ML13311A055

Fluence Extrapolation in Support of NMP2 P-T Cure Update, MPM-913991, September 30, 2013, Attachment 3

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ATTACHMENT 3 FLUENCE EXTRAPOLATION IN SUPPORT OF NMP2 P-T CURVE UPDATE, MPM-913991, SEPTEMBER 30, 2013 Nine Mile Point Nuclear Station, LLC November 4, 2013

Report Number MPM-913991 Fluence Extrapolation in Support of NMP2 P-T Curve Update

,a&

TESTINIL September, 2013

MPM Technical Report Fluence Extrapolation in Support of NMP2 P-T Curve Update MPM Report Number MPM-913991 Final Report September, 2013 MP Machinery and Testing, LLC 2161 Sandy Drive State College, PA 16803 USA MP Machinery and Testing, LLC

• 2161 Sandy Drive, State College, Pennsylvania 16803 - USA 814-234-8860

DISCLAIMER OF WARRANTIES AND LIMITATION OF LIABILITIES THIS DOCUMENT WAS PREPARED BY MP MACHINERY AND TESTING, LLC AS AN ACCOUNT OF WORK SPONSORED OR COSPONSORED BY CONSTELLATION ENERGY GROUP. NEITHER MPM, ANY COSPONSOR, NOR ANY PERSON ACTING ON BEHALF OF ANY OF THEM:

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Nuclear Quality Assurance Certification This document certifies that MPM has performed all work under Constellation Energy Group Purchase Order Number 7734669 in accordance with the requirements of the Purchase Order. All work has been performed under the MPM Nuclear Quality Assurance Program.

M. P. Manahan, Sr.

President 10/22/2013 Date J. Nemet QA Manager 10/22/2013 Date ii

EXECUTIVE

SUMMARY

This work was undertaken to calculate projected best estimate neutron fluence data in support of the Nine Mile Point Unit 2 (NMP2) Pressure-Temperature (P-T) curve update.

Previous neutron transport analyses by MPM reported fluence projections to 22 EFPY. The current work uses more recent transport calculations to project the fluence to 32 EFPY for the Pressure and Temperature Limits Report (PTLR).

Summary of Vessel Fluence Extrapolations P-T curves require the peak fluence at the vessel wetted surface to conservatively calculate the limiting vessel material ARTNDT. The key vessel peak fluence results are summarized in Table ES-1. This table includes the calculated fluence attenuation through the vessel as well as the RG 1.99 (Rev 2) dpa-based attenuation. The fluence was evaluated throughout the vessel to indicate the areas likely to exceed a fluence of 1.0E 17 n/cm 2 . It was found that with operation to 54 EFPY, this fluence was exceeded from about 7 inches below the core region to about 10 inches above the core region. Fluence to selected reactor nozzles was calculated and only the nozzles near the top of the core (N6 and N12) were found to exceed 1.OE+17 n/cm 2 in the future.

Uncertaintyand Bias Analysis Fluence values for the vessel and shroud in the beltline region (except for the very top and bottom of the core) are estimated to have uncertainties of about 15% and 16%, respectively.

These uncertainties are within the value of +/- 20% specified by RG 1.190. Moreover, benchmarking of the MPM calculational methodology is provided by comparisons of previous calculations using the same methodology with NMPI capsule and shroud measurements (boat samples were cut from the shroud), NMP2 capsule measurements, River Bend Station (RBS) capsule measurement, and Grand Gulf Nuclear Station (GGNS) cycle 1 dosimetry measurements. It is concluded that the calculations of vessel and capsule fluence meet all the requirements of RG 1.190.

iii

Table ES-1 Projected NMP2 Maximum Vessel Fluence and Fluence with dpa Attenuation.

Calculated Attenuation using Attenuation usinm Position Fluence (E > 1 MeV) Calculated dpa RG1.99( Rev 2)

(n/cm2 ) (n/cm2 ) (n/cm 2 )

At 32 EFPY Vessel Clad IR 8.20E+ 17 8.20E+ 17 8.20E+ 17 Vessel IR 8.07E+ 17 8.05E+ 17 7.84E+ 17 1/4 T 5.42E+17 5.61E+17 5.33E+17 3/4 T 1.81E+17 2.22E+17 2.46E+17 At 54 EFPY Vessel Clad IR 1.53E+18 1.53E+18 1.53E+18 Vessel IR 1.50E+18 1.50E+18 1.46E+ 18 1/4 T 1.O1E+18 1.05E+ 18 9.92E+17 3/4 T 3.36E+17 4.14E+17 4.58E+17

1. Calculated fluence at the vessel inner wetted surface (clad IR) times the ratio of calculated dpa at the vessel interior points to the dpa at the inner wetted surface.

2. Calculated fluence at the vessel inner wetted surface (clad IR) times exp(-0.24*x), where x is the distance into the vessel in inches.

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CONTENTS 1 INTRO DUCTIO N ........................................................................................................... 1-1 1.1 Chapter 1 References ................................................................................................. 1-2 2 VESSEL FLUENCE EXTRAPO LATIO N ................................................................................ 2-1 2.5 Chapter 2 References ................................................................................................. 2-2 3 UNCERTA INTY A ND BIAS A NA LYSIS ................................................................................. 3-1 3.1 Uncertainty Assum ptions ............................................................................................. 3-1 3.2 Uncertainty Evaluation ................................................................................................. 3-3 3.4 Chapter 3 References ................................................................................................. 3-4 4 SUM MA RY A ND CO NCLUSIO NS ......................................................................................... 4-1 5 NO MENCLATURE ................................................................................................................. 5-1 V

LIST OF FIGURES Figure 2-1 NMP2 Beltline Weld Seam and Plate Locations .................................................... 2-12 vi

LIST OF TABLES Table 1-1 NMP2 Cycle Operating History (Reference [1-6] ...................................................... 1-3 Table 2-1 Axial Variation of Maximum Fluence at the Vessel Wetted Surface ......................... 2-3 Table 2-2 Relative Exposure Through the Vessel at the Maximum Fluence Point Averaged over Cycles 1 to 10 ............................................................................................ 2-6 Table 2-3 Calculated Maximum Vessel Exposure at the End of Cycle 14 (21.49 EFPY) and Projected to 32 and 54 EFPY ...................................................................................... 2-7 Table 2-4 Calculated Maximum Vessel Lower Course Exposure at the End of Cycle 14 (21.49 EFPY) and Projected to 32 and 54 EFPY ............................................................... 2-8 Table 2-5 Azimuthal Locations of Vessel Welds that Extend into the Beltline .......................... 2-9 Table 2-6 Calculated Maximum Vessel Beltline Weld Exposures at the End of Cycle 14 (21.49 EFPY) and Projected to 32 and 54 EFPY ............................................................... 2-9 Table 2-7 Calculated Peak Vessel Fluence and Fluence Determined Using dpa Atte n uatio n ....................................................................................................................... 2 -10 Table 2-8 NMP2 Nozzle Locations Near the Core Region ..................................................... 2-11 Table 2-9 NMP2 Nozzle Maximum Fluence Values ............................................................... 2-11 Table 3-1 Nine Mile Point Unit 2 Shroud, Capsule, and Vessel Fluence Calculational Un ce rta inty ......................................................................................................................... 3 -6 vii

I INTRODUCTION This work was undertaken to calculate the best estimate neutron fluence to the Nine Mile Point Unit 2 (NMP2) pressure vessel in support of the P-T curve update effort. The neutron fluence calculations in the active fuel region were carried out using a synthesis of two dimensional neutron transport calculations, including plant-specific R-0 and R-Z calculations.

