ML20217A867

Responds to 970909 RAI to Support TS Change Request 268, Submitted on 970812.Encl Provides Clarification & Addl Justification for Changes Requested

ML20217A867
Person / Time
Site:	Three Mile Island
Issue date:	09/15/1997
From:	Langenbach J GENERAL PUBLIC UTILITIES CORP.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
6710-97-2401, NUDOCS 9709230030
Download: ML20217A867 (12)

Text

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j GPU Nucleaf. Inc,

( Foute 441 South NUCLEAR V* *' 0"" B" **

Mddletown. PA 17057-0480 Tel 717 944 7621 September 15, 1997 6710 97-2401 U. S. Nuclear Regulatory Commission Attention: Document Control Desk ,.

Washington, DC 20555 7

i

Dear Sir:

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Subject:

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

Operating License No. DPR-50 Docket No. 50-289 Response to NRC Request for Additional information Regarding Technical Specification Change Request (TSCR) No. 268 In accordance with 10 CFR 50.4(b)(1), enclosed is the GPU Nuclear response to a request for additional information to support TSCR No. 268, which was submitted on August 12, 1997 This information is submitted in response to a request by the NRC dated September 9,1997.

Also enclosed is the Certificate of Service for this request certifying senice to the chief executives of the township and county in which the facility is located, as well as the designated oflicial of the Commonwealth of Pennsylvania, Bureau of Radiation Protection.

The purpose of TSCR No. 268 is to request changes to the Surveillance Specification for Once Through Steam Generator (OTSG) insenice inspections for TM1-1 Cycle 12 Refueling (12R) examinations applicable to TMI l Cycle 12 operation. GPU Nuclear has requested that an amendment be issued on or before October 3,1997.

The enclosure provides clarification and additional justification for the changes requeste:! by GPU Nuclear in TSCR No. 268. TSCR No 268 included the revised technical specification pages and our analysis using the standards in 10 CFR 50.92 to conclude that the proposed 9709230030 970915 @lylll$ %ll(lhll-PDR ADOCK 05000289 P PDR

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, ' changes would not constitute a significant hazards consideration. Neither the significant i hazards consideration analysis nor the proposed revised technical specitication pages are affected by this additional information.

Sinceiety, vtts)0]fdn< 2,1htt' 7

James W. Langenbach Vice President and Director, TMl Attachment MRK l

cc: Administrator NRC Region I l

! TMI Senior NRC Resident inspector l TMI.1 Senior NRC Project Manager l l

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3 UNITED STATES OF Ah1 ERICA NUCLEAR REGULATORY COhihilSSION l i

IN Tile hiATTER OF DOCKET NO. 50 289

GPU NUCLEAR INC. LICENSE NO. DPR 50

CERTIFICATE OF SERVICE j i This is to certify that a copy of this response to an NRC request for additional information to l l- support Technical Specification Change Request No. 268 to Appendix A of the Operating j j License for Three hiile Island Nuclear Station Unit 1, has, on the date given below, been

filed with executives of Londonderry Township, Dauphin County, Pennsylvania; Dauphin 1 County, Pennsylvania; and the Pennsylvania Department of Environmental Resources, Bureau i

of Radiation Protection, by deposit in the United States mail, addressed as follows:

Mr. Darryl LeHew, Chairman his. Sally Klein, Chairman 1oard of Supervisors of Board of County Commissioners Londonderly Township of Dauphin County

. R. D. #1, Geyers Church Road Dauphin County Courthouse

{

hilddletown, PA 17057 liarrisburg, PA 17120 f

Director, Bureau of Radiation Protection PA Dept. of Environmental Resources Rachael Carson State Office Building P,0. Box 8469 11arrisburg, PA 17105 8469 Att

hir. Stan hiaingi GPU NUCLEAR INC. -

BY: 4161 44d I pce President anc(pirector, Thil I

DATE: /F/97 4

j

i METROPOLITAN EDISON COMPANY JERSEY CENTRAL POWER & LIGilT COMPANY AND PENNSYLVANIA ELECTRIC COMPANY TilREE MILE ISLAND NUCLEAR STATION, UNIT 1 Operating License No. DPR 50 Docket No. 50 289 Technical Specification Change Request No. 268 Response to NRC Request for Additional information COMMONWEALTil OF PENNSYLVANIA )

) SS:

COUNTY OF DAUP111N )

This response to an NRC request for additional information is submitted in support of ,

Licensee's request to change Appendix A to Operating License No. DPR-50 for Three Mile Island Nuclear Station, Unit 1. All statements contained in this submittal have been reviewed, and all such statements made and matters set forth therein are tme and correct to the best of my knowledge, GPU NUCLEAR, INC.

