ML19327A214
| ML19327A214 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 07/30/1980 |
| From: | Clayton F ALABAMA POWER CO. |
| To: | Schwencer A Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8008010285 | |
| Download: ML19327A214 (79) | |
Text
.
Alabama Power Company 4
600 North 18tn Street
~~
Post Offica Box 2641 Birmingham, Alabaml 35291 Telephone 205 323-5341 k
n pekjWc"?,"; fin, Alabama Power me soumem eiectre sys:em July 30, 1980 Docket No. 50-364 Director of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D.C.
20555 Attention:
Mr. A. Schwencer Gentlemen:
J. M. FARLEY NUCLEAR PLANT - UNIT NO. 2 10CFR50, APPENDICES G AND H In response to your letter dated June 19, 1980, Alabama Power Company hereby submits the additional information requested concerning the subject issue.
Should you have any questions regarding this matter, please advise.
Very truly yours,
..te:.,
F. L. Clayton, Jr.
/
FLCJr/JGS:bhj Enclosures cc:
Mr. R. A. Tho=as Mr. G. F. Trowbridge Mr. L. L. Kintner (W/ Enclosure)
M. W. H. Bradford f
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ENCLOSURE RESP 0bSE TO NRC OVESTIONS ON APPENDICES G AND H
i.,
QUESTION 122.1 According to FSAR Table 5.2-20, SA 533 Class 2 steel is used in the steam generator and pressurize:r.
Demonstrate the fracture toughness adequacy of this. material as required by Paragraph I.A, Appendix G,10 CFR Part 50.
RESPONSE
SA 508 Class 2a material and SA 533 Class 2 material was used in the Farley Unit 2 pressurizer. Neither of these materials was used in primary side (RCPB) pressure retaining application of the Farley Unit 2 steam 9enerators. The fracture toughness data for these materials are included in the tabulation provided in response to Question 122.2.
The adequacy of the fracture toughness properties of these materials has been documented in WCAP-9292; responses to questions, received by Westinghouse from the NRC, will be provided to the NRC by the end of September,1980.
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QUESTION 122.2 Provide sufficient fracture toughness data for the ferritic materials in the steam generator and pressurizer.
(Data should include values for NDTT, TNDT, and upper shelf energy.) Define and determine an RTNDT for each material specimen as required by Paragraph IV.A.2(a) of Appendix G,10 CFR Part 50.
RESPONSE
The Farley Unit 2 steam generators and pressurizer were designed and fabricated in accordance with the requirements of the 1971 Edition of the ASME Code Section III and the 1971 Edition of the ASME Code Section III through the Winter 1970 Addendum.
The current 10CFR50 Appendix G require-ments, which became effective on August 16, 1973, are more stringent than the applicable Code requirements for Farley Unit 2.
The actual' fracture toughness data for reactor coolant pressure boundary pressure retaining applications in the steam generators and pressurizer are tabulated in Table 122.2-1.
In all cases, the applicable ASME Code require-ments, as well as the intent of 10CFR50 Appendix G, are satisfied.
TABLE 122.2-1 Nj Charpy lateral T
RT Component Test Material V-Notch Expansion Temperature NOT NOT-COMPONENT Part Number Speci fication (ft-lb)
(in)
(OF)
(OF) (OF)-
Steam Generator (1551) Channel llead TO 3848 SA 216 Gr.WCC 61.4,59.0,7'8.7
.070,.066,.089 70 10 10 86.2,84.2,86.6
.127 124,.135 70 Tube Sheet TO 3254 SA 508 C1.2 88, 88, 74
.057,.057,.051 10 84, 58, 81
.057,.041,.055 10 Manway Cover TO 3729 SA 533 Gr.AC1.1 53, 51, 52
.051,.047,.049 100 10 40 52, 53, 54
.049,.052,.052 100 Steam Generator (1552) Channel llead TO 3843 SA 216 Gr.WCC 56, 68, 57
.046,.057,.043 70 0
10 '
80, 80, 80
.061,.062,.068 70 Tube Sheet TO 3302 SA 508 C1.2 80.5,74.0,85.0
.054,.047,.063 10 86.0.108.0,83.0
.059. 067,.052 10 Manway Cover TO 3729 SA 533 Gr.AC1.1 53, 51, 52
.051,.047,.049 100 10 40 52, 53, 54
.049,.052,.052 100 Steam Generator (1553) Channel llead TO 3909 SA 216 Gr.WCC 88.9,70.1,73.7
.075. 064,.062 70 62.9,67.9,67.9
.058,.063,.062 70 Tube Sheet TO 3307 SA 508 C1.2 67, 55, 65
.045,.036,.046 10 75, 51, 54
.051,.035,.037 10 Manway Cover TO 3729 SA 533 Gr.AC1.1 53, 51, 52
.051,.047,.049 100 10 40 52, 53, 54
.049,.052,.052 100 Pressurizer (1561)
Lower llead TO 3626 SA 533 Gr.AC1.2 75, 83, 76
.060 064,.062 70 10 10 Surge Nozzle TO 3386 SA 508 C1.2 88, 100, 96
.067,.082,.078 70 10 10 Forging Upper llead TO 3748 SA 533 Gr.AC1.2 63, 75, 72
.060,.056,.062 70 10 10 Manway Nozzle TO 3336 SA 508 C1.2 113, 129, 120
.085,.086,.077 70 10 10 Forging Safety Nozzle TO 3381-3 SA 508 C1.2 82, 82, 78
.065,.066,.063 70 10 10 Forging Safety Nozzle TO 3284-1 SA 508 C1.2 64, 64, 55
.042,.042,.036 10 Forging Safety Nozzle TO 4281-10 SA 508 C1.2a 139, 136, 141
.089,.079,.087 120 60 60 forging Relief Nozzle TO 3380-3 SA 508 C1.2 84, 83, 92
.067,.070,.076 70 10 10 Forging Spray Nozzle TO 3722 SA 508 C1,2 74, 86, 71
.064,,072, 059 70 10 10 Forging
TABLE 122.2-1 Cont'd T
'kT Charpy Lattral Component Test Mat:: rial V-Notch Expansien Temperature NDT NDT:
COMPONENT Part Number Specification (ft-lb)
(in)
(DF)
(DF)
(OF)'
Pressurizer (1561)
Manway Cover TO 4405 SA 533 Gr.AC1.1 75, 79, 81
.069 068 064 120 60 60 (Cont'd)
Shell Barrel TO 3630 SA 533 Gr.AC1.2 58, 58, 62
.054,.054,.054 70 10 10 Shell Barrel TO 3741 SA 533 Gr.AC1.2 51, 54, 54
.049 044,.050 80
_10 20 Shell Barrel TO 3355 SA 533 Gr.AC1.2 60, 64, 66
.046,.058,.058 70 10 10
- drop weight test r'esults not available 4
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QUESTION 122.3
' According to FSAR Table 5.2-20, SA 540 Grades B23 and B24 steel is used for closure bolting in the reactor coolant pump. To establish the actual material yield strength, as required by Paragraph I.C, Appendix G, 10CFR Part 50, identify the class (es) of SA 540 Grades B23 and B24 used.
RESPONSE
The RCP main flange bolts are SA 540 Grade B24 Class 4.
The RCP seal housing bolts are SA 193 Grade B7.
9 6
QUESTION 122.4 Provide details of the program for calibrating temperature instru-ments and Charpy V-notch test machines. This information should be sufficient to demonstrate that the program is in compliance with Paragraph NB-2360 of the ASME Code as required by Paragraph III.B.3, Appendix G, 10CFR Part 50.
RESPONSE
The Farley Unit 2 reactor vessel, steam generators, and pressurizer were fabricated to ASME Code Section III requirements in effect prior to the issuance of 10CFR50, Appendix G.
However, the following discussion demonstrates that the intent of the Appendix G requirements is satisfied.
Reactor Vessel - Combustion Engineering calibrated Charpy V-notch test machines in accordance with Watertown Arsenal Standards every six months.
Temperature instruments, calibrated in accordance with ASTM-E-23, were purchased every three months.
These calibrations were performed in accordance with the requirements of the ASME Code 1968 Edition through Sumer 1970 Addenda (Appendix IX-221 and 260), which is the applicable Code for the Farley Unit 2 reactor vessel.
The Charpy V-notch test machine calibrations were recorded. The temperature instrument calibrations were not recorded; however, thermometers qualified to ASTM standards were purchased, used for the certified time period, and replaced with new qualified thermometers.
Combustion Engineering required that all of its vendors who furnished materials or parts (for Farley Unit 2) to be on an approved vendors list.
Each vendor was required to have a quality control system in accordance with
- N-335 of the 1968 ASME Code through Sumer 1970 Addenda.
Periodic audits of these vendors were performed by CE QA personnel.
- is QUESTION 122.4 (Cont'd)
Page 2 It should be noted that the Farley Unit 2 reactor vessel was partially furnished by B&W. Material furnished by B&W was accepted on the basis of material certifications; therefore, no QA audits were performed for those by CE.
Steam Generators and Pressurizers - Charpy V-notch test machine calibration at Westinghouse Tampa plant was performed yearly using samples obtained from Watertown Arsenal. Temperature instrument calibration was performed with standards traceable to the National Bureau of Standards.
All material suppliers have been either surveyed by ASME auditors or Westinghouse Tampa Plant Product Assurance to obtain supplier certifications.
A sampling of one of the major material suppliers indicated that Charpy V-notch test machine calibrations were recorded and that calibrated tempera-ture instruments were purchased (as replacements) on a yearly basis.
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QUESTION 122.5 Provide information, including training and experience, to demonstrate that the qualification of individuals who performed fracture toughness tests are in compliance with the requirements of Paragraph III.B.4, Appendix G,10 CFR Part 50.
