ML18215A178

WCAP-18169-NP, Rev 1, Arkansas Nuclear One Unit 2 Heatup and Cooldown Limit Curves for Normal Operation.

ML18215A178
Person / Time
Site:	Arkansas Nuclear
Issue date:	06/30/2018
From:	Mays B Westinghouse, Westinghouse
To:	Office of Nuclear Reactor Regulation
References
CALC-ANO2-EP-17-00002, Rev 0 WCAP-18169-NP, Rev 1
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CALC-AN02-EP-17-00002 Rev O Page 3 of 88 Westinghouse Non-Proprietary Class 3 WCAP-18169-NP June 2018 Revision 1 Arkansas Nuclear One Unit 2 Heatup and Cooldown Limit Curves for Normal Operation

@ Westinghouse

CALC-AN02-EP-17-00002 Rev O Page 4 of 88 WESTINGHOUSE NON-PROPRIETARY CLASS 3 WCAP-18169-NP Revision 1 Arkansas Nuclear One Unit 2 Heatup and Cooldown Limit Curves for Normal Operation Benjamin E. Mays*

Structural Design & Analysis III June 2018 Reviewers: D. Brett Lynch* Approved: Lynn A. Patterson*, Manager Structural Design & Analysis III Structural Design & Analysis III Eugene T. Hayes* Laurent P. Houssay*, Manager Radiation Engineering & Analysis Radiation Engineering & Analysis

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

Westinghouse Electric Company LLC 1000 Westinghouse Dr.

Cranberry Township, PA 16066

© 2018 Westinghouse Electric Company LLC All Rights Reserved

CALC-AN02-EP-17-00002 Rev O Page 5 of 88 Westinghouse Non-Proprietary Class 3 II RECORD OF REVISION Revision 0: Original Issue Revision 1: Revised issue. The purpose of this rev1s1on is to remove utilization of the TLR-RES/DE/CIB-2013-01 report methodology. Therefore, calculated LlRTNDT values less than or equal to 25°F will not be reduced to zero. The pressure-temperature (P-T) limit curves are not affected by the changes. Changes are indicated with change bars.

WCAP-18 169-NP June 2018 Revision 1

CALC-AN02-EP-17-00002 Rev O Page 6 of 88 Westinghouse Non-Proprietary Class 3 lll TABLE OF CONTENTS LIST OF TABLES ... ....... .......... ..... .... .. .............................................. ......... .. ...... ....... ............. ... .................. iv LIST OF FIGURES ..................................................................................................................................... vi EXECUTIVE

SUMMARY

......................................................................................................................... vii INTRODUCTION ...................................... .. .. .. ........................................ ........................... ...... ... 1-1 2 CALCULATED NEUTRON FLUENCE .................................................... ............. .................... 2-1

2.1 INTRODUCTION

.... ............... ... ................................................. ........... .................. ....... 2-l 2.2 DISCRETE ORDINATES ANALYSIS ... ............................................. ................. .......... 2-1 2.3 CALCULATIONAL UNCERTAINTIES ......................................... ............... ................ 2-3 3 FRACTURE TOUGHNESS PROPERTIES .. .......... ........ ............................................................. 3-1 4 SURVEILLANCE DATA ........................... .. ................................................................................ 4-1 5 CHEMISTRY FACTORS .................................. ........ ....................... .. .. ........................................ 5-1 6 CRITERIAFORALLOWABLE PRESSURE-TEMPERATURE RELATIONSHIPS ... ............. 6-1

6. 1 OVERALLAPPROACH ................. ..................... ....... ............... .... ............... ..... ............. 6-1 6.2 METHODOLOGY FOR PRESSURE-TEMPERATURE LIMIT CURVE DEVELOPMENT ............... ........ ......................................... ............................... ....... ...... 6-1 6.3 CLOSURE HEADNESSEL FLANGE REQUIREMENTS ................ ......... ... ............... 6-5 6.4 LOWEST SERVICE TEMPERATURE REQUIREMENTS ..... ......... .......................... ... 6-5 6.5 BOLTUP TEMPERATURE REQUIREMENTS ..... .. .. .............. .. ............................... ... .. 6-5 7 CALCULATION OF ADWSTED REFERENCE TEMPERATURE ........... ... .............. .............. 7-1 8 HEATUP AND COOLDOWN PRESSURE-TEMPERATURE LIMIT CURVES ....... ................ 8-1 9 REFERENCES ... ..................... ..... .............. ...................... ............... ....... .. ............. ....... ................ 9-1 APPENDIXA THERMAL STRESS INTENSITY FACTORS (K11) ***** **** *** **** ***** *************** *** ** A-1 APPENDIXB REACTOR VESSEL INLET AND OUTLET NOZZLES ............. ................... B-1 APPENDIXC NON-REACTOR VESSEL FERRlTIC COMPONENTS ........... ........ .............. C-1 APPENDIXD CREDIBILITY EVALUATION OF THE WELD HEAT# 10137 SURVEILLANCE DATA .............. .......................... ....... ......... .................... ...... D-1 WCAP-18169-NP June2018 Revision 1

CALC-AN02-EP-17-00002 Rev O Page 7 of 88 Westinghouse Non-Proprietary Class 3 IV LIST OF TABLES Table 2-1 Pressure Vessel Material Locations ...................... .................. .. ...... ....... ......... ............ ...... 2-5 Table 2-2 Calculated Maximum Fast Neutron (E > 1.0 MeV) Fluence of the Pressure Vessel Clad/Base Metal Interface ....... ..... ... ..................... .......... ............... ......................... .......... 2-6 Table 2-3 Calculated Maximum Iron Atom Displacements at the Pressure Vessel Clad/Base Metal Interface ................... ..... ........... ........ ....... ......... ... .. ... ..... ......... ............................... ........... 2-7 Table 2-4 Calculated Azimuthal Variation of the Maximum Fast Neutron (E > 1.0 MeV) Fluence Rate at the Reactor Vessel Clad/Base Metal Interface ................................................... .. 2-8 Table 2-5 Calculated Azimuthal Variation of the Maximum Fast Neutron (E > 1.0 MeV) Fluence at the Reactor Vessel Clad/Base Metal Interface ...... .. .. ..... ... .. ............. ...... ..... .. ...................2-9 Table 2-6 Calculated Iron Atom Displacement Rate at the Pressure Vessel Clad/Base Metal Interface ..................... ....................... ......... ........................ ............. .. ... ..... ....... .. ............ 2-10 Table 2-7 Calculated Iron Atom Displacements at the Pressure Vessel Clad/Base Metal Interface .....

......................... ........ .................. .... ................... .. ........... ..... .......... .............. ................... 2-l 1 Table 2-8 Calculational Uncertainties ........ ................. ........ ........ ... ................... .. ..... ......... .. ... ........ 2-12 Table 3-1 Summary of the Best-Estimate Chemistry and Initial RT NDT Values for the Arkansas Nuclear One Unit 2 Reactor Vessel Materials ............. .. .. .. ............................. .. ......... ...... 3-2 Table 3-2 Summary of Arkansas Nuclear One Unit 2 Reactor Vessel Closure Heads, Vessel Flange and Balance of RCS Initial RTNDT Values ..... ............................ ..................................... .. 3-4 Table 4-1 Arkansas Nuclear One Unit 2 Surveillance Capsule Data ...................... ... ..................... .4-2 Table 4-2 Calvert Cliffs Unit 2 and Millstone Unit 2 Surveillance Capsule Data for Weld Heat #

10137 ...... ............ ... .. ..... ............. ............... ......... ....... ........ ..... ... ... .. ....................... ... ... .. ...4-3 Table 4-3 J.M. Farley Unit 2 Surveillance Capsule Data for Weld Heat# BOLA. ......................... .4-3 Table 5-1 Calculation of Arkansas Nuclear One Unit 2 Chemistry Factor Value for Intermediate Shell Plate C-8009-3 Using Surveillance Capsule Data ................ ... ...... ....... ........... .. ..... 5-2 Table 5-2 Calculation of Arkansas Nuclear One Unit 2 Chemistry Factor Value for Weld Heat #

83650 Using Surveillance Capsule Data ..... ........ ....................... ............... .. ...... .. .. .... .. ..... 5-2 Table 5-3 Calculation of Arkansas Nuclear One Unit 2 Chemistry Factor Value for Weld Heat #

10137 Using Surveillance Capsule Data ...... ...... ................................. ...................... .... .. . 5-3 Table 5-4 Calculation of Arkansas Nuclear One Unit 2 Chemistry Factor Value for Weld Heat #

BOLA Using Surveillance Capsule Data ..................... .. .... .............................................. 5-4 Table 5-5 Summary of Arkansas Nuclear One Unit 2 Positions 1.1 and 2.1 Chemistry Factors ..... 5-5 Table 7-1 Fluence Values and Fluence Factors for the Vessel Surface, l /4T and 3/4T Locations for the Arkansas Nuclear One Unit 2 Reactor Vessel Materials at 54 EFPY ..... ................... 7-3 WCAP-18169-NP June 2018 Revision I

CALC-AN02-EP-17-00002 Rev O Page 8 of 88 Westinghouse Non-Proprietary Class 3 V Table 7-2 Adjusted Reference Temperature Evaluation for the Arkansas Nuclear One Unit 2 Reactor Vessel Beltline Materials through 54 EFPY at the l /4T Location .................... .. 7-4 Table 7-3 Adjusted Reference Temperature Evaluation for the Arkansas Nuclear One Unit 2 Reactor Vessel Beltline Materials through 54 EFPY at the 3/4T Location ... .......... ......... 7-6 Table 7-4 Summary of the Increased Limiting ART Values Used in the Generation of the Arkansas Nuclear One Unit 2 Heatup and Cooldown Curves at 54 EFPY ..................................... 7-8 Table 8-1 Arkansas Nuclear One Unit 2 54 EFPY Heatup Curve Data Points using the 1998 through the 2000 Addenda App. G Methodology (w/ K1c, w/ Flange and LST Requirements, and w/o Margins for Instrumentation Errors) ... .. ...... ........ .. .. ...................... ............................ 8-6 Table 8-2 Arkansas Nuclear One Unit 2 54 EFPY Cooldown Curve Data Points using the 1998 through the 2000 Addenda App. G Methodology (w/ K1c, w/ Flange and LST Requirements, and w/o Margins for Instrumentation Errors) ... ............... .. .. .................... 8-8 Table 8-3 Arkansas Nuclear One Unit 2 54 EFPY Inservice Hydrostatic and Leak Test Curve Data Points using the 1998 through the 2000 Addenda App. G Methodology (w/ K1c, w/ Flange and LST Requirements, and w/o Margins for Instrumentation Errors) ......................... .. 8-9 Table A-1 K1t Values for Arkansas Nuclear One Unit 2 at 54 EFPY 80°F/hr Heatup Curves (w/

Flange and LST Requirements, and w/o Margins for Instrument Errors) ....... ................ A-2 Table A-2 K 11 Values for Arkansas Nuclear One Unit 2 at 54 EFPY 100°F/hr Cooldown Curves (w/

Flange and LST Requirements, and w/o Margins for Instrument Errors) .. ........ ............. A-3 Table B-1 Summary of the Arkansas Nuclear One Unit 2 Reactor Vessel Nozzle Material Initial RT NDT, Chemistry, and Fluence Values at 54 EFPY .......................................... ... .......... B-3 Table B-2 Summary of the Limiting ART Values for the Arkansas Nuclear One Unit 2 Inlet and Outlet Nozzle Materials .......................... ............................................................... ......... B-3 Table D-1 Mean Chemical Composition and Operating Temperature for Calvert Cliffs Unit 2 and Millstone Unit 2 .:.................. ............................................. .. ............ .. .. ....... ..... ............... D-4 Table D-2 Operating Temperature Adjustments for the Calvert Cliffs Unit 2 and Millstone Unit 2 Surveillance Capsule Data ........... ... ...... ........................... .. ...... .... ... ................. .. .. ........... D-4 Table D-3 Calculation of Weld Heat # 10137 Interim Chemistry Factor for the Credibility Evaluation Using Calvert Cliffs Unit 2 and Millstone Unit 2 Surveillance Capsule Data ...

............ .. ......................................... ............ ..................................................................... D-5 Table D-4 Best-Fit Evaluation for Surveillance Weld Metal Heat # 10137 Using Calvert Cliffs Unit 2 and Millstone Unit 2 Data ................................................... ............. .. .......................... D-5 WCAP-181 69-NP June 2018 Revision 1

CALC-AN02-EP-17 -00002 Rev O Page 9 of 88 Westinghouse Non-Proprietary Class 3 VI LIST OF FIGURES Figure 2-1 Arkansas Nuclear One Unit 2 r,e Reactor Geometry Plan View at the Core Midplane with Surveillance Capsules ................ ............................................................... ..................... 2-13 Figure 2-2 Arkansas Nuclear One Unit 2 r,e Reactor Geometry Plan View at the Core Midplane without Surveillance Capsules .. ................ ..................................................................... 2-14 Figure 2-3 Arkansas Nuclear One Unit 2 r,z Reactor Geometry Section View ..... .............. ......... ...2-15 Figure 8-1 Arkansas Nuclear One Unit 2 Reactor Coolant System Heatup Limitations (Heatup Rates of 50, 60, 70, and 80°F/hr) Applicable for 54 EFPY (with Flange and LST Requirements and without Margins for Instrumentation Errors) using the 1998 through the 2000 Addenda App. G Methodology (w/ K1c) .................................... ...................................... 8-3 Figure 8-2 Arkansas Nuclear One Unit 2 Reactor Coolant System Cooldown Limitations (Cooldown Rates of 0, 25 , 60, and 100°F/hr) Applicable for 54 EFPY (with Flange and LST Requirements and without Margins for Instrumentation Errors) using the 1998 through the 2000 Addenda App. G Methodology (w/ K1c) ................................. ........................... 8-4 Figure 8-3 Arkansas Nuclear One Unit 2 Reactor Coolant System Inservice Hydrostatic and Leak Test Limitations Applicable for 54 EFPY (with Flange and LST Requirements and without Margins for Instrumentation Errors) using the 1998 through the 2000 Addenda App. G Methodology (w/ K1c) .......................................................................................... 8-5 Figure B-1 Comparison of Arkansas Nuclear One Unit 2 Beltline P-T Limits to Inlet Nozzle Limits

............ ........... ............... ........... ...... ................................................................................. B-6 Figure B-2 Comparison of Arkansas Nuclear One Unit 2 Beltline P-T Limits to Outlet Nozzle Limits

....................................................... ...... ............ ................... ................................... ....... .. B-7 WCAP-181 69-NP June 2018 Revision 1

CALC-AN02-EP-17-00002 Rev O Page 10 of 88 Westinghouse Non-Proprietary Class 3 vii EXECUTIVE

SUMMARY

This report provides the methodology and results of the generation of heatup and cooldown pressure-temperature (P-T) limit curves for normal operation of the Arkansas Nuclear One Unit 2 reactor vessel.

The heatup and cooldown P-T limit curves were generated using the limiting Adjusted Reference Temperature (ART) values for Arkansas Nuclear One Unit 2. The limiting ART values were those of Lower Shell Plate C-8010-1 (Position 1.1) at both 1/4 thickness (1 /4T) and 3/4 thickness (3/4T) locations.

The P-T limit curves were generated using the K,c methodology detailed in the 1998 through the 2000 Addenda Edition of the ASME Code, Section XI, Appendix G.

The P-T limit curves were generated for 54 effective full-power years (EFPY) using heatup rates of 50, 60, 70, and 80°F/hr, and cooldown rates of O (steady-state), -25, -60, and -100°F/hr. The curves were developed with the flange and lowest service temperature (LST) requirements and without margins for instrumentation errors. They can be found in Figures 8-1 and 8-2.

Appendix A contains the thermal stress intensity factors for the maximum heatup and cooldown rates at 54EFPY.

Appendix B contains a P-T limit evaluation of the reactor vessel inlet and outlet nozzles based on a 1/4T flaw postulated at the inside surface of the reactor vessel nozzle comer, where T is the thickness of the vessel at the nozzle comer region. As discussed in Appendix B, the P-T limit curves generated based on the limiting cylindrical beltline material (Lower Shell Plate C-8010-1) bound the P-T limit curves for the reactor vessel inlet and outlet nozzles for Arkansas Nuclear One Unit 2 at 54 EFPY.

Appendix C contains discussion of the other non-reactor vessel ferritic Reactor Coolant Pressure Boundary (RCPB) components relative to P-T limits. As discussed in Appendix C, all of the non-reactor vessel ferritic RCPB components meet the applicable requirements of Section III of the ASME Code.

Appendix D contains a credibility evaluation for weld Heat# 10137 considering all applicable sister plant surveillance program data.

WCAP-18169-NP June 2018 Revision I

CALC-AN02-EP-17-00002 Rev O Page 11 of 88 Westinghouse Non-Proprietary Class 3 1-1 1 INTRODUCTION Heatup and cooldown P-T limit curves are calculated using the adjusted RTNDT (reference nil-ductility temperature) corresponding to the limiting beltline region material of the reactor vessel. The adjusted RT NDT of the limiting material in the core region of the reactor vessel is determined by using the unirradiated reactor vessel material fracture toughness properties, estimating the radiation-induced dRT NDT, and adding a margin. The unirradiated RTNDT is designated as the higher of either the drop weight nil-ductility transition temperature (NDTT) or the temperature at which the material exhibits at least 50 ft-lb of impact energy and 35-mil lateral expansion (normal to the major working direction) minus 60°F. For the purposes of this report, heatup is defined as the process of heating the reactor coolant system (RCS) from ambient temperature to operating temperature. Cooldown is defined as the process of cooling the RCS from operating temperature to ambient temperature. Steady-state is defined as a 0°F/hr cooldown or heatup rate. Under steady-state, the thermal stress intensity factor is considered to be zero.

The steady-state curve is necessary from an engineering perspective for comparison with heatup and cooldown.

RT NDT increases as the material is exposed to fast-neutron radiation. Therefore, to find the most limiting RTNDT at any time period in the reactor's life, dRTNDT due to the radiation exposure associated with that time period must be added to the unirradiated RT NDT (RTNDT(U)). The extent of the shift in RTNDT is enhanced by certain chemical elements (such as copper and nickel) present in reactor vessel steels. The U.S. Nuclear Regulatory Commission (NRC) has published a method for predicting radiation embrittlement in Regulatory Guide 1.99, Revision 2 [Ref. l]. Regulatory Guide 1.99, Revision 2 is used for the calculation of Adjusted Reference Temperature (ART) values (RTNDT(U) + dRTNDT + margins for uncertainties) at the l /4T and 3/4T locations, where Tis the thickness of the vessel at the beltline region measured from the clad/base metal interface.

The heatup and cooldown P-T limit curves documented in this report were generated using the most limiting ART values (plus an additional margin to account for future perturbations such as an uprate or surveillance capsule results) and the NRC-approved methodology documented in WCAP-14040-A, Revision 4 [Ref. 2]. Specifically, the Kie methodology of the 1998 through the 2000 Addenda Edition of ASME Code, Section XI, Appendix G [Ref. 3] was used. The Kie curve is a lower bound static fracture toughness curve obtained from test data gathered from several different heats of pressure vessel steel.

The limiting material is indexed to the Kic curve so that allowable stress intensity factors can be obtained for the material as a function of temperature. Allowable operating limits are then determined using the allowable stress intensity factors.

The purpose of this report is to present the calculations and the development of the Arkansas Nuclear One Unit 2 heatup and cooldown P-T limit curves for 54 EFPY. This report documents the calculated ART values and the development of the P-T limit curves for normal operation. The calculated ART values for 54 EFPY are documented in Section 7 of this report. The fluence projections used in calculation of the ART values are provided in Section 2 of this report.

The P-T limit curves herein were generated without instrumentation errors. The reactor vessel flange requirements of 10 CFR 50, Appendix G [Ref. 4] have been incorporated in the P-T limit curves, along with the lowest service temperature (LST) requirements of ASME Code, Section III [Ref. 9]. As discussed in Appendix B, the P-T limit curves generated in Section 8 bound the P-T limit curves for the reactor vessel inlet and outlet nozzles for Arkansas Nuclear One Unit 2 at 54 EFPY. Discussion of the other non-reactor vessel ferritic RCPB components relative to P-T limits is contained in Appendix C.

WCAP-18169-NP June 2018 Revision 1

CALC-AN02-EP-17-00002 Rev O Page 12 of 88 Westinghouse Non-Proprietary Class 3 2-1 2 CALCULATED NEUTRON FLUENCE

2.1 INTRODUCTION

A discrete ordinates (SN) transport analysis was performed for the Arkansas Nuclear One Unit 2 reactor to determine the neutron radiation environment within the reactor pressure vessel. In this analysis, radiation exposure parameters were established on a plant- and fuel-cycle-specific basis. An evaluation of the dosimetry sensor sets from the 284° and 97° surveillance capsules is provided in WCAP-18166-NP

[Ref. 21]. The dosimetry analysis documented in WCAP-18166-NP showed that the +/-20% (lcr) acceptance criterion specified in Regulatory Guide 1.190 [Ref. 5] is met. The validated calculations form the basis for providing projections of the neutron exposure of the reactor pressure vessel for operating periods extending to 54 EFPY.

All of the calculations described in this section were based on nuclear cross-section data derived from the Evaluated Nuclear Data File (ENDF) database (specifically, ENDF/B-Vl). Furthermore, the neutron transport evaluation methodologies follow the guidance of Regulatory Guide 1.190 [Ref. 5]. Additionally, the methods used to develop the calculated pressure vessel fluence are consistent with the NRC-approved methodology described in WCAP-14040-A, Revision 4 [Ref. 2].

2.2 DISCRETE ORDINATES ANALYSIS In performing the fast neutron exposure evaluations for the Arkansas Nuclear One Unit 2 reactor vessel, a series of fuel-cycle-specific forward transport calculations were performed using the following three-dimensional fluence rate synthesis technique:

cp(r,0,z) = cp(r,0) x cp(r,z) cp(r) where cp(r, 0,z) is the synthesized three-dimensional neutron fluence rate distribution, cp(r,0) is the transport solution in r,0 geometry, cp(r,z) is the two-dimensional solution for a cylindrical reactor model using the actual axial core power distribution, and cp(r) is the one-dimensional solution for a cylindrical reactor model using the same source per unit height as that used in the r,0 two-dimensional calculation.

This synthesis procedure was carried out for each operating cycle at Arkansas Nuclear One Unit 2.

For the Arkansas Nuclear One Unit 2 transport calculations, the r,0 model depicted in Figure 2-1 and Figure 2-2 were utilized since, with the exception of the capsules, the reactor is octant symmetric. These r,8 models included the core, the reactor internals, octants with surveillance capsules at 7° and 14° and octants without surveillance capsules, the pressure vessel cladding and vessel wall, the insulation external to the pressure vessel, and the primary biological shield wall. These models formed the basis for the calculated results and enabled making comparisons to the surveillance capsule dosimetry evaluations.

