ML20217J220

Provides Corrected Copy of Response to GL 97-05, SG Tube Insp Techniques. Response Submitted on 980316 Omitted Two Lines of Type from Top of Page 3

ML20217J220
Person / Time
Site:	Crane
Issue date:	03/26/1998
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 9804060233
Download: ML20217J220 (6)

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l CORRECTED COPY GPU Nuclear,Inc.

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Route 441 South NUCLEAR Post Office Box 480 -

Middletown PA17057 0480 Tel717 944 7621 March - 26.1998 1920-98-20131 U. S. Nuclear Regulatory Commission

' Attn.: Document Control Desk

. Washington, DC 20555 '

Dear Sir:

l

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" This " corrected copy" is provided to correct'an error in the March' 16,1997 GPU Nuclear response to GL 97-05 because two lines oftype.were missing from the top of page 3 of the letter that was sent to the NRC on that date.

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 sten generator tubes with indications in service based upon sizing, and (2) If the response to (1) is affinnative, those liceasees should submit -

. a written report that includes, for each type ofindication a description of the asscximated nondestructive examination method being used and the technical basis for acceptability of the technique used.

INTRODUCTION:

The nuclear power indusiry recently voted to adopt an initiative requiring each utility to meet the intent of the guidance provided by the Nuclear Energy Institute (NEI) in 'NEl 97-06, " Steam Generator :

. Program Guidelines," no later than the first refueling outage starting after January 1,1999. ' Adherence to these NEI guidehnes 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 Exammation," 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 y

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mechanisms are divided into the following categories for quali6 cation: thinning, pitting, weai, outside 3

diameter Inter-Granular Attack / Stress Corrosion Cracking (IGA / SCC), primary-side SCC, and -

impingement damage.

'lest samples are used to evaluate detection and sizing capabilities for qualificatien purposes. While pulled tube samples are preferred, fabricated _ samples may be used. Iffabricated test samples are used, the samples are ven6ed 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 veri 6ed as part of the Appendix H qual!6 cation process.

The eddy current detection and sizmg procedures qualified in accordance with Appendk H specify the '

essential variables for each technique. These essential variables ~are associated with an individual -

instrument, probe, cable, or particular on-site equipment con 6gurations. Additionally, certain l techniques have undergone testing and review to quantify sizing perfonnance The sizing data set

. includes the detection data set for the technique with additional requirements for number and composition of the grading units.

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~ GENERAL'INFORMATION:

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,.

Inconel 600.

. The TMI-1 Quality Assurance Program implements the guidance found in the American Society of '

Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Sections V and XL The plant adheres to the 1986 Code, with no Addenda. Steam generator tube indications left in service are those :

which are sized at less than the Technical Specification value of 40 % TW (or repair limit with a consideration for growth).

RESPONSE TONRC QUESTION 1:

1

.At TMI-1, sizing techniques are used during steam generator inspections to leave flaws with the i

following dagradation modes in service:

,.a) Mechanical wear at tube support plates, -

b) Inside Diameter (ID)1GA, in accordance with TMI-l License Amendment No. 206, dated October 16,1997, and 1

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) g Tube Inspection Report with ASME NIS Data Reports for inservice Inspection (ISI)".

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1920-98-20131 Page 3 of 5

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RESPONSE TO NRC QUESTION 2:

The basis for application of these siz:ng techniques is the conduct of the examinations in accordance ivith the TMI-l Quality Assurance Program following the requirements of Sections V and XI of the ASME Code 1986 Edition with no Addenda, and Regulatory Guide 1.83. Additional support for sizing degradation specific mechanisms is p ovided by the EPRI Appendix H qualification data sets.

a) Mechanical Wear at Tube Support Plates Mechanical wear indications are identified by the use of the bobbin coil examination at TMI-1. All

'l 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 i

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.

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The sizing procedure is based on the analysis of 26 sample data points. The samples ranged in

. depth from 22% to 10J%. This database has been reviewed to ensure that application of the sizing 1

. 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 Examination Guidehnes, Revision 5.

