ML22067A150

WCAP-18660-NP, Revision 0, Analysis of Capsule N from the Xcel Energy Prairie Island Unit 1 Reactor Vessel Radiation Surveillance Program, Section 6 Through Appendix B

ML22067A150
Person / Time
Site:	Prairie Island
Issue date:	11/30/2021
From:	Benson R, Long M, Mays B Westinghouse
To:	Office of Nuclear Reactor Regulation
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ML22067A147	List: L-PI-22-005, WCAP-18660-NP, Revision 0, Analysis of Capsule N from the Xcel Energy Prairie Island Unit 1 Reactor Vessel Radiation Surveillance Program, Appendix C, Part 5 Through Appendix D ML22067A148 ML22067A149 ML22067A150 ML22067A151 ML22067A152 ML22067A153 ML22067A154
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L-PI-22-005 WCAP-18660-NP, Rev 0
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6 RADIATION ANALYSIS AND NEUTRON DOSIMETRY

6.1 INTRODUCTION

This section describes a discrete ordinates (Sn) transport analysis performed for the Prairie Island Unit 1 reactor to determine the neutron radiation environment within the reactor pressure vessel and surveillance capsules. In this analysis, fast neutron exposure parameters in terms of fast neutron (E > 1.0 MeV) fluence and iron atom displacements (dpa) were established on a plant-and fuel-cycle-specific basis. An evaluation of the most recent dosimetry sensor set from Capsule N, withdrawn at the end of the 31st plant operating cycle, is provided. Comparisons of the results from the dosimetry evaluations with the analytical predictions served to validate the plant-specific neutron transport calculations. These validated calculations subsequently form the basis for projections of the neutron exposure of the reactor pressure vessel for operating periods extending to 60 effective full-power years (EFPY).

The use of fast neutron (E > 1.0 MeV) fluence to correlate measured material property changes to the neutron exposure of the material has traditionally been accepted for the development of damage trend curves as well as for the implementation of trend curve data to assess the condition of the vessel. However, it has been suggested that an exposure model that accounts for differences in neutron energy spectra between surveillance capsule locations and positions within the vessel wall could lead to an improvement in the uncertainties associated with damage trend curves and improved accuracy in the evaluation of damage gradients through the reactor vessel wall.

Because of this potential shift away from a threshold fluence toward an energy-dependent damage function for data correlation, ASTM E853-18, Standard Practice for Analysis and Interpretation of Light-Water Reactor Surveillance Neutron Exposure Results [17], recommends reporting displacements per iron atom along with fluence (E > 1.0 MeV) to provide a database for future reference. The energy-dependent dpa function to be used for this evaluation is specified in ASTM E693-94, Standard Practice for Characterizing Neutron Exposures in Iron and Low Alloy Steels in Terms of Displacements per Atom (DPA), E706 (ID)

[18]. The application of the dpa parameter to the assessment of embrittlement gradients through the thickness of the reactor vessel wall has been promulgated in Revision 2 to Regulatory Guide 1.99, Radiation Embrittlement of Reactor Vessel Materials [1].

All of the calculations and dosimetry evaluations described in this section and in Appendix A were based on nuclear cross-section data derived from ENDF/B -VI. Furthermore, the neutron transport and dosimetry evaluation methodologies follow the guidance of Regulatory Guide 1.190, Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence [ 19]. Additionally, the methods used to develop the calculated pressure vessel fluence are consistent with the NRC-approved methodology described in WCAP-18124-NP-A, Fluence Determination with RAPTOR -M3G and FERRET [20].

6.2 DISCRETE ORDINATES ANALYSIS

The arrangement of the surveillance capsules in the Prairie Island Unit 1 reactor vessel is shown in Figure 4-1. Six irradiation capsules attached to the thermal shield are included in the reactor design that constitutes the reactor vessel surveillance program. Capsules S, T, V, N, P, and R are located at azimuthal angles of 57°, 67°, 77°, 237°, 247°, and 257°, respectively. These full-core positions correspond to the following octant symmetric locations represented in Figure 6-1 and Figure 6-2: 13°, 23°, and 33° from the core

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cardinal axes. The stainless steel specimen containers are approximately 1-inch square in cross section and are approximately 63 inches in height. The containers are positioned axially such that the test specimens are centered on the core midplane, thus spanning the central five feet of the 12-foot high reactor core.

From a neutronic standpoint, the surveillance capsules and associated support structures are significant.

The presence of these materials has a significant effect on both the spatial distribution of neutron exposure rate and the neutron spectrum in the vicinity of the capsules. However, the capsules are far enough apart that they do not interfere with one another. In order to determine the neutron environment at the test specimen location, the capsules themselves must be included in the analytical model.

In performing the fast neutron exposure evaluations for the Prairie Island Unit 1 reactor vessel and surveillance capsules, plant-specific 3D forward transport calculations were carried out to directly solve for the space-and energy-dependent neutron exposure rate, (r,,z,E).

For the Prairie Island Unit 1 transport calculations, the model depicted in Figure 6-1 and Figure 6-2 was utilized. The model contained a representation of the reactor core, the reactor internals, the pressure vessel cladding and vessel wall, the insulation external to the pressure vessel, and the primary biological shield wall. This model formed the basis for the calculated results. In developing this analytical model, nominal design dimensions were generally employed for the various structural components. In addition, water temperatures, and hence, coolant densities in the reactor core and downcomer regions of the reactor were taken to be representative of full-power operating conditions. The coolant densities were treated on a fuel-cycle-specific basis. Table 6-10 contains the cycle-specific power levels, inlet coolant temperatures, and core average temperatures used in this analysis. The reactor core itself was treated as a homogeneous mixture of fuel, cladding, water, and miscellaneous core structures, such as fuel assembly grids, guide tubes, etc.

Section views of the model are shown in Figure 6-3 and Figure 6-4. The model extends radially from the centerline of the reactor core out to a location interior to the outer radius of the primary biological shield and over an axial span from an elevation more than five feet below the active fuel to more than five feet above the active fuel.

The RAPTOR-M3G model consisted of 186 radial mesh, 200 azimuthal mesh, and 435 axial mesh. Mesh sizes were chosen to assure that proper convergence of the inner iterations was achieved on a pointwise basis. The pointwise inner iteration flux convergence criterion utilized in the calculations was set at a value of 0.001.

The core power distributions used in the plant-specific transport analysis for the first 32 fuel cycles at Prairie Island Unit 1 included cycle-dependent fuel assembly initial enrichments, burnups, radial and axial power distributions. Actual operating characteristics through Cycle 32 have been evaluated; projections beyond Cycle 32 were based on Cycle 32 spatial power distributions, water temperatures, and reactor power level with a 10% bias on the peripheral and re-entrant corner relative powers as directed by Xcel Energy. Note that two projections were evaluated: one assuming 60% flexible power operation (FPO) of Cycle 32 and one assuming full-power operation of Cycle 32. The full-power operation projection bounds the FPO fluence values and thus only the full-power operation projections are reported herein. The cycle-dependent fuel assembly initial enrichments, burnups, radial and axial power distributions were used to develop spatial-and energy-dependent core source distributions averaged over each individual fuel cycle.

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Therefore, the results from the neutron transport calculations provided data in terms of fuel-cycle-averaged neutron exposure rate, which when multiplied by the appropriate fuel cycle length, generated the incremental fast neutron exposure for each fuel cycle. In constructing these core source distributions, the energy distribution of the source was based on an appropriate fission split for uranium and plutonium isotopes based on the initial enrichment and burnup history of individual fuel assemblies. From these assembly-dependent fission splits, composite values of energy release per fission, neutron yield per fission, and fission spectrum were determined.

All of the transport calculations supporting this analysis were carried out using the RAPTOR-M3G discrete ordinates code and the BUGLE-96 cross-section library, as described in Westinghouse Report WCAP-18124-NP-A [20]. The BUGLE-96 library provides a coupled 47-neutron, 20-gamma-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 angular discretization was modeled with an S16 order of angular quadrature. Energy-and space-dependent core power distributions, as well as system operating temperatures, were treated on a fuel-cycle-specific basis.

Results of the discrete ordinates transport analyses pertinent to the surveillance capsule evaluations are provided in Table 6-1 through Table 6-3. In Table 6-1, the calculated fast neutron fluence rate and fluence (E > 1.0 MeV) are provided at the geometric center of the capsules, as a function of irradiation time for the Prairie Island Unit 1 reactor. Similar data presented in terms of iron atom displacement rate and integrated iron atom displacements are given in Table 6-2. Note that the fluence values for the surveillance capsules are different than the previous report. This is largely due to the updated methodology used in determining the fluence values. The previous values were determined using 2D adjoint transport methods. This analysis employed 3D forward transport methods.

In Table 6-3, lead factors associated with surveillance capsules are provided as a function of operating time for the Prairie Island Unit 1 reactor. 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.

Neutron exposure data pertinent to the pressure vessel clad/base metal interface are given in Table 6-4 and Table 6-5 for neutron fluence (E > 1.0 MeV) rate and fluence (E > 1.0 MeV), respectively, and in Table 6-6 and Table 6-7 for dpa/s and dpa, respectively. In each case, the data are provided for each operating cycle of the Prairie Island Unit 1 reactor. Neutron fluence (E > 1.0 MeV) and dpa are also projected to future operating times extending to 60 EFPY. The vessel exposure data are presented in terms of the maximum exposure experienced by the pressure vessel at azimuthal angles of 0°, 15°, 30°, and 45°, and at the azimuthal location providing the maximum exposure relative to the core cardinal axes.

In Table 6-8 and Table 6-9, maximum projected fluences and dpa, respectively, of the various pressure vessel materials are given.

These data tabulations include both plant-and fuel-cycle-specific calculated neutron exposures at the end of Cycle 32 and projections to 60 EFPY. The projections beyond Cycle 32 were based on full-power operation of Cycle 32 spatial power distributions, water temperatures, and reactor power level with a 10%

bias on the peripheral and re-entrant corner assemblies.

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6.3 NEUTRON DOSIMETRY

The validity of the calculated neutron exposures reported in Section 6.2 is demonstrated by a direct comparison against the measured sensor reaction rates and a least-squares evaluation performed for each of the capsule dosimetry sets. However, since the neutron dosimetry measurement data merely serve to validate the calculated results, only the direct comparison of measured-to-calculated results for the most recent surveillance capsule removed from Prairie Island Unit 1, Capusle N, is provided in this section of the report. For completeness, the assessment of all measured dosimetry removed from Prairie Island Unit 1 up to date based on both direct and least-squares evaluation comparisons is documented in Appendix A.

The direct comparison of measured versus calculated fast neutron threshold reaction rates for the sensors from Capsule N, that was withdrawn from the reactor at the conclusion of Cycle 31, is summarized below.

Reaction Rate (rps/atom)

Reaction Measured Calculated M/C

63Cu (n,) 60Co 3.56E-17 4.28E-17 0.83

54Fe (n,p) 54Mn 3.63E-15 4.64E-15 0.78

58Ni (n,p) 58Co 5.71E-15 6.40E-15 0.89

238U(Cd) (n,f) 137Cs

237 137 Rejected Np(Cd) (n,f) Cs Average of M/C Results 0.83

Standard Deviation (%) 6.6

Note that the fission monitor results were determined to be non-credible and rejected. The measured-to-calculated (M/C) reaction rate ratios for the Capsule N threshold reactions range from 0.78 to 0.89, and the average M/C ratio is 0.83 6.6% (1). This direct comparison falls within the 20% criterion specified in Regulatory Guide 1.190. This comparison validates the current analytical results described in Section 6.2; therefore, the calculations are deemed applicable for Prairie Island Unit 1.

6.4 CALCULATIONAL UNCERTAINTIES

The uncertainty associated with the calculated neutron exposure of the Prairie Island Unit 1 surveillance capsule and reactor pressure vessel is based on the recommended approach provided in Regulatory Guide 1.190. In particular, the qualification of the methodology was carried out in the following four stages:

1. Simulator Benchmark Comparisons: Comparisons of calculations with measurements from simulator benchmarks, including the Pool Critical Assembly (PCA) simulator at the Oak Ridge National Laboratory (ORNL) and the VENUS-1 Experiment.

2. Operating Reactor and Calculational Benchmarks: Comparisons of calculations with surveillance capsule and reactor cavity measurements from the H.B. Robinson power reactor

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benchmark experiment. Also considered are comparisons of calculations to results published in the NRC fluence calculation benchmark.

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

4. Plant-Specific Benchmarking: Comparisons of the plant-specific calculations with all available dosimetry results from the Prairie Island Unit 1 surveillance program.

