ML11223A012

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Calculation OSC-9863, Oconee, Units 1, 2 and 3 - License Exemption Using BAW-2308
ML11223A012
Person / Time
Site: Oconee  Duke Energy icon.png
Issue date: 09/22/2009
From:
Duke Energy Carolinas
To:
Office of Nuclear Reactor Regulation
References
BAW-2308 86-9109752-000, OSC-9863
Download: ML11223A012 (28)
v • d • e


Text

ATTACHMENT CALCULATION OSC-9863 Oconee Units 1, 2 and 3 License Exemption Using BAW-2308

Form 101.1 (R06.09)

CERTIFICATION OF ENGINEERING CALCULATION Station and Unit Number. Oconee Nuclear Station Units 1, 2 and 3 Title of Calculation: Oconee Units 1, 2 and 3 License Exemption Using BAW-2308 (AREVA Document No. 86 - 9109752 - 000)

Calculation Number: OSC - 9863 Total Original Pages: A. and 1 Through 20 Total Supporting Documentation Attachments: 1 (38 pages) Total Microfiche Attachments: 0 Total Volumes: 1 Active Calculation/Analysis: Yes U No []

Microfiche Attachment List: Yes E3 No 09 If Active, is this a Type I Calculation/Analysis: Yes rl No Ul (See Form 101.4)

These engineering Calculations cover QA Condition 1 Items. In accordance with established procedures, the quality has been assured and I certify that the above Calculation has been Originated, Checked, Inspected or Approved as noted below:

Originated By: See Vendor Cover Sheet Date:

Checked By: See Vendor Cover Sheet Date:

Verification Method: Method 1 E' Method 2 0I Method 3 jl Other -'

Approved By: David E. Whitaker . */./i. Date: /2/,2eO*

Issued To DORM: E"-' Date: J opq, Received By DCRM: Date:

Co)'fte the Spaces Below for Documentation of Multiple Originators or Checkers Pages: Through _

Originated By: Date:

Checked By: Date:

Verification Method: Method I El Method 2 L] Method 3 0- Other [LI Pages: Through Originated By: Date:

Checked By: Date:

Verification Method: Method 1 [] Method 2 0I Method 3 E] Other -I Pages: __ Through Originated By: Date:

Checked By: Date:

Verification Method: Method 1 E] Method 2 E] Method 3 E] Other E]

FIGURE -I CERTIFICATION OF ENGINEERING CALCULATION

Form 101.2 (113-03) Calculation Number OSC - 9863 REVISION DOCUMENTATION SHEET Revision Revision Description Number 0 Original with pages "A"(EM 4.9 Calculation Impact Assessment) added to AREVA Document No.

86 - 9109752 - 000. Calculation OSC - 9863 Revision 0 is for Limited Use and shall be used only for requesting a License Exemption using BAW-2308 from the NRC, and as inputs by AREVA in calculating new (54 EFPY) Reactor Vessel (RV) Pressure-Temperature (P-T) limits for Oconee Nuclear Station.

I Revision 1 is a complete replacement of AREVA Document No. 86 - 9109752 - 000 with AREVA Document No. 86 - 9109752 - 002. Calculation OSC - 9863 Revision 1 is still for Limited Use and shall be used only for requesting a License Exemption using BAW-2308 from the NRC, and as inputs by AREVA in calculating new (54 EFPY) Reactor Vessel (RV) Pressure-Temperature (P-T) limits for Oconee Nuclear Station.

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FIGURE ERROR! NO TEXT OF SPECIFIED STYLE IN DOCUMENT.-2 REVISION DOCUMENTATION SHEET

En*,ineerini, Manual 4.9 Rev. 10 Enpineerine Manual 4.9 Rev. 10 CALCULATION IMPACT ASSESSMENT (CIA)

Station /Unit Oconee/l. 2and 3 Calculation No. S46 Rev. I Page iA PIP No. (if applicable)__ By .. _ f Z54C Date AM/2100 Prob. No. (stress &slr use only) Checked By ........ N/A Date _

NEDL reviewed to Identify calculations? 0 Yes No Note: a NEDL search is NOT required for calculation originations (i.e., Rev.0's) I Identify in the blocks below, the groups consulted for an Impact Assessment of this calculation origination/revision.

Indv. Contacted/Date Indv. Contacted/Date

"] RES -] NGO

[Power, I & C, ERRT, Reactor] [QA Tech. Services (ISI); Severe Accident Analysis; Elect. Sys. &

Equip.; Design & Reactor Sup.;

Civil-Structural; Core Mech. & T/H MCE Analysis; Mech. Sys. & Equip.;

[Primary Systems, Balance of Plant, Nuclear Design and Safety Analysis; Rotating Equipment, Valves & Heat Materials/Metallurgy & Piping]

Exchangers, Civil]

L1 MOD LI Training

[Mechanical Engr., Electrical Engr.,

Civil Engr.]

El Local IT

[-' Operations - OPS Support D] Regulatory Compliance nI Maintenance - Tech. Support El Chemistry E] Work Control - Program Sup. I] Radiation Protection

[--] Other Group N No Group required to be consulted Listed below are the identified documents (ex: TECHNICAL SPECIFICATION SECTIONS, UFSAR SECTIONS, DESIGN BASIS DOCUMENTS, STATIONS PROCEDURES', DRAWINGS, OTHER CALCULATIONS, ETC.) that may require revision as a result of the calculation origination or revision, the document owner/group and the change required (including any necessary PIP Corrective Actions).

  • Note: Any design changes,which require changes to Station Procedures,must be transmittedas Design DeliverableDocuments.

DOCUMENT GROUP CHANGE REQUIRED Original (Attach Additional Sheets As Required)

Engineering Manual 4.9 Rev. 10 CALCULATION IMPACT ASSESSMENT (CIA)

Station/Unit Oconee/l .2and3 Calculation No. 0 Rev. 0 Page A PIP No. (if applicable) By _t9AedZZ. 1/ L Date V2 &CP Prob. No. (stress &s/r use only) - Checked By N/A Date __

NEDL reviewed to identify calculations? ' Yes [ No Note: a NEDL search is NOT required for calculation originations (i.e., Rev.0's)

Identify in the blocks below, the groups consulted for an Impact Assessment of this calculation origination/revision.

D[] E Indv. Contacted/Date

________ E NGo 1ndv. Contacted/Date

[Power, I & C, ERRT, Reactor) [QA Tech. Services (ISI); Severe Accident Analysis; Elect. Sys. &

Equip.; Design & Reactor Sup.;

Civil-Structural; Core Mech. & T/H "PriMCE Analysis; Mech. Sys. & Equip.;

[Primary Systems, Balance of Plant, Nuclear Design and Safety Analysis; Rotating Equipment, Valves & Heat Materials/Metallurgy & Piping]

Exchangers, Civil]

El MOD El Training

[Mechanical Engr., Electrical Engr.,

Civil Engr.] -ElLocal IT El Operations - OPS Support El Regulatory Compliance

-- Maintenance - Tech. Support E'_ Chemistry El Work Control - Program Sup. __ Radiation Protection El Other Group [_ No Group required to be consulted Listed below are the identified documents (ex: TECHNICAL SPECIFICATION SECTIONS, UFSAR SECTIONS, DESIGN BASIS DOCUMENTS, STATIONS PROCEDURES', DRAWINGS, OTHER CALCULATIONS, ETC.) that may require revision as a result of the calculation origination or revision, the document owner/group and the change required (including any necessary PIP Corrective Actions).

  • Note: Any design changes,which requirechanges to Station Procedures,must be transmittedas Design DeliverableDocuments.

