ML15061A277

WCAP-17954-NP, Revision 0, Indian Point Unit 3 Heatup and Cooldown Limit Curves for Normal Operation.

ML15061A277
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Site:	Indian Point
Issue date:	12/31/2014
From:	Freed A, Hunter M Westinghouse
To:	Document Control Desk, Office of Nuclear Reactor Regulation
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ML15061A270	List: ML15061A277 NL-15-014, Indian Point, Unit 3 - Proposed Changes to Technical Specifications Regarding Reactor Vessel Heatup and Cooldown Curves and Low Temperature Overpressure Protection System Requirements NL-15-014, Proposed Changes to Technical Specifications Regarding Reactor Vessel Heatup and Cooldown Curves and Low Temperature Overpressure Protection System Requirements
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NL-15-014 WCAP-17954-NP, Rev. 0
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ENCLOSURE 2 TO NL-15-014 WCAP-17954-NP, Revision 0 "Indian Point Unit 3 Heatup and Cooldown Limit Curves for Normal Operation,"

Westinghouse Electric Co, December 2014.

(Non-Proprietary)

ENTERGY NUCLEAR OPERATIONS, INC.

INDIAN POINT NUCLEAR GENERATING UNIT NO. 3 DOCKET NO. 50-286

Westinghouse Nofl-ProprietarY~

Ciass 3 WCAP-1 954-NP Revision 70 December 2014 Indian Point Unit Cooldown Limit 3 HeatuP and Operation Curves for Normal

• )WestinlgiOuse

WESTINGHOUSE NON-PROPRIETARY CLASS 3 WCAP-17954-NP Revision 0 Indian Point Unit 3 Heatup and Cooldown Limit Curves for Normal Operation Amy E. Freed*

Materials Center of Excellence Melissa A. Hunter*

Radiation and Engineering Analysis December 2014 Reviewers: Elliot J. Long* Approved: Frank C. Gift*, Manager Materials Center of Excellence Materials Center of Excellence Nathan L. Glunt* John L. McFadden*, Manager Piping Analysis and Fracture Mechainics Piping Analysis and Fracture Mechanics Jesse J Klingensmith* Laurent P. Houssay*, Manager Radiation and Engineering Analysis Radiation and Engineering Analysis

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

Westinghouse Electric Company LLC 1000 Westinghouse Dr.

Cranberry Township, PA 16066

© 2014 Westinghouse Electric Company LLC All Rights Reserved

Westinghouse Non-Proprietary Class 3 ii RECORD OF REVISION Revision 0: Original Issue WCAP-17954-NP December 2014 Revision 0

Westinghouse Non-Proprietary Class 3 iii TABLE OF CONTENTS L IST O F TA B L ES ....................................................................................................................................... iv L IST O F F IG U R E S ..................................................................................................................................... vi EX EC UT IV E SU M M A RY ......................................................................................................................... vii I IN T R O D U C T ION ........................................................................................................................ 1-1 2 CALCULATED NEUTRON FLUENCE ..................................................................................... 2-1 2.1 IN T RO D U C T ION ........................................................................................................... 2-I 2.2 DISCRETE ORDINATES ANALYSIS ...................................................................... 2-1 2.3 CALCULATIONAL UNCERTAINTIES ........................................................................ 2-3 3 FRACTURE TOUGHNESS PROPERTIES ............................................................................. 3-1 4 SU RV E ILLA N C E DATA ............................................................................................................. 4-1 5 C H EM IST RY FA C TO R S ............................................................................................................. 5-1 6 CRITERIA FOR ALLOWABLE PRESSURE-TEMPERATURE RELATIONSHIPS ........ 6-1 6.1 OVERALLAPPROACH ............................................................................................ 6-1 6.2 METHODOLOGY FOR PRESSURE-TEMPERATURE LIMIT CURVE DEV E LO PM EN T ............................................................................................................ 6-1 6.3 CLOSURE HEAD/VESSEL FLANGE REQUIREMENTS ........................................... 6-5 7 CALCULATION OF ADJUSTED REFERENCE TEMPERATURE ......................................... 7-1 8 HEATUP AND COOLDOWN PRESSURE-TEMPERATURE LIMIT CURVES ....................... 8-1 9 R E F E R E N C E S ............................................................................................................................. 9-1 APPENDIX A THERMAL STRESS INTENSITY FACTORS (K1 t).................................... A-1 APPENDIX B REACTOR VESSEL INLET AND OUTLET NOZZLES ................................ B-I APPENDIX C NON-REACTOR VESSEL FERRITIC COMPONENTS .............................. C-1 WCAP- 17954-NP December 2014 Revision 0

Westinghouse Non-Proprietary Class 3 iv LIST OF TABLES Table 2-I Pressure Vessel M aterial Locations .................................................................................. 2-5 Table 2-2 Indian Point Unit 3 Calculated Neutron Fluence Projections at the Reactor Vessel Clad/Base Metal Interface at 36, 42, 48, and 54 EFPY ................................................... 2-6 Table 2-3 Indian Point Unit 3 Calculated Neutron Fluence at the Reactor Vessel Clad/Base Metal Interface for Cycles I through 19 and Future Projections ............................................... 2-7 Table 2-4 C alculational U ncertainties .............................................................................................. 2-7 Table 3-1 Summary of the Best-Estimate Cu and Ni Weight Percent and Initial RTNDT Values for the Indian Point Unit 3 Reactor Vessel M aterials .................................................................. 3-2 Table 4-1 Indian Point Unit 3 Surveillance Capsule Plate Data ...................................................... 4-2 Table 4-2 Diablo Canyon Unit I and Palisades Surveillance Capsule Data for Weld Heat 27204 ....

......................................................................................................................................... 4 -2 Table 4-3 H.B. Robinson Unit 2, Palisades, and Indian Point Units 2 and 3 Surveillance Capsule D ata for W eld H eat W 5214 .............................................................................................. 4-3 Table 5-1 Calculation of Indian Point Unit 3 Chemistry Factor Value for LS Plate B-2803-3 Using Surveillance C apsule Data ............................................................................................... 5-2 Table 5-2 Calculation of Indian Point Unit 3 Chemistry Factor Value for Weld Heat 27204 Using Surveillance C apsule Data ............................................................................................... 5-2 Table 5-3 Calculation of Indian Point Unit 3 Chemistry Factor Value for Weld Heat W52 14 Using Surveillance C apsule Data ............................................................................................... 5-3 Table 5-4 Summary of Indian Point Unit 3 Positions 1.1 and 2.1 Chemistry Factors ..................... 5-4 Table 7-1 Fluence Values and Fluence Factors for the Vessel Surface, 1/4T and 3/4T Locations for the Indian Point Unit 3 Reactor Vessel Materials at 37 EFPY ......................................... 7-3 Table 7-2 Adjusted Reference Temperature Evaluation for the Indian Point Unit 3 Reactor Vessel Beltline Materials through 37 EFPY at the 1/4T Location .............................................. 7-4 Table 7-3 Adjusted Reference Temperature Evaluation for the Indian Point Unit 3 Reactor Vessel Beltline Materials through 37 EFPY at the 3/4T Location .............................................. 7-5 Table 7-4 Summary of the Limiting ART Values Used in the Generation of the Indian Point Unit 3 Heatup and Cooldown Curves at 37 EFPY ...................................................................... 7-6 Table 8-1 Indian Point Unit 3 37 EFPY Heatup Curve Data Points using the 1998 through the 2000 Addenda App. G Methodology (w/ K1,, w/ Flange Notch, and w/o Margins for Instrum entation Errors) .................................................................................................... 8-5 Table 8-2 Indian Point Unit 3 37 EFPY Cooldown Curve Data Points using the 1998 through the 2000 Addenda App. G Methodology (w/ KI., w/ Flange Notch, and w/o Margins for Instrum entation Errors) .................................................................................................... 8-7 WCAP- 17954-NP December 2014 Revision 0

Westinghouse Non-Proprietary Class 3 V Table A-I Kit Values for Indian Point Unit 3 at 37 EFPY 100 0F/hr Heatup Curves (w/ Flange Requirements and w/o Margins for Instrument Errors) .................................................. A-2 Table A-2 Krt Values for Indian Point Unit 3 at 37 EFPY I00°F/hr Cooldown Curves (w/ Flange Requirements and w/o Margins for Instrument Errors) .................................................. A-3 Table B- I ART Calculations for the Indian Point Unit 3 Reactor Vessel Nozzle Materials Using Conservative Fluence Bounding 54 EFPY ................................................................. B-4 Table B-2 Summary of the Limiting ART Values for the Indian Point Unit 3 Inlet and Outlet Nozzle M aterials ......................................................................................................................... B-5 WCAP- 17954-NP December 2014 Revision 0

Westinghouse Non-Proprietary Class 3 vi LIST OF FIGURES Figure 2-1 Indian Point Unit 3 r-0 Reactor Geometry at the Core Mid-plane ................................... 2-8 Figure 2-2 Indian Point Unit 3 r-z Reactor Geometry ....................................................................... 2-9 Figure 8-1 Indian Point Unit 3 Reactor Coolant System Heatup Limitations (Heatup Rates of 60 and 100 0 F/hr) Applicable for 37 EFPY (with Flange Requirements and without Margins for Instrumentation Errors) using the 1998 through the 2000 Addenda App. G Methodology (w / K i.) ............................................................................................................................. 8-3 Figure 8-2 Indian Point Unit 3 Reactor Coolant System Cooldown Limitations (Cooldown Rates of 0, -20, -40, -60, and -100 0 F/hr) Applicable for 37 EFPY (with Flange Requirements and without Margins for Instrumentation Errors) using the 1998 through the 2000 Addenda A pp. G M ethodology (w /K jj .......................................................................................... 8-4 Figure B-1 Comparison of Indian Point Unit 3 Beltline P-T Limits to Inlet Nozzle Limits ............. B-7 Figure B-2 Comparison of Indian Point Unit 3 Beltline P-T Limits to Outlet Nozzle Limits .......... B-8 WCAP- 17954-NP December 2014 Revision 0

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 Indian Point Unit 3 reactor vessel. The heatup and cooldown P-T limit curves were generated using the limiting Adjusted Reference Temperature (ART) values for Indian Point Unit 3. The limiting ART values were those of Lower Shell Plate B-2803-3 (using credible surveillance data) at both 1/4 thickness (1/4T) and 3/4 thickness (3/4T) locations. The P-T limit curves were generated using the K1I 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 37 Effective Full Power Years (EFPY) using heatup rates of 60 and I00 0 F/hr, and cooldown rates of 0, -20, -40, -60, and -I 00°F/hr. The curves were developed with the flange 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 37 EFPY.

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 corner. As discussed in Appendix B, the P-T limit curves generated based on the limiting cylindrical beltline material (Lower Shell Plate B-2803-

3) bound the P-T limit curves for the reactor vessel inlet and outlet nozzles for Indian Point Unit 3 at 37 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.

December 2014 WCAP- 17954-NP WCAP-17954-NP December 2014 Revision 0

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 RTNDT 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 ARTNDT, 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 607F.

RTNDT 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, ARTNDT due to the radiation exposure associated with that time period must be added to the unirradiated RTNDT (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) + ARTNDT + margins for uncertainties) at the 1/4T and 3/4T locations, where T is 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 and the NRC-approved methodology documented in WCAP-14040-A, Revision 4

[Ref. 2]. Specifically, the K1, methodology of the 1998 through the 2000 Addenda Edition of ASME Code,Section XI, Appendix G [Ref. 3] was used. The K1, curve is a lower bound static fracture toughness obtained from test data gathered from several different heats of pressure vessel steel. The limiting material is indexed to the K1, 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 Indian Point Unit 3 heatup and cooldown P-T limit curves for 37 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 37 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. 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 Indian Point Unit 3 at 37 EFPY. Discussion of the other non-reactor vessel ferritic RCPB components relative to P-T limits is contained in Appendix C.

