ML19329E426

Interim Deficiency Rept Re Small Break Analysis.Small Break Applicability to Specific Facility Design Being Evaluated. Operator Action or Design Mods Feasibility to Mitigate Accident Consequences Being Investigated

ML19329E426
Person / Time
Site:	Midland
Issue date:	05/12/1978
From:	Howell S CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
To:
Shared Package
ML19329E424	List: ML19329E425 ML19329E426 ML19329E430
References
HOWE-74-78, NUDOCS 8006120796
Download: ML19329E426 (2)

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f CORSilmCES POV!Br i m";"-" j C0mpany General Omces: 212 West Micnigen Avenue. Jackson. M6cn6gan 40201 May 12, 1978 Hove-74-78 Mr J. G. Keppler, Regional Director Office of Inspection & Enforcement Region III US Nuclear Regulatory Commission 799 Roosevelt Road Glen Ellyn, IL 60137 MIDLAND NUCLEAR PLANT -

UNIT NO. 1, DOCKET NO. 50-329 UNIT NO. 2, DOCKET NO. 50-330 SMALL BREAK ANALYSIS -

Reference:

1) Babcock & Wilcox (James H. Taylor) letter to the Nuclear Regulatory Commission (Dr Ernst Volgenau), dated April 14, 1978.

2) Babcock & Wilcox (James H. Taflor) letter to the Nuclear Regulatory Commission (Mr Robert L. Baer), dated April 25, 1978.

3) Babcock & Wilcox (James H. Taylor) letter to the Nuclear Regulatory Commis'sion (Mr Robert L. Baer) dated May 1, 1978..

The reference 1 letter reported under 10 CFR Part 21, a condition potentially affecting safety for which it was postulated that for B&W 177FA Lovered-Loop Plants, the analysis presented in BAW-10103A, "ECCS Analysis for B&W's 177FA Lovered-Loop NSS," may be non-conservative for a small break in the reactor coolant pump discharge. The subsequent reports, reference letters 2 and 3, describe the methods used and results obtained from B&W's generic studies of small b'reaks. A worst case break has been determined to be at the reactor coolant pump discharge and those evaluation reports show that operation up to at least 2568 MWt is possible within the criteria of 10 CFR 50.46 and Appendix K. Those reports also show that it is necessary to use operator action during the early stages of the postulated accident to achieve sufficient and balanced flow through all four HPI injection lines to effec-tively mitigate the accident consequences.

This letter is an interim 50 55(e) report of the condition as applicable to the 2568 MWt Midland Nuclear Plant. An evaluation is in progress on the applicability of the condition to the specific design of the Midland Plant.

The feasibility of using operator action or de' sign modifications is under preliminary investigation as a means of mitigating the consequences of an accident.

8006120 7 f 5 N.AY15I978

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.'- , i. 2 Another interim report will be provided by July 21, 1978.

CC: Dr Ernst Volgenau, USNRC (15)

Director, Office of Management Information and Program Control, USNRC (1)

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