ENCLOSURE 3 Tennessee Valley Authority Sequoyah Nuclear … · ENCLOSURE 3 Tennessee Valley...

ENCLOSURE 3

Tennessee Valley AuthoritySequoyah Nuclear Plant Units I and 2

WCAP-17539-NP, Revision 0, "Sequoyah Units 1 and 2 Time-Limited AgingAnalysis on Reactor Vessel Integrity"

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Westinghouse Non-Proprietary Class 3

WCAP-17539-NP ..arc.~Revision 0

Sequoyah Units 1 and 2Time-Limited Aging Analysison Reactor Vessel Integrity

Westinghouse

h 201211

WESTINGHOUSE NON-PROPRIETARY CLASS 3

WCAP-17539-NPRevision 0

Sequoyah Units 1 and 2 Time-Limited Aging Analysis onReactor 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 ManagerAging Management and License Renewal Services

*Electronically approved records are authenticated in the electronic document management system.

Westinghouse Electric Company LLC1000 Westinghouse Drive

Cranberry Township, PA 16066

© 2012 Westinghouse Electric Company LLCAll Rights Reserved

Westinghouse Non-Proprietary Class 3

RECORD OF REVISION

iii

Revision 0: Original Issue

WCAP-17539-NP March 2012Revision 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 .......................................... 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


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 ReactorVessel Clad/Base Metal Interface for Sequoyah Unit 1 Beltline Materials ..................... 2-3

Table 2-2 Calculated Neutron Fluence Projections at the Peak Location on the ReactorVessel Clad/Base Metal Interface for Sequoyah Unit 1 Extended BeltlineM aterials .......................................................................................................................... 2 -4

Table 2-3 Calculated Neutron Fluence Projections at the Peak Location on the ReactorVessel Clad/Base Metal Interface for Sequoyah Unit 2 Beltline Materials ..................... 2-5

Table 2-4 Calculated Neutron Fluence Projections at the Peak Location on the ReactorVessel Clad/Base Metal Interface for Sequoyah Unit 2 Extended BeltlineM aterials .......................................................................................................................... 2 -6

Table 2-5 Summary of the Sequoyah Units 1 and 2 Maximum RPV Fluence on the ReactorVessel 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, andInitial USE Values for the Sequoyah Unit 1 RPV Beltline and Extended BeltlineM aterials .......................................................................................................................... 3 -5

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

Table 3-3 Calculation of Position 2.1 CF Values using Sequoyah Unit 1 SurveillanceC apsule Test R esults ....................................................................................................... 3-9

Table 3-4 Calculation of Position 2.1 CF Values using Sequoyah Unit 2 SurveillanceC apsule Test R esults ...................................................................................................... 3-10

Table 3-5 Summary of the Sequoyah Unit 1 RPV Beltline and Extended Beltline MaterialChemistry Factor Values based on Regulatory Guide 1.99, Revision 2, Position1.1 and Position 2.1 ................................................................................................... 3-11

Table 3-6 Summary of the Sequoyah Unit 2 RPV Beltline and Extended Beltline MaterialChemistry Factor Values based on Regulatory Guide 1.99, Revision 2, Position1.1 and P osition 2 .1 ........................................................................................................ 3-11

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


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


E F P Y ................................................................................................................................ 6 -5


E F P Y ................................................................................................................................ 6 -6


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 32E F P Y .............................................................................................................................. 6 -11



E F P Y .............................................................................................................................. 6 -13


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 UsingSequoyah 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 SurveillanceCapsule Locations for Sequoyah Unit 1 .................................................................... B-1

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

WCAP- 17539-NP March 2012Revision 0

Westinghouse Non-Proprietary Class 3 vii

Table B. 1-3 Sequoyah Unit 1 Potential Capsule Withdrawal Times Associated with CapsuleRelocation from the 40 to the 400 Azimuthal Location ................................................... B-3

Table B.2-1 Projected Neutron Fluence Values at the Geometric Center of the SurveillanceCapsule Locations for Sequoyah Unit 2 ......................................................................... B-4

Table B.2-2 Sequoyah Unit 2 Projected Capsule Neutron Fluence Values Associated withCapsule Relocation from the 40 to the 40' Azimuthal Location ..................................... B-5

Table B.2-3 Sequoyah Unit 2 Potential Capsule Withdrawal Times Associated with CapsuleRelocation 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


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 aFunction 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 aFunction of Copper and Fluence for Sequoyah Unit 2 .................................................... 5-6


Westinghouse Non-Proprietary Class 3 ix

EXECUTIVE SUMMARY

This report presents the Time-Limited Aging Analyses (TLAA) for the Sequoyah Units 1 and 2reactor pressure vessels in accordance with the requirements of the License Renewal Rule, 10CFR Part 54. Time-Limited Aging Analyses are calculations which evaluate some safety-relatedaspects of the reactor pressure vessel within the bounds of the current 40-year license that mustbe 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 fullpower 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 52EFPY, which is deemed end-of-license extension (EOLE). Updated neutron fluence evaluationswere 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 vesselintegrity (RVI) evaluations in support of license renewal.

In addition to the RVI TLAA evaluations, the credibility of the Sequoyah Units 1 and 2surveillance materials was also evaluated. Conclusions for the surveillance data credibilityevaluations are contained in Appendix A of this report. Appendix B contains recommendationsfor capsule relocations in order to obtain meaningful metallurgical data for the future. AppendixC contains the Emergency Response Guideline (ERG) limits classification for Sequoyah Units1 and 2. The ERG limits were developed in order to establish guidance for operator action in theevent of an emergency situation, such as a PTS event. Conclusions for the ERG limitsevaluations 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 theresults of this TLAA evaluation, it is concluded that the Sequoyah Units 1 and 2 reactor vesselswill 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 vesselsare below the RTPTs screening criteria values of 2707F, for forgings, and 3007F, forcircumferentially oriented welds (Per 10 CFR 50.61), through EOLE (52 EFPY). See Section 4for more details.

EOLE Upper-Shelf Energy

All of the beltline and extended beltline materials in the Sequoyah Units 1 and 2 reactor vesselsare 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.


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 factorvalues based on surveillance data, and the consideration of TLAA fluence projections, theapplicability of the Sequoyah Units 1 and 2 pressure-temperature limit curves may either remainunchanged 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 (32EFPY). Several additional capsules for each Unit should be relocated to higher lead factorlocations. One of these relocated capsules in each Unit should be withdrawn from the reactorvessels in order to achieve 60-year (52 EFPY) fluence data prior to EOLE. See Section 7 formore details.


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 theSSC 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 withindication of whether or not they meet the six criteria of 10 CFR 54.3 (Reference 1) for TLAAs.


1-2 Westinghouse Non-Proprietary Class 31-2 Westinghouse Non-Proprietary Class 3

Table 1-1 Evaluation of Time-Limited Aging Analyses Per the Criteria of 10 CFR54.3

Pressure-

Calculated Pressurized Upper- TemperatureTime-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-LimitedAssumptions Defined by the YES YES YES YES

Current Operating Term

Involves SSC Within the Scope of YES YES YES YESLicense Renewal

Involves Conclusions or Providesthe Basis for Conclusions Related

to the Capability of SSC to PerformIts Intended Function

Determined to be Relevant by theLicensee in Making a Safety YES YES YES YES

Determination

Contained or Incorporated by YES YES YES YESReference 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. SeeAppendix C for additional information.



2" CALCULATED FLUENCE

At currently licensed service times and operating conditions, the Sequoyah Units 1 and 2 RPVfracture toughness properties provide adequate margins of safety against vessel failure. However,as a vessel accumulates more and more service time, neutron irradiation (fluence) reducesmaterial fracture toughness and initial safety margins. Prevention of RPV failure dependsprimarily on maintaining RPV material fracture toughness at levels that resist brittle fractureduring plant operation. The first step in the TLAA of vessel embrittlement is the calculation ofthe neutron fluence that causes the embrittlement to increase with time.

The reactor vessel beltline neutron fluence values applicable to a postulated 20-year licenserenewal 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 satisfythe requirements set forth in Regulatory Guide 1.190, "Calculational and Dosimetry Methods forDetermining Pressure Vessel Neutron Fluence" (Reference 2). These methodologies have beenapproved by the US NRC and are described in detail in WCAP-14040-A, Revision 4 (Reference3) 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 materialsexceeding 1.0 x 1017 n/cm 2 (E > 1.0 MeV) must be monitored to evaluate the changes in fracturetoughness. RPV materials that are not traditionally thought of as being plant limiting because oflow levels of neutron radiation must now be evaluated to determine the accumulated fluence atEOLE. Therefore, fluence calculations were performed for the Sequoyah Units 1 and 2 RPVinlet nozzle to upper shell welds, upper to intermediate shell circumferential welds, lower shell tobottom head ring circumferential welds, and the bottom head ring to bottom head circumferentialwelds, 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 notevaluated because they experience lower fluence levels, as comparedto the inlet nozzle to uppershell welds, due to a higher elevation relative to the active core. The materials that exceed the1.0 x 1017 n/cm 2 (E > 1.0 MeV) threshold are referred to as extended beltline materials in thisreport 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 through18 for Unit 1 and cycles 1 through 17 for Unit 2, and projections for future operation throughEOLE, which is 60 years of plant life or 52 EFPY of operation. In all cases, the maximumexposure 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 werebased on the continued use of the average core data of Cycles 16, 17, and 18 for Unit 1 andCycles 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 foreach of the reactor pressure vessel beltline and extended beltline materials, respectively. Similardata for Sequoyah Unit 2 are provided in Tables 2-3 and 2-4.