Further details on the analytical model are contained in references [1-1 ] and [1-2]. This report uses the outputs from the References [1-1, 1-21 to provide projections of the peak fluence to the vessel wetted surface at 32 and 54 effective full power years (EFPY) using the operating history as of September, 2013 (Table 1-1). The methodology used in the evaluation of the NMP2 vessel fluence is completely consistent with that used in the evaluation performed for the NMP2 3-degree surveillance capsule [1-3]. The NMP2 surveillance capsule dosimetry data was used, together with that from other BWR plants to benchmark the methodology [1-4] for fluence evaluation for NMP2. This benchmark evaluation establishes that the methodology meets all the requirements of Regulatory Guide 1.190 [1-5] and that no bias need be applied to the calculational results.

The present fluence evaluation uses a detailed cycle-by-cycle evaluation of fluence throughout the reactor system for the first 10 cycles of operation. The fluence for cycles 11 through 13 is estimated using the final cycle 10 results in [1-2]. The effect of changing the depleted uranium blanket height from 12 inches to 6 inches starting with fuel inserted in cycle 11 is also accounted for in the evaluation. Starting with cycle 14, NMP2 is operating at an uprated power level 15% higher. A calculation of a typical cycle for this extended power uprate was documented in [1-2] and this is used for evaluation of the cycle 14 fluence as well as the extrapolation to 32 and 54 EFPY of operation.

1-1

Introduction 1.1 Chapter 1 References

[1-1] "Neutron Transport Analysis for Nine Mile Point Unit 2", Report MPM-1205889, MPM Technologies, Inc., 2161 Sandy Drive, State College, PA 16803-2283, January, 2006.

[1-2] "Neutron Transport Analysis for Nine Mile Point Unit 2 Uprate", Report MPM-1 108779, MPM Technologies, Inc., 2161 Sandy Drive, State College, PA 16803-2283, December, 2008.

[1-3] "Nine Mile Point Unit 2 3-Degree Pressure Vessel Surveillance Capsule Report", Report MPM-1200676, MPM Technologies, Inc., 2161 Sandy Drive, State College, PA 16803-2283, December, 2000.

[1-4] "Benchmarking of Nine Mile Point Unit I and Unit 2 Neutron Transport Calculations,"

Report Number MPM-402781 (Revision 1), MPM Technologies, Inc., 2161 Sandy Drive, State College, PA 16803-2283, September, 2003.

[1-5] Regulatory Guide 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence," U. S. Nuclear Regulatory Commission, March 2001.

[1-6] Constellation Engineering Change Notice ADC-13-001142-CN-001, "Update Effective Full Power Year (EFPY) Data with EOC 10 and EOC 13 Conditions", Revision 00006, 2013.

1-2

Introduction Table 1-1 NMP2 Cycle Operating History (Reference [1-61.

Fuel Cycle Rated Power EFPSa Cumulative EFPY Fuel___yle (MWt) 1 3323 4.41E+07 1.398 2 3323 2.69E+07 2.251 3 3323 3.65E+07 3.406 4 3323 3.87E+07 4.634 5 3467 3.96E+07 5.887 6 3467 4.24E+07 7.231 7 3467 4.70E+07 8.720 8 3467 5.43E+07 10.442 9 3467 5.63E+07 12.225 10 3467 5.88E+07 14.087 11 3467 5.88E+07 15.951 12 3467 6.07E+07 17.875 13 3467 5.95E+07 19.759 14 3988 5.4 7 E+07b 21.492b

a. Effective full power seconds (EFPS) calculated using rated power for each cycle.

b. Projected to end of cycle 14.

1-3

2 VESSEL FLUENCE EXTRAPOLATION The neutron fluence (E>1 MeV) to the reactor vessel has been evaluated as of the projected end of cycle 14 and projected to 32 and 54 EFPY of operation. Results for the wetted surface of the vessel (clad IR) are given as a function of height from 24 inches below the bottom of the fuel to 12 inches above the top of active fuel in Table 2-1. These results are calculated at the azimuthal angle with the maximum fast neutron flux, approximately 26.4 degrees in the first octant. The radial fall-off through the vessel is given in Table 2-2 evaluated at the maximum fluence point. Included are the attenuation of flux E>0. 1 MeV and dpa, as well as flux E>1.0 MeV. Table 2-3 evaluates the calculated maximum vessel fluence and dpa values at the wetted surface, the vessel IR, 1/4T, and 3/4T positions.

The reactor vessel in the beltline region is made up of plates that are welded together as shown in Figure 2-1. The figure shows that there are two shells that extend within the beltline region and each is made up of 3 plates. The boundary between these two shells is located at 60.9 inches above BAF. Thus, the maximum fluence to the upper of these two shells, the lower intermediate course, is equal to the maximum for the vessel. Each of the 3 plates extends over a sufficient region that each plate will contain at least one point at the maximum fluence. The lower of these shells, the lower course, will have a maximum fluence at the boundary circumference between the shells. The maximum fluence values for this course are tabulated in Table 2-4. These maximum fluence values also apply to the circumferential weld separating the two courses.

Figure 2-1 also indicates the approximate locations of the 6 vertical welds (3 in each shell) that extend into the beltline region. The exact locations are given in Table 2-5. The maximum fluence for the lower course welds is at the top of each weld, 60.9 inches above BAF.

The maximum fluence to the lower intermediate course vertical welds is at the vessel axial peak, about 95 inches above BAF. The maximum fluence values for each of these vertical welds and for the beltline circumferential weld are given in Table 2-6.

Radiation embrittlement effects are often correlated with fast fluence (E > 1.0 MeV).

However, it is generally thought that dpa might be a better correlation parameter since it accounts for spectral effects and, if this is correct, the use of the fast fluence (E > 1.0 MeV) values within the vessel might under predict the radiation damage at locations within the vessel.

Therefore, the NRC requires the use of a dpa attenuation in the vessel for preparation of pressure-temperature operating curves. The fluence attenuation factors within the vessel can be evaluated using calculated dpa attenuation from Table 2-3, or using a dpa formulation specified in RG 1.99 (Rev 2) [2-1]. In accordance with RG 1.190, fluence values reported in Table 2-7 are recommended for use since they are the result of an absolute fluence calculation and it has been established that there is no bias that needs to be applied.

2-1

Vessel Fluence Extrapolation The dpa values in this report are calculated using the ASTM E693-94 Standard dpa cross-sections [2-2]. This evaluation of the dpa cross section is based on the ENDF-IV cross-section file. A new dpa cross-section evaluation based on ENDF-VI (consistent with the cross-sections in BUGLE-96) has recently been introduced [2-3]. The new standard was not used here in order to be consistent with the data used for the formulation of the accepted damage correlations and the procedures recommended by the NRC [2-1]. Change to the new cross section set would result in at most a few percent change in the dpa results. For example, a comparison given in the new standard indicates that the effect at the I/4T position in a PWR is less than 2% [2-3].

The vessel has a number of nozzle penetrations and exposure to the nozzles is of interest for damage analysis. In NMP2, five sets of these nozzles were evaluated. The locations of the axial centerline of each nozzle together with the axial location of the maximum fluence points are given in Table 2-8. In the case of the nozzles below core, the maximum fluence point is at the top of each nozzle. For the nozzles near the top of the core, the reverse is true. For the LPCI nozzles (N6), the nozzle extends over a span greater than + 8 degrees from the center. Thus, since the fluence increases with decreasing angle and decreases with increasing height, it is not apparent which location on the nozzle is at the maximum fluence point. Accordingly, the fluence was evaluated at 8 points around the bottom of the nozzle. It was found that the maximum fluence is indeed at the bottommost location, so the maximum fluence point is at the bottom of both the N6 amd N12 nozzles.

The nozzle maximum fluence results are given in Table 2-9. The calculations indicate that the upper nozzles have exceeded a fluence of 1.OE 17 n/cm 2 by the present time. The lower nozzles are not expected to exceed this fluence during the life of the plant.