DOCUMENT GROUP CHANGE REQUIRED Original (Attach Additional Sheets As Required)

Controlled Document0402-01-FOI (20697) (Rev. 014, 04/13/2009)

A CALCULATION

SUMMARY

SHEET (CSS)

AREVA Document No. 86 9109752 - 002 Safety Related: Z Yes [:]No Title Oconee Units 1. 2 and 3 License Exemption Using BAW-2308 PURPOSE AND

SUMMARY

OF RESULTS:

Purpose:

The purpose of this document is to summarize the pressurized thermal shock reference temperature (RTpTs) [1] and adjusted reference temperature (ART) [2] calculations at 54 EFPY for Oconee Units 1, 2 and 3 when using alternate initial RTNDT (nil ductility reference temperature) and ao (standard deviation of the initial RTNDT) values from BAW-2308, Revision 2-A. This document also provides the text that Duke Energy will need in their request for exemption to 10CFR50.61 and 10CFR50 Appendix G.

The purpose of Revision 2 is to report ART values calculated at additional locations to aid the pressure-temperature limits analysis, which is performed separately.

Results:

When using the initial nil ductility reference temperature and a, values from BAW-2308, Revision 2-A, the RTPTS values were below the screening criteria for all Oconee beltline materials. The controlling beltline material for the Oconee Unit 1 reactor vessel is the Upper Shell Longitudinal Weld, SA-1493, with a predicted RTPTS value of 196.0°F. Screening criterion from this material is 2700 F. The controlling beltline material for the Oconee Unit 2 reactor vessel is the Upper Shell to Lower Shell Circumferential Weld, WF-25, with a predicted RTprs value of 226.1 F. Screening criterion from this material is 300 0 F.

The controlling beltline material for the Oconee Unit 3 reactor vessel is the Upper Shell to Lower Shell Circumferential Weld ID (75%), WF-67, with a predicted RTPTS value of 222.51F. Screening criterion from this material is 3001F.

ART values were calculated using the initial RTNOT and at values from BAW-2308, Revision 2-A and Regulatory Guide 1.99 Revision 2, for all the Oconee beltline materials. The results are presented in section 4.0.

THE DOCUMENT CONTAINS ASSUMPTIONS THAT SHALL BE THE FOLLOWING COMPUTER CODES HAVE BEEN USED IN THIS DOCUMENT: VERIFIED PRIOR TO USE CODENERSIONIREV CODENERSIONJREV r Y SYES Z]NO AREVA NP Inc., an AREVA and Siemens company Page I of 23

A 0402-01-FOI (20697) (Rev. 014, 04/13/2009)

AREVA AREVANPInm, Document No. 86-9109752-002 in ARE VA and le.,mna campmny Oconee Units 1, 2 and 3 License Exemption Using BAW-2308 Review Method: 1! Design Review (Detailed Check)

D Alternate Calculation Signature Block P/R/A Name and Title and Pages/Sections (printed or typed) Signature LPILR Date Prepared/ReviewedlApproved R. S. Hosler 9 - P 0l,, All Engineer IH "lA D , R.( IO,.A ll S. B .D avidsaver I"OUWa R 01 0 l Engineer IV F. M. Gregory, Jr. A All Acting Engineering ,l-.. 6-Manager Note: P/R/A designates Preparer (P), Reviewer (R), Approver (A);

LP/LR designates Lead Preparer (LP), Lead Reviewer (LR)

Project Manager Approval of Customer References (N/A If not applicable)

Name Title (printed or typed) (printed or typed) Signature Date N/A N/A N/A N/A MentorIng Information (not required per 0402-01)

Page 2

fn At, ID A 0402-01-FOl (20697) (Rev. 014, 04/13/2009)

AREVA AREVA NPInc.,

Document No. 86-9109752-002 an AREVA and Siemens company Oconee Units 1, 2 and 3 License Exemption Using BAW-2308 Record of Revision Revision PageslSectionsl No. Date Paragraphs Changed Brief Description I Change Authorization 000 June 2009 All Original Submittal 001 April 2010 Section 1.0 Removed statement about when the Oconee fluence calculations were last updated Section 2.0 Removed statement about the 32 EFPY fluence values being compliant with RG 1.190 Tables 3-1, 3-2 and 3-3 Plate and forging type corrected; footnotes added Section 7.0 References 1, 2 and 8 updated 002 June 2010 Tables 4-2 and 4-3 Changed 55 EFPY to 54 EFPY Section 7.0 Reference 2 updated, added references 11-16 and renumbered all references Sections 2.0 and 4.0 Perform an ART calculation at an additional location on the LNB forging (Location 2) at Oconee Unit 1.

Added Figures 2-1 and 2-2.

Sections 2.0 and 4.0 Performed an ART calculation at an additional location on the LNB forging (Location 4) at Oconee Units 2 and 3.

Tables 3-1, 3-2, and 3-3 Added Location designation to the lower nozzle belt forging Page 3

A AR EVA Document No. 86-9109752-002 AREVA NP Inc..

an ARE VA and Siemens company Oconee Units 1, 2 and 3 License Exemption Using BAW-2308 Table of Contents Page SIG NATURE BLOCK ............................................................................................................................. 2 RECO RD O F REVISIO N ....................................................................................................................... 3 LIST OF TABLES .................................................................................................................................. 5 LIST OF FIGURES ................................................................................................................................ 6 1.0 INTRODUCTIO N ........................................................................................................................ 7 2.0 PROJECTED REACTOR VESSEL FLUENCE ...................................................................... 7 3.0 PRESSURIZED THERMAL SHOCK .................................................................................... 11 4.0 ADJUSTED REFERENCE TEM PERATURE ........................................................................ 15 5.0 RECENT LICENSE EXEM PTION REQ UESTS ................................................................... 19 6.0 LICENSE EXEM PTION JUSTIFICATION ............................................................................. 19 6.1 Introduction ...................................................................................................................................... 19 6.2 Background ..................................................................................................................................... 19 6.3 Proposed Exemption ....................................................................................................................... 20

7.0 REFERENCES

......................................................................................................................... 22 Page 4

A AREVA Document No. 86-9109752-002 AREVA NP Inc..

an ARE VA and Siemens oompany Oconee Units 1. 2 and 3 License Exemption Using BAW-2308 List of Tables Page Table 2-1: Inner Wetted Surface Fluence (E>1.0 MeV) Values for the Oconee Unit I Vessel Beltline Materials ....................................................................................................................................... 10 Table 2-2: Inner Wetted Surface Fluence (E>1.0 MeV) Values for the Oconee Unit 2 Vessel Beltline Mate ria ls ....................................................................................................................................... 10 Table 2-3: Inner Wetted Surface Fluence (E>I.0 MeV) Values for the Oconee Unit 3 Vessel Beltline Materials ....................................................................................................................................... 10 Table 3-1: Oconee Unit I Pressurized Thermal Shock Reference Temperature at 54 EFPY ........... 12 Table 3-2: Oconee Unit 2 Pressurized Thermal Shock Reference Temperature at 54 EFPY ........... 13 Table 3-3: Oconee Unit 3 Pressurized Thermal Shock Reference Temperature at 54 EFPY ....... 14 Table 4-1: Adjusted Reference Temperature Evaluation for the Oconee Unit I Reactor Vessel Beltline Materials at 54 EFPY ........................................................................................... ...... 16 Table 4-2: Adjusted Reference Temperature Evaluation for the Oconee Unit 2 Reactor Vessel Beltline Materials at 54 EFPY .................................................................................................................... 17 Table 4-3: Adjusted Reference Temperature Evaluation for the Oconee Unit 3 Reactor Vessel Beltline Materials at 54 EFPY .................................................................................................................... 18 Page 5

Con b-olled. DoCurn nt A Document No. 86-9109752-002 AREVA AREVA NP Inc.,

an AREVA and Siemens company Oconee Units 1, 2 and 3 License Exemption Using BAW-2308 List of Figures Page Figure 2-1: RPV Configuration for Oconee Unit I [133.................................

