ML20094L811
| ML20094L811 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 11/14/1995 |
| From: | Stetz J CENTERIOR ENERGY |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| 2343, GL-92-01, GL-92-1, NUDOCS 9511210137 | |
| Download: ML20094L811 (5) | |
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- " CENTERf0R 4l@
ENERGY 300 Modison Ausnue John P. Slett Toledo, OH 43652M1 yice p,eeldent - Nuclear 419-249 2300 m.go Docket Number 50-346 License Number NPF-3 Serial Number 2343 November 14, 1995 United States Nuclear Regulatory Commission Document Control Desk Washington, D. C. 20555-0001
Subject:
Response to Generic Letter 92-01, Revision 1, Supplement 1,
" Reactor Vessel Structural Integrity", for the Davis-Besse Nuclear Power Station Ladies and Gentlemen:
This letter provides Toledo Edison's (TE) response for the Davis-Besse Nuclear Power Station, Unit 1 (DBNPS), to Nuclear Regulatory Commission (NRC) Generic Letter 92-01, Revision 1, Supplement 1, " Reactor Vessel Structural Integrity," dated May 19, 1995. By letter dated August 16, 1995, (TE Serial Number 2314), Toledo Edison provided its initial response to Part 1 of the Generic Letter.
In the response, Toledo Edison identified its support of the B&W Owners Group Reactor Vessel Working Group efforts to obtain additional data which could be relevant to the determination of reactor pressure vessel (RPV) integrity, and provided acceptance of B&W Owners Group report BAW-2257, " Response to Part (1) of Generic Letter 92-01, Revision 1, Supplement 1",
submitted to the NRC by letter dated August 1, 1995 (reference: OG-95-1527).
The Generic Letter required the following information also be provided to l
the NRC:
l.
"(2) an assessment of any change in best-estimate chemistry based on i
consideration of all relevant data; (3) a determination of the need for use of the ratio procedure in accordance with the established Position 2.1 of Regulatory Guide 1.99, Revision 2, for those licensees that use surveillance data to provide a basis for the RFV integrity evaluation; and j
i 9511210137 951114 I
PDR ADDCK 05000346
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._tVQi oprating companies-t Cleveland Elecinc illuminating Toledo Edison
Docket Number 50-346 License Number NPF-3 Serial Number 2343 Page 2 (4) a written report providing any newly acquired data specified above and (1) the results of any necessary revisions to the evaluation of RPV integrity in accordance with the requirements of 10 CFR 50.60, 10 CFR 50.61, Appendices G and H to 10 CFR Part 50, and any potential impact on the LTOP or P-T limits in the technical specifications or (2) a certification that previously submitted evaluations remain valid. Revised evaluations and certifications should include consideration of Position 2.1 of Regulatory Guide 1.99, Revision 2, as applicable, and any new data."
Toledo Edison has accepted the attached B&W Owners Group Topical Report BAW-2257, Revision 1, " Response to Generic Letter 92-01. Revision 1, Supplement 1", submitted to the NRC by letter dated November 1, 1995, (reference OG-95-1552). This response was prepared by the B&W Nuclear Technologies Company for the B&W Owners Group Reactor Vessel Working Group; provides responses for Parts 2, 3 and 4; and completes the response for Part 1.
Toledo Edison also certifies that previously submitted DBNPS reactor vessel integrity calculations remain valid.
These evaluations include low temperature overpressure protection (LTOP) and pressure-temperature (P-T) limit curves in the Technical Specifications (reference: DBNPS License Amendment 199 dated July 20, 1995) and previous responses to GL 92-01 (reference: Toledo Edison Serial Number 2060 dated July 1, 1992, and Serial Number 2233 dated June 30, 1994).
Should you have any questions or require additional information, please contact Peter W. Smith, acting Manager - Regulatory Affairs, at (419) 321-7744.
Very truly yours,
/
DHL i
Enclosure Attachment cc:
L. L. Gundrum, DB-1 NRC/NRR Project Manager H. J. Miller, Regional Administrator, NRC Region III S. Stasek, DP-1 NRC Senior Resident Inspector Utility Radiological Safety Board 4
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Docket Number-50-346 i
License Number NPF-3 Serial Number 2343 Enclosure i
Page 1 4
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RESPONSE TO GENERIC LETTER 92-01 REVISION 1, SUPPLEMENT 1 i
FOR e
DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 i
j-This letter is submitted in conformance with Section 182a of the Atomic j
Energy Act of 1954 as amended, and 10CFR50.54(f). Enclosed is Toledo j
Edison's Response to Generic Letter 92-01, Revision 1, Supplement 1, Reactor Vessel Structural Integrity,-for the Davis-Besse Nuclear Power i
Station.
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4I By:
I John P [Stetz, Vice'PreVident - Nuclear l
Sworn to and subscribed before me this 14th day of November, 1995.
Q s.
Notary Publ',
tate of Ohio e
LORI J. STRAUSS j
Notary Public. State of Ohio My Commission Empires 3/22/98 1
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Docket Number 50-346 License Number NPF-3 Serial Number 2343 Attachment Topical Report BAW-2257, Revision 1 B&W Owners Group Reactor Vessel Working Group Response to Generic Letter 92-01, Revision 1, Supplement 1 l
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MP-95-02600 BAW-2257 Revision 1 October 1995 hSbbh!$W:..l-lS$
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"aOWNERS GROUP
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B&W Owners Group Reactor Vessel Working Group l
Response to Generic Letter 92-01, l
Revision 1, Supplement 1 i
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'B&WNUCLEAR
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' TECHNOLOGIES I
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B&W Owners Group Reactor Vessel Working Group l
Response to Generic Letter 92-01, Revision 1, Supplement 1 by M. J. DeVan BWNT Document No. 43-2257-01 (See Section 5 for document signatures.)
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Prepared for B&W Owners Group Reactor Vessel Workine Group Conunonwealth Edison Company Duke Power Company Entergy Operations, Inc.
Florida Power Corporation Florida Power & Light Company i
GPU Nuclear Corporation Rochester Gas and Electric Corporation Toledo Power Company Virginia Power Wisconsin Electric Power Company Prepared by B&W Nuclear Technologies, Inc.
Engineering and Project Services Div,isi'on 3315 Old Forest Road i
P. O. Box 10935 Lynchburg, Virginia 24506-0935 Qr %<rt9'=%=h o P
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CONTENTS i
1.0 INTRODUCTION
1-1 2.0 STATEMENT OF RESPONSE.................................
2 2.1 Part (1) to NRC Generic Letter 92-01, Revision 1, Supplement 1.......
2-1 2.2 Part (2) to NRC Generic Letter 92-01, Revision 1, Supplement 1.......
2-2 2.3 Part (3) to NRC Generic Letter 92-01, Revision i, Supplement 1.......
2-3 2.4 Part (4) to NRC Generic Letter 92-01, Revision 1, Supplement 1.......
2-3 3.0 CONC LUS I ON............................................
3-1 1
4.0 REFERENCES
4-1 S.0 CERTIFICATION..........................................
5-1 A P PEN DI X A................................................... A-1 i
i A P P ENDI X B................................................... B-1 c.nu 1
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1.0 INTRODUCTION
1-This document provides the Owners Group response to the Nuclear Regulatory Commission (NRC) Generic Letter 92-01, Revision 1, Supplement 1, for the B&W Owners Group Reactor Vessel Working Group member plants listed below.
4 Generic Letter 92-01, Revision 1, Supplement 1, was issued by the NRC on May 19,1995 and j
' addressed to all holders of nuclear power plant operating licensees. The generic letter was issued to ensure that all licensees have all of the data relevant to the evaluation of structural integrity i
of their reactor vessels and that all such data is appropriately included in their assessments of compliance with regulatory requirements regarding reactor vessel integrity. The requests for information in Generic Letter 92-01, Supplement 1, are divided into four parts, and written responses are required as follows:
Part 1:
within 90 days of the issue date of the generic letter j
(deadline of August 17,1995), and i.
)
Parts 2,3, and 4:
within 6 months of the issue date of the generic letter i
(deadline of November 19, 1995).
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-B This document provides the required information in Generic Letter 92-01, Revision I, Supplement 1, for the following plants:
Elant Owner Arkansas Nuclear One Unit 1 Entergy Operations, Inc.
Crystal River Unit 3 -
Florida Power Corporation Davis-Besse '
Toledo Edison Company R. E. Ginna Rochester Gas and Electric Corporation Oconee Unit 1 Duke Power Company Oconee Unit 2
. Duke Power Company Oconec Unit 3 Duke Power Company Point Beach Unit 1 Wisconsin Electric Power Company i
Point Beach Unit 2 Wisconsin Electric Power Company Surry Unit i Virginia Power Surry Unit 2 Virginia Power Three Mile Island Unit 1 GPU Nuclear Corporation Turkey Point Unit 3 Florida Power & Light Company Turkey Point Unit 4 Florida Power & Light Company Zion Unit 1 Commonwealth Edison Company Zion Unit 2 Commonwealth Edison Company Revision 1 of this report provides the Owners Group response to Parts (2), (3), and (4). The response to Part (1), which is included herein, was provided in the Revision 0 of this report dated July 1995.
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e 2.0 STATEMENT OF RESPONSE 2.1 Part (1) to NRC Generic Letter 92-01, Revision 1, Supplement 1 Part (1) of the Generic Letter requires the Addressees to provide the following information:
"a description of those actions taken or planned to locate all data relevant to the determination of RPV integrity, or an explanation of why the existing data base is considered complete as previously submitted" B&W Owners Group Reactor Vessel Workine Group Response to Part (1)
The B&W Owners Group's Materials Committee initiated efforts in 1977 to resolve the reactor vessel integrity issue for high copper, Linde 80, automatic submerged arc welds. Its successor organization, the B&W Owners Group's Reactor Vessel Working Group (RVWG), has continued these efforts to acquire and document relevant data for determining reactor vessel integrity ofits member plants. Appendix A provides a list of documents that are in the possession of the NRC and were considered in the development of the previous RVWG responses to Generic Letter 92-01, Revision 1 and associated requests for additional information.
Information for these documents was obtained from the following sources:
Babcock & Wilcox fabrication documentation (e.g., weld qualification reports)
Investigations sponsored by:
B&W Owners Group Electric Power Research Institute (EPRI)
NRC (at the Oak Ridge National Laboratory)
Reactor vessel material surveillance program reports The RVWG's Master Integrated Reactor Vessel Material Surveillance Program (MIRVP) links all the operating PWRs that contain high copper, Linde 80 welds and makes careful note of where 2-1
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the same weld, or its surrogate, is used in more than one reactor vessel. These data are shared for regulatory analyses regarding reactor vessel integrity.
A review of the list of reactor vessels fabricated by Babcock & Wilcox indicates that additional weld chemistry and initial Charpy V-notch and Drop Weight impact toughness data from a few domestic BWR and foreign PWR vessels may be available. Nevertheless, the RVWG believes that the available data, relevant to the assessment of its member plant vessels, have been 1
appropriately considered in the submitted reactor vessel integrity evaluations.
- A review of available data sources for B&W-fabricated reactor vessels (domestic and foreign) shows all available data have been considered in establishing the chemistry values for the Linde 80, ASA welds. Therefore, no change in the previously reported weld metal chemistry values is required for Linde 80 weld wires.
2.2 Part (2) to NRC Generic Letter 92-01, Revision 1, Supplement 1 Part (2) of the Generic Letter requires the Addressees to provide the following information:
"an assessment of any change in best-estimate chemistry based on consideration of relevant data;"
B&W Owners Group Reactor Vessel Working Group Response to Part (2)
The B&W Owners Group (B&WOG) performed extensive work to establish the chemical composition of beltline welds using the automatic submerged-arc (ASA) process, copper-plated Mn-Mo-Ni filler wire, and Linde 80 flux.'33 The work included collecting existing sources of chemistry data, performing extensive chemical analyses on archive production reactor vessel weld metals (from nozzle belt " dropouts" and surveillance welds) and developing predictive methods with the aid of statistical analyses. The results of the above work are considered to be the appropriate "best-estimate" chemical composition values representative of the high copper ASA/Linde 80 beltline welds.
A review of available data sources for B&W-fabricated reactor vessels (domestic and foreign) shows all available data have been considered in establishing the chemistry values for the Linde 2-2 e
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q 80, ASA welds. Therefore, no change in the previously reported weld metal chemistry values is required for Linde 80 weld wires.
2.3 Part (3) to NRC Generic Letter 92-01, Revision 1, Supplement 1 Part (3) of the Generic Letter requires the Addressees to provide the following information:
"a determination of the need for use of the ratio procedure in accordance with the established Position 2.1 of Regulatory Guide 1.99, Revision 2, for those licensees that use surveillance data to provide a basis for the RPV integrity evaluation;"
B&W Owners Group Reactor Vessel Workine Groun Resoonse to Part G)
Position 2.1 of Regulatory Guide 1.99, Revision 2, states "if there is clear evidence that the copper or nickel contents of the surveillance weld differs from that of the vessel weld, measured values of ARTwr hould be s
adjusted by multiplying them by the ratio of the chemistry factor for the vessel l
weld to that for the surveillance weld."
