ML20072C064
| ML20072C064 | |
| Person / Time | |
|---|---|
| Site: | Braidwood  |
| Issue date: | 08/08/1994 |
| From: | Saccomando D COMMONWEALTH EDISON CO. |
| To: | Russell W NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM), Office of Nuclear Reactor Regulation |
| References | |
| GL-92-01, GL-92-1, NUDOCS 9408170055 | |
| Download: ML20072C064 (11) | |
Text
i Edison
_Z_. /) Comm:nw:c th 1400 Opus Place Downers Grove. Ilknois 60515 August 8,1994 Mr. William Russell, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington D.C. 20555 Attention:
Document Control Desk
Subject:
GL 92-01, " Reactor Vessel Structural Integrity" Data Update Braidwood Station Units 1 and 2 (NPF-72/77; NRC Docket Nos. 50-456/457)
Reference:
1)
J. Bauer letter to W. Russell dated July 22,1994, transmitting update on GL 92-01 2)
R R. Assa letter to D. L. Farrar dated June 24,1994, requesting verification of data pertaining to GL 92-01,
" Reactor Vessel Structural Integrity
Dear Mr. Russell:
By letter dated July 2,1992 and November 19,1993, Commonwealth Edison Company (Com Ed) provided its response to GL 92-01, Revision 1. In the referenced letter 2, the Nuclear Regulatory Commission (NRC) Staff requested the Comed verify the previously supplied information, as this data will be entered into a Reactor Vessel Integrity Database.
In reference 1, Comed requested that the Braidwood data verification / update be delayed until August 8,1994, to allow for a complete and accurate verification of the subject data.
The attachment includes Braidwood's response to the reference 2 correspondence. In the summary tables, the changes from the data tr asmitted in reference 2 are clearly indicated.
Please address any comments or questions regarding this matter to this office.
Si carely, N
,,, 4 Denise M. Sac - nando Nuclear Licensing Administrator ec:
R. R. Assa, Braidwood Project Manager, NRR S. G. Dupont, Senior Resident Inspector - Braidwood B Clayton, Branch Chief-Region III Office of Nuclear Facility Safety - IDNS j00 9 0
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lI' Attachment Braidwood Units 1 and 2 Response to NRC Generic Letter 92-01, Revision 1 s
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4 Introduction i
i This report provides a response to the Generic Letter 92-01, Revision 1, closure letter recently issued by the NRC for Commonwealth Edison Company's Braidwood Units 1 and 2.
The following is the full Data Summary Tables for Pressurized Thermal Shock and Upper-Shelf Energy.
Those values that are unchanged are shown in the shaded boxes.
Revised values are indicated in the unshaded boxes.
1 1
1.
l ktnla % l\\g19201r\\3
)
~
L Table 2-1.
Braidwood Unit 1 -- Data Summary for Pressurized Thermal Shock Calculation
~
h IS Neutron Method of Method of Beltline Fluence at IRT,
Determin.
Chemistry Determin.
Material Heat No.
32 EFPY F
IRT,
Factor CF
%Cu
%Ni Lower SP-7016 6.82E+18
+10*
Plant 26 RG1.99 0.04*
0.71N Nozzle Specific Table 2 Belt Forging Upper 49C344-1-1/
3. 0 3 E +19 '*3
-30N Plant 31 RG1.99 0.05" 0.73N-Shell 49D383-1-1 Specific Table 2 Forging Lower 49D867-1-1/
3. 0 3 E+19
-20" Plant 20 RG1.99 0.03" 0.73N Shell 49C813-1-1 Specific Table 2 Forging HF-645 H4498 6. 8 2 E + 18
-30*
Plant 41 RG1.99 0.03*
0.50W Upper Specific Table 1 Cire. Weld WF-562 442011 3. 0 3 E+19 '*3
+40*
Plant 41 RG1.99 0.03*
0.65N Middle Specific Table 1 Cire. Weld WF-653 31401
< 1. 0 0 E + 17 '*3
-40 Plant 150.8 RG1.99 0.19N~
0.56N N
Lower Specific Table.1 Cire. Weld NOTES:
a.
Fluence data are from WCAP-12685, " Analysis of Capsule U from the Commonwealth Edison Company Braidwood Unit 1 Reactor Vessel Surveillance Program," August 1990.
b.
Chemical compositions and initial RTa data for all materials are from the July 2, 1992 letter from M.
A.
Jackson to T.
E. Murley,
Subject:
Braidwood Station, Units 1 and 2.
Prepared By:
M.
J.
DeVan Date:
7/1/94 77-1234176-00 Reviewed By:
L.
B.
Gross Date:
7/1/94 Page 3
^
Table 2-2.
Braidwood Unit 1 -- Data Summarv for Uoner-Shelf Enerov Calculation 1/4T Method of
(
1/4T USE Neutron Determin.
l Beltline Material at 32 Fluence at Unirrad.
Unirrad.
(
Material Heat No.
