05000315/LER-1997-001, :on 970112,while in Mode 1 at 99.8% Rated Thermal Power,Si Pump Discharge cross-tie Train B Shutoff Valve Was Closed to Support Filing SI Sys Accumulators. Caused by Incorrect Sequence of Procedure Step

:on 970112,while in Mode 1 at 99.8% Rated Thermal Power,Si Pump Discharge cross-tie Train B Shutoff Valve Was Closed to Support Filing SI Sys Accumulators. Caused by Incorrect Sequence of Procedure Step

ML17333A782
Person / Time
Site:	Cook
Issue date:	02/11/1997
From:	Blind A, Gillespie R AMERICAN ELECTRIC POWER CO., INC.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
LER-97-001, LER-97-1, NUDOCS 9702200157
Download: ML17333A782 (9)

LER-1997-001, on 970112,while in Mode 1 at 99.8% Rated Thermal Power,Si Pump Discharge cross-tie Train B Shutoff Valve Was Closed to Support Filing SI Sys Accumulators. Caused by Incorrect Sequence of Procedure Step

Event date:
Report date:
Reporting criterion:	10 CFR 50.73(a)(2) 10 CFR 50.73(a)(2)(ii) 10 CFR 50.73(a)(2)(iii) 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(viii)(A) 10 CFR 50.73(a)(2)(viii)(B) 10 CFR 50.73(a)(2)(x)
3151997001R00 - NRC Website

text

CATEGORY j.

REGULATOR INFORMATION DISTRIBUTION TEM (RIDS)

ACCESSI'ON NBN:9702200157

, DOC.DATE: 97/02/11 NOTAR25ED:

NO FACIL:50-315 Donald. C.

Cook Nuclear Power Plant, Unit 1, Indiana M

AUTH.NAME AUTHOR AFFILIATION GIJ LRsPIEER.

American Electric Power co., Inc.

BLIN1),A.A.

American Electric Power Co., Inc.

RECIP.NAME RECIPIENT AFFILIATION DOCKET I 05000315

SUBJECT:

LER 97-001-00:on 970112,while in mode 1 at 99.8% rated thermal power,SI pump discharge cross-tie train "B" shutoff valve was closed to support filing SI sys accumulators.

'aused by incorrect sequence of procedure step.W/970211 ltr.

DISTRIBUTION CODE: IE22T COPIES RECEIVED:LTR ENCL SIZE:

TITZE: 50.73/50.9 Licensee Event Report (LER), Incident Rpt, etc.

NOTES:

T t

Q RECIPIENT

'ID CODE/NAME.

PD3-3 PD INTERNAL: AEODQQ~RAB Q~IZ E CENTER NRR/DE/EELB NRR/DRCH/HHFB NRR/DRCH/HOLB NRR/DRPM/PECB NRR/DSSA/SRXB RGN3 FILE 01 EXTERNAL: L ST LOBBY WARD NOAC POOREEW NRC PDR COPIES LTTR ENCL 1

1 2

2 1

1 1

1 1

1 1

1 1

1 1

1 1

1 1

1 1

1 1

1 RECIPIENT ID CODE/NAME HICKMANEJ AEOD/SPD/RRAB NRR/DE/ECGB NRR/DE/EMEB NRR/DRCH/HICB NRR/DRCH/HQMB NRR/DSSA/SPLB RES/DET/EIB LITCO BRYCEEJ H

NOAC QUEENER,DS NUDOCS. FULL TXT COPIES LTTR ENCL 1

1 1

1 1

1 1

1 1

1 1

1 1

1 1

1 1

1 1

1 1

1 D

0 L

C E

N NOTE TO ALL "RIDS" RECIPIENTS:

PLEASE HELP US TO REDUCE WASTE!

CONTACT THE DOCUMENT CONTROL DESK, ROOM OWFN 5D-5(EXT. 415-2083)

TO ELIMINATE YOUR NAME FROM DISTRIBUTION LISTS FOR DOCUMENTS YOU DON'T NEED!

FULL TEXT CONVERSION REQUIRED TOTAZ NUMBER OF COPIES REQUIRED:

ZTTR 24 ENCZ 24

American Electric Pos Cook Nuclear Plant One Cook Place Bndgman, Ml49106 616 465 5901 Z

AMERICAN EI.ECTRIC POUTER February 11, 1997 United States Nuclear Regulatory Commission Document Control Desk Rockville, Maryland 20852 Document Control Manager:

Operating Licenses DPR-58 Docket No. 50-315

'hi lych 97-001-00 Sincerely, A. A. Blind Site Vice President

/mbd Attachment A. B.

