ML063560114

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Reactor Pressure Vessel Fluence Evaluation at End of Cycle 15 and 54 EFPY (Non-Proprietary)
ML063560114
Person / Time
Site: Pilgrim
Issue date: 10/25/2006
From: Watkins K
TransWare Enterprises
To:
Entergy Nuclear Operations, Office of Nuclear Reactor Regulation
References
2.06.099 ENT-FLU-001-R-003, Rev 0
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PILGRIM NUCLEAR POWER STATION REACTOR PRESSURE VESSEL FLUENCE EVALUATION AT END OF CYCLE 15 AND 54 EFPY (NON-PROPRIETARY)

TransWare Document Number: ENT-FLU-001-R-003 Revision 0 TransWare Enterprises Inc.

5450 Thornwood Drive, Suite M San Jose, California 95123 October 2006

ENT-FLU-OO.1-R-003 Revision 0 Page i-i of i-viii PILGRIM NUCLEAR POWER STATION REACTOR PRESSURE VESSEL FLUENCE EVALUATION AT END OF CYCLE 15 AND 54 EFPY (NON-PROPRIETARY).

Document Number: ENT-FLU-001-R-003 Revision 0 Preparing Organization:

TransWare Enterprises Inc.

5450 Thornwood Dr., Ste. M San Jose, California 95123 Prepared By: /0e 2 K. E. Watkins, Pro eer Date Reviewed By: e Eg Date ioJect

.N.Jnes, Reviewed By
7 V. K. JonQA Specialist Date By:

Approved es, Project Manager Da e I Prepared For:

Entergy Nuclear Operations, Inc.

Pilgrim Nuclear Power Station 600 Rocky Hill Road Ur)o\+r0 It-1eed c1wy Plymouth, Massachusetts 02360 Fo~r 1 xor~o-+/-U¶M OK.(\

Raymond Pace, Project Manager Controlled Copy Number:

This document represents the non-proprietary version of TransWare Enterprises Inc. document number ENT-FLU-001-R-001, Revision 1. Proprietary information from ENT-FLU-001-R-001, Revision 1 that has been removed is identified by enclosure in blue double brackets.

ENT-FL U-OO1-R-003 Revision 0 Page i-ii of i-viii DISCLAIMEROF WARRANTIES AND LIMITATION OF LIABILITIES THE INFORMATION CONTAINED IN THIS REPORT IS BELIEVED BY TRANSWARE ENTERPRISES INC. TO BE AN ACCURATE AND TRUEREPRESENTATION OF THE.

FACTS KNOWN, OBTAINED OR PROVIDED TO TRANSWARE ENTERPRISES INC. AT THE TIME THIS REPORT WAS PREPARED. THE USE OF THIS INFORMATION BY ANYONE OTHER THAN THE CUSTOMER OR FOR ANY PURPOSE OTHER THAN THAT FOR WHICH IT IS INTENDED, IS NOT AUTHORIZED; AND WITH RESPECT TO ANY UNAUTHORIZED USE, TRANSWARE ENTERPRISES INC. MAKES NO REPRESENTATION OR WARRANTY AND ASSUMES NO LIABILITY AS TO THE COMPLETENESS, ACCURACY OR USEFULNESS OF THE INFORMATION CONTAINED IN THIS DOCUMENT. IN NO EVENT SHALL TRANSWARE ENTERPRISES INC. BE LIABLE FOR ANY LOSS OF PROFIT OR ANY OTHER COMMERCIAL DAMAGE, INCLUDING BUT NOT LIMITED TO SPECIAL, CONSEQUENTIAL OR OTHER DAMAGES.

QUALITY REQUIREMENTS This document has been prepared in accordance with the requirements of 10CFR50 Appendix B, 10CFR21, and TransWare Enterprises Inc.'s 10CFR50 Appendix B quality assurance program.

ENT-FLU-OO1-R-003 Revision 0 Page i-iii of i-viii ACKNOWLEDGMENTS TransWare wishes to acknowledge Mr. George Mileris and Mr. Richard Weader of Entergy Nuclear Operations, Inc. for their diligence and resolve to provide the vast amount of mechanical design and operating history data detail for the Pilgrim Nuclear Power Station reactor fluence evaluation. TransWare also wishes to extend an acknowledgment to Mr. Raymond Pace for his support and management of this project.

ENT-FLU-001-R-003 Revision 0 Page i-v of i-viii CONTENTS I Introduction ............................................................................................................ 1-1 2 Summary of Results ............................................................................  :......................... 2-1 3 Description of the Reactor System .............................................................................. 3-1 3.1 Reactor System Mechanical Design Inputs ......................................................... 3-1 3.2 Reactor System Material Compositions ............................................................... 3-3 3.3 Reactor Operating Data Inputs ............................................................................ 3-4 3.3.1 Power History Data .................................................................................. 3-5 3.3.2 Reactor State-Point Data ......................................................................... 3-6 3.3.2.1 Beginning of Operation through Cycle 15 State Points ............... 3-6 3.3.2.2 Projected Operation through End of Design Life State Points .... 3-6 3.3.3 Core Loading Pattern ............................................................................... 3-7 4 Calculation Methodology .............................................................................................. 4-1 4.1 Description of the RAMA Fluence Methodology .................................................. 4-1 4.2 The RAMA Geometry Model for the Pilgrim Nuclear Power Station ................... 4-2 4.3 RAMA Calculation Parameters...................................... 4-6 4.4 RAMA Neutron Source Calculation ..................................................................... 4-7 4.5 RAMA Fission Spectra ......................................................................................... 4-7 4.6 Parametric Sensitivity Analyses ........................................................................... 4-7 5 Surveillance Capsule Activation and Fluence Results .............................................. 5-1 1))

6 Reactor Pressure Vessel Uncertainty Analysis ...................................................... 6-1 1]

7 Calculated Neutron Fluence for Reactor Pressure Vessel ........................................ 7-1 8 Re fe re n c e s ....................................................................................................................... 8 -1

ENT-FLU-OO1-R-003 Revision 0 Page i-vi of i-viii A Cycle 4 Surveillance Capsule Evaluation ................................................................ A-1 A.1 Comparison of Predicted Activation to Plant-Specific Measurements ........... A-i

))

ENT-FLU-OO1-R-003 Revision 0 Page i-vii of i-viii LIST OF FIGURES Figure 3-1 Planar View of the Pilgrim Nuclear Power Station at the Core Mid-plane Elevation 3-2 Figure 3-2 Pilgrim Nuclear Power Station RPV Shell Plates and Weld Location Identifiers ...... 3-3 1]

ENT-FLU-OO1-R-003 Revision 0 Page i-viii of i-viii LIST OF TABLES Table 2-1 Maximum >1.0 MeV Neutron Fluence for Pilgrim Nuclear Power Station RPV Weld and Shell Locations at the Inner Surface ......................................................... 2-1 Table 3-1 Summary of Material Compositions by Region for the Pilgrim Nuclear Power S ta tio n ........................................................................................................................ 3 -4 Table 3-2 Number of State-point Data Files for Each Cycle in the Pilgrim Nuclear Power S ta tio n ........................................................................................................................ 3 -5 Table 3-3 Summary of the Pilgrim Nuclear Power Station Core Loading Pattern ..................... 3-8

((I 1]

Table 7-1 Maximum >1.0 MeV Neutron Fluence in Pilgrim Nuclear Power Station RPV S hells at EO C 15 (20.7 E FPY) ................................................................................... 7-2 Table 7-2 Maximum >1.0 MeV Neutron Fluence in Pilgrim Nuclear Power Station RPV Welds at EO C 15 (20.7 EFPY) .................................................................................. 7-2 Table 7-3 Maximum >1.0 MeV Neutron Fluence in Pilgrim Nuclear Power Station RPV S hells at 54 E F PY ................................................................................................ . . 7-3 Table 7-4 Maximum >1.0 MeV Neutron Fluence in Pilgrim Nuclear Power Station RPV W e lds at 54 E F PY ...................................................................................................... 7-3 Table 7-5 Neutron Fluence Profile for Pilgrim Nuclear Power Station RPV Circumferential Welds 1-344 and 3-339 B at the Inner Surface for Energy >1.0 MeV at EOC 15 (2 0 .7 E F PY ) ................................................................................................................ 7 -4 Table 7-6 Neutron Fluence Profile for Pilgrim Nuclear Power Station RPV Circumferential Welds 1-344 and 3-339 B at the Inner Surface for Energy >1.0 MeV at 54 EFPY... 7-6 Table 7-7 Neutron Fluence Profile for Pilgrim Nuclear Power Station RPV Vertical Welds 1-338 A, 1-338 B, and 1-338 C at the Inner Surface for Energy >1.0 MeV at E O C 15 (20 .7 E FP Y) .................................................................................................. 7-8 Table 7-8 Neutron Fluence Profile for Pilgrim Nuclear Power Station RPV Vertical Welds 1-338 A, 1-338 B, and 1-338 C at the Inner Surface for Energy >1.0 MeV at 54 E F P Y ................................................................................................................... 7-1 5 11

