ML18142A864

R. E. Ginna - Response to Request of May 23, 1977 for Additional Information on the Ginna Reactor Vessel Material Surveillance Program

ML18142A864
Person / Time
Site:	Ginna
Issue date:	09/13/1977
From:	White L Rochester Gas & Electric Corp
To:	Schwencer A US Atomic Energy Commission (AEC)
References
Download: ML18142A864 (19)

Text

DISTRIBUTION AFTER ISSUANCE OF OPERATING LICENSE 1

U.S. NUCI.EAR REG'ULATORY CO >SSION DOCKET NUMBER NRC FoRM 195 (2-18 I I' 1

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kj'.'-~-NRC FILE NUMBER DISTRIBUTION FGR PART 50 DOCKET MATERIAL FROM: OATE OF OOC)MF)T A Schw neer Rochester Gas & Elec Corp Rochester, NY DATE RECEIVED L D White 9 77 LETTER QNOTORIZEO PROP INPUT FORM NUMBER OF COPIES RECEIVEO

/ORIGINAL OCOPY g UN C I ASS I F I E 0 DESCRIPTION ENCLOSURE Pattial response to NRC ltr dtd 5-23-77 which concerned reactor vessel material surveillance program.. ~....... ", ~

Sp PLANT NAME: Ginna 9-15-77 hf SAFETY FOR ACTION/INFORMATION BRANCH CHIEF: (7)

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/ Zuu(Z~ZE /IIIIIIIII -I g ROCHESTER GAS AND ELECTRIC CORPORATION o 89 EAST AVENUE, ROCHESTER, N.Y. 14649 LEON D. WHITE, JR. TKLSPHON S VICE PRESIDENT ARK* COOS 7ld 546.2700 September 13, 1977~ <

Director of Nuclear Reactor Regulation rE')"=..

ATTN: Mr. A. Schwencer, Chief SEP) S l977 Operating Reactors Branch 41 h4tl S'CIion" U. S. Nuclear Regulatory Commission Washington, D. C. 20555 '4~0 lr

Dear Mr. Schwencer:

This letter is in response to your request of May 23, 1977 for additional information on the Ginna reactor vessel material surveillance program. In our letter of July 14, 1977 we answered questions 1, 2 and 4.b(12) of your letter as well as requested additional time for our NSSS supplier, Westinghouse, to develop the answers to your other questions.

Since then we have received the answers to your remaining questions and are transmitting them to you as Attachment A to this letter.

Also you will find a description of our surveillance program as found in Ginna Station Technical Specification Section 4.3 as Attachment B to this letter.

As requested we are enclosing one signed original and 39 copies of this letter and its attachments for your use.

Very truly yours, L. D. White, Jr.

Attachments

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ATTACHMENT A ROBERT E. GINNA UNIT NO.'

REACTOR VESSEL tlATERIAL SURVEILLANCE PROGRAM 1.) The estimated maximum'fluence (E > 1 Yiev) at the inner surface of the reactor vessel wall as of March 31, 1977 is 5.26 x 10'8 n/cm~.

2.) The effective full power years (EFPY) of operation accumulated as of March 31, 1977 is 4.55 EFPY.

3.) Fabrication of the reactor vessel was performed by Babcock 5 'Hi lcox Co.

4.) a.) Sketch o the reactor vessel showing base material and welds in the beltline region is shown in Figure l.

b.) Information on each of the welds in the beltline region is shown in Tables 1 through 4.

c.) Information on each of the shell forgings in the beltline region is

.shown'in, Tables 4 through 7.

5,) Information relative to weld and forging material included in the material surveillance program is shown in Tables 1 through 3 and 5 through 7.

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FIGURE 1 IDENTIFICATIw AND LOCATION QF BELTLIHE REGI is l@TERIAL ROBERT E. GINNA UNIT NO. 1 REACTOR VESSEL QJ Forging 123PP18VAl M

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SA-1101 Forging 125S255VAl Core SA-847 Forging 125P666VAl

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TA8LE 1 IDENTIFICATION OF REACTOR YESSEL BELTLINE REGION MELD MATERIAL Meld Mire Flux Meld Location Meld Process Control No. T~e Heat No. ~T e Lot No. Post Meld Heat Treatment Nozzle Shell to Submerged Arc SA-1101 bin-No-Ni 71249 Linde 80 8445 1100-1125'F-48 Hrs.-FC Inter. Shell inter. Shell to Submerged Arc SA-847 Pin-Mo-Ni 61782 Linde 80 8350 1100-1125'F-48 Hrs.-FC Lo~!er Shell Surveillance Meld Subsserged Arc SA-1036 Nn-No-Ni 6'l782 Linde 80 8436 ll00'F-11-1/4 Hrs.-FC TA8LE 2 MELD tQTERIAL CHEMICAL COMPOSITION Meld Meld Mire Flux Mei ht Percent Control No. ~T e Heat No. ~Te Lot No. C P S i~in Si llo Ni Cr c.