2-2 Westinghouse Non-Proprietary Class 3

From Table 2-2, it is noted that, although the upper shell course and the upper shell tointermediate shell circumferential weld are projected to exceed the 1.0 x 101 n/cm2 (E > 1.0MeV) threshold neutron exposure defining the beltline region, the inlet nozzle to upper shellweld remains below 1.0 x 1017 n/cm 2 through 52 EFPY of operation. Likewise, the bottom headring to bottom head circumferential weld remains outside of the extended beltline region through52 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 thecalculations contained within this report. The peak neutron fluence at 32 EFPY and 52 EFPY forthe beltline materials corresponds to the intermediate to lower shell forgings. The peak neutronfluence at 52 EFPY for the extended beltline materials corresponds to the lower shell to bottomhead ring circumferential weld along with the bottom head forgings. These maximum neutronfluence values are summarized in Table 2-5.

Four surveillance capsules have been withdrawn from each of the Sequoyah Plants. Thecalculated fast neutron fluences at the 40' azimuthal surveillance capsule location are shown inTables 2-6 and 2-7 for Units 1 and 2, respectively.


Westinghouse Non-Proprietary Class 3 2-3Westinghouse Non-Proprietary Class 3 2-3

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

Intermediate Shell Intermediate Shell Lower ShellOperating 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+182.85 2.15E+18 2.15E+18 2.15E+184.03 2.85E+18 2.85E+18 2.85E+185.26 3.60E+18 3.59E+18 3.60E+186.26 4.11E+18 4.10E+18 4.11E+187.49 4.80E+ 18 4.78E+18 4.80E+ 188.72 5.46E+18 5.42E+18 5.46E+1810.02 6.19E+18 6.16E+18 6.19E+1811.38 6.88E+18 6.85E+18 6.88E+1812.82 7.64E+18 7.61E+18 7.64E+1814.12 8.42E+ 18 8.40E+ 18 8.42E+ 1815.44 9.12E+ 18 9.09E+ 18 9.12E+ 1816.80 9.83E+18 9.80E+18 9.83E+1818.17 1.06E+19 1.06E+19 1.06E+1919.51 1.15E+19 1.15E+19 1.15E+1920.90 1.21E+19 1.21E+19 1.21E+1922.14 1.27E+19 1.27E+19 1.27E+1923.47 1.33E+19 1.33E+19 1.33E+1924.00 1.36E+19 1.35E+19 1.36E+1928.00 1.54E+19 1.54E+19 1.54E+1932.00 1.73E+19 1.72E+19 1.73E+1936.00 1.92E+19 1.91E+19 1.92E+1940.00 2.10E+19 2.09E+19 2.10E+1944.00 2.29E+19 2.28E+19 2.29E+1948.00 2.47E+19 2.46E+19 2.47E+1952.00 2.66E+19 2.65E+19 2.66E+19



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

Lower Shell Bottom Head

Operating Inlet Nozzle to Upper to Bottom Bottom Ring toTime to Upper Intermediate Shell Head Bottom Head

Shell Welds Shell Circ. Forging Circ. Weld Ring Circ. WeldWeld


1.07 5.78E+14 1.03E+16 1.03E+16 7.37E+16 7.37E+16 4.73E+131.90 1.23E+15 2.25E+16 2.25E+16 1.38E+17 1.38E+17 8.76E+132.85 1.98E+15 3.65E+16 3.65E+16 2.35E+17 2.35E+17 1.49E+144.03 2.70E+15 4.99E+16 4.99E+16 3.20E+17 3.20E+17 2.03E+145.26 3.56E+15 6.49E+16 6.49E+16 4.13E+17 4.13E+17 2.66E+146.26 4.17E+15 7.49E+16 7.49E+16 4.77E+17 4.77E+17 3.13E+147.49 5.06E+15 8.92E+16 8.92E+16 5.64E+17 5.64E+17 3.78E+148.72 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+1411.38 7.61E+15 1.31E+17 1.31E+17 8.10E+17 8.10E+17 5.59E+1412.82 8.70E+15 1.49E+17 1.49E+17 9.08E+17 9.08E+17 6.32E+1414.12 9.75E+15 1.66E+17 1.66E+17 1.01E+18 1.01E+18 7.03E+1415.44 1.07E+16 1.81E+17 1.81E+17 1.10E+18 1.10E+18 7.72E+1416.80 1.17E+16 1.98E+17 1.98E+17 1.19E+18 1.19E+18 8.39E+1418.17 1.28E+16 2.15E+17 2.15E+17 1.28E+18 1.28E+18 9.07E+1419.51 1.41E+16 2.37E+17 2.37E+17 1.40E+18 1.40E+18 9.93E+1420.90 1.50E+16 2.51E+17 2.51E+17 1.48E+18 1.48E+18 1.05E+1522.14 1.58E+16 2.64E+17 2.64E+17 1.55E+18 1.55E+18 1.11E+1523.47 1.67E+16 2.78E+17 2.78E+17 1.63E+18 1.63E+18 1.17E+1524.00• 1.71E+16 2.83E+17 2.83E+17 1.67E+18 1.67E+18 1.20E+1528.00 1.97E+16 3.26E+17 3.26E+17 1.91E+18 1.91E+18 1.38E+1532.00 2.24E+16 3.69E+17 3.69E+17 2.15E+18 2.15E+18 1.56E+1536.00 2.51E+16 4.12E+17 4.12E+17 2.39E+18 2.39E+18 1.74E+1540.00 2.78E+16 4.55E+17 4.55E+17 2.63E+18 2.63E+18 1.93E+1544.00 3.05E+16 4.98E+17 4.98E+17 2.88E+18 2.88E+18 2.11E+1548.00 3.31E+16 5.41E+17 5.41E+17 3.12E+18 3.12E+18 2.29E+1552.00 3.58E+16 5.84E+17 5.84E+17 3.36E+18 3.36E+18 2.48E+15



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

Intermediate Shell Intermediate Shell Lower ShellOperating Time Forging to Lower Shell Forging

Circ. Weld F


1.07 7.85E+17 7.38E+17 7.85E+17

1.88 1.50E+18 1.45E+18 1.50E+182.91 2.07E+18 2.01E+18 2.07E+184.15 2.87E+18 2.82E+18 2.87E+185.36 3.64E+18 3.58E+18 3.64E+186.63 4.41E+18 4.33E+18 4.41E+187.95 4.98E+18 4.89E+ 18 4.98E+189.16 5.66E+ 18 5.54E+ 18 5.66E+ 1810.55 6.40E+ 18 6.28E+18 6.40E+ 1811.98 7.09E+ 18 6.96E+ 18 7.09E+ 1813.38 7.79E+18 7.66E+ 18 7.79E+1814.75 8.51E+I 8 8.38E+1 8 8.51E+1 816.04 9.22E+18 9.10E+18 9.22E+1817.51 9.99E+18 9.87E+18 9.99E+1818.84 1.07E+19 1.06E+19 1.07E+1920.17 1.13E+19 1.11E+19 1.13E+1921.60 1.19E+19 1.18E+19 1.19E+1922.97 1.25E+19 1.24E+19 1.25E+1924.00 1.30E+19 1.29E+19 1.30E+1928.00 1.48E+19 1.47E+19 1.48E+1932.00 1.66E+19 1.65E+19 1.66E+1936.00 1.84E+19 1.83E+19 1.84E+1940.00 2.02E+19 2.01E+19 2.02E+ 1944.00 2.20E+19 2.19E+ 19 2.20E+1948.00 2.38E+19 2.37E+19 2.38E+1952.00 2.57E+ 19 2.55E+19 2.57E+19



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

Inlet Upper Shell Lower Bottom

Operating Nozzle to to Upper Shell to Bottom Head RingTime 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+131.88 1.04E+15 1.91E+16 1.91E+16 1.42E+17 1.42E+17 8.75E+132.91 1.47E+15 2.68E+16 2.68E+16 1.93E+17 1.93E+17 1.21E+144.15 2.29E+15 4.11E+16 4.11E+16 2.85E+17 2.85E+17 1.83E+145.36 2.97E+15 5.30E+16 5.30E+16 3.66E+17 3.66E+17 2.38E+146.63 3.83E+15 6.75E+16 6.75E+16 4.61E+17 4.61E+17 3.05E+147.95 4.51E+15 7.84E+16 7.84E+16 5.33E+17 5.33E+17 3.58E+149.16 5.29E+15 9.11E+16 9.11E+16 6.12E+17 6.12E+17 4.16E+1410.55 6.24E+15 1.07E+17 1.07E+17 6.98E+17 6.98E+17 4.80E+1411.98 7.18E+15 1.22E+17 1.22E+17 7.81E+17 7.81E+17 5.41E+1413.38 8.20E+15 1.38E+17 1.38E+17 8.73E+17 8.73E+17 6.10E+1414.75 9.21E+15 1.55E+17 1.55E+17 9.67E+17 9.67E+17 6.80E+1416.04 1.02E+16 1.71E+17 1.71E+17 1.05E+18 1.05E+18 7.44E+1417.51 1.13E+16 1.88E+17 1.88E+17 1.15E+18 1.15E+18 8.17E+1418.84 1.22E+16 2.03E+17 2.03E+17 1.24E+18 1.24E+18 8.81E+1420.17 1.31E+16 2.17E+17 2.17E+17 1.31E+18 1.31E+18 9.36E+1421.60 1.41E+16 2.33E+17 2.33E+17 1.40E+18 1.40E+18 1.00E+1522.97 1.50E+16 2.47E+17 2.47E+17 1.48E+18 1.48E+18 1.06E+1524.00. 1.56E+16 2.58E+17 2.58E+17 1.54E+18 1.54E+18 1.11E+1528.00 1.83E+16 3.OOE+17 3.00E+17 1.77E+18 1.77E+18 1.28E+1532.00 2.09E+16 3.42E+17 3.42E+17 2.OOE+18 2.00E+18 1.46E+1536.00 2.35E+16 3.84E+17 3.84E+17 2.23E+18 2.23E+18 1.63E+1540.00 2.61E+16 4.26E+17 4.26E+17 2.47E+18 2.47E+18 1.80E+1544.00 2.87E+16 4.68E+17 4.68E+17 2.70E+18 2.70E+18 1.98E+1548.00 3.13E+16 5.10E+17 5.10E+17 2.93E+18 2.93E+18 2.15E+1552.00 3.40E+16 5.52E+17 5.52E+17 3.16E+18 3.16E+18 2.33E+15



