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WCAP-17539-NP, Revision 0, Sequoyah, Units 1 and 2 Time-Limited Aging Analysis on Reactor Vessel Integrity
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Site:	Sequoyah
Issue date:	01/10/2013
From:	Wang S S; Westinghouse
To:	; Office of Nuclear Reactor Regulation
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ENCLOSURE 3 Tennessee Valley Authority Sequoyah Nuclear Plant Units I and 2 WCAP-17539-NP, Revision 0, "Sequoyah Units 1 and 2 Time-Limited Aging Analysis on Reactor Vessel Integrity"

------ -----Westinghouse Non-Proprietary Class 3 WCAP-17539-NP

..arc.~Revision 0 Sequoyah Units 1 and 2 Time-Limited Aging Analysis on Reactor Vessel Integrity Westinghouse h 201211 WESTINGHOUSE NON-PROPRIETARY CLASS 3 WCAP-17539-NP Revision 0 Sequoyah Units 1 and 2 Time-Limited Aging Analysis on Reactor Vessel Integrity Amy E. Freed*Aging Management and License Renewal Services Sylvia S. Wang*Radiation Engineering and Analysis March 2012 Reviewers:

Elliot J. Long*Aging Management and License Renewal Services Stanwood L. Anderson*Radiation Engineering and Analysis Approved:

Michael G. Semmler*, Acting Manager Aging Management and License Renewal Services*Electronically approved records are authenticated in the electronic document management system.Westinghouse Electric Company LLC 1000 Westinghouse Drive Cranberry Township, PA 16066© 2012 Westinghouse Electric Company LLC All Rights Reserved Westinghouse Non-Proprietary Class 3 RECORD OF REVISION iii Revision 0: Original Issue WCAP-17539-NP March 2012 Revision 0 iv Westinghouse Non-Proprietary Class 3 TABLE OF CONTENTS TA B LE O F C O N TEN T S ..................................................................................................................

iv L IST O F TA B L E S .............................................................................................................................

v L IST O F F IG U R E S .........................................................................................................................

viii EX ECU TIV E SU M M A RY ...............................................................................................................

ix 1 TIME-LIMITED AGING ANALYSIS .............................................................................

1-1 2 CA LCU LATED FLU EN CE ..................................................................................................

2-1 3 MATERIAL PROPERTY INPUT ....................................................................................

3-1 4 PRESSURIZED THERMAL SHOCK ....................................

4-1 5 U PPER-SHELF EN ERGY ................................................................................................

5-1 6 HEATUP AND COOLDOWN PRESSURE-TEMPERATURE LIMIT CURVES A P P L IC A B IL IT Y .................................................................................................................

6-1 6.1 SE Q U O Y A H U N IT 1 .................................................................................................

6-3 6.2 SEQ U O Y A H U N IT 2 ................................................................................................

6-9 7 SURVEILLANCE CAPSULE WITHDRAWAL SCHEDULES

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7-1 8 R E F E R E N C E S ......................................................................................................................

8-1 APPENDIX A CREDIBILITY EVALUATION OF THE SEQUOYAH UNITS 1 AND 2 SURVEILLANCE PROGRAMS...................................................................................

A-1 A .1 SEQ U O Y A H U N IT 1 ..........................................................................................

A -1 A .2 SEQ U O Y A H U N IT 2 ..............................................................................................

A -8 APPENDIX B SURVEIILANCE CAPSULE RELOCATION EVALUATION FOR SEQU OYAH UN ITS 1 AN D 2 ........................................................................................

B-1 B .I SEQU O Y A H UN IT 1 ..........................................................................................

B-1 B .2 SEQ U O Y A H U N IT 2 ..............................................................................................

B -4 APPENDIX C EMERGENCY RESPONSE GUIDELINE LIMITS .................................

C-1 WCAP-17539-NP March 2012 Revision 0 Westinghouse Non-Proprietary Class 3 V LIST OF TABLES Table 1-1 Evaluation of Time-Limited Aging Analyses Per the Criteria of 10 CFR 54.3 ...............

1-2 Table 2-1 Calculated Neutron Fluence Projections at the Peak Location on the Reactor Vessel Clad/Base Metal Interface for Sequoyah Unit 1 Beltline Materials

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2-3 Table 2-2 Calculated Neutron Fluence Projections at the Peak Location on the Reactor Vessel Clad/Base Metal Interface for Sequoyah Unit 1 Extended Beltline M aterials ..........................................................................................................................

2 -4 Table 2-3 Calculated Neutron Fluence Projections at the Peak Location on the Reactor Vessel Clad/Base Metal Interface for Sequoyah Unit 2 Beltline Materials

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2-5 Table 2-4 Calculated Neutron Fluence Projections at the Peak Location on the Reactor Vessel Clad/Base Metal Interface for Sequoyah Unit 2 Extended Beltline M aterials ..........................................................................................................................

2 -6 Table 2-5 Summary of the Sequoyah Units 1 and 2 Maximum RPV Fluence on the Reactor Vessel Clad/Base Metal Interface at EOL and EOLE ......................................................

2-7 Table 2-6 Calculated Fluence for the Withdrawn Sequoyah Unit 1 Surveillance Capsules (40" A zim uthal L ocation) .................................................................................................

2-7 Table 2-7 Calculated Fluence for the Withdrawn Sequoyah Unit 2 Surveillance Capsules (400 A zim uthal L ocation) .................................................................................................

2-7 Table 3-1 Best-Estimate Cu, Ni, P, and Mn Weight Percents, Initial RTNDT Values, and Initial USE Values for the Sequoyah Unit 1 RPV Beltline and Extended Beltline M aterials ..........................................................................................................................

3 -5 Table 3-2 Best-Estimate Cu, Ni, P, and Mn Weight Percents, Initial RTNDT Values, and Initial USE Values for the Sequoyah Unit 2 RPV Beltline and Extended Beltline M aterials ..........................................................................................................................

3 -7 Table 3-3 Calculation of Position 2.1 CF Values using Sequoyah Unit 1 Surveillance C apsule Test R esults .......................................................................................................

3-9 Table 3-4 Calculation of Position 2.1 CF Values using Sequoyah Unit 2 Surveillance C apsule Test R esults ......................................................................................................

3-10 Table 3-5 Summary of the Sequoyah Unit 1 RPV Beltline and Extended Beltline Material Chemistry Factor Values based on Regulatory Guide 1.99, Revision 2, Position 1.1 and Position 2.1 ...................................................................................................

3-11 Table 3-6 Summary of the Sequoyah Unit 2 RPV Beltline and Extended Beltline Material Chemistry Factor Values based on Regulatory Guide 1.99, Revision 2, Position 1.1 and P osition 2 .1 ........................................................................................................

3-11 Table 4-1 Calculation of Sequoyah Unit 1 RTPTS Values for 52 EFPY (EOLE) at the C lad/B ase M etal Interface

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4-3 WCAP-17539-NP March 2012 Revision 0 vi Westinghouse Non-Proprietary Class 3 Table 4-2 Calculation of Sequoyah Unit 2 RTPTS Values for 52 EFPY (EOLE) at the C lad/B ase M etal Interface

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4-4 Table 5-1 Predicted USE Values at 52 EFPY (EOLE) for Sequoyah Unit 1 ....................................

5-3 Table 5-2 Predicted USE Values at 52 EFPY (EOLE) for Sequoyah Unit 2 ....................................

5-5 Table 6.1-1 Calculation of the Sequoyah Unit 1 ART Values at the 1/4T Location for 32 E F P Y .......................................

........................................................................................

6 -4 Table 6.1-2 Calculation of the Sequoyah Unit 1 ART Values at the 3/4T Location for 32 E F P Y ................................................................................................................................

6 -5 Table 6.1-3 Calculation of the Sequoyah Unit 1 ART Values at the 1/4T Location for 52 E F P Y ................................................................................................................................

6 -6 Table 6.1-4 Calculation of the Sequoyah Unit 1 ART Values at the 3/4T Location for 52 E F P Y ................................................................................................................................

6 -7 Table 6.1-5 Summary of the Sequoyah Unit 1 Limiting ART Values used in the Applicability Evaluation of the Current Reactor Vessel Heatuip and Cooldown Curves .......................

6-8 Table 6.2-1 Calculation of the Sequoyah Unit 2 ART Values at the 1/4T Location for 32 E F P Y ..............................................................................................................................

6 -11 Table 6.2-2 Calculation of the Sequoyah Unit 2 ART Values at the 3/4T Location for 32 E F P Y ..............................................................................................................................

6 -12 Table 6.2-3 Calculation of the Sequoyah Unit 2 ART Values at the 1/4T Location for 52 E F P Y ..............................................................................................................................

6 -13 Table 6.2-4 Calculation of the Sequoyah Unit 2 ART Values at the 3/4T Location for 52 E F P Y ..............................................................................................................................

6 -14 Table 6.2-5 Summary of the Sequoyah Unit 2 Limiting ART Values used in the Applicability Evaluation of the Current Reactor Vessel Heatup and Cooldown Curves .....................

6-15 Table 7-1 Sequoyah Unit 1 Surveillance Capsule Withdrawal Summary ........................................

7-1 Table 7-2 Sequoyah Unit 2 Surveillance Capsule Withdrawal Summary ........................................

7-2 Table A. 1-1 Calculation of Interim Chemistry Factors for the Credibility Evaluation Using Sequoyah Unit 1 Surveillance Capsule Data Only ........................................................

A-4 Table A. 1-2 Sequoyah Unit 1 Surveillance Capsule Data Scatter about the Best-Fit Line ...........

A-5 Table A.2-1 Calculation of Interim Chemistry Factors for the Credibility Evaluation Using Sequoyah Unit 2 Surveillance Capsule Data Only ...................................................

A-11 Table A.2-2 Sequoyah Unit 2 Surveillance Capsule Data Scatter about the Best-Fit Line ...... A-12 Table B.1-1 Projected Neutron Fluence Values at the Geometric Center of the Surveillance Capsule Locations for Sequoyah Unit 1 ....................................................................

B-1 Table B. 1-2 Sequoyah Unit 1 Projected Capsule Neutron Fluence Values Associated with Capsule Relocation from the 4' to the 40' Azimuthal Location .....................................

B-2 WCAP- 17539-NP March 2012 Revision 0 Westinghouse Non-Proprietary Class 3 vii Table B. 1-3 Sequoyah Unit 1 Potential Capsule Withdrawal Times Associated with Capsule Relocation from the 40 to the 400 Azimuthal Location ...................................................

B-3 Table B.2-1 Projected Neutron Fluence Values at the Geometric Center of the Surveillance Capsule Locations for Sequoyah Unit 2 .........................................................................

B-4 Table B.2-2 Sequoyah Unit 2 Projected Capsule Neutron Fluence Values Associated with Capsule Relocation from the 40 to the 40' Azimuthal Location .....................................

B-5 Table B.2-3 Sequoyah Unit 2 Potential Capsule Withdrawal Times Associated with Capsule Relocation from the 40 to the 40' Azimuthal Location ...................................................

B-6 Table C- 1 Evaluation of Sequoyah Units 1 and 2 ERG Limit Category .....................................

C-1 WCAP-17539-NP March 2012 Revision 0 viii Westinghouse Non-Proprietary Class 3 LIST OF FIGURES Figure 3-1 RPV Material Identification for Sequoyah Units 1 and 2 ................................................

3-4 Figure 5-1 Regulatory Guide 1.99, Revision 2 Predicted Decrease in Upper-Shelf Energy as a Function of Copper and Fluence for Sequoyah Unit 1 ....................................................

5-4 Figure 5-2 Regulatory Guide 1.99, Revision 2 Predicted Decrease in Upper-Shelf Energy as a Function of Copper and Fluence for Sequoyah Unit 2 ....................................................

5-6 WCAP-17539-NP March 2012 Revision 0 Westinghouse Non-Proprietary Class 3 ix EXECUTIVE

SUMMARY

This report presents the Time-Limited Aging Analyses (TLAA) for the Sequoyah Units 1 and 2 reactor pressure vessels in accordance with the requirements of the License Renewal Rule, 10 CFR Part 54. Time-Limited Aging Analyses are calculations which evaluate some safety-related aspects of the reactor pressure vessel within the bounds of the current 40-year license that must be re-evaluated to account for an extended period of operation.

The Sequoyah Units 1 and 2 current 40-year licenses are applicable through 32 effective full power years (EFPY) of operation, which is deemed end-of-license (EOL). Therefore, with a 20-year license extension, the license renewal is applicable through 60 years of operation or 52 EFPY, which is deemed end-of-license extension (EOLE). Updated neutron fluence evaluations were performed as part of this TLAA evaluation, and are summarized in Section 2 of this report.The fluence values were used to identify the Sequoyah Units 1 and 2 extended beltline materials, which are summarized in Section 3 of this report, and were used as input to the reactor vessel integrity (RVI) evaluations in support of license renewal.In addition to the RVI TLAA evaluations, the credibility of the Sequoyah Units 1 and 2 surveillance materials was also evaluated.

Conclusions for the surveillance data credibility evaluations are contained in Appendix A of this report. Appendix B contains recommendations for capsule relocations in order to obtain meaningful metallurgical data for the future. Appendix C contains the Emergency Response Guideline (ERG) limits classification for Sequoyah Units 1 and 2. The ERG limits were developed in order to establish guidance for operator action in the event of an emergency situation, such as a PTS event. Conclusions for the ERG limits evaluations are contained in Appendix B of this report.A summary of results for the Sequoyah Units 1 and 2 TLAA is provided below. Based on the results of this TLAA evaluation, it is concluded that the Sequoyah Units 1 and 2 reactor vessels will remain adequate through the extended period of operation.

EOLE Pressurized Thermal Shock All of the beltline and extended beltline materials in the Sequoyah Units 1 and 2 reactor vessels are below the RTPTs screening criteria values of 2707F, for forgings, and 3007F, for circumferentially oriented welds (Per 10 CFR 50.61), through EOLE (52 EFPY). See Section 4 for more details.EOLE Upper-Shelf Energy All of the beltline and extended beltline materials in the Sequoyah Units 1 and 2 reactor vessels are projected to remain above the USE screening criterion value of 50 ft-lb (per 10 CFR 50, Appendix G), through EOLE (52 EFPY). See Section 5 for more details.WCAP-17539-NP March 2012 Revision 0 X Westinghouse Non-Proprietary Class 3 Applicability of Existing Pressure-Temperature Limit Curves With a re-evaluation of surveillance data credibility, a recalculation of the chemistry factor values based on surveillance data, and the consideration of TLAA fluence projections, the applicability of the Sequoyah Units 1 and 2 pressure-temperature limit curves may either remain unchanged or be extended.

See Section 6 for more details.Surveillance Capsule Withdrawal Schedules Sequoyah Units 1 and 2 have satisfied the surveillance capsule requirements through EOL (32 EFPY). Several additional capsules for each Unit should be relocated to higher lead factor locations.

One of these relocated capsules in each Unit should be withdrawn from the reactor vessels in order to achieve 60-year (52 EFPY) fluence data prior to EOLE. See Section 7 for more details.WCAP-17539-NP March 2012 Revision 0 Westinghouse Non-Proprietary Class 3 1-1 1 TIME-LIMITED AGING ANALYSIS Time-limited aging analyses (TLAAs) are those licensee calculations that:* Consider the effects of aging" Involve time-limited assumptions defined by the current operating term (e.g., 40 years)* Involve systems, structures, and components (SSCs) within the scope of license renewal* Involve conclusions or provide the basis for conclusions related to the capability of the SSC to perform its intended functions* Were determined to be relevant by the licensee in making a safety determination: " Are contained or incorporated by reference in the current licensing basis (CLB)The potential TLAAs for the reactor pressure vessel (RPV) are identified in Table 1-1 along with indication of whether or not they meet the six criteria of 10 CFR 54.3 (Reference

1) for TLAAs.WCAP-17539-NP March 2012 Revision 0 1-2 Westinghouse Non-Proprietary Class 3 1-2 Westinghouse Non-Proprietary Class 3 Table 1-1 Evaluation of Time-Limited Aging Analyses Per the Criteria of 10 CFR 54.3 Pressure-Calculated Pressurized Upper- Temperature Time-Limited Aging Analysis Thermal Shelf Limits for Fluence Shock(a) Energy Heatup and Cooldown Considers the Effects of Aging YES YES YES YES Involves Time-Limited Assumptions Defined by the YES YES YES YES Current Operating Term Involves SSC Within the Scope of YES YES YES YES License Renewal Involves Conclusions or Provides the Basis for Conclusions Related to the Capability of SSC to Perform Its Intended Function Determined to be Relevant by the Licensee in Making a Safety YES YES YES YES Determination Contained or Incorporated by YES YES YES YES Reference in the CLB Note: (a) The limiting Pressurized Thermal Shock (PTS) values are used to determine the appropriate Emergency Response Guideline (ERG) Limits category for Sequoyah Units 1 and 2 through the end of the potential 20-year license extension period. However, the ERG Limit categories themselves are not a TLAA. See Appendix C for additional information.

WCAP-17539-NP March 2012 Revision 0 Westinghouse Non-Proprietary Class 3 2-1 2" CALCULATED FLUENCE At currently licensed service times and operating conditions, the Sequoyah Units 1 and 2 RPV fracture toughness properties provide adequate margins of safety against vessel failure. However, as a vessel accumulates more and more service time, neutron irradiation (fluence) reduces material fracture toughness and initial safety margins. Prevention of RPV failure depends primarily on maintaining RPV material fracture toughness at levels that resist brittle fracture during plant operation.

The first step in the TLAA of vessel embrittlement is the calculation of the neutron fluence that causes the embrittlement to increase with time.The reactor vessel beltline neutron fluence values applicable to a postulated 20-year license renewal period were calculated for each of the Sequoyah Units 1 and 2 RPV beltline materials.

The analysis methodologies used to calculate the Sequoyah Units 1 and 2 vessel fluences satisfy the requirements set forth in Regulatory Guide 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence" (Reference 2). These methodologies have been approved by the US NRC and are described in detail in WCAP-14040-A, Revision 4 (Reference

3) and WCAP-16083-NP-A, Revision 0 (Reference 4).In accordance with Item IV.A2.R-84 of NUREG-1801, Revision 2 (Reference 5), any materials exceeding 1.0 x 1017 n/cm 2 (E > 1.0 MeV) must be monitored to evaluate the changes in fracture toughness.

RPV materials that are not traditionally thought of as being plant limiting because of low levels of neutron radiation must now be evaluated to determine the accumulated fluence at EOLE. Therefore, fluence calculations were performed for the Sequoyah Units 1 and 2 RPV inlet nozzle to upper shell welds, upper to intermediate shell circumferential welds, lower shell to bottom head ring circumferential welds, and the bottom head ring to bottom head circumferential welds, along with the associated forging materials to determine if they will exceed 1.0 x 1017 n/cm 2 (E > 1.0 MeV) at EOLE. Note that the outlet nozzle to upper shell welds were not evaluated because they experience lower fluence levels, as comparedto the inlet nozzle to upper shell welds, due to a higher elevation relative to the active core. The materials that exceed the 1.0 x 1017 n/cm 2 (E > 1.0 MeV) threshold are referred to as extended beltline materials in this report and are evaluated to determine their impact to the proposed license renewal period.The fluence evaluations included a plant and fuel cycle specific analysis for fuel cycles 1 through 18 for Unit 1 and cycles 1 through 17 for Unit 2, and projections for future operation through EOLE, which is 60 years of plant life or 52 EFPY of operation.

In all cases, the maximum exposure occurs at the 450 azimuthal location of the pressure vessel clad/base metal interface.

Data is given for the nominal end of Cycle 18 for Unit 1 (22.1 EFPY) and Cycle 17 for Unit 2 (21.6 EFPY) as well as for projections through 52 EFPY. Projections for future operation were based on the continued use of the average core data of Cycles 16, 17, and 18 for Unit 1 and Cycles 15, 16, and 17 for Unit 2 and a core power level of 3455 MWt.Tables 2-1 and 2-2 summarize the maximum projected neutron fluence at Sequoyah Unit 1 for each of the reactor pressure vessel beltline and extended beltline materials, respectively.