Beach, Region III E. E.

Fitzpatrick P. A.

Barrett S. J.

Brewer J.

R.

Padgett D.

Hahn Records Center, INPO NRC Resident Inspector 9702200i57 9702ii PDR ADQCK 050003i5 S

PDR

>>0J0V

HRC FORM 366 (5-92)

S.

NUCLEAR REGULATORY COMMISSION LICENSEE EVENT REPORT (LER)

PROVED BY OHB NO. 3'l50-0104 EXP IRES 5/31/95 I

ESTIHATED BURDEN PER

RESPONSE

TO COHPLY liITH THI INFORMATION COLLECTION REQUEST: 50.0 HRS.

FORNAR COMMENTS REGARDING BURDEN ESTIMATE To TN INFORHATION AND RECORDS MANAGEMENT BRANCH (MNB 7714),

UPS.

NUCLEAR REGULATORY COMMISSION, liASHIHGTON, DC 20555-0001, AND TO THE PAPERNOR REDUCTION PROJECT (3150-0104),

OFFICE 0

HANAGEHENT AND BUDGET liASHINGTON DC 20503.

FACILITY NAHE (1)

Donald C. Cook Nuclear Plant - Unit 1 DOCKET NUMBER (2) 50-315 Page 1 of 5 TITLE (4)

Technical Specification 3.03 Entered On Loss of Four Loop Injection Capability Due to Incorrect Procedural Guidance MONTH DAY YEAR YEAR SEQUEHTIAL NUMBER REVISION NUMBER HONTH DAY YEAR FACILITY HAME DOCKET NUMBER 01 12 97 97

001 02 11 FACILITY NAHE 97 DOCKET NUHBER OPERATING MODE (9)

PONER LEVEL (10)

20. 2201(b) 20.2203(8)(2)(I) 20.2203(a)(2)(iii) 20.2203(a)(2)(iv) 20.2203(a)(2)(v)

X 20.2203(a)(3)(I) 20.2203(a)(4) 50.36(c)(2) 50.73(a)(2)(I) 50.73(a)(2)(ii) 50.73(a)(2)(iii) 50.73(a)(2)(v) 50.73(a)(2)(viii)(A) 50.73(a)(2)(viii)(B) 50.73(a)(2)(x) 73.71(b)

OTHER (Specify in Abstract below and

>n Text,NAHE Robert Gillespie, Operations Superintendent TELEPHONE NUMBER (Include Area Code) 616 465-5901, X2535

CAUSE

SYSTEM COMPOHENT MANUFACTURER REPORTABLE To NPRDS

CAUSE

SYSTEH COMPONENT HANUFACTURER REPORTABLE TO NPRDS YES X

No EXPECTED SUBMISSION DATE 15 ABSTRACT (Limit to 1400 spaces, i.e., approximately 15 single-spaced typeMritten lines) (16)

On January 12, 1997, with Unit 1 in Mode 1 at 99.8 percent Rated Thermal Power, the Safety Injection (Sl) pump discharge cross-tie train "B"shutoff valve was closed to support fillingthe Sl system accumulators.

The East and West Residual Heat Removal (RHR) pump discharge cross-tie valves were already closed at this time. With the unit in this configuration, all four loop injection was lost and Technical Specification 3.0.3 was entered.

This condition existed for approximately one minute. The event was terminated when the RHR cross-tie valves were opened per procedure.

The problem was discovered on January 16, 1997 during a subsequent simulator training scenario on four loop injection where itwas determined that procedure steps had been incorrectly sequenced.

A one hour notification was made in accordance with 10 CFR 50.72(b)(1)(ii)(B), as a condition outside of the design basis, at 1653 hours0.0191 days <br />0.459 hours <br />0.00273 weeks <br />6.289665e-4 months <br /> on the same day.

This event was a result of incorrect procedural guidance.

The affected procedures were corrected to direct the proper sequence for operation ofthe Sl pump and RHR pump discharge cross-tie valves. Other actions were taken to ensure that source documentation, performance of 10CFR50.59 reviews and functional reviews are properly performed.

