ML20101L502
| ML20101L502 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 07/01/1992 |
| From: | Woodward J SOUTHERN NUCLEAR OPERATING CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| GL-92-01, GL-92-1, NUDOCS 9207070153 | |
| Download: ML20101L502 (38) | |
Text
S t nN ar O ating Ccmpany Damingham, Atat>ama 35?01 lelephone ?05 868 SOBG Southem Nudear Operating Company 1 o, y, w,,,
$y July 1, 1992 10 CFR 50.54(f)
Docket Nos. 50-348 50-364 U. S. Nuclear Regulatory Commission 1
ATTN: Document Control Desk Washington, DC 20555 Joseph M. Farley Nuclear Plant Generic Letter 92-01. Reactor Vessel Struttural Intearity Gentlemen:
On. February 28,.1992 the NRC issued Generic letter (G.L.) 92-01, " Reactor-Vessel Structural Integrity" which wF subsequently replaced by Revision I dated March 6.-1992.
The stated purpose of this generic letter is to obtain information needed to assess compliance with requirements and commitments regarding reactor vessel integrity in light of events associated with Yankee Nuclear Power Station. 'In G.L. 92-01, the NRC requested information regarding compliance with the requirements of 10 CFR 50.60 and 50.61 and. commitments made in response to C.L. 88-11. "NRC Position on Radiation Embrittlement of Reactor Vessel Materials and its Impact on Plant Operations." Specific responses to each of the requests listed in G.L. 92-01~ for Farley Nuclear Plant Units 1 and 2 are provided as Attachments =1 and 2, respectively, in summary, Southern Nuclear _ Operating Company has reviewed the recuirements and commitments related to reactor vessel structural integrity anc has verified continued compliance.
It should be noted that the information providad in Attachments 1 and 2 was extracted from the material certification reports provided by the manufacturer of the reactor vessels and is the most accurate available to Southern Nuclear Operating Company.
If there are any questions, please_ advise.
Respectfully submitted, h
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_ \\ Q. Woodard JDW/BHW: map 2531 Attachments-SWORN TO AND SUBSCRIBED BEFORE ME cc: Mr. S. D. Ebneter y/
Mr. S. T. Hoffman THIS /
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ATTACHMENT 1 l-.
JOSEPH M. FARLEY UNIT 1 RESPONSE T0-GENERIC LETTER 92-01 1
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ATTACHMENT 1 Page 1 i
- 1. - 120._Bgquell Addressees who do not have a surveillance program meeting ASTM E 185-73, -79, or -82 and who-do not have an integrated ' surveillance program approved by the NRC, are requested to describe actions taken or te be taken to ensure compliance with Appendix H to 10 CFR Part 50.
SNC Response l
As-stated in NUREG-75/034, " Safety Evaluation Report Joseph M. Farley Nuclear Plant Units 1 and 2," dated May 2, 1975, the toughness properties of the reactor vessel beltline m:terial will be monitored throughout service life in accordance with the requirements of ASTM E-185-73 and 10 CFR Part 50, Appendix H.
The SER concludes that the reactor vessel material surveillance program constitutes an acceptable basis for monitoring radiation induced changes in the future toughness of the reactor vessel material, and satisfies the requirements of General Design Criterion 31 of 10 CFR Part 50, 2.
NRC Reouest Addressees of plants for which the Charpy upper shelf energy is predicted to be less than 50 foot-pounds at the end of their licenses using the guidance in Paragraphs C.1.2 or C.2.2 in Regulatory Guide 1.99, Revision 2, are requested to provide to the NRC the Charpy upper shelf _ energy predicted for December 16, 1991, and for the end of their current license for the limiting beltline weld and the plate or forging and are requested to describe the actions taken pursuant to Paragraphs IV.A.1 or V.C of Appendix G to 10 CFR Part 50.
SNC Response Table 1 contains the 32 EFPY Charpy upper shelf energy for Farley Unit 1 beltline materials. The calculated 32 EFPY Charpy upper she',f energy for all the beltline region plates are: predicted to be above the 50 foot-pound criteria.
TABLE 1 J. M. FARLEY UNIT 1 CALCULATED 32 EFPY UPPER SHELF ENERGY E0L USE MATERIAL DESCRIPTION (Ft-Lbs)
Inter. Shell, B6903-2 71.8 Inter. Shell, B6903-3 66 Lower Shell, 86919-1 61.92 Lower Shell, B6919-2 61.92 Inter. Shell Long. Welds (a)
Lower Shell Long. Welds (a)
Inter./ Lower Circ. Weld (a)
Surveillance Weld 98,34 (a) No estimate since upper shelf energy values are not available because unirradiated Charpy '/-notch impact tests were not performed.
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ATTACHMENT 1 Page 2 w
3.
NRC Reouest
' Addressees whose-reactor vessels were constructed to an ASME Code earlier' than the Summer 1972 Addenda of the 1971 Edition are requested to describe the consideration given to the following material properties in their evaluations performed pursuant to 10 CFR 50.61 and Paragraph III. A of 10 CFR Part 50, Appendix G:
(1) the results from all Charpy and drop weight tests for all unirradiated beltline materials, the unirradiated reference temperature for each beltline material, and the method of determining the unirradiated reference temperature from the Charpy and drop weight test; (2) the heat treatment received by all beltline and surveillance materials; (3) tne heat number for each beltline plate or forging and the heat number of wire and flux lot number used to fabricate eech beltline weld; (4)
"i ' '+ number for each surveillanc; plate or forging and the heat number of w flux lot number used to fabricate the surveillance weld;
- )
t'
+
al composition, in particular the weight in percent of copper, nickel, s, and sulfur for each beltline and surveillance material; and
. umber of the wire used for determining the weld metal chemical
pn if different than item (3) above.
SNC Response The J. M. Farley Unit I reactor vessel was constructed to the 1970 Summer Addenda to Section III of the ASME Code.
Thus, the J. M. Farley Unit I reactor vessel was constructed to an ASME Code earlier than the Summer 1972 Addenda of the 1971 Edition.
Tables 2 through 12 document the unirradiated data (Charpy and drop weight test results, reference temperature, upper shelf energy), heat treatment, heat numbers, weld flux lot number, and chemical composition for all beltline region and surveillance materials. These values were developed using the current reactor pressure vessel material tesi requirements and acceptance standards at the time of fabrication.
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age 3 TABLE 2 ATTACHMENT 1 Joseph M. Farley Unit I litterials Certification 1nformat193 Materia'is C The following information was taken from the 19, 1970.