RESPONSE
The Farley Unit 2 reactor vessel, steam generators, and pressurizer were fabricated to ASME Code Section III requirements in effect prior to the issuance of 10 CFR 50, Appendix G.
However, the following discussion demonstrates that the intent of the Appendix G requirements is satisfied.
Reactor Vessel - The personnel performing the Charpy testing at Combustion Engineering were qualified by schooling, training, and many years of experience. Their qualifications to perfom this work have been certified by qualified supervisory personnel. This meets the requirements of the applicable ASME Code 1968 Edition through Summer 1970 Addenda (Appendix IX 221d.).
Steam Generators and Pressurizer - Charpy impact tests are performed at Westinghouse Tampa Plant by Level III and Level II personnel who have a minimum of five years directly related testing experience.
9
QUESTION 122.6 Paragraph III.C.2, Appendix G,10 CFR Part 50, specifies that every fracture toughness test specimen from the reactor vessel beltline be subjected to a heat treatment that produces metallurgical effects equivalent to those produced in the vessel material throughout its fabrica-tion process.
Identify all specimens that do not meet this requirement and provide technical justification for use of such specimens in establish-ing fracture toughness properties of the reactor vessel beltline.
RESPONSE
Materials used to prepare fracture toughness test specimens from the reactor vessel beltline region were subjected to a heat treatment in accordance with the requirenents of the Summer 1970 Addenda to the 1968 ASME Code Section III. The heat treatment applied to the specimens con-forns with paragraph NB-2211 of the present ASME Code Section III require-ments and, therefore, the heat treated condition of the test specimens is in compliance with paragraph III.C.2 of 10 CFR 50, Appendix G.
I
QUESTION 122.7 In weld seam 10-923, two of the nine specimens tested had impact 0
energies less than 75 ft-lbs at a test temperature of 10 F.
To demonstrate compliance with the upper shelf impact energy requirements of paragraph IV.B of 10CFR Part 50, the applicant must supply additional data, possibly from either baseline surveillance material or information available in the literature, and/or analyses to define the minimum upper shelf energy for weld seam 10-923. The data and/or analyses used to demonstrate compliance with the upper shelf requirements of Paragraph IV.B must include variables that affect upper shelf toughness, e.g., chemical compcsition, fabrication history, weld wire and heat of filler metal. The applicant must also supply the individual data points obtained frcm the Cy impact tests for each of the base metal heats in the reactor vessel beltline.
RESP 0NSE Although two test specimens for weld metal used in weld seam 10-923 0
exhibited impact energies of less than 75 ft-lb at a test temperature of 10 F, it is expected that the upper shelf impact energy requirement of 75 ft-lb identified in paragraph IV.B of 10CFR50 Appendix G would easily be exceeded if tests had been perfomed at test temperatures representative of the upper shel f.
A review of many weld test certificates provided by the vessel fabricator indicates that the upper shelf energy of welds of chemical composition and fabrication history similar to weld seam 10-923 and fabricated with the same type of wire and flux (Type B-4 weld wire and Linde 0091 Flux) used in seam 10-923 exceeds 75 ft-lb by a considerable margin. Four examples of the vessel fabrication test results for weld material similar to that of seam 10-923 are shown in Table 122.7-1. Like weld seam 10-923, two of these four examples U
did not exhibit 75 f t-lb for all test specimens at 10 F; however, at higher temperatures, 75 ft-lb was exceeded.
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. QUESTION 122.7 (Cont'd)
Page 2 Individual data points obtained from Charpy V notch impact tests for each of the base metal heats in the Farley Unit 2 reactor vessel beltline are presented in Tables'122.7-2,122.7-3, and 122.7-4.
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TABLE 122.7-1 Example 1 - Type B-4 Weld Wire and Linde 0091 Flux (Test #1302)
IMPACT AND/0R FRACTUPC tiSTS TYPE TEMP. UF VALUES TEMP.0F
\\ALUES NOT FT/LB % SHEAR MilsLatExp DROP WEIGHTS CVN
-80 3
0 1
- 50 1F
-400F
-80 3
0 2
-40 1F
-80 9
0 4
-30 2 NF
-40 26 10 19
-40 37 15 25
-40 38 15 24
+10 69 35 46
+100 117 90 83
+10 50 25 38
+100 114 90 82
+10 66 30 44
+100 120 90 83
. 20 66 35 46
+160 124 100 83
+
+20 81 50 57
+160 136 100 89
+20 90 60 63
+160 135 100 88 Example 2 - Type B-4 Weld Wire and Linde 0091 Flux (Test #1388)
IMPACT AND/0R FRACTURE TESTS TYPE TEMP. UF VALUES TEMP.0F 1 VALUES NDT FT/LB % SHEAR MilsLatExp DROP WEIGHTS 0
CVN
-80 11 0
3
-60 1F
- 60 F
-80 11 0
3
-50 2 NF
-80 13 0
4
-40 1 NF
-40 30 15 17
-40
'27 15 15
-40 25 10 11 0
77 50 45
+100 143 100 84 0
72 50 40
+100 133 100 82 0
70 50 41
+100 145 100 86
+10 76 50 41
+180 143 100 82
+10 74 50 46
+180 149 100 86
+10 82 60 45
+180 148 100 85
+60 116 70 76
+60 118 70 74
+60 1 21 70 71 l
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TABLE 122.7-1 (Continued)
Example 3 - Type B-4 Weld Wire and Linde 0091 Flux (Test #1389)
IMPACT AND/OR FRACTURE TESTS TYPE TEMP. UF VALUES TEMP. OF VALUES NDT FT/LB % SHEAR MilsLatExp DROP WEIGHTS
-CVN
-60 16 0
9
-60 1F
- 60 F
-60 15 0
7
-50 2 NF
-60 19 0
11
-40 1 NF
-40 20 5
11
-40 28 10 16
-40 32 15 22
-20 85 50 53
+60 132 80 77
-20 88 50 56
+60 149 100 84
-20 76 40 47
+60 123 80 74 0
77 40 47
+100 142 100 82 0
75 40 45
+100 148 100 84 0
99 60 52
+100 140 100 82
+20 117 70 74
+20 105 60 65
+20 114 70 74 Example 4 - Type B-4 Weld Wire and Linde 0091 Flux (Test #1386)
IMPACT AND/OR FRACTURE TESTS TYPE TEMP UF VALUES TEMP. vF VALUES NOT FT/LB % SHEAR MilstatExp DROP WEIGHTS CVN
-80 16 0
7
-60 1F
-80 18 0
8
-50 2 NF
-80 18 0
7
-40 1 NF
-40 38 20 26
-40 32 15 17
-40 34 15 19 0
79 40 52
+100 137 100 82 0
61 70 39
+100 132 100 82 0
95 70 60
+100 1 41 100 83
+10 96 70 62
+180 142 100 82
+10 101 70 60
+180 145 100 85
+10 84 60 58
+180 143 100 83
+60 118 80 78
+60 130 90 80
+60 117 80 75 l
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TABLE 122.7-2 LOWER SHELL COURSE CHARPY V-NOTCH DATA
- Plate Code No. B7210-1 Plate Code No. B7210-2
-Test Energy)
Lat. Exp Shear Test Energy)
(Mils)
(%)
Lat. Exp Shear Temp. (D )
(Ft-Lb (Mils)
(%)
Temp. (OF)
(Ft-Lb F
-50 10 6
9
-50 15 11 9
-50 14.5 8
15
-50 12.5 8
9
-50 11 7
9
-50 11 8
9 20 33 25 29 0
26 24 30 20 47 35 34 0
27.5 27 34 20 46 33 34 0
45 35 32 75 48.5 38 59 30 51 39 30 75 50 40 59 30 40 34 34 75 62 47 64 30 47 39 30 110 86 67 80 100 67 52 79 110 75 57 75 100 80.5 58 75 110 69.5 54 67 100 85 60 75 150 100 69 100 150 100 76 100 150 95 71 100 150 101 74 100 150 93 67 100 150 97 75 100 210 96 70 100 210 98 69 100 210 105.5 74 100 210 102 76 100 210 107 75 100 210 95.5 72 100
- Normal to major rolling direction of the plate t
TABLE 122.7-3 INTERMEDIATE SHELL COURSE CHARPY V-NOT'CH DATA
- Plate Code No. B7203-1 Plate Code No. B7212-1 Test Energy Lat. Exp Shear Test Energy Lat. Exp Shear Temp. (O )
(Ft-Lb)
(Mils)
(%)
Temp. (OF)
(Ft-Lb)
(Mils)
(%)
F
-50 13.5 9
15
-50 18.5 11 12
-50 19 11 15
-50 15.5 11 12
-50 14 8
15
-50 19 11 12 0
28 25 30 0
35 27 27 0
34 26 28 0
34.5 27 25 0
44 36 34 0
30 27 25 20 55 41 40 30 43 35 32 20 51 38 45 30 48 36 35 20 43 32 34 30 52 39 43 75 50.5 50 56 100 76.5 55 73 75 61.5 40 52 100 74 56 73 75 65 46 61 100 70 54 69 150 91 68 100 150 95 67 100 150 97 76 100 150 98 68 100 150 92 70 100 150 106 76 100 210 105.5 69 100 210 89 68 100 210 97.5 74 100 210 94 70 100 210 95.5 72 100 210 88 69 100 1
(Normal to major rolling direction of the plate L-
TABLE 122.7-4 N0ZZLE SHELL COURSE CHARPY V-NOTCH DATA
- Forgoing Code No. B7216-1 Test Energy)
Lat. Exp Shear Temp. (oF)
(Ft-Lb (Mils)
(%)
-80 2
0 0
-80 4
0 0
-80 8
4 0
-20 68 53 29
-20 37 25 16
-20 66 52 29 10 99 76 64 10 103 76 64 10 110 81 70 10 95 77 52 10 55 41 29 le 78 63 40 30 72 57 23 30 93 70 55 30 87 65 46 100 147 91 100 100 123 77 75 100 110 80 70 180 146 90 100 180 151 88 100 180 149 88 100
- Major working direction of forgoing
QUESTION 122.8 The materials surveillance program uses six specimen capsules containing reactor vessel steel specimens of the limiting base material, weld metal-material, and heat affected zone material. To demonstrate compliance with Appendix H 10 CFR Part 50, provide a table that includes the following information for all the surveillance specimens:
1.
