ML20216B298
| ML20216B298 | |
| Person / Time | |
|---|---|
| Site: | Three Mile Island  |
| Issue date: | 08/28/1997 |
| From: | Langenbach J GENERAL PUBLIC UTILITIES CORP. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| 6710-97-2376, NUDOCS 9709050286 | |
| Download: ML20216B298 (17) | |
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l GPU Nucleer. Inc.
( Route 441 south Mdde wn 17057 0480 1e1717-944 7621 6710-97-2376 August 28e 1997 i
! U. S. Nuclear Regulatory Commission l Attention: Document Control Desk Washington, DC 20555
Dear Sir:
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 additionalinformation to support TSCR No. 268, which was submitted on August 12,1997.
This information is submitted in response to a request by the NRC Staff during a conference call en August 14,1997. Also enclosed is the Certificate of Senice for this request certifying senicel to the chief executives of the township and county in which the facility is located, as well as the '}
designated ollicial of the Commonwealth of Pennsylvania,13ureau of Radiation Protection.
The purpose of TSCR No. 268 is to request changes to the Surveillance Specification for Once g, Through Steam Generator (OTSG) inservice inspections for TMI-l Cycle 12 Refueling (12R) / $
examinations applicable to TM1-1 Cycle 12 operation. GPU Nuclear has requested that an /
amendment be issued on or before October 3,1997.
The enclosure provides clarification and additionaljustification for the changes requested 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 changes would i **., '. i. ' .,.,
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- Q710 97-2376
, Page 2 of 2 not constitute a significant hazards consideration. Neither the significant hazards consideration analysis nor the proposed revised technical specification pages are affected by this additional
, information.
Sincerely, 4M M James W. Langenba Vice President and Director, Thil Attachment MRK cc: Administrator NRC Region i TMI Senior NRC Resident inspector TMI-l Senior NRC Project hianager
hiETROPOLITAN EDISON COhtPANY
,liRSEY CENTRAL POWER & LIGIIT COh1PANY AND PENNSYLVANIA ELECTRIC COh1PANY TliREE hilLE ISLAND NUCLEAR STATION, UNIT 1 Operating License No. DPR.50
{
Docket No. 50 289 l
Technical Specification Change Request No. 268, Additional Information COhihiONWEALTil OF PENNSYLVANIA )
) SS:
COUNTY OF DAUPillN )
This additionalinformation is submitted in support of Licensee's request to change Appendix A to Operating License No. DPR-50 for Three hiile 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 true and correct to the best of my knowledge.
GPU NUCLEAR, INC.
BY:
ViceVresident and Director, Thil Sworn a d Subscribed to before me thi iday org [ ,1997.
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'Notarf7Public tenu sw Sparnfi C. M Mo w , N2fy PUD!iC Londondary Twp , DayJun County l My Commsor. E: pns un R M9 Membet, Fenr4yivaa.e k.waawn ct tuanes
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UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION IN Tile MATTER OF DOCKET NO. 50-289 GPU NUCLEAR INC. LICENSE NO. DPR-50 I
CERTiflCATE OF SERVICE This is to certify that a copy of this response to a request for additional infonnation to support l Technical Specification Change Request No. 268 to Appendix A of the Operating License for Three Mile 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 Department of Environmental Resources, Bureau of Radiation Protection, by deposit in the United States mail, addressed as follows:
Mr. Darryl Leitew, Chairman Ms. Sally Klein, Chairman
- Board ofSupenisors of Board of County Commissioners Londonderry Township of Dauphin County
, R. D. #1, Geyers Church Road Dauphin County Courthouse Middletown, PA 17057 11arrisburg, PA 17120 Director, Bureau of Radiation Protection PA Dept. of Emironmental Resources Rachael Carson State Office Building P.O. Box 8469 liarrisburg, PA 17105 8469 Att: Mr. Stan Maingi GPU NUCLEAR INC.