Combustion Engineering, Inc. on June lin1 Jig a C6294-1 Intertnediate Shell Plate, B6903-2 fompo_nanti tillLChamkal Analy111 n
I f
Mo Cu f
S Si Ni 0.011l0.013j 0.21_ j 0.60 j 0.55 l0.13 l 0.01 C
Mn P
0.20 1.32_
j
[httgy_Jmprt_and Fratture Tetti f% Shear lMilslat.E Temp. *F l
Ft-lbs I
23 jf
/
10
]l f
31
[
40 l
20 l
10 l
40 l
28 I
17 O
jl 40 j
19 l
61_
I 40 l
85 l
64
+ 10 l
45 lI_
l 93
+ 10 f
70 45 l
96 lf
+ 10 65 45
]l f
96 l
jf ll
+ 40 70 f
50
+ 40 l
105 66 50_
f 103 jl
+ 40 91 99 141 ll
+110 90 95 lf
+110
/
135 fl f
85 l
86
)l I
+110 j
132 I
88 100 I
+160 l
153 j
}l 88 l
100 l
155
+1E0 87 0_0d_
147
+
DT ig Temp.
f jl l[l_
1-F f
f(
- 20 l
l-F f
0 F jl f_
10 hF 1-NF 0
+ 10
)
2-NF hat _T_tutme_nt 0
Water quenched, 1125 F t 25"F, 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
Furnace cooled to 600 F.
l 1550-1650 F, 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
5 F, 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br />.
ATTACHMENT 1 TABLE 2 Page 3 Joseph H. Farley Unit I Materials Certification Information The following information was taken from the Materials Certification Report prepared by Combustion Engineering, Inc. on June 19, 1970, Component:
Intermediate Shell Plate, B6903-2 Heat No.:
C6294-1 MILL Chemical Analysis C
Mn P
S Si Ni Mo Cu Al Co 0.20 1.32 0.011 0.013 0.21 0.60 0.55 0.13 0.017 0.015 Charov Imoact and Fracture Tests Temp. 'F Ft-Lbs
% Shear Mils Lat. Exp.
- 40 31 10 23
- 40 28 10 20
- 40 19 0
17
+ 10 85 40 61
+ 10 93 45 64
+ 10 96 45 70
+ 40 96 45 65
+ 40 105 50 70
+ 40 103 50 66
+110 141 99 91
+110 135 95 90
+110 132 85 86
+160 153 100 88
+160 155 100 88
+160 147 100 87 Temp. 'F Drop Weights NDT l
- 20 1-F
- 10 1-F l
0 1-F 1-NF 0F
+ 10 2-NF Heat Treatment 1550-1650 F, 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
Water quenched.
1125 F 25 F, 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
1150 F i 25 F, 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br />.
Furnace cooled to 600 F.
ATTACHMENT 1 TABLE 3 Page 4 Joseph M. - farley Unit 1 Materials Certification Informatig.D The following information was..taken from the Materials Certification Report prepared by Combustion Engineerin9, Inc. on June 19, 1970.
Cpmoonent.1 Intermediate Shell Plate, B6903-3 Heat N u C6308 Mill Chemical Analysis i
C Mn P
S Si Ni Mo Cu t1,_j.
Co 0.21 1.29 0.014 0.015 0.16 0.56 0.56 0.12 0.019 0.023 Charoy Impact and Fracture Tests Temp.
F Ft-Lbs
% Shear Nils !at. Exp.
- 80 12 0
15
- 80 8
0 10
- 40 43 15 30
- 40 23 10 16
- 40 50 20 36
+ 10 65 25 49
+ 10 90 40 62
+ 10 66 25 47
+ 40 72 30 56
+ 40 96 45 62
+ 40 81 35 56
+110 109 85 78
+110 126 90 86
+160 132 100 84
+160 137 100 90
_ Temp.
F Drop Weights NDT 0
1-F
+ 10 1-F
+10 F
+ 20 2-NF Heat Treatment 1550-1650 F, 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. Water quenched.
1225 F i 25 F, 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, 1150 F i 25 F, 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br />.
Furnace cooled to 600 F.
. ATTACHMENT 1~
TABLE 4 Page 5 Joseph M. Farley Unit 1 Materials Certification Information The following information was taken from the Materials Certification Report prepared by Combustion Engineering, Inc on July. 30, 1970.
ComDonent:
Lower Shell Plate, B6919-1 Heat No.:
C6940-1 MILL Chemical Analyih I
C Mn P
S Si Ni Cu i
Co 0.20 1.39 0.015 0.015 0.18 0.55 0.56 0.14 0.025 0.008 CharDY ImDact and Fracture Tests Temp. 'F Ft-lbs
% Shear Mils Lat. Exp.
- 80 9
0 15
- 80 6
0 6
- 40 42 20 30
- 40 27 15 21
+ 10 82 40 61
+ 10 62 30 48
+ 10 58 25 44
+ 40-77 40 58
+ 40 80 40 60
+ 40 87 45 62
+110 116 80 82
+110 140 100 86
+110 123 100 84
+160 130 100 87
+160 136 100 88 Temp.
F Drop Weights NDT
- 20 1-F
- 10 2-NF
- 20of 0
1-NF Heat Treatment 1550-1650 F, 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
Water quenched.
1225 F i 25 F, 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
1150 F t 25 F, 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br />.
Furnace cooled to 600oF.
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- ATTACHMENT 1 TABLE 5 Pa9e 6 Joseph M. Farley Unit I l
Materials Certificat. ion Information
- The following information _was taken from the Materials Certification Report prepared by Combustion Engineering, Inc. on July 30, 1970.
Component:
Lower Shell Plate, B6919-2 Heat No.1 C6697-2 MILL Chemical Analysis
_ C Mn P
S Si Ni Mo Cu Al Co
_ 0.20 1.39 0.015 0.018 0.19 0.56 0.53 0.14 0.018 0.010 Charoy Imoact and Fracture Tests Temp.
F Ft-Lbs
% Shear Mils Lat. Exp.
- 80 11 0
4
- 80 7
0 3
- 40 4J, 20 34
- 40 30 15 23
+ 10 67 30 49
+ 10 67 30 50
+ 10 57 25 40
+ 40 84 40 60
+ 40 86 40 63
+ 40 81 35 58
+110 111 90 87
+110 112 80 82
+110 132 100 88
+160 135 100 82
-+160 133 100 84 Temp.
F Drop Weights NDT
- 20 1-F
- 10 1-F
-10 F 0
2-NF Heat Treatment 1550-1650 F, 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. Water quenched.
1225 F 25 F, 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, 1150 F i 25 F, 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br />.
Furnace cooled to 600 F.
