ML20062N029
| ML20062N029 | |
| Person / Time | |
|---|---|
| Site: | Wolf Creek, Callaway  |
| Issue date: | 12/14/1981 |
| From: | Petrick N STANDARDIZED NUCLEAR UNIT POWER PLANT SYSTEM |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| References | |
| SLNRC-81-129, NUDOCS 8112180352 | |
| Download: ML20062N029 (24) | |
Text
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ -
SNUPPS Standardized Nuclear Unit Power Plant System 5 Choke Cherry Road Nicholas A. Petrick Rockville, Maryland 20650 Executi.e Director (301) 869-8010 December 14, 1981 SLNRC 81 129 FILE: 0541 SUBJ:
NRC Request for Additional Information - Materials E_ngineering e
D Mr.HaroldR.Denton,DirectorJ A
O Office of Nuclear Reactor Regulation liEC, i'ED B
U. S. Nuclear Regulatory Commission q
Washington, D. C.
20555 2
DEC171981> rS' Docket Nos:
STN 50-482 and STN 50-483 u m:mrwa
- b inun:u.mun
Reference:
NRC (Youngblood) letter to KGE (Koester), dated b
October 28, 1981:
Same subject N
iq;i 3
Dear Mr. Denton:
The referenced letter requested additional information on the Wolf Creek olant in the area of materials engineering. The 251-series NRC questions for Wolf Creek were identical to 123-series questions previously issued for Callaway. Therefore, the responses for the Wolf Creek questions are provided in the form of modified responses to the Callaway questions which were already included in the SNUPPS FSAR.
Certain of the question responses for Callaway contained a reference to a Westinghouse surveillance capsule report (WCAP-9842). The surveillance capsule report for Wolf Creek (which is currently scheduled to be available in mid-1982) will be identical in scope to the Callaway report. The material to be used in the Wolf Creek surveillance capsules is plate R2508-3, along with a test weldment fabricated from plates R2508-3 and R2508-1 welded with E3.16 weldment. The fracture toughness properties of these limiting materials are given in FSAR Tables 5.3-3, 123.7-5, and 123.7-6.
The above and attached information should be sufficient for the NRC Staff to evaluate the Wolf Creek plant in this review area.
Should the surveillance capsule report for Wolf Creek not be available prior to the issuance of the NRC's SER, a confirmatory item could be the review of the report.
In addition to the modified responses discussed above, the attached FSAR changes reflect the cancellation of Callaway Unit 2.
The attached FSAR changes will be included in the next revision to the SNUPPS FSAR.
Ve trul
- yours, 8112180352 811214
- l N. A. Petr c RLS/mtk4al3 AO lg[ky Attachment cc:
G. L. Koester KGE D. F. Schnell UE D. T. McPhee KCPL W. Hansen NRC/ Cal T. Vandel NRC/WC
A A A
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4 st.u w e.
Ci - Ic2 9 SNUPPS i
Y l,
l Q123.3 To den.onstrate compliance with the beltline (G 7.si.1 Wc.)
material test requirements of Paragraph III.C.2 of Appendix G, 10 CFR Part 50:
a.
Provide a schematic for the reactor vessel showing all welds, plates and/or forgings in the beltline.
Welds should be identified by shop control number, weld procedure quali-fication number, the heat of filler metal, and type and batch of flux.
Provide the chemical composition for these welds (partic-ularly Cu, P, and S content).
b.
Indicate the post-weld heat treatment used in the fabrication of the test welds.
c.
Indicate the plates used to fabricate the test welds.
d.
Indicate whether the test specimen for the longitudinal seams was removed from excess material and welds in the vessel shell course following completion of the longitudinal weld joint.
RESPONSE
Figure 123.3-1 identifies the location of the beltline materials and welds for the Callaway Unit 1 reactor vessel.
Weld identifi-cation information for these welds is given in Table 123.3-1.
Information concerning the fabrication and post-weld heat treat-ment of the surveillance test specimen weld is identified in WCAP-9842 for Callaway Unit 1.
Similar information will be provided in the surveillance program WCAF[for F#w/ e m 71 ptsto Wolf Creek Unit 1 at a later date.
The test weldment is fabricated as a separate weld, not as an extension of a longitudinal weld seam.
idenki@ies %e locntion OS 4he belfhne Figwe 1Z3 3 -2 mcae.ncds and welcis for +he WoIF 6eek Unif i rmcJer Vessel, Weld iden+ificcchon :nformedion -kr Wese welds
_is gwen m Th. late 123.3 -2.
