ML17262A845
| ML17262A845 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 03/25/1988 |
| From: | Snow B ROCHESTER GAS & ELECTRIC CORP. |
| To: | Russell W NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| References | |
| IEB-88-002, IEB-88-2, NUDOCS 9205180176 | |
| Download: ML17262A845 (23) | |
Text
ACCELERATED DI RIBUTION DEMONST TION SYSTEM C
D'EGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)
ACCESSXON NBR:9205180176 DOC.DATE: 88/03/25 NOTARIZED: YES DOCKET FACIL:50-244 Robert Emmet Ginna Nuclear Plant, Unit 1, Rochester G 05000244 AUTH. NAME AUTHOR AFFXLIATION SNOW,B.A. Rochester Gas 6 Electric Corp.
RECIP.NAME RECIPIENT AFFILIATION RUSSELL,W.T. Region 1 (Post 820201)
SUBJECT:
Forwards documentation re susceptibility of SG tubes to high cycle fatigue DISTRIBUTION CODE: DFOID TITLE: Direct Flow NOTES:License Exp 6
potential for subj failure,per NRC COPIES RECEIVED:LTR Distribution: 50 Docket (PDR Avail)l description of actions taken to minimize Bulletin ENCL 88-002 request.
date in accordance with 10CFR2,2.109(9/19/72).
I SIZE:
05000244 D
A RECIPIENT COPIES "
RECIPIENT COPIES ID CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL D INTERNAL: NUDOCS-ABSTRACT 1 1 G FIL 01 1 1 D EXTERNAL: NRC PDR 1 1 NSIC 1 1 R
D D
D NOTE TO ALL "RIDS" RECIPIENTS:
PLEASE HELP US TO REDUCE WASTE! CONTACT THE DOCUMENT CONTROL DESK, ROOM P 1-37 (EXT. 20079) TO ELIMINATEYOUR NAME FROM DISTRIBUTION LIS13 FOR DOCUMENTS YOU DON'T NEED!
TOTAL NUMBER OF COPIES REQUIRED: LTTR ~ ENCL
I t
ROCHESTER GAS AND ELECTRIC CORPORATION ~ 89 EAST AVENUE, ROCHESTER, N.Y. 14649-0001 Narcri 25, 1988 Q eece <<oc f fe 0'ic ro 5~6? rOO U.S. Nuclear Regulatory Commission Mr. William T. Russell, Regional Administrator 475 Allendale Road King of Prussia, PA 19406
Subject:
NRC Bulletin No. 88-02: Rapidly Propagating Fatigue Cracks in Steam Generator Tubes R.E. Ginna Nuclear Power Plant
~
Docket No. 50-244
Dear Mr. Russell:
NRC Bulletin 88-02 requested that holders of operating licenses or construction permits for Westinghouse (W) - designed nuclear power reactors with steam generators having carbon steel support plates implement actions to minimize the potential for a steam generator tube rupture event caused by a rapidly propagating fatigue crack such as occurred at North Anna Unit 1 on July 15, 1987. Attached is documentation regarding the susceptibility of R.E. Ginna steam generator tubes 'to high cycle fatigue failure and a description of actions taken to minimize the potential for such a failure.
Very truly yours, Bruce A. Snow Subscribed and sworn to before me this 25th day of March, 1988.
Notary Public LYNN L HAUCK NoffrfyPobbc or rbe Stere d New York MONROE COUNTY f~frrorr Exprree Nov. 30. 19.X~
Attachment xc: Document Control Desk (original}
xc: Mr. Carl Stahle, PWR Project Directorate No. 1 xc: NRC Resident Inspector 9205i80i76 880325 PDR ADOCK 05000244 gr. Inc/
~D: P/lsn MhnMn
SUSCEPTIBILITY TO HIGH CYCLE FATIGUE FAILURE OF STEAM GEMDHTOR TUBES R.E. GINNA NUCLEAR PLANT This evaluation was prepared in accordance with the actions requested by NRC Bulletin No. 88-02: "Rapidly Propagating Fatigue Cracks in Steam Generator Tubes." The initiation of the circumferential crack in the tube at the top of the top tube support plate at North Anna 1 has been attributed to limited displacement, fluid elastic instability. The unstable condition prevailed in the R9C51 tube when the tube experienced denting at the support plate. A combination of conditions were present that led to the rupture. The tube is riot supported by an anti-vibration bar (AVB), has a higher flow field due to the uneven insertion depths of AVB's, has reduced damping because of denting at the top support plate, and has reduced fatigue properties in an environment of the all volatile treatment (AVT) chemistry of the secondary water because of the additional mean stress from the denting.