The first phase of the methods qualification (simulator benchmark comparisons) addressed the adequacy of basic transport calculation and dosimetry evaluation techniques and associated cross-sections. This phase, however, did not test the accuracy of commercial core neutron source calculations nor did it address uncertainties in operational or geometric variables that impact power reactor calculations. The second phase of the qualification (operating reactor and calculational benchmark comparisons) addressed uncertainties in these additional areas that are primarily methods-related and would tend to apply generically to all fast neutron exposure evaluations. The third phase of the qualification (analytical sensitivity study) identified the potential uncertainties introduced into the overall evaluation due to calculational methods approximations, as well as to a lack of knowledge relative to various plant-specific input parameters. The overall calculational uncertainty applicable to the Prairie Island Unit 1 analysis was established from results of these three phases of the methods qualification.

The fourth phase of the uncertainty assessment (comparisons with Prairie Island Unit 1 measurements) was used solely to demonstrate the validity of the transport calculations and to confirm the uncertainty estimates associated with the analytical results. The comparison was used only as a check and was not used in any way to modify the calculated surveillance capsule and pressure vessel neutron exposures described in Section 6.2. As such, the validation of the Prairie Island Unit 1 analytical model based on the measured plant dosimetry is completely described in Appendix A.

The following summarizes the uncertainties developed from the first three phases of the methodology qualification. Additional information pertinent to these evaluations is provided in Westinghouse Report WCAP-18124-NP-A [20].

Description Capsule and Vessel IR Simulator Benchmark Comparisons 3%

Operating Reactor and Calculational Benchmarks 5%

Analytic Uncertainty Analysis 11%

Additional Uncertainty for Factors not Explicitly Evaluated 5%

Net Calculational Uncertainty 13%

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 described in Appendix A support these uncertainty assessments for Prairie Island Unit 1.

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The NRC-issued Safety Evaluation for WCAP-18124-NP-A appears in Section A of [20]. The NRC identified two Limitations and Conditions associated with the application of RAPTOR-M3G and FERRET, which are reproduced here for convenience:

1. Applicability of WCAP-18124-NP, Revision 0 is limited to the RPV region near the active height of the core based on the uncertainty analysis performed and the measurement data provided.

Additional justification should be provided via additional benchmarking, fluence sensitivity analysis to the response parameters of interest (e.g., pressure-temperature limits, material stress/strain), margin assessment, or a combination thereof, for applications of the method to components including, but not limited to, the RPV upper circumferential weld and the reactor coolant system inlet and outlet nozzles and reactor vessel internal components.

2. Least squares adjustment is acceptable if the adjustments to the M/C radios and to the calculated spectra values are within the assigned uncertainties of the calculated spectra, the dosimetry measured reaction rates, and the dosimetry reaction cross sections. Should this not be the case, the user should re-examine both measured and calculated values for possible errors. If errors cannot be found, the particular values causing the discrepancy should be disqualified.

The primary purpose of this report is to describe the evaluation of a surveillance capsule. The neutron exposure values applicable to the surveillance capsules and the maximum reactor pressure vessel neutron exposure values used to derive the surveillance capsule lead factors are completely covered by the benchmarking and uncertainty analyses in WCAP-18124-NP-A. Therefore, Limitation # 1 does not strictly apply. Note, however, that this report does contain neutron exposure values for materials that are outside the qualification basis of WCAP-18124-NP-A (i.e., extended beltline materials). Should values outside the qualification basis of WCAP-18124-NP-A be cited in future evaluations of reactor vessel integrity, additional justification should be supplied, as stated in Limitation # 1.

Limitation # 2 applies in situations where the least-squares analysis is used to adjust the calculated values of neutron exposure. In this report, the least-squares analysis is provided only as a supplemental check on the results of the dosimetry evaluation. The least-squares analysis was not used to modify the calculated surveillance capsule or reactor pressure vessel neutron exposure. Therefore, Limitation # 2 does not apply.

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Table 6-1 Calculated Maximum Fast (E > 1.0 MeV) Neutron Fluence Rate and Fluence at Surveillance Capsule Locations

Cycle Total Fluence Rate (n/cm2-s)

Cycle Length Time 13° 23° 33° (EFPY) (EFPY) 1 1.40 1.40 1.38E+11 7.87E+10 7.59E+10 2 0.78 2.18 1.43E+11 8.53E+10 8.25E+10 3 0.86 3.03 1.53E+11 8.47E+10 7.90E+10 4 0.89 3.92 1.50E+11 8.91E+10 8.60E+10 5 0.99 4.91 1.55E+11 8.72E+10 8.29E+10 6 0.87 5.79 1.57E+11 8.96E+10 8.57E+10 7 1.01 6.80 1.38E+11 8.47E+10 9.03E+10 8 0.89 7.69 1.70E+11 9.61E+10 9.30E+10 9 0.94 8.63 1.34E+11 8.91E+10 8.85E+10 10 0.92 9.55 1.78E+11 9.07E+10 8.15E+10 11 0.93 10.48 1.83E+11 1.06E+11 9.80E+10 12 1.18 11.65 1.32E+11 9.06E+10 8.77E+10 13 1.24 12.89 9.65E+10 7.01E+10 6.70E+10 14 1.21 14.11 7.82E+10 5.70E+10 5.66E+10 15 1.25 15.36 7.91E+10 5.59E+10 5.54E+10 16 1.29 16.65 9.15E+10 6.86E+10 6.29E+10 17 1.47 18.12 9.34E+10 6.88E+10 6.13E+10 18 1.55 19.68 8.21E+10 5.66E+10 5.35E+10 19 1.21 20.89 8.25E+10 6.71E+10 6.41E+10 20 1.61 22.50 8.97E+10 6.58E+10 5.97E+10 21 1.60 24.09 8.88E+10 5.82E+10 5.52E+10 22 1.72 25.81 9.07E+10 6.26E+10 6.19E+10 23 1.36 27.17 9.97E+10 6.72E+10 6.37E+10 24 1.61 28.78 8.79E+10 5.96E+10 5.60E+10 25 1.43 30.21 8.92E+10 6.34E+10 6.25E+10 26 1.42 31.63 9.10E+10 6.08E+10 5.84E+10 27 1.33 32.96 8.95E+10 5.59E+10 5.36E+10 28 1.74 34.71 8.79E+10 5.60E+10 5.37E+10 29 1.69 36.40 9.11E+10 6.01E+10 5.87E+10 30 1.79 38.19 8.87E+10 6.13E+10 6.30E+10 31 1.87 40.06 9.16E+10 5.93E+10 5.81E+10 32 2.03 42.08 8.77E+10 5.88E+10 5.96E+10

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Table 6-1 Calculated Maximum Fast (E > 1.0 MeV) Neutron Fluence Rate and Fluence at Surveillance Capsule Locations (cont.)

Total Fluence (n/cm2)

Cycle Time V P R S N T (EFPY) 1 1.40 6.09E+18 3.47E+18 6.09E+18 3.35E+18 3.35E+18 3.47E+18 2 2.18 5.57E+18 9.60E+18 5.38E+18 5.38E+18 5.57E+18 3 3.03 7.86E+18 1.37E+19 7.51E+18 7.51E+18 7.86E+18 4 3.92 1.04E+19 1.79E+19 9.93E+18 9.93E+18 1.04E+19 5 4.91 1.31E+19 2.28E+19 1.25E+19 1.25E+19 1.31E+19 6 5.79 2.71E+19 1.49E+19 1.49E+19 1.55E+19 7 6.80 3.15E+19 1.78E+19 1.78E+19 1.83E+19 8 7.69 3.63E+19 2.04E+19 2.04E+19 2.09E+19 9 8.63 4.02E+19 2.30E+19 2.30E+19 2.36E+19 10 9.55 2.54E+19 2.54E+19 2.62E+19 11 10.48 2.82E+19 2.82E+19 2.93E+19 12 11.65 3.15E+19 3.15E+19 3.27E+19 13 12.89 3.41E+19 3.41E+19 3.54E+19 14 14.11 3.63E+19 3.63E+19 3.76E+19 15 15.36 3.85E+19 3.85E+19 3.98E+19 16 16.65 4.10E+19 4.10E+19 4.26E+19 17 18.12 4.39E+19 4.39E+19 4.58E+19 18 19.68 4.65E+19 4.86E+19 19 20.89 4.90E+19 5.12E+19 20 22.50 5.20E+19 5.45E+19 21 24.09 5.48E+19 5.74E+19 22 25.81 5.81E+19 6.08E+19 23 27.17 6.08E+19 6.37E+19 24 28.78 6.37E+19 6.67E+19 25 30.21 6.65E+19 6.96E+19 26 31.63 6.91E+19 7.23E+19 27 32.96 7.14E+19 7.47E+19 28 34.71 7.44E+19 7.78E+19 29 36.40 7.75E+19 8.10E+19 30 38.19 8.10E+19 8.44E+19 31 40.06 8.45E+19 8.79E+19 32 42.08 9.17E+19 48 1.04E+20 51 1.10E+20 54 1.16E+20 60 1.28E+20 Note:

1. Values beyond Cycle 32 are projected based on full-power operation of Cycle 32 with a 10% bias on the peripheral and re-entrant corner assemblies.

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Table 6-2 Calculated Iron Atom Displacement Rate and Iron Atom Displacements at Surveillance Capsule Locations

Cycle Total Iron Atom Displacement Rate Cycle Length Time (dpa/s)

(EFPY) (EFPY) 13-degree 23-degree 33-degree 1 1.40 1.40 2.53E-10 1.38E-10 1.34E-10 2 0.78 2.18 2.61E-10 1.50E-10 1.46E-10 3 0.86 3.03 2.80E-10 1.49E-10 1.39E-10 4 0.89 3.92 2.75E-10 1.56E-10 1.52E-10 5 0.99 4.91 2.84E-10 1.53E-10 1.46E-10 6 0.87 5.79 2.87E-10 1.57E-10 1.51E-10 7 1.01 6.80 2.53E-10 1.49E-10 1.59E-10 8 0.89 7.69 3.12E-10 1.69E-10 1.64E-10 9 0.94 8.63 2.44E-10 1.56E-10 1.56E-10 10 0.92 9.55 3.27E-10 1.59E-10 1.43E-10 11 0.93 10.48 3.37E-10 1.86E-10 1.73E-10 12 1.18 11.65 2.40E-10 1.58E-10 1.55E-10 13 1.24 12.89 1.75E-10 1.22E-10 1.18E-10 14 1.21 14.11 1.42E-10 9.90E-11 9.93E-11 15 1.25 15.36 1.43E-10 9.72E-11 9.71E-11 16 1.29 16.65 1.66E-10 1.19E-10 1.11E-10 17 1.47 18.12 1.69E-10 1.20E-10 1.08E-10 18 1.55 19.68 1.49E-10 9.84E-11 9.37E-11 19 1.21 20.89 1.49E-10 1.17E-10 1.13E-10 20 1.61 22.50 1.63E-10 1.14E-10 1.05E-10 21 1.60 24.09 1.61E-10 1.01E-10 9.68E-11 22 1.72 25.81 1.65E-10 1.09E-10 1.09E-10 23 1.36 27.17 1.81E-10 1.17E-10 1.12E-10 24 1.61 28.78 1.59E-10 1.04E-10 9.82E-11 25 1.43 30.21 1.62E-10 1.10E-10 1.10E-10 26 1.42 31.63 1.65E-10 1.06E-10 1.03E-10 27 1.33 32.96 1.63E-10 9.74E-11 9.40E-11 28 1.74 34.71 1.60E-10 9.76E-11 9.42E-11 29 1.69 36.40 1.66E-10 1.05E-10 1.03E-10 30 1.79 38.19 1.61E-10 1.07E-10 1.11E-10 31 1.87 40.06 1.66E-10 1.03E-10 1.02E-10 32 2.03 42.08 1.59E-10 1.02E-10 1.05E-10

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Table 6-2 Calculated Iron Atom Displacement Rate and Iron Atom Displacements at Surveillance Capsule Locations (cont.)

