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{{#Wiki_filter:,  ,  Southern Nuclear Operating Company Post Office Box 1295 Birmingham, Alabama 35201 Telephone (205) 868-5066 L k J. o. Woodard Executive Vice President Southern Nudear Operating Company the southem elecinc System June 21, 1994 10 CFR 50.54(f)
{{#Wiki_filter:Southern Nuclear Operating Company Post Office Box 1295 Birmingham, Alabama 35201 Telephone (205) 868-5066 k
Docket Nos. 50-348 50-364 TAC Nos.         83461 83462 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555 Joseph M. Farley Nuclear Plant Responses to Open Issues Regarding Generic Letter 92-01, Revision 1 Reactor Vessel Structural Integrity Gentlemen:
L Southern Nudear Operating Company J. o. Woodard Executive Vice President the southem elecinc System June 21, 1994 10 CFR 50.54(f)
Docket Nos. 50-348 50-364 TAC Nos.
83461 83462 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555 Joseph M. Farley Nuclear Plant Responses to Open Issues Regarding Generic Letter 92-01, Revision 1 Reactor Vessel Structural Integrity Gentlemen:
On March 6,1992, the NRC issued Generic Letter (GL) 92-01, Revision 1, " Reactor Vessel Structural Integrity." The purpose of the GL was to obtain information needed to assess compliance with requirements and commitments regarding reactor vessel integrity due to events associated with the Yankee Nuclear Power Station. Southern Nuclear Operating Company (SNC) provided a response to GL 92-01, Revision 1, by {{letter dated|date=July 1, 1992|text=letter dated July 1,1992}}. On October 7,1993, the NRC issued SNC a request for additional information (RAI) in order to complete the review of the SNC response to GL 92-01, Revision 1. SNC provided a response to the RAI by {{letter dated|date=November 23, 1993|text=letter dated November 23,1993}}.
On March 6,1992, the NRC issued Generic Letter (GL) 92-01, Revision 1, " Reactor Vessel Structural Integrity." The purpose of the GL was to obtain information needed to assess compliance with requirements and commitments regarding reactor vessel integrity due to events associated with the Yankee Nuclear Power Station. Southern Nuclear Operating Company (SNC) provided a response to GL 92-01, Revision 1, by {{letter dated|date=July 1, 1992|text=letter dated July 1,1992}}. On October 7,1993, the NRC issued SNC a request for additional information (RAI) in order to complete the review of the SNC response to GL 92-01, Revision 1. SNC provided a response to the RAI by {{letter dated|date=November 23, 1993|text=letter dated November 23,1993}}.
The NRC notified SNC, by {{letter dated|date=May 20, 1994|text=letter dated May 20,1994}}, of two open issues regarding GL 92-01 and requested verification of the NRC summary data file information for Farley Nuclear Plant.
The NRC notified SNC, by {{letter dated|date=May 20, 1994|text=letter dated May 20,1994}}, of two open issues regarding GL 92-01 and requested verification of the NRC summary data file information for Farley Nuclear Plant. provides SNC's responses to the open items identified by the NRC associated with GL 92-01, Revision 1. Attachment 2 provides the data used to determine a statistical value for the unirradiated upper shelf energy for type B4 weld filler material. provides a marked-up copy of the summary data file for the Farley Unit 1 and Unit 2 vessels to reflect changes to the information with references supporting each change.
Attachment 1 provides SNC's responses to the open items identified by the NRC associated with GL 92-01, Revision 1. Attachment 2 provides the data used to determine a statistical value for the unirradiated upper shelf energy for type B4 weld filler material.
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Attachment 3 provides a marked-up copy of the summary data file for the Farley Unit 1 and Unit 2 vessels to reflect changes to the information with references supporting each change.
PDR ADOCK 05000348 p
k     9406280120 940621 PDR                                                                                                Af I
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P ADOCK 05000348                                                                           p   l PDR
PDR