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Westinghouse Non-Proprietary Class 3 2-1 2 CALCULATED NEUTRON FLUENCE

2.1 INTRODUCTION

A discrete ordinates SN transport analysis was performed for the Indian Point Unit 3 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 first three surveillance capsules is provided in Reference 5. The dosimetry analysis documented in Reference 5 showed that the +/-20% (1a) acceptance criteria specified in Regulatory Guide 1.190 [Ref. 6] is met. The results of this analysis are consistent with those of Reference 5 within the uncertainty of the methodology. Therefore, the acceptance criterion continues to be 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 ENDF/B-VI.3 and made use of the latest available calculational tools. Furthermore, the neutron transport evaluation methodologies follow the guidance of Regulatory Guide 1.190 [Ref. 6]. 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 Indian Point Unit 3 reactor vessel, a series of fuel-cycle-specific forward transport calculations were carried out using the following three-dimensional flux synthesis technique:

(p(r, 0, z) = (p(r, 0) x (p(r) where (p(r,0,z) is the synthesized three-dimensional neutron flux distribution, qp(r,0) is the transport solution in rO geometry, (p(r,z) is the two-dimensional solution for a cylindrical reactor model using the actual axial core power distribution, and (p(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 Indian Point Unit 3.

For the Indian Point Unit 3 transport calculations, the r,O model depicted in Figure 2-1 was utilized since the reactor is octant symmetric. This rO model includes the core, the reactor internals, the thermal shield.-

including explicit representations of the surveillance capsules at 40 and 400 - the pressure vessel cladding and vessel wall, the insulation external to the pressure vessel, and the primary biological shieldwall.The symmetric rO model was utilized to perform both the surveillance capsule dosimetry evaluations, subsequent comparisons with calculated results, and to generate the maximum fluence levels at the pressure vessel wall. In developing this analytical model, nominal design dimensions were employed for the various structural components. Likewise, 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 coolant densities were treated on a fuel-cycle-specific basis. The reactor core itself was WCAP- 17954-NP December 2014 Revision 0

Westinghouse Non-Proprietary Class 3 2-2 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 consisted of 170 radial by 67 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 flux convergence criterion utilized in the r,0 calculations was set at a value of 0.001.

The r,z model used for the Indian Point Unit 3 calculations is shown in Figure 2-2. The model extends radially from the centerline of the reactor core out to a location interior to the primary biological shield and over an axial span from an elevation approximately one foot below to five feet above the active fuel.

As in the case of the r,0 models, nominal design dimensions and full-power coolant densities were employed in the calculations. In this case, the homogenous core region was treated as an equivalent cylinder with a volume equal to that of the active core zone. The stainless steel former plates located between the core baffle and core barrel regions were also explicitly included in the model. The r,z geometric mesh description of these reactor models consisted of 153 radial by 148 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 flux convergence criterion utilized in the rz calculations was also set at a value of 0.001.

The one-dimensional radial model used in the synthesis procedure consisted of the same 153 radial mesh intervals included in the rz model. Thus, radial synthesis factors could be determined on a meshwise basis throughout the entire geometry.

The data utilized for the core power distributions in plant-specific transport analyses included cycle-dependent fuel assembly initial enrichments, burnups, and axial power distributions. 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 flux, 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 bumup 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 carried out using the' DORT discrete ordinates code [Ref. 7] and the BUGLE-96 cross-section library [Ref. 8]. 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 P5 Legendre expansion and angular discretization was modeled with an S16 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. .

In Table 2-1, locations of the Indian Point Unit 3 vessel welds and plates are provided. The axial position of each material is indexed to z = 0.0 cm, which corresponds to the mid-plane of the active fuel stack.

Selected results from the neutron transport analyses are provided in Tables 2:-2 and 2-3.. In Table 2-2, calculated fast neutron (E > 1.0 MeV) fluence for reactor vessel materials, on the pressure vessel WCAP- 17954-NP . December2014 Revision 0

Westinghouse Non-Proprietary Class 3 2-3 clad/base metal interface, is provided at future projections to 36, 42, 48 and 54 EFPY. Cycle-specific calculations were performed for Cycles I through 19, where a core thermal power of 3025.0 MWt was used in Cycles I through 11, 3030.6 MWt was used in Cycle 12, 3067.4 MWt was used in Cycle 13, 3180.0 MWt was used in Cycle 14, and 3188.4 MWt was used in Cycles 15 through 19. The projections were based on the assumption that the core power distributions and associated plant operating characteristics from Cycles 14 through 19 were representative of future plant operation. In Table 2-3, calculated fast neutron (E > 1.0 MeV) fluence on the pressure vessel clad/base metal interface is provided for Cycles I through 19 and future projections at various azimuthal locations.

2.3 CALCULATIONAL UNCERTAINTIES The uncertainty associated with the calculated neutron exposure of the Indian Point Unit 3 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:

I. 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 Indian Point Unit 3 surveillance program.

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 Indian Point Unit 3 analysis was established from results of these three phases of the methods qualification.

The fourth phase of the uncertainty assessment (comparisons with Indian Point Unit 3 measurements) was used solely to demonstrate the validity of the transport calculations andto 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 2-4 summarizes the uncertainties developed from the first three phases of the methodology qualification. Additional information pertinent to these evaluations is provided. in Reference 2. The net calculational uncertainty was determined by combining the individual components in quadrature.

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Westinghouse Non-Proprietary Class 3 2-4 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 Section 7.5 of Reference 5 support these uncertainty assessments for Indian Point Unit 3.

17954-NP December 2014 WCAP-WCAP-17954-NP December 2014 Revision 0

Westinghouse Non-Proprietary Class 3 2-5 Westinghouse Non-Proprietary Class 3 2-5 Table 2-1 Pressure Vessel Material Locations Azimuthal Material Axial Location Location (cm) (degrees)

Outlet Nozzle Forging to Upper Shell Welds - Lowest Extent Nozzle 1(a) 272.57 22.0 Nozzle 2(a" 272.57 158.0 Nozzle 3Pa) 272:57 202.0 Nozzle 4(a" 272.57 338.0 Inlet Nozzle Forging to Upper Shell Welds - Lowest Extent Nozzle 1(a) 267.49 67.0 Nozzle 2(a) 267.49 113.0 Nozzle 3Pa) 267.49 247.0 Nozzle 4(ý) 267.49 293.0 Upper Shell Plate 1 237.81 to 485.93 7.0 to 127.0 Plate 2 237.81 to 485.93 127.0 to 247.0 Plate 3 237.81 to 485.93 247.0 to 7.0 Upper Shell Axial Welds Weld 1 237.81 to 485.93 7.0 Weld 2 237.81 to 485.93 127.0 Weld 3 237.81 to 485.93 247.0 Upper Shell to Intermediate Shell Circumferential Weld 234.47 to 237.81 0.0 to 360.0 Intermediate Shell Plate 1 -38.66 to 234.47 300.0 to 60.0 Plate 2 -38.66 to 234.47 60.0 to 180.0 Plate 3 -38.66 to 234.47 180.0 to 300.0 Intermediate Shell Axial Welds Weld 1 -38.66 to 234.47 60.0 Weld 2 -38.66 to 234.47 180.0 Weld 3 -38.66 to 234.47 300.0 Intermediate Shell to Lower Shell Circumferential Weld -41.99 to -38.66 0.0 to 360.0 Lower Shell Plate 1 -309.25 to -41.99 345.0 to 105.0 Plate 2 -309.25 to -41.99 105.0 to 225.0 Plate 3 -309.25 to -41.99 225.0 to 345.0 Lower Shell Axial Welds Weld 1 -309.25 to -41.99 105.0 Weld 2 -309.25 to -41.99 225.0 Weld 3 - -309.25 to -41.99 345.0 Lower Shell to Lower Vessel Head Circumferential -309.25 0.0 to 360.0 Weld Note:

(a) Lowest extent WCAP- 17954-NP December 2014 Revision 0

Westinghouse Non-Proprietary Class 3 2-6 Table 2-2 Indian Point Unit 3 Calculated Neutron Fluence Projections at the Reactor Vessel Clad/Base Metal Interface at 36, 42, 48, and 54 EFPY Fluence(a)

Reactor Vessel Material (n/cm 2, E > 1.0 MeV) 36 EFPY 42 EFPY 48 EFPY 54 EFPY Outlet Nozzle to Upper Shell Welds (Lowest Extent) 1, 2, 3, 1.46E+16 1.73E+16 1.99E+16 2.26E16 and 4 Inlet Nozzle to Upper Shell Welds (Lowest Extent) 1.97E+16 2.33E+16 2.69E+16 3.06E+16 1, 2, 3, and 4 Upper Shell Plates 1.47E+17 1.73E+17 1.98E+17 2.24E+17 Upper Shell Axial Weld 1 (b) 7.45E+16 8.73E+16 1.00E+17 1.13E+17 Upper Shell Axial Weld 2 (b) 1.35E+17 1.59E+17 1.83E+17 2.07E+17 Upper Shell Axial Weld 3 (b1 1.22E+17 1.43E+17 1.65E+17 1.87E+17 Upper Shell to Intermediate Shell Circumferential Weld 1.89E+17 2.22E+17 2.54E+17 2.87E+17 Intermediate Shell Plates 1.22E+19 1.42E+19 1.61E+19 1.81E+19 Intermediate Shell Axial Welds I and 3 (b) 9.87E+18 1.15E+19 1.31E+19 1.48E+19 Intermediate Shell Axial Weld 1b) 2 5.49E+18 6.34E+18 7.19E+18 8.04E+18 Intermediate Shell to Lower Shell Circumferential Weld 1.22E+19 1.41E+19 1.61E+19 1.80E+19 Lower Shell Plates 1.22E+19 1.41E+19 1.61E+19 1.80E+19 Lower Shell Axial Welds I and 3 b) 8.53E+18 9.88E+18 1.12E+19 1.26E+19 Lower Shell Axial Weld 2 (b) 1.22E+19 1.41E+19 1.61E+19 1.80E+19 Notes: -

(a) Extended beltline materials are currently interpreted to be the reactor vessel materials that will be exposed to a neutron fluence greater than or equal to I x l0ol n/cm 2 (E > 1.0 MeV). Only the materials that are projected to experience a fluence value of at least I x 1017 n/cm 2 (E > 1.0 MeV) will be included in the subsequent evaluations contained within this report.

(b) The fluence value at each individual weld location (1, 2, 3) was reported here for documentation purposes; however, the maximum fluence value across all three axial welds in each reactor vessel shell will be used as the bounding fluence value in the subsequent calculations contained within this report.