Total Iron Atom Displacements (dpa)

Cycle Time V P R S N T (EFPY) 1 1.40 1.12E-02 6.09E-03 1.12E-02 5.91E-03 5.91E-03 6.09E-03 2 2.18 9.76E-03 1.76E-02 9.49E-03 9.49E-03 9.76E-03 3 3.03 1.38E-02 2.52E-02 1.32E-02 1.32E-02 1.38E-02 4 3.92 1.82E-02 3.29E-02 1.75E-02 1.75E-02 1.82E-02 5 4.91 2.29E-02 4.18E-02 2.21E-02 2.21E-02 2.29E-02 6 5.79 4.97E-02 2.62E-02 2.62E-02 2.73E-02 7 6.80 5.78E-02 3.13E-02 3.13E-02 3.20E-02 8 7.69 6.65E-02 3.59E-02 3.59E-02 3.67E-02 9 8.63 7.38E-02 4.06E-02 4.06E-02 4.14E-02 10 9.55 4.47E-02 4.47E-02 4.60E-02 11 10.48 4.98E-02 4.98E-02 5.14E-02 12 11.65 5.55E-02 5.55E-02 5.73E-02 13 12.89 6.02E-02 6.02E-02 6.21E-02 14 14.11 6.40E-02 6.40E-02 6.59E-02 15 15.36 6.78E-02 6.78E-02 6.97E-02 16 16.65 7.23E-02 7.23E-02 7.46E-02 17 18.12 7.73E-02 7.73E-02 8.01E-02 18 19.68 8.19E-02 8.49E-02 19 20.89 8.62E-02 8.94E-02 20 22.50 9.15E-02 9.52E-02 21 24.09 9.64E-02 1.00E-01 22 25.81 1.02E-01 1.06E-01 23 27.17 1.07E-01 1.11E-01 24 28.78 1.12E-01 1.17E-01 25 30.21 1.17E-01 1.21E-01 26 31.63 1.22E-01 1.26E-01 27 32.96 1.26E-01 1.30E-01 28 34.71 1.31E-01 1.36E-01 29 36.40 1.36E-01 1.41E-01 30 38.19 1.43E-01 1.47E-01 31 40.06 1.49E-01 1.53E-01 32 42.08 1.60E-01 48 1.81E-01 51 1.91E-01 54 2.02E-01 60 2.23E-01 Note:

1. Values beyond Cycle 32 are projected based on full-power operation of Cycle 32 with a 10% bias on the peripheral and re-entrant corner assemblies.

WCAP-18660-NP November 2021 Revision 0

• • • This record was final approved on 12/1/2021 8:24:23 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 6-11

Table 6-3 Calculated Surveillance Capsule Lead Factors

Total Lead Factor Time V P R S N T Cycle (EFPY) 1 1.40 2.981 1.70 2.98 1.64 1.64 1.70 2 2.18 1.76 3.04 1.70 1.70 1.76 3 3.03 1.75 3.06 1.67 1.67 1.75 4 3.92 1.77 3.07 1.70 1.70 1.77 5 4.91 1.762 3.07 1.69 1.69 1.762 6 5.79 3.07 1.69 1.69 1.76 7 6.80 3.11 1.75 1.75 1.80 8 7.69 3.10 1.74 1.74 1.79 9 8.63 3.083 1.76 1.76 1.80 10 9.55 1.72 1.72 1.77 11 10.48 1.71 1.71 1.77 12 11.65 1.73 1.73 1.80 13 12.89 1.76 1.76 1.83 14 14.11 1.79 1.79 1.86 15 15.36 1.81 1.81 1.88 16 16.65 1.84 1.84 1.91 17 18.12 1.864 1.86 1.94 18 19.68 1.87 1.96 19 20.89 1.90 1.99 20 22.50 1.92 2.01 21 24.09 1.92 2.01 22 25.81 1.93 2.02 23 27.17 1.93 2.02 24 28.78 1.94 2.03 25 30.21 1.95 2.04 26 31.63 1.95 2.05 27 32.96 1.95 2.04 28 34.71 1.95 2.04 29 36.40 1.95 2.04 30 38.19 1.97 2.05 31 40.06 1.975 2.05 32 42.08 2.05 48 2.05 51 2.06 54 2.06 60 2.06 Notes:

1. Capsule V was removed after Cycle 1.

2. Capsule P was removed after Cycle 5.

3. Capsule R was removed after Cycle 9.

4. Capsule S was removed after Cycle 17.

5. Capsule N was removed after Cycle 31.

6. The projections beyond Cycle 32 are based on full-power operations with 10% bias on peripheral and re-entrant corner assemblies.

WCAP-18660-NP November 2021 Revision 0

• • • This record was final approved on 12/1/2021 8:24:23 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 6-12

Table 6-4 Calculated Maximum Fast (E > 1.0 MeV) Neutron Fluence Rate at the Pressure Vessel Clad/Base Metal Interface

Cycle Total Fluence Rate (n/cm2-s)

Cycle Length Time Elevation (EFPY) (EFPY) 0° 15° 30° 45° Maximum of Max.

(cm) 1 1.40 1.40 4.61E+10 2.74E+10 1.82E+10 1.60E+10 4.64E+10 -73 2 0.78 2.18 4.77E+10 2.95E+10 2.03E+10 1.73E+10 4.79E+10 121 3 0.86 3.03 4.96E+10 2.93E+10 1.88E+10 1.70E+10 4.98E+10 -1 4 0.89 3.92 4.81E+10 2.90E+10 2.03E+10 1.75E+10 4.83E+10 3 5 0.99 4.91 5.05E+10 3.01E+10 1.97E+10 1.78E+10 5.08E+10 67 6 0.87 5.79 5.06E+10 3.06E+10 2.03E+10 1.75E+10 5.09E+10 -73 7 1.01 6.80 4.13E+10 2.68E+10 2.07E+10 1.92E+10 4.16E+10 5 8 0.89 7.69 5.57E+10 3.24E+10 2.20E+10 1.81E+10 5.60E+10 5 9 0.94 8.63 4.60E+10 2.66E+10 2.09E+10 1.73E+10 4.62E+10 3 10 0.92 9.55 5.84E+10 3.36E+10 1.95E+10 1.86E+10 5.87E+10 5 11 0.93 10.48 5.98E+10 3.52E+10 2.36E+10 1.73E+10 6.02E+10 5 12 1.18 11.65 4.43E+10 2.65E+10 2.10E+10 1.63E+10 4.45E+10 5 13 1.24 12.89 2.94E+10 2.00E+10 1.61E+10 1.44E+10 2.96E+10 5 14 1.21 14.11 2.39E+10 1.64E+10 1.35E+10 1.27E+10 2.40E+10 73 15 1.25 15.36 2.36E+10 1.63E+10 1.31E+10 1.26E+10 2.37E+10 5 16 1.29 16.65 2.76E+10 1.91E+10 1.54E+10 1.29E+10 2.77E+10 3 17 1.47 18.12 2.75E+10 1.95E+10 1.51E+10 1.23E+10 2.76E+10 5 18 1.55 19.68 2.51E+10 1.69E+10 1.29E+10 1.17E+10 2.52E+10 5 19 1.21 20.89 2.31E+10 1.77E+10 1.55E+10 1.36E+10 2.32E+10 67 20 1.61 22.50 2.75E+10 1.92E+10 1.50E+10 1.25E+10 2.76E+10 71 21 1.60 24.09 2.94E+10 1.80E+10 1.33E+10 1.19E+10 2.96E+10 5 22 1.72 25.81 2.87E+10 1.88E+10 1.49E+10 1.34E+10 2.88E+10 67 23 1.36 27.17 3.17E+10 2.04E+10 1.54E+10 1.39E+10 3.18E+10 67 24 1.61 28.78 2.76E+10 1.81E+10 1.36E+10 1.21E+10 2.78E+10 67 25 1.43 30.21 2.73E+10 1.82E+10 1.48E+10 1.30E+10 2.74E+10 5 26 1.42 31.63 2.83E+10 1.87E+10 1.40E+10 1.24E+10 2.85E+10 -73 27 1.33 32.96 2.86E+10 1.82E+10 1.29E+10 1.17E+10 2.88E+10 -73 28 1.74 34.71 2.79E+10 1.79E+10 1.29E+10 1.18E+10 2.80E+10 -73 29 1.69 36.40 2.93E+10 1.86E+10 1.40E+10 1.22E+10 2.95E+10 -73 30 1.79 38.19 2.73E+10 1.82E+10 1.47E+10 1.37E+10 2.75E+10 -73 31 1.87 40.06 2.94E+10 1.85E+10 1.38E+10 1.22E+10 2.96E+10 -73 32 2.03 42.08 2.80E+10 1.79E+10 1.40E+10 1.30E+10 2.81E+10 -73

WCAP-18660-NP November 2021 Revision 0

• • • This record was final approved on 12/1/2021 8:24:23 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 6-13

Table 6-5 Calculated Maximum Fast (E > 1.0 MeV) Neutron Fluence at the Pressure Vessel Clad/Base Metal Interface (Full-Power Operations)

Cycle Total Fluence (n/cm2)

Cycle Length Time Elevation (EFPY) (EFPY) 0° 15° 30° 45° Maximum of Max.

(cm) 1 1.40 1.40 2.04E+18 1.21E+18 8.04E+17 7.08E+17 2.05E+18 -73 2 0.78 2.18 3.14E+18 1.89E+18 1.28E+18 1.11E+18 3.16E+18 -73 3 0.86 3.03 4.47E+18 2.66E+18 1.78E+18 1.57E+18 4.49E+18 -3 4 0.89 3.92 5.81E+18 3.48E+18 2.35E+18 2.06E+18 5.84E+18 -3 5 0.99 4.91 7.38E+18 4.38E+18 2.96E+18 2.61E+18 7.42E+18 -1 6 0.87 5.79 8.77E+18 5.21E+18 3.52E+18 3.09E+18 8.81E+18 -1 7 1.01 6.8 1.01E+19 6.07E+18 4.18E+18 3.71E+18 1.01E+19 -1 8 0.89 7.69 1.16E+19 6.97E+18 4.80E+18 4.22E+18 1.17E+19 1 9 0.94 8.63 1.30E+19 7.76E+18 5.42E+18 4.73E+18 1.31E+19 1 10 0.92 9.55 1.47E+19 8.74E+18 5.98E+18 5.27E+18 1.48E+19 3 11 0.93 10.48 1.65E+19 9.77E+18 6.68E+18 5.78E+18 1.65E+19 3 12 1.18 11.65 1.81E+19 1.08E+19 7.45E+18 6.38E+18 1.82E+19 3 13 1.24 12.89 1.93E+19 1.15E+19 8.09E+18 6.95E+18 1.94E+19 3 14 1.21 14.11 2.02E+19 1.22E+19 8.60E+18 7.43E+18 2.03E+19 3 15 1.25 15.36 2.11E+19 1.28E+19 9.12E+18 7.93E+18 2.12E+19 3 16 1.29 16.65 2.22E+19 1.36E+19 9.75E+18 8.45E+18 2.23E+19 3 17 1.47 18.12 2.35E+19 1.45E+19 1.04E+19 9.02E+18 2.36E+19 3 18 1.55 19.68 2.47E+19 1.53E+19 1.11E+19 9.59E+18 2.48E+19 3 19 1.21 20.89 2.56E+19 1.60E+19 1.17E+19 1.01E+19 2.57E+19 3 20 1.61 22.5 2.70E+19 1.69E+19 1.24E+19 1.07E+19 2.71E+19 5 21 1.60 24.09 2.84E+19 1.78E+19 1.31E+19 1.13E+19 2.86E+19 5 22 1.72 25.81 3.00E+19 1.88E+19 1.39E+19 1.21E+19 3.01E+19 5 23 1.36 27.17 3.13E+19 1.97E+19 1.45E+19 1.26E+19 3.15E+19 5 24 1.61 28.78 3.27E+19 2.06E+19 1.52E+19 1.33E+19 3.29E+19 5 25 1.43 30.21 3.39E+19 2.14E+19 1.59E+19 1.38E+19 3.41E+19 5 26 1.42 31.63 3.52E+19 2.23E+19 1.65E+19 1.44E+19 3.54E+19 5 27 1.33 32.96 3.64E+19 2.30E+19 1.70E+19 1.49E+19 3.66E+19 5 28 1.74 34.71 3.79E+19 2.40E+19 1.78E+19 1.55E+19 3.81E+19 5 29 1.69 36.4 3.95E+19 2.49E+19 1.85E+19 1.62E+19 3.97E+19 5 30 1.79 38.19 4.10E+19 2.59E+19 1.93E+19 1.69E+19 4.12E+19 5 31 1.87 40.06 4.27E+19 2.70E+19 2.01E+19 1.77E+19 4.29E+19 3 32 2.03 42.08 4.45E+19 2.81E+19 2.10E+19 1.85E+19 4.47E+19 3 48 5.02E+19 3.17E+19 2.39E+19 2.11E+19 5.05E+19 3 51 5.32E+19 3.36E+19 2.53E+19 2.24E+19 5.34E+19 3 54 5.61E+19 3.54E+19 2.67E+19 2.38E+19 5.63E+19 3 60 6.19E+19 3.90E+19 2.96E+19 2.65E+19 6.22E+19 3 Note:

1. Projections are based on full-power operations of Cycle 32 with a 10% bias on the peripheral and re-entrant corner assemblies.

WCAP-18660-NP November 2021 Revision 0

• • • This record was final approved on 12/1/2021 8:24:23 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 6-14

Table 6-6 Calculated Maximum Iron Atom Displacement Rate at the Pressure Vessel Clad/Base Metal Interface

Cycle Total Iron Atom Displacement Rate (dpa/s)

Cycle Length Time Elevation (EFPY) (EFPY) 0° 15° 30° 45° Maximum of Max.

