ULNRC-03772, Submits 90-day Response to GL 97-05, Steam Generator Tube Inspection Techniques
| ML20216H762 | |
| Person / Time | |
|---|---|
| Site: | Callaway  |
| Issue date: | 03/17/1998 |
| From: | Passwater A UNION ELECTRIC CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| GL-97-05, GL-97-5, ULNRC-03772, ULNRC-3772, NUDOCS 9803230168 | |
| Download: ML20216H762 (7) | |
Text
i Union EI:ctric One Ameren Ylaza 1901 Chouteau Avenue P0 Box 66149 St. Louis, MO 63166-6149 314.6211222 March 17,1998 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Mail Station: PI-137 Washington, D.C. 20555-0001 Gentlemen:
ULNRC-03772 hYL.
77/ggg DOCKET NUMBER 50-483 CALLAWAY PLANT
- E uN10N EtterRic cOuPANv RESPONSE TO GENERIC LETTER 97-05
Reference:
Generic Letter 97-05, " STEAM GENERATOR TUBE INSPECTION TECHNIQUES," dated December 17,1997 The above reference requested information on whether or not it is our practice to leave steam generator tubes with indications in service based on sizing and if so to submit a written repon that includes, for each type ofindication, a description of the associated nondestructive examination method being used and the technical basis for the acceptability of the technique used. provides Union Electric's 90-day response to Generic Letter 97-05. The information provided in Attachment I describes programs in place as of the date of this letter. It is not intended to preclude subsequent changes following normal administrative procedures or to require NRC notification or consent for such changes other than those currently required. This letter contains no new commitments.
Should you have any questions or need additional information concerning this matter, please contact us.
Sincerely, t
(DDI i
Kf a-cumf Alan C. Passwater Manager, Licensing and Fuels JMC/pir f
Attachment i
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9803230168 980317 PDR ADOCK 05000483 lillm!IlllI.n.l.la.)lly
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STATE OF MISSOURI f
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l Alan C.
Passwater, of lawful age, being first duly sworn upon oath says that he is Manager, Licensing and Fuels (Nuclear) for Union Electric Company; that he has read the foregoing document and knows the content thereof; that he has executed the same for and on behalf of said company with full power and authority to do so; and that the facts therein stated are true and correct to the best of his j
knowledge, information and belief.
By Alan C. Passwater Manager, Licensing and Fuels l
Nuclear SUBSCRIBED and sworn to before me this
/7dI day I
of
- ht<'M 1998.
8 lN suur. M amme WEAAAMM M M E M
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cc:
M..H.
Fletcher professional Nuclear Consulting, Inc.
19041'Raines Drive Derwood, MD 20855-2432 Regional Administrator U.S. Nuclear Regulatory Commission Region IV
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611 Ryan Plaza Drive Suite 400 Arlington, TX 76011-8064 Senior Resident Inspector Callaway Resident Office U.S. Nuclear Regulatory Commission 8201 NRC Road Steedman, MO.65077_
Barry C. Westreich (2)
Office _of Nuclear Reactor Regulation U.S. Nuclear Regulatory. Commission 1 White Flint, North, Mail Stop 13E16 11555 Rockville Pike Rockville, MD 20852-2738 Manager, Electric Department _.
H Missouri Public Service-Commission P.O. Box 360 Jefferson City, MO 65102 Ron Kucera Department of Natural' Resources P.O. Box-176 Jefferson City, MO 65102 i
Denny Buschbaum
'TU Electric P.O. Box 1002 Glen Rose, TX 76043 s.
Pat Nugent Pacific Gas & Electric Regulatory Services P.O. Box 56 Avila Beach, CA 93424
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bec: J. Brandt/A160.761
/QA Record (CA-758)
E210.01 J. V. Laux-4 G. L. Randolph' R. J.
Irwin P. M.
Barrett J. D. Blosser I
A. C.~Passwater D.
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Shafer W.,E.