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Westinghouse Non-Proprietary Class 3 2-7 Table 2-3 Indian Point Unit 3 Calculated Neutron Fluence at the Reactor Vessel Clad/Base Metal Interface for Cycles 1 through 19 and Future Projections Cumulative Fluence (n/cm', E > 1.0 MeV)

CyIle Cycle Time

!D (EP)00 150 3f0° 450 (EFPY) 1 1.4 2.58E+17 4.11E+17 5.15E+17 7.68E+17 2 2.3 4.47E+17 7.11E+17 8.96E+17 1.38E+18 3 3.2 6.79E+17 1.08E+18 1.32E+18 1.98E+18 4 4.4 9.29E+17 1.44E+18 1.71E+18 2.51E+18 5 5.5 1.15E+18 1.76E+18 2.05E+18 2.99E+18 6 6.7 1.36E+18 2.08E+18 2.39E+18 3.43E+18 7 7.8 1.53E+18 2.38E+18 2.69E+18 3.75E+18 8 8.9 1.75E+18 2.71E+18 2.97E+18 4.10E+18 9 10.5 2.01E+18 3.07E+18 3.33E+18 4.52E+18 10 12.3 2.24E+18 3.42E+18 3.71E+18 5.00E+18 1I 13.7 2.41E+18 3.69E+18 4.02E+ 18 5.35E+18 12 15.5 2.63E+18 4.01E+18 4.42E+18 5.86E+18 13 17.4 2.86E+18 4.37E+18 4.81E+18 6.29E+18 14 19.3 3.12E+18 4.79E+18 5.32E+18 6.85E+18 15 21.1 3.39E+18 5.23E+18 5.86E+18 7.47E+18 16 22.9 3.64E+18 5.62E+18 6.33E+18 8.09E+18 17 24.8 3.90E+18 6.04E+18 6.85E+18 8.66E+18 18 26.7 4.18E+18 6.47E+18 7.33E+18 9.26E+18 19 28.6 4.45E+ 18 6.91E+18 7.87E+18 9.88E+18 36.0 5.49E+18 8.57E+18 9.87E+18 1.23E+19 42.0 6.34E+18 9.92E+18 1.15E+19 1.42E+19 48.0 7.19E+18 1.13E+19 1.31E+19 1.61E+19 54.0 8.04E+18 1.26E+19 1.48E+19 1.81E+19 Table 2-4 Calculational Uncertainties Uncertainty Description Capsule Vessel IR PCA Comparisons 3% 3%

H. B. Robinson Comparisons 3% 3%

Analytical SensitivityStudies o 10% 11%

Additional Uncertainty for Factors not Explicitly Evaluated 5% 5%

Net Calculational Uncertainty 12% 13%

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Westinghouse Non-Proprietary Class 3 2-8 Westinghouse Non-Proprietary Class 3 2-8 Indian Point 3 Reactor R,T Model Meshes: 170R, 678 OR=~uRegiuw Dmowmri siAnu Steel

=

Ai Concret.

M IN - Cwbunsf C D-The stainless steel regions include the core baffle, core barrel, thermal shield, and vessel clad 5 263.8 290.2 316.6 3 R Figure 2-1 Indian Point Unit 3 r-O Reactor Geometry at the Core Mid-plane WCAP-1 7954-NP December 2014 Revision 0

Westinghouse Non-Proprietary Class 3 2-9 Indian Point Unit 3 Reactor R,Z Model Meshes: 153X,148Y CoreRegim - Region owa~omr - Si*hm Steel Al' Comwef - WY- Cwbn f

-Y1PlowW~on Fuel FhIa andEnd G

TopNzzh ui Rgd UKFRe*A OJU PmmoRei.

Y 0

61 The stainlesssteel regions include the core baffle, formers, core barrel,thermal shield, pressure vessel clad, E

)N: and upper core plate B, which is located below outlet plenum B.

43.0 Figure 2-2 Indian Point Unit 3 r-z Reactor Geometry December 2014 17954-NP WCAP- 17954-NP December 2014 Revision 0

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 consideredin the selection of the most limiting materialwith regardto radiationdamage."

The Indian Point Unit 3 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. 9], any reactor vessel materials that are predicted to experience a neutron fluence exposure greater than 1.0 x 1017 n/cm 2 (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 nozzle shell plates and associated welds.

The reactor vessel inlet and outlet nozzle materials experience a neutron fluence exposure less than 1.0 x 1011 n/cm 2 (E > 1.0 MeV) at 37 EFPY; however, per NRC RIS 2014-11, these materials are still evaluated for their potential effect on P-T limit curves - See Appendix B for more details.

Summary of the best-estimate copper (Cu) and nickel (Ni) contents, in units of weight percent (wt. %),as well as initial RTNDT values for the reactor vessel beltline and extended beltline materials are provided in Table 3-1 for Indian Point Unit 3.

. WCAP-17954-NP December 2014 Revision 0

Westinghouse Non-Proprietary Class 3 3-2 Table 3-1 Summary of the Best-Estimate Cu and Ni Weight Percent and Initial RTNDT Values for the Indian Point Unit 3 Reactor Vessel Materials Fracture Toughness Property Chemical Composition Reactor Vessel Material Heat Property and Identification Number Number Cu Ni I (~t. %)

(Wt.  % (Wt %)

I(Wt.  %) Initial RTNDT(C) (0 F)

Reactor Vessel Extended Beltline Materials(a)

Nozzle Shell (NS) Plate B-2801-1 B5391-1 0.21 0.48 34 NS Plate B-2801-2 B5394-1 0.20 0.50 44 NS Plate B-2801-3 A0516-1 0.22 0.54 3 NS Axial Weld Seams 1-042A, B, C W5214 0.213 1.007 -56 NS to IS Circumferential Weld Seam 8-042 27204 0.203 1.018 -56 Reactor Vessel Beltline Materials(b)

Intermediate Shell (IS) Plate B-2802-1 B5394-2 0.20 0.50 5 IS Plate B-2802-2 A0516-2 0.22 0.53 -4 IS Plate B-2802-3 B5391-2 0.20 0.49 17 Lower Shell (LS) Plate B-2803-1 A0495-2 0.19 0.47 49 LS Plate B-2803-2 C 1397-3 0.22 0.52 -5 LS Plate B-2803-3 A0512-2 0.24 0.52 74 IS Axial Weld Seams 2-042A, B, C 34B009 0.192 1.007 -56 LS Axial Weld Seams 3-042A, B, C 34B009 0.192 1.007 -56 IS to LS Circumferential Weld Seam 9-042 13253 0.221 0.732 -54 Surveillance Weld Data(d)

Diablo Canyon Unit 1 0.198 0.999 - - -

27204 Palisades 0.194 1.067 - - -

H.B. Robinson Unit 2 0.34 0.66 -- -

Palisades 0.307 1.045 - - -

W5214 Indian Point Unit 2 0.20 1.03 -- -

Indian Point Unit 3 0.16 1.12 ---

Notes:

(a) Information for the extended beltline plate and weld materials was taken from Entergy letter NL-08-143 [Ref. 10].

(b) Information for the beltline plate and weld materials was taken from Entergy letter NL-08-014 [Ref. II].

(c) The initial RTNDT values for the plate materials as well as the initial RTNDT value for the IS to LS circumferential weld seam 9-042 are based on measured data. The initial RTNDT values for the extended beltline welds and the beltline axial welds are generic values from 10 CFR 50.61 [Ref. 12].

(d) Surveillance data exists for weld heat numbers 27204 and W5214 from multiple sources; See Section 4 for more details. The data for Diablo Canyon Unit I and Palisades surveillance weld metal heat 27204 was taken from Table A-2 of Structural Integrity Associates (SIA), Inc. Report No. 1000915.401, Revision I [Ref. 131. The data for H, B.

Robinson Unit 2, Palisades and Indian Point Units 2 and 3 for surveillance weld metal heat W5214 was taken from Table 6 of SIA, Inc. Report No. 0901132.401, Revision 0 [Ref. 14].

WCAP-17954-NP December 2014 Revision 0

Westinghouse Non-Proprietary Class 3 4-1 4 SURVEILLANCE DATA Per Regulatory Guide 1.99, Revision 2, 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 the same limiting 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 Indian Point Unit 3 is from Lower Shell Plate B-2803-3. Table 4-1 summarizes the Indian Point Unit 3 surveillance data for the plate material that will be used in the calculation of the Position 2.1 chemistry factor value. The results of the last withdrawn and tested surveillance capsule, Capsule X, were documented in WCAP-16251-NP [Ref. 15]. Appendix D of WCAP- 1625 1-NP concluded that the surveillance plate data is credible; therefore, a reduced margin term will be utilized in the ART calculations contained in Section 7.

The Indian Point Unit 3 reactor vessel NS to IS circumferential weld seam was fabricated using weld heat 27204. Weld heat 27204 is contained in the Diablo Canyon Unit I surveillance program along with the two supplemental surveillance capsules at Palisades. Thus, the Diablo Canyon Unit I and Palisades data will be used in calculation of the Position 2.1 chemistry factor value for Indian Point Unit 3 weld heat 27204. Table 4-2 summarizes the applicable surveillance capsule data pertaining to weld heat 27204.

Furthermore, Appendix A of SIA, Inc. Report No. 1000915.401, Revision I [Ref. 13] concluded that the combined surveillance data for weld heat 27204 is credible; therefore, a reduced margin term will be utilized in the ART calculations contained in Section 7.

The Indian Point Unit 3 reactor vessel NS axial weld seams were fabricated using weld heat W5214.

Weld heat W5214 is contained in the H. B. Robinson Unit 2 and Indian Point Units 2 and 3 surveillance programs along with the two supplemental surveillance capsules at Palisades. Thus, the H.B. Robinson Unit 2, Palisades, and Indian Point Units 2 and 3 data will be used in calculation of the Position 2.1 chemistry factor value for Indian Point Unit 3 weld heat W5214. Table 4-3 summarizes the applicable surveillance capsule data pertaining to weld heat W5214. Furthermore, Section 5 of SIA, Inc. Report No.

0901132.401, Revision 0 [Ref. 14] concluded that the combined surveillance data for weld heat W5214 is not fully credible; therefore, a full margin term will be utilized in the ART calculations contained in Section 7.

WCAP- 17954-NP December 2014 Revision 0

Westinghouse Non-Proprietary Class 3 4 -2 Westinghouse Non-Proprietary Class 3 4-2 Table 4-1 Indian Point Unit 3 Surveillance Capsule Plate Data Capsule Fluence(a) Measured 30 ft-lb Transition Material Capsule~ (x0~9 n/cm 2, E > 1.0 MeV) Temperature Shift(a) (IF)

T 0.263 139.4 Lower Shell Plate B-2803-3 (Longitudinal) Z 1.04 167.8 X 0.874 159.6 T 0.263 105.9 Y 0.692 148.9 Lower Shell Plate B-2803-3 (Transverse)

Z 1.04 157.9 X 0.874 158.2 Note:

(a) Surveillance plate data was taken from Table 5-10 ofWCAP-16251-NP [Ref. 15].

Table 4-2 Diablo Canyon Unit 1 and Palisades Surveillance Capsule Data for Weld Heat 27204 Capsule Fluence(a) Measured 30 ft-lb Transition Inlet Temperature Material Capsule(a) (xl0 19 n/cm 2, E > 1.0 MeV) Temperature Shiftd') (OF) Temperature(a) Adjustment(c)

(OF) (OF)

S 0.283(b, 110.79 544 4.0 Diablo Canyon Y 1.05 232.59 542 2.0 Unit 1 Data V 1.3 6 "b) 201.07 541.5 1.5 SA-60-1 1.50 253.1 535 -5.0 Palisades Data SA-240-1 2.38 267.8 535.7 -4.3 Notes:

(a) Data pertaining to the Diablo Canyon Unit I and Palisades surveillance weld heat 27204 were taken from Table A-2 of SIA, Inc. Report No. 1000915.401, Revision I [Ref. 13].

(b) The fluence values for Diablo Canyon Unit 1 Capsules S and V were updated in WCAP-17315-NP, Revision 0 [Ref. 16].

(c) Temperature adjustment = 1.0*(Tcapsul. - Tpoam), where Tplant = 540'F for Indian Point Unit 3 (applied to the weld ARTNDT data for each of the Diablo Canyon Unit I and Palisades capsules in the Position 2.1 chemistry factor calculation - See Section 5 for more details).

WCAP- 17954-NP December 2014 Revision 0

Westinghouse Non-Proprietary Class 3 4-3 Table 4-3 H.B. Robinson Unit 2, Palisades, and Indian Point Units 2 and 3 Surveillance Capsule Data for Weld Heat W5214 Capsule Fluence(a) Measured 30 ft-lb Transition Inlet - Temperature Material Capsulet a) (xlO' 9 n/cm 2, E > 1.0 MeV) Temperature Shift(")(°F) Temperature(')(F) Adjustment(b) (0F)

T 3.87 289.1 547 7.0 H. B. Robinson Unit 2 Data V 0.530 208.8 547 7.0 X 4.49 265.6 547 7.0 SA-60-1 1.50 259 535.0 -5.0 Palisades Data SA-240-1 2.38 280.1 535.7 -4.3 V 0.492 197.5 524.0 -16.0 Indian Point Unit 2 Data Y 0.455 193.9 529.1 -10.9 T 0.263 149.8 539.4 -0.6 Y 0.692 171.1 539.5 -0.5 Indian Point Unit 3 Data Z 1.04 228.3 538.9 -1.1 X 0.874 192.5 539.7 -0.3 Notes:

(a) Data pertaining to the H.B. Robinson Unit 2, Palisades, and Indian Point Units 2 and 3 surveillance weld heat W5214 were taken from Table 6 of SIA, Inc. Report No.