U. S. Nuclear Regulatory Commission                                                     Page 2 As stated in previous submittals regarding GL 92-01, Revision 1, SNC continues to comply with the requirements of 10 CFR 50.61 and 10 CFR 50, Appendix G.
U. S. Nuclear Regulatory Commission Page 2 As stated in previous submittals regarding GL 92-01, Revision 1, SNC continues to comply with the requirements of 10 CFR 50.61 and 10 CFR 50, Appendix G.
If there are any questions, please advise.
If there are any questions, please advise.
Respectfully submitted, SOUTHERN NUCLEAR OPERATING COMPANY
Respectfully submitted, SOUTHERN NUCLEAR OPERATING COMPANY
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                                                        'oodard SWORN TO AND SUBSCRIBED BEFORE ME THIS dek)AY OF b                     ,1994 ht& LA L0ne-Notary'Public           (/
'oodard SWORN TO AND SUBSCRIBED BEFORE ME THIS dek)AY OF b
My Commission Expires: 7/dua. /, /997 DRC; cit 920lRAl2. doc Attachments cc:   Mr. S. D. Ebneter Mr. B. L Siegel Mr. T. M. Ross i
,1994 ht& LA L0ne-Notary'Public
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My Commission Expires: 7/dua. /, /997 DRC; cit 920lRAl2. doc Attachments cc:
Mr. S. D. Ebneter Mr. B. L Siegel Mr. T. M. Ross i
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ATTACIIMENT 1 SNC RESPONSES TO GL 92-01, REVISION 1, OPEN ISSUES   ,
ATTACIIMENT 1 SNC RESPONSES TO GL 92-01, REVISION 1, OPEN ISSUES
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SNC Response to Open Issue (1)
SNC Response to Open Issue (1)
SNC stated in our response to NRC Request for Additional Information (RAI) regarding GL 92-01, Revision 1, dated November 23,1993, that the nickel content for beltline welds was taken from earlier WOG programs. Specifically, the 0.20% value was determined from a 1982 program performed to calculate operating and near term operating limit vessel RTndt values and the 0.21%
SNC stated in our response to NRC Request for Additional Information (RAI) regarding GL 92-01, Revision 1, dated November 23,1993, that the nickel content for beltline welds was taken from earlier WOG programs. Specifically, the 0.20% value was determined from a 1982 program performed to calculate operating and near term operating limit vessel RTndt values and the 0.21%
value was detennined from a program to develop a materials database aimed at filling in gaps where data was not available. As stated in our response to the RAI, the nickel content of the
value was detennined from a program to develop a materials database aimed at filling in gaps where data was not available. As stated in our response to the RAI, the nickel content of the beltline welds is not available from the material certifications and the values listed are based on engineeringjudgement. However, a footnote to Table 2 and Table 8 of the RAI response identified these nickel values as mean values taken from the WOG materials database.
* beltline welds is not available from the material certifications and the values listed are based on engineeringjudgement. However, a footnote to Table 2 and Table 8 of the RAI response                                                         ;
identified these nickel values as mean values taken from the WOG materials database.
Subsequent conversations with Westinghouse have verified that the WOG database value for nickel content of the beltline welds is based solely on engineering judgement described below and not actual material certifications.
Subsequent conversations with Westinghouse have verified that the WOG database value for nickel content of the beltline welds is based solely on engineering judgement described below and not actual material certifications.
The rationale used to determine the WOG database nickel value for weld wire heats 33 A277, 6329637, 90099, 5P5622, and 83640 in the vessel beltline welds considered that both the Farley Unit I and Unit 2 vessels were fabricated between 1971 and 1973 by Combustion Engineering (CE)in Chattanooga, Tennessee. The vessels were fabricated using the automatic submerged arc                                                 ;
The rationale used to determine the WOG database nickel value for weld wire heats 33 A277, 6329637, 90099, 5P5622, and 83640 in the vessel beltline welds considered that both the Farley Unit I and Unit 2 vessels were fabricated between 1971 and 1973 by Combustion Engineering (CE)in Chattanooga, Tennessee. The vessels were fabricated using the automatic submerged arc welding (SAW) process and type B4 weld filler material supplied by the Reid Avery Company.
welding (SAW) process and type B4 weld filler material supplied by the Reid Avery Company.
Nickel was not added as an alloying element to the B4 weld filler material manufactured by the Reid Avery Company; therefore, CE did not require the Reid Avery Company to perfonn or submit a chemical analysis value for tickel in type B4 weld filler wire. A chemical analysis of the Farley Unit I surveillance weld ma;erial, heat number 33 A277, indicated a nickel content of 0.19 weight percent, consistent with the estimated value of 0.20 weight percent reported in our response to the RAI.
Nickel was not added as an alloying element to the B4 weld filler material manufactured by the Reid Avery Company; therefore, CE did not require the Reid Avery Company to perfonn or submit a chemical analysis value for tickel in type B4 weld filler wire. A chemical analysis of the Farley Unit I surveillance weld ma;erial, heat number 33 A277, indicated a nickel content of 0.19 weight percent, consistent with the estimated value of 0.20 weight percent reported in our response to the RAI.
SNC is a charter member of the Combustion Engineering Reactor Vessel Group (CERVG) and joined the group with the expectation of obtaining additional information to augment existing vessel chemistry and toughness data for SNC vessels. Based on reviews performed to date, SNC does not expect the nickel content of weld wire heats 33 A277,6329637, 90099, SP5622, and 83640 to exceed 0.20 weight percent. The CERVG effort is currently targeted for completion by December 1994.
SNC is a charter member of the Combustion Engineering Reactor Vessel Group (CERVG) and joined the group with the expectation of obtaining additional information to augment existing vessel chemistry and toughness data for SNC vessels. Based on reviews performed to date, SNC does not expect the nickel content of weld wire heats 33 A277,6329637, 90099, SP5622, and 83640 to exceed 0.20 weight percent. The CERVG effort is currently targeted for completion by December 1994.
l Attachment 1                                                                                                                          Page1of2
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Open Issue (2)
Open Issue (2)
The second open issue for Farley involved a NRC staff concern that surveillance data was used to determine the unirradiated upper shelf energy (USE) values for weld wire heat numbers 6? 29637,         i 90099, SP5622,83640, and HODA. The staffs letter stated that since the surveillance data are           !
The second open issue for Farley involved a NRC staff concern that surveillance data was used to determine the unirradiated upper shelf energy (USE) values for weld wire heat numbers 6? 29637, i
from a different heat, a statistical analysis addressing heat variability may be appropriate and that when the unirradiated USE for a particular heat of material has not been determined, the USE can be set equal to the lower tolerance limit calculated for the group of similar materials. The staff stated that the unirradiated USE should be determined such that there exists 95% confidence that       ,
90099, SP5622,83640, and HODA. The staffs letter stated that since the surveillance data are from a different heat, a statistical analysis addressing heat variability may be appropriate and that when the unirradiated USE for a particular heat of material has not been determined, the USE can be set equal to the lower tolerance limit calculated for the group of similar materials. The staff stated that the unirradiated USE should be determined such that there exists 95% confidence that at least 95% of the population is greater than the !ower tolerance limit and if the lower tolerance limit results in a projected USE at EOL orless than 50 fl-lb, then SNC must demonstrate, in accordance with Appendix G,10 CFR Part 50, that equivalent lower values of USE will provide margins of safety against fracture equivalent to those required by Appendix G of Section III of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code.
at least 95% of the population is greater than the !ower tolerance limit and if the lower tolerance limit results in a projected USE at EOL orless than 50 fl-lb, then SNC must demonstrate, in accordance with Appendix G,10 CFR Part 50, that equivalent lower values of USE will provide margins of safety against fracture equivalent to those required by Appendix G of Section III of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code.
SNC Response to Open Issue (2)
SNC Response to Open Issue (2)
A statistical analysis was performed by Westinghouse, in accordance with the NRC staffs guidance, to determine a generic unirradiated USE applicable to heats 6329637,90099, SP5622, and 83640. Attachment 2 provides 61 values of Charpy V-notch energy at 100% shear where data for full Charpy curves are available for type B4 weld wire. The lower tolerance limit determined from the data in Attachment 2 resulted in a generic unirradiated USE value of 82.5 ft-Ib. For weld wire heat number 90099, Farley will utilize the generic unirradi ated USE value of 82.5 fl-lb determined by the statistical analysis. Use of the generic unirradiated USE value for heat 6329637 would result in a projected EOL USE less than 50 fl-lb. However, for heat 6329637 the Farley specific Charpy value at +10 F is known and can be used as a conservative estimate for the unirradiated USE since the unirradiated USE is known to be higher than the
A statistical analysis was performed by Westinghouse, in accordance with the NRC staffs guidance, to determine a generic unirradiated USE applicable to heats 6329637,90099, SP5622, and 83640. Attachment 2 provides 61 values of Charpy V-notch energy at 100% shear where data for full Charpy curves are available for type B4 weld wire. The lower tolerance limit determined from the data in Attachment 2 resulted in a generic unirradiated USE value of 82.5 ft-Ib. For weld wire heat number 90099, Farley will utilize the generic unirradi ated USE value of 82.5 fl-lb determined by the statistical analysis. Use of the generic unirradiated USE value for heat 6329637 would result in a projected EOL USE less than 50 fl-lb. However, for heat 6329637 the Farley specific Charpy value at +10 F is known and can be used as a conservative estimate for the unirradiated USE since the unirradiated USE is known to be higher than the
      +10 F Charpy value. The known Farley specific +10 F Charpy values are also being used for heats SP5622 and 83640. Since the statistical analysis is only applicable to type B4 weld wire, the
+10 F Charpy value. The known Farley specific +10 F Charpy values are also being used for heats SP5622 and 83640. Since the statistical analysis is only applicable to type B4 weld wire, the
      +10 F Charpy value is also being used for the SMAW filler metal, heat HODA.
+10 F Charpy value is also being used for the SMAW filler metal, heat HODA.
Based on the above information, the projected EOL USE for the Farley Unit I and 2 belthne welds exceeds 50 fl-lb, as shown in Attachment 3; therefore, Farley Units 1 and 2 will continue to comply with the requirements of Appendix G of 10 CFR Part 50.
Based on the above information, the projected EOL USE for the Farley Unit I and 2 belthne welds exceeds 50 fl-lb, as shown in Attachment 3; therefore, Farley Units 1 and 2 will continue to comply with the requirements of Appendix G of 10 CFR Part 50.
Attachment 1                                                                              Page 2 of 2 l
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ATTACIIMENT 2 Charpy V-Notch Energy (fl-lb) Data at 100% Shear for Determination of Generic Unirradiated USE for Welds with Type B4 Weld Filler Wire 138                                                                     159           151           152           159       174 156                                                                     150           151         162.5           144       154 139                                                                     140           149           153             135       140 141                                                                     144           154           158             135       144 145                                                                     143           143           144           122       122 l 127                                                                     125           125           125             125       172 154                                                                     162           158           87           90.5     84.5 109                                                                     116           108           126             121       99 118                                                                     129           123           113             119       86 88                                                                   82             90             94             99       100 102 Number of Points = 61 Sample Mean = 130.467 Sample Deviation = 24.4896 Lower Tolerance Bound = 82.5
ATTACIIMENT 2 Charpy V-Notch Energy (fl-lb) Data at 100% Shear for Determination of Generic Unirradiated USE for Welds with Type B4 Weld Filler Wire 138 159 151 152 159 174 156 150 151 162.5 144 154 139 140 149 153 135 140 141 144 154 158 135 144 145 143 143 144 122 122 l
127 125 125 125 125 172 154 162 158 87 90.5 84.5 109 116 108 126 121 99 118 129 123 113 119 86 88 82 90 94 99 100 102 Number of Points = 61 Sample Mean = 130.467 Sample Deviation = 24.4896 Lower Tolerance Bound = 82.5