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INTRODUCTION This report presents the results of the reactor pressure vessel (RPV) fluence evaluation performed for the Pilgrim Nuclear Power Station (hereinafter referred to as the "Reactor"). Reactor pressure vessel fluence is determined for energy >1.0 MeV at welds and shells in the reactor pressure vessel beltline region. Projected fluence values are presented for energy >1.0 MeV at two points in time: 1) the end of the Reactor's operating cycle 15; and 2) the end of the design lifetime of 54 effective full power years (EFPY). This evaluation was performed in accordance with guidelines presented in U. S. Nuclear Regulatory Guide 1.190 [1] for use in evaluating the effects of embrittlement on RPV material in the reactor beltline region as required in 10CFR50 Appendix G. The reactor beltline region, as defined in Appendices G and H of 10CFR50, includes the region that directly surrounds the effective height of the reactor core as well as those areas of the RPV that exceed a neutron fluence (E>1.0 MeV) of 1.00E+17 n/cm2 .

Neutron fluence profiles are presented azimuthally in one degree increments at the inner surface of the RPV wall for the RPV circumferential welds in the reactor beltline region. In addition,

,inner surface values are presented for the RPV beltline vertical weld locations for the entire height of the weld in the RAMA model in one inch increments.

The fluence values presented in Sections 2 and 7 of this report represent the RAMA best estimate values, assuming no statistical bias. Appendix A reports the adjustments needed to obtain best estimate fluence values with a bias, as outlined in U. S. Regulatory Guide 1.190 [1], if the Reactor's cycle 4 surveillance capsule evaluation is included.

The fluence values presented in this report were calculated using the RAMA Fluence Methodology [2]. The RAMA Fluence Methodology (hereinafter referred to as the Methodology) has been developed for the Electric Power Research Institute, Inc. (EPRI) and the Boiling Water Reactor Vessel and Internals Project (BWRVIP) for the purpose of calculating neutron fluence in Boiling Water Reactor (BWR) components.

The Methodology has been approved by the U. S. Nuclear Regulatory Commission [3] for application in accordance with U. S. Regulatory Guide 1.190 [ 1]. Benchmark testing has been performed using the Methodology for several surveillance capsule and reactor pressure vessel fluence evaluations. ((

))

The information and associated evaluations provided in this report have been performed in accordance with the requirements of 10CFR50 Appendix B.

ENT-FLU-OO1-R-003 Revision 0 Page 2-1 of 2-2 2

SUMMARY

OF RESULTS This section provides a summary of the results of the reactor pressure vessel fluence evaluation for the Reactor. The primary purpose of this evaluation is to determine the reactor pressure vessel fluence for energy range >1.0 MeV at selected welds and shells in the reactor pressure vessel beltline region. Fluence values are calculated at the end of operating cycle (EOC) 15 and projected through the end of design life at 54 EFPY. Fluence is calculated at the inner surface (OT) and 1/4T locations for each RPV weld and shell. Detailed tables of all results are presented in Section 7 of this report. Section 7 also contains detailed tables of the predicted RPV neutron fluence at one degree increments in the azimuth for the RPV circumferential welds in the core beltline region; and fluence profiles for RPV vertical welds in the core beltline at one inch increments over the entire height of the weld included in the RAMA model at the inner surface of the RPV wall.

Table 2-1 summarizes the peak fluence results for this evaluation for energy >1.0 MeV at EOC 15 (20.7 EFPY) and 54 EFPY. One value represents the peak fluence for the weld locations and the other represents the value at the shell locations. Note that the peak fluence for both the RPV welds and shells occurs at the inner surface (OT) at the point closest to the edge of the core. The peak fluence for the weld locations is in vertical welds 1-338 A and 1-338 C with a value of 1.14E+1 8 n/cm 2 at 54 EFPY. The peak fluence for the RPV shells is in the lower intermediate shell with a value of 1.28E+18 n/cm 2 at 54 EFPY. Figure 3-2 illustrates the positioning of the shells in the reactor.

2 It was observed that the threshold fluence value of 1.OOE+I 7 n/cm was reached prior to the end of operating cycle 15 (20.7 EFPY) at the OT and 1/4T thickness in a majority of the lower intermediate shell locations. Several locations in the lower shell exceed the threshold value, but no welds in upper shell are expected to exceed 1.00E+1 7 n/cm 2 during the Reactor's design life.

The elevation range over which the fluence value exceeds 1.OOE+17 n/cm 2 is 512.13 cm (201.63 inches) to 929.95 cm (366.12 inches) for 54 EFPY.

It was determined that the recirculation inlet (jet pump) nozzle, with a central axis elevation of 508 cm, was positioned within the bounds of the 1.OE+17 n/cm 2 threshold range. The peak fluence value is along the upper edge of the nozzle with a value of 1.30E+17 n/cm 2 for EOC 15 and 2.8 1E+17 n/cm 2 for 54 EFPY.

ENT-FLU-OO1-R-003 Revision 0 Page 2-2 of 2-2 Table 2-1 Peak >1.0 MeV Neutron Fluence for Pilgrim Nuclear Power Station RPV Weld and Shell Locations at the Inner Surface Peak Fluence for Elevation P Peak Fluence for 2

Weld/Shell Location EOC 15 (20.7 EFPY) M )

[cm (in)] (n/cm2) 54 EFPY (n/cm_)

Weld 1-338 A / C 776.94 (305.88) 4.96E+17 1.14E+18 Lower(250 Intermediate Shell azimuh) 791.21 (311.50) 5.51 E+17 1.28E+18 (250 azimuth)

I((

I]

This neutron fluence evaluation for the Pilgrim Nuclear Power Station RPV shell and weld locations has been performed in accordance with the guidelines presented in Regulatory Guide 1.190.

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DESCRIPTION OF THE REACTOR SYSTEM This section describes the Reactor fluence model used in the reactor pressure vessel fluence evaluation. The fluence model is based on plant-specific basic design inputs that include component mechanical designs, material compositions, and reactor operating history. Plant-specific mechanical design drawings and structural material data were provided by Entergy Nuclear Operations, Inc. [5] and were used to build the Pilgrim Nuclear Power Station RAMA geometry model. Detailed operating history data from core simulator models was provided for cycles 1 through 15 [6].

3.1 Reactor System Mechanical Design Inputs The Reactor is modeled with the RAMA Fluence Methodology. The Methodology employs a three-dimensional modeling technique to describe the reactor geometry for the neutron transport calculations. Detailed mechanical design information is used in order to build an accurate three-dimensional RAMA computer model of the reactor system. The mechanical design information for Pilgrim Nuclear Power Station was provided by Entergy Nuclear Operations, Inc. [5].

Pilgrim Nuclear Power Station is a General Electric BWR/3 class reactor with core loading of 580 fuel assemblies. The rated thermal power output of the reactor was 1998 MWt for cycles 1-14. A power up-rate was achieved in cycle 15, raising the power to 2028 MWt.

Figure 3-1 illustrates the basic planar geometry configuration of the reactor at the axial elevation corresponding to the core mid-plane. This figure identifies the positioning of the surveillance capsules relative to the inside surface of the reactor pressure vessel wall. All radial regions comprising the fluence model are illustrated. Beginning at the center of the reactor and projecting outwards, the regions include: the core region; core reflector region (bypass water); central shroud wall; downcomer water region including the jet pumps; reactor pressure vessel (RPV) wall; mirror insulation; biological shield (concrete wall); and cavity regions between the RPV and biological shield. Also shown are the azimuthal positions of the surveillance capsules in the downcomer region at 120, 210, and 300 degrees and the jet pump assemblies at 30, 60, 90, 120, 150, 210, 240, 270, 300, and 330 degrees. Shroud repair tie rods were installed in the reactor at the beginning of cycle 11 and are positioned at 45, 135, 225 and 315 degrees.

ENT-FLU-OOI-R-003 Revision 0 Page 3-2 of 3-8 0* North 398.78 cm 270* -

1800 F = Fuel assembly locations.

Biological Shield (Locations shown only for the northeast quadrant.)