0901132.401, Revision 0 [Ref. 14].

(b) Temperature adjustment = 1.0*(Tc,,pu~ - Tplan), where Tpla, = 540'F for Indian Point Unit 3 (applied to the weld ARTNDT data for each of the H.B. Robinson Unit 2, Palisades, and Indian Point Units 2 and 3 capsules in the Position 2.1 chemistry factor calculation - See Section 5 for more details).

WCAP- 17954-NP December 2014 Revision 0

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 I and 2 of Regulatory Guide 1.99, Revision

2. The best-estimate copper and nickel weight percent values for the Indian Point Unit 3 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 Indian Point Unit 3 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 5-3 for the Indian Point Unit 3 reactor vessel materials that are 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. 17]. 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 5-4 for Indian Point Unit 3.

December 2014 WCAP- 17954-NP WCAP- 17954-NP December 2014 Revision 0

Westinghouse Non-Proprietary Class 3 5-2 Table 5-1 Calculation of Indian Point Unit 3 Chemistry Factor Value for LS Plate B-2803-3 Using Surveillance Capsule Data LS Plate B-2803-3 Capsule f' FF(b) ARTNDT,(c) FF*ARTNDI 2 19 2 (OF) F DataBCapsule (x10 n/cm , E > 1.0 MeV) (OF)

T 0.263 0.637 139.4 88.76 0.405 Longitudinal Z 1.04 1.011 167.8 169.64 1.022 Orientation X 0.874 0.962 159.6 153.57 0.926 T 0.263 0.637 105.9 67.43 0.405 Transverse Y 0.692 0.897 148.9 133.53 0.804 Orientation z 1.04 1.011 157.9 159.63 1.022 X 0.874 0.962 158.2 152.23 0.926 SUM: 924.78 5.511 2 0 CFLs Plate B-2803-3 E(FF

• ARTNDT) (FF ) = (924.78) + (5.511) = 167.8 F Notes:

(a) f= fluence.

(b) FF = fluence factor = f0o.28 -0.o*log 0 (c) ARTNDT 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 Indian Point Unit 3 Chemistry Factor Value for Weld Heat 27204 Using Surveillance Capsule Data Weld Metal Heat Capsule Capsule FF (b) ARTNDToC) FF*ARTNT FF2 27204 (xl019 n/cm 2 , E > 1.0 MeV) F (0F) (OF)

S 0.283 0.655 117.09 (110.79) 76.73 0.429 Diablo Canyon Unit 1 Data Y 1.05 1.014 239.28 (232.59)

Data 242.55 1.027 V 1.36 1.085 206.62 (201.07) 224.28 1.178 SA-60-1 1.50 1.112 248.10 (267.80) 275.95 1.237 Palisades Data SA-240-1 2.38 1.234 263.50 (253.10) 325.12 1.522 SUM: 1144.62 5.395 2

CFWeld Heat 27204 (FF

• ARTNDT) + Z(FF ) = (1144.62) - (5.395) = 212.2-F Notes:

(a) f= fluence.

(b) FF = fluence factor = 0.28. 0-.olog f).

(c) ARTNDT values are the measured 30 ft-lb shift values. The ARTNDT 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 Diablo Canyon Unit I surveillance 0 0 data = CFvqse, Weld / CFsur Weld = 226.8 F / 222.26 F = 1.02. Ratio applied to the Palisades surveillance data = CFvessel 0

Weld / CFsu 0 v Weld = 226.87F / 227.8 F = 1.0.

WCAP- 17954-NP December 2014 Revision 0

Westinghouse Non-Proprietary Class 3 5-3 Table 5-3 Calculation of Indian Point Unit 3 Chemistry Factor Value for Weld Heat W5214 Using Surveillance Capsule Data Weld Metal Heat Capsule Capsule 0"' FF(b) ARTNDT(c) FF*ARTNDT FFE W5214 (xl01 n/cm 2, E > 1.0 MeV) 9 (OF) (OF)

T 3.87 1.349 313.87 (289.1) 423.41 1.820 H. B. Robinson V 0.530 0.823 228.75 (208.8) 188.17 0.677 X 4.49 1.381 288.96 (265.6) 398.95 1.906 SA-60-1 1.50 1.112 220.98 (259.0) 245.79 1.237 SA-240-1 2.38 1.234 239.95 (280.1) 296.06 1.522 Indian Point Unit 2 V 0.492 0.802 185.13 (197.5) 148.50 0.643 Data Y 0.455 0.781 186.66 (193.9) 145.75 0.610 T 0.263 0.637 167.10 (149.8) 106.40 0.405 Indian Point Unit 3 Y 0.692 0.897 191.07 (171.1) 171.34 0.804 Data Z 1.04 1.011 254.46 (228.3) 257.26 1.022 X 0.874 0.962 215.26 (192.5) 207.13 0.926 SUM: 2588.76 11.573 CFWeld Heat W5214 = I(FF

• ARTNDT) + E(FF 2) = (2588.76) +(11.573) = 223.7 0F Notes:

(a) f= fluence.

(b) FF = fluence factor = 28- 0. 10*log f.

(c) ARTNDT values are the measured 30 ft-lb shift values. The ARTNDT 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 adjustments are listed in Table 4-3. Ratio applied to the H.B. Robinson Unit 2 surveillance data = CFvessel Weld / CFsurv Weld = 230.7 0 F / 217.7 0 F = 1.06. Ratio applied to the Palisades surveillance data = CFvessel Weld / CFsur, Weld 230.70 F / 266.5 0 F = 0.87. Ratio applied to, the Indian Point Unit 2 surveillance data = CFvessel Weld /

CFsurv Weld = 230.7-F / 226.30 F = 1.02. Ratio applied to the Indian Point Unit 3 surveillance data = CFvessel Weld / CFsr¢ 0 0 Weld = 230.7 F / 206.2 F = 1.12.

WCAP-1 7954-NP December 2014 Revision 0

Westinghouse Non-Proprietary Class 3 5-4 Table 5-4 Summary of Indian Point Unit 3 Positions 1.1 and 2.1 Chemistry Factors Heat Chemistry Factor (F)

Reactor Vessel Material and Identification Number Position 1.1l Position 2.1 Reactor Vessel Extended Beltline Materials NS Plate B-2801-1 B5391-1 139.4 ---

NS Plate B-2801-2 B5394-1 137.0 ---

NS Plate B-2801-3 A0516-1 152.9 ---

NS Axial Weld Seams 1-042A, B, C W5214 230.7 223.7")

NS to IS Circumferential Weld Seam 8-042 27204 226.8 212.2(c)

Reactor Vessel Beltline Materials IS Plate B-2802-1 B5394-2 137.0 ---

IS Plate B-2802-2 A0516-2 151.6 -- -

IS Plate B-2802-3 B5391-2 135.8 ---

LS Plate B-2803-1 A0495-2 127.7 ---

LS Plate B-2803-2 C 1397-3 150.2 -- -

LS Plate B-2803-3 A0512-2 160.4 167.8(')

IS Axial Weld Seams 2-042A, B, C 34B009 221.3 -- -

LS Axial Weld Seams 3-042A, B, C 34B009 221.3 ---

IS to LS Circumferential Weld Seam 9-042 13253 189.1 -- -

Surveillance Weld Data Diablo Canyon Unit I 222.26 ---

27204 Palisades 227.8 - --

H.B. Robinson Unit 2 217.7 -- -

Palisades 266.5 ---

W52l4 Indian Point Unit 2 226.3 ---

Indian Point Unit 3 206.2 - --

Notes:

(a) Position 1.1 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 5-3 of this report. As discussed in Section 4, the surveillance weld data for heat W5214 is not fully credible...

(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 27204 is credible.

(d) 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. -

WCAP-17954-NP - December 2014 Revision 0

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, Kl,, for the metal temperature at that time. Kjc 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 Kr¢ curve is given by the following equation:

02 K c = 33.2 + 20.734

• e[0 (T- RTNDr)] (1)

where, Kic (ksi4in.) = reference stress intensity factor as a function of the metal temperature T and the metal reference nil-ductility temperature RTNDT This K1, curve is based on the lower bound of static critical K, 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:

C'* Kr. + K1 , < Kic (2)

where, Kim stress intensity factor caused by membrane (pressure) stress K = stress intensity factor caused by the thermal gradients Kf = reference stress intensity factor as a function of the metal temperature T and the metal reference nil-ductility temperature RTNDT 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 December 2014 WCAP- 17954-NP WCAP- I17954-NP December 2014 Revision 0

Westinghouse Non-Proprietary Class 3 6-2 For membrane tension, the corresponding K, for the postulated defect is:

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

Mm = 1.85 for 4t < 2, Mm = 0.926 f for 2:<Ft <* 3.464, Mm = 3.21 for t > 3.464 and, Mm for an outside axial surface flaw is given by:

Min = 1.77 for ft < 2, Mm = 0.8934I for 2<4ft7 3.464, Mm = 3.09 for 47 > 3.464 Similarly, Mm for an inside or an outside circumferential surface flaw is given by:

Mm = 0.89 for ft < 2, Mm = 0.4434-7t for 2<5 ft _<3.464, Mm = 1.53 for 7 > 3.464 Where:

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

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

KIb = Mb

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

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

Ki, = 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 Ki, = 0.753x103 x HU x t25 (6) where HU is the heatup rate in °F/hr.

WCAP- 17954-NP December 2014 Revision 0

Westinghouse Non-Proprietary Class 3 6-3 The through-wall temperature difference associated with the maximum thermal K, can be determined from ASME 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 KI.

(a) The maximum thermal K, 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, for radial thermal gradient can be calculated for any thermal stress distribution and at any specified time during cooldown for a '/4-thickness axial or circumferential inside surface defect using the relationship:

Kj, = (1.0359Co + 0.6322CJ + 0.4753C 2 + 0.3855C3)

• V (7) or similarly, Kit during heatup for a 1/4-thickness outside axial or circumferential surface defect using the relationship:

K, = (1.043 Co + 0.630 C, + 0.481C2+ 0.401 C,3)

• r (8) where the coefficients Co, Ch, C2 and C3 are determined from the thermal stress distribution at any specified time during the heatup or cooldown using the form:

oa(x) = Co + Ci(x / a) + C2(x /a) 2 + C3(x /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 2.6.24 and 2.6.3-1).

At any time during the heatup or cooldown transient, Kic 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 RTNDT, 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 intensity factors, K1,, for the reference flaw are computed. From Equation 2, the pressure stress intensity factors are obtained and, from these, the al.lowable pressures are calculated.

For the calculation of the allowable pressure versus coolant temperature during cooldown, the reference I/4T flaw of Appendix G to Section XI of the ASME Code is assumed to exist at the inside of the vessel WCAP- 17954-NP December 2014 Revision 0

Westinghouse Non-Proprietary Class 3 6-4 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 AT (temperature) across the vessel wall developed during cooldown results in a higher value of K1, at the i/4T location for finite cooldown rates than for steady-state operation. Furthermore, if conditions exist so that the increase in K1 c 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 I/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 KI, for the inside 1/4T flaw during heatup is lower than the K1, 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 K1 , 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 1/4T 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 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 WCAP- 17954-NP December 20140 Revision

Westinghouse Non-Proprietary Class 3 6-5 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 HEAD/VESSEL 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 RTNDT by at least 120'F for normal operation when the pressure exceeds 20 percent of the preservice hydrostatic test pressure, which is calculated to be 636 psia. The limiting unirradiated RTNDT of 38'F is associated with the vessel flange of the Indian Point Unit 3 vessel, so the minimum allowable temperature of this region is 158°F at pressures greater than 636 psia (without margins for instrument uncertainties). This limit is shown in Figures 8-1 and 8-2.