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ATTACHMENT 1 Page 7 TABLE 6 Joseph M. Farley Unit 1 Materials Certification Information The' following information was taken from the Materials Certification Report prepared by Combustion Engineering, Inc. on October 19, 1971.
Component:
Intermediate to Lower Shell Circle Seam,11-894 Weld Wire:
Type B4, Heat No. 6329637 Flux 1 Linde 0091, Lot No. 3999 MILL Chemical Analysis C
Mn P
S Si Mo Cu 0.14 1.15 0.011 0.014 0.19 0.53 0.24 Charov Imoact Tests Temp.
F Ft-Lbs NDT F
+ 10 101
+ 10 108 0 (a)
+ 10 103 (a) Estimated per NRC Standard Review Plan Section 5.3.2 Hgat TreatmeD1 1150 F t 25 F, 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br />.
Furnace cooled to 600 F.
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ATTACHMEMT 1 TABLE 7 Page 8 Joseph M. Farley Unit 1 Materials Certification Informallan The following information was taken from the Materials Certification Report prepared by Combustion Engineering, Inc on Octolser 27, 1969.
Component:
Intermediate Shell Long. Seams, 19-894A&B Weld Wire:
Type B4, Heat No. 33A277 Flux:
Linde 1092, Lot No. 3889 MILL Chemical Analysis C
Mn P
S Si Mo Cu 0.11 1.27 0.015 0.010 0.14 0.49 0.27 Charov Impact Tests Temp.
F Ft-Lbs NDT F
+ 10 103
+ 10 105 0 (a)
+ 10 108-(a) Estimated per NRC Standard Review Plan Section 5.3.2 Heat Treatment 1125 F 25 F, 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br />.
Furnace cooled to 600of.
ATTACHMENT'l' TABLE 8 Page 9 Joseph M. Farley linit-1 Materials Certification Information 1The following information was taken from the Materials Certification Report prepared by Combustion Engineering, Inc. on October 7, 1970.
[paponent:
Lower Shell Long. Seams, 20-894A&B Weld Wire:
Type B4, Heat No. 90099 Flux:
Linde 0091, Lot No. 3977 d]J.L Chemical Analysis C
Mn P
S Si Mo Cu 0.15 1.12 0.022 0.012 0.23 0.49 0.17 Charoy Imoact Tests Temp.
F Ft-Lbs NOT *F
+ 10 56
+ 10 30 0 (a)
+ 10 52 (a) Estimated per NRC Standard Review Plan Section 5.3.2 Heat Treatment 1150 F t 25 F, 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br />. -Furnace cooled to 600 F.
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ATTACHMENT 1 TABLE 9 Page 10 Joseph M. Farley Unit 1 Surveillance Caosule Materials Data The following information was taken-from WCAP-8810. " Alabama Power Company Joseph M. Farley Nuclear Plant Unit No.1 Reactor Vessel Radiation Surveillance Program,"
December 1976.
. Component:
Surveillance Material SA533 Grade B Class 1 plate used in the core region lower shell plate, B6919-1.
Chemical Analysis Lower Shell Plate, B6919-1 Chemical analyses were performed by both Combustion Engineering and Westinghouse.
The results of both analyses are given below.
Element Combustion Engineering Westinghouse Analysis Analysis C
0.20 S
0.015 0.013 N
0.00?
Co 0.008 0.016 Cu 0.14 0.10 Si 0.18 0.28 Mo 0.56 0.51 Ni 0.55 0.56 Mn 1.39 1.40 Cr 0.13 V
<0.001 P
0.015 0.015 Sn 0.008 Al 0.025 Heat Treatment Lower Shell Plate, 86919-1
-1550 F/1650 F, 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
W9.ter quenched.
1225 F 25 F, 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
1150 F 25 F,-40 hours.
Furnace cooled to 600 F.
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l-Drooweicht Tests Lower Shell Plate, B6919-1 The nil-ductility transition temperature (NDTT) was determined for plate B6919-1 to be -20 F by drop weight tests (ASTM E-208) performed by Combustion Engineering.
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ATTACHMENT 1-Page 11 TABLE 9 continued Charov Inoact and Fracture Tests Lower Shell. Plate, 86919-1 (Longitudinal)
Temp. 'F Ft-Lbs
% Shear Mils Lat. Exp.
- 50 26 14 18
- 50 12 10 8
- 50 11 14 8
-0 59 30 44 0
48 27 37 0
56 27 38
+ 40 80 50 62
+ 40 52 35 44
+ 40 68 42 53
+ 80 107 80 70
+ 80 100 80 73
+ 80 106 80 71
+130 135 100 80
+130 14C
-100 90
+130 145 100 88
+210 131 100 83
+210 129 100 85
+210 142 100 85
ATTACHMENT 1 Page 12 TABLE 9 continued l
Charov impact and Fracture Tests lower Shell Plate, B6919-1 (Transverse)
Il Temp.
F Ft-Lbs
% Shear Mils Lat. Ex L
- 40 12 14 7
- 40 25 14 17
- 40 29 20 18 0
30 25 21 0
27 25 24 0
35 25 26
& 40 26 37 26
+ 43 37 29 30
+ 40 44 52 37 RT 60 55 50 RT 53 64 45 RT 53 50 46
+110 75 77 56
+110 69 79 53
+110 80 90 52
+210 92 100 71
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+210 91 100 72
+210-89 100 68 I
ATTACHMENT 1 TABLE 10 Page 13 Joseph M. Farley Unit 1 Surveillance Caosule Materials Data Component:. Surveillance. Weld Material made from sections of lower shell plate, B6919-1 and adjoining intermediate shell plate B6903-2, using weld wire representative of tnat used in the original fabrication.
Chemical Analysis The following data was obtained from Westinghouse Analysis.
C Mn P
S Si N i_.
No Cu Al Co 0.13 1.06 0.016 0.009 0.27 0.19 0.50 0.14 0.009 0.018 N
Cr V
Sn 0.005 0.063 0.003 0.005 Cha_rp.y_Imgact and Fracture Tests r
Temp. *F Ft-Lbs
% Shear Mils Lat. Exp.
-100 5
15 1
-100 14 25 11
-100 18 20 11
- 40 53 43 43
- 40 60 32 44
- 40 75 50 55
+ 10 79 65 54
+ 10 86 73 63
+ 10 80 65 58
+ 72 117 100 82
+ 72 123 100 80
+ 72 113 100 79
+150 151 100 90
+150 144 100 89
+150 118 100 81
+210 138 100 88
+210 159 100 85
+210 151 100 85 Heat Treatment y
1150 F i 25 F, 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br />.
Furnace cooled.