Rev. 7 123.3-1 9/81
v SNUPPS Y
TABLE 123.3-1 CALLAWAY UNIT 1 VESSEL BELTLINE RECION WELD METAL IDENTIFICATION INFORMATION Weld Weld Procedure Weld Wire Flux Weld Seam !centification l
Control No.
Oual. No.
Ty Heat No.
Ty Lot No.
Int. shell long veld seam 1-124A.
B.
and C C2.03 SAA-SMA-12.12-102 B4 90077 Linde 0091 0842 Lower shell long weld seam 101-142A. B, and C C2.03 SAA-SMA-12.12-102 B4 90077 Linde 0091 0842 Inter. to lower st.111 girth seam 101-171 E3.14 SAA-SMA-3.3-107 84 90077 Lande 124 1061 Surveillance test weld E3.14 SAA-SMA-3.3-107 B4 90077 Linde 124 1061 W Metal Chemical Composition (t *)
M P
S S
C N
M C
V Weld Control No
-6 p -
p i
n y
C2.03
.16
.1
.008
.010
.19
.w,
.06
.53
.04
.007 E3.14
.08 1.30
.006
.007
.52
.03
.04
.52
.04
.004 liev. 7 9/81
4 SNUPPS b
MM MMT TABLE 123.3 %
WM UNIT 1 VESSEL BELTLINE REGION WEL$ METAL IDDfTIFICATION INFORMATICM Weld Weld Procedure Weld Wire riuit e
Weld 5+am Identification
[
Control No.
ICual. No.
g Heat No.
M Lot No.
de 46/We Int. shall long weld seam 1-124A. 8.
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I.inde 0091 0842 JJM fc/s/for Inter. to lower shell girth seam 101-171 E3.I.E SAA-SMA-3.3-te 84 9ee**
Linde 124 1061 Surveillance test weld E3.
SAA-SMA-3.3-84 t.inde 124 1061
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Rev.7 9/81 SNUPPS ss x\\s s,f;gure 123.3-1.
v Callaway Unit 1 HNttor Vessel Beltline Region Material PJtintification And Location
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SNUPPS s
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s L;ausrea Unit 13eactor Vessel Beltline Region s
s Material (dentification And Location e,
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SNUPPS l
y 9
Q123.4 To demonstrate compliance with the fracture
( G Z 51.~4 W C.)
toughness requirements of Paragraph IV.A.1 of Appendix G, 10 CFR Part 50:
a.
Provide the RT for all RCPB welds which maybelimitingDIoroperationo'fthereactor vessel.
b.
Indicate whether there are any RCPB heat-affected zones which require CVN impact testing per paragraph NB-4335.2 of the 1977 ASME Code.
Provide CVN impact test data for these heat-affected zones which may be limiting for operation of the reactor vessel.
c.
Indicate that there are no ferritic RCPB base metals' other than in vessels which require fracture toughness testing to NB-2300 of the ASME. Code.
If there are ferritic RCPB base metals other than in vessels which require fracture toughness testing to NB-2300 of the ASME Code, provide CVN impact and drop weight 1
data for all materials which will be limiting for operation of the reactor vessel.
RESPONSE
Charpy V-notch test data for the heat-affected zone of the limiting beltline region plate is presented in WCAP-9842 for Callaway Unit 1.
Similar information will be provided for the limiting materials'of w m q - w N o Wolf Creek Unit 1 at a later date.
There are no other heat-affected zones which require impact
'/
testin'g per Paragraph NB-4335.2 of the 1977 ASME Code.
There are no ferritic base metals other than in vessels in the reactor coolant pressure boundary.
9 s
k'
'e
/
Rev. 7 123.4-1 9/81
..s.
SNUPPS Q123.6 Provide actual pressure-temperature limits for Callaway Unit 1 (Wolf Creek) based upon the b 2M
- 3 limiting fracture toughness of the reactor vessel material and the predicted shift in the adjusted The pr$ Iu,re sulting from reference temperature, RT re radiation damage.
temperature limits for the following conditions must be included in the technical specifications *.: hen they are sub-mitted:
a.
Preservice hydrostatic tests, b.
Inservice leak and hydrostatic tests, c.
Heatup and cooldown operations, and d.
Core operation.