BY:
- Vice 14dident and Director, TMI DATE:
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- 4710-97-2376
. Attachment Page1 of7 Response to NRC Request for Additional Infonnation Regarding Technical Specification Change Request No. 268 Byletter dated August 12,1997, GPU Nuclear proposed a revision to the TMI l Technical l Specifications (TS) to permit the use of dimensional based steam generator tube repair criteria to disposition tube indications during inspections during the Cy:le 12 Refueling (12R) Outage. The proposed tube repair criteria apply only to inside diameter (ID) volumetric intergranular attack i
(IGA) indications. The proposed TS changes would allow the disposition ofinside diameter volumetric IGA indications based on bobbin coil voltage response 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.
As discussed in a conference call on August 14,1997 between GPU Nuclear and members of the NRC Staff, this letter provides additional information in support of our recently submitted Technical Specification Change Request (TSCR) No. 268 (Reference 1). 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.
A. DACKGROUND TMI l currently has several hundred steam generator tubes in senice with small eddy current indications on their inside diameter (ID) surfaces. These small indications are volumetric intergranular attack (IGA)" remnants" of the sulfur intrusion damage that occurred to the generators (primary side)in 1981 and which necessitated an extensive repair program. The sulfur damage was repaired by kinetic expansion of all of the insenice tubes within the steam generator upper tubesheets and the plugging of more than 1000 tubes. The predominant damage mechanism was intergranular Stress Assisted Cracking (IGSAC) at roll transition locations, where residual stresses within the upper tubesheet are higher. A large number of tubes were also damaged by volumetric, " patch" 1GA at other locations.
In conjunction with the TMI l steam generator repair program, which was reviewed and approved by the NRC in TMI-l Technical Specification Amendment No,103, GPU Nuclear developed an eddy current program able to identify the flaws and demonstrated the assurance of the integrity of the steam generators to the satisfaction of the NRC (Reference 2). Twenty-nine tubes were pulled from the generators to characterize the damage morphologies and develop and benchmark the eddy current program. After the restart of the plant, three more tubes were pulled (in 1986) to re-verify the ability of the eddy current techniques to accurately identify structurally significant flaws and confirm that the IGSAC mechanism is no longer active. Some of the patches oflGA that now remain in ser ice are so small that it is difficult to reliably determine their through-wall extent because of the low amplitude eddy current signals that they produce.
, 4710 97 2376 Attachment Page 2 of 7 GPU Nuclear has inspected and monitoied known patches of IGA in its tubes during each of the outages since the original repairs were performed. Among the proposed TS revisions for Cycle 12 in TSCR No. 268 are additional limits for the axial and circumferential extents of these flaws to ensure that any stmeturally significant ID IGA flaws continue to be removed from senice.
Additional repair limits for defective tubes and limits at which tubes will be considered degraded i are provided for this ID IGA degradation. The proposed changes are written specifically for l flaws which have a volumetric morphology as confirmed by a hiotorized Rotating Pancake !
(hiRPC) probe, in accordance with the proposed TS changes in TSCR No. 268, GPU Nuclear l will perform both bobbin and hiRPC inspections of any ID IGA indications that may continue to remain in senice.
l EMQSED EXTENT 1 lhilTS l
l The proposed tube repair criteria include dimensionallimits. The dimensionallimits are imposed to prevent tube burst during both normal operation and postulated accident loading conditions.
The repair criteria limit the allowable axial and circumferential lengths of the volumetric degradation. The axial extent of tube degradation is limited to 0.25 inches as measured on the ID surface. The circumferential length of any volumetric indication must not exceed 0.57 inches as measured on the ID surface. Tubes with indications that have measured dimensions in excess of either the axial or circumferential dimensional limits are considered defective and will be removed from senice or repaired per the proposed TS.