Drooweicht Tests The nil-ductility transition temperature (NDTT) was determined for the surveillance weld metal to be -60 F by drop weight tests (ASTM E-208) performed by Combustion Engineering.
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ATTACHMENT 1 Page 14 TABLE 11 e -
iloseph M. Farley Unit 1 Surveilla_nce Caosule Materials-Data Component: Surveillance Weld-Heat-Affected Zone (HAZ) Material were obtained from the weld-heat-affected zone of lower shell plate, 86919-1.
Charny Imoact and Fracture Tests Temp. 'F Ft-Lbs
% Shear Mils 1.at. Exp.
-150 15 18 7
-150 11 18 6
-150 58 45 29
-100 103 55 54 1
-100 33 32 19
-100 67 45 40
- 75 101 65 62
- 75 110 65 57
- 20 122 80 72 l
- 20 120 80 71
,_- 20 84 65 52
+ 50 141 100 83
+ 50 142 100 85
+ 75 (Es 100 74
+210 132 100 85
+210 170 100 83 j
+210 163 100 85 Drooweiaht Tests rae nil-ductility transition temperature (NDTT) was determined for the surveillance weld-heat-affected ' zone material to be -10oF by drop weight tests (ASTM E-208) performed by Combustion Engineering.
I The initial RT,,, values for the beltline region plates are calculated based on the information given in the Combustion Engineering Material Certification Reperts using the methodology presented in the Branch Technical Position MTEB 5-2 contained in the Standard Review Plan.
The initial RT,., values for the beltline region welds are generic mean values given in 10 CFR 50.61.
The upper shelf energy (USE) values for the beltline region plates are calculated based on the information given in the Combustion Engineering Material Certification Reports using the methodology presented in the Branch Technical Position MTEB 5-2 contained in the Standard Review Plan.
The USE values for the weld metal were not determined by the fabricator.
The beltline region welds only had fabrication Charpy tests performed at 10 F, so initial USE values could not be determined.
However, the surveillance weld, which is representative of the intermediate shell longitudinal welds, had an unirradiated USE of 149 ft-lb.
Evaluation according to Regulatory Guide 1.99, Revision 2, of the intermediate shell longitudinal welds assuming the initial USE of 149 ft-lb results in the 32 EFPY USE prediction of 98 ft-lb.
The initial RT,,1 and USE values are presented in Table 12.
TABLE 12 INITIAL RT,,, AND UPPER SHELF ENERGY VALUES FOR J. M. FARLEY UNIT 1 INITIAL RT,c1 USE MATERIAL DESCRIPTION F
LF_t_-J bM Inter Shell, B6903-2 0
152 (99)
Inter Shell, 86903-3 10 134.5 (87)
Lower Shell, B6919-1 15 132 (86)
Lower Shell, B6919-2 20 133 (86)
Inter Shell Long Welds
-56 (a)
(b)
Lower Shell Long Walds
-56 (a)
(b)
Inter / Lower Circ Weld
-56 (a)
(b)
Surveillance Plate (c) 140 (90.7)
Surveillance Weld (c) 149 Notes:
Numbers in () are for transverse data - estimated per MTEB 5-2.
(a) Calculated based on 10 CFR 50.61 methodology.
(b) Upper Shelf energy values are not available because unirradiated Charpy V-notch impact tests were not performed.
(c)
Initial RT,c1 not determined for surveillance material.
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ATTACHMENT 1 Page 16 4.
NRC Reauest Describe how-the embrittlement effects of operating et en irradiation temperature
-(cold leg or recirculation suction temperature) below 525'F were considered.
Describe consideration given to determining the effect of lower irradiation temperature on the reference temperature and-on the Charpy upper shelf energy.
SNC Response The requirements of Technical Specification 3/4.1.1.4, " Minimum Temperature for Criticality," prohibits plant criticality below 541"F.
A review of plant operating records indicates that RCS average temperature has never dropped below 525 F while the core was critical.
Inerefore, the concern with increased embrittlement effects due to operation below 525 F is not applicable to Farley Nuclear Plant, Unit 1.
5.
NRC Reauett Describe how surveillance results on the predicted amount of embrittlement were considered.
SNC Response As stated in Alabama Power Company's response to NRC Generic Ietter 88-11 for the Farley Unit I reactor vessel dated November 23, 1988, the piessure-temperature limits developed for J. M. Farley Unit I were based upon the methoc' ology of the Proposed Regulatory Guide 1.99, Revision 2.
This Regulatory Guide was made effective May, 1988 and includes the same information as the proposed rule.
The current f arley Unit I heatup and cooldown curves (16 EFPY) are applicable up to 18 EFPY.
These results were based on the limiting material in the beltline region, lower shell plate B6919-2 using surveillance capsule data.
The RT,,, values calculated for 32 EFPY are within the screening critern of 10 CFR 50.61 as indicated in Table 13. These values were calculated using the data presented in Tables 2 - 8 and Table 12.
The nickel content for the welds were not available from the material certificaticns, therefore the nickel values of 0.20 and 0.21 were used as presented in the FSAR for the welds. The copper values used in calculating the RT,1, values given in Table 13 are those given in the material certifications.
TABLE 13 J. M. FARLEY UNIT 1 RT,13 VALUES FOR 32 EFPY RT,,,
tLATERI AL DESCRIPTION F
Inter Shell, B6903-2 156
-Inter Shell, B6903-3 154 Lower Shell, B691921 180 (168)
Lower Shell, B6919-2 186 Inter Shell Long Welds 153 Lower Shell Long Welds 108 Inter / Lower Circ Weld 172
()lndicates RT,,, value based on surveillance capsule data.
m ATTACHMENT 1 Page l'/
6.
NRC Reouelt.
lf a measured increase in reference temperature exceeds the mean-plus-two standard deviations predicted by Regulatory Guide 1.99, Revision 2, or if a measured decrease f
in Charpy upper shelf energy exceeds the value predicted using the guidance in Paragraph C l.2 in Regulatory Guide 1.99, Revision 2, reoort the information and describe the effect of the surveillance results on the adjusted reference temperature and Charpy upper shelf energy for each beltline material as predicted for December 16, 1991, and for the end of its current license.
SNC Response The measured increase in reference temperature does not exceed the mean-plus-two standard deviations predicted by the Regulatory Guide 1.99, Revision 2, for any of the surveillance capsule materials as indicated in Table 14.