Eddy Current Data Evaluation During the Spring 1988 Refueling Outage the tubes in the R.E.
Ginna steam generators, which could be affected by the North Anna phenomenon, were inspected by eddy current probes to evaluate tube denting at the top support plate and to determine the location of the AVBs.
The eddy current examination consisted of 100% examination of rows 9 through 12 with a few additional larger row tubes (approximately 11% of each Steam Generator). The eddy current test data showed that 10 tubes in SG A and 6 in SG B are dented at the sixth support plate as shown on Tables 1 and 2 (in rows 9 through 12). It is judged that support plate corrosion is present at the top support plate for the remaining tubes. In the analysis all the tubes were assumed to be dented and were treated as clamped with maximum mean stress (due to the denting deformation) .
The eddy current test data also indicates that no tubes have wall thinning at the AVB locations. It is unlikely that tubes with AVB support have been unstable.
AVB Insertion Depths The eddy current inspection data were used to determine the AVB insertion depths and the extent of tube support for each column of tubes within the region. Where clear indications of the presence of an AVB are found, the AVB positions are defined.
Where clear indications are not found, eddy current results from outer rows are used to determine the AVB insertion depth. This is done by projecting the inner row position of the AVB from its known location in the outer rows, using the geometry of the AVB
and U-bend. The results are summarized in Figures 1 and 2. A tube is considered supported only when the AVB penetrates to the centerline of the tube. Virturally all of row 12 and row 13 tubes are supported except for a few peripheral column of row supported.
ll tubes are supported. Some row 10 tubes are tubes'ost ll Tube Vibration and Fatigue 'Analysis has developed conservative acceptance criteria to
.-"'estinghouse determine susceptibility to high cycle fatigue of steam generator tubes. The general approach is to specify that tube vibration displacements and total stresses are sufficiently small such that the high cycle fatigue would not be expected to occur.
Specifically, the acceptance criteria for a tube which is dented at the top tube support plate are 1) AVB support or a stability ratio which is less than or equal to 90 percent of the stability ratio of North Anna R9C51 and, 2) a stress ratio relative to North Anna R9C51 which, after applying the ten percent reduction in the stability ratio, is less than or equal to 1.0. The ten percent reduction in stability ratio reflects tube vibration stresses of less than 4 KSI based on the maximum calculated stresses for North Anna R9C51. At 4 KSI the fatigue usage is less than 0.021 per year. Even assuming a tube developed a through-wall crack up to 125 mils long, such a tube would not experience crack growth at these stress levels..
The stability ratios for R.E. Ginna tubing, the corresponding stress amplitude, and the resulting cumulative fatigue usage must be evaluated relative to the ruptured tube at Row 9, Column 51, North Anna 1, Steam Generator C, for two reasons. The local effect on the flow field due to various AVB insertion depths is not within the capability of available analysis techniques and must be determined by test. In addition, an analysis and examination of the ruptured tube provided a range of initiating stress amplitudes, but can only bound the possible stability ratios that correspond to these stress amplitudes. Therefore, the evaluation of R.E. Ginna tubing has been based on relative stability ratios, relative flow peaking factors and stress ratios.
The determination of stability ratio is the evaluation of a ratio of velocities, the effective velocity divided by the critical velocity. A value greater than unity (1.0) indicates instability. The stress ratio is the expected stress amplitude in an R.E. Ginna tube divided by the stress amplitude attributable to the North Anna 1 tube rupture.
