Safety Hazard or Failure to Comply pursuant to 10 CFR 21.

The affected Emergency Operating Procedure was revised to caution the operator to establish minimum charging flow to protect the operating CCP if cold leg recirculation has been initiated, and to verify proper operation of the CCP emergency leakoff valves, during SI conditions.

The Westinghouse letter provided several potential means for resolving the issue. One specific option was to review plant specific EOPs to ensure that procedures are provided to plant operators to depressurize and cooldown during a postulated small break LOCA, and to modify the EOPs to ensure that the charging pump miniflow lines are isolated during recirculation.

Following conversations with Westinghouse that preceded receipt of the letter described above, we made modifications to our EOPs that meet the intent of the Westinghouse recommendation. The EOP, ES-1.3 (Transfer to Cold Leg Recirculation) has been modified such that it instructs the operator to open a pressurizer PORV, as necessary, to reduce RCS pressure and to maintain minimum charging pump flow. The procedure instructs the operator to close the charging pump miniflow valves as part of the switchover from injection to recirculation. (Precautions are included, however, that instruct the operator to shut off a redundant charging pump or to open the miniflow valves that ensure adequate pump protection are not met.)

if conditions Failed Com onent Identified None Previous Similar Events None NRC FomI 366A (64)9)

NRC FORM 355A U.S. NUCLEAR REGULATORY COMMISSION (SS)91 APPROVED OMB NO. 31504)0i 5 X PI A 5SI i/30/92 ESTIMATED BURDEN PER RESPONSE TO COMPLY WTH THIS LICENSEE EVENT REPORT (LER) INFORMATION COLLECTION REOUESTI 500 HRS. FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE RECORDS TEXT CONTINUATION AND REPORTS MANAGEMENT BRANCH IP430), U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON, OC 20555. AND TO THE PAPERWORK REDUCTION PROJECT 13150410i). OFFICE OF MANAGEMENTAND BUDGET, WASHINGTON, DC 20503.

FACILITY NAME 11) DOCKET NUMBER 12) LEA NUMBER LE) PAGE IS)

YEAR SEQUENTIAL REVISION NUMEER NUMEER D. C. Cook Nuclear Plant Unit 2 o s o o o 316 9 1 0 0 7 0 1 0 6 OF 0 6 TEXT /// mors spsss is rsr/vtrsrL vss sA//aarM//YRC farm 35EA's/ I )7)

FIGURE 1 75k TO CVCS HOLD-UP TAHRS PRON LETDOUH 150f CLOSED CLOSED FROM SEAL ON SI RETURN PROM RHR PUNPS ELO DURING ECCS RECIRC.

To CHG HEADER CHG PUMPS TO BIT NRC Form 355A ISSQ)