Table 2-5 Summary of the Sequoyah Units 1 and 2 Maximum RPV Fluence onthe 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 1Surveillance 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 2Surveillance Capsules (40* Azimuthal Location)

Capsule EFPY Neutron Fluence(Cycle Withdrawn) (n/cm2, 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)



3 MATERIAL PROPERTY INPUT

The Sequoyah Units 1 and 2 reactor pressure vessels were fabricated by Rotterdam DrydockCompany (RDM). The Sequoyah Units 1 and 2 beltline materials consist of the IntermediateShell (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 40weld wire type, heat # 25295 and SMIT 89 flux type, lot # 2275. The weld material in theSequoyah Unit 1 surveillance program was made of the same material as the Unit 1 reactorvessel beltline circumferential weld, and is not in any other plant's surveillance program. TheSequoyah Unit 2 reactor vessel beltline circumferential weld was fabricated using Arcos weldwire type, heat # 4278 and SMIT 89 flux type, lot # 1211. The weld material in the SequoyahUnit 2 surveillance program was made of the same material as the Unit 2 reactor vessel beltlinecircumferential 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 2extended beltline materials and are evaluated to determine their impact on the proposed licenserenewal period. The Sequoyah Units 1 and 2 extended beltline materials consist of the UpperShell (US) Forging 06, Bottom Head Ring 03, US to IS Circumferential Weld W06, and the LSto Bottom Head Ring Weld W04. The Sequoyah Unit 1 US to IS Circumferential Weld W06was 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 withSMIT 40 weld wire type, heat # 25295 and SMIT 89 flux type, lot # 1135, which is the samematerial 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 numbers25006 and 721858.

The identification of the reactor vessel beltline and extended beltline materials are included inFigure 3-1 for Sequoyah Units 1 and 2. The material property inputs used for the subsequentRVI evaluations contained in this report are described in this section. Note that the sources andmethods used to determine the extended beltline material properties are consistent with thoseused in the past to determine the initial properties for the beltline materials. The sources andmethods used in the determination of the chemical compositions and the fracture toughnessproperties are summarized below.

Chemical Compositions

The best-estimate copper (Cu), nickel (Ni), phosphorus (P), and manganese (Mn) chemicalcompositions for the Sequoyah Units 1 and 2 beltline and extended beltline materials arepresented in Tables 3-1 and 3-2, respectively. The best-estimate weight percent copper andnickel values for the beltline materials were previously reported, and were used in past RVIevaluations. The best-estimate weight percent copper and nickel values for the extended beltlinematerials, along with the best-estimate manganese and phosphorus for the beltline and extendedbeltline materials were determined as part of this TLAA effort. Note that the best-estimate



manganese and phosphorus values are reported for information purposes only, and are not usedin any subsequent RVI evaluations contained within this report.

Except for the weight percent copper values, Certified Material Test Report (CMTR) data wasused to determine the chemical compositions for all of the Sequoyah Units 1 and 2 beltline andextended beltline forging materials. Weight percent copper values were not reported in theCMTRs for the extended beltline forging materials; therefore, the maximum weight percentcopper value for A508 Class 2 forging materials was conservatively applied based on the genericdata 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 weldmaterials (heat # 25295) were previously documented in WCAP-15293, Revision 2 (Reference7). The best-estimate phosphorus and manganese for these weld materials were determinedusing test records from the Rotterdam Weld Files as well as WCAP-8233 (Reference 8). Limitedinformation was available for the Sequoyah Unit 1 extended beltline US to IS circumferentialweld (heat # 25006) in the Rotterdam weld certification records. Except for the weight percentnickel value, the chemical compositions were taken from a chemical analysis performed on theweld wire (heat # 25006) included in the Rotterdam weld certification records. Weight percentnickel was not reported in the weld certification records for heat # 25006; therefore, a value of1.0 was conservatively assumed per 10 CFR 50.61 (Reference 9). The LS to Bottom Head Ringcircumferential weld was fabricated using the same weld wire heat number and flux type as theIS to LS circumferential weld. Therefore, the chemical compositions of the IS to LScircumferential 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 weldmaterials (heat # 4278) were previously documented in WCAP-15321, Revision 2 (Reference10). The best-estimate phosphorus and manganese for these weld materials were determinedusing Rotterdam test records as well as WCAP-8513 (Reference 11). The weight percent copperand manganese values for the Sequoyah Unit 2 extended beltline welds (heat # 721858) weretaken 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 wiresince an analysis for phosphorus was not performed on the as-deposited weld. Weight percentnickel 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 pressureboundary 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 1and 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 inWCAP-15293, Revision 2 (Reference 7) and WCAP-15321, Revision 2 (Reference 10),respectively. The fracture toughness properties for the extended beltline forging materials are



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 Ringmaterials as well as the Unit 2 Upper Shell Forging material were reduced to 65% of the originalvalues 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 noreported 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 testtemperature 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 resultsfrom the first surveillance capsule withdrawn from Sequoyah Unit 1 were used in accordancewith Section B.1.2 of NUREG-0800 Branch Technical Position 5-3. Weld heat # 25295 is aRotterdam weld of the same type (SMIT 40 with SMIT 89 flux). All surveillance weld datapoints 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 resultsfrom 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 forthis weld heat. In accordance with Section B. 1.1(4) of NUREG-0800 Branch Technical Position5-3, this test temperature was used as an estimate of the initial RTNDT since at least 45 ft-lbs wasobtained. Furthermore, in absence of USE data for weld heat # 721858, the lowest initial USEvalue 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 RegulatoryGuide 1.99, Revision 2 (Reference 14). Position 1.1 uses Tables 1 and 2 from the RegulatoryGuide along with the best-estimate copper and nickel weight percents, which are presented inTables 3-1 and 3-2 of this report for Sequoyah Units 1 and 2, respectively. Position 2.1 uses thesurveillance capsule data from all capsules withdrawn and tested to date. The calculated fluencevalues at the surveillance capsule locations are provided in Tables 2-6 and 2-7 and are used todetermine the CFs in Tables 3-3 and 3-4 for Sequoyah Units 1 and 2, respectively. Tables 3-5and 3-6 summarize the Positions 1.1 and 2.1 CF values determined for the Sequoyah Units 1 and2 RPV beltline and extended beltline materials.



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



Table 3-1 Best-Estimate Cu, Ni, P, and Mn Weight Percents, Initial RTNDT Values, and InitialUSE Values for the Sequoyah Unit 1 RPV Beltline and Extended Beltline Materials

Chemical Composition Fracture ToughnessChemial 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 1.40Weld (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 8 117 7 /2 8 8 8 72 )(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 113W04 (Heat # 2 52 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 materialswere 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 theweld 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.



(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.



Table 3-2 Best-Estimate Cu, Ni, P, and Mn Weight Percents, Initial RTNDT Values, and InitialUSE Values for the Sequoyah Unit 2 RPV Beltline and Extended Beltline Materials

Chemical Composition Fracture ToughnessProperties

RPV Material(a) InitialCu 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


Upper Shell (US) Forging 06(Heat # 9 8 12 0 1/2 8 5 84 9)(d) 0.16 0.84 0.016 0.72 5 68

Bottom Head Ring 03(Heat # 9 8 1 17 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



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 inSection 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 USEvalue for heat # 721858. This initial USE value of 78 ft-lbs is associated with the Unit 1 US to IS circ. weld.



Table 3-3 Calculation of Position 2.1 CF Values using Sequoyah Unit I SurveillanceCapsule Test Results

Capsule

RPV Material Capsule Fluenc6(a) FF(b) ARTNDT(c) FF*ARTNDT FF2

(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 0.805Metal (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# 2 529 5= X(FF * ARTNDT) + I(FF2) = (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.



Table 3-4 Calculation of Position 2.1 CF Values using Sequoyah Unit 2 SurveillanceCapsule Test Results

CapsuleFluence(a) FF(b) ARTNDT(c) FF*ARTNDT FF 2

RPV Material Capsule (x10 19 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 0.776Metal (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 27 8 = X(FF * ARTNDT) - EFF2) = (288.82) (3.663) = 78.9°FNotes:

(a)

(b)(c)

f = calculated fluence values from Table 2-7.

FF = fluence factor = 0.2 8

-0.10l og(f)).

ARTNDT (0F) 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.



Table 3-5 Summary of the Sequoyah Unit 1 RPV Beltline and ExtendedBeltline Material Chemistry Factor Values based on RegulatoryGuide 1.99, Revision 2, Position 1.1 and Position 2.1

RPV Material Chemistry Factor (IF)

Position 1.1 Position 2.1Reactor Vessel Beltline Materials

IS Forging 05 115.6 - - -

LS Forging 04 95.0 109.3IS to LS Circ. Weld W05 161.3 139.3

(Heat # 25295)Sequoyah Unit 1 Surveillance Weld 178.7

(Heat # 25295) 178.7Reactor 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 ExtendedBeltline Material Chemistry Factor Values based on RegulatoryGuide 1.99, Revision 2, Position 1.1 and Position 2.1

RPV Material Chemistry Factor (IF)

Position 1.1 Position 2.1Reactor Vessel Beltline Materials

IS Forging 05 95.0 91.1LS 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 MaterialsUS Forging 06 123.4 ---

Bottom Head Ring 03 122.3US to IS Circ. Weld W06 108.0

(Heat # 721858) 108.0LS to Bottom Head Ring Weld W04 108.0 ---

(Heat # 721858) 1 _ _ _1



4 PRESSURIZED THERMAL SHOCK

A limiting condition on RPV integrity known as Pressurized Thermal Shock (PTS) may occurduring 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: severeovercooling of the inside surface of the vessel wall followed by high repressurization; significantdegradation of vessel material toughness caused by radiation embrittlement; and the presence ofa 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) thatestablished screening criteria on PWR vessel embrittlement, as measured by the maximumreference 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 ofplant license. All domestic PWR vessels have been required to evaluate vessel embrittlement inaccordance with the criteria through the end-of-license. The U.S. NRC revised 10 CFR 50.61 in1991 and 1995 to change the procedure for calculating radiation embrittlement. These revisionsmake 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 thefollowing RTPTs values for the Sequoyah Units 1 and 2 RPV materials at 52 EFPY (EOLE). TheEOLE 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 (seeTable 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 crediblesurveillance data.

