L-98-133, Forwards Response to Request for Addl Info Re GL 92-01,Rev 1,Suppl 1 for Reactor Vessel Structural Integrity

Forwards Response to Request for Addl Info Re GL 92-01,Rev 1,Suppl 1 for Reactor Vessel Structural Integrity

ML20236Q052
Person / Time
Site:	Beaver Valley
Issue date:	07/09/1998
From:	Jain S DUQUESNE LIGHT CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
GL-92-01, GL-92-1, L-98-133, NUDOCS 9807170288
Download: ML20236Q052 (20)

Text

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Beave Valley Power Station l Shippingport, PA 15077-0004

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i S r esident - Fax 4-Nuclear Power Divis6on July 9, 1998 L-98-133 U. S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555-0001 1

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l.

Subject:

Beaver Valley Power Station, Unit No. I and No. 2

! BV-1 Docket No. 50-334, License No. DPR-66 l BV-2 Docket No. 50-412, License No. NPF-73 Response to Request for Additional Information  !

Generic Letter 92-01, Revision 1, Supplement 1 l

! Reactor Vessel Structural Integrity The enclosed information is provided in response to a request for additional l - information (RAI) dated April 8,1998, regarding reactor pressure vessel (RPV) integrity l st Beaver Valley Power Station Unit Nos. I and 2. This response is a follow-up to

' earlier submittals by Duquesne Light Company (DLC)' on August 6,1996, and _

! supplemented on March 14,1997 and June 5,1997, in which DLC performed a revised pressurized thermal shock (PTS) analysis for the Unit I reactor pressure vessel per the requirements 'of ~ 10 CFR 50.61. This evaluation determined the PTS reference temperature (RTm) value for Unit I to be 264.5 F at end oflife (EOL). On October 7, 1997, the NRC sent a safety evaluation report (SER) on the PTS assessment for Unit 1  ;

that did not agree with the assumptions for fluence and margin term used in the DLC submittal. The NRC SER evaluation produced a slightly higher RTm value of 267.8 F i

l. for the limiting plate at EOL.

l .

Following these submittals, additional RPV information has become available from

. the Combustion Engineering Owners Group (CEOG) for vessel welds fabricated by  !

Combustion Engineering as documented in CEOG report, CE NPSD-1039, Revision 02.

As a follow-up to the previous submittals and the CEOG report, DLC has been requested  :

to assess the impact of revised best-estimate chemistries for vessel beltline welds and to consider the need to revise previous evaluations of RPV integrity. For comparison, this i response also includes the requested tables for Unit I using the assumptions from the

' NRC SER.

, DEllVERING Y C 9807t70288 9807o9 0UALITY

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DR ADOCK 05000334 I

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[., Beaver Valley Power Station, Unit No. I and No. 2

, Response to Request for Additional Information i Generic Letter 92-01, Revision 1, Supplement I Reactor Vessel StmeturalIntegrity Page 2 The enclosure provides the requested information and reanalysis for Beaver Valley _

Power Station Unit Nos. I and 2. Because the additional CEOG weld data does not

, affect the limiting plates in either vessel, the previous evaluations for PTS, PT limits and l overpressure protection remain valid for both units.

Sincerely,

~

~

J_ ^ "-

Sushil C. Jain

, Enclosure l

l l c: Mr. D. S. Brinkman, Sr. Project Manager Mr. D. M. Kern, Sr. Resident Inspector l Mr. H. J. Miller, NRC Region I Administrator i

i l

l l

l

9 Enclosure

, Beaver Valley Power Station, Unit Nos. I and 2 i

Response to Request for AdditionalInformation Generic Letter 92-01, Revision 1, Supplement 1 Reactor Vessel Structural Integrity l

Introduction:

The NRC issued Generic Letter (GL) 92-01, Revision 1, Supplement 1 (GL 92-01, ,

, Rev.1, Supp.1), " Reactor Vessel Structural Integrity" in May 1995. The purpose of this GL was to request licensees to identify, collect, and report any new data pertinent to the analysis of the structuralintegrity of their reactor pressure vessels (RPVs) and to assess the impact of those data on their RPV integrity analyses relative to the requirements of Section 50.60 of Title 10 of the .

Code of Federal Regulations (10 CFR Part 50.60),10 CFR 50.61, Appendices G and H to l 10 CFR Part 50 (which encompass pressurized thermal shock (PTS) and upper shelf energy

(USE) evaluations), and any potential impact on low temperature overpressure protection (LTOP) limits or pressure-temperature (PT) lhnits.

