ML20002D895
| ML20002D895 | |
| Person / Time | |
|---|---|
| Site: | Big Rock Point File:Consumers Energy icon.png |
| Issue date: | 09/11/1970 |
| From: | Haueter R CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| To: | Morris P US ATOMIC ENERGY COMMISSION (AEC) |
| References | |
| NUDOCS 8101230327 | |
| Download: ML20002D895 (10) | |
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Dr. Peter A. Morris, Director C
/s/ Re : Docket No 50-155 United States Atomic Energy Commission i '$1qff DPR-6 Division of Reactor Licensing Regulate y Fi,.e Cy.
Washington, DC 20545
Dear Dr. Morris :
Enclosed are LO copies of the response to your letter of August 6, 1970.
Listed below are Consumers Power Company's answers to the nine questions asked in your letter of August 6, 1970:
Question 1 - Identify all known sensitized stainhss steel components of and within the reactor coolant pressur9 boundary,* including portions of piping. Include furnace-sens1cized components affected by substantial field stress relienng, but not the heat affected zones caused by field welds. State location, type of material and sensitization process.
Answer - Please refer to the attached data sheets, Columns 1, 5, 6 and 7 Question 2 - Specify the maximum stress levels (calculated or measured, if known) these components receive in service.
Indicate whether the calculations or measurerents of stress level were based on the "as-built" condition, including effects of "as-insta' led" piping hangers and restraints. Summarize the results of any field measurements of piping displacement that have been per-formed, including the system ct.nditions for which the measure-ments were made.
Answer - Stress level information will be supplied by January 15, 1971.
The acceptance test for the nuclear steam supply system included measuring constant support hanger deflections in the vertical direction, the movement of the steam drum by use of temporary trams at each end of the drum, the horizontal movement of each reacto-ecirculating pumn by use of a plum bob and the movement of the horizontal cowncomer headere also by using plum boba. These measurements were obtained from the cold and empty condition to the filled-with-water 365 condition. The results of these tests are as follows :
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Ty Constant Support Hanger Data 3~
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Travel With Drum Total 3*t Hanger Equipment IMsign Cold &
Filled to Normal Level Travel-g'>
M Number Supported Travel Empty 1C00 1500 2000 250 E 350 365_
Flooded H-lOIA Drum Northwest 2.25" o.0"
-0 3" o.0" o.2" 0.2" o.6" o.5" o.9"
- 1. 2 "
E&n H-101B Drum Northside 2.25" O.0"
-0.2"
-0.1" O.2" O.2" O.6" O.8" o.9" 1.1" g ;;-
Y H-lOlC Drum Northside 2.25" O.0"
-0.2"
-0.1" 0.1" O.2" O.h" O.7" O.8" 1.0" H-101D Drum Northeast 2.25" O.0"
-0.2" 0.0" O.2" 0.2" O.6" 0.8" O.7" O.9" H-101E Drum Southwest 2.25" 0.0" O.0" O.2" O.2" 0.2" O.7" 0 7" 09" 0 9" H-101F Drum Southside 2.25
