ML20080H663
| ML20080H663 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 02/28/1995 |
| From: | ENTERGY OPERATIONS, INC. |
| To: | |
| Shared Package | |
| ML20080H660 | List: |
| References | |
| NUDOCS 9502230211 | |
| Download: ML20080H663 (89) | |
Text
Awarhment to 2CAN029505 Page1of89 ENTERGY ARKANSAS NUCLEAR ONE l
UNIT TWO
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I STEAM GENERATOR CIRCUMFERENTIAL CRACKING EVALUATION February 1995
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~ Page 2 of 89 Table ofContents 1.0 Introduction l 2.0 Summary
. 3.0 ANO-2 Steam Generator Description 4.0 Steam GeneratorlaW~dRepair Estory l 4.1 OperatonalEstory 4.2 Eddy Current / Repair Summary 5.0 Technical Evaluation 5.1 Inspection Techniques / Improvements 5.2 Statistical Analysis 5.2.1 Crack SizeIndication Distnbution 5.2.2 Projection ofNumber of Cracks 5.2.3 Statistical Projection of Degradation Expected at the End of Cycle 11 5.3 StructursJ Analysis 5.3.1 Regulatory Guide 1.121 Evaluaton 5.3.2 Finite Element Model ;
5.3.3 In-Situ Pressure Testing 5.4 Leak Rate Analysis 5.5 Crack Growth Rate 5.6 Laboratory Crack Project 5.6.1 Nondestructive Exammation 54.2 Residual Stress Measurements '
5.6.3 Metallurgical Analysis 5.6.4 Leak /BurstTesting 6.0 Safety Assessment l l
6.1 Safety Analysts !
6.2 Probabilistic Safety Analysis 6.2.1 FLB/SLB Analysis 6.2.2 ATWS Analysis 6.2.3 PSA Analysis Results i
7.0 Operational Response
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1 7.1 Inservice Leakage Detection / Response 8.0 Conclurions 9.0 References d Appendix A l
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1.0 - IhTRODUCTION !
Arkansas Nuclear One, Unit Two (ANO-2), first experienced circumferential cracking in March 1992. Cracking was discovered as a result of primary-to-secondary leakage from a ;
crack in the "A" steam generator (SG). "Ihe first inspection for circumferential cracking was !
performed during this forced outage Since that time, four additional comprehensive i exanunanons for circumferential crmiLg (during two refueling and two planned mspection outages) have been conducted. In order to assess the safety implicanons of continued operanon ,
of the unit, a comprehensive evalumnon was performed to assess the impact of the circumferential cracks on the structural integrity of the SG tubing. 'Ihis evaluation, which is a ,
compilation of several related engineering calculanons, follows.
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SUMMARY
l He ANO-2 steam generators have experienced several forms of tubing degradation, most !
notably stress cormsion cracking (SCC) at the tube support plates (axially oriented) and at the expansion transition region (ETR) at the top of the hot leg (HL) tubesheet in the sludge pile i' region. ne indications at the tubesheet have primardy been circumferentially oriented, but have also included axially oriented, as well as volumetric indications. His evaluation is limited ;
to the circumferential cracks at the ETR, since these have been the only flac detected in the SGs which could potentially challenge struct' rial margin requirements.
As a result of comprehensive inspections, application of NRC Regulatory Guide (RG) 1.121 ]
safety factors, use of statistically valid (95/95) material properties, statistically based !
degradation projections, and tube burst test data, it is concluded that the ANO-2 steam generators can safely operate for the remainder of the current cycle with an acceptable level of .
risk of compromising the structural integrity requirements for the tubing. From a risk '
perspective, full cycle operation is also assured at an acceptable level. Le next cycle's operation will be evaluated agam following data collection from the September 1995 refueling outage (2R11). Details of the evaluations performed are described in the remauung sections of ,
this report.
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Page 5 of 89 i 3.0 ANO-2 STEAM GENERATOR DESCRFI1ON i The ANO-2 steam generators are of the U-tube design manufactured by Combuston !
Engir=ams (model 2815). Each steam generator contains 8411 tubes constructed of high r temperature mill annealed (HTMA) Inconel alloy 600 material with an outside di=aw of 3/4 l mehes and a wall thickness of 0.048 inches h tubes are explosively expanded to the full !
depth of the tube sheet. There are seven full eggerate tube support plates (TSP), two partial i eggerate TSPs, and two partial drilled TSPs. L SG layout is shown in Figure 3.01. The I reactor went into commercial operaten in March 1980, and utihzes all volatile treatment '
(AVT) #=. :-Ty. Secondary side boric acid addition was initiated in 1983 to arrest denting at :
the partial drilled TSPs. The hot leg operstmg temperature was initially 607* F, but was }
reduced to ~400* F following the ninth rd= ling outage in the fall of 1992.
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/ 4.0 STEAM GENERATOR INSPECTION / REPAIR HISTORY 4.1 OperanonalHistory he ANO-2 SGs began commercial operation with 44 tubes plugged (15 in "A" and 29 in "B").; nose preservice repairs were done primarily as a result of s," post rim cut" of the dnlled TSPs around the periphery to prevent crackmg of the solid hgaments i between the dnlied holes of the TSPs due to denting. An inspecnon of the penphery was performed to ensure that any tubes 'ci danng the moddicanon were repaired.
ANO-2 historically performed the todmical spacirnei,wi mmimum inspection sample size of 3% per SG through the eighth refinehng outage Early on, denting of the tubes at the parnal drilled TSPs (TSPs 10 and 11) was of concem. Changes to the secondary plant, including removal of the CuNi tubing in the feedwater heaters,. oxygen control
, modifications to the credaa==*a storage tanks, boric acid treatment, and adoption of the Electric Power research Institute (EPRI) Secondary Water (%a=3=+ y Guidehnes, arrested the denting rate and all but -L=i==*~i that concem. Dese changes /er a an==aats were completed by the end of the third refueling outage (1983).
The next area ofinterest that dew,' sped was Sow induced vibration in the form of wear at the upper bt .Je vertical and diagonal support straps (s--osely referred to as "batwmss"). In the sixth refuehng outage (1988), expanded tesung was performed in the "B" SG due to wear induced damage. De initial sample plan was 8% in "A" and 5% in "B." In accordance with the technical speci6 cations, a second inspection sample i (6%) for a total of 11% was pafurw.ed in the "B" SG. Six tubes were plugged as a l result. He "A" SG had no repairable indications in the first sample, and thus no '
expansion was required.
In August 1990, the Erst indication of primary-secondary leakage occurred. Evidence ofleakage was detected following a full power reactor trip and decreased upon retum to full power to a low level (<0.01 GPM).
In 1991, during the eighth refueling outage, the initial inspection of the "B" SG (3%)
was categorized as C-3 (>1% of the sample defective per the technical specifications) due to TSP indications, and thus a 100% inspection was required. Additionally, a second sample inspection was required in the "A" SG as a result of the C-3 categorization in the "B" SG. The 9% total sample of"A" did not yield defective tubes and no further ia=padons were performed. An attempt was made in 2RS to locate the leaker with a pressure / helium test, but no indication ofleakage was found.
In October 1991, following a full power reactor trip, indicanons of primary-=~~dary leakage again occurred. Following a retum to full power, the leakage gradually decreased to a very small level (<0.01 GPM).
On March 9,1992, primary-secondary leakage took a step increase and the plant was j subsequently shut down. De calculated maximum leak rate was 0.25 GPM. During l
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the outage, the leak was desermined to be from a circumferennal crack at the top of L u tubesheet (ITS) ETR on the hot leg of the "A" steam generator. Rotanng pancake coil ;
(RPC) inspections were performed on 100% of the hot leg ETR of both SGs, and 20% l of the sludge pile region in the cold leg (CL). In addition, all tubes with potential !
indications were tested full length with a bobbin coil probe. In total,421 tubes were :
repaired in "A" and 67 in "B". Of those numbers, 392 tubes in "A" and 56 in B" i were repaired via i.d.g. h remainder of the tubes were removed from service i with plugs and stabihzers. Subsequent detailed evaluations indicated that only 208 tubes in "A" and 11 in "B" had circumferential cracks, with the remainder beirig ;
' volumetric or axial indications h results of the outage were presented to the NRC ;}
staffin a meeting on April 16,1992m, j f
Five tubes were removed during this outage. Dree contained circumferential cracks ,
and two contained axial indications at three support elevations h results of the tube i pull and additional analysis were presuited to the NRC staffin a meeting on August i 26,' 1992m, The unit operates formwing the forced outage for approximately four months until the .
ninth refuehng outage in. September 1992 (2R9). During that outage, 25 l circumferential em:ks (17 in "A" and 8 in "B") were detected as a result of an i isg+1,a of 100% of the hot :eg ETR and 20% of the cold leg sludge pile region in ;
each SG.. He largest flaw was pressure tested prior to removal. ' He test ' i demonstrated the flaw was able to withstand three times the normal operatirig l differential pressure (3AP). Two tubes with circumferential cracks were removed j during the outage. h inrea and pulled tube results were presented to the NRC ;
staffin a meeting on December 3,1992W. l l
As a result of concem for possible high growth rate and potential for leakage, a mid- .
cycle outage (2P93-1) was planned. h outage was conducted after approximately ,
six months of operation from the refueling outage. Dunng that outage, inspecuans of !
both SGs were performed g = e ,1 in the sludge pile region (100% of sludge pile ;
area, ~70% of the total bundle). Forty-five circumferential cracks were found in the "A" SG and three in the "B" SG. In addition, two volumetric indications were ;
discovered in the "A" SG. All indications were plugged and stabilized. h results of ;
the outage were presented to the NRC staffin a meetmg on August 30,1993W. .j i !
ANO-2 operated until the next refuehng outage (2R10, beginrung in March 1994) with i
no detectable leakage. During this outage,170 circumferential cracks were discovered (147 in "A" and 23 in "B") as a result of an ia@ of 100% of the hot leg ETR and 20% of the cold leg sludge pile region in each SG. . In-situ pressure . testing was .
performed on three tubes. A detailed discussion of the pressure testing is pmvided in !
Section 5.3.3. Subsequent evaluations indicated that all tubes met RG 1.121 structural margin requirements, and probabilistic analysis showed that full cycle operation was ,
acceptable (see Section 6). However, in order to provide additional data, a subsequent ;
mid-cycle outage was scheduled. h results of the 2R10 outage were presented to the NRC staffin a meeting on July 14,1994W. l l
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ANO-2 operated from 2R10 until the mid-cycle outage in January 1995 (2P95-1), with ;
no detectable leakage. h mid-cycle outage was performed with a planned scope of .
RPC testag at the ETR of the entire hot leg sludge pile region of both steam generators h results were 283 circumferential cracks (203 in "A" and 80 in "B")..
In addition,17 other flaws (12 in "A" and 5 in "B") were detected. These flaws were .
generally axially oriented or volumetnc in nature All tubes c- ==---.g detected flaws ,
were plugged and those tubes containing circumferential cracks were also stabilized. i A summary hstmg of the 2P95-1 results is included in Appendix A. The hstmg contains all tubes plugged during the outage and satisfies the ANO-2 technical specificaten reporting requirement.
W largest flaw detected was sized at 304* arc length, 81% maximum depth, and ;
calculated average depth of 69%. In-situ pressure testmg was performed ca this tube and two additional tubes A detailed discussion of the pressure testmg is provided in Section 5.3.3. Based on the RPC data and the in-situ pressure testing, all flaws met the RG 1.121 structural requirements. Additional probabilistic and statistical analysis show that operation for the remainder of the cycle is acceptable (see Sections 5.2 and 6).
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Aaar hma=* to i . 2CAN029505' Page 9 of 89 4.2 Eddy Current / Repair Summary The followmg tables summarues the ANO 2 inspectum history since 1991 (Table 4.2-1) and the rejair summary since pre @ (Table 4.2-2) for all defects. l Table 4.2-1 ANO-2 Inspectum History ,
"A" Steam Generator % Tubes IneM i
Tvoe ofInsoection 2R8 2F92 2R9 2P93-1 2R10 2P95-1 HL TTS RPC(ETR) 0 100 100 71* 100 66*
CLTFS RPC (ETR) 0 20** 20 " 0 20** O Bobbin FullLength 9 5 100 0 100 0 DentInspection (RPC) 0 0 0 0 8 0 i Sleeve Inspection N/A 100 0 0 100 0 ;
"B" Steam Generator % Tubes InwM Tvoc ofInsoection 2R8 2F92 2R9 2P93-1 2R10 2P95-1 HL TFS RPC (ETR) 0 100 100 72* 100 71*
CL TTS RPC (ETR) 0 0 20 " 0 20 " O Bobbin Full Length 100 1 100 0 100 0 DentInspection(RPC) 0 0 0 0 33 0 SleeveInspection N/A 100 0 0 100 0 l
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Percentage ofnon-plugged and non-sleeved tubes in the entire bundle. 100% of the sludge pile region was inspected.
" Percentage of the sludge pile region i
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Table 4.2-2 ;
Reper Summary ,
SG "A" SG "B" YEAR EFPY OUTAGE PLUGS RTFFVES PLUGS SLEEVES 1978 0.00 MFG 15 0 29 0 1981 0.89 2R1 0 0 0 0 1982 1.69 2R2 0 0 1 0 1983 2.33 2R3 0 0 0 0 1985 3.31 2R4 0 0 0 0 1986 4.16 2RS 0 0 0 0.
1988 5.38 2R6 0 0 6 0 t 1980 6.52 2R7 0 0 0 0 .
1991 7.67 2R8 0 0 73 0 1992 8.51 2F92 29 392 11 56 l 1992 8.85 2R9 67* 0* 132 0 ;
1993 9.36 2P93-1 47 0 3 0 1994 10.16 2R10 170 0 77 0 1995 10.86 2P95-1 215 0 85 0 Total # Repaired: 543 388* 417 56 ,
Fo tu p evi ly leeved at the tubesh were plugged due to flaws at tube support plates. The total number ofinservice sleeves in the "A" SG is 388.
Note: Eighteen sleeves are equivalent to one plug in reduction of reactor coolant system (RCS) flow. l
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-r 5.0 'IECHNICAL EVALUATION 1 5.1 InspectionTechniquashpe .
i Nondestmenve Examination (NDE) is a critical element in assessing the structural l signi6cance of SG tubing flaws, and the inspartian method must provide a rehable measurement of crack dunensions. Additinnally, the inspection process can provide crack growth data which is rmired to estabhsh alkmable crack sizes and operating intervals forvarious conditions ,
Conventenal bobbin coil technology, used for routine steam generator examinations, has limitations in inspecting expansion transitans, and circumferentially oriantad flaws -
are essentially urhhla with the bobbin coil. h use of RPC eddy current technology provides an important field diagnostic tool for detectmg and measuring .
circumferential cracks.
l RPC tube inspection is accomphshed using a surface-riding coil which is rotated I around the tube axis. b coil is spring loaded to maintain contact with the tube inner i surface as it moves through the expansion transinon region. Lift-off variations between ,
the coil and the tube surface, caused by a change in tube diameter due to the expansion, ,
are signi6cantly reduced when compared with the bobbin coil. As the probe is j translated and rotated through the tube it describes a helical path. A hnear ;
discontinuity is scanned once during each rotation of the probe. 'Ihe coil output voltage l from a given rotation is used to generate a line scan which represents signal amplitude -l as a function of coil position around the tube -w.J.-s'O. Multiple line scans are '
typically displayed in a three dimensional format (known as C-Scan) to provide a tool for evaluating potential flaws. :
i Entergy Operations has previously utihzed an approach to predict the average l circumferential crack depth as the product of the maximum depth from the RPC phase angle analysis and the circumferential crack length divided by 360*: ;
d = d***/ i 360 i i
where: dmax = crack maximum depth (based on phase analysis), % throughwall ('IW) 1 = crack are length (largest of clip plot or crack map), degrees ,
The intent of the average depth approach is to define the flawed area in order to assess ;
, structural integrity. Information related to this issue was presented to the NRC staffin !'
a meeting on July 15,1993M. .
l In Belgium, a similar approach is used where the " crack area" is calculated by using the percent depth times the length in degrees as determmed by ultrasonic testing *. ;
Average depth or crack area is the desirable product onthe NDE because of its ;
correlation to the structural adequacy of a cir:cumferentiality cracked tube. In the !
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circumferential cracks can be catated by equating the axial load due to pressure (Ib,) ' !
to the not section remaining tube area (in2) times a suitable flow strength (psi). :
i Figures 5.1-1 and 5.1-2 illustrate that the observed top of tubesheet circumferential . !
crsGg at ANO-2 is oAen comprised of a band of circumferential cracks Small axial i cracks are sometimes observed, with a resultmg crack network morphology that has' been termed cellular corrosion. De sound maserial which resists axial pressure forces ;j is composed of uncracked material from the landmg edge.of partial throughwall j '
circumferential cracks at different axial lacanians (see Figure 5.1-3). Machined circumferential flaws only simulate this fust type of hamment. %e ligaments of ~
material between circumferential cracks at different axial lacatiam in a bed of ;
circumferential cracks are not present in machined flaw geometrier, which have been <
tested to data?).
i Both average depth and crack area methods have shown reasonable agreement with the f
observations on pulled tubes; however, they appear to provide conservative assessments of burst strength of circumferential cracks as demonstrated by the use of l in-situ pressure tests at ANO-2 and other sites. Assessmos of the structural :
significance of circumferential cracks can be performed usmg we length alone, as long !
as the length is short enough to support the necessary assumptus. When the crack ;
lengths are longer, some other parameter must be factored in, ar.d fhus the average !
depth concept was created However, the success of the averag6 RPC depth hinges on the balance of overestimates and underestimates. He RPC maxacum depth is ,
typically an overestimate of the actual crack depth and the datactM lengdi is typically :
an underestimate of the actual crack length. It is expected that this procedure would be 1 increasingly overconservative for very long cracks, since a detected length of about l 360* cannot be an underestunate However, the data available to date for crack lengths l
near 360* show that the RPC calculated average depth and the measured averase depth ;
are about equal. He present agreement between average crack depth and estimates ;
from RPC may be fortuitous at very long crack lengths For cracks of 360*, the j cracked area estimate, and thus the burst strength estimate, is critically dependent on !
estimated maximum depth from RPC data. Sizing of nearly throughwall cracks must t be considered a region of considerable uncertainty. While the RPC ' average depth approach is expected to be reasonably successful when cracking is not detected over !
the full tube circumference, this technique is subject to being overly conservative for !
very long and deep cracks, mostly due to the planar flaw presentation versus the flaw !
geometry contauung ligaments described above ne RPC average depth concept is but one technique utilized at ANO to evaluate the structural signi6cance of a given ,
flaw. It is used to identify those flaws needing further evaluation via other means such l as in-situ pressure testmg, ultrasonic testmg (UT), or RPC deconvolution analysis to i evaluate the flawm. !
