ML20217G020

Proposed Tech Specs Re Revised Pages 4-80 & 4-81 Previously Submitted

ML20217G020
Person / Time
Site:	Three Mile Island
Issue date:	10/03/1997
From:	GENERAL PUBLIC UTILITIES CORP.
To:
Shared Package
ML20217G013	List: ML20217G007 ML20217G016 ML20217G020
References
NUDOCS 9710090154
Download: ML20217G020 (11)

Text

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ATTACHMENT I

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SUPPLEMENTAL TMI-I TECHNICAL SPECIFICATION REVISED PAGES i;

N 9710090154 971003 PDR ADOCK 05000289 ,

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• 4.193 ]mpnlion Frequency (Continued)

- c. ' Additional, unscheduled inservice inspections shall be performed on each steam generator in accordance with the first sample inspection specified in Table 4.19-2 during the shutdown subsequent to any of the following conditions: .

1. A seismic occurrence greater than the Operating Basis Earthquake.

2 A loss of coolant accident requiring actuation of engineering safeguards, or

3. A major main steam line or feedwater line break.

d. After primary-to-secondary tube leakage (not including leaks originating from tube-to-mbe sheet welds) in excess of the limits of Specification 3.1.6.3, an inspection of the affected steam generator will be performed in accordance with the following criteria:

1. If the leak is above the 14th tube support plate in a Group as defined in Section 4.19.2.a.4(1) all of the tubes in this Group in the affected steam generator will be inspected above the 14th tube support plate. If the restilts of this inspection fall into the C-3 category, additional inspections will be perfcrmed in the same Group in the other steam generator.

2. If the leaking tube is not as defined in Section 4.19.3.d.1, then an inspection will be performed on the affected steam generator (s)in accordance with Table 4.19-2.

4.19.4 Acceptance Criteria

a. As used in the Specification:

1. Imperfection means an exception to the dimensions, finish, or contour of a tube from that required by fabrication drawing or specifications. Eddy current testing indications less than degraded tube criteria specified in a.3 below may be considered l imperfections.

2. Drgra. ation d it.eans a service-induced cracking, wastage, wear or general corrosion occurring on either inside or outside of a tube.

3. Degraded Tube means a tube containing :

l (a) an inside diameter (l.D.) IGA indication with a bobbin coil indication 2 0.5 volt or 2 0.13 inches axial extent or 2 0.26 inches circumferential extent (for 12R outage examinations and Cycle 12 operation only), or (b) imperfections 2 20% of the nominal wall thickness caused by degradation.

4. % Degndation means the percentage of the tube wall thickness afTected or removed by degradation.

4-80 Aiaendment No. l#,--1497 #3

4.19.4 Acceptance Criteria (Continued)

5. Defect means an imi 'rfection of such severity that it exceeds the repair limit. A tube containing a defect is defective.

6. Repair Limit means the extent of degradation at or beyond which the tube shall be repaired or removed from service because it may become unserviceable prior to the next inspection.

This limit is equal to 40% of the nominal tube wall thickness. For Outage 12R -

examinations a id Cycle 12 operation only, inside diameter 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 through wall degradation

, dimension of > 40% if assigned. l 2

7. Unserviceable describes the condition of a tube ifit leaks or contains a defect large enough to affect its structural integrity in the event of an Operating Basis Earthquake, a loss of coolant accident, or a steam line or feedwater line break as speciGed in 4.19.3.c.,

above.

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8. Ip_b_e_ Inspection means an inspection of the steam generator tube from the bottom of the upper tubesheet completely to the top of the lower tubesheet, except as permitted by 4.19.2.b.2, above.

9. Inside Diameter Inter-Granular Attack (IGA) Indication means a bobbin coil indication initiating on the inside diameter surface and conGrmed by diagnostic ECT to have a volumetric morphology characteristic of IGA.

b. The steam generator shall be determined OPERABLE aner completing the corresponding actions (removal from service by plagging, or repair by kinetic expansion, sleeving, or other methods, of all tubes exceedir.g the repair limit and all tubes containing throughwall cracks) required by Table 4.19-2.

4.19.5 Rep _ qts

a. After the completion of each inservice inspection of steam generator tubes, prior to exceeding a reactor coolant system (RCS) temperature of 250 F, the NRC shall be notified of the following:

1) The number of tubes repaired or removed from service in each steam generator,

2) An assessment of growth ofinside diameter IGA degradation, and

3) Results ofin-situ pressure testing,if performed.