2.5 Chapter 2 References

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

[2-2] ASTM Designation E693-94, Standard Practice for Characterizing Neutron Exposures in Iron and Low Alloy Steels in Terms of Displacements Per Atom (DPA), E706(ID), in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 2000.

[2-3] ASTM Designation E693-01, Standard Practice for Characterizing Neutron Exposures in Iron and Low Alloy Steels in Terms of Displacements Per Atom (DPA), E706(ID), in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 2004.

2-2

Vessel Fluence Extrapolation Table 2-1 Axial Variation of Maximum Fluence at the Vessel Wetted Surface.

2 Fluence (E>1.0 MeV) n/cm Distance from BAF (in) EOC 14 32 EFPY 54 EFPY (21.49 EFPY)

-24 4.40E+ 15 8.11 E+15 1.59E+16

-22 6.01E+15 1.11E+16 2.17E+16

-20 8.09E+15 1.49E+16 2.93 E+ 16

-18 1.08E+16 1.99E+16 3.90E+ 16

-16 1.41E+16 2.61E+16 5.13E+16

-14 1.84E+16 3.41E+16 6.69E+16

-12 2.38E+16 4.41E+16 8.66E+16

-10 3.09E+16 5.73E+16 1.13E+17

-8 3.95E+16 7.34E+16 1.44E+ 17

-6 5.01E+16 9.31E+16 1.83E+17

-4 6.23E+16 1.16E+17 2.28E+17

-2 7.60E+16 1.42E+17 2.79E+17 0 9.09E+16 1.70E+17 3.35E+17 2 1.06E+17 1.98E+17 3.91E+17 4 1.22E+17 2.28E+17 4.50E+17 6 1.37E+17 2.57E+17 5.09E+17 8 1.53E+17 2.88E+17 5.69E+17 10 1.70E+17 3.19E+17 6.31E+17 12 1.85E+17 3.48E+17 6.89E+17 14 2.OOE+17 3.76E+17 7.45E+17 16 2.14E+17 4.03E+17 7.99E+17 18 2.28E+17 4.29E+17 8.49E+ 17 20 2.40E+ 17 4.52E+17 8.95E+17 22 2.51E+17 4.73E+17 9.37E+17 24 2.62E+17 4.92E+17 9.75E+17 26 2.71E+17 5.1OE+17 1.01E+18 28 2.80E+17 5.26E+17 1.04E+18 30 2.88E+17 5.41E+17 1.07E+18 32 2.96E+ 17 5.54E+17 1.1OE+18 34 3.03E+17 5.67E+17 1.12E+18 36 3.1OE+17 5.79E+17 1.14E+ 18 38 3.18E+17 5.92E+17 1.17E+18 40 3.26E+17 6.06E+17 1.19E+18 2-3

Vessel Fluence Extrapolation Table 2-1 Axial Variation of Maximum Fluence at the Vessel Wetted Surface (continued).

2 Fluence (E>1.0 MeV) n/cm Distance from BAF (in) EOC 14 32 EFPY 54 EFPY (21.49 EFPY) 42 3.33E+17 6.18E+17 1.21E+18 44 3.40E+ 17 6.29E+17 1.23E+18 46 3.47E+ 17 6.40E+ 17 1.25E+18 48 3.54E+ 17 6.52E+17 1.28E+18 50 3.61E+17 6.63E+17 1.29E+18 52 3.68E+17 6.74E+17 1.31E+18 54 3.75E+17 6.84E+ 17 1.33E+18 56 3.82E+17 6.95E+17 1.35E+18 58 3.89E+17 7.05E+17 1.37E+18 60 3.96E+17 7.15E+17 1.38E+18 62 4.02E+17 7.25E+17 1.40E+18 64 4.09E+17 7.34E+17 1.41E+18 66 4.16E+17 7.43E+17 1.43E+18 68 4.22E+17 7.52E+17 1.44E+ 18 70 4.28E+17 7.60E+ 17 1.45E+18 72 4.34E+17 7.68E+17 1.47E+18 74 4.40E+ 17 7.75E+17 1.48E+18 76 4.46E+17 7.82E+17 1.49E+18 78 4.51E+17 7.89E+17 1.50E+18 80 4.56E+17 7.95E+17 1.50E+18 82 4.61E+17 8.OOE+17 1.51E+18 84 4.66E+17 8.05E+17 1.52E+18 86 4.70E+17 8.09E+17 1.52E+18 88 4.73E+17 8.13E+ 17 1.52E+18 90 4.76E+17 8.16E+17 1.53E+18 92 4.79E+17 8.18E+17 1.53E+18 94 4.81E+17 8.19E+17 1.53E+18 96 4.83E+17 8.20E+17 1.53E+18 98 4.84E+ 17 8.20E+17 1.52E+18 100 4.85E+17 8.19E+17 1.52E+18 102 4.85E+17 8.17E+17 1.51E+18 104 4.84E+17 8.14E+ 17 1.50E+18 106 4.82E+17 8.09E+17 1.49E+18 2-4

Vessel Fluence Extrapolation Table 2-1 Axial Variation of Maximum Fluence at the Vessel Wetted Surface (continued).

2 Fluence (E>1.0 MeV) n/cm Distance from BAF (in) EOC 14 32 EFPY 54 EFPY (21.49 EFPY) 108 4.80E+17 8.03E+17 1.48E+18 110 4.76E+17 7.96E+17 1.46E+18 112 4.72E+17 7.87E+17 1.45E+18 114 4.67E+17 7.77E+17 1.43E+18 116 4.60E+ 17 7.64E+ 17 1.40E+ 18 118 4.52E+17 7.50E+17 1.37E+18 120 4.43E+17 7.33E+17 1.34E+18 122 4.32E+17 7.14E+17 1.30E+18 124 4.19E+ 17 6.92E+17 1.26E+18 126 4.05E+17 6.69E+17 1.22E+18 128 3.89E+17 6.42E+17 1.17E+18 130 3.64E+17 6.01E+17 1.1OE+18 132 3.37E+17 5.57E+17 1.02E+18 134 3.12E+17 5.16E+17 9.43E+17 136 2.85E+17 4.72E+17 8.64E+17 138 2.91E+17 4.82E+17 8.81E+17 140 2.62E+17 4.34E+17 7.95E+17 142 2.33E+17 3.87E+17 7.1OE+17 144 2.04E+17 3.40E+17 6.24E+17 146 1.72E+17 2.87E+17 5.28E+17 148 1.36E+17 2.27E+17 4.19E+ 17 150 1.04E+17 1.76E+17 3.25E+17 152 9.44E+ 16 1.60E+17 2.97E+ 17 154 8.30E+16 1.41E+17 2.63E+17 156 7.09E+16 1.21E+17 2.25E+17 158 5.72E+16 9.78E+16 1.83E+17 160 4.60E+16 7.88E+16 1.47E+ 17 162 3.94E+ 16 6.76E+16 1.27E+17 2-5

Vessel Fluence Extrapolation Table 2-2 Relative Exposure Through the Vessel at the Maximum Fluence Point Averaged over Cycles 1 to 10.

