ML18142A864
| 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
NRC FoRM 195 U.S. NUCI.EAR REG'ULATORY CO
>SSION (2-18 I I'
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kj'.'-~-NRC DISTRIBUTION FGR PART 50 DOCKET MATERIAL DOCKET NUMBER FILE NUMBER A Schw neer LETTER
/ORIGINAL OCOPY DESCRIPTION QNOTORIZEO g UN C I ASS I F IE0 FROM:
PROP Rochester Gas
& Elec Corp Rochester, NY L D White INPUT FORM ENCLOSURE OATE OF OOC)MF)T DATE RECEIVED 9 77 NUMBER OF COPIES RECEIVEO 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 BRANCH CHIEF: (7)
FOR ACTION/INFORMATION gREG FI NRC PDR I G E (2)
OELD HANAUER CHECK STELLO EISENHUT SHAO BAER BUTLER GRIMES COLLI S INTERNALDISTRIBUTION LPDR:
TIC NSIC 16 CYS ACRS SENT CATE 0 EXTERNALDISTRIBUTION CONTROL NUMBER
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/IIIIIIIII ROCHESTER GAS AND ELECTRIC CORPORATION NCT LLi~ 0 ILAOg 5o avf
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89 EAST AVENUE, ROCHESTER, N.Y. 14649 LEON D. WHITE, JR.
VICE PRESIDENT TKLSPHON S ARK*COOS 7ld 546.2700 Director of Nuclear Reactor Regulation ATTN:
Mr. A. Schwencer, Chief Operating Reactors Branch 41 U. S. Nuclear Regulatory Commission Washington, D. C.
20555 September 13, 1977~ <
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SEP) S l977 h4tl S'CIion"
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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 willfind 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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ATTACHMENTA 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 M
M O
Forging 123PP18VAl SA-1101 Forging 125S255VAl Core SA-847 Forging 125P666VAl
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TA8LE 1
IDENTIFICATION OF REACTOR YESSEL BELTLINE REGION MELD MATERIAL Meld Location Nozzle Shell to Inter. Shell inter. Shell to Lo~!er Shell Surveillance Meld Meld Mire Flux Meld Process Control No.
T~e Heat No.
~T e
Lot No.
Post Meld Heat Treatment Submerged Arc SA-1101 bin-No-Ni 71249 Linde 80 8445 1100-1125'F-48 Hrs.-FC Submerged Arc SA-847 Pin-Mo-Ni 61782 Linde 80 8350 1100-1125'F-48 Hrs.-FC 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 Control No.
~T e
Heat No.
SA-1101 Hn-Ho-Ni 71249 SA-847 Hn-tIo-i'eii 61782 Survei 1 1 ance Meld Flux
~Te Lot No.
Linde 80 8445 Linde 80 8350 C
.070
.082-,
.075 Mei ht Percent P
S i~in Si llo Ni Cr
.021
.014 1.28
.52
.36
.57
.17 c.
.21
.012
.012 1-.34
.45
.39
.39
.06
.20
.012
.016 1.31
.59
.36
.56
.59
.23
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TABLE 3 MECHANICAL PROPERTIES OF WELD HATERIAL Weld Weld Wire Control No.
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Heat No.
SA-1484 Hn-Ho-Ni 71249 SA-1101 Hn-Ho-Ni 61782 Surveillance Weld Energy at 10'F fl-'lb 45, 45, 46 58, 60, 36 0*
54, 66.5, 71:*
Flux I ilPT
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Lot No.
'F Linde 80 8445 0"'inde 80 8350 0*
Shel f Energy ft-1b NDT oF 0*
0*
0*
79,.0 YS ksi 68.63 67.00 73.52 UTS Elong.
- ksi 84.26
- 28.5 81.88 29.5 87.35 22.8
- Estimated based on NRC Standard Review Plan Section 5.3.2 and HTEB 5-2
"-* Energy at 60'F TABLE 4 MXIHUH END'OF LIFE FLUENCE AT VESSEL'HAL'L L'OCATIONS Nozzle Shell to Inter. Shell Weld Inter. Shell to Low Shell Weld Nozzle Shell Forging 'i23P118VAl.
Inter Shell Forging 125S255YAl Lower Shell Forging 125P666YA1 Fluence n/cm
~2.0 x 10 3.7 x 10 19
~2.0 x 10 18 3.7.x 10 19 3.7 x 10
TABLE 5 IDENTIFICATION OF REACTOR VESSEL BELTLINE FORGING MATERIAL
~tom anent Nozzle Shell Inter. Shell Lower Shell Surveillance Forgings
. Forging No.
123P118VA1 125S255VA1 125P666VAl 125S255VAl 125P666VAl.
Material Heat No.
~Sec.
123P118 A336 125S255 A508 CL2 125P666 A508 CL2 125S255 A508 CL2 125P666 A508 CL2
~Su 1 1 er Bethl chem Steel Bethlehem Steel Bethlehem Steel Bethlehem Steel Bethlehem Steel Austenitize 1550'F-11 Hrs-klQ 1550'F-15-1/2 Hrs-ilQ 1550'F-9 Hrs-MQ 1550'F-15-1/2 Hrs-ilQ 1550'F-9 Hrs-.AC Heat Treatment
~Tem er 1220'F-22 Hrs-AC 1210'F-18 Hrs-AC 1220'F-12 Hrs-AC 1210'F-18 Hrs-AC 1220'F-12 Hrs-AC Stress Relief 1125'F.-'30 Hrs-FC 1125'F-30 Hrs-FC 1125'F-30 Hrs-FC 1100'F-11-1/4 Hrs-~
1100'F-ll-Hrs-FC W TABLE 6 BELTLINE FORGING MATERIAL CHEMICAL COMPOSITION Forqin No.
123P118VA1 125S255VA1 125P666VAl C
.19
.18
.19 P
S Mn
.010
.009
.65
.010
.007
.66
.012
.011
.67 Mei ht Percent Si
.23
.23
.20 Mo
.60
.58
.57 Ni
.69
.69
.69 Cr
.42
.33
.37
.07
.05
.02
.02
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TABLE 7 NECHANICAL PROPERTIES OF BELTLINE FORGINGS For in No.
123P118VAl 125S255VA1 125P666VA1 NDT oF 40 RT.,
NDI oF 40~
Upper Shel f Energy ft-lb 117*
20 20*
106*
40 40*
114*
YS ksi 66.87 67.25 63.50 UTS ksi 88.00 88.25 85.00 Elong.
cj 25.50 26.25 26.25 RA 73.50 70.10 71.05 125S255VA1 125P666VAl 20 40 20" 40*
91x120'8.
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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e ATTACHMENTB 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.
4.3-1
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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