ML17261B033

Forwards Info Re Reactor Vessel Issues,Per 900305 Telcon Concerning Change of License Expiration Date

ML17261B033
Person / Time
Site:	Ginna
Issue date:	03/28/1990
From:	Mecredy R ROCHESTER GAS & ELECTRIC CORP.
To:	Andrea Johnson Office of Nuclear Reactor Regulation
References
NUDOCS 9004130175
Download: ML17261B033 (17)

Text

ACCELERATED DISTRIBUTION DEMONSTIRATION SYSTEM

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REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS) 4 ACCESSION NBR:9004130175 DOC.DATE: 90/03/28 NOTARIZED: NO DOCKET FACIL:50-244 Robert Emmet Ginna Nuclear Plant, Unit 1, Rochester- G 05000244 AUTH. NAME AUTHOR AFFILIATION MECREDY,R.C. Rochester Gas & Electric Corp.

RECIP.NAME RECIPIENT AFFILIATION Project Directorate I-3 'R

SUBJECT:

Forwards info re reactor vessel issues,per 900305 telcon concerning change of license expiration date.

DISTRIBUTION CODE: AOOID TITLE: OR COPIES RECEIVED:LTR Submittal: General. Distribution

( ENCL 2 SIZE:

S NOTES:License Exp date in accordance with 10CFR2,2.109(9/19/72). 05000244, RECIPIENT COPIES RECIPIENT ,, COPIES "A ID CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL PD1-3 LA 1 1 PD1-3 PD 1 1 JOHNSON,A 5 5

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INTERNAL: NRR/DET/ECMB 9H 1 1 NRR/DOEA/OTS B1 1 1 NRR/DST 8E2 1 1 NRR/DST/SELB 8D 1 1 NRR/DST/SICB 7E 1 1 NRR/DST/SRXB 8E 1 1 NUDOCS-ABSTRACT 1 1 1 0 OGC/HDS2 1 0 E 01 1 1 RES/DSIR/EIB 1 1 EXTERNAL: LPDR 1 1 NRC PDR 1 1 NSIC 1 1 t ~W e'~

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D-D NOTE TO ALL "RIDS" RECIPIENTS:

PLEASE HELP US TO REDUCE WASIZl CONTACT THE. DOCUMENT CONTROL DESK, ROOM Pl-37 (EXT. 20079) TO ELIMINATEYOUR NAME FROM DISTRIBUTION LISIS FOR DOCUMENTS YOU DON'T NEEDl /

TOTAL NUMBER OF COPIES REQUIRED: LTTR 21 ENCL 19

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ROCHESTER GAS AND ELECTRIC CORPORATION o 89 EAST AVENUE, ROCHESTER, N K 14649 0001 March 28, 1990 TCLEPHONK ARKA coDK 7ld 546.2700 IJ.S. Nuclear Regulatory Commission Document Control Desk Attn: Allen R. Johnson Project Directorate I-3 Washington, D.C. 20555

Subject:

Change of License Expiration Date Additional Information R.E. Ginna Nuclear Power Plant Docket No. 50-244

Reference:

(1) Letter from Bruce Snow (RG&E) to Carl Stahle (NRC),

dated January 19, 1988 of same subject (2) Letter from Robert Smith (RG&E) to Allen Johnson (NRC), dated October 5., 1989 of same subject (3) Telephone Conference of March 5, 1990 concerning issues related to change in license expiration

Dear Mr. Johnson:

On March 5, 1990, a telephone conference (Reference 3) was held with Allen Johnson (NRC),- John Tsao (NRC), Peter Nagata (EG&G), John Smith (RG&E), George Wrobel (RG&E), Mike Saporito (RG&E), William Galloway (RG&E), Robert Eliasz (RG&E) and John Jorgensen (RG&E). During the course of the conversation, the NRC and your contractor asked a number of questions which centered.

around. reactor vessel issues. The attached data is being supplied to you and directly. to your EG&G contractor (per your instructions) in an attempt to address your concerns.

Very truly yours, Robert C. Me redy Division Manager Nuclear Production JRJN096 Attachments 90041301'g~~ 900328 go/

PDR ADOCK 05000244 yqoe<<<4'~

P PDC

xc: Mr. Allen R. Johnson (Mail Stop 14D1)

Project Directorate I-3 Washington, D.C. 20555 U.S. Nuclear Regulatory Commission Region I 475 Allendale Road King of Prussia, PA 19406 Peter Nagata (Mail Stop 2218)

EG&G Idaho P.O. Box 1625 Idaho Falls, ID 83415 Ginna Senior Resident Inspector

LICENSE AMENDMENT CHANGE OF EXPIRATION DATE estions from March 5 1990 NRC/RGGE Telecon

1. Question RG&E's October 5, 1989 submittal has only one sentence in Attachment B regarding the issue of Reactor Vessel.

