ML20216J430
| ML20216J430 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 03/16/1998 |
| From: | Jeffery Wood CENTERIOR ENERGY |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| 2520, GL-97-05, GL-97-5, NUDOCS 9803230439 | |
| Download: ML20216J430 (4) | |
Text
1 P=.- -- n.r=
Davis-Besse Nuclear Power Statron g gy MC 5501 North State Route 2 m
Oak Harbor. Ohio 43449-9760 n
John K. Wbod 419-249-2300 Vice President Nuclear Fax: 419-321-8337 Docket Number 50-346 License Number NPF-3 Serial Number 2520 March 16, 1998 United States Nuclear Regulatory Commission Document Control Desk Washington, D. C. 20555-0001
)
Subject:
Response to NRC Generic Letter 97-05, Steam Generator Tube Inspection Techniques Ladies and Gentlemen:
On December 17,1997, the Nuclear Regulatory Commission (NRC) issued Generic Letter (GL) 97-05. That letter requested licensees, such as those for the Davis-Besse Nuclear Power Station (DBNPS), to respond within 90 days and to address the following issues:
(1) whether it is the practice of the licensee to leave steam generator tubes with indications in service based on sizing, and (2) if the response to item (1) is affirmative, the licensee 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.
The DBNPS response to Generic Letter 97-05 follows:
INDUSTRY INITIATIVES:
The nuclear power industry recently adopted an initiative requiring each licensee to meet the intent of the guidance provided in the Nuclear Energy Institute (NEI) document NEI 97-06, Steam Generator Program Guidelines, no later than the first refueling outage starting after January 1,1999. For the DBNPS, the 12th Refueling in the spring of 2000 will be the outage that meets this criterion. Adherence to these NEI guidelines will require the technical basis for f4 the inspections to be upgraded to the then-current revision (currently Revision 5) of the EPRI t
4 PWR Steam Generator Examination Guidelines.
Appendix H, Performance Demonstration for Eddy Current Examination, of the EPRI PWR Steam Generator Examination Guidelines, Revisions 3 through 5, provides guidance on the
, y ;) l; g
M 9803230439 980316 PDR ADOCK 05000346 P
7.
Docket Number 50-346 License Number NPF-3 Serial Number 2520 1
- Page;2 qualification of steam generator tubing examination techniques and equipment used to detect and i
size flaws.
Damage mechanisms are divided into the following categories for qualification: thinning, pitting, wear, outside diameter intergranular attack / stress corrosion cracking (IGA / SCC),
primary-side SCC, and impingement damage.
Test samples are used to evaluate detection and sizing capabilities for qualification purposes.
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 i
the field in terms of signal characteristics, signal amplitude, and signal-to-noise ratio. Sample actual through-wall defect measurements must be verified as part of the Appendix H qualification process.
The eddy current detection and sizing procedures qualified in accordance with Appendix H specify the essential variables for each technique. These essential variables are associated with an individual instrument, probe, cable, or particular on-site equipment configurations.
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.
GENERAL INFORMATION:
The DBNPS steam generators are Babcock and Wilcox Model 177F Once Through Steam Generators, commonly referred to as OTSGs. The tubing material in the OTSGs is 0.625" nominal diameter,0.037" nominal wall thickness Inconel 600. Steam generator tube indications allowed by Technical Specifications to be left in service are those which are sized at less than the Technical Specification value of 40 % throughwall (or a lesser through-wall limit with a consideration for growth).
RESPONSE TO NRC QUESTION 1:
At the DBNPS, sizing techniques are used during steam generator inspections to determine which flaws produced by mechanical wear at the tube support plates (TSPs) may be left in service. All other damage mechanisms are plugged or repaired in accordance with the requirements of the DBNPS Operating License Technical Specifications.
RESPONSE TO NRC QUESTION 2:
For wear identified at broached TSPs in the DBNPS OTSGs,300/100 kHz TSP mix signals from a mid-range 0.080" pancake coil are used to size the depth of the wear flaw. An amplitude-based calibration curve is established using the 0%,20% and 50% manufactured wear marks of the
Docket NumbeE 50-346 License Number NPF-3 Serial Number 2520
,Page3 calibration standard. The size of the flaw is measured using the largest amplitude signal from the 300/100 kHz mix channel. Filters are not used during this depth sizing.
The sizing procedure was based on the analysis of 26 sample data points. The samples ranged in depth from 22% to 100%. The database of sample data points has been reviewed to ensure that i
application of the sizing procedure is consistent with the steam generator conditions at the
{
DBNPS. Therefore, the sizing procedure for wear is site-qualified for the DBNPS in accordance with paragraph 6.2.4 of the EPRI PWR Steam Generator Examination Guidelines, Revision 5.
The current OTSG wear sizing qualification, using the mid-range 0.080" pancake coil, is a site-specific qualification due to the use of the mix channel instead of the raw data frequency for sizing. Examination data acquired using the mix channel evaluation method on the same samples used for the original qualification ndicate that the mix channel sizing with either a 0.115" i
pancake coil or a Plus Point ccil also performs with sizing errors equivalent to the original raw frequency channel and could be used. This sizing technique has been submitted to the EPRI Peer review group for industry peer review during 1998.
The steam generator inspection sizing techniques are performed under the DBNPS Quality Assurance Program. These techniques are based on the requirements of Sections V and XI of the ASME Code 1986 Edition, no Addenda, and Regulatory Guide 1.83, Inservice Inspection of Pressurized Water Reactor Steam Generator Tubes. Additional support for sizing specific 1
degradation mechanisms is provided by the EPRI Appendix H qualification de.ta sets.
Should you have any questions or require additional information, please contact Mr. James L.
Freels, Manager-Regulatory Affairs, at (419) 321-8466.
Very truly yours, i
FWK/laj cc:
A. B. Beach, Regional Administrator, NRC Region 111 S. J. Campbell, NRC Region III, DB-1 Senior Resident Inspector A. G. IIansen, DB-1 NRC/NRR Project Manager Utility Radiological Safety Board
e Docket Number 50-346 License Number NPF-3 Serial Number 2520
. Page,4
RESPONSE
TO NRC GENERIC LETTER 97-05 FOR THE DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 This letter is submitted pursuant to 10 CFR 50.54(f) and contains information pursuant to NRC Generic Letter 97-05, Steam Generator Tube Inspection Techniques for the L) avis-Besse Nuclear Power Station, Unit Number 1.
By:
John K/ Wood, Vice President - Nuct' ear Sworn to and subscribed before me this 16th day of March, 1998.
At.-
4%
4 Notary Public, State offhio Nora Lynn Flood My commission expires September 4, 2002 s