NRC 2013-0036, To License Amendment Request 252 Technical Specification 5.6.5, Reactor Coolant System (RCS) Pressure and Temperature Limits Report

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To License Amendment Request 252 Technical Specification 5.6.5, Reactor Coolant System (RCS) Pressure and Temperature Limits Report
ML13113A008
Person / Time
Site: Point Beach  NextEra Energy icon.png
Issue date: 04/18/2013
From: Meyer L
Point Beach
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
NRC 2013-0036, TAC MF0532, TAC MF0533, TAC MF0534, TAC MF0535
Download: ML13113A008 (98)
v • d • e


Text

{{#Wiki_filter:April 18, 2013 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555 Point Beach Nuclear Plant, Units 1 and 2 Dockets 50-266 and 50-301 Renewed License Nos. DPR-24 and DPR-27 Supplement 2 to License Amendment Request 252 Technical Specification 5.6.5, Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR) NEXTera" ENERGY~ ,/ POINTBEA~ NRC 2013-0036 10 CFR 50.90

References:

(1) NextEra Energy Point Beach, LLC letter to NRC, dated January 15, 2013, License Amendment Request 252 Technical Specification 5.6.5, Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR) (ML13016A028) (2) NRC E-Mail to NextEra Energy Point Beach, dated March 27, 2013, Information Needed for Review (TAC Nos. MF0532, MF0533, MF0534, and MF0535) (ML13098B070) In Reference (1), NextEra Energy Point Beach, LLC (NextEra) submitted a license amendment request to amend renewed Facility Operating License Nos. DPR-24 and DPR-27 for Point Beach Nuclear Plant (PBNP), Units 1 and 2, respectively. The proposed amendments would revise the PBNP Technical Specifications (TS) to allow the use of two new methodologies: Framatome ANP Topical Report BAW-2308, Revisions 1-A and 2-A, "Initial RT NOT of Linde 80 Weld Materials," and Westinghouse Owners Group (WOG) WCAP-14040-A, Revision 4, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves." The revision would add BAW-2308, Revisions 1-A and 2-A and WCAP-14040-A, Revision 4, as approved methodologies to TS 5.6.5, "Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR)," for determining RCS pressure-temperature (PT) limits. In Reference (2), the NRC informed NextEra that supplemental information would be required in order for the LAR to meet acceptance review criteria. This letter provides the requested information. provides the requested supplemental information. Enclosure 2 provides a mark-up of TRM 2.2, Pressure Temperature Limits Report. The mark-up of TRM 2.2 in Reference (1) had incorrect temperatures listed for the RT PTS values on page 2.2-3. Enclosure 2 in this letter replaces in its entirety the mark-up of TRM 2.2, Pressure Temperature Limits Report, provided in Reference (1). NextEra Energy Point Beach, LLC, 6610 Nuclear Road, Two Rivers, WI 54241

Document Control Desk Page 2 The information contained in this letter does not alter the no significant hazards consideration contained in Reference (1) and continues to satisfy the criteria of 10 CFR 51.22 for categorical exclusion from the requirements of an environmental assessment. Approval of the proposed amendment is requested by January 1, 2014. NextEra will implement the amendment within 180 days of Commission Approval. This letter contains no new Regulatory Commitments and no revisions to existing Regulatory Commitments. The supplemental information to the LAR has been reviewed by the Plant Operations Review Committee. In accordance with 10 CFR 50.91, a copy of this letter is being provided to the designated Wisconsin Official. I declare under penalty of perjury that the foregOing is true and correct. Executed on April 18, 2013. Very truly yours, NextEra Energy Point Beach, LLC /~~. Lar Meyer Ite Vice President Enclosures cc: Administrator, Region III, USNRC Project Manager, Point Beach Nuclear Plant, USNRC Resident Inspector, Point Beach Nuclear Plant, USNRC PSCW

ENCLOSURE 1 NEXTERA ENERGY POINT BEACH, LLC POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2 SUPPLEMENT 2 TO LICENSE AMENDMENT REQUEST 252 TECHNICAL SPECIFICATION 5.6.5, REACTOR COOLANT SYSTEM (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR) SUPPLEMENTAL INFORMATION

1.0 INTRODUCTION

License Amendment Request (LAR) 252 (Reference 1) proposes to amend the Point Beach Nuclear Plant (PBNP) Technical Specifications' (TS) to allow the use of two new methodologies; Framatome ANP Topical Report BAW-2308, Revisions 1-A (Reference 2) and 2-A (Reference 3), "Initial RT NDT of Linde 80 Weld Materials," and Westinghouse Owners Group (WOG) WCAP-14040-A, Revision 4, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves" (Reference 4). The revision would add BAW-2308, Revisions 1-A and 2-A and WCAP-14040-A, Revision 4, as approved methodologies to TS 5.6.5, "Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR)," for determining RCS pressure-temperature (PT) limits. The NRC staff had indicated via an email dated March 27, 2013, (Reference 5) that additional information was required to evaluate use of the new methodologies at PBNP. On April 4, 2013, a teleconference between the NRC and PBNP staffs was held to discuss the NextEra Energy Point Beach, LLC (NextEra) submittal. The NRC reviewer informed NextEra that the calculations used to generate the adjusted reference temperature (ART) and reference temperature - pressurized thermal shock (RT-PTS) would be required for the review of the LAR. Additionally, the information submitted should document that the LAR meets the limitations discussed in the NRC safety evaluation report (SER) approving the BAW-2308 reports.

2.0 BACKGROUND

10 CFR 50.61 (a)(5) and 10 CFR 50, Appendix G(II)(D)(i), require that the pre-service or unirradiated condition reference nil-ductility temperature (RT NDT) be evaluated according to the procedures in the American Society for Mechanical Engineers (ASME) Code, Section III, Paragraph NB-2331, which requires Charpy V-notch impact tests and drop weight tests. Topical Report BAW-2308, Revisions 1-A and 2-A, provide an NRC-approved alternate method for determining the initial, unirradiated material reference temperatures of the Linde 80 weld materials present in the beltline region of the PBNP reactor pressure vessel (RPV). BAW-2308, Revisions 1-A and 2-A, were approved by the NRC for referencing in plant-specific license amendments in NRC Safety Evaluations (SEs) dated August 4,2005 (Reference 6) and March 24, 2008 (Reference 7), respectively. BAW-2308, Revision 2-A, is a supplement to Revision 1-A, and incorporated additional test data and a re-evaluation of the reference temperature, To, determination, as requested by the NRC in the SE for Revision 1-A. The SE for Revision 2-A of BAW-2308, states that the Conditions and Limitations, Items (1) through (4), contained in the SE for Revision 1-A of BAW-2308, must be addressed in all future plant-specific applications referencing Topical Report BAW-2308, Revisions 1-A and 2-A. Conditions and Limitations Item (1) contained in the SE for Revision 1-A of BAW-2308 states: The IRTTOand m values given in Table 3 of this SE may be used by a licensee to define the initial heat-specific or generic properties of its facility's Linde 80 welds. For those Linde 80 weld wire heats for which heat-specific values are given, those values must be used when applying TR BAW-2308, Revision 1 if the heat-specific IRhovalue is more conservative than the generic "all heats" IRho value. Page 1 of 3

Conditions and Limitations Item (2) contained in the SE for Revision 1-A of BAW-230B states: When the values from Table 3 of this SE are used by a licensee, the methodology of RG 1.99, Revision 2 may be used for the purpose of assessing the shift in initial properties due to irradiation, even though the RG 1.99, Revision 2 methodology is based upon Charpy V-notch 30 ft-Ib energy level shift data. However, based on the information in TR BAW-2308, Revision 1 (see Figure 3 of this SE), a minimum chemistry factor of 167 of must be applied when using initial properties given in Table 3 of this SE. A higher chemistry factor may be required if weld wire heat-specific chemical composition or Charpy V-notch surveillance data indicate, via the methodology of RG 1.99, Revision 2, that a higher chemistry factor should apply. Conditions and Limitations Item (3) contained in the SE for Revision 1-A of BAW-230B states: When the values from Table 3 of this SE are used by a licensee, a value of af::. = 28 of must be used to determine the overall margin term, when the margin term per TR BAW-2308, Revision 1 is defined as: Conditions and Limitations Item (4) contained in the SE for Revision 1-A of BAW-230B states: Any licensee who wants to utilize the methodology of TR BAW-2308, Revision 1 as outlined in items (1) through (3) above, must request an exemption, per 10 CFR 50.12, from the requirements of Appendix G to 10 CFR Part 50 and 10 CFR 50.61 to do so. As part of a licensee's exemption request, the NRC staff expects that the licensee will also submit information which demonstrates what values the licensee proposes to use for,1RTNDT and the margin term for each Linde 80 weld in its RPV through the end of its facility's current operating license. 3.0 REQUESTED INFORMATION Attachments 1 and 2 of this Enclosure contain the calculations used to generate the adjusted reference temperature (ART) and reference temperature - pressurized thermal shock (RT-PTS), respectively. The four Conditions and Limitations required by the NRC SE have been met in the calculations as described below: Conditions and Limitations (1), (2) and (3) contained in the SE for Revision 1-A of BAW-230B were utilized in the two attached calculations. o, Table 1 shows compliance with the IRTrovalue portion of Conditions and Limitations (1) and also shows compliance with Conditions and Limitations (2). o, Table 7 shows compliance with the m values portion of Conditions and Limitations (1) and also shows compliance with Conditions and Limitations (3). Conditions and Limitations (4) is met in the original submittal to the NRC (Reference 1), which included the exemption request. Page 2 of 3

4.0 REFERENCES

(1) NextEra Energy Point Beach, LLC letter to NRC, dated January 15, 2012, License Amendment Request 252 Technical Specification 5.6.5, Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR) (ML13016A028) (2) Framatome ANP Topical Report BAW-2308, Revision 1-A, "Initial RT NOT of Linde 80 Weld Materials," approved August 2005 (3) Framatome ANP Topical Report BAW-2308, Revision 2-A, "Initial RT NOT of Linde 80 Weld Materials," approved March 2008 (4) Westinghouse Owners Group (WOG) WCAP-14040-NP-A, Revision 4, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves", dated May 2004 (5) NRC E-Mail to NextEra Energy Point Beach, dated March 27, 2013, Information Needed for Review (TAC Nos. MF0532, MF0533, MF0534, and MF0535) (ML13098B070) (6) NRC letter to Framatome ANP, dated August 4, 2005, Final Safety Evaluation for Topical Report BAW-2308, Revision 1, "Initial RT NOT of Linde 80 Weld Materials" (TAC No. MB6636) (ML052070408) (7) NRC letter to Westinghouse Electric Company, dated March 24, 2008, Final Safety Evaluation for Pressurized Water Reactor Owners Group (PWROG) Topical Report (TR) BAW-2308, Revision 2, "Initial RTNOT of Linde 80 Weld Materials" (TAC No. MD4241) (ML080770349) Page 3 of 3

ENCLOSURE 1 ATTACHMENT 1 NEXTERA ENERGY POINT BEACH, LLC POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2 SUPPLEMENT 2 TO LICENSE AMENDMENT REQUEST 252 TECHNICAL SPECIFICATION 5.6.5, REACTOR COOLANT SYSTEM (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR) ART VALUES FOR POINT BEACH UNIT 1 AND UNIT 2 51 pages follow

20697-10 (3/30106) A CALCULATION

SUMMARY

SHEET (CSS) AREVA Document Identifier 32-9019240-000 Title ART Values for Point Beach Unit 1 and Unit 2 PREPARED BY: REVIEWED BY: METHOD: ~ DETAILED CHECK 0 INDEPENDENT CALCULATION NAME S. B. Davidsaver NAME J. B. Hall SIGNATURE ~.~. DuAtkJau61 SIGNATURE (413 ~. TITLE Engineer II DATE ~Lob TITLE Princ. E'neer DATE {J..o -(j6 COST REF. TMSTATEMENT, ~ CENTER 41324 PAGE(S) 51 REVIEWER INDEPENDENCE ~ NAME IS.. ('~I\\'\\ 1Ja,~ PURPOSE AND

SUMMARY

OF RESULTS: The purpose of this analysis is to determine the reactor vessel adjusted reference temperatures (ART) at the y. thickness (y.n and % thickness (%n wall locations for Point Beach Units 1 and 2 using the projected fluences for eight "cases" for each Unit The ART values are calculated for the Point Beach Units 1 and 2 and the limited beltline materials for each Unit and each "case" are shown in the table below. Y.TART %TA Value V ? Hf Rods? EFPY Case Unit Limiting Beltline Material Heat Number (OF) Yes Yes 53 1 1 Intermediate Shell LonQitudinal Weld SA-812/SA-775 218.8 183.3 Yes Yes 43 2 1 Intermediate Shell Longitudinal Weld SA-812/SA-775 209.7 173.6 No Yes 53 3 1 Intermediate Shell Longitudinal Weld SA-812/SA-775 217.2 181.6 No Yes 43 4 1 Intermediate Shell LonQitudinal Weld SA-812/SA-775 208.7 172.4 Yes Removal 10/08 53 5 1 Intermediate Shell Longitudinal Weld SA-812/SA-775 220.0 184.6 Yes Removal 10/08 43 6 1 Intermediate Shell Longitudinal Weld SA-812/SA-775 210.5 174.6 No Removal 10/08 53 7 1 Intermediate Shell Longitudinal Weld SA-812/SA-775 218.3 182.8 No Removal 10/08 43 ~---...L_...!!:I.~~_r:.J!I_~!~~~_?_~~!!-_~~ll!~~inal Weld __ SA-812/SA-775 209.3 173.2

yes------- ----53----

c-SA-14Sr----------r-f55.S--- ---2T8~4----- Yes 1 2 Intermediate to Lower Shell Weld Yes Yes 43 2 2 Intermediate to Lower Shell Weld SA-1484 247.7 209_5 No Yes 53 3 2 Intermediate to Lower Shell Weld SA-1484 254.5 216.9 No Yes 43 4 2 Intermediate to Lower Shell Weld SA-1484 246.8 208.5 Yes Removal 04/08 53 5 2 Intermediate to Lower Shell Weld SA-1484 265.5 229.2 Yes Removal 04/08 43 6 2 Intermediate to Lower Shell Weld SA-1484 254.9 217.3 No Removal 04/08 53 7 2 Intermediate to Lower Shell Weld SA-1484 263.4 227.0 No Removal 04/08 43 8 2 Intermediate to Lower Shell Weld SA-1484 253.3 215.7 THE DOCUMENT CONTAINS ASSUMPTIONS THAT MUST BE VERIFIED PRIOR TO USE ON THE FOLLOWING COMPUTER CODES HAVE BEEN USED IN THIS DOCUMENT: SAFETY-RELATED WORK CODENERSION/REV CODENERSION/REV D YES ~ NO AREVA NP Inc., an AREVA and Siemens company Page_1_of~

A AREVA NON-PROPRIETARY 32-9019140-000 RECORD OF REVISIONS Revision Description Date 000 Original Release June 2006 Page 2 of 51

A AREVA NON-PROPRIETARY 32-9019240-000 TABLE OF CONTENTS 1.0 Introduction............................................................................................................. 8 2.0 Smnmary of Results................................................................................................ 9 3.0 Assumptions.......................................................................................................... 10 4.0 Reactor Vessel Fluence......................................................................................... 10 4.1 Reactor Vessel Inner Surface Fluences............................................................. 10. 4.2 Attenuation Through Reactor Vessel Wall....................................................... 11 5.0 Adjusted Reference Temperature Where No Surveillance Data Is Available...... 24 5.1 Initial RTNDI..................................................................................................... 24 5.2 ARTNDT............................................................................................................. 24 5.2.1 Chemistry Factor....................................................................................... 25 5.2.2 Fluence Factor........................................................................................... 25 5.2.3 ARTNDI Calculation.................................................................................. 30 5.3 Margin............................................................................................................... 33 5.4 Calculation of Adjusted Reference Temperature (ART).................................. 38 6.0 Adju*sted Reference Temperature Calculation Where Surveillance Data is Available........................................................................................................................... 43 6.1 Surveillance Data Credibility Assessment........................................................ 43 6.2 Credible Surveillance Data............................................................................... 44 6.3 Non Credible Surveillance Data....................................................................... 44 6.4 Assessment of the Weld Wire Heat Surveillance Data..................................... 44 7.0 References...................................................... :...................................................... 51 Page 3 of 51

A. AREVA NON-PROPRIETARY 32-9019240-000 LIST OF TABLES Table 1. Case Descriptions................................................................................................ 9 Table 2. Summary of Results for Point Beach Unit 1 and Unit 2 Adjusted Reference Temperatures..................................................................................................................... 10 Table 3. Adjusted Reference Temperature Evaluation for the Point Beach Unit 1 and Unit 2 Reactor Vessel Beltline Materials with Uprate, with hafnium rods, through 53 EFPY (Case 1).................................................................................................................. 12 Table 4. Adjusted Reference Temperature Evaluation for the Poirit Beach Unit 1 and Unit 2 Reactor Vessel Beltline Materials with Uprate, with hafnium rods, through 43 EFPY (Case 2).................................................................................. ~............................... 13 Table 5. Adjusted Reference Temperature Evaluation for the Point Beach Unit 1 and Unit 2 Reactor Vessel Beltline Materials without Uprate, with hafuium rods, through 53 EFPY (Case 3).................................................................................................................. 14 Table 6. Adjusted Reference Temperature Evaluation for the Point Beach Unit 1 and Unit 2 Reactor Vessel Beltline Materials without Uprate, with hafuium rods, through 43 EFPY (Case 4).................................................................................................................. 15 Table 7. Adjusted Reference Temperature Evaluation for the Point Beach Unit 1 and Unit 2 Reactor Vessel Beltline Materials with Uprate, with hafnium removal October 2008 (Unit 1) and April 2008 (Unit 2), through 53 EFPY (Case 5)................................. 16 Table 8. Adjusted Reference Temperature Evaluation for the Point Beach Unit 1 and Unit 2 Reactor Vessel Beltline Materials with Uprate, with hafnium removal October 2008 (Unit 1) and April 2008 (Unit 2), through 43 EFPY (Case 6)................................. 17 Table 9. Adjusted Reference Temperature Evaluation for the Point Beach Unit 1 and Unit 2 Reactor Vessel Beltline Materials without Uprate, with hafnium removal October 2008 (Unit 1) and April 2008 (Unit 2), through 53 EFPY (Case 7)................................. 18 Table 10. Adjusted Reference Temperature Evaluation for the Point Beach Unit 1 and Unit 2 Reactor Vessel Beltline without Uprate, with hafnium removal October 2008 (Unit

1) and April 2008 (Unit 2), through 43 EFPY (Case 8).................................................... 19 Table 11. Fluence (E>1.0 MeV) Values for the Point Beach Unit 1 and Unit 2 Vessel Beltline Materials with Uprate, with hafuium rods, through 53 EFPY (Case 1).............. 20 Table 12. Fluence (E> 1.0 MeV) Values for the Point Beach Unit 1 and Unit 2 Vessel Beltline Materials with Uprate, with hafnium rods, through 43 EFPY (Case 2).............. 20 Table 13. Fluence (E> 1.0 MeV) Values for the Point Beach Unit 1 and Unit 2 Vessel Beltline Materials without Uprate, with hafnium rods, through 53 EFPY (Case 3)......... 21 Table 14. Fluence (E> 1.0 MeV) Values for the Point Beach Unit 1 and Unit 2 Vessel Beltline Materials without Uprate, with hafnium rods, through 43 EFPY (Case 4)......... 21 Table 15. Fluence (E> 1. 0 MeV) Values for the Point Beach Unit 1 and Unit 2 Vessel Beltline Materials with Uprate, with hafnium rod removal October 2008 (Unit 1) and Apri12008 (Unit 2); through 53 EFPY (Case 5)............................................................... 22 Table 16. Fluence (E>1.0 MeV) Values for the Point Beach Unit 1 and Unit 2 Vessel Beltline Materials with Uprate, with hafnium rod removal October 2008 (Unit 1) and April 2008 (Unit 2), through 43 EFPY (Case 6)............................................................... 22 Page 4 of 51

A AREVA NON-PROPRIETARY 32-9019240-000 Table 17. Fluence (E> 1.0 MeV) Values for the Point Beach Unit 1 and Unit 2 Vessel Beltline Materials without Uprate, with hafnium rod removal October 2008 (Unit 1) and April 2008 (Unit 2), through 53 EFPY (Case 7)............................................................... 23 Table 18. Fluence (E> 1.0 MeV) Values for the Point Beach Unit 1 and Unit 2 Vessel Beltline Materials without Uprate, with hafnium rod removal October 2008 (Unit 1) and April 2008 (Unit 2), through 43 EFPY (Case 8)............................................................... 23 Table 19. Initial RT NDT Values for the Point Beach Unit 1 and Unit 2 Reactor Vessel Beltline Materials.............................................................................................................. 24 Table 20. Regulatory Guide 1.99, Revision 2 Chemistry Factors for the Point Beach Unit 1 and Unit 2 Reactor Vessel Beltline Materials................................................................ 25 Table 21. Fluence Factors for the ~T and %T Locations of the Point Beach Unit 1 and Unit 2 Reactor Vessel Beltline Materials with Uprate, with hafnium rods, through 53 EFPY (Case 1)................................................................................................................... 26 Table 22. Fluence Factors for the ~T and %T Locations of the Point Beach Unit 1 and Unit 2 Reactor Vessel Beltline Materials with Uprate, with hafnium rods, through 43 EFPY (Case 2).................................................................................................................. 26 Table 23. Fluence Factors for the ~T and %T Locations of the Point Beach Unit 1 and Unit 2 Reactor Vessel Beltline Materials without Uprate, with hafnium rods, through 53 EFPY (Case 3).................................................................................................................. 27 Table 24. Fluence Factors for the ~T and %T Locations of the Point Beach Unit 1 and Unit 2 Reactor Vessel Beltline Materials without Uprate, with hafnium rods, through 43 EFPY (Case 4).................................................................................................................. 27 Table 25. Fluence Factors for the Y4 T and % T Locations of the Point Beach Unit I and Unit 2 Reactor Vessel Beltline Materials with Uprate, with hafnium rod removal October 2008 (Unit 1) and April 2008 (Unit 2), through 53 EFPY (Case 5)................................. 28 Table 26. Fluence Factors for the ~T and 314T Locations of the Point Beach Unit 1 and Unit 2 Reactor Vessel Beltline Materials with Uprate, with hafnium rod removal October 2008 (Unit 1) and April 2008 (Unit 2), through 43 EFPY (Case 6)................................. 28 Table 27. Fluence Factors for the Y4T and %T Locations of the Point Beach Unit 1 and Unit 2 Reactor Vessel Beltline Materials without Uprate, with hafnium rod removal October 2008 (Unit 1) and April 2008 (Unit 2), through 53 EFPY (Case 7)................... 29 Table 28. Fluence Factors for the Y4T and %T Locations of the Point Beach Unit 1 and Unit 2 Reactor Vessel Beltline Materials without Uprate, with hafnium rod removal October 2008 (Unit 1) and April 2008 (Unit 2), through 43 EFPY (Case 8)................... 29 Table 29. ~RTNDT Values for the Y4T and %T Wall Locations of the Point Beach Unit 1 and Unit 2 Beltline Materials with Uprate, with hafnium rods, through 53 EFPY (Case 1) ........................................................................................................................................... 30 Table 30. ~RTNDT Values for the ~T and %T Wall Locations of the Point Beach Unit 1 and Unit 2 Beltline Materials with Uprate, with hafnium rods, through 43 EFPY (Case 2) ........................................................................................................................................... 30 Table 31. ~RTNDT Values for the Y4T and %T Wall Locations of the Point Beach Unit 1 and Unit 2 Beltline Materials without Uprate, with hafnium rods, through 53 EFPY (Case

3)....................................................................................................................................... 31 PageS ofS1

A AREVA NON-PROPRIETARY 32-9019240-000 Table 32. Lill.TNDT Values for the 'l4T and %T Wall Locations of the Point Beach Unit 1 and Unit 2 Beltline Materials without Uprate, with hafuium rods, through 4.3 EFPY (Case

4).......................................... :............................................................................................ 31 Table 33. ~RTNDT Values for the 'l4T and %T Wall Locations of the Point Beach Unit 1 and Unit 2 Beltline Materials with Uprate, with hafnium rod removal October 200S (Unit
1) and Apri1200S (Unit 2), through 53 EFPY (Case 5).................................................... 32 Table 34. Lill.TNDT Values forthe 'l4T and %T Wall Locations of the Point Beach Unit 1 and Unit 2 Beltline Materials with Uprate, with hafnium rod removal October 200S (Unit
1) and April200S (Unit 2), through 43 EFPY (Case 6).................................................... 32 Table 35. ~RTNDT Values for the 'l4T and %T Wall Locations of the Point Beach......... 33 Table 36. ~RTNDI Values for the 'l4T and 3i4T Wall Locations of the Point Beach......... 33 Table 37. Margin Values for the 'l4T and %T Wall Locations ofthe Point Beach Unit 1 and Unit 2 Reactor Vessel Be1tline Materials with Uprate, with hafnium rods, through 53 EFPY (Case 1).................................................................................................................. 34 Table 3S. Margin Values for the 'l4T and %T Wall Locations ofthe Point Beach Unit 1 and Unit 2 Reactor Vessel Belt1ine Materials with Uprate, with hafnium rods, through 43 EEPY (Case 2).................................................................................................................. 35 Table 39. Margin Values for the 'l4T and %T Wall Locations ofthe Point Beach Unit 1 and Unit 2 Reactor Vessel Beltline Materials without Uprate, with hafnium rods, through 53 EFPY (Case 3)............................................................................................................. 35 Table 40. Margin Values for the 'l4T and %T Wall Locations of the Point Beach Unit 1 and Unit 2 Reactor Vessel Beltline Materials without Uprate, with hafnium rods, through 43 EFPY (Case 4)............................................................................................................. 36 Table 41. Margin Values for the ~T and %T Wall Locations of the Point Beach Unit 1 and Unit 2 Reactor Vessel Beltline Materials with Uprate, with hafnium rod removal October 200S (Unit 1) and Apri1200S (Unit 2), through 53 EFPY (Case 5)................... 36 Table 42. Margin Values for the 'l4T and %T Wall Locations of the Point Beach Unit 1 and Unit 2 Reactor Vessel Beltline Materials with Uprate, with hafnium rod removal October 200S (Unit 1) and April 2008 (Unit 2), through 43 EFPY (Case 6)................... 37 Table 43. Margin Values for the ~T and %T Wall Locations of the Point Beach Unit 1 and Unit 2 Reactor Vessel Beltline Materials without Uprate, with hafnium rod removal October 2008 (Unit 1) and April200S (Unit 2), through 53 EFPY (Case 7)................... 37 Table 44. Margin Values for the ~T and %T Wall Locations of the Point Beach Unit 1 and Unit 2 Reactor Vessel Beltline Materials without Uprate, with hafnium rod removal October 200S (Unit 1) and April200S (Unit 2), through 43 EFPY (Case 8)................... 3S Table 45. ART Values for the ~T and %T Wall Locations of the Point Beach Unit 1 and Unit 2 Reactor Vessel Beltline Materials with Uprate, with hafnium rods, through 53 EFPY (Case 1).................................................................................................................. 39 Table 46. ART Values for the ~T and %T Wall Locations of the Point Beach Unit 1 and Unit 2 Reactor Vessel Beltline Materials with Uprate, with hafnium rods, through 43 EFPY (Case 2).................................................................................................................. 39 Table 47. ART Values for the ~ T and % T Wall Locations of the Point Beach Unit 1 and Unit 2 Reactor Vessel Beltline Materials without Uprate, with hafnium rods, through 53 EFPY (Case 3).................................................................................................................. 40 Page 6 of 51

A AREVA NON-PROPRIETARY 32-9019240-000 Table 48. ART Values for the V4T and %T Wall Locations of the Point Beach Unit 1 and Unit 2 Reactor Vessel Beltline Materials without Uprate, with hafnium rods, through 43 EFPY (Case 4).................................................................................................................. 40 Table 49. ART Values for the V4T and %T Wall Locations of the Point Beach Unit 1 and Unit 2 Reactor Vessel Beltline Materials with Uprate, with hafnium rod removal October 2008 (Unit 1) and April 2008 (Unit 2), through 53 EFPY (Case 5)................................. 41 Table 50. ART Values for the V4T and %T Wall Locations of the Point Beach Unit 1 and Unit 2 Reactor Vessel Beltline Materials with Uprate, with hafnium rod removal October 2008 (Unit 1) and April 2008 (Unit 2), through 43 EFPY (Case 6)................................. 41 Table 51. ART Values for the V4T and %T Wall Locations of the Point Beach Unit 1 and Unit 2 Reactor Vessel Beltline Materials without Uprate, with hafuium rod removal October 2008 (Unit 1) and April 2008 (Unit 2), through 53 EFPY (Case 7)................... 42 Table 52. ART Values for the V4T and %T Wall Locations of the Point Beach Unit 1 and Unit 2 Reactor Vessel Beltline Materials without Uprate, with hafuium rod removal October 2008 (Unit 1) and April 2008 (Unit 2), through 43 EFPY (Case 8)................... 42 Table 53. Credibility Assessment for Weld Wire Heat Number 71249.......................... 45 Table 54. Table Chemistry Factor Non-Conservatism Assessment for Weld Wire Heat Number 71249.................................................................................................................. 46 Table 55. Credibility Assessment for Weld Wire Heat Number 61782.......................... 47 Table 56. Table Chemistry Factor Non-Conservatism Assessment for Weld Wire Heat Number 61782.................................................................................................................. 48 Table 57. Credibility Assessment for Weld Wire Heat Number 72442[]........................ 49 Table 58. Table Chemistry Factor Non-Conservatism Assessment for Weld Wire Heat Number 72442.................................................................................................................. 50 Page 7 of 51

t\\ AREVA NON-PROPRIETARY 32-9019240-000 1.0 Introduction The purpose of this analysis is to detennine the reactor vessel adjusted reference temperatures (ART) at the 'l4-thiclmess ('l4T) and %-thiclmess (%T) wall locations for Point Beach Unit 1 and Unit 2 using the project~d 60 year fluences. The ART values are calculated for Pojnt Beach Unit 1 and Unit 2 reactor vessel beltline materials applicable to both 43 and 53 effective full power years (EFPY). The ART values for eight "cases" will be calculated, for each Unit. The cases are summarized below in Table 1. Page 8 of 51