(cm) 1 1.40 1.40 7.56E-11 4.56E-11 3.00E-11 2.61E-11 7.59E-11 -73 2 0.78 2.18 7.83E-11 4.88E-11 3.30E-11 2.81E-11 7.87E-11 119 3 0.86 3.03 8.14E-11 4.97E-11 3.11E-11 2.77E-11 8.17E-11 -1 4 0.89 3.92 7.89E-11 4.93E-11 3.36E-11 2.85E-11 7.92E-11 3 5 0.99 4.91 8.30E-11 5.07E-11 3.25E-11 2.89E-11 8.33E-11 67 6 0.87 5.79 8.30E-11 5.10E-11 3.35E-11 2.85E-11 8.34E-11 -73 7 1.01 6.80 6.79E-11 4.55E-11 3.42E-11 3.13E-11 6.82E-11 5 8 0.89 7.69 9.14E-11 5.51E-11 3.63E-11 2.95E-11 9.18E-11 5 9 0.94 8.63 7.54E-11 4.51E-11 3.44E-11 2.82E-11 7.57E-11 3 10 0.92 9.55 9.59E-11 5.71E-11 3.23E-11 3.02E-11 9.63E-11 5 11 0.93 10.48 9.83E-11 5.98E-11 3.90E-11 2.83E-11 9.87E-11 5 12 1.18 11.65 7.26E-11 4.48E-11 3.46E-11 2.67E-11 7.29E-11 5 13 1.24 12.89 4.82E-11 3.37E-11 2.65E-11 2.35E-11 4.84E-11 5 14 1.21 14.11 3.92E-11 2.74E-11 2.22E-11 2.06E-11 3.94E-11 73 15 1.25 15.36 3.87E-11 2.75E-11 2.17E-11 2.04E-11 3.88E-11 5 16 1.29 16.65 4.52E-11 3.22E-11 2.53E-11 2.09E-11 4.54E-11 3 17 1.47 18.12 4.50E-11 3.28E-11 2.49E-11 2.00E-11 4.52E-11 5 18 1.55 19.68 4.11E-11 2.84E-11 2.12E-11 1.90E-11 4.12E-11 5 19 1.21 20.89 3.79E-11 2.98E-11 2.56E-11 2.22E-11 3.81E-11 67 20 1.61 22.50 4.51E-11 3.22E-11 2.46E-11 2.04E-11 4.53E-11 73 21 1.60 24.09 4.81E-11 3.04E-11 2.19E-11 1.93E-11 4.84E-11 5 22 1.72 25.81 4.70E-11 3.16E-11 2.45E-11 2.18E-11 4.72E-11 67 23 1.36 27.17 5.18E-11 3.43E-11 2.53E-11 2.26E-11 5.20E-11 67 24 1.61 28.78 4.52E-11 3.04E-11 2.24E-11 1.97E-11 4.54E-11 67 25 1.43 30.21 4.47E-11 3.06E-11 2.44E-11 2.11E-11 4.49E-11 5 26 1.42 31.63 4.64E-11 3.10E-11 2.30E-11 2.01E-11 4.66E-11 -73 27 1.33 32.96 4.68E-11 3.01E-11 2.11E-11 1.90E-11 4.70E-11 -73 28 1.74 34.71 4.56E-11 2.97E-11 2.12E-11 1.91E-11 4.57E-11 -73 29 1.69 36.40 4.79E-11 3.08E-11 2.30E-11 1.99E-11 4.82E-11 -73 30 1.79 38.19 4.47E-11 3.02E-11 2.42E-11 2.22E-11 4.49E-11 -3 31 1.87 40.06 4.81E-11 3.08E-11 2.26E-11 1.98E-11 4.84E-11 -73 32 2.03 42.08 4.58E-11 2.97E-11 2.29E-11 2.10E-11 4.60E-11 61

WCAP-18660-NP November 2021 Revision 0

• • • This record was final approved on 12/1/2021 8:24:23 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 6-15

Table 6-7 Calculated Maximum Iron Atom Displacements at the Pressure Vessel Clad/Base Metal Interface (Full-Power Operations)

Cycle Total Iron Atom Displacements (dpa)

Cycle Length Time Elevation (EFPY) (EFPY) 0° 15° 30° 45° Maximum of Max.

(cm) 1 1.40 1.40 3.34E-03 2.01E-03 1.32E-03 1.15E-03 3.35E-03 -73 2 0.78 2.18 5.15E-03 3.14E-03 2.10E-03 1.81E-03 5.17E-03 -73 3 0.86 3.03 7.33E-03 4.49E-03 2.94E-03 2.55E-03 7.36E-03 -3 4 0.89 3.92 9.54E-03 5.87E-03 3.88E-03 3.35E-03 9.58E-03 -3 5 0.99 4.91 1.21E-02 7.44E-03 4.89E-03 4.25E-03 1.22E-02 -1 6 0.87 5.79 1.44E-02 8.85E-03 5.81E-03 5.04E-03 1.45E-02 -1 7 1.01 6.80 1.66E-02 1.03E-02 6.90E-03 6.04E-03 1.66E-02 -1 8 0.89 7.69 1.91E-02 1.18E-02 7.92E-03 6.87E-03 1.92E-02 -1 9 0.94 8.63 2.14E-02 1.32E-02 8.94E-03 7.70E-03 2.14E-02 -1 10 0.92 9.55 2.41E-02 1.48E-02 9.88E-03 8.58E-03 2.42E-02 3 11 0.93 10.48 2.70E-02 1.66E-02 1.10E-02 9.41E-03 2.71E-02 3 12 1.18 11.65 2.97E-02 1.83E-02 1.23E-02 1.04E-02 2.98E-02 3 13 1.24 12.89 3.16E-02 1.96E-02 1.33E-02 1.13E-02 3.17E-02 3 14 1.21 14.11 3.31E-02 2.06E-02 1.42E-02 1.21E-02 3.32E-02 3 15 1.25 15.36 3.46E-02 2.17E-02 1.51E-02 1.29E-02 3.48E-02 3 16 1.29 16.65 3.65E-02 2.30E-02 1.61E-02 1.38E-02 3.66E-02 3 17 1.47 18.12 3.85E-02 2.45E-02 1.72E-02 1.47E-02 3.87E-02 5 18 1.55 19.68 4.06E-02 2.59E-02 1.83E-02 1.56E-02 4.07E-02 5 19 1.21 20.89 4.20E-02 2.71E-02 1.93E-02 1.65E-02 4.22E-02 5 20 1.61 22.50 4.42E-02 2.87E-02 2.05E-02 1.75E-02 4.44E-02 5 21 1.60 24.09 4.66E-02 3.02E-02 2.16E-02 1.85E-02 4.68E-02 5 22 1.72 25.81 4.92E-02 3.19E-02 2.29E-02 1.96E-02 4.94E-02 5 23 1.36 27.17 5.13E-02 3.33E-02 2.40E-02 2.06E-02 5.16E-02 5 24 1.61 28.78 5.36E-02 3.49E-02 2.51E-02 2.16E-02 5.38E-02 5 25 1.43 30.21 5.56E-02 3.63E-02 2.62E-02 2.25E-02 5.59E-02 5 26 1.42 31.63 5.77E-02 3.76E-02 2.72E-02 2.34E-02 5.79E-02 5 27 1.33 32.96 5.96E-02 3.89E-02 2.81E-02 2.42E-02 5.99E-02 5 28 1.74 34.71 6.21E-02 4.05E-02 2.93E-02 2.53E-02 6.24E-02 5 29 1.69 36.40 6.47E-02 4.22E-02 3.05E-02 2.63E-02 6.50E-02 5 30 1.79 38.19 6.72E-02 4.39E-02 3.18E-02 2.76E-02 6.75E-02 5 31 1.87 40.06 7.00E-02 4.57E-02 3.32E-02 2.87E-02 7.03E-02 5 32 2.03 42.08 7.30E-02 4.76E-02 3.46E-02 3.01E-02 7.33E-02 5 48 8.23E-02 5.37E-02 3.93E-02 3.43E-02 8.27E-02 3 51 8.71E-02 5.67E-02 4.17E-02 3.65E-02 8.74E-02 3 54 9.18E-02 5.98E-02 4.41E-02 3.87E-02 9.22E-02 3 60 1.01E-01 6.60E-02 4.88E-02 4.30E-02 1.02E-01 3 Note:

1. Projections are based on full-power operations of Cycle 32 with a 10% bias on the peripheral and re-entrant corner assemblies.

WCAP-18660-NP November 2021 Revision 0

• • • This record was final approved on 12/1/2021 8:24:23 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 6-16

Table 6-8 Calculated Maximum Fast Neutron Fluence (E > 1.0 MeV) at Pressure Vessel Welds and Shells

Fast Neutron (E > 1.0 MeV) Fluence (n/cm2)

Material 42.1 EFPY 48 EFPY 54 EFPY 60 EFPY

Upper Shell Forging 2.63E+19 3.00E+19 3.37E+19 3.74E+19 Intermediate Shell Forging 4.47E+19 5.05E+19 5.63E+19 6.22E+19 Lower Shell Forging 4.37E+19 4.95E+19 5.53E+19 6.12E+19 Inlet Nozzle to Nozzle Shell Weld 2.56E+16 2.96E+16 3.36E+16 3.76E+16

- Lowest Extent(1)

Upper to Intermediate Shell Weld 2.84E+19 3.23E+19 3.63E+19 4.02E+19 Intermediate to Lower Shell Weld 4.38E+19 4.95E+19 5.53E+19 6.12E+19 Lower Shell to Lower Closure 1.49E+16 1.70E+16 1.92E+16 2.14E+16 Head Weld Note:

1. The outlet nozzle weld fluence is bounded by the inlet nozzle location fluence.

2. See description of Limitation 1 described in Section 6.4 relative to the use of fluence values generated for "extended beltline" materials.

WCAP-18660-NP November 2021 Revision 0

• • • This record was final approved on 12/1/2021 8:24:23 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 6-17

Table 6-9 Calculated Maximum Iron Atom Displacements at Pressure Vessel Welds and Shells

Material Iron Atom Displacements (dpa) 42.1 EFPY 48 EFPY 54 EFPY 60 EFPY Upper Shell Forging 4.33E-02 4.93E-02 5.54E-02 6.15E-02 Intermediate Shell Forging 7.33E-02 8.27E-02 9.22E-02 1.02E-01 Lower Shell Forging 7.16E-02 8.10E-02 9.05E-02 1.00E-01 Inlet Nozzle to Nozzle Shell Weld 1.25E-04 1.43E-04 1.62E-04 1.80E-04

- Lowest Extent(1)

Upper to Intermediate Shell Weld 4.67E-02 5.31E-02 5.96E-02 6.61E-02 Intermediate to Lower Shell Weld 7.17E-02 8.10E-02 9.06E-02 1.00E-01 Lower Shell to Lower Closure 7.60E-05 8.70E-05 9.81E-05 1.09E-04 Head Weld Note:

1. The outlet nozzle weld dpa is bounded by the inlet nozzle location dpa.

WCAP-18660-NP November 2021 Revision 0

• • • This record was final approved on 12/1/2021 8:24:23 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 6-18

Table 6-10 Summary of Reactor Power and RCS Temperatures

Core Core Inlet Cycle Core Power Average Coolant Temperature (MWth) Temperature (°F)

(°F) 1 1650 571 536 2 1650 571 536 3 1650 571 536 4 1650 571 536 5 1650 571 536 6 1650 571 536 7 1650 571 536 8 1650 571 536 9 1650 571 536 10 1650 571 536 11 1650 571 536 12 1650 571 536 13 1650 571 536 14 1650 571 536 15 1650 571 536 16 1650 571 536 17 1650 571 536 18 1650 571 536 19 1650 571 536 20 1650 571 536 21 1650 571 536 22 1650 571 536 23 1650 563 531 24 1650 563 531 25 1650 563 531 26 1660(1) 563 531 27 1677 563 530 28 1677 562 532 29 1677 562 532 30 1677 562 531 31 1677 562 531 32 1677 562 531 Note(s):

1. Note that there was a mid-cycle power uprate during Cycle 26 from 1650 MWth to 1677 MWth. This uprate was done at a burnup of 12,015 MWD/MTU. The burnup-weighted average thermal power of 1660 MWth was used for the Cycle 26 transport calculations.

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Figure 6-1 Prairie Island Unit 1 Plan View of the Reactor Geometry at the Core Midplane

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Figure 6-2 Prairie Island Unit 1 Plan View of the Reactor Geometry at the Nozzle Centerline

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Figure 6-3 Prairie Island Unit 1 Section View of the Reactor Geometry at 0-Degrees

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Figure 6-4 Prairie Island Unit 1 Section View of the Reactor Geometry at 33-Degrees

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7 SURVEILLANCE CAPSULE REMOVAL SCHEDULE

The following surveillance capsule removal schedule (Table 7-1) meets the requirements of ASTM E185-82

[10] with consideration of [21]. It is noted that the Capsule N fluence bounds the projected fluence of the Prairie Island Unit 1 RV through 80 years of operation (peak vessel fluence of 7.40 x 1019 n/cm2 at 72 (EFPY).