BY: _ b 4kJJ h , f/Ns !dt gee President a'nd Dpctor, TM1 Swom and Subscribed to before me Jl this /5Aday of 3/P7/wdu_. 1997.

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. 6710 97 2401 Attachment Page 1 of 8 GPU Nuclear Response to an NRC Request for Additional Inibrmation Regarding Technical Specification Change Request No. 268 Ily letter dated August 12, 1997, GPU Nuclear submitted a proposed revision to the TMI l Technical Specifications (TS) to permit the use of dimensional based steam generater tube repair criteria to disposition tube indications during the Cycle 12 Refueling (12R) Outage

.I inspections. The p oposed tube repair criteria apply only to inside diameter (ID) volumetric intergranular attack (IGA) indications. The proposed TS changes would allow the disposition of inside diameter volumetric IGA indications based on bobbin coil depth measurement, if assigned, and motorized rotating pancake coil (MRPC) probe dimensional measurements. The

, proposed TS changes would be applicable for one cycle only (Cycle 12) until the Cycle 13 Refueling (13R) Outage which is planned for September 1999.

This letter provides the GPU Nuclear response to the NRC's request for additional information dated September 9,1997 to support NRC review of Technical Specification Change Request (TSCR) No. 268. This supplemental information does not affect the analysis of no significant hazards consideration nor does it affect the revised Technical Specification pages submitted with TSCR No. 268. The GPU Nuclear response is provided along with the statement of the NRC's question as follows:

NRC Guestion No.1:

1he proposed amendment woodd ajydy only to intergramdar attack (IGA) degradation initiatingfrom the inside diameter (ID) of a steam generator tube, 1his degradation was a consequence of the sodium thiosulphate intniston in the early 1980's, and the licensee has concluded snat its growth has arrested. Describe how these indications are trackedfrom outage to outage, and discuss the provisions included in the data pnalysis guidelines to ensure that the proposed repair criteria are only atydied to those indications dating back to the original thiosulphate intrusion? Explain how new IGA degradation (i.e., not a result of thiosulphate intrusion) will be ident@ed and dispositioned during the course of the tube examinations. Discuss whether new indications will be dispositioned using the proposed i' udtageklimensional based limits and whether the tube would be counted in the class @ cation ofinspection results per 1S 4.19.3. ,

GPU Nuclear Response:

Tubes with known ID IGA are examined by at least the bobbin coil probe each outage after identification. If bobbin coil examination identifies an apparent change in the indication (e.g.,

significant increase in voltage or percent through wall), it has been our practice to perform a pancake coil probe examination The degraded tube population has been evaluated for potential trends ofincreasing voltage and/or indication depth penetration; the results of these evaluations have been reponed to the NRC in the reports submitted twelve months after each refueling outage in accordance with TS 419.5.

The analysis guidelines for the unexpanded region of tubing require that, for bobbin coil examinations, the indication be < 30' phase angle to be considered an ID indication. This i

, 6710 97 2401 Attachment Page 2 of 8 requirement allows only those indications which fall into the ID phase plane of the ECT calibration curve to be considered tD indications on a first screen. During the 12R Outage, all indications of apparent tube wall degradation, as identified by bobbin coil examination, will be examined with a rotating pancake coil probe. The 12R Outage Examination Guidelines require that ID IGA less than the poposed 0.25" axial length and less than the 0.57" circumferential length be assigned a code of"VOL" (ID Volumetric). This will assure only ID IGA indications are dispositioned using the proposed TS acceptance criteria.

During the 12R Outage, newly identified ID IGA indications will be treated as though they are the result of the original thiosulphate intrusion if they follow the pattern of the original damage. This pattern has been that they are present in the upper elevations of the steam generator (generally above the 15* support plate), are ID initiated, volumetric in nature, and have both a small axial and small circumferential extent. Newly identified ID IGA indications will be counted toward the classification ofinspection results in accordance with the proposed TS 4.19,3 and will be dispositioned in accordance with proposed TS 4.19.4.a.6. Where ECT data exist subsequent to the kinetic expansion repairs (back to and including 1986 SM Outage), that ECT data will be used as necessary to assist in assuring that the ID IGA indications observed are related to the original thiosulfate intrusion.

New ID IGA would have to be caused by a new chemical contaminant excursion on the primary side of the OTSGs; and no reduced sulfur or other significant chemistry excursions have been noted during the plant's recent operating cycles. The sodium thiosulphate tank has been drained and physically removed from the TMI l Building Spray System design, so a repeat of the original damage is not possible. The likelihood of a new ID initiating damage mechanism (causing volumetric IGA) is extremely small and would be very evident when growth rate studies are performed for operational assessment of the OTSGs. The results of growth rate studies will be reported to the NRC in accordance with TS 4.19.5.