Since then, additional RPV information has become available from the Combustion Engineering Owners Group (CEOG) for vessel welds fabricated by Combustion Engineering as documented in a CEOG report [1] and a recently updated CEOG analysis [10]. As a follow-up to i the Generic Letter and the CEOG report, Duquesne Light Company (DLC) has been requested to )

assess the impact of revised best-estimate chemistries for vessel beltline welds and to consider the  ;

need to revise previous evaluations of RPV integrity. If there is no impact or if questions in this i request for additional information (RAI) do not apply for Beaver Valley Unit Nos. I and 2, the l NRC requests certification that previously submitted evaluations remain valid. j l

Background:

By letter dated August 2,1996, as supplemented March 14,1997 and June 5,1997, DLC submitted a reassessment of the PTS analysis for the Beaver Valley Unit 1 vessel [2]. A Safety Evaluation regarding PTS assessment for Unit I was issued on October 7, 1997 [3]. In that evaluation, the NRC staff did not concur with all aspects of the DLC analysis. However, the NRC staff did concur with the conclusion that the PTS reference temperature (RTrrs) value for the Unit I limiting material remains below the PTS screening criteria limit through end-of-life (EOL).

An evaluation of pressurized thermal shock for Unit 2 was performed in February 1997 [4].

This response analyzes the information in the CEOG report for the Unit I and Unit 2 vessel beltline welds to determine the impact on these prior vessel integrity evaluations.

Enclosure

.' Response to Request for AdditionalInformation Generic Letter 92-01, Revision 1, Supplement 1

, Reactor Vessel Structural Integrity l

Page 2

• l Response to NRC PAI Regarding Reactor Pressure Vessel Integrity for Beaver Valley Power Station Unit Nos. I and 2:

RAISection 1.0 Assessment of Best-Estimate Chemistry Based on the information in the CEOG Report CE NPSD-1039, Revision 2, "Best-Estimate Copper and Nickel Values in CE Fabricated Reactor Vessel Welds," and in accordance with the provisions of Generic Letter 92-01, Revision 1, Supplement 1, the NRC requests the following:

Reauest 1 An evaluation of the information in the reference above and an assessment ofits applicability to the determination of the best-estimate chemistry for all of your RPV beltline welds. Based upon this reevaluation, supply tle information necessary to completely fill out the data requested in Table 1 for each RPV l beltline weld material. A!so provide a discussion for the copper and nickel values chosen for each weld wire heat noting what heat-specific data were  !

I included and excluded from the analysis and the analysis method chosen for determining the best-estimate. If the limit;ng material for your vessel's PTS /PT limits evaluation is not a weld, include the information requested in Table 1 for j the limiting material also. Furthermore, you should consider the information l provided in Section 2.0 of this RAI on the use of surveillance data when  !

responding.

Response 1 Urut 1 There are three weld heats in the Unit I vessel beltline, 305424, 305414, and 90136.

Best-estimate copper and nickel chemistries for these weld heats were reevaluated using the ,

method recommended in the CEOG report (see Tables 4-1 through 4-3). A comparison with the i previously-reported vahas is shown below:

Reactor Vessel Weld and Previous Values [2i CECG Values tu Beltline R.egion Material Cu wt% Ni wt% Cu wt% Ni wt%

Interm. Shell Long Welds 19-714A/B Heat #305424, Linde 1092 T,9e Flux 0.263 0.632 0.273 0.629 Lower shell Long. Welds 20-714A/B Heat #305414, Linde 1092 Type Flux 0.338 0.606 0.337 0.609 Circumferential Weld Il-714 l Heat #90136, Linde 0091 Type Flux 0.278 0.071 0.269 0.070 l

l m

Enclosure I

.?- , Response to Request for Additional Information Generic Letter 92-01, Revision 1, Supplement 1 Reactor Vessel StaucturalIntegnty i Page 3 I

The revised best-estimate chemistries are included in Table 1-1. The revised chemistries are slightly different than the previously reported values. These differences are not significant and have no effect on the limiting plate. Updated Adjusted Reference Temperature values are computed for these welds along with the information and calculated values for limiting plate heat C6317-1. The surveillance data results were used to determine the chemistry factor (CF) value for the surveillance weld and plate materials. The results in Table 1-1 reflect the revised fluence and proposed margin term from the recent PTS nbmittal for Unit 1[2].

For comparison, the NRC best-estimate values from the recent safety evaluation report (SER) are presented in Table 1-2 [3]. The best-estimate chemistries in this table have been updated to include the CEOG recommended values [1]. There is no change in the RTm value for liuiting plate heat C6317-1.

Unit 2 There is one weld heat in the Unit 2 vessel beltline, heat no. 83642. Best-estimate copper and nickel chemistries for this weld heat were evaluated here using the method recommended in the CEOG report with a revision based on reanalysis of data validity. A comparison with the previously-calculated values is shown below:

Reactor Vessel Weld and Previous Values W Revised Values Beltline Region Material Cu wt% Niwt% Cu wt% Ni wt%

Axial and Circumferential Welds Heat #83642, Linde 0091 Type Flux 0.05 0.u7 0.047 0.085 The revised best-estimate chemistries are included in Table 1-3. The revised chemistries are slightly different from the previously reported values, and are different from the CEOG recommended value for this heat because the measured chemistries from the Unit 2 surveillance weld material (excluded from the CEOG analysis) were determined to be valid (see Table 4-4).

These differences do not affect the weld CF since the surveillance data results were used to determine the CF value for the surveillance weld material. Updated Adjusted Reference Temperature values are computed for this weld along with the information and calculated values forlimiting plate heat C0544-1.