Actual surveillance material; 2.
Beltline material from which the specimen was obtained; 3.
Test specimen type and orientation; 4.
Fabrication history of each test specimen; 5.
Chemical composition of each test specimen; and, 6.
Heat of filler material, production welding conditions, and base metal combinations for weld specimens.
Provide the lead factor for each specimen capsule calculated with respect to the vessel inner wall.
RESP 0NSE 1
Refer to Table 122.8-1 for the information pertaining to Items 1, t
2, and 4.
Refer to Table 122.8-2 for Item 3 and Table 122.8-3 for Item 5.
Regarding Item 6, the surveillance weldment was fabricated with the same type of wire and the same heat of wire (Wire Type E8018C3 and wire heat no. BOLA) as was used to fabricate the longitudinal weld seal (19-9238) in the intermediate shell course of the vessel. The same welding procedures were used to fabrica'te the surveillance weldment and the vessel j
weld seam (19-9238).
Table 122.8-4 identifies the lead factor for each specimen capsule.
l iAs indicated, the weld metal representative of the intermediate shell longitudinal weld seam.is included in the Farley Unit 2 surveillance program; the surveillance weldment has a copper content of 0.03%.
It should be noted that the vessel girth weld has a copper content of 0.13%, and therefore is more limiting than the surveillance weldment. However, the base metal in
QUESTION 122.8 (Cont'd)
Page 2-the surveillance program (from intermediate shell plate B7212-1) has a 0
copper content of 0.20% and a predicted RT shift of 390 ; the base metal NDT is the most limiting material in the beltline region and will be used to define operating limits for Farley Unit 2.
In conclusion, although the most limiting weld metal in the beltline region is not included in the surveillance program, this is not of consequence since the weld metal is not the controlling beltline region material.
O
O TABLE 122.8-1 SURVEILLANCE MATERIAL BELTLINE LOCATION AND FABRICATIONS HISTORY Surveillance Beltline Location of Material Surveillance Material Heat Treatment Base Metal Intermediate Shell Plate B7212-1 1550-1650"F-4Hr.-WQ,1200-1250 F-4Hr.-AC, ll25-ll75"F-18Hr.-FC U
0 Weld Metal
- Intermediate Shell Longitudinal Weld Seam 1125-ll75 F-13Hr.-FC HAZ Metal Intermediate Shell Plate B7212-1 1125-1175 F-13Hr.-FC OSurveillance weldment fabricated using Plate 87212-1 and B7203-1. Surveillance weldment was fabricated using the same type of wire (E8018C3) and the same heat of wire (Heat No. BOLA) as was used to fabricate the intermediate shell longitudinal weld seam (19-9238) in the vessel. The same welding procedures (MA-Sil-D and A-244-110-8) were used by 4
the vessel supplier to fabricate the surveillance weldment and the intermediate shell longitudinal weld seam (19-9238).
- WQ - Water Quench AC - Air Cooled FC - Furnace Cooled 4
TABLE 122.8-2 SURVEILLANCE TEST SPECIMENS TYPE, ORIENTATION, AND QUANTITY PER TEST CAPSULE Surveillance Specimen Material Orientation Charpy V Tensile 1/2 T-CT Base Metal (Plate B7212-1)
Transverse 15 3
4 Base Metal (Plate B7212-1)
Longitudinal 15 3
4 Weld Metal Transverse 15 3
4 HAZ Metal (Plate B7212-1)
Longitudinal 15 i
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m
.mm--
,9-y
a o.
TABLE 122.8-3 SURVEILLANCE MATERIAL CHEMICAL COMPOSITION (WT. %)
Element Plate B7212-1 Weld Metal C
O.21 0.086 Mn 1.30 0.95 P
0.01 8 0.004 S
0.016 0.014 Si 0.24 0.34 Ni 0.60 0.90 Cr 0.15
<0.01 Mo 0.49 0.23 Cu 0.20 0.03 V
0.003 0.006 Co 0.027 0.010 Sn 0.011 0.002 A1 0.040 0.003 N
0.006 0.007 2
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a',
TABLE 122.8-4 SPECIMEN CAPSULE LEAD FACTOR For the six surveillance program capsules, the lead factor calcu-lated with respect to the vessel inner wall is as follows:
Capsule Lead Factor 0
V (107 )
3.5 0
X (287 )
3.5 j
U(3430) 3.5 0
W (110 )
2.9 0
Y(290) 2.9 0
Z(340) 2.9 9
9
QUESTION 122.9 To demonstrate the integrity of the reactor coolant pump flywheels, supply the Charpy V-notch impact and tensile data for each flywheel, explicitly stating the material used for each flywheel.
Also, confirm that welding, including repair welding, was not performed on any finished flywheel.
If welding were performed, identify the flywheel (s) and location of the welds.
RESPONSE
The reactor coolant pump flywheels consist of two discs bolted together with three keyways to key the flywheel to the motor shaf t, No welding, including repair welding, is perfomed on the flywheels.
Each flywheel disc is fabricated of ASTM A533 Grade B Class 1 steel. The Charpy V-notch and tensile data (from material certifications) for the flywheel discs are tabulated in Tables 122.9-1 and 122.9-2.
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TABLE 122.9-1, YIELD TENSILE MOTOR MELT STRENGTH STRENGTH UNIT NUMBER (PSI)
(PSI) 01 R0317 74,300 91,100 69,500 91,100 R0187 65,400 87,000 70,500 90,000 02 R0174 62,600 87,100 69,800 91,100 R0187 65,400 87,000 70,500 90,000 03 R0194 64,300 86,200 63,400 84,700 R0187 65,400 87,000 70,500 90,000 4
e
- + -.
y e
,,-----.-m-w-- - - -
TABLE 122.9-2 CHARPY LATERAL TEST Tgy*
RTNOT*
MOTOR MELT V-NOTCH EXPANSION TEMPERATURE D
(F )
(F )
UNIT NUM8ER (FT-LB)
(IN.)
O (F )
01 R0317 58,52,52(T)
.050,.043,.044(T) 70 10 70 140,178,182 (L)
.080,.079,.086 (L) 70 R0187 128,128,118 (T)
.089,.087,.086 (T) 70 10 10 135,150,165 (L)
.098,.088,.098 (L) 70 02 R0174 120,103,120 (T)
.082,.091,.088 (T) 70 10 10 110,131,123 (L)
.090,.083,.087(L) 70 R0187 128,128,118 (T)
.089,.087,.086 (T) 70 10 10 135,150,165 (L)
.098,.088,.098(L) 70 03 R0194 90,100,110 (T)
.083,.072,.086(T) 70 10 10 112, 120, 130 - (L)
.091,.086,.085 (L) 70 R0187,
128,128,118 (T)
.089,.087,.086(T 70 10
_10 135,150,165 (L)
.098,.088,.098 (L 70
- Based on two drop weight tests exhibiting no break performance at 20 F.
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e ENCLOSURE RESPONSE TO NRC QUESTIONS ON APPENDICES G AND H 1
o QUESTION 122.1 According to FSAR Table 5.2-20, SA 533 Class 2 steel is used in the steam generator and pressurizer. Demonstrate the fracture toughness adequacy i
of this material as required by Paragraph I.A. Appendix G,10 CFR Part 50.
RESPONSE
SA 508 Class 2a material and SA 533 Class 2 material was used in the Farley Unit 2 pressurizer. Neither of these materials was used in primary side (RCPB) pressure retaining application of the Farley Unit 2 steam generators. The fracture toughness data for these materials are included in the tabulation provided in response to Question 122.2.
The adequacy of the fracture toughness properties of these materials has been documented in WCAP-9292; responses to questions, received by Westinghouse from the NRC, will be provided to the NRC by the end of September,1980.
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QUESTION 122.2 Provide sufficient fracture toughness data for the ferritic materials in the steam generator and pressurizer.
(Data should include values for NDTT, TNDT, and upper shelf energy.)
Define and determine an RTNDT for each material specimen as required by Paragraph IV.A.2(a) of Appendix G,10 CFR Part 50.
RESPONSE
The Farley Unit 2 steam generators and pressurizer were designed and fabricated in accordance with 'the requirements of the 1971 Edition of the ASME Code Section III and the 1971 Edition of the ASME Code Section III through the Winter 1970 Addendum.
The current 10CFR50 Appendix G require-ments, which became effective on August 16, 1973, are more stringent than the applicable Code requirements for Farley Unit 2.
The actual fracture toughness data for reactor coolant pressure boundary pressure retaining applications in the steam generators and pressurizer are tabulated in Table 122.2-1.
In all cases, the applicable ASME Code require-ments, as well as the intent of 10CFR50 Appendix G, are satisfied.