+ = Control rod locations.

Figure 3-1 Planar View of the Pilgrim Nuclear Power Station Reactor at the Core Mid-plane Elevation As the primary interest in this fluence evaluation is the determination of the neutron fluence at specified RPV welds and shells, Figure 3-2 identifies these specific weld locations. These weld locations are referenced in the tables in Sections 2 and 7 of this report showing the RPV weld fluence results by their identification numbers. ((

ENT-FLU-OOI-R-003 Revision 0 Page 3-3 of 3-8 180" 210" 240" 270" 300" 330" 0! 30" 60° 90! 120' 150" 1806 3-339 A 1 - 1295.25 cm Q Upper Shell 3-339 B " -- 1010.92 cm U

0 Lower Intermediate Shell 1-344 616.92 cm 00 00 I - Lower Shell 00 00 0 9-338 -- 270.21 cm 180' 210' 240" 270" 300" 330" 0' 30" 60" 90" 120" 150" 180" Inside View Figure 3-2 Pilgrim Nuclear Power Station RPV Shell Plates and Weld Location Identifiers 3.2 Reactor System Material Compositions Each region of the reactor is comprised of materials that include reactor fuel, steel, water, insulation, concrete, and air. Accurate material information is essential for the fluence evaluation as the material compositions determine the neutron source, scattering, and absorption of neutrons throughout the reactor system and, thus, affect the determination of neutron fluence in the reactor components.

Table 3-1 provides a summary of the material compositions in the various components and regions of the Reactor. The attributes for the steel, insulation, concrete, and air compositions (i.e., material densities and isotopic concentrations) are assumed to remain constant for the operating life of the reactor. The attributes for the ex-core water compositions will vary with the operation of the reactor((

)). The attributes of the fuel compositions in the reactor core region change continuously during an operating cycle due to changes in power level, fuel burnup, control rod movements, and changing moderator density levels (voids).

((

ENT-FLU-OOI-R-003 Revision 0 Page 3-4 of 3-8 Table 3-1 Summary of Material Compositions by Region for the Pilgrim Nuclear Power Station Region Material Composition 2 35 23 9 24 0 2 42 Reactor Core U, 238U, pu, Pu, 241Pu, pU, Ofuel, Zr, Water Core Reflector Water Fuel Support Piece Stainless Steel Lower Tie Plate Stainless Steel, Zr, Inconel Top Guide Stainless Steel Upper Tie Plate Stainless Steel, Zr, Inconel Shroud Stainless Steel Shroud Repair Tie Rods (*) Stainless Steel Downcomer Region Water Jet Pump Riser and Mixer Flow Areas Water Jet Pump Riser and Mixer Metal Stainless Steel Surveillance Capsule Specimens Carbon Steel Reactor Pressure Vessel Clad Stainless Steel Reactor Pressure Vessel Wall Carbon Steel Cavity Regions Air Insulation Clad Stainless Steel Insulation Stainless Steel, Aluminum, Air Biological Shield Clad Carbon Steel Biological Shield Reinforced Concrete

(*) The shroud repair tie rods were introduced into the reactor at the start of cycle 11.

3.3 Reactor Operating Data Inputs An accurate evaluation of fluence in the reactor requires an accurate accounting of the reactor operating history. The primary reactor operating parameters that affect neutron fluence evaluations for BWR's include the reactor power level, core power distribution, core void fraction distribution (or equivalently, water density distribution), and fuel material distribution.

ENT-FL U-OO1-R-003 Revision 0 Page 3-5 of3-8 3.3.1 PowerHistory Data The reactor power history used in the Reactor component fluence evaluation was obtained from Entergy Nuclear Operations, Inc. for operating cycles 1 through 15 [6f. The power history data accounts for the Reactor shutdown periods. The shutdowns were primarily due to the refueling outages between cycles. Table 3-2 provides the accumulated effective full power years of power generation at the end of each cycle in this fluence evaluation.

Table 3-2 Number of State-point Data Files for Each Cycle in the Pilgrim Nuclear Power Station Number of State- Rated Thermal Power Accumulated Effective Full Point Data Files MWt Power Years (EFPY) 1 1998 0.9 2 1998 1.7 3 1998 2.6 4 1998 4.2 5 1998 5.3 6 1998 6.7 7 1998 7.7 8 1998 9.1 9 1998 10.4 10 1998 11.8 11 1998 13.4 12 1998 15.3 13 1998 17.0 14 1998 18.8 15 2028 20.7

>15 (Projected) )) (( )) 54.0

ENT-FLU-OO1-R-003 Revision 0 Page3-6 of 3-8 3.3.2 Reactor State-PointData Reactor operating data for the Pilgrim Nuclear Power Station RPV fluence evaluation was provided as state-point data files by Entergy Nuclear Operations, Inc. [6]. The state-point files are generated by three-dimensional core simulator models and provide a best-available representation of the operating conditions of the unit over the lifetime of the Reactor. The data files include three-dimensional data arrays that describe the fuel materials, moderator materials, and the relative power distribution in the core region.

A separate neutron transport calculation was performed for each of the available state points. The calculated neutron flux for each state point was combined with the appropriate power history data described in Section 3.3.1 to calculate the neutron fluence in the reactor pressure vessel.

3.3.2.1 Beginning of Operation through Cycle 15 State Points A total of (( )) state-point data files were used to represent the first fifteen operating cycles of the Pilgrim Nuclear Power Station. Table 3-2 shows the number of state points used for each cycle in this fluence evaluation. The rated thermal power output of the Reactor for operating cycles 1 through 14 is specified as 1998 MWt. For operating cycle 15 the rated thermal power output is specified as 2028 MWt due to a power up-rate at the beginning of cycle 15.

3.3.2.2 Projected Operation through End of Design Life State Points

((

))

This analysis predicts fluence at the end of the last completed operating cycle at the time of the analysis and projects fluence to the end of the reactor licensed lifetime. With these two data points, it is common practice to use linear interpolation techniques to determine RPV and component fluence at any time in between. While the fluence at the end of the last completed operating cycle is based upon historical operating conditions, the projected fluence assumes an "equilibrium cycle" strategy. If future reactor cycles deviate from the equilibrium cycle, the use of linear interpolation techniques may produce inaccurate results. It is recommended that each new operating cycle be evaluated for potential impact on the projected fluence and that the fluence analysis be updated accordingly. Deviations from equilibrium cycle conditions can be incurred as the result of, for example, changes in core management strategies, power uprates, new fuel designs, and revised heat balances.

ENT-FLU-OO1-R-003 Revision 0 Page 3-7 of 3-8 3.3.3 Core Loading Pattern It is common in BWRs that more than one fuel assembly design will be loaded in the reactor core in any given operating cycle. For fluence evaluations, it is important to account for the fuel assembly designs that were loaded in the core in order to accurately represent the neutron source distribution at the core boundaries (i.e., peripheral fuel locations, the top fuel nodes, and the bottom fuel nodes).

Four different fuel assembly designs are used in the Reactor during cycles 1 through 15. Table 3-3 provides a summary of the fuel designs loaded in the reactor core for these operating cycles.

The cycle core loading patterns provided by Entergy Nuclear Operations, Inc. are used to identify the fuel assembly designs in each cycle and their location in the core loading pattern.

For each cycle, appropriate fuel assembly models are used to describe the reactor core region of the RAMA fluence model for the Reactor.

ENT-FLU-OO1-R-003 Revision 0 Page 3-8 of 3-8 Table 3-3 Summary of the Pilgrim Nuclear Power Station Core Loading Pattern Number of Number of Number of Number of Dominant General Electric General Electric General Electric General Electric Peripheral Cycle (GE) 7x7 Fuel (GE) 8x8 Fuel (GE) 9x9 Fuel (GE) Ox1O Fuel Fuel Design Assembly Assembly Assembly Assembly in the RAMA Designs Designs Designs Designs Model 1 580 0 0 0 GE7x7 2 560 20 0 0 GE 7x7 3 428 152 0 0 GE 7x7 4 0 580 0 0 GE 8x8 5 0 580 0 0 GE 8x8 6 0 580 0 0 GE 8x8 7 0 580 0 0 GE 8x8 8 0 580 0 0 GE 8x8 9 0 580 0 0 GE 8x8 10 0 580 0 0 GE 8x8 11 0 444 136 0 GE 8x8 12 0 236 344 0 GE 8x8 13 0 76 504 0 GE 8x8 14 0 0 436 144 GE 9x9 15 0 0 272 308 GE 9x9

>15 ((] ((] ,((]((] ((]

.1]

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CALCULATION METHODOLOGY The Reactor RPV fluence evaluation was performed using the RAMA Fluence Methodology software package [2]. The Methodology and the application of the Methodology to the Reactor are described in this section.