MISecelon 10 Eval = dan ami Ure ofSurnillance Data 1.

I Since the evaluation of surveillance data rely on both the best-estimate chemical composition of the RPV weld and the surveillance weld, the information in the CEOG report may result in the need to revise previous evaluations ofRPV integrity (including LTOP setpoints and PT limits) per the requirements of 10 CFR 50.60,10 CFR 50.61, and Appendices G and H to 10 CFR Part 50.

Enclosure

? Response ts Request for Additional Information Generic Letter 92-01, Revision 1, Supplement 1 l

Reactor Vessel Structural Integrity Page 4 Based on this information and consistent with the provisions of Generic Letter 92-01, Revision 1, Supplement 1, the NRC requests the following:

Reauest 2 that (1) the information listed in Table 2, Table 3, and the chemistry factor from the surveillance data be provided for each heat of material for which surveillance weld data are available and a revision in the RPV integrity analyses j (i.e., current licensing basis) is needed or (2) a certification that previously submitted evaluations remain valid. Separate tables should be used for each heat of material addressed. If the limiting material for your vessel's PTS /PT limits evaluation is not a weld, include the information requested in the tables for the limiting material (if surveillance data are available for this material).

Resnonse 2 Unit 1 DLC believes that the previously submitted suneillance results for Unit I remain valid. The results for surveillance weld heat 305424 are given in Tables 2-1 and 3-1. The results for surveillance plate heat C6317-1 are given in Tables 2-2 and L2. These tables include the utility best-estimate fluence values and TANIi curve-fitted results (2].

Because of differences in the NR>." staffs evaluation of fluence [3], the results for surveillance weld heat 305424 according to the SER are given in Tables 2-3 and 3-3. The results for surveillance plate heat C6317-1 using NRC best-estimate fluences are given in l Tables 2-4 and 3-4.

Unit 2 The results for suneillance weld heat 83642 are given in Tables 2-5 and 3-5. This is the same information that was presented in the PTS evaluation for Unit 2 [4]; DLC believes l that these suncillance results remain valid.  !

RMSecdon 3.0 PTS /PTLimit Evaluadon Reauest 3 If the limiting material for your plant changes or if the adjusted reference l temperature (ART) for the limiting material increases as a resuh of the above 1 evaluations, provide the revised RTrrs value for the limiting material in j accordance with 10 CFR 50.61. In addition, if the adjusted RTwr value l increased, provide a schedule for revising the PT and LTOP limits. The l schedule should ensure that compliance with 10 CFR Part 50, Appendix G, is tesmiestmi,ned,

Enclosure

? Response to Request for Additional Inform:. tion Generic Letter 9241, Revision 1, Supplement 1 Reactor Vessel Structural Integrity Page5 '

Resnonse 3 {

Based on this evaluation, the vessel weld chemistry factors and ART values are only slightly changed by the revised best-estimate chemistries, and no new information has become '

available on the limiting plate materials of either Unit's reactor vessel. Therefore, the previously submitted analyses for PTS, PT limits and LTOP remain valid for Beaver Valley Unit Nos. I and 2.

References:

[1] "Best Estimate Copper and Nickel Values in CE Fabricated Reactor Vessel Welds,"

Prepared for the C-E Owners Group, CE NPSD-1039, Revision 2, June 1997.

[2] " Evaluation of Pressurized Thermal Shock for the Beaver Valley Unit i Reactor Vessel," i Westinghouse' Electric Corp., WCAP-14543, June 1996.

[3] Safety Emluation Reganiing Pressurized 7hermal Shock (PTS) Assessment for Beaver

' ValleyPower Station, Unit No.1, US Nuclear Regulatory Commission, October 7,1997.

[4' . Evaluation of Pressurized Thermal Shock for Beaver Valley Unit 2," West'mghouse Electric Corp., WCAP-14784, Rev. 2, Februaiy 1997.

[5] " Analysis of Capsule V from the Duquesne Light Company Beaver Valley Unit No.1 J Reactor Vessel Radiation Surveillance Program," Westinghouse Electric Corp., WCAP-9860, January 1981. ,

[6] " Analysis of Capsule U from the Duquesne Light Company Beaver Valley Unit 1 Reactor Vessel Radiation Surveillance Program," Westinghouse Electric Corp., WCAP-10867, September 1985.

[7] " Analysis of Capsule W from the Duquesne Light Company Beaver Valley Unit 1 Reactor Vessel Radiation Surveillance Program," Westinghouse Electric Corp., WCAP-12005, November 1988.  !

[8] " Analysis of Capsule U from the Duquesne Light Company Denver Valley Unit 2 Reactor Vessel Radiation Surveillance Program," Westinghouse Electric Corp., WCAP-124%,

September 1989.

[9] " Analysis of Capsule V from the Duquesne Light Company Beaver Valley Unit 2 Reactor Vessel' Radiation Surveillance Program," Westinghouse Electric Corp., WCAP-14484, l February 1996. i l

[10] ." Updated Analysis for Combustion Engineering Fabricated Reactor Vessel Welds Best i Estimate Copper and Nickel Content," Prepared for the C-E Owners Group, CENPSD-1119, Revision 00, Draft Report, June 1998.  !