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Pago 3 of 8 j NRC Question No. 2:

Given the limitathms in relying on eddy current phase angle to vert,& the initiation surface of tube degradation, describe the cdh current data analysis methodology used to distinguish between ID IGA and degradation initiatingfrom the outside diameter of a tube. In addition, provide details on how the technique was quallfled to make such a charactert:ation.

- GPU Nuclear Response:

The bobbin coil examination technique was qualiGed and reviewed by the NRC as described  ;

in the Safety Evaluation Report for License Amendment No.103, This document concluded i that the examination technique could identify and quantify ID IGA indications appropriately.

The 12R Outage examination guidelines restrict the indications being considered ID IGA to those that provide a bobbin coil phase angle of < 30' (ID flaw plane on the phase angle

depth calibration curve).

< Follow up 12R Outage MRPC examinations, required for all tube wall degradation indications, will further substantiate whether a flaw is both volumetric and ID initiated. The pancake coil responses from the volumetric ID IGA indications exhibit unmistakable ID surface initiating characteristics. The peak to peak phase angle is in the ID flaw plane (< 40'

phase angle with most of the indications being less than 20 or even rotating into the less than i

0' axis (270 to 360* quadrant)). Review of the lissajous responses from the edges of the flaws show a distinct formation from the ID flaw plane, which is about 120* out of phase

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angle rotation from the responses for OD flaws. -In addition, the pancake coil terrain plots have a distinct " volcano like" appearance, typical of volumetric indications.

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' NRC Question No. 3:

1he proposed changes to 15 4.19.4.fa.3] include a provision to class # a tube as degraded if the w>ltage qf an ID IGA indication is equal to or exceeds 0.5 w>hs. Erplain the basis for establishing a 0.5 udt ddgradation threshold and discuss its telationship to the existing 20-percent deinh based threshold.

GPU Nuclear Response:

Proposed TS 4.19.4.a.3 classifies a tube as degraded if the ID IGA bobbin coil indication exceeds 0.5 volts or 0.13 inches axial extent or 0.28 inches circumferential extent (for 12R Outage examinations). The current TS categorization of a tube as degraded when it contains imperfections > 20 percent through wall (T.W.) is also retained for those tubes where the eddy current signal is sutTicient to reliably perform depth sizing by phase angle analysis. The voltage and length criteria are intended to provide a similar level for classifying degraded tubes where the eddy current indication signal does not present suflicient amplitude for conventional percent T.W. depth sizing. The circumferential and axial extent values proposed are approximately half the repair limit values, similar to the existing TS definitions which use 40 percent T.W. for defective tubes and 20 percent T.W. for degraded tubes. Additionally, a tube would also be considered degraded ifit contained an eddy current indication that equals or exceeds 0.5 volt amplitude, even ifit does not meet either of the proposed length criteria or the existing TS 20 pe cent T.W. value. The 0.5 volt value was chosen to be a conservative value that will ensure continued monitoring of tubes with small measurable eddy current indications for growth in future outage inspections.

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l 'NRC Question No. 4:

j i Has a growth rate study ofID IGA indications been performed considering rotating pancake

} call (RPC) data (e.g., w>ltage, length, width? If so, provide the restats. If not, discuss the l usefidness of such a stuh.

1 l GPU Nuclear Response:

! GPU Nuclear has not performed a growth study ofID IGA indications using MRPC data to date. On completion of the 12R Outage examinations, a growth rate study will be performed j to further confirm that these indications are inactive (not growing). The new growth rate

study, which is expected to include approxin'ately 100 indications, will compare the 10R and 11R Outage MRPC examination results for ID IGA indications with the 12R Outage MRPC

examinations for ID IGA. - A summary of this study will be included in the outage report to be submitted in accordance with TS 4.19.5.b.

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'NRC {hiestion No. 3:

1he triformation provided on the assessment of growth rate for the ID IGA degradation does not appear to address how do[ferences in data acquisition between inspections were considered in the studier Discuss how changes in the acquisition of edh current data (e.g.,

probes, cables, testers, etc.) affect the udtages ofindications measured in previous inspections.

, GPU Nuclear Response:

For all the data used in the growth study discussed in Appendix A of our August 28,1997 submittal, bobbin coil examinations were performed with high frequency probes. GPU Nuclear began using "high performance" data cables during the 10R Outage when they became available; but they would not have had an impact on voltage measurement because voltages are assigned us ng the complete examination system. Voltage measurements (normalization) for all tho outage examinations since 1982 have been performed using the 400kilz ailferential responte fiom the four 20 percent drilled holes of the ash 1E calibration standard. Thus a 2.0 volt indication in 1982 is equivalent to a 2.0 volt indication in 1997.