Relative Fluence Relative Fluence Location (E>1.0 MeV) (E>0.1 MeV) Relative dpa Clad IR 1.016 0.978 1.019 Fraction of Distance through Vessel 0.00 1.000 1.000 1.000 0.05 0.944 1.000 0.944 0.10 0.878 0.982 0.883 0.15 0.809 0.956 0.820 0.20 0.739 0.923 0.758 0.25 0.672 0.886 0.698 0.30 0.609 0.846 0.641 0.35 0.551 0.805 0.589 0.40 0.496 0.764 0.539 0.45 0.446 0.722 0.493 0.50 0.400 0.680 0.450 0.55 0.358 0.638 0.411 0.60 0.320 0.596 0.374 0.65 0.286 0.555 0.339 0.70 0.253 0.513 0.307 0.75 0.224 0.472 0.276 0.80 0.197 0.430 0.247 0.85 0.172 0.387 0.219 0.90 0.148 0.343 0.191 0.95 0.124 0.296 0.163 1.00 0.105 0.254 0.139 2-6

Vessel Fluence Extrapolation Table 2-3 Calculated Maximum Vessel Exposure at the End of Cycle 14 (21.49 EFPY) and Projected to 32 and 54 EFPY.

Location Radius (cm) 21.49' EFPY 32 EFPY 54 EFPY 2

Fluence (E>1.0 MeV) n/cm Vessel Clad IR 321.310 4.85E+17 8.20E+17 1.53E+18 Vessel IR 321.786 4.77E+17 8.07E+ 17 1.50E+18 Vessel 1/4T 325.874 3.21E+17 5.42E+ 17 1.O1E+18 Vessel 3/4T 334.050 1.07E+17 1.81E+17 3.36E+17 2

Fluence (E>O.1 MeV) n/cm Vessel Clad IR 321.310 8.77E+17 1.49E+18 2.77E+18 Vessel IR 321.786 8.97E+17 1.52E+18 2.83E+18 Vessel 1/4T 325.874 7.94E+ 17 1.35E+18 2.51E+18 Vessel 3/4T 334.050 4.23E+17 7.18E+17 1.34E+18 Displacements per Atom (dpa)

Vessel Clad IR 321.310 7.60E-04 1.29E-03 2.39E-03 Vessel IR 321.786 7.46E-04 1.26E-03 2.35E-03 Vessel 1/4T 325.874 5.20E-04 8.80E-04 1.64E-03 Vessel 3/4T 334.050 2.06E-04 3.48E-04 6.49E-04

1. EFPY at expected end of cycle 14.

2-7

Vessel Fluence Extrapolation Table 2-4 Calculated Maximum Vessel Lower Course Exposure at the End of Cycle 14 (21.49 EFPY) and Projected to 32 and 54 EFPY.

Location Radius (cm) 21.491 EFPY 32 EFPY 54 EFPY 2

Fluence (E>1.0 MeV) n/cm Vessel Clad IR 321.310 3.99E+17 7.20E+17 1.39E+18 Vessel IR 321.786 3.93E+17 7.09E+ 17 1.37E+18 Vessel 1/4T 325.874 2.64E+17 4.76E+17 9.20E+17 Vessel 3/4T 334.050 8.84E+ 16 1.59E+17 3.07E+ 17 2

Fluence (E>O.1 MeV) n/cm Vessel Clad IR 321.310 7.23E+17 1.31E+18 2.52E+18 Vessel IR 321.786 7.39E+17 1.33E+18 2.58E+18 Vessel 1/4T 325.874 6.57E+17 1.18E+18 2.29E+18 Vessel 3/4T 334.050 3.54E+17 6.36E+17 1.23E+18 Displacements per Atom (dpa)

Vessel Clad IR 321.310 6.27E-04 1.13E-03 2.18E-03 Vessel IR 321.786 6.15E-04 1.11E-03 2.14E-03 Vessel 1/4T 325.874 4.29E-04 7.74E-04 1.49E-03 Vessel 3/4T 334.050 1.71E-04 3.08E-04 5.94E-04 I. EFPY at expected end of cycle 14.

2-8

Vessel Fluence Extrapolation Table 2-5 Azimuthal Locations of Vessel Welds that Extend into the Beltline.

Azimuthal Equivalent Azimuthal Weld ID' Angle (degrees) Angle (degrees) in First Octant Lower Course BA 77.5 12.5 BB 197.5 17.5 BC 317.5 42.5 Lower Intermediate Course BD 90 0 BE 210 30 BF 330 30

1. See Figure 2-1.

Table 2-6 Calculated Maximum Vessel Beltline Weld Exposures at the End of Cycle 14 (21.49 EFPY) and Projected to 32 and 54 EFPY.

Weld ID 21.491 EFPY 1 32 EFPY 54 EFPY 2

Fluence (E>1.0 MeV) n/cm Lower Course Vertical Welds BA 2.61E+17 4.71E+17 9.11E+17 BB 3.37E+17 6.09E+17 1.18E+ 18 BC 2.83E+17 5.11E+17 9.88E+17 Lower Intermediate Course Vertical Welds BD2 2.07E+ 17 3.50E+17 6.52E+17 BE 4.39E+17 7.42E+17 1.38E+18 BF 4.39E+17 7.42E+17 1.38E+18 Circumferential Weld Between Lower and Lower Intermediate Course 60.9 inches above 3.99E+17 J 7.20E+17 1.39E+18 BAF II I. EFPY at expected end of Cycle 14.

2. Weld BD fluence is slightly conservative since the jet pump centered at 90 degrees is not present in the first octant model.

2-9

Vessel Fluence Extrapolation Table 2-7 Calculated Peak Vessel Fluence and Fluence Determined Using dpa Attenuation.

Calculated Attenuation using Attenuation using Position Fluence (E > 1 MeV) Calculated dpal RG1.99( Rev 2)2 (n/cm 2 ) (n/cm 2 )

(n/cm 2 )

3 End of Cycle 14 (21.49 EFPY)

Vessel Clad IR 4.85E+17 4.85E+17 4.85E+17 Vessel IR 4.77E+17 4.76E+17 4.63E+17 1/4 T 3.21E+17 3.32E+17 3.15E+17 3/4 T 1.07E+17 1.31E+17 1.45E+17 At 32 EFPY Vessel Clad IR 8.20E+ 17 8.20E+ 17 8.20E+ 17 Vessel IR 8.07E+17 8.05E+17 7.84E+17 1/4 T 5.42E+17 5.61E+17 5.33E+17 3/4 T 1.81E+17 2.22E+ 17 2.46E+ 17 At 54 EFPY Vessel Clad IR 1.53E+18 1.53E+18 1.53E+18 Vessel IR 1.50E+18 1.50E+18 1.46E+18 1/4 T 1.O1E+18 1.05E+18 9.92E+17 3/4 T 3.36E+17 4.14E+17 4.58E+17

1. Calculated fluence at the vessel inner wetted surface (clad IR) times the ratio of calculated dpa at the vessel interior points to the dpa at the inner wetted surface.

2. Calculated fluence at the vessel inner wetted surface (clad IR) times exp(-0.24*x), where x is the distance into the vessel in inches.

3. EFPY at expected end of Cycle 14.

2-10

Vessel Fluence Extrapolation Table 2-8 NMP2 Nozzle Locations Near the Core Region.

Height above Height above Height of Max.

Nozzle vessel zero (in)1 BAF (in) OD (in) Fluence Point Angle in First above BAF (in) Octant NI 175.5 -40.813 50.875 -15.375 0 N2 181.5 -34.813 35.375 -17.125 30 N9 152.0 -64.313 18.875 -54.875 15 N6 372.5 156.187 35.375 138.500 45 N 12 366.0 149.687 3.375 148.000 20

1. Height of the nozzle centerline.

Table 2-9 NMP2 Nozzle Maximum Fluence Values.

2 Height above 4Fluence (E > 1 MeV) n/cm BAF (in)' (21.49 EFPY)2 32 EFPY 54 EFPY NI -15.375 6.33E+15 1.14E+16 2.19E+16 N2 -17.125 1.09E+16 2.02E+ 16 3.96E+16 N9 -54.875 6.08E+13 1.08E+14 2.08E+14 N6 138.500 1.52E+17 2.51E+17 4.57E+17 N12 148.000 1.04E+17 1.68E+17 3.O1E+17

1. Height of maximum fluence point.

2. EFPY at expected end of Cycle 14.

2-11

Vessel Fluence Extrapolation 00 900 1800 2700 3600 I

428". U.