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Enclosure Response to Request for AdditionalInformation Generic Letter 92-01, Revision 1, Supplement 1 Reactor Vessel Structural Integrity

• Page'9 Facility: Beaver Vallev 1 Table 2-1: Weld Heat 305424 CapsuleID Cu Ni Irradiation Fluence Measured Data Used in (including Temperatum 2 (x1028n/cm ) ARTuur Assessing Vessel source) (*F) (*F) (Y or N)

VI51 0.26 0.62 546 0.3160) 157.80) Y' I UM1 0.26 0.62 546 0.690n) 164.40) Y WM 0.26 0.62 546 0.9150) 185.60) Y CF fitted = 196.4 'F Table 3-1: Weld Heat 305424 Capsule ID Cu NI Irradiation Fluenm Measured Adjusted Predicted (Adjusted -

(including Temperature Factor ARTwor ARTwer ARTuur Prodicted) source) (O (F) ff) (F) ARTNur (F)

~

Visi 0.26 0.62 546' O.684 157.8. 157.8 134.3 23.5 UM1 0.26 0.62 546 0.8 % . 164.4 164.4 175.9 -11.5 4

WR 0.26 0.62 546 0.975 185.6 185.6 191.5 -5.9

") capsule fluences determined from Beaver Valley 1 PTS evaluation [2]

C) measured ARTurrvalues from TANH curve-fit [2]

l l

\ _ _ . _____-_______________ - - -

Enclosure

. Response to Requ:st for AdditionalInformation Generic Letter 92-01, Revision 1, Supplement 1 Reactor Vessel Structural Integrity

-' Page'10 Facility: Beaver Valley 1 Table 2-2: Plate Heat C6317-1 Capsule ID Cu Ni Irradiation Fluence Measured Data Used in (including Temperature 2 (x1018n/cm ) ARTNur Assessing Vessel source) (*F) (*F) (Y or N)

VrLM 0.20 0.54 546 0.316m 137.8m Y Vorm 0.20 0.54 546 0.316m 128.2m Y.

Urt i 'l 0.20 0.54 546 0.690m 131.8m Y UtrM 0.20 0.54 546 0.690m 118.9m Y Wnm 0.20 0.54 546 0.915m 179.9m Y Worm 0.20 0.54 546 0.915m 147.8m Y CF fitted = 163.4 F Table 3-2: Plate Heat C6317-1 Capsule ID Cu Ni Irradiation Fluence Measured Adjusted Predicted (Adjusted -

(induding Temperature Factor ARTwar ARTNor ARTNur Predicted) source) (*F) (*F) (*F) (*F) ARTNDT ( F)

VnN 0.20 0.54 546 0.684 137.8 137.8 111.7 26.1 i

VtrW 0.20 0.54 546 0.684 128.1 128.1 111.7 16.4 j UrtN 0.20 0.54 546 0.8% 131.8 131.8 146.4 -14.6 UtrN 0.20 0.54 546 0.8% 118.9 118.9 146.4 -27.5 ,

4 Wnn 0.20 0.54 546 0.975 179.9 179.9 159.3 20.6  ;

l Wtin 0.20 0.54 546 0.975 147.7 147.7 159.3 -11.6 L

l W capsule fluences determined from Beaver Valley 1 PTS evaluation [2]

• measured ARTmyr values from TANH curve-fit [2]

L--

Encl sure

.'. Response to Request for Additional Information Generic Letter 92 01, Revision 1, Supplement 1 Reactor Vessel Structural Integrity

• Page'l1 Facility: Beaver Valley 1 Table 2-3: Weld Heat .305414.

Capsule ID Cu Ni Irradiation Fluence Measured Data Usedin (including Temperature (x101' n/cm2) ARTNor Assessing Vessel source) (*F) (*F) (Y or N)

V 0.26 0.62 546 0.3400) 157.8 Y U 0.26 0.62 546 0.6880) 164.4 Y W 0.26 0.62 546 1.0580) 185.6 Y CF fitted = 191.92 'F Table 3-3: Weld Heat 305424 CopsuleID Cu Ni Irraaation Fluenm Measured Adjusted Predicted (Adjusted-(including Temperature Factor ARTNur ARTNur ARTNur Predicted) source) (*F) (*F) (*F) (*F) ARTuur (*F)

V 0.26 0.62 546 0.703 157.8 157.8 134.9 22.9 U 0.26' O.62 546 0.895 164.4 164.4 171.8 -7.4 W 0.26 0.62 546 1.016 185.6 185.6 194.9 -9.3

( ) capsule fluences determined from NRC SER Regarding PTS Assessment for BV-1 [3]

_ - _ . - - - _ _ _ _ - . . _ - - _ _ _ _ _ _ _ _ _ _ _ - .. w

Enclosure

..L Response to Request f;r Additional Information Generic Letter 92-01, Revision 1, Supplement 1 Reactor Vessel Structural Integrity