SA-1101 Hn-Ho-Ni 71249 Linde 80 8445 .070 .021 .014 1.28 .52 .36 .57 .17 .21 SA-847 Hn-tIo-i'eii 61782 Linde 80 8350 .082-, .012 .012 1-.34 .45 .39 .39 .06 .20 Survei 1 1 ance Meld .075 .012 .016 1.31 .59 .36 .56 .59 .23

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TABLE 3 MECHANICAL PROPERTIES OF WELD HATERIAL Weld Weld Wire Flux I Energy Shel f ilPT at 10'F NDT Energy YS UTS - Elong.

Control No. ~T e Heat No. ~T e Lot No. 'F fl-'lb oF ft-1b ksi -

ksi SA-1484 Hn-Ho-Ni 71249 Linde 80 8445 45, 45, 46 0* 68.63 84.26 -

28.5 0"'inde SA-1101 Hn-Ho-Ni 61782 80 8350 0* 58, 60, 36 0* 67.00 81.88 29.5 Surveillance Weld 0* 54, 66.5, 71:* 0* 79,.0 73.52 87.35 22.8

- Estimated based on NRC Standard Review Plan Section 5.3.2 and 5-2 HTEB

"-* Energy at 60'F TABLE 4 MXIHUH END'OF LIFE FLUENCE AT VESSEL'HAL'L L'OCATIONS Fluence n/cm Nozzle Shell to Inter. Shell Weld ~2.0 x 10 Inter. Shell to Shell Weld 19 Low 3.7 x 10 Nozzle Shell Forging 'i23P118VAl. ~2.0 x 10 18 Inter Shell Forging 19 125S255YAl 3.7.x 10 Lower Shell Forging 125P666YA1 3.7 x 10

TABLE 5 IDENTIFICATION OF REACTOR VESSEL BELTLINE FORGING MATERIAL

. Forging Material Heat Treatment

~tom anent No. Heat No. ~Sec. ~Su 1 1 er Austenitize ~Tem er Stress Relief Nozzle Shell 123P118VA1 123P118 A336 Bethl chem Steel 1550'F-11 Hrs-klQ 1220'F-22 Hrs-AC 1125'F.-'30 Hrs-FC Inter. Shell 125S255VA1 125S255 A508 CL2 Bethlehem Steel 1550'F-15-1/2 Hrs-ilQ 1210'F-18 Hrs-AC 1125'F-30 Hrs-FC Lower Shell 125P666VAl 125P666 A508 CL2 Bethlehem Steel 1550'F-9 Hrs-MQ 1220'F-12 Hrs-AC 1125'F-30 Hrs-FC Surveillance 125S255VAl 125S255 A508 CL2 Bethlehem Steel 1550'F-15-1/2 Hrs-ilQ 1210'F-18 Hrs-AC 1100'F-11-1/4 Hrs-~

Forgings 125P666VAl. 125P666 A508 CL2 Bethlehem Steel 1550'F-9 Hrs-.AC 1220'F-12 Hrs-AC 1100'F-ll- Hrs-FC W TABLE 6 BELTLINE FORGING MATERIAL CHEMICAL COMPOSITION Mei ht Percent Forqin No. C P S Mn Si Mo Ni Cr 123P118VA1 .19 .010 . .009 , .65 .23 .60 .69 .42 125S255VA1 .18 .010 .007 .66 .

.23 .58 .69 .33 .07 .02 125P666VAl .19 .012 .011 .67 .20 .57 .69 .37 .05 .02

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TABLE 7 NECHANICAL PROPERTIES OF BELTLINE FORGINGS Upper RT., Shel f NDT NDI Energy YS UTS Elong. RA For in No. oF oF ft-lb ksi ksi cj 123P118VAl 40 40~ 117* 66.87 88.00 25.50 73.50 125S255VA1 20 20* 106* 67.25 88.25 26.25 70.10 125P666VA1 40 40* 114* 63.50 85.00 26.25 71.05 20" 125S255VA1 125P666VAl 20 40 40* 120'8.

91x 22 62.72 97.19 83.65 23.30 26.35 66.85 Surveillance Test Results

• Estimated Based on NRC Standard Review Plan Section 5.3.2 and NTEB 5.2

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4.3.0 REACTOR VESSEL MATERIAL SURVEXLLANCE PROGRAM Applies to the tests of the metallurgical specimens taken from the reactor beltline region.

To provide data for the determination of the fracture

'toughness of the reactor vessel.

4.3.1 The reactor vessel material surveillance testing.

program is designed to meet the requirements of Appendix H to 10 CFR Part 50. This program consists of the metallurgical specimens receiving the follow-ing test: tensile, charpy impact and the WOL test.

These test of the Radiation, Capsule Speciments shall be performed as follows:

Casu le Time Tested End of 1st core cycle End of 3rd core cycle 10 years, at nearest refueling 20 years, at nearest refueling 30 years, at nearest. refueling Standby 4.3.2 The report of the Reactor Vessel Material Surveillance shall be written as a Summary Technical Report as required by Appendix H to 10CFR Part 50.

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Basis: wh,I II iw Thi material surveillance program monitors changes I) in the fracture toughness properties of ferritic materials in the reactor. vessel beltline region of the reactor resulting from exposure to neutron irradiation and the thermal environment. The test data obtained from this program will be used to determine the conditions under which the reactor vessel can be operated with adequate margins of safety against fracture throughout its service life.

4.3-2

RECElVED PPCUgENy PitOC:.SSSG U~,'ip 19/f $ g'5 Nf 9 f0