ML20101J787

Provides Summary of Data Applicable to Point Beach,Units 1 & 2 from Encl B&W Rept BAW-2166, B&Wog Response to GL 92-01, in Response to GL 92-01, Reactor Vessel Structural Integrity, Per Commitments in Response to GL 88-11

ML20101J787
Person / Time
Site:	Point Beach
Issue date:	06/25/1992
From:	Link B WISCONSIN ELECTRIC POWER CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
CON-NRC-92-070, CON-NRC-92-70 GL-88-11, GL-92-01, GL-92-1, VPNPD-92-231, NUDOCS 9207010243
Download: ML20101J787 (10)

Text

myw Wisconsin - -

Electnc POWER COMPANY 231 w Metagon. Po Bem 204uwaukee.wi 53201 m 2212345 VPNPD-92-231 NRC-92-070 June 25, 1992 U. S. NUCLEAR REGULATORY COMMISSION Document Control Desk Mail Station P1-137 Washington, DC 20555 Gentlemen:

DOCKETS 50-266 AND 50-301 RESPONSE TO NRC GENERIC LETTER 92-01. REVISION 1 REACTOR VESSEL srRUCTURAL INTEGRITY. 10 CFR-50.54(fl NRC Generic Letter 92-01, " Reactor-Vessel Structural-Integrity,"

dated March 6, 1992, was issued to obtain information-from licensees to enable the NRC_to assess compliance with regulatory requirements and commitments regarding reactor _ vessel integrity. A response-to Generic Letter 92-01 was requested within 120 days of the date of-issuance. We. understand that Generic Letter 92-01 was' issued in view of certain concerns raised during NRC staff's review of reactor vessel integrity for the Yankee Nuclear Power Station.

The Babcock and Wilcox Owners Group's Reactor Vessel Working Group, under the direction of Wisconsin Electric and other member utilities, developed report BAW-2166, '_'B&W: Ownefs Group Response to Generic Letter 92-01," which is enclosed.- This report provides the information requested by Generic Letter 92-01 and was-forwarded:to the NRC by B&W Nuclear Service Company on June 17,'1992. .

Generic Letter 92-01 presents the information: requested in three sections (1, 2, items. and 3) which_-are further divided into a number of

'A tabular response' format:is'u' sed--in BAW-2166 to respond to the individual sections and items-contained in the Generic-Letter.

The response format-is delineated in Chapters 3-and 6.of.BAW-2166.

Wisconsin Electric sponsored and directed theidevelopment of.BAW-2166 and has endorsed the data contained in.BAW-2166 regarding our Point Beach Nuclear Plant. We'believe the. data contained in BAW-2166 regarding Point Beach Nuclear Plant satisfactorily' responds lto the information requested in Generic Letter 92-01. :A summary of-the

. data from that report, which is applicable to Point-Beach, Units 1 and 2,-is provided in-the following_ paragraphs.

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Document Control Desk

-June 25, 1992 Page 2 Section 1 of the Generic Letter requests information related to the licensee's surveillance program pursuant to Appendix H to 10 CFR Part 50. Section 1 is not applicable to Point Beach, Units 1 and 2 because they are currently part of an NRC-approved integrated surveillance program as listed in Enclosure 2 to the Generic Letter.

Table 1 in the Point Beach, Units 1 and 2 chapters of BAW-2166 addresses the issue identified in Section 1.

Section 2 of Generic Letter 92-01 is divided into Items a and b.

Item b contains a number of subitems - 1 through 6. Section 2, Item a discusses plants for which the Charpy' upper shelf energy is predicted to be less than 50 foot-pounds at the end of their current license period using the guidance of Regulatory Guide 1.99, Revision

2. Item a, asks addressees to provide upper shelf energy values for the limiting beltline weld and plate or forging. This data is provided in able 2 of BAW-2166 for the chapters applicable to PBNP Units 1 and 2. As noted in Table 2, both Point Beach units are projected to drop below 50 foot-pounds prior to the end of their current licensed life, using the guidance provided in Regulatory Guide 1.99 Revision 2. An analysis for our Point Beach Nuclear Plants to demonstrate adequate margins of safety to that_ required in ASME Section III-Appendix G is scheduled to be performed in 1993 under the sponsorship of the B&W Owners Group Reactor Vessel Working Group. The owners Group has completed analyses that have demonstrated the required margin of safety for the Zion and Turkey Point plants and have submitted the results to the NRC. The results of these-analyses are anticipated to bound the outcome of both Point Beach units. Additionally, as previously reported in our Point Beach Nuclear Plant Unit 2 surveillance capsule S report dated October 15, 1991, correlations.have been developed by the Owners Group for predicting the effects of neutron irradiation on Linde 80 Submerged Arc Welds. These results were reported in BAW-1803, Revision 1, " Correlations for Predicting the Effects of Neutron

~

Irradiation on Linde 80 Submerged-Arc Welds," which was transmitted directly from B&W Nuclear Service Company to the NRC on October 4, 1991. This report demonstrates that for both Point Beach Nuclear Plant Unit 1 and Unit 2, the mean value of the upper shelf energy for the controlling weld metal will-not decrease below 50 ft-lbs during the current 40-year. license.

Section 2, Item b, requests licensees whose reactor vessels were constructed to an ASME code earlier than the Summer 1972 Agenda of the 1971 Edition to describe considerations given-to certain material properties described in Subitems 1 through 6. As stated in Chapter 5 of BAW-2166, both Point Beach reactor vessels were constructed to the 1965 Edition. The answers to the questions of Subitems 1 through 6 along with the associated references are-

_ i

Document Control Desk June 25, 1992 Page 3 i provided in Tables 2 through 7 of BAW-2166 chapters applicable to each Point Beach unit.

Sectinn 3 requests licensees to provide information regarding commitments made to respond-to Generic Letter 88-11. Section 3 is divided into Items a,'b, and c.

Section 3, Item a, requests information regarding how the embrittlement effects of operating at a temperature below 525 'F were considered for Charpy upper shelf energy and reference temperature. This part-is only applicable to Point Beach, Unit 1.

Unit 1 was operated at reduced power and temperature for approximately four years because of steam generator concerns. Once the Unit 1 steam generators were replaced,'the Unit was returned to normal power and temperature operation, as shown in. Figure 4-4 of BAW-2166. As discussed in Chapter 4, Section 4.4 Land Table 8 of BAW-2166, a surveillance capsule was removed both before and after the period of low power and low temperature operation. The results from these capsules show that the actual material behavior is conservatively estimated Fy Regulatory Guide 1.99, Revision 2.

Therefore,_due to this conservatism, this period.of low temperature operation was not considered in determination of embrittlement effects.

Section 3, Item b, requests information regarding-how' surveillance results on the predicted amount of embrittlement were considered.

Surveillance results from Point Beach' Surveillance Program have not been used to predict the embrittlement of the Point. Beach reactor vessels as stated in BAW-2166 Table 9 for both Point Beach units.

In general, the predicted amounts of embrittlement have been-determined by-the-generic methods outlined-in regulatory guides and appropriate 10 CFR Part 50 regulations.- Additionally, correlation techniques developed by the B&W Owners Group Reactor Vessel Working Group have been used.

Section 3, Item c, requests information regarding whether the measured shift in reference temperature exceeds the mean-plus-two standard deviations predicted by Regulatory Guide 1.99, Revision 2, or if the measured decrease in Charpy upper shelf-energy exceeds the value predicted using the guidance in Regulatory Guide 1.99, Revision 2. As depicted in BAW-2166 Table 10 for each Point Beach unit, no measured changes have exceeded these limits.

In addition to the enclosed BAW-2166 report, we have attached additional information which will contribute t, your review of our reactor vessel integrity program. Attachment 1 provides a listing I'

Dt ument Control Desk Ju..e 25, 1992 Page 4 of our overall reactor vessel integrity program and Attachment 2-provides Unit 1 and 2 reactor vessel sketches of the beltline-region material.

We believe this response has demonstrated our continued compliance with 10 CFR 50.60 and conformance to our commitments made in response to Generic Letter 88-11.

Please contact us should you have questions or require additional information regarding this response.

Sincerely, L

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,,.4 Bob Link Vice President Nuclear Power Enclosure (BAW-2166)

Copies to NRC Regional Administrator, Region III NRC Resident Inspector Subscribed and sworn to before me this M day of \%t , , 1992.

/

Notary Public, State of Wisconsin My Commission expires 3I-22-97.

1 i

ATTACKMENT 1 POINT BEACH NUCLEAR PLANT REACTOR VEBBEL INTEGRITY PROGRAM 1984 TO PRESENT PROJECT DATE COMPLETE

1. Neutron exposure evaluativu of Point Beach reactor vessels. December 1984

2. Tested Unit 1 Surveillanco Capsule T. December 1984

3. 10 CFR 50.61 - Pressurized Thermal Shock (PTS)

Submitt, January 1986 Correction to PTS submittal. March 1986 Safety evaluation report received from NRC. July 1986

4. Reactor Vessel Life Extension Study.

Initiated study in May 1986.

Evaluation of fuel management techniques and internals modifications (shielding) to meet flux reduction goals. September 1987 Identification of critical components in NSSS, including the reactor vessel and compilation of transient data associated with these components. October 1987 Comprehensive scoping risk assessment to examine Point Beach specific concerns and the proprie:ty of the flux reduction goals. December 1987 Developed bases and specifications for a plantwide on-line fatique monitoring system. December 1987 1

k ______-__-__ __ . _ _ .

5. Inservice Inspection

a. Second Unit 1 Reactor Vessel Ten-Year Exam.

Performed ASME Coda exam utilizing S' II standard data acquisition system, including 50/70 tandem near surface search units. May 1987 Periormed exam using NES/Dynacon Ultrasonic Data Recording and Processing System (UDRPS) concurrent with ASME Code exam above. May 1987

b. Second Unit 2 Reactor Vessel Ten-year Exam.

SWRI Enhanced Data Acquisition System (EDAS) was utilized. October 1989

6. Joined Babcock and Wilcox Owner's Group (BWOG)

Materials Committee. August 1988

, Full participant in BWOG Reactor Vessel Integrity Program (RVIP). August 1988 Participant in BWOG Reactor Vessel Life Extension Surveillance Program (RVSP). 1989 Developing master integrated reactor vessel >

surveillance program to include Westinghouse utilities with Linde 80 welds in their reactor vessels. (BAW-1543) March 1989 Submitted BAW-1543 Revision 3 to NRC. October 1989 Safety evaluation report received from NRC for BAW-1543. June 1991 2

1 N

i I 7. . Installation of excore neutron dosimetry

(radiometric monitors and solid state track

' recorders) over one octant of each unit's reactor vessel. Analysis of_ sensor sets and-4 correlation of cavity measurements with j- transport calculations will be performed after

each fuel cycle for first three sets, j Thereafter, a three year interval will be used until sufficient data'is'obtained to increase i the interval, l-

. Install mounting hardware and first set of l- dosimetry in-Unit 2. November 1988 l

l Install mounting hardware and first. set of j dosimetry in Unit 1. May 1989 4

3 First sensor set-analyzed for-Unit.2. November 1990 l'

First sensor set analyzed for Unit 1. December 1990 Second sensor set analyzed-for Unit 2. October 1991

! Second sensor set analyzed for Unit 1. March 1992

} 8. Pilot project: On-line fatigue monitoring of j Unit 2 pressurizer surge nozzle (related--to i reactor vessel' life extension study fatigue-

~

evaluation).

! November 1988

?

' I

9. Implement super Low Leakage-Loading Pattern

! (L4P) cores and axially-zoned hafnium inserts in

the guide tubes of peripheral ~ assemblies.

Unit 1 May 1989 l- Unit 2 November 1989 i

l j -10 . Performed image enhancement of-selected 1

radiographs of important reactor coolant _ system

, ' components (reactor. vessels, piping,: steam generators, etc.) and retained radiograph image; '

on media more permanent than original media. 1989 3

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11. Submit revised heatup and cooldown curves using the guidance of Regulatory Guide 1.99, Revision

2. August 1989 '

Technical Specification change approved by NRC. January 1990

12. Tested Unit 2 Surveillance Capsule S. August 1991

13. Unit 1 &.2 Charpy Upper Shelf Energy Status and Unit 2 PTS Submittal Oc tob Ai- 1991

14. Generic Letter 92-01, " Reactor Vessel Structural Integrity" Submittal June 1992 l

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Attachment E .

FIGURE 1 IDENTIFICATION AND LOCATION OF BELTLINE REGION t%TERIAL FOR THE POINT BEACH UNIT NO.1 REACTOR VESSEL

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PIGNRE 2 IDENTIFICATION AND LOCATION 0F BEtTLINE REGION MATERIAL '

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BAW-2166 JUNE 1992 EKEEES? tic 2 Bad D 2?E505FE.

e e 1 i DEZIF"ZKE" TNT 23BifMF4L 6 B&W OWNERS GROUP RESPONSE TO GENERIC LETTER 92-01 W%!?fgB&W NUCLEAR EEk.v SERVICE COMPANY lYVW) f

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l B&W OWNERS GROUP RESPONSE TO GENERIC LETTER 92-01

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BAW-2166 June 1992 B&W OWNERS-GROUP RESPONSE TO GENERIC LETTER 92-01 by N.-J. DeVan, L. B. Gross, and A. L. Lowe, Jr.

BWNS Document No. 77-2166-00

',See Section 7 for document signatures.)

Prepared for B&W Owners Group Reactor Vessel Workina Group Commonwealth Edison Company-

-Duke Power Company.

Entergy Operations, Inc.

.5

. Florida Power Corporation- ~

Florida Power & Light Company GPU Nuclear Corporations Rochester Gas and Electric Corporation Toledo -Edison Company-Virginia Power Company Wisconsin Electric Power. Company.-

s Prepared byf B&W Nuclear Service Cor..pany-Engineering and Plant: Services Division 3315 Old Forest Road-

-P. 0. Box 10935 Lynchburg, Virginia 24506-0935 l

CONTENTS Page

1. INTRODUCTION . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-1

2. GENERIC LETTER . . . . . . . . . . . . . . . . . . . . . . . . . . 2-1

3. METHOD OF RESPONSE . . . . . . . . . . . . . . . . . . . . . . . . 3-1 3.1. Organization . . . . . . . . . . . . . . . . . . . . . . . . 3-1 3.2. Response Details . . . . . . . . . . . . . . . . . . . . . . 3-2

4. IRRADIATION TEMPERATURE . . . . . . . . . . . . . . . . . . . . . . 4-1 4.1. B&W-Designed 177-FA Plants . . . . . . . . . . . . . . . . . 4-1 4.2. Davis-Besse . . . . . . . . . . . . . . . . . . . . . . . . . 4-2 4.3. R. E. Ginna . . . . . . . . . . . . . . . . . . . . . . . . 4-2 4.4. Point Beach Units 1 and 2 . . . . . . . . . . . . . . . . . . 4-2 4.5. Surry Units 1 and 2 . . . . . . . . . . . .-. . . . . . . . 4-4 4.6. Turkey Point Units 3 and 4 . . . . . . . . . . . . . . . . . 4-4 4.7. Zion Units 1 and 2 . . . . . . . . . . . . . . . . . . . . . 4-4

5. SUPPLEMENTARY INFORMATION . . . . . . . . . . . . . . . . . . . . . 5-1 5.1. Construction Code . . . . . . . . . . . . . . . . . . . . . . 5-1 5.2. Fl uence Predictions . . . . . . . . . . . . . . . . . . . . . 5-2

6. RESPONSE TO GENERIC LETTER 92-01 . . . . . . . . . . . . . . . . . 6-1

7. REFERENCES . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7-1

8. CERTIFICATION . . . . . . . . . . . . . . . . . . . . . . . . . . 8-1

- ii -

List of Fioutgt figure Page 4 1. Reactor Coolant System Temperatures as a. Function of Power for B&W 177 FA Plants Except Davis Besse . . . . . . . . . . . . . . . 45 4 2. Reactor Coolant System Temperatures as a function of Power for Davis-Besse . . . . . . . . . . . . . . . . . . . . . . . . . . . 45

• 4-3. Reactor Coolant System Temperatures as a function of Power for Westinghouse Designed Plants for R. E. Ginna Unit 1. . . . . . . 46 4-4. Reactor Coolant System Temperatures as a function of Power for Westinghouse Designed Plants for Point Beach Units 1 and 2 . . . . 46 4-5. Reactor Coolant System Temperatures as a function of Power for Westinghouse Designed Plants for Surry Units 1 and 2 . . . . . . . 4 7- '

4 6. Reactor Coolant System Temperatures as a function of Power for Westinghouse Designed Plants for-Turkey Point Units 3 and 4 . . . 47 4-7. Reactor Coolant System Temperatures as a function of Power for Westinghouse Designed Plants for Zion Units 1 and 2 . . . . . . . 4-8 1

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1. INTRODUCTION This report provides a response to the Nuclear Regulatory Commission (NRC)

Generic letter 92-01 for those nuclear power plants that are members of the B&W Owners Group Reactor Vessel Working Group.

Generic letter 92-01, Revision 1. shown in Section 2 of this report, was issued by the NRC on March 6,1992 and addressed to all holders of nuclear power plant opers ng licenses. The generic letter was issued to obtain information from the licensees to enable the NRC to assess the degree of compliance with regulatory requirements regarding reactor vessel integrity. Response is required within 120 days of the issue date; this comes to July 4,1992. This document provides the required information, insofar as it is available, for the following plants:

Elnt 03(Dft Arkansas Nuclear One Unit 1 Entergy Operations, Inc.

Crystal River Unit 3 Florida Power Corporation Davis Besse Unit 1 Toledo Edison Company R. E. Ginna Unit 1 Rochester Gas & Electric Corp.

Ocoi;ae Unit 1 Duke Power Company Oconee Unit 2 Duke Power Company Oconee Unit 3 Duke Power Company Point Beach Unit 1 Wisconsin Electric Power Co.

Point Beach Unit 2 Wisconsin Electric Power Co.

Surry Unit 1 Virginia Electric & Power Co.

Surry Unit 2 Virginia Electric & Power Co.

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

? ion Unit 2 Commonwealth Edison Company 1-1

J

2. GENERIC LETTER i

Generic Letter 92 01, Revision 1, is shown below. (Enclosuro 2 does not include Crystal River Unit 3; discussions with the NRC staff indicated that this is an l inadvertent omission and t. hat Crystal River Unit 3 is to be considered as if it ,

)

is included in Enclosure 2.)

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[*,, UNft D 8tAfs3 NUCLEAR RE3ULATORY COMMISSION vsAa.HNetos n c mass k,*** Hatch 6, 1992 i

TO: ALL HOL0tP1 07 OPERAf!N', LICEN$t$ OR CONSTRUCT!DN PERMITS FOR NUCLEAR POWER l'LAki$ (f! CEPT YANrt! AT XIC [LtCTRIC ComPAuf. LICEN$tt FOR THE YANs.It NUCLEAR POWER STAfl04)

$UBJtCT: REACTOR yt$1tL STRUCTURAL IN7tCRITT,10 CFR 50.54(f)

(CEN!RICLETTER9201,RtV1110F1)

This letter replaces Ceneric letter 92 01 dated February 28, 1992. The background inforutton concerning hPC's assesseent of es6rittlement in the f antee Nuclear Power Station reactor vessel was draf ted by staff some senths ago and has now been clarified and updated to better reflect the licenste's entensive technical effort' regarding reactor vessel integrity. The section pertaining to *equired information has not changed.

The U.s. Nuclear Regulatory Commission (hRC) is issuing this generic letter to obtain inforestion needed to assess compliance with requirements and cosmitments regarding reactor vessel integrity in view of certain concerns raised in the staff's review of reactor vessel integrity for the Yankee Nuclear Power Station.

In Section 50.60f a) of Title 10 of the Code of Federal Regulations (10 CFR 50.60(a)), the NRC requires that licensees for all Itght water nuclear power reactors sett fracture toughness requirements and beve a esterial surveillance progree for the reactor coolant pressure boundary. These reoutrements are set forth in Appendices G and H to 10 CFR Part 50. In 10 CFR 50.60(b), where the requirements of Appendices G and H to 10 CFR Part 60 cannot be emet, an esee9 tion is necessary pursuant to 10 CFR 50.!!. In 10 CFR $0.61 the NRC also provided fracture toughness requirements for protectieg pressurited water reactors against pressurtled thermal shock events. Licensees and permit holders have alsostadecommitsentsinresponsetoGenericLetter(til)8811,"NRCPosition on Radiation fabrittlement of Reactor Vessel Materials and its !apact on Plant Operations ' to use the methodology in R' Radiation fabrittissent of Reactor 7 esse 1v1 story of ?*Jtron trradiation as reeutree by Paragraph V. A o,f 10 CFR Part 50, Appendte G. c 10 CFR 50.60 and 10 CFR $0.61 requirements and GL 6d.11 are in the overall regulatory program to maintain the structural integrity of the reactor tesfel.

This gtneric Intter is part of a progree to evaluate reactor vessel integrity and take regulatory actions if needed to ensure that licensees and persit holders are complying with IO CFR 50.68 and 10 CFR 50.61, and are fulfitting commitments Ede in response to GL 8811. Enclosure 1 is a disevstion of tne applicable regulatory requirements. The NRC is rgquiring information on comp 11 ann under the provistons of 10 CFR 60.54(f).

2-1 9203060147

2 Assessment bessel of toerittlement for the Yankee Nuclear Power StatioMeactor j in an effort to resolve concerns regarding the neutron ee6rittlement of the i Yankee reactor vessel, ths staff performed a safety assessment of the Yankee reactor vessel. The staff found that the licensee for the Yankee Nuclear j Power Station might not be in coepliance with 10 CFR 50.60 and 10 CFR 50.61.

The staff found that the Charpy upper shelf energy of the Yankee reactor vessel raterial could be as low as 35.5 foot. pounds which is less than the 50 foot pound value required in Appendix G to 10 CFR Part 50. However, the licenset for the Yankee Nuclear Power Station had not performed the actions required in Paragraphs

!Y. A.1 or V.C of Appendix G to 10 CFR Part 50. Since then, the licensee has performed an analysis in accordance with Paragraph !Y.A.1 of Appendix G to 10 r 4

• art 50 using criteria being developed by the American Society of Mechanical (69*2 trs (ASME) tu demonstrate Nrgins of- safety equivalent to those in the W Sode.

The NRC expressed a concern regarding compliance with the requirements of Appendix H to 10 CFR Part 50. Section E 185 of the American Society for Testing and Materials (ASTM) Code requires that the licensee tako sample specimens from actual material used in fab"icating the beltline of the reactor vessel. Yhese surveillance materials shall include one heat of base metal, one butt weld, and one weld " heat affected zone." The licensee for the Yankee Nuclear Power Station terminated the meterial surveillance program in 1965.

Therefore, the Yankee Nuclear Power Station had no material surveillance program on July 26, 1983, when Appendix H to 10 CFR Part 50 l,ecame effecthe.

Further, the samples irradiated at Yankee Rowe before 1965 were comprised only of base metal.

The licensee for the Yankee Nuclear Power Station had used the methodology in Revision 2, to predict the effects of neutron draf t Regulatory erbrittler:ent. Guide The staff ra1.99,ised concerns regarding the licensee's application of the methodology. The specific issues were (1) the irradiation temperature, (2) the chemistry composition of reactor vessel material, and (3) the results of the material surveillance program.

The irradiation temperature at the Yankee Nuclear Power Station is between 454 'F and 520 'F, which is below the nominal irradiation temperature of 550 'F used in developing Regulatory Guide 1.99, Revision 2. A lower irradiation temperature increases the effect of neutron embrittlement. The regulatory ,

guide int'icates that for irradiation temperatures less than 525 'F l es6rittlement effects should be considered to be greater than predicted by the .

methods of the guide. Adjustments that were made by the licensee were insufficient to account for this effect..

The results of the surveillance program from the Yankee Nuclear Power Station indicated that the increase in the reference temperature exceeds the mean-plus-two standard2.deviations Revision as predicted The regulatory guide by the that states procedures the Itcensee in Regulatory should use Guide credib 1.99 le surveillance data to prsciet the increase in reference temperature resulting from neutron irradiation, i

2-2

i 1

3 i

The staff implemen,ed RG 1.99, Revision 2, t'y issuing GL 88 11. In comitti*g to GL 8811, license's have comitted to calculats radiation embrittlement in accordance with th6 procedures documented in RG 1.99, devision 2. To meet

. the limitations in Section 1.3 of the regulatory guide, the licensee should consider the effects on irradiation embrittlement during core critical operation with irradiation temperatures less than $25 'F. Section 2 of the re9ulatory gH de states thct the licensees should const1er the effects of the results from its surveillance capsules.

i The Susener 1972 Addenda of the 1971 Edition of Section !!! of the ASME Boiler and Pressure Yessel Code are the earliest code requirements for testing i materials to determine their unirradiated reference temperature. The Yankee J reactor vessel was constructed in 1959 to ASME Code, Section Vil!. Therefore, j the unirradiated reference temperature could not be established in accordance j vith the requirements of the Summer 1972 Addenda. The licensee for the Yarkee Nuclear Power $tation extrapolated the available test results to determine an 4

unirradiated reference temperature. The staff determined that the licenste's extrapolation was not conservative.

The chemical composition of the Yankee reactor vessel welds is unknown. The seterial's sensitivity to neutron embrittlement depends on its chemical e

content. The licensee assumed that the chemistry of its welds was equivalent to that cf the BR.3 reactor vessel in Nol, Belgium. The heat nuabar of the wire used to fabric 6te the Yankee welds was not available. The licensee was j assuming a chemical composition that was not based on its plant specific information, since the chemical composition, in particular, the amount of copper, depends upon the heat number of the weld wire.

These factors prompted the staff to find that the licensee for the Yankee Nuclear Power Station had not fully' considered plant specific information in assessing compliance with 10 CFR 50.61. When plant-specific information is considered the Yankee reactor vessel may have exceeded the screening criteria in10CFR$0.61.

Upon conducting the Yankee Nuclear Power Station review, the staff became concerned about other licensee's compliance with 10 CFR 50.60 and 10 CFR 50.61 and fulfillment of cosuitmenu made in resporse to GL 8811. Thus, the staff

, is issuing this generic letter to obtain information to assess compliance with 4

these regulations and fulfillment of commitments. The staff is continuing to pursue this concern with the Yankee Atomic Electric Company. Therefore, the Yankee Atomic Electric Company need not respond to this generic letter.

Required.Information Portions of the following information requested are not applicable to all 1

addressees. The responses provided should, in these cases, indicate that the requested information is not applicable and why it is not applicable.

2-3

... -. ~ .- - - -- - _ _ - _ - - , . -. . .,

4 1

1. Certain addressees are requested to provide the following information regarding Appendix H to CFR Part 50

- Addressees who do not have a surveillance program metting ASTM E 185 73, .79, or .82 and who do not have an integrated surveillance program approved by the NRC (see Enclosure 2), are requested to describe actions taken er to be taken to ensure compliance with Appendix H to 10 CFR Part 50. Addressees who plan to revise the surveillance program to meet Appendix H to 10 CFR Part 50 are requested to indicate when the revised program will be submitted to the NRC staff for review. If the surveillance program is not to b2 revised to meet Appendix H to 10 CFR Part 50, addressees are requested to indicate when they p*an to request an exemption from AppendixHto10CFRPart50under10CFR50.60(b).

2. Certain addressees are requested to provide the following information regarding Appendix G to 10 CFR Part 50:

6. Addressets of plants for which the Charpy upper shelf energy is predicted to be less than 50 foot-pounds at the end of their licenses using the guidance in Paragraphs C.1.2 or C.2.2 in Revision 2, are requested to provide to the t Regulatory and for the end Guide NRC the Charpy upper s 1.99.helf of their current licenseenergy for the limiting predicted be t inefor Deces6e 1991,

~

weld and the plate or forging and are rewested to describe the actions taken pursuant to Paragraphs IV.A.1 or V.C of Appendix G to 10 CFR Part 50.

b. Addressees whose reactor vessels were constructed to an ASME Code earlier than the Sussner 1972 Addenda of the 1971 Edition are requested to describe the consideration given to the following material properties in their evaluations perforwed pursuant to 10 CFR 50161 and Paragraph !!!.A of 10 CFR Part 50, Appendix G (1) the results from all Charpy and drop weight tests for all )

unirradiated beltline materials, the unirradiated reference temperature for each beltline material, and the method of determining the unirradiated reference temperature from the Charpy and drop weight testl .

(2) the heat treatment received by all beltline and surveillance materials; (3) the heat number for each beltline plate or forging and the heat number of wire and flux lot nusber used to fabricate each beltline weld; 2-4

d i

i 5-(4) the heat number for each surveillance plate or forging and the i heat number of wire and flux lot nun 6er used to fabricate the  !

surveillance weld: '

(5) the chemical composition, in particular the weight in percent of coppe? nickel phosphorous, and sulfur for each beltline and survedlancema,terial: and (6) the heat number of the wire used for detemining the weld metal chemical composition if different than Item (3) above.

3. Addressees are requested to provide the following information regarding consnitments rede to respond to GL, 88-1L:

L' a. How the embrittlement effects of operating at an irrediation

• temperature (cold leg or recirculation suction temperature) below 525

'T were considered, in particular licensees are requested to describe ,

)

' consideration given to determining the effect of lower irradiation temperature on the reference teinperature and on the Charpy upper i

shelf energy.

l b. How their s wveillance results on the predicted amount of

{ esbrittlerent were considered.

! c. If a measured increase in reference temperature exceeds the

' mean-plus.two standard deviations predicted by Regulatory Guide i

1.99, Revision 2, or if a measured decrease in Charpy upper shelf i

energy exceeds the value predicted usira the guidance in Paragraph '

the licensee is requested C.1.2 to report inthe Regulatory informationGuide and describe 1.99, the Revision e 2, ffect of the survet11ance j results on the adjusted reference temperature and Charpy upper shelf energy for each beltline material as predicted for December 16,

1991, and for the end of its current license.

_teporting.tequirements 1

Pursuant to Section 182a of the Atomic Energy Act of 1954, as amended, and 10 CFR 50.54(f),'each addressee shall submit a letter within 120 days of the

date of this generic letter providing the infomation described under ' Required Information." The letter shall be addressed to the U.S. Nuclear Regulatory Commission, ATTNt Document Control Desk, Washington, DC 20555, under oath or i

affirmation. A copy shall also be subsitted to the appropriate Regional i Administrator. This generic letter requests information that will enable the i

NRC to verify'that the licensee is complying with its curient licensing basis

' regarding reactor vessel fracture toughness and material surveillance for the reactor coolant pressure boundary. Accordingly an evaluation justifying this information request is not necessary under 10 EFR 50.54(f).

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Backfit. Discussion

This generic letter requests information that will enable the NRC staff to determine whether licensees are complying with their prior commitments and any license conditions regarding 10 CFR $0.60 10 CFR $0.61 and GL 88 11.

The staff is not establishing a new position for such comp 1Iance in this generic letter. The staff is requesting information to verify that the licensee is complying with its previously established commitsents and is not establishing any new position. Therefore, this generic letter does not ,

constitute a backfit and no documented evaluation or backfit analysis need be prepared.

Request for Volunta*y $ubmittal of Impact, Data This reouest is covered by Office of Menagement and Budget' Clearance Number -

3150-0011, which expires May 31, 1994. The estimated everage number of burden hours is 200 person hours for each addressee's response, including the time required to assess the requirements search data sources, gather and analyze thedata,andpreparetherequiredletters. This estimated everage nos6er of burden hours pertains only to the identified response-related matters and does not include the time to implement the actions required by the regulations.

Coments en the accuracy of this estimate and suggestions to reduce the burden may be directed to Ronald Minsk. Office of Information and Regulatory Affairs (3150-0011), NE08 3019, Office of Management and Budget, Washington, DC 20503, and to the U.S. Nuclear Regulatory Cosmission, Information and Records Management Branch, Division of Information Support Services, Office of Information and Resources Management, Washington, DC 20$55.

Although no specific request or requirement is intended, the following information would assist the NRC in evaluating the cost of complying with this generic letter:

, (1) the licensee staff's time and costs to perform requested inspections, corrective actions, and associated testing:

(2) the licensee staff's time and costs to prepare the requested reports and documentation; (3) the additional short-ters costs incurred to addrest the inspection findings I such as the costs of the corrective actions or the costs of down times and (4) an estimate of the additional long-term costs that will be incurred as a result of implementing cosmitments such as the estimated costs of conducting future inspections or increased maintenance.

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4 If you have any questions about this matter, please contact one of the NRC j technical contacts or the lead project manager listed below.  !

L i

$1ncerely, i 1

} Jaies s. Partlow t i Associate Director for Projects j

- Office of Nuclear Reactor Regulation  !

l Enclosures

1. .tpplicable Regulatory Requirements

2. Plants with Integrated Progrees

! 3. List of Recently Issued

, Generic Letters 1

i Technical Contactst '

Barry J. Elliot, NRR (301)504-2709

]

, Keith R. Wichman, NRR

+

(301)504-2757 4

Lead Project Managert Daniel G. Mcdonald, NRR I- (301)504-1408 4 <

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[nclosure 1 ,

Regulatory Requirementdypticable to

!!actorVesseL5tructural, Integrity 10 CFR.50.60 Pursuant to 10 CFR 50.60, all light water nuclear power reactors sust meet the ,

fracture toughness and material survetilance program requirements for the  ;

reactor coolant pressure boundary set forth in Appendices 6 and H to 10 CFR Part 50.-

The fracture toughness of the reactor coolant pressure boundary required by 10 CFR 50.60 is necessary to provide adequate margins of safety during any condition of normel plant operation, including anticipated operational 1 occurrences and system hydrostatic tests. The material surveillance program required by 10 CFR 50.60 monitors changes in the fracture toughness properties of ferritic materials in the ' reactor vessel beltline region of light water i nuclear power reactors resulting from exposure of these materials-to neutron .

irradiation and the thermal eny' ronment. Under the program,' fracture toughness test data are obtained from material specimens exposed in surveillance capsules, which are withdrawn periodically from the reactor vessel.

Appendix G to 10 CFR Part 50 requires th'at the reactor vessel beltline materials must have Charpy upper shelf energy of no less than 50 ft-Ib throughout the life of the vessel. Otherwise licensees are required to providedemonstrationofequivalentmarginsofsafetyinaccordancewith Paragraph IV.A.1 of Appendix G to 10 CFR Part 50 or perform actions in accordance with Paragraph V.C of Appendix G to 10 CFR Part 50.

Appendix H to 10 CFR Part 50 requires the survei' lance program to meet the-American Society for Testing and Materials (ASTM) Standard E 185, ' Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Noelear-Power Reactor Vessels. Further, Appendix H to 10 CFR Part 50 spec 1ti^5 the ,

, applicable edition of ASTM E 185. Appendix H to 10 CFR Part 50, as amended on i July 26 1983, requires that the part of the surveillance program conducted  !

beforelhefirstcapsuleiswithdrawnmustmeettherequirementsofthe1973, l the 1979, or the 1982 edition of ASTM E 285 that is current on the issue date  !

oftheAmericanSocietyofMechanicalEngineers(ASME)SollerandPressure l Vessel Code under which tha reactor vessel was purchased. The licensee may.

also use later editions of ASTF E 185 which have been endorsed by the NRC. ,

The test procedures and reporting requirements for each capsule withdrawal: i 1

after July 26, 1983 must meet the requirements of the 1982 edition of ASTM-E 185 to the extent practical for the configuration of the specimens in the capsule. The licensee may use either the 1973 the 1979, or the 1982 edition of ASTM E 185 for each capsule withdrawal befor,e July 26,=1983.

2-8 -)

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4 l Licensees, especially those with reactor vessels purchased before ASTM i- issued the 1973 edition of ASTN E 185, say have surveillance programs that do

[ not meet the requirements of Appendix H to 10 CFR Part 50 but may have i alternative surveillance programs. The licensee may use these alternative

surveillance programs in accordance with 10 CFR 50.60(b) if the licensee has '

l been granted an exemption by the Cosmission under 10 CFR 50,12.  !

i l The licensee must monitor the test results from the material surveillance

} program. According to Paragraph !!!.C of Appendix H te 10 CFh Part 50, the

! rasults of the surveillance program may indicate that a technical-j specifications change is required, either in the pressure-temperature limits or in the operating procedures required to meet the limits. >

l

j. 10.CFR 50.61 4 .

j Pursuant to 10 CFR 50.61, there are fracture toughness requirements for -

i protection against pressurized themal shock events for pressurized water i reacters. Licensees are required to perfom an assessment of the projected values of reference temperature. If the projected reference temperature exceeds the screening criteria established in 10 CFR 50.61, licensees are i

required to submit an analysis and schedule for such flux reduction programs -

as are reasonably practicable to avoid exceeding the screening criteria. If i

a no reasonably practicable flux reduction program will avoid exceeding the j screening criteria licensees shall submit a safety analysis to determine 4

what actions are ne,cessary to prevent potential failure of the reactor vessel l if continued operation beyond the screening criteria is allowed. In 10 CFR .t

! 50.61(b)(1) 14, 1991 (56 Fed Reg 22300 et. seq.,

May15,1991)asamendedeffectiveJune, Itcensees are required to submit their assess 1 Deceaser 16, 1991, if the projected reference temperature will exceed the '

screening criteria before the expiration of the operating license.

I Plant specific information is required to be coesidered in assessing the level 4

' ofneutrones6rittlementasspecifiedin10CFR50.61(b)(3). This information includes but is not limited to the rea-tor vessel operating temperature and surveillance results.

Prediction.of -!rradiation Embrittlement Paragraph V.A of Appendix G-to 10 CFR Part 50 requires the prediction of the effects cf neutron irradiation on reactor vessel materials. The extent of l neutron embrittlement depends on the material properties._ thermal environment, and resdits of the material surveillance program. In Generic Letter 88-11, j 'NRC 4 Irpact Position on Planton 0 Radiation Embrittlement of Reactor Vessel Materials and its ,

j Regulatory Guide $99, Revlsion-2,- ' Radiation Embrittlement of Reactor Vesselrations ' t

Materials

• in estimating the embrittlement of the materials in the reactor j vesselbeltline.- All. licensees and permittees have responded to Generic j 1.etter 88-11 cosmitting to use the methodology in-Regulatory Guide 1.99,

a. .

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u. a - .. - . .,.,,.:.-- --.-_-.,_n-.-.- - . . - .

I Revision 2, in predicting the effects of neutron irradiation as required by Paragraph V.A of 10 CFR Part 50, Appendix G. The methodology in Regulatory Guide 1.99, Revision 2, is also the basis in 10 CFR 50.61 in projecting the reference temperature.

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Enclosure 2 i ,

)

1 Plants With. Integrated Surveillance; Programs, Approved,By The,NRC i

j Oconee Units 1, 2 and 3

, Arkansas Nuclear One Unit 1

Rancho Seco Three Mile Island Unit 1

, Davis-Besse Ginna Point Beach Units 1 and 2 Surry Uni .J 1 and 2 Turkey Point Units 3 and 4 Zion Units 1 and 2 l-l l

4 2-11 rM* S* *+M gr -7 7ya* g- y- 9 w- 7ewm--yet'-.me-w-we- --y ws -g+ E7-'S-v -me-*w----pg.- 'e geq't-A s 9g qq -M g g- wnwLiwwm W f i- e-reW WT-"ref*f V+ *4W Er4 "' 41'pM"' @ *%-P-WF T'-~M"-rk-'****W*'

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3. METHOD OF RESPONSE 3.1. Oraanization The Generic Letter presents the information requests in three sections (1, 2, and

3) further divided into a number of items. Ten distinct sections /ite.u were identified, each of which are presented in a table. The tables are identified as follows:

GL 92-01 Ighlg Reference Egbjerl (1) Section 1 10CFR50, Appendix Hi Adherence to RVSP Requirements (2) Section 2, 10CFR50, Appendix G; CvuSE Requirements

Item a (3) Section 2, 10CFR50.61 and 10CFR50, Appendix G, III.A; Material Item b, Properties Related to PTS and Fracture Toughness Require-1 (1) ments (unirradiated Charpy and RT a values)

(4) Section 2, 10CFR50.61 and 10CFR50, Appendix G, III.A; Material item b, Properties Related to PTS and Fracture Toughness Require-1 (2) ments (material heat treatment)

(5) Section 2, 10CFR50.61 and 10CFR50, Appendix G, III.A; Material Item b, Properties Related to PTS and Fracture Toughness Require-1 (3) ments (beltline material identification)

(6) Section 2, 10CFR50.61 and 10CFR50, Appendix G, III.A; Material Item b, Properties Related to PTS and Fracture Toughness Require-1 (4) ments (surveillance material identification)

(7) Section 2, 10CFR50.61 and 10CFR50, Appendix G, III.A; Material Item b, Properties Related to PTS and Fracture Toughness Require-1 (5) ments [ chemical composition) 3-1

i GL 92 01 f Table Reference Sub.iect (8) Section 3 Generic Letter 88-11 Response Commitments: Effect of item a Irradiation Temperature-(9) Section 3 Generic Letter 88-11 Response Commitments; Utilization of Item b Surveillance Results 4

(10) Section 3, Generic Letter 8811 Response Commitments; Difference Item c Between Measured-and Predicted (Regulatory Guide 1.99, '

Revision 2) Embrittlement Effects Each of the above ten tables were prepared for each of the sixteen plants covered l by this report. These tables are presented in Section 5 of this report.  !

3.2. Response Details 1 3.2.1. Abbreviations used in-the responso are as follows:

ARTuoi Adjusted reference temperature CvVSE Charpy upper shelf energy EOL End of life EST Estimated value l NA Not applicable '

ND Not determined PTS Pressurized thermal shock RVSP Reactor vessel surveillance program RTuor Reference temperature 1 1

ARTuor_ Reference temperature shift 3 i

l o Standard deviation 3.2.2. Material properties were determined at the I-thickness. location, in ,

accordance with 10CFR50, Appendix G, 1 V.B. Footnote 2.. Effects of neutron embrittlement were determined in accordance with the methods of 3-2 l

_.-__..a._._....___..____.-._... . _ _ . . . , _ . _ _ _ _ . . . . . ,._.u.;__,,._,--_._.,;.,._ _ , . . - . _ . . . _ _

_ . _ . . __ _ . . . _ ._ ____ _ __ _ . _. ._ -__ - ~ _ _ _ _ . . . _ _ . _ _

i Regulatory Guide 1.99. Revision 2. The drop in C,USE was determined in accordance with Position 1 unless otherwise stated in the response tables. The end of-life is taken as the time when 32 EfPY is achieved l unless otherwise stated in the response tables. ,

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4. IRRADIATION TEMPERATURE 4

Material sensitivity to irradiation embrittlement is directly affected by irradiation terperature. Over the temperature range that most light-water cooled reactors operate, the irradiation embrittlement is inversely related to irradiation temperature. However, since current generation pressurized water cooled reactors operate over a the relatively narrow temperature range (i.e. 529-556F RV inlet temperature), the relative-sensitivity of the beltline materials as a function of temperature is easily overshadowed by other parameters such as variations in material properties and Charpy impact testing techniques. The development of Regulatory Guide 1.99, Revision 2, was based solely on surveil-lance data in the irradiation temperature range of 525 to 575F. Normally, the Regulatory Guide 1.99, Rev. 2 data is applied directly in the evaluation of a reactor vessel on the assumption that the reactor vessel temperature was always within this temperature range. However, as can be seen from a review of reactor coolant system temperature as a function of power, the inlet temperature can vary. This does not affect the monitoring of irradiation embrittlement of the 4

reactor vessel because the surveillance capsules are located in the downcomer region of the reactor vessel and experience the same temperature history as the reactor vessel.