• Page'12 Facility: Beaver Vallev 1 Table 2-4: Plate Heat C6317-1 Capsule ID Cu Ni Irradiation Fluence Measured Data Used in (including Temperature 2 (x101' n/cm ) ARTm Assessin6Vessel source) (*F) (*F) (Y or N)

Vn. 0.20 0.54 546 0.3400) 137.8 Y Vtr 0.20 0.54 546 0.3400) 128.1 Y Un. 0.20 0.54 546 0.6880) 131.8 Y Ut.r 0.20 0.54 546 0.6880) 118.9 Y W n. 0.20 0.54 546 1.0580) 179.9 Y Wtr 0.20 0.54 546- 1.0580) 147.7 Y CF fitted = 159.9 F Table 3-4: Plate Heat C6317-1 Capsule ID Cu - Ni Irradiation Fluence Measured Adjusted Predicted (Adjusted -

(including Temperature Factor ARTm ARTm ART m Predicted) source) (*F) (*F) (*F) (*F) ARTm (*F)

Vn. 0.20 0.54 546 0.703 137.8 137.8 112.40) 25.4 Vor 0.20 0.54 546 0.703 128.1 128.1 112.40) .15.7 Un. 0.20 0.54 546 0.895 131.8 131.8 143.10) -11.3 Utr 0.20 - 0.54 546 0.895 118.9 118.9 143.10) -24.2 W n. 0.20 0.54 546 1.016 179.9 179.9 162.40) 17.5 i

Wtr 0.20 0.54 546 1.016 147.7 147.7 162.40) -14.7 "3 determined from NRC SER Regarding PTS Assessment for BV-1 [3]

I h

b

l.

Enclosure l . Response to Request for Additional Information i Generic Letter 92-01, Revision 1, Supplement I  !

Reactor Vessel Structural Integrity

• Page'13 Facility: Braver Vallev 2 Table 2-5: Weld Heat 83642 l

Capsule ID Cu Ni Irradiation Fluence Measured Data Used in (br.luding Temperature (x10" n/cm2) .iRTer Assessing Vessel terxe) (*F) (*F) (Y or N)

UM 0.08 0.07 546 0.6010) 3B2) Y f VM 0.08 0.07 546 2.640) 25.50) Y CF fitted = 15.2 F l

Table 3-5: Weld Heat 83642 l Capsule ID Cu Ni Irradiation Fluence Mansured Adjusted Predicted (Adjusted -

(including Temperature Factor ARTer ARTwr ARTer Predicted) source) (*F) (*F) (*F) (*F) ARTer (*F)

UM 0.08 0.07 546 0.857 3.6 3.6 13.0 -9.4 VM 0.08 0.07 546 1.26 25.5 25.5 19.1 6.4

") capsule fluences determined from Beaver Valley 2 PTS evaluation [4]

(23 measured ARTmyr values from TANH curve-fit [4]

.D

,. Enclosure .  !

Response to Request for Additional Information l Generic Letter 92-01, Revision 1, Supplement 1

, . Reactor Vessel Structural Integrity

. Page 14

' Table 4-1: Evaluation of Chemistry for Weld Wire Heat No. 305424 3s5424 Jos. sees Type sies new muchmeios ce se sta Les Mugher 3000 Au Co Au PN BoL # at ut% wen CE955003E5 l OOURCEla l NEDORT/AMALYS5 osee Tao we% we% Caes crue s asse eme n ,

0.m G e4 VADD l WDCe# l DB180 e l oree a e.aes I sme I  : l 0 30 _0 se . ,_w uD 1 WDC46s0 1 Ones b 1 osee e e.Ses I asan I e l J 0 36 0 62 VAUD WDC4651 WCAb100s7 o en O m? vAuD WDC4ss2 WCAbl0Off e 0 31 Oml VAUD WDC-183p MET 46-305 e 0.31 0 $34 VAUD WDC 1840 MBTh305 e O s2 Osu vAUD WDC48:1 herT46-303 o Ost asia vAuD Winspes narT4s-203 e 01os 0 as vAUD WDC-ISO MFT46-M e 019$ 0 St VAUD WDC-IS44 MET 46-303 a 0 240 0 61 VAUD WDC4841 - WTed 203 a 0 34s 0.00 VAUD WDC 1846 METM305 e 02w O sse vAUD WDC4847 MET 46-303 e 0 234 0 991 YADO WDCISS MET 46-203 e 0 341 O e04 VAUD WDMSW MFT46-303 e O 241 0 SSB VAUD WDC-lWO latTM303 e

]

0.23s O sei vAuD WDC-lost *sT4s-3e e j 02M 0 194 VAUD WDC-1892 MET 46233 e 0 233 0 001 YAUD WDC-1953 MET 46-303 e f

0.235 0 Sp VAUD WDC4354 MET 46-303 e j

0 235 0 sep VAUD WDC4055 MBT 4305 e i 0 25$ 0 Ses VAUD WDC-1856 MET 46-303 e  !