O.0" 0.1" 0.2" O.2" 03" 07" 0.T" o.9" o.8" H-10lG Drum Southside 2.25 0.0" 0.1" 0 3" 03" 0 5" 0 9" 1.0" 1.1" 1.0" H-101H Drum Southeast 2.25" O.0" 0.1" O.3" 03" 03" 0.T" o.8" o.9" o.8" H-lC2
- 1 Riser Above 1.63" 0.0" 0.1"
-0.1"
-0.1" 0.2" o.2" o.3" o.2" c.1" H-lO3
!6 Riser Above 1.63" 0.0"
-0 3"
-0.1"
-0.1" 0.2" 0.2" o.2" O.3" o.6" H-104
- 3 Riser Above 1.63" 0.0"
-0 3" 0.1" 0.2 "
0.2" 0 5" o.7" o.6" 0 9" H-lOS
- 4 Riser Above 1.63" 0.0" 0.0" 03" o.2" 0.h" 0.6" c.8" o.7" 0 7" H-lc6
- 5 Riser Above 1.63" 0.0" 0.0"
-0.1"
-0 2 '
00" 0.T" 03" 03" 03" 0.0"
-0.2" o.0" 0.2" 0.2" 0 5" 0.8" 0.8" 1.0" fg H-107
- 2 Riser Above 1.63",,
O.0" 0.1" 0 3" 0 5" 0.6" o.8" o.9" o.9" o.8" H-1A fl Pump Suction 1 75"'
H-1B Y2 Pu: p Suction 1 75" O.0" O.1" 0.3" 0 5" 0.7" 0.8" o.9" o.9" O.8" H-2A
- 1 Suct Valve 1.69" o.0" o.2" 0 5" o.7" o.7" o.8" 1.0" 1.0" o.8" H-2B y2 Suct Valve 1.69" o.0" o.2" o.7" o.5" o.7" 1.1" 1.0" 1.2" 1.0" H-3A
- 1 Pump Inlet 1.69" O.0" 0.6" O.6" O.8" 0.8" 1.1" 1.1" 1.1" O.5" H-3B
- 2 Pump Inlet 1.69" O.0" 03" o.6" 0.6" 0.7" 1.0" 1.0 "
1.0" o.7" H hA fl Pu p outlet 1.63" 0.0" 0 7" 0.8" o 9" 1.0" 1.1" 1.2" 1.h" o.7" m
H-hB
- 2 Pu p Outlet 1.63" 0.0" o.6" 03" o.6" 0 7" 0.8" 1.0" 1.0" o.h" H-5 El Butterfly 1.63" o.o" o.6" o.3" c.6" 0.7" 0.8 1.l~
1.1" o.5" Above travels in vertical directicn only, based on pcsition of hanger movement indicators.
Negative values indicate overtravel in cold directica upon filling system with water.
Note:
L Travel With Drum Total
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Hanger Equipment Design Cold'&
Filled to Normal Ievel Travel-Number Supported Travel Empty 1000 150 2 coo 250 300 350 365 Flooded Q
H-6A
- 1 Disch Valve 1.63" O.0" O.2" O.8" O.6" o.9" 1.0" 1.0" 1.2" 1.0" I(
H-6B
- 2 Disch Valve 1.63" o.o" o.6' o.6" 0 7" o.7" o.8" 1.0" 1.0" O.h" g"
H #2 Butterfly 1.63" 0.0" o.h" o.8" o.8" o.8" 1.0" 1.2" 1.2" o.8"
-y' e-H-8A
- 1 Disch Pipe 1.19" O.0" o.5" o.6" o.7" 0 9" -0 9" o.9" 1.2" o.7" J=-
H-8B 82 Disch PJpe 1.19" o.o" o.6" o.5" o.7" o.9" 1.0" 1.2 "
1.2" O.6" e
H-9A
- 1 Bypass Valve 1.44" O.0" O.6" O.7" O.7" O.7" O.8" O.9" O.9" U.3" H-9B
- 2 Bypass Valve 1.hh" O.0" O.6" O.8" O.6" O.7" O.8" O.9" O.9" O.3" H-LOA
- 1 Disch Pipe 1.00" O.0" O.1" O.5" O.5" O.5" O.6" O.6" O.7" o.6" H-10B
- 2 Disch Pipe 1.00" O.0" O.1" O.h" O.h" O.5" O.5" O.6" O.7" O.6" H-11
- 1 Bypass Pipe 1.h4" O.0" O.3" O.6" O.T" O.6" O.6" O.9" O.5 0.6" H-12
- 2 Bypass Pipe 1.44" 0.0" O.3" 0.1" O.h" O.7" O.7" o.9" o.9" 0.6" H-13A
- 1 Pump Mtg 0.0" O.0" O.0" O.0" O.0" O.0 "
O.0" O.0" O.0" O.0" H-13B
- 2 Pump Mtg 0.0" O.0" O.0" O.0" O.0" O-1" O.1" O.1" O.3" 03" H-lhA
- 1 Pump Mtg 0.0" O.0" O.0" O.0" O.1" O.1" O.1" O.2" O.3" 03" 2
H-14B
- 2 Pump Mtg 0.0" 0.0" O.0" O.0" O.1" O.1" O.1" O.1" O.3"'
O 3" H-15A
- 3 Riser Below 0 38" o.o':
0.0" o.1" o.2" o.2" o.2" o.2" o.2" o.2" H-15B
- 4 Riser Below 0 38" o.0" 0.2" o.2" o.2" o.2" O.2" O.2" 0.2" o.2"
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H-16
- 2 Riser Below 0 50" O.0" O.1" O.5" O.2" O.0" O.0" 0.1" O.2" O.1" H.17
- 5 Riser Belov 0 38" o.o" o.1" 0.h" O.o" o.o" o.o" o.o" o.o"
-o.1" H-18A
- 1 Riser Below 0 38" 0.0" 0.0" 0.0" 0.0" O.0" 0.0" 0.1" 0.1" O.1".