The reactor coolant system temperatures as a function of power for each plant included in this report are reviewed below. These data were provided by each plant owner and are as stated in their respective FSAR's, 4.1. B&W-Desianed 177-FA P131t,1

~

Figure 4-1 shows the reactor vessel outlet temperature (Tu) and the reactor vessel inlet temperature (Tc,w) for the B&W 177-FA reactor vessels. This is 4-1

representative of all 177-FA plants except Davis-Besse. These operating limits are characterized by a constant system average temperature and an increase in the inlet temperature (Tem) to 580F with a reduction in operating power. These temperature characteristics result from the fact that initial approach to power is controlled by the water level in the steam generator followed by a change in operation to maintain the system average temperature constant. The increase in inlet temperature may have the effect of minimizing irradiation embrittlement for these plants.

4.2. Davis-Besse Figure 4-2 shows the reactor vessel outlet temperature (Tn,) and the reactor vessel inlet temperature (Tem) for the Davis-Besse reactor vessel. The system behavior is similar to that of the other 177 FA plants with the exception that the change from level control to control of system average t,emperature is at approximately 28% power.

4.3. R. E. Ginna Figure 4-3 shows the reactor vessel outlet temperature (Tn,) and the reactor vessel inlet temperature (Tew) as a function of power for the R. E. Ginna reactor vessel. These operating limits are characterized by an increasing average temperature and a near constant reactor vessel inlet temperature for all power levels.

4.4. Point Beach Units 1 and 2 Figua 4-4 shows the reactor vessel outlet temperature (Tn,) and the reactor vessel inlet temperature (Tem).as a function of power for the Point Beach Units 1 and 2 reactor vessels. These operating limits are characterized by an increasing average temperature and a small decrease in reactor vessel inlet temperature as power increase to 100%.

The Point Beach Unit 1 operated at a reduced power from approximately December 1, 1979 to October 1, 1983, as shown in Figure 4-4. During this period, the reactor vessel was operated at a temperature of 511F at 80% to 522F at 0% power.

4-2 l

This reduced operating temperature does not appear to have affected the irradiation embrittlement characteristics of the materials. Fortunately, a surveillance capsule was removed and evaluated prior to the reduced temperature operation. This capsule (Capsule R, WCAP 9357 and BAW.lB03, Rev.1) experienced a fluence of 2.10 x 10" n/cm' (E > 1 MeV) and the weld metal exhibited an irradiation induced 30 ft lb Charpy temperature shift of 165F. A similar capsule (Capsule T, WCAP-10736 and BAW 1803, Rev.1) was removed and evaluated after the reduced temperature operation. The capsule experienced a fluence of 2.11 x 10" n/cm' (E > 1 HeV) and the weld metal exhibited an irradiated induced Charpy 30 ft-lb temperature shift of 175F. Although it might be argued that this difference was caused by the reduced temperature exposure, the values are well within the expected scatter of Charpy impact test data. The comparable Regulatory Guide 1.99, Rev. 2 estimate for a fluence of 2.11 x 10" n/cm', based on the weld metal chemical composition, is a shift of 196F. Consequently, the Regulatory Guide 1.99, Rev. 2 conservatively estimated the weld metal response to irradiation without the margin. Based on the Regulatory Guide 1.99, Revision 2, Position 2, and the data from four surveiliance capsules estimates a shift value at 2.11 x 10" n/cm' (E > 1 MeV) of 176F (without margin). Therefore, the reactor vessel material shift behavior as a result of exposure to irradiation is conservatively estimated by Regulatory Guide 1.99, Revision 2, both Position 1 and Position 2. Similar evaluation of the Charpy upper shelf energy showed a value of 53 ft-lbs at a fluence of 2.10 x 10" n/cm' (E > 1 MeV) for the capsule removed prior to the reduced temperature operation and-a value of 55 ft-lbs at a fluence of 2.11 x 10" n/cm' (E > 1 HeV) for the capsule removed after the reduced temperature operation. These values are well within the expeued scatter of Charpy impact test data. The comparable Regulatory Guide 1.99, Revision 2 estimate for a fluence of 2.11 x 10" n/cm' is an upper-shelf value of 37 ft-lbs.

The Regulatory Guide 1.99, Revision 2 conservatively estimates the upper-shelf energy of the weld metal.

4-3

4.5. Surry Units 1 and 2 Figure 4 5 shows the reactor vessel outlet temperature (Tno,) and the reactor vessel inlet temperature (Tc ) as a function of power for the Surry Units 1 and 2 reactor vessels. These operating limits are characterized by an increasing average temperature and near constant reactor vessel inlet temperature for all p wer levels.

4.6. Turkey Point Units 3 and 4 Figure 4-6 shows the reactor vessel outlet temperature (in,,) and the reactor vessel inlet temperature (Tew) as a function of power for ths Turkey Point Units 3 and 4 reactor vessels. These operating limits are characterized by an increasing average temperature and a near constant reactor vessel inlet

- temperature for all power levels.

4.7. Zion Units 1 and 2 figure 4-7 shows the reactor vessel outlet temperature (Tso,) and the reactor vessel inlet temperature (Tew) as a function of. power for the Zion Units 1 and 2 reactor vessels. These operating limits are characterized by ar increasing

, average temperature with increasing power levels. The inlet temperature decreases with increasing power and reaches a minimum at 100% power.

T 44 e , , - , . , , _ a.,.-, .y.,-a. -

e- ..w- + ~ - .

figure 4-1. Reactor Coolant System Temperatures as a Function of Power for B&W 177-FA Plants Except Davis-Besse y, 620  ;

i i 602F 2 600 -

7 g ,,

E $80F T A ,,,.g. $79p 5 580 Nominalf CL I Cold 560 -

c H 556F

@ 540 -

l ~5 532F (Hot Standby) o o 520 '-

~

5 500 - Level -

@ Control T s,,,... control o

[ I ' '

480 0 25 50 75 100 Nuclear System Power, Percent figure 4-2, Reactor Coolant System Temperatures as a Function of Power for Davis Besse

u. I I I 602F

$ 600 -

s 580F 7 H

~

Nominal

$ 580 -

---

S*- - - - 67D ct / Tcois E

o 560 -

H $56F

~

@ 540 -

~

o 532F (Hot Standby) o 520 -

~

E 500 -

Level -

@ Control T4 ,,,... Control e

T I_ I I 480 O 25 50 75 100 Nuclear System Power, Percent 45

Figure 4-3. Reactor Coolant System Temperatures as a function of Power for Westinchouse-Desioned Plants for R. E. Ginna Unit I 620

