ML20235P297
| ML20235P297 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 03/13/1987 |
| From: | SACRAMENTO MUNICIPAL UTILITY DISTRICT |
| To: | |
| Shared Package | |
| ML20235P295 | List: |
| References | |
| AP.56, TAC-60632, TAC-64798, NUDOCS 8707200434 | |
| Download: ML20235P297 (21) | |
Text
_
c EFFECTIVE DATE 03-13-87
. Rev. 3 f~ WP2553P D-0040P
{ AP.56 CORE DAMAGE ASSESSMENT 1.0 PURPOSE l
l To provide a method for estimating the extent of core damage utilizing data obtained from the Post Accident Sampling System (PASS) and supported by other appropriate station indicators (e.g., containment high range monitors, in-core temperature monitors, containment hydrogen concentration).
2.0 REFERENCES
2.1 " Post-Accident Sampling Guide for Preparation of a Procedure to Estimate Core Damage," NRC Generic Memo, 1982 2.2 NUREG-0772, " Technical Bases for Estimating Fission Product Behavior During LWR Accidents," USNRC, June 1981 2.3 NUREG/CR-1237, "Best Estimate LOCA Radiation Signature," L. L. Bonzon, Sandia Labs..N.A. Lurie IRT Corp., Jan.,1980 2.4 NUREG/CR-1288, " Fission Product Source Terms for the LWR Loss-of-Coolant
. Accident," Lorenz et. al. (ORNL), July 1980 2.5 Rancho Seco USAR, Chapters I and XIV 2.6 AP.500, Rancho Seco Emergency Plan 2.7 AP.313-1, " Emergency Sampling Using the Post Accident Sampling System (PASS)"
2.6 Stone and Webster Calculation, Technical Bas'is for Radioisotopic Determination of Extent of Core Damage Following a loss of Coolant Accident, Rancho Seco, June 24, 1983 2.9 Bechtel Calculation, Z-RDH-M-1810, Response of High Range Containment Atmosphere Monitors, Rancho Seco, June 16, 1986 Rev. 3 Os AP.56-1 g 7Pgg Q S70706poo {gla P
l
3.0 PROCEDURE 3.1 Limitations and Precautions 1
3.1 .1 This procedure is for accident analyses and may not be applicable i
for determining a fuel cladding failure of less than or equal to 1%.
3- 3.1 .2 When conducting an isotopic analysis from PASS always use data from several isotopes since the sample measurement error on a given isotope can range up to around 30%.
3.1 .3 The method given in this procedure assumes that the activity released to the containment and the radiation monitors response is
- proportional to the extent of core damage.
3.1 .4 Results obtained from radioisotope analyses procedures should be "
evaluated in conjunction with other plant indicators. This will provide a more realistic estimate of core damage.
3.1 .5 When analyzing the halogen data, care should be exercised to consider the " iodine spiking" effect after shutdown on the core damage estimates derived from iodine isotopes.
. 3.1 .6 If samples are taken more than a few hours after shutdown, isotopes Kr87 and 1134 may have decayed to low levels, adversely affecting the accuracy of the fuel failure predictions.
3.1 .7 Boron and chloride concentrations are recorded on Enclosure 4.1 for bha information only.
3- 3.1 .8 Accuracy of the results from this procedure can be affected significantly if the plant conditions are changir.g rapidly.
3.2 The Emergency Coordinator has determined the need to estimate the extent of core damage.
3.3 Collect and record station data in Enclosure 4.1.
3.3 .1 Evaluate actual (accident condition) data as compared to normal station conditions.
3.3 .2 Determine preliminary indication of core damage from Enclosure 4.1.
Rev. 3 "
AP.56-2
r
+
PROCEDURE (Continued) 3.4 Containment Area Radiation Monitor Indications 3.4 .1 Record area radiation monitor, R-15049 and R-15050 readings on
' Enclosure 4.2 Table 1.