H-18B
- 6 Riser Below 0 38" 0.0" 0.0" O.0" o.0" 0.1" 0.1" O.1" o.2" o.2 "
H-19A West Downcomers 1 75" O.0"
-0 3" 0 5" 0 7"
'O 7" 09" 1.0" 1.0" 13" H-19B East Downcomers 1 75" O.0" O.1" O.3" 0.h" 07" 07" 0.7" o.9" o.8" H-20 Suct Crossover 1.87" O.0" O.1" O.7" O.7" O.7" 1.1" 1.1" 1.1" 1.0" Note: Above travels in vertical direction only, based on position of hanger movement indicators.
Negative values indicate overtravel in cold direction upon filling system with water.
M Dr. Petar A. Morris k
'Septrmber 11, 1970
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Steam Drum Movement West End Movement East End Movement Condition West South Up East South Up System Cold and Empty Zero Zero Zero Zero Zero Zero Normal Drum Level, loo Zero 0.19"
.06" Zero 0.25"
.06" Normal Drum Level,150 0.06" o.25" o.37" o.12" o.37" o.12" Normal Drum Level, ' 200 0.19" o.37" o.31" o.12" -0 50" o.31" Normal Drum Level, 250 0 37" o.56" o.50" o.25" o.69" o.56 Normal Drum Level, 300 0.44" o.75" o.75" o.37" o.88" o.69" Normal Drum Level, 350 0 50" o.88" o.94" o.50" 1.00" o.81" Normal Drum Level, 365 0.62" o.88" 1.00" o.50" 1.00" o.81" Design Movement at 600 1.88" 2.25" 1.88" 2.25" c.
Recirculating Pump and Ibwncomer Header Movement No 1 Pump No 2 Pump West Pipe East Pipe Temperature West South East South West South East South 100 0.06" o.06" o.06" Zero Zero s
150 0.12" o.06" o.12" o.12" o.12" Zero Zero 200 0.25" o.12" o.37" o.37" o.25" o.25" o.25" o.12" 200 0.25" o.12" o.37" o.37" o.25" o.25" o.25" o.12" 250 0 50" o.31" o.50" o.k3" 0.19" o.50"-
0.19" o.25" 300 0.43" o.56" o.69" o.37" o.19" o.50" o.19" o.37" 350 0 75" 0 56" o.50" o.50" o.19" o.50" o.25" o.37" 365 1.00" o. %"
1.00" o.50" o.25" o.50" o.25" o.50" During the hot function testing p2 7 gram, the constant support hangers and sway braces were inspected to insure thit they did not impede piping travel caused by thermal expansion. The results af this inspection were satisfactory.
In addition, linear _ motion transducers were used in June 1970 to determine the upward movement of the steam drum from the cold condition (no pump flow) to the hot operating condition at 164 MW. These transducers t
indicated an upward motion of 1 75 inches at each end of the drum.
Question 3 - Specify the normal external operating environment of the components listed above.
Discuss the probability of external surface contact with corrodents.
Indicate the normal water chemistry that has been maintained within the reactor coolst system during both operating and shutdown con-ditions, including the range of values for materials whose concen-trations have varied appreciably. Include measured values of oxygen and halide concentre.tions.
Answer _- Please rer r to Column 8 for the normal external operating i
environment. As all theca nozzles are insulated, the skin temperature should be approximately that of operating temperatures.