u. i i i

. 602.0F S 600 -

3

$ TH' 573.5F-5 580 -

~~~~

a TAv..o.

E y

560 -

""""~ , , ,

Teeio 545.0F fo 540 ja7F (Hot Standby) _

o o 520 - -

E -

U 500 -

e m ' ' '

E 480 O 25 50 75 100 Nuclear System Power, Percent figure 4-4. Reactor Coolant System Temperatures as a Function of Power for Westinahouse-Desianed Plants for Point Beach Units 1 and 2

u. I i i 1

• 600 -

59BF -

! o l

g Twoi

'; 580 - -

T 570F ,

g ~ ~ g ,. e o 560 -

""' ~~

557F 7

540 547F (Hot Standby) gp_

]m -- --**~ Aver.01. . . . . 534 F o

o - -~~ ~~~.....

O 329 -

Teolo

~'

522F (Hot Standby) - 511F hm 500 - -

o E ' I '

480 -

O 25 50 75 100 Nuclear System Power, Perc.ent l

4-6 i

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! Figure 4 5. Reactor Coolant System Temperatures as a function of Power j for Westinahouse Desianed Plants for Surry Units 1 and 2 4

1 620 'I

u. I I 605.8F

. 8 600 - -

o

! g THot-  !

! 5 580 -

574.4F~ -

! o. TA,.,.g. ~ ,,.

- ~

1 E ~

j y 560 -

Teolo 543.0F l 54C 547.0F (Hot Standby) -

5 O

O .

! o 520 - -

1, '

i O E 500 - -

m s c) i c ' I I 480- ,

O 25 50 75 100 t

1 Nuclear System Power, Percent-I Figure 4-6. Reactor coolant System Temperatures as a Function of Power  !

' for Westinahouse-Desianed Plants for Turkey Point Units 3 and 4

u. l l l i .

2 600 - . 602.0F i

.< .o I Hot l m j 5 580 -

573.0F ~

~ ~

[ T vereg ,.,,,

A

y 560 -

~,,

~

3 Tema 546.0F

i. c 547F (Hot Standby) m 540 -

4- -

t O 1 0 4

o 520 -

+

O L U 500' -

m -

i Q) g 180 I I -

I 0 25 50 75 100 1

Nuclear System oower, Percent ,

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, _ . . . - _.....;_m.m._.,, .. , _. - ..-,,-... - .. _ .... .m .... - _. ..,_., - . ,-,.. , _. ..,m.... . . . , _ . - , . . . . . . . . . - - . _ . , - , , . . , .

Figure 4-7. Reactor Coolant System Temperatures as a Function of Power for Westinahouse-Desianed Plants for Zion Units 1 and 2 620 , , ,

LL N 600 - -

3 589.4 F m

5 580 -

Ts.i -

c.

s 560 -

TA80s __g, 59 c 540 547.0F Tcold

$ (Hot Standby) $29.4F _

o -

$ 520 - -

8 3 500 - -

m o

E ' I '

480 0 25 50 75 100 Nuclear System Power, Percent 48 u___ _ . _ . . _

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5. SUPPLEMENTARY INFORMATION 5.1. Constructicn Code The reactor vessels for the following plants were fabricated'in accordance with Section 111 of the ASME Boiler and Pressure Vessel Code.- The Edition and Addenda- l l (where applicable) of the Code are noted. ,

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_ Plant Section III Edition and Addenda Ark.nsas Nuclear One Unit 1 1965 Edition, Sumer 1967 Addenda Crystal River 'J nit 3 1965 Edition, Sumer 1967 Addenda Davis-Besse Unit 1 1968 Edition, Summer 19Es Addenda -!

R. E. F4ana Unit 1 1965 Edition

! Oconee Unit 1- 1965 Edition, Sumer 1967 Addenda-Oconee Unit 2 _1965 Edition, Sumer 1967 Addenda-Oconee Unit 3 1965 Edition, Sumer:1967 Addenda' l Point Beach Unit 1 1965 Edition Point Beach Unit 2 1965 Edition--

-Surry Unit 1 Not available, final assembly by Rotterdam Surry Unit 2 Not available,- finalt assembly by Rotterdam .

Three Mile Island Unit 1 1965 Edition, Sumer 1967 Addenda' Turkey Point Unit 3 1965 Edition, Summer 1966 Addenda Turkey Point Unit 4 1965 Edition,- Sumer 1966 Addenda Zion Unit 1 1965 Edition, Sumer 1966 Addenda Zion Unit 2 1965 Edition, Sumer 1966 Addenda :

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5.2. fluence Predictions Peak fluence predictions for the beltline materials for each plant are presented in Table 5.2 1.

P 9

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5-2 y.-,. v-., . - - . . . - . - - . --e , .-p,,, , .,g..

i Table 5.2-1. Fluence Predictions for Beltline Reaion Materials Arkansas Nuclear One Unit 1 Fluence. 12/16/91 Fluence. 32 EFPY

! Material location is T/4 IS T/4 AYN 131 Lower Nozzle 3.19E+18 1.92E+18 8.62E+18 5.18E+18 j b it forging L5120-2 Upper Shell 3.63E418 2.18E+18 9.79E+18 5.88E+18 j Plate I

C5114-2 Upper Shell 3.63E+18 2.18E+18 9.79E+18 5.88E+18 Plate

! C5120-1 Lower Shell 3.48E+18 2.09E+18 9.40E+18 5.64E+18 Plate l

i C5114 1 Lower Shell 3.48E+18 2.09E+18 9.40E+18 5.64E+18 Plate WF-182-1 Nozzle 8elt to 3.19E+18 1,92E+18 8.62E+18 5.18E+18

) Upper Shell

• Circ. Weld WF-ll2 Upper Shell to 3.48E+18 2.09E+18 P 'A18 5.64E+18 Lower Shell i Cire. Weld SA-1788 Lower Shell to 2.03E+16 1.22E+16 5.48E+16 3.29E+16 Dutchman Cire.

. Weld WF-18 Upper Shell 2.61E+18 1.57E+18 7.05E+18 4.23E+18 Longit. Weld WF-18 Lower Shell 2.58E4!) 1.55E+18 6.95E+18 4.17E+18 Longit. Weld Crystal River Unit 3 Fluence. 12/16/91 Fluence. 32 EFPY

Material Location IS T/4 IS T/4 AJZ 94 Lower Nozzle 2.39E+18 1.44E+18 7.53E+18 4.52E+18 Belt Forging

53 1

(

i L

Table 5.2-1. Fluence Predictions for .Seltline Region Materials (Cont.1 r Crystal River Unit 3 (Cont.)

Fluence. 12/16/91 Fluence. 32 EFPY Material location IS T/4 IS T/4 C4344 1 Upper Shel'. 2.72E+18 1.63E+18 8.56E418 5.14E418 Plate ,

C4344 2 Upper Shell 2.72E+18 1.63E+18 8.56E+18 5.14E+18 .

Plate C4347-1 Lower Shell 2.61E+18 1.57E+18 8.22E+18 4.94E+18 4

Plate C4347-2 Lower Shell 2.61E+18 1.57E+18 8.22E+18 4.94E+18 j Plate SA 1769 Nozzle Belt to 2.39E+18 1.44E+18 7.53E+18 4.52E+18 Upper Shell ,

Cire. Weld (40% ID) ,

WF-169-1 Nozzle Belt to -- - --- --- --

Upper Shell Cire. Weld (60% 00) &

WF-70 Upper Shell to 2.61E+18 1.57E+18 8.22E+18 4.94E+18 Lower Shell -

Cire. Weld WF-154 Lower Shell to 1.52E+16 9.15E+15 4.79E+16 2.88E+16 Dutchman Circ.

Weld <

WF-18 Upper Shell 2.53E+18 1.52E+18 7.96E+18 4.78E+18 i Longit. Weld l 1

WF-8 Upper Shell 2.53E+18 1.52E+18 7.96E+18 4.78E+18 H '

-Longit. Weld- ,

SA 1580 Lower Shell 2.22E+18 1.33E+18 6.98E+18 4.19E+18 Longit. Weld .

b 5-4  :

l i

I

! . :__ - ._ . _ _ , . - . . _ . - , . _ - _ , _ - - _ _ - . . ~ . . . . . . . _ . _ . - . _ . - . . . . . - .

i.: l i

1 j Table 5.2-1. Fluence Predictions for Beltline Reaion Materials-(Cont.1-3 Davis Besse Unit 1 4

! Fluence. 12/16/91- Fluence. 32 EFPY-

Material location IS T/4 IS T/4  ;

j ADB 203 Nozzie Belt 3.92E+17 2.35E+17 - 1.50E+18- 9.01E+17 Forging

[

i AKJ 233 Upper Shell 2.80E+18 1.68E+18_ l.07E+19 6.43E+18 j- Forging l r

i BCC 241 Lower Shell_ 2.80E+18 1.68E+18 1.07E+19 6.43L T Forging

WF-232 Norrie Belt to - 3.92E+17 ---

1.50E+18 ---

Upper Shell Cire. Weld '

2 (9% 10) j WF-233 Nozzle Belt'to ---

2.35Er17 ---

9. 01 E+ 17 -

i Upper Shell i Cire. Weld l (91% 00) 4 WF-182-1 Upper Shell to 2.80E418 1.6BE+18 1.07E+19 6.43E+18 l Lower Shell l' Circ. Weld i

j WF-232 Lower Shell to . l.57E+16 ---

6.00E+16 ---

Dutchman Cire.

Weld (12% ID) i i WF-233 Lower Shell to ---

9.42E+15 ---

-3.60E+16-l Dutchman Circ.

I Weld (88% 00)

! R. E. Ginna Unit 1 jir.'nce. 12/16/91 Fluence. 3E EFPY-Material location ,,'3 1 T/4 IS T/4

123Pll8 val Nozzle 8elt - 2.05E+18 1.39E+18 3.69E+18' 2.50E+181 -

-- 8elt Forging L

1255255 val Interm. Shell 1.86E+19 - l.26E+19- 3.35E+19- 2.27E+19 i Forging i

i 5 i

1 i

L v m % , - c r w y- e,- .e,+, av mm -,-r---,+ rw w -e-4 ~-.-r -*---r-',..w.=ece.-v. _ m w.w . w.. .w w~ ~ w barr .-m+..-wb*-.Y-*

Table 5. 2-1. Fluence Predictions for Beltline Reaion Materials (Cont.1 R. E. Ginna Unit 1 (Cont.t Fluence. 12/16/91 Fluence. 32'EFPY Material location IS- T/4 IS T/4 125P666 val Lower Shell 1.86E+19- 1.26E+19- 3.35E+19 2.27E+19 Forging-SA-1101 Nozzle Belt to 2.05E+18 1.39E+18 3.72E+18 2.52E+18 Interm. Shell Circ. Weld SA 847 Interm. Shell 1.86E+19 1.26E+19 3.35E+19 2.27E+19 to Lower Shell Circ Weld SA-848- Lower Shell to_ NA NA_ NA NA Dutchman Cire. Weld Oconee Unit 1 Fluence, 12/16/91 Fluence. 32 EFPY Material location IS T/4 IS- T/4-AHR 54 Lower Nozzle 6.20E+17 3.72E+17- 1.18E+18- 7.09E+17 Belt Forging

,.)? . Interm. Shell 4.20E+18 2.52E+18 7.96E+18 :4.78E+18 Plate 3265 U57ar Shell 4.77E+18 2.86E+18 9.04E+18 -5.43E+18 Plate LSL :-l Upper Shell _4.77E+18 2.'86E+18 - 9.04E+18 5.43E+18 Plate-C2800-1 Lower Shell 4.58E+18' 2.75E+18 8.68E+18 5.21E+18 Plate C2800-2 -Lower-Shell 4.58E+18 2.75E+18 -8.68E+18 5.21E+18-Plate SA-1135 Nozzle Belt to 6.20E+17 3.72E+17 1.18E+18 7.09E+17 Interm Shell Cire. k;'d i

4 5-6 i

1 l

Table 5.2-1. Fluence Predictions for Beltline Reaion Materials (Cqat 1 Oconee Unit 1 (Cont.1 Fluence. 12/16/91 Fluence. 32 EFPY Material Location IS T/4 IS T/4 4

SA-1229 Interm. Shell 4.20E+18 2.52E+18 7.96E+18 4.78E+18 to Upper Shell Cire. Weld (61% ID)

WF-25 Interm. Chell --- --- --- ---

to Uppur Shell Cire. Weld (39% OD)

, SA-1585 Upper Shell to 4.58E+18 2.75E+18 8.68E+18 5.21E+18 Lower Shell Cire. Weld WF-9 Lower Shell to 2.67E+16 1.60E+16 5.06E+16 3.04E+16

Dutchman Cire.

Weld SA-1073 Interm. Shell 3.32E+18 1.99E+18 6.28E+18 3.77E+18 Longit. Weld SA-1493 Upper Shell 3.82E+18 2.29E+18 7.23E+18 4.34E+18 Longit. Weld SA-1426 Lower Shell 3.85E+18 2.31E+18 7.29E+18 4.38E+18 Longit. Weld SA-1430 Lower Shell 3.85E+18 2.31E+18 7.29E+18 4.38E+18 Longit. Weld Oconee Unit 2 Fluence. 12/16/91 Fluence. 32 EFPY Material location IS T/4 IS T/4 AMX 77 Lower Nozzle 3.88E+18 2.33E+18 8.42E+18 5.06E+18 Belt Forging AAW 163 Upper Shell 4.41E+18 2.65E+18 9.57E+18 5.75E+18 Forging

, 5-7

Table 5.2-1. Fluence Predictions for Beltline Reaion Materials (Cont.)

• Oconee Unit 2 (Cont.)

Fluence. 12/16/91- Fluence. 32 EFPY-Material Location IS T/4 IS T/4 i

I AWG 164 Lower Shell 4.23E+18 2.54E+18 . 9.19E+18 5.52E+18 Forging i WF-154 Nozzle Belt.to 3.88E+18 2.33E+18 8.42E+18 5.06E+18 1 Upper Shell Cire. Weld WF-25 Upper Shell to 4.23E+18 2.54E+18 9.19E+18' 5.52E+18.

Lower Shell Cire. Weld WF-112 Lower-Shell to 2.47E+16 'l.48E+16 - 5.36E+16 3.22E+16-Dutchman Cire.

Weld Oconee Unit 3 Fluence. 12/16/91 Fluence. 3? EFPY Material Location IS T/4 IS T/4 i 4680 Lov er Nozzle 3.85E+18 2.31E+18L 8.26E+18- 4.96E+18 8elt Forging ,

AWS 192 Upper Shell 4.37E+18 2.62E+18 9.39E+18 5.64E+18 Forging ANK 191 Lower Shell 4.20E+18 2.52E+18 9.01E+18- 5.41E+18-Forging

.WF-200 Nozzle Belt to' 3.85E+18 2.31E+18 8.26E+18 4.96E+18 Upper Shel1~

Cire. Weld WF-67 Upper Shell to 4.20E+18 2.52E+18 ' 9.01E+18: 5.41E+18 2 I

Lower Shell Circ. Weld )

(75% ID):

9

1. A

i 1

I Table 5.2-1. Fluence Predictions for Beltline Reuion Materials (Cont.)

Oconee Unit 3 (Cont.)

Fluence. 12/16/91 Fluence. 32 EFPY Material Location- IS T/4 IS . T/4 WF-70 Upper Shell to --- --- --- ---

Lower Shell Cire. Weld (05%OD)

WF-169-1 Lower Shell to 2.45E+16 1.47E+16 5.26E+16 3.16E+16 Dutchman Cire.

Weld 4

Point Beach Unit 1 Fluence. 12/16/91 . Fluence. 32 EFPV Material Location IS T/4 IS T/4 I 122P237 val Nozzle Belt 1.71E+18 1.16E+18 2.95E+18 2.00E+18 Forging A9811-1 Interm. Shell 1.55E+19 1.05E+19 2.68E+19 1.81E+19 Plate

C1423-1 Lower Shell 1.52E+19 1.03E+19 2.33E419 1.58E+19 Plate SA-1426 Nozzle Belt to 1.71E+18 1.16E+18 2.95E+18 2.00E+18 Interm. Shell Circ. Weld SA-1101 Interm. Shell 1.52E+19 1.03E+19 2.33E+19 1.58E+19 to lower Shell Circ. Weld

. SA-1101 Lower Shell to --- --- --- ---

Dutchman Cire.

Weld 4

5-9 4

4 4

Table 5. 2-1. Fluence Predictions for Beltline Reaion Materials (Cont.) i Point Beach Unit 1 (Cont.)

Fluence. 12/16/91 Fluence. 32 EFPY Material location IS T/4 IS T/4 SA-812 Interm. Shell 9.64E+18 6.53E+18 1.71E+19 1.16E+19 Longit. Weld (27% ID)

SA-775 Interm. Shell -- --- -> ---

Longit. Weld (73% OD)

SA-847 Lower Shell 9.53E+18 6.45E+18 1.56E+19 1.06E+19 Longit. Weld Point Beach Unit 2 Fluence. 12/16/91 Fluence. 32 EFPY Material location IS T/4 IS T/4 Not avail. Nozzle Belt 2.00E+18 1.35E+18 3.50E+18 2.37E+18 Forging 123V500 val Interm. Shell 1.67E+19 1.13E+19 2.92E+19 1.98E+19 Forging 122W195 val Lower Shell 1.65E+19 1.12E+19 2.66E+19 1.80E+19 Forging Not avail. Nozzle Belt to 2.00E+18 1.35E+18 3.50E+18 2.37E+18 Interm. Shell Circ. Weld l

l SA-1484 Interm. Shell 1.64E+19- 1.11E+19 2.56E+19 1.73E+19 to Lower Shell Circ. Weld Not, avail. Lower Shell to --- --- --- ---

Dutchman Cire.

Weld l 5-10

Table 5.2-1. Fluence Predictions for Beltline Reaion Materials (Cont.1 Surry Unit 1 Fluence. 12/16/91 Fluence. 32 EFPY Material location IS T/4 IS ,

T/4 122V109 val Nozzle Belt 2.48E+18 1.56E+18 5.27E+18 3.31E+18 Forging

• C4326-1 Interm. Shell 2.07E+19 1.30E+19 4.39E+19 2.76E+19 Plate C4326-2 Interm. Shell 2.07E+19 1.30E+19 4.39E+19 2.76E+19 Plate C4415-1 Lower Shell 2.07E+19 1.30E+19 4.39E+19 2.76E+19 Plate C4415-1 Lower Shell 2.07E+19 1.30E+19 4.39E+19 2.76E+19 Plate J726 Nozzle Belt to 2.48E+18 1.56E+18 S.27E+18 3.31E+18 Interm. Shell Cire. Weld SA-1585 Interm. Shell 2.07E+19 1.30E+19 4.39E+19 2.76E+19 to Lower Shell Circ. Weld (40% 10)

SA-1650 Interm. Shell --- --- --- ---

to Lower Shell Circ. Weld (60% 00)

SA-1494 Interm. Shell 3.34E+18 2.10E+18 7.08E+18 4.45E+18 2

Longit. Weld SA-1494 Lower Shell to 3.34E+18 2.10E+18 7.08E+18 4.45E+18 Longit. Weld

SA-1526 Lower Shell 3.34E+18 2.10E+18 7.08E+18 4.45E+18

Longit. Weld a

5-11 d

J

Table 5.2-1. Fluence Predictions for Beltline Reaion Materials (Cont.)

Surry Unit 2 Fluence. 12/16/91 Fluence. 32 EFPY Material location IS T/4 IS T/4 123V303 val Nozzle Belt 2.48E+18 1.56E+18 4.45E+18 2.80E+18 Forging C4208-2 Interm. Shell 2.07E+19 1.30E+19 3.71E+19 2.33E+19 Plate C4339-1 Interm. Shell 2.07E+19 1.30E+19 3.71E+19 2.33E+19 Plate C4331-1 Lower Shell 2.07E+19 1.30E+19 3.71E+19 2.33E+19 Plate C4339-2 Lower Shell  ?.07E+19 1.30E+19 3.71E+19 2.33E+19 Plate L737 Nozzle Belt to 2.48E+18 1.56E+18 4.45E+18 2.80E+18 Interm. Shell Cire. Weld R3008 Interm. Shell 2.07E+19 1.30E+19 3.71E+19 2.33E+19 to Lower Shell Circ. Weld SA-1585 Interm. Shell 4.32E+18 2.71E+18 7.75E+18 4.87E+18 Longit. Weld SA-1585 Interm. Shell 4.32E+18 2.71E+18 7.75E+18 4.87E+18 Longit. Weld (50% 10)

WF-4 Interm. Shell --- --- --- ---

(50 Ob)

WF-4 Lower Shell 4.32E+18 2.71E+18 7.75E+18 4.87E+18 Longit. Weld WF-4 Lower Shell 4.32E+18 2.71E+18 7.75E+18 4.87E+18 Longit. Weld (63% ID) l 5-12

i i

i Table 5.2-1. Fluence Predictions for Beltline Reaion Materials (Cont.)

Surry Unit 2 (Cont.)

Fluence. 12/16/91 Fluence. 32 EFPY Material Location 15 T/4 IS T/4 WF-4 Interm. Shell --- --- --- ---

Longit. Weld *

(37% OD)

Three Mile Island Unit 1 Fluence. 12/16/91 Fluence. 26.17 EFPY Material location ___JJ___ T/4 IS T/4 ARY 59 Lower Nozzle 2.68E+18 1.61E+18 6.60E+18 3.96E+18 Belt Forging C2789-1 Upper Shell 3.04E+18 1.83E+18 7.50E+18 4.50E+18 Plate C2789 2 Upper Shell 3.04E+18 1.83E+18 7.50E+18 4.50E+18 Plate C3307-1 lower Shell 2.92E+18 1.75E+18 7.20E+18 4.32E+18 Plate C3251-1 Lower Shell 2.92E+18 1.75E+18 7.20E+18 4.32E+18 Plate WF-70 Nozzle Belt to 2.68E+18 1.61E+18 6.60E+18 3,96E+18 Interm. Shell Circ. Weld WF-25 Upper Shell to 2.92E+18 1.75E+18 7.20E+18 4.32E+18 to Lower Shell Cire. Weld WF-67 Lower Shell to 1.70E+16 1.02E+16 4.20E+16 2.52E+16

, Dutchman Circ.

Weld (50% ID)

WF-70 Lower Shell to --- --- --- ---

Dutchman Circ.

Weld (50% 00) 5-13

1 Table 5.2-1. Fluence Predictions for Beltline Reaion Materials (Cont.)

Three Mile Island Unit 1 (Cont.)

Fluence. 12/16/91 Fluence. 26.17 EFPY Material location 15 T/4._ IS T/4 WF-8 Upper Shell 3.04E+18 1.83E+18 7.50E+18 4.50E+18 (

Longit. Weld SA-1526 Lower Shell 2.69E+18 1.62E+18 6.50E+18 3.90E+18 Longit. Weld SA-1526 Lower Shell 2.69E+18 1.62E+18 6.50E+18 3.90E+18

.I Longit. Welo (37% 10)

SA-1494 Lower Shell --- --- --- ---

Longit. Weld (63% 00)

• Turkey Point Unit 3 Fluence. 12/16/91 Fluence. 32 EFPY Material location IS T/4 IS T/4 122S146 val Nozzle Belt 1.80E+18 1.13E+18 3.17E+18 1.99E+18 Forging 123P461 val Interm. Shell 1.50E+19 9.42E+18 2.64E+19 1.66E+19 Forging ,

123S266 val Lower Shell 1.50E+19 9.42E+18 2.64E+19 1.66E+19 Forging SA-1484 Nozzle Belt to 1.80E+18 1.13E+18 3.17E+18 1.99E+18 Interm. Shell Circ. Weld SA-1101 Interm. Shell 1.50E+19 9.42E+18 2.64E+19 1.66E+19 to Lower Shell Circ. Weld SA-ll35 Lower Shell to --- --- --- ---

Dutchman Cire.

Weld i

l 5-14 l

I

l

! Table 5.2-1. Fluence Predictions for Beltline Reaion Materials (Cont.)

i Turkey Point Unit 4 i Fluence. 12/16/91 Fluence. 32 EFPY Material location IS T/4 IS- T/4 I

124S309 val Nozzle Belt 1.64E+18 1.03E+18 3.04E+18 1.91E+18 Forging -

i 123P481 val Interm. Shell 1.37E+19 8.60E+18- 2.53E+19 l~.59E+19 Forging i 122S180 val Lower Shell 1.37E+19 8.60E+18 2.53E+19 1.59E+19 i Forging WF-67 Nozzle Belt to 1.64E+18 1.03E+18 3.04E+18 1.91E+18

Interm. Shell i- Cire. Weld j (67%ID)

WF-70 Nozzle Belt to --- --- --- ---

Interm. Shell Circ. Weld (33% OD)

SA-Il01 Interm. Shell 1.37E+19- 8.60E+18 2.53E+19 ~1.59E+19 to Lower Shell e Circ. Weld SA-ll35 Lower Shell~to --- --- ---- ---

Outchman Circ.-

Weld Zion Unit 1 Fluence; 12/16/91 ' Fluence. 32 EFPY Material Location IS T/4 IS T/4 ANA 102 Lower Nozzle 3.22E+18 1;93E+18 -8.65E+18 5.19E+18 Belt Forging-C3795-2 Interm.- Shell 6.44E+18 3.87E+18 1.73E+19- 1.04E+19-Plate B7835-1 Interm. Shell 6.44E+18 3.87E+18 1.73E+19. 1.04E+19' Plate 5-15

Table 5.2-1. Fluence Predictions for Beltline Reaion Materials (Cont.) ,

I liRR_Lnit i 1 (Cont.)

Fluence. 12/16/91__ Fluence. 32 EFPY Material location IS T/4 i: T/4 l C3799-2 Lower Shell 6.44E+18 3.87E+18 1.73E+19 1.04E+19 Plate B7823-1 Lower Shell 6.44E+18 3.87E+18 1.73E+19 1.04E+19 Plate WF-154 Nozzle Belt to 3.22E+18 1.93E+18 8.65E+18 5.19E+18 Interm. Shell Cirt. Weld (82% ID)

SA-1769 Nozzle Belt to --- --- --- ---

Interm. Shell Cire. Weld (18% OD)

WF 70 Interm. Shell 6.44E+18 3.87E+18 1.~3E+19 1.04E+19 to Lower Shell Cire. Weld WF-154 Lower Shell to -- --- --- ---

Dutchman Cire.

Weld WF-4 Interm. Shell 2.34E+18 1.41E+18 6.29E+18 3.78E+18 Longit. Weld WF-8 Interm. Shell 2.34E+18 1.41E+18 6.29E+18 3.78E+18 Longit. Weld (39% ID)

WF-4 Interm. Shell --- --- --- ---

Longit. Weld (61% 00)

WF-8 Lower Shell 2.34E+18 1.41E+18 6.29E+18 3.78E+18 Longit. Weld l 5-16

l Table 5.2-1. Fluence Predictions for Beltline Reaion Materials (Cont.)

Zion Unit 2

! Fluence. 12/16/91 Fluence;'32 EFPY _.

Material location IS T/4 IS T/4 ZV 3855 Lower Nozzle 3.22E+18- 1.93E+18 8.45E+18 - 5.07E+18'-

Belt Forging l-l B8006-1 Interm. Shell 6.43E+18 3.86E+18 1.69E+19 1.01E+19 Plate I

B8040-1 Interm. Shell 6.43E+18 3.86E+18 1.69E+19 1.01E+19 j Plate i- C4007-1 Lower-Shell 6.43E+18 3.86E+18 1.69E419_ l.01E+19 i Plate i

. B8029-1 Lower Shell~ 6.43E+18 3.86E+18 1.69E+19 1.01E+19' Plate WF-200 -Nozzle Belt to 3.22E+18 1.93E+18 8.45E+18 . 5.07E+18 i Interm. Shell

! Circ. Weld

! SA-1769 Interm. Shell 6.43E+18 - 3.86E+18 1.69E+19 - 1 ole +19-

! to Lower Shell i -Circ. Weld l WF-154 Lower Shell to --- --- --- . ---

j- -Dutchman Circ.

! Weld a .

1- WF-70 Interm. Shell 2.30E+18- 1.38E+18 6.04E+18- 3.63E+18-l Longit. Weld i

i WF Lower Shell- 2.30E+18 - -

1.38E+18 - 6.04E+18 - 3.63E+18 l- Longit. Weld l

L l

c 5-17 4

,y.- - - - - -

nm.w, --,yy,,,,y

, ,--,7--,y,%.,ew[...,,g,,.w_,-., .,..,gm,y.,.,.. y, 3 9, p , ..w-. ,,s.,y,.., ,, - - ,s. y ,g _4

6. RESPONSE TO GENERIC LETTER 92-01 The following tables are submitteo in response to the information requested in Generic Letter 92 01.

Arkansas Nuclear One Unit 1 Table 1. Adherence to RYSP Requirements Table 2. Cv0SE Requirements Table 3. Unirradiated Charpy and RTuar Values Table 4. Material Heat Treatment Table 5. Beltline Mater sl Identification Table 6. Surveillance Material Identification Table 7. Chemical Composition Table 8. Effect of Irradiation Temperature Table 9. Utilization of Surveillance Results Table 10. Difference Between Measured and Predicted Embrittlement Effects Crystal River Unit 3 Table 1. Adnerence to RVSP Requirements Table 2. Cv0 SE Requirements Table 3. Unirradiated Charpy and RTuor Values Table 4. Material heat Treatment Table 5. Beltline Material Identification Table 6. Surveillance Material Identification Table 7. Chemical Composition Table 8. Effect of Irradiation Temperature Table 9. Utilization of Surveillance Results Table 10. Difference Between Measured and Predicted Embrittlement Effects 6-1

Davis-Besse Unit._1 Table 1. Adherence to RVSP Requirements Table 2. Cvu SE Requirements Table 3. Unirradiated Charpy and RTuo7 Values Table 4. Material Heat Treatment Table 5. Beltline Material Identification Table 6. Surveillance Material Identification Table 7. Chemical Composition Table 8. Effect of Irradiation Temperature Table 9. Utilization of Surveillance Results Table 10. Difference Between Measured and Predicted Embrittlement Effects R. E. Ginna Unit 1 Table 1. Adherence to RVSP Requirements Table 2. Cvu SE Requirements Table 3. Unirradiated Charpy and RTuo7 Values Table 4. Material Heat Treatment Table 5. Beltline Material Identification Table 6. Surveillance Material Identification Table 7. Chemical Composition Table 8. Effect of Irradiation Temperature Table 9. Utilization of Surveillance Results Table 10. Difference Between Measured and Predicted Embrittlement Effects Qgonee Unit 1 Table 1. Adherence to RVSP Requirements Table 2. CvUSE Requirements Table 3. Unirradiated Charpy and RTuoy Values  !

Table 4. Material Heat Treatment Table 5. Beltline Material Identification Table 6. Surveillance Material Identification Table 7. Chemical Composition 6-2 l

l l

\

4 l

Table 8. Effect of Irradiation Temperature Table 9. Utilization of Surveillance Results Table 10. Difference Between Measured and Predicted Embrittlement Effects Oconee Unit 2 Table 1. Adherence to RVSP Requirements Table 2. CvUSE Requirements Table 3. Unirradiated Charpy and RTuor Values I

'able 4. Material Heat Treatment i

Table 5. Beltline Material Identification Table 6. Surveillance Material Identification Table 7. Chemical Composition Table 8. Effect of Irradiation Temperature Table 9. Utilization of Surveillance Results Table 10. Difference Between Measured and Predicted Embrittlement Effects l Oconee Unit 3 Table 1. Adherence to RVSP Requirements l Table 2. Cv0SE Requirements l Table 3. Unirradiated Charpy and RTuoy Values

! Table 4. Material Heat Treatment I

Table 5. Beltline Material Identification

Table 6. Surveillance Material Identification

Table 7. Chemical Composition

Table 8. Effect of Irradiation Temperature Table 9. Utilization of Surveillance Results Table 10. Difference Between Measured and Predicted Embrittlement Effects Point Beach Unit 1 Table 1. Adherence to RVSP Requirements

Table 2. Cvu SE Requirements Table 3. Unirradiated Charpy and RTuor Values 6-3

4

Table 4. Material Heat Treatment

-Table 5. Beltline Material Identification Table 6. Surveillance Material Identification Table 7. Chemical Composition Table 8. Effect of Irradiation Temperature Table 9. Utilization of Surveillance Results Table 10. Difference Between Measured and Predicted Embrittlement Effects ;

1

, Eoint Beach Unit 2

] Table 1. Adherence to RVSP Requirements Table 2. Cvu SE Requirements Table 3. Unirradiated Charpy and RTuor Values ,

Table 4. Material Heat Treatment

. Table 5. Beltline Material Identification Table 6. Surveillance Material Identification Table 7. Chemical Composition Table 8. Effect of Irradiation Temperature Table 9. Utilization of Surveillance Results i

Table 10. Difference Between Measured and Predicted Embrittlement Effects Surry Unit 1 Table 1. Adherence to RVSP Requirements Table 2. Cvu SE Requirements Table 3. Unirradiated Charpy and RTuor Values Table 4. Material Heat Ireatment Table 5. Beltline Material identification Table 6. Surveillance Material Identification Table 7. Chemical Composition )

Table 8. Effect of Irradiation Temperature Table 9. Utilization of Surveillance Results Table 10. Difference Between Measured and Predicted Embrittlement Effects 6-4

i Surry Unit 2 i - Table 1. Adherence to.RVSP. Requirements Table 2. Cvu SE Requirenents a l

[ Table 3. Unirradiated Charpy and RTuor Values -

l Table 4. Material Heat Treatment Table 5. Beltline Material Identification' l

l Table 6. Surveillance Material Identification j Table 7. Chemical Composition j Table 8. Effect of Irradiation Temperature-l Table 9. Utilization of Surveillance Results-l Table 10. Difference Between Measured and Predicted Embrittlement Effects!

Three Mile Island Unit 1 ,

i j Table 1. Adherence to RVSP Requirements l- Table 2. Cvu SE Requirements Table 3. Unirradiated Charpy and RTuor Values i

3 Table 4. Material Heat Treatment

Table 5. Beltline Material Identification ~  ;

1-

{' Table 6. Surveillance Material Identification-

[ Table 7. Chemical Composition Table 8. Effect of Irradiation Temperature i Table C. Utilization of Surveillance Results '

' Table 10. Difference Between Measured and Predicted Embrittlement Effects 1

Turkey Point Unit 3

[

i

Table 1. Adherence-to RVSP Requirements Table 2. Cv0SE Requirements ,

4 .

i Table 3. Unirradiated Charpy and RTuor Values -

h Table 4. Material Heat Treatment h Table 5. Beltline Material Identification i

Table 6. Surveillance Material Identification

; Table 7. Chemical Composition' I

6-5 i

n

- - , - r + - - - - - - , v-- %, w- er-r+4 y , r- - , , ,--=,,, .<y w -y- p e me , -y ,

• w+ ggy e, r e-y m m- r %--e n p aw-vw-wwww.wp- w v e-

Table 8. Effect of Irradiation Temperature Table 9. Utilization of Surveillance.Results Table 10. Difference Between Measured and Predicted Embrittlement Effects Turkey Point Unit 4 Table 1. Adherence to RVSP Requirements Table 2. Cvu SE Requirements Table 3. Jnirradiated Charpy and RTuoy Values Table 4. Material Heat Treatment Table 5. Beltline Material Identification Table 6. Surveillance Material Identification Table 7. Chemical Composition Table 8. Effect of Irradiation Temperature Table 9. Utilization of Surveillance Results Table 10. Difference Between Measured and Predicted Embrittlement Effects Zion Unit 1 Table 1. Adherence to RVSP Requirements Table 2. Cv0SE Requirements Table 3. Unirradiated Charpy and RTuor Values Table 4. Material Heat Treatment Table 5. Beltline Material Identification Table 6. Surveillance Material Identification Table 7. Chemical Composition Table 8. Effect of Irradiation Tempera' re Table 9. Utilization of Surveillance Results Table 10. Difference Between Measured and Predicted Embrittlement Effects Zion Unit &

Table 1. Adherence to RVSP Requirements Table 2. Cvu SE Requirements Table 3. Unirradiated Charpy and RTuor Values 1

6-6

_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ 1

1 Tabic 4. Material Heat Treatment Table 5. Beltline Material Identification Table 6. Surveillance Material Identification Table 7. Chemical Composition Table 8. Effect of Irradiation Temperature Table 9. Utilization of Surveillance Results Table 10. Difference Between Measured and Predicted Embrittlement Effects 1

-i 6-7 -

c

TA8LE I. GENERIC LETTER 92-01 RESPONSE: SECTICN 1

Subject:

10CFR50, Appendix H; Adherence to RVSP Requirements Plant: Arkansas Nuclear One Unit 1 Does RVSP meet ASTM E 185-73, E 185-79, or E 185-82? Yes a No /

Question I:

Question II: Is plant one of the following? ANO-1, Crystal River-3, Davis Besse, R. E. Giana, Oconee-1, Oconee-2, Oconee-3, Point Beach-1, Point Beach-2, Rancho Seco, Surry-1, Surry-2, Turkey Point-3, Turkey Point-4, Zion-1, Zion-2. Yes / No o IF ANSWER IS "YES" TO EITHER QUESTION I OR QUESTION II, PROCEED TO TABLE 2.

IF ANSWER IS "N0" TO BOTH QUESTION I AND QUESTION II, PROCEED TO QUESTION III AND QUESTION IV.

Question III: If plan is to revise RVSP to meet requirements of 10CFR50, Appendix H, when will revised RVSP be submitted to NRC?

Response

Not applicable (see. Question I and II above)

Question IV: If plan is not to revise RVSP to meet requirements of 10CFR50, Appendix H, when will exemption from 10CFR50, Appendix H be requested from NRC?

Response

Not applicable (see Question I and II abeve)

NOTES: BAW-10006A, Revision 3: Surveillance Program Description (ASTM E 185-70)

i l

' TABLE 2. GENERIC LETTER-92-01 RESPONSE: SECTION 2, ITEM a

- Subjects 10CFR50, Appendix G, C,USE Requirements Plantf Arkansas Nuclear One Unit I Column I Column 2 Column 3 Column 4 Limiting Initial EFPY to reach If Column 2 is within license Action taken Material- USE Cy USE<50 ft-lb period: C USE at indicated time per IV.A.1 ft-lb Column 3A Column'3B i

12/16/91 E0L LIMITING Analysis per 10CFR50, ,

BELTLINE WELD Appendix G,Section V.C.3.  !

is scheduled for 1993 WF-Il2 70 (6) 8, approx. 48 43 under the sponsorship of B&WOG Reactor Vessel Working Group.

1 LIMITING BELTLINE PLATE .

OR FORGING-C5120 123'(7)- >32 NA NA NA NOTES FOR TABLE 2 ARE ON THE FOLLOWING PAGE.

_.__.r-_-__.

- ..c,.,ww, -m _ , _r m.m-- . .-__ ,

TABLE 2 (CONTINUED)

NOTES: (1) Fluence values taken at %-thickness.

(2) C,USE values for 12/16/91 and EOL (Column 3) calculated per requirements outlined in Regulatory Guide 1.99, Revision 2, Paragraph C.I.2.

(3) Analyses that have deconstrated the required margin of safety for the Zion and Turkey Point plants at load level A and B conditions have been submitted to the NRC (Reports BAW-2138P and BAW-2148P).

The results of these two analyses are anticipated to bound the outcome of the ANO Unit I analysis.

(4) C yUSE is calculated on the basis of RGl.99, Rev. 2, Position 1. SECY 91-333' states that this

. procedure is " inadequate." Recognizing this and having provided for the uniqueness of Linde 80 -

welds in BAW-1803, Rev.1, the method for calculating C,USE in BAW-1803, Rev.1, is put forward as more representative and is intended to be used for predicting the behavior of these welds in licensing applications.

(5) Result of fracture analysis presented in BAW-2075, Revision 1, demonstrate that the most limiting low upper-shelf welds have irradiated fracture toughness chrracteristics which will assure adequate margins of safety in accordance with the requirements of 10CFR50, Appendix G.

(6) BAW-1803 ,

(7) BAW-1829 i

-TABLE'3. GENERIC' LETTER 92-01 RESPONSE: SECTION 2, ITEM'b,-1 (1) 7

Subject:

10CFR50.61 and 10CFR50, Appendix G, III.A; Material Properties Related to

'  : PTS and Fracture Toughness Requirements Plant: Arkansas Nuclear One Unit I ,

Column 1 Column 2 Column 3 Column 4 ' Column 5 C.6-Beltline Unirradiated Charpy Test Results Unirrad. Unirrad. Method of Notes Materials Dropweight RT , Determng

. Col. 2a Col. 2b Col. 2c Col. 2d Test f RT ,

Results

.C C C T, .,

C y' 10*F. .30fl-lb- 50fl-lb 35 MLE cft-lb' .F F F i

FORGING

! AYN 131' 74,33,62 ND- ND ND ND +3 Est. (2) (1,3) ,

4 42,58,69 PLATE C5120-2 55,53,49; -16 +12 +5 -10 -10' NB-2331 (1,3,6) "

C5114-2 40,50,36- +14 +42- +30 -10 -10 NB-2331 (1,3,6).

C5120-1 56,48,54 -6 +19 +20. -10 -10 NS-2331 (1,3,6)

! .C5114-1 57,40,57 +10 +40 +35 0 0, NB-2331 (1,3,6) i

! WELD WF-182-1 36,33,44 ND ND ND ND -5 Est. (4) (1,5)

WF-ll2 35,40,30 'ND- ND ND ND- -5 Est. (4) (1,5)

.SA-1788 40,38,36 - ND ND ND ND -5 Est. (4) (1,5) '4 WF 45,46,38 ND ND ND ND -5 Est. (4) (1,5) j l

NOTES FOR TABLE 3 ARE ON THE FOLLOWING'PAGE.

_ t - - -

4 - .

TABLE 3 (CONTINUED)

NOTES TO TABLE 3:

(1) BAW-1820 (2) BAW-10046P, pp 3-17, -18; mean of most conservative value for each of 24 cases.

(3) Cy (+10F) values are for 60 hr stress-relief; other values for 40 hr stress-relief.

(4) BAW-1803, Revision 1, Tables 3-1 and 3-2; mean of RT , values for 34 Linde 80 welds.

(5) C (+10F) values are for 48 hr stress-relief.

(6) M}..Vernonqualificationtestdata.

S 4

TABLE 4. GENERIC LETTER 92-01 RESPONSE: SECTION 2, ITEM b, 1 (2)

Subject:

10CFR50.61 and 10CFR50, Appendix G, III.A; Material Properties Related to PTS and Fracture Toughness Requirements -- APPLICABLE ONLY TO REACTOR VESSELS CONSTRUCTED TO AN ASME CODE EARLIER THAN THE 1971 EDITION, SUMMER 1972 ADDENDA Plant: Arkansas Nuclear One Unit 1 Column 1 Column 2 Col. 3 Material Heat Treatment Notes BELTLINE t1ATERIALS AYN 131 1580120F-5h/WQ; 1250i20F-14h/WQ; 1100-Il50F-20:06h/FC (cumul.) (1,2)

C5120-2 1550-1600F-44h/BQ; 1200-1225F-5h/BQ; 1100-1150F-28%h/FC (cumul.)

C5114-2 1550-1600F-4%h/BQ; 1200-1225F-Sh/BQ; 1100-1150F-28%h/FC (cumul.)

C5120-1 1550-1600F-4%h/BQ; 1200-1225F-Sh/BQ; 1100-Il50F-26ih/FC (cumul.)

C5114-1 1550-1600F-4%h/BQ; 1200-1225F-Sh/BQ; 1100-Il50F-26\h/FC (cumul.)

WF-182-1 Il00-Il50F-19h/FC (cumul.)

WF-ll2 Il00-1150F-24h/FC_(cumul.)

SA-1788 Il00-Il50F-20h/FC (cumul.)

WF-18 (US Long.) 1100-Il50F-28;h/FC (cumul.)

WF-18 (LS Long.) 1100-1150F-26%h/FC (cumul.)

SURVEILLANCE MATERIALS-C5114-1 1550-1600F-4%h/BQ: 1200-1225F-5h/BQ; 1100-Il50F-29h/FC (1) 05114-2 1550-1600F-4%h/BQ; 1200-1225F-5h/BQ; 1100-1150F-29h/FC WF-193 Il00-Il50F-29h/FC NOTES:

(1) BAW-1820 (2) Additional stress relief information per Mt. Vernon process drawing.

(3) WQ - water quench BQ - brine quench FC - furnace cool

._ . . . _ _ . ~ . _ . . _ _ _ . _ , . - - _ . _ _ . - _ _ _ . _ . . _ . _ . _ _ _ _ . _ _ . _ _ . _ _ _ _ _ _ _ . _ . _ _

a

^

i

-TABLE 5. GENERIC LETTER 92-01 RESPONSE: SECTION 2, ITEM b, 1 (3) -

Subject:

"10CFR50.61 and 10CFR50, Appendix G, III.A; Material Properties Related to PTS and Fracture Toughness Requirements -- APPLICABLE ONLY TO REACTOR VESSELS CONSTRUCTED TO AN ASME CODE i

EARLIER'THAN THE 1971 EDITION SUMMER 1972 ADDENDA Plant

' Arkansas Nuclear One Unit 1 ,

Column 1 Column 2 Column 3 Column 4 Column'5 C. 6 c Beltline Heat Beltline Weld Wire Weld Flux Notes Plate or Number Weld Heat Lot Forging Lower NB forging .AYN 131, 528360 NB to US Circ.:'WF-182-1 821T44 8754 (1) i US to LS Circ.: WF-ll2.

US Plate' C5120-2 406L44- 8688 US Plate .C5114-2 LS to Dutch Circ.: SA-1788 61782 8754 -

LS Plate C5120-l' US Longit.: WF-18 8T1762 8650 LS Plate C5114-1 LS-Longit.: WF-18 8T1762 8650 4

NOTES: (1).. BAW-1820 (2) NB '- Nozzle Belt US - Upper Shell LS . Lower Stell i

b' l

i

i- i i

' TABLE 6.' GENERIC LETTER 92-01 RESPONSE: SECTION 2, ITEM b, T'(4) .

Subject:

10CFR50.61.and'10CFR50, Appendix G, III.A; Material Properties Related to PTS and Fracture

, Toughness .Requiremersts -- APPLICABLE ONLY TO REACTOR VESSELS CONSTRUCTED TO AN ASME CODE  ;

__ EARLIER ~THAN THE 1971 EDITION SUMMER 1972 ADDENDA [

i Plant: Arkansas Muclear One Unit 1 [

Column'l Column 2 Column 3 Column 4 Column 5

. t j Surveillance Surveillance Weld Wire Weld Flux Notes ~;

Plate or Weld Heat Lot I . Forging Heat Number I

'C5114-1 WF-193 406L44 8773 (1)  ;

C5114-2 I  :!

I i

l I NOTES: (1) BAW-1820  ;

i '

i e

i

! t d- '

4

'e

. ~ . - , . . .. . .

_ _ _ _ _ . . _ . _ . . _ . _ _ . . . _ _ _ _ .. .m. . . _ _ _ . _ . _ - _ _ _ . _ _ _ _ _ _ _ _ . _ _

l, TABLE

-s,

7. GENERIC LETTER 92-01 RESPONSE: SECTION 2. ITEM b. 1 (5) [,

Subject:

10CFR50.61 and 10CFR50, Appendix G, III.A; Material Pmperties Related to PTS and Fracture j Toughness Requirements -- APPLICABLE ONLY TO REACTOR VESSELS CONSTRUCTED TO AN ASME CODE j . EARLIER THAN THE 1971 EDITION, StHMER 1972 ADDENDA  ;

Plant: Arkansas Nuclear One Unit I i Column 1 Column 2 C. 3

' \

l' Material- Chemical Composition Meight Percent Notes

! C Mn P S Si Cr Ni No Cu BELTLINE.

MATERIAL i AYN 131 0.27 0.64 0.009 0.015 0.21 0.32 0.70 0.66 0.03 (1) 05120-2 0.22: 1.41 0.018 0.013 0.18 0.18 0.55 0.53 0.17 (1) ,

C5114-2.- 0.21 1.32 0.01c- 0.016 0.20 0.19 0.52 0.57 0.15 (1)  ;

C5120-1 0.22 1.41 0.014 0.013 0.18 0.18 0.55 0.53 0.17 (1)  ;

C5114-1 0.21 1.32 0.010 0.016 0.20 0.19 0.52- 0.57 0.15 (1) j' WF-182-1 0.08 1.69 0.014 0.013 0.45 0.14 0.63 0.40 0.24 (2) i WF-Il2 0.08 1.47 0.016 0.015 0.54- 0.07 0.59 0.40 0.31 (2) ;jl'

} WF-18 0.09- 1.45 0.004 0.017 0.39 0.12 0.55 0.41 1.20 (2)  !

SURVEILLANCE

( MATERIALS

{ 05114-1 0.21- 1.32- 0.010 0.016 0.20 0.19 0.52 0.57 0.15 (3)

C5114-2 0.21 1.32 0.010 0.016 0.20 0.19 0.52 0.57 0.15 (3) j WF-193  ?.09 1.49 0.016- 0.016 0.51 0.06 0.59 0.39 0.28 (3)

! REQUIRED: _ 5 tate heat number of weld wires used for cletermining above chemical composition if different from j that in.1 (3). -- Not applicaM e --

I NOTES:

(1) BAW-1820 j_. (2) BAW-2121P (3) BAW-1543, Revision 3 l

~

i i  !

L

, TABLE 8. GENERIC LETTER 92-01 RESPONSE: SECTION 3. ITEN a '!

)

Subject:

Generic Letter 88-11 Response Commitments; E7 xt of Irradiation Temperature.

4 Plant: Arkansas Nuclear One Unit I  !

i;

! Cold leg Temperature (T_,,): 565 F (See Figure 4-1) 1 ,. r i -

is <525 F, state how this was considered in determination of embrittlement e"fects i If T (, RT,) in accordance with Regislatory Guide 1.99, Revision 2:

(C,0$

Not applicable 1

t i

i t

1

1. . t

j.

References:

r a i None 4

1 1

6 t

1 i

i 4 .

l- TABLE 9. GENERIC LETTER 92-01 RESPONSE: SECTION 3, ITEM b

Subject:

Generic letter 88-11 Response Connitments; Utilization of Surveillance Results

]

, Plant: Arkansas 7 sear

.c One Unit I Were surveillance results used in determining C,USE? Yes a No /

Were surveillance results used in determining RT ,? Yes / No o

. If any "yes" boxes were checked above, state how the wweillar.ce results were used:

i l per Rec'latory Guide 1.99, Revision 2,. Position 2, for preparation of 4 Determination of RT, limit cu.ves for WF-182-1 and WF-ll2 weld materials only.

pressure-temperature 1.

i-

