ML20207T179

Provides Background,Specifies Safety Analyses Scope & Summarizes Current Results & Conclusions Re Operation of Plant at Higher Steam Generator Tube Plugging Levels,Per NRC Request & Util 861210 Tech Spec Change Request

ML20207T179
Person / Time
Site:	Summer
Issue date:	03/17/1987
From:	Nauman D SOUTH CAROLINA ELECTRIC & GAS CO.
To:	Harold Denton Office of Nuclear Reactor Regulation
References
NUDOCS 8703230285
Download: ML20207T179 (5)

Text

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10CFR50.36 na Electric & Gas Company n A. n

.. gth Nuclear Operations Co ugbg,29218

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March 17, 1987 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D. C. 20555

Subject:

Virgil C. Summer Nuclear Station Docket No. 50/395 Operating License No. NPF-12 Heat Flux Hot Channel Factor (FQ)

Technical Specification Change

Dear Mr. Denton:

In a letter from Mr. D. A. Nauman to Mr. H. R. Denton dated December 10, 1986, South Carolina Electric & Gas Company requested an amendment to the Virgil C. Summer Nuclear Station's (VCSNS) Technical Specifications (T.S.),

Section 3/4.2.2, " Heat Flux Hot Channel Factor - FQ (E)," and its bases. The

. proposed T.S. change would lower the maximum allowable Fq from 2.32 to 2.25 beginning with Cycle 4 and is requested to support ongoing safety analyses and evaluations to justify steam generator (SG) tube plugging levels in excess of the current analysis value of 6%. Specifically, the FQ reduction is intended to create additional peak clad temperature margin in the large break LOCA analysis which will be used to offset the increase in peak clad temperature due to higher SG tube plugging.

Based upon subsequent technical review, your staff has requested additional information regarding the ongoing safety analyses and evaluations. The attachment to this letter provides some background, specifies the safety analyses scope, and summarizes the current results and conclusions. This information is preliminary pending completions of the engineering work in accordance with SCE&G and Westinghouse quality assurance requirements. On a preliminary basis, however, the results and conclusions of the safety evaluation are expected to permit SCE&G to perform an evaluation of steam generator tube plugging up to 16%, in accordance with 10CFR50.59.

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Mr. Harold R. D nton March 17, 1987 Page 2 The statements and' matters set forth in this submittal are true and correct to the best of my knowledge, information, and belief. In addition, SCE&G will provide you additional information should these preliminary results or conclusions change, following completion of the safety analyses and evaluations.

If you should have any question, please advise.

Very truly yours,

. A. Nau an DAN /wfp Attachment c: 0. W. Dixon, Jr./T. C. Nichols, Jr.

E. C. Roberts

0. S. Bradham J. G. Connelly, Jr.

D. R. Moore W. A. Williams, Jr.

J. Nelson Grace Group Managers W. R. Baehr C. A. Price C. L. Ligon (NSRC)

R. M. Campbell, Jr.

K. E. Nodland R.'A. Stough G. O. Percival R. L. Prevatte J. B. Knotts, Jr.

H. G. Shealy NPCF File

-Attachment to Mr. Harold R. D::nton Letter March 17, 1987 Page 1 of 3 SAFETY EVALUATION SUPPORTING ' OPERATION OF VIRGIL C. SLSOER NUCLEAR STATION AT HIGHER STEAM GENERATOR TUBE PLUGGING LEVELS

. BACKGROUND The LOCA and non-LOCA safety -analyses supporting operation of the Virgil C.

Summer Nuclear Station (VCSNS) are provided in Chapter 15 of the FSAR and Reference 1. The analyses and supporting evaluations are valid for steam generator (SG) tube ' plugging levels up to 6%. The most limiting analysis condition is the Large Break LOCA accident which produces a maximum peak clad 4

temperature of 2189.2*F based on a LOCA FQ of 2.32.

Because the model D3 steam generators at VCSNS have experienced primary water stress corrosion cracking, tube plugging levels in excess of 6% are a possibility. Consequently, SCE&G initiated a safety evaluation to justify higher levels of SG tube plugging, should they be required in Cycle 4 or subsequent cycles.