WCAP- 17954-NP December 2014 Revision 0

Westinghouse Non-Proprietary Class 3 7-1 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 RTNDT + ARTNDT + 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. 18]. 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.

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

ARTNDT = CF

• f1.

2 1S-0.10oglo (11)

To calculate ARTNDT 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.

f(depth x) = fsuirface

• e (-0.24x) (12) where x inches (reactor vessel cylindrical shell beltline thickness is 8.625 inches) is the depth into the vessel wall measured from the vessel clad/base metal interface. The resultant fluence is then placed in Equation II to calculate the ARTNDT 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 37 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, per Regulatory Guide 1.99, Revision 2. The values in these tables will be used to calculate the 37 EFPYART values for the Indian Point Unit 3 reactor vessel materials.

Margin is calculated as M = 2 ý0-2 + A . The standard deviation for the initial RTNDT margin term (as) is 07F when the initial RTNDT is a measured value, and 177F when a generic value is available. The standard deviation for the ARTNDT margin term, aA, is 17°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, aA is equal to 280 F when surveillance capsule data is not used or is non-credible, and is 140 F (half the value) when credible surveillance capsule data is used. The value for CaAneed not exceed 0.5 times the mean value of ARTNDT.

Contained in Tables 7-2 and 7-3 are the 37 EFPY ART calculations at the 1/4T and 3/4T locations for generation of the Indian Point Unit 3 heatup and cooldown curves.

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Westinghouse Non-Proprietary Class 3 7-2 The limiting ART values for Indian Point Unit 3 to be used in the generation of the P-T limit curves are based on Lower Shell Plate B-2803-3 (using credible surveillance data, Position 2.1). The limiting ART values, using the "Axial Flaw" methodology, for Lower Shell Plate B-2803-3 are summarized in Table 7-4.

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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 Indian Point Unit 3 Reactor Vessel Materials at 37 EFPY Surface Fluence, 1/4T f 3/4T f Reactor Vessel Region f "I (n/cm 2, (n/cm2 IFF (n/cm2'

~E > 1.0 ncFF E > 1.0 MeV) MeV)

E > 1.0 MeV)

Nozzle Shell Plates 1.51 x 107 9.02 x 1016 0.1022 3.20 x 1016 0.0478 Nozzle Shell Axial Welds 1.39 x 10 17'b) 8.28 x 1016 0.0963 2.94 x 1016 0.0447 Nozzle Shell to Intermediate 1.95 x 101 7 1.16 x 101 7 0.1211 4.12 x 101 7 0.0580 Shell Circumferential Weld Intermediate Shell Plates 1.25 x 10' 9 7.47 x 1018 0.9182 2.65 x 1018 0.6390 Intermediate Shell Axial Welds 1.01 X 10 19(b) 6.04 x 1018 0.8590 2.15 x 10"8 0.5865 Intermediate to Lower Shell 1.25 x 10'9 7.46 x 10"s 0.9178 2.65 x 1018 0.6386 Circumferential Weld Lower Shell Plates 1.25 x 10' 9 7.46 x l0IS 0.9178 2.65 x 1018 0.6386 Lower Shell Axial Welds 1.25 x 1 0 19(b) 7.46 x 1018 0.9178 2.65 x 1018 0.6386 Notes:

(a) 37 EFPY fluence values were interpolated from the 36 and 42 EFPY fluence values documented in Table 2-2.

(b) The axial weld fluence values are taken to be the peak value across all three azimuthal locations for each shell elevation (nozzle, intermediate and lower).

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Westinghouse Non-Proprietary Class 3 7-4 Table 7-2 Adjusted Reference Temperature Evaluation for the Indian Point Unit 3 Reactor Vessel Beitline Materials through 37 EFPY at the 1/4T Location CF/4T f 1/4T (aIRT0AT0 M ART(d)

Reactor Vessel Material (n/cm 2 /4T RTNDTU ARTNDT a R(F) E>1.0MeV) FF (OF) (OF) (OF) (OF) (OF) (OF)

Reactor Vessel Extended Beltline Materials NS Plate B-2801-1 139.4 9.02 x 1016 0.1022 34 14.2 0 7.1 14.2 62.5 NS Plate B-2801-2 137.0 9.02 x 1016 0.1022 44 14.0 0 7.0 14.0 72.0 NS Plate B-2801-3 152.9 9.02 x 1016 0.1022 3 15.6 0 7.8 15.6 34.2 NS Axial Weld Seams 1-042A, B, C 230.7 8.28 x 1016 0.0963 -56(b) 22.2 17(') 11.1 40.6 6.8 Using surveillance data 223.7 8.28 x 1016 0.0963 -501 21.5 17(b) 10.8 40.3 5.8 NS to IS Circumferential Weld Seam 8-042 226.8 1.16 x 1017 0.1211 -56(b) 27.5 17(b) 13.7 43.7 15.2 Using surveillance data 212.2 1.16 x 10"- 0.1211 -56 M) 25.7 17(b) 12.9 42.6 12.3 Reactor Vessel Beltline Materials IS Plate B-2802-1 137.0 7.47 x 10"8 0.9182 5 125.8 0 17.0 34.0 164.8 IS Plate B-2802-2 151.6 7.47 x 10"s 0.9182 -4 139.2 0 17.0 34.0 169.2 IS Plate B-2802-3 135.8 7.47 x 1018 0.9182 17 124.7 0 17.0 34.0 175.7 LS Plate B-2803-1 127.7 7.46 x 10"8 0.9178 49 117.2 0 17.0 34.0 200.2 LS Plate B-2803-2 150.2 7.46 x 10"s 0.9178 -5 137.9 0 17.0 34.0 166.9 LS Plate B-2803-3 160.4 7.46 x 1018 0.9178 74 147.2 0 17.0 34.0 255.2 Using surveillance data 167.8 7.46 x lO0" 0.9178 74 154.0 0 8.5 17.0 245.0 IS Axial Weld Seams 2-042A, B, C 221.3 6.04 x 1018 0.8590 -56Nb) 190.1 17(b) 28.0 65.5 199.6 LS Axial Weld Seams 3-042A, B, C 221.3 7.46x 1018 0.9178 -56(bN 203.1 17 (b) 28.0 65.5 212.6 IS to LS Circumferential Weld Seam 9-042 189.1 7.46 x 1018 0.9178 -54 173.6 0 28.0 56.0 175.6 Notes:

(a) Initial RTNDT values are measured values, unless otherwise noted.

(b) Initial RTNDT value is generic; hence a, = 7'F.

(c) As discussed in Section 4, the surveillance plate data was deemed credible. The surveillance weld data for heat 27204 was deemed credible; whereas, for heat W5214, it was deemed not fully credible. Per the guidance of Regulatory Guide 1.99, Revision 2, the base metal OA = 17'F for Position 1.1 and, with credible surveillance data, GA

= 8.5°F for Position 2.1; the weld metal GA - 28'F for Position 1.1 and 2.1 with not fully credible surveillance data, and with credible surveillance data, aA = 14'F for Position 2.1. However, GA need not exceed 0.5*ARTNDT.

(d) The Regulatory Guide 1.99, Revision 2 methodology was used to calculate ART values. ART = RTNDT(U) + ARTNDT + Margin.

WCAP- 17954-NP December 2014 Revision 0

Westinghouse Non-Proprietary Class 3 7-5 Table 7-3 Adjusted Reference Temperature Evaluation for the Indian Point Unit 3 Reactor Vessel Beltline Materials through 37 EFPY at the 3/4T Location CF 3/4T f 3/4T RTNDT(U (a) ARTNDT C51(a A( M ART(d)

Reactor Vessel Material (OF) (E/cM, FF (OF) (OF) (OF) (OF) (OF) (OF)

_____________________E_ ____ E1.0

> MeV) I____ I_____

___ I____i__

Reactor Vessel Extended Beltline Materials NS Plate B-2801-1 139.4 3.20 x 1016 0.0478 34 6.7 0 3.3 6.7 47.3 NS Plate B-2801-2 137.0 3.20 x 1016 0.0478 44 6.5 0 3.3 6.5 57.1 NS Plate B-2801-3 152.9 3.20 x 1016 0.0478 3 7.3 0 3.7 7.3 17.6 NS Axial Weld Seams 1-042A, B, C 230.7 2.94 x 1016 0.0447 -56(" 10.3 17(b) 5.2 35.5 -10.2 Using surveillance data 223.7 2.94 x 1016 0.0447 - 56 bN 10.0 17 (b) 5.0 35.4 -10.6 NS...........................................................................................................................................................

to IS Circumferential Weld Seam 8-042 226.8 4.12 x 1017 0.0580 -5 6 (b) 13.1 17 (b) 6.6 36.5 -6.4 Using surveillance data 212.2 4.12 x 1017 0.0580 -56(b) 12.3 17 (b) 6.1 36.2 -7.5 Reactor Vessel Beltline Materials IS Plate B-2802-1 137.0 2.65 x 1018 0.6390 5 87.5 0 17.0 34.0 126.5 IS Plate B-2802-2 151.6 2.65 x 10I 0.6390 -4 96.9 0 17.0 34.0 126.9 IS Plate B-2802-3 135.8 2.65 x 1018 0.6390 17 86.8 0 17.0 34.0 137.8 LS Plate B-2803-1 127.7 2.65 x 10 " 0.6386 49 81.6 0 17.0 34.0 164.6 LS Plate B-2803-2 150.2 2.65 x 10"s 0.6386 -5 95.9 0 17.0 34.0 124.9 LS Plate B-2803-3 160.4 2.65 x 1018 0.6386 74 102.4 0 17.0 34.0 210.4 Using surveillance data 167.8 2.65 x 10"8 0.6386 74 107.2 0 8.5 17.0 198.2 IS Axial Weld Seams 2-042A, B, C 221.3 2.15 x 1018 0.5865 - 56(b) 129.8 17 (b) 28.0 65.5 139.3 LS Axial Weld Seams 3-042A, B, C 221.3 2.65 x 1018 0.6386 -56(b) 141.3 17(b) 28.0 65.5 150.8 IS to LS Circumferential Weld Seam 9-042 189.1 2.65 x 1018 0.6386 -54 120.8 0 28.0 56.0 122.8 Notes:

(a) Initial RTNDT values are measured values, unless otherwise noted.

(b) Initial RTNDT value is generic; hence a, = 17YF.

(c) As discussed in Section 4, the surveillance plate data was deemed credible. The surveillance weld data for heat 27204 was deemed credible; whereas, for heat W5214, it was deemed not fully credible. Per the guidance of Regulatory Guide 1.99, Revision 2, the base metal aA = 17'F for Position 1.1 and, with credible surveillance data, 0a

= 8.5'F for Position 2.1; the weld metal aA = 28'F for Position 1.1 and 2.1 with not fully credible surveillance data, and with credible surveillance data, 0a= 14'F for Position 2.1. However. 0a need not exceed 0.5*ARTNoT.

(d) The Regulatory Guide 1.99, Revision 2 methodology was used to calculate ART values. ART = RTNDT(*) + ARTNDT + Margin.

WCAP- 17954-NP December 2014 Revision 0

Westinghouse Non-Proprietary Class 3 7-6 Table 7-4 Summary of the Limiting ART Values Used in the Generation of the Indian Point Unit 3 Heatup and Cooldown Curves at 37 EFPY I/4T Limiting ART 3/4T Limiting ART 245.0°F 198.20F Lower Shell Plate B-2803-3 (using credible surveillance data)

WCAP- 17954-NP December 2014 Revision 0

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.

