ML17263A718

Forwards Summary for Pressurized Thermal Shock Calculation Re GL 92-01 Closeout

ML17263A718
Person / Time
Site:	Ginna
Issue date:	06/30/1994
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 9407120310
Download: ML17263A718 (10)

Text

R.I C) R.IMY PCCELERATED RIDS PROCESSING)

REGULATORY INFORMATION DXSTRIBUTION SYSTEM (RIDS)

ACCESSION NBR: 9407120310 DOC. DATE: 94/06/30 FACIL:50-244 Robert Emmet Ginna Nuclear Plant, Unit 1, Rochester NOTARIZED: NO DOCKET I G 05000244 AUTH. NAME AUTHOR AFFILIATION MECREDY,R.C. Rochester Gas 6 Electric Corp. P RECIP.NAME RECIPXENT AFFILXATION JOHNSON,A.R. Document Control Branch (Document Control Desk) R

SUBJECT:

Forwards summary for pressurized thermal shock calculation re GL 92-01 closout. I DISTRIBUTION CODE: A028D COPXES RECEIVED:LTR ENCL SIZE:

TTTLE: Generic Letter 92-01 Responses (Reactor Vessel Structural Tntegrrty 1 NOTES:License Exp date in accordance with 10CFR2,2.109(9/19/72). 05000244 RECIPIENT COPIES RECIPIENT COPIES ID CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL PD1-3 PD 1 1 JOHNSON,A 2 2 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 D/ CB 1 0 OGC/HDS3 1 0 REG FIL 01 1 1 RES/DE/MEB 1 1 EXTERNAL: NRC PDR 1 1 NSIC 1 1 D

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U NOTE TO ALLeRIDS" RECIPIENTS:

PLEASE HELP US TO REDUCE WASTE! CONTACT THE DOCUMENT CONTROL DESK, ROOM PI-37 (EXT. 504-2083 ) TO ELIMINATEYOUR NAME FROM DISTRIBUTIONLISTS I'OR DOCUMENTS YOU DON'T NEED!

TOTAL NUMBER OF COPIES REQUIRED: LTTR 15 ENCL 13

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AND ROCHESTER GAS AND E'IECTRIC CORPORATION ~ 89 FAST AVENUE, ROCHESTER, N.Y. 146d9-0001 AREA CODE7T6 546-2700 ROBERT C. MECREDY Vice Prertdent June 30, 1994 Ctnna Nuctear Production U.S. Nuclear Regulatory Commission Document Control Desk Attn: Allen R. Johnson Project Directorate I-3 Washington, D.C. 20555

Subject:

Response to Generic Letter 92-01 Request for Closure 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),

Generic Letter (GL) 92-01, Revision 1, Reactor Structural Integrity", dated April 12, 1994 (b): Letter from R. C. Mecredy (RG&E) to A. R. Johnson (NRC),

"Generic Letter 92-01 Revision 1 Reactor Structural Integrity Response for Additional Information", dated May 16, 1994 (c): Letter from R. C. Mecredy (RG&E) to A. R. Johnson (NRC),

"Analysis of Capsule "S" from the Rochester Gas and Electric Corporation R. E. Ginna Reactor Vessel WCAP 13902 December 1993", dated March 29, 1994

Dear Mr. Johnson:

The reference (a) letter requested information to support closeout of issues addressed in Generic Letter 92-01. Specifically, in that it was requested that data listed in two tables attached reference be evaluated and updated .as required. Rochester Gas and Electric Corporation (RG&E) provided its initial response in reference (b).

It is the intent'of this letter to complete RG&E's response.

The data applicable to the R. E. Ginna reactor vessel are provided in Tables 1 and 2 attached., These data reflect results derived from the latest surveillance capsule (ref. c) and information derived through the B&W Owners Group. The attached data supersedes the data provided in reference (b).

t; The B&W Owners 'Group has, provided Topical Reports BAW-2178PA and BAW-2192PA which present equivalent margin analyses to address conditions of low upper-shelf energy. Though the R. E. Ginna reactor vessel does not, demonstrate low upper shelf energy for its limiting SA-847'eld material, RG&E endorses the results and conclusions ggesented in these reports and may choose to apply 9407120310 940M'0 PDR P

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these reports to the Ginna reactor vessel should conditions later warrant.

Very truly yours, Robert C. Mecredy REJ5339 xc: U.S. Nuclear Regulatory Commission Region I 475 Allendale Road King of Prussia, PA 19406 Ginna Senior Resident Inspector

Table 1. R. E. Ginna Data Summar for Pressurized Thermal Shock Calculation IS Neut. Method of Method of Beltline Fluence at IRTN~ Determin. Chemistry Determin.