The B&WOG RVWG position is that the variability in chemical composition between the 1
individual surveillance weld sources for a particular weld wire heat is representative of the chemical variability in the reactor vessel beltline welds; therefore, the RVWG will continue to use surveillance data when available in accordance with Position 2.1 of Regulatory Guide 1.99, Revision 2, without applying the ratio procedure.
j The B&WOG RVWG has reviewed the potential impact of using the ratio procedure on the limiting weld in their reactor vessels and concludes that even though the chemistry factors may differ from the chemistry factors calculated based on the average weld wire chemistry, the RT rs e
i values do not change significantly as defined in 10CFR50.61 (i.e., does not exceed the screening criteria). These differences are presented in Table 2-1 along with their respective RTers values; j
these values are presented for information only and do not necessarily reflect plant-specific data previously submitted for plant licensing. (The supporting information for the development of Table 2-1 is presented in Appendix B.)
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2.4 Part (4) to NRC Generic Letter 92-01, Revision 1, Supplement I Part (4) of the Generic Letter requires the Addressees to provide the following information:
j "a written report providing any newly acquired data as specified above and (1) the results of any necessary revisions to the evaluation of RPV integrity in accordance with the requirements of 10CFR50.60,10CFR50.61, Appendices G and H to 10CFR Part 50, and any potential impact on the LTOP or P-T limits in the technical specifications or (2) a certification that previously submitted evaluations remain valid. Revised evaluations and certifications should include consideration l
of Position 2.1 of Regulatory Guide 1.99, Revision 2, as applicable, and any new data."
4 B&W Owners Group Reactor Vessel Workinn Group Response to Part (4)
Since the chemical variability in the Linde 80 weld surveillance data is represented by the variability observed in the Charpy surveillance data used to establish the chemistry factors for the weld wire heats in the RVWG beltline welds, and no new chemistry or mechanical information is available for these weld metals, the previously submitted reactor vessel integrity evaluations (i.e., plant-specific LTOP and P-T limit curves in plant technical specifications and RVWG evaluations of reactor vessel integrity in accordance with the requirements of 10CFR50.60,
.b 10CFR50.61, and Appendices G and H to 10CFR Part 50) remain valid.
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Table 2-1. Effect on Projected Values of RTm of Applying Ratio Procedure i
of Regulatory Guide 1.99, Revision 2, Position 2.1 i
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Ratio..
" Weld '. Wire _
Chem.
Chem.
.' Ratio.
> Screening Plant' (Weld Id.f Factor.:
Factor RTm
- RTm(
Criteria -
l (1);
(2)?
(3))
- (4)2 T(5)~
.'(6)'
(7)'
ANO-1 821T44 (WF-182-1) 162.1 170.7 156.3 164.5 300 406L44 (WF-112) 175.0 184.5
'173.0 182.4 300 CR3 71249 (SA-1769)
N/A' N/A' N/A' N/A' 300 72105 (WF-70) 138.4 136.6 132.3 130.6 300
}j DB 821T44 (WF-182-1) 162.1 170.7 166.2 174.9 300 l
REG 71249 (SA-1101)
N/A' N/A' N/A' N/A' 300 j
61782 (SA-847) 147.2 152.1 198.0 204.5 300 OC1 61782 (SA-1135) 147.2 152.1 67.4 69.6 300 i
71249 (SA-1229)
N/A*
N/A' N/A*-
N/A*
300 72445 (SA-1585) 149.8 143.8 144.8 139.0 300 i
OC2 406L44 (WF-154) 175.0 184.5 167.6 176.6 300 299L44 (WF-25) 216.9 231.4 212.7 226.8 300 OC3 821T44 (WF-200) 162.1 170.7 154.3 162.5 300 72442 (WF-67)
N/A*
N/A*
N/A' N/A*
300 PB1 71249 (SA-1101)
N/A' N/A' N/A' N/A' 300 i
61782 (SA-847) 147.2 152.1 168.1 173.6 270 PB2 72442 (SA-1484)
N/A*
N/A*
N/A*
N/A' 300 S1 72445 (SA-1585) 149.8 143.8 203.8 195.7 300 299L44 (SA-1526) 216.9 231.4 190.6 203.2 270 i
S2 72445 (SA-1585) 149.8 143.8 136.7 131.3 270 TMI-l 72105 (WF-70) 138.4 136.6 130.8 129.1 300 299L44 (WF-25) 216.9 231.4 208.8 222.7 300 299L44 (SA-1526) 216.9 231.4 201.8 215.3 270 TP3 72442 (SA-1484)
N/A' N/A' N/A' N/A' 300 71249 (SA-1101)
N/A*
N/A*
N/A' N/A*
300 TP4 72442 (WF-67)
N/A' N/A' N/A' N/A' 300 71249 (SA-1101)
N/A' N/A' N/A*
N/A*
300 Z1 406L44 (WF-154) 175.0 184.5 185.3 195.3 300 72105 (WF-70) 191.5 227.8 221.9 263.7 300 Z2 821T44 (WF-200) 162.1 170.7 174.9 184.2 300 72105 (WF-70) 191.5 227.8 166.0 197.2 270 71249 (SA-1769)
N/A*
N/A*
N/A' N/A*
300
- Surveillance data not used to calculate chemistry factor /RTm value.
2-5
b Notes to Table 2-1:
)
(1)
ANO-1
- Arkansas Nuclear One Unit 1 CR3
- Crystal River Unit 3 i
- Davis-Besse REG-
- R. E. Ginna
~
OCl
- Oconee Unit I OC2
- Oconee Unit 2 OC3
- Oconee Unit 3 i
PB1
- Point Beach Unit 1
~
PB2
- Point Beach Unit 2 4
SI-
- Surry Unit 1
)
S2
- Surry Unit 2
- TMl-1
- Three Mile Island Unit 1 TP3
- Turkey Point Unit.3 4
TP4
- Turkey Point Unit 4 4
Z1
- Zion Unit 1 Z2
- Zion Unit 2 3
4 (2)
Weld wire heat number with weld identifications where surveillance data is available.
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(3)
Chemistry factor in accordance with 10CFR50.61 without using ratio procedure (see Appendix B, Table A-1, pages B-27 through B-29).
l (4)
Chemistry factor with application of Regulatory Guide 1.99, Revision 2, Position j
2.1, ratio procedure (see Appendix B, Table B-1, pages B-31 through B-33).
(5)
Predicted RTm value in accordance with proposed 10CFR50.61, Federal Register, October 4,1994 (see Appendix B, Table 7, pages B-17 and B-18).
(6)
Predicted RT value in accordance with proposed 10CFR50.61, Federal Register, m
i October 4,1994 and using "ratioed chemistry factor" in accordance with Regulatory Guide 1.99, Revision 2, Position 2.1 (see Appendix B, Table 10, pages
)
B-24 and B-25).
]
i (7)
Screening criteria in accordance with 10CFR50.61.
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3.0 CONCLUSION
4 The B&W Owners Group (B&WOG) Reactor Vessel Working Group (RVWG) written responses for the requested information in Parts (1), (2), (3), and (4) of NRC Generic Letter 92-01, Revision 1, Supplement 1, are provided in Section 2 of this document. Based on the content of these responses, Generic Letter 92-01, Revision 1, Supplement 1, is considered complete for the B&WOG RVWG.
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4.0 REFERENCES
1.
K. E. Moore and A. S. Heller, " Chemistry of 177-FA B&W Owners' Group Reactor Vessel Beltline Welds," BAW-1500P. Babcock & Wilcox's Power Generation Group, 1
Nuclear Power Generation Division, Lynchburg, Virginia, September 1978.*
2.
K. E. Moore and A. S. Heller, "B&W 177-FA Reactor Vessel Beltline Weld Chemistry Study," BAW-1799, Babcock & Wilcox's Utility Power Generation Division, Lynchburg, Virginia, July 1983.*
3.
L. B. Gross, " Chemical Composition of B&W Fabricated Reactor Vessel Beltline Welds,"
BAW-2121P, B&W Nuclear Technologies, Inc., Lynchburg, Virginia, April 1991.
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l 5.0 CERTIFICATION -
This report accurately responds to the information requested in Generic Letter 92-01, Revision 1, Supplement 1.
WO iW I O/31l 9f M. J. IIeVan, Engineer III Date Materials & Structural Analysis Unit This report has been reviewed for technical content and accuracy.
ocT. 3I, /9'i
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L. b. Gross, Advisory Engineer Date Materials & Structural Analysis Unit Verification of independent review.
Y t441-A6 -Sb9<~
K. E. M6cre, Manager Date Materials & Structural Analysis Unit This report is approved for release.
Wawd) ichtMs D. L. Howell Date Program Manager l
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APPENDIX A DOCUMENT
SUMMARY
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Perrin, J.S., et al., " Final Report on Point' Beach Nuclear Plant Unit No.1. Pressure Vessel i
Surveillance Program: Evaluation of Capsule V," BMI-0671, Battelle Columbus Laboratories, Columbus, Ohio,' June 15,1973.
Perrin, J.S., et al.,." Final Report on Point Beach Nuclear Plant Unit No. 2 Pressure Vessel Surveillance Program:. Evaluation of Capsule V," BMI-0675. Battelle Columbus Laboratories, Columbus, Ohio, June 10,1975.
Perrin, J.S., et al., "Surry Unit No.1 Pressure Vessel Irradiation Capsule Program: Examination 2
l and Analysis of Capsule T," Battelle Columbus Laboratories, Columbus, Ohio, June 24,1975.
Lowe, A.L., Jr., et al., " Analysis of Capsule OCl-F from Duke Power Company Oconee Unit
}
1 Reactor Vessel Materials Surveillance Program," BAW-1421. Rev.1,' Babcock & Wilcox, Lynchburg, Virginia, September 1975.*
f Lowe, A.L., Jr., et al., " Analysis of Capsule OCl-E Duke Power Company Oconee Nuclear I-Station - Unit 1 Reactor Vessel Materials Surveillance Program," B AW-1436, Babcock & Wilcox, Lynchburg, Virginia, September 1977.*
)
Lowe, A.L., Jr., et al., " Analysis of Capsule OCII-C from Duke Power Company Oconee Nuclear i
Station, Unit 2 Reactor Vessel Material Surveillance Program," BAW-1437. Babcock & Wilcox, l
Lynchburg, Virginia, May 1977.*
1 I
l Lowe, A.L., Jr., et al., " Analysis of Capsule OCIII-A from Duke Power Company Oconee i
Nuclear Station, Unit 3 Reactor Vessel Materials Surveillance Program," BAW-1438, Babcock
& Wilcox, Lynchburg, Virginia, July 1977.*
Lowe, A.L., Jr., et al., " Analysis of Capsule TMI-lE from Metropolitan Edison Company Three i
Mile Island Nuclear Station-Unit 1 Reactor Vessel Materials Surveillance Program," B AW-1439, l
Babcock & Wilcox, Lynchburg, Virginia, January 1977.*
l Lowe, A.L., Jr., et al., " Analysis of Capsule ANI-E from Arkansas Power & Light Company i
Arkansas Nuclear One -- Unit 1 Reactor Vessel Materials Surveillance Program," BAW-1440, Babcock & Wilcox, Lynchburg, Virginia, April 1977.*
Moore, K.E., and A.S. Heller, " Chemistry of 177-FA B&W Owners' Group Reactor Vessel Beltline Welds," BAW-1500. Babcock & Wilcox, Lynchburg, Virginia, September 1978.*
Harbison, L. S., " Master Integrated Reactor Vessel Surveillance Program," B AW-1543. Revision 4, B&W Nuclear Technologies, Inc., Lynchburg, Virginia, February 1993.