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Because of previous problems with data quality, three maior hardware changes were j incorporated for the 2R10 inspection to improve the data and crihance the detection and !
sizing of flaws. He Srst change was the use of a 0.115" diameter pancake coil, and an i axial and circumferential coil. Previously, the pancake coil was 0.080". %e 3-coil ;
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RPC 0.080" shielded coil, which was da-i, ped for the detection and sizing ofinside {
diameter (ID) initiated creJaug, provides hmited detecten of low level outside !
diameter (OD) initiated creJagg. His is, in part, due to the relatively high optimum j frequency which results in the reduced density of eddy currents at the OD of the tube wall and because shaHow OD flaws, coupled with the decreased current flow at the i OD, cause nunor perturbations of the eddy current field; these concoquandy limit !
detection of shallow OD flaws. W use of the 0.115" coil, coupled with overall system l optin6meiari has allowed for better d**=i -: ;4 * ). De second change was the use of a low loss cable from the probe pusher to the MIZ-18 tester to provide a i better signal to noise ratio. He third change involved the use of d~hr=*~1 power l supplies for each SG. None of these changes can be directly attributed to dannitive ,
improvements, although subjective evalustens have been performed indicabrig that the ;
1994 data quality appeared to be siysL4 better than that acquired in 1993(").
One area related to analysis that was changed during the 2R10 outage involved the use ;
of C-scan terrain plots to assess the data. He Eddy Current Tesang (ECT) Guidehnes l did not previously require analysts to utihze the C-scan plots, even though it was !
common practice. A concem is that C-scan plots can produce flaw-like signals in good j tubes (i.e., simdar signals exist in steam generators never placed in service). l nerefore, both C-scan and strip chart (hsajous) analysis must be performed in order to !
provide the best evaluation of the data. Analysts had previously relied on hsajous ;
phase correlation between the base frequencies to evaluate signals. However, in many cases, the phase rotates completely out of a flaw plane. One would normally expect the ,
phase spread between the base frequencies to be relatively narrow. For Tube 48-50 in ;
2R10, the spread was significant (from 62 degrees at 400 kHz to 353 degrees at 100 l
. kHz). In addition, for that particular tube, a multi-frequency mix removes most of the ;
signal response such that there is little to no evidence of a flaw being present. His is j shown in Figure 5.1-4. Because of the above concems with individual analysis tools, :
multiple analysis tools must be used to ensure that a thorough analysis of the data is -
performed. j l
For 2P95-1, the following changes were made
. Data was acquired while pushmg vs. pulhng the probe to eliminate probe snap through the transition
. Rotational speed was increased from 300 RPM to 900 RPM e Push speed was increased from 0.2 inches per second to 0.5 inches per second
. Terrain maps were required to be plotted for every tube.
These changes were made based on previous ANO and industry experience and yielded the best quality data to date (qualitative assessment). All of the above changes are enhancements Entergy Operations believes results in a comprehensive and conservative inspection.
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For Tube 32-126, whidi was the largest flaw at 360* in 2R10, the prenous outage !
results were classified as no heaceahle defect (NDD). De data does, however, l pmvide some maight as to the formation of a flaw. %e hsgious data for 2F92 and 2R9 (shown in Figure 5.1-5) at the base frequency with the pancake coil shows a flat response ne C-scan data does in&cate a " ridge-hke" response, but this is potentially a result of the effects of the expansion transiten, moductive deposits, and the top of tubesheet (see Figme 5.14). In 2P93-1, whde the hsgious by itself does not provide '
any indication of a flaw, it does show a change from the previous outage (see Figme 5.1-5). Whde this type of response could still be due to e--le-se deposits, it would be a cause for fwther evaluation. When conductive deposits exist in coriunction with - 'I the explosive expansen and top of tubesheet geometric &seortions, comparison to j previous data, even if the ECT response is not clearly in&cative of a flaw, may be the best way to evaluate the data. Entergy Operatens currendy compares all tubes flagged l with potential in& cations with the previous outage results j i
A signal processing operation referred to as " deconvolution" has been used to obtain j improved angular resolution of ANO-2 RPC data obtamed from circumferential cracks I at the top of the tubesheet. While deconvolution is not a new concept, its applicaten to rotating probe eddy current data represents a first time applicaten. Deconvolution provides a more detailed and accurate assessment of multiple closely spaced eddy current indications which, when coupled with leak rate data, can be used to provide a better estimate of tube integrity in the context ofleakage and burst susceptibility. De deconvoluted eddy current data has also been compared with in-situ UT data with excellent agreement M. Figures 5.1-7 through 5.1-9 show the deconvoluted ECT data as compared to the UT.
ANO-2 has tested a total of 151 tubes with ultrasonic techniques and has received mixed results. Le exams were performed in the 2F92,2R10, and 2P95-1 outages.
During 2F92, 85 tubes were tested De largest flaw was 360*,100%TW maximum depth, and had an average depth of 67%, which is less than the RG 1.121 maximum of 79%.
UT was performed on 52 tubes with potential indications during 2R10. Of those,13 of the 52 had no indicatmn present. %e largest flaw by Ur was 360',100%TW maximum depth, and had an average depth of 76%, which is less than the RG 1.121 maximum of 79%, thus indicatmg the flaw would have sustained the required pressures with the margin included. A UT plot of the largest flaw in 2R10 is shown in Figure 5.1-10.
During 2P95-1,14 tubes with indicatens as called by RPC were examined with UT.
De results of the UT provided consistently poor correlation with the RPC results. De largest flaw by UT was 45',46%TW maximum depth, and an average depth of 3%.
He same flaw by RPC was sized at 304', 81%TW maximum depth and an average depth of 69%.
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Attachment to I 2CAN029505 I Page 17 of 89 i l I Depth Direction in- ' Plane Ligament
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Attachment to 2CAN029505 3 Page 22 of 89
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1 Attachment to l 2CAN029505 Page 23 of 89
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1: l Attachment to 2CAN029505 Page 24 of 89 i TUBE Haz2112e ARKANSAS UNIT 2 STEAM GENERATOR *A* HOT led 50" 180* 270' RELATIVE CRACK POSmON SKETCH BOLD CRACK $ .100 % DEEP o- 2.o-l 1 525mm 524 nun v i 523mm __.1 _. 9 8 - 10 _ 522mm
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Anachment 2
- 2CAN029505.
; Page 25 of 89 -
5.2 Statmucal Analysis i In order to adequately assess the margin of safety provided by the steam generator j tubes, it is necessay to be able to conservatively predict the pmbabahty of the existence ! of cracks the we of sufEcient size to challenge the tube integrity during design basis events This can be ==;"I by statisacally analyzing data fmm past inspection campaigns and industry data where applicable.. Sp+4'="y, for ANO-2, this is ! accomphshed by combining the most pmbable distribution of crack sizes and the l expected number of cracks based upon plant operanon time. l
?
He use of statistical evalushons to support safe operation of ANO-2 was initially. j applied following the forced outage in the spring of 1992. Since this was the first ; inspection conducted at ANO-2 speci6cally to detect circumferential flaws at the ETR, l the initial data base was limited. His data base was supplemented by. using l Millstone-2 (MP-2) plant data. MP-2 is a similar Combustion F=p -xiig (CE) I designed plant that had expenenced circumferential cracking prior to ANO-2. - MP-2 characterized the average %TW degradation as exhibiting the form of a gamma
]
distribution (similar to Figure 5.2-1). De ANO-2 data demonstrated this same '! statistical distribution. Because of the limited plant speci6c data, ANO-2's data was ! combined with that of MP-2 to provide a reasonable assessment of the distribution of the flaws vad to be present at the end of the operating cycle (approximately four-additional months of operation) to predict the probability of a flaw in the SGs which . ; would exceed the RG 1.121 acceptance criteria. His analysis sic;;d that ANO-2 was safe to operate the remainder of that cycle with a greater than 95% probability that no tube would have a flaw in excess of the acceptance criteria of RG 1.121, l This same basic statistical approach has beea applied following each subsequent SG inspection. Entergy Operations has now performed Sve major inspections for
'{
circumferential cracks at ANO-2. Because of the expanded site spec 6c data base, j recent predictions have been based on ANO-2 speci6c data. - As'the data base l increases, statistical confidence increases. Sections 5.2.1 and 5.2.2 describe this : statistical approach.' Section 6 combines this information with a plant specific failure mode analysis to assess core melt probabilities. t 5.2.1 Crack SizeIndication Distribution 1 This section provides a review of the circumferential crack data for ANO-2. ! Appendix A contains summary data tables of all cracks identi6ed in the five l i inspections Random variations in steam generator tube =~ hanical properties, local stresses, chemistry conditions, and individual crack growth rates' result in - l variations in the circumferential length and depth of these cracks. ~ ne frequency of occurrence of specific crack lengths and depths can be' i characterized by statistical probability distributions nese distributions are not ! well characterized by the " standard" Ga==an probability distribution. His fact is consistent with the experience at other sites where SG circumferential i cracking has occurred U. Standard non-Gaussian probability distribunons can l l l r I l i
g s 2CAN029505 Page 26 of 89 be St to the data which provide consistent and statistically valid :=rlais of the i variabihty of circumfamntial crack lerigth, maximum depth, and average depth. !
- A' statistical procedure known as a goodness-of-St test is commonly used to l
- test whether- a psi.es set 'of measurements can be concluded to' be. l represented by a speedied probabihty distnbution. In these methods,' a " null- i hypd 4." speci6es an assumed probabihty detribution function. Various ;
methods can then be'used to assess whether the empirical distribution is l consistent with the hypothesized distnbution. If the tests indicate that there is i . not good greement, then the hypothesized &stnbution is abandoned in favor of -
, an ahomativdu)
- l Three candidate famihes of probability distributions were considered for representmg the distribution beta, samma, and extreme value. De resuhs indicate that the gamma probability distribution has the best goodness-o?-fit for :
crack length and average depth. Figures 5.2-1 and 5.2-2 show the distribution of all flaws for average depth and are length, respectively. Figures 5.2-3 and - ! 5.2-4 show the distribution of flaws for the 2R10 outage for average depth and arc length, respectively. Figures 5.2-5 and 5.2-6 show the distribution of flaws - for the 2P95-1 outage for average depth and arc length, respectively. 5.2.2 Projection ofNumber ofCracks ; I For most of the degradation mechanisms, it was assumed that the time ! dependence of the best estunate number of tubes requiring repair as the result - , of that degradation mechanism is described by a Weibull probability ; distribution. ne Weibull distribution is pediaps the most widely used lifetime . distribution modelM. De basic Weibull distribution for failure tunes is: l F = 1 - exp[-(TAB)*] ! where F is the fraction of the tube population that has failed at time t. l Characteristic time (0) and Weibull slope (b) are adjustable parameters of the Weibull distribution. Dese parameters are determined by fitting observed data l for the plant being analyzed, or from analyzed industry experience for a given i degradation mechanisms. l l Because the actual number of cracks found in 2R10 was signi6cantly higher j than predicted (170 found vs.110 predicted), a more conservative approach ! using a 95% con 6dence level has been adopted. His work built upon the ; extensive work originally performed which used industry data and ANO-2 data , up to 2R9M. Current efforts concentrate on the ANO-2 specific data and j includes the results o' 311 inspection campaigns to date. He observed data for , all five ANO-2 inspections (for both the "A" and "B" SGs) is plotted in Figure 5.2-7. He best fit function to the last two inspection campaigns (2R10 and 2P95-1) conservatively results in 0 and b values of 16.436 and 6.197 for the '
"A" SG and 14.228 and 15.539 for the "B" SG, respectively.
~
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.: AhG 2CAN029505 Page 27 of 89 l 5.2.3 Statistical Pmjection ofDegradation Expected at the End of Cycle 11 Utihzing the Weibull distnbution derived above and applying it to the ~ nine ;
months of operation remainag in Cycle 11, the number of anticipated circumferential cracks at the next refuehng outage (2Rll), which occurs at , 11.6 EFPY, is anticipated to be 276 in the "A" SG and 200 in the "B" SG. : Applying this projecten to the gamma distribution derived imm the data plotted in Figure 5.2-1 results in the conclusion, with a greater than 95% con 6dence level, that no flaw is anticipated to exist in ather the "A" or "B" SGs at the end of Cycle 11 which would exceed the structural hmits - established by RG 1.121. { 5 y r i b I i I
l
- A*= L .=t to 2CAN029505 Page 28 of 89 l
Frequency Distribution for Avg. Crack Depth j 45 - Five Tubesheet inspections 40 - M ANO-2 Data 35 - -Gamma Dist.
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15 - 10 - 5-0 Qhr1 , I . .. . 0 10 20 30 40 50 60 70 80 90 100 Avg. % TW Figure 5.21 i l i Frequency Distribution for Arc Length ) Five Tubesheet inspections ; 100 - 30 so . i ro - l eo . 30 - 20 10-0-A d d " + sd u -a : 1814: I-
- w R888 !R!!BRHERRRRRR Cract A#t Length Figure 5.2-2
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, g , , , 2P95-1 Circ Crack Data
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Attachment to 2CAN029505 Page 30 of 89 ANO-2 STATISTICAL PROJECTIONS 1800 1600 1@ E 1200 1000 i W 800 $ 600 o
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t Attachment to'
~ - 2CAN029505 j Page 31 of 89 ;
5.3 Structural Analysis j 5.3.1 Regulatory Guide 1.121 Evalusten A. Summary f W following is a brief summary of the requirements contained in RG 1.121 , relatmg to the analysis and testmg to determine the pe=wy margms against i steam generator tube rupture:
- 1) Normal Operaten Tubes with throughwall or part throughwall cracks should have a i margin of at least three to burst, determined analytically either by tests or by refined finite element or fracture mechanics techniques. h l material stress / strain cheractenstics at temperature, fracture egh , i stress intensity factors, and material flow properties should be i considered in making this determination.' h American Society of 1 Mechanical Engineers (ASME) Section III code requirements must i also be met. For ANO-2,3AP is 4050 psi. -l I
- 2) Accident Conditions l
l h margin of safety against failure under accident conditions, i concurrent with safe shutdown earthquake (SSE), should be consistent ! with NB-3225 of Section III, and be accommodated by the ultimate l tube burst strength de*:rmined expenmentally at - the operating ! temperature. l i NB-3225 suggests use of Appendix F, " Rules for Evaluation of Faulted l Conditions"- ! I Appendix F allows the use of any of the following methods, a) Elasticanalysit b) Stress ratio c) Collapseloaddetermmation t d) Plastic instability load or stress determmation 2 e) Strain limit loss or stress determmation f) Inelastic analysis B. Approach Options for RG Analysis ! A number of different analytical and test methods may be used to det.mine !' allowable tube degradation per the above RG requirements. Accordmgly, some variation can be expected in the final answer, dependmg on the methodology chosen 'Ihese variations are typically small, and are i l 4
,- c , --- - . , --
ce. Attachment to 12CAN029505 Page 32 of 89
-W-*u with margm, by the requirements of RG 1.121, e.g., by the safety factor of three required for normal operating pressure.
Among the factors which can in8uence the outcome of the results are the following:
- 1) Analysis methods:
a) ne type of defect analyzed (e.g., the directen of contralhng stress can be diferent for a circumferential defect than for an axial defect, depending on the size of the defect: testing has shown that circumferential defects up to about 58% throughwall, 360 degree extent, or 100% throughwall for 90 degrees will fail with axial splitting due to -cu.-J.-. Gal stresses, at pressures equivalent to a virgm tube)". Also, an OD carcumferential defect will have a higher allowable percent throughwall penetration than an ID defect, due to the diference in remaining cross sectional area, and the additional load imposed on an ID defect. b) ne type of analysis technique chosen (e.g., ' Appendix F attematives). c) ne method chosen to m- ----i#= variations of material properties (lot variations) which occur in the steam generator tubing. d) ne method of determining flow stress (burst strength). e) ne method of adjusting material properties from room temperature to operating temperatures.
- 2) Use ofBurst Test Results a) The method of accc.. ..cd4u.g variations in test specimen material properties.
b) The method of accommodatmg variations in tubing wall thickness and diameter (e.g., ANO-2 tubing is 0.048" wall, 3/4" OD tubing and Westinghouse tubing [ largest data base) is 0.043"). c) Burst test configuration: for circumferential cracks at the face of the tubesheet, testmg has shown that use of a simalated tube support structure above the defect will essentially alimia=*- bending loads". Testing with
- m. . . .. _ _ _ _ _ _ _ _ _ _ .. _ _ .. _ _ _ _ __ __ _.
i Attachment to !
' 2CAN029505, l
- Page 33 of 89 !
i bendmg support, as would be provided within'the steam j generator, will significandy increase tube rupture pressure,- , compared to an unsupported test specimen. j C. Original ANO-2 Analysis ; Le original ANO-2 RG evaluationM is based upon analysis performed by CE 1 and independently reviewed by MPR A=acia*= nis evaluaton utilizes the following methodologies for r.r.dg..g circumferential cracks.
- 1) Normal Operation An evaluation of burst strength for circumferential cracks was done i based on burst test data from PNL-2684, " Steam Generator Tube i Integrity Program," Annual Progress Report, January 1 - December 31, 1977. CE evaluated the results of a 0.875" diameter tube with a 77%
throughwall circumferential crack that burst at 5100 psi (> three times ; R normal operating AP). He ratio of wall thickness to diameter was 1 compared to ANO-2 tubing and it was determined that the test was l boundmg; and ucordmgly, the allowable defect penetration was l greater than 77% for ANO-2. MPR performed a more specific l analysis, in which the flow stress was calculated for the burst tube, then j correamns were calculated to account for the variations in wall thickrass and diameter, with mimmum ANO-2 material properties assumed. It was also conservatively assumed that the cracks initissed l from the ID. The equivalent flow stress for ANO-2 tubing was them limited to 1/3 (i.e., to attain the margin of three .to burst), and ibie j muumum wall thickness was calculated which corresponds to this ) stress. His stress corresponds to a remaining wall thickness of 21%, or an allowable defect depth of 79% throughwall.
- 2) Accident Conditions The 77% penetration was analyzed in accordance with ASME Section III, Appendix F, by the elastic analysis method for components, to confirm that the resultant stresses for accident conditions were less than the allowable; i.e., stresses were less than 2.4 S, and less than 0.7 S,.