If a tube has support from an AVB, the analyst can eliminate from further consideration. The criteria for establishing that a it tube has support from an AVB is that an AVB is present, on at
'least one side, that penetrates to, at least, the tube centerline. This has been established by analysis of eddy current (EC) measurements by either an EC indication of both of the AVB or by projecting the depth of insertion knowing thelegs geometry of the AVB and the location of its EC indications on larger radius U-bends in the same column of tubes.
The tube evaluation for R.E. Ginna steam generators was performed by Westinghouse Electric Corporation and included the following interrelated analyses:
- 1) The three dimensional flow analysis code (ATHOS) was used to calculate flow conditions in regions of interest. The ATHOS model was set up with geometry parameters specific to R.E.
Ginna steam generators. Detailed secondary side velocity, density and void fraction distributions were calculated for tube rows 8 through 12. The data from the ATHOS analysis provided input to the vibration analysis codes.
- 2) Tube stiffness, frequency, and fluidelastic stability ratios were obtained by dynamic analysis for tubes in rows 8 tnrough 12 using the FLOVIB code. The inputs for this evaluation were the flow field characteristics of velocity, density, and void fraction obtained from the ATHOS analysis.
The dynamic analysis results were then used to calculate tube stress ratios to identify tubes which were potentially susceptible to high cycle fatigue similar to North Anna R9C51 event.
- 3) Axisymmetric finite element analysis was used to determine the mean stress at the tube intersection with the top tube support plate during 100% power conditions.
- 4) Concurrent with the evaluations in 1 through 3 above, air model tests were perrormed to determine the effects on fluidelastic instability of columnwise variations in AVB insertion depths. Simulated AVBs were inserted at different depths in adjacent tube columns to evaluate various configurations of tubes and AVBs. The instability thresholds were determined for specific tubes of interest.
These characteristics were then compared to the North Anna R9C51 value to determine tube specific relative stability peaking factors for R.E. Ginna. The direct test measurements for peaking factors are adjusted by uncertainties developed from test and analysis data prior to application to the tube vibration and fatigue analyses.
The data obtained from the various analyses, together with the tube denting data from the eddy current evaluation and the AVB insertion maps, were combined to provide disposition of the tubes within the potentially susceptible region of rows 9 or larger.
The analyses identified the following tubes for further evaluation based on unsupported tubes which either had relative stability ratios greater than approximately 0.90 or stress ratios
greater than 1.0 (assuming local flow peaking conditions to be the same as for North Anna R9C51). These were as follows:
Steam Generator A Steam Generator B Row Column Row Column 13 5 15 3 13 6 15 90' 12 2 14 3 12 3 14 90O 12 4 13 3 12 5 13 9Q' 12 6 12 12 7 12 3 12 8 17 12 9 12 5 12 10 12 6 12 11 12 87 12 91 12 88 11 12 89 12 90 12 91 11 73 11 89 11 QQ
$ 1 91 10 67 10
'indicates tube already plugged The local flow peaking due to actual AVB insertion depths was then factored into the evaluation for each of the above tubes. This resulted in the relative stability and stress ratios shown below. All ratios are in comparison to North Anna R9C51.
Steam Generator A Steam Generator B Stabilit Stress Stress Tube Ratio Ratio ~b'I.ube Ratio Ratio R13C5 1.29 >1.0' R15C3 >1.0*
R13C6 1.23 >1.0* R15C90 p**
R12C2 0.86 0.58 R14C3 >1.0*
R12C3 1.09 >1.0* R14C90 >1 Pk*
R12C4 1.04 >1.0* R13C3 0.91 0.67 R12C5 1.01 >1.0* R13C90 0.91 0.67 R12C6 0.98 1.31* R12C2 0.86 0.58 R12C7 0.96 1.03* R22C3 1.09 >1.0+
R12C8 0.944 0.99 R12C4 1.04 >1.0*
R12C9 0.937 0.92 R12C5 1.01 >1.0*
R12C10 0.801 0.45 R1? 0.98 1.31+
R12C11 C6'12C87 0.98 1.31*
R12C91 R12C88 1.01 >1.0*
P.11C3 P12C89 1.04 p*
R12C90 1.09 >1 0*
R12C91 0.86 0.58 R11C73 1.012 >1.0*
R11C89 0.78 0.38 R11C90 0.80 0.45 R11C91 0.63 0.12 R10C67 0.69 0.22 R10C51 0.60 0.11 The current fatigue usage was calculated for tube R12C8 in SG A, which has the highest stress ratio, 0.99 of those tubes left inservice. This tube is currently not dented.
