ML17321A577

Advises of Rev to Westinghouse Loca/Eccs Evaluation Model. Rev Effects Flooding Rate Info Generated by Wreflood Code as Input to Bart Code.Rev Discussed in Encl Westinghouse 850403 & Util s

ML17321A577
Person / Time
Site:	Cook
Issue date:	04/19/1985
From:	Alexich M INDIANA MICHIGAN POWER CO. (FORMERLY INDIANA & MICHIG
To:	Harold Denton Office of Nuclear Reactor Regulation
References
AEP:NRC:0745W, AEP:NRC:745W, NUDOCS 8504260100
Download: ML17321A577 (9)

Text

REGULATOR.

NFORMATION DISTRIBUTION S EM (RIDS)

ACC'iSSION NBR: 8504260100 DOC, DATE: 85/00/19 NOTARIZED:

NO DOCKET

,FACIL:50-315 Donald C ~

Cook Nuclear Power Plantg Unit li Indiana 8

05000315 AUTH,NAME AUTHOR AFFILIATION ALEXICHiH,P-.

Indiana L Michigan Electric Co, RECIP ~ NAME<<RECIPIENT AFFILIATION DENTONgH,-Ri

.Office<< of Nuclear Reactor Regulationi Director SUBJECT Advises of <<rev to Westinghouse LOCA/ECCS evaluation models Rev effects flooding rate i.nfo generated by HREFLOOD code as input to BART code,Rev discussed in encl westinghouse!

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INDIANA8 MICHIGAN ELECTRIC COMPANY P.O. BOX 16631 COLUMBUS, OHIO 43216 April 19, 1985 AEP:NRC:0745W Donald C. Cook Nuclear Plant Unit No.

1 Docket No. 50-315 License No. DPR-58 INPUT METHODOLOGY REVISION TO WESTINGHOUSE LOCA/ECCS EVALUATION MODEL Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D. C.

20555

Dear Mr. Denton:

This letter is submitted to inform you of an input methodology revision, which is described in the two attachments:

letter 85AE1I-G-009, dated April 3,

1985, and letter NS-NRC-85-3025, dated March 22, 1985.

The

~

input methodology revision, which we have discussed with your staff, affects the flooding rate information generated by the WREFLOOD code which is input to the BART code.

This revision is expected to result in, an increase in peak clad temper-ature.

However, Westinghouse has informed us that their evaluation of BART results for D. C. Cook Nuclear Plant Unit 1 indicates that the 2200 F limit of 10CFR50.46 will not be exceeded as a result of this revision.

A revised LOCA/ECCS analysis will be performed for D. C.

Cook Unit 1 in the near future.

D. C. Cook Unit 1 is currently shut down for an In Service Inspection outage and not expected to return to service until July, 1985 at the earliest.

This revised analysis will be reviewed and sent to you for review by your staff prior to startup of the next cycle.

No amendment to Technical Specifications is expected to result from the revised analysis.

8504260100 850419 PDa nDOCK 050003i5 P

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Mr. Harold R. Denton AEP:NRC:0745M This information in this letter will be reviewed by the Plant Nuclear Safety Review Committee (PNSRC) and the Nuclear Safety and Design Review Committee (NSDRC) at a future meeting.

This document has been prepared following Corporate procedures which incorporate a reasonable set of controls to insure its accuracy and completeness prior to signature by the undersigned.

Very truly yours, I

M. F. AlexiohI~)<qI~+

Vice President JLB:wj Attachments cc: John E. Dolan M. G. Smith, Jr. - Bridgman George Bruchmann R. C. Callen R. Charnoff NRC Resident Inspector - Bridgman

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Westinghouse Electric Corporation Water Reactor Divisions Nuclear Fuel Oivision Box 3912 Pittsburgh Pennsylvania 15230 85AE*-G-009 April 3, 1985 Keywords:

AEP COOK-1 LOCA-BART

Reference:

NS-NRC-85-3025 W-AEP/01 Indiana and Michigan Electric Company c/o Joseph L. Bell

Engineer, Nuclear Materials and'Fuel Management American Electric Power Service Corporation One Riverside Plaza, 20th Floor

Columbus, OH 43215 NMFM DOCUMENT DOC SS-ODZr DAT

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'EYWORDS i.

p. I os S. -EP

4. aN-F "Z~VD EXP O'F'7 I, II,.

II:~PB AMERICAN ELECTRIC POWER CORPORATION UN 1

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'IBIOII ZoN This letter is to inform you of an input methodology revision in the interface between two computer codes used in the Westinghouse Emergency Core Cooling System (ECCS) evaluation model used to demonstrate compliance with Appendix K to 10CFR50.46.

The code interface methodology revision may have an impact on the ECCS analysis for the D.

C.

Cook Unit 1 Plant, Specifically, the input methodology revision concerns the input interface between the BART code and the WREFLOOD code in the large break loss-of-coolant-accident, (LOCA) analyses.

The WREFLOOD code calculates the thermal-hydraulic response for the reactor coolant system during the refill and reflood period following the large break LOCA blowdown.

The WREFLOOD code calculates the core reflooding rate which is used as an transfer coefficients used to determine the hot rod peak cladding temperature.

The revision in the input methodology may result in an increase in calculated peak cladding temperature for

'nalyses which have used the BART computer code.

This problem has been discussed with Dr. Brian Sheron and Mr. Norman Lauben of the Nuclear Regulatory Commission (NRC) staff and a letter, Reference 1,

has been sent describing the problem.

Additional details regarding this problem may be found in the reference which is provided as an attachment.

Westinghouse has reviewed the D.C.

Cook Unit 1

LOCA analysis which has been performed with the Westinghouse ECCS evaluation model incorporating the BART code and WREFLOOD code interface.

We believe the effect of the input methodology revision will not result in the analysis exceeding the 2220 F regulatory limit on peak cladding temperature for'D. C; Cook Unit 1.

A reanalysis of D.

C'.

Cook Unit 1 is in progress, and you will be notified of the results at the earliest date possible.

0 Apri 1985 If you have any questions concerning these modifications and the status of your reanalysis, please contact Mr. Brian McIntyre of the Nuclear Safety Department at (412) 374-5506 Very truly yours, Project Manager, NFD Projects BMB:mid ATTACHMENT cc:

M. P. Alexich J.

M. Cleveland G. John V.

D. Vanderburg W. L. Zimmermann