A AREVA NON-PROPRIETARY 32-9019240-000 Table 1. Case Descriptions LTR-REA-04-64 Case Unit Power (MWth). EFPY Hafnium Rods? Case Number 1518.5 staItup to 2/3/2003 1 1 1 1540.02/3/2003 to 10/2008 53 Yes 1678.010/2008 to 53 EFPY 1518.5 startup to 2/3/2003 5 1 2 1540.02/3/2003 to 4/2008 53 Yes 1678.04/2008 to 53 EFPY 1518.5 startup to 2/3/2003 1 2 1 1540.02/3/2003 to 10/2008 43 Yes 1678.0 1012008 to 43 EFPY 1518.5 strutup to 2/3/2003 5 2 2 1540.0 2/3/2003 to 412008 43 Yes 1678.04/2008 to 43 EFPY 2 3 1 1518.5 startup to 2/3/2003 53 Yes 1540.02/312003 to 53 EFPY 6 3 2 1518.5 staItup to 2/3/2003 53 Yes 1540.02/3/2003 to 53 EFPY 2 4 1 1518.5 strutup to 2/3/2003 43 Yes 1540.02/3/2003 to 43 EFPY 6 4 2 1518.5 strutup to 2/3/2003 43 Yes 1540.0 2/3/2003 to 43 EFPY 1518.5 strutup to 2/3/2003 3 5 1 1540.0 2/3/2003 to 10/2008 53 Removal October 2008 1678.0 10/2008 to 53 EFPY 1518.5 startup to 2/3/2003 7 5 2 1540.0 2/312003 to 412008 53 Removal April 2008 1678.04/2008 to 53 EFPY 1518.5 startup to 2/3/2003 3 6 1 1540.0 2/312003 to 10/2008 43 Removal October 2008 1678.0 10/2008 to 43 EFPY 1518.5 staItup to 2/3/2003 7 6 2 1540.02/3/2003 to 412008 43 Removal April 2008 1678.04/2008 to 43 EFPY 4 7 1 1518.5 startup to 2/3/2003 53 Removal October 2008 1540.02/312003 to 53 EFPY 8 7 2 1518.5 startup to 2/3/2003 53 Removal April 2008 1540.0 2/3/2003 to 53 EFPY 4 8 1 1518.5 strutup to 2/3/2003 43 Removal October 2008 1540.02/3/2003 to 43 EFPY 8 8 2 1518.5 strutup to 2/3/2003 43 Removal April 2008 1540.0 2/3/2003 to 43 EFPY 2.0 Summary of Results The Y4T and %T ART values for the Point Beach Unit 1 and Unit 2 reactor vessel beltline materials applicable to the EFPY are listed in Tables 3 - 10. These values were calculated in accordance with Regulatory Guide 1.99, Revision 2. [1] Based on the analysis, the limiting beltline material for the Point Beach Unit 1 and Unit 2 reactor vessel is shown below in Table 2, for each case. Page 9 of 51

A AREVA NON-PROPRIETARY 32-9019240-000 Table 2. Summary of Results for Point Beach Unit 1 and Unit 2 Adjusted Reference Temperatures Y4T ART %TART Value Value Uprate? HfRods? EPPY Case Unit Limiting.Beltline Material Heat Number (OP) (OP) Yes Yes 53 1 1 Intermediate Shell Longitudinal Weld SA-SI2/SA-775 21S.S IS3.3 Yes Yes 43 2 1 Intermediate Shell Longitudinal Weld SA-SI2/SA-775 209.7 173.6 No Yes 53 3 1 Intermediate Shell Longitudinal Weld SA-SI2/SA-775 217.2 181.6 No Yes 43 4 1 Intermediate Shell Longitudinal Weld SA-S12/SA-775 208.7 172.4

  • Yes Removal 10/08 53 5

1 Intermediate Shell Longitudinal Weld SA-SI2/SA-775 220.0 IS4.6 Yes Removal 10/0S 43 6 1 Intermediate Shell Longitudinal Weld SA-SI2/SA-775 210.5 174.6 I No Removal 10/0S 53 7 1 Intermediate Shell Longitudinal Weld SA-SI2/SA-775 21S.3 182.8 No Removall010S 43 S 1 ~~~_c:l~ate Shell Longitudin~!_!V eld SA-SI2/SA-775 209.3 173.2 ~---- Yes Yes 53 1 2 Intermediate to Lower Shell Weld SA-1484 255.S 21S.4 Yes Yes 43 2 2 Intermediate to Lower Shell Weld SA-1484 247.7 209.5 No Yes 53 3 2 Intermediate to Lower Shell Weld SA-1484 254.5 216.9 No Yes 43 4 2 Intermediate to Lower Shell Weld SA-1484 246.8 208.5 . Yes Removal 04/0S 53 5 2 Intermediate to Lower Shell Weld SA-1484 265.5 229.2 Yes Removal 04/0S 43 6 2 Intermediate to Lower Shell Weld SA-1484 254.9 217.3 No Removal 04/08 53 7 2 Intermediate to Lower Shell Weld SA-1484 263.4 227.0 No Removal 04/08 43 8 2 Intermediate to Lower Shell Weld SA-1484 253.3 215.7 3.0 Assumptions No major assumptions are contained in this calculation. 4.0 Reactor Vessel Fluence 4.1 Reactor Vessel Inner Surface Fluences The inner surface neutron fluence is the calculated value defined at clad/low alloy steel interface of the Point Beach Unit 1 and Unit 2 reactor vessels. The projected 43 and 53 EPPY inner surface fluences for the Point Beach Unit 1 and Unit 2 reactor vessel beltline materials are listed in Table 11-Table 18.[2] Page 10 of 51 . -----.,.,--.......... -.. --- **.. -1'-**_*----_******-********** '.. -.-.-. -~--'.-... '.,............

A AREVA NON-PROPRIETARY 32-9019240-000 4.2 Attenuation Through Reactor Vessel Wall In accordance with Regulatory Guide 1.99, Revision 2, the neutron fluence at the 'l4T and %T wall locations in the vessel is determined as follows: where fsurf (10 19 n/cm2, E> 1.0 Me V) is the calculated value of the neutron fluence at the clad/low alloy steel interface, and "x" (in inches) is the depth into the vessel wall measured from the clad/low alloy steel interface. The Point Beach Unit 1 and Unit 2 reactor vessel thickness is reported to be 6.50 inches. [3] Therefore, the depth into the vessel wall measured from the vessel inner (wetted) surface, "x", is for 'l4T = [6.5*0.25] = 1.625 inches and x for %T = [6.5*0.75] = 4.875 inches. The 'l4T ART value for the intermediate shell longitudinal weld for Unit 1 will be calculated for SA-812 (ID 27%) and the %T ART value for the intermediate shell longitudinal weld for Unit 1 will be calculated for SA-775 (OD 73%). Using these vessel wall depths and the neutron fluence at the inner wetted surface of the vessel, the 'l4T and %T fluence values for the Point Beach Unit 1 and Unit 2 reactor vessel beltline materials are calculated in accordance with Equation 1, and these values are listed in Tables 11 - 18. Page 11 of 51 (1)

A AREVA NON-PROPRIETARY 32-9019240-000 Table 3. Adjusted Reference Temperature Evaluation for the Point Beach Unit 1 and Unit 2 Reactor Vessel Beltline Materials 122P237 A9SII-' CI423-' 122P237 A9SII-' CI423-. 123V352 123V500 123Wl95 21935 72442 with with hafnium 53 EFPY Type SA-50S Cl. 2 SA-50S Cl. 2 SA-50S Cl. 2 Linde 80 Flux Linde gO Flux Chenucal ComDosition Cu wt% 0.11 0.20 0.12 0.18 0.26 Ni wt% 0.S2 0.06 0.07 [J - Controlling values. of the adjusted reference temperatures.

  • -Determined from surveillance data.

IJ 1.2 225.0 3/4T T/4 Margm 3/4T ART, of .t 53EFPY T/4 3/4T Page 12 of 51

A AREVA NON-PROPRIETARY 32-9019240-000 Table 4. Adjusted Reference Temperature Evaluation for the Point Beach Unit 1 and Unit 2 Reactor Vessel Beltline Materials with Uprate, with hafnium rods, through 43 EFPY (Case 2) ChemIcal 6R.TNDT* OF ART,oF Material D~scription Composition 43 EFPY Fluenee, nlem' .t43 EFPY .,43 EFPY Reactor Vessel 3/4T T/4 3/4T T/4 3/4T T/4 3/4T Bcltline Rcgton Location Location Location Location Point Beach Unit 1 Evaluation Nozzle Bell Forging (NB) 12ZP237 I 22P237 SA-508 CI. 2 27,6 124.0 Intermedi.'e Shell Plate (IS) A9SII-l A98I1-' SA-508 CI. 2 56.4 157.6 Lower Shell Plate (LS) CI423-. CI423-' SA-S08 CI. 2 35.4 56.4 100.5 NB to IS Cire. Weld (100%) SA-1426 STl762 Linde 80 Flux 0.19 0.57 1.61£+18 IS Long. Weld (10 27"10) SA-SI2 IP0815 Linde 80 Flux 0.17 0.52 1.71£+19 IS Long. Weld (00 73%) SA-77S IP0661 Linde 80 Flux 0.17 0.64 N/A Intermc,,',diate to LS eire. Weld (100%) SA-llOI 71249 Linde 80 Flux 0.23 0.59 LS Lon', Weld (100% SA-847 61782 Linde 80 Flux 0.23 0.52 Point Beach Unit 2 Evaluation No""le Belt Forging (NB) 123V3S2 123V352 SA-508 CI. 2 0.11 0.73 Intermediate Shell Forging (IS) 123VSOO 123V500 SA-50S CI. 2 0.09 0.70 Lower Shell Forging (LS) 123WI95 123WI95 SA-50SCI. 2 0.05 0.72 42.8* NB to IS Cire. Weld (100%) 21935 21935 Linde SO Flux O.IS 0.70 _56 170.5 3.42E+1S 2.32E+18 l.o6£+IS 103.2 73.1 65.5 65.5 Intermediate to LS eire. Weld {IOO% SA-14S4 72442 Linde SO Flux 0.26 0.60 -30 IS0.0 3.14£+19 2.I3E+19 9.75£+18 216.9 17S.7 60.S 60.S [] - Controlling values of the adjusted reference temperatures,

  • - Determined from surveillance data, Page 13 of 51

A AREVA NON-PROPRIETARY 32-9019240-000 Table 5. Adjusted Reference Temperature Evaluation for the Point Beach Unit 1 and Unit 2 Reactor Vessel Beltline Materials without Uprate, with hafnium rods, through 53 EFPY (Case 3) Chemical Material Description Composition Reactor Vessel Mati. Heat Cu Ni Initial Chemistry Bc:ltline Region Location Ident. Number Type wt% W!"10 RTNDT Factor ':':j;,.:,:;.*.:'.'i'~'\\l, I:.';:!l; POLnt Beach Unit 1 EvalUation <!:~!-:::j!.::l

  • C.

Nozzle Belt Forging (NS) I 22P237 122P237 SA-50S CI. 2 0.11 0.S2 50 In,ermediate Shell Plate (IS) A9SII-' A9811-' SA-50S Cl. 2 0.20 0.06 1 Lower Shell Pla'e (LS) C1423-1 C1423-1 SA-50S CI. 2 0.12 0.07 I NB to IS Circ. Weld (100%) SA-1426 ST1762 Linde SO Flux 0.19 0.57 -47.6 IS Long. Weld (ID 27%) SA-SI2 IP0815 Linde SO Flux 0.17 0.52 -47.6 IS Long. Weld (00 73%) SA-775 IP0661 Linde 80 Flux 0.17 0.64 -47.6 In(ermedint'to LS Cire. Weld (100%) SA-1l01 71249 Linde 80 Flux 0.23 0.59

  • 47.4 LS Lon '. Weld {I 00%)

SA-847 61782 Linde SO Flux 0.23 0.52 -47.6' Point Beadl Unit 2 EYllluation 0:'-,".- .',,; "".-...*,,:.),, *,:.~'if':~!c!~: Nozzle Belt Forging (NB) 123V352 123V352 SA-508 CI. 2 0.11 0.73 40 Intermediate Shell ForgIng (is) 12JV500 123Y500 SA*50S Cl. 2 0.09 0.70 40 Lower Shell Forging (LS) 123WI95 123W195 SA-50S CI. 2 0.05 0.72 40 NB (0 IS Cire. Weld (100%) 21935 21935 Linde 80 Flux. O.IS 0.70

  • 56 Inlermediate to LS eire. Weld (100%)

SA-1484 72442 Linde 80 Flux 0.26 0.60 -30 [] - Controlling values of the adjusted reference temperatures.

  • -Determined from surveillance data.

77.0 79.3' 35.S* 167.0 167.0 167.0 167.6 167.0 CO:. :':'. 76.0 5S.0 42.S" 170.5 ISO.O ARTNOT* OF ART, OF 53 EFPY Fluence, nlcm'l .153 EFPY . MarSIn at 53 EFPY Clad/Low T/4 3/4T T/4 3/4T T/4 3/4T T/4 3/4T Alloy Stee, Interface LocatIon locatIon Location Location Location Location Location Location >~ ~~:';.:h.i~l<<fj,~;:!i i:~:-::> '.;:,', j;;:§:~~i~:h:';: ~i~j(:*~~t~;\\,:. ,~~~;~~~, 1:';fo.,.::;:;,,[(., -,), 2.75E+IS 1.S6E+18 S.54E+17 42.5 29.7 34.0 29.7 126.5 109.4 4.69E+19 3.ISE+19 1.46E+19 \\03.4 S7.6 56.4 56.4 160.S 145.0 3.59E+l9 2.43E+19 !.lIE+19 44.4 36.9 56.4 56.4 101.S 94.3 2.75E+IS l.86E+18 8.54E+17 92.2 64.5 65.7 65.7 \\l0.3 82.6 2.99E+19 2.02E+19 NlA 199.1 N/A 65.7 N/A [217.2] N/A 2.99E+19 N/A 9.2SE+18 N/A 163.5 NlA 65.7 N/A [ISI.6] 3.59E+19 2.43E+19 t.11E+19 207.7 172.6 61.7 61.7 222.0 IS6.9 2.5IE+19 1.70E+19 7.79E+lS 191.4 155.3 65.7 65.7 209.5 173.6 1?'4;~'D:'~';;' !'.

  • !,igif,ltf:f~::; " ::.':-<".
':,i Ii.;;:;':'

..,**.. *.. *\\;*;*";1 ".'!"")

~~
Nf~
; l*i~&)ti:f~!Hit'it 3.99E+1S 2.70E+1S 1.24E+IS
  • 4S.9 35.0 34.0 34.0 122.9 109.0 4.49E+19 3.04E+19 l.39E+19 75.1 63.3 34.0 34.0 149.1 137.3 4.02E+19 2.72E+19 1.25E+19 54.2 45.5 17.0 17.0 111.2 102.5 3.99E+18 2.70E+18 1.24E+1S 109.6 7S.6 65.5 65.5 119.1 88.1 3.64E+l9 2.46E+19 t.13E+19 223.7 IS6.1 60.S 60.8

. [254.5] [216.9] Page 14 of 51

A AREVA NON-PROPRIETARY 32-9019240-000 Table 6. Adjusted Reference Temperature Evaluation for the Point Beach Unit 1 and Unit 2 Reactor Vessel Beltline Materials without Uprate, with hafnium rods, through 43 EFPY (Case 4) Chemical ARTI'IDT. of ART,oF Material Description Composition 43 EFPY Ftuence, nJcm2 0143 EFPY Margin 0143 EFPY Reactor Vessel 314T TI4 314T TI4 314T TI4 314T Beltliut! Region Locatlon POint Beach Umt J Eva!uatIOn No:a:le Belt ForsJng (NB) I 22P237 122P237 SA*508 CI. 2 Inlerm,>diate Shell Plate (IS) A9BtI*1 A9811*1 SA*508 CI. 2 79.3* Lower Shell Plate (LS) C1423*1 C1423*1 SA*508CI. 2 35.S* NB to IS Cire. Weld (100%) SA*1426 BTl762 Linde BO Flux 0.19 0.57 -47.6 167.0 2.33E+18 1.5BE+18 7.23E+17 85.8 59.3 65.7 65.7 IS Long. Weld (10 27%) SA*BI2 IP0815 Linde BO Flux 0.17 0.52 -47.6 167.0 2.46E+19 1.67E+19 NIA 19Q.6 NlA 65.7 NIA IS Long. Weld (00 73%) SA*775 IP066 I UndeBO Flux 0.17 0.64 -47.6 167.0 2.46E+19 NIA 7.64E+18 N/A 154.3 N/A 65.7 Jntenncdiate to LS eire. Weld (100%) SA*1I01 71249 Linde 80 Flux 0.23 0.59 -47.4 167.6 3.03E+19 9.47E+18 200.4 164.B 61.7 61.7 LS Lon '. Weld 100% SA*B47 61782 Linde 80 Flux 0.23 0.52 -47.6 167.0 2.10E+19 6.52E+18 183.4 147.0 65.7 65.7 POint Beach Unit 2 Evaluation Nozzle Belt Forging (NB) 123V352 123V352 SA*508CI. 2 0.11 0.73 40 76.0 3.34E+18 2.26E+18 I.04E+18 45.5 32.2 34.0 32.2 Intermediate Shell Forging (IS) 123V500 123V500 SA*50B CI. 2 0.09 0.70 40 5S.0 3.71E+19 2.51E+19 1.15E+19 72.3 60.3 34.0 34.0 Lower Shell Forging (LS) 123WI95 123W195 SA*50S Cl. 2 0.05 0.72 40 42.S* 3.37E+19 2.28E+19 1.05E+19 52.3 43.4 17.0 17.0 NB to IS Cire. Weld (100%) 21935 21935 Linde BO Flux O.IS 0.70

  • 56 170.5 3.34E+1B 2.26E+IB 1.00E+18 102.1 72.3 65.5 65.5 lnrermediate to LS eire. Weld (100%

SA*1484 72442 Linde SO Flux 0.26 0.60

  • 30 ISO.O 3.0SE+19 2.09E+19 9.56E+18 216.0 177.7 60.B 60.B

[] - Controlling values of the adjusted reference temperatures.

  • - Determined from surveillance data.

Page 15 of 51

A AREVA NON-PROPRIETARY 32-9019240-000 Table 7. Adjusted Reference Temperature Evaluation for the Point Beach Unit 1 and Unit 2 Reactor Vessel Beltline Materials with Uprate, with hafnium removal October 2008 (Unit 1) and April 2008 (Unit 2), through 53 EFPY (Case 5) Chemical Material Descnpllon Composition Reactor Vessel Mati. Heat Cu Ni Initial Chemistry Belliine Region LocatIOn Iden!. Number Type wt% wt% RT"OT Factor POll1l Bench Umt 1 Evaluation i[';;~f}~k:,&: *EJ:i;~Ji~t~*}' ~iri!~:!~i;':... i:.L~:;':~:::n~ Nozzle Belt Forging eNB) I 22P237 122P237 SA-50S CL 2 0.11 0.S2 SO 77.0. Inlermediate Shell Plale (IS) A98II-' A9SlI-' SA-50S CL2 0.20 0.06 I 79.3' Lower Shell Plate (LS) C1423-' CI423-1 SA-50S CL 2 0.12 0,07 I 35.S' NB.o IS Cire. Weld (100%) SA-1426 8T1762 Linde SO Flux 0.19 0.57 -47.6 167.0 IS Long. Weld (10 27%) SA-812 IPOSI5 Linde SO Flux 0.17 0.52 -47.6 167.0 IS Long. Weld (00 73%) SA-775 IP0661 Linde 80 Flux 0.17 0.64 -47.6 167.0 Intermediate to LS eire. Weld (100%) SA-llOl 71249 Linde SO Flux 0.23 0.59 -47.4 167.6 LS Lon '. Weld (100%) SA-S47 61782 Linde 80 Flux 0.23 0.52 -47.6 167.0

_(I~~~:~: I:c:~;::' *... ;
',;0"1 Point Beach UnIt 2 EvaluatIon

"*."Im," Nozzle Bell ForSlng eNB) 123V352 123V352 SA-50S CL 2 0.11 0.73 40 76.0 Intermediate Shell Forgmg (IS) 123V500 123V500 SA-508 CL 2 0.09 0.70 40 58.0 Lower Shell For~lng (LS) 123WI95 123WI95 SA-50S CL 2 0.05 0.72 40 42.8" NB to IS Cire. Weld (100%) 21935 21935 Linde &0 Flux 0.18 0.70 -56 170.5 Intermediate to LS eire. Weld (100%) SA-1484 72442 Linde 80 Flux 0.26 0.60 -30 ISO.O [] - Controlling values of the adjusted reference temperatures.

  • -Determined from surveillance data.

dRTNDT. OF ART,oF 53 EFPY Fluence, niem' at 53 EFPY Margin .t 53 EFPY Clad/Low ~ T/4 3/4T T/4 3/4T 3/4T Alloy Steel T/4 Interface Location Location Locahon Locatlon LocatJon Location Locabon LocatIon I~f~:;. ::;':;;:: lit\\ti;~;i::; ii: 1;;' 1.:;\\,!,:V0)~~:: : i..... ~:.'::~ ~'**,*,:";:i;1H;J; I,;**... ",**,!"*'.,::**,*;'::: 3.5SE+IS 2.42E+IS J.lIE+IS 47.4 33.7 34.0 33.7 13l.4 117.4 4.90E+19 3.32E+19 L52E+19 104.1 SS.5 56.4 56.4 161.5 145.9 4.55E+19 3.09E+19 IAIE+19 46.4 39.2 56.4 56.4 103.8 96.6 3.58E+lS 2.42E+lS 1.1lE+18 102.9 73.2 65.7 65.7 121.0 91.3 3.19E+19

2. I 6E+19 N/A 201.9 NlA 65.7 N/A

[220.0] NlA 3.19E+19 NlA 9.90E+18 N/A 166.5 NlA 65.7 N/A [184.6] 4,43E+19 3.00E+19 L3SE+19 216.4 182.3 61.7 61.7 230.7 196.6 3.0SE+19 2.07E+19 9.47E+18 199.9 164.5 65.7 65.7 218.1 IS2.7 !:*i;il!:~!t*,:,,,. ~>::i,;il~ji~ ~~y,i.i:*. I::~*:*:).iq:~""i' iJ:~t;~t'I;t:!!':

,,1;'.:;:'\\';>;;;;,
1. *':'i'r*".:,.

J.'?,. '.:ii.: ~":;-,C::i." I,*;*',: !i"(,i"!",i~,, 5.04E+IS 3.4IE+l8 J.56E+1S 53.5 3S.9 34.0 34.0 127.5 112.9 5.05E+19 3,42E+19 L57E+19 76.6 65.2 34.0 34.0 150.6 139.2 4.90E+19 3.32E+19 1.52E+19 56.2 47.S 17.0 17.0 113.2 104.S S.04E+1S 3.4IE+IS 1.56E+IS 120.0 S7.3 65.5 65.5 129.5 96.S 4.65E+l9 3.15E+19 1.44E+19 234,4 198.4 60.S 60.8 r265.21 r229.21 Page 16 of 51

A AREVA NON-PROPRIETARY Chemical Material Description Composition Reactor Vessel Cu Beltline Region Location Type wt% Point Beach Umt 1 Evaluation Nozzle Belt Forging (NB) 122P237 I 22P237 SA-508 Cl. 2 0.11 Intermediate Shell Plate (IS) A9811-1 A9811-1 SA-508CI. 2 0.20 Lower Shell Plate (LS) C1423-1 C1423-1 SA-508 Cl. 2 0.12 NB to IS Cire. Weld (100%) SA-1426 STI762 LiMe SO Flux 0.57 -47.6 IS Long. Weld (10 27%) SA-812 IP0815 Linde SO Flux 0.52 -47.6 IS Long. Weld (00 73%) SA-775 IP0661 Linde SO Flux 0.64 -47.6 lnlermedinte to LS Cire. Weld (IOO%) SA-lIOI 71249 Linde 80 Flux 0.59 -47.4 LS Lon '. Weld (100%) SA-847 61782 Lincle 80 Flux 0.52 -47.6 POlOt Beach Unit 2 Evaluation Nozzle Belt Forging (NB) 123V352 123V352 SA-508Cl.2 0.11 0.73 40 tnlermediate Shell Forging (IS) 123V500 123V500 SA-508CI. 2 0.09 0.70 40 t...ower Shell Forging (LS) 123WI95 123WI95 SA-508CI. 2 0.05 0.72 40 NB to IS Cire. Weld (100%) 21935 21935 Linde 80 Flu" O.IS 0.70 -56 Jntermediale to LS eire. Weld (100%) SA-1484 72442 Linde 80 Flux 0.26 0.60 -30 [] - Controlling values of the adjusted reference temperatures.

  • -Determined from surveillance data.

79.3' 35.S' S.69E+17 NlA 7.98E+l8 1.09E+19 7.51E+18 J.23E+18 1.2SE+19 1.20E+19 170.5 3.95E+lS 2.67E+lS 1.23E+18 180.0 3.67E+19 2.49E+19 1.14E+19 ...... ****.. *.. **t** 32-9019240-000 ART, of at43 EFPY T/4 T/4 3/4T 93.0 65.0 65.7 65.7 liLt 192.4 N/A 65.7 N/A [210.5] N/A 156.5 NlA 65.7 N/A 206.7 171.6 61.7 61.7 221.0 189.7 153.6 65.7 65.7 207.8 48.7 34.9 34.0 34.0 122.7 73.5 61.7 34.0 34.0 147.5 53.9 45.0 17.0 17.0 110.9 109.3 7S.3 65.5 65.5 224.1 186.5 60.8 60.S Page 17 ofSI

A AREVA NON-PROPRIETARY 32-9019240-000 Table 9. Adjusted Reference Temperature Evaluation for the Point Beach Unit 1 and Unit 2 Reactor Vessel Beltline Materials without Up rate, with hafnium removal October 2008 (Unit 1) and April 2008 (Unit 2), through 53 EFPY (Case 7) Chemical Malenal Descnption Composition Reactor Vessel Cu Ni Bl:hline Region Location Point Beach Unit l EvaluatIon Nozzle Belt Forging (NB) 122P237 122P237 SA-50S CI. 2 tnterlnediate Shell Plate (IS) A9Sl1-1 A9S11-1 SA-50S CI. 2 Lower Shell PI.te (LS) .CI423-1 C1423-1 SA-50S CI. 2 NB to IS Cire. Weld (100%) SA-1426 STI762 Linde SO Flux 0.19 IS Lon£. Weld (ID 27%) SA-SIl IPOS15 Linde SO Flux 0.17 IS Long. Weld (OD 73%) SA-775 IP06GI Linde SO Flux 0.17 Intermediate to LS Cire. Weld (100%) SA-I 101 71249 Linde SO Flux. LS Lon' Weld (100% SA-S47 61782 Ljnde 80 Flux POint Bench Unit 2 Evaluation Nozzle Belt ForgUlg (NB) 123V352 123VJ52 SA-50S CI. 2 Intermediate Shl:li Forging (IS) 113V500 123V500 SA-SOB CI. 2 LoweI' Shdl Forging (LS) 123WI95 123WI95 SA-SOB CI. 2 NB to IS Cire. Weld (100%) 21935 21935 Linde so Flux O.IS 0.70 -56 1I1lcrmcdi:l.te to LS Cire. Weld 100%) SA-14S4 72442 Linde 80 Flux 0.26 0.60 ~30 [ ] - Controlling values of the adjusted' reference temperatures.

  • -Detennined from surveillance data.

.6RTNDT* OF ART, DF 53 EFPY Fluence, n/cm' .t53 EFPY Margin .t 53 EFPY T/4 3/4T T/4 3/4T T/4 T/4 3/4T 79.3" 35.S" 170.5 4.S3E+IS 3.17E+IS 1.50E+IS l1S.2 B5.S 65.5 65.5 ISO.O 4.46£+19 3.02E+19 1.3SE+19 232.6 196.2 60.S 60.S Page 18 of 51

A AREVA NON-PROPRIETARY 32-9019240-000 Table 10. Adjusted Reference Temperature Evaluation for the Point Beach Unit 1 and Unit 2 Reactor Vessel Beltline without. Uprate, with hafnium removal October 2008 (Unit 1) and April 2008 (Unit 2), through 43 EFPY (Case 8) Matenal Descnption ChemIcal Composition [] - Controlling values of the adjusted reference temperatures.

  • - Determined from surveillance data.

