ML19273B187
| ML19273B187 | |
| Person / Time | |
|---|---|
| Site: | Duane Arnold  |
| Issue date: | 03/05/1979 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19273B186 | List: |
| References | |
| NUDOCS 7904030008 | |
| Download: ML19273B187 (15) | |
Text
b ONITED STATES j
.:' N NUCLEAR REGULATORY COMMISSION W ASHINGTON, D. C. 20665 s.~....j SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 49 TO LICENSE NO. DPR-49 IWO ELECTRIC LIGHT AND POWER COMPANY CENTRAL IOWA POWER COMPANY CORN BELT POWER COOPERATIVE DOCKET NO. 50-331 DUANE ARNOLD ENERGY CENTER 1.0 Introduction On January 8,1979, the Nuclear Regulatory Commission (NRC) issued a Safety Evaluation Report (SER) pertaining to the safe-end cracking that occurred at the Duane Arnold Energy Center (DAEC) on June 17,1978 and to the repair program initiated by the licensee to replace the damaged safe-ends with safe-ends of an improved design. This SER was transmitted January 8, l(lectric Light and power Company by a letter also dated to the Iowa W9.
There re,aained, however, three items requiring resolution before operation under the amendment could commence; these are (1) a finding that the testing had been conducted in conformance with approved procedures; (2) satisfactory completion of a hydrostatic test of the repaired safe-ends in accordance with applicable ASME Code requirements; and (3) completion of the licensee's audit of the safe-end repair work and resolution of any discrepancies identified in this audit. These three items were reiterated in a letter to the licensee dated January 16, 1979 from the NRC's Region III Office of Inspection and Enforcement.
Item (1) and the part of item (3) regarding the licensee's audit of the safe-end repair work have been completed to the satisfaction of the Office of Inspection and Enforcement; this will be documented in a subsequent inspection report.
Item (2) must be completed to the satisfaction of the NRC's Region III Office of Inspection and Enforcement before the facility is authorized to exceed 5% rated power.
In relation to those aspects of item (3), not completed by I&E there are several matters that required evaluation because of questions concerning certain discrepancies in the actual performance of the safe-end repair program This safety evaluation assesses the safety s;gnificance of the following information:
'tadiographic (RT) examination of the new pressure boundary welds A.
indicated that some of the weld root inside surfaces are very irregular, possibly caused by difficulty with uniform melting of In addition, obvious the insert used for the first welding pass.
oxidation in local areas was noted in the radiographs, indicating that the inert gas purge was not uniformly effective.
7 9 04 0 3 0 00f
. B.
During the initial leak test, performed after all repair operations, a flow blockage was noticed in one recirculation riser and jet pump assembly.
An investigation showed that the flow blockage was caused by a temporary radiation shield plug used to protect personnel during the welding operation and inadvertently left in the pipe near the nozzle.
This plug consisted of a thin aluminum and carbon steel can filled with shaped lead blocks.
During the leak test, water flow in the line pushed the plug into the jet pump assembly where it came apart.
Retrieval operations recovered most of the can and all of the lead blocks.
One small lead block went through the jet pump ard down to the bottom of the reactor vessel, but it was re-covered also.
Because all the lead blocks were recovered, the only potential safety significance of this event is that Alloy 600* may be subject to stress corrosion cracking at reactor operating temperature due to residual lead smears or high concentrations of lead in the In addition to the Alloy 600 nozzle safe end and thermal water.
sleeve that could have been contaminated by contact with the lead blocks, the control rod drive stub tubes in the lower head of the reactor vessel also are made of Alloy 600 and may have been contami-nated by the lead block that fell to this region.
All other hardware that could have been contaminated by contact with the lead blocks is made of stainless steel, which is not affected by lead contamination.
Several small metal tabs of carbon steel and a part of the thin aluminum backing plate from the can used to contain the lead blocks were not recovered.
This SER addresses the potential effects of the relatively poor weld root geometry of some welds, the effect of possible lead contamination, and the Dossible effects of the loose pieces of metal on the safe operation of the plant. These issues were addressed in the licensee's submittal of February 22, 1979, as supplemented by letters dated March 1, 1979 and March 3,1979.
