ML20216J021

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

ML20216J021
Person / Time
Site:	Crane
Issue date:	03/16/1998
From:	Langenbach J GENERAL PUBLIC UTILITIES CORP.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
1920-98-20131, GL-97-05, GL-97-5, NUDOCS 9803230279
Download: ML20216J021 (6)

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GPU Nuclear. inc.

Route 441 south Post Office Box 480 NUCLEAR uddletown. PA 17057-0480 Tel 717-944-7821 March 16,1998 1920-98-20131 U. S. Nuclear Regulatory Commission Attn.: Document Control Desk Washington,DC 20555

Subject:

Three Mile Island Nuclear Station, Unit 1 (TMI-1)

Operating License No. DPR-50 Docket No. 50-289 GPU Nuclear Response to NRC Generic Letter (GL) 97-05, " Steam Generator Tube Inspection Techniques" NRC Generic Letter (GL) 97-05 requested that licensees submit a written response that includes the following information: (1) Whether it is their practice to leave steam generator tubes with indications in senice based upon sizing, and (2) If the response to (1) is aflirmative, those licensees should submit a written report that includes, for each type ofindication a desenption of the associated nondestructive examination method being used and the technical basis for acceptability of the technique used.

INTRODUCTION:

The nuc! car power industry recently voted to adopt an initiative requiring each utility to meet the intent of the guidance provided by the Nuclear Energy Institute (NEI) in NEI 97-06, " Steam Generator Program Guidelines," no later than the first refueling outage starting afler January 1,1999. Adherence to these NEI guidelines will require the technical basis for the inspections to be upgraded to the current revision (currently Revision 5) of the Electric Power Research Institute (EPRI) Pressurized Water Reactor (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 qualification of steam generator tubing examination techniques and equipment used to detect and size flaws. Damage mechanisms are divided into the following categories for qualification: thinning, pitting, wear, outside diameter Inter-Granular Attack / Stress Corrosion Cracking (IGA / SCC), primary-side SCC, and impingement damage.

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1920-98-20131 Page 2 of 5 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 veri 6ed to produce signals similar to those being observed in the field in terms of signal characteristics, signal amplitude, and signal-to-noise ratio. Sample actual Through Wall (TW) 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 indisidual instrument, probe, cable, or panicular 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 ofthe grading units.

GENERALINFORMATION:

The TMI-l steam generators are Babcock and Wilcox Model 177F Once Through Steam Generators (OTSGs). The OTSG tubing material is 0.625" nominal diameter,0.037" nominal wall thickness, inconel600.

The TMI-l Quahty Assurance Program implements the guidance found in the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Sections V and XI. The plant adheres to the 1986 Code, with no Addenda. Steam generator tube indications left in senice are those which are sized at less than the Technical Specification value of40 % TW (or repair limit with a consideration for growth).

RESPONSE TO NRC QUESTION 1:

. At TMI-1, sizing techniques are used during steam generator inspections to leave flaws with the following degradation modes in senice:

a) Mechanical wear at tube suppon plates, b) Inside Diameter (ID) IGA, in accordance with TMl-1 License Amendment No. 206, dated October 16,1997, and c) Kinetic expansion region flaws dispositioned as described in a January 12,1998 GPU Nuclear

- submittal to the NRC, " Cycle (12R) Refueling Outage Once Through Steam Generator (OTSG)

Tube Inspection Repon with ASME NIS Data Repons for Insenice Inspection (ISI)"

- RESPONSE TO NRC QUESTION 2:

The basis for application of these sizing techniques is the conduct of the examinations in accordance

- with the TMI-l Quality Assurance Program following the requirements of Section XI and V of the

1920-98-20131

. Page 4 of 5 equivalent of the B&W Owners Group Mother Standard, four 20% TW ASME drilled holes. This setting was stored to all other channels.

This bobbin coil sizing qualification was established using 21 TMI-l steam generator pulled tube and two laboratory grown flaw samples ranging in depth from 16% to 100% TW. Three of the -

flaws were volumetric IGA while 18 of the flaws were ID SCC. The technique was further

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validated during the September 1997 outage using in-situ pressure testing.

All volumetric IGA indications detected during the bobbin examinations were examined with a motorized rotating pancake (MRPC) probe. Axial and circumferential length measurements were assigned using a 0.080" high frequency pancake coil 600 kHz clip plot measurement. The eddy current test (ECT) measured dimension ofIGA has been determined to be consistently conservative when compared to 23 machined OTSG tube samples. These samples ranged from 0.020" to 0.160"in diameter and from 20% to 80% TW. In-situ pressure testing further validated the examination findings.

One tube with multiple ID IGA flaws was removed from rhe TMI-l "A" steam generator in October 1997 for laboratory analysis. Preliminary leak and burst test results (with and without 1400 lbs applied axial load at normal operating pressure, main steam line break pressure, and draft Regulatory Guide 1.121 conditions) have recently been obtained on four sections of this tube.

These sections contained several eddy current indications ofsuspected ID IGA. No detectable

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leakage was noted during any of the leak tests and the lovcest burst pressure was measured at 10,800 psig. The findings from laboratory analy6 will be reported to the NRC by June 1,1998.

c.