AC 0 w U.

Ca 366.31" SHELL 2 C3065 -1 C3121-2 C3147-1 117 . ' . ..

A8 SHELL 1 216.31'-.

I BELTLINE C3066-2 C3066-2 C3147-2 CD cc 126.5" AA Figure 2-1 NMP2 Beltline Weld Seam and Plate Locations.

2-12

3 UNCERTAINTY AND BIAS ANALYSIS Detailed uncertainty analyses have been performed in the past for NMP2 to estimate each source of uncertainty in the calculated fluence values. These analysis made use of defined uncertainties and tolerances where possible, but many of the uncertainty estimates had to be based on estimates derived from data variation, such as the detailed power distribution and void fraction calculations used for the transport models. Discussion of each uncertainty assumption is briefly discussed below. The uncertainty assessments for the shroud, capsule, and vessel are summarized in Table 3-1.

3.1 Uncertainty Assumptions Nuclear Data Nuclear data input to the transport calculations includes the multigroup cross sections and neutron spectrum. Uncertainties in the cross sections are complicated because of the large number of cross section values and the correlations between these values. Although the uncertainties in individual cross section values may be relatively large, the total effect of cross section uncertainties is limited by adjustments made by cross section evaluators to agree with benchmark data. The approach taken here is to limit the cross section uncertainty effects to just the total cross section and to evaluate this by varying the material densities (see below).

Uncertainty in the multigroup fission source arises from uncertainty in the fission spectra for each fissioning isotope, the distribution of fission among the fissioning isotopes, the energy release per fission (K), and the number of neutrons produced per fission (u). Uncertainty in the fission spectrum is mainly at the higher energies, and this has little effect on the fluence above 1 MeV except for very deep penetrations. The uncertainty was represented as an uncertainty in burnup, which was taken as 10,000 MWdiMTU (megawatt days per metric ton of uranium). The transport calculation used a burnup value for each calculated cycle that was an average of the outer two rows of fuel bundles at core axial midplane averaged over the cycle. The uncertainty of 10,000 MWd/MTU is a conservative estimate of the burnup variation that affects the assemblies important for fluence to the shroud and vessel. A 1-D calculation was performed for a typical case to determine the spectral effect and it was found to affect the flux above 1 MeV by between 0.2% in the core to 1.8% at the outside of the vessel [3-1].

The parameters u and K both increase with burnup, but the source normalization is proportional to the ratio u / K. Thus, since the change in these parameters partially offset each other, the variation with burnup is small. For an uncertainty of 10,000 MWd/MTU, the normalization uncertainty is 1.1%. Since this is in the same direction as the spectrum uncertainty, it is added to the spectrum contribution to give the values in Table 3-1.

3-1

Uncertainty andBias Analysis Normalization In addition to the normalization uncertainty due to u/K, there is an overall normalization uncertainty in reactor power as measured by the heat balance. This uncertainty is estimated by the plant staff to be 2% [3-2].

Geometry Geometry uncertainties are taken from Reference [3-2]. The vessel inner radius uncertainty was taken to be a typical value [3-1]. The uncertainty in the shroud inner radius was based on as-built measurements of the inner diameter. These measurements indicate a range of 203.062 to 203.250 inches. The radius will then have a maximum to minimum range of 101.531 to 101.625 inches (a range of 0.094 inches). The important distance is, however, the distance from the core edge to the shroud, and if the shroud is slightly off-center, then this uncertainty could be larger. To be conservative, an uncertainty of 0.188 inches was used and assumed to be 1 standard deviation. The shroud radius used in the calculation was actually not the center of the range, but was the design radius of 101.56 inches, which is only 0.029 inches from the minimum as-built value. Thus, the true uncertainty is more in the direction of lower fluence than higher fluence, an additional conservatism. The tolerance on the shroud thickness of 0.042 is a conservative value taken from [3-1]. The shroud thickness uncertainty has no impact on the inside of the shroud, but has a 0.5% effect on the outside.

Jet Pumps The jet pumps could not be exactly modeled in the calculations due to limitations on mesh size. The steel from the jet pumps was placed appropriately in the mesh in the downcomer region in the R,O calculation. The jet pumps could not be included in the R,Z calculation. To estimate the uncertainty introduced by the jet pump model, a separate R,0 calculation was made with the jet pumps omitted. This had no effect on the shroud or surveillance capsule fluence, but the maximum fluence at the vessel inner radius increased up to 13.9%. For fluence at the maximum fluence points, a reasonable estimate of the 1-Y uncertainty from the imperfect modeling of the jet pumps is 25% of this value, or 3.5%. Near the top of the core, the jet pump structure varies. At locations where there is more steel present, this was ignored and the vessel fluence will be slightly overestimated. For the upper 12 inches of the core region, the vessel fluence was calculated with no jet pumps included in the model.

MaterialDensities The material density uncertainty was treated differently for the water density and the steel density. The water density in the core decreases with height as the void fraction increases.

Based on the variation in the void fraction at NMP1 [3-1], on a comparison of the Unit 2 data for the various cycles, and on the necessity for the heat generation in the core to produce a certain rate of steam, the void fraction uncertainty was estimated to be 5%. The bypass water is not thought to have any void volume, but the temperature may vary from the value that was assumed. For conservatism, the bypass water density was assumed to be the density of the core water. The uncertainty was estimated by taking one half of the difference between the inlet water density and the assumed bypass water density. This indicates an uncertainty of 1.3%.

This is consistent with the value of 1.4% in Reference [3-1 ], which was estimated using a slightly different method. The slightly higher value of 1.4% was adopted. The uncertainty in the downcomer water density was calculated from a temperature uncertainty of 5 F [3-1].

3-2

Uncertaintyand Bias Analysis The uncertainty in steel density is less than about 1%. However, as noted above, the cross section uncertainty was included as an addition to the steel density uncertainty. An estimate for this uncertainty was derived by considering vessel mockup benchmark results [3-3],

comparisons of reactor cavity and surveillance capsule measurements [3-4, 3-5, 3-6, 3-7], and comparisons of cross section evaluations [3-8]. It was concluded that uncertainties due to the iron cross section contribute a 10% effect on fluence through a reactor vessel. This translates into a cross section uncertainty of 3.5%. This value was adopted as the density variation and uncertainties were calculated based on this uncertainty estimate. In addition, the core cross sections for the fuel and cladding were also assumed to have this uncertainty. This estimate includes effects due to the core homogenization.

Source Uncertainty Source uncertainties were estimated based on the variation of the calculated power distributions for various depletion calculations [3-2]. This produced estimated uncertainties of 6% radially and 3.7% axially. Larger differences were observed between cases and these uncertainties were included in the flux history uncertainty (see below).

Methods Uncertainty The neutron transport was calculated using a model of the reactor and SN code. This is only an approximation to the solution of the Boltzmann transport equation and thus also contributes uncertainty. Two components of this uncertainty were considered. First, the uncertainty of the fuel model was considered. From the VENUS benchmark measurements, it was found that a typical range of C/E results was about 10%. Thus the standard deviation was about 5% and this value was used here. The second component was the adequacy of the S8 calculation. To test this, S16 calculations were performed to indicate the accuracy. Differences of 1.4% were observed in the shroud and as high as 3% at the outside of the vessel. The differences were added in quadrature to the 5% from the first modeling effect.

