ML17263A718
| ML17263A718 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 06/30/1994 |
| From: | MECREDY R C ROCHESTER GAS & ELECTRIC CORP. |
| To: | JOHNSON A R 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
P R.I C)R.IMY CCELERATED RIDS PROCESSING)
REGULATORY INFORMATION DXSTRIBUTION SYSTEM (RIDS)DOCKET I 05000244 ACCESSION NBR: 9407120310 DOC.DATE: 94/06/30 NOTARIZED:
NO FACIL:50-244 Robert Emmet Ginna Nuclear Plant, Unit 1, Rochester G AUTH.NAME AUTHOR AFFILIATION MECREDY,R.C.
Rochester Gas 6 Electric Corp.RECIP.NAME RECIPXENT AFFILXATION JOHNSON,A.R.
Document Control Branch (Document Control Desk)P NOTES:License Exp date in accordance with 10CFR2,2.109(9/19/72).
05000244 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 RECIPIENT ID CODE/NAME PD1-3 PD INTERNAL: NRR/DE/EMCB NRR/DRPE/PDI-1 NUDOCS-ABSTRACT OGC/HDS3 RES/DE/MEB EXTERNAL: NRC PDR COPIES LTTR ENCL 1 1 2 2 1 1 1 1 1 0 1 1 1 1 RECIPIENT ID CODE/NAME JOHNSON,A NRR/DORS/OGCB NRR/DRPW D/CB REG FIL 01 NSIC COPIES LTTR ENCL 2 2 1 1 1 1 1 0 1 1 1 1 D 0 C U NOTE TO ALL eRIDS" RECIPIENTS:
PLEASE HELP US TO REDUCE WASTE!CONTACT THE DOCUMENT CONTROL DESK, ROOM PI-37 (EXT.504-2083)TO ELIMINATE YOUR NAME FROM DISTRIBUTION LISTS I'OR DOCUMENTS YOU DON'T NEED!TOTAL NUMBER OF COPIES REQUIRED: LTTR 15 ENCL 13
't 1e AND ROCHESTER GAS AND E'IECTRIC CORPORATION
~89 FAST AVENUE, ROCHESTER, N.Y.146d9-0001 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 AREA CODE7T6 546-2700
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, it was requested that data listed in two tables attached in that 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 ADOCK 05000244 P PDR pgrP4 ZD//K/A~~T ggS',
lr f 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 Beltline Material Upper Shell Forging Interm.Shell Forging Heat No.123P118VA1 1258255VA1 IS Neut.Fluence at EOL/EFPY 3 69E+18 3.68E+19s IRTN~F+305 (a)=0)+203 (ai=o)Method of Determin.IRTypp Plant Specific Plant Specific Chemistry Factor 223.6 16.26 Method of Determin.CF RG1.99 Table 2 Calculated
%Cu O.3S'07m Lower Shell Forging SA-1101 US to IS Circ.Weld SA-847 IS to LS Circ.Weld SA-848 LS to Dutch.Circ.Weld 125P666VA1 71249 61782 61782 3.68E+19i 3.72E+18i 3.68E+19~N/A+40'.(a)=0)+10 (a)=0)55 (a,=19.7)5$(a,=19.7)Plant Specific Plant Specific Generic Generic 27.80 173.567 147 19s 147 19s Calculated 0.05'alculated 0.26" Calculated 0.25" Calculated 0.25>>
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 Beltline Material Upper Shell Forging Interm.Shell Forging Lower Shell Forging SA-1101 US to IS Circ.Weld SA-847 IS to LS Circ.Weld SA-848 LS to Dutch.Circ.Weld Heat No.123P118VA1 1258255VA1 125P666VA1 71249 61782 61782 Material Type SA-336 A 508-2 A 508-2 Linde 80 SAW Linde 80 SAW Linde 80 SAW 1/4T USE at EOL 78.8'2.6 94.2'MA>50 ft-lbs~N/A4 1/4T Neutron Fluence at EOL 2.71E+18~2.49E+19'~
2.49E+19'~
2.71E+18~2.49E+191'1.00E+17~
Unirrad.USE 117 91 114 70 70 70 Method of Determin.Unirrad.USE MTEB 5-2: 65%(Matl.Cert.)MTEB 5-2~: 65%(Surv.Matl.)MTEB 5-2: 65%Surv.Matl.Generic~Generic~Generic~
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 N~4 P C