TABLE 14 J. M. FARLEY UNIT 1 MEASURED INCREASE IN REFERENCE TEMPERATURE VERSUS REGULATORY GUJDE 1.99, REVISION 2, PREDICTIONS ART, ( F)
SURVEILLANCE R.G. 1.99, REV 2 MATERIAL DESCRIPTION CAPSULE MEASURED PREDICTION 4 2a_
Plate, B6919-1 Y
85 115 (Longitudinal)
U 105 144 X
135 158 Plate, B6919-1 Y
55 115 (Transverse)
U 90 144 X
105 158 Weld Metal Y
80 121 0
80 163 X
100 155 The measured decrease in Charpy upper shelf energy does not exceed the value predicted using methodology specified in Regulatory Guide 1.99, Revision 2, for any of the surveillance capsule materials as indicated in Table 15.
TABLE 15 J. M. FARLEY UNIT 1 MEASURED DECREASE IN CHARPY UPPER SHELF ENERGY VERSUS REGULATORY GUIDE 1.99, REVISION 2, PREDICTIONS DECREASE IN UPPER SHELF ENERGY (%)
SURVEILLANCE R.G. 1.99, REV 2 MATERIAL DESCRIPTION CAPSULE MEASURED PREDICTION Plate, B6919-1 Y
9 20 (Longitudinal)
U 21 26 X
19 29 Plate, B6919-1 Y
0 20 (Transverse)
U 9
26 X
11 29 Weld Metal Y
13 24 V
28 31 X
23 35
ATTACllMENT 2 JOSEPH M. FARLEY UNIT 2 RESPONSE TO GENERIC LETTER 92-01 t-
ATTACHMENT 2 Page 1 1.
MR.C.bguest
' Addressees who do not have a surveillance program meeting ASTM E 185-73, -79, or -82 and who do not have an-integrated surveillance program approved by the NRC, are requested to describe actions taken or to be taken to ensure compliance with Appendix H to-10 CFR Part 50.
SNC Response As stated in NUREC-Oll7, Supplement 5, " Safety Evaluation Report Related to the Operation of Joseph M. Farley Nuclear Plant Unit 2," dated Maren 1981, all requirements of 10 CFR Part 50, Appendix H, are met with the exception of Paragraph li.B.
Paragraph II.B requires the beltline region of the reactor vessel to be monitored by a surveillance program complying with ASTM Standard E-185-73.
The weld material included in the Farley Unit 2 surveillance program is not the most limiting weld in the reactor vessel beltline region. Alabama Power Company )rovided sufficinnt justification for this exemption and the NRC subsequently granted t11s exemption.
The reasons stated for granting this exemption are: 1) the surveillance program includes the beltline material predicted to be most limiting (base plate B7212-1), and 2) conservative analysis methods contained in Regulatory Guide 1.99 are available to determine the radiation characteristics of the limiting beltline weld.
2.
NRC Reouest Addressees of plants for which the Charpy upper shelf energy is predicted to be less than 50 foot-pounds at the end of their licenses using the guidance in Paragraphs C.l.2 or C.2.2 in Regulatory Guide 1.99, Revision 2, are requested to provide to the NRC the-Charpy upper shelf energy predicted for December 16, 1991, and for the end of their current license for the limiting beltline weld and the plate or forging and are requested to describe the actions taken pursuant to Paragraphs IV.A.1 or V.C of Appendix G to 10 CFR Part 50.
SNC Responig Table I contains the 32 EFPY Charpy upper shelf energy for Farley Unit 2 beltline material s.
The calculated 32_ EFPY Charpy upper shelf energy for all the beltline region plates are predicted to be above the 50 foot-pound criteria.
TABLE 1 J. M. FARLEY UNIT 2 CALCULATED 32 EFPY UPPER SHELF ENERGY E0L USE MATERIAL DESCRIPTION (Ft-lbs)
Inter. Shell, B7203-1 71.6 Inter. Shell, B7212-1 64.8 Lower Shell, B7210-1 75.2 Lower Shell, 87210-2 71.3 Inter. Shell Long. Weld:
(a)
Lower Shell Long. Welds (a)
Inter./ Lower Circ. Weld (a)
(a) No estimate since upper shelf energy values are not available because unirradiated Charpy V-notch impact tests were not performed.
L ATTACHMENT 2 Page 2
+
3, NRC Reauest Addressees whose reactor vessels were constructed to an ASME Code earlier than the Summer 1972 Addenda of the 1971 Edition are requested to describe the consideration given to the following material properties in their evaluations performed pursuant to 10 CFR 50.61 and Paragraph III.A of 10 CFR Part 50, Appendix G:
(1) the results from all Charpy and drop weight tests for all unirradiated beltline materials, the unirradiated reference temperature for each beltline material, and the method of determining the unirradiated reference temperature from the Charpy and drop weight test; (2) the heat treatment received by all beltline and surveillance materials; (3) the heat number for each beltline plate or forging and the heat number of wire and flux lot number used to fabricate each beltline weld; (4)_ the heat nuwber for each surveillance plate or forging and the heat number of wire and flux lot number used to fabricate the surveillance weld; (5) the chemical composition, in particular the weight in percent of copper, nickel, phosphorous, and sulfur for each beltline and surveillance material; and (6) the heat numoe* of the wire used for determining the weld metal chemical composition if different than Ite, (3) above.
SNC Response
- The J. M. Farley Unit 2 reactor vessel was constructed to the 1970 Summer Addenda to Section 111 of the ASME Code.
Thus, the J. M. Farley Unit 2 nuclear power plant was constructed to an ASME Code earlier than the St.mmer 1972 Addenda of the 1971 Edition.
Tables 2 through 13 document the unirradiated data (Charpy and drop weight test results, reference temperature, upper shelf energy), heat treatment, heat numbers, weld flux lot number, and chemical composition for all beltline region and surveillance materials. These values were developed using the current reactor pressure vessel material test requirements and acceptance standards at the time of fabrication.
L L
ATTACHMENT 2 TABLE 2 Page 3 Joseph M. Farley Unit-2 Materials Certification Information The following information was taken from the Materials Certification Report prepared by
~
Combustion En91neering, Inc. on September 23, 1969.
Component:
Intermediate Shell Plate, B7203-1 Heat No.:
C6309-2 MILL-Chemical Analysis
==
C Mn P
Si Ni Mo Cu Al Co 0.20 1.31 0.010 0 013 0.19 0.60-0.55 0.14 0.020 0.025 Charoy Impact and Fracture Tests Temp. 'F Ft-Lbs
% Shear Mils Lat. Exp.
- 40 11 0
9
- 40 23 5
18
- 40 21 5
14
+ 10 70 30 51
+ 10 76 35 53
+ 10 66 30 46
+ 40 81 40 56
+ 40 77 40 53
+ 40 89 50 61
+110 120 90 77
+110 128 90 82
+110 120 90 78
+160 142 100 86
+160 143 100 84
+160 136 100 83 Temp. 'F Drop Weights NDT i
- 40 1-F
- 30 2-NF
- 20 1-NF
-40 F_
0 1-NF Heat Treatment 1
1600 F 25 F, 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. Water quenched.
1225 F 25 F, 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
1150 F i 25 F, 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br />.
Furnace cooled to 600 F.