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o TABLE 122.2-1 Charpy Lateral T
RT Component Test Material V-Notch Expansion Temperature NDT NDT COMP 0NENT Part Number Speci fica tion
( f t-lb)
(in)
(OF)
(F)
(OF)
Steam Generator (1551) Channel llead TO 3848 SA 216 Gr.WCC 61.4,59.0,7'8.7
.070,.066. 089 70 10 10 86.2,84.2,86.6
.127 124,.135 70 Tube Sheet TO 3254 SA 508 C1.2 88, 88, 74
.057. 057,.051 10 84, 58, 81
.057. 041. 055 10 Manway Cover TO 3729 SA 533 Gr.AC1.1 53, 51, 52
.051 047,.049 100 10 40 52, 53, 54
.049 052,.052 100 Steam Generator (1552) Channel llead TO 3843 SA 216 Gr.WCC 56, 68, 57
.046,.057,.043 70 0
10 80, 80, 80
.061. 062,.068 70 Tube Sheet TO 3302 SA 508 C1.2 80.5,74.0,85.0
.054,.047 063 10 86.0,108.0,83.0
.059,.067,.052 10 Manway Cover TO 3729 SA 533 Gr.AC1.1 53, 51, 52
.051,.047,.049 100 10 40 52, 53, 54
.049,.052,.052 100 Stcam Generator (1553) Channel llead TO 3909 SA 216 Gr.WCC 88.9,70.1,73.7
.075. 064,.062 70 62.9,67.9,67.9
.058,.063,.062 70 Tube Sheet TO 3307 SA 508 C1.2 67, 55, 65
.045,.036 046 10 75, 51, 54
.051 035,.037 10 Manway Cover TO 3729 SA 533 Gr.AC1.1 53, 51, 52
.051,.047 049 100 10 40 52, 53, 54
.049,.052. 052 100 Pressurizer (1561)
Lower llead TO 36E6 SA 533 Gr.AC1.2 75, 83, 76
.060. 064,.062 70 10 10 Surge Nozzle TO 3386 SA 508 C1.2 88, 100, 96
.067,.082,.078 70 10 10 Forging Upper llead TO 3748 SA 533 Gr.AC1.2 63, 75, 72
.060 056,.062 70 10 10 Manway Nozzle TO 3336 SA 508 C1.2 113, 129, 120
.085,.086,.077 70 10 10 Forging Safety Nozzle TO 3381-3 SA 508 C1.2 82, 82, 78
.065,.066. 063 70 10 10 Forging Safety Nozzle TO 3284-1 SA 508 C1.2 64, 64, 55
.042,.042,.036 10 Forging Safety Nozzle TO 4281-10 SA 508 C1.?a 139, 136, 141
.089,.079,.087 120 60 60 Forging Relief Nozzle TO 3380-3 SA 508 C1.2 84, 83, 92
.067,.070,.076 70 10 10 Forging Spray Nozzle TO 3722 SA 508 C1.2 74, 86, 71
.064, 072,.059 70 10 10 Forging
-TABLE 122.2-1 Cont'd Charpy Lateral T
RT Component Test Material V-Notch Expansion Temperature NDT NDI:
COMPONENT Part Number Specification (ft-lb)
(in)
(CF)
(CF) (0.F) i Pressurizer (1561)
Manway Cover TO 4405 SA 533 Gr.AC1.1 75, 79, 81
.069,.068. 064 120 60 60 (Cont'd)
Shell Barrel TO 3630 SA 533 Gr.AC1.2 58, 58, 62
.054,.054,.054 70 10 10 Shell Barrel TO 3741 SA 533 Gr.AC1.2 51, 54, 54
.049 044. 050 80 10 20 Shell Barrel TO 3355 SA 533 Gr.AC1.2 60, 64, 66
.046,.058. 058 70 10 10
- drop weight test r'esults not available a
e 4
4
QUESTION 122.3 According to FSAR Table 5.2-20. SA 540 Grades B23 and B24 steel is used for closure bolting in-the reactor coolant pump. To establish the actual material yield strength, as required by Paragraph I.C Appendix G, 10CFR Part 50, identify the class (es) of SA 540 Grades B23 and B24 used.
RESPONSE
The RCP main flange bolts are SA 540 Grade B24 Class 4.
The RCP seal housing bolts are SA 193 Grade B7.
1 e
=,
QUESTION 122.4 Provide details of the program for calibrating temperature instru-ments and Charpy V-notch test machines. This information should be sufficient to demonstrate that the program is in compliance with Paragraph NB-2360 of the ASME Code as required by Paragraph III.B.3, Appendix G, 10CFR Part 50.
RESPONSE
The Farley Unit 2 reactor vessel, steam generators, and pressurizer were fabricated to ASME Code Section III requirements in effect prior to the issuance of 10CFR50, Appendix G.
However, the following discussion demonstrates that the intent of the Appendix G requirements is satisfied.
Reactor Vessel - Combustion Engineering calibrated Charpy V-notch test machines in accordance with Watertown Arsenal Standards every six months.
Temperature instruments, calibrated in accordance with ASTM-E-23, were purchased every three months.
These calibrations were performed in accordance with the requirements of the ASME Code 1968 Edition through Summer 1970 Addenda (Appendix IX-221 and 260), which is the applicable Code for the Farley Unit 2 reactor vessel.
The Charpy V-notch test machine calibrations were recorded. The temperature instrument calibrations were not recorded; however, thermometers qualified to ASTM standards were purchased, used for the certified time period, and replaced with new qualified thermometers.
Combustion Engineering required that all of its vendors who furnished materials or parts (for Farley Unit 2) to be on an appmved vendors list.
Each vendor was required to have a quality control systen in accordance with
- N-335 of the 1968 ASME Code through Sunraer 1970 Addenda.
Periodic audits l
l of these vendors were performed by CE QA personnel.
i
QUESTION 122.4 (Cont'd)
Page 2 r
It should be noted that the Farley Unit 2 reactor vessel was partially furnished by B&W. Material furnished by B&W was accepted on the basis of material certifications; therefore, no QA audits were performed for those by CE.
Steam Generators and Pressurizers - Charpy V-notch test machine calibration at Westinghouse Tampa plant was performed yearly using samples obtained from Watertown Arsenal. Temperature instrument calibration was performed with standards traceable to the National Bureau of Standards.
All material suppliers have been either surveyed by ASME auditor's or Westinghouse Tampa Plant Product Assurance to obtain supplier certifications.
A sampling of one of the major material suppliers indicated that Charpy V-notch test machine calibrations were recorded and that calibrated tempera-ture instruments were purchased (as replacements) on a yearly basis.
_.w_,
QUESTION 122.5
. Provide information, including training and experience, to demonstrate that the qualification of individuals who performed fracture toughness tests are in compliance with the requirements of Paragraph III.B.4, Appendix G,10 CFR Part 50.
RESPONSE
The Farley Unit 2 reactor vessel, steam generators, and pressurizer i
were fabricated to ASME Code Section III requirements in effect prior to the issuance of 10 CFR 50, Appendix G.
However, the following discussion demonstrates that the intent of the Appendix G requirements is satisfied.
Reactor Vessel - The personnel performing the Charpy testing at Combustion Engineering were qualified by schooling, training, and many years of experience. Their qualifications to perform this work have been certified by qualified supervisory personnel. This meets the requirements of the applicable ASME Code 1968 Edition through Summer 1970 Addenda 1
(AppendixIX221d.).
Steam Generators and Pressurizer - Charpy impact tests are performed
.i at Westinghouse Tampa Plant by Level III and Level II personnel who have i
a minimum of five years directly related testing experience.
e I
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- - --e-
J QUESTION 122.6 Paragraph III.C.2, Appendix G,10 CFR Part 50, specifies that every fracture toughness test specimen from the reactor ve.' el beltline be subjected to a heat treatment that produces metallurgical effects equivalent to those produced in the vessel material throughout its fabrica-tion process.
Identify all specimens that do not meet this requirement and provide technical justification for use of such specimens in establish-ing fracture toughness properties of the reactor vessel beltline.
RESP 0NSE Materials used to prepare fracture toughness test specimens from the reactor vessel beltline region were subjected to a heat treatment in accordance with the requirements of the Summer 1970 Addenda to the 1968 ASME Code Section III. The heat treatment applied to the specimens con-forms with paragraph NB-2211 of the present ASME Code Section III require-ments and, therefore, the heat treated condition of the test specimens is in compliance with paragraph III.C.2 of 10 CFR 50, Appendix G.
9 4
QUESTION 122.7 In weld seam 10-923, two of the nine specimens tested had impact energies less than 75 f t-lbs at a test temperature of 10 F.
To demonstrate i
compliance with the upper shelf impact energy requirements of paragraph IV.B of 10CFR Part 50, the applicant must supply additional data, possibly from either baseline surveillance material or information available in the literature, and/or analyses to define the minimum upper shelf energy for weld seam 10-923. The data and/or analyses used to demonstrate compliance with the upper shelf requirements of Paragraph IV.B must include variables that affect upper shelf toughness, e.g., chemical composition, fabrication history, weld wire and heat of filler metal. The applicant must also supply the individual data points obtained from the Cy impact tests for each of the base metal heats in the reactor vessel beltline.
RESPONSE
Although two test specimens for weld metal used in weld seam 10-923 0
exhibited impact energies of less than 75 ft-lb at a test temperature of 10 F, it is expected that the upper shelf impact energy requirement of 75 ft-lb identified in paragraph IV.B of 10CFR50 Appendix G would easily be exceeded i
if tests had been perfomed at test temperatures representative of the upper shel f.
A review of many weld test certificates provided by the vessel fabricator indicates that the upper shelf energy of welds of chemical composition and fabrication history similar to weld seam 10-923 and fabricated with the same t;ype of wire and flux (Type B-4 weld wire and Linde 0091 Flux) used in seam 10-923 exceeds 75 ft-lb by a considerable margin. Four examples of the vessel fabrication test results for weld material similar to that of seam 10-923 are shown in Table 122.7-1. Like weld seam 10-923, two of.these four examples did not exhibit 75 ft-lb' for all test specimens at 10 F; however, at higher temperatures ~, 75 ft-lb was exceeded.
l L
QUESTION 122.7 (Cont'd)
Page 2 Individual data points obtained from Charpy V notch impact tests for each of the base metal heats in the Farley Unit 2 reactor vessel beltline are presented in Tables 122.7-2,122.7-3, and 122.7-4.