GPU Nuclear used the Zetec hilz-18 for the 9R and 10R 6t.. ages and the hilZ-30 during the llR Outage. Voltege normalization was identical for all three outages, so again a 2.0 volt indication in the 9R Outage is equivalent to a 2.0 volt indication during the llR Outage.

The hilZ 30 drive and voltage settings used during the llR Outage provided equivalent or superior performance to the hilZ-18, but had no effect on ID IGA voltage or phase angle measurements.

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6710 97 2401 Attachmcnt Page 7 of 8

'NRC Question No. 6:

lhe discussion on in situ pressure testing inchtded in the subrnitial dated August 28,1997, states th. criteriafor tube selectionfor the upcorning outage, lhe criteria spec (fy that tubes with the highest bobbin coil voltages coupled with the lowest dirnensional extent as sneasured b)* RPC and those tubes containing the indications with the highest arial and'or circuntferential lengths will be consideredfor testing. The sta[{ notes that, based on previous experience with in-situ pressure testing, tube selection inay need to address other quantitative ineasurernents (e.g., RPC voltage, phase angle depth measurernents) and qualitative aspects (e.g., signal quality) of the eddy current data in order to deterrnine the best candidatesfor testing. Discuss whether suchfactors will be considered in the tube selection criteria during the upcorning refueling outage. If criteria other than that included in the submittal are not going to be consideredfor tube selection, discuss the basisfor this decision.

GPU Nuclear Response:

GPU Nuclear intends to use the ECT examination results to identify those ID IGA indications that exhibit the most degradation (e g., those that would have the expected lowest burst pressure based on ECT results). Both the bobbin coil and RPC data will be reviewed to make this determination. One flaw parameter such as voltage may be outweighed by longer flaw length or higher percent T.W. in a different flaw. Flaws will be selected for in situ pressure testing which exhibit the theoretically highest average T.W. penetration over a given length (similar to a percent degraded area measurement for circumferential crack measurement) because these would have the lowest expected burst pressure.

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NRC Question No. 7:

The proposid to use in-situ pressure testing to assess the leakage integrity of tubes inydicitly relles upon the ability of the test to adequately sirnulate tube loads during postulated accident conditions. According to the subrnittal dated August M 1997, "lt}he in-situ pressure test causes a higher princit xd stress in the circurnferential direction than that caused by the MSLB hxid in the axial direction. " Ahhough the test hxsdingfor arially orientedflaws should be consent 1th'e with respect to accident induced hxuls provided fernperature related e[{ects are also considered, the staff cannot determine whether the huidingfor circurnferentially oriented ddects will also be consenutive during the in-situ pressure tests. Discuss the quahfication of the in-sint pressure test device and its ability to adequately sirnulate tube loads.

GPU Nuclear Response:

GPU Nuclear has proposed to use in-situ pressure testing to assess the leakage integrity of tubes having indications of volumetric intergranular attack (IGA) on the inside diameter (ID) surfaces. Volumetric IGA is " patch like," thus it is characterized by both axial and circumferential extent. The " patch-like" morphology is further substantiated when reviewing the eddy current test (ECT) lissajous response behavior.

The proposed in situ pressure testing will subject the individual tubes to a pressure that will satisfy R.G.1.121 criteria and result in a higher principal stress in the circumferential direction than that applied in the axial direction during a postulated MSLB accident. The in-situ pressure test creates a more severe condition than the postulated faulted condition with regard to the potential for causing tube leakage because a higher principal stress is applied across the smaller of the two limiting dimensions that are identified in the dispositioning criteria for ID IGA damage. The axial extent of these volumetric flaws is limiting, for example, the proposed dispositioning criteria permit potential ID IGA damage extent that is

r. bout twice as long in the circumferential direction (0.57") than in the axial direction (0.25")

A ten percent pressure increase is included as a correction to account for changes in the inconel 600 tube properties with temperature so that in-situ pressure testing will accurately predict the capability of the tubes under the postulated accident pressure and temperature conditions.

The in situ pressure test design criteria requires that 1) the process be designed to induce normal and faulted loads across a tube defect location, and 2) the process be able to deliver suflicient pressure to satisfactorily envelope NRC Regulatory Guide 1.121 structural limits and uncertainty margins The qualification determined that the in situ pressure test process / system provides good correlation of actual to predicted stress values and is suitable for use in steam generator tubing to verify structural integrity l