All ID IGA indications detected with the bobbin probe will undergo an hiRPC probe examination during the 12R Outage to confirm the degradation mode of the indication (e.g., volumetric). The hiRPC probe will also be used to measure the length and width ofindications per the proposed repair criteria. If these dimensional measurements are within the proposed limits, then the degradation in the tube is considered acceptable for continued senice. This is a further enhancement of the dual probe (bobbin plus absolute coil) technique reviewed by the NRC in Reference 2 for dispositioning ID IGA degradation. Tube stmetural integrity under accident conditions is assured via the proposed dimensional limits; indications with axial lengths less than 0: equal to 0.25 inches and circumferential lengths no greater than 0.57 inches are considered to have adequate margins to preclude burst under design basis tube differential pressure loading.
The proposed axial and circumferential extent repair limits are based on structural analyses of OTSG tubing per Reg. Guide 1.121. This analysis was previously submitted to the NRC by Florida Power Cwporation as part of Crystal River 3 Technical Specification Change Request No.198 (Reference 3) and has been reviewed by GPU Nuclear and verified as applicable to Thil-l's OTSGs as weli(Reference 4). Reg. Guide 1.121 requires that any potential through wall cracks / flaws must not propagate or burst under postulated accident conditions, including allowance for conservative margins of safety. Both the analyses and the proposed new limits for the Tech. Specs. (i.e. 0.25" and 0.57") are based on the conservative assumption that the flaws may be 100% through-wall The Crystal River-3 analysis was originally performed to assess OD-initiated flaws rather than ID-initiated flaws.110 wever, since the flaws are assumed to be 100%
through wall there is no analytical di'"erence between these two conditions. (The type of flaws
, 4710 97 2376
. Attachment Page 3 of 7 that were analyzed to determine these dimensions were axial" slots" whose sides were perpendicular to the axes of the tube; there were no aspect ratios or angled fracture faces assumed for the flaws.) GPU Nuclear has evaluated the Crystal River tube structuralintegrity analysis and
! confirmed that the analysis can be applied conservatively with respect to Thil 1 TSCR No. 268.
Thil 1 OTSG minimum tube wall yield strength is higher than that of the Crystal River tubing material by nearly 20%, which provides additional conservatism (Reference 5).
The proposed dimensional limits are based on the RG 1.121 assessment assuming linear (i.e.,
crack like) Caw geometries. htRPC probe inspection results ofID IGA indications have characterized the indications as volumetric rather than crack like. Potential tube mptures due to volumetric indications are evaluated considering perpendicular cracks, one reprunting the length and the other representing the width of the volumetric indication, rather than using a hole as the representation. This is consistent with established pressure vessel failure theory that unites failure appearance with the highest applied principal stress. Under burst conditions, as an example, the highest principal stress occurs in the circumferential direction, the second highest principal stress occurs in the axial direction, and the lowest principal stress occurs in the radial direction. The highest principal stress, i.e., circumferential, would exercise the axial extent of the volumetric defect to open first, and the second highest principal stress, i.e., axial, would exercise the circumferential extent of the volumetric defect to open next. A hole would not open because the radial principal stress is the lowest. The logic that the geometry of a potential rupture is determined by the magnitude of the applied principal stress is applicable to all service conditions.
Therefore, GPU Nuclear has completed analytical calculations providing the basis for the proposed dimensional limits that assume a linear (i.e., crack like) Daw geometry. This results in conservative dimensional repair limits with respect to the actual volumetric nature of the IGA degradation. GPU Nuclear has also demonstrated the ability to adequately assess the dimensional extent of tube degradation through measurements with htRPC probes (Reference 6).
The proposed axial and circumferential (L 5" axial extent and 0.57" circumferential extent) limits for flaws which require repair were halved in order to determine the proposed limits for flaws which would be considered degraded (i.e.,0.13" axia :xtent and 0.28" circumferential extent).