Flux History The flux calculations were made for all cycles through 10. The detailed data for cycles 8 through 10 enabled confidence in the integrated fluence for these cycles. Previous evaluations based on data for cycles 1 to 7 determined an estimated flux history uncertainty of 7% [3-9] and 8% [3-10]. The cycles 8 to 10 variation indicates that these estimates may be somewhat high.

However, the conversion to GE14 fuel introduces additional uncertainty for extrapolation. Based on these considerations, a flux history uncertainty of 7% is a reasonable approximation.

3.2 Uncertainty Evaluation The results for the uncertainty evaluation at midplane are summarized in Table 3-1, based on the assumptions just discussed. This table is applicable to the shroud, vessel, and surveillance capsule at axial midplane and nearby locations including the peak axial fluence point. Some of the uncertainty factors in this table vary with azimuth, and in this case, the maximum value was used for the shroud and vessel. A total uncertainty was derived by combining the independent individual contributors in quadrature. This gives an uncertainty for the maximum vessel fluence of 15.1%. The main contributors to the vessel fluence uncertainty were found to be the methods uncertainty, the core void fraction and source distributions, and the vessel radius uncertainty.

3-3

Uncertainty andBias Analysis The vessel fluence uncertainty is evaluated at the maximum fluence point, but the variation in vessel uncertainty with position is relatively small.

The uncertainty for the capsule fluence was found to be 14.8%. The contributors to the uncertainty in the surveillance capsule fluence are similar to that for the reactor vessel inner radius except for the capsule axial and azimuthal position uncertainty, which must be added, and the jet pump uncertainty which is not present at this angle. The uncertainty in the capsule radial location was taken to be the same as the vessel radius uncertainty since the uncertainty in capsule location relative to the vessel is small.

The overall conclusion is that the uncertainty is within acceptable tolerance. Based on the results of the benchmark report, the C/M ratio is within 20% as required by the Regulatory Guide 1.190 for no bias application.

3.4 Chapter 3 References

[3-1] "Nine Mile Point Unit 1 Shroud Neutron Transport and Uncertainty Analysis," Report Number MPM-108679, MPM Technologies, Inc., 2161 Sandy Drive, State College, PA 16803-2283, October, 1998.

[3-2] Letter, George Inch (Constellation Energy) to Dr. Michael Manahan (MPM Consulting),

"Reference Transmittal for NMP Unit 1 and Unit 2 Fluence Analysis," November 23, 2005.

[3-3] McElroy, W.N., Ed., "LWR-PV-SDIP: PCA Experiments and Blind Test", NUREG/CR-1861, 1981.

[3-4] Lippincott, E.P., "Consumers Power Company Palisades Nuclear Plant Reactor Vessel Fluence Analysis", WCAP-13348, May 1992.

[3-5] Maerker, R.E., et. al., "Application of LEPRICON Methodology to LWR Pressure Vessel Dosimetry", Reactor Dosimetry: Methods, Applications, and Standardization, ASTM STP 1001, 1989, pp 405-414.

[3-6] Anderson, S.L., "Westinghouse Fast Neutron Exposure Methodology for Pressure Vessel Fluence Determination and Dosimetry Evaluation", WCAP-13362, May 1992.

[3-7] Lippincott, E.P., et. al., "Evaluation of Surveillance Capsule and Reactor Cavity Dosimetry from H. B. Robinson Unit 2, Cycle 9", WCAP-1 1104, NUREG/CR-4576, February 1987.

[3-8] Haghighat, A. and Veerasingam, "Comparison of the Different Cross Section Libraries used for Reactor Pressure Vessel Fluence Calculations", Trans. Amer. Nuclear Society, 64, p. 357, 1991.

[3-9] "Nine Mile Point Unit 2 Shroud Neutron Transport and Uncertainty Analysis," Report MPM-200623, MPM Technologies, Inc., 2161 Sandy Drive, State College, PA 16803-2283, February 2000.

3-4

Uncertainty andBias Analysis

[3-10] "Nine Mile Point Unit 2 3-Degree Pressure Vessel Surveillance Capsule Report", Report MPM-1200676, MPM Technologies, Inc., 2161 Sandy Drive, State College, PA 16803-2283, December, 2000.

3-5

Uncertainty andBias Analysis Table 3-1 Nine Mile Point Unit 2 Shroud, Capsule, and Vessel Fluence Calculational Uncertainty.

Shroud IR Vessel IR Capsule Fluence Fluence Fluence Uncertainty Assigned Uncertainty % Uncertainty % Uncertainty %

Contributor Uncertainty (lo) (la) (lO)

Fission Spectrum and 10000 Mwd/MTU 2.1 2.9 2.9 nu/kappa Heat Balance 2% 2.0 2.0 2.0 Shroud IR 0.188 inches 7.5 0.2 0.2 Shroud Thickness 0.042 inches 0.0 1.0 1.0 Vessel IR 0.125 inches 0.0 3.2 3.2 Core Midplane Water 5% 6.4 4.4 4.4 Density Bypass Water Density 1.40% 4.4 3.2 2.9 Downcomer Water 5F 0.0 3.9 3.9 Temperature Steel Density (total 3.5% 1.4 2.3 2.3 cross section)

Core Fuel Density 3.5% 2.7 3.0 3.0 Radial Source Dist. 6% 6.0 6.0 6.0 Axial Source Dist. 3.7% 3.7 3.7 3.7 Methods Uncertainty 5% + 5.2 5.8 5.8 Flux History 7% 7.0 7.0 7.0 Capsule Radial 0.125 inches 0.0 0.0 2.0 Location Capsule Azimuthal and 1 inch 0.0 0.0 1.0 Axial Location Jet Pump Model 25% ofjet pump 0.0 3.5 0.0 mass Total 16.1 15.1 14.8 3-6

4

SUMMARY

AND CONCLUSIONS An updated evaluation of the neutron exposure to the NMP2 reactor vessel, including each vessel plate and weld and the vessel nozzles has been completed based on reactor operation through the estimated EOC 14 and projected to 32 and 54 EFPY. This evaluation is based entirely on methodology that has been extensively benchmarked to satisfy all the requirements of Nuclear Regulatory Guide 1.190. A detailed uncertainty analysis has shown that all results within the reactor beltline region fall within the allowable 20% uncertainty bound.

4-1

5 NOMENCLATURE BWR boiling water reactor C/M calculated-to-measured ratio D dimension dpa displacements per atom EFPS effective full power seconds EFPY effective full power years EOC end-of-cycle EOL end-of-license GGNS Grand Gulf Nuclear Station ID inner diameter IR inner radius MWd/MTU megawatt days per metric ton of uranium MOC middle-of-cycle MPM MP Machinery and Testing, LLC NMP2 Nine Mile Point Unit 2 NRC U. S. Nuclear Regulatory Commission OD outer diameter OR outer radius ORNL Oak Ridge National Laboratory PTLR Pressure and Temperature Limits Report RBS River Bend Station RG Regulatory Guide (NRC)

RSICC Radiation Safety Information Computational Center T vessel or shroud wall thickness TAF top of active fuel 5-1

ATTACHMENT 4 RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION, EVIB - RAI 6 (NON-PROPRIETARY)

This attachment provides supplemental information from Enclosure 2 to GEH letter GE-PPO-1GYEF-KGI-720, in response to the NRC request for additional information that was provided in an email from the NRC to NMPNS on September 27, 2013; specifically, EVIB - RAI 6. The NRC question is repeated (in italics), followed by the NMPNS response.