The Sequoyah Unit 2 limiting RTPTs value for forging materials at 52 EFPY is 142.3°F (seeTable 4-2), which corresponds to the Lower Shell Forging 04. The limiting RTPTs value for theUnit 2 circumferentially oriented welds at 52 EFPY is 150.7°F (see Table 4-2), whichcorresponds 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 2reactor 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 Registerby the NRC in 2010. This alternate rule is less restrictive than the Mandatory PTS Rule (10 CFR50.61) and is intended to be used for situations where the 10 CFR 50.61 criteria cannot be met.



Sequoyah Units 1 and 2 currently meet the criteria for the Mandatory PTS Rule through EOLEand therefore do not need to utilize the Alternate PTS Rule at this time.



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 (OF) (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


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 measuredvalues.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 =170F 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 Position2.1 with credible surveillance data. However, GA need not exceed 0.5*ARTNDT.



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 MaRPV 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 17.0 141.2LS 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 56.0 150.7Reactor Vessel Extended Beltline Materials

US Forging 06 123.4 0.0552 0.3087 5. 38.1 0 17.0 34.0 77.1Bottom 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 measuredvalues.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. Perthe 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.



5 UPPER-SHELF ENERGY

The decrease in Charpy upper-shelf energy (USE) is associated with the determination ofacceptable RPV toughness during the license renewal period when the vessel is exposed toadditional 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 USEof any RPV material is predicted to drop below 50 ft-lb, as measured by Charpy V-notchspecimen 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 RegulatoryGuide 1.99, Revision 2. For vessel beltline materials that are not in the surveillance program orare non-credible, the Charpy USE (Position 1.1) is assumed to decrease as a function of fluenceand 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 beused to determine the Charpy USE of the surveillance material. The surveillance data are thenused in conjunction with the Regulatory Guide to predict the change in USE (Position 2.2) of theRPV material due to irradiation.

The 52 EFPY (EOLE) Position 1.2 USE values of the vessel materials can be predicted using thecorresponding 1/4T fluence projection, the copper content of the materials, and Figure 2 inRegulatory Guide 1.99, Revision 2.

The predicted Position 2.2 USE values are determined for the reactor vessel materials that arecontained in the surveillance program by using the reduced plant surveillance data along with thecorresponding 1/4T fluence projection. The reduced plant surveillance data was obtained fromTable 5-10 of WCAP-15224 (Reference 15) and WCAP-15320 (Reference 16) for SequoyahUnits 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 surveillancecapsule fluence values documented in Tables 2-6 and 2-7 of this report for Sequoyah Units 1 and2, respectively. This data was fitted by drawing a line parallel to the existing lines as the upperbound of all the surveillance data. These reduced lines were used instead of the existing lines todetermine the Position 2.2 EOLE USE values.

The projected USE values were calculated to determine if the Sequoyah Units 1 and 2 beltlineand extended beltline materials remain above the 50 ft-lb limit at 52 EFPY (EOLE). Thesecalculations are summarized in Tables 5-1 and 5-2.



USE Conclusion

For Sequoyah Unit 1, the limiting USE value at 52 EFPY is 52.5 ft-lb (see Table 5-1); this valuecorresponds to the Bottom Head Ring 03. For Sequoyah Unit 2, the limiting USE value at 52EFPY 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 vesselsare projected to remain above the USE screening criterion value of 50 ft-lb (per 10 CFR 50Appendix G) through EOLE (52 EFPY).



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

RPV Material


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


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) 80.2Notes:

(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 RegulatoryGuide. The percent USE decrease values that corresponded to each material's specific Cu wt. % value weredetermined 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 fromTable 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 fordetermination of ARTNDT, "they may be credible for determining decrease in upper-shelf energy if the uppershelf 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 ofthe Guide) through the surveillance data points should be used in preference to the existing graph lines fordetermining the decrease in USE.



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



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

Cuba) EOLE 1/4T Initial USE EOLECu~a)UUS

RPV Material . %) Fluence(b) USE(a) Decrease(c) USE

(xlO09 n/cm2) (ft-lb) (%) (ft-lb)


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


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 fordetermination 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 ofthe Guide) through the surveillance data points should be used in preference to the existing graph lines for

determining the decrease in USE.



* Surveillance Material: IS Forging 05

A Surveillance Material: Weld Heat# 4278

100

wl ine

forging lineU.'

CI0.0.

1m

(L

10

11.00E4-17 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 forSequoyah Unit 2



6 HEATUP AND COOLDOWN PRESSURE-TEMPERATURE LIMITCURVES 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 thebeltline region of the RPV. The most limiting RTNDT of the material in the core (beltline) regionof the RPV is determined by using the unirradiated RPV material fracture toughness propertiesand estimating the irradiation-induced shift (ARTNDT).

RTNDT increases as the material is exposed to fast-neutron irradiation; therefore, to find the mostlimiting RTNDT at any time period in the reactor's life, ARTNDT due to the radiation exposureassociated with that time period must be added to the original unirradiated RTNDT. Using theadjusted reference temperature (ART) values, pressure-temperature (P-T) limit curves aredetermined 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 forthe period of extended operation (52 EFPY) do not need to be submitted as part of the SequoyahLicense Renewal Application since P-T limit curves are available through the current license (32EFPY). However, new P-T limit curves will need to be developed prior to the expiration of thecurrent curves as specified in the Sequoyah licensing basis. Therefore, only the applicability ofthe 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 forSequoyah 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 limitcurves are based on the limiting beltline material ART values, which are influenced by both thefluence and the initial material properties of that material. The Sequoyah Units 1 and 2 P-T limitcurves were developed by calculating ART values utilizing the clad/base metal interface fluencethat 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) forSequoyah Unit 2, the limiting reactor vessel material ART values with consideration of theupdated TLAA fluence values must be shown to be less than the limiting beltline material ARTvalues used in development of the existing 32 EFPY P-T limit curves contained in References 7and 10. The Regulatory Guide 1.99, Revision 2 (Reference 14) methodology was used alongwith the surface fluence of Section 2 to calculate ART values for the Sequoyah Units 1 and 2reactor vessel materials at 32 EFPY and 52 EFPY. The ART calculations are summarized inTables 6.1-1 through 6.1-4 for Sequoyah Unit 1 and in Tables 6.2-1 through 6.2-4 for SequoyahUnit 2.


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 thoseused in calculation of the existing P-T limit curves are contained in Tables 6.1-5 and 6.2-5 forSequoyah Units 1 and 2, respectively. With a re-evaluation of surveillance data credibility, arecalculation of the Position 2.1 chemistry factor values, and the consideration of TLAA fluenceprojections, the applicability of the Sequoyah Units 1 and 2 P-T limit curves may either remainunchanged or can be extended. For more detailed conclusions, refer to Sections 6.1 and 6.2below for Sequoyah Units 1 and 2, respectively.



6.1 SEQUOYAH UNIT 1

Tables 6.1-1 through 6.1-4 below summarize the 1/4T and 3/4T ART calculations for SequoyahUnit 1. The limiting 32 EFPY and 52 EFPY ART values for Sequoyah Un'it 1 corresponds to theLS Forging 04 using credible surveillance data (Position 2.1).

The applicability of the existing 32 EFPY P-T limit curves, contained in WCAP-15293, Revision2 (Reference 7) for Sequoyah Unit 1, is evaluated by comparing the updated ART valuescontained in this section with those used in the Reference 7 calculations. The existing 32 EFPYP-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 theTLAA fluence projections, the 1/4T and 3/4T ART values were recalculated in Tables 6.1-1through 6.1-4 as part of this applicability evaluation for Sequoyah Unit 1. Since the capsulefluence values were also updated as part of the TLAA effort, the Position 2.1 chemistry factorvalues were revised in Section 3 of this report. Furthermore, the credibility evaluationconclusions 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 limitcurves were developed. The comparison of limiting ART values is contained in Table 6.1-5 forSequoyah Unit 1.

Table 6.1-5 below compares the TLAA limiting ART values at 32 EFPY and 52 EFPY to thelimiting ART values used in development of the existing 32 EFPY P-T limit curves that aredocumented in WCAP-15293, Revision 2 (Reference 7). The limiting ART values used todevelop 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 ARTvalues used to develop the existing 32 EFPY P-T limit curves. This is primarily due to therevised credibility evaluation, which is performed in Appendix A of this report, and updatedfluence data of the Sequoyah Unit 1 surveillance capsules. Therefore, the existing SequoyahUnit 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 areduction of the applicability date. Since the P-T limit curves remain valid through the originalEFPY period, the current low temperature overpressure protection (LTOP) setpoints also remainapplicable through 32 EFPY.

Furthermore, based on the TLAA evaluation, Tennessee Valley Authority may instead choose toextend the applicability of the existing Sequoyah Unit 1 P-T limit curves. The new applicabilitydate with consideration of the TLAA credibility and fluence evaluations is 52 EFPY. Note thatan evaluation would have to be performed to increase the LTOP setpoints applicability period.