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Septtmber 11,.1970 A search' of the construction. records yielded no evidence of external surface contamination by corrodents prior to. plant operation. Since plant
-operation commenced ~in 1962, there has been.no known contamination of these
-components by'corrodents. It is very unlikely that these surfaces vill ever
= become contaminated by corrodents because:
1.
They arf covered by insulation.
2.
No known corrodent source exists in the steam drum and vessel cavities to contaminate them.
-3 Mest are physically inaccessible.
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The normal operating and ' shutdown water chemistry conditions are summarized as follows:
Operating Conditions pH 71 3
Conductivity 0 5 pmho/cm Turbidity-
<0.1 APHA Units (ppm)
U Chloride
<20 ppb Boron.
.25 ppm 1
. Silica 0.15' ppm
- Iodine Activity,
,y DCi[cc 1 x 10 Filtrate, Gross Gamma at 2 Hours, 5
i cp:/ml 3 x 10
- Crud, Gross Gamma-at 2~ Hours, cpm / Turbidity
. Unit 1 x 10 Dissolved Oxygen Measured at the
- Clean-Up Deminer-alizer Influent 180 ppb i
- Based on Efficiency of I-131, 2 Hours After Sampling
- Based on APHA Units (Turbidity) and 500 ml of filtered sample.
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September 11, 1970 t
Shutdown Conditions Minimum Average Maximum pH 6.1 6.8 7.h 3
Conducticity, umho/cm 03 0.6 09 Turbidity, APHA Units (ppm)
<0.1 0.2 15 0 Chloride, ppb
<20
<20 40.0 Boron 0.1 0.1 0.1 Silica 0.15 0.2
5 at 2 Hours, cpm /mi 7 x 10 5 x 10 3 x 10
- Crud, Gross Gamma at 2 Hours,com/TurbidityUnit 6x103 5
h x 10
-9 x lo'
- Based on Efficiency of I-131, 2 Hours After Sampling
- Based on APHA Units (Turbidity) and 500 ml of Filtered Sample 1
Question h - For Aach component listed, indicate whether the internal surface 1:: r.ormally in contact with flowing water, stagnant water or steam, and indicate whether the configuration and operating con-ditions are such that a possibility exists of entrapment of gases within the sensitized portion. Also, discuss whether possible corrodents could have come into contact with the internal surfaces during cleaning or other preoperational exposure of these surfaces.
Answer - Please refer to Column 10 on the attached data sheets for normal internal component environmental conditions.
No construction records exist for monitoring of possible corrodents on component surfaces during construction. The nuclear steam supply system was chemically cleaned following construction with a solution consisting of 25,000 gallons of demineralized water, 250 pounds NaOH, 250 pounds Na3P0h>
3 9 gallons detergent (Triton 100) and 55g gallons of tri-ethanol-amine (chelating agent) at a temperature of 150. Periodic samples showed the PO4 and OH residuals remained steady and th!re was no buildup of S O. Rinses i2 were conducted with a solution consisting of 10,000 gallons of demineralized water, 21 pounds monosodium phosphate, 21 pounds disodium phosphate and h2 pounds sodium nitrate. The first rinse solution was first used as the final rinse solution for the feed-water and heater drain system. Following the nuclear steam supply rinse, samples showed a low iron color residual measured by colormetric analysis. A second rinse solution similar to the first was prepared and recycled. Analysis showed that a complete rinse of the nuclear steam supply system was obtained.
- Question 5 - Specify the nondestructive tests that have been performed inter.
nally and externally on each component listed cince its instal-lation. Indicate the acceptance criteria established for each type of test, the sensitivity in terms of flaw detection and the results of these tests.
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7 Dr'. Peter A. Morris September 11, 1970 l
Answer' - Ultrasonic examinations were performed in early 1970 on i
. one' steam drum downcomer nozzle extension, one steam drum riser extension and
. three "J" welds between.the stub tube and control rod drive housing assembly.
For the steam drum nozzle extensions, the ultrasonic inspection system was calibrated from a 3% of wall notch cut into the inner diameter of a 3-inch
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schedule,' 80-pipe section. The signal from this notch was adjusted to equal.
