ML17264A194

Provides Pressurized Thermal Shock Assessment for Reactor Pressure Vessel Limiting Circumferential Weld for Review & Approval

ML17264A194
Person / Time
Site:	Ginna
Issue date:	10/11/1995
From:	Mercredy R ROCHESTER GAS & ELECTRIC CORP.
To:	Andrea Johnson NRC (Affiliation Not Assigned), NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 9510180239
Download: ML17264A194 (9)

Text

I PRIORITY 3y 7

(ACCELERATED RZDS PROCESSING)

REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)

ACCESSION NBR:9510180239 DOC.DATE: 95/10/11 NOTARIZED: NO DOCKET FACIL:50-244 Robert Emmet Ginna Nuclear Plant, Unit 1, Rochester G 05000244 P AUTH. NAME AUTHOR AFFILIATION MERCREDY,R.C. Rochester Gas & Electric Corp.

RECIP.NAME RECIPIENT AFFILIATION JOHNSON,A.R.

SUBJECT:

Provides pressurized thermal shock assessment for reactor pressure vessel limiting circumferential weld for review &

approval. 0 DISTRIBUTION CODE: A049D COPIES RECEIVED:LTR ENCL SIZE:

TITLE: OR Submittal: Thermal Shock to Reactor Vessel R NOTES:License Exp date in accordance with 10CFR2,2.109(9/19/72). 05000244 RECIPIENT COPIES RECIPIENT COPIES 1D CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL PD1-1 PD INTERNAL: AEOD/DOA NRR/DE OGC/HDS3 1

1 1 0

JOHNSO F

NRR/DSSA/SRXB 1

1' 1

1 1

1 1 0 RES DE 1 1 RES/DE/MEB/MES 1 1 RES/DSIR/EIB 1 1 RES/DSR/RPSB 1 1 RES/DST/PRAB 1 1 RGN1 ADMSTR 1 1

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EXTERNAL: NOAC 1 1 NRC PDR t'g 5'3/~i N

NOTE,TO ALL "RZDS" RECIPIENTS:

PLEASE HELP US TO REDUCE WASTE! CONTACT THE DOCUMENT CONTROL DESK, ROOM OWFN 5D8 (415-2083) TO ELIMINATE YOUR NAME FROM DISTRIBUTION LISTS FOR DOCUMENTS YOU DON'T NEEDI TOTAL, NUMBER OF COPIES REQUIRED: LTTR 15 ENCL 13

0 4 AND ROCHE/TER GAS AND ELECTRIC CORPORATION ~ 89 FASTAYENLIE, ROCHESTER, N.Y. Idio-000I AREA CODE 716 Sd6-2700 ROBERT C. MECREDY Vice President N vcleor Operotions, October 11, 1995 U.S. Nuclear Regulatory Commission Document Control Desk Attn: Allen R. Johnson Project Directorate I-1 Washington, D.C. 20555

Subject:

Pressurized Thermal Shock Assessment for Ginna Reactor Vessel R.E. Ginna Nuclear Power Plant Docket No. 50-244 Ref. (a): Letter from Allen R. Johnson (NRC), to Robert C. Mecredy (RG&E), "Summary of Meeting with Rochester Gas and Electric Corporation on May 16, 1995 Pressurizer Thermal Shock Assessment of Reactor Vessels," dated June 5, 1995.

Dear Mr. Johnson:

The purpose of this correspondence is to provide Rochester Gas and Electric Corporation's (RG&E) pressurized thermal shock (PTS) assessment for the R.E. Ginna reactor pressure vessel (RPV) limiting circumferential weld for your review and approval. The assessment, as presented in the attached, is provided in response to the referenced letter and is based on the guidance provided in 10CFR50.61 and Regulatory Guide 1.99 revision 2. As provided for by this guidance, the RG&E assessment includes the use of the results of the RPV surveillance capsule program. As the attached PTS determination shows, the Ginna RPV can be expected to operate past the end of license (EOL) of 2009 without exceeding the PTS screening criterion of 10CFR50.61 for the limiting case weld.

Very truly yours, Robert C. Mecredy REJtl,393 i.70jg J 95i0180239 'st5lOii PDR ADOCK 05000244 P PDR pyre+ p 8'K/ prtl.