This event was evaluated against the analyses for large break LOCA, small break LOCA, and containment, and found to have no safety significance.

HRC FORM 366A U.

CLEAR REGULATORY COMHISSION LICENSEE EVENT CONTINUATION PROVED BY OMB HO. 3150-0104 EXP IRES 5/31/95 1

ESTIMATED BURDEN PER

RESPONSE

TO COMPLY 'WITH THI IHFORMATION COLLECTION REQUEST: 50.0 HRS.

FORNAR COHMENTS REGARDING BURDEN ESTIMATE TO TH INFORMATION AND RECORDS MANAGEMENT BRANCH (MNB 7714),

U.S.

NUCLEAR REGULATORY COMMISSION, MASHINGTON~

DC 20555 0001 ~

AND TO THE PAPERNOR REDUCTION PROJECT (3150-0104),

OFF ICE 0

MAHAGEMENT AND BUDGET MASHINGTON DC 20503.

Cook Nuclear Plant - Unit 1 50-315 YEAR SEQUENTIAL 97

001 REVISION 2OF 5 TEXT (if more space is required, use additional(17)

Co ditions P 'or to Occurrence Unit 1 in Mode 1 at 99.8 percent Rated Thermal Power.

escri ion of Even On January 12, 1997, with Unit 1 at 99.8 percent Rated Thermal Power, during performance of 01-OHP 4021.008.004, "Adjusting the Level of Accumulators", four loop injection capability was momentarily lost due to the simultaneous closure of the Safety Injection (SI) (EIIS-BQ) and Residual Heat Removal (RHR) (EIIS-BO) discharge cross-tie valves. Procedure Step 3.8.1 directed the operators to close at least one SI pump discharge cross-tie valve. The valve selected was 1-IMO-275, Sl pump discharge cross-tie train "B"shutoff valve. Atthis time, the East and West RHR pump discharge valves, 1-IMO-314 and 1-IMO-324, were already closed. The resultant configuration precluded four loop injection capability and placed the unit in Technical Specification 3.0.3. Step 3.8.2 opened both 1-IMO-314 and 1-IMO-324 restoring four loop injection capability. Four loop injection capability was unavailable for approximately one minute.

This condition was discovered on January 16, 1997, by the same Operations shift who had performed the accumulator level adjustment, during a subsequent simulator training scenario on four loop injection. During this simulator training itwas determined that the steps in 01-OHP 4021.008.004 had been incorrectly sequenced.

Cause o ve This event is attributable to incorrect procedure guidance resulting from personnel error on the part of the procedure writer and supporting reviewers.

Anal sis ofthe Eve This event is being reported under 10 CFR 50.72(b)(1) (ii)(B), as a condition outside of the design basis.

With both sets of cross-ties closed, the Sl system was in a configuration where four-loop injection capability was not available in the event of a worst-case single failure. In this configuration, each Sl and RHR pump would flowto the RCS through its own header only, two-loop injection for each pump. This is acceptable ifboth trains of ECCS are operating. However, ifone train of. ECCS were to fail, then flowwould only be delivered to two loops, and at a flow consistent with the piping resistance of one header, as the piping resistance of one header causes the operating point of the pump to shift to a lower flowrate. This results in a reduced ECCS flowduring injection from a given Sl pump than would be available ifthat pump were open to'both headers.

Large Break LOCA (LBLOCA)Analysis The Unit 1 LBLOCAwas analyzed at several different sets of operating conditions. One case was analyzed at 3250 MWtwith the RHR cross-tie valves closed and the Sl cross-tie valves open. The ECCS configuration that, existed during the performance of the accumulator fillprocedure had both sets of cross-ties closed for approximately one minute. The LBLOCAwas not analyzed for the case of all cross-ties closed. Therefore the ECCS configuration on January 12, 1997 placed the plant in an unanalyzed condition viith respect to LBLOCA.

NRc FDRM 366A U.

CLEAR REGULATORY COMMISSION PROVED BY OMB NO. 3150 0104 EXPIRES 5/31/95 LICENSEE EVENT CONTINUATION ESTIMATED BURDEN PER

RESPONSE

To COMPLY WITH THI INFORMATIOH COLLECTIOH REQUEST: 50.0 HRS.