O 391 0 Spl VAUD WDC-Its? MET 46-30$ e I 0 295 0 613 VAUD WDC-teSS kaufT46-10B e 0 333 0 68 VAUD WDC-1859 - 68748-28 e 0 39 A NS VAUD WDC4000 MET 46-303 o 0W 0 617 VAUD WDC4061 MET 48-305 e 0 302 ~ 0 424 YAUD WDC-1882 MET 40-309 e 9 300 0 436 VAUD WDC4883 hET46-20S e 03 O SS VAUD WDC-1964 ,

DEIT46-305 e 0 34 0 807 VAUD WDC-10$$ ,, nST48-303 e 0.23p e 804 VAUD WDC-1988 MET 48-20B e 0 346 Os VAUD WDC4857 h5T430S o 0 134 O ses VAUD WDC-18EB MET 46-303 e anos e . sowse.Weemiss Auer. 1 (se Weese s Am As On 75  ; Ce PG Ce fle SJ55 8.80 . 0373 4.415 ,

E305 S.480 l I

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. A e

, Enc 1ssure Response to Request for Additional Information Generic Letter 92-01, Revision 1, Supplement 1

. Reactor Vessel Structural Integrity )

Page 15 Table 4-2: Evaluation of Chemistry for Weld Wire Heat No. 305414 3ss414 IJmes test Type Pts New Evatmamm l Cu M Shslad Neuber and? As Cu Au Mt Est e of we% we% CEFEBeDREE l SDURCE LEL l REPORTsANALYes Creep Ten we% we% Cass aeen, e uM e.mo i 0.D 0 Se VAUD l WDC4el3 l 12320 e crew b , eJoe I e.ses l  : 1 0 se O Se vAUD WDC46M TR444CM402 b 0 30 0 Se VAUD WDC4637 TRC4dCM-e02 b 0 30 0 00 VAIJD WDC463e TRO4dCW.002 b 0 30 0 en VAMD WDC4630 TROMCM402 b 6M 0 GO VAUD WDC4640 M44dCM402 b 0 31 0 00 VAUD WDC464) TR440CW.002 b 0 t' 0 60 VAUD WDC4642 TR44dCM402 b D .1 c el VAIJD WDC4643 TR44dCM402 b 0 30 0 61 VAUD WDC4644 MoMCM402 b 0 31 0 61 YAUD WDC4649 TROMCM402 6 O SS 0 65 -VAUD WDC4647 IEsde b 0 35 0 00 VAUD , WDC464e IMD49 b cese e eJee 1 e.ees 1 I OM 06 VAUD l WDC4634 l 13923 e '

ceem e eJes I e.eae 1  : 1 0 35 0 64 VAUD l WDC46BS l Dess e Q SampleWsWess Aw CesWeMass Ageren At s Co M Co M C. x e.u, un A. C. A. M eJt7 e.As Table 4-3: Evaluation of Chemnistry for Weld Wire Heat No. 90136 ,

I sol 36 IJmes tese Type Peu New Deutsamms Co - fe Vestems plum Ems Ne$se As Ce Am MI Est e of we% ush G IBROGREE I 80tillCE LE REpoRTiANALYeS Geese Tag we% we% CoAs Osse e SJe8 0 30 VAUD 1 W3C-lO4 l INF3 e (sem b eJPe 7 I i l 0 30 VAUD ] W ml435 l imer b Graus e

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GJ l 1 8 l VAUD WDC-t 579 DOM e 0 16 VAUD' VA!JD WDC-ISe0 WDC-1182 D 0997 D'03 8 a

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~

0 16 VAUD WDC4h3 D OM6 e

~

0 17 VAUD WDC-IS84 D10389 e 0 le VAUD WDC-1 $$$ D OM2 e e le VAUD WDC- 506 DOM 4 e 0 20 VAUD W30- 557 D0961 e (eem s eJee l l t _j 0 30 VAuD i Wm une i - , . e (ses po 8.316 l 4A1e i I I O 22 0M VAJD W3C- G6 Die 646 e ,

0 21 eM VAJD E 3C- 47 Dies 30 e  :

Gene f 834 1 I t I e 37 VAUD 1 wpG-15e l DisDl4 f I Geno m eJes 1 I t I

( 0 32 VAUD 1 WUC-15ee l UIU1B a ~

[ ereas b eJee 1 l l 1 eM l VAUD I WDC-ISeo I Diasp h l l

1 i

l Enclosure Response to Request for AdditionalInformation i Generic Letter 92-01, Revision 1, Supplement 1 Reactor Vessel StmeturalIntegrity l Page 16 l {

Table 4-3: Evaluation of Chemistry for Weld Wire Heat No. 90136 (cont.)