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

ATTACHMENT R.E. Ginna Reactor Vessel EOL PTS Assessment Discussion The purpose of this assessment is to show that the expected reference temperature PTS (RT~,) at EOL for the Ginna RPV limiting weld remains below the screening criteria when using surveillance program data. The limiting material for the Ginna RPV is the intermediate to lower shell weld having weld wire heat number 61782. The material used for the surveillance capsule program is weld wire SA-1036 having material heat number 61782 which has been determined to be acceptable for use as a surveillance material+. The determination of the reference temperature for RT~ for the RPV-PTS assessment uses the 10CFR50.61(b)(3) plant specific surveillance program method which includes the surveillance data adjustments per Regulatory Guide 1.99 revision 2 position 2, where:

(1) RTpgg = I + M + aRTggg (2) I = initial RT~~~

(4) ~RT~Y = reference temperature shift">

8 II. Material Values Used A. RPV material heat 61782 1~ I= the mean of the test values where: 1 o F(()

i=1 n

i1 n

n =

=

2

-38'F)

~ I= -19.5'F and o', = 18.5'F 2 ~ fluence 9 EOL = 3.68E19("

3 ~ Chemistry(')

Cu = ~ 25 Ni = .54 4 ~ CF('):

167.6oF 5 ~ M = 2ltr~ + e,~ where e, = 18.5'F (II.A.1) and 14 o F(4)

~ M = 46.4'F B. Surveillance Capsules Weld SA 1036 1~ ~RTND V 5.56E18 140oF R 1.15E19 165oF T 1.97E19 150oF S 3.87E19 205oF 2 ~ Mean Value Chemistry+(')+:

Cu = .214 Ni = .505 3 ~ CF>>:

150 'oF

IXX. Ca'lculations A. Chemistry factor ratio<4>:

CF of RPV (II.A.4 above)  : CF of capsule (II.B.3 above)

~ Ratio = 167.6 = 1.11 150.9 B. Best fit CF<"

1. Summing adjusted capsule ~RTNDT Adjusted V 140 X 1.11 155.4 R 165 X 1.11 183.2 T 150 X 1.11 166.5 S 205 X 1.11 227.6 2 ~ Determine sum of fluence factor x adjusted aRT~~

where fluence factor (ff) = f< "

Adjusted

~Ca sule FF ~RT m Product V 0.836 155. 4 129.9.

R 1.039 183.2 190.3 T 1.185 166.5 197.3 S 1.349 227.6 307.

sum = 824.5 3 ~ Summing squares of fluence factors (FF) for capsules using fluence factor values from II.B.1:

~ca sule V 0. 836 0. 699 R 1. 039 1. 080 T 1.185 1. 405 S 1.349 1.820 sum = 5.004 4 Sum of adjusted aRTN T

sum of squares of fluence 824.5 (XII.B.2)  : 5.004 (III.B.3) = 164.8 F

C.'RT~~ for surveillance fit 9 EOL EOL gRT~~ (CF) frogs-oioi~o where f= 3.68

( I.B.4) x 3.68 8 F) (1. 338) = 220. '164.

5 D. RTrrs ~ EOL RT~ = I+ M + aRTNg~

-19 ~ 5 F (II.A.1) + 46.4 F (II.A.5) + 220.5 (III.C) 247.4oF IV. Results The RT~ value at EOL for the R.E. Ginna vessel is 247.4 F.

This is 52.64F below the PTS screening criterion described in 10CFR50.61.

REFERENCES:

(1) BAW 1803 Revision 1, "Correlations for Predicting the Effects of Neutron Radiation on Linde 80 Submerged Arc Welds."

(2) WCAP-13902, "Analysis of Capsule S from the Rochester Gas and Electric Corporation R.e. Ginna Reactor Vessel Radiation Surveillance Program," dated December 1993.

(3) Code of Federal Regulations, 10CFR50.61, "Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events. "

(4) Regulatory Guide 1.99, revision 2, "Radiation Embrittlement of Reactor Vessel Materials."

(5) BAW 1920P, "Analysis of Capsule DB1-LG1," October 1986.

(6) Letter to Roger W. Kober from Morton B. Fairtile, "Safety Evaluation by the Office of Nuclear Reactor Regulation Related to Amendment No. 15 to Facility Operating License No. DPR-18 Rochester Gas and Electric Corporation R.E.

Ginna Nuclear Power Plant Docket No. 50-244," dated June 12( 1986.

(7) BAW-1500, "Chemistry of 177-FA B& W Owner's Group Reactor Vessel Beltline Welds," September 1978.

(8) 10CFR50.61, "Fracture toughness requirements for protection against pressurized thermal shock events."

(9) BAW-2121P, "Chemical Composition of B&W Fabricated Reactor Vessel Beltline Welds," April 1991.