4.1 Description of the RAMA Fluence Methodology The RAMA Fluence Methodology is a system of codes that is used to perform fluence evaluations in light water reactor components. The Methodology includes a transport code, model builder codes, a fluence calculator code, an uncertainty methodology, and a nuclear data library. The transport code, fluence calculator, and nuclear data library are the primary software components for calculating the neutron flux and fluence. The transport code uses a deterministic, three-dimensional, multigroup nuclear particle transport theory to perform the neutron flux calculations. The transport code couples the nuclear transport method with a general geometry modeling capability to provide a flexible and accurate tool for calculating fluxes in light water reactors. The fluence calculator uses reactor operating history information with isotopic production and decay data to estimate activation and fluence in the reactor components over the operating life of the reactor. The nuclear data library contains nuclear cross-section data and response functions that are needed in the flux, fluence, and reaction rate calculations. The cross sections and response functions are based on the BUGLE-96 nuclear data library [7]. The Methodology and procedures for its use are described in the following reports: Theory Manual

[8] and Procedures Manual [9].

The primary inputs for the RAMA Fluence Methodology are mechanical design parameters and reactor operating history data. The mechanical design inputs are obtained from reactor design drawings (or vendor drawings) of the plant. The reactor operating history data is obtained from reactor core simulation calculations, system heat balance calculations, and daily operating logs that describe the operating conditions of the reactor.

The primary outputs from the RAMA Fluence Methodology calculations are neutron flux, neutron fluence, and uncertainty determinations. The RAMA transport code calculates the neutron flux distributions that are used in the determination of neutron fluence. Several transport calculations are typically performed over the operating life of the reactor in order to calculate neutron flux distributions that accurately characterize the operating history of the reactor. The post-processing code (RAFTER) is then used to calculate component fluence and nuclide activations using the neutron flux solutions from the transport calculations and daily operating history data for the plant. The fluence calculated by RAFTER may then be adjusted in accordance with the calculational bias to determine the best estimate fluence and uncertainty in accordance with the intent of U. S. Nuclear Regulatory Guide 1.190.

ENT-FLU-OO1-R-003 Revision 0 Page 4-2 of 4-8 4.2 The RAMA Geometry Model for the Pilgrim Nuclear Power Station The RAMA Fluence Methodology uses a flexible three-dimensional modeling technique to describe the reactor geometry. The geometry modeling technique is based on the Cartesian coordinate system in which the (x,y) coordinates describe an axial plane of the reactor system and the z-axis describes elevations of the reactor system.

)) The pressure vessel has cladding on the wall inner surface. The biological shield has cladding on the inner and outer surfaces. The downcomer region includes representations for the jet pumps, surveillance capsules, and, from cycle 11 onward, shroud repair tie rods.

E[

Each of the components and regions that extend outward from the core region are modeled in their correct geometrical form. ((

)) The riser pipe is correctly situated on a curvilinear path between the centers of the mixer pipes.

[R

)) Downcomer water surrounds the capsule on all sides.

1[

)) The tie rods, springs, and stabilizers are surrounded by downcomer water on all sides.

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((E 1]

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)) The 10 jet pump assemblies are positioned azimuthally at 30, 60, 90, 120, 150, 210, 240, 270, 300, and 330 degrees. ((

1]

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ENT-FLU-OO1-R-003 Revision 0 Page 4-6 of 4-8 I[

4.3 RAMA Calculation Parameters The RAMA transport code uses a three-dimensional deterministic transport method to calculate neutron flux distributions in reactor problems. ((

1]

The RAMA transport calculation also uses information from the RAMA nuclear data library to determine the scope of the flux calculation. This information includes a Legendre expansion of the scattering cross sections that is used in the treatment of anisotropy of the problem. ((

))

The neutron flux is calculated using an iterative technique to obtain a converged solution for the problem. ((

The impact of these calculation parameter selections on the RAMA fluence evaluation for the Pilgrim Nuclear Power Station is presented in Section 4.6.

ENT-FLU-OO1-R-003 Revision 0 Page 4-7 of 4-8 4.4 RAMA Neutron Source Calculation The neutron source for the RAMA transport calculation is calculated using the input relative power density factors for the different fuel regions and data from the RAMA nuclear data library.

The core neutron source is determined using the cycle-specific three-dimensional burnup distributions. ((

I))

4.5 RAMA Fission Spectra

[4 1]

4.6 Parametric Sensitivity Analyses Several sensitivity analyses were performed to evaluate the stability and accuracy of the RAMA transport calculation for the Pilgrim Nuclear Power Station model. ((

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+

  • 4
  • 11

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SURVEILLANCE CAPSULE ACTIVATION AND FLUENCE RESULTS

)) This section addresses the evaluation of the Pilgrim Nuclear Power Station surveillance capsule flux wires and the comparison to measurements. The flux wires were installed at the start of commercial operation and were removed at the end of cycle 4 after being irradiated for a total of 4.17 effective full power years (EFPY). ((

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REACTOR PRESSURE VESSEL UNCERTAINTY ANALYSIS

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ENT-FLU-OO1-R-003 Revision 0 Page 6-2 of 6-3 t I t *1 I .1 I- 4 I- I I- 4 I 4 I. 4 I- I I- I

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ENT-FLU-OO1-R-003 Revision 0 Page 7-1 of 7-21 7

CALCULATED NEUTRON FLUENCE FOR REACTOR PRESSURE VESSEL The neutron fluence for the reactor pressure vessel at the inner vessel wall (OT) and at 1/4T is determined by the RAMA Fluence Methodology for two points in time: at the end of cycle 15 (EOC 15) and through the projected end of normal operating life at 54 EFPY. The results of the fluence evaluation for energy >1.0 MeV are presented in the tables that follow.

Tables 7-1 and 7-4 report the >1.0 MeV fluence in the RPV shell and weld locations for the RPV beltline region. The location and identification of the RPV welds and shells are shown in Figure 3-2. It is observed that the fluence is greatest at the inner surface (OT) at the point closest to the edge of the core for all RPV welds and shells. The maximum fluence for the RPV welds is at the inner diameter of vertical welds 1-338 A and 1-338 C with a value of 1.14E+18 n/cm 2 at 54 EFPY. The maximum fluence for the RPV shells is at the lower intermediate shell with a value of 1.28E+18 n/cm 2 at 54 EFPY.

It was observed that the threshold fluence value of 1.00E+l 7 n/cm2 was reached prior to the end of operating cycle 15 (20.7 EFPY) at the OT and 1/4T thickness in a majority of the lower intermediate shell locations. Several locations in the lower shell exceed the threshold value, but no welds in the upper shell are expected to exceed 1.OOE+I 7 n/cm 2 during the Reactor's design life. The elevation range over which the fluence value exceeds 1.OOE+17 n/cm2 is 512.13 cm (201.63 inches) to 929.95 cm (366.12 inches) for 54 EFPY.

Tables 7-5 and 7-6 present the fluence profiles for the RPV circumferential welds 1-344 and 3-339 B for energy >1.0 MeV at EOC 15 and 54 EFPY, respectively. The fluence values for the circumferential welds are determined from azimuth 0 to 45 degrees in one degree increments at the inner surface of the RPV wall (OT). In Tables 7-5 and 7-6 the peak fluence values are shown in bold text.

The fluence values for each RPV vertical weld in the RPV beltline region is determined for the entire height of the weld represented in the RAM\A model. Fluence is calculated in one degree increments at the inner surface of the RPV wall. Tables 7-7 and 7-8 present the fluence profiles for vertical welds 1-338 A, 1-338 B and 1-338 C for energy >1.0 MeV at EOC 15 and 54 EFPY, espectively. In Tables 7-7 and 7-8 the peak fluence values are shown in bold text.

It was determined that the recirculation inlet (jet pump) nozzle, with a central axis elevation of 508 cm, was positioned within the bounds of the 1.0E+17 n/cm 2 threshold range. The peak fluence value is along the upper edge of the nozzle with a value of 1.30E+17 n/cm 2 for EOC 15 and 2.81E+17 n/cm 2 for 54 EFPY.

