ML20138G945

Application for Amends to Licenses NPF-2 & NPF-8,reflecting Guidance in GL 95-05, SG Tube Support Plate Voltage-Based Repair..., to Use Revised Accident Leakage Limit of 20 Gpm & Include 50% of Rpc Ndds in Calculation.W/O Unit 2 TSs

ML20138G945
Person / Time
Site:	Farley
Issue date:	12/26/1996
From:	Dennis Morey SOUTHERN NUCLEAR OPERATING CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20138G950	List: ML20138G945 ML20138G958
References
GL-95-05, GL-95-5, NUDOCS 9701030074
Download: ML20138G945 (7)

Text

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'u- ' D;ve Morey S:uthern Nuclear ,

l-d Vice President Opera 2ng Compzy  !

Farley Project P.O. Box 1295

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Birrningham. Alabama 35201

' Tel205.992.5131 I. '

! December 26, 1996 SOUTHERNM COMPANY Energyto Sen4YourWorld*

! Docket Nos.: 50-348 50-364 l

U. S. Nuclear Regulatory Commission

. ATTN: Document Control Desk -

Washington, DC 20555 l l >

Joseph M. Farley Nuclear Plant Steam Generator Tube Suonort Plate Voltaae-Based Repair Criteria t

Ladies and Centlemen:

- On August 3,1995, Generic Letter 95-05, Voltage-Based Repair Criteria for Westinghouse Steam j Generator Tubes Affected by Outside Diameter Stress Corrosion Cracking, was issued. Farley l Unit I has an interim tube repair criteria that expires at the end of the current operating cycle. As ,

a result, Southern Nuclear is submitting a technical specification amendment for Unit I reflecting ,

the guidance contained in the Generic Letter. Although some clarifications to the Generic Lette r i requirements are proposed, the technical specification requested continues to fulfill all safety .,

requirements for a voltage-based steam generator tube repair criteria. NRC approval of the Farby voltage-based criteria is requested by March 1,1997, based on a Unit 1 outage start date of March 14,1997. A similar amendment has already been approved for Unit 2.  !

In addition to the Unit I tec' aal specification amendment, Southern Nuclear proposes the  ;

following for Farley Units 1 and 2: ,

1. Use of a revised accident leakage limit of 20 gpm; I

2. Inclusion of 50% of the RPC NDDs in the calculation instead of 100%; and

3. Use of probability of detection that is voltage dependent, called POPCD (probability of prior cycle detection), instead of a constant 60%.

The safety analyses to support this amendment have been previously docketed. These analyses include -

1. WCAP-12871, Revision 2, J. M. Farley Units I and 2 SG Tube Plugging Criteria for I ODSCC at Tube Support Plates, February 1992;

2. EPRI Report TR-100407, Revision 1, PWR Steam Generator Tube Repair Limits - 1 Technical Support Document of Outside Diameter Stress Corrosion Cracking at Tube j Support Plates; and l l 3. Southern Nuclear to NRC letter dated December 9,1993, and asp + ' technical specification amendment and NRC safety evaluation dated April - ~4.

l , l i 9701030074 961226 PDR ADOCK 05000348 fi()D l i P PDR 010034

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U. S. Nuclear Regulatory Commission Page 2 l

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Additional analyses exist in Generic Letter 95-05. l Attachment I contains responses to the Generic Letter. Attachment 2 contains the proposed  ;

changed tachale=! specification pages in suppon of the voltage-based plugging criteria.

Attachment 3 provides a significant hazards evaluation for the proposed voltage-based repair criteria in accordance with 10 CFR 50.92. Attachments 1,2, and 3 are applicable ta Farley Unit 1 only, AM=chment 4 provides justification for use of the revised accident leakage limit, 50% of the RPC NDDs, and the voltage dependent probability of detection. Attachment 4 is applicable to both Farley Units I and 2.

Southern Nuclear has performed an assessment of the impact of the proposed revision to the ted: kal specifications on the environment and has determined that there is no impact on the environment. The proposed revision does not affect the types or amounts of any radiological or non-radiological effluents that may be released offsite. No increase in individual or cumulative occupational radiation exposures will result fror: this revision. Additionally, the revis:an does not involve the use of any resources not previously considered in the Final Environmental Statement related to the operation of Farley Nuclear Plant.

A copy of these proposed changes is being sent to Dr. D. E. Williamson, the Alabama State Designec, in accordance with 10 CFR 50.91(b)(1).

if there are any questions, please ad5ise.