- 2. 0 Discussion and Evaluation 2.1 Weld roots 2.1.1 Characterization of weld root conditons Although there were differe.ces of opinion regarding the detailed weld rest geometry and conditons as determined from interpretation of the radiographs, it was agreed that the evidence clearly showed a wide range of conditions existed from weld to weld and even from place to place on some individual welds.
Therefore, the licensee charac-terized the root conditions at weld numbers 2 and 6
- The Alloy also is know as Inconel 600.
. (see Figure 1) by assuming and analyzing six different notch configurations (see Figures 2 and 3).
These notch configura-tions were chosen by the licensee to conservatively bound, relative to mechanical strength, the worst root conditions that could be inferred from the interpretation of the radi-ographic films.
The worst-case condition enveloping all the stated variations is a sharp cup-shaped intrusion, with a depth of about 1/10 the wall thickness.
This condition is designated as Case E* in the licensee's analyses.
There are three possible adverse effects of this assumed worst-case root condition; adequate fatigue life, initiation of stress corrosion cracking, and initiation of brittle or unstable fracture without prior warning.
These three possible effects are discussed below.
2.1.2 Fatigue analysis The licensee performed detailed stress analyses for weld locations 2 and 6 in accordance with the requirements of Section III of the ASME boiler and Pressure Vessel Code assuming the notch configurations shown in Figures 2 and 3.
The results of the original stress analyses were modified to account for the reduced wall thickness at the assumed notches and to obtain primary stress intensities and primary plus secondary stress ranges.
No credit was taken for any weld reinforcement at the outer weld surface.
In addition, theoretical stress concentration factors were determined for Cases A through D and F for the calculation of the fatigue usage factors.
For Case E, a fatigue strength reduction factor of 4.0 was used which is the highest value required by the ASME Boiler and Pressure Vessel Code for a partial penetration weld.
The analysis of weld numbers 2 and 6 for each of the assumed r.otch configurations, considering the reduced wall thickness and the fatigue reduction factors, show that the calculated stresses and fatigue usage factors are within the limits s
prescribed by Section III of the ASME Boiler and Pressure Vessel Code.
The staf f has evaluated the fatique analysis and agrees with the licensee's evaluation.
T The letters A through F refer to the notch configurations analyzed by the licensee as shown in Figures 2 and 3.
, 2.1.3 Stress Corrosion Because there were concerns that the weld root irregular-ities could be severe enough to act like crevices to initiate stress corrosion cracking, the licensee performed a stress corrosion crack growth rate analysis.
In this analysis he assumed that stress corrosion cracks would be initiated, even though the worst case root condition postulated is not nearly as severe a crevice as was present in the original design.
The results of these analyses show that the crack growth rate in terms of fractional depth through the wall as a function of time would be slower than in the old design.
The staff has considered the possibility of stress cor-rosion cracking initiating at the irregular weld roots, and considers that it is a fairly remote possibility, because the root irregularities are very unlikely to be deep enough and tight enough to cause the crevice chemistry conditions necessary to initiate stress corrosion cracking.
Even the worst postulated case, that of a cusp-shaped defect 1/10 the wall thickness at local areas around the circumference, is nowhere near as severe a crevice from a stress corrosion standpoint as the built in deep crevice of the original design.
Although the staff has reservations regaraing the bases for the stress corrosion crack growth predictions submitted by the licensee, we conclude that even if stress corrosion cracking should initiate at local spots around the circum-ference, crack propagation would be no faster than the original stress corrosion cracks, and most likely will be slower, because the nominal stress levels at velds 2 and 6 are lower than those at the cracked section of the original safe ends.
2.1.4 Brittle or Unstable Fracture Alloy 600 is a very ductile and tough material that is very resistant to brittle or unstable fracture.
The licensee performed analyses to justify the assumption that complete failure of the new safe end and pipe welds will not occur unless an extremely large crack is present, and that the worst case weld root irregularities could not initiate failure of the weldments.
This was done using a net-section stress analysis of the new safe end design with a postulated crack.