ENT-FLU-001-R-003 Revision 0 Page 7-2 of 7-21 Table 7-1 Maximum >1.0 MeV Neutron Fluence in Pilgrim Nuclear Power Station RPV Shells at EOC 15 (20.7 EFPY)

Shell Location Fluence (n/cm 2) Fluence (n/cm2 )

OT 1/4T Lower 4.87E+17 2.93E+17 Lower Intermediate 5.51 E+ 17 3.63E+17 Upper 2.45E+15 1.71E+15 Table 7-2 Maximum >1.0 MeV Neutron Fluence in Pilgrim Nuclear Power Station RPV Welds at EOC 15 (20.7 EFPY)

Fluence (n/cm2) Fluence (n/cm2 )

OT at weld 114T at weld Lower 2-338 A 2.93E+17 1.80E+17 2-338 B 3.94E+17 2.41 E+17 2-338 C 2.99E+17 1.83E+17 1-344 (Circumferential) 4.19E+17 2.74E+ 17 Lower Intermediate 1-338 A 4.96E+17 3.31E+17 1-338 B 2.22E+17 1.50E+17 1-338 C 4.96E+ 17 3.31E+17 3-339 B (Circumferential) 2.45E+15 1.71 E+15 Upper 2-339 A 2.06E+15 1.44E+15 2-339 B 2.45E+15 1.71E+15 2-339 C 2.36E+15 1.66E+15

ENT-FLU-OO1-R-003 Revision 0 Page 7-3 of 7-21 Table 7-3 Maximum >1.0 MeV Neutron Fluence in Pilgrim Nuclear Power Station RPV Shells at 54 EFPY Shell Location Fluence (n/cm2) Fluence (n/cm 2)

OT 1/4T Lower 1.01E+18 6.09E+17 Lower Intermediate 1.28E+18 8.40E+17 Upper 5.12E+15 3.59E+15 Table 7-4 Maximum >1.0 MeV Neutron Fluence in Pilgrim Nuclear Power Station RPV Welds at 54 EFPY Shell Location Weld Fluence (n/cm 2) Fluence (n/cm 2)

OT at weld 114T at weld Lower 2-338 A 6.08E+17 3.74E+ 17 2-338 B 8.17E+17 5.OOE+17 2-338 C 6.72E+17 4.11E+17 1-344 (Circumferential) 8.69E+17 5.70E+17 Lower Intermediate 1-338 A 1.14E+18 7.63E+17 1-338 B 5.17E+17 3.51EE+ 17 1-338 C 1.14E+18 7.63E+17 3-339 B (Circumferential) 5.12E+15 3.59E+15 Upper 2-339 A 4.28E+15 3.02E+15 2-339 B 5.12E+15 3.59E+15 2-339 C 5.00E+15 3.52E+15

ENT-FLU-OO1-R-003 Revision 0 Page 7-4 of 7-21 Table 7-5 Neutron Fluence Profile for Pilgrim Nuclear Power Station RPV Circumferential Welds 1-344 and 3-339 B at the Inner Surface for Energy >1.0 MeV at EOC 15 (20.7 EFPY)

Azimuth (degrees) Weld 1-344 Fluence (n/cm2 ) Weld 3-339 B Fluence (nlcm 2) 0 1.71E+17 1.94E+15 1 1.74E+17 1.97E+15 2 1.76E+17 2.01E+15 3 1.80E+17 2.03E+15 4 1.84E+17 2.06E+15 5 1.88E+17 2.07E+15 6 1.94E+17 2.07E+15 7 2.OOE+17 2.07E+15 8 2.09E+17 2.08E+15 9 2.19E+17 2.11E+15 10 2.29E+17 2.13E+15 11 2.40E+17 2.14E+15 12 2.54E+17 2.17E+15 13 2.69E+17 2.20E+15 14 2.84E+17 2.22E+15 15 2.99E+17 2.25E+15 16 3.14E+17 2.28E+15 17 3.30E+17 2.30E+15 18 3.42E+17 2.33E+15 19 3.53E+17 2.36E+15 20 3.62E+17 2.37E+15 21 3.73E+17 2.38E+15 22 3.87E+17 2.38E+15 23 4.07E+17 2.42E+15 24 4.16E+17 2.44E+15 25 4.19E+17 2.45E+15 26 4.13E+17 2.45E+15 27 4.03E+17 2.43E+15 28 3.92E+17 2.41E+15 29 3.82E+17 2.41E+15

ENT-FLU-OO1-R-003 Revision 0 Page 7-5 of7-21 Table 7-5 (Continued)

Neutron Fluence Profile for Pilgrim Nuclear Power Station RPV Circumferential Welds 1-344 and 3-339 B at the Inner Surface for Energy >1.0 MeV at EOC 15 (20.7 EFPY)

Azimuth (degrees) Weld 1-344 Fluence (n/cm 2) Weld 3-339 B Fluence (n/cm 2) 30 3.70E+17 2.40E+15 31 3.58E+17 2.38E+15 32 3.43E+17 2.37E+15 33 3.31E+17 2.36E+15 34 3.21E+17 2.36E+15 35 3.12E+17 2.36E+15 36 3.03E+17 2.35E+15 37 2.92E+17 2.33E+15 38 2.81E+17 2.29E+15 39 2.73E+17 2.26E+15 40 2.67E+17 2.24E+15 41 2.61E+17 2.23E+15 42 2.59E+17 2.20E+15 43 2.58E+17 2.17E+15 44 2.54E+17 2.15E+15 45 2.50E+17 2.14E+15

ENT-FLU-OO1-R-003 Revision 0 Page 7-6 of 7-21 Table 7-6 Neutron Fluence Profile for Pilgrim Nuclear Power Station RPV Circumferential Welds 1-344 and 3-339 B at the Inner Surface for Energy >1.0 MeV at 54 EFPY Azimuth (degrees) Weld 1-344 Fluence (nlcm 2) Weld 3-339 B Fluence (n/cm 2) 0 3.59E+17 4.04E+15 1 3.63E+17 4.11E+15 2 3.68E+17 4.18E+15 3 3.75E+17 4.24E+15 4 3.84E+17 4.28E+15 5 3.92E+17 4.31E+15 6 4.04E+17 4.31E+15 7 4.18E+17 4.31E+15 8 4.37E+17 4.34E+15 9 4.57E+17 4.39E+15 10 4.78E+17 4.45E+15 11 5.OOE+17 4.47E+15 12 5.28E+17 4.52E+15 13 5.58E+17 4.58E+15 14 5.88E+17 4.62E+15 15 6.19E+17 4.69E+15 16 6.51E+17 4.76E+15 17 6.85E+17 4.78E+15 18 7.10E+17 4.85E+15 19 7.33E+17 4.92E+15 20 7.52E+17 4.95E+15 21 7.74E+17 4.95E+15 22 8.01E+17 4.97E+15 23 8.41E+17 5.04E+15 24 8.61E+17 5.09E+15 25 8.69E+17 5.12E+15 26 8.59E+17 5.12E+15 27 8.41E+17 5.10E+15 28 8.21E+17 5.07E+15 29 8.02E+17 5.08E+15

ENT-FLU-OO1-R-003 Revision 0 Page 7-7 of 7-21 Table 7-6 (Continued)

Neutron Fluence Profile for Pilgrim Nuclear Power Station RPV Circumferential Welds 1-344 and 3-339 B at the Inner Surface for Energy >1.0 MeV at 54 EFPY Azimuth (degrees) Weld 1-344 Fluence (n/cm2) Weld 3-339 B Fluence (n/cm2 )

30 7.83E+17 5.06E+15 31 7.62E+17 5.03E+15 32 7.37E+17 5.03E+15 33 7.17E+17 5.01E+15 34 6.99E+17 5.OOE+15 35 6.85E+17 5.01E+15 36 6.68E+17 4.99E+15 37 6.49E+17 4.94E+15 38 6.30E+17 4.85E+15 39 6.16E+17 4.76E+15 40 6.03E+17 4.66E+15 41 5.91E+17 4.58E+15 42 5.83E+17 4.49E+15 43 5.74E+17 4.40E+15 44 5.55E+17 4.32E+15 45 5.36E+17 4.25E+15

ENT-FLU-OOI-R-003 Revision 0 Page 7-8 of 7-21 Table 7-7 Neutron Fluence Profile for Pilgrim Nuclear Power Station RPV Vertical Welds 1-338 A, 1-338 B, and 1-338 C at the Inner Surface for Energy >1.0 MeV at EOC 15 (20.7 EFPY)