Table 7-1 Surveillance Capsule Withdrawal Schedule Fluence Capsule Capsule Location Lead Factor Withdrawal EFPY (1) (n/cm2, E > 1.0 MeV)

V 77° 2.98 1.40 6.09 x 1018 (EOC 1)

P 247° 1.76 4.91 1.31 x 1019 (EOC 5)

R 257° 3.08 8.63 4.02 x 1019 (EOC 9)

S 57° 1.86 18.12 4.39 x 1019 (EOC 17)

N 237° 1.97 40.06 8.45 x 1019 (EOC 31)

T 67° 2.05 Standby(2) - - -

Notes:

1. Effective full-power years (EFPY) from plant startup. EOC = end-of-cycle.

2. It is recommended that Capsule T be removed at approximately 51 EFPY, which is the projected peak reactor vessel fluence at 120 years (1.09 x 1020 n/cm2 at 108 EFPY). The need for an alternative form of neutron dosimetry should be assessed when the last capsule is withdrawn.

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8 REFERENCES

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

[Agencywide Documents Access and Management System (ADAMS) Accession Number ML003740284]

2. Westinghouse Report WCAP-8086, Rev. 0, Northern States Power Co. Prairie Island Unit No. 1 Reactor Vessel Radiation Surveillance Program, June 1973.

3. ASTM E185-70, Standard Recommended Practice for Surveillance Tests for Nuclear Reactor Vessels, 1970.

4. Appendix G of the ASME Boiler and Pressure Vessel (B&PV) Code,Section XI, Division 1, Fracture Toughness Criteria for Protection Against Failure.

5. ASTM E208, Standard Test Method for Conducting Drop -Weight Test to Determine Nil-Ductility Transition Temperature of Ferritic Steels, ASTM.

6. ASTM E399, Standard Test Method for Linear-Elastic Plane-Strain Fracture Toughness Klc of Metallic Materials, ASTM.

7. Westinghouse Drawing 685J394, Rev. 3, Large Pressurized Water Reactor Irradiation Sample Assembly.

8. Westinghouse Report WCAP-11006, Rev. 0, Analysis of Capsule R from the Northern States Power Company Prairie Island Unit 1 Reactor Vessel Radiation Surveillance Program, February 1986.

9. 10 CFR 50, Appendix H, Reactor Vessel Material Surveillance Program Requirements, U.S.

Nuclear Regulatory Commission, Federal Register, October 2, 2020.

10. ASTM E185-82, Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels, American Society for Testing and Materials, 1982.

11. ASTM E23-18, Standard Test Methods for Notched Bar Impact Testing of Metallic Materials, 2018.

12. ASTM E2298-18, Standard Test Method for Instrumented Impact Testing of Metallic Materials, 2018.

13. ASTM A370-18, Standard Test Methods and Definitions for Mechanical Testing of Steel Products, 2018.

14. ASTM E8/E8M-16a, Standard Test Methods for Tension Testing of Metallic Materials, 2016.

15. ASTM E21-17, Standard Test Methods for Elevated Temperature Tension Tests of Metallic Materials, 2017.

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16. Westinghouse Report WCAP-14779, Rev. 2, Analysis of Capsule S from the Northern States Power Company Prairie Island Unit 1 Reactor Vessel Radiation Surveillance Program, February 1998.

17. ASTM E853-18, Standard Practice for Analysis and Interpretation of Light-Water Reactor Surveillance Neutron Exposure Results, 2018.

18. ASTM E693-94, Standard Practice for Characterizing Neutron Exposures in Iron and Low Alloy Steels in Terms of Displacements Per Atom (DPA), E706 (ID), 1994.

19. U.S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, Regulatory Guide 1.190, Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence, March 2001. [ADAMS Accession Number ML010890301]

20. Westinghouse Report WCAP-18124-NP-A, Rev. 0, Fluence Determination with RAPTOR -M3G and FERRET, July 2018. [ADAMS Accession Number ML18204A010]

21. NUREG-1801, Rev. 2, Generic Aging Lessons Learned (GALL) Report, December 2010, U.S.

Nuclear Regulatory Commission Report. [ADAMS Accession Number ML103490041]

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APPENDIX A VALIDATION OF THE RADIATION TRANSPORT MODELS BASED ON NEUTRON DOSIMETRY MEASUREMENTS

NEUTRON DOSIMETRY

Comparisons of measured dosimetry results to both the calculated and least-squares adjusted values for Capsules V, P, R, S, and N are provided in this appendix. The sensor sets 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 [A-1]. One of the main purposes for providing 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 reported in Section 6.2.

A.1.1 Sensor Reaction Rate Determinations

In this section, the results of the evaluations of Capsules V, P, R, S, and N are presented. The capsules designation, locations within the reactor, and time of withdrawal are as follows:

Capsule Azimuthal Withdrawal Irradiation Time Location Time (EFPY)

V 77º End of Cycle 1 1.40 P 247º End of Cycle 5 4.91 R 257º End of Cycle 9 8.63 S 57º End of Cycle 17 18.12 N 237º End of Cycle 31 40.06 T 67º Standby ---

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The passive neutron sensors included in these evaluations are summarized as follows:

Sensor Material Reaction of Interest Capsule V Capsule P Capsule R Capsule S Capsule N Copper Cu-63 (n,) Co-60 X X X X X Iron Fe-54 (n,p) Mn-54 X X X X X Nickel Ni-58 (n,p) Co-58 X(2) X X X X Uranium-238 U-238 (n,f) Cs-137 X X X X X(2)

Neptunium-237 Np-237 (n,f) Cs-137 X X X X X(2)

Cobalt-Aluminum(1) Co-59 (n,) Co-60 X X X X X

Notes:

1. The cobalt-aluminum sensors include both bare and cadmium-covered sensors.

2. These dosimeter sensor results are rejected.

The design of the in-vessel surveillance capsules places the individual neutron sensors at several radial locations within the test specimen array. As a result of the various radial locations, gradient correction factors are applied to the measured reaction rates to index all of the neturon sensor measurements to a common geometric location (the center of the capsule) prior to use in the least-squares adjustment procedure. Pertinent physical and nuclear characteristics of the passive neutron sensors analyzed are listed in Table A-1.

The use of passive monitors does not yield a direct measure of the energy-dependent neutron exposure 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 exposure rate has on the target material over the course of the irradiation period. An accurate assessment of the average neutron exposure 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.

• The neutron energy spectrum at the monitor location.

The radiometric counting of the sensors from Capsule N was carried out by Pace Analytical Services, Inc.

The radiometric counting followed established ASTM procedures. The prevously withdrawn in-vessel Capsules V, P, R, and S were re-evaluated using the current calculational model.

The operating history of the reactor over the irradiation periods was based on the monthly power generation of Prairie Island Unit 1 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 irradiation history for Cycle 1 through Cycle 17 is in [A-2]. The monthly thermal generation data for Cycle 18 through Cycle 31 were provided by Xcel Energy.

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The irradiation history for Cycle 18 through Cycle 31 is summarized in Table A-2.

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 N FY Pj C [1 etj][etd,j]

0 Pref j

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).

N0 = 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.

Pj = Average core power level during irradiation Period j (MW).

Pref = Maximum or reference power level of the reactor (MW).

Cj = 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 (1/sec).

tj = Length of irradiation Period j (sec).

td,j = Decay time following irradiation Period j (sec).

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]/[Pref] accounts for month-by-month variation 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 discussed in Section 6.2, accounts for the change in sensor reaction rates caused by variations in exposure rate level 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, the additional Cj term should be employed. The impact of changing exposure rate levels for constant power operation can be quite significant for sensor sets that have been irradiated for many cycles in a reactor that has transitioned from

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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 exposure rate values are used to compute cycle-dependent values for Cj values at the radial and azimuthal center of the respective capsules at core midplane.

Prior to using the measured reaction rates in the least-squares evaluations of the dosimetry sensor sets, additional corrections were made to the U-238 measurements to account for the presence of 235U 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 U-238 and Np-237 sensor reaction rates to account for gamma-ray-induced fission reactions that occurred over the course of the surveillance capsule irradiations. The correction factors corresponding to the Prairie Island Unit 1 fission sensor reaction rates are summarized as follows:

Correction Capsule V Capsule P Capsule R Capsule S Capsule N U-235 Impurity/Pu Build-in 0.861 0.833 0.742 0.730 N/A U-238 (,f) 0.955 0.960 0.955 0.960 (Fission Net U-238 Correction 0.822 0.799 0.708 0.700 Monitors 0.985 0.986 0.985 0.986 Rejected)

Np-237 (,f)

The correction factors were applied in a multiplicative fashion to the decay-corrected cadmium-covered fission sensor reaction rates.

Results of the sensor reaction rate determinations for the in-vessel Capsules V, P, R, S, and N are given in Table A-3 through Table A-7, where the measured specific activities, decay-corrected saturated specific activities, and computed reaction rates for each sensor are listed.

A.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 fluence rate (E > 1.0 MeV) or dpa/s along with their uncertainties are then easily obtained from the adjusted spectrum. In general, the least-squares method, as applied to 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,

relates a set of measured reaction rates, Ri, to a single neutron spectrum, g, through the multigroup dosimeter reaction cross-sections, ig, each with an uncertainty. The primary objective of the least-squares evaluation is to produce unbiased estimates of the neutron exposure parameters at the location of the measurement.

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For the least-squares evaluation of the Prairie Island Unit 1 dosimetry, the FERRET code [A-3] was employed to combine the results of the plant-specific neutron transport calculations and sensor set reaction rate measurements to determine the best-estimate values of exposure parameters (fluence rate (E > 1.0 MeV) and dpa) and their associated uncertainties.

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 sensor set.

3. The energy-dependent dosimetry reaction cross-sections and associated uncertainties for each sensor contained in the multiple sensor set.

For the Prairie Island Unit 1, the calculated neutron spectrum was obtained from the results of plant-specific neutron transport calculations described in Section 6.2. The sensor reaction rates were derived from the measured specific activities using the procedures described in Section A.1.1. The dosimetry reaction cross-sections and uncertainties were obtained from the SNLRML dosimetry cross-section library [A-4].

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 covar iances. The assignment of the input uncertainties followed the guidance provided in ASTM E944, Standard Guide for Application of Neutron Spectrum Adjustment Methods in Reactor Surveillance [A -5].

The following provides a summary of the uncertainties ass ociated with the least-squares evaluation of the Prairie Island Unit 1 surveillance capsule sensor sets.

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 ensured 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:

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Reaction Uncertainty

63Cu (n,) 60Co 5%

54Fe (n,p) 54Mn 5%

58Ni (n,p) 58Co 5%

59Co (n,) 60Co 5%

238U (n,f) FP 10%

237Np (n,f) FP 10%

These uncertainties are given at the 1 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 recent cross-section evaluations, and they have been tested for accuracy and consistency for least -

squares evaluations. Further, the library has been empirically tested for use in fission spectra determination, as well as in the fluence and energy characterization of 14 MeV neutron sources.

For sensors included in the Prairie Island Unit 1 surveillance program, the following uncertainties in the fission spectrum averaged cross -sections are provided in the SNLRML documentation package:

Reaction Uncertainty Cu-63 (n,) Co-60 4.08-4.16%

Fe-54 (n,p) Mn-54 3.05-3.11%

Ni-58 (n,p) Co-58 4.49-4.56%

Co-59 (n,) Co-60 0.79-3.59%

U-238 (n,f) 0.54-0.64%

Np-237 (n,f) 10.32-10.97%

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 inputs 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).

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Therefore, within the constraints of the assigned uncertainties, the calculated data were treated equally with the measurements.

While the uncertainties associated with the reaction rates were obtained from the measurement procedures and counting benchmarks and the dosimetry cross-section uncertainties were 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:

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 ( specifies the strength of the latter term).

The value of is 1.0 when g = g, and is 0.0 otherwise.

The set of parameters defining the input covariance matrix for the Prairie Island Unit 1 calculated spectra was as follows:

Exposure Rate Normalization Uncertainty (Rn) 15%

Exposure Rate 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 ()

(E > 0.0055 MeV) 0.9 (0.68 eV < E < 0.0055 MeV) 0.5 (E < 0.68 eV) 0.5

Exposure Rate Group Correlation Range ()

(E > 0.0055 MeV) 6 (0.68 eV < E < 0.0055 MeV) 3 (E < 0.68 eV) 2

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A.1.3 Comparisons of Measurements and Calculations

Results of the least-squares evaluations are provided in Table A-8 through Table A-12. In these tables, measured, calculated, and best-estimate values for sensor reaction rates are given. Also provided in these tabulations are ratios of the measured reaction rates to both the calculated and least-squares adjusted reaction rates. These ratios of measured-to-calculated (M/C) and measured-to-best estimate (M/BE) illustrate the consistency of the fit of the calculated neutron energy spectra to the measured reaction rates both before and after adjustment. Additionally, comparisons of the calculated and best-estimate values of neutron fluence rate (E > 1.0 MeV) and iron atom displacement rate are tabulated along with the best -

estimate-to-calculated (BE/C) ratios observed for each of the capsules.