Similar data for Sequoyah Unit 2 are provided in Tables 2-3 and 2-4.WCAP-17539-NP March 2012 Revision 0 2-2 Westinghouse Non-Proprietary Class 3 From Table 2-2, it is noted that, although the upper shell course and the upper shell to intermediate shell circumferential weld are projected to exceed the 1.0 x 101 n/cm 2 (E > 1.0 MeV) threshold neutron exposure defining the beltline region, the inlet nozzle to upper shell weld remains below 1.0 x 1017 n/cm 2 through 52 EFPY of operation.

Likewise, the bottom head ring to bottom head circumferential weld remains outside of the extended beltline region through 52 EFPY. Similar observations are noted from Table 2-4 for Sequoyah Unit 2 also.The material-specific neutron fluence values at 32 EFPY and 52 EFPY will be used for the calculations contained within this report. The peak neutron fluence at 32 EFPY and 52 EFPY for the beltline materials corresponds to the intermediate to lower shell forgings.

The peak neutron fluence at 52 EFPY for the extended beltline materials corresponds to the lower shell to bottom head ring circumferential weld along with the bottom head forgings.

These maximum neutron fluence values are summarized in Table 2-5.Four surveillance capsules have been withdrawn from each of the Sequoyah Plants. The calculated fast neutron fluences at the 40' azimuthal surveillance capsule location are shown in Tables 2-6 and 2-7 for Units 1 and 2, respectively.

WCAP-17539-NP March 2012 Revision 0 Westinghouse Non-Proprietary Class 3 2-3 Westinghouse Non-Proprietary Class 3 2-3 Table 2-1 Calculated Neutron Fluence Projections at the Peak Location on the Reactor Vessel Clad/Base Metal Interface for Sequoyah Unit 1 Beltline Materials Intermediate Shell Intermediate Shell Lower Shell Operating Time to Lower Shell Forging Circ. Weld EFPY n/cm2, E > 1.0 MeV 1.07 7.64E+17 7.41E+17 7.64E+17 1.90 1.31E+18 1.31E+18 1.31E+18 2.85 2.15E+18 2.15E+18 2.15E+18 4.03 2.85E+18 2.85E+18 2.85E+18 5.26 3.60E+18 3.59E+18 3.60E+18 6.26 4.11E+18 4.10E+18 4.11E+18 7.49 4.80E+ 18 4.78E+18 4.80E+ 18 8.72 5.46E+18 5.42E+18 5.46E+18 10.02 6.19E+18 6.16E+18 6.19E+18 11.38 6.88E+18 6.85E+18 6.88E+18 12.82 7.64E+18 7.61E+18 7.64E+18 14.12 8.42E+ 18 8.40E+ 18 8.42E+ 18 15.44 9.12E+ 18 9.09E+ 18 9.12E+ 18 16.80 9.83E+18 9.80E+18 9.83E+18 18.17 1.06E+19 1.06E+19 1.06E+19 19.51 1.15E+19 1.15E+19 1.15E+19 20.90 1.21E+19 1.21E+19 1.21E+19 22.14 1.27E+19 1.27E+19 1.27E+19 23.47 1.33E+19 1.33E+19 1.33E+19 24.00 1.36E+19 1.35E+19 1.36E+19 28.00 1.54E+19 1.54E+19 1.54E+19 32.00 1.73E+19 1.72E+19 1.73E+19 36.00 1.92E+19 1.91E+19 1.92E+19 40.00 2.10E+19 2.09E+19 2.10E+19 44.00 2.29E+19 2.28E+19 2.29E+19 48.00 2.47E+19 2.46E+19 2.47E+19 52.00 2.66E+19 2.65E+19 2.66E+19 WCAP-17539-NP March 2012 Revision 0 2-4 Westinghouse Non-Proprietary Class 3 Table 2-2 Calculated Neutron Fluence Projections at the Peak Location on the Reactor Vessel Clad/Base Metal Interface for Sequoyah Unit 1 Extended Beltline Materials Lower Shell Bottom Head Operating Inlet Nozzle to Upper to Bottom Bottom Ring to Time to Upper Intermediate Shell Head Bottom Head Shell Welds Shell Circ. Forging Circ. Weld Ring Circ. Weld Weld EFPY n/cm2, E > 1.0 MeV 1.07 5.78E+14 1.03E+16 1.03E+16 7.37E+16 7.37E+16 4.73E+13 1.90 1.23E+15 2.25E+16 2.25E+16 1.38E+17 1.38E+17 8.76E+13 2.85 1.98E+15 3.65E+16 3.65E+16 2.35E+17 2.35E+17 1.49E+14 4.03 2.70E+15 4.99E+16 4.99E+16 3.20E+17 3.20E+17 2.03E+14 5.26 3.56E+15 6.49E+16 6.49E+16 4.13E+17 4.13E+17 2.66E+14 6.26 4.17E+15 7.49E+16 7.49E+16 4.77E+17 4.77E+17 3.13E+14 7.49 5.06E+15 8.92E+16 8.92E+16 5.64E+17 5.64E+17 3.78E+14 8.72 5.87E+15 1.03E+17 1.03E+17 6.48E+17 6.48E+17 4.40E+14 10.02 6.84E+15 1.18E+17 1.18E+17 7.36E+17 7.36E+17 5.05E+14 11.38 7.61E+15 1.31E+17 1.31E+17 8.10E+17 8.10E+17 5.59E+14 12.82 8.70E+15 1.49E+17 1.49E+17 9.08E+17 9.08E+17 6.32E+14 14.12 9.75E+15 1.66E+17 1.66E+17 1.01E+18 1.01E+18 7.03E+14 15.44 1.07E+16 1.81E+17 1.81E+17 1.10E+18 1.10E+18 7.72E+14 16.80 1.17E+16 1.98E+17 1.98E+17 1.19E+18 1.19E+18 8.39E+14 18.17 1.28E+16 2.15E+17 2.15E+17 1.28E+18 1.28E+18 9.07E+14 19.51 1.41E+16 2.37E+17 2.37E+17 1.40E+18 1.40E+18 9.93E+14 20.90 1.50E+16 2.51E+17 2.51E+17 1.48E+18 1.48E+18 1.05E+15 22.14 1.58E+16 2.64E+17 2.64E+17 1.55E+18 1.55E+18 1.11E+15 23.47 1.67E+16 2.78E+17 2.78E+17 1.63E+18 1.63E+18 1.17E+15 24.00* 1.71E+16 2.83E+17 2.83E+17 1.67E+18 1.67E+18 1.20E+15 28.00 1.97E+16 3.26E+17 3.26E+17 1.91E+18 1.91E+18 1.38E+15 32.00 2.24E+16 3.69E+17 3.69E+17 2.15E+18 2.15E+18 1.56E+15 36.00 2.51E+16 4.12E+17 4.12E+17 2.39E+18 2.39E+18 1.74E+15 40.00 2.78E+16 4.55E+17 4.55E+17 2.63E+18 2.63E+18 1.93E+15 44.00 3.05E+16 4.98E+17 4.98E+17 2.88E+18 2.88E+18 2.11E+15 48.00 3.31E+16 5.41E+17 5.41E+17 3.12E+18 3.12E+18 2.29E+15 52.00 3.58E+16 5.84E+17 5.84E+17 3.36E+18 3.36E+18 2.48E+15 WCAP-17539-NP March 2012 Revision 0 Westinghouse Non-Proprietary Class 3 2-5 Table 2-3 Calculated Neutron Fluence Projections at the Peak Location on the Reactor Vessel Clad/Base Metal Interface for Sequoyah Unit 2 Beltline Materials Intermediate Shell Intermediate Shell Lower Shell Operating Time Forging to Lower Shell Forging Circ. Weld F EFPY n/cm 2 , E > 1.0 MeV 1.07 7.85E+17 7.38E+17 7.85E+17 1.88 1.50E+18 1.45E+18 1.50E+18 2.91 2.07E+18 2.01E+18 2.07E+18 4.15 2.87E+18 2.82E+18 2.87E+18 5.36 3.64E+18 3.58E+18 3.64E+18 6.63 4.41E+18 4.33E+18 4.41E+18 7.95 4.98E+18 4.89E+ 18 4.98E+18 9.16 5.66E+ 18 5.54E+ 18 5.66E+ 18 10.55 6.40E+ 18 6.28E+18 6.40E+ 18 11.98 7.09E+ 18 6.96E+ 18 7.09E+ 18 13.38 7.79E+18 7.66E+ 18 7.79E+18 14.75 8.51E+I 8 8.38E+1 8 8.51E+1 8 16.04 9.22E+18 9.10E+18 9.22E+18 17.51 9.99E+18 9.87E+18 9.99E+18 18.84 1.07E+19 1.06E+19 1.07E+19 20.17 1.13E+19 1.11E+19 1.13E+19 21.60 1.19E+19 1.18E+19 1.19E+19 22.97 1.25E+19 1.24E+19 1.25E+19 24.00 1.30E+19 1.29E+19 1.30E+19 28.00 1.48E+19 1.47E+19 1.48E+19 32.00 1.66E+19 1.65E+19 1.66E+19 36.00 1.84E+19 1.83E+19 1.84E+19 40.00 2.02E+19 2.01E+19 2.02E+ 19 44.00 2.20E+19 2.19E+ 19 2.20E+19 48.00 2.38E+19 2.37E+19 2.38E+19 52.00 2.57E+ 19 2.55E+19 2.57E+19 WCAP-17539-NP March 2012 Revision 0 2-6 Westinghouse Non-Proprietary Class 3 Table 2-4 Calculated Neutron Fluence Projections at the Peak Location on the Reactor Vessel Clad/Base Metal Interface for Sequoyah Unit 2 Extended Beltline Materials Inlet Upper Shell Lower Bottom Operating Nozzle to to Upper Shell to Bottom Head Ring Time Upper Intermediate Shell Bottom Head to Bottom Shell Shell Circ. Forging Head Ring Ring Head Circ.Welds Weld Circ. Weld Weld EFPY n/cm2 , E > 1.0 MeV 1.07 4.50E+14 8.08E+15 8.08E+15 6.95E+16 6.95E+16 4.37E+13 1.88 1.04E+15 1.91E+16 1.91E+16 1.42E+17 1.42E+17 8.75E+13 2.91 1.47E+15 2.68E+16 2.68E+16 1.93E+17 1.93E+17 1.21E+14 4.15 2.29E+15 4.11E+16 4.11E+16 2.85E+17 2.85E+17 1.83E+14 5.36 2.97E+15 5.30E+16 5.30E+16 3.66E+17 3.66E+17 2.38E+14 6.63 3.83E+15 6.75E+16 6.75E+16 4.61E+17 4.61E+17 3.05E+14 7.95 4.51E+15 7.84E+16 7.84E+16 5.33E+17 5.33E+17 3.58E+14 9.16 5.29E+15 9.11E+16 9.11E+16 6.12E+17 6.12E+17 4.16E+14 10.55 6.24E+15 1.07E+17 1.07E+17 6.98E+17 6.98E+17 4.80E+14 11.98 7.18E+15 1.22E+17 1.22E+17 7.81E+17 7.81E+17 5.41E+14 13.38 8.20E+15 1.38E+17 1.38E+17 8.73E+17 8.73E+17 6.10E+14 14.75 9.21E+15 1.55E+17 1.55E+17 9.67E+17 9.67E+17 6.80E+14 16.04 1.02E+16 1.71E+17 1.71E+17 1.05E+18 1.05E+18 7.44E+14 17.51 1.13E+16 1.88E+17 1.88E+17 1.15E+18 1.15E+18 8.17E+14 18.84 1.22E+16 2.03E+17 2.03E+17 1.24E+18 1.24E+18 8.81E+14 20.17 1.31E+16 2.17E+17 2.17E+17 1.31E+18 1.31E+18 9.36E+14 21.60 1.41E+16 2.33E+17 2.33E+17 1.40E+18 1.40E+18 1.00E+15 22.97 1.50E+16 2.47E+17 2.47E+17 1.48E+18 1.48E+18 1.06E+15 24.00. 1.56E+16 2.58E+17 2.58E+17 1.54E+18 1.54E+18 1.11E+15 28.00 1.83E+16 3.OOE+17 3.00E+17 1.77E+18 1.77E+18 1.28E+15 32.00 2.09E+16 3.42E+17 3.42E+17 2.OOE+18 2.00E+18 1.46E+15 36.00 2.35E+16 3.84E+17 3.84E+17 2.23E+18 2.23E+18 1.63E+15 40.00 2.61E+16 4.26E+17 4.26E+17 2.47E+18 2.47E+18 1.80E+15 44.00 2.87E+16 4.68E+17 4.68E+17 2.70E+18 2.70E+18 1.98E+15 48.00 3.13E+16 5.10E+17 5.10E+17 2.93E+18 2.93E+18 2.15E+15 52.00 3.40E+16 5.52E+17 5.52E+17 3.16E+18 3.16E+18 2.33E+15 WCAP-17539-NP March 2012 Revision 0 Westinghouse Non-Proprietary Class 3 2-7 Table 2-5 Summary of the Sequoyah Units 1 and 2 Maximum RPV Fluence on the Reactor Vessel Clad/Base Metal Interface at EOL and EOLE Maximum Neutron Fluence(a)

Operating Time (n/cm 2 , E > 1.0 MeV)(EFPY)Unit 1 Unit 2 32 1.73E+19 1.66E+19 52 2.66E+19 2.57E+19 (Beltline Materials) 52 3.36E+18 3.16E+18 (Extended Beltline Materials)

Note: (a) Peak fluence values taken from Tables 2-1 & 2-3 for 32 EFPY and 52 EFPY (Beltline) and from Tables 2-2& 2-4 for 52 EFPY (Extended Beltline).

Table 2-6 Calculated Fluence for the Withdrawn Sequoyah Unit 1 Surveillance Capsules (40° Azimuthal Location)Capsule EFPY Neutron Fluence (Cycle Withdrawn) (n/cm2, E > 1.0 MeV)T 1.07 2.41E+18 (EOC 1)U 2.85 6.93E+18 (EOC 3).X 5.26 1.16E+ 19 (EOC 5)F 10.02 1.97E+19 (EOC 9)Table 2-7 Calculated Fluence for the Withdrawn Sequoyah Unit 2 Surveillance Capsules (40* Azimuthal Location)Capsule EFPY Neutron Fluence (Cycle Withdrawn) (n/cm 2 , E > 1.0 MeV)T 1.07 2.44E+18 (EOC 1)U 2.91 6.54E+18 (EOC 3)x ( 5.36 1.16E+19 (EOC 5)Y 10.55 2.02E+19 (EOC 9)WCAP- 17539-NP March 2012 Revision 0 Westinghouse Non-Proprietary Class 3 3-1 3 MATERIAL PROPERTY INPUT The Sequoyah Units 1 and 2 reactor pressure vessels were fabricated by Rotterdam Drydock Company (RDM). The Sequoyah Units 1 and 2 beltline materials consist of the Intermediate Shell (IS) Forging, Lower Shell (LS) Forging 04, and the IS to LS Circumferential Weld W05.The Sequoyah Unit 1 reactor vessel beltline circumferential weld was fabricated using SMIT 40 weld wire type, heat # 25295 and SMIT 89 flux type, lot # 2275. The weld material in the Sequoyah Unit 1 surveillance program was made of the same material as the Unit 1 reactor vessel beltline circumferential weld, and is not in any other plant's surveillance program. The Sequoyah Unit 2 reactor vessel beltline circumferential weld was fabricated using Arcos weld wire type, heat # 4278 and SMIT 89 flux type, lot # 1211. The weld material in the Sequoyah Unit 2 surveillance program was made of the same material as the Unit 2 reactor vessel beltline circumferential weld, and is not in any other plant's surveillance program.Based on the results of Section 2 of this report, the materials that exceeded the 1 x 1017 n/cm 2 (E> 1.0 MeV) threshold at 52 EFPY (EOLE) are considered to be the Sequoyah Units 1 and 2 extended beltline materials and are evaluated to determine their impact on the proposed license renewal period. The Sequoyah Units 1 and 2 extended beltline materials consist of the Upper Shell (US) Forging 06, Bottom Head Ring 03, US to IS Circumferential Weld W06, and the LS to Bottom Head Ring Weld W04. The Sequoyah Unit 1 US to IS Circumferential Weld W06 was fabricated with SMIT 40 weld wire type, heat # 25006 and SMIT 89 flux type, lot # 8985.The Sequoyah Unit 1 LS to Bottom Head Ring Circumferential Weld W04 was fabricated with SMIT 40 weld wire type, heat # 25295 and SMIT 89 flux type, lot # 1135, which is the same material as the Unit 1 reactor vessel beltline circumferential weld and surveillance material.Both of the Unit 2 extended beltline welds were fabricated using Arcos weld wire type, heat #721858 and SMIT 89 flux type, lot # 1197. No surveillance data exists for weld heat numbers 25006 and 721858.The identification of the reactor vessel beltline and extended beltline materials are included in Figure 3-1 for Sequoyah Units 1 and 2. The material property inputs used for the subsequent RVI evaluations contained in this report are described in this section. Note that the sources and methods used to determine the extended beltline material properties are consistent with those used in the past to determine the initial properties for the beltline materials.

The sources and methods used in the determination of the chemical compositions and the fracture toughness properties are summarized below.Chemical Compositions The best-estimate copper (Cu), nickel (Ni), phosphorus (P), and manganese (Mn) chemical compositions for the Sequoyah Units 1 and 2 beltline and extended beltline materials are presented in Tables 3-1 and 3-2, respectively.

The best-estimate weight percent copper and nickel values for the beltline materials were previously reported, and were used in past RVI evaluations.

The best-estimate weight percent copper and nickel values for the extended beltline materials, along with the best-estimate manganese and phosphorus for the beltline and extended beltline materials were determined as part of this TLAA effort. Note that the best-estimate WCAP-17539-NP March 2012 Revision 0 3-2 Westinghouse Non-Proprietary Class 3 manganese and phosphorus values are reported for information purposes only, and are not used in any subsequent RVI evaluations contained within this report.Except for the weight percent copper values, Certified Material Test Report (CMTR) data was used to determine the chemical compositions for all of the Sequoyah Units 1 and 2 beltline and extended beltline forging materials.

Weight percent copper values were not reported in the CMTRs for the extended beltline forging materials; therefore, the maximum weight percent copper value for A508 Class 2 forging materials was conservatively applied based on the generic data provided in Appendix G of the Oak Ridge National Laboratory Report (Reference 6).The best-estimate copper and nickel for the Sequoyah Unit 1 beltline and surveillance weld materials (heat # 25295) were previously documented in WCAP-15293, Revision 2 (Reference 7). The best-estimate phosphorus and manganese for these weld materials were determined using test records from the Rotterdam Weld Files as well as WCAP-8233 (Reference 8). Limited information was available for the Sequoyah Unit 1 extended beltline US to IS circumferential weld (heat # 25006) in the Rotterdam weld certification records. Except for the weight percent nickel value, the chemical compositions were taken from a chemical analysis performed on the weld wire (heat # 25006) included in the Rotterdam weld certification records. Weight percent nickel was not reported in the weld certification records for heat # 25006; therefore, a value of 1.0 was conservatively assumed per 10 CFR 50.61 (Reference 9). The LS to Bottom Head Ring circumferential weld was fabricated using the same weld wire heat number and flux type as the IS to LS circumferential weld. Therefore, the chemical compositions of the IS to LS circumferential weld were applied to the LS to Bottom Head Ring circumferential weld.The best-estimate copper and nickel for the Sequoyah Unit 2 beltline and surveillance weld materials (heat # 4278) were previously documented in WCAP-15321, Revision 2 (Reference 10). The best-estimate phosphorus and manganese for these weld materials were determined using Rotterdam test records as well as WCAP-8513 (Reference 11). The weight percent copper and manganese values for the Sequoyah Unit 2 extended beltline welds (heat # 721858) were taken from the as-deposited weld analysis in the Rotterdam weld certification records; however, the weight percent phosphorus was taken from a chemical analysis performed on the weld wire since an analysis for phosphorus was not performed on the as-deposited weld. Weight percent nickel was not reported in the Rotterdam weld certification records for heat # 721858; therefore, a value of 1.0 was conservatively assumed per 10 CFR 50.61 (Reference 9).Fracture Toughness Properties The fracture toughness properties of the ferritic materials in the reactor coolant pressure boundary were determined in accordance with NUREG-0800 Branch Technical Position 5-3 (Reference 12). The beltline and extended beltline material properties of the Sequoyah Units 1 and 2 reactor vessels are presented in Tables 3-1 and 3-2, respectively.