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) MCF M24 1/4T RTNDT(U) (c) ARTNDT l(c) A Margin ARTRPV Material (OF) (xl019 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


IS to LS Circ. Weld W05 (Heat # 25295) 161.3 1.036 1.0099 -40 162.9 0 28.0 56.0 178.9


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 = 8.50F for Position 2.1 with credible surveillance data; the weld metal GA = 28°F for Position 1.1and CA = 14'F for Position 2.1 with credible surveillance data. However, GA need not exceed 0.5*ARTNDT.




CF(a) 3/4T Fluenceb 3/4T RTNDT(U) (c) ARTNDT 6I(0 gA(d) Margin ARTRPV Material -(OF) (xlO09 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




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 0F 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.



Table 6.1-3 Calculation of the Sequoyah Unit 1 ART Values at the 1/4T Location for 52 EFPYf 1=

RPV MaterialCF(a)

(OF)

1/4T Fluence(b)

(xl019 n/cm 2,E > 1.0 MeV)


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


IS to LS Circ. Weld WO5 161.3 1.596 1.1291 -40 182.1 0 28.0 56.0 198.1(Heat # 25295)



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 56.0 108.3eat # 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 = 170F 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.




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)


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


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) ........................



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)_



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 = 00 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.



Table 6.1-5 Summary of the Sequoyah Unit 1 Limiting ART Values used in the Applicability Evaluation of the CurrentReactor Vessel Heatup and Cooldown Curves

1/4T Location 3/4T Location

Existing 32 Eitn TLAAEFPY Curves EFPY Curvesdocumented 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 04Using Non- LS Forging 04 Using Credible Using Non- LS Forging 04 Using CredibleLimiting Material Credible Surveillance Data Credible Surveillance Data

Surveillance Data Surveillance Data



6.2 SEQUOYAH UNIT 2

Tables 6.2-1 through 6.2-4 below summarize the 1/4T and 3/4T ART calculations for SequoyahUnit 2. The limiting 32 EFPY and 52 EFPY ART values for Sequoyah Unit 2 corresponds to theIS 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, Revision2 (Reference 10) for Sequoyah Unit 2, is evaluated by comparing the updated ART valuescontained in this section with those used in the Reference 10 calculations. The existing 32 EFPYP-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 theTLAA fluence projections, the 1/4T and 3/4T ART values were recalculated in Tables 6.2-1through 6.2-4 as part of this applicability evaluation for Sequoyah Unit 2. Since the capsulefluence values were also updated as part of the TLAA effort, the Position 2.1 chemistry factorvalues were revised in Section 3 of this report. Furthermore, the credibility evaluationconclusions 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 weredeveloped. The comparison of limiting ART values is contained in Table 6.2-5 for SequoyahUnit 2.

Table 6.2-5 below compares the TLAA limiting ART values at 32 EFPY and 52 EFPY to thelimiting ART values used in development of the existing 32 EFPY P-T limit curves that aredocumented in WCAP-15321, Revision 2 (Reference 10). The limiting ART values used todevelop 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 todevelop the existing 32 EFPY P-T limit curves. Therefore, the existing Sequoyah Unit 2 P-Tlimit curves remain valid through 32 EFPY.

Furthermore, the TLAA limiting 1/4T ART value at 52 EFPY is bounded by the limiting ARTvalue used to develop the existing 32 EFPY P-T limit curves; however, the TLAA limiting 3/4TART value at 52 EFPY is not bounded by the limiting ART value used to develop the existing 32EFPY 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 updatedTLAA fluence evaluation as well as the updated credibility analysis of the Sequoyah Unit 2surveillance materials.

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



Surveillance data is available for the IS to LS Circumferential Weld W05. Therefore, thePosition 2.1 chemistry factor, initial RTNDT, and margin terms from Table 6.2-4 were used todetermine the 3/4T fluence value when the 3/4T ART equals 115'F (Table 6.2-5) for thismaterial. This fluence value is approximately 2.22 x 1019 n/cm2 (E > 1.0 MeV), which was usedto 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 WeldW05 (Position 2.1) is 44.7 EFPY. Therefore, the existing Sequoyah Unit 2 P-T limit curves maybe 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 areduction of the applicability date. Since the P-T limit curves remain valid through the originalEFPY period, the current LTOP setpoints also remain applicable through 32 EFPY.

Furthermore, based on the TLAA evaluation, Tennessee Valley Authority may instead choose toextend the applicability of the existing Sequoyah Unit 2 P-T limit curves. The new applicabilitydate with consideration of the TLAA credibility and fluence evaluations is 44.7 EFPY. Note thatan evaluation would have to be performed to increase the LTOP setpoints applicability period.




CF(a) 1/4T Fluence(b) 1/4T RTNDT(U) (c) ARTNDT 61t(C) FA(d) Margin ARTRPV 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


LS Forging 04 104.0 1.000 0.9999 -22 104.0 0 17.0 34.0 116.0


Using non-credible surveillance data 78.9 0.994 0.9983 -4 78.8 0 28.0 56.0 130.8


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 = 170F for Position 1.1 and A = 8.50F for Position 2.1 with crediblesurveillance 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 exceed0.5*ARTNDT.

WCAP-1 7539-NP March 2012




CF(a) 3/T Fluence(b) 3/4T RTNDT(U) (c) ARTNDT Fgl(c) GA(d) Margin ARTRPV 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


LS Forging 04 104.0 0.363 0.7199 -22 74.9 0 17.0 34.0 86.9




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, = 00 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 crediblesurveillance 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 exceed0.5*ARTNDT.




CF(a) 1/4T Fluence(b) 1/4T RTNDT(U)(c) ARTNDT 01(c) (A(d) Margin ARTRPV Material (OF) (X__019 n/cm2' FF(b) (OF) (OF) (OF) (OF) (OF) (OF)

E > 1.0 MeV)


IS Forging 05 95.0 1.548 1.1208 10 106.5 0 17.0 34.0 150.5


LS Forging 04 104.0 1.548 1.1208 -22 116.6 0 17.0 34.0 128.6




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)


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 a1 = 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.




Reactor Vessel MaterialCF(a)(01F)

3/4T Fluence (

(xl01 9 n/cm 2,

E > 1.0 MeV)


IS Forging 05 95.0 0.562 0.8386 10 79.7 0 17.0 34.0 1,23.7


LS Forging 04 104.0 0.562 0.8386 -22 87.2 0 17.0 34.0 99.2




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


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.



Table 6.2-5 Summary of the Sequoyah Unit 2 Limiting ART Values used in the Applicability Evaluation of the CurrentReactor Vessel Heatup and Cooldown Curves

1/4T Location 3/4T Location

Existing 32 Existing 32EFPY Curves EFPY Curves

documented Evaluation at Evaluation at Evaluation at Evaluation at32 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 UsingSimtig atehutence D Non-Credible Surveillance Data S utvellnc Non-Credible Surveillance DataSurveillance Data Surveillance Data



7 SURVEILLANCE CAPSULE WITHDRAWAL SCHEDULES

Contained in Tables 7-1 and 7-2 are the Sequoyah Units 1 and 2 recommended surveillancecapsule withdrawal schedules, respectively. These schedules meet the recommendations ofASTM E185-82 (Reference 21) as required by 10 CFR 50, Appendix H (Reference 22). Withthe withdrawal of Capsule Y, Sequoyah Units I and 2 fulfilled the surveillance capsulewithdrawal 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 recommendedthat several remaining capsules be relocated to higher lead factor locations for each Unit. One ofthese relocated capsules in each Unit should be subsequently withdrawn from the reactor vesseland tested at the time when the accumulated neutron fluence of the capsule corresponds to notless 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 (xl019 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 Zshould be withdrawn so that the capsule fluence corresponds to at least one times the 60-year EOL vesselfluence (2.66 x 10'9 n/cm-), but less than two times the 60-year EOL vessel fluence:(5.32 xj j010' n/cm2 ).However, none of these remaining capsules are predicted to experience a neutron fluence of 2?.66 x" 1019

c•rn ffi2 to EO LEini'their current locations; therefore, it is recommended to relocate several of theseremaining 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 timewhen the capsule experiences the peak EOLE vessel fluence value (2.66 x 1019 n/cm 2) is approximately32.5, 33.4, or 34.4 EFPY, respectively. See Appendix B for further details on capsule relocationrecommendations.



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 Zshould be withdrawn so that the capsule fluence corresponds to at least one times the 60-year EOL vesselfluence, (.57 x'10 9)n0i•n )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 theseremaining 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 timewhen the capsule experiences the peak EOLE vessel fluence value (2.57 x 1019 n/cm2) is approximately32.6, 33.7, or 34.7 EFPY, respectively. See Appendix B for further details on capsule relocationrecommendations.



8 REFERENCES

1. Code of Federal Regulations, 10 CFR Part 54.3, "Definitions," U.S. Nuclear RegulatoryCommission, Washington, D.C., Federal Register, Volume 72, dated August 28, 2007.

2. Regulatory Guide 1.190, "Calculational and Dosimetry Methods for DeterminingPressure Vessel Neutron Fluence," U:S. Nuclear Regulatory Commission, Office ofNuclear Regulatory Research, March 2001.

3. WCAP-14040-A, Revision 4, "Methodology Used to Develop Cold OverpressureMitigating 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 LeastSquares 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 BasedCorrelation 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 forNormal Operation and PTLR Support Documentation," J. H. Ledger, July 2003.

8. WCAP-8233, Revision 0, "Tennessee Valley Authority Sequoyah Unit No. 1 ReactorVessel Radiation Surveillance Program," S. E. Yanichko et al., December 1973.

