ML20073K545

Forwards Response to NRC Requesting Verification of Data Pertaining to GL 92-01,Rev 1, Reactor Vessel Structural Integrity

ML20073K545
Person / Time
Site:	Byron
Issue date:	10/04/1994
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 9410110282
Download: ML20073K545 (17)

Text

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. N C mmonw:alth Edison

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c' . 1400 Opus Place Downers Grove, lihnois 60515 October 4,1994 Mr. William Russell, Director OfIice 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 Byron Station Units 1 and 2 NPF-37/66; NRC Docket Nos. 50-454/455

References:

1. J. Bauer letter to W. Russell transmitting Commonwealth Edison Company response to NRC inquiry to GL 92-01 dated July 22,1994

2. R. Assa letter to D. Farrar dated June 24,1994, requesting verification of data pertaining to GL 92-01,

" Reactor Vessel Structural Integrity"

Dear Mr. Russell:

By letters dated July 2,1992, and November 19,1993, Commonwealth Edison Company (Comed) provided a response to GL 92-01, Revision 1. In the above reference letter 2, the Staff requested the Comed verify the previously supplied information by July 24,1994, as this data will be entered into a Reactor Vessel Integrity Database.

In reference letter 1, Comed requests that the Byron data verification / update be delayed until September,1994 due to completion of Capsule W results. The attachment includes Byron's response to the reference 2 correspondence. In the summary tables, the changes from the data transmitted in reference 2 are clearly indicated.

Please address any comments or questions regarding this matter to this office.

Sincerely, hDenise n M./ff b Shecomando Nuclear Licensing Administrator ec: G. Dick, Byron Project Manager, NRR II. Peterson, Senior Resident Inspector - Byron B. Clayton, Branch Chief- Regina III Office of Nuclear Facility Safety - IDNS m:nla\ byron \g19201\1 j

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Attachment Byron Units 1 and 2  !

Response to NRC Generic Letter 92-01, Revision 1 e

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Introduction 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 Byron Units 1 and 2. The fbilowing 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. >

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Table 2-1. Braidwood Unit 1 -- Data Surary for Pressurized Thermal Shock Calculation ,

IS Neutron Method of Method of Beltline Fluence at IR'Im Determin. Chemistry Determin.

Heat No. 32 EFPY F IRTm Factor CF %Cu %Ni Material 1 Lower SP-7016 6 . 8 2 E +18 +10

• Plant 26 RG1.99 0.04* 0.71

• Nozzle Specific Table 2 Belt Forging Upper 49C344-1-1/ 3 . 0 3E +19 -30* Plant 31 RG1.99 0.05* 0.73"'I Shell 49D383-1-1 Specific Table 2 Forging

-20* , Plant 20 RG1.99 0.03* 0,73

• Lower 49D867-1-1/ 3 . 0 3 E +19

Shell 49C813-1-1 Specific Table 2 Forging WF-645 H4498 6 . 8 2 E+18 -30* Plant 41 RG1.99 0.03* 0.50

• Upper Specific Table 1 Cire. Weld WF-562 442G11 3 . 0 3E + 19 - +40* Plant 41 RG1.99 0.03* 0.65

• Middle Specific Table 1 Cire. Weld WF-653 31401 < 1. 0 0E + 17 '** -40* Plant 150.8 RG1.99 0.19* 0.56 "

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 RT m 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 V-- _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _

Table 2-2. Braidwood Unit 1 -- Data Sumr.ary for Uroer-Shelf Enercy Calculation 1/4T Method of 1/4T USE Neutron Determin.

Beltline Material at 32 Flttence at UniirOd. Unirrad.

Material Heat No. Type EFPY 32 EFPY USE USE Lower 5P-7016 A 508-2 13 7 4 . 0 9E +18 'd' 16 2 Direct Nozzle Belt Forging ,

Upper 49C344-1-1/ A 508-3 9 3 '*' 1.66E+19'd' 118"' Direct Shell 49D383-1-1 Forging Lower 49D867-1-1/ A 509-3 10 7 1.66E+19 488 13 6 Direct-Shell 49C813-1-1 Forging WF-645 M4498 Linde 80, 7 5 4 . 0 9 E + 18 '*' 87(*1 Direct Upper SAW Circ. Weld WF-562 442011 Linda 80, 5 5 'b' 1.66E+19 i d' 7 0 Direct Middle SAW Cire. Weld WF-653 31401 Linde 80, ---' cl.00E+17 88) 7 9

5 Direct Lower SAW Circ. Weld 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: ,

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, assuming the lower limiting value of 0.05% copper for welds. l

c. EOL fluence is below the limits of the Figure 2 curves defined in Regulatory Guide 1.99, Revisian 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.