Elevation cm Weld 1-338 A Fluence Weld 1-338 B Fluence Weld 1-338 C Fluence and (in) (n/cm 2) (n/cm 2) (nlcm 2) 616.92 37E1 (242 3.70E+17 1.71E+17 3.70E+17 (242.88) 619.46 (24. 3.74E+17 1.73E+17 3.74E+17 (243.88) 622.00 3.77E+17 1.74E+17 3.77E+17 (244.88) 624.543.0+7 (245. 3.80E+17 1.75E+17 3.80E+17 (245.88) 627.08 3.83E+17 1.77E+17 3.83E+17 (246.88) 629.6238617 (24.88 3.86E+17 1.78E+17 3.86E+17 (247.88) 632.16 3.90E+17 1.79E+17 3.90E+1 7 (248.88) 634.70 39E1 (24.88 3.93E+17 1.81E+17 3.93E+17 (249.88) 637.24 3.96E+17 1.82E+17 3.96E+17 (250.88) 639.78 3.99E+17 1.84E+17 3.99E+17 (251.88) 642.32 (252 4.02E+17 1.85E+17 (252.88) 4.02E+17 644.86 (25.88 4.05E+17 1.86E+17 4.05E+17 (253.88) 647.40 40E1 (4.48 4.08E+17 1.88E+17 4.08E+17 (254.88) 649.94 625.88 4.11 E+17 1.89E+17 (255.88) 4.11E+17 652.48 (2.88 4.14E+17 1.90E+17 (256.88) 4.14E+17 655.02 625.88 4.16E+17 1.91E+17 (257.88) 4.16E+17 657.56 41E1 (58.88 4.19E+17 1.93E+17 (258.88) 4.19E+17 660.10 259.88 4.21E+17 1.94E+17 4.21E+17 (259.88) 662.64 (260 4.24E+17 1.95E+17 (260.88) 4.24E+17 665.18 (618 4.26E+17 1.96E+17 (261.88) 4.26E+1 7 667.72 262.88 4.29E+17 1.97E+17 (262.88) 4.29E+17 670.26 (26 4.32E+17 1.98E+17 (263.88) 4.32E+17 672.80434+7 (2.88 4.34E+17 2.OOE+17 4.34E+17 (264.88)1

ENT-FLU-001-R-003 Revision 0 Page 7-9 of 7-21 Table 7-7 (Continued)

Neutron Fluence Profile for Pilgrim Nuclear Power Station RPV Vertical Welds 1-338 A, 1-338 B, and 1-338 C at the Inner Surface for Energy >1.0 MeV at EOC 15 (20.7 EFPY)

Elevation cm Weld 1-338 A Fluence Weld 1-338 B Fluence Weld 1-338 C Fluence and (in) (n/cm2) (n/cm2) (n/cm2) 675.34 4.37E+17 2.01E+17 4.37E+17 (265.88) 677.88 4.40E+17 (266.88) 680.42 4.42E+17 2.03E+17 4.42E+17 (267.88) 682.96 68.88 4.44E+17 2.04E+17 4.44E+17 (268.88) 685.50 69.88 4.46E+17 2.05E+17 4.46E+17 (269.88) 688.04 (88) 4.49E+17 2.05E+17 4.49E+17 (270.88) 690.58 4.51E+17 2.06E+17 4.51E+17 (271.88) 693.12 45E1 (272 4.52E+17 2.07E+17 4.52E+17 (272.88) 695.66 4.52E+17 2.08E+17 4.52E+17 (273.88) 698.20 4.53E+17 2.09E+17 4.53E+17 (274.88) 700.74453+7 75.88 4.53E+17 2.10E+17 4.53E+17 (275.88) 703.28 76.88 4.53E+17 2.10E+17 4.53E+17 (276.88) 705.82 4.53E+17 2.11E+17 4.53E+17 (277.88) 708.36 78.88 4.53E+17 2.12E+17 4.53E+17 (278.88) 710.90 45E1 79.88) 4.53E+17 2.13E+17 4.53E+17 (279.88) 713.44 (28.88 4.54E+17 2.14E+17 4.54E+17 (280.88) 715.98 45E1 21.88 4.54E+17 2.15E+17 4.54E+17 (281.88) 718.52 45E1 (282 4.54E+17 2.15E+17 4.54E+17 (282.88) 721.06 4.55E+17 2.16E+17 4.55E+17 (283.88) 723.60 (28.88 4.57E+17 2.17E+17 4.57E+17 (284.88) 726.14 (28.88 4.58E+17 2.17E+17 4.58E+1 7 (285.88) 728.68460+7 (2868 4.60E+17 2.18E+17 4.60E+17 (286.88) 731.22 46E1 7.88 4.62E+17 2.18E+17 4.62E+17 (287.88)

ENT-FLU-OO1-R-003 Revision 0 Page 7-10 of 7-21 Table 7-7 (continued)

Neutron Fluence Profile for Pilgrim Nuclear Power Station RPV Vertical Welds 1-338 A, 1-338 B, and 1-338 C at the Inner Surface for Energy >1.0 MeV at EOC 15 (20.7 EFPY)

Elevation cm Weld 1-338 A Fluence Weld 1-338 B Fluence Weld 1-338 C Fluence and (in) (n/cm 2) (nlcm 2) (nlcm 2) 733.76 4.64E+17 2.19E+17 4.64E+17 (288.88) 736.304.8+7 (28.88 4.68E+17 2.19E+17 4.68E+17 (289.88) 738.84 4.73E+17 2.20E+17 4.73E+17 (290.88) 741.38 4.78E+17 2.20E+17 4.78E+17 (291.88) 743.92 (292 48E1 4.83E+17 2.20E+17 4.83E+17 (292.88) 746.46 4.88E+17 2.21E+17 4.88E+17 (293.88) 749.00 4.93E+17 2.21E+17 4.93E+17 (294.88) 751.54 (5.88 4.94E+17 2.21E+17 4.94E+17 (295.88) 754.08494+7 (29.88 4.94E+17 2.21E+17 4.94E+17 (296.88) 756.62 4.95E+17 2.21E+17 4.95E+17 (297.88) 759.16495+7 (29.88 4.95E+17 2.22E+17 4.95E+17 (298.88) 761.70 4.96E+17 2.22E+17 4.96E+17 (299.88) 764.24 (30.88 4.96E+17 2.22E+17 4.96E+17 (300.88) 766.78 (30.88 4.96E+17 2.22E+17 4.96E+17 (301.88) 769.32 (32 4.96E+17 2.21E+17 4.96E+17 (302.88) 771.86 (30.88 4.96E+17 2.21 E+17 4.96E+17 (303.88) 774.40 304.88 4.96E+17 2.21E+17 (304.88) 4.96E+17 776.94 (30.88 49E1 4.96E+17 2.21E+17 4.96E+17 (305.88) 779.48 (30.88 4.94E+17 2.21E+17 4.94E+17 (306.88) 782.02 7.88 4.93E+17 2.20E+17 (307.88) 4.93E+17 784.56 (8.88 49E1 4.91E+17 2.20E+17 4.91E+17 (308.88) 787.10 (30 4.89E+17 2.19E+17 4.89E+17 (309.88) 789.64 (31.88 4.87E+17 2.19E+17 (310.88)1 4.87E+17

ENT-FLU-OO1-R-003 Revision 0 Page 7-11 of 7-21 Table 7-7 (continued)

Neutron Fluence Profile for Pilgrim Nuclear Power Station RPV Vertical Welds 1-338 A, 1-338 B, and 1-338 C at the Inner Surface for Energy >1.0 MeV at EOC 15 (20.7 EFPY)

Elevation cm Weld 1-338 A Fluence Weld 1-338 B Fluence Weld 1-338 C Fluence and (in) 2 (n/cm ) 2 (n/cm ) (nlcm 2) 792.18 (3118 4.83E+17 2.18E+17 4.83E+17 (311.88) 794.72 (312 4.78E+17 2.17E+17 4.78E+17 (312.84)_