In addition, any ID IGA indication confirmed by h1RI" and having bobbin coil voltage 2 0.5 Volts would also be considered " degraded" even ifit did not exceed 0.13" axial extent or 0.28" circumferential extent. This is a conservative method by which to determine the " degraded" criteria for volumetric ID IGA indications where through wall depth sizing may be unreliable as a result oflow eddy current amplitude. In addition, this method of halving the repair limit to set a 2 " degraded" limit is consistent with the plant's TS where 40% through-wall and 20% through-wall are the repair and degraded limits, respectively.
GROWTil EVALUATION Thil-l's ID IGA flaws monitored in previously degraded tubes have exhibited essentially no growth over the operating cycles since the restart of the unit in 1985. This is expected behavior since the original damage mechanism was a " cold plant" damage mechanism in which the tubes were attacked while awaiting restart aller the Thil-2 accident. The sodium thiosulfate solution
^
, 4710 97 2376 Attachment Page 4 of 7 which entered the primary system and caused the attack has been removed from the plant's design and is no longer present, Changes in the plant's operation, primary water chemical treatments, and design were implemented in order to prevent further damage to the tubing from this I mechanism (Reference 2).
The bobbin coil voltage response to similar volumetric defects varies with overall size of an indication. Bobbin coil voltage response should increase as a defect becomes larger or, in the case ofIGA, as a loss of material or a weakening of the grain boundaries occurs. Thus, an increase in the bobbin voltage amplitude between successive inspections may be an indication that some growth occurred. GPU Nuclear's submittal to the NRC following the llR Outage (Reference 7) contains the results of a growth rate evaluation ofID IGA indications. The study compared the bobbin coil voltage amplitudes obtained during the 1993 and 1995 inspections for all tubes inspected in both outages. No significant changes in voltage, phase angle or noise component of the signals were noted. On this basis, GPU Nuclear concluded that there was no growth of the ID IGA indications in previously degraded tubes during Operating Cycle 10. Appendix A compares the changes in measured voltages between the last three inspections for a number of these ID IGA indications detected with the bobbin coil probe.
The growth analyses do not reveal a well-defined increase or decrease in the mean voltage for the population ofindications studied. This indicates that growth, if any, of the IGA defects is below the level of scatter in the voltage measurements due to NDE uncertainty. Since the potential growth for the next cycle would be expected to proceed at a rate similar to prior operational cycles, GPU Nuclear concludes that there is reasonable assurance that tubes leil in senice as a result of using the proposed repair criteria should maintain similar structural and leakage integrity margins through the end of the next operating cycle.
Based on the results of growth studies, GPU Nuclear has concluded that the degradation mechanism for the ID IGA defects in previously degraded tubes is inactive. Tubes with indications exhibiting a voltage growth of at least 0.6 Volts would be considered degraded and would be included in the classification of the inspection results (i.e., C-1, C-2, or C-3) per the proposed TS 4.19. Tubes exceeding the length criteria (2 0.57" circumferential extent or > 0.25" axial extent) will be removed from senice.
The proposed voltage growth threshold of 0.6 Volts was determined through a review of the outage-to-outage scatter in bobbin coil voltages for a large number ofindications with no increase in mean voltage. The voltage change between inspections for a majority of the indications varied within 10.6 Volts (i.e.,2 sigma deviation). The proposed 0.6 Volt " growth trigger" is derived based upon technicaljudgement and is intended to provide assurance that, should actual degradation growth occur between the inspections for certain indications, those indications will be identified by a voltage change above an appropriate threshold. It is noteworthy that GPU Nuclear normalizes its bobbin coil voltages to 10 volts on the four (4) 20% through wall holes of an ASME standard using the 400 kilz differential channel, so that a 0.6 Volt amplitude increase for TMl would correspond to a 0.24 Volt amplitude increase for the inspection program at many other plants (where the normalization of bobbin coil voltage on the four (4) 20% Through-wall holes of the ASME Standard is to 4 VoltsL)
4710 97 2376
. Attachment Page 5 of 7 The proposed revisions to the TS Section 4.19.5 require that GPU Nuclear evaluate the growth rate of the ID IGA indications, and notify the NRC of the results prior to the plant's attaining a primary coolant temperature of 250 'F. This revision requires that the status of this degradation be assessed in a timely manner so that the above "no growth" assumption is evaluated prior to full power operation.