Kinetic Expansion Region Flaws The TMl-1 OTSG upper tubesheet uppermost tube region was damaged in 1981 by a sulfur intrusion. The tubes were repaired by kinetically expanding the uppermost 17" or 22" of tube inside the upper tubesheet. The expansion process and improvements in ECT examination technology led to identification of residual tube degradation from the 1981 sulfur intrusion during our most recent 1997 outage. These indications are not from an active damage mechanism. The indications were dispositioned based on leakage calculation and stmetural integrity. Indications measured as > 67% TW by the mid-frequency plus point probe were considered potential leakage contributors and were considered to be 100% TW for the leakage calculations. The structural integrity evaluatiormasiders all measured flaw circumfereratial and/or axial extents to be 100%

TW. (For volumetric indications, the structural iutegrity evaluation considers a volumetric indication as consisting of a 100% TW axial crack of tength corresponding to the axial extent of j

the volumetric indication, as well as a 100% TW circumferential crack corresponding to the

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circumferential extent of the volumetric indication.) Further details on the dispositioning of kinetic expansion indications may be found in the GPU Nuclear submittd dated January 12,1998, " Cycle 12 Refueling (12R) Outage Once Through Steam Generator (OTSG) Tube Inspection Report with ASME NIS Data Reports for Insenice Inspections (ISI)," Enclosure 1, Attachment 1,Section II.C.2.

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• 1920-98-20131 Page 3 of 5 a) Mechanical Wear at Tube Support Plates

Mechanical wear indications are identified by the use of the bobbin coil examination at TMI-1. All wear indications are then examined with a rotating coil ' probe to provide more accurate sizing -

information.-

For wear at broached Tube Support Plates (TSPs), the 300/100 kHz TSP mix signals from the mid-range 0.115" pancake coil are used to size the depth of the wear flaw. A calibration curve is established using the 0%,20% and 50% land contact points of the 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.

The sizing procedure is based on the analysis of 26 sample data points. The samples ranged in depth from 22% to 100%. This database has been reviewed to ensure that application of the sizing.

procedure is consistent with the steam generator conditions at TMI-1. Therefore, the sizing procedure for wear is site-qualified for TMI-l in accordance with paragraph 6.2.4 of the PWR

' Steam Generator Exarvination Guidelines, Revision 5.

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.The current OTSG wear sizing qualification 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 indicate that mix channel sizing with either a 0.115" pancake coil or a plus point coil performs with sizing errors i

equivalent to the original raw frequency channel qualification. This sizing technique has been submitted for industry peer review during 1998.

b) IGA in Unexpanded Tubing and Expansion Transitions 1

On October 16,1997, the NRC issued TM1-1 License Amendment No. 206, which authorizes an I

alternate tube repair criteria for ID IGA for one operating cycle, Cycle 12. The amendment requires that "lD IGA indications" shall be repaired or removed from service if they exceed an axial extent of 0.25 inches, or a circumferential extent of 0.52 inches, or a TW degradation dimension of 240% if assigned." The amendment requires that any ID indications be confirmed as both ID initiated and volumetric in nature.

The TW dimension, if assigned, was assigned using a 0.540" high frequency bobbin coil probe. ID 1GA bobbin coil indications were assigned a measured TW dimension if the signal was > 5? and

< 30* and was > 3:1 signal-to-noise ratio or 21 volt. ID GA signals < 3:1 signal-to-noise ratio

' and < 1 volt were assigned a three letter code, "BVC." A TW dimension was assigned using the 400 kHz differential channel or, for signals in the tubesheet cresice that were not clear with the 400 kHz channel, the 400/200 kHz mix channel was used. Both channels employed a phase angle measurement technique established from the 20%,60%, and 100% TW ASME drilled holes.

Voltage normalization was established by setting the 400 kHz differential channel to a 10 volt

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. Page 5 of 5 For Primary Water Stress Corrosion Cracking (PWSCC) circumferential cracks in the kinetic expansion region, the 300 kHz signal from the mid-range plus point is used to size the depth and

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length. A calibration curve is established using the 40%, 60%, and 100% TW ID notches of the calibration standard.- Voltage normalization is performed from the 100% TW hole at 10 volts.

This sizing procedure is based on the analysis of 16 sample data points. Five of the' data points are from pulled tubes. The samples ranged in depth from 32% to 100% TW.

For volumetric ID IGA in the kinetic expansion region, the 300 kHz signal from the mid-range plus -

point is used to size the depth of the degradation. Axial and circumferential length measurements were assigned using the 0.080" high frequency pancake coil 600 kHz clip plot measurement. The ECT measured axial and circumferential extents ofIGA have been determined to be consistently conservative when compared to 23 machined OTSG tube samples. These samples ranged from 0.020" to 0.160" in diameter and from 20% to 80% TW.

In-situ pressure tests performed in the freespan of the OTSG tubing in October 1997 further validate the conservatism of the accident leakage volumes calculated for flaws in the kinetic expansion region.

Sincerely, James W. Lang ach Vice President and Director, TMI Attachment MRK-

- cc:

Administrator, Region I 1

TMI Senior Resident Inspector TMl Senior NRC Project Manager I

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• 1920-98-20131 Attachment METROPOLITAN EDISON COMPANY JERSEY CENTRAL POWER AND LIGHT COMPANY PENNSYLVANIA POWER AND LIGHT COMPANY GPU NUCLEAR INCORPORATED Three Mile Island Nuclear Station, Unit 1 (TMI-1)

Operating License No. DPR-50 Docket No. 50-289 1, James W. Langenbach being duly sworn, state that I am a Vice President of GPU Nuclear, Inc. and that I am duly authorized to execute and file this response on behalfofGPU Nuclear. To the best of my knowledge and belief, the statements contained in this document are tme and correct. To the extent that these statements are not based on my personal knowledge, they are based upon information by other GPU Nuclear employees and/or consultants. Such information has been reviewed in accordance with company practices and I believe it to be reliable.

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James W. Langen ach Vice President and Director, TMI Signed and sworn before me this

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