(Note: the factor of 0.7 also has been used by others to iridhm;dy define a requirement for a 1.4 margin for accidents (1/0.7 = 1.43) but this value is not specified within the RG, as is the factor of three for normal operation). He calculated stress levels were well below the allowable, such that the 79% value for normal operation also covered accident conditions. Note: the RG states that the margin should also be accommodated by the ultunate tube burst strength determined experimentally at the operating temperature). However, using the Code values for S, and S, conservatively bounds this consideration. , i Although these Code values do not vary with temperature for Alloy I
~
Attachment to 2CAN029505
- Page 34 of 89
' 600 material, they are lower than the actual burst stress values at ;
operating temperatures (is) , ,
. D. Subsequent ANO-2 Analysis ;
Additional analysis has been unhzed to confirm the validity of the original ANO-2 approach ( 8) . 1)- Analysis Techniques ; i a) A statistical analysis was performed of actual ANO-2 tube properties (certified material test reports) to establish the - proper input for analysis, providing a 95/95 confidence : level. Results showed that the properties used in the , original analysis were conservative. , 1 ANO-2 lower bound material properties (room temperature): l Category Ulumste Strength Yield Strenath l SG "A" 91.92 39.83 SG "B" 91.68 39.95 - avg. 91.8 39.9 b) An evaluation of Alloy 600 flow stress was performed to refine the necessary analysis input. 'Ihis evaluation , considered all burst testing whose configuranons simulate r the correct ANO-2 configuration. It was done prior to the : secondary screening discussed in 2) below such that it includes 1. significant number of known low (Westinghoun) data. 'Ihis evaluation determmed that the best correlaien of burst stress to material properties at room temperature (rt)is: Flow Stress, = 0.507 (Ultimate, + Yield,) c) An evaluation of material property corrections from room l temperature to operating temperature (ot) was performed. i This evaluation determmed that-Ultimate, = .97 Ultimate, Yield, = .87 Yield, l Note that yield strength (Y) is more dependent on : temperature than is ultimate (U). Othces (e.g., Mainte l Yankee) have used the correction of yield strength to adjust :
Q A=h==e to 2CAN029505 n Page 35 of 89 flow stress (FS); however, burst is controlled more by l
, ultimate strength. 'Ihe correction discussed below more - !
closely approximates the actual effects of temperature on l burst. d) 'Ihese corrections are applied together, such that burst strength is adjusted for effects on both yield and ultimate j ~ strengths, as follows ; FS, = 0.507(.97U, + .87Y) 'l l Note: Based on actual burst test results (see 2 below), this approach is additionally conservative, as expected, due to , the inclusion of a large number of known low data. Using j this approach, the wie failure stress for analysis, i-l=Eag the lower bound 95/95 ANO-2 material ; properties, and mPMag for operating temperature is:
.FS, = 0.507 (.97(91.8) + .87(39.9)) = 62,745 psi l Ushig this failure stress to calculate allowable wall i penetration (avg. 360 degree OD crack) yields 77% j For comparison, flow stress from the original ANO-2 RG 1.121 analysis was assumed to be 0.85(best estimate :
Ultimate) = 0.85(90,000) = 76,500. Using this failure .; stress yields an allowable (360 degree ID crack) of 79%. i
- 2) Tube BurstTest Data a) Industry burst data was reviewed and screened to eliminate non-representative data. Applicable test results are shown lj on Figure 5.3-1. The largest set of non-representative data !
was detemuned to be the Westinghouse data for MP-2.
'Ihis data was elimmated based on Reference 17 where it l was determined that s' conservative methodology for l determining remanung wall thickness a8er electrical i discharge madd.deg (EDM) resulted in low apparent burst ,
pressures. The remaining burst test results were then ! normalized to account for lower bound ANO-2 properties : and corrected for operating temperature. Also included in '! Figure 5.3-1 is tihe ANO-2 pulled tube which was burst i tested (aAer correenon for bendmg stresses). A tube from the ANO 2 lab crack program that had sufficient cracking to fall within the area ofinterest is also included in the CE i data shown on the Sgure.
]
I e
- - - _ , -- - -__ ______________J
. . .. . ~ .
Attachment to - 2CAN029505 l Page 36 of 89 5.3.2 Finite Element Model To augment the analytiul RG evalumnon, an approach to evaluate the response of certain flaws utilizmg finite element techniques was desi, ped. Two types ofmodels were deci, ped: 1
- 1) 2-D wi.. ..e:ric model for analysis of 360* circumferential l crack
- 2) 3-D models for analysis of specific flaws (e.g., Tube 32-126 l containing thelargest 2R10 flaw) ;
1 The failure criteria used assumed failure when the average effective stress across a given secnon is equal to the flow stress at temperature. He effective : stress from the non-hnear analysis is smular to von Mises stresses of a knear , analysis. A. 2-D AxisymmetricModel 2-D axisymmetric finite element models (FEMs) are appropriate for modehng f 360* circumferential cracks. The following crack sizes (average %TW) were considered in this analysis: 1) 74%, 2) 79%, and 3) 82%. ! i The results were as follows: I Ave. %TW Burst Pressure for o% 74 5330 [ i 79 4740 j 82 4090 ! B. 3-D Finite Element Analysis 3 D finite element analysis (FEA) is appropriate for modeling a tube with flaws that are not symmetric. Tube 32-126 is considered in this analysis. Two 3-D - models were developed for this tube: 1) a model consistmg of first order ; elements, and 2) a model consisting ofsecond order elemets. !
- 1) First Order 3-D FEA for Tube 32-126 f
l In this case, four cracks were modeled on three planes. De stress contours for a pressure of 3AP (4050 psi) demonstrate the tube would , not be expected to burst at 3AP since e, < o,, However, a more
._ . -. .. . .. ~.. . - .
N' ! x ., ' Attachment to - $'"' 2CAN029505 y , Page 37 of 89. i ~ (l' , * . 1 i accurate solution (and higher stresses) will be raalwM with the model i
.r consistag ofsecond order elements.
I
- 2) Second OrderFEA forTube 32-126 l
De model consistmg of second order elements is bemg run in steps at
~
the time of this writmg A more accurate prediction of the burst ! pressure can be made when these results are available. : i C Conclusion j i ne finite element model is an additional technique used to assess the structural ! signi6cance of flaws. %e results of the FEM work demonstrate tbt the j original limit of 79% is reahstically conservative for the ANO-2 circumfet mitial cracks, and will be maintuned as the appropriate RG limit. Based on th 3 UT data and RPC deconvolution analysis evaluation results, all flaws from the ; ANO-2 steam generators have met the RG 1.121 margin requirements. His is i based on the largest flaw being 76% average depth (UT Average %TW). UT l has been successfully utilized by others to evaluate the presence ofligaments ; for the purpose ofp forni.g structural evaluations (28). ! 4 e i 4 l 1 I i i l l l 1 j l
- i Attachmesnt to - 1 N 2CAN029505. '!
Page 38 of 89 '
.F In-Situ Pressure Testag l 5.3.3
.)
A. Purpose j Per the requirements of RG 1.121, the effects of defects on steam generator l tube burst strength must be determined and found to be within the margms : defined in the RG. Both analysis technique and NDE inaccuracies must be j accommodated in assessing actual defects against the allowable defect size. j " Traditionally, to demonstrate this, defective tubes have been pulled from steam j generators and burst testag is then performed in the laboraeory. However, the removal of tubes for testmg has the following drawbacks 1) the radiation dose i during the tube pulls, and the subsequent handling of the pulled tube durvig l shipping, handimg, and testing,2) the costs ==aari=* art with the extraction and : hanAI;ai .3) the maccuracy of burst test results due to damage that occurs to - l the tube during the pulhng process,4) the potential problems which can occur ! associated with aborted tube pulls (e.g., if a tube breaks or becomes stuck , 1 during the pull, the potential for loose parts or vibration induced wear of adjacent tubes is signi6cantly increased), and 5) tube lateral support is often , missing or is roughly simulated in lab tests. In order to alimia* these ! problems, an in-situ pressure test device has been developed to allow testing of- l a defective tube within the steam generator without the need to pull the tube, l thereby avoiding the above drawbacks. The process discussed below is bemg l used for this purpose at ANO-2. 'Ihis discussion is limited to the use of the CE ! supplied in-situ testmg device which is used for. testmg of ANO-2 steam ! generator tubes with circumferential defects at the secondary face of the tube sheet. However, similar evaluations can be used to justify this method for .i other types of defects. l l 3 B. TubeI.cadmg : l For the deep penetration circumferential defects ofinterest, the main loadmg l which is significant for burst testing is the axial load unposed by the tube i intemal/extemal pressure differential times the tube cross sectional (nondefective ID) area 00 Accord. . gly, the in-situ pressure test device must transmit this axial load across the defected region for the test to be valid. This j is accomplished by the CE tool, which has a slip joint design between the two ! sealing bladders as shown in Figure 5.3-2. This design allows each end of the ; device to' move independently of the other when the chamber betweat the : bladders is pressurized, such that the pressure load is transmitted to the tube in j a manner equivalent to a full tube hydro. This is also equivalent to the AP ! loadmg for a laboratory burst test, where each end of the tube is capped 'Ihis j equivalency was demonstrated by CE in a companson test, where stram gauges : were installed in a tube that was hydro tested in a manner equivalent to the lab : type burst test, and then w.i.gwed to strain gauge results using the in-situ ! , device. The test showed essentially identical results for axial loads in each l configuration"M. This is shown in Figure 5.3-3. Accordmgly, the in-situ i device can impose the proper loading on the tube. ! 1
- ., . . ~ . - .
j, hn
- Attachment to 2CAN029505 " - Page 39 of 89-G C. Evaluation ofPotential PEects ofLocked Support Plates a
Eyefs with various tube support plate designs has shown that a non-protective magnetite formation in the crevice between the tube'and plate can grow to the extent that the tube can be deformed (called dentmg). h conversion ofiron to magneate is accomphshed by a two-fold volume increase This increased volume squeezes the ' tube resultag in tube dents (relatrvely uniform contraction of the d=nvear) at support plates and ov=hrahan of tubes j at eggcrate suppons De presence of these corrosion products can potentudly ! lock the tubes into the support. Because of the open nature of the eggerate 3 support, the lockmg forces are confined to the four line contact cremces. .nus, i tubes will be less severely locked into egscrate supports than dnlied supports, ) but if signi6 cant dentirig has w.d, senificant loads may be required to- j remove tubes from the eggerate In one modnt boiler test where corrosion was ' so extensive that the oggerate had begun so breakup, a hydraulic jack was ! required to pull tubes imm the eggcrateM. I~Irad supports can cause several j negative effects. h poiential effects on in-situ pressure testag will be j discussed here. , 4 If a tube is locked in the supports, then the full axial load imposed by the in-situ ! test device may not be carned by the defectrve tube. b non-protective i magnetite that produces denting (and locked tubes) develops only at operstmg j temperatures. Because of the greater coefEcient of thermal expansion for i Alloy 600, when cooled, the tubes will shrink more than' the carbon steel parts ! of the generator (shroud, etc.) which would result in a tensile load during l shutdown. However, the eggerates and the partial drilled support plates are ! flexible in the axial direction. A recent CE analysis indicates that with wide 'j scale tube lockup the axial stress for most tubes is near zeroM. h exception < to this is the normal hard spots a4acent to tie-rods and the lugs where the i eggcrates are attached to the shroud; axial stresses at these locations may be ! 2000 psi during shutdown. ! For those tubes with near zero shutdown load, the axial load imposed by the l pressure test would be reduced by the resistance provided by a locked support, ; i.e., the adjacent tubes which would not be pressurized would cany some of the in-situ test load that would not be carried if all the tubes were pressurizcd. l l While this effect is thought to be small, its signi6cance is ad6tionally mitigated i by the lack of any indication oflocked tube supports at ANO-2, as discussed i further below. j For ANO-2, the maximum loadmg would result from main steam line break (MSLB). During this -4M a concern exists that if a signi6 cant number of
)<
excessively defective and locked tubes were loaded the TSP could fail, I resulting in multiple tube failures, perhaps with the failed support causirs even l more failures than those which had been excessively defective. A part of this j concem is that an in-situ pressure test might pass an excessively defected and ; 1
)
l
o
' Attachment to :
2CAN029505 - Page 40 of 89 locked individual tube, leadmg to an incorrect conclusion regarding the vad behavior of the steam generator during an accident loadmg, when all the tubes would see the maximum loadmg. His potential condition has been evaluated, and it has been determined that it is not a factor which afects the intended functon of performing the in-situ pressure test at ANO-2. His is based on the followmg 1) the determmation that ANO-2 tubes are not hkely locked in their suppons, and 2) the hmited number of excessively defected tubes which would afect the supports, as discussed below D. Assessment ofPotential for ANO-2 Iacked Supports For the evaluation of chemical cleaning of the ANO-2 steam generators, CE performed an assessment of the efects of cleaning on tube supports, considering existmg cormsiorF8). nis evaluation concluded that corrosion of the ANO-2 eggerates was not signi6 cant (based primarily on hmited denting, as umfgir,6d by NDE). Signi6 cant denting has not been had at ANO-2 (very small dents have bem observed in less than ten tubes out of 16,822) in the lower eggerates his indicates that it is unlikely that the tubes are locked within the lower supports, where denting could afect the in-situ results. Although the drilled upper partial supports at ANO-2 are dented, these supports encompass only the outer periphery of the steam generator and are outside the region of the magority of the detected circumferential cracks Also, - the drilled plates are well above the in-situ tested defects ofinterest at the tubesheet. In addition, two tubes (19-55 and 96-166) from the "B" steam generator were pulled through the 1st and 2nd eggerates. De absence oflocking is further evidenced by the results of these tube pulls:
- 1) Examination of the tubes in the area of the eggerate supports shows no evidence oflocking: the tubes are not dented, and crevice corrosion indications are limited to a small region at the lines ofcontact with the eggerates.
- 2) De pull forces required to remove these tubes were not significantly diferent from those pulled from below the eggerates, suggesting that no locking had occurredczz) ,
However, the forces required to pull all the tubes was quite high, indicatmg the tubes were locked hi the tubesheet, which may have masked any efects fmm the eggerates. E. Assessment ofEfects ofIacked Supports In the unlikely evet a few tubes could be locked, the possible efects were considered. As diead elsewhere in this report, the number of large circumferential defects which could occur between ia=,We is very small. Also note that the in-situ pressure test loads the tube well above the actual
7j $p , lA==b==f to j
. 2CAN029505 Page 41 of 89.
loads that would be imposed during an accident.. AM.igly, if the small number of tubes whidt are excessively degraded also happen to be kekad in place, the load transfer to the a4acent tubes during the in-situ testing is not hkely to occur, since clustering oflocked tubes is not W_ nerefore, the results of the in-situ pressure test will be w :Gdy indicative of the tube's i actual behavior during an accident condman. 2 F. Test Pressure Correcnon he in-situ pressure test should be run such that the RG requirements are l demonstrated for normal operanon, and for accident (faulted) condmons.
- 1) Normal Operanon l I
For normal operation, the required pressure is three times normal operating pressure differennal. Normal operstmg pressure is 2250 - , 900 = 1350 psi, and the required pressure is 3(1350) = 4050 psi. His I pressure must be corrected for 1) temperature effects, and 2) ! mstrument error. a) TemperatureF5ects l Matenal strength is somewhat reduced at operating ! temperature hus, the in-situ test nm at room temperature must be increased to account for this effect. From Reference 15, this correcnon is 6.5%, or (4050)(1.065) = 4314 psi. ; I b) InstrumentError i His correenon depends on the particular pressure gauge used f tor the test and should be reviewed at each applicanon For the i 2R10 use, the correction was %% of span (5000 psi), or an additional 25 psi ( 8) ) l Based on the above, the required test pressure for three times normal l operating pressure is 4339 psi. During 2R9 and 2R10,4700 psi was [ selected as a tarpt pressure for the test. His value is high and was ! chosen to ensure the temperature and instrument corrections were l adequate. l
- 2) Accident Conditions l De maximum accident loadmg occurs during a MSLB, where the differential pressure is 2500 psi ( 5). Here is no dynamic amplificanon of the tube loads involved since the pressure change rate inside the l steam generator is not sufBeiently rapid. In essence, the saturated temperature water on the secondary side acts sinular to that in a I
.__ ______I
Y Attachment to l 2CAN029505 l
' Page 42 of 89 !
t i presswiser and retards the rate of pressure reduchon during a MSIR -l accident. De RG requirements for anddaw condmons impose ASME Section HI, NB-3225, which in turn refers to Appendix F, " Rules for l Evalunhan of Fauhed Conditions." Appenduc F ofers several analysis options for the evaluation. For the method normally chosen for tubing l (elastic component analysis), Appendix F imposes a design limit of the lesser of 2.4Sm or 0.7Su. Neither of these limits varies with j temperature for Alloy 600 material. A 4. gly, a temperature ; correction is not required for this limit. From Reference.15, the i required pres ,ure is 3500 psi, a) InstrumentError , j his presswe should be corrected for instrument error as l discussed under the " Normal Operation" section above i A 4#j, the required test pressure for accident loads is : 3525 psi. l 1 G. Conclusion i h in-situ pressure test is an acceptable means of demonstrating structural : integrity of defective tubes in accordance with the requirements of RG 1.121. I When considenng the advantages of the in-stu test over a t,'oe pua as discussed above, the in-situ test becomes the preferred methmi ne test should be performed at the pressures indicated above .) i H. Experience In-situ pressure testing was first conducted in the industry by Westi #ouse at the San Onofre Nuclear Generahng Station in the early 1980s. De second j known use was by ABB-CE at ANO-2 during the ninth refueling outage, ' where the entire tube was pressunzed folkd.ag pluggmg at both ends. nat tube held a pressure of 4700 psi for ten minutes with no leakage. During 2R10, three tubes were selected for testing. h pressure test device had been redesigned prior to the outage to allow for more tubes to be tested in a shorter time frame h new device unhzes two expandable bladders app-o.dy 3.5" apart, and allows the chamber between the bladders to be pressunzed up to 5000 psi. Eree tubes were selected for testag in 2R10 based on their defect size and physical location in the SG to allow testing from one fixture location due to ALARA consideranons & results of all ANO-2 in-situ pressure testmg are summarized at the end of this section. De in-situ pressure test system unhzed a small capacity positive displacement 1 pump which could maintain a leak rate of up to 0.6 GPM at 4700 psi (0.94 ! GPM at 0 psi.). Tube 32-126 leaked in excess of the makeup capacity of the l
- -. , e, - - - , - . , . - . -- .- - , - . . ,-- - t
i Attachment to'
- _2CAN029505 l Page 43 of 89 j
.j system wbm tested, but did not best as evidenced by the tube holding 2000 ;
psi for nr.aswoment of aleak rate.
]
i The lealage values shown below have not been corrected for operating l temperature, but operstmg temperature values are esumated to be ! approximately 25% of the room temperstwe rates. The leakage for Tubes 24- l 132 and'32-126 is consulered high due to the tube sweihng during the l pressunaanon to the higher value causing the crack to open shghtly. ' I Laboratory experience indicates this does not represent a decrease in the burst strength of the tube. h applied pressure ddferenmal leads to plastic deformation of the crack face such that any subsequent leak tests exceeded the .