4-81 Amendment No. 17,83,91,103,12M4974437467

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ATI'ACHMENT II GPU Nuclear Response to

- NRC Request for Additional Information Regarding Technical Specification Change Request No. 268 i

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l GPU Nuclear Rerponse to NRC Request for Additional Information Regarding Technical SpeciScation Change Request No. 268 By letter dated August 12,1997, GPU Nuclear submitted a proposed revision to the TMI-l Technical Speci6 cations (TS) to permit the use of dimensional based steam generator tube repair criteria to disposition tube indications during the Cycle 12 Refueling (12R) Outage inspections.

The proposed tube repair criteria apply only to inside diameter (ID) volumetric intergranular attack (lGA) indications. The proposed TS changes would allow the disposition ofinside diameter volumetric IGA indications based on both 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 October 1,1997 to support NRC review of Technical Specincation Change Request (TSCR) No. 268. The GPU Nuclear response is provided along with the statement of the NRC's question as follows:

NRC Question No. I:

The response to Question 7 of the NRC 's requestfor additionalinformation in the licensee 's submittal dated September 15,1997 did not appear to adequately respond to the staff's inquiry.

Specifically, the staffrequested that the licensee discuss the loads imparted to a predominantly circumferential defect imposed by the in-situ preswre testing. Does the in-situ pressure test device induce axial stresses into the tube that are consermtive with respect to the load stresses postulated to occur under the most limiting conditions (i.e., design basis)? Discuss u hether am' tubes with limiting circumferentially oriented degradation are being consideredfor in-situ pressure testing in the current antage. Provide details on the most sigmficant circumferential indications identified in the inspections.

GPU Nuclear Response; in situ pressure, test devices currently used at TMI-l may induce incidental axial stresses into the tube, depending on the specinc combinations of test device probes used and the location and length of the tube being tested. The principal design feature of these test devices is to achieve high internal test pressures with a capability to quantitutively measure leakage at ihe test pressures; these devices are not speci6cally designed to achieve axial loads equivalent to postulated Main Steam Line Break (MSLB) axial loacis and they do not conservatively approach those loads.

There were no tubes found during the 12R Outage inspections that exceed the circumferential repair limit (> 0.52 inches). The average indication size found was approximately 0.15 inches in circumferential extent with a correspoading average axial extent of 0.15 inches representative of volumetric indications. The most signincant circumferential indication identined in the current outage inspections is sized at 0.51 inches long by plus point MRPC. It is located within the upper tubesheet approximately 1.70 inches above the upper tubesheet secondary face in "B" OTSG, Row 118, Tube 38. The tube containing this indication will be in situ pressure tested (and then plugged) along with six other tubes which are planned for in situ tests. The other tubes being tested have indications with circumferential extents ranging from 0.24 inches to 0.50 inches.

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.. 1 6710-97-2440'-

Enclosure II- l Page.2 of 5 - l 1

F NRC Question N'o.- 2:

( Other licensee 's that have used in-situ pressure testing to assess degraded or defective tubes -

i often adjust the testpressure to accountfor tube locking within support structures. Discuss whether these effects were consideredin establishing he in-situ testpressures. If these effects

were taken into account, state where it was assumed the tube was locked (e.g., tube support

plate, tubesheet, etc.). Did the quahfication of the test device consider locked tube ejfects?

i i GPU Nuclear Response:

_The planned in situ testing at TMI-l takes into account the effect of locked tubes. A factor is q applied to the test pressures to account for tube being fixed into both the upper and lower tubesheets.1The tubes are assumed to be free to move at each support location. The OTSG_-

L incorporates broached tube support plates and TMI-l was recently chemically cleaned on the

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secondary side. Tube pulls at all OTSG sites have also verified freedom of movement through l the support plates to support this assumption. _ Qualification testir.g of the in situ test device to be

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4 used at TMI-l did consider and account for locked tube effects.