ATTACHhiENT 3 h1ARKED-UP SUhihiARY DATA FILE FOR FARLEY UNITS 1 & 2 VESSELS The NRC stafTrequested that SNC verify the summary data file information contained in Enclosures 1,2, and 3 of the NRC's hiay 20,1993, letter. Accordingly, SNC provides the following marked-up copics of the summary data file with reference information to support the changes.
ATTACHhiENT 3 h1ARKED-UP SUhihiARY DATA FILE FOR FARLEY UNITS 1 & 2 VESSELS The NRC stafTrequested that SNC verify the summary data file information contained in Enclosures 1,2, and 3 of the NRC's hiay 20,1993, letter. Accordingly, SNC provides the following marked-up copics of the summary data file with reference information to support the changes.


Sumary File for Pressurized Thermal Shock P la.itt       Betttlne         Meet No.                                               10 Neut.         IRT. Method of   Chemistry   Method of   %Cu     %Ni kame           Ident.           Ident.                                                 Fluence et           Detersin. Factor     Determin.
Sumary File for Pressurized Thermal Shock P la.itt Betttlne Meet No.
EOL/EFPY             1 R T,                 CF F2rtey 1       Int. Shett       C6294 1                                                 3.75E19     0*F     MTEB 5 2   91           febte       0.13   0.60 86903 2 EOL:           Int. Shett       C6308 2                                                 3.75E19       10'F   MTEB 5 2   82.2         Table       0.12   0.56 6/25/2017       86903 3 Lower             C6940 1                                                 3.75E19       15'F   MTER 5 2   88.831       Calculated   0.14   0.55   j Shett B6919 1                                                                                                                                             l j
10 Neut.
Lower             C6897 2                                                 3. 75E19     5'F     MTEs 5-2   98.2         Table       0.14   0.56 shelt 86919 2 Int. Shett       33A277                                                 1.24E19       56'F   Generic   78.689       Calculated   0.25
IRT.
* 0.21 Axial Welds Cire. Weld         6329637                                                 3.75E19       56'F   Generic   -443-Table       -4430. W 0.20 '
Method of Chemistry Method of
tower             90099                                                   1.24E19     -56*F   Generic   92           Table       0.17   0.20 '
%Cu
Sheit Axlet Welds REFEREhCES FOR FARLEY 1:
%Ni kame Ident.
Ident.
Fluence et Detersin.
Factor Determin.
EOL/EFPY 1 R T, CF F2rtey 1 Int. Shett C6294 1 3.75E19 0*F MTEB 5 2 91 febte 0.13 0.60 86903 2 EOL:
Int. Shett C6308 2 3.75E19 10'F MTEB 5 2 82.2 Table 0.12 0.56 6/25/2017 86903 3 Lower C6940 1 3.75E19 15'F MTER 5 2 88.831 Calculated 0.14 0.55 j
Shett l
B6919 1 j
Lower C6897 2
: 3. 75E19 5'F MTEs 5-2 98.2 Table 0.14 0.56 shelt 86919 2 Int. Shett 33A277 1.24E19 56'F Generic 78.689 Calculated 0.25 0.21 Axial Welds
-4430. W Cire. Weld 6329637 3.75E19 56'F Generic
-443-Table 0.20 tower 90099 1.24E19
-56*F Generic 92 Table 0.17 0.20 Sheit Axlet Welds REFEREhCES FOR FARLEY 1:
Fluence, IRT and chemistry values f rom hovember 23, 1993, letter from D. Morey (suoC) to UskRC Docet Control Desk, select: Responses to Regaeste for Additionet Information Regarding GL 92-01.
Fluence, IRT and chemistry values f rom hovember 23, 1993, letter from D. Morey (suoC) to UskRC Docet Control Desk, select: Responses to Regaeste for Additionet Information Regarding GL 92-01.
j (0      Tbc          vnnr ked- up                                               circ.. weld % Cu ei o.zz5 a 'd cW misby &clor i
j vnnr ked-up circ.. weld % Cu ei o.zz5 a 'd cW misby &clor (0
ed   114.5 ( n om Table) were de valoc s submitied                                                             b y h e_                     l l
Tbc i
tlovem be r 23, 1993 \ceer refe r en ce d nbov e. ,
ed 114.5 ( n om Table) were de valoc s submitied b h e_
l y
l tlovem be r 23, 1993 \\ceer refe r en ce d nbov e.,
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is an esW M e d value- beed on eg oe er ing
is an esW M e d value-beed on eg oe er ing
                                                            ' Chemical composition -from mean ;;lue of L'OC dat.                                                             #ddition:1 "forn tier rem. ired. jg                                                               ,,,g, g% g, 4e q , ,, , , , ;               pa$,,,a is A c>riib e ) in A h ch ,er+                                                           1.
' Chemical composition -from mean ;;lue of L'OC dat.
#ddition:1 "forn tier rem. ired.
jg,,,g, g% g, 4e q,,,,,, ;
pa$,,,a is A c>riib e ) in A h ch,er+
1.