..................................... 8 Figure 2-2: RPV Configuration for Oconee Units 2 & 3 [13] ................................................................... 9 P 6 Page 6

A Document No. 86-9109752-002 AR EVA AREVA NP Inc.,

an AREVA and Siommns company Oconee Units 1, 2 and 3 License Exemption Using BAW-2308

1.0 INTRODUCTION

Lower initial RTNDT values for the high copper Linde 80 welds B&W used in fabricating reactor vessels have been justified in topical report BAW-2308, Rev. 2. Technical approval by the NRC has been completed and is contained in BAW-2308 Rev. 2-A [3]. As stated in the NRC safety evaluation, each utility must submit a request for exemption per 10CFR50.12 [4] from the requirements of 10CFR50.61 [5] and/or IOCFR50 Appendix G [6] to use the lowered initial RTNr values. The submittal must include pressurized thermal shock (PTS) and/or adjusted reference temperature (ART) calculations. A significant input to these calculations is the projected fluence at the reactor vessel. Fluence projections that were submitted as part of the Oconee license renewal application for 48 EFPY are extrapolated to 54 EFPY. The use of 54 EFPY for the 60-year life was a conservative estimate and may be updated based on specific plant data at a later time. The PTS and ART values are calculated in support of the exemption request. Once the full pedigree fluence projection is completed, which takes into account the change to 2 year cycles and other fuel changes, PTS and ART values can then be recalculated for input to the pressure-temperature curve effort.

2.0 PROJECTED REACTOR VESSEL FLUENCE The inner surface neutron fluence is the calculated value defined at the inside wetted surface of the Oconee Units 1, 2 and 3 reactor vessels. The projected 54 EFFY inner surface fluences for the Oconee Units 1, 2 and 3 reactor vessel beltline materials are listed in Table 2-1, Table 2-2 and Table 2-3. These 54 EFPY fluences were calculated by extrapolation using the previously reported projected 32 EFPY [7] and 48 EFPY [8] inner surface fluence values. The 48 EFPY fluence values were calculated in accordance with the requirements of U. S. NRC Regulatory Guide 1.190 [9], using a methodology described in detail in AREVA fluence topical report BAW-2241P-A [10]. For Oconee Unit 1, the projected fluence values for the lower nozzle belt (LNB) forging and the LNB to intermediate shell (IS) circumferential weld decreased from 32 EFPY to 48 EFPY. This was due to the change in the fluence calculation methodology for components above the core from an overly conservative method to a more accurate method. The 54 EFPY fluences for the Oconee Unit I LNB and the LNB to IS circumferential weld were calculated using the average percent increase in fluence from 48 EFPY to 54 EFPY of the remaining Oconee Unit I beltline materials.

The thickness of the lower nozzle belt (LNB) forging at Oconee Unit I varies from 8.44 [11] to 12.0 inches [12],

as shown in Figure 2-1 [13]. The peak RPV fluence at Location 1 was calculated using the extrapolation method discussed above. The fluence at Location 2 was estimated to be equal to the fluence at Location 1, as shown in Table 2-1. This estimation is conservative because Location I is closer to the core. ART values were calculated for both Locations 1 and 2, as shown in Section 4.0.

The inside wetted fluence of the LNB forging at Oconee Units 2 and 3 were estimated at two locations because of the thickness change of the forging. These two locations are shown in Figure 2-2 [13]. Location 3 represents the location of the peak inside fluence of the 8.44 inches [11] thick portion of the LNB forging, which was calculated using the extrapolation method described above. The fluence at Location A is used as a conservative estimate of the fluence at Location 4; this estimate is conservative because Location 4 is farther from the core than Location A. Location A at Oconee Units 2 and 3 is at the same location as the LNB to IS circumferential weld (SA-1 135) at Oconee 1, which is 11.0 feet from the reactor vessel flange mating surface. This distance of 11.0 feet is obtained from Reference 13 for Oconee Unit 1, References 14 and 15 for Oconee Unit 2 and References 15 and 16 for Oconee Unit 3. The projected 54 EFPY fluence at Oconee Unit I IS to US circumferential weld (SA-1229) is greater than the similarly located Oconee Unit 2 LNB to US circumferential weld (WF-154) and Oconee Unit 3 LNB to US circumferential weld (WF-200), as shown in Table 2-1, Table 2-2 and Table 2-3. These three welds are all 13.5 feet from the reactor vessel flange mating surface [13]. Based on this information, the flux at the

.. ... ..... .......... .... e...

Page 7

A Document No. 86-9109752-002 AREVA AREVA NP Inc..

an AREVA and Siemens company Oconee Units 1, 2 and 3 License Exemption Using BAW-2308 Oconee Unit I LNB to IS circumferential weld (SA-l 135) will likely be equal to or greater than the flux at Location A at Oconee Unit 2 and 3. Therefore the 54 EFPY fluence at Location A for Oconee Unit 2 and 3 is projected to be 1.25x1018 n/cm 2, which is equivalent to the similar location at Oconee Unit 1. ART values were calculated for both Locations 3 and 4, as shown in Section 4.0.

Upper Nozzle Belt Forging AFM 53; ZV 288 Outlet Nozzle Forging Location 2: Beginning of 12 In. thickness - Weld SA 1036 and SA 1101 Location 1: 8.44 in. thicknes -- Lower Nozz e Belt Forging AHR4: ZV 2861

- Weld SA 1135 Intermediate Shell C2197-2 Weld SA 1229 Inside 61%; WF 25 Outside 39%

Upper Shell C3265-1 and C3278-1 LowerHead A-0973e2 Figure 2-1: RPV Configuration for Oconee Unit 1 [13]

Page 8

A Document No. 86-9109752-002 AR EVA AREVA NP Inc.,

an AREVA and Siemens company Oconee Units 1, 2 and 3 License Exemption Using BAW-2308 Upper Nozzle Belt Forging

- Outlet Nozzle Forging Location 4: Beginning of 12.0 in.

Weld portion of the forging

- Lower Nozzle Belt Forging Location A. Peak fluence used for_..-

the 12.0 In. section (Location 4)

Location 3: Peak fluence of 8.44 in. Upper Shell Forging portion of forging Lower Shell Forging Lower Head Figure 2-2: RPV Configuration for Oconee Units 2 & 3 [13]

Page 9

ýj A Document No. 86-9109752-002 AR EVA AREVA NP Inc.,

an AREVA and Siemens company Oconee Units 1, 2 and 3 License Exemption Using BAW-2308 Table 2-1: Inner Wetted Surface Fluence (E>1.0 MeV) Values for the Oconee Unit 1 Vessel Beltllne Materials Beltline Materials Material Ident.

1 32 EFPYInner Wetted Fluence (n/cm )

[48 EFPY 1 54 EFPY 2

LN-B Forging (Locaion 1) AHR-54 1.18E+18 1.11E+18 1.25E+18-LNB Forging (Location 2) AHR-54 1.18E+18 1.1]E+18.. 1.25E+18*

Intermediate Shell Plate (IS) C2197-2 7.96E+18 1.18E+19 1.32E+19 Upper Shell Plate (US) C3265-1 9.04E+18 1.31E+19 1.46E+19 Upper Shell Plate C3278-1 9.04E+18 1.3 1E+19 1.46E+19 Lower Shell Plate (LS) C2800-1 8.68E+18 1.31E+19 1.48E+19 Lower Shell Plate C2800-2 8.68E+18 1.31E+19 1.48E+19 LNB to IS Circ Weld 100% SA-1135 1.18E+18 1.11E+18 1.25E+18*

IS Long Weld 100% SA-1073 6.28E+18 9.24E+18 1.04E+19 IS to US Circ Weld ID 61% SA-1229 7.96E+18 1.19E+19 1.34E+19 IS to US Circ Weld OD 39% WF-25 N/A N/A N/A US Long Weld 100% SA-1493 7.23E+18 1.12E+19 1.27E+19 US to LS Circ Weld'1001/6 SA-1585 8.68E+ 18 1.27E+ 19 1.42E+19 LS Long Weld 100% SA-1426 7.29E+18 1.08E+19 1.21E+19 LS Long Weld 100% SA-1430 7.29E+18 1.08E+19 1.21E+19

  • The )rojected fluence values for these materials decreased from 32 EFPY to 48 EFPY. This was due to the change in the fluence calculation methodology for components above the core. The 54 EFPY fluences for these components were calculated applying the average percent increase in fluence from 48 EFPY to 54 EFPY of the remaining Oconee Unit I beltline materials.