The data comparisons provided in Table A-8 through Table A-12 show that the adjustments to the calculated spectra are relatively small and within the assigned uncertainties for the calculated spectra, measured sensor reaction rates, and dosimetry reaction cross-sections. Further, these results indicate that the use of the least-squares evaluation results in a reduction in the uncertainties associated with the exposure of the surveillance capsules. From Section 6.4, the calculational uncertainty is specified as 13% at the 1 level.

Further comparisons of the measurement results with calculations are given in Table A-13 and Table A-14.

In Table A-13, calculations of individual threshold sensor reaction rates are compare d directly with the corresponding measurements. These threshold reaction rate comparisons provide a good evaluation of the accuracy of the fast neutron portion of the calculated energy spectra. In Table A-14, calculations of fast neutron exposure rates in terms of fast neutron (E > 1.0 MeV) fluence rate and dpa/s are compared with the best-estimate results obtained from the least-squares evaluation of the capsule dosimetry results. These comparisons yield consistent and similar results with all measurement -to-calculation comparisons falling within the 20% limits specified as the acceptance criteria in Regulatory Guide 1.190.

In the case of the direct comparison of the measured and calculated sensor reaction rates, for the individual threshold sensors considered in the least-squares analysis, the M/C comparisons of the fast neutron threshold reactions range from 0.86 to 1.06. The overall average M/C ratio is 0.94 with an associated standard deviation of 9.8%.

In the case of the comparison of the best-estimate and calculated fast neutron exposure parameters, the BE/C comparisons are 0.92 and 0.94 for fast neutron (E > 1.0 MeV) fluence rate and iron atom displacement rate, respectively.

Based on these comparisons, it is concluded that the calculated fast neutron exposures provided in Section 6.2 are valid for use in the assessment of the condition of the materials comprising the beltline region of the Prairie Island Unit 1 reactor pressure vessel.

The uranium and neptunium monitors from Capsule N have been excluded from the least-squares analysis since the reaction rates differed significantly from normalized database averages and the measured results are significantly different than expected values based on calculated cycle flux data.

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Table A-1 Nuclear Parameters Used in the Evaluation of Neutron Sensors

Atomic Target Product Fission 90%

Reaction of Weight Atom Half-life Yield Response Interest Range(1)

(g/g-atom) Fraction (days) (%) (MeV)

Cu-63 (n,) Co-60 63.546 0.6917 1925.28 - 4.53-11.0 Fe-54 (n,p) Mn-54 55.845 0.05845 312.13 - 2.27-7.54 Ni-58 (n,p) Co-58 58.693 0.68077 70.86 - 1.98-7.51 Co-59 (n,) Co-60 58.933 0.0015 1925.28 - Non-threshold U-238 (n,f) Cs-137 238.051 1.00 10975.76 0.0602 1.44-6.69 Np-237 (n,f) Cs-137 237.048 1.00 10975.76 0.0627 0.68-5.61 Note:

1. Energies between which 90% of activity is produced (U-235 fission spectrum) [A-6]

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Table A-2 Monthly Thermal Generation at Prairie Island Unit 1 Cycles 18 through 31

Cycle 18 Cycle 19 Cycle 20 Cycle 21 Thermal Thermal Thermal Thermal Month Generation Month Generation Month Generation Month Generation

[MW -Hr] [MW -Hr] [MW -Hr] [MW -Hr]

3/1/1996 0 12/1/1997 0 6/1/1999 140000 3/1/2001 72600 4/1/1996 1030000 1/1/1998 585000 7/1/1999 1190000 4/1/2001 1230000 5/1/1996 1190000 2/1/1998 1150000 8/1/1999 1230000 5/1/2001 1170000 6/1/1996 1220000 3/1/1998 1110000 9/1/1999 1230000 6/1/2001 1230000 7/1/1996 1140000 4/1/1998 1230000 10/1/1999 1190000 7/1/2001 1180000 8/1/1996 1160000 5/1/1998 1190000 11/1/1999 1230000 8/1/2001 1230000 9/1/1996 1220000 6/1/1998 1230000 12/1/1999 1190000 9/1/2001 17400 10/1/1996 1190000 7/1/1998 634000 1/1/2000 1210000 10/1/2001 748000 11/1/1996 1210000 8/1/1998 1230000 2/1/2000 1230000 11/1/2001 1230000 12/1/1996 1190000 9/1/1998 1220000 3/1/2000 1150000 12/1/2001 1190000 1/1/1997 1230000 10/1/1998 1190000 4/1/2000 1210000 1/1/2002 1220000 2/1/1997 1220000 11/1/1998 1150000 5/1/2000 1190000 2/1/2002 1220000 3/1/1997 1110000 12/1/1998 465000 6/1/2000 1230000 3/1/2002 1110000 4/1/1997 1230000 1/1/1999 1230000 7/1/2000 1190000 4/1/2002 1230000 5/1/1997 1110000 2/1/1999 960000 8/1/2000 1230000 5/1/2002 1160000 6/1/1997 1230000 3/1/1999 1110000 9/1/2000 1230000 6/1/2002 1230000 7/1/1997 589000 4/1/1999 1230000 10/1/2000 1170000 7/1/2002 1190000 8/1/1997 1230000 4/17/1999 654000 11/1/2000 1230000 8/1/2002 1220000 9/1/1997 1190000 12/1/2000 1190000 9/1/2002 1230000 10/1/1997 1190000 1/1/2001 736000 10/1/2002 1190000 10/18/1997 600000 1/19/2001 639000 11/1/2002 1220000 11/15/2002 568000

Total 22479000 Total 17568000 Total 23235000 Total 23086000 EFPS 4.905E+07 EFPS 3.833E+07 EFPS 5.069E+07 EFPS 5.037E+07 EFPD 567.65 EFPD 443.64 EFPD 586.74 EFPD 582.98 EFPY 1.55 EFPY 1.21 EFPY 1.61 EFPY 1.60

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Table A-2 Monthly Thermal Generation at Prairie Island Unit 1 Cycles 18 through 31 (cont.)

Cycle 22 Cycle 23 Cycle 24 Cycle 25 Thermal Thermal Thermal Thermal Month Generation Month Generation Month Generation Month Generation

[MW -Hr] [MW -Hr] [MW -Hr] [MW -Hr]

1/1/2003 963000 11/1/2004 0 6/1/2006 0 4/1/2008 248000 2/1/2003 1220000 12/1/2004 218000 7/1/2006 928000 5/1/2008 1180000 3/1/2003 1110000 1/1/2005 1210000 8/1/2006 1190000 6/1/2008 1210000 4/1/2003 1190000 2/1/2005 1200000 9/1/2006 1220000 7/1/2008 1170000 5/1/2003 949000 3/1/2005 704000 10/1/2006 1180000 8/1/2008 1190000 6/1/2003 1230000 4/1/2005 1090000 11/1/2006 1200000 9/1/2008 1120000 7/1/2003 1180000 5/1/2005 1180000 12/1/2006 1180000 10/1/2008 1180000 8/1/2003 1220000 6/1/2005 1220000 1/1/2007 1220000 11/1/2008 1220000 9/1/2003 1220000 7/1/2005 1180000 2/1/2007 1220000 12/1/2008 1180000 10/1/2003 1190000 8/1/2005 1220000 3/1/2007 1110000 1/1/2009 1220000 11/1/2003 1220000 9/1/2005 1220000 4/1/2007 1220000 2/1/2009 1220000 12/1/2003 1180000 10/1/2005 1150000 5/1/2007 1100000 3/1/2009 1100000 1/1/2004 1220000 11/1/2005 1220000 6/1/2007 684000 4/1/2009 1220000 2/1/2004 1230000 12/1/2005 1180000 7/1/2007 1180000 5/1/2009 1180000 3/1/2004 1100000 1/1/2006 1220000 8/1/2007 1220000 6/1/2009 980000 4/1/2004 1230000 2/1/2006 1100000 9/1/2007 1220000 7/1/2009 1180000 5/1/2004 1190000 3/1/2006 1100000 10/1/2007 1180000 8/1/2009 1220000 6/1/2004 1230000 4/1/2006 1220000 11/1/2007 1190000 9/1/2009 1220000 7/1/2004 1130000 4/28/2006 1010000 12/1/2007 1180000 9/11/2009 413000 8/1/2004 1220000 1/1/2008 1080000 9/1/2004 1130000 2/1/2008 1220000 9/10/2004 290000 2/13/2008 419000

Total 24842000 Total 19642000 Total 23341000 Total 20651000 EFPS 5.420E+07 EFPS 4.286E+07 EFPS 5.093E+07 EFPS 4.506E+07 EFPD 627.32 EFPD 496.01 EFPD 589.42 EFPD 521.49 EFPY 1.72 EFPY 1.36 EFPY 1.61 EFPY 1.43

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Westinghouse Non-Proprietary Class 3 A-12

Table A-2 Monthly Thermal Generation at Prairie Island Unit 1 Cycles 18 through 31 (cont.)

Cycle 26 Cycle 27 Cycle 28 Cycle 29 Thermal Thermal Thermal Thermal Month Generation Month Generation Month Generation Month Generation

[MW -Hr] [MW -Hr] [MW -Hr] [MW -Hr]

11/1/2009 0 6/1/2011 0 12/1/2012 0 12/1/2014 258000 12/1/2009 232000 7/1/2011 713000 1/1/2013 0 1/1/2015 525000 1/1/2010 1220000 8/1/2011 1160000 2/1/2013 1090000 2/1/2015 1040000 2/1/2010 1220000 9/1/2011 1240000 3/1/2013 1130000 3/1/2015 643000 3/1/2010 1100000 10/1/2011 1210000 4/1/2013 1250000 4/1/2015 1240000 4/1/2010 1220000 11/1/2011 1250000 5/1/2013 1210000 5/1/2015 261000 5/1/2010 1180000 12/1/2011 1210000 6/1/2013 1210000 6/1/2015 858000 6/1/2010 1230000 1/1/2012 1230000 7/1/2013 1210000 7/1/2015 1060000 7/1/2010 1190000 2/1/2012 1250000 8/1/2013 1250000 8/1/2015 1250000 8/1/2010 1230000 3/1/2012 1170000 9/1/2013 1250000 9/1/2015 1250000 9/1/2010 1230000 4/1/2012 1250000 10/1/2013 1170000 10/1/2015 1210000 10/1/2010 1190000 5/1/2012 1210000 11/1/2013 1250000 11/1/2015 1240000 11/1/2010 1240000 6/1/2012 1250000 12/1/2013 1210000 12/1/2015 1210000 12/1/2010 1150000 7/1/2012 1180000 1/1/2014 1250000 1/1/2016 1250000 1/1/2011 1250000 8/1/2012 1250000 2/1/2014 1250000 2/1/2016 1250000 2/1/2011 1250000 9/1/2012 957000 3/1/2014 1130000 3/1/2016 1170000 3/1/2011 1130000 10/1/2012 1210000 4/1/2014 1250000 4/1/2016 1210000 4/1/2011 1250000 10/23/2012 879000 5/1/2014 1150000 5/1/2016 1210000 4/29/2011 1140000 6/1/2014 1250000 6/1/2016 1250000 7/1/2014 1190000 7/1/2016 1210000 8/1/2014 1250000 8/1/2016 1250000 9/1/2014 1250000 9/1/2016 1250000 10/1/2014 1190000 10/1/2016 1210000 10/8/2014 259000 10/15/2016 556000 Total 20652000 Total 19619000 Total 25649000 Total 24861000 EFPS 4.479E+07 EFPS 4.212E+07 EFPS 5.506E+07 EFPS 5.337E+07 EFPD 518.37 EFPD 487.45 EFPD 637.27 EFPD 617.70 EFPY 1.42 EFPY 1.33 EFPY 1.74 EFPY 1.69

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Westinghouse Non-Proprietary Class 3 A-13

Table A-2 Monthly Thermal Generation at Prairie Island Unit 1 Cycles 18 through 31 (cont.)