cumulative fatigue usage to date of 0.059. Even assuming It has a it becomes dented during the next cycle, the total usage during the remaining term of the operating license would be 0.417, a value of 1.0 being the limit of acceptability.
Based on the current fatigue usage and the stress ratios given above, it is concluded that those tubes indicated
(*) could possibly experience the by asterisks fatigue similar to the North Anna occurance. As a preventative measure, these tubes have been plugged with standard Combustion Engineering (CE) mechanical plugs in the hot leg and CE sentinel plugs in the cold leg of the steam generators. Tubes marked with two asterisks (*") could also experience fatigue similar to North Anna, however, they have been previously plugged. In addition, tubes R13C90, R15C89, R16C89, and R17C89 were also previously plugged. Because of this prior plugging, box plugging was required as shown on Figure 3.
Tubes box plugged are listed below:
R13 C89 R13 CSS 814 C89 i14 CSS R15 CSS R16 CSS R17 C89 (already plugged)
R17 CSS R18 CSS A summary of all North Anna plugging for each steam generator is shown on Tables 3 and 4.
Combustion Engineering Sentinel Tube Plugs The CE sentinel tube plug is a derivative of the standard CE mechanical tube plug. The only difference between the standard mechanical plug and the sentinel plug is the addition of a small drilled hole approximately 3 inches from the flared end of the plug as 'shown on Figure 4. The small hole in the plug allows approximately 200 gallons per day of primary to secondary leakage in the event of a tube failure.
Enhanced Leakage Monitoring Program Pending completion of the NRC staff review and approval of the program as previously described, an enhanced primary to secondary leakage monitoring program will be implemented. The program is described below:
- 1) New alert alarm setpoints have been established for the two air ejector monitors (R-15 and R-15A) to maximize sensitivity to primary to secondary leakage prior to power operation following an outage.
- 2) During power escalation, leak rates will be evaluated using data from the continuous monitors and radiochemistry analysis perrormed during periods of steady state operation.
Based on these evaluations, alarm setpoints are revised, required.
if
- 3) During full power operation, the leak rate monitoring consists of continuous monitoring, periodic radiochemistry analyses, and leak rate calculations. The radioactivity concentrations from the air ejector monitors are trended and reviewed daily. These provide direct indications of primary to secondary leakage. Data from the failed fuel monitor is trended and reviewed daily to assess changes in the reactor coolant activity. Primary to secondary leak rates are calculated three times per week as a function of the blowdown radioiodine concentration, blowdown rate and primary coolant radioiodine concentration. Leak rates are calculated monthly as a function of air ejector noble gas
'ctivity, air ejector flow rate and primary coolant noble gas concentration.
- 4) If theand calculated primary to secondary leak rate exceeds 20 gpd is less than 60 gpd or an alert alarm is received from an air ejector monitor, the indicated change is evaluated by other means. These may include, but a'e not limited to, reviewing the trends of the other continuous monitors, reviewing changes in operating conditions, or performing radiochemistry analysis to allow independent leak rate calculation. If the leak rate is verified between 20 and 60 gpd, the frequency of the .radiochemistry analysis of the blowdown and the leak rate calculation is increased to daily. The air ejector alarm is reset to a value corresponding to an increase of 30 gpd.