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-- NRC Question No. 3:

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! .For the current outage, were allirufications detected by bobbin coilinspected with a rotating l probe? Ifno, discuss which irulications were not consideredfor additionalinspections arul l provide the basisfor diqx>sitioning the tubes with these trufications, i'

I - GPU Nuclear Response:

During the 12R Outage all suspected ID IGA and otherwise " flaw-like" indications, were I' examined with MRPC. In additioni dents and dings above the lower tubesheet were examined with MRPC and all dents at the lower tubesheet > 40 volts as recorded by bobbin _ coil probe were i examined by MRPC (50 tubes in OTSG "A" and 196 tubes in OTSG "B"); ' Dents less than 40 volts at the lower tubesheet were not examined because no PWSCC or ODSCC was detected in o the larger voltage dents examined. The higher voltage dents are considered the most likely to L exhibit service-induced cracking because they exhibit a higher voltage as a result of greater reduction in tube diameter.

L-p Non " flaw-like" indications were not examined with a rotating probe. Inside diameter (ID) chatter (pilgering) was not examined with MRPC because this is a tube wall anomaly which is L recorded to note where it occurs. (This condition has been recorded during previous .

Lexaminations and is not considered to interfere with indication detection.) A three frequency turbo mix was also used to further suppress this condition. ID chatter is not associated with a

' flaw mechanism. Permeability variations were not examined with MRPC because these signals L are associated with the magnetic properties of the tube wall and are not associated with tube wall l degradation.

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'6710-97-2440-Enclosure 11 Page 3 of 5 NRC Question No. 4:

Ahhough there is a theoretical basisfor concluding that edh current inspection techniques can distinguish between inside diameter (ID) and outside diameter (OD) initiated ddects, the staffis aware ofseveral occurrences at other PWRfacilities uhere ID defects were called OD during inspections and vice-versa. Ghen the difficadties associated with resahing the initiating surface for tube degradatio o, discuss how data discrepancies regarding the ID or OD nature of indications between bobbin coil and rotatingprobes will be resohrd during the inspections. The staffnotes that the licensee 's response to Question 2 in the submittal dated September 15,1997, indicates that the staffagreed with the original quahpcation of the bobbin coil examination technique. The staffhas reruviewed the original assessment and concluded that the scope of the original quahpcation did not address the resolution ofID and OD initiated defects. 1herefore, the statements included in the staff's enduation as documented in NUREG-1019 do not apply.

GPU Nuclear Response:

The Thil-l ID initiated Daws generally occur in the freespan of tubing where examinations with both the bobbin coil and hiRPC probes provide information to evaluate the initiating surface.

The TMI-l guidelines require that any bobbin coil signals outside the ID indication phase plane be assigned an "NQI" code which requires that the indication signal be treated as an OD flaw so that the repair criteria incorporated by TSCR No. 268 would not be applied. The Thil-1 guidelines require that ID initiated volumetric indications with extents less than the proposed axial length and circumferential length repair limits be assigned a "VOL" code signifying the 4 signals' ID origin. ID initiated indications with greater lengths or OD indications require assigning a code ending in "1" and those indications will be repaired if they are located in freespan tubing or the tubesheet unexpanded crevice region. This is a " dual probe" approach requiring that both the hiRPC and bobbin coil examinations indicate that the flaw is ID initiating in order to be considered for evaluation using the TSCR No. 268 repair criteria. This " dual probe" approach is similar to bobbin and 8xl absolute coil examination approach which was evaluated and accepted by the NRC in NUREG-1019, Supplement 1, the safety evaluation report (SER) which approved Technical Specification Amendment No.103 following the Thil-1 steam generator repairs.

Dimculty in determining the surface origin of flaws may occur if the flaws are 100%

through-wall. This could be extremely dimcult if these indications were to occur in an expansion transition at the tubesheet seconday face where expansion transition, tubesheet interface, deposit, dant, and flaw signals could all be present in the area being evaluated. The Thil-1 ID IGA flaws do not occur at such dimcult areas; these indications are not 100% th.ough wall; and bobbin coil data is available for freespan indications to further aid evaluation.

Therefore, since Thil-l does not have these dimculties, ID indications and OD indications can be differentiated reliably.

NRC Question No. 5:

Discuss how the bobbin coil an.:rotatingprobes will be calibratedfor the inspections.

Specifically, address the phase angle associated with a 100% hole or notchfor each probe type.

'6710-97-2440 Enclosure 11 Page 4 of 5 Also, provide additional hqformation regarding the criteria establishedfor determining u hether an indication isID or OD.