Figure 8-1 presents the limiting heatup curves without margins for possible instrumentation errors using heatup rates of 60 and 100°F/hr applicable for 37 EFPY, with the flange 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, -20, -40, -60, and -100°F/hr applicable for 37 EFPY, with the flange 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 straight-line portion of the criticality limit is at the minimum permissible temperature for the 2500 psia inservice hydrostatic test as required by Appendix G to 10 CFR Part 50. The governing equation for the hydrostatic test is defined in the 1998 through the 2000 Addenda ASME Code Section XI, Appendix G as follows:

1.5 Kim <Klc

where, Klm is the stress intensity factor covered by membrane (pressure) stress, KI, = 33.2 + 20.734 e [0.02 (T-RTNT)I T is the minimum permissible metal temperature, and RTNDT 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 407F 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 leak tests for the Indian Point Unit 3 reactor vessel at 37 EFPY is 305'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.

WCAP- I 7954-NP December 2014 Revision 0

Westinghouse Non-Proprietary Class 3 8-2 Figures 8-1 and 8-2 define all of the above limits for ensuring prevention of non-ductile failure for the Indian Point Unit 3 reactor vessel for 37 EFPY with the flange 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 P-T limit curves shown in Figures 8-1 and 8-2 were generated based on the limiting ART values for the cylindrical beltline and extended beltline reactor vessel materials. 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 Indian Point Unit 3 at 37 EFPY.

WCAP- 17954-NP December 2014 Revision 0

Westinghouse Non-Proprietary Class 3 8-3 MATERIAL PROPERTY BASIS LIMITING MATERIAL: Lower Shell Plate B-2803-3 using credible surveillance data, Position 2.1 LIMITING ART VALUES AT 37 EFPY: 1/4T, 245.07F (Axial Flaw) 3/4T, 198.2°F (Axial Flaw) 2500 2250 2000 1750 1500 U) 1250 1000 750 500 250 0

300 350 400 450 500 550 Moderator Temperature (Deg. F)

Figure 8-1 Indian Point Unit 3 Reactor Coolant System Heatup Limitations (Heatup Rates of 60 and 1001F/hr) Applicable for 37 EFPY (with Flange Requirements and without Margins for Instrumentation Errors) using the 1998 through the 2000 Addenda App.

G Methodology (w/ K)

WCAP- 17954-NP December 2014 Revision 0

Westinghouse Non-Proprietary Class 3 8-4 MATERIAL PROPERTY BASIS LIMITING MATERIAL: Lower Shell Plate B-2803-3 using credible surveillance data, Position 2.1 LIMITING ART VALUES AT 37 EFPY: I/4T, 245.07F (Axial Flaw) 3/4T, 198.2'F (Axial Flaw) 2500 2250 2000 1750 1500

()

2 1250 C-S1000 750 500 250 0 I . . . . ,. . . . . . . .. i. . . .. . . . . . ., . . . . , .. . - . . . . ,. . . . i *..

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

Figure 8-2 Indian Point Unit 3 Reactor Coolant System Cooldown Limitations (Cooldown Rates of 0, -20, -40, -60, and -100°F/hr) Applicable for 37 EFPY (with Flange Requirements and without Margins for Instrumentation Errors) using the 1998 through the 2000 Addenda App. G Methodology (w/ K,)

WCAP-17954-NP December 2014 Revision 0

Westinghouse Non-Proprietary Class 3 8-5 Table 8-1 Indian Point Unit 3 37 EFPY Heatup Curve Data Points using the 1998 through the 2000 Addenda App. G Methodology (w/ Kl,, w/

Flange Notch, and w/o Margins for Instrumentation Errors) 60°F/hr Heatup 60°F/hr Criticality 100 0F/hr Heatup 100IF/hr Criticality T (OF) P (psia) T (-F) P (psia) T (IF) P (psia) T (IF) P (psia) 60 0 305 0 60 0 305 0 60 574 305 1185 60 517 305 942 65 574 310 1244 65 517 310 994 70 574 315 1305 70 517 315 1051 75 574 320 1358 75 517 320 1114 80 574 325 1417 80 517 325 1183 85 574 330 1482 85 517 330 1260 90 574 335 1553 90 517 335 1345 95 574 340 1632 95 517 340 1438 100 574 345 1719 100 517 345 1541 105 574 350 1815 105 517 350 1655 110 574 355 1921 110 517 355 1780 115 574 360 2037 115 517 360 1918 120 576 365 2166 120 517 365 2032 125 578 370 2309 125 517 370 2148 130 581 375 2465 130 517 375 2276 135 585 135 517 380 2417 140 590 140 517 145 595 145 518 150 602 150 520 155 609 155 522 160 617 160 525 165 627 165 529 170 637 ..170 534 175 649 175 540 180 662 180 547 185 676 185 555 190 692 190 564 -

195 710 195 .575 200 730 200 587 "

205 752 205 600 210 776 210 615 215 803 215 632 220 832 220 651 -

225 865 225 671 WCAP- 17954-NP December 2014 Revision 0

Westinghouse Non-Proprietary Class 3 8-6 Table 8-1 Indian Point Unit 3 37 EFPY Heatup Curve Data Points using the 1998 through the 2000 Addenda App. G Methodology (w/ K1 ,, w/

Flange Notch, and w/o Margins for Instrumentation Errors) 60IF/hr Heatup 60 0F/hr Criticality 100 0F/hr Heatup 100IF/hr Criticality T (IF) P (psia) T (OF) P (psia) T (IF) P (psia) T (IF) P (psia) 230 901 230 694 235 930 235 720 240 963 240 748 245 999 245 780 250 1039 250 814 255 1082 255 853 260 1131 260 895 265 1185 265 942 270 1244 270 994 275 1305 275 1051 280 1358 280 1114 285 1417 285 1183 290 1482 290 1260 295 1553 295 1345 300 1632 300 1438 305 1719 305 1541 310 1815 310 1655 315 1921 315 1780 320 2037 320 1918 325 2166 325 2032 330 2309 330 2148 335 2465 335 2276 340 2417 Leak Test Limit T (OF) P (psia) 288 2015 305 2500 WCAP-17954-NP December 2014 Revision 0

Westinghouse Non-Proprietary Class 3 8-7 Table 8-2 Indian Point Unit 3 37 EFPY Cooldown Curve Data Points using the 1998 through the 2000 Addenda App. G Methodology (w/ Kl,, w/ Flange Notch, and w/o Margins for Instrumentation Errors)

Steady State -20°F/hr. -40°F/hr. -60°F/hr. -100°F/hr.

T P T P T P T P T P (OF) (psia) (OF) (psia) (OF) (psia) (OF) (psia) (OF) (psia) 60 0 60 0 60 0 60 0 60 0 60 630 60 579 60 526 60 473 60 361 65 631 65 580 65 527 65 473 65 361 70 632 70 580 70 528 70 474 70 362 75 633 75 582 75 529 75 475 75 362 80 634 80 583 80 530 80 476 80 363 85 636 85 584 85 531 85 477 85 365 90 636 90 586 90 533 90 479 90 366 95 636 95 588 95 535 95 480 95 368 100 636 100 589 100 536 100 482 100 370 105 636 105 592 105 539 105 484 105 372 110 636 110 594 I10 541 110 487 110 375 115 636 115 597 115 544 115 490 115 378 120 636 120 600 120 547 120 493 120 382 125 636 125 603 125 551 125 497 125 386 130 636 130 607 130 555 130 501 130 391 135 636 135 611 135 559 135 506 135 397 140 636 140 616 140 564 140 511 140 403 145 636 145 621 145 569 145 517 145 410 150 636 150 627 150 576 150 524 150 418 155 636 155 633 155 582 155 531 155 427 158 636 158 636 160 590 160- 539. 160 437 158 687 158 637 165 599 165 549 165 448 160 690 160 640 170 608 170 559 170 461 165 697 165 648 175 618 175 570 175 475 170 705 170 657 180 630 180 583- 180 490 175 714 175 666 185 643 185 597 185 508 180 723 180 677 190 657 190 613 190 527 185 734 185 689 195 673 195 631 195 549 190 746 190 702 200 691 200 650 200 573 195 760 195 716 205 710 205 672 205 600 WCAP- 17954-NP December 2014 Revision 0

Westinghouse Non-Proprietary Class 3 8-8 Table 8-2 Indian Point Unit 3 37 EFPY Cooldown Curve Data Points using the 1998 through the 2000 Addenda App. G Methodology (w/ K1,, w/ Flange Notch, and w/o Margins for Instrumentation Errors)

Steady State -20°F/hr. -40IF/hr. -60°F/hr. -100IF/hr.

T P T P T P T P T P (IF) (psia) (IF) (psia) (OF) (psia) (OF) (psia) (OF) (psia) 200 774 200 732 210 732 210 696 210 630 205 790 205 750 215 756 215 723 215 664 210 808 210 770 220 783 220 752 220 701 215 828 215 791 225 812 225 785 225 742 220 850 220 815 230 845 230 822 230 787 225 874 225 842 235 881 235 862 235 838 230 901 230 872 240 921 240 907 240 894 235 930 235 904 245 966 245 956 245 956 240 963 240 940 250 1015 250 1011 250 1011 245 999 245 980 255 1069 255 1069 255 1069 250 1039 250 1024 260 1127 260 1127 260 1127 255 1082 255 1073 265 1185 265 1185 265 1185 260 1131 260 1127 270 1244 270 1244 270 1244 265 1185 265 1185 275 1310 275 1310 275 1310 270 1244 270 1244 280 1382 280 1382 280 1382 275 1310 275 1310 285 1462 285 1462 285 1462 280 1382 280 1382 290 1551 290 1551 290 1551 285 1462 285 1462 295 1649 295 1649 295 1649 290 1551 290 1551 300 1757 300 1757 300 1757 295 1649 295 1649 305 1876 305 1876 305 1876 300 1757 300 1757 310 2009 310 2009 310 2009 305 1876- 305 1876 315. 2155 315 2155 315 2155 310 2009 310 2009 320 2316 320 2316 320 2316 315 2155 315 2155 325 2494 325. 2494 325 2494 320 2316 320 2316 .

325 2494 325 2494 WCAP- 17954-NP December 2014 Revision 0

Westinghouse Non-Proprietary Class 3 9-1 9 REFERENCES

1. Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials," U. S.

Nuclear Regulatory Commission, 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. Westinghouse Report WCAP-16212-P, Revision 0., "Indian Point Nuclear Generating Unit No. 3 Stretch Power Uprate NSSS and BOP Licensing Report," June 2004

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

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

8. RSICC Data Library Collection DLC- 185, "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.

9. 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 14, 2014. [NRC ADAAIS Accession Numnber ML14149A165]

10. Entergy Letter NL-08-143, "Additional Information Regarding License Renewal Application -

Reactor Vessel Fluence Clarification," September 24, 2008. [NRC ADAMS Accession Number ML082760402]

11. Entergy Letter NL-08-014, "Clarification to Reactor Vessel Surveillance Program and Neutron Embrittlement Time-Limited Aging Analyses and Audit Item #105; and Revision to License Renewal Regulatory Commitment List," January 17, 2008. fNRC ADAMS Accession Number ML0802500271

12. 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 1.9;. 1995. effective January 18, 1996. ..-

13. Structural Integrity Associates, Inc. Report No. 1000915.401, Revision 1, "Revised Pressurized Thermal Shock Evaluation for the Palisades- Reactor Pressure Vessel," November 2010. [NRC ADAMS Accession Number ML110060694]

14. Structural Integrity Associates, Inc. Report No. 0901132.401, Revision 0, "Evaluation of Surveillance Data for Weld Heat No. W5214 for Application to Palisades PTS Analysis," April 2010. [NRC ADAMS-A c&ession Numb er AML 10060693].;

15. Westinghouse Report WCAP-16251-NP, Revision 0, "Analysis of Capsule X from Entergy's Indian Point Unit 3 Reactor Vessel Radiation Surveillance Program," July 2004.

16. Westinghouse Report WCAP-17315-NP, Revision 0, "Diablo Canyon Units I and 2 Pressurized Thermal Shock and Upper-Shelf Energy Evaluations," July 2011.