Specifically, the r,8 model with surveillance capsules was utilized to perform capsule dosimetry evaluations and subsequent comparisons with calculated results, while the r,0 model without surveillance capsules was used to generate the maximum fluence levels at the pressure vessel wall. In developing these analytical models, nominal design dimensions were generally employed for the various structural components. Note that for the pressure vessel inner radius, however, the average of the as-built inner radii WCAP-181 69-NP June 2018 Revision 1

CALC-AN02-EP-17-00002 Rev O Page 13 of 88 Westinghouse Non-Proprietary Class 3 2-2 was used. In addition, water temperatures and, hence, coolant densities in the reactor core and downcomer regions of the reactor were taken to be representative of full-power operating conditions. The reactor core itself was treated as a homogeneous mixture of fuel, cladding, water, and miscellaneous core structures such as fuel assembly grids, guide tubes, et cetera. The geometric mesh description of the r,0 reactor model in Figure 2-1 consisted of 166 radial by 123 azimuthal intervals. The geometric mesh description of the r,0 reactor model in Figure 2-2 consisted of 166 radial by 116 azimuthal intervals. Mesh sizes were chosen to assure that proper convergence of the inner iterations was achieved on a pointwise basis. The pointwise inner iteration fluence rate convergence criterion utilized in the r,0 calculations was set at a value of0.001.

The r,z model used for the Arkansas Nuclear One Unit 2 calculations is shown in Figure 2-3. The model extends radially from the centerline of the reactor core out to the primary biological shield and axially from an elevation approximately 4.9 feet below to 6 feet above the active core. As in the case of the r,0 models, nominal design dimensions, with the exception of the pressure vessel inner radius, and full-power coolant densities were employed in the calculations. In the r,z model, the homogenous core region was treated as an equivalent cylinder with a volume equal to that of the active core zone. The stainless steel girth ribs located between the core shroud and core barrel regions were also explicitly included in the model. The geometric mesh description of the r,z reactor model in Figure 2-3 consisted of 161 radial by 222 axial intervals. As in the case of the r,0 calculations, mesh sizes were chosen to assure that proper convergence of the inner iterations was achieved on a pointwise basis. The pointwise inner iteration fluence rate convergence criterion utilized in the r,z calculations was set at a value of0.001.

The one-dimensional radial model used in the synthesis procedure consisted of the same 161 radial mesh intervals included in the r,z model. Thus, radial synthesis factors could be determined on a meshwise basis throughout the entire geometry.

The core power distributions used in the plant-specific transport analysis for each of the first 25 fuel cycles at Arkansas Nuclear One Unit 2 included cycle-dependent fuel assembly initial enrichments, burnups, and axial power distributions (note that Cycles 1-24 have been completed; Cycle 25 is based on the expected core design for this cycle and an assumed cycle length of 1.37 EFPY). This information was used to develop spatial- and energy-dependent core source distributions averaged over each individual fuel cycle. Therefore, the results from the neutron transport calculations provided data in terms of fuel cycle-averaged neutron fluence rate, which, when multiplied by the appropriate fuel cycle length, generated the incremental fast neutron exposure for each fuel cycle. In constructing these core source distributions, the energy distribution of the source was based on an appropriate fission split for uranium and plutonium isotopes based on the initial enrichment and burnup history of individual fuel assemblies.

From these assembly-dependent fission splits, composite values of energy release per fission, neutron yield per fission, and fission spectrum were determined.

All of the transport calculations supporting this analysis were performed using the DORT discrete ordinates code [Ref. 23) and the BUGLE-96 cross-section library [Ref. 7). The BUGLE-96 library provides a 67-group coupled neutron-gamma ray cross-section data set produced specifically for light-water reactor (LWR) applications. In these analyses, anisotropic scattering was treated with a P 5 Legendre expansion and angular discretization was modeled with an S 16 order of angular quadrature. Energy- and space-dependent core power distributions, as well as system operating temperatures, were treated on a fuel-cycle -specific basis.

WCAP-1 8169-NP June 20 18 Revision I

CALC-AN02-EP-17 -00002 Rev O Page 14 of 88 Westinghouse Non-Proprietary Class 3 2-3 The locations of the Arkansas Nuclear One Unit 2 vessel welds and plates are provided in Table 2-1. The axial position of each material is indexed to z = 0.0 cm, which corresponds to the rnidplane of the active fuel stack.

These data tabulations include both plant- and fuel-cycle-specific calculated neutron exposures at the end of Cycle 24, at the end of projected Cycle 25, and at further projections to 54 EFPY. The calculations account for the uprate from 2815 MWt to 3026 MWt that occurred at the beginning of Cycle 16. The projections are based on the assumption that the core power distributions and associated plant operating characteristics from Cycle 23, Cycle 24, and the design of Cycle 25 are representative of future plant operation. The future projections are based on the current reactor power level of 3026 MWt.

Selected results from the neutron transport analyses are provided in Table 2-2 through Table 2-7. In Table 2-2, the calculated maximum fast neutron (E > 1.0 MeV) fluence values for the reactor pressure vessel materials are provided at future projections to 32, 36, 40, 48 and 54 EFPY. The projections are based on the assumption that the core power distributions and associated plant operating characteristics from Cycles 23-25 are representative of future plant operation. In Table 2-3 , the calculated maximum iron atom displacement values for the reactor pressure vessel materials are provided at future projections to 32, 36, 40, 48 and 54 EFPY.

The calculated fast neutron (E > 1.0 MeV) fluence rate, fast neutron (E > 1.0 MeV) fluence , iron atom displacement rate, and iron atom displacements are provided in Table 2-4 through Table 2-7, respectively, for the reactor pressure vessel inner radius at four azimuthal locations, as well as the maximum exposure observed within the octant. The vessel data given in Table 2-4 through Table 2-7 were taken at the clad/base metal interface and represent maximum calculated exposure levels on the vessel.

2.3 CALCULATIONAL UNCERTAINTIES The uncertainty associated with the calculated neutron exposure of the Arkansas Nuclear One Unit 2 reactor pressure vessel materials is based on the recommended approach provided in Regulatory Guide 1.190. In particular, the qualification of the methodology was carried out in the following four stages:

1. Comparison of calculations with benchmark measurements from the Pool Critical Assembly (PCA) simulator at the Oak Ridge National Laboratory (ORNL).

2. Comparisons of calculations with surveillance capsule and reactor cavity measurements from the H.B. Robinson power reactor benchmark experiment.

3. An analytical sensitivity study addressing the uncertainty components resulting from important input parameters applicable to the plant-specific transport calculations used in the neutron exposure assessments.

4. Comparisons of the plant-specific calculations with all available dosimetry results from the Arkansas Nuclear One Unit 2 surveillance program.

WCAP-18 169-NP June2018 Revision 1

CALC-AN02-EP-17-00002 Rev O Page 15 of 88 Westinghouse Non-Proprietary Class 3 2-4 The first phase of the methods qualification (PCA comparisons) addressed the adequacy of basic transport calculation and dosimetry evaluation techniques and associated cross-sections. This phase, however, did not test the accuracy of commercial core neutron source calculations nor did it address uncertainties in operational or geometric variables that impact power reactor calculations.

The second phase of the qualification (H.B. Robinson comparisons) addressed uncertainties in these additional areas that are primarily methods related and would tend to apply generically to all fast neutron exposure evaluations.

The third phase of the qualification (analytical sensitivity study) identified the potential uncertainties introduced into the overall evaluation due to calculational methods approximations as well as to a lack of knowledge relative to various plant-specific input parameters. The overall calculational uncertainty applicable to the Arkansas Nuclear One Unit 2 analysis was established from results of these three phases of the methods qualification.

The fourth phase of the uncertainty assessment (comparisons with Arkansas Nuclear One Unit 2 measurements) was used solely to demonstrate the validity of the transport calculations and to confirm the uncertainty estimates associated with the analytical results. The comparison was used only as a check and was not used in any way to modify the calculated surveillance capsule or pressure vessel neutron exposures.

Table 2-8 summarizes the uncertainties developed from the first three phases of the methodology qualification. Additional information pertinent to these evaluations is provided in Reference 6. The net calculational uncertainty was determined by combining the individual components in quadrature.

Therefore, the resultant uncertainty was treated as random and no systematic bias was applied to the analytical results. The plant-specific measurement comparisons given in Appendix A of Reference 21 support these uncertainty assessments for Arkansas Nuclear One Unit 2.

WCAP-18169-NP June 2018 Revision I

CALC-AN02-EP-17-00002 Rev O Page 16 of 88 Westinghouse Non-Proprietary Class 3 2-5 Table 2-1 Pressure Vessel Material Locations Axial Location Relative to Core Azimuthal Location Material Midplane at Ocm (degrees)

(cm)

Inlet Nozzle to Upper Shell Welds - Lowest Extent Nozzle I 301.625(*) 60 Nozzle 2 30 1.625(*) 120 Nozzle 3 30 1.625(*) 240 Nozzle 4 301.625(*) 300 Outlet Nozzle to Upper Shell Welds - Lowest Extent Nozzle I 301.625(*) 0 Nozzle 2 301.625<*> 180 Upper Shell to Intermediate Shell Circumferential Weld 8-203 248.722 0 to 360 Intermediate Shell Plates C-8009-1 , -2, -3 -33.973 to 248 .722 0 to 360(b)

Intermediate Shell Longitudinal Welds 2-203 A -33.973 to 248.722 90 2-203 B -33 .973 to 248.722 210 2-203 C -33 .973 to 248.722 330 Intermediate Shell to Lower Shell Circumferential Weld 9-203 -33.973 0 to 360 Lower Shell Plates C-8010-1 , -2, -3 -3 15.892 to -33.973 0 to 360(c)

Lower Shell Longitudinal Welds 3-203 A -3 I 5.892 to -33.973 90 3-203 B -315.892 to -33.973 210 3-203 C -3 I 5.892 to -33 .973 330 Lower Shell to Bottom Head Circumferential Weld 10-203 -315 .892 0 to 360 Notes:

(a) This axial location corresponds to the bottom of the vessel support pad of the inlet nozzle, instead of the nozzle to upper shell weld. This provides a bounding fluence for the nozzle to upper shell weld.

(b) Intermediate shell plates C-8009- 1, -2, and -3 extend from azimuthal angles of330° to 90°, 90° to 210°,

and 210° to 330°, respectively.

(c) Lower shell plates C-80 10- 1, -2, and -3 extend from azimuthal angles of210° to 330°, 330° to 90°, and 90° to 210°, respectively.

WCAP- 18 169-NP June2018 Revision I

CALC-AN02-EP-17 -00002 Rev O Page 17 of 88 Westinghouse Non-Proprietary Class 3 2-6 Table 2-2 Calculated Maximum Fast Neutron (E > 1.0 MeV) Fluence of the Pressure Vessel Clad/Base Metal Interface Fluence(b) (n/cm2)

Material 32 EFPY 36 EFPY 40 EFPY 48 EFPY 54 EFPY Inlet Nozzle to Upper Shell Welds - Lowest Extent Nozzle 1<*) 4.78E+ l6 5.36E+l6 5.93E+16 7.09E+l6 7.96E+ l 6 Nozzle 2<*l 4.78E+l6 5.36E+l6 5.93E+l6 7.09E+l6 7.96E+l6 Nozzle 3(a) 4.78E+ 16 5.36E+16 5.93E+16 7.09E+l6 7.96E+l6 Nozzle 4(*) 4.78E+ l6 5.36E+ 16 5.93E+ l6 7.09E+ l6 7.96E+ 16 Outlet Nozzle to Upper Shell Welds - Lowest Extent Nozzle 1<*) 6.05E+ 16 6.73E+ 16 7.41E+ 16 8.77E+ l6 9.80E+16 Nozzle 2<*) 6.05E+ 16 6.73E+ 16 7.4 1E+ l6 8.77E+ l6 9.80E+ l6 Upper Shell to Intermediate Shell Circumferential Weld 8-203 3.53E+ l7 3.96E+ l7 4.38E+17 5.24E+ l7 5.89E+ l7 Intermediate Shell Plates C-8009-1 , -2, -3 3.02E+ 19 3.36E+19 3.70E+ 19 4.39E+l9 4.91E+19 Intermediate Shell Longitudinal Welds 2-203 A 2.89E+l9 3.21E+19 3.53E+ 19 4.16E+l9 4.64E+ l9 2-203 B 2.22E+ 19 2.48E+19 2.75E+ 19 3.28E+ l9 3.68E+ 19 2-203 C 2.22E+l9 2.48E+ 19 2.75E+ 19 3.28E+\9 3.68E+ l9 Intermediate Shell to Lower Shell Circumferential Weld 9-203 3.00E+ 19 3.35E+ l9 3.69E+ l9 4.38E+l9 4.89E+ l 9 Lower Shell Plates C-8010- 1, -2, -3 3.02E+ l9 3.37E+19 3.73E+ l9 4.44E+ 19 4.98E+19 Lower Shell Longitudinal Welds 3-203 A 2.89E+ I9 3.22E+ 19 3.55E+l9 4.2IE+ I9 4.71E+ 19 3-203 B 2.22E+l9 2.49E+19 2.77E+19 3.32E+ l9 3.73E+ l9 3-203 C 2.22E+ 19 2.49E+ l9 2.77E+ l 9 3.32E+ l 9 3.73E+ 19 Lower Shell to Bottom Head Circumferential Weld 10-203 5.41E+ l6 6.06E+ 16 6.71E+l6 8.0IE+ l 6 8.98E+ l 6 Notes:

(a) The axial location used corresponds to the bottom of the vessel support pad of the inlet nozzle, instead of the nozzle to upper shell weld. This provides a bounding fluence for the nozzle to upper shell weld.

(b) Values are based on the average power distributions and core operating conditions of Cycles 23-25.

WCAP-18169-NP June 2018 Revision I

CALC-AN02-EP-17-00002 Rev O Page 18 of 88 Westinghouse Non-Proprietary Class 3 2-7 Table 2-3 Calculated Maximum Iron Atom Displacements at the Pressure Vessel Clad/Base Metal Interface Iron Atom Displacements<bl (dpa)

Material 32 EFPY 36 EFPY 40 EFPY 48 EFPY 54 EFPY Inlet Nozzle to Upper Shell Welds - Lowest Extent Nozzle 1<*l 2.83E-04 3. 17E-04 3.51E-04 4.19E-04 4.70E-04 Nozzle i*l 2.83E-04 3.17E-04 3.51 E-04 4.19E-04 4.70E-04 Nozzle 3<*l 2.83E-04 3.17E-04 3.51E-04 4.19E-04 4.70E-04 Nozzle 4<*l 2.83E-04 3. 17E-04 3.51E-04 4.19E-04 4.70E-04 Outlet Nozzle to Upper Shell Welds - Lowest Extent Nozzle 1<*l 3.37E-04 3.75E-04 4.14E-04 4.91E-04 5.48E-04 Nozzle 2<*l 3.37E-04 3.75E-04 4.14E-04 4.91E-04 5.48E-04 Upper Shell to Intermediate Shell Circumferential Weld 8-203 6.41E-04 7. 18E-04 7.95E-04 9.50E-04 1.07E-03 Intermediate Shell Plates C-8009-1, -2, -3 4.59E-02 5. l lE-02 5.64E-02 6.69E-02 7.47E-02 Intermediate Shell Longitudinal Welds 2-203 A 4.42E-02 4.90E-02 5.39E-02 6.36E-02 7.09E-02 2-203 B 3.39E-02 3.80E-02 4.20E-02 5.02E-02 5.63E-02 2-203 C 3.39E-02 3.80E-02 4.20E-02 5.02E-02 5.63E-02 Intermediate Shell to Lower Shell Circumferential Weld 9-203 4.57E-02 5. IO E-02 5.62E-02 6.67E-02 7.45E-02 Lower Shell Plates C-8010-1 , -2, -3 4.59E-02 5. I 3E-02 5.67E-02 6.76E-02 7.58E-02 Lower Shell Longitudinal Welds 3-203 A 4.41E-02 4.92E-02 5.42E-02 6.43E-02 7.19E-02 3-203 B 3.39E-02 3.81E-02 4.23E-02 5.08E-02 5.71E-02 3-203 C 3.39E-02 3.8 1E-02 4.23E-02 5.08E-02 5.71E-02 Lower Shell to Bottom Head Circumferential Weld 10-203 2.85E-04 3.19E-04 3.53E-04 4.22E-04 4.73E-04 Notes:

(a) The axial location used corresponds to the bottom of the vessel support pad of the inlet nozzle, instead of the nozzle to upper shell weld. This provides a bounding fluence for the nozzle to upper shell weld.

(b) Values are based on the average power distributions and core operating conditions of Cycles 23-25.

WCAP-18169-NP June2018 Revision 1

CALC-AN02-EP-17-00002 Rev O Page 19 of 88 Westinghouse Non-Proprietary Class 3 2-8 Table 2-4 Calculated Azimuthal Variation of the Maximum Fast Neutron (E > 1.0 MeV) Fluence Rate at the Reactor Vessel Clad/Base Metal Interface Cumulative Fluence Rate (n/cm 2-s)

Cycle Operating Cycle Length Time oo 15° 30° 45° Maximum (EFPY)

(EFPY) 1 0.89 0.89 3.62E+ l0 3.60E+ IO 2.71E+ IO 2.63E+IO 3.81E+IO 2 0.80 1.69 4.59E+ l0 4.51E+l0 3.38E+l0 3.22E+10 4.80E+l0 3 0.64 2.33 4.34E+l0 4.21E+l0 3.18E+l0 3.12E+l0 4.52E+ IO 4 0.97 3.31 4.36E+l0 4.12E+l0 3.34E+l0 3.33E+10 4.46E+l0 5 0.85 4.16 4.68E+ l0 4.52E+l0 3.34E+l0 3.30E+10 4.86E+IO 6 1.22 5.38 3.40E+ IO 3.40E+l0 2.54E+ l0 2.34E+10 3.64E+l0 7 1.13 6.51 3.l 7E+ l0 3.03E+ IO 2.39E+ l0 2.20E+l0 3.30E+ l0 8 1.15 7.66 3.25E+ IO 3.20E+IO 2.42E+ l0 2.17E+l0 3.44E+l0 9 1.18 8.84 3.20E+ l0 3.15E+IO 2.38E+IO 2.09E+IO 3.38E+l0 10 1.32 10.16 2.98E+l0 2.98E+ l0 2.35E+ l0 2.26E+l0 3.19E+IO II 1.33 11.49 2.36E+ IO 2.06E+IO l.73E+IO 1.83E+IO 2.36E+IO 12 1.31 12.8 1 2.41E+ l0 2.30E+ l0 l.77E+ l0 l.75E+ l0 2.48E+l0 13 1.47 14.27 2.29E+l0 l.99E+ l0 l.63E+ l0 l.78E+ l0 2.29E+l0 14 1.41 15.69 2.35E+ l0 2.28E+ 10 l.79E+ l0 l.78E+l0 2.42E+IO 15 1.29 16.98 2.31E+10 2.23E+10 l.90E+ 10 l.85E+ l0 2.34E+ l0 16 1.35 18.33 2.73E+l0 2.49E+ 10 1.90E+ l0 1.92E+ l0 2.75E+ l0 17 1.36 19.69 2.56E+ l0 2.47E+IO l.96E+ l0 l.89E+ l0 2.63E+10 18 1.43 21.12 2.77E+ l0 2.83E+ l0 2.21E+l0 2.13E+l0 2.95E+l0 19 1.34 22.46 2.71E+ l0 2.72E+ l0 2.09E+l0 2.06E+l0 2.85E+l0 20 1.36 23.82 2.72E+ l0 2.72E+ l0 2.lOE+ lO 2.1 lE+ lO 2.85E+l0 21 1.35 25.17 2.72E+l0 2.72E+l0 2.12E+IO 2.12E+l0 2.86E+ IO 22 1.45 26.61 2.59E+IO 2.59E+ IO 2.03E+IO 2.05E+IO 2.72E+ IO 23 1.36 27.98 2.50E+ l0 2.68E+ l0 2.13E+ IO 2.16E+l0 2.76E+ l0 24 1.26 29.24 2.40E+ IO 2.57E+ 10 2.14E+ l0 2.20E+l0 2.64E+ l0 25<*) 1.37 30.60 2.97E+ IO 2.96E+l0 2.29E+ l0 2.32E+l0 3.1 lE+ lO Note:

(a) Cycle 25 is the current operating cycle. Values listed for this cycle are projections based on the Cycle 25 design.