The initial reference nil-ductility transition temperature (RTNDT) and initial upper-shelf energy (USE) values for the Sequoyah Units 1 and 2 beltline materials were previously documented in WCAP-15293, Revision 2 (Reference

7) and WCAP-15321, Revision 2 (Reference 10), respectively.

The fracture toughness properties for the extended beltline forging materials are WCAP-17539-NP March 2012 Revision 0 Westinghouse Non-Proprietary Class 3 3-3 based on the values documented in Table B 3/4.4-1 of the Units 1 and 2 Technical Specification (TS) Bases. In accordance with Section B. 1.2 of NUREG-0800 Branch Technical Position 5-3, the initial USE values reported in the TS Bases for the Units 1 and 2 Bottom Head Ring materials as well as the Unit 2 Upper Shell Forging material were reduced to 65% of the original values in order to estimate the initial USE values associated with the weak direction.

The weld certification records for the Sequoyah Unit 1 extended beltline weld (heat # 25006)reports only six Charpy V-notch impact energy values at a single test temperature (10°F) with no reported shear data. No other Charpy impact energy information is available for this weld heat.In accordance with Section B.1.1(4) of NUREG-0800 Branch Technical Position 5-3, this test temperature was used as an estimate of the initial RTNDT since at least 45 ft-lbs was obtained.Furthermore, in absence of USE data for weld heat # 25006, the weld heat # 25295 test results from the first surveillance capsule withdrawn from Sequoyah Unit 1 were used in accordance with Section B.1.2 of NUREG-0800 Branch Technical Position 5-3. Weld heat # 25295 is a Rotterdam weld of the same type (SMIT 40 with SMIT 89 flux). All surveillance weld data points that achieved greater than 95% shear in Table 5-2 of WCAP-10340, Revision 1 (Reference

13) were averaged to calculate the USE value for weld heat # 25006 based on results from the first capsule tested.Similarly, the weld certification records for the Sequoyah Unit 2 extended beltline welds (heat #721858) reports only three Charpy V-notch impact energy values at a single test temperature (107F) with no reported shear data. No other Charpy impact energy information is available for this weld heat. In accordance with Section B. 1.1(4) of NUREG-0800 Branch Technical Position 5-3, this test temperature was used as an estimate of the initial RTNDT since at least 45 ft-lbs was obtained.

Furthermore, in absence of USE data for weld heat # 721858, the lowest initial USE value from all of the Sequoyah Units 1 and 2 welds was conservatively applied to heat # 721858.Chemistry Factor Values The chemistry factor (CF) values were calculated using Positions 1.1 and 2.1 of Regulatory Guide 1.99, Revision 2 (Reference 14). Position 1.1 uses Tables 1 and 2 from the Regulatory Guide along with the best-estimate copper and nickel weight percents, which are presented in Tables 3-1 and 3-2 of this report for Sequoyah Units 1 and 2, respectively.

Position 2.1 uses the surveillance capsule data from all capsules withdrawn and tested to date. The calculated fluence values at the surveillance capsule locations are provided in Tables 2-6 and 2-7 and are used to determine the CFs in Tables 3-3 and 3-4 for Sequoyah Units 1 and 2, respectively.

Tables 3-5 and 3-6 summarize the Positions 1.1 and 2.1 CF values determined for the Sequoyah Units 1 and 2 RPV beltline and extended beltline materials.

WCAP-17539-NP March 2012 Revision 0 3-4 Westinghouse Non-Proprietary Class 3 ak Upper Shell Forging 06 Upper Shell to Intermediate Shell Circumferential Weld W06 4 1 -Intermediate Shell Forging 05 Intermediate Shell to Lower Shell Circumferential Weld W05.1 -Lower Shell Forging 04 04-Lower Shell to Bottom Head Ring Weld W04 Bottom Head Ring 03 Figure 3-1 RPV Material Identification for Sequoyah Units 1 and 2 WCAP-17539-NP March 2012 Revision 0 Westinghouse Non-Proprietary Class 3 3-5 Table 3-1 Best-Estimate Cu, Ni, P, and Mn Weight Percents, Initial RTNDT Values, and Initial USE Values for the Sequoyah Unit 1 RPV Beltline and Extended Beltline Materials Chemical Composition Fracture Toughness Chemial CmpostionProperties RPV Material(a)

Cu Ni P Mn Initial Initial USE (Wt. %) (Wt. %) (Wt. %) (Wt. %) RTNDT (b) (OF) (ft-lb)Reactor Vessel Beltline Materials(c)

Intermediate Shell (IS) Forging 05 0.15 0.86 0.011 0.70 40 79 (Heat # 980807/281489)

Lower Shell (LS) Forging 04 0.13 0.76 0.015 0.62 73 72 (Heat # 980919/281587)

IS to LS Circ. Weld W05 0.35 0.11 0.021 1.47 -40 113 (Heat # 25295) 1_1_1 Sequoyah Unit 1 Surveillance 0.39 0.11 0.021 1.40 Weld (Heat # 25295) 1 --- ---Reactor Vessel Extended Beltline Materials Upper Shell (US) Forging 06 (Heat H 9 8 0 9 5 0/2 8 2 7 5 8)(d) 0.16 0.89 0.011 0.70 23 83 Bottom Head Ring 03 (Heat # 9 8 1 1 7 7/2 8 8 8 7 2)(d) 0.16 0.77 0.016 0.73 5 64 US to IS Circ. Weld W06 0.17(e) 1.0(e) 0.013(e) 1.90(e) 10(f 78(f (Heat # 25006)LS to Bottom Head Ring Weld 113 W04 (Heat # 2 5 2 9 5)(g) 0.35 0.11 0.021 1.47 -40 Notes: (a) The heat numbers for the forging materials are the charge numbers taken from the CMTR. Note that the heat numbers listed for these forging materials in the Sequoyah Unit 1 TS Bases Table B 3/4.4-1 are the ingot numbers from the CMTR.(b) Initial RTNDT (RTNDT(U))

values are based on measured data for all beltline and extended beltline materials.(c) Except for the best-estimate P and Mn weight percent values, the beltline material properties were taken from WCAP-15293, Revision 2 (Reference 7). The weight percent P and Mn values for the beltline forging materials are based on Sequoyah Unit 1 CMTR data. The weight percent P and Mn values for the beltline and surveillance weld materials were determined using Rotterdam weld certification records as well as WCAP-8233 (Reference 8).(d) Except for the weight percent copper values, the chemical compositions for the extended beltline forging materials are based on Sequoyah Unit 1 CMTR data. No weight percent copper values were reported in the CMTRs for the extended beltline forging materials; therefore, the maximum weight percent copper value for A508 Class 2 forging materials is conservatively applied based on the generic data provided in Appendix G of the Oak Ridge National Laboratory Report (Reference 6). The initial fracture toughness properties are based on the data contained in Table B 3/4.4-1 of the Unit 1 TS Bases, and in accordance with Section B.1 of NUREG-0800 Branch Technical Position 5-3. Note that the USE value for the Bottom Head Ring has been reduced to 65% of the USE value associated with the strong orientation in order to approximate the value associated with the weak orientation.(e) Except for the weight percent nickel, the chemical compositions were taken from a chemical analysis performed on the weld wire (heat # 25006) included in the Rotterdam weld certification records. No weight percent nickel value was reported in the weld files for heat # 25006; therefore, a value of 1.0 was conservatively assumed per 10 CFR 50.61.WCAP- 17539-NP March 2012 Revision 0 3-6 Westinghouse Non-Proprietary Class 3 (f) The initial RTNDT was determined using all available measured data for heat # 25006 and the method described in Section B.1.1(4) of NUREG-0800 Branch Technical Position 5-3. In absence of USE data for weld heat # 25006, weld heat # 25295 test results from the first surveillance capsule withdrawn from Sequoyah Unit 1 were used in accordance with Section B. 1.2 of NUREG-0800 Branch Technical Position 5-3 to conservatively estimate the initial USE value for weld heat # 25006.(g) The LS to Bottom Head Ring Weld was fabricated using the same weld wire heat number and flux type as the IS to LS Circ. Weld. Therefore, the chemical and fracture toughness properties of the IS to LS circumferential weld are applied to this weld material.WCAP-17539-NP March 2012 Revision 0 Westinghouse Non-Proprietary Class 3 3-7 Table 3-2 Best-Estimate Cu, Ni, P, and Mn Weight Percents, Initial RTNDT Values, and Initial USE Values for the Sequoyah Unit 2 RPV Beltline and Extended Beltline Materials Chemical Composition Fracture Toughness Properties RPV Material(a)

Initial Cu Ni P Mn Initial (Wt. %) (Wt. %) (Wt. %) (Wt. %) RTNDT (b) (OF) US I (ft-lb)Reactor Vessel Belitline Materials(c)

Intermediate Shell (IS) Forging 05 0.13 0.76 0.014 0.70 10 93 (Heat # 288757/981057)

Lower Shell (LS) Forging 04 0.14 0.76 0.012 0.68 -22 100 (Heat # 990469/293323)

IStoLSCirc.WeldW 05 0.12 0.11 0.016 1.50 -4 102 (Heat # 4278)Sequoyah Unit 2 Surveillance Weld 0.13 0.11 0.016 1.50 (Heat # 4278) 1 Reactor Vessel Extended Beltline Materials Upper Shell (US) Forging 06 (Heat # 9 8 1 2 0 1/2 8 5 8 4 9)(d) 0.16 0.84 0.016 0.72 5 68 Bottom Head Ring 03 (Heat # 9 8 1 1 7 7/2 8 8 8 7 2)(d) 0.16 0.77 0.016 0.73 5 64 US to IS Circ. Weld W06 0.08(e) 1.0(e) 0.019(e) 1.52(e) 10(f 78(0 (Heat # 721858)LS to Bottom Head Ring Weld W04 0.08(e) 1.0(e) 0.019(e) 1.52(e) 1 0 (f) 78(0 (Heat # 721858)Notes: (a) The heat numbers for the forging materials are the charge numbers taken from the CMTR. Note that the heat numbers listed for these forging materials in the Sequoyah Unit 2 TS Bases Table B 3/4.4-1 are the ingot numbers from the CMTR.(b) Initial RTNDT (RTNDT(U))

values are based on measured data for all beltline and extended beltline materials.(c) Except for the best-estimate P and Mn weight percent values, the beltline material properties were taken from WCAP-15321, Revision 2 (Reference 10). The weight percent P and Mn values for the beltline forging materials are based on Sequoyah Unit 2 CMTR data. The weight percent P and Mn values for the beltline and surveillance weld materials were determined using Rotterdam weld certification records as well as WCAP-8513 (Reference 11).(d) Except for the weight percent copper values, the chemical compositions for the extended beltline forging materials are based on Sequoyah Unit 2 CMTR data. No weight percent copper values were reported in the CMTRs for the extended beltline forging materials; therefore, the maximum weight percent copper value for A508 Class 2 forging materials is conservatively applied based on the generic data provided in Appendix G of the Oak Ridge National Laboratory Report (Reference 6). The initial fracture toughness properties are based on the data contained in Table B 3/4.4-1 of the Unit 2 TS Bases, and in accordance with Section B.1 of NUREG-0800 Branch Technical Position 5-3. Note that these USE values have been reduced to 65% of the USE value associated with the strong orientation in order to approximate the values associated with the weak orientation.(e) Except for the weight percent nickel, the chemical compositions were taken from chemical analyses performed on the weld wire (heat # 721858) along with the as-deposited weld included in the Rotterdam weld certification records. No WCAP-17539-NP March 2012 Revision 0 3-8 Westinghouse Non-Proprietary Class 3 weight percent nickel value was reported in the weld files for heat # 721858; therefore, a value of 1.0 was conservatively assumed per 10 CFR 50.61.(f) The initial RTNDT was determined using all available measured data for heat # 721858 and the method described in Section B. 1.l1(4) of NUREG-0800 Branch Technical Position 5-3. In absence of USE data for weld heat # 721858, the lowest initial USE value from all the Sequoyah Units 1 and 2 welds was conservatively assumed to be the initial USE value for heat # 721858. This initial USE value of 78 ft-lbs is associated with the Unit 1 US to IS circ. weld.WCAP-17539-NP March 2012 Revision 0 Westinghouse Non-Proprietary Class 3 3-9 Table 3-3 Calculation of Position 2.1 CF Values using Sequoyah Unit I Surveillance Capsule Test Results Capsule RPV Material Capsule Fluenc6(a)

FF(b) ARTNDT(c)

FF*ARTNDT FF 2 (x10 1 9 n/cm2, (OF) (OF)E > 1.0 MeV)T 0.241 0.615 67.52 41.52 0.378 LS Forging 04 U 0.693 0.897 109.7 98.42 0.805 (Tangential)

X 1.16 1.041 145.12 151.13 1.085 Y 1.97 1.185 129.87 153.92 1.405 T 0.241 0.615 50.59 31.11 0.378 LS Forging 04 U 0.693 0.897 67.59 60.64 0.805 (Axial) X 1.16 1.041 103.34 107.62 1.085 Y 1.97 1.185 133.35 158.04 1.405 SUM: 802.39 7.344 CFLS Forging 04 E(FF

ARTNiDT) + E(FF 2) = (802.39) + (7.344) = 109.3°F T 0.241 0.615 115.01 70.72 0.378 (127.79)Surveillance Weld U 0.693 0.897 130.43 117.01 0.805 Metal (144.92)(Heat # 25295) X 1.16 1.041 143.12 149.05 1.085_____ __ _ ____ _____ __ _ ____ (159.02) 190 .8 Y 1.97 1.185 147.42 174.72 1.405 (163.8)SUM: 511.50 3.672 CFHeat#2 5 2 9 5= X(FF

ARTNDT) + I(FF 2) = (511.50) + (3.672) = 139.3°F Notes: (a)(b)(c)f = calculated fluence values from Table 2-6.FF = fluence factor = t0.28-10.1°g(t).

ARTNDT (7F) values are the measured 30 ft-lb shift values taken from Table 5-10 of WCAP-15224, Revision 0 (Reference 15). The surveillance weld ARTNDT values have been adjusted by the ratio of 0.90 to account for the chemistry differences between the vessel weld material and the surveillance weld material.Pre-adjusted values are listed in parentheses.

Ratio = CFvessel Weld / CFsurv. Weld = 161.3'F / 178.7'F = 0.90.WCAP-17539-NP March 2012 Revision 0 3-10 Westinghouse Non-Proprietary Class 3 Table 3-4 Calculation of Position 2.1 CF Values using Sequoyah Unit 2 Surveillance Capsule Test Results Capsule Fluence(a)

FF(b) ARTNDT(c)

FF*ARTNDT FF 2 RPV Material Capsule (x10 1 9 n/cm2, (OF) (OF)E > 1.0 MeV)T 0.244 0.618 63.65 39.33 0.382 IS Forging 05 U 0.654 0.881 79.31 69.87 0.776 (Tangential)

X 1.16 1.041 85.7 89.25 1.085 Y 2.02 1.192 134.12 159.83 1.420 T 0.244 0.618 48.73 30.11 0.382 IS Forging 05 U 0.654 0.881 66.06 58.20 0.776 (Axial) X 1.16 1.041 110.04 114.60 1.085 Y 2.02 1.192 89.21 106.31 1.420 SUM: 667.51 7.325 CFIs Forging 05 = X(FF *ARTNDT) .(FF 2) = (667.51) -(7.325) = 91.1°F T 0.244 0.618 69.34 42.85 0.382 (74.56)121.25 Surveillance Weld U 0.654 0.881 106.82 0.776 Metal (130.38)(Heat # 4278) X 1.16 1.041 41.12 42.83 1.085_______ ______ _______ ______ (44.22) _ _ _ _ _ _ _ _Y 2.02 1.192 80.83 96.32 1.420 (86.91) 1 SUM: 288.82 3.663 CFHeat # 4 2 7 8 = X(FF

ARTNDT) -EFF2) = (288.82) (3.663) = 78.9°F Notes: (a)(b)(c)f = calculated fluence values from Table 2-7.FF = fluence factor = 0.2 8-0.10l og(f)).ARTNDT (0 F) values are the measured 30 fi-lb shift values taken from Table 5-10 of WCAP-15320, Revision 0 (Reference 16). The surveillance weld ARTNDT values have been adjusted by the ratio of 0.93 to account for the chemistry differences between the vessel weld material and the surveillance weld material.Pre-adjusted values are listed in parentheses.

Ratio = CFvessel Weld / CFUs. Weld = 63.0'F / 67.9'F = 0.93.WCAP-17539-NP March 2012 Revision 0 Westinghouse Non-Proprietary Class 3 3-11 Table 3-5 Summary of the Sequoyah Unit 1 RPV Beltline and Extended Beltline Material Chemistry Factor Values based on Regulatory Guide 1.99, Revision 2, Position 1.1 and Position 2.1 RPV Material Chemistry Factor (IF)Position 1.1 Position 2.1 Reactor Vessel Beltline Materials IS Forging 05 115.6 ---LS Forging 04 95.0 109.3 IS to LS Circ. Weld W05 161.3 139.3 (Heat # 25295)Sequoyah Unit 1 Surveillance Weld 178.7 (Heat # 25295) 178.7 Reactor Vessel Extended Beltline Materials US Forging 06 123.9 ---Bottom Head Ring 03 122.3 ---US to IS Circ. Weld W06 (Heat # 25006) 207.0 LS to Bottom Head Ring Weld W04 161.3 139.3 (Heat # 25295) 161.3 139.3 Table 3-6 Summary of the Sequoyah Unit 2 RPV Beltline and Extended Beltline Material Chemistry Factor Values based on Regulatory Guide 1.99, Revision 2, Position 1.1 and Position 2.1 RPV Material Chemistry Factor (IF)Position 1.1 Position 2.1 Reactor Vessel Beltline Materials IS Forging 05 95.0 91.1 LS Forging 04 104.0 -- -IS to LS Circ. Weld W05 63.0 78.9 (Heat # 4278) 63.0 78.9 Sequoyah Unit 2 Surveillance Weld 67.9 (Heat # 4278) 67.9 Reactor Vessel Extended Beltline Materials US Forging 06 123.4 ---Bottom Head Ring 03 122.3 US to IS Circ. Weld W06 108.0 (Heat # 721858) 108.0 LS to Bottom Head Ring Weld W04 108.0 ---(Heat # 721858) 1 _ _ _1 WCAP- 17539-NP March 2012 Revision 0 Westinghouse Non-Proprietary Class 3 4-1 4 PRESSURIZED THERMAL SHOCK A limiting condition on RPV integrity known as Pressurized Thermal Shock (PTS) may occur during a severe system transient such as a loss-of-coolant accident (LOCA) or steam line break.Such transients may challenge the integrity of the RPV under the following conditions:

severe overcooling of the inside surface of the vessel wall followed by high repressurization; significant degradation of vessel material toughness caused by radiation embrittlement; and the presence of a critical-size defect anywhere within the vessel wall.In 1985, the U.S. NRC issued a formal ruling (10 CFR 50.61) on PTS (Reference

9) that established screening criteria on PWR vessel embrittlement, as measured by the maximum reference nil ductility transition temperature in the limiting beltline component at the end-of-license, termed RTPTs. RTP-s screening values were set by the U.S. NRC for beltline axial welds, forgings or plates, and for beltline circumferential weld seams for plant operation to the end of plant license. All domestic PWR vessels have been required to evaluate vessel embrittlement in accordance with the criteria through the end-of-license.

The U.S. NRC revised 10 CFR 50.61 in 1991 and 1995 to change the procedure for calculating radiation embrittlement.