*W leakage at the lower presswe ddferennalsM. l l
Based on their peak pressure, Tubes 64-48, 24-132 and 48-50 met the j requirement of 3AP. Tube 32-126 was unable to be pressunzod to the 3AP ! pressure due to leakage, but did exceed the 1.4 times MSLB pressure of 3500 _ . psi. Based on the Ur results, coupled with finite element modeling and the in- -l situ pressure test results where applicable, all ANO-2 circumferential cracks l tested met the RG 1.121 structural margin requirements. i During outage 2P95-1, three tubes were in-situ pressure tested using an , improved tool. h new tool had the capability to perform pressure tesang of l tubes experiencing leakage in excess of makeup purnp capacity by elMag a bladder over the crack area (as is done in laborasory pressure test). All three of the tubes tested in 2P95-1 held the maximum pressure with no leakage, and . i thus the use of the bladder to seal the crack was not required. -j Table 5.3-1 In-Situ Pressure Test Results Outane Tube RPC Ave %1W Max. Pra==m I+=kaae 2R9 64-48 58 ~4700 psi 0 j 2R10 24-132 51 -4600 psi 0.05 GPM @ 2000 psi l j 2R10 32-126 97 ~3600 psi 0.15 GPM @ 2000 psi 2R10 48-50 48 ~4700 psi 0.04 GPM @ 4700 psi 2P95-1 23-135 69 ~4550 psi 0 2P95-1 77-97 34 ~4550 psi 0 2P95-1 24-134 39 ~4550 psi 0 1
.jeenume w 2CAN029505 L Page 44.of 89 , i
- LEGEND: !
t c . CE Test Data i O PNLTest Data l .
- ANO 2 Pulled Tube h64-L48 t 12 '
t 11 - ------------------------------------------J--------- t i 10 -- - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - 1 i i g-.--------.---.--.-------------- 3----.--.------------. ; 6 \ I h8- -------------------------------- \e9------------------ , n
\- '
h 7-
-----------------------------------\---------------- l 3 .
\
i 6- --- l 4 \ ud ! e 5-----------------------------------------\
= \' t a
f
\
4-- .-,.-,,-__1X h .,- . ,4-. j -- -- -- 1.4 x SLB -- 3- --------------------------------------------l------- s i j l
\
sts y ; t 2- -----------------------------------------------\----- ' i .
.\ ,
. ,. . \ t 1- ----.- - - - - - - - - - - .- - - - - . - - - - - - - - - , -----------------\--
i i . t
. 3
. g
\
I 1 . ! ' 6 o . . . . . . ', 0 10 20 30 40 50 60 80 90 70 100 1 Average Depti(% Through-Was for 36(f Defect) j Figure 5.3-1 Comparison of Burst Tests to RG Limits i This figure indicates a RG limit of 82% for best St, or a lower bound limit of 79%. ; 1 I i i i
Attachmerdia i 2CAN029505 , Page 45 of 89 I steam generator l g - { seals (typ) grippers # spnng m v b e
, pressure
/ chamber
/ \
/ O C \
/ \ :
/ \ l
/ \ '
tubesheet \ test pressure t connection Figure 5.3-2 > Schematic Drasg of CE In-Situ Pressure Test Device
+ M-AttachmentO g 2CAN029505 l c - Page 46 of 89. Average Strains a 600- ' i r
/ .
- a.i.i _
Microstrain
--*- Hoop a00 .
g , r 200 0 , o imo 2000 3000 ; 4000 P$l Capped Tube hydrostatic Test Strain Results Average Strains l 1000 800 - Y W r-- Assal
--*-- Hoop Macrostrain 400 l
200 m a e ' O 1000 2000 3000 4000 5000 5 PSI ' 1 In-Situ Pressure Test Tool Strain Results -{ i Figure 5.3-3 l
2 g 2CAN029505 - '
?Page 47 of 89 '
5.4 Laak Rate Analysis
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lAak rates for indrvidual cracks were calc'Jabed using a basic calcidehal technique j which is presented in Reference 6 and deals with leak rates through axial cracks caused , by primary water stress corrosion cacking (PWSCC) in expansaan transinons. De ! leak rate, expressed in gallons per minute at the primmy fluid temperature, is . -l proportional to the crack opening area, a flow &sdiarge coefficient and the square soot i of the "effecove ddierential pressure " j Crack opening areas fer circumferential cracks were cale=1=*ad and then benchmarked versus test measurements on a tube with a 200* (total) throughwall crack. De i presence of structures in the steam generator which limit the later61 esplacement of i tubes was considered. ; ne flow discharge coef5cient is a function of the total crack opening. De semi-empirical equation of Reference 6 was used. De total opening is mraaa ad of the 3 initial crack opening under no load ar.d the opening under a differential pressure. A' tight PWSCC ' crack was assumed. Cracks peduced in the laboratory may have : substantial initial openings and this has to be considered in the comparison of measured l and calculationalleak rates. ; t While the crack opening is a funcnon of the primary to secondary pressure efferential, i the " effective differential pressure" driving the leak rate is a hydraulic parameter. If the primary fluid flashes to vapor, a back pressure opposing the flow will be created. If heat transfer / friction is mimmal during the flow process, the " effective differential q pressure" will be essentially the primmy pressure minus the saturation pressure at the ; primary temperature. j l his is the condition assumed in calculatirig leak rates under steam line break ) conditions. If flow is not isentropic, then the " effective &fferennal pressure" will be j somewhat higher than that assumed since the back pressure will be lower. For steam j line break conditions and larger crack sizes, the assumption of isentropic flow is . 2 appropriate. For smaller cracks, the assumption of isentropic flow should lead to calculated leak rates which are low by about 20%. His difference is M.ly overwhelmed by the uncertainties in leak rates resulting from variations in crack j
)
morphology, and the uncertainties in crack sizi@. This basic leakage model was nedwad in coriunction with the statistical model to provide an assessment of potential end of cycle (EOC) leakage vad under MSIB loads. Le distribution curve of arc lengths for 2R10 was used with the upper bound (95%) estimate for number of cracks at 2Ril. For each crack it was assumed that % of the detected length was 100%TW.. His is considered very conservative based on tube pull data where the average 100%'IW extent has been ~ 20% of the total length, with the largest - 47%A. It is also conservative since the leakage calculations are based on the length of a single crack; whereas, the ANO-2 cracking is comprised of multiple smaller cracks. At EOC, the total summed leakage from all cracks under 1 _ -- .- -~ _ ,,.-_..,.y, _ , . , , . . . _ _ _
Attachment to 2CAN029505' - Page 48 of 89 i MSLB loads (2500 psid at operating temperature), is ~ 9 GPM. This would lead to an j offsite dose well below the applicable 10CFR100 limit. i 1 k r i I l i i l
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Attachment to : ' i 2CAN029505
$'E Page 49 of 89 i 5.5 - Crack Growds Rate l
Crack growth during the three operatag periods between four inspections was used to j develop growth ratss expected for Cycle 11. Growth rate data for the interval between ; 2R10 and 2P95-1 are being calculated and were not available at the time of this report. l t. i 5 A review of the ECT data from the inspections was used to determine the maximum ! and average growth rate, wtach could then be used to' assess the postulated end of. > cycle conditions. Growth rate pre &cnons are difficult due to the fact that all detected flaws are repaired : when dwar*~I One must rely on in& cations not being daar*=d in previous outages to ! obtain growth information. Based on knowing the existence of cracks, a review of ' previous outage data shows evidence of a flaw for a number ofindications A review j of the 1993 outage data showed that only 22 of 147 circumferential cracks found in the ;
"A" SG were clearly new. 'Ihe largest circumferential crack in 2R10 does appear to be present in the previous outage, although quantification is difficuh due to the influence 1 of OD deposits. This flaw was conservatively assumed to have grown from no crack in l 2P93-1 to its value in 2R10. Overall, however, the average growth ~ rate indicates no i signi5 cant change in the ANO-2 damage progression. This is shown in Figures 5.5-1 l and 5.5-2, where 95% of the indications grew s35% average depth and/or s110' in arc .;
length over the ten month operating interval Further work is ongoing to assess the ; growth between all of the previous operstmg intervals. ! i I As a result of the inherent difBculties in growth rate analysis, Entergy Operations believes flaw growth should be evaluated statistically. Such analysis was performed l' and is described in Section 6.2. i I i j 1 1 1 l
O k' Numtw of indications # of Indications D5 i-O b b b b b b om o$O S8M8N 10 N mmm O M h$h 8 :::::::-- ::- - g ..... s 5 10 333 3 gl 60 'm 15 MMMMM g 8 h 20 M3333 1 k 110~l 30 ..33 9 g I N 35 g F 140 I g 40 g jt [ 9 6 h E 45 h b 8 -l 50 'l p 55 "E Ig 200 g, y N 210 "I ~ 5 60 -* i 65 k 70 g O j{ 75 $ 280 80
$$0 85
!!8 i so 99 8
350 100 l ggg3gggggg 508888888 8
% Cumulative Dish % Cumulative Distribution
- ~ . __
- m. . ;
Attachment to i 2CAN029505 - I
- Page 51 of 89 i
5.6 Laboratory Crack Project , pt . Because of the imuted amount of 6 eld data on circumferennal cracks, a program was ! l initiated to genuate and analyze circumferential cracks typical of those found in the ! ANO-2 steam generators l ne defects were produced in 3/4 inch Alloy 600 tubing at the secondary face of a ! simulated tube sheet. The tubing was explosively expanded in the simulated t ~besheet l
. with the same process used for the initial fabncanon of the steam generators Seven f j tubesheet blocks containing six tubes each (for a total of 42) were assembled. ]
- De objectives of the program were to: l
. Provide additional data to support the RPC depth vs. actual depth correlation j e Pmvide insight into crack mitisbon and propagation j l
. Provide additional structural integrity,j fu... .ce data to support further RG !
1.121 correlations
. Provide insight into compansons of various NDE techniques =
i The following summarizes the mapr areas of the project: ; 5.6.1 Nondestructive Examination l Eddy current and ultrasonic nond-structive testmg was performed on the tube ! samples in an e5 art to assess NDB sizmg error. Vendors with field ready f r+-A=1uctive testing were invited to test the tube samples. . Vendors performing eddy current testing included /28L Combustion Er.gir i..g Nuclear Operations, Babcock & Wilcox Nuclear Technologies, We@gha>= Electric Corpvin;cri, Electric Power Research Institute, Zetoc, and M Ridge i National Laboratories. nose performing UT were B&W Huclear Technologies, Wsghause Electric, and NUSON. In addition to the NDE techniques normally utilized, a dye penetrant system l was used to measure the 100% throughwall extent of the circumferennal ! cracks in several of the defects. He tubes were ia-arm prior to leak and - l burst testmg. l 5.6.2 Residual Stress Measurements ! Residual stress measurements were conducted to determine the stress level in the explosive expansions of the tubes. All residual stress measurements were performed by an X-ray diffraction technique. Eis technique has been ) rengaiW for some years as a reliable means for measuring surface residual ! stresses and, by electropolishing to remove layers of metal, the variation of I I
1 Attachment to 2CAN029505 i
. Page 52 of 89 j stress with depth. In X-ray d ffraction ramWal stress measurement, the strain !
in the crystal lattice is measured, and the residual stress producing the strain is j calculated, assunung a hnear elastic &storten of the crystal lattice. 5.6.3 Metallurgical Analysis In order to evaluate the real size of the flaws, destructive exanunatens were p fvia.ed. W tubes were pulled apart using a tensile machme. h depth ! and length of the defect was determmed by measuring the depth of the crack every 10' around the tube using a light optical stereo microscope (LOM) with graduated eye piece. The depths at 10' intervals were plotted on a schematic sketch of a tube cross-section to provide a graphic representation of the size , and shape ofeach flaw. 5.6.4 Leak / Burst Testmg All of the samples were leak tested at normal operstmg pressure and at peak . accident pressure. Some of the tubes were leak tested to a pressure of 3 AP to i provide informaton on crack openmg under the higher pressure for comparison i to in-situ pressure tests previously performed All samples were sum ==*Iy l burst tested. While laboratory cracks are questmnably simdar to actual cracks in the SG, I certain data is valuable and presents conservative results. Information on NDE performance and leakrates may be dependent on crack dimmsions such as crack width and may provide misleadmg results. 'Ihe complete results of the program are cunently under final review and should be available later in 1995. . i i I f 5
p. 1
^
Attachment to', 2CAN029505-
-Page 53 of 89 6.0 SAFETY ASSESSMENT g 6.1 Safety Analysis L
An important aspect of any safety analysis is the demonstration that calculated offsite doses are _within the NRC staff criteria given in the Standard Review Plan (SRP), NUREG-0800. Steam generator tube leakage is especially signi6 cant since it has the potential to lead to r. containment bypassM. For ANO-2, the analysis utikzed guidance fmm Reference' 25,' SRP assumptions, reahstic assumptions for RCS activity, and a best estunate methodology (ANO-2 speedic CEPAC Model), examining both premisting and event-generated iodine spike
~ (PIS and GIS) consequences, L Using best estimate methodologies, the following were evaluated o
e Valid emergency operating procedure (EOP) guidance e Acceptable offsite dose consequences e Adequate refueling water tank (RWT) inventory The evaluation was performed for the following events:
. Steam generator tube rupture (SGTR) - single tube e SGTR- multiple tubes
- MSLB induced tube leak (s) or rupture (s)*
De MSLB induced single and multiple tube rupture event is limiting. The general methodology was to use the ANO-2 simulator and operating crews to determine the best estimate operator response times for representative MSLB/SGTR scenarios. %ese results were then inwipcid into the ANO-2 specdic CEPAC evaluations of various MSLB/SGTR scenarios. He CEPAC results were then compared to " hand" calculations. The safety analysis performed for ANO-2 yielded the followmg results:
. EOP guidance was valid for event induced tube ruptures; therefore, no significant changes or improvements to the existing EOPs were needed.
m 3 4 y : Attachmentto ~ ) d" ' 2CAN029505
; Page 54 of 89 '
. The offsite dose imuts of 10CFR100 and the contml mom dose hmits of General Design Criterion (GDC) 19 can be satis 6ed usmg renhstic assumptuns e Based on generic system e= leal =daan and demonttraten in the Ah0-2 I simulator, there is sufficient RWT inventory to shut down and I depressurize the plant before depleting the RWT and sustammg core I I
damage. I Details of this analysis were presented to the NRC staff on August 30,1993W. I l 6.2 Probabdistic Safety Analysis he avoidance of a severe accident leadmg to core damage is an important part of ' l assuring adequate protecten of the public health. Such severe accidents are important ! because, ahhough m....dj unlikely, they have the potential of releasing large s quantities of fission products to the environment. A probabilistic safety analysis (PSA) was performed in order to assess the impact of the SG tube degradaten mechanism on ; the probability of such a severe accident at ANO-2M. He objective of this PSA was , to identify the optimum ANO-2 SG inspection interval: one which assures adequate protection to the public, yet minimizes ANO-2 operational costs and radianon worker l exposure. ; He severe accident safety impact was evaluated by Wadag the change in the ANO-2 core damage frequency (CDF) as a functon of the time between SG i inspections since the !=gianing of Cycle 11 (BOCll) steam generator iam ! Several proposed inspection interval lengths were considered: (1) a half-cycle interval ! at the middle of Cycle 11 (MOCll), (2) a full cycle interval at the end of Cycle 11 !
. (EOC11), and (3) strictly for comparison purposes, a two cycle interval at the end of I Cycle 12 (EOCl2). De formal analysis was based upon steam generator irt@ ;
results through 2R10 (BOCll). He results of these analyses were qualitatively ! extended to include the 2P95-1 SG tube iaca-*= Gadinge in order to assess the risk : of operating for the remainder of Cycle 11. The CDF estimates were developed via the use of a modified version of the l ANO-2 PSA plant model ne ANO-2 SG mspection interval risk analysis was l performed via the development and quantificatan of event tree and fault tree models in - ! a manner similar to that done in the ANO-2 individual plant exanunanon (IPE)/PSA - l analysisA.' The subject ANO-2 SG inspection interval safety analysis differed from i
.the ANO-2 IPFlPSA analysis in that the subject analysis was limited to accidents j involving SGTRs, since its intent was to estunate the change in ANO-2 CDF for ;
several inspection interval options. In addition, the subject analysis accounted for the ; risk contributions due to both spontaneous SGTR initiators (R) and SGTRs induced by ; other initiators. He ANO-2 IPFJPSA did not account for SGTRs induced by other { initiators. l 1 l i I
- Attachment G ' '!
2CAN029505 Page 55 of 89 h ANO-2 SGTR CDF analysis included the following steps: {
- 1. Review and identinemeinri of events which could lead to spontaneous or ;
induced SGTRs (either initiators or subsequent events), !
- 2. Assessment of the SGIR conditional probability (CP) givri an imtsator l or subsequent event which could cause a SGIR, !
- 3. Identification of the safety functions important to assuring adequate l core coolmg, j
- 4. Development of event tree logic winch accounts for combinations of j safety function failures which lead to core damage (i.e., core damage accident sequence),
l 1
- 5. Development of system fault tree logic to account for canaaaaaat ,
failures which contribute to safety function failures, and -i
- 6. Quanti 6 cation of the above event and fault trees to estimate the frequency of core damage involving SGTAs.
The frequency of spontaneous SGTRs (i.e., thor occurnng dunng power operation which are not due to significant changes in the pnmary-to-secondary pressure diferential) was estimated to be s.77E-3/rx-yr per the ANO-2 FE/PSAA. W j frequency ofinduced SGTRs (i.e., those occumng during power operation which are a J result of a signi6 cant change in the primary-to-secondary pressure diferential) required i the review of the transient and loss of coolant accidents (LOCAs) described in the ANO-2 Safety Analysis Report (SAR)N, NUREG-0844M, and other sources of information. The potential for a SGTR event in each of these accidents was assessed by estimating the maximum primary-to-secondary diferential pressure (PSdP) occurring in each and estimdag tha probability that one or more SG tubes will fail as a a result of this diferential pressure. An accident was considered a candidate for inducing a SGTR only ifits maximum PSdP exceeded the nominal operating PSdP of 1350 psid(2250 psia- 900 psia). i Based on a review of the ANO-2 SAR and other sources, three accident initiators were I identified to produce PSdPs signi6cantly greater than the nominal 1350 psid PSdP: the l steam line break (SLB), the feed line break (FLB), and the anticipated transient without scram (ATWS). Other initiators, includmg LOCAs, were not considered signi6 cant SGTR initiators. Due to diferences in the plant response, the FLB/SLB accidents are assessed together followed by the ATWS-induced MTR accider,fs. 6.2.1 FLB/SLB Analysis In order to account for the ANO-2 plant response dependencies on the SLB location and to distinguish the FLB from the SLB, the ANO-2 IPE/PSA 1 i
, . - _ -_, . - - - , _ __ _ ._ J
m [r. m ' L Attachment to. l C ' 2CAN029505.- ; Page 56 of 89 j e
- combined SLB/FLB initiator (TS) _was cplit into four parts and swen a unique initiating event designator:
V H
- 1. Steam line piping outside of the main steam molation valves (MSIVs) 1
. (T5-1),
- 2. Steam line pipeg inside of MSIVs and outside of the containment (CNMT)on both SGs (T5-2), !