Table 2-2: Inner Wetted Surface Fluence (E>1.0 MeV) Values for the Oconee Unit 2 Vessel Beltline Materials Material Inner Wetted Fluence (n/cmr)

Beltline Materials Ident. 32 EFPY 1 48 EFPY 54 EFPY LNB Forging (Location 3) AMX-77 8.42E+18 1.19E+19 1.32E+19 LNB Forging (Location 4) AMX-77 N/A N/A 1.25E+ 18 US Forging AAW-163 9.57E+18 1.28E+19 1.40E+1 9 LS Forging AWG-164 9.19E+18 1.27E+19 1.40E+19 LNB to US Circ Weld 100% WF-154 8.42E+18 1.19E+19 1.32E+19 US to LS Circ Weld 100% WF-25 9.19E+18 1.23E+19 1.35E+19 Table 2-3: Inner Wetted Surface Fluence (E>1.0 MeV) Values for the Oconee Unit 3 Vessel Beitline Materials Material Inner Wetted Fluence (n/cm)

Beltline Materials Ident. 32 EFPY 48 EFPY 54 EFPY LNB Forging (Location 3) 4680 8.26E+18 1.14E+19 1.26E+19 LNB Forging (Location 4) 4680 N/A N/A 1.25E+18 US Forging AWS-192 9.39E+18 1.26E+19 1.38E+19 LS Forging ANK-191 9.01E+18 1.26E+19 1.39E+19 LNB to US Cirt Weld 100% WF-200 8.26E+18 1.14E+19 1.26E+19 US to LS Ci*Vic- iD- 75% WF-67 9.01E+18 1.22E+19 1.34E+19 LUS to LS Circ. Weld (OD 25%):...-,  :,,WF-70 N/A N/A N/A Page 10

Qcointrol:led..Dac. irrld t A

AR EVA Document No. 86-9109752-002 AREVA NP bc..

an AREVA and Siemens company Oconee Units 1, 2 and 3 License Exemption Using BAW-2308 3.0 PRESSURIZED THERMAL SHOCK The RTprs values applicable at 60 calendar years (54 EFPY) for the Oconee Units 1, 2 and 3 reactor vessel beltline materials are listed in Table 3-1, Table 3-2 and Table 3-3. These values were calculated in accordance with the requirement specified in the Code of Federal Regulations, Title 10, Part 50.61 (10 CFR 50.61) [5] and using the initial RTNDT values and the corresponding el values from BAW-2308, Revision 2-A [3].

The controlling beltline material for the Oconee Unit 1 reactor vessel is the Upper Shell Longitudinal Weld, SA-1493, with a predicted RTprs value of 196.0"F. The screening criterion from this material is 2700 F.

The controlling beltline material for the Oconee Unit 2 reactor vessel is the Upper Shell to Lower Shell Circumferential Weld, WF-25, with a predicted RTrrs value of 226.1 °F. The screening criterion from this material is 300'F.

The controlling belthne material for the Oconee Unit 3 reactor vessel is the Upper Shell to Lower Shell Circumferential Weld ID (75%), WF-67, with a predicted RTprs value of 222.5"F. The screening criterion from this material is 300 0F.

Page 11

faontro.61ied."b extrenet.

A Document No. 86-9109752-002 AR EVA AREVA NP Inc.,

an AREVA and Siemens company Oconee Units 1, 2 and 3 License Exemption Using BAW-2308 Table 3-1: Oconee Unit 1 Pressurized Thermal Shock Reference Temperature at 54 EFPY Chemical 54 EFPY Material Description Composition Fluence at Inside Initial Wetted Screening Reactor Vessel Heat Chem. RTNm- Surface ART*mT Margin RTPrs Criteria Beltline RqiaonMad. Mart Ident.* Number. Type, Cuw ., . .. (n/cm ., -1

.. (T)) - (*F) (°F) (F)

LNB Fo'rg'ing ZV-2861 AHR-54 A-508, CI. 2' 0.16 0.65 119.3 +3 1.25E+18 55.2 31.0 17.0 70.7 128.9 270 (Location 1) ____________________

Intermediate Shell (IS) C2197-2 C2197-2 SA-302, Or. B, Mod.b 0.15 0.50 104.5 +1 1.32E+19 112.7 26.9 17.0 63.6 177.3 270

Plate C 4 6 _ O._0 0 6+ . 929365 Upper Shell (US) Plate C3265-1 C3265-1 SA-302, Or. B. Mod.' 0.10 0.50 65.0 +1 1.46E+19 71.8 26.9 17.0 63.6 136.5 270 Upper Shell Plate C3278-i C3278-1 SA-302, Or. B. Mod.b 0.12 0.60 83.0 +1 1.46E+19 91.7 26.9 17.0 63.6 156.4 270 Lower Shell (LS) Plate C2800-2 C2800-2 SA-302, Or. B, Mod.' 0.11 0.63 74.5 +1 1.48E+19 82.5 26.9 17.0 63.6 147.2 270 Lower Shell Plate C2800-2 C2800-2 SA-302, Or. Bp Mod.' 0.1I1 0.63 74.5 +1 1.48E+19 82.5 26.9 17.0 63.6 147.2 270 LNB to IS Circ. WVld SA-1 135 61782 Linde 80 0.23 0.52 167.0' -58.5 1.25E+18 77.2 15.4 28.0 63.9 82.6 300 (100%)

IS Long. Weld (100%) SA-1073 IP0962 Linde 80 0.21 0.64 170.6 -48.6 1.04E+19 172.2 18.0 28.0 66.6 190.2 270 IS to US Circ. Weld SA-1229 71249 Linde 80 0.23 0.59 167.6 -53.5 1.34E+19 181.2 12.8 28.0 61.6 189.2 300 (ID61%)

IS to US Circ. Weld WF-25 2991A4 Linde 80 0.34 0.68 N/A N/A NIA NIA NIA N/A N/A N/A 300 (OD 39%) 1_1__

US Long. Weld SA-1493 8T1762 Linde 80 0.19 0.57 167.0' -48.6 1.27E+19 178.1 18.0 28.0 66.6 (196.0] 270 (100%) _________ ___ ____ ____

UStoLSCirc.Weld SA-1585 72445 Linde 80 0.22 0.54 167.0' -72.5 1.42E+19 183.3 12.0 28.0 60.9 171.7 300 (100%) 1 LS Long. Weld SA-1426 8T1762 Linde 80 0.19 0.57 167.0' -48.6 1.21E+19 175.9 18.0 28.0 66.6 193.9 270 (100%)

LS Long. Wl (100%) SA-1430 8T1762 Linde 80 0.19 0.57 167.0' -48.6 1.21E+19 175.9 18.0 28.0. 66:6 .193.9 270