Respectfully submitted, SOUTHERN NUCLEAR OPERATING COMPANY

/1  ??wu Dave Morey REM:maf fnplarcl. doc Attachments SWORN TO AND SUBSCRIBED BEFORE ME cc: Mr. S. D. Ebneter THIS o7[ DAY OF " I w l996 Mr. J.1. Zimmemian Mr. T. M. Ross Dr. D. E. Williamson Notary Putilic Mr. T. A. Reed My Commission Expires: / /,/O/ /9 7

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8 Attachment 1 Responses :o Generic letter 9545 Guidance ,

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, , Responses to Generie Letter 95-05 Guidance 1

l Southern Nuclear Operating Company (SNC) will implement the requested actions of Generic Letter 95-05 l l with the following comments: i (1) He applicability requirements discussed in Section I of Attachment I of Generic letter 95-05 will l be implemented.

1.b.1 - Concerning the deformation or collapse of steam generator tubes following a loss of coolant accident plus a safe shutdown earthquake event, a Farley specific analysis was docketed under WCA.P-12871, Revision 2 dated February 1992. As a result of this analysis, no tubes will be excluded from using the voltage repair criteria. l l

l (2) He inspection guidance discussed in Section 3 of Attachment 1 of the Generic Letter will be implemented in accordance with the Appendix A guidelines last submitted to the NRC by letter dated February 23,1994, and with the following responses:

3.b - SNC will utilize a motorized rotating coil probe, e.g., pancake or + Point, instead of specifying a rotating pancake coil. His wording change is made to ensure that the + Point i probe can be used as an attemative to the rotating pancake coil.

3.b.1 - SNC will inspect all bobbin flaw indications with voltages greater than 2.0 volts with a motorized rotating coil probe.

l 3.b.2 - SNC will inspect all intersecticas where copper signals interfere with the detection of flaws with a motorized rotating coil probe. Any indication found with the motorized rotating coil will result in repair of the tube.

3.b.3 - SNC will inspect all intersections with dent signals greater than 5 volts with a motorized rotating coil probe. If circumferential cracking or primary water stress corrosion cracking is detected at the tube support plates intersections, a sampling plan will be implemented in I accordance with the PWR Steam Generator Tube Exammation Guidelines, Resision 4. If  !

indications are found at dents with voltages near 5 volts, the flaw will be characterized. If l the flaw exceeds the structural requirement of Regulatory Guide 1.121, the sampling plan I will be expanded to intersections with dents less than 5 volts. If the flaw is evaluated as not significant, the sampling plan will not be expanded.

3.b.4 - SNC will inspect all intersections with large mixed residuals that could be expected to mask a 1.0 volt bobbin flaw signal with a motorized rotating coil probe. l 3.c.2 - The *10% limit on new probe variability will be implemented using the guidance included in Nuclear Energy Institute to NRC letter dated January 23,1996, conceming "New Probe Variability for Use in the ODSCC Alternate Repair Criteria, as discussed in the NRC to the Nuclear Energy Institute letter dated February 9,1996. Furthermore, SNC will venfy l

j that both the primary and mix frequencies will meet the 110% variability requirement-l I

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. ,- , Responses to Draft Generic leter 95-05 Guidance Page 2

• Voltage-Based Repair Criteria

._ 3.c.3 - De limits on probe wear will be implemented using the guidance included in Nuclear

. Energy to NRC letter dated January 23,1996, conceming " Eddy Current Probe Replacement Criteria for Use in ODSCC Alternate Repair Criteria"; as discussed in the NRC to the Nuclear Energy Institute letter dated February 9,19%; NEI to NRC letter dated February 23,1996; and NRC to NEl letter dated March I8,1996. De following summarizes this guidance as agreed to by the NRC Staffi e For all tubes identified with indications above 1.5 volts (i.e.,75% of the 2 voit repair limit for 7/8 inch tubes is 1.5 volts) since the last successful probe wear check (< 15% wear), the whole tube (i.e., all hot-leg tube support plate intersections to the lowest cold leg TSP intersection with known ODSCC) will be re-inspected with an acceptable probe (<l5% probe wear) and all oddy current data from the acceptable probe will be evpluated. If a large indicatmn (greater than approximately i voit for 7/8 inch tubes) is detected which was previously missed with the failed probe, an assessment of the significance will be performed during the outage. His assessment, along with the description of actions taken, will be provided to the NRC in the 90-day report.

De inspection described above will be modified slightly for tubes which would require a double entry to inspect the entire tube. For low row tubes in which the U-bend radms precludes passing a full size bobbin coil over the U-bend or for tubes with sleeves which preclude passing a full size bobbin through the sleeve, the portion of the tube with the indication above 75% of the repair limit will be re-inspected. De second entry for inspection of the remamder of the tube is not required provided there is not an indication above 75% of the repair limit.

• Actions will be taken to minbuu the potential for tubes to be inspected with probes that fail the probe wear check. His includes replacing a probe immediately upon finding that it fails d.a probe wear check.