Summary File for Pressurized Thermal Shock 1
Summary File for Pressurized Thermal Shock l
      .                                                                                                                                    l Pitnt         BettlIne     Heat No. 1D Neut.         I R T. Method of     Chemistry Method of   %Cu         Emi Name           Ident.       Ident.       Fluence at                 Detersin.     Factor     Determin.                         {
Pitnt BettlIne Heat No.
E0L/EFPT                   IR T.                     CF                               '
1D Neut.
Farley 2       Int. shett   C6309 2     3.8E19       15*7         Plant         100       Table       0.14         0.60 87203 1                                                 Specific EOL:           Int. Shett   C7466-1       3.8E19         10'F         Plant         145.0     Calculated 0.20         0.60 3/31/2021     e7212 1                                                 specific Lower         C6888 2       3.8E19       18'F         Plant         89.8       Table       0.13         0.56 shett                                                   Specific 87210-1 Lower         C6293 1       3.8E19       10'F           Plant         98.7       fable       0.14         0.57
I R T.
                        $helt                                                   Specific 87210-2 Cire. Weld   5P5622       3.8E19       -40*F         Plant         76         fable       0.13         0.20
Method of Chemistry Method of
* 11-923                                                 specific Lower         83640         1.23E19     -70'F         Plant         49         Table       0.05         0.20 '
%Cu Emi Name Ident.
Shett                                                 Specific Axlet Welds 20 923 A/S Int, shett   H00A         1.23E19     -56*F         Generic       10.01       Cateutated 0.02         0.%
Ident.
Axial         $ MAW Welds 19 923A Int. Shell   BOLA         1.23E19     -60'F         Plant         10.01       Calculated 0.02         0.93 Axist         SMAW                                     Specific Welds 19 9238 REFERENCES FOR FARLET 2:
Fluence at Detersin.
Fluence, IRT , and chemistry values from Novenber 23, 1993, letter from D. Morey (SNOC) to USNRC Doctament Control Desk, s@ ject: Responses to Requests for Additional Information Regarding 92-01.
Factor Determin.
{
E0L/EFPT IR T.
CF Farley 2 Int. shett C6309 2 3.8E19 15*7 Plant 100 Table 0.14 0.60 87203 1 Specific EOL:
Int. Shett C7466-1 3.8E19 10'F Plant 145.0 Calculated 0.20 0.60 3/31/2021 e7212 1 specific Lower C6888 2 3.8E19 18'F Plant 89.8 Table 0.13 0.56 shett Specific 87210-1 Lower C6293 1 3.8E19 10'F Plant 98.7 fable 0.14 0.57
$helt Specific 87210-2 Cire. Weld 5P5622 3.8E19
-40*F Plant 76 fable 0.13 0.20 11-923 specific Lower 83640 1.23E19
-70'F Plant 49 Table 0.05 0.20 Shett Specific Axlet Welds 20 923 A/S Int, shett H00A 1.23E19
-56*F Generic 10.01 Cateutated 0.02 0.%
Axial
$ MAW Welds 19 923A Int. Shell BOLA 1.23E19
-60'F Plant 10.01 Calculated 0.02 0.93 Axist SMAW Specific Welds 19 9238 REFERENCES FOR FARLET 2:
Fluence, IRT, and chemistry values from Novenber 23, 1993, letter from D. Morey (SNOC) to USNRC Doctament Control Desk, s@ ject: Responses to Requests for Additional Information Regarding 92-01.
l 0
l 0
13 an es wah d vale basr A en engineer-in3
wah d vale basr A en engineer-in3 13 an es
                            , Chemical composition from meanwalue-of-WOG data . Additional
, Chemical composition from meanwalue-of-WOG data. Additional
                  -informa t i c" required-to--confirm v a l u e -
-informa t i c" required-to--confirm v a l u e -
pdcmed.
pdcmed.
3            ~n c bd3 for A eng.ncericg y d3cment is Jc w a cJ i,. Att.stbnen+ 1
~n c bd3 for A 3
d eng.ncericg y 3cment is Jc w a cJ i,. Att.stbnen+ 1


