ML19274G102

Forwards Request for Addl Info to Util Proposed Steam Break Protection Sys & N-1 Loop Operation

ML19274G102
Person / Time
Site:	Beaver Valley
Issue date:	07/31/1979
From:	Schwencer A Office of Nuclear Reactor Regulation
To:	Darrell Dunn DUQUESNE LIGHT CO.
References
TAC-10386, TAC-11010, NUDOCS 7908290642
Download: ML19274G102 (6)

Text

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July 31,1979 Docket No. 50-334 Mr. C. N. Dunn Vice President Operations Division Duquesne Light Company 435 Sixth Avenue Pittsburgh, Pennsylvania 15219

Dear Mr. Dunn:

On October 27, 1978, you submitted a request for amendment to the operating license for the Beaver Valley Power Station, Unit No. 1.

In that request, you proposed a new steam break protection system and (N-1) loop ope'.ation.

In order for us to proceed with our review of the steam break protection system, we will need the additional infor-mation requested in the enclosure. The questions on the (N-1) loop operation are related to our steamline break protection system review and we will eventually need answers to these questions as well.

Your response to the steamline break protection system questicns should be submitted within 30 days upon receipt of this letter.

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hg46t/&J A. Schwencer, Chief Operating Reactors Branch =l Division of Operating Reactors

Enclosure:

Request for Additional Information cc: w/ enclosure 2QQg

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See next page 7 9 082 90(oQ,.

Mr. C. N. Dunn Duquesne Light Company July 31,1979 cc: Gerald Charnoff, Esquire Mr. James A. Werling Jay E. 511 berg, Esquire Plant Superintendent Shaw, Pittman, Potts and Trewbridge Beaver '/ alley Power Station 1800 M Street, N.W.

P. O. 3cx 4 Washington, D. C.

20036 Shippingport, Pennsylvania 15077 Karin Carter, Esquire Special Assistant Attorney General Bureau of Administrative Enforcement 5th Floor, Executive House Harrisburg, Pennsylvania 17120 M r. Roger Tapan Stone and Webster Engineering Cor: ora tion P. O. Scx 2325 Easten, Massachusetts 02107 M r. J. D. Woodward R 3 D Center Westinghouse Electric Corporation Svilding 7-303 Dittsburgh, Pennsylvania 15230

3. F. Jones Memorial Library

$63 Franklin Avenue Aliquippa, Pennsylvania 15001 Mr. Jack Carey Technical Assistant Duquesne Light Ccmpany P. O. Sex 4 Shippingport, Dennsylvania 15077 Mr. R. E. Martin Duquesne Light Company 435 Sixth Avenue Pittsburgh, Pennsylvania 15219 Marvin Fein Utility Counsel City of Dittsburgh 313 City-County Building 3ft sourgh, Pennsylvania 15219 2008 <?18

STEAMLINE BREAK PROTECTION SYSTE1 BEAVER VALLEY UNIT NO. 1 REQUEST FOR ADDITIONAL INFOR!1ATION Reactor Systems 1.

Proposed Steamline Break Protection System Determine the ability for core heat removal following a steam-line break with the RCS initially in a hot shutdown condition.

Assume the RCS has been borated to the cold shutdown Ax prior to the event, the SIS actuation signals have been disabled (Lp and low pressure), and assume the new SLBPS is in the "cooldown/heatup" mode.

Two calculations should be done: a steamline break inside and outside containment.

For the steamline break outside containment, the break location and size should be such that the "High Negative Pressure Rate" trip setpoint is not reached.

For the steamline break inside containment, the availability of the high-contain-ment pressure signal (SIS actuation) should be consistent with allowable plant operations (If the signal can be disabled, assume it is unavailable).

The RCP status should be consistent with the plant status.

That is, if any protective feature shuts down the RCP (or its supporting systems) as a result of the steamline break, then assume the RCPs have stopped.

If natural circulation is assumed, justify its con-tinuance considering the RCS status (pressure, temperature and void content from steam and/or non-condensibles).

Each calculation should be done for two cases with and without charging flow (normal makeup). The following parameters of the transients should be determined:

1.

RCS temperature, pressure and pressurizer level.

2.

The status of steam or void formation in the RCS.

3.

The core AT and flow.