• If a probe fails prior to peiforming a probe wear check, it will be assumed that the probe failed the probe wear check and the probe wear criteria approved by l the Staff will be followed. ,

j e ne effects of probe wear will be explicitly assessed as a potential l contributing factor if significant differences between the actual and end-of cycle l' projections exist in the 90-day report.

. The 90-day report will address if a non-proportionate number of new l indications have been detected in tubes which were inspected in the previous j outage with a probe that failed the probe wear check.  ;

3.c.4 - Data analysts will be trained and qualified in the use of the analyst's guidelines and  !

procedures. At Farley Nuclear Plant, a minimal number of analysts are used for  :

deternunation of voltage. He use of a small number of analysts is intended to minimize j the effect of analyst variability on determination ofgrowth rate, resulting in as accurate a prediction for the next operating cycle as possible. We believe this results in a more  !

accurate growth rate determination; however, it is time consuming and can result in ,

difficulty in performing the calculations prior to returning the steam generators to senice..  !

t 3.c.5 - Quantitative noise criteria have historically been applied and will be incorporated in the j Farley Nuclear Plant Data Acquisition procedures. His enables noise levels due to ,

electrical noise, tube noise, calibration standard noise, etc., to be addressed at the initial l

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. Responses to DraA Generic letter 95-05 Guidance Page 3 l Voltage-Based Repair Criteria

. . point ofir,spection which has nummized the need for re-inspection. Probes are typically .

replaced prior to exceeding the noise criteria. If, upon measurement, the probe in use fails to meet the criteria, tubes tested with that probe since the last satisfactory measurement are re-inspected in addition, the Farley Nuclear Plant Analysis procedures allow the analyst  !

to require re-inspection due to noise on a " qualitative" basis. l l

3.c.6 - Data analysts will review the mixed residuals on the standard itself and take actions as necessary to mainuze these residuals.

l 3.c.8 - Data analysts will be tramed on the potential for pnmary water stress corrosion crackmg l to occur at tube support plate intersections. The discovery of PWSCC at tube support l plate intersections will be reported to the NRC Staff prior to startup. l (2) Calculations of the main steam line break leakage will be per the guidance of Sections 2.b and 2.c j

'of Attachment 1 of Generic letter 95-05 with the following responses  :

i 2.b - Calculations performed in support of the voltage-based repair criteria will follow the l probabilistic methodology described in WCAP-14277, SLB leak Rate and Tube Burst Probability Analysis Methods for ODSCC at TSP Intersections, January 1995.

2.b.2(1) - No distribution cutoff will be applied to the voltage measurement variability distribution.  ;

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- 2.b.4 - In order to preclude the possible need for rapid tum around of a technical specification j amendment for reactor coolant system specific iodine actisity, the Farley technical  :

specification will remain at 0.5 pCi/ gram. A leakage limit for Farley Unit I and 2 of 20 .

gpm is justified in Attachment 4.

2.c - Reference is made to the use of an RPC probe. SNC will utilize a motorized rotating coil  !

probe, e.g., pancake or + Point, instead of specifying a rotating pancake coil. It is SNC's l intent (and desire) to always perform the calculations on the projected EOC distributions. }

In the event that the growth rate determmations cannot be completed prior to retuming the j steam generators to service, the calculations will be based on the actual EOC distributions  !

as allowed in Section 2.c. However, even if the calculation made prior to returning the .!

steam generators to service is based on the actual measured voltage distribution, the '

calculation based on the projected EOC voltage distribution will be provided to the NRC in the 90 day report following the outage.

(3) Calculation of the conditional burst probability will be per the guidance of Section 2.a of Attachment I of Generic Letter 95-05 with the following responses:

2.a - Calculations performed in support of the voltage-based repair criteria will follow the methodology described in WCAP-14277, SLB Leak Rate and Tube Burst Probability Analysis Methods for ODSCC at TSP Intersections, January 1995.

2.a.2 - The upper voltage repair limit will be determined 2 months prior to the outage using the most recently approved NRC database The database proposed by NEI letter dated

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. . Responses to DraA Generie letter 95-05 Guidance Page 4 Voltage-Based Repair Criteria

. . September 18,1996, will be used for the Unit I outage. Following this outage, the database used will be based on the NRC/ industry protocol.

l (4) Farley leakage monitoring measures provide guidance on trending and response to rapidly increasing Icaks. Guidance is provided not only for the absolute leakage measured, but also on the rate of change of the leak rate. Timely detection ofleaks is ensured by the N-16 monitors on both units. Farley has also implemented the guidelines contained in EPRI topical report "PWR Primary-to-Secondary leak j l Guidelines," EPRI TR-104788, May 1995.-

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t (5) Tube pull guidance of Section 4 of Attachment I of Generic letter 95-05 will be followed.

N) Results will be reported per the guidance of Section 6 of Attachment 1 of Generic Letter 95-05.

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Attachment 2 Unit i Technical Specification Pages P

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