9. Code of Federal Regulations, 10 CFR Part 50.61, "Fracture Toughness Requirements forProtection Against Pressurized Thermal Shock Events," U.S. Nuclear RegulatoryCommission, Washington, D.C., Federal Register, Volume 60, No. 243, dated December19, 1995, effective January 18, 1996.

10. WCAP-15321, Revision 2, "Sequoyah Unit 2 Heatup and Cooldown Limit Curves forNormal Operation and PTLR Support Documentation," J. H. Ledger, July 2003.

11. WCAP-8513, Revision 0, "Tennessee Valley Authority Sequoyah Unit No. 2 ReactorVessel 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 AnalysisReports for Nuclear Power Plants: LWR Edition, NUREG-0800, March 2007.

13. WCAP-10340, Revision 1, "Analysis of Capsule T From the Tennessee Valley AuthoritySequoyah Unit 1 Reactor Vessel Radiation Surveillance Program," S. E. Yanichko et al.,February 1984.



14. Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor VesselMaterials," U.S. Nuclear Regulatory Commission, Office of Nuclear RegulatoryResearch, May 1988.

15. WCAP- 15224, Revision 0, "Analysis of Capsule Y from the Tennessee Valley AuthoritySequoyah Unit 1 Reactor Vessel Radiation Surveillance Program," T. J. Laubham, June1999.

16. WCAP-15320, Revision 0, "Analysis of Capsule Y from the Tennessee Valley AuthoritySequoyah Unit 2 Reactor Vessel Radiation Surveillance Program," T. J. Laubham,December 1999.

17. Code of Federal Regulations, 10 CFR 50.61a, "Alternate Fracture ToughnessRequirements for Protection Against Pressurized Thermal Shock Events," U.S. NuclearRegulatory Commission, Washington, D.C., Federal Register, Volume 75, No. 1, datedJanuary 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 ToughnessRequirements," U.S. Nuclear Regulatory Commission, Washington, D.C., FederalRegister, Volume 60, No. 243, dated December 19, 1995.

19. Appendix G to the 1998 through the 2000 Addenda Edition of the ASME Boiler andPressure Vessel (B&PV) Code, Section XI, Division 1, "Fracture Toughness Criteria forProtection Against Failure.

20. NUREG- 1800, Revision 2, "Standard Review Plan for Review of License RenewalApplications for Nuclear Power Plants," U.S. Nuclear Regulatory Commission,December 2010.

21. ASTM E185-82, "Standard Practice for Conducting Surveillance Tests for Light-WaterCooled Nuclear Power Reactor Vessels," American Society for Testing and Materials,1982.

22. Code of Federal Regulations, 10 CFR 50, Appendix H, "Reactor Vessel MaterialSurveillance Program Requirements," U.S. Nuclear Regulatory Commission,Washington, D.C., Federal Register, Volume 60, No. 243, dated December 19, 1995.


Westinghouse Non-Proprietary Class 3 A-1

APPENDIX A CREDIBILITY EVALUATION OF THE SEQUOYAHUNITS 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 tothe NRC staff for calculating the effects of neutron radiation embrittlement of the low-alloysteels currently used for light-water-cooled reactor vessels. Position C.2 of Regulatory Guide1.99, Revision 2, describes the method for calculating the adjusted reference temperature andCharpy 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 datasets become available from the reactor in question.

To date there have been four surveillance capsules removed and tested from the Sequoyah Unit 1reactor vessel. To use these surveillance data sets, they must be shown to be credible. Inaccordance with Regulatory Guide 1.99, Revision 2, the credibility of the surveillance data willbe 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 thatsurveillance data is credible.

EVALUATION

Criterion 1: Materials in the capsules should be those judged most likely to be controlling withregard to radiation embrittlement.

The beltline region of the reactor vessel is defined in Appendix G to 10 CFR Part 50, "FractureToughness Requirements" (Reference A. 1-2), as follows:

"the region of the reactor vessel (shell material including welds, heat affected zones, andplates or forgings) that directly surrounds the effective height of the active core andadjacent regions of the reactor vessel that are predicted to experience sufficient neutronradiation damage to be considered in the selection of the most limiting material withregard 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 SeamW05 (fabricated with SMIT 40 weld wire type, heat # 25295 and SMIT 89 fluxtype, lot # 2275)


A-2 Westinghouse Non-Proprietary Class 3

The Sequoyah Unit 1 surveillance program utilizes tangential and axial test specimens fromLower Shell Forging 04. The surveillance weld metal was fabricated with SMIT 40 weld wiretype, heat # 25295 and SMIT 89 flux type, lot # 1103.

Per WCAP-8233 (Reference A.1-3), the Sequoyah Unit 1 surveillance program was based onASTM E185-73 (Reference. A.1-4). Per Section 4.1 of ASTM E185-73, "The base metal andweld metal to be included in the program should represent the material that may limit theoperation of the reactor during its lifetime. The test material should be selected on the basis ofinitial transition temperature, upper shelf energy level, and estimated increase in transitiontemperature considering chemical composition (copper (Cu) and phosphorus (P)) and neutronfluence.

At the time when the surveillance program was developed, it was believed that copper andphosphorus were the elements most important to embrittlement of reactor vessel steels. LowerShell Forging 04 had the highest initial RTNDT and lowest initial upper-shelf energy out of thetwo beltline forgings in the Sequoyah Unit 1 reactor vessel. In addition, Lower Shell Forging 04had 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 materialas the reactor vessel beltline circumferential weld. In accordance with the definition of thereactor 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 andunirradiated conditions should be small enough to permit the determination of the30 ft-lb temperature and upper-shelf energy unambiguously.

Plots of Charpy energy versus temperature for the unirradiated and irradiated conditions for all ofthe Sequoyah Unit 1 surveillance materials are presented in Section 5 and Appendix C of thelatest 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 topermit the determination of the 30 ft-lb temperature and the upper-shelf energy of the SequoyahUnit 1 surveillance materials unambiguously.

Hence, Criterion 2 is met for the Sequovah Unit 1 surveillance program.


Westinghouse Non-Proprietary Class 3 A-3Westinghouse Non-Proprietary Class 3 A-3

Criterion 3: When there are two or more sets of surveillance data from one reactor, the scatterof ARTNDT values about a best-fit line drawn as described in Regulatory Position2.1 normally should be less than 28°F for welds and 17'F for base metal. Even ifthe fluence range is large (two or more orders of magnitude), the scatter shouldnot exceed twice those values. Even if the data fail this criterion for use in shiftcalculations, they may be credible for determining decrease in upper-shelf energyif the upper shelf can be clearly determined, following the definition given inASTM E185-82 (Reference A. 1-6).

The functional form of the least squares method as described in Regulatory Position 2.1 will beutilized 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 forcredibility. The weld is made from weld wire heat # 25295. This weld metal is not in any othersurveillance program.

Following is the calculation of the best-fit line as described in Regulatory Position 2.1 ofRegulatory Guide 1.99, Revision 2. In addition, the recommended NRC methods fordetermining credibility will be followed. The NRC methods were presented to the industry at ameeting held by the NRC on February 12 and 13, 1998 (Reference A.1-7). At this meeting theNRC presented five cases. Of the five cases, Case 1 ("Surveillance data available from plant butno other source") most closely represents the situation listed above for the Sequoyah Unit 1surveillance forging and weld materials.



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 inTable A.1-1.

Table A.1-1 Calculation of Interim Chemistry Factors for the Credibility EvaluationUsing Sequoyah Unit 1 Surveillance Capsule Data Only

CapsuleFluence(a) FF b) ARTNDT(c) FF*ARTNDT FF2

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 0.805Metal

(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 # 25 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 includethe adjustment ratio procedure of Regulatory Guide 1.99, Revision 2, Position 2.1 since thiscalculation 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.



The scatter of ARTNDT values about the functional form of a best-fit line drawn as described inRegulatory 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 0FCF 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 NoSurveillance U 154.8 0.693 0.897 144.92 138.9 6.1 YesWeld 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 Guide1.99, Revision 2, Position 2.1, should be less than 17'F for base metal. Table A. 1-2 indicatesthat six of the eight surveillance data points fall within the +/- la of 17'F scatter band forsurveillance 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 Guide1.99, Revision 2, Position 2.1, should be less than 28°F for weld metal. Table A. 1-2 indicatesthat three of the four surveillance data points fall within the +/- lC of 28°F scatter band forsurveillance weld materials; therefore, the weld material is deemed "credible" per the thirdcriterion.


A-6 Westinghouse Non-Proprietary Class 3A- esigoueNn-rpieayCls-

Criterion 4: The irradiation temperature of the Charpy specimens in the capsule should matchthe 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 shieldand the vessel wall and are positioned opposite the center of the core. The test capsules are inbaskets attached to the thermal shield. The location of the specimens with respect to the reactorvessel beltline provides assurance that the reactor vessel wall and the specimens experienceequivalent 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 shouldfall 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 surveillanceforging 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 VesselMaterials," 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. 1Reactor Vessel Radiation Surveillance Program," S. E. Yanichko et al., December1973.

A. 1-4 ASTM E185-73, "Standard Recommended Practice for Surveillance Tests forNuclear Reactor Vessels," American Society for Testing and Materials, 1973.

A.1-5 WCAP-15224, Revision 0, "Analysis of Capsule Y from the Tennessee ValleyAuthority Sequoyah Unit 1 Reactor Vessel Radiation Surveillance Program," T. J.Laubham et al., June 1999.



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 andMaterials, 1982.

A. 1-7 K. Wichman, M. Mitchell, and A. Hiser, USNRC, Generic Letter 92-01 and RPVIntegrity Workshop Handouts, NRC/Industry Workshop on RPV Integrity Issues,February 12, 1998.



