ML20248L421
| ML20248L421 | |
| Person / Time | |
|---|---|
| Site: | Waterford  |
| Issue date: | 03/17/1998 |
| From: | Ewing E ENTERGY OPERATIONS, INC. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| GL-97-05, GL-97-5, W3F1-98-0043, W3F1-98-43, NUDOCS 9803200212 | |
| Download: ML20248L421 (10) | |
Text
.
y Ent6. y perations, Inc.
Killona. LA 70066 Tel 504 739 6242 Earl C. Ewing, Ill ac ear Safety & Regulatory AHarrs W3F1-98-0043 A4.05 PR March 17,1998 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555
Subject:
Waterford 3 SES Docket No. 50-382 License No. NPF-38 90-Day Response to NRC Generic Letter 97-05, " Steam Generator Tube inspection Techniques" Gentlemen:
On December 17,1997, the NRC issued Generic Letter (GL) 97-05. The generic letter requested information which would allow a determination on whether licensees are in compliance with the current licensing basis for their respective facilities given their steam generator tube inservice inspection practices. By this letter, EOl is providing the requested 90-day response for the Waterford 3 Steam Electric Facility.
Specifically, the GL requested a written response to the following questions:
(1) whether it is their 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 which includes, for each type indication, a description of the associated nondestructive examination method being used and the technical basis for the acceptability of the technique used.
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9803200212 980317 PDR ADOCK 05000302 P
- r; 90-Day' Response to NRC Generic Letter 97-05, " Steam Generator Tube
- Inspection Techniques" W3F1-98-0043 '
Page 2
- March 17,1998.
It is Waterford 3's current practice to leave steam generator tubes with indications in service based on sizing. Therefore, in accordance with the Generic Letter request, a written report including the qualification techniques employed to size indications is provided in the attached report. The information provided justifies that Waterford 3 complies with 10 CFR Part 50 Appendix B Criterion IX, " Control of Special Processes" and plant Technical Specifications as well as provides a reasonable level of assurance that steam generator tube integrity margins have been maintained for those indications that remain in service. This information is being submitted under oath and affirmation in accordance with 10CFR50.54(f). Should you have any questions regarding this matter, please contact Mr. T.J. Gaudet at (504) 739-6666 or me at (504) 739-6242.
Very truly yours, s-l E.C. Iwing Director Nuclear Safety & Regulatory Affairs ECE/PRSfjmm L
Enclosures:
Affidavit Response to Generic Letter 97-05 cc:
E.W. Merschoff, NRC Region IV C.P. Patel, NRC-NRR J. Smith
' N.S. R.eynolds NRC Resident inspectors Office
_____.________________m--_
l:
1, 1
UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION l
In the matter of
)
)
Entergy Operations, Incorporated
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Docket No. 50-382 -
Waterford 3 Steam Electric Station
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AFFIDAVIT Early Cunningham Ewing, being duly sworn, hereby deposes and says that he is Director, Nuclear Safety and Regulatory Affairs-Waterford 3 of Entergy Operations, Incorporated; that he is duly authorized to sign and file with the Nuclear Regulatory Commission the attached 90 Day Response to NRC Generic Letter 97-05; that he is familiar with the content thereof; and that the matters set forth therein are true and correct to the best of his knowledge, information and belief.
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/Earlyfnning win N
Diredor, Nuclear Safe Regulatory Affairs -
Waterford 3 STATE OF LOUISIANA
)
) ss PARISH OF ST. CHARLES
)
Subscribed and sworn to before me, a Notary Public in and for the Parish and State above named this e day of me ml
.1998.
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Notary Public My Commission expires a AM i
( to W3F1-98-0043 Page 1 of 7 Waterford 3 Response to Generic Letter 97-05
" Steam Generator Tube inspection Techniques" l
Introduction This document is the Waterford 3 Safety Assessment justifying compliance with 10 CFR Part 50 Appendix B Criterion IX, " Control of Special Processes" and
. plant Technical Specifications with a reasonable level of assurance that steam generator (S/G) tube integrity margins have been maintained for those indications that remain in service.
The December 17,1997 release of NRC Generic Letter 97-05, " Steam Generator Tube inspection Techniques," required a written response that
-includes the following:
1 (1)
Whether it is the licensees 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 I
technique used.
Response to item (1):
1 Presently, Waterford 3 leaves steam generator tubes with indications in service based on sizing. The Waterford 3 S/G eddy current testing program monitors j
and subsequently re-examines degraded tubes (220% - 539% Through Wall) that have been detected / sized and remain in service. This is accomplished in accordance with Waterford 3's S/G Eddy Current Data Analysis Guidelines and EPRI 's Pressurized Water Reactor (PWR) Steam Generator Examination Guidelines, Appendix H.
