ML19308C930

Forwards Request for Addl Info Re Postulated Main Steam Line Break Inside Containment to Demonstrate Plant Safety from Both Core & Containment Integrity Standpoints

ML19308C930
Person / Time
Site:	Crane
Issue date:	11/21/1975
From:	Kniel K Office of Nuclear Reactor Regulation
To:	Arnold R METROPOLITAN EDISON CO.
References
TASK-TF, TASK-TMR NUDOCS 8002110670
Download: ML19308C930 (4)

Text

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UNITED 'ST/.TES OLEAR REGULATOHV C O r.t.1 t * "s.

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w asmu cto re, o. c. nass s Oscket l:o. 50-320 flova.ber 21, 1975 Metropolitan Edison Company ATTf!:

fir. R. C. Arnold Vice President P. O. Box 542 Reading, Pennsylvania 19503 Gentlemen:

As a result of our continuing review 'of the Three l'.ile Island ? ucle:r Station Final S&fety Analysis Report, we have identified the enclosed request fcr additional analysis.

The necd for and basic requirstits of this analysis ucre conveyed to your staff in a meeting in Cetheida on October 30, 1975 during a discussion of the analy:;is witich had pre-viously been parformed in response to second round question 042.7.

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In order to ct3intain our present expected licensing schedule, us recuire your response by Jan'uary 20, 1976.

If you cennot r.:ct this dato, olease inform us within 10 days of the receipt of this letter, s

Sincerely,/ / _.

//

S Karl Knicl, Chief ~

Light Cater Reactors Branch 2-2 w

Division of Reactor Licensing Enclosura:

Request for Additional Analysis ccs:

See page 2

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2 ecs:

George F. Troubridge, Esquire Shaw, Pittman, Potts & Trowbridge 910 17th Street,fl.'W.'

Washington, D. C.20006 Chauncey R. Kepford, Esquire Chai rman York Committee for a i

Safe Environment General Delivery York, Pennsylvania 17401 1

Mr. Richard H. Howard Project lianager GPU Service Corporation 260 Cherry Hill Rond Parsippany,ilow Jersey 07054 1

Mr. Thomas M. Crin:ains, Jr.

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Safety and Licensing !! nager l

~ GPU Service Corporation 260 Cherry Hill Road-i Parsippany, New Jersey 07054 2

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REQUEST FOR ADDITIU'!AL A!!ALYSIS_

ThRu IIILE iSL!..'iO, U:iIT'd 21.50 The staff requires the' folicwing additional analyses of the Three Mile Island 2 plant relative to a postulated mcin steam line break inside. containment to demonstrate: that the plant is safe from both core and containment integrity standpoints.

The following parameters must be considered in the analysos:

1.

A spectrum of steam line breaks inside of containment; 2.

The most reactive rod stuck out; 3.

The v!orst single active failure affecting:

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core behavior b.

containment integrity 4.

Availability of offsite pcwer, i.e., with and without offsite e

power.

5.

As a consequence of the accident, consider the possibility of other equip:aent failure, e.g., loss of reacter ccolant pumps or valve operation due to steca environment.

For those sequence of_ events determined to be most severe relctive to the core behavior and containment integrity, the following results are to be presented for the cases where operotor cction r;;y and cay not be required.

a) reactor coolant system pressure; b) steam generator pressure; c) fluid temperature d) fuel and clad temperature; e) discharge flow rate; f) steam line and feedwater ficw rates ;

i g) safety and relief valve flow rates; h) pressurizer and steam generator water levels; i

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2-i) mass and energy. transfer within containment; j) containment pressure; k) reactor power; 1) total core reactivity; m) hot and average channel heat flux; and n) minimum departure from nucleate boiling ratio (D tBR).

With each analysis provide a reference of pertinent events and actions including operator actions.

A table of pertinent parameters for each of the accidents analyzed which were not determined to be limiting will be satisfactory.

Include minimua DitCR, containment pressure, and a description of of the accident.

In the event the consequences of a steam line break using the above assumptions are unacceptable, a probabilistic anclysis of the seduance of events that uvuld occur, including any necessary operator action, should be performed to quantify the degree of risk involved.

In the event your analyses indicate unacceptable results for either the core or containment ir.tegrity, indicate what design modifications could be performad to assure safety.

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