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Lowe, A.L., Jr., et al., " Analysis of Capsule CR3-B Florida Power Corporation Crystal River Unit 3 Reactor Vessel Materials Surveillance Program," BAW-1679. Rev.1. Babcock & Wilcox, Lynchburg, Virginia, June 1982.*
Lowe, A.L., Jr., et al., " Analysis of Capsule OCIII-B from Duke Power Company Oconee Nuclear Station, Unit 3 Reactor Vessel Materials Surveillance Program," BAW-1697. Babcock
& Wilcox, Lynchburg, Virginia, October 1981.*
Lowe, A.L., Jr., et al., " Analysis of Capsule ANI-B from Arkansas Power & Light Company's Arkansas Nuclear One, Unit 1 Reactor Vessel Materials Surveillance Program," BAW-1698, Babcock & Wilcox, Lynchburg, Virginia, November 1981.*
Lowe, A.L., Jr., et al., " Analysis of Capsule OCll-A from Duke Power Company's Oconee Nuclear Station, Unit 2 Reactor Vessel Material Surveillance Program," BAW-1699. Babcock &
Wilcox, Lynchburg, Virginia, December 1981.*
Lowe, A.L., et al., " Analyses of Capsule TEl-F The Toledo Edison Company Davis-Besse Nuclear Power Station Unit 1 Reactor Vessel Materials Surveillance Program," BAW-1701, Babcock & Wilcox, Lynchburg, Virginia, January 1982.*
Lowe, A.L., Jr., et al., " Analysis of Capsule RSI-B Sacramento Municipal Utility District Rancho Seco Unit i Reactor Vessel Material Surveillance Program," BAW-1702, Babcock & Wilcox, Lynchburg, Virginia, February 1982.*
Lowe, A.L., Jr., et al., " Analysis of Capsule RSI-D Sacramento Municipal Utility District Rancho Seco Unit 1 Reactor Vessel Material Surveillance Program," BAW-1792. Babcock & Wilcox, Lynchburg, Virginia, October 1983.*
Lowe, A.L., Jr., and J.W. Pegram, " Correlations for Predicting the Effects of Neutron Radiation on Linde 80 Submerged-Arc Welds," BAW-1803. Rev.1, B&W Nuclear Service Company, Lynchburg, Virginia, May 1991.*
l Aadland, J.D., " Babcock & Wilcox Owners' Group 177-Fuel Assembly Reactor Vessel and Surveillance Program Materials Information," BAW-1820. Babcock & Wilcox, Lynchburg, Virginia, December 1984.*
i Lowe, A.L., et al., " Analyses of Capsule TEl.B The Toledo Edison Company Davis-Besse Nuclear Power Station Unit 1 Reactor Vessel Material Surveillance Program," BAW-1834, Babcock & Wilcox, Lynchburg, Virginia, May 1984.*
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A-3 i
e Lowe, A.L., Jr., et al.," Analysis of Capsule ANI-A Arkansas Power & Light Company Arkansas Nuclear One, Unit 1 Reactor Vessel Materials Surveillance Program," BAW-1836. Babcock &
Wilcox, Lynchburg, Virginia, July 1984.*
Aadland, J.D., et al., " Analysis of Capsule OCl-A Duke Power Company Oconee Nuclear Station
- Unit 1 Reactor Vessel Materials Surveillance Program," BAW-1837. Babcock & Wilcox, Lynchburg, Virginia, August 1984.*
Lowe, A.L., et al., " Analyses of Capsule TEl-A The Toledo Edison Company Davis-Besse Nuclear Power Station Unit 1 Reactor Vessel Material Surveillance Program," BAW-1882. Rev.-
1, Babcock & Wilcox, Lynchburg, Virginia, June 1989.*
Lowe, A.L., Jr., et al.," Analysis of Capsule CR3-C Florida Power Corporation Crystal River Unit 3 Reactor Vessel Materials Surveillance Program," DAW-1898, Babcock & Wilcox, Lynchburg, Virginia, March 1986.*
Lowe, A.L., Jr., et al., " Analysis of Capsule CR3-D Florida Power Corporation Crystal River Unit 3 Reactor Vessel Materials Surveillance Program," BAW-1899, Babcock & Wilcox, Lynchburg, Virginia, March 1986.*
Lowe, A.L., Jr., et al., '. Analysis of Capsule TMil-C GPU Nuclear Three Mile Island Nuclear Station-Unit 1 Reactor Vessel Material Surveillance Program," BAW-1901. Babcock & Wilcox, Lynchburg, Virginia, March 1986.*
Lowe, A.L., Jr., et al., " Analysis of Capsule CR3-LG1 -- Babcock & Wilcox Owners Group Integrated Reactor Vessel Materials Surveillance Program," BAW-1910P. Babcock & Wilcox, Lynchburg, Virginia, August 1986.*
Lowe, A.L., Jr., et al., " Analysis of Capsule DB1-LG1 -- Babcock & Wilcox Owners Group Integrated Reactor Vessel Materials Surveillance Program," BAW-1920P. Babcock & Wilcox, Lynchburg, Virginia, October 1986.*
Lowe, A.L., Jr., et al.," Analysis of Capsule CR3-F Florida Power Corporation Crystal River Unit 3 Reactor Vessel Materials Surveillance Program," BAW-2049, Babcock & Wilcox, Lynchburg, Virginia, September 1988.*
Lowe, A.L., Jr., et al., " Analysis of Capsule OCl-C Duke Power Company Oconee Nuclear Station Unit-1 Reactor Vessel Materials Surveillance Program," BAW-2050, Babcock & Wilcox, Lynchburg, Virginia, October 1988.*
A-4
Lowe, A.L., Jr., et al., " Analysis of Capsule OCII-E Duke Power Company Oconee Nuclear Station Unit-2 Reactor Vessel Material Surveillance Program," BAW-2051, Babcock & Wilcox, Lynchburg, Virginia, October 1977.*
Lowe, A.L., Jr., et al.," Analysis of Capsule RSI-F Sacramento Municipal Utility District Rancho Seco Unit 1 Reactor Vessel Material Surveillance Program," BAW-2074, Babcock & Wilcox, Lynchburg, Virginia, April 1989.*
Lowe, A.L., Jr., et al.," Analysis of Capsule ANI-C Arkansas Power & Light Company Arkansas Nuclear One, Unit 1 Reactor Vessel Materials Surveillance Program," BAW-2075. Rev.1.
Babcock & Wilcox, Lynchburg, Virginia, October 1989.*
Lowe, A.L., Jr., et al., " Analysis of Capsule Y Commonwealth Edison Company Zion Nuclear Plant Unit 1 Reactor Vessel Material Surveillance Program," BAW-2082. B&W Nuclear Service Company, Lynchburg, Virginia, March 1990.*
Gross, L.B., " Chemical Composition of B&W Fabricated Reactor Vessel Beltline Welds," BAW-2121 P, 77-2121-00, B&W Nuclear Service Company, Lynchburg, Virginia, April 1991.*
Lowe, A.L., et al., " Analysis of Capsule TEl-D The Toledo Edison Company Davis-Besse Nuclear Power Station Unit 1 Reactor Vessel Material Surveillance Program," BAW-2125, Babcock & Wilcox, Lynchburg, Virginia, December 1990.*
,. w Lowe, A.L., Jr., et al., " Analysis of Capsule OCIll-D Duke Power Company Oconee Nuclear Station Unit-3 Reactor Vessel Material Surveillance Program," BAW-2128. Rev.1, B&W Nuclear Service Company, Lynchburg, Virginia, May 1992.*
Lowe, A.L., Jr., et al.," Analysis of Capsule S Wisconsin Electric Power Company Point Beach Nuclear P! ant Unit No.1 Reactor Vessel Material Surveillance Program," BAW-2140, B&W Nuclear Service Company, Lynchburg, Virginia, August 1991.*
Ouellette, C.A., " Materials Information for Westinghouse-Designed Reactor Vessels Fabricated by B&W", DAW-2150, B&W Nuclear Service Company, Lynchburg, Virginia, December 1990.*
DeVan, M.J., Gross, L. B., and Lowe, A. L. Jr., "B&W Owners Group Response to Generic Letter 92-01," BAW-2166. B&W Nuclear Service Company, Lynchburg, Virginia, June 1992.
Yoon, K. K., " Low Upper-Shelf Toughness Fracture Mechanics Analysis of Reactor Vessels of B&W Owners Reactor Vessel Working Group for Level C & D Service Loads," BAW-2178PA, B&W Nuclear Technologies, Inc., Lynchburg, Virginia, April 1994.
This report is available from B&W Nuclear Technologies, Inc., Lynchburg, Virginia.
A-5
Lowe, A.L., Jr., et al., " Evaluation of Capsules TMI2-C and TMI2-E Irradiated in Davis Besse Nuclear Power Station Unit 1 -- EPRI Reactor Vessel Embrittlement Program," BAW-2190, B&W Nuclear Technologies, August,1993.
Yoon, K. K., " Low Upper-Shelf Toughness Fracture Analysis of Reactor Vessels of B&W Owner., Group Reactor Vessel Working Group for Level A & B Conditions," BAW-2192PA, D&W Nuclear Technologies, Inc., Lynchburg, Virginia, April 1994.
Yoon, K.K., " Fracture Toughness Characterization of WF-70 Weld Metal," BAW-2202, B&W Nuclear Technologies, Inc., Lynchburg, Virginia, September 1993.
DeVan, M.J., and K.K. Yoon, " Response to Closure Letters to Generic Letter 92-01, Revision 1," BAW-2222, B&W Nuclear Technologies, Inc., Lynchburg, Virginia, June 1994.
Palme, ll.S., et al., " Methods of Compliance with Fracture Toughness and Operational Requirements of 10 CFR 50, Appendix G," BAW-10046P, Babcock & Wilcox, Lynchburg, Virginia, March 1976.*
Moore, K.E., et al.," Evaluation of the Atypical Weldment," BAW-10144A, Babcock & Wilcox, Lynchburg, Virginia, February 1980.*
Fromm, E.O., et al., "Surry Unit No. 2 Nuclear Plant Reactor Pressure Vessel Surveillance Program:
Examination and Analysis of Capsule W," BCL-585-026, Battelle Columbus Laboratories, Columbus, Ohio, February 1981.
Perrin JA, et al., " Zion Nuclear Plant Reactor Pressure Vessel Surveillance Program: Unit No.
1 Capsule T, and Unit No. 2 Capsule U," BCL-585-4, Battelle Columbus Laboratories, Columbus, Ohio, March 25,1978.
Van Der Sluys, W.A., et al., "An Investigation of Mechanical Properties and Chemistry Within a Thick MnMoNi Submerged Arc Weldment," EPRI NP-373, Prepared by Babcock & Wilcox for the Electric Power Research Institute, Palo Alto, California, February 1977.
Letter from J. W. Williams, Florida Power & Light Company, to D. G. Eisenhut, Nuclear Regulatory Commission, Turkey Point Units 3 & 4, Docket Nos. 50-250 & 50-251, Pressurized Thermal Shock - Reactor Vessel Materials Data, dated February 10,1984.
Mager, T.R., et al.," Analysis of Capsule V from the Rochester Gas and Electric R.E. Ginna Unit No.1 Reactor Vessel Radiation Surveillance Program," FP-RA-1, Westir:ghouse Electric Corporation, Pittsburgh, Pennsylvania, March 1,1973.
A-6
O Nanstad, R.K., and R.G. Berggren, " Irradiation Effects on Charpy Impact and Tensile Properties of Low Upper Shelf Welds, HSSI Series 2 and 3," NUREG/CR-5696, Prepared by Oak Ridge National Laboratory for the U.S. Nuclear Regulatory Commission, Washington, DC, August j
1991.
Nanstad, R.K., et al., " Chemical Composition and RTwr Determinations for Midland Weld WF-70," NUREG/CR-5914. Prepared by Oak Ridge National Laboratory for the U.S. Nuclear Regulatory Commission, Washington, DC, December 1992.
Letter from S. A. Varga, Nuclear Regulatory Commission, to J. W. Williams. Florida Power &
Light Company,
Subject:
Evaluation of Reactor Vessel Materials Data for Turkey Point Plant Units 3 and 4 Reactor Vessels, dated April 26,1984.
Letter from G.E. Edison, Nuclear Regulatory Commission, to W.F. Conway, Florida Power and Light Company, Turkey Point Units 3 and 4 -- Issuance of Amendment RE: Pressure and Temperature (P/F) Limits (TAC Nos. 69390 and 69391), dated January 10,1989.
Norris, E.B., " Reactor Vessel Material Surveillance Program for Turkey Point Unit No. 4 Analysis of Capsule T," SwRI 02-4221. Southwest Research Institute, San Antonia, Texas, June 14, 1976.
Norris, E.B., " Reactor Vessel Material Surveillance Program for Capsule S -- Turkey Point Unit No. 3, Capsule S -- Turkey Point Unit No. 4," SwRI-02-5131, SwRI-02-5380, Southwest Research Institute, San Antonio, Texas, May 1979.
l Norris, E.B., "Reacter Vessel Material Surveillance Program for Zion Unit No. 2 Analysis of Capsule T," SwRI-06-6901-001, Southwest Research Institute, San Antonio, Texas, July 6,1983.
Norris, E.B., " Reactor Vessel Material Surveillance Program for Zion Unit No.1 Analysis of Capsule X," SwRI-06-7484-001. Southwest Research Institute, San Antonio, Texas, March 1984.
Nair, P.K., and E.B. Norris, " Reactor Vessel Material Surveillance Program for Turkey Point Unit No. 3: Analysis of Capsule V," SwRI-06-8575, Southwest Research Institute, San Antonio, j
Texas, August 1986.
i Yanichko, S.E., et al., " Analysis of Capsule R from the Rochester Gas and Electric R.E. Ginna Unit No.1 Reactor Vessel Radiation Surveillance Program," WCAP-8421, Westinghouse Electric Corporation, Pittsburgh, Pennsylvania, November 1974.