(4 pages attached)

Nine Mile Point Nuclear Station, LLC November 4, 2013

ENCLOSURE 2 GE-PPO-1GYEF-KG1-720 Response to EVIB RAI 6 Non-proprietary Information- Class I (Public)

NON-PROPRIETARY NOTICE This is a non-proprietary version of Enclosure 1 of GE-PPO-1GYEF-KG1-720 which has the proprietary information removed. Portions of the document that have been removed are indicated by an open and closed bracket as shown here ((

Enclosure 2 Non-proprietary Information-Class I (Public)

GE-PPO- 1GYEF-KG 1-720 Page 1 of 3

Background

Nine Mile Point Nuclear Station, Unit No. 2, Docket No. 50-410, License Amendment Request Pursuantto 10 CFR 50.90: Relocation of Pressureand TemperatureLimit Curves to the Pressureand TemperatureLimits Report (TA C NO. MF0345)

By letter dated November 21, 2012, as supplemented by letter dated March 25, 2013, Nine Mile Point Nuclear Station, LLC (NMPNS, the Licensee), submitted a license amendment request (LAR) for Nine Mile Point Unit 2. The proposed amendment would modify Technical Specification (TS) Section 3.4.11, "RCS Pressureand Temperature (PIT)Limits, " by replacing the existing reactorvessel heatup and cooldown rate limits and the pressureand temperature (P-T)limit curves with references to the Pressureand TemperatureLimits Report (PTLR). In addition, a new definitionfor the PTLR would be added to TS Section 1.1, "Definitions,"and a new section addressingadministrativerequirementsfor the PTLR would be added to TS Section 5.0, "Administrative Controls." By letters datedJuly 31, and September 6, 2013, the licensee provided the responses, to the RAIs transmittedby NRC on June 20, 2013 (ADAMS Accession Package No. ML13214A396). The Licensee's Responses containedProprietary information. (Agencywide Document Access and Management System (ADAMS) Accession Nos.

ML123380336for November 21, 2012, submission, and ML13214A396 for July 31, 2012, submission, and ML13254A156 respectively).

The Nuclear Regulatory Commission (NRC) staff has determined that additionalinformation is needed to complete its review.

Based on the telephone discussions with the licensee on September 27, 2013, to clarify the ADDITIONAL DRAFT RAIS (ML13268A558), dated September 25, 2013, the NRC staffs Revised Additional request for additionalinformation (RAI) is provided below (Revisions are indicated in Bold Red).

NRC EVIB RAI 6 The conclusions of Appendix J to NEDC-33178P-A indicate that the ARTfrom the adjacentRPV shell materialis used in conjunction with the T-RTNDT values determined in the appendix to generate the P-T curvefor the instrument nozzle. For the NMP2 water level instrumentation (WLI) nozzle, the calculation of the P-T limits provided in the response to EVIB-RAI-5 uses an ART of 39YF. Appendix B to the PTLR provides a 32 EFPYARTfor the WLI nozzle forging of 39YF. The PTLR indicates that in addition to the limiting beltline plate material,the limiting ARTfor the beltline LPCI N6 and Water Level InstrumentationN12 nozzle forgings and welds are also consideredin the development of the beltline PT curves. The PTLR also indicates that the plant-specific copper and nickel contentfor the WLI nozzle forging were not available,so copper and nickel values were determined based on a bounding estimateforforgingsfabricated from SA508 Class 1 material,and that this was defined based on a search of availableBWR vessel purchase recordsfor SA508 Class I materials. The PTLR further indicates that representative valuesfor copper and nickel contentfor the WLI nozzle were developed using the mean values plus one standarddeviation.

Non-proprietary Information-Class I (Public)

GE-PPO- 1GYEF-KG 1-720 Page 2 of 3 Pleaseprovide the following Information (a) Clarify whether the WLI nozzle ART or the adjacentRP V shell materialART was used as the basisfor generatingthe P-T limitsfor the WLI nozzle.

(b) If the WLI nozzle ART was used as the basisfor generatingthe WLI nozzle P-T limits, explain how this is consistent with the methodology describedin NEDC-33178P-A, Appendix J.

(c) If the WLI nozzle ART was used, provide technicaljustificationfor the use of copper and nickel values basedon mean phls one standarddeviation ratherthat mean plus two standarddeviations. Also, pleaseprovide reference(s) to previouspressure-temperature limits approvedby NRC that considereda partial-penetrationtype nozzle in the beitline region, and/or determinationsof RP V materialcopper and nickel values that used a one standarddeviation upper bound of available data.

Response

NMP2 Response to (a) and (b)

The pressure-temperature (PT) curve licensing topical report (PTLR, Reference 1) states that the Er

)) The intent of this statement was

)). The PT curves for Nine Mile Point Nuclear Station, Unit 2 (NMP2) meet the intent of this statement by using the Er )) adjusted reference temperature (ART) determined for the ((

)). Application of this (( )) for development of the beltline PT curve is Er )) than use of the ART for the ((

The ART for the (( )), as shown in the PTLR is calculated using a Er )) with the fluence determined as bounding for the ((

)), equal to (( )). The resulting ART is 39°F. At the same time, the ART for the (( )), as shown in the ART table in the PTLR, is 51 °F. However, the ART of 51 °F was calculated using the peak 1/4T fluence in the lower-intermediate shell of 6.62e 17 n/cm 2, which is much greater than the fluence ((

)). Calculating the ART for the shell material using the CF = 74.5°F with the 1/4T fluence (( )), results in an ART of 22°F. It can therefore be seen that the NMP2 PT curve was developed using the limiting ART of 39°F, which results in a conservative PT curve.

NMP2 Response to (c)

Previous submittals for Duane Arnold Energy Center (Reference 2) and Monticello Nuclear Generating Plant (Reference 3) were approved by the NRC using Mean + 1 sigma for beltline nozzles. These nozzles did not have sufficient plant-specific data available to permit determination of the copper content. All other submittals prepared by GE or GEH had sufficient certified material test report (CMTR) information and/or had WLI partial penetration forgings fabricated from non-ferritic materials. Where the ((

)). PTLRs including partial penetration nozzles for Grand Gulf Nuclear Station Non-proprietary Information-Class I (Public)

GE-PPO- 1GYEF-KG 1-720 Page 3 of 3 (Reference 4) and Peach Bottom Atomic Power Station (Reference 5) were submitted to the NRC for review; all have non-ferritic (( )) and were evaluated considering the adjacent shell material.

For Duane Arnold Energy Center, copper content of 0.18% was used for both the low pressure coolant injection (LPCI) and WLI nozzles; both of these nozzles are full penetrations. The value of 0.18% was obtained using data points defined as:

)), resulting in a Mean + 1 sigma of 0.18% Cu. The same data was used for the Monticello Nuclear Generating Plant recirculation inlet nozzle that occurs in the beltline; this is also a full penetration nozzle.

New calculations were performed more recently to determine both the copper content and nickel content of the (( )). Available CMTRs were reviewed with fourteen (14) copper data points and thirty-five (35) nickel data points located. Mean + 1 sigma calculations for the (( )). This value is ((

)) than the value previously used and approved by the NRC. Combining the original data points that resulted in 0.18% Cu with the additional 14 data points resulted in a Mean + 1 sigma of((

In conjunction with the response to items (a) and (b), it has been shown that the chemistry factor (CF) of (( )), obtained using the Mean + 1 sigma chemistry, is significantly greater than the CF of(( )), required for the limiting (( )). This results in conservatism of (( )) over the use of the (( )). It is also noted that Regulatory Guide 1.99, Revision 2 contains the following text in Position 1.1 regarding chemistry: "If not available, conservative estimates (mean plus one standard deviation) based on generic data may be used if justification is provided." It is therefore justified that the use of Mean + 1 sigma is reasonable and bounding for the NMP2 [ ]

References

1. GE Hitachi Nuclear Energy, "Licensing Topical Report GE Hitachi Nuclear Energy Methodology for Development of Reactor Pressure Vessel Pressure-Temperature Curves,"

NEDC-33178P-A, Revision 1, June 2009.