43 EFPY Fluence. n/cm' ARTNDT* of .143 EFPY ART,'F .143 EFPY Page 19 of 51

A AREVA NON-PROPRIETARY 32-9019240-000 Table 11. Fluence (E> 1.0 MeV) Values for the Point Beach Unit 1 and Unit 2 Vessel Beltline M~terials with Uprate, with hafnium rods, through S3 EFPY (Case 1) 53 EFPY Fluence, nJcm2 Material Inner Wetted ~TLocation %TLocation Beltline Materials Unit Ident. Surface (x=1.625) (x""4.875) Nozzle Belt Forging (NB) 1 I 22P237 2.84E+18 1.92E+l8 8.81E+17 ~djate Shell (IS) 1 A981l-1 4.868+19 3.29E+19 1.51E+19 Shell (LS) 1 C1423-1 3.70E+19 2.51E+19 1.I5E+19 SCire. Weld (100%) 1 SA-1426 2.84E+18 1.92E+18 8.81E+17 IS Long. Weld (ID 27%) 1 SA*812 3.10E+19 2.10E+19 NfA IS Long. Weld (OD 73%) 1 SA-775 3.10E+l9 NlA 9.62E+18 Intermediate to LS Cire. Weld (100%) 1 SA*lIOl 3.7IE+19 2.5 1E+1 9 1.15E+19 l:§ L~I!..ll:..We~1!QQ.~L._. ___.* _ I SA*847 2.6~~,!".!2..._ _.J2E!l.+/-'12...._. 8.07E+l8 -128E+is"-' Nozzle Belt Forging (NB) 2 l23V352 4.12E+18 2.79E+18 ~Shell Forging (IS) 2 123V500

4. 66E+19 3.16E+19 1.44E+19 Forging (LS) 2 123W195 4.15E+19 2.81E+19 1.29E+19 NB to IS Cire. Weld (100%)

2 21935 4.12E+18 2.79E+18 1.28E+18 ~ Intermediate to LS Cire. Weld (LOO%) 2 SA-1484 3.75E+19 2.54E+19 1.16E+19 Table 12. Fluence (E> 1.0 MeV) Values for the Point Beach Unit 1 and Unit 2 Vessel Beltline Materials with Uprate, with hafnium rods, through 43 EFPY (Case 2) 43 EFPY Fluence, n1cm2 Material Inner Wetted !!.IT Location %TLocation Beltline Materials Unit Ident. Surface (x=1.625) (x~.875) Nozzle Belt Forging (NB) 1 122P237 2.388+18 1.61E+18 7.39E+17 Intermediate Shell (IS) I A9811-1 3.95E+19 2.67E+19 1.23E+19 Lower Shell (LS). 1 C1423-1 3.1lE+19 2.11E+19 9.658+18 NB to IS eire. Weld (100%) I SA-1426 238E+18 L6lE+18 7.39E+17 IS Long. Weld (ID 27%) 1 SA.BI2 2.52E+19 1.71E+19 NfA IS Long. Weld (OD 73%) 1 SA-775 2.52E+19 N/A 7.82E+18 Intermediate to LS Cire. Weld (l00%) 1 SA-HOI 3.10E+19 2.10E+19 9.62E+18 -:I&l:~~ll.:.W~ld (100~L..** _____. __. I SA-841 2.15E+19 1.46E+19 6.61E+18 Nozzle Belt Forging (NB) "'2---' 'T:f3V352 -_. '--'"i42E+lf"'" -----2.3z"E+I!C--*_* "***r66E+lS"**** Intermediate Shell Forging (IS) 2 123V500 3.81E+19 2.58E+19 1.18E+19 Lower Shell Forging (LS) 2 123W195 3.45E+19 2.34E+19 1.07E+19 NB to IS Cire. Weld (100%) 2 21935 3.42E+18 2.32E+18 1.06E+18 Intermediate to LS Cire. Weld (100%) 2 SA*1484 3.14E+19 2.13E+19 9.75E+18 Page 20 ofSI

A AREVA NON-PROPRIETARY 32-9019240-000 Table 13. Fluence (E> 1.0 MeV) Values for the Point Beach Unit 1 and Unit 2 Vessel Beltline Materials without Uprate, with hafnium rods, through S3 EFPY (Case 3) 53 EFPY Fluence, nlcm2 Material Inner Wetted Y4T Location %TLocation Beltline Materials Unit Ident. Surface (x=1.625) (x=4.875) Nozzle Belt Forging iNB) 1 122P237 2.75E+IS I.S6E+IS 8.54E+17 Intennediate Shell (IS) 1 A981l-1 4.69E+19 3.ISE+19 1.46E+19 Lower Shell (LS) I C1423-1 3.59E+19 2.43E+19 l.IIE+19 NB to IS Cire. Weld (100%) 1 SA-1426 2.75E+18 1.86E+18 S.54E+17 IS Long. Weld (ID 27%) I SA-SI2 2.99E+19 2.02E+19 N/A IS Long. Weld (00 73%) 1 SA-775 2.99E+19 N/A 9.2SE+18 Intennediate to LS Cire. Weld (100%) 1 SA-I 101 3.59E+19 2.43E+19 l.l1E+19 ~~~~eft~g~~~ifJB)------------ I SA-S47 f-____ 2.5IE+ 19 ____ . 1.70E+19 7.79E+18 2 -Ti3V352-- 3.99E+18

2.70E+18-**--- ----1:24E+18--

Intennediate Shell Forging (IS) 2 123V500 4.49E+19 3,04E+19 1.39E+19 Lower Shell Forging (LS) 2 123W195 4.02E+19 2.72E+19 1.25E+19 NB to IS Cire. Weld (100%) 2 21935 3.99E+18 2.70E+18 1.24E+18 Intermediate to LS Cire. Weld (100%) 2 SA-1484 3.64E+19 2.46E+19 1.13E+19 Table 14. Fluence (E>1.0 MeV) Values for the Point Beach Unit 1 and Unit 2 Vessel Beltline Materials without Uprate, with hafnium rods, through 43 EFPY (Case 4) 43 EFPY Fluence, nlcm2 Material Inner Wetted Y4T Location %T Location Beltline Materials Unit Ident. Surface (x=1.625) (x=4.875) Nozzle Belt Forging iNB) 1 1 22P237 2.33E+18 1.58E+18 7.23E+17 lntennediate Shell (IS) 1 A981l-1 3.85E+19 2.6IE+19 1.20E+19 Lower Shell (LS) 1 C1423-1 3.05E+19 2.07E+19 9.47E+18 NB to IS Cire. Weld (100%) I SA-1426 2.33E+18 1.58E+18 7.23E+17 IS Long. Weld (ID 27%) 1 SA-812 2.46E+19 1.67E+19 N/A IS Long. Weld (00 73%) I SA-775 2.46E+19 N/A 7.64E+18 Intennediate to LS Cire. Weld (100%) I SA-I 101 3.03E+19 2.05E+19 9.40E+18 LS Long. Weld (100%) 1 __ ~~..:~~_L _____ 2.IOE+19 f--___ LQE+ 1 ? ____ 1--_~)2E+ 1 ~ _____ NoZZle-Belt Forgiilg(NBj-*---------* 2 123V352

3.34E+i8---

2.26E+18 1.04E+18 Intennediate Shell Forging (IS) 2 123V500 3.. 7IE+19 2.5IE+19 1.I5E+19 Lower Shell Forging (LS) 2 123W195 3.37E+19 2.2SE+19 1.05E+19 NB to IS Cire. Weld (100%) 2 21935 3.34E+18 2.26E+IS 1.04E+18 Intcnnediate to LS Cire. Weld (100%) 2 SA-1484 3.0SE+19 2.09E+19 9.56E+1S Page 21 of 51

A AREVA NON-PROPRIETARY 32-9019240-000 Table 15. Fluence (E>1.0 MeV) Values for the Point Beach Unit 1 and Unit 2 Vessel Beltline Materials with Uprate, with hafnium rod removal October 2008 (Unit 1) and April 2008 (Unit 2), through 53 EFPY (Case 5) 53 EFPY Fluence, nlcm2 Material hmerWetted

4T Location

%TLocation Beltline Matelials Unit Ident. Surface (x==1.625) (x=4.875) =B). 1 122P237 3.58E+18 2.42E+18 .1.11E+18 Intermediate Shell (IS I A9811*1 4.90E+19 3.32E+19 1.52E+19 Lower Shell (LS) 1 C1423-1 4.55E+19 3.08E+19 1.41E+19 B' Wcld (100%) I SA-1426 3.58E+18 2.42E+18 l.llE+18 ong. Weld (10 27%) I SA*gI2

3. 19E+19 2.16£+19 N/A ong. Weld (00 73%)

1 SA*775 3.19E+19 N/A 9.90E+18 ~~~Q~i~~~-~~~~--- 1 SA*lIO! 4.43E+l9 3.00E+19 1.38E+19 1 SA-847 3.05E+19 2.07E+19 9.47E+18 -TnV352 ----- ---5~04E+18---- c----:---:-:-------------------- Nozzle Belt Forging ) 2 3.41E+18 1.56E+18 Intermediate Shell Forging (IS) 2 123V500 5.05E+19 3.42E+19 1.57E+19 Lower Shell Forging (LS) 2 123W195 4.90E+19 3.32E+19 1.52E+19 NB to IS eire. Weld (100%) 2 21935 5.04E+18 3.41E+18 1.56E+18 Intermediate to 18 Cire. Weld (100%) 2 SA-1484 4.65E+19 3.15E+19 1.44E+19 Table 16. Fluenee (E>1.0 MeV) Values for the Point Beach Unit 1 and Unit 2 Vessel BeltIine Materials with Up rate, with hafnium rod removal Octob~r 2008 (Unit 1) and April 2008 (Unit 2), through 43 EFPY (Case 6) 43 EFPY Fluence, n/cm2 Material Inner Wetted

4T Location

%TLocation Beltline Materials Unit Ident. Surface (x=1.625) (x=4_875) B) 1 122P237 2.80E+!8 1.90E+18 8.69E+l7 Intermediate Shell (IS) 1 A9811-1 3.97E+l9 2.69E+19 1.23E+!9 Lower Shell (18) 1 C1423-1 3.59E+19 2.43E+19 UIE+!9 Cire. Weld (100%) I SA-[426 2.80E+18 1.90E+18 8.69E+17 Weld (ID 27%) 1 SA*812 2.57E+19 1.74E+19 N/A IS Long. Weld (OD 73%) 1 SA-775 2.57E+19 N/A 7.98E+18 Intermediate to LS eire. Weld (100%) 1 SA-lIOl 3.51E+19 2.38E+19 1.09E+19 ~_!:~lL?!_~!~1 0D.'r?L ___ I _§'~:H? _______ 2.42E+19 1.64E+19 7.5IE+18

T95E+18"-- r---;---------- -EZ3E+iil----

Nozzle Belt Forging (NB) I 23V352 2.67E+18 Intermediate Shell Forging (IS) 2 I 23V500 4.04E+19 2.74E+!9 1.25E+19 Lower Shell Forging (LS) 2 123WI95 3.88E+!9 2.63E+l9 1.20E+l9 NB to IS Cire. Weld (100%) 2 21935 3.95E+18 2.67E+18 1.23E+18 Intermediate to LS Cire. Weld (100%) 2 SA-I 484 3.67E+19 2.49E+19 1.14E+l9 Page 22 of 51

A AREVA NON-PROPRIETARY 32-9019240-000 Table 17. Fluence (E>1.0 MeV) Values for the Point Beach Unit 1 and Unit 2 Vessel Beltline Materials without Uprate, with hafnium rod removal October 2008 (Unit 1) and April 2008 (Unit 2), through 53 EFPY (Case 7) 53 EFPY Fluence, nlcm 2 Material Inner Wetted If4T Location %T Location Beltline Materials Unit Ident. Surface (x=1.625) (x=4.875) Nozzle Belt Forging (NB) 1 122P237 3.43E+18 2.32E+1.8 1.07E+18 Intermediate Shell (IS) 1 A98lt-l 4.72E+19 3.20E+19 1.47E+19 Lower Shell (LS) 1 C1423-1 4.36E+19 2.95E+19 1.35E+19 NB to IS Cire. Weld (100%) 1 SA-1426 3.43E+18 2.32E+18 1.07E+18 IS Long. Weld (ID 27%) I SA-812 3.07E+19 2.08E+19 N/A IS Long. Weld (00 73%) I SA-775 3.07E+19 N/A 9.53E+18 Intermediate to LS Cire. Weld (100%) 1 SA-ltOI 4.25E+19 2.88E+19 1.32E+19 -~~~~~'e~~l!r~~~i~Bj--------------------J------ SA-847 2.93E+19 1.98E+19 9.09E+18 2 --123V352 ---- ----4]'3"E+l8---- -327£+18--

i:sOE+~-

Intermediate Shell Forging (IS) 2 123V500 4.85E+19 3.28E+19 1.51£+19 Lower Shell Forging (LS) 2 123W195 4.71£+19 3.19E+19 1.46E+19 NB to IS Cire. Weld (100%) 2 21935 4.83E+18 3.27E+18

1. 5 OE+ I 8 Intermediate to LS Cire. Weld (100%)

2 SA-1484 4.46E+19 3.02E+19 1.38E+19 Table 18. Fluence (E>1.0 MeV) Values for the Point Beach Unit 1 and Unit 2 Vessel BeltIine Materials without Uprate, with hafnium rod removal October 2008 (Unit 1) and April 2008 (Unit 2), through 43 EFPY (Case 8) 43 EFPY Fluence, nlcm2 Material Inner Wetted If4T Location %TLocation Beltline Materials Unit Ident. Surface (x=1.625) (x=4.875) Nozzle Belt Forging (NB) 1 122P237 2.72E+18 1.84E+18 8.44E+17 Intermediate Shell (IS) I A9811-1 3.87E+19 2.62E+19 1.20E+19 Lower Shell (LS) 1 C1423-1 3.49E+19 2.36E+19 1.08E+19 NB to IS Cire. Weld (100%) 1 SA-1426 2.72E+18 1.84E+18 8.44E+17 IS Long. Weld (ID 27%) 1 SA-812 2.50E+19 1.69E+19 N/A [S Long. Weld (0073%) 1 SA-775 2.50E+19 N/A 7.76E+18 Intermediate to LS Cire. Weld (100%) I SA-llOI 3.41E+19 2.3IE+19 1.06E+19 . LS Long. Weld (100%) 1 -§.~:!~..?----- 2.34E+19 1.588+19 7.26E+18 -i~iozzle-B-eitForging (NBj------------ ---259E+ 18-- -----iT9E+-is------ 2 123V352 3.83E+18 Intermediate Shell Forging (IS) 2 123VSOO 3.92E+19 2.658+19 1.22E+19 Lower Shell Forging (LS) 2 123WI95 3.77E+19 2.55E+19 1.17E+19 NB to IS Cire. Weld (100%) 2 21935 3.83E+18 2.598+18 1.19E+18 Intermediate to LS Cire. Weld (100%) 2 SA-1484 3.55E+19 2.40E+19 1.I0E+19 Page 23 of 51

A AREVA NON-PROPRIETARY 32-9019240-000 5.0 Adjusted Reference Temperature Where No Surveillance Data Is Available The following information is required for determination of the ART in accordance with Regulatory Guide 1.99, Revision 2. 5.1 Initial RTNDT The initial RTNDT is the reference temperature for the reactor vessel be1tline material in the unirradiated condition, evaluated in accordance with Paragraph NB-2331 of Section III of the ASME Boiler and Pressure Vessel Code[4] or Code Case N-629.[51 Table 19 lists the initial RTNDT values for the Point Beach Unit 1 and Unit 2 reactor vessel be1t1ine materials. [5,6] Table 19. Initial RTNDT Values for the Point Beach Unit 1 and Unit 2 Reactor Vessel Beltline Materials Material Initial RTNDT Beltline Materials Unit Ident. COF) Reference Nozzle Belt Forging (NB) 1 122P237 50 6 Intermediate Shell (IS) 1 A9811-1 1 6 Lower Shell (LS) 1 C1423-1 I 6 NB to IS Cire. Weld (100%) I SA-1426 -47.6 5 IS Long. Weld (ID 27%) 1 SA-812 -47.6 5 IS Long. Weld (OD 73%) 1 SA-775 -47.6 5 Intermediate to LS Cire. Weld (100%) I SA-1101 -47.4 5 _!:§}:~()~g. Wel~J!g_Q~L ______________ 1 SA-847 -47.6 5

.--- -----------.---- --------------------- -----*6-Nozzle Belt Forging (NB) 2 123V352 40 Intermediate Shell Forging (IS) 2 123V500 40 6

Lower Shell Forging (LS) 2 123W195 40 6 NB to IS Cire. Weld (100%) 2 21935 -56 6 Intermediate to LS Cire. Weld (100%) 2 SA-1484 -30 5 5.2 ~RTNDT ~R T NOT is the mean value of the adjustment in reference temperature caused by irradiation and is calculated as follows: where CF ff = Chemistry Factor = fluence factor ART NDT = (CF)

  • tff)

Page 24 of 51 (2)

A AREVA 5.2.1 Chemistry Factor The chemistry factor (CF) is determined from the copper and nickel content for each reactor vessel beltline region material. Using the copper and nickel contents for the Point Beach Unit 1 and Unit 2 reactor vessel beltline materials, [6J the CF is determined from Table 1 (for weld metals) and Table 2 (for base metals) in Regulatory Guide 1.99, Revision 2. Linear interpolation is permitted. When determining the CF, the "weight percent copper" and "weight percent nickel" are best estimate values for the material, which will normally be the mean of the measured values for the material. If RT NDT values from BAW-2308 are used, the CF cannot be less than 167.0. Using Tables 1 and 2 in Regulatory Guide 1.99, Revision 2, the CF values for the Point Beach Unit 1 and Unit 2 reactor vessel beltline materials are listed in Table 20. Table 20. Regulatory Guide 1.99, Revision 2 Chemistry Factors for the Point Beach Unit 1 and Unit 2 Reactor Vessel BeItline Materials Material Cu Ni Chemistry Beltline Materials Unit Ident. wt%[6] wt%[6] Factor Nozzle Belt Forging (NB) 1 122P237 0.11 0.82 77.0 Intermediate Shell (IS) 1 A9811-1 0.20 0.06 79.3* Lower Shell (LS) 1 C1423-1 0.12 0.07 35.8* NB to IS Circ. Weld (100%) 1 SA-1426 0.19 0.57 167.0 IS Long. Weld (JD 27%) 1 SA-812 0.17 0.52 167.0 IS Long. Weld (OD 73%) 1 SA-775 .0.17 0.64 167.0 Intermediate to LS Circ. Weld (100%) 1 SA-1101 0.23 0.59 167.6 __ ~§..b~!!g~W~ld <!Q9~) ____________________ 1 SA-847 0.23 0.52 167.0 Nozzle Belt Forging (NB) 2 123V352 O.ll 0.73 76.0 Intermediate Shell Forging (IS) 2 123V500 0.09 0.70 58.0 Lower Shell Forging (LS) 2 123W195 0.05 0.72 42.8* NB to IS Circ. Weld (100%) 2 21935 0.18 0.70 170.5 Intermediate to LS Circ. Weld (100%) 2 SA-1484 0.26 0.60 180.0

  • -determined from surveillance data, see Section 6.0 5.2.2 FIuence Factor In accordance with Regulatory Guide 1.99, Revision 2, the fluence factor (ff) is determined as follows:

.ff = j(O.28-0.1010g f) Tablse 21 - 28 lists the fluence factors for the V4T and %T locations of the Point Beach Unit 1 and Unit 2 reactor vessel beltline materials. Page 25 of 51 (3)

A AREVA NON-PROPRIETARY 32-9019240~OOO Table 21. Fluence Factors for the %T and %T Locations of the Point Beach Unit 1 and Unit 2 Reactor Vessel Beltline Materials with Uprate, with hafnium rods, through 53 EFPY (Case 1) V.T Location %TLocation

Fluence, Fluence, Material nlem2 Fluence nlcm2 Fluence Beltline Material Unit Ident.

(x 1019) Factor (xl 019) Factor Nozzle Belt Forging (NB) 1 122P237 1.92E+l8 0.560 8.81E+17 0.392 Intelmediate Shell (IS) 1 A9811-1 3.29E+l9 1.312 1.51E+19 1.114 Lower Shell (LS) I C1423-1 2.51E+19 1.247 1.15E+19 1.039 NB to IS Cire. Weld (100%) 1 SA-1426 1.92E+1& 0.560 8.81E+17 0.392 IS Long. Weld (ID 27%) 1 SA-gI2 2.10E+19 1.202 N/A N/A IS Long. Weld (OD 73%) I SA-775 NlA N/A 9.62E+18 0.989 InteIIDediate to LS Cire. Weld (100%) 1 SA-lIOl 2.51E+19 1.247 1.I5E+19 1.039 1 SA-847 1.76E+19 1.155 8.07£+18 0.940 LS Long. Weld (100%) -NozzieBeitForghlgCNBj""------------ ------- --.. --------- -----~------------- --------------- 2 123V352 2.79E+18 0.652 1.28E+18 0.468 Intermediate ShelI Forging (rS) 2 123V500 3.16£+19 1.303 1.45E+19 1.102 Lower Shell Forging (LS) 2 123W195 2.81E+19 1.275 1.29E+19 1.070 NB to IS Cire. Weld (100%) 2 21935 2.79E+18 0.652 1.28E+18 0.468 InteIIDediate to LS Cire. Weld (100%2. 2 SA-1484 254E+19 1250 116E+19 1042 Table 22. Fluence Factors for the %T and %T Locations ofthe Point Beach Unit 1 and Unit 2 Reactor Vessel Beltline Materials with Uprate, with hafnium rods, through 43 EFPY (Case 2) V.T Location %TLocation

FJuence, Fluence, Material nlcm2 Fluenee nlcm2 Fluence Beltline Material Unit Ident.

(xlO I9) Factor (x 1019) Factor Nozzle Belt Forging (NB) 1 122P237 1.6 lE+l 8 0.519 7.39E+l7 0.359 : IntcIIDediate Shell (IS) 1 A9811-I 2.67E+19 1.263 1.23E+19 1.057 Lower Shell (LS) 1 C1423-1 2.l1E+19 1.203 9.65E+18 0.990 NB to IS Cire. Weld (100%) 1 SA-1426 1.6 lE+ 18 0.519 7.39E+17 0.359 IS Long. Weld (ID 27%) 1 SA-gI2 1.71E+19 1.147 N/A N/A IS Long. Weld (OD 73%) 1 SA-775 NfA N/A 7.82£+18 0.931 Intermediate to LS Cire. Weld (100%) 1 SA-llOI 2.10E+19 (202 9.62E+18 0.989 J&~.E~g. 'Y.~!!J.. Q_QOO/?l _________ 1 SA-847 1.46E+19 1.104 6.67E+l8 0.887

~-~-~-- --------~ ---_.----- ~-----~-...... --.. ---

Nozzle Belt Forging (NB) 2 123V352 2.32E+18 0.605 1.06E+18 0.429 Intermediate Shell Forging (IS) 2 123V500 2.58E+19 1.254 1.18E+19 1.047 Lower Shell Forging (LS) 2 123Wl95 2.34E+19 1.229 1.07E+19 1.019 NB to IS Cire. Weld (100%) 2 21935 2.32E+18 0.605 1.06£+18 0.429 InteIIDediate to LS Cire. Weld (100%) 2 SA-1484 2.13E+19 1.205 9.75E+18 0.993 Page 26 of 51

A AR EVA NON-PROPRIETARY 32-9019240-000 Table 23. Fluence Factors for the ~T and %T Locations ofthe Point Beach Unit 1 and Unit 2 Reactor Vessel Beltline Materials without Uprate, with hafnium rods, through 53 EFPY (Case 3) Y4T Location %TLocation

Fluence, Fluence, Material nlcm2 Fluence nlcm2 Fluence Beltline Material Unit Ident_

(XI019) Factor (xlOI9) Factor Nozzle Belt Forging (NB) I I 22P237 1.86E+18 0.552 8.54E+17 0.386 Intermediate Shell (IS) 1 A9811-1 3.18E+19 1.304 1.46E+19 1.104 Lower Shell (LS) I CI423-1 2.43E+19 1.239 1.11E+19 1.030 NB to IS Circ. Weld (100%) 1 SA-I426 1.86E+18 0.552 8.54E+17 0.386 IS Long_ Weld (ID 27%) I SA-8I2 2.02E+19 1.192 N/A N/A IS Long. Weld (OD 73%) I SA-775 N/A N/A 9.28E+18 0.979 Intermediate to LS Circ_ Weld (100%) I SA-lIOI 2.43E+19 1.239 1.11E+19 1.030 LS Long. Weld (100%) I SA-847 1.70E+19 1.146 7.79E+18 0.930 NozZie-Beit Forgini(NBj"---------------- ------- ----------- -.. ------------------ ------- 2 123V352 2.70E+18 0.643 1.24E+18 0.461 Intermediate Shell Forging (IS) 2 123V500 3.04E+19 1.294 1.394E+19 1.092 Lower Shell Forging (LS) 2 123W195 2.72E+19 1.267 1.25E+19 1.062 NB to IS Circ. Weld (100%) 2 21935 2.70E+18 0.643 1.24E+18 0.461 Intermediate to LS Circ_ Weld (100%) 2 SA-1484 2.46E+19 1243 1.13E+19 1034 Table 24. Fluence Factors for the ~T and %T Locations of the Point Beach Unit 1 and Unit 2 Reactor Vessel Beltline Materials without Uprate, with hafnium rods, through 43 EFPY (Case 4) Y4T Location %TLocation

Fluence, Fluence, Material nlcm2 Fluence n/cm2 Fluenee Belt1ine Material Unit Ident (XIOl~

Factor (XI019) Factor Nozzle Belt Forging (NB) I 122P237 1.58E+18 0.514 7.23E+17 0.355 Intermediate Shell (IS) I A9811-1 2.61E+19 1.257 1.20E+19 1.050 Lower Shell (LS) I C1423-1 2.07E+19 1.197 9.47E+18 0.985 NB to IS Circ_ Weld (I 00%) I SA-I 426 1.58E+18 0.514 7.23E+17 0.355 IS Long. Weld (ID 27%) 1 SA-8I2 1.67E+19 1 141 N/A N/A IS Long. Weld (OD 73%) 1 SA-775 NlA N/A 7.64E+18 0.924 Intermediate to LS Circ. Weld (100%) I SA-IIOI 2.05E+19 1.196 9.40E+18 0.983 LS Long. Weld (100%) I SA-847 1.42E+19 1.098 6.52E+18 0.880 -NozZie-Beit-Forging (NBj""--------------------- 1---------.-------.. ---- -.. _------------- ---------------- ------------- ---------- 2 123V352 2.26E+18 0.599 1.04E+18 0.424. Intermediate Shell Forging (IS) 2 123V500 2.5IE+19 1.247 I.l5E+19 1039 Lower Shell Forging (LS) 2 l23W195 2.28E+19 1.223 1.05E+19 1013 NB to IS Cire. Weld (100%) 2 21935 2.26E+18 0.599 104E+18 0.424 Intermediate to LS Cire. Weld (100%) 2 SA-I484 209E+19 1.200 9.56E+18 0.987 Page 27 of 51

A. AREVA NON*PROPRIETARY 32-9019240-000 Table 25. Fluence Factors for the % T and % T Locations of the Point Beach Unit 1 and Unit 2 Reactor Vessel Beltline Materials with Uprate, with hafniu~ rod removal October 2008 (Unit 1) and April 2008 (Unit 2), through 53 EFPY (Case 5) 'l4T Location %TLocation

Fluence, Fluence, Matetial nlcm2 Fluenee n/em2 Fluenee Beltline Materia}

Unit Ident. (Xl01~ Factor (x 1019) Factor Nozzle Belt Forging (NB) 1 122P237 2,42E+18 0.616 I.IIE+l8 0.438 Intermediate Shell (IS) 1 A9811-1 3.32E+19 1.314 1.52E+l9 UI6 Lower Shell (LS) 1 C1423-1 3.08E+19 1.297 1.41E+l9 1.096 NB to IS Cire. Weld (100%) 1 SA-1426 2.42E+18 0.616 l.l1E+18 0.438 IS Long. Weld (ID 27%) 1 SA-gI2 2.16E+19 1.209 N/A N/A IS Long. Weld (aD 73%) 1 SA-775 N/A N/A 9.90E+18 0.997 Intermediate to LS Circ. Weld (100%) 1 SA-lIOl 3.00E+19 1.291 l.38E+19 1.088 J:~_~~!!g: W ~!~Sl OO~L_. ______....... 1 SA-847 2.07E+19 1.197 9,47E+18 0.985 2 __ ~.~ ______ w_~

~----.----.---- t---:--~.. ----- ---~---

Nozzle Belt Forging (NB) 123V352 3.41E+18 0.704 1.56E+18 0.512 Intermediate Shell Forging (IS) 2 123V500 3.42E+19 1.321 I.57E+19 1.124 Lower Shell Forging (LS) 2 '123W195 3.32B+19 1.314

1. 52E+19 1.116 NB to IS Cire. Weld (100%)

2 21935 3.4-1E+18 0.704 1.56E+l8 0.512 Intermediate to LS Circ. Weld (100%) 2 SA-1484 3 15E+19 1302 1.44E+19 1.102 Table 26. Fluence Factors for the ~T and %T Locations of the Point Beach Unit 1 and Unit 2 Reactor Vessel Beltline Materials with Up rate, with hafnium rod removal October 2008 (Unit 1) and April 2008 (Unit 2), through 43 EFPY (Case 6) 'l4T Location %TLocation

Fluence, Fluence, Material n/cm2 Fluenee n/em2 Fluenee Beltline Material Unit Ident.

(X1019) Factor (XlO I9) Factor Nozzle Belt Forging (NB) 1 122P237 1.90E+18 0.557 8.69E+17 0.389 Intermediate Shell (IS) I A9811-1 2.69E+19 1.264 l.23E+19 1.05& Lower Shell (LS) 1 C1423~1 2.43E+19 1.239 1J1E+19 1.030 NB to IS Cire. Weld (100%) 1 SA~1426 1.90E+l8 0.557 8.69E+17 0.389 IS Long. Weld (ID 27%) 1 SA-812 1.74E+19 1.152 N/A NlA IS Long. Weld (aD 73%1. 1 SA-775 N/A N/A 7.98E+18 0.937 Intermediate to LS Cire. Weld (100%) 1 SA-I 101 2.38E+19 1.233 1.09E+l9 1.024 J.$ !:-~!%... W ~9.{~.Qg~. ___________,, __ 1 SA-847 1.64E+19 1.136 7.5lE+18 0.920 "123-Y352-- ...... ~~-~--.... -.. ~- ~-----~-- --~-----.--.. _--------_.... Nozzle Belt Forging (NB) 2 2.67E+18 0.641 l.23E+l8 0.459 Intermediate Shell Forging (IS) 2 123V500

2. 74E+l 9 1.268 1.25E+19 1.063 Lower Shell Forging (LS) 2 123W195 2.63E+19 1.259 1.20E+19 1.D52 NB to IS Cire. Weld (100%)

2 21935 2.67E+18 0.641 1.23E+18 @tJ Intermediate to LS Cire. Weld (100%) 2 SA-1484 2.49E+19 L245 1.14E+19 Page2S of 51

A AREVA NON-PROPRIETARY 32-9019240-000 Table 27. Fluence Factors for the ¥.sT and %T Locations of the Point Beach Unit 1 and Unit 2 Reactor Vessel Beltline Materials withont Uprate, with hafnium rod removal October 2008 (Unit 1) a:nd April 2008 (Unit 2), through 53 EFPY (Case 7) V4T Location %T Location

Fluence, Fluence, Material nlcm2 Fluence nlcm2 Fluence BeltHne Material Unit Ident.