Date: 7/1/94 77-1234176-00 Prepared By:. M. J. DeVan Date: 7/1/94 Page S Reviewed By: L. B. Gross

Table 2-3. Braidwood Unit 2 -- Data Summarv for Pressurized Themal Shock Calculation IS Neutron Method of Method of Beltline Fluence at IRTm Determin. Chemistry Determin.

F Factor CF %Cu %Ni Material Heat No. 32 EFPY IRTc Lower SP-7056 6 . 82E+18 +30*' Plant 26 RG1.99 0.04*' O.90*'

Nozzle Specific Table 2 Belt Forging Upper 49D963-1-1/ 3 . 0 3E + 19 -30*' Plant 20 RG1.99 0.03*' O.71*'

Shell 49C904-1-1 Specific Table 2 Forging Lower 50D102-1-1/ 3 . 0 3 E+ 19 -30*' Plant 37 RG1.99 0.06*' O.75*'

Shell 50C97-1-1 Specific Table 2 Forging WF-645 H4498 6 . 82 E + 18 '*i -30*' Plant 41 RG1.99 0.03*' O.50*'

Upper Specific Table 1 Cire. Weld WF-562 442011 3 . 0 3E+19 +40*' Plant- 41 RG1.99 0.03*' O.65*'

Middle Specific Table 1 Cire. Weld 1084-18 < 1. 0 0 E+ 17 (*' -16*' Plant 54 RG1.99 0 . 0 4 *' O.60*'

hF-696 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 RTm 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.

M. J. DeVan Date: 7/1/94 77-1234176-00 Prepared By: Page 6 Reviewed By: L. B. Gross Date: 7/1/94 l

Table 2-4, Braidwood Unit 2 -- Data Sumarv for Upper-Shelf Enerov 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 5P-7056 A 508-2 10 9 '*) 4.09E+18* 128* Direct Nozzle Belt Forging Upper 49D963-1-1/ A 508-3 9 4 1.66E+19'd' 119'" Direct shell 49C904-1-1 j Forging Lower 50D102-1-1/ A 508-3 118 1.66E+19'*) 15 0 Direct Shell 50C97-1-1 Forging WF-645 H4498 Linde 80, 75

• 4.09E+18* 87

• Direct Upper SAW Cire. Weld WF-562 442011 Linde 80, 55

• 1.66E+19* 70

• Direct-Middle SAW Circ. Weld WF-696 1084-18 Linde 80, --- * <1.00E+17* 78

• Direct Lower SAW-Circ. Weld l

l Prepared By: M. J. DeVan Date: 7/1/94 77-1234176-00 Reviewed By: L. B. Gross Date: 7/1/94 Page 7

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 2, 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.

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.

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.

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' Prepared By: M. J. DeVan Date:

Date: 7/1/94 Page 8 f Reviewed By: L. B. Gross l

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s Table 2-5. Byron Unit 1 -- Data Summary for Pressurized Thermal Shock Calculation .

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IS Neutrc n Method of Method of j

Beltline Fluence at IRTm Determin. Chemistry Determin.

F IRTm Factor CF %Cu %Ni I Material Heat No. 32 EFPY l

Lower. 123J218 4 . 8 6 E+18 +30* Plant 31 RG1.99 0.05

• 0.72'*3 Specific Table 2 Nozzle Belt Forging Upper 5P-5933 2.159E+19 +40* Plant 19.14 Calculated 0.05

• 0. 73 * '

Shell Specific Forging Lcwer SP-5951 2.159E+19 +10* Plant 26 RG1.99 0.04* 0.64

• Shell Specific Table 2 Forging 4 . 8 6 E +18 0* Generic 41 RG1.99 0.03* 0 . 6 3 'b' WF-501 442011 Table 1 Upper Circ. Weld 442002 2 .15 9 E+19 -30* Plant 22.33 Calculated 0.03

• 0.46

• WF-336 Middle Specific Cire. Weld

< 1. 0 0 E+ 17 +10" Plant 164.65 RG1.99 0.23* 0.57 "

WF-472 31401 Table 1 Lower Specific Cire. Weld NOTES:

a. Fluence data are from WCAP-13880, " Analysis of Capsule X from the Commonwealth Edison Company Byron Unit 1 Reactor Vessel Surveillance Program," January 1994.