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h to W3F1-98-0043 Page 2 of 7
)-
Response to item (2):
~
The following provides the requested written report which includes Background l-Information, Types of Indications, a Description of the Nondestructive l:
Examination Method, and the Technical Basis for Technique Acceptability.
1 Background Information L
L
. As required by Waterford 3 Technical Specification 3/4.4.4.4, Steam Generator tubes with through wall indications 240% are removed from service by plugging.
The Waterford 3 S!G eddy current testing program monitors and re-examines i
. degraded tubes (220% - s39% Through Wall) that have been detected / sized and 1
f l
remain in service.
The nuclear power industry recently voted to adopt an initiative requiring each -
utility to meet the intent of the guidance provided in NEl 97-06,- Steam Generator -
Program Guidelines, no later than the first refueling outage starting after January 1,1999. As required by NEl 97-06, each utility is required to follow the l
inspection guidelines contained in the latest revision of the EPRI PWR Steam Generator Examination Guidelines.
l Waterford 3's Refuel #8 (Spring 1997) S/G eddy current examination program -
. I was administered in compliance with plant Technical Specifications and EPRI i
PWR Steam Generator Examination Guidelines, Revision.4. Specifically, Waterford 3 followed EPRl's Appendix H, Performance Demonstration for Eddy I
- Current Examination which 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 stress corrosion cracking (ODSCC), intergranular
. attack (IGA), primary water stress corrosion cracking (PWSCC), and impingement damage for qualification.
As per Appendix H requirements, the steam generator tubing examination techniques and equipment used to detect and size flaws shall be qualified by performance demonstration. For qualification purposes, data set test samples are used to evaluate detection and sizing capabilities. While pulled tube 1
samples are preferred, fabricated samples may be utilized. Test samples l
fabricated using mechanical (electric discharge machining (EDM)) or chemical methods may be used. The flaws should produce signals similar to those being observed in the field in terms of signal characteristics, signal amplitude, and j
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.. to W3F1-98-0043 Page 3 of 7 e
signal-to-noise ratio.'. Samples are metallurgically analyzed and examined to determine the actual through wall defect depth measurements as part of the Appendix H qualification process.
The acquisition and analysis procedures are developed in accordance with Appendix H requirements for each technique which correlates to essential variables. The following list of essential variables define the limits of each H
technique specific to detection, sizing, and data analysis applicable to the I
I damage mechanism:2 l
Equipment Variables l-Instrument Manufacturer Probe Size and Type Cable Lengths Extension Cables l
l Technique Variables Examination Frequencies Voltage and Gain Settings i
Coil Excitation (Absolute or Differential)
Minimum Data to be Recorded (Channels)
Method of Data Recording (Digital)
Digitizing Rates (Samples per Inch)
Scan Pattern (Direction, Push / Pull)
Analysis Variables Method of Calibration (Calibration Curves and Mixes) 4 Data Review Requirements Reporting Requirements
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Instrument / Computerized System Algorithms Waterford 3 has qualified techniques to size wear degradations.' Other types of degradation have been detected but not sized at Waterford 3. Those S/G tubes l
having degradation mechanisms detected for which no qualified Appendix H sizing technique is available are plugged on detection.
u i
The Waterford 3 S/G eddy current testing program is performed in accordance i
with Regulatory Guide 1.83, inservice inspection of Pressurized Water Reactor Steam Generator Tubes, Revision 1, July 1975. The program is implemented
' EPRI-TR-107569-V1, PWR Steam Generator Tube Examination Guidelines, Rev. 5 2
EPRI-TR-107569-V1, PWR Steam Generator Tube Examination Guidelines, Rev. 5
L, c.
. to W3F1-98-0043 Page 4 of 7 through Waterford 3 Procedure NOECP-252, Steam Generator Eddy Current inservice Testing. Regulatory Guide 1.83 prescribes the use of ASME Code Section XI as the examination methodology. ASME Section XI references ASME Section V, Article 8, Appendix 1,1-865 which provides the correlation of eddy 1
current testing signal by phase angle to depth of discontinuity. A relationship of reference comparator depths versus signal phase angle for the examination being performed develops a phase angle to flaw depth percentage. S/G eddy current testing flaw detection and sizing are performed in the field utilizing ASME calibration' standards which are fabricated from tubing of the same nominal size and material as tubing in the S/G.