Surveillance. In particular, the NRC is interested in how the requirements of 10CFR50, Appendix H are being addressed.

~Res ense The purpose of RGGE's- Material Surveillance Program is to monitor changes in the fracture toughness properties of*

ferritic materials in the reactor vessel beltline region resulting from exposure of these materials to neutron irradiation and the thermal environment. The information from the Material Surveillance Program is used as input to the 10CFR50, Appendix G Fracture Toughness Requirements. As stated.

in Section 5.3.3.2 of the Ginna UFSAR and WCAP-7254, the Material Surveillance Program is designed to meet the requirements of Appendix H and ASTM E-185-73. It consists of six surveillance capsules positioned in the reactor vessel between the thermal shield and the vessel wall. The vertical center of each capsule is opposite the vertical center of the core. Each capsule contains tensile, Charpy V-notch and wedge opening loading specimens from the forgings (heats 125P666 and 125S255), and weld metal and Charpy V-notch specimens from the heat affected zone material. The surveillance capsules contain dosimeter wires of copper, nickel and aluminum-cobalt. They also contain cadmium-shielded dosimeters of Neptunium-237 and Uranium-238. The dosimeters permit evaluation of the neutron flux seen by the various specimens.

The surveillance capsules can be removed when the vessel head is removed. The capsules can be replaced when the vessel internals are removed. Surveillance capsules are periodically removed and. tested in accordance with Technical Specifications and test results are documented. Test results -are analyzed, the shift in transition temperature is compared to the predicted shift and pressure-temperature limit curves are revised accordingly.

Capsules withdrawn after July 26, 1983 will be tested and results reported in accordance with the 1982 revision of ASTM E-185 as required by Appendix H requirements.

The Ginna cap e withdrawal schedule is 8F accordance with E-185 and Technical Specification Section 4.3.1.1 and is as follows:

~Ca sule Time Removed for Testin V Removed. in 1971 R Removed in 1974 T Removed in 1980 p 17EFPY at nearest refueling S Standby N Standby The above capsule withdrawal schedule may change based on the Integrated Reactor Vessel Surveillance Program of Babcock and Wilcox in which RG&E participates.

it If such a change is will be covered by a separate submittal.

requested, The results of reactor vessel material surveillance capsules V, R and T have been submitted. to the NRC in accordance with ASTM E-185 requirements. The report numbers for these capsules are listed below:

~Ca sule Re ort No.

V FP-RA-1 R WCAP 8421 T WCAP 10086

2. Question The NRC wishes to have a commitment that the Charpy upper shelf energy of the Ginna vessel is above 50 ft-lb. as required by 10CFR50, Appendix G.

R~es nse The beltline weld is the limiting material in the Ginna reactor vessel. Three capsules have been removed and tested and the Charpy upper shelf energy results for the weld material is summarized as follows:

Capsule V: Upper shelf energy decreased from 74.0 ft-lbs. to 50.8 ft-lbs. at a fluence of 4.9 x 10"'eutrons/cm (E > 1 Mev).

Capsule R: Upper shelf energy decreased from 80.0 ft-lbs. to 50.0 ft-lbs.. at a fluence of 7.6 x 10'" neutrons/cm (E > 1 Mev).

Capsule T: Upper shelf energy decreased from 80.0 ft-lbs. to 55.0 ft-lbs. at a fluence of 1.75 x 10'eutrons/cm (E > 1 Mev).

In addition, reconstituted Charpy tests were performed on weld metal from Capsule T. The upper shelf energy decreased from 80.0 ft-lbs. to 55.0 ft-lbs. at a fluence of 1.75 x 10" neutrons/cm (E > 1 Mev).

Based. on the rveillance capsule result to date, concluded that the reactor it can be beltline region material is not as sensitive to radiation as predicted and a saturation of radiation damage may be occurring.

In addition to the above, a 100% examination of the reactor vessel beltline weld was performed during the 1989 refueling outage with no recordable indications found. (See letter of May 4, 1989 from R.E. Smith (RG&E) to NRC concerning Reactor Vessel Examination of 1989.)

RG&E has joined the Babcock and Wilcox Owners Group (BWOG) which was formed to address the upper shelf toughness issue for all B&W fabricated. reactor vessels with Linde 80 welds. The principal objective of this group is to assure the continued.

licensability of all eleven participants and. their seventeen reactor vessels. The group will provide the data required of Appendix G, 10CFR50 to accomplish this objective.

In the event that the Charpy upper shelf energy falls below 50 ft-lbs. in the future, an analysis would be performed that would conservatively demonstrate, making appropriate allowances for all uncertainties, the existence of, equivalent margins of safety for continued operation.

Finally, if there is no indication that an equivalent safety margin exists, then the reactor vessel may have to be thermally annealed to recover the fracture toughness of the material.