[' 3. Steam line inside of the CNMT on both SGs (T5-3), and 3
- 4. Feedwater. line inside of the ' feedwater check valves on both SGs (T5-4).
He frequency of each initiation was taken as the fraction of the total lengths of' the steam line (SL) and feedwater line (FL) that each secten represents times q the total ANO-2 IPE/PSA SLB/FLB frequency. . A summary of these i calculated SGTR frequencies is provided in Table 6.2-1 below. s l Table 6.2-1 -I Summary of SGTR Initiating Event Frequencies , ANO-2 Fracten of SGTR ! Initiating IPE/PSA Total SIJFL Analysis . l Event Frequency Length Frequency .l Urx-yr) Urx-yr) ! R 9.77E-3 not appl. 9.77E-3 T51 1.1E-3 0.699 7.690E-4 T5-2 1.lE-3 0.107 1.182E-4 T5-3 1.1E-3 0.114 1.258E-4 TS-4 1.lE-3 0.079 8.710E-5 he probability of a tube ruptunng in a given accident was usessed by developMg an estimate of the probability of a tube wall failure as a bnction of average tube wall defect ?qth (i.e., %TW) at selected ANO-2 bunes and comparing each of these " fragility curves" with the expected population of ANO-2 SG tube defects at each of these burnups. De expected populaten of ANO-2 SG tube defects, i.~e., the number of tube defects as a function of defect size (%TW), was estimated using SG tube inspection data collected in past ANO-2 SG iaW campaigns through 2RIO. It has been shown in Reference 14 that the Weibull functon can be used to predict the total number of defective tubes as a function of operstmg time (see Secten 5.2). Furthermore, the data collected at ANO-2 through j 2R10 and includmg the recent 2P95-1 mspection indicates that the sizes of the ! defects (avg %TW) can be described by a Gamma probability distribution j function (see Secten 5.2). Using these two relationships, estimates of the l l 1
- - - . - - ~
4 p .a.-- , ..a... + e . . .a an. .- --, - a n.. a a a _ a ..rs Attachment to
~
2 CANT 29505 ! Page 57 of 89 defect population aAer a half-cycle, a full-cycle, and two cycles of operation i were developed. *Ihe results of these analyses were quahtstively extended to l include the 2P95-1 SG tube inspecten findmgs in order to assess the risk of operating for the remainder of Cycle 11. i
. 'Ihe SG tube "fragihty curves" were based on expenmental SG tube burst l pressure test data. 'Ihese burst tests were performed for a wide range of tube l wall defect sizes based upon metallurgical examination of the tubes aner !
failure. For use in this study, these test results were si.c.ed for temperature ! and to an average RPC %TW indication (see Section 5.3). 'Ihe resultag data : was used to estimate the probability a tube will, as a function of its defect average depth, fail for a given PSdP. h likely number of SG tube failures , resulting from an accidait is the combination (i.e., convolution) of the SG tube i defect population and the SG tube fragihty distnbution for' a given bumup and ! PSdP and is depicted graphically in Figure 6.2-1. j The conditiorial probability of tube failure for a PSdP of 2500 psid based on the expected number of tube defects at 2Rll obtamed by this convolution l l' technique veas approximately 2.2E-03 which is less than the 1.0E-02 threshold - value provided in the dran Generic Letter for Voltage-Based Repair CriteriaN. I This value is the sum of the EOCll SG "A" and SG "B" conditional ; probabilities at the 95% con 6dence level (0.87846 for SG "A" plus 0.14689 - for SG "B" per Table 6.2-2, below) divided by the sum of the WM number of tube defects at 2Rll (276 for SG "A" plus 200 for SG "B" per Section 5.2). In order to assure conservative results, the EOCll SG "A" and l SG "B" tube rupture probabilities (the numerator) assume that a single j defective tube randomly distributed ber;ca 60%TW and 100%TW was not i detected during the BOCll tube inspecten campaign. ' Ibis single defective i tube assumpties dominates the SGTR conditiemal faumre probability estimate. For SLB- and FLB-induced SGTRs, the PSdP was assumed to be 2500 psid. This is the maximum credible differential pressure between the primary and secondary systems and represents the primary pressure at approximately the primary code safety relief valve setpomt with the secondary pressure at atmospheric conditions. For these conditions, if the number of tubes susceptible to rupture for a given accident was estimated to be less than one . tube, this value was conservatively. interpreted as the conditional probability of a single tube rupture an the PSA analysis of CDF. A summary of these SGTR conditional probabilities are provided in Table 6.2-2 below. l 1 l
A w h tto 2CAN029505 , Page 58 of 89 t Table 6.2-2 SGTR Con &tional Probabilities for Accidents L,civing a SLB or FLB Event SG SGTR Condmanal fa m PmbabahtyUsed . (+1 tube @ 95% CL) ! SG "A" SG "B" BOCll 0.3688 0 i MOCll 0.61434 0.06532 EOC11 0.87846 0.14689 ; EOCl2 1.0 - 0.41861 hse SGTR conditional probabilities were applied directly to the SLB and : FLB initiating event frequencies provided in Table 6.2-1 to obtain the ; frequency of SGTR in a given SG following a SLB or FLB at the specified time aAer the BOC11 SGinspe:: tion. ! Consistent with that performed in the ANO-2 IPE/PSA, the spontaneous ; SGTR, the SLB-induced SGTR, and the FLB-induced SGTR CDF analyses - were performed via use of event trees and fault trees. Event trees speedic to .; the SGTR accident were developed in order to provide a more detailed account of the accident progression than that in the ANO-2 IPF/PSA. h safety : functions listed in Table 6.2-3 below were identified to be important for the SGTR accidents involving spontaneous, SLB-induced, and FLB-induced ! SGTRs - l t I 1
Attachment to -
~ 2CAN029505 l Page 59 of 89 i P
Table 6.2-3 . i
. Spontaneous and SIE/FLB-Induced SGIR Safety Functions j Safety Function Desenpeor Desenpten !
Reactivity control - K Insert siMw-it negative reactmty to stop nuclear reaction and maintain reactor subentical . SGTRisolable with RCS Iam Ruptured SG can beisolated fmm ! intact environment and Containment i SGTRisolable withm Iomr Ruptured SG can beisolated from ! containment environment Feedwater(FW) avadable to Bn Main feedwater(MFW), emergency . ; intact SG feedwater (EFW), or auxihary feedwater (AFW)provides feed tointact SG FW available tointact or Bm MFW, EFW,or AFW provides feed to . ruptured SG eitherintact or ruptured SG l Secondary system pressure Bn Operation of atmospheric dump valves l control viaintact SG (ADVs) and/or turbine bypass valves .. (TBVs)onintact SG ! Secondary system pressure Bm Operation of ADVs and/or TBVs on either l controlviaintact or ruptured intact or ruptured SG i SG l RCS inventony control U High pressure safety injection (HPSI) or, ! ifsufficient, charsma l Once through cooling F RCS depressur=*vwn via emergency core ! cooling system vent valve orlow- : temperature overpressure protection vent - : valves Long-term cooling X Shutdown coohng(SDC)orHPSI recirculation j These safety functions were used to develop a spontaneous and SLB/FLB-induced SGTR event tree. 'Ihis event tree identifies the combinations of. ! initiating events and functional failures W to lead to core damage - resulting in fifty-two (52) accident sequences afimportance. Using these event ; trees and Boolean algebra, the ANO-2 CDF due to spontaneous and : SLB/FLB-induced SGTR was quantdied at BOCll, MOC11, EOCll, and ! EOCl2. These CDF values are " point estimates," i.e., they correspond to their . respective bumup conditions. 'Ihese CDF values were averaged over the intervals between BOCl1, MOCl1, EOCl1, and EOC12 in order to determine I the average ANO-2 CDF due to spontaneous and SLB/FLB-induced SGTRs in i cach interval 'these CDF values are reported in Section 6.2.3. 4 4s -e v. s es r- p,- ., , , - - -
*-we-- - ,,----,m- -.em - --- --v -- = - - ~ -*
I Attachment to ) 2CAN029505: 1 Page 60 of 89
]
l i 6.2.2 ATWS Analysis i Due to its severe * -- *W " challenge lo the plant, the ATWS-induced l SGTR safety analysis was performed separmaly from that of other initutors j As in the ANO-2 IPF/PSA, the dominant ATWS-induced oore damage risk ! was assumed to be me=*=I with only three transient groups: turbine trips j (TI), loss of MFW (T2), and loss of off-site power (T3) initiators. b safety l functions hsted in Table 6.2-4 below were identdied to be important in the ! ATWS-induced SGTR core damage accidents Note that these funcnons were ! selected to be relatively independant of each other in order to ensure that the' ! A1WS-induced SGTR CDF is mnservatively quantdied. % Table 6.2-4 i A1WS-Induced SGIR Safety Functions i Safety Function Descriptor Description Reactivity control K Control rods insert sufficient negative ! reactivity to stop nuclear reaction and i maintain reactor subcritical l Emergency AC power AC Emergency AC power i Turbine trip (IT) successful IT. TurbmeTrip successful ! MFW available MFW MFW avadable l EFW available E EFW avadable Moderator 6..wiware MTC Moderator temperature coefficient for coefficient TT/MFW/E and bumup condition i produces RCS-SG pressure differential .( less than 3700 psid i SGintegrity SGI SG tube integnty remams atact following ! ATWS pressure excursion Primary reliefsecured W Pnmary relief followmg ATWS pressure excursion secured Borated water BW Borated water injechon mserts sufficient l negative reactivity to stop nuclear reaction l and maintain reactor suberitical i Long-term cooling LTC SDC successful These safety fi=A were used to develop three ATWS-induced SGTR event : trees which identify forty-two (42) accident sequences ofimportance Using these event trees, the ANO-2 CDF due to ATWS-induced SGTR was j arithmetically quanti 6ed for the intervals between BOCll, MOCll, EOCll. - I and EOCl2.1he arithmetic quanti 6 cation of the ATWS-induced SGTR CDF is consistent with that performed in the ANO-2 IPF/PSA and with the screming nature of the ATWS analysis. & ATWS-induced SGTR CDF values are averaged over the intervals between BOCll, MOCll, EOCll, and f i
. _ _ _ . _ . _ _ ~ _-
g' ~ . TAttachment to .
~ 2CAN029505
- Page 61 of 89 EOC12 (the averaging is accounted in the ATWS event trees). These CDF values are included in the combined CDF values reported in Section 6.2.3.
6.2.3 PSA Analysis Results The results of the ANO-2 SG inspecton inearval safety analyses are hsted in Table 6.2-5 below and are graphicaDy depicted in Figure 6.2-2. These results include the combined core damage conenbutions of the spontaneous SGTR, the SLB- and FLB-induced SGTR, and the ATWS-induced SGTR accidents Table 6.2-5 ANO-2 Spontaneous, SLB/FLB-Induced,'and ATWS-Induced SGTR CDF Results Averaged @ MOCl1, and EOClI and EOCl2 w/95% +1 Confidencelevel Case BOCll to BOCll to BOCll to MOC11 EOC11 EOCl2 Spontaneous andInduced SGTR CDF 3.79E-07 3.98E-07 5.25E-07 w/ SGInsp
@ BOCl1 only Spontaneous andInduced SGTR CDF 3.79E-07 3.39E-07 3.39E-07 w/ SGInsp
@ Mid-Cycles CDFIncrease 0 5.92E-8 1.86E-7
'Ihese results, slightly revised from those presented to the NRC staff on July 14,1994"), show that the extension of the ANO-2 SG inspection interval from .
one half cycle (MOC11) to a full cycle (EOCll) interval has a negligible increase on the average ANO-2 CDF due to a SGTR. SpaciAcany, using the 95% confidence level CDF, the increase in ANO-2 CDF due to spontaneous and induced SGTRs is only 0.59E-07/rx-yr or about 0.2% of the ANO-2 CDF due to intemal events reported in the ANO-2 IPE/PSAW). This increase in CDF is small and is well widun the NO ACriON recommendation c :-.;wy per SECY 91-270 5) . In addition, it is. below the NO ACTION ran===A=G s per NUliARC 91-04% and within the NON-RISK SIGNIFICANT classification per the draA PSA Applications Guide. Since the increase in the ANO-2 SGTR CDF over a half cycle is small, the greater than WM SG defect population found during 2P95-1 at MOCl1 indicates that the continued operanon of ANO-2 for the remamder of Cycle 11 is not risk significant.
i Atta' chment to L 4 C, (2CAN029505, + N '. Page 62 of 89 l l , l 2500 PSID at 2R11 (95% Confidence with 1 , !
. tube between 60 and 100 MTW) l l
.v I
-i' 12 -- l' 10 -- SGB -
0.8 -
'8- SGA -- 0.6 Burst Prob. o
['6- 0.4 E ' o 4-2 eggs 88 i 2- 0.2 g 0 , , , , 0 0.00 20.00 40.00 60.00 80.00 100.00
%TW I I
Figun 6.2-1 Change in ANO-2 SGTR-Related CDF vs. SG Inspection interval -
, l 2.00E-07 ;
1.80E-07 1.60E-07 { a u 1.40c-07 / : T I O 1.20E-07 ! t I 1.00E-07 9 4 8.00E-08 / 6.00E-08 4.00E-08 2 E.0 , // Os l
.0.00E+00 1 DOC 11 to MOC11 BOC11 to EOC11 BOC11 to EOC12 j SG Inspection Interval Figure 6.2-2 l
1 =. . _ . .- - . _ . . ._ - .. . - . ., .-. . - .
K
; Attachment to' l
- 2CAN029505 E, . Page 63 of 89 7.0 - OPERATIONAL RESPONSE l An operatonal leak rate limit is established to provide reasonable assurance that flaws either missed during inspecten or growing more rapidly than expected will not render the tube vulnerable to tube rupture in the event of a MSLB. He ANO-2 Technical Speci6caten limit is 0.5 GPM per SG, but an administrative procedural limit of 0.1 GPM (144 GPD) exists to provide for added margin against burst. In Waaa, rate of change hmits exist to ensure rapidly propagating cracks or damage will be addressed at the earliest possible stages.
Upon any control room alarm indicatmg primary to secondary leakage, the Operations and Chemistry Departments enter abnormal operatung procedures. If the leak rate is 20.1 GPM, a plant shutdown is procedurally required. In addition, a plant shutdown is procedurally required if the leak rate is projected to be 20.1 GPM in one hour. Stable leak rates of >0.01 GPM procedurally require management awareness for contmued plant operations 7.1 Inservice Leakage Detecton/ Response There are several methods and techniques utdized by the Operations and Chemistry Departments to identify the presence of a primary to secondary leak. He need to provide leakage monitoring is given in the Bases for Secten 3.4.6.2 of the ANO-2 Technical Speci6 cations and denotes that a maximum total leakage of 1.0 GPM (0.5 GPM per steam generator) provides assurance that off-site release hmits will not be exceeded (small fraction of 10CFR100 hmits) in the event of a steam line break or a steam generator tube rupture, and guards against tube severance or rupture under the extreme differential pressure conditions imposed during a steam line break or LOCA. ANO-2 Technical Speci6 canon 3.4.6.2.c requires that when primary to secondary leakage is greater than 0.5 GPM through one SG or 1.0 GPM through both SGs the reactor is to be in hot standby within 6 hours and in cold shutdown within the followirng 30 hours. De plant is administratively shutdown at 20.1 GPM. The Operations Department trends the steam, condenser off-gas, and steam generator sample systems in determining indicatmn of steam generator tube leak. Steam lines are monitored via radiation monitors and nitrogen sixteen (N-16) gamma detectors, which provide the chemists and operators with the capability of quantifying leakage. Le amount of N-16 present in the secondary system is influenced by the size of the leak, the location of the leak in the steam generator, and the power level. He ANO-2 N-16 monitors have the capability of correctmg the leak rate for changes in power level, and are set to detect or discriminate $x the high energy gamma rays emitted from N-16 only; therefore, other iso; oyes that leak into the secondary system will not have an effect on the N-16 values. Because of the short half-life of N-16, the amount of N-16 seen by the detector will vary d-=viing on the location of the leak. ANO-2 N-16 monitors can be adjusted to display a leak rate for the hot leg bottom of the tube, the J U-bend, the cold leg bottom of the tube, or an average and thus make it possible to gather more accurate leak rates" ubr.iatim. Condenser off-gas monitonng provides a means to measure the gaseous activity levels released to the system vent (located on spent fuel pool floor) using a Geiger Mueller tube (h.dr.es gross levels of beta and gamma radiation) and an off-line gaseous sampler. Le steam generator monitoring
~
Attachment to 2CAN029505
- Page 64 of 89 system contmually takes steam generator blowdown samples downstream of their
' respective sample coolers to detect for fimeinn product activity (p.L. sily Cs-137) and allows for determinmg which generator is leakmg. Procedures (e.g., EOP 2204.004 and abnormal operstmg procedure (AOP) 2203.038) are utilmed when the monitors or trend recorders for the aforementioned systems exhibit increasing trends he Operations Department enters these procedurus to place the' plant in a stable condition and to mitigate the consequences of a steam generator tube leak. AOP 2203.038 (Primary to Secondary Leakage) is used when any of the folkd.g entry conditions exist:
(Increasing trends on any of the following)
. Mam steam radiatiant monitors or secondary syst m trend recorders Steam generator sample line monitors (samples tdan off blowdown)
. Ch.= off-gas monitors
)
. Steam generator tube leak N-16 monitors 7
. Unexplained " Secondary Systeen Radiatinen HI" annunciator alarm
. Increasing fission product activity through chr -J.tiy sampling
. Unexplained " Trouble / Leak rate HI" annunciator alarm l
. Unexplained " Rate of Change HI" annunciator alarm EOP 2204.004 (Steam Generator Tube Rupture) is used when any of the following i conditions exist:
(Increasing trends on any of the following) [
. Main steam radiation monitors or secondary system trend recorders ;
. Steam generator sample line monitors ;
e CA=ar off-gas monitors l
. Steam generator tube leak N-16 monitors i
ne Chemistry Department routinely samples both the primary water and .~aadary { water systems to identify primary to secondary leakage and trends the sample results to 3 identify possible primary to secondary occurrences Off-gas samples (extracted from ; the condenser vacuum pump discharge) are taken for Ar-41 and samples for tritium { (taken from condensate as liquid sample) are taken to quantify activity levels in the .,i secondary system Argon-41 levels yield more of an indicative measure of ; instantaneous levels of primary to secondary leakage. Tritium levels (in the secondary t system) will increase linearly over time during a primary to secondary leak. By ! knowing the tritium level in the primary system and the time delay from the initial ; indication of a leak to the tritium analysis in the secondary system, a primary to l secondary leakrate can be determined. Secondary liquid samples are taken to measure i fission product activity on such radioisotopes as I-131, I-132, I-133, I-134, Xe-133, , and Cs-134 via gamma pimscopy, which can identify the presence of rachoisotopes ; in the sample based on the energies of the incident radiation particles. Resin from the ; startup & blowdown denunerahzers is also sampled for increased activity (gamma
- spectroscopy is used to identify and quantify activity levels). -
r
i Attachment to
---2CAN029505 Page 65 of 89 ,
'Ihe following is a list of primary to secondary monitoring devices and their current -
setpomts:
. SG sample (2RTIS-5854,5864) ~ 130 cpm
..- Steam line rad;ation (2RR-1007,1057) - 1 mr/hr !