ATTACHMENT 2 TABLE 3 Page 4 Joseph M. Farley Unit 2 Eaterials Certification Information The following information was taken from the Materials Certification Report prepared by Combustion Engineering, Inc. on May 20, 1970.
Canonent:
Inteimediate Shell Plate, B7212-1 Heat No.:
C7466-1 Mill Chemical Analvd C
Mn P
S Si Ni Mo Cu Al Co 0.21 1.30 0.018 0.016 0.24 0.60 0.49 0.20 0.040 0.027 Charov Imoact and Fracture Tests Temp. 'F Ft-Lbs
% Shear Mils Lat. Exp.
- 40 13 0
9
- 40 14 0
9
- 40 22 5
16
+ 10 65 30 43
+ 10 76 40 53
+ 10 49 25 37
+ 40 70 40 49
+ 40 80 50 55
+ 40 89 60 63
+110 110 80 78
+110 114 80 77
+110 118 80 80
+160 133 100 83
+160 130 100 83
+160 139 100 84 Temp.
F Drop Weights NDT
- 40 1-F
- 30 1-F
- 20 2-NF
-30 F 0
1-NF Heat Treatment 1600 F i 25 F, 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. Water quenched.
1225 F 25 F, 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, 1150 F 25 F, 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br />.
Furnace cooled to 600 F.
1 TABLE 4 AT. TAC.liMENT 2 Page 5 Joseph M. Farley Unit 2 tiit erials Certification Information t
The following information was taken from the Materials Certification Report prepared by Combustion Engineering, Inc. on August 1, 1972.
Component:
Lower Shell Plate, B7210-1 litat No. :
C6888-2 MILL Chemical Analysis C
Mn P
S Si Ni Mo Cu Al Co 0.24 1.28 0.010 0.014 0.20 0.56 0.56 0.13 0.020 0.012 Charov Impact and Fracture Tests Temp.
F Ft-Lbs
% Shear Mils Lat. Exp.
- 40 9
0 7
- 40 6
0 6
+ 10 28 15 16
+ 10 33 15 20
+ 10 31 15 19
+ 40 49 20 32
+ 40 36 15 25
+ 40 62 25 43 i
+100 101 70 72 l
+100 97 70 69
+100 84 60 61
+160 134 100 83
+160 125 99 82
+160 124 95 79 l
Temp.
F Drop Weights NDT
- 40 1-F
- 30 2-NF i
- 20 1-NF
-40 F
- 10 1-NF 0
1-NF Heat Treatment 1600 F i 25 F, 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
Water quenched.
1225 F i 25 F, 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
1150of i 25 F, 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br />.
Furnace cooled to 600 F.
L ATTACHMENT 2 TABLE 5 Page 6 1
Joseph M. Farley Unit 2 Materials Certificatior. Information The following information was taken from the Materials Certification Report prepared.by Combustion Engineering, Inc. on April 30, 1970.
Comooi.ent:
Lower Shell Plate, B7210-2 Heat No.:
C6293-1 MILL Chemical Analysis C
Mn P
S Si Ni Mo Cu Al Co 0.19 1.30 0.015 0.015 0.18 0.57 0.59 0.14 0.026 0.021 Charov Impact and Fracture Tests Temp.
F Ft-Lbs
% Shear Mils Lat. Exp.
- 40 13 0
- 1. 0
- 40 11 0
9
+ 10 67 30 45
+ 10 78 35 54
_ + 10 69 30 47
+ 40 85 35 57
+ 40 100 45 66
+ 40 95 45 63
+100 142 99 78
+100 127 90 76
+100 119 80 74
~
+160 141 100 82
+160 146 100 86
+160 147 100 38 Temp.
F Drop Weights NDT
- 30 1-F
- 20 2-NF
- 10 1-NF
-30 F 0 NF Heat Treatment
-1600 F i 25 F, 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. Water quenched.
1225 F 1 25 F, 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
1150 F i 25 F, 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br />.
Furnace cooled to 600of.
l l
ATTACHMENT 2 Page 7 TABLE 6 Joseph H. Farley Unit 2 Materials Certification Information The following information was taken from the Materials Certification Report prepared by Combustion Engineering, Inc. on February 13, 1973.
[omponent:-
Intermediate to Lower Shell Circle Seam,11-923 Weld Wire:
Type B4, Heat No, SP5622 Flux:
Lindo 0091, Lot No. 1122 MILL Chemical Analysis C
Mn P
S Si Mo Cu V
0.17 1.29 0.016 0.008 0.19 0.57 0.13 0.009 Charov Imoact-Tests l
Temp. *F Ft-lbs Mils Lat. Exp.
+ 10 100 62
+ 10 99 60 l
+ 10 108 63
+ 20 100 64
+ 20 105 66 l
+ 20 98 63 Temp.
F Drop Weights NDT
- 40 1-F q
- 30 2-NF
-40 F
- 20 1-NF L
Beat Treatment 1150 F 25 F, 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br />.
Furnace cooled to 600 F.
ATTACHMENT 2 TABLE 7 Page 8 Joseph M. Farley Unit 2 Materials Certification Information The following information was taken from the Materials Certification Report prepared by Combustion Engineering, Inc. on September 15, 1971.
Component:
Intermediate Shell long. Seams, 19-923A Reld Wire:
Type E8018C3, Heat No. H00A tiin 8066 l
MILL Chemical Analysis C
Mn P
S Si Mo Cu Ni V
0.09 1.00 0.009 0.010 0.38 0.25 0.02 0.96 0.010
[haroy Impact Testi Temp. *F
- c-Lbs
+ 10 130
+ 10 135 l
+ 10 128 l-Heat' Treatment 1150 F
- 25 F, 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br />.
Furnace cooled to 600 F.
l
ATTACHMENT 2 TABLE 8 P890 9 Joseph H. farley Unit 2 tiaterials Certification In[ornLips The following information was taken from the Haterials Certification Report prepared by Combustson Engineering, Inc. on April 26, 1973.
Cnmngneatl Intermediate Shell Long. Seam, 19-9238 Rel.d_WAni Type E801803 Heat No. BOLA f_h n Not Available MILL Chemical Anah111 C
Hn P
S Si Mo Cu Ni V
0.081 1.02 0.0!0 0.016 0.41 0.24
,0.02 0.93 0.005 Charny impact Tests Temp. 'T Ft-lbs
% Shear Hils Lat. Exp.