Westinghouse and EPRI have been working on a program to demonstrate the feasibility of this type of procedure as documented in EPRI Report No. NP-6113-M, Thermal Annealing of an Embrittled Reactor Vessel.

3. Question The NRC is interested in RG&E's. response to Generic Letter 88-11, "NRC Position on Radiation Embrittlement of Reactor Vessel Materials and it's Impact on Plant Operations".

R~es nse Our response was provided by letter dated January 23, 1989.

In accordance with our commitments in that response, RG&E expects new heatup and cooldown curves to be implemented by May 1991.

4. Question There is a discrepancy in WCAP-8421 concerning data for the Charpy upper shelf energy of forging 125666VA1 which is the lower shell forging. Table 5-6 on Page 5-13 lists an upper shelf energy of approximately 183 ft-lbs. while Table A-1 on

-

Page A-4 lists an upper shelf energy of approximately 164 ft-lbs. Which is correct?

R~es onse The average value of approximately 183 ft-lbs. for forging 125P666 is derived from Figure 5-1 on Page 5-8 of the Capsule Report. ,Unirradiated values of 202 ft-lbs., 186 ft-lbs., and 162 ft-lbs. are averaged to give 183 ft-lbs. The average value of approximately 183 ft-lbs. for forging 125P666 is correct and is consistent with values reported in Capsules V and T, as well as Capsule R. Table A-1 on Page A-4 of Capsule R was apparently copied from the report of Capsule V which has the same incorrect value of approximately 164 on Page 45.

5. question Page 4-1 of Capsule R Report appears to have the lower and.

intermediate forging numbers reversed.

~Res onse The sentence from Page 4-1 reads in part ...intermediate and lower shell forgings (heats 125P666 and 125S255).... This sentence is confusing and does have the numbers in the wrong order. This should read ...intermediate forging (125S255) and lower forging (125P666VA1).... The correct forging numbers are shown .on Ginna UFSAR Figure 5.3-2, which is attached.

6. question NRC requested a copy of Westinghouse Letter NSID WOG-37, dated November 16, 1982 to R. Mecredy regarding Ginna Reactor Vessel Material.

~Res onse Letter attached.

7. ~estion What material is limiting?

R~es nse As stated in Section 5.3.1.2 of the Ginna UFSAR, the beltline weld, SA-847, is limiting. The location of this weld is shown on the attached Figure 5.3-2 of the Ginna UFSAR.

8. Question Forging. 125S255 has showed no shift in RT for Capsule R.

R~es nse There was a 0 F shift in RT for Capsule R, this was also true for Capsule V and Capsule T. RG&E can only speculate'as to the reason for the O'F shift, however, it appears that for some reason this material is insensitive to radiation.

Westinghouse Water Reactor Nuclear Service Oivision Electric Corporation Divisions Box 2728 Pittsburgh Pennsylvania 15230 November 16, 1982 NSID/WOG-37 Dr. Robert Mecredy Rochester Gas 8 Electric 89 East Avenue Rochester, New York 14649

Dear Dr. Mecredy:

C Robert E. Ginna Reactor Vessel Material Attached is a complete specification of the base material and weld material for the Robert E. Ginna reactor pressure vessel. The source of this data is the original material certification supplied by Babcock and Wilcox with the vessel. This is the data forming the basis for the Westinghouse RTNDT calculations.

Very truly yours, R. A. Muench, Manager Westinghouse Owners Group

/elr Attachment cc: G. W. Dillon

ROBERT E. GINNA UNIT NO. 1 REACTOR YESSEL MATERIAL'SURVEILLANCE PROGRAM 1.) The estimated maximum fluence (E > 1 Mev) at the inner surface of the reactor vessel wall as of March 31, 1977 is 5.26 x 10 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 v'ess'el was performed by Babcock & 'Llilcox Co.

4.) a.) Sketch of 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.

FIGURE 1 IDENl'IFICA ~ AND IOCATION OF DELTLINE REG MATERIAL ROBERT E. GINNA UNIT NO. 1 REACIOR VESSEL FORGING 123PIIGVAI SA-1101 UJ I FORGING 125S255VA1 C5 UJ CL UJ I

CORE SA-847 (LIMITING WELD)

FORGING 125P666VA1 ROCHESTER GAS AND ELECTRIC CORPORATION R.E. GINNA NUCLEAR POWER PLANT UPOATED FINAL SAFETY ANALYSIS REPORT Figure 5.3-2 Identification and Location of Beltline Region Material