WCAP-17954-NP December 2014 Revision 0

Westinghouse Non-Proprietary Class 3 9-2

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

18. ASME Boiler and Pressure Vessel (B&PV) Code, Section 1I1,Division 1, Subsection NB, Section NB-2300, "Fracture Toughness Requirements for Material."

WCAP- 17954-NP December 2014 Revision 0

Westinghouse Non-Proprietary Class 3 A-I APPENDIX A THERMAL STRESS INTENSITY FACTORS (Kit)

Tables A-i and A-2 contain the thermal stress intensity factors (K,,) for the maximum heatup and cooldown rates at 37 EFPY for Indian Point Unit 3. The reactor vessel cylindrical shell radii to the I/4T and 3/4T locations are as follows:

• I/4T Radius = 89.062 inches

• 3/4T Radius = 93.375 inches WCAP- 17954-NP December 2014 Revision 0

Westinghouse Non-Proprietary Class 3 A-2 Table A-1 K1, Values for Indian Point Unit 3 at 37 EFPY 100*F/hr Heatup Curves (w/ Flange 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) 1OO0 F/hr Heatup (OF) (KSI Iin.) 100°F/hr Heatup (*F) (KSI 4in.)

60 55.985 -0.995 55.043 0.473 65 58.558 -2.452 55.294 1.438 70 61.621 -3.712 55.962 2.426 75 64.898 -4.910 57.100 3.356 80 68.449 -5.945 58.655 4.190 85 72.111 -6.892 60.590 4.938 90 75.955 -7.714 62.865 5.599 95 79.899 -8.465 65.440 6.192 100 83.973 -9.122 68.281 6.719 105 88.135 -9.721 71.357 7.189 110 92.392 -10.247 74.637 7.609 115 96.722 -10.728 78.098 7.988 120 101.123 -11.154 81.717 8.327 125 105.583 -11.544 85.476 8.634 130 110.095 -11.891 89.359 8.910 135 114.656 -12.210 93.352 9.160 140 119.256 -12.495 97.441 9.387 145 123.895 -12.759 101.616 9.593 150 128.564 -12.996 105.866 9.781 155 133.264 -13.217 110.183 9.953 160 137.986 -13.416 114.558 10.111 165 142.734 -13.603 118.986 10.257 170 147.498 -13.773 123.459 10.391 175 .152.283 -13.934 127.974 10.515 180- 157.080 -14.081 132.524 10.631 185 161.893 -14.221 137.106 10.739 190 166.717 -14.350 141.717 10.840 195 171.553 -14.474 146.352 10.936 200 176.397 -14.589 151,009 11.025 205 181.250 -14.700 155.686 11.111 210 186.110 -14.804 160.380 11.192 WCAP-17954-NP December 2014 Revision 0

Westinghouse Non-Proprietary Class 3 A-3 Table A-2 K1, Values for Indian Point Unit 3 at 37 EFPY 100 0F/hr Cooldown Curves (w/ Flange Requirements and w/o Margins for Instrument Errors)

Water Vessel Temperature at 1/4T -100*F/hr Cooldown Temp. Location for -100*F/hr l/4T Thermal Stress (OF) Cooldown (*F) Intensity Factor (KSI 4in.)

210 237.042 17.138 205 231.957 17.069 200 226.871 17.001 195 221.784 16.932 190 216.698 16.863 185 211.611 16.794 180 206.524 16.725 175 201.437 16.655 170 196.350 16.586 165 191.262 16.516 160 186.175 16.447 155 181.087 16.377 150 175.999 16.308 145 170.912 16.238 140 165.824 16.169 135 160.737 16.099 130 155.649 16.030 125 150.561 15.961 120 145.474 15.892 115 140.387 15.823 110 135.299 15.754 105 130.212 15.685 100 125.125 15.616 95 120.038 15.547 90 114.951 15.479 85 109.864 15.411 80 104.777 15.342 75 99.691 15.274 70 94.604 15.206 65 89.518 15.138 60 84.433 15.070 WCAP-17954-NP December 2014 Revision 0

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-i], 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 (RTNDT) 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 Indian Point Unit 3 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 comer 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 Indian Point Unit 3 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 corner 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 I/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 I/4T flaw.

B.1 CALCULATION OF ADJUSTED REFERENCE TEMPERATURES The fracture toughness (KIt) 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 K1c 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 RTNDT value, and projected neutron fluence as inputs. The ART values for each of the reactor vessel inlet and outlet nozzle forging materials are documented in Table B-I

- and a summary of the limiting inlet and outlet nozzle ART values for Indian Point Unit 3 is presented in Table B-2.

Nozzle Chemistry Data The nickel weight percent values were obtained directly from the material-specific analyses documented in Combustion Engineering report MISC-PENG-ER-010 [Ref. B-4] for each of the Indian Point Unit 3 reactor vessel inlet and outlet nozzles. The MISC-PENG-ER-010 report did not contain copper weight WCAP- 17954-NP December 2014 Revision 0

Westinghouse Non-Proprietary Class 3 B-2 percent values because at the time that the Indian Point Unit 3 nozzles were manufactured, it was not required for SA-508, Class 2 low-alloy steel. Therefore, a best-estimate copper weight percent value of

(( )) from Section 4 of the NRC-approved BWRVIP report, BWRVIP-173-A [Ref. B-5], was utilized for the Indian Point Unit 3 inlet and outlet nozzles. A mean plus two standard deviations methodology was applied to the data in BWRVIP-I 73-A to determine the conservative copper weight percent value of

(( )). The data in the BWRVIP report was tabulated from an industry-wide database of SA-508, Class 2 forging materials.

Position 1.1 chemistry factors for each of the reactor vessel nozzle materials are calculated using the copper and nickel weight percent of the material and Table 2 of Regulatory Guide 1.99, Revision 2. The copper and nickel weight percent values as well as the Position 1.1 chemistry factors for the Indian Point Unit 3 reactor vessel materials are provided in Table B-I.

Nozzle Fracture Toughness Properties The initial RTNDT values were determined for each of the Indian Point Unit 3 reactor vessel inlet and outlet nozzle forging materials using the BWRVIP-173-A, Alternative Approach 2 methodology, contained in Appendix B of that report. For all eight of the Indian Point Unit 3 inlet and outlet nozzle materials, CVGraph Version 5.3 was utilized to plot the material-specific Charpy V-Notch impact energy data from MISC-PENG-ER-010 to determine the transition temperatures at 35 ft-lb and 50 ft-lb, as specified in the Alternative Approach 2 methodology. The 35 ft-lb and 50 ft-lb temperatures were then evaluated per the Alternative Approach 2 methodology presented in BWRVIP-173-A, to determine the initial RTNDT values for the inlet and outlet nozzle materials for Indian Point Unit 3. Based on the information provided in MISC-PENG-ER-010, the Charpy V-Notch forging specimens for the inlet nozzles were oriented in the strong direction, whereas the outlet nozzle specimens were oriented in the weak direction. Therefore, the 50 ft-lb transition temperatures for the inlet nozzles were increased by 30'F to provide conservative estimates for specimens oriented in the weak direction per the Alternative Approach 2 methodology in BWRVIP-173-A.

Nozzle Calculated Neutron Fluence Values The maximum fast neutron (E > I MeV) exposure of the Indian Point Unit 3 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 corner of the nozzle, for conservatism.

Using linear interpolation of the 36 and 42 EFPY fluence values provided in Table 2-2, the inlet nozzles are determined to receive a projected maximum fluenceof 2.03 x 1016 n/cm 2 (E > I MeV) at the lowest extent of the nozzles at 37 EFPY. Similarly, the outlet nozzles are projected to achieve a maximum fluence value of 1.51 x 1016 n/cm 2 (E > I MeV) at the lowest extent of the nozzles at 37 EFPY. The maximum neutron fluence values for the inlet and outlet nozzle materials are not projected to exceed a fluence greater than I x 1017 n/cm 2 at 37 EFPY. Nevertheless, the projected 37 EFPY fluence values at the lowest extent of the nozzles were increased by a factor of two to conservatively bound the projected fluence at 54 EFPY. Finally, per NRC RIS 2014-011, embrittlement of reactor vessel materials, with projected fluence values less than 1 x 1017 n/cm 2, does not need to be considered; however, for conservatism, embrittlement of the Indian Point Unit 3 nozzle materials was considered in this summary WCAP- I7954-NP December 2014 Revision 0

Westinghouse Non-Proprietary Class 3 B-3 report. The inclusion of neutron embrittlement for the nozzle ART calculations, coupled with the factor of two increase to the projected fluence values, is meant to add conservatism such that this evaluation may be demonstrated to be bounding in the future, in the event that the nozzle fluence projections should increase (e.g. due to a power uprate).

The neutron fluence values used in the ART calculations for the Indian Point Unit 3 inlet and outlet nozzle forging materials are summarized in Table B-1.

WCAP- 17954-NP December 2014 Revision 0

Westinghouse Non-Proprietary Class 3 B-4 Table B-1 ART Calculations for the Indian Point Unit 3 Reactor Vessel Nozzle Materials Using Conservative Fluence Bounding 54 EFPY Wt. % Wt. % CF°( Fluence at Lowest (A)0 ARTNDT OU GA Margin ART Reactor Vessel Material Extent of Nozzle(d) FF(d) RTNDT(U)(e)

Cu(S) Ni(b) (OF) (n/cm 2, E > 1.0 MeV) (OF) (OF) (OF) (OF) (OF) (OF)

Inlet Nozzle B-2808-1 (( )) 0.72 138.2 4.06 x 1016 0.0573 -72 7.9 0 4.0 7.9 -56.2 Inlet Nozzle B-2808-2 0.67 136.5 4.06 x 1016 0.0573 -66 7.8 0 3.9 7.8 -50.3 Inlet Nozzle B-2808-3 (( ] 0.62 134.7 4.06 x 1016 0.0573 -73 7.7 0 3.9 7.7 -57.6 Inlet Nozzle B-2808-4 (( ] 0.70 137.5 4.06 x 1016 0.0573 -53 7.9 0 3.9 7.9 -37.2 Outlet Nozzle B-2809-1 ((J] 0.70 137.5 3.01 X 1016 0.0455 5 6.3 0 3.1 6.3 17.5 Outlet Nozzle B-2809-2 ((J]L 0.74 138.9 3.01 X 1016 0.0455 -40 6.3 0 3.2 6.3 -27.4 Outlet Nozzle B-2809-3 (( 0.75 139.3 3.01 X 1016 0.0455 -62 6.3 0 3.2 6.3 -49.3 Outlet Nozzle B-2809-4 (( 0.65 135.8 3.01 X 1016 0.0455 -45 6.2 0 3.1 6.2 -32.6 Notees:

(a) Cu wt. % values are the best-estimate values for SA-508, Class 2 low-alloy steel as documented in BWRVIP-173-A. This is considered Electric Power Research Institute (EPRI) proprietary information.

(b) The Ni wt. % values are material-specific values as documented in MISC-PENG-ER-010.

(c) CF values were calculated using the Cu and Ni wt. % values and Table 2 of Regulatory Guide 1.99, Revision 2.

(d) The 37 EFPY fluence values increased by a factor of 2. These are considered conservative fluence values that bound the maximum nozzle fluence values at 54 EFPY. FF values were calculated using Regulatory Guide 1.99, Revision 2.

(e) RTNDT(Li) values were determined using the Alternative Approach 2 methodology as described in Appendix B of BWRVIP-173-A.

(M Per Regulatory Guide 1.99, Revision 2, the base metal nozzle forging materials eYa= 17'F for Position 1.1 without surveillance data. However, ea need not exceed 0.5

• ARTNDT.

