ML20070N803

Provides Response to NRC 940317 Request That Util Confirm Applicability & Accuracy of Info Previously Provided by Util & B&Wog Re GL 92-01,Rev 1, Reactor Vessel Structural Integrity

ML20070N803
Person / Time
Site:	Point Beach
Issue date:	05/02/1994
From:	Link B WISCONSIN ELECTRIC POWER CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
CON-NRC-94-037, CON-NRC-94-37 GL-92-01, GL-92-1, VPNPD-94-051, VPNPD-94-51, NUDOCS 9405090224
Download: ML20070N803 (18)

Text

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. i Wisconsin Electnc POWER COMPANY 231 W Mchgan, PO Box 2046. Mdwoukee.WI 53201-2046 (414)221 2345 VPNPD-94-051 NRC-94-037 May 2, 1994 l Document Control Desk  !

U.S. NUCLEAR REGULATORY COMMISSION i Mail Station P1-137 l Washington, DC 20555 Gentlemen; DOCKETS 50-266 AND 50-301 GENERIC LETTER 92-01, REVISION 1, i

" REACTOR VESSEL STRUCTURAL INTEGRITY" l POINT BEACH NUCLEAR PLANT, UNITS 1 AND ?

Nuclear Regulatory Commission (NRC) Generic Letter (GL) 92-01,  !

" Reactor Vessel Structural Integrity," dated March 6, 1992, was issued to obtain information from licensees to enable the NRC to assess compliance with regulatory requirements and commitments regarding reactor vessel integrity. Our response to GL 92-01 was provided to the NRC on June 25, 1992, and supplemented on July 30, 1992, and November 1, 1993 (VPNPD-93-186). On March 17, 1994, the NRC requested that Wisconsin Electric (WE) confirm the applica-bility and accuracy of information previously provided by WE and the Babcox & Wilcox Owners Group. This letter provides our response to your request.

In letters dated May 21, 1993, and November 1, 1993 (VP-NPD 185), WE referred to B&W Nuclear Technologies (BWNT) topical reports BAW-2178P and BAW-2192P, respectively, for Point Beach Units 1 and 2. These topical reports were previously submitted to the NRC by BWNT on behalf of the B&W Owners Group. We have reviewed and confirmed the accuracy of the limiting material properties in these reports that applies to the Point Beach units.

However, we noted during our review that the cold leg temperature for the Point Beach units was incorrectly listed as 552.5"F in BAW-2178P instead of 542 F. Wisconsin Electric has confirmed with BWNT that Point Beach Units 1 and 2 remain bounded by the fracture mechanics analyses presented in BAW-2178P at this lower operating temperature (Attachment 1). The introduction to the final version of BAW-2178P will include reference to the correct cold leg temperature as shown in Attachment 1. We request that the NRC review and approve the final versions of BAW-2178P and BAW-2192P as the basis for demonstrating compliance with 10 CFR Part 50, Appendix G, Paragraph IV.A.1 for Point Beach Units 1 and 2.

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Document Control Desk May 2, 1994 Page 2 l

Wisconsin Electric has also reviewed the information contained in Enclosures 1 and 2 of your March 17, 1994, letter. We have confirmed the accuracy of the material chemistry and property data as presented in your enclosures. However, since the time of the submittal of our response to GL 92-01, we have performed a further analysis of reactor vessel dosimetry as part of our reactor cavity neutron monitoring program. The results of this analysis (Westinghouse WCAP-12794, Rev. 2 and WCAP-12795, Rev. 2) provided updated fluence values that differ slightly from those provided in Enclosures 1 and 2 of your letter. The most current end-of-license fluence projections for Point Beach Units 1 and 2 are presented in Attachment 2.

Please contact us if you have any questions or require additional information regarding this response.

Sincerely,

',./

p A Q *'

Bob Link Vice President Nuclear Power JRP/kmc Attachments: 1) BWNT Letter to Wisconsin Electric dated 4/25/94

2) Fluence Predictions for the Point Beach Units "

at 32 EPPY cc: NRC Regional Administrator, Region III NRC Resident Inspector

_ _ _ _ - - - 1

BW B&WNUCLEAR TECHNOLOGIES _

3315 Oldfarvst Road D1.1. Box 10935 Lynchburg, VA 24506-0935 Telephone: 804-385-2000 Telecopy: 804-385-3663 April 25, 1994 Mr. J. R. Pfefferle UIMb-Wisconsin Electric Power Company g h3- -

231 W. Michigan St.

P. O. Box 2046 Milwaukee, WI

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43201

Dear Jim:

In BAW-2178P report, the cold leg temperatures of Point Beach Units 1 and 2 are listed as 552.5F referring to BAW-1543, Rev.3.

