ML20041C714

Submits Addl Info Supporting 820204 Cycle 5 non-LOCA Transient Analyses Results for Addl Plugged Steam Generator Tubes,Per Request.Increasing Number of Tubes Do Not Require Tech Spec Mods

ML20041C714
Person / Time
Site:	Millstone
Issue date:	02/23/1982
From:	Counsil W NORTHEAST NUCLEAR ENERGY CO.
To:	Clark R Office of Nuclear Reactor Regulation
References
820224, B10433, TAC-47355, TAC-47389, TAC-47473, NUDOCS 8203020498
Download: ML20041C714 (2)

Text

.

$10RTDIEAerg. UTII.ITIES *"*'d "'

• 5 * 'd*" 5 ""'- '"C*""''

P11.? C,1',% "7f?:' T ~ feffo%7 connecticur asioi 5

T? :::;;%;'.'l".' ) '

• A 903) see-esti L L J  ::TZ4T ,

co RECEIVED i

ck MAR 011982d '

' a ssun tenu tamma February 23, 1982 mannwrx a e t

• _ Docket No. 50-336 s @ B10433 Director of Nuclear Reactor Regulation Attn: Mr. Robert A. Clark, Chief Operating Reactors Branch #3 U. S. Nuclear Regulatory Commission Washington, D.C. 20555

Reference:

(1) W. G. Counsil letter to R. A. Clark, dated February 4,1982.

Gentlemen:

Millstone Nuclear Power Station, Unit No. 2 Additional Information Regarding Cycle 5 Non-LOCA Transient Review Northeast Nuclear Energy Company (NNECO) informed the NRC Staff in Reference (1) of the continued validity of the Cycle 5 non-LOCA transient analyses results with additional plugged steam generator tubes. This determination was made following a detailed review of each non-LOCA transient scenario with regards to potential effects due to the plugging of steam generator tubes during the current refueling outage. At the request of the NRC Staff, NNECO hereby provides the following discussion to support the conclusions presented in Reference (1).

A review of the non-LOCA safety analyses for Millstone Unit No. 2 was performed to determine the impact of plugging additional steam generator tubes. The review entailed a comparison of the key parameters for each transient with the additional tubes plugged as compared to the docketed analyses. The most significant parameters affected are number of tubes plugged, primary coolant flow, primary system temperatures, reactor coolant system volume, steam generator heat transfer area, and secondary system operating conditions. The intent was to determine whether there were any significant impacts in the results or conclusions of the previous analyses. In performing this review, the non-LOCA analyses can be classified into transients which involve secondary system effects and those which involve primary system effects. This type of categorization is appropriate since the major impact of plugging steam generator tubes is to alter the steam generator heat transfer effects (a secondary side effect) and to alter the primary system coolant flow (a primary side effect).

r203020499 820223 lg PDR ADOCK 05000336 P PDR g

t The original non-LOCA analyses assumed a total of 1,000 steam generator tubes were plugged in both generators. This evaluation assumes 1,550 total tubes are plugged. The total number of plugged steam generator tubes at Millstone Unit No. 2 is 1,504.

The primary system transients are impacted by the reduced core coolant i

flow rate (and subsequently higher core a T) associated with the tube plugging and indirectly due to the decreased heat transfer area in the steam generator. However, it has been estimated that the core coolant flow after plugging will be greater than 114% of design minimum coolant fl ow. All safety analyses conservatively assume a primary flow (and corresponding core a T) of 114% of design. Therefore, a slight reduction in flow does not adversely impact any of the results of these analyses.

Confirmation that the reactor coolant flow is greater than 114% of design will be accomplished during the power ascension test program for Cycle 5. The small reduction in heat transfer area has a negligible effect upon these transients and the conclusions remain valid. The analyses included in this category are CEA Withdrawal, Boron Dilution, RCS Depressurization, Loss of Coolant Flow, CEA Drop, CEA Ejection, and Seized Rotor.

The secondary system transients are affected by the reduced steam generator heat transfer area, the smaller volume of water in the steam generator tubes, and the reduced secondary system pressure caused by the plugging.

The effect of these changes on the primary system cooldown and heatup transients is negligible. This is because the change in total heat transfer area and volume of water in the steam generator tubes is small (less than 4%). The transients included in this category are Loss of Load, Loss of Feedwater, Excess Load, Malfunctioning of One Steam Generator, Steam Line Break, and Steam Generator Tube Rupture. For the Steam ,

Generator Tube Rupture transient, the slight difference in the initial secondary pressure has a negligible effect on the total primary-to-secondary leakage and the activity releases would be no more adverse than those reported in the previous analyses.

In summary, the conclusions and results of all previously docketed non-LOCA analyses remain valid. The increase in the number of tubes plugged does not require modifications to the Technical Specifications for non-LOCA transients.

We trust you find this information responsive to your request.

Very truly yours, NORTHEAST NUCLEAR ENERGY COMPANY P MMt W.'G. Counsil Senior Vice President ,

_-