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i OBJECTIVES OF THE SAFETY EVALUATION

1. Estimate the' SG tube plugging level for which RC flow measurements should continue to verify thecurrentthermaldesignflow(TDF).

2. Confirm the validity of current VCSNS Non-LOCA safety analyses for tube plugging levels which maintain TDF.

3. Demonstrate conformance to 10CFR50.46 for SG tube plugging levels which maintain TDF.

4. Complete the safety analyses and evaluations to support a potential 10CFR50.59 ' evaluation for SG tube plugging in excess of the current safety analysis value of 6%.

r PRELIMINARY RESULTS

1. TDF - Plugging Limit Using a flow measurement uncertainty of 2.1% (Reference 2), field verification of TDF is expected to be successful for SG tube plugging levels up to 16%.

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Attachment to Mr. Harald R. Dent::n Litter March 17, 1987 Page 2 of 3

2. Asymmetric Plugging In lieu of more detailed analyses, the maximum plugging differential between any two SG's will be limited to 10% or 467 plugged tubes in order to maintain the validity of simplifying assumptions (e.g.,

complete reactor vessel mixing) within the VCSNS safety analyses.

3. Non-LOCA Safety Analyses (Chapter 15'- FSAR)

SG tube plugging up to 16% with no reduction in TDF will not adversely impact the conclusions of the non - LOCA safety analyses.

4. Large Break LOCA The limiting large break (DECLG, CD=0.4) was reanalyzed using an approved ECCS Evaluation Model (References 3 & 4) with 16% SG tube plugging. To offset the peak clad temperature (PCT) penalty for increased plugging, the LOCA FQ was reduced from 2.32 to 2.25. In addition, the reanalysis included two plant specific input changes: a lower initial fuel pin pre-pressurization and a lower locked rotor K-factor in WREFLOOD. Both input changes and the use of the most recent BART-WREFLOOD revision (References 5 & 6) also tended to decrease PCT.

The analysis results at 16% SG tube plugging meet the ECCS acceptance criteria of 10CFR50.46. The calculated PCT is 2023.6*F.

5. Small Break LOCA Based on both WFLASH and NOTRUMP analyses on Westinghouse plants, the small break analysis results are insensitive to SG tube plugging levels up to 20%. Therefore, VCSNS current small break analysis using WFLASH will remain valid for plugging levels up to 16%.

CONCLUSIONS Based on the results of this safety evaluation, SG tube plugging levels up to 16% can be justified from a safety analysis standpoint beginning with Cycle 4, provided:

1. The total reactor vessel flow must be equal to or greater than 288,600 gpm, the current TDF.

2. The VCSNS Technical Specifications are revised to decrease the RC flow measurement uncertainty from 3.5% (current value) to 2.1%.

3. The VCSNS Technical Specifications are revised to decrease the LOCA FQ from 2.32 (current value) to 2.25.

4. The maximum difference in percent (%) tube plugging between any two steam generators does not exceed 10% or 476 plugged tubes.

Attachment to Mr. HIrold R. Denton trtt"r March 17, 1987 Page 3 of 3 REFERENCES

1. SCE&G 1etter, O. W. Dixon, Jr. to H. R. Denton (USNRC), " Reactor Coolant -

System Flow", March 6, 1985.

~2. SCE&G 1etter, D. A. Nauman to H. R. Denton (USNRC), " Reactor Coolant System Flow", June 27, 1986.

3. Rahe, E. P., " Westinghouse ECCS Evaluation Model, '1981 Version", WCAP-9220-P-A, Revision 1, 1981.

4. Young, M. Y. et al . , "BART-AI: A Computer Code for the Best Estimate Analysis of Reflood Transients", WCAP-9561-P-A and Addendum 3, Revision 1.

5. Westinghouse letter, E. P. Rahe to D. G. Eisenhut (USNRC) NS-NRC 3147, July 16, 1986.

6. USNRC letter, C. E. Rossi to E. P. Rahe, "Exceptance For Referencing of Licensing Topical Report WCAP-9561", Addendum 3, Revision 1, August 25, 1986.

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