In this approach, it is assumed
. that a pipe (or a safe end) of highly ductile material with a crack is at point of incipient failure when the stress in the remaining ligament ahead of th crack reaches the flow stress.* Numerous test have been performed by Battelle, General Electric and others using pre-cracked pipes under both pressure and bending loads.
The results of these tests validate the efficacy of this approach and indicate that a typical pipe of tough material can tolerate a crack of sufficient size to make detection of a crack by inservice inspection of by the leak detection system highly likely before a critical crack size is reached.
The staff agrees with the method of analysis and with the licensee's conclusions.
The staff also used information available from the recent Pipe Crack Study Group report.
In this report, a more sophisticated analysis was performed utilizing the tearing stability concept and the associated tearing modulus stability criterion.
Based on this analysis and the analyses performed by the licensee and the staff, we conclude that is is unlikely that safe end cracks or postu-lated cracks emanating from irregular weld root geometries, should they occur and be missed by inservice inspection, will result in unstable crack growth and excessive loss of coolant.
In addition to the confidence gained from the above analyses, the inservice inspection program and the leak detection system, the emergency core cooling system is designed to keep the plant in a safe condition even postu-lating the failure of the largest pipe in the reactor system.
2.2 Lead contamination The radiation shields used in the nozzles to protect welders from radiation emanating from the reae. tor vessel were made of shaped lead blocks enclosed in a thin sheet metal can.
The circular end piece was made of 0.016" thick aluminum and the cylindrical portion of 0.016" thick carbon steel sheet metal.
A cause for concern about lead smear contamination results from the one shield plug that was not removed before the pipe closure welds were made.
The plug was forced by water flow into the elbow and jet pump assembly where it came apart.
Because some of the lead blocks fell back down the jet pump riser pipe, there is a possibility that lead came into contact with the Alloy 600 portion of the thermal sleeve.
In the case of the single lead block that passed through the jet pump and was found in the bottc.n head of the vessel, contact between it and the Alloy 600 control rod drive stub tubes must be assumed.
Although contact with stainless steel components could also result in smears of lead contamination, there is no concern regarding
- lhe flow stress has been shown by various experiments to be approximately the average of the yield stress and te ultimate stress of the material,
. these components other than that the smears would add to the total inventory of lead in the system.
Tests were run by General Electric to determine the amount of lead that could be deposited on such components by forceful impact.
They reported that the maximum smear was measured to be 0.0001 inch thick.
It is unlikely that significantly thicker deposits would have occurred in the actual incident because of the cushioning effect of the water in the vessel and the comparatively small mass of the actual lead blocks.
At plant operating temperatures (about 550 F) in flowing water the lead smears, if present, will dissolve in about two days.
The resulting concentration of lead in the reactor water would be very low due to the large mass of water and the continuous clean-up.
Except for Alloy 600, all reactor and fuel materials (carbon steel, stainless stael and zircaloy) are not affected by the presence of lead or lead compounds in water.
Although lead causes stress corrosion cracking in nickel-base alloys, relatively high concentrations are As determined from the open literature, the measured time necessary.
to initiate SCC in Alloy 600 loaded to stresses above yield in BWR conditions is three weeks or more.
Thus the smears will be oxidized or dissolved in less time than that neces:,ary to influence the Allcy 600 corrosion behavior.
The reactor water clean-up system will remove the lead or its compounds.
We conclude that with the successful retrieval of all tne lead blocks that any minor amounts of lead that may have been left in the system as lead smears will not represent a safety concern.
2.3 Loose Parts Evaluation The licensee wishes to commence operations with some loose objects that cannot be accounted for and are assumed to be somewhere in the vessel.
The objects consist of a fragment of aluminum, which were 16 small carbon steel tabs, torn from the protective cannister.
2.3.1 Aluminum Fragment The aluminum section of the can used to contain the lead blocks has the shape of a segment of a circle of diameter 9.75" with a chord length of 5.25" and a thickness of 0.016" This section is presumed to have passed through one of the jet pumps into the bottom region of the reactor vessel.
It may have done so as one deformed piece or as two or more fractured pieces.
