ML20206K630

Provides Addl Info Re TMI-1 LOFW Accident re-analysis Assumptions for 20% Average SG Tube Plugging as Discussed on 990421

ML20206K630
Person / Time
Site:	Three Mile Island
Issue date:	05/07/1999
From:	Langenbach J GENERAL PUBLIC UTILITIES CORP.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
1920-99-20222, NUDOCS 9905130220
Download: ML20206K630 (2)

Text

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GPU Nuclear,inc. -

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L Route 441 South NUCLEAR fy'l&*,**,,.g Tel 717 944-7621 '

May 07,1999 .

1920-99-20222 U S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555 Ladies and Gentlemen:

Subject:

Three Mile Island Nuclear Station, Unit 1 (TMI-1)

Operating License No. DPR-50 Docket No. 50-289 Additional Information - Technical Specification Change Request No. 279 Core Protection Safety Limit This letter provides the additional information regarding the TMI-l Loss of Feedwater (LOFW) accident re-analysis assumptions for 20% average steam generator tube plugging as discussed on April 21,1999.

As described in Technical Specification Change Request No. 279, submitted to NRC on December 3,1998, the proposed change does not affect the normal reactor coolant system (RCS) operating pressure. The re-analysis of the LOFW accident assumed a maximum reactor j protection system (RPS) trip error of +32 psi, and nominal initial RCS pressure. This is O consistent with the assumptions for the existing design basis accident analyses performed in -  ;

Chapter 14 of the TMI-l Updated Final Safety Analysis Report (UFSAR), and is conservative as  !

explained below. ] !

The +32 psi RPS trip error is the calculated value of the error for an abnormal containment environment, which does not exist during a loss of feedwater accident. The RPS trip and RCS -

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o pressure control circuits utilize the same pressure transmitters. The RPS trip function utilizes J four hot leg pressure transmitters,'while the RCS pressure control function utilizes a single pressure transmitter out of these four transmitters. The major contributor to the RPS trip or the

- RCS pressure control error is the potential error of the pressure transmitter.

9905130220 990507 M^ )

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1920-99-20222 Page 2 of 2.

The normal two-sigma difference between the RPS trip and the RCS pressure control is equal to L

the root mean square of the uncertainty in these two variables, and is less than 32 psi. Thus, the .

instrument error in the RPS trip setpoint and the RCS initial pressure, are both accounted for by using a trip setpoint error of+32 psi.

- If any additional information is needed, please contact Mr. David J. Distel, Nuclear Licensing

- and Regulatory Affairs at (973) 316-7955.

Sincerely, h.

James W. Langenbach Vice President and Director, TMI DJD/gba Cc: Administrator, Region I TMI-l Senior Project Manager TMI-l Senior Resident Inspector File 'No. 98195 9

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