Material Heat No. EOL/EFPY F IRTypp Factor CF %Cu Upper Shell 123P118VA1 3 69E+18 +305 Plant 223. 6 RG1.99 O.3S' Forging (a)=0) Specific Table 2 Interm. Shell 1258255VA1 3.68E+19s +203 Plant 16.26 Calculated 07m Forging (ai=o) Specific Lower Shell 125P666VA1 3.68E+19i +40'. Plant 27.80 Calculated Forging (a)=0) Specific 0.05'alculated SA-1101 US to 71249 3.72E+18i +10 Plant 173. 567 0.26" IS Circ. Weld (a)=0) Specific SA-847 IS to LS 61782 3.68E+19~ 55 Generic 147 19s Calculated 0.25" Circ. Weld (a,=19. 7)

SA-848 LS to 61782 N/A 5$ Generic 147 19s Calculated 0.25>>

Dutch. Circ. (a,=19. 7)

Weld

Table 1. cont. R. E. Ginna Data Summar for Pressurized Thermal Shock Calculations NOTES:

1. Values from July 2, 1992 letter from R. C. Mecredy (RGGE) to A. R. Johnson (USNRC)

Subject:

Reactor Vessel Structural Integrity, 10CFR50.54(f), Response to Generic Letter 92-01, Revision 1, R. E. Ginna Nuclear Power Plant.

2. Values determined from WCAP-13902 and WCAP-13893.

3. Values determined from data in Material Test Report.

4. Value determined from data in EPRI NP-373.

5. Mean value from data in BAW-1803, Revision 1.

6. Chemistry Factors for forging 125S255VA1 and forging 125P666VA1 were determined using REG surveillance data as reported in WCAP-13902 and WCAP 13893.

7. Chemistry Factor for weld metal SA-1101 was determined using TP3 surveillance data for weld metal SA-1101. The TP3 30 ft-lb transition temperature shift data were obtained from BAW-1803, Revision 1, while the fluence data for the capsules were obtained from BAW-1803, Revision 1 and NUREG CR-3319, Revision 1.

8. Chemistry Factor for weld metal SA-847 and weld metal SA-848 was determined using BGWOG surveillance data for weld metal SA-1135 and REG surveillance data for weld metal SA-1036. These surveillance welds were fabricated with the same wire heat as weld metal SA-847 and weld metal SA-848. The BGWOG surveillance data were obtained from BAW-1803, Revision 1. The REG surveillance data were obtained from WCAP-13902.

9. No data available for this material, therefore, 0.35% is specified as defined in Regulatory Guide 1.99, Revision 2~

10. Values obtained from BAW-2150.

11. Values obtained from BAW-2121P.

12. Values obtained from BAW-1500.

Table 2. R. E. Ginna Data Summar for U er-Shelf Ener Calculation 1/4T Neutron Method of Beltline 1/4T USE Fluence at Unirrad. Determin.

Material Heat No. Material Type at EOL EOL USE Unirrad. USE Upper Shell 123P118VA1 SA-336 78. 2.71E+18~ 117 MTEB 5-2: 65%

Forging 8'2.6 (Matl. Cert.)

Interm. Shell 1258255VA1 A 508-2 2.49E+19'~ 91 MTEB 5-2~: 65%

Forging (Surv. Matl.)

Lower Shell 125P666VA1 A 508-2 2.49E+19'~ 114 MTEB 5-2: 65%

Forging 94.2'MA Surv. Matl.

SA-1101 US to IS 71249 Linde 80 SAW 2.71E+18~ 70 Generic~

Circ. Weld SA-847 IS to LS 61782 Linde 80 SAW > 50 70 Generic~

Circ. Weld ft-lbs~ 2.49E+191'1.00E+17~

SA-848 LS to 61782 Linde 80 SAW N/A4 70 Generic~

Dutch. Circ.

Weld

Table 2. cont. R. E. Ginna - Data Summar for U er-Shelf Ener Calculation NOTES:

I

1. Values determined using Regulatory Guide 1.99, Revision 2, guidelines.

2. USE issue covered by the approved equivalent margins analysis in the Topical Reports BAW-2192PA and BAW-2178PA.

3. Values obtained from BAW-2192PA

4. Not applicable due to fluence being below threshold

5. Unirradiated USE is 65% of the USE from a longitudinal oriented specimens as defined in MTEB 5-2.

6. Unirradiated USE is determined using data from other plants with similar materials to the beltline material (BAW-1803, Table 3-5).

7. Values determined using capsule surveillance results WCAP-13902

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