According to Rev.4 of BAW-1543, these temperatures should be 542F. This correction is made to the approved version of this report as shown in the attached. This temperature doe not affect the results of this analysis since Point Beach vessels were not used for a lower-bounding analysis.

Attached page shows an addition to the introduction of BAW-2178P.

This will be issued as BAW-2178PA as the approved version of the report showing the NRC SER in the report.

Sincerely, Kenneth K con

Attachment:

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Addition to the end of INTRODUCTION section on page 1-1.

Following receipt of the NRC SER, this report is reissued as BAW-2178PA with a typographical error correction in paragraph 4.1 and the changes in cold leg temperature column of Table 5-1 to reference a later version of BAW-1543 (Revision 4), to be consistent with BAW-2192PA. These changes do not affect the results of this analysis.

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l Attachment 2 Fluence Predictio'ns for Point Beach Unit 1 at 32 EFPY W

Beltline Heat No. ID Neutron 1/4T Neutron Identification Ider tification Fluence at EOL Fluence at EOL sm (n/cm2) (n/cm) 2 Nozzle Belt 1222237VA1 3.17E18 2.15E18 Forging Int. Shell A-9811-1 2.78E19 1.88E19 Plate Lower Shell C-1423-1 2.43E19 1.65E19 Plate Int. Shell 1P0815 1.78E19 1.21E19 Axial Welds SA-812 Lower Shell 61782 1.63E19 1.10E19 Axial Welds SA-847 Circ. Weld 71249 2.43E19 1.65E19 SA-1101 Nozzle Belt to 8T1762 3.17E18 2.15E18 Int. Shell Cire. Weld SA-1426

t Attachment 2 Fluence Predictions for Point Beach Unit 2 at 32 EFPY Beltline Heat No. ID Neutron 1/4T Neutron Identification Identification Fluence at EOL Fluence at EOL 2

2 (n/cm ) (n/cm)

Nozzle Belt 123V352VA1 3.70E18 2.51E18 Forging Int. Shell 123V500VA1 2.88E19 1.95E19 Forging Lower Shell 122W195VA1 2.62E19 1.77E19 Forging Int./ Lower 72442 2.52E19 1.71E19 Shell Weld SA-1484 Nozzle Belt to CE Weld 3.70E18 2.51E18 Int. Shell Circ. Weld J

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• • • • • March 17, 1994 h$@ q2 -01~O' Docket Nos. 50-266 and 50-301 W &/

KvA kd Mr. Robert E. Link, Vice President Nuclear Power Department Wisconsin Electric Power Company 231 West Michigan Street, Room P379 Milwaukee, Wisconsin 53201

SUBJECT:

GENERIC LETTER (GL) 92-01, REVISION 1, " REACTOR VESSEL STRUCTURAL INTEGRITY," POINT BEACH, UNITS 1 AND 2, (TAC NOS. M83737 AND M83738)

By letter dated June 25, 1992, as supplemented July 30, 1992, and November 1, 1993, you provided your response to GL 92-01, Revision 1. The NRC staff has completed its review of your response and determined that: (1) based on the available surveillance data, the Point Beach reactor vessels will be below the pressurized thermal shock (PTS) screening criteria in 10 CFR 50.61 when their operating licenses expire; and (2) based on the analyses in BAW-2192 and BAW-2178P, the Point Beach reactor vessels will be able to satisfy the upper shelf energy (USE) requirements of 10 CFR 50, Appendix G, throughout the term of their operating licenses.

The GL is part of the staff's program to evalwe 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 and permittees 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, Revision 1. These data have been entered into a computerized data base designated Reactor Vessel Integrity Database (RVID). The RVID contains the following tables: A PTS table for PWRs, a pressure-temperature limit table for BWRs and a USE table for PWRs and BWRs. Enclosure 1 provides the PTS and/or pressure temperature tables, Enclosure 2 provides the USE tables for your facilities, and Enclosure 3 provides a key for the nomenclature used in the tables. The tables include the data necessary to perform USE, pressure-temperature limit, and RT g evaluations. These data were taken from your response to GL 92-01 and previously docketed information. The information in the RVID for your facilities will be considered accurate at this point in time, and will be used in the staff's assessments related to vessel structural integrity. References to the specific source of the data are provided in the tables.