4.

The steam generator (broken) secondary water level.

Also provide a listing of the sequence of operator actions nor-mally taken during an RCS shutdown and cooldown which impact the assumptions used in the analyses, or the results.

2008 219 If any operator action is taken subsequent to the steamline break, list those actions, when they are taken, and the indica-tions available to the operator to warn him to take those actions and to confirm them.

Discuss the impact if these actions had not been taken. Also, discuss the impact on system performance and/or accident results if the operator takes improper action in the course of following emergency operating procedures.

2.

Comparison to Current Steamline Break Protection System Compare the results of question (1) to the current steamline break protection system.

For any areas of difference, justify on the basis of relative safety between the two systems.

2008

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N-1 LOOP OPERATI0ft REQUEST FOR ADDITIONAL INFORMATION 1.

Describe the modifications you propose to make to the reactor protection system to facilitate changing of the overpower trip set points when changing to operation modes with less than a full comp-lement of reactor cooling loops (N-1).

Describe how these modif,i-cations will satisfy the requirements of IEEE Std. 279 - 4.15 and Branch Technical Position EICSB 12 (attached) regarding multiple set points and i. hair control, and how administrative control of the control cabinets is aaintained both for the initial modification and for each subsequent change over.

2.

What instrument readings and indications in the control room will change for (N-1) cooling loop mode of operation and what recali-brations, adjustments and temporary notices will be required to avoid anomalous indications as required by IEEE Std. 279 - 4.20?

2008 221

r SEANCH TECHNICAL FCSITICN EIC!3 12 FROTECTION SYSTEM TRIP 70 INT CHAN3ES FCR CEERATICN WITH REACTCR C0CLANT PUMPS CUT GF SERVICE A.

BACK ;CUND F:r the past several years, including a time price to the devel:;mant of IEEE 5:d 279, the staff has retuired aut::stic adjustment to more ttstrictive settings Of tri;s affecting reacter safety by ceans of circuits satisfying the single failure criterien. The basis for this requirement is that the function can be a::: ;11shed m:re reliab*.y by autcmatic circuity than by a human c; erat:r. This design pra::!:e, which has also been adopted inde;endentiv by the national laborat: ries and by much Of industry, served as the basis for paragra;h 4.15, " Multiple set F:ints," cf IEEE Std 279.

M:re recently, all a;plicants have stated that their prote::icn systens were designed to rett IEEE Std 279. Paragraph 4.15 of IEEE Std 279s;ecified that where a e:de of ren:ter 0;erati:n re:uires a more restrictive set ;oint, tne means f:r insuring the r:re restrictive set ;: int shall be ; sitive and must ree the etner re:wiretents f IEEE Std 279. A nur:er Of designs have been pr:seted and a::e;ted which relia:ly ar: sic:ly satisfy this re;uirement. Curing the review Of s me a;;licatiers, h:.ever :er: sin design deficiencies have been f:urd, The pur;;;e Of this ;:siti:n is to ;revide at:iti:nal guidance on the a;;11:ation Of Se:: fen 4.15 cf IEEE Std 279.

E.

SRANCH TECHNICAL PCSITION 1.

If c:re restrictive safety tri; ;oints are re:uired f r ::eration wi*'. a react:r c :lant pues Out of service, and if :;erati:n with a rea:t:r ::clant pu ; cut of service is Of sufficient likelih: d to be a planned :de of 0;erati:n, the change to t..e m:re restrictive tri; points sn:uld ce a:::::11shed aut:tatically.

2.

Plants with designs n:t in ac: rdan:e with the at:ve sh:uld have included in the plant te:hnical s;e:ifications a re:uirement that the reacter be shut down prier to changing the set points manually.

C.

REFERENCES 1.

Report to the ACRS en the prote: tion system trip :cint chan;es f:r 0;eration with the reacter :: lant pum:s cut of service, July 23, 1970.

2.

Mem:ran:wm to R. C. CeY::.n; fr:m V. Stell:, Se;;em:er 14, 1973 (RESAR).

3.

Millst:-e-3 Safe:y Evaluatien Re;;rt, Se: tem:er 24, 1973.

4

!eaver Valley-Z Safety Evaluatier te: Ort, C::::er 10 1973.

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