i l
i l
Sumary File for Upper Shelf Energy Plant Name       Settline     Neat No. Material     1/4T USE     1/4T           unirred.     Method of               !
Sumary File for Upper Shelf Energy Plant Name Settline Neat No.
Ident.                 Type         at           Neutron         USE           Deternin.               i EOL/EFPY     Fluence at                   Unfrred.                 l EOL/EFPY                     USE Farley 1         Int. SheLL   C6294 1     A 5338 1     73           2.34E19         99           65%
Material 1/4T USE 1/4T unirred.
B6903 2 EOL:             Int. Shell C6308 2     A 5338-1     65           2.34E19         87           651 6/25/2017         B6903-3 Lower       C6940 1     A 5338 1     -         2.34E19         86           65%
Method of Ident.
shall 86919-1 g N Lower       C6897 2     A 5338 1     62           2.34E19         86           65%
Type at Neutron USE Deternin.
shott 86919 2 Int. Shett 33A2TT     Linde         115         7.73E18         149         surv.
i EOL/EFPY Fluence at Unfrred.
Axlet                   1092, SAW                                               Weld Welds Cire. Weld   6329637     Linde                     2.34E19       #            5;!r. t:   +10*F    (1) 0091, SAW   -44585- '*+)                              !=-           ^'IY 104 '*I    mid       value.
EOL/EFPY USE Farley 1 Int. SheLL C6294 1 A 5338 1 73 2.34E19 99 65%
Lower       90099       Linde       #            7.73E18       #            E;!- t:   Stati,+l cal Shell                   0091, SAW           (3)
B6903 2 EOL:
Axist                                   58                           82. 5(3) h gg sis y
Int. Shell C6308 2 A 5338-1 65 2.34E19 87 651 6/25/2017 B6903-3 Lower C6940 1 A 5338 1 -
Welds References for Farley i Fluence, heet ruber and (AJ$E vetues f rom {{letter dated|date=November 23, 1993|text=November 23, 1993, letter}} f rom D. Morey (sw0C) to USNRC Docunent Control Desk, subject: Responses to Requests for Additional Information Regarding GL 92 01.
2.34E19 86 65%
(1)     The Mococ e , *'. Co , and onirra dde d USE values are ideMic AI Or lower Shell pides T> 6919 -1 ne d 3 6919 - 2. ; iherefore.,4he co r r e c t value for V4T USE cd E,0L / Er Py for plM e. 3G919 - 1 is G2 .
shall N
(7.)   The c irc. , w eld unirr adiMed USE                 and V4 T OSE         n+     EoL/ EFPY ar e- ebAn$ed io re flec3 ik e- a ppr oach o$ osia$ ike +10*F C.har py value. a <, w co n se rva b e_.
g 86919-1 Lower C6897 2 A 5338 1 62 2.34E19 86 65%
estimai c d Ak e. onir radi Me d USE i o d e m o n d e d e. Abd YT                4 USE 4 EOL/EFP'/
shott 86919 2 Int. Shett 33A2TT Linde 115 7.73E18 149 surv.
i5 $ realer ihan 50 if       The lnwer shell Adal wcils onirradiaf ed OSE and VT                           4  OSE at EoL/EFPY ct e c    chan gc cl   ao reflec3 an unirradla4c d OSE determined by 3Anlis4ical a n ,,1 y sis . ne vauwd analysis is described in Miachmed                               2. .
Axlet 1092, SAW Weld Welds
l
+10*F (1)
            ' Sur+eillance weld-4s-free-d4-f-f+ rent he:t th:n belt 1 %e > elds,
Cire. Weld 6329637 Linde 2.34E19 5;!r. t:
    -add 4ti on a l-i nfo rma t4 on-requ ired-to-con f4 rm44 tense e ' : value-1 1
-44 5- +)
1
58 '*
104 '*I 0091, SAW
!=-
^'IY mid value.
Lower 90099 Linde 7.73E18 E;!- t:
Stati,+l cal Shell 0091, SAW (3) h gg sis Axist 58
: 82. 5(3) %
y Welds References for Farley i Fluence, heet ruber and (AJ$E vetues f rom {{letter dated|date=November 23, 1993|text=November 23, 1993, letter}} f rom D. Morey (sw0C) to USNRC Docunent Control Desk, subject: Responses to Requests for Additional Information Regarding GL 92 01.
(1)
The Mococ e, *'. Co, and onirra dde d USE values are ideMic AI Or lower Shell pides T> 6919 -1 ne d 3 6919 - 2. ; iherefore.,4he co r r e c t value for V4T USE cd E,0L / Er Py for plM e. 3G919 - 1 is G2.
ebAn$ed (7.)
The c irc., w eld unirr adiMed USE and V T OSE n+
EoL/ EFPY ar e-4 io re flec3 ik e-a ppr oach o$ osia$ ike +10*F C.har py value.
co n se rva b e_.
a <, w YT USE 4 EOL/EFP'/
estimai c d Ak e. onir radi Me d USE i o d e m o n d e d e.
Abd 4
i5 $ realer ihan 50 if The lnwer shell Adal wcils onirradiaf ed OSE and VT OSE at EoL/EFPY 4
chan gc cl unirradla4c d OSE determined by 3Anlis4ical ao reflec3 an ct e c a n,,1 sis.
ne vauwd analysis is described in Miachmed
: 2..
y l
' Sur+eillance weld-4s-free-d4-f-f+ rent he:t th:n belt 1 %e > elds,
-add 4ti on a l-i nfo rma t4 on-requ ired-to-con f4 rm44 tense e ' : value-1


Sumary File for Upper Shelf Energy Plant Name   Baltiltw       Meet No.     Material     1/4T U$E   1/4T           Unirred. Method of Ident.                     Type         at         Woutron         USE         Determin.
Sumary File for Upper Shelf Energy Plant Name Baltiltw Meet No.
EOL/EFPT     Fluence at                 Leirred.
Material 1/4T U$E 1/4T Unirred.
EOL/EFPY                   USE Farley 2     Int. SheLL   C6309 2       A $338 1     72           2.369E19         100 -    Direct 87203-1 EOL:         Int. Shell   CT466 1       A 5334-1     62           2.369E19         100       Direct 3/31/2021   s7212-1 Lower         C6888 2       A 5338 1     76           2.369E19       103         Direct Shell B7210-1 Lower         C6293 1       A 5338 1     72           2.369E19       99         0iroct shell 8T'10 2 circ. Weld                                                            #                      + 10
Method of Ident.
* F bI SP5622       Linde       -.nis.f*-     2.369E19                   5;F *:
Type at Woutron USE Determin.
11 923                       0091, sAW       g (0                                 ~8'**~~
EOL/EFPT Fluence at Leirred.
t01                    [o[
EOL/EFPY USE Farley 2 Int. SheLL C6309 2 A $338 1 72 2.369E19 100 Direct 87203-1 EOL:
Lower         83640         Linde       #            7.67E18       #            5 ;'". ': + IO'F (0 Shell                       0091, SAW                                             -surv Antal                                        103(0                       116(0     y,44 ,_   tu], Y Welds 20 923A/s Int. Shell     NODA         SMAW         -4M +         7.67E18       -448 +-     " - ' " .. +LO*F N Axial                                                                             r..='
Int. Shell CT466 1 A 5334-1 62 2.369E19 100 Direct 3/31/2021 s7212-1 Lower C6888 2 A 5338 1 76 2.369E19 103 Direct Shell B7210-1 Lower C6293 1 A 5338 1 72 2.369E19 99 0iroct shell 8T'10 2
Clnar ry Welds                                         107 (0                      13)(Q    yo;4_.     ygge 19-923A Int. Shell   BOLA         SMAW         131         7.67E18         148         sury.
+ 10
Axial                                                                             Weld Welds 19-9238 REFERENCES FOR FARLEY 2:
* F bI circ. Weld SP5622 Linde
-.nis.f*-
2.369E19 5;F *:
11 923 0091, sAW g (0 t01
~8'**~~
[o[
Lower 83640 Linde 7.67E18 5 ;'". ':
+ IO'F (0 Shell 0091, SAW 103(0 116(0
-surv tu], Y Antal y,44,_
Welds 20 923A/s
+LO*F N Int. Shell NODA SMAW
-4M +
7.67E18
-448 +-
Axial 107 (0 13)(Q Clnar ry r..='
Welds yo;4_.
ygge 19-923A Int. Shell BOLA SMAW 131 7.67E18 148 sury.
Axial Weld Welds 19-9238 REFERENCES FOR FARLEY 2:
Fluence, heat ruber and LA;$[ values f rom Novenber 23, 1993, letter from D. Morey (SWOC) to USNRC Docunent Control Desk., subject: Responses to ReqJests for Additional Information Regarding GL 92-01.
Fluence, heat ruber and LA;$[ values f rom Novenber 23, 1993, letter from D. Morey (SWOC) to USNRC Docunent Control Desk., subject: Responses to ReqJests for Additional Information Regarding GL 92-01.
(t)    ~T b c cir c , w eld ( 11 - 9 2.3) , lower Skell adal Weldt ( 20 - 9 2 3 A /3) , ^n d id , sh c Il a x ial weld (19-923 A) unkradiat e j USE and PT                     4  USE ai LOL/Er ry are        ebe g c cl do r e dec + 4ke approach c4 vsA3 ike +to*F C%necy valoe      as a eme r va%e edade of on,rrata;ca use so demonstede ibd VT           4  USE al EOL/ Erry 15 gred er 4kan 50-l
w eld ( 11 - 9 2.3), lower Skell adal Weldt ( 20 - 9 2 3 A /3),
                '"S u r v e i ' l ance-weld-4s-frw-a-4tMerent-heat-tha n-be4414ne-we146-
^n d (t)
    --add 444onal informat4onsequired-40-conf 4rm-14cencee's nlue t
~T b c cir c,
id, sh c Il a x ial weld (19-923 A) unkradiat e j USE and PT USE ai 4
ebe g c cl do r e dec + 4ke approach c4 vsA3 LOL/Er ry are ike
+to*F eme r va%e edade of on,rrata;ca use so C%necy valoe as a demonstede ibd VT USE al EOL/ Erry 15 gred er 4kan 50-4 l
'"S u r v e i ' l ance-weld-4s-frw-a-4tMerent-heat-tha n-be4414ne-we146-
--add 444onal informat4onsequired-40-conf 4rm-14cencee's nlue t
1 L}}
1 L}}