(xlOI9) Factor (x 10 I!) Factor Nozzle Belt Forging (NB) I 122P237 2.32E+18 0.606 1.07E+l8 0.430 Intermediate Shell (IS) 1 A981H 3.90E+19 1.306 1.47E+l9 1.106 Lower Shell (LS) 1 C1423-1 2.95E+l9 1.287 1.35E+19 1.084 NB to IS Cire. Weld (100%) 1 SA-1426 2.32E+18 0.606 1.07E+18 0.430 IS Long. Weld (ID 27%) 1 SA-BI2 2.08E+19 1.199 N/A N/A IS Long. Weld (OD 73%) 1 SA-775 N/A N/A 9.53E+18 0.986 Intermediate to LS Cire. Weld (100%) 1 SA-IIOl 2.88E+19 1.281 1.32E+19 1.077 f-:~~-~~!!g~-W eld_Q22~1 __________ i SA-847 1.98E+19 1.187 9.09E+]8 0.973 -.----~.... ------ .~-------.... ~.----~------- --_ Nozzle Belt Forging (NB) 2 123V352 3.27E+18 0.693 l.50E+) 8 0.503 Intermediate Shell Forging (IS) 2 123V500 3.28E+19 1.312 1.5IE+19 1.113 Lower Shell Forging (LS) 2 123W195 3.19E+19 1.305 1.46E+19 UOS NB to IS eirc. Weld (100%) 2 21935 3.27E+18 0.693 1.50E+18 0.503 Intermediate to LS Cire. Weld (100%) 2 SA-1484 3.02E+\\9 1292 138E+l9 1090 Table 28. Fluence Factors for the ¥.sT and %T Locations of the Point Beach Unit 1 and Unit 2 Reactor Vessel Beltline Materials without Uprate, with hafnium rod removal OCtober 2008 (Unit 1) and Apri12008 (Unit 2), through 43 EFPY (Case 8) ~T Location %TLocation

Fluence, Fluence, Material nlcm2 Fluenee nlcm2 Fluence Beltline Material Unit Ident.

(x 1019) Factor (xl 019) Factor Nozzle Belt Forgin~ (NB) 1 122P237 1.84E+18 0.550 8.44E+17 0.384 Intermediate Shell (IS) 1 A9811-1 2.62E+19 1258 120E+19 1.051 Lower Shell (LS) 1 C1423-1 2.36E+19 1.232 1.08£+19 1.022 NB to IS eire. Weld (100%) 1 SA-1426 1.84E+18 0.550 8.44E+17 0.384 ti Long, Wold (ID 27%) 1 SA-gI2 1.69E+19 1.145 N/A N/A Long. Weld (OD 73%) 1 SA-775 N/A N/A 7.76E+18 0.929 Intermediate to LS Cire. Weld (100%) 1 SA-IIOl 2.31E+19 1.226 1.06E+19 1.016 '!&~£!!&_:!y~ld (1 QQ~L ___________. __ } SA-847 1.58E+l9 1127 726E+}8 0.910

~~---

~""-~----~-- -_.. _ .. ----1----------------0:4"52-- Nozzle Belt Forging (NB) 2 123V352 2.59E+18 0.633 1.19E+18 Intermediate Shell Forging (IS) 2 123V500 2.65E+19 1.261 1.22E+19 l.055 Lower Shell Forging (LS) 2 123W195 2.55E+19 1.251 1.I7E+l9 1.044 NB to IS Cire. Weld (100%) 2 21935 2.59E+18 0.633 1.19E+18 0.452 Intermediate to LS eirc. Weld (100%) 2 SA-1484 2.40E+19 1.236 1.10E+19 1.027 Page 29 of 51

A AREVA NON-PROPRIETARY 32-9019240-000 5.2.3 ARTNDT Calculation The ~RT NDT values for the Point Beach Unit 1 and Unit 2 reactor vessel beltline materials at the V4T and %T wall locations are calculated by multiplying the chemistry factors and fluence factors. The ~RT NDT values for the Point Beach Unit 1 and Unit 2 reactor vessel beltline materials are presented in Tables 29 - 36. Table 29. ARTNDT Values for the Y-.T and %T Wall Locations ofthe Point Beach Unit 1 and Unit 2 Beltline Materials with Uprate, with hafnium rods, through 53 EFPY (Case 1) Material V4T Location 3i4T Location Beltline Materials Unit Ident. CF ff ~RTNDr CF ff ~RTNDr Nozzle Belt Forging (NB) 1 122P237 77.0 0.560 43.1 77.0 0.392 30.2 Intermediate Shell (IS) 1 A981l-1 79.3* 1.312 104.0 79.3* 1.114 88.3 Lower Shell (LS) 1 C1423-1 35.8* 1.247 44.6 35.8* 1.039 37.2 NB to IS Cire. Weld (100%) 1 SA-1426 167.0 0.560 93.5 167.0 0.392 65.5 IS Long. Weld (JD 27%) 1 SA-812 167.0 1.202 200.7 167.0 N/A N/A IS Long. Weld (OD 73%) 1 SA-775 167.0 N/A N/A 167.0 0.989 165.2 Intermediate to LS Cire. Weld (100%) 1 SA-llOl 167.6 1.247 209.1 167.6 1.039 174.1 LS Long. Weld (100%) 1 SA-847 167.0 1.155 192.9 167.0 0.940 157.0 Nozzle-Belt'Forging -(NBy-------------- ~--------- 1--0--*--- ------------- 2 123V352 76.0 0.652 49.6 76.0 0.468 35.6 Intermediate Shell Forging (IS) 2 123V500 58.0 1.303 75.6 58.0 1.102 63.9 Lower Shell Forging (LS) 2 123W195 42.8* 1.275 '54.6 42.8* 1.070 45.8 NB to IS Cire. Weld (100%) 2 21935 170.5 0.652 111.2 170.5 0.468 79.8 Intermediate to LS Cire. Weld (100%) 2 SA-1484 180.0 1.250 225.0 180.0 1.042 187.6

  • -determmed from surveillance data Table 30. ARTNDT Values for the Y-.T and %T Wall Locations of the Point Beach Unit 1 and Unit 2 Beltline Materials with Uprate, with hafnium rods, through 43 EFPY (Case 2)

Material V4T Location 'Y4T Location Beltline Materials Unit Ident. CF ff ~RTNDI CF ff ~TNDI Nozzle Belt Forging (NB) 1 122P237 77.0 0.519 40.0 77.0 0.359 27.6 Intermediate Shell (IS) 1 A9811-1 79.3* 1.263 100.2 79.3* 1.057 83.8 Lower Shell (LS) 1 C1423-1 35.8* 1.203 43.1 35.8* 0.990 35.4 NB to IS Cire. Weld (100%) 1 SA-1426 167.0 0.519 86.7 167.0 0.359 60.0 IS Long. Weld (JD 27%) 1 SA-812 167.0 1.147 191.6 167.0 N/A N/A IS Long. Weld (OD 73%) 1 SA-775 167.0 N/A NIA 167.0 0.931 155.5 Intermediate to LS Circ. Weld (100%) 1 SA-l 101 167.6 1.202 201.5 167.6 0.989 165.8 LS Long. Weld (100%) 1 SA-847 167.0 1.104 184.4 167.0 0.887 148.1 Nozzle-B~it F oiging (NB)------------------- ---~-.. --- -------- ~------- --'46-:0-- 2 123V352 76.0 0.605 76.0 0.429 32.6 Intermediate Shell Forging (IS) 2 123V500 58.0 1.254 72.7 58.0 1.047 60.7 Lower Shell Forging (LS) 2 123W195 42.8* 1.229 52.6 42.8* 1.019 43.6 NB to IS Circ. Weld (100%) 2 21935 170.5 0.605 103.2 170.5 0.429 73.1 Intelmediate to LS Cire. Weld (100%) 2 SA-1484 180.0 1.205 216.9 180.0 0.993 178.7

  • -determmed from surveIllance data Page 30 of 51

A AREVA NON-PROPRIETARY 32-9019240-000 Table 31. ~TNDT Values for the %T and %T WaD Locations ofthe Point Beach Unit 1 and Unit 2 Beltline Materials without Uprate, with hafnium rods, through 53 EFPY (Case 3) Material ViT Location %T Location Beltline Materials Unit Ident. CF ff L'1RTNDI CF ff L'1RTNDT Nozzle Belt Forging (NB) 1 122P237 77.0 0.552 42.5 77.0 0.386 29.7 Intermediate Shell (IS) 1 A9811-1 79.3* 1.304 103.4 79.3* 1.104 87.6 Lower SheIl (LS) 1 C1423-1 35.8* 1.239 44.4 35.8* 1.030 36.9 NB to IS Circ. Weld (100%) 1 SA-1426 167.0 0.552 92.2 167.0 0.386 64.5 IS Long. Weld (ID 27%) 1 SA-812 167.0 1.192 199.1 167.0 N/A N/A IS Long. Weld (OD 73%) 1 SA-775 167.0 N/A N/A 167.0 0.979 163.5 Intermediate to LS Circ. Weld (100%) 1 SA-Il0t 167.6 1.239 207.7 167.6 1.030 172.6 _~~~~Qg~~eld D~QJjl. _______._. ___._ I SA-847 167.0 1.146 191.4 167.0 0.930 155.3 ---.-----f---.-...... --

f---.--.. --

.---~-------

Nozzle Belt Forging (NB) 2 123V352 76.0 0.643 48.9 76.0 0.461 35.0 Intermediate Shell Forging (IS) 2 123V500 58.0 1.294 75.1 58.0 1.092 63.3 Lower Shell Forging (LS) 2 123W195 42.8* 1.267 54.2 42.8* 1.062 45.5 NB to IS Cire. Weld (100%) 2 21935 170.5 0.643 109.6 170.5 0.461 78.6 Intermediate to LS Cire. Weld (100%) 2 SA-1484 180.0 1.243 223.7 180.0 1.034 186.1

  • -determined from surveillance data Table 32. ARTNDT Values for the %T and %T WaD Locations ofthe Point Beach Unit 1 and Unit 2 Beltline Materials without Uprate, with hafnium rods, through 43 EFPY (Case 4)

Material ViT Location 3i4T Location Be1tline Materials Unit Ident. CF ff L'1RTNDT CF ff L'1RTNDT Nozzle Belt Forging (NB) 1 122P237 77.0 0.514 39.6 77.0 0.355 27.3 Intermediate Shell (IS) 1 A9811-1 79.3* 1.257 99.7 79.3* 1.050 83.3 Lower Shell (LS) 1 C1423-1 35.8* 1.197 42.9 35.8* 0.985 35.3 NB to IS Circ. Weld (100%) 1 SA-1426 167.0 0.514 85.8 167.0' 0.355 59.3 IS Long. Weld (ID 27%) 1 SA-812 167.0 1.141 190.6 167.0 N/A N/A IS Long. Weld (OD 73%) 1 SA-775 167.0 N/A N/A 167.0 0.924 154.3 Intermediate to LS Circ. Weld (100%) 1 SA-I101 167.6 1.196 200.4 167.6 0.983 164.8 LS Long. Weld (100%) 1 SA-847 167.0 1.098 183.4 167.0 0.880 147.0 --_. __.-... _--_.. _-_.. _.----_._._-_.. _._------- 1--.... _... - -c-*-*-*--****

---C.;........ _

Nozzle Belt Forging (NB) 2 123V352 76.0 0.S99 4S.5 76.0 0.424 32.2 Intermediate Shell Forging (IS) 2 123VSOO 58.0 1.247 72.3 58.0 1.039 60.3 Lower Shell"Forging (LS) 2 123W195 42.8* 1.223 52.3 42.8* 1.013 43.4 NB to IS Cire. Weld (100%) 2 21935 170.5 0.599 102.1 170.S 0.424 72.3 Intermediate to LS Circ. Weld (100%) 2 SA-1484 180.0 1.200 216.0 180.0 0.987 177.7

  • -determined from survetllance data Page 31 of 51

A AREVA NON-PROPRIETARY 32-9019240-000 Table 33. ARTNDT Values for the %T and %T Wall Locations ofthe Point Beach Unit 1 and Unit 2 Beltline Materials with Uprate, with hafnium rod removal October 2008 (Unit 1 and April 2008 (Unit 2), through 53 EFPY (Case 5) I ~~~~;~1 ~TLoeation 'AT~ati~ Beltline Materials YT CF ff LlRTNDl CF Nozzle Belt Forging (NB) 1 122P237 77.0 0.616 47.4 77.0 0.438 33.7 Intermediate Shell (I~ 1 A9811-1 79.3* 1.313 104.1 79.3* 1.116 88.5 Lower Shell (LS) 1 C1423-J 35.8* 1.297 46.4 35.8* 1.096 39.2 NB to IS Cire. Weld (100%) 1 SA-1426 167.0 0.616 102.9 167.0 0.438 73.2 IS Long. Weld (ID 27%) 1 SA-BI2 161.0 1.209 201.9 167.0 N/A N/A IS Long. Weld (OD 73%) 1 SA-775 167.0 N/A N/A 167.0 0.997 166.5 Intermediate to LS Cire. Weld (100%) 1 SA-1l01 167.6 1.291 216.4 167.6 1.088 182.3 1 SA-847 167.0 1.197 199.9 0.985 164.5 .L~ l:~~g:. W.e1d (1.Q.~2. __... _..... _ f-.... _...... ~~-~--~.---- .... -------.. ----~--~.. _~67:.Q.. ------_...... -.. ----"'-----. Nozzle Belt ForgingiN!3) 2 123V352 76.0 0.704 53.5 76.0 0.512 Intermediate Shell Forging OS) 2 123V500 58.0 1.321 76.6 58.0 1.124 Lower Shell Forging (LS) 2 123W195 42.8* 1.314 56.2 42.8* 1.116 NB to IS Circ. Weld (100%) 2 21935 170.5 0.704 120.0 170.5 0.512 Intermediate to LS Circ. Weld (100%) 2 SA-I484 180.0 1.302 234.4 180.0 Ll02

  • -determined from surveillance data Table 34. ARTNDT Values for the %T and %T Wall Locations of the Point Beach Unit 1 and Unit 2 Beitiine Materials with Uprate, with hafnium rod removal October 2008 (Unit 1 and April 2008 (Unit 2), through 43 EFPY (Case 6) 38.9 65.2 47.8 87.3 198.4 Material 1--_-,V,_4T_L_o-'c ___ at,-io_n'--_-+-_--.._74_T_L_oc...:,a;.=ti.;;.con"--_....-J

~--~~B;e~ltl~i~n~e~M~m~e~ri:ru~s------~U~n~i~t~~Id~e~n~t~.~~C~F~ __ ~~ __ ~~A\\~D

    • ~~~~~~~~_~

Nozzle Belt ForginglNB) I 122P237 77.0 0.557 42.9 77.0 0.389 30.0 Intermediate Shell (IS) 1 A9811-1 79.3* 1.264 100.2 79.3* 1.058 83.9 Lower Shell (LS) 1 C1423-1 35.8* 1.239 44.4. 35.8* l.030 36.9 NB to IS Circ. Weld (100%) 1 SA-1426 167.0 0.557 93.0 167.0 0.389 65.0 IS Long. Weld (ID 27%) 1 SA-812 167.0.1.152 192.4 167.0 N/A N/A IS Long. Weld (OD 73%) 1 SA-775 161.0 N/A N/A 167.0 0.937 156.5 Intermediate to LS Cire. Weld{100%) 1 SA-llOl 167.6 1.233 206.7 167.6 1.024 171.6 J:§_1:.2~£..weld i!Q2~1._.. _.. _.. _... ___.... J._.. §'~-=~41.. _ 167.0 l:.P~.....18?:.'L. 16~ 0.920 _.. I 53:.?..... Nozzle Belt Forging (NB) 2 123V352 76.0 0.641 48.7 76.0 0.459 34.9 Intermediate Shell Forging (1S) 2 123V500 58.0 1.268 73.5 58.0 1.063 61.7 Lower Shell Forging (LS) 2 123Wl95 42.8* 1.259 53.9 42.8* 1.052 45.0 NB to IS Cire. Weld (100%) 2 21935 170.5 0.641 109.3 170.5 0.459 78.3 Intermediate to LS Circ. Weld (100%) 2 SA-1484 180.0 1.245 224.1 180.0 1.036 186.5

  • - determined from surveillance data Page 32 of 51

A AREVA NON-PROPRIETARY 32-9019240-000 Table 35. ARTNDT Values for the %T and %T Wall Locations of the Point Beach Unit 1 and Unit 2 Beltline Materials without Uprate, with hafnium rod removal October 2008 (Unit 1 and April 2008 (Unit 2), throueh 53 EFPY (Case 7) Material 14T Location %T Location Beltline Materials Unit Ident. CF ff ARTNDI CF ff ARTNDT Nozzle Belt Forging (NB) 1 122P237 77.0 0.606 46.7 77.0 0.430 33.1 Intermediate Shell (IS) 1 A981 1-1 79.3* 1.306 103.6 79.3* 1.106 87.7 Lower Shell (LS) 1 C1423-1 35.8* 1.287 46.1 35.8* 1.084 38.8 NB to IS Cire. Weld (100%) 1 SA-1426 167.0 0.606 101.2 167.0 0.430 71.8 IS Long. Weld (lD 27%) 1 SA-8I2 167.0 1.199 200.2 167.0 N/A N/A IS Long. Weld (OD 73%) 1 SA-775 167.0 N/A N/A 167.0 0.986 164.7 Intermediate to LS Cire. Weld (100%) 1 SA-ll01 167.6 1.281 214.7 167.6 1.077 180.5 ~ng. ~~d O_~0~2_. __... _..... _ 1 SA-847 167.0 1.187 198.2 167.0 0.973 162.5 -O~693-f--._.-- "o.5ci3"* r----..... Nozzle Belt Forging (NB) 2 123V352 76.0 52.7 76.0 38.2 Intermediate Shell Forging (IS) 2 123V500 58.0 1.312 76.1 58.0 1.113 64.6 Lower Shell Forging (LS) 2 123W195 42.8* 1.305 55.9 42.8* 1.105 47.3 NB to IS Cire. Weld (100%) 2 21935 170.5 0.693 118.2 170.5 0.503 85.8 Intermediate to LS Cire. Weld (100%) 2 SA-1484 180.0 1.292 232.6 180.0 L090 196.2

  • - detennined from surveIllance data Table 36. ARTNDT Values for the %T and %T Wall Locations of the Point Beach Unit 1 and Unit 2 Beltline Materials without Uprate, with hafnium rod removal October 2008 (Unit 1) and April 2008 Unit 2), throueh 43 EFPY (Case 8)

Material 14T Location %TLocation Beltline Materials Unit Ident. CF ff~ rF ff ' TVT' Nozzle Belt Forging (NB) 1 122P237 no 0.550 42.4 77.0 0.384 29.6 Intermediate Shell (IS) t A9811-1 79.3* 1.258 99.8 79.3* 1.051 83.3 Lower Shell (LS) 1 C1423-1 35.8* 1.232 44.1 35.8* 1.022 36.6 NB to IS Circ. Weld (100%) 1 SA-1426 167.0 0.550 91.9 167.0 0.384 64.1 IS Long. Weld (ID 27%) 1 SA-8I2 167.0 1.145 191.2 167.0 N/A N/A IS Long. Weld (OD 73%) I SA-775 167.0 N/A N/A 167.0 0.929 155.1 Intermediate to LS Cire. Weld (100%) 1 SA-llOl 167.6 1.226 205.5 167.6 1.016 170.3 .'!:.~ Lol!&_~ el~J 1 0~2 __.... _.. __._... _._ 1 SA-847 167.0 1.127 188.2 167.0 0.910 152.0 --~~---~- -_.... _.. _- o.6i:f*... _---.. ---- Nozzle Belt Forging cNB) 2 I 23V352 76.0 48.1 76.0 0.452 34.4 InteImediate Shell Forging (IS) 2 123V500 58.0 1.261 73.1 58.0 LOSS 61.2 Lower Shell Forging (LS) 2 123W195 42.8* 1.251 53.5 42.8* 1.044 44.7 ~irc. Weld (100%) 2 21935 170.5 0.633 107.9 170.5 0.452 77.1 te to LS Cire. Weld (l 00%) 2 SA-1484 180.0 1.236 222.5 180.0 1.027 184.9

  • -detennmed from surveIllance data 5.3 Margin The "margin" is the quantity that is added to obtain conservative, upper-bound values of the ART. The margin is detennined by the following expression:

Page 33 of 51

A AREVA where ar aLl NON-PROPRIETARY

standard deviation for the initial RT NDT

= standard deviation for l'!RTNDI 32-:9019240-000 If a measured value of the initial RT NDT for the material in question is available, O"r is to be estimated from the precision ofthe test method. If generic mean values are used, ar is the standard deviation obtained from the set of data used to establish the mean. The standard deviation for l'!RT NDT, 0',1, is 28°F for welds and 17°F for base metals, except that (jtl need not exceed 0.50 times l'!RT NDT. For cases in which the results from a credible plant-specific surveillance program are used, the value of (jt\\ to be used is 14°F for welds and 8.5°F for base metal; the value of need not exceed one-half of l'!RT NDT. Tables 37 44 list the margin values calculated for the Point Beach Unit 1 and Unit 2 reactor vessel beltline materials for Cases 1 - 8. Table 37. Margin Values for the %T and %T Wall Locations of the Point Beach Unit 1 and Unit 2 Reactor Vessel Beltline Materials with Uprate, with hafnium rods, through 53 EFPY (Case 1) Material a[ I af1 i'lRT 12 I Margin B61tIine Materials Unit Ident. Y4T %T I Y4T %T Nozzle Belt Forging (NB) 1 122P237 0 17.0* 21.55 15.10* 34.0 30.2 Intermediate Shell (IS) 1 A9811-1 26.9 8.5* 50.00 44.15 56,4 56.4 Lower Shell (LS) 1 C1423-1 26.9 85* 22.30 18.60 56,4 56,4 NB to IS Cire. Weld(lOO%) 1 SA-1426 17.2 28.0* 46.75 32.75 65.7 65.7 IS Long. Weld (ID 27%) 1 SA-812 17.2 28.0* 100.35 N/A 65.7 N/A IS Long. Weld (OD 73%) 1 SA-775 17.2 28.0* N/A 82.60 N/A 65.7 Intermediate to LS Cire. Weld 1 SA-1I01 12.9 28.0* 104.55 87.05 61.7 61.7 (100%) 1&~2~g~~~!=!lJ OOr?2. ________ 1 SA-847 17.2 28.0* 96.45 78.50 65.7 65.7 ---~--~---- --~~,.--------- --j7.0*- --24~80---' 1----':--------- ---------- ----------- Nozzle Belt Forging (NB) 2 123V352 0 17.80 34.0 34.0 Intennediate Shell Forging (IS) 2 123V500 0 17.0* 37.&0 31.95 34.0 34.0 Lower Shell Forging (LS) 2 123W195 0 8.5* 27.30 22.90 17.0

17.

NB to IS Cire. Weld (100%) 2 21935 17.0 2.8.0* 55.60 39.90 65.5 65.5 Intennediate to LS Cire. Weld 2 SA-1484 11.9 28.0* 112.50 93.80 60.8 60.8 (100%)

  • -Used to calculate margm term Page 34 of 51

A AREVA NON-PROPRIETARY 32-9019240-000 Table 38. Margin Values for the %T and %T Wall Locations of the Point Beach Unit 1 and Unit 2 Reactor Vessel BeitIine Materials with Uprate, with hafnium rods, through 43 EFPY (Case 2) Material .!lli. T NDI / 2 Margin Beltline Materials Unit Ident. 0'1 0'6 YlT ~T YlT ~T Nozzle Belt Forging (NB) 1 122P237 0 17.0* 20.00 13.80* 34.0 27.6 Intermediate Shell (IS) 1 A98Il-1 26.9 8.5* 50.10 41.90 56.4 56.4 Lower Shell (LS) 1 C1423-1 26.9 8.5* 21.55 17.70 56.4 56.4 NB to IS Cire. Weld (100%) 1 SA-1426 17.2 28.0* 43.35 30.00 65.7 65.7 IS Long. Weld (ID 27%) 1 SA-812 17.2 28.0* 95.80 N/A 65.7 N/A IS Long. Weld (OD 73%) 1 SA-775 17.2 28.0* N/A 77.75 N/A 65.7 Intermediate to LS Cire. Weld 1 SA-I 101 12.9 28.0* 100.75 82.90 61.7 61.7 J100%) LS Long. Weld (100%) 1 SA-847 17.2 28.0* 92.20 74.05 65.7 65.7 i-::---.------------------.---.- ----.----- ---------------- _._--_._-----f--.------. -------- --------- -------_.. - --------------- Nozzle Belt Forging (NB) 2 123V352 0 17.0* 23.00 16.30* 34.0 32.6 Intermediate Shell Forging (IS) 2 123V500 0 17.0* 36.35 30.35 34.0 34.0 Lower Shell Forging (LS) 2 123W195 0 8.5* 26.30. 21.80 17.0 17.0 NB to IS Cire. Weld (100%) 2 21935 17.0 28.0* 51.60 36.55 65.5 65.5 Intermediate to LS Cire. Weld 2 SA-1484 11.9 28.0* 108.45 89.35 60.8 60.8 (100%)

  • -Used to calculate margin tenn Table 39. Margin Values for the %T and %T WaD Locations ofthe Point Beach Unit 1 and Unit 2 Reactor Vessel Beltline Materials without Uprate, with hafnium rods, through 53 EFPY (Case 3)

Material ~RTNDT/2 Margin Beltline Materials Unit Ident. 0'1 0'6 YlT ~T YlT 3i4T Nozzle Belt Forging (NB) 1 122P237 0 17.0* 21.25 14.85* 34.0 29.7 Intermediate Shell (1S) 1 A9811-1 26.9 8.5* 51.70 43.80 56.4 56.4 Lower Shell (LS) 1 C1423-1 26.9 8.5* 22.20 18.45 56.4 56.4 NB to IS Cire. Weld (100%) 1 SA-1426 17.2 28.0* 46.10 32.25 65.7 65.7 IS Long. Weld (ID 27%) 1 SA-812 17.2 28.0* 99.55 N/A 65.7 N/A IS Long. Weld (OD 73%) I SA-775 17.2 28.0* NIA 81.75 N/A 65.7 Intermediate to LS Cire_ Weld 1 SA-UOI 12.9 28.0* 103.85 863 61.7 61.7 (100%) LS Long. Weld (100%) 1 SA-847 17.2 28.0* 95.70 77.65 65.7 65.7 r:.----------------.. ------------------. ---------- r-i23V352 ---_.-------- -:-::-,------ ------------- ---------- -------- ------- Nozzle Belt Forging (NB) 2 0 17.0* 24.45 17.50 34.0 34.0 Intermediate SheIl Forging (IS) 2 123V500 0 17.0* 37.55 31.65 34.0 34.0 Lower Shell Forging (LS) 2 123WI95 0 8.5* 27.10 22.75 17.0 17.0 NB to IS Cire. Weld (100%) 2 21935 17.0 28.0* 54.80 39.30 65.5 65.5 Intermediate to LS Cire. Weld 2 SA-1484 11.9 28.0* 111.85 93.05 60.8 60.8 (100%)

  • -Used to calculate margin tenn Page 35 of 51

A AREVA NON-PROPRIETARY 32-9019240-000 Table 40. Margin Values for the %T and %T Wall Locations of the Point Beach Unit 1 and Unit 2 Reactor Vessel Beltline Materials without Uprate, with hafnium ro d th h 43 EFPY (C

4) s, rougl ase Material

~RTNDr 12 Margin Beltline Materials Unit Ident_ <Y] <Yll. Y.tT %T !.iT %T Nozzle Belt Forging (NB) 1 122P237 0 17_0* 19.80 13.65* 34.0 27.3 InteImediate Shell (IS) 1 A9811-1 26.9 8.5* 49.85 41.65 56.4 56.4 Lower Shell (LS) 1 C1423-1 26.9 8.5* 21.45 17.65 56.4 56.4 NB to IS Cire. Weld (100%) 1 SA-1426 172 28.0* 42.90 29.65 65.7 65.7 IS Long. Weld (ID 27%) 1 SA-812 17.2 28.0* 95.30 N/A 65.7 N/A IS Long. Weld (OD 73%) 1 SA-775 17.2 28.0* N/A 77.15 N/A 65.7 Intermediate to LS Cire_ Weld I 8A-1101 12.9 28_0* 100..20 82.40 61.7 61.7 (100%) L8 Long. Weld (100%) I SA-847 17.2 28.0* 91.70 73.50 65.7 65.7 Nozzie-Beii"Forging -(NBy------ ------2----r--------- 123V352 0 17.0* 22.75 16.10* 34.0 32.2 Inteiinediate Shell Forging (IS) 2 123V500 0 17.0* 36.15 30.15 34.0 34.0 Lower Shell Forging (LS) 2 123W195 0 8.5* 26.15 21.70 17.0 17.0 NB to IS Cire. Weld (100%) 2 21935 17.0 28.0* 51.05 36.15 65.5 65.5 Intermediate to LS Cire. Weld 2 SA-1484 11.9 28_0* 108_00 88_85 60_8 60.8 (100%)

  • -Used to calculate margm term Table 41. Margin Values for the %T and %T Wall Locations of the Point Beach Unit 1 and Unit 2 Reactor Vessel Beltline Materials with Uprate, with hafnium rod removal October 2008 (!Lnit 1) and Ajlril 2008 (Unit 2), through 53 EFPY (Case 5)

Material .6.RTNDT /2 Margin Beltline Materials Unit Ident. <Y] Gll. Y.tT %T Y.tT YoT Nozzle Belt Forging (NB) I 122P237 0 17.0* 23.70 16.85* 34.0 33.7 Intermediate Shell (IS) I A9811-1 26.9 8.5* 52.05 44.25 56.4 56.4 Lower Shell (LS) 1 C1423-1 26.9 8.5* 23.20 19.60 56.4 56.4 NB to IS Cire. Weld (100%) I SA-1426 17.2 28.0* 51.45 36.60 65.7 65.7 IS Long. Weld (ID 27%) 1 SA-812 17.2 28.0* 100.95 N/A 65.7 N/A IS Long. Weld (OD 73%) 1 SA-775 17.2 28_0* NIA 83_25 N/A 65.7 Intermediate to LS Cire. Weld 1 SA-1101 12_9 28.0* 108.20 91.15 61.7 61.7 (100%) _!=.~_!:g_I].&:_~~_~_ (100%) ______________ I SA-847 17.2 28.0* 99_95 82_25 65.7 65.7 Nozzle Belt Forging (NB) 2 123V352 0 17.0* 26.75 19.45 34.0 34.0 Intermediate Shell Forging (IS) 2 123V500 0 17.0* 38.30 32.60 34.0 34.0 Lower Shell Forging (LS) 2 123W195 0 8.5* 28.10 23.90 17.0 17.0 NB to IS Cire_ Weld (100%) 2 21935 17.0 28.0* 60.00 43.65 65.5 65.5 Intermediate to LS Cire. Weld 2 SA-1484 11.9 28.0* 117.20 99.20 60.8 60.8 (100%)