Yanichko, S.E., et al., " Analysis of Capsule T from the Florida Power and Light Company Turkey Point Unit No. 3 Reactor Vessel Radiation Surveillance Program," WCAP-8631,
)
Westinghouse Electric Corporation, Pittsburgh, Pennsylvania, December 1975.
J 1
Yanichko, S.E., and S.L. Anderson," Analysis of Capsule S from the Wisconsin Electric Power Company and Wisconsin Michigan Power Company Point Beach Nuclear Plant Unit No.1 A-7
e Reactor Vessel Radiation Surveillance Program," WCAP-8739. Westinghouse Electric Corporation, Pittsburgh, Pennsylvania, November 1976.
Davidson, J.A., et al.," Analysis of Capsule T from the Wisconsin Electric Power Company Point Beach Nuclear Plant Unit No. 2 Reactor Vessel Radiation Surveillance Program," WCAP-9331, Westinghouse Electric Corporation, Pittsburgh, Pennsylvania, August 1978.
Yanichko, S.E., and S.L. Anderson, " Analysis of Capsule R from the Wisconsin Electric Power Company Point Beach Nuclear Plant Unit No.1 Reactor Vessel Radiation Surveillance Program,"
WCAP-9357, Westinghouse Electric Corporation, Pittsburgh, Pennsylvania, August 1978.
Yanichko, S.E., et al.," Analysis of Capsule R from the Wisconsin Electric Power Company Point Beach Nuclear Plant Unit No. 2 Reactor Vessel Radiation Surveillance Program," WCAP-9635, Westinghouse Electric Corporation, Pittsburgh, Pennsylvania, December 1979.
Yanichko, S.E., et al., " Analysis of Capsule U from the Commonwealth Edison Company Zion Nuclear Plant Unit 1 Reactor Vessel Radiation Surveillance Program," WCAP-9890.
Westinghouse Electric Corporation, Pittsburgh, Pennsylvania, March 1981.
Yanichko, S.E., et al., " Analysis of Capsule T from the Rochester Gas and Electric R.E. Ginna Nuclear Plant Reactor Vessel Radiation Surveillance Program," WCAP-10086. Westinghouse Electric Corporation, Pittsburgh, Pennsylvania, April 1982.
..n Yanichko, S.E., et al.," Analysis of Capsule T from the Wisconsin Electric Power Company Point Beach Nuclear Plant Unit No.1 Reactor Vessel Radiation Surveillance Program," WCAP-10736, Westinghouse Electric Corporation, Pittsburgh, Pennsylvania, December 1984.
Yanichko, S.E., and V.A. Perone, " Analysis of Capsule V from the Virginia Electric and Power Company Surry Unit 1 Reactor Vessel Radiation Surveillance Program," WCAP-11415.
Westinghouse Electric Corporation, Pittsburgh, Pennsylvania, February 1987.
Yanichko, S.E., and V.A. Perone, " Analysis of Capsule V from the Virginia Electric and Power Company Surry Unit 2 Reactor Vessel Radiation Surveillance Program," WCAP-11499.
Westinghouse Electric Corporation, Pittsburgh, Pennsylvania, June 1987.
Terek, E., et al., " Analysis of Capsule Y from the Commonwealth Edison Company Zion Unit 2 Reactor Vessel Radiation Surveillance Program," WCAP-12396. Westinghouse Electric Corporation, Pittsburgh, Pennsylvania, September 1989.
Chicots, J.M., et al., " Analysis of Capsule S from the Rochester Gas and Electric Corporation R.E. Ginna Reactor Vessel Radiation Surveillance Program," WCAP-13902, Westinghouse Electric Corporation, Pittsburgh, Pennsylvania, December 1993.
A-8
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APPENDIX B i
SUPPORTING INFORMATION FOR EFFECT ON PROJECTED VALUES OF RTm OF APPLYING RATIO PROCEDURE 4
4 0
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TABLE OF CONTENTS
1.0 INTRODUCTION
........................................... B-3 2.0
SUMMARY
OF RESULTS..................................... B-3 3.0 A S S UM PTI ONS............................................ B-3 4.0 B ASIS OF INPUT DATA...................................... B-6 f
4.1 Linde 80 Weld Wire Chemical Compositions................... B-6 l
4.1.1 Chemical Compositions for Reactor Vessel Beltline Welds...... B-6 4.1.2 Surveillance Data Chemical Compositions................. B-6
' 4.2 Neutron Fluence Estimates.................................. B-6 4.3 Initial Reference Nil-Ductility Temperature.................... B-11 1
4.4 Linde 80 Weld Metal Surveillance Data Available................ B-11
.s 5.0 PRESSURIZED THERMAL SHOCK REFERENCE TEMPERATURE...... B-11 5.1 Pressurized Thermal Shock Reference Temperature, Surveillance Data Available........................................ B-1 1 5.1.1 Initial Nil-Ductility Reference Temperature............... B-16 5.1.2 Irradiation Inducted Change in Reference Temperature........ B-16 5.1.2.1 Chemistry Factor........................... B-16 5.1.2.2 Fluence Factor............................. B 19 5.1.3 M argi n........................................ B-19 5.2 Regulatory Guide 1.99, Revision 2, Position 2.1, Ratio Procedure..... B-20 ATTACHMENT A - Chemistry Factor Determination Using Surveillance Data...... B-26 ATTACllMENT B - Chemistry Factor Determination Using Ratio Procedure and Surveillance Data in Accordance With Regulatory Guide 1.99, Revision 2, Position 2.1..................................... B-3 0 ATTACH MENT C - References...................................... B-34 i
B-2 i
i
1.0 INTRODUCTION
The purpose of this Appendix is to provide the supporting information regarding the calculations of the chemistry factors and pressurized thermal shock reference temperature (RTm) values contained in Table 2-1 of this report relative to the use of Regulatory Guide 1.99, Revision 2, Position 2.l ' ratio procedure.
S l
The RTm values are calculated in accordance with the proposed revision of 10CFR50.61.S' These calculations are applied to the Linde 80 beltline welds in B&W Owners Group (B&WOG) Reactor Vessel Working Group (RVWG) plants.
2.0
SUMMARY
OF RESULTS r.*
Chemistry factor and projected RTm values by applying the Regulatory Guide 1.99, Revision 2, Position 2.1, ratio procedure are shown in Table 1.
3.0 ASSUMPTIONS No major assumptions are contained in this calculation.
B-3
I o
r Table 1. Effect on Projected Values of RTm of Applying Ratio Procedure of Regulatory Guide 1.99, Revision 2, Position 2.1 Ratio
. Weld Wire Chem.
Chem.
Ratio.
Screening Plant (Weld Id.)
Factor Factor RTm RTm Criteria (1)'
(2)
-(3)
(4)
(5)
(6)
(7)
ANO-1 821T44 (WF-182-1) 162.1 170.7 156.3 164.5 300 406L44 (WF-112) 175.0 184.5 173.0 182.4 300 CR3 71249 (SA-1769)
N/A*
N/A*
N/A*
N/A' 300 72105 (WF-70) 138.4 136.6 132.3 130.6 300 DB 821T44 (WF-182-1) 162.1 170.7 166.2 174.9 300 REG 71249 (SA-110l)
N/A' N/A' N/A' N/A' 300 61782 (SA-847) 147.2 152.1 198.0 204.5 300 OCl 61782 (SA-1135) 147.2 152.1 67.4 69.6 300 71249 (SA-1229)
N/A*
N/A*
N/A' N/A*
300 72445 (SA-1585) 149.8 143.8 144.8 139.0 300 OC2 406L44 (WF-154) 175.0 184.5 167.6 176.6 300 299L44 (WF-25) 216.9 231.4 212.7 226.8 300 OC3 821T44 (WF-200) 162.1 170.7 154.3 162.5 300 72442 (WF-67)
N/A*
N/A*
N/A' N/A' 300 PB1 71249 (SA-1101)
N/A' N/A' N/A*
N/A*
300 61782 (SA-847) 147.2 152.1 168.1 173.6 270 PB2 72442ISA-1484)
N/A' N/A' N/A' N/A' 300 S1 72445 (SA-1585) 149.8 143.8 203.8 195.7 300 299L44 (SA-1526) 216.9 231.4 190.6 203.2 270 S2 72445 (SA-1585) 149.8 143.8 136.7 131.3 270 TMI-I 72105 (WF-70) 138.4 136.6 130.8 129.1 300 299L44 (WF-25) 216.9 231.4 208.8 222.7 300 299L44 (SA-1526) 216.9 231.4 201.8 215.3 270 TP3 72442 (SA-1484)
N/A' N/A' N/A' N/A' 300 71249 (SA-1101)
N/A*
N/A*
N/A*
N/A*
300 TP4 72442 (WF-67)
N/A' N/A' N/A' N/A' 300 71249 (SA-1101)
N/A*
N/A*
N/A*
N/A' 300 Z1 406L44 (WF-154) 175.0 184.5 185.3 195.3 300 72105 (WF-70) 191.5 227.8 221.9 263.7 300 Z2 821T44 (WF-200) 162.1 170.7 174.9 184.2 300 72105 (WF-70) 191.5 227.8 166.0 197.2 270 71249 (SA-1769)
N/A*
N/A*
N/A*
N/A*
300
- Surveillance data not used to calculate chemistry factor /RTm value.
B-4
g.
Notes to Table 1:
(1)
ANO-1 Arkansas Nuclear One Unit 1 CR3 Crystal River Unit 3 DB Davis-Besse REG R. E. Ginna OCl Oconee Unit 1 OC2 Oconee Unit 2 OC3 Oconee Unit 3 l
PB1 Point Beach Unit 1 PB2 Point Beach Unit 2 Surry Unit 1 S1 Surry Unit 2 S2 TMI-l Three Mile Island Unit 1 Turkey Point Unit 3 i
TP3 Turkey Point Unit 4 TP4 Z1 Zion Unit 1 Z2 Zion Unit 2 (2)
Weld wire heat number with weld identifications where surveillance data is available.
-m:
(3)
Chemistry factor in accordance with 10CFR50.61 without using ratio procedure (see Table A-1, pages B-27 through B-29).
(4)
Chemistry factor with application of Regulatory Guide 1.99, Revision 2, Position 2.1, ratio procedure (see Table B-1, pages B-31 through B-33).
(5)
Predicted RTm value in accordance with proposed 10CFR50.61, Federal Register, j
October 4,1994 (see Table 7, pages B-17 and B-18).
(6)
Predicted RTm value in accordance with proposed 10CFR50.61, Federal Register, j
October 4,1994 and using "ratioed chemistry factor" in accordance with Regulatory Guide 1.99, Revision 2, Position 2.1 (see Table 10, pages B-24 and B-l 25).
i (7)
Screening criteria in accordance with 10CFR50.61.
a B-5
-r m-w ---
4.0 BASIS OF INPUT DATA 4.1 Linde 80 Weld Wire Chemical Compositions 4.1.1 Chemical Compositions for Reactor Vessel Beltline Welds The best-estimate Linde 80 weld metal compositions are based on work reported in BAW-1500P -' and BAW-2121P " Table 2 presents the 8
8 copper and nickel chemical compositions of the Linde 80 weld wire where surveillance data are available for reactor vessel integrity evaluations.
4.1.2 Surveillance Data Chemical Compositions The copper and nickel chemical compositions including supplemental analyses performed on tested surveillance specimens for the each Linde 80 weld metal included in the B&W Owners Group RVWG Master Integrated Surveillance Program are shown in Table 3.
4.2 Neutron Fluence Estimates The end-of-life (i.e.,32 EFPY) neutron fluences for the RVWG reactor vessel weld metals for which surveillance data is available are shown in Table 4. The end-of-life inside surface fluences were taken from BAW-2222a.5 with the exception of R. E. Ginna where the end-of-life inside surface fluence was taken from WCAP-13902."