A.2 SEQUOYAH UNIT 2

INTRODUCTION

Regulatory Guide 1.99, Revision 2 (Reference A.2-1) describes general procedures acceptable tothe NRC staff for calculating the effects of neutron radiation embrittlement of the low-alloysteels currently used for light-water-cooled reactor vessels. Position C.2 of Regulatory Guide1.99, Revision 2, describes the method for calculating the adjusted reference temperature andCharpy 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 datasets become available from the reactor in question.

To date there have been four surveillance capsules removed and tested from the Sequoyah Unit 2reactor vessel. To use these surveillance data sets, they must be shown to be credible. Inaccordance with Regulatory Guide 1.99, Revision 2, the credibility of the surveillance data willbe 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 thatsurveillance data is credible.

EVALUATION

Criterion 1: Materials in the capsules should be those judged most likely to be controlling withregard to radiation embrittlement.

The beltline region of the reactor vessel is defined in Appendix G to 10 CFR Part 50, "FractureToughness Requirements" (Reference A.2-2), as follows:

"the region of the reactor vessel (shell material including welds, heat affected zones, andplates or forgings) that directly surrounds the effective height of the active core andadjacent regions of the reactor vessel that are predicted to experience sufficient neutronradiation damage to be considered in the selection of the most limiting material withregard 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 SeamW05 (fabricated with Arcos weld wire type, heat # 4278 and SMIT 89 flux type,lot # 1211)



The Sequoyah Unit 2 surveillance program utilizes tangential and axial test specimens fromIntermediate Shell Forging 05. The surveillance weld metal was fabricated with Arcos weld wiretype, heat # 4278 and SMIT 89 flux type, lot # 1211.

Per WCAP-8513 (Reference A.2-3), the Sequoyah Unit 2 surveillance program was based onASTM E185-73 (Reference. A.2-4). Per Section 4.1 of ASTM E185-73, "The base metal andweld metal to be included in the program should represent the material that may limit theoperation of the reactor during its lifetime. The test material should be selected on the basis ofinitial transition temperature, upper shelf energy level, and estimated increase in transitiontemperature considering chemical, composition (copper (Cu) and phosphorus (P)) and neutronfluence."

At the time when the surveillance program was developed, it was believed that copper andphosphorus 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 energyout of the two beltline forgings in the Sequoyah Unit 2 reactor vessel. In addition, IntermediateShell Forging 05 had approximately the same copper and phosphorus content of the otherbeltline 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 materialas the reactor vessel beltline circumferential weld. In accordance with the definition of thereactor 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 andunirradiated conditions should be small enough to permit the determination of the30 ft-lb temperature and upper-shelf energy unambiguously.

Plots of Charpy energy versus temperature for the unirradiated and irradiated conditions for all ofthe Sequoyah Unit 2 surveillance materials are presented in Section 5 and Appendix C of thelatest 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 topermit the determination of the 30 ft-lb temperature and the upper-shelf energy of the SequoyahUnit 2 surveillance materials unambiguously.




Criterion 3: When there are two or more sets of surveillance data from one reactor, the scatterof ARTNDT values about a best-fit line drawn as described in Regulatory Position2.1 normally should be less than 287F for welds and 17'F for base metal. Even ifthe fluence range is large (two or more orders of magnitude), the scatter shouldnot exceed twice those values. Even if the data fail this criterion for use in shiftcalculations, they may be credible for determining decrease in upper-shelf energyif the upper shelf can be clearly determined, following the definition given inASTM E185-82 (Reference A.2-6).

The functional form of the least squares method as described in Regulatory Position 2.1 will beutilized 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 forcredibility. The weld is made from weld wire heat # 4278. This weld metal is not in any othersurveillance program.

Following is the calculation of the best-fit line as described in Regulatory Position 2.1 ofRegulatory Guide 1.99, Revision 2. In addition, the recommended NRC methods fordetermining credibility will be followed. The NRC methods were presented to the industry at ameeting held by the NRC on February 12 and 13, 1998 (Reference A.2-7). At this meeting theNRC presented five cases. Of the five cases, Case 1 ("Surveillance data available from plant butno other source") most closely represents the situation listed above for the Sequoyah Unit 2surveillance forging and weld materials.



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 inTable A.2-1.

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

CapsuleFluence(a) FF(b) ARTNDT(c) FF*ARTNDT FFr2

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 0.776Metal

(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 # 4278= X(FF * ARTNDT) + I(FF2) = (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 includethe adjustment ratio procedure of Regulatory Guide 1.99, Revision 2, Position 2.1 since thiscalculation is based on the actual surveillance weld metal measured shift values. In addition, onlySequoyah Unit 1 data is being considered; therefore, no temperature adjustment is required.



The scatter of ARTNDT values about the functional form of a best-fit line drawn as described inRegulatory 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 FCF Fluence

RPV Material Capsule (OF) (x1019 n/cm 2, FF ARTNDT ARTNDT ARTNDT (Base Metal)

E > 1.0 MeV) (OF) (OF) (OF) <280F (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 NoWeld 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 values1.99, Revision 2, Position 2.1,

about the best-fit line, drawn as described in Regulatory Guideshould 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 forsurveillance 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 Guide1.99, Revision 2, Position 2.1, should be less than 28°F for weld metal. Table A.2-2 indicatesthat two of the four surveillance data points fall within the +/- la of 28°F scatter band forsurveillance weld materials; therefore, the weld material is deemed "non-credible" per the thirdcriterion.



Criterion 4: The irradiation temperature of the Charpy specimens in the capsule should matchthe 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 shieldand the vessel wall and are positioned opposite the center of the core. The test capsules are inbaskets attached to the thermal shield. The location of the specimens with respect to the reactorvessel beltline provides assurance that the reactor vessel wall and the specimens experienceequivalent operating conditions such that the temperatures will not differ by more than 25°F.


Criterion 5: The surveillance data for the correlation monitor material in the capsule shouldfall 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 surveillanceforging 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 VesselMaterials," 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. 2Reactor Vessel Radiation Surveillance Program," J. A. Davidson et al., November1975.

A.2-4 ASTM E185-73, "Standard Recommended Practice for Surveillance Tests forNuclear Reactor Vessels," American Society for Testing and Materials, 1973.

A.2-5 WCAP-15320, Revision 0, "Analysis of Capsule Y from the Tennessee ValleyAuthority Sequoyah Unit 2 Reactor Vessel Radiation Surveillance Program," T. J.Laubham et al., December 1999.

WCAP- 17539-NP March 20 12Revision 0


A.2-6 ASTM E185-82, "Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels," American Society for Testing andMaterials, 1982.

A.2-7 K. Wichman, M. Mitchell, and A. Hiser, USNRC, Generic Letter 92-01 and RPVIntegrity Workshop Handouts, NRC/Industry Workshop on RPV Integrity Issues,February 12, 1998.

•WCAP- 17539-NP March 2012Revision 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 vesseland tested, as recommended by ASTM E185-82 (Reference B.1-1). With the withdrawal ofCapsule Y, Sequoyah Unit 1 fulfilled the surveillance capsule withdrawal recommendationscontained in ASTM E185-82, as required by 10 CFR 50, Appendix H (Reference B.1-2), for a40=year EOL (32 EFPY). Since Sequoyah Unit 1 is applying for a 20-year license extension, anadditional capsule is expected to provide metallurgical data corresponding with an EOL fluenceof 60 years (52 EFPY). Currently, there are four remaining capsules (W, V, S, and Z) in theSequoyah Unit 1 reactor vessel.

Capsules T, U, X and Y in the Sequoyah Unit 1 reactor vessel were positioned at the 400azimuthal 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 areconsidered to be radiologically equivalent. Note that the 40 azimuthal location is a lag (less thanone) factor location; therefore, at this time, the Sequoyah Unit 1 reactor vessel is being irradiatedslightly faster than the remaining capsules. In order for Sequoyah Unit 1 to have meaningfulmetallurgical capsule data in the future, it is recommended that several of the remaining capsulesbe 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 40and 400 azimuthal capsule locations.

Table B.1-1 Projected Neutron Fluence Values at the Geometric Centerof the Surveillance Capsule Locations for Sequoyah Unit 1

Capsule FluenceCycle EFPY (x10 1 9 n/cm 2 , E > 1.0 MeV)

40 Azimuthal 40' Azimuthal

Location Location18 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


B-2 Westinghouse Non-Proprietary Class 3

The fluence values listed in Table B.1-1 are used to determine neutron fluence projectionsassuming 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 theremaining Sequoyah Unit 1 40 capsules assuming they are relocated to any of the 400. locations atvarious relocation times.

Table B.1-2 SequoyahIUnit 1IProjected Capsule Neutron Fluence Values Associated withCapsule Relocation from the 40 to the 400 Azimuthal Location

Capsule Fluence

Cycle EFPY (xW0 19 n/cm 2, E > 1.0 MeV)

Relocation at the Relocation at the Relocation at theEOC 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 isexpected to satisfy the same criteria as the EOL capsule, as described in ASTM E185-82, withthe EOL fluence at 60 years (52 EFPY). Therefore, a capsule should be withdrawn so that thecapsule fluence corresponds to at least one times the 60-year EOL vessel fluence (2.66 x .1019n/cm2, per Table 2-5), but less than two times the 60-year EOL vessel fluence (5.32 x 1019n/cm2). Based on the fluence projections in Table B.1-1, none of the remaining Sequoyah Unit 1capsules, in their current azimuthal locations (40), would experience a neutron fluence of 2.66 x1019 n/cm 2 prior to EOLE.

However, based on the fluence projections in-Table B. 1-2, the peak 52,EPY calculated vessefluence 4,2.66 x 10.14, n/CM2F•>(,e•.0 V), would occur at approximately 32.5, 33.4, or 34.4EFPY, 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 peak52 EFPY calculated vessel fluence of 5.32 x 10 n/cm 2 (E > 1.0 MeV) would occur atapproximately 52 EFPY for a relocated capsule, assuming the capsule was relocated at the EOCs18; 19, oIr620.'-.0-to a 400 azimuthal location.