O i
i
TABLE 122.7-1 Example 1 - Type B-4 Weld Wire and Linde 0091 Flux (Test #1302)
IMPACT AND/OR FRACTURE TESTS TYPE TEMP. uF VALUES TEMP.0F VALUES NDT FT/LB % SHEAR MilsLatExp DROP WEIGHTS CVN
-80 3
0 1
- 50 1F
-400F
-80 3
0 2
-40 1F
-80 9
0 4
-30 2 NF
-40 26 10 19
-40 37 15 25
-40 38 15 24
+10 69 35 46
+100 117 90 83
+10 50 25 38
+100 114 90 82
+10 66 30 44
+100 120 90 83
.+20 66 35 46
+160 124 100 83
+20 81 50 57
+160 136 100 89
+20 90 60 63
+160 135 100 88 1
Example 2 - Type B-4 Weld Wire and Linde 0091 Flux (Test #1388)
IMPACT AND/0R FRACTURE TES'S TYPE TEMP. uf VALUES
- TEMP.0F TiALUES NDT FT/LB % SHEAR MilsLatExp DROP WEIGHTS CVN
-80 11 0
3
-60 1F
- 60 F
-80 11 0
3
-50 2 NF
-80 13 0
4
-40 1 NF
-40 30 15 17
-40 27 15 15
-40 25 10 11 0
77 50 45
+100 143 100 84 0
72 50 40
+100 133 100 82 0
70 50 41
+100 145 100 86
+10 76 50 41
+180 143 100 82
+10 74 50 46
+180 149 100 86
+10 82 60 45
+180 148 100 85
+60 116 70 76
+60 118 70 74
+60 1 21 70 71 l
l
TABLE 122.7-1 (Continued)
Example 3 - Sype B-4 Weld Wire and Linde 0091 Flux (Test #1389)
IMPACT AND/OR FRACTURE TESTS TYPE TEMP. UF VALUES TEMP. OF VALUES NDT FT/LB % SHEAR MilsLatExp DROP WEIGHTS 0
CVN
-60 16 0
9
-60 1F
- 60 F
-60 15 0
7
-50 2 NF
-60 19 0
11
-40 1 NF
-40 20 5
11
-40
-28 10 16
-40 32 15 22
-20 85 50 53
+60 132 80 77
-20 88.
50 56
+60 149 100 84
-20 76 40 47
+60 123 80 74 0
77 40 47
+100 142 100 82 0
75 40 45
+100 148 100 84 0
99 60 52
+100 140 100 82
+20 117 70 74
+20
'105 60 65
+20 114' 70 74 J
Example 4 - Type B-4 Weld Wire and Linde 0091 Flux (Test #1386)
IMPACT AND/OR FRACTURE TEST 5 TYPE TEMP. UF VALUES TEMP. UF VALUES NDT FT/LB % SHEAR MilsLatExp DROP WEIGHTS l
CVN
-80 16 0
7
-60 1F
-80 18 0
8
.50 2 NF
-80 18 0
7
-40 1 NF
-40 38 20 26
-40 32 15 17
-40 34 15 19 0
79 40 52
+100
-137 100 82 0
61 70 39
+100 132 100 82 0
95 70 60
+100 1 41 100 83
+10 96 70 62
+180 142 100 82
+10 101 70 60
+180 145 100 85
+10-84 60 58
+180 143 100 83
+60 118 80
'78
+60 130 90 80
+60 117 80 75 l
I l
TABLE 122.7-2 LOWER SHELL COURSE CHARPY V-NOTCH DATA
- Plate Code No. B7210-1 Plate Code No. B7210-2 Test Energy)
Lat. Exp Shear Test Energy Lat. Exp Shear (D )'
(Ft-Lb (Mils)
(%)
Temp. (OF)
(Ft-Lb)
(Mils)
(%)
Temp.
F
-50 10 6
9
-50 15 11 9
-50 14.5 8
15
-50 12.5 8
9
-50 11 7
9
-50 11 8
9 20 33 25 29 0
26 24 30 20-47 35 34 0
27.5 27 34 20 46 33 34 0
45 35 32 75 48.5 38 59 30 51 39 30 75 50 40 59 30 40 34 34 75 62 47 64 30 47 39 30 110 86 67 80 100 67 52 79 110 75 57 75 100 80.5 58 75 110 69.5 54 67 100 85 60 75 150 100 69 100 150 100 76 100 150 95 71 100 150 101 f7-4 100 150 93 67 100 150 97 75 100 210 96 70 100 210 98 69 100 210 105.5 74 100 210 102 76 100 210 107 75 100 210 95.5 72 100
$ Normal to major rolling direction of the plate l'
I
TABLE 122.7-3 INTERMEDIATE SHELL COURSE CHARPY V-NOTCH DATA
- Plate Code No. B7203-1 Plate Code No. B7212-1 Test Energy Lat. Exp Shear Test Energy)
Lat. Exp Shear
. Temp. (O )
(Ft-Lb)
(Mils)
(%)
Temp. (OF)
(Ft-lb (Mils)
(%)
F
-50 13.5 9
15
-50 18.5 11 12
-50 19 11 15
-50 15.5 11 12
-50 14 8
15
-50 19 11 12 0
28 25 30 0
35 27 27 0
34 26 28 0
34.5 27 25 0
44 36 34 0
30 27 25 20 55 41 40 30 43 35 32 20 51 38 45 30 48 36 35 20 43 32 34 30 52 39 43 75 50.5 50 56 100 76.5 55 73 75 61.5 40 52 100 74 56 73 75 65 46 61 100 70 54 69 150 91 68 100 150 95 67 100 150 97 76 100 150 98 68 100 150 92 70 100 150 106 76 100 210 105.5 69 100 210 89 68 100
.210 97.5 74 100 210 94 70 100 210 95.5 72 100 210 88 69 100
- Nonnal to major-rolling direction of the plate
TABLE 122.7-4 N0ZZLE SHELL COURSE CHARPY V-NOTCH DATA
- Forgoing Code No. B7216-1 Test Energy)
Lat.' Exp Shear Temp. (oF)
(Ft-Lb (Mils)
(%)
-80 2
0 0
-80 4
0 0
-80 8
4 0
-20 68 53 29
-20 37 25 16
-20 66 52 29 10 99 76 64 10 103 76 64 10 110 81 70 10 95 77 52 10 55 41 29 10 78 63 40 30 72 57 23 30 93 70 55 30 87 65 46 100 147 91 100 100 123 77 75 100 110 80 70 180 146 90 100 180 151 88 100 180 149 88 100
- Major working direction of forgoing
.w y
o QIIESTION 122.8 The materials surveillance program uses six specimen capsules containing reactor vessel steel specimens of the limiting base material, weld metal material, and heat affected zone material. To demonstrate compliance with Appendix H,10 CFR Part 50, provide a table that includes the following information for all the surveillance specimens:
1.
Actual surveillance material; 2.
Beltline material from which the specimen was obtained; 3.
Test specimen type and orientation; 4.
Fabrication history of each test specimen; 5.
Chemical composition of each test specimen; and, 6.
Heat of filler material, production welding conditions, and base metal combinations for weld specimens.
Provide the lead factor for each specimen capsule calculated with respect to the vessel inner wall.
RESPONSE
Refer to Table 122.8-1 for the information pertaining to Items 1, 2, and 4.
Refer to Table 122.8-2 for Item 3 and Table 122.8-3 for Item 5.
Regarding Item 6, the surveillance weldment was fabricated wf th the same type of wire and the same heat of wire (Wire Type E8018C3 ant' wire heat no. BOLA) as was used to fabricate the longitudinal weld seal (19-923B) in the intermediate shell course of the vessel. The same welding procedures were used to fabrica'te the surveillance weldment and the vessel weld seam (19-9238).
Table 122.8-4 identifies the lead factor for each specimen capsule.
As indicated, the weld metal representative of the intermediate shell longitudinal weld seam is included in the Farley Unit 2 surveillance program; the surveillance weldment has a copper content of 0.03%.
It should be noted that the vessel girth weld has a copper content of 0.13%, and therefore is more limiting than the surveillance weldment. However, the base metal in
e e
QUESTION 122.8 (Cont'd)
Page 2 the surveillance program (from intermediate shell plate B72121) has a 0
copper content of 0.20% and a predicted RT shif t of 390 ; the base metal NDT is the most limiting material in the beltline region and will be used to define operating limits for Farley Unit 2.
In conclusion, although the most limiting weld metal in the beltline region is not included in the surveillance program, this is not of consequence since the weld metal is not the controlling beltline region material.
e
?
TABLE 122.8-1 SURVEILLANCE MATERIAL BELTLINE LOCATION AND FABRICATIONS HISTORY Surveillance Beltline Location of Ma terial Surveillance Material Heat Treatment e
0 Base Metal Intennediate Shell Plate B7212-1 1550-1650 F-4Hr.-WQ,1200-1250 F-4Hr.-AC,1125-1175 F-18Hr.-FC Weld Metal
- Intermediate Shell Longitudinal Weld Seam ll 25-ll 75,F-13Hr.-FC HAZ Metal Intermediate Shell Plate B7212-1 1125-1175 F-13Hr.-FC
- Surveillance weldment fabricated using Plate B7212-1 and B7203-1. Surveillance weldment was fabricated using the same type of wire (E8018C3) and the same heat of wire (Heat No. BOLA) as was used to fabricate the intermediate shell longitudinal weld seam (19-9238) in the vessel. The same welding procedures (MA-511-D and A-244-110-8) were used by the vessel supplier to fabricate the surveillance weldment and the intermediate shell longitudinal weld seam (19-9238).