B-6 l
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Table 2. Best-Estimate Chemical Composition of Linde 80 Weld Wires for Which Surveillance Data is Available j
Weld '
Weld Metal Chemical. Composition, wt%
Wire Identification Cu-Ni 299L44 SA-1526, WF-25 0.35 0.68 406L44 WF-112, WF-154, WF-193 0.31 0.59 61782 SA-847, SA-848, SA-1036, SA-1135 0.25 0.54 71249 SA-1094, SA-1101, SA-1229, SA-1769 0.26 0.60 72105 WF-70, WF-209-1 0.35 0.59 72442 SA-1484, WF-67 0.24 0.60 72445 SA-1263, SA-1585, SA-1650, WF-9 0.21 0.59 821T44 WF-182-1, WF-200 0.24 0.63 g
Table 3. Chemical Composition of Linde 80 Surveillance Weld Metals Weld Weld Metal Chemical Wire Identification Composition, Reference WtYO Cu-Ni 84 "~8 299L44 SA-1526 - B&W Owners Group 0.37 0.70 BAW-1543, Rev. 4 SA-1526 - Surry-1 0.25 0.68 WCAP-11415" 0.243 0.643 WCAP-il415 WF Three Mile Island-l 0.33 0.66 BAW-1820"
WF-25(6) - B&W Owners Group 0.35 0.67 BAW-1543, Rev. 4 WF-25(9) - B&W Owners Group 0.35 0.70 BAW-1543, Rev. 4 406L44 WF-112 - Oconee-1 0.32 0.59 BAW-1820 WF-112 - B&W Owners Group 0.32 0.59 BAW-1543, Rev. 4 WF-193 - Arkansas Nuclear One-1 0.28 0.59 BAW-1820 WF-193 - Rancho Seco-1 0.31 0.59 BAW-1820 WF-193 - Point Beach-2 0.25 0.59 WCAP-7712"'"
61782 SA-1036 - R. E. Ginna 0.23 0.56 WCAP-10086"-"
0.22 0.50 WCAP-10086 SA-1135 - B&W Owners Group 0.27 0.59 BAW-1543, Rev. 4 B-7
c Table 3 (cont'd). Chemical Composition of Linde 80 Surveillance Weld Metals 1 Weld Metal?,
1 Chemical.
' Weld.
!Idehtification?
- CompositionT-
! Reference
- Wire -
^
fwt% -
,y c,;
= g;;
71249-SA-1094 - Turkey Point-4 0.30 0.60 WCAP-7660 -"
8 SA-1101 - Turkey Point-3 0.31 0.57' WCAP-7656 '"
8 72105 WF-70(N) - B&W Owners Group 0.42 0.59 BAW-1543, Rev. 4 WF-209 Oconee-2 0.36 0.58 BAW-1820 WF-209 Oconee-3 0.30 0.58 BAW-1820 WF-209 Zion-1 0.35 0.57 BAW-2082 -"
8 0.216 0.53 SwRI-7484-001/l" 0.27 0.57 SwRI-7484-001/1 0.218 0.545 SwRI-7484-001/1 0.25 0.49 SwRI-7484-001/1 0.26 0.56 SwRI-7484-001/1 0.26 0.54 SwRI-7484-001/1 y,,,
0.24 0.55 SwRI-7484-001/1 0.26 0.53 SwRI-7484-001/1 0.28 0.56 SwRI-7484-001/1 0.25 0.54 SwRI-7484-001/1 0.25 0.55 BAW-2082 0.22 0.55 BAW-2082 0.22 0.54 BAW-2082 0.23 0.54 BAW-2082 0.23 0.54 BAW-2082 l
0.22 0.54 BAW-2082 0.24 0.55 BAW-2082 O.24 0.53 BAW-2082 4
l 1
1 B-8
Table 3 (cont'd). Chemical Composition of Linde 80 Surveillance Weld Metals Weld l Weld Metal;
_ ' Chemical Wire Identification Composition,
~ Reference wt%
. Cu -
Ni-72105 WF-209 Zion-2 0.28 0.55 WCAP-12396 "
S (cont'd) 0.19 0.52 SwRI-6901" 0.23 0.52 SwRI-6901 0.23 0.54 SwRI-6901 0.25 0.53 SwRI-6901 0.27 0.53 SwRI-6901 j
0.21 0.48 SwRI-6901 0.17 0.53 SwRI-6901 0.26 0.54 SwRI-6901 0.23 0.47 SwRI-6901 d
0.22 0.52 SwRI-6901 0.20 0.56 SwRI-6901 0.26 0.53 SwRI-6901 1
0.31 0.52 SwRI-6901 0.28 0.55 SwRI 6901 0.26 0.57 WCAP-12396 l
0.27 0.60 WCAP-12396 0.26 0.59 WCAP-12396 j
0.28 0.60 WCAP-12396 0.26 0.60 WCAP-12396 I
0.27 0.60 WCAP-12396 4
0.26 0.56 WCAP-12396 1
0.23 0.59 WCAP-12396 72442 WF B&W Owners Group 0.22 0.60 BAW-1543, Rev. 4 72445 SA-1263 - Point Beach-1 0.24 0.57 WCAP-10736" 0.22 0.66 WCAP-10736 SA-1585 - B&W Owners Group 0.21 0.59 BAW-1543, Rev. 4 821T44 WF-182 Davis-Besse 0.21 0.63 BAW-1820 1
4 B-9
i f
Table 4. Predicted Inside Surface Fluence of B&WOG RVWG Reactor
- Vessel Weld Metals for Which Surveillance Data is Available.
I
~
. Inside Surface iPlant # '
4 Weld Id/
FluencefA/cth2 Arkansas Nuclear One Unit 1 821T44 (WF-182-1) 8.62E+18 406L44 (WF-112) 9.40E+18 Crystal River Unit 3 71249 (SA-1769)
N/A' 72105 (WF-70) 8.22E+18 Davis-Besse 821T44 (WF-182-1) 1.07E+19 R. E. Ginna 71249 (SA-1101)
N/A*
61782 (SA-847) 3.68E+19 Oconee Unit 1 61782 (SA-1135) 1.18E+18 71249 (SA-1229)
N/A*
72445 (SA-1585) 8.68E+18 Oconee Unit 2 406L44 (WF-154) 8.42E+18 299L44 (WF-25) 9.19E+18 Oconee Unit 3 821T44 (WF-200) 8.26E+18
^
72442 (WF-67)
N/A*
[
l Point Beach Unit 1 71249 (SA-1101)
N/A*
61782 (SA-847) 1.63E+19 i
Point Beach Unit 2 72442 (SA-1484)
N/A*
l Surry Unit 1 72445 (SA-1585) 3.96E+19 1
299L44 (SA-1526) 6.39E+18 Surry Unit 2 72445 (SA-1585) 7.14E+18 Three Mile Island Unit 1 72105 (WF-70) 7.89E+18 l
299L44 (WF-25) 8.61E+18 299L44 (SA-1526) 7.67E+18 Turkey Point Unit 3 72442 (SA-1484)
N/A*
71249 (SA-1101)
N/A*
Turkey Point Unit 4 72442 (WF-67)
N/A*
71249 (SA-1101)
N/A' Zion Unit 1 406L44 (WF-154) 1.21E+19 72105 (WF-70) 1.73E+19 l
Zion Unit 2 821T44 (WF-200) 1.30E+19 72105 (WF-70) 6.04E+18 i
71249 (SA-1769)
N/A*
i 4
- Surveillance data not used to calculate chemistry factor /RTvrs value.
l B-10 i
f 4
r 5
e 4.3 Initial Reference Nil-Ductility Temperature Table 5 lists the initial reference nil-ductility temperature (IRTuor) values and their standard deviations (oi) for the RVWG reactor vessel beltline materials for which surveillance data is available. The IRTwor values for the Linde 80 weld metals was determined using an alternative method based on fracture toughness in the transition range. This method is described in BAW-2202 -2 and BAW-2245, 8
Revision 1.8 2i 4.4 Linde 80 Weld Metal Surveillance Data Available The available power reactor surveillance data through October 1995 for the Linde 80 weld metals are listed in Table 6. The data contained in this Table include capsule fluence and 30 ft-lb transition temperature.
5.0 PRESSURIZED THERMAL SHOCK REFERENCE TEMPERATURE 5.1 Pressurized Thermal Shock Reference Temperature, Sun'elllance Data Available In accordance with proposed rule,10CFR50.61, the pressurized thermal shock reference temperature (RT,n) is determined by the following expression:
i RT,,n = Initial RTgo7 + ARTuor + Margin (1) where:
Initial RTuor = Initial nil-ductility reference temperature ARTwnr
= ltradiation induced change in reference temperature Margin
= 2 sigma (c) standard deviation B-11
e
\\
g Table 5. Initial Reference Temperature for B&WOG RVWG Reactor Vessel Weld Metals for Which Surveillance Data is Available Initial RTuor, Plant Weld Id.
F ai Arkansas Nuclear One Unit 1 821T44 (WF-182-1)
-27 0
406L44 (WF-112)
-27 0
Crystal River Unit 3 71249 (SA-1769)
N/A*
72105 (WF-70)
-26.5 0
Davis-Besse 821T44 (WF-182-1)
-27 0
R. E. Ginna 71249 (SA-1101)
N/A*
61782 (SA-847)
-27 0
Oconee Unit 1 61782 (SA-1135)
-27 0
71249 (SA-1229)
N/A*
0 72445 (SA-1585)
-27 0
Oconee Unit 2 406L44 (WF-154)
-27 0
299L44 (WF-25)
-27 0
Oconee Unit 3 821T44 (WF-200)
-27 0
72442 (WF-67)
N/A*
Point Beach Unit 1 71249 (SA-1101)
N/A*
61782 (SA-847)
-27 0
Point Beach Unit 2 72442 (SA-1484)
N/A' Surry Unit 1 72445 (SA-1585)
-27 0
299L44 (SA-1526)
-27 0
Surry Unit 2 72445 (SA-1585)
-27 0
Three Mile Island Unit 1 72105 (WF-70)
-26.5 0
299L44 (WF-25)
-27 0
299L44 (SA-1526)
-27 0
Turkey Point Unit 3 72442 (SA-1484)
N/A*
71249 (SA-1101)
N/A*
Turkey Point Unit 4 72442 (WF-67)
N/A*
71249 (SA-1101)
N/A*
Zion Unit 1 406L44 (WF-154)
-27 0
72105 (WF-70)
-26.5 0
Zion Unit 2 821T44 (WF-200)
-27 0
72105 (WF-70)
-26.5 0
71249 (SA-1769)
N/A*
- Surveillance data not used to calculate chemistry factor /RTp73 value.
B-12
m_.
l Table 6. Surveillance Data from B&W Integrated Reactor Vessel Surveillance Program t
L30 ft-lb Transition -
M
+
+
s L
Weld:
Weld Metal Capsule
- Fluence,
' Temperature, F LWire
. Identification '
Ident.
~
. ' cm -
Refemnce n/ 2 Initial Irradiated Delta.
299L44 SA-1526 - B&W Owners Group TMI2-LG1 8.30E+18 9
191 182 BAW-2253PS 22 SA-1526 - Surry Unit 1 T
2.81E+18
-15 150 165 WCAP-11415 V
1.94E+19
-15
.225 240 WCAP-ll415 WF Three Mile Island Unit 1 E
1.07E+18
-56 68-124 BAW-1901 l
5 23 C
8.66E+18
-56 147 203 BAW-1901 WF-26(6) - B&W Owners Group TMI2-LG1 9.68E+18 22 244 222 BAW-2253P i
i WF-25(9) - B&W Owners Group CR3-LG1 7.79E+18
-20 194 214 BAW-1910P5 24 406L44 WF-112 - Oconee Unit 1 E
1.50E+18
-5 73 78 BAW-2050 5 25 A
8.95E+18
-5 186 191 BAW-2050
[
C 9.86E+18
-5 180 185 BAW-2050 t
WF-112 - B&W Owners Group DBI-LG1 8.21E+18
-52 152 204 BAW-1920P s
{
S2 WF-193 - Arkansas Nuclear One E
7.27E+17 5
110 105 BAW-2075/R15 27 Unit 1 A
1.03E+19 5
156 151 BAW-2075/R1 C
1.46E+19 5
190 185 BAW-2075/R1 l
WF-193 - Rancho Seco Unit 1 B
3.99E+18
-14 85 99 BAW-207452s D
6.60E+18
-14 138 152 BAW-2074 F
1.42E+19
-14 152 166 BAW-2074
[
WF-193 - Point Beach Unit 2 V
7.12E+18 0
165 165 BAW-214052'; WCAP-12795/R2 '*
[
S T
8.97E+18 0
150 150 BAW-2140; WCAP-12795/R2 R
2.33E+19 0
235 235 BAW-2140; WCAP-12795/R2
- I S
3.47E+19 0
231 231 BAW-2140; WCAP-12795/R2 5
B-13 e
?
i Table 6 (cont'd). Surveillance Data from B&W Integrated Reactor Vessel Surveillance Program
- 30 ft-lb Transition JWeld -
' Weld Metal.
~ Capsule
- Fluence,
- Temperature, F
~
2
_ Wire '
Identification Ident.
