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Enclosure 5 - WCAP-18455-NP, Revision 1, D.C. Cook Unit 1 Heatup and Cooldown Limit Curves for Normal Operation, Westinghouse Electric Company, February 2020. (Non-Proprietary)
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Enclosure 5 to AEP-NRC-2020-01 WCAP-18455-NP, Revision 1, "D.C. Cook Unit 1 Heatup and Cooldown Limit Curves for Normal Operation," Westinghouse Electric Company, February 2020. (Non-Proprietary)

Westinghouse Non-Proprietary Class 3 WCAP-18455-NP February 2020 Revision 1 D.C. Cook Unit 1 Heatup and Cooldown Limit Curves for Normal Operation

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 WCAP-18455-NP Revision 1 D.C. Cook Unit 1 Heatup and Cooldown Limit Curves for' Normal Operation Donald M. McNutt Ill*

RV/CV Design & Analysis Andrew E. Hawk*

Nuclear Operations & Radiation Analysis (NORA)

February 2020 Reviewers: D. Brett Lynch* Approved: Lynn A. Patterson*, Manager RV/CV Design & Analysis RV/CV Design & Analysis Jianwei Chen* Laurent P. Houssay*, Manager NORA NORA

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

Westinghouse Electric Company LLC 1000 Westinghouse Dr.

Cranberry Township, PA 16066

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Westinghouse Non-Proprietary Class 3 ii RECORD OF REVISION Revision Description Completed 0 Original Issue September 2019 1 The in-vessel surveillance capsule fission monitor target atom fractions used for See PRIME Revision O were corrected and all affected dosimetry results were updated. See Corrective Action Program (CAP) Issue Report (IR) 2020-1051. There were no changes made to the calculated RPV neutron exposures reported in Revision 0 as a result of this issue.

WCAP-18455-NP February 2020 Revision 1

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Westinghouse Non-Proprietary Class 3 iii TABLE OF CONTENTS LIST OF TABLES ....................................................................................................................................... iv LIST OF FIGURES .................................................................................................................................... vii EXECUTIVE

SUMMARY

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

2.1 INTRODUCTION

........................................................................................................... 2-1 2.2 DISCRETE ORDINATES ANALYSIS ........................................................................... 2-1 2.3 CALCULATIONAL UNCERTAINTIES ........................................................................ 2-4 3 FRACTURE TOUGHNESS PROPERTIES ................................................................................. 3-1 4 SURVEILLANCE DATA ............................................................................................................. 4-1 5 CHEMISTRY FACTORS ............................................................................................................. 5-1

6 CRITERIA FOR ALLOWABLE PRESSURE-TEMPERATURE RELATIONSHIPS ................ 6-1 6.1 OVERALL APPROACH ................................................................................................. 6-1 6.2 METHODOLOGY FOR PRESSURE-TEMPERATURE LIMIT CURVE DEVELOPMENT ............................................................................................................ 6-1 6.3 CLOSURE HEAD/VESSEL FLANGE REQUIREMENTS ........................................... 6-5 6.4 BOLTUP TEMPERATURE REQUIREMENTS ............................................................. 6-5 7 CALCULATION OF ADJUSTED REFERENCE TEMPERATURE .......................................... 7-1 8 HEATUP AND COOLDOWN PRESSURE-TEMPERATURE LIMIT CURVES ....................... 8-1 9 REFERENCES ............................................................................................................................. 9-1 APPENDIX A THERMAL STRESS INTENSITY FACTORS (Krt) ...................................................... A-1 APPENDIX B OTHER RCPB FERRITIC COMPONENTS ................................................................. B-1 APPENDIX C D.C. COOK UNIT 1 SURVEILLANCE PROGRAM CREDIBILITY EVALUATION C-1 APPENDIX D VALIDATION OF THE RADIATION TRANSPORT MODELS BASED ON NEUTRON DOSIMETRY MEASUREMENTS ................................................................................ D-1 WCAP-18455-NP February 2020 Revision 1

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Westinghouse Non-Proprietary Class 3 iv LIST OF TABLES Table 2-1 RPV Material Locations .................................................................................................. 2-5 Table 2-2 Reactor Core Power Level ............................................................................................... 2-6 Table 2-3 Calculated Maximum Fast Neutron Fluence Rate (E > 1.0 MeV) at the Pressure Vessel Clad/Base Metal Interface ................................................................................................ 2-7 Table 2-4 Calculated Maximum Fast Neutron Fluence (E > 1.0 MeV) at the Pressure Vessel Clad/Base Metal Interface ................................................................................................ 2-8 Table 2-5 Calculated Maximum Iron Atom Displacement Rate at the Pressure Vessel Clad/Base Metal Interface ................................................................................................................. 2-9 Table 2-6 Calculated Maximum Iron Atom Displacements at the Pressure Vessel Clad/Base Metal Interface .......................................... :.............................................................................. 2-10 Table 2-7 Calculated Maximum Fast Neutron Fluence (E > 1.0 MeV) at the Pressure Vessel Welds and Shells ...................................................................................................................... 2-11 Table 2-8 Calculated Maximum Iron Atom Displacements at the Pressure Vessel Welds and Shells

- - - - 2-12 Table 2-9 Calculated Fast Neutron Fluence Rate and Fluence (E > 1.0 MeV) at the Surveillance Capsule Positions ........................................................................................................... 2-13 Table 2-10 Calculated Iron Atom Displacement Rate and Iron Atom Displacements at the Surveillance Capsule Positions ............................ _. .............................................................................. 2-14 Table 2-11 Calculated Surveillance Capsule Lead Factors .............................................................. 2-15 Table 2-12 Projected Fast Neutron Fluence Rate (E > 1.0 MeV) at the Surveillance Capsule Positions (Future Operation) ......................................................................................................... 2-16 Table 2-13 Calculational Uncertainties ............................................................................................ 2-17 Table 3-1 Summary of the Best-Estimate Chemistry and Initial RTNoT Values for the D.C. Cook Unit 1 Reactor Vessel Materials ............................................................................................... 3-2 Table 3-2 Initial RTNoT Values for the D.C. Cook Unit 1 Reactor Vessel Closure Head and Vessel Flange Materials .............................................................................................................. 3-3 Table 4-1 D.C. Cook Unit 1 Surveillance Capsule Data ................................................................. .4-2 Table 5-1 D.C. Cook Unit 1 Reactor Vessel Intermediate Shell Plate B4406-3 Chemistry Factor Calculation Using Surveillance Capsule Data ................................................................. 5-2 Table 5-2 D.C. Cook Unit 1 Reactor Vessel Intermediate to Lower Shell Circumferential Weld Chemistry Factor Calculation Using Surveillance Capsule Data .................................... 5-3 Table 5-3 Summary ofD.C. Cook Unit I Position 1.1 and 2.1 Chemistry Factors ......................... 5-4 WCAP-18455-NP February 2020 Revision 1

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Westinghouse Non-Proprietary Class 3 V Table 7-1 Fluence Values and Fluence Factors for the Vessel Surface, 1/4T, and 3/4T Locations for the D.C. Cook Unit 1 Reactor Vessel Beltline and Extended Beltline Materials at 48 EFPY

.............................................................................................. ,.......................................... 7-3 Table 7-2 Adjusted Reference Temperature Evaluation for the D.C. Cook Unit 1 Reactor Vessel Beltline and Extended Beltline Materials through 48 EFPY at the 1/4T Location .......... 7-4 Table 7-3 Adjusted Reference Temperature Evaluation for the D.C. Cook Unit 1 Reactor Vessel Beltline and Extended Beltline Materials through 48 EFPY at the 3/4T Location .......... 7-6 Table 7-4 Limiting ART Values for D.C. Cook Unit 1 at 48 EFPY ................................................. 7-8 Table 8-1 ART Values To Be Used In P-T Limit Curves Development for D.C. Cook Unit 1 at 48 EFPY ........................................................................................................................... 8-l Table 8-2 D.C. Cook Unit 1 48 EFPY Heatup Curve Data Points using the 1998 through the 2000 Addenda App. G Methodology (w/ K1c, w/ Flange Requirements, and w/o Margins for Instrumentation Errors) .................................................................................................... 8-5 Table 8-3 D.C. Cook Unit 1 48 EFPY Leak Test Curve Data Points using the 1998 through the 2000 Addenda App. G Methodology (w/ K1c, w/ Flange Requirements, and w/o Margins for Instrumentation Errors) .................................................................................................... 8-6 Table 8-4 D.C. Cook Unit 1 48 EFPY Cooldown Curve Data Points using the 1998 through the 2000 Addenda App. G Methodology for Steady-state (0°F/hr), -20°F/hr, -40°F/hr, -60°F/hr, and

-100°F/hr (w/ K1c, w/ Flange Requirements, and w/o Margins for Instrumentation Errors)

- - - - 8-7 TableA-1 K1t and Vessel Temperature Values for D.C. Cook Unit 1 at 48 EFPY 60°F/hr Heatup Curves (w/o Margins for Instrument Errors) .................................................................. A-2 TableA-2 K1t and Vessel Temperature Values for D.C. Cook Unit 1 at 48 EFPY -100°F/hr Cooldown Curves (w/o Margins for Instrument Errors) .................................................................. A-3 Table C-1 Calculation of Interim Chemistry Factors for the Credibility Evalu1tion Using D.C. Cook Unit 1 Surveillance Data ................................................................................................. C-4 Table C-2 D.C. Cook Unit 1 Calculated Surveillance Capsule Data Scatter about the Best-Fit Line

- - - - C-5 Table C-3 Calculation of Interim Weld Chemistry Factor for the Credibility Evaluation Using All Available Surveillance Data ............................................................................................ C-7 Table C-4 D.C. Cook Unit 1 Calculated Surveillance Weld Metal Data Scatter about the Best-Fit Line Using All Available Surveillance Data............................................................................ C-8 Table C-5 Calculation of Residual versus Fast Fluence .................................................................. C-9 Table D-1 Nuclear Parameters Used in the Evaluation of the In-Vessel Surveillance Capsule Neutron Sensors .......................................................................................................................... D-11 Table D-2 Startup and Shutdown Dates ......................................................................................... D-12 Table D-3 Measured Sensor Activities and Reaction Rates for Surveillance Capsule T ............... D-13 WCAP-18455-NP February 2020 Revision 1

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Westinghouse Non-Proprietary Class 3 vi Table D-4 Measured Sensor Activities and Reaction Rates for Surveillance Capsule X .............. D-14 Table D-5 Measured Sensor Activities and Reaction Rates for Surveillance Capsule Y ............... D-15 Table D-6 Measured Sensor Activities and Reaction Rates for Surveillance Capsule U .............. D-16 Table D-7 Comparison of Measured and Calculated Threshold Foil Reaction Rates for the In-Vessel Capsules ........................................................................................................................ D-17 Table D-8 Comparison of Calculated and Best-Estimate Exposure Rates for the In-Vessel Capsules

.......................................................................... .'........................................................... D-17 WCAP-18455-NP February 2020 Revision 1

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Westinghouse Non-Proprietary Class 3 vii LIST OF FIGURES Figure 2-1 Plan View of the Reactor Geometry at the Core Midplane ............................................ 2-18 Figure 2-2 Section View of the Reactor Geometry - 0° Azimuth .................................................... 2-19 Figure 2-3 Section View of the Reactor Geometry - 4° Azimuth .................................................... 2-20 Figure 8-1 D.C. Cook Unit 1 Reactor Coolant System Heatup Limitations (Heatup Rate of 60°F/hr)

Applicable for 48 EFPY (with Flange Requirements and without Margins for Instrumentation Errors) using the 1998 through the 2000 Addenda App. G Methodology (w/ K1c) ............................................................................................................................. 8-3 Figure 8-2 D.C. Cook Unit 1 Reactor Coolant System Cooldown Limitations (Cooldown Rates of 0,

-20, -40, -60, and -100°F/hr) Applicable for 48 EFPY (with Flange Requirements and without Margins for Instrumentation Errors) using the 1998 through the 2000 Addenda App. G Methodology (w/ Kic) .......................................................................................... 8-4 WCAP-18455-NP February 2020 Revision 1

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Westinghouse Non-Proprietary Class 3 Vlll 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 D.C. Cook Unit 1 reactor vessel. The P-T limit curves were generated using the K1c methodology detailed in the 1998 through the 2000 Addenda Edition of the ASME Code,Section XI, Appendix G. This P-T limit curve generation methodology is consistent with the U.S. Nuclear Regulatory Commission (NRC) approved methodology documented in WCAP-14040-A, Revision 4. The heatup and cooldown P-T limit curves utilize the Adjusted Reference Temperature (ART) values for D.C. Cook Unit 1 calculated using Regulatory Guide 1.99, Revision 2. The limiting ART values in material with a postulated axial flaw were those of the Lower Shell Longitudinal Welds 3-442 A, B, and C (Position 1.1) at both 1/4 thickness (1/4T) and 3/4 thickness (3/4T) locations. The limiting ART values in material with a postulated circumferential flaw were those of the Intermediate to Lower Shell Circumferential Weld at both 1/4T and 3/4T locations. The axially oriented flaw cases are limiting; therefore, only the axially oriented flaw curves are presented in this report.

The P-T limit curves were generated for 48 effective full-power years (EFPY) using a heatup rate of 60°F/hr, and cooldown rates of0° (steady-state), -20°, -40°, -60°, and -100°F/hr. The curves were developed with the flange requirements of 10 CFR 50, Appendix G, but the curves were developed without margins for instrumentation errors. The curves 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 48 EFPY.

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

Appendix C contains the credibility evaluation of the D.C. Cook Unit 1 reactor vessel surveillance data per the requirements of Regulatory Guide 1.99, Revision 2. D.C. Cook Unit 1 fluence values, described in Section 2.0, were used to complete the evaluation.

Appendix D provides the validation of the radiation transport calculation models based on neutron dosimetry measurement.

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Westinghouse Non-Proprietary Class 3 1-1 1 INTRODUCTION Heatup and cooldown P-T limit curves are calculated using the adjusted RTNoT (reference nil-ductility temperature) of the beltline region material of the reactor vessel. The adjusted RTNoT is determined by using the unirradiated reactor vessel material fracture toughness properties, estimating the radiation-induced LlRTNOT, and adding a margin. The unirradiated RTNOT is designated as the higher of either the drop weight nil-ductility transition temperature (TNoT) or the temperature at which the material exhibits at least 50 ft-lb of impact energy and 35-mil lateral expansion (normal to the major working direction) minus 60°F.

RT NOT increases as the material is exposed to fast-neutron radiation. Therefore, to find the most limiting RT NOT at any time period in the reactor's life, LlRT NOT due to the radiation exposure associated with that time period must be added to theunirradiated RTNoT (RTNoT(UJ). The extent of the shift in RTNoT is enhanced by certain chemical elements (such as copper and nickel) present in reactor vessel steels. The NRC has published a method for predicting radiation embrittlement in Regulatory Guide 1.99, Revisipn 2 [1].

Regulatory Guide 1.99, Revision 2 is used for the calculation of ART values (RT NOT(U) + LlRTNOT+ margins for uncertainties) at the 1/4T and 3/4T locations, where Tis the thickness of the vessel at the beltline region measured from the clad/base metal interface.

The heatup and cooldown P-T limit curves documented in this report were generated using the NRC-approved methodology documented in WCAP-14040-A, Revision 4 [2]. Specifically, the K1c methodology from Section XI, Appendix G of the 1998 through the 2000 Addenda Edition of the ASME Code [3] was used. The K1c curve is a lower bound static fracture toughness curve obtained from test data gathered from several different heats of pressure vessel steel. The limiting material is indexed to the K1c 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 D.C. Cook Unit 1 heatup and cooldown P-T limit curves for 48 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 48 EFPY are documented in Section 7 of this report. The fluence projections used in the calculation of the ART values are provided in Section 2 of this report, and a validation of the radiation transport calculation model based on neutron dosimetry measurements is contained in Appendix D.

The P-T limit curves herein were generated without instrumentation errors. The reactor vessel flange requirements of 10 CFR 50, Appendix G [4] have been incorporated in the P-T limit curves. Discussion of the other reactor coolant pressure boundary (RCPB) ferritic components relative to P-T limits is contained in Appendix B.

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

2.1 INTRODUCTION

Discrete ordinates (SN) transport analyses were performed for the D.C. Cook Unit 1 reactor to determine the neutron radiation environment within the reactor pressure vessel (RPV). In these analyses, radiation exposure parameters were established on a plant- and fuel-cycle-specific basis. The dosimetry analysis documented in Appendix D shows that the +/-20% (1 o") acceptance criteria specified in Regulatory Guide 1.190, "Calculational and Dosimetry Methods for Detennining Pressure Vessel Neutron Fluence" [5], is met, based on the measurement-to-calculation (MIC) comparison results for the in-vessel surveillance capsules withdrawn and analyzed to-date. Additional information regarding compliance with Regulatory Guide 1.190 is provided in Appendix D. These validated calculations form the basis for providing projections of the neutron exposure of the RPV through the end oflicense extension (EOLE).

All of the calculations described in this section were based on nuclear cross-section data derived from the Evaluated Nuclear Data File (ENDF) database (specifically, ENDF/B-VI). Additionally, the methods used to develop the calculated pressure vessel fluence are consistent with the NRC-approved methodology described in WCAP-18124-NP-A, Revision 0, "Fluence Determination with RAPTOR-M3G and FERRET" [6]. The neutron transport evaluation methodology described in [6] is based on the guidance of Regulatory Guide 1.190. Note, however, that the NRC Safety Evaluation Report (SER) in [6] states that the applicability of the methodology described in [6] is limited to the traditional RPV beltline region approximated by the RPV region near the active height of the core.

2.2 DISCRETE ORDINATES ANALYSIS In performing the fast neutron exposure evaluations for the RPV, a series of fuel-cycle-specific forward transport calculations were performed using the three-dimensional discrete ordinates code, RAPTOR-M3G

[6], and the BUGLE-96 cross-section library [7]. The BUGLE-96 library provides a coupled 47-neutron and 20-gamma-ray group cross-section data set produced specifically for light water reactor (LWR) applications. In these analyses, anisotropic scattering was treated with a P3 Legendre expansion and the angular discretization was modeled with an S12 order of angular quadrature. Energy- and space-dependent core power distributions were treated on a fuel-cycle-specific basis.

The D.C. Cook Unit 1 reactor is a standard Westinghouse 4-loop design employing reactor internals that include 1.125-inch-thick baffle plates and a fully circumferential thermal shield. The model of the reactor (and reactor cavity) geometry used in the plant-specific evaluation is shown in Figure 2-1 through Figure 2-3.

The model extends radially from the center of the core to 349.89 cm, azimuthally from 0° to 45° (taking advantage of the octant symmetry of the reactor configuration), and axially from -380.26 cm to 358.75 cm with respect to the midplane of the active core. Elevations of key RPV materials relative to the model geometry are provided in Table 2-1.

A plan view of the model geometry at the core midplane is shown in Figure 2-1. In this figure, a single octant is depicted showing the arrangement of the core, reactor internals, core barrel, thermal shield, WCAP-18455-NP February 2020 Revision 1

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Westinghouse Non-Proprietary Class 3 2-2 downcomer, cladding, RPV, reactor cavity, reflective insulation, and bioshield. Depictions of the in-vessel surveillance capsules, including their associated support structures, are also shown.

From a neutronics standpoint, the inclusion of the surveillance capsules and associated support structures in the geometric model is significant. Since the presence of the capsules and support structures has a marked impact on the magnitude of the neutron fluence rate and relative neutron and gamma ray spectra at dosimetry locations within the capsules, a meaningful evaluation of the radiation environment internal to the capsules can be made only when these perturbation effects are accounted for in the transport calculations.

Section views of the model geometry are shown in Figure 2-2 and Figure 2-3. Note that the stainless steel former plates located between the core baffle and barrel regions are shown in these figures.

When developing the reactor model shown in Figure 2-1 through Figure 2-3, 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. These coolant temperatures were varied on a cycle-specific basis. The reactor core itself was treated as a homogeneous mixture of fuel, cladding, water, and miscellaneous core structures such as fuel assembly grids, and guide tubes.

The geometric mesh description of the reactor model shown in Figure 2-1 through Figure 2-3 consisted of 220 radial by 188 azimuthal by 374 axial intervals. Mesh sizes were chosen to ensure sufficient resolution of the stair-step-shaped baffle plates as well as an adequate number of meshes throughout the radial and axial regions of interest. The pointwise inner iteration convergence criterion utilized in the calculations was set at a value of0.001.

The core power distributions used in the plant-specific transport analysis were taken from nuclear design documentation. The data extracted included fuel assembly-specific initial enrichments, beginning-of-cycle bumups and end-of-cycle bumups. Appropriate axial power distributions were also obtained.

For each fuel cycle of operation, fuel-assembly-specific enrichment and bumup data were used to generate the spatially dependent neutron source throughout the reactor core. This source description included the spatial variation of isotope-dependent (U-235, U-238, Pu-239, Pu-240, Pu-241, and Pu-242) fission spectra, neutron emission rate per fission, and energy release per fission based on the bumup history of individual fuel assemblies. These fuel-assembly-specific neutron source strengths derived from the detailed isotopics were then converted from fuel pin Cartesian coordinates to the spatial mesh arrays used in the discrete ordinates calculations.

In Table 2-1, axial and azimuthal locations of the RPV materials are provided. The axial position of each material is indexed to z = 0.0 cm, which corresponds to the midplane of the active fuel stack.

Cycle-specific calculations were performed for Cycles 1-29, with core thermal powers given in Table 2-2.

Note that future fluence projection data beyond Cycle 29 are based on the average core power distributions and reactor operating conditions of Cycles 25-27, but include a 1.1 bias on the core thermal power. Note, at the time of development of the fluence model, Cycles 28 and 29 were yet to be completed and, thus, the results for these cycles are based on the cycle design data, whereas Cycles 25-27 had completed and, as WCAP-18455-NP February 2020 Revision 1

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Westinghouse Non-Proprietary Class 3 2-3 such, the results for these cycles are based on actual operating data. Therefore, only Cycles 25-27, being the most recently completed cycles, are used for projections per [33].