- 5) If the primary to secondary leak rate exceeds 60 gpd, radiochemistry analysis of the blowdown and leak rate calculations are increased to a frequency of once a shift and trending of the continuous monitors is performed once a shift.
- 6) If the primary to secondary leak rate exceeds 100 gpd, an evaluation of the need to reduce power or shutdown will be performed. This evaluation will include the leak rate history and the rate of change of the leak rate. This should provide an additional 5 to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> Co reduce power or shutdown in addition to the 5 hour5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> margin indicated on Figure 1 of Bulletin 88-02, should a high cycle fatigue failure of a tube be in progress.
- 7) Air ejector monitor R-15A is the most sensitive to primary to secondary leakage. Therefore, if R-15A is out of service, daily blowdown activity analyses and primary to econdary leak rate calculations will be performed.
Trending and review of other parameters will continue as described in 3) above.
Future Analysis and Inspections In the analysis of the support conditions for the inner rows in the U-bend region of the Ginna steam generators, several tubes in the outer columns were found to be unsupported and have stability ratios just above the acceptability criterion. The flow analysis was performed using a cylindrical coordinate model in the ATHOS code, as has been done in the past.
Recent experience from the l1odel 51 steam generator analyses has shown that the Cartesian coordinate version of ATHOS permits a better simulation of the U-bend tubes and the AVBs. A plot of stability ratio as a function of column number for a given tube row shows a peak in the peripheral columns at both ends of the row. t 's suspected that inadequate definition of the U-bend geometry in the cylindr'cal model may be responsible for this.
It is likely that the results of an analysis using the Cartesian
coordinate version will show lower stability ratios for these peripheral tubes. Consequently, the outer tubes in question could be stable. In this event, plugs installed in up to 12 end column tubes, during the refueling outage, could be removed.
In addition to the Cartesian analysis, it is anticipated that tubes on the periphery of the steam generator will be visually inspected during the next outage to determine if some of the previously plugged tubes have supports that could not be verified by eddy current. Tubes having proper support could then be unplugged.
.Summary In an effort to minimize the potential for a stea'm generator tube rupture event similar to that which occurred at North Anna Unit 1, an ehgineering analysis of the R.E. Ginna steam generators has been completed. As a result of this analysis, 29 (3 of these tubes were previously plugged) tubes were identified as being potentially susceptible to high cycle fatigue or susceptible to the consequences of fatigue failure and long term corrective actions were taken accordingly i.e., all active tubes (26) were preventively plugged utilizing sentinel plugs. All unsupported tubes remaining in service are projected to have acceptable cumulative fatigue usage at the end of the plant life cycle. In addition, on an interim basis, a leakage monitoring program has been implemented at R.E. Ginna with an effectiveness consistent with the time dependent leakage curve given in Figure 1 of NRC Bulletin 88-02.
Based on the above, the return to power and subsequent operation of R.E. Ginna will not result in a previously unanalyzed accident and is not expected to increase the probability of a previously analyzed accident. Additionally, as the tube bundle has been effectively restored such that tubes potentially susceptible to fatigue failures have been removed from service, the margin of safety of the primary pressure boundary which, in part, is provided by the ASME Code is not reduced. Therefore, subsequent plant operation of R.E. Ginna does not represent an unreviewed safety question as defined in the criteria of 10 CFR 50.59.