_ n/cm Initial Irradiated
- Delta Reference 61782 SA-1036 - R. E. Ginna V
5.56E+13
-25 115 140 WCAP-13902 R
1.15E+19
-25 140 165 WCAP-13902 T
1.97E+19
-25 125 150 WCAP-13902 S
3.87E+19
-25 180 205 WCAP-13902 SA-1135 - B&W Owners Group DBI-LG1 1.03E+19
-39 103 142 BAW-1920P 71205 WF-70(N) - B&W Owners Group TMI2-LG1 5.85E+18 45 168 123 BAW-2253P DB1-LG1 6.63E+18 45 180 135 BAW-1920P S
l CR3-LG2 1.19E+19 45 170 125 BAW-2254P
WF-209 Oconee Unit 2 C
1.02E+18 4
49
-45 BAW-20515 32 A
3.37E+18 4
118 114 BAW-2051 E
1.21E+19 4
183 179 BAW-2051 i
WF-209 Oconee Unit 3 A
8.10E+17 45-93 48 BAW-2128/RI "
S B
3.12E+18 45 109 64 BAW-2128/R1 D
1.45E+19 45 185 140 BAW-2128/R1 WF-209 Zion Unit 1 T
2.87E+18 4
116 112 BAW-2082; WCAP-10962/R352' U
9.50E+18 4
203 199 BAW-2082; WCAP-10962/R3 X
1.16E+19 4
203 199 BAW-2082; WCAP-10962/R3 Y
1.57E+19 4
209 205 BAW-2082; WCAP-10962/R3 WF-209 Zion Unit 2 U
2.65E+18
-10 118 128 WCAP-123%; WCAP-10962/R3 T
7.68E+18
-10 165 175 WCAP-123%; WCAP-10962/R3 Y
1.45E+19
-10 210 220 WCAP-12396; WCAP-10962/R3 O
B-14
~
V
~
s t Table 6 (cont'd). Surveillance Data from B&W Integrated Reactor Vessel Surveillance Program
' 30 ft-lb Transition i
Weld Weld Metal Capsule
- Fluence, Temperature, F.
Wire Identification Ident.
.n/cm.f Initial Irradiated Delta.
. Reference 2
72445 SA-1263 - Point Beach Unit 1 V
5.02E+18
-45 65 110 WCAP-10736;WCAP-12794/R2 3' S
S 8.29E+18
-45 120 165 WCAP-10736; WCAP-12794/R2 l
R 2.38E+19
-45 120 165 WCAP-10736; WCAP-12794/R2 T
2.42E+19
-45 135 180 WCAP-10736; WCAP-12794/R2
~
SA-1585 - B&W Owners Group CR3-LG1 5.10E+18
-27 121 148 BAW-1910P CR3-LG2 1.67E+18
-27 141 168 BAW-2254P 821T44 WF-182 Davis-Besse F
1.96E+18
-11 116 127 BAW-212553' B
5.92E+18
-11 114 125 BAW-2125 A
1.29E+19
-11 164 175 BAW-2125 D
9.62E+18
-11 139 150 BAW-2125 P
P B-15
.m a
=
_.._w.,
The RTm values, where surveillance data are available, are presented in Table 7 for the B&WOG RVWG reactor vessel beltline weld materials. The results were calculated in accordance with the proposed 10CFR50.61 rule, and the calculational methods are briefly discussed below.
5.1.1 Initial Nil-Ductility Reference Temperature The IRTuor values used to calculate the RTm are listed in Table 5.
5.1.2 Irradiation Induced Change in Reference Temperature The irradiation induced change in reference temperature (ARTwor) is defined as the mean value of the adjustment in reference temperature caused by irradiation and is calculated as fo!!ows:
- m..
ARTuor = (cry x (g)
(2) where CF
= Chemistry Factor ff
= Fluence Factor 1
l 5.1.2.1 Chemistry Factor
\\
The data obtained from reactor vessel surveillance programs are 7
i used to develop a material-specific chemistry factor which is used f
in the calculation of ARTwor. The chemistry factor is calculated by multiplying each adjusted ARTuorvalue reported in the surveillance report by its corresponding fluence factor. The products of these
~
results are then summed and divided by the sum of the squares of the fluence factors. See Attachment A for the determination of the chemistry factors using surveillance data.
B-16
Table 7. Predicted RTm Values for the B&WOG RVWG Reactor Vessels for Which Surveillance Data is Available i
Weld' Wire Chemistry 1 Fluence
'ARTm.
Screening L
Plant *'
(Weld Identification)
Factor ~
Factor :
? (CF*ff) '
- IRTm.
Margin
- RTm CriteriaL 1
ANO-1 821T44 (WF-182-1) 162.1 0.958 155.3
-27 28 156.3 300 4%L44 (WF-112) 175.0 0.983 172.0
-27 28 173.0 300 CR3 71249 (SA-1769)
N/A" 300 j
i 72105 (WF-70) 138.4 0.945 130.8
-26.5 28 132.3 300 l
t DB 821T44 (WF-182-1) 162.1 1.019 165.2
-27 28 166.2 300 REG 71249 (SA-1101)
N/A" 300 61782 (SA-847) 147.2 1.338 197.0
-27 28 198.0 300 OCl 61782 (SA-1135) 147.2 0.451 66.4
-27 28 67.4 300 71249 (SA-1229)
N/A" 300-72445 (SA-1585) 149.8 0.960 143.8
-27 28 144.8 300 i
OC2 406L44 (WF-154) 175.0 0.952 166.6
-27 28 167.6 300 299L44 (WF-25) 216.9 0.976 211.7
-27 28 212.7 300 OC3 821T44 (WF-200) 162.1 0.946 153.3
-27 28 154.3 300 j
72442 (WF-67)
N/A" 300 l
PB1 71249 (SA-Il01)
N/A" 300 i
=
61782 (SA-847) 147.2 1.135 167.1
-27 28.
168.1 270 PB2 72442 (SA-1484)
N/A" 300 S1 72445 (SA-1585) 149.8 1.354 202.8
-27 28 203.8 300 299L44 (SA-1526) 216.9 0.874 189.6
-27 28 190.6 270 I
S2 72445 (SA-1585) 149.8 0.906 135.7
-27 28 136.7 270 1
B-17 i
f
\\
4
+
i
l.
l l
l Table 7 (cont'd). Predicted RTm Values for the B&WOG RVWG Reactor Vessels for Which Surveillance Data is Available
- Su Q
. Weld Wire -
Chemictry
-Fluence ART,a -
l Plant'
^(Weld Identification)
Fadtor; iFactor i T (CF'ff)
IRT m Margin
). RTm :
iCriteriaY TMI-1 72105 (WF-70) ~
138.4 0.934
.129.3
-26.5 28 130.8 300 299L44 (WF-25) 216.9 0.958 207.8
-27 28 208.8 300 299L44 (SA-1526) 216.9 0.926 200.8
-27 28
' 201.8 270 l
TP3 72442 (SA-1484)
N/A**
300 l
71249 (SA-1101)
N/A**
-300 j
TP4 72442 (WF-67)
N/A**
300 l
71249 (SA-1101)
N/A**
300 Z1 406L44 (WF-154) 175.0 1.053 184.3
-27 28 185.3 300 72105 (WF-70) 191.5 1.151 220.4
-26.5 28 221.9 300 Z2 821T44 (WF-200) 162.1 1.073 173.9
-27 28 174.9 300 l
72105 (WF-70) 191.5 0.859 164.5
-26.5 28 166.0 270' l
71249 (SA-1769)
N/A' 300
- Plant Identification:
ANO-1
- Arkansas Nuclear One Unit 1 PB2
- Point Beach Unit 2 CR3
- Crystal River Unit 3 S1
- Surry Unit 1 DB
- Davis-Besse S2
- Surry Unit 2 REG
- R. E. Ginna TMI-I
- Three Mile Island Unit 1 OCl
- Oconee Unit 1 TP3
- Turkey Point Unit 3 OC2
- Oconee Unit 2 TP4
- Turkey Point Unit 4 OC3
- Oconee Unit 3 Z1
- Zion Unit 1 PB1
- Point Beach Unit 1 Z2
- Zion Unit 2
- Surveillance data not used to calculated chemistry factor value.
8 B-18.
s
,...u
b 5.1.2.2 Fluence Factor The fluence factor (ff) is determined as follows:
g, fo.2s - c.10 les />
(3) 2 where f
= fluence x 10 (n/cm, E > 1 MeV) 5.1.3 Margin The " margin" term is the quantity that is added to obtain conservative, upper-bound values of the PTS reference temperature for the calculations required by 10CFR50, Appendix G. The margin is determined by the following expression:
2 2
(4)
Margin = 2\\
f+og o
where e,
= standard deviation for the initial RTuor o,
= standard deviation for the ARTuor If a measured value of initial RTuor for the material in question is available, o, is zero because the measured initial value is an absolute value and it is assumed to have no error. If generic values ofinitial RTuor are used, o, is the standard deviation obtained from the set of data used to establish the mean value.
When surveillance data is used to calculate the ARTuor, the c, is 14*F for 3
weld metals, except that c3 need not exceed 0.50 times the mean value of ARTuo7 B-19
5.2 Regulatory Guide 1.99, Revision 2, Position 2.1, Ratio Procedure Position 2.1 of Regulatory Guide 1.99, Revision 2, states that if there is clear evidence that the copper and nickel contents of the surveillance weld differs from that of the vessel weld,' measured values of ARTuor should be adjusted by multiplying them by the ratio of the chemistry factor for the vessel weld to that for the surveillance weld.
To determine the ratios for adjusting the measured values of ARTwor, the chemistry factors for the available surveillance weld metals and the weld wire best-estimate means were calculated using 10CFR50.61, Table 1. The chemistry factors for the surveillance weld metals were calculated using the mean copper and nickel contents of the reported values listed in Table 3, while the chemistry factors for the best-estimate weld wire means were calculated using the data in Table 2.
The data used to determine the ratios for the vessel weld (best-estimate mean) to that for the surveillance weld are presented in Table 8. To adjust the measured values of ARTuor for the available surveillance data, the measured shift in the 30 ft-lb transition temperature were multiplied by their respective chemistry factor ratio. The adjusted measured ARTwor are presented in Table 9, and these values are used to determine the chemistry factor in accordance with Position 2.1 of Regulatory Guide 1.99, Revision 2.
Using the available surveillance data and the ratio procedure stated in Regulatory Guide 1.99, Revision 2, Position 2.1, the RTrrs values for the B&WOG RVWG reactor vessel beltline weld metals are presented in Table 10. The results were calculated in accordance with the proposed 10CFR50.61 rule, and the calculational methods are identical to those discussed in Section 5.1 with the exception of calculating the chemistry factor in which a ratio adjustment was made to the measured ARTwor values.
See Attachment B for the determination of the chemistry factors after applying the ratio adjustment.
f B-20 i
y..
Table 8. Ibtio of the Weld Wire Best-Estimate Mean Chemistry Factor to That of the Surveillance Weld Chemistry Factor Chemistry Factor from Ratio Weld 10CFR50.61, (Vessel to Wire Weld Id.
Cu Ni Table 1 Sury. Data) 299L44 SA-1526 (B&W Owners Group) 0.37 0.70 234.0 0.956 SA-1526 (Surry-1) 0.25' O.66*
185.9 1.203 WF-25 (Three Mile Island-1) 0.33 0.66 213.7 1.046 WF-25(6) (B&W Owmers Group) 0.35 0.67 222.2 1.007 WF-25(9) (B&W Owmers Group) 0.35 0.70 226.5 0.987 Weld Wire Heat Best-Estimate 0.35 0.68 223.6 406L44 WF-112 (Oconee-1) 0.32 0.59 200.6 0.980 WF-112 (B&W Owners Group) 0.32 0.59 200.6 0.980 WF-193 (Arkansas Nuclear One-1) 0.28 0.59 185.6 1.060 WF-193 (Rancho Seco-1) 0.31 0.59 196.7 1.000 WF-193 (Point Beach-2) 0.25 0.59 174.6 1.127 Weld Wire Heat Best-Estimate 0.31 0.59 196.7 61782 SA-1036 (R. E. Ginna) 0.23' O.53*
158.9 1.055 SA-1135 (B&W Owners Group) 0.27 0.59 182.6 0.918 Weld Wire Heat Best-Estimate 0.25 0.54 167.6 72105 WF-70(N) (B&W Owners Group) 0.42 0.59 229.8 0.917 WF-209-1 (Oconee-2) 0.36 0.58 213.5 0.987 WF-209-1 (Oconee-3) 0.30 0.58 191.3 1.102 WF-209-1 (Zion-1) 0.28**
0.55**
180.3 1.169 WF-209-1 (Zion-2) 0.26" 0.5 5**
172.8 1.220 Weld Wire Heat Best-Estimate 0.35 0.59 210.8 72445 SA-1263 (Point Beach-1) 0.23' O.62*
172.4 0.942 SA-1585 (B&W Owmers Group) 0.21 0.59 162.4 1.000 Weld Wire Heat Best-Estimate 0.21 0.59 162.4 821T44 WF-182-1 (Davis-Besse) 0.21 0.63 169.0 1.053 l
Weld Wire Heat Best-Estimate 0.24 0.63 178.0
- Mean value from data in Table 3.
" Mean value based on the mean from individual capsule chemical analyses.