In any event, while the recirculation system is in operation, the water velocities and degree of turbulance in the bottom region of the vessel are tufficient to keep the aluminum piece or pieces levitated so c. hat most likely it or they would follow the streams of water flow toward the fuel assemblies and either be jammed in the fuel support assembly or further broken up by hydraulic forces to pieces small enough to pass into the fuel.
. Aluminum loses its strength at elevated temperatures and, at 540 F operating temperatures, is approximately as strong as, but more flexible than, typing paper.
Because of this low strength, we estimate that approximately one pound of force would drive the fragment through a fuel support casting orifice.
Actual hydraulic forces are about a factor of twenty greater than this, and therefore we conclude that the fragment, even if crumpled and wedged such that it had sufficient area to eclipse an orifice, would not be likely to obstruct orifice flow more than momentarily.
Once past the orifice, the fragment woJld come to rest against the lower tie plate of a fuel assembly.
If com-pletely flattened out, the fragment has an area of about 3 inches square.
Thus, the fragment would not cause the 86%
area restriction GE has determined necessary to cause ONB at full power.
[NED0-10174, rev. 1]
It is expected that the fragment will break up into smaller particles which would be drawn up into the fuel bundle.
Such a particle, small enough to pass the lower tie plate, would not cause any significant change in local flow.
Cross flows would eliminate any perturbation, and in any case the flow perturbation would be of the same order or smaller than normal perturbation caused by steam bubbles.
If a small particle were to become caught at a grid spacer, no problem would likely result since the grid spacers normally reduce power in fuel rods in their locality.
Particles carried through to the core exit need not be considered further, as they would at worst be carried through to the lower plenum again.
In any case, the licensee has demonstrated, by autocicve tests of samples of aluminum recovered from the shield cannister and samples of stock from which the cannister was faorv.ated, that the aluminum fragment will completely corrou awcy to Al 03 (as a fine powder) after exposure to wateratBWRopera$ingtemperaturesinabout156hoursor less.
The lic7nsee has proposed to operate first at 5%
power (to dissolve the lead smears discussed earlier), then at 25% power until 156 hours0.00181 days <br />0.0433 hours <br />2.579365e-4 weeks <br />5.9358e-5 months <br /> at 540 F have been accumulated.
Only then will power be increased to rated.
On the basis of the above discussion we conciude that the aluminum fragment will not cause a safety p oblem due to flow blockage.
. 2.3.2 Carbon Steel Tabs The carbon steel tabs will retain their integrity and will not corrode away for several years.
They are light enough to be carried up to the fuel support casting orifices, just like the aluminum fragment.
Therefore, flow blockage at the orifices and at the lower tie plates must be considered.
Except in the case of peripheral assemblies, the tabs are too small to block the orifices and should pass freely through.
The licensee has conseratively calculated the probability of peripheral crifice blockage by one tab to be 4% (peripheral orifices are smaller).
One tab can obstruct up to 81% of the flow area, which is slightly greater than the 79% DNB threshold discussed in NED0-10174, rev. 1).
However, peripheral bundles operate at much lower power, generally 2/3 or less of that of the "avrcage" bundle.
Therefore, transition boiling is unlikely even if a tab does obstruct a peripheral assembly's orifice.
We therefore conclude that local fuel damage is highly unlikely to result from orifice blockage.
Once past an orifice, a tab would remain at the lower tio plate.
GE has calculated (NE00-10174 rev. 1) that 78% area blockage at the lower tie plate is necessary to cause transi-tion boiling in an assembly.
This tolerance to blockage is due to the automatic reduction in oower caus.d by reduced flow, due to the other structures and orifices also contri-buting to hydraulic resistance, and also due to some leakage flow to the assembly.
Although we have not yet accepted
[NE00-10174, rev. 1], we note that all the tabs collectcd at one assembly in the worst assumed distribution would cause at most a 66% blockage, much less than the 78% calculated to cause transition boiling.
The low probability of such a blockage occurring in a limiting assembly at the worst time of life simultaneous with an abnormal operational occurrence is low.
Therefore, we find the conclusion that no adverse flow blockage will occur to be acceptable.