Latest revision as of 10:10, 16 December 2024

Responds to Open Issues Re GL-92-01,rev 1, Reactor Vessel Structural Integrity
ML20070A012
Person / Time
Site: Farley  Southern Nuclear icon.png
Issue date: 06/21/1994
From: Woodard J
SOUTHERN NUCLEAR OPERATING CO.
To:
NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
GL-92-01, GL-92-1, TAC-M83461, TAC-M83462, NUDOCS 9406280120
Download: ML20070A012 (11)
v • d • e


Text

Southern Nuclear Operating Company Post Office Box 1295 Birmingham, Alabama 35201 Telephone (205) 868-5066 k

L Southern Nudear Operating Company J. o. Woodard Executive Vice President the southem elecinc System June 21, 1994 10 CFR 50.54(f)

Docket Nos. 50-348 50-364 TAC Nos.

83461 83462 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555 Joseph M. Farley Nuclear Plant Responses to Open Issues Regarding Generic Letter 92-01, Revision 1 Reactor Vessel Structural Integrity Gentlemen:

On March 6,1992, the NRC issued Generic Letter (GL) 92-01, Revision 1, " Reactor Vessel Structural Integrity." The purpose of the GL was to obtain information needed to assess compliance with requirements and commitments regarding reactor vessel integrity due to events associated with the Yankee Nuclear Power Station. Southern Nuclear Operating Company (SNC) provided a response to GL 92-01, Revision 1, by letter dated July 1,1992. On October 7,1993, the NRC issued SNC a request for additional information (RAI) in order to complete the review of the SNC response to GL 92-01, Revision 1. SNC provided a response to the RAI by letter dated November 23,1993.

The NRC notified SNC, by letter dated May 20,1994, of two open issues regarding GL 92-01 and requested verification of the NRC summary data file information for Farley Nuclear Plant. provides SNC's responses to the open items identified by the NRC associated with GL 92-01, Revision 1. Attachment 2 provides the data used to determine a statistical value for the unirradiated upper shelf energy for type B4 weld filler material. provides a marked-up copy of the summary data file for the Farley Unit 1 and Unit 2 vessels to reflect changes to the information with references supporting each change.

k 9406280120 940621 Af I

PDR ADOCK 05000348 p

l P

PDR

U. S. Nuclear Regulatory Commission Page 2 As stated in previous submittals regarding GL 92-01, Revision 1, SNC continues to comply with the requirements of 10 CFR 50.61 and 10 CFR 50, Appendix G.

If there are any questions, please advise.

Respectfully submitted, SOUTHERN NUCLEAR OPERATING COMPANY

\\

\\h

'oodard SWORN TO AND SUBSCRIBED BEFORE ME THIS dek)AY OF b

,1994 ht& LA L0ne-Notary'Public

(/

My Commission Expires: 7/dua. /, /997 DRC; cit 920lRAl2. doc Attachments cc:

Mr. S. D. Ebneter Mr. B. L Siegel Mr. T. M. Ross i

i

+

r P

ATTACIIMENT 1 SNC RESPONSES TO GL 92-01, REVISION 1, OPEN ISSUES

?

b I

I i

I l

i

OperLIntttu)

The first open issue identified for Farley was associated with weld wire heats 33 A277,6329637, and 90099 in Unit 1, and weld wire heats 5P5622 and 83640 for Unit 2. Specifically, the NRC staff stated that the nickel content for these heats was determined as mean values from the Westinghouse Owners Group (WOG) database and requested that SNC provide the WOG data that was used to determine the amount of nickel. Additionally, SNC was requested by the stafTto determine the best-estimate amount of nickel in accordance with the PTS Rule,10 CFR 50.61.

SNC Response to Open Issue (1)

SNC stated in our response to NRC Request for Additional Information (RAI) regarding GL 92-01, Revision 1, dated November 23,1993, that the nickel content for beltline welds was taken from earlier WOG programs. Specifically, the 0.20% value was determined from a 1982 program performed to calculate operating and near term operating limit vessel RTndt values and the 0.21%

value was detennined from a program to develop a materials database aimed at filling in gaps where data was not available. As stated in our response to the RAI, the nickel content of the beltline welds is not available from the material certifications and the values listed are based on engineeringjudgement. However, a footnote to Table 2 and Table 8 of the RAI response identified these nickel values as mean values taken from the WOG materials database.

Subsequent conversations with Westinghouse have verified that the WOG database value for nickel content of the beltline welds is based solely on engineering judgement described below and not actual material certifications.

The rationale used to determine the WOG database nickel value for weld wire heats 33 A277, 6329637, 90099, 5P5622, and 83640 in the vessel beltline welds considered that both the Farley Unit I and Unit 2 vessels were fabricated between 1971 and 1973 by Combustion Engineering (CE)in Chattanooga, Tennessee. The vessels were fabricated using the automatic submerged arc welding (SAW) process and type B4 weld filler material supplied by the Reid Avery Company.