90136 tJue,see Type nas msw paiemenen Os PG Mas 14 Namsher 3947 Aug Co Ave. NI Est. # at

_ we% we% Cf PEDIGREE l SOL'IICE I.R l ktPORTIANALYS5 Grew Tam we% we% Cess Grung j e.3437 3.gseg 3 0 20 0 05 VAUD WDC-1467 114-MCM 004 1 0 20 0 05 VAUD WDC-1448 TRS MCM-0D4 6 0 30 0 05 VAUD WDC-1449 TRf-MCM4lD4 j I 0 21 OW VAUD WDC-1470 TRJ-WCM4104 j 0 22 1 005 VAUD WDC-1471 TRf-WCM41De l 0 22 06S VAUD WDC-1472 TR#-MCM4lD4 j j 0 197 W VAUD WDC-1473 WCAP 12751 1 j 0 21 0 06 VAUD WDC-1414 TR4-MCM-004 1 0 22 0 06 VAUD WDC 1475 TR4 4eJ4-004 j 0 22 0 06 VAUD WDC-1476 TRf-MCM4]D4 i 0 32 0 06 VAUD WDC 1477 TRJ MCM4104 1 0 32 0 06 VAUD WDD 1478 TRS-MCM-004 i l 0 44 0 06 VAUD WDC-1479 TRf-MCM-004 j 0 20s 1 06 VAUD WDC-1400 WCAP 12751 1 0 238 0 07 VAUD WDC-14sl WCAP 12751 j 0 341 0 07 VAUD WDC-1482 WCAP-12751 1 0 23 0 11 VAUD WDC-1484 Diod86 )

0.25 0 06 VAUD WDC-14B* D10GB7 1 0 20 0 06 VAUD WDC 1490 D64374 1 0 22 0 06 VAUD WDC 1491 Ds4394 j 0 23 0 06 VAUD WDC-14F2 D44395 j -

0 23 0 06 VAUD WDC-1493 D64396 I l 0 23 0 06 VAUD WDC-1494 D44398 1 0 24 0 06 VAUD WDC-1493 D44397 0 32 0 06 VAUD WDC-1496 D44310 0 15 0 07 VAUD WDC-1497 D44379 0 16 0 07 VAUD WDC-1498 D44379 1 0 18 0 07 VAUD WDC-1499 D64377 j 0 15 0 07 VAUD WDC-IS00 D44380 1 0 19 0 07 VAUD WINJ-lM1 D64392 i 0 20 0 07 VAUD WDC-IS02 D44376 1 0 20 0 07 VAUD WDC_lS4 Dt4390 1 0 20 0 07 VALID WDC-1904 D64391 t 0 30 0 07 VAUD WDC-I SOS D443P3 t 0 23 0 07 VAUD WDC-IS06 D64375 j 0 23 0 07 VAUD WDC-1$07 Ds43e6 6 0 29 0 07 VAUD WDC-1SGB D44346 i 0 29 0 07 VAUD WDC-1909 Ds4101 i 0.3 6.01 VAUD WDC-ISIO D64349 1 0 31 0 07 VAUD WDC-ISil D64382 1 0 32 0 05 VAUD WDC-IS12 D44371 }

0 23 0 0D VAUD WDC-IS13 D64387 1 j 0.29 0 00 VAUL, WDC-ISid D64364 j l 03 0 00 VAUD WDC-lSIS D64MS l 0 32 0 00 VAUD WDC-lS16 D64372 0 23 010 VAUD WDC-Isl1 De4384 0 23 0 :0 VAUD WDQl3 D64381 1 0 29 0 0 VAUD WDC-ISl9 De4MB j

, 0 23 0 t VAUD WDC-1320 D6435 j

__0 24 0$ VAUD WDC-IS21 D64363 1 42 0 06 VAUD WDC-IS26 DsetED j 0 22 0 06 VAUD WDC-IS27 D44501 f 0 22 0 06 VAUD WDC-IS2B D64932 j 0 22 0 06 VAUD WDC-1$29 D440B4 1 0 22 0 06 VAUD WDC-1S30 D49004 1 0 23 0 06 VAUD --W )C-IS$1 Deess3 i 0 23 0 06 VAUD W)C-IS$2 DaS003 1 0 21 0 06 VAUD W )C-I S33 DeSO49 1 0 23 0 86 VAUD WDC-IS34 D450SO 1 l

?

, Enclosure Response to Request for Additional Information Generic Letter 92-01, Revision 1, Supplement 1

. Reactor Vessel St;ucturalIntegrity Page 17 Table 4-3: Evaluation of Chemnistry for Weld Wire Heat No. 90136 (cont.)

90ln IJm6e test Typs Hus New Eveimonen (a NI Mus IAt Nimmber 3947 Avs. Co Asa M1 Esf 8 et we% es% ' g PEDIGREE SDURCE I.E REPORTIANALYSIB Grew Tag we% we% Cots 0 24 0 06 ,,