Type EFPY 32 EFPY USE USE Lower SP-7016 A 508-2 13 7
4. 0 9E &l 8 '88 16 2
Direct Nozzle Belt Forging Upper 49c344-1-1/
A 508-3 9 3 '*'
1.66E+19'd' 110'"
Direct Shell 49D383-1-1 Forging Lower 49D867-1-1/
A 508-3 10 7
1.66E+19'83 13 6
Direct Shell 49C813-1-1 l
Forging WF-645 H4498 Linde 80, 75'*8 4.09E+18'8 87'*3 Direct 5
Upper SAW l
Cire. Weld WF-562 442011 Linde 80, 5 5 *'
- 1. 6 6 E+ 19 '83 708 Direct l
l Middle SAW Circ. Weld WF-653 31401 Linde 80,
< 1. 0 0E+ 17 'd8 7 9
Direct Lower SAW Circ. Weld l
l j
Prepared By:
M.
J.
DeVan Date:
7/1/94 77-1234176-00 Reviewed By:
L.
B.
Gross Date:
7/1/94 Page 4
NOTES FOR TABLE 2-2:
i a.
EOL USE values for the forgings were calculated using Regulatory Guide 1.99, Revision 2, Figure 2,
assuming the lower limiting value of 0.1% copper for base metal.
b.
EOL USE values for the welds were calculated using Regulatory Guide 1.99, Revision 2, Figure 2, l
assuming the lower limiting value of 0.05% copper for welds.
c.
EOL fluence is below the limits of the Figure 2 curves defined in Regulatory Guide 1.99, Revision 2.
d.
Fluence data are from WCAP-12685, " Analysis of Capsule U from the Commonwealth Edison Company Braidwood Unit 1 Reactor Vessel Surveillance Program," August 1990.
e.
UUSE data are from the November 19,1993 letter from T.
W.
Simpkin to T. E. Murley, Braidwood Station Units 1 and 2, Response to Request for Additional Information Regarding NRC Generic Letter 92-01.
f.
UUSE data for forging 49C344-1-1/49D383-1-1 is from the July 2, 1992 letter from M. A. Jackson to T.
E. Murley,
Subject:
Braidwood Station, Units 1 and 2.
Prepared By:
M.
J.
DeVan
.Date:
7/1/94 77-1234176-00 Reviewed By:
L.
B.
Gross Date:
7/1/94 Page 5 i
i
Table 2-3.
Braidwood Unit 2 -- Data Summary for Pressurized Thermal Shock Calculation IS Neutron Method of Method of Beltline Fluence at IRTm Determin.
Chemistry Determin.
Material Heat No.
32 EFPY F
%Cu
%Ni m
Lower SP-7056 6.82E+18
+30*'
Plant 26 RG1.99 0.04*'
O.90*3 Nozzle Specific Table 2 Belt Forging Upper 49D963-1-1/
3. 0 3 E + 19
-30*'
Plant 20 RG1.99 0.03*'
O.71*3 Shell 49C904-1-1 Specific Table 2 Forging-S' O.75*'
Lower-50D102-1-1/
3. 0 3 E+ 19
-30*'
Plant 37 RGl.99 0.06 Shell 50C97-1-1 Specific Table 2 Forging WF-645 H4498 6. 8 2 E+ 18
'-30**
Plant 41.
RG1.99 0.03*'
O.50*)
Upper Specific Table 1 Cire. Weld WF-562
-442011 3. 03 E + 19 '*8
+40*8 Plant 41 RG1.99 0.03**
- 0. 6 5 *!
Middle Specific Table 1 Cire. Weld WF-696
_1084-18
< 1. 0 0 E + 17
-16*'
Plant 54 RG1.99 0.04*3 0.60*8 Lower Specific Table 1 Cire. Weld NOTES:
a.
Fluence data are from WCAP-12845, " Analysis of Capsule U from the Commonwealth Edison Company Braidwood Unit 2 Reactor Vessel Surveillance Program," March 1991.
b.
Chemical compositions and initial RT data for all materials are from the July 2, 1992 letter from m
M.
A. Jackson to T.
E. Murley,
Subject:
Braidwood Station, Units 1 and 2.
Prepared By:
M.
J.
Devan Date:
7/1/94 77-1234176-00 Reviewed By:
L.
B.
Gross Date:
7/1/94 Page 6
\\
?
l Table 2-4.
Braidwood Unit 2 -- Data Summary for Upper-Shelf Energy Calculation
(
1/4T Method of 1/4T USE Neutron Determin.
Beltline Material at 32 Fluence at Unirrad.
Unirrad.
Material Heat No.
Type EFPY 32 EFPY USE USE Lower SP-7056 A 508-2 10 9 '**
4. 0 9 E +18
128*'
Direct 8
Nozzle Belt Forging 8
Upper 49D963-1-1/
A 508-3 94 *'
- 1. 6 6 E +19 'd' 119'"
Direct Shell 49C904-1-1 Forging Lower 50D102-1-1/
A 508-3 118'*8
- 1. 6 6 E + 19
15 0
l Direct Shell 50C97-1-1 Forging WF-645 H4498 Linde 80, 7 5 'b' 4. 0 9E+18 :42 8 7
Direct Upper SAW Cire. Weld WF-562 442011 Linde 80, 5 5 'b' 1.66E+19888 7 0
Direct Middle SAW Circ.~ Weld WF-696 1084-18 Linde 80,
--'*3
< 1. 0 0E+ 17 'd' 78 *8-Direct 8
Lower SAW-Circ. Weld i
t-Prepared By:
M.