Neutron fluence rate and fluence for the RPV are given in Table 2-3, Table 2-4, and Table 2-7. Similarly, iron atom displacement rate and iron atom displacements for the RPV are provided in Table 2-5, Table 2-6, and Table 2-8. The data presented represent the maximum neutron exposures experienced by RPV materials. The reported data also consider both the inner and outer radius of the RPV base metal, and account for the possibility of higher neutron exposure values occurring on the outer surface of the RPV (as compared to the inner surface) for materials that are distant from the active core. In each case, the data are provided for each operating cycle of the reactor. Note that, for any given fuel cycle, the location of the maximum neutron exposure rate may or may not coincide with the location of the maximum neutron exposure.

Calculated neutron exposure projections of the RPV are provided in Table 2-4 and Table 2-6 through Table 2-8. These projections were based on the average spatial power distributions and reactor operating conditions of Cycles 25-27, but included a 1.1 bias on the core thermal power. The projected results will remain valid as long as future plant operation is consistent with these assumptions.

Results of the discrete ordinates transport analyses pertinent to the surveillance capsule evaluations are provided in Table 2-9 through Table 2-11. In Table 2-9, the calculated fast neutron fluence rate and fluence (E > 1. 0 Me V) are provided at the geometric center of the capsules and at core midplane as a function of operating time. Similar data presented in terms of iron atom displacement rate (dpa/s) and integrated iron atom displacements (dpa) are given in Table 2-10.

In Table 2-11, lead factors associated with the surveillance capsules are provided as a function of operating time. The lead factor is defined as the ratio of the neutron fluence (E > 1.0 MeV) at the geometric center of the surveillance capsule to the maximum neutron fluence (E > 1.0 MeV) at the pressure vessel clad/base metal interface.

All surveillance capsules at the 40. 0 first-octant-equivalent (FOE) azimuthal locations have been removed from the RPV, so neutron exposure data at the 40° FOE azimuthal positions beyond Cycle 10 are unnecessary (because there are no capsules receiving any fluence). However, if any capsules were to be re-inserted or re-located to the 40° FOE azimuthal locations, it would be necessary to know the fast neutron fluence rate at the surveillance capsule holder position(s). To allow for the determination of potential fast fluence accumulation, the projected fast fluence rate (E > 1.0 MeV) at each surveillance capsule location is provided in Table 2-12. Projections of future operation are based on the spatial power distributions and reactor operating conditions of Cycles 25-27, but include a 1.1 bias on the core thermal power. This bias is intended to account for cycle-to-cycle variations in peripheral fuel assembly relative powers that are expected to occur during the time period of future operation evaluated in this report. Note that RPV neutron exposure rates are dominated by neutron leakage from the peripheral fuel assemblies. The additional fast fluence accumulated for any re-inserted/re-located capsule can be determined by multiplying the fast fluence rate value in Table 2-12 for the appropriate capsule position with the irradiation duration in effective full-power seconds (EFPS).

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Westinghouse Non-Proprietary Class 3 2-4 2.3 CALCULATIONAL UNCERTAINTIES The uncertainty associated with the calculated neutron exposure of the RPV is based on the recommended approach provided in Regulatory Guide 1.190. In particular, the qualification of the methodology used in the plant-specific neutron exposure evaluation is carried out in the following four stages:

1. Comparisons of calculations with benchmark measurements from the. Pool Critical Assembly (PCA) simulator (NUREG/CR-6454, "Pool Critical Assembly Pressure Vessel Facility Benchmark" [8]) at the Oak Ridge National Laboratory (ORNL) and the VENUS-I experiment.

2. Comparison of calculations with surveillance capsule and reactor cavity measurements from the H.B. Robinson power reactor benchmark experiment (NUREG/CR-6453, "H.B. Robinson-2 Pressure Vessel Benchmark" [9]).

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 (WCAP-18124-NP-A, "Fluence Determination with RAPTOR-M3G and FERRET"

[6]).

4. Comparison of the calculations with all available dosimetry results from the RPV measurement programs carried out at the D.C. Cook Unit I (Appendix D).

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 and geometric variables that impact power reactor calculations.

The second phase of the qualification (H.B. Robinson comparisons) addressed uncertainties 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 method approximations as well as to a lack of knowledge relative to various plant-specific parameters. The overall calculational uncertainty applicable to the D.C. Cook Unit 1 analyses was established from the results of these three phases of the methods qualification.

The fourth phase of the uncertainty assessment (comparisons of plant-specific dosimetry measurements) was used solely to demonstrate the adequacy of the transport calculations and to confirm the uncertainty estimates associated with the analytical results. The comparison was used only as a check and was not used to bias the final results in any way.

Table 2-13 summarizes the uncertainties developed from the first three phases of the methodology qualification. Additional information pertinent to these evaluations is provided in WCAP-18124-NP-A. The net calculational uncertainty was determined by combining the individual components in quadrature.

Therefore, the resultant uncertainty was treated as random and no systematic bias was applied to the analytical results. The plant-specific measurement comparisons given in Appendix D support these uncertainty assessments for D.C. Cook Unit 1.

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Westinghouse Non-Proprietary Class 3 2-5 Table 2-1 RPV Material Locations Material Axial Elevation<al Azimuth<bl (cm) (Degrees)

Outlet Nozzle to Upper Shell Weld -

Lowest Extent 274.30 22.0 Inlet Nozzle to Upper Shell Weld -

Lowest Extent 269.22 23.0 Upper Shell 239.45 to 358.75(c) 0.0 to 45.0 Upper Shell to Intermediate Shell Circumferential Weld 235.96 to 239.45 0.0 to 45.0 Intermediate Shell -37.01 to 235.96 0.0 to 45.0 Intermediate Shell Longitudinal Welds Weld 1 (60°)Cctl,(eJ -37.01 to 235.96 29.5 to 30.5 Weld 2 (180°)Cd),(e) -37.01 to 235.96 0.0 to 0.5 Weld 3 (300°)Cctl,CeJ -37.01 to 235.96 29.5 to 30.5 Intennediate Shell to Lower Shell Circumferential Weld -40.50 to -37.01 0.0 to 45.0 Lower Shell -307.44 to -40.50 0.0 to 45.0 Lower Shell Longitudinal Welds Weld 1 (0°)CctJ.<eJ -307.44 to -40.50 0.0 to 0.5 Weld 2 (120°)CctJ,Cel -307.44 to -40.50 29.5 to 30.5 Weld 3 (240°)Cd),(e) -307.44 to -40.50 29.5 to 30.5 Lower Shell to Lower Head Circumferential Weld -311.25 to -307.44 0.0 to 45.0 Notes:

(a) Values listed are indexed to Z = 0.0 at the midplane of the active fuel stack.

(b) Azimuthal angles are given relative to the cardinal axes at 0°, 90°, 180°, and 270°.

(c) Elevation given is equal to the maximum elevation of the reactor model.

(d) Azimuthal angles are given relative to 0° as shown on reactor vessel drawing (e) This weld is approximately 1.375 inches in width. At the RPV inner radius (86.719 inches), this corresponds to -1 °.

WCAP-18455-NP February 2020

. Revision 1

- - This record was final approved on 2/14/2020 11 :56:36 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 2-6 Table 2-2 Reactor Core Power Level Core Thermal Power Cycle (MWt) 1 3250 2 3250 3 3250 4 3250 5 3250 6 3250 7 3250 8 3250 9 3250 10 3250 11 3250 12 3250 13 3250 14 3250 15 3250 16 3250 17 3250 18 3278.5(*)

19 3304 20 3304 21 3304 22 3304 23 3304 24 3304 25 3304 26 3304 27 3304 28 3304 29 3304 Note:

(a) A reactor power uprate from 3250 MWt to 3304 MWt was implemented at a burnup of 7575 MWD/MTU, where the total cycle length was 16046 MWD/MTU. Therefore, a burnup-weighted average power level of 3278.5 MWt is used for this cycle.

WCAP-18455-NP February 2020 Revision 1

- - This record was final approved on 2/14/2020 11 :56:36 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietal)' Class 3 2-7 Table 2-3 Calculated Maximum Fast Neutron Fluence Rate (E > 1.0 MeV) at the Pressure Vessel Clad/Base Metal Interface Cycle Cumulative Fluence Rate (n/cm 2-s)

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

1 1.27 1.27 6.20E+09 1.04E+!O l.31E+!O 2.02E+!O 2.03E+!O 2 0.78 2.04 7.33E+09 1.24E+!O 1.55E+l0 2.59E+!O 2.59E+!O 3 0.70 2.75 6.91E+09 l.16E+!O 1.47E+l0 2.44E+!O 2.45E+!O 4 0.73 3.48 6.76E+09 l.14E+!O 1.44E+I0 2.36E+!O 2.36E+!O 5 0.74 4.22 6.74E+09 l.14E+!O 1.47E+ 10 2.45E+!O 2.46E+!O 6 0.72 4.95 6.74E+09 l.14E+!O 1.42E+!O 2.31E+!O 2.31E+!O 7 0.73 5.67 6.78E+09 l.15E+!O 1.43E+!O 2.32E+IO 2.33E+!O 8 1.12 6.80 6.90E+09 1.02E+!O 9.73E+09 l.32E+!O l.33E+!O 9 1.19 7.98 6.62E+09 8.77E+09 9.19E+09 l.30E+!O l.31E+!O 10 1.19 9.17 5.75E+09 9.72E+09 9.76E+09 1.25E+!O 1.25E+!O II 1.14 10.32 5.71E+09 9.14E+09 9.78E+09 1.27E+!O 1.28E+!O 12 1.19 11.51 5.55E+09 8.70E+09 9.46E+09 l.29E+!O l.30E+!O 13 1.18 12.68 4.70E+09 7.41E+09 9.32E+09 l.33E+!O l.33E+!O 14 1.04 13.72 4.45E+09 7.07E+09 9.02E+09 l.23E+!O l.24E+!O 15 1.16 14.88 4.51E+09 8.24E+09 l.13E+!O 1.78E+!O 1.79E+!O 16 0.35 15.23 3.92E+09 6.66E+09 9.69E+09 1.56E+!O l.57E+!O 17 1.21 16.44 4.90E+09 8.19E+09 1.00E+IO l.52E+!O 1.52E+!O 18 1.18 17.62 4.53E+09 7.67E+09 1.12E+!O 1.84E+!O 1.84E+!O 19 1.29 18.91 3.83E+09 6.28E+09 1.02E+IO 1.55E+!O l.56E+!O 20 1.34 20.26 3.74E+09 6.48E+09 9.96E+09 1.45E+!O 1.46E+!O 21 1.34 21.60 3.93E+09 6.94E+09 1.02E+!O 1.57E+!O l.57E+IO 22 0.58 22.18 3.57E+09 5.75E+09 9.64E+09 1.51E+!O 1.51E+!O 23 1.39 23.57 3.87E+09 6.64E+09 8.57E+09 1.28E+!O l.29E+IO 24 1.39 24.96 3.87E+09 6.70E+09 1.06E+!O 1.62E+!O 1.62E+!O 25 1.33 26.29 3.96E+09 6.21E+09 8.85E+09 1.52E+!O 1.52E+!O 26 1.23 27.52 4.13E+09 6.96E+09 l.05E+IO 1.65E+!O 1.66E+!O 27 1.37 28.89 4.83E+09 7.16E+09 1.04E+!O 1.58E+!O l.59E+!O 28(b) 1.25 30.14 4.98E+09 7.67E+09 1.05E+!O 1.64E+!O 1.65E+!O 29(c) 1.37 31.51 4.96E+09 7.81E+09 1.12E+ 10 1.71E+!O l.72E+!O Note(s):

(a) Values correspond to an azimuthal angle of 44°.

(b) Cycle 28 was the current operating cycle at the time these neutron exposures were determined. Values listed for this cycle are projections based on the Cycle 28 design data.

(c) Values listed for this cycle are projections based on the Cycle 29 design data.

WCAP-18455-NP February 2020 Revision 1

- - This record was final approved on 2/14/2020 11 :56:36 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 2-8 Table 2-4 Calculated Maximum Fast Neutron Fluence (E > 1.0 MeV) at the Pressure Vessel Clad/Base Metal Interface Cycle Cumulative Fluence (n/cm 2)

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

(EFPY)

I 1.27 1.27 2.48E+l7 4.14E+l7 5.23E+l7 8.08E+l7 8.IOE+l7 2 0.78 2.04 4.27E+l7 7.14E+l7 8.96E+l7 l.43E+l8 1.43E+l8 3 0.70 2.75 5.80E+l 7 9.72E+l7 1.22E+l8 1.97E+l8 1.97E+l8 4 0.73 3.48 7.37E+l7 l.24E+l8 l.56E+l8 2.51E+l8 2.52E+l8 5 0.74 4.22 8.95E+l7 1.50E+l8 l.90E+l8 3.09E+18 3.09E+l8 6 0.72 4.95 l.05E+l8 1.76E+18 2.22E+18 3.61E+18 3.62E+18 7 0.73 5.67 1.20E+18 2.03E+18 2.55E+18 4.15E+l8 4.15E+18 8 1.12 6.80 1.45E+18 2.38E+18 2.89E+18 4.61E+18 4.62E+18 9 1.19 7.98 1.68E+18 2.70E+18 3.23E+18 5.09E+18 5.10E+18 10 1.19 9.17 1.90E+18 3.07E+18 3.59E+18 5.56E+l8 5.57E+18 11 1.14 10.32 2.10E+l8 3.40E+18 3.95E+18 6.01E+18 6.03E+l8 12 1.19 11.51 2.31E+l8 3.72E+18 4.29E+l8 6.49E+l8 6.51E+l8 13 1.18 12.68 2.49E+l8 3.99E+18 4.64E+18 6.98E+l8 7.00E+l8 14 1.04 13.72 2.63E+18 4.22E+18 4.93E+l8 7.39E+l8 7.41E+18 15 1.16 14.88 2.80E+l8 4.51E+18 5.33E+l8 8.0IE+18 8.04E+18 16 0.35 15.23 2.84E+l8 4.58E+18 5.43E+18 8.18E+18 8.21E+l8 17 1.21 16.44 3.03E+l8 4.90E+18 5.81E+l8 8.76E+18 8.78E+l8 18 1.18 17.62 3.20E+l8 5.18E+18 6.23E+l8 9.43E+l8 9.46E+18 19 1.29 18.91 3.35E+l8 5.44E+l8 6.64E+l8 1.0!E+l9 1.0IE+l9 20 1.34 20.26 3.51E+l8 5.71E+l8 7.07E+l8 l.07E+l9 l.07E+19 21 1.34 21.60 3.68E+l8 6.0IE+18 7.50E+18 1.13E+19 l.14E+19 22 0.58 22.18 3.74E+l8 6.11E+l8 7.67E+l8 l.16E+19 l.16E+l9 23 1.39 23.57 3.91E+l8 6.40E+18 8.05E+l8 l.22E+l9 l.22E+19 24 1.39 24.96 4.08E+18 6.70E+l8 8.51E+18 l.29E+19 1.29E+l9 25 1.33 26.29 4.25E+l8 6.96E+18 8.89E+18 1.35E+19 1.35E+19 26 1.23 27.52 4.41E+18 7.23E+l8 9.29E+18 1.41E+l9 1.42E+19 27 1.37 28.89 4.62E+18 7.54E+18 9.74E+l8 1.48E+!9 1.49E+19 28<bl 1.25 30.14 4.81E+18 7.84E+18 1.02E+19 1.55E+19 1.55E+19 29(c) 1.37 31.51 5.03E+l8 8.18E+l8 1.06E+l9 1.62E+19 1.62E+19 Future(d) -- 36.00 5.70E+l8 9.23E+18 l.22E+l9 1.87E+19 1.87E+19 Future<d) -- 42.00 6.60E+18 1.06E+19 1.43E+19 2.20E+19 2.20E+19 Future<dl -- 48.00 7.49E+18 1.21E+19 1.63E+19 2.53E+l9 2.53E+19 Future<d) -- 54.00 8.39E+18 1.35E+19 1.84E+l9 2.86E+19 2.86E+19 Future(dl -- 60.00 9.28E+18 1.49E+19 2.05E+19 3.19E+!9 3.20E+19 Note(s):

(a) Values correspond to an azimuthal angle of 44°.

(b) Cycle 28 was the current operating cycle at the time these neutron exposures were determined. Values listed for this cycle are projections based on the Cycle 28 design data.

(c) Values listed for this cycle are projections based on the Cycle 29 design data.

(d) Values beyond Cycle 29 are based on the average core power distributions and reactor operating conditions of Cycles 25-27, but include a 1.1 bias on the core thermal power.

WCAP-18455-NP February 2020 Revision 1

- - This record was final approved on 2/14/2020 11 :56:36 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 2-9 Table 2-5 Calculated Maximum Iron Atom Displacement Rate at the Pressure Vessel Clad/Base Metal Interface Cycle Cumulative Displacement Rate (dpa/s)

Cycle Length Operating Time oo 15° 30° 45° Maximum<*)

(EFPY) (EFPY)

I 1.27 1.27 l.OOE-11 l.66E-11 2.llE-11 3.27E-11 3.28E-11 2 0.78 2.04 1.18E-11 1.99E-l 1 2.SIE-11 4.18E-11 4.19E-l 1 3 0.70 2.75 1.12E-11 1.86E-11 2.38E-l 1 3.95E-11 3.96E-11 4 0.73 3.48 l.09E-11 l.83E-11 2.32E-11 3.81E-11 3.82E-11 5 0.74 4.22 1.09E-11 l.82E-11 2.37E-11 3.97E-11 3.98E-11 6 0.72 4.95 l.09E-11 l.83E-11 2.29E-11 3.73E-11 3.74E-11 7 0.73 5.67 1.IOE-11 1.84E-11 2.31E-11 3.76E-11 3.77E-11 8 1.12 6.80 I.I IE-I I l.63E-11 l.57E-11 2.12E-11 2.13E-11 9 1.19 7.98 l.07E-11 l.40E-11 l.48E-11 2.l lE-11 2.l lE-11 10 1.19 9.17 9.24E-12 I.SSE-I I l.57E-11 2.02E-11 2.03E-11 11 1.14 10.32 9.23E-12 l.46E-11 l.58E-11 2.06E-11 2.06E-11 12 1.19 11.51 8.89E-12 l.39E-11 l.52E-11 2.07E-11 2.08E-11 13 1.18 12.68 7.59E-12 1.19E-11 I.SOE-I I 2.14E-11 2.lSE-11 14 1.04 13.72 7.14E-12 l.13E-11 1.45E-11 1.99E-1 l l.99E-l l 15 1.16 14.88 7.24E-12 1.32E-11 l.81E-ll 2.86E-11 2.86E-11 16 0.35 15.23 6.34E-12 l.06E-11 l.56E-ll 2.52E-l 1 2.53E-11 17 1.21 16.44 7.87E-12 1.31E-11 l.61E-11 2.43E-11 2.44E-11 18 1.18 17.62 7.27E-12 l.23E-11 l.81E-11 2.95E-1 l 2.95E-11 19 1.29 18.91 6.ISE-12 1.0IE-11 l.64E-11 2.49E-11 2.SOE-11 20 1.34 20.26 6.0IE-12 l.04E-11 l.60E-11 2.33E-11 2.34E-11 21 1.34 21.60 6.32E-12 1.! IE-11 l.64E-11 2.SIE-11 2.52E-l 1 22 0.58 22.18 5.77E-12 9.22E-12 I.SSE-I I 2.43E-1 l 2.44E-l 1 23 1.39 23.57 6.21E-12 l.06E-11 1.38E-11 2.06E-11 2.06E-11 24 1.39 24.96 6.21E-12 l.07E-11 l.70E-11 2.60E-11 2.61E-11 25 1.33 26.29 6.35E-12 9.93E-12 l.42E-11 2.44E-11 2.44E-11 26 1.23 27.52 6.62E-12 I.I IE-I I l.69E-11 2.66E-11 2.66E-11 27 1.37 28.89 7.74E-12 1.ISE-11 l.68E-11 2.55E-11 2.56E-11 23<h> 1.25 30.14 7.98E-12 l.23E-11 l.69E-l l 2.64E-11 2.65E-11 29(c) 1.37 31.51 7.97E-12 l.25E-11 l.80E-11 2.76E-11 2.76E-11 Note(s):

(a) Values correspond to an azimuthal angle of 44°.

(b) Cycle 28 was the current operating cycle at the time these neutron exposures were determined. Values listed for this cycle are projections based on the Cycle 28 design data.

(c) Values listed for this cycle are projections based on the Cycle 29 design data.