Steam generator eddy current testing is conducted in accordance with Waterford 3's Quality Assurance program specific to review and acceptance of analysts' qualifications / certifications, procedures, and calibration of equipment.
Types of Indications Wear is currently the only steam generator tube degradation mechanism detected and sized which remains in service at Waterford 3.
Description of Nondestructive Examination Method At Waterford 3, the wear sizing technique was utilized during steam generator eddy current inspections to leave tubes with wear indications in the degraded range (220% - s39% Through Wall) in service. The wear indications are located at upper bundle diagonal and vertical straps.d Bobbin coil Acquisition Technique Specification Sheet (ACTS) for tube flaw detection and Analyais Technique Specification Sheet (ANTS) for sizing of wear at diagonal and vertical straps has been accomplished utilizing the following techniques:5 4
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ASME B&PV Section V, Nondestructive Examination, Article 8, Appendix l f
' Rockridge Technologies /FTI, Final Report, Volume 1, April 1997 8 Waterford 3 Steam Generator Eddy Current Data Analysis Guidelines, Rev. 4 1
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W3F1-98-0043 Page 5 of 7 DETECTION:
te Mix #1 (P1) 400/100kHz Differential for Strip Display during the
. Initial Screening of the Data. (Utilizing ASME Calibration Standard)
I l
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SIZING:
l Mix'#2 (P2) 400/100kHz Differential for Assigning Depth Estimates e
to Indications of Mechanical Wear.' (Utilizing Wear Scar Standard)
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.
l
. Technical Basis for Technique Acceptability l
The Waterford 3 Appendix H wear sizing qualification was based on utilizing l
0.580" and 0.560" diameter probes which have Zetec, Inc. catalog designations of MULC and SFRM, respectively. The qualification technique utilized seventeen
}
(17) wear scar standards as the data set. The range of wear indications for this qualification were 10% - 65% through wall. The wear qualification for differential.
and absolute methods utilizing a 0.580" MULC probe resulted in calculated respective root mean square errors of 4.24% and 2.56%. Additionally, the wear L
qualification for differential and absolute methods utilizing the 0.560" SFRM.
probe resulted in calculated respective root mean square errors of 3.67% and
- 4.6%. These results are well within the 25% root mean square error limit-prescribed by EPRI's Appendix H qualification.8 Therefore, the Waterford 3 sizing technique for wear is qualified in accordance with EPRI's PWR Steam Generator Examination Guidelines.
The Waterford 3 Refuel #8100% full length eddy current testing program of both S/Gs used the following three bobbin coil probes and with respective fill factors 7
' (FF):
WATERFORD 3 REFUEL #8 BOBBIN COIL PROBES (1) 0.600" MULC, (FF) (0.600)'/(0.654")' = 83.7%, 92.8% of tubes.
'(2) 0.580" MULC, (FF) (0.580)'/(0.654")2 = 78.6%,4.3% of tubes.
(3) 0.560" SFRM, (FF) (0.560)'/(0.654")' = 73.3%, 2.8% of tubes.
The use of the 0.600" diameter MULC bobbin coil increased the fill factor 5.1%
above the differential method utilized to size wear indications. Waterford 3's l
- Rockridge Technologies /FTI, Appendix H Wear Qualification CE Steam Generators
' Rockridge Technologies /FTI, Final Report Volurne 1, April 1997 l
, to W3F1-98-0043 Page 6 of 7 utilization of the higher fill factor (0.600" dia.) probe further improved the signal amplitude response for wear indications.
Conclusion The Waterford 3 S/G eddy current testing program methods utilized during the inspection and analysis processes ensures that degraded tubes that remain in service as a result of wear have been detected and sized in accordance with EPRI's PWR S/G Tube Examination Guidelines, Appendix H.
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- to W3F1-98-0043 Page 7 of 7
- References Waterford 3 Plant Technical Specification 3/4.4.4 Steam Generators ASME B&PV Code Section XI, Rules for inspection of Nuclear Power Plant Components,1992 Edition ASME B&PV Code Section V, Nondestructive Examination, Article 8, Appendix I NOECP-252, Steam Generator Eddy Current inservice Testing EPRI TR-107589-V1, PWR Steam Generator Tube Examination Guidelines, Revision 4 EPRI TR-107569-V1, PWR Steam Generator Tube Examination Guidelines, Revision 5 Rockridge Technologies /Framatome Technologies, Inc., Waterford 3 Final Report, Volume 1, April 1997 Rockridge Technologies /FTI, Appendix H Wear Qualification CE Steam L
- Generators, December 1996 Waterford 3 Steam Generator Eddy Current Data Analysis Guidelines, Rev. 4 l
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