Nickel was not added as an alloying element to the B4 weld filler material manufactured by the Reid Avery Company; therefore, CE did not require the Reid Avery Company to perfonn or submit a chemical analysis value for tickel in type B4 weld filler wire. A chemical analysis of the Farley Unit I surveillance weld ma;erial, heat number 33 A277, indicated a nickel content of 0.19 weight percent, consistent with the estimated value of 0.20 weight percent reported in our response to the RAI.

SNC is a charter member of the Combustion Engineering Reactor Vessel Group (CERVG) and joined the group with the expectation of obtaining additional information to augment existing vessel chemistry and toughness data for SNC vessels. Based on reviews performed to date, SNC does not expect the nickel content of weld wire heats 33 A277,6329637, 90099, SP5622, and 83640 to exceed 0.20 weight percent. The CERVG effort is currently targeted for completion by December 1994.

l Page1of2

Open Issue (2)

The second open issue for Farley involved a NRC staff concern that surveillance data was used to determine the unirradiated upper shelf energy (USE) values for weld wire heat numbers 6? 29637, i

90099, SP5622,83640, and HODA. The staffs letter stated that since the surveillance data are from a different heat, a statistical analysis addressing heat variability may be appropriate and that when the unirradiated USE for a particular heat of material has not been determined, the USE can be set equal to the lower tolerance limit calculated for the group of similar materials. The staff stated that the unirradiated USE should be determined such that there exists 95% confidence that at least 95% of the population is greater than the !ower tolerance limit and if the lower tolerance limit results in a projected USE at EOL orless than 50 fl-lb, then SNC must demonstrate, in accordance with Appendix G,10 CFR Part 50, that equivalent lower values of USE will provide margins of safety against fracture equivalent to those required by Appendix G of Section III of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code.

SNC Response to Open Issue (2)

A statistical analysis was performed by Westinghouse, in accordance with the NRC staffs guidance, to determine a generic unirradiated USE applicable to heats 6329637,90099, SP5622, and 83640. Attachment 2 provides 61 values of Charpy V-notch energy at 100% shear where data for full Charpy curves are available for type B4 weld wire. The lower tolerance limit determined from the data in Attachment 2 resulted in a generic unirradiated USE value of 82.5 ft-Ib. For weld wire heat number 90099, Farley will utilize the generic unirradi ated USE value of 82.5 fl-lb determined by the statistical analysis. Use of the generic unirradiated USE value for heat 6329637 would result in a projected EOL USE less than 50 fl-lb. However, for heat 6329637 the Farley specific Charpy value at +10 F is known and can be used as a conservative estimate for the unirradiated USE since the unirradiated USE is known to be higher than the

+10 F Charpy value. The known Farley specific +10 F Charpy values are also being used for heats SP5622 and 83640. Since the statistical analysis is only applicable to type B4 weld wire, the

+10 F Charpy value is also being used for the SMAW filler metal, heat HODA.

Based on the above information, the projected EOL USE for the Farley Unit I and 2 belthne welds exceeds 50 fl-lb, as shown in Attachment 3; therefore, Farley Units 1 and 2 will continue to comply with the requirements of Appendix G of 10 CFR Part 50.

Page 2 of 2 l

l

ATTACIIMENT 2 Charpy V-Notch Energy (fl-lb) Data at 100% Shear for Determination of Generic Unirradiated USE for Welds with Type B4 Weld Filler Wire 138 159 151 152 159 174 156 150 151 162.5 144 154 139 140 149 153 135 140 141 144 154 158 135 144 145 143 143 144 122 122 l

127 125 125 125 125 172 154 162 158 87 90.5 84.5 109 116 108 126 121 99 118 129 123 113 119 86 88 82 90 94 99 100 102 Number of Points = 61 Sample Mean = 130.467 Sample Deviation = 24.4896 Lower Tolerance Bound = 82.5

ATTACHhiENT 3 h1ARKED-UP SUhihiARY DATA FILE FOR FARLEY UNITS 1 & 2 VESSELS The NRC stafTrequested that SNC verify the summary data file information contained in Enclosures 1,2, and 3 of the NRC's hiay 20,1993, letter. Accordingly, SNC provides the following marked-up copics of the summary data file with reference information to support the changes.

Sumary File for Pressurized Thermal Shock P la.itt Betttlne Meet No.

10 Neut.

IRT.

Method of Chemistry Method of

%Cu

%Ni kame Ident.

Ident.

Fluence et Detersin.

Factor Determin.

EOL/EFPY 1 R T, CF F2rtey 1 Int. Shett C6294 1 3.75E19 0*F MTEB 5 2 91 febte 0.13 0.60 86903 2 EOL:

Int. Shett C6308 2 3.75E19 10'F MTEB 5 2 82.2 Table 0.12 0.56 6/25/2017 86903 3 Lower C6940 1 3.75E19 15'F MTER 5 2 88.831 Calculated 0.14 0.55 j

Shett l

B6919 1 j

Lower C6897 2

3. 75E19 5'F MTEs 5-2 98.2 Table 0.14 0.56 shelt 86919 2 Int. Shett 33A277 1.24E19 56'F Generic 78.689 Calculated 0.25 0.21 Axial Welds

-4430. W Cire. Weld 6329637 3.75E19 56'F Generic

-443-Table 0.20 tower 90099 1.24E19

-56*F Generic 92 Table 0.17 0.20 Sheit Axlet Welds REFEREhCES FOR FARLEY 1:

Fluence, IRT and chemistry values f rom hovember 23, 1993, letter from D. Morey (suoC) to UskRC Docet Control Desk, select: Responses to Regaeste for Additionet Information Regarding GL 92-01.

j vnnr ked-up circ.. weld % Cu ei o.zz5 a 'd cW misby &clor (0

Tbc i

ed 114.5 ( n om Table) were de valoc s submitied b h e_

l y

l tlovem be r 23, 1993 \\ceer refe r en ce d nbov e.,

1 I

is an esW M e d value-beed on eg oe er ing

' Chemical composition -from mean ;;lue of L'OC dat.

  1. ddition:1 "forn tier rem. ired.

jg,,,g, g% g, 4e q,,,,,, ;

pa$,,,a is A c>riib e ) in A h ch,er+

1.

Summary File for Pressurized Thermal Shock l

Pitnt BettlIne Heat No.

1D Neut.

I R T.

Method of Chemistry Method of

%Cu Emi Name Ident.

Ident.

Fluence at Detersin.

Factor Determin.

{

E0L/EFPT IR T.

