ML20077E346

Application for Amends to Licenses NPF-2 & NPF-8,revising TS to Reflect Guidance in Draft GL 94-XX, Voltage-Based Repair Criteria for Repair of Westinghouse SG Tubes Affected by Outside Diameter Stress Corrosion Cracking

ML20077E346
Person / Time
Site:	Farley
Issue date:	12/07/1994
From:	Dennis Morey SOUTHERN NUCLEAR OPERATING CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20077E352	List: ML20077E346 ML20077E354
References
NUDOCS 9412120246
Download: ML20077E346 (7)

Text

e Southern Nuctrar Operating Company Post Otfica Box 1295 f

Birmingham, Alabama 35201 l.

Telephone (205) 868 5131 j

L Southern Nudcar Operating Company Dave Morey Vice President Farley Project the Southern eleC!nc System December 7, 1994 Docket Nos.:

50-348 10 CFR 50.90 50-364 U. S. Nuclear Regulatory Commission A'ITN.: Document Control Desk Washington, D. C. 20555 Joseph M. Farley Nuclear Plant Technical Specification Changes Associated With Steam Generator Tube Support Plate Vpitane-Based Repair Criteria Gentlemen:

In Federal Register notice 41520 published on August 12,1994, th: NRC issued draft Generic Letter 94-XX, Voltage-Based Repair Criteria for the Repair of Westinghouse Steam Generator Tubes Affected by Outside Diameter Stress Corrosion Cracking, for comment. Both Farley units currently have steam generator interim tube repair criteria that expire at the end of their current operating cycles. As a result, Southern Nuclear is submitting technical specification amendments for both Farley Units I and 2 reflecting the guidance contained in the draft Generic Letter.

Although some exceptions to the Generic Letter are requested, the technical specifications requested continue to fulfill all safety requirements for a voltage-based steam generator tube repair criteria. NRC approval of the Farley voltage-based criteria is requested by February 1,1995, based on a Unit 2 outage start date of March 10,1995.

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

1.

WCAP-12871, Revision 2, J. M. Far!cy Units 1 and 2 Steam Generator Tube Plugging Criteria for ODSCC at Tube Support Plates, February 1992; 2.

EPRI Report TR-100407, Revision 1, PWR Steam Generator Tube Repair Limits -

Technical Support Document of Outside Diameter Stress Corrosion Cracking at Tube Support Plates; and 3.

Southern Nuclear to NRC letter dated December 9,1993 and associated technical specification amendment and NRC safety evaluation dated April 5,1994.

Additional analyses exist in draft Generic Letter 94-XX, Voltage-Based Repair Criteria for the Repair of Westinghouse Steam Generator Tubes Affected by Outside Diameter Stress Corrosion Cracking.

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I' iU. S. Nuclear Regulatory Commission Page 2 4

Attachment I contains responses to and exceptions taken to the draft Generic Letter. Attachment 2 q

contains the proposed changed technical specification pages in support 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.

Southern Nuclear Operating Company has performed an assessment of the impact of the proposed '

4 revision to the technical specifications on the environment and has determined that there is no impact on the environment.l He proposed revision does not affect the types or amounts of any radiological or non-radiological effluents that may be released o5fsite. No increase in individual or cumulative occupational radiation exposures will result from this revision. Additionally, the revision 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 advise.

Respectfully submitted, l

SOUTilERN NUCLEAR OPERATING COMPANY b!Y ])k; Dave Morey 1

4 REM / cit:NRCVBRC1. DOC 4

SWORN TO AND SUBSCRIBED BEFORE ME Attachments Tills dS DAY OFbN;xIu.), 1994 1

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Af r. S. D. Ebneter Mr. T. A. Reed N/N6, d.ab gu.r#

Mr. B. L. Siegel Notary Publif

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Mr. T. M. Ross Dr. D. E. Williamson My Commission Expires: 7/82c/Jo b [/N97 u

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Responses / Exceptions to Draft Generic Letter 94-XX Guidance 1

