ML20029E745

Requests Confirmation of plant-specific Applicability to Topical Repts BAW-2178P & BAW-2192P by Comparing Limiting Matl Properties for Davis-Besse Nuclear Power Station to Values Reported in Topical Repts,Per GL 92-01

ML20029E745
Person / Time
Site:	Davis Besse
Issue date:	05/13/1994
From:	West G Office of Nuclear Reactor Regulation
To:	Shelton D CENTERIOR ENERGY
References
GL-92-01, GL-92-1, TAC-M83732, NUDOCS 9405200035
Download: ML20029E745 (8)

Text

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3 sSS REGO ye

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f'lS UNITED STATES NUCLEAR REGULATORY COMMISSION g-j#

WASHINGTON, D.C. 20555-0001 Bhy 13,1994 Docket No. 50-346 Mr. Donald C. Shelton Senior Vice President, Nuclear Centerior Service Company c/o Toledo Edison Company Davis-Besse Nuclear Power Station 5501 North State Route 2 Dak Harbor, Ohio 43449

SUBJECT:

DAVIS-BESSE NUCLEAR POWER STATION - GENERIC LETTER (GL) 92-01, REVISION 1, " REACTOR VESSEL STRUCTURAL INTEGRITY," (TAC NO. M83732)

By letter dated July 1,1992, Toledo Edison Company provided its response to GL 92-01, Revision 1.

The NRC staff has completed its review of your response.

Based on its review, the staff has determined that Toledo Edison Company has provided the information requested in GL 92-01.

The GL is part of the staff's program to evaluate reactor vessel integrity for Pressurized Water Reactors (PWRs) and Boiling Water Reactors (BWRs).

The information provided in response to GL 92-01, including previously docketed information, is being used to confirm that licensees satisfy the requirements and commitments necessary to ensure reactor vessel integrity for their facilities.

A substantial amount of information was provided in response to GL 92-01, i

Revision 1.

These data have been entered into a computerized database j

designated the Reactor Vessel Integrity Database (RVID).

The RVID contains the following tables: A pressurized thermal shock (PTS) table for PWRs, a pressure-temperature limits tablo for BWRs and an upper-shelf energy (USE) table for PWRs and BWRs. provides the PTS table, Enclosure 2 provides the USE table for your facility, and Enclosure 3 provides a key for the nomenclature used in the tables. The tables include the data necessary to perform USE and RT,iously docketed information.

evaluations. These data were taken from your response to GL 92-01 and prev References to the specific source of the data are provided in the tables.

We request that, within 30 days, you provide confirmation of the plant-specific applicability of the Topical Reports BAW-2178P and BAW-2192P and submit a request for approval of the topical reports as the basis for demonstrating compliance with 10 CFR Part 50, Appendix G, Paragraph IV.A.I.

To demonstrate that the topical reports are applicable to Davis-Besse Nuclear Power Station, you must compare the limiting material properties for the Davis-Besse Nuclear Power Station vessel to the values reported in the topical reports. This review will be a plant-specific licensing action. We further T>

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l request that you verify that the information you have provided for your facility has been accurately entered in the summary data file.

If no comments j

are made in your response to the last request, the staff will use the j

information in the tables for future NRC assessments of your reactor pressure j

vessel. Once your confirmation of the applicability of the topical reports j

and request for approval are received, the staff will consider your actions related to GL 92-01, Revision 1, to be complete.

I The information requested by this letter is within the scope of the overall i

burden estimated in GL 92-01, R*. vision 1, " Reactor Vessel Structural 1

Integrity, 10 CFR 50.54(f)." The estimated average number of burden hours is j

200 person hours for each addressee's response.

This estimate pertains only 3

to the identified response-related matters and does not include the time i

required to implement actions required by the regulations. This action is i

covered by the Office of Management and Budget Clearance Number 3150-0011, l'

which expires June 30, 1994.

j Sincerely, 2

l ORIGINAL SIGNTl) BY l

Garmon West, Jr., Acting Project Manager i

Project Directorate III-3 Division of Reactor Projects III/IV d

Office of Nuclear Reactor Regulation

Enclosures:

1.

Pressurized Thermal Shock Table 2.

Upper-Shelf Energy Table 3.

Nomenclature Key 1

f cc w/ enclosures:

j See next page f

DISTRIBUTION LDocket; File:

Local & NRC PDRs D. Mcdonald 3

PDIII-3 r/f J. Roe E. Hackett J. Zwolinski J. Hannon K. Bohrer 2

J. Hopkins M. Rushbrook OGC ACRS(10)

E. Greenman, RIII G. West

.l OFFICE LA:PDIII-)

(A)PM:PDIII-3 PM:PDIII-3 PD : PDI_IIs-3 MRus M b GWest:dy JHopkins N Jdnnon NAME kj 94 9 //l/94

[ /p /94 I / O/94 DATE i

0FFICIAL RECORD COPY FILENAME: G:\\DAVISBES\\DBM83732.LTR i

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l Mr. Donald C. Shelton Davis-Besse Nuclear Power Station l

Toledo Edison Company Unit No. 1 cc:

Mary E. O'Reilly Attorney General Centerior Energy Corporation Department of Attorney General j

300 Madison Avenue 30 East Broad Street i

Toledo, Ohio 43652 Columbus, Ohio 43216 Mr. William T. O'Connor, Jr.