WCAP-1 8169-NP June2018 Revision I

CALC-AN02-EP-17-00002 Rev O Page 20 of 88 Westinghouse Non-Proprietary Class 3 2-9 Table 2-5 Calculated Azimuthal Variation of the Maximum Fast Neutron (E > 1.0 MeV) Fluence at the Reactor Vessel Clad/Base Metal Interface Cumulative Fluence (n/cm2 Cycle Operating Cycle Length Time oo 15° 30° 45° Maximum (EFPY)

(EFPY) 1 0.89 0.89 1.02E+l8 1.01 E+ 18 7.61E+17 7.37E+ l7 l.07E+18 2 0.80 1.69 2. 16E+l8 2.14E+ 18 l.6 1E+18 1.54E+18 2.27E+ 18 3 0.64 2.33 3.04E+18 2.99E+18 2.25E+18 2.1 7E+18 3.18E+18 4 0.97 3.31 4.38E+18 4.26E+18 3.27E+18 3. 19E+1 8 4.55E+18 5 0.85 4.16 5.64E+18 5.48E+1 8 4.17E+18 4.08E+ 18 5.86E+l 8 6 1.22 5.38 6.95E+18 6.78E+18 5.15E+18 4.98E+ 18 7.26E+ l 8 7 1.13 6.51 8.08E+18 7.86E+18 6.00E+18 5.77E+18 8.44E+ 18 8 1.15 7.66 9.26E+ l 8 9.02E+18 6.88E+18 6.56E+ l 8 9.68E+ l 8 9 1.18 8.84 1.04E+ 19 1.02E+19 7.76E+ l 8 7.33E+ 18 l.09E+ 19 10 1.32 10.16 l.1 7E+ 19 l.14E+ 19 8.74E+ 18 8.27E+ 18 l.23E+ 19 11 1.33 11.49 l.27E+ I9 l.23E+19 9.47E+ 18 9.04E+ 18 1.32E+ 19 12 1.31 12.81 l.37E+ l9 1.32E+19 l. 02E+ 19 9.76E+ 18 1.43E+19 13 1.47 14.27 l .47E+ l9 1.42E+19 1.09E+ 19 l.06E+ 19 l. 53E+19 14 1.41 15.69 1.57E+ 19 l.51E+l9 l.17E+ 19 1.14E+19 l.63E+ l 9 15 1.29 16.98 1.67E+ l9 l. 60E+l 9 l. 25E+ 19 l.21E+ l 9 l.73E+ 19 16 1.35 18.33 l.78E+19 l.7IE+19 1.33E+ l 9 l.29E+ l 9 l. 84E+ l9 17 1.36 19.69 l. 89E+ 19 l.81E+l9 l.41E+ l 9 1.37E+ I 9 l. 95E+ l 9 18 1.43 21.12 2.01E+ l9 1.94E+ l9 l.51E+ l 9 l.46E+ l 9 2.08E+ 19 19 1.34 22.46 2.12E+ l9 2. 05E+ l 9 1.59E+ l 9 l.55E+ l9 2.20E+ l 9 20 1.36 23 .82 2.23E+ 19 2. 16E+ l 9 1.68E+ l 9 l. 64E+ l9 2.32E+ 19 21 1.35 25.17 2.35E+l9 2.27E+ l 9 l.77E+ l 9 l.72E+ l 9 2.43E+ 19 22 1.45 26.6 1 2.46E+l9 2.39E+ l9 l. 86E+ l 9 l.81E+ l 9 2.55E+ I 9 23 1.36 27.98 2.57E+ l9 2.50E+l9 l.95E+ l 9 l.91E+ l 9 2.67E+ l 9 24 1.26 29.24 2.66E+l9 2. 60E+ l 9 2.03E+ l 9 l.99E+ l 9 2.77E+ l9 25<*) 1.37 30.60 2.78E+ l9 2.72E+ l 9 2. 12E+l9 2.08E+ l 9 2.90E+ l 9 Future<b) 32.00 2.89E+ l9 2.84E+l9 2.22E+ l 9 2. 18E+ l 9 3.02E+ l 9 Future(b) 36.00 3.22E+ l9 3. 18E+l9 2.49E+ l 9 2.46E+ l 9 3.37E+ l 9 Future<b) 40.00 3.55E+ 19 3.53E+19 2.77E+l9 2.74E+ l 9 3.73E+ l 9 Future(b) 48.00 4.21E+l9 4.22E+ l 9 3.32E+l9 3.30E+l 9 4.44E+ l 9 Future(b) 54.00 4.71E+l9 4.74E+ l 9 3.73E+l 9 3.72E+ l 9 4.98E+ l9 Notes:

(a) Cycle 25 is the current operating cycle. Values listed for this cycle are proj ections based on the Cycle 25 des ign.

(b) Values beyond Cycle 25 are based on the average power distributions and core operating conditions of Cycles 23-25 .

WCAP-18169-NP June201 8 Revision 1

CALC-AN02-EP-17-00002 Rev O Page 21 of 88 Westinghouse Non-Proprietary Class 3 2-10 Table 2-6 Calculated Iron Atom Displacement Rate at the Pressure Vessel Clad/Base Metal Interface Cumulative Iron Atom Displacement Rate (dpa/s)

Cycle Operating Cycle Length Time oo 15° 30° 45° Maximum (EFPY)

(EFPY) l 0.89 0.89 5.52E-11 5.47E-11 4.14E-ll 4.0 l E-11 5.79E-11 2 0.80 1.69 6.99E-l l 6.86E-11 5.16E-11 4.91E-1 l 7.29E- l l 3 0.64 2.33 6.62E-11 6.40E-11 4.85E- 11 4.76E-l l 6.87E-1 l 4 0.97 3.31 6.65E- l l 6.27E-1 l 5.l OE-11 5.08E-11 6.78E-ll 5 0.85 4.16 7.13E-11 6.87E- l l 5.l OE-1 1 5.04E-l l 7.39E-l l 6 1.22 5.38 5.20E-11 5.17E-1 1 3.88E- l l 3.58E- l 1 5.53E- l l 7 1.13 6.51 4.84E-1 l 4.61E-11 3.66E-l l 3.37E-11 5.02E-l l 8 1.15 7.66 4.97E-l l 4.86E-l l 3.69E-l 1 3.33E-11 5.24E-l l 9 1.18 8.84 4. 89E-l 1 4.79E-1 l 3.64E- 11 3.21E-l 1 5.1 5E- l l 10 1.32 10.16 4.56E-l l 4.54E-l l 3.59E-l l 3.45E-11 4.84E-l l 11 1.33 11.49 3.60E-11 3. 14E- l l 2.65E-11 2.80E- l l 3.60E-11 12 1.31 12.81 3.68E-11 3.51E-ll 2.71E- l l 2.68E-l l 3.78E-l l 13 1.47 14.27 3.50E- l l 3.03E- l l 2.50E-11 2.73E-ll 3.50E-11 14 1.41 15.69 3.59E-1 l 3.48E- l l 2.73E- l l 2.73E-ll 3.69E-1 l 15 1.29 16.98 3.53E- l l 3.41E- l l 2.91E- l 1 2.83E- l l 3.57E-11 16 1.35 18.33 4. l 7E-l l 3.80E- l l 2.91E-ll 2.95E-l l 4.19E-11 17 1.36 19.69 3.91E-l l 3.76E-1 1 3.0lE-11 2.89E- l l 4.0lE-11 18 1.43 21.12 4.23E-1 l 4.31E- l 1 3.38E-1 l 3.26E-11 4.49E-1 l 19 1.34 22.46 4.13E-11 4.14E- l l 3.20E- ll 3.16E-11 4.34E-11 20 1.36 23 .82 4.15E-ll 4.l3E-11 3.21E-11 3.23E-l 1 4.34E-11 21 1.35 25 .17 4.16E- ll 4.14E-ll 3.24E-11 3.25E-l l 4.35E-1 l 22 1.45 26.61 3.95E-l l 3.95E-l l 3.lOE-1 1 3.13E-l l 4.14E-l l 23 1.36 27.98 3.82E-l l 4.08E-l 1 3.26E- l l 3.29E-11 4.20E-l 1 24 1.26 29.24 3.67E-1 l 3.91E-l l 3.27E-1I 3.36E-11 4.02E-11 25<*) 1.37 30.60 4.53E-1 1 4.51E-1 1 3.50E- 11 3.55E-l 1 4.74E-11 Note:

(a) Cycle 25 is the current operating cycle. Values listed for this cycle are projections based on the Cycle 25 design.

WCAP-1 8169-NP June 201 8 Revision 1

CALC-AN02-EP-17-00002 Rev O Page 22 of 88 Westinghouse Non-Proprietary Class 3 2-11 Table 2-7 Calculated Iron Atom Displacements at the Pressure Vessel Clad/Base Metal Interface Cumulative Iron Atom Displacements (dpa)

Cycle Operating Cycle Length Time oo 15° 30° 45° Maximum (EFPY)

(EFPY) 1 0.89 0.89 l.55E-03 l.54E-03 1.16E-03 l.13E-03 l.63E-03 2 0.80 1.69 3.30E-03 3.25E-03 2.45E-03 2.36E-03 3.45E-03 3 0.64 2.33 4.63E-03 4.54E-03 3.43E-03 3.32E-03 4.83E-03 4 0.97 3.31 6.68E-03 6.47E-03 5.00E-03 4.88E-03 6.91E-03 5 0.85 4.16 8.60E-03 8.32E-03 6.37E-03 6.24E-03 8.90E-03 6 1.22 5.38 l .06E-02 l .03E-02 7.86E-03 7.61E-03 1.IOE-02 7 1.13 6.51 1.23E-02 l .20E-02 9.17E-03 8.82E-03 l .28E-02 8 1.15 7.66 l.41E-02 1.37E-02 l.05E-02 l.OOE-02 l.47E-02 9 1.18 8.84 1.59E-02 l .55E-02 l .19E-02 l.12E-02 l.66E-02 10 1.32 10.16 l.78E-02 l.74E-02 l .34E-02 1.27E-02 l.86E-02 11 1.33 11.49 l.94E-02 l.87E-02 l.45E-02 l .38E-02 2.0lE-02 12 1.31 12.81 2.09E-02 2.02E-02 l .56E-02 l.49E-02 2. l 7E-02 13 1.47 14.27 2.25E-02 2.15E-02 l .67E-02 l.62E-02 2.32E-02 14 1.41 15.69 2.40E-02 2.31E-02 l.79E-02 1.74E-02 2.48E-02 15 1.29 16.98 2.54E-02 2.44E-02 l.91E-02 l .85E-02 2.63E-02 16 1.35 18.33 2.72E-02 2.60E-02 2.03E-02 1.97E-02 2.80E-02 17 1.36 19.69 2.89E-02 2.76E-02 2.16E-02 2.lOE-02 2.97E-02 18 1.43 21.12 3.07E-02 2.95E-02 2.31E-02 2.24E-02 3.l 7E-02 19 1.34 22.46 3.24E-02 3.12E-02 2.44E-02 2.37E-02 3.35E-02 20 1.36 23 .82 3.41E-02 3.29E-02 2.57E-02 2.50E-02 3.53E-02 21 1.35 25 .17 3.58E-02 3.46E-02 2.70E-02 2.63E-02 3.70E-02 22 1.45 26.61 3.76E-02 3.64E-02 2.84E-02 2.77E-02 3.89E-02 23 1.36 27.98 3.92E-02 3.81E-02 2.98E-02 2.91E-02 4.06E-02 24 1.26 29.24 4.06E-02 3.96E-02 3.l lE-02 3.04E-02 4.22E-02 25<*) 1.37 30.60 4.25E-02 4.14E-02 3.25E-02 3.19E-02 4.41E-02 Future<bl 32.00 4.42E-02 4.32E-02 3.39E-02 3.33E-02 4.59E-02 Future(b) 36.00 4.92E-02 4.84E-02 3.81E-02 3.76E-02 5.13E-02 Future<b) 40.00 5.42E-02 5.37E-02 4.23E-02 4 .19E-02 5.67E-02 Future<b) 48.00 6.43E-02 6.42E-02 5.08E-02 5.05E-02 6.76E-02 Future<b) 54.00 7.19E-02 7.21E-02 5.71E-02 5.69E-02 7.58E-02 Notes:

(a) Cycle 25 is the current operating cycle. Values listed for this cycle are projections based on the Cycle 25 design.

(b) Values beyond Cycle 25 are based on the average power distributions and core operating conditions of Cycles 23-25 .

WCAP-18 169-NP June2018 Revision 1

CALC-AN02-EP-17-00002 Rev O Page 23 of 88 Westinghouse Non-Proprietary Class 3 2-12 Table 2-8 Calculational Uncertainties Description Capsule and Vessel IR PCA Comparisons 3%

H.B. Robinson Comparisons 3%

Analytical Sensitivity Studies 11%

Additional Uncertainty for Factors not Explicitly Evaluated 5%

Net Calculational Uncertainty 13%

WCAP-18169-NP June 2018 Revision 1

CALC-AN02-EP-17-00002 Rev O Page 24 of 88 Westinghouse Non-Proprietary Class 3 2-13

c. --*- __ _ - -C.*- --

AN0 - 2 Vessel Model - OORT - r,t Geometry with Capsules

-- -*'-'°" ....

Meshes: 166R,123S Figure 2-1 Arkansas Nuclear One Unit 2 r,0 Reactor Geometry Plan View at the Core Midplane with Surveillance Capsules WCAP-18169-NP June 2018 Revision 1

CALC-AN02-EP-17 -00002 Rev O Page 25 of 88 Westinghouse Non-Proprietary Class 3 2-14

_ - (,*-

AN0-2 Vessel Model -

DORT -

r ,t Geometry without Capsules Meshes: 166R,1169

- ,._, *- °" -

Figure 2-2 Arkansas Nuclear One Unit 2 r,0 Reactor Geometry Plan View at the Core Midplane without Surveillance Capsules WCAP-18169-NP June 2018 Revision 1

CALC-AN02-EP-17-00002 Rev O Page 26 of 88 Westinghouse Non-Proprietary Class 3 2-15 AN0 - 2 Vessel Model - DORT - r, z Geometry Meshes: 161R,222Z

- r,* - C<<t llrool - c.-. ir.....

- CH

- ,,_, v.... 0:,6 - , - ,, .....

0,

'°"'

e

~

N "1

N I

I;!

I ci

!51 I

"'I N

",\:).O 9U 182.9 270 365.8 R

[cml Figure 2-3 Arkansas Nuclear One Unit 2 r,z Reactor Geometry Section View WCAP-18169-NP June 2018 Revision 1

CALC-AN02-EP-17-00002 Rev O Page 27 of 88 Westinghouse Non-Proprietary Class 3 3-1 3 FRACTURE TOUGHNESS PROPERTIES The requirements for P-T limit curve development are specified in 10 CFR 50, Appendix G [Ref. 4]. The beltline region of the reactor vessel is defined as the following in 10 CFR 50, Appendix G:

"the region of the reactor vessel (shell material including welds, heat affected zones and plates or forgings) that directly surrounds the effective height of the active core and adjacent regions of the reactor vessel that are predicted to experience sufficient neutron radiation damage to be considered in the selection of the most limiting material with regard to radiation damage. "

The Arkansas Nuclear One Unit 2 beltline materials traditionally included the intermediate and lower shell plate and weld materials; however, as described in NRC Regulatory Issue Summary (RIS) 2014-11

[Ref. 8], any reactor vessel materials that are predicted to experience a neutron fluence exposure greater than 1.0 x 10 17 n/cm2 (E > 1.0 MeV) at the end of the licensed operating period should be considered in the development of P-T limit curves. The materials that exceed this fluence threshold are referred to as extended beltline materials and are evaluated to ensure that the applicable acceptance criteria are met. As seen from Table 2-2 of this report, the extended beltline materials include the upper shell plates, upper shell longitudinal welds, and the upper to intermediate shell girth weld. Note that for reactor vessel welds, the terms "girth" and "circumferential" are used interchangeably; herein, these welds shall be referred to as girth welds. Similarly, for reactor vessel welds, the terms "axial" and "longitudinal" are used interchangeably; herein, these welds shall be referred to as longitudinal welds. Although the reactor vessel nozzles are not a part of the extended beltline, per NRC RIS 2014-11 , the nozzle materials must be evaluated for their potential effect on P-T limit curves regardless of exposure - See Appendix B for more details.

As part of this P-T limit curve development effort, the methodology and evaluations used to determine the initial RTNOT values for the Arkansas Nuclear One Unit 2 reactor vessel beltline and extended beltline base metal materials were reviewed and updated, as appropriate. The initial RTNDT value of Intermediate Shell Plate C-8009-3 was determined per ASME Code, Section III, Subsection NB-2300 [Ref. 9]. The initial RT NOT values for each of the other eight reactor vessel plates were determined per BTP 5-3, Paragraph B l.1(3) [Ref. 10] in conj unction with ASME Code, Section III, Subsection NB-2300 [Ref. 9].

These initial RTNDT values were determined using both BTP 5-3 Position l.1(3)(a) and Position 1.1(3)(b),

and the more limiting initial RT DT value was chosen for each material. A summary of the best-estimate copper (Cu), nickel (Ni), Manganese (Mn), and Phosphorus (P) contents, in units of weight percent (wt.

%), as well as initial RT DT values for the reactor vessel beltline and extended beltline materials are provided in Table 3-1 for Arkansas Nuclear One Unit 2. Table 3-2 contains a summary of the initial RTNDT values of the reactor vessel flange, reactor vessel closure head, replacement reactor vessel closure head, and balance of the reactor coolant system (RCS). These values serve as input to the P-T limit curves "flange-notch" and LST per Appendix G of 10 CFR 50 and ASME Code, Section Ill, respectively

- See Sections 6.3 and 6.4 for details.

WCAP-18169-NP June2018 Revision l

CALC-AN02-EP-17 -00002 Rev O Page 28 of 88 Westinghouse Non-Proprietary Class 3 3-2 Table 3-1 Summary of the Best-Estimate Chemistry and Initial RT NDT Values for the Arkansas Nuclear One Unit 2 Reactor Vessel Materials Fracture Chemical Composition<bl Toughness Reactor Vessel Material Property Heat Number<*>

and Identification Number<*> (c)

Wt.% Wt.% Wt.% Wt.% Initial RT NDT Cu Ni Mn p (OF)

Reactor Vessel Beltline Materials Intermediate Shell Plate C-8009-1 C8161-3 0.098 0.605 1.35 0.010 -1.4 Intermediate Shell Plate C-8009-2 C8161-l 0.085 0.600 1.37 0.010 0.5 Intermediate Shell Plate C-8009-3 C8182-2 0.096 0.580 1.36 0.012 o.o<s)

Lower Shell Plate C-8010-1 C8161-2 0.085 0.585 1.33 0.009 12.0 Lower Shell Plate C-8010-2 B2545- l 0.083 0.668 1.36 0.008 -16.7 Lower Shell Plate C-80 I 0-3 B2545-2 0.080 0.653 1.36 0.008 -22.6 Intermediate Shell Longitudinal Welds 1.oo<d> l.16(d)

Multiple(d) o.05<d> o.014<d> -56 2-203A, B, & C Lower Shell Longitudinal Welds 0.046(e) 10120 0.08zC 0 > l.21 0.01 2 -56 3-203A, B, & C Intermediate to Lower Shell Girth .40<&)

83650 0.045<*) 0.08i 0 > 1.24 0.006 Weld 9-203 Reactor Vessel Extended Beltline Materials Uooer Shell PlateG> C-8008-1 C8182-I 0.13 0.60 1.36 0.011 12.2 Upper Shell Plate C-8008-2 C7605-l 0.13 0.55 1.36 0.013 60.5 Uooer Shell Plate C-8008-3 C8571-2 0.08 0.55 l.29 0.014 27.3 Upper Shell Longitudinal Welds -60(g)

BOLA 0.02 0.93 1.02 0.010 l-203A, B, & C 10137 0.22 0.02 0.94 0.015 -56 Upper to Intermediate Shell Girth 0.11 <1) 6329637 0.21 1.22 0.011 -56 Weld 8-203 FAGA 0.03 0.95 1.00 0.008 -24<hJ Surveillance Weld Data<il Arkansas Nuclear One Unit 2 83650 0.045 0.083 1.33 0.007 ---

Calvert Cliffs Unit 2 0.2 1 0.06 --- --- - --

101 37 Millstone Unit 2 0.2 1 0.06 --- --- ---

J.M. Farley Unit 2 BOLA 0.028 0.89 --- --- ---

Notes on following page.

WCAP-1 8169-NP June2018 Revision I

CALC-AN02-EP-17-00002 Rev O Page 29 of 88 Westinghouse Non-Proprietary Class 3 3-3 Notes:

(a) The reactor vessel plate and weld material identification and heat numbers were taken from the Arkansas Nuclear One Unit 2 Certified Material Test Reports (CMTRs) and/or A-PENG-ER-002 [Ref. 11], unless otherwise noted.

(b) All chemistry values obtained from A-PENG-ER-002 and/or the Arkansas Nuclear One Unit 2 CMTRs, unless otherwise noted. Chemistry values for plates are the average of all available analyses. Chemistry values for welds are the average of all coated electrode deposit chemistry (CEDC) for the E-8018 stick electrodes or weld flux deposit chemistry (WFDC) for Linde 0091 welds, unless otherwise noted. Where avai lable, additional chemistry analysis results from BMI-0584 [Ref. 12]

were also included in the average. The chemistry values for the beltline plates reported in this table are identical to those previously reported in the Arkansas Nuclear One Unit 2 License Renewal Application (LRA) [Ref. 13] and the previous capsule report, BAW-2399, Revision I [Ref. 14].

(c) The RT NOT(UJ values for the plates are based on drop-weight data, longitudinally-oriented Charpy V-notch test data and NUREG-0800, BTP 5-3 Position 1.1(3)(a) and (b) [Ref. 10], with the more limiting RTNoT(UJ value being selected, un less otherwise noted. The RT NOT(UJ values for welds are the generic value for Linde 0091 flux type welds (-56°F) per IO CFR 50.61 [Ref. 15], unless otherwise noted.

(d) The material Heat numbers for the Intermediate Shell Longitudinal Welds 2-203A, B, & Care unclear in the historical data.

For conservatism, the material properties for the Intermediate Shell Longitudinal Welds 2-203A, B, & C reported are the most limiting values from welds relevant to the Intermediate Shell Longitudinal Welds 2-203A, B, & C per A-PENG-ER-002. These welds include Heat# 10120, Flux Type Linde 0091 , Lot# 3999 (sister plant weld), Heat# 10120, 10120, Heat#

AAGC, and the analysis of the in-process weld deposit chemistry.

(e) The Cu and Ni wt. % values for the Lower Shell Longitudinal Welds 3-203A, B, & C and the Intermediate to Lower Shell Girth Weld 9-203 are consistent with BAW-2399, Revision I [Ref. 14] and were originally taken from CE NPSD-1039, Revision 2 [Ref. 16].

(t) Weld Heat# 6329637 does not contain any WFDC Ni wt.% values, thus the bare wire chemical analysis (BWCA) value of 0.11 % from A-PENG-ER-002 [Ref. 11 ] was used.

(g) This RT NOT(U) value for the surveillance plate, Intermediate Shell Plate C-8009-3 , is based on drop-weight data, transverse orientation Charpy V-notch test data taken from the baseline capsule test report, TR-MCD-002 [Ref. 17] and ASME Code Section Ill Subarticle NB-2331 [Ref. 9]. The RT NOT(UJ value for the two weld materials, Heat numbers 83650 and BOLA, is based on drop-weight data, Charpy V-notch test data taken from A-PENG-ER-002 and ASME Code Section Ill Subarticle NB-233 1 [Ref. 9].

(h) Drop-weight test data is not available for this E-8018 weld heat. Therefore, to assign an RT NOT(UJ value to this E-8018 stick electrode weld, the data in Table 8 of A-PENG-ER-002 was analyzed. The average T NOT value for the 17 E-8018 weld heats is -57"F with a standard deviation of 16.5"F. This yields a bounding value of -24"F using a mean plus two sigma model; therefore, a value of -24"F is acceptable for the initial RT NOT value of this weld material, with consideration that its Charpy impact energy at IO"F, which is less than TNoT + 60"F, was greater than 100 ft-lb. Furthermore, -24°F bounds all of the E-8018 stick electrode T NOT values present in A-PENG-ER-002 [Ref. 11).

(i) Surveillance data exists for weld Heat# 83650, # IO 137, and# BOLA from multiple sources; see Section 4 for more detai ls.