Cycle 30 Cycle 31 Thermal Thermal Month Generation Month Generation

[MW -Hr] [MW -Hr]

12/1/2016 361000 11/1/2018 137000 1/1/2017 1250000 12/1/2018 1210000 2/1/2017 1250000 1/1/2019 1250000 3/1/2017 1130000 2/1/2019 1250000 4/1/2017 1220000 3/1/2019 1130000 5/1/2017 1090000 4/1/2019 1250000 6/1/2017 1040000 5/1/2019 1150000 7/1/2017 1130000 6/1/2019 1250000 8/1/2017 1250000 7/1/2019 1210000 9/1/2017 1250000 8/1/2019 1250000 10/1/2017 1170000 9/1/2019 1250000 11/1/2017 1250000 10/1/2019 1200000 12/1/2017 1210000 11/1/2019 1250000 1/1/2018 1250000 12/1/2019 1210000 2/1/2018 1250000 1/1/2020 1250000 3/1/2018 1130000 2/1/2020 1250000 4/1/2018 1200000 3/1/2020 1170000 5/1/2018 1210000 4/1/2020 1240000 6/1/2018 1250000 5/1/2020 1190000 7/1/2018 1210000 6/1/2020 1250000 8/1/2018 1250000 7/1/2020 1210000 9/1/2018 1230000 8/1/2020 1250000 9/21/2018 710000 9/1/2020 1050000 9/18/2020 649000 Total 26291000 Total 27506000 EFPS 5.644E+07 EFPS 5.905E+07 EFPD 653.23 EFPD 683.41 EFPY 1.79 EFPY 1.87

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Westinghouse Non-Proprietary Class 3 A-14

Table A-3 Measured Sensor Activities and Reaction Rates for Surveillance Capsule V

Corrected Average Average Measured Saturated Reaction Reaction Reaction Activity Activity Rate Rate Rate Sensor Location (dps/g) (dps/g) (rps/atom) (rps/atom) (rps/atom)

Cu Top Middle 5.61E+04 4.09E+05 6.25E-17 6.657E-17 6.657E-17 Cu Bottom Middle 6.35E+04 4.63E+05 7.07E-17 Iron Top 2.23E+06 4.79E+06 7.60E-15 Iron Top Middle 2.05E+06 4.41E+06 6.99E-15 Iron Middle 2.09E+06 4.49E+06 7.13E-15 7.404E-15 7.404E-15 Iron Bottom Middle 2.17E+06 4.66E+06 7.40E-15 Iron Bottom 2.32E+06 4.99E+06 7.91E-15 Nickel Middle Rejected per [A-2]

U-238 (Cd)2 Middle 2.44E+05 7.79E+06 5.12E-14 5.115E-14 4.203E-14 Np-237 (Cd)2 Middle 1.88E+06 6.00E+07 3.77E-13 3.768E-13 3.712E-13 Bare Co Top 1.90E+07 1.13E+08 7.35E-12 7.984E-12 7.984E-12 Bare Co Bottom 2.23E+07 1.32E+08 8.62E-12 Co (Cd) Top 8.25E+06 6.16E+07 4.02E-12 3.877E-12 3.877E-12 Co (Cd) Bottom 7.68E+06 5.73E+07 3.74E-12 Notes:

1. Measured activity is decay corrected to 8/23/1976.

2. Note that the fission monitors were decay corrected to 8/25/1976 but they were analyzed using the 8/23/1976 date. This two day difference does not have a significant impact on the results since the half-life of Cs-137 is 10976 days.

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Westinghouse Non-Proprietary Class 3 A-15

Table A-4 Measured Sensor Activities and Reaction Rates for Surveillance Capsule P

Corrected Average Average Measured Saturated Reaction Reaction Reaction Activity Activity Rate Rate Rate Sensor Location (dps/g) (dps/g) (rps/atom) (rps/atom) (rps/atom)

Cu Top Middle 1.27E+05 3.89E+05 5.94E-17 5.730E-17 5.730E-17 Cu Bottom Middle 1.18E+05 3.62E+05 5.52E-17 Iron Top 1.08E+06 3.65E+06 5.79E-15 Iron Top Middle 8.41E+05 2.84E+06 4.51E-15 Iron Middle Wire Not Received 5.419E-15 5.419E-15 Iron Bottom Middle 1.00E+06 3.38E+06 5.36E-15 Iron Bottom 1.12E+06 3.79E+06 6.01E-15 Nickel Middle 3.77E+05 5.61E+07 8.03E-15 8.033E-15 8.033E-15 U-238 (Cd) Middle 5.55E+05 5.40E+06 3.55E-14 3.548E-14 2.836E-14 Np-237 (Cd) Middle 4.28E+06 4.16E+07 2.61E-13 2.613E-13 2.575E-13 Bare Co Top 2.64E+07 6.40E+07 4.17E-12 4.576E-12 4.576E-12 Bare Co Bottom 3.15E+07 7.63E+07 4.98E-12 Co (Cd) Top 9.34E+06 2.86E+07 1.87E-12 1.923E-12 1.923E-12 Co (Cd) Bottom 9.92E+06 3.04E+07 1.98E-12 Note:

1. Measured activity is decay corrected to 12/14/1981.

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Westinghouse Non-Proprietary Class 3 A-16

Table A-5 Measured Sensor Activities and Reaction Rates for Surveillance Capsule R

Corrected Average Average Measured Saturated Reaction Reaction Reaction Activity Activity Rate Rate Rate Sensor Location (dps/g) (dps/g) (rps/atom) (rps/atom) (rps/atom)

Cu Top Middle 2.42E+05 4.86E+05 7.42E-17 7.539E-17 7.539E-17 Cu Bottom Middle 2.50E+05 5.02E+05 7.66E-17 Iron Top 2.74E+06 5.65E+06 8.97E-15 Iron Top Middle 2.47E+06 5.10E+06 8.08E-15 Iron Middle 2.56E+06 5.28E+06 8.38E-15 8.508E-15 8.508E-15 Iron Bottom Middle 2.56E+06 5.28E+06 8.38E-15 Iron Bottom 2.67E+06 5.51E+06 8.74E-15 Nickel Middle 3.84E+06 8.26E+07 1.18E-14 1.182E-14 1.182E-14 U-238 (Cd) Middle 1.96E+06 1.12E+07 7.38E-14 7.379E-14 5.225E-14 Np-237 (Cd) Middle 1.41E+07 8.10E+07 5.09E-13 5.088E-13 5.012E-13 Bare Co Top 7.41E+07 1.21E+08 7.88E-12 Bare Co Bottom 8.13E+07 1.33E+08 8.65E-12 8.406E-12 8.406E-12 Bare Co Bottom 8.16E+07 1.33E+08 8.68E-12 Co (Cd) Top 2.96E+07 6.08E+07 3.97E-12 4.001E-12 4.001E-12 Co (Cd) Bottom 3.01E+07 6.18E+07 4.04E-12 Note:

1. Measured activity is decay corrected to 10/11/1985.

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Westinghouse Non-Proprietary Class 3 A-17

Table A-6 Measured Sensor Activities and Reaction Rates for Surveillance Capsule S

Corrected Average Average Measured Saturated Reaction Reaction Reaction Activity Activity Rate Rate Rate Sensor Location (dps/g) (dps/g) (rps/atom) (rps/atom) (rps/atom)

Cu Top Middle 1.70E+05 2.93E+05 4.46E-17 4.634E-17 4.634E-17 Cu Bottom Middle 1.83E+05 3.15E+05 4.81E-17 Iron Top 1.30E+06 2.98E+06 4.72E-15 Iron Top Middle 1.13E+06 2.59E+06 4.11E-15 Iron Middle 1.17E+06 2.68E+06 4.25E-15 4.418E-15 4.418E-15 Iron Bottom Middle 1.20E+06 2.75E+06 4.36E-15 Iron Bottom 1.28E+06 2.93E+06 4.65E-15 Nickel Middle 2.28E+06 4.31E+07 6.17E-15 6.171E-15 6.171E-15 U-238 (Cd) Middle 1.67E+06 5.20E+06 3.42E-14 3.417E-14 2.393E-14 Np-237 (Cd) Middle 1.13E+07 3.52E+07 2.21E-13 2.210E-13 2.178E-13 Bare Co Top 3.87E+07 5.27E+07 3.44E-12 3.444E-12 3.444E-12 Bare Co Bottom 3.88E+07 5.28E+07 3.45E-12 Co (Cd) Top 1.41E+07 2.42E+07 1.58E-12 1.592E-12 1.592E-12 Co (Cd) Bottom 1.43E+07 2.46E+07 1.60E-12 Note:

1. Measured activity is decay corrected to 9/18/1996.

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Westinghouse Non-Proprietary Class 3 A-18

Table A-7 Measured Sensor Activities and Reaction Rates for Surveillance Capsule N

Corrected Average Average Measured Saturated Reaction Reaction Reaction Activity Activity Rate Rate Rate Sensor Location (dps/g) (dps/g) (rps/atom) (rps/atom) (rps/atom)

Cu Top Middle 1.46E+05 2.24E+05 3.42E-17 3.557E-17 3.557E-17 Cu Bottom Middle 1.58E+05 2.42E+05 3.70E-17 Iron Top 1.55E+06 2.44E+06 3.88E-15 Iron Top Middle 1.33E+06 2.10E+06 3.33E-15 Iron Middle 1.42E+06 2.24E+06 3.55E-15 3.626E-15 3.626E-15 Iron Bottom Middle 1.45E+06 2.29E+06 3.63E-15 Iron Bottom 1.50E+06 2.37E+06 3.75E-15 Nickel Middle 6.45E+06 3.99E+07 5.71E-15 5.706E-15 5.706E-15 U-238 (Cd) Middle REJECTED Np-237 (Cd) Middle Bare Co Top 3.53E+07 4.28E+07 2.79E-12 2.568E-12 2.568E-12 Bare Co Bottom 2.96E+07 3.59E+07 2.34E-12 Co (Cd) Top 1.39E+07 2.13E+07 1.39E-12 1.378E-12 1.378E-12 Co (Cd) Bottom 1.37E+07 2.10E+07 1.37E-12 Note:

1. Measured activity is decay corrected to 2/15/2021.

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Westinghouse Non-Proprietary Class 3 A-19

Table A-8 Least-Squares Evaluation of Dosimetry in Surveillance Capsule V (13° Position, Core Midplane, Irradiated During Cycle 1)

2/DOF = 0.096 Reaction Rate (rps/atom)

Best-Measured Calculated Estimate Reaction (M) (C) (BE) M/C M/BE BE/C 63Cu (n,) 60Co 6.66E-17 7.10E-17 6.54E-17 0.94 1.02 0.92

54Fe (n,p) 54Mn 7.40E-15 8.58E-15 7.63E-15 0.86 0.97 0.89

58Ni (n,p) 58Co Rejected

238U(Cd) (n,f) 137Cs 4.20E-14 4.56E-14 4.08E-14 0.92 1.03 0.90

237Np(Cd) (n,f) 137Cs 3.71E-13 4.01E-13 3.66E-13 0.93 1.01 0.91

59Co (n,) 60Co 7.98E-12 1.03E-11 8.01E-12 0.78 1.00 0.78

59Co(Cd) (n,) 60Co 3.88E-12 3.91E-12 3.85E-12 0.99 1.01 0.99 Average of Fast Energy Threshold Reactions 0.91 1.01 0.91 Percent Standard Deviation 3.9 2.6 1.4 Best-Calculated Estimate Integral Quantity (C) % Unc. (BE) % Unc. BE/C Neutron Fluence Rate (E > 1.0 MeV) 1.39E+11 13 1.25E+11 6 0.90 (n/cm2-s)

Displacement Rate 2.49E-10 13 2.27E-10 7 0.91 (dpa/s)

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Westinghouse Non-Proprietary Class 3 A-20

Table A-9 Least-Squares Evaluation of Dosimetry in Surveillance Capsule P (23° Position, Core Midplane, Irradiated During Cycles 1 through 5)

2/DOF = 0.242 Reaction Rate (rps/atom)

Best-Measured Calculated Estimate Reaction (M) (C) (BE) M/C M/BE BE/C 63Cu (n,) 60Co 5.73E-17 5.62E-17 5.57E-17 1.02 1.03 0.99

54Fe (n,p) 54Mn 5.42E-15 6.03E-15 5.70E-15 0.90 0.95 0.95

58Ni (n,p) 58Co 8.03E-15 8.28E-15 7.98E-15 0.97 1.01 0.96

238U(Cd) (n,f) 137Cs 2.84E-14 2.92E-14 2.81E-14 0.97 1.01 0.96

237Np(Cd) (n,f) 137Cs 2.57E-13 2.33E-13 2.43E-13 1.10 1.06 1.04

59Co (n,) 60Co 4.58E-12 5.14E-12 4.59E-12 0.89 1.00 0.89

59Co(Cd) (n,) 60Co 1.92E-12 2.00E-12 1.92E-12 0.96 1.00 0.96 Average of Fast Energy Threshold Reactions 0.99 1.01 0.98 Percent Standard Deviation 7.5 4.0 3.7 Best-Calculated Estimate Integral Quantity (C) % Unc. (BE) % Unc. BE/C Neutron Fluence Rate (E > 1.0 MeV) 8.46E+10 13 8.22E+10 6 0.97 (n/cm2-s)