These revisions make the procedure for calculating the reference temperature for pressurized thermal shock (RTpTs) values consistent with the methods given in Regulatory Guide 1.99, Revision 2 (Reference 14).These accepted methods were used with the surface fluence of Section 2 to calculate the following RTPTs values for the Sequoyah Units 1 and 2 RPV materials at 52 EFPY (EOLE). The EOLE RTPTs calculations are summarized below in Tables 4-1 and 4-2 for Units 1 and 2, respectively.

PTS Conclusion The Sequoyah Unit 1 limiting RTPTS value for forging materials at 52 EFPY is 227.9°F (see Table 4-1), which corresponds to the Lower Shell Forging 04 using credible surveillance data.The limiting RTPTS value for the Unit 1 circumferentially oriented welds at 52 EFPY is 163.6°F (see Table 4-1), which corresponds to the IS to LS Circumferential Weld W05 using credible surveillance data.The Sequoyah Unit 2 limiting RTPTs value for forging materials at 52 EFPY is 142.3°F (see Table 4-2), which corresponds to the Lower Shell Forging 04. The limiting RTPTs value for the Unit 2 circumferentially oriented welds at 52 EFPY is 150.7°F (see Table 4-2), which corresponds to the IS to LS Circumferential Weld W05 using non-credible surveillance data.Therefore, all of the beltline and extended beltline materials in the Sequoyah Units 1 and 2 reactor vessels are below the RTPTS screening criteria values of 270'F, for forgings, and 300'F, for circumferentially oriented welds through EOLE (52 EFPY).The Alternate PTS Rule (10 CFR 50.61a (Reference 17)) was published in the Federal Register by the NRC in 2010. This alternate rule is less restrictive than the Mandatory PTS Rule (10 CFR 50.61) and is intended to be used for situations where the 10 CFR 50.61 criteria cannot be met.WCAP-17539-NP March 2012 Revision 0 4-2 Westinghouse Non-Proprietary Class 3 Sequoyah Units 1 and 2 currently meet the criteria for the Mandatory PTS Rule through EOLE and therefore do not need to utilize the Alternate PTS Rule at this time.WCAP- 17539-NP March 2012 Revision 0 Westinghouse Non-Proprietary Class 3 4-3 Westinghouse Non-Proprietary Class 3 4-3 Table 4-1 Calculation of Sequoyah Unit 1 RTPTS Values for 52 EFPY (EOLE) at the Clad/Base Metal Interface RPV Material (a) Fluence(b)

FF() TNT(U)(d)

ARTNDT(e) cru(d) cArf) Margin RTPTS (OF) (x1019 n/cm 2) J (O F) (OF) (OF) (OF) (OF) (OF)Reactor Vessel Beltline Materials IS Forging 05 115.6 2.66 1.2616 40 145.8 0 17.0 34.0 219.8 LS Forging 04 95.0 2.66 1.2616 73 119.8 0 17.0 34.0 226.8 Using credible surveillance data 109.3 2.66 1.2616 73 137.9 0 8.5 17.0 227.9 IStoLSCirc.WeldW05 161.3 2.65 1.2607 -40 203.3 0 28.0 56.0 219.3 ( H e a t f# 2 5 2 9 5 ) --------------------------------------------------. -------------- --. ------------------------------. --------- ------. ---- ---------. ..............Using credible surveillance data 139.3 2.65 1.2607 -40 175.6 0 14.0 28.0 163.6 Reactor Vessel Extended Beltline Materials US Forging 06 123.9 0.0584 0.3180 23 39.4 0 17.0 34.0 96.4 Bottom Head Ring 03 122.3 0.336 0.6997 5 85.6 0 17.0 34.0 124.6 UStoISCirc.WeldW06 207.0 0.0584 0.3180 10 65.8 0 28.0 56.0 131.8 (Heat # 25006)LStoBottomHeadRing Weld W04 161.3 0.336 0.6997 -40 112.9 0 28.0 56.0 128.9..............(HI! eat _#_ 2_5_29.5.)

Using credible surveillance data 139.3 0.336 0.6997 -40 9°7.5 0 14.0 28.0 85.5 Notes: (a)(b)(c)(d)(e)(f)Data taken from Table 3-5 of this report.Data taken from Tables 2-1 and 2-2 of this report.FF = fluence factor = f.28-0.10log(f)).

Initial RTNDT values were taken from Table 3-1 of this report and are based on measured data for all of the materials.

Note that c¢j = 0°F for measured values.ARTNDT = CF

FF.Per Appendix A of this report, the surveillance data of the forging and weld were deemed credible.

Per the guidance of 10 CFR 50.61, the base metal CA =17 0 F for Position 1.1 and 0 A = 8.5°F for Position 2.1 with credible surveillance data; the weld metal oA = 28°F for Position 1.1 and GA = 14'F for Position 2.1 with credible surveillance data. However, GA need not exceed 0.5*ARTNDT.

WCAP-17539-NP March 2012 Revision 0 4-4 Westinghouse Non-Proprietary Class 3 Table 4-2 Calculation of Sequoyah Unit 2 RTPTS Values for 52 EFPY (EOLE) at the Clad/Base Metal Interface CF(a) EOLE (d 1 Ma RPV Material Fluence(b)

FF(C) RTNDT(U)(d)

ARTNDT(e) cru(d) TA( Margin RTPTS (OF) (x1019 n/cm2) (OF) (oF) (OF) (OF) (OF) (OF)Reactor Vessel Beltline Materials-I-S__F-or-gin-g 05 95.0 2.57 1.2531 10 119.0 0 17.0 34.0 163.0 Using credible surveillance data 91.1 2.57 1.2531 10 114.2 0 8.5 17.0 141.2 LS Forging 04 104.0 2.57 1.2531 -22 130.3 0 17.0 34.0 142.3 IS to LS Circ. Weld W05 63.0 2.55 1.2511 -4 78.8 0 28.0 56.0 130.8 (Heat if 4278)Using non-credible surveillance data 78.9 2.55 1.2511 -4 98.7 0 28.0 56.0 150.7 Reactor Vessel Extended Beltline Materials US Forging 06 123.4 0.0552 0.3087 5. 38.1 0 17.0 34.0 77.1 Bottom Head Ring 03 122.3 0.316 0.6837 5 83.6 0 17.0 34.0 122.6 US to IS Circ. Weld W06 108.0 0.0552 0.3087 10 33.3 0 16.7 33.3 76.7 (Heat # 721858)LS to Bottom Head Ring Weld W04 108.0 0.316 0.6837 10 73.8 0 28.0 56.0 139.8 (Heat # 721858)Notes: (a)(b)(c)(d)(e)(f)Data taken from Table 3-6 of this report.Data taken from Tables 2-3 and 2-4 of this report.FF = fluence factor = f(0.28-0.10*log(f).

Initial RTNDT values were taken from Table 3-2 of this report and are based on measured data for all of the materials.

Note that cau = 0°F for measured values.ARTNDT = CF

FF.Per Appendix A of this report, the surveillance data of the forging was deemed credible and the surveillance data of the weld was deemed non-credible.

Per the guidance of 10 CFR 50.61, the base metal cA = 17'F for Position 1.1 and aA = 8.5°F for Position 2.1 with credible surveillance data; the weld metal 0 A =28°F for Position 1.1 and for Position 2.1 with non-credible surveillance data. However, CFA need not exceed 0.5*ARTNDT.

WCAP-17539-NP March 2012 Revision 0 Westinghouse Non-Proprietary Class 3 5-1 5 UPPER-SHELF ENERGY The decrease in Charpy upper-shelf energy (USE) is associated with the determination of acceptable RPV toughness during the license renewal period when the vessel is exposed to additional irradiation.

The requirements on USE are included in 10 CFR 50, Appendix G (Reference 18). 10 CFR 50, Appendix G requires utilities to submit an analysis at least 3 years prior to the time that the USE of any RPV material is predicted to drop below 50 ft-lb, as measured by Charpy V-notch specimen testing.There are two methods that can be used to predict the decrease in USE with irradiation, depending on the availability of credible surveillance capsule data as defined in Regulatory Guide 1.99, Revision 2. For vessel beltline materials that are not in the surveillance program or are non-credible, the Charpy USE (Position 1.1) is assumed to decrease as a function of fluence and copper content, as indicated in Regulatory Guide 1.99, Revision 2 (Reference 14).When two or more credible surveillance sets become available from the reactor, they may be used to determine the Charpy USE of the surveillance material.

The surveillance data are then used in conjunction with the Regulatory Guide to predict the change in USE (Position 2.2) of the RPV material due to irradiation.

The 52 EFPY (EOLE) Position 1.2 USE values of the vessel materials can be predicted using the corresponding 1/4T fluence projection, the copper content of the materials, and Figure 2 in Regulatory Guide 1.99, Revision 2.The predicted Position 2.2 USE values are determined for the reactor vessel materials that are contained in the surveillance program by using the reduced plant surveillance data along with the corresponding 1/4T fluence projection.

The reduced plant surveillance data was obtained from Table 5-10 of WCAP-15224 (Reference

15) and WCAP-15320 (Reference

16) for Sequoyah Units 1 and 2, respectively.

The surveillance data was plotted on Regulatory Guide 1.99, Revision 2, Figure 2 (see Figures 5-1 and 5-2 of this report) using the updated surveillance capsule fluence values documented in Tables 2-6 and 2-7 of this report for Sequoyah Units 1 and 2, respectively.

This data was fitted by drawing a line parallel to the existing lines as the upper bound of all the surveillance data. These reduced lines were used instead of the existing lines to determine the Position 2.2 EOLE USE values.The projected USE values were calculated to determine if the Sequoyah Units 1 and 2 beltline and extended beltline materials remain above the 50 ft-lb limit at 52 EFPY (EOLE). These calculations are summarized in Tables 5-1 and 5-2.WCAP- 17539-NP March 2012 Revision 0 5-2 Westinghouse Non-Proprietary Class 3 USE Conclusion For Sequoyah Unit 1, the limiting USE value at 52 EFPY is 52.5 ft-lb (see Table 5-1); this value corresponds to the Bottom Head Ring 03. For Sequoyah Unit 2, the limiting USE value at 52 EFPY is 53.1 ft-lb (see Table 5-2); this value corresponds to Bottom Head Ring 03. Therefore, all of the beltline and extended beltline materials in the Sequoyah Units 1 and 2 reactor vessels are projected to remain above the USE screening criterion value of 50 ft-lb (per 10 CFR 50 Appendix G) through EOLE (52 EFPY).WCAP- 17539-NP March 2012 Revision 0 Westinghouse Non-Proprietary Class 3 5-3 Table 5-1 Predicted USE Values at 52 EFPY (EOLE) for Sequoyah Unit 1 RPV Material Reactor Vessel Beltline Materials IS Forging 05 0.15 1.602 79 27 57.7 LS Forging 04 0.13 1.602 72 25 54.0 Using surveillance data 0.13 1.602 72 26 (d) .533.IStoLSCirc.WeldW 05 0.35 1.596 113 46 61.0 (Heat # 25295)Using surveillance data 0.35 1.596 113 61.0 Reactor Vessel Extended Beltline Materials US Forging 06 0.16 0.035 83 12 73.0 Bottom Head Ring 03 0.16 0.202 64 18 52.5 US to IS Circ. Weld W06 0.17 0.035 78 15 66.3 (Heat # 25006)LS to Bottom Head Ring Weld W04 0.35 0.202 113 34 74.6... ... ... ..(H e a t # 2 5 2 9 5 )I --------------------------------------------------Using surveillance data 0.35 0.202 113 29(d) 80.2 Notes: (a) Data taken from Table 3-1 of this report.(b) The 1/4T fluence was calculated using the Regulatory Guide 1.99, Revision 2 correlation, and the Sequoyah Unit 1 reactor vessel wall thickness of 8.45 inches.(c) Unless otherwise noted, percentage USE' decrease values are based on Position 1.2 of Regulatory Guide 1.99, Revision 2, and were calculated by plotting the 1/4T fluence values on Figure 2 of the Regulatory Guide. The percent USE decrease values that corresponded to each material's specific Cu wt. % value were determined using interpolation between the existing Weld or Base Metal lines on Figure 2.(d) Percentage USE decrease is based on Position 2.2 of Regulatory Guide 1.99, Revision 2 using data from Table 5-10of WCAP-15224 (Reference 15). Credibility Criterion 3 in the Discussion section of Regulatory Guide 1.99, Revision 2, indicates that even if the surveillance data are not considered credible for determination of ARTNDT, "they may be credible for determining decrease in upper-shelf energy if the upper shelf can be clearly determined, following the definition given in ASTM E 185-82." Regulatory Guide 1.99, Revision 2, Position 2.2 indicates that an upper-bound line drawn parallel to the existing lines (in Figure 2 of the Guide) through the surveillance data points should be used in preference to the existing graph lines for determining the decrease in USE.WCAP-17539-NP March 2012 Revision 0 5-4 Westinghouse Non-Proprietary Class 3 Figure 5-1 Regulatory Guide 1.99, Revision 2 Predicted Decrease in Upper-Shelf Energy as a Function of Copper and Fluence for Sequoyah Unit 1 WCAP-17539-NP March 2012 Revision 0 Westinghouse Non-Proprietary Class 3 5-5 Table 5-2 Predicted USE Values at 52 EFPY (EOLE) for Sequoyah Unit 2 Projected Cuba) EOLE 1/4T Initial USE EOLE Cu~a)UUS RPV Material .%) Fluence(b)

USE(a) Decrease(c)

USE (xlO09 n/cm2) (ft-lb) (%) (ft-lb)Reactor Vessel Beltline Materials IS Forging 05 0.13 1.548 93 25 69.8 Using surveillance data 0.13 1.548 93 21id) 73.5 LS Forging 04 0.14 1.548 100 26 74.0 IS to LS Circ. Weld W05 0.12 1.536 102 29 72.4-- -- -- -- -- -(H eat # 4 2 7 8)_ ------------------------------------------------------------------------------------------------------.

Using surveillance data 0.12 1.536 102 38(d) 63.2 Reactor Vessel Extended Beltline Materials US Forging 06 0.16 0.033 68 12 59.8 Bottom Head Ring 03 0.16 0.190 64 17 53.1 US to IS Circ. Weld W06 0.08 0.033 78 10 70.2 (Heat # 721858)LS to Bottom Head Ring Weld W04 0.08 0.190 78 15 66.3 (Heat # 721858)Notes: (a) Data taken from Table 3-2 of this report.(b) The 1/4T fluence was calculated using the Regulatory Guide 1.99, Revision 2 correlation, and the Sequoyah Unit 2 reactor vessel wall thickness of 8.45 inches.(c) Unless otherwise noted, percentage USE decrease values are based on Position 1.2 of Regulatory Guide 1.99, Revision 2, and were calculated by plotting the 1/4T fluence values on Figure 2 of the Regulatory Guide. The percent USE decrease values that corresponded to each material's specific Cu wt. % value were determined using interpolation between the existing Weld or Base Metal lines on Figure 2.(d) Percentage USE decrease is based on Position 2.2 of Regulatory Guide 1.99, Revision 2 using data from Table 5-10 of WCAP-15320 (Reference 16). Credibility Criterion 3 in the Discussion section of Regulatory Guide 1.99, Revision 2, indicates that even if the surveillance data are not considered credible for determination of ARTNDT, "they may be credible for determining decrease in upper-shelf energy if the upper shelf can be clearly determined, following the definition given in ASTM E 185-82." Regulatory Guide 1.99, Revision 2, Position 2.2 indicates that an upper-bound line drawn parallel to the existing lines (in Figure 2 of the Guide) through the surveillance data points should be used in preference to the existing graph lines for determining the decrease in USE.WCAP-17539-NP March 2012 Revision 0 5-6 Westinghouse Non-Proprietary Class 3 5-6 Westinghouse Non-Proprietary Class 3* Surveillance Material:

IS Forging 05 A Surveillance Material:

Weld Heat# 4278 100 wl ine forging line U.'CI 0.0.1m (L 10 1 1.00E4-17 1.OOE+18 1.OOE+19 1.00E+20 Neutron Fluence, n/cm 2 (E > 1 MeV)Figure 5-2 Regulatory Guide 1.99, Revision 2 Predicted Decrease in Upper-Shelf Energy as a Function of Copper and Fluence for Sequoyah Unit 2 WCAP- 17539-NP March 2012 Revision 0 Westinghouse Non-Proprietary Class 3 6-1 6 HEATUP AND COOLDOWN PRESSURE-TEMPERATURE LIMIT CURVES APPLICABILITY Heatup and cooldown limit curves are calculated using the most limiting value of RTNDT (reference nil ductility transition temperature) corresponding to the limiting material in the beltline region of the RPV. The most limiting RTNDT of the material in the core (beltline) region of the RPV is determined by using the unirradiated RPV material fracture toughness properties and estimating the irradiation-induced shift (ARTNDT).RTNDT increases as the material is exposed to fast-neutron irradiation; therefore, to find the most limiting RTNDT at any time period in the reactor's life, ARTNDT due to the radiation exposure associated with that time period must be added to the original unirradiated RTNDT. Using the adjusted reference temperature (ART) values, pressure-temperature (P-T) limit curves are determined in accordance with the requirements of 10 CFR Part 50, Appendix G (Reference 18), as augmented by Appendix G to Section XI of the ASME Boiler and Pressure Vessel Code (Reference 19).According to Section 4.2.2.1.3 of NUREG-1800, Revision 2 (Reference 20), P-T limit curves for the period of extended operation (52 EFPY) do not need to be submitted as part of the Sequoyah License Renewal Application since P-T limit curves are available through the current license (32 EFPY). However, new P-T limit curves will need to be developed prior to the expiration of the current curves as specified in the Sequoyah licensing basis. Therefore, only the applicability of the existing P-T limit curves is assessed in this report.The P-T limit curves for normal heatup and cooldown of the primary reactor coolant system for Sequoyah Units 1 and 2 were previously developed in WCAP-15293, Revision 2 (Reference 7)and WCAP-15321, Revision 2 (Reference

10) for 32 EFPY. The existing 32 EFPY P-T limit curves are based on the limiting beltline material ART values, which are influenced by both the fluence and the initial material properties of that material.

The Sequoyah Units 1 and 2 P-T limit curves were developed by calculating ART values utilizing the clad/base metal interface fluence that corresponded to each reactor vessel beltline material.To confirm the applicability of the P-T limit curves developed in WCAP-15293, Revision 2 (Reference

7) for Sequoyah Unit 1 and in WCAP-15321, Revision 2 (Reference

10) for Sequoyah Unit 2, the limiting reactor vessel material ART values with consideration of the updated TLAA fluence values must be shown to be less than the limiting beltline material ART values used in development of the existing 32 EFPY P-T limit curves contained in References 7 and 10. The Regulatory Guide 1.99, Revision 2 (Reference

14) methodology was used along with the surface fluence of Section 2 to calculate ART values for the Sequoyah Units 1 and 2 reactor vessel materials at 32 EFPY and 52 EFPY. The ART calculations are summarized in Tables 6.1-1 through 6.1-4 for Sequoyah Unit 1 and in Tables 6.2-1 through 6.2-4 for Sequoyah Unit 2.WCAP-17539-NP March 2012 Revision 0 6-2 Westinghouse Non-, Proprietary Class 3 Existing P-T Limit Curves Applicability Conclusions Comparisons of the limiting ART values calculated as part of this RVI TLAA evaluation to those used in calculation of the existing P-T limit curves are contained in Tables 6.1-5 and 6.2-5 for Sequoyah Units 1 and 2, respectively.

With a re-evaluation of surveillance data credibility, a recalculation of the Position 2.1 chemistry factor values, and the consideration of TLAA fluence projections, the applicability of the Sequoyah Units 1 and 2 P-T limit curves may either remain unchanged or can be extended.

For more detailed conclusions, refer to Sections 6.1 and 6.2 below for Sequoyah Units 1 and 2, respectively.