RESPONSE
The pressure-temperature limits will be included with,the proposed technical specifications.
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suures I
v Q123.7 Provide full CVN impact curves for each weld
(& 251.4 Wt) and plate in the beltline region.
Provide the data in tabulated and graphical form.
RESPONSE
Complete Charpy test results for each weld and plate in the Callaway Unit 1 reactor vessel beltline region are provided in Tables 123.7-1 through 123.7-3.
C i.i l a r i.n f o...a ; _ v ii ;v& c olla aay ^-
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l' Rev. 7 9/81 123.7-1 L
SNUPPS TABLI 123.7 1 CALLAWAY UNIT 1 BELTLINE REGION INTERMEDIATE SHELL PLATE TOUCHNESS Plate R2707-1 Plate R2707-2 Plate p2707-3 Toop.
Energy Shear Lat. Exp.
Temp.
Energy Shear Lat. Exp.
Temp.
Energy Shear Lat. Exp.
tri (ft 1b)
- _Lil, fetts)
(F)
(ft Ib)
- _ill, (mtis)
(F)
(ft Ib)
,[11, (m:Is)
-40 15 0
10
-40 10 0
5
-40 14 0
7
-40 10 0
6
-40 19 0
11
-40 16 0
8
-40 9
0 6
-40 10 0
4
-40 12 0
5 30 29 15 20 10 2b 10 20 20 35 15 26 30 31 15 23 10 37 15 25 20 35 15 28 20 37 20 26 10 21 5
13 20 33 15 24 00 44 20 32 60 53 25 38 40 55 30 40 60 43 20 33 60 45 20 36 40 46 25 34 00 44 20 35 60 46 20 35 40
!O 10 46 80 50 40 40 70 52 25 40 50 64 40 42 80 58 60 47
't o 62 40 46 50 51 30 38 80 47 50 42 70 59 30 45 50 50 25 37 00 47 50 39 80 62 40 48 60 79 35 53 90 59 60 42 80 60 40 48 60 54 25 38 00 56 60 43 80 66 40 47 60 62 30 44 100 70 80 52 100 74 60 54 100 79 40 59 103 67 70 51 100 88 80 62 100 80 50 58 100 61 70 49 100 65 60 51 100 76 40 57 160 76 100 58 160 100 100 71 160 96 100 67 160 78 100 62 160 98 100 65 160 103 100 69 160 81 100 61 160 103 Iv0 72 160 90 100 68 NOT
-40 F NDT
-50 F NDT
-40 F j0F NDT
9/81 1
SNUPPS i
Y TABLE 123.7-2 CALLA =AY LHIT 1 SEL*LINE REGICN LCwT.R SHELL PLATE TOUCHNESS Plate R2'08-1 Plate R2700-2 Plate R2'00-3 Terp.
Energy Shear Lat. Exp.
Temp.
Energy Shear Lat. Exp.
Temp.
Energy Shear Lat. Exp.
sF)_
(ft !b) fi)
(mils)
(F)
(ft ib)
It)
- mils)
(F)
(ft Ibf g
(mils) 20 10 0
5
-4C 7
0 3
-40 6
0 2
20 9
0 4
-40 8
0 3
-40
?
O 3
20 11 0
5
-40 7
0 3
-40 6
0 3
60 18 0
13 30 29 10 19 0
14 0
11 60 16 0
13 30 22 5
16 0
13 0
9 60 18 0
15 30 21 5
14 0
15 0
11 80 22 5
18 60 43 20 29 50 32 15 26 80 27 10 23 60 43 20 28 50 33 10 28 80 27 10 22 60 42 20 28 50 47 20 32 100 45 20 35 70 51 25 35 70 46 20 31 100 41 15 32 70 57 30 39 70 52 25 35 100 40 15 33 70 51 25 36 70 45 20 31 110 58 25 41 100 68 30 47 80 52 25 37 110 57 25 41 100 57 25 40 80 54 25 38 110 50 25 35 100 81 40 51 00 57 30 39 160 79 60 52 160 90 90 60 100 71 25 54 160 39 70 56 160 98 95 62 100 64 30 46 160 78 60 52 160 101 90 65 100 67 30 48 212 78 100 57 212 105 100 63 160 94 80 60 212
'6 100 58 212 110 100 74 160 92 80 66 212 74 100 58 212 100 100 64 160 100 90 68 212 95 100 61 212 100 100 62 212 109 100 60 n
+
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9/81
SNUPPS TABLE 123.7-3 CALLAWAY UNIT 1 BELTLINE REGION WELD METAL TOUGHNESS Weld Control No. C2.03 Weld Control No. E3.14 Tmp.
