ML20214S179

Resubmits Application for Amends to Licenses NPF-2 & NPF-8, Increasing Steam Generator Tube Plugging Limit to 10% & Increasing Fq Coeficient to 2.32 for Greater than 50% & 4.64 for Less than or Equal to 50% Rated Thermal Power

ML20214S179
Person / Time
Site:	Farley
Issue date:	06/02/1987
From:	Mcdonald R ALABAMA POWER CO.
To:	NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM)
Shared Package
ML20214S183	List: ML20214S179 ML20214S210 ML20214S233
References
TAC-62283, TAC-62284, NUDOCS 8706090169
Download: ML20214S179 (4)

Text

e A!!bim2 Power Company

. 600 North 18th Street Post Office Box 2641..

' Dirmingham, Alabama 35291-0400 Telephone 205 250-1835 Li%c."'Loen, AlabamaPower ttn southem ehrtrC System A.,..

10CFR50.90 June 2, 1987 3

Docket Nos. 50-348 50-364 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington. D. C.

20555 s

t Gentlemen:

Joseph M. Farley Nuclear Plant - Units 1 and 2

/,

Proposed Steam Generator Tube Plugging Limit and Heat Flux Hot Channel Factor Technical Specification Changes By letter dated August 25, 1986, Alabama Power Company submitted proposed changes to the Technical Specifications which would allow a change in the steam generator tube plugging limit from b% to 10% and a change in the Heat Flux' Hot Channel Factor (Fo) limit from 2.31 to 2.32 for greater than 50% Rated Thermal Power (RTP)' and from 4.62 to 4.64 for less than or equal to 50% RTP. These proposed changes were supported by a revised large break LOCA analysis based on the Westinghouse 1981 ECCS Large Break Evaluation Model with BART.

The version of BART used for this analysis did not account for the effects of control rod thimble filling during reflood or for an inappropriately applied hot assembly power adjustment as discussed in the Westinghouse letter to the NRC (NS-NRC-86-3130) dated June 2, 1986.

The concerns raised by Westinghouse regarding the BART code methodology were recognized by Alabama Power Company in the August 25, 1986 submittal and were addressed qualitatively. However, on January 15, 1987, the NRC Staff informed Alabama Power Company that the proposed l

changes to the Technical Specifications would not be evaluated on the l

basis that the supporting Large Break LOCA analysis had been performed using the uncorrected BART code methodology.

l l

l 8706070169 870602 PDR ADOCK 05000348 P

PDR 00l Ill

U. S. Nuclear Regulatory Commission June 2,1987 l

Page 2 l

Accordingly, Alabama Power Company is resubmitting the proposed changes to the Technical Specifications which would increase the steam generator tube plugging limit from 5% to 10% and which would increase the Fg coefficient from 2.31 to 2.32 for greater than 50% RTP and from 4.62 to 4.64 for less than or equal to 50% RTP. A new large Break LOCA analysis has been performed by Westinghouse for the Farley Nuclear Plant utilizing the 1981 Evaluation Model (WCAP-9220-P-A and WCAP-9221) with f

BASH (WCAP-10266, Rev. 2). The use of the BASH methodology has been l

approved by the NRC Staff in a letter dated November 13, 1986 from Mr.

Charles E. Rossi to Mr. E. P. Rahe, Jr. This submittal supersedes the August 25, 1986 submittal.

Subsequent to the completion of the Farley Large Break LOCA analysis with BASH, Westinghouse notified Alabama Power Company of enhancements to the BASH code and methodology that were made to improve the reliability and performance of the code in certain circumstances. The modifications to the BASH methodology which incorporates these enhancements are described in Addendum 2 to WCAP-10266, Revision 2 and were submitted to the NRC via letter NS-NRC-87-3212 dated March 26, 1987. Alabama Power Company in concert with Westinghouse has evaluated the impact of the BASH code modifications on the Farley Large Break LOCA analysis with BASH and concluded the Farley analysis remains conservative and bounding.

Farley huclear Plant currently has a steam generator tube plugging limit of 5% as shown on Technical Specification Figure 2.1-1.

This limit is based on the Large Break LOCA/ECCS analysis in the FSAR Section 15.4 which assumes 5% steam generator tube plugging. Approximately 2.9% of the steam generator tubes have been plugged in Unit 1 and approximately 3.7% of the steam generator tubes have been plugged in Unit 2.

This level of steam generator tube plugging includes all row 1 tubes in each steam generator. Based on degradation identified during the last Unit 2 inspection (LER 86-004-00), expected tube pluggings during the Fall 1987 Unit 2 Refueling Outage could exceed the current margin to 5%,

especially if F* and sleeving options are not available. Alabama Power Company does not desire to risk a potential delay of plant startup due to this 5% Technical Specification limit should defectise tubes require plugging that exceeds the 5% limit. Therefore, the proposed change to increase the steam generator tube plugging limit to 10f. will provide additional margin to the limit.