797.26 4.72E+17 2.16E+17 4.72E+17 (313.88) 799.80 (31.88 4.67E+17 2.15E+17 4.67E+17 (314.88) 802.34 (35 4.61 E+17 2.14E+17 4.61E+17 (315.88) 804.88 (31.88 4.56E+17 2.13E+17 4.56E+17 (316.88) 807.42 (37.88 4.50E+17 2.12E+17 4.50E+17 (317.88) 809.96 (318.88) 4.43E+17 2.10E+17 4.43E+17 812.50 (319.88) 4.36E+17 2.08E+17 4.36E+17 815.04 4.30E+17 2.07E+17 4.30E+17 (320M8) 817.58 (1.88 4.24E+17 2.05E+17 4.24E+17 (321.88) 820.12 (322.88) 4.18E+17 2.03E+17 4.18E+17 822.66 (323.88) 4.12E+17 2.01E+17 4.12E+17 825.20 40E1 (24.88 4.06E+17 1.98E+17 4.06E+17 (324.88) 827.74 (25.88 4.OOE+17 1.96E+17 4.00E+17 (325.88) 830.28 (328 3.94E+17 1.93E+17 3.94E+17 (326.88) 832.82 (327.88) 3.89E+17 1.91E+17 3.89E+17 835.36 38E1 (32.88 3.81E+17 1.88E+17 3.81E+17 (328.88) 837.90 (3298 3.73E+17 1.84E+17 3.73E+17 (329.88) 840.44 3.66E+17 1.81E+17 3.66E+17 (330.88) 842.98 (33.88 3.58E+17 1.78E+17 3.58E+17 (331,88),

845.52 (332 3.51E+17 1.75E+17 3.51E+17 (332.88) 848.06 (33.88 3.43E+17 1.71E+17 3.43E+17 (333.84)_17E7

ENT-FLU-OO1-R-003 Revision 0 Page 7-12 of 7-21 Table 7-7 (continued)

Neutron Fluence Profile for Pilgrim Nuclear Power Station RPV Vertical Welds 1-338 A, 1-338 B, and 1-338 C at the Inner Surface for Energy >1.0 MeV at EOC 15 (20.7 EFPY)

Elevation cm Weld 1-338 A Fluence Weld 1-338 B Fluence Weld 1-338 C Fluence and (in) (n/cm 2) (nlcm 2) (nlcm 2) 850.60 3.38E+17 1.66E+17 3.38E+1 7 (334.88) 853.14 (33.88 3.33E+17 1.61E+17 3.33E+17 (335.88) 855.68 3.28E+17 1.56E+17 3.28E+17 (336.88) 858.22 (33.88 3.23E+17 1.51E+17 3.23E+17 (337.88) 860.76 3.18E+17 (338.88) 863.30 (339 3.13E+17 (339.88) 1.41E+17 3.13E+17 865.84 (30.8) 3.07E+17 (340.88) 1.36E+17 3.07E+17 868.38 (38 3.02E+17 1.30E+17 3.02E+17 (341.88) 870.92 (342 2.97E+17 1.25E+17 2.97E+17 (342.88) 873.46 2.92E+17 1.20E+17 2.92E+17 (343.88) 876.00 (344.88) 2.87E+17 1.15E+17 2.87E+17 878.54 345.88 2.73E+17 1.09E+17 2.73E+17 (345.88) 881.08 (34.88 2.59E+17 1.02E+17 2.59E+17 (346.88) 883.62 (3.88 2.45E+17 9.61E+16 2.45E+17 (347.88) 886.16 348.88 2.33E+17 9.04E+16 2.33E+17 888.70 (34.88 2.21E+17 8.51E+16 (349.88) 2.21E+17 891.24 2.07E+17 7.92E+16 2.07E+17 (350.88) 893.78 (3.88 1.89E+17 7.24E+16 1.89E+17 (351.88) 896.32 352.88 1.73E+17 6.62E+16 1.73E+17 (352.88) 898.86 835.88 1.58E+17 6.05E+16 1.58E+17 (353.88) 901.40 (3548 1.44E+17 5.53E+16 1.44E+17 (354.88) 903.94 (355.88) 1.32E+17 5.05E+16 1.32E+17 906.48 936.88 1.19E+17 4.56E+16 1.19E+17 (356.88)

ENT-FL U-OO1-R-003 Revision 0 Page 7-13 of 7-21 Table 7-7 (continued)

Neutron Fluence Profile for Pilgrim Nuclear Power Station RPV Vertical Welds 1-338 A, 1-338 B, and 1-338 C at the Inner Surface for Energy >1.0 MeV at EOC 15 (20.7 EFPY)

Elevation cm Weld 1-338 A Fluence Weld 1-338 B Fluence Weld 1-338 C Fluence and (in) (n/cm 2) (n/cm 2) (n/cm 2) 909.02 (35.08 1.07E+17 4.11E+16 1.07E+17 (357.88) 911.56 (358 9.60E+16 3.70E+16 9.60E+16 (358.88) 914.10 8.63E+16 3.34E+16 8.63E+16 (359.88) ,.

916.64 (36.88 7.75E+16 3.01E+16 7.75E+16 (360.88) 919.18 6.95E+16 (361.88) 921.72 (2.88 6.17E+16 2.45E+16 6.17E+16 (362.88) 924.26 5.47E+16 2.22E+16 5.47E+16 (363.88) 926.80 (36.88 4.85E+16 2.01E+16 4.85E+16 (364.88) 929.34 (36 43E1 4.30E+16 1.81E+16 4.30E+16 (365.88) 931.88 3.82E+16 1.64E+16 3.82E+16 (366.88) 934.42 (36.88 3.44E+16 1.52E+16 3.44E+16 (367.88) 936.96 (36.88 3.13E+16 1.42E+16 3.13E+16 (368.88) 939.50 (39.88 2.84E+16 1.32E+16 2.84E+16 (369.88) 942.04 (3708 2.59E+16 1.23E+16 2.59E+16 (370.88) 944.582.35E+16 1.15E+16 2.35E+16 (371.88) 947.12 (372 21E1 2.14E+16 1.08E+16 2.14E+16 (372.88) 949.66 1.97E+16 1.03E+16 1.97E+16 (373.88) 952.20 (37.88 1.82E+16 (374.88) 9.79E+15 1.82E+16 954.74 (5.88 1.67E+16 (375.88) 9.33E+15 1.67E+16 957.28 (37.88 1.54E+16 (376.88) 8.89E+15 1.54E+16 959.82 (37.88 1.42E+16 (377.88) 8.48E+15 1.42E+16 962.36 (37 13E1 1.31E+16 8.08E+15 1.31E+16 (378.88) 964.90 1.22E+16 7.69E+ 15 1.22E+16 (379.88)

ENT-FLU-OO1-R-003 Revision 0 Page 7-14 of 7-21 Table 7-7 (continued)

Neutron Fluence Profile for Pilgrim Nuclear Power Station RPV Vertical Welds 1-338 A, 1-338 B, and 1-338 C at the Inner Surface for Energy >1.0 MeV at EOC 15 (20.7 EFPY)

Elevation cm Weld 1-338 A Fluence Weld 1-338 B Fluence Weld 1-338 C Fluence and (in) (n/cm 2) (n/cm 2) (nlcm 2) 967.44 1.13E+16 7.33E+15 1.13E+16 (380.88) 969.98 (38.88 1.04E+16 6.98E+15 1.04E+16 (381.88) 972.52 9.68E+15 6,65E+15 9.68E+15 (382.88) 975.06 8.97E+15 6.34E+15 8.97E+15 (383.88) 977.60 (38.88 8.31E+15 5.99E+15 8.31 E+15 (384.88) 980.14 (85.88 77E1 7.70E+15 5.66E+15 7.70E+15 (385.88) 982.68 7.13E+1 5 5.35E+15 7.13E+15 (386.88) 985.226.61E+15 5.05E+15 6.61E+15 (387.88) 987.76 (88.88 61E1 6.13E+15 4.78E+15 6.13E+15 (388.88) 990.30 5.64E+1 5 4.46E+15 5.64E+15 (389.88) 992.84 90.88 5.11E+15 4.07E+15 5.11 E+15 (390.88) 995.38 (91.88 (391.88) 4.64E+15 3.71E+15 4.64E+15 997.92 (392 (392.88) 4.21E+15 3.38E+15 4.21 E+15 1000.46 (393.88 3.82E+15 3.08E+15 3.82E+15 (393.88) 1003.O00 (3.88 3.47E+15 2.81E+15 3.47E+15 (394.88) 1005.54 (395.88 3.09E+15 2.51E+15 (395.88) 3.09E+15 1008.08 (396.88 (396.88) 2.74E+15 2.22E+15 2.74E+15 1010.62 (397.88 (397.88) 2.43E+1 5 1.96E+15 2.43E+15 1010.92 1398.00 (398.00) 2.40E+15 1.94E+15 2.40E+15

ENT-FL U-OO1-R-003 Revision 0 Page 7-15 of 7-21 Table 7-8 RPV Vertical Welds 1-338 A, 1-338 B, Neutron Fluence Profile for Pilgrim Nuclear Power Station

