ML20150A641

Forwards Response to 880225 Request for Addl Info Consisting of Six Questions on NUSCO-151.Info Will Enable NRC Complete Review of NUSCO-151

ML20150A641
Person / Time
Site:	Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png
Issue date:	03/08/1988
From:	Mroczka E CONNECTICUT YANKEE ATOMIC POWER CO.
To:	NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM)
References
B12594, NUDOCS 8803150361
Download: ML20150A641 (4)

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CONNECTICUT YANKEE AT O MIC POWER COMPANY B E R L I N. CONNECTICUT P O BOX 270 e HARTFORD. CONNECTICUT 061414270 March 8, 1988 T Et t PHONE

'"" Docket No. 50-213 812594 Re: 10CFR50.46 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D.C. 20555 Gentlemen:

Haddam Neck Plant Response to Request for Additional Information on NUSCO-151 On February 25, 1988, a request for additional information consisting of six questions on NUSCO-151 was forwarded to Connecticut Yankee Atomic Power Company (CYAPC0). Attachment 1 provides our response to each of these questions.

We trust this information will enable the NRC staff to complete their review of NUSCO-151. Should you have any further questions please contact us.

Very truly yours, CONNECTICUT YANKEE ATOMIC POWER COMPANY H ~J E.NfHroczkae' Senior Vice President cc: W. T. Russell, Region I Administrator A. B. Wang, NRC Project Manager, Haddam Neck Plant J. T. Shedlosky, Resident Inspector, liaddam Neck Plant d803150361 DR soo30s ADOCK 05000213

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Docket No. 50-213 B12594 Attachment 1 Response to Request for Additional Information on NUSCO-151

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March, 1988 i

ATTACHMENT 1 Responses to Request for Additional Information

1. Do you have a CYAPC0 submittal dated (roughly) July 14, 1986 which reportedly presented revised minimum AFW flowrates (given as Table 2) used in loss of Normal Feedwater analysis? We would like to get a copy of this submittal.

The revised minimum AFW flowrates were contained in an October 14, 1986 CYAPC0 submittal. Appendix A is a copy of this submittal.

2. Provide the values of peak pressurizer pressures for 4-loop and 3-loop '

operation (Figs. 4.12-1 and 4.12-8) in the March 10, 1987 submittal-Reanalysis of Loss of Normal Feedwater.

The peak pressurizer pressures are 2527 psia and 2528 psia for 4-loop and 3-loop operation, respectively.

3. In addition, provide the pressurizer safety valve setpoint value actually

used in the above analysis. Need a statement to the fact that the safeties lifted during the event.

The pressurizer safety valve setpoint used was 2525 psia. Table 4.12 of the March 10, 1987 submittal provides the time at which the pressurizer safety valves first opened.

4. Provide the times at which MDNBR was computed for 4-loop and 3-loop cases to support their statement that AFW flow does not impact the MONBR since it comes on later.

Steady state cases were run to calculate the MDNBR. These runs utilized conservative core temperature, pressure, flow and power parameters. The values were based on the RETRAN calculations from the initiation of the event until the control rods were fully inserted. The times for this were ten and eleven seconds for 4-loop and 3-loop operation, respective-ly.

5. Were PORVs credited during this event?

The analyses did not credit the pressurizer PORVs. Only the pressuriter safety valves were modeled.

6. What is the SG mass inventory required to keep the tube sheet covered?

The criterion utilized in the analyses to determine adequate inventory to keep the tube sheet covered was 500 lbm.

O Docket No. 50-213 B12594 Appendix A l Transmittal of Reference l on Auxiliary Feedwater Flow Rates l

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ATOMIC POWE R C OM P A tJ Y C CONNECTICUT YANKEE f tLIN%t P o Box 270 B E R LIN CONNECTICUT HARif o4C CONNECT CUT C4141-C270 203 465 5000 October 14, 1986 Docket No. 50-213 A05679 Of fice of Nuclear Reactor Regulation Attn: Mr. Christopher 1. Gritnes, Director Integrated Safety Assessment Project Directorate D; vision of PWR Licensing - B U.S. Nuclear Regulatory Commission Wr.hington, D.C. 20555 References (1) J. F. Opeka letter to C.1. Grimes, Response to Request for Additional Information Concerning the Auxiliary Feedwater System, dated June 30,1936.

Centlement Haddam Neck Plant Auxiliary Feedwater Flowrates in Reference (1) a commitinent was made to provide a table of revised minimum expected auxiliary feedwater flowrates as determined by calculations which were being refined. As such, the intent was to provide an update to Table 2 or the enclosure included as part of Reference (1). Attachment 1 provices this update. Because these revisions, in some cases, are less conservative !!owrates than those assumed in design basis analyses, as documented in our June 30,19S6 letter,(1) a revision will tse provided to the affected design basis analyses on or before February 15,1987.

Very truly yours,  ;

CONNECTICUT YANKEE ATOMIC POWER COMPANY S %.DrA J. F. Opeks '

Senior Vice IMsident l L o.vL By: C. F. Sears Vice President (1) See J. F. Opeks letter to " 1. Grimes, dated June 30,19S6 which forwarded the reanalysis for non-LOCA design basis accidents.

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Attachment 1 Minirnutn Expected AFW Flowrates Total Flow to AllIntact Steam Generators (SGs)-

One Pump Two Pump Condition Op(eration GPM) Op(eration GPM) -

1 Rupture SG and 3 Intact SGs @ 1000 PSIG O(l) O(l) 4 Intact SGs @ 1000 PSIG ,297.8 447.4 3 Intact SGs Q 1000 PSIG, One isolated 283.5 409.6 2 Intact SGs Q 1000 PSIG, Two isolated 253.0 341.3 1 Intact SG @ 1000 PSIG, Three isolated 179.6 213.2 (1) All flow is assumed to be lost through the rupture.

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