Energy Shear Lat. Exp.
Temp.
Energy Shear Lat. Exp.
(F)
(ft lb)
A (mils)
(F)
(ft lb) h (mils)
-80 16 0
7
-80 16 0
9
-80 18 0
8
-80 17 0
13
-80 18 0
7
-80 21 0
15
-40 38 20 26
-40 38 15 29
-40 32 15 17
-40 59 30 42-
-40 34 15 19
-40 50 25 36 0
79 40 52 0
84 60 59 0
61 70 39 0
83 60 57 0
95 70 60 0
72 40 48 10 96 70 62 60 108 90 76 10 101 70 60 60 106 90 76 10 84 60 58 60 109 90 79 60 118 80 78 100 109 100 77 60 130 90 80 100 114 100 80 60 117 80 75 100 109 100 79
.100 137 100 82 160 115 100 83 100 132 100 82 160 108 100 80 100 141 100 83 160 114 100 81 180 142 100 82 180 145 100 85 180 143 100 83 I
T Rev. 7 NDT
-60 F NDT
-60 F NDT
-60 F
SNUPPS
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Plate "I'
' I Pl a te #f%a 9 Temp.
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Temp.
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Temp.
Energy Shear Lat. Exp.
(F)
(ft Ib)
A (mils)
(F)
(ft Ib)
A (mals)
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A&44 d8664 TABLE 123.7-X N
Y UNIT 1 BELTLINE REGION WELD METAL TOUGHNESS 6
b Weld Control No. C2.of Weld Control No. E3.!X TF.r. p.
Energy Shear Lat. Exp.
Temp.
Energy Shear Lat. Exp.
(F)
(ft lb)
(t)
(mils)
(F)
(ft lb)
(%)
(mils)
.-Jh0 -(,o 0
-Ws D O o
-+-/D
-8o
-tv //
o 9
.-ae -6o :-e o3
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SNUPPS Q123.8 To demonstrate the surveillance capsule program complies with Paragraph II.C.3 of Appendix H:
4 a.
Provide the withdrawal schedule for each capsule.
b.
Provide the lead factors for each capsule.
c.
Indicate the estimated reactor vessel end of life fluence at the 1/4 wall thickness as measured from the ID.
RESPONSE
The requested material is provided in WCAP-9842 for Callaway Unit 1. Agimilar report #for Wolf Creek Unit 1 Mw124wavl h 21will be available later.
Rev. 7 3xiL&-1
@ M 1'
SNUPPS V
Q123.9 Identify the location of each material surveillance capsule and the materials in each capsule.
a.
For each base metal and heat-affected zone surveillance specimen provide the specimen type, the orientation of the specimen relative to the principal rolling direction of the plate, the heat number, the component code number from which the sample was removed, the chemical composition especially the copper (Cu) and phosphorus (P) contents, the melting practice and the heat treatment received by the sample material.
b.
For each weld metal surveillance specimen provide the weld identification from which the sample was removed, the weld wire type and heat identification, flux type and lot identification, weld process and heat treatment used for fabrication of the weld sample.
c.
Provide a sketch which indicates the azimuthal location for each capsule relative to the reactor core.
RESPONSE
The requested material is provided in WCAP-9842 for Callaway Unit 1. A /imilar report /~for Wolf Creek Unit 1 EKdhwp"y1 Vnrt/21will be available later.
~
Rev. 7 123.9-1
'9/81 m
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SWPFS I
V Q123.10 Indicate the normal operating temperature of the gg,7 flywheels and provide CVN impact and drop weight test data from each flywheel tgat indicates the RT of the flywheels are 100 F less than their noESIl operating temperatures.
RESPONSE
As stated in WCAP-8163 (Reference 1 to Section 5.4), the normal operating temperature of the reactor coolant pump motor fly-wheels is 120 F.
The Westinghouse specifications require a maximum RT of 10 F, as discussed in Section 5.4.1.5.2.2.
NgT otch and dropweight tests confirm that the normal The Charpy N
operating temperature is in excess of 100 F above the RTNDT the flywheel material.