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I We request that you confirm within 30 days the plant-specific applicability of 3 the NRC staff approved revisions of the topical reports BAW-2192 and ,i

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l BAW-2178P, and request review and approval in accordance with 10 CFR Part 50, Appendix G. This review will be a plant-specific action. We further request i that you verify that the information you have provided for your facilities has y been accurately entered in the data base. If no comments are made in your l response to the last request, the staff will use the information in the tables -y l for future NRC assessments of your reactor pressure vessel. Once your confirmation is received, the staff will consider your actions related to GL 92-01, Revision 1, to be complete.

If you desire to use surveillance data from another plant to determine the chemistry factor and margin value for welds in your beltline, you must compare the actual beltline irradiation temperatures (cold leg temperatures) of the

, two plants to determine whether a temperature correction is required. In addition, the data must: (1) be from a weld fabricated using weld wire with the same heat number and with the same type of flux as the beltline weld; and, (2) meet the credibility criteria of Regulatory Guide 1.99, Revision 2 (the l scatter of the data about the best-fit line should normally be less than 28 F).

l The information requested by this letter is within the scope of the overall burden estimated in Generic Letter 92-01, Revision 1, " Reactor Vessel

, 3tructural Integrity,10 CFR 50.54(f)." The estimated average number of l burden hours is 200 person hours for each addressee's response. This estimate pertains only to the identified response-related matters and does not include the time required to implement actions required by the regulations. This action is covered by the Office of Management and Budget Clearance Number 3150-0011, which expires June 30, 1994.

Sincerely, f w. I s.y .,

Allen G. Hansen, Project Manager Project Directorate III-3 Division of Reactor Projects Ill/IV/V Office of Nuclear Reactor Regulation

Enclosures:

! 1. Pressurized Thermal Shock or Pressure-Temperature Limit Tables

2. Upper-Shelf Energy Tables

3. Nomenclature Key cc w/ enclosures: 1 See next page i 1

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Mr. Robert E. Link Point Beach Nuclear Plant Wisconsin Electric Power Company Unit Nos. I and 2 l

cc:

Ernest L. Blake, Jr. 1 Shaw, Pittman, Potts & Trowbridge 1 2300 N Street, N.W. l Washington, DC 20037 .

l Mr. Gregory J. Maxfield, Manager l l

Point Beach Nuclear Plant '

Wisconsin Electric Power Company 6610 Nuclear Road Two Rivers, Wisconsin 54241 Town Chairman Town of Two Creeks ,

l Route 3 l l Two Rivers, Wisconsin 54241 l

l Chairman Public Service Commission of Wisconsin Hills Farms State Office Building Madison, Wisconsin 53702 l i

Regional Administrator, Region III U.S. Nuclear Regulatory Commission 801 Warrenville Road ,

Lisle, Illinois 60532-4351  !

\

Resident Inspector's Office U.S. Nuclear Regulatory Commission 6612 Nuclear Road i l

Two Rivers, Wisconsin 54241 '

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ENCLOSURE 1 Summary File for Pressurized Thermal Shock Plant Beltline Heat No. 10 Neut. IRT Method of Chemistry Method of %Cu %Ni Name Ident. Ident. Fluence at Determin. Factor Determin.

EOL/EFPY I R T, CF Point Nozzle 122P237VA1 2.95E18 50*F Plant 115.2 Table 0.11 0.82 BQach 1 Belt Specific Forging l Ect. : Int. Shell A 9811-1 2.68E19 1*F Generic 92.556 Calculated 0.20 0.056 10/5/2010 Plate Lower C-1423 1 2.33E19 1'F Generic 47.466 Calculated 0.12 0.065 Shell Plate Int. Shell ;PC815 1.71E19 5'F Generic 138.2 Yable 0.17 0.52

, Axial Welds SA-812 tower 61782 1.56E19 5'F Generic 167.6 Table 0.25 0.54

, $hett Axial Welcs

$A 847 Cire. Weld 71249 2.33E19 10'F Plant 180.0 Table 0.26 0.60 54 1101 soecific Nezzle 871762 2.95E18 -5'F Generic 152.25 Table 0.20 0.55 Belt to In: Shell Cire. 'wetd SA '426 Seferences l

l Chemical composition, fluence, and IRt.., data are from June 25, 1992, letter from B. Link (WEPCo) to USNRC Doe n nt Controt Oesk, subject: Response to Generic Letter 92-01, Revision 1, Reactor vessel structural Integrity, 10 CFR l 50.54(f)  !

4' IR4, f er plates A 9811-1 and C-14231 were determined per pp 318 of BAW 10042P. The values are conservative relative 4

to MTES 5-2.

Percent coeper for Nozzle Belt. Forging determined by the NRC staff from data from similar forgings in letter dated hove-cer 1, 1993. The value is a tolerance limit with 95 percent confidence that at least 95 percent of the peputation is less than the tolerance limit (TL).