Reference:

BAW-2075, Revision 1

].

4 5

Y l

4 i

I 1

2 1

i-  ;

1. i i

~

, t 1-

TABLE 10. GENERIC LETTER 92-01 RESPONSE

SECTION 3, ITEM c i .  !

Subject:

Generic Letter 88-11 Response Commitments; Difference Between Measured and  !

q Predicted (Regulatory Guide 1.99, Revision 2) Embrittlement Effects i

} Plant: Arkansas ~ Nuclear One Unit 1 ,

i  ?

!.- Question'I. Does measured ART , exceed A% + 20 predicted by Regulatory Guide 1.99,  ;

j Revision 2?

! . _, Question II. Does measured yC USE drop exceed that obtained from Regulatory Guide 1.99,  !

j' P.evision 2, Figure 2? i Column 1- Column 2 Column 3 Column 4 Column 5 Column 6 Column 7 l 5 Beltlin?' Fluenge Neasured Predicted Question I- Neasured Predicted Ouestion II Material n/cm ART., ART.,+2a If "yes" C,USE C,USE If "yes" i (3) see Note (5) Drop Drop see Note (5) j 4

AYN 131' ---

ND ND' --

ND- ND --

i C5120-2 ----

ND ND --

ND ND --

i 4.28E+18 C5114-2 0(1) 115 No 17(1) 22(1) No 05120-1 ---

-ND ND --

ND ND - --

l l~ C5114-1 7.27E+17 10(2) 72 No 14(2) 19(2) No  !

' 1.03E+19 66(2) 140 No 11(2) 23(2) No  !

l.46E+19 . 38(2) 151 No 16(2) 25(2) No  ;

WF-182-1 1.96E+18 - 127(3). 151.- No 6(3) 17(4) No-5.92E+18 125(3) 200- No 13(3) 22(4) No j 1.29E+19 175(3) :237 No 8(3) 26(4) No .

9.62E+18 150(4) , 223 No 16(6) 24(4) No WF-112 1.50E+18- 78(3) . 157- No: 9(3) 19(6) No i- 8.95E+18 191(3i j 250 No 12(3) 27(6) No ,

9.86E+18 i- - 185(3) 256- No 12(3) 27(6) No ,

? . ..

8.21E+18 204(3 .246 No 29(3) 33(7) No

SA-1788 ---

ND ND --

ND ND --

WF-18 ---

ND ND --

ND ND , --

l

! NOTES FOR TABLE 10 ARE ON THE FOLLOWING PAGE.-

f I

g --r- c. r - e- , *= <r-- ~ ,e n- w 3 * , - -- .*-----c.,.i,y

, ..< ,,,e . ----.m ,.,_,. -- -- ,_ . . . -

l TABLE 10 (CONTINUED)

NOTES: (1) 'BAW-1698 (2) BAW-2075, Revision I (3) BAW-1803, Revision 1 (4) BAW-2125 3 (5) No statement required.

1 (6) BAW-2050 1

(7) BAW-1920P i

4 4

S J

1

l TABLE 1. GENERIC LETTER 92-01 RESPONSE: SECTION 1

Subject:

10CFR50, Appendix H: Adherence to RVSP Requirements Plant: Crystal River Unit 3 Question I: Does RVSP meet ASTM E 185-73. E 185-79, or E 185-82? Yes / No a Question II: L Is plant one of the following? ANO-1, Crystal River-3, Davis Besse, R. E. Ginna, Oconee-1, Oconee-2, Oconte-3, Point Beach-1, Point Beach-2, Rancho Seco, Surry-1,

-Surry-2, Turkey Point-3, Turkey Point-4, Zion-1, Zion-2 Yes / No a IF ANSWER IS "YES" TO EITHER QUESTION I OR QUESTION II, PROCEED TO TABLE 2.

IF ANSWER IS "N0* TO BOTH QUESTION I AND QUESTION II. PROCEED TO QUESTION III AND QUESTION IV.

Question III: If plan is to revise RVSP to meet requirements of 20CFR50, Appendix H, when will revised RVSP be submitted to NRC? _ , _ _ _ , ,

Response

Not applicable (see Question I an II .Luef.'

=> w. a Question IV: If plan is not to revise RVSP to meet requin33 eats of 10CFR50, Appendix H, when will exemption from 10CFR50, Appendix H be requeste um NRC?

' Response:

Not applicable (see Question I an II above)

NOTES: BAW-10100A:~ Surveillance Program Description (ASTM E-185-73)

I L_. . . _ _ _ _ _ _ . __

, -3 1 i i l

]-

t r

a  ;

i i

'. t i TABLE 2. GENERIC LETTER 92-Ol' RESPONSE: SECTION 2, ITEM a i 1, t

Subject:

10CFR50, Appendix G. C,USE Requirements

Plant

Crystal River Unit 3 l

Column 1 Column 2 Column 3 Column 4 ,

If Column 2 is within license Limiting Initial EFPY to reach Action taken -

Material USE Cy USE<50 ft-lb period: C,USE at indicated time per IV.A.I ,

i ft-lb t Column 3A Column 3B l t

12/16/91 EOL f LIMITING Analysis per 10CFR50, BELTLINE WELD Appendix G,Section V.C.3. f is scheduled for 1993

) WF-70 70 (5) 5, approx. 48 43 under the sponsorship of  ;

B&WOG Reactor Vessel  ;

l Working Group. [

j, i

! LIMITING I

!' BELTLINE PLATE l OR FORGING C4344-1 123 (6) >32 NA. NA NA

I

. NOTES FOR-TADLE 2 ARE ON THE FOLLOWING PAGE.

4 i t 1

i

[  !

. t i  !

c t

_ . _ . . . , _ , -~. _ , , - . - . , ~ . . . , _ _-..___ . , . . . - . . . _ , J

P.

TABLE 2 (CONTINUED)

NOTES: (1) Fluence values taken at 1.-thickness.

(2) C,USE values for 12,'16/91 and EOL (Column 3) calculated per requirements outlined in Regulatory Guide 1.99, Revision 2, Paragraph C.I.2.

(3) Analyses that have demonstrated the required margin of safety for the Zion and Turkey Point plants at load level A and B conditions have been submitted to the NRC (Reports BAW-2118P and BAW-2148P).

The results of these two analyses are anticipated to bound the outccme of the Crystal River Unit 3 analysis.

(4) C,USE is calculated on the basis of RGl.99, Rev. 2, Position 1. SECY 91-333 states that this procedure is " inadequate." Recognizing this and having provided for the uniqueness of Linde 80

. welds in BAW-1803, Rev.1, the method for calculating C,USE in BAW-1803, Rev.1, is put forward as more representative and is intended to be used for predicting the behavior of these welds in licensing applications.

~

(5) BAW-1803 (6) BAW-1820

TABLE 3. GENERIC LETTER 92-01 RESPONSE: SECTION 2, ITEM b. 1 (1)

Subjec': 10CFR50.61 and 10CFR50, Appendix G, III.A; Material Properties Related to PTS and Fracture Toughness Requirements Plant: Crystal River Unit 3-Column 1 Column 2 Column 3 Column 4 Column 5 C.6 Beltline Unirradiated Charpy Test Results Unirrad. Unirrad. Method of Notes Materials Dropweight RT., Determng Col. 2a Col. 2b Col. 2c Col. 2d Test F RT ,

'Results

%  %  %  % ['

10 F 30 ft-lb 50 ft-lb 35 MLE

'ft-lb F F F FORGING AJZ 94 103,96,97 ND ND ND ND +3 Est. (2) (1,3) 101,111,91 PLATE:

C4344-1 39,40,36- +50 +80 Not avail. -10 +20 NB-2331 (1,3,7)

C4344-2 42,40,30 +30- +80 Not avail.- -10 +20 NB-2331 (1,3,7)

C4347-1' 53,54,47 . +20 +50 45 -20 -10 NB-2331 (1,3,7)

C4347-2 43,53,63 +85 +105 +95 +45- .N8-2331 (1,3,7)

WELD SA-1769 36,35,38. ND- ND ND . ND -5 ,Est. (4) (1,5)

WF-169-1 36,43,42 ND ND ND ND -5 Est. (4) (1,5) 42,29,46 WF-8. 45,38,30 ND ND ND. ND -5 Est. (4) (1,5)

WF-18 45,46,38 ND ND ND ND -5 Est. (4) (1,5)

WF-70 39,35,44 ND ND ND ND +18 Eval.(6) (1,5,8)

SA-1580 31,29,25 NO HD ND ND -5 Est. (4) (1,5) 49,41,40.

WF-154 41,37,43 ND ND ND ND -5 Est. (4) (1,5)

NOTES FOR TABLE 3 ARE-ON THE FOLLOWING PAGE.

TABLE 3 (CONTINUED)

NOTES TO TABLE 3:

(1) BAW-1820 (2) BAW-10046P, pp 3-17, -18; mean of most conservative.value for each of 24 cases.

(3) Cy (+10F) values are for 60 hr stress-relief; other values for 27-40 hr stress-relief.

(4) BAW-1803, Revision 1, Tables 3-1 and 3-2; mean of RT , values for 34 Linde 80 welds.

(5) Cy (+10F) values are for 48-80 hr stress-relief.

(6) BAW-2100 (7) Supplementary Mt. Vernon test of surveillance material.

(8) RT , value for 40 hr stress-relief maximum. l O

I

(

, TABLE 4. ' GENERIC LETTER 92-01 RESPONSE: SECTION 2. ITEM b. 1 (2) l

Subject:

10CFR50.61 and 10CFR50, Appendix G, III.A; Material Properties Related to PTS and Fracture i Toughness Requirements -- APPLICABLE ONLY TO REACTOR VESSELS CONSTRUCTED TO AN ASME CODE

! EARLIER THAN THE 1971 EDITION, SUMMER 1972 ADDENDA i

j Plant: Crystal River Unit 3 Column'1 Column 2 Col. 3 1 Material Heat Treatment Notes i

BELTLINE i- MATERIALS

AZJ.94 1590120F-7h/WQ; 1270120F-14h/WQ; 1100-Il50F-22%h/FC (cumul.) (1,2)

~

C4344-1 1550-1600F-4%h/BQ; ll75-1200F-6h/BQ; Il00-1150F-27h/FC (cumul.)

C4344-2 1550-1600F-4%h/BQ; ll75-1200F-6h/BQ: Il00-Il50F-27h/FC (cumul.)

C4347-1 1550-1600F-4 h/BQ; 1250-1275F-Sh/BQ; 1100-Il50F-24%h/FC (cumul.)

C4347-2 1550-1600F-4%h/BQ; 1250-1275F-Sh/BQ; 1100-Il50F-24%h/FC (cumul.)

. SA-1769 Il00-1150F-19%h/FC-(cumul.)

i WF-169-1 Il00-1150F-19%h/FC (cumul.) .  ;

WF-70 1100-Il50F-20%h/FC (cumul.)

• i WF-154 1100-Il50F-27h/FC (cumul.)

j WF-8 Il00-Il50F-27h/FC (cumul.)

f WF-18 Il00-1150F-27h/FC (cumul.)  !

SA-1580 Il00-Il50F-24%h/FC (cumul.)

. . i t

SURVEILLANCE-MATERIALS-C4344-1 .1550-1600F-4%h/BQ; Il75-1200F-6h/BQ; 1100-Il50F-27h/FC (1) i j C4344-2 . 1550-1600F-4%h/BQ; ll75-1200F-6h/BQ; 1100-Il50F-27h/FC i 1- WF-209-1 .

1100-1150F-27h [ cooling not reported] 4 Atypical weld 1100-Il50F-27h (cooling not reported]

', NOTES FOR-TABLE 4 ARE ON'FOLLOWING PAGE.

{-

i i

L  !

i i

, . _ , _ . . . . ,._....c... _ __ . . . . _. _ ,_ , - _ , ._. . - - ,_,_ . , . . . _ . .-

TABLE 4. (CONTINUED)

NOTES:

(1) BAW-1820 (2) Additional stress relief information per Mt. Vernon process drawing.

(3) WQ - water quench BQ - brine quench FC - furnace cool O'

l

~

- - . -i-..----

j-i 4

{- t g- TABLE 5. GENERIC LETTER 92-01 RESPONSE: SECTION 2, ITEM b. 1 (3)

.; t

Subject:

10CFR50.61 and 10CFRSC, Appendix G, III.A; Material Properties Related to PTS and Fracture j l Toughness Requirements -- APPLICABLE ONLY TO REACTOR VESSELS CONSTRUCTED TO AN ASME CODE EARLIER THAN THE 1971 EDITION, SUMMER 1972 ADDENDA Plant: C'rystal River Unit 3

' t Column'I Column 2 Column 3 Column 4 Column S C. 6 i t

j Beltline Heat Beltline Weld Wire Weld Flux Notes Plate or-Number Weld Heat Lot

i. Forging

lower NB Forging AZJ 94, 123Vl90 NB to US Circ.(ID 40%)

SA-1769 71249 8738 (I)

US Plate C4344-1 NB to US Circ.(00 60%): WF-169-1 8T1554 8754

US Plate C4344-2, US to LS Circ.

WF-70 72105 8669 i LS to Dutch. Circ.: WF-154 LS Plate C4347-1 406L44 8720 LS Plate. C4347-2 US Longit.: WF-8 8T1762 8632 i US Longit.: WF-18 8T1762 8650  !

LS Longit.: SA-1580 8T1762 8596 i

! NOTES: (1) BAW-1820 (2) NB.- Nozzle Belt i US - Upper Shell  !

LS - Lower Shell i

i 4 -- .t 4-e '

7 i

i:

, . . - . .- ,-. , . , - _ - . - . - . - .. a

TABLE 6. GENERIC LETTER 92-01 RESPONSE: SECTION 2, ITEM b, 1 (4)

Subject:

10CFR50.61 and 10CFR50, Appendix G, III.A; Material Properties Related to PTS and Fracture Toughness Requirements -- APPLICABLE ONLY TO REACTOR VESSELS CONSTRUCTED TO AN ASME CODE EARLIER THAN THE 1971 EDITION, SUMMER 1972 ADOENDA Plant:' Crystal River Unit 3 Column I Column 2 Column 3 Column 4 Column 5 Surveillance Surveillance Weld Wire Weld Flux Notes Plate or Weld Heat Lnt Forging.

Heat Number C4344-1 WF-209-1 72105 8773 (1)

C4344-2 Atypical Atypical 8773 NOTES: (1) BAW-1820 l

l' l

i l

i 1 TABLE 7. GENERIC LETTER 92-01 RESPONSE: SECTION 2, ITEM b. 1 (5)

Subject:

10CFR50.61.and 10CFR50, Appendix G, III.A; Material Properties Related to PTS and Fracture Toughness Requirements -- APPLICABLE ONLY TO REACTOR VESSELS CONSTRUCTED TO AN ASME CODE i EARLIER THAN THE 1971 EDTTION. SUMMER 1972 ADDENDA Plant: Crystal' River Unit 3 Column 1 Column 2 C. 3 Material Chemical Composition. Weight Percent Notes C Mn P S Si Cr Ni Mo Cu BELTLINE MATERIALS  !

AZJ 94 0.26 0.65 0.007 0.016 0.24 0.34 l0.72 0.62 ND (1) ,

04344-1 0.23 1.30 0.008 0.016 0.22 0.11 'O.54 0.55 0.20 (1)

C4344-2 0.23 1.30 0.008 0.016 0.22 0.11 0.54 0.55 0.20 (1)

C434', 1 0.22 1.32 0.013 0.015 0.24 0.11 0.58 0.55 0.12 (1)

C4347-2 0.22 1.32. 0.013 0.015 0.24 0.11 0.58 0.55 0.12 (1)

SA-1580 0.07 1.45 0.015 0.013' O.43 0.12 0.55 0.41 0.20 (2)

SA-1769 , 0.09' I.49 0.020 0.014 0.56 0.16 0.61 0.37 0.26 (2)

WF-8 0.06 1.45 0.009 0.009- 0.53 0.12 0.55 0.41 0.20 (2)

WF-18 0.09 1.45 0.004 0.017 0.39 0.12 0.55 0.41 0.20 (2)

WF 0.09 :1.63 0.018 0.009 0.54 0.10 0.59 0.40 0.35 (2)

WF-169-1 0.08 1.56 0.016 0.016 0.45 0.08 0.63 0.37 0.18 (2) ,

SURVEILLANCE ,

MATERIALS -

C4344-1 0.23- 1.30 0.008 0.016 0.22 0.11 0.54 0.55 0.20 (3)

C4344-2 0.23 1.30' O.008 0.016 0.22- 0.11 0.54- 0.55 0.20 (3)

Atypical 0.08 1.65 0.021 0.013 1.00 0.07 0.10 0.45 0.41 (3)

REQUIRED: State heat number of weld wires used for determining above chemical composition if different f rom that in 1 (3).

-- Not applicable --

NOTES FOR TABLE 7 ARE ON THE FOLLOWING PAGE.

i

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i

! TABLE 8. GENERIC LETTER 92-01 RESPONSE: SECTION 3 ITEM.a

! . 1

Subject:

Generic Letter 88-11 Response Comitments: Effect of Irradiation Temperature

!- Plant: Crystal River Unit 3 ColdlegTemperature(T,,,$): 556 F (See Figure 4-1)

- is <525 F, state how this was considered in determination of embrittlement effects '

If (CUS T y (, RT,) in accordance with Regulatory Guide 1.99, Revision 2: -

1 .

! Not applicable-1 i i

i.

I f'

i l

i-i, References-None 4

t

t i

i

' -TABLE 9. GENERIC LETTER 92-01 RESPONSE: SECTION 3, ITEM b t i

Subject:

. Generic Letter 88-11 Response Commitments; Utilization of Surveillance  !

Results Plant: Crystal River Unit 3 Were surveillance results used in determining C.USE? Yes a No / I Were surveillance results used in determining RT_,7 Yes / No o ,

i If any "yes" boxes were checked above, state how the surveillance results were used- t i

l Determination of RT., using measured values for " atypical

• weld material only. RT., was also used for preparation of pressure-temperature limit curves.  ;

i i

i

Reference:

- BAW-2049 l t

I f

i t

l TABLE 10. GENERIC LETTER 92-01 RESPONSE: SECTION 3 ITEM c

Subject:

Generic Letter 88-11 Response Commitments: Difference Between Measured and Predicted (Regulatory Guide 1.99, Revision 2) Embrittlement Effects Plant: Crystal River Unit 3

. Question I.- Does measured ART , exceed ART , + 20 predicted by Regulatory Guide 1.99, Revision 2?

Question II. Does measured yC USE drop exceed that obtained from Regulatory Guide 1.99, Revision 2. Figure 2?

Column 1 Column 2 Column 3 Column 4 Column 5 Column 6 Column 7 Bel tl'ine fluenge Measured Predicted Question I Measured Predicted Question II Material ART., If "yes" CyUSE CyUSE If "yes" n/cm ART 1+2a (1,3) see Note (5) Drop Drop see Note (5)

AZJ 94- ---

ND ND --

ND ND --

No' 5(1). 17(1) No C4344-1 L1.17E+18 '21(1) _ 98

'6.56E+18'- 159 No 18(1)- 25(1) No 126(1) 164 No 22(1) 26(1) No 7.50E+18 .97(1) 1.08E+19' 179 No 23(1) 28(1) No 128(1)

6.56E+18 159 No- 17(2) 25(2) No C4344-2 127(2)

C4347-1 --- ND ND --

ND ND --

C4347 ---

ND ND --

ND ND -

SA-1769 ---

ND- ND -- ND ND --

WF-169-1 ---

ND ND -- ND ND --

6.63E+18' 259 No 13(3) 25(4) No WF-70 135(3)

WF-154- ---

ND ND- -- ND ND -

WF-8 ---

ND ND- --

ND ND --

WF-10 --- ND ND --

ND ND -

SA-1580- ---

ND- ND --

ND ND -

Atypical l.17E+18 138 No 9(1) 25(1) No 28(1)-

6.56E+18 216' No 16(1) 32(1) No 122(1) 7.50E+18 223 No 11(1) 32(1) No 119(1) 1.08E+19 242 No 15(1) ~34(1) No' 120(1)

NOTES FOR TABLE 10 ARE ON THE FOLLOWING PAGE.

---

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TABLE I. GENERIC LETTER 92-01 RESPONSE: SECTION I

Subject:

10CFR50, Appendix H; Adherence to RYSP Requirements Plant:' Davis Besse Unit I Question I: Does RVSP meet ASTM E 185-73, E 185-79, or E 185-82? Yes / No a Question II: Is plant one of the following? ANO-1, Crystal River-3, Davis Besse, R. E. Ginna, Oconee-1, Oconee-2, Oconee-3, Point Beach-1, Point Beach-2, Rancho Seco, Surry-1, Surry-2. Turkey Point-3, Turkey Point-4, Zion-1, Zion-2. Yes [ _ No o IF ANSWER'IS "YES" TO EITHER QUESTION I OR QUESTION II, PROCEED TO TABLE 2.

IF ANSWER'IS "N0* TO BOTH QUESTION I AND QUESTION II, PROCEED TO QUESTION III AND QUESTION IV.

Question III: If plan is to revise RVSP to meet requirements of 10CFR50, Appendix H, when will revised RVSP be submitted to NRC?

Response

Nol applicable (see Question I and II above) s Question IV: If plan is not' to revise RVSP to meet requirements of 10CFR50, Appendix H, when will exemption from IOCFR50, Appendix H be requested from NRC?

Response

Not applicable (see Question I and II above)

NOTES: BAW-10100A: Surveillance Program Description (ASTM E 185-73)

L=

i i

TABLE 2. GENERIC LETTER 92-01 RESPONSE:-SECTION 2, ITEM'a

Subject:

10CFR50, Appendix G, C,USE Requirements Plant: Davis-Besse Unit 1 t

[

i

'Cclumn 1 Column 2 Column 3 Column 4 limiting Initial EFPY to reach If Column 2 is within license Action taken Material USE C,USE<50 ft-Ib period: C.USE at indicated time per IV.A.I  :

ft-lb 4 Column 3A Column 38 1 t

12/16/91 EOL t

{. LIMITING -

BELTLINE WELD .

i  !

4 WF-233 70 (2) >32 NA NA - NA f LIMITING BELTLINE PLATF OR FORGING l BCC 241 118 (3)' >32 NA NA NA i

NOTES: (1) Fluence values taken at L.-thickness.

~

) (2) BAW-1803 (3)- BAW-1820

~

i i i i

i ,

I i 1 ,

3 i l  :

i i

I

TABLE 3. GENERIC LETTER 92-01 RESPONSE: SECTION 2. ITEM b. 1 (1)

Subject:

10CFR50.61 ar.d 10CFR50, Appendix G, III.A; Material Properties Related to PTS and Fracture Toughness Requirements Plant: Davis-Besse 'Jnit 1 Column 1 Column 2 Column 3 Column 4 Column 5 l C.6 Beltline Unirradiated Charpy Test Results Unirrad. Unirrad Method of Notes Dropweight Determng Materials .

Col. 2a Col. 2b Col. 2c Col. 2d Test RT , RT ,7 Results F T

C, C C, C, yr 10 F 30f[-lb 50 ft-lb 35 MLE ft-lb F F F FORGING ADB 203 71,70,67 +48 465 Not avail. +50 +50 NS-2331 (1,3,6) 118.113,102 AKJ 233 ND -15 +30 +15 +20 +20 NB-2331 (1,4,6)

BCC 241 ND -14 +27 +5 +50 +50 NB-2331 (1,4,6)

WEID WF-232 25,31,35 ND ND ND ND -5 Est. (2) (1,5) l WF-233 43.30,26 ND ND ND ND -5 Est. (2) (1,5) l WF-IS2-1 36,33,44 +5 +62 Not avail. -20 +2 NB-2331 (I,5)

NOTES FOR TABLE 3 ARE ON THE FOLLOWING PAGE.

l

TABLE 3 (CONTINUED)

NOTES TO TABLE 3:

(1) BAW-1820 (2) BAW-1803, Revision 1, Tables 3-1 and 3-2; mean of RT , values for 34 Linde 80 welds.

(3) Cy (+10F) values are for 40 hr stress-relief; other values for unknown stress-relief.

(4) Values are for 15h hr stress-relief.

(5) C,(+10F) values are for 48 hr stress-relief.

(6) Supplementary Mt. Vernon tests of excess surveillance program material.

i b

s -TABLE 4. GENERIC LETTER 92-01 RESPONSE: SECTION 2, ITEM b. 1 (2)

Subject:

10CFR50.61 and 10CFR50, Appendix G, III.A; Material Properties Related to PTS and Fracture Toughness Requirements -- APPLICABLE ONLY TO REACTOR VESSELS CONSTRUCTED TO AN ASME CODE EARLIER THAN DIE 1971 EDITION. SUPMER 1972 ADDENDA ,

Plant: Davis-Besse Unit 1 Column 2 Col. 3 Column 1 Heat Treatment Notes Material BELTLINE MATERIALS (1,2)

- ADB 203 1590110F-6h/WQ: 1240110F-14h/WQ: Il00-Il50F-15%h/FC (cumul.)

AKJ 233 1590110F-4h/WQ; 1240110F-6h/AC; 1100-Il50F-15h/FC (cumul.)

BCC 241 1590110F-4h/WQ;.1240110F-Sh/AC; 1100-Il50F-15h/FC (cumul.)

WF-232 Il00-Il50F-14h/FC (cumul.)

WF-233 1100-Il50F-14h/FC (cumul.)

~

l

-; WF-IE2 Il00-Il50F-15h/FC (cumul.) l WF -232 1100-Il50F-14%h/FC (cumul.) l Il00-Il50F-14%h/FC (cumul.)

jy-233

,ifSURVEILLANCE

<IATERIALS J

SCC 241 1590110F-4h/WQ; 1240110F-Sh/AC; 1100-Il50F-15%h/FC (1)

AKJ '233 1590110F-4h/WQ; 1240110F-6h/AC; 1100-Il50F-15)h/FC 1100-Il50F-15%h/FC WF-182-{

NOTES:

(1) BAW-1820' (2) Additional. stress relief information per MT. Vernon process Crawing.

(3) WQ - water quench AC - air cool FC - furnace cool-

~

I' i

i i

TABLE 5. GENERIC LETTER 92-01 RESPONSE: SECTION 2, ITEM b. 1 (3)

Subject:

1GCFR50.61 and 10CFR50, Appendix G, III.A; Material Properties Related to PTS and Fracture

, Toughness Requirements -- APPLICABLE ONLY TO REACTOR VESSELS CONSTRUCTED TO AN ASME CODE EARLIER THAN THE 1971 EDITION. SINER 1972 ADDENDA Plant: Davis-Besse Unit I -

, i Column'l Column 2 Column 3 Column 4 Column 5 C. 6 Beltline. Heat Beltline Weld Wire Weld Flux Notes i Plate or Number Weld Heat tot Forging  ;

NB Forging ADB 203, 123Y317 NB to US Circ.(ID 9%): WF-232 8T3914 8790 (1) ,

, i US Forging AKJ 233, 123X244 NB to US Circ.(00 91%): WF-233 T29744 8790

, LS Forging .BCC 241, SP4086 US to LS Circ.: WF-182-1 821T44 8754 j LS to Dutch Circ.(ID 12%): WF-232

~

'8T3914 8790 LS to Dutch Circ.(00 88%): WF-233 T29744 8790 2

. NOTES: (1) BAW-1820 l- (2) NB -: Nozzle Belt ,

i. US - Upper Shell F LS:- Lower Shell 4 ,

i a

t T

4

TABLE 6. GENERIC LETTER 92-01 RESPONSE: SECTION 2, ITEM b. 1 (4)

Subject:

. 10CFR50.61 and 10CFR50, Appendix G, III.A; Material Properties Related to PTS and Fracture Toughness Requirements -- APPLICABLE ONLY TO REACTOR VESSELS CONSTRUCTED TO AN ASME CODE EARLIER THAN THE 1971 EDITION. SUPMER 1972 ADDENDA '

Plant: Davis-Besse Unit I Column 1 Column 2 Column 3 Column 4 Column 5 Surveillance Surveillance Weld Wire Weld Flux Notes Plate or Weld Heat Lot-Forging Heat Number BCC 241, SP4085 WF-182-1 821T44 8754 (1)

'AKJ 233, 123X244 i

NOTES: (1) BAW-1820

(

t TABLE 7. GENERIC LETTER 92-01 RESPONSE: SECTION 2 ITEM b. 1 (5)

Subject:

10CFR50.61 and 10CFR50, Appendix G, III.A; Material Properties Related to PTS and Fracture Toughness Requirements -- APPLICABLE ONLY TO REACTOR VESSELS CONSTRUCTED TO AN ASME CODE futLIER THAN THE 1971 EDITION, SUMMER 1972 ADDENDA Plant: Davi 'asse Unit 1 Column I . Column 2 C. 3 Material Chemical Composition, Weight Percent Notes ,

C Mn P S Si Cr Ni Mo Cu BELTLINE MATERIALS ADB 203 0.23 0.70 0.007 0.009 0.29 0.39 0.68 0.63 0.04 (1)

AKJ 233 0.26 0.68 0.004 0.006 0.30 0.38 0.77 0.64 0.04 (1)

BCC 241 0.22 0.63 0.01' O.011 0.27 0.32 0.81 0.63 0.02 (1) i WF-232 0.06 1.30 0.016 0.011 0.47 0.11 0.64 0.37 0.18 (2)

WF-233 0.05 1.45 0.021 0.015 0.42 0.08 0.68 0.44 0.29 (2)

WF-182-1 0.08 1.69 0.014 0.013 0.45 0.14 0.63 0.40 0.24 (2)

SURVEILLANCE MATERIALS BCC ,241 0.22 0.63 0.011 0.011 0.27 0.32 0.81 0.63 0.02 (3) '

AKJ 233 0.26 0.68 0.004 0.006 0.30 0.38 0.77 C 64 0.04 (3)

WF-182-1 0.09 1.69 0.014 0.013 0.41 0.15 0.63 , la.40 0.21 (3)

REQUIRED: State heat number of weld wires used for determining above chemical composition if dif ferent from that'in T (3). -- Not applicable --

NOTES:

(1) BAW-1820 (2) BAW-2121P (3) BAW-1543, Revision 3

TABLE 8. GENERIC LETTER 92-01 RESPONSEi SECTION 3. ITEM a

Subject:

-Generic Letter 88-11 Response Commitments: Effect of Irradiation Temperature Plant: Davis-Besse Unit 1 Cold Leg Temperature (T,,,,,): 556 F (See Figure 4-2)

If T is <525 F,' state how this was considered in determination of embrittlement effccts (Cy U$$,, RT,37) i.o accordance with Regulatory Guide 1.99, Revision 2:

I Not applicable

< 4 Referencesi None:

_u 4:

._._m.-_-

TABLE 9. GENERIC LETTER 92-01 RESPONSJ: SECTION 3, ITEM b

Subject:

Generic Letter 88-11 Response Commitments; Utilization of Surveillance Results Plant: Davis-Besse Unit 1 Were surveillance results used in determining CoUSE? Yes a No /

Were surveillance results used in determining RT ,,7 Yes / No o If any "yes" boxes were checked above, stat- ilow the surveillance results were used:

per Regulatory Guide 1.99, Re tision ?, Position 2, for preparation of

. Determination of RT, limit curves for WF-182-1 and WF-233 weld materials only.

pressure-temperature

Reference:

BAW-2125

. _ _ _ _.. _ _ . _ ..~._ . __ _ . _. . . _ . _ _ . _ _ _ _ . . _ _ _ . _ - . . . . _ _ . _ . _ _ _ _ _ _ . _ . . . _ . _ _ _ _ _ _ - .

i

. . l TABLE 10. GENERIC LETTER 92-01 RESPONSE: SECTIDN 3, ITEM c-

Subject:

Generic letter 88-11 Response Commitments; Difference Between Measured and Predicted (Regulatory Guide 1.99, Revision 2) Embrittlement Effects Plant: Davis-Besse Unit 1 Question I.- Does measured ART ,g exceed ART,g + 2a predicted by Regulatory Guide 1.99, Revision 2?

Question II. Does measured yC USE drop exceed that obtained from Regulatory Guide 1.99, Revision 2, Figure 2?

Column 1 Column 2 Column 3 Column 4' Column 5 Column 6 Column 7 Beltline -Fluenge Measured Predicted Question I Measured Pred' ted Question II Material n/cm ART,n ART,3+2a If "yes" CyOSE C,2ii If "yes" l

(3) see Note (4) Drop D. + > see Note (4)

ADB 203. ---

ND ND --

ND -- --

AKJ 233 1.29E+19 2(1) 56 No 13(1) 17(1) No BCC 241 1.96E+18 0(2) 24 --

9(2) 9(2) No 5.92E+18 0(2) 34- --

9(2) 11(2) No 1.29E+19 28(2) 44 No 4(2) 14(2) No 9.62E+18 3(2) 40 No- 5(2) '12(2)' No.

WF-232 ---

. ND ND --

ND ND --

WF-233 4.67E+18 191(3) 211 No 18(3) 23- No 28 l.08E+19 187(3) '257 No 24(3) No

, 1.21E+19; 222(3) 263 No 19(3) 28 No-WF-182-1 1.96E+18 127(3) 152- No 6(3) 17(2) No-5.92E+18 125(3) 200 No 13(3). 22(2) No 1.29E+19 175(3) 237 No 8(3) '26(2) Nc 9.62E+18' 150(2)' 223 No 16(2) 25(2) No NOTES FOR TABLE 10 ARE ON THE FOLLOWING PAGE.

t

TABLE 10 (CONTINUED)

NOTES: ( 1 ) .. .BAW-1882, Revision 1

~

(2) BAW-2125 (3) BAW-1803, Revision 1 (4) No Statement required.

4 .

t e

-_1

J E

i i

TABLE.1. GENERIC' LETTER 92-01 RESPONSE:'SECTION 1- ^^

N 1

i

Subject:

10CFR50, Appendix H; Adherence to RVSP Requirements- -

-Plant: R. E. Ginna 4

Question I: Does RVSP meet ASTM E 185-73, E 185-79, or E 185-82? Yes f(l)No a Question II: Is plant one of the following? ANO-1, Crystal . Rive -3, Davis Besse, R. E. Ginna,

'Oconee-1, Oconee-2, Oconee-3, Point Beach-1, Point Beach-2, Rancho Seco, Surry-1, Surry-2, Turkey Point-3, Turkey Point-4, Zion-1, Zion-2. ' Yes / No a IF' ANSWER IS "YES" TO.EITHER' QUESTION I OR QUESTION II, PROCEED TO TABLE 2.

IF ANSWER IS "N0" TO BOTH QUESTION I AND QUESTION II, PROCEED TO QUESTION III AND QUESTION IV. j c- a Question!III:. If plan .is to revise RVSP to meet requirements.of 10CFR50, Appendix H, when will

-revised RVSP be' submitted to NRC?

. Response:  ;

i Not applicable.(see Question I and II above)

I Question IV: If plan 'is not to revise RVSP to meet requirements of 10CFR50, Appendix H, when will exemption from 10CFR50, Appendix H be requested from NRC?

Response

, Not applicable _(see Question I and II above) '!

NOTES: . (1)' Robert E. Ginna Final Safety Analysis' Report, Revssion 6, Docket No. 50-244, December 1999.

4 1

i:

• 4 l

.. . . . m , . m . - _

TABLE"2. GENERIC LETTER 92-01 RESPONSE: SECTION 2, ITEM a

Subject:

10CFR50, Appendix G, CuUSE Requirements Plant: R. E. Ginna Column 1 Column 2 Column 3 Column 4 Limiting Initial EFPY to reach If Column 2 is within license Action taken Material USE. Cy USE<50 ft-lb period: C USE at' indicated time per IV.A.1 ft-lb Column 3A Column 38 12/16/91 EOL Analysis per 10CFR50, LIMITING.

BELTLINE WELD Appendix G,:Section V.C.3.

is scheduled for 1993 -

41 37 under the sponsorship of SA-847 70 (5) 4, approx.

B&WOG Reactor Vessel Working Grcup.

LIMITING' BELTLINE-PLATE.

OR FORGING

>32 'NA NA NA lESS255~ 124'(6)

NOTES FOR: TABLE 2 ARE ON THE FOLLOWING PAGE.

. i

_-.a.m _ .

l l

TABLE 2 (CONTINUED)

NOTES: (1) Fluence values taken at \-thickness.

(2) C,USE values for 12/16/91 and EOL (Column 3) calculated per requirements outlined in Regulatory Guide 1.99, Revision 2, Paragraph C.1.2.

(3) Analyses that have demonstrated the required margin of safety for the Zion and Turkey Point plants at load level A and B conditions have been submitted to the NRC (Reports BAW-2118P and BAW-2148P).

The results of these two analyses are anticipated to bound the outcome of the R. E. Ginna analysis.

(4) C yOSE is calculated on the basis of RG1.99, Rev. 2, Position 1. SECY 91-333 states that this procedure is " inadequate." Recognizing.this and having provided for the uniqueness of Linde 80 welds in BAW-1803, Rev.1, the method for calculating C yUSE in BAW-1803, Rev.1, is put forward as more representative. and is intended to be used for predicting the behavior of these welds in licensing applications.

(5) BAW-1803 (6) CAW-10046P t

TABLE 3. GENERIC LETTER 92-01 RESPONSE: SECTION 2, ITEM b, 1 (1)

Subject:

10CFR50.61 and 10CFR50, Appendix G, Ill.A; Material Properties Related to PTS and Fracture Toughness Requirements Plant: R. E. Ginna Column 2 Column 3 Column 4 Column 5 C.6 Column 1 Beltline Unirradiated Charpy Test Results Unirrad. Unirrad. Method of Notes Dropweight Determng Materials RT[ ,

Col. 2b Col. 2c Col. 2d Test RT,,,

Col. 2a Results T,

Cy C C Cy y l

10 F 30ff-lb 50fl-lb 35 MLE ft-lb F F F FORGING 123PII8 30, average -3 +9 +4 +30 +30 NB-2331 (1,4) 1255255 45,121,112 -56 -48 -47 +20 +20 NB-2331 (1,2,4) 79,102,60, 97 125P666 105,115,112 -23 -5 -1 +40 +40 NB-2331 (1,2,4) 108,77,112 WELD SA-Il01 45,45,46 ND +70 ND -70 +10 N8-2331 (2,5,6)

SA-847 58,60,36 ND ND ND ND -5 Est. (3) (2,5)

'SA-848 54,56,59 ND ND ND ND -5 Est. (3) (2,5)

NOTES FOR TABLE 3 ARE ON THE FOLLOWING PAGE.

TABLE 3 (CONTINUED)

NOTES TO TABLE 3:

(1) Supplier test report data.

(2) BAW-2150 (3) BAW-1803, Revision 1, Tables 3-1 and 3-2; mean of RT., values for 34 Linde 80 welds.

(4) Values are for 30 hr stress

• elief.

(5) Cy (+10F) values are for 8 - 6 hr cycles stress relief.

(6) EPRI NP-373;yC 50 ft-lb, Drop Weight, and RT , values.

4 a

a 4

i

TABLE 4. GENERIC LETTER 92-01 RESPONSE: SECTION 2, ITEM b, 1 (2)

Subject:

10CFR50.61 and 10CFR50, Appendix G, III.A; Material Properties Related to PTS and Fracture Toughness Requirements -- APPLICABLE ONLY TO REACTOR VESSELS CONSTRUCTED TO AN ASME CODE EARLIER THAN THE 1971 EDITION, SUMMER 1972 ADDENDA Plant: R. E. Ginna Column 1 Column 2 Col. 3 Material Heat Treatment Notes BELTLINE.

MATERIALS 123Pil8 val 1550F-11h/WQ; 1220F-22h; ll25F-11h (min)/FC (1,2) 1255255 val 1550F-15%h/WQ: 1210F-18h/AC; ll25F-10%h (min)/FC 125P666 val 1550F-9h/WQ; 1220F-12h/AC; Il25F-10%h (min)/FC j SA-1101 Il25F-9h (min)/FC SA-847 ll25F-10%h (min)/FC SA-848 ll25F-9%h (min)/FC SURVEILLANCE ,

MATERIALS 125S255 val 1550F- 15%h/WQ; 1220F-18h/AC; 1100F-11%h/FC (1) 125P666 val 1550F-9h/WQ; 1220F-12h/AC; 1100F-ll%h/FC SA-1035 1100F-Il%h/FC NOTES:

(1) BAW-2150 l (2) Additional stress relief information per Mt. Vernon fabrication process sheets (3) WQ - water quench AC - air cool FC - furnace cool

.p.__

.m.. . - _ _ . _ . _ _ . _ . _ _ _ .... . . ..._.. _ _ _ _ . . . . . _ . _ _ _ . _ _ . _ _ . _ _ . _ . . _ _ . _ . _ . . . . _ . . . _ . . . _ . . _

L i

i TABLE 5. : GENERIC LETTER 92-01 RESPONSE: SECTION 2, ITEM b, 1 (3) i

Subject:

10CFR50.61 and 10CFR50, Appendix G, Ill.A; Material Properties Related to PTS and Fracture

. ' Toughness Requirements --. APPLICABLE ONLY TO REACTOR VESSELS CONSTRUCTED TO AN ASME CODE EARLIER THAN THE 1971 E0lTION, SUMMER 1972 ADDENDA

4 Plant: R. E. Ginna ,

Column 1 Column 2 Column 3 Column 4 Column 5 C. 6 l

1 ' Bel tline Heat Bel tl.ine . Weld Wire Weld Flux Notes  !

Plate or Number Weld Heat Lot-i Forging  !

NB Forging 123Pil8 NB to IS Circ.: SA-Il01 71249 8445 (1,2) }

IS Forging 125S255 IS to LS Circ.: SA-847 61782 8350 t LS Forging 125P666 LS to Dutch Circ.: SA-848 61782 8373 g ,

[

t

! NOTES: (1) .BAW-2150 (2) Mt. Vernon fabrication process sheets -!

(3) -NB - Nozzle Belt ,

l IS - Intermediate Shell <

LS - Lower..Shell 1

4 I

N

}

2 t

i i -

4 s e i 1 4 i i

TABLE 6. GENERIC LETTER 92-01 RESPONSE: SECTION 2, ITEM b, 1 (4)

Subject:

'10CFR50.61 and 10CFR50, Appendix G, III.A; Material Properties Related to PTS and Fracture Toughness' Requirements -- APPLICABLE ONLY TO REACTOR' VESSELS CONSTRUCTED TO AN ASME' CODE

EARLIER THAN THE-1971 EDITION, SUMMER 1972 ADDENDA Plant: R. E. Ginna

-Column l

~

Column 2 Column 3 Column 4 Column 5'-

~. Surveillance Surveillance Weld Wire Weld Flux Notes' Plate or, Weld Heat Lot Forging <

Heat Number 125S255 SA-1036 61782- 8436 (1)

'125P666 i

. , . I NOTES: . (1) BAW-2150'

h TABLE 7. GENERIC LETTER 92-01 RESPONSE: SECTION 2. ITEM b, 1 (5)-

[

Subject:

10CFR50.61 and 10CFR50, Appendix G, III.A; Material Properties Related to PTS and Fracture ,

i '.. Toughness Requirements -- APPLICABLE ONLY TO REACTOR VESSELS CONSTRUCTED TO AN ASME CODE  :

i. EARLIER THAN THE 1971 EDITION. SUMMER-1972 ADDENDA .

l

! P1 ant: R. E. Ginna  ;

- Column'l Column 2 C. 3 Material Chemical Composition, Weight Percent Notes C Mn'- P- S Si Cr Ni Mo' Cu BELTLINE

MATERIALS. .t

.123P118 val 0.20 0.64 0.010 0.008 0.23 0.41 0.68 0.60- ND (1)

1255255 Vale 0.18- 0. 66 __ 0.010 0.006 0.23 0.34' O.68 0.58- 0.07 (2,5) '

125P666 val 0;19 0.67- 0.010 0.011- 0.20. 0.37 0.68 0.57: 0.05 (2,5) i 'SA-1101 0.07 . l .28 . 0.021 0.014 0.52- 0.16' 0.60. 0.37' O.26 (3) "i SA-847- 0.08 1.34 0.012 0.012' O.45 0.08 0.54 0.38 0.25 (3)  !

0.25 j- SA-848 0.08 1. 4 4 -- 0.012 0.011 0.51 0.08 0.54- 0.38 (2) i l- SURVEILLANCE

! MATERIALS .

j 125S255VAll 0.18 0.66 0.010 0.007 0.23 0.33 0.69 0.58 0.07 (4) ,

! 125P666 val- 0.19 _0.67. 0.010 0.011 0.20 0.37 0.69 0.57 0.05 (4).  !

SA-1036 0.08 1.41 0.012 0.016 0.59 0.09 0.56 0.36 0.23- (4)

REQUIRED: State heat number of weld wires used for determining above chemical composition if different from i that in j (3). -- Not applicable --

i- ,

i NOTES; 1

[ (1) Supplier Material Test Report. ' '

i .(2) -BAW-2150 ,

j (3) BAW-2121P-j (4) BAW-1543,' Revision 1-  !

3 (5). Copper content based on surveillance material data.

i <

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3 0 - n1 s 4 i 2 t e 9 n e dd e r ei R m u g

ru E t eG T

i i d T m F iy E m sr L o e no C e ot C e S

(

ca I l P s su E n ag N o p

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. ti 8 1 w 1  : w E - ) oe L 8 a, hc B 8 , n A , ea T r T td e ( ar t t o t e sc e e a r c n L n u ,a o n t N c i a F n i G r 5i r e 2 e e . p 5) l n E m < n b e e g a  :

G . T sT c s R iR i e

g l c t e d , p n c  : L t E p e e t S a r j n d TV e b a l y t f u l o fC o e S P C I( N R l

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. .- - . . . , . . - _ . ... -......- ~., - ..- -- .- - . - - - . _ . . . _ - . . . . - . - - . . - . - . - - _ - . - . . . ~ . . ~ . _ . .

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TABLE 9. GENERIC LETTER 92-01 RESPONSE: SECTION 3, ITEM b'  ;

i

Subject:

. Generic' Letter 88-11 Response Commitments: Utilization of Surveillance i Results  ;

t

, Plant: R. E. Ginna- .;

I Were surveillance results used in determining C USE? Yes a- No / j 4

Were' surveillance results used in determining RT ,?' Yes / No o if any "yes" boxes were checked above, state how the surveillance results were used: i 1  !

! R. E. Ginna - Application for Amendment Docket 50-244.

L ,

a i t

y

-)

References:

Letter'to A. R. Johnson from R. C. Mecredy dated February 15, 1991.

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4 - _ . . . _ . . . _ . . _ . _ . . . -.

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. TABLE 10. GENERIC LETTER'92-01 RESPONSE: SECTION 3, ITEM c

Subject:

Generic Letter 88-11' Response Commitments; Difference Between Measured and Predicted (Regulatory Guide 1.99, Revision 2) Embrittlement Effects Plant: R. E. Ginna LQuestion-I. . Does measured ART., exceed ART , + 2a predicted by Regulatory Guide 1.99, Revision 2?

Question II. Does measured 'Cy USE ' drop exceed that obtained from Regulatory Guide' l.99, Revision 2,. Figure 2?

Column 2 Column 3 Column 4 Column 5 Column 6 Column 7 Column 1-Beltline Measured Predicted Ouestion I Measured Predicted Question II-Fluenge- CyOSE . CyOSE. If "yes" Material n/cm ART , ART ,+2a If "yes" see Note (3) Drop Drop see Note-(3)

(2)

ND' ND. --- ND ND ---

123P118 . .---

.6.53E+18- 73 No 0(1) 20 No 125S255- :0(1) No 78' No. .0(1) 23 1.02E+19 0(1)

'1.78E+19 0(1) .85 No 0(1)- 26 .No 6.53E+18 55 'No 23(1). 23 No 125P666 25(1)'

1.02E+19 25(1), 62 No '13(1)- 26 No 70 -No 40(1) 29 Yes 1.78E+19 30(1). No 195 No. 4(2) 21 SA-1101 7.01E+18 '-164(2) 220' No- 18(2) 24 No 1 23E+19

. 178(2)

---

ND. ND

-. ND ND --

SA-847 ND ND -- ND ND --

SA-848 - - =

NOTES'FOR'. TABLE 10 (IN-PARENTHESIS AB0VE) ARE ON'THE;FOLLOWING PAGE.

TABLE 10 (CONTINUED)

NOTES: (1) WCAP-10086

-(2) BAW-1803, Revision 1 (3) The only instance where a measured " shift" or " drop" exceeds that predicted by calculation performed in accordance with Regulatory Guide 1.99, Revision 2, is for a " drop" for base metal. The requirements of 10CFR50, Appendix G, were not violated. The use of " drop" data is '.only to indicate if beltline material has fallen below 50 ft-lb. Since this has not 4

occurred, the effect of these surveillance results are not significant.

4.

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. _ . _ . . _ . . _ . . m. . . _

i ll TABLE-1. GENERIC' LETTER 92-01 RESPONSE:tSECTION l'

Subject:

10CFR50, Appendix H: Adherence to RVSP Requirements Plant: Oconee Unit 1-

Question ~I

Does RVSP meet ASTM E 185-73, E 185-79, or E 185-82? Yes a No /

! Question II: Is plant one of the following? ANO-1, Crystal River-3, Davis Besse, R. E. Ginna,

! Oconee-1, Oconee-2, Oconee-3, Point Beach-1,' Point Beach-2, Rancho Seco, Surry-1,  ;

i Surry-2, Turkey Point-3,. Turkey Point-4, Zion-1, Zion-2. Yes / No o IF ANSWER IS "YES" TO EITHER QUESTION'I OR QUESTION II, PROCEED TO TABLE 2.

-IF ANSWER IS "N0" TO BOTH QUESTION I AND QUESTION II, PROCEED TO QUESTION'III AND QUESTION IV.

4 .

. Question III

If plan is to revise RVSP to meet requirements of 10CFR50, Appendix H,l when will

' revised RVSP be submitted to NRC? j j Response: j

LNot applicable. (see Question I and II above) 1 i

Question IV:. If plan is not to revise RVSP to meet requirements of 10CFR50, . Appendix H, when will i

exemption from 10CFR50, Appendix H be requested from NRC? -l

Response

1 L

Not applicable.(see Question I and II above) i l' NOTES: BAW-10006A, Revision 3: Surveillance Program Description  :(

(ASTM E 185-70) i t

t i

2 .

TABLE 2. GENERIC LETTER 92-01 RESPONSE: SECTION 2, ITEM a

Subject:

10CFR50, Appendix G, C,USE Requirements Plant: Oconee Unit 1-Column 1 Column 2 Column 3 Column 4 l Limiting Initial EFPY to reach If Column 2 is within license Action taken Material USE Cy OSE<50 ft-lb I period: Co0SE at indicated time per IV.A.1 ft-lb Column 3A Column 3B 12/16/91 E0L ,