CF Farley 2 Int. shett C6309 2 3.8E19 15*7 Plant 100 Table 0.14 0.60 87203 1 Specific EOL:

Int. Shett C7466-1 3.8E19 10'F Plant 145.0 Calculated 0.20 0.60 3/31/2021 e7212 1 specific Lower C6888 2 3.8E19 18'F Plant 89.8 Table 0.13 0.56 shett Specific 87210-1 Lower C6293 1 3.8E19 10'F Plant 98.7 fable 0.14 0.57

$helt Specific 87210-2 Cire. Weld 5P5622 3.8E19

-40*F Plant 76 fable 0.13 0.20 11-923 specific Lower 83640 1.23E19

-70'F Plant 49 Table 0.05 0.20 Shett Specific Axlet Welds 20 923 A/S Int, shett H00A 1.23E19

-56*F Generic 10.01 Cateutated 0.02 0.%

Axial

$ MAW Welds 19 923A Int. Shell BOLA 1.23E19

-60'F Plant 10.01 Calculated 0.02 0.93 Axist SMAW Specific Welds 19 9238 REFERENCES FOR FARLET 2:

Fluence, IRT, and chemistry values from Novenber 23, 1993, letter from D. Morey (SNOC) to USNRC Doctament Control Desk, s@ ject: Responses to Requests for Additional Information Regarding 92-01.

l 0

wah d vale basr A en engineer-in3 13 an es

, Chemical composition from meanwalue-of-WOG data. Additional

-informa t i c" required-to--confirm v a l u e -

pdcmed.

~n c bd3 for A 3

d eng.ncericg y 3cment is Jc w a cJ i,. Att.stbnen+ 1

i l

Sumary File for Upper Shelf Energy Plant Name Settline Neat No.

Material 1/4T USE 1/4T unirred.

Method of Ident.

Type at Neutron USE Deternin.

i EOL/EFPY Fluence at Unfrred.

EOL/EFPY USE Farley 1 Int. SheLL C6294 1 A 5338 1 73 2.34E19 99 65%

B6903 2 EOL:

Int. Shell C6308 2 A 5338-1 65 2.34E19 87 651 6/25/2017 B6903-3 Lower C6940 1 A 5338 1 -

2.34E19 86 65%

shall N

g 86919-1 Lower C6897 2 A 5338 1 62 2.34E19 86 65%

shott 86919 2 Int. Shett 33A2TT Linde 115 7.73E18 149 surv.

Axlet 1092, SAW Weld Welds

+10*F (1)

Cire. Weld 6329637 Linde 2.34E19 5;!r. t:

-44 5- +)

58 '*

104 '*I 0091, SAW

!=-

^'IY mid value.

Lower 90099 Linde 7.73E18 E;!- t:

Stati,+l cal Shell 0091, SAW (3) h gg sis Axist 58

82. 5(3) %

y Welds References for Farley i Fluence, heet ruber and (AJ$E vetues f rom November 23, 1993, letter f rom D. Morey (sw0C) to USNRC Docunent Control Desk, subject: Responses to Requests for Additional Information Regarding GL 92 01.

(1)

The Mococ e, *'. Co, and onirra dde d USE values are ideMic AI Or lower Shell pides T> 6919 -1 ne d 3 6919 - 2. ; iherefore.,4he co r r e c t value for V4T USE cd E,0L / Er Py for plM e. 3G919 - 1 is G2.

ebAn$ed (7.)

The c irc., w eld unirr adiMed USE and V T OSE n+

EoL/ EFPY ar e-4 io re flec3 ik e-a ppr oach o$ osia$ ike +10*F C.har py value.

co n se rva b e_.

a <, w YT USE 4 EOL/EFP'/

estimai c d Ak e. onir radi Me d USE i o d e m o n d e d e.

Abd 4

i5 $ realer ihan 50 if The lnwer shell Adal wcils onirradiaf ed OSE and VT OSE at EoL/EFPY 4

chan gc cl unirradla4c d OSE determined by 3Anlis4ical ao reflec3 an ct e c a n,,1 sis.

ne vauwd analysis is described in Miachmed

2..

y l

' Sur+eillance weld-4s-free-d4-f-f+ rent he:t th:n belt 1 %e > elds,

-add 4ti on a l-i nfo rma t4 on-requ ired-to-con f4 rm44 tense e ' : value-1

Sumary File for Upper Shelf Energy Plant Name Baltiltw Meet No.

Material 1/4T U$E 1/4T Unirred.

Method of Ident.

Type at Woutron USE Determin.

EOL/EFPT Fluence at Leirred.

EOL/EFPY USE Farley 2 Int. SheLL C6309 2 A $338 1 72 2.369E19 100 Direct 87203-1 EOL:

Int. Shell CT466 1 A 5334-1 62 2.369E19 100 Direct 3/31/2021 s7212-1 Lower C6888 2 A 5338 1 76 2.369E19 103 Direct Shell B7210-1 Lower C6293 1 A 5338 1 72 2.369E19 99 0iroct shell 8T'10 2

+ 10

  • F bI circ. Weld SP5622 Linde

-.nis.f*-

2.369E19 5;F *:

11 923 0091, sAW g (0 t01

~8'**~~

[o[

Lower 83640 Linde 7.67E18 5 ;'". ':

+ IO'F (0 Shell 0091, SAW 103(0 116(0

-surv tu], Y Antal y,44,_

Welds 20 923A/s

+LO*F N Int. Shell NODA SMAW

-4M +

7.67E18

-448 +-

Axial 107 (0 13)(Q Clnar ry r..='

Welds yo;4_.

ygge 19-923A Int. Shell BOLA SMAW 131 7.67E18 148 sury.

Axial Weld Welds 19-9238 REFERENCES FOR FARLEY 2:

Fluence, heat ruber and LA;$[ values f rom Novenber 23, 1993, letter from D. Morey (SWOC) to USNRC Docunent Control Desk., subject: Responses to ReqJests for Additional Information Regarding GL 92-01.

w eld ( 11 - 9 2.3), lower Skell adal Weldt ( 20 - 9 2 3 A /3),

^n d (t)

~T b c cir c,

id, sh c Il a x ial weld (19-923 A) unkradiat e j USE and PT USE ai 4

ebe g c cl do r e dec + 4ke approach c4 vsA3 LOL/Er ry are ike

+to*F eme r va%e edade of on,rrata;ca use so C%necy valoe as a demonstede ibd VT USE al EOL/ Erry 15 gred er 4kan 50-4 l

'"S u r v e i ' l ance-weld-4s-frw-a-4tMerent-heat-tha n-be4414ne-we146-

--add 444onal informat4onsequired-40-conf 4rm-14cencee's nlue t

1 L

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