WCAP-17539-NP March 2012 Revision 0 Westinghouse Non-Proprietary Class 3 6-3 6.1 SEQUOYAH UNIT 1 Tables 6.1-1 through 6.1-4 below summarize the 1/4T and 3/4T ART calculations for Sequoyah Unit 1. The limiting 32 EFPY and 52 EFPY ART values for Sequoyah Un'it 1 corresponds to the LS Forging 04 using credible surveillance data (Position 2.1).The applicability of the existing 32 EFPY P-T limit curves, contained in WCAP-15293, Revision 2 (Reference

7) for Sequoyah Unit 1, is evaluated by comparing the updated ART values contained in this section with those used in the Reference 7 calculations.

The existing 32 EFPY P-T limit curves for Sequoyah Unit 1 are based on the limiting beltline material ART values, which are influenced by both fluence and initial material properties of that material.

Using the TLAA fluence projections, the 1/4T and 3/4T ART values were recalculated in Tables 6.1-1 through 6.1-4 as part of this applicability evaluation for Sequoyah Unit 1. Since the capsule fluence values were also updated as part of the TLAA effort, the Position 2.1 chemistry factor values were revised in Section 3 of this report. Furthermore, the credibility evaluation conclusions contained in Appendix A of this report have changed (from non-credible to credible)for the Sequoyah Unit 1 surveillance weld and forging materials since the current P-T limit curves were developed.

The comparison of limiting ART values is contained in Table 6.1-5 for Sequoyah Unit 1.Table 6.1-5 below compares the TLAA limiting ART values at 32 EFPY and 52 EFPY to the limiting ART values used in development of the existing 32 EFPY P-T limit curves that are documented in WCAP-15293, Revision 2 (Reference 7). The limiting ART values used to develop the existing P-T limit curves are documented in Table 10 of Reference 7.The TLAA limiting ART values at 32 EFPY and 52 EFPY are bounded by the limiting ART values used to develop the existing 32 EFPY P-T limit curves. This is primarily due to the revised credibility evaluation, which is performed in Appendix A of this report, and updated fluence data of the Sequoyah Unit 1 surveillance capsules.

Therefore, the existing Sequoyah Unit 1 P-T limit curves may be deemed applicable through 52 EFPY.P-T Limits Applicability Conclusion For Sequoyah Unit 1, it is concluded that the existing 32 EFPY P-T limit curves do not require a reduction of the applicability date. Since the P-T limit curves remain valid through the original EFPY period, the current low temperature overpressure protection (LTOP) setpoints also remain applicable through 32 EFPY.Furthermore, based on the TLAA evaluation, Tennessee Valley Authority may instead choose to extend the applicability of the existing Sequoyah Unit 1 P-T limit curves. The new applicability date with consideration of the TLAA credibility and fluence evaluations is 52 EFPY. Note that an evaluation would have to be performed to increase the LTOP setpoints applicability period.WCAP-17539-NP March 2012 Revision 0 6-4 Westinghouse Non-Proprietary Class 3 Table 6.1-1 Calculation of the Sequoyah Unit 1 ART Values at the 1/4T Location for 32 EFPY ,F(a) 1/4T Fluence (b) () (d) M CF M24 1/4T RTNDT(U) (c) ARTNDT l(c) A Margin ART RPV Material (OF) (xl01 9 n/cm, FF(b) (OF) (OF) (OF) (OF) (0F) (OF)E > 1.0 MeV)IS Forging 05 115.6 1.042 1.0115 40 116.9 0 17.0 34.0 190.9 LS Forging 04 95.0 1.042 1.0115 73 96.1 0 17.0 34.0 203.1 Using credible surveillance data 109.3 1.042 1.0115 73 110.6 0 8.5 17.0 200.6 IS to LS Circ. Weld W05 (Heat # 25295) 161.3 1.036 1.0099 -40 162.9 0 28.0 56.0 178.9 Using credible surveillance data 139.3 1.036 1.0099 -40 140.7 0 14.0 28.0 128.7 Notes: (a) Data taken from Table 3-5 of this report.(b) The 1/4T fluence and FF were calculated using the Regulatory Guide 1.99, Revision 2 correlations, and the Sequoyah Unit 1 reactor vessel wall thickness of 8.45 inches.(c) Initial RTNDT values were taken from Table 3-1 of this report and are based on measured data for all of the materials.

Note that aT = 0°F for measured values.(d) Per Appendix A of this report, the surveillance data of the forging and weld were deemed credible.

Per the guidance of Regulatory Guide 1.99, Revision 2, the base metal OA = 17'F for Position 1.1 and aA = 8.5 0 F for Position 2.1 with credible surveillance data; the weld metal GA = 28°F for Position 1.1 and CA = 14'F for Position 2.1 with credible surveillance data. However, GA need not exceed 0.5*ARTNDT.

WCAP- 17539-NP March 2012 Revision 0 Westinghouse Non-Proprietary Class 3 6-5 Westinghouse Non-Proprietary Class 3 6-5 Table 6.1-2 Calculation of the Sequoyah Unit 1 ART Values at the 3/4T Location for 32 EFPY CF(a) 3/4T Fluenceb 3/4T RTNDT(U) (c) ARTNDT 6I(0 gA(d) Margin ART RPV Material -(OF) (xlO0 9 n/cm 2 , FF(b) (OF) (OF) (OF) (OF) (OF) (OF)E > 1.0 MeV)IS Forging 05 115.6 0.378 0.7309 40 84.5 0 17.0 34.0 158.5 LS Forging 04 95.0 0.378 0.7309 73 69.4 0 17.0 34.0 176.4 Using credible surveillance data 109.3 0.378 0.7309 73 79.9 0 8.5 17.0 169.9 IS to LS Circ. Weld W05 (Heat # 25295) 161.3 0.376 0.7293 -40 117.6 0 28.0 56.0 133.6 Using credible surveillance data 139.3 0.376 0.7293 -40 101.6 0 14.0 28.0 89.6 Notes: (a) Data taken from Table 3-5 of this report.(b) The 3/4T fluence and FF were calculated using the Regulatory Guide 1.99, Revision 2 correlations, and the Sequoyah Unit 1 reactor vessel wall thickness of 8.45 inches.(c) Initial RTNDT values were taken from Table 3-1 of this report and are based on measured data for all of the materials.

Note that (71 0°F for measured values.(d) Per Appendix A of this report, the surveillance data of the forging and weld were deemed credible.

Per the guidance of Regulatory Guide 1.99, Revision 2, the base metal 0 A = 17'F for Position 1.1 and oA = 8.5 0 F for Position 2.1 with credible surveillance data; the weld metal cA = 28'F for Position 1.1 and ca= 14'F for Position 2.1 with credible surveillance data. However, ovA need not exceed 0.5*ARTNDT.

WCAP-17539-NP March 2012 Revision 0 6-6 Westinghouse Non-Proprietary Class 3 Table 6.1-3 Calculation of the Sequoyah Unit 1 ART Values at the 1/4T Location for 52 EFPY f 1=RPV Material CF(a)(OF)1/4T Fluence(b)(xl01 9 n/cm 2 , E > 1.0 MeV)Reactor Vessel Beltline Materials IS Forging 05 115.6 1.602 1.1301 40 130.6 0 17.0 34.0 204.6 LS Forging 04 95.0 1.602 1.1301 73 107.4 0 17.0 34.0 214.4 Using credible surveillance data 109.3 1.602 1.1301 73 123.5 0 8.5 17.0 213.5 IS to LS Circ. Weld WO5 161.3 1.596 1.1291 -40 182.1 0 28.0 56.0 198.1 (Heat # 25295)Using credible surveillance data 139.3 1.596 1.1291 -40 157.3 0 14.0 28.0 145.3 Reactor Vessel Extended Beltline Materials US Forging 06 123.9 0.035 0.2408 23 29.8 0 14.9 29.8 82.7 Bottom Head Ring 03 122.3 0.202 0.5722 5 70.0 0 17.0 34.0 109.0 US to IS Circ. Weld W06 207.0 0.035 0.2408 10 49.8 0 24.9 49.8 109.7 (Heat # 25006) 1 LS to Bottom Head Ring Weld W04 161.3 0.202 0.5722 -40 92.3 0 28.0 56.0 108.3 eat # 25295_)------------------------------------------------------------------------

Using credible surveillance data 139.3 0.202 0.5722 -40 79.7 *0 14.0 28.0 67.7 Notes: (a)Data taken from Table 3-5 of this report.(b) The 1/4T fluence and FF were calculated using the Regulatory Guide 1.99, Revision 2 correlations, and the Sequoyah Unit 1 reactor vessel wall thickness of 8.45 inches.(c) Initial RTNDT values were taken from Table 3-1 of this report and are based on measured data for all of the materials.

Note that c7 = 0°F for measured values.(d) Per Appendix A of this report, the surveillance data of the forging and weld were deemed credible.

Per the guidance of Regulatory Guide 1.99, Revision 2, the base metal GA = 17 0 F for Position 1.1 and GA = 8.5°F for Position 2.1 with credible surveillance data; the weld metal GA = 28'F for Position 1.1 and GA =14'F for Position 2.1 with credible surveillance data. However, CA need not exceed 0.5*ARTNDT.

WCAP-17539-NP March 2012 Revision 0 Westinghouse Non-Proprietary Class 3 6-7 Table 6.1-4 Calculation of the Sequoyah Unit 1 ART Values at the 3/4T Location for 52 EFPY I 3/4T3/4T RTNDT(U)(c)

ARTNDT (c) (A (d) Margin ART (xE019 n/cM2, FF(b) (OF) (OF) (OF) (OF) (OF) (OF)E > 1.0 MeV)Reactor Vessel Beltline Materials IS Forging 05 115.6 0.581 0.8481 40 98.0 0 17.0 34.0 172.0 LS Forging 04 95.0 0.581 0.8481 73 80.6 0 17.0 34.0 187.6 Using credible surveillance data 109.3 0.581 0.8481 73 92.7 0 8.5 17.0 182.7 IS to LS Circ. Weld W05 161.3 0.579 0.8471 -40 136.6 0 28.0 56.0 152.6'(H eat # 25295) ........................

Using credible surveillance data 139.3 0.579 0.8471 -40 118.0 0 14.0 28.0 106.0 Reactor Vessel Extended Beltline Materials US Forging 06 123.9 0.013 0.1291 23 16.0 0 8.0 16.0 55.0 Bottom Head Ring 03 122.3 0.073 0.3579 5 43.8 0 17.0 34.0 82.8 US to IS Circ. Weld W06 207.0 0.013 0.1291 10 26.7 0 13.4 26.7 63.4 (Heat # 25006) 1 _LS to Bottom Head Ring Weld W04 161.3 0.073 0.3579 -40 57.7 0 28.0 56.0 73.7-- (Heat # 25295)_Using credible surveillance data 139.3 0.073 0.3579 -40 49.9 0 14.0 28.0 37.9 Notes: (a)Data taken from Table 3-5 of this report.(b) The 3/4T fluence and FF were calculated using the Regulatory Guide 1.99, Revision 2 correlations, and the Sequoyah Unit 1 reactor vessel wall thickness of 8.45 inches.(c) Initial RTNDT values were taken from Table 3-1 of this report and are based on measured data for all of the materials.

Note that (7I = 0 0 F for measured values.(d) Per Appendix A of this report, the surveillance data of the forging and weld were deemed credible.

Per the guidance of Regulatory Guide 1.99, Revision 2, the base metal crA = 17'F for Position 1.1 and GA= 8.5°F for Position 2.1 with credible surveillance data; the weld metal CA = 28°F for Position 1.1 and CA =14'F for Position 2.1 with credible surveillance data. However, oA need not exceed 0.5*ARTNDT.

WCAP-17539-NP March 2012 Revision 0 6-8 Westinghouse Non-Proprietary Class 3 6-8 Westinghouse Non-Proprietary Class 3 Table 6.1-5 Summary of the Sequoyah Unit 1 Limiting ART Values used in the Applicability Evaluation of the Current Reactor Vessel Heatup and Cooldown Curves 1/4T Location 3/4T Location Existing 32 Eitn TLAA EFPY Curves EFPY Curves documented Evaluation at Evaluation at Evaluation at Evaluation at 32 EFPY 52 EFPY 32 EFPY 52 EFPY Revision 2 (Table 6.1-1) (Table 6.1-3) Revision 2 (Table 6.1-2) (Table 6.1-4)Limiting ART (IF) 216 200.6 213.5 186 169.9 182.7 LS Forging 04 LS Forging 04 Using Non- LS Forging 04 Using Credible Using Non- LS Forging 04 Using Credible Limiting Material Credible Surveillance Data Credible Surveillance Data Surveillance Data Surveillance Data WCAP- 17539-NP March 2012 Revision 0 Westinghouse Non-Proprietary Class 3 6-9 6.2 SEQUOYAH UNIT 2 Tables 6.2-1 through 6.2-4 below summarize the 1/4T and 3/4T ART calculations for Sequoyah Unit 2. The limiting 32 EFPY and 52 EFPY ART values for Sequoyah Unit 2 corresponds to the IS to LS Circumferential Weld W05 using non-credible surveillance data (Position 2.1).The applicability of the existing 32 EFPY P-T limit curves, contained in WCAP-15321, Revision 2 (Reference

10) for Sequoyah Unit 2, is evaluated by comparing the updated ART values contained in this section with those used in the Reference 10 calculations.

The existing 32 EFPY P-T limit curves for Sequoyah Unit 2 are based on the limiting beltline material ART values, which are influenced by both fluence and initial material properties of that material.

Using the TLAA fluence projections, the 1/4T and 3/4T ART values were recalculated in Tables 6.2-1 through 6.2-4 as part of this applicability evaluation for Sequoyah Unit 2. Since the capsule fluence values were also updated as part of the TLAA effort, the Position 2.1 chemistry factor values were revised in Section 3 of this report. Furthermore, the credibility evaluation conclusions contained in Appendix A of this report have changed (from non-credible to credible)for the Sequoyah Unit 2 surveillance forging material since the current P-T limit curves were developed.

The comparison of limiting ART values is contained in Table 6.2-5 for Sequoyah Unit 2.Table 6.2-5 below compares the TLAA limiting ART values at 32 EFPY and 52 EFPY to the limiting ART values used in development of the existing 32 EFPY P-T limit curves that are documented in WCAP-15321, Revision 2 (Reference 10). The limiting ART values used to develop the existing P-T limit curves are documented in Table 10 of Reference 10.The TLAA limiting ART values at 32 EFPY are bounded by the limiting ART values used to develop the existing 32 EFPY P-T limit curves. Therefore, the existing Sequoyah Unit 2 P-T limit curves remain valid through 32 EFPY.Furthermore, the TLAA limiting 1/4T ART value at 52 EFPY is bounded by the limiting ART value used to develop the existing 32 EFPY P-T limit curves; however, the TLAA limiting 3/4T ART value at 52 EFPY is not bounded by the limiting ART value used to develop the existing 32 EFPY P-T limit curves. Since there is a slight difference between the limiting 3/4T ART values, the extended applicability of the existing P-T limit curves is determined considering the updated TLAA fluence evaluation as well as the updated credibility analysis of the Sequoyah Unit 2 surveillance materials.

Due to the revised credibility evaluation, which is performed in Appendix A of this report, and updated fluence data of the Sequoyah Unit 2 surveillance capsules, the IS to LS Circumferential Weld W05 using non-credible surveillance data (Position 2.1) has become the limiting material based on the calculations presented in Tables 6.2-1 through 6.2-4. Note that this limiting material has changed since the existing 32 EFPY P-T limit curves were developed.

WCAP- 17539-NP March 2012 Revision 0 6-10 Westinghouse Non-Proprietary Class 3 Surveillance data is available for the IS to LS Circumferential Weld W05. Therefore, the Position 2.1 chemistry factor, initial RTNDT, and margin terms from Table 6.2-4 were used to determine the 3/4T fluence value when the 3/4T ART equals 115'F (Table 6.2-5) for this material.

This fluence value is approximately 2.22 x 1019 n/cm2 (E > 1.0 MeV), which was used to calculate an associated EFPY based on the updated fluence values (Table 2-3) for this material.The EFPY associated with a 3/4T ART value of 115'F for the IS to LS Circumferential Weld W05 (Position 2.1) is 44.7 EFPY. Therefore, the existing Sequoyah Unit 2 P-T limit curves may be deemed applicable through 44.7 EFPY.P-T Limits Applicability Conclusion For Sequoyah Unit 2, it is concluded that the existing 32 EFPY P-T limit curves do not require a reduction of the applicability date. Since the P-T limit curves remain valid through the original EFPY period, the current LTOP setpoints also remain applicable through 32 EFPY.Furthermore, based on the TLAA evaluation, Tennessee Valley Authority may instead choose to extend the applicability of the existing Sequoyah Unit 2 P-T limit curves. The new applicability date with consideration of the TLAA credibility and fluence evaluations is 44.7 EFPY. Note that an evaluation would have to be performed to increase the LTOP setpoints applicability period.WCAP-17539-NP March 2012 Revision 0 Westinghouse Non-Proprietary Class 3 6-11 Westinghouse Non-Proprietary Class 3 6-11 Table 6.2-1 Calculation of the Sequoyah Unit 2 ART Values at the 1/4T Location for 32 EFPY CF(a) 1/4T Fluence(b) 1/4T RTNDT(U) (c) ARTNDT 61t(C) FA(d) Margin ART RPV Material (OF) (0 /c FF(b) (OF) (OF) (OF) (OF) (OF) (OF)E > 1.0 MeV)IS Forging 05 95.0 1.000 0.9999 10 95.0 0 17.0 34.0 139.0 Using credible surveillance data 91.1 1.000 0.9999 10 91.1 0 8.5 17.0 118.1 LS Forging 04 104.0 1.000 0.9999 -22 104.0 0 17.0 34.0 116.0 IS to LS Circ. Weld W05 (Heat # 4278) 63.0 0.994 0.9983 -4 62.9 0 28.0 56.0 114.9 Using non-credible surveillance data 78.9 0.994 0.9983 -4 78.8 0 28.0 56.0 130.8 Notes: (a) Data taken from Table 3-6 of this report.(b) The 1/4T fluence and FF were calculated using the Regulatory Guide 1.99, Revision 2 correlations, and the Sequoyah Unit 2 reactor vessel wall thickness of 8.45 inches.(c) Initial RTNDT values were taken from Table 3-2 of this report and are based on measured data for all of the materials.

Note that oi = 0°F for measured values.(d) Per Appendix A of this report, the surveillance data of the forging was deemed credible and the surveillance data of the weld was deemed non-credible.

Per the guidance of Regulatory Guide 1.99, Revision 2, the base metal 0 A = 17 0 F for Position 1.1 and A = 8.5 0 F for Position 2.1 with credible surveillance data; the weld metal cA = 28°F for Position 1.1 and for Position 2.1 with non-credible surveillance data. However, CA need not exceed 0.5*ARTNDT.

WCAP-1 7539-NP March 2012 WCAP-17539-NP March 2012 Revision 0 6-12 Westinghouse Non-Proprietary Class 3 6-12 Westinghouse Non-Proprietary Class 3 Table 6.2-2 Calculation of the Sequoyah Unit 2 ART Values at the 3/4T Location for 32 EFPY CF(a) 3/T Fluence(b) 3/4T RTNDT(U) (c) ARTNDT Fgl(c) GA(d) Margin ART RPV Material (OF) (xl0 9 n/cm2, FF(b) (OF) (OF) (OF) (OF) (OF) (OF)E > 1.0 MeV)IS Forging 05 95.0 0.363 0.7199 10 68.4 0 17.0 34.0 112.4 Using credible surveillance data 91.1 0.363 0.7199 10 65.6 0 8.5 17.0 92.6 LS Forging 04 104.0 0.363 0.7199 -22 74.9 0 17.0 34.0 86.9 IS to LS Circ. Weld W05 (Heat # 4278) 63.0 0.361 0.7183 -4 45.3 0 22.6 45.3 86.5 Using non-credible surveillance data 78.9 0.361 0.7183 -4 56.7 0 28.0 56.0 108.7 Notes: (a) Data taken from Table 3-6 of this report.(b) The 3/4T fluence and FF were calculated using the Regulatory Guide 1.99, Revision 2 correlations, and the Sequoyah Unit 2 reactor vessel wall thickness of 8.45 inches.(c) Initial RTNDT values were taken from Table 3-2 of this report and are based on measured data for all of the materials.