WCAP- 17954-NP December 2014 Revision 0

Westinghouse Non-Proprietary Class 3 B-5 Table B-2 Summary of the Limiting ART Values for the Indian Point Unit 3 Inlet and Outlet Nozzle Materials EFPY Nozzle Material and ID Limiting ART Value Number (OF)

> 54 Inlet Nozzle B-2808-4 -37.2 (Conservative Projection) Outlet Nozzle B-2809-1 17.5 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 Indian Point Unit 3 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-6], and are consistent with ASME PVP2011-57015 [Ref. B-7]. 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 stresses used for the Indian Point Unit 3 nozzle comers are consistent with those used in the 4-loop Pressurized Water Reactor (PWR) nozzle analyses in ORNL/TM-2010/246, which was based on a three-dimensional (3-D) finite element model (FEM).

The through-wall stresses at the nozzle comer location were fitted based on a third-order polynomial of the form:

2 3 Y= A0 + Aix+A 2x +A3x

where, a = through-wall stress distribution x = through-wall distance from inside surface A0, A,, A2, 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:

K,=.r- [0.706Ao+ 0.537 (2) A 1+ 0.448 (-) A 2+ 0.393 4a3l A 3]

WCAP- I 7954-NP December 2014 Revision 0

Westinghouse Non-Proprietary Class 3 B-6

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

The Indian Point Unit 3 inlet and outlet nozzle P-T limit curves are shown in Figures B-i 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- I00°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-i 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 comer.

Conclusion Based on the results shown in Figures B-I 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 37 EFPY remain limiting for the beltline and non-beltline reactor vessel components.

WCAP- 17954-NP December 2014 Revision 0

Westinghouse Non-Proprietary Class 3 B-7 2500 2250 2000 1750 oL 1500 1250 C 1 S1000 M.

Cu 750 -... . - RC°°Iown

~Deg. F/Hr

• steady-state 500 .. .. 0.........

i*.......*.*-6051000i 0 . ... .. .

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

Figure B-4 Comparison of Indian Point Unit 3 Beltline P-T Limits to Inlet Nozzle Limits WCAP- 17954-NP December 2014 Revision 0

Westinghouse Non-Proprietary Class 3 B-8 2500 ____ _ _ _ _ _ _

22 0I. i_ _

2i i Outlet Nozzle _

2000 Cooldown ___

-100 Deg. F/Hr " . _

1750 .Outlet Nozzle Steady State C-T I 1500 .... .

4) i  !

'1 I

i I i 1250 r_... .. ........

70~

750  !- - Cooldown - ......... ..

' ~Rates l

~steady-state Deg. F/Hr

~-40 500 - _

-60

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

Figure B-2 Comparison of Indian Point Unit 3 Beltline P-T Limits to Outlet Nozzle Limits December 2014 WCAP- 17954-N 17954-NPP December 2014 Revision 0

Westinghouse Non-Proprietary Class 3 B-9 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 14, 2014. [NRC ADAMS Accession Number ML14149A165]

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 Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials," U.S.

Nuclear Regulatory Commission, May 1988.

B-4 Combustion Engineering Report MISC-PENG-ER-0 10, Revision 0, "The Reactor Vessel Group Records Evaluation Program Phase II Final Report for the Indian Point Unit 3 Reactor Pressure Vessel Plates, Forgings, Welds and Cladding," October 1995.

B-5 BWRVIP-1 73-A: BWR Vessel and Internals Project: Evaluation of Chemistry Data for BWR Vessel Nozzle ForgingMaterials. EPRI, Palo Alto, CA: 2011. 1022835.

B-6 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.

B-7 ASME PVP2011-57015, "Additional Improvements to Appendix G of ASME Section XI Code for Nozzles," G. Stevens, H. Mehta, T. Griesbach, D. Sommerville, July 2011.

WCAP-17954-NP December 2014 Revision 0

Westinghouse Non-Proprietary Class 3 C-1 APPENDIX C NON-REACTOR VESSEL FERRITIC COMPONENTS 10 CFR Part 50, Appendix G [Ref. C-i], 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-2332(b) 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-2332(b) of the Section III ASME Code is applicable to material for ferritic piping, pumps and valves with a nominal wall thickness greater than 2 V2 inches [Ref. C-2]. Note that the Indian Point Unit 3 reactor coolant system does not have ferritic materials in the Class 1 piping, pumps or valves. Therefore, the LST requirements of NB-2332(b) are not applicable to the Indian Point Unit 3 P-T limits and the only ferritic RCPB components that are not part of the reactor vessel consist of the Model 44F Replacement Steam Generators (RSGs) and the original 44 Series Pressurizer.

The 44 Series Indian Point Unit 3 Pressurizer was constructed to the 1965 Edition Section III ASME Code through the Summer 1966 Addenda, 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.

Similarly, the 44F RSGs were designed and evaluated to the 1965 Edition Section III ASME Code through the Summer 1966 Addenda, and met all applicable requirements at the time of construction. A comprehensive 1983 Edition through Summer 1984 Addenda ASME Section III, Appendix G analysis has additionally been performed for the RSGs at Indian Point Unit 3. Furthermore, the RSGs 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.A 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 B&PV Code Section 111, Division I, NB-2332, "Material for Piping Pumps, and Valves, Excluding Bolting Material."

WCAP- 17954-NP December 2014 Revision 0

ENCLOSURE 3 TO NL-15-014 EPRI LETTER TO NRC "REQUEST FOR WITHOLDING OF THE FOLLOWING PROPRIETARY INFORMATION INCLUDED IN:"

DATED JANUARY 21,2015.

ENTERGY NUCLEAR OPERATIONS, INC.

INDIAN POINT NUCLEAR GENERATING UNIT NO. 3 DOCKET NO. 50-286

ELECTRIC POWER SIRESEARCH INSTITUTE Kurt Edsinger Director, PWR &

BWR Materials January 21, 2015 Document Control Desk Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555-0001

Subject:

Request for Withholding of the following Proprietary Information Included in:

Proposed Changes to Indian Point Unit 3 Technical Specifications Regarding Reactor Vessel Heatup and Cooldown Curves and Low Temperature Overpressure Protection System Requirements, WCAP-17954-P, Revision 0, December 2014 To Whom It May Concern:

This is a request under 10 C.F.R. §2.390(a)(4) that the U.S. Nuclear Regulatory Commission ("NRC") withhold from public disclosure the report identified in the enclosed Affidavit consisting of the proprietary information owned by Electric Power Research Institute, Inc. ("EPRI") identified in the attached report. Proprietary and non-proprietary versions of the Response and the Affidavit in support of this request are enclosed.

EPRI desires to disclose the Proprietary Information in confidence to assist the NRC review of the enclosed submittal to the NRC by Entergy. The Proprietary Information is not to be divulged to anyone outside of the NRC or to any of its contractors, nor shall any copies be made of the Proprietary Information provided herein. EPRI welcomes any discussions and/or questions relating to the information enclosed.

Ifyou have any questions about the legal aspects of this request for withholding, please do not hesitate to contact me at (650) 855-2271. Questions on the content of the Report should be directed to Andy McGehee of EPRI at (704) 502-6440.

Sincerely, Together . . . Shaping the Future of Electricity PALO ALTO OFFICE 3420 Hillview Avenue, Polo Alto, CA 94304-1395 USA

• 650.855.2000

• Customer Service 800.313.3774

• www.epri.com

ELECTRIC POWER e IRESEARCH INSTITUTE AFFIDAVIT RE: Request for Withholding of the Following Proprietary Information Included In:

Proposed Changes to Indian Point Unit 3 Technical Specifications Regarding Reactor Vessel Heatup and Cooldown Curves and Low Temperature Overpressure Protection System Requirements, WCAP-17954-P, Revision 0, December 2014 I, Kurt Edsinger, being duly sworn, depose and state as follows:

I am the Director of PWR and BWR Materials at Electric Power Research Institute, Inc. whose principal office is located at 3420 Hillview Avenue, Palo Alto, CA. ("EPRI") and I have been specifically delegated responsibility for the above-listed report that contains EPRI Proprietary Information that is sought under this Affidavit to be withheld "Proprietary Information". I am authorized to apply to the U.S. Nuclear Regulatory Commission ("NRC") for the withholding of the Proprietary Information on behalf of EPRI.

EPRI Proprietary Information is identified inthe above referenced report by a solid underline inside double square brackets. An example of such identification is as follows:

((This sentence is an example.{E}))

Tables containing EPRI Proprietary Information are identified with double square brackets before and after the object. In each case, the superscript notation {E} refers to this affidavit as the basis for the proprietary determination.

EPRI requests that the Proprietary Information be withheld from the public on the following bases:

Withholding Based Upon Privileged And Confidential Trade Secrets Or Commercial Or Financial Information (see e.g., 10 C.F.R. § 2.390(a)(4):

a. The Proprietary Information is owned by EPRI and has been held in confidence by EPRI. All entities accepting copies of the Proprietary Information do so subject to written agreements imposing an obligation upon the recipient to maintain the confidentiality of the Proprietary Information. The Proprietary Information is disclosed only to parties who agree, inwriting, to preserve the confidentiality thereof.

b. EPRI considers the Proprietary Information contained therein to constitute trade secrets of EPRI. As such, EPRI holds the Information inconfidence and disclosure thereof is strictly limited to individuals and entities who have agreed, in writing, to maintain the confidentiality of the Information.

c. The information sought to be withheld is considered to be proprietary for the following reasons. EPRI made a substantial economic investment to develop the Proprietary Information and, by prohibiting public disclosure, EPRI derives an economic benefit in the form of licensing royalties and other additional fees from the confidential nature of the Proprietary Information. Ifthe Proprietary Information were publicly available to consultants and/or other businesses providing services inthe electric and/or nuclear power industry, they would

be able to use the Proprietary Information for their own commercial benefit and profit and without expending the substantial economic resources required of EPRI to develop the Proprietary Information.

d. EPRI's classification of the Proprietary Information as trade secrets isjustified by the Uniform Trade Secrets Act which California adopted in 1984 and a version of which has been adopted by over forty states. The California Uniform Trade Secrets Act, California Civil Code §§3426 - 3426.11, defines a "trade secret" as follows:

"'Trade secret' means information, including a formula, pattern, compilation, program device, method, technique, or process, that:

(1) Derives independent economic value, actual or potential, from not being generally known to the public or to other persons who can obtain economic value from its disclosure or use; and (2) Is the subject of efforts that are reasonable under the circumstances to maintain its secrecy."

e. The Proprietary Information contained therein are not generally known or available to the public. EPRI developed the Information only after making a determination that the Proprietary Information was not available from public sources. EPRI made a substantial investment of both money and employee hours in the development of the Proprietary Information. EPRI was required to devote these resources and effort to derive the Proprietary Information. As a result of such effort and cost, both in terms of dollars spent and dedicated employee time, the Proprietary Information is highly valuable to EPRI.

f. A public disclosure of the Proprietary Information would be highly likely to cause substantial harm to EPRI's competitive position and the ability of EPRI to license the Proprietary Information both domestically and internationally. The Proprietary Information can only be acquired and/or duplicated by others using an equivalent investment of time and effort.

I have read the foregoing and the matters stated herein are true and correct to the best of my knowledge, information and belief. I make this affidavit under penalty of perjury under the laws of the United States of America and under the laws of the State of California.

Executed at 3420 Hillview Avenue, Palo Alto, CA being the premises and place of business of Electric Power Research Institute, Inc Date: /2 / /*

Kurt Edsinger

V Civil Code 1189 A notary public or ott'er officer completing this certificate verifies only the identityof the individual who signed the document to which this certificate is attached, and rot the truthfulness, accuracy, or-validity of that document.

(State of Califo nia)/-

(County of Subscribed and s or .* (or afirmed) before me on this 20 j.._hday y

/ ,proved to me on the basiof satisfac evidence to be the person(s) who a pear befor me.

Signature - _ , h.) , - (Seal)

My Commission Expires c ay of ,k20..=,,

~*BERTE A.DAHL SCOMMJ# 1926383 NOTARY PUBLIC

.CALI*C RNIAi

~:i SANTACLARA COUNTY myCokmI EXP.MAR 20.2015 1

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