3.4 .2 Record time when the readings were taken. ,
3.4 .3 Record the time of reactor shutdown.
3.4 .4 Estimate the time interval between reactor shutdown and the time monitor readings were taken, and enter the time interval in Table 1.
3.4 .5 Select the monitor giving higher readings for further evaluation (R-15050 is expected to read higher).
3- 3.4 .6 Are the containment emergency Coolers, 500A and 5008, are operating:
3.4 [ ] Yes [ ] No [ ] Not Known 3.4 .7 If status of the containment emergency Coolers, 500A and 500B, is not known, assume they are. operating.
3.4 .8 Go to Enclosures 4.10, 4.11, 4.12, and 4.13 and select one of the four Enclosures as following:
Enclosure 4.10: If R-15050 selected and containment emergency Coolers, 500A and 500B operating.
Enclosure 4.11: If R-15050 selected but containment emergency Coolurs 500A and 5006 nat operating.
Enclosure 4.12: If R-15049 selected and containment emergency Coolers, 500A and 5008 operating.
Enclosure 4.13: If R-15049 selected but containment emergency Coolers, 500A and 500B not operating.
3.4 .9 Enter the selected Enclosure number in Table 2 of Enclosure 4.2.
3.4 .10 The selected Enclosure contains four curves ('A, B, C and D) showing monitor response (R/Hr) as a function of time since reactor shutdown. Plot the reading of the selected _ monitor for the time interval recorded in Table 1 in Enclosure 4.2.
Nata: If the data point is between two intersecting curves, ignore the curve which gives lower core damage.
3.4 .11 Complete Table 2 in Enclosure 4.2 to estimate the extent of core damage.
Rev. 3 O -
AP.55-3
-l PROCEDURE (Continued) 3.4 .12 If the indicated core damage assessment in Enclosure 4.2 is significantly lower than the indications from Enclosure 4.1, the ;
following two conditions could exist:
~ i 3.4 .12.1 RCS was not breached during the event. ,
3.4 .12.2 The radiation monitor is not giving a true representation of the containment radiation environment.
3.4 .13 Compare the core damage indicated by Table 2 in Enclosure 4.2 with the indications of hydrogen monitor response in Enclosure 4.1.
3.4 .14 If the two indications in Step 3.4.13 are comparable, but both are l significantly lower. than the other indications in Enclosure 4.1 (particularly in-core temperature data) evaluate the possibility that the RCS is not breached.
3.4 .15 If the indicated damage by Table 2 in Enclosure 4.2 is significantly less than the hydrogen monitor indications, or the RCS breach is verified by other means, assume the radiation monitors, R-15049 and R-15050 are not representing radiation from failed fuel escaping the RCS through a break.
3.4 .16 If the actual power history is significantly different from steady state full power operation before reactor trip, use engineering judgement to include the results of Enclosure 4.2 for further @5 evaluation of core damage. l 3.5 Determine the need for PASS samples. Notify the Radiological Assessment Coordinator which samples should be taken and analyzed. j 3.5 Record PASS results on Enclosure 4.3 Column 2 (measured sample concentration in pC1/ml) and normalize results as follows:
3.6 .1 If the measured reactor coolant sample included non-reactor coolant !
water, such as safety injection water (V1 ) or watgr located in !
the sump prior to the accident (V2), then use the .following
~
dilution factor:
f - (Vc + V 1 + V2 )/Ve, ,
where Vc is the reactor coolant volume. ;
3.6 .2 Calculate Temperature / Pressure Normalization (TPN) factor as follows:
Rev. 3 V l
AP.56-4
4 PROCEDURE (Continued) 13 V 3.6 .2.1 Containment Atmosphere Pressure Pc = psig - atm (atm = 1 + psig/14.7)
Temp ~ Tc - 'F - *R ('R = 460 + *F) -
3.6 .2.2 Sample (AP.313-1)
Pressure Ps " psi 9 " atm Temp Ts - *F - 'R 3.6 .2.3 TPN Factor . Pc ata x Ts 'R Ps ata x Tc *R .