[]- Limiting reactor vessel beltline region material in accordance with 10 CFR 50.61 a ASTM A-508-64 Cl. 2 Mod. By ASME Code Case 1332-2 b ASME SA-302 Gr. B Mod. To ASME Code Case 1339 r Per BAW-2308 Rev. 1-A [17]

d See Figure 2-1 Page 12

controlled Emoc Ument A Document No. 86-9109752-002 AREVA AREVA NP Inc.,

an AREVA and Siemens company Oconee Units 1. 2 and 3 License Exemption Using BAW-2308 Table 3-2: Oconee Unit 2 Pressurized Thermal Shock Reference Temperature at 54 EFPY Chemical 54 EFPY Material Description Composition Fluence al Reactor Vessel Initial Inside Screening Beltline Region Matt. Heat Chem. RTNDT Surface ARTNDT Margin RTpTs Criteria Matt. Ident. Number Type Cu wth Ni wt% Factor (OF) (n/core) (oF) 0_ CA (IF) (_F) (OF)

N Forging

    • LNB Fogin AMX 77 123T382 A-508, CI. 2- 0.13 0.76 95.0 +3 1.32E+19 102.3 31.0 17.0 70.7 176.1 270 (Location 3YC______ ____________________________ _________________ ___________

US Forging AAW 163 3P2359 A-508. C1. 2b 0.04 0.75 26.0 +20 1.40E+19 28.4 0.0 17.0 28.4 76.9 270 LS Forging AWG 164 4PI885 A-508, C 2 0.02 0.80 20.0 +20 1.40E+19 21.9 0.0 17.0 21.9 63.7 270 LNB to US Circ. Weld WF- 154 4061.44 Linde 80 0.27 0.59 182.6 -98.0 1.32E+19 196.7 11.6 28.0 60.6 159.3 300 (100o/)

USto LS Circ. Weld US WF-25 2991.44 Linde 80 0.34 0.68 220.6 -74.3 1.35E+19 238.8 12.8 28.0 61.6 [226.2] 300

[I]- Limiting reactor vessel beltline region material in accordance with 10 CFR 50.61 a ASTM A-508-64 Cl. 2 Mod. By ASME Code Case 1332-2 b ASTM A-508-64 CI. 2 Mod. By ASME Code Case 1332-4

' See Figure 2-2 Page 13

Conitrolled ~Ocurnent, A Document No. 86-9109752-002 AREVA AREVA NP Inc..

an AREVA and Siemens company Oconee Units 1, 2 and 3 License Exemption Using BAW-2308 Table 3-3: Oconee Unit 3 Pressurized Thermal Shock Reference Temperature at 54 EFPY Chemical 54 EFPY Material Description Compo ition Fluence at Vessel 1Reactor Initial Inside Screening Wetted u-DTMrn (TF) Criteria Beltline Region Mard. Heat Chem. RT"MT Surface ARTNDT Margin RTpTs Criteria 2

Matd. Ident. Number Type Cu wt%/. Ni wt% Factor (OF) (n/cm ) (OF) _"_ TA (OF) (_F) (OF)

LNB Forging 4680 4680 A-508, CI. 2° 0.13 0.91 96.0 +3 1.26E+19 102.1 31.0 17.0 70.7 175.8 270

'(Location 3

)' _____ ____ _____ _____

  • US Forging AWS 192 522314 A-508. Cl. 2' 0.01 0.73 36.00 +40 1.38E+19 39.2 0.0 17.0 34.0 113.2 270

.LS Forging ANK 191 522194 A-508 C). 2' 0.02 0.76 17.4' +40 1.39E+19 19.0 0.0 14.0 19.0 78.0 270 LNB tUS Ciro. Weld WF-200 821T44 Linde 80 0.24 0.63 178.0 -84.2 1.26E+19 189.4 9.6 28.0 59.2 164.4 300 (100%) ____ __________

US toLS Cire. Weld WF-67 72442 Linde 80 0.26 0.60 180.0 -33.2 1.34E+ 19 194.6 12.2 28.0 61.1 [222.5] 300 (ID75%) 1 1 1 1 1 US to LS Cire. Weld WF-70 72105 Linde 80 0.32 0.58 199.3 -31.1 N/A N/A 13.7 28.0 N/A INA 300 (01) 25%) ____ _____________ __________ _______

r]- Limiting reactor vessel beltline region material in accordance with 10 CFR 50.61 ASTM A-508-64 Cl. 2 Mod. By ASME Code Case 1332-3 b ASTM A-508-64 Cl. 2 Mod. By ASME Code Case 1332-4 This Chemistry Factor was determined from surveillance data.

d See Figure 2-2 Page 14

A AR EVA Document No. 86-9109752-002 AREVA NPInc.,

an AREVA and Siemons company Oconee Units 1, 2 and 3 License Exemption Using BAW-2308 4.0 ADJUSTED REFERENCE TEMPERATURE The 1/44T and 3T ART values for the Oconee Units 1, 2 and 3 reactor vessel beltline materials applicable at 60 calendar years (54 EFPY) are calculated using the alternate initial RTrDT values and the corresponding a( values from BAW-2308, Revision 2-A [3] (for weld metals) and are listed in Table 4-1, Table 4-2 and Table 4-3. The ART values are calculated in accordance with Regulatory Guide 1.99, Revision 2 [18]. The material type, chemical compositions, chemistry factors, and initial RTNDT values used in the ART calculations shown in Table 4-1, Table 4-2 and Table 4-3 are the same as shown in Table 3-1, Table 3-2 and Table 3-3 respectively and are thus not repeated.

The circumferential welds with the highest ART values for the Oconee Unit 1 reactor vessel are the Intermediate Shell to Upper Shell Circumferential Weld ID (61%), SA-1229, with an ART value at 54 EFPY of 164.2'F at the 1/T wall location and Intermediate Shell to Upper Shell Circumferential Weld OD (39%), WF-25, with an ART value at the 54 EFPY of 132. 1IF at the 3/T wall location. Considering the base metal and the longitudinal welds, the materials with the highest ART values are the Upper Shell Longitudinal Weld, SA-1493, with an ART value at 54 EFPY of 171.0'F at the 'AT wall location and the Intermediate Shell Plate, C-2197-2, with an ART value at the 54 EFPY of 132.9'F at the 1/T wall location.

The circumferential weld with the highest ART values for the Oconee Unit 2 reactor vessel is the Upper Shell to Lower Shell Circumferential Weld, WF-25, with an ART value at 54 EFPY of 193.1°F at the %hTwall location and 132.5`F at the 3/T wall location. Considering the base metal (there are no longitudinal welds in Oconee Unit 2), the material with the highest ART values is the Lower Nozzle Belt Forging, AMX-77, with an ART value at 54 EFPY of 161.8°F at the 1/4Twall location and 135.7 0F at the %T wall location. The ART values for the Lower Nozzle Belt Forging correspond to Location 3, which is the location of peak fluence for the 8.44 inches thick portion of the forging.

The circumferential welds with the highest ART values for the Oconee Unit 3 reactor vessel are the Upper Shell to Lower Shell Circumferential Weld ID (75%), WF-67, with an ART value at 54 EFPY of 195.6*F at the 'AT wall location and the Upper Shell to Lower Shell Circumferential Weld OD (25%), WF-70, with an ART value at the 54 EFPY of 162. 10F at the 3/T wall location. Considering the base metal (there are no longitudinal welds in Oconee Unit 2), the material with the highest ART values is the Lower Nozzle Belt Forging, 4680, with an ART value at 54 EFPY of 161.4'F at the 1AT wall location and 135.2°F at the %T wall location. The ART values for the Lower Nozzle Belt Forging correspond to Location 3, which is the location of peak fluence for the 8.44 inches thick portion of the forging.