Displacement Rate 1.46E-10 13 1.44E-10 7 0.99 (dpa/s)

WCAP-18660-NP November 2021 Revision 0

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Westinghouse Non-Proprietary Class 3 A-21

Table A-10 Least-Squares Evaluation of Dosimetry in Surveillance Capsule R (13° Position, Core Midplane, Irradiated During Cycles 1 through 9)

2/DOF = 0.316 Reaction Rate (rps/atom)

Best-Measured Calculated Estimate Reaction (M) (C) (BE) M/C M/BE BE/C 63Cu (n,) 60Co 7.54E-17 7.61E-17 7.37E-17 0.99 1.02 0.97

54Fe (n,p) 54Mn 8.51E-15 9.18E-15 8.74E-15 0.93 0.97 0.95

58Ni (n,p) 58Co 1.18E-14 1.28E-14 1.22E-14 0.92 0.97 0.95

238U(Cd) (n,f) 137Cs 5.22E-14 4.88E-14 4.80E-14 1.07 1.09 0.98

237Np(Cd) (n,f) 137Cs 5.01E-13 4.29E-13 4.67E-13 1.17 1.08 1.09

59Co (n,) 60Co 8.41E-12 1.10E-11 8.48E-12 0.76 0.99 0.77

59Co(Cd) (n,) 60Co 4.00E-12 4.19E-12 3.99E-12 0.95 1.00 0.95 Average of Fast Energy Threshold Reactions 1.02 1.03 0.99 Percent Standard Deviation 10 5.6 5.9 Best-Calculated Estimate Integral Quantity (C) % Unc. (BE) % Unc. BE/C Neutron Fluence Rate (E > 1.0 MeV) 1.48E+11 13 1.48E+11 6 1.00 (n/cm2-s)

Displacement Rate 2.67E-10 13 2.71E-10 7 1.02 (dpa/s)

WCAP-18660-NP November 2021 Revision 0

• • • This record was final approved on 12/1/2021 8:24:23 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 A-22

Table A-11 Least-Squares Evaluation of Dosimetry in Surveillance Capsule S (33° Position, Core Midplane, Irradiated During Cycles 1 through 17)

2/DOF = 0.293 Reaction Rate (rps/atom)

Best -

Measured Calculated Estimate Reaction (M) (C) (BE) M/C M/BE BE/C 63Cu (n,) 60Co 4.63E-17 4.79E-17 4.46E-17 0.97 1.04 0.93

54Fe (n,p) 54Mn 4.42E-15 5.27E-15 4.59E-15 0.84 0.96 0.87

58Ni (n,p) 58Co 6.17E-15 7.28E-15 6.34E-15 0.85 0.97 0.87

238U(Cd) (n,f) 137Cs 2.39E-14 2.62E-14 2.31E-14 0.91 1.04 0.88

237Np(Cd) (n,f) 137Cs 2.18E-13 2.14E-13 2.05E-13 1.02 1.06 0.96

59Co (n,) 60Co 3.44E-12 4.70E-12 3.47E-12 0.73 0.99 0.74

59Co(Cd) (n,) 60Co 1.59E-12 1.86E-12 1.59E-12 0.86 1.00 0.86 Average of Fast Energy Threshold Reactions 0.92 1.01 0.90 Percent Standard Deviation 8.4 4.5 4.5 Best-Calculated Estimate Integral Quantity (C) % Unc. (BE) % Unc. BE/C Neutron Fluence Rate (E > 1.0 MeV) 7.69E+10 13 6.82E+10 6 0.89 (n/cm2-s)

Displacement Rate 1.33E-10 13 1.21E-10 7 0.91 (dpa/s)

WCAP-18660-NP November 2021 Revision 0

• • • This record was final approved on 12/1/2021 8:24:23 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 A-23

Table A-12 Least-Squares Evaluation of Dosimetry in Surveillance Capsule N (33° Position, Core Midplane, Irradiated During Cycles 1 through 31)

2/DOF = 0.394 Reaction Rate (rps/atom)

Best-Measured Calculated Estimate Reaction (M) (C) (BE) M/C M/BE BE/C 63Cu (n,) 60Co 3.56E-17 4.28E-17 3.56E-17 0.83 1.00 0.83

54Fe (n,p) 54Mn 3.63E-15 4.64E-15 3.82E-15 0.78 0.95 0.82

58Ni (n,p) 58Co 5.71E-15 6.40E-15 5.46E-15 0.89 1.04 0.85

238U(Cd) (n,f) 137Cs 237 137 Rejected Np(Cd) (n,f) Cs 59Co (n,) 60Co 2.57E-12 4.07E-12 2.60E-12 0.63 0.99 0.64

59Co(Cd) (n,) 60Co 1.38E-12 1.60E-12 1.37E-12 0.86 1.01 0.85 Average of Fast Energy Threshold Reactions 0.83 1.00 0.83 Percent Standard Deviation 6.6 4.5 1.8 Best-Calculated Estimate Integral Quantity (C) % Unc. (BE) % Unc. BE/C Neutron Fluence Rate (E > 1.0 MeV) 6.70E+10 13 5.71E+10 8 0.85 (n/cm2-s)

Displacement Rate 1.16E-10 13 9.93E-11 9 0.86 (dpa/s)

WCAP-18660-NP November 2021 Revision 0

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Westinghouse Non-Proprietary Class 3 A-24

Table A-13 Measured-to-Calculated (M/C) Reaction Rates - In-Vessel Capsules

Capsule % Std.

Reaction V P R S N Average Dev.

63Cu (n,) 60Co 0.94 1.02 0.99 0.97 0.83 0.95 7.7

54Fe (n,p) 54Mn 0.86 0.90 0.93 0.84 0.78 0.86 6.7

58Ni (n,p) 58Co - 0.97 0.92 0.85 0.89 0.91 5.6

238U(Cd) (n,f) 137Cs 0.92 0.97 1.07 0.91 - 0.97 7.6

237Np(Cd) (n,f) 137Cs 0.93 1.10 1.17 1.02 - 1.06 9.8 Average of M/C Results 0.94 9.8

Table A-14 Best-Estimate-to-Calculated (BE/C) Exposure Rates - In-Vessel Capsules

Neutron Fluence (E > 1.0 MeV)

Rate Iron Atom Displacement Rate Capsule BE/C BE/C

V 0.90 0.91 P 0.97 0.99 R 1.00 1.02 S 0.89 0.91 N 0.85 0.86 Average 0.92 0.94

% Std. Dev. 6.7 7.0

WCAP-18660-NP November 2021 Revision 0

• • • This record was final approved on 12/1/2021 8:24:23 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 A-25

REFERENCES

A-1 U.S. Nuclear Regulatory Commission Regulatory Guide 1.190, Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence, March 2001. [ADAMS Accession Number ML010890301]

A-2 Westinghouse Report WCAP-14779, Rev. 2, Analysis of Capsule S from the Northern States Power Company Prairie Island Unit 1 Reactor Vessel Radiation Surveillance Program, February 1998.

A-3 A. Schmittroth, FERRET Data Analysis Core, HEDL -TME 79-40, Hanford Engineering Development Laboratory, Richland, WA, September 1979.

A-4 RSICC Data Library Collection DLC-178, SNLRML Recommended Dosimetry Cross-Section Compendium, July 1994.

A-5 ASTM Standard E944-19, Standard Guide for Application of Neutron Spectrum Adjustment Methods in Reactor Surveillance, 2019.

A-6 ASTM Standard E844-18, Standard Guide for Sensor Set Design and Irradiation for Reactor Surveillance, 2018.

WCAP-18660-NP November 2021 Revision 0

• • • This record was final approved on 12/1/2021 8:24:23 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 B-1

APPENDIX B LOAD-TIME RECORDS FOR CHARPY SPECIMEN TESTS FROM CAPSULE N

• NXX denotes Intermediate Shell Forging C, Tangential Orientation

• SXX denotes Intermediate Shell Forging C, Axial Orientation

• WXX denotes Surveillance Weld material

• HXX denotes Heat-Affected Zone (HAZ) material

• RXX denotes Correlation Monitor material

Note that the instrumented Charpy data is not required per ASTM Standards E185-82 or E23-18.

WCAP-18660-NP November 2021 Revision 0

• • • This record was final approved on 12/1/2021 8:24:23 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 B-2

N49: Tested at 77°F

N54: Tested at 90°F

WCAP-18660-NP November 2021 Revision 0

• • • This record was final approved on 12/1/2021 8:24:23 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 B-3

N51: Tested at 110°F

N55: Tested at 120°F

WCAP-18660-NP November 2021 Revision 0

• • • This record was final approved on 12/1/2021 8:24:23 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 B-4

N60: Tested at 130°F

N50: Tested at 140°F

WCAP-18660-NP November 2021 Revision 0

• • • This record was final approved on 12/1/2021 8:24:23 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 B-5

N53: Tested at 150°F

N58: Tested at 160°F

WCAP-18660-NP November 2021 Revision 0

• • • This record was final approved on 12/1/2021 8:24:23 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 B-6

N56: Tested at 180°F

N52: Tested at 200°F

WCAP-18660-NP November 2021 Revision 0

• • • This record was final approved on 12/1/2021 8:24:23 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 B-7

N57: Tested at 250°F

N59: Tested at 300°F

WCAP-18660-NP November 2021 Revision 0

• • • This record was final approved on 12/1/2021 8:24:23 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 B-8

S60: Tested at 60°F

S54: Tested at 77°F

WCAP-18660-NP November 2021 Revision 0

• • • This record was final approved on 12/1/2021 8:24:23 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 B-9

S57: Tested at 90°F

S55: Tested at 100°F

WCAP-18660-NP November 2021 Revision 0

• • • This record was final approved on 12/1/2021 8:24:23 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 B-10

S59: Tested at 110°F

S58: Tested at 130°F

WCAP-18660-NP November 2021 Revision 0

• • • This record was final approved on 12/1/2021 8:24:23 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 B-11

S49: Tested at 160°F

S56: Tested at 170°F

WCAP-18660-NP November 2021 Revision 0

• • • This record was final approved on 12/1/2021 8:24:23 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 B-12

S51: Tested at 180°F

S53: Tested at 200°F

WCAP-18660-NP November 2021 Revision 0

• • • This record was final approved on 12/1/2021 8:24:23 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 B-13

S52: Tested at 250°F

S50: Tested at 300°F

WCAP-18660-NP November 2021 Revision 0

• • • This record was final approved on 12/1/2021 8:24:23 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 B-14

W34 : Tested at 110°F

W40 : Tested at 125°F

WCAP-18660-NP November 2021 Revision 0

• • • This record was final approved on 12/1/2021 8:24:23 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 B-15

W36 : Tested at 150°F

W38 : Tested at 175°F

WCAP-18660-NP November 2021 Revision 0

• • • This record was final approved on 12/1/2021 8:24:23 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 B-16

W37 : Tested at 190°F

W39 : Tested at 200°F

WCAP-18660-NP November 2021 Revision 0

• • • This record was final approved on 12/1/2021 8:24:23 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 B-17

W 35: Tested at 250°F

W 33: Tested at 300°F

WCAP-18660-NP November 2021 Revision 0

• • • This record was final approved on 12/1/2021 8:24:23 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 B-18

H35: Tested at -50°F

H40: Tested at -25°F

WCAP-18660-NP November 2021 Revision 0

• • • This record was final approved on 12/1/2021 8:24:23 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 B-19

H37 Tested at 10°F

H33: Tested at 50°F

WCAP-18660-NP November 2021 Revision 0

• • • This record was final approved on 12/1/2021 8:24:23 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 B-20

H36: Tested at 120°F

H34: Tested at 200°F

WCAP-18660-NP November 2021 Revision 0

• • • This record was final approved on 12/1/2021 8:24:23 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 B-21

H38: Tested at 250°F

H39: Tested at 300°F

WCAP-18660-NP November 2021 Revision 0

• • • This record was final approved on 12/1/2021 8:24:23 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 B-22

R33: Tested at 200°F

R39: Tested at 225°F

WCAP-18660-NP November 2021 Revision 0

• • • This record was final approved on 12/1/2021 8:24:23 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 B-23

R34: Tested at 250°F

R40: Tested at 275°F

WCAP-18660-NP November 2021 Revision 0

• • • This record was final approved on 12/1/2021 8:24:23 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 B-24

R38: Tested at 300°F

R37: Tested at 325°F

WCAP-18660-NP November 2021 Revision 0

• • • This record was final approved on 12/1/2021 8:24:23 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 B-25

R35: Tested at 325°F

R36: Tested at 350°F

WCAP-18660-NP November 2021 Revision 0

• • • This record was final approved on 12/1/2021 8:24:23 AM. (This statement was added by the PRIME system upon its validation)