In addition, the new large Break LOCA analysis assumes an FQ of 2.40.

The present Fg coefficient of 2.31 for greater than 50% RTP and 4.62 for less than or equal to 50% RTP was required as a result of penalties assessed by the NRC against the 1978 version of the Westinghouse ECCS Evaluation Model. Since the current Small Break LOCA analysis assumes an Fg of 2.32 and the proposed increase in Fg is conservatively bounded by the assumptions of the non-LOCA transient analyses, the proposed changes to increase the FD coefficients of Technical Specification 3.2.2 to 2.32 for greater than 50% RTP and 4.64 for less than or equal to 50%

RTP are consistent with the design / licensing basis for Farley Nuclear Plant.

U. S. Nuclear Regulatory Commission June 2,1987 Page 3 The proposed technical specification changes to the tube plugging limit and FQ are provided in Attachment 1.

Alabama Power Company has determined that the proposed changes do not involve a significant hazards consideration.

In accordance with 10CFRSO.92 a significant hazards evaluation is provided as Attachment 2.

To support these proposed changes to the Technical Specifications, Westinghouse has performed the required Large Break LOCA analysis utilizing the Westinghouse 1981 ECCS Large Br ik Evaluation Model with BASH for Alabama Power Company. A descriptio, of this analysis, including the methodology, assumptions, references and results, is provided in Attachment 3.

The new analysis has calculated a worst-case peak clad temperature (PCT) of 2013*F and confirms that the Farley Nuclear Plant ECCS cooling performance meets the acceptance criteria of 10CFRSO.46. As discussed earlier, modifications to the BASH code methodology were made subsequent to the completion of the Farley Large Break LOCA analysis with BASH. An evaluation of the impact of the BASH modifications on the Farley analysis is provided in Attachment 4.

This evaluation concludes that the Farley Large Break LOCA analysis with BASH remains conservative and bounding.

An additional analysis has been performed to determine the effects on core flow due to steam generator tube plugging. This analysis determined that 10% steam generator tube plugging would not decrease RCS flow below the thermal design flow (T0F) for Farley Nuclear Plant. Pump coastdown characteristics are based on TDF, which does not change for 10% steam generator tube plugging. Therefore, the modeled pump coastdown in the current non-LOCA analyses will not become more severe.

Since the non-LOCA DNB transients are based on TDF, a 10% steam generator tube plugging limit was determined to have no impact on the non-LOCA DNB transients. The effect of 10% steam generator tube plugging upon those non-LOCA accidents which are not DNB related or for which DNB is not the only safety criteria were also evaluated. The only accident analysis of this group which is affected by 10% steam generator tube plugging is the boron dilution analysis. The boron dilution analysis uses the active RCS volume.

Increasing the plugging levels to 10% will reduce the active volume by approximately 4%. However, from the boron dilution analysis done for Farley Nuclear Plant, it can be shown that the RCS active volume can be reduced by more than 4% and operator action time (of at least 15 minutes) will still be met.

Therefore, the 10% tube plugging level for Farley Nuclear Plant will not change the conclusions of the safety analysis. Furthermore, RCS structural integrity is not impacted by the increase in steam generator tube plugging.

In addition, as has already been stated, the licensing basis Small Break LOCA analysis is based on a limiting power shape contained within an envelope of peaking factors with a maximum allowable total peaking factor (F ) of 2.32.

At low steam generator plugging levels (up to O

20%), Small Break LOCA transients would not be affected by the tube

U. S. Nuclear Regulatory Commission June 2,1987 Page _4 plugging. The proposed changes to the Technical Specifications will not impact or invalidate the current licensing basis Small Break LOCA analysis as represented in the Farley FSAR.

Alabama Power Company's Plant Operations Review Committee has reviewed these proposed changes and the Nuclear Operations Review Board will review these proposed changes at a future meeting.

It is requested that these proposed changes be approved by September 18, 1987.

Pursuant to 10CFR170.21, the required License Amendment Application fee of $150.00 was enclosed with Alabama Power Company's August 25, 1986 submittal. A copy of these proposed changes has been sent to Dr. C. E. Fox, the Alabama State Designee, in accordance with 10CFR50.91(b)(1).

If there are any questions, please advise.

Respectfully submitted, ALABAMA POWER OIhh Y f

1 R. P. Mcdonald RPH/ JAR:ks-T.S.7 Attachments cc:

Mr. L. B. Long SWORN TO AND SUBSCRIBED BEFORE ME Dr. J. N. Grace Mr. E. A. Reeves THIS 7 *S' DAY OF

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, 1987 g

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Mr. W. H. Bradford Dr. C. E. Fox K -l>. ] A l' /]W

~ ' Notary Publid Jn, s/,/pq My Commission Expires:

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