WCAP-18455-NP February 2020 Revision 1

- - This record was final approved on 2/14/2020 11 :56:36 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 2-10 Table 2-6 Calculated Maximum Iron Atom Displacements at the Pressure Vessel Clad/Base Metal Interface Cycle Cumulative Displacements (dpa)

Cycle Length Operating Time oo 15° 30° 45° Maximum<*)

(EFPY) (EFPY) 1 1.27 1.27 4.0IE-04 6.63E-04 8.44E-04 1.31 E-03 l.31E-03 2 0.78 2.04 6.86E-04 l.14E-03 I.45E-03 2.31E-03 2.31 E-03 3 0.70 2.75 9.34E-04 l.56E-03 l.98E-03 3.18E-03 3. l 9E-03 4 0.73 3.48 l.19E-03 l.98E-03 2.51E-03 4.07E-03 4.08E-03 5 0.74 4.22 l.44E-03 2.40E-03 3.07E-03 5.00E-03 5.0IE-03 6 0.72 4.95 l.69E-03 2.82E-03 3.59E-03 5.85E-03 5.86E-03 7 0.73 5.67 l.94E-03 3.24E-03 4.12E-03 6.71 E-03 6.72E-03 8 1.12 6.80 2.33E-03 3.81E-03 4.66E-03 7.46E-03 7.47E-03 9 1.19 7.98 2.72E-03 4.33E-03 5.21E-03 8.23E-03 8.25E-03 IO 1.19 9.17 3.06E-03 4.90E-03 5.80E-03 8.99E-03 9.0IE-03 II 1.14 10.32 3.40E-03 5.43E-03 6.36E-03 9.73E-03 9.75E-03 12 1.19 11.51 3.72E-03 5.94E-03 6.93E-03 l.05E-02 I.05E-02 13 1.18 12.68 4.0IE-03 6.38E-03 7.48E-03 l.13E-02 l.13E-02 14 1.04 13.72 4.24E-03 6.75E-03 7.95E-03 l.19E-02 l.20E-02 15 1.16 14.88 4.49E-03 7.21E-03 8.59E-03 l.30E-02 l.30E-02 16 0.35 15.23 4.56E-03 7.33E-03 8.76E-03 l.32E-02 l.33E-02 17 1.21 16.44 4.86E-03 7.82E-03 9.36E-03 l.42E-02 l .42E-02 18 1.18 17.62 5.13E-03 8.28E-03 I.OOE-02 l.53E-02 l.53E-02 19 1.29 18.91 5.38E-03 8.69E-03 I.07E-02 l.63E-02 l.63E-02 20 1.34 20.26 5.63E-03 9.13E-03 l.14E-02 l.73E-02 I.73E-02 21 1.34 21.60 5.90E-03 9.60E-03 1.21E-02 l.83E-02 l.84E-02 22 0.58 22.18 6.0lE-03 9.76E-03 l.23E-02 l.88E-02 I.88E-02 23 1.39 23.57 6.28E-03 l.02E-02 l.29E-02 l.97E-02 l.97E-02 24 1.39 24.96 6.55E-03 l.07E-02 l.37E-02 2.08E-02 2.08E-02 25 1.33 26.29 6.82E-03 l.l lE-02 l.43E-02 2.l 8E-02 2.19E-02 26 1.23 27.52 7.07E-03 l.15E-02 l.49E-02 2.28E-02 2.29E-02 27 1.37 28.89 7.41 E-03 l.20E-02 l.57E-02 2.39E-02 2.40E-02 23<bJ 1.25 30.14 7.72E-03 l.25E-02 l.63E-02 2.50E-02 2.50E-02 29(c) 1.37 31.51 8.07E-03 l.31E-02 1.71 E-02 2.62E-02 2.62E-02 Future(d) -- 36.00 9.14E-03 l.48E-02 I.96E-02 3.0IE-02 3.02E-02 Future(d) -- 42.00 l.06E-02 l.70E-02 2.29E-02 3.54E-02 3.55E-02 FutureCd) -- 48.00 l.20E-02 l.93E-02 2.63E-02 4.07E-02 4.08E-02 FutureCd) -- 54.00 I .35E-02 2.15E-02 2.96E-02 4.60E-02 4.61E-02 FutureCd) -- 60.00 l.49E-02 2.38E-02 3.29E-02 5.13E-02 5.14E-02 Note(s):

(a) Values correspond to an azimuthal angle of 44°.

(b) Cycle 28 was the current operating cycle at the time these neutron exposures were determined. Values listed for this cycle are projections based on the Cycle 28 design data.

(c) Values listed for this cycle are projections based on the Cycle 29 design data.

(d) Values beyond Cycle 29 are based on the average core power distributions and reactor operating conditions of Cycles 25-27, but include a 1.1 bias on the core thermal power.

WCAP-18455-NP February 2020 Revision 1

- - This record was final approved on 2/14/2020 11 :56:36 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 2-11 Table 2-7 Calculated Maximum Fast Neutron Fluence (E > 1.0 MeV) at the Pressure Vessel Welds and Shells<a>

Fast Neutron Fluence 1n/cm2 )

Material 31.51 EFPY 36EFPY 42 EFPY Outlet Nozzle Forging to 8.39E+l5 9.62E+15 1.13E+16 Upper Shell Welds Inlet Nozzle Forging to 1.13E+16 l.29E+16 l.51E+16 Upper Shell Welds Upper Shell l.03E+l 7 1.18E+17 l.39E+l 7 Upper Shell to Intermediate Shell l.38E+l7 l.59E+17 l.87E+l 7 Circumferential Weld Intermediate Shell l.62E+19 l.87E+19 2.20E+19 Intermediate Shell to Lower Shell l.58E+l9 l.82E+l9 2.14E+19 Circumferential Weld Lower Shell l.62E+19 l.87E+19 2.20E+19 Lower Shell to Lower Vessel Head1 2.93E+ 15 3.35E+l5 3.92E+15 Circumferential Weld Intermediate Shell Longitudinal Weld at 0° 4.99E+18 5.64E+l8 6.52E+18 Intennediate Shell Longitudinal Weld at 30° l.07E+19 l.23E+19 l.43E+19 Lower Shell Longitudinal Weld at 0° 5.06E+18 5.74E+18 6.64E+l8 Lower Shell Longitudinal Weld at 30° l.08E+ 19 l.24E+19 l.45E+l9 Fast Neutron Fluence (n/cm 2)

Material 48 EFPY 54 EFPY 60 EFPY Outlet Nozzle Forging to l.29E+16 l.45E+l6 l.62E+16 Upper Shell Welds Inlet Nozzle Forging to l.73E+ 16 l.95E+16 2.l 7E+16 Upper Shell Welds Upper Shell l.60E+l 7 l.81E+17 2.02E+17 Upper Shell to Intermediate Shell 2.15E+17 2.43E+l 7 2.71E+17 Circumferential Weld Intermediate Shell 2.53E+19 2.85E+19 3.18E+19 Intermediate Shell to Lower Shell 2.47E+19 2.79E+19 3.11E+19 Circumferential Weld Lower Shell 2.53E+19 2.86E+19 3.20E+19 Lower Shell to Lower Vessel Head 4.49E+15 5.05E+15 5.62E+15 Circumferential Weld Intermediate Shell Longitudinal Weld at 0° 7.40E+18 8.28E+18 9.16E+18 Intennediate Shell Longitudinal Weld at 30° l.64E+19 l.85E+19 2.05E+19 Lower Shell Longitudinal Weld at 0° 7.54E+18 8.44E+18 9.34E+18 Lower Shell Longitudinal Weld at 30° l.66E+19 l.87E+19 2.08E+19 Note:

(a) Fluence projection 'for future cycles are based on the average core power distributions and reactor operating conditions of Cycles 25-27, but include a 1.1 bias on the core thermal power.

WCAP-18455-NP February 2020 Revision 1

- - This record was final approved on 2/14/2020 11 :56:36 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 2-12 Table 2-8 Calculated Maximum Iron Atom Displacements at the Pressure Vessel Welds and Shells<a>

Displacements (d pa)

Material 31.51 EFPY 36 EFPY 42EFPY Outlet Nozzle Forging to 4.47E-05 5.09E-05 5.92E-05 Upper Shell Welds Inlet Nozzle Forging to 5.22E-05 5.95E-05 6.93E-05 Upper Shell Welds Unner Shell* 1.98E-04 2.28E-04 2.68E-04 Upper Shell to Intermediate Shell 2.61E-04 3.0lE-04 3.54E-04 Circumferential Weld Intennediate Shell 2.62E-02 3.02E-02 3.55E-02 Intermediate Shell to Lower Shell 2.56E-02 2.95E-02 3.47E-02 Circumferential Weld Lower Shell 2.61E-02 3.0lE-02 3.54E-02 Lower Shell to Lower Vessel Head 2.22E-05 2.54E-05 2.96E-05 Circumferential Weld Intermediate Shell Longitudinal Weld at 0° 8.06E-03 9.12E-03 l.OSE-02 Intennediate Shell Longitudinal Weld at 30° 1.73E-02 1.97E-02 2.31E-02 Lower Shell Longitudinal Weld at 0° 8.llE-03 9.20E-03 l.06E-02 Lower Shell Longitudinal Weld at 30° 1.74E-02 1.99E-02 2.33E-02 Displacements (dpa)

Material 48 EFPY 54 EFPY 60 EFPY Outlet Nozzle Forging to 6.75E-05 7.58E-05 8.42E-05 Unner Shell Welds Inlet Nozzle Forging to 7.90E-05 8.88E-05 9.86E-05 Unner Shell Welds Unner Shell 3.08E-04 3.48E-04 3.88E-04 Upper Shell to Intermediate Shell 4.06E-04 4.59E-04 5.12E-04 Circumferential Weld Intennediate Shell 4.08E-02 4.61E-02 5.14E-02 Intermediate Shell to Lower Shell 3.99E-02 4.51E-02 5.03E-02 Circumferential Weld Lower Shell 4.07E-02 4.60E-02 5.13E-02 Lower Shell to Lower Vessel Head 3.39E-05 3.81E-05 4.24E-05 Circumferential Weld Intermediate Shell Longitudinal Weld at 0° l.20E-02 l.34E-02 1.48E-02 Intermediate Shell Longitudinal Weld at 30° 2.64E-02 2.98E-02 3.3 lE-02 Lower Shell Longitudinal Weld at 0° 1.21E-02 1.35E-02 l.SOE-02 Lower Shell Longitudinal Weld at 30° 2.67E-02 3.00E-02 3.34E-02 Note:

(a) Fluence projection for future cycles are based on the average core power distributions and reactor operating conditions of Cycles 25-27, but include a 1.1 bias on the core thermal power.

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Westinghouse Non-Proprietary Class 3 2-13 Table 2-9 Calculated Fast Neutron Fluence Rate and Fluence (E > 1.0 MeV) at the Surveillance Capsule Positions Cycle Cumulative Fluence Rate (n/cm 2-s) Fluence (n(cm 2)

Cycle Length Operating Time 40 40° 40 40° (EFPY) (EFPY)

I 1.27 1.27 l.97E+IO 6.83E+IO 7.87E+17 2.73E+18 2 0.78 2.04 2.29E+IO 8.54E+l0 1.35E+18 4.82E+18 3 0.70 2.75 2.20E+IO 8.25E+IO l.84E+18 6.66E+18 4 0.73 3.48 2.13E+IO 7.96E+IO 2.33E+18 8.50E+18 5 0.74 4.22 2.15E+IO 8.30E+IO 2.83E+l8 l.04E+19 6 0.72 4.95 2.15E+IO 7.8IE+IO 3.32E+18 l.22E+19 7 0.73 5.67 2.16E+IO 7.87E+IO 3.81E+I8 1.40E+19 8 1.12 6.80 2.13E+l0 4.39E+IO 4.57E+l8 l.56E+19 9 1.19 7.98 2.06E+IO 4.31E+IO 5.34E+l8 l.72E+19 10 1.19 9.17 l.81E+l0 4.23E+l0 6.02E+l8 l.88E+l9 11 1.14 10.32 l.81E+IO 4.29E+IO 6.67E+l8 2.03E+l9 12 1.19 11.51 l.73E+l0 4.39E+IO 7.32E+l8 2.20E+l9 13 1.18 12.68 l.51E+IO 4.56E+IO 7.89E+l8 2.37E+l9 14 1.04 13.72 l.39E+IO 4.24E+IO 8.34E+l8 2.51E+l9 15 1.16 14.88 1.37E+IO 5.83E+IO 8.85E+l8 2.72E+l9 16 0.35 15.23 l.26E+l0 5.34E+IO 8.98E+l8 2.78E+l9 17 1.21 16.44 l.53E+IO 5.07E+IO 9.57E+l8 2.97E+19 18 1.18 17.62 l.40E+l0 6.17E+IO l.OIE+l9 3.20E+l9 19 1.29 18.91 l.18E+l0 5.23E+IO l.06E+l9 3.42E+19 20 1.34 20.26 l.16E+IO 4.93E+IO 1.IIE+l9 3.62E+l9 21 1.34 21.60 l.22E+IO 5.23E+l0 1.16E+l9 3.85E+l9 22 0.58 22.18 1.13E+IO 5.14E+10 1.18E+l9 3.94E+l9 23 1.39 23.57 l.18E+IO 4.20E+IO 1.23E+l9 4.12E+l9 24 1.39 24.96 1.19E+IO 5.45E+10 1.28E+l9 4.36E+l9 25 1.33 26.29 l.22E+IO 5.05E+l0 l.33E+l9 4.58E+19 26 1.23 27.52 l.28E+IO 5.56E+IO l.38E+l9 4.79E+l9 27 1.37 28.89 l.45E+IO 5.17E+l0 l.45E+l9 5.0IE+l9 28<*) 1.25 30.14 l.50E+IO 5.36E+l0 l.51E+l9 5.23E+19 29(b) 1.37 31.51 l.50E+IO 5.60E+IO l.57E+l9 5.47E+19 Future<c) -- 36.00 -- -- l.78E+l9 6.29E+19 Future(c) -- 42.00 -- -- 2.05E+l9 7.38E+l9 Future<cl -- 48.00 -- -- 2.32E+19 8.48E+19 Future(c) -- 54.00 -- -- 2.60E+l9 9.57E+I9 Future<c) -- 60.00 -- -- 2.87E+19 l.07E+20 Note(s):

(a) Cycle 28 was the current operating cycle at the time these neutron exposures were determined. Values listed for this cycle are projections based on the Cycle 28 design data.

(b) Values listed for this cycle are projections based on the Cycle 29 design data.

(c) Values beyond Cycle 29 are based on the average core power distributions and reactor operating conditions of Cycles 25-27, but include a 1.1 bias on the core thermal power.

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Westinghouse Non-Proprietary Class 3 2-14 Table 2-10 Calculated Iron Atom Displacement Rate and Iron Atom Displacements at the Surveillance Capsule Positions Cycle Cumulative Displacement Rate (dpa/s) Displacements (dpa)

Cycle Length Operating Time 40 40° 40 40° (EFPY) (EFPY)

I 1.27 1.27 3.19E-l 1 1.16E-10 l.27E-03 4.63E-03 2 0.78 2.04 3.71E-11 l.45E-10 2.18E-03 8.19E-03 3 0.70 2.75 3.55E-ll l.40E-10 2.97E-03 l.13E-02 4 0.73 3.48 3.45E-l 1 1.35E-10 3.77E-03 l.44E-02 5 0.74 4.22 3.47E-11 l.41E-10 4.58E-03 l.77E-02 6 0.72 4.95 3.48E-11 1.33E-10 5.38E-03 2.08E-02 7 0.73 5.67 3.49E-11 1.34E-10 6.18E-03 2.38E-02 8 1.12 6.80 3.44E-11 7.42E-11 7.40E-03 2.65E-02 9 1.19 7.98 3.33E-11 7.28E-11 8.65E-03 2.92E-02 10 1.19 9.17 2.94E-11 7.15E-11 9.75E-03 3.19E-02 11 1.14 10.32 2.92E-l 1 7.25E-11 1.08E-02 3.45E-02 12 1.19 11.51 2.79E-11 7.40E-11 1.19E-02 3.73E-02 13 1.18 12.68 2.44E-l 1 7.69E-l 1 1.28E-02 4.0lE-02 14 1.04 13.72 2.25E-11 7.15E-l 1 1.35E-02 4.25E-02 15 1.16 14.88 2.22E-11 9.86E-l l 1.43E-02 4.61E-02 16 0.35 15.23 2.03E-l 1 9.0lE-11 1.45E-02 4.71E-02 17 1.21 16.44 2.47E-11 8.57E-11 l.55E-02 5.03E-02 18 1.18 17.62 2.27E-11 l.04E-10 1.63E-02 5.42E-02 19 1.29 18.91 l.90E-11 8.83E-11 1.71E-02 5.78E-02 20 1.34 20.26 l.87E-11 8.31E-11 l.79E-02 6.14E-02 21 1.34 21.60 l.97E-11 8.82E-11 l.87E-02 6.51E-02 22 0.58 22.18 I.82E-l l 8.68E-1 l l.91E-02 6.67E-02 23 1.39 23.57 l.91E-l 1 7.09E-11 l.99E-02 6.98E-02 24 1.39 24.96 l.93E-11 9.20E-l 1 2.07E-02 7.38E-02 25 1.33 26.29 l.97E-11 8.54E-l l 2.16E-02 7.74E-02 26 1.23 27.52 2.07E-l 1 9.39E-l l 2.24E-02 8.1 lE-02 27 1.37 28.89 2.34E-l 1 8.74E-l l 2.34E-02 8.48E-02 23<*> 1.25 30.14 2.43E-l 1 9.08E-l 1 2.43E-02 8.84E-02 29(b) 1.37 31.51 2.42E-11 9.48E-11 2.54E-02 9.25E-02 Future<c) -- 36.00 -- -- 2.87E-02 1.06E-01 FutureCc) -- 42.00 -- -- 3.3 lE-02 l.25E-01 FutureCc) -- 48.00 -- -- 3.76E-02 1.43E-01 Future(c) -- 54.00 -- -- 4.20E-02 l.62E-01 FutureCc) -- 60.00 -- -- 4.64E-02 l.80E-01 Note(s):

(a) Cycle 28 was the current operating cycle at the time these neutron exposures were determined. Values listed for this cycle are projections based on the Cycle 28 design data.

(b) Values listed for this cycle are projections based on the Cycle 29 design data.

(c) Values beyond Cycle 29 are based on the average core power distributions and reactor operating conditions of Cycles 25-27, but include a I. I bias on the core thermal power.

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Westinghouse Non-Proprietary Class 3 2-15 Table 2-11 Calculated Surveillance Capsule Lead Factors Cycle Cumulative Lead Factor Cycle Length Operating Time 40 40° Capsule s<*l

<EFPY) <EFPY) 1 1.27 1.27 0.97 3.37 (Capsule T) 0.97 2 0.78 2.04 0.94 3.38 0.94 3 0.70 2.75 0.93 3.37 0.93 4 0.73 3.48 0.92 3.37 (Capsule X) 0.92 5 0.74 4.22 0.92 3.37 0.92 6 0.72 4.95 0.92 3.38 (Capsule Y) 0.92 7 0.73 5.67 0.92 3.38 0.92 8 1.12 6.80 0.99 3.37 0.99 9 1.19 7.98 1.05 3.37 1.05 10 1.19 9.17 1.08 3.37 (Capsule U) 1.08 11 1.14 10.32 I.I I 3.37 I.I I 12 1.19 11.51 1.13 3.38 1.13 13 1.18 12.68 1.13 3.38 1.13 14 1.04 13.72 1.13 3.38 1.13 15 1.16 14.88 1.10 3.38 1.30 16 0.35 15.23 1.09 3.39 1.35 17 1.21 16.44 1.09 3.38 1.48 18 1.18 17.62 1.07 3.38 1.62 19 1.29 18.91 1.05 3.38 1.73 20 1.34 20.26 1.03 3.39 1.82 21 1.34 21.60 1.02 3.38 1.91 22 0.58 22.18 1.01 3.38 1.95 23 1.39 23.57 1.01 3.38 1.90 24 1.39 24.96 0.99 3.38 1.84 25 1.33 26.29 0.99 3.38 1.79 26 1.23 27.52 0.98 3.38 1.74 27 1.37 28.89 0.97 3.38 1.71 28(b) 1.25 30.14 0.97 3.37 1.67 29(<) 1.37 31.51 0.97 3.37 1.64 Future(d) -- 36.00 0.95 3.36 1.53 Future<d) -- 42.00 0.93 3.35 1.42 Future(d) -- 48.00 0.92 3.35 1.35 Future<d) -- 54.00 0.91 3.34 1.29 Future<d) -- 60.00 0.90 3.34 1.24 Note(s):

(a) Capsule S was moved from the 4° location to the 40° location following the completion of Cycle 14. It was then moved back to the 4° location following the completion of Cycle 22.

(b) Cycle 28 was the current operating cycle at the time these lead factors were determined. Values listed for this cycle are projections based on the Cycle 28 design data.

(c) Values listed for this cycle are projections based on the Cycle 29 design data.

(d) Values beyond Cycle 29 are based on the average core power distributions and reactor operating conditions of Cycles 25-27, but include a I.I bias on the core thermal power.

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Westinghouse Non-Proprietary Class 3 2-16 Table 2-12 Projected Fast Neutron Fluence Rate (E > 1.0 MeV) at the Surveillance Capsule Positions (Future Operation)

Capsule Position Fluence Rate (n/cm 2-s) 40 1.45E+10 40° 5.78E+10 WCAP-18455-NP February 2020 Revision 1

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Westinghouse Non-Proprietary Class 3 2-17 Table 2-13 , Calculational Uncertainties Uncertainty Description Vessel Inner Capsule Radius PCA Comparisons 3% 3%

H. B. Robinson Comparisons 5% 5%

Analytical Sensitivity Studies 9% 11%

Additional Uncertainty for Factors not Explicitly Evaluated 5% 5%

Net Calculational Uncertainty 12% 13%

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Westinghouse Non-Proprietary Class 3 2-18 N

E u

~

0 0

279.9 349.9 R lcml Figure 2-1 Plan View of the Reactor Geometry at the Core Mid plane WCAP-18455-NP February 2020 Revision I

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Westinghou e Non-Proprietary Class 3 2-19 E(J co en N

I 0

tO .,.......,-...-----,---,--.,...

, -.-..----,- ' I '

10.0 70.0 140.0 209.9 279.9 3-49.9 R (cm ]

Figure 2-2 Section View of the Reactor Geometry - 0° Azimuth WCAP-18455-NP February 2020 Revision 1

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Westinghou e Non-Proprietary Class 3 2-20 Eu 0

~

N I

<O O>

N I

ci co +-...--.--r-,----,---,---r--,-....,..,...,......,.... j ..

7D.o 70.0 140.0 209.9 279.9 3-49.9 R

[cm)

Figure 2-3 Section View of the Reactor Geometry - 4° Azimuth WCAP-18455-NP February 2020 Revision 1

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Westinghouse Non-Proprietary Class 3 3-1 3 FRACTURE TOUGHNESS PROPERTIES The requirements for P-T limit curve development are specified in 10 CFR50,Appendix G [4]. The beltline region of the reactor vessel is defined as the following in 10 CFR 50, Appendix G:

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

The D.C. Cook Unit 1 beltline materials traditionally included the Intermediate Shell Plates, Lower Shell Plates, the Intermediate to Lower Shell Circumferential Weld, the Intermediate Shell Longitudinal Welds, and the Longitudinal Shell Welds. However, as described in NRC Regulatory Issue Summary (RIS) 2014-11 [10], any reactor vessel materials that are predicted to experience a neutron fluence exposure greater than 1.0 x 10 17 n/cm2 (E > 1.0 MeV) at the end of the licensed operating period should be considered in the development of P-T limit curves. The additional 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.

The extended beltline materials include the Upper Shell Plates, Upper Shell Longitudinal Welds, and Upper Shell to Lower Shell Circumferential Weld. As seen from Table 2-7 of this report, the fluence for the both inlet/outlet nozzle to upper shell welds are less than 1.0 x 10 17 n/cm 2 (E > 1.0 MeV) at 48 EFPY. Therefore, these materials do not need to be considered in the extended beltline.