2.3.3 Other Loose Object Con,siderations The loose objects are too small to cause mechanical damage due to impacting or abrading the rather massive components located in the lower plenum.
Coolant velocities vary, but average about 15 teet/second.
The impact of a steel tab or aluminum fragment moving with t.he coolant is therefore equivalent to dropping it several feet in air.
When carried up into the chamber below a lower tie plate, a loose object can more credibly cause damage due to the increased turbulence.
However, test data indicates that one fuel rod can absorb 250 ft-lb without damage when struck from the end.
[NEDO-240ll, " Generic Reload Fuel Application,"
May 1977] The parts of concern would have impact forces of less than 1 ft-lb.
We therefore conclude that mechanical impact damage will not pose a problem.
Mechanical interference with moving parts must also be considered.
The only moving parts within the lower plenum and in contact with coolant are the control blades and their drives.
The blades are located within guide tubes, the inside of which does not communicate with lower plenum water.
Thus, there is no way for a loose object to enter the support column, and mechanical interference is precluded.
Similarly, the control blade drives do not communicate with lower plenum water except through a ball check valve which is too small to admit the tabs and which is normally closed in any case.
Therefore, there is no significant possibility of mechanical interference with the control blades and their drives.
Based upon the conclusions above, we conclude that the presence of the loose objects will not pose a safety problem when the plant is operated.
3.0 Augmented ISI The repair and redesign of the safe end thernial sleeve assembly elimi-nated the primary cause of the original stress corrosion cracking, namely, the built-in crevice between the non-welded portion of the facing surfaces between the thermal sleeve and the safe end.
Nevertheless, to provide additional assurance that unrecognized factors resulting from irregularities in the root passes of the welds do not exist that could cause future cracking of the pressure boundary welds in the new safe end assembly, the staff has required, and the licensee has committed to, an augmented inservice inspection program for these welds.
All pressure boundary welds (designated as nutbers 2, 6 and 7) were subject to on ultrasonic examination to provide a base line for future examinations.
Complete recordings were made of these examina-tions to ensure that any changes in ultrasonic results indicative of cracking during service will be identified.
The specific program that will be followed will be to ultrasonically examine all three welds in one half (four) of the eight safe end assemblies every refueling outage.
This program will continue at least until every weld involved will have been inspected twice.
As the Duane Arr.old plant is on a yearly refueling schedule, this means that welds in four of the safe end assemblies will be inspected after one and three years of operation, the remaining four will be inspected after two and four years of operation.
The NRC staff will require that the licensee submit his criteria for each of the above inspections at least 30 days prior to the inspections.
We will require that the safe end assemblies selected for the first inspection be those deemed to have the poorest weld root conditions.
The license will be modified by changing the Technical Specifications to reflect these coaclusions.
A decision on whether or not it will be necessary to continue or man;fy this program will be made after all welds have been examined twice, - cording to the schedule described above.
The plan for the augmented inservice inspection of the safe end volume away from pres-sure boundary welds, will incorporate conclusions reached after analysis of the Pipe Crack Study Group report (to be issued) as it relates to Alloy 600 safe ends.
- 4. 0 Leak Detection The technical specifications governing nuclear facility operation require that certain leak detection systems be functioning during operation and imposes limits on the amount of leakage that may be permitted. When these conditions 6 annot be met, timely remedial measures are required, including possible shutdown of the facility.
In NUREG-0313*, the NRC staff recommended that facilities that cannot meet the guidelines stated in Part lI of that document (i.e., facili-ties with piping materials susceptible to ISSCC) have augmented leak detection requirements.
Specifically, plant shutdown should be initi-ated for inspection and corrective action when the leakage system indicates, within a period of four hours or less, an increase in the rate of unidentified leakage in excess of 2 gpm, or when the total unidentified leakage attains a rate of 5 gpm, whichever occurs first.
The recent Pipe Crack Study Group reconsidered this requirement and concluded that it is still appropriate.
The licensee states that since i N uary 1975 the DAEC has implemented the criteria as required in NRC IE Bulletin No.74-10B which are essentially in agreement with the NUREG-0313 augmented leak detection recommendations.