IN-SITU PRESSURILTESI]FO GPU Nuclear has proposed in the Technical Specification Change Request that in situ pressure testing be utilized to confirm the ability of the ID IGA flaws to withstand the postulated accident-induced loads such that accident induced leakage will not occur. Tubes containing indications will be selected for in situ pressure testing during the 12R Outage. Tubes selected for testing will be determined based on: 1) those tubes that contain the indications with the highest bobbin coil voltage and the lowest axial and circumferential extent as measured by h1RPC inspection, and 2) those tubes with highest axial and/or circumferential extent as measured by h1RPC. Since bobbin coil voltage is related to degradation volume for similar defects, the greater the voltage, the higher the indication volume. By selecting tubes containing indications with maximum voltages and minimal axial and circumferential extents, GPU Nuclear will test those tubes which could contain the degradation with the most extensive through wall penetration. The ID IGA at Thil has also been shown to develop greater through-wall penetration as the (axial and) circumferential extents increase, so those indications with greater circumferential (and/or axial) extents will also provide a basis for selecting tubes for in situ pressure testing.
GPU Nuclear believes that the subject small ID IGA patches would not contribute to increased primary to-secondary leakage during a postulated hiain Steam Line Break (htSLB) accident. We intend to demonstrate the integrity of flawed tubes during an MSLB by performing in situ pressure tests of several flawed tubes during the 12R Outage. The proposed in-situ pressure testing will subject the individual tubes to pressures that will satisfy R.G.1.121 and represents a higher principal stress than that applied during a postulated MSLB accident. While the tests are performed at essentially room temperature when the plant is shut down, corrections to account for the changes in the inconel 600 tube propenies with temperature enable the tests to accurately-predict the capability of the tubes under the postulated accident pressures and temperatures.
Based on industry experience to date for the size ID IGA indications that are expected, no measurable leakage is expected in any of the in situ leakage tests. These tests are expected to demonstrate that tubes containing indications near the proposed repair limits will have adequate <
margins to ensure that accident induced leakage will not occur. The in-situ pressure test causes a J higher principal stress in the circumferential direction than that caused by the hiSLB load in the axial direction. The in situ pressure test creates a more severe condition than the postulated faulted condition with regard to the potential for causing tube leakage. It can be concluded that finding no measurable leakage during in-situ pressure testing from tubes containing ID IGA, selected as described above, shows that tubes with ID IGA will not leak during the postulated htSLB. As discussed above, the ID IGA indications are experiencing essentially no growth 1
, 6710 97 2376 Attachment Page 6 of 7 between inspections and therefore are expected to retain adequate margin against leakage until the next refueling outage.
Given the above, GPU Nuclear has included text in the new Hun of the Technical Specifications stating that " serviceability for accident leakage will be evaluated by successful 12R in situ pressure testing of a sample of these degraded tubes to evaluate their leakage potential."
incorporation of this statement ensures that, if the results of the in situ pressure testing do not demonstrate the serviceability of the tubing, repairs will be undertaken. In addition, GPU Nuclear has proposed that the results of the in situ pressure testing be communicated to the NRC prior to l the temperature of the primary coolant reaching 250 'F, when the OTSGs are required to be operable per TS 3.1.