A summary of the best-estimate copper (Cu) and nickel (Ni) contents, in units of weight percent (wt. %),

as well as the initial RTNOT values for the reactor vessel beltline, extended beltline, nozzle, and sister-plant materials are provided in Table 3-1 for D.C. Cook Unit 1. Table 3-2 provides the initial RTNoT values for the reactor vessel closure head and vessel flange materials for D.C. Cook Unit 1.

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Westinghouse Non-Proprietary Class 3 3-2 Table 3-1 Summary of the Best-Estimate Chemistry and Initial RTNDT Values for the D.C. Cook Unit 1 Reactor Vessel Materials Wt.% Wt.% RTNDT(u)(e)

Reactor Vessel Material Heat Number (OF)

Cu Ni Reactor Vessel Beltline MaterialsCa)

Intermediate Shell Plate B4406-1 C1260-2 0.12 0.52 5 Intermediate Shell Plate B4406-2 C3506-2 0.15 0.50 33 Intermediate Shell Plate B4406-3 C3506-1 0.15 0.49 40 Lower Shell Plate B4407-1 C3929-l 0.14 0.55 28 Lower Shell Plate B4407-2 C3932-1 0.12 0.59 -12 Lower Shell Plate B4407-3 C3929-2 0.14 0.50 38 Intermediate Shell Longitudinal Welds 13253/12008 -56(!)

0.21 0.873 2-442 A, B, and C (Linde 1092 flux, Lot# 3791)

Lower Shell Longitudinal Welds 13253/12008 -56(!)

0.21 0.873 3-442 A, B, and C (Linde 1092 flux, Lot # 3 791)

Intermediate Shell to Lower Shell 1P3571 0.287(d) 0.756(d) -56(!)

Circumferential Weld 9-442 (Linde 1092 flux, Lot# 3958) 13253 D.C. Cook Unit 1 Surveillance Weld (Linde 1092 flux, Lot# 3791) 0.27 0.74 ---

Reactor Vessel Extended Beltline Materials<hl Upper Shell Plate B4405-1 C3594-1 0.14 0.46 10 Upper Shell Plate B4405-2 C3594-2 0.14 0.45 26 Upper Shell Plate B4405-3 C3872-1 0.14 0.48 30 Upper Shell Longitudinal Welds 13253/12008 -56(!)

0.21 0.873 1-442 A, B, and C (Linde 1092 flux, Lot # 3791) (h)

Upper Shell to Intermediate Shell 20291 Circumferential Weld 8-442 0.216(i) 0_737(i) -56(!)

(Linde 1092 flux, Lot# 3833)

Reactor Vessel Nozzle Materials(hl Inlet Nozzle B-4403-1 ZV-3209-1 Note (g) 0.66 -31 Inlet Nozzle B-4403-2 ZV-3209-2 Note (g) 0.67 -7 Inlet Nozzle B-4403-3 123T394VA1 Note (g) 0.71 -72 Inlet Nozzle B-4403-4 123T394VA2 Note (g) 0.72 -99 Outlet Nozzle B4401-1 EV-9514 Note (g) 0.65 -77 Outlet Nozzle B4401-2 EV-9703 Note (g) 0.69 -61 Outlet Nozzle B4401-3 EV-9895 Note (g) 0.87 -91 Outlet Nozzle B4401-4 EV-9921 Note (g) 0.66 -69 Reactor Vessel Sister-Plant MaterialsCc) 1P3571 Kewaunee Surveillance Weld 0.219 0.724 ---

(Linde 1092 flux, Lot# 3958) 1P3571 Maine Yankee Surveillance Weld (Linde 1092 flux, Lot# 3958) 0.351 0.771 ---

Notes (continued on following page):

(a) All data taken from [12], unless noted. Heat numbers, flux type, and lot numbers are from [14], unless otherwise noted.

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Westinghouse Non-Proprietary Class 3 3-3 (b) All data taken from [13], unless noted. Heat numbers, flux type, and lot numbers are from [14], unless otherwise noted.

(c) Data taken from [31].

(d) Data are the industry average values from [31].

(e) The initial RTNDT values are based on measured data for all beltline and extended beltline materials, unless otherwise noted.

(t) This initial RTNDT value is a generic value from [19] and requires use of CH= 17.0°F.

(g) Generic wt.% Cu values for SA-508, Class 2 low-alloy steel are available in [15].

(h) All information for the inlet and outlet nozzle forgings is taken from certified material test report (CMTR), and is based on measured data, unless otherwise noted. The initial RTNDT values were determined using the CMTR data along with the methodology in BWRVIP-173-A [15].

(i) Best-estimate copper and nickel content based on [16].

Table 3-2 Initial RTNnT Values for the D.C.

Cook Unit 1 Reactor Vessel Closure Head and Vessel Flange Materials Unit 1 Initial RTNDT Reactor Vessel Material

{°F)

Replacement Closure Head -4o<*J Vessel Flange 28(b)

Notes:

(a) Data taken from [33].

(b) Data taken from [12].

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

The surveillance capsule plate material for D.C. Cook Unit 1 is from Intermediate Shell Plate B4406-3.

Since this material shares a Heat number (Heat # C3506) with Intermediate Shell Plate B4406-2, the surveillance plate data is also applicable to this reactor vessel plate. Per Appendix C, the surveillance data are deemed credible for D.C. Cook Unit 1; therefore, a reduced margin term will be utilized in the ART calculations contained in Section 7 for the D.C. Cook Unit 1 Intermediate Shell Plate.

It is noted that the D.C. Cook Unit 1 surveillance weld is not directly applicable to any of the D.C. Cook Unit 1 reactor vessel beltline or extended beltline weld seams. As described in [12], the D.C. Cook Unit 1 Surveillance Weld was fabricated from weld Heat# 13253, while the lower, intermediate, and upper shell longitudinal welds were fabricated with a tandem heat weld wire (Heat# 13253 / 12008). Thus, the D.C.

Cook Unit 1 surveillance weld data is not applicable to these welds and the data is not used. As a result, no Position 2.1 CF or credibility calculations are performed for this material.

For the Intermediate Shell to Lower Shell Circumferential Weld material (Heat # 1P3571), surveillance capsule data from the Kewaunee and Maine Yankee plants is used to calculate the Position 2.1 CF and credibility calculations.

Table 4-1 summarizes the surveillance data available for the D.C. Cook Unit 1 plate and weld materials that will be used in the calculation of the Position 2.1 chemistry factor values.

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Westinghouse Non-Proprietary Class 3 4-2 Table 4-1 D.C. Cook Unit 1 Surveillance Capsule Data Capsule Fluence<*> Measured 30 ft-lb Transition Material Capsule (x 10 19 n/cm2, E > 1.0 MeV) Temperature Shift(b) (°F)

T 0.273 60 Intermediate Shell X 0.850 90 Plate B4406-3 y 1.22 105 (Longitudinal) u 1.88 115 T 0.273 70 Intermediate Shell X 0.850 110 Plate B4406-3 y 1.22 115 (Transverse) u 1.88 115 V 0.586 175 Kewaunee R 1.76 235 Surveillance Weld(c) p 2.61 230 (Heat# 1P3571) s 3.67 250 T 5.62 271 W-263 0.567 222 Maine Yankee W-253 1.25 260 Surveillance Weld(d)

(Heat# 1P3571) A-25 1.76 270 A-35 7.13 345 Notes:

(a) Fluence values are from Section 2, unless otherwise noted.

(b) Measured dRTNDT values are from [12], unless otherwise noted.

(c) The capsule, 6.RTNDT and fluence values for Kewaunee surveillance weld are taken from [30].

(d) The capsule, 6.RTNDT and fluence values for Maine Yankee surveillance weld are taken from [31].

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Westinghouse Non-Proprietary Class 3 5-1 5 CHEMISTRY FACTORS The chemistry factors (CFs) were calculated using Regulatory Guide 1.99, Revision 2 [1], Positions 1.1 and 2.1. Position 1.1 CFs for each reactor vessel material are calculated using the best-estimate copper and nickel weight percent of the material and Tables 1 and 2 of[l]. The best-estimate copper and nickel weight percent values for the D.C. Cook Unit 1 reactor vessel materials are provided in Table 3-1 of this report.

The Position 2.1 CFs are calculated for the materials that have available surveillance program results. The calculation is performed using the method described in [1]. The D.C. Cook Unit 1 and sister-plant surveillance data are summarized in Section 4 of this report and will be utilized in the Position 2.1 CF calculations in this Section. Table 5-1 and Table 5-2 calculate the D.C. Cook Unit 1 Position 2.1 CFs.

Position 1.1 and Position 2.1 CFs are summarized in Table 5-3 for D.C. Cook Unit 1. Adjustment of the

~RTNDT values due to chemistry differences between the industry average and the surveillance material from sister-plants was required per [ 1]. The Position 1.1 CF for the intermediate shell to lower shell forging circumferential weld seams is 214°F, while the surveillance program weld Position 1.1 CF is 187.2°F for Kewaunee and 237.2°F for Maine Yankee. Therefore, the chemistry adjustment factor would be equal to 214 / 187.2 = 1.14 for Kewaunee and 214 / 237.2 = 0.90 for Maine Yankee. No temperature adjustments were undertaken since Kewaunee, Maine Yankee, and D.C. Cook Unit 1 have similar operating temperatures.

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Westinghouse Non-Proprietary Class 3 5-2 Table 5-1 D.C. Cook Unit 1 Reactor Vessel Intermediate Shell Plate B4406-3 Chemistry Factor

Calculation Using Surveillance Capsule Data Capsule Fluence<*l FFChl ARTNoicl FF*ARTNDT Material Capsule FF 2 (x 10 19 n/cm2, E > 1.0 MeV) (OF) (DF)

T 0.273 0.646 60 38.77 0.42 Intermediate Shell X 0.850 0.954 90 85.90 0.91 Plate B4406-3 y (Longitudinal) 1.22 1.055 105 110.82 1.11 u 1.88 1.173 115 134.88 1.38 T 0.273 0.646 70 45.23 0.42 Intermediate Shell X 0.850 0.954 110 104.99 0.91 Plate B4406-3 y 1.22 1.055 115 121.38 1.11 (Transverse) u 1.88 1.173 115 134.88 1.38 SUM: 776.84 7.64 CF B4406-3 = I:(FF * ~RTNm) + I:(FF2) = (776.84) + (7.64) = 101.7°F Notes:

(a) The capsule fluence values are from Table 4-1.

(b) FF= fluence factor= f(O.lS-O.lO*log(fll.

(c) 6RTNDT values are the measured 30 ft~lb shift values taken from [12].

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Westinghouse Non-Proprietary Class 3 5-3 Table 5-2 D.C. Cook Unit 1 Reactor Vessel Intermediate to Lower Shell Circumferential Weld Chemistry Factor Calculation Using Surveillance Capsule Data Measured Adjusted Capsule Fluence<*l FF*~TNDT Material Capsule FF<bl ~TNDT(a) ~TNoic) FF2 (x 10 19 n/cm2, E > 1.0 MeV) (OF) (OF)

(OF)

V 0.586 0.850 175 199.5 169.66 0.72 Kewaunee Surveillance R 1.76 1.155 235 267.9 309.52 1.33 Weld p 2.61 1.257 230 262.2 329.56 1.58 (Heat# 1P3571) s 3.67 1.337 250 285.0 381.12 1.79 T 5.62 1.425 271 308.9 440.14 2.03 W-263 0.567 0.841 222 199.8 168.08 0.71 Maine Yankee W-253 1.25 1.062 260 234.0 248.55 1.13 Surveillance Weld A-25 1.76 1.155 270 243.0 280.75 1.33 (Heat# 1P3571)

A-35 7.13 1.466 345 310.5 455.15 2.15 SUM: 2782.53 12.78 CF IP357I = :E(FF

LlRTNnT) -;- :E(FF2) = (2782.53)-;- (12.78) = 217.8°F Notes:

(a) The capsules fluence and L'l.RTNDT values are taken from Table 4-1.

(b) FF= fluence factor= fCO.is-O.IO*Iog(!J>.

(c) The surveillance weld L'l.RTNDT values have been adjusted using the ratio procedure of Reg. Guide 1.99, Rev. 2. For Kewaunee the L'l.RTNDr values have. been increased by a factor of 1.14, and for Maine Yankee the L'l.RTNDT values have been decreased by a factor of 0.90.

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Westinghouse Non-Proprietary Class 3 5-4 Table 5-3 Summary of D.C. Cook Unit 1 Position 1.1 and 2.1 Chemistry Factors Chemistry Factor (°F)

Reactor Vessel Material Heat Number Position 1.1 <*> Position 2.1 (b)

Reactor Vessel Beltline Materials Intermediate Shell Plate B4406- l Cl260-2 81.4 Intermediate Shell Plate B4406-2 C3506-2 104.5 101.7 Intermediate Shell Plate B4406-3 C3506-l 104.0 101.7 Lower Shell Plate B4407-l C3929-l 97.8 Lower Shell Plate B4407-2 C3932-l 82.8 Lower Shell Plate B4407-3 C3929-2 95.5 Intermediate Shell Longitudinal Welds 2-442 A, 13253/12008 208.7 B, and C Lower Shell Longitudinal Welds 3-442 A, B, 13253/12008 208.7 and C Intermediate Shell to Lower Shell 1P3571 214.0 217.8 Circumferential Weld 9-442 Reactor Vessel Extended Beltline Materials Upper Shell Plate B4405-l C3594-l 93.7 Upper Shell Plate B4405-2 C3594-2 93.3 Upper Shell Plate B4405-3 C3872-l 94.6 Upper Shell Longitudinal Welds 1-442 A, B, 13253/12008 208.7 and C Upper Shell to Intermediate Shell Circumferential 20291 188.4 Weld 8-442 Reactor Vessel Sister-Plant Materials Kewaunee Surveillance Weld 1P3571 187.2 '*

Maine Yankee Surveillance Weld 1P3571 237.2 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 factors were taken from Table 5-1 and Table 5-2 of this report. As discussed in Appendix C, the surveillance plate data is deemed credible. Also, as discussed in Appendix C, the surveillance weld data is deemed credible.

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Westinghouse Non-Proprietary Class 3 6-1 6 CRITERIA FOR ALLOWABLE PRESSURE-TEMPERATURE RELATIONSHIPS 6.1 OVERALLAPPROACH The ASME (American Society of Mechanical Engineers) 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, Kie, for the metal temperature at that time. K1c is obtained from the reference fracture toughness curve, defined in the 1998 Edition through the 2000 Addenda of Section XI, Appendix G of the ASME Code [3]. The K1c curve is given by the following equation:

Kie =33.2+20. 734*e[0.02(T-RTNDT )] (1)

where, K1c (ksi'1in.) reference stress intensity factor as a function of the metal temperature T and the metal reference nil-ductility temperature RTNOT This K1c curve is based on the lower bound of static critical K1 values measured as a function of temperature on specimens of SA-533 Grade B Class 1, SA-508-1, SA-508-2, and SA-508-3 steel.

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

(2)

where, Kim stress intensity factor caused by membrane (pressure) stress K1t stress intensity factor caused by the thermal gradients K1c reference stress intensity factor as a function of the metal temperature T and the metal reference nil-ductility temperature RTNOT C 2.0 for Level A and Level B service limits C = 1.5 for hydrostatic and leak test conditions during which the reactor core is not critical WCAP-18455-NP February 2020 Revision 1

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Westinghouse Non-Proprietary Class 3 6-2 For membrane tension, the corresponding K 1 for the postulated defect is:

Kim =Mmx(pRi/t) (3)

Axial Flaw Methodology For plates, forgings, and longitudinal welds, Mm for an inside axial surface flaw is given by:

Mm 1.85 for -Ji < 2, Mm 0.926 -Ji for 2 ~ -Ji ~ 3.464, Mm 3.21 for -Ji > 3.464 and, Mm for an outside axial surface flaw is given by:

Mm 1.77 for -Ji < 2, Mm 0.893 -Ji for 2 s; Ji s; 3.464, Mm 3.09 for -Ji > 3.464 Circumferential Flaw Methodology Similarly, for circumferential welds, Mm for an inside or an outside circumferential surface flaw is given by:

Mm 0.89 for -Ji < 2, Mm 0.443 -Ji for 2 s; Ji s; 3.464, Mm 1.53 for -Ji > 3.464 Where:

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

For bending stress, the corresponding K1 for the postulated axial or circumferential defect is:

K1b = Mb

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

The maximum K1 produced by radial thermal gradient for the postulated axial or circumferential inside surface defect of G-2120 is:

Ka= 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 K1t = 0.753xl0-3 x HU x t25 (6)

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Westinghouse Non-Proprietary Class 3 6-3 where HU is the heatup rate in °F/hr.

The through-wall temperature difference associated with the maximum thermal Ki 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, Figure G-2214-2 for the maximum thermal Ki.

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

(b) Alternatively, the Ki for radial thermal gradient can be calculated for any thennal stress distribution and at any specified time during cooldown for a 1/4T axial or circumferential inside surface defect using the relationship:

Ki1 = (l.0359Co + 0.6322Ci + 0.4753C2 + 0.3855CJ)

J;i; (7) or similarly, Ku during heatup for a l/4T outside axial or circumferential surface defect using the relationship:

Ku= (1.043Co + 0.630Ci + 0.481C2 + 0.401C3) *.j;; (8) where the coefficients Co, C1, C2 and C3 are determined from the thermal stress distribution at any specified time during the heatup or cooldown using the form:

o-(x) =Co+ Ci(x I a)+ C2(x I a) 2 + CJ(x I a)3 (9) and xis 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 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" [2] Section 2.6 (Equations 2.6.2-4 and 2.6.3-1). Finally, the reactor vessel metal temperature at the crack tip of a postulated flaw is determined based on the methodology contained in Section 2.6.l ofWCAP-14040-A, Revision 4 (Equation 2.6.1-1). This equation is solved utilizing values for thermal diffusivity of 0.518 ft2/hr at 70°F and 0.379 ft 2/hr at 550°P and a constant convective heat-transfer coefficient value of 7000 Btu/hr-ft2-0 F.

At any time during the heatup or cooldown transient, Kie 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 RTNnT, 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, Ku, for the reference flaw are computed. From Equation 2, the pressure stress intensity factors are obtained, and from these, the allowable pressures are calculated.

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Westinghouse Non-Proprietary Class 3 6-4 For the calculation of the allowable pressure versus coolant temperature dur:ing cooldown, the reference 1/4T flaw of Appendix G to Section XI of the ASME Code is assumed to exist at the inside of the vessel wall. During cooldown, the controlling location of the flaw is always at the inside of the vessel wall because the thermal gradients, which increase with increasing cooldown rates, produce tensile stresses at the inside surface that would tend to open (propagate) the existing flaw. Since an inside surface flaw has a higher tensile stress than an outside flaw and is subject to more neutron embrittlement than an outside surface flaw in the beltline region, postulation of outside flaw for cooldown conditions is unnecessary. Allowable P-T 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 L1T (temperature) across the vessel wall developed during cooldown results in a higher value of K1c at the 1/4T location for finite cooldown rates than for steady-state operation. Furthermore, if conditions exist so that the increase in K1c exceeds Ka, 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 P-T relationships are developed for steady-state conditions as well as finite heatup rate conditions assuming the presence of a 1/4T defect at the inside of the wall. The heatup results in compressive stresses at the inside surface that alleviate the tensile stresses produced by internal pressure. The metal temperature at the crack tip lags the coolant temperature; therefore, the K1c for the inside 1/4T flaw during heatup is lower than the Kie for the flaw during steady-state conditions at the same coolant temperature. During heatup, especially at the end of the transient, conditions may exist so that the effects of compressive thermal stresses and lower K1c values do not offset each other, and the P-T 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 P-T 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 thennal stresses at the outside are tensile and increase with increasing heatup rates, each heatup rate must be analyzed on an individual basis.

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Westinghouse Non-Proprietary Class 3 6-5 Following the generation of P-T curves for the steady-state and finite heatup rate situations, the final limit curves are produced by constructing a composite curve based on a point-by-point comparison of the steady-state and finite heatup rate data. At any given temperature, the allowable pressure is taken to be the least of the three values taken from the curves under consideration. The use of the composite curve is necessary to set conservative heatup limitations because it is possible for conditions to exist wherein, over the course of the heatup ramp, the controlling condition switches from the inside to the outside, and the pressure limit must at all times be based on analysis of the most critical criterion.

6.3 CLOSURE HEADNESSEL FLANGE REQUIREMENTS 10 CFR Part 50, Appendix G [4] addresses the metal temperature of the closure head flange and vessel flange regions. This rule states that the metal temperature of the closure head regions must exceed the material unirradiated RT NOT by at least 120°F for nonnal operation when the pressure exceeds 20 percent of the preservice hydrostatic test pressure, which is calculated to be 621 psig. The initial RTNOT values of the reactor vessel closure head and vessel flange are documented in Table 3-2. The limiting unirradiated RTNoT of28°F is associated with the vessel flange of the D.C. Cook Unit 1 reactor vessel, so the minimum allowable temperature of this region is 148°F at pressures greater than 621 psig (without margins for instrument uncertainties). This limit is shown in Figures 8-1 and 8-2.

6.4 BOLTUP TEMPERATURE REQUIREMENTS The minimum boltup temperature is the minimum allowable temperature at which the reactor vessel closure head bolts can be preloaded. It is determined by the highest reference temperature, RTNOT, in the closure flange region. This requirement is established in Appendix G to 10 CFR 50 [4]. Per the NRC-approved methodology in WCAP-14040-A, Revision 4 [2], the minimum boltup temperature should be 60°F or the limiting unirradiated RTNDT of the closure flange region, whichever is higher. Since the limiting unirradiated RTNoT of this region is below 60°F per Table 3-2, the minimum boltup temperature for the D.C. Cook Unit 1 reactor vessel is 60°F. This limit is shown in Figure 8-1.

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Westinghouse Non-Proprietary Class 3 7-1 7 CALCULATION OF ADJUSTED REFERENCE TEMPERATURE From Regulatory Guide 1.99, Revision 2 [l], the adjusted reference temperature (ART) for each material in the beltline region is given by the following expression:

ART= Initial RTNOT+ LiRTNOT+ Margin (10)

Initial RTNOT is the reference temperature for the unirradiated material as defined in Paragraph NB-23 31 of Section III of the ASME Boiler and Pressure Vessel Code [ 11]. If measured values of the initial RTNOT 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.