WQ - Water Quench AC - Air Cooled FC - Furnace Cooled
+
l,
. a TABLE 122.8-2 SURVEILLANCE TEST SPECIMENS TYPE, ORIENTATION, AND QUANTITY PER TEST CAPSULE Surveillance Specimen Material Orientation Charpy V Tensile 1/2 T-CT Base Metal (Plate B7212-1)
Transverse 15 3
4 Base Metal (Plate B7212-1)
Longitudinal 15 3
4 Weld Metal Transvers a 15 3
4 HAZ Metal (Plate B7212-1)
Longitudinal 15 6
4
TABLE 122.8-3 SURVEILLANCE MATERIAL CHEMICAL COMPOSITION (WT. %)
Element Plate B7212-1 Weld Metal C
O.21 0.086 Mn 1.30 0.95 P
0.018 0.004 S
0.016 0.014 Si 0.24 0.34 Ni 0.60 0.90 Cr 0.15
<0.01 Mo-0.49 0.23 Cu 0.20 0.03 V
0.003 0.006 Co 0.027 0.01 0 Sn 0.011 0.002 Al 0.040 0.003 N
0.006 0.007 2
4
TABLE 122.8-4 SPECIMEh CAPSULE LEAD FACTOR For the six surveillance program capsules, the lead factor calcu-lated with respect to the vessel inner wall is as follows:
Capsule Lead Factor 0
V (107 )
3.5 0
X (287 )
3.5 U(3430) 3.5 0
W (110 )
2.9 0
Y (290 )
2.9 0
Z(340) 2.9 9
e e
t
s QUESTION 122.9 To demonstrate the integrity of the reactor coolant pump flywheels, supply the Charpy V-notch impact and tensile data for each flywheel, explicitly stating the material used for each flywheel.
Also, confirm that welding, including repair welding, was not performed on any finished flywheel.
If welding were performed, identify the flywheel (s) and location of the welds.
RESPONSE
The reactor coolant pump flywheels consist of two discs bolted together with three keyways to key the flywheel to the motor shaft. No welding, including repair welding, is performed on the flywheels.
Each flywheel disc is fabricated of ASTM A533 Grade B Class 1 steel. The Charpy V-notch and tensile data (from material certifications) for the flywheel discs are tabulated in Tables 122.9-1 and 122.9-2.
9 9
TABLE 122.9-1 YIELD TENSILE MOTOR MELT STRENGTH STRENGTH UNIT NUMBER (PSI)
(PSI) 01.
R0317 74,300 91,100 69,500 91,100
^
R0187 65,400 87,000 70,500 90,000 02 R0174 62,600 87,100 69,800 91,100 R0187 65,400 87,000 i
70,500 90,000 I
03 R0194 64,300 86,200 j
63,400 84,700 l
R0187 65,400 87,000 70,500 90,000 1
i e
i.--.i.
TABLE 122.9-2 CHARPY LATERAL TEST Tgy*
RTNDT MOTOR MELT V-NOTCH EXPANSION TEMPERATURE O
(F )
UNIT NUMBER (FT-LB)
(IN.)
(F )
O (F )
01 R0317 58, 52, 52 (T)
.050,.043,.044(T) 70
-10 10 140,178,182 (L)
.080,.079
.086 (L) 70 j
R0187 128,128,118 (T)
.089,.087,.086(T) 70 10 10 135,150,165 (L)
.098,.088,.098 (L) 70 02 R0174 120,103,120 (T)
.082,.091,.088 (T) 70 10 10 110,131,123 (L)
.090,.083,.087(L) 70 R0187 128,128,118 (T)
.089,.087,.086 (T) 70 10 10 135,150,165 (L)
.098,.088,.098 (L) 70 I
90,100,110 (T)
.083,.072,.086 (T) 70 10 10 l
03 R0194 4
112,120,130 (L)
.091,.086,.085 (L) 70 i
R0187,
128,128,118 (T)
.089,.087,.086(T) 70 10 10 135,150,165 (L)
.098,.088,.098 (L) 70
- Based on two drop weight tests exhibiting no break performance at 20 F.
ENCLOSURE RESPONSE TO NRC OUESTIONS ON APPENDICES G AND H 9
i l
E QUESTION 122.1 According to FSAR Table 5.2-20, SA 533 Class 2 steel is used in the steam generator and pressurizer. Demonstrate the fracture toughness adequacy of this material as required by Paragraph I.A, Appendix G,10 CFR Part 50.
RESPONSE
SA 508 Class 2a material and SA 533 Class 2 material was used in the Farley Unit 2 pressurizer. Neither of these materials was used in primary side (RCPB) pressure retaining application of the Farley Unit 2 steam 9enerators. The fracture toughness data for these materials are included in the tabulation provided in response to Question 122.2.
l The adequacy of the fracture toughness properties of these materials has been documented in WCAP-9292; responses to questions, received by Westinghouse from the NRC, will be provided to the NRC by the end of September,1980.
9
QUESTION 122.2 Provide sufficient fracture toughness data for the ferritic materials in the steam generator and pressurizer.
(Data should include values for NDTT, TNDT, end upper shelf energy.) Define and determine an RTNDT for each material specimen as required by Paragraph IV.A.2(a) of Appendix G,10 CFR Part 50.
RESPONSE
The Farley Unit 2 steam generators and pressurizer were designed and fabricated in accordance with the requirements of the 1971 ~ Edition of the ASME Ccde Section III and the 1971 Edition of the ASME Code Section III through the Winter 100 Addendum. The current 10CFR50 Appendix G require-ments, which became effective on August 16, 1973, are more stringent than the applicable Code requirements for Farley Unit 2.
The actual fracture toughness data for reactor coolant pressure boundary pressure retaining applications in the steam generators and pressurizer are tabulated in Table 122.2-1.
In all cases, the applicable ASME Code require-ments, as well as the intent of 10CFR50 Appendix G, are satisfied.
TABLE 122.2-1 Charpy Lateral T
Rh Component Test Material V-Notch Expansion Temperature NOT NDT COMPONENT Part Number Specification (ft-lb)
(in)
(OF)
Steam Generator (1551) Channel Head TO 3848 SA 216 Gr.WCC 61.4,59.0,78.7
.070,.066,.089 70 10 10 86.2,84.2,86.6
.127 124,.135 70 Tube Sheet TO 3254 SA 508 C1.2 88, 88, 74
.057,.057,.051 10 84, 58, 81
.057,.041,.055 10 Manway Cover TO 3729 SA 533 Gr.AC1.1 53, 51, 52
.051,.047,.049 100 10 40 52, 53, 54
.049,.052,.052 100 Steam Generator (1552) Channel Head TO 3843 SA 216 Gr.WCC 56, 68, 57
.046,.057,.043 70 0
10 80, 80, 80
.061,.062 068 70 Tube Sheet TO 3302 SA 508 C1.2 80.5,74.0,85.0
.054 047,.063 10 86.0,108.0,83.0
.059,.067,.052 10 Manway Cover TO 3729 SA 533 Gr.AC1.1 53, 51, 52
.051,.047,.049 100 10 40 52, 53, 54
.049,.052,.052 100 Steam Generator (1553) Channel Head TO 3909 SA 216 Gr.WCC 88.9,70.1,73.7
.075,.064,.062 70 62.9,67.9,67.9
.058,.063,.062 70 Tube Sheet TO 3307 SA 508 C1.2 67, 55, 65
.045,.036,.046 10 75, 51, 54
.051,.035 037 10 Manway Cover TO 3729 SA 533 Gr.AC1.1 53, 51, 52
.051,.047,.049 100 10 40 52, 53, 54
.049 052,.052 100 Pressurizer (I'dbi) tower Head TO 3626 SA 533 Gr.AC1.2 75, 83, 76
.060 064,.062 70 10 10 surge Nozzle TO 3386 SA 508 C1.2 88, 100, 96
.067,.082,.078 70 10 10 Forging Upper Head TO 3748 SA 533 Gr.AC1.2 63, 75, 72
.060,.056,.062 70 10 10 Manway Nozzle TO 3336 SA 508 C1.2 113, 129, 120
.085,.086,.077 70 10-10 Forging Safety Nozzle TO 3381-3 SA 508 C1.2 82, 82, 78
.065,.066,.063 70 10 10 Forging Safety Nozzle TO 3284-1 SA 508 C1.2 64, 64, 55
.042,.042,.036 10 Forging Safety Nozzle TO 4281-10 SA 508 C1.2a 139, 136, 141
.089,.079,.087 120 60 60 Forging Relief Nozzle TO 3380-3 SA 508 C1.2 84, 83, 92
.057,.070,.076 70 10 10 Forging Spray Nozzle TO 3722 SA 508 C1,2 74, 86, 71 064,.072, 059 70 10 10 Forging
TABLE 122.2-1 Cont'd Charpy Lattral T
RT Compon:nt Tcst Material V-Notch Expansicn Temperature NDT NDT-COMPONENT Part Number Specification (ft-lb)
(in)
(DF)
(DF) (OF)
Pr ssurizer (1561)
Manway Cover TO 4405 SA 533 Gr.AC1.1 75, 79, 81
.069,.068 064 120 60 60 (Cont'd)
Shell Barrel TO 3630 SA 533 Gr.AC1.2 58, 58, 62
.054,.054,.054 70 10 10 Shell Barrel TO 3741 SA 533 Gr.AC1.2 51, 54, 54
.049,.044,.050 80 10-20 Shell Barrel
-T0 3355 SA 533 Gr.AC1.2 60, 64, 66
.046,.058,.058 70 10 10
- drop weight test results not avt.lable 9
QUESTION 122.3 According to FSAR Table 5.2-20, SA 540 Grades 823 and B24 steel is used for closure bolting in the reactor coolant pump. To establish the actual material yield strength, as required by Paragraph I.C, Appendix G, 10CFR Part 50, identify the class (es) of SA 540 Grades B23 and B24 used.
RESPONSE
The RCP main flange bolts are SA 540 Grade B24 Class 4.
The RCP seal housing bolts are SA 193 Grade 87.