DXERVIEW GPU Nuclear intends to perform a bobbin coil eddy current exam of 100% of Thil l's insenice steam generator tubes during Outage 12R. The bobbin coil probe will be used as the primary method for detecting tube indications. In addition, GPU Nuclear has committed in TSCR No. 268 to use the h1RPC probe for the assessment of morphology of all ID IGA indications which may be dispositioned under the new repair criteria and degraded tube criteria. The proposed new definition ofID IGA indications at Th11 1 in Section 4.19.4, item 9, requires the indication / flaw to be "... confirmed by diagnostic ECT to have a solumetric morphology characteristic of(ID) IGA." This is a restrictive definition that requires that the indications be examined by diagnostic ECT (i.e., h1RPC). Thus, the proposed dispositioning criteria are not based on a single eddy current probe or a single parameter. In addition to the MRPC probe dimensional (i.e., length and width) measurements, supplemental in-situ pressure testirg is proposed to assure that tubes left in service as a result of applying the repair criteria have adequate structural and leakage integrity for continued senice through the end of 0perating Cycle 12. To provide defense-in depth assurance ofleakage integrity, the allowable primary-to-secondary leakage of 0.1 gpm through any one steam generator above baseline (currently 0 gpm) has already been established at TMl-1 by TS Amendment 103 license condition. Th11-1 has two independent condenser exhaust radiation monitors to provide for an early indication of tube leakage and timely unit shutdown in the event of significant leakage.
In summary, the revisions to Technical Specification Section 4.19 proposed by TSCR No. 268 are for one cycle only and provide reasonable assurance that defective OTSG tubes will be either repaired or removed from senice. The proposed changes are more restrictive in several respects (e g., reporting requirements, required use of diagnostic h1RPC, limitation of serviceable length and width extents) and are updates of the TS to proside for the use of newer NDE technology and in-situ pressure tests as part of the inspections. The ID IGA degradation that is unique to the Thil 1 sulfur intrusion event in 1981, and subsequent repairs, have previously been reviewed and found acceptable by the NRC in their Safety Evaluation Report that approved TM1-1 Technical Specification Amendment No.103.
g 6710 97 2376
. Attachment i Page 7 of 7 REFER 13NCES
- 1. GPU Nuclear letter, Langenbach to NRC Document Control Desk," Technical Specification Change Request (TSCR) No. 268," August 12,1997.
- 2. NRC Letter, Stolz to liukill,"TMI l Technical Specifications Amendment No.103 " dated December 21,1984.
- 3. Florida Power Corporation Letter, Heard to NRC Document Control Desk," Technical Specification Change Request (TSCR) No.198," Appendix A,"MPR Structural Analysis,"
dated March 4,1994.
- 4. GPU Nuclear Memorandum No. E520 97-024, Leshnoff to File,"Use of CR 3 Burst Strength Calculation for Damaged OTSG Tubes for TM1 License Change for BVC Indications," dated August 7,1997.
BAW-1760, Revision 1, (GPU Nuclear TR 007, RcWinn 11, dated March 1983.
- 6. Framatome Technologies Inc. (FTI), Document identifici 51 5000345-00, "PWSCC and Primary Side IGA Sizing Performance of OTSG Rotating Coil Examinations," Revision 0, dated July 2,1997.
- 7. GPU Nuclear letter, Knubel to NRC Document Control Desk,"TMI l Cycle 11 Refueling (llR) Outage OTSG Tube Inspection Report," dated August 1,1996.
I
. 6710 97 3376 Appendix A Page 1 of 6 TMI-l OTSG ID IGA Growth (Outages 9R,10R, andllR)
ID IGA at TMI l is not growing based on the results of a statistical evaluation of voltage variations between Outages 9R and 10R and between Outages 10R and llR. Regression analysis of voltages for ID IGA indications from 9R through IIR shows that, in general, voltage difTerences from one outage to the next are not significant. This is because voltages are related linearly with the slope of the statistical model very nearly equal to one and with a corresponding intercept equal to zero in each case. Observations that are related statistically by a slope of unity and with an intercept equal to zero are usually considered to be identical. At twice the root mean squared error (RMSE) the range o.
variation in voltage does not exceed about 0.6 V so that the' change in voltage does not exceed the magnitude of the voltage responses that are being compared. A range of variation using twice the RMSE corresponds to about two standard deviations at any particular voltage. The results of the statistical model provide support for the conclusion that no physical growth of the ID IGA indications has occurred between outages.