Additionally, it is anticipated at this time that if an additional 20-year license extension wassought, another capsule would be needed to be withdrawn from the reactor vessel in order tosatisfy the same criteria as the EOL capsule with an EOL fluence at 80 years (72 EFPY). Theextrapolated maximum neutron fluence value at 72 EFPY for Sequoyah Unit 1 is approximately3.61 x 1019 n/cm2 (E > 1.0 MeV). Based on the fluence projections in Table B.1-2, the peak 72EFPY calculated vessel fluence of 3.61 x 1019 n/cm2 (E > 1.0 MeV) would occur atapproximately 39.0, 39.9, or 40.9 EFPY, assuming the capsule was relocated to a 40' azimuthallocation at the EOC 18, 19, or 20, respectively.

In summary, it is recommended that several of the Sequoyah Unit 1 remaining capsules berelocated to higher lead factor locations. One of these relocated capsules should be subsequentlywithdrawn from the reactor vessel and tested at the time when the accumulated neutron fluenceof the capsule corresponds to not less than once or greater than twice the peak 60-year vesselfluence. Another relocated capsule could be used for future testing, if additional licenseextensions are sought. Table B. 1-3 summarizes potential removal times for the relocatedcapsules based on license extension out to 60 and 80 years of operation. These dates are basedon the capsule fluence being equivalent to one times the peak vessel fluence at 60 years (2.66 x1019 n/cm 2) as well as one times the peak vessel fluence at 80 years (3.61 x 1019 n/cm2).

Table B.1-3 Sequoyah Unit 1 Potential Capsule Withdrawal Times Associated with CapsuleRelocation 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 39.9EOC 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 1019 n/cm2) 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 andMaterials, 1982.

B. 1-2 Code of Federal Regulations, 10 CFR 50, Appendix H, "Reactor Vessel MaterialSurveillance Program Requirements," U.S. Nuclear Regulatory Commission,Washington, D.C., Federal Register, Volume 60, No. 243, dated December 19,1995.



B.2 SEQUOYAH UNIT 2

Four capsules (T, U, X and Y) have been withdrawn from the Sequoyah Unit 2 reactor vesseland tested, as recommended by ASTM E185-82 (Reference B.2-1). With the withdrawal ofCapsule Y, Sequoyah Unit 2 fulfilled the surveillance capsule withdrawal recommendationscontained in ASTM E185-82, as required by 10 CFR 50, Appendix H (Reference B.2-2), for a40-year EOL (32 EFPY). Since Sequoyah Unit 2 is applying for a 20-year license extension, anadditional capsule is expected to provide metallurgical data corresponding with an EOL fluenceof 60 years (52 EFPY). Currently, there are four remaining capsules (W, V, S, and Z) in theSequoyah Unit 2 reactor vessel.

Capsules T, U, X and Y in the Sequoyah Unit 2 reactor vessel were positioned at the 400azimuthal 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 areconsidered to be radiologically equivalent. Note that the 4' azimuthal location is a lag (less thanone) factor location; therefore, at this time, the Sequoyah Unit 2 reactor vessel is being irradiatedslightly faster than the remaining capsules. In order for Sequoyah Unit 2 to have meaningfulmetallurgical capsule data in the future, it is recommended that several of the remaining capsulesbe 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 40and 400 azimuthal capsule locations.

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

Capsule FluenceCycle EFPY (x10 19 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



The fluence values listed in Table B.2-1 are used to determine neutron fluence projectionsassuming 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 theremaining Sequoyah Unit 2 40 capsules assuming they are relocated to any of the 400 locations atvarious relocation times.

Table B.2-2 Sequoyah Unit 2 Projected Capsule Neutron Fluence Values Associated withCapsule Relocation from the 4' to the 400 Azimuthal Location

Capsule Fluence

Cycle EFPY (xl0 19 n/cm 2, E > 1.0 MeV)

Relocation at the Relocation at the Relocation at theEOC 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 isexpected to satisfy the same criteria as the EOL capsule, as described in ASTM E185-82, withthe EOL fluence at 60 years (52 EFPY). Therefore, a capsule should be withdrawn so that thecapsule fluence corresponds to at least one times the 60-year EOL vessel fluence (2.57 x 1019n/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 2capsules, in their current azimuthal locations (40), would experience a neutron fluence of 2.57 x1019 n/cm 2 prior to EOLE.

However, based on the fluence projections in Table •B.2-2, the peak 52 EFPY calculated vesselfluence ofl2.57 x10 - n/cm E> 10 MeV) would occur at approximately 32.6, 33.7, or 34.7EFPY, assuming the capsule was relocated to a 400 azimuthal location at the EOCs 18,•:19; Lor 20;respectively. Furthermore, based on the fluence projections in Table B.2-2, two times the peak52 EFPY calculated vessel fluence of 5.14 x 10 n/cm 2 (E > 1.0 MeV) would occur atapproximately 52 EFPY for a relocated capsule, assuming the capsule was relocated at the EOCs18, 19, or 20 to a 400 azimuthal location.



Additionally, it is anticipated at this time that if an additional 20-year license extension wassought, another capsule would be needed to be withdrawn from the reactor vessel in order tosatisfy the same criteria as the EOL capsule with an EOL fluence at 80 years (72 EFPY). Theextrapolated maximum neutron fluence value at 72 EFPY for Sequoyah Unit 2 is approximately3.52 x 1019 n/cm 2 (E > 1.0 MeV). Based on the fluence projections in Table B.2-2, the peak 72EFPY calculated vessel fluence of 3.52 x 1019 n/cm2 (E > 1.0 MeV) would occur atapproximately 39.3, 40.4 or 41.4 EFPY, assuming the capsule was relocated to a 400 azimuthallocation at the EOC 18, 19, or 20, respectively.

In summary, it is recommended that several of the Sequoyah Unit 2 remaining capsules berelocated to higher lead factor locations. One of these relocated capsules should be subsequentlywithdrawn from the reactor vessel and tested at the time when the accumulated neutron fluenceof the capsule corresponds to not less than once or greater than twice the peak 60-year vesselfluence. Another relocated capsule could be used for future testing, if additional licenseextensions are sought. Table B.2-3 summarizes potential removal times for the relocatedcapsules based on license extension out to 60 and 80 years of operation. These dates are basedon the capsule fluence being equivalent to one times the peak vessel fluence at 60 years (2.57 x1019 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 CapsuleRelocation 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 41.4Notes:

(a) These dates are based on the capsule fluence being equivalent to one times the peak vessel fluence at 60years (2.57 x 1019 n/cm2) as well as one times the peak vessel fluence at 80 years (3.52 x 1019 n/cm2).

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 andMaterials, 1982.

B.2-2 Code of Federal Regulations, 10 CFR 50, Appendix H, "Reactor Vessel MaterialSurveillance Program Requirements," U.S. Nuclear Regulatory Commission,Washington, D.C., Federal Register, Volume 60, No. 243, dated December 19,1995.


Westinghouse Non-Proprietary Class 3 C-1

APPENDIX CEMERGENCY RESPONSE GUIDELINE LIMITS

The Emergency Response Guideline (ERG) limits were developed to establish guidance foroperator 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 insidesurface RTNDT. These generic categories were conservatively generated for the WestinghouseOwners Group (WOG) to be applicable to all Westinghouse plants.

The highest value of RTNDT for which the generic category ERG limits were developed is 2507Ffor a longitudinal flaw and 300'F for a circumferential flaw. Therefore, if the limiting vesselmaterial has an RTNDT that exceeds 2507F for a longitudinal flaw or 3007F for a circumferentialflaw, plant-specific ERG P-T limits must be developed.

The ERG category is determined by the magnitude of the limiting RTNDT value, which iscalculated the same way as the RTprs values are calculated in Section 4 of this report. Thematerial with the highest RTNDT defines the limiting material, which for Sequoyah Unit 1 is LSForging 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 forSequoyah 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°Fwith Non-Credible Surveillance Data

Notes:

(a) Longitudinally oriented flaws are applicable only up to 250'F; circumferentially oriented flaws areapplicable up to 300'F.

(b) Values taken from Tables 4-1 and 4-2 for Sequoyah Units 1 and 2, respectively.


C-2 Westinghouse Non-Proprietary Class 3

Per the ERG limit guidance document (Reference C-i), some vessels do not change categoriesfor operation through the end-of-license. However, when a vessel does change ERG categoriesbetween the beginning and end of operation, a plant-specific assessment must be performed todetermine at what operating time the category changes. Thus, the ERG classification need not bechanged until the operating cycle during which the maximum vessel value of actual or estimatedreal-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. Thisvalue corresponds to the Lower Shell Forging 04. Thus, the limiting material RTNDT valueexceeds the ERG Category I criterion (RTNDT < 2007F) prior to 52 EFPY. The transition occurswhen RTNDT = 200'F. The operating cycle at which the ERG category transitioned fromCategory I to Category II was determined to be prior to Cycle 15. Sequoyah Unit 1 will remainin ERG Category Unit II through EOLE.

Unit 2

Per Table C-i, the limiting material for Sequoyah Unit 2 (Intermediate Shell to Lower ShellCircumferential Weld) has an RTNDT less than 200'F through 52 EFPY. Therefore, SequoyahUnit 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 ERGCategory Unit II through EOLE (52 EFPY). Sequoyah Unit 2 is currently in ERG Category Iand remains in ERG Category I through EOLE (52 EFPY).

Appendix C Reference

C-1 "Background Information for Westinghouse Owners Group Emergency ResponseGuidelines, Critical Safety Function Status Tree, F-0.4 Integrity, HP/LP-Rev. 2,"Westinghouse Owners Group, April 30, 2005.


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