  • -Used to calculate margm term Page 36 of 51

A AREVA NON-PROPRIETARY 32-901 9240-000 Table 42. Margin Values for the %T and %T Wall Locations ofthe Point Beach Unit 1 and Unit 2 Reactor Vessel Beltline Materials with Uprate, with hafnium rod removal October 2008 (Unit 1) and April 2008 (Unit 2), through 43 EFPY (Case 6) Material ~RTNDT 12 Margin Beltline Materials Unit Ident al a~ Y4T %T Y4T %T Nozzle Belt Forging (NB) 1 122P237 0 17.0* 21.45 15.00* 34.0 30.0 Intermediate Shell (IS) 1 A9811-1 26.9 8.5* 50.10 41.95 56.4 56.4 Lower Shell (LS) 1 C1423-1 26.9 8.5* 22.20 18.45 56.4 56.4 NB to IS Cire. Weld (100%) 1 SA-1426 17.2 28.0* 46.50 32.50 65.7 65.7 IS Long. Weld (ID 27%) 1 SA-812 17.2 28.0* 96.20 NIA 65.7 NIA IS Long. Weld (OD 73%) 1 SA-775 17.2 28.0* NIA 78.25 NIA 65.7 Intennediate to LS Cire. Weld 1 SA-II01 12.9 28.0* 103.35 85.8 61.7 61.7 (100%) .~~.!:-.<?I?:g.:.~~ld (.!.OO~L._. ___ 1 SA-847 17.2 28.0* 94.85 76.80 65.7 65.7 r--'-'-.. _.... __.- ---~~--------

_. r---...... - --.... _._- _...... -.-

Nozzle Belt Forging (NB) 2 123V352 0 17.0* 24.35 17.45 34.0 34.0 Intennediate Shell Forging (IS) 2 123V500 0 17.0* 36.75 30.85 34.0 34.0 Lower Shell Forging (LS) 2 123W195 0 8.5* 26.95 22.50 17.0 17.0 NB to IS Cire. Weld (100%) 2 21935 17.0 28.0* 54.65 39.15 65.5 65.5 Intennediate to LS Cire. Weld 2 SA-1484 11.9 28.0* 112.05 93.25 60.8 60.8 (100%)

  • -Used to calculate margm tenn Table 43. Margin Values for the %T and %T Wall Locations of the Point Beach Unit 1 and Unit 2 Reactor Vessel Beltline Materials without Up rate, with hafnium rod removal October 2008 (Unit 1) and April 2008 (Unit 2), through 53 EFPY (Case
7)

Material ~RTNDT 12 Margin Beltline Materials Unit Ident. al a~ Y4T 3i4T Y4T 3i4T Nozzle Belt Forging (NB) 1 I 22P237 0 17.0* 23.35 16.55* 34.0 33.1 Intennediate Shell (IS) 1 A9811-1 26.9 8.5* 51.80 43.85 56.4 56.4 Lower Shell (LS) 1 C1423-1 26.9 8.5* 23.05 19.40 56.4 56.4 NB to IS Cire. Weld (100%) 1 SA-1426 17.2 28.0* 50.60 35.90 65.7 65.7 IS Long. Weld (ID 27%) 1 SA-812 17.2 28.0* 100.10 N/A 65.7 N/A IS Long. Weld (OD 73%) I SA-775 17.2 28.0* NIA 82.35 N/A 65.7 Intennediate to LS Cire. Weld 1 SA-II01 12.9 28.0* 107.35 90.25 61.7 61.7 (100%) .!:-.~.!:-.<!.I?:g: W:~.!~j.!g.Q.~L.... _.... _. 1 SA-847 17.2 28.0* 99.10 81.25 65.7 65.7 Nozzle Belt Forging (NB) 2 123V352 0 17.0* 26.35 19.10 34.0 34.0 Intennediate Shell Forging (IS) 2 123V500 0 17.0* 38.05 32.30 34.0 34.0 Lower Shell Forging (LS) 2 123W195 0 8.5* 27.95 23.65 17.0 17.0 NB to IS Cire. Weld (100%) 2 21935 17.0 28.0* 59.10 42.90 65.5 65.5 Intennediate to LS Cire. Weld 2 SA-1484 11.9 28.0* 116.3 98.10 60.8 60.8 (100%)

  • -Used to calculate margm tenn Page 37 of 51

A AREVA NON-PROPRIETARY 32-9019240-000 Table 44. Margin Values for the YtT and %T Wall Locations of the Point Beach Unit 1 and Unit 2 Reactor Vessel Beltline Materials without Uprate, with hafnium rod removal October 2008 (Unit 1) and April 2008 (Unit 2), throngh 43 EFPY (Case

8)

Material MTNDr/2 Margin Beltline Materials Unit Ident Gl Gil !4T %T !4T %T Nozzle Belt Forging (NB) 1 1 22P237 0 17.0* 21.20 14.80* 34.0 29.6 Intelmediate Shell (IS) 1 A9811-1 26.9 8.5* 49.90 41.65 56.4 56.4 Lower Shell (LS) 1 C1423-1 26.9 8.5* 22.05 18.30 56.4 56.4 NB to IS Cire. Weld (100%) 1 SA-I 426 17.2 28.0* 45.95 32.05 65.7 65.7 IS Long. Weld (ill 27%) 1 SA-8I2 17.2 28.0'" 95.60 N/A 65.7 N/A IS Long. Weld (OD 73%) 1 SA-775 17.2 28.0* NJA 77.55 NJA 65.7 Intermediate to LS Cire. Weld 1 SA-llOI 12,9 28.0* 102.75 85.15 61.7 61.7 (100%) 65.7 !:-.~ L<?E.S: W~~~J.100.~L.. _.. ___ 1 SA-847 17.2 28.0* 94.10 76.00 65.7 123'1352-1--------.--.. "17.0* -_.. -_.---:.'-- -~-~--~~--- -._--------'----------- Nozzle Belt Forging (NB) 2 0 24.05 17.20 34.0 Intermediate Shell Forging (IS) 2 123V500 0 17.0* 36.55 30.60 34.0 Lower Shell Forging (LS) 2 123W195 0 8.5* 26.75 22.35 17.0 NB to IS Cire.Weld (100%) 2 21935 17.0 28.0* 53.95 38.55 65.5 Intermediate to LS Cire. Weld 2 SA-1484 11.9 28.0* 111.25 92.45 60.8 (100%)

  • -Used to calculate margin term 5.4 Calculation of Adjusted Reference Temperature (ART)

The ART is given by the following expression: ART== Initial RTNDT + ART NDT + Margin Tables 45 - 52 list the ART values at the 1;4T and %T wall1ocations calculated for the Point Beach Unit 1 and Unit 2 reactor vessel beltline materials for Cases 1 - 8. 34.0 34.0 17.0 65.5 60.8 Page 38 of 51 (5)

A AREVA NON-PROPRIETARY 32-9019240-000 Table 45. ART Values for the VaT and %T Wall Locations ofthe Point Beach Unit 1 and Unit 2 Reactor Vessel BeJtJine Materials with Uprate, with hafnium rods, through 53 EFPY (Case 1) Initial ARINDT, of Margin, of Adjusted Reference Reference Temperature, of Material I emperature, Y.T Y<T Y.T Y<T Y.T Y.T Beltline Materials Unit Ident. of Location Location Location Location Location Location Nozzle Belt Forging (NB) 1 122P237 50 43.1 30.2 34.0 30.2 127.1 IlO.4 Intermediate Shell (IS) 1 A981l-1 1 104.0 88.3 56.4 56.4 161.4 145.7 Lower Shell (LS) 1 C1423-1 1 44.6 37.2 56.4 56.4 102.0 94.6 NB to IS Cire. Weld (100%) 1 SA-1426 -47.6 93.5 65.5 65.7 65.7 111.6 83.6 IS Long. Weld (ID 27%) 1 SA-812 -47.6 200.7 N/A 65.7 N/A 218.8 N/A IS Long. Weld (OD 73%) 1 SA-775 -47.6 N/A 165.2 N/A 65.7 N/A 183.3 InteImediate to LS Cire. Weld I SA-lIOl -47.4 209.1 174.1 61.7 61.7 223.4 188.4 (100%) ~~l:0n!L~:!~.L~2Q~<i. ______._ I SA-847 -47.6 192.9 157.0 65.7 65.7 211.1 175.3 r--------* Nozzle Belt Forging (NB) 2 123V352 40 49.6 35.6 34.0 34.0 123.6 109.6 Intennediate Shell Forging (IS) 2 123V500 40 75.6 63.9 34.0 34.0 149.6 137.9 Lower Shell Forging (LS) 2 123W195 40 54.6 45.8 17.0 17.0 111.6 102.8 NB to IS Cire. Weld (100%) 2 21935 -56 111.2 79.8 65.5 65.5 120.7 89.3 Intermediate to LS Cire. Weld 2 SA-1484 -30 225.0 187.6 60.8 60.. 8 255.8 218.4 (100%) Table 46. ART Values for the ¥4T and %T Wall Locations of the Point Beach Unit 1 and Unit 2 Reactor Vessel Beltline Materials with Uprate, with hafnium rods, through 43 EFPY (Case 2) Initial ARIND1, OF Margin, OF Adjusted Reference Reference Temperature, OF Material Temperature, Y.T Y.T Y.T %T Y.T Y.T Beltline Materials Unit Ident. OF Location Location Location Location Location Location Nozzle Belt Forging (NB) 1 122P237 50 40.0 27.6 34.0 27.6 124.0 105.2 Intermediate Shell (IS) 1 A98Il-l 1 100.2 83.8 56.4 56.4 157.6 141.2 Lower Shell (LS) 1 C1423-1 1 43.1 35.4 56.4 56.4 100.5 92.8 NB to IS Cire. Weld (100%) I SA-l426 -47.6 86.7 60.0 65.7 65.7 104.8 78.1 IS Long. Weld (ID 27%) I SA-812 -47.6 191.6 N/A 65.7 N/A 209.7 N/A IS Long. Weld (OD 73%) I SA-775 -47.6 N/A 155.5 N/A 65.7 N/A 173.6 InteImediate to LS Cire. Weld I SA-I 101 -47.4 201.5 165.8 61.7 61.7 215.8 180.1 (100%) _~~_ Lo~~: W~!~J~ OO%L... _______. I SA-847 -47.6 184.4 148.1 65.7 65.7 202.5 166.4 ,---- --------- -------------------------- ------------ ------1--::-------- ----_._.-1--,----------- Nozzle Belt Forging (NB) 2 123V352 40 46.0 32_6 34.0 32.6 120.0 105.2 Intennediate Shell Forging (IS) 2 123V500 40 72.7 60.7 34.0 34.0 146.7 134.7 Lower Shell Forging (LS) 2 123W195 40 52.6 43.6 17.0 17.0 109.6 100.6 NB to IS Cire. Weld (100%) 2 21935 -56 103.2 73.1 65.5 65_5 112.7 82.6 InteImediate to LS Circ_ Weld 2 SA-l484 -30 216.9 178.7 60.8 60.8 247,7 209.5 (100%) Page 39 of 51

A AREVA NON-PROPRIETARY 32-9019240-000 Table 47. ART Values for the %T and %T Wall Locations ofthe Point Beach Unit 1 and Unit 2 Reactor Vessel BeJtline Materials without Uprate, with hafnium rods, through 53 EFPY (Case 3) Initial ARTNDT, of Margin, OF Adj usted Reference Reference Temperature, OF Material Temperature, 14T '/4T V.T %T V.T %T Beltline Materials Unit Ident OF Location cation Location Location Location Location Nozzle Belt Forging (NB) 1 122P237 50 42.5 29.7 34.0 29.7 ~ 109.4 Intemlediate Shell (IS) 1 A9811-1 1 103.4 87.6 56.4 56.4 160 145.0 Lower Shell (LS) 1 C1423-1 1 44.4 36.9 56.4 56.4 101.8 94.3 NB to IS Cire. Weld (100%) 1 SA-1426 -47.6 92.2 64.5 65.7 65.7 110.3 82.6 IS Long. Weld (lD 27%) I SA-812 -47.6 199.1 NlA 65.7 NIA 217.2 N/A IS Long. Weld (OD 73%) 1 SA-775 -47.6 N/A 163.5 N/A 65.7 N/A 181.6 Intelmediateto LS Cire. Weld 1 SA-I 101 -47.4 207.7 172.6 61.7 61.7 222.0 186.9 (100%) J& Long. _~~~{1 OO~2 __________ 1 SA-847 -47.6 191.4 155.3 65.7 65.7 209.5 173.6 ~----------- ---~~.-~- ---48.9 --~----- ~-------~- ---------:---'-- Nozzle Belt Forging (NB) 2 123V352 40 35.0 34.0 34.0 122.9 109.0 Intennediate Shell Forging (IS) 2 123V500 40 75.1 63.3 34.0 34.0 149.1 l37.3 Lower SheIl Forging (L8) 2 123Wl95 40 54.2 45.5 17.0 17.0 111.2 102.5 NB to IS Cire. Weld (100%) 2 21935 -56 109.6 78.6 65.5 65.5 119.1 88.1 InteImediate to L8 Cire. Weld 2 SA-1484 ~30 223.7 186.1 60,8 60.. 8 254.5 216.9 (100%) Table 48. ART Values for the %T and %T Wall Locations of the Point Beach Unit 1 and Unit 2 Reactor Vessel BeJtJine Materials without Uprate, with hafnium rods, through 43 EFPY (Case 4) lnitial ARTND1, OF Margin, OF Adjusted Reference Reference Temperature, OF Material Temperature, 14T %T !4T %T V.T %T Beltline Materials Unit Ident. OF Location Location Location Location Location Location Nozzle Belt Forging (NB) 1 122P237 50 39.6 27.3 34.0 27.3 123.6 104.6 Intermediate Shell (IS) t A98l1-1 1 99.7 83.3 56.4 56.4 157.1 140.7 Lower Shell (L8) I C1423-1 1 42.9 353 56.4 56.4 100.3 92.7 NB to IS Cire. Weld (100%) 1 SA-1426 -47.6 85.8 59.3 65.7 65.7 103.9 77.4 IS Long. Weld (ID 27%) 1 SA-8I2 -47.6 190.6 N/A 65.7 N/A 208.7 N/A IS Long. Weld (OD 73%) I SA-775 -47.6 N/A 154.3 N/A 65.7 N/A 172.4 Intermediate to LS Cire_ Weld I SA-llOI -47.4 200A 164.8 6L7 61.7 214.7 179.1 (100%) LS Long. Weld (100%) 1 SA-847 -47.6 183.4 147.0 65.7 65.7 201.4 165.3 123V352' -------- ---._------ -----------r------------ ---~------.. - ~----.... --- :------------- Nozzle Belt Forging (NB) 2 40 45.5 32.2 34.0 32.2 119.5 104.4 Intennediate Shell Forging (IS) 2 123V500 40 72.3 60.3 34.0 34.0 146.3 134.3 Lower Shell Forging (LS) 2 123W195 40 52.3 43.4 17.0 17.0 109.3 100.4 NB to IS Cire. Weld (100%) 2 21935 -56 102.1 72.3 65.5 65.5 111.6 81.8 InteImediate to LS Cire_ Weld 2 SA-1484 ~30 216_0 ]77.7 60_8 60.8 246_8 208_5 (100%) Page 40 of 51

A AREVA NON-PROPRIETARY 32-9019240-000 Table 49. ART Values for the ~T and %T Wall Locations of the Point Beach Unit 1 and Unit 2 Reactor Vessel Beltline Materials with Uprate, with hafnium rod removal October 2008 (Unit 1) and April 2008 (Unit 2), through 53 EFPY (Case 5) Initial ~RTNDT, OF Margin, OF Adjusted Reference Reference Temperature, OF Material Temperature, YoT 7'.T YoT 7'.T \\I.oT 7'.T Beltline Materials Unit Ident. OF Location Location Location Location Location Location Nozzle Belt Forging (NB) 1 122P237 50 47.4 33.7 34.0 33.7 131.4 117.4 Intennediate Shell (IS) 1 A9811-1 1 104.1 88.5 56.4 56.4 161.5 145.9 Lower Shell (LS) 1 C142J-1 1 46.4 39.2 56.4 56.4 103.8 96.6 NB to IS Circ. Weld (100%) 1 SA-I 426 -47.6 102.9 73.2 65.7 65.7 121.0 91.3 IS Long. Weld (ID 27%) 1 SA-SI2 -47.6 201.9 N/A 65.7 N/A 220.0 N/A IS Long. Weld (OD 73%) 1 SA-775 -47.6 N/A 166.5 N/A 65.7 N/A 184.6 Intermediate to LS Gire. Weld 1 SA-11 01 -47.4 216.4 182.3 61-7 61-7 230.7 196.6 (100%) LS Long. Weld (100%) 1 SA-847 -47.6 199.9 164.5 65.7 65.7 I-?~L 182.7 --~--- r--,---'o- -----1----------- ---------- ~~---------- Nozzle Belt Forging (NB) 2 123V352 40 53.5 38.9 34.0 34.0 127.5 112.9 InteImediate Shell Forging (IS) 2 I 23V500 40 76.6 65.2 34.0 34.0 150.6 139.2 Lower Shell Forging (LS) 2 123WI95 40 56.2 47.8 17.0 17.0 113.2 104.8 NB to IS Circ. Weld (100%) 2 21935 -56 120.0 87.3 65.5 65.5 129.5 96.8 Intermediate to LS Cire. Weld 2 SA-I 484 -30 234.4 198.4 60.8 60.8 265.2 229.2 (100%) Table 50. ART Values for the ~T and %T Wall Locations of the Point Beach Unit 1 and Unit 2 Reactor Vessel Beltline Materials with Uprate, with hafnium rod removal October 2008 (Unit 1) and April 2008 (Unit 2), through 43 EFPY (Case 6) Initial ~RTNoI' OF Margin, OF Adjusted Reference Reference Temperature, OF Material Temperature, YoT 7'.T YoT 7'.T YoT Y.T Beltline Materials Unit Ident. OF Location Location Location Location Location Location Nozzle Belt Forging (NB) 1 122P237 50 42.9 30.0 34.0 30.0 126.9 110.0 Intermediate Shell (IS) 1 A9SII-1 1 100.2 83.9 56.4 56.4 157.6 141.3 Lower Shell (LS) 1 C1423-1 1 44.4 36.9 56.4 56.4 101.8 94.3 NB to IS Cite. Weld (100%) 1 SA-1426 -47.6 93.0 65.0 65.7 65.7 111.1 83.1 IS Long. Weld (ID 27%) I SA-SI2 -47.6 192.4 N/A 65.7 N/A 210.5 NIA IS Long. Weld (OD 73%) I SA-775 -47.6 N/A 156.5 N/A 65.7 N/A 174.6 Intennediate to LS Circ. Weld 1 SA-I 101 -47.4 206.7 171.6 61.7 61-7 221.0 185_9 (100%) _~~_~0~~:..~~!~JlQ2~L ___________ 1 SA-S47 -47.6 189.7 153.6 65.7 65.7 207.8 171.9

--------------- ------------ ------_._---- r------------ -------- --------- ---------------

Nozzle Belt Forging (NB) 2 123V352 40 48.7 34.9 34.0 34.0 122.7 108.9 InteImediate Shell Forging (IS) 2 123V500 40 73.5 61.7 34.0 34.0 147.5 i35.7 Lower Shell Forging (LS) 2 123W195 40 53.9 45.0 17.0 17.0 110.9 102.0 NB to IS Cire. Weld (100%) 2 21935 -56 109.3 78.3 65.5 65.5 118.8 87.8 Intennediate to LS Circ. Weld 2 SA-14S4 -30 224.1 186.5 60.8 60.8 254.9 217.3 (100%) Page 41 of 51

A AREVA NON-PROPRIETARY 32-9019240-000 Table 51. ART Values for the %T and %T Wall Locations ofthe Point Beach Unit 1 and Unit 2 Reactor Vessel BeltJine Materials without Uprate, with hafnium rod removal October 2008 (Unit 1) and Aj!riI2008 (Unit 2), through 53 EFPY (Case 7) Initial ~RTNDT, of Margin, of Adjusted Reference Reference Te~erature, of Material Temperature, 'l.T %T 'l.T %T 'AT %T Beltline Materials Unit Ident of Location Location Location Location Location Location Nozzle Belt Forging (NB) I 122P237 50 46.7 33.1 34.0 33.1 130.7 116.2 Intennediate Shell (IS) 1 A9811-1 1 103.6 87.7 56.4 56.4 161.0 145.1 Lower Shell (LS) I C1423-1 1 46.1 38.8 56.4 56.4 103.5 96.2 NB to IS Cire. Weld (100%) 1 SA* 1426 -47.6 101.2 71.8 65.7 65.7 119.3 89.9 IS Long. Weld (10 27%) 1 SA*812 -47.6 200.2 N/A 65.7 N/A 218.3 N/A IS Long. Weld (OD 73%) I SA-775 -47.6 N/A 164.7 N/A 65.7 N/A 182.8 Intermediate to LS Cire. Weld 1 SA-llOI -47,4 214.7 180.5 61.7 61.7 229.0 194.8 (100%) !:§}:,~E~:~.~!QQ°O/~. __._....... _ 1 SA*847 -47.6 198.2 162.5 65.7 65.7 216.3 180.9 -.. --........... -.-1---'_._--' ~---~-- ------- -----------....... _ .... _-f-.............. -.-. Nozzle Belt Forging (NB) 2 123V352 40 52.7 38.2 34.0 34.0 126.7 112.2 Intermediate Shell Forging (IS) 2 123V500 40 76.1 64.6 34.0 34.0 150.1 138.6 Lower Shell Forging (LS) 2 123W195 40 55.9 47.3 17.0 17.0 112.9 104.3 NB to ISCire. Weld (100%) 2 21935 -56 118.2 85.8 65.5 65.5 127.7 95.3 Intermediate to LS Cire. Weld 2 SA-1484 -30 232.6 196.2 60.8 60.8 263.4 227.0 (100%) Table 52. ART Values for the %T and %T Wall Locations of the Point Beach Unit 1 and Unit 2 Reactor Vessel Beltline Materials without Uprate, with hafnium rod removal October 2008 (Unit 1) and Aj!ril2008 (Unit 2), through 43 EFPY ( Case 8) Initial ~RTNDr, OF Margin, OF Adjusted Reference Reference Temperature, OF Material Temperature, 'l.T 'AT Y<T

v..r Y.r

%T Belt1ine Materials Unit Ident OF Location Location Location Location Location Location Nozzle Belt Forging (NB) I 122P237 50 42.4 29.6 34.0 29.6 126.4 109.2 Intermediate Shell (IS) 1 A9811-1 1 99.8 83.3 56.4 56.4 157.2 140.7 Lower Shell (LS) 1 C1423-1 1 44.1 36.6 56.4 56.4 101.5 94.0 NB to IS Cire. Weld (100%) I SA-l 426 -47.6 91.9 64.1 65.7 65.7 110.0 82.2 IS Long. Weld (ID 27%) I SA-8I2 -47.6 191.2 N/A 65.7 N/A 209.3 N/A IS Long. Weld (OD 73%) I SA-775 -47.6 N/A 155.1 N/A 65.7 N/A 173.2 Intelmediate to LS eire. Weld 1 SA-IIOI -47.4 205.5 170.3 61.7 61.7 219.8 184.6 (100%) .. ~§..!:.?E&.~~!!t!QQ%).. _"_" __ I SA-847 -47.6 188.2 152.0 65.7 65.7 206.3 170.3 ._.... - --------------- --.----.-------- ----------.--...... --1-...... _-- Nozzle Belt Forging (NB) 2 123V352 40 48.1 34.4 34.0 34.0 122.1 108.4 Intermediate Shell Forging (IS) 2 123V500 40 73.1 61.2. 34.0 34.0 147.1 135.2 Lower Shell Forging (LS) 2 123W195 40 53.5 44.7 17.0 17.0 110.5 101.7 NB to IS Cire. Weld (100%) 2 21935 -56 107.9 77.1 65.5 65.5 117.4 86.6 Intennediate to LS Cire. Weld 2 SA-1484 -30 222.5 184.9 60,8 60.8 253.3 215.7 (100%) Page 42 of 51

A AREVA NON-PROPRIETARY 32-9019240-000 6.0 Adjusted Reference Temperature Calculation Where Surveillance Data is Available The adjusted reference temperature may be calculated when two or more credible surveillance data sets are available. Using the Lill.TNDT and its corresponding fluence, the chemistry factor may be calculated by multiplying each adjusted ~RT NDT by the corresponding fluence factor, summing the products, and dividing by the sum of the squares of the fluence factors. The Master Integrated Reactor Vessel Surveillance Program (MIRVP) described in BAW-1543, Revision 4[3), provides surveillance data for the Point Beach Unit 1 weld metals SA-11Ol and SA-847. The base metal evaluations completed in Reference 5 for all base metals pertinent to Point Beach Unit 1 (A9811-1 and C1423-1) and Unit 2 (123W195) are valid. 6.1 Surveillance Data Credibility Assessment When assessing credibility for surveillance data from several sources, the capsule data may have to be adjusted to account for the irradiation environment and chemical composition differences. Temperature Adjusted ARTNDT,normaliud = ART NDT,measu,ed + 1.0 * (Tcapsule - Tcapsulemean) Additionally, if the surveillance data are from multiple sources, it is necessary to adjust the capsule data for chemical composition (copper and nickel content) differences. For the credibility determination, the surveillance data are "normalized" to the mean copper and nickel contents of the surveillance materials using the following equation: ( CF ) Table, Surv. Avg. Chem.

  • RatIO Adjusted ART NDT,no,malized -

ART NDT,measu,ed CFTable, Surv. Chem. A best-fit line (least squares regression) is then determined from the adjusted ~RT NOT capsule surveillance data as a function of the capsule fluence factor. The data are considered credible if the difference between the adjusted Lill.TNoT (i.e., chemistry adjusted) and the predicted ~RTNOT (from the best-fit line) for all the data is within +/-28°F for weld metals and +/-17°F for base metals. Page 43 of 51

A AREVA NON-PROPRIETARY 32-9019240-000 6.2 Credible SurveilIance Data In accordance with Regulatory Guide 1.99, Revision 2 and 10 CFR 50.61, credible surveillance data are used to determine material-specific chemistry factor values for use in reactor vessel integrity assessments. The chemistry factor is determined from a best-fit line through the surveillance data adjusted to account for differences in chemical composition (i.e., copper and nickel contents) and irradiation environment (Le., irradiation temperature) between the capsules and the vessel. The surveillance data are adjusted in the same manner as for the credibility determination except that the 30 ft-Ib transition temperature values are "normalized" to the best estimate copper and nickel contents and the irradiation temperature of the vessel being assessed. 6.3 Non Credible Surveillance Data If the surveillance data are determined to be non-credible, the chemistry factor value is calculated from the generic Tables in 10 CFR 50.61 and Regulatory Guide 1.99, Revision 2 unless the chemistry factor determined from the surveillance data is significantly greater than that from the generic Tables, indicating that the Table chemistry factor is non-conservative. To determine if the generic Table chemistry factor is non-conservative, the following steps are performed:

1.

Determine the chemistry factor from the generic Tables based on the surveillance specimen chemical composition; use this chemistry factor to determine the predicted ilRTNDI for each capsule: (Predicted ART NDT :::: CF Table, Surv. Avg. Chern * !!capsule)

2.