The data for Arkansas Nuclear One Unit 2 weld metal Heat # 83650 was taken as the average of the data available from TR-MCD-002, as well as the subsequent analyses completed during testing of the first capsule, BMl-0584 [Ref. 12]. The data for the Calvert Cliffs Unit 2 weld metal Heat # IO 137 was taken from Table 4-3 of WCAP-1750 I-NP [Ref. 18). The data for the Millstone Unit 2 weld metal Heat# 10137 was taken from Table 4-1 ofWCAP-16012 [Ref. 19]. The data for the J.M.

Farley Unit 2 weld metal Heat # BOLA was taken from Table 4-1 of WCAP-16918, Revision I (Ref. 20].

U) Upper Shell Plate C-8008-1 shares the same material heat number as the Arkansas Nuclear One Unit 2 surveillance plate material, Intermediate Shell Plate C-8009-3; therefore, surveillance program test results apply to this material as well.

WCAP-18169-NP June 2018 Revision I

CALC-AN02-EP-17-00002 Rev O Page 30 of 88 Westinghouse Non-Proprietary Class 3 3-4 Table 3-2 Summary of Arkansas Nuclear One Unit 2 Reactor Vessel Closure Heads, Vessel Flange and Balance of RCS Initial RT NDT Values Initial RT NDT Reactor Vessel Material Methodology (OF)

Current Closure Head 10 BTP 5-3, Paragraph B1.1(3)(a) and (b) [Ref. 10]

(Heat# 125B404)

Replacement Closure Head<*J ASME Code, Section III, Subsection NB-2300

-22 (Heat# R378 l/R3782) [Ref. 9]

Vessel Flange 30 BTP 5-3, Paragraph B1.1(3)(a) and (b) [Ref. 10]

(Heat# 122A440)

Balance of RcsCb> 50 Note (b)

Notes:

(a) A replacement, si ngle forging, closure head has been fabricated; however, it has not yet been installed at Arkansas Nuclear One Unit 2. The vessel flange initial RTNoT value is higher than both the current and replacement closure head initial RT NDT values. Thus, the results contained herein are conservative for the current and replacement closure heads.

(b) 50°F was conservatively assigned to all RCS material not specifically tested per Section 5.2.4.3 of the Arkansas Nuclear One Unit 2 updated Final Safety Analysis Report (FSAR).

WCAP-18169-NP June 2018 Revision 1

CALC-AN02-EP-17-00002 Rev O Page 31 of 88 Westinghouse Non-Proprietary Class 3 4- 1 4 SURVEILLANCE DATA Per Regulatory Guide 1.99, Revision 2 [Ref. 1], calculation of Position 2.1 chemistry factors requires data from the plant-specific surveillance program. In addition to the plant-specific surveillance data, data from surveillance programs at other plants which include a reactor vessel beltline or extended beltline material should also be considered when calculating Position 2.1 chemistry factors. Data from a surveillance program at another plant is often called 'sister plant' data.

The surveillance capsule plate material for Arkansas Nuclear One Unit 2 is from Intermediate Shell Plate C-8009-3 . Surveillance results from this plate also apply to Upper Shell Plate C-8008-1 , because the two plates were made from the same heat of material (Heat # C8182). The surveillance capsule weld material for Arkansas Nuclear One Unit 2 is Heat # 83650, which is applicable to the intermediate to lower shell girth weld. Table 4-1 summarizes the Arkansas Nuclear One Unit 2 surveillance data for the plate material and weld material (Heat # 83650) that will be used in the calculation of the Position 2.1 chemistry factor values for these materials. The results of the last withdrawn and tested surveillance capsule, Capsule 284°, were documented in WCAP-18166-NP [Ref. 21]. Appendix D of WCAP-18166-NP concluded that the surveillance plate and weld (Heat # 83650) data are credible; therefore, a reduced margin term will be utilized in the ART calculations contained in Section 7.

The Arkansas Nuclear One Unit 2 reactor vessel upper to intermediate shell girth weld seam was fabricated using weld Heat# 10137. Weld Heat # 10137 is contained in the Calvert Cliffs Unit 2 and Millstone Unit 2 surveillance programs. Thus, the Calvert Cliffs Unit 2 and Millstone Unit 2 data will be used in the calculation of the Position 2.1 chemistry factor value for Arkansas Nuclear One Unit 2 weld Heat # 10137. Note that no surveillance data is available for the other two Heats (# 6329637 and #

FAGA) which were also used to make the Upper to Intermediate Shell Girth Weld 8-203 . Table 4-2 summarizes the applicable surveillance capsule data pertaining to weld Heat # 10137. The combined surveillance data is deemed credible per Appendix D; however, as a result of the Mi llstone Unit 2 surveillance data including both weld Heat # 10137 and 90136, the Position 2.1 chemistry factor calculations for weld Heat# 10137 will utilize a full margin term for conservatism. See Appendix D for details.

The Arkansas Nuclear One Unit 2 reactor vessel upper shell longitudinal weld seams were fabricated using weld Heat # BOLA. Weld Heat # BOLA is contained in the J.M. Farley Unit 2 surveillance program. Thus, the J.M. Farley Unit 2 data will be used in the calculation of the Position 2.1 chemistry factor value for Arkansas Nuclear One Unit 2 weld Heat# BOLA. Table 4-3 summarizes the applicable surveillance capsule data pertaining to weld Heat # BOLA. Per Appendix D of WCAP-1 6918-NP, Revision 1 [Ref. 20], the J.M. Farley Unit 2 surveillance weld data is deemed non-credible. Since the J.M.

Farley Unit 2 surveillance weld is not analyzed with any additional surveillance capsule material herein, this credibility conclusion is applicable to the Arkansas Nuclear One Unit 2 weld Heat # BOLA.

Therefore, a full margin term will be utilized in the ART calculations contained in Section 7.

WCAP-18169-NP June2018 Revision 1

Westinghouse Non-Proprietary Class 3 4-2 Table 4-1 Arkansas Nuclear One Unit 2 Surveillance Capsule Data Capsule Fluence<*> Measured 30 ft-lb Transition Material Capsule(*>

(x 10 19 n/cm2, E > 1.0 MeV) Temperature Shift<*>(°F) 97° 0.303 23.5 Intermediate Shell Plate C-8009-3 (Longitudinal) 284° 3.67 85.7 97° 0.303 33.4 Intermediate Shell Plate C-8009-3 (Transverse) 104° 2.15 52.9 284° 3.67 85.6 (")

97° 0.303 13 .2 l>

r Surveillance Weld Material (Heat# 83650) (")

104° 2.15 16.1 )>

284° 3.67 12.0 z 0

No te: "'m I

(a) Surveillance data was taken from Table 5- 10 of WCAP-18166-NP [Ref. 21]. ~

I

-.,j c:,

0 0

0

0 CD 0

~

Ill IQ CD

(,)

"'0 co co WCAP-18169-NP June 2018 Revision 1

Westinghouse Non-Proprietary Class 3 4-3 Table 4-2 Calvert Cliffs Unit 2 and Millstone Unit 2 S urveillance Capsule Data for Weld Heat# 10137 Capsule Fluence<*> Measured 30 ft-lb Transition Inlet Temperature(bl Temperature Material Capsule<*>

(x 10 19 n/cm2, E > 1.0 MeV) Temperature Shift<*> (°F) {°F) Adjustment<<) (°F) 263 ° 0.825 72.7 550 -1.0 Calvert Cliffs Unit 2 97° 1.95 82.9 549 -2.0 Data 104° 2.44 69.7 548 -3.0 97° 0.324 65.93 544.3 -6.7 Millstone Unit 2 104° 0.949 52.12 547.6 -3.4 0 Data )>

83° 1.74 56.09 548.0 -3 .0 r 0

Notes : :i:,.

z (a) For surveillance weld Heat# 10137, data pertaining to Calvert Cliffs Unit 2 were taken from Table 5- 10 of WCAP- 17501-NP [Ref. 18]. Data pertaining to 0 N

Millstone Unit 2 were taken from Table 5-10 ofWCAP-16012 [Ref. 19). I (b) Inlet temperatures were calculated as the average inlet temperature from all the previously completed cycles at the time of capsule withdrawal.

m

"'CJ (c) Temperature adjustment= l .O*(Tcapsute - Tp1an1), where Tptan, = 55 l .0°F for Arkansas Nuclear One Unit 2 (applied to the weld ~TNDT data for each of the Calvert .....

I Cliffs Unit 2 and M illstone Unit 2 capsules in the Position 2.1 chemistry factor calculation - See Section 5 for more details). 0 I

0 0

0 Table 4-3 J .M. Farley U nit 2 S urveillance Capsule D ata for Weld Heat# BOLA N

(D Capsule Fluence<*> Measured 30 ft-lb Transition Inlet Temperature<hl Temperature 0 Material Capsule<*>

(x 10 19 n/cm2, E > 1.0 MeV) Temperature Shift<*> (0 F) {°F) Adjustment<c) (°F) "'CJ Ill IQ (D

u 0.605 -28.4 544 -7.0 w w

J.M. Farley Unit 2

w X

z 1.73 2.98 4.92 7.0

-15.6 10.2 542 543 543

-9.0

-8.0

-8.0 0

co co y 6.79 69.l 543 -8.0 V 8.73 56.5 542 -9.0 Notes:

(a) For surveillance weld Heat# BOLA, data pertaining to J.M. Farley Unit 2 were taken from Table 5-10 ofWCAP- 16918-NP, Revision I [Ref. 20].

(b) Inlet temperatures were calculated as the average inlet temperature from all the previously completed cycles at the time of capsule withdrawal.

(c) Temperature adjustment= J.O*(Tcapsule -Tp1an,), where Tplant = 551.0°F for Arkansas Nuclear One Unit 2 (applied to the weld li.RTNDT data for each of the J.M.

Farley Unit 2 capsules in the Posi tion 2.1 chemistry factor calculation - See Section 5 for more details).

WCAP-18169-NP June 2018 Revision I

CALC-AN02-EP-17-00002 Rev O Page 34 of 88 Westinghouse Non-Proprietary Class 3 5-1 5 CHEMISTRY FACTORS The chemistry factors (CFs) were calculated using Regulatory Guide 1.99, Revision 2, Positions 1.1 and 2.1. Position 1.1 chemistry factors for each reactor vessel material are calculated using the best-estimate copper and nickel weight percent of the material and Tables 1 and 2 of Regulatory Guide 1.99, Revision

2. The best-estimate copper and nickel weight percent values for the Arkansas Nuclear One Unit 2 reactor vessel materials are provided in Table 3-1 of this report.

The Position 2.1 chemistry factors are calculated for the materials that have available surveillance program results. The calculation is performed using the method described in Regulatory Guide 1.99, Revision 2. The Arkansas Nuclear One Unit 2 surveillance data as well as the applicable sister plant data was summarized in Section 4 of this report, and will be utilized in the Position 2.1 chemistry factor calculations in this Section.

The Position 2.1 chemistry factor calculations are presented in Tables 5-1 through 5-4 for Arkansas Nuclear One Unit 2 reactor vessel materials that have associated surveillance data. These values were calculated using the surveillance data summarized in Section 4 of this report. All of the surveillance data is adjusted for irradiation temperature and chemical composition differences in accordance with the guidance presented at an industry meeting held by the NRC on February 12 and 13, 1998 [Ref. 22].

Margin will be applied to the ART calculations in Section 7 according to the conclusions of the credibility evaluation for each of the surveillance materials, as documented in Section 4.

The Position 1.1 chemistry factors are summarized along with the Position 2.1 chemistry factors in Table 5-5 for Arkansas Nuclear One Unit 2.

WCAP-181 69-NP June 2018 Revision 1

CALC-AN02-EP-17 -00002 Rev O Page 35 of 88 Westinghouse Non-Proprietary Class 3 5-2 Table 5-1 Calculation of Arkansas Nuclear One Unit 2 Chemistry Factor Value for Intermediate Shell Plate C-8009-3 Using Surveillance Capsule Data (d)

Intermediate Shell (IS) Capsule J ARTNDT FF*ARTNDT Capsule FF<cl FF 2 Plate C-8009-3 Data<*> (x 10 19 n/cm2, E > 1.0 MeV) (OF) (OF) 97° 0.303 0.6728 23.5 15.8 1 0.453 Longitudinal Orientation 284° 3.67 1.3373 85.7 114.60 1.788 97° 0.303 0.6728 33.4 22.47 0 .453 Transverse Orientation 104° 2.15 1.2079 52.9 63 .90 1.459 284° 3.67 1.3373 85.6 114.47 1.788 SUM: 33 1.26 5.941 2

CF,s Plate C-8009-3 = L(FF

• L'1RT NOT)-;- L(FF ) = (33 1.26) -;- (5.94 1) = 55.8°F Notes:

(a) This surveillance data applies to both Intermediate Shell Plate C-8009-3 and Upper Shell Plate C-8008-1, since the two plates were made from the same heat of material (Heat# C8 l 82).

(b) f= fluence.

(c) FF= fluence factor = t<0 *2 s-o.io*iog ()_

(d) tiRTNDT values are the measured 30 ft-lb shift values. All values are taken from Table 4-1 of this report.

Table 5-2 Calculation of Arkansas Nuclear One Unit 2 Chemistry Factor Value for Weld Heat

1. 83650 Using Surveillance Capsule Data (c)

Weld Metal Capsule J<*l ARTNDT FF*ARTNDT Capsule FF(bl FF 2 19 Heat# 83650 (x 10 n/cm2, E > 1.0 MeV) (OF) (O F) 97° 0.303 0.6728 13.3 (13.2) 8.97 0.453 Arkansas Nuclear 104° 2.15 1.2079 16.3 (16.1) 19.64 1.459 One Unit 2 Data 284° 3.67 1.3373 12. 1 (12.0) 16.21 1.788 SUM: 44.82 3.700 2

CFweldHeat #83650= L(FF

• L'1RTNoT)-;- L(FF ) = (44.82)-;-(3.70) = 12.1°F Notes:

(a) f= fluence.

(b) FF= fluence factor = t<0 *2s- o.io*iosf)_

(c) tiRT NDT values are the measured 30 ft-lb shift values. The tiRT NDT values are adjusted using the ratio procedure to account for differences in the surveillance weld chemistry and the reactor vessel weld chemistry (pre-adj usted values are listed in parentheses and were taken from Table 4- 1 of this report). Ratio applied to the Arkansas N uclear One Unit 2 surveillance data = CF vessel Weld / Cf surv. Weld= 34. 1°F / 33.7°F = 1.0 I.

WCAP-18 169-NP June 20 18 Revision 1

_j

CALC-AN02-EP-17 -00002 Rev O Page 36 of 88 Westinghouse Non-Proprietary Class 3 5-3 Table 5-3 Calculation of Arkansas Nuclear One Unit 2 Chemistry Factor Value for Weld Heat# 10137 Using Surveillance Capsule Data (c)

Weld Metal Capsule t<*> L\.RTNDT FF*L\.RTNDT 2 Capsule FF<bl FF 19 Heat# 10137 (x 10 n/cmZ, E > 1.0 MeV) (OF) (OF) 263 ° 0.825 0.9460 73 .1 (72 .7) 69.19 0.895 Calvert Cliffs Unit 2 97° 1.95 1.1825 82.5 (82 .9) 97.58 1.398 Data 104° 2.44 1.2401 68.0 (69.7) 84.37 1.538 97° 0.324 0.6902 60.4 (65.93) 41.70 0.476 Millstone Unit 2 Data(dJ 104° 0.949 0.9853 49.7 (52 .12) 48.97 0.971 83° 1.740 1.1523 54.2 (56.09) 62.40 1.328 SUM: 404.20 6.606 2

CF weld Heat # 10137 = }.:(FF

• LlRT NDT).;.. l:(FF ) = (404.20).;.. (6.606) = 61.2°F Notes:

(a) f = tluence.

(b) FF = tluence factor = t<02 B-o.io*iog I)_

(c) ti.RT NDT values are the measured 30 ft-lb shift values. The ti.RT NDT values are adjusted first by the difference in operating temperature then using the ratio procedure to account for differences in the surveillance weld chemistry and the reactor vessel weld chemistry (pre-adjusted values are listed in parentheses and were taken from Table 4-2 of this report). The temperature adjustments are listed in Table 4-2. Ratio applied to the Calvert Cliffs Unit 2 surveillance data = CFvessel Weld / CFsurv. Weld = 98.5 °F / 96.8°F = 1.02. Ratio applied to the Millstone Unit 2 surveillance data

= CF vessel Weld I CFsurv. Weld = 98.5°F / 96.8°F = 1.02.

(d) Millstone Unit 2 surveillance data contains specimens from both weld Heat# 10137 and weld Heat # 90136. However, this inclusion of an additional heat is not expected to negatively impact the subsequent reactor vessel integrity calculation results, as additional conservatisms are in place. See Appendix D for more details.

WCAP-18169-NP June 2018 Revision 1

CALC-AN02-EP-17-00002 Rev O Page 37 of 88 Westinghouse Non-Proprietary Class 3 5-4 Table 5-4 Calculation of Arkansas Nuclear One Unit 2 Chemistry Factor Value for Weld Heat# BOLA Using Surveillance Capsule Data Weld Metal Capsule f' 3 > FF{bl ~RTNDT

{c)

FF*~RTNDT 2 Capsule FF Heat#BOLA (x 10 19 n/cm2, E > 1.0 MeV) (OF) (OF) u 0.605 0.8593 o.o<d> (-28.4) 0.00 0.738 w 1.73 1.1508 o.o<d) (7.0) 0.00 1.324 J.M. Farley X 2.98 1.2891 o.o<d) (-15.6) 0 .00 1.662 Unit2 z 4.92 1.3992 2.2 (10.2) 3 .08 1.958 y 6.79 1.4579 61.1 (69. 1) 89.08 2.125 V 8.73 1.4960 47 .5 (56 .5) 71.06 2.238 SUM: 163 .22 10.046 2

CF Heat # BOLA= I:(FF * ~RTN DT) I:(FF ) = (163 .22) -c- (10.046) = 16.2°F Notes:

(a) f= fluence.

(b) FF = fluence factor= t< 0*2s-O.IO'logfl_

(c) ~RTNDT values are the measured 30 ft-lb shift values. The ~RT NOT values are adjusted first by the difference in operating temperature then using the ratio procedure to account for differences in the surveillance weld chemistry and the reactor vessel weld chemistry (pre-adjusted values are listed in parentheses and were taken from Table 4-3 of this report). The temperature adj ustments are listed in Table 4-3. A ratio of 1.00 was conservatively applied to the J.M.

Farley Unit 2 surveillance data, since CF vessel Weld < CFsurv. Weld*

(d) A negative ~RTNOT value was calculated after temperature adj ustment. Physically, this should not occur; thus a conservative value of0.0°F was used.

WCAP-18169-NP June2018 Revision 1

CALC-AN02-EP-17-00002 Rev O Page 38 of 88 Westinghouse Non-Proprietary Class 3 5-5 Table 5-5 Summary of Arkansas Nuclear One Unit 2 Positions 1.1 and 2.1 Chemistry Factors Reactor Vessel Material Chemistry Factor (°F)

Heat Number and Identification Number Position 1.1 <*> Position 2.1 Reactor Vessel Beltline Materials Intermediate Shell Plate C-8009-1 C8161-3 63.6 ---

Intermediate Shell Plate C-8009-2 C8161-l 54.5 ---

Intermediate Shell Plate C-8009-3 C8182-2 62.2 55.8(b)

Lower Shell Plate C-8010-1 C8161-2 54.5 ---

Lower Shell Plate C-8010-2 B2545-l 53.1 ---

Lower Shell Plate C-8010-3 B2545-2 51.0 ---

Intermediate Shell Longitudinal Welds Multiple 68.0 ---

2-203A, B, & C Lower Shell Longitudinal Welds 3-203A, B, & C 10120 34.0 ---

Intermediate to Lower Shell Girth Weld 9-203 83650 34. 1 12.1 (c)

Reactor Vessel Extended Beltline Materials Upper Shell Plate C-8008-1 C8182-l 91.0 55.8(b)

Upper Shell Plate C-8008-2 C7605-l 89.5 - --

Upper Shell Plate C-8008-3 C8571-2 51.0 ---

Upper Shell Longitudinal Welds BOLA 27.0 16.2(d) l-203A, B, & C 10137 98.5 6 l .2(e)

Upper to Intermediate Shell Girth Weld 8-203 6329637 100.8 ---

FAGA 41.0 ---

Surveillance Weld Data Arkansas Nuclear One Unit 2 83650 33.7 -- -

Calvert Cliffs Unit 2 96.8 ---

10137 Millstone Unit 2 96.8 ---

J.M. Farley Unit 2 BOLA 38.2 ---

Notes:

(a) Position I.I chemistry factors were calculated using the copper and nickel weight percent values presented in Table 3-1 of this report and Tables I and 2 of Regulatory Guide 1.99, Revision 2.

(b) Position 2.1 chemistry factor was taken from Table 5-1 of this report. As discussed in Section 4, the surveillance plate data is credible and applicable to both Intermediate Shell Plate C-8009-3 and Upper Shell Plate C-8008-1 .

(c) Position 2.1 chemistry factor was taken from Table 5-2 of this report. As discussed in Section 4, the surveillance weld data for Heat # 83650 is credible.

(d) Position 2.1 chemistry factor was taken from Table 5-4 of this report. As discussed in Section 4, the surveillance weld data for Heat# BOLA is not credible.

(e) Position 2.1 chemistry factor was taken from Table 5-3 of this report. As discussed in Section 4, the surveillance weld data for Heat # l O137 is credible; however no reduction in the margin term wi ll be taken.

WCAP-18169-NP June2018 Revision 1

CALC-AN02-EP-17-00002 Rev O Page 39 of 88 Westinghouse Non-Proprietary Class 3 6-1 6 CRITERIA FOR ALLOWABLE PRESSURE-TEMPERATURE RELATIONSHIPS 6.1 OVERALL APPROACH The ASME approach for calculating the allowable limit curves for various heatup and cooldown rates specifies that the total stress intensity factor, K1, for the combined thermal and pressure stresses at any time during heatup or cooldown cannot be greater than the reference stress intensity factor, K,c, for the metal temperature at that time. K1c is obtained from the reference fracture toughness curve, defined in the 1998 Edition through the 2000 Addenda of Section XI, Appendix G of the ASME Code [Ref. 3]. The K1c curve is given by the following equation:

K le =33 .2+ 20 .734 *e[0 .02 (T - RTNDT )] (1)

where, K 1c (ksi-Vin.) reference stress intensity factor as a function of the metal temperature T and the metal reference nil-ductility temperature RT NDT This K1c curve is based on the lower bound of static critical K 1 values measured as a function of temperature on specimens of SA-533 Grade B Class 1, SA-508-1, SA-508-2, and SA-508-3 steel.