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7-9/81 man _am_,
SNUPPS (especially copper content), initial upper shelf energy, and
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fluence to assure that a 50-foot-pound shelf energy, as required by Appendix G of 10 CFR 50 is maintained throughout the life of the vessel.
The specimens are oriented as required by NB-2300 of Section III of the ASME Code..
The vessel fracture toughness data is provided in Tables 5.3-3g* ond 5.3-4 W r W a for Wolf Creek Unit 1 Callaway Unit 1, tirm M am Srt'EZ) respectively.
d 5.3.1.6 Material Surveillance In the surveillance program, the evaluation of radiation damage is based on preirradiation testing of Charpy V-notch and tensile specimens and postirradiation testing of Charpy V-notch, tensile, and 1/2 T (thickness) compact tension (CT) fracture mechanics test specimens.
The program is directed toward evaluation of the effect of radiation on the fracture toughness of reactor vessel steels based on the transition temperature approach and the fracture mechanics approach.
The program will conform with ASTM E-185 " Recommended Practice for Surveillance Tests for Nuclear Reactor Vessels," and 10 CFR 50, Appendix H.
The reactor vessel surveillance program uses six specimen capsules.
The capsules are located in guide baskets welded to the outside of the neutron shield pads and positioned directly opposite the center portion of the core.
The capsules can be removed when the vessel head is removed and can be replaced when the internals are removed.
The six capsules contain reactor yessel steel specimens, oriented both parallel and normal (longitudinal and transverse) to the principal rolling direction of the limiting base material located in the core region of the reactor vessel and associated weld metal and weld heat-affected zone metal.
The six capsules contain 54 tensile specimens, 360 Charpy V-notch specimens (which include weld metal and weld heat-affected zone material), and 72 CT specimens.
Archive material sufficient for two additional capsules will be retained.
Dosimeters, as described below, are placed in filler blocks drilled to contain them.
The dosimeters permit evaluation of the flux seen by the specimens and the vessel wall.
In addition, thermal monitors made of low melting point alloys are included to monitor the maximum temperature of the specimens.
The specimens are inclosed in a tight-fitting stainless steel sheath to prevent corrosion and ensure good thermal conductivity.
The complete capsule is helium leak tested.
As part of the surveillance program, a report of the residual elements in weight percent to the nearest 0.01 percent will be made for surveillance material and as deposited weld metal..
' ~
Rev. 5 5.3-5 7/81
SNUPPS Similar expressions may be developed for points within the pres-srre vessel wall; and, thus, together with the surveillance program dosimetry, serve to correlate the radiation induced damage to test specimens with that of the reactor v.essel.
5.3.1.7 Reactor Vessel Fasteners The reactor vessel closure studs, nuts, and washers are designed and fabricated in accordance with the requirements of the ASME Code,Section III.
The closure studs are fabri-cated of SA-S40, Class 3, Grade B24.
The closure stud material meets the fracture toughness requirements of the ASME Code,Section III and 10 CFR 50, Appendix G.
Compliance with Regulatory Guide 1.65, " Materials and Inspections for Reactor Vessel Closure Studs," is discussed in Appendix 3A.
Nondestructive examinations are performed in accordance with the ASME Code,Section III.
Bolting materials fracture toughness data is provided in Tables R
- H --
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, uzu m e e um te) respectively.
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Refueling procedures require that the studs, nuts, and washers be removed from the reactor closure and be placed in storage racks during preparation for. refueling.
The storage racks are then removed from the refueling cavity and stored at convenient locations on the containment operating deck prior to removal of the reactor closure head and refueling cavity flooding.
Therefore, the reactor closure studs are never exposed to the borated refueling cavity water.
Addi-tional protection against the possibility of incurring corrosion effects is assured by the use of a manganese base phosphate surfacing treatment.
The stud holes in the reactor flange are sealed with special plugs before removing the reactor closure, thus preventing leakage of the bora ted refueling water into the stud holes.
5.3.2 PRESSURE - TEMPERATURE LIMITS 5.3.2.1 Limit Curvqs Startup and shutdown operating limitations will be based on the properties of the reactor pressure vessel beltline materials.
Actual material property test data will be used.
l The methods outlined in Appendix G to Section III of the ASME Code will be employed for the shell regions in the analysis of protection against nonductile failure.
The initial operating curves are calculated, assuming a period of reactor operation such that the beltline material will be limiting.
The heatup and cooldown curves are given in the Technical Specifications.