-[ ] x [ ] -
l 3.7 Verify that the PASS samples are corrected for isotopic decay from the time of reactor' shutdown to the time the sample is analyzed per procedure AP.313-1.
3.8 Compute the Power Level Correction as follows:
3.8 .1 Determine thermal power history within 30 days preceding shutdown,
, and record as follows:
3.8 .1.1 During 24-hours preceding shutdown MHt hrs.
3.8 .1.2 During 14-days' preceding shutdown MHt days. j 3.8 .1.3 During 30-days preceding shutdown MHt days.
3.8 .2 For isotopes with half-life T1/2 1 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />s:
F- 2772 MHt x 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Actual MHt Hr summed over 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. preceding ' reactor trip
2772 X 24
(Above Item) 3.8 .3 For isotopes with half-life T1/2 between 1 and 3 days: !
l F- 2772 MHt x 14 davs '
Actual MHt Days summed over 14 days preceding reactor trip
. 2772 x 14 .
(Above Item)
Rev. 3 1 O AP.56-5 I
-_ __ _ __ __)
m PROCEDURE (Continued) 4 3.8 .4 For isotopes with T1/2 > 3 days:
4 F- 2772 MHt x 30 days Actual MHt' Days summed over 30 days preceding reactor. trip
. 2772 x 30 =
(Above Item)
NOTE: The above designations are for reactor operating time greater than 30 days. If the reactor operating time is less than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, use 3.8.2 for all isotopes. If the reactor i
operating time is greater ' han t 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and less than 14 .
I days, use 3.8.3.for all isotopes. If the reactor operating time is greater than 14 days and less than 30 days use 3.8.3 for isotopes with half-life T 1/2124 hours and use 3.8.4 for isotopes with half-life T 1/2 > 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
3.8 .5 Use Enclosure 4.9 to determine the half-life of various isotopes.
3-- 3.9 Record all power level correction factors computed from Steps 3.8.2, 3.8.3 and 3.8.4 on Column 5 of Enclosure 4.3.
3-~ 3.10 Compute the Normalized Concentration (pC1/ml) and record in ,
Column 6 of Enclosure 4.3.
3-~ 3.11 Using data from Enclosure 4.3 Column 6 refer to Enclosures 4.4, 4.5, 4.6, and 4.7 as appropriate and record data on Columns 2 and 3 of kh.
Enclosure 4.8.
3.12 Evaluate the results of Enclosures 4.1, 4.2, and 4.8 to estimate the extent of core damage, and enter the best estimate in Enclosure 4.14.
l l
1 1
I Rev. 3 AP.56-6 I
4.0 ENCLOSURES 4.1 Preliminary Indication of Core Damage Utilizing Various Station Indicators 4.2 Containment Area Radiation Monitor Indications 4.3 Normalized Post-Accident Radionuclides Samples 1
4.4 Extent of Core Damage for Xe133 Concentration in Containment Atmosphere or Reactor Coolant at Rancho Seco Full Power History i 4.5 Extent of Core Damage for I131 Concentrations in Reactor Coolant at '
Rancho Seco Full Power History 4.6 Extent of Core Damage for Cs Isotope Concentrations in Reactor Coolant l at Rancho Seco Full Power History 4.7 Extent of Fuel Helting or High Overheating for Ho99 and Rul03 Concentrations in Reactor Coolant at Rancho Seco Full Power History 4.8 Radioisotope Determination of Core Damage from Corrected and Normalized .
