ML17264A270

Provides Response to Items 2 & 3 of GL 92-01,Rev 1,Suppl 1, Reactor Vessel Structural Integrity.

ML17264A270
Person / Time
Site:	Ginna
Issue date:	11/20/1995
From:	Mecredy R ROCHESTER GAS & ELECTRIC CORP.
To:	Andrea Johnson NRC (Affiliation Not Assigned), NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
GL-92-01, GL-92-1, NUDOCS 9512050295
Download: ML17264A270 (10)

Text

REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)

ACCESSION NBR:9512050295 DOC.DATE: 95/11/20 NOTARIZED: NO DOCKET FACIL:50-244 Robert Emmet Ginna Nuclear Plant, Unit, 1, Rochester G 05000244 AUTH. NAME AUTHOR AFFILIATION MECEDY,R.C. Rochester Gas & Electric Corp. P RECIP.NAME RECIPIENT AFFILIATION JOHNSON,A.R.

SUBJECT:

Provides response to items 2 & 3 of GL 92-01,Rev 1,Suppl 1, "Reactor Vessel Structural Integrity." I DISTRIBUTION CODE: A028D COPIES RECEIVED:LTR ENCL SIZE:

TITLE: Generic Letter 92-01, Rev 1, Suppl 1 Responses (Reactor Vessel Struct 0 NOTES:License Exp date in accordance with 10CFR2,2.109(9/19/72). 05000244 R

RECIPIENT COPIES RECIPIENT COPIES ID CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL PD1-1 PD 1 1 JOHNSON,A 1 1 INTERNAL. ILE CENTER 01 1 1 NRR/DE/EMCB 2 2 NRR/DRPE/PD1-1 1 1 NUDOCS-ABSTRACT 1 1 OGC/HDS3 1 0 RES/DE/MEB 1 1 RES/DET/EMMEB 1 1 EXTERNAL: NOAC 1 1 NRC PDR D

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N NOTE TO ALL"RIDS" RECIPIENTS:

PLEASE ISELP US TO REDUCE 4V%STE! CONTACTTHE DOCL'CLIENT CONTROL DESK, ROOK! PI-37 (EXT. 504-20S3 I TO ELIXII.'PATE YOUR NAiILFROil DISTRIBUTION LISTS I'OR DOCI,'NIEN I'S YO! 'ON"I'l.'ED!

TOTAL NUMBER OF COPIES REQUIRED: LTTR 12 ENCL 11

AND ROCHESTER GAS AND EIECTRIC CORPORATION e 89 EASTAYENUE, ROCHESTER, N. Y Id&f9-0001 AREA CODE716 $ 16-2700 ROBERT C. MECREDY Vice President Nvcteor operations November 20, 1995 U.S. Nuclear Regulatory Commission Document Control Desk Attn: Allen R. Johnson Project Directorate I-1 Washington, D.C. 20555

Subject:

Six Month Response to NRC Generic Letter 92-01, Revision 1, Supplement 1, "Reactor Vessel Structural Integrity" R.E. Ginna Nuclear Power Plant Docket No. 50-244 Ref.(a): NRC Generic Letter 92-01, Revision 1, Supplement 1, "Reactor Vessel Structural Integrity", dated May 19, 1995 (b): Letter from R. C. Mecredy (RGGE) to A. R. Johnson (NRC)

"Pressurized Thermal Shock Assessment for Ginna Reactor Vessel", R.E. Ginna Nuclear Power Plant, Docket No. 50-244, dated 10/11/95

Dear Mr. Johnson:

The purpose of this letter is to provide the Rochester Gas and Electric response to items (2) and (3) of the reference (a) letter as follows:

Item ~Res ense (2) after consideration of all relevant data, there is no change in best estimate chemistry (3) a pressurized thermal shock assessment has been provided to NRC for approval which includes the ratioing technique described in Position 2.1 of Regulatory Guide 1.99, Rev.

2.

Item (4) of Reference (a) requires that a written report be generated which addresses RPV integrity, LTOP or P/T limits as required. The RPV integrity has been addressed by reference (b) for NRC review and approval. LTOP and P/T limits have been addressed by the RGGE Improved Technical Specification submittal per NUREG-1431. Completion of these activities will constitute completion of item (4).

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The attached tables provide the data applicable to the R.E. Ginna RPV integrity.

Very ruly yours, Robert C. Mecredy Attachment REJ4399 xc: Mr. Allen R. Johnson (Mail Stop 14B2)

Project Directorate I-1 Washington, D.C. 20555 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 IRTwo~ Determin. Chemistry Determin.

Material Heat No. EOL/EFPY F I Ropy Factor CF %Cu %Ni Upper Shell 123P118VA1 Plant 223.6 RG1.99 0.35 0.68

+30'~i=0)

Forging 3.69E+18'25S255VA1 Specific Table 2 Interm. Shell 3.68E+19 Plant 16.2

+20'cr,=0)

Forging Specific 0.05'.68'alculated Lower Shell t40 Plant Calculated 0.07'.68'A-1101 125P666VAl 3.68E+19 27.80 Forging (o,=0) Specific US to 71249 3.72E+18'10 Plant 173.56 Calculated 0.26" 0.60" IS Circ. Weld (~i=0) Specific SA-847 IS to LS 61782 3. 68E+19~ -19.5 Plant TBD* Cal cul ated 0. 25" 0. 54" Circ. Weld (al=l8.5) Specific SA-848 LS to 61782 N/A'19.5 Plant TBD* Calculated 0.25" 0.54" Dutch. Circ. (.,=18.5) Specific Weld

• To Be Determined - Currently under NRC review.

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

1. Values from July 2, 1992 letter from R. C. Hecredy (RG&E) 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 Haterial Test Report.

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

5. Hean value from data in BAW-1803, Revision 1 and BAW 1920P.

6. Chemistry Factors for forging 125S255VAl and forging 125P666VAl were determined using REG surveillance data as reported in MCAP-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 BLMOG surveillance data for weld metal SA-1135 and R.E. Ginna 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 BKWOG surveillance data were obtained from BAW-1803, Revision 1. The R.E. Ginna surveillance data were obtained from WCAP-13902 and used ratioing guidance per Regulatory Guide 1.99 Revision 2.

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 BAM-2150.

11. Values obtained from BAW-2121P and BAM-1500.

Table 2. R. E. Ginna - - Data Summar for U er-Shelf Ener Calculation Neutron Method of Beltline 1/4T USE Fluence at 'etermin.

Material Upper Shell Forging Heat No.

123P118VA1 Haterial SA-336 Type 2'/4T49E+19'nirrad.

at EOL EOL

2. 71E+18 117 USE Unirrad.

HTEB 5-2 Hatl. Cert.

USE

65%

78.8'2.6'4.

Interm. Shell 125S255VA1 A 508-2 2. 49E+19'~ 91 -MTEB 5-2: 65%

Forging Surv. Hatl.

Lower Shell 125P666VA1 A 508-2 2. 114 MTEB 5-2: 65%

Forging (Surv. Matl.)

SA-1101 US to IS 71249 Linde 80 SAW EHA 2. 71E+18 70 Generi Circ. Weld IS to LS 61782 Linde 80 SAW > 50 2. 49E+19'0 Generic c'A-847 Circ. Weld ft-lbs SA-848 LS to 61782 Linde 80 SAW N/A <1.00E+17 70 Generic' Dutch. Circ.

Weld

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

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 HTEB 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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