B-21
.5 Table 9. Adjustment of Measured Surveillance Data ARTwor Using Ratio Procedure of Regulatory Guide 1.99, Revision 2 Measured Normalized -
Weld Wire Heat Number -
ARTwor, F~
. CF, Ratio :
- ARTwor for (Weld Identifications)
Cap.* '
1(see Table 6);
(See. Table 8)
Weld Wire -
299L44 Tl 182 0.956 174.0 (SA-1526 & WF-25)
SI-T 165 1.203 198.5 SI-V 240 1.203 288.7 TMlE 124 1.046 129.7 TMI-C 203 1.046 212.3 C1 214 0.987 211.2 Tl 222 1.007 223.6 406L44 OCl-E 78 0.980 76.4 (WF-112 & WF-193)
OCl-A 191 0.980 187.2 OCl-C 185 0.980 181.3 D1 204 0.980 199.9 AN1-E 105 1.060 111.3 AN1-A 151 1.060 160,1 AN1-C 185 1.060 196.1 RS1-B 99 1.000 99.0 RSI-D 152 1.000 152.0 RSI-F 166 1.000 166.0 PB2-V 165 1.127 186.0 PB2-T 150 1.127 169.1 PB2-R 235 1.127 264.8 PB2-S 231 1.127 260.3 61782 REG-V 140 1.055 147.7 (SA-1036 & SA-1135)
REG-R 165 1.055 174.1 REG-T 150 1.055 158.3 REG-S 205 1.055 216.3 D1 142 0.918 130.4 72105 (B&W Design Only)
Tl 123 0.917 112.8 (WF-70 & WF-209-1)
D1 135 0.917 123.8 C2 125 0.917 114.6 OC2-C 45 0.987 44.4 OC2-A 114 0.987 112.5 OC2-E 179 0.987 176.7 OC3-A 48 1.102 52.9 OC3-B 64 1.102 70.5 OC3-D 140 1.102 154.3 i
B-22
i Table 9 (cont'd). Adjustment of Measured Surveillance Data ARTum Using Ratio Procedure of Regulatory Guide 1.99, Revision 2 i
i
' Normalized Measured l
Weld Wire Heat Number
~
ARTum,F.
- CF Ratio
'ARTum for -
j (Weld Identifications) '
Cap.*
- (see Table 6)s (SeeTable 8)
Weld Wire
)
72105 (W Design Only)
Zl-T 112 1.169 130.9 i
j (WF-70 & WF-209-1)
Zl-U 199 1.169 232.6 Zl-X 199 1.169 232.6 i
Zl-Y 205 1.169 239.6 Z2-U 128 1.220 156.2 Z2-T 175 1.220 213.5 Z2-Y 220 1.220 268.4 72445 PB1-V 110 0.942 103.6 (SA-1263 & SA-1585)
PBl-S 165 0.942 155.4
}
PB1-R 165 0.942 155.4 PB1-T 180 0.942 169.6 i
Cl 148 1.000 148.0 C2 168 1.009 168.0 821T44 TEl-F 127 1.053 133.7 L
(WF-182-1)
TEl-B 125 1.053 131.6 TEl-A 175 1.053 184.3 TEl-D 150 1.053 158.0
- - Irradiation Capsule Identification:
T1 = B&WOG Capsule TM12-LG1 S1 = Surry Unit 1 Cl = B&WOG Capsule CR3-LG1 TMI = Three Mile Island Unit 1 ANI = Arkansas Nuclear One Unit 1 OCl = Oconee Unit 1 RS1 = Rancho Seco Unit 1 D1 = B&WOG Capsule dbl-LG1 PB2 = Point Beach Unit 2 REG = R. E. Ginna i
OC2 = Oconee Unit 2 OC3 = Oconee Unit 3 Z1 = Zion Unit 1 Z2 = Zion Unit 2 PB1 = Point Beach Unit 1 TEl = Davis-Besse B-23
\\
1 i
Table 10. Predicted RT Values for the B&WOG RVWG Reactor Vessels Using m
Ratio Procedure and Surveillance Data Screening -
Weld Wire Adjusted
' Fluence'-
ARTor; i
Plant (Weld Identification)
Chem. Factor Factor,
- (CF'ff)
IRTer _
Margin; J RTm tCriteria' ANO-1 821T44 (WF-182-1) 170.7 0.958 163.5
-27 28 164.5 300 406L44 (WF-112) 184.5 0.983 181.4
-27 28 182.4 300 i
CR3 71249 (SA-1769)
N/A" 300 72105 (WF-70) 136.6 0.945 129.1
-26.5 28 130.6 300 DB 821T44 (WF-182-1) 170.7 1.019 173.9
-27 28 174.9 300 REG 71249 (SA-1101)
N/A" 300 61782 (SA-847) 152.1 1.338 203.5
-27 28 204.5 300 OCl 61782 (SA-ll35) 152.1 0.451 68.6
-27 28 69.6 300 71249 (SA-1229)
N/A" 300 72445 (SA-1585) 143.8 0.960 138.0
-27 28 139.0 300-OC2 406L44 (WF-154) 184.5 0.952 175.6
-27 28 176.6 300 299L44 (WF-25) 231.4 0.976 225.8
-27 28 226.8 300 OC3 821T44 (WF-200) 170.7 0.946 161.5
-27 28 162.5 300 72442 (WF-67)
N/A" 300 PB1 71249 (SA-1101)
N/A" 300 61782 (SA-847) 152.1 1.135 172.6
-27 28 173.6 270 PB2 72442 (SA-1484)
N/A" 300 S1 72445 (SA-1585) 143.8 1.354 194.7
-27 28 195.7 300 299L44 (SA-1526) 231.4 0.874 202.2
-27 28 203.2 270 S2 72445 (SA-1585) 143.8 0.906 130.3
-27 28 131.3 270 e
B-24
(
m Table 10 (cont'd). Predicted RTm Values for the B&WOG RVWG Reactor Vessels Using Ratio Procedure and Surveillance Data Weld Wire Adjusted Fluence ARTer Screening Plant (Weld Identification)
Chem. Factor Factor
' (CF*ff)
IRTer
- Margin RTm
. Criteria TMl-1 72105 (WF-70) 136.6 0.934 127.6
-26 5 28 129.1 300 299L44 (WF-25) 231.4 0.958 221.7
-27 28 222.7 300 299L44 (SA-1526) 231.4 0.926 214.3
-27 28 215.3 270 TP3 72442 (SA-1484)
N/A" 300 71249 (SA-1101)
N/A" 300 TP4 72442 (WF-67)
N/A" 300 71249 (SA-1101)
N/A" 300 Z1 406L44 (WF-154) 184.5 1.053 194.3
-27 28 195.3 300 72105 (WF-70) 227.8 1.151 262.2
-26.5 28 263.7 300 Z2 821T44 (WF-200) 170.7 1.073 183.2
-27 28 184.2 300 72105 (WF-70) 227.8 0.859 195.7 26.5 28 197.2 270 71249 (SA 1769)
N/A*
300
^
- Plant Identification:
ANO-1
- Arkansas Nuclear One Unit 1 PB2
- Point Beach Unit 2 CR3
- Crystal River Unit 3 SI
- Surry Unit 1 DB
- Davis-Besse S2
- Surry Unit 2 REG
- R. E. Ginna TMI-l
- Three Mile Island Unit i OC1
- Oconee Unit 1 TP3
- Turkey Point Unit 3 OC2
- Oconee Unit 2 TP4
- Turkey Point Unit 4 OC3
- Oconee Unit 3 Z1
- Zion Unit 1 PB1
- Point Beach Unit 1 Z2
- Zion Unit 2 Surveillance data not used to calculated chemistry factor value.
i B-25 I
l I
l l
ATTACIIMENT A Chemistry Factor Determination Using Surveillance Data l
l l
l l
l I
1 l
l 1
I B-26 l
l
~
v Table A-1. Calculation of Weld Metal Chemistry Factors Using Surveillance Data Weld Wire Heat Number ARTer (Weld Identifications)
Cap."
Fluence ARTer ff (x) ff ff' 299L44 Tl 8.30E+18 182 0.948 172.5 0.899 (SA-1526 & WF-25)
St-T 2.81E+18 165 0.653 107.7 0.426 SI-V 1.94E+19 240 1.181 283.4 1.395 Cl 7.79E+18 214 0.930 199.0 0.865 Tl 9.68E+18 222 0.991 220.0 0.982 TMI-E 1.07E+18 124 0.431 53.4 0.186 TMI-C 8.66E+18 203 0.960 194.9 0.922 SUM 1,230.9 5.675
)
CF = E(ART or*ff) / E(ff*) = 216.9 f
u 406L44 AN1-E 7.27E+17 105 0.356 37.4 0.127 (WF-112, WF-154, &
AN1-A 1.03E+19 151 1.008 152.2 1.016 WF-193)
AN1-C 1.46E+19 185 1.105 204.4 1.221 OCl-E 1.50E+18 78 0.503 39.2 0.253 OCl-A 8.95E+18 191 0.969 185.1 0.939 OCl-C 9.86E+18 185 0.996 184.3 0.992 RSI-B 3.99E+18 99 0.745 73.8 0.555 RSI-D 6.60E+18 152 0.884 134.4 0.781 RSI-F 1.42E+19 166 1.097 182.1 1.203 D1 8.21 E+18 204 0.945 192.8 0.893 PB2-V 7.12E+18 165 0.905 149.3 0.819 PB2-T 8.97E+18 150 0.970 145.5 0.941 PB2-R 2.33E+19 235 1.229 288.8 1.510 PB2-S 3.47E*19 231 1.325 306.1 1.756 SUM 2,275.4 13.006 2
CF = E(ART,snr*ff) / E(ff) = 175.0 B-27
Weld Wire Heat Number ARTwor (Weld Identifications) -
Cap.*
Fluences
. ARTyor ff
. (x) ff ffS 61782 D1 1.03E+19 142 1.008 143.1 1.016 (SA-847, SA-848, SA-1036, REG-V 5.56E+18 140 0.836 117.0 0.699 SA-1135, & SA-1788)
REG-R 1.15E+19 165 1.039 171.4 1.080 REG-T 1.97E+19 150 1.185 177.8 1.404 REG-S 3.87E+19 205 1.349 276.5 1.820 SUM 885.8 6.019 2
CF = E(ARTsor*ff) / E(ff) = 147.2 72105 - B&W Design Only OC2-C 1.02E+18 45 0.421 18.9 0.177 (WF-70 & WF-209-1)
OC2-A 3.37E+18 114 0.701 79.9 0.491 OC2-E 1.21E+19 179 1.053 188.5 1.109 OC3-A 8.10E+17 48 0.376 18.0 0.141 OC3-B 3.12E+18 64 0.680 43.5 0.462 OC3-D 1.45E+19 140 1.103 154.4 1.217 Tl 5.85E+18 123 0.850 104.6.