Chemical compositions and initial RTmy data for all materials are from the July 2, 1992 letter from b.

M. A. Jackson to T. E. Murley,

Subject:

Byron S'.ation, Units 1 and 2.

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7/1/94 77-1234176-00 Prepared By: M. J. DeVan Date:

Date: 7/1/94 Page 9 Reviewed By: L. B. Gross

Table 2-6. Byron Unit 1 -- Data Summary for Urner-Shelf Enerav Calculation .

1/4T Method of 1/4T USE Neutron Determin.

Beltline Material at 32 Fluence at Unirrad. Unirrad.

Material Heat No. Type EFPY 32 E7PY USE USE Lower 123J218 A 508-2 119 '*' 2.92E+18" 13 8 Direct Nozzle Belt Forging Upper SP-5933 A 508-2 111 1.179E+19'd' 138" Direct Shell Forging Lower SP-5951 A 508-2 12 0 1.179E+19' 15 0 Direct Shell I Forging WF-501 442011 Linde-80, 63

• 2 . 9 2 E+ 18 '8' 73'S' Direct Upper SAW Circ. Weld WF-336 442002 Linde 80, 60

• 1.17 9 E + 19 '*' 74'S' Direct Middle SAW Circ. Weld WF-472 314Ol' Linde 80, -- cl.00E+17'd' 72M' . Direct Lower SAW Circ. Weld Prepared By: M. J. DeVan Date: 7/1/94 77-1234176-00 Reviewed By: L. B. Gross Date: 7/1/94 Page 10

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. i NOTES FOR TABLE 2-6:

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. ,

f b. EOL USE values for the welds were calculated using Regulatory Guide 1.99, Revision 2, Figure 2, assuming the lower limiting value of 0.05% copper for welds.

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! c. EOL fluence is below the limits of the Figure 2 curves defined in Regulatory Guide 1.99, l I Revision 2.

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d. Fluence data are from WCAP-13380, " Analysis of Capsule X from the Commonwealth Edison Company l

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Byron Unit 1 Reactor Vessel Surveillance Program," January 1994.

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e. UUSE data for forgings 123J218 and SP-5951 are from WCAP-11651, " Analysis of Capsule U from the l

l Commonwealth Edison Co. Byron Unit 1 Reactor Vessel Radiation Surveillance Program," November-1987.

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f. UTTSE data for forging 5P-5933 is from the July 2, 1992. letter from M. A. Jackson to T. E.

2 Murley,

Subject:

Byron Station, Units 1 and 2.

g. UUSE data for the welds are from the November 19,199s letter from T. W. Simpkin to T. E. Murley, Byron Station Units 1 and 2, Response to Request for Additional Information Regarding NRC Generic Letter 92-01.

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t M. J. DeVan Date: 7/1/94 77-1234176-00 Prepared By: Page 11 .

Reviewed By: L. B. Gross Date: 7/1/94' 1

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Table 2-7. Byron Unit 2 -- Data Summarv for Pressurized Thermal Shock Calculation ,

a IS Neutron Method of Method of Fluence at IRTm Determin. Checistry Determin.

Beltline Factcr CF tCu tNi Material Heat No. 32 EFPY F IRTm Plant 31 RG1.99 0 . 0 5 - 0.74'*8 Lower 4P-6107 6 . 8 2 E+18 + 10

Specific Table 2 Nozzle Belt Forging

- 2 0 Plant 20 RG1.99. 0.01'*8 0 . 7 0

Upper 49D329-1-1/_ 2,192 E+19 '*' Specific Table 2 Shell 49C297-1-1 Forging 31 RGl'.99 ' O . 0 5 - 0.73 *) 8 Lower 49D330-1-1/ 2.192E+19 - 2 0 - Plant .