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

L'iRT NOT = CF

f (0.28 - 0.10 log f) (11)

To calculate LiRTNoT 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) -_ £surface

e (-0.24x) (12) where x inches (reactor vessel cylindrical shell beltline thickness is 8.5 inches) is the depth into the vessel wall measured from the vessel clad/base metal interface. The resultant fluence is then placed in Equation 11 to calculate the LiRT NOT 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" [2].

Table 7-1 contains the surface fluence values at 48 EFPY for D.C. Cook Unit 1. These values are used for the development of the P-T limit curves contained in this report. Table 7-1 also contains the 1/4T and 3/4T calculated fluence values and fluence factors (FFs ), per Regulatory Guide 1.99, Revision 2 [ 1]. The values in this table are used to calculate the 48 EFPY ART vaiues for the D.C. Cook Unit 1 reactor vessel materials.

Margin is calculated as M = 2 ~ a; + a~ . The standard deviation for the initial RTNOT margin term (cr 1) is 0°F when the initial RTNOT is a measured value. When a generic value is used, the cr 1 is obtained from the set of data used to establish the mean. The standard deviation for the LiRT NOT margin term, cr,.., is l 7°F for plates or forgings when surveillance data is not used or is non-credible, and 8.5°F (half the value) for plates or forgings when credible surveillance data is used. For welds, cr,.. is equal to 28°F when surveillance capsule data is not used or is non-credible, and is 14°F (half the value) when credible surveillance capsule data is used. Per [l], the value for cr,.. need not exceed 0.5 times the mean value of LiRTNoT-WCAP-18455-NP February 2020 Revision 1

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Westinghouse Non-Proprietaiy Class 3 7-2 Contained in Tables 7-2 and 7-3 are the 48 EFPY ART calculations at the 1/4T and 3/4T locations for generation of the D.C. Cook Unit 1 heatup and cooldown curves. The limiting ART values for D.C. Cook Unit 1 are summarized in Table 7-4.

Per Table 2-7, both the inlet and outlet nozzle forgings and welds for D.C. Cook Unit 1 have projected fluence values at the lowest extent of the nozzle welds that do not exceed the 1 x 10 17 n/cm 2 fluence threshold at 48 EFPY. Consistent with NRC RIS 2014-11 [ 1OJ, neutron radiation embrittlement need not be considered herein for either the inlets or outlet nozzle materials.

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Westinghouse Non-Proprietal)' Class 3 7-3 Table 7-1 Fluence Values and Fluence Factors for the Vessel Surface, 1/4T, and 3/4T Locations for the D.C. Cook Unit 1 Reactor Vessel Beltline and Extended Beltline Materials at 48 EFPY Surface Fluence, f (a) l/4T f 3/4Tf Surface l/4T 3/4T Reactor Vessel Region (x10 19 n/cm2, (x10 19 n/cm2, (x10 19 n/cm2, FF FF FF E> 1.0MeV) E > 1.0 MeV)

E > 1.0 MeV)

Reactor Vessel Beltline Materials Intermediate Shell Plate B4406-1 2.53 1.249 1.52 1.116 0.548 0.832 Intermediate Shell Plate B4406-2 2.53 1.249 1.52 1.116 0.548 0.832 Intermediate Shell Plate B4406-3 2.53 1.249 1.52 1.116 0.548 0.832 Lower Shell Plate B4407-1 2.53 1.249 1.52 1.116 0.548 0.832 Lower Shell Plate B4407-2 2.53 1.249 1.52 1.116 0.548 0.832 Lower Shell Plate B4407-3 2.53 1.249 1.52 1.116 0.548 0.832 Intermediate Shell Longitudinal Welds 1.64 1.136 0.985 0.996 0.355 0.714 Lower Shell Longitudinal Welds 1.66 1.140 0.997 0.999 0.359 0.718 Intermediate to Lower Shell 2.47 1.243 1.48 1.109 0.535 0.825 Circumferential Weld Reactor Vessel Extended Beltline Materials Upper Shell Plate B4405-l 0.0160 0.149 0.00961 0.107 0.00346 0.051 Upper Shell Plate B4405-2 0.0160 0.149 0.00961 0.107 0.00346 0.051 Upper Shell Plate B4405-3 0.0160 0.149 0.00961 0.107 0.00346 0.051 Upper Shell Longitudinal Welds o.0215<h> 0.180 0.0129 0.130 0.00466 0.064 Upper to Intermediate Shell 0.0215 0.180 0.0129 0.130 0.00466 0.064 Circumferential Weld Notes:

(a) 48 EFPY surface fluence values are documented in Table 2-7.

(b) The Upper Shell Longitudinal Welds fluence values are conservatively set equal to the Upper to Intermediate Shell Circumferential Weld.

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Westinghouse Non-Proprietary Class 3 7-4 Table 7-2 Adjusted Reference Temperature Evaluation for the D.C. Cook Unit 1 Reactor Vessel Beltline and Extended Beltline Materials through 48 EFPY at the 1/4T Location R.G. 1.99, Rev. 2 CF<a> 1/4T Fluence<h) 1/4T RTNDT(Ulc) ARTNDT O'( <l'd(d) Margin ART<*>

Reactor Vessel Material FF<b)

Position (OF) (x 10 19 n/cm2, E > 1.0 MeV) (OF) (OF) {°F) (OF) (OF) (OF)

Reactor Vessel Beltline Materials Intermediate Shell Plate B4406-1 1.1 81.4 1.52 1.116 5 90.8 0 17.0 34.0 129.8 Intermediate Shell Plate B4406-2 1.1 104.5 1.52 1.116 33 116.6 0 17.0 34.0 183.6 Intermediate Shell Plate B4406-2 Using Credible D.C. Cook Unit 1 Surveillance 2.1 101.7 1.52 1.116 33 113.5 0 8.5 17.0 163.5 Data Intermediate Shell Plate B4406-3 1.1 104.0 1.52 1.116 40 116.0 0 17.0 34.0 190.0 Intermediate Shell Plate B4406-3 Using Credible D.C. Cook Unit 1 Surveillance 2.1 101.7 1.52 1.116 40 113.5 0 8.5 ,17.0 170.5 Data Lower Shell Plate B4407-1 1.1 97.8 1.52 1.116 28 109.1 0 17.0 34.0 171.1 Lower Shell Plate B4407-2 1.1 82.8 1.52 1.116 -12 92.4 0 17.0 34.0 114.4 Lower Shell Plate B4407-3 1.1 95.5 1.52 1.116 38 106.6 0 17.0 34.0 178.6 Intermediate Shell Longitudinal Welds 1.1 208.7 0.985 0.996 -56 207.8 17 28.0 65.5 217.3 (Heat# 13253/12008)

Lower Shell Longitudinal Welds (Heat #

1.1 208.7 0.997 0.999 -56 208.5 17 28.0 65.5 218.0 13253/12008)

Intermediate to Lower Shell 1.1 214.0 1.48 1.109 -56 237.4 17 28.0 65.5 246.9

_Circumferential Weld (Heat# 1P3571) ___ ---------------------------- ---------- ------------------------------------------

Intermediate to Lower Shell Circumferential Weld Using Credible 2.1 217.8 1.48 1.109 -56 241.6 17 14.0 44.0 229.6 Kewaunee and Maine Yankee Surveillance Data (Heat# 1P3571)

Reactor Vessel Extended Beltline Materials Upper Shell Plate B4405-1 1.1 93.7 0.00961 0.107 10 10.0 0 5.0 10.0 3.0.0 Upper Shell Plate B4405-2 1.1 93.3 0.00961 0.107 26 10.0 0 5.0 10.0 45.9 Upper Shell Plate B4405-3 1.1 94.6 0.00961 0.107 30 10.1 0 5.0 10.1 50.2 Upper Shell Longitudinal Welds 1.1 208.7 0.0129 0.130 -56 27.1 17 13.6 43.5 14.7 (Heat# 13253/12008)

Upper to Intermediate Shell 1.1 188.4 0.0129 0.130 -56 24.5 17 12.3 41.9 10.4 Circumferential Weld (Heat # 20291)

Notes contained on the following page.

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Westinghouse Non-Proprietary Class 3 7-5 Notes:

(a) Values are taken from Table 5-3.

(b) Values are taken from Table 7-1.

(c) Values are taken from Table 3-1.

(d) Per Appendix C, both the intermediate shell plate material surveillance data and the weld Heat# IP3571 surveillance data are determined to be credible. Therefore, per the guidance of Regulatory Guide 1.99, Revision 2 [I], the base metal <rd= l 7°F for Position 1.1 and <rd= 8.5°F for Position 2.1 with credible surveillance data, and the weld metal <rd=

28°F for the Position 1.1 data and <rd= 14°F for Position 2.1 with credible surveillance data. However, <rd need not exceed O.S*ll.RTNDr per regulatory guidance in[!].

(e) ART values are calculated in accordance with[!].

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Westinghouse Non-Proprietary Class 3 7-6 Table 7-3 Adjusted Reference Temperature Evaluation for the D.C. Cook Unit 1 Reactor Vessel Beltline and Extended Beltline Materials through 48 EFPYat the 3/4T Location R.G. 1.99, Rev. 2 CF<*> 3/4T Fluence<h) 3/4T RTNDT(U)(c) ARTNDT O"I O"a(d) Margin ART<e>

Reactor Vessel Material (OF)

Position (OF) (x 10 19 n/cm2, E > 1.0 MeV) FF (OF) (OF) (OF) (OF) (OF)

Reactor Vessel Beltline Materials Intermediate Shell Plate B4406-1 1.1 81.4 0.548 0.832 5 67.7 0 17.0 34.0 106.7 Intermediate Shell Plate B4406-2 1.1 104.5 0.548 0.832 33 86.9 0 17.0 34.0 153.9 Intermediate Shell Plate B4406-2 Using Credible D.C. Cook Unit 1 Surveillance 2.1 101.7 0.548 0.832 33 84.6 0 8.5 17.0 134.6 Data Intermediate Shell Plate B4406-3 1.1 104.0 0.548 0.832 40 86.5 0 17.0 34.0 160.5 Intermediate Shell Plate B4406-3 Using Credible D.C. Cook Unit 1 Surveillance 2.1 101.7 0.548 0.832 40 84.6 0 8.5 17.0 141.6 Data Lower Shell Plate B4407-1 1.1 97.8 0.548 0.832 28 81.3 0 17.0 34.0 143.3 Lower Shell Plate B4407-2 1.1 82.8 0.548 0.832 -12 68.9 0 17.0 34.0 90.9 Lower Shell Plate B4407-3 1.1 95.5 0.548 0.832 38 79.4 0 17.0 34.0 151.4 Intermediate Shell Longitudinal Welds 1.1 208.7 0.355 0.714 -56 149.1 17 28.0 65.5 158.6

<Heat# 13253/12008)

Lower Shell Longitudinal Welds (Heat#

1.1 208.7 0.359 0.718 -56 149.7 17 28.0 65.5 159.3 13253/12008)

Intermediate to Lower Shell 1.1 214.0 0.535 0.825 -56 176.6 17 28.0 65.5 186.1

_Circumferential Weld_(Heat # 1P3571) *-- ---------------------------- -------------------------------------------- ----------------- ----------------- --------- --------- -------------- -------------

Intermediate to Lower Shell Circumferential Weld Using Credible 2.1 217.8 0.535 0.825 -56 179.7 17 14.0 44.0 167.8 Kewaunee and Maine Yankee Surveillance Data (Heat# 1P3571)

Reactor Vessel Extended Beltline Materials Upper Shell Plate B4405-1 1.1 93.7 0.00346 0.051 10 4.8 0 2.4 4.8 19.5 Upper Shell Plate B4405-2 1.1 93.3 0.00346 0.051 26 4.7 0 2.4 4.7 35.5 Upper Shell Plate B4405-3 1.1 94.6 0.00346 0.051 30 4.8 0 2.4 4.8 39.6 Upper Shell Longitudinal Welds (Heat#

1.1 208.7 0.00466 0.064 -56 13.3 17 6.6 36.5 -6.2 13253/12008)

Upper to Intermediate Shell 1.1 188.4 0.00466 0.064 -56 12.0 17 6.0 36.0 -8.0 Circumferential Weld (Heat # 20291)

Notes contained on the following page.