J.
DeVan Date:
7/1/94 77-1234176-00 Reviewed By:
L.
B.
Gross Date:
7/1/94 Page 7
g NOTES FOR TABLE 2-4:
a.
EOL USE values for the forgings were calculated using Regulatory Guide 1.99, Revision 2, Figure 2,
assuming the lower limiting value of 0.1% copper for base metal.
b.
EOL USE values for the welds were calculated using Regulatory Guide 1.99, Revision 3, Figure 2, assuming the lower limiting value of 0.05% copper for welds.
c.
EOL fluence is below the limits of the Figure 2 curves defined in Regulatory Guide 1.99, Revision 2.
i d.
Fluence data are from WCAP-12845, " Analysis of Capsule U from the Commonwealth Edison Company Braidwood Unit 2 Reactor Vessel Surveillance Program," March 1991.
e.
UUSE data are from the November 19,1993 letter from T.
W.
Simpkin to T.
E. Murley, Braidwood Station Units 1 and 2, Response to Request for Additional Information Regarding NRC Generic Letter 92-01.
1 f.
UUSE data for forging 49D963-1-1/49C904-1-1 is from the July 2, 1992 letter from M. A. Jackson to T.
E.
Murley,
Subject:
Braidwood Station, Units 1 and 2.
i i
l Prepared By:
M.
J.
DeVan Date:
7/1/94 77-1234176-00 Reviewed By:
L.
B.
Gross Date:
7/1/94 Page 8
PRESSURIZED THERMAL SHOCK AND USE TABLES FOR ALL PWR PLANTS HOMENCLATURE Pressurized Thermal Shock Table Column 1:
Beltline material location identification.
Column 2:
Beltline material heat number; some welds that a single-wire or tandem-wire process has been reported, (s) indicates single wire was used in the SAW process, (T) indicates i
tandem wire was used in the SAW process.
1 Column 3:
End-of-life (EOL) neutron fluence at vessel inner wall; cited directly from inner diameter (ID) value reported in the latest submittal (GL 92-01, PTS, or P/T limits submittals).
Column 4:
Unirradiated reference temperature.
Column 5:
Method of determining unirradiated reference temperature (IRT).
Plant-Soecific This indicates that the IRT was determined from tests on material removed from the same heat of the beltline material.
Generic This indicates that the unirradiated reference temperature was determined from the mean value of tests on material of similar types.
Column 6:
Chemistry factor for irradiated reference temperature evaluation.
Column 7:
Method of determining chemistry factor.
RG1.99 Table 1 or 2 This indicates that the chemistry factor was determined from the chemistry factor tables in Regulatory Guide 1.99, Revision 2.
Calculated This indicates that the chemistry factor was determined from surveillance data via procedures described in Regulatory Guide 1.99, Revision 2.
Column 8:
Copper content; cited directly from licensee value except when more than one value was reported.
(Staff used the average value in the latter case.)
Column 9:
Nickel content; cited directly from licensee value except when more than one value was reported.
(Staff used the average value in the latter case.)
Prepared By:
M.
J.
DeVan Date:
7/1/94 77-1234176-00
)
Reviewed By:
L.
B.
Gross Date:
7/1/94 Page 15 l
l
.a.
/
Upper-Shalf Energy Table Column 1:
Beltline material location identification.
Column 2:
Beltline material heat number; some welds that a singl3-wire or tandem-wire process has been reported, (s) indicates single wire was used in the SAW process, (T) indicates tandem wire was used in the SAW process.
Column 3:
Material type; plate types include A 533B-1, A 302B, A 302B Mod.; forging types include A 508-2 and A508-3; weld types include SAW welds using Linde 80, 0091, 124, 1092, ARCOS-B5 flux, Rotterdam welds using Grau Lo, SMIT 89, LW 320, and SAF 89 flux, and SMAW welds using no flux.
Column 4:
EOL upper-shelf energy (USE) at T/4; calculated by using the EOL fluence and either the copper value or the surveillance data.
(Both methods are described in Regulatory Guide 1.99, Revision 2.)
Column 5:
EOL neutron fluence at T/4 from vessel inner wall; cited directly from T/4 value or calculated by using Regulatory Guide 1.99, Revision 2, neutron fluence attenuation methodology from the ID value reported in the latest submittal (GL 92-01, PTS, or P/T limits submittals).
Column 6:
Unirradiated USE Column 7:
Method of determining unirradiated USE.
Direct For forgings, this indicates that the unirradiated USE was from specimens oriented in the weak direction.
For welds, thia indicates that the unirradiated USE was from test data.
l 4
Prepared By:
M.
J. DeVan Date:
7/1/94 77-1234176-00 Reviewed By:
L.
B.
Gross Date:
7/1/94 Page 16