GPU Nuclear Response:

The bobbin coil probe examination requires that only indication signals > 5 and 5 30 be treated as ID initiated indications because these are indications with signals located in the ID indication phase plane. These indications are recorded from the 400 kHz differential channel, mix channel P1, or mix channel P2. The calibration setup for the bobbin coil probe is as listed below:

Phase Channel / Mix Phase Rotation Signal Rotation (Oegrees) 400 kHz Differential 100% Through-Wall Hole 30 P1 (400/200 ditTerential broached TSP mix) 100% Through-Wall Hole 30 P2 (400/200 differential drilled TSP mix) 100% Through-Wall Hole 30 The Plus Point MRPC Technique Qualification and TMI-l Analysis Guidelines require the 40%

ID axial notch be set at 15 for axial sensitive channels and a +180 rotation from this setting for circumferential sensitive channels. The 100% through-wall notch is not set to a specific rotation for this technique because it is most important to assure that ID initiated flaws are rotated appropriately for detection purposes (adjustment of the rotation using the 100% through-wall notch would not assure correct rotation for detection of shallow ID flaws). The 100%

- through-wall notch generally appears at about 26 Degrees. Only indications which display a nhase rotation in the ID flaw plane are considered ID initiated indications.

NRC Question No. 6:

The proposed amendment establishes a circumferentiallength repair limit of 0.57-inches.

However, Appendix D in TDR 423, Revision 1, appears to state that the limitingflaw lengthfor circumferentialdefects is 0.52-inches. Please clanfy this apparent discrepancy between the current proposed TS amendment and the limiting defect length discussed in the previously docketedlicensee analysis.

GPU Nuclear Response:

As discussed with the NRC in a conference call on September 30,1997, GPU Nuclear intends to accept the more restrictive circumferential extent repair limit of 0.52 inches for IGA indications as it appears in the 1983 report (TDR 423, Revision 1) rather than providing additional justification to support the 0.57 inch limit we requested in our original submittal of TSCR No. 268.

We are also reducing the corresponding degraded tube criteria limit from 0.28 inches to C.26 inches because it was our intent that the degraded tube criteria limit for circumferential extent be

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~ 6710-97-2440 Enclosure 11 ,

Page 5 of 5 1 one half the repair limit. Since the 12R eddy current examination results are unaffected by these 1 changes, which are being requested for TMI-l Cycle 12 Refueling (12R) Outage examinations I and Operating Cycle 12 only, GPU Nuclear prefers to expedite NRC approval of the amendment by accepting the more restrictive limits.

NRC Question No. 7:

The licensee 's response to Question 4 as given in the submittal dated September 15,1997 states that 100 indications will be selectedfor a growth rate stuh using rotatingprobe eddy current data. Discuss the criteriafor selecting the indicationsfor this stuh.

I GPU Nuclear Response:

The indications to be used for this study will be ID initiated flaws examined during the 12R Outage which were examined with the 0.115" pancake coil probe during either the 11R or 10R Outages. There will be no selection process; all available data (estimated to be approximately 100 tubes) from the three outages will be used in this study, i

A"ITACllMI:NT Ill Certillcate of Se:Tice for Supple:ncat to TMI l Technical Specification Change itequest No. 268

4 UNITED STATES OF AhiERICA -

, NUCLEAR REGULATORY C0hihilSS10N IN Tile h1ATTER OF DOCKET NO. 50 289 GPU NUCLEAR INC. LICENSE NO. DPR.50 GRTIFICATE OF SERVIG This is to cenify that a copy of the Supplement to License Amendment Request No. 268 for Three hiite Island Nuclear Station Unit 1, has, on the date given below, been filed with -

executives of Londonderry Township, Dauphin County, Pennsylvania; Dauphin County, Pennsylvania; and the Pennsylvania Depanment of Environmental Resources, Bureau of Radiation Protection, by deposit in the United States mall, addressed as follows:

hir. Darryl Let tew, Chairman his. Sally S. Klein, Chairman Board of Supervisors of Doard of County Commissioners Londonderry Township of Dauphin County R. D, #1, Geyers Church Road Dauphin County Counhouse hilddletown, PA 17057 liarrisburg, PA 17120 Director, Bureau of Radiation Protection PA Dept. of Environmental Resources Rachael Carson State Oflice Buil ling P.O. Box 8469

. Ilarrisburg, PA 17105 8469 Att: htr. Stan hialngi GPU NUCLEAR INC.

BY: M1(l/ //8/B Vice President and Dfector, Thil DATE: /0 f/97

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