. N-16 monitors PRE-0200, O'01) 0.01 GPM,2X currentlenkrate .
. CW off-gas moniers - - 600 cpm a
-l a
P h e [ a 6 1
AttachmentG i 2CAN029505 i Page 66 of 89 )
8.0 CONCLUSION
S Entergy Operanons has performed an extensive investigation into the circumferential cracidng t occurring at ANO-2. 'Ihe investiganon includes comprshensive inspections, applicanon of apprcpiate safety factors, use of statistically valid (95/95) material' properties, degradanon i projections, and tube burst test data. Probabihsuc evalushnna were pelformed to assess the l validity for continued operation for the unit, and concluded that the ANO-2 steam generators can safely operate for the remainder of the fuel cycle with an acceptable level of risk. The : acceptable level is based on not comprennsing the structural mtegnty margin requirements for the tubing and a mmimal change in the SGTR-related core damage fmquency The evaluations also considered the previous large flaws found at ANO-2 during na_& Based on a combination of analytical evaluations, in-situ pressure tests, laboratory testing, finite element modeling, and use of advanced hTE techniques (UT and RPC deconvolution analysis), all of the circumferential cracks previously femd would have met the RG 1.121 structural margin requirements. . l t [ l 1
.I l
d'" Attadunant to -
. 2CAN029505 ~
- Page 67 of 89
~ 9.0 ~ REFERENCES
- 1) ' Information Update ANO-2 Steam Generator Tube Leak Outage, April 16,1992.
~ 2) Information Update Concerning the ANO-2 Steam Generseors, August 26,1992. !
3)- Information Update Conceming the ANO-2 Steam Generators, December 3,~ 1992. !
- 4) Information Update Conceming the ANO-2 Steam Generators,' August 30,1993. !
- 5) Information Update Conceming the ANO-2 Steam Generators, July 14,1994. ,
- 6) "PWR Steam Generator Tube Repair Lmuts Techmcal Support Document for ;
Expansion Zone PWSCC in Roll Transitions," EPRI NP-6864L - Rev. 2.
- 7) Summary of Meetmg on July 15,1993 C===i-g Eddy Current Methodologies Used :
at Arkansas Nuclear One, Unit 2 (ANO-2), dated August 6,1993.
- 8) "Circumferential IGSCC: UT 'Ihe Only Choice?," by D. Dobleni and D. Degreve, i presented at the 12th EPRI Steam Generator NDE Workshop, Park City, Utah, August 1993.
- 9) "Some NDE, I.mak Rate and Burst Pressure Considershons in Support of the Analysis !
of Circumferential Cracking Incidents at ANO Unit 2," by S.D. Brown and J.A. ~ ; Begley, Packer F=g!=es g Report No. B51677-RI, Revision 0, SveW 1994. ;
- 10) " Coil Design and Technique Parameter Development for Early Detechon of O.D.
Crackmg," by Jeff Siegel and Marty Klatt, ZETEC, Inc., presented at the EPRI 13th SGNDE Workshop, July 25-27,1994. I1) NRC Inspection Report 50-313/94-17; 50-368/94-17, Section 6.2.3, July 13,1994, i
- 12) " Statistical Evaluatum of ANO-2 Circumferential Crack Data,". Tetra Engineering :
Group, Inc., Report TR-94-7201, January 1994. 1 i
- 13) " Statistical Analysis of Steam Genrrator Tube Degradatiari," EPRI NP-7493.
- 14) "ANO-2 Steam Generator Tube Life Prediction Analysis," Dominion Er.girs..g, j Inc., February 1993, Revision 1. :
- 15) "ANO-2 Steam Generator Tube Burst Pressures for Circumferential Defects at Top of Tubesheet," MPR Associates, November 11,1994. i
- 16) "ANO-2 Steam Generator Degraded Tube Analysis Per Reg. Guide 1.121," ANO !
Engineering Report 92-R 2025-01. _
- 17) Packer Engmeering memo dated June 3,1993, to A. Buford.
[ 5
Attachmentb
' 2CAN029505 Page 68 of 89
- 18) Westinghouse Owners Group /NRC SG Circumferential Cracking Issue Meeting, September 26,1994.
- 19) " Final Report for the Steam Generssor Tube In-Situ Hydrostatic Test Tool," CE Report TR-ESE-1030.
l
- 20) . "In-Situ Pressure Test Informaton," CE Memo A-PENG-94-025, November 20, 1994.
- 21) " Tasks 3 and 6 Report: Effects of Chemical Cleanirig on Steam Generator Defects; Chemical Cleaning Quah5caten Testag for ANO-2," CE Report TR-MCC-274, Revision 0, December 1,1993.
~
'22) " Arkansas Nuclear One Unit 2 Leaker Outage Steam Generator Tube Repair Summary (Final Report)," B&W Report.
23). "ANO-2 Eddy Current Data Evalusten and leak Rate Estimaton," Wsgham Report, NSD-JLH-3347, SG-93-09-011, CARK-501, September 20,1993.
- 24) "NRC Integrated Program for the Resolution of Unresolved Safety Issues A-3, A-4, and A-5 Regarding Steam Generator Tube Integnty, NUREG-0844, September 1988.
- 25) " Voltage-Based Interim Plugging Criteria for Steam Generator Tubes - Task Group Report," DraA NUREG-1477, June 1,1993.
- 26) "ANO-2 SG Inspection Interval Risk Analysis Based on !& Data through 2R10," Calc. 93-E-0079-05, Rev. O.
- 27) "ANO-2 Probabilistic Risk Assessment (PRA), Individual Plant Exi.rJ.r.: ion (IPE)
Submittal," Report 94-R-2005-01, Revision 0.
- 28) ANO-2 Safety Analysis Report, Amendnunt 12.
- 29) " Interim Guidance on StaffImplementation of the Commission's Safety Goal Policy,"
Letter form J.M. Taylor (NRC) to the Commissioners, SECY 91-270, August 27, 1991.
- 30) " Severe Accident Issue Closure Guidelines," NUMARC 91-04, January 1992.
Attachment to i 2CAN029505 Page 69 of 89 Appendix A DataTables i Steam Ger-.ius "A" TOTAL 2L91 2B1 2E22-J. 2811 222E1
- of Cracks 620 208 17 45 147 203 Crack Imagth Average = 97.04 115.79 102.76 90.56 85.57 86.98 Std. Dev. = 54.31 61.43 63.85 54.45 50.60 42.17 Mimmum = 4 4 30 21 18 25 i
Maximum = 360 360 239 252 360 304 i Max %TW > Average = 71.33 72.73 71.29 71.76 71.56 69.66 Std. Dev. = 21.50 20.34 15.08 17.09 21.% 23.60 Muumum = 0 2 41 27 0 1 Maximum = 100 100 88 97 97 100 i Average %TW i Average = 19.77 24.30 21.48 17.89 17.52 17.02 l Std. Dev. = 13.57 15.95 15.93 11.60 12.45 10.41 Minimum = 0.00 0.33 5.24 4.48 0.00 0.15 Maximum = 96.88 89.00 57.76 51.31 %.88 68.40 I i
Attachment O 2CAN029505 Page 70 of 89 Steam Generator"B" TOTAL 2.D1 2B1 2Ha:1 2R11 2DE::1
# of Cracks 125 11 8 3 23 80
. Crack langth Average = 75.51 82.91 63.13 67.27 56.98 81.36 - Std. Dev. = 44.62 93.51 40.18 41.58 20.% 39.36 - Minimum = 25 35 30 30 25 30 Maximum = 360 360 139 112 93 220 ! i Max %'IW Average = 69.80 64.55 58.50 52.33 63.61 74.09 : Std. Dev. = 23.50 22.63 20.44 36.96 22.21 23.13 Mmimum = 2 9 24 30 17 2 Maximum = 100 91 84 95 93 100 - Average %TW
- Average = 15.01 15.% 11.83 7.51 10.17 16.87 Std. Dev. = 11.21 21.68 10.90 2.02 5.09 10.31 Muumum = 1.00 1.28 2.00 5.33 1.49 1.00 Maximum = 80.00 80.00 32.43 9.33 20.44 47.27 b
i f I 6 f
. . . .- . . . - - . ~ . - .
~ AttachmentCD 2CAN029505 Page 71 of 89 l l l l
- Both Steam Generators l
IDIAL 151 221 2Dh1 2Rif 1HE1
# of Cracks 745 219 25 48 170 283 Crack IAngtb {
Average = 93.42 114.14 90.08 89.10 81.66 85.39 Std. Dev. = 53.39 63.53 59.54 53.68 -48.61 41.40 Mmimum = 4 4 30 21 18 25 ! Maximum = 360 360 239- 252 360 304 Max %TW < Average = 71.07 72.32 67.20 70.54 70.45 70.91 , Std. Dev. = 21.84 20.48 17.62 18.82 22.10 23.51 Minimum = 0 2 24 27 0 1 Maximum = 100 100 88 97 97 100 Avera8e %1W ; Average = 18.97 23.88 18.39 17.24 16.53 16.98 Std. Dev. = 13.31 16.32 15.00 11.52 11.99 10.36 , Mmimum = 0.00 0.33 2.00 4.48 0.00 0.15 Maximum = %.88 89.00 57.76 51.31 96.88 68.40 , i i s 1 l l I l l l
Awkmant O 2CAN029505 Page 72 of 89 Circumferential Crack Data OUTAGE SG ROW IJNE VOLTS MAX LENGTH- AVG
%TW %1W 2F92 A 63 59 1,77 2 60 0.33 2F92- A 114 72 1.53 4 44 0.49 2F92 A 55 65 1.76 87 4 0.97 2F92 A 102 72 3.69 83 5 1.15 2F92 B 100 106 4.43 9 51 1.28 2F92 A 18 34 1.75 11 54 1.65 2F92 A 110 62 1.87 12 75 2.50 2F92 A 60 66 3.84 91 11 2.78 2F92 A 67 77 1.90 19 58 3.06 2F92 A 66 58 2.27 20 56 3.11 2F92 A 32 50 2.98 20 68 3.78 2F92 A 113 71 1.06 34 40 3.78 2F92 A 37 37 5.86 82 19 4.33 2F92 A 36 50 1.43 21 81 4.73 2F92 A 8 16 1.15 20 88 4.89 2F92 .A 46 54 1.70 24 74 4.93 2F92 A 50 38 1.05 33 54 4.95 2F92 A 34 42 2.12 37 56 5.76 2F92 A 105 63 1.12 41 54 6.15 2F92 A 101 79 4.12 87 28 6.77 2F92 B 103 65 2.56 60 42 7.00 2F92 A 81 127 2.51 83 33 7.61 2F92 A 45 49 1.60 86 32 7.64 2F92 A 14 152 1.27 45 62 7.75 2F92 A 63 77 2.34 20 140 7.78 2F92 A 41 39 0.98 61 47 7.96 2F92 A 76 80 1.71 82 35 7.97 2F92 A 107 71 1.94 66 44 8.07 2F92 A 11 45 1.35 55 54 8.25 2F92 B 106 82 2.66 86 35 8.36 2F92 B 84 42 2.82 69 44 8.43 2F92 A 86 124 8.38 73 42 8.52 2F92 A 95 119 5.07 96 32 8.53 2F92 A 15 31 1.04 35 90 8.75 2F92 A 5 53 1.94 44 72 8.80 2F92 '. 113 67 3.18 43 74 8.84 2F92 B 104 60 4.08 49 65 8.85 2F92 A 71 73 0.89 61 53 8.98 2F92 B 104 68 , 2.19 54 60 9.00 2F92 B 105 89 3.34 80 42 9.33 2F92 A 24 34 2.87 43 79 9.44 2F92 A 64 76 1.97 51 67 9.49
n . . . -_ .-. - . . .. _ r
- AttachmentCD -
2CAN029505 j Page 73 of 89 l OUTAGE SG ROW IJNE VOLTS MAX LENGTH AVG i
%TW ~%TW
'2F92 A 13 151 1.02 61 57 9.66 .
2F92 B 84 42 3.01 66 53 9.72' ? 2F92 A 23 53 3.21 40 88 9.78 ; 2F92 A 32 46 1.10 55 65 . 9.93 2F92 A 90 122 1.60 45 80 10.00 2F92 A 34 98 2.24- 78 47 10.18 . 2F92- A 27 39 1.79 51 72 10.20 ~ ! 2F92- A 56 54 3.22 36 102 10.20 2F92 A 67 81 2.18 81 46 10.35 2F92 A 36 38 2.20 72 53 10.60 2F92 A 106 82 1.45 48 83 11.07 2F92 A 36 44 1.29 79 54 11.85 2F92 A 21 149 6.30 74 58 11.92 2F92- A 81 63 1.35 85 51 12.04 ; 2F92 A 88 46 0.93 55 79 12.07 ! 2F92 A -14 28 1.21 41 107 12.19 2F92 A 13 27 2.74 72 61 12.20
-2F92 A 87 43 2.22 85 53 12.51 ;
2F92 A- 87 43 5.20 89 51 12.61 : 2F92 A 94 46 1,63 65 70 12.64 l 2F92 A 105 77 1.22 73 63 12.78 2F92 A 46 74 0.95 55 86 13.14 1 2F92 A 9 151 6.95 77 62 13.26 2F92 A 32 130 3.21 58 83 13.37 ; 2F92 A 8 54 2.33 55 88 13.44 2F92 A 19 35 5.50 73 67 13.59 ' 2F92 A 60 68 4.39 93 53 13.69 2F92 A 107 81- 1.09 62 83 14.29 ! 2F92 A 111 95 4.12 87 60 14.50 i 2F92 A 70 74 2.48 39 135 14.63 l 2F92 A 4 38 1.65 53 100 14.72 ; 2F92 A 59 57 6.64 82 65 14.81 ! 2F92 A 30 114 2.07 89 60 14.83 , 2F92 B 113 99 3.81 91 61 15.42 ; 2F92 A 15 149 3.17 67 83 15.45 2F92 A 27 33 2.47 85 67 15.82 ! 2F92 A 32 20 2.75 85 67 15.82 l 2F92 A 34 24 1.66 60 97 16.17 I 2F92 A 62 64 5.10 97 60 16.17 ) 2F92 A 104 % 3.92 72 82 16.40 l 2F92 A 51 45 6.49 63 95 16.63 i 2F92 A 59 63 1.92 56 107 16.64 l 2F92 A 21 53 5.79 77 79 16.90
( J - ,. Anachmerd zo
'* # 2CAN029505 ,1 Page 74 of 89 -
JJ , ,. L : OUTAGE ' 6G1 ROW- LINE VOLTS MAX IENGUI AVG' , g %'IW %TW ' 2F92 'A' 65 '77 .l.54 63 98- 17.15 l i 2F92 .B .111 .73 2.06 66 99. .18.15 l 2F92 A 11 149 1.20 - 58 113- 18.21- i
~ 2F92 A 79 81 4.32 94 70 'I8.28 .l 2F92 A -31 41 3.80 65 102 '
18.42 ; 2F92 'A '63 111 2.00 42 158 -18.43 l 2F92 A 71 107 2.54 - 85 79 18.65 j 2F92~ A 37 39 7.07 80 84 18.67- .i 2F92 A 41 33 1.18 54 125 18.75 : 1 2F92 A 41 35 6.65 97 70 18.86 .I 2F92 A 13 153 0.99 50 137 19.03 ! 2F92 'A 32 124 7.78 92 75 19.17 g 2F92 A 113 -73 4.07 76 91 19.21 2F92 A -100 122 6.67 73 95 19.26 i 2F92 A 106 98 5.88 95 73 19.26 l 2F92 A 17 151 1.71 67 104 19.36 l 2F92 A 101 - 73 2.34 87 81 19.58 2F92 A 78. 130 8.19 94 75 19.58 2F92 A % 46 2.50 85 84 19.83 ! 2F92- A 84 126 4.17 56 128 19.91- 1 2F92 A 13 17 5.03 87 84 20.30 l 2F92 A 28 112 1.72 53 140 20.61 ( 2F92 A 27 35 4.05 73 . 102 20.68 ~! 2F92 A 79 79 1.74 90 83 20.75 2F92 A _68 32 6.58 87 86 20.78 3 2F92 A' 31 123 4.70 90 84 21.00-2F92 A 5 55- 2.69 ' 73 104 21.09 ! 2F92 A 68 103 1.29 61 125 21.18 l 2F92 A 28 126 3.46 77 100 21.39 i 2F92 A 43 119 2.13 83 93 21.44- l 2F92 A 108 74 2.59 79 98 21.51 l 2F92 A 31 51 2.41 81 97 21.83 ; 2F92 A 21 151 0.91 69 114 21.85 2F92 A 46 58 3.84 83 43 9.91 2F92 A 55 77 9.57 82 % 21.87 t 2F92 A 55 77 5.08 54 146 21.90 j 2F92 A 105 67 2.14 79 102 22.38 ; 2F92 A 46 4% 7.44 87 93 22.48 : 2F92 A 51 H. 6.87 73 111 22.51 1 2F92 A 82 80 2.46 73 112 22.71 2F92 A 25 131 3.41 86' '97 23.17 ! 2F92 A 47 121- 6.32 87 % 23.20 ! 2F92 A 70 78 3.30 67 125 23.26' ! t
. _ ~ - --_ _ _ _ - _ _ - ._ .- --
.. - . . -- ~. _ _
y 2CAN029505 : Page 75 of 89 l OUI' AGE SG ROW LINE VOLTS MAX LENGTH AVG-
%TW ' %7W 2F92 A 16 148 1.41 63 134 23.45 2F92 A' 35 127 2.42 80 106 23.56 +
2F92 A 44 46 2.39 68 125 23.61 2F92 A 101 69 4.27 92 93 23.77 l 2F92 A 92 46 10.02 87 100 24.17 2F92 A 51 119 14.30 75 118 24.58 l 2F92 A 59 65 1.59 86 104 24.84 2F92 A 32 48 4.90 94 97 25.33 -l 2F92 A 12 146 3.39 61 150 25.42 i 2F92 A 20 52 1.64 79 116 25.46
~
2F92 A 57 65 5.00 73 126 25..,5 2F92 A 15 145 5.65 91 102 25.78 2F92 A 68 76 10.13 78 123 26.65 ! 2F92 A 97 121 8.49 75 128 26.67 ! 2F92 A 55 53 20.30 81 121 27.23 , 2F92 A 107 69 2.27 79 128 28.09 > 2F92 A 52 106 5.35 71 143 28.20 [ A 1,66 2F92 18 138 83 125 28.82 2F92 .A 11 151 11.39 82 128 29.16 j 2F92 A 45 119 1.34 75 140 29.17 ! 2F92 A % 118 16.64 92 115 29.39 l 2F92 A 109 77 7.14 92 116 29.64 ' 2F92 A 47 49 3.27 83 130 29.97 ; 2F92 A 19 137 7.78 90 120 30.00 ; 2F92 A 58 56 10.50 88 123 30.07 , 2F92 A !.08 94 9.19 76 143 30.19 : 2F92 A 26 52 5.71 90 121 30.25 i 2F92 A 71 105 17.46 75 146 30.42 ; 2F92 A 83 127 9.97 83 133 30.66 ! 2F92 A 22 146 1.37 75 148 30.83 2F92 A 11 19 2.98 77 146 31.23 . 2F92 A 28 128 7.99 81 139 31.28 i 2F92 A 42 118 3.26 81 139 31.28 ! 2F92 A 49 49 3.51 85 135 31.88 A 2F92 103 109 5.76 87 132 31.90 ' 2F92 A 64 68 4.20 95 121 31.93 2F92 A 108 76 6.79 83 139 32.05 2F92 A 102 106 10.55 89 130 32.14 2F92 A 93 45 6.52 86 135 32.25 2F92 A 105 69 4.90 87 135 32.63 2F92 A 106 64 9.62 91 130 32.86 I 2F92 A 106 64 6.16 87 137 33.11 2F92 A 23 133 9.05 89 134 33.13
Asch=aat to r 2CAN029505 - : Page 76 of 89 OUTAGE SG ROW LINE VOLTS adAX LENGTH AVG i
%TW %TW '~
2F92 A 108 72 2.39 72 170~ 34.00 - 2F92 A 106 76 13.95 84 149 34.77
- 2F92 A 105 61 6.24 95 132 34.83 2F92 A 62 66 5.93 100 126 35.00 2F92 A 103 107 6.78 85 156 36.83 2F92 A 60 112 9.92 94 142 37.08 2F92 A 56 62 7.21 84 160 37.33 !