( TL = 2+Ka , where R=0 . 0 6 , K=3 .187 , a=0.015) 4

Summary File for Pressurized Thermal Shock Plant Bettline Heat No. ID Neut. I R T, Method of Chevni s try Method of 10u %Ni herre Ident. Ident. Fluence at Determin. Factor Determin.

EOL/EFPY IRT., CF Point Negate 123v352vA1 3.50E18 40'F Plant 113.25 Table 0.11 0.73 Beach 2 Bett Specific Forging EOL: Int. Shell 123v500VA1 2.92E19 40'F Plant 49.771 Calculated 0.09 0.70 3/8/2013 Forging specific Lowe* 122W195vA1 2.66E19 40'F plant 28.63 Calculated 0.05 0.72

$helt Specific Forging Int./ Lower 72442 2.56E19 5'F Generic 1 73 Table 0.24 0.60 shett

$A 1484 hettle CE Weld 3.50E18 56'F Generie 232.5 Table 0.27 0.90 Belt to Irt Shett Cire weld

1. 0'e*eares Chemical corposition. fluence, and Iti, data are f rom June 25, 1992, letter f rom B. Link (WEPCo) to USNRC Occunent Controt Desu, subject: Response to Generic Letter 92 01, Revision 1, Reactor vessel Structural Integrity,.10 CFR 50.5 (f)

Peaceat cepc<r fer hozzle Belt. Forging determined by the NRC staff from data from similar forgings in letter cated hovemer 1, 1993. The value is a tolerance limit mitn 95 percent conficence that at least 95 percent of the population is less ' Fan the tolerance limit (TL).

( ~'I = 2+ Ko , wh er e 2= 0 . 0 6 , K= 3 .187 , o=0.015) ll

1

. ENCLOSURE 2 Sumary File for Upper Shelf Energy l Plamt kame Beltline Heat No. Material 1/4T USE 1/47 Unirred. Method of ident. Type at EOL Neutron USE Determin.

Fluence at Unirrad.

EOL USE Point Nozzle 122P237vA1 A 508 2 51 2.0E18 59 NRC i Beach 1 Belt Generic Forging )

l EOL: Int. Shell A-9611 1 A 302B 54 1.81E19 70 65%

10/5/2010 Lower C 1423 1 A 3028 60 1.58E19 77 65%

shell Int. Shell 1P0815 Lirde 80, E MA' 1.16E19 EMA' Generic Axial SAW Welds SA 812 Lo.ee 61782 Linde 80, EMA' 1.06E19 E MA' Generic Smell SAW Axial wetos SA-847 Cire. Weld 71249 Lirce 80, E MA' 1.58E19 EMA' Generic SA-1101 SAW hozzle 871762 Lirce 30, EMA' 2.0E18 E MA' Generic Belt / Int. SAw Shell Cire. held

, SA + 1I.26 sede*emees

% Orco in USE f or plate A 9311 1 determined f rom surveillance data in accordance with RG 1.99, Rev. 2, paragaspn 2.2.

Cremcal corocsition and fluence data are f rom June 25, 1992, letter frcn B. Link (WEPCo) to USk2C Ooe n nt Control Desk, sucject: Response to Generic Letter 92 01, Revision 1, Reactor vessel Structur al Integrity, 10 CFR 50.54(f) 4 UUSE data for weld 54 8!.7 are from Jme 25, 1992, letter from B. Link (WEPCo) to USERC Doctrwnt C: ate:! Cesk, sebject: Response to Generic Letter 92 01, Revision 1, Reactor vessel Structural Irtegrity, 10 CFR 50.54(f) ,

j UUSE f o plate C 1423 and A 9811-1 are f rom WCAP 10736 UUSE data for other welds are f rom BAW 1803, Revision 1 UUSE for hozzle Belt. Forging determined by using 65 percent correction factor on data from i

similar forging reported in a letter dated hovecter 1,1993 f rom B. Link (WIPCo) to UShtC. The l

UUSE value is a tolerance limit with 95 percent confidence that at least 95 percent of the peculation is greater than the tolerance lifnit (TL)

( ~'I = 2-Ko wh er e : 2=9 9, o = 12. 6 9, K = 3.187 )

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Licensee must confirm applicability of Topical Reports BAW 2192 and BAW 2178P.

