ML17263A477

Forwards Addl Info in Response to AR Johnson 930924 Ltr Re Generic Ltr 92-01, Reactor Vessel Structural Integrity, 10CFR50.54(f).

ML17263A477
Person / Time
Site:	Ginna
Issue date:	11/29/1993
From:	Mecredy R ROCHESTER GAS & ELECTRIC CORP.
To:	Andrea Johnson NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM), Office of Nuclear Reactor Regulation
References
GL-92-01, GL-92-1, NUDOCS 9312080099
Download: ML17263A477 (8)

Text

ACCELERATED DIS UTION DEMONS TION SYSTEM REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)

ACCESSION NBR:9312080099 DOC.DATE: 93/11/29 NOTARIZED: NO DOCKET 5 FACXL:50-244 Robert Emmet Ginna Nuclear Plant, Unit 1, Rochester G 05000244 AUTH. NAME AUTHOR AFFILIATION MECREDY,R.C. Rochester Gas 6 Electric Corp.

RECIP.NAME RECIPIENT AFFILIATION JOHNSON,A.R. Project Directorate I-3

SUBJECT:

Forwards addi info in response to AR Johnson 930924 ltr re Generic Ltr 92-01, "Reactor Vessel Structural Integrity, 10CFR50.54(f)." D DISTRIBUTION CODE: A028D COPIES RECEIVED:LTR ENCL SIZE:

TITLE: Generic Letter 92-01 Responses (Reactor essel Structural ntegrxty 1 8 NOTES:License Exp date in accordance with 10CFR2,2.109(9/19/72). 05000244 /

A RECIPIENT COPIES RECXPIENT COPIES ID CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL D PD1-3 PD 1 1 JOHNSON,A 2 2 D

INTERNAL: NRR/DE/EMCB 2 2 NRR/DORS/OGCB 1 1 NRR/DRPE/PDI-1 1 1 NRR/DRPW 1 1 NUDOCS-ABSTRACT 1 1 OC ~Q3C 1 0 OGC/HDS1 1 0 G FILE 01 1 1 RES/DE/MEB 1 1 EXTERNAL: NRC PDR 1 1 NSIC D

D D

NOTE TO ALL "RIDS" RECIPIENTS:

S PLEASE HELP US TO REDUCE WASTE! CONTACT THE DOCUMENT COblTROL DESK, ROOM P 1-37 (EXT. 20079) TO ELIMINATEYOUR NAME FROM DISTRIBUTION LISIS FOR DOCUMENTS YOU DON'T NEED!

TOTAL NUMBER OF COPIES REQUIRED: LTTR 15 ENCL 13

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i IIIII I IIIIIIIIII ROCHESTER GAS AND ELECTRIC CORPORATION o 89 EAST AVENUE, ROCHESTER N.Y. 14649-0001 ROBERT C. MECREDY TELEPHONE Vice President AREA CODE71B 546 2700 Cinna Nuctear Production November 29, 1993 U.S. Nuclear Regulatory Commission Document Control Desk Attn: Allen R. Johnson Project Directorate I-3 Washington, D.C. 20555

Subject:

Reactor Vessel Integrity Response to Request for Additional Information R.E. Ginna Nuclear Power Plant Docket No. 50-244 Ref.(a): Letter from A. R. Johnson (NRC) to R. C. Mecredy (RG&E),

>>Reactor Vessel Structural Integrity Request for Additional Information," dated September 24, 1993 (b): Letter from R. C. Mecredy (RG&E) to A. R. Johnson (NRC),

"Reactor Vessel Structural Integrity, 10CFR50.54(f)

Response to Generic Letter 92-01, Revision 1,>> dated July 2/ 1992

Dear Mr. Johnson:

The purpose of this letter is to provide additional information, as requested by reference (a), regarding the Rochester Gas and Electric response to Generic Letter 92 01/ reference (b). The requested information is provided in Table A to this letter. Table B provides preliminary Charpy upper-shelf energy (USE) data obtained from the latest (1993) surveillance capsule tests which demonstrate that the Ginna Station reactor vessel continues to exceed the 50 ft. lb. requirement of 10CFR50, Appendix G. The final results for capsule >>S>> will be provided when available.