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Responses / Exceptions to Draft Generic Letter 94-XX Guidance Southern Nuclear will implement the requested actions of the draft Generic Letter with the following comments / exceptions:

(1)

The inspection guidance discussed in Section 3 of Enclosure 1 of the draft Generic Letter will be implemented with the following responses / exceptions:

3.b.2 / 3.b.3 - Both Farley units were built with copper in the secondary systems. As such, it is difficult to argue that copper is not present in the tube support plate deposits; however, as described in Appendix A Inspection Guidelines submitted by letter dated February 23,1994, no plated copper has been found on the tube outside diameter within the support plate crevice. As noted in the guidelines, no pulled tubes have been identified with copper deposits on the tube at TSP intersections. Therefore, it is not expected that copper interference will significantly influence the TSP signals in the Farley steam generators. If copper interference is observed at either Farley unit, the existing rules and procedures for comp!ying with the technical specifications plugging limit based on depth of wall penetration will apply.

3.b.5 - The definition of"large mixed residuals", i.e., those that could cause a 1-volt bobbin signal to be missed or misread, is broad and difficult to implement in inspection guidelines. Farley will continue to include large mixed residuals in the sample plan implemented under paragraph 3.b.6 of the draft Generic Letter to ensure that unacceptable flaws are not left in service. This is a continuation of the practice first submitted in a letter to the NRC dated February 20,1992 and is included in the latest Appendix A Guidelines submitted to the NRC by letter dated February 23,1994.

3.c.2 - In order to perform data acquisition in a manner consistent with the methode!ogy utilized to develop the voltage limits, bobbin coil probes will continue to be calibrated against the 20% holes in the ASME calibration standard instead of the 100% through wall holes. No benefit has been identified for use of the 100% through wall holes for calibration; however, use of 100%

through wall holes for calibration would introduce an unnecessary inaccuracy for comparison of new eddy current data to historical data by requiring the use of an adjustment factor for all data points currently included in the data bases.

3.c.3 - The requirement for 10% variability on new probes will be implemented provided an NRC/ industry accepted method is agreed upon for satisfying this requirernent.

3.c.4 - The requirement to re-inspect all tubes if the wear measurement exceeds 15% is unnecessary. As acknowledged in the draft Generic Letter, a 5.6 volt repair criterion is justified; however, the repair criterion is limited to 2.0 volts. To require re-inspection of all tubes inspected with a specific bobbin probe if probe wear exceeds 15% is not necessary from a safety standpoint and could affect critical path outage time, l

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Responses / Exceptions to Draft Gencric Letter 94-XX Guidance Page 2 Voltage-Based Repair Criteria As presented at the November 3,1994 meeting between the industry and the NRC Staff, an indication will be re-inspected with a probe meeting the il5% wear limit ifit exceeds 70% of the repair limit, i.e.,1.4 volts for 7/8" tubes.

3.c.6 - Quantitative noise criteria will be included in data analysis procedures. However, it is expected that these criteria will be evoking over the inspection and, as a result, are subject to change.

Inspections will be performed in accordance with the Appendix A guidelines last submitted to the NRC by letter dated February 23,1994.'

(2)

Calculations of the leakage will be per the guidance of Section 2.b of Enclosure 1 of the draft Generic Letter with the following responses / exceptions:

2.b -

The latest description of the calculation of the total leak rate during a steam line break was submitted in the "Farley 1 Cycle 12 Assessment and Projected EOC-13 SLB Leakage" document docketed by letter dated October 4,1994.

2.b.1 - Use of a 0.6 probability of detection and including indications that were not detected by MRPC is unnecessarily conservative. This calculation should be revised to reduce the conservatism. NRC/ industry discussions on this topic are planned for the immediate future. Farley will utilize NRC approved probabilitics of detection or methods for obtaining probabilities of detection, as available at the time analyses are performed. Including indications that are not confirmed by RPC, i.e., RPC NDDs, in the frequency distribution of detected indications is also unnecessarily conservative. This requirement should be revised based on NRC/ industry discussions on this topic.