0 82 50 58 0
101 60 70 0
108 70 75
+ 10 106 70 66
+ 10 108 70 72
'1
+ 10 105 70 71 l
(nTeinp. of Drop Weights NDT
- 60 1-F
- 50 2-NF
- 40 1-NF
-60 f
}
- 20 1-NF Heat Treatment 1150 f i 25 f, 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br />, furnace cooled to 600 f.
ATTAtllMENT 2 TABLE 9 Page 10 4
4 Joseph H. Farley Unit 2 tinie.tials Certification Information The following information was taken from the Haterials Certification Report prepared by Combustion Enginearing. Inc. on August 17, 1972.
Component:
Lower Shell Long. Seam. 20-923A&B Meld Wirei Type B4, Heat No. 83640
[hn Linde 0091, Lot No. 3490 MILL Chemical Analysis C
Hn P
S Si Ho Cr V
0.16 1.22 0.006 0.011 0.19 0.57 0.05 0.006 Charny Imoact Tests Temp. 'F Ft-lbs Mils Lat. Exp.
- 10 119 85
- 10 118 74
- 10 110 78
+ 10 126 86
+ 10 124 85
+ 10 129 87 Temp. "F Drop Weights NDT
- 70 1-F
- 60 2-NF
- 40 1-NF
-70*F
- 20 1-NF Heat Treatment 1150 F 25'F, 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br />.
Furnace cooled to 600'F.
ATTACilMENT 2 89" II TABLE 10 Joseph M. f arley Unit 2 Surveillance Capiule Materials Dt(A The following information was taken from WCAP-8956,
- Alabama Power Company Joseph M. Farley Nuclear Plant Unit No. 2 Reactor Vessel Radiation Surveillance Program,'
August 1977.
(smoonenil Surveillance Material SA533 Grade B Class 1 plate used for the intermediate shell plate, B7212-1.
Chemical Analysis The following dr.ta was obtained from Combustion Engineering Analysis, unless otherwise noted.
C Mn P
S Si Hi Mo Cu Al Co 0.21 1.30 0.018 0.016 0.24 0.60 0.49 0.20 0.040 0.027 N*
Cr*
V*
Sn*
0.006 0.15,
0.003 0.011
- Westinghouse Analysis.
Charov Im0jtti and fracture Testi Intermediate Shell Plate, B7212-1 (Longitudinal)
Temp. of Ft-Lbs
% Shear Mils Lat. Exp.
- 60 12 8
9
- 50 15 12 11
- 25 19 22 16
- 25 34 18 26
- 10 47 25 35 0
50 30 40
+ is 63 40 49
+ 50 72 60 60
+ 60 67 48 53
+ 75 103 63
'i
+100 110.0 85 80
+125 123 100 84
+150 132 100 91
+210 134 100 87
+210 130 100 90 x
+210 132 100 89
ATTACHMENT 2 TABLE 10 continued Page 12
,4 8
[Aarov Imonst and Fracture Tests Intermediate Shell Plate, B7212-1 (Transverse)
Temp. *F Ft-Lbs
% Shear Hils Lat. Exp.
- 50 18.5 12 11
- 50 15.5 le 11
- 50 19 12 11 0
35 27 27 0
34.5 25 27 0
30 25 27
+ 30 43 32 35
+ 30 48 35 36
+ 30 52 43-39
+100 76.5 73 55
+100 74 73 56
+100 70 69 54
+150 95 100 67
+150 98 100 68
+150 106 100 76
+210 89 100 68
+210 94 100 70
- 210 88 100 69fuLt Treatmen_t i
Intermediate Shell Plate, B7212-1 1550"F/1650*F, 4 hours. Water quenched. 1225 F
- 25 F, 4 hours.
Air cooled. Il50*F
- 25 F, 18 hours.- Furnace cooled to 600 F.
Drooweiaht Tests Intermediate Shell Plate, 87212-1 The nil-ductility transition temperature (NDTT) was determined for plate B7212 ' to be -30*F by drop weight tests-(ASTM E-208) performed by Combustion Engineering. l l
ATTACHMENT 2 TABLE 11 Page 13 Joseph M. Farley Unit 2 Surveillance Capsule Materials Data Component: Surveillance Weld Material made from sections of intermediate shell plate, B7212-1 and adjoining intermediate shell plate, B7203-1. using weld wire representative of that used in the original fabrication. Chemical Analysis The following data was obtained from Combustion Engineering Analysis. C Hn P S Si Ni Ho Cu Al Co <0.086 0.95 0.004 0.014 0.34 0.90 0.23 0.03 0.003 0.010 N Cr V Sn 0.007 <0.01 0.006 0.002 Charfd Imoact and Fracture Test.1 h Temp. 'F Ft-Lbs % Shear Mils Lat. Exp. -100 8 12 7 - 50 16 30 15 - JE 42 40 35 - 10 50 43 44 + 10 109 80 73 + 25 68 72 60 + 40 124 85 80 + 50 133 94 89 + 75 144 100 92 +100 132 100 91 +150 131 100 92 +150 150 100 91 +175 154.5 100 90 +210 154 100 91 +210 153 100 93 +210 137 100 92 p Heat Tregimaat 1150 F t 25 F, 13 hours. Furnace cooled. l Drooweiaht Tests The nil-ductility transition temperature (NDTT) was determined for the surveillance weld metal to be -70 F by drop weight tests (ASTM E-208) performed by . Combustion Engineering.
ATTACHMENT 2 TABLE 12 Page 14 Joseph H. farley Unit 2 Surveillance Caosule Materials Data (9Enantali Surveillance Weld-Heat-Affected Zone (HAZ) Haterial were obtained from the weld-heat-affected zone of intermediate shell plate, B7212-1. Charov lmoact and Fracture Telli Temp 'F Ft-Lbs % Shear Hils Lat. Exp. -200 11 8 5 -150 159 100 95 -100 98 45 57 -100 40 42 33 - 50 98 60 60 - 50 71 34 46 1 + 20 44 82 73 25 113.5 83 79 50 136 100 91 e + 75 123 100 87 +100 191 100 81 +150 159 100 95 +210 151 100 91 +210 86 93 73 +210 189 100 87 DroDweicht Tests The nil-ductility transition temperature (NDTT) was determined for the surveillance weld-heat-affected zone material to be -80 F by drop weight tests (ASTM E-208) performed by Combustion Engineering.