Table 2 49C298-1-1 Specific Shell Forging RG1.99 0.03'*8 0. 6 5 tel WF-562 442011 6 . 8 2 E+ 18 '*8 +40 '*8 ! . Plant 41 Specific Table 1 Upper Circ. Wald Plant 68 RG1.99 0. 0 5 '*8 -

'O.62*8 8 WF-447 442002 2.192E+19 + 10 -

Specific Table 1 Middle Cire. Weld

+ 4 0 Plant 144.4 RG1.99 0 .18 . ' O . 5 4 '*8 WF-614 31401 < 1. 0 0 E + 17

Specific Table 1 Lower j Circ. Weld NOTES:

a. Fluence data are from WCAP-12431, "Arnalysis of Capsule U from the Commonwealth Edison Company Byron Unit 2 Reactor Vessel Surveillance Program," October 1989.

b. Fluence data for weld WF-614 is from the January 17, 1986 letter from G. L. Alexander to H. R.

Denton,

Subject:

Zion Station Units 1 and 2; Byron Station Units 1 and 2; Braidwood Station Units 1 and 2; Pressurized Thermal Shock.

1992 letter from

c. Chemical compositions and initial RTm data for all materials are from the July 2, M. A. Jackson to T. E. Murley, S"bject: Byron Station, Units 1 and 2.

77-1234176-00 Prepared By: M. J. DeVan Date: 7/1/94 Page 12 Reviewed By: L. B. Gross Date: 7/1/94 5

Table 2-8. Byron 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 Nc. Type EFPY 32 EFPY USE USE Lower 4P-6107 A 508-2 131 '* ' 4 . 0 9E+18 '*' 155"' Direct Nozzle Belt Forging Upper 49D329-1-1/ A 508-3 117 '** 1.66E+19'** 149'" Direct Shell 49C297-1-1 Forging Lower 49D330-1-1/ A 508-3 9 9 1. 6 6 E + 19 127'55 Direct Shell 49C298-1-1 Forging WF-562 442011 Linde 80, 60

• 4 . 0 9 E +18 70

• Direct Upper SAW Cire. Weld WF-447 442002 Linde 80, 53

• 1.66E+19 67

• Direct Middle SAW Circ. Weld WF-614 31401 Linde 80i --- < 1. 0 0 E+17 74

• Direct Lower SAW Circ. Weld Prepared By: M. J. DeVan Date: 7/1/94 77-1234176-00 Reviewed By: L. B. Gross Date: 7/1/94 Page 13

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• l NOTES FOR TABLE 2-8:

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, 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-12845, " Analysis of Capsule U from the Commonwealth Edison Company Braidwood Unit 2 Reactor Vessel Surveillance Program," March 1991.

I e. Fluence data for weld WF-614 is from the January 17, 1986 letter from G. L. Alexander to H. R.  ;

Denton,

Subject:

Zion Station Units 1 and 2; Byron Station Units 1 and 2; Braidwood Station Units 1 and 2; Pressurized Thermal Shock. l

f. UUSE data for forgings 4P-6107 and 49D329-1-1/49C297-1-1 are from WCAP-12431, " Analysis of Capsule U from the Commonwealth Edison Company Byron Unit 2 Reactor Vessel Radiation i Surveillance Program," October 1989.

g. UUSE data for forging 49D330-1-1/49C298-1-1 is from the July 2, 1992 letter from M. A. Jackson to T. E. Murley,

Subject:

Byron Station, Units 1 and 2.

h. UUSE data for the welds 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.

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Prepared By: M. J. DeVan Date: 7/1/94 77-1234176-00 ,

Reviewed By: L. B. Gross Date: 7/1/94 Page 14 M

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PRESSURIZED THERMAL SHOCK AND USE TABLES FOR ALL PWR PLANTS NOMENCLATURE Pressurized Thermal Shock Table  !

Column 1: Beltline material location identification.  !

I Column 2: Beltline material heat 2. umber; 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 ,

tandem wire was used in the SAW process. l 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-Specific 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 I similar types.

l 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 i 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 4 when more than one value was reported. (Staff used the l average value in the latter case.) l Prepared By: M. J. DeVan Date: 7/1/94 77-1234176-00 Reviewed By: L. B. Gross Date: 7/1/94 Page 15 1

Upper-$helf Energy 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 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, this indicates that the unirradiated USE was from test data.

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Prepared By: M. J. DeVan Date: 7/1/94 77-1234176-00 Reviewed By: L. B. Gross Date: 7/1/94 Page 16 w