FORWAR COMMENTS REGARDING BURDEN ESTIMATE TO TH INFORMATION AND RECORDS MANAGEMENT BRANCH (MNB 7714),

U.S.

NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555-0001, AND TO THE PAPERWOR REDUCTION PROJECT (3150.0104),

OFFICE OF MAHAGEMENT AHD BUDGET WASHINGTOH DC 20503.

Cook Nuclear Plant - Unit 1 50415 YEAR SEQUEHT IAL REVISION 97 3OF 5 TEXT (if more space is required, use additional(11)

Even though the ECCS configuration was outside of its design limit,there are conservative assumptions in the LBLOCAanalysis which would help offset the reduced flow condition associated with the Sl cross-ties closed. The LBLOCAanalysis assumes that the Sl pumps are degraded by 10%. Actual plant test data plotted against the Sl pump head curve show that the SI pumps have very little degradation. This is important since the flow rate through one header at higher RCS pressures, early in the LBLOCAtransient, is in the portion of the pump head curve where a 10% degradation results in a significant drop-off in delivered flow. The result is that the Sl pump could have delivered more flowthan the assumptions in the accident analysis would have allowed, particularly earlier in the blowdown before the RHR pumps begin to deliver flowto the RCS. In addition, since the RHR pumps are also assumed to be degraded by 10%, their nominal flowwould have also been higher than assumed in the accident analysis, and could have been a significant source of additional water. The RHR pumps cannot begin to deliver flow until the RCS depressurizes below 170 psig, so the extra flowfrom these pumps willbe delayed. Since the peak cladding temperature (PCT) does not occur until approximately 65 seconds after the break, and the RHR pumps begin delivering flowat approximately 24 seconds, they will be at fullflowwell before the PCT is reached. Finally, our current analysis of record for this case results in a calculated PCT of2075 'F. Therefore there is some additional margin to the maximum allowable PCT of 2200 'F.

It should also be noted that the Appendix K evaluation models have substantial built-in margin that allow the analysis to envelope all expected operating and plant conditions. In addition, the analysis models postulate a worst case single failure, loss of one ECCS train with a loss of offsite power, that further limits the analysis ECCS flow.

The risk to which Unit 1 was exposed to due to the configuration ofthe ECCS is very limited due to the short time involved. An informal estimate of the increase in core damage probability for a 15 minute period with one train of ECCS inoperable is approximately 1 x 10~. Loss of an entire train of ECCS would easily bound the loss of flowfrom the Sl pumps with the Sl cross-ties closed. Therefore, the risk impact to the core for one minute was minimal. In addition, Technical Specification 3.0.3 requires that the plant initiate actions within one hour to place the plant in at least Hot Standby within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> ifthe requirements of an LCO or its Action statement cannot be met. This requirement recognizes the relative risk associated viith a condition that violates a Technical Specification, and the rapid shutdown of the plant, by allowing a certain amount of time to initiate orderly actions to shut the plant down. The risk due to design basis accidents inherent in the one hour time allotted to initiate the shutdown of the unit is accepted in the Technical Specifications, and bounds the 1 minute exposure time in which the Sl cross-ties were closed.

In conclusion, the configuration of the ECCS system with both cross-ties closed is outside of the LBLOCAanalysis assumptions for Unit 1 ofthe Donald C. Cook nuclear plant. However, the conservative assumptions in the LBLOCA analysis help to offset the reduction in flowthat results from the loss of four loop injection capability given a postulated worst-case single failure. In addition, the brief period oftime that the plant was in this configuration limits the, actual risk to which the plant was exposed to within the window recognized in the Technical Specifications for loss of ECCS capability.U.

CLEAR REGULATORY COMMISSION PROVED BY OMB No. 3'150-0104 EXPIRES 5/31/95 LICENSEE EVENT CONTINUATION ESTIMATED BURDEN PER

RESPONSE

To COMPLY MITH THI INFORMATION COLLECTION REQUEST: 50.0 MRS.

FORNAR COMMENTS REGARDING BURDEN ESTIMATE TO TH INFORMATIOH AND RECORDS MANAGEMENT BRANCH (MHB 7714),

U.S.

NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555-0001, AHD TO THE PAPERMOR REDUCTION PROJECT (3150-0104),

OFFICE 0

MAHAGEHENT AND BUDGET WASHINGTON DC 20503.