LIMITING Analysis per 10CFR50, BELTLINE WELD Appendix G,Section V.C.3.

's scheduled for 1993 SA-1229 70 (5) 14, approx. 49 46 ender the sponsorship of B&WOG Reactor Vessel Wo.-king Group.

LIMITING BELTLINE PLATE OR FORGING

>32 NA NA NA C2197-2 91 (6)

NOTES FOR TABLE 2 ARE ON THE FOLLOWING PAGE.

. ..s .. . . . .

TABLE 2 (CONTINUED)

NOTES: (1) Fluence values taken at %-thickness.

(2) C USE y values for 12/16/91 and E0L (Column 3) calculated per requirements outlined in Regulatory Guide 1.99, Revision 2, Paragraph C.I.2.

(3) Analyses that have demonstrated the required margin of safety for the Zion and Turkey Point plants at load level A and B conditions have been submitted to the NRC (Reports BAW-2118P and BAW-2148P).

The results of these two analyses are anticipated to bound the outcome of the Oconee Unit I analysis.

(4) C yUSE is calculated on the basis of RGl.99, Rev. 2, Position 1. SECY 91-333 states that this procedure is " inadequate." Recognizing this and having provided for the uniqueness of Linde 80 welds in BAW-1803, Rev.1, the method for calculating C yUSE in BAW-1803, Rev.1, is put fonvard as more representative and is intended to be used for predicting the behavior of these welds in licensing applications.

f (5) BAW-1803 (6) BAW-10046P t

-,.-.s . ..-- . ... .. i. .. r - i . . .. . ..-. . . . -

..,...4.--. .

1,

.......-....i.i . . ..g i .

s. i . i . . ii . . -

. .. ....i .-- - - .... . .

TABLE 3. GENERIC LETTER 92-01 RESPONSE: SECTION 2, ITEM b, 1 (1)

Subject:

10CFR50.61 and 10CFR50, Appendix G, III.A; Material Properties Related to PTS and Fracture Toughness Requirements Plant: Oconee Unit 1 Column 1 Column 2 Column 3 Column 4 Column 5 C.5 Beltline Unirradiated Charpy Test Results Unirrad. Unirrad. Method of Notes Materials Dropweight RT , Determng Col. 2a Col. 2b Col. 2c Col. 2d Test f RT ,

Results Cy C C Cy T;

y 10 F 30fl-lb 50ff-lb 35 MLE ft-lb F F F FORGING AHR 54 87,54,112 ND ND ND ND +3 Est. (2) (1) 80,95,107 PLATE C2197-2 54,58,65 ND ND ND 5+10 +1 Est. (3) (1,4) 39,45,26 C3265-1 34,64,27 ND ND ND s+10 +1 Est. (3) (1,5) 37,65,63 C3278-1 35,29,53 ND ND ND $+10 +1 Est. (3) (1,5) 65,94,60 C2800-1 44,39,36 ND ND ND $+10 +1 Est. (3) (1,4) 36,39,39 C2800-2 ND ND ND ND 0 +1 Est. (3) (1)

WELD (1,7)

SA-ll35 56,44,55 ND ND ND ND -5 Est. (6)

SA-1229 55,45,40 ND ND ND ND -5 Est. (6) {l,7) 38,28,49 ND ND ND ND -5 Est. (6) (1,7)

WF-25 (1,8)

SA-1585 31,32,31 ND ND ND ND -5 Est. (6) 50,54,51 40,45,39 ND ND ND ND -5 Est. (6) (1,9)

SA-1073 (1,7)

SA-1493 41,35,40 ND ND ND 0 -5 Est. (6) 54,52,53 ND ND ND ND -5 Est. (6) (1,7)

SA-1430 (1,10)

SA-1426 46,31,36 ND ND ND ND -5 Est. (6) 35,45,45 NOTES TO TABLE 3:

(1) BAW 1820 (2) BAW-10046P, pp 3-17, -18; mean of most conservative value for each of 24 cases.

(3) BAW-10046P, pp 3-18; mean of most conservative value for each of 13 casts.

(4) Values are for 60 hr stress-relief.

; Values are for 40 hr stress-relief.

, L.) BAW-1803, Revision 1, Taoles 3-1 and 3-i'; mean of RT, values for 34 Linde 80 wel .s.

(7) Values are for.48 hr stress-relief.

(8) C y(+10F) test results from center and surface of test block; 80 hr stress-relief.

(9) C y(+10F) values are for 8.6 hr cycles stress relief.

(10) Cy (+10F) test results from center and surface of test block; 48 hr stress-relief.

l

l TABLE 4. GENERIC LETTER 92-01 RESPONSE: SECTION 2, ITEM b, T (2)

Subject:

10CFR50.61 and 10CFR50, Appendix G, III.A; Material Properties Related to PTS and Fracture Toughness Requirements -- APPLICABLE ONLY TO REACTOR VESSELS CONSTRUCTED TO AN ASME CODE EARLIER THAN THE 1971 EDITION. SUMt1ER 1972 ADDENDA Plant: Oconee Unit I Column 1 Column 2 Col. 3 Heat Treatment Notes Material BELTLINE MATERIALS 1600120F-Sh/WQ; 1250120F-15h/WQ; 1100-1150F-78%h/FC (cumul.) (1,2)

AHR 54 C2197-2 1600-1650F-9%h/BQ; 1200-1225F-9%h/BQ; 1100-Il50F-46%h/FC (cumul.)

03265-1 1600-1650F-9%h/BQ; 1200-1220F-9%h/BQ; 1100-1150F-50h/FC (cumul.)

C3278-1 1600-1650F-9%h/BQ; 1200-1225F-9%h/BQ; 1100-Il50F-50h/FC (cumul.)

C2800-1 1600-1650F-9)h/BQ; 1230-1225F-9%h/Bq: 1100-1150F-49h/FC (cumul.)

C2800-2 1600-1650F-9%h/BQ; 1200-1225F-9%h/Bw; 1100-1150F-49h/FC (cumul.)

SA-Il35 1100-Il50F-43%h/FC (cumul.)

SA-1229 Il00-Il50F-43%h/FC (cumul.)

WF-25 1100-1150F-43%h/FC (cumul.)

SA-1585 1100-1150F-48h/FC (cumul .)

WF-9 1100-Il50F-43%h/FC.(cumul.)

SA-1073 1100-Il50F-46%h/FC (cumul.)

SA-1493 Il00-ll50F-50h/FC (cumul.)

SA-1430 1100-Il50F-49h/FC (cumul.)

SA-1426 Il00-Il50F-49h/FC (cumul.)

SURVEILLANCE MATERIALS C3265-1 1600-1650F-9%h/BQ; 1200-1220F-9%h/BQ; 1100-Il50F-40h/FC (1)

C2800-2 1600-1650F-9%h/BQ; 1200-1225F-9%h/SQ; 1100-Il50F-40h/FC WF-ll2 1100-1150F-40h/FC NOTES FOR TABLE 4 ARE ON FOLLOWING PAGE.

TABLE 4 (CONTINUED)

NOTES:

(1) BAW-1820 (2) Additional stress relief information per Mt. Vernon process drawing.

(3) WQ - water quench BQ - brine quench FC - furnace cool l

TABLE 5. GENERIC LETTER 92-01 RESPONSE: SECTION 2, ITEM b, 1 (3) _

Subject:

10CFR50.61 and 10CFR50, Appendix G, III.A; Material Properties Related to PTS and Fracture Toughness Requirements -- APPLICABLE ONLY TO RE.^CTOR VESSELS CONSTRUCTED TO AN ASME C0Cr EARLIER THAN THE 1971 EDITION, SUMMER 1972 ADDENDA Plant: Oconee Unit 1 Column 1 Column 2 Column 3 Column 4 Column 5 C. 6 Beltline Heat Beltline Weld Wire Weld Flux Notes Plate or Number Weld Heat Lot Forging l

Lower NB Forging AHR 54, ZV 2861 NB to IS Circ.: SA-1135 51782 8457 (1)

'IS Plate C2197-2 IS to US Circ.(ID 61%): SA-1229 11249 8492 US Plate C3265-1 IS to US Circ.(OD 39%): WF-25 299L44 8650 US Plate C3278-1 US to LS Circ.: SA-1585 72445 8597 i LS Plate C2800-1 LS to Dutch Circ.: WF-9 72445 8632 LS Plate C2800-2 IS Longit.: SA-1073 IP0962 8445 US Longit.: SA-1493 8T1762 8578 LS Longit.: SA-1430 8T1762 8553 LS Longit.: SA-1426 8T1762 8553 NOTES: (1) BAW-1820 (2) NB - Nozzle Belt IS - Intermediate Shell US - Upper Shell LS - Lower Shell

. _ . . _ . _ _ _ . _ . _ _ _ . _ . . . _ _ . _ _ _ _ _ . . . _ . _ - . . - . . ~ _ . . _ . _ _ . . . . _ _ _ _ . _ . . . _ _ _ _ . . _ . . _ _ . _ . . . _ _ _ _ . ._. _.

3 i

i t

l' . . I

j. -TABLE-6. GENERIC LETTER 92-01 RESPONSE: SECTION 2, ITEM b,-1 (4) 1

Subject:

'L10CFR50.61 and 10CFR50, Appendix G, III.A; Material Properties Related to PTS and Fracture a Toughness Requirements'-- APPLICABLE ONLY=TO REACTOR VESSELS CONSTRUCTED TO AN ASME CODE EARLIER THAN THE 1971' EDITION. SLM ER 1972 ADDENDA t

Plant: Oconee Unit 1

! Column 1-~ Column 2 Column 3 Column 4 Column 5 e '

Surveillance'- Surveillance Weld Wire Weld Flux Notes

, Plate or' . Weld- Heat Lot

!- Forging n Heat' Number  !

I C3265-1 .WF-112 406L44' -8688 (1)

. -C2800-2 I

r l-1 i

t. NOTES: ~(1) BAW-1820 i:

J-l 4

4 i

i i

t

TABLE 7. GENERIC LETTER 92-01 RESPONSE: SECTION 2, ITEM b, 1 (5)

Subject:

10CFR50.61 and 10CFR50, Appendix G, III.A; Material Properties Related to PTS and Fracture Toughness Requirements -- APPLICABLE ONLY TO REACTOR VESSELS CONSTRUCTED TO AN ASME CODE E_ LIER THAN THE 1971 EDITION. SUMMER 1972 ADDENDA Plant: Oconee Unit 1 Column 2 C. 3 Column 1 Chemical Composition, Weight Percent Notes Material C Mn P S Si Cr Ni Mo Cu BELTLINE MATERIALS 0.18 0.64 0.006 0.010 0.29 0.31 0.65 0.57 0.16 (1)

AHR 54 C2197-2 0.21 1.28 0.008 0.010 0.17 ---

0.50 0.46 0.I5 (1)

C3265-1 0.21 1.42 0.015 0.015 0.23 0.17 0.50 0.49 0.10 (1) 0.19 1.'26 0.010 0.016 0.23 0.11 0.60 0.47 0.12 (1)

C3278-1 0.20 .1.40 0.012 0.017 0.20 0.13 0.63 0.50 0.11 (1)

C2800-1 C2800-2 0.20 1.40 0.012 0.017 0.20 0.13 0.63 0.50 0.11 (1) 0.10 1.38 0.025 0.017 0.51 0.11 0.64 0.43 0.21 (2)

SA-1073 0.08 1.45 0.011 0.013 0.49 0.08 0.54 0.38 0.25 (c,i SA-ll35 0.37 0.26 SA-1229 0.06 1.56 0.021 0.012 0.43 0.16 0.61 (2)

SA-1426 0.08 1.53 0.017 0.013 0.43 0.12 0.55 0.41 0.20 (2)

SA-1430 0.08 1.43 0.017 0.015 0.43 0.12 0.55 0.41 0.20 (2)

SA-1493 0.08 1.51 0.017 0.010 0.46 0.12 0.55 0.41 0.20 (2)

SA-1585 0.08 1.45 0.016 0.016 0.51 0.09 0.59 0.38 0.21 (2) 0.08 1.45 0.016 0.016 0.51 0.09 0.59 0.38 0.21 (1)

WF-9 WF-25 0.09 1.60 0.015 0.016 0.50 0.09 0.68 0.42 0.35 (2)

SURVEILLANCE MATERIALS C3265-1 0.21 1.42 0.015 0.015 0.23 0.17 0.50 0.49 0.10 (3) 0.20 1.40 0.012 0.017 0.20 0.13 0.63 0.50 0.11 (3)

C2800-2 0.08 1.47 0.016 0.015 0.54 0.07 0.59 l0.40 0.32 (3) l WF-ll2 REQUIRED:

State heat number of weld wires used for determining above chemical composition if different from that in 1 (3). - Not applicable -

TABLE 7. (CONTINUED)

NOTES:

(1) BAW-1820 (2) BAW-2121P (3) BAW-1543, Revision 3

__.______..m t

i- t i

• s- ,

! ' TABLE-8. ' GENERIC LETTER 92-01 RESPONSE: SECTION 3, ITEM'a f

Subject:

Generic Letter 88-11 Response Commitments: Effect of 'rradiation Temperature Plant: Oconee Unit 1

• i, .

Cold leg Temperature (T_,,): 556 f . . -e Figure 4-1) l 1 If T is <525 F, state how this was considered in determination of embrittlement effects ,.

i (C,U$U,N RT ) in accordance with Regulatory Guide 1.99, Revision 2:

j. Not applicable  ;

4

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i i

i i

References:

[

- None. f

t

i i

TABLE 9. GENERIC LETTER 92-01-RESPONSE: SECTION 3. ITEM b

Subject:

Generic Letter 88-11 Response Comnitments; Utilization of Surveillance Results Plant: Oconee Unit 1 Were' surveillance results used in determining C USE? Yes a No /

Were surveillance results used in determining RT ,? Yes / No a 1

If any *yes" boxes were checked above, state how the surveillance results were used: i 2

l per Regulatory Guide 1.99, Revision 2, Position 2, for prepartation of 1-Determination of RT , limit curves for WF-25 weld material only.

pressure-temperature t

3

Reference:

' BAW-2050 ,

i I  !

l i

f i

j.

i

~

4 4

V i l i

4

i.  ;

L b

TABLE 10. GENERIC' LETTER 92-01 RESPONSE: SECTION 3, ITEM c i

Subject:

Generic Letter 88-11 Response Commitments: Difference Between Measured and  !

l Predicted (Regulatory Guide 1.99, Revision 2) Embrittlement Effects j Plant: Oconee Unit 1 i

t i Question I. Does measured ART, exceed ART, + 2a predicted by Regulatory Guide 1.99, Revision 27 t j Question II. Does measured C,USE drop exceed that obtained from Regulatory Guide 1.99, l

! Revision 2, Figure 2? j i .

Column 1  ! olumn C 2 Column 3 Column 4 Column 5 Column 6 Column 7  !

l Beltline Fluenge. Measured Predicted Question I Neasured Predicted Question II  :

[' Material n/cm . ARr, ART.+2a If "yes" C,USE C,USE If "yes*  ;

! (3) see Note (4) . Drop Drop see Note (4) [

ND ND AHR 54 --- ND ND -- --

C2197-2 -- . ND ND --

ND ND --

l l '

. C3265-1 1.50E+18 15(1) 65 No 1(1) 13(1) No i 8.95E+18 34(1) 97 No 2(1) 20(1) No j, 9.8FE+18 39(1). 99 No 12(1) 21(1) No

C3278-1 -- . ND ND --

1ND ND --

l C2800-l' --- ND ND --

MD. ND --

[

C2800-2 SA-Il35-8.30E+17.

1.03E+19 18(2).

142(3) 57 240 No-No 0(2) 21(3) 13(2) 31(5)

No No

[

}- SA-1229 ---

ND ND --

ND ND --

l I WF-25 1.07E+18 124(6) 148 No 17(3) 24(6) No .!

F j 8.66E+18: .- 203(6) 261 - No 31(3) 34(6) No 7.79E+18 214(3) 263 No 25(3) 30(7) No 5.10E+18 148(3) 188 No 22(3) 24(7) No i

[ SA-1585..

j WF-9 e ---

ND: ;ND --

ND ND -- i ND ND ND --  !

i SA-1073 --- ND --

^

SA-1493- --- ND ND --

ND .ND --

SA-1430: ---

.ND ND --

ND MD --

ND ND ND -- l i- SA-1426- ---- ND --

I WTES FOR IMLE 10 ARE ON THE. FOLLOWING PAGE. }

r l

- - -. . _ - _ . __ - - - . . . . - - _. - _ _ - . - . _ _ - _ _ - . - . D

TABLE 10 (CONTINUED)

NOTES: (1) BAW-2050 (2) BAW-1421, Revision 1 (3) BAW-1803, Revision 1 (4) Statement not required.

(5) BAW-1920P (6) BAW-1901 (7) BAW-1910P

f TABLE 1. GENERIC LETTER 92-01 RESPONSE: SECTION 1  !

Subject:

10CFR50, Appendix H; Adherence to RVSP Requirements Plant: Oconee Unit 2 ,

Question I: Does RVSP meet ASTM E 185-73 E 185-79, or E 185-82? Yes a- No /

Question II: Is plant one of the following? ANO-1, Crystal River-3, Davis Besse, R. E. Ginna, Oconee-1, Oconee-2, Oconee-3, Point Beach-1, Point Beach-2, Rancho Seco, Surry-1, Surry-2, Turkey Point-3 Turkey Point-4. Zion-1, Zion-2. Yes / No a  !

IF ANSWER IS "YES" TO EITHER QUESTION I OR QUESTION II, PROCEED TO TABLE 2.

IF ANSWER IS "NO" TO BOTH QUESTION I AND QUESTION II, PROCEED TO QUESTION III AND QUESTION IV.

t Question III:. If plan is to revise RVSP to meet requirements of 10CFR50, Appendix H, when will l revised RYSP be submitted to NRC?  :

Response: [

f Not applicable (see Question I and II above).

Question'IV: If plan is not to revise RVSP to meet requirements of 10C050, Appendix H, when will I exemption from 10CFR50, Appendix H be requested from NRC?

Response

l Not applicable (see Question I and II above) I NOTES: BAW-10006A, Revision 3: _ Surveillance Program Description [

(ASTM E 185-70) l l

l e

f i

t

l TABLE 2. GENERIC LETTER 92-01 RESPONSE: SECTION 2. ITEM a- f

Subject:

'10CFR50, Appendix G, C.,USE Requirements Plant:- Oconee Unit 2 Column 1 Column 2 Column 3 Column 4 Limiting- Initial EFPY to reach If Column 2 is within license Action taken Material USE Cy USE<50 ft-lb period:~ C,USE at indicated time per IV.A.1 ft-lb Column 3A Column 3B 12/16/91 EOL LIMITING Analysis per 10CFR50, BELTLINE WELD Appendix G,Section V.C.3.

is scheduled for 1993 WF-25 70 (5) 4, approx. 46 43 under the sponsorship of B&WOG Reactor Vessel l

Working Group.

LIMITING.

B_El.TLINE ' PLATE OR FORGING-

>32- NA NA NA AMX 77 124 (6)

- NOTES FOR TABLE'2 ARE ON THE FOLLOWING PAGE.

4-h___..____._______,___._____.. b

TABLE 2 (CONTINUED)

NOTES: (1) Fluence values taken at %-thickness.

(2) CfUSE values for 12/16/91 and EOL (Column 3) calculated per requirements outlined in Regulatory Guide 1.99, Revision 2, Paragraph C.I.2.  ;

(3) Analyses that have demonstrated the required margin of safety for the Zinn and Turkey Point plants at load level A and B conditions have been submitted to the NRC (Reports BAW-2118P and BAW-2148P).

The results of these two analyses are anticipated to bound the outcome of the Oconee Unit 2 analysis.

(?) CyVSE is calculated on the basis of RGl.99, Rev. 2, Position 1. SECY 91-333 states that this procedure is " inadequate." Recognizing this and having provided for the uniqueness of Linde 80 '

welds in BAW-1803, Rev.1, the method for calculating C,USE in BAW-1803, Rev.1, is put forward as more representative and is intended to be used for predicting the behavior of these welds in licensing applications.

-(5) BAW-1803 (6) BAW-10046P

P TABLE 3. GENERIC LETTER 92-01 RESPONSE: SECTION 2, ITEM b. 1 (1)

Subject:

.10CFR50.61 and 10CFR50, Appendix G, III.A; Material Properties Related to PTS and Fracture Toughness Requirements Plant: Oconee Unit 2 Column I Column 2 Column 3 Column 4 Column 5 C.6 Beltline Unirradiated Charpy Test Results Unirrad. Unirrad. Method of Notes Materials Dropweight RT., Determng Col. 2a Col. 2b Col. 2c Col. 2d. Test F RT ,

Results C, C C, C, T,y 10 F 30fl-.lb 50 ft-lb 35 MLE ft-lb F- .F F FORGING AMX 77- 90,121,106 ND ND ND ND +3 Est. (2) (1,4) 103,91,128 AAW 163 ND -40 -10 -15 +20 +20 NB-2331 (1,5,7)

AWG 164 ND -75 -45 -50 +20 +20 NB-2331 (I,5,7)

WELD WF-154. 41,37,43~ ND ND ND T- -5 Est. (3) (1,6)

(1,6)

~

WF-25 38,28,49 ND. ND ND -5 Est. (3)

WF-112 35,40,30 ND ND ND ND -5 Est. (3) (1,6)

NOTES FOR TABLE 3 ARE ON THE FOLI.0 WING PAGE.

i

?

TABLE 3 (CONTINUED) j NOTES TO TABLE 3:  ;

(1) BAW-1820 t (2) BAW-10046P, pp 3-17, -18; mean of most conservative value for each of 24 cases.  !

(3) BAW-1803, Revision I. Tables 3-1 and 3-2; mean of RT,,, values for 34 Linde 80 welds. j (4) Values are for 60 hr stress-relief. ,

(5) Values are for 40 hr stress-relief. '

(6) Cy (+10F) values are for 48 hr stress-relief. t (7) Supplier test report data.

t i

l t

i I

4 V- uw w _^-- - -

- - - - - - - - - - . _ . - . . -.x-- ,,-

I TABLE'4. GENERIC LETTER 92-01 RESPONSE: SECTION 2, ITEM b. 1 (2)

] .

Subject:

10CFR50.61 and 10CFR50, Appendix G, III.A; Material Properties Related to PTS and Fracture  :

i Toughness Requirements -- APPLICABLE ONLY TO REACTOR VESSELS CONSTRUCTED TO AN ASME CODE ~

EARllER THAN THE 1971 EDITION, SUMMER 1972 ADDENDA

[

Plant: Oconee Unit 2  ;

' Column I Colu=n 2 Col. 3 I l-Material Heat Treatment Notes BELTLINE:-

, MATERIALS AMX'77 1580120F-7h/WQ; 1240i20F-14h/WQ; 1100-Il50F-53%h/FC (cumul.) (1,2) i AAW:163' 1590120F-4h/WQ; 1260120F10h/WQ: Il00-Il50F-41h/FC (cumul.)

• i AWG 164 1590 10F-4h/WQ: 1260i20F-10h/WQ; 1100-Il50F-4th/FC (cumul.)

WF-154 1100-Il50F-32%h/FC (cumul.)  !

WF-25 Il00-Il50F-41h/FC (cumul.) i j WF-112 Il00-Il50F-39%h/FC (cumul.)  ;

e  !

I' SURVEILLANCE i

!- MATERIALS  !

1 AAW:163 1590120F-4h/WQ; 1260120F-10h/WQ: Il00-Il50F-33h/FC (1)  !

j. AWG 164 1590120F-4h/WQ: 1260120F-10h/WQ; Il00-Il50F-33h/FC  !

WF-209-1 Il00-1150F-33h/FC j i

i NOTES-

, (1) BAW-1820 ._

[

Additional stress relief information per Mt. Vernon process drawing. i

~. : (2) f (3) WQ - water quench  !

j FC - furnace cool 3

( f l  !

I t

l I  !

i

- - . , . . - , . .s-~ ., _ _ _ _ . . _ . . . , . - . - _ _ _

l l

TABLE 5. GENERIC LETTER 92-01 RESPONSE: SECTION 2, ITEM b. 1 (3) t

Subject:

10CFR50.61 and 10CFR50, Appendix G, III.A; Material Properties Related to PTS and Fracture  !

Toughness Requirements -- APPLICABLE ONLY TO REACTOR VESSELS CONSTRUCTED TO AN ASME CODE  !

EARLIER THAN THE 1971 EDITION, SUMMER 1972 ADDENDA  !

i Plant: Oconee Unit 2  ;

Column 1 Column 2 Column 3 Column 4 Column 5 C. 6 Beltline Heat Beltline Weld Wire Weld Flux Notes ,

Plate or Number Weld Heat Lot [

Forging t Lower NB forging AMX 77, 123T382 NB to US Circ.: WF-154 406L44 8720 (1)  !

US Forging ' AAW 163, 3P2359 US to LS Circ.: WF-25 299L44 8650 [

j LS Forging AWG 164, 4PI885 LS to Dutch Circ.: WF-ll2 406L44 I 8688 l,

!; [

~

NOTES: (1) BAW-1820 I

.(2) NB - Nozzle Belt  !

j. US - Upper Shell LS - Lower Shell  !

e I

i t

4 i

i 1

p i

i i i

t

[

i l  !

TABLE 6. GENERIC LETTER 92-01 RESPONSE: SECTION 2, ITEM b. T (4)

Subject:

10CFR50.61 and 10CFR50, Appendix G, III.A; Material Properties Related to PTS and Fracture Toughness Requirements -- APPLICABLE ONLY TO REACTOR VESSELS CONSTRUCTED TO AN ASME CODE EARLIER THAN THE 1971 EDITION, SUMMER 1972 ADDENDA Plant:- Oconee Unit 2 Column 1 Column 2 Column 3 Column 4 Column 5 Surveillance Weld Wire Weld Fiux Notes Surveillance Plate or Weld Heat tot-Forging Heat Number AAW 163, 3P2359 WF-209-1 72105 8773 (1)

AWG 164, 4P1885 NOTES: (I) BAW-1820

TABLE 7. GENERIC LETTER 92-01 RESPONSE: SECTION 2, ITEM b, 1 (5)

Subject:

~10CFR50.61 and 10CFR50, Appendix G, III.A; Material Properties Related to PTS and Fracture Toughness Requirements -- APPLICABLE ONLY TO REACTOR VESSELS CONSTRUCTED TO AN ASME CODE EARLIER THAN THE 1971 EDITION, SUPMER 1972 A00ENDA Plant: Oconee Unit-2 Colnmn 1 Column 2 C. 3 Material Chemical Composition, Weight Percent Notes C Mn P S Si Cr Ni Mo Cu BELTLINE MATERIAL.S AMX 77 0.25 0.65 0.006 0.009 0.23 0.36' O.76 0.64 0.06 (1)

AAW 163' O.24 0.63 0.006- 0.012 0.25 0.36 0.75 0.62 0.04 (1)

AWG 164 0.21 0.62 0.010 0.010 0.23 0.39 0.80 0.58 0.02 (1)

WF-25 0.09 1.60 0.015 0.016 0.50 0.09 0.68 0.42 0.35 (2)

WF-112- 0.08 1.47 0.016 0.015 0.54 0.07 0.59 0.40 0.31 (2)

WF-154 0.07 1.54 0.013 0.016 0.42 0.07- 0.59 0.40 0.31 (2)

SURVEILLANCE MATERIALS-AAW 163. '0.24 0.63 0.006' O.012 0.25 0.36 0.75 0.62 0.04 (3)

AWG 164 0.21 0.62 0.010 0.010 0.23 0.39 0.80 0.58 0.02 (3)

WF-209-1 0.11 1.55 0.022' O.010 0.65 0.09 0.58 0.39 0.36 (3)

REQUIRED: State heat number of. weld wires used for cletermining above chemical composition if different from that in 1-(3). -- Not applicable --

NOTES:

(1) 'BAW-1820 (2) BAW-2121P

-(3) BAW-1543, Revision 3

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TABLE.10. GENERIC-LETTER 92-01 RESPONSE: SECTION 3. ITEM c

Subject:

Generic Letter 88-11 Response Commitments; Difference Between Measured and Predicted (Regulatory Guide 1.99, Revision 2) Embrittlement Effects Plant: Oconee Unit 2 Question I. Does measured RA.T , exceed ART , + Zu predicted by Regulatory Guide 1.99, Revision 2?

Question II. Does measured C,USE drop exceed that.obtained from Regulatory Guide 1.99, Revision 2, figure 2?

Column 1 Column 2- Column 3 Column 4 Column 5 Column 6 Column 7 Beltline I Flue e Measured Predicted Question I Measured Predicted Question II Material ART , . ART ,+2a If "yes* C,USE C,USE If "yes" n/c Drop (2) see Note (3) Drop see Note (3)

AMX 77 ---

ND ND -- ND . ND --

AAW 163 1.02E+18 0(1). 22 No 12(1) 11(1) Yes 3.37E+18 0(1) 36 No --

15(1) --

55 No 19(1) 20(1) No 1.21E+19 0(1)

AWG 164 --- ND ND -- ND ND --

WF-154. --- ND ND -- ND ND --

1.07E+18 148 No 17(2) 24(4) No WF-25 124(2) 8.66E+18 261 No 31(2) '34(4) No 203(2) 7.79E+18 263 No 25(2) 30(4) No 214(2)

WF-Il2 1.50E+18 -78(2) 157 No '9(2) '19(5) No 8.95E+18 250 No 12(2) 27(5) No 191(2) 9.86E+18 256 No ~12(2) 27(5) No 185(2) 8.21E+18 204(2) 246 No 29(2) 33(6) No NOTES FOR TABLE 10 ARE ON THE FOLLOWING PAGE.

TABLE 10 (CONTINUED)

NOTES: (1) BAW-2051 (2) BAW-1803, Revision 1 (3) The only instance where a measured " shift" or " drop" exceeds that predicted by calculation performed in accordance with Regulatory Guide 1.99, Revision 2, is for a "drcp" for base metal and that was by I ft-lb and is not considered to be significant. At a higher fluence, the predicted " drop" for that material did not exceed the measured value. The requirements of 10CFR50, Appendix G, were not violated, and there being no further application of the

" drop" data, the effect of these surveillance results are therefore not significant.

(4) BAW-1901

_(5) BAW-2050 (6) BAW-1920P i

~ " ~ - --

-- - _ , . . ~ , . . , . - _ _ , --

I-i I

-TABLE I. ' GENERIC LETTER 92-01 RESPONSEE SECTION 1 j

Subject:

10CFR50, Appendix H: Adherence to RVSP Requirements

Plant

Oconee Unit 3 Question I: Does RVSP meet ASTM E.185-73, E 185-79, or E 185-82? Yes a No /

'Is plant'one of the following? ANO-1, Crystal River-3, Davis Besse, R. E. Ginna, Question II: .

j Oconee-1, Oconee-2, Oconee-3, Point Beach-1, Point Beach-2, Rancho Seco, Surry-1,

Surry-2, Turkey Point-3 Turkey Point-4, Zion-I, Zion-2. Yes / No o i IF ANSWER IS "YES* TO EITHER QUESTION I OR' QUESTION II, PROCEED TO TABLE 2.

I IF ANSWER IS "N0"'TO BOTH QUESTION I AND QUESTION II,' PROCEED TO QUESTION III AND QUESTION IV.

i

Question III: .' If plan is to revise RVSP to meet requirements of 10CFR50, Appendix H, when will l revised RVSP be submitted to NRC?

Response

j Not applicable (see Que, tion I and II above) i.

Question IV: If plan is not to revise RVSP to meet requirements of 10CFR50, Appendix H, when will exemption from 10CFR50, Appendix H be requested from NRC?

Response: .

Not applicable (see Question I and II above) i i

NOTES: ' -BAW-10006A, Revision 3: Surveillance Program Description

! (ASTM E.185-70) 5 i

d 4

1

TABLE'2. -GENERIC' LETTER 92-01 RESPONSE:ESECTION 2,' ITEM a

Subject:

10CFR50, Appendix G, C,,USE Requirements Plant: Oconee Unit 3 Column 1 Column 2 Column 3 Column 4 Limiting Initial. EFPY to reach If Column 2 is within license Action taken Material USE C,USE<50 ft-lb period: C USE at indicated time per IV.A.1

~

Column 3A Column 38 12/16/91' EOL LIMITING Analysis per 10CFR50, BELTLINE WELO Appendix G,.Section V.C.3.

is scheduled for 1993 47 under the sponsorship of l WF-67 70 (5) 17, approx. 50 B&WOG Reactor Vessel Working Group.

LIMITING BELTLINE PLATE OR FORGING

>32 NA NA NA AWS 192 90 (6)-

NOTES FOR TABLE 2.ARE ON THE FOLLOWING PAGE.

TABLE 2 (CONTINUED)

NOTES: (1) Fluence values taken at %-thickness.

(2) CyVSE values for 12/16/91 and EOL (Column 3) calculated per requirements outlined in Regulatory Guide 1.99, Revision 2, Paragraph C.I.2.

(3) Analyses that have demonstrated the required margin of safety for the Zion and Turkey Point plants at load level A and B conditions have been submitted to the NRC (Reports BAW-2118P and BAW-2148P).

The results of these two analyses are anticipated to bound the outcome of the Oconee Unit 3 analysis.

(4) CyOSE is calculated on the basis of RGl.99, Rev. 2, Position 1. SECY 91-333 states that this procedure is " inadequate." Recognizing this and having provided for the uniqueness of Linde 80 welds in BAW-1803, Rev.1, the method for calculating C,USE in BAW-1803, Rev.1, is put forward as more representative and is intended to be used for predicting the behavior o'f these welds in licensing applications.

(5) BAW-1803 )

(6) BAW-1820 (

i

______-- \

TABLE 3. GENERIC LETTER 92-01 RESPONSE: SECTION 2. ITEM b. T (1)

Subject:

10CFR50.61 and 10CFR50, Appendix G, III.A; Material Properties Related to PTS and Fracture Tougnness Requirements Plant: Oconee Unit 3 Column 2 Column 3 Column 4 Colu=n 5 C.6 Column 1 Unirradiated Charpy Test Results Unirrad. Unirrad. Method of Notes Beltline Dropweight RT , Determng Materials f Col. 2b Col. 2c Col. 2d Test RT ,

Col. 2a Results T

10 F 30 f -lb 50 f -lb 35 LE ft-lb F F F .

FORGING 117,111,109 ND ND ND ND +3 Est. (2) (1,4) 4680 113,49,101 ND -55 -30 -40 +40 +40 NB-2331 (1,5,8)

AWS 192 ND -2 +20 -3 +40 +40 NB-2331 (1,5,8)

ANK 191 WELD (1,6)

WF-200 36,25,26 ND ND ND ND -5 Est. (3) 29,35,30 ND ND ND ND -5 Est. (3) (1,6)

WF-67 WF-70 39,35,44 ND ND ND ND +18 Eval.(7) (1,6,9) 42,29,46 ND ND ND ND -5 Est. (3) (1,6)

WF-169-1 NOTES FOR TABLE 3 ARE ON TiiE F9LLOWING PAGE.

TABLE 3 (CONTINUED)

NOTES TO TABLE 3:

(1) BAW-1820 (2) BAW-10046P, pp 3-17 -18; mean of most conservative value for each of 24 cases..

(3) BAW-1803, Revision 1, Tables 3-1 and 3-2; mean of RT, values for 34 Linde 80 welds.

(4) Cy (+10F) values are for 60 hr stress-relief.

(5) Values are for 25 hr stress-relief.

(6) C. (+10F) values are for 48 hr stress-relief.

(7) BkW-2100 (8) Supplier test report data.

(9) RT, value are for 40 hr stress-relief maximum.

4

- m w

^

i t

TABLE 4. GENERIC LETTER 92-01 RESPONSE: SECTION 2. ITEM b. 1 (2)

Subject:

10CFR50.61 and 10CFR50, Appendix G, III.A; Material Properties Related te PTS and Fracture Toughness Requirements -- APPLICABLE ONLY TO REACTOR VESSELS CONSTRUCTED TO AN ASME CODE l EARLIER THAN THE 1971 EDITION. SUPMER 1972 ADDENDA Plant: Oconee Unit 3  ;

Column 1 Column 2 Col. I !

Material Heat Treatment Notes I BELTLINE ~  !

1 MATERIALS  !

i. 4680 1675F-7h/WQ; 1220F-15h/AC; 1100-Il50F-26)h/FC (cumul.) (1,2) l l AWS 192 1590120F-4h/WQ; 1240120F-10h/WQ; 1100-Il50F-29%h/FC (cumul.)  ;

l~ ANK 191 1590120F-4h/WQ; 1250120F-10h/WQ; 1100-Il50F-29%h/FC (cumul.) [

WF-200 1100-Il50F-25h/FC (cumul.) t

- WF-67 1100-Il50F-29%h/FC-(cumul.). 5

WF-70 1100-Il50F-29%h/FC

(cumul.)

, WF-169-1 Il00-1150F-284h/FC~(cumul.) ,

SURVEILLANCE i MATERIALS ,

ANK 191 1590120F-4h/WQ; 1250120F-10h/WQ; 1100-Il50F-30h/FC (1)

AWS 192 1590120F-4h/WQ; 1240120F-10h/WQ; Il00-Il50F-30h/FC WF-209-1 Il00-Il50F-30h/FC

NOTES: .

t (1) BAW-1820 ,

, (2) . Additional stress relief information per Mt. .Vernon process drawing. i (3) WQ - water quench ,

F AC - air cool -

i FC - furnace cool i-4

\

TABLE 5. GENERIC LETTER 92-01 RESPONSE: SECTION 2, ITEM b, 1 (3)

Subject:

10CFR50.61 and 10CFR50, Appendix G, III.A; Material Properties Related to PTS and Fracture Toughness Requirements -- APPLICABLE ONLY TO REACTOR VESSELS CONSTRUCTED TO AN ASME CODE EARLIER THAN THE 1971 EDITION. SUPMER 1972 ADDENDA Plant: Oconee Unit 3 Column 2 Column 3 Column 4 Column 5 C. 6 Column 1 Beltline Weld Wire Weld Flux Notes Beltline Heat Heat tot Plate or Number Weld Forging NB to US Circ.: WF-200 821T44 8773 (1)

Lower NB Forging 4680 AWS 192, 522314 US to LS Circ.(ID 75%): WF-67 72442 8669 US Forging ANK 191, 522194 US to LS Circ.(00 25%): WF-70 72105 8669 LS Forging 8754 LS to Dutch Circ.: WF-169-I STI554

- NOTES: (1) BAW-1820 (2) NB - Nozzle Belt ,l US - Upper Shell LS - Lower Shell l

)

r

i I ,

I I

L I

r

+_  !

TABLE 6. GENERIC LETTER 92-01 RESPONSE: SECTION 2 ITEM b. 1 (4)

Subject:

10CFR50.61 and 10CFR50, Appendix G, III.A; Material Properties Related to PTS and Fracture  ;

Toughness Requirements -- APPLICABLE ONLY TO REACTOR VESSELS CONSTRUCTED TO AN ASME CODE  !

EARLIER THAN THE 1971 EDITION, SUMMER 1972 ADOENDA

-l f

Plant:- Oconee Unit 3  !

' Column 1 Column 2 Column 3 . Column 4 Column 5 Surveillance Surveillance Weld Wire Weld Flux Notes  !

Plate or Weld Heat tot

• Forging Heat Number  !

r

! ANK 191, 522194 WF-209-1 72105 8773 (1) l AWS 192, 522314 i I

l l

l i

I

NOTES

(1) BAW-1820 I

i

! I k

.f L

l'

.- - - . . _ - - - - - - - - - - . _ . . - - _ _ . = . - _ - - . _ - -- . _

I l

l ~

TABLI 7. GENERIC LETTER 92-01 RESPONSE: SECTION 2, ITEM'b,-1 (5) l l- .

Subject:

10CFR50.61 and 10CFR50, Appendix G, III.A; Material Properties Related to PTS and Fracture Toughness Requirements -- APPLICABLE ONLY TO REACTOR VESSELS CONSTRUCTED TO AN ASME CODE i

! EARLIER THAN THE 1971 EDITION SUPMER 1972 ADDENDA Plant: Oconee Unit 3'  !

1 i

Column 1 Column 2 C. 3

i. Material Chemical Composition, Weight Percent Notes C Mn P S Si Cr Ni Mo Cu

BELTLINE MATERIALS  !

4680 0.21 0.67 0.009 0.012 0.22 0.36 0.91 0.56 ---

(1)

AWS 192 0.21 0.58 0.011 0.015 0.24 0.30 0.73 0.60 0.01 (1) I j ANK 191 0.24 0.72 0.014 0.012 0.21 0.34 0.76 0.62 0.02 (1)

WF-67 0.08 1.55 0.021 0.016 0.58 0.09 0.60 0.39 0.24 (2)

WF-70 0.09 1.63 0.018 0.009 0.54 0.10 0.59 0.40 0.35 (2) ,

WF-169-1 0.08 1.56 0.016 0.016 0.45 0.08 0.63 0.37 0.18 (2) t WF-200 0.07 1.60 0.010 0.015 0.48 0.14 0.63 0.40 0.24 (2)

SURVEILLANCE i MATERIALS

.. ANK 191 0.24 0.72 0.014 0.012 0.21 0.34 0.76 0.62 0.02 (3) ,

i AWSil92 0.21 0.58 0.011 0.015 0.24 0.30 0.73 0.60 0.01 (3)

WF-209-1 0.08 1.63 0.017 0.012 0.61 0.10 0.58 0.39 0.30 (3)

, REQUIRED: . State. heat number of weld wires used for determining above chemical composition if different from 't j that in 1 (3). --.Not applicable --

2 NOTES.

j (1) BAW-1820

! (2) BAW-2121P  ;

] (3) BAW-1543, Revision 3 i

i. .

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t ,

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R e9 E 1 d9 1 -

4 n1 0 s i 2 t: e e 9 . r dd e u ei R t m g ru E i i eG T F d T m iy E m e sr L o e no C S ot C e

( ca I l R s su E n ag N op F we E s R G 6 s e 5 ih R 5 ht

. ti 1

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i a E t l S a z a a l i

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e Y r S Y E t u R U s 1

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- t S t 2 n e U, T m w 9 m C R o t h R i g g E n n e T e n i i t T o n n a E C i i t L s m m C e r r I s e e ,

R n t t e E o e e v N

p d d o E s b G e n n a R i i d

1 d d e 9 1 e e k

- s s c E 8 u u e L 8 h B s s c A r t t T e 3 l l e t u u r t t s s e e i e e w L n r r U s cs e e e e it e c c x n rl e n n o o eu n a a b N ns o l l ee c l l GR O i i "s e e e  :

v v y e

r r " c t u u n c  : s s y e e t n r j

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a e

r r e a f e

u l e e f e S P W W I R

_ f1 1L(l l

TABLE 10. GENERIC LETTER 92-01 RESPONSE: SECTION 3, ITEM c .

Subject:

Generic Letter 88-11 Response Comitments; Difference Between Measured and Predicted (Recziatory Guide 1.99, Revision 2) Embrittlement Effects Plant: Oconee Unit 3 Question 1. Does measured ART., exceed ARTer + 20 predicted by Regulatory Guide 1.99, Revision 2?

Question 11. Does measured yC USE drop exceed that obtained from Regulatory Guide 1.99, I l Revision 2, Figure 2? _

Column 1 Column 2 Column 3 Column 4 Column 5 Colv 6 Column 7 Beltline Measured Predicted Question I Measured Prr.dic ed Question II l Fluenge If "yes" CyUSE CyUSE If "yes" Material n/cm ART , ART r+2a (2) see Note (3) Drop Dron see Note (3)

ND ND --

ND ND --

4680 ---

AWS 192 8.10E+17 63(1) 16 Yes 12(1) 4(i) Yes 28 No 13(1) 9(1) Yes 3.12E+18 19(1) 44 Yes 19(1) 12(1) Yes 1.45E+19 45(1) 16 No 14(1) 8(1) Yes ANY, 191 8.10E+17 9(1) Yes 3.12E+18 32(1) 28 Yes 21(1) 13(1) 44 No _19(1) 17(1) Yes 1.45E+19 31(1)

ND ND -- ND ND --

WF-200 ---

200 No 15(2) 23(a) No WF-67 6.09E+18 160(2) 259 No 13(2) 22(5) No WF-70 6.63E+18 135(2)

ND ND -- ND ND --

WF-169-1 --- ,

NOTES FOR TABLE 10 ARE ON THE FOLLOWING PAGE.

TABLE 10 (CONTINUED)

NOTES: (1) BAW-2128 (2) BAW-1803, Revision 1 (3) (a) The only instances where a measured " drop" exceeds that ,'redicted by calculation performed in accordance with Regulatory Guide 1.99, Revision 2, is for base metal. The requirements of 10CFR50, Appendix G, were not violated.

(b) The only instances where a measured " shift" exceeds that predicted by calculation performed in accordance with Regulatory Guide 1.99, Revision 2, is for base metal.

For for irradiated to 8.1x10'ging nyt. AWS 192, the first such finding was for materialThis 'was not observed for Since the measured shift of the naterial irradiated to a higher neutron fluence did not exceed the predicted value, it is safe to conclude that the conservativeness of the Regulatory Guide method was not compromised. The second such finding, for material ' irradiated to 1.4x10", showed a difference of one degree F and is not taken as significant.

For forging ANK 191, there was one finding, for material irrap'iated to 3.1x10'8 nyt. This was not observed for material irradiated to 1.4x10 nyt. Since the measured shift of the material irradiated to a higher neutron fluence did not exceed the predicted value, it is safe to conclude that the conservativeness of the Regulatory Guide method was not compromised.

(c). As noted above, the " drop" data did not violate regulatory requirements, and there is no further application of the " drop" data. In the instances where the measured " shift" values exceeded the predicted values, it was shown above that the conservativeness of the Regulatory Guide was not compromised. It is concluded,. therefore, that the effect of these surveillance results is not significant.

(4) BAW-1910P (5) BAW-1920P

_m_ _____-____..-____--_________._m_____

TABLE 1. GENERIC LETTER 92-0  : SE8 ' 24 1

. a--..~.

Subject:

'10CFR50, Appendix H; Adherence to RVSP Requir,'

Plant: Point Beach Unit 1 .~.- , ,w a Question I: Does RVSP meet ASTM E 185-73, E 185-79, or t ... Yes o No /

Question II: Is plant one of the following? ANO-1, Crystal River-3, Davis Besse, R. E. Ginna, Oconee-1, Oconee-2, Oconee-3, Point Beach-1, Point Beach-2, Rancho Seco, Surry-1, Surry-2, Turkey Point-3, Turkey Point-4, Zion-1, Zion-2. Yes / No o IF ANSWER IS "YES" TO EITHER QUESTION I OR QUESTION II, PROCEED TO TABLE 2.

IF ANSWER IS "N0" TO BOTH QUESTION I AND QUESTION II, PROCEED TO QUESTION III AND QUESTION IV.

Queetion III: If plan is to revise RVSP to meet requirements of 10CFR50, Appendix H, when will revised RVSP be submitted to NRC?

Response

Not applicable (see Question I and II above) 4 Question IV: If plan is not to revise RVSP to meet requirements of 10CFR50, Appendix H, when will exemption from 10CFR50, Appendix H be requested from NRC? .

Response

Not applicable (see Question I and II above)

NOTES: WCAP-7513: Surveillance Program Description (ASTM E 185-66)

)

-TABLE 2. GENERIC LETTER 92-01 RESPONSE: SECTION 2, ITEM a

Subject:

10CFR50, Appendix G. C,USE Requirements Plant: Point Beach Unit 1

-Column 1 Column 2 Column 3' Column 4 Initial- EFPY to reach If Column 2 is within license Action taken Limitin3 per IV.A.1-Material USE Cy USE<50 ft-lb ' period: C,USE at indicated time ft lb Column 3A Column 38 12/16/91 EOL LIMITING Analysis per 10CFR50, i

. BELTLINE WELD Appendix G,Section V.C.3.

is scheduled for 1993

'SA-II01 70 (5) 5, approx. 42- 39 under the sponsorship of B&WOG Reactor Vessel Working Group.

LIMITING BELTLINE PLATE

'OR' FORGING

>32 .NA NA NA A9811-1 91 (6)- <

NOTES FOR. TABLE 2 ARE ON THE FOLLOWING PAGE.-

l :.

..__ _ ____._. _ _ . 1 . m .. . _ _ _ _ _ .

_ _ _ . .._..._..n _._ ..._ . . _ .

, TABLE 2 (CONTINUED)

NOTES: (1) Fluence values taken at %-thickness.

(2) C VSE y values for 12/16/91 and E0L (Column 3)' calculated per requirements outlined in Regulatory Guide 1.99,. Revision 2, Paragraph C.I.2.

(3) Analyses that have demonstrated the required margin of safety for the Zion and Turkey Point plants at load level A and B conditions have been submitted to the NRC (Reports BAW-2118P and BAW-2148P).

The results of these two analyses are anticipated to bound the outcome of the Point Beach Unit .1 analysis.

(4) C yOSE is calculated on the basis of RGl.99, Rev. 2, Position 1. SECY 91-333 states that this procedure is " inadequate." Recognizing this and having provided for the uniqueness of Linde 80 welds in BAW-1803, Rev.1, the method for calculating C vuSE in BAW-1803, Rev.1, is put forward as more representative and is intended to be used for predicting the behavior of these welds in licensing applications.

t (5) BAW-1803 4

(6) BAW-10046P 1

i.

i I

[-  !

LTABLE 3. GENERIC LETTER 92-01 RESPONSE: SECTION 2,. ITEM b 1 (1) ,

Subject:

'10CFR50.61 and 100FR50, Appendix G, III.A; Material Properties Related to PTS and Fracture Toughness Requirements  :

Plant: Point Beach Unit I Column 1 Column 2 Column 3 Column 4 Column 5 C.6 Beltline Unirradiated Charpy Test Results Unirrad. Unirrad. Method of Notes  !

Materials Dropweight RT gg Determng Col. 2a Col. 2b Col. 2c Col. 2d Test F RT ,g Results T,

y i-C* ( C 50fE-lb C

35 MLE i 10 F 30 ft-lb-ft-lb- F F F FORGING ,

122P237 63,66,67 -22 +4 +15 +50 +50 NB-2331 (1,5) 62,95,42  :

t PLATE .

(1,2,6).

A9811-1 50,41,51 -4. +24 ND -30 +1 Est. (3) .;

C14?3-1 103,51,88 -38 -16 ND- -20 +1 Est. (3) (1,2,6)  !

WELD SA-1426- 35,45,45 ND ND ND ND -5 Est. (4)

(2,7,9) 46,31,36 .

i SA-1101 45,45,46 ND +70 ND -70' +10 53-2331 (2,8)

SA-812 43,40,36l ND- ND ND ND -5 Est. (4) (2,7)

SA-775- 48,45,44 ND ND ND ND -5 Est. (4) (2,7)

SA-847 58,60,36' ND ND 'ND 'ND -5 lEst. (4) (2,7)

NOTES FOR' TABLE 3'ARE ON THE FOLLOWING PAGE.

1 N

w

TABLE 3 (CONTINUED)

NOTES TO TABLE 3:

(1) Supplier test report data.

(2) BAW-2150 (3) . BAW-10046P, pp 3-18; mean of most conservative value for each of 13 cases.

(4) BAW-1803, Revision 1, Tables 3-1 and 3-2; mean of RT,g values for 34 Linde 80 welds.

(5) Values.are for 30 hr stress-relief.

(6) Values are' for 50 hr stress-relief.

(7)' Cy (+10F) values are for 8 - 6 hr cycles stress relief.

(8) EPRI NP-373; yC 50 ft-lb, Drop Weight, and RT ,n values.

(9) Cy (+10F) test results from center and surface of test block.

?

b

TABLE 4. GENERIC LETTER 92-01~ RESPONSE: SECTION 2, ITEM b, 1 (2)>

Subject:

-10CFR50.61 and 10CFR50, Appendix G, Ill.A; Material Properties Related to PTS and Fracture

. Toughness Requirements -- APPLICABLE ONLY TO REACTOR VESSELS CONSTRUCTED TO AN ASME CODE-EARLIER THAN THE 1971 EDITION. SUMMER 1972 ADDENDA Plant: Point'8each Unit I Column't Column 2 Col. 3 Material Heat Treatment. Notes BELTLINE MATERIALS 122P237 val 1550F-Ilh/WQ; 1220F-22h/FC; 1125F-10%h/FC (1,2,3) j l A9811-1 1625-1675F-lh/in (min)/WQ; 1200-1250F-lh/in (min)/AC; l 1100-1150F-10%h (min)/FC

. C1423-1 1625-1675F-lh/in (min)/WQ; 1200-1250F-lh/in (min)/AC; 1

1100-1150F-10%h (min)/FC SA-1426 Il25F-9h-(min)/FC SA-Il01. 1125F-10%.(min)/FC SA-812/SA-775 1125F-10%h (min)/FC SA-847 1125F-10%h (min)/FC SURVEILLANCE MATERIALS A9811-1 1650F-7h/WQ;-1225F-7h/AC; 1125F-Il%h/FC (1)

C1423-1 1650F-7h/WQ; 1225F-7h/AC; ll25F-10)h/FC l

SA-1263 1125F-11%h/FC NOTES'FOR TABLE 4 ARE ON FOLLOWING PAGE.

2-_ _ _ _ _ _ _ _ . _. . _ _ _ _ _ _ _^ . _ - -4e-_ui _ x-. _M:

TABLE 4. (CONTINUED)

NOTES:

(1). BAW-2150 (2) ' Supplier Material Test Report (3) Additional stress relief information per Mt. Vernon fabrication process sheets.

(4) WQ - water quench AC - air cool FC - furnace cool i

! l i b TABLE 5. GENERIC LETTER 92-01 RESPONSE: SECTION 2, ITEM b, j (3) i

Subject:

10CFR50.61 and 10CFR50, Appendi:: G, III.A; Material Properties Related to PTS and Fracture ,

Toughness Requirements -- APPLICABLE ONLY TO REACTOR VESSELS CONSTRUCTED TO AN ASME CODE EARLIER THAN THE 1971 EDITION, SUMMER 1972 ADDENDA Plant: Point Beach Unit 1 Column 1 Column 2 Column 3 Column 4 Column 5 C. 6

.< Beltline Heat Beltline Weld Wire. Weld Flux Notes Plate or Number Weld- Heat tot ,

Forging i NB Forging 122P237 NB to IS Circ.::SA-1426 8T1762 8553 (1,2)

IS Plate A9811-1 IS to LS Circ.

'SA-1101 71249 8445 LS Plate' Cl423-1 IS Longit.(ID 27%): SA-812 IP0815 8350' IS Longit.(OD 73%): SA-775 IP0661 8304 LS Longit.: SA-847 61782 8350 NOTES': .(1) BAW-2150 (2) Mt. Vernon fabrication process sheets (3) ' NB - Nozzle Belt i IS - Intermediate Shell j LS.--Lower Shell

\;

,! 'f

)

i

_ _ _ _ _ _ _ _ _ _ _ _ - _. - _. _ _ _ _ _ _ _ _ _ _ _ _ . _ . _ s . .

. TABLE 6. GENERIC LETTER 92-01 RESPONSE: SECTION 2, ITEM b,'T (4)

Subject:

10CFR50.61 and 10CFR50, Appendix G, III.A; Material Properties Related to PTS and Fracture Toughness Requirements -- APPLICABLE ONLY TO REACTOR VESSELS CONSTRUCTED TO AN ASME CODE EARLIER-THAN THE 7971 EDITION, SUMMER 1972 ADDENDA:

F Plant: Point Beach Unit I Column I Column 2 Column 3 Column 4 Column 5 Surveillance Weld Wire Weld' Flux Notes Surveillance-Plate or ' Weld- Heat Lot Forging Heat Number A9811-1 .SA-1263 72445 8504 (1)

C1423-1 NOTES: (1) BAW-2150 t,

I -

I

. ~ . . . _ _ . . . - . - - . . . . . . . . . . . . . . . . . . . . . , _ . _ - -

l TABLE 7. GENERIC LETTER 92-01 RESPONSE: SECTION 2, ITEM b, 1 (5) f

Subject:

10CFR50.61 and 10CFR50, Appendix G, III.A; Material Properties Related to PTS and Fracture Toughness Requirements -- APPLICABLE ONLY TO REACTOR VESSELS CONSTRUCTED TO AN ASME CODE EARLIER THAN THE 1971 EDITION, SUMMER 1972 ADDENDA Plant: Point Beach Unit 1 Column 1 Column 2 C. 3 Material Chemical Composition, Weight Percent Notes C Mn P S Si Cr Ni Mo Cu 3

BELTLINE MATERIALS 122P237 val 0.21 0.65 0.010 0.008 0.22 0.33 0.82 0.62 0.15(7) (1)

A9811-1 0.20 1.42 0.010 0.020 0.25 ND 0.056(6) 0.49 0.20 (2,5)

C1423-1 0.20 1.36 0.016 0.020 0.25 ND 0.065(6) 0.46 0.12 (2,5)

SA-1426 0.08 1.53 0.017 0.013 0.43 0.12 0.55 0.41 0.20 (3)

SA-812 0.08 1.54 0.017 0.015 0.40 0.07 0.52 0.38 0.17 (3)

SA-775 0.08 1.52 0.024 0.019 0.46 0.06 0.63 0.45 0.19 (3)

SA-1101 0.07 1.28 0.021 0.014 0.52 0.16 0.60 0.37 0.26 (3)

SA-847 0.08 1.34 0.012 0.012 0.45 0.08 0.54 0.38 0.25 (3)

SURVEILLANCE MATERIALS A9811-1 0.19 1.42 0.010 0.020 0.25 ND 0.056(6) 0.48 0.20 (4)

C1423-1 0.21 1.37 0.014 0.019 0.25 ND 0.065(6) 0.46 0.12 (4)

SA-1263 0.09 1.47 0.019 0.024 0.49 0.13 0.57 0.39 0.22 (4)

REQUIRED: State heat number of weld wires used for determining above chemical composition if different from that in 1 (3).

-- Not applicable --

NOTES FOR TABLE 7 ARE ON THE FOLLOWING PAGE.

- - - .. _ _ _ _ = _ _ _ _ _ _ _ _ _ ._ _ _ ___________ ____ __- _ _ _ _ _ - --

i TABLE 7. (CONTINUED)

NOTES:

(1) Supplier Material Test Report (2) BAW-2150 (3) -BAW-2121P (4) BAW-1543, Revision 3 (5) Copper and nickel contents based on surveillance material data.-

,(6) These values are suspect; verification of this information is planned.

(7) Estimated value based on review of similar materials.

4 d

O e i

a a

I

1 4

TABLE 8. GENERIC LETTER 92-01 RESPONSE:fSECTION 3, ITEM a

Subject:

Generic-Letter 88-11 Response Commitments: Effect of Irradiation Temperature Plant: Point Beach Unit'l Cold Leg Temperature (Tm,a): 542 F (See Figure 4-4) q is <525 F, state how this was considered in determination of embrittlement effects If T'(,I RT,,,) in accordance with Regulatory Guide 1.99, Revision .2:

(C,0$

During the time span from approximately December 1,1979 to October 1,1983, the plant. operated at approximately.80% power and a reduced system average temperature. This operation produced a cold leg temperature of approximately 511.F. This temperature was not considered in determination of embrittlement effects since the surveillance capsule results spanning this interval did not exhibit any significant variation from the expected values. See Section 4.

I 1

References:

--- - - ' - ' ' ' - - -v- - -

. - _ _ . . . _ _ 2. -

TABLE 9. GENERIC LETTER 92-01 RESPONSE: SECTION 3, ITEM b

Subject:

Generic letter 88-11 Response Commitments; Utilization of Surveillance Results Plant: Point Beach Unit 1 Were surveillance results used in determining C VSE?

o Yes a No /

Were surveillance results used in determining RT,,? Yes a No /

If any "yes" boxes were checked above, state how the surveillance results were used:

References:

None

v. 2 Tf.BLE 10. G- '. LETTER 92-01 RESPONSE: SECTION 3 ITEM c

Subject:

Generic letter 88-11 Response C': anitments; Difference Between Measured and Predicted (Regulatory Guide 1.99, Revision 2) Embrittlement Effects Plant: Point Beach Unit 1 Question 1. Does measured ART,o, exceed ART., + 20 predicted by Regulatory Guide 1.99, Revision 2?

' Question II. Does measured yC USE drop exceed that obtained from Regulatory Guide 1.99, Revision 2, Figure 2?

l Column 1 Column 2 Column 3 Column 4 Column 5 Column 6 Column 7 Beltline Measured Predicted Ouestion I Measured Predicted Ouestion Il Fluenge If "yes" CyOSE CyOSE If "yes" Material n/cm ART,o, ART,o,+2a (2) see Note (3) Drop Drop see Note (3)

ND ND --

ND ND --

122P237 -- -

6.20E+18 90(1) 110 No 18(1) 28 No A9811-1 115 No 15(1) 29 No 7.58E+18 90(1) 2.10E+19 140 No 7(1) 37 No 105(1) 140 No 12(1) 37 No 2.11E+19 100(1) 82 No 0(1) 22 No C1423-1 6.20E+18. 50(1) 85 No 0(1) 23 No 7.58E+18 50(1)

'2.10E+19 100 No 0(1) 30 No 50(1) 2.llE+19 101 No 0(1) 30 No 50(1)

ND ND --

ND ND --

SA-1426 ---

195 No 4(2) 21 No SA-1101 7.01E+18 164(2) 220 No 18(2) 24 No 1.23E+19 178(2)

ND ND --

ND ND --

SA-812 ---

ND ND -- ND ND --

SA-775 ---

ND ND -- ND ND --

SA-847 ---

NOTES FOR TABLE 10 ARE ON THE FOLLOWING PAGE.

TABLE 10 (CONTINUED) 4 NOTES: (1) WCAP-10736 (2). BAW-1803, Revision 1 (3) Statement not required.

J I

-. . . - . . _ _ = _ . - - - _ . . _ _ _ _ .

TABLE 1. GENERIC LETTER 92-01 RESPONSE: SECTION 1

Subject:

10CFR50, Appendix H; Adherence to RVSP Requirements Plant: Point Reach Unit 2 Does RVSP meet ASTM E 185-73, E 185-79, or E 185-82? Yes a No /

Question I:

Question II: Is plant one of the following? AND-1, Crystal River-3, Davis Besse, R. E. Ginna, Oconee-1, Oconee-2, Oconee-3, Point Beach-1, Point Beach-2, Rancho Seco, Surry-1, Surry-2, Turkey Point-3, Turkey Point-4, Zion-1, Zion-2. Yes / No o IF ANSWER IS "YES" TO EITHER QUESTION I OR QUESTION II, PROCEED TO TABLE 2.

IF ANSWER IS "N0" TO BOTH QUESTION I AND QUESTION II, PROCEE0 TO QUESTION III AND QUESTION IV.

Question III: If plan is to revise RVSP to meet requirements of 10CFR50, Ap9endix H, when will revised RVSP be submitted to NRC?

Response

Not applicable (see Question I and II above)

Question IV: If plaa is not to revise RVSP to meet requirements of 10CFR50, Appendix H, when will exemption from 10CFR50, Appendix H be requested from NRC?

Response

Not applicable (see Question I and II above)

NOTES: WCAP-7712: Surveillance Program Descripti,n (ASTM E 185-66)

TABLE'2. GENERIC LETTER 92-01 RESPONSE: SECTION 2, ITEM a

Subject:

10CFR50, Appendix .G,- Cy USE Requirements Plant: Point Beach Unit 2 Column 1 ' Column 2 Column 3 Column 4 i Limiting Initial EFPY to reach If Column 2-is within license Action taken

-Material- USE Cy VSE<50 ft-lb _a'riod: CsVSE at indicated time per IV.A.1 ft-lb Column 3A Column 38 12/16/91 EOL LIMITING Analysis per 10CFR50,

. BELTLINE WELD Appendix G,Section V.C.3. *

!' is scheduled for 1993 - '

SA-1484 70 (5) 5, approx. 43 40 under the sponsorship of B&WOG Reactor Vessel-Working Group.

LIMITING

• t BELTLINE PLATE OR FORGING' ,

123V500 124 (6) >32 NA NA- NA HOTES FOR TABLE 2 ARE ON THE FOLLOWING PAGE. -

1-r

-k

_ - . _ - _ _ _ . , - _ . - . - . _ _ - - .__=. - - --..a. _ _ - _ _ t--- , e u v- ,, .p

TABLE 2 (CONTINUED)

NOTES: (1) Fluence values taken at \-thickness.

(2) C OSE y values for 12/16/91 and E0L (Column 3) calculated per requirements outlined in Regulatory Guide 1.99, Revision 2, Paragraph C.I.2.

(3) Analyses that have demonstrated the required margin of safety for the Zion and Turkey Point plants at load level A and B conditions have been submitted to the NRC (Reports BAW-2118P and BAW-2148P).

The results of these two analyses are anticipated to bound the outcome of the Point Beach Uni? ?

analysis.

(4) C yOSE is calculated on the basis of RGl.99, Rev. 2, Position 1. SECY 91-333 states that this procedure is " inadequate." Recognizing this and having provided for the uniqueness of Linde 80 welds in BAW-1803, Rev.1, the method for calculating C yUSE in BAW-1803, Rev.1, is put forward as ,

more representative and is intended to be used for predicting the behavior of these welds in licensing applications.

(5) BAW-1803 (6) BAW-10046P ,

3 i

b

'I

_ - - - - - - - - - - - _ _ _ _ _ _ r- p

' TABLE 3. GENERIC LETTER 92-01 RESPONS'E: SECTION 2,. ITEM b, 1 (1)

Subject:

10CFR50.61 and 10CFR50, Appendix G, III.A; Material Properties Related to PTS and Fracture Toughness Requirements Plant: Point Beach Unit 2 Column 1 Column 2 Column 3 l Column 4 Colume 5 C.6

- Unirradiated Charpy Test Results Unirrad. Unirrad. Method of- Notes Beltline. Determng Materials Dropwt. RT.,

Col. 2a ~ Col. 2b Col. 2c Col. 2d Test F RT.,

Results C C C, Ty,- 1 Cy L 10 F 30f[-lb 59f[-lb 35 MLE ft-lb F F F l

FORGING '

123V352. 75,58,61' ' 27

+15 +25 +40 +40' NB-2331- (2,4) 60,55,62-123V500 111,80.81 -50 -25 -57 +40 +40 NB-2331 (1,2,4) 111,108,118 123W195- 38,49,46 -30 -7 -10 +40' +40 NB-2331 (1,2,4) 66,54,75,78-WELD CE Weld- 'Not avail. Not avail Not" avail Not avail. Not: avail -56 10CFR50.61 SA-1484 40,52,41 ;ND ND ND ND ,5 -

'Est. (3) (1)

NOTES FOR TABLE 3 ARE ON THE FOLLOWING PAGE.

m_________.__m__ _ __ ...i.....__m g

TABLE 3 (CONTINUED)  ;

NOTES TO TABLE 3: i (1) BAW-2150 f (2) Supplier Test Reports '

(3) BAW-1803, Revision 1, Tables 3-1 and 3-2; mean of RT, values for 34 Linde 80 welds.

(4) Values are for 30 hr stress relief.

h l

4 4

e

F i

i

~TABLEL4. ' GENERIC LETTER 92-01 RESPONSE: SECTION 2 ITEM b, 1 (2) r'

Subject:

10CFR50.61 and 10CFR50, Appendix G, III.A; Material Properties Related to PTS and Fracture Toughness Requirements -- APPLICABLE ONLY TO REACTOR VESSELS CONSTRUCTED T0 AN SME CODE' 3}

EARLIER THAN'THE 1971 EDITION, SUMMER 1972 ADDENDA ,

i Plant: Point Beach Unit 2  !

Column 1 Column'2- Col. 3

• i Material Heat Treatment - Notes ,

I

' t BELTLINE MATFRIALS .

] 123V352 val' -1550F-ll%h/WQ; 1220F-12h/FC; 1125F-%h (min)/FC (1,2) l j 123V500VA1 1550F-9%h/WQ; 1200F-12h/AC; 1125F-%h (min)/FC 122W195 val' 1550F-8h/WQ; :-1200F-12h/AC; ll25F-%h (min)/FC l

CE Weld Not'available- 1 SA-1484 Il25F-%h (min)/FC SURVEILLANCE-  !

MATERIALS ,;

s ,

! 123V500 val '

1550F-9)h/WQ; 1200F-12h/AC; ll25F-12h/FC (1) i j 122W195 val' 1550F-8h/WQ; 1200F-12h/AC; ll25F-12h/FC WF-193: ll25F-Il%h/FC: j :

l NOTES: ';

j

! (1) BAW-2150 . . .. .

(2) Additional stress relief information per Mt. Vernon 'abrication process -sheets. l

-(3) WQ - water-quench .  !

AC - air' cool

. FC - furnace cool i

i 3

r-' - - -

-,,, b

- _ - _ - . , _ _ - - a

n . .

TABLE 5. GENERIC LETTER 92-01 RESPONSE: SECTION 2, ITEM b,.1 (3)

Subject:

10CFR50.61 and 10CFR50, Appendix G, III.A; Material Properties Related t'o PTS and Fracture Toughness Requirements -- APPLICABLE ONLY TO REACTOR VESSELS CONSTRUCTED TO AN ASME CODE EARLIER THAN THE 1971 EDITION SUMMER 1972 ADDENDA Plant: Point Beach Unit 2 Column 2 Column 3 Column 4 Column 5 C. 6 Column 1 Beltline Heat Beltline Weld Wire Weld Flux Notes Number. Weld Heat tot

. Plate or.

Forging 123V352- NB to IS Circ.: CE Weld Not avail. Not avail. (1,2)

NB-Forging IS to LS Circ.: SA-1484 72442 8579 IS Forging 123V500' LS Forging 122W195 3 NOTES: ..(1) BAW-2150:

(2) Mt. Vernon fabrication process sheets (3) NB - Nozzle Belt IS - Intermediate Shell LS - Lower Shell

i l

i f

+

h.

TABLE 6. GENERIC LETTER 92-01 RESPONSE: SECTION:2, ITEM b, 1 (4) -[;

i

Subject:

.10CFR50.61 and-10CFR50, Appendix G,'III.A; Material Properties Related to PTS and Fracture ,

. . Toughness Requirements -- APPLICABLE ONLY TO REACTOR VESSELS CONSTRUCTED TO AN ASME CODE i EARLIER THAN THE 1971 EDITION. SUMMER 1972 ADDENDA i Plant: Point Beach-Unit 2 Column 1 Column 2 Column 3 Column 4 Column 5

Surveillance- Surveillance ' Weld Wire' ..W eld Flux Notes' 'f

. Plate or' Weld Heat' Lot -;

Forgir.g -

Heat Number

• o.

. 123V500 :WF-193- 406L44 8773' (I) "

122W195 i

f i

4 :-

. NOTES': '.(1) BAW-2150' 4

.l 3

TABLE 7. GENERIC LETTER 92-01 RESPONSE:' SECTION'2, ITEM b,1 (5) i

Subject:

-10CFR50.61 and 10CFR50, Appendix G, III.A; Material Properties Related to PTS and Fracture' .i Toughness Requirements - APPLICABLE ONLY TO REACTOR VESSELS CONSTRUCTED TO AN ASME COGE i o

EARLIER THAN THE 1971 EDITION. SUMMER 1972 ADDENDA t

Plant: Point Beach Unit 2 Column'l- Column 2 C. 3  ;

Material Chemical Composition, Weight Percent Note i

C Mn P S Si Cr Ni Mo Cu i

BELTLINE MATERIALS ,

123V352 val' O.20 0.68 0.010: 0.010 0.24 0.34 0.73L 0.59 0.15(3)' (1) 123V500 val 0.18 0.66' O.010 0.008 0.24 0.34 0.70 0.57 0.09 (4,7) 122W195 val 0.24 0.58 0.010 0.008 0.22 0.33 0.72 0;57- 0.05- (4,7) "

4 NB to IS NA NA NA NA NA NA 0.90- NA. 0.27 *2,41 SA-1484 0.08 1.52 0.018- 0.015 0.42 0.09- 0.60' O.39 0.24 (5) l r

SURVEILLANCE-MATERIALS 0.009 0.009 0.24 0.35- 0.59 0.09 123V500 val 0.20 !0.65 0.71_ (6) 122W195 val 0.22 0.59. 0.010' O.008 0.23 0.33 0.70 0.60 0.05 (6).  !

WF-193' O.08 1.40' O.014 0.013- 0.55- 0.07 0.59 0.39 0.25 (6)

REQUIRED: State heat number of weld wires used for determining above chemical composition if different from . t

-- Not applicable --

'that-in 1 (3).

NOTES FOR TABLE-7 ARE ON THE FOLLOWING PAGE.

i I

_. 2-._ - --- *

,.BLE 7. (CONTINUED)

NOTES:

'(1) Supplier Material Test Report (2) Nozzle belt-to-intermediate .shell weld was fabricated by Combustion Enginnering; information is not available (NA).

l '(3) Estimated value based on review of similar materials. ,

'- ( 4 ) BAW-2150 l '(5) BAW-2121P (6) BAW-1543, Revision 3 j l

l (7) . Copper content based on surveillance material data.

l l

- - - - - = _ . _ _ _ . _ _ _ . . _ . ~ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _

s t

c e

f e f r e u

t t a n a r e ep m

M e E m l T e t I T t i

, n r 3 o b' i

m' N t a

e O i f I

T d o:

C a 2 E r n S r on I io

ti E f as S o ni N iv O t c me rR P e S f e

t ,

E f R e9 E ) d5 1 4

- n1 0 s 4 i-2 t n

e 9 e dd e r ei R m u ru E t g eG T

i i d T m F iy E m sr L o e no C e ot C e S

(

ca I l R s su E 'n ag N o p

F weR E s G 2 s e 4 ih R 5 ht

. ti 8 1 w 1  : r E 8 2

) oe L ,a h c B 8 , n A r t , ea T

i T td e n ( ar t U t o t e sc e e h r c n L c u ,a o a t F N .

c e a n i B r 5i r e 2' e e t p 5) . l n n m < m, b

e i e a  :

G o T sT c s P iR i e

g l c t e ,. p n c  : L [5 p e e t a r j n d T 0y e b a l 't f u l o fC o e S P C I( N R p

a h

m -

w TABLE,9. GENERIC LETTER 92-01 RESPONSE: SECTION 3 ITEM b

Subject:

Generic Letter s -11 Response Commitments; Utilization of Surveillance Results Plant: Point Beach Unit 2

' Were surveillance results used in,d?[ormining CLUSE? Yes a No /

Were surveillance results used in determining RT,,? Yes a No /

If.any "yes" boxes were checked above, state how the surveillance results were used:

I i

References:

None

i i 4

i i

j;  ; TABLE 10. GENERIC LETTER 92-01 RESPONSE: SECTION 3. ITEM c l

i-

[,

I-

Subject:

' eneric G letter 88-11 Response Conunitments; Difference Between Measured and

-i Predicted (Regulatory Guide 1.99, Revision 2) Embrittlement Effects i Plant: Point Beach Unit 2 1,

L Question I. Does measured ART, exceed ART, + 2a predicted by Regulatory Guide 1.99, i Revision 2?

] '

i i Question II. Does measured C,USE drop exceed that obtained from Regulatory Guide 1.99, Revision 2, Figure 2?

! Column 1 Column 2 Column 3 Column 4 Column 5 Column 6 Column 7

. Beltline Fluenge Measured Predicted Question 1 Measured Predicted Question II ,

[ Material n/cm ART , ART ,+2a If *yes" C,USE C,USE If "yes* l l' (3) see Note (2) Drop Drop see Note (2) i

123V352. ---

ND ND --

ND ND --

l s

123V500 6.14E+18, 30(1) 84 No 0 29(1) No j 1

^"

8.36E+18 . 30(1) 89 No 0 31(1) No  !

2.15E+19 . 70(1). 104 No .0 38(1) No i

{' 3.47E+19 76(1) Ill No 17 41(1) No

, 122W195 6.14E+18 10(1) 54 No 10' 17(1) No  ;

8.36E+18 59 17(1) No 0 19(1) No i

,'t 2.15E+19- ' 35(1) 71 No 5 23(1) No i 3.47E+19 47(1) 75 No 11 26(1) No  :

i CE Weld ---

ND ND --

ND ND --

SA-1484~ ---

ND ND --

ND ND --

l l, CE Weld- ---

l ND ND --

ND ND --

t

i. ~

4 NOTES FOR TABLE'10 ARE ON THE FOLLOWING PAGE.  !

S [

I 3

i 1 i i:  :

r - u -.. w ,-m-v ..

-e-- p , -en.. nnv v . ,. ,,.-.r...,.,-..m-e

_ .wm-- masom----se-,gALem_a, ,,A---,a-- ,,.a, ,4W,..m,n__y-,a_-, , _ u _-L x-~n-a ,-~,--W~e, -,e,L--,

l 1

l 1

l 1

l I

b b

w emas M

E O

V w

O.e W

J CO CC t-m

'3 C C7 O Ww b .M M >

0@

C CC

+b e OCM eWO

• E CO N W*=*

s M e 3 m3 .

• T 4 4 iQMM mmm MNM www ee W

t==

O Z

. -, .. __ , _ _ . . . ..___ , ._,__. . - . . ~ , _.. -., - .---, . . ,.- .---. -._.-

TABLE I. GENERIC LETTER 92-01 RESPONSE: SECTION 1

Subject:

10CFR50, Appendix H: Adherence to RYSP Requirements Plant: Surry Unit 1 Does RVSP meet ASTM E 185-73, E 185-79, or E 185-82? Yes a No /

Question I:

Question II: Is plant one of the following? AND-I, Crystal River-3, Davis Besse, R. E. Sinna, Oconee-1, Oconee-2, Oconee-3, Point Beach-1, Point Beach-2, Rancho Seco, Surry-1, Surry-2, Turkey Point- T, Turkey Point-4, Zion-1, Zion-2. Yes / No o IF ANSWER IS 'YES" TO EITHER QUESTION I OR QUESTION II, PROCEED TO TABLE 2..

IF ANSWER IS "M0" TO BOTH QUESTION I AND QUESTION II, PROCEED TO QUESTION III AND QUESTION IV.

Question III:

J If plan is to revise RYSP to meet requirements of 10CFRSO, Appendix H, when will revised RVSP be submitted to NRC? .

l

Response

Not applicable (see Question I and II above)

Question IV: If plan is not to revise RYSP to meet requirements of 10CFR50, Appendix H, whte will exemption from 10CFR50, Appendix H be requested from NRC?

Response:-

Not aoplicable (see Question I and II above) ,

NOTES: WCAP-7723: Surveillance Program Description (ASTM E 185-66)

L

TABLE 2. GENERIC LETTER 92-01' RESPONSE: SECTION 2. ITEM a j

Subject:

10CFR50, Appendix G, C USE Reovirements Plant: 'Surry Unit I l

! Column 1 Column 2 Column 3 Column 4 ,

t

)3 - Limiting Initial :EFPY to reach If Column 2 is within license Action taken  :

1 Material USE Cy USE<50 ft-lb period: Cf.SI at indicated time per IV.A.1 j j ft-lb Column 3A' Column 38  !

12/16/91 EOL . l f- LIMITING 2 Analysis per 10CFR50, f BELTLINE WELD Appendix G,Section V.C.3.  !

is scheduled for 1993 l SA-1585 70 (6) 5, approx. 44(2) 39(2)- under the sponsorship of  !

j 54(3) 51(3). B&WOG Reactor Vessel  !

I Working Group. L l LIMITING i

BELTLINE PLATE  !

- OR FORGING

C4326-1 91 (7) >32 NA NA NA ,

. C4326-2 91-(7)- >?2 NA NA NA l C4415-1 91 (7) >32 NA NA NA  ;

C4415-2 91 (7) >32 NA NA NA l

NOTES FOR TABLE.2 ARE ON THE FOLLOWING PAGE.  !

t  ;

+' f

TABLE 2 (CONTINUED)

NOTES: (1) Fluence values taken at %-thickness.

(2) C,USE values for 12/16/91 and EOL (Column 3) calcolated per requirements outlined in Regulatory Guide 1.99, Revision 2, Paragraph C.I.2.

(3) C,USE values for 12/16/91 and EOL (Column 3) calculated per requirements outlined in Regulatory Guide 1.99, Revision 2, Paragraph C.2.2. (See also letter to U. S. Nuclear Regulatory Commission from W. L. Stewart dated December 1, 1989.

Title:

" Virginia Electric and Power Company, Surry Unit I and 2: Response to Request for Additional Information, Upper-Shelf Energy of Reactor Vessel Materials." Docket Nos. 50-280 and 50-281, License Nos. DPR-32 and DPR-37)

(4) Analyses that have demo'.istrated the required manjin of safety for the Zion and Turkey Point plants at load level A and B conditions have been submitted to the NRC (Reports BAW-2118P and BAW-2148P).

The results of these two analyses are anticipated to bound the outcome of the Surry Unit I analysis.

(5) C,USE is calculated on the basis of RGl.99, Rev. 2, Position 1. SECY 91-333 states that this precedure is " inadequate." Recognizing this and having provided for the uniqueness of Linde 80 welds in BAW-1803, Rev.1, the spethod for calculating C,USE in BAW-1803, Rev.1, is put forward as more representative and is intended to be used for predicting the behavior of these welds in licensing applications.

(6) BAW-1803 (7) BAW-10046P

TABLE 3. GENERIC LETTER 92-01 RESPONSE: SECTION 2, ITEM b. T (1) .-

Subject:

'10CFR50.61 and 10CFR50, Appendix G, Ill.A; Material Properties Related to PTS and Fracture Toughness Requirements Plant: Surry Unit 1 Column 2 Column 3 Column 4 Column 5 C.6 Column 1 Untrradiated Charpy Test Results Unirrad. Unirrad. Method of Notes Beltline Dropwt. RT , Detencng Materials Test f RT.,

Col. 2a Col. 2b Col. 2c Col. 2d Results T,

( C C ( y 10 F 30f[-lb 50ff-lb 35 MLE ft-lb F F F FORGING 48,55,41 -35 0 -15 +40 +40 NS-2331 (3,6) l 122V109 l

42,62,47 PLATE NB-2331 (1,2,7) 28,39,47 +5 +40 +37 +10 +IC C4326-1 NB-2331 (1,2,7) 58,34,60 -5 +25 +20 0 0 C4326-2 NB-2331 (1,2,7) 40,32,34 +10 +45 +35 +20 +20 C4415-1 NB-2331 (1,2,7) 50,55,46 0 +20 +17 0 0 C4415-2 WELD (1,8)

J726 54,77,51 ND ND ND ND 0 Est. (3) 31,32,31 ND ND ND ND -5 Est. (5) (1,9,10)

SA-1585 50,54,5!

48,40,40 ND ND ND ND -5 Est. (5) (1,11)

SA-1650 (1,11)

SA-1494 54,25,44 ND ND ND NG -5 Est. (5) 33,33,33 ND ND ND ND -5 Est. (5) (1,11)

SA-1526 NOTES FOR TAB'E 3 ARE ON THE FOLLOWING PAGE.

TABLE 3 (CONTINUED)

NOTES TO TABLE 3:

(1) BAW-2150 (2) Supplier Test Report.

(3) BAW-1909, Revision 1 (4) BAW-10046P, pp 3-17, 18; mean of most conservative value for each of 24 cases.

(5) BAW-1803, Revision 1, Tables 3-1 and 3-2; mean of RT (6) Values from BWNS Drawing 02-IIS7427E-00, Sheet 2 of I., values for 34 Linde 80 welds.

(7) Valces are for 60 hr stress relief.

(8) Values are for 30 hr stress relief.

(9) C (+10F) y values; specimens from center and surface of test block.

(10) Cy (+10F) values are for 80 hr stress relief.

(11) Cy (+10F) values are for 48 hr stress relief.

i r

TABLE 4. GENERIC LETTER 92-01 RESPONSE: SECTION 2. ITEM b 1 (2)

Subject:

10CFR50.61 and 10CFR50, Appendix G, III.A; Material Properties Related to PTS and Fracture Toughness Requirements -- APPLICABLE ONLY TO REACTOR VESSELS CONSTRUCTED TO AN ASME CODE EARLIER THAN THE 1971 EDITION, SUMMER 1972 ADDENDA Plant: Surry Unit 1 ,

Column 1 Column 2 Col. 3 Material Heat Treatment Notes r BELTLINE MATERIALS 122V109 val 1550F-Ilh/WQ; 1200F-22h/AC; ll25F-40h/FC (1)

C4326-1 1675125F-9h/WQ; 1210F-9h/AC; ll25125F-60h/FC C4326-2 1675125F-9h/WQ; 1210F-9h/AC; ll25125F-60h/FC C4415-1 1675125F-9h/WQ; 1200-1225F-9h/AC; ll25125F-60h/FC ,

C4415-2 1675125F-9h/WQ; 1200-1225F-9h/AC; ll25125F-60h/FC t J726 Il30F-30h/FC -

SA-1585 ll25125F-80h/FC SA-1650 1125125F-48h/FC SA-1494 ll25125F-48h/FC SA-1526 ll25125F AQb'f(,

SURVEILLANCE MATERIALS C4326-1 1650-1700F-9h/WQ; 1210F-9h/AC; II25F-15%h/FC (1)

C4415-1 1650-1700F-9h/WQ; 1200F-9h/AC; ll25F-15%h/FC SA-1526 Il25F-15%h/FC NOTES:

(1) BAW-1909, Revision 1 (2) 'WQ - water quench AC - air cool FC - furnace cool 3

O

TABLE 5. GENERIC LETTER 92-01 RESPONSE: SECTION 2. ITEM b. 1 (3)

Subject:

10CFR50.61 and 10CFR50, Appendix G, III.A; Material Properties Related to PTS and Fracture Toughness Requirements -- APPLICABLE ONLY TO REACTOR VE5SELS CONSTRUCTED TO AN ASME CODE

' EARLIER THAN THE 1971 EDITION, SUMMER 1972 AD0ENDA Plant: Surry Unit I Column 1 Column 2 Column 3 Column 4 Column 5 C. 6 Beltline Heat Beltline Weld Wire Weld Flux Notes Plate or Number Weld Heat Lot

' Forging NB Forging .'

122V109 NB to IS Circ.: J726 25017 1197 (1,2)

IS Plate C4326-1 IS to LS Circ.(10 407.): SA-1585 72445 8597 IS Plate C4326-2 IS to LS Circ.(00 40".): SA-1650 72445 8632 LS Plate C4415-1 IS Longit.: SA-1494 8T1554 8579 LS Plate C4415-2 LS Longit.: SA-1494 BT1554 8579 LS Longit.: SA-1526 299L44 85 %

NOTES: (1) BAW-2150 (2) BAW-1909, Revision 1 (3) NB - Nonle Belt IS - Intermediate Shell LS - Lower Shell m _. . -, _

r

~

TABLE 6. GENERIC LETTER 92-01 RESPONSE: SECTION 2, ITEM b. 1 (4)

Subject:

10CFR50.61 and 10CFRSO, Appendix G, III.A; Material Properties Related to PTS and Fracture Toughness Requirements -- APPLICABLE ONLY TO REACTOR VESSELS CONSTRUCTED TO AN ASME CODE EARLIER THAN THE 1971 EDITION, SUMMER 1972 AD9ENDA Plant: Surry Unit 1 Column 1 Column 2 Column 3 Column 1 Column S Surveillance Surveillance Weld Wire Weld Flux Notes Plate or Weld Heat Lot Forging Heat Number C4326-1 SA-1526 299L44 8596 (1,2)

C4415-1 i

NOTES: (1) BAW-2150

(2) BAW-1909, Revision 1 1

E

TABLE 7. -GENERIC LETTER'92-01 RESPONSE:-SECTION'2. ITEM b. 1 (5)

Subject:

10CFR50.61 and 10CFR50, Appendix G, III.A; Material Prol:erties Related to PTS and Fracture i Toughness Requirements -- APPLICABLE ONLY TO REACTOR VESSELS CONSTRUCTED TO AN ASME CODE EARLIER THAN THE 1971 EDITION, SUMMER 1972 ADDENDA Plant: Surry Unit I Column 1 Column 2 C. 3 Material Chemical Composition, Weight Percent Notes C Mn P S Si Cr Ni Mo Cu BELTLINE MATERIALS 122V109 0.22 0.70 0.010 0.011 0.24 0.36 0.74 0.60 0.09 (2)

C4326-1 0.23 1.35 0.008 0.015 0.23 0.07 0.55 0.55 0.11 (2)

C4326-2 0.23 1.35 0.008 0.015 0.23- 0.07 0.55 0.55 0.11 (2)

C4415-1 0.22 1.33 0.014 0.014 0.23 0.08 0.50 0.55 0.11 (2)

C4415-2 0.22 1.33 0.014 0.014 0.23 0.08 0.50 0.55 0.11 (2)

J726 0.09- 1.67 ND ND 0.27 ND 0.10 0.44 0.33 (2)

SA-1494 0.09 1.52 0.015 0.012 0.44 0.08 0.63 0.37 0.18 (3)

SA-1585 0.08 1.45 0.016 0.016 0.51 0.09 0.59 0.38 0.21 (3)

SA-1650 0.08 1.43 0.018 0.014 0.40 0.09 0.59 0.38 0.21 (3)

SA-1526 0.09 1.53 0.013 0.017 0.53 0.09 0.68 0.42 0.35 (3)

LS TO Dutchman NA NA NA NA NA NA NA NA NA (1)

I SURVEILLANCE j MATERIALS C4326-1 0.23 1.35 0.008 0.015 0.23 0.07 0.55 0.55 0.11 (4)

C4415-1 0.22 1.33 0.014 0.014 0.23 0.08 0.50 0.55 0.11 (4)

SA-1526 0.09 1.53 0.013 0.017 0.53 0.09 0.68 0.42 0.35 (4)

REQUIRED:

State heat nuabar of weld wires used for determining above chemical composition if dif ferent from that in 1 (3). -- Not applicable --

NOTES FOR TABLE"7 ARE ON THE FOLLOWING PAGE.

- - - .ii--

. .i. -

L t

TABLE 7. (CONTINUED)

NOTES:

(1) Lower shell-to-dutchmen weld was fabricated by Rotterdam; information is not available. <

l (2) BAW-1909, Revision 1 (3) BAW-2121P (4) .BAW-1543, Revision 3 P

i l';;t -

r ' [! i lfi} tii?j !f!i i!i ; f!!! t !

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ti ,

E f as S o ni N iv O t c me rR .

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t ,

E f R ) e9 E 5 d9 1 -

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. E t i eG T

i F d T m iy E m e sr L o e no C S ot C e

( ca I l R s su E n ag N o p

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_ l n r m b e r e a  :

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g l c t e d , p n c  : L 'E p e e t S a r j n d T U, e b e l t f u l o fC o e S P C I( N R a j  ! .

i I

TA8LE 9. GEhEPIC LETTER 92-01 RESPONSE: SECTION 3, ITEM b t

Subject:

Generic Letter 88-11 Response Ccunitments; Utilization of Surveillance  !

Results Plant: Surry Unit I t Were surveillance results used in determining C,05t? Yes / No a i

Were surveillance results used_ in determining RT ,7 Yes a No / j If any "yes" boxes were checked above, state how the surveillance results were used:  !

i I

Determination of Upper-Shelf Energy per Regulatory Guide 1.99, Revision 2 Position 2, for SA-1585 j weld materials only. j i

l

References:

Letter to U. S. Nuclear Regulatory Commission from W. L. Stewart dated December 1, 1989.

Title:

' Virginia Electric and Power Company, Surry Unit I and 2: Response to  ;

. Request for Additional Information, Upper-Shelf Energy of Reactor Vessel Materials." [

Docket Nos. 50-280 and 50-281 License Nos. DPR-32 and DPR-37. .

i I

r i

t I

i TABLE 10. GENERIC LETTER 92-01 RESPONSE: SECTION 3, ITEM c

Subject:

Generic letter 88-11 Response Comitments: Difference Between Measured and Predicted (Regulatory Guide 1.99, Revision 2) Embrittlement Effects Plant: Surry Unit 1 Question I. Does measured ART er exceed ART., + 2a predicted by Regulatory Guide 1.99,

! Revision 27 Question II. Does measured C,USE drop exceed that obtained from Regulatory Guide 1.99, Revision 2, Figure 27 __

Column 1 Column 2 Column 3 Column 4 Column 5 Column 6 Column 7 Beltline Fluenge Measured Predicted Ouestion 1 Measured Predicted Ouestion 11 Material n/cm ART , ART.,+2a C,USE C,USE l yes/no Drop Drop yes/no 122V109 NA ND ND NA ND ND NA C4326-1 NA ND ND NA ND ND NA C4326-2 NA ND ND NA ND ND NA C4415-1 2.86E+18(2) 50(1) 82 No 5(1) 19 No 1.94E+19(1) 110(1) 120 No 9(1) 29 No C4415-2 NA ND ND NA ND ND NA J726 NA ND ND NA ND ND NA SA-1585 5.10E+18(2) 148(2) 188 No 22(2) 24(3) No SA-1650 NA ND ND HA ND ND NA ,

SA-1494 NA ND ND NA ND ND NA SA-1526 2.86E+18(2) .167(2) 203 No 17(2) 25 No ,

1.94E+19(1) 240(1) 320 No _

. 20(1) 33 No NOTES: (1) WCAP-il415 .

(2) BAW-1803, Revision 1 (3) BAW-1910P i

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TABLE I. GENERIC LETTER 92-01 RESPONSE: SECTION I

Subject:

10CFR50, Appendix H: Adherence to RYSP Requirements Plant: Surry Unit 2 Quest bn I: Does RVSP meet ASTM E 185-73, E 185-79, or E 185-82? Yes a No /

Question II: Is plant one of the following? ANO-I, Crystal River-3, Davis Besse, R. E. Ginna, Oconee-1, Oconee-2, Oconee-3, Point Beach-I, Point Beach-2, Rancho Seco, Surry-I, Surry-2 Turkey Point-3, Turkey Point-4, Zior.-1, Zion-2. Yes / No o IF ANSWER IS "YES" TO EITHER QUESTION I GR QUESTION II, PROCEE0 TO TABLE 2.

IF ANSWER IS "NO" TO BOTH QUESTION I AND QUESTION II, PROCEED TO QUESTION III AND QUESTION IV.

Question III: If plan is to revise RVSP to meet requirements of 10CFR50, Appendix H, when will revised RVSP be submitted to NRC?

f Response: ,

j- Not applicable (see Question I and II above) i j! ' Question IV: If plan is not to revise RYSP to meet requirements of 10CFR50, Appendix H, when will j exemptian fror 10CFR50, Appendix H be requested from NRC? i 1 i

' Response:

i Not applicable (see Question I and II aoove) l

! NOTES: NCAP-8085: Surveillance Program Description (ASTM E 185-70) 1 i

5

I i

TABLE 2. GENEPIC LETTER 92-01 RESPONSE: SECTION 2. ITEM a l fi

Subject:

10CFR50, Appendix G, C,1USE t Requirements. i

} Plant: Surry Unit 2 1

Column I Column 2 Column 3 Column 4 t

i-Limiting Initial EFPY to reach If Column 2 is within license Action taken t'

Material. USE C,USE<50 ft-lb period: C,USE at indicated time per IV.A.1

' ft-lb Column 3A Column 3B i i-4 12/16/91 EOL 1

! LIMITING Analysis per 10CFR50, i

BELTLINE WELD Appendix G,Section V.C.3. i' is scheduled for 1993

{ SA-1585 70 (5) 32, approx. NA NA under the sponsorship of

B&WOG Reactor Vessel j ,

Working Group.

2 LIMITING BELTLINE PLATE OR FORGING r C4208-2 91 (6) >32 NA NA i NA I

NOTES FOR TABLE 2 ARE ON THE FOLLOWING PAGE.

J i i

4 i.

i -

TABLE 2 (CONTINUED)

NOTES: (1) fluence values taken at %-thickness.

(2) C USE y values for 12/16/91 and E0L (Column 3) calculated per requirements outlined in Regulatory Guide 1.99, Revision 7 Eareyrapn C.I.2.

(3) Analyses that have demonstrated the required margin of safety for the Zion and Turkey Point plants at load level A and B conditions have been submitted to the NRC (Reports BAW-2118P and BAW-2148P).

The results of these two analyses are anticipated to bound the outcome of the Surry Unit 2 analysis.

(4) C yUSE is calculated on the basis of RGl.99, Rev. 2, Position I. SECY 91-333 states that this orocedure is " inadequate." Recognizing this and having provided for the uniqueness of Linde 80 welds in BAW-1803, Rev.1, the meth i for calculating C USE y in BAW-1803, Rev.1, is put forward as more representative and is intended to be used for predicting the behavice of these welds in licensing applications.

(5) BAW-1803 (6) BAW-10046P i

i t

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! f

. i TABLE'3. ' GENERIC LETTER 92-01 RESPONSE

SECTION 2, ITEM b. 1 (1)

Subject:

10CFR50.61 and 100FR50, Appendix G, III.A; Material Properties Related to t PTS and Fracture Toughness Requirements

~

i Plant: Surry Unit 2 .

l Column 1 Column 2 Column 3 Column 4 Column 5 C.6  !

Beltline Unirradiated Charpy Test Results Unirrad. Unirrad. Method of Notes I Materials Dropwt. RT Determng l Col. 2a Col. 2b Col. 2c Col. 2d Test fr RT, Results  :

i~ T Cy C, C, C, pT ,

10 F 30 ft-lb 50 ft-lb 35 MLE

' ft-lb F F F j FORGING i i

j 123V303 142,83,122 -20 0 +5 +30 +30 NB-2331 (3,6) i 110,90,168 ('

j- 105

PLATE j' C4208-2 64,63,75 -45 -20 -20 -30 -30 NB-2331 (1,2,7)  !

C4339-1 NO' +25 +50 +45 -10 -10 MB-2331 ;1,2,7)  !

C4331-2 46,60,25 +5 +35 +32 -10 -10 NE-2331 (1,2,7)  !

(1,2,7)

C4339-2' 48,45,25 -5 +25 +10 -20 -20 NB-2331 t

i WELD 4

L737 75,62,66 NO NO NO ND 0 Est. (3) (1,8)  !

SA-1585 31,32,31 NO NO NO ND -5 Est. (5) (1,9,10) l t 50,54,51  !

- R3008 66,51,46 NO NC NO ND 0 Est. (3) (1,11) l WF-4 40,31,34 NO NO NO NO -5 Est. (5) (1,10)  !

WF-8 45,38,30 NO NO NO ND Est. (5) (1,12) l l

NOTES FOR TABLE 3 ARE ON THE FOLLOWING PAGE. .

l o r

?

TABLE 3 (CONTINUED)

NOTES TO TABLE 3:

(1) BAW-2150 (2) Supplier Test Report.

(3) BAW-1909, Revision 1 (4) BAW-10046P, pp 3-17, -18; mean of most conservative value for each of 24 cases.

(5) BAW-1803, Revision 1, Tables 3-1 and 3-2; mean of RT values for 34 Linde 80 welds.

(6) Values from BWNS Drawing 02-1167428E-00, Sheet 2ofI.r (7) Values are for 60 hr stress relief. ,

(8) Values are for 24 hr stress relief. '

(9) C (+10F) y values; specimens from center and surface of test block.

(10) Cy (+10F) values are for 80 hr stress relief.

(11) C (+10F) values are for 25 hr stress relief.

(12) C;(+10F) values are for 48 hr stress relief.

k e