Page 15

Con01Q;iee t AREVA Document No. 86-9109752-002 AREVA NP Inc, an ARVA and Siamens company Oconee Units 1, 2 and 3 License Exemption Using BAW-2308 Table 4-1: Adjusted Reference Temperature Evaluation for the Oconee Unit I Reactor Vessel Beltline Materials at 54 EFPY

&RT"T(OF)at 54 Material Description 54 EFPY Fluence (n/cm') EFPY Std. Deviation Mar (*F) ART (:F) at 54 EFPY Inner Reactor Vessel Beltline Region Mat]. Wetted 1/4AT 'AT '/T 3A/T 1/4T 'AT %T %T Location Ident. Surface Location Location Location Location o0 oa Location Location Location Location LNB Forging (Location 1) AHR-54 1.251+18 7.29E+17 2.65E+17 42.5 24.3 31.0 17.0 70.7 66.6 116.2 93.9 LNB Forzing (Location 2Y AHR-54 1.25E+ 18 5.90E+17 1.401+17 38.2 16.4 31.0 17.0 70.7 64.1 111.9 83.5 Internediate Shell (IS) Plate C2197-2 1.32E+19 7.74E+18 2.81E+18 97.0 68.3 26.9 17.0 63.6 63.6 161.6 (132.91 Upper Shell (US) Plate C3265-1 1.46E+19 8.55E+18 3.11E+18 62.1 44.2 26.9 17.0 63.6 63.6 126.7 108.8 Upper Shell Plate C3278-1 1.46E+19 8.55E+18 3.11E+18 79.4 56.4 26.9 17.0 63.6 63.6 144.0 121.0 Lower Shell (LS) Plate C2800-1 1.48E+19 8.63E+18 3.13E+18 71.4 50.8 26.9 17.0 63.6 63.6 136.0 115.4 Lower Shell Plate C2800-2 t.ASE+19 8.631+18 3.13E+18 71.4 50.8 26.9 17.0 63.6 63.6 136.0 115.4 LNB to IS Circ. Weld (100%) SA-I 135 1.25E+18 7.29E+17 2.65E+17 59.5 34.1 15.4 28.0 63.9 63.9 64.9 39.5 IS Long. Weld (100%) SA-1073 1.041+19 6.05E+18 2.20E+18 146.6 101.0 18.0 28.0 66.6 66.6 164.6 119.0 IS to US Circ. Weld(D) 61%) SA-1229 1.34E+19 7.82E+18 N/A 156.1 N/A 12.8 28.0 61.6 N/A [164.21 N/A IS to US Circ. Weld (OD 39%) WF-25 1.34E+19 N/A 2.84E+18 N/A 144.8 12.8 28.0 N/A 61.6 N/A f 132.1]

US Long. Weld (100%) SA-1493 1.27E+19 7.42E+18 2.70E+18 153.0 107.4 18.0 28.0 66.6 66.6 (171.0) 125.4 US to LS Circ. Weld(100%) SA-158S5 1.42E+19 8.311+18 3.02E+18 158.3 112.2 12.0 28.0 60.9 60.9 146.7 100.6 LS Long. Weld (100%) SA-1426 1.21E+19 7.091+18 2.57E+I8 150.9 105.4 18.0 28.0 66.6 66.6 168.9 123.4 LS Long. Weld (100%) SA-1430 1.21E+19 7.091+18 2.57E+18 150.9 105.4 18.0 28.0 66.6 66.6 168.9 123.4 aSee Figure 2-1

[]-Highest values of the adjusted reference temperatures for circumferential welds.

{ }- Highest values of the adjusted reference temperatures for base metal or longitudinal welds.

Page 16

Controllcd ~octiment

,-1 AREVA Document No. 86-9109752-002 AREVA NP Inc.

an AREVA and Siemens company Oconee Units 1, 2 and 3 License Exemption Using BAW-2308 Table 4-2: Adjusted Reference Temperature Evaluation for the Oconee Unit 2 Reactor Vessel Beltline Materials at 54 EFPY 0

ARTNDr ( F) at 54 Material Description 54 EFPY Fluence (n/cm) EFPY Std. Deviation Margin (*F) ART (-F) 54 EFPY Inner Reactor Vessel Belthine Region Mat]. Wetted 4T %AT V.T Y.T %AT /'T '/AT %T Location [dent. Surface Location Location Location Location al C(A Location Location Location Location LNB Forging (Location 3" AMX77 1.32E+19 7.72E418 2.80E+18 88.1 62.0 3.0 LO 17.0 70.7 70.7 (161.81 1135.7)

LNB Forging (Location 4)' AMX 77 1.25E3+18 5.90E+17 1.40E+ 17 30.4 13.0 31.0 17.0 69.0 63.4 102.4 79.4 IUSForging AAW 163 1.40E+19 8.19E+18 2.98E+18 24.5 17.4 0.0 17.0 24.5 17.4 69.0 54.8 S Forging: AWG 164 1.40E+19 8.20E+18 2.98E+18 18.9 13.4 0.0 17.0 18.9 13.4 57.8 46.8 LNB to US Circ. Weld (100%) WF-154 1.32E+19 7.721+18 2.80E+18 169.4 119.2 11.6 28.0 60.6 60.6 132.0 81.8 US toLS Circ. Weld(100%) WF-25 1.35E+19 7.88E+18 2.86E+18 205.8 145.2 12.8 28.0 61.6 61.6 [193.1] [132.51 a See Figure 2-2

[]-Highest values of the adjusted reference temperatures for circumferential welds.

} - Highest values of the adjusted reference temperatures for base metal (no longitudinal welds in Oconee Unit 2).

Page 17

Controlled lprurne'n AREVA Document No. 86-9109752-002 AREVA NP Inc.

an AREVA and Siemens company Oconee Units 1, 2 and 3 License Exemption Using BAW-2308 Table 4-3: Adjusted Reference Temperature Evaluation for the Oconee Unit 3 Reactor Vessel Beltilne Materials at 54 EFPY ARTNTr (IF) at 54 Std.

Material Description 54 EFPY Fluence (n/cm') EFPY Deviation Margin (*F) ART (IF) at 54 EFPY Inner Reactor Vessel Beltline Region Mat). Wetted %T %AT 1/T %Y.T AT W'/T Y.T 1AT Location [dent. Surface Location Location Location Location o, Location La Location Location Location LNB Forging (Location 3)° 4680 1.26E+19 7.36E+18 2.67E+18 87.7 61.5 31.0 17.0 70.7 70.7 {161.4) (135.21 LNB Forging (Location 4)" 4680 1.25E+18 5.90E+17 1.40E+17 30.7 13.2 31.0 17.0 69.2 63.4 102.9 79.5 US Forging AWS 192 1.38E+19 8.07E+18 2.93E+18 33.8 23.9 0.0 17.0 34.0 34.0 107.8 97.9 LS Forging ANK 191 1.39E+19 8.16E+18 2.96E+18 16.4 11.6 0.0 8.5 16.4 11.6 72.8 63.2 LNB to US Circ. Weld (100%) WF-200 1.26E1+19 7.36E+18 2.67E+18 162.7 114.0 9.6 28.0 59.2 59.2 137.7 89.0 US to LS Circ. Weld (ID 75%) WF-67 1.34E+19 7.83E+ 18 NIA 167.7 N/A 12.2 28.0 61.1 N/A (3956] N/A US to LS Circ. Weld (OD 25%) WF-70 1.34E+19 N/A 2.85E+18 N/A 130.9 13.7 28.0 N/A 62.3 N/A [162.11 a See Figure 2-2

[]-Controlling values of the adjusted reference temperatures for circumferential welds.

{ } - Highest values of the adjusted reference temperatures for base metal (no longitudinal welds in Oconee Unit 3).