I TABLE 1 IDENTIFICATION Of REACTOR VESSEL BELTLINE REGION WELD MATERIAL e Weld Wire Flux Wel d Weld Location Weld Process Control Ho. ~T e Heat Ho. ~T e Lot Ho. Post Weld Heat'Treatment Nozzle Shell- to Submerged Arc :SA-'101 Hn-Ho-Ni 71249 .Linde 80 8445 1100-1125'F-48 Hrs.-FC Inter. Shell Inter. Shell to Submerged Arc SA-847 . Hn-Mo-Ni 61782 Linde 80 8350 1100-1125'F-48 Hrs.-FC Lower Shell I Surveillance Weld Submerged Arc SA-1036 Hn-Ho-Ni 61782 Linde 80 8436 1100'F-ll-l/4 Hrs.-f TABLE 2 WELD MATERIAL CHEMICAL COMPOSITION Weld Wire Flux lle1 ht Percent Control No. ~T e Heat Ho. ~Te: Lot No. C ~

P S Hn Si Ho Ni Cr Cu SA-1101 Mn-Ho-Ni 71249 Linde 80 "8445 .070 .021 '014 1.28 .52',36 .57 .17 .21 SA-847 Hn-Ho-Ni 61782 Linde 80 8350 .082 .012 i .012 1.34 .45 .39 .39 .06 .20 Surveillance Weld .075 .012 .016 1.31 .59 .36 .56 .59 .23

I TABLE 3 MECHANICAL PROPERTIES OF N/LO MATERIAL Weld Wire Flux Energy Shel f Weld THDT at 10'F RTNP T Energy YS UTS El ong. RA Control No. ~T e Heat No. ~Te ~

'Lot No. 'F ft-Ib OF ft-Ib ksi ksi x SA-1484 Mn-Mo-Hi 71249 Linde 80 8445 0* 45, 45, 46 0* 68.63 84.26 28.5 SA-1101 Hn-Mo-Ni 61782 Linde 80 8350 0* 58i 60, 36 0* 67.00 81.88 29.5 Surveillance Meld ~ ~

P 0* 54, 66.5, 71** 0* 79.0 73.52 87.35 '2.8 62.0

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

• • Energy at 60'F . I TABLE 4 MAXIMUM'END OF LIFE FLUENCE AT VESSEL WALL'LOCATIONS Fluence n/cm Nozzle Shell to Inter. Shell Weld ~2.0 x 10 Inter. Shell to Low Shell Weld 3.7 x 10 18 Nozzle Shell Forging 123P118VAl ~2.0 x 10 Inter Shell Forging 125S255VAl 3.7 x 10 Lower Shell Forging 125P666VAl 3.7 x, 10

TABLE 5

.IDENTIFICATION OF REACTOR VESSEL BELTLINE FORGING MATERIAL Heat Treatment Forging Material

~Com onent No. Heat No. , ~S ec. ~Su 1( er . Austenitize ~Tem er Stress Relief Nozzle Shell 123P118YAl 123P118 A336 Bethlehem Steel 1550'F-ll Hrs-NQ 1220'F-22 Hrs-AC 1125'F-30 Hrs-FC Inter. Shell 125S255VAl 125S255 A508 CL2 Bethlehem Steel 1550'F-15-1/2 Hrs-llQ 1210'F-18 Hrs-AC 1125'F-30 Hrs-FC Lower Shell 125P666VA1 125P666 A508 CL2 Bethlehem Steel 1550'F-9 Hrs-MQ 1220'F-12 Hrs-AC 1125'F-30 Hrs-FC Surveillance 125S255VA1 '25S255 A508 CL2 Bethlehem Steel 1550'F-15-1/2 Hrs-llQ 1210'F-18 Hrs-AC 1100'F-ll-l/4 Hr C Forgings 125P666VA1 125P666 A508 CL2 Bethlehem Steel 1550'F-9 Hrs-.AC 1220'F-12 Hrs-AC 1100'F-11 Hrs-FC TABLE 6 BELTLINE FORGING MATERIAL CHEMICAL COMPOSITION

'Hei ht Percent For in No. C P ~

S t1n Si: Mo Ni Cr -Cu V 123P118YA1 .19 .010 . 009 .65 .23 .60 .69 .42 125S255VA1 .18 .010 ..007 .66 23, .58 .69 .33 .07 .02 125P666VAl .19 .012 .bll .67 .20( .57 .69 .37 .05 .02 I

I I

TABLE 7 MECHANICAL PROPERTIES OF BFLTLINE FORGINGS

>'..1.l f11i, ~

Upper Shel f RTNPT En'er gy Elong.

'or NDT YS UTS RA in No. oF "oF ft-'1b ksi I x 123P118YA1 40 4p* 117* 66.87 88.00 25.50 73.50 125S255YA1 20 20* 106* 67.25 88.25 26.25 70.10 125P666VA1 40 40* 114* 63.50 85,00 26.25 71. 05 f'i) 125S255VA1 20 20*. 91* 78. 22 97.19 23.30 66.85 Surveillance 125P666VA1 '40k 120* 62.72 83. 65 26.35 7p 75 Test Resul ts 4p

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