2F92 A 13 149 3.01 72 188 37.60 2F92 A 52 $4 3.09 73 188 38.12 2F92 A 106 96 9.84 68 202 38.16 , 2F92 A 88 126 9.94 80 172 38.22 2F92 A 21 135 6.78 88 157 38.38 2F92 A 15 147 17.59 78 178 38.57 2F92 A 14 144 5.35 84 167 38.97 ' 2F92 A 63 69 4.29 84 167 38.97 2F92 A 15 143 3.05 80 176 39.11 2F92 A 104 108 19.90 77 186 39.78 2F92 A 12 28 9.93 67 221 41.13 2F92 A 20 136 4.04 85 183 43.21 2F92 A 45 121 14.89 86 181 43.24 . 2F92 A 97 115 1.90 82 194 44.19 2F92 A 106 68 16.37 87 184 44.47 2F92 A 107 73 6.58 88 183 44.73 2F92 A~ 63 61 4.39 89 183 45.24 2F92 A 17 147 1.51 81 202 45.45 2F92 A 30 120 1.70 94 175 45.69 2F92 A 28 130 7.77 90 183 45.75 t 2F92 A 31 127 14.42 84 197 45.97 2F92 A 29 129 5.21 81 209 47.03 2F92 A 110 74 21.86 80 212 47.11 2F92 A 52 52 18.11 88 193 47.18 ; 2F92 A 101 109 8.48 7' 218 47.23 l 2F92 A 101 109 6.80 87 196 47.37 : 2F92 A 31 125 6.92 95 183 48.29 2F92 A 104 78 6.14 57 309 48.93 ! 2F92 A 30 124 12.69 100 180 50.00 l 2F92 A 66 106 2.27 86 218 52.08 I 2F92 A 51 51 4.87 91 207 52.33 l 2F92 A 14 142 6.80 88 225 55.00 2F92 A 34 40 36.46 83 240 55.33 2F92 A 52 116 3.37 91 230 58.14 2F92 A 110 68 12.81 88 238 58.18 2F92 A 28 132 19.44 82 257 58.54
Q ( i, Attachment to 2CAN029505 ; [7 Page 77 of 89 OUTAGE SG' ; ROW - LINE VOLTS MAX LENGTH AVG
- c. %TW %TW ;
- , 2F92 A . 44 - 48 3.30 91- 270 68.25 !
2F92 B 36 130 7.09 80 360 80.00 l A 2F92 68 78 13.10 89- 347 85.79 i 2F92 A 55 63 39.80- 88 360 88.00 1 2F92 A 13 147 80.29 89 360 89.00 ; 2R9 B 92 .122 0.95 24: 30 2.00 , 2R9 B 18 18 1.16 40 41 4.56 - l 2R9 B 108 64 0.51 48 36 4.80 ! 2R9 A 93 117 1.39 41 46 5.24 ; 2R9 A 47 57 1.18 43 55 6.57 ! , 2R9 A % 120 2.39 49 51 6.94 j 2R9 A 50 50 2.31 88 30 7.33 ! 2R9 B 66 136 0.63 66 43 7.88 ; 2R9 A 57 51 1.46 76 39 8.23 j 2R9 B 25 147 1.48 69 46 8.82 ; , 2R9 B 84 80 1.13 57 57 9.03 { 2R9 A 60 64 0.55 69 51 9.78 j 2R9 A 104 110 1.36 76 58 12.24 2R9 A 11 17 1.66 74 78 16.03 l 2R9 A 98 122 1.80 74 92 18.91. 2R9 A 58 62 0.56 -83 88 20.29 ! 2R9 A 69 131 2.76 69 107 20.51 I 2R9 A 95 117 2.63 62 123 21.18 2R9 B 101 65 2.61 80 113 25.11 i 2R9 A %- 116 0.83 67 148 27.54 2R9 B 108 86 3.38 84 139 32.43 ! 2R9 A 102 110 3.32 88 143 34.% j 2R9 A 79 83 0.99 87 181 43.74 ! 2R9 A 67 79 2.78 79 218 47.84 ; 2R9 A 64 48 6.34 87 239 57.76 ! 2P93-1 A 111 91 0.89 46 35 4.48 -l 2P93-1 A 24 38 1.55 35 53 5.15 ; 2P931 B 90 48 0.83 32 60 5.33 ; 2P93-1 A 98 86 3.87 97 21 5.66 l 2P93-1 A 45 45 0.91 66 33 6.05 l 2P93-1 A 53 79 1.25 36 65 6.50 l l 2P93-1 A 114 90 1.2 74 35 7.20 2P93-1 A 52 118 0.95 27 % 7.23 2P93-1 A 101 77 2.79 70 39 7.50 2P93-1 A 54 50 1.79 85 33 7.79 ! 2P93-1 B 66 128 3.04 95 30 7.86 2P93-1 A 33 129 1.46 56 54 8.40 ;
~
2P93-1 A 53 51 1.69 75 43 8.96 i
. i
7.
~ '
, ,,,0 2CAN029505
, Page 78 of 89 ;-
[ 9 OUTAGE SG. -ROW IJNE VOLIS MAX LENGTH AVG
%1W %1W 2P93 'A 114 96 1.19 66 51 9.32 2P93-l_ B 82 44 0.97. .30 112 9.33 2P93-1 A 62 50- ,1.93~ 92 37 9.46 2P93 A. 49- 115 2.26 58 63 10.15 2P93-1: A 54 52 1.83 67 58. 10.79
- 2P93-1 A 27 127 3.14- 90 44, 11.00 2P93 A 15 33 2.42 87 49. I1.84 2P93-1 .A 30 126 2.09 70 63 12.25 3 2P93-1 A 10 130 2.56 85 ~$3 12.41 2P93-1 A 105. 91 0.91 78 '58 12.53 2P93-1 A 103- 75 2.59 82 60 13.57 2P93-1 A 64 108 2.11- 48 102 13.60 2193 1 A 101 81 4.54 86 58 13.81' 2P93-1 'A 104 88 2.4 90 56 14.02 2P93-1 A 106 90 0.87 45 117 '14.68 2PP3-1 A 27 131 2.80 60 95 15.83 2P93-1 A 31 129 1.% 60 102 '16.94 '
2P93-1 A 34 130 2.55 80 81 18.00 2P93-1 A 72 130 4.27 87 75 18.13 2P93-1 A 30 52 2.% 84 79. 18.43 2P93-1 A 77. 77 2.22 72 % 19.27 2P93-1 A- 20 34 1.21 74 100 20.56 2P931 A 104 90 5.76 88 84 20.56 2P93-1 'A 4 56 1.74 57 144 22.80 2P93-1 A 111 79 3.09 89 105 25.99 2P93-1 A 105 89 1.67 83 121 27.87' 2P93-1 A 101 85 2.94 77 137 29.30 2P93-1 A 109 73 1.19 67 159 29.59 2P93-1 A 106 102 2.11 80 137 30.44 2P93-1 A 77 79 2.64 50 252 35.04 2P93-1 A 16 30 3.72 75 179 37.29 2P931 A 100 56 1.47 75 194 40.52 2P93-1 A 42 46 1.72 76 193 40.74 2P93-1 A 35 45 4.8 93 163 42.11 2P93-1 A 82 126 5.92 91 -203 51.31 2R10 A 6 110 2.7 93 96 24.90 2R10 A 8 34 1.0 N/A 33 20.00 2R10 A 8 40 2.0 88' 95 23.13 2R10 A 12 30 7.1 86 130 30.97 2R10 A 13 29 3.6 91 49 12.40 2R10 A 15 29 8.0 94 163 42.55 2R10 A 13 145 0.7 41 72 8.18 2R10 A 17 29 1.8 87 75 18.21
p. o 3 2CAN0295J5 Pase 79 of 89 - OUTAGE SG ROW LINE VOLTS MAX LENGTH AVG
%TW %1W , 2R10 .A 17 31 1.2 71 110 21.77 2R10 'A 17 139. 1.5 48 165 21.96 2R10 A 19- 53 4.7 91' 95 23.92 2R10' A 20 138 3.7 89 . 56 - 13.86-2R10 A 20 150 2.4 82 68 15.57.
2R10' A 21 19 2.9 89 .70 17.33 2R10 'A 21 29 2.8 72 142 28.39 2R10 .A 21 137 2.7 76 184 38.84 2R10 A 22 54 . 4.0 89 72 17.76 2R10 A 22 -114 2.1 39 54 5.88 2R10 A 22 150 2.4 72 135 26.98 2R10 A 23 ' 39 2.1 88 72 17.56 2R10 A 24 130 2.0 84 138- 32.30 2R10 A 24 132 9.3 90 205 51.25 2R10 A 25 133 4.0 87 124 30.06 2R10 A 27 37 1.6 78 79: 17.08 2R10 A 29 39 4.0 88 124 30.41 2R10 A 31 39 7.8 88 142 34.69 2R10 A 31 43 3.6 88 98 23.99 2R10 A 32 40 5.1 97 67 17.94 2R10 A 32 42 1.6 72 84 16.82 2R10 A 32 122 7.4 90 151 37.67 2R10 A 32 126 6.8 97 360 96.88 2R10. A 33 43 1.5 86 35 '8.37 2R10 A 33 53 2.7 71 81 15.90 2R10 A 33 123 0.9 71 154 30.41 2R10 A 33 125 2.6 55 105 16.06 2R10 A 34 48 1.4 71 42 8.29 2R10 A 34 58 1.7 75 67 13.87 2R10 A 35 41 4.2 84 N/A 20.00 2R10 A 35 53 2.9 67 89 16.63 2R10 A 36 42 1.1 74 37 7.56 2R10 A 36 128 1.9 93 112 28.97 2R10 A 37 35 2.3 76 63 13.32 2R10 A 37 43 1.7 N/A 65 20.00 2R10 A 38 36 1.8 83 47 10.91 2R10 A 38 38 1.0 83 88 20.20 2R10 A 38 40 7.3 94 109 28.37-2R10 A 38 42 1.7 86 56 13.39 2R10 A 38 44 4.5 92 63 16.12 2R10 A 44 118 1.4 80 42 9.34 2R10 A 46 42 3.3 87 75 18.21 2R10 A 48 48 1.5 70 56 10.90
Nu ' +
, Asenchment to
- - 4 ' 2CAN029505 ' "' , Pate 80 of 89 j
-)
. OUTAGE. 4 SG ROW IJNE VOLTS MAX LENGTH AVG !
j,-- . %1W %'IW j 2R10 A, 48 50 4.2 79 217 47.68 l
;2R10 A 48 120 1.5 58 63 10.16 .;
2R10 -- A '49 41 0.4 51 70 9.93 l - ~2R10- A 49 107 - 1.8 N/A 142 20.00. i 2R10 A 50 144 0.7 50 30 4.14 1 2R10 A 51 43 N/A N/A N/A 20.00
.2R10 A 51 113 1.2 67 65- 12.07 . j 2R10 A 51 115 0.8 61 42 7.13 [
2R10 A 52 104 1.0 48 42 5.61 -l 2R10 A 52 120 1.6 61 % 16.33 ! 2R10 A 54 64 3.0 82 68 15.57 ! 2R10 A 54 106 2.2 N/A 65 20.00 ! 2R10 A 55 105 1.2 40- 39 4.28 l 2R10 A 57 63 2.1 73 98 19.90 i 2R10 A 58 54 1.2 33 39 3.53 j 2R10 A 58 66- 9.8 % 96 25.70 1 2R10 A 59 55 1.8 41 88 9.98 ; 2R10 A 59 113 1.0 25 54 3.77 j 2R10 A 59 115 2.5 97 46 12.27 l 2R10. A 61 133 0.5 70 18 3.41 i 2R10 A 62 60 1.9 50 67 9.25 i 2R10 A 62 82 1.5 50 98 13.63 -l 2R10 A 63 63 1.3 24 74 4.91 i 2R10 A 63 137 1.5 85 21 4.% . ! 2R10 'A 64 60 3.2 ~ 79 116 25.38 ; 2R10 A 65 61' l.5 49 37 5.01- l 2R10 A 65 107 1.6 60 79 13.14 j 2R10 A 65 109 1.2 32 51 4.52 1 2R10 A 66 60 1.1 59 228 37.33 2R10 A 66 110 5.7 86 47 11.30 i 2R10 A 67 61 1.6 58 149 23.99 ! 2R10 A 67 63 0.9 35 149 14.48 ! 2R10 A 67 107 1.4 62 179 30.78 : 2R10 A 68 66 4.0 95 84 22.19 ! 2R10 A 68 104 2.0 75 126 26.28 ; 2R10 A 70 106 2.5 85 121 28.55 l 2R10 A 71 35 9.2 81 223 50.07 l 2R10 A 71 103 1.7 0 98 0.00 ; 2R10 A 71 111 2.7 92 47 12.09 ; 2R10 A 73 41 1.5 81 33 7.49 2R10 A 74 36 1.8 3 51 0.42 l 2R10 A 75 37 4.0 94 61 16.01 l 2R10 A 75 79 3.6 76 68 14.43
y .. .. V , fAttachment to . 2CAN029505 Page 81 of 89 I I OUTAGE- SG ROW LINE VOLTS MAX LENGTH AVG
%1W %TW ,
i 2R10 A 75 81 1.6 3 63 0.53 2R10 A 75 133 1.0 66 26 4.82 ; 2R10 A 76 36 1.0 84 39 8.99 2Rio A 76 130 3.2 84 117 27.39 l 2R10 A 77 129 4.2 91 91 23.03
- 2R10 A 77 131 3.3 94 49 12.81 !
2R10 A 78 80 1.2 54 84 12.62 2R10 A 79 127 1.1 48 46 6.07 i 2R10 A 82 82 7.0 90 131 32.85 ; 2R10 A 84 84 3.6 90 46 11.39 l 2R10 A 87 125 6.7 82 154 35.12 : 2R10 A 88 44 3.4 89 51 12.56 : 2R10 A 89 125 0.8 73 39 7.82 , 2R10 A 90 124 2.1 79 40 8.84 ! 2R10 A 93 85 3.1 76 79 16.65 ! 2R10 A 94 84 11.2 88 123 29.98 l 2R10 A 95 115 4.7 92 63 16.12 l 2R10 A % 52 7.1 87 96 23.29 l 2R10 A 97 45 0.7 63 46 7.97 2R10 A 97 51 1.3 68 44 8.27 2R10 A 97 79 2.3 88 39 9.42 ; 2R10 A 97 83 2.9 90 86 21.46 - 2R10 A 98 118 4.8 86 67 15.91 l 2Rio A 99 55 1.5 44 91 11.14 : 2R10 A 100 84 2.1 73 72 14.57 2R10 A 102 78 6.7 97 131 35.41 j 2R10 A 102 86 2.5 80 39 8.57 ! 2R10 A 102 92 1.5 4 103 1.15 2R10 A 102 108 5.5 78 131 28.47 2R10 A 103 61 1.2 63 28 4.91 ! 2R10 A 103 77 2.4 97 28 7.55 l 2R10 A 103 81 1.4 46 51 6.49 j 2R10 A 103 83 1.4 69 51 9.74 i 2R10 A 103 85 1.6 61 128 21.67 2R10 A 103 87 0.7 33 21 1.93 2R10 A 104 74 2.1 95 42 11.10 2R10 A 104 76 1.3 65 53 9.49 ) 2R10 A 104 84 2.8 60 217 36.21 2R10 A 105 65 1.4 74 46 9.36 2R10 A 105 75 1.4 91 137 34.55 2R10 A 105 87 1.2 58 88 14.11 2R10 A 105 109 1.1 71 79 15.55 2R10 A 106 60 1.3 1 81 0.22 I ____._i
- 7. .