6.2 METHODOLOGY FOR PRESSURE-TEMPERATURE LIMIT CURVE DEVELOPMENT The governing equation for the heatup-cooldown analysis is defined in Appendix G of the ASME Code as follows :

(2)

where, stress intensity factor caused by membrane (pressure) stress stress intensity factor caused by the thermal gradients reference stress intensity factor as a function of the metal temperature T and the metal reference nil-ductility temperature RTN DT C 2.0 for Level A and Level B service limits C 1.5 for hydrostatic and leak test conditions during which the reactor core is not critical WCAP-181 69-NP June 2018 Revision I J

CALC-AN02-EP-17-00002 Rev O Page 40 of 88 Westinghouse Non-Proprietary Class 3 6-2 For membrane tension, the corresponding K 1 for the postulated defect is:

Kim= Mmx(pR / t) (3) where, Mm for an inside axial surface flaw is given by:

Mm 1.85 for Ji < 2, Mm 0.926 Ji for 2::; Ji ::; 3.464, Mm 3.2 1 for Ji > 3.464 and, Mm for an outside axial surface flaw is given by:

Mm 1. 77 for Ji < 2, Mm 0.893 Ji for 2::; Ji ::; 3 .464, Mm 3.09 for Ji > 3.464 Similarly, Mm for an inside or an outside circumferential surface flaw is given by:

Mm 0.89 for Ji < 2, Mm 0.443 Ji for 2::; Ji ::; 3 .464 ,

Mm 1.53 for Ji > 3.464 Where:

p = internal pressure (ksi), Ri = vessel inner radius (in), and t = vessel wall thickness (in).

For bending stress, the corresponding K 1 for the postulated axial or circumferential defect is:

K 1b = Mb

• Maximum Stress, where Mb is two-thirds of Mm (4)

The maximum K 1 produced by radial thermal gradient for the postulated axial or circumferential inside surface defect ofG-2120 is:

K11 = 0.953x10-3 x CR x t2*5 (5) where CR is the cooldown rate in °F/hr., or for a postulated axial or circumferential outside surface defect K11 = 0.753x10-3 x HU x t25 (6) where HU is the heatup rate in °F /hr.

WCAP-18169-NP June2018 Revis ion 1

CALC-AN02-EP-17-00002 Rev O Page 41 of 88 Westinghouse Non-Proprietary Class 3 6-3 The through-wall temperature difference associated with the maximum thermal K 1 can be determined fromASME Code, Section XI, Appendix G, Fig. G-2214-1. The temperature at any radial distance from the vessel surface can be determined from ASME Code, Section XI, Appendix G, Fig. G-2214-2 for the maximum thermal K 1*

(a) The maximum thermal K 1 relationship and the temperature relationship in Fig. G-2214-1 are applicable only for the conditions given in G-2214.3(a)(l) and (2).

(b) Alternatively, the K 1 for radial thermal gradient can be calculated for any thermal stress distribution and at any specified time during cooldown for a 11,i-thickness axial or circumferential inside surface defect using the relationship :

!01 = (1.0359Co + 0.6322Ci + 0.4753C2 + 0.3855C3) * .fi;. (7) or similarly, Ku during heatup for a 11,i-thickness outside axial or circumferential surface defect using the relationship:

K1t = (l .043Co + 0.630C, + 0.481C2 + 0.401C3) * .f;; (8) where the coefficients C 0 , C 1, C 2 and C3 are determined from the thermal stress distribution at any specified time during the heatup or cooldown using the form:

o-(x) = Co+ C,(x I a)+ C2(x I a)2 + C3(x I a)3 (9) and x is a variable that represents the radial distance (in) from the appropriate (i.e., inside or outside) surface to any point on the crack front, and a is the maximum crack depth (in).

Note that Equations 3, 7, and 8 were implemented in the OPERLIM computer code, which is the program used to generate the pressure-temperature (P-T) limit curves. The P-T curve methodology is the same as that described in WCAP-14040-A, Revision 4, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves" [Ref. 2] Section 2.6 (equations 2.6.2-4 and 2.6.3-1 ). Finally, the reactor vessel metal temperature at the crack tip of a postulated flaw is determined based on the methodology contained in Section 2.6.1 of WCAP-14040-A, Revision 4 (equation 2.6.1-1 ). This equation is solved utilizing values for thermal diffusivity of 0.518 ft2/hr at 70°F and 0.379 ft 2/hr at 550°F and a constant convective heat-transfer coefficient value of 7000 Btu/hr-ft2 -°F.

At any time during the heatup or cooldown transient, K1c is determined by the metal temperature at the tip of a postulated flaw (the postulated flaw has a depth of 1/4 of the section thickness and a length of 1.5 times the section thickness per ASME Code, Section XI, Paragraph G-2120), the appropriate value for RT NOT, and the reference fracture toughness curve (Equation 1). The thermal stresses resulting from the temperature gradients through the vessel wall are calculated and then the corresponding (thermal) stress WCAP-18169-NP June 2018 Revision I

CALC-AN02-EP-1 7-00002 Rev O Page 42 of 88 Westinghouse Non-Proprietary Class 3 6-4 intensity factors, K 1i, for the reference flaw are computed. From Equation 2, the pressure stress intensity factors are obtained and, from these, the allowable pressures are calculated.

For the calculation of the allowable pressure versus coolant temperature during cooldown, the reference 1/4T flaw of Appendix G to Section XI of the ASME Code is assumed to exist at the inside of the vessel wall. During cooldown, the controlling location of the flaw is always at the inside of the vessel wall because the thermal gradients, which increase with increasing cooldown rates, produce tensile stresses at the inside surface that would tend to open (propagate) the existing flaw. Allowable pressure-temperature curves are generated for steady-state (zero-rate) and each finite cooldown rate specified. From these curves, composite limit curves are constructed as the minimum of the steady-state or finite rate curve for each cooldown rate specified.

The use of the composite curve in the cooldown analysis is necessary because control of the cooldown procedure is based on the measurement of reactor coolant temperature, whereas the limiting pressure is actually dependent on the material temperature at the tip of the assumed flaw. During cooldown, the 1/4T vessel location is at a higher temperature than the fluid adjacent to the vessel inner diameter. This condition, of course, is not true for the steady-state situation. It follows that, at any given reactor coolant temperature, the ~T (temperature) across the vessel wall developed during cooldown results in a higher value of Kie at the 1/4T location for finite cooldown rates than for steady-state operation. Furthermore, if conditions exist so that the increase in K1c exceeds K1t, the calculated allowable pressure during cooldown will be greater than the steady-state value.

The above procedures are needed because there is no direct control on temperature at the 1/4T location and, therefore, allowable pressures could be lower if the rate of cooling is decreased at various intervals along a cooldown ramp. The use of the composite curve eliminates this problem and ensures conservative operation of the system for the entire cooldown period.

Three separate calculations are required to determine the limit curves for finite heatup rates. As is done in the cooldown analysis, allowable pressure-temperature relationships are developed for steady-state conditions as well as finite heatup rate conditions assuming the presence of a 1/4T defect at the inside of the wall. The heatup results in compressive stresses at the inside surface that alleviate the tensile stresses produced by internal pressure. The metal temperature at the crack tip lags the coolant temperature; therefore, the K 1e for the inside 1/4T flaw during heatup is lower than the K1c for the flaw during steady-state conditions at the same coolant temperature. During heatup, especially at the end of the transient, conditions may exist so that the effects of compressive thermal stresses and lower K1c values do not offset each other, and the pressure-temperature curve based on steady-state conditions no longer represents a lower bound of all similar curves for finite heatup rates when the 1/4T flaw is considered. Therefore, both cases have to be analyzed in order to ensure that at any coolant temperature the lower value of the allowable pressure calculated for steady-state and finite heatup rates is obtained.

The third portion of the heatup analysis concerns the calculation of the pressure-temperature limitations for the case in which a l/4T flaw located at the 1/4T location from the outside surface is assumed. Unlike the situation at the vessel inside surface, the thermal gradients established at the outside surface during heatup produce stresses which are tensile in nature and therefore tend to reinforce any pressure stresses present. These thermal stresses are dependent on both the rate of heatup and the time (or coolant WCAP-18169-NP June 2018 Revision I

CALC-AN02-EP-17-00002 Rev O Page 43 of 88 Westinghouse Non-Proprietary Class 3 6-5 temperature) along the heatup ramp. Since the thermal stresses at the outside are tensile and increase with increasing heatup rates, each heatup rate must be analyzed on an individual basis.

Following the generation of pressure-temperature curves for the steady-state and finite heatup rate situations, the final limit curves are produced by constructing a composite curve based on a point-by-point comparison of the steady-state and finite heatup rate data. At any given temperature, the allowable pressure is taken to be the least of the three values taken from the curves under consideration. The use of the composite curve is necessary to set conservative heatup limitations because it is possible for conditions to exist wherein, over the course of the heatup ramp, the controlling condition switches from the inside to the outside, and the pressure limit must at all times be based on analysis of the most critical criterion.

6.3 CLOSURE HEADNESSEL FLANGE REQUIREMENTS 10 CFR Part 50, Appendix G [Ref. 4] addresses the metal temperature of the closure head flange and vessel flange regions. This rule states that the metal temperature of the closure head regions must exceed the material unirradiated RT NDT by at least l 20°F for normal operation when the pressure exceeds 20 percent of the preservice hydrostatic test pressure, which is calculated to be 622 psig. The initial RT NOT values of the reactor vessel closure head, replacement reactor vessel closure head, and vessel flange are documented in Table 3-2. The limiting unirradiated RT NOT of 30°F is associated with the vessel flange of the Arkansas Nuclear One Unit 2 vessel, so the minimum allowable temperature of this region is l 50°F at pressures greater than 622 psig (without margins for instrument uncertainties). This limit is shown in Figures 8-1 and 8-2.

6.4 LOWEST SERVICE TEMPERATURE REQUIREMENTS The lowest service temperature (LST) is the minimum allowable temperature at which pressure can exceed 20% of the pre-service hydrostatic test pressure (3110 psig). This temperature is defined by Paragraphs NB-3211 and NB-2332 of ASME Code Section lil [Ref. 9] as the most limiting RTNOT for the balance of the RCS components plus 100°F. The balance of the reactor coolant system components includes consideration of the ferritic materials outside the reactor vessel cylindrical shell beltline, nozzle comer (see Appendix B), closure head, and vessel flange regions, but within the primary system. Per Table 3-2, the most limiting RT NDT for the balance of RCS is 50°F. Therefore, without margins for instrument errors, the LST for Arkansas Nuclear One Unit 2 is 150°F. For Arkansas Nuclear One Unit 2, this limit is identical to the vessel flange limit described in Section 6.3 and is shown in Figures 8-1 and 8-2.

6.5 BOLTUP TEMPERATURE REQUIREMENTS The minimum boltup temperature is the minimum allowable temperature at which the reactor vessel closure head bolts can be preloaded. It is determined by the highest reference temperature, RT NOT, in the closure flange region. This requirement is established in Appendix G to 10 CFR 50 [Ref. 4]. Per the NRC-approved methodology in WCAP-14040-A, Revision 4 [Ref. 2], the minimum boltup temperature should be 60°F or the limiting unirradiated RT NDT of the closure flange region, whichever is higher. Since the limiting unirradiated RT NDT of this region is below 60°F per Table 3-2 , the minimum boltup temperature for the Arkansas Nuclear One Unit 2 reactor vessel is 60°F (without margins for instrument uncertainties). This limit is shown in Figures 8-1 and 8-2.

WCAP-18169-NP June2018 Revision I

CALC-AN02-EP-17-00002 Rev O Page 44 of 88 Westinghouse Non-Proprietary Class 3 7-l 7 CALCULATION OF ADJUSTED REFERENCE TEMPERATURE From Regulatory Guide 1.99, Revision 2, the adjusted reference temperature (ART) for each material in the beltline region is given by the following expression:

ART = Initial RT NDT + Lill.T NDT + Margin (10)

Initial RTNDT is the reference temperature for the unirradiated material as defined in Paragraph NB-2331 of Section III of the ASME Boiler and Pressure Vessel Code [Ref. 9]. If measured values of the initial RTNDT for the material in question are not available, generic mean values for that class of material may be used, provided if there are sufficient test results to establish a mean and standard deviation for the class.

Lill.TNDT is the mean value of the adjustment in reference temperature caused by irradiation and should be calculated as follows:

~RTNDT =CF* f (0.28 -0.IO!ogf) (11)

To calculate Lill.T NDT at any depth (e.g., at 1/4T or 3/4T), the following formula must first be used to attenuate the fluence at the specific depth.

£(depth x) = Lsu-'ace

• e (-O.Z 4x) (12) where x inches (reactor vessel cylindrical shell beltline thickness is 7.875 inches) is the depth into the vessel wall measured from the vessel clad/base metal interface. The resultant fluence is then placed in Equation 11 to calculate the Lill.T NDT at the specific depth.

The projected reactor vessel neutron fluence was updated for this analysis and documented in Section 2 of this report. The evaluation methods used in Section 2 are consistent with the methods presented in WCAP-14040-A, Revision 4, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves" [Ref. 2].

Table 7-1 contains the surface fluence values at 54 EFPY, which were used for the development of the P-T limit curves contained in this report. Table 7-1 also contains the 1/4T and 3/4T calculated fluence values and fluence factors (FFs), per Regulatory Guide 1.99, Revision 2 [Ref. l]. The values in this table will be used to calculate the 54 EFPY ART values for the Arkansas Nuclear One Unit 2 reactor vessel materials.

Margin is calculated as M = 2 ~ cr; + cr ! . The standard deviation for the initial RT NDT margin term (cri) is 0°F when the initial RTNDT is a measured value, and l 7°F when a generic value is available. The standard deviation for the ~RT NDT margin term, a,.., is l 7°F for plates or forgings when surveillance data is not used or is non-credible, and 8.5°F (half the value) for plates or forgings when credible surveillance data is used. For welds, a,.. is equal to 28°F when surveillance capsule data is not used or is non-credible, and is 14°F (half the value) when credible surveillance capsule data is used. The value for cr,.. need not exceed 0.5 times the mean value of ~RTNDT*

WCAP-181 69-NP June2018 Revision l

CALC-AN02-EP-17-00002 Rev O Page 45 of 88 Westinghouse Non-Proprietary Class 3 7-2 Contained in Tables 7-2 and 7-3 are the 54 EFPY ART calculations at the 1/4T and 3/4T locations for generation of the Arkansas Nuclear One Unit 2 heatup and cooldown curves.

The inlet and outlet nozzle forging materials for Arkansas Nuclear One Unit 2 have projected fluence values that do not exceed the 1 x 10 17 n/cm2 fluence threshold at 54 EFPY per Table 2-2 at the lowest extent of the nozzle; therefore, per NRC RIS 2014-11 [Ref. 8], neutron radiation embrittlement need not be considered herein for these materials. Thus, ART calculations for the inlet and outlet nozzle forging materials utilizing the 1/4T and 3/4T fluence values are excluded from Tables 7-2 and 7-3 , respectively.

Limiting ART values for the nozzle materials are contained in Appendix B.

The limiting ART values for Arkansas Nuclear One Unit 2 to be used in the generation of the P-T limit curves are based on Lower Shell Plate C-8010-1 (Position 1.1). In order to provide an additional margin of conservatism, the limiting calculated ART values were rounded up and increased by 5°F. The increased limiting ART values, using the "Axial Flaw" methodology, for Lower Shell Plate C-8010-1 are summarized in Table 7-4.

WCAP-18169-NP June 2018 Revision l

CALC-AN02-EP-17-00002 Rev O Page 46 of 88 Westinghouse Non-Proprietary Class 3 7-3 Table 7-1 Fluence Values and Fluence Factors for the Vessel Surface, 1/4T and 3/4T Locations for the Arkansas Nuclear One Unit 2 Reactor Vessel Materials at 54 EFPY Surface Fluence, 1/4T f 3/4T f 1/4T 3/4T Reactor Vessel Region f <*> (n/cm2, (n/cm2, (n/cm2, FF FF E > 1.0 MeV) E > 1.0 MeV) E > 1.0 MeV)

Reactor Vessel Beltline Materials Intermediate Shell Plates 4.91 X 10 19 3.06 X 10 19 1.2955 1.19 X 10 19 1.0485 Lower Shell Plates 4.98 X 10 19 3.10 X 10 19 1.2988 1.21 X 10 19 1.0524 Intermediate Shell Longitudinal 4.64 X 10 19 2.89 X 10 19 1.2820 1.12 X 10 19 1.0327 Welds Lower Shell Longitudinal 4.71 X 10 19 2.94 X 10 19 l.2856 1.14 X 10 19 1.0369 Welds Intermediate to Lower Shell 4.89 X 10 19 3.05 X 10 19 1.2945 1.18 X 10 19 l.0474 Girth Weld Reactor Vessel Extended Beltline Materials Upper Shell Plates 5.89 X 10 17 3.67 X 10 17 0.2467 1.43 X 10 17 0.1389 Upper Shell Longitudinal Welds 5.89 X 10 17 3.67 X 10 17 0.2467 1.43 X 10 17 0.1389 Upper to Intermediate Shell 5.89 X 10 17 3.67 X 10 17 0.2467 1.43 X 10 17 0.1389 Girth Weld Note:

(a) 54 EFPY fluence values are documented in Table 2-2 .

WCAP-18169-NP June 2018 Revision 1

Westinghouse Non-Proprietary Class 3 7-4 Table 7-2 Adjusted Reference Temperature Evaluation for the Arkansas Nuclear One Unit 2 Reactor Vessel Beltline Materials through 54 EFPY at the 1/4T Location Reactor Vessel Material and ID Heat CF l/4T Fluence l/4T RT NDT(U)(a) ARTNDT a/*> (JA(b) Margin ART<c>

Number Number {°F) (n/cm2, E > 1.0 MeV) FF (OF) (OF) (OF) (OF) (OF) (OF)

Reactor Vessel Beltline Materials Intermediate Shell Plate C-8009-1 C8161 -3 63 .6 3.06 X 10 19 1.2955 -1.4 82.4 0.0 17.0 34.0 115.0 19 Intermediate Shell Plate C-8009-2 C8161 -l 54.5 3.06 X 10 1.2955 0.5 70.6 0.0 17.0 34.0 l 05.1 Intermediate Shell Plate C-8009-3 C8182-2 62.2 3.06 X 10 19 1.2955 0.0 80.6 0.0 17.0 34.0 114.6 Using credible Arkansas Nuclear C')

C8182-2 55.8 3.06 X 10 19 1.2955 0.0 72.3 0.0 8.5 17.0 89.3 )>

One Unit 2 surveillance data r C')

Lower Shell Plate C-8010-1 C8161-2 54.5 3.10 X 10 19 1.2988 12.0 70.8 0.0 17.0 34.0 116.8 )>

Lower Shell Plate C-8010-2 B2545 -l 53. l 3.10 X 10 19 1.2988 -16 .7 69.0 0.0 17.0 34.0 86.3 z 0

Lower Shell Plate C-80 l 0-3 B2545-2 51.0 3.10 X 10 19 1.2988 -22 .6 66.2 0.0 17.0 34.0 77.6 "',,ri, Intermediate Shell Longitudinal .....

I Multiple 68.0 2.89 X 10 19 1.2820 -56 87.2 17.0 28.0 65.5 96.7 "'-I Welds 2-203A, B, & C 6

Lower Shell Longitudinal Welds 0 19 0 10120 34.0 2.94 X 10 1.2856 -56 43.7 17.0 21.9 55.4 43.1 0 3-203A, B, & C

c Intermediate to Lower Shell Girth 19 (!)

83650 34.1 3.05 X 10 1.2945 -40 44 .1 0.0 22.1 44.1 48.3 <

Weld 9-203 Using credible Arkansas Nuclear 19 0

Cl 83650 12.1 3.05 X 10 1.2945 -40 15.7 0.0 7.8 15.7 -8.7 (C One Unit 2 surveillance data (!)

~

Reactor Vessel Extended Beltline Materials "'-I Upper Shell Plate C-8008-1 C8182-1 91.0 0.0367 X 10 19 0.2467 12.2 22.5 0.0 11.2 22.5 57.1 0

co co Using credible Arkansas Nuclear C8182-l 55.8 0.0367 X 10 19 0.2467 12.2 13 .8 0.0 6.9 13.8 39.7 One Unit 2 surveillance data Upper Shell Plate C-8008-2 C7605- l 89.5 0.0367 X 10 19 0.2467 60.5 22.1 0.0 11.0 22.1 104.7 19 Upper Shell Plate C-8008-3 C8571 -2 51.0 0.0367 X 10 0.2467 27.3 12.6 0.0 6.3 12.6 52.5 Upper Shell Longitudinal Welds BOLA 27.0 0.0367 X ] 0 19 0.2467 -60 6.7 0.0 3.3 6.7 -46.7 1-203A, B, & C Using non-credible J.M Farley BOLA 16.2 0.0367 X 10 19 0.2467 -60 4.0 0.0 2.0 4.0 -52.0 Unit 2 surveillance data WCAP-18169-NP June 2018 Revision 1

Westinghouse Non-Proprietary Class 3 7-5 Table 7-2 Adjusted Reference Temperature Evaluation for the Arkansas Nuclear One Unit 2 Reactor Vessel Beltline Materials through 54 EFPY at the l/4T Location Reactor Vessel Material and ID Heat CF I/4T Fluence 1/4T RTNDT(U)

(a)

ARTNDT e1/*l G,1(b) Margin ART<c>

Number Number (OF) (n/cm2, E > 1.0 MeV) FF (OF) {°F) {°F) (OF) {°F) (OF) 10137 98.5 0.0367 x 10 19 0.2467 -56 24.3 17.0 12.2 41.8 I 0.1 Using credible Calvert Cliffe Upper to Intermediate Shell Girth 61.2 0.0367 X 10 19 0.2467 -5 6 I 5. I I 7.0 28.0 65.5 24.6 Unit 2 and Weld 8-203 Millstone n

)>

Unit 2 data r 6329637 100.8 0.0367 X 10 19 0.2467 -56 24.9 17.0 12.4 42.1 I 1.0 n

)>

FAGA 41.0 0.0367 X 10 19 0.2467 -24 IO.I 0.0 5.1 10.l -3 .8 z 0

Notes: "'

(a) The plate material initial RT NDT values are measured values. For weld materials with generic initial RTNOT values, cr1 = l 7°F. For weld materials with measured initial RT NOT values, OJ = 0°F. "......