Page 18

A AR EVA Document No. 86-9109752-002 AREVA NP Inc.,

an AREVA and Siemnens company Oconee Units 1, 2 and 3 License Exemption Using BAW-2308 5.0 RECENT LICENSE EXEMPTION REQUESTS Virginia Electric and Power Company (Dominion) requested an exemption from the requirements of 10 CFR 50.61 and 10 CFR 50 Appendix G to revise the initial RTNrr and associated o] values of the Linde 80 weld materials present in the beltline region of Surry Unit I and Unit 2 reactor pressure vessels using AREVA Topical Report BAW-2308, Revision 1-A [19]. This exemption request was accompanied by RTprs and ART calculations, which utilized the revised initial RTNDT and associated oa values in BAW-2308, Revision 1-A. Attachment 2 of Reference 19 details the Surry exemption request, to which the Oconee license exemption request below was based. The U.S. Nuclear Regulatory Commission approved this license exemption request for the Surry Power Station Units I and 2 [20].

Florida Power and Light Company has requested an exemption from the requirements of 10 CFR 50.61 and 10 CFR 50 Appendix G for Turkey Point Units 3 and 4 [21]. The exemption request incorporates the methodology of BAW-2308, Revision 2-A.

FirstEnergy Nuclear Operating Company has requested an amendment of the operating license for the Davis-Besse Nuclear Power Station [22]. The amendment incorporates the methodology of BAW-2308, Revision 2-A for 10 CFR 50.61 and 10 CFR 50 Appendix G.

6.0 LICENSE EXEMPTION JUSTIFICATION 6.1 Introduction In accordance with the provisions of 10 CFR 50.60(b) and 10 CFR 50.12, Duke Energy is submitting a request for exemption from certain requirements of 10 CFR 50.61, "Fracture Toughness Requirements for Protection Against Thermal Shock Events," and 10 CFR 50, Appendix G, "Fracture Toughness Requirements." The requested exemption would allow use of alternate initial RTNDT (reference nil ductility temperature), as described in AREVA NP Topical Report BAW-2308, Revision I-A and Revision 2-A, for determining the adjusted RTNDT of the Linde 80 weld materials present in the beltline region of the Oconee Units 1, 2 and 3 reactor pressure vessels.

6.2 Background 10 CFR 50.61 (a)(5) and 10 CFR 50, Appendix G (II)(D)(i), require that the pre-service or unirradiated condition RTNDT be evaluated according to the procedures in the ASME Code,Section III, Paragraph NB-233 1, from Charpy V-notch impact tests and drop weight tests.

AREVA NP Topical Report BAW-2308, Rev. 2-A provides an NRC-approved alternate initial RT-or and associated a1 values of the Linde 80 weld materials present in the beitline region of the reactor pressure vessels at Oconee Units 1, 2 and 3.

The following Condition and Limitation is stated in the NRC's Safety Evaluation for Topical Report BAW-2308, Rev. 1-A [17]:

"Any licensee who wants to utilize the methodology of TR BAW-2308, Revision I as outlinedin items (1) through (3) above, must request an exemption, per 10 CFR 50.12,from the requirements ofAppendix G to 10 CFR Part 50 and 10 CFR 50.61 to do so."

Page 19

AR EVA Document No. 86-9109752-002 AREVA NP Inc.,

an AREVA end Siemens company Oconee Units 1, 2 and 3 License Exemption Using BAW-2308 In the above quotation, Condition and Limitation (1) pertains to NRC-accepted values of initial (unirradiated) reference temperature, IRTTo, and the corresponding uncertainty term, a7, for Linde 80 weld materials based on the Master Curve methodology using direct testing of fracture toughness in accordance with ASTM Standard Test Method E-1921.

Condition and Limitation (2) requires that a minimum chemistry factor of 167.0*F be applied when the methodology of Regulatory Guide 1.99, Revision 2, and 10 CFR 50.61 is used to assess the shift in nil-ductility transition temperature due to irradiation.

Condition and Limitation (3) requires that a value of aa = 28.0°F be used to determine the margin term, as defined in Topical Report BAW-2308, Revision 2-A, and Regulatory Guide 1.99, Revision 2.

6.3 Proposed Exemption The exemption requested by Duke Energy addresses portions of the following regulations:

(1) Appendix G to 10 CFR Part 50, which sets forth fracture toughness requirements for ferritic materials of pressure-retaining components of the reactor coolant pressure boundary of light water nuclear power reactors to provide adequate margins of safety during any condition of normal operation, including anticipated operational occurrences and system hydrostatic tests, to which the system may be subjected over its service lifetime; (2) 10 CFR 50.61, which sets forth fracture toughness requirements for protection against pressurized thermal shock (PTS).

The exemption from Appendix G to 10 CFR 50 is to replace the required use of the existing Charpy V-notch and drop-weight-based methodology with the use of an alternate methodology that incorporates the use of fracture toughness test data for evaluating the integrity of the Linde 80 weld materials present in the Oconee Units 1, 2 and 3 RPV beltline regions. The alternate methodology employs direct fracture toughness testing per the Master Curve methodology based on use of ASTM Standard Method E 1921 (1997 and 2002 editions), and ASME Code Case N-629. The exemption is required since Appendix G to 10 CFR 50 requires that for the pre-service or unirradiated condition, RTNDT be evaluated by Charpy V-notch impact tests and drop weight tests according to the procedures in the ASME Code, Paragraph NB-233 1.

The exemption from 10 CFR 50.61 is to use an alternate methodology to allow the use of direct fracture toughness test data for evaluating the integrity of the Linde 80 weld materials present in the Oconee Units 1, 2 and 3 RPV beltline regions, based on the use of ASTM E 1921 (1997 and 2002 editions) and ASME Code Case N-629. The exemption is required because the methodology for evaluating RPV material fracture toughness in 10 CFR 50.61 requires that the pre-service or unirradiated condition be evaluated using Charpy V-notch impact tests and drop weight tests according to the procedures in the ASME Code, Paragraph NB-2331.

Additionally, the NRC's Safety Evaluation for Topical Report BAW-2308, Revision I-A, concludes that an exemption is required to address issues related to 10 CFR 50.61 inasmuch as the methodology presented in Topical Report BAW-2308, Revision I-A, as modified and approved by the NRC staff, represents a significant change to the methodology specified in 10 CFR 50.61 for determining the PTS reference temperature (RTprs) value for Linde 80 weld material. The changes in the methodology described in BAW-2308, Revision I-A, with respect to the methodology per 10 CFR 50.61, include the requirements for use of a minimum chemistry factor of 167*F and a value of oa = 28.0*F for Linde 80 weld materials.

Page 20

AREVA Document No. 86-9109752-002 AREVA NPInc.,

an AREVA and Siemons company Oconee Units 1, 2 and 3 License Exemption Using BAW-2308 10 CFR 50.12 states that the Commission may grant an exemption from requirements contained in 10 CFR 50 provided that: 1) the exemption is authorized by law, 2) the exemption will not result in an undue risk to public health and safety, 3) the exemption is consistent with the common defense and security, and 4) special circumstances, as defined in 10 CFR 50.1 2(a)(2) are present. The requested exemption to allow the use of Topical Report BAW-2308, Revision 1-A and Revision 2-A (Revision 2-A is a supplement to Revision I-A), as the basis for the Linde 80 weld material initial properties at Oconee Units 1, 2 and 3 satisfy these requirements as described below.

1. The requested exemption is authorized by law, No law exists which precludes the activities covered by this exemption request. 10 CFR 50.60(b) allows the use of alternatives to 10 CFR 50, Appendix G when an exemption is granted by the Commission under 10 CFR 50.1 2.

In addition, 10 CFR 50.61 permits other methods for use in determining the initial material properties provided such methods are approved by the Director, Office of Nuclear Reactor Regulation.