Mr. James W. Harris, Director Manager '- Regulatory Affairs Division of Power Generation Toledo Edison Company Ohio Department of Industrial i

Davis-Besse Nuclear Power Station Regulations 1

5501 North State - Route 2 P. O. Box 825 Oak Harbor, Ohio 43449 Columbus, Ohio 43216 4

Gerald Charnoff, Esq.

Ohio Environmental Protection Agency j

Shaw, Pittman, Potts DERR--Compliance Unit l

and Trowbridge ATTN:

Zack A. Clayton 2300 N Street, N. W.

P. O. Box 1049 Washington, D. C. 20037 Columbus, Ohio 43266-0149 Regional Administrator, Region III State of Ohio

{

U. S. Nuclear Regulatory Commission Public Utilities Commission j

801 Warrenville Road 180 East Broad Street Lisle, Illinois 60532-4351 Columbus, Ohio 43266-0573 3

Mr. Robert B. Borsum Mr. James R. Williams Babcock & Wilcox State Liaison to the NRC Nuclear Power Generation Division Adjutant General's Department 1700 Rockville Pike, Suite 525 Office of Emergency Management il Rockville, Maryland 20852 Agency 2825 West Granville Road Resident Inspector Columbus, Ohio 43235-2712 U. S. Nuclear Regulatory Commission 5503 N. State Route 2 j

Oak Harbor, Ohio 43449 j

Mr. John K. Wood, Plant Manager i

Toledo Edison Company i

Davis-Besse Nuclear Power Station 5501 North State Route 2 Oak Harbor, Ohio 43449 Robert E. Owen, Chief Bureau of Radiological Health Services Ohio Department of Health Post Office Box 118 Columbus, Ohio 43266-0118 4

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ENCLOSURE 1 l

Summary File for Pressurized Thermal Shock i

f Plant Beltline Nest No.

ID Neut.

147 Nethod of Chemistry Method of

%Cu

%Ni j

kame Ident.

Ident.

Fluence et Determin.

Factor

Determin, j

EOL/EFPY 1 R Y.,

CF Davis-Nozzle 123Y317 1.50E18 50*F Plant 26 Table 0.04 0.68 Besse Belt (ADB 203)

Specific a

i Forging 1

l EOL Upper 123x244 1.07E19 20*F Plant 26 Table 0,04 0.77 4/22/2017 shett (AKJ 233) specific Forging Lower

$P4086 1.07E19 50*F Plant 20.00 Calculated 0.02 0.81

]

Shell (BCC 241)

Specific 4

Forging i

Upper /

821764 1.07E19 2*F Plant 162.09 Calculated 0.24 0.63 i

Lower Specific

$ hell e

j Cire. Weld i

WF 182 1 i

f Nortle T29744 5'F Generic 203.4 Table 0.29 0.68 i.

8elt/ Upper l

Shelt i

Cire. Weld I

(outside 91%)

l WF 233 1

Nozzle 873914 1.50E18 5'F Generie 160.6 Table 0.18 0.64 l

Belt /Umer i

Shell Cire. Weld l

(inside j

g M)

WF 232

'l j

References 5

1 The chemical corposition, fluence, and IRT. date are f rom July 1,1992, letter f rom D. C. Shelton (TEco) to Usuc i

i Docant Control Desk, subject: Response to NRC Generic Letter 92 01, Revision 1, Reactor vessel structural Integrity, j

for the Davis Besse Nuclear Power $tetion.

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ENCLOSURE 2 l

Sumary File for Upper Shelf Energy Plant Name Belttine Heat No.

Material 1/4T USE 1/47 Unirred.

Method of Ident.

Type at EOL Weutron USE Determin.

Fluence at Unirred.

EOL USE Davis-Nozzle 123Y317 A $08 2 119 9.0E17 134 Direct Beste Belt (ADI 203)

Forging EOL:

Upper 123x244 A 508 2 119 6.43E18 144 Direct 4/22/2017 shett (AKJ 233)

I Forging Lower 5P4086 A 508 2 102 6.43E18 118 Direct shett (BCC 241)

Forging upper /

821T44 Linde 80, 66 6.43E18 81 Direct Lower SAW Shell Cire. Weld WF 182 1 Nozzle T29744 Linde 80, EMA' 9.0E17 EMA*

Generic Belt / Upper SAW i

Shell Circ. Weld 4

(outside 91%), WF-233 Norste 8T3914 Linde 80, EMA' 9.0E17 EMA' Generic Belt / Upper

$AW Shell Cire. Weld (inside 9%), WF-232 References The cheni st cocposit'on, and fluence data are from July 1,1992, letter from D. C. Shelton (TEco) to UShRC DocmA Control Desk, s@ ject: Response to NRC Generic Letter 92 01, Revision

1. Reactor Vessel Structuret Integrity, for the Davis Besse Nuclear Power Station.

The UUSEs for forgings ADB 203, SCC 241 and AKJ 233 med weld WF 182-1 are from BAW 2125, which described the analysis of surveillance capsute D.