Note that cy, = 0 0 F for measured values.(d) Per Appendix A of this report, the surveillance data of the forging was deemed credible and the surveillance data of the weld was deemed non-credible.

Per the guidance of Regulatory Guide 1.99, Revision 2, the base metal 0 A = 17'F for Position 1.1 and GA = 8.5°F for Position 2.1 with credible surveillance data; the weld metal 0 A = 28'F for Position 1.1 and for Position 2.1 with non-credible surveillance data. However, cYA need not exceed 0.5*ARTNDT.

WCAP- 17539-NP March 2012 Revision 0 Westinghouse Non-Proprietary Class 3 6-13 Westinghouse Non-Proprietary Class 3 6-13 Table 6.2-3 Calculation of the Sequoyah Unit 2 ART Values at the 1/4T Location for 52 EFPY CF(a) 1/4T Fluence(b) 1/4T RTNDT(U)(c)

ARTNDT 01(c) (A(d) Margin ART RPV Material (OF) (X__019 n/cm2' FF(b) (OF) (OF) (OF) (OF) (OF) (OF)E > 1.0 MeV)Reactor Vessel Beltline Materials IS Forging 05 95.0 1.548 1.1208 10 106.5 0 17.0 34.0 150.5 Using credible surveillance data 91.1 1.548 1.1208 10 102.1 0 8.5 17.0 129.1 LS Forging 04 104.0 1.548 1.1208 -22 116.6 0 17.0 34.0 128.6 IS to LS Circ. Weld W05 (Heat # 4278) 63.0 1.536 1.1187 -4 70.5 0 28.0 56.0 122.5 Using non-credible surveillance data 78.9 1.536 1.1187 -4 88.3 0 28.0 56.0 140.3 Reactor Vessel Extended Beltline Materials US Forging 06 123.4 0.033 0.2331 5 28.8 0 14.4 28.8 62.5 Bottom Head Ring 03 122.3 0.190 0.5576 5 68.2 0 17.0 34.0 107.2 US to IS Circ. Weld W06 (Heat # 721858) 108.0 0.033 0.2331 10 25.2 0 12.6 25.2 60.4 LS to Bottom Head Ring Weld W04 108.0 0.190 0.5576 10 60.2 0 28.0 56.0 126.2 (Heat # 721858)Notes: (a) Data taken from Table 3-6 of this report.(b) The 1/4T fluence and FF were calculated using the Regulatory Guide 1.99, Revision 2 correlations, and the Sequoyah Unit 2 reactor vessel wall thickness of 8.45 inches.(c) Initial RTNDT values were taken from Table 3-2 of this report and are based on measured data for all of the materials.

Note that a 1 = 0°F for measured values.(d) Per Appendix A of this report, the surveillance data of the forging was deemed credible and the surveillance data of the weld was deemed non-credible.

Per the guidance of Regulatory Guide 1.99, Revision 2, the base metal A = 17'F for Position 1.1 and GA = 8.5°F for Position 2.1 with credible surveillance data;the weld metal GA = 28°F for Position 1.1 and for Position 2.1 with non-credible surveillance data. However, 0 A need not exceed 0.5*ARTNDT.

WCAP- 17539-NP March 2012 Revision 0 6-14 Westinghouse Non-Proprietary Class 3 Table 6.2-4 Calculation of the Sequoyah Unit 2 ART Values at the 3/4T Location for 52 EFPY Reactor Vessel Material CF(a)(01F)3/4T Fluence ((xl0 1 9 n/cm 2 , E > 1.0 MeV)Reactor Vessel Beltline Materials IS Forging 05 95.0 0.562 0.8386 10 79.7 0 17.0 34.0 1,23.7 Using credible surveillance data 91.1 0.562 0.8386 10 76.4 0 8.5 17.0 103.4 LS Forging 04 104.0 0.562 0.8386 -22 87.2 0 17.0 34.0 99.2 IS to LS Circ. Weld W05 (Heat # 4278) 63.0 0.557 0.8364 -4 52.7 0 26.3 52.7 101.4 Using non-credible surveillance data 78.9 0.557 0.8364 -4 66.0 0 28.0 56.0 118.0 Reactor Vessel Extended Beltline Materials US Forging 06 123.4 0.012 0.1244 5 15.3 0 7.7 15.3 35.7 Bottom Head Ring 03 122.3 0.069 0.3469 5 42.4 0 17.0 34.0 81.4 US to IS Circ. Weld W06 (Heat # 721858) 108.0 0.012 0.1244 10 13.4 0 6.7 13.4 36.9 LS to Bottom Head Ring Weld W04 108.0 0.069 0.3469 (Heat _ 721858) 1 10 37.5 0 18.7 37.5 84.9 Notes: (a) Data taken from Table 3-6 of this report.(b) The 3/4T fluence and FF were calculated using the Regulatory Guide 1.99, Revision 2 correlations, and the Sequoyah Unit 2 reactor vessel wall thickness of 8.45 inches.(c) Initial RTNDT values were taken from Table 3-2 of this report and are based on measured data for all of the materials.

Note that cy, 0°F for measured values.(d) Per Appendix A of this report, the surveillance data of the forging was deemed credible and the surveillance data of the weld was deemed non-credible.

Per the guidance of Regulatory Guide 1.99, Revision 2, the base metal Ga = 17'F for Position 1.1 and aA = 8.5°F for Position 2.1 with credible surveillance data;the weld metal CA = 28°F for Position 1.1 and for Position 2.1 with non-credible surveillance data. However, 0 A need not exceed 0.5*ARTNDT.

WCAP- 17539-NP March 2012 Revision 0 Westinghouse Non-Proprietary Class 3 6-15 Westinghouse Non-Proprietary Class 3 6-15 Table 6.2-5 Summary of the Sequoyah Unit 2 Limiting ART Values used in the Applicability Evaluation of the Current Reactor Vessel Heatup and Cooldown Curves 1/4T Location 3/4T Location Existing 32 Existing 32 EFPY Curves EFPY Curves documented Evaluation at Evaluation at Evaluation at Evaluation at 32 EFPY 52 EFPY 32 EFPY 52 EFPY WCAP-15321, (Table 6.2-1) (Table 6.2-3) WCAP-15321, (Table 6.2-2) (Table 6.2-4)Revision 2 _ _ _ _ _ __ Revision2 2_ _ _Limiting ART (IF) 142 130.8 140.3 115 108.7 118.0 IS Forging 05 IS to LS Circ. Weld W05 Using IS Forging 05 IS to LS Circ. Weld W05 Using Simtig atehutence D Non-Credible Surveillance Data S utvellnc Non-Credible Surveillance Data Surveillance Data Surveillance Data WCAP- 17539-NP March 2012 Revision 0 Westinghouse Non-Proprietary Class 3 7-1 7 SURVEILLANCE CAPSULE WITHDRAWAL SCHEDULES Contained in Tables 7-1 and 7-2 are the Sequoyah Units 1 and 2 recommended surveillance capsule withdrawal schedules, respectively.

These schedules meet the recommendations of ASTM E185-82 (Reference

21) as required by 10 CFR 50, Appendix H (Reference 22). With the withdrawal of Capsule Y, Sequoyah Units I and 2 fulfilled the surveillance capsule withdrawal recommendations contained in ASTM E185-82 for their 40-year EOL (32 EFPY).Since Sequoyah Units 1 and 2 are applying for a 20-year license extension, it is recommended that several remaining capsules be relocated to higher lead factor locations for each Unit. One of these relocated capsules in each Unit should be subsequently withdrawn from the reactor vessel and tested at the time when the accumulated neutron fluence of the capsule corresponds to not less than once or greater than twice the peak 60-year vessel fluence.Table 7-1 Sequoyahb Unit 1 Surveillance Capsule Withdrawal Summary Fluence(a)

Capsule Capsule Lead Withdrawal (xl01 9 n/cm 2 , Location Factor(a)

EFPY(b)E > 1.0 MeV)T 400 3.15 1.07 0.241 U 1400 3.23 2.85 0.693 X 2200 3.22 5.26 1.16 Y 3200 3.18 10.02 1.97 5 (ý40 0.90 ~(c)_ __ __ __ __ __ __V 1760 0.90 (c) (c)W 1840 0.90 (c) (c)Z 3560 0.90 (c) (c)Notes: (a)(b)(c)Updated as part of the TLAA fluence evaluation.

EFPY from plant startup.Capsules,;

S V, W and Z are currently in the Sequoyah Unit I reactor vessel. Either Capsule S, V, W, or Z should be withdrawn so that the capsule fluence corresponds to at least one times the 60-year EOL vessel fluence (2.66 x 10'9 n/cm-), but less than two times the 60-year EOL vessel fluence:(5.32 xj j010' n/cm 2).However, none of these remaining capsules are predicted to experience a neutron fluence of 2?.66 x" 1019ffi 2 to EO LEini'their current locations; therefore, it is recommended to relocate several of these remaining capsules to higher lead factor locations in order to achieve higher capsule fluence data.Assuming a capsule was relocated at the end of cycle 18, 19, or 20, the EFPY that corresponds to the time when the capsule experiences the peak EOLE vessel fluence value (2.66 x 1019 n/cm 2) is approximately 32.5, 33.4, or 34.4 EFPY, respectively.

See Appendix B for further details on capsule relocation recommendations.

WCAP-17539-NP March 2012 Revision 0 7-2 Westinghouse Non-Proprietary Class 3 Table 7-2 Sequoyah Unlt2 Surveillance Capsule Withdrawal Summary Fluence(a)

Capsule Lead Withdrawal Fx10n9 (a)Capsule Location Factor(a)

EFPY(b) (X19 n/cm 2 , E > 1.0 MeV)T 400 3.11 1.07 0.244 U 1400 3.17 2.91 0.654 X 2200 3.18 5.36 1.16 Y 3200 3.15 10.55 2.02 S4 ~0.94ý _________(V 1760 0.94 (c) (c)W 1840 0.94 (c) (c)Z 3560 0.94 (c) (c)Notes: (a)(b)(c)Updated as part of the TLAA fluence evaluation.

EFPY from plant startup.Capsules S, V, W and Z are currently in the Sequoyah Unit 2 reactor vessel. Either CapsuleS, V, W, or Z should be withdrawn so that the capsule fluence corresponds to at least one times the 60-year EOL vessel fluence, (.57 x'10 9)but less than two times the 60-year EOL vessel fluence,(5.14 x 1019 nlmM))However, none of these remaining capsules are predicted to experience a neutron fluence of 2.57 x 1019 it/cm 2 prior to EOLE in their current locations; therefore, it is recommended to relocate several of these remaining capsules to higher lead factor locations in order to achieve higher capsule fluence data.Assuming a capsule was relocated at the end of cycle 18, 19, or 20, the EFPY that corresponds to the time when the capsule experiences the peak EOLE vessel fluence value (2.57 x 1019 n/cm 2) is approximately 32.6, 33.7, or 34.7 EFPY, respectively.

See Appendix B for further details on capsule relocation recommendations.

WCAP-17539-NP March 2012 Revision 0 Westinghouse Non-Proprietary Class 3 8-1 8 REFERENCES

1. Code of Federal Regulations, 10 CFR Part 54.3, "Definitions," U.S. Nuclear Regulatory Commission, Washington, D.C., Federal Register, Volume 72, dated August 28, 2007.2. Regulatory Guide 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence," U:S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, March 2001.3. WCAP-14040-A, Revision 4, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves," J. D.Andrachek et al., May 2004.4. WCAP-16083-NP-A, Revision 0, "Benchmark Testing of the FERRET Code for Least Squares Evaluation of Light Water Reactor Dosimetry," S. L. Anderson, May 2006.5. NUREG-1801, Revision 2, "Generic Aging Lessons Learned (GALL) Report," U.S.Nuclear Regulatory Commission, December 2010.6. Oak Ridge National Laboratory document ORNL/TM-2006/530, "A Physically Based Correlation of Irradiation-Induced Transition Temperature Shifts for RPV Steels," E. D.Eason et al., November 2007.7. WCAP-15293, Revision 2, "Sequoyah Unit 1 Heatup and Cooldown Limit Curves for Normal Operation and PTLR Support Documentation," J. H. Ledger, July 2003.8. WCAP-8233, Revision 0, "Tennessee Valley Authority Sequoyah Unit No. 1 Reactor Vessel Radiation Surveillance Program," S. E. Yanichko et al., December 1973.9. Code of Federal Regulations, 10 CFR Part 50.61, "Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events," U.S. Nuclear Regulatory Commission, Washington, D.C., Federal Register, Volume 60, No. 243, dated December 19, 1995, effective January 18, 1996.10. WCAP-15321, Revision 2, "Sequoyah Unit 2 Heatup and Cooldown Limit Curves for Normal Operation and PTLR Support Documentation," J. H. Ledger, July 2003.11. WCAP-8513, Revision 0, "Tennessee Valley Authority Sequoyah Unit No. 2 Reactor Vessel Radiation Surveillance Program," S. E. Yanichko et al., November 1975.12. "Fracture Toughness Requirements," Branch Technical Position 5-3, Revision 2, Contained in Chapter 5 of Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition, NUREG-0800, March 2007.13. WCAP-10340, Revision 1, "Analysis of Capsule T From the Tennessee Valley Authority Sequoyah Unit 1 Reactor Vessel Radiation Surveillance Program," S. E. Yanichko et al., February 1984.WCAP-17539-NP March 2012 Revision 0 8-2 Westinghouse Non-Proprietary Class 3 14. Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials," U.S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, May 1988.15. WCAP- 15224, Revision 0, "Analysis of Capsule Y from the Tennessee Valley Authority Sequoyah Unit 1 Reactor Vessel Radiation Surveillance Program," T. J. Laubham, June 1999.16. WCAP-15320, Revision 0, "Analysis of Capsule Y from the Tennessee Valley Authority Sequoyah Unit 2 Reactor Vessel Radiation Surveillance Program," T. J. Laubham, December 1999.17. Code of Federal Regulations, 10 CFR 50.61a, "Alternate Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events," U.S. Nuclear Regulatory Commission, Washington, D.C., Federal Register, Volume 75, No. 1, dated January 4, 2010, with corrections dated February 3, 2010 (No. 22), March 8, 2010 (No.44), and November 26, 2010 (No. 227).18. Code of Federal Regulations, 10 CFR Part 50, Appendix G, "Fracture Toughness Requirements," U.S. Nuclear Regulatory Commission, Washington, D.C., Federal Register, Volume 60, No. 243, dated December 19, 1995.19. Appendix G to the 1998 through the 2000 Addenda Edition of the ASME Boiler and Pressure Vessel (B&PV) Code,Section XI, Division 1, "Fracture Toughness Criteria for Protection Against Failure.20. NUREG- 1800, Revision 2, "Standard Review Plan for Review of License Renewal Applications for Nuclear Power Plants," U.S. Nuclear Regulatory Commission, December 2010.21. ASTM E185-82, "Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels," American Society for Testing and Materials, 1982.22. Code of Federal Regulations, 10 CFR 50, Appendix H, "Reactor Vessel Material Surveillance Program Requirements," U.S. Nuclear Regulatory Commission, Washington, D.C., Federal Register, Volume 60, No. 243, dated December 19, 1995.WCAP-17539-NP March 2012 Revision 0 Westinghouse Non-Proprietary Class 3 A-1 APPENDIX A CREDIBILITY EVALUATION OF THE SEQUOYAH UNITS 1 AND 2 SURVEILLANCE PROGRAMS A.1 SEQUOYAH UNIT 1 INTRODUCTION Regulatory Guide 1.99, Revision 2 (Reference A. 1-1) describes general procedures acceptable to the NRC staff for calculating the effects of neutron radiation embrittlement of the low-alloy steels currently used for light-water-cooled reactor vessels. Position C.2 of Regulatory Guide 1.99, Revision 2, describes the method for calculating the adjusted reference temperature and Charpy upper-shelf energy of reactor vessel beltline materials using surveillance capsule data.The methods of Position C.2 can only be applied when two or more credible surveillance data sets become available from the reactor in question.To date there have been four surveillance capsules removed and tested from the Sequoyah Unit 1 reactor vessel. To use these surveillance data sets, they must be shown to be credible.

In accordance with Regulatory Guide 1.99, Revision 2, the credibility of the surveillance data will be judged based on five criteria.The purpose of this evaluation is to apply the credibility requirements of Regulatory Guide 1.99, Revision 2, to the Sequoyah Unit 1 reactor vessel surveillance data and determine if that surveillance data is credible.EVALUATION Criterion 1: Materials in the capsules should be those judged most likely to be controlling with regard to radiation embrittlement.

The beltline region of the reactor vessel is defined in Appendix G to 10 CFR Part 50, "Fracture Toughness Requirements" (Reference A. 1-2), as follows: "the region of the reactor vessel (shell material including welds, heat affected zones, and plates or forgings) that directly surrounds the effective height of the active core and adjacent regions of the reactor vessel that are predicted to experience sufficient neutron radiation damage to be considered in the selection of the most limiting material with regard to radiation damage." The Sequoyah Unit 1 reactor vessel consists of the following beltline region materials:

1. Intermediate Shell (IS) Forging 05 2. Lower Shell (LS) Forging 04 3. Intermediate Shell Forging to Lower Shell Forging Circumferential Weld Seam W05 (fabricated with SMIT 40 weld wire type, heat # 25295 and SMIT 89 flux type, lot # 2275)WCAP-17539-NP March 2012 Revision 0 A-2 Westinghouse Non-Proprietary Class 3 The Sequoyah Unit 1 surveillance program utilizes tangential and axial test specimens from Lower Shell Forging 04. The surveillance weld metal was fabricated with SMIT 40 weld wire type, heat # 25295 and SMIT 89 flux type, lot # 1103.Per WCAP-8233 (Reference A.1-3), the Sequoyah Unit 1 surveillance program was based on ASTM E185-73 (Reference.

A.1-4). Per Section 4.1 of ASTM E185-73, "The base metal and weld metal to be included in the program should represent the material that may limit the operation of the reactor during its lifetime.

The test material should be selected on the basis of initial transition temperature, upper shelf energy level, and estimated increase in transition temperature considering chemical composition (copper (Cu) and phosphorus (P)) and neutron fluence.At the time when the surveillance program was developed, it was believed that copper and phosphorus were the elements most important to embrittlement of reactor vessel steels. Lower Shell Forging 04 had the highest initial RTNDT and lowest initial upper-shelf energy out of the two beltline forgings in the Sequoyah Unit 1 reactor vessel. In addition, Lower Shell Forging 04 had approximately the same copper and phosphorus content of the other beltline forging. Thus, it was selected as the surveillance base metal.The weld material in the Sequoyah Unit 1 surveillance program was made of the same material as the reactor vessel beltline circumferential weld. In accordance with the definition of the reactor vessel beltline at that time, this was the only weld in the beltline region.Based on the above discussion, Criterion 1 is met for the Sequovah Unit 1 surveillance program.Criterion 2: Scatter in the plots of Charpy energy versus temperature for the irradiated and unirradiated conditions should be small enough to permit the determination of the 30 ft-lb temperature and upper-shelf energy unambiguously.

Plots of Charpy energy versus temperature for the unirradiated and irradiated conditions for all of the Sequoyah Unit 1 surveillance materials are presented in Section 5 and Appendix C of the latest surveillance capsule report, WCAP-15224 (Reference A. 1-5).Based on engineering judgment, the scatter in the data presented in these plots is small enough to permit the determination of the 30 ft-lb temperature and the upper-shelf energy of the Sequoyah Unit 1 surveillance materials unambiguously.