2. The requested exemption does not present an undue risk to the public health and safety.

The proposed material initial properties basis described in Topical Report BAW-2308 Revision 2-A represents an NRC-approved methodology for establishing weld wire specific and generic IRTT. values for Linde 80 welds.

Topical Report BA-2308, Revision 2-A, includes appropriate conservatisms to ensure that use of the proposed initial material properties basis does not increase the probability of occurrence or the consequences of an accident at Oconee Units 1, 2 and 3, and will not create the possibility for a new or different type of accident that could pose a risk to public health and safety.

The use of this proposed approach ensures that the intent of the requirements specified in 10 CFR 50 Appendix G and 10 CFR 50.61 are satisfied.

The requested exemption is consistent with the NRC staff requirements specified in the Safety Evaluation for the approved Topical Report BAW-2308, Revision 1-A and Revision 2-A; consequently, the exemption does not present an undue risk to the public health and safety.

3. The requested exemRtion will not endanger the common defense and security, The requested exemption is specifically concerned with RPV material properties and is consistent with NRC staff requirements specified in the Safety Evaluation for approved Topical Report BAW-2308, Revision 2-A.

Consequently, the requested exemption will not endanger the common defense and security.

4. Special circumstances are present which necessitate the request for an exemption to the regulations of 10 CFR 50.61 and 10 CFR 50 Appendix G.

Pursuant to 10 CFR 50.1 2(a)(2), the NRC will not consider granting an exemption to the regulations unless special circumstances are present. The requested exemption meets the special circumstances of paragraph 10 CFR 50.1 2(a)(2)(ii) since application of the methodology in BAW-2308, Revision I-A and Revision 2-A, in this particular circumstance serves the underlying purpose of the regulations.

The underlying purpose of 10 CFR 50.61 and 10 CFR 50 Appendix G is to protect the integrity of the reactor coolant pressure boundary by ensuring that each reactor vessel material has adequate fracture toughness.

Application of paragraph NB-2331 of ASME Section III in the determination of initial material properties was Page 21

nrC~iL~ Dcc.Ll GD iment A Document No. 86-9109752-002 AR EVA AREVA NPInc.,

ar AREVA and Siemens company Oconee Units 1, 2 and 3 License Exemption Using BAW-2308 conservatively developed based on the level of knowledge existing in the early 1970s concerning RPV materials and the estimated effects of operation. Since the early 1970s, the level of knowledge concerning these topics has greatly expanded. This increased knowledge level permits relaxation of the ASME III NB-2331 requirements via application of Topical Report BAW-2308, Revision 2-A, while maintaining the underlying purpose of the ASME Code and NRC regulations to ensure an acceptable margin of safety is maintained.

This submittal presents the reactor vessel integrity assessments for Oconee Units 1, 2 and 3 utilizing the methodology of Topical Report BAW-2308, Revision 2-A for Linde 80 weld materials. The assessment documents the integrity of the RPV for Oconee Units 1, 2 and 3 relative to the requirements and underlying purpose of 10 CFR 50.61 and 10 CFR 50 Appendix G.

Therefore, the intent of 10 CFR 50.61 and 10 CFR 50 Appendix G will continue to be satisfied for the proposed change in reactor vessel material initial properties basis, thus justifying the exemption request. Issuance of an exemption from the criteria of these regulations to permit the use of Topical Report BAW-2308, Revision 2-A for Oconee Units 1, 2 and 3 will not compromise the safe operation of the reactors, and will ensure that RPV integrity is maintained.

7.0 REFERENCES

1. AREVA Document 32-9109428-001, "RTPTS Values for Oconee Units 1, 2 and 3 at 60 Calendar Years,"

April 2010.

2. AREVA Document 32-9108770-002, "ART Values for Oconee Units 1, 2 and 3 at 60 Calendar Years,"

June 2010.

3. AREVA Document 43-2308-004, "Initial RTNDT of Linde 80 Weld Materials," (BAW-2308, Revision 2-A), March 2008.
4. Code of Federal Regulations, Title 10, "Domestic Licensing of Production and Utilization Facilities," Part 50.12, "Specific Exemptions," Effective Date: December 12, 1995.
5. Code of Federal Regulations, Title 10, "Domestic Licensing of Production and Utilization Facilities," Part 50.61, "Fracture Toughness Requirements for Protection against Pressurized Thermal Shock," Effective Date: August 28, 1996.
6. Code of Federal Regulations, Title 10, "Domestic Licensing of Production and Utilization Facilities," Part 50 Appendix G, "Fracture Toughness Requirements," Effective Date: December 19, 1995.
7. AREVA NP Document 43-2325-01, "Response to Request for Additional Information (RAI) Regarding Reactor Pressure Vessel Integrity," (BAW-2325, Revision 1), January 1999.
8. Letter from Duke Energy Corporation forwarding application for renewal of operating licenses for the Oconee Nuclear Station, Unit Nos. 1, 2, and 3, U. S. Nuclear Regulatory Commission, ACN:

9807200136, Fiche: A4344:001-A4347:255, July 6, 1998.

9. U. S. Nuclear Regulatory Commission, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence," Regulatory Guide 1.190, March 2001.
10. J. R. Worsham, et al., "Fluence and Uncertainty Methodologies," BAW-2241P-A, Revision 1, AREVA NP, Inc., Lynchburg, Virginia, April 2000.
11. AREVA NP Document 43-1543-04, "Master Integrated Reactor Vessel Surveillance Program," BAW-1543, Revision 4, February 1993.

Page 22

AR EVA Document No. 86-9109752-002 AREVA NP Inc.,

an AREVA and Siemens company Oconee Units 1, 2 and 3 License Exemption Using BAW-2308

12. AREVA Drawing 02-128706-09, "Core Flooding Nozzle," Oconee Unit 1.
13. BAW-1 820, "Babcock & Wilcox Owner's Group 177-Fuel Assembly Reactor Vessel and Surveillance Program Materials Information," December 1984.
14. AREVA Drawing 02-128737-06, "Upper Shell Assembly," Oconee Unit 2.
15. AREVA Drawing 02-99319-04, "Upper Shell Forging," Oconee Units 2 and 3.
16. AREVA Drawing 02-149904-05, "Upper Shell Assembly," Oconee Unit 3.
17. A.REVA NP Document 43-2308-002, "Initial RTNDT of Linde 80 Weld Materials," (BAW-2308, Revision I-A), August 2005.
18. U. S. Nuclear Regulatory Commission, "Radiation Embrittlement of Reactor Vessel Materials,"

Regulatory Guide 1.99, Revision 2, May 1988.

19. Letter to NRC, "Virginia Electric and Power Company Surry Power Station Units I and 2 Update to NRC Reactor Vessel Integrity Database and Exemption Request for Alternate Material Properties Basis Per 10 CFR 50.60(b)," ML061650080, June 2006.
20. Letter from NRC, "Surry Power Station, Unit Nos. 1 and 2, Exemption from the Requirements of 10 CFR Part 50, Appendix G and 10 CFR Part 50, Section 50.61," ML071160287, June 2007.
21. Letter to NRC, "Turkey Point, Units 3 and 4, Update to NRC Reactor Vessel Integrity Database and Exemption Request for Alternate Material Properties Bases Per 10 CFR 50.12 and 10 CFR 50.60 (b),"

ML090920408, March 2009.

22. Letter to NRC, "Davis-Besse Nuclear Power Station, Unit No. 1 Docket No. 50-346, License No. NPF-3 License Amendment Request to Incorporate the Use of Alternate Methodologies for the Development of Reactor Pressure Vessel Pressure-Temperature Limit Curves, and Request for Exemption From Certain Requirements Contained in 10 CFR 50.61 and 10 CFR 50, Appendix G," ML091130228, April 2009.

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