VAUD WDC-ISSS des 003 1 0 24 0 06 VAUD WDC 1536 des 004 1 0 24 0 06 VAUD WDC-t S37 D45020 i 0 24 0 06 VAUD WDC-1535 D45028 j 0 24 0 06 VAUD WDC-1939 des 029 i 0 24 0 06 VAUD WDC-1540 D45029 i O!1 0 06 VAUD WDC-IS41 D85049 1 0 24 0 06 VAUD WDC-1$42 De5010 1 0 15 0 07 VAUD WDC IS43 DN980 j O ll 0 07 VAUD WDC-IS44 D45000 1 0 16 0 07 VAUD WDC-1349 des 000 1 0 18 0 07 VAUD WDC-IS46 DN970 i 0 18 0 07 VALID WDC-1947 D44979 1 0.19 0 07 VAUD WDC-154W D44978 j 0 19 0 01 VAUD WDC-1549 D49002 1 0 19 0 07 VAI.lD WDC-1990 DeSO45 j 0 20 0 07 VAUD WDC-ISSI DN977 1 0 20 0 07 VAUD WDC-IS$2 D89001 j 0 20 0 07 VAUD Wr-1953 des 00: 1 0 20 0 07 VAUD W3C-14e D49FM 1 0 20 0 07 ~

VAUD WX'-IS$$ D49027 1 0 20 0 07 VAUD Wr-ISS6 DeSee? i 0 20 0 07 VALID WE-l S S1 De9048 1 0 21 0 07 VAUD Wr ISSS des 001 i 0 24 0 07 VAUD W 3C-IS$9 DeSO26 j 02% 0 01 VAUD WDC-IS60 des 026 1 0 21 0 07 VAUD W3C-IS61 DeSO47 i 0 39 0 07 VAUD W3C-IS62 , DN971 0 38 0 07 vAuD W )c-ISO i Dae72 0 13 00s VAUD WDC-IS64 D49046 0 16 0 Os val y W r-1%$ fx1025 0 16 0 08 VAUD Wr sS66 D49025 0 16 0 03 VAUD W $\l567 DeSO46 0 23 0 10 VAUD WDC-ISS DNeid 0 24 0 10 VAuD WDC-ISe 944P75 0.38 VAUD WDC-IS72 D1025$

Cave e SJ2e l l 1 l

- 0 32 VAUD l WDC-ISW1 [ Ulurw I cree e e.se l l 1 l 0 16 VAus l WDC-tw2 l DiawS m Gene o e.ase 1 l t 1 0 15 VAUD l wuc-IS7S l U50er2 m Gree e 4 968 l l t l 0 36 VAIJD l WDC-1574 l U15p13 e Gree p SJte  ! I a j 0 31 VAUD I WDC-IS75 1 DIO14 p Grau e e.tst l l 4 1 0 15 VAUD I WDC-IS76 i Dies 22 e Ske pts SammteWeW Aw Cd WeW Aserey As s. os is ou pe

c. " T """ e.an ' e.m em eve MI1"7. Om l

l o

. Encl::sure I Response to Request for Additional Information Generic Letter 92-01, Revision 1, Supplement 1

. Reactor Vessel Structural Integrity Page 18 l

Table 4-4: Evaluation of Chemistry for Weld Wire Heat No. 83642 83642 unde see Type n.: New r h.se.

c. re nur 34s Namber seM Avg ce A4 Ni est. e af )

- e c erDu:we I soenceta i urosmmves G, r.s e e cwe G, s- e ees e.e= i

_ 0 08 0 07 INVAUD* WDC-1175 WCAP 1240s s' Group a e.ees l S.ese l t l 0 05 0 06 VAUD WDC-1196 D23932 s 0 00 0 07 VAUD WDC 1197 l D23S49 s Greep b e.e43 l e.133 l 3 l 0 04 0 19 VAUD WDC-1176 Dl4234 b 0 04 0 06 VAUD WDC-1177 Dl4235 b 0 05 0 16 VAUD WDC-Il78 D14236 b Greep a r. 940 j e.870 l 1 l 0 04 0 04 VAUD WDC-Il19 D14925 e 0 04 01 VAJJD WDC 1100 D14926 e Gesey e e.ees l e ses l t l 0 0$ 0 04 VAUD l WDC-lle2 l WCAP 9228 e Group i e.eJ7 l 4.40 l t l 0 04 0 06 VAUD WDCII4 Dl64 S9 f 6 01 0 06 VAUD WDC-lle$ Dl6460 f 0 04 0 01 VAUD WDLallV, D14461 f Geear s 4.ees I a ees I a 1 0 04 0 00 VAUD WDC llt? D137e4 g 0 04 0 09 VALID, WDC-Ilse D13783 g cree, b e.ede I e se7 i n l 0 05 0 07 V Al.lD WDC-lie 0 Dl41el h 0 04 0 00 VAUD WDC-ll90 D141$ 8 h 0 01 Ol VAUD WDC-Il91 Di m b 0 01 0 11 VAUD WDC-Il92 D23e%8 h 0 03 0 12 VAUD WDC-1191 D23ee$ h 0 03 0 07 VAUD WDC ll94 Dl41&2 h 0 03 0 13 VAUD WDC-119$ D14132 h Greup a 4.e44 l l 8 l 0 04 VAUD j W1X'-119e l Dl3139 i Nuta mammred data kan flueve Valley 3 sovellenes weld desernmand to be wahd duen Stampes Smauple teW Aurer Ces wow Average A4 Ce N1 ou M Ce PIB 4A44 e.e77 4 ede e.e77 4 447 4.ees i

I

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