Hence, Criterion 2 is met for the Sequovah Unit 1 surveillance program.WCAP-17539-NP March 2012 Revision 0 Westinghouse Non-Proprietary Class 3 A-3 Westinghouse Non-Proprietary Class 3 A-3 Criterion 3: When there are two or more sets of surveillance data from one reactor, the scatter of ARTNDT values about a best-fit line drawn as described in Regulatory Position 2.1 normally should be less than 28°F for welds and 17'F for base metal. Even if the fluence range is large (two or more orders of magnitude), the scatter should not exceed twice those values. Even if the data fail this criterion for use in shift calculations, they may be credible for determining decrease in upper-shelf energy if the upper shelf can be clearly determined, following the definition given in ASTM E185-82 (Reference A. 1-6).The functional form of the least squares method as described in Regulatory Position 2.1 will be utilized to determine a best-fit line for this data and to determine if the scatter of these ARTNDT values about this line is less than 287F for welds and less than 177F for the forging.The Sequoyah Unit 1 Lower Shell Forging 04 and surveillance weld will be evaluated for credibility.

The weld is made from weld wire heat # 25295. This weld metal is not in any other surveillance program.Following is the calculation of the best-fit line as described in Regulatory Position 2.1 of Regulatory Guide 1.99, Revision 2. In addition, the recommended NRC methods for determining credibility will be followed.

The NRC methods were presented to the industry at a meeting held by the NRC on February 12 and 13, 1998 (Reference A.1-7). At this meeting the NRC presented five cases. Of the five cases, Case 1 ("Surveillance data available from plant but no other source") most closely represents the situation listed above for the Sequoyah Unit 1 surveillance forging and weld materials.

WCAP-17539-NP March 2012 Revision 0 A-4 Westinghouse Non-Proprietary Class 3 Following the NRC Case 1 guidelines, the Sequoyah Unit 1 surveillance forging and weld metal (Heat # 25295) will be evaluated using the Sequoyah Unit 1 data. This evaluation is contained in Table A.1-1.Table A.1-1 Calculation of Interim Chemistry Factors for the Credibility Evaluation Using Sequoyah Unit 1 Surveillance Capsule Data Only Capsule Fluence(a)

FF b) ARTNDT(c)

FF*ARTNDT FF 2 RPV Material Capsule (x1019n/cm2, (OF) (OF)E > 1.0 MeV)T 0.241 0.615 67.52 41.52 0.378 LS Forging 04 U 0.693 0.897 109.7 98.42 0.805 (Tangential)

X 1.16 1.041 145.12 151.13 1.085 Y 1.97 1.185 129.87 153.92 1.405 T 0.241 0.615 50.59 31.11 0.378 LS Forging 04 U 0.693 0.897 67.59 60.64 0.805 (Axial) X 1.16 1.041 103.34 107.62 1.085 Y 1.97 1.185 133.35 158.04 1.405 SUM: 802.39 7.344 CFLs Forging 04 = E(FF

ARTNDT) -E(FF 2) = (802.39) -(7.344) 109.3°F T 0.241 0.615 127.79 78.57 0.378 Surveillance Weld U 0.693 0.897 144.92 130.02 0.805 Metal (Heat # 25295) X 1.16 1.041 159.02 165.61 1.085 Y 1.97 1.185 163.8 194.13 1.405 SUM: 568.33 3.672 CFHeat # 2 5 29 5= X(FF

ARTNDT) E(FF 2) = (568.33) + (3.672) 154.8°F Notes: (a) f= capsule fluence taken from Table 2-6 of this report.(b) FF = fluence factor = f0.28-0.10*log f)(c) ARTNDT values are the measured 30 ft-lb shift values taken from Table 5-10 of WCAP-15224 (Reference A. 1-5). These measured ARTNDT values for the surveillance weld metal do not include the adjustment ratio procedure of Regulatory Guide 1.99, Revision 2, Position 2.1 since this calculation is based on the actual surveillance weld metal measured shift values. In addition, only Sequoyah Unit 1 data is being considered; therefore, no temperature adjustment is required.WCAP-17539-NP March 2012 Revision 0 Westinghouse Non-Proprietary Class 3 A-5 The scatter of ARTNDT values about the functional form of a best-fit line drawn as described in Regulatory Position 2.1 is presented in Table A. 1-2.Table A.1-2 Sequoyah Unit 1 Surveillance Capsule Data Scatter about the Best-Fit Line Capsule Measured Predicted Scatter <17 0 F CF Fluence RPV Material Capsule (OF) (x1019 n/cm2, FF ARTNDT ARTNDT ARTNDT (Base Metal)E > 1.0 MeV) (OF) (OF) (OF) <281F (Weld)T 109.3 0.241 0.615 67.52 67.2 0.3 Yes LS Forging 04 U 109.3 0.693 0.897 109.7 98.0 11.7 Yes (Tangential)

X 109.3 1.16 1.041 145.12 113.8 31.3 No Y 109.3 1.97 1.185 129.87 129.5 0.4 Yes T 109.3 0.241 0.615 50.59 67.2 16.6 Yes LS Forging 04 U 109.3 0.693 0.897 67.59 98.0 30.4 No (Axial) X 109.3 1.16 1.041 103.34 113.8 10.4 Yes Y 109.3 1.97 1.185 133.35 129.5 3.9 Yes T 154.8 0.241 0.615 127.79 95.2 32.6 No Surveillance U 154.8 0.693 0.897 144.92 138.9 6.1 Yes Weld Metal (Heat # 25295) X 154.8 1.16 1.041 159.02 161.2 2.2 Yes Y 154.8 1.97 1.185 163.8 183.4 19.6 Yes The scatter of ARTNDT values about the best-fit line, drawn as described in Regulatory Guide 1.99, Revision 2, Position 2.1, should be less than 17'F for base metal. Table A. 1-2 indicates that six of the eight surveillance data points fall within the +/- la of 17'F scatter band for surveillance base metals; therefore, the forging data is deemed "credible" per the third criterion.

The scatter of ARTNDT values about the best-fit line, drawn as described in Regulatory Guide 1.99, Revision 2, Position 2.1, should be less than 28°F for weld metal. Table A. 1-2 indicates that three of the four surveillance data points fall within the +/- lC of 28°F scatter band for surveillance weld materials; therefore, the weld material is deemed "credible" per the third criterion.

WCAP-17539-NP March 2012 Revision 0 A-6 Westinghouse Non-Proprietary Class 3 A- esigoueNn-rpieayCls-Criterion 4: The irradiation temperature of the Charpy specimens in the capsule should match the vessel wall temperature at the cladding/base metal interface within +/- 25°F.The Sequoyah Unit 1 capsule specimens are located in the reactor between the thermal shield and the vessel wall and are positioned opposite the center of the core. The test capsules are in baskets attached to the thermal shield. The location of the specimens with respect to the reactor vessel beltline provides assurance that the reactor vessel wall and the specimens experience equivalent operating conditions such that the temperatures will not differ by more than 25TF.Hence, Criterion 4 is met for the Sequoyah Unit 1 surveillance program.Criterion 5: The surveillance data for the correlation monitor material in the capsule should fall within the scatter band of the database for that material.The Sequoyah Unit 1 surveillance program does not contain correlation monitor material;therefore, this criterion is not applicable to the Sequoyah Unit 1 surveillance program.CONCLUSION:

Based on the preceding responses to the five criteria of Regulatory Guide 1.99, Revision 2, Section B, and the application of engineering judgment, the Sequoyah Unit 1 surveillance forging and weld materials are deemed credible.Appendix A.1 References A.1-1 U.S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials," May 1988.A. 1-2 10 CFR 50, Appendix G, "Fracture Toughness Requirements," Federal Register, Volume 60, No. 243, December 19, 1995.A.1-3 WCAP-8233, Revision 0, "Tennessee Valley Authority Sequoyah Unit No. 1 Reactor Vessel Radiation Surveillance Program," S. E. Yanichko et al., December 1973.A. 1-4 ASTM E185-73, "Standard Recommended Practice for Surveillance Tests for Nuclear Reactor Vessels," American Society for Testing and Materials, 1973.A.1-5 WCAP-15224, Revision 0, "Analysis of Capsule Y from the Tennessee Valley Authority Sequoyah Unit 1 Reactor Vessel Radiation Surveillance Program," T. J.Laubham et al., June 1999.WCAP- 17539-NP March 2012 Revision 0 Westinghouse Non-Proprietary Class 3 A-7 A. 1-6 ASTM E 185-82, "Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels," American Society for Testing and Materials, 1982.A. 1-7 K. Wichman, M. Mitchell, and A. Hiser, USNRC, Generic Letter 92-01 and RPV Integrity Workshop Handouts, NRC/Industry Workshop on RPV Integrity Issues, February 12, 1998.WCAP-17539-NP March 2012 Revision 0 A-8 Westinghouse Non-Proprietary Class 3 A.2 SEQUOYAH UNIT 2 INTRODUCTION Regulatory Guide 1.99, Revision 2 (Reference A.2-1) describes general procedures acceptable to the NRC staff for calculating the effects of neutron radiation embrittlement of the low-alloy steels currently used for light-water-cooled reactor vessels. Position C.2 of Regulatory Guide 1.99, Revision 2, describes the method for calculating the adjusted reference temperature and Charpy upper-shelf energy of reactor vessel beltline materials using surveillance capsule data.The methods of Position C.2 can only be applied when two or more credible surveillance data sets become available from the reactor in question.To date there have been four surveillance capsules removed and tested from the Sequoyah Unit 2 reactor vessel. To use these surveillance data sets, they must be shown to be credible.

In accordance with Regulatory Guide 1.99, Revision 2, the credibility of the surveillance data will be judged based on five criteria.The purpose of this evaluation is to apply the credibility requirements of Regulatory Guide 1.99, Revision 2, to the Sequoyah Unit 2 reactor vessel surveillance data and determine if that surveillance data is credible.EVALUATION Criterion 1: Materials in the capsules should be those judged most likely to be controlling with regard to radiation embrittlement.

The beltline region of the reactor vessel is defined in Appendix G to 10 CFR Part 50, "Fracture Toughness Requirements" (Reference A.2-2), as follows: "the region of the reactor vessel (shell material including welds, heat affected zones, and plates or forgings) that directly surrounds the effective height of the active core and adjacent regions of the reactor vessel that are predicted to experience sufficient neutron radiation damage to be considered in the selection of the most limiting material with regard to radiation damage. " The Sequoyah Unit 2 reactor vessel consists of the following beltline region materials:

1. Intermediate Shell (IS) Forging 05.2. Lower Shell (LS) Forging 04 3. Intermediate Shell Forging to Lower Shell Forging Circumferential Weld Seam W05 (fabricated with Arcos weld wire type, heat # 4278 and SMIT 89 flux type, lot # 1211)WCAP-17539-NP March 2012 Revision 0 Westinghouse Non-Proprietary Class 3 A-9 The Sequoyah Unit 2 surveillance program utilizes tangential and axial test specimens from Intermediate Shell Forging 05. The surveillance weld metal was fabricated with Arcos weld wire type, heat # 4278 and SMIT 89 flux type, lot # 1211.Per WCAP-8513 (Reference A.2-3), the Sequoyah Unit 2 surveillance program was based on ASTM E185-73 (Reference.

A.2-4). Per Section 4.1 of ASTM E185-73, "The base metal and weld metal to be included in the program should represent the material that may limit the operation of the reactor during its lifetime.

The test material should be selected on the basis of initial transition temperature, upper shelf energy level, and estimated increase in transition temperature considering chemical, composition (copper (Cu) and phosphorus (P)) and neutron fluence." At the time when the surveillance program was developed, it was believed that copper and phosphorus were the elements most important to embrittlement of reactor vessel steels.Intermediate Shell Forging 05 had the highest initial RTNDT and lowest initial upper-shelf energy out of the two beltline forgings in the Sequoyah Unit 2 reactor vessel. In addition, Intermediate Shell Forging 05 had approximately the same copper and phosphorus content of the other beltline forging. Thus, it was selected as the surveillance base metal.The weld material in the Sequoyah Unit 2 surveillance program was made of the same material as the reactor vessel beltline circumferential weld. In accordance with the definition of the reactor vessel beltline at that time, this was the only weld in the beltline region.Based on the above discussion, Criterion 1 is met for the Sequoyah Unit 2 surveillance program.Criterion 2: Scatter in the plots of Charpy energy versus temperature for the irradiated and unirradiated conditions should be small enough to permit the determination of the 30 ft-lb temperature and upper-shelf energy unambiguously.

Plots of Charpy energy versus temperature for the unirradiated and irradiated conditions for all of the Sequoyah Unit 2 surveillance materials are presented in Section 5 and Appendix C of the latest surveillance capsule report, WCAP- 15320 (Reference A.2-5).Based on engineering judgment, the scatter in the data presented in these plots is small enough to permit the determination of the 30 ft-lb temperature and the upper-shelf energy of the Sequoyah Unit 2 surveillance materials unambiguously.

Hence, Criterion 2 is met for the Sequoyah Unit 2 surveillance program.WCAP- 17539-NP March 2012 Revision 0 A-10 Westinghouse Non-Proprietary Class 3 Criterion 3: When there are two or more sets of surveillance data from one reactor, the scatter of ARTNDT values about a best-fit line drawn as described in Regulatory Position 2.1 normally should be less than 287F for welds and 17'F for base metal. Even if the fluence range is large (two or more orders of magnitude), the scatter should not exceed twice those values. Even if the data fail this criterion for use in shift calculations, they may be credible for determining decrease in upper-shelf energy if the upper shelf can be clearly determined, following the definition given in ASTM E185-82 (Reference A.2-6).The functional form of the least squares method as described in Regulatory Position 2.1 will be utilized to determine a best-fit line for this data and to determine if the scatter of these ARTNDT values about this line is less than 28°F for welds and less than 17'F for the forging.The Sequoyah Unit 2 Intermediate Shell Forging 05 and surveillance weld will be evaluated for credibility.

The weld is made from weld wire heat # 4278. This weld metal is not in any other surveillance program.Following is the calculation of the best-fit line as described in Regulatory Position 2.1 of Regulatory Guide 1.99, Revision 2. In addition, the recommended NRC methods for determining credibility will be followed.

The NRC methods were presented to the industry at a meeting held by the NRC on February 12 and 13, 1998 (Reference A.2-7). At this meeting the NRC presented five cases. Of the five cases, Case 1 ("Surveillance data available from plant but no other source") most closely represents the situation listed above for the Sequoyah Unit 2 surveillance forging and weld materials.

WCAP-17539-NP March 2012 Revision 0 Westinghouse Non-Proprietary Class 3 A-11 Following the NRC Case 1 guidelines, the Sequoyah Unit 2 surveillance forging and weld metal (Heat # 4278) will be evaluated using the Sequoyah Unit 2 data. This evaluation is contained in Table A.2-1.Table A.2-1 Calculation of Interim Chemistry Factors for the Credibility Evaluation Using Sequoyah Unit 2 Surveillance Capsule Data Only Capsule Fluence(a)

FF(b) ARTNDT(c)

FF*ARTNDT FFr 2 RPV Material C(x1019 n/cm2, (OF) (OF)E > 1.0 MeV)T 0.244 0.618 63.65 39.33 0.382 IS Forging 05 U 0.654 0.881 79.31 69.87 0.776 (Tangential)

X 1.16 1.041 85.7 89.25 1.085 Y 2.02 1.192 134.12 159.83 1.420 T 0.244 0.618 48.73 30.11 0.382 IS Forging 05 U 0.654 0.881 66.06 58.20 0.776 (Axial) X 1.16 1.041 110.04 114.60 1.085 Y 2.02 1.192 89.21 106.31 1.420 SUM: 667.51 7.325 CFIs Forging 05 = Y(FF *ARTNDT) + E(FF2) = (667.51) -(7.325) 91.1°F T 0.244 0.618 74.56 46.07 0.382 Surveillance Weld U 0.654 0.881 130.38 114.86 0.776 Metal (Heat # 4278) X 1.16 1.041 44.22 46.05 1.085 Y 2.02 1.192 86.91 103.57 1.420 SUM: 310.56 3.663 CFHeat # 42 7 8= X(FF

ARTNDT) + I(FF 2) = (310.56) + (3.663) = 84.8°F Notes: (a) f = capsule fluence taken from Table 2-7 of this report.(b) FF = fluence factor = (°0.28- 0.10*log f).(c) ARTNDT values are the measured 30 ft-lb shift values taken from Table 5-10 of WCAP-15320 (Reference A.2-5). These measured ARTNDT values for the surveillance weld metal do not include the adjustment ratio procedure of Regulatory Guide 1.99, Revision 2, Position 2.1 since this calculation is based on the actual surveillance weld metal measured shift values. In addition, only Sequoyah Unit 1 data is being considered; therefore, no temperature adjustment is required.WCAP-17539-NP March 2012 Revision 0 A-12 Westinghouse Non-Proprietary Class 3 The scatter of ARTNDT values about the functional form of a best-fit line drawn as described in Regulatory Position 2.1 is presented in Table A.2-2.Table A.2-2 Sequoyah Unit 2 Surveillance Capsule Data Scatter about the Best-Fit Line Capsule Measured Predicted Scatter <17 0 F CF Fluence RPV Material Capsule (OF) (x101 9 n/cm 2 , FF ARTNDT ARTNDT ARTNDT (Base Metal)E > 1.0 MeV) (OF) (OF) (OF) <28 0 F (Weld)T 91.1 0.244 0.618 63.65 56.3 7.3 Yes IS Forging 05 U 91.1 0.654 0.881 79.31 80.3 1.0 Yes (Tangential)

X 91.1 1.16 1.041 85.7 94.9 9.2 Yes Y 91.1 2.02 1.192 134.12 108.6 25.5 No T 91.1 0.244 0:618 48.73 56.3 7.6 Yes IS Forging 05 U 91.1 0.654 0.881 66.06 80.3 14.2 Yes (Axial) X 91.1 1.16 1.041 110.04 94.9 15.1 Yes Y 91.1 2.02 1.192 89.21 108.6 19.4 No T 84.8 0.244 0.618 74.56 52.4 22.2 Yes Surveillance U 84.8 0.654 0.881 130.38 74.7 55.7 No Weld Metal (Heat # 4278) X 84.8 1.16 1.041 44.22 88.3 44.1 No Y 84.8 2.02 1.192 86.91 101.0 14.1 Yes The scatter of ARTNDT values 1.99, Revision 2, Position 2.1, about the best-fit line, drawn as described in Regulatory Guide should be less than 17'F for base metal. Table A.2-2 indicates that six of the eight surveillance data points fall within the +/- ly of 17'F scatter band for surveillance base metals; therefore, the forging data is deemed "credible" per the third criterion.

The scatter of ARTNDT values about the best-fit line, drawn as described in Regulatory Guide 1.99, Revision 2, Position 2.1, should be less than 28°F for weld metal. Table A.2-2 indicates that two of the four surveillance data points fall within the +/- la of 28°F scatter band for surveillance weld materials; therefore, the weld material is deemed "non-credible" per the third criterion.

WCAP- 17539-NP March 2012 Revision 0 Westinghouse Non-Proprietary Class 3 A-13 Criterion 4: The irradiation temperature of the Charpy specimens in the capsule should match the vessel wall temperature at the cladding/base metal interface within +/- 25°F.The Sequoyah Unit 2 capsule specimens are located in the reactor between the thermal shield and the vessel wall and are positioned opposite the center of the core. The test capsules are in baskets attached to the thermal shield. The location of the specimens with respect to the reactor vessel beltline provides assurance that the reactor vessel wall and the specimens experience equivalent operating conditions such that the temperatures will not differ by more than 25°F.Hence, Criterion 4 is met for the Sequoyah Unit 2 surveillance program.Criterion 5: The surveillance data for the correlation monitor material in the capsule should fall within the scatter band of the database for that material.The Sequoyah Unit 2 surveillance program does not contain correlation monitor material;therefore, this criterion is not applicable to the Sequoyah Unit 2 surveillance program.CONCLUSION:

Based on the preceding responses to the five criteria of Regulatory Guide 1.99, Revision 2, Section B, and the application of engineering judgment, the Sequoyah Unit 2 surveillance forging material is deemed credible, and the weld material is deemed non-credible.