2.b.3(1)/2.b.3(2) -

The exclusion of data should be resolved by the NRC Staff as soon as possible. No new revelations affecting the exclusion of data is expected.

Farley will utilize the latest NRC approved databases at the time leak rate and burst probabilities are analyzed.

2.b.4 - In order to preclude the possible need for rapid turn around of a technical specification amendment for reactor coolant system specific iodine activity, Farley will revise its technical specification to.5 pCi/ gram.

(3)

Calculation of the conditional burst probability will be per the guidance of Section 2.a of of the draft Generic Letter with following responses / exceptions:

2. -

A limit of I x 102 on the calculated conditional burst probability of burst based on the projected EOC voltage distribution is too restrictive. The goal for probability of burst should be 2.5 x 10 2 based on guidance contained in NUREG-0844, NRC Integrated Program for the Resolution of Unresolved Safety issues A-3, A-4, and A-5 Regarding Steam Generator Tube Integrity. The 2.5 x 10-2 goal was voluntarily implemented by Farley in WCAP-12871, Revision 2, J. M. Farley Units 1 and 2 Steam Generator Tube Plugging Criteria for ODSCC at Tube j

Support Plates, February 1992, wiiinut comment from the NRC Staff.

4 Responses / Exceptions to Draft Generic Letter 94-XX Guidance Page 3 Voltage-Based Repair Criteria 2.a -

he latest description of the calculation of the probability of burst during a steam line break was submitted in the "Farley 1 Cycle 12 Assessment and Projected EOC-13 SLB Leakage" document docketed by letter dated October 4,1994. He analysis methods are Monte Carlo analyses that account for parameter uncertainties.

2.a.1 - The exclusion of data should be resolved by the NRC Staff as soon as possible.

No new revelations affecting the exclusion ofdata are expected. Farley will utilize the latest NRC approved databases at the time leak rate and burst probabilities are analyzed.

(4)

The operational leakage limits identified in Section 5 of Enclosure 1 of the draft Generic Letter are impicmented by the proposed technical specification amendments. (Note: Unit I will remain at 140 gallons per day through only one steam generator as previously approved by the NRC.)

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(5)

Farley leakage monitoring measures provide guidance on trending and response to rapidly J

increasing leaks. 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 installation of N-16 monitors on both units.

Farley continues to participate in the industry effort for developing primary-to-secondary leakage guidelines. Upon NRC concurrence with the industry guideline document on leakage monitoring, Farley will implement the industry guidelines.

(6)

Tube pull guidance of Section 4 of Enclosure 1 of the draft Generic Letter will be followed with the following responses / exceptions:

4.a -

A total of 5 steam generator tube intersections from Farley Unit I and 13 from Farley Unit 2 have been pulled and evaluated since 1985. Included in the alternate repair criteria's databases are 3 intersections from Farley Unit I and 6 intersections from Farley Unit 2. Two intersections from Unit I with relatively

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high voltages, i.e.,3.3 and 3.2 volts, were removed in 1992 and destructively i

examined to support Farley interim plugging criteria / alternate repair criteria applications. The existing tube pull results substantially confirm that the dominant flaw feature affecting the integrity at Farley is axially oriented outside diameter stress corrosion cracking.

The industry has submitted a proposed industry tube pull program. It is Farley's intent to participate in the industry tube pull program once it is approved by the NRC Staff. However, Farley sees no benefit to pulling additional tubes until the tube pull program is endorsed by the NRC.

1 As always, if unusual indications are found in the Farley steam generators, tube pulls will be considered.

i (7)

Results will be reported per the guidance of Section 6 of Enclosure 1 of the draft Generic Letter.

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' Responses / Exceptions to Droft Gencric Letter 94-XX Guidance Page 4

, Voltage-Based Repair Criteria.

t-1 Additional requested information:

1.b.1 - Conceming the deformation or collapse of steam generator tubes following a loss I

of coolant accident plus a safe shutdown earthquake event, a Farley specific analysis was docketed under WCAP-12871, Revision 2 dated February 1992. As a result of this analysis, no tubes will be excluded from using the voltage repair criteria.

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