Determine difference between the predicted ilRTNDT and the measured ilRTNDT. If the difference between the predicted LlRTNDT and the measured ilRTNDI values exceeds 2 standard deviations (i.e., 56°P for weld metals and 34°P for base metals), the Table chemistry factor is considered non-conservative. When the Table chemistry factor is determined to be non-conservative, the chemistry factor determined from the "non-credible" surveillance data is used in the assessment of reactor vessel integrity using the "full" value of crt, in calculating the Margin term. 6.4 Assessment ofthe Weld Wire Heat Surveillance Data Tables 53 - 58 provide the credibility assessment of the weld wire heats 71249, 61782, and 72442. NOTE: The original Charpy V-notch impact data are based on hand-fit Charpy curves using engineering judgment; these data were using a hyperbolic tangent curve fitting program to achieve consistency in the interpretation of the available surveillance test data. Page 44 of 51

A AREVA NON-PROPRIETARY Table 53. Credibility Assessment for Weld Wire Heat Number 71249 Irrad. Capsule Chern. Fluence Cuwt% Niwt% Factor Factor 0.33 0.57 201.3 546 0.739 0.915 0.33 0.57 201.3 546 1.530 1.118 0.29 0.60 191.0 546 0.708 0.903 0.33 0.57 201.3 546 2.900 1.283 where Predicted ARTNDT = (SlopebestfzJ * (Fluence Factor) and Slopebestjit = best fit line relating Adjusted ART NDT to the Fluence Factor (i.e., 172.7) Temp. Meas. Adj. t.RTNDT t.RTNDT 166 165 179 178 211 210 191 190 These data points are not credible since the scatter is greater than +/-28°P for two surveillance points. 32-9019240-000 Pred. Ratio Adj. t.RTNDT (OF) (Adjusted-t.RTNDT from best fit Predicted) 163.3 158.0 5.3 176.2 193.1 -16.9 219.0 156.0 63.0 188.0 221.6 -33.6 Page 45 of 51

A AREVA NON-PROPRIETARY 32-9019240-000 Table 54. Table Chemistry Factor Non-Conservatism Assessment for Weld Wire Heat Number 71249 TableChem. Capsule Adjusted (Adjusted - Capsule Factor (Surv. Fluence LlRTNDT Predicted Predicted) t.RT NDT Designation Avg.) Factor (OF) LIR T NDT (OF) (OF) Turkey Pomt Umt 3: Capsule T 199.2 0.915 163.3 182.3 -19.0 SA-I 101 : Plant Specific RVSP Malena! Turkey Pomt Unit 3: Capsule V 199.2 1.118 176.2 222.7 -46.5 SA-I 101 : Plant Specific RVSP Materia! Turkey Point Unit 4: Capsule T 199.2 0.903 219.0 179.9 39.1 SA-I094: Plant Specific RVSP Matenal Turkey Pomt Umt 3: Capsule X 199.2 1283 188.0 255.6 -67.6 SA-II 01: Plant Specific RVSP MaterIal B&WOG: CapsuleA5 199.2 1.253 216.7 249.6 -32.9 SA-1I01: TP-3 Plant Specific RVSP MatI. The above assessment results indicate that the generic Table chemistry factor for the surveillance data grossly over-predicts the adjusted measured data. Therefore, the Table chemistry factor calculated using the weld wire heat best-estimate copper and nickel contents is considered conservative. Page 46 of 51

A AREVA NON-PROPRIETARY Table 55. Credibility Assessment for Weld Wire Heat Number 61782 Irrad. Fluence Capsule Ni Chern. Temp. (xl 0 19n/c Fluence 0.24 0.52 161.4 545 0.5028 0.808 0.24 0.52 161.4 545 1.105 1.028 0.24 0.52 161.4 545 1.864 1.171 0.24 0.52 161.4 545 3.746 1.342 0.27 0.59 182.6 556 1.030 1.008 where Predicted ifRTNDT = (SlopebestjiJ * (Fluence Factor) and Slopebestjit = best fit line relating Adjusted ifRT NDT to the Fluence Factor (i.e., 152.8) Temp: Meas. Adjusted LlRTNDT LlRTNDT 146 142 167 163 169 165 223 219 138 145 These data points are not credible since the scatter is greater than +/-28°P for two surveillance points. 32-9019240-000 Ratio Predicted Adjusted LlRT NDT (OF) (Adjusted - LlRTNDT from Best Fit Predicted) 148.2 123.5 24.7 170.2 157.1 l3.i 172.3 179.0 -6.7 228.6 205.1 23.5 133.8 154.0 -20.2 Page 47 of 51

A AREVA NON-PROPRIETARY 32-9019240-000 Table 56. Table Chemistry Factor Non-Conservatism Assessment for Weld Wire Heat Number 61782 TableChem. Capsule Adjusted (Adjusted-Capsule Factor (S urv. Fluence L!.RTNDT Predicted Predicted) L!.RT NOT DeSlgJl8.tIon Avg.) Factor (OF) L!.RTNDT(°F"t (OF) R. E. Ginna: Capsule V 168.5 0.808 149 136.2 12.8 SA-1036: Plant Specific RVSP Material. R. E. Ginna: Capsule R 168.5 1.028 170 173.2 -3.2 SA-I036: Plant Specific RVSP Material R. E. Ginna: Capsule T 168.5 1.171 173 197.3 -24.3 SA-I036: Plant Specific RVSP Material R. E. Ginna: Capsule S 168.S 1.342 229 226.1

2.9 SA-I036

Plant Specific RVSP Materiai B&WOG: Capsule DBI-LG! 168.5 1.008 134 169.8 -3S.8 SA-I 13S: ONS-2 Nozzle Belt Dropout Mati. B&WOG: Capsule DBI-LG2 168.S 1.136 141 191.4 -SO.4 SA-l13S: ONS-2 Nozzle Belt Dropout Mati. Since the scatter for all data points is-less than 2 standard deviations (56°P), the Table chemistry factor is conservative. Page 48 of 51

A AREVA NON-PROPRIETARY Table 57. Credibility Assessment for Weld Wire Heat Number 72442[7] Cu Ni Chern. Irrad. Fluence wflo wt% Factor Tern (OF) (xlO '9 nlcrn2) 0.609 B&WOG: Capsule CR3-LGI 0.22 0.60 167.0 556 WF-67: MO 1 Nozzle Belt Oro out MatI. 1.950 B&WOG: Capsule CR3-LG2 0.22 0.60 167.0 556 WF-67: MDl Nozzle Belt Oro out Mati 1.140 EPRI: PWR-5 0.22 0.60 167.0 556 WF-67 Surv. Av. 0.22 0.60 167.0 556 where Predicted ~TNDT = (Slopebestf,J * (Fluence Factor) and Slopebestfa = best fit line relating Adjusted ART NDT to the Fluence Factor (i.e., 147.4) Meas. Fluence RTNDT Factor (OF 0.861 167 1.182 138 1.037 161 --~~---- These data points are not credible since the scatter is greater than +/-28°P for two surveillance points. 32-9019240-000 (Measured-Predicted Predicted) RTNDTfrorn tlRTNDT best fit line (OF) (OOF 126.9 40.1 174.2 -36.2 152.9 8.1 Page 49 of 51

A AREVA NON-PROPRIETARY 32-9019240-000 Table 58. Table Chemistry Factor Non-Conservatism Assessment for Weld Wire Heat Number 72442 TableChem. Capsule Adjusted (Adjusted-Capsule Factor (Surv. Fluence Lill.TNDT Predicted Predicted) ilRT NOT DeS]gnatlOn Avg.) Factor (OF) Ll.RTNDT (OF) (OF) B&WOG: Capsule CR3-LGl 167.0 0.861 167 143.8 23.2 WF-67: MDl Nozzle Belt Dropout Mati. B&WOG: Capsule CR3-LG2 167.0 1.182 138 197.4 -59.4 WF-67: MD 1 Nozzle Belt Dropout Matl EPRI. PWR-5 167.0 1:037 161 173.2 -12.2 WF-67 The above assessment results indicate that the generic Table chemistry factor for the surveillance data over-predicts the adjusted measured data. Therefore, the Table chemistry factor calculated using the weld wire heat best-estimate copper and nickel contents is considered conservative. Page 50 of 51

A AREVA NON-PROPRIETARY 32-9019240-000 7.0 References

1. U. S. Nuclear Regulatory Commission, "Radiation Embrittlement o/Reactor Vessel Materials," Regulatory Guide 1.99, Revision 2, May 1988.
2. AREVA Document 38-9008745-000, H. P. Gunawardane, "NMC Letter Dated June 29, 2004 with Westinghouse Attachment (LTR-REA-04-64), JJ December 2005.
3. AREVA Document 43-1543-04, M. J. DeVan, "Master Integrated Reactor Vessel Surveillance Program, JJ (BAW-1543-04), February 1993.
4. American Society of Mechanical Engineers Boiler and Pressure Vessel Code, Section III, Nuclear Power Plant Components, Subsection NB, Class 1 Components.
5. AREVA Document 43-2308-02, K. K. Yoon, "Initial RTNDT of Linde 80 Weld Materials, JJ (BAW-2308, Revision I-A), August 2005.
6. AREVA Document 43-2325-01,.M. J; Devan, "Response to Request for Additional Information (RAJ) Regarding Reactor Vessel Pressure Integrity, B& W Owner's Group -

Reactor Vessel Working Group, "(BAW-2325-01), January 1999.

7. AREVA Document 77-2313-005, "B&WOG Reactor Vessel Working Group, Reactor Vessel Materials and Surveillance Data Information," (BAW-2313, Revision 5),

December 2005. Page 51 of 51

ENCLOSURE 1 ATTACHMENT 2 NEXTERA ENERGY POINT BEACH, LLC POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2 SUPPLEMENT 2 TO LICENSE AMENDMENT REQUEST 252 TECHNICAL SPECIFICATION 5.6.5, REACTOR COOLANT SYSTEM (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR) RTPTS VALUES FOR POINT BEACH UNIT 1 AND UNIT 2 22 pages follow

20697-10 (3/30/06) A CALCULATION

SUMMARY

SHEET (CSS) AREVA Document Identifier 32-9019238-000 ~~~~~~---------------------- Title RTPTS Values for Point Beach Unit 1 and Unit 2 PREPARED BY: NAME S. B. Davidsaver SIGNATURE 1. 11;. Q;JJIqt,~ I TITLE Engineer II DATE $/ov REVIEWED BY: METHOD: [gj DETAILED CHECK D INDEPENDENT CALCULATION NAME J. B. Hall SIGNATURE Q P ~ TITLE Princ. E~eer DATE 6~)1 ~ 0 (, CENTER _4--'--1-'--3=-2--'4___________ PAGE(S) _2_2_______ REVIEWER INDEPENDENCE

0 COST REF.

TMSTATEMENT: ~~~ NAME 15K IY1 b,.,<A-PURPOSE AND

SUMMARY

OF RESULTS: The RTPTS values applicable to 53 EFPY for the Point Beach Unit 1 and Unit 2 reactor vessel beltline materials are presented in this document. These values were calculated in accordance with the requirements specified in the Code of Federal Regulations, Title 10, Part 50.61. The RTPTS values were calculated using projected f1uence values for 60 years. The controlling beltline material for the point Beach Unit 1 reactor vessel is the intermediate shelliongitudirial weld heat number 1 P0815 with a predicted RTPTS value of 236.0F. Screening criterion for this material is 270F. The controlling beltline material for the Point Beach Unit 2 reactor vessel is the intermediate to lower shell circumferential shell weld heat number 72442 with a predicted RTPTS value of 280.6F. Screening criterion for this material is 300F. THE FOLLOWING COMPUTER CODES HAVE BEEN USED IN THIS DOCUMENT: CODENERSION/REV CODENERSION/REV AREVA NP Inc., an AREVA and Siemens company THE DOCUMENT CONTAINS ASSUMPTIONS THAT MUST BE VERIFIED PRIOR TO USE ON SAFETY-RELATED WORK DYES I2$J NO Page_1_of~

A AREVA NON-PROPRIETARY 32-9019238-000 RECORD OF REVISIONS Revision Description Date 000 Original Release May 2006 Page 2 of22

A AREVA NON-PROPRIETARY 32-9019238-000 TABLE OF CONTENTS 1.0 Introduction............................................................................................................. 5 2.0 Summary of Results................................................................................................ 5 3.0 Assumptions............................................................................................................ 5 4.0 Reactor Vessel Fluence........................................................................................... 5 5.0 Pressurized Thermal Shock Reference Temperature Calculation Where No Surveillance Data Is Available............................................................................................ 7 5.1 Initial RTNDT....................................................................................................... 7 5.2 LlRTNDI............................................................................................................... 8 5.2.1 Chemistry Factor......................................................................................... 8 5.2.2 Fluence Factor................................ ;............................................................ 9 5.2.3 dRTND1 Calculation.................................................................................. 10 5.3 Margin............................................................................................................... 10 5.4 Calculation of Pressurized Thermal Shock Reference Temperature (RTPTs)... 11 6.0 Pressurized Thermal Shock Reference Temperature Calculation Where Surveillance Data is Available.......................................................................................... 12 6.1 Surveillance Data Credibility Assessment........................................................ 13 6.2 Credible Surveillance Data............................................................................... 14 6.3 Non-Credible Surveillance Data....................................................................... 14 6.4 Assessment of Weld Wire Heat Surveillance Data........................................... 15 7.0 References................................................................................................................... 22 Page 3 of22

A AREVA NON-PROPRIETARY 32-9019238-000 LIST OF TABLES Table 1. Point Beach Unit 1 and Unit 2 Pressurized Thermal Shock Reference Temperature Applicable to 53 EFPY.................................................................................. 6 Table 2. Fluence CE> 1.0 MeV) Values at 53 EFPY for the Point Beach Unit 1 and Unit 2 Reactor Vessel Beltline Materials....................................................................................... 7 Table 3. Initial RT NDT for the Point Beach Unit 1 and Unit 2 Reactor Vessel Beltline Materials............................................................................................................................. 8 Table 4. 10 CFR 50.61 Chemistry Factors for the Point Beach Unit 1 and Unit 2 Reactor Vessel Beltline Materials.................................................................................................... 9 Table 5. Fluence Factors Through 53 EFPY for the Point Beach Unit 1 and Unit 2 Reactor Vessel Beltline Materials....................................................................................... 9 Table 6. Lill.TNDT Values for the Point Beach Unit 1 and Unit 2 Reactor Vessel Beltline Materials........................................................................................................................... 10 Table 7. Margin Values for the Point Beach Unit 1 and Unit 2 Reactor Vessel Beltline Materials........................................................................................................................... 11 Table 8. Pressurized Thernial Shock Reference Temperature for Point Beach Unit 1 and Unit 2 Reactor Vessel Beltline Materials Through 53 EFPY........................................... 12 Table 9. Credibility Assessment for Weld Wire Heat Number 71249[6,7,8]..................... 16 Table 10. Table Chemistry Factor Non-Conservatism Assessment for Weld Wire Heat Number 71249.................................................................................................................. 17 Table 11. Credibility Assessment for Weld Wire Heat Number 61782 [6,9].................... 18 Table 12. Table Chemistry Factor Non-Conservatism Assessment for Weld Wire Heat Number 61782.................................................................................................................. 19 Table 13. Credibility Assessment for Weld Wire Heat Number 72442[10]...................... 20 Table 14. Table Chemistry Factor Non-Conservatism Assessment for Weld Wire Heat Number 72442............................................. :.................................................................... 21 Page 4 of22

A AREVA NON-PROPRIETARY 32-9019238-000 1.0 Introduction The purpose oftrus analysis is to determine the reactor vessel pressurized thermal shock* reference temperatures (RTpTs) applicable to 53 effective full power years (53 EFPy) for the Point Beach Unit 1 and Unit 2 Nuclear Power Plant using projected fluence values for 60 years. 2.0 Summary of Results The RTPTS values applicable to 53 EFPY for the Point Beach Unit 1 and Unit 2 reactor vessel beltline materials are listed in Table 1. These values were calculated in accordance with the reRuirements specified in Code of Federal Regulations, Title 10, Part 50.61 (10 CFR 50.61).[] The RTpTs values were calculated using projected fluence values for 60 years. The controlling beltline material for the Point Beach Unit 1 reactor vessel is the intermediate shell longitudinal weld heat number 1 P0815 with a predicted RTpTs value of236.0°F. Screening criterion for this material is 270°F. The controlling beltline material for the Point Beach Unit 2 reactor vessel is the intermediate to lower shell circumferential shell weld heat number 72442 with a predicted RTpTs value of 280.6°F. Screening criterion for this material is 300°F. 3.0 Assumptions No major assumptions are contained in this calculation. 4.0 Reactor Vessel Fluence The projected (53 EFPY) fluences for the Point Beach Unit 1 and Unit 2 reactor vessel beltline materials are listed in Table 2. [2] These values are at the clad-base metal interface on the inside surface of the vessel at the location where the material in question receives the highest fluence. The fluence values are for an uprate to 1678 MWth and hafnium rod removal in October 2008 for Point Beach Unit 1 and an uprate to 1678 MWth and hafnium rod removal in April 2008 Point Beach Unit 2. PageS of 22

A AREVA NON-PROPRIETARY 32-9019238-000 Table 1. Point Beach Unit l' and Unit 2 Pressurized Thermal Shock Reference Temperature Applicable to 53 EFPY 53 EFPY Fluence Chemical at Clad-Material Descri )tion Composition Initial Base Metal Reactor Vessel MatI. Heat Cu Ni Chern.

RTNDT, Interface, Fluence

~TPTs,

Margin, RTPTs, Screening Beltline Region MatI.

Ident. Number Type wt% wt% Factor OF nJcm2 Factor OF OF OF Criteria RTpTS Calculation for Unit I Nozzle Belt Forging 122P237 122P237 SA-508 Cl. 2 0.11 0.82 77.0 50 3.58E+18 0.716 55.1 34.0 l39.1 270 Intermediate Shell Plate A9811-1 A9811-1 SA-508 C1. 2 0.20 0.06 79.3* 1 4.90E+l9 1.398 110.9 56.4 168.3 270 Lower Shell Plate C1423-1 C1423-1 SA-508 CI. 2 0.12 0.07 35.8* 1 4.55E+19 1.383 49.5 56.4 106.9 270 NB to IS Circ. Weld SA-I426 '8Tl762 Linde 80 Flux 0.19 0.57 167.0 -47.6 3.58E+18 0.716 119.6 65.7 137.7 300 (100%) IS Long. Weld (10 27%) SA-812 IP0815 Linde 80 Flux 0.17 0.52 167.0 -47.6 3.19E+19 1.305 217.9 65.7 [236.0] 270 IS Long. Weld (OD 73%) SA-775 IP0661 Linde 80 Flux 0.17 0.64 167.0 -47.6 N/A N/A N/A N/A N/A 270 Intermediate to LS Circ. SA-lIOI 71249 Linde 80 Flux 0.23 0.59 167.6 -47.4 4.43E+l9 1.378 231.0 61.7 245.3 300 Weld (100%) LS Lorig. Weld (100%) SA-847 61782 Linde 80 Flux 0.23 0.52 167.0 -47.6 3.05E+l9 1.295 216.3 65.7 234.4 270 RTpTS Calculation for Unit 2 Nozzle Belt Forging 123V352 123V352 SA-508 Cl. 2 0.11 0.73 76.0 40 5.04E+lS 0.S09 61.5 34.0 l35.5 270 Intermediate Shell Forging 123V500 123V500 SA-508 Cl. 2 0.09 0.70 58.0 40 5.05E+19 1.404 81.4 34.0 155.4 270 Lower Shell Forging; 122W195 122W195 SA-508 Cl. 2 0.05 0.72 42.8* 40 4.90E+19 1.398 59.8 17.0 116.8 270 NB to IS Circ. Weld 21935 21935 Linde 1092 0.18 0.70 170.5 -56 5.04E+18 0.809 137.9 65.5 147.4 300 (100%) Flux Intermediate to LS Circ. SA-14M 72242 Linde 80 Flux 0.26 0.60 180.0 -30 4.65E+19 1.388 249.8 60.8 [280.6] 300 Weld (100%)

  • -Determined from Surveillance Data

[] - Limiting reactor vessel beltline region material in accordance with 10 CPR 50.61 Page 6 of22

A AREVA NON-PROPRIETARY 32-9019238-000 Table 2. Fluence (E> 1.0 MeV) Values at 53 EFPY for the Point Beach Unit 1 and Unit 2 Reactor Vessel BeltHne Materials 53 EFPY Fluenee, nlem2 at the CladlBase Metal Material Interface on the Inside Beltline Materials Ident. Unit Surface Nozzle Belt Forging (NB) 122P237 1 3.58E+18 Intermediate Shell Plate (IS) A9811-1 1 4.90E+19 Lower Shell Plate (LS) C1423-1 1 4.55E+l9 NB to IS Cire. Weld (100%) SA-1426 1 3.58E+l8 IS Long. Weld (ID 27%) SA-812 1 3.19E+19 IS Long. Weld (OD 73%) SA-775 1 N/A Intermediate to LS Cire. Weld (100%) SA-llOl 1 4,43E+l9 __ ~§}~~ng. W~!~_ (1 OQ~l ____________ SA-847 1 3.05E+l9 r-------------- _ Nozzle Belt Forging (NB) 123V352 2 5.04E+l8 Intennediate Shell Forging (IS) 123V500 2 5_05E+l9 Lower Shell Forging (LS) 122W195 2 4.90E+19 NB to IS Cire. Weld (100%) 21935 2 5.04E+18 Intermediate to LS Cire. Weld (100%) SA-1484 2 4.65E+l9 5.0 Pressurized Thermal Shock Reference Temperature Calculation Where No Surveillance Data Is Available The following information is required for detennination of the pressurized thermal shock reference temperature in accordance with 10 CFR 50.61. 5.1 Initial RTNDT The initial RTNDT is the reference temperature for the reactor vessel beltline material in the unirradiated condition, evaluated in accordance with Paragraph NB-2331 of Section III ofthe ASME Boiler and Pressure Vessel Code. (3J An alternative initial reference temperature for the Linde 80 beltline welds in the Babcock & Wilcox fabricated reactor vessels is specified in BAW-2308, Revision I-A, "Initial RT NDT of Linde 80 Weld Materials". [4] This report was reviewed and approved for use by the Nuclear Regulatory Commission.[5] The alternative initial RINDT values in BAW-23 08, Revision I-A were determined based on brittle-to-ductile transition range fracture toughness test data ofthese welds obtained in accordance the ASTM Standard EI92I and using ASME Boiler and Pressure Vessel Code Case N-629. Due to the generally low Charpy V -notch upper shelf energy behavior of Linde 80 welds, the testing specified in ASME Code, Section III, Paragraph NB-2331 has been shown to be overly conservative when used to predict the transition from the ductile to brittle failure in Linde 80 welds. Table 3 lists the initial RT NOT values for the Point Beach Unit 1 and Unit 2 reactor vessel beltIine materials. [4,6] Page 7 of22

A AREVA NON-PROPRIETARY 32-9019238-000 Table 3. Initial RT NDT for the Point Beach Unit 1 and Unit 2 Reactor Vessel Beltline Materials Material Initial R T NOT, Beltline Materials Ident. Unit of Reference Nozzle Belt Forging (NB) 122P237 1 50 6 Intermediate Shell Plate (IS) A981l-1 I 1 6 Lower Shell Plate (LS) C1423-1 1 I 6 NB to IS Circ. Weld (100%) SA-1426 1 -47.6 4 IS Long. Weld (m 27%) SA-SI2 I -47.6 4 IS Long. Weld (OD 73%) SA-775 I -47.6 4 Intermediate to LS Cire. Weld (100%) SA-llOI I -47.4 4 LS Long. Weld (100%) SA-S47 I -47.6 4 No~l~BeitF~lii;g-(NB) ------------- ---------------- ---------r-------------- ---------------- 12~V352 2 40 6 Intermediate Shell Forging (IS) 123V500 2 40 6 Lower Shell Forging (LS) 122WI95 2 40 6 NB to IS Cire. Weld (100%) 21935 2 -56 6 Intermediate to LS Cire. Weld (100%) SA-1484 2 -30 4 5.2 i1RTNDT L1RT NDT is the mean value of the adjustment in reference temperature caused by irradiation and is calculated as follows: where CF ff 5.2.1 Chemistry Factor MlTNDT = (CF) * (ff) = Chemistry Factor = fluence factor (1) The chemistry factor (CF) is determined from the copper and nickel content for each reactor vessel beltline region material. Using the copper and nickel contents for the Point Beach Unit 1 and Unit 2 reactor vessel beltline materials,[6] the CF is determined from Table 1 (for weld metals) and Table 2 (for base metals) in 10 CFR 50.61. Linear interpolation is permitted. When determining the CF, the "weight percent copper" and "weight percent nickel" are best estimate values for the material, which will normally be the mean of the measured values for the material. Table 4 lists the chemistry factors for the Point Beach Unit 1 and Unit 2 reactor vessel beltline materials. Page 8 of22

A AREVA NON-PROPRIETARY 32-9019238-000 Table 4. 10 CFR 50.61 Chemistry Factors for the Point Beach Unit 1 and Unit 2 Reactor Vessel Beltline Materials Material Cu Ni Chemistry Beltline Materials Ident. Unit wt% wtOlo Factor Nozzle Belt Forging (NB) . 122P237 1 O.ll 0.82 77.0 Intennediate Shell Plate (IS) A98ll-1 1 0.20 0.06 79.3* Lower Shell Plate (LS) C1423-1 1 0.12 0,07 35.8* NB to IS Cire. Weld (100%) SA-1426 1 0.19 0.57 167.0 IS Long. Weld (ID 27%) SA-812 1 0.17 0.52 167.0 IS Long. Weld (OD 73%) SA-775 1 0.17 0.64 167.0 Intennediate to LS Cire. Weld (100%) SA-llOI 1 0.23 0.59 167.6 .~.~.. ~~~j};:. W.~.~.QQQ.~2.. _____. SA-847 1 0.23 I-.Q.52.... 167.0 f--.... -... -...... -. Nozzle Belt Forging (NB) 123V352 2 0.11 0.73 76.0 Intennediate Shell Forging (IS) 123V500 2 0.09 0.70 58.0 Lower Shell Forging (LS) 122W195 2 0.05 0.72 42.8* NB to IS Cire. Weld (l00%) 21935 2 0.18 0.70 170.5 Intennediate to LS Cire. Weld (100%) SA-1484 2 0.26 0.60 180.0 .lbj - detennmed from surveIllance data 5.2.2 Fluence Factor In accordance with 10 CFR 50.61, the fluence factor Cft) is detennined as follows: ff = /(0.28-0.10 log f ) Table 5 lists the fluence factors for the Point Beach Unit 1 and Unit 2 reactor vessel beltline materials at 53 EFPY. Table 5. Fluence Factors Through 53 EFPY for the Point Beach Unit 1 and Unit 2 Reactor Vessel Beltline Materials Material Fluenee, nJem2 Fluenee Fa:torll Beldine Materials Ident. Unit (x 1019) Nozzle Belt Forging (NB) 122P237 1 3.58E+18 0.716 Intermediate Shell Plate (IS) A9811-1 1 4.90E+19 1.398 Lower Shell Plate (LS) C1423-1 I 4.55E+19 1.383 NB to IS Cire. Weld (100%) SA-1426 1 3.58E+18 0.716 IS Long. Weld (ID 27%) SA-812 1 3.19E+l9 1.305 IS Long. Weld (OD 73%) SA-775 1 NIA N/A Intermediate to LS Cire. Weld (l00%) SA-llOl 1 4.43E+19 1.378 .. LS Lo~g:.. ~~J.<U!QQ~L_. ___.. _..... SA-847 1 3.05E+19 1.295 Nozzle Belt Forging (NB) 123V352 2 5.04E+18 0.809 Intennediate Shell Forging (IS) 123V500 2 5.05E+ 19 1.404 Lower Shell Forging (LS) 122W195 2 4.90E+19 1.398 NB to IS Cire. Weld (100%) 21935 2 5.04E+18 0.809 Intennediate to LS Cire. Weld (l00%) SA*1484 2 4.65E+19 1.388 Page 9 of22

A AREVA NON-PROPRIETARY 32-9019238-000 5.2.3 ~R T NDI Calculation The ~T NDI values for the Point Beach Unit 1 and Unit 2 reactor vessel beltline materials are calculated by multiplying the chemistry factors and fluence factors. The 53 EFPY ~TNDT values for the Point Beach Unit 1 and Unit 2 reactor vessel beltline materials are presented in Table 6. Table 6. ~TNDT Values for the Point Beach Unit 1 and Unit 2 Reactor Vessel Beltline Materials Material Chemistry Fluence ~TNDr, Beltline Materials Ident. Unit Factor Factor of Nozzle Belt Forging (NB) 122P237 1 77.0 0.716 55.1 Intennediate Shell Plate (IS). A9811-1 1 79.3 1.398 110.9 Lower Shell Plate (LS) C1423-1 1 35.8 1.383 49.5 NB to IS Circ. Weld (100%) SA-1426 1 167.0 0.716 119.6 IS Long. Weld (10 27%) SA-812 1 167.0 1.305 217.9 IS Long. Weld (OD 73%) SA-775 1 167.0 N/A N/A Intennediate to LS Circ. Weld (100%) SA-11 01 1 167.6 1.378 231.0 ~S Long. y{~!~_Q Oo~t ________________ SA-847 1 167.0 1.295 216.3 Nozzle Belt Forging (NB) 123V352 2 76.0 0.809 61.5 Intennediate Shell Forging (IS) 123V500 2 58.0 1.404 81.4 Lower Shell Forging (LS) 122W195 2 42.8 1.398 59.8 NB to IS Circ. Weld (100%) 21935 2 170.5 0.809 137.9 Intermediate to LS Circ. Weld (100%) SA-1484 2 180.0 1.388 249.8 5.3 Margin The "margin" is the quantity that is added to obtain conservative, upper-bound values of the adjusted reference temperature. The margin is determined by the following expression: Margin = 2~ cr; + a! where crr = standard deviation for the initial RT NDT cr~ = standard deviation for ~RTNDI (3) If a measured value of initial RT NDT for the material in question is available, crI is to be estimated from the precision of the test method. If generic mean values are used, crI is the standard deviation obtained from the set of data used to establish the mean. The standard deviation for ~RTNDT, cr~, is 28°F for welds and 17°F for base metals, except that a~ need not exceed 0.50 times the mean value of ~RT NOT. For cases in which the results from a credible plant-specific surveillance program are used, the value of cr~ to be used is 14°F for welds and 8.5°F for base metal; the value of need not exceed one-half ofRTNDT. Page 10 of22

A AREVA NON-PROPRIETARY 32-9019238-000 Table 7 lists the margin values calculated for the Point Beach Unit 1 and Unit 2 reactor vessel beltline materials through 55 EFPY. Table 7. Margin Values for the Point Beach Unit 1 and Unit 2 Reactor Vessel BeltIine Materials Material BeltIine Material Ident. Unit crT crt> <'ill. T NDT /2 Margin Nozzle Belt Forging (NB) 122P237 1 0 17.0* 27.55 34.0 Intermediate Shell Plate (IS) A9811-1 I 26.9 8.5* 55.45 56.4 Lower Shell Plate (LS) C1423-1 1 26.9 8.5* 24.75 56.4 NB to IS Cire. Weld (100%) SA-1426 1 17.2 28* 59.80 65.7 IS Long. Weld (10 27%) SA-812 1 17.2 28* 108.95 65.7 IS Long. Weld (OD 73%) SA-775 1 N/A N/A N/A N/A Intermediate to LS Cire. Weld (100%) SA-II01 1 12.9 28* 115.50 61.7 ~!":-_~E~:..~eld (~ OO~L

SA-847 1

17.2 28* 108.15 65.7 1-23V3-S2-- ----------- /----------- ---------- 1---30.75------ ----------.. --- Nozzle Belt Forging (NB) 2 0 17.0* 34.0 Intennediate Shell Forging (IS) 123V500 2 0 17.0* 40.70 34.0 Lower Shell Forging (LS) 122W195 2 0 8.5* 29.90 17.0 NB to IS Cire. Weld (100%) 21935 2 17.0 28* 68.95 65.5 Intennediate to LS Cire. Weld (100%) SA-1484 2 11.9 28* 124.90 60.8

  • -Used to calculate margin term.

5.4 Calculation of Pressurized Thermal Shock Reference Temperature (RTm), The RTpTs is given by the following expression: RT PTS = Initial RT NDT + I1RT NDT + Margin (4) Table 8 lists the RTpTs calculated for the Point Beach Unit 1 and Unit 2 reactor vessel beltline materials through 55 EFPY. Page 11 of22

A AREVA NON-PROPRIETARY 32-9019238-000 Table 8. Pressurized Thermal Shock Reference Temperature for Point Beach Unit 1 and Unit 2 Reactor Vessel Beltline Materials Through 53 EFPY Material Initial ~RTNDr,

Margin, RTpys, Beltline Materials Ident.