Beltline material properties degrade with radiation exposure, and this degradation is measured in terms of -the adjusted reference nil-ductility temperature, which includes a refer-ence nil-ductility temperature shift (aRTNDT)*
Rev. 5 5.3-12 7/81
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(Mils) bilN 84730 SAS40, B24 SOS 339.0 155.0 17.0 52.2
$2, 52, SI 30, 31, 25 353 84730 SAS40, B24 505-1 142.2 157.0 16.0
$1.4 So, 53, So 30, lu, 26 338 84730 SAS40, B24 SIO 144.7 ISS.O 16.0
$1.9 49, 50, 49 30, 18, 27 311 84730 SAS40, 824 Slo-1 141.0 156.0 16.0 52,6 54, 51, 52 34, 30, 28 321 84730 SAS40, B24 512 144.5 160.0 16.0 51.4 40, 49, to 29 2n, 30 344 84730 SAS40, B24 512-1 141.0 ISS.S 15.S 58.5 S3, 51, $1 32, 36, 32 331 84730 SAS40, b24 515 841.5 157.5 16.0 51.7 51, 51, SS 31, 29, 38 324 84730 SA540, B24 SIS-1 140.S 15S?S 16.5 53.8
$4, S4 36, 34, 33 128 84730 SAS40, B24 52I 135.5 153.0 17.0 56.0 53, $5, 56 31, 15, 35 338 84730 SAS40 B24 521-1 143.7 860.0 17.5 53.8 49, 49 SO 26, 27, 27 334 84730 SAS40, B24 528 143.0 159.0 17.5 55.7 Si, 54, SS 31, 13 33 331 84113 SAS40, B24 528-1 143.0 150.0 17.5 53.8 56, SS. SS 35, 31, 34 323 Closuse llead Nuts 6 Washees 63182 SAS40. B24 132 148.0 162.0 17.5 57.3 SI, 52, 53 31, 32, 30 331 63382 SAS40, B24 832-1 148.7 162.0 l '7, 0 54.7 49, 48, 49 29, 26, 29 334 63382 SA540, B24 133 147.2 168.0 17.0 55.2
$2, 50, 51 31, 3[),
30 328 63382 SAS40, B24 133-1 149.2 162.5 17.5 54.7 St. 51, 49 29 31, 27 138 63182 SAS40 B24 135 147.5 168.0 37.0 53.0 49, 49, 51 28, 29, 30 321 63182 SAS40, 824 135-8 143.2 157.0 17.5 55.2 55, 54, S2 31, 32, 31 328 63182 SAS40 B24 137 145.0 159.0 16.5 54.8 54, 54, S3 31, 33, 29
~131 63182 S AS 4'),
B24 137-1 147.0 160.0 17.0 55.7 54, 55, 54 34, 36
~3 3 328 63182 SAS40, B24 143 145.0 159.0 18.0 58.1 S5, 54, 54 33, 32, 32 338 63182 SAS40, B24 143-1 147.0 160.0 17.0 57.3 54, 50, 52 31, 29, 30 328 63182 SAS40, B24 145 I45.0 159.0 17.0 56.0 54, 54, 55 34, 35, 34 328 63182 SAS40 B24 845-1 146.2 IS9.7 17.0
$7.0 56, 55, S4 36, 35, 36 331 63182
%AS40, B24 14e 444.0 157.5 17.5 56.5 56, SS, SS 33, 34, 14 334 63882 SAS40 B24 148-1 148.5 162.0 17.0 55.6 52, St. S2 33 Jo, 30 321 63182 SAS40, B24 150 144.7 158.0 17.5 55.7 55, 55, S4 33, 30, 38 331 63182 SAS40, B24 150-1 145.7 160.0 17.0 56.5 53, 50, 52 33, 30, 31 33a hev. 5 7/88
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4 63182 SAS40, B24 133 147.2 461.0 17.0 55.2 S2. 50, 51 31, 30, 30 324
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63382 SAS40, B24
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63382 SAS40, 824 148 144.0 157.5 17.5 56.5 56, SS, 55
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- 52. St. 52 33, 28, 30 321 63382 SAS40, B24 150 144.7 158.0 37.5 SS.7 SS. 55, 54 31, 30, 31 338 63382 SAS40, B24 150-1 145.7 160.0 37.0 S6.5 53, 50 52 33 30, 31 338.
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