Cook Nuclear Plant - Unit 1 50415 YEAR 97 SEQUENTIAL REVISION 40F5 TEXT (if nore space is required, use addirional<1'7i Small Break LOCA (SBLOCA) Analysis The analysis for the SBLOCA for Unit 1 includes a case in which the Sl cross-tie is closed and the RHR cross-tie is open. Asimilar case was also analyzed for the main steam safety valve setpoint tolerance relaxation from 1% to 3%.

The event described in the condition report stated that both sets of cross-ties were closed for one minute. Since the analysis models the Sl cross-tie closed, there is no difference in the ECCS configuration with respect to Sl flow. The analysis also modeled the RHR cross-tie open. However, the RHR pumps do not have a significant impact on the analysis results since the peak cladding temperature for all ofthe cases analyzed occurs at RCS pressures much higher than the RHR pump shutoff head. Therefore, the configuration ofthe ECCS system with the SI and RHR cross-ties closed does not significantly affect the results of the SBLOCA at 3250 MWt.

Containment Analysis The long term containment pressure analysis assumes that minimum safeguards is available for core cooling. Other assumptions are that the RHR cross-tie valves are closed, the RHR and Sl pumps are degraded by 10%, and the core power is at 3425 MWt. However, the peak containment pressure occurs after the ice bed meltout, which in turn is after the transfer to containment sump recirculation. The Sl cross-ties are closed when recirculation is established.

Since the Sl cross-ties are closed well before the peak containment pressure occurs, their closure earlier has no significant affect on the long term containment pressure analysis. Note that before ice bed meltout, the containment pressure is at or below 8 psig, well below the containment design pressure of 12 psig.

The peak containment temperature analysis is based on a main steam line break inside of containment, and is not affected by this ECCS configuration.

Conclusion r

The inadvertent closure of the Sl cross-tie valves for approximately one minute placed the plant in an unanalyzed condition with respect to the LBLOCAanalysis. However, the conservative assumptions in the LBLOCAanalysis help to offset the reduction in flowthat results from the loss offour loop injection capability (assuming a worst-case single failure. In addition, the brief period of time that the plant was in this configuration limits the actual risk to which the plant was exposed to within the one hour window recognized in the Technical Specifications (e.g. TlS 3.03) for loss of ECCS capability. The ECCS configuration does not significantly impact the results ofthe SBLOCA and the containment integrity analysis.

Based on the above evaluations, the event was determined to have no safety significance.

HRC FORM 366A U.

CLEAR REGULATORY COMMISSION LICENSEE EVENT CONTINUATION PROVED BY OMB NO. 3150-0104 EXPIRES 5/31/95

'I 1

ESTIMATED BURDEN PER

RESPONSE

TO COMPLY IIITH THI INFORMATION COLLECTION REQUEST: 50.0 HRS.

FORIIAR COMMENTS REGARDING BURDEN ESTIMATE TO TH INFORMATION AHD RECORDS MANAGEMENT BRAHCH (MHB 7714),

U.S.

NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555-0001, AHD TO THE PAPERNOR REDUCTION PROJECT (3150-0104),

OFFICE 0

MAHAGEMEHT AND BUDGET HASHINGTON DC 20503.

Cook Nuclear Plant - Unit 1 50-315 YEAR 97 SEQUENTIAL REVISION 5OF5 TEXT iif morc space is required, uso additional NRC I'orm 366A's)

(17)

Corrective Actions

The RHR pump discharge cross-tie valves were opened during the next procedural step.

Procedure 01/02-OHP 4021.008.004 was corrected to direct the proper sequence for operation ofthe Sl pump and RHR pump discharge cross-tie valves. Appropriate administrative actions were taken to ensure that accurate guidance is implemented in the operating procedures.

Training was provided to all Operations Department procedure writers on source document requirements, 10CFR50.59 reviews and functional reviews.

The reactor operator requalification program is currently providing training on the proper operation ofthe safety injection pump and residual heat removal pump discharge cross-tie valves. AIIoperating crews were also provided with a 'Lessons Learned" training document describing barriers that failed to identify this event before it occurred.

Operator Aids were placed on the main control boards to provide guidance to the operators on Sl pump and RHR pump discharge cross-tie configuration control.

ailed Co o e Identificatio Not Applicable

Previous Similar Events

315/95-004-00