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i QUESTION 122.4 Provide details of the program for calibrating temperature instru-ments and Charpy V-notch test machines. This information should be sufficient.to demonstrate that the program is in compliance with paragraph NB-2360 of the ASME Code as required by Paragraph III.B.3, Appendix G,
-10CFR Part 5
RESPONSE
The Farley Unit 2 reactor vessel, steam generators, and pressurizer were fabricated to ASME Code Section III requirements in effect prior to the issuance of 10CFR50, Appendix G.
However, the following discussion demonstrates that the intent of the Appendix'G requirements is satisfied.
Reactor Vessel - Combustion Engineering calibrated Charpy V-notch test machines in accordance with Watertown Arsenal Standards every six months.
Temperature instruments, calibrated in accordance with ASTM-E-23, were purchased every three months.
These calibrations were performed in accordance with the requirements of the ASME Code 1968 Edition through Summer 1970 Addenda (Appendix IX-221 and 260), which is the applicable Code for the Farley Unit 2 reactor vessel.
The Charpy V-notch test machine calibrations were recorded. The temperature instrument calibrations were not recorded; however, thermometers qualified to ASTM standards were purchased, used for the certified time period, and replaced with new qualified thermometers.
Combustion Engineering required that all of its vendors who furnished materials or parts (for Farley Unit 2) to be on an approved vendors list.
Each vendor was required to have a quality control system in accordance with
- N-335 of the 1968 ASME Code through Summer 1970 Addenda. Periodic audits of these vendors were performed by CE QA personnel.
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QUESTION 122.4 (Cont'd)
Page 2-It should be noted that the Farley Unit 2 reactor vessel was partially furnished by B&W. Material furnished by B&W was accepted on the basis of material certifications; therefore, no QA audits were performed for those by CE.
Steam Generators and Pressurizers - Charpy V-notch test machine calibration at Westinghouse Tampa plant was performed yearly using samples obtained from Watertown Arsenal. Temperature instrument calibration was performed with standards traceable to the National Bureau of Standards.
Nll material suppliers have been either surveyed by ASME auditors or Westinghouse Tampa Plant Product Assurance to obtain supplier certifications.
4 A' sampling of one of the major material suppliers indicated that Charpy V-notch test machine calibrations were recorded and that calibrated tempera-ture instruments were purchased (as replacements) on a yearly basis.
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QUESTION 122.5 Provide information, including training and experience, to demonstrate that the qualification of individuals who performed fracture toughness tests are in compliance with the requirements of Paragraph III.B.4, Appendix G,10 CFR Part 50.
RESPONSE
The Farley Unit 2 reactor vessel, steam generators, and pressurizer were fabricated to ASME Code Section III requirements in effect prior to the issuance of 10 CFR 50, Appendi.x G.
However, the following discussion demonstrates that the intent of the Appendix G requirements is satisfied.
Reactor Vessel - The personnel perfonning the Charpy testing at Combustion Engineering were qualified by schooling, training, and many years of experience. Their qualifications to perform this work have been certified by qualified supervisory personnel. This meets the requirements of the applicable ASME Code 1968 Edition through Summer 1970 Addenda (Appendix IX 221d.).
Steam Generators and pressurizer - Charpy impact tests are performed at Westinghouse Tampa Plant by Level III and Level II personnel who have a minimum of five years directly related testing experience.
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o QUESTION 122.6 Paragraph III.C.2, Ap?endix G,10 CFR Part 50, specifies that every fracture toughness test specir'en from the reactor vessel beltline be subjected to a heat treatmer;."at produces metallurgical effects equivalent to those produc,J ir. the vessel material throughout its fabrica-tion process.
Identify all specimens that do not meet this requirement and provide technical justification for use of such specimens in establish-ing fracture toughness properties of the reactor vessel beltline.
RESPONSE
Materials used to prepare fracture toughness test specimens from the reactor vessel beltline region were subjected to a heat treatment in accordance with the requirements of the Summer 1970 Addenda to the 1968 ASME Code Section III. The heat treatment applied to the specimens con-forms with paragraph NB-2211 of the present ASME Code Section III require-ments and, therefore, the heat treated condition of the test specimens is in compliance with paragraph III.C.2 of 10 CFR 50, Appendix G.
QUESTION 122.7 In weld seam 10-923, two of the nine specimens tested had impact energies less than 75 ft-lbs at a test temperature of 10 F.
To demonstrate compliance with the upper shelf impact energy requirements of paragraph IV.B of 10CFR Part 50, the applicant must supply additional data, possibly from either baseline surveillance material or information available in the literature, and/or analyses to define the minimum upper shelf energy for weld seam 10-923. The data and/or analyses used to demonstrate compliance with the upper shelf requirements of paragraph IV.B must include variables that affect upper shelf toughness, e.g., chemical composition, fabrication history, weld wire and heat of filler metal. The applicant must also supply the individual data points obtained from the Cy impact tests for each of the base metal heats in the reactor vessel beltline.
RESP 0NSE Although two test specimens for weld metal used in weld seam 10-923 0
exhibited impact energies of less than 75 ft-lb at a test temperature of 10 F, it is expected that the upper shelf impact energy requirement of 75 ft-lb identified in naragraph IV.B of 10CFR50 Appendix G would easily be exceeded if tests had been performed at test temperatures representative of the upper shel f.
A review of many weld test certificates provided by the vessel fabricator indicates that the upper shelf energy of welds of chemical composition and fabrication history similar to weld seam 10-923 and fabricated with the same 1;ype of wire and flux (Type B-4 weld wire and Linde 0091 Flux) used in seam 10-923 exceeds 75 ft-lb by a considerable margin. Four examples of the vessel fabrication' test results for weld material similar to that of seam 10-923 are shown in Table 122.7-1. Like weld seam 10-923, two of these four examples did not exhibit 75 ft-lb for all test specimens at 10 F; however, at higher temperatures, 75 ft-lb was exceeded.
QUESTION 122.7 (Cont'd).
Page 2 Individual data points obtained from Charpy V notch impact tests for each of *he base metal heats in the Farley Unit 2 reactor vessel beltline are presented in Tables 122.7-2,122.7-3, and 122.7-4.
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TABLE 122.7-1 Example 1 - Type B-4 Weld Wire and Linde 0091 Flux (Test #1302)
IMPACT AND/0R FRACTURE TESTS TYPE TEMP. vF VALUES TEMP.0F VALUES NDT FT/L8 % SHEAR MilsLatExp DROP WEIGHTS CVN
-80 3
0 1
- 50 1F
-400 F
-80 3
0 2
-40 1F
-80 9
0 4
-30 2 NF
-40 26 10 19
-40 37 15 25
-40 38 15 24
+10 69 35 46
+100 117 90 83
+10 50 25 38
+100 114 90 82
+10 66 30 44
+100 120 90 83
+20 66 55 46
+160 124 100 83
+20 81 50 57
+160 136 100 89
+20 90 60 63
+160 135 100 88 Example 2 - Type B-4 Weld Wire and Linde 0091 Flux (Test #1388)
IMPACT AND/0R FRACTURE TESTS TYPE TEMP. "F VALUES TEMP.0F VALUES NDT FT/LB % SHEAR MilsLatExp DROP WEIGHTS 0
CVN
-80 11 0
3
-60 1F
- 60 F
-80 11 0
3
-50 2 NF
-80 13 0
4
-40 1 NF
-40 30 15 17
-40 27 15 15
-40 25 10 11 0
77 50 45
+100 143 100 84 0
72 50 40
+100 133 100 82 0
70 50 41
+100 145 100 86
+10 76 50 41
+180 143 100 82
+10 74 50 46
+180 149 100 86
+10 82 60 45
+180 148 100 85
+60 116 70 76
+60 118 70 74
+60 1 21 70 71
TABLE 122.7-1 (Continued)
Example 3 - Type B-4 Weld Wire and Linde 0091 Flux (Test #1389)
IMPACT AND/OR FRACTURE TESTS TYPE TEMP. UF VALUES TEMP. CF VALUES NDT FT/LB % SHEAR MilsLatExp DROP WEIGHTS CVN
-60 16 0
9
.-60 1F
- 60 F
-60 15 0
7
-50 2 NF
-60 19 0
11
-40 1 NF
-40 20 5
11
-40 28 10 16
-40 32 15 22
-20 85 50 53
+60 132 80 77
-20 88 50 56
+60 149 100 84
-20 76 40 47
+60 123 80 74 0
77 40 47
+100 142 100 82 0
75 40 45
+100 148 100 84 0
99 60 52
+100 140 100 82
+20 117 70 74
+20 105 60 65
+20 11'4.