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Normal scanning was performed 9 divisions on the tester's display scope.
with the sensitivity increaced by a facto-of two.
Evaluation of indications received during normal scanning were performed at " times one" sensitivity.
All defects above the background noise level vere noted (none). For the "J" welds, the system was. calibrated using 10% ana 20% notches machined into a mockup of the assemblies to be inspected. As ex amination on a half skip or-full skip basis was not1possible for accessibili / reasons, a distance Jamplitude correction curve was developed. Both anning and evaluation were performed at " times one" sensitivity. All 2 teations above background 1-noise level vould be considered a flaw if it coul at be otherwise evaluated.
There.were none.
Eydrostatic tests at 1450 psi are condue during each refueling While the pressure is held, a visual inspc. on of all accessible outage.
piping, valver and other equipment is made. The acce '.ance criteria is no l
visible leakage. In addition, during this hydrostati. test, a pressure 1
drop' test is conducted. The acceptance criteria for t.is drop test is approximately 100 psi or less pressure drop occur in ti 30-minute hold period. As the *.est is conducted above ambient tempera res, precise temperature control is difficult.
A reactor vessel surveillance inspection of vesw.1 internals is conducted during each refueling outage prior to loading fuel. This inspection
[
is visual and utilizes viewing aids, mirrors and an underwater television camera to observe the various components 'nside the reactor vessel. This inspection will be continued at each refueling outage.
j Question 6 - Indicate whether any destructive metallurgical examinations 1
l have been performed on sensitized material removed from the reactor coolant pressure boundary, or samples thereof, and the results of such tests.
Answer - No sensitized materia 1 1as been removed from the reactor coolant pressure boundary; therefore, no destructive metallurgical examin-2 ations have been performed.
4 Question 7 - Discuss the operating performance of leak detection systems l
during plant operation to date. Indicate the current senci-J
.tivity of each system.
Answer - A dev cell with a remote recorder is installed in an ex-haust duct from the steam drum cavity. A significant increase in the dew The increase point temperature alerts the operator to a possible steam leak.
j
-in dew point temperature considered significant is that which is caused by a moderate valve packing leak. The presence of a steam leak is confirmed 4
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8 September 11, 1970 by taking a grab sample for air particulate activity on the steam drgm cavity exhaust plenum. The minimum sensitivity of this sample is 5 2 x 10- gpm based on a reactor water iodine activity of'T x 10-2 pCi/cc and 10% of the activity in the leak being carried away by the ventilation stream.
The dirty vaste collection system for the Big Rock Plant typically runs 15 gallons per hour and doubling of this rate for no known reason will be reported by an operator. If this increase in collection rate cannot be explained by plant operation, a. grab sample for air particulate activity is taken to confirm or deny the presence of a leak. The sensitivity of this sample is as discussed in the preceding paragraph.
Very-small leaks in the control rod drive room can be heard )n inspection rounds as the background noise level is very low.
An air particulate sample is routinely taken weekly on the steam drum enclosure exhaust line. The sensitivity of this is 5 2 x 10-h gpm as discussed above. This' method allows detection of very small valve packing leaks.
Question 8 - For each component listed, indicate the degree of accessibility which presently exists for the performance of nondestructive tests and inspections.
Answer - Please refer to Column h of the attached data sheets.
Question 9 - Describe the plans you have developed for surveillance and non-destructive tests of the sensitized stainless steel components of and within the reactor coolant pressure boundary, including a proposed timetable. In this connection, the recent experience with furnace-sensitized stainless steel components indicates that unless a considerable amount of evidence attests to the current integrity of such components or unless valid technical reasons would preclude performing nondestructive tests, the performance of a program of nondestructive testing of a sizeable sample of such components may be appropriate at an early date.
These examinations should include dye penetrant testing and either ultrasonic testing or radiography.