1 ATTACHMENT 2 Page 15 ,, o. The initial RT., values for the beltline region plates are calculated based on the information provided in Nuclear Safety Task Sheet 17276 (transverse data). The initial RT., values for the beltline region welds are calculated based on the information given in the Combustion Engineering Haterial Certification Reports using the methodology of the Branch Technical ?osition MTEB 5-2. The upper shelf energy (USE) values for the beltline region plates were calculated using the data provided in Nuclear Safety Task Sheet 17267. Date was not available to calculate upper shelf energy values for the welds. The beltline region welds only had fabrication Charpy tests performed at 10'F, so initial USE values could not be determined. However, the surveillance weld, which is representative of the intermediate shell longitudinal weld seam, 19-9238, had an unirradiated USE of 148 ft-lb. Evaluation acording to Regulatory Guide 1.99, Revision 2, of the intermediate shell longitudtc1 weld scam 19-921B assuming the initial USE of 148 ft-lb results in the 32 EFPY USE prediction of 121 ft-lb. The initial RT,,o and USE values are presented in Table 13. l-TABLE 13 INiil AL RTmb,R J. M. FARLEY UNIT 2 AND UPPER SHELF ENERGY VALUES F INITIAL RT,,a USE MATERIAL DESCRIPT1Q!f f IfblbM Inter Shell, B7203-1 15 99.5 Inter Shell, B7212-1 -10 99.7 Lower Shell, B7210-1 18 103 Lower Shell, B7210-2 0 99 Long Weld Seam, 19-923A -56 (a) (b) { Long Weld Seam, 19-9238 -60 (b) long Weld Seams, -70 (b) 20-923A&B Inter / Lower Circ Weld -40 (b) Surveillance Plate (c) 100 Surveillance Weld (c) 148 Notes: (a) Estimated per 10 CFR 50.61. (b) Upper Shelf energy values are not available because unirradiated Charpy V-notch impact tests were not performed. (c) Initial RT., not determined for surveillance material.
ATTACHMENT 2 Page 16 l 4. liRC_Agunt Describe how the embrittlement affects of operating at an irradiation temperature (cold leg or recirculation suction temperature) below 525 f were considered. Describe consideration given to determining the effect of lower irradiation temperature on the reference temperature and on the Charpy upper ;helf energy. SNC_Remonig The requirements of Technical Specification 3/4.1.1.4, Tinimum Temperature for Criticality," prohibits plant criticality below 541*f. A review of plant operating records indicates that RCS average temperature has never dropped below 525'f while the core was critical. Therefore, the concern with increased embrittlement effects due to operation below 525'r is not applicable to Farley Nuclear Plant, Unit 2. 5. NRC Request Describe how surveillance results on the predicted amount of imbrittlement were considered. $NC Response i The pressure-temperature limits developed for J. H. Farley Unit 2 as part of the Surveillance Capsulo X lesting Program were based upon the methodology of Regulatory Guide 1.99, Revision 2. These curves were based on the limiting material in the beltline region, intermediate shell plate B7212-1 using surveillance capsule data. The RT,,, values calculated for 32 EfPY are within the screening criteria of 10 CfR 50.61 as indicated in Table 14. These values were calculated using the data presented in Tables 2 - 9 and Table 13. The nickel content for the lower shell longitudinal welds and the intermediate to lower shell circumferential weld were not available from the material cer ifications, therefore the nickel value of 0.20 was used as presented in the FSAR for i se welds. TABLE 14 J. M. FARLEY UNIT 2 R1,,, VALUES FOR 32 EfPY RT,,, MAi RIAL DESCRIPTION 'F Inter Shell, B7203-1 184 Inter Shell, B7212-1 224 (217) Lower Shell, 87210-1 R3 Lower Shell, 87210-2 177 Long Weld, 19-923A 39 Long Weld, 19-923B 25 Long Weld, 20-923A&B 38 Inter / Lower Circ Weld 118 () Indicates RT,,, value based on surveillance crpsule data.
ATTACHMENT 2 Page 17 6. NRC.Reguait if a measured increase in reference temperature exceeds the mean-plus-two standard deviations predicted by Regulatory Guide 1.99, Revision 2, or if a measured decrease in Charpy upper shelf energy exceeds the value predicted using the guidance in Paragraph C.).2 in Regulatory Guide 1.99, Revision 2, report the information and describe the effect of the surveillance results on the adjusted reference temperature and Charpy upper shelf energy for each beltline material as predicted for December 16, 1991, and for the end of its current license. SNC Response The measured increase in reference temperature does not exceed the mean-plus-two standard deviations predicted by the Regulatory Guide 1.99, Revision 2, for any of the surveillance capsule materials as indicated in Table 15. TABLE 15 J. H. FARLEY VNIT 2 MEASURED INCREASE IN REFERENCE TEMPERATURE VERSUS REGULATORY GUIDE 1.99, REVISION 2, PREDICTIONS oRim, (*f) SURVElllANCE R.G. 1.99, REV 2 BalLRIAL DESCRIPilQB CAPSULE HEASURED PREDICTION + 2g_ Plate, B7212-1 U 103 156 (longitudinal) W 165 198 X 180 228 Plate, B7212-1 U 133 156 (Transverse) W 165 198 X 150 228 Weld Metal U 10 90 W 10 101 X 10 109 w e .,,e,,- ,m-, .--re,. ~,,,,e.y._._ .,,,_me. .,,,my -,- e y-,- -,m., m,,-
r ATTACHMENT 2 Page 18 The measured decrease in Charpy rpper shelf energy does not e., reed the value predicted using methodology specifies in Regulatory Guide 1.99, Revisicn 2, for the surveillance capsule materials tested in the Capsule W and X analyses as indicated in Table 16. The measured decrease in Charpy USE for plate 07212-1 was greater than the Regulatory Guide 1.99, Revision 2, )rediction in the Capsule U analysis. This decrease was less than 3 ft-lb more than tie predictions and is most likely due to the uncertainty in the analysis methods. The predicted 32 EFPY upper shelf energy values are predicted to be above the 50 ft-lb criteria as shown in Table 1. TABLE 16 J. H. FARLEY UNIT 2 MEASURED DECREASE IN CHARPY UPPER SHELF ENERGY VERSUS REGULATORY GUIDE 1.99, REVISION 2, PREDICTIONS DECREASE IN UPPER SHELF ENERGY (El_ SURVEILLANCE R.G. 1.99, REV 2 MATERIAL DESCRIPTION CAPSULE ___ MEASURED PREDICTION Plate, B7212-1 U 27.7 25 (Longitudinal) W 21.5 32 X 27.7 36 Plate, 87212-1 0 27.0 25 (Transverse) W 20.0 32 X 27.0 36 Weld Metal V 8.3 16 W 0 21 X 0 24 i _}}