-...I I

I (b) As discussed in Section 4, the surveillance plate and weld Heat # 83650 data were deemed credible, while the weld Heat # BOLA data were deemed non-credible. The 0 0

surveillance weld data for Heat# IO 137 was deemed credible; however, per Section 4 and Appendix D, a full margin term will be used. Per the guidance of Regulatory Guide 0 0

1.99, Revision 2 [Ref. I], the base metal cr6 = l 7°F for Position 1.1 and cr6 = 8.5°F for Position 2.1 with credible surveillance data. The weld metal cr6 = 28°F for Position 1.1 "'

o and 2.1 with non-credible surveillance data (Heat # BOLA), and the weld metal cr6 = 14°F for Position 2.1 with credible surveillance data (Heat# 83650). Since a full margin (I) term will be used for Heat# IO 137, cr6 = 28°F with credible surveillance data for Position 2.1 for this weld heat. However, cr6 need not exceed 0.5*ll.RTNOT*

0 (c) The Regulatory Guide 1.99, Revision 2 methodology was used to calculate ART values. ART= RT NDT(U) + Li.RTNOT + Margin.

"II) cc (I)

""co0

-co co WCAP-1 8169-NP June 2018 Revision I

Westinghouse Non-Proprietary Class 3 7-6 Table 7-3 Adjusted Reference Temperature Evaluation for the Arkansas Nuclear One Unit 2 Reactor Vessel Beltline Materials through 54 EFPY at the 3/4T Location Reactor Vessel Material and ID Heat CF 3/4T Fluence RTNDT(U)

(*)

iiRTNDT o}*> Ga (b) Margin ART<c>

3/4T FF Number Number (OF) (n/cm2, E > 1.0 MeV) (OF) (OF) (OF) (OF) (OF) (OF)

Reactor Vessel Beltline Materials Intermediate Shell Plate C-8009-1 C8161-3 63.6 1.19 X 10 19 1.0485 -1.4 66.7 0.0 17.0 34.0 99.3 19 Intermediate Shell Plate C-8009-2 C8161 -1 54.5 l.19x 10 1.0485 0.5 57.1 0.0 17.0 34.0 91.6 19 Intermediate Shell Plate C-8009-3 C8182-2 62.2 1.19 X 10 1.0485 0.0 65 .2 0.0 17.0 34.0 99.2 Using credible Arkansas Nuclear 0 C8182-2 55.8 1.19 X 10 19 1.0485 0.0 58.5 0.0 8.5 17.0 75.5 )>

One Unit 2 surveillance data r-19 0 Lower Shell Plate C-80 I 0-1 C8161 -2 54.5 1.21 X 10 1.0524 12.0 57.4 0.0 17.0 34.0 103.4 )>

Lower Shell Plate C-8010-2 B2545-l 53.1 1.21 X 10 19 1.0524 -16.7 55.9 0.0 17.0 34.0 73 .2 z 0

Lower Shell Plate C-80 l 0-3 B2545-2 51.0 1.21 X 10 19 1.0524 -22.6 53.7 0.0 17.0 34.0 65 .1 "'m I

Intermediate Shell Longitudinal '"ti Multiple 68.0 1.12 X 10 19 1.0327 -56 70.2 17.0 28.0 65 .5 79.7 .....

I Welds 2-203A, B, & C -..I 6

Lower Shell Longitudinal Welds 0 10120 34.0 1.14 X 10 19 1.0369 -56 35.3 17.0 17.6 49.0 28.2 0 0

3-203A, B, & C

c Intermediate to Lower Shell Girth (!)

83650 34.1 l.18xl0 19 1.0474 -40 35.7 0.0 17.9 35.7 31.4 <

Weld 9-203 0 Using credible Arkansas Nuclear '"ti

~

83650 12.1 1.( 8 X 10 19 1.0474 -40 12.7 0.0 6.3 12 .7 -14.7 (C One Un it 2 surveillance data (!)

~

Reactor Vessel Extended Beltline Materials

<D 0

Upper Shell Plate C-8008-1 C8 182-1 91.0 0.0143 X 10 19 0.1389 12.2 12.6 0.0 6.3 12 .6 37.5 00 00 Using credible Arkansas Nuclear C8182-1 55.8 0.0143 X 10 19 0.1389 12.2 7.8 0.0 3.9 7.8 27.7 One Unit 2 surveillance data Upper Shell Plate C-8008-2 C7605-1 89.5 0.0143 X 10 19 0.1389 60.5 12.4 0.0 6.2 12.4 85.4 19 Upper Shell Plate C-8008-3 C8571 -2 51.0 0.0143 X I0 0.1389 27.3 7.1 0.0 3.5 7.1 41.5 Upper Shell Longitudinal Welds BOLA 27.0 0.0143 X 10 19 0.1389 -60 3.8 0.0 1.9 3.8 -52.5 1-203A, B, & C Using non-credible J.M Farley BOLA 16.2 0.0143 X 10 19 0.1389 -60 2.3 0.0 1.1 2.3 -55.5 Unit 2 surveillance data WCAP-1 8169-NP June 2018 Revision 1

Westinghouse Non-Proprietary Class 3 7-7 Table 7-3 Adjusted Reference Temperature Evaluation for the Arkansas Nuclear One Unit 2 Reactor Vessel Beltline Materials through 54 EFPY at the 3/4T Location Reactor Vessel Material and ID Heat CF 3/4T Fluence RTNDT(U)

(a)

ARTNDT (J?) <J,1(b) Margin ART<c>

3/4T FF Number Number (OF) (n/cm2, E > 1.0 MeV) (OF) (OF) (OF) (OF) (OF) {°F) 19 10137 98 .5 0.0143 X 10 0.1389 -56 13.7 17.0 6.8 36.6 -5.7 Using credible Calvert Cliffs Upper to Intermediate Shell Girth 61.2 0.0143 X 10 19 0.1389 -56 8.5 17.0 28.0 65.5 18.0 Unit 2 and Weld 8-203 Millstone 0

)>

Unit 2 data r-19 0 6329637 100.8 0.0143 10 0.1389 -56 14.0 17.0 7.0 36.8 -5 .2 FAGA 41.0 0.0143 X

X 10 19 0.1389 -24 5.7 0.0 2.8 5.7 -12.6 z

0 Notes: "'rn (a) The plate material initial RT NOT values are measured values. For weld materials with generic initial RT NOT values, cr1 = l 7°F. For weld materials with measured initial RTNDT values, cr 1 = 0°F.

...."ti......

I (b) As discussed in Section 4, the surveillance plate and weld Heat # 83650 data were deemed credible, while the weld Heat # BOLA data were deemed non-credible. The 6 0

surveillance weld data for Heat# 10137 was deemed credible; however, per Section 4 and Appendix D, a full margin term will be used. Per the guidance of Regulatory Guide 0 0

1.99, Revision 2 [Ref. I], the base metal cr6 = l 7°F for Position 1.1 and cr6 = 8.5°F for Position 2.1 with credible surveillance data. The weld metal cr6 = 28°F for Position 1.1 "'::tJ and 2.1 with non-credible surveillance data (Heat# BOLA), and the weld metal cr6 = 14°F for Position 2.1 with credible surveillance data (Heat# 83650). Since a full margin CD term will be used for Heat# IO 137, cr6 = 28°F with credible surveillance data for Position 2.1 for this weld heat. However, cr6 need not exceed 0.5*~RTNOT* 0 (c) The Regulatory Guide 1.99, Revision 2 methodology was used to calculate ART values. ART= RTNoT(U) + ~RTNoT + Margin. "ti II)

(C CD 0,

0

....000 00 WCAP-18169-NP June 2018 Revision 1

CALC-AN02-EP-17 -00002 Rev O Page 51 of 88 Westinghouse Non-Proprietary Class 3 7-8 Table 7-4 Summary of the Increased Limiting ART Values Used in the Generation of the Arkansas Nuclear One Unit 2 Heatup and Cooldown Curves at 54 EFPY l/4T Limiting ART<*> 3/4T Limiting ART<*>

122°F 109°F Lower Shell Plate C-8010-1 (Position 1.1)

Notes:

(a) The ART values used for P-T limit curve development in this report are the limiting ART values calculated in Tables 7-2 and 7-3 rounded up and increased by 5°F to add additional margin; this approach is conservative.

WCAP-181 69-NP June2018 Revision I

CALC-AN02-EP-17-00002 Rev O Page 52 of 88 Westinghouse Non-Proprietary Class 3 8-1 8 HEATUP AND COOLDOWN PRESSURE-TEMPERATURE LIMIT CURVES Pressure-temperature limit curves for normal heatup and cooldown of the primary reactor coolant system have been calculated for the pressure and temperature in the reactor vessel cylindrical beltline region using the methods discussed in Sections 6 and 7 of this report. This approved methodology is also presented in WCAP-14040-A, Revision 4 [Ref. 2].

Figure 8-1 presents the limiting heatup curves without margins for possible instrumentation errors using heatup rates of 50, 60, 70, and 80°F/hr applicable for 54 EFPY, with the flange and lowest service temperature requirements and using the "Axial Flaw" methodology. Figure 8-2 presents the limiting cooldown curves without margins for possible instrumentation errors using cooldown rates of 0, 25, 60, and 100°F/hr applicable for 54 EFPY, with the flange and lowest service temperature requirements and using the "Axial Flaw" methodology. The heatup and cooldown curves were generated using the 1998 through the 2000 Addenda ASME Code Section XI, Appendix G.

Allowable combinations of temperature and pressure for specific temperature change rates are below and to the right of the limit lines shown in Figures 8-1 and 8-2. This is in addition to other criteria, which must be met before the reactor is made critical, as discussed in the following paragraphs.

The reactor must not be made critical until pressure-temperature combinations are to the right of the criticality limit line shown in Figure 8-1 (heatup curve only). The first straight-line portion of the criticality limit is at the minimum permissible temperature for the 2485 psig inservice hydrostatic test as required by Appendix G to 10 CFR Part 50 (see Figure 8-3 and Table 8-3). The governing equation for the hydrostatic test is defined in the 1998 through the 2000 Addenda ASME Code Section XI, Appendix G as follows:

where, K 1m is the stress intensity factor covered by membrane (pressure) stress [see page 6-2, Equation (3)],

K,c = 33 .2 + 20.734 e [O.OZ(T- RTNoTll [see page 6-1 , Equation (l)],

T is the minimum permissible metal temperature, and RT NDT is the metal reference nil-ductility temperature.

The criticality limit curve specifies pressure-temperature limits for core operation in order to provide additional margin during actual power production. The pressure-temperature limits for core operation (except for low power physics tests) are that: 1) the reactor vessel must be at a temperature equal to or higher than the minimum temperature required for the inservice hydrostatic test, and 2) the reactor vessel must be at least 40°F higher than the minimum permissible temperature in the corresponding pressure-temperature curve for heatup and cooldown calculated as described in Section 6 of this report. For the heatup and cooldown curves without margins for instrumentation errors, the minimum temperature for the inservice hydrostatic test at 2485 psig for the Arkansas Nuclear One Unit 2 reactor vessel at 54 EFPY is WCAP-181 69-NP June20I8 Revision I

CALC-AN02-EP-17-00002 Rev O Page 53 of 88 Westinghouse Non-Proprietary Class 3 8-2 l 79°F. The vertical line drawn from these points on the pressure-temperature curve, intersecting a curve 40°F higher than the pressure-temperature limit curve, constitutes the limit for core operation for the reactor vessel.

Figures 8-1, 8-2, and 8-3 define all of the above limits for ensuring prevention of non-ductile failure for the Arkansas Nuclear One Unit 2 reactor vessel for 54 EFPY with the flange and lowest service temperature requirements and without instrumentation uncertainties. The data points used for developing the heatup and cooldown P-T limit curves shown in Figures 8-1 and 8-2 are presented in Tables 8-1 and 8-2. The data points used for developing the inservice hydrostatic and leak test P-T limit curve shown in Figure 8-3 are presented in Table 8-3. The P-T limit curves shown in Figures 8-1 , 8-2, and 8-3 were generated based on the limiting ART values for the cylindrical beltline and extended beltline reactor vessel materials rounded up and increased by 5°F to add additional margin; this approach is conservative.

As discussed in Appendix B, the P-T limits developed for the cylindrical beltline region bound the P-T limits for the reactor vessel inlet and outlet nozzles for Arkansas Nuclear One Unit 2 at 54 EFPY.

WCAP-18169-NP June 2018 Revision 1

CALC-AN02-EP-17-00002 Rev O Page 54 of 88 Westinghouse Non-Proprietary Class 3 8-3 MATERIAL PROPERTY BASIS LIMITING MATERIAL: Lower Shell Plate C-8010-1 using Regulatory Guide 1.99 Position 1.1 data LIMITING ART VALUES AT 54 EFPY: l /4T, 122°F (Axial Flaw) 3/4T, 109°F (Axial Flaw) 2500 '

loperlimAnalysis Version:5.4 Run:24301 Operlim .xlsm Version: 5.4.1 I II ,_

2250 2000

~ IUnacceptable I Ooeration I I ~

u '

IHeatup Rater 1750 50°F/Hr ~

L.---"' "' !Critical Limit!

(!)

en I

IHeatup Rater 60°F/Hr I I f ~

I 50°F/Hr I

a. 1500 I Q)

I,.

IHeatup Rate 70°F/Hr j' "'

!Critical Limitj 60°F/Hr t/j t/j Q) 1250

~ IHeatup RateV,...

80°F/Hr ,---

I i'...

_!Critical Limitl 70°F/Hr I,.

I I ll.

!Critical Limitl

-0

~

Q)

~ 80°F/Hr

, 1000

~

cu

(.) 11 Lowest Service ~

Temp.= 150°F I Acceptable I 750 Operation 500

+- Minimum Boltup Criticality Lim it based on Temo. = 60°F I+- - inservice hydrostatic test 250 temperature (179°F) for the service period up to 54 EFPY I I 0

I I I I 0 50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Deg. F)

Figure 8-1 Arkansas Nuclear One Unit 2 Reactor Coolant System Heatup Limitations (Heatup Rates of 50, 60, 70, and 80°F/hr) Applicable for 54 EFPY (with Flange and LST Requirements and without Margins for Instrumentation Errors) using the 1998 through the 2000 Addenda App. G Methodology (w/ K1c)

WCAP-18169-NP June 2018 Revision I

CALC-AN02-EP-17-00002 Rev O Page 55 of 88 Westinghouse Non-Proprietary Class 3 8-4 MATERJAL PROPERTY BASIS LIMITING MATERJAL: Lower Shell Plate C-8010-1 using Regulatory Guide 1.99 Position 1.1 data LIMITING ART VALUES AT 54 EFPY: l /4T, 122°F (Axial Flaw) 3/4T, 109°F (Axial Flaw) 2500 ' I loperlimAnalysis Version :5.4 Run:24301 Operlim .xlsm Version: 5.4.1 2250 2000 f-1 Unacceptable Operation I I

1750 J

(!)

en c... 1500 Q)

, I Ill Ill Q) c...

1250

~

"C Q)

~

, 1000 ~ - Cooldow n

(.)

cu

(.) I Lowest Service ~ " ~ Rates steady-state IAcceptable I Temp.= 150"F -25°F/Hr Operation

-60°F/H r 750 I

r.

500 ~ I Cooldown Ratel 11-100°F/Hr I I I I

250 Minimum

+-- Boltup  ! I I Temp.= 60' F  !

0 I

' I  !

0 50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Deg. F)

Figure 8-2 Arkansas Nuclear One Unit 2 Reactor Coolant System Cooldown Limitations (Cooldown Rates of O, 25, 60, and 100°F/hr) Applicable for 54 EFPY (with Flange and LST Requirements and without Margins for Instrumentation Errors) using the 1998 through the 2000 Addenda App. G Methodology (w/ K1c)

WCAP- 181 69-NP June2018 Revision 1

CALC-AN02-EP-17-00002 Rev O Page 56 of 88 Westinghouse Non-Proprietary Class 3 8-5 MATERIAL PROPERTY BASIS LIMITING MATERIAL: Lower Shell Plate C-8010-1 using Regulatory Guide 1.99 Position 1.1 data LIMITING ART VALUES AT 54 EFPY: 1/4T, 122°F (Axial Flaw) 3/4T, 109°F (Axial Flaw) 2500 '

loperlimAnalysis Version:5.4 Run:24301 Operlim.xlsm Version : 5.4.1 I 2250 2000

._j Unacceptable Ooeration I . llnservice Hydrostatic!

and Leak Test Limit I

I 1750 C) en 0...

Q) 1500 II)

II)

.... 1250 Q) 0...

i:,

Q) ns u 1000 cu I Lowest Service ~

IAcceptable I 0 Tem:J. =150'F Operation 750 500 Minimum 250 I+- Boltup Temo. =60'F I I

l I I

0 l I I I 0 50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Deg. F)

Figure 8-3 Arkansas Nuclear One Unit 2 Reactor Coolant System Inservice Hydrostatic and Leak Test Limitations Applicable for 54 EFPY (with Flange and LST Requirements and without Margins for Instrumentation Errors) using the 1998 through the 2000 Addenda App. G Methodology (w/ K1c)

WCAP-1 8169-NP June201 8 Revision I

Westinghouse Non-Proprietary Class 3 8-6 Table 8-1 Arkansas Nuclear One Unit 2 54 EFPY Heatup C urve Data Points using the 1998 through the 2000 Addenda App. G Methodology (w/ Kic, w/ Flange and LST Requirements, and w/o Margins for Instrumentation Errors) 50°F/hr 60°F/hr 70°F/hr 80°F/hr 50°F/hr Heatup 60°F/hr Heatup 70°F/hr Heatup 80°F/hr Heatup Criticality Criticality Criticality Criticality p p p p p p p p T (OF) T (°F) T (°F) T (OF) T (°F) T (°F) T (°F) T (°F)

(psig) (psig) (psig) (psig) (psig) (psig) (psig) (psig) 60 0 179 0 60 0 179 0 60 0 179 0 60 0 179 0 60 622 179 622 60 622 179 622 60 622 179 622 60 622 179 622 C')

)>

65 622 180 622 65 622 180 622 65 622 180 622 65 622 180 622 r C')

70 622 185 622 70 622 185 622 70 622 185 622 70 622 185 622 )>

z 75 622 190 622 75 622 190 622 75 622 190 622 75 622 190 622 0 N

80 622 190 1176 80 622 190 1114 80 622 190 1057 80 622 190 1007 m 85 622 195 1238 85 622 195 11 70 85 622 195 1109 85 622 195 1054 ........."'C' 90 622 200 1308 90 622 200 1234 90 622 200 1166 90 622 200 1106 b 0

0 95 622 205 1384 95 622 205 1303 95 622 205 1230 95 622 205 11 64 0 N

JOO 622 210 1468 100 622 210 1380 100 622 210 1300 100 622 210 1228  ::0 (I) 105 622 215 1561 105 622 215 1466 105 622 215 1378 105 622 215 1299 <

0 1664 110 622 "'C 110 622 220 220 1560 110 622 220 1464 11 0 622 220 1378 Ill (C

115 622 225 1778 115 622 225 1664 115 622 225 1559 115 622 225 1465 (I)

U1 120 125 130 622 622 622 230 235 240 1903 2042 2195 120 125 130 622 622 622 230 235 240 1778 1905 2045 120 125 130 622 622 622 230 235 240 1664 1780 1908 120 125 130 622 622 622 230 235 240 1561 1667 1784 0

CIO CIO 135 622 245 2363 135 622 245 2200 135 622 245 2050 135 622 245 1913 140 622 140 622 250 2370 140 622 250 2206 140 622 250 2056 145 622 145 622 145 622 255 2378 145 622 255 22 14 150 622 150 622 150 622 150 622 260 2388 150 1176 150 1114 150 1057 150 1007 155 1238 155 1170 155 1109 155 1054 WCAP-18169-NP June 20 18 Revision l

Westinghouse Non-Proprietary Class 3 8-7 Table 8-1 Arkansas Nuclear One Unit 2 54 EFPY Heatup Curve Data Points using the 1998 through the 2000 Addenda App. G Methodology (w/ K1c, w/ Flange and LST Requirements, and w/o Margins for Instrumentation Errors) 50°F/hr 60°F/hr 70°F/hr 80°F/hr 50°F/hr Heatup 60°F/hr Heatup 70°F/hr Heatup 80°F/hr Heatup Criticality Criticality Criticality Criticality p p p p p p p p T (°F) T (°F) T (°F) T (°F) T (°F) T (°F) T (°F) T (°F)

(psig) (psig) (psig) (psig) (psig) (psig) (psig) (psig) 160 1308 160 1234 160 1166 160 1106 165 1384 165 1303 165 1230 165 1164 0

)>

170 1468 170 1380 170 1300 170 1228 r 0

175 1561 175 1466 175 1378 175 1299 )>

z 180 1664 180 1560 180 1464 180 1378 0 N

185 1778 185 1664 185 1559 185 1465 m "ti 190 1903 190 1778 190 1664 190 1561 .....

I

.....i I

195 2042 195 1905 195 1780 195 1667 0 0

0 200 2195 200 2045 200 1908 200 1784 0 N

205 2363 205 2200 205 2050 205 1913  :;o l'D 210 2370 210 2206 210 2056 0 "ti 215 2378 215 2214 Q) u:i 220 2388 l'D UI 00 0

00 00 WCAP-18169-NP June 2018 Revision 1

CALC-AN02-EP-17 -00002 Rev O Page 59 of 88 Westinghouse Non-Proprietary Class 3 8-8 Table 8-2 Arkansas Nuclear One Unit 2 54 EFPY Cooldown Curve Data Points using the 1998 through the 2000 Addenda App. G Methodology (w/ K1c, w/ Flange and LST Requirements, and w/o Margins for Instrumentation Errors)

Steady-State -25°F/hr. -60°F/hr. -100°F/hr.

T (OF) P (psig) T (°F) P (psig) T (°F) P (psig) T (°F) P (psig) 60 0 60 0 60 0 60 0 60 622 60 622 60 622 60 576 65 622 65 622 65 622 65 594 70 622 70 622 70 622 70 613 75 622 75 622 75 622 75 622 80 622 80 622 80 622 80 622 85 622 85 622 85 622 85 622 90 622 90 622 90 622 90 622 95 622 95 622 95 622 95 622 100 622 100 622 100 622 100 622 105 622 105 622 105 622 105 622 110 622 110 622 110 622 llO 622 115 622 11 5 622 115 622 11 5 622 120 622 120 622 120 622 120 622 125 622 125 622 125 622 125 622 130 622 130 622 130 622 130 622 135 622 135 622 135 622 135 622 140 622 140 622 140 622 140 622 145 622 145 622 145 622 145 622 150 622 150 622 150 622 150 622 150 1321 150 132 1 150 132 1 150 1321 155 1393 155 1393 155 1393 155 1393 160 1474 160 1474 160 1474 160 1474 165 1562 165 1562 165 1562 165 1562 170 1660 170 1660 170 1660 170 1660 175 1768 175 1768 175 1768 175 1768 180 1888 180 1888 180 1888 180 1888 185 2020 185 2020 185 2020 185 2020 190 2166 190 2166 190 2166 190 2166 195 2328 195 2328 195 2328 195 2328 WCAP-1 8169-NP June 201 8 Revision 1

CALC-AN02-EP-17-00002 Rev O Page 60 of 88 Westinghouse Non-Proprietary Class 3 8-9 Table 8-3 Arkansas Nuclear One Unit 2 54 EFPY lnservice Hydrostatic and Leak Test Curve Data Points using the 1998 through the 2000 Addenda App. G Methodology (w/ K1c, w/ Flange and LST Requirements, and w/o Margins for Instrumentation Errors)

T (°F) P (psig) 60 0 60 622 65 622 70 622 75 622 80 622 85 622 90 622 95 622 100 622 105 622 110 622 115 622 120 622 125 622 130 622 135 622 140 622 145 622 150 622 150 1761 155 1858 160 1965 162 2000 165 2083 170 22 14 175 2358 179 2485 WCAP-18 169-NP June 20 18 Revision l

CALC-AN02-EP-17 -00002 Rev O Page 61 of 88 Westinghouse Non-Proprietary Class 3 9-1 9 REFERENCES

1. U.S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials," May 1988.

2. Westinghouse Report WCAP-14040-A, Revision 4, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves," May 2004.

3. Appendix G to the 1998 through the 2000 Addenda Edition of the ASME Boiler and Pressure Vessel (B&PV) Code, Section XI, Division 1, "Fracture Toughness Criteria for Protection Against Failure."