Answer - During the February-March 1970 refueling outage, consumers Power Company inspected ultrasonically one steam drum downcomer nozzle exten-sion, one steam drum riser nozzle extension and three control rod drive assem-bly "J" selds in addition to other welds associated with the primary coolant pressurt boundary but considered nonsensitized. Original plans for the March 19il refueling outage were for a similar program. However, since the telephone conversation of March 16, 1970 between Mr. Dennis L. Ziemann of the Division of Reactor Licensing and Mr. Gerald J. Walke of Consumers Power Company,
-Consumers Power has modified these plans to include ultrasonic examination of all the remaining uninspected steam drum riser and downcomer extensions, the steam drum vent-nozzle extension (piece No 10h-7), the two reactor head vent nozzles and three more control rod. drive assembly "J" welds. In addition, Consumers Power is attempting to develop equipment to volumetrically examine
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1 Dr. ? ster A. Morris
'Septtner 11,'1970
'the reactor vessel steam outlet nozzle from the inside of the vessel.
If this equipment is successfully developed, at least one of the reactor vessel steam outlet nozzles villibe inspected. These plans are contirgent on the allowable' radiation exposure of the inspector; the items discussed abeve vill receive
. preferential treatment.
A drum manway,and seal plate vill be removed during th> next re-fueling outage so that the drum internals may be visually inspec;ed. However, it is doubtful.that much information can be gained by an inspection of this nature due to the high radiation dose rate estimated to exist inside the drum (5 to 10 R/hr).
The visual reactor vessel surveillance inspection of reac'or vessel internals discussed in the answer to Question 5 vill be continued at each refueling outage.
When sufficient experience has been cained concerning which com-ponents can be inspected volumetrically and/or surface and/or visually.nnd the radiation dose associated with the inspection is determined, Consumers
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Power Company vill formulate a program for recurring inspections.
Yours very truly, i
RBS/dmb Robert L. Haueter Electric Production Superintendent -
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Et App 11 able 343 Piete 14., 4 9 NA
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34 C3-)c4 23*
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su tev4 latt estU.a icstle 14F 4;) j)46 1 es
- e h147 A
faseal (83-1M)
("Wr MPftA 1C 146 4A SS-K 4 NJe Sct Apt 11coble
- 3. eeve atvurt Ping Ir l H MA famsel (Sl%)
Et Applicable 34 port Fle*,
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Det Arp11%ble agecer l':) 1 RA
@ )C4 50*
Et ApplJrable 1 - 2*
33 6 G b y earl-ate E430 1C4 lied A
Ne s tle 1%-2 thconel (ES-1M)
CM PPEA 1
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3860 Vent 24)C-.1C4 LCe-$
A 24:41e It h-6 lawnel 39-1M)
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3860 Welt.46FW ht. %er it 4-)
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Sh 6 G Inst a sale. fewer 14304C4 1t440 t e:Le ich-J 1 A
facueel (381M)
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SS 6 0 Inst hesle
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% avle l'.b 41 incomel (33-1%)
CJT PrRA
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Segle % sale
- JN4C4 10447 A
Ossle 1% -19 Inmn*1 (3F1M)
CM PFPA 21*
as & G Yect E s t le 14K-104 104 21 a
Watle 1% 42 Inoonel (334%)
CM f7hA 6. J' 3360
$sfety pelter E4)C-LC4 104 4 S A
.'se n s le ich.2fa taconel ( 634M)
C9F RrRA 2
W, S8 6 0 hvoy Asey 54 7-102 102-1 A
Fest Piste 102 4 33.f 4 33*
EAI l'k eer Dsffle Aesy et t **r Cleepe 1-2)C4M 64 SS-)C4 5*
Sil M App 11:able 16 W
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nat Applicable 32 m ser lug 1 230 108 79 4A W
Feet *eter beter SuA Anales B430400 O$A AA SS-)C 4 8+
sn Not App 1!<eble deevel + 9e 4 o wettised M e kneenettiset 1.
CLatrol 4rtee housing J weld e e cmfireet secessible caly.
+ e Aase et by BFR 8t est ema 2.
Staen outlet nostle esisy.
- 8W e Stee.e b Water ettemptias to seeeley W e weer epipment to 1 tapect.
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a-t atteesit ces be se te to AM e Al sit.us conced Type sted11p inspect tiy remete JOT e Celetum Silicate Type Fsrnsee te olttsed esene.
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