Appendix A.2 References A.2-1 U.S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials," May 1988.A.2-2 10 CFR 50, Appendix G, "Fracture Toughness Requirements," Federal Register, Volume 60, No. 243, December 19, 1995.A.2-3 WCAP-8513, Revision 0, "Tennessee Valley Authority Sequoyah Unit No. 2 Reactor Vessel Radiation Surveillance Program," J. A. Davidson et al., November 1975.A.2-4 ASTM E185-73, "Standard Recommended Practice for Surveillance Tests for Nuclear Reactor Vessels," American Society for Testing and Materials, 1973.A.2-5 WCAP-15320, Revision 0, "Analysis of Capsule Y from the Tennessee Valley Authority Sequoyah Unit 2 Reactor Vessel Radiation Surveillance Program," T. J.Laubham et al., December 1999.WCAP- 17539-NP March 20 12 Revision 0 A-14 Westinghouse Non-Proprietary Class 3 A.2-6 ASTM E185-82, "Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels," American Society for Testing and Materials, 1982.A.2-7 K. Wichman, M. Mitchell, and A. Hiser, USNRC, Generic Letter 92-01 and RPV Integrity Workshop Handouts, NRC/Industry Workshop on RPV Integrity Issues, February 12, 1998.*WCAP- 17539-NP March 2012 Revision 0 Westinghouse Non-Proprietary Class 3 B-1 APPENDIX B SURVEILLANCE CAPSULE RELOCATION EVALUATION FOR SEQUOYAH UNITS 1 AND 2 B.1 SEQUOYAH UNIT 1 Four capsules (T, U, X and Y) have been withdrawn from the Sequoyah Unit 1 reactor vessel and tested, as recommended by ASTM E185-82 (Reference B.1-1). With the withdrawal of Capsule Y, Sequoyah Unit 1 fulfilled the surveillance capsule withdrawal recommendations contained in ASTM E185-82, as required by 10 CFR 50, Appendix H (Reference B.1-2), for a 40=year EOL (32 EFPY). Since Sequoyah Unit 1 is applying for a 20-year license extension, an additional capsule is expected to provide metallurgical data corresponding with an EOL fluence of 60 years (52 EFPY). Currently, there are four remaining capsules (W, V, S, and Z) in the Sequoyah Unit 1 reactor vessel.Capsules T, U, X and Y in the Sequoyah Unit 1 reactor vessel were positioned at the 400 azimuthal location, and were considered to be radiologically equivalent.

Similarly, Capsules W, V, S, and Z are currently located at the 40 azimuthal location in the Unit 1 reactor vessel, and are considered to be radiologically equivalent.

Note that the 40 azimuthal location is a lag (less than one) factor location; therefore, at this time, the Sequoyah Unit 1 reactor vessel is being irradiated slightly faster than the remaining capsules.

In order for Sequoyah Unit 1 to have meaningful metallurgical capsule data in the future, it is recommended that several of the remaining capsules be relocated to any of the empty 400 azimuthal capsule locations.

Capsule neutron fluence projections are summarized in Table B. I-1 for the Sequoyah Unit 1 40 and 400 azimuthal capsule locations.

Table B.1-1 Projected Neutron Fluence Values at the Geometric Center of the Surveillance Capsule Locations for Sequoyah Unit 1 Capsule Fluence Cycle EFPY (x10 1 9 n/cm 2 , E > 1.0 MeV)40 Azimuthal 40' Azimuthal Location Location 18 22.14 1.14 4.02 19 23.47 1.20 4.21 20 24.80 1.25 4.41--- 28.00 1.38 4.88--- 32.00 1.54 5.46--- 36.00 1.70 6.05--- 40.00 1.87 6.64--- 44.00 2.03 7.23--- 48.00 2.19 7.81--- 52.00 2.35 8.40 WCAP-17539-NP March 2012 Revision 0 B-2 Westinghouse Non-Proprietary Class 3 The fluence values listed in Table B.1-1 are used to determine neutron fluence projections assuming capsule relocation from a 40 to a 400 location beginning at end-of-cycles (EOC) 18, 19, and 20. Table B.1-2 below summarizes the projected neutron fluence values for any of the remaining Sequoyah Unit 1 40 capsules assuming they are relocated to any of the 400. locations at various relocation times.Table B.1-2 SequoyahIUnit 1IProjected Capsule Neutron Fluence Values Associated with Capsule Relocation from the 40 to the 400 Azimuthal Location Capsule Fluence Cycle EFPY (xW0 1 9 n/cm 2 , E > 1.0 MeV)Relocation at the Relocation at the Relocation at the EOC 18. EOC 19 EOC 20 18 22.14 1.14 1.14 1.14 19 23.47 1.33 1.20 1.20 20 24.80 1.53 1.40 1.25 28.00. 2.00 1.87 1.72--- 32.00 2.58 2.45 2.30--- 36.00 3.17 3.04 2.89 40.00 3.76 3.63 3.48--- 44.00 4.35 4.22 4.07--- 48.00 4.93 4.80 4.65 52.00 5.52 5.39 5.24 Since Sequoyah Unit I is applying for a 20-year license extension, an additional capsule is expected to satisfy the same criteria as the EOL capsule, as described in ASTM E185-82, with the EOL fluence at 60 years (52 EFPY). Therefore, a capsule should be withdrawn so that the capsule fluence corresponds to at least one times the 60-year EOL vessel fluence (2.66 x .1019 n/cm2, per Table 2-5), but less than two times the 60-year EOL vessel fluence (5.32 x 1019 n/cm 2). Based on the fluence projections in Table B.1-1, none of the remaining Sequoyah Unit 1 capsules, in their current azimuthal locations (40), would experience a neutron fluence of 2.66 x 1019 n/cm 2 prior to EOLE.However, based on the fluence projections in-Table B. 1-2, the peak 52,EPY calculated vesse fluence 4,2.66 x 10.14, V), would occur at approximately 32.5, 33.4, or 34.4 EFPY, assuming the capsule was relocated to a 400 azimuthal location at the EOCs 18, 19, or 20;,'respectively.

Furthermore, based on the fluence projections in Table B. 1-2, two times the peak 52 EFPY calculated vessel fluence of 5.32 x 10 n/cm 2 (E > 1.0 MeV) would occur at approximately 52 EFPY for a relocated capsule, assuming the capsule was relocated at the EOCs 18; 19, oIr620.'-.0-to a 400 azimuthal location.WCAP-17539-NP March 2012 Revision 0 Westinghouse Non-Proprietary Class 3 B-3 Additionally, it is anticipated at this time that if an additional 20-year license extension was sought, another capsule would be needed to be withdrawn from the reactor vessel in order to satisfy the same criteria as the EOL capsule with an EOL fluence at 80 years (72 EFPY). The extrapolated maximum neutron fluence value at 72 EFPY for Sequoyah Unit 1 is approximately 3.61 x 1019 n/cm 2 (E > 1.0 MeV). Based on the fluence projections in Table B.1-2, the peak 72 EFPY calculated vessel fluence of 3.61 x 1019 n/cm2 (E > 1.0 MeV) would occur at approximately 39.0, 39.9, or 40.9 EFPY, assuming the capsule was relocated to a 40' azimuthal location at the EOC 18, 19, or 20, respectively.

In summary, it is recommended that several of the Sequoyah Unit 1 remaining capsules be relocated to higher lead factor locations.

One of these relocated capsules should be subsequently withdrawn from the reactor vessel and tested at the time when the accumulated neutron fluence of the capsule corresponds to not less than once or greater than twice the peak 60-year vessel fluence. Another relocated capsule could be used for future testing, if additional license extensions are sought. Table B. 1-3 summarizes potential removal times for the relocated capsules based on license extension out to 60 and 80 years of operation.

These dates are based on the capsule fluence being equivalent to one times the peak vessel fluence at 60 years (2.66 x 1019 n/cm 2) as well as one times the peak vessel fluence at 80 years (3.61 x 1019 n/cm 2).Table B.1-3 Sequoyah Unit 1 Potential Capsule Withdrawal Times Associated with Capsule Relocation from the 40 to the 400 Azimuthal Location Capsule Capsule Time (EFPY) Corresponding to Vessel Life(a)Relocation Time 60 Years of Operation 80 Years of Operation (52 EFPY) (72 EFPY)EOC 18 32.5 39.0 EOC 19 33.4 39.9 EOC 20 34.4 40.9 Notes: (a) These dates are based on the capsule fluence being equivalent to one times the peak vessel fluence at 60 years (2.66 x 101 9 n/cm 2) as well as one times the peak vessel fluence at 80 years (3.61 x 1019 n/cm 2).Appendix B.1 References B.1-1 ASTM E185-82, "Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels," American Society for Testing and Materials, 1982.B. 1-2 Code of Federal Regulations, 10 CFR 50, Appendix H, "Reactor Vessel Material Surveillance Program Requirements," U.S. Nuclear Regulatory Commission, Washington, D.C., Federal Register, Volume 60, No. 243, dated December 19, 1995.WCAP- 17539-NP March 2012 Revision 0 B-4 Westinghouse Non-Proprietary Class 3 B.2 SEQUOYAH UNIT 2 Four capsules (T, U, X and Y) have been withdrawn from the Sequoyah Unit 2 reactor vessel and tested, as recommended by ASTM E185-82 (Reference B.2-1). With the withdrawal of Capsule Y, Sequoyah Unit 2 fulfilled the surveillance capsule withdrawal recommendations contained in ASTM E185-82, as required by 10 CFR 50, Appendix H (Reference B.2-2), for a 40-year EOL (32 EFPY). Since Sequoyah Unit 2 is applying for a 20-year license extension, an additional capsule is expected to provide metallurgical data corresponding with an EOL fluence of 60 years (52 EFPY). Currently, there are four remaining capsules (W, V, S, and Z) in the Sequoyah Unit 2 reactor vessel.Capsules T, U, X and Y in the Sequoyah Unit 2 reactor vessel were positioned at the 400 azimuthal location, and were considered to be radiologically equivalent.

Similarly, Capsules W, V, S, and Z are currently located at the 40 azimuthal location in the Unit 1 reactor vessel, and are considered to be radiologically equivalent.

Note that the 4' azimuthal location is a lag (less than one) factor location; therefore, at this time, the Sequoyah Unit 2 reactor vessel is being irradiated slightly faster than the remaining capsules.

In order for Sequoyah Unit 2 to have meaningful metallurgical capsule data in the future, it is recommended that several of the remaining capsules be relocated to any of the empty 400 azimuthal capsule locations.

Capsule neutron fluence projections are summarized in Table B.2-1 for the Sequoyah Unit 2 40 and 400 azimuthal capsule locations.

Table B.2-1 Projected Neutron Fluence Values at the Geometric Center of the Surveillance Capsule Locations for Sequoyah Unit 2 Capsule Fluence Cycle EFPY (x10 1 9 n/cm 2 , E > 1.0 MeV)40 Azimuthal 400 Azimuthal Location Location 18 22.97 1.18 3.91 19 24.34 1.24 4.13 20 25.70 1.30 4.32--- 28.00 1.40 4.65--- 32.00 1.58 5.22--- 36.00 1.76 5.78--- 40.00 1.94 6.35--- 44.00 2.12 6.92--- 48.00 2.30 7.48 52.00 2.48 8.05 WCAP-17539-NP March 2012 Revision 0 Westinghouse Non-Proprietary Class 3 B-5 The fluence values listed in Table B.2-1 are used to determine neutron fluence projections assuming capsule relocation from a 40 to a 40' location beginning at end-of-cycles (EOC) 18, 19, and 20. Table B.2-2 below summarizes the projected neutron fluence values for any of the remaining Sequoyah Unit 2 40 capsules assuming they are relocated to any of the 400 locations at various relocation times.Table B.2-2 Sequoyah Unit 2 Projected Capsule Neutron Fluence Values Associated with Capsule Relocation from the 4' to the 400 Azimuthal Location Capsule Fluence Cycle EFPY (xl0 1 9 n/cm 2 , E > 1.0 MeV)Relocation at the Relocation at the Relocation at the EOC 18 EOC 19 EOC 20 18 22.97 1.18 1.18 1.18 19 24.34 1.40 1.24 1.24 20 25.70 1.59 1.43 1.30 28.00 1.92 1.76 1.62--- 32.00 2.49 2.33 2.19--- 36.00 3.05 2.89 2.75--- 40.00 3.62 3.46 3.32--- 44.00 4.19 4.03 3.89--- 48.00 4.75 4.59 4.45--- 52.00 5.32 5.16 5.02 Since Sequoyah Unit 2 is applying for a 20-year license extension, an additional capsule is expected to satisfy the same criteria as the EOL capsule, as described in ASTM E185-82, with the EOL fluence at 60 years (52 EFPY). Therefore, a capsule should be withdrawn so that the capsule fluence corresponds to at least one times the 60-year EOL vessel fluence (2.57 x 1019 n/cm2, per Table 2-5), but less than two times the 60-year EOL vessel fluence (5.14 x 1019 n/cmr). Based on the fluence projections in Table B.2-1, none of the remaining Sequoyah Unit 2 capsules, in their current azimuthal locations (40), would experience a neutron fluence of 2.57 x 1019 n/cm 2 prior to EOLE.However, based on the fluence projections in Table the peak 52 EFPY calculated vessel fluence ofl2.57 x10 -n/cm E> 10 MeV) would occur at approximately 32.6, 33.7, or 34.7 EFPY, assuming the capsule was relocated to a 400 azimuthal location at the EOCs Lor 20;respectively.

Furthermore, based on the fluence projections in Table B.2-2, two times the peak 52 EFPY calculated vessel fluence of 5.14 x 10 n/cm 2 (E > 1.0 MeV) would occur at approximately 52 EFPY for a relocated capsule, assuming the capsule was relocated at the EOCs 18, 19, or 20 to a 400 azimuthal location.WCAP-17539-NP March 2012 Revision 0 B-6 Westinghouse Non-Proprietary Class 3 Additionally, it is anticipated at this time that if an additional 20-year license extension was sought, another capsule would be needed to be withdrawn from the reactor vessel in order to satisfy the same criteria as the EOL capsule with an EOL fluence at 80 years (72 EFPY). The extrapolated maximum neutron fluence value at 72 EFPY for Sequoyah Unit 2 is approximately 3.52 x 1019 n/cm 2 (E > 1.0 MeV). Based on the fluence projections in Table B.2-2, the peak 72 EFPY calculated vessel fluence of 3.52 x 1019 n/cm 2 (E > 1.0 MeV) would occur at approximately 39.3, 40.4 or 41.4 EFPY, assuming the capsule was relocated to a 400 azimuthal location at the EOC 18, 19, or 20, respectively.

In summary, it is recommended that several of the Sequoyah Unit 2 remaining capsules be relocated to higher lead factor locations.

One of these relocated capsules should be subsequently withdrawn from the reactor vessel and tested at the time when the accumulated neutron fluence of the capsule corresponds to not less than once or greater than twice the peak 60-year vessel fluence. Another relocated capsule could be used for future testing, if additional license extensions are sought. Table B.2-3 summarizes potential removal times for the relocated capsules based on license extension out to 60 and 80 years of operation.

These dates are based on the capsule fluence being equivalent to one times the peak vessel fluence at 60 years (2.57 x 1019 n/cm 2) as well as one times the peak vessel fluence at 80 years (3.52 x 1019 n/cm 2).Table B.2-3 Sequoyah Unit 2 Potential Capsule Withdrawal Times Associated with Capsule Relocation from the 40 to the 400 Azimuthal Location Capsule Capsule Time (EFPY) Corresponding to Vessel Life(*)Relocation Time 60 Years of Operation 80 Years of Operation (52 EFPY) (72 EFPY)EOC 18 32.6 39.3 EOC 19 33.7 40.4 EOC 20 34.7 41.4 Notes: (a) These dates are based on the capsule fluence being equivalent to one times the peak vessel fluence at 60 years (2.57 x 1019 n/cm 2) as well as one times the peak vessel fluence at 80 years (3.52 x 1019 n/cm 2).Appendix B.2 References B.2-1 ASTM E185-82, "Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels," American Society for Testing and Materials, 1982.B.2-2 Code of Federal Regulations, 10 CFR 50, Appendix H, "Reactor Vessel Material Surveillance Program Requirements," U.S. Nuclear Regulatory Commission, Washington, D.C., Federal Register, Volume 60, No. 243, dated December 19, 1995.WCAP-17539-NP March 2012 Revision 0 Westinghouse Non-Proprietary Class 3 C-1 APPENDIX C EMERGENCY RESPONSE GUIDELINE LIMITS The Emergency Response Guideline (ERG) limits were developed to establish guidance for operator action in the event of an emergency situation, such as a PTS event (Reference C-i).Generic categories of limits were developed for the guidelines based on the limiting inside surface RTNDT. These generic categories were conservatively generated for the Westinghouse Owners Group (WOG) to be applicable to all Westinghouse plants.The highest value of RTNDT for which the generic category ERG limits were developed is 2507F for a longitudinal flaw and 300'F for a circumferential flaw. Therefore, if the limiting vessel material has an RTNDT that exceeds 2507F for a longitudinal flaw or 3007F for a circumferential flaw, plant-specific ERG P-T limits must be developed.

The ERG category is determined by the magnitude of the limiting RTNDT value, which is calculated the same way as the RTprs values are calculated in Section 4 of this report. The material with the highest RTNDT defines the limiting material, which for Sequoyah Unit 1 is LS Forging 04 (Position 2.1) and for Sequoyah Unit 2 is IS to LS Circ. Weld W05 (Position 2.1).Table C-i identifies ERG category limits and the limiting material RTNDT values at 52 EFPY for Sequoyah Units 1 and 2.Table C-1 Evaluation of Sequoyah Units 1 and 2 ERG Limit Category ERG Pressure-Temperature Limits (Reference C-i)Applicable RTNDT Value a) ERG P-T Limit Category RTNDT < 200°F Category I 200OF < RTNDT < 250°F Category II 250°F < RTNDT < 300OF Category III b Limiting RTNDT Values(b)Reactor Vessel Material RTNDT Value @ 52 EFPY Unit 1 LS Forging 04 with Credible Surveillance Data 227.9°F Unt 2 IS to LS Circ. Weld W05 (Heat # 4278) 150.7°F with Non-Credible Surveillance Data Notes: (a) Longitudinally oriented flaws are applicable only up to 250'F; circumferentially oriented flaws are applicable up to 300'F.(b) Values taken from Tables 4-1 and 4-2 for Sequoyah Units 1 and 2, respectively.

WCAP-17539-NP March 2012 Revision 0 C-2 Westinghouse Non-Proprietary Class 3 Per the ERG limit guidance document (Reference C-i), some vessels do not change categories for operation through the end-of-license.

However, when a vessel does change ERG categories between the beginning and end of operation, a plant-specific assessment must be performed to determine at what operating time the category changes. Thus, the ERG classification need not be changed until the operating cycle during which the maximum vessel value of actual or estimated real-time RTNDT exceeds the limit on its current ERG category.Unit 1 Per Table C-i, the limiting material RTNDT for Sequoyah Unit 1 is 227.97F at 52 EFPY. This value corresponds to the Lower Shell Forging 04. Thus, the limiting material RTNDT value exceeds the ERG Category I criterion (RTNDT < 2007F) prior to 52 EFPY. The transition occurs when RTNDT = 200'F. The operating cycle at which the ERG category transitioned from Category I to Category II was determined to be prior to Cycle 15. Sequoyah Unit 1 will remain in ERG Category Unit II through EOLE.Unit 2 Per Table C-i, the limiting material for Sequoyah Unit 2 (Intermediate Shell to Lower Shell Circumferential Weld) has an RTNDT less than 200'F through 52 EFPY. Therefore, Sequoyah Unit 2 remains in ERG Category I through EOLE (52 EFPY).Conclusion of ERG P-T Limit Categorization As summarized above, Sequoyah Unit 1 is currently in ERG Category II and will remain in ERG Category Unit II through EOLE (52 EFPY). Sequoyah Unit 2 is currently in ERG Category I and remains in ERG Category I through EOLE (52 EFPY).Appendix C Reference C-1 "Background Information for Westinghouse Owners Group Emergency Response Guidelines, Critical Safety Function Status Tree, F-0.4 Integrity, HP/LP-Rev.

2," Westinghouse Owners Group, April 30, 2005.WCAP- 17539-NP March 2012 Revision 0

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