PASS Results I 4.9 Isotopic Data 4.10 Area Radiation Monitor R15050 Response with Coolers On 1
't 4.11 Area Radiation Monitor R15050 Responsa with Coolers Off
}
4.12 Area Radiation Monitor R15049 Response with Coolers On~
4.13 Area Radiation Monitor R15049 Response with Coolers Off 4.14 Best Estimate Conclusion for Extend of Core Damage ,
i l I Rev. 3 O AP.56-7
1
. 3 ENCLOSURE 4.1 q
~
PRELIMINARY INDICATION OF CORE DAMAGE l UTILIZING VARIOUS STATION INDICATORS i j
INDICATES {
INDICATOR NORMAL RANGE ACTUAll NO POSSIBLE i l
DAMAGE CLAD / FUEL DAMAGE 3- Containment Hydrogen 3 0 Monitor l
l In-Core Temperature 590 - 630 *F i Monitor i Reactor Coolant 15-100 cc/kg Dissolved Gases oH of Water 2 4.8 - 8.5 For Information Oniv Iodine Spiking Estimate up to 50 pci/cc Information to evaluate PASS samole results Coolant Boron 4 0 - 2000 ppm Information to evaluate criticality Coolant Chloride 5 0 - 0.15 ppm Information to evaluate
- Lena Term Reactor Status CONFIRMING INDICATORS OF CORE DAMAGE I Indicator Rance Clad Damace Fuel Overheat Fuel Melt In Core 750*-1300*.F 101 - -
Temp 1300*-1650*F 50% 10% -
l 1650*-2200*F 100% 50% >10%
Containment .21% 1% - -
H2 Concentration 2.1 % 10% 1% -
10.5 % 50% 10% - j 21.0 % 100% 50% >10% i Footnotes: '
- 1. If all indicators are " normal" no core damage is indicated.
- 2. Equivalent pH at 600*F is 6.8 to 7.8 l 3- 3. 1% Zr-H 2 O reaction will result in 0.21% H2 in containment l
atmosphere assuming uniform mixing. A Zr-H2 O reaction is not expected below 1400*F.
- 4. For boron concentrations'less than 2000 ppm..a shutdown calculation may be required.
l S. Chloride concentration is not an immediate concern, but may cause future l corrosion.
ENCLOSURE 4.1 PAGE 10F 1 ._-
Rev. 3 AP.56-8
.---________________Q
. ENCLOSURE 4.2
- CONTAINMENT AREA RADIATION MONITOR INDICATIONS TABLE 1 MONITOR RESPONSE R-15049 R-15050 Time Interval since Time R/Hr R/Hr Reactor Shutdown, Hours l
TABLE 2 DATA FROM ENCLOSURES 4.10 THROUGH 4.13 i
Enclosure # Response Clad Damage '
Fuel Damage Selected Rance Rance Rance-1 1 R/Hr 1%; minor RCS leak None Between. Curve D 1%; minor to major and 1R/Hr RCS break None Between Curves C Partial Clad damage and D mostly heatup Between Curves B 100% Fuel Damage C Possible l
Between Curves A 100% , Significant B Fuel Damage Above A 100% Meltdown n
HQII:
If the time interval since reactor shutdown is more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> extrapolate the data to the time of shutdown (10-2 on X-axis) by plotting a curve through the data point parallel to the neighboring curve (s). Enter extrapolated data-in Table 1.