The current OTSG wear sizing qualification is a site specific qualification due io the use of the mix channel instead of the raw data frequency for sizing. Examination data acquired using the mix

• l 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 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 E' xpansion Transitions On October 16,1997, the NRC issued TMI-l License Amendment No. 206, which authorizes an.

~ lternate tube repair criteria for ID IGA for one operating cycle. Cycle 12. The amendment a

requires that "ID IGA indications shall be repaired or iemoved 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 volumetricin nature

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~ The TW dimension, if assigned, was assigned using a 0.540" high frequency bobbin coil probe. ID IGA 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 IGA signals < 3:1 signal-to-noise ratio I

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. and < 1 volt were assigned a three letter code, "BVC." A TW dimension was assigned usir.g the 400 kHz differential channel or, for signals in the tubesheet crevice that were not clear with the 400 kHz channel, the 400/200 kHz mix channel was used. Both channels employed a phase angle i

Voltage normalization was established by setting the 400 kHz differential channel to a 10 volt measurement technique established from the 20%,60%, and 100% TW ASME drilled holes.

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

This bobbin coil sizing qualification was established using 21 TMI-1 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 funher.

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

All volumetric IGA indications detected during the bobbin exammations were examined with a motodzed 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 i

conservative when compared to 23 machined OTSG tube samples. These sample 2 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 the 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 of suspected ID IGA. No detectable

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leakage was noted dudng any of the leak tests and the lowest burst pressure was measured at

'.10,800 psig. The findings from laboratory analysis will be reported to the NRC by June 1,1998.

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c. Kinetic Expansion Region Flaws

- The TMI-l 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 i

our most recent 1997 outage. These indications are apt from an active damage mechanism. The i

~ 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 contdbutors and were considered to be 100% TW for the leakage calculations. The structural integdty evaluation considers all measured flaw circumferential and/or axial extents to be 100%

..TW (For vo'.umetric indications, the structural integrity evaluation considers a volumetric indication as consisting of a 100% TW axial crack oflength corresponding to the axial extent of

' the volumetric indication, as well as a 100% TW circumferential crack corresponding to the circumferential extent of the volumetric indication.) Further details on the dispositioning of kinetic l

6we b

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1920-98-20131

' Page 5 of 5 expansion indications m'ay be found in the GPU Nuclear submittal dated January 12,1998," Cycle 12 Refueling (12R) Outage Once Through Steam Generator (OTSG) Tube Inspection Report with

- ASME NIS Data Reports for Inservice Inspections (ISI)," Enclosure 1, Attachment 1, Section

. II.C.2.

For Primary Water Stress Corrosion Cracking (PWSCC) circumferential ca.cks in the kinetic expansion region, the 300 kHz signal from the mid-range plus point is used to size the depth and length. A calibration curve is established using the 40%,60%, and 100% TW ID notches of the calibration standard. Vohage normahzation is performed from the 100% TW hole at 10 voks.

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% ta 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 conseivatism of the accident leakage volumes calculated for flaws in the kinetic expansion region.

Sincerely, p.

G A

' James W. Langenbach Vice President and Director, TMI L

' Attachment-MRK cc:. _ Administrator, Region I TMI Senior Resident Inspector TMI Senior NRC Project Manager j

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1920-98-20131

~ Attachment i

METROPOLITAN EDISON COMPANY JERSEY CENTRAL POWER AND LIGHT COMPANY PENNSYLVANIA POWER AND LIGHT COMPANY GPUNUCLEARINCORPORATED L

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

Operating License No. DPR-50 Docket No. 50-289 i

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 behalf of GPU Nuclear. To the best of my knowledge and belief, the statements cc.,*ained in this document are tme and correct. To the extent that these statements are not based ca ray personal knowledge, they are based upon information by other GPU Nuclear employees and/er cc.multants. Such information has been reviewed in accordance with company practices end I believe it to be reliable.

1 James W. Langenbach j

Vice President and Director, TMI Signed and sworn before me this i

.i day of

,1998.

26th March i

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