2. GE Nuclear Energy, "Pressure-Temperature Curves for Duane Arnold Energy Center,"

GE-NE-A22-00100-08-0 I-R2, Revision 2, August 2003.

3. GE Hitachi Nuclear Energy, "Pressure-Temperature Curves for Nuclear Management Company LLC Monticello Nuclear Generating Plant," NEDC-33307P, Revision 0, February 2008.

4. Alan B. Wang (NRC) to Grand Gulf Nuclear Station, "Grand Gulf Nuclear Station, Unit 1 - Issuance of Amendment RE: Extended Power Uprate (TAC No. ME4679),"

July 18, 2012 (ADAMS Accession No. ML121210020).

5. Richard B. Ennis (NRC) to Michael J. Pacilio (Exelon Generation Company), "Peach Bottom Atomic Power Station, Units 2 and 3 - Issuance of Amendments RE: Relocation of Pressure and Temperature Limit Curves to the Pressure and Temperature Limits Report (TAC Nos. ME8535 and ME8536)," April 1, 2013 (ADAMS Accession No. ML13079A219).

ATTACHMENT 5 AFFIDAVIT FROM GE-HITACHI NUCLEAR ENERGY AMERICAS LLC (GEH) JUSTIFYING WITHHOLDING PROPRIETARY INFORMATION CONTAINED IN ATTACHMENT 6 Nine Mile Point Nuclear Station, LLC November 4, 2013

GE-Hitachi Nuclear Energy Americas LLC AFFIDAVIT I, Peter M. Yandow, state as follows:

(1) 1 am the Vice President, Nuclear Plant Projects/Services Licensing, Regulatory Affairs, of GE-Hitachi Nuclear Energy Americas LLC ("GEH"), and have been delegated the function of reviewing the information described in paragraph (2) which is sought to be withheld, and have been authorized to apply for its withholding.

(2) The information sought to be withheld is contained in Enclosure 1 of GEH letter, GE-PPO-IGYEF-KG1-720, "GEH Response to NMP2 EVIB RAI 6," dated October 16, 2013. The GEH proprietary information in Enclosure 1, which is entitled "Response to EVIB RAI 6,"

is identified by a dotted underline inside double square brackets. ((Th.i se.ntence is an exa.mple..!3] In each case, the superscript notation 13) refers to Paragraph (3) of this affidavit, which provides the basis for the proprietary determination.

(3) In making this application for withholding of proprietary information of which it is the owner or licensee, GEH relies upon the exemption from disclosure set forth in the Freedom ofInformation Act ("FOIA"), 5 U.S.C. Sec. 552(b)(4), and the Trade Secrets Act, 18 U.S.C.

Sec. 1905, and NRC regulations 10 CFR 9.17(a)(4), and 2.390(a)(4) for trade secrets (Exemption 4). The material for which exemption from disclosure is here sought also qualifies under the narrower definition of trade secret, within the meanings assigned to those terms for purposes of FOIA Exemption 4 in, respectively, Critical Mass Energy Proiect v. Nuclear Regulatory Commission, 975 F.2d 871 (D.C. Cir. 1992), and Public Citizen Health Research Group v. FDA, 704 F.2d 1280 (D.C. Cir. 1983).

(4) The information sought to be withheld is considered to be proprietary for the reasons set forth in paragraphs (4)a. and (4)b. Some examples of categories of information that fit into the definition of proprietary information are:

a. Information that discloses a process, method, or apparatus, including supporting data and analyses, where prevention of its use by GEH's competitors without license from GEH constitutes a competitive economic advantage over other companies;

b. Information that, if used by a competitor, would reduce their expenditure of resources or improve their competitive position in the design, manufacture, shipment, installation, assurance of quality, or licensing of a similar product;

c. Information that reveals aspects of past, present, or future GEH customer-funded development plans and programs, resulting in potential products to GEH;

d. Information that discloses trade secret or potentially patentable subject matter for which it may be desirable to obtain patent protection.

(5) To address 10 CFR 2.390(b)(4), the information sought to be withheld is being submitted to NRC in confidence. The information is of a sort customarily held in confidence by GEH, and is in fact so held. The information sought to be withheld has, to the best of my Affidavit for GE-PPO- 1GYEF-KG 1-720 Page I of 3

GE-Hitachi Nuclear Energy Americas LLC knowledge and belief, consistently been held in confidence by GEH, not been disclosed publicly, and not been made available in public sources. All disclosures to third parties, including any required transmittals to the NRC, have been made, or must be made, pursuant to regulatory provisions or proprietary or confidentiality agreements that provide for maintaining the information in confidence. The initial designation of this information as proprietary information, and the subsequent steps taken to prevent its unauthorized disclosure, are as set forth in the following paragraphs (6) and (7).

(6) Initial approval of proprietary treatment of a document is made by the manager of the originating component, who is the person most likely to be acquainted with the value and sensitivity of the information in relation to industry knowledge, or who is the person most likely to be subject to the terms under which it was licensed to GEH. Access to such documents within GEH is limited to a "need to know" basis.

(7) The procedure for approval of external release of such a document typically requires review by the staff manager, project manager, principal scientist, or other equivalent authority for technical content, competitive effect, and determination of the accuracy of the proprietary designation. Disclosures outside GEH are limited to regulatory bodies, customers, and potential customers, and their agents, suppliers, and licensees, and others with a legitimate need for the information, and then only in accordance with appropriate regulatory provisions or proprietary or confidentiality agreements.

(8) The information identified in paragraph (2), above, is classified as proprietary because it contains the detailed GEH methodology for pressure-temperature curve analysis for the GEH Boiling Water Reactor (BWR). These methods, techniques, and data along with their application to the design, modification, and analyses associated with the pressure-temperature curves were achieved at a significant cost to GEH.

The development of the evaluation processes along with the interpretation and application of the analytical results is derived from the extensive experience databases that constitute a major GEH asset.

(9) Public disclosure of the information sought to be withheld is likely to cause substantial harm to GEH's competitive position and foreclose or reduce the availability of profit-making opportunities. The information is part of GEH's comprehensive BWR safety and technology base, and its commercial value extends beyond the original development cost.

The value of the technology base goes beyond the extensive physical database and analytical methodology and includes development of the expertise to determine and apply the appropriate evaluation process. In addition, the technology base includes the value derived from providing analyses done with NRC-approved methods.

The research, development, engineering, analytical and NRC review costs comprise a substantial investment of time and money by GEH. The precise value of the expertise to devise an evaluation process and apply the correct analytical methodology is difficult to quantify, but it clearly is substantial. GEH's competitive advantage will be lost if its competitors are able to use the results of the GEH experience to normalize or verify their Affidavit for GE-PPO- 1GYEF-KG 1-720 Page 2 of 3

GE-Hitachi Nuclear Energy Americas LLC own process or if they are able to claim an equivalent understanding by demonstrating that they can arrive at the same or similar conclusions.

The value of this information to GEH would be lost if the information were disclosed to the public. Making such information available to competitors without their having been required to undertake a similar expenditure of resources would unfairly provide competitors with a windfall, and deprive GEH of the opportunity to exercise its competitive advantage to seek an adequate return on its large investment in developing and obtaining these very valuable analytical tools.

I declare under penalty of perjury that the foregoing affidavit and the matters stated therein are true and correct to the best of my knowledge, information, and belief.

Executed on this 16th day of October 2013.

Peter M. Yandow Vice President, Nuclear Plant Projects/Services Licensing, Regulatory Affairs GE-Hitachi Nuclear Energy Americas LLC 3901 Castle Hayne Rd.

Wilmington, NC 28401 Peter.Yandow @ge.com Affidavit for GE-PPO- 1GYEF-KG 1-720 Page 3 of 3