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- TABLE 4.- GENERIC LETTER 92-01 RESPONSE

SECTION 2, ITEM b. 1 (2)

Subject:

10CFR50.61 and'10CFR50, Appendix G, III.A; Material Properties Related to PTS and Fracture Toughness Requirements -- APPLICABLE ONLY TO REACTOR YESSELS CONSTRUCTtD TO AN ASME CODE i i

EARLIER THAN THE 1971 EDITION. SUPMER 1972 ADDENDA i

. P1 ant: Surry Unit 2 r 1

l Column 1 Column 2 . _ _ .

Col. 3 i  :

Material Heat Treatment Notes l BELTLINE i

! MATERIALS I 123V303VA1 155f6 12h/WQ; 1200F-22h/AC; ll25F-40h/FC (1)  !

C4331-2 160t, soSOF-9h/BQ; 1200-1225F-9h/BQ; 1]?SF-60h/FC -

! C4339-2~ 1600-1650F-9h/BQ;- 1200-1225F-9h/BQ; 1425F-60h/FC  !

J C4339-1 1600-1650F-9h/BQ; 1200-1225F-9h/BQ; 1125F-60h/FC C4208-2 '. 1600-1650F-9h/BQ: 1200-1225F-9h/BQ; ll257-60h/FC 3 l i

! L737 Il30F-24h/FC R3008 ll30F-25h/FC l SA-1585 ll25125F-80h/FC ,

-l WF.4 II25125F-805/FC WF-8 ll25125F-48h/FC

' I SURVEILLANCE MATERIALS l '

l C4339 1625F-9h/BQ: 1212F-9h/BQ; Il40F-15\h/FC (1)

R3008 Il40F-15%h/FC i 4

i i: NOTES:-

il .(1) BAW-1909. Revision 1

(2) BQ - brine quench l- FC - furnace cool

\ l

_.J

TABLE 5. GENERIC LETTER 92-01 RFSPONSE: SECTION 2. ITEM b, 1 (3)

Subject:

IG;sR50.61 and 10CFR50, Appendix G III.A; Material Properties Related to PTS and Fracture Toughness Requirements -- APPLICABLE ONLY TO REACTOR VESSELS CONSTRUCTED TO AN ASME CODE EARLIER THAN THE 1971 EDITION. SUMMER 1972 ADDENDA Plant: Surry Unit 2 Column 1 Column 2 Column 3 Column 4 Column S C. 6 Beltline Heat Beltline Weld Wire Weld Flux Notes Plate or Number Weld Heat Lot Forging __

NB Forging 123V303 NB to IS Circ.: L737 4275 02275 (1,2, IS Plate C4331-2 IS to LS Circ.: R3008 0227 LW320 3)

IS Longit.(ID 50%): WF-4 8T1762 8597 IS Plate C4339-2 8579 LS Plate C4208-2 IS Longit.(DD 50%): SA-1585 72445 C4339-1 LS Longit.: WF-4 8TI762 8597 LS Plate 8597 LS Longit.(ID 63%): WF-4 8T1762 LS Longit.(OD 37%): WF-8 8T1762 8632 j NOTES: (1) BAW-2150 (2) BAW-1909, Revision 1 (3) Mt. Vernon fabrication process sheets (4) NB - Nozzle Belt IS - Intermediate Shell LS ' Lower shell g

..m_____

i TABLE 6. GENERIC LETTER 92-01 RESPONSE: SECTION 2, ITEM b. 1 (4)

Subject:

10CFR50.61 and 10CFR50, Appendix G, III.A; Material Properties Related to PTS and Fracture Toughness Requirements -- APPLICABLE ONLY TO REACTOR VESSELS CONSTRUCTED TO AN ASME CODE EARLIER THAN THE 1971 EDITION, SUMMER 1972 ADDENDA Plant: Surry Unit 2 Column 1- Column 2 Column 3 Column 4 Column 5 Surveillance Surveillance Weld Wire Weld Flux Notes Plate or Weld Heat Lot Forging Heat Number 1 C4339-1 R3008 0227 LW320 (1,2)

I.

NOTES: (1) BAW-2150 (2) BAW-1909, Revision 1

TABLE 7. GENERIC LETTER 92-01 RESPONSE: SECTION 2, ITEM b, 1 (5)

Subject:

10CFR50.61 and 10CFR50, Appendix G, III.A; Material Properties Related to PTS and Fracture Toughness Requirements -- APPLICABLE ONLY TO REACTOR VESSELS CONSTRUCTED TO AN ASME CODE EARLIER THAN THE 1971 EDITION. SUMMER 1972 ADDENDA Plant: Surry Unit 2 Column 2 C. 3 Columr. 1 Chemical Composition, Weight Percent Notes Material C Mn P S 'Si Cr Ni Mo Cu BELTLINE MATERIALS 0.20 0.63 0.010 0.010 0.24 0.36 0.73 0.58 0.09 (2) 123V303 0.12 0.23 1.42 0.009 0.015 0.22 ND 0.60 0.55 (2) ,

C4331-2 0.54 0.11 C4339-2 0.23 1.30 0.012 0.014 0.25 ND 0.54 (2) 0.21 1.28 0.008 0.013 0.24 ND 0.55 0.55 0.15 (2)

C4208-2 0.23 1.30 0.012 0.014 9.25 ND' O.54 0.54 0.11 (2)

C4339-1 0.38 0.35 L737 0.08 1.74 ND ND 0.35 ND 0.10 (2) 0.08 1.45 0.016 0.016 0.51 0.09 0.59 0.38 0.21 (3)

SA-1585 0.19 0.09 1.51 0.017 0.016 0.46' !0.10 0.56 0.41 (2)

R3008 0.07 1.48 0.017 0.011 0.51 'O.12 0.55 0.41 0.20 (3)

WF-4 0.06 1.45 0.009 0.009 0.53 0.12 0.55 0.41 0.20 (3)

WF-8 LS-to-Dutchman NA NA NA NA NA NA NA NA NA (1)

SURVEILLANCE MATERIALS 0.23 1.30 0.012 0.014 0.25 0.08 0.54  ;).54 0.11 (4)

C4339-1 0.19 0.09 1.51 0.017- 0.016 0.46 0.10 0.56 0.41 (4)

R3008 REQUIRED:

State heat number of weld wires used for determining above chemical composition if different from '

that in 1_(3). .-- Not applicable --

NOTES FOR. TABLE 7 ARE ON THE FOLLOWING PAGE.

l

.i...... , , . . . . . .

..i. ..

. . . . . . . . . . . ~ . . .

l TABLE 7. (CONTINUED)

NOTES:

(1) Lower shell-to-dutchman weld was fabricated by Rotterdam; infor1mation is not available (NA).

(2) BAW-1909, Revision 1 (3) BAW-2121P (4) BAW-1543, Revision 3

! l i-1

F j i ,-  !

.. i-

. i

, TABLE 8. GENERIC LETTER 92-01 RESPONSE: SECTION 3, ITEM a

i. .

i i, .

Subject:

Generic Letter 88-11 Response Commitments; Effect of Irradiation Temperature i

' I

. Plant: Surry Unit 2  ;

i' Cold leg Temperature (T,,,,): 543 F (See Figure 4-5)

' l

If yT ,, RT,) in accordance with Regulatory Guide 1.99,' Revision 2(C USE, is '<525 F, state how t ,

i i:

j. Not applicable i  !

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I I

- s 3- .

!. I l

References:

i' i None

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Il

+, - -- .. - - - _ .- - _ - - _ - - . _ _ - - . _ _ . _ _ . _ - _ . - - - - - . . . - - _ - !

TABLE 9. GENERIC LETTER 92-01 RESPONSE: SECTION 3, ITEM b

Subject:

Genavic Letter 88-11 Response Commitments; Utilization of Surveillance R m its Plant: Surry Unit 2 Were surveillance results used in determining CuUSE? Yes_ o No /

I Were surveillance results used in determining RT.,7 Yes a No /

If any "yes" boxes were checked above, state how the surveillance' results were used:

References:

None

-- n _

a-_

2 TABLE 10. GENERIC LETTER 92-01 RESPONSE: SECTION 3, ITEM c

Subject:

Generic Letter 88-11 Response Commitments; Difference Between Measured and Predicted (Regulatory Guide 1.99, Revision 2) Embrittlement Effects Plant: Surry Unit 2 Question I. Does measured ART , exceed ART , + 20 predicted by Regulatory Guide 1.99, Revision 2?

Question II. Does measured yC VSE drop eyteed that obtained from Regulatory Guide 1.99, Revision 2, Figure 2?

Column 1 Column 2 Column 3 Column 4 Column 5 Column 6 Column 7 Beltline Fluenge Measured Predicted Ouestion I Measured Preditted Qugstion II Material n/cm ART , ART ,+2a Cy UF C,USE Yes/No '

Drr a Drop yes/no 123V303 NA ND ND NA NL ND NA C4208-2 NA ND ND NA ND ND NA C4339-1 3.02E+18(1) 45(1) 83 No 10(1) 16 No 1.88E+19(1) 75(1) 120 No 11(1) 24 No C4331-2 NA ND ND NA ND  ; ND NA C4339-2 NA ND ND NA ND ND NA I

L737 NA ND ID NA ND ND NA R3008 3.02E+18(1) 95(1) 157 No 20(1) 22 No 1.88E+19(1) 145(1) 233 No 30(1) 34 No SA-1585 5.10E+18(2) 148(2) 188 No 22(2) 24(3) No WF-4 NA ND ND NA ND ND NA i

~

WF-8 NA ND ND NA { ND ND NA l

NOTES: (1) WCAP-11499 (2) BAW-180' Revision 1 (3) BAW-19 a

=' ' ' "

• 6 @uk,-sm,, , ' 'd'dA d-E-- wX5- h h-es-6_. 4p,.. . ,- ",'-M"#

-* ' h- f 4 d-4,Jg_a .;mp4,4 g , ,,, "O*r*&-m?,ah--'4 a __, yy,, _

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-,. ve,ar e. , e..r,--r--e e---,,,--e,, s ,--,..,.# e w w w r ,y. . ;e.,m.,,,-we,-,eee.,,sr--,- ,w w.w y, -,%m..-,,-wmy.&. , ,,,w.y.-, - - ,r-c evre p e -tg~ e ,v w ,,.c,,.%, #

TABLE 1. GENERIC LETTER 92-01 RESPONSE: SECTION 1

Subject:

10CFR50, Appendix H; Adherence to RVSP Requirements Plant: Three Mile Island Unit 1 Does RVSP meet ASTM E 185-73, E 185-79, or E 185-82? Yes a No /

Question I:

Question II: Is plant one of the following? ANO-1, Crystal River-3, Davis Besse, R. E. Ginna, Oconee-1, Oconee-2, Oconee-3, Point Beach-1, Point Beach ~i, Rancho Seco, Surry-1, Surry-2, Turke, Point-3, Turkey Point-4, Zion-1, Zion-2. Yes / No o IF ANSWER IS "YES" TO EITliER QUESTION I OR QUESTION II, PROCEED TO TABLE 2.

IF ANSWER IS "N0" TO BOTH QUESTION I AND QUESTION II, PROCEED TO QUESTION III AND QUESTION IV.

Question III: If plan is to revise RVSP to meet requirements of 10CFR50, Appendix H, when will revised RVSP be submitted to NRC?

Response

Not applicable (see Question I and II above)

Ouestion IV: If plan is not to revise RVSP to meet requirements of 10CFR50, Appendix H, when will exemption from 10CFR50, Appendix H be reonested from NRC?

i

Response

Not applicable (see Question I and II above)

_~

NOTES: BAW-10006A, Revision 3: Surveillance Program Description (ASTM E 185-70)

d i

-TABLE 2. GENERIC LETTER:92-01' RESPONSE: . SECTION 2,: ITEM a.

i

Subject:

10CFR50, Appendix G, C USE Requirements-Plant: Three Mile Island Unit'l

, Column 1 Column 2 Column 3 Column 4 j

, Limiting Initial EFPY to reach If Column 2 is within licente Action taken  !  :

l Material USE Cy OSE<50. ft-lb period: C.USE at- indicated time per IV.A.1 l ft-lb '

Column 3A Column 3B l i

12/16/91 End of License  !

(26.17) i IIMITING Analysis per 10CFR50, 22LTLINE WELD Appendix G,Section V.C.3.  ;

.is scheduled for 1993 l

WF-25 70'(5) 4, approx. 47 44 under the sponsorshipfof- l

, B&WOG Reactor Vessel  !

!, - Working Group.: l:

. I i '

LIMITING BELTLINE PLATE OR FORGING l C3307-l' 91 (6) >32 NA NA NA ,

I NOTES FOR TABLE 2 ARE ON THE FOLLOWING PAGE. l I

t i

a

+

TABLE 2 (CONTINUED)

NOTES: (1) Fluence values taken at \-thickness.

(2) C,USE values for 12/16/91 and End of License (Column 3) calculated per requirements outlined in Regulatory Guide 1.99, Revision 2, Paragraph C.I.2.

(3) Analyses that have demonstrated the required margin of safety for the Zion and Turkey Point plants at load level A and B conditions have been submitted to the NRC (Reports BAW-2118P and BAW-2148P).

The results af these two analyses are anticipated to bound the outcome of the Three Mile Island Unit I analysis.

(4) C yUSE is calculated on the basis of RGl.99, Rev. 2, Position 1. SECY 91-333 states'that this procedure is " inadequate." Recognizing this and having provided for the uniqueness of Linde 80 welds in BAW-1803, Rev.1, the method for calculating C OSE y in BAW-1803, Rev.1, is put forward as n: ore representative and is intended to be used for predicting the behavior of these welds in licensing applications.

(5) BAW-1803 (6) BAW-10046P l

l 1

a 4

l 2

4 TABLE 3. GENERIC LETTER 92-01 RESPONSE: SECTION 2, ITEM b, 1 (1) ,

q

Subject:

10CFR50.61 and 10CFR50, Appendix G, III.A; Material Properties Related to i PTS and Fracture Toughness Requirements ,

Plant: 'Three Mile Island Unit 1 Column 1 Column 2 Column 3 Column 4 Column 5 C.6 Beltline Unirradiated Charpy Test ResuPs Unirrad. Unirrad. Method of Notes

. Materials Dropweight RT 'Determng

' Test f, RT ,

Col. 2a Col. 2b Col. 2c. Col. 2d Results T

Cy .C C C, yr

10 F 30fl-lb 50ff-lb 35 MLE j ft-lb F F F FORGING 1 ARY 59 117,110,101 ND ND ND ND' +3 Est. (2) (1,6)

, 120,122,123 PLATE- .

C2789-1 50,36,33 -ND ND ND 0 +1 Est. (3) (1,7)

C2789-2 42,37,35 ND, ND ND +10 +1 Est.'(3) (1,7) ,

, C3307-1 ND ND ND ND. +10 +1 Est.-(3) (I,7) i

C3251-1 43,40,29 ND. ND ND -10 +1 Est. (3) (1,7). ,

71,59,26 -i 4-i WELD "

WF-70. 39,35,44 ND ND ND ND . +18 Eval . (4)' (1,7,9)

E WF-25 38,28,49 ND ND ND' ND -5' Est.-(5)' (1,8) t WF-67 29,35,30 ND ND ND. ND -5 Est. (5) (1,8)  !

Est. (5) (1,8)

WF-8 45,38,30' ND ND ND -ND -5  ;

i SA-1526 33,33,33 ND. 'ND ND ND- -5 Est.-(5). I1,8)  !

l. SA-1494 54,25,44 ND ND ND ND- -5 Est. (5) l (1,8) i- ' NOTES' FOR TABLE 3 ARE ON THE FULLOWING PAGE.

j j

l

TABLE 3 (CONTINilED)

NOTES TO TABLE 3:

(1) BAW-1820 (2) BAW-10046P, pp 3-17, -18; mean of most conserva'ive value for each of 24 cases.

(3) BAW-10046P, pp 3-18; metn of most conservative value for each of 13 cases.

(4) BAW-2100 (5) BAW-1803, Revision 1, Tables 3-1 and 3-2; mean of RT , values for 34 Linde 80 welds.

(6) C,(+10F) values are for 60 hr stress-relief.

(7) Cy (+10F) values are for 40 hr stress-relief.

(8) Cy (+10F) values are for 48 hr stress-relief.

(9) RI , value- for 40 hr stress-relief maximum.

i l

~

e

I' -

TABLE 4. GENERIC LETTER 92-01 RESPONSE: SECTION 2, ITEM b, 1 (2) ,

Subject:

10CFR50.61 and 10CFR50, Appendix G, III.A; Material Properties Related to PTS and Fracture Toughness Requirements -- APPLICABLE ONLY TO REACTOR VESSELS CONSTRUCTED TO AN ASME CODE

--EARLIER THAN THE 1971 EDITION. SUMMER 1972 ADDENDA Plant: Three Mile: Island Unit 1 Column 1 Column 2 Col. 3 Material Heat ireatment Hotes I <

BELTLINE MATERIALS ARY 59 1600i20F-7h/WQ; 1230120F-14h/WQ; 1100-Il50F-45th/FC (cumul.) (1,2)

C2789-1 '1510-1535F-Sh/BQ; 1200-1225F-Sh/BQ; 1100-1150F-36h/FC.(cumul.)  :

C2789-2 1510-1535F-Sh/BQ; 1200-1225F-Sh/BQ; 1100-1150F-36h/FC (cumul.)

C3307-l' 1600-1650F-9%h/BQ; 1200-1225F-9%h/BQ; 1225-1250F-9%h/BQ; 1100-1150F-37%h/FC (cumul.)

3 C3251-1 1600-1650F-9%h/BQ 1200-1225F-9)h/BQ; 1100-Il50F-37%h/FC (cumul.)

} WF-70' 1100-Il50F-27%h/FC 4 - WF-25 1100-1150F-35h/FC (cumul.)

WF-67/WF-70 1100-Il50F-35n/FC (cumul.)

4 WF-8 Il00-Il50F-36h/FC (cumul.) . i SA-1526/SA-1494 Il00-Il50F-37%h/FC (cumul.)

SURVEILLANCE MATERIALS. . . .

C2789-2 1510-1535F-Sh/BQ; 1200-1225F-Sh/BQ; 1100-Il50F-27%h/FC (1) .;

t. C3307 1600-1650F-9%h/BQ; 1200-1225F-9%h/BQ: 1225-1250F-9%h/BQ;.

. Il00-1150F-27%h/FC WF-25 1100-Il50F-27%h/FC NOTES FOR TABLE 4'ARE ON_FOLLOWING PAGE.

t e

P h

a

I i

TABLE 4. (CONTINUED)

NOTES: e i

, (1) BAW-1820 (2) Additional stress relief information per Mt. Vernon process drawing. '

(3) WQ - water quench BQ - brine quench'

! FC - furnace cool i

4 i

,l 1

4

r

, ac.me j TABLE'5.' JENERIC LETTER 92-01 RESPONSE: 'SECTION 2, ITEM b, 1 (3)

Subject:

10CFR50.61 and 10CFR50, Appendix G,.III.A;' Material Propertier Related to PTS and Fracture j i ~ Toughness Requirements -- APPLICABLE ONLY TO REACTOR VESSELS CONSTRUCTED TO AN ASME CODE.  !

!. EARLIER THAN THE 1971 EDITION. SUMMER 1972 ADDENDA Plant: . Three Mile Island Unit.1 i

l. Column I Column 2- Column 3- Column 4 Column 5 C. 6 .

. Beltline Heat Beltline Weld Wire Weld Flux ' Notes l

Plate or Number Weld Heat Lot'

Forging _

Lower NB-Forging.. ARY 59, 123S454' .NB to US Circ.: WF-70: 72105 '8669 -(l) i US' Plate C2789-1 US to LS Circ.: WF-25 299L44 8650 l' US Plate'. C2789-2 LS to Dutch Circ.(ID 50%): WF-67 72442 ' 8669 ,

l LS Plate C3307 LS to Dutch Circ.(0D 50%): WF-70 '72105 8669 1S Plate C3251-1 US Longit.: WF-8 8T1762 8632 299L44, i

LS Longit.

SA-1526' 8596 l LS Longit.: SA-1494' 8T1554- 8579' I

NOTES: (1) BAW-1820 i i

(2) NB - Nozzle Belt'

! US - Upper Shell l LS - Lower Shell  ;

I - ,

i i

I 4

! j

,_ -.. . ~ . ,

TABLE 6. GENERIC LETTER 92-01 RESPONSE: SECTION 2, ITEM b, 1 (4)

Subject:

10CFR50.61 and 10CFR50, Appendix G, III.A; Material Properties Related to PTS and Fracture Toughness Requirements -- APPLICABLE ONLY TO REACTOR VESSELS CONSTRUCTED TO AN ASME CODE EARLIER THAN THE 1971 EDITION, SUMMER 1972 ADDENDA Dlant: Three Nile Island Unit 1 Column 1 Column 2 Column 3 Column 4 Column 5 Surveillance Surveillance Weld Wire Weld Flux Notes Plate or Weld Heat Lot Forging Heat Number _ _ _ _

' 8650 l C2789-2 WF-25 299L44 (1)

C3307-1 NOTES: (1) BAW-1820

r!

t TABLE 7. GENERIC LETTER 92-01 RESPONSE: 'SECTION 2, ITEM b, 1-(5)

Subject:

10CFR50.61 and 10CFR50, Appendix G, III.A; Material Properties Related.to PTS and Fracture  ;

Toughness-Requirements -- APPLICABLE ONLY TO REACTOR VESSELS CONSTRUCTED TO AN ASME CODE EARLIER THAN THE 1971 EDITION. SUMMER 1972 ADDENDA Plant: Three Mile Island Unit 1 Column 1 Column 2 C. 3 r Material Chemical Composition, Weight Percent Notes C Mn P S Si Cr Ni Mo Cu

' BELTLINE MATERIALS

., ARY 59 0.26 0.63 0.006- 0.008 0.28 0.34 0.72 0.64 0.08 (1)

C2789-1 0.24 ,1.36 0.010 0.017 0.23 0.19 0.57' O.51_ 0.09 (1)' J C2789-2 0.24 1.36 0.010 _ 0.017 0.23 0.19 0.57 0.51 0.09 (1) ,

, C3307-l 0.21- 1.24_ 0.010 0.016 0.27 0.12 0.55 0.47 0.12 (1) 0.23 -1.41- 0.012 0.013 0.21- 0.14 0.50- 0.47 0.11 C3251-1 '(1)

SA-1494- 0.09 '1.52 0.015 0.012 0.44 0.08 0.63 0.37 0.18 (2)

SA-1526. 0.09 1.53: 0.013- 0.017 0.53 0.09 0.68 0.42 0.35 (2)

WF-8 0.06 1.45 0.009 0.009 0.53 0.12- 0.55 0.41 0.20 (2)

WF-25 0.09 1.60 0.015 0.016 0.50 0.09 0.68 0.42. 0.35' .(2)

WF-67 0.08 1.55 0.021 0.016 0.58' O.09 0.60 0.39 0.24 (2)

WF-70 0,09- 1.63 0.018 0.009 0.54 0.10 0.59 0.40 0.35 (2)

SURVEILLANCE

MATERIALS- . . .

C2789-2 0.24 1.36- '0.010 0.017.- 0.23- -0.19 0.57 0.51 0.09 -(3)

C3307-1 0.21- 1.24 0.010.' O.016 0.27 0.12 0.55- 0.47 0.12- (3)- .

'WF-25 0.09 1.62 0.014 _ 0.015- 0.46 0.10 0.66 0.40 0.33 -(3)  :

REQUIRED: State heat number of weld wires used for. determining above chemical composition if different from l that in.1 (3). - .Not applicable.-- l

! NOTE'S FOR TABLE 7 ARE ON THE'FOLLOWING PAGE.

t

-i i

~ - _ . - - _ . _ _ . - - - - ._

-. . . - - . . . - . _. . . = .

- . ~. . - - _ .- .. ... _~ . .. -..-. - -. . .- ... _,- -

!i l TABLE 9. GENERIC LETTER 92-01 RESPONSE: SECTION 3, ITEM b

Subject:

Generic Letter 88-11 Response Commitments; Utilization of Surveillance Results l Plant: 'Three Mile Island Unit 1 Were surveillance results used in determining Co0SE? Yes o No /

Were surveillance results used in determining RT,n,? Yes a No /

If any "yes" boxes were checked above, state how the surveillance results were used:

1 1

References:

j None .i t

e a 8 a

T

]

}

s

i TABLE'10. -GENERIC LETTER 92-01 RESPONSE: SECTION 3,' ITEM c '

]

4

Subject:

Generic Letter 88-11 Response Commitments; Difference Between Measured and Predicted (Regulatory Guide 1.99,- Revision 2) Embrittlement Effects ,

. Plant: Three Mile Island Unit 1 Question I. Does measured ART , exceed ART , + 20 predicted by Regulatory Guide 1.99, Revision 2?

i Question II. Does measured yC USE drop exceed that obtained from Regulatory Guide 1.99, Revision 2, Figure 2? '

Column 1 Column 2 Column 3 Column 4 Column 5 Column 6 Column 7 Beltline Fluenge Measured Predicted Question I Measured Predicted Question II-Material n/cm ' ART , . ART ,+2a If "yes" CyUSE. CyUSE If "yes" .

'(2,4) see Note (5) Drop ' Drop see Note (5) l i

ARY 59 ---

ND ND --

ND ND --

C2789-1 .--- ND ND- --

ND ND --

4 C2789-2 ' l.07E+18' 5(1) 50 No 10(1) 10(1) No 8.66E+18 13(1) 90 No 19(1)- 17(1) Yes l C3307-1 ---

ND ND .-- ND ND --

( C3251-1 ---

ND ND --

ND ND -- '- i WF-70 ' 6.63E+18 -135(2) 259 No' 13(2) 22(6) No .

WF-25 - 1.07E+18 ' 124 (2) '- 148. No 17(2) 24(1) No 8.66E+18 203(2). -261 No 31(2) 34(1) No 7./9E+18- '214(2)- 263 No 25(2)- 30(7) No.

WF-67 6.09E+18 160(2) 200 No. 15(2) 23(7) Na WF-8 ---

NDL ND --

ND ND . --

SA-1526 - 2.86E+18' 167(2) 203 No 17(2): 25' No l" 320 No 20(3) 33 Nc-

'1.94E+19 240(3) 4 SA-1494' .h .

-- I ND- ND --

ND ND --

Atypical' l.17E+18 -28(4) 138 No '9(4) 25(4) No

6.56E+18- 122(4) 216 No 16(4) 32(4) No- i

! - 7.50E+18 119(4) 223 No 11(4) 32(4) Nu

,, 1.08E+19 120(4) 242 No 15(4) 34(4) No

~

NTTESFORTABLE10AREONTHEFOLLOWINGPAGE.

1' e w.- m - . - - _ _ _..----%_____- - - -

• _ _ _ _ _ . - ---.--.____._-__S'e- v..-m- m _ . -

TABLE 7. (CONTINUED)

NOTES:

(1) BAW-1820 (2) BAW-2121P i (3) BAW-1543, Revision 3 i

f

}

I l j l

. t I

I

{

I i.

i

i i -i

TABLE 8. GENERIC LETTER 92-01 RESPONSE: SECTION 3, ITEM a

Subject:

Generic Letter 88-11 Response Commitments; Effect of Irradiation Tencerature Plant: Three Mile Island Unit 1 Cold Leg Temperature (T_,a): 556 F (See Figure 4-1) if T t is <525 F, state how this was considered in determination of embrittlement effects (Cy U$E,o RTer) in accordance with Regulatory Guide 199, Revisien 2:

Not applicable L

I 1

4 m -

References:

None

4 TABLE 10 (CONTINUED)

NOT'ES: (1) BAW-1901 3

(2) BAW-1803, Revision 1 -

(3) WCAP-II415 *

, (4) BAW-2049 '

(5) The only instance where a measured " shift" or " drop" exceedt that predicted by calculation performed in accordance with Regulatory Guide 1.99, Revisi n 2, is for a " drop" for base metal tad- the predicted " drop" exceeds the measured "dr .jp" by 2 ft-lbs which is not considered significant. The requirements of 10CFR50, Appandix G, were not violated, and

! there being no further application of the " drop" data, the effect of these surveillance i results are therefore not significant.

(6) BAW-1920P (7) BAW-1910P i

i J

i

+

8

q

, i e

f I

i l 4 f '

j [_ TABLE 1. ' GENERIC LETTER 92-01 RESPONSE: SECTION 1

Subject:

'10CFR50, Appendix H; Adherence'to RYSP Requirements 'l l- .

Plant: Turkey Point Unit 3 l

l -Question.I: Does RVSP meet ASTM'E 185-73, E 185-79, or E 185-82? Yes a No / -

i Question II: Is plant.one of the following? ANO-1, Crystal River-3, Davis 8 esse, R. E. Ginna, '

l Oconee-1, Oconee-2, Oconee-3, Point Beach-1, Point Beach-2, Rancho Seco,' Surry-1,  ;

Surry-2, Turkey Point-3, Turkey Point-4, Zion-1, Zion-2. Yes / No o IF' ANSWER IS "YES" TO'EITHER QUESTION I OR QUESTION II, PROCEED TO TABLE-2.  ;

i IF' ANSWER IS "M0" TO BOTH QUESTION I AND QUESTION II, PROCEED TO QUESTION III- AND OVESTION IV.  ;

~!

Question III: If plan is to revise RVSP to meet requirements of 10CFR50, Appendix H, when will.

1 revised RVSP be submitted to NRC? '

j I Response:

2

.Not applicable (see Question I and II-above).

If plan. is not to revise RVSP to meet. requirements of .10CFR50, Appendix H, when will

~

j Question IV:

exemption from 10CFR50, Appendix:H be requested from NRC?

1

! Response:-

i '%ot-app?icable (see-Question I and II above)_

1 i

NOTES: WCAP-7656: -Surveillance Program. Description  ;

(ASTM E 185-66) t I

! f L '

1 i" -i

) <

1

- . m + _

TABLE 2. GENERIC LETTER 92-01 RESPONSE: SECTION 2, ITEM a

Subject:

10CFR50, Appendix G, C,USE Requirements Plant: Turkey Point Unit 3 j Column 1 Column 2 Column 3 Column 4

, limiting Initial EFPY to reach If Column 2 is within license Action taken Material USE Cy VSE<50 ft-lb period: C USE at indicated time per IV.A.1 ft lb Column 3A Column 3B 12/16/91 E0L LIMITING An analysis which BELTLINE WELD demenstrates that this material provides margin SA-1101 65 (4) 2, approx. 39 36 of safety against fracture equivalent to that required by ASME Section III, Appendix G, has been 4

performed under the sponsorship of the B&W Owners Group's Reactor Vessel Working Group and submitted to the NRC as report BAW-2118P.

LIMITING BELTLINE PLATE OR FORGING 123S266 154 (4) >32 NA NA NA NOTES FOR TABLE 2 ARE ON THE FOLLOWING PAGE.

. . - . . _-- .. . . .-. . - . . . . . .= . . . - ~.- _- _ ~ . , _ . - . .. ..

TABLE 2 (C0f4TIf4UED)

NOTES: (1) Fluence values taken at \-thickness. .

(2) C,USE values for 12/16/91 and EOL (Column 3) calculated per requirements outlined in Regulatory Guide 1.99, Revision 2, Paragraph C.l.2.

(3) C VSE y is calculated on the basis of RGl.99, Rev. 2, Position 1. SECY 91-333 states that this procedure is " inadequate." Recognizing this and having provided for the uniqueness of Linde 80 welds 'in BAW-1803, Rev.1, the method for calculating C yUSE in BAW-1803, Rev.1, is put' forward as more representative and is intended to be used for predicting the behavior of these welds in licensing applications.

(4) BAW-2150; Based on TP-3 surveillance material test data.

1 3

f

. TABLE 3. GENERIC LETTER 92-01 RESPONSE: SECTION2,IT{Mb,1(1) ,

Subject:

10CFR50.61 and 10CFR50, Appendix G, III.A; Material Properties Related to PTS and Fracture Toughness Requirements Plant: Turkey Point Unit 3 Column 1 Column 2 Column 3 Column 4 Column 5 C.6 Beltline Unirradiated Charpy Test Results Unirrad. Unirrad. Method of Notes ,

Materials Dropweight RT er Determng Col. 2a Col. 2b Col. 2c Col. 2d Test F RT ,

Results C C C C T, y

10*F 30f[-lb 50ff-lb 35ItLE ft-lb F F F FORGING 122S146 73,52,62 0 +15 +15 +50 +50 NS-2331 (2,4) 97,65,88 123P461 99,63,86 -28 -10 -16 +40 +40 NB-2331 (1,2,4) 18,83,84 123S266 88,38,87 -48 -27 -31 +30 +30 NB-2331 (1,2,4) 93,130,89 WELD SA-1484 40,52,41 ND ND ND ND -5 Est. (3) (1)

SA-1101 45,45,46 ND +70 ND -70 +10 NB-2331 (1,5,6)

SA-ll35 56,44,55 ND ND ND ND -5 Est. (3) (1)

NOTES FOR TABLE 3 ARE ON THE FOLLOWING PAGli.

TABLE 3 (CONTINUED)

NOTES TO TABLE 3:

(1) BAW-2150 (2) Supplier Test Report (3) BAW-1803, Revision 1, Tables 3-1 and 3-2; mean of RT., values for 38 Linde 80 welds.

(4) Values are for 40 hr stress-relief.

(5) Cy (+10F) values are for 8 - 6 hr cycles stress relief.

(6) EPRI NP-373;yC 50 ft-lb, Drop Weight, and RT., values.

i TABLE 4. GENERIC _ LETTER 92-01 RESPONSE: SECTION 2, ITEM b, 1 (2)

Subject:

10CFR50.61 and 10CFR50, Appendix G, Ill.A; Material Properties Related to PTS and Fracture Toughness Requirements -- APPLICABLE ONLY TO REACTOR VESSELS CONSTRUCTED TO AN ASME CODE EARLIER THAN THE 1971 EDITION SUMMER 1972 ADDENDA Plant: Turkey Point Unit 3 I Column 2 Col. 3 Column 1 Heat Treatment Notes Material l

~

BELTLINE l MATERIALS (1,2) 122S146VA] 1550F-11h/WQ; 1220F-22h/AC; 1125F-11h (min)/FC l

123P461VA) 1550F-13h/WQ; 1210F-18h/AC; ll25F-9%h (min)/FC 123S266 val 1550F-13h/WQ; 1210F-18h/AC; ll25F-9%h (min)/FC SA-1484 ll25F-9%h (min)/FC SA-1101 ll25F-9)h (min)/FC SA-1135 1125F-9\h (min)/FC SURVEILLANCE MATERIALS 123P461 val 1550F-13h/WQ; 1210F-8h/AC; ll25F-10%h/FC (1) 123S266 val 1550F-13h/WQ; 1210F-8h/AC; ll25F-10%h/FC j

SA-1101 ll25F-10\h/FC NOTES:

(1) 8AW-2150 (2) Additional stress relief information per Mt. Vernon fabrication process sheets.

(3) WQ - water quench AC - air cool FC - furnace cool

.. _ _ _ _ . _ _ _ _ , . . . _ . . _ _ _ _ _ . _ _ . _ _ _ _ . . . _ . . _ _ - _ . _ _ . _ . . . _ _ _ . . _ . . _ _ . _ _ . _ _ _ _ _ . . _ . _ _ _ ~ _ _ . . . . _ . . _ . _ -

R TABLE 5. GENERIC LETTER 92-01-RESPONSE: SECTION 2, ITEM b, 1 (3) l

Subject:

10CFR50.61 and 10CFR50, Appendix G, III.A; Material Properties Related to PTS and Fracture

' Toughness Requirements -- APPLICABLE ONLY TO REACTOR VESSELS CONSTRUCTED TO AN ASME CODE' i

! EARLIER THAN THE 1971 EDITION, SUMMER 1972 ADDENDA ,

Plant: Turkey Point Unit 3 Column 1 Column 2 Column 3 Column 4. Column 5' C. 6 ,

Beltline Heat Beltline Weld Wire Weld Flux Notes  !

Plate or Number Weld Heat- Lot  ;

Forging _. -!

(1,2)

NB Forging 122S146 NB to IS Circ.: SA-1484 72442 8579 IS' Forging 123P461 IS to LS Circ.: SA-1101 71249 8445

- LS Forging 123S266 LS to Dutch Circ.: SA-ll35 61782 8457 NOTES: (1) BAW-2150 .

(2) Mt. Vernon fabrication process sheets ,

(3) .NB - Nozzle Belt t IS Intermediate Shell LS - Lower Shell f

f I 1

t 4.

4.

l. t

-i i

k i

I

TABLE 6. GENERIC LETTER 92-01 RESPONSE: SEC}.pN2,ITEMb,T(4)

Subject:

10CFR50.61 and 10CFR50, Appendix G, Ill.A; Material /roperties Related to PTS and Fracture Toughness Requirements -- APPLICABLE ONLY TO REACTOR VESSELS CONSTRUCTED TO AN ASME CODE EARLIER THAN THE 1971 EDITION, SUMMER 1972 ADDENDA Plant: Turkey Point Unit 3 Column 1 Column 2 Column 3 Column 4 Column 5 Surveillance Surveillance Weld Wire Wa!d Flux Notes Plate or Weld Heat Lot forging Heat Number _

123P461 SA-Il01 71249 8445 (1) 123S266 NOTES: (1) BAW-2150

TABLE 7. GENERIC LETTER 92-01 RESPONSE: SECTION 2, ITEM b, 1 (5)

Subject:

10CFR50.61 and 10CFR50, Appendix G, Ill.A; Material Properties Related to PTS and Fracture Toughness Requirements -- APPLICABLE ONLY TO REACTOR VESSELS CONSTRUCTED TO AN ASME CODE EARLIER THAN THE 1971 EDITION, SUMMER 1972 AUDENDA Plant: Turkey Point Unit 3 Column 1 Column 2 C. 3 Material Chemical Com)osition, Weight Percent Notes C Mn P S Si Cr Ni Mo Cu BELTLINE MATERIALS 122S146 val 0.22 0.64 0.010 0.013 0.25 0.34 0.68 0.58 ND (1) 123P461 val 0.20 0.64 0.010 0.010 0.26 0.40 0.70 0.62 0.06 (2,3) 123S266 val 0.20 0.62 0.010 0.008 0.20 0.38 0.67 0.58 0.08 (2,3)

SA-1484 0.08 1.52 0.018 0.015 0.42 0.09 0.60 0.39 0.24 (4)

SA-1101 0.07 1.28 0.021 0.014 0.52 0.16 0.60 0.37 0.26 (4)

SA-1135 0.08 1.45 0.011 0.013 0.49 0.08 0.54 0.38 0.25 (4)

SURVEILLANCE MATERIALS 125P461VA1 0.20 0.64 0.010 0.010 0.26 0.40 0.70 0.62 0.06 (5) 123S266 val 0.20 0.62 0.010 0.008 0.20 0.38 0.67 0.58 0.08 (5)

SA-1101 0.08 1.51 0.020 0.013 0.57 0.16 0.60 0.37 0.26 (5)

REQUIRED: State heat number of weld wires used for determining ibove c!.emical composition if different from that in 1 (3). -- Not applicable --

NOTES:

(1) Supplier Material Test Report (2) BAW-2150 (3) Copper content based on surveillance material data.

(4) BAW-2121P (5) BAW-1543, Revision 3

!l l i ilj!li' <l l l s

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i r 5i r y e 2 e e e p 5). l n k m < b e r e m a  :

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t e o, p n c : l t E p e e t S A r j n .d T Uy e b a l t f u l o fC o e __

S P C I( N R _

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TABLE 9. GENERIC IETTER 92-01 RESPONSE: SECTION 3. ITEM b l

Subject:

Generic Letter 88-11 Response Commitments; Utilization of Surveillance Results Plant: Turkey Point Unit 3 Were er veillance results used in determining C USE? Yes o No / __

Were surveillance results used in determining RT,,.? Yes / No o If any "yes" boxes were checked above, state how the surveillance resu?ts were used:

Turkey Point Units 3 and 4 - Issuance of Amendments RE: Pressure and Temperature (P/T) Limits (TAC Nos. 69390 and 69391).

References:

Letter to W. F. Conway from G. E. Edison dated January 10, 1989.

1 L_ .

i TABLE 10. GENERIC LETTER 92-01 RESPONSE: SECTION 3 ITEM c i

Subject:

Generic Letter 88-11 Response Commitments; Difference Between Measured and Predicted

. (Regulatory Gcide 1.99, Revision 2) Embrittlement Effects Plant: Turkey Point Unit 3 Question I. Does measured ART , exceed ART , + 20 predicted by Regulatory Guide 1.99, Revision 2? j Question II. Does measured yC USE drop exceed that obtained from Regulaiory Guin 1.99, Revision 2, Figure 2?

j' Column I Column 2 Column 3 Column 4 Column 5 Column 6 Column 7 Beltline Fluenge Measured Predicted Question I Measured Predicted Ouestion II Material n/cm ART , ART ,+2a If "yes" CyUSE C,USE If "yes" seo' Note (6) Drop Drop see Note (6) 122S146 '--- ND ND --

ND ND --

0(1,5) 123P461 7.01E+18(4) 67 No 0(1) 20 No i

1.41E+19(2) 23(2,i,) 75 No 17(2) 24 No

, 123S266 1.41E+19(2) 45(2,5) 90 No 32(2) 29 Yes 1.23E+19(4) 45(3) 88 No 0(3) 28 No SA-1484 ---

ND ND --

ND ND --

l SA-1101' 7.01E+18(4) 164(4) 195 No 4(4) 21 No 1.23E+19(4) 178(4) 220 No 18(4) 24 No ,

SA-1135 1.03E419(4) 142(4) 240 No l 21(4) 31(7) No t ,

1 j NOTES FOR TABLE 10 ARE ON THE FOLLOWING PAGE.

l' f

i

-- - . - = . . - . . . _ - . - . . . . . - . _ -

l l

TABLE 10 (CONTINUED) h NOTES: (1) WCAP-8631 (2) SWRI 02-5131 and SWRI 02-5380 (3) SWRI 06-8575 (4) BAW-1803, Revision 1 (5) 50 ft-lb transition temperature (6) The only instance where a measured " shift" or " drop" exceeds that predicted by calculation i performed in accordance with Regulatory Guide I.99, Revision 2, is for a " drop" for base ,

metal. .ne requirements of 10CFR50, Appendix G, were not violated, and there being no  ;

further application of the " drop" data, the effect of these surveillance results are i therefore not significant. .

(7) BAW-1920P t

i I

l _, ~ .

i TABLE 1. GENERIC LETTER 92-01 RESPONSE: SECTION 1

Subject:

10CFR50, Appendix H; Adherence to RVSP Regtirements I Plant: Turkey Point' Unit 4 t

Question I: Does RVSP meet ASTM E 185-73, E 185-79, or E 185-82? Yes a No /

1 Question II: Is plant one of the following? ANO-I, Crystal River-3, Davis Besse, R. E. Ginna, Oconee-1, Oconee-2, Oconee-3, Point Beach-1, Point Beach-2, Rancho Seco, Surry-1, Surry-2, Turkey Point-3, lurkey Point-4, Zion-l. Zion-2. Yes / No a IF ANSWER IS "YES" TO EITHEk QUESTION I OR QUESTION II, PROCEED TO TABLE 2.

I l

IF ANSWER IS "NO" TO BOTH QUESTION I AND QUESTION II, PROCEED TO QUESTION III AND QUE3' ION IV. l 3 Question III: If plan is to revise RVSP to meet requirements of 10CFR50, Appendix H, den will ,

i' revised RVSP be submitted to NRC?

Response

Not applicable (see Question I and II above)

! Question IV: If plan is not to revise RVSP to meet requirements of 10CFR50, Appendix H, when will  ;

exemption from 10CFR50, Appendix H be requested from NRC? _

~

l Response:

l Not applicable (see Question I and II above) i

, NOTES: WCAP-7660: Surveillance Program Description (ASiM E 165-66) 4 I

4 2

f i

i i

j -

TABLE 2. GENERIC LETTER 92-01 RESPONSE: SECTION 2. ITEM a i

Subject:

10CFR50, Appendix G C OSEy Requirements 4

Plant: Turkey Point Unit 4 Column 1 Column 2 Column 3 Column 4 Limiting Initial EFPY to reach If Column 2 is within license Action taken Material USE C,USE<50 ft-lb period: C USE at indicated time per IV.A.1 .

ft-lb i

. Column 3A Column 3B 12/16/91 EOL LIMITING An analysis which BELTLINE WELD demonstrates that this material provides margin SA-1101 65 (4) 2, approx. 40 36 of safety against fracture ,

equivalent to that  !

required by ASME Section III, Appendix G, has been performed under the j sponsorship of the B&W i Owners Group's Reactor Vessel Working Group and ,

submitted to the NRC as i

j report BAW-2118P. '

I LIMITING i- BELTLINE PLATE i OR FORGING l22S180 132 (4) { >32 NA i NA NA NOTES FOR' TABLE 2 ARE ON THE FOLLOWING PAGE.

.._...___..m_..__

t t

l TABLE 2 (CONTINUED)

I' NOTES: (1) Fluence values taken at k.-thickness. .

(2) C 0SE v values for 12/16/91 and EOL (Column 3) calculated per requirements outlined in Regulatory Guide 1.99, Revision 2, Paragraph C.I.2.

i (3) C USE y 'is calculated on the basis of RGI.99, Rev. 2, Position 1. SECY 91-333 states that this  !

4 procedure is " inadequate." Recognizing this and having provided for the uniqueness of Linde 80 i i welds in BAW-1803, Rev. I, the method for calculating C,USE in BAW-1803, Rev.1, is put forward as more representative and is intended to be used for predicting the behavior of these welds in 1icensing applications.-

1 1

(4) BAW-2150; Based on TP-3 surveillance test data.

. i a

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TABLE 3. GENERIC LETTER 92-01 RESPONSE: SECTION 2, ITEM b, 1 (1) v

Subject:

10CFR50.61 and 10CFR50, Appendix G, III. A; Material Properties Related to PTS and Fracture Toughness Requirements Plant: Turkey Point Unit 4 Columt, !  ! Column 2 Column 3 Column 4 Column 5 C.6 Beltline Unirradiated Charpy Test Results Unirrad. Unirrad. Hethod of Notes Materials Oropweight RT, Detereng

. Col. 2a Col. 2b Col. 2c rol. 2d Test F RT, Results ,

T, C, C C Cy y 10 F 30ff-lb 50ff-lb 35 MLE i t

ft-lb F F F FORGING t 1245309 80,101,89 0 +20 +18 440 +40 NB-2331 (2,5) 49,74,88,64 123P481 44,62,28 +15 +45 +40 +50 +50 NB-2331 (1,2,5) 30,58,46 1 1235180 91,59,64 -37 -15 -25 +40 +40 NB-2331 (1,2,5) +

62,56,53 iWELD WF-67 29,35,30. ND ND ND ND -5 Est. (3) (1,6)

WF-70 39,35,44 ND ND ND ND +18 Eval.{4) (1,6,9) .

SA-Il01 45,45,46 ND +70 ND -70 +10 NB-2331 (1,7,8)  !

! SA-ll35 56,44,55 ND NO ND ND -5 Est. (3) (1)

NOTES FOR TABLE 3 ARE ON THE FOLLOWING PAGE.

, t

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f

TABLE 3 (CONTIMJED)

NOTES TO TABLE 3:

(1) BAW-2150 -

(2) Supplier Test Report .

(3) -BAW-1803, Revision I, Tables 3-1 and 3-2; mean of RT , values for 34 Linde 80 welds.

(4) BAW-2100 (5) Values are for 40 hr stress-relief.

(6)' Cy (+10F) values are for 48 hr stress-relief.

'(7) Cy (+10F) values are for 8. - 6 hr cycles stress-relief.

(8)- EPRI NP-373; yC 50 ft-lb, Drop Weight, and RT,,, values.

(9) RT ,' value for 40 hr stress-relief maximum.

b

I f

TABLE 4. GENERIC LETTER 92-01 RESPONSE: SECTION 2, ITEM b. T (2)

Subject:

10CFR50.61 and 10CFR50, Appendix G, llI.A; Material Properties Related to PTS and Fracture Toughness Requirements -- APPLICABLE ONLY TO REACTOR VESSELS CONSTRUCTED TO AN ASME CODE EARLIER THAN THE 1971 EDITION, SUMMER 1972 ADDENDA Plant: Turkey Point Unit 4 5 Column 1 Column 2 Col. 3 Material Heat Treatment Notes i

j BEL (LINE MATERIALS 1245309 val 1550F-15h/WQ; 1220F-22h/FC; ll25F-10th/FC (1,2) 123P481VA1 1550F-10%h/WQ; 1210F-18h/AC; ll25F-10%h (min)/r0 4 122S180 val 1550F-10%h/WQ: 1200F-lCh/FC; 1125F-10%h (min)/FC j WF-67/WF-70 ll25F-9th (min)/FC

SA-1101 Il25F-10%h (min)/FC j SA-Il35 1125F-10\h (min)/FC SURVEILLANCE MATERIALS l

.'3 3481 val 1550F-104h/WQ; 1200F-18h/AC; ll25F-10)h/FC (1) i 122S180 val 1550F-10\h/WQ; 1210F-18h/AC; ll25F-10)h/FC ,

SA-1094 Il25F-10th/FC  !

NOTES:

i (1) BAW-2150 (2) Additional stress relief information per Mt. Vernon fabrication process sheets.

(3) WQ - water quench AC - air coal FC - furnace cool i

i 1

"m - - __- --. _ _ _, _- __s -

_m

TABLE 5. GENERIC LETTER 92-01 RESP 0flSE: SECTION 2. ITEM b, 1 (3)

Subject:

10CFR50.61 and 10CFR50, Appendix G, III.A; Material Properties Related to PTS and Fracture Toughness Requirements -- APPLICABLE ONLY TO iEACTOR VESSELS CONSTRUCTED TO AN ASME CODE EARLIER THAN THE 1971 EDITION, SUMMER 1972 A'JiENDA Plant: Turkey Point Unit 4 Column 1 Column 2 Column 3 Column 4 Column 5 C. 6 Beltline Heat Beltline Weld Wire Weld Flux Notes Plate or Number Weld Heat Lot Forging NB Forging 1245309 NB to IS Circ.(ID 67%): WF-67 72442 8669  !(1,2)

IS Forging 123P481 NB to IS Circ.(00 33%): WF-70 72105 8669 i LS Forging 123S180 IS to LS Circ.: SA-Il01 71249 8445 LS to Dutch Circ.: SA-ll35 61782 8457 NOTES: (1) BAW-2150 (2) Mt. Vernon fabrication process sheets (3) NB - Nozzle Belt IS - Intermediate Shell LS - Lower Shell 4

'l i' I r

3-  !

t L i

l TABLE 6. GENERIC LETT:R 92-01 RESPONSE: SECTION 2 ITEM b 1 (4)  :

t

Subject:

10CFR50.61 and '.0CFR50, Appendix G, III.A; Material Properties Related to PTS and Fracture Toughness Requirements -- APPLICABLE ONLY TO REACTOR VESSELS CONSTRUCTED TO AN ASME CODE i EARLIER THAN THE 1971 EDITION, SUMMER 1972 ADDENDA Phnt: Turkey Point Unit 4 j Column I Column 2 Column 3 Column 4 Colurrn S L Surveillance Surveillance Weld Wire Weld Flux Notes [

Plate or Weld Heat Lot  !

l Forging  !

l' Heat Number  ;

i I i

, 123P481 SA-1094 71249 8457 (1)  ;

. 1225180 ,

j I ' .

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e i  !

l u NOTESh (1) BAW-2150  !

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i

. . . ___ ~___ - _ , - - __ . . - _ _ . - . . __ . . . _ . _ _ , _ _ _ ..._

i i

i l TABLE 7. GENERIC LETTER 92-01 CSPONSE: SECTION 2. ITEM b. 1 (5) l

Subject:

10CFR50.GI and 10CFR53, Appendix G, III.A; Material Propertie: Related to PTS and Fracture ,

Toughness Requirements -- APPLICABLE ONLY TO REACTOR VESSELS CONSTRUCTED TO AN ASME CODE EARLIER lHAN THE 1971 EDITION, SUMMER 1972 ADDENDA Plant: Turkey Point Unit 4 Column 1 Column 2 C. 3 Material Chemical Composition, Weight Percent Notes 1 C Mn P S Si Cr Ni Mo Cu BELTLINE MATERIALS

, 1245309 val 0.20 0.60 0.010 0.012 0.26 0.33 0.70 0.56 ND (1) 123P481 val 0.20 0.65 0.010 0.010 0.24 0.32 0.68 0.59 0.05 (2) 122S180 val 0.22 0.60 ,0.010 0.009 0.22 0.34 0.74 0.60 0.06 (2)

WF-70 0.09 1.63 0.018 0.009 0.54 0.10 0.59 0.40 0.35 (3) i WF-67 0.08 1.55 0.021 0.016 0.58 0.09 0.60 0.39 0.24 (3)

SA-1101 0.07 1.28 0.021 0.014 0.52 0.16 0.60 0.37 0.26 (3)

SA-ll35 0.08 1.45 0.011 0.013 0.49 0.08 0.54 0.38 0.25 (3)

SURVEILLANCE MATERIALS 123P481 val 0.22 0.67 0.010 0.009 0.20 0.33 0.71 0.56 0.05 (4) 122SIPOVA! 0.21 0.67 0.011 0.009 0.23 0.31 0.70 0.56 0.06 (4)

SA-1094 0.10 1.44 0.014 0.011 0.50 0.14 0.60 0.36 0.26 (4)

REQUIRED: State heat number of weld wires used for determining above chemical composition if different f rom that in 1 (3). -- Not applicable --

NOTES:

(1) Supplier Material Test Report.

(2) BAW-2150 1

(3) BAW-2121P (4) BA1-1543, Revision 3

't

4

. I 4 . TABLE 8. GENERIC LETTER 92-01 RESPONSE: SECTION 3, ITEM a

Subject:

Generic letter 88-11 Response Commitments; Effect of Irradiation Temperatt.re Plant: Turkey Point Unit 4 Cold Leg Temperature (T,,,,):

546 F (See Figure 4-6) t If T i is <525 F, state how this was considered in determination of embrittlement effects (Cy VSE,o RT ,) in accordance with Regulatory Guide 1.99, Revision 2.

. l l

Not applicable  ;

i

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i 4

6

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i i  !

4

References:

, i j None i i i

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_ ~ _ _ . - - ~ . _. '

1 l

TABLE 9. GENERIC LETTER 92-01 RESPONSE: SECTION 3. ITEM b i

(

Subject:

Generic Letter 88-11 Response Comitments; Utilization of Surveillance r

Results i

Plant: Turkey Point Unit 4 Were surveillance results used in determining C USE? Yes a No /

1 Were surveillance results used in determining RTm? Yes / ho o j If any "yes" boxes were checked above, state how the surveillance results were used:

. t

Turkey Point Units 3 and 4 - Issuance of Amedendments RE

Pressure and Temperature (P/T) Limi's l

(TAC Nos. 69390 and 69391)

References:

Letter to W. F. Conway frore G. E. Edison dated January 10, 1989.

i 4

i 6

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TABLE 10. GENERIC LETTER 92-01 RESPONSE: SECTION 3, ITEM c 7

Subject:

Generic Letter 88-11 Response Commitments; Difference Between Measured and Predicted (Regulatory Guide 1.99, Revision 2) Embrittlement Ef fects _

Plant

Turkey Point Unit 4 Question I. Does measured ART , exceed ART , + 2a predicted by Regulatory Guide 1.99, t

. Revision 2? i i i

Questian II. Does measured C,USE drop exceed that obtained from Regulatory Guide 1.99, Revision 2, figure 2?

l+

t l

Column I Column 2 Column 3 Column 4 Column 5 Column _6 Column 7

• Beltline Fluence Measured Predicted Question I Measured Predicted Question 11 2

Material n/cm ARTer ART ,+2a If "yes* C,USE C,USE If "yes" r see Note (5) Drop Drop see Note (5) -

1245309 ---

ND ND --

ND ND --

123P481 1.25E+19(1). 35(1,4) 66 No 12(1). 20 No 123S180 '7.54E+18(3) 10(2,4) 68 No 0(2) 18 No ,

5 1.25E+19(1) 11(1,4) 73 No 10(1) El No WF-67 6.09E+18(3) 160(3) 200 He 15(3) 23(6) No WF-70 6.63E418(3) 135(3) 259 No 13(3) 22(7) No j SA-Il01 7.01E+18(3) 164(3) 195 No 4(3) 21 No i 1.23E+19(3) 178(3) 220 No 18(3) 24 No j SA-1135 1.03E+19(3) 142(3) 240 No 21(3) 31(7) No i'

NOTES FOR TABLE 10 ARE ON THE FOLLOWING PAGE.

i i

i 5

- ___.._ ._. __.= _ .._ _ __.. _ ._._.-.__. _ . . _ . _ __ _ .--..____. ___ _- -

TABLE 10 (CONTINUED) i NOTES: (1) SWRI 02-5131 and SWRI G2-5380 4- - (2) SWRI 02-4221 (3) BAW-1803, Revision 1 i 3

(4) 50 ft-lb transition ten.perature  :

(5) Statement not required.

(6) BAW-1910P ,

(7) BAW-1920P 1

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1 L_. _ TABLE 1. GENERIC LETTER 92-01 RESPONSE: SECTION 1 1

Subject:

10CFR50, Appendix H: Adherence to RVSP Requirements Plant: Zion Unit 1 Question I: Does RVSP meet ASTM E 185-73, E 185-79 or E 185-82? Yes a No /

l Question II- Is plant one of the following? ANO ., Crystal River-3, Davis Besse, R. E. Ginna, Oconee-1, Oconee-2, Oconee-3, Point Beach-1, Point Beach-2, Rancho Seco, Surry-I,

, Surry-2, Turkey Point-3, Turkey Point-4, Zion-1, Zion-2. Yes / No o l IF ANSWER IS "YES" TO EITHER QUESTION I OR OUESTION II, PROCEED TO TABLE 2.

IF ANSWER IS *N0" TO BOTH QUESTION I AND QUESTION II, PROCEED TO QUESTION III AND QUESTION IV.

! Question III: If plan is to revise RVSP to meet requirements of 10CFR50, Appendix H, when will ,

revised RVSP be submitted to NRC?

P sponse:

I i I

Not applicable (see Question I and II above)

~

Question IV: If plan is not to revise RVSP to meet requirements of 10CFR50, Appendix H, when will exemption from 10CFR50, Appendix H be requested from NRC?

i i i Response:

j Not applicable (see Question I and II above)

NOTFS: WCAP-8064: Surveillance Program Description

(ASTM E 185-70) 1 4

4 i

I l

1

! TABLE 2. GENERIC LETTER 92-01 RESPONSE: SECTION 2, ITEM a j

Subject:

10CFR50, Appendix G, C,USE Requirements l I

Plant: Zion Unit I I

i Column I Column 2 Column 3 Column 4 Limiting Initial EFPY to reach If Column 2 is within license Action taken Material USE Cy USE<50 f t-lb period: C USE at indicated time per IV.A.1 ft-lb Column 3A Column 3B

! 12/16/91 EOL ,

l LIMITING An analysis which BELTLINE VELD demonstrates that this material provides margin i j WF-70 70 (4) 2, approx. 44 40 of safety against fracture ,

, equivalent to that l required by ASME Section  ;

i III, Appendix G, has been

- performed under the t 4 sponsorship of the B&W

, Owners Group's Reactor Vessel Working Group and submitted to the NRC as report BAW-2148P.  :

LIMITING BELTLINE PLATE

,- OR FORGING l

B7823-1 91 (5) >32 NA NA NA NOTES FOR TABLE 2 ARE ON THE FOLLOWING PAGE.

j  !

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4 i

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TABLE 2 (CONTINUED)

)

NOTES: (1) Fluence values taken at L.-thickness.

i (2) C,USE values for 12/16/91 and EOL (Column 3) calculated per requirements outlined in Regulatory i

Guide 1.99, Revision 2, Paragraph C.1.2. I (3) C yUSE is calculated on the basis of RG1.99, Rev. 2, Position I. SECY 91-333 states that this

i. procedure is " inadequate " Recognizing this and having provided for the uniqueness of Linde 80  ;

welds in BAW-1803, Rev.1, the method for calculating C,USE in BAW-1803, Rev.1, is put forward as i

, more representative and is intended to be used for predicting the behavior of these welds in 4 2

licensing applications. .

(4) BAW-1803 [

l (5) BAW-10046P l ~' i i .

I i

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4 o

4 i

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i.  ;

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i L

TABLE 3. GENERIC LETTER 92-01 RESPONSE: SECTION 2, ITEM b, 1 (1)

Subject:

10CFR50.61 and 10CFR50, Appendix G, III.A; Material Properties Related to -

PTS and Fracture Toughness Requirements  !

Plant: Zion Unit 1 Column 1 Column 2 Column 3 Column 4 Column 5 C.6 Beltline Unirradiated Charpy Test Results Unirrad. Unirrad. Method of Notes Materials Dropweight RT*, Determng j Col. 2a Col. 2b Col. 2c Col. 2d Test F RT, .

Results '

C C C T, i 10*F 30fl-lb 50 l-lb 35 MLE l ft-lb F F F FORGING

+

ANA 102 ND +15 +45 +35 +20 +20 NB-2331 (2)

PLATE C3795-2 42,44,39 -10 +25 +15 -10 -10 NB-2331 (1,2,5)  !

B7835-1 ND +15 +32 +25 -20 -20 NB-2331 (1,2,5) l C3799-2 40,35,43 -2 +33 +25 -20 -20 NB-2331 (1,2,5) l B7823-1 27,40,33 +5 +27 +20 -20 -20 NB-2331 (I.2,5) l WELD NF-154 41,37,43 ND ND ND ND -5 Est. (3) (1,6) l SA-1769 36,35,38 ND ND ND ND -5 Est. (3) (1,7)

WF-70 39,35,44 ND ND ND ND +18 Eval.(4) (1,6,9) '

WF-4 40,31,34 ND N9 ND ND -5 Est. (3) (I,8)

. WF-8 45,38,30 ND ND ND .ND -5 Est. (3) (1,6) l t

j' NOTES FOR TABLE 3 ARE ON THE FOLLOWING PAGE. [

[ .

i i

i

TABLE 3 (CONTINUED)

NOTES TO TABLE 3:

(1) BAW-2150 (2) Mt. Vernon Qualification Test Paport (3) 9AW-1803, Revision 1, Tables 3-1 and 3-2; mean of RT, values for 34 Linde 80 welds.

(4) BAW-2100 (5) Values are for 60 hr stress-relief. ,

(6) Cy (+10F) values are for 48 hr stress-relief.

(7) Cy (+10F) values are for 8 - 6 hr cycles stress-relief.

(8) Cy (+10F) values are for 80 hr stress-relief.

(9) RT, value for 40 hr stress-relief maximum.

l l

f i

L

TABLE 4. GENERIC LETTER 92-01 RESPONSE: SECTION 2 ITEM b, 1 (2)

Subject:

10CFR50.61 and 10CFR50, Appendix G, Ill.A; Material Properties Related to PTS and Fracture Toughness Requirements -- APPLICABLE ONLY TO REACTOR VESSELS CONSTRUCTED TO AN ASME CODE EARLIER THAN THE 1971 EDITION, SUMMER 1972 ADDENDA Plant: Zion Unit 1 Column I Column 2 Col. 3 Material Heat Treatment Notes BELTLINE MATERIALS ANA 102 Not available (1,2)

C3795-2 1600-1650F-9%h/BQ; 1200-1225F-9%h/BQ; 1125F-26th/FC (cumul.)

B7835-1 1600-1650F-9%h/BQ: 1200-1225F-9%h/BQ; ll25F-26th/FC (cumul.)

C3799-2 1600-1650F-9%h/BQ; 1200-1225F-9%h/BQ; 1125F-23%h/FC (cumul.)

B7823-1 1600-1650F-9%h/BQ; 1200-1225F-9%h/BQ; ll25F-23%h/FC (cumul.)

WF-154/SA-1769 1100-Il50F-18%h/FC (cumul.)

WF-70 1100-1150F-23h/FC (cumul.)

WF-154 Il00-Il50F-18%h/FC (cumul.)

WF-4/WF-8 1100-Il50F-26(h/FC (cumul.)

WF-8 Il00-Il50F-23%h/FC (cumul.)

SURVEILLANCE MATERIALS B7835-1 1625F-9%h/BQ; 1212F-9%/h/BQ; ll25F-25h/FC (1)

WF-209-1 Il25F-23h/FC  ;

NOTESi (1) BAW-2150 (2) Additional stress relief information per Mt. Vernon process drawing.

(3) BQ - brine quench FC - furnace cool

TABLE S. GENERIC LETTER 92-01 RESPONSE: SECTION 2. ITEM b, 1 (3)

Subject:

10CFR50.61 and 10CFR50, Appendix G, III.A; Material Properties Related to PTS and Fracture Toughness Requirements -- APPLICABLE ONLY TO REACTOR VESSELS CONSTRUCTED TO AN ASME CODE EARLIER THAN THE 1971 EDITIOL lUMMER 1972 ADDENDA Plant: Zion Unit 1 Column 2 Column 3 Column 4 Column 5 C. 6 Column 1 Beltline Weld Wire Weld Flux Notes Beltline Heat Heat Lot Plate or Number Weld Forging Lower NB Forging ANA 102 NB to IS Circ.(ID 82%): WF-154 406L44 8720 (1,2)

NB to IS Circ.(OD 18%): SA-1769 71249 8738 IS Plate B7835-1 7210S 8669 IS Plate C3795-2 IS to LS Circ.: WF-70 B7823-1 LS to Dutch Circ.: WF-154 406L44 8720 LS Plate 8597 C3799-2 IS Longit.: WF-4 8T1762 LS Plate 8T1762 8632 IS Longit.(ID 39%): WF-8 If 4ongit.(00 61%): WF-4 8TI762 8597 ,

LS Longit.: WF-8 8T1762 8632 l NOTES: (1) BAW-2150 (2) Mt. Vernon process drawing (3) NB - Nozzle Belt IS - Intermediate Shell LS - Lower Shell I

TABLE 6. GENERIC LETTER 92-01 RESPONSE: SECTION 2, ITEM b. T (4)

Subject:

10CFR50.61 and 10CFR50, Appendix G, III.A; Material Properties Related to PTS and Fracture Toughness Requirements -- APPLICABLE ONLY TO REACTOR VESSELS CONSTRUCTED TO AN ASME CODE EARLIER THAN THE 1971 EDITION, SUMMER 1972 ADDENDA Plant: Zion Unit I Column 1 Column 2 Column 3 Column 4 l Column 5 Surveillance Surveillance Weld Wire Weld Flux Notes Plate er Weld Heat Lot Forging Heat Number B7835-1 Wr-209-1 72105 8773 (1) l NOTES: (I) BAW-2150

TABLE 7. GENERIC LETTER 92-01 RESPONSE: SECTION 2, ITEM b, 1 (5)

Subject:

10CFR50.61 and 10CFR50, f.ppendix G, llI.A; Material Properties Related to PTS and Fracture Toughtiess Requirements -- APPLICABLE ONLY TO REACTOR VESSELS CONSTRUCTED TO AN ASME CODE EARLIER THAN THE 1971 EDITION. SUMMER 1972 ADDENDA ._

Plant: Zion Unit 1 Column 1 Column 2 0.3 Material Chemical Composition, Weight Percent Notes C Mn P S Si Cr Ni Mo Cu BELTLINE MATERIALS ANA 102 X 0.74 X 0.012 0.26 0.45 X 0.60 0.06 (1)

C3795-2 I 0.21 1.50 0.010 0.015 0.23 ND 0.49 0.49 0.12 (1,2) 87835-1 3. .T 1.30 0.010 0.011 0.20 ND 0.49 0.47 0.12 (1,2)

C3799-2 OAl 1.33 0.010 0.014 0.24 ND 0.50 0.46 0.15 (1,2)

(1,2) 87823-1 '

• Il

. 1.36 0.013 0.016 0.21 ND 0.48 0.46 0.13 WF-154 0.07 1.54 0.013 0.016 0.42 0.07 0.59 0.40 0.31 (3)

SA-1769 0.09 1.49 0.020 0.014 0.56 0.16 0.61 0.37 0.26 (3)

WF-4 0.07 1.48 0.017 0.011 0.51 0.12 0.55 0.41 0.20 (3)

WF-8 0.06 1.45 0.009 0.009 0.53 0.12 0.55 0.41 0.20 (3)

WF-70 0.09 1.63 0.018 0.009 0.54 0.10 0.59 0.40 0.35 (3)

SURVEILLANCE MATERIt.LS B7835-1 0.20 1.30 0.010 0.011 0.20 ND 0.49 0.47 0.11 (4)

WF-209-1 0.09 1.51 0.020 0.013 0.68 0.06 0.57 0.39 0.35 (4) i REQUIRED: State heat number of weld wires used for determining above chemical composition if dif ferent from that in 1 (3). -- Not applicable -

NOTES:

(1) Supplier Material Test Report (X - chemical contents are not legible)

(2) BAW-2150 (3) BAW-2121P (4) BAW-1543, Revision 3

TABLE 8. GENERIC LETTER 92-01 RESPONSE: SECTION 3, ITEM a

Subject:

Generic Letter 88-11 Response Commitments; Effect of Irradiation Temperature Plant: Zion Unit 1 Cold Leg Temperature (T,m,a): 529.4 F. (See Figure 4-7)

If T i is <525 F, state how this was considered in determination of embrittlement effects (Cy 05E,s RT,) in. accordance with Regulatory Guide 1.99, Revision 2:

Not applicable l

References:

None

i i

TABLE 9. GENERIC LETTER 92-01 RE*,PONSE

SECTION 3, ITEM b

Subject:

Generic Letter 88-11 Response Commitments; Utilization of Surveillance Results '

Plant: Zion Unit I Were surveillance results used in determining C USE? Yes a No /

Were surveillance results used in determining RTm? Yes / No o

! If any "yes" boxes were checked above, state how the surveillance results were used: I j Initial RTg, value for WF.70 weld metal.

1

i

! References-BAW-2100 4  :

i 4

4 b

i i

i 4

TABLE 10. GENERIC LETTER 92-01 RESPONSE: SECTION 3, ITEM c

Subject:

Generic Letter 88-11 Response Commitments; Difference Between Measured and Predicted (Regulatory Guide 1.99, Revision 2) Embrittlement Effects Plant: Zion Unit I _

Question I. Does measured ART , exceed ART., + 20 predicted by Regulatory Guide 1.99, Revision 2?

Question 11. Does measured yC OSE drop exceed that obtained from Regulatory Guide 1.99, Revision 2, Figure 2?

Column I Column 2 Column 3 Column 4 Column 5 Column 6 Column 7 Beltline I Fluen e Measured Predicted Ouestion I Measured Predicted Ouestion 11 ART , ART ,+2a If "yes" C,USE CyUSE If "yes" Material l n/cm

! (1,2,3) see Note (5) Drop Drop see Note (5)

ANA'102 l --

ND' ND --

ND ND --

C3795-2 --- ND ND --

ND ND --

2.53E+18 84 No 3(1) 18(1) No 87835-1 25(1) 8.49E+18 111 No 15(1) 24(1) No 60(1)

,112. No 24(1) 26(1) No 1.26E+19 80(1) 116 No 21(1) 27(1) No 1.56E+19 94(1)

C3799-2 --- ND ND --

ND ND --

87823-1 --

ND ND --

ND. ND --

WF-154 ---

ND ND --

'ND ND --

SA-1769 ---

ND ND --

ND ND --

6.63E+18 242 No 13(2) 22(4) No WF-70 135(2)

WF-4 --- ND ND --

ND NO --

WF-8 --- ND ND --

ND ND --

9(3) 25(3) No Atypical 1.17E+18 28(3) 138 No 216 No l 16(3) 32(3) No 6.56E+18 122(3) 223 No 11(3) 32(3) No 7.50E+18 119(3) 242 No 15(3) 34(3) No 1.08E+19 120(3)

NOTES FOR TABLE 10 ARE ON THE FOLLOWING PAGE.

f I _

~;. .

g__~

I< .

TABLE 10 (CONTIN'JED)

NOTES: (I) Bli -2082 (2) BAW-1803, Revision ]

(3) BAW-2049 (4) BAW-1920P (5) Statement not required.

f f

1 1

l l

l l

t

/

e

4 S. E. Yanichko et 41, " Analysis of Capsule T from the Rochester Gas and Electric

Corporation R. E. Ginna Nuclear Plant Reactor Vessel Radiation Surveillance Program," WCAP-10086, Westinghouse Electric Corporation, Pittsburgh, Pennsylva-nia, April 1982.

S. E. Yanichko et al, " Analysis of Capsule T f. i the Wisconsin Electric Power Company Point Beach Nuclear Plant Unit No. 1 Reactor Vessel Radiation Surveil-

lance Program," WCAP-1073,5, Westinghouse Electric Corporation, Pittsburgh, Pennsylvania, December 1984.

S. E. Yanichko and V. A. Perone, " Analysis of Capsule V from the Virginia Electric and Power Company Surry Unit 1 Reactor Vessel Radiation Surveillance Program," W[AP-11415, Westinghouse Electric Corporation, Pittsburgh, Pennsylva-nia, February 1987.

, , S. E. Yanichko and V. A. Perone, " Analysis of Capsule V from the Virginia Electric &au Power Company Surry Unit 2 Reactor Vessel Radiation Surveillance

^

Prog ~ ram," FCAP-11499, Westinghouse Electric Corporation, Pittsburgh, Pennsylva-nia, June 1987.

S. E. Yanichko et al, " Analysis of Capsule Y from the Commonwealth Edison Company Zion Unit 2 Reactor Vessel Radiation Surveillance Program," " CAP-12396, Westinghouse Electric Corporation, Pittsburgh, Pennsylvania, Septetber .939, i

i 7-E

8. CERTIFICATION 4

This report accurately responds to the request for information stated in Generic Letter 92-01.

7/Xb)[

H. Jf DeVan, Engineer 11 6 //6/92 Date Materials and Structural Analysis Unit i b ' VA fa. E. Lowe, Jr., Aqffsory Engineer' Date Materials and Structural An41ysis Unit This report was reviewed and found to be a curate.

LB w d i d ,-

L. B/. Gross, Advisory Engin6er Date Matdria; and Structural Analysis Unit Verification of independent review.

WD b>!lbl90,

'K. E. Moore, Manager ' ate D

Materials and Structural Analysis Unit l

1 This report is approved for release.

&flC l92.

D. L. Howell, Project Manager Date Owners Group Engineering Servicts i 1

8-1 l s  !

TABtE 1. GENERIC LETTER 92-01 RESPONSE: SECTION 1

Subject:

10CFR50, Appendix H; Adherence to RVSP Requirements Plant: Zion Unit 2 Question I: Does RVSP meet ASTH 185-73, E 185-79, or E 185-827 Yes o No /

Question II: Is plant one of the following? ANO-1, Crystal River-3, Davis Besse, R. E. Ginna, .

Oconee-1, Oconee-2, Oconee-3, Point Beach-1, Point Beach-2, Rancho Seco, Surry-1, l Surry-2, Turkey Point-3, Turkey Point-4, Zion-1, Zion-2. Yes / No o IF ANSWER IS "YES" TO EITHER QUESTION I OR QUESTION II, PROCEED TO TABLE 2.

IF ANSWER IS "N0" TO BOTH QUESTION I AND QUESTION II, PROCEED TO QUESTION III AND QUESTION IV.

Question III: If plan is to revise RVSP to meet requirements of 10CFR50, Appendix H, when will revised RVSP be submitted to NRC?

. Response:

Not applicable (see Question I and II above)

Question IV: If plan is not to revise RVSP to meet requirements of 10CFR50. Appendix H, when will exemption from 10CFR50, Appendix H be requested froo NRC?

Response

Not applicable (see Question I and II above) i

, NOTES: WCAP-8132: Surveillance Program Description (ASTM E 185-70) i I

TABLE 2. GENERIC LETTER 92-01 RESPONSL: SECTION 2, ITEM a

Subject:

10CFR50, Appendix G, C,USE Requirements Plant: Zion Unit 2 ,

Column 1 Column 2 Column 3 Column 4 Limiting Initial EFPY to reach if Column 2 is within license Action taken Material USE C,Ud2<50 f t-lb period: C,05E at indicated time per IV.A.1 ft-lb Column 3A Column 3B 12/16/91 EOL LIMITING An analysis which BELTLINE WELD demonstrates that this material provides margin SA-1769 70 (5) 7, approx. 48 42 of safety against fracture

, equivalent to that required by ASME Section III, Appendix G, has been gerformed under the sponsorship of the B&W Owners Group's Reactor Vessel Working Group and submitted to the NRC as report BAW-2148P.

LIMITi 3 BELTLINE PLATE OR FORGING 88006-1 91 (6) >32 NA NA NA C4007-1 91 (6) >32 NA NA NA 88029-1 91 (6) ! >32 NA NA NA NOTES FOR TABLE 2 ARE ON THE FOLLOWING Pf3E.

i TABLE 2 (CONTINUED) ,

NOTES: (1) Fluence values taken at \-thickness.

(2) C USE y values for 12/16/91 and E0L (Column 3) calculated per requirements outlined in Regulatory Guide 1.99, Fevision 2, Paragraph C.I.2.

(3) C yOSE is calculated on the basis of RGl.99,1ev. 2, Posit an 1. SECY 91-333 states that this procedure is " inadequate." Recognizing this and having provided for the uniqueness of Linde 80 welds in BAW-1803, Rev.1, the method for calculating C yUSE in BAW-1803, Rev.1, is put forward as more representative and is intended to be used for predicting the behavior of these welds in licensing applications.

(4) The analysis was based on the worst case weld chemical composition (WF-70) combined with peak ,

fluences seen in the Zion vessels (Units 1 an_d 2), and is a conservative analysis.  !

(5) BAW-1803 (6)' BAW-10046P l

1 i

.[

t 4

i n

1

_ _ _ _ _ _ _ _ _ _ - _ - - _ _ _ _ _ - - _______mm a- -_ .

r 4

i TABLE 3. GENERIC LETTER 92-01 RESPONSE: SECTION 2, ITEM b, 1 (1)

Subject:

10CFR50.61 and 10CFR50, Appendix G, III.A; Material Properties Related to l PTS and Fracture Toughness Requirements

. Plant: Zion Unit 2-  ;

a i  !

~

I Column 1 Column 2 Column 3 Column 4 Column 5 C.6 i

[ Beltline Unirradiated Charpy Test Resu1M Unirrad. Unirrad. Method of Notes-Materials ---

Dropweight RT , Determng i Col. 2a Col. 2b Col. 2c Cel. 2d Test f RT, 'i

.Results '!

C C C Cy T, j

10,F .30f[-lb 50f[-lb 35 MLE i j ft-lb F F F -

2 FORGING l

ZV-3855 42,49,32 -7 +10 +25 +10 +10 NB-2331 (1,2,5) 8 i

! 72,34,32 i-PLATE l- 88006-1 38,36,26 +5 -+27 +20 +10 510 NB-2331- (1,2,5)  !

32,38,35 08040-1 36,64,38 -5 +25 +15' -10' -10 NB-2331 (1,2,5)  !

C4007 ND +30 +68 +65' +10 +10 NB-2331 (1,2,5) i B8029-1 -ND +12 +35 +35, -10 -

-10 NB-2331 (1,2,5) l t

WELD  :

WF-200 36,35,26 ND ND ND ND -5 Est. (3) .(1,6)  !

SA-1769 36,35,38  ! ND ND. ND ND -5 Est.-(3) (1,7) l WE-154 41,37,43 NO ND HD ND Est. (3) (1,6)  ;

WF-70 39,35,44 ND ND ND .ND' 418 Eval.(4) ( 1,6,8) l t WF-29 49,39,45 .. ND ND ND ND i5 Est. (3) (1,6)-

[ NOTES FOR TABLE 3 ARE ON THE FOLLOWING PAGE.

t 1

- . ~- - __ . -_ _ __. _

r-

.A - . . . .

TABLE 3 (C00TINUED)

NOTES TO TABLE 3:

(1) BAW-2150 (2) Mt. Vernon Qualification Test Report (3). .8AW-1803, Revision 1, Tables 3-1 and 3-2; mean of RT , values for 34 Linde 80 welds.

(4) BAW-2100 (5) Values are for 60 hr stress-relief.

l (6) Cy (+10F) values are for 48 hr stress-relief.

(7)

C,I(,+10F) value forvalues are for 8maximum.

- 6 hr cycles stress-relief.

1 (8) R 40 hr stress-relief d

- ' -- - ' r -

J TABLE 4. GENERIC LETTER 92-01 RESPONSE: SECTION 2 ITEM b, 1 (2)

Subjoct: 10CFR50.61 and 10CFR5G, Appendix G, III.A; Material Properties Related to PTS and Fracture Toughness Requirements -- APPLICABLE ONLY TO REACTOR VESSELS CONSTRUCTED TO AN ASME CODE EARLIER THAN THE 1971 EDITION, SUMMER 1972 ADDENDA Plant: Zion Unit 2 Column 1 Column 2 Col. 3 t Material Heat Treatment Notes BELTLINE MA'IERIALS ZV3855 1525F-12h/WQ; 1215F-7h/AC; ll25F-29%h/FC (1,2) 88006-1 1600-1650F-9%h/BQ: 1200-1225F-9%h/BQ; 1100-1150F-31h/FC (cumul.)

B8040-1 1600-1650F-9%h/BQ; 1200-1225F-9%h/BQ; 1100-Il50F-31h/FC (cumul.)

C4007-1 1600-1650F-9%h/BQ; 1200-1225F-9%h/BQ; 1100-Il50F-?9h/FC (cumul.)

B8029-1 1600-1650F-9%h/BQ; 1200-1225F-9%h/BQ; 1100-1150F .!9h/FC (cumul.)

WF-200 1100-ll50F-24\h/FC (cumul.)

SA-1769 1100-Il50F-26h/FC (cumul.) i WE-154 1100-Il50F-24%h/FC (cumul ) I WF-70 1100-Il50F-31h/FC (cumul.) s WF-29 Il00-ll50F-29h/FC (cumul .)

SURVEILLANCE

. MATERIALS

'C4007-1 1600-1650F-9%h/BQ; 1200-1225F-9%h/BQ; 1100-Il50F-30h/FC (1)

WF-2v9-1 1100-Il50F-30h/FC NOTES:

(1) BAW-2150 (2) Additional stress relief information per Mt. Vernon p'.ocess drawing.

(3) BQ - brine quench FC - furnace cool ,

TABLE 5. GENERIC LETTER 92-01 RESPONSE: SECTION 2, ITEM b, 1 (3)

Subject:

10CFR50.61 and 10CFR50, Appendix G, III.A; Material Properties Related to PTS and Fracture Toughness Requirements -- APPLICABLE ONLY TO REACTOR VESSELS CONSTRUCTED TO AN ASME CODE EARLIER THAN THE 1971 EDITION, SUMMER 1972 ADDENDA Plant: Zion Unit 2

[0, sma 1 Column 2 Column 3 Column 4 Column 5 C. 6 Be t u ine Heat Beltline Weld Wire Weld Flux Notes Plate or Number Weld Heat Lot Forging l Lower NB Forging ZV3855 NB to IS Circ.: WF-200 821T44 8773 (1,2)

IS Plate 88006-1 IS to LS Circ.: SA-1769 71249 8738 IS Plate 88040-1 LS to Dutch Circ.: WF-154 406L44 8720 LS Plate B8029-1 IS Longit.: WF-70 72105 8669 LS Plate C4007-1 LS Longit.: WF-29 72102 8650 NOTES: (1) BAW-2150 (2) Mt. Vernon process drawing (3) NB - Nozzle Belt IS - Intermediate Shell LS - Lower Shell 1

4

TABLE 6. GENERIC LETTER 92-01 RESPONSE: SECTION 2, ITEM b, 1 (4) l i

Subject:

10CFR50.61 and 10CFR50, Appendix G, III.A; Material Properties Related to PTS and Fracture Toughness Requirements -- APPLICABLE ONLY TO REACTOR VESSELS CONSTRUCTED TO AN ASME CODE EARLIER TilAN Tile 1971 EDITION, SUMMER 1972 ADDENDA Plant: Zion Unit 2 '

l Column 2 Column 3 Column 4 Column 5 Column 1 Survaillance Weld Wire Weld Flux Notes Surveillance Plate or Weld lleat Lot Forging lieat Number _-

C4007-1 l WF-209-1 72105 8773 (1) i 9

m-NOTES: (1) BAW-2150

TABLE 7. GENERIC LETTER 92-01 RESPONSE: SECTION 2, ITEM b, 1 (5)

Subject:

10Cr'R50.61 and 10CFR50, Appendix G, III.A; Material Properties Relate <i to PTS and Fracture Toughness Requirements -- APPLICABLE ONLY TO REACTOR VESSELS CONSTRUC ED TO AN ASME CODE EARLIER TriAN THE Ic71 EDITION SUMMER 1972 ADDENDA Plant: Zion Unit 2 Column 1 Column 2 C. 3 Material Chemici. Composition, Weight Percent Notes C Mn P S __ S i Cr Ni Mo Cu BELTLINE MATERIAL I

{ ZV3855 0.22 0.67 0.008 0.006 0.35 0.42 0.66 0.62 0.09 (1)

B8006-1 0.21 1.35 0.010 0.015 0.24 ND 0.54 0.53 0.12 (2)

B8040-1 0.23 1.35 0.008 0.014 0.25 ND 0.52 0.54 0.14 (2)

C4007-1 0.23 1.39 3.010 6.016 0.22 ND 0.53 0.54 0.12 (2)

B8029-1 0.23 1.38 0.010 0.014 0.21 ND 0.51 0.52 0.12 (2)

WF-200 0.07 1.60 0.010 0.015 0.48 0.14 0.63 0.40 0.24 (3)

WF-70 0.09 1.63 0.018 0.009 0.54 0.10 0.59 0.40 0.35 (3)

SA-1769 0.09 1.49 0.020 0.014 0.56 0.16 0.61 0.37 0.26 (3)

WF-29 0.08 1.65 0.015 0.012 U.42 0.05 0.63 0.38 0.23 (3)

WF-154 0.07 1.54 0.013 0.016 0.42 0.07 0.59 0.40 0.31 (3)

SURVEILLANCE MATERIALS C4007-1 0.23 1.39 0.010 0.016 0.22 0.065 0.53 0.54 C.12 (4)

WF-209-1 0.08 1.51 0.til7 0.013 0.68 0.06 U.57 0.39 0.30 (4)

REQUIRED: State heat number of weld wires used for determining above chemical composition if different from that in 1 (3). -- Not applicable --

NOTES:

(1) Supplier Material Test Report (2) BAW-2150 ,

(3) BAW-2121P (4) BAW-1543, Revision 3

TABLE 8. GENERIC LETTER 92-01 RESPONSE: SECTION 3, ITEM a

Subject:

Generic letter 88-11 Resanse Commitments; Effect of Irradiation Temperature Plant: Zion Unit 2 Cold leg Temperatuse (T,,,a): 529.4 F (See Figure 4-7)

If T is <525 F, state how this was considered in determination of embrittlement effects (CUSEg y , RT,) in accordance with Regulatory Guide 1.99, Revision 2:

Not applicable

References:

None

. . _ _ . ...___ ._ _ - . _ _ . _ _ . . _ _ _ _ _ . _ - - - ......________..-____._____.__._.._._m..

1 1

TABLE 9. GENERIC LETTER 92-01 RESPONSE: SECTION 3.' ITEM b-

i.

Subject:

Gereric Letter 88-11' Response Comitments; Utilization of Surveillance  ;

Recults

- Plant

Zion Unit 2

Were surveillance results used in determining C.USE? Yes a No /

Were'surveillarce results used.in determining RT ,7 Yes / No o i

!, If any "yes" boxes were checked above, state how the surveillance results were used:

! Initial RT, value for WF-70 weld metal.

J i

i l

References:

BAW-2100

i i

1 1 .

l i

j '.

t e

r'

_ _ _ _ _ _ _ _ =- _ - _ _ _ - _ - - _ _ - - - _ _ _ . . -

. ._. .__ -. - . = . . -

TABLE.10. GENERIC LETTER 92-01 RESPONSE: SECTION 3, ITEM c

Subject:

Generic Letter 88-11 Response Commitments; Difference Between Measured and ,

Predicted (Regulatory Guide 1.99, Revision. 2) Embrittlement Effects 4

Plant: Zion Unit 2 Question I. Does measured ART,o, exceed ART,oy + 20 predicted by Regulatory Guide 1.99, Revision 2?

Question II. Does measured yC USE drop exceed that obtained from Regulatory Guide 1.99, Revision 2, Figure 27 Column 1 Column 2 Column 3 Column 4 I Column 5 Column 6 Column 7 Beltline Fluenge Measured Predicted question I Measured Predicted Ouestion II Haterial n/cm ART,o, ART,or+2a If "yes" CyUSE CyUSE If "yes" (1,2,3) see Note (4) Drop Drop see Note (4)

ZV-3855 l ---

ND ND --

ND ND --

B8006-1 --

ND ND --

ND ND --

88040-1 ---

C ,

ND --

ND ND --

C4007-1 2.57E+18 49(1) 86 No 0(1) 14 No 8.04E+18 g 90(1) 110 No 14(1) 19 No 1 1.48E+19 121(1) 124 No 0(1) 22 No 88029-1 ---

ND ND --

ND ND --

WF-200 ---

ND ND --

ND ND --

SA-1769 ---

ND HD --

ND ND --

WF-154 ---

ND ND --

ND ND --

WF-70 6.63E+18 135(2) 259 No 13(2) 22(5) No WF-29 ---

ND ND --

ND ND --

Atypical 1.17E+18 28(3). 138 No 9(3) 25(3) No 6.56E+18 122(3) 216 No 16(3) 32(3) No 7.50E+18 119(3) 223 No 11(3) 32(3) No 1.08E+19 120(3) 242 No 15(3) 34(3) No NOTES FOR TABLE 10 ARE ON THE FOLLOWING PAGE.

1 TABLE 10 (CONTINUED)

NOTES: (1) WCAP-12396 (2) BAW-1803, Revision 1 (3) BAW-2049 (4)_ Statement not required.

(5) BAW-1920P

7. REFERENCES-

'A. L. Lowe, Jr. <t al, " Analysis of Capsule OCl-F - from- Duke Power Company Oconee Unit 1 Reactor Vessel Materials Surveillance Program," BAW-1421. Revision 1, Babcock & k'ilcox, Mclear Power Generction Division, Lynchburg, Virgini ,

September 1975.

'A. L. Lowe, Jr. et al, " Analysis of Capsule ANI-B from Arkansas Powm.r & Light ,

Company's Arkansas Nuclear One, Unit 1 Reactor- Vessel Materials Surveillance Prodram," BAW-1694. Babcock- &- Wilcox,, Nuclear Power Generation' Division,=

Lynchburg, Virginia, November 1981.

'A. S. Heller and A. L. Lowe, Jr., " Correlations for Predicting the Effects of Neutron Radiation on Linde 80 Submerged-Arc Welds," BAW-1803, Babcock & Wilcox, ,

Utility Power Generation Division, Lynchburg, Virginia, January 1984.

A. L. Lowe, Jr. and J. W. Pegram, " Correlations for Predicting the-Effects of Neutron Radiation on Linde' 80 Submerged-Arc Welds, BAL'-1803. Revision 1, B&W -

Nuclear Service Company, Lynchburg, Virginia, May 1991.

'J. D. Aadland, " Babcock & Wilcox Owners' Group 177-fuel Assembly Reactor Vessel and Surveillance Program Materials 'Information," BAW-1820, Babcock.-& Wilcox, Nuclear Power Division, Lynchburg, Virginia, December 1984.

'A. L. Lowe, Jr. et al, "A.ialysis of Capsule TEl-A, The Toledo' Edison Company, Davis Besse Nuclear Power Station Unit 1, Reactor Vessel Materials Surveillance Program," -BAW-1882. Revision 1, Babcock & Wilcox, Nuclear -Power Divisio?,

Lynchburg, Virginia, June 1989.

'This report is available from B&W- Nuclear Service Company, Lynchburg, Virginia.

7-1

'A. L. Lowe, Jr. et al, " Analysis of Capsule CR3-C, Florida Power Corpo,ation, Crystal River Unit 3, Reactor Vessel Materials Surveillance Program," BAW-1898, Babcock & Wilcox, Nuclear Power Division, Lynchburg, Virginia, March 1986.

'A. L. Lowe, Jr. et al, " Analysis of Capsule CR3-D, Florida Power Corporation, Crystal River Unit 3, Reactor Vessel Materials Surveillance Program," BAW-lqH, Babcock & Wilcox, Nuclear Power Division, Lynchburg, Virginia, March 1986.

'A. L. Lowe, Jr. et al, " Analysis of Cap le TMll-C, GPU Nuclear, Three Mile Island Nuclear Station-Unit 1, Reactor Vessel Materials Surveillance Program,"

BAW-1901, Babcock & Wilcox, Nuclear Power Division, Lynchburg, Virginia, March 1986.

'A. L. Lowe, Jr., " Reactor Pressure Vessel and Surveillance Program Materials Information for Surry Units 1 and 2, North Anna Units 1 and 2," BAW-1908, Babcock

& Wilcox. Nuclear Power Division, Lynchburg, Virginia, February 1986.

'A. L. Lowe, Jr., " Reactor Pressure Vessel and Surveillance Program Materials Licensing Information for Surry Units 1 tnd 2," BAW-1909. Revisign_1, Babcock &

Wilcox. Nuclet' Power Division, Lynchburg, Virginia, August 1986.

'A. L. Lowe, Jr. at al, " Analysis of Capsule CR3-LG1, Babcock & Wilcox Owners Group, Integrated Reactor Vessel haterials Surveillance Program," BAW-1910P, Babcock & Wilcox, Nuclear F w r Division, Lynchburg, Virginia, August 1986.

'A. L. Lowe, Jr. et al, " Analysis of Capsule DB1-LG1, Babceck & Wilcox Owners Group, Integrated Reactor Vessel Materials Surveillance Program," BAW-1920P, Babcock & Wilcox, Nuclear Power Division, Lynchburg, Virginia, October 1986.

'A. L. Lowe, Jr. et al, " Analysis of Capsule CP.3-F, Florida Power Corporation, Crystal River Unit 3, Reactor Vessel Materials Surveillance Program," BAW-2049, Babcock & Wilcox, Nuclear Power Division, Lynchburg, Virginia, September-1988.

'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-20FQ, Babcock & Wilcox, Nuclear Power Division, Lynchburg, Virgir.ia, October 1988.

7-2

'A. L. Lowe, Jr. et al, " Analysis of Capsule OCII-E, Duke Power Company, Oconee Nuclear Station Unit 2, Reactor Vessel Materials Surveillance Program," BAW-2051, Babcock & Wilcox, Nuclear Power Division, Lynchburg, Virginia, October 1988.

'A. L. Lowe, Jr. et al, " Analysis of Capsule AN1-C, Arkansas Power & Light

! Company, Arkansas Nuclear One, Unit 1, Reactor Vessel Materials Surveillance Program," BAW-2075. Revision 1, Babcock & Wilcox, Nuclear Power Division, l Lynchburg, Virginia. October 1989.

A. L. Lowe, Jr. et al, " Analysis of Capsule Y, Commonwealth Edison company, Zion Nuclear Plant Unit 1, Reactor Vessel Materials Surveillance Program," BAW-1Q32, B&W Nuclear Service Company, Lynchburg, Virginia, March 1990.

A. L. Lowe, Jr. " Properties of Weld Wire Heat Number 72105 (weld Metals WF-70 and WF-209-1)," BAW-?100, B&W Nuclear Service. Company, Lyncht m , Virginia, To be published. .

K. K. Yoon and A. L. Lowe, Jr., " Low Upper-Shelf Toughness Fracture Analysis of Reactor Vessels of Turkey Point Units 3 and 4 for load Level A and B Condition,"

BAW-2118P, B&W Nuclear Service Company, Lynchburg, Virginia, November 1991.

l L. B. Gross, " Chemical Composition of B&W Fabricated Reactor Vessel Beltline Welds," BAW-2121P, BSW Nuclear Service Company, Lynchburg, Virginia, April 1991.

A. L. Lowe, Jr et al, " Analysis of Capsule TEl-0, The Toledo Edison Company, Davis Besse Nuclear Power Station Unit 1, Reactor Vessel Materials Surveillance Program," BAW-2125, B&W Nuclear Service Company, Lynchburg, Virgini4, December 1990.

A. L. Lowe, Jr. et al, " Analysis of Capsule OCIII-D, Duke Power Company, Oconee l Nuclear Station Unit-3, Reactor Vessel Materials Surveillance Program," BAW-2128, l B&W Nuclear Service Company, Lynchburg, Virginia, May 1991, i

A. L. Lowe, Jr. et al, " Analysis of Capsule S, Wisconsin Electric Power Company, l Point Beach Nuclear Plant Unit No. 2, Reactor Vessel M3terials Surveillance l Program," BAW-2140, B&W Nuclear Service Company, Lynchbarg, Virginia, August 1991.

l 7-3

K. K. Yoon and L. B. Gross, " Low Upper-Shelf Toughness Fracture Analysis of Reactor Vessels of Zion Units 1 and 2 for Load Level A and B Conditions," BAW-2148P, B&W Nuclear Service Company, Lynchburg, Virginia, March 1992.  !

'H. S. Palme et al, " Methods of Compliance with Fracture Toughness and Operational Requirements of 10CFR50, Appendix G," BAW-10046P, Babcock & Wilcox,  !

Nuclear Power Generation Division, Lynchburg, Virginia, March 1976.

W. A. VanDerSluys et al, "An Investigation of Mechanical Properties and Chemistry Within a Thick Hn-Mo-Ni Submerged Arc Weldment," EPRI NP-373, Electric Power Research Institute, Palo Alto, California, February 1977.

E. B. Norris, " Reactor Vestel Material Surveillance Program for Turkey Point Unit No. 4: Analysis of Capsule T," Final Ret > ort SWRI Project No. 02-4221, Southwest Research Institute, San Antonio, Texas, June 1976.

E. B. Norris, " Reactor Vessel Material Surveillance Program for Capsule S -

Turkey Point Unit No. 3 (and] Capsule S- Turkey Point Unit No. 4," Final Report SWRI Project Nos. 02-5131 and 02-5380, Southwest Research Institute, San Antonio, Texas, May 1979.

P. K. Nair and E. 8. Norris, " Reactor Vessel Material Surveillance Program for Turkey Poin' Unit No. 3: Analysis of Capsule V," Final Reoort SWRI Project No.

06-8575, Southwest Research Institute, San Antonio, Texas, August 1986.

S. E. Yanichko et al, " Analysis of Capsule T from the Florida Power and Light Company Turkey Point Unit Na. 3 Reactor Vessel Radiation Surveillance Program,"

WCAP-8631, Westinghouse Electric Corporation, Pittsburgh, Pennsylvania, December 1975.

S. E. Yanichko and S. L. Anderson, " Analysis of Capsule R from the Wisconsin Electric Power Company Point Beach Nuclear Plant Unit No. 1 Reactor Vessel Surveillance Program," WCAP-9357, Westinghouse Electric Corporation, Pittsburgh,

! Pennsylvania, August 1978.

l l

7-4

-- - -