4. Code of Federal Regulations, 10 CFR Part 50, Appendix G, "Fracture Toughness Requirements,"

U.S. Nuclear Regulatory Commission, Federal Register, Volume 60, No. 243, dated December 19, 1995.

5. Regulatory Guide 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence," U.S. Nuclear Regulatory Commission, March 2001.

6. Westinghouse Report WCAP-16083-NP, Revision 1, "Benchmark Testing of the FERRET Code for Least Squares Evaluation of Light Water Reactor Dosimetry," April 2013.

7. RSICC Data Library Collection DLC-1 85, "BUGLE-96: Coupled 47 Neutron, 20 Gamma-Ray Group Cross-Section Library Derived from ENDF/B-VI for LWR Shielding and Pressure Vessel Dosimetry Applications," March 1996.

8. NRC Regulatory Issue Summary 2014-11, "Information on Licensing Applications for Fracture Toughness Requirements for Ferri tic Reactor Coolant Pressure Boundary Components," U.S. Nuclear Regulatory Commission, October 2014. [Agencywide Document Management System (ADAMS)

Accession Number ML14149Al65}

9. ASME Boiler and Pressure Vessel (B&PV) Code, Section III, Division 1, Subsection NB, "Class 1 Components."

10. NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, Chapter 5 LWR Edition, Branch Technical Position 5-3, "Fracture Toughness Requirements,"

Revision 2, U.S. Nuclear Regulatory Commission, March 2007.

11. Combustion Engineering Report A-PENG-ER-002, Revision 0, "The Reactor Vessel Group Records Evaluation Program Phase II Final Report for the ANO 2 Reactor Pressure Vessel Plates, Forgings, Welds and Cladding," October 1995.

12. Battelle - Columbus Report BMI-0584, "Final Report on Examination, Testing, and Evaluation of Irradiated Pressure Vessel Surveillance Specimens from the Arkansas Nuclear One Unit 2 Generating Plant to Arkansas Power and Light Company," May 1984.

13. Arkansas Nuclear One - Unit 2 License Renewal Application, October 2003. [Available on the NRC website}

14. AREVA NP, Inc. Report BAW-2399, Revision 1, "Analysis of Capsule W-104 Entergy Operations, Inc. Arkansas Nuclear One Unit 2 Power Plant Reactor Vessel Material Surveillance Program,"

February 2005.

15 . Code of Federal Regulations, 10 CFR 50.61 , "Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events," Federal Register, Volume 60, No. 243, dated December 19, 1995, effective January 18, 1996.

16. Combustion Engineering Owners Group Report CE NPSD-1039, Revision 2, "Best Estimate Copper and Nickel Values in CE Fabricated Reactor Vessel Welds," June 1997.

WCAP-18169-NP June2018 Revision 1

r CALC-AN02-EP-17 -00002 Rev O Page 62 of 88 Westinghouse Non-Proprietary Class 3 9-2

17. Combustion Engineering Report TR-MCD-002, "Arkansas Power & Light Arkansas Nuclear One -

Unit 2 Evaluation of Baseline Specimens Reactor Vessel Materials Irradiation Surveillance Program,"

March 1976.

18. Westinghouse Report WCAP-17501-NP, Revision 0, "Analysis of Capsule 104° from the Calvert Cliffs Unit No. 2 Reactor Vessel Radiation Surveillance Program," February 2012.

19. Westinghouse Report WCAP-16012, Revision 0, "Analysis of Capsule W-83 from the Dominion Nuclear Connecticut Millstone Unit 2 Reactor Vessel Radiation Surveillance Program," February 2003 .

20. Westinghouse Report WCAP-16918-NP, Revision 1, "Analysis of Capsule V from the Southern Nuclear Operating Company Joseph M. Farley Unit 2 Reactor Vessel Radiation Surveillance Program," April 2008.

21. Westinghouse Report WCAP-18166-NP, Revision 0, "Analysis of Capsule 284° from the Entergy Operations, Inc. Arkansas Nuclear One Unit 2 Reactor Vessel Radiation Surveillance Program,"

September 2016.

22. K. Wichman, M. Mitchell, and A. Hiser, U.S. NRC, Generic Letter 92-01 and RPV Integrity Workshop Handouts, NRC/Industry Workshop on RPV Integrity Issues, February 12, 1998. [ADAMS Accession Number MLJ 10070570]

23. RSICC Computer Code Collection CCC-650, "DOORS 3.2: One, Two- and Three Dimensional Discrete Ordinates Neutron/Photon Transport Code System," April 1998.

WCAP-1 8169-NP June2018 Revision I

CALC-AN02-EP-17-00002 Rev O Page 63 of 88 Westinghouse Non-Proprietary Class 3 A-1 APPENDIX A THERMAL STRESS INTENSITY FACTORS (Ku)

Tables A-1 and A-2 contain the thermal stress intensity factors (K11) for the maximum heatup and cooldown rates at 54 EFPY for Arkansas Nuclear One Unit 2. The reactor vessel cylindrical shell radii to the l/4T and 3/4T locations are as follows :

• l /4T Radius= 81.688 inches

• 3/4T Radius= 85.625 inches WCAP-18169-NP June 2018 Revision 1 I

I l_

CALC-AN02-EP-17-00002 Rev O Page 64 of 88 Westinghouse Non-Proprietary Class 3 A -2 TableA-1 Ku Values for Arkansas Nuclear One Unit 2 at 54 EFPY 80°F/hr Heatup Curves (w/

Flange and LST Requirements, and w/o Margins for Instrument Errors)

Water Vessel Temperature 1/4T Thermal Stress Vessel Temperature 3/4T Thermal Stress Temp. at 1/4T Location for Intensity Factor at 3/4T Location for Intensity Factor (OF) 80°F/hr Heatup (°F) (ksi -Vin.) 80°F/hr Heatup (°F) (ksi -Vin.)

60 56.321 - 1.028 55.105 0.545 65 59.387 -2.397 55.641 1.540 70 62.751 -3 .449 56.862 2.407 75 66.385 -4.400 58.707 3.149 80 70.275 -5. 155 61.052 3.764 85 74.297 -5.8 16 63.832 4.283 90 78.513 -6.35 1 66.967 4.7 16 95 82.818 -6.820 70.400 5.083 100 87.260 -7.203 74.080 5.390 105 91.763 -7.540 77.960 5.654 11 0 96.358 -7.818 82.009 5.878 11 5 100.997 -8.066 86.196 6.070 120 105 .699 -8.271 90.500 6.236 125 11 0.432 -8.456 94.900 6.3 79 130 11 5.207 -8.611 99.381 6.5 03 135 120.006 -8.753 103 .929 6.6 13 140 124.832 -8.872 108.533 6.709 145 129.675 -8.984 113.183 6.794 150 134.537 -9.080 117.872 6.87 1 155 139.412 -9.170 122.594 6.940 160 144.298 -9.249 127.343 7.002 165 149.195 -9.324 132.114 7.060 170 154.099 -9.39 1 136.904 7. 11 3 175 159.011 -9.457 141.710 7.162 180 163.927 -9.5 16 146.529 7.209 185 168.849 -9.574 151.358 7.253 190 173 .774 -9.627 156.198 7.294 195 178 .703 -9.68 1 161.044 7.334 200 183 .634 -9.730 165 .897 7.373 205 188.568 -9.780 170.756 7.410 2 10 193.503 -9.826 175 .6 19 7.446 WCAP- 18169-NP June201 8 Revision 1

CALC-AN02-EP-17-00002 Rev O Page 65 of 88 Westinghouse Non-Proprietary Class 3 A-3 TableA-2 Ktt Values for Arkansas Nuclear One Unit 2 at 54 EFPY 100°F/hr Cooldown Curves (w/ Flange and LST Requirements, and w/o Margins for Instrument Errors)

Water Vessel Temperature at l/4T 100°F/hr Cooldown Temp. Location for 100°F/hr l/4T Thermal Stress (OF) Cooldown (°F) Intensity Factor (ksi --./in.)

210 232.413 13.504 205 227.339 13.448 200 222.265 13 .392 195 217.192 13 .336 190 212.118 13.281 185 207.044 13.225 180 201.970 13.169 175 196.897 13.113 170 191.823 13 .058 165 186.749 13 .002 160 181.676 12.947 155 176.602 12.891 150 171.529 12.836 145 166.456 12.781 140 161.383 12.726 135 156.310 12.671 130 151.237 12.616 125 146.164 12.561 120 141.091 12.507 115 136.018 12.452 110 130.946 12.398 105 125.874 12.343 100 120.801 12.289 95 115.729 12.235 90 110.657 12.182 85 105.585 12.128 80 100.514 12.074 75 95.442 12.020 70 90.371 11.967 65 85.299 11.914 60 80.229 11.860 WCAP-18169-NP June 2018 Revision 1

CALC-AN02-EP-17 -00002 Rev O Page 66 of 88 Westinghouse Non-Proprietary Class 3 B-1 APPENDIX B REACTOR VESSEL INLET AND OUTLET NOZZLES As described in NRC Regulatory Issue Summary (RIS) 2014-11 [Ref. B-1], reactor vessel non-beltline materials may define pressure-temperature (P-T) limit curves that are more limiting than those calculated for the reactor vessel cylindrical shell beltline materials. Reactor vessel nozzles, penetrations, and other discontinuities have complex geometries that can exhibit significantly higher stresses than those for the reactor vessel beltline shell region. These higher stresses can potentially result in more restrictive P-T limits, even if the reference temperatures (RTNOT) for these components are not as high as those of the reactor vessel beltline shell materials that have simpler geometries.

The methodology contained in WCAP-14040-A, Revision 4 [Ref. B-2] was used in the main body of this report to develop P-T limit curves for the limiting Arkansas Nuclear One Unit 2 cylindrical shell beltline material; however, WCAP-14040-A, Revision 4 does not consider ferritic materials in the area adjacent to the beltline, specifically the stressed inlet and outlet nozzles. Due to the geometric discontinuity, the inside corner regions of these nozzles are the most highly stressed ferritic component outside the beltline region of the reactor vessel; therefore, these components are analyzed in this Appendix. P-T limit curves are determined for the reactor vessel nozzle comer region for Arkansas Nuclear One Unit 2 and compared to the P-T limit curves for the reactor vessel traditional beltline region in order to determine if the nozzles can be more limiting than the reactor vessel beltline as the plant ages and the vessel accumulates more neutron fluence. The increase in neutron fluence as the plant ages causes a concern for embrittlement of the reactor vessel above the beltline region. Therefore, the P-T limit curves are developed for the nozzle inside comer region since the geometric discontinuity results in high stresses due to internal pressure and the cooldown transient. The cooldown transient is analyzed as it results in tensile stresses at the inside surface of the nozzle comer.

A 1/4T axial flaw is postulated at the inside surface of the reactor vessel nozzle comer and stress intensity factors are determined based on the rounded curvature of the nozzle geometry. The allowable pressure is then calculated based on the fracture toughness of the nozzle material and the stress intensity factors for the l/4T flaw.

B.1 CALCULATION OF ADJUSTED REFERENCE TEMPERATURES The fracture toughness (Kie) used for the inlet and outlet nozzle material is defined in Appendix G of the Section XI ASME Code, as discussed in Section 6 of this report. The K,c fracture toughness curve is dependent on the Adjusted Reference Temperature (ART) value for irradiated materials. The ART values for the inlet and outlet nozzle materials are determined using the methodology contained in Regulatory Guide 1.99, Revision 2 [Ref. B-3], which is described in Section 7 of this report, and weight percent (wt.

%) copper (Cu) and nickel (Ni), initial RT NDT value, and projected neutron fluence as inputs. The material properties for each of the reactor vessel inlet and outlet nozzle forging materials are documented in Table B- 1 and a summary of the limiting inlet and outlet nozzle ART values for Arkansas Nuclear One Unit 2 is presented in Table B-2.

Nozzle Material Properties The Arkansas Nuclear One Unit 2 nozzle material properties are provided in Table B-1. Nickel (Ni),

Manganese (Mn), and Phosphorus (P) weight percent (wt. %) values were obtained as the average of the material-specific analyses documented in Combustion Engineering reportA-PENG-ER-002 [Ref. B-4] for WCAP-18169-NP June2018 Revision l

CALC-AN02-EP-17 -00002 Rev O Page 67 of 88 Westinghouse Non-Proprietary Class 3 B-2 each of the Arkansas Nuclear One Unit 2 reactor vessel inlet and outlet nozzles. Copper weight percent values for each of the Arkansas Nuclear One Unit 2 outlet nozzles were also taken to be the average of all available analyses contained in Combustion Engineering report A-PENG-ER-002 [Ref. B-4]. However, the A-PENG-ER-002 report did not contain copper weight percent values for the inlet nozzles, because at the time that the Arkansas Nuclear One Unit 2 nozzles were manufactured, these values were not required to be documented for SA-508, Class 2 low-alloy steel. Therefore, no material-specific copper weight percent value is available for the Arkansas Nuclear One Unit 2 inlet nozzles. Per NRC RIS 2014-11 [Ref.

B-1], a copper weight percent value is not required for calculation of the Arkansas Nuclear One Unit 2 nozzle material ART values, because the nozzles have fluence values less than 1 x 10 17 n/cm2* However, if a copper weight percent value is ever needed, a best-estimate copper weight percent value is available from Section 4 of the NRC-approved Boiling Water Reactor Vessel and Internals Project (BWRVlP

[proprietary]) report, BWRVlP-173-A [Ref. B-5], and this value could be utilized for the Arkansas Nuclear One inlet nozzles. A mean plus two standard deviations methodology was applied to the data in BWRVIP-173-A to determine a conservative copper weight percent value. The data in the BWRVIP report was tabulated from an industry-wide database of SA-508, Class 2 forging materials.

The Charpy V-Notch forging specimen orientation for the inlet and outlet nozzles was not reported in A-PENG-ER-002 ; thus, it was conservatively assumed that the orientation was the "strong direction" for each nozzle forging. The initial RTNOT values were therefore determined for each of the Arkansas Nuclear One Unit 2 reactor vessel inlet and outlet nozzle forging materials using the Branch Technical Position (BTP) 5-3 , Position 1.1(3) methodology [Ref. B-6]. The initial RTNOT values for all of the nozzle materials were determined directly from the data or by using a CVGRAPH, Version 6.02 hyperbolic tangent curve fit through the minimum data points, in accordance with ASME Code Section III, Subarticle NB-2331, Paragraph (a)(4) [Ref. B-7]. The initial RTNoT values were determined using both BTP 5-3 Position 1.1(3)(a) and Position 1.1(3)(b), and the more limiting initial RTNoT value was chosen for each nozzle forging material. The Arkansas Nuclear One Unit 2 initial RTNOT values for the inlet and outlet nozzles materials are summarized in Table B-1.

Nozzle Calculated Neutron Fluence Values The maximum fast neutron (E > 1 MeV) exposure of the Arkansas Nuclear One Unit 2 reactor vessel materials is discussed in Section 2 of this report. The fluence values used in the inlet and outlet nozzle ART calculations were calculated at the lowest extent of the nozzles (i.e., the nozzle to nozzle shell weld locations) and were chosen at an elevation lower than the actual elevation of the postulated flaw, which is at the inside comer of the nozzle, for conservatism.

Per Table 2-2, the inlet nozzles are determined to receive a projected maximum fluence of 7.96 x 10 16 n/cm2 (E > 1 MeV) at the lowest extent of the nozzles at 54 EFPY. Similarly, the outlet nozzles are projected to achieve a maximum fluence value of 9.80 x 10 16 n/cm2 (E > 1 MeV) at the lowest extent of the nozzles at 54 EFPY. Thus, the maximum neutron fluence values for the nozzle materials are not projected to exceed a fluence of 1 x 10 17 n/cm2 at 54 EFPY. Per NRC RIS 2014-11 [Ref. B-1] ,

embrittlement of reactor vessel materials, with projected fluence values less than 1 x 10 17 n/cm 2, does not need to be considered. Therefore, the initial RTNOT values documented in Table B-1 are identical to the nozzle ART values.

The neutron fluence values used in the ART calculations for the Arkansas Nuclear One Unit 2 inlet and outlet nozzle forging materials are summarized in Table B-1.

WCAP-18169-NP June 2018 Revision I

CALC-AN02-EP-17 -00002 Rev O Page 68 of 88 Westinghouse Non-Proprietary Class 3 B-3 Table B-1 Summary of the Arkansas Nuclear One Unit 2 Reactor Vessel Nozzle Material Initial RT NDT, Chemistry, and Fluence Values at 54 EFPY Fluence at Lowest Chemical Composition<*> (c)

RTNDT(U) Extent of Nozzle

1.0 MeV) Cu Ni Mn p Inlet Nozzle C-8015-1 Note (b) 0.69 0.65 0.007 30 7.96 X 10 16 Inlet Nozzle C-8015-2 Note (b) 0.60 0.70 0.010 10 7.96 X 10 16 Inlet Nozzle C-8015-3 Note (b) 0.69 0.65 0.006 10 7.96 X 10 16 Inlet Nozzle C-8015-4 Note (b) 0.65 0.70 0.007 30 7.96 X 10 16 Outlet Nozzle C-8016-1 0.12 0.63 0.64 0.006 0 9.80 X 10 16 Outlet Nozzle C-8016-2 0.17 0.76 0.83 0.017 13 .5 9.80 X 10 16 Notes: (a) Chemistry values are the average of all available material-specific chemical analyses, unless otherwise noted. (b) The Arkansas Nuclear One Unit 2 copper weight percent values are not documented in the historical records. This value is not needed for the current analysis per NRC RIS 20 14-11 [Ref. 8-1], since the fluence values for these materials are below 1.0 x 10 17 n/cm2 . If a copper weight percent value is needed in the future, a best-estimate copper weight percent value is available from Section 4 of the NRC-approved 8WRVIP (proprietary) report, 8WRVIP-l73-A [Ref. 8-5] . (c) RTNDT(U) values were determined using NUREG-0800, 8TP 5-3 Position 1.1(3)(a) and (b) [Ref. 8-6] methodology with the more limiting RT NDT(UJ value being selected for each nozzle material. (d) Fluence values conservatively correspond to 54 EFPY fluence values at the lowest extent of the nozzle weld. Table B-2 Summary of the Limiting ART Values for the Arkansas Nuclear One Unit 2 Inlet and Outlet Nozzle Materials Nozzle Material and ID Limiting ART Value EFPY Number (OF) Inlet Nozzle C-8015-1 30 54 and C-8015-4 Outlet Nozzle C-8016-2 13 .5 The use of the embrittlement conclusion of NRC RlS 2014-11 [Ref. B-1], and thus the limiting ART values summarized in Table B-2, will remain unchanged as long as the fluence values assigned to the inlet and outlet nozzles remain below 1.0 x 10 17 n/cm 2 (E > 1.0 MeV). If these fluence values are reached, the Arkansas Nuclear One Unit 2 nozzle material ART values should be re-evaluated. WCAP-18169-NP June2018 Revision 1 CALC-AN02-EP-17 -00002 Rev O Page 69 of 88 Westinghouse Non-Proprietary Class 3 B-4 B.2 NOZZLE COOLDOWN PRESSURE-TEMPERATURE LIMITS Allowable pressures are determined for a given temperature based on the fracture toughness of the limiting nozzle material along with the appropriate pressure and thermal stress intensity factors . The Arkansas Nuclear One Unit 2 nozzle fracture toughness used to determine the P-T limits is calculated using the limiting inlet and outlet nozzle ART values from Table B-2. The stress intensity factor correlations used for the nozzle comers are provided in ORNL study, ORNL/TM-2010/246 [Ref. B-8], and are consistent with ASME PVP2011-57015 [Ref. B-9]. The methodology includes postulating an inside surface 1/4T nozzle comer flaw, and calculating through-wall nozzle comer stresses for a cooldown rate of 100°F/hour. The through-wall stresses at the nozzle corner location were fitted based on a third-order polynomial of the form:

where, CJ = through-wall stress distribution x = through-wall distance from inside surface Ao, A 1, A 2, A3 = coefficients of polynomial fit for the third-order polynomial, used in the stress intensity factor expression discussed below The stress intensity factors generated for a rounded nozzle corner for the pressure and thermal gradient were calculated based on the methodology provided in ORNL/TM-2010/246. The stress intensity factor expression for a rounded corner is:

where, K1 stress intensity factor for a circular corner crack on a nozzle with a rounded inner radius comer a crack depth at the nozzle corner, for use with 1/4T (25% of the wall thickness)

The Arkansas Nuclear One Unit 2 reactor vessel inlet and outlet nozzle P-T limit curves are shown in Figures B-1 and B-2, respectively, based on the stress intensity factor expression discussed above; also shown in these figures are the traditional beltline cooldown P-T limit curves from Figure 8-2. The nozzle P-T limit curves are provided for a cooldown rate of 100°F/hr, along with a steady-state curve. An outside surface flaw in the nozzle was not considered because the pressure stress is significantly lower at the outside surface than the inside surface. A heatup nozzle P-T limit curve is also not provided since it would be less limiting than the cooldown nozzle P-T limit curve in Figures B-1 and B-2 for an inside surface flaw. Additionally, the cooldown transient is more limiting than the heatup transient since it results in tensile stresses at the inside surface of the nozzle corner. WCAP-18169-NP June2018 Revision 1 CALC-AN02-EP-17-00002 Rev O Page 70 of 88 Westinghouse Non-Proprietary Class 3 B-5 Conclusion Based on the results shown in Figures B-1 and B-2, it is concluded that the nozzle P-T limits are bounded by the traditional beltline curves. Therefore, the P-T limits provided in Section 8 for 54 EFPY remain limiting for the beltline and non-beltline reactor vessel components. WCAP-18169-NP June 2018 Revision I CALC-AN02-EP-17-00002 Rev O Page 71 of 88 Westinghouse Non-Proprietary Class 3 B-6 2500 I Inlet Nozzle Cooldown L t I 2250 -lOO"F/Hr I Inlet Nozzle r Steady State 2000 -l-I I 1750 C) (/J 1500 r-- Q. ~ 1250 -- - - 4 ~ 1 t - - -1

s 1/)

