ML20216J489

Submits 90-day Response to GL 97-05, Steam Generator Tube Insp Techniques

ML20216J489
Person / Time
Site:	Wolf Creek
Issue date:	03/16/1998
From:	Muench R WOLF CREEK NUCLEAR OPERATING CORP.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
ET-98-0019, ET-98-19, NUDOCS 9803230460
Download: ML20216J489 (4)

Text

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WG) LF CREEK OPERAT NUCLEAR i

Richard A. Muench j

Vice President Engineering March 16, 1998

(

ET 98-0019 l

U. S. Nuclear Regulatory Commission j ATTN: Document Control Desk 1 Fail Station Pl-137 I hashington, D. C. 20555

Reference:

NRC Generic Letter 97-05, " Steam Generator Tube Inspection Techniques," dated December 17, 1997

Subject:

Docket No. 50-482: Response to NRC Generic Letter 97-05 l l

Gentlemen:

Attached is Wolf Creek Nuclear Operating Corporation's (WCNOC) response to NRC Generic Letter 97-05, " Steam Generator Tube Inspection Techniques." WCNOC has evaluated the reference as it applies to Wolf Creek Generating Station (WCGS),

and has reviewed plant information to address the requests made in the Generic Letter. Based on the reviews completed to develop this response, WCNOC has i reasonable assurance we are in compliance with our current licensing basis. )

l If you have any questions concerning this response, please contact me at (316) {

364-8831, extension 4034, or Mr. Michael J. Angus, at extension 4077. l Very truly yours,

(,0 Richard . Muench Attachment RAM /jad j

cc: W. D. Johnson (NRC), w/a E. W. Merschoff (NRC), w/a '/

J. F. Ringwald (NRC), w/a g/d )

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K. M. Thomas (NRC),w/a *

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9803230460 980316 PDR p

ADOCK 05000482 PDR

. , , , , k P.O Box 411/ Burlington, KS 66839 / Phone- (316) 364-8831 An Equal Opportundy Employer WF/HCNET

STATE OF KANSAS )

) SS COUNTY OF COFFEY )

Richard A. Muench, of lawful age, being first' duly sworn upon oath says that he is Vice President Engineering of Wolf Creek Nuclear Operating Corporation; that he has read the foregoing document and knows the content thereof; that he has executed that same for and on behalf of said Corporation with full .

power and authority to do so; and that the facts therein stated are true and correct to the best of his knowledge, information and belief. .

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By Richardg.Muench Vice President Engineering SUBSCRIBED and sworn to before me this b day of MhQfCh , 1998.

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, Nota Public j JUUE A. DALE '

NotaryPubHe StateolKansas l A W. EnWu J D /,4 DNP ' f b!2O Expiration Date /7 l

• l i

• Attachment to ET 98-0019

. Page 1 of 2 Response to Generic Letter 97-05

" Steam Generator Tube Inspection Techniques" Scope NRC Generic Letter 97-05 (GL), " Steam' Generator Tube Inspection Techniques" was issued to

(1) emphasize to the addressees the importance of performing steam generator tube in service inspections using qualified techniques in accordance with the requirements of Appendix B to 10 CFR Part 50, and (2) require certain information from addressees to determine whether they are in compliance with the current licensing basis for their respective facilities given their steam generator tube in service inspection practices.

Requested Information This response provides information for Wolf Creek Nuclear Operating Corporation (WCNOC) requested by GL 97-05. The information requested includes:

(1) whether it is the licensee's practice to leave steam generator tubes with indications in service based on sizing, (2) if the response to item (1) is affirmative, those licensees should submit a written report that includes, for each type of indication, a description of the associated nondestructive examination method being used and the technical basis for the acceptability of the technique used.

Response to Item (1) :

It .is the practice of WCNOC to leave steam generator tubes with wear indications in service based on sizing.

Written report required by Item (2) :

WCNOC leaves steam generator tube indications in service which can be sized at less than the Technical Specification 4.4.5.4.a value of 40 percent. Wolf Creek Generating Station's steam generators are Westinghouse Model F. The tubes were thermally treated.

The only steam generator tubes left in service, based on qualified sizing techniques,~are those experiencing wear degradation. Any tube that exceeds the Technical Specification repair limit of 40 percent of the nominal tube ,

wall thickness will be plugged. l WCNOC has qualified techniques to size wear degradations. WCNOC's program requires any other degradation mechanisms identified, that can not be appropriately sized with qualified techniques, be plugged on detection. In the future, should other types of degradation be detected, other sizing techniques may be qualified to evaluate the degradation.

The basis for application of sizing techniques is to conduct the examinations under the WCNOC Quality Assurance Program. The examinations follow the requirements of Section XI and V of the ASME Code, 1989 Edition, and Regulatory Guide 1.83. Additional . support for sizing degradation-specific mechanisms is provided by qualification data sets provided in Electric Power Research Institute's (EPRI) " Pressurized Water Reactor (PWR) Steam Generator Examination Guidelines," Revisions 3 through 5, Appendix H, entitled

" Performance Demonstration for Eddy Current Examination." The particular technique number from the EPRI Qualified Techniques Performance Demonstration Database is ETSS #96004 for wear.

For wear at flow distribution baffle plates and anti-vibration bars, sizing is accomplished using the 300/150 absolute mix of the bobbin probe. A

~

Attachment to ET 98-0019 a Paga 2 of 2 calibration curve for amplitude vertical maximum is determined based on the applicable s,tandards replicating the damage mechanism type and quantity. The calibrdtion curve must represent the full range of expected depths.

This sizing qualification is based on the equivalency of 64 sample data points. The samples ranged in depth from 4 percent to 78 percent through wall depth. WCNOC believes that the methodology being used is technically sound and consistent with our current licensing basis.

The nuclear power industry recently voted to adopt an initiative requiring each utility to implement the guidance provided in Nuclear Energy Institute (NEI) 97-06, " Steam Generator Program Guidelines," no later than the first refueling outage starting after January 1, 1999. As specified in NEI 97-06, each utility is to follow the inspection guidelines contained in the latest revision of the EPRI "PWR Steam Generator Examination Guidelines."

EPRI's "PWR Steam Generator Examination Guidelines," Revisions 3 through 5, Appendix H, provides guidance on the qualification of steam generator tubing examination techniques and equipment used to detect and size flaws. Damage mechanisms are divided into the following categories: thinning, pitting, wear, outside diameter intergranular attack / stress-corrosion cracking (IGA / SCC), primary-side SCC, and impingement damage for qualification.

For qualification purposes, test samples are used to evaluate detection and sizing capabilities. While pulled tube samples are preferred, fabricated samples may be used. If fabricated test samples are used, the samples are verified to produce signals similar to those being observed in the field in terms of signal characteristics, signal amplitude, and signal-to-noise ratio.

Samples are examined to determine the actual through wall defect measurements as part of the Appendix H qualification process.

The procedures developed in accordance with Appendix H of the EPRI guidance document specify the essential variables for each procedure. These essential variables are associated with an individual instrument, probe, cable, or particular on-site equipment configuration. Additionally, certain techniques have undergone testing and review to quantify sizing performance. The sizing data set includes the detection data set for the technique, with additional requirements for number and composition of the grading units.