ML18086B594

Responds to 820420 Request for Addl Info on Postulated Main Steam Line Break Inside Containment (IE Bulletin 80-04).Addl Analysis Beyond Design Basis Not Necessary

ML18086B594
Person / Time
Site:	Salem, Hope Creek, 05000000, 05000355
Issue date:	07/26/1982
From:	Liden E Public Service Enterprise Group
To:	Varga S Office of Nuclear Reactor Regulation
References
REF-SSINS-6820 IEB-80-04, IEB-80-4, NUDOCS 8208020112
Download: ML18086B594 (2)

Text

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Public Service Electric and Gas Company 80 Park Plaza, Newark, NJ 07101 / 201 430-7000 MAILING ADDRESS/ P.O. Box 570, Newark, NJ 07101 July 26, 1982 Mr. Steven A. Varga, Chief

.operating Reactors Branch #1 Division of Licensing U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Gentlemen:

ADDITIONAL INFORMATION RELATED TO NRC BULLETIN 80-04 This response is directed to your letter dated April 20, 1982 which requested additional information pertaining to a postulated main steam line break inside the containment.

The enclosure to your letter contains a statement indicating that our use of a ten minute operator response time for the accident may be unrealistic.

Existing Westinghouse containment analyses use 10 minutes for isolation of a faulted steam generator and do not project con-tainment peak pressure beyond such assumption. This ten minute isolation time criteria is the design basis for Westinghouse units, inclusive of Salem. In the event that the postulated main steam line break in the containment occurs, auxiliary feedwater to the affected steam generator is manually terminated by push-button operation in the control room. The control room operator, upon evaluating the alarms and indications symptomatic of this very recognizable accident, must depress the 'shut' pushbuttons for either one or two electrically operated control valves supplying the faulted steam generator. The number of valves to be closed depends upon the number of pumps feeding the affected steam generator at that time.

• The instrumentation available to alert the operator of the need to isolate auxiliary feedwater to the affected steam generator is mounted on the control console in the control room. The pressure in each steam generator is monitored and displayed by several independent channel~ of instrumentation. Also, pen recorders indicate steam and feedwater flows for each steam generator; this allows the control room operator to readily view and compare the flows of one steam generator with the others.

The suction and discharge pressures of each auxiliary feedwater pump are indicated on the control console~ The auxiliary feed-water flow indications for each steam generator are mounted on th~ control console ne~t to each other, allowing the operator to easily view and compare flows.

• The Enerav Peoole _

8208020112 820726 PDR ADOCK 05000272 95-2001(400M)10*81 G PDR

Mr. Steven A. Varga U.S. Nuclear Regulatory Commission

• 7/26/82 In addition to the above mentioned indications, high steam flOw, low steam pressure, and steam/feed flow deviation conditions for each steam generator are alarmed on the main control console in the control room. Alarms for these conditions are also provided on the overhead annunciator.

Based on the number of class lE indications provided to monitor plant conditions, the number of alarms provided to annunciate this accident, and the minimal operator actions required to terminate.auxiliary feedwater to the faulted steam generator, operator action within 10 minutes is easily achievable and is justifiable as a design base.

Therefore, we feel that additional analyses beyond the design basis are not necessary.

If you have any questions in this regard, please do not hesitate to contact us.

v~

~ .

E. A. L1den, Manager Nuclear Licensing and

.,/ Regulation CC Mr. Leif Norrholm NRC Senior Resident Inspector Mr. William J. Ross NRC Licensing Project Manager