0.723 D1 6.63E+18 135 0.885 119.5 0.783 [-
C2 1.19E+19 125 1.049 131.1 1.100 SUM 858.4 6.203 2
CF = E(ARTuor*ff) / E(ff ) = 138.4 72105 E Design Only Zl-T 2.87E+18 112 0.659 73.8 0.434 (WF-70 & WF-209-1)
Zl-U 9.50E+18 199 0.986 196.2 0.972 i
Zl-X 1.16E+19 199 1.041 207.2 1.084 i
j Zl-Y 1.57E+19 205 1.125 230.6 1.266 l
Z2-U 2.65E+18 128 0.639 81.8 0.408 i
Z2-T 7.68E+18 175 0.926 162.1 0.857 Z2-Y 1.45E+19 220 1.103 242.7 1.217 SUM 1,194.4 6.238 2
CF = E(ART nr*ff) / E(ff ) = 191.5 3
i B-28
~
l l
Weld Wire Heat Number ARTor u
(Weld Identifications)
Cap.*
Fluence ARTwor ff (x) ff ff' 72445 Cl 5.10E+18 148 0.812 120.2 0.659 (SA-1263, SA-1585, C2 1.67E+19 168 1.141 191.7 1.302 SA-1650, & WF-9)
PBI-V 5.02E+18 110 0.808 88.9 0.653 PBI-S 8.29E+18 165 0.947 156.3 0.897 PB1-R 2.38E+19 165 1.234 203.6 1.523 PB1-T 2.42E+19 180 1.238 222.8 1.533 SUM 983.5 6.567 CF = E(ARTuor*ff) / E(ff) = 149.8 821T44 TEl-F 1.96E+18 127 0.565 71.8 0.319 (WF-182-1 & WF-200)
TEl-B 5.92E+18 125 0.853 106.6 0.728 TEl-A 1.29E+19 175 1.071 187.4 1.147 TEl-D 9.62E+18 150 0.989 148.4 0.978 SUM 514.2 3.172 CF = E(ARTsorq / RM) =.J62.1
- - Irradiation Capsule Identification:
Tl = B&WOG Capsule TM12-LG1 S1 = Surry Unit 1 C1 = B&WOG Capsule CPJ-LG1 TMI = Three Mile Island Unit 1 ANI = Arkansas Nuclear One Unit 1 OCl = Oconee Unit 1 RS1 = Rancho Seco Unit 1 D1 = B&WOG Capsule dbl-LG1
)
I B2 = Point Beach Unit 2 REG = R. E. Ginna OC2 = Oconee Unit 2 OC3 = Oconee Unit 3 Z1 = Zion Unit 1 Z2 = Zion Unit 2 PB1 = Point Beach Unit 1 TEl = Davis-Besse a
B-29
ATTACHMENT B a'
Chemistry Factor Determination Using Ratio Procedure and Surveillance Data in Accordance with Regulatory Guide 1.99, Revision 2, Position 2.1 i
l 3
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B-30
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l Table B-1. Calculation of Weld Metal Chemistry Factors Using Ratio Procedure and Surveillance Data Weld Wire Heat Number Ratio Adjusted ARTwor (Weld Identifications)'
Cap.*
Fluence
. ARTwor ff (x) ff ff' 299L44 Tl 8.30E+18 174.0 0.948 165.0 0.899 (SA-1526 & WF-25)
SI-T 2.81E+18 198.5 0.653 129.6 0.426 j
SI-V 1.94E+19 288.7 1.181 341.0 1.395 Cl 7.79E+18 211.2 0.930 196.4 0.865 T1 9.68E+18 223.6 0.991 221.6 0.982 TMI-E 1.07E+18 129.7 0.431 55.9 0.186 TMI-C 8.66E+18 212.3 0.960 203.8 0.922 SUM 1,313.3 5.675 2
CF = E(ARTuor*ff) / E(ff) = 231.4 406L44 AN1-E 7.27E+17 111.3 0.356 39.6 0.127 (WF-112, WF-154, &
AN1-A 1.03E+19 160.1 1.008 161.4 1.016 WF-193)
AN1-C 1.46E+19 196.1 1.105 216.7 1.221 OCl-E 1.50E+18 76.4 0.503 38.4 0.253 OCl-A 8.95E+18 187.2 0.969 181.4 0.939 OC1-C 9.86E+18 181.3 0.996 180.6 0.992 RSl-B 3.99E+18 99.0 0.745 73.8 0.555 RS1-D 6.60E+18 152.0 0.884 134.4 0.781 RSI-F 1.42E+19 166.0 1.097 182.1 1.203 D1 8.21E+18 199.9 0.945 188.9 0.893 PB2-V 7.12E+18 186.0 0.905 168.3 0.819 PB2-T 8.97E+18 169.1 0.970 164.0 0.941 PB2-R 2.33E+19 264.8 1.229 325.4 1.510 l
PB2-S 3.47E+19 260.3 1.325 344.9 1.756 h
SUM 2,399.9 13.006 2
CF = E(ARTuor*ff) / E(ff ) = 184.5 t
B-31
'.3 Weld Wire Heat Number Ratio Adjusted ARTuor (Weld Identifications)
Cap.*
Fluence ARTwor ff (X) ff ff 61782 D1 1.03E+19 130.4 1.008 131.4 1.016 (SA-847, SA-848, SA-1036, REG-V 5.56E+18 147.7 0.836 123.5 0.699 SA-1135, & SA-1788)
REG-R 1.15E+19 174.1 1.039 180.9 1.080 l
l REG-T 1.97E+19 158.3 1.185 187.6 1.404 REG-S 3.87E+19 216.3 1.349 291.8 1.820 SUM 915.2 6.019 CF = E(ART
- ff) / E(ff) = 152.1 331 72105 - B&W Design Only OC2-C 1.02E+18 44.4 0.421 18.7 0.177 (WF-70 & WF-209-1)
OC2-A 3.37E+18 112.5 0.701 78.9 0.491 OC2-E 1.21E+19 176.7 1.053 186.1 1.109 OC3-A 8.10E+17 52.9 0.376 19.9 0.141 4
OC3-B 3.12E+18 70.5 0.680 47.9 0.462 OC3-D 1.45E+19 154.3 1.103 170.2 1.217 Tl 5.85E+18 112.8 0.850 95.9 0.723 i
D1 6.63E+18 123.8 0.885 109.6 0.783 o
C2 1.19E+19 114.6 1.049 120.2 1.100 SUM 847.4 6.203 I
CF = E(ARTuor*fi) / E(ff') = 136.6 72105_W Design Only Zl-T 2.87E+18 130.9 0.659 86.3 0.434 (WF-70 & WF-209-1)
Zl-U 9.50E+18 232.6 0.986 229.3 0.972 Zl-X 1.16E+19 232.6 1.041 242.1 1.084 Zl-Y 1.57E+19 239.6 1.125 269.6 1.266 Z2-U 2.65E+18 156.2 0.639 99.8 0.408 Z2-T 7.68E+18 213.5 0.926 197.7 0.857 I
Z2-Y 1.45E+19 268.4 1.103 296.0 1.217 SUM 1,420.8 6.238 2
CF = E(ART or*ff) / E(ff) = 227.8 u
4 4
J B-32
Weld Wire Heat Number Ratio Adjusted ARTuor (Weld Identifications)
Cap.*
Fluence ARTuor ff (x) ff ff 72445 Cl 5.10E+18 148.0 0.812 120.2 0.659 (SA-1263, SA-1585, C2 1.67E+19 168.0 1.141 191.7 1.302 SA-1650, & WF-9)
PB1-V 5.02E+18 103.6 0.808 83.7 0.653 PB1-S 8.29E+18 155.4 0.947 147.2 0.897 PB1-R 2.38E+19 155.4 1.234 191.8 1.523 PB1-T 2.42E+19 169.6 1.238 210.0 1.533 SUM 944.6 6.567 i
CF = E(ARTuor*ff) / E(ff ) = 143.8 821T44 TEl-F 1.96E+18 133.7 0.565 75.5 0.319 (WF-182-1 & WF-200)
TEl-B 5.92E+18 131.6 0.853 112.3 0.728 TEl-A 1.29E+19 184.3 1.071 197.4 1.147
~
TEl-D 9.62E+18 158.0 0.989 156.3 0.978 SUM 541.5 3.172 i
CF = E(ARTuor*ff) / E(ff ) = 170.7
- - Irradiation Capsule Identification:
Tl = B&WOG Capsule TM12-LG1 S1 = Surry Unit 1 C1 = B&WOG Capsule CR3-LG1 TMI = Three Mile Island Unit 1 ANI = Arkansas Nuclear One Unit 1 OCl = Oconee Unit 1 RSl = Rancho Seco Unit 1 Di = B&WOG Capsule dbl-LG1 PB2 = Point Beach Unit 2 REG = R. E. Ginna OC2 = Oconee Unit 2 OC3 = Oconee Unit 3 21 = Zion Unit 1 Z2 = Zion Unit 2 PB1 = Point Beach Unit 1 TEl = Davis-Besse B-33
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ATTACIIMENT C j
l References 4
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i B-34 J
-- --.- - ~. -.. - - -. - - - - -
B-1.
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B-2.
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{
Against Pressurized Thermal Shock Events," Federal Register, October 4,1994.
i
}
B-3.
K. E. Moore and A. S. Heller, " Chemistry of 177-FA B&W Owners' Group Reactor Vessel Beltline Welds," BAW-1500. Babcock & Wilcox's Nuclear Power _ Generation.
Division, Lynchburg, Virginia, September 1978.*
)
B-4.. L. B. Gross, " Chemical Composition of B&W Fabricated Reactor Vessel Beltline Welds,"
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j B-5.
M. J. DeVan and K. K. Yoon, " Response to Closure Letters to Generic Letter 92-01, l
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]-
1994.
I B-6.
J. M. Chicots, S. L. Anderson, and A. Madeyski, " Analysis of Capsule S from the Rochester Gas and Electric Corporation R. E. Ginna Reactor Vessel Radiation i
Surveillance Program," WCAP-13902, Westinghouse Electric Corporation, Pittsburgh, i
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l B-7.
L. S. Harbison, " Master Integrated Reactor Vessel Surveillance Program," BAW-1543.
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L. S. Harbison, " Supplement to the Master Integrated Reactor Vessel Surveillance Program," BAW-1543. Revision 4. Suoolement 1. B&W Nuclear Technologies, Inc.,
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i B-9.
S. E. Yanichko and V. A. Perone, " Analysis of Capsule V from the Virginia Electric and l
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i i
}
B-35 1
4 i
i
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B-12. S. E. Yanichko, et al.. " Analysis of Capsule T from the Rochester Gas and Electric Corporation R. E. Ginna Nuclear Plant Reactor Vessel Radiation Surveillance Program,"
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1 r
B-13. S. E. Yanichko, " Florida Power and Light Company Turkey Point Unit No. 4 Reactor Vessel Radiation Surveillance Program," WCAP-7660. Westinghouse Electric Corporation, Pittsburgh, Pennsylvania, May 1971.
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WCAP-10736, Westinghouse Electric Corporation, Pittsburgh, Pennsylvania, December 1984.
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B&W Nuclear Technologies, Inc., Lynchburg, Virginia, September 1993.
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B&W Nuclear Technologies, Inc., Lynchburg, Virginia, (Scheduled for publication).
B-36 e
B-23. A. L. Lowe, Jr., st_al, " Analysis of Capsule TMil-C GPU Nuclear Three Mile Island
)
i Nuclear Station Unit 1, Reactor Vessel Materials Surveillance Program," BAW-1901.
Babcock & Wilcox's Nuclear Power Division, Lynchburg, Virginia, March 1986.*
4 i
B-24. A. L. Lowe, Jr., flaL, " Analysis of Capsule CR3-LG1 Babcock & Wilcox Owners Group, i
Integrated Reactor Vessel Materials Surveillance Program," BAW-1910P. Babcock &
Wilcox's Nuclear Power Division, Lynchburg, Virginia, August 1986.*
B-25. A. L. Lowe, Jr., et al., " Analysis of Capsule OCl-C Duke Power Company Oconee Nuclear Station Unit 1, Reactor Vessel Materials Surveillance Program," BAW-2050.
j Babcock & Wilcox's Nuclear Power Division, Lynchburg, Virginia, October 1988.*
l j
B-26. A. L. Lowe, Jr., et al., " Analysis of Capsule DB1-LG1 Babcock & Wilcox Owners Group, J
Integrated Reactor Vessel Materials Surveillance Program," BAW-1920P. Babcock &
Wilcox's Nuclear Power Division, Lynchburg, Virginia, October 1986.*
i l
B-27. A. L. Lowe, Jr., et al., " Analysis of Capsule AN1-C Arkansas Power & Light Company l
Arkansas Nuclear One Unit 1, Reactor Vessel Materials Surveillance Program," BAW-2075. Revision 1, Babcock & Wilcox's Nuclear Power Division, Lynchburg, Virginia, j
October 1989.*
j B-28. A. L. Lowe, Jr., st_L, " Analysis of Capsule RSI-F Sacramento Municipal Utility District i
Rancho Seco Unit 1, Reactor Vessel Materials Surveillance Program," BAW-2074.
j l
Babcock & Wilcox's Nuclear Power Division, Lynchburg, Virginia, April 1989.*
l B-29. A. L. Lowe, Jr., et al,. " Analysis of Capsule S Wisconsin Electric Power Company Point Beach Unit No. 2, Reactor Vessel Materials Surveillance Program," BAW-2140, B&W i
Nuclear Technologies, Inc., Lynchburg, Virginia, August 1991, i
B-30. S. L. Anderson and A. H. Fero, " Reactor Cavity Neutron Measurement Program for Wisconsin Electric Power Company Point Beach Unit 2, WCAP-12795. Revision 2, j
Westinghouse Electric Corporation, Pittsburgh, Pennsylvania, September 1992.
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Owners Group Master Integrated Reactor Vessel Surveillance Program," BAW-2254P, B&W Nuclear Technologies, Inc., Lynchburg, Virginia, (Scheduled for publication).
B-32. A. L. Lowe, Jr., et al., " Analysis of Capsule OCII-E Duke Power Company Oconee Nuclear Station Unit 2, Reactor Vessel Materials Survcillance Program," BAW-2051.
i Babcock & Wilcox's Nuclear Power Division, Lynchburg, Virginia, October 1988.*
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Revision 1, B&W Nuclear Technologies, Inc., Lynchburg, Virginia, May 1992.
l~
B-37 L
- w.,
)
i B-34. J. M. Chicots, et al.. " Zion Units 1 and 2 Reactor Vessel Fluence and RTm Evaluations,"
WCAP-10962. Revision 3. Westinghouse Electric Corporation, Pittsburgh, Pennsylvania,-
September 1991.
B-35. S. L. Anderson, " Reactor Cavity Neutron Measurement Program for Wisconsin Electric Power Company Point Beach Unit 1, WCAP-12794. Revision 2, Westinghouse Electric Corporation, Pittsburgh, Pennsylvania, December 1992.
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B-38 r
,