The primary leakage detection system for the DAEC containment is the Drywell Floor Drain Sump system for which the licensee states a four-hour interval is required to obtain aa accurate leak rate measurement.
Based on information obtained via this sys.em, the DAEC operators
- Technical Report on Material Selection and Processing Guidelines for BWR Coolant Pressure Boundary Piping.
, concluded that the June, 1978, safe end crack resulted in approximately a two gpm increase in unidentified leakage over about a twelve-hour period (not within a period of four hours or less) and hence immediate shutdown was not required.
Information obtained from other detection systems, although not quantitative in terms of gpm, could be inter-pretated to indicate that the actual safe end leak developed over a period less than twelve hours and possibly less than four hours.
There are inherent time limitations associated with the various types of leak detection systems.
For instance, assume the instantaneous development of a two gpm leak from the crimary system at operating tempera'.ure and pressure.
A portion of this two gpm will flash to steam und be absorbed by the atmosphere or condense on cooler struc-tures and equ'ement and most likely will take some time to reach the sump.
Even that portion of leakage exiting directly as water may have to traverse a long a tortuous path to the sump.
Altnough other leakage detection systems can provide more prompt information, they too have limits such as associated with mixing times of steam and its associated entrained radioactive materials with the containment atmosphere.
The DAEC licensee has committed by letter dated March 1, 1979, to modify his present operating procedure by changing the words "
.witnin a period of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or less.
" to read "
.within a period of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or less.
" and not alter the other words.
The sump level will continue to be monitored at four-hour intervals.
This interim comn:it-ment will remain in effect until the licensee can satisfy the NRC that modifications or additions to his present leakage detection systems, which 5e may propose, will increase the sensitivity of his systems and aid i discriminating against non-critical sources of leakage such is throush valve stem packings, or unless modified in accordance with '.h:
provisions of the applicable Technical Specifications.
5.0 Evaluation Summary Except for the increase in the ISI, the matters discussed abnyc do not affect the presently approved design and operating license amendment authori zation.
They relate in the main toward further evaluation performed concerning certain questions which arose in the completion of the repair work.
The changes in ISI discussed above involve an amendment to the Techr; cal Specificctions.
. 6.0 Environmental Consideration _s We have determined that this amendment does not authorize a change in effluent types or total amounts nor an increase in power level and will not result in any significant environmental impact.
Having made this determination, we have further concluded that this amendaent involves an action which is insignificant from the stand-point of environmental impact and pursuant to 10 CFR 51.5(d)(4) that an environmental impact statement, or negative declaration and environ-mental impact appraisal need not be prepared in connection with the issuance of this amendment.
7.0 Conclusion We have concluded: (1) because the amendment does not involve a significant increase in the probability or consequences of accidents previously considered and does not involve a significant decreese in a safety margin, the amendment does not involve ' significant hazards consideration, (2) there is reasonable assurance t.at the health and safety of the public will not be endangered by operat'on in the proposed manner, and (3) such activities will be conducted in com-pliance with the Commission's regulations and the issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public.
Date : March 5, 1979
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FIGURE 2:
Analyzed Hotch Configurations at Root of Safe End to Piping Wcld (Weld #6) g CASE NOTCH CONFIGURATION
% OF WALL STRESS CONC.
FACTOR 2.14%
2.93 A
,7a:
I
)
L
.015"/7\\
d I
.015"r B
.7" 2.86%,
2.90
~
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.02"r 5%
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.7a iTs77%
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.035"r 10%
3.3 l
.035"r
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.07I 1
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. 7'1 10 4,0 E
.07" A I
t i
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l.7" 3.86%
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6 k
.027" [
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.087"r.
FIG E 3:
Analyzed Notch Configuraticas at Roct cf Nozzle to Safe End Weld (Weld #2)
CASE NOTCH CONFIGURATION
% OF WALL STRESS CONC.
FACTOR i
.96" I
1.56%
3.1 A
.015 Q I
.015"r T
.96"*
2.08%
2.9 B
.02" @
I
.02"r
.96" 5%
2.8 C
.048"/~/N
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.048"r
.96" 10%
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.048"r l
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.125"
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