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} l j Summary File for Upper Shelf Energy i

Plant Name Beltline Heat No. Material 1/47 USE 1/47 Unirred. Method of Ident. Type at EOL Neutron USE Determin.

lj Fluence at Unirred.

j EOL USE Point Nozzle 123v352 A 508 2 74 2.37E18 89 65%

1 Beach 2 Belt-Forging  !

EOL: Int. Shell 123v500VA1 A 508 2 108 1.98E19 117 65%

3/8/2013 i Lower 122W195VA1 A 508 2 85 1.80E19 94 65%

j shell 4 cire Weld 72442 Linde 80, E MA' 1.73E19 EMA' Generic i $A 148/. SAW hczzle Not Wo data 53 2.37E 18 75 WRC i Belt / Int, provided available Generic i Shell I Cire. Weld 1

s e der,-ees t

l Chemical corposition and fluence data are from June 25, 1992, letter from B. Link (WEPCo) to Ush4C Cocteent Control Desk, subject: Response to Generic Letter 92 01, Revision 1. Reactor vessel structural Integrity, 10 CFR 50.54(f)

UUSE data for weld SA 1484 are from Jtne 25, 1992, letter from 8. Link (WEPto) to UsWRC j OccJwnt Control Desk, subject: Response to Generic letter 92 01, Revision 1, Reactor vessel  :

Structural Integrity, 10 CFR 50.54(f) l 9 '

, UUSE for forging 122W195vA1 and 123v500VA1 is from BAW 2140, which analyzed capsule s 1

q UUSE for forging 123V352 reported in a letter dated Novenber 1, 1993 from B. Link (WEPCo) to i

USkRC.

l.

UUSE for hczzle Belt / Int. Shell Cire. Weld is the NRC staff value for Cereustion Engineering i' faericated welds that was reported in a letter dated Decencer 3, 1993 to T.L. Patterson (OPPD) l from S. Bloom (USWRC). '

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i Licensee must confirm applicability of Topical Reports BAW 2192 and BAW 2178P J

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i ENCL 0SURE 3 PRESSURIZED-THERMAL SH0CK TABLES AND USE TABLES NOMENCLATURE Pressurized Thermal Shock Table Column 1: Plant name and date~ of expiration of license.

Column ?: 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.

Column 4: End-of-life (E0L) neutron fluence at vessel inner wall; cited directly from inner diameter (ID) value or calculated by using 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 l This indicates that the unirradiated reference temperature was

! determined from following MTEB 5-2 guidelines for cases where l the IRT was not determined using ASME 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 I evaluation. I Column 8: Method of determining chemistry factor Table This indicates that the chemistry factor was determined from l the chemistry factor tables in RG 1.99, revision 2. l l

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l Calculated This indicates that the chemistry factor was determined from surveillance data via procedures described in RG 1.99, revision 2.

Column 9: Copper content; cited directly from licensee value except when more than one value were 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, Rev. 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).

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

Upper Shelf Energy Table 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-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.

l Column 4: Material type; plate types include A 5338-1, A 302B, 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.

l Column 5: E0L upper-shelf energy (USE) at T/4; calculated by using the

EOL fluence and either the copper value or the _ surveillance l data (both methods are described in RG 1.99, Revision 2.)

ffB l This indicates that the USE issue may be covered.by the approved equivalent margins analysis.

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.. . C'olumn 6: E0L neutron fluence at T/4 from vessel inner wall; cited J

directly from T/4 value or calculated by using RG.1.99, Revision 2 neutron fluence attenuation methodology from the ID 3 value reported in the latest submittal (GL 92-01, PTS, or P/T-limits submittals).

j Column 7: Unirradiated USE.

! EMA i .

I This indicates that the USE issue may be covered by the 7 approved equivalent margins analysis.

Column 8: Method.of determining unirradiated USE' 2

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Direct i This indicates'that the unirradiated USE was from a transverse j specimen, 65%

This indicates that the unirradiated USE was 65% of the USE from a longitudinal specimen.

j Generic l This indicates that the unirradiated USE was reported by the j licensee from other plants with similar materials to the j beltline material.

NRC aeneric This indicates that the unirradiated USE was derived by the

staff from other plants with similar materials to the l beltline material.

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

j i Surv. Weld 1

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

l Eaui. to Surv. Weld i This indicates that the unirradiated USE was from the surveillance weld having different weld wire heat number, i

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Sister Plant This indicates that the unirradiated USE was derived by using the reported value from other plants with the same weld wire heat number.

Blank This indicates that there is insufficient data to determine the unirradiated USE. These licensees may need to utilize an approved equivalent margins analysis to demonstrate USE compliance to Appendix G, 10 CFR 50.

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1 We request that, within 30 days, you provide confirmation of the plant-i specific applicability of the Babcock & Wilcox 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 2

IV.A.1. To demonstrate that the topical reports are applicable to , you must compare the limiting material properties of the M reacto sselrto

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