ENCLOSURE 4.2 PAGE 1 GT 1 Rev. 3 O AP.56-9 1
L______________ _
i
{
ENCLOSURE 4.3 ']--
NORMALIZED POST-ACCIDENT RADIONUCLIDES SAMPLES (1) (2) (3) . (4) (5) (6)
MEASURED SAMPLE DILUTION TEMP / PRESS POWER NORMALIZED 3-- 0 ISOT.PE CONC. X FACTOR X CORRECTION X CORR. - CONCENTRATION pCi/ml (f) FACT. pCi/mi From Step 3.8 l Containment Samples N/A l Xe133 N/A l 135 N/A l KR85m N/A l 85 N/A l 87 N/A -l' 88 N/A ,
Coolant Samples N/A Il31 N/A 132 N/A 133 N/A 134 N/A 135 N/A -
pg Csl34 N/A >
137 N/A 138 N/A ,
Te132 N/A Sr91 N/A 92 N/A Bal40 N/A Mo99 N/A Rul03 N/A Y92 N/A La140 N/A Cel41 N/A 144 N/A Zr95 N/A 97 N/A ENCLOSURE 4.3 PAGE 1 0F 1 Rev. 3 AP.56-10
\ ._ _ ____ _ _ ___ _ -
\
i i
ENCLOSURE 4.4 EXTENT OF CORE DAMAGE FOR Xe 133 CONCENTRATIONS IN CONTAINMENT ATMOSPHERE OR REACTOR C00LAFf AT RANCHO SECO FULL POWER HISTORY l
l
- 0 d 10 2 10*I 10 0 10 I 10 2 10 0 10 10 3 10 10 6 3n 5 3o 10'3
, , 100 100 , , , , , , ,
f ACCIDENT
- to to -
- 1 1 _ _ _ _ _ _ _ . _ _ _ _ _ '
CNE
} l
==--- ASSEMBLY ,
l * -
10 '
10'l j
l
" 10 2 e 10 2 e? CNE -
5 A' :
A 10'3 10 3 E
$ 10*
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/ \
RANGE " 10
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d O l g
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E 10'5 -
- . l 2
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- 10'8 10'8 1
10*I 10'I -
l
. . . . l , ,
5 6 10 0 10' 10- 10-4 1 10 10 10 0 10 5 10 10'3 10'2 10*l 10 NCRMALJC2D SAMP'.5 ACT1Vl*Y CCNC3NTRATICN Xe 1 3 y C*/mi ENCLOSURE 4.4 PAGE 1 0F 1
- Rev. 3 -
AP.56-11
4 i
. k i l
i 3
.. S )
l ENCLOSURE 4.5 EXTENT OF CORE DAMAGE FOR i I 131 CONCENTRATIONS IN REACTOR COOLANT AT {
RANCHO SECO FULL POWER HISTORY I i
0 2 3 # 5 10 3
10 6 $g.5 10 4 10'3 10 2 go 1 10 10 1
10 10 10 10 100 , , , , , , , , . .
- 10 to - ACCIDEN-3 ______---
~ '
T _,
ASSEMBLY-I . 10 1
- 10 1 ,
a l
m , o.2 _
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y
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a
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10 2 10 10 1 o.6 10 5 10 j
NORMALIZED SAMPLE ACTIVlW CONCENTRATION I 131 ;;Cimi i
ENCLOSURE 4.5 PAGE 1 0F 1 I l
Rev. 3 AP.56-12
1 l
3-ENCLOSURE 4.6
., EXTENT OF CORE DAMAGE FOR l' Cs ISOTOPE CONCENTRATIONS IN REACTOR COOLANT AT RANCHO SECO FULL POWER HISTORY 10 6 10 5 lo d 10'3 10 2 10'l 10 0
10 I 10 2
10 0
10 10 3
10 6
100 , , 100 10 ACO! CENT - 10 l
1 _ _ _ . _ Y y 3 ;
l ,
CNE j i e ASSEMBLY ;
10'l -
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10
p9 l r y -
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' ~
.4CRMALJZED SAMPLE AC*1VITY CONCENTRATION Cs :: C*/mi ENCLOSURE 4.6 PAGE 1 0F 1 Rev. 3 AP.56-13