Kahl S. Wideman (WCNOC)
A. J. DiPerna, (Bechtel)
H. D. Bono g
NSRB.(Patty Reynolds) 1 J. M. Chapman j
T.-E. Hermann A160.412 (97-05)
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ATI'ACHMENT 1 PAGE l OF 3 ULNRC-3772 l
RESPONSE TO GENERIC LETTER 97-05 i
" STEAM GENERATOR TUBE I'NSPECTION TECHNIOUES" INTRODUCTION Eddy cunent inspection techniques have traditionally been used to assess the condition of steam generator tubes, The ability to depth size certain specific modes of tube degradation has allowed utilities to avoid unnecessary repair of tubes which remain.
structurally sound and capable of performing their operational and safety functions. As inspection techniques and industry understanding of damage mechanisms have evolved, the limits of sizing certain defects have become apparent. This has led to the issuance of NRC Generic Letter 97-05, Steam Generator Tube Inspection Techniques. The following information addresses the requested actions of Generic Letter 97-05 as they pertain to the Westinghouse designed and manufactured steam generators installed at Callaway Plant.
Callaway Plant is equipped with Westinghouse Model F steam generators, the first of this type manufactured by the supplier. The units are tubed with 11/16 inch O.D. Inconel 600 material. The 10 inner (short radius) tube rows are thermally treated; all remaining rows are mill annealed. Tubing is installed with full depth hydraulic expansion in the tube j
sheet and all tube suppott plates are stainless steel with broached quatrefoil tube passages having circular lands.
REOUIRED INFORMATION Within 90 days of the date of this generic letter, addressees are required to submit a written response that includes the following information.
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(1) whether it is their practice to leave steam gemrator tubes with indications in service based on sizing, (2) if the response to item (1) is affirmative, those licensees should submit a
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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.
REOUIRED RESPONSE It is the current practice at Callaway Plant to leave steam generator tubes with j
wear indications in service based on sizing. Steam generator tube wear indications that are sized at less than the Technical Specification value of 40% through-wall are left m i
service. Under the current program, recently revised, the only type of indication for which a percent through-wall depth will be assigned is wear at the anti-vibration bar (AVB) intersections. Prior to this program revision, benign type free span signals with a
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ATTACHMENT 1 PAGE 2 OF 3 ULNRC-3772 differential' response that could be historically tracked back to startup were frequently assigned a depth size and left in service. These type indications will no longer be assigned a size, but will require historical verification and, if appropriate, Rotating Pancake Coil testing.
Technical Basis for Leavine in Service The nuclear power industry recently voted to adopt an initiative requiring each utility to implement the guidance provided in 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 required to follow the inspection guidelines contained in the latest revision of the EPRI PWR Steam Generator Examination Guidelines.
Appendix H, " Performance Demonstration for Eddy Current Examination," of the PWR Steam Generator Examination Guidelines, Revisions 3 through 5, 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 qualification categories: thinning, pitting, wear, outside diameter intergranular attack / stress-corrosion cracking (IGA / SCC), primary-side SCC, and impingement damage.
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 specify the essential variables for each procedure. These essential variables are associated with an individual instrument, probe, cable, or particular on-site equipment configurations. Additionally, J
certain techniques have undergone testing and review to quantify sizing performance.
I The sizing data set includes the detection data set for the technique with additional requirements for number and composition of the grading units.
At Callaway Plant, sizing techniques are used during steam generator inspections to leave AVB wear flaws in service. Application of these sizing technique examinations are conducted under the Callaway Plant Quality Assurance Program following the requirements of Sections XI and V of the ASME Code,1989 Edition and Regulatory Guide 1.83. Additional support for sizing degradation-specific mechanisms is provided by the EPRI Appendix H qualification data sets.
For wear at the anti-vibration bars, sizing is accomplished using an appropriate differential or absolute frequency mix of the bobbin coil in accordance with the qualified
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.I ATTACHMENT 1 PAGE 3 OF 3 ULNRC-3772 EPRI Appehdix H technige (ETSS #96004). A calibration curve for amplitude vertical maximum is determined based on the applicable standards replicating the damage mechanism type and quantity. The calibration curve must represent the full range of expected depths. This sizing qualification is based on 64 sample data points. The samples ranged in depth from 4% to 78% through-wall depth.
No other types ofindications are currently allowed to remain in service pmely based on depth sizing. However, Callaway Plant would consider the use of sizing for other type damage mechanisms based on Appendix H qualified techniques and accepted industry practice, should the need arise.
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