ML20203N140

Forwards Request for Addl Info to Assist in NRC Review of Util 860829 Submittal Re LOCA Calculations for Facility. Response within 30 Days of Ltr Date Requested

ML20203N140
Person / Time
Site:	Millstone
Issue date:	09/18/1986
From:	Jaffe D Office of Nuclear Reactor Regulation
To:	Opeka J NORTHEAST NUCLEAR ENERGY CO.
References
TASK-2.K.3.30, TASK-2.K.3.31, TASK-TM TAC-48179, NUDOCS 8609230162
Download: ML20203N140 (3)

Text

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Stptember 18, 1986 b

Docket No. 50-336 OI Mr. John F. Opeka, Senior Vice President Nucl_ ear Engineering and Operations Northeast Nuclear Energy Company P. O. Box 270 Hartford, Connecticut 06141-0270

Dear Mr. Opeka:

We are in the process of reviewing your August 29, 1986 submittal regarding Loss-of-Coolant Accident calculations for Millstone Unit 2. In order that we may continue our review, we request that you respond to the enclosed questions within 30 days following receipt of this letter.

This request for additional information affects fewer than 10 respondents, therefore, OMB clearance is not required under P.L.96-511.

Sincerely,

/S/

David H. Jaffe, Project Manager PWR Project Directorate #8 Division of PWR Licensing-B

Enclosure:

Request for Additional Information cc w/ enclosure:

See next page Distribution:

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Mr. John F. Opeka Millstone Nuclear Power Station Northeast Nuclear Energy Company Unit No. 2 cc:

Gerald Garfield, Esq. Mr. Wayne D. Romberg Day, Berry & Howard Superintendent Counselors at Law Millstone Nuclear Power Station City Place P. O. Box 128 Hartford, Connecticut 06103-3499 Waterford, Connecticut 06385 Regional Administrator, Region I Mr. Edward J. Mroczka U.S. Nuclear Regulatory Commission Vice President, Nuclear Operations Office of Executive Director for Northeast Nuclear Energy Company Operations P. O. Box 270 631 Park Avenue Hartford, Connecticut 06141-0270 King of Prussia, Pennsylvania 19406 Mr. Charles Brinkman, Manager Washington Nuclear Operations C-E Power Systems Combustion Engineering, Inc.

7910 Woodmont Avenue

.Bethesda, Maryland 20814 Mr. Lawrence Bettencourt, First Selectman Town of Waterford Hall of Records - 200 Boston Post Road Waterford, Connecticut 06385 Northeast Utilities Service Company ATTN: Mr. Richard R. Laudenat, Manager Generation Facilities Licensing Post Office Box 270 Hartford, Connecticut 06141-0270 Kevin McCarthy, Director f Radiation Control Unit l Department of Environmental Protection -

i State Office Building Hartford, Connecticut 06106 Mr. Theodore Rebelowski U.S. NRC

, P. O. Box 615 Waterford, Connecticut 06385-0615 Office of Policy & Management ATTN: Under Secretary Energy Division 80 Washington Street Hartford, Connecticut 06106 L

REQUEST FOR ADDITIONAL INFORMATION ON MILLSTONE 2 LOCA EVALUATIONS Small Break Analysis

1. Describe.the basis for the core axial power distribution used in the analysis. Justify that this power shape is the worst axial shape j allowed by the Technical Specifications.

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2. It is stated that the small break spectrum analyses, documented in WCAP-10054, Addendum 1, is based upon the Millstone 2 plant. The i Millstone 2 limiting break analysis resulted in a peak cladding temp-erature of 2135'F or approximately 160*F higher than the results in WCAP-10054, Addendum 1. Describe and justify the differences between the models used in these two analyses and discuss the relative effects of these differences on the temperature increase.

3. The staff is not convinced that the 4 inch cold leg pump discharge break is the worst case small break. It is noted that, prior to accumulator actuation, cladding temperature was continuously increasing. The brief accumulator actuation resulted in an approximate 2 foot level increase in the core mixture level which terminated the cladding temperature increase. It appears that the worst case break would be a slightly smaller break which does not rely upon accumulator injection to terminate the transient. Provide additional spectrum analyses to demonstrate that the worst case break has been identified.

Large Break LOCA Analysis

1. On June 2, 1986, Westinghouse notified the staff of errors in its 1981 ECCS evaluation model with respect to modeling of the control rod thimbles. Determine whether this model error is present in your analysis.

, If this error is present, assess its impact on your plant to demonstrate compliance with 10 CFR 50.46.

2. Describe whether the steam generator tube plugging was modeled symme-trically or assymmetrically and justify the approach used.

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