70 74 Example 4 - Type 8-4 Weld Wire and Linde 0091 Flux (Test #1386)
IMPACT AND/OR FRACTURE TEST 5 TYPE TEMP. UF VALUES TEMP. uF VALUES NDT FT/LB % SHEAR MilstatExp DROP WEIGHTS CVN
-80 16 0
7
-60 1F
-80 18 0
8
-50 2 NF
-80 18 0
7
-40 1 NF
-40 38 20 26
, 32 15-17
-40 34 15 19 0
79 40 52
+100 137 100 82 0
61 70 39
+100 132 100 82 0
95 70 60
+100 1 41 100 83
+10 96 70 62
+180 142 100 82
+10 101 70 60
+180 145 100 85
+10 84 60 58
+180 143 100 83 i
+60 118 80 -
78
+60 130 90 80 l
+60 117 80 75
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t TABLE 122.7-2 LOWER SHELL COURSE CHARPY V-NOTCH DATA
- Plate Code No. B7210-1 Plate Code No. 87210-2 Test Energy Lat. Exp Shear Test Energy Lat. Exp Shear Temp. (O )
(Ft-Lb)
(Mils)
(%)
Temp. (OF)
(Ft-Lb)
(Mils)
_lj]__
F
-50 10 6
9
-50 15 11 9
-50 14.5 8
15
-50 12.5 8
9
-50 11 7
9
-50 11 8
9 20 33 25 29 0
26 24 30 20 47 35 34 0
27.5 27 34 20 46 33 34 0
45 35 32 75 48.5 38 59 30 51 39 30 i
75 50 40 59 30 40 34 34 75 62 47 64 30 47 39 30 110 86 67 80 100 67 52 79 110 75 57 75 100 80.5 58 75 110 69.5 54 67 100 85 60 75 150 100 69 100 150 100 76 100 150 95 71 100 150 101 74 100 150 93 67 100 150 97 75 100 210 96 70 100 210 98 69 100 l
210 105.5 74 100 210 102 76 100 210 107-75 100 210 95.5 72 100 i
- Normal to major rolling direction of the plate
TABLE 122.7-3 INTERMEDIATE SHELL COURSE CHARPY V-NOTCH DATA
- Plate Code No. B7203-1 Plate Code No. B7212-1 Test Energy Lat. Exp Shear Test Energy Lat. Exp Shear-Temp. (O )
(Ft-Lb)
(Mils)
(%)
Temp. (OF)
(Ft-Lb)
(Mils)
(%)
F
-50 13.5 9
15
-50 18.5 11 12
-50 19 11 15
-50 15.5 11 12
-50 14 8
15
-50 19 11 12 0
28 25 30 0
35 27 27 0
34 26 28 0
34.5 27 25 0
44 36 34 0
30 27 25 20 55 41 40 30 43 35 32 20 51 38 45 30 48 36 35 20 43 32 34 30 52 39 43 75 50.5 50 56 100 76.5 55 73 75 61.5 40 52 100 74 56 73 75 65 46 61 100 70 54 69 150 91 68 100 150 95 67 100 150 97 76 100 150 98 68 100 150 92 70 100 150 106 76 100 210 105.5 69 100 210 89 68 100 210 97.5 74 100 210 94 70 100 210 95.5 72 -
100 210 88 69 100
- Normal to major rolling direction of the plate
TABLE 122.7-4 N0ZZLE SHELL COURSE CHARPY V-NOTCH DATA
- Forgoing Code No. B7216-1 Test Energy Lat. Exp Shear Temp. (oF)
(Ft-Lb)
(Mils)
(%)
-80 2
0 0
-80 4
0 0
-80 8
4 0
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-20 68 53 29
-20 37 25 16
-20 66 52 29 10 99 76 64 10 103 76 64 10 110 81 70 10 95 77 52 10 55 41 29 10 78 63 40 30 72 57 23 30 93 70 55 30 87 65 46 100 147 91 100 100 123 77 75 100 110 80 70 180 146 90 100 l
l 180 151 88 100 l
180 149 88 100
- Major working direction of forgoing y-y-I
4 QUESTION 122.8 The materials surveillance program uses six specimen capsules containing reactor vessel steel specimens of the limiting base material, weld metal material, and heat affected zone material. To demonstrate compliance with Appendix H,10 CFR Part 50, provide a table that includes the following information for all the surveillance specimens:
1.
Actual surveillance material; 2.
Beltline material from which the specimen was obtained; 3.
Test specimen type and orientation; 4.
Fabrication history of each test specimen; 5.
Chemical composition of each test specimen; and, 6.
Heat of filler material, production welding conditions, and base metal combinations for weld specimens.
Provide the lead factor for each specimen capsule calculated with respect to the vessel inner wall.
Resp 0NSE Refer to Table 122.8-1 for the information pertaining to Items 1, 2, and 4.
Refer to Table 122.8-2 for Item 3 and Table.122.8-3 for Item 5.
Regarding Item 6, the surveillance weldment was fabricated with the same type of wire and the same heat of wire (Wire Type E8018C3 and wire heat no. BOLA) as was used to fabricate the longitudinal weld seal (19-9238) in the intermediate shell course of the vessel. The same welding procedures were used to fabrica'te the surveillance weldment and the vessel weld seam-(19-9238).
Table 122.8-4 identifies the lead factor for each specimen capsule.
As indicated, the weld metal representative of the intermediate shell longitudinal weld seam is included in the Farley Unit 2 surveillance program; the surveillance weldment has a copper content of 0.03%. It should be noted that the vessel girth weld has a copper content of 0.13%,. and therefore is more limiting than the turveillance weldment. However, the base metal in l
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QUESTION 122.8 (Cont'd)
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i the surveillance program (from intermediate shell plate B7212-1) has a 0
copper content of 0.20% and a predicted RT shif t of 390 ; the base metal NDT is the most limiting material in the beltline region and will be used to define operating limits for Farley Unit 2.
In conclusion, although the most limiting weld metal in the beltline region is not included in the surveillance program, this is not of consequence since the weld metal is not the controlling beltline region material.
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TABLE 122.8-1 SURVEILLANCE MATERIAL BELTLINE LOCATION AND FABRICATIONS HISTORY Surveillance Beltline Location of Ma terial Surveillance Material' Heat Treatment 0
0 Base Metal Intermediate Shell Plate B7212-1 1550-1650 F-4Hr.-WQ,1200-1250 F-4Hr.-AC,1125-ll75 F-18Hr.-FC Weld Metal
- Intermediate Shell Longitudinal Weld Seam 1125-1175 F-13Hr.-FC 0
HAZ Metal Intermediate Shell Plate B7212-1 ll25-1175 F-13Hr.-FC
- Surveillance weldment fabricated using Plate B7212-1 and B7203-1. Surveillance weldment was fabricated using the same type of wire (E8018C3) and the same heat of wire (Heat No. BOLA) as was used to fabricate the intermediate shell longitudinal weld seam (19-9238) in the vessel. The same welding procedures (MA-511-D and A-244-110-8) were used by the vessel supplier to fabricate the surveillance weldment and the intermediate shell longitudinal weld seam (19-9238).
WQ - Water Quench AC - Air Cooled FC - Furnace Cooled 1
TABLE 122.8-2 SURVEILLANCE TEST SPECIMENS TYPE, ORIENTATION, AND QUANTITY PER TEST CAPSULE Surveillance Specimen Material Orientation Charpy V Tensile 1/2 T-CT Base heril (Plate B7212-1)
Transverse 15 3
4 Base Metal (Plate B7212-1)
Longitudinal 15 3
4 Weld Metal Transverse 15 3
4 HAZ Metal (Plate B7212-1)
Longitudinal 15
TABLE 122.8-3 SURVEILLANCE MATERIAL CHEMICAL COMPOSITION (WT. ")
Element Plate B7212-1 Weld Metal C
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0.21 0.086 Mn 1.30 0.95 P
0.018 0.004 S
0.01 6 0.014 Si 0.24 0.34 Ni 0.,60 0.90 Cr 0.15
<0.01 Mo 0.49 0.23 Cu 0.20 0.03 V
0.003 0.006 Co' O.027 0.010 Sn 0.011 0.002 A1 0.040 0.003 N
0.006 0.007 2
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n-TABLE 122.8-4 SPECIMEN CAPSULE LEAD FACTOR For the six surveillance program capsules, the lead factor calcu-lated with respect to the vessel inner wall is as follows:
Capsule Lead Factor 0
V (107 )
3.5 0
X (287 )
3.5 U (3430) 3.5 0
W (110 )
2.9 9
Y (290 )
2.9 0
Z (340 )
2.9
4 QUESTION 122.9 To demonstrate the integrity of the reactor coolant pump flywheels, supply the Charpy V-notch impact and tensile data for each flywheel, explicitly stating the material used for each flywheel.
Also, confirm that welding, including repair welding, was not performed on any finished flywheel.
If welding were performed, identify the flywheel (s) and location of the welds.
RESPONSE
The reactor coolant pump flywheels consist of two discs bolted together with three keyways to key the flywheel to the motor shaft. No welding, including repair welding, is perfomed on the flywheels.
Each flywheel disc is fabricated of ASTM A533 Grade B Class 1 steel. The Charpy V-notch and tensile data (from material certifications) for the flywheel discs are tabulated in Tables 122.9-1 and 122.9-2.
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TABLE 122.9-1 YIELD TENSILE MOTOR MELT STRENGTH STRENGTH UNIT NUMBER (PSI)
(PSI) 01 R0317 74,300 91,100 69,500 91,100 RCi87 65,400 87,000 70,500 90,000 02 R0174 62,600 87,100 69,800 91,100 R0187 65,400 87,000 70,500 90,000 03 R0194 64,300 86,200 63,400 84,700 R0187 65,400 87,000 70,500 90,000
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4 TABLE 122.9-2 CHARPY LATERAL TEST l
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RTNDT MOTOR MELT V-NOTCH EXPANSION TEMPERATURE O
(F )
(F )
UNIT NUMBER (FT-LB)
(IN.)
O (F )
f 01 R0317 58,52,52(T)
.050,.043,.044(T) 70 10 10 140,178,182(L)
.080,.079,.086 (L) 70 R0187 128,128,118 (T)
.089,.087,.086(T) 70 10 10 135,150,165 (L)
.098,.088,.098 (L) 70 02 R0174 120,103,120 (T)
.082,.091,.088(T) 70 10 10 110,131,123 (L)
.090,.083,.087 (L) 70 R0187 128,128,118 (T)
.089,.087,.086 (T) 70 10 10 l
135,150,165 (L)
.098,.088,.098 (L) 70 03 R0194 90,100,110 (T)
.083,.072,.086(T) 70 10 10 112,120,130 (L)
.091,.086,.085 (L) 70 R0187,
128,128,118 (T)
.089,.087,.086(T) 70 10 10 135,150,165 (L)
.098,.088,.098(L) 70
- Based on two drop weight tests exhibiting no break performance at 20 F.
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