ML20113D122

Submits Response to RAI Re 960329 Submittal for Permanent Alternative SG Tube Support Plate voltage-based Repair Criteria IAW GL 95-05

ML20113D122
Person / Time
Site:	Farley
Issue date:	06/27/1996
From:	Dennis Morey SOUTHERN NUCLEAR OPERATING CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
GL-95-05, GL-95-5, NUDOCS 9607030025
Download: ML20113D122 (5)

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Southern Nucisar Op: rating Company

,. Post Offica Box 1295

• Birmingham. Alabama 35201 Telephone (205) 868 5131 L k Dave Morey Southern Nudear Operating Company Vice President Farley Project the Southem electnc system June 27, 1996 l

Docket No.: 50-364 U. S. Nuclear Regulatory Commission ATTN: Document ControlDesk Washington, DC 20555 Joseph M. Farley Nuclear Plant - Unit 2 j Responses to Request for AdditionalInformation Ladies and Gentlemen:

By letter dated May 30,1996, the NRC requested additional information concerning Southern Nuclear's March 29,1996 submittal for a permanent alternative steam generator tube support plate voltage-based repair criteria in accordance with Generic Letter 95-05.

Responses to the request for additional information are attached.

If there are any questions, please advise.

Respectfully submitted, )

$f blev Dave Morey Sworn to andsubscribedbefor me thisd 9 day of AU) 1996 W Olu My Commission i es t

O Yb/WV REM / cit:NRCRAIR1. DOC Attachment cc: Mr. S. D. Ebneter, Region II Administrator Mr. B. L. Siegel, NRR Senior Project Manager Mr. T. M. Ross, FNP Sr. Resident Inspector g Dr. D. E. Williamson, State Department ofPublic Health  %\

Mr. T. A. Reed, NRR - Materials and Chemical Engineering Branch 9607030025 960627 PDR ADOCK 05000364 \

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ATTACHMENT Responses to the Request for Additional Information Related to the Farley Unit 2 Steam Generator Tube Support Plate Voltage-Based Repair Criteria Technical Specification Amendment i

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ATTACHMENT Responses to the Request for AdditionalInformation Related to the Farley Unit 2 Steam Generator Tube Support Plate Voltaae-Based Repair Criteria Technical Specification Amendment

1. On page 2, you stated that calculations of the main steam line break leakage will be based on the guidance of Section 2.b of Attachment I to Generic Letter (GL) 95-05. You provided responses to Sections 2.b,2.b.2(1) and 2.b.4, but not 2.b.1,2.b.2,2.b.2(2),2.b.3, 2.b.3(1) and 2.b.3(2). Please provide clarification regarding whether or not you will comply with those sections of GL 95-05 for which a response was not provided.

Southern Nuclear Response: Farley Nuclear Plant will couply with sections 2.b.1, 2.b.2, 2.b.2(2), 2.b.3, 2.b.3(1), and 2.b.3(2) of Generic Letter 95-05.

2. In your response to Section 2.b, it was stated that calculations performed in support of the voltage-based repair criteria will follow the methodology described in WCAP-14277.

WCAP-14277 specifies deterministic and probabilistic methods; however, the NRC has approved the probabilistic method, but not the deterministic method. Please provide clarification as to how this WCAP is being used.

1 Southern Nuclear Response: As stated in WCAP-14277, deterministic analyses may be '

performed for screening purposes. However, for all regulatory commitments, the probabilistic methodology that has been approved by the NRC will be used.

3. Provide clarification to your response to Section 2.c, regarding whether you intend to use a projected end-of-cycle (EOC) distribution or actual measured bobbin voltage to perform the probabilistic calculations.

Southern Nuclear Response: It is Southern Nuclear's intent (and desire) to always perform the calculations on the projected EOC distributions. In the event that the growth rate determinations cannot be completed prior to returning the 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.

4. Confirm that in Sections 3.b.2,3.b.3, and 3.b.4 of Attachment I to GL 95-05, your intent is that "any indications found at such intersections with rotating pancake coil should cause the tube to be repaired." It should be noted that this is a condition of the GL 95-05 alternate repair criteria.

i Southern Nuclear Response: Concerning Sections 3.b.2,3.b.3, and 3.b.4 of Attachment 1 of Generic Letter 95-05, any indications found at such intersections with a motorized rotating coil probe will cause the tube to be repaired.

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Attachment Response to RAI

5. Section 3.b.3 of Attachment I to GL 95-05, states that the licensee should submit a specific sampling plan if circumferential cracking or primary water stress corrosion cracking indications are detected. Also, as noted in the Generic Letter, ifindications are found at dents with voltages near 5 volts, it may be necessary to expand the rotating pancake coil sampling plan to include dents less than 5.0 volts. Please provide clarification of your plans in this regard.

Southern Nuclear Response: If circumferential cracking or primary water stress corrosion cracking is detected at the tube support plates intersections, a sampling plan will be implemented in accordance with the PWR Steam Generator Tube Examination Guidelines, Revision 4. Ifindications are found at dents with voltages near 5 volts, the flaw will be characterized. If the flaw exceeds the structural requirement ofRegulatory Guide 1.121, the sampling plan 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.

6. Confirm that related to Sections 3.c.2 and 3.c.3, you commit to the probe wear and variability guidance in the Generic Letter or in (1) the Nuclear Energy Institute to NRC letter dated February 23,1996,

Subject:

Eddy Current Probe Replacement Criteria for Use in ODSCC Alternate Repair Criteria (Project No. 689) and (2) the NRC to Nuclear Energy Institute letter dated March 18,1996.

Southern Nuclear Response: Southern Nuclear will implement the guidance in NEI to NRC letter dated February 23,1996 and NRC to NEI letter dated March 18,1996.

Furthermore, Southern Nuclear will verify that both the primary and mix frequencies will meet the i10% variability requirement.

7. Your submittal stated that the inspection guidance discussed in Section 3 of Attachment 1 of GL 95-05 will be ' 'plemented in accordance with the Appendix A guidelines submitted to the NRC by letter dated February 23,1994. However, Appendix A did not prescribe guidelines for data analyst qualifications. Please provide responses to Sections 3.c.4 and 3.c.6 of Attachment 1 to GL 95-05 that addresses this issue.

Southern Nuclear Response: The requirements of 3.c.4 and 3.c.6 will be met. 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 determination of voltage. The use of a small number of analysts is intended to minimize the effect of analyst variability on determination of growth 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 discussed in 3 above prior to returning the r. team generators to service.

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Attachment Response to RAI

8. In your response to Section 3.c.5, which deals with quantitative noise criteria, you stated that " data analysts will use qualitative guidelines in the evaluation of the data." Clarify ;

how Section 3.c.5 is being satisfied and what is meant by " qualitative guidelines."

Southern Nuclear Response: The intent of Section 3.c.5 is satisfied as follows:

Quantitative noise criteria have historically been applied and will be incorporated in the FNP Data Acquisition procedures. This enables noise levels due to electrical noise, tube noise, calibration standard noise, etc., to be addressed at the initial point ofinspection which has minimized 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 FNP Analysis procedures allow the analyst to require re-inspection due to noise on a " qualitative" basis.

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9. Operational Leakage is covered in Section 5.0 and stated that you have implemented the current industry guidelines for leakage monitoring measures. Please provide the reference ,

to the industry guidelines that have been implemented.

Southern Nuclear Response: The requirements of Section 5.0 will be met by the implementation of the operating guidelines for primary-to-secondary leakage contained in the EPRI topical report "PWR Primary-to-Secondary Leak Guidelines," EPRI TR-104788, May 1995.

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