1 2Licensee must confirce applicability of Topical Reports BAW-2178P and BAW-2192P

l l

ENCLOSURE 3 l

N0MENCLATURE Pressurized Thermal Shock Table l

Column 1:

Plant name and date of expiration of license.

i Column 2:

Beltline material location identification.

Column 3:

Beltline material heat number; for some welds that a single-l wire or tandem-wire process has been reported, (S) indicates single wire was used in the SAW process, (T) indicates tandem wire was used in the SAW process.

Column 4:

End-of-life (E0L) neutron fluence at vessel inner wall; cited directly from inner diameter (ID) value or calculated by using Regulatory Guide (RG) 1.99, Revision 2, neutron fluence attenuation methodology from the quarter thickness (T/4) value reported in the latest submittal (GL 92-01, PTS, or P/T limits submittals).

Column 5:

Unirradiated reference temperature.

Column 6:

Method of determining unirradiated reference temperature (IRT).

Plant-Soecific This indicates that the IRT was determined from tests on material removed from the same heat of the beltline material.

MTEB 5-2 This indicates that the unirradiated reference temperature was determined from following MTEB 5-2 guidelines for cases where the IRT was not determined using American Society of Mechanical Engineers Boiler and Pressure Vessel Code,Section III, NB-2331, methodology.

Generic This indicates that the unirradiated reference temperature was determined from the mean value of tests on material of similar types.

Column 7:

Chemistry factor for irradiated reference temperature evaluation.

Column 8: Method of determining chemistry factor.

Table This indicates that the chemistry factor was determined from 1.

+he chemistry factor tables in RG 1.99, Revision 2.

Calculated This indicates that the chemistry factor was determined from surveillance data via procedures described in RG 1.99,

. Revision 2.

l 1 Column 9:

Copper content; cited directly from licensee value except when more than one value was reported.

(Staff used the average value in the latter case.)

No Data This indicates that no copper data has been reported and the default value in RG 1.99, Revision 2, will be used by the staff.

Column 10: Nickel content; cited directly from licensee value except when more than one value was reported.

(Staff used the average value in the latter case.)

d No Data i

This indicates that no nickel data has been reported and the default value in RG 1.99, Revision 2, will be used by the staff.

Upper Shelf Energy Table Column 1:

Plant name and date of expiration of license.

Column 2:

Beltline material location identification.

Column 3:

Beltline material heat number; for some welds that a single-wire or tandem-wire process has been reported, (S) indicates single wire was used in the SAW process.

(T) indicates tandem wire was used in the SAW process.

j Column 4: Material type; plate types include A 533B-1, A 3028, A 302B Mod., and forging A 508-2; weld types include SAW welds using Linde 80, 0091, 124, 1092, ARCOS-B5 flux, Rotterdam welds using Graw Lo, SMIT 89, LW 320, and SAF 89 flux, and SMAW welds using no flux.

Column 5:

E0L upper-shelf energy (USE) at T/4; calculated by using the E0L fluence and either the cooper value or the surveillance data.

(Both methods are described in RG 1.99, Revision 2.)

EMA This indicates that the USE issue may be covered by the approved equivalent margins analysis in the B&W Owners Group Topical Reports: BAW-2178P.and BAW-2192-P.

Column 6:

E0L neutron fluence at T/4 from vessel inner wall; cited directly from T/4 value or calculated by using RG 1.99, Revision 2, neutron fluence attenuation methodology from the ID value reported in the latest submittal (GL 92-01, PTS, or P/T limits submittals).

1 i

i*.

l 4 Column 7: Unirradiated USE.

US This indicates that the USE issue may be covered by the approved equivalent margins analysis in the B&W Owners Group Topical Reports: BAW-2178P and BAW-2192P.

i Column 8: Method of determining unirradiated USE.

Direct For plates, this indicates that the unirradiated USE was from i

a transverse specimen.

For welds, this indicates that the j

unirradiated USE was from test date, j

65%

This indicates that the unirradiated USE was 65% of the USE 4

j from a longitudinal specimen.

I Generic l

This indicates that the unirradiated USE was reported by the licensee from other plants with similar materials to the l

beltline material.

i 1

2 -

NRC aeneric This indicates that the unirradiated USE was derived by the staff from other plants with similar materials to the beltline material.

10. 30. 40. or 50 'F This indicates that the unirradiated USE was derived from Charpy test conducted at 10, 30, 40, or 50 'F.

Surv. Weld i

This indicates that the unirradiated USE was from the surveillance weld having the same weld wire heat number.

Eauiv. to Surv. Weld This indicates that the unirradiated USE was from the surveillance weld having different weld wire heat number.

Sister Plant i

This indicates that the unirradiated USE was derived by using j

the nported value from other plants with the same weld wire heat number.

3 Blank Indicates that there is insufficient data to determine the unirradiated USE.

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