. I ENCLOSURE 4.7 EXTENT OF FUEL MELTING OR HIGH OVERHEATING FOR Mo AND Ru 103 CONCENTRATIONS IN REACTOR COOLANT AT RANCHO SECO FULL POWER HISTORY 3 #
10'I 10-6 10 5 40 10*3 10 2 10*1 10 0 10 1
10 2 10 10 10 5 100 , ,
100 10 -
10 1 -
-1 ACC: CENT -: CNE ASSEMBLY- >
01 -
10 <
. E
,0 2 -
p -
e g ,0
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g I
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2 $g 3 # 3 10*7 10 6 3g 5 10 4 10*3 10 2 10*l 10 0 10 I 10 10 10 NCRMALCED SAMPLE AC";VITY CCNCENTRATICN u C:/mi FC:1 Mo 59 AND Ru 102 ENCLOSURE 4.7 PAGE 1 0F 1 i Rev. 3 AP.S6-14
, '. j
.. A ;
ENCLOSU.RE 4.8 l RADI0 ISOTOPE DETERMINATION OF CORE DAMAGE FROM CORRECTED AND NORMALIZED PASS RESULTS (a)
(1) (2) (3)
Grachical Estimates (b)
Clad Fuel Melt 'I Isotone Ruoture +0verheat 1
Containment Air Only Xe133 135 l
Kr85m 85 87 88 Reactor Coolant Only 1131 132 133 134 135 Cs134 137 138 Te132 Sr91
(- '
92 Ba140 '
Mo99 Rul03 Y92 La140 Ce141 ;
144 Zr95 97 NOTES:
(a) From Col. 6. Enclosure 4.3 (b) Express results in percent; assume all activity due to clad rupture in ,
Col. 2, then assume all activity due to fuel melt-overheat in Col. 3; )
use Enclosures 4.4 through 4.7.
I ENCLOSURE 4.8 PAGE 1 0F 1 I l
Rev. 3 O AP.56-15
t ENCLOSURE 4.9 ,
ISOTOPIC DATA HALF-LIFE ISOTOPE (nr.)
Epble Gases Xe133 1.26 + 2 (>3 days) 131m 2.86 + 2 (>3 days) 133m 5.26 + 1 (>24 hours) 135m 2.55 - 1 (<24 hours) 135 9.09 + 0 (<24 hours) '
137 6.42 - 2 (<24 hours) 138 2.37 - 1 (<24 hours)
Kr83m 1.86 + 0 (<24 hours) 85m 4.48 + 0 (<24 hours) l 85 9.40 + 4 (>3 days) 87 1.27 + 0 (<24 hours) 88 2.84 + 0 (<24 hours) 89 5.25 - 2 (<24 hours)
Haloaens I131 1.93 + 2 (>3 days) 129 1,39 + 11 (>3 days) 130 1.236 + 1 (<24 hours) gC7g 132 2.29 + 0 (<24 hours) M(dv
(<24 hours)
~
133 2,08 + 1 134 8.77 - 1 (<24 hours) 135 .6.58 + 0 (<24 hours) 136 2.36 - 2 (<24 hours) i Br83 2.40 + 0 (<24 hours) l 84 5.30 - 1 (<24 hours)
! 85 4.78 - 2 (<24 hours) 87 1.56 - 2 (<24 hours) 3- Other Isotoos ,
Cs 134 1.80 + 4 (>3 days)
-Cs 137 2.64 + 5 (>3 days)
Te 132 7.8 +1 (>3 days) !
Mo 99 6.60 + 1 (>24 hours) j Ru 103 9.46 + 2 (>3 days) !
l a
i 4
ENCLOSURE 4.9 PAGE 1 0F 1 l Rev. 3 'q AP.56-16 ;
1 _ _ _ _ _ _ _ _ - _ _ _ _
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Rev. 3 AP.56-20
e 9 ENCLOSURE 4.14 O '
Q BEST ESTIMATE CONCLUSIONS FOR EXTENT OE CORE DAMAGE i
~
CORE DAMAGE RESULTS OF PASS AND SPECTRUM OTHER PLANT INDICATORS l
No Damage l Clad Damage i
Minor (510%) .1 Intermediate (10% - 50%) .
Major (>50%)
1 Fuel Overheat
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Minor-(<10%) {
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'l 3-~ Signature: Date J
ENCLOSURE 4.14 PAGE 1 0F 1 END-Rev. 3 O AP.56-21 I 1
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