Table 1 Stewn Generator I, Dented Tubes in Rows 9 throu h 12 Bow Col Indication Volts Location 2 DENT 75.25 H6 10 DENT 15.94 H6 12 2 DENT 26.82 H6 10 3 DENT 17.44 H6 12 3 DENT 22.60 H6 10 4 DENT 6.64 H6 12 DENT 9.78 H6 12 5 DENT 6.48 H6 12 90 DENT 10.93 C6 12 91 DENT 11.33 C6 H6 sixth support plate, hot leg C6 sixth support plate, cold leg
Table 2 Steam Generator 3 Denced Tubes in Rows 9 throu h 12 Col indication Volts Location 12 DENT 9.15 C6 12 81 DENT 14.21 Ho 11 89 DENT 14.26 C6 12 89 DENT 13.94 C6 11 90 DENT 17.12 C6 12 90 DENT 20.05 C6 H6 sixth support plate, hot leg C6 sixth support plate, cold leg
"able 3 A Steam Generator North Anna Plua in Summar TYPE OF PLUG INSTALLED ROW COL HOT LEG COLD LEG COMMENTS 12 CE-MECH CE-SENTINEL NO AVB SUPPORT 12 CE-MECH CE-SENTINEL NO AVB SUPPORT 12 CE-MECH CE-SENTINEL NO AVB SUPPORT 12 CE-MECH CE-SENTINEL NO AVB SUPPORT 12 CE-MECH CE-SENTINEL NO AVB SUPPORT 13 CE-MECH CE"SENTINEL NO AVB SUPPORT 13 CE-MECH CE-SENTINEL NO AVB SUPPORT
Table 4 8 Steam Generator North Anna Plu in Summar TYPE OF PLUG INSTALLED ROW COL HOT LEG COLD LEG COMMENTS 12 3 ~ CE-MECH CE"SENTINEL NO AVB SUPPORT X2 4 CE-MECH CE"SENTINEL NO AVB SUPPORT 12 5 CE-MECH CE-SENTINEL NO AVB SUPPORT 12 6 CE-MECH CE-SENTINEL NO AVB SUPPORT 14 3 CE-MECH CE-SENTINEL NO AVB SUPPORT 15 3 CE-MECH CE-SENTINEL NO AVB SUPPORT 12 87 CE-MECH CE-SENTINEL NO AVB SUPPORT 12 88 CE-MECH CE-SENTINEL NO AVB SUPPORT 12 89 CE-MECH CE-SENTINEL NO AVB SUPPORT 12 90 CE-MECH CE-SENTINEL NO AVB SUPPORT 13 88 .E-MECH CE-SENTINEL NG AVB SUPPORT 14 . 88 CE-MECH CE-SENTINEL NO AVB SUPPORT 15 88 'E-MECH CE-SENTINEL NO AVB SUPPORT 16 88 CE-MECH CE-SENTINEL NO AVB SUPPORT 17 88 CE-MECH CE-SENTINEL NO AVB SUPPORT 18 88 CE-MECH CE-SENTINEL NO AVB SUPPORT 13 89 CE-MECH CE-SENTINEL NO AVB SUPPORT 14 89 CE-MECH CE-SENTINEL NO AVB SUPPORT 73 CE-MECH CE-SENTINEL NO AVB SUPPORT
FIGURE 1 GINNA: SG - A AVB POSITJONS 12 2 4 4 4 4 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 4 2 2 2 2 2 2 2 2 s 2 2 2 2 2 2 2 s 2 2 2 OoOo 2 10
- 680880'0'080~80'0'8 ~ 0'OLO'88 O'O'000'IOIO~OIO~O>>
Cotunn 82 8088880000880080000$
81 80 80 88 87 88 8584 8S 82 81 80 79 78 77 78 75 74 7S 72 00000$ 0$ $
7'I 70 88 88 87 88 85 84 2 S S 2 12 2 2 2 2 2 2 0 0 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 s 2 01010108080'000.0'O. 010'0101010101 2 2 2 2 2 2 10 800 88880O08888688O8888886O888
$ 8888888$ $ 8$ $ $ $ 0$ 80880000'$00888 Ccetnon 82 el 50 se 57 58 5$ 54 $2 51 50 40 48 47 40 4$ 44 45 42 40 se 57 se 55 Si SS 22 11.2 11.5 114 11.
1 Q1Q1 2 2 S 2 2 2 2 2 0 0000001ee 2 0 0 0 0 ~
2 S 2 2 2 2
- GOSSOOOOO80808808 10 00080000080000$ 00880 $ 00 000 009e
~ 800880$ 0080$ 6000 ~ 0000080008000$ Insense Sl SO 28 28 2l 28 25 24 2S 22 21 20 18 I~ 17 18 1514 15 12 11 10 8. 8 7 8 5 Aojsdksl 120 'IS.O 12,5 12 12.5 12. IS. 1S. IS. 12.