L Attachment'to
^-
- 2CAN029505
- Page 82 of 89 OUTAGE _ SG ROW LINE VOLTS MAX LENGTH AVG )
%'IW %TW l I
2R10 'A 107 61 2.4 87 51 12.28 j
'2R10 A 107 83 2.9 91 82 20.82 2R10 A -107 85 1.1 57' 60 9.43 ,
2R10 A 108 70 1.2 70' 42 8.18 i 1- - 2R10 A 108 96 5.4 97 58 15.58 ( 2R10 A 108 102 2.1 82~ 63 14.37 ! 2R;0 A 111 73 - 2.1 63 147 25.76 2R10 A 111 75 1.3 48 32' 4.21 2R10 A 111 107 3.6 83 93 21.41 2R10 A 112 92 1.8 73 70 14.21 ; 2R10 B 18 150 3.3 91 53 13.29 2R10 B 18 152 1.7 91 67 16.83 .j 2R10 B 20 18 1.4 17 32 1.49 ! 2R10 B 20 114 1.1 50 35 4.87- I 2R10 B 24 146 1.1 83 63 14.54 I 2R10 B 40 126 2.5 93 25 6.34 l 2R10 B 42 66 1.0 57 93 14.70 ; 2R10 B 43 135 3.0 80 46 10.12 ! 2R10 B 71 55 1.8 21 47 2.76 l 2R10 B 83 95 2.0 44 SL 7.07 ! 2R10 B 93 121 0.9 62 37 6.34 l 2R10 B 94 116 2.0 90 29 9.64- ! 2R10 B 97 115 2.1 41 44 4.99 ; 2R10 B 98 50- 0.8 52 _65' 9.36 , 2R10 B 101 53 0.6 53 28 4.13 2R10 B 102 62 0.8 - ~67 49 9.13 2R10 B 104 70 1.6 69 89 17.13 l 2R10 B 106 84 0.7 35 89 8.69 2R10 B 106 98 1.0 57 65 10.26 'l' 2R10 B 107 83 2.3 84 88 20.44 2R10 B 107 85 3.1 71 72 14.17 .l 2R10 B 110- % i.5 74 81 16.57 , 2R10 B 114 72 7.4 81 49 11.04 2P95-1 A 46 52 1.1 83 41 9 2P95-1 A 50 52 0.72 66 90 17 l 2P95-1 A 32 66 1.8 92 39 10 ! 2P95-1 A 57 55 1.38 72 74 15 ! 2P95-1 A 57 57 2.32 98 48 13 l
' 2P95-1 A 53 63 5.9 88 88 22 j 2P95-1 A 61 67 1.39 45 55 7 1 2P95-1 A 60 60 1.92 59 83 14 2P95-1 A 64 62 1.13 37 65 7 2P95-1 A 54 62 0.77 58 53 9 l q
l
Attachment to 2CAN029505 Page 83 of 89 OUTAGE SG ROW IlNE VOLTS MAX LENGTH AVG
%TW' %TW i
2P95-1 A 66 64 3.27 93 129 33 . 2P95-1 A 56 66 0.75 79 42 9 2P95-1 A 64 66 1.3 90 48 12 2P95 A 23 51 2.22 11 122 4 2P95-1 A 10 152 0.69 3 62 1 2P95-1 A 22 152 0.62 39 51 6 2P95-1 A 24 52 2.99 42 143 17 '
~2P95-1 A 12 152 1.63 67 50 9
'2P95-1 A 22 32 2.03 89 170 42 2P95-1 A 16 54 2.03 74 106 22 .;
2P95-1 A 20 54 1.08 19 111 6 ; 2P951 A 17 153 1.89' 88 25 6 ! 2P95-1 A 27 '3 2.27 80 94 21 2P95-1 A 16 10 1.78 91 67 17 , 2P95-1 A 28 146 2.97 88 102 25 i 2P95-1 A 16 144 1.75 % 78 21 2P95-1 A 12 148 2.71 88 95 23 2P95-1 -A 23 135 2.48 81 304 68 2P95-1 A 19 139 1.06 60 138 23 2P95-1 A 11 147 0.97 50 163 23 2P95-1 A 66 138 1,62 95 55 15 2P95-1 A 61 137 2.01 94 71 ~19 , 2P95-1 A 57 141 3.64 99 74 20 2P95-1 A 90 118 0.64 89 50 12 2P95-1 A 90 126 1,66 97 57 15 2P95-1 A 82 132 1.11 1 40 1 2P95-1 A 68 136 1.01 25 75 5 2P95-1. A 72 132 0.85 67 117 22 2P95-1 A 96 122 1.61 88 113 28 2P95-1 A 21 111 0.86 69 92 18 ! 2P95-1 A 21 113 3,49 71 95 19 < 2P95-1 A 29 113 1.17 61 46 8 2P95-1 A 31 113 1.06 55 53 8 2P95-1 A 15 153 1.08 44 64 8 2P95-1 A 99 121 2.37 95 95 25 2P951 A 71 135 1.54 75 118 25 2P95-1 A 34 122 0.86 59 79 13 2P95-1 A 49 119 1.5 68 151 29 2P95-1 A 4 118 1.93 70 58 11 2P95-1 A 5 127 1.62 76 58 12 2P95-1 A 33 127 2.73 88 132 32 2P95-1 A 35 129 1.47 91 117 30 2P95-1 A 29 131 2.15 23 71 5 S
- Attachmentto 2CAN029505 Page 84 of 89
' OUTAGE SG ROW IJNE VOLTS MAX LENGTH AVG'
%TW %IW j 2P95-1 A 33 131 3.91- 97 .104 28 -
2P95-1 A 32 128 0.57 15 60 3 ; 2P95-1 A 30 130 1.57 1 55 0 2P95-1 A 32 132 3.35 92 67 17 ; 2P95-1 A 26 134 2.5 % 87 23 ; 1 2P95-1 A 20 134 7.11 91 134 34 2P95-1 A 24 134 4.4 97 143 39 ! 2P95-1 A 11 127 0.69 87 53 13 ! 2P95-1 A 43 131 4.2 99 48 13 ! 2P95-1 A 21 133 2.31 72 62 12 , 2P95-1 A 41 133 3.51 98 48 13 ! 2P95-1 A 24 126 2.74 79 58 13 2P95-1 A 14 128 3.84 72 85 l'7 2P95-1 A 8 130 0.8 65 64 12 I 2P95-1 A 71 69 2.67 92 58 15 2P95-1 A. 83 81 2.7 89 74 18 2P95-1 A 81 83 1.75 62 134 23 2P951 A 85' 83 2.14 46 112 14 , 2P95-1 A 72 .68 1.18 36 90 9 ! 2P95-1 A 88 68 1.77 45 101 13 2P95-1 A 72 80 2.05- 35 115 11 2P95-1 A 78 82 2.15 84 138 32 2P95-1 A 76 84 0.85 30 103 9 2P951 A 80 74 4.26 1 56 1 - 2P95-1 A 74 76 3.84 92 83 21 2P95-1 A 78 76 2.74 77 165 35 2P95-1 A 76 78 7.15 69 83 16 2P95-1 A 70 80 1.73 59 69 11 i 2P95-1 A 81 91 1.35 27 69 5 2P95-1 A 71 101 3.39 SO 71 16 2P95-1 A 75 65 1.95 76 64 14 2P95-1 A 70 82 2.66 38 83 9 2P95-1 A 69 83 1.4 64 66 12 i 2P95-1 A 72 102 2.1 89 94 23 2P95-1 A 76 98 1.77 14 92 4 - 2P95-1 A 80 88 1.36 47 102 13 L 2P95-1 A 76 100 1.01 39 43 5 2P95-1 A 95 45 2.86 84 120 28 ! 2P95-1 A 95 53 0.68 76 64 14 , 2P95-1 A 107 59 0.85 41 69 8 , 2P95-1 A 79 91 2.71 93 143 37 , 2P95-1 A 83 91 1.32 52 109 16 2P95-1 A 77 97 1.42 51 241 34 3
i L Attachment to G 2CAN029505 ; Page 85 of 89 ' i OUTAGE SG ROW IJNE VOLTS MAX LENGTH AVG
%TW %TW 2P95-1 A 97 53 1.14 53 51 8 !
2P95-1 A 98 48 1.2 71- 46 9 l 2P95-1 A 98 52 1.28 68 43 8 ! 2P95-1 A 92 44 1.84 81 120 27 2P95-1 A 103 79 1.88 79 64 14 2P95-1 A 107 79 1.97 81 97 22 2P95-1 A 95 81 0.75 77 83 18 2P95-1 A 107 103 5.05 92 62 16 : 2P95-1 A 111 83 0.91 56 81 13 2P95-1 A 104 68 1.97 83 76 18 ; 2P95-1 A 104 70 0.84 43 34 4 : 2P95-1 A 112 70 1.02 64 92 16 , 2P95-1 A 100 76 4.49 97 108 29 2P95-1 A 109 78 0.94 59 60 10 2P95-1 A 104 80 0.97 81 62 14 ;
-2P95-1 A 100 86 1,58 71 76 15 )
2P95-1 A 112 90 4.17 93 167 43 2P95-1 A 112 98 3.66 96 69 18 I 2P95-1 'A 97 77 2.05 82 115 26 2P95-1 A 93 83 1.7 99 123 34 ! 2P95-1 A 101 83 4.67 % 113 30 2P95-1 A 105 83 1.74 74 112 23 , 2P95-1 A 109 83 2.98 90 92 23 l 2P95-1 A 97 85 1.66 81 182 41 ! 2P93-1 A 101 87 0.77 66 98 18 i 2P95-1 A 97 95 1.26 60 88 15 { 2P95-1 A 110 70 1.38 87 51 12 2P95-1 A 106 72 2.8 97 76 20 2P95-1 A 110 96 1.04 70 155 30 l 2P95-1 A 114 100 2.62 93 44 11 1 2P95-1 A 106 94 1.9 86 95 23 > 2P95-1 A 98 80 3.29 92 150 38 I 2P95-1 A 102 80 6.88 98 136 37 2P95-1 A 106 84 0.75 68 74 14 2P95-1 A 110 84 3.75 97 78 21 2P95-1 A 98 92 2.77 85 104 25 2P95-1 A 67 95 1.12 75 46 10 i 2P95-1 A 52 88 1.3 83 43 10 l 2P95-1 A 68 94 0.99 40 39 4 2P95-1 A 42 74 2 37 150 15 2P95-1 A 50 74 1.15 69 46 9 2P95-1 A 66 78 1.5 37 115 12 2P95-1 A 46 72 0.75 89 53 13 l i
,L f Attachment to x 2CAN029505 5
4 , Page 86 of 89 l OUTAGE SG: ROW- LINE VOLTS MAX LENGTH AVG
%'IW %TW j 2P95 A 39 69 1.84 77 76 16 .{
2P95-1 A 82 110 3.18 86 46- 11 -i
- 2P95-1 A 26 124 0.58 58 41 7 l 2P95-1 A 29 125 0.58 50 58 8- 3 2P95-1 A 72- 78 1.21 82 69 16 ,
2P95-1 ~A 77 85 1.58- 59 133 22 l t 2P95-1 .A 68 82 1.81 73 133 27 ! 2P95-1 A 43 101 1.75 86 65 16 ! 2P95-1 A 34 116 0.71 42 48 6 'l 2P95-1 A 62 .I10 0.97 74 58 12 2P95-1 A 56 116 0.62 71 35 7 2P95-1 A 48 110 0.91 62 39 7 2P95-1 A $3 103 1,42 25 76 5 l 2P95-1 A 61 113 0.88 67 158 29 2P95-1 A 64 98 1.32 62 46 8 2P95-1 A 51 103 1.28 36 55 6 i 2P95-1 A 25- 35 2.17 94 124 - 32 I 2P95-1 A 21 37 2.26 95 -43 11 2P95-1 A 25 37 2.59 % 134 36 2P95-1 A 33 41 4.24 98 145 39 I 2P95-1 A 23 35 3.53 98 218 59 2P95-1 A 11 37 2 79 36 8 - 2P95-1 A 23 37 2.61 85 134 32 l 2P95-1 A 35 39 1.36 66 113 21 , 2P95-1 A 27 41 1.63 90 57 14' 2P951 A 11 43 11.17 54 67 10 2P95-1 A 32 34 1.53 97 64 17 ) 2P95-1 A 24 36 1.03 74 48 10 2P95-1 ~A 28 36 1.32 43 99 12 2P95-1 A 28 38 5.43 100 87 24 2P95-1 A 40 36 0.88 46 37 5 2P951 A 22 34 0.92 85 110 26 2P95-1 A 26 38 1 88 39 10 2P95-1 A 49 45 0.8 52 83 12 2P95-1 A 37 47 2.87 52 170 25 2P95-1 A .37 49 0.77 31 39 3 2P95-1 A 35 43 1.81 68 36 7 2P95-1 A 35 47 1.53 50 83 12 l 2P95-1 A 43 47 0.97 45 164 21 2P95-1 A 47 47 2.49 30 210 18 2P95-1 A 36 46 1.12 39 111 12 ; 2P95-1 A 52 46 3.24 99 51 14 i 2P95-1 A 34 46 2.31 68 152 29 , l l I
w. l Attachment to
>' l2CAN029505
' Page 87 of 89 .!
~
~
. OUTAGE SG_ ROW- IJNE VOLTS MAX LENGTH- AVG :
p %TW %TW- l 2P95-1 A 42 48 0.87 41 67 8 i 2P95 1 A 14 26. 1.62 91 92 24 l
~ 2P95-1 A. 22 26 4.05 84 93 22 2P95-1 A 16 28 1.99 50 .115 16 2P95-1 A 14 30 1.62 78 108 24 .;
2P95-1 A 18 30 2.08 81 72 17 f 2P95-1 A 12 32 1.65 89 48 12 2P95-1 A 37 21 1.05 40 109 13 , 2P95-1 A 23 33 3.1 93 92 24 ! 2P95-1 A 22 32 1.32 80 - 76 17 l 2P95-1 A 65 79 2.27 69 -160 31 -i 2P95-1 A 70 34 1.64 93 55 15 ) 2P95-1 A 73 35 0.74 56 52 8 2P95-1 A 80 40 1.2 40 62 7 5 2P95-1 A 55 39 2.06 94 52' 14 ! 2P95-1 A 81 41 4.65 % 76 21 ! 2P95-1 A 83 41 0.99 57 41 7 i 2P95-1 A 58 40 0.4 57 37 6 l 2P95-1 A 82 40 2.41 86 71 17 i 2P95-1 A 86 40 0.88 70 28 6 ; 2P951 A 12 18 3.32 92 96 25 2P95-1 B 83 57 2.4 84 51 12 ' 2P951 B 71 59 1.5 87 155 37 2P95-1 B 69 59 3.76 92 122 31 l 2P95-1 B 73 59 2.75 % 74 20 2P95-1 B 28 58 3.51 87 50 12 7 2P95-1 B 95 113 0.83 61 83 14 2P95-1 B 94 124 1.89 81 93 21 r 2P95-1 B 20 106 1 80 120 27 j 2P95-1 B 36 126 1.33 18 59 3 ! 2P95 1 B 11 133 0.86 20 30 2 : 2P951 B 87 65 0.39 51 35 5 ! 2P95-1 B 87 53 1.66 88 71 17 2P95-1 B % 52 0.73 21 67 4 2P95-1 B 100 52 2.52 88 103 25 , 2P95-1 B 92 54 0.77 21 112 7 s 2P95-1 B 96 54 2.21 14 74 3 2P95-1 B 100 54 3.28 87 73 18 2P951 B 81 47 2.82 94 34 9 ; 2P95-1 B 93 55 1.16 73 78 16 ; 2P95-1 B 101 63 0.8 73 93 19 2P95-1 B 91 49 2.37 93 99 26 , 2P95-1 B 100 48 2.64 91 97 25
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~ A*n=mt to - {
2CAN029505 - Page 88 of 89 OUTAGE - SG ROW LINE VOLTS MAX LENGTH AVG'
%TW %TW 2P95-1 B 104 62 1.46 95 106 28 2P95-1 B 100 62 1.33 91 164 41 2P95-1 B 100 60 1.68 76 100 21 i 2P95-1 B 111 75 2.19 93 60 16 i 2P95-1 B 107 65 0.35 52 60 9 2P95-1 B 103 77 3.11 91 120 30 l 2P95-1 B 91 83 1.25 68 43 8 t 2P95-1 B 111 87 0.64 64 57 10 ;
2P95-1 B 107 97 1.31 87 55 13 2P95-1 B 103 103 1.5 77 74 16 - 2P95-1 B 107 105 0.8 89 48 12 2P95-1 B 107 93 3.66 83 158 36 2P95-1 B 101 67 0.8 2 81 1 2P95-1 B 109 69 0.75 56 58 9 2P95-1 B 113 73 0.85 98 67 18 l 2P95-1 B 97 81 2.01 88 65 16 2P95-1 B 109 85 1.93 82 143 33 2P95-1 B 105 91 1.21 76 62 13 2P95-1 B 113 97 2.4 91 187 47 2P95-1 B 104 64 1.4 87 106 26 2P95-1 B 100 66 0.89 59 104 17 2P95-1 B 108 70 1.35 71 67 13 2P95-1 B 108 80 1.13 70 44 9 2P95-1 B 108 84 3.22 99 109 30 2P95-1 B 108 88 1.51 84 50 12 2P95-1 B 108 % 0.79 79 95 21 2P95-1 B 100 100 1.42 57 183 29 ! 2P95-1 B 104 104 2.19 88 57 14 2P95-1 B 106 80 0.66 47 92 12 2P95-1 B 106 88 2.33 85 85 20 2P95-1 B 110 98 1.39 85 109 26 2P95-1 B 110 100 2.45 91 48 12 2P95-1 B 106 104 0.59 47 39 5 2P95-1 B 98 100 3.38 100 64 18 2P95-1 B 100 98 1.73 84 41 10 2P95-1 B 112 % 0.93 85 41 10 2P95-1 B 66 112 0.94 75 53 11 2P951 B 40 34 1.43 75 78 16 2P95-1 B 17 37 0.7 17 99 5 2P95-1 B 16 36 1.41 83 58 13 2P95-1 B 22 42 3.27 85 108 26 2P95-1 B 23 45 2.49 85 126 30 2P95-1 B 24 48 2.48 76 220 46 I i
y, , Attachment to
!; 2CAN029505 l Page 89 of 89 l l
OUTAGE SG ROW LINE VOLTS MAX LENGTH AVG
%TW %TW 2P95-1 B 79 43 0.82 59 64 10 !
2P95-1 B 83 43 2.14 88 123 30 2P95-1 B 87 43 1.11 77 46 10 I 2P95-1 B 21 21 0.83 45 85 11 2P95-1 B 32 26 1.37 69 39 7 2P95-1 B 21 31 2.49 95 41 11 2P95-1 B 32 32 1.85 44 48 6 2P95-1 B 52 34 2.03 95 30 8 2P95-1 B 64 34 2.89 79 57 13 2P95-1 B 78 38 2.42 93 60 16 2P95-1 B 74 38 3.81 90 126 30 2P95-1 B 75 39 1.46 46 42 6 2P95-1 B 59 41 3.17 84 48 12 2P95-1 B 19 17 1.23 41 106 12 2P95-1 B 71 41 1.17 91 37 10 In addition, the following tubes with volumetric / axial indications were plugged in 2P95-1 OUTAGE SG ROW LINE 2P95-1 A 31 101 2P95-1 A 59 51 , 2P95-1 A 88 84 2P95-1 A 74 86 2P95-1 A 69 97 2P95-1 A 79 61 , 2P95-1 A 99 83 2P95-1 A 48 72 2P95-1 A 33 103 2P95-1 A 34 104 2P95-1 A 51 49 4 2P95-1 A 117 65 + 2P95-1 B 74 84 2P95-1 B 79 85 2P95-1 B 64 % 2P95-1 B 67 75 2P95-1 B 66 98 i i f
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