Figures I through 4 below provide scatter plots for each steam generator comparing voltage measurements of one outage to the subsequent outage and as can be seen in the plots in each case the linear trend line intercepts at zero.
O T S O A O u ta g e 9 R V e rs u s 10 R ID IG A O ro w th f o r B o b b in C o il P ro b e 6
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6710-97 3376 Appendix A Page ? of 6 O T S G A O u ta g e 10 R V ers u s 11 R ID IG A G ro w th f o r B o b bin C oll P ro b e 6
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I OTSG B O u ta g e D R V e rs u s 10 R ID IG A G row th f o r B o b bin C oil P ro b e 2.5
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, , 6710 97 2376
,- Appendix A Page 3 of 6 OTSG B O utage 10R Versus 11R ID IG A G rowth for Bobbin Coll Probe 3
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. . 6710-97 2376
. Appendix A Page 4 of 6 4
Figures 5 through 8 provide voltage variation histograms ofID IGA recorded during Outages 9R,10R, and llR as an ahsuate view of the data. These histograms also substantiate no growth conclusions. The histograms indicate a peak in the "no-growth" or zero voltage change column as would be expected when no growth is occurring.
OTSG A Outage 9R Versus 10R ID IG A Growth for Bobbin Coil Probe j :::
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6710 97 4376 Appendix A Page 5 of 6 OTSG B Outage 9R Versus 10R ID IG A Growth for Bobbin Coil Probe
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Bobbin Cell Voltage Verlation Figure 7 - Histogram of Voltage Variation oflD IGA Indications Recorded in OTSG-B During Both Outages 9R and 10R t
OTSG B Outage 10R Versus 11R ID IGA Growth for Bobbin Coil l' robe l
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During Both Outages 10R and 11R i
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/ , 6710 97 2376 Appendix A Page 6 of 6 ID IGA indications where both percent through wall and amplitude have been assigned -
during two consecutive outages have been statistically evaluated and indicate no growth trend for either through wall dimension or amplitude (voltage). The mean growth values have not exceeded the apparent " negative growth" seen over the seven outage-to-outage comparisons, indicating that the changes seen represent data scatter from normal eddy current examination technique error and do not represent flaw growth. See Table 1 below:
Table i COMPARISON OF STATISTICAL EVALUATION TO PREVIOUS EVALUATIONS Period Number of Mean STD Mean STD (Year-Outage No.) Indications Change Deviation Change Deviation
% T.W. % T.W. Volts Volts 1984/1986-5M 152 -2.6 6.1 -0.2 0.3 1986-5M/1986-6R 118 + 1. I 6.6 +0.0 0.2 l 1986-6R/1988-7R 119 +2.6 5.5 +0.2 0.3 1988-7R/1990-8R 291 -0.2 7.43 -0.25 0.35 -
1990 8R/1991-9R 229 -2.0 6.96 +0.07 0.31 1991-9R/1993-10R 207 -0.6 6.62 +0.16 0.28 1993-10R/1995-llR 197 +0.9 6.39 -0.26 0.40 Data extracted from GPUN Letter 6710-96-2264 submitted to the NRC on August 1, 1996, Reporting the Outage 11R OTSG Eddy Current Examination Results.
NOTE For Outages 1984 - SR through 1988 -7R, all indications were examined with both the 8xl and the .540" bobbin coil probes and only indications
> 20% T.W. and confirmed by both probes were included. This criteria conservatively biased the data by eliminating any indications which showed a decrease in % T.W., thereby dropping them below 20% T.W.,
or were afTected by the variability of the 8xt-ABS probe. For 1990-8R through 1995-10R, all indications > 16% T.W. were included, which allows for the full variability of both the T. W. phase angle analysis and voltage recording processes and accounts for the increased number of indications.
e JM a