Unit RTNDhoF of of of Nozzle Belt Forging (NB) 122P237 1 50 55.1 34.0 139.1 Intermediate Shell Plate (IS) A9811-1 1 1 110.9 56.4 168.3 Lower Shell Plate (LS) C1423-1 1 1 49.5 56.4 106.9 NB to IS Cire. Weld (100%) SA-1426 1 -47.6 119.6 65.7 137.7 IS Long. Weld (ID 27%) SA-812 1 -47.6 217.9 65.7 236.0 IS Long. Weld (OD 73%) SA-775 1 -47.6 N/A N/A N/A Intermediate to LS Cire. Weld (100%) SA-lIOl 1 -47.4 231.0 61.7 245.3 SA-847 1 -47.6 216.3 65.7 234.4 _~§ __ !:~~FL!" el~_.Q_QQ~l ________________ -i23v352 --------------- ----------- ------------1----------- ----i3-5.5 -- Nozzle Belt Forging (NB) 61.5 2 40 34.0 Intermediate Shell Forging (IS) 123V500 2 40 81.4 34.0 Lower Shell Forging (LS) 122W195 2 40 59.8 17.0 NB to IS Cire. Weld (100%) 21935 2 -56 137.9 65.5 Intermediate to LS Cire. Weld (100%) SA-1484 2 -30 249.8 60.8 6.0 Pressurized Thermal Shock Reference Temperature Calculation Where Surveillance Data is Available 155.4 116.8 147.4 280.6 To verify that the RTPTs for each vessel beltline material is a bounding value for the reactor vessel, plant specific information shall be considered. This information includes, but is not limited to, the reactor vessel operating temperature and surveillance program results. The results from the plant-specific surveillance program must be integrated into the RTPTS estimate if the plant-specific surveillance data has been deemed credible as judged by the following criteria:

1.

The materials in the surveillance capsules must be those which are the controlling materials with regard to radiation embrittlement.

2.

Scatter in the plots of Charpy energy versus temperature for the irradiated and unirradiated conditions must be small enough to permit the determination ofthe 30 ft-lb temperature unambiguously.

3.

Where there are two or more sets of surveillance data from one reactor, the scatter of .1RT NDT values must be less than 28°F for welds and 17°F for base metal. Even if the range in the capsule fluences is large (two or more orders of magnitude); the scatter may not exceed twice those values.

4.

The irradiation temperature of the Charpy specimens in the capsule must equal the vessel wall temperature at the cladding/base metal interface within +/-25°P. Page 12 of22

A AREVA NON-PROPRIETARY 32-9019238-000

5.

The surveillance data for the correlation monitor material in the capsule, if present, must fall within the scatter band of the data base for the material. The surveillance data deemed credible according to the criteria specified above must be used to determine a material-specific value of CF for use in the following equation: MlTNDT = CF

  • ff A material-specific value of CF is determined from the following equation:

where: n Ai fii n LIA/

  • ffJl CF = -=-1=-=-1 ___

n Lff/

=}

= number of surveillance data points = measured value of llRTNDT = fluence factor for each surveillance data point. (5) (6) For cases in which the results from a credible plant-specific surveillance program are used, the value of 0'/\\ to be used is 14°F for welds and 8.5°F for base metals; however the value of 0'/\\ may not exceed one-half ~RTNDT. The base metal evaluations done in reference 6 for base metals pertinent to this document are still valid. There have been 6 RVSP Capsule reports reporting Charpy shift data for Linde 80 welds since reference 6 was issued in 1999. The applicable Linde 80 welds for Point Beach Unit 1 and Unit 2 are evaluated in Section 6.4. 6.1 Surveillance Data Credibility Assessment Temperature Adjusted ART NDT, normalked = ART NDT, measured + 1.0 * (Tcap.lUle - Tcapsule mean) (7) For the reactor vessels, numerous surveillance data are available for evaluation of irradiation embrittlement. When assessing credibility for surveillance data from several sources, the capsule data may have to be adjusted to account for the irradiation environment and chemical composition differences. Additionally, if the surveillance data are from multiple sources, it is necessary to adjust the capsule data for chemical composition (copper and nickel content) differences. For the credibility determination, the Page 13 of22

A AREVA NON-PROPRIETARY 32-9019238-000 surveillance data are normalized to the mean copper and nickel contents of the surveillance materials using the following equation: ( eF ) Table,Surv.Avg.Chem.

  • Ratio Adjusted ART NDT, normalized -

ART NDT, measured CFTable, Surv. Chem. (8) A best-fit line (least squares regression) is then determined from the adjusted ~RTNDI capsule surveillance data as a function ofthe capsule fluence factor. The data are considered credible if the difference between the adjusted ~RTNDI (i.e., chemistry adjusted) and the predicted ~RTNDT (from the best-fit line) for all the data is within +/-28°F for weld metals and +/-17°F for base metals. 6.2 Credible Surveillance Data In accordance with Regulatory Guide 1.99, Revision 2 and 10 CFR 50.61, credible surveillance data are used to determine material-specific chemistry factor values for use in reactor vessel integrity assessments. The chemistry factor is determined from a best-fit line through the surveillance data adjusted to account for differences in chemical composition (Le., copper and nickel contents) and irradiation environment (Le., irradiation temperature) between the capsules and the vessel. The surveillance data are adjusted in the same manner as for the credibility determination except that the 30 ft-lb transition temperature values are normalized to the best estimate copper and nickel contents and the irradiation temperature of the vessel being assessed. 6.3 Non-Credible Surveillance Data If the surveillance data are determined to be non-credible, the chemistry factor value is calculated from the generic Tables in 10 CFR 50.61 and Regulatory Guide 1.99, Revision 2 unless the chemistry factor determined from the surveillance data is significantly greater than that from the generic Tables, indicating that the Table chemistry factor is non-conservative. To determine if the generic Table chemistry factor is non-conservative, the following steps are performed:

1.

Determine the chemistry f~ctor from the generic Tables based on the surveillance specimen chemical composition; use this chemistry factor to determine the predicted ~RT NDT for each capSUle: (Predicted ART NDT = CF Table, Surv. Avg. Chem. * /lcapsule)

2.

Determine difference between the predicted ~RTNDI and the measured ~RTNDT. Page 14 of 22

A AREVA NON-PROPRIETARY 32-9019238-000 If the difference between the predicted Llli.TNDT and the measured i1RTNDT values exceeds 2 standard deviations (i.e., 56°P for weld metals and 34°P for base metals), the Table chemistry factor is considered non-conservative. When the Table chemistry factor is determined to be non-conservative, the chemistry factor determined from the "non-credible" surveillance data is used in the assessment of reactor vessel integrity using the "full" value of (J~ in calculating the Margin term. 6.4 Assessment of Weld Wire Heat Surveillance Data The following tables provide the credibility assessment of the weld wire heats 71249, 61782, and 72442. NOTE: The original Charpy V -notch impact data are based on hand-fit Charpy curves using engineering judgment; these data were using a hyperbolic tangent curve fitting program to achieve consistency in the interpretation ofthe available surveillance test data. Page 15 of22

A AREVA NON-PROPRIETARY Table 9. Credibili Assessment for Weld Wire Heat Number 71249[6,7,8) Irrad. Capsule Chern. Temp. Desi nation Cu wt"10 Ni wt"lo Factor (OF Fluence Fluence (xl019n1cm2) Factor Turkey Point Umt 3: Capsule T 0.33 0.57 201.3 546 SA-lIOl: Plant S ecific RVSP Material 0.739 0.915 Turkey Point Unit 3: Capsule V 0.33 0.57 201.3 546 SA-I \\0 I: Plant S ecific RVSP Matenal 1.530 1.118 Turkey Point Unit 4: Capsule T 0.29 0.60 191.0 546 SA-I 094: Plant S ecific RVSP Material 0.708 0.903 Turkey Pomt Umt 3: Capsule X 0.33 0.57 201.3 546 SA-II 0 I: Plant S ecific RVSP Matenal 2.900 1.283 B&WOG: Capsule AS 0.33 0.57 201.3 551 SA-ll 0 I: TP-3 Plant S ecific RVSP MatI. 2.572 1.253 Surv. Avg. 0.322 0.576 199.2 547 where Predicted jjRTNDT = (Slopebestfzd * (Fluence Factor) and Slopebestflt = bestfit line relating Adjusted jjRTNDTto the Fluence Factor (i.e., 172.8) Temp. Meas. Adj. 8RTNDT LlRTNDT (OF) (0 166 165 179 178 211 210 191 190 215 219 These data points are not credible since the scatter is greater than +/-28°F for two surveillance points. ~. _... Ratio Adj. 8RTNDT (OF) 163.3 176.2 219.0 188.0 216.7 32-9019238-000 Pred. LlRTNOT (OF) (Adjusted - from best fit Predicted) line llRT NOT (OF) 158.1 5.2 193.1 -16.9 156.0 63.0 221.6 -33.6 216.5 0.2 Page 16 of 22

A AREVA NON-PROPRIETARY 32-9019238-000 Table 10. Table Chemistry Factor Non-Conservatism Assessment for Weld Wire Heat Number 71249 TableChem. Capsule Adjusted (Adjusted-Capsule Factor (Surv. Fluence Iill.TNDT Predicted Predicted) Iill.T NOT DesIgnation Avg.) Factor (OF) ilRTNOr (OF) (oF) Turkey Pomt DOlt 3: Capsule T 199.2 0.915 163.3 182.5 -19.2 SA-I 10 I: Plant Specific RVSP Matenal Turkey Point DOlt 3: Capsule V 1992 l.U8 176.2 222.7 -46.5 SA-110I: Plant Specific RVSP Material Turkey Point DOlt 4: Capsule T 199.2 0.903 219.0 179.9 39.1 SA-I 094: Plant Specific RVSP Material Turkey Point DOlt 3: Capsule X 199.2 1.283 188.0 255.6 -67.6 SA-I 101: Plant Specific RVSP Matenal B&WOG: Capsule AS 199.2 1.253 216.7 249.6 -32.9 SA-I 101: TP-3 Plant Specific RVSP Mat!. The above assessment results indicate that the generic Table chemistry factor for the surveillance data grossly over-predicts the adjusted measured data. Therefore, the Table chemistry factor calculated using the weld wire heat best-estimate copper and nickel contents is considered conservative. Page 17 of 22

A AREVA NON-PROPRIETARY Table 11. Credibility Assessment for Weld Wire Heat Number 61782 [6,9) Irrad. Fluence Capsule Ni Chern. Temp. (xl019n!c Fluence Factor 0.24 0.52 161.4 545 0.5028 0.808 0.24 0.52 161.4 545 1.105 1.028 0.24 0.52 161.4 545 1.864 1.171 0.24 0.52 161.4 545 3.746 1.342 0.27 0.59 182.6 556 1.030 1.008 where Predicted ARTNDT = (Slopebestf,J * (Fluence Factor) and Slopebestflt = best fit line relating Adjusted ART NDT to the Fluence Factor (i.e., 153.1) Meas. Adjusted L'.RTNDT L'.RTNDT 146 142 167 163 169 165 223 219 138 145 These data points are not credible since the scatter is greater than +/-28°F for two surveillance points. Adjusted ARTNDT 149 170 173 229 134 32-9019238-000 Predicted L'.RTNDT (OF) (Adjusted - from Best Fit Predicted) Line L'.RT 124 25 157 13 179 -7 205 23 154 -20 Page 18 of22

A AREVA NON-PROPRIETARY 32-9019238-000 Table 12. Table Chemistry Factor Non-Conservatism Assessment for Weld Wire Heat Number 61782 Capsule Des! ation R. E. Oinna: Capsule V 168.5 0.808 149 136.1 12.9 SA-1036: Plant S ecific RVSP Matenal R. E. Oinna: Capsule R 168.5 1.028 170 173.2 -3.2 SA-I036: Plant S ecific RVSP Matenal R. E. Oinna: Capsule T 168.5 1.171 173 197.3 -24.3 SA-I036: Plant S ecific RVSP Material . R. E. Oinna: Capsule S 168.5 1.342 229 226.1

2.9 SA-I036

PlantS ecificRVSP Matenal B&WOO; Capsule DBI-LOI 168.5 1.008 134 169.8 -35.8 SA-l 135: ONS-2 Nozzle Belt Dro out Mati. B&WOO: Capsule DBI-L02 168.5 1.136 141 191.4 -50.4 SA-II3S: ONS-2 Nozzle Belt Dropout Mati. Since the scatter for all data points is less than 2 standard deviations (56°F), the Table chemistry factor is conservative. Page 19 of22

A AREVA NON-PROPRlETARY 32-9019238-000 Table 13. Credibility Assessment for Weld Wire Heat Number 72442[10J I Cu Ni Chern. Irrad. Fluence C<lj)sule Des~ation wt"10 wt"10 Factor Temp (oF) (xlO t?nlcm'2j B&WOO: Capsule CR3-LOI 0.22 0.60 167.0 556 0.609 Wf-67L: MDI Nozzle Belt DfQQout Mati. B&WOO: Capsule CR3-LCJI 0.22 0.60 167.0 556 1.950 WF-67L: MD I Nozzle Belt Dropout MatI B&WOO: PWR-5 0.22 0.60 167.0 556 Ll40 WF*67 Surv. AVI!:. 0.22 0.60 167.0 556 where Predicted.t1RT NDT = (SlopebestpJ * (Fluence Factor) and Slopebestjit = best fit line relating Adjusted LfRT NDT to the Fluence Factor (i.e., 147.4) Meas. Fluence RTNDT Factor (OF) 0.861 167 Ll82 138 1.037 161 These data points are not credible since the scatter is greater than +/-28°F for two surveillance points. f'*"*----"- (Measured-Predicted Predicted) RTNDTfrom t.RTNDT best fit line (oF) (OOF) 126.9 40.1 174.2 -36.2 152.9 8.1 Page 20 of22

A AREVA NON-PROPRIETARY 32-9019238-000 Table 14. Table Chemistry Factor Non-Conservatism Assessment for Weld Wire Heat Number 72442 TableChem. Capsule Adjusted (Adjusted-Capsule Factor (Surv. Fluence 6.RTNDT Predicted Predicted) MT NOT Deslgnabon Avg.) Factor (OF) ~RTNDT (OF) (OF) B&WOG: Capsule CRJ-LGI 167.0 0.861 167 143.8 23.2 WF-67L MDI Nozzle Belt Dropout MatI. B&WOG: Capsule CRJ-LGl 167.0 1.182 138 197.4 -59.4 WF-67L MD I Nozzle Belt Dropout Mati B&WOG: PWR-5 167.0 1.037 161 173.2 -12.2 WF-67 The above assessment results indicate that the generic Table chemistry factor for the surveillance data grossly over-predicts the adjusted measured data. Therefore, the Table chemistry factor calculated using the weld wire heat best-estimate copper and nickel contents is considered conservative. Page 21 0(22

A AREVA NON-PROPRIETARY 32-9019238-000 7.0 References

1. Code of Federal Regulations, Title 10, "Domestic Licensing of Production and Utilization Facilities,"

Part 50.61, "Fracture Toughness Requirements for Protection Against Pressurized Thennal Shock," Effective Date: August 28, 1996.

2. AREVA Document 38-9008745-000, "NMC Letter Dated June 29, 2004 with Westinghouse Attachment," December 2005.
3. American Society of Mechanical Engineers Boiler and Pressure Vessel Code, Section III, Nuclear Power Plant Components, Subsection NB, Class 1 Components.
4. AREVA Document 43-2308-02, "Initial RTNDT of Linde 80 Weld Materials," (BAW-2308 Rev. I-A),

August 2005.

5. Nuclear Regulatory Commission, "Safety Evaluation for Topical Report BAW-2308, Revision 1, 'Initial RTNDT of Linde 80 Weld Materials' (TAC No. MB6636)," August 2005.
6. AREVA Document 77-2355-01, "B&WOG Reactor Vessel Working Response to Request for Additional Information (RAI) Regarding Reactor Pressure Vessel Integrity," (BAW-2355 Rev. 1), January 1999.
7. WCAP-15916, "Analysis of Capsule X from the Florida Power and Light Company Turkey Point Unit 3 Reactor Vessel Radiation Surveillance Program," September 2002, NRC public document number ML022940497.
8. AREVA Document 77-2350-00, "Test Results of WI Capsule, B&WOG Owner's Group, Master Integrated Reactor Vessel Surveillance Program," (BAW-2350), April 1999.
9. AREVA Document 43-2486-000, "Analysis of the B&W Owner's Group Capsule DBI-LG2," (BAW-2486), December 2005.
10. AREVA Document 77-2313-005, "B&WOG Reactor Vessel Working Group, Reactor Vessel Materials and Surveillance Data Information," (BA W -2313, Revision 5), December 2005.

Page 22 of22

ENCLOSURE 2 NEXTERA ENERGY POINT BEACH, LLC POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2 SUPPLEMENT 2 TO LICENSE AMENDMENT REQUEST 252 TECHNICAL SPECIFICATION 5.6.5, REACTOR COOLANT SYSTEM (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR) MARK-UP OF TRM 2.2 PRESSURE TEMPERATURE LIMITS REPORT 16 pages follow

POINT BEACH NUCLEAR PLANT TRM 2.2 TECHNICAL REQUIREMENTS MANUAL PRESSURE TEMPERATURE LIMITS REPORT Note: Applicability limits for pressure temperature limits are discussed in Section 2.0, "Operating Limits." 1.0 RCS PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR) This RCS Pressure and Temperature Limits Report (PTLR) for Point Beach Nuclear Plant Units 1 and 2 has been prepared in accordance with the requirements of Technical Specification 5.6.5. The analytical methods used to determine the RCS pressure and temperature limits shall be those previously reviewed and approved by the NRC; specifically those described in NRC Safety Evaluations dated October 6,2000, July 23,2001,---aRG October 18, 2007, and XXXXXXX.. The PTLR shall be provided to the NRC upon issuance for each reactor vessel fluence period and for any revision or supplement thereto (Ref 5.19). Based upon fluence values in Westinghouse report LTR REA 08 144 WCAP-16983-P (Ref 5.15), this PTLR is effective for ~ 50 EFPY (approximately June 2014 2029). (Ref 5.8) The Technical SpeCifications addressed in this report are listed below: 1.1 3.4.3 Pressure/Temperature (P-T) Limits 1.2 3.4.12 Low Temperature Overpressure Protection (L TOP) System 2.0 OPERATING LIMITS The cycle-specific parameter limits for the specifications listed in Section 1.0 are presented in the following subsections. Changes to these limits must be developed using the NRC approved methodologies specified in Technical Specification 5.6.5. These limits have been determined such that applicable limits of the safety analysis are met. Items that appear in capitalized type are defined in Technical Specification 1.1, "Definitions." 2.1 RCS Pressure and Temperature Limits (LCO 3.4.3) 2.1.1 The RCS temperature rate-of-change limits are:

a. A maximum heatup rate of 100°F in anyone hour.
b. A maximum cooldown rate of 100°F in anyone hour.
c. An average temperature change of :::;1 O°F per hour during inservice leak and hydrostatic testing operations.

2.1.2 The RCS P-T limits for heatup and cooldown are specified by Figures 1 and 2, respectively. (Ref 5.2) POINT BEACH TRM 2.2 -1 REV. 8 MARK-UP

POINT BEACH NUCLEAR PLANT TECHNICAL REQUIREMENTS MANUAL PRESSURE TEMPERATURE LIMITS REPORT TRM 2.2 2.1.3 The minimum temperature for pressurization or bolt up, using the methodology, is 60°F, which when corrected for possible instrument uncertainties is a minimum indicated RCS temperature of 78°F (as read on the RCS cold leg meter) or 70°F using the hand-held, digital pyrometer. 2.2 Low Temperature Overpressure Protection System Enable Temperature (LCO 3.4.6.3.4.7.3.4.10 and 3.4.12) The enable temperature for the Low Temperature Overpressure Protection System is 285°F (includes instrument uncertainty for RCS Tc wide range). (Ref 5.4) 2.3 Low Temperature Overpressure Protection System Setpoints (LCO 3.4.12) Pressurizer Power-Operated Relief Valve Lift Setting Limits The limiting trip setpoint (Ref 5.26) for the pressurizer power-operated relief valves (PORVs) is :::::420 psig (includes instrument uncertainty). The following operating restrictions ensure continued operability of the L TOP system: 2.3.1 RCP Operating Restriction - No more than one RCP in operation for RCS temperature <180°F. (Ref 5.20 to 5.24) 2.3.2 Charging Pumps - Limit the number of operating charging pumps to two when L TOP is in service. (Ref 5.20 to 5.24) 2.4 Criticality and Hydrostatic Leak Test Limits 2.4.1 Criticality and hydrostatic leak test limits are shown on the RCS Pressure Temperature Limits for heatup, Figure 1. (Ref 5.2) 3.0 REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM The reactor vessel material irradiation surveillance specimens shall be removed and examined to determine changes in material properties. The removal schedules for Units 1 and 2 are provided in Tables 1 and 2, respectively. For the period of the renewed facility operating license, all capsules in the reactor vessel that are removed and tested shall meet the test procedures and reporting requirements of ASTM E 185-82. Any changes to the capsule withdrawal schedule, including spare capsules, shall be approved by the NRC prior to implementation. (Ref 5.16 and 5.17) POINT BEACH TRM 2.2 -2 REV. 8 MARK-UP

POINT BEACH NUCLEAR PLANT TRM2.2 TECHNICAL REQUIREMENTS MANUAL PRESSURE TEMPERATURE LIMITS REPORT The pressure vessel surveillance program is in compliance with Appendix H to 10 CFR 50, entitled, "Reactor Vessel Radiation Surveillance Program." The material test requirements and the acceptance standard utilize the nil-ductility temperature, RT NOT, which is determined in accordance with ASTM E208. The empirical relationship between RT NOT and the fracture toughness of the reactor vessel steel is developed in accordance with Appendix G, "Protection Against Non-Ductile Failure," to Section XI of the ASME Boiler and Pressure Vessel Code. The surveillance capsule removal schedule meets the requirements of ASTM E185-82. Surveillance specimens for the limiting materials for the PBNP reactor vessels are not included in the plant specific surveillance program. Therefore, the results of the examinations of these specimens do not meet the credibility criteria of Regulatory Guide 1.99, Revision 2, for PBNP Units 1 and 2. During the period of extended operation, reactor vessel surveillance capsules will be removed and tested in accordance with the schedule contained in the most recently NRC-approved Pressurized Water Reactor Owners Group (PWROG) Master Integrated Reactor Vessel Surveillance Program (MIRVSP) Document. (Ref. 5.5)( Ref 5.25) 4.0 SUPPLEMENTAL DATA INFORMATION The limiting RT PTS values for the PBNP limiting beltline materials at ~ 50 EFPYare: Unit 1 - Intermediate to Lower Shell Circ Weld = ~ 236.0 OF; Lower Shell Axial Weld = ~ 234.4 OF (Ref. 5.8, AttaohmentA 5.8 Table 1 Unit 2 - Intermediate to Lower Shell Circ Weld = 29§.4 280.6 °F~termediate Shell Forging = ~ 155.4 OF (Ref. 5.8, /\\tlaohmenl A 5.8 Table 1) POINT BEACH TRM 2.2 -3 REV. 8 MARK-UP

POINT BEACH NUCLEAR PLANT TRM 2.2 TECHNICAL REQUIREMENTS MANUAL PRESSURE TEMPERATURE LIMITS REPORT

5.0 REFERENCES

5.1 WCAP-14040-NP-A, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves," Revision 2, January 1 996 4, May 2004 5.2 VI/CAP 15976, WCAP-16669, "Point Beach Units 1 and 2 Heatup and Cooldown Limit Curves for Normal Operation," Revision 1, Marsh 2008 January 2009 5.3 WEPCO Calculation Addendum No. 98-0156-00-A, Revision 0, "Evaluation of New Surveillance Data on Chemistry Factor for Weld Wire Heat 61782, Point Beach Unit 1," 9/22/1999 5.4 Westinghouse Letter VVEP 08 25, "Transmittal of LTOPS Setpoint Evaluation, " dated March 14,2008 Low Temperature Overpressure Protection System (L TOPS) Setpoint Analysis, January 2007 5.5 PWR Owner Group Topical Report BAW-1543(NP), Revision 4, Supplement 6-A, "Supplement to the Master Integrated Reactor Vessel Surveillance Program" (TAC No. MC9608), June 2007 5.6 BAW-2325, "Response to Request for Additional Information (RAI) Regarding Reactor Pressure Vessel Integrity," May 1998 5.7 CEOG Report "Best Estimate Copper and Nickel Values in CE Fabricated Reactor Vessel Welds," CE NPSD-1039, Revision 2, Final Report, June 1997 5.8 Westinghouse Letter LTR PCAM 08 57, "Point Beach Units 1 and 2 EPU P T Limit Curve Applicability Determination and Related Calculations," dated December 2008 Areva Calculation 32-9019238-000, RTPTS Values for Point Beach Unit 1 and Unit 2, Revision 0, May 2006 5.9 ASME B&PVC Code Case N-641, "Alternative Pressure-Temperature Relationship and Low Temperature Overpressure Protection System Requirements, Section XI, Division 1" 5.10 NRC Letter, "Point Beach Nuclear Plant, Units 1 and 2 - Exemption from the Requirements of 10CFR50.60 (TAC NOS. MA9680 and MA9681)," dated October 6, 2000

5. 11 NRC Letter, "Point Beach Nuclear Plant, Units 1 and 2 - Acceptance of Methodology for Referencing Pressure Temperature Limits Report (TAC Nos. MA8459 and MA8460)," dated July 23,2001
5. 12 NRC Letter, "Point Beach Nuclear Plant, Units 1 and 2 - Issuance of Amendments RE: The Conversion to Improved Technical Specifications (TAC Nos. MA7186 and MA7187)," dated August 8,2001 5.13 Deleted Areva Calculation 32-9019240-000, ART Values for Point Beach Unit 1 and Unit 2, Revision 0, May 2006 POINT BEACH TRM 2.2 -4 REV. 8 MARK-UP

POINT BEACH NUCLEAR PLANT TECHNICAL REQUIREMENTS MANUAL PRESSURE TEMPERATURE LIMITS REPORT TRM 2.2 5.14 NRC SE "Amendment Nos. 229/234 to Facility Operating Licenses DPR-24 and DPR-27, (approving use of FERRET Code as approved methodology for determining RCS pressure and temperature limits)," dated October 18, 2007 5.15 VVestinghouse Letter LTR REA 08 144, "Summary of Neutron Fluence Evaluations for the Point Beach Units 1 and 2 Extended Power Uprate," dated January 2009 WCAP-16983-P, Point Beach Units 1 and 2 Extended Power Uprate (EPU) Engineering Report 5.16 Renewed Facility Operating License DPR-24, Point Beach Nuclear Plant Unit 1 5.17 Renewed Facility Operating License DPR-27, Point Beach Nuclear Plant Unit 2 5.18 Deleted 5.19 Root Cause Evaluation 01092944, "Apparent Non-compliance with TS 5.6.5.c," Corrective Action to Prevent Recurrence (CATPR) 2 Root Cause (RC)2. 5.20 CL 4C, Low Temperature Overpressurization Protection Unit 1 5.21 CL 4C, Low Temperature Overpressurization Protection Unit 2 5.22 OP 3C, Hot Standby to Cold Shutdown 5.23 OP 4B, Reactor Coolant Pump Operation 5.24 OP 1A, Cold Shutdown to Hot Standby 5.25 NextEra Point Beach Letter, "Reactor Vessel Surveillance Program Request to Change Reactor Vessel Surveillance Specimen Withdrawal Schedule," dated January 19, 2010 5.26 Point Beach Nuclear Plan Design Guide DG-101, Instrument Setpoint Methodology POINT BEACH TRM 2.2 -5 REV. 8 MARK-UP

Ci

  • iii Co -

<II...

l III III f!!

a.. (jj III III <II > "iij

l -

(J <C POINT BEACH NUCLEAR PLANT TECHNICAL REQUIREMENTS MANUAL PRESSURE TEMPERATURE LIMITS REPORT REPLACE FIGURE WITH FIGURE 5-7 OF WCAP-16669-NP Figure 1 RCS PRESSURE-TEMPERATURE LIMITS FOR HEATUP PBNP 100°F/hr Heatup Limits 2500~~~~~~===s~==~======~--~========~--~r---~------~~~--------1 HEATUP LIMIT USE: - Unacceptable Operation above and/or to of curves Leak Test Limit - Acceptable Operation below and/or to right 0 es

  • L TOP Setpoint includes corrections for instrument 2000+-----------~-----------r----~----_+----------~~----~~~------~~--r_--------~

1500 1000 500+-~--------4_~--------~----------_r----------_+----~--~_1----------~r_--------__; Maximum LTOP Setpoint = 420 psig with Rep restriction 100 150 200 250 300 350 Moderator Temperature (OF) POINT BEACH TRM 2.2 - 6 REV. 8 MARK-UP TRM 2.2

MATERlAL PROPERTY BASIS LIMITING MATERIAL: Intermediate Shell Longitudinal Welds SA-8I2 (ID) and SA-775 (OD) LIMITING ART VALUES AT 53 EFPY (Hafuium Removal):

1I4T, 220.0°F 2500 2250 2000 1750 0-

~ 1500 e

l lZ e 1250 0-1

'5 ~ 1000 (J 750 500 250 o 3/4T, 184.6°F , Oparllm V"rsion:5,2 Run:17762 Operllm.xls Version: 5.21 J I Leak Test Llml!J..- ~Crl!lCal Limltl 60 Oeg. F/HrJ ........ ti Critical Limit.! !------ -I Unacceptable r ~ 100 De9. FIHr Operatlon 1--' 1--- j j Acceptable ! "I Heatup Rat~ I I Operation 60 Deg',F/Hr H Heatup Rate I 100 De9* F/H:h~ ~ 'l Criticality Limit based on , inservice hydrostatic test -- temperature (270 F) for the service period up to 53 EFPY I;;t Boltup Temp 60°F o 50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Deg. F)

  • 5-2]

Figure 5-7 Point Beach Units 1 and 2 Reactor Coolant System JIeatup Limitations (Heatup Rates of 60 and 1000F/hr) Applicable for 53 EFPY (with Hafnium Removal and without Margins for Instrumentation Errors) Using 1998 App. G Methodology (wIKle) WCAP-16669-NP January 2009 Revision 1

Ci Ui Co - POINT BEACH NUCLEAR PLANT TECHNICAL REQUIREMENTS MANUAL PRESSURE TEMPERATURE LIMITS REPORT REPLACE FIGURE WITH FIGURE 5-8 OF WCAP-16669-NP Figure 2 RCS PRESSURE-TEMPERATURE LIMITS FOR COOLDOWN PBNP 100°F/hr Cooldown Limits 2500~~~~~~========~====~---r---------r--------~------~~---------r - Unacceptable Operation abo dlor to left of curves - Acceptable Operation below andfo right of curves - L TOP Setpoint includes corrections for I rument uncertainty 2000+-----------~--------~~---------+----------~----~~~~----------+---------~ ~ 1500+-----------~--------~-----------+~~~~--_hF_----_+--~----------+_--------_4

I II)

II) ~ D.. a; II) II) L TOP Enable Temperature = 285 0 F ~ 1000+-----------~--------~----~----~~~------~--~---+--~----------+---------_4 iii

I

~ 500+-~------~--~------~----------+---------~------~--~--~~----~--------~ 100 150 POINT BEACH TRM 2.2 -7 Maximum L TOP Setpoint = 420 psig with Rep restriction 200 250 Moderator Temperature (OF) REV. 8 MARK-UP 300 350 TRM 2.2

MATERIAL PROPERTY BASIS LIMITING MATERIAL: Intennediate Shell Longitudinal Welds SA-812 (ill) and SA-775 (OD) LIMITING ART VALUES AT 53 EFPY (Hafnium Removal):

1I4T, 220.0°F 3/4T, 184.6°F 2500 2250 2000 1750 en 1500 e:.

e!