HoNe:
PI~has
~ 14mbne
~
ebthe te&d AV8e Abb h ctereepondny tubs beeee Iet ereerlnedotshv8 Iebooe eooe cenleots et
~ Ruooed Tube DenNd Ttite PLUGGING STATUS SHOWN IS PRE 1988 OUTAGE Q Ih~
FlGURE 2 GINNA: SIG - B AVB POSITIONS
~ I I I 2 222442 2
12 eQ$ Q$
~ I I ~ 4 2 2 4 2 2 2 ee08488088800 2 2
0$
01 I 5 2 2 I 2 l 1 10 00088880800888' ' *
- 0$
- F 85 '5QQOQ0088888888888888 O'O'0'Qe
$2 81 $0 8$ $ $ 87 $ 8 85 $ 4 8S 82 ~I 80 7$ 77 78 75 74 7S 72 7$ 8$ $$ 87 88 8$ 84 (i) IS. $ 2.$ 12.2 1.$ 11.$ 1.7 12.1 $ .7 $ .$ 1.2 0.7 $ .5 12 2 2 2 4 t 2 2 2 2 2 2 2 2 2 2 I 2 2 t 4
- 8888808888*
10 8Q**OQQQQQQOQ888888'806 *880868 8888888888888888884800808808608 82 $1 $0 5$ 5$ 57 58 $$ 54 55 51 50 4$ 4$ 47 48 4$ 44 4S 42 41 40 SO $$ S7 5$ S5 54 SS 52
$ .4 '11.5 11.2 10.4 102 lb.1 11 04
'084 1$ I I I Qb 12 1 4 2 4 4 2 2 2 4 2 2 I 5 1 4 2 2 2 88 F 006880 10 0000000600000000 00080000000600 I I SO I
2$
I 28 I
8088888888885888000058480$
27 I
28 I
25 24 I
2S I
22 I
2$
8 20 8
I~
8 1$
~
17
~
18
~
1$
~
14
~
1$ 12
~
11
~
10
~
8
~
8 8
7
$ 2.$
0$
~
8 12.1 8
~
5 12.
'0 4 12.1 1$ .4
~
S 2 I'1.8 Ihhe:
~
~ Wmbere h cbceee eppeeenl AV8e vldble h ccerseponc$ ny bkeL proJeclione se booed on oer!1%no elAvB lelelbe lo csrser$ ne et Mrs,
~ Plugged Tube Oereed lc4e Q 8$ $ $ 41 5 oe Ieugpsd, eupporl csnnol be esrbhd
~
lhdds
~
p) Icecionne hr o)ecllonInleryreeedio be rehher rh0; eupporl le nal seecesd.
PLUGGING STATUS SHOWN IS PRE 1958 OUTAGE Q
Row 18 OS Sentinel Plug Row 17 Plugged S Row 16 Row 15 s Row 14 s s Row 13 S S S Row 12 Row 11 Row10 Columns 91 90 89 88 Figure 3. Tubes Involved in Postulated Chain of Damage in Ginna SG B.
The Posulated Chain Propagation would Involve only plugged tubes without affecting any other tubes.
~~~~~ Iflf iCALe
~ NAWN Q<~
iT 4J~ O4C~
~ ATe CWCCNCO t o TNACti OT ~TRO TNH ONAWINi Ii Tm feOftaTT Or COMIUSTN CNOHCNNP. NC,
%Nest&, COHk.
AGIO N WOT TO M Nlf&OOQCCO. ON II%CD TO tVNNINIIAIIT ~ONQATIOII
~ 00 IAAKIINlIÃ ONAWINOi ON At%A~
MTMO CXCCTT WIICNC POOVIDCD 0'T AONCCltCIIT WITII OLIO COEIPANY.
FlGURE 4 AW
0 1