WCAP-18455-NP February 2020 Revision 1

This record was final approved on 2/14/2020 11 :56:36 AM. (This statement was added by the PRIME system upon its validation)
~
Westinghouse Non-Proprietary Class 3 7-7 Notes:
(a) Values are taken from Table 5-3.
(b) Values are taken from Table 7-1.
(c) Values are taken from Table 3-1.
(d) Per Appendix C, both the intermediate shell plate material surveillance data and the weld Heat# IP3571 surveillance data are determined to be credible. Therefore, per the guidance of Regulatory Guide 1.99, Revision 2 [1], the base metal cr,-. = 17°F for Position 1.1 and cr,-. = 8.5°F for Position 2.1 with credible surveillance data, and the weld metal cr,-. =
28°F for the Position 1.1 data and cr,-. = 14°F for Position 2.1 with credible surveillance data. However, cr,-. need not exceed 0.5*6.RTNDT per regulatory guidance in [l].
(e) ART values are calculated in accordance with [l].
WCAP-18455-NP February 2020 Revision 1
This record was final approved on 2/14/2020 11 :56:36 AM. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 7-8 Table 7-4 Limiting ART Values for D.C. Cook Unit 1 at 48 EFPY<a>
Limiting Limiting 1/4T ART 3/4T ART Limiting Material Value {°F) Value {°F)
"Axial Flaw" Method 218.0 159.3 Lower Shell Longitudinal Welds 3-442 A, B, and C "Circumferential Flaw" 229.6(b) Intermediate to Lower Shell Circumferential Weld 167.8(bl Method Seam 9-442 using Credible Position 2.1 Data Notes:
(a) Values are the limiting values from Tables 7-2 and 7-3. For the materials listed in Table 7-2 through Table 7-3 circumferential flaws are considered in the circumferential weld materials and axial flaws are considered in all other materials.
(b) For the Intermediate to Lower Shell Weld materials, the ART values calculated using Position 2.1 CFs are used instead of the Position 1.1 results because credible surveillance data is available.
WCAP-18455-NP February 2020 Revision 1
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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 [2].
The highest ART values for D.C. Cook Unit 1 correspond to the Intermediate to Lower Shell Forging Circumferential Weld (Heat# 1P3571). However, since this material is a "Circumferential Flaw" material, the applied membrane (pressure) stress and resulting stress intensity factor at the postulated flaw location are much lower than for the most limiting "Axial Flaw" material. Consequently, this material does not produce the most limiting P-T limit curves. The most limiting P-T limit curves for D.C. Cook Unit 1 are produced by using the "Axial Flaw" methodology and the limiting "Axial Flaw" material ART values.
Thus, the limiting ART values for D.C. Cook Unit 1 used in the generation of the P-T limit curves are based on the Lower Shell Longitudinal Welds (Position 1:1) from Table 7-4. For P-T limit curve development, the limiting ART values are conservatively rounded up to the nearest whole number and increased by 3°F as shown below in Table 8-1. This additional margin is added to account for potential future increases to D.C. Cook Unit 1 fluence projections.
Table 8-1 ART Values To Be Used In P-T Limit Curves Development for D.C. Cook Unit 1 at 48 EFPY<al Limiting 1/4T Limiting 3/4T Limiting Material ART Value (°F) ART Value (°F)
Lower Shell Longitudinal Welds 3-442 A, B, and C 221 163 Note:
(a) Values correspond to the limiting "Axial Flaw" Method ART values in Table 7-4 rounded up to the nearest whole number and have an additional 3°F of margin added.
Figure 8-1 presents the limiting heatup curves without margins for possible instrumentation errors using a heatup rate of 60°F/hr applicable for 48 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 48 EFPY, with the flange requirements and using the "Axial Flaw" methodology. The heatup and cooldown curves were generated using the 1998 through the 2000 AddendaASME Code Section XI, Appendix G. As discussed in previously, the use of the "Axial Flaw" methodology and the limiting "Axial Flaw" ART values produce the most limiting P-T limit curves for D.C. Cook Unit 1.
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 P-T 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 2485 psig inservice hydrostatic test as required by Appendix G to 10 CPR WCAP-18455-NP February 2020 Revision 1
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Westinghouse Non-Proprietary Class 3 8-2 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 Krm < K1c (13)
where, Kim is the stress intensity factor covered by membrane (pressure) stress [see page 6-2, Equation (3)],
K1c = 33.2 + 20.734 e [o.oz (T-RTNDTll [see page 6-1 Equation (1)],
T is the minimum permissible metal temperature, and RTNDT is the metal reference nil-ductility temperature.
The criticality limit curve specifies P-T limits for core operation in order to provide additional margin during actual power production. The P-T limits for core operation (except for low power physics tests) are that: 1) the reactor vessel must be at a temperature equal to or higher than the minimum temperature required for the inservice hydrostatic test, and 2) the reactor vessel must be at least 40°F higher than the minimum permissible temperature in the corresponding P-T 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 D.C. Cook Unit 1 reactor vessel at 48 EFPY is 281 °F; this temperature value is calculated based on Equation (13). The vertical line drawn from these points on the P-T curve, intersecting a curve 40°F higher than the pressure-temperature limit curve, constitutes the limit for core operation for the reactor vessel.
Figures 8-1 and 8-2 define all of the above limits for ensuring prevention of non-ductile failure for the D.C.
Cook Unit 1 reactor vessel for 48 EFPY 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 8-2, 8-3, and 8-4. Vacuum refill limits for the Reactor Coolant System (RCS) are included in Figures 8-1 and 8-2.
Nozzle P-T limit curves have previously been developed for D.C. Cook Unit I in [13] and compared to the D.C. Cook Unit 1 32 EFPY beltline P-T limit curves from [12]. The 32 EFPY beltline P-T limit curves were shown to be bounding. These nozzle P-T limit curves from [13] remain applicable through 48 EFPY, because the projected nozzle forging fluence is less than 1 x 10 17 n/cm2 at 48 EFPY and therefore embrittlement effects need not be considered consistent with RIS 2014-11 [ 1OJ. Since the 48 EFPY beltline P-T limit curves developed herein are based on higher ART values and produce more limiting pressure-temperature combinations than the 32 EFPY curves from [12], the curves developed herein are also more limiting than the nozzle P-T limit curves in [13]. Therefore, the issue raised by RIS 2014-11 concerning the stresses associated with the geometry of the inlet and outlet nozzles have been addressed and it is concluded that the nozzle P-T limits remain non-bounding compared to the beltline P-T limits developed herein.
WCAP-18455-NP February 2020 Revision 1
This record was final approved on 2/14/2020 11 :56:36 AM. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 8-3 MATERIAL PROPERTY BASIS LIMITING MATERIAL: D.C. Cook Unit 1 Lower Shell Longitudinal Welds 3-442 A, B, and C using Regulatory Guide 1.99 Position 1.1 data LIMITING ART VALUES AT 48 EFPY: I/4T, 221 °F (Axial Flaw) 3/4T, 163°F (Axial Flaw) 2500 I IOperlimAnalysis Version:5.4 2250 I I ILeakTestllmlt~
I U ,
I '. Run:12992 Operlim.xlsm Version:
54
_ _ ,.,__I_ *_*1--+------+----+----1 I
2000 - I!unacceptable!
Operation !, _
__,__ _.,___ j 11*1 1750 C!> 1500
-+-------+-----+---+----+---+----*
-+-----+--,I Heatup Rate I G0°F/Hr I rN Critical Limitl G0°F/Hr
-*---+----+I_ __,__ _ _ _ _ ,
U) a..
f
, 1250 -t----;-------;-----1------+---*- - -
~ '
_L_______,___,_,_,.~----~---1
"'~
a..
"g 1000
~
-J------t------t----l---f-,-1/ v __ *. LL. . . I Operation I 1,--+----r-*
~
(.)
iij 750 -t-----+---1----+.-~~--l---1--l---r-~-~-~-~---,_ I I u V Criticality Limit based on inservice hydrostatic test
.---'!""""'--~ temperature (281°F) for the service period up to 48 EFPY _
500 -+--------~
~ I~~ 1Te~~*:1 250 +---t-l--+----t------t---t---l-t---+----t-----1---t----l 0
~f-.-_J RCS Vacuum V~
I l -14.1 psig r
-250 0 50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Deg. F)
Figure 8-1 D.C. Cook Unit 1 Reactor Coolant System Heatup Limitations (Heatup Rate of 60°F/hr) Applicable for 48 EFPY (with,Flange Requirements and without Margins for Instrumentation Errors) using the 1998 through the 2000 Addenda App. G Methodology (w/ K1c)
WCAP-18455-NP February 2020 Revision 1
This record was final approved on 2/14/2020 11 :56:36 AM. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 8-4 MATERIAL PROPERTY BASIS LIMITING MATERIAL: D.C. Cook Unit 1 Lower Shell Longitudinal Welds 3-442 A, B, and C using Regulatory Guide 1. 99 Position 1.1 data LIMITING ART VALUES AT 48 EFPY: l/4T, 221 °F (Axial Flaw) 3/4T, 163°F (Axial Flaw) 2500 .. - . -*
I IOperlimAnalysis Version:5.4 Run:12992 i Operlim.xlsm Version: 5.4.1 I 2250 i
I I L 2000 IUnacceptable I I Operation I 1750 . I I J
(!)
CJ) 1500 I c..
Q)
I 1250 V *I Acceptable I
1--
/ l__
II) Operation II)
Q) c..
"C 1000 .
.... I I
(
Q)
I I
(.J iii 750 ~
Cooldown Rate, ~
{)
("F/H,~ - ~
0 - ~
20 I 500 -.40-~ ~
I I
V 100 - -
250 .
0 ..._______
N RCS Vacuum -14.7 psig
-250 I ' I I 0 50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Deg. F}
Figure 8-2 D.C. Cook Unit 1 Reactor Coolant System Cooldown Limitations (Cooldown Rates of O, -20, -40, -60, and -100°F/hr) Applicable for 48 EFPY (with Flange Requirements and without Margins for Instrumentation Errors) using the 1998 through the 2000 Addenda App. G Methodology (w/ K1c)
WCAP-18455-NP February 2020 Revision 1
This record was final approved on 2/14/2020 11 :56:36 AM. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 8-5 Table 8-2 D.C. Cook Unit 1 48 EFPY Heatup Curve Data Points using the 1998 through the 2000 Addenda App. G Methodology (w/ K1c, w/ Flange Requirements, and w/o Margins for Instrumentation Errors) 60°F/hr Heatup 60°F/hr Criticality T (°F) P (psig) T (°F) P (psig) 60 -14.7 281 -14.7 60 595 281 1170 65 595 285 1211 70 595 290 1271 75 595 295 1323 80 595 300 1381 85 595 305 1445 90 595 310 1516 95 595 315 1594 100 596 320 1680 105 598 325 1775 110 601 330 1879 115 605 335 1995 120 611 340 2122 125 618 345 2263 130 621 350 2418 135 621 - -
140 621 - -
145 621 - -
148 621 - -
148 669 - -
150 675 - -
155 691 - -
160 709 - -
165 725 - -
170 738 - -
175 753 - -
180 768 - -
185 786 - -
190 805 - -
195 826 - -
200 850 - -
205 876 - -
210 905 - -
215 936 - -
WCAP-18455-NP February 2020 Revision 1
This record was final approved on 2/14/2020 11 :56:36 AM. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 8-6 60°F/hr Heatup 60°F/hr Criticality T {°F) P (psig) T {°F) P (psig) 220 971 - -
225 1010 - -
230 1053 - -
235 1101 - -
240 1153 - -
245 1211 - -
250 1271 - -
255 1323 - -
260 1381 - -
265 1445 - -
270 1516 - -
275 1594 - -
280 1680 - -
285 1775 - -
290 1879 - -
295 1995 - -
300 2122 - -
305 2263 - -
310 2418 - -
Table 8-3 D.C. Cook Unit 1 48 EFPY Leak Test Curve Data Points using the 1998 through the 2000 Addenda App. G Methodology (w/ K1c, w/ Flange Requirements, and w/o Margins for Instrumentation Errors)
Leak Test Limits T {°F) P (psig) 264 2000 281 2485 WCAP-18455-NP February 2020 Revision 1
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Westinghouse Non-Proprietary Class 3 8-7 Table 8-4 D.C. Cook Unit 1 48 EFPY Cooldown Curve Data Points using the 1998 through the 2000 Addenda App. G Methodology for Steady-state (0°F/hr), -20°F/hr, -40°F/hr, -60°F/hr, and -100°F/hr (w/ K1c, w/ Flange Requirements, and w/o Margins for Instrumentation Errors)
Steady-State -20°F/hr Cooldown -40°F/hr Cooldown -60°F/hr Cooldown -100°F/hr Cooldown T (°F) P (psig) T {°F) P (psig) T (°F) P (psig) T {°F) P (psig) T {°F) P (psig) 60 -14.7 60 -14.7 60 -14.7 60 -14.7 60 -14.7 60 618 60 569 60 520 60 469 60 364 65 619 65 571 65 521 65 470 65 365 70 621 70 572 70 523 70 472 70 367 75 621 75 574 75 525 75 474 75 369 80 621 80 576 80 527 80 476 80 371 85 621 85 579 85 529 85 478 85 374 90 621 90 581 90 532 90 481 90 377 95 621 95 584 95 535 95 484 95 380 100 621 100 588 100 538 100 488 100 384 105 621 105 591 105 542 105 492 105 389 110 621 110 595 110 546 110 497 110 394 115 621 115 600 115 551 115 502 115 400 120 621 120 605 120 557 120 507 120 407 125 621 125 611 125 563 125 514 125 414 130 621 130 617 130 569 130 521 130 423 135 621 135 621 135 577 135 529 135 432 140 621 140 621 140 585 140 538 140 443 145 621 145 621 145 594 145 548 145 455 148 621 148 621 150 604 150 559 150 468 148 690 148 645 155 615 155 571 155 483 150 694 150 649 160 628 160 585 160 500 155 703 155 659 165 642 165 600 165 519 160 714 160 671 170 657 170 617 170 539 165 725 165 683 175 674 175 636 175 562 170 738 170 697 180 693 180 656 180 588 175 753 175 713 185 714 185 680 185 616 WCAP-18455-NP February 2020 Revision 1
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Westinghouse Non-Proprietary Class 3 8-8 Steady-State -20°F/hr Cooldown -40°F/hr Cooldown -60°F/hr Cooldown -100°F/hr Cooldown T (°F) P (psig) T (°F) P (psig) T {°F) P (psig) T (°F) P (psig) T (°F) P (psig) 180 768 180 730 190 737 190 705 190 648 185 786 185 749 195 763 195 734 195 683 190 805 190 770 200 791 200 765 200 722 195 826 195 794 205 823 205 800 205 765 200 850 200 820 210 858 210 839 210 814 205 876 205 848 215 897 215 882 215 867 210 905 210 880 220 940 220 930 220 926 215 936 215 915 225 987 225 982 225 982 220 971 220 954 230 1040 230 1040 230 1040 225 1010 225 997 235 1096 235 1096 235 1096 230 1053 230 1044 240 1153 240 1153 240 1153 235 1101 235 1096 245 1211 245 1211 245 1211 240 1153 240 1153 250 1275 250 1275 250 1275 245 1211 245 1211 255 1345 255 1345 255 1345 250 1275 250 1275 260 1424 260 1424 260 1424 255 1345 255 1345 265 1510 265 1510 265 1510 260 1424 260 1424 270 1605 270 1605 270 1605 265 1510 265 1510 275 1711 275 1711 275 1711 270 1605 270 1605 280 1827 280 1827 280 1827 275 1711 -, 275 1711 285 1956 285 1956 285 1956 280 1827 280 1827 290 2099 290 2099 290 2099 285 1956 285 1956 295 2256 295 2256 295 2256 290 2099 290 2099 300 2430 300 2430 300 2430 295 2256 295 2256 - - - - - -
300 2430 300 2430 - - - - - -
WCAP-18455-NP February 2020 Revision 1
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Westinghouse Non-Proprietary Class 3 9-1 9 REFERENCES
1. U.S. Nuclear Regulatory Commission, Office ofNuclear Regulatory Research, Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials," May 1988. [Agencywide Documents Access and Management System (ADAMS) Accession Number ML003740284}
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 Regulations, 10 CFR 50, Appendix G, "Fracture Toughness Requirements," Federal Register, December 12, 2013.
5. Regulatory Guide 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence," U.S. Nuclear Regulatory Commission, March 2001.
6. Westinghouse Report WCAP-18124-NP-A, Revision 0, "Fluence Determination with RAPTOR-M3G and FERRET," July 2018.
7. RSICC Data Library Collection DLC-185, "BUGLE-96: Coupled 47 Neutron, 20 Gamma-Ray Group Cross-Sectioll' Library Derived from ENDF/B-VI for LWR Shielding and Pressure Vessel Dosimetry Applications," July 1999.
8. ORNL Report ORNL/TM-13205, "Pool Critical Assembly Pressure Vessel Facility Benchmark,"
(NUREG/CR-6454), July 1997.
9. ORNL Report ORNL/TM-13204, "H. B. Robinson-2 Pressure Vessel Benchmark," (NUREG/CR-6453), October 1997 (published February 1998).
10. NRC Regulatory Issue Summary (RIS) 2014-11, "Information on Licensing Applications for Fracture Toughness Requirements for Ferritic Reactor Coolant Pressure Boundary Components," U.S. Nuclear Regulatory Commission, October 2014. [ADAMS Accession Number ML14149A165]
11. ASME Boiler and Pressure Vessel (B&PV) Code,Section III, Division 1, Subsection NB, "Class 1 Components."
12. Westinghouse Report, WCAP-15878, Revision 0, "D.C. Cook Unit 1 Heatup and Cooldown Limit Curves For Normal Operation For 40 Years And 60 Years," December 2002.
13. Westinghouse Letter, MCOE-LTR-15-82, Revision 0, "D.C. Cook Units 1 and 2 Pressure-Temperature Limits Amendment Request: NRC Request for Additional Information and License Condition Response," September 2, 2015.
14. AEP Document, DIT-B-03394-00, "Reactor Vessel Integrity Data Request for the SPU Conceptual Design Phase," April 15, 2010.
15. EPRI Document, BWRVIP-173-A: BWR Vessel and Internals Project, "Evaluation of Chemistry Data for BWR Vessel Nozzle Forging Materials," July 2011.
16. Combustion Engineering Owners Group Report, CE NPSD-1119, Revision 1, "Updated Analysis for Combustion Engineering Fabricated Reactor Vessel Welds Best Estimate Copper and Nickel Content,"
July 1998.
17. ASTM E 185-82, "Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels E706 (IF)," American Society for Testing and Materials, 1982.
WCAP-18455-NP February 2020 Revision 1
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Westinghouse Non-Proprietary Class 3 9-2
18. K. Wichman, M. Mitchell, and A. Hiser, NRC, Generic Letter 92-01 and RPV Integrity Assessment Workshop Handouts, "NRC/Industry Workshop on RPV Integrity Issues," February 12, 1998. [ADAMS Accession Number MLJ 10070570}
19. Code of Federal Regulations 10 CFR 50.61, "Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events," Federal Register, January 4, 2010.
20. Southwest Research Institute (SWRI) Project 02-4770, "Reactor Vessel Material Surveillance Program for Donald C. Cook Unit No. 1 Analysis of Capsule T," December 1977.
21. Southwest Research Institute (SWRI) Project 02-6159, "Reactor Vessel Material Surveillance Program for Donald C. Cook Unit No. 1 Analysis of Capsule X," June 1981.
22. Southwest Research Institute (SWRI) Project 06-7244-001, "Reactor Vessel Material Surveillance Program for Donald C. Cook Unit No. I Analysis of Capsule Y," January 1984.
23. Westinghouse Report WCAP-12483, Revision I, "Analysis of Capsule U from the American Electric Power Company D.C. Cook Unit 1 Reactor Vessel Radiation Surveillance Program," December 2002.
24. A. Schmittroth, FERRET Data Analysis Core, HEDL-TME 79-40, Hanford Engineering Development Laboratory, Richland, WA, September 1979.
25. RSICC Data Library Collection DLC-178, SNLRML Recommended Dosimetry Cross-Section Compendium, July 1994.
26. ASTM Standard El018-09, Application of ASTM Evaluated Cross-Section Data File, Matrix E706 (JIB), American Society for Testing and Materials, 2013.
27. ASTM Standard E944-13, Application of Neutron Spectrum Adjustment Methods in Reactor Surveillance, E 706 (JJJA), American Society for Testing and Materials, 2013.
28. Chart of the Nuclides, "Nuclides and Isotopes," 171h Edition, Lockheed Martin, 2009.
29. ASTM Standard EI005-16, Standard Test Method for Application and Analysis of Radiometric Monitors for Reactor Vessel Surveillance, American Society for Testing and Materials, 2016.
30. Westinghouse Report, WCAP-16641-NP, Revision 0, "Analysis of Capsule T from Dominion Energy Kewaunee Power Station Reactor Vessel Radiation Surveillance Program," October 2006.
31. Westinghouse Report, WCAP-15074, Revision 1, "Evaluations of the 1P3571 Weld Metal from the Surveillance Programs for Kewaunee and Maine Yankee," August 2006.
32. NUREG/CR-6413, ORNL/TM-13133, "Analysis of the Irradiation Data for A302B and A533B Correlation Monitor Materials," April 1996.
33. AEP Letter, DIT-B-03771-00, "Transmittal of Design Inputs in support of the P-T Limit Curve Project," March 14, 2019.
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Westinghouse Non-Proprietary Class 3 A-1 APPENDIX A THERMAL STRESS INTENSITY FACTORS (Kn)
Tables A-1 and A-2 contain the thermal stress intensity factors (K1t) and vessel temperatures for the maximum heatup and cooldown rates at 48 EFPY for D.C. Cook Unit 1. The reactor vessel cylindrical shell radii to the 1/4T and 3/4T locations are as follows:
1/4T Radius= 88.845 inches
3/4T Radius= 93.095 inches WCAP-18455-NP February 2029
Revision 1
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Westinghouse Non-Proprietary Class 3 A-2 TableA-1 K1t and Vessel Temperature Values for D.C. Cook Unit 1 at 48 EFPY 60°F/hr Heatup Curves (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) 60°F/hr Heatup (°F) (ksi -./in.) 60°F/hr Heatup (°F) (ksi -./in.)
60 56.443 -1.093 55.138 0.598 65 59.668 -2.491 55.806 1.642 70 63.128 -3.514 57.259 2.493 75 66.891 -4.424 59.374 3.200 80 70.899 -5.119 62.007 3.769 85 75.050 -5.719 65.080 4.238 90 79.389 -6.188 68.501 4.620 95 83.811 -6.594 72.207 4.935 100 88.364 -6.915 76.139 5.195 105 92.967 -7.195 80.254 5.413 110 97.658 -7.419 84.516 5.593 115 102.379 -7.617 88.897 5.747 120 107.159 -7.777 93.377 5.876 125 111.958 -7.921 97.934 5.987 130 116.797 -8.038 102.556 6.082 135 121.648 -8.147 107.230 6.165 140 126.525 -8.237 111.946 6.237 145 131.411 -8.322 116.696 6.301 150 136.313 -8.393 121.475 6.359 155 141.222 -8.462 126.276 6.411 160 146.140 -8.522 131.095 6.458 165 151.064 -8.580 135.929 6.502 170 155.994 -8.632 140.776 6.543 175 160.927 -8.684 145.633 6.582 180 165.864 -8.730 150.498 6.618 185 170.805 -8.777 155.369 6.654 190 175.747 -8.820 160.246 6.687 195 180.691 -8.864 165.127 6.720 200 185.636 -8.905 170.012 6.752 205 190.584 -8.947 174.899 6.783 210 195.531 -8.986 179.789 6.814 WCAP-18455-NP February 2020 Revision 1
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Westinghouse Non-Proprietary Class 3 A-3 Table A-2 K11 and Vessel Temperature Values for D.C. Cook Unit 1 at 48 EFPY -100°F/hr Cooldown Curves (w/o Margins for Instrument Errors)
Water Vessel Temperature at 1/4T 100°F/hr Cooldown Temp. Location for 100°F/hr 1/4T Thermal Stress (OF) Cooldown (°F) Intensity Factor (ksi .../in.)
210 236.226 16.489 205 231.142 16.422 200 226.058 16.356 195 220.974 16.289 190 215.889 16.222 185 210.805 16.155 180 205.720 16.088 175 200.635 16.021 170 195.550 15.954 165 190.465 15.887 160 185.380 15.820 155 180.294 15.753 150 175.209 15.686 145 170.124 15.618 140 165.039 15.552 135 159.954 15.485 130 154.869 15.418 125 149.783 15.351 120 144.699 15.285 115 139.614 15.218 110 134.529 15.152 105 129.444 15.085 100 124.360 15.019 95 119.275 14.953 90 114.191 14.888 85 109.107 14.822 80 104.023 14.756 75 98.939 14.690 70 93.855 14.625 65 88.771 14.560 60 83.689 14.494 WCAP-18455-NP February 2020 Revision 1
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Westinghouse Non-Proprietary Class 3 B-1 APPENDIXB OTHER RCPB FERRITIC COMPONENTS 10 CPR Part 50, Appendix G [4] requires that all Reactor Coolant Pressure Boundary (RCPB) components meet the requirements of Section III of the ASME Code. The lowest service temperature (LST) requirement for all RCPB components, which is specified in NB-2332(b) and NB-3211 of the ASME Code,Section III [11], is the relevant requirement that would affect the P-T limits. This requirement is applicable to ferritic materials outside of the RV with a nominal wall thickness greater than 2 Yi inches, such as piping, pumps and valves [11].
The D.C. Cook Unit 1 reactor coolant system does not have ferritic materials in the Class 1 piping, pumps, and valves (fabricated instead with stainless steel). Therefore, the LST requirements of the ASME Code,Section III, NB-2332(b) and NB-3211 [11] for these components do not need to be considered.
RIS 2014-11 [ 1OJ also addresses other ferritic components of the reactor coolant system relative to P-T limit, and states the following:
As specified in Sections I and IV.A of 10 CFR Part 50, Appendix G, ferritic RCPB components outside of the reactor vessel must meet the applicable requirements of ASME Code,Section III, "Rules for Construction ofNuclear Facility Components."
