ML20099C936

Requests That Encl Be Used as Response to Request for Addl Info Re Draft SER Open Item 34 Re Steamline Break Analysis

ML20099C936
Person / Time
Site:	Vogtle
Issue date:	03/04/1985
From:	Bailey J GEORGIA POWER CO.
To:	Adensam E Office of Nuclear Reactor Regulation
References
GN-540, NUDOCS 8503110479
Download: ML20099C936 (1)

Text

Georgia Power Company Project Management Routa 2. Box 299A Waynesboro, Georgia 30830 Telephone 404 724-8114 404 554 9961 Vogtle Project March 4, 1985 Director of Nuclear Reactor Regulation File: X7BC35 Attention: Ms. Elinor G. Adensam, Chief Log: GN-540 Licensing Branch 44 Division of Licensing U. S. Nuclear Regulatory Commission Washington, D.C. 20555 NRC DOCKET NUMBERS 50-424 AND 50-425 CONSTRUCTION PERMIT NUMBERS CPPR-108 AND CPPR-109 VOGTLE ELECTRIC GENERATING PLANT - UNITS 1 AND 2 REQUET FOR SUPPLEMENTAL INFORMATION DSER OPEN ITEM 34 - STEAMLINE BREAK DNBR

Dear Mr. Denton:

On February 19, 1985, Westinghouse sent a letter (attached) to M.:. Cecil O. Thomas of your staff. As resolution to the VEGP DSER open item 34, it is requested that we be allowed to reference the attached letter.

If your staff requires any additional information, please do not hesitate to contact me.

Sincerely, L.tk./I$

J. A. Bailey Project Licensing Manager JAB /sw Attachment xc: D. O. Poster R. A. Thomas G. F. Trowbridge, Esquire J. E. Joiner, Esquire C. A. Stangler l

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J Westinghouse Water Reactor Nuclear Technology Division Electric Corporation Olvisions 80,333 PittsburghPennsylvania15230 February 19, 1985 NS-NRC-85-30CTI Mr. Cecil 0. Thomas, Chief Division of Licensing U.S. Nuclear Regulatory Comission Washington, D. C. 20555 Ref: 1) Westinghouse Licensing Topical Report, WCAP-9226(P)/9227(NP),

, " Reactor Core Response to Excessive Secondary Stean Releases."

2) Letter from C. O. Thomas, to E. P. Rahe, dated April,1984.

Subject:

Request Ntaber 3 for Additional Infonnation on WCAP-9226(P)/9227(NP) .

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Dear Mr. Thomas:

RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION WCAP 4226 (P) 1Q227 (NP) i i Attached is one (1) copy of the Westinghouse response to the Steamline Break Topical Report (WCAP-9226) question 440.18 (Reference 2) concerning the use of the W-3 CHF correlation below its original pressure range of 1000 to 2300 l psia. As discussed in the attachment, justification for its use is provided by

! the results of an analysis of low pressure data (700 - 1000 psia) using the W-3 correlation.

Please note that this letter will be referenced as part of the license application for Georgia Power Company (Vogtle).

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Mr. Harold R. Denton NS-NRC-85-3007 l

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Please feel free to contact M. P. Osborne, Manager, Plant Transient Analysis l (412/374-4481) if you have any questions concerning this matter. '

l Very truly yours, l

E. P. Rahe, Jr4 Manager Neulear Safety Department Attachment NR/mh

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ATTACHENT QUESTION ON GENERIC STEAMLINE-BREAK TOPICAL REPORT (WCAP-9226 M0.18)

In a letter from T. M. Anderson to J. Stolz, dated May 7,17T9, Westinghouse responded to NRC's questions relating to WCAP-9226. Question 222.9 requested references for the W-3 correlation which cover the applicable range of the calculated coolant parameters.

We have reviewed those references and find them incomplete for the range of coolant parameters calculated in WCAP-9226. The references limit the applicability of the W3 correlation down to 1000 psia. Steamline break analyses, calculated by Westinghouse, have shown decrease in primary system pressure to the range of 500 psia. Provide additional justification for the applicablility of the W3 correlation below 1000 psia. What is the minimm pressure for which the W3 correlation and corresponding 1.3 minimm DNBR limit can be used?

Resoonse:

Reference 1 describes the results of applying the W-3 correlation over its original pressure range,1000 to 2300 psia. The mean measured-to-predicted critical heat flux ratio and sample standard deviation from that analysis are shown in Table 1. -

Reference 2 contains the results of an analysis of low pressure (700 - 1000 psia) data using the W3 correlation. Those data were taken from the same sources as those used in the development of the W3 correlation. As shown in the attached for the figure (taken icw pressure data.from Reference 2), no anomalous behavior is observed The W-3 correlation statistics have been recalculated for the extended database (P = 700 - 2300 psia). The revised statistics are essentially mchanged from the original values (Table 1). The limit DNBR was also recalculated using the method of Owen (Reference 3.) As shown in Table 1, the revised correlation statistics demonstrate that there is a 95% probability with 95% confidence that DNB will not occur if the minimm DNBR is maintained in excess of 131. Again, this value original is essentially mchanged from the limit DNBR associated with the database.

This evaluation demonstrates that extending the pressure range of the W-3 correlation database by a significant amount has a negligible effect on the correlation statistics. Therefore, continued use of the 1.3 limit DNBR for steambreak analyses of Westinghouse pressurized water reactors is justified.

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TABLE 1 - W-3 CHF CORRELATION STATISTICS i

Sample Pressure Range Nunber of Standard (nsis) Data Points Jill Deviation Limit DNBR 1000 - 2300 809 0.9% 0.132 1.30

. . 700 - 2300 885 1.001 0.137 1.31 II i

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, REFERENCES

1) Tong, L. S., " Prediction of Departure from Nucleate Boiling for Axially Non-Unifonn Heat Flux Distribution," J. Ntri ear Energy, Vol. 21, pp 241-248 (1%7) .

2) Prairie Island FSAR Amendnent 20, p.14.2-30, Docket #50-282, August 4, 1972.

3) Owen, D. B., " Factors for One-Sided Tolerance Limits and for Variable Sanpling Plans," SCR-607, March 1%3.

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(Predicted / Measured) DNB Heat Flux vs Pressure Figure 14.2-11 l

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