Very truly yours, Robert C. Mec y 060lsg 9312080099 931129 PDR ADOCK 05000244 ' "I I Qp PDR

RE J43 11 Attachments xc: Mr. Allen R. Johnson (Mail Stop 14D1)

Project Directorate I-3 Washington, D.C. 20555 U.S. Nuclear Regulatory Commission Region I 475 Allendale Road King of Prussia, PA 19406 Ginna Senior Resident Inspector

TABLE A Additional information in response to NRC-RAI dated September 24, 1993:

A. Unirradiated CUSE Values 1~ Forging CUSE ( ft. lb) 125P666VA1 114 (NOTE 1) 123P118VA1 117 (NOTE 2) 2 ~ Weld SA-1101 CyUSE (ft lb) 70 (NOTE 3)

SA-848 70 (NOTE 3)

NOTE 1: The Charpy impact data reported in the Materials Test Report (Bethlehem Steel Corp., Report of Tests, Report No. 2203, October 24, 1966) for forging 125P666VA1 indicates an initial CUSE values of 176 ft-lb. The specimens were oriented such that the break is in the strong direction. This value is below the strong direction CUSE value (183 ft-lb) reported in WCAP-7254 and WCAP-10086 for the forging 125P666VA1 surveillance material. Therefore, for conservatism 176 ft-lb will be used to represent. the CUSE value in the strong direction for the beltline forging-125P666VAl. In accordance with the Standard Review Plan, Section 5.3.2, 654 of the strong direction CyUSE value may be taken as the weak direction CUSE value. Therefore, the weak direction CUSE value is 114 ft-lb (65> of 176 ft-lb). This value is well in excess of the minimum 75 ft-lb that is initially required for reactor vessel beltline materials in accordance with 10CFR50, Appendix G.

NOTE 2: The Charpy impact data reported in the Materials Test Report (Bethlehem Steel Corp., Report of Tests, Report No. 679, May 24, 1966) for forging 123P118VA1 indicates an initial CUSE value of 180.7 ft-lb. The specimens were oriented such that the break is in the strong direction. In accordance with the Standard Review Plan, Section 5.3.2, 65% of the strong direction CUSE value may be taken as the weak direction CUSE value.

Therefore, the weak direction CUSE value is 117 ft-lb (654 of 180.7 ft-lb). This value is well in excess of the minimum of 75 ft-lb that is initially required for reactor vessel beltline materials in accordance with 10CFR50, Appendix G.

NOTE 3 The unirradiated for SA-1101 and SA-848 is 70 ft-is rounded from 69.7 CUSE lb. This value is taken from Table 3-5 of BAW-1803; ft-lb.

it This value was statistically derived from the entire population of Linde 80 welds and

F is considered to be the most representative value obtainable, since it was obtained using valid and applicable information for Linde 80 weld material.

Analyses have demonstrated that material with this low CUSE value will provide margins of safety against fracture toughness. These analyses are documented in reports BAW-2192P and BAW-2178P which have been submitted to the NRC for acceptance.

Copper Content for Forging 123P118VA1 The copper content for forging 123P118VA1 is not available.

This forging was fabricated to the ASME SA-336 specification which does not require the reporting of a copper composition.

Therefore, as specified in Regulatory Guide 1.99, Revision 2, a copper content of 0.35% is assumed. This of course is greatly conservative since typical copper content for reactor vessel forging materials is in the range of 0.01 to 0.10<.

Fluence Values Capsule Fluence (n/cm~)

(reported in response to GL 92-01)

V 6.53E18 R 1.02E19 T 1.78E19 The fluence values specified in the response to GL 92-01, Revision 1, were obtained from BAW-1803, Revision 1, which restate the revised fluence values reported in WCAP-11026. To date, four surveillance capsules have been withdrawn from the R. E. Ginna reactor and the dosimetry from these irradiations has been evaluated by Westinghouse using current evaluation techniques. Two of these capsules were withdrawn at the conclusion of Fuel Cycles 2 and 3 in Fall 1972 and Spring 1974, respectively; while the third capsule was withdrawn following the completion of Fuel Cycle 9 in Spring 1980.

These revised fluence values, being larger than those documented in the earlier surveillance reports, are considered more conservative. The fourth capsule, Capsule S, was removed in Spring 1993 and is currently being analyzed. Preliminary results are provided in Table B.

TABLE B Latest weld capsule surveillance test results:

Capsule AUSE (ft lb) Fluence (n/cm~)

T 55.4 1.78E19 S (preliminary) 54.5 3.87E19