The other ferritic RCPB components that are not part of the RV beltline or extended beltline for D.C. Cook Unit 1 consist of the RV closure head, steam generators, and pressurizer. The D.C. Cook Unit 1 primary system components are analyzed to the following ASME Code Section III Editions and met all applicable requirements at the time of construction. Therefore, no further consideration of these components is necessary.
Reactor Vessel Closure Head - ASME Code Section III 1995 Edition through the 1996 Addenda.
Steam Generator - Portions of the original steam generators were replaced. The replacement steam generator components consist of the lower replacement steam generator subassembly (RSGSA),
replacement steam drum internals, and replacement feedring, and the re-used original components consist of the steam drum pressure boundary. Their designs are as follows:
o Unit 1 Original Steam Generator Components - ASME Code Section III 1965 Edition through Winter 1966 Addenda o Unit 1 Replacement Steam Generator Components - ASME Code Section III 1989 Edition
Pressurizer - ASME Code Section III 1965 Edition through Winter 1966 Addenda WCAP-18455-NP February 2020 Revision 1
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Westinghouse Non-Proprietary Class 3 C-1 APPENDIXC D.C. COOK UNIT 1 SURVEILLANCE PROGRAM CREDIBILITY EVALUATION Regulatory Guide 1.99, Revision 2 [1] describes general procedures acceptable to the NRC staff for calculating the effects of neutron radiation embrittlement of the low-alloy steels currently used for light-water-cooled reactor vessels. Position 2.1 of [1], describes the method for calculating the adjusted reference temperature of reactor vessel beltline materials using surveillance capsule data. The methods of Position 2.1 can only be applied when two or more credible surveillance data sets become available from the reactor in question.
To date there have been four surveillance capsules removed and tested from the D.C. Cook Unit 1 reactor vessel. To use the surveillance data, the data must be shown to be credible. In accordance with [l], the credibility of the surveillance data will be judged based on five criteria.
The purpose of this evaluation is to apply the credibility requirements of [1], to the D.C. Cook Unit 1 reactor vessel surveillance data, including fluence values updated in Section 2, to determine if the surveillance data is credible.
Additionally, sister-plant surveillance data relevant to the D.C. Cook Unit 1 Intennediate to Lower Shell Circumferential Weld (Heat# 1P3571) is evaluated for credibility in this Appendix. However, since the Heat # 1P3571 evaluation contains only sister-plant data which has previously been deemed credible in
[30] and [31], only Criterion 3 must be analyzed for this material. Each of the other criterion would be unaffected by use of the surveillance data for D.C. Cook Unit 1 as shown herein.
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Westinghouse Non-Proprietary Class 3 C-2 C.1 D.C. COOK UNIT 1 CREDIBILITY EVALUATION Criterion 1: Materials in the capsules should be those judged most likely to be controlling with regard to radiation embrittlement.
The beltline region of the reactor vessel is defined in Appendix G to 10 CFR Part 50, "Fracture Toughness Requirements" [4], as follows:
"the region of the reactor vessel (shell material including welds, heat affected zones, and plates or forgings) that directly surrounds the effective height of the active core and adjacent regions of the reactor vessel that are predicted to experience sufficient neutron radiation damage to be considered in the selection of the most limiting material with regard to radiation damage."
At the time of the design of the D.C. Cook Unit 1 surveillance program, the D.C. Cook Unit 1 reactor vessel beltline region was considered to consist of the following materials:
1. Intennediate Shell Plates B4406-1, B4406-2, and B4406-3
2. Lower Shell Plates B4407-l, B4407-2, and B4407-3
3. lntennediate Shell Axial Welds 2-442A, 2-442B, and 2-442C (Weld Wire Heat #
13253/12008, Linde 1092 Flux Type, Flux Lot# 3791)
4. Lower Shell Axial Welds 3-442A, 3-442B, and 3-442C (Weld Wire Heat# 13253/12008, Linde 1092 Flux Type, Flux Lot# 3791)
5. Intermediate Shell Plates to Lower Shell Plates Circumferential Weld Seam 9-442 (Weld Wire Heat# 1P3571, Linde 1092 Flux Type, Flux Lot# 3958)
The D.C. Cook Unit 1 surveillance program utilizes longitudinal and transverse test specimens from the Intermediate Shell Plate B4406-3. This plate was chosen due to the fact it has the highest initial RTNDT, as well as a higher copper content. It is noted that this plate is applicable to both Intermediate Shell Plate B4406-2 and B4406-3.
The D.C. Cook Unit 1 surveillance weld was fabricated with weld wire Heat # 13253. Since the Intermediate and Lower Shell longitudinal welds were fabricated with a tandem heat weld wire (13253/12008), the surveillance weld data from D.C. Cook Unit 1 is not directly applicable to the D.C.
Cook Unit 1 reactor vessel and no further credibility evaluations are perfonned for this material. However, the Intermediate Shell to Lower Shell Circumferential Weld Seam 9-442 was fabricated using weld wire Heat# 1P3571. Surveillance data for this heat number is available from Kewaunee and Maine Yankee plants, and is applicable to D.C. Cook Unit 1 (see Criterion 3).
Therefore, the materials selected for use in the D.C. Cook Unit 1 surveillance program were those judged to be most likely limiting with regard to radiation embrittlement according to the accepted methodology at the time the surveillance program was developed.
Based on the discussion above, Criterion 1 is met for the D.C. Cook Unit 1 surveillance program.
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Westinghouse Non-Proprietal)' Class 3 C-3 Criterion 2: Scatter in the plots of Charpy energy versus temperature for the irradiated and unirradiated conditions should be small enough to permit the determination of the 30 ft-lb temperature and upper-shelf energy unambiguously.
The credibility evaluation in [12] reviewed the plots of Charpy energy versus temperature for the unirradiated and irradiated conditions. This review concluded that, based on engineering judgment, the scatter in the data presented in these plots is small enough to permit the determination of the 30 ft-lb temperature and the upper-shelf energy of the D.C. Cook Unit 1 surveillance materials unambiguously.
Hence, the D.C. Cook Unit 1 surveillance program meets this criterion.
Criterion 3: When there are two or more sets of surveillance data from one reactor, the scatter of
~RTNDT values about a best-fit line drawn as described in Regulatory Position 2.1 nonnally should be less than 28°F for welds and l 7°F for base metal. Even if the fluence range is large (two or more orders of magnitude), the scatter should not exceed twice those values. Even if the data fail this criterion for use in shift calculations, they may be credible for determining decrease in upper-shelf energy if the upper shelf can be clearly determined, following the definition given in ASTM E185-82 [17].
The functional form of the least squares method as described in Regulatory Position 2.1 will be utilized to determine a best-fit line for this data and to determine if the scatter of these ~RTNDT values about this line is less than 28°F for the weld and less than l 7°F for the plate.
Following is the calculation of the best-fit line as described in Regulatory Position 2.1 of [1]. D.C. Cook Unit 1 has one circumferential weld that will be evaluated for credibility based on sister-plant data. This weld is Intermediate to Lower Shell Forging Circumferential Weld Seam and is fabricated from weld wire Heat# 1P3571. This weld metal heat is contained in both the Kewaunee and Maine Yankee surveillance programs. Since the weld in question utilizes data from other surveillance programs, the recommended NRC methods for determining credibility will be followed. The NRC methods were presented to industry at a meeting held by the NRC on February 12 and 13, 1998 [18]. At this meeting the NRC presented five cases. Of the five cases, Case 5 ("Surveillance Data from Other Sources Only") most closely represents the situation for the D.C. Cook Unit 1 surveillance weld metal. Note that for the plate material, the straightforward method in [1] will be followed, which is equivalent to Case 1 from [18] ("Surveillance Data Available from Plant but No Other Source").
Intermediate Shell Plate B4406-3 Following the NRC Case 4 guidelines, the D.C. Cook Unit 1 data will be evaluated. Table C-1 provides the calculation of the interim CF for D.C. Cook Unit 1. Note that when evaluating the credibility of the plate data, the measured ~RTNDT values for the plate metal do not include the adjustment ratio procedure of Regulatory Guide 1. 99, Revision 2, Position 2.1, since this calculation is based on the actual plate metal measured shift values. In addition, only D.C Cook Unit 1 data is being considered; therefore, no temperature adjustment is required.
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Westinghouse Non-Proprietary Class 3 C-4 Table C-1 Calculation oflnterim Chemistry Factors for the Credibility Evaluation Using D.C.
Cook Unit 1 Surveillance Data Capsule Fluence<a> ARTNDT(c) FF*ARTNDT Material Capsule FF(bl FF 2 (x 10 19 n/cm2, E > 1.0 MeV) {°F) (OF)
T 0.273 0.646 60 38.77 0.42 Intermediate Shell Plate B4406-3 X 0.850 0.954 90 85.90 0.91 (Longitudinal) y 1.22 1.055 105 110.82 1.11 u 1.88 1.173 115 134.88 1.38 T 0.273 0.646 70 45.23 0.42 Intermediate Shell Plate B4406-3 X 0.850 0.954 110 104.99 0.91 (Transverse) y 1.22 1.055 115 121.38 1.11 u 1.88 1.173 115 134.88 1.38 SUM: 776.84 7.64 CFs44o6-3 = L(FF * ~RTNor) --;- L(FF 2) = (776.84)--;- (7.64) = 101.7°F Notes:
(a) Taken from Table 4-1.
(b) FF = fl uence factor = ft 0 *28 - o. IO*log (f)).
(c) Measured values are 30 ft-lb ~RTNDT values from [12].
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Westinghouse Non-Proprietary Class 3 C-5 The scatter of L').RTNDT values about the functional form of a best-fit line drawn as described in Regulatory Position 2.1 is presented in Table C-2.
Table C-2 D.C. Cook Unit 1 Calculated Surveillance Capsule Data Scatter about the Best-Fit Line CF Capsule Measured(*> Predicted Scatter <17°F Material Capsule (Slope best-fit) Fluence FF ARTNDT ARTNDT ARTNDT(b) (Base Metal)
(OF) (x 10 19 n/cm2) (OF) (OF) (OF) <28°F (Weld)
T 0.273 0.646 60 65.7 5.7 Yes Intermediate Shell Plate X 0.850 0.954 90 97.1 7.1 Yes B4406-3 y 1.22 1.055 105 107.3 2.3 Yes (Longitudinal) u 1.88 1.173 115 119.3 4.3 Yes 101.7 Intermediate T 0.273 0.646 70 65.7 4.3 Yes Shell Plate X 0.850 0.954 110 97.1 12.9 Yes B4406-3 y 1.22 1.055 115 107.3 7.7 Yes (Transverse) u 1.88 1.173 115 119.3 4.3 Yes Notes:
(a) Measured values are 30 ft-lb L'iRTNDr values from [12].
(b) Scatter L'iRTNDr = Absolute Value [Predicted L'iRTNDr- Measured L'iRTNDr].
From a statistical point of view, +/- I cr would be expected to encompass 68% of the data. The scatter of L').RTNDT values about the best-fit line, drawn as described in [I], Position 2.1, should be less than I 7°F for base metal. Table C-2 indicates that eight of the eight surveillance data points fall inside the+/- I cr of l 7°F scatter band for surveillance base metals ( I 00% within the scatter band); therefore, the plate data is deemed "credible" per the third criterion.
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Westinghouse Non-Proprietary Class 3 C-6 Evaluation of Maine Yankee and Kewaunee Weld Data (Heat# 1P3571, Case 5)
Next, data from all sources is considered in order to evaluate the credibility of Heat# 1P3571 used in the D.C. Cook Unit 1 Intermediate to Lower Shell Circumferential Weld Seam using the NRC Case 5 guidelines. Data for the Kewaunee surveillance weld material and Maine Yankee surveillance weld material are available. Since data are from multiple sources, the data must be adjusted for chemical differences.
Adjustment for irradiation temperature differences is not completed consistent with [31], which concludes that the irradiation environment is similar for all capsules.
It is noted that Case 5 recommends first determining the credibility of one set of plant data separately; however, since [31] concluded that both the Kewaunee and Maine Yankee data are credible when treated separately, the data is not re-analyzed separately herein. The analysis herein considers all surveillance data together.
In accordance with the NRC Case 5 guidelines, the data from all sources should be adjusted to the mean chemical composition of all the data. This is performed as follows:
Kewaunee surveillance weld metal Cu Wt.%= 0.219, Ni Wt.%= 0.724, Position 1.1 CF= 187.2°F (from Table 3-1 and Table 5-3)
Maine Yankee surveillance weld metal Cu Wt.%= 0.351, Ni Wt.%= 0.771, Position 1.1 CF= 237.2°F (from Table 3-1 and Table 5-3)
Surveillance Data average composition (considering all available capsules)
Cu Wt.%= (5
0.219 + 4
0.351) I 9 = 0.278 Ni Wt.%= (5
0.724 + 4
0.771) / 9 = 0.745 Position 1.1 CF= 209.0°F (when using Cu Wt. % =0.278 and Ni Wt. % = 0.745)
The ratio procedure is then applied considering the industry average chemical composition. Therefore, the following ratios are applied to the LlRTNDT in Table C-3:
RatiOKewaunee = CFAverage/ CF Kewaunee Weld= 209.0 / 187.2 = 1.12 RatiOMaineYankee = CFAverage/ CF Maine Yankee Weld= 209.0 / 237.2 = 0.88 Table C-3 provides the summary of the weld interim CF considering all available data.
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Westinghouse Non-Proprietary Class 3 C-7 Table C-3 Calculation of Interim Weld Chemistry Factor for the Credibility Evaluation Using All Available Surveillance Data Measured Adjusted Capsule FluenceC*) FF*ARTNDT Material Capsule FF(b) .ARTNoi*> ARTNDT(c) FF 2 19 (x 10 n/cm2, E > 1.0 MeV) (OF)
(OF) (OF)
V 0.586 0.850 175 196 167 0.72 Kewaunee R 1.76 1.155 235 263 304 1.33 Surveillance Weld Material p 2.61 1.257 230 258 324 1.58 (Heat# 1P3571) s 3.67 1.337 250 280 374 1.79 T 5.62 1.425 271 304 432 2.03 Maine Yankee W-263 0.567 0.841 222 195 164 0.71 Surveillance Weld W-253 1.25 1.062 260 229 243 1.13 Material A-25 1.76 1.155 270 238 275 1.33 (Heat# 1P3571)
A-35 7.13 1.466 345 304 445 2.15 SUM: 2728.32 12.78 CF Surv. Weld= L(FF * ~RTNDT) 7 L(FF2 ) = (2728.32) 7 (12.78) = 213.6°F Notes:
(a) Taken from Table 4-1.
(b) FF= fluence factor= f 0-2s-O.IO*Iog(t))_
(c) Measured 8RTNDT values multiplied by the calculated ratios for the average composition compared to the Kewaunee (1.12) and Maine Yankee (0.88) compositions.
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Westinghouse Non-Proprietary Class 3 C-8 The scatter of LlRTNDT values about the functional form of a best-fit line drawn as described in Regulatory Position 2.1 is presented in Table C-4.
Table C-4 D.C. Cook Unit 1 Calculated Surveillance Weld Metal Data Scatter about the Best-Fit Line Using All Available Surveillance Data CF Capsule Measured(*> Predicted Scatter <17°F Material Capsule (Slope best-tit) Fluence FF ARTNDT ARTNDT ARTNDT(b) (Base Metal)
(OF) (x 10 19 n/cm 2) (OF) (°F) (°F) <28°F (Weld)
V 0.586 0.850 196 185.2 10.8 Yes Kewaunee R 1.76 1.155 263 251.6 11.6 Yes Surveillance p 2.61 1.257 258 273.8 16.2 Yes Weld Material s 3.67 1.337 280 291.3 11.3 Yes T 213.6 5.62 1.425 304 310.3 6.8 Yes W-263 0.567 0.841 195 183.2 12.1 Yes Maine Yankee W-253 1.25 1.062 229 231.3 2.5 Yes Surveillance Weld Material A-25 1.76 1.155 238 251.6 14.0 Yes A-35 7.13 1.466 304 319.3 15.7 Yes Notes:
(a) Measured values are 30 ft-lb L'lRTNDT values from [30] for Kewaunee and [31] for Maine Yankee.
(b) Scatter t.RTNDT = Absolute Value [Predicted t.RTNDT- Measured t.RTNDT].
The scatter of LlRTNDT values about the best-fit line, drawn as described in [l], Position 2.1, should be less than 28°F for weld metal. Table C-4 indicates that all nine surveillance data points (100%) fall within the
+/- 1cr of28°F scatter band for surveillance weld materials. Therefore, the weld material is deemed "credible" per the third criterion when all available data for the D.C. Cook Unit 1 weld is considered.
Hence, Criterion 3 is met for the D.C. Cook Unit 1 surveillance program materials.
Criterion 4: The irradiation temperature of the Charpy specimens in the capsule should match the vessel wall temperature at the cladding/base metal interface within+/- 25°F.
The D.C. Cook Unit 1 capsule specimens are located in the reactor between the core barrel and the vessel wall and are positioned opposite the center of the core. The test capsules are in guide tubes attached to the thermal shields. The location of the specimens with respect to the reactor vessel beltline provides assurance that the reactor vessel wall and the specimens experience equivalent operating conditions and will not differ by more than 25°F.
Hence, Criterion 4 is met for the D.C. Cook Unit 1 surveillance program.
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Westinghouse Non-Proprietary Class 3 C-9 Criterion 5: The surveillance data for the correlation monitor material in the capsule should fall within the scatter band of the database for that material.
The D.C. Cook Unit 1 surveillance program does contain A533B-1 correlation monitor material. Figure 11 of [32] contains a plot of residual versus fast fluence for the correlation monitor material. The data used in Figure 11 is contained in Table 14 of[32], identified as product SRM. The data found in the report contains the four surveillance capsules that have been removed and tested from D.C. Cook Unit 1; however, the fluence values have been updated. Table C-5 contains an updated calculation of the residual versus fast fluence.
Table C-5 Calculation of Residual versus Fast Fluence Capsule Fluence<a) Measured RG 1.99 Shift Residual Measured Shift Capsule FF (x 10 19 n/cm 2, E > 1.0 MeV) Shift<bl (CF*FF)Ccl (OF)
(OF) {°F)
T 0.273 0.646 60 82.7 -22.7 X 0.850 0.954 100 122.2 -22.2 y 1.22 1.055 110 135.1 -25.1 u 1.88 1.173 120 150.1 -30.1 Note:
(a) Taken from Table 4-1.
(b) Taken from [32].
(c) Per [32], the Cu and Ni values for the correlation monitor material are 0.170 Cu and 0.640 Ni. This equates to a CF of 128°F from[!].
Table C-5 shows a 2cr uncertainty of less than 50°F, which is the allowable scatter in Figure 11 of [32].
Hence, Criterion 5 is met for the D.C. Cook Unit 1 surveillance program.
Conclusion:
Based on the preceding responses to all fiv{l criteria of Regulatory Guide 1.99, Revision 2, Section B:
The D.C. Cook Unit 1 surveillance plate data are deemed "credible"
The D.C. Cook Unit 1 sister-plant surveillance weld data for Heat # 1P3571 are deemed "credible" WCAP-18455-NP February 2020 Revision 1
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Westinghouse Non-Proprietary Class 3 D-1 APPENDIXD VALIDATION OF THE RADIATION TRANSPORT MODELS BASED ON NEUTRON DOSIMETRY MEASUREMENTS D.1 NEUTRON DOSIMETRY Comparisons of measured dosimetry results to both the calculated and least-squares adjusted values for all surveillance capsules withdrawn from service to-date are described herein. The sensor sets from these capsules have been analyzed in accordance with the current dosimetry evaluation methodology described in Regulatory Guide 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence" [5]. One of the main purposes for presenting this material is to demonstrate that the overall measurements agree with the calculated and least-squares adjusted values to within +/- 20% as specified by Regulatory Guide 1.190, thus serving to validate the calculated neutron exposures previously reported in Section 2 of this report.
D.1.1 Sensor Reaction Rate Determinations In this section, the results of the evaluations of the in-vessel neutron sensor sets withdrawn and analyzed to-date as part of the reactor vessel materials surveillance program are presented.
Eight irradiation capsules attached to the thermal shield were included in the reactor design to constitute the reactor vessel surveillance program. The capsules were located at azimuthal angles of 4° (Capsule S),
176° (Capsule V), 184° (Capsule W), and 356° (Capsule Z) that are 4° from the core cardinal axes and 40° (Capsule T), 140° (Capsule U), 220° (Capsule X), and 320° (Capsule Y) that are 40° from the core cardinal axes. The irradiation history of each of these eight in-vessel surveillance capsules is summarized as follows:
Capsule Location Irradiation History T 40° Cycle 1 (withdrawn for analysis)
X 40° Cycles 1-4 (withdrawn for analysis) y 40° Cycles 1-6 (withdrawn for analysis) u 40° Cycles 1-10 (withdrawn for analysis) w 40 In the reactor s 40 In the reactor (moved to 40° location in 1995 and then moved back to its original 4°location in 2010)
V 40 In the reactor z 40 In the reactor The azimuthal locations included in the above tabulation represent the FOE azimuthal angle of the geometric center of the respective surveillance capsules.
The passive neutron sensors included in the evaluations of the surveillance capsules are summarized as follows:
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Westinghouse Non-Proprietary Class 3 D-2 Sensor Material Reaction Of Capsule T CapsuleX Capsule Y Capsule U Interest Copper 63 Cu(n,a) 6°Co X X X X Iron 54Fe(n,p)s4Mn X X X X Nickel 58Ni(n,p ) 58 Co X X X Uranium-238 mu(n,f)137Cs X X X X Neptunium-237 237Np(n,f) mes X X X X Cobalt-Aluminum* 59 Co(n,y)6°Co X X X X
The cobalt-aluminum measurements include both bare wire and cadmium-covered sensors.
Pertinent physical and nuclear characteristics of the in-vessel surveillance capsule passive neutron sensors are listed in Table D-1.
The use of passive monitors such as those listed above does not yield a direct measure of the energy-dependent neutron fluence rate at the point of interest. Rather, the activation or fission process is a measure of the integrated effect that the time- and energy-dependent neutron fluence rate has on the target material over the course of the irradiation period. An accurate assessment of the average neutron fluence rate incident on the various monitors may be derived from the activation measurements only if the irradiation parameters are well known. In particular, the following variables are of interest:
The measured specific activity of each monitor,
The physical characteristics of each monitor,
The operating history of the reactor,
The energy response of each monitor, and
The neutron energy spectrum at the monitor location.
Results from the radiometric counting of the neutron sensors from the in-vessel capsules are documented in [20-23], and re-evaluated in this appendix using the RAPTOR-M3G model described in Section 2. In all cases, the radiometric counting followed established ASTM procedures. Following sample preparation and weighing, the specific activity of each sensor was determined by means of a high-resolution gamma spectrometer. For the copper, iron, nickel, and cobalt-aluminum sensors, these analyses were performed by direct counting of each of the individual samples. In the case of the uranium and neptunium fission sensors, the analyses were carried out by direct counting preceded by dissolution and chemical separation of cesium from the sensor material.
The irradiation history of the reactor over the irradiation periods experienced by the in-vessel capsules was based on monthly power generation data from initial reactor criticality through the end of the dosimetry evaluation period. For the sensor sets utilized in the surveillance capsules, the half-lives of the product isotopes are long enough that a monthly histogram describing reactor operation has proven to be an adequate representation for use in radioactive decay corrections for the reactions of interest in the exposure evaluations. The startup and shutdown dates for each cycle of operation used in the evaluations are given in Table D-2.
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Westinghouse Non-Proprietary Class 3 D-3 Having the measured specific activities, the physical characteristics of the sensors, and the operating history of the reactor, reaction rates referenced to full-power operation were determined from the following equation:
R=~~~~~~~~~~-
A P*1 Cj[J-e-AtjJ re-Atd,j)
NoFY"'i..,-
Pref where:
R = Reaction rate averaged over the irradiation period and referenced to operation at a core power level of Pref (rps/nucleus).
A Measured specific activity (dps/g).
No Number of target element atoms per gram of sensor.
F Atom fraction of the target isotope in the target element.
y Number of product atoms produced per reaction.
Average core power level during irradiation period j (MW).
Pref Maximum or reference power level of the reactor (MW).
Calculated ratio of ~(E > 1.0 MeV) during irradiation period j to the time weighted average ~(E > 1.0 MeV) over the entire irradiation _period.
Decay constant of the product isotope (11sec).
Length of irradiation periodj (sec).
Decay time following irradiation periodj (sec).
and the summation is carried out over the total number of monthly intervals comprising the irradiation period.
In the equation describing the reaction rate calculation, the ratio [Pj]/[PrerJ accounts for month-by-month variations of reactor core power level within any given fuel cycle as well as over multiple fuel cycles. The ratio Cj, which was calculated for each fuel cycle using the transport methodology described in Section 2, accounts for the change in sensor reaction rates caused by variations in fluence rate induced by changes in core spatial power distributions from fuel cycle to fuel cycle. For a single-cycle irradiation, Cj is normally taken to be 1.0. However, for multiple-cycle irradiations, particularly those employing low-leakage fuel management, the additional Cj term should be employed. The impact of changing flux levels for constant WCAP-18455-NP February 2020 Revision 1
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Westinghouse Non-Proprietary Class 3 D-4 power operation can be quite significant for sensor sets .
that, have been irradiated for many cycles in a reactor that has transitioned from non-low-leakage to low-leakage fuel management or for sensor sets contained in surveillance capsules that have been moved from one capsule location to another. The fuel-cycle-specific neutron fluence rate values are used to compute cycle-dependent Cj values at the radial and azimuthal center of the respective capsules at the axial elevation of the active fuel midplane.
Prior to using the measured reaction rates in the least-squares evaluations of the dosimetry sensor sets, additional corrections were made to the 238 U measurements to account for the presence of 235 U impurities in the sensors as well as to adjust for the build-in of plutonium isotopes over the course of the irradiation.
Corrections were also made to the 238 U and 237Np sensor reaction rates to account for gamma-ray-induced fission reactions that occurred over the course of the capsule irradiations. The correction factors applied to the fission sensor reaction rates are summarized as follows:
Correction Capsule T CapsuleX CapsuleY Capsule U 235 U Impurity!Pu Build-in 0.874 0.852 0.837 0.813 238U(y,t) 0.957 0.957 0.957 0.957 Net 238U Correction 0.836 0.815 0.801 0.778 231Np(y,t) 0.984 0.984 0.984 0.984 These factors were applied in a multiplicative fashion to the decay-corrected uranium and neptunium fission sensor reaction rates.
Results of the sensor reaction rate determinations for the in-vessel capsul~s are given in Table D-3 through Table D-6.
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Westinghouse Non-Proprietary Class 3 D-5 D.1.2 Least-Squares Evaluation of Sensor Sets Least-squares adjustment methods provide the capability of combining the measurement data with the corresponding neutron transport calculations, resulting in a best-estimate neutron energy spectrum with associated uncertainties. Best estimates for key exposure parameters such as ~(E > 1.0 MeV) or dpa/s along with their uncertainties are then easily obtained from the adjusted spectrum. In general, the least-squares methods, as applied to surveillance capsule dosimetry evaluations, act to reconcile the measured sensor reaction rate data, dosimetry reaction cross sections, and the calculated neutron energy spectrum within their respective uncertainties. For example, Ri +/-oR, = L(crig +/-8.,,.)(q,g +/-o(j).)
g relates a set of measured reaction rates, Ri, to a single neutron spectrum, ~g, through the multigroup dosimeter reaction cross section, <Jig, each with an uncertainty o. The primary objective of the least-squares evaluation is to produce unbiased estimates of the neutron exposure parameters at the location of the measurement.
For the least-squares evaluation of the surveillance capsule dosimetry, the FERRET Code [24] was employed to combine the results of the plant-specific neutron transport calculations and sensor set reaction rate measurements to determine best-estimate values of exposure parameters (~(E > 1.0 MeV) and dpa) along with associated uncertainties for the in-vessel capsules analyzed to-date.
The application of the least-squares methodology requires the following input:
1. The calculated neutron energy spectrum and associated uncertainties at the measurement location.
2. The measured reaction rates and associated uncertainty for each sensor contained in the multiple foil set.
3. The energy-dependent dosimetry reaction cross sections and associated uncertainties for each sensor contained in the multiple foil sensor set.
For the plant-specific application of the least-squares methodology, the calculated neutron spectrum was obtained from the results of the neutron transport calculations described in Section 2 of this report. The sensor reaction rates were derived from the measured specific activities using the procedures described in Section D.1.1. The dosimetry reaction cross sections and uncertainties were obtained from the Sandia National Laboratories Radiation Metrology Laboratory (SNLRML) dosimetry cross-section library [25].
The SNLRML library is an evaluated dosimetry reaction cross-section compilation recommended for use in LWR evaluations by ASTM Standard El O18, "Application of ASTM Evaluated Cross-Section Data File, Matrix E706 (IIB)" [26].
The uncertainties associated with the measured reaction rates, dosimetry cross sections, and calculated neutron spectrum were input to the least-squares procedure in the form of variances and covariances. The assignment of the input uncertainties followed the guidance provided in ASTM Standard E944, "Application of Neutron Spectrum Adjustment Methods in Reactor Surveillance" [27].
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Westinghouse Non-Proprietary Class 3 D-6 The following provides a summary of the uncertainties associated with the least-squares evaluation of the surveillance capsule sensor sets withdrawn and analyzed to-date.
Reaction Rate Uncertainties The overall uncertainty associated with the measured reaction rates includes components due to the basic measurement process, irradiation history corrections, and corrections for competing reactions. A high level of accuracy in the reaction rate determinations is assured by utilizing laboratory procedures that conform to the ASTM National Consensus Standards for reaction rate determinations for each sensor type.
After combining all of these uncertainty components, the sensor reaction rates derived from the counting and data evaluation procedures were assigned the following net uncertainties for input to the least-squares evaluation:
Reaction Uncertainty 63 Cu(n,a)6°Co 5%
54Fe(n,p)54Mn 5%
ssNi(n,p )ssco 5%
238 U(n,f) 137 Cs 10%
mNp(n,f)'37Cs 10%
59Co(n,y)6oco 5%
These uncertainties are given at the 1cr level.
Dosimetry Cross-Section Uncertainties The reaction rate cross sections used in the least-squares evaluations were taken from the SNLRML library.
This data library provides reaction cross sections and associated uncertainties, including covariances, for 66 dosimetry sensors in common use. Both cross sections and uncertainties are provided in a fine multigroup structure for use in least-squares adjustment applications. These cross sections were compiled from the most recent cross-section evaluations, and they have been tested with respect to their accuracy and consistency for least-squares evaluations. Further, the library has been empirically tested for use in fission spectra detennination as well as in the fluence and energy characterization of 14 MeV neutron sources.
For sensors included in the plant-specific reactor vessel surveillance program, the following uncertainties in the fission spectrum averaged cross sections are provided in the SNLRML documentation package.
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Westinghouse Non-Proprietary Class 3 D-7 Reaction Uncertainty 63 Cu(n,a)6°Co 4.08-4.16%
54Fe(n,p)54Mn 3.05-3.11%
ssNi(n,p )ssco 4.49-4.56%
23sU(n,t)l37Cs 0.54-0.64%
237Np(n,t)l37Cs 10.32-10.97%
59 Co(n, y) 6°Co 0.79-3.59%
These tabulated ranges provide an indication of the dosimetry cross-section uncertainties associated with the sensor sets used in LWR irradiations.
Calculated Neutron Spectrum The neutron spectra input to the least-squares adjustment procedure were obtained directly from the results of plant-specific transport calculations for each surveillance capsule irradiation period and location. The spectrum for each capsule was input in an absolute sense (rather than as simply a relative spectral shape).
Therefore, within the constraints of the assigned uncertainties, the calculated data were treated equally with the measurements.
Using the uncertainties associated with the reaction rates obtained from the measurement procedures and counting benchmarks and the dosimetry cross-section uncertainties supplied directly with the SNLRML library, the uncertainty matrix for the calculated spectrum was constructed from the following relationship:
where Rn specifies an overall fractional normalization uncertainty and the fractional uncertainties Rg and Rg* specify additional random groupwise uncertainties that are correlated with a correlation matrix given by:
where (g- g')2 H=----
2y2 The first term in the correlation matrix equation specifies purely random uncertainties, while the second term describes the short-range correlations over a group range y (8 specifies the strength of the latter term).
The value of o is 1.0 when g = g', and is 0.0 otherwise.
The set of parameters defining the input c;ovariance matrix for the calculated spectra was as follows:
Flux Normalization Uncertainty (Rn) 15%
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Westinghouse Non-Proprietary Class 3 D-8 Flux Group Uncertainties (Rg, Rg*)
(E > 0.0055 MeV) 15%
(0.68 eV < E < 0.0055 MeV) 25%
(E < 0.68 eV) 50%
Short Range Correlation (8)
(E > 0.0055 MeV) 0.9 (0.68 eV < E < 0.0055 MeV) 0.5 (E < 0.68 eV) 0.5 Flux Group Correlation Range (y)
(E > 0.0055 MeV) 6 (0.68 eV < E < 0.0055 MeV) 3 (E < 0.68 eV) 2 WCAP-18455-NP February 2020 Revision I
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Westinghouse Non-Proprietary Class 3 D-9 D.1.3 Comparisons of Measurements and Calculations This section provides comparisons of the measurement results from each of the sensor set irradiations with corresponding analytical predictions at the measurement locations. These comparisons are provided on two levels. In the first level, calculations of individual sensor reaction rates are compared directly with the measured data from the counting laboratories. This level of comparison is not impacted by the least-squares evaluations of the sensor sets. In the second level, calculated values of neutron exposure rates in terms of fast neutron fluence rate ~ (E > 1.0 MeV) and iron atom displacement rate are compared with the best-estimate exposure rates obtained from the least-squares evaluation.
In Table D-7, comparisons of MIC ratios are listed for the threshold sensors contained in the in-vessel capsules. From Table D-7, it is noted that for the individual threshold sensors, the average MIC ratio ranges from 0.97 to 1.09 with an overall average of 1.02 and an associated standard deviation of 8.4%. In this case, the overall average was based on an equal weighting of each of the sensor types with no adjustments made to account for the spectral coverage of the individual sensors.
In Table D-8, best-estimate-to-calculation (BE/C) ratios for fast neutron fluence rate (E > 1.0 MeV) and iron atom displacement rate resulting from the least-squares evaluation of each dosimetry set. For the in-vessel capsules, the average BE/C ratio is seen to be 0.99 with an associated uncertainty of 6.8% for neutron fluence rate (E > 1.0 MeV) and 0.99 with an associated uncertainty of 6.1 % for the iron atom displacement rate.
The MIC comparisons based on individual sensor reactions without recourse to the least-squares adjustment procedure are summarized as follows:
In-Vessel Capsules Reaction Avg.MIC % Unc. (lo-)
63 Cu(n,a) 1.09 9.7%
54 Fe(n,p) 0.98 8.4%
58 Ni(n,p) 1.03 7.6%
238 U(Cd) (n,f) 0.97 0.7%
237 Np(Cd)(n,f) 1.05 6.7%
Linear Average 1.02 8.4%
A similar comparison for exposure rate expressed in terms of neutron fluence rate (E > 1.0 MeV) and iron atom displacement rate (dpa/s) are summarized as follows:
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Westinghouse Non-Proprietary Class 3 D-10 In-Vessel Capsules Parameter Avg.BE/C % Unc. (lo')
Fast Neutron Fluence Rate (E > 1.0 MeV) 0.99 6.8%
Iron Atom Displacement Rate (dpa/s) 0.99 6.1%
These data comparisons show similar and consistent results, with the linear average MIC ratio of 1.02 in good agreement with the resultant least-squares BE/C ratios of 0.99 for neutron fluence rate (E > 1.0 MeV) and 0.99 for iron atom displacement rate. The comparisons demonstrate that the calculated results provided in Section 2 of this report are validated within the context of the assigned 13% uncertainty and, further, show that the +/-20% (la) agreement between calculation and measurement required by [5] is met.
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Westinghouse Non-Proprietary Class 3 D-11 Table D-1 Nuclear Parameters Used in the Evaluation of the In-Vessel Surveillance Capsule Neutron Sensors Reaction Atomic Target Product Fission of Weight<*l Atom Half-lifeCbJ,(cJ,Cctl YieldCctl Interest fa/2:-atom) Fraction(b),(c) (days) (%)
63 Cu (n,a) 6°Co 63.546 0.6917 1925.28 n/a 54Fe (n,p) 54Mn 55.845 0.05845 312.13 n/a 58Ni (n,p) 58 Co 58.6934 0.68077 70.86 n/a 238 U (n,f) mes 238.051 1.00 10975.76 6.02 z31Np (n,f) mes 237.048 1.00 10975.76 6.27 59 eo (n;y) 60eo 58.933 o.0015(e) 1925.28 n/a Note(s):
(a) Atomic weight data were taken from the Chart of the Nuclides, 17'h Edition, dated 2010 [28].
(b) Half-life and target atom fraction data for 63 Cu (n,a), 54 Fe (n,p), and 58 Ni (n,p), reactions were taken from ASTM Standard EIOOS-16 [29].
(c) The half-life for the 59 Co (n,y) reaction was taken from ASTM Standard E 1005-16 [29]. The target atom fractions for the 59 Co (n,y), 238 U (n,f), and 237 Np (n,f) reactions are reflective of standard Westinghouse surveillance capsule dosimeter values.
(d) Half-life and fission yield data for the 238 U (n,f) and 237Np (n,f) reactions were taken from ASTM Standard EIOOS-16
[29].
(e) The 6°Co measured activities for Capsules T, X, and Y were reported in units ofBq/mg of Co in the Co-Al sensor (i.e.,
the target atom fraction of Co in the Co-Al alloy was divided out when these activities were reported). Therefore, the Co target atom fraction for these capsules was input as 1.0.
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Westinghouse Non-Proprietary Class 3 D-12 Table D-2 Startup and Shutdown Dates Cycle Startup Date Shutdown Date 1 01/18/1975 12/23/1976 2 02/01/1977 04/07/1978 3 06/24/1978 04/06/1979 4 07/17/1979 05/30/1980 5 08/05/1980 05/29/1981 6 08/03/1981 07/03/1982 7 09/28/1982 07/16/1983 8 10/23/1983 04/07/1985 9 11/15/1985 06/27/1987 10 10/05/1987 03/19/1989 WCAP-18455-NP February 2020 Revision 1
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I I
Westinghouse Non-Proprietary Class 3 D-13 Table D-3 Measured Sensor Activities and Reaction Rates for Surveillance Capsule T Radially Radially Corrected Average Measured Corrected Adjusted Average Sample Target Reaction Description Activity Saturated Reaction Reaction ID Isotope Rate (dps/g)<al Activity Rate Rate (rps/atom)
(dps/g) (rps/atom) (rps/atom) 1 63 Cu (n,a) 6°Co Top Mid 5.14E+04 3.26E+05 4.98E-17 2 63 Cu (n,a) 6°Co Mid 5.27E+04 3.35E+05 5.lOE-17 5.31E-17 5.31E-17 3 63 Cu (n,a) 6°Co Bot Mid 6.04E+04 3.83E+05 5.85E-17 4 . 54Fe (n,p) 54Mn Top Wire 1.93E+06 3.54E+06 5.62E-15 5 54Fe (n,p) 54Mn Top Mid l.69E+06 3.10E+06 4.92E-15 6 54Fe (n,p) 54Mn Mid 1.69E+06 3.10E+06 4.92E-15 5.13E-15 5.13E-15 7 54Fe (n,p) 54Mn Bot Mid 1.69E+06 3.10E+06 4.92E-15 8 54Fe (n,p) 54Mn Bot 1.80E+06 3.30E+06 5.24E-15 9 58 Ni (n,p) 58 Co Top Mid 3.83E+07 5.19E+07 7.43E-15 10 58Ni (n,p) 58 Co Mid 3.77E+07 5.11E+07 7.31E-15 7.47E-15 7.47E-15 11 58Ni (n,p) 58 Co Bot Mid 3.95E+07 5.35E+07 7.66E-15 12 238 U(Cd) (n,f) mes Mid l.20E+06 4.19E+07 2.75E-13 2.75E-13 2.30E-13 13 237Np(Cd) (n,f) mes Mid 4.53E+06 1.58E+08 9.92E-13 9.92E-13 9.77E-13 14 59 Co (n,y) 6°Co Top 4.87E+o9<bl 3.12E+10 3.06E-12 3.llE-12 3.1 lE-12 15 59Co (n,y) 6oco Bottom 5.03E+o9<bl 3.23E+10 3.16E-12 16 59 Co(Cd) (n,y) 6°Co Top (Cd) 1.83E+o9<bl 1.42E+10 1.39E-12 59 Co(Cd) (n,y) 6°Co 1.32E-12 1.32E-12 17 Bottom (Cd) 1.64E+o9<bl 1.27E+ 10 1.24E-12 Note(s):
(a) Measured activities are decay corrected to December 23, 1976.
(b) Measured activities were reported in units ofBq/mg of Co in the Co-Al sensor (i.e., the target atom fraction of Co in the Co-Al alloy was divided out when these activities were reported).
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Westinghouse Non-Proprietary Class 3 D-14 Table D-4 Measured Sensor Activities and Reaction Rates for Surveillance Capsule X Radially Radially Corrected Average Measured Corrected Adjusted Average Sample Target Reaction Description Activity Saturated Reaction Reaction ID Isotope Rate (dps/g)<*l Activity Rate Rate (rps/atom)
(dps/g) (rps/atom) (rps/atom) 1 63 Cu (n,a) 6°Co Top Mid l.14E+05 3.20E+05 4.88E-17 2 63 Cu (n,a) 6°Co Mid l.15E+05 3.23E+05 4.93E-17 4.96E-17 4.96E-17 3 63 Cu (n,a) 6°Co Bot Mid l.18E+05 3.31E+05 5.06E-17 4 54Fe (n,p) 54Mn Top Wire 2.21E+06 3.16E+06 5.02E-15 5 54Fe (n,p) 54Mn Top Mid 2.28E+06 3.26E+06 5.17E-15 6 54Fe (n,p) 54Mn Mid 2.23E+06 3.19E+06 5.06E-15 5.20E-15 5.20E-15 7 54Fe (n,p) 54Mn Bot Mid 2.34E+06 3.35E+06 5.3 lE-15 8 54Fe (n,p) 54Mn Bot 2.41E+06 3.45E+06 5.47E-15 9 ssNi (n,p) ssco Top Mid 3.98E+07 5.14E+07 7.36E-15 10 ssNi (n,p) ssco Mid 4.02E+07 5.19E+07 7.43E-15 7.SlE-15 7.SlE-15 11 ssNi (n,p) ssco Bot Mid 4.18E+07 5.40E+07 7.73E-15 238 12 U(Cd) (n,f) mes Mid 3.64E+05 4.79E+06 3.lSE-14 3.lSE-14 2.57E-14 13 237Np(Cd) (n,f) mes Mid 2.57E+06 3.38E+07 2.12E-13 2.12E-13 2.09E-13 14 59 Co (n,y) 6°Co Top 1.44E+ 1oCbl 4.09E+10 4.00E-12 3.94E-12 3.94E-12 15 59Co (n,y) 6oco Bottom 1.40E+ 1QCbl 3.97E+10 3.89E-12 16 59 Co(Cd) (n,y) 6°Co Top (Cd) 5.52E+Q9Cbl 1.89E+10 1.85E-12 59 Co(Cd) (n,y) 6°Co 1.85E-12 1.85E-12 17 Bottom (Cd) 5.53E+o9Cbl 1.90E+10 1.86E-12 Note(s):
(a) Measured activities are decay corrected to May 30, 1980.
(b) Measured activities were reported in units ofBq/mg of Co in the Co-Al sensor (i.e., the target atom fraction of Co in the Co-Al alloy was divided out when these activities were reported).
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Westinghouse Non-Proprietary Class 3. D-15 Table D-5 Measured Sensor Activities and Reaction Rates for Surveillance Capsule Y Radially Radially Corrected Average Measured Corrected Adjusted Average Sample Target Reaction Description Activity Saturated Reaction Reaction ID Isotope Rate (dps/g)<*l Activity Rate Rate (rps/atom)
(dps/g) (rps/atom) (rps/atom) 63 1 Cu (n,a) 6 °Co Top Mid 1.47E+05 3.26E+05 4.98E-17 63 6 2 Cu (n,a) °Co Mid 1.48E+05 3.29E+05 5.0lE-17 5.07E-17 5.07E-17 63 3 Cu (n,a) 6 °Co Bot Mid 1.54E+05 3.42E+05 5.22E-17 4 54 Fe (n,p) 54Mn Top Wire 2.32E+06 3.29E+06 5.22E-15 5 54Fe (n,p) 54Mn Top Mid 2.39E+06 3.39E+06 5.38E-15 6 54Fe (n,p) 54Mn Mid 2.41E+06 3.42E+06 5.42E-15 5.39E-15 5.39E-15 7 54Fe (n,p) 54Mn Bot Mid 2.46E+06 3.49E+06 5.53E-15 8 54 Fe (n,p) 54Mn Bot 2.40E+06 3.40E+06 5.40E-15 9 58 Ni (n,p) 58 Co Top Mid 3.80E+07 5.35E+07 7.66E-15 10 58Ni (n,p) 58 Co Mid 3.81E+07 5.37E+07 7.68E-15 7.82E-15 7.82E-15 11 58 Ni (n,p) 58 Co Bot Mid 4.03E+07 5.68E+07 8.12E-15 238 12 U(Cd) (n,f) mes Mid 5.26E+05 4.99E+06 3.28E-14 3.28E-14 2.62E-14 237 13 Np(Cd) (n,f) mes Mid 3.99E+06 3.78E+07 2.38E-13 2.38E-13 2.34E-13 14 59 Co (n,y) 6 °Co Top 1.69E+ 1QCbl 3.79E+10 3.71E-12 59 Co (n,y) 6 °Co 3.68E-12 3.68E-12 15 Bottom l .66E+ 1o<bl 3.72E+10 3.64E-12 16 59 Co(Cd) (n,y) 6°Co Top (Cd) 6.80E+o9<bl 1.84E+10 l.SOE-12 59 Co(Cd) (n,y) 6°Co 1.84E-12 1.84E-12 17 Bottom (Cd) 7.07E+o9<bl 1.92E+10 1.88E-12 Note(s):
(a) Measured activities are decay corrected to July 3, 1982.
(b) Measured activities were reported in units ofBq/mg of Co in the Co-Al sensor (i.e., the target atom fraction of Co in the Co-Al alloy was divided out when these activities were reported).
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Westinghouse Non-Proprietary Class 3 D-16 Table D-6 Measured Sensor Activities and Reaction Rates for Surveillance Capsule U Radially Radially Corrected Average Measured Corrected Adjusted Average Sample Target Reaction Description Activity Saturated Reaction Reaction ID Isotope Rate (dps/g)<*l Activity . Rate Rate (rps/atom)
(dps/g) (rps/atom) (rps/atom) 1 63 Cu (n,a) 6°Co Top Mid l.30E+05 2.73E+05 4.16E-17 2 63 Cu (n,a) 6°Co Mid l.26E+05 2.64E+05 4.03E-17 4.15E-17 4.15E-17 3 63 Cu (n,a) 60 Co Bot Mid l.33E+05 2.79E+05 4.26E-17 4 54Fe (n,p) 54Mn Top Wire 7.10E+05 2.47E+06 3.92E-15 5 54 Fe (n,p) 54Mn Top Mid 7.34E+05 2.55E+06 4.05E-15 6 54Fe (n,p) 54Mn Mid 7.31E+05 2.54E+06 4.03E-15 3.98E-15 3.98E-15 7 54Fe (n,p) 54Mn Bot Mid Not Recovered NIA NIA 8 54Fe (n,p) 54Mn Bot 7.10E+05 2.47E+06 3.92E-15 9 ssNi (n,p) ssco Top Mid 2.30E+06 4.09E+07 5.85E-15 10 58Ni (n,p) 58 Co Mid 2.27E+06 4.03E+07 5.77E-15 5.87E-15 5.87E-15 11 58Ni (n,p) 58 Co Bot Mid 2.36E+06 4.19E+07 6.00E-15 238 137 12 U(Cd) (n,f) Cs Mid 4.49E+05 2.57E+06 1.69E-14 1.69E-14 1.31E-14 13 237Np(Cd) (n,f) 137 Cs Mid 2.87E+06 l.64E+07 l.03E-13 l.03E-13 l.OlE-13 14 59 Co (n;y) 6°Co Top 2.06E+07 4.37E+07 2.85E-12 2.76E-12 2.76E-12 15 59Co (n,y) 6oco . Bottom l.93E+07 4.09E+07 2.67E-12 16 59 Co(Cd) (n,y) 6°Co Top (Cd) 8.66E+06 2.20E+07 l.44E-12 59 Co(Cd) (n,y) 6°Co l.38E-12 l.38E-12 17 Bottom (Cd) 8.01E+06 2.03E+07 l.33E-12 Note(s):
(a) Measured activities are decay corrected to October 10, 1989.
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Westinghouse Non-Proprietary Class 3 D-17 Table D-7 Comparison of Measured and Calculated Threshold Foil Reaction Rates for the In-Vessel Capsules Capsule Average Std. Dev.
Reaction T X y u 63 Cu (n,a) 60 eo 1.24 1.04 1.05 1.01 1.09 9.7%
54Fe (n,p) 54Mn 1.07 0.97 0.99 0.87 0.98 8.4%
58Ni (n,p) 58 eo 1.13 1.01 1.04 0.94 1.03 7.6%
238 U (n,f) mes __ (a) 0.96 0.97
__ (a) 0.97 0.7%
__ (a) __(a) 237Np (n,f) mes 1.00 1.10 1.05 6.7%
Average of MIC Results 1.02 8.4%
Note:
(a) The measurement for these sensors were rejected in final evaluation.
Table D-8 Comparison of Calculated and Best-Estimate Exposure Rates for the In-Vessel Capsules Fast (E > 1.0 Me V) Fluence Rate Iron Atom Displacement Rate Capsule BE/C Std. Dev. BE/C Std. Dev.
T 1.06 7.0% 1.05 8.0%
X 0.98 6.0% 0.98 7.0%
y 1.01 6.0% 1.02 7.0%
u 0.90 7.0% 0.91 9.0%
Average 0.99 6.8% 0.99 6.1%
WCAP-18455-NP February 2020 Revision 1
This record was final approved on 2/14/2020 11 :56:36 AM. (This statement was added by the PRIME system upon its validation)

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ML20108F000

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