J rn rn e 1250 Q.

~ 'E u 1000 ~ 750 500 250 o IOperlim Vers/on:5.2 Run:17762 Operlim.xls VersIon: 5.2 I -J..-..--. I Unacceptable I Operation ~ I Acceptable I J Operation I------ II, Cooldown Rates F/Hr steady*state I

  • 20

-40 -60 I-- !oo""

  • 100

.-~ ---- I--- -~ Bo/tup Temp I 60°F I o 50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Oeg. F) 5-22 Figure 5-8 Point Beach Units 1 and 2 Reactor Coolant System Cooldown Limitations (Cooldown Rates up to lOOOF/hr) Applicable for 53 EFPY (with Hafnium Removal and without Margins for Instrumentation Errors) Using 1998 App. G Methodology (wiKle) WCAP-16669-NP January 2009 Revision 1

POINT BEACH NUCLEAR PLANT TECHNICAL REQUIREMENTS MANUAL PRESSURE TEMPERATURE LIMITS REPORT TABLE 1 (**) POINT BEACH NUCLEAR PLANT UNIT 1 REACTOR VESSEL SURVEILLANCE CAPSULE REMOVAL SCHEDULE Capsule Identification Letter Approximate Removal Oate* V September 1972 (actual) S December 1975 (actual) R October 1977 (actual) T March 1984 (actual) P April 1994 (actual) N Standby TRM 2.2 The actual removal dates will be adjusted to coincide with the closest scheduled plant refueling outage or major reactor plant shutdown.

    • During the period of extended operation, reactor vessel surveillance capsules will be removed and tested in accordance with the schedule contained in the most recently NRC-approved Pressurized Water Reactor Owners Group (PWROG) Master Integrated Reactor Vessel Surveillance Program (MIRVSP) Document. (Ref. 5.5)( Ref 5.25)

TABLE 2 (**) POINT BEACH NUCLEAR PLANT UNIT 2 REACTOR VESSEL SURVEILLANCE CAPSULE REMOVAL SCHEDULE Capsule Identification Letter Approximate Removal Oate* V November 1974 (actual) T March 1977 (actual) R April 1979 (actual) S October 1990 (actual) P June 1997 (actual) N Standby A April 2022 The actual removal dates will be adjusted to coincide with the closest scheduled plant refueling outage or major reactor plant shutdown.

    • During the period of extended operation, reactor vessel surveillance capsules will be removed and tested in accordance with the schedule contained in the most recently NRC-approved Pressurized Water Reactor Owners Group (PWROG) Master Integrated Reactor Vessel Surveillance Program (MI RVSP) Document. (Ref. 5.5)( Ref 5.25).

POINT BEACH TRM 2.2 - 8 REV. 8 MARK-UP

POINT BEACH NUCLEAR PLANT TECHNICAL REQUIREMENTS MANUAL PRESSURE TEMPERATURE LIMITS REPORT TRM 2.2 TABLE 3 POINT BEACH UNIT 1 RPV BEL TLiNE ~ 50 EFPY VALUES(Ej Based on Westinghouse Report WC/\\p 15976,WCAP-16669, Revision 1, "Point Beach Units 1 and 2 Heatup and Cooldown Limit Curves for Normal Operation," (Ref 5.2). Note that the estimated f1uenoe at a speoifio point in time is not linearly interpolated between zero and the estimated f1uenoe at 39.9 EF"PY, due to ohanges in oore design at oertain points in the operating history of the unit. Although the analysis in WCAP 15979 is based on 39.9 EF"PY, the applioability of the analysis is no'.... 35.9 EF"PY per LTR PCAM OS 57, "Point Beaoh Units 1 and 2 EPU P T Limit Cur.'e /\\pplioability Determination and Related Caloulations" (Ref 5.8). Vessel Manufacturer: Babcock & Wilcox Plate and Weld Thickness (without cladding): 6.5", without clad 35.9 EF"PY tet 35.9 EF"PY tet 35.9 EF"P¥-tet 35.9 EFPY tel 35.9 EFPY tet Component Description Heat or Heat/Lot Inside Surface 1/4T Fluence 1/4T Fluence 3/4T Fluence 3/4T Fluence Fluence (E19 n/cm2) (E19 n/cm2) (B) Factor (C) (E19 n/cm2) (B) Factor (C) Nozzle Belt Forging 122P237 ~0.36 Q.4l. 0.24 ~0.62 G,.Qg 0.11 Q.,J.7 0.44 Intermediate Shell Plate A9811-1 ~4.90 ~3.32 ~1.31 :()§ 1.52 ~1.12 Lower Shell Plate C1423-1 ~4.55 2-,W 3.08 ~1.30 (:},.94. 1.41 Q.,gg 1.10 Nozzle Belt to Intermed. Shell 8T1762 ~0.36 Q.4l. 0.24 ~0.62 G,.Qg 0.11 Q.,J.7 0.44 Circ Weld (100%) (SA-1426) Intermediate Shell Long 1P0815 (SA-812) 649 3.19,.4g 2.16 A-:t-1.21 N/A N/A Seam (ID 27%) Intermediate Shell Long \\1-1.) 1 P0661 (SA-775) 649 3.19 N/A N/A Q,.eg 0.99 Q.,gg 1.00 Seam (OD 73%) Intermed. to Lower Shell Circ. 71249 (SA-1101) ~4.43 ~3.00 ~1.29 g.,ga 1.38 Q.,991.09 Weld (100%) Lower Shell Long Seam (Ai 61782 (SA-847) 2-,00 3.05 441-2.07 :-14 1.20 g.,ea 0.95 G,.gg 0.99 (100%) Footnotes: (A) limiting material (8) From an inside surface f1uence value (not including cladding), f1uence is attenuated to a desired thickness using equation (3) of Regulatory Guide 1.99, Revision 2: f = fsurf x e*O.24X, where fsurf is expressed in units of E19 n/cm2, E>1 MeV, and x is the desired depth in inches into the vessel wall. For example, fer the noale selt ferging, heat no. 122P237, at 35.9 eFP¥, at a depth of 1/4 of the 6.5" vessel wall (1.625"), f Q.25 H e~ Q.17 e19 n,lsm2~ (e) The dimensionless f1uence factor is calculated using the f1uence fac:tor formula from equation_(2) of Regulato (D) Instruction Manual, 132-lnch 1.0. Reactor Pressure Vessel, Babcock & Wilcox, September 1969 Guide 1.99, Revision 2: ff = rO.2S

  • 0.10 109 f), where f is the f1uence in units of E19 n/cm2*

~ (E) EFPY value listed here is based on various reactor fuel management strategies and reactor power levels. See VI/CAP 15976 Revision 1 (Ref 5.2) fer dissussion of eFP¥ values. The 36.9 eF'P¥ values listed in 'NCAP 15976, Revision 1, are no..... applisasle to 35.9 eFP¥. POINT BEACH TRM 2.2 - 9 REV. 8 MARK-UP

POINT BEACH NUCLEAR PLANT TECHNICAL REQUIREMENTS MANUAL PRESSURE TEMPERATURE LIMITS REPORT TRM2.2 TABLE 4 POINT BEACH UNIT 2 RPV BEL TLiNE ~ 50 EFPY VALUES(E) Based on Westinghouse Report WCI\\P 1597l,wCAP-16669, Revision 1, "Point Beach Units 1 and 2 Heatup and Cool down Limit Curves for Normal Operation," (Ref 5.2). ~lote that the estimated fiuence at a specific point in time is not linearly interpolated between zero and the estimated fiuence at 36.9 EFPY, due to changes in core design at certain points in the operating history of the unit. Although the analysis in '1,/CAP 1 a976 is based on 36.9 EFPY, the applicability of the analysis is no'.... 3a.9 EFPY per LTR PC/\\M gg a7, "Point Beach Units 1 and 2 EPU P T Limit Curve Applicability Determination and Related Calculations" (Ref 5.S). Vessel Manufacturer: Babcock & Wilcox and Combustion Engineering Plate and Weld Thickness (without cladding): 6.5", without clad (V) 35.9 EFPYw 35.9 EFPY~ 35.9 EFPY~ 35.9 EFPY~ 35.9 EFPY~ Component Description Heat or Inside Surface 1/4T Fluence 1/4T Fluence 3/4T Fluence 3/4T Fluence Heat/Lot Fluence (E19 n/cm2) (B) Factor (C) (E19 n/cm2) (B) Factor (C) (E19 n/cm2) Nozzle Belt Forging 123V3S2 ~0.50 ~0.34 G-:-W 0.70 G-:44 0.16 G44 0.51 Intermediate Shell Forging (A) 123V500 ~S.05 ~3.42 ~1.32 ~1.57 4:-{f.i 1. 1 2 Lower Shell Forging 122W195 ~4.90 ~3.32 ~1.31 .:t-:G2: 1.52 4:-{f.i 1. 1 2 Nozzle Belt to Intermed. Shell 21935 ~0.50 ~0.34 G-:-W 0.70 G-:44 0.16 G44 0.51 Circ Weld (100%) Intermed. to Lower Shell Circ 72442 ~4.65 ~3.15 ~1.30 {h9+ 1.44 .fh991.10 Weld (100%) (Al (SA-1484) Footnotes: (A) Limiting Material (8) From an inside surface f1uence value (not including cladding), f1uence is attenuated to a desired thickness using equation (3) of Regulatory Guide 1.99, Revision 2: f = fsur! x e*O.24" where fsur! is expressed in units of E19 n/cm2, E>1 MeV, and x is the desired depth in inches into the vessel wall. for e*aA'lJ3le, for the noale eelt forging, heat no. 123V352, at 35.9 EfPY, at a deJ3th of 1/4 of the l.5" vessel 'Nail (1.l25"), f 0.34)( e~ 0.23 E19 n/sA'li!~ (e) The dimensionless f1uence factor is calculated using the f1uence factor formula from equation (2) of Regulatory Guide 1.992 Revision 2: ff = 1'0.28.0.10 log I), where f is the f1uence in units of E19 n/cm2* for m(aA'lJ3le, the 35.9 EfPY 1/4T fliolense fastor for noale eelt forging, Reat no. 123V352, ff 0.23(9.28 9.191.~ 9. a)~ (D) (E) Instruction Manual, Reactor Vessel, Point Beach Nuclear Plant No.2, Combustion Engineering, CE Book #4869, October 1970. EFPY value listed here is based on various reactor fuel management strategies and reactor power levels. See 'NCAP 1597l Revision 1 (Ref 5.2) for disSlolssion of EfPY valloles. The 3l.9 EfPY values listed in 'NCAP 1597l, Revision 1, are now aJ3J3lisaele to 35.9 EfPY. POINT BEACH TRM 2.2 - 10 REV. 8 MARK-UP

POINT BEACH NUCLEAR PLANT TECHNICAL REQUIREMENTS MANUAL PRESSURE TEMPERATURE LIMITS REPORT TRM 2.2 TABLE 5 POINT BEACH UNIT 1 RPV 1/4T BELTLINE MATERIAL ADJUSTED REFERENCE TEMPERATURES AT ~ 50 EFPV.(H) Unless otherwise noted, all ART input data obtained from SAW 222§, "RespeAse te Request fer AdditieAallAfoFFAatieA (RAI) Re§ardiA§ Reaster Pressure VesseIIAte§rity," May 1998 (Ref. §.6) aAd 'NGAP 1 §976,WCAP-16669. Revision 1, "Point Beach Units 1 and 2 Heatup and Cooldown Limit Curves for Normal Operation," (Ref 5.2). Altheu§h the aAalysis iA 'NGAP 1 §976 is based eA 26.9 eF"PY, the applisabilil,y efthe aAalysis is Am..., 2§.9 eF"PY per LTR PGAM Q8 §7, "PeiAt Beash UAits 1 aAd 2 ePU P T biFRit GUNe Applisability DeteFFAiAatieA aAd Related GalsulatieAs" (D"f'" Q\\ Vessel Manufacturer: I Babcock & Wilcox I Plate and Weld Thickness (without cladding): I 6.5", without cladU-) I Heat or Initial CF "-'AT ':II:; Q t:t:~ ARTNDT Margin Component Description Heat/Lot RTNDT (OF) %Cu %Ni CF Method Fluenc~ Factor(A) eFt 0"1 0"<1. (OF) Nozzle Belt Forging 122P237 +50 0.11 0.82 77 Table ~0.62 4Q.,8 0 17 34 47.4 Intermediate Shell Plate A9811-1 +1 0.20 0.06 gg +able ~1.31 4Q+.:4 ~ 47- ~ 79.3 Surv: W:+ 8.5 56.4 Oata(B) 104.1 Lower Shell Plate C1423-1 +1 0.12 0.07 55.3 Table ~1.30 W4 ~ 47- ~ 35.8 Surv. ~ 8.5 56.4 Data(B) 46.4 Nozzle Belt to Intermed. Shell 8T1762 -a -47.6 0.19 0.57 452-:4 Table ~0.62 w.,g 49,.7 28 e&-:4+ Circ Weld (100%) (SA-1426) 167.0 102.9 17.2 65.7 Intermediate Shell Long Seam 1 P0815 -a -47.6 0.17 0.52 ~ Table 4-4-4 1.21 4§d.4 49,.7 28 e&-:4+ (1027%) (SA-812) 167.0 201.9 17.2 65.7 Intermediate Shell Long Seam 1P0661 -a -47.6 0.17 0.64 ~ Table N/A N/A 49,.7 28 N/A (0073%) (SA-775) 167.0 17.2 Intermed. to Lower Shell Circ. Weld 71249 ~-47.4 0.23 0.59 167.6 Table(C) ~1.29 ~ G 28 W (100%) (SA-11 01) 216.4 12.9 61.7 Lower Shell Long Seam (100%) 61782 -a-47.6 0.23 0.52 4-eM Table .:I4G 1.20 ~ 49,.7 28 e&-:4+ (SA-847) 167.0 199.9 17.2 65.7 ~ ~ ~ " -14 48M Gata Footnotes. iAi See Table 3 (8) Credible Surveillance Data; see BAW-2325 for evaluation. (e) Non-credible surveillance data; see BAW-2325 for evaluation. Table CF conservative because difference between ratio-adjusted measure ilRT NOT and predicted LlRT NOT based on Table CF is less than 2er (56°F). (0) Credible Surveillance Data; see WE Calculation Addendum 98-0156-00-A, "Evaluation of New Surveillance Data on Chemistry Factor for Weld Wire Heat 61782, Point Beach Unit 1," (Ref.5.3) utilizing latest time-weighted temperature data for Point Beach Unit 1, which supersedes BAW-2325. (E) Adjusted reference temperature (ART) calculated per Regulatory Guide 1.99, Rev. 2. ART = Initial RT NOT + LlRT NOT + Margin, where LlRT NOT = Chemistry Factor x Fluence Factor, and Margin = 2(d + erA 2)0.5, with er] defined as the standard deviation of the Initial RT NOT and erA defined as the standard deviation of LlRT NOT. Galsulated ART values are reuAded te the Aearest oF" iA asserdaAse with the reuAdiA§ eff FRethed ef ASTM Prastise e29. (F) Instruction Manual, 132-lnch 1.0. Reactor Pressure Vessel, Babcock & Wilcox, September 1969. (G) Deleted. (H) EFPY value listed here is based on various reactor fuel management strategies and reactor power levels. See 'NGAP 1 §976 RevisioA 1 (Ref §.2) for dissussieA of eF"PY values. The aM eF"PY values listed iA 'NGAP 1 §976, RevisioA 1, are AOW applisable to 2§.9 eF"PY. POINT BEACH TRM 2.2 - 11 REV. 8 MARK-UP ART eF)(E) ~ 131.4 ~ 4.§4 161.5 4M 400 103.8 444-121.0 ~ 220.0 N/A ~ 230.7 ~ 218.1 ~

POINT BEACH NUCLEAR PLANT TECHNICAL REQUIREMENTS MANUAL PRESSURE TEMPERATURE LIMITS REPORT TABLE 6 POINT BEACH UNIT 2 RPV 1/4T BELTLINE MATERIAL ADJUSTED REFERENCE TEMPERATURES AT ~50EFPy.(I) TRM 2.2 Unless otherwise noted, all ART input data obtained from B/W'l 2d2a, "Response to Request for Aelelitional Information (RAI) Regareling Reastor Pressure Vessel Integrity," May 1999 (Ref. a.e) anel'NCAP 1 a97e,WCAP-16669, Revision 1, "Point Beach Units 1 and 2 Heatup and Cooldown Limit Curves for Normal Operation," (Ref 5.2). Although the analysis in WCAP 1 a97e is saseel on de.9 eFPY, the applisasility of the analysis is nO'N da.9 eFPY per LTR PCAM Qg a7, "Point Beash Units 1 anel 2 ePU P T Limit Curve Applisasility Determination anel Relateel Calsulations" (Refa.g). Vessel Manufacturer: Babcock & Wilcox and Combustion Engineering Plate and Weld Thickness (without cladding): 6.5", without clad H4+ 35.950 Component Description Heat or Initial %Cu %Ni CF CF tIl--F-P¥ dRTNOT Margin Heat/Lot RT NOT (OF) Method Fluence (OF) 0"1 0".1. (OF) Factor(A) Nozzle Belt Forging 123V352 +40 0.11 0.73 76 Table fl.,.W 0.70 4&:e 0 17 34 53.5 Intermediate Shell Forging 123V500 +40 0.09 0.70 58 Table(B) 4421.32 .:ro.,g 0 17 34 76.6 Lower Shell Forging 122W195 +40 0.05 0.72 ~ +able 4421.31 ~ Q 4+ ~ 42.8 Surv. ~ 0 8.5 17 Data(C) 56.2 Nozzle Belt to Intermed. Shell Circ Weld 21935 -56 0.18 0.70 170 Table(H) fl.,.W 0.70 400 17 28 65.5 (100%) 120.0 Intermed. to Lower Shell Circ. Weld 72442 (SA-1484) -a-30 0.26 0.60 180 Table(D) ~1.30 ~ 49-:7 28 e&A+ (100%) 234.4 11.9 60.8 Footnotes. 1A) See Table 4 (8) Non-credible surveillance data; see BAW-2325 for evaluation. Table CF conservative because difference between measured ~RT NOT and predicted ~RT NOT based on Table CF is less than 2a (34°F) (e) Credible surveillance data; see BAW-2325 for evaluation. (D) Non-credible surveillance data; Table CF value based on best-estimate chemistry is higher than best fit calculated using surveillance data, and therefore, conservative. (E) Adjusted reference temperature (ART) calculated per Regulatory Guide 1.99, Rev. 2. ART = Initial RT NOT + ~RT NOT + Margin, where ~RT NOT = Chemistry Factor x Fluence Factor, and Margin = 2( a]2 + ad 2)0.5, with a] defined as the standard deviation of the Initial RT NOT, and ad defined as the standard deviation of ~RT NOT. For example, for no;;;zle selt forging, heat no. 12dVda2, ART 4Q'" (7e)( Q.eQ) '" d4 12QoF. Calsulateell\\RT values are rouneleel to the nearest OF in assorelanse with the rouneling off methoel of ASTM Prestise 1229. (F) Instruction Manual, Reactor Vessel, Point Beach Nuclear Plant Unit 2, Combustion Engineering, CE Book #4869, October 1970. (G) Deleted. (H) Table CF value based on best-estimate chemistry data from CEOG Report "Best Estimate Copper and Nickel Values in CE Fabricated Reactor Vessel Welds," CE NPSD-1039, Revision 2, Final Report, June 1997 (Ref.5.7). (/) EFPY value listed here is based on various reactor fuel management strategies and reactor power levels. See WCAP 1 a97e Revision 1 (Ref a.2) for elissussion of eFPY values. The de.9 eFPY values listeel in 'NCAP 1 a97e, Revision 1, are now applisasle to da.9 eFPY. POINT BEACH TRM 2.2 - 12 REV. 8 MARK-UP ART (OF)(E) 42G 127.5 44a 150.6 1-2 44Q 113.2 1-2 129.5 2gQ 265.2

POINT BEACH NUCLEAR PLANT TECHNICAL REQUIREMENTS MANUAL PRESSURE TEMPERATURE LIMITS REPORT TABLE 7 POINT BEACH UNIT 1 RPV 3/4T BELTLINE MATERIAL ADJUSTED REFERENCE TEMPERATURES AT ~50 EFPy(H) TRM 2.2 Unless otherwise noted, all ART input data obtained from BAW-2325, "Response to Request for Additional Information (RAI) Regarding Reactor Pressure Vessel Integrity," May 1998 (Ref. 5.6) and 'NCAP 15976,WCAP-16669. Revision 1, "Point Beach Units 1 and 2 Heatup and Cool down Limit Curves for Normal Operation," (Ref 5.2). Althololgh the analysis in 'NCAP 15976 is eased on a6.9 eF"PY, the applisaeility of the analysis is now a5.9 eF"PY per bTR PCAM gg 57, "Point Beash Units 1 and 2 ePU P T birnit Clolrve Applisaeility Determination and Related Calslollations" LO~f"'--!ll Vessel Manufacturer: I Babcock & Wilcox I Plate and Weld Thickness (without cladding): I 6.5", without clad1r; I Initial CF 3l4T 35.9 aRTNDT Margin Component Description Heat or Heat/Lot RTNDT eF) %Cu %Ni CF Method ~-Fluence (OF) 0"1 O"d (OF) Factor(A) Nozzle Belt Forging 122P237 +50 0.11 0.82 77 Table Q.,3+ 0.44 ~ 0 17 ~ 33.7 33.7 Intermediate Shell Plate A9811-1 +1 0.20 0.06 gg +aeIe ~1.12 gg.,g ~ 4+ eaM 79.3 Surv. 8M 8.5 56.4 Data(B) 88.5 Lower Shell Plate C1423-1 +1 0.12 0.07 55.3 Table g.,gg 1.10 ~ ~ 4+ eaM 35.8 Surv. ~ 8.5 56.4 Data(B) 39.2 Nozzle Belt to Intermed. Shell Circ Weld 8T1762 (SA-1426) -e -47.6 0.19 0.57 4§.2.4 Table Q.,3+ 0.44 W4 49-:+ 28 eM+ (100%) 167.0 73.2 17.2 65.7 Intermediate Shell Long Seam (1027%) 1P0815 (SA-812) -e -47.6 0.17 0.52 ~ Table N/A N/A 49-:+ 28 eM+ 167.0 17.2 N/A Intermediate Shell Long Seam (00 73%) 1 P0661 (SA-775) -e -47.6 0.17 0.64 ~ Table g.,gg 1.00 ~ 49-:+ 28 eM+ 167.0 166.5 17.2 65.7 Intermed. To Lower Shell Circ. Weld (100%) 71249 (SA-1101) ~-47.4 0.23 0.59 167.6 Table(C) fh991.09 ~ Q 12.9 28 W 182.3 61.7 Lower Shell Long Seam (100%) 61782 (SA-847) -e -47.6 0.23 0.52 4-6+:4 Table (}.gg. 0.99 ~ 49-:+ 28 eM+ 167.0 164.5 17.2 65.7 ~ ~ -QatatGJ 44&7 44 48.,M Footnotes: (A) See Table 3. (B) Credible Surveillance Data; see BAW-2325 for evaluation. (e)

  • Non-credible surveillance data; see BAW-2325 for evaluation. Table CF conservative because difference between ratio-adjusted measured ~RT NOT are predicted ~RT NOT based on Table CF is less than 2a (56°F).

(0) (E) (F) (G) (H) Credible Surveillance Data; see WE Calculation Addendum 98-0156-00-A, "Evaluation of New Surveillance Data on Chemistry Factor for Weld Wire Heat 61782, Point Beach Unit 1," utilizing latest time-weighted temperature data for Point Beach Unit 1, which supersedes BAW-2325. Adjusted reference temperature (ART) calculated per Regulatory Guide 1.99, Rev. 2. ART = Initial RT NOT + ~RT NOT + Margin, where ~RT NOT = Chemistry Factor x Fluence Factor, and Margin = 2( d + a" 2)0.5, with al defined as the standard deviation of the Initial RT NOT, and a" defined as the standard deviation of ~RT NOT. F"or exarnple, for no~le eelt forging, heat no. 122P2a7, ART 5g.. (77)( g.a7)" a4 11a°F". Calslollated ART valloles are rOlolnded to the nearest oF" in assordanse with the FOlolnding off rnethod of ASTM Prastise e29. Instruction Manual, 132-lnch 1.0. Reactor Pressure Vessel, Babcock & Wilcox, September 1969. Deleted. EFPY value listed here is based on various reactor fuel management strategies and reactor power levels. See VVCAP 15976 Revision 1 (Ref 5.2) for disslolssion of eF"PY valloles. The a6.9 eF"PY valloles listed iR WCAP 15976, Re\\'isioR 1, are ROW applisaele to a5.9 eF"PY. POINT BEACH TRM 2.2 - 13 REV. 8 MARK-UP ART (oF)(E) ~ 117.4 454 ~ 145.9 449 W 96.6 4:W 91.3 N/A ~ 184.6 ~ 196.6 ~ 182.7 4S+

POINT BEACH NUCLEAR PLANT TECHNICAL REQUIREMENTS MANUAL PRESSURE TEMPERATURE LIMITS REPORT TABLE 8 POINT BEACH UNIT 2 RPV 3/4T BELTLINE MATERIAL ADJUSTED REFERENCE TEMPERATURES AT .ge.,.g 50 EFPY (I) TRM 2.2 Unless otherwise noted, all ART input data obtained from 9AW 2325, "Response to ReEllolest for Additional Information (RAI) Re§ardin§ Reastor Presslolre Vessellnte§rity," May 1998 (Ref. 5.6) and VI/CAP 15976,wCAP-16669. Revision 1, "Point Beach Units 1 and 2 Heatup and Cooldown Limit Curves for Normal Operation," (Ref 5.2). Altholol§h the analysis in '1VCAP 15976 is eased on 36.9 E:FPY, the applisaeility of the analysis is now 35.9 E:FPY per LTR PCAM 08 57, "Point geash Units 1 and 2 E:PU P T bimit Cwve Applisaeility Determination and Related Calslollations" (Ref 5.8). Vessel Manufacturer: Babcock & Wilcox and Combustion Engineering Plate and Weld Thickness (without cladding): 6.5", without clad 3/4+ 35.9 Component Description Heat or Initial %Cu %Ni CF CF Method ~ ARTNDT Margin Heat/Lot RTNDT eF) Fluence (OF) 0'1 O't. (OF) Factor(AJ Nozzle Belt Forging 123V352 +40 0.11 0.73 76 Table GA4 0.51 ~ 0 17 34 38.9 Intermediate Shell Forging 123V500 +40 0.09 0.70 58 Table(B) 441-1.12 ~ 0 17 34 65.2 Lower Shell Forging 122W195 +40 0.05 0.72 ~ +aa!e 441-1.12 ~ Q 47-d4 42.8 Surv. 46A 0 8.5 17 Data(C) 47.8 Nozzle Belt to Intermed. Shell Circ 21935 -56 0.18 0.70 170 Table(H) GA4 0.51 -74,S 17 28 65.5 Weld (100%) 87.3 Intermed. to Lower Shell Circ. Weld 72442 30 0.26 0.60 180 Table (D) .Q.,.99 1. 1 a ~ 49-:+ 28 e&4+ (100%) (SA-1484) 198.4 11.9 60.8 Footnotes: See Table 4. (A) (8) (C) (D) (E) Non-credible surveillance data; see BAW-2325 for evaluation. Table CF conservative because difference between measured ~RTNOT and predicted ~RTNOT based on Table CF is less than 2cr (56°F). (F) (G) (H) (I) Credible surveillance data; see BAW-2325 for evaluation. Non-credible surveillance data; Table CF value based on best-estimate chemistry is higher than best fit calculated using surveillance data, and therefore, conservative. Adjusted reference temperature (ART) calculated per Regulatory Guide 1.99, Rev. 2. ART = Initial RT NOT + ~RT NOT + Margin, where ~RT NOT = Chemistry Factor x Fluence Factor, and Margin = 2( crl2 + a/)O.5, with al defined as the standard deviation of the Initial RT NOT, and cr" defined as the standard deviation of ~RT NOT. Calslollated ART '.'alloles are FOIoInded to the nearest OF in assordanse with the rOlolndin§ off method of ASTM Prastise E:29. Instruction Manual, Reactor Vessel, Point Beach Nuclear Plant No.2, Combustion Engineering, CE Book #4869, October 1970. Deleted. Table CF value based on best-estimate chemistry data from CEOG Report "Best Estimate Copper and Nickel Values in CE Fabricated Reactor Vessel Welds," CE NPSO-1039, Revision 2, Final Report, June 1997 EFPY value listed here is based on various reactor fuel management strategies and reactor power levels. See 'NCAP 15976 Revision 1 (Ref 5.2) for disslolssion of E:FPY valloles. The 36.9 E:FPY \\'alloles listed in VI/CAP 15976, Revision 1, are nO'N applisaele to 35.9 E:FPY. POINT BEACH TRM 2.2 -14 REV. 8 MARK-UP ART (OFiEJ 4W 112.9 ~ 139.2 4-Ga 400 104.8 84-96.8 ~ 229.2}}

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