1/) ~ Q. I E 1000 --+ .2l1/) Cooldown >, Rates Acceptable (/J 0 r:: C'l:J 750 steady-state -25"F/Hr -60"F/Hr 0 eration I 0 0 ...0 t, Cooldown Rate

E 500 - ~ ~ -100°F/Hr 0::

250 Minimum Boltup Tern . = 60°F 0 50 100 150 200 250 300 350 400 450 500 550 Averaged Reactor Coolant System Temperature (°F) Figure B-1 Comparison of Arkansas Nuclear One Unit 2 Beltline P-T Limits to Inlet Nozzle Limits WCAP-181 69-NP June 2018 Revision l CALC-AN02-EP-17-00002 Rev O Page 72 of 88 Westinghouse Non-Proprietary Class 3 B-7 2500 -,------,--,~-.,------,-----,-----,- - - , - - - - - - , Outlet Nozzle I M.l.------t- - -t Cooldown l 2250 -100°F/Hr t Outlet Nozzle - - ----t------f 2000 Steady State I 1750 -+-- - - - - - - - ---+---+ -en C) -0.. ~ 1500 1250 t -+--- - --+- (I) (I) ~ 0.. E 1000 Cooldown ~ Rates Acceptable en ~ C: 0 750 0 eration 0 0 Lowest Service ...0 Tern . = 150"F 0 Cooldown Rate cu Q) 500 -100°F/Hr a::: 250 0 0 50 Minimum Boltup Tern . = 60"F 100 150 t 200 250 300 350 400 450 500 550 Averaged Reactor Coolant System Temperature (°F) Figure B-2 Comparison of Arkansas Nuclear One Unit 2 Beltline P-T Limits to Outlet Nozzle Limits WCAP-181 69-NP June 20 18 Revision 1 CALC-AN02-EP-17-00002 Rev O Page 73 of 88 Westinghouse Non-Proprietary Class 3 B-8 B.3 REFERENCES B-1 NRC Regulatory Issue Summary 2014-11 , "Information on Licensing Applications for Fracture Toughness Requirements for Ferritic Reactor Coolant Pressure Boundary Components," U.S . Nuclear Regulatory Commission, October 2014. [ADAMS Accession Number ML 14149AJ65] B-2 Westinghouse Report WCAP-14040-A, Revision 4, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves," May 2004. B-3 U.S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials," May 1988. B-4 Combustion Engineering Report A-PENG-ER-002, Revision 0, "The Reactor Vessel Group Records Evaluation Program Phase II Final Report for the ANO 2 Reactor Pressure Vessel Plates, Forgings, Welds and Cladding," October 1995. B-5 BWR VIP-173 -A: B WR Vessel and Internals Project: Evaluation of Chemistry Data for BWR Vessel Nozzle Forging Materials. EPRI, Palo Alto, CA: 2011. 1022835. B-6 NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, Chapter 5 LWR Edition, Branch Technical Position 5-3, "Fracture Toughness Requirements," Revision 2, U.S. Nuclear Regulatory Commission, March 2007. B-7 ASME Boiler and Pressure Vessel (B&PV) Code, Section III, Division 1, Subsection NB, "Class 1 Components." B-8 Oak Ridge National Laboratory Report, ORNL/TM-2010/246, "Stress and Fracture Mechanics Analyses of Boiling Water Reactor and Pressurized Water Reactor Pressure Vessel Nozzles - Revision 1," June 2012. [ADAMS Accession Number ML110060164] B-9 ASME PVP2011-57015, "Additional Improvements to Appendix G of ASME Section X1 Code for Nozzles," G. Stevens, H. Mehta, T. Griesbach, D. Sommerville, July 2011 . WCAP-18169-NP June 2018 Revision I CALC-AN02-EP-17-00002 Rev O Page 74 of 88 Westinghouse Non-Proprietary Class 3 C-1 APPENDIX C NON-REACTOR VESSEL FERRITIC COMPONENTS 10 CFR Part 50, Appendix G [Ref. C-1], requires that all Reactor Coolant Pressure Boundary (RCPB) components meet the requirements of Section III of the ASME Code. The lowest service temperature requirement for all RCPB components, which is specified in NB-3211 and NB-2332 of the Section III ASME Code, is the relevant requirement that would affect the pressure-temperature (P-T) limits. The lowest service temperature (LST) requirement of NB-3211 and NB-2332 of the Section III ASME Code is applicable to material for ferritic piping, pumps and valves with a nominal wall thickness greater than 2 Yi inches [Ref. C-2]. Arkansas Nuclear One Unit 2 reactor coolant system piping contains ferritic materials in the Class 1 piping; the pumps and valves do not contain ferritic material. The LST requirements of NB-3211 and NB-2332 are considered in Section 6.4 of this report. The other ferritic RCPB components that are not part of the reactor vessel consist of the replacement closure head, the pressurizer and the replacement steam generators. The replacement closure head is considered in the cylindrical beltline P-T limit curves as described in Section 6 of this report. Furthermore, the replacement closure head has not undergone neutron embrittlement that would affect P-T limits. Therefore, no further consideration is necessary for this component with regards to P-T limits. The pressurizer was constructed to the 1968 Edition through 1970 Summer Addenda Section III ASME Code and met all applicable requirements at the time of construction and is original to the plant. Furthermore, the pressurizer has not undergone neutron embrittlement that would affect P-T limits. Therefore, no further consideration is necessary for this component with regards to P-T limits. The replacement steam generators were constructed to the 1989 Edition Section III ASME Code and met all applicable requirements at the time of construction. Furthermore, the replacement steam generators have not undergone neutron embrittlement that would affect P-T limits. Therefore, no further consideration is necessary for these components with regards to P-T limits. C.1 REFERENCES C-1 Code of Federal Regulations, 10 CFR Part 50, Appendix G, "Fracture Toughness Requirements," U.S. Nuclear Regulatory Commission, Federal Register, Volume 60, No. 243 , December 19, 1995. C-2 ASME Boiler and Pressure Vessel (B&PV) Code, Section III, Division 1, Subsection NB, "Class 1 Components." WCAP-18169-NP June 2018 Revision l CALC-AN02-EP-17-00002 Rev O Page 75 of 88 Westinghouse Non-Proprietary Class 3 D-1 APPENDIX D CREDIBILITY EVALUATION OF THE WELD HEAT# 10137 SURVEILLANCE DATA D.1 INTRODUCTION Regulatory Guide 1. 99, Revision 2 [Ref. D-1] describes general procedures acceptable to the NRC staff for calculating the effects of neutron radiation embrittlement of the low-alloy steels currently used for light-water-cooled reactor vessels. Position 2.1 of Regulatory Guide 1.99, Revision 2, describes the method for calculating the adjusted reference temperature of reactor vessel beltline materials using surveillance capsule data. The methods of Position 2.1 can only be applied when two or more credible surveillance data sets become available from the reactor in question. The credibility of all surveillance program data applicable to the Arkansas Nuclear One Unit 2 beltline was assessed in WCAP-18166-NP [Ref. D-2]. However, the Arkansas Nuclear One Unit 2 extended beltline contains two welds with sister plant data, the Upper Shell Longitudinal Welds l-203A, B, & C (Heat# BOLA) and the Upper to Intermediate Shell Girth Weld 8-203 (Heat# 10137). Note that no surveillance data is available for the other two Heats (# 6329637 and # FAGA) which were also used to make the Upper to Intermediate Shell Girth Weld 8-203. The weld Heat # BOLA sister plant data is available from the J.M. Farley Unit 2 surveillance program. Since this surveillance data is analyzed by itself, the credibility conclusion documented in Appendix D of WCAP-16918, Revision 1 [Ref. D-3] is applicable to Arkansas Nuclear One Unit 2; thus, the credibility conclusion of the Heat# BOLA data need not be updated. The J.M. Farley Unit 2 surveillance weld data (Heat# BOLA) is non-credible in regard to the Arkansas Nuclear One Unit 2 reactor vessel materials. The weld Heat# 10137 sister plant data is available from both the Calvert Cliffs Unit 2 and Millstone Unit 2 surveillance programs. The Millstone Unit 2 surveillance program includes two distinct welds, Heat# 10137 and Heat# 90136. In previous analyses, this weld surveillance data was treated as one combined weld and subsequently analyzed together. However, these two weld metal heats were not melted together into a tandem weld; they were individually deposited. It cannot be determined with full confidence how much of the overall surveillance weld is which weld metal heat and, furthermore, exactly which weld heat specimens are contained in which surveillance capsules in the Millstone Unit 2 program. The Millstone Unit 2 (combined) surveillance weld data met the second and third credibility criteria of Regulatory Guide 1.99, Revision 2 [Ref. D-1]. Additionally, Table D-2 of WCAP-16012 [Ref. D-4] indicates that all of the measured weld L'iRT NDT values were within the 1-sigma scatter band; therefore, suggesting that there is good agreement between the measured capsule data and the embrittlement correlations. If the two heats of weld material were evaluated individually, one would expect that the scatter in the data would decrease since the irradiated material would embrittle differently for the two separate welds with different, as-measured, copper and nickel contents. However, since the (combined) weld material already passes the Regulatory Guide 1.99, Revision 2 credibility analysis, a re-evaluation of the material (as two separate heats) is not expected to significantly change the overall results of the subsequent reactor vessel integrity analyses. Thus, the surveillance weld metal will be considered to be only Heat# 10137 for the evaluations contained herein. All currently determined input data for Position 2.1 chemistry factor determination (See Section 5) and surveillance data credibility assessment WCAP-18169-NP June 2018 Revis ion I CALC-AN02-EP-17-00002 Rev O Page 76 of 88 Westinghouse Non-Proprietary Class 3 D-2 documented in this Appendix will be used "as-is," as documented in the Millstone Unit 2 surveillance capsule analyses of record. For conservatism, no reduction in the margin term of Regulatory Guide 1.99, Revision 2 [Ref. D-1] was taken to account for the additional uncertainties, despite the data remaining credible (see Section D.2). Note that this approach should also be followed when completing analyses per 10 CFR 50.61 [Ref. D-5]. Despite this additional conservatism, the Arkansas Nuclear One Unit 2 Upper to Intermediate Shell Girth Weld 8-203 (Heat# 10137) was not the limiting material for the Arkansas Nuclear One Unit 2 P-T limit curves. D.2 EVALUATION Per Appendix D of WCAP-17501-NP [Ref. D-6], the Calvert Cliffs Unit 2 surveillance weld data (Heat# 10137) was deemed credible, and per Appendix D of WCAP-16012 [Ref. D-4], the Millstone Unit 2 surveillance weld data (Heat# 10137) was also deemed credible. Thus, when analyzed individually, these surveillance welds pass all five of the Regulatory Guide 1.99, Revision 2 [Ref. D-1] credibility criterion. The only credibility criterion that must be updated as a result of analyzing the two surveillance welds together is Criterion 3. This evaluation is documented herein. Criterion 3: When there are two or more sets of surveillance data from one reactor, the scatter of LlRT NOT values about a best-fit line drawn as described in Regulatory Guide 1.99, Revision 2 [Ref. D-1] normally should be less than 28°F for welds and l 7°F for base metal. Even if the fluence range is large (two or more orders of magnitude), the scatter should not exceed twice those values. Even if the data fail this criterion for use in shift calculations, they may be credible for determining decrease in upper-shelf energy if the upper shelf can be clearly determined, following the definition given in ASTM E185-82 [Ref. D-7]. The functional form of the least-squares method as described in Regulatory Guide 1.99, Revision 2 will be utilized to determine a best-fit line for this data and to determine if the scatter of these LlRTNDT values about this line is less than 28°F for the weld. Following is the calculation of the best-fit line as described in Reference D-1. In addition, the recommended NRC methods for determining credibility will be followed. The NRC methods were presented to industry at a meeting held by the NRC on February 12 and 13, 1998 [Ref. D-8]. At this meeting the NRC presented five cases. Of the five cases, Case 5 ("Surveillance Data from Other Sources Only") most closely represents the situation for the Arkansas Nuclear One Unit 2 reactor vessel Upper to Intermediate Shell Girth Weld 8-203 (Heat # 10137) as described below: Heat# 10137 {Case 5) - This weld heat pertains to the Upper to Intermediate Shell Girth Weld 8-203 in the Arkansas Nuclear One Unit 2 reactor vessel. This weld heat is not contained in the Arkansas Nuclear One Unit 2 surveillance program. However, it is contained in the Calvert Cliffs Unit 2 and Millstone Unit 2 surveillance programs. NRC Case 5 per Reference D-8 is entitled "Surveillance Data from Other Sources Only" and most closely represents the situation for Arkansas Nuclear One Unit 2 weld Heat# 10137. WCAP-18169-NP June 2018 Revision I CALC-AN02-EP-17 -00002 Rev O Page 77 of 88 Westinghouse Non-Proprietary Class 3 D-3 Credibility Assessment Case 5: Weld Heat# 10137 (Calvert Cliffs Unit 2 Data Only) Following the NRC Case 5 guidelines, the Calvert Cliffs Unit 2 and Millstone Unit 2 surveillance weld metal (Heat # 10137) will be evaluated for credibility. Weld Heat# 10137 pertains to Arkansas Nuclear One Unit 2 reactor vessel Upper to Intermediate Shell Girth Weld 8-203 , but is not contained in the Arkansas Nuclear One Unit 2 surveillance program. In accordance with the NRC Case 5 guidelines, the data from only Calvert Cliffs Unit 2 will be analyzed first, since the irradiation environment for Calvert Cliffs Unit 2 is judged closer to that of Arkansas Nuclear One Unit 2 as evidenced by the temperature adjustments documented in Table 4-2. This assessment was performed in Appendix D of WCAP-17501 -NP [Ref. D-6] and concluded that the surveillance data for Heat # 10137 from Calvert Cliffs Unit 2 only was credible. Therefore, in accordance with Case 5, the combined data from both Calvert Cliffs Unit 2 and Millstone Unit 2 will now be assessed to determine the credibility conclusion for all applicable data for weld Heat# 10137. WCAP-18169-NP June 2018 Revision l CALC-AN02-EP-17-00002 Rev O Page 78 of 88 Westinghouse Non-Proprietary Class 3 D-4 Credibility Assessment Case 5: Weld Heat# 10137 (All data) In accordance with the NRC Case 5 guidelines, the data from Calvert Cliffs Unit 2 and Millstone Unit 2 will now be analyzed together. Data is adjusted to the mean chemical composition and operating temperature of the surveillance capsules. This is performed in Table D-1. Table D-1 Mean Chemical Composition and Operating Temperature for Calvert Cliffs Unit 2 and Millstone Unit 2 Cu Ni Inlet Temperature during Material Capsule Wt.%<*> Wt.%<*> Period of Irradiation (0 F)(b) 263° 550 Weld Metal Heat # 10137 97° 0.21 0.06 549 (Calvert Cliffs Unit 2 Data) 104° 548 97° 544.3 Weld Metal Heat# 10137 104° 0.21 0.06 547.6 (Millstone Unit 2 Data) 83° 548.0 MEAN 0.21 0.06 547.8 Note: (a) Chemistry data obtained from Table 3-1. (b) Temperature data obtained from Table 4-2. Since the mean chemical composition of the surveillance capsule data is identical to the actual chemical composition data for each capsule, no chemistry adjustment is necessary. However, since the Calvert Cliffs Unit 2 and Millstone Unit 2 surveillance capsule operating temperatures are not identical to the mean operating temperature, the surveillance capsule data will be adjusted to the mean operating temperature. The capsule-specific temperature adjustments are as shown in Table D-2. Table D-2 Operating Temperature Adjustments for the Calvert Cliffs Unit 2 and Millstone Unit 2 Surveillance Capsule Data Inlet Temperature during Mean Operating Temperature Material Capsule Period of Irradiation (°F) Temperature (°F) Adjustment (°F) 263° 550 2.2 Weld Metal Heat# 10137 97° 549 1.2 (Calvert Cliffs Unit 2 Data) 104° 548 0.2 547.8 97° 544.3 -3.5 Weld Metal Heat# 10137 104° 547.6 -0.2 (Millstone Unit 2 Data) 830 548.0 0.2 Using the chemical composition and operating temperature adjustments described and calculated above, an interim chemistry factor is calculated for weld Heat # 10137 using the Calvert Cliffs Unit 2 and Millstone Unit 2 data. This calculation is shown on the following page in Table D-3 . WCAP-18 169-NP June 2018 Revision I CALC-AN02-EP-17-00002 Rev O Page 79 of 88 Westinghouse Non-Proprietary Class 3 D-5 Table D-3 Calculation of Weld Heat # 10137 Interim Chemistry Factor for the Credibility Evaluation Using Calvert Cliffs Unit 2 and Millstone Unit 2 Surveillance Capsule Data Capsule t<*l .1RTNDT (c) FF*.1RTNDT Material Capsule 19 FF<bl FF 2 (x 10 n/cm2, E > 1.0 MeV) (OF) {°F) Weld Metal Heat # 263° 0.825 0.9460 74.9 (72.7) 70.86 0.895 10137 (Calvert 97° 1.95 1.1825 84. 1 (82.9) 99.45 1.398 Cliffs Unit 2 Data) 104° 2.44 1.2401 69.9 (69.7) 86.68 1.538 Weld Metal Heat # 97° 0.324 0.6902 62.4 (65.93) 43.09 0.476 1013 7 (Millstone 104° 0.949 0.9853 51.9 (52.12) 51.16 0.971 Unit 2 Data) 830 1.740 1.1523 56.3 (56.09) 64.86 1.328 SUM: 416.10 6.606 2 CF Heat # 10137 = L(FF * ~RT NDT) .;- L(FF ) = (416.10) .;- (6.606) = 63.0°F Notes: (a) f = fluence ; (b) FF = fluence factor= t<0 *28 -O. IO'log I). (c) t>RT NOT values are the measured 30 ft-lb shift values. Each t>RT NOT value has been adjusted according to the temperature adjustments summarized in Table 0-2. The t>RT NOT values for each surveillance weld data point are not adjusted by the ratio procedure, because the mean chemical composition is identical to each capsule chemical composition (pre-adjusted values are listed in parentheses and were taken from Table 4-2). The scatter of ~RT NDT values about the functional form of a best-fit line drawn as described in Regulatory Guide 1.99, Revision 2, Position 2.1 [Ref. D-1] is presented in Table D-4. Table D-4 Best-Fit Evaluation for Surveillance Weld Metal Heat# 10137 Using Calvert Cliffs Unit 2 and Millstone Unit 2 Data CF Capsule f Measured Predicted Residual Material Capsule (Slopebest-fit) (x 10 19 n/cm2, FF .1RTNDT .1RTNDT .1RTNDT <28°F (Weld) (OF) E > 1.0 MeV) (OF) (OF) (OF) Weld Metal Heat# 263 ° 63 .0 0.825 0.9460 74.9 59.6 15.3 Yes 10137 (Calvert 97° 63.0 1.95 1.1825 84.l 74.5 9.6 Yes Cliffs Unit 2 Data) 104° 63.0 2.44 1.240 1 69.9 78.1 8.2 Yes Weld Metal Heat# 97° 63 .0 0.324 0.6902 62.4 43 .5 18.9 Yes 10137 (Millstone 104° 63.0 0.949 0.9853 51.9 62. 1 10.2 Yes Unit 2 Data) 830 63.0 1.740 1.1523 56.3 72.6 16.3 Yes The scatter of ~RTN DT values about the best-fit line, drawn as described in Regulatory Guide 1.99, Revision 2, Position 2.1 [Ref. D-1 ], should be less than 28°F for weld metal. Table D-4 indicates that 100% (six out of six) of the surveillance data points fall within the +/- la of 28°F scatter band for surveillance weld materials. Therefore, the surveillance weld material (Heat # 1013 7) is deemed "credible" per the third criterion when all available data is considered. WCAP- 18169-NP June 2018 Revision 1 CALC-AN02-EP-17-00002 Rev O Page 80 of 88 Westinghouse Non-Proprietary Class 3 D-6 In conclusion, the combined surveillance data from Calvert Cliffs Unit 2 and Millstone Unit 2 for weld Heat# 10137 may be applied to the Arkansas Nuclear One Unit 2 reactor vessel weld. The Position 2.1 chemistry factor calculation, as applicable to the Arkansas Nuclear One Unit 2 reactor vessel weld, is contained in Section 5. This Position 2.1 CF value could be used with a reduced margin term in the ART calculations contained in Section 7. However, consistent with the discussion in Section D.1 of this Appendix, the ART values calculated with the Position 2.1 CF value for weld Heat# 10137 utilize a full margin term for conservatism. D.3 REFERENCES D-1 U.S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials," May 1988. D-2 Westinghouse Report WCAP-1 8166-NP, Revision 0, "Analysis of Capsule 284° from the Entergy Operations, Inc. Arkansas Nuclear One Unit 2 Reactor Vessel Radiation Surveillance Program," September 2016. D-3 Westinghouse Report WCAP-16918-NP, Revision 1, "Analysis of Capsule V from the Southern Nuclear Operating Company Joseph M. Farley Unit 2 Radiation Surveillance Program," April 2008. D-4 Westinghouse Report WCAP-16012, Revision 0, "Analysis of Capsule W-83 from the Dominion Nuclear Connecticut Millstone Unit 2 Reactor Vessel Radiation Surveillance Program," February 2003. D-5 Code of Federal Regulations, 10 CFR 50.61, "Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events," Federal Register, Volume 60, No. 243, dated December 19, 1995, effective January 18, 1996. D-6 Westinghouse Report WCAP-17501-NP, Revision 0, "Analysis of Capsule 104° from the Calvert Cliffs Unit No. 2 Reactor Vessel Radiation Surveillance Program," February 2012. D-7 ASTM El85 -82, "Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels," ASTM, July 1982. D-8 K. Wichman, M. Mitchell, and A. Hiser, U.S. NRC, Generic Letter 92-01 and RPV Integrity Workshop Handouts, NRC/lndustry Workshop on RPV Integrity Issues, February 12, 1998. [ADAMS Accession Number MLI 10070570] WCAP-18169-NP June 2018 Revision I