>1.0 MeV at 54 EFPY and 1-338 C at the Inner Surface for Energy Weld 1-338 B Fluence Weld 1-338 C Fluence Elevation cm Weld 1-338 A Fluence (n/cm 2) (n/cm 2) and (in) (n/cm 2) 616.92 6242 7.83E+17 3.59E+17 7.83E+17 (242.88) 619.46 (243 7.90E+17 3.62E+17 7.90E+17 (243.88) 622.00 (2.88 7.97E+17 3.65E+17 7.97E+17 (244.88) 624.54 (2458 8.04E+17 3.68E+17 8.04E+17 (245.88) 627.08 (246.8 8.11 E+17 3.72E+17 8.11E+1 7 (246.88) 629.62 (24.88 8.18E+17 3.75E+17 8.18E+17 (247.88) 632.16 2.88 8.25E+17 3.78E+17 8.25E+17 (248.88) 634.70 (24.80 8.32E+17 3.81E+17 8.32E+17 (249.88) 637.24 (25 8.39E+17 3.85E+17 8.39E+17 (250.88) 639.78 8.46E+17 3.88E+17 8.46E+17 (251.88) 642.32 (252 8.53E+17 3.92E+17 8.53E+17 (252.88)

,.644.86 (25.88 8.61E+17 3.95E+17 8.61E+17 (253.88) 647.40 (4.88 8.68E+17 3.98E+17 8.68E+17 (254.88) 649.94 625.88 8.75E+17 4.02E+17 8.75E+17 (255.88) 652.48 (2.88 8.81E+17 4.05E+17 8.81E+17 (256.88) 655.02 8.87E+17 4.09E+ 17 8.87E+1 7 (257.88) 657.56 8.93E+17 4.12E+17 8.93E+1 7 (258.88) 660.10 (25.88 9.OOE+17 4.15E+17 9.00E+17 (259.88) 662.64 (268 9.06E+17 4.19E+17 9.06E+17 (260.88) 665.18 2618 ______________________90E1 9.13E+17 4.22E+17 9.13E+17 (261.88) 667.72 (62.88 9.19E+17 4.25E+17 9.19E+17 (262.88) 670.26 (26 ______________________91E1 9.26E+17 4.28E+17 9.26E+17 (263.88) 672.80 9.33E+17 4.31E+17 9.33E+17 (264.88)

ENT-FLU-OO1-R-003 Revision 0 Page 7-16 of 7-21 Table 7-8 (Continued)

Neutron Fluence Profile for Pilgrim Nuclear Power Station RPV Vertical Welds 1-338 A, 1-338 B, and 1-338 C at the Inner Surface for Energy >1.0 MeV at 54 EFPY Elevation cm Weld 1-338 A Fluence Weld 1-338 B Fluence Weld 1-338 C Fluence and (in) (n/cm 2) (n/cm 2) (nlcm 2) 675.34 9.39E+17 4.34E+17 9.39E+17 (265.88) 677.88 9.46E+1 7 (266.88) 680.42 9.52E+17 4.40E+17 9.52E+17 (267.88) 682.96 9.59E+17 4.43E+17 9.59E+17 (268.88) 685.50 69.88 9.65E+17 4.46E+17 9.65E+17 (269.88) 688.04 (88) 9.71E+17 4.49E+17 9.71E+17 (270.88) 690.58 9.77E+17 4.52E+17 9.77E+17 (271.88) 693.12 9.82E+17 4.54E+17 9.82E+17 (272.88) 695.66984+7 (27.88 (273.88) 9.84E+17 4.57E+17 9.84E+17 698.20 9.85E+17 4.60E+17 9.85E+17 (274.88) 700.74 75.88 (275.88) 9.87E+17 4.62E+17 9.87E+17 70328 .9.89E+17 4.65E+17 9.89E+17 (276.88) 705.82 9.91E+17 4.67E+17 9.91 E+1 7 (277.88) 708.36 78.88 9.93E+17 4.70E+17 9.93E+17 (278.88) 710.90 79.88) 9.96E+17 4.73E+17 9.96E+17 (279.88) 713.44 9.99E+17 4.76E+17 9.99E+17 (280.88) 715.98 1.OOE+18 4.78E+17 1.OOE+18 (281.88) _ _ _ _ _ _ _ _ _ _ __ _ _ _ _ _ _ _ _ _ _

718.52 (282 1.00E+18 4.81E+ 17 1.OOE+18 (282.88) 721.06 (283.88) 1.01E+18 4.84E+17 1.01E+18 (284.88) 723.60 1OE1 (284 1.02E+18 4.86E+17 1.02E+18 (285.88) 726.14 728.88 1.02E+18 4.88E+17 1.02E+18 (286.88) 728.68 (268)1.02E+18 4.91E+17 1.02E+18 731.22 7.88 10E1 (287.88) 1.03E+18 4.93E+17 1.03E+18 1

ENT-FLU-001-R-003 Revision 0 Page 7-17 of 7-21 Table 7-8 (continued)

Neutron Fluence Profile for Pilgrim Nuclear Power Station RPV Vertical Welds 1-338 A, 1-338 B, and 1-338 C at the Inner Surface for Energy >1.0 MeV at 54 EFPY Elevation cm Weld 1-338 A Fluence Weld 1-338 B Fluence Weld 1-338 C Fluence and (in) (n/cm 2) (n/cm 2) (n/cm 2) 733.76 (28.88 1.04E+18 4.95E+17 (288.88) 1.04E+18 736.30 (289.88) 1.05E+18 4.97E+17 1.05E+18 738.84 (290.88) 1.06E+18 4.99E+17 1.06E+18 741.38 1.07E+18 (291.88) 743.92 (292 1.08E+18 5.02E+17 1.08E+18 (292.88) 746.46 (29.88 1.1OE+18 5.04E+17 1.1OE+18 (293.88) 749.00 (29.88 1.11E+18 5.05E+17 1.11E+18 (294.88) 751.54 71.11 E+1 8 (295.88) 754.08 (29.8 1.12E+18 5.08E+17 1.12E+18 (296.88) 756.62 1.12E+18 5.09E+17 1.12E+18 (297.88) 759.16 11E1 (298.88) 1.12E+18 5.11E+17 1.12E+18 761.70 (299.88) 1.13E+18 5.12E+17 1.13E+18 766.78 (30.88 1.13E+18 5.13E+17 (300.88) 1.13E+18 76.1.13E+18 5.14E+17 1.13E+18 (301.88) 76 9.3 2 1 4 + . 5 + 71 1 E 1 771.86 1.14E+18 (303.88) 774.40 1.14E+18 (304.88)114+856E7 776.94 1.14E+18 5.17E+17 1.14E+18 (305.88) 779.48 1.14E+18 5.17E+17 1.14E+18 (307.88)

(306.88) 782.02 1.14E+18 (308.88) 784.56 784.58 1.14E+18 5.17E+17 1.14E+18 (8.88 1.14E+18 5.17E+17 1.14E+18 789.564 (3108.88)

(309.88)

(31.88 1.13E+18 5.17E+17 1.13E+18

ENT-FLU-OO1-R-003 Revision 0 Page 7-18 of 7-21 Table 7-8 (continued)

Neutron Fluence Profile for Pilgrim Nuclear Power Station RPV Vertical Welds 1-338 A, 1-338 B, and 1-338 C at the Inner Surface for Energy >1.0 MeV at 54 EFPY Elevation cm Weld 1-338 A Fluence Weld 1-338 B Fluence Weld 1-338 C Fluence and (in) (n/cm 2) (n/cm 2) (n/cm 2) 792.18 1.13E+18 (311.88) 794.72 1.12E+18 (312.88) 797.26 1.11E+18 5.14E+17 1.11E+18 (313.88) 799.80 (31.88 1.IOE+18 5.13E+17 1.10E+18 (314.88) 802.34 1.08E+18 (315.88) 804.88 (31.88 1.07E+18 5.11E+17 1.07E+18 (316.88) 807.42 1.06E+18 5.08E+17 1.06E+18 (317.88) 10E1 809.96 8.88 1.05E+18 5.05E+17 1.05E+18 (318.88) 812.50 319.88 1.03E+18 5.02E+17 1.03E+18 (319.88) 815.04 1.02E+18 4.99E+17 1.02E+18 (320.88) 817.58 (1.88 1.01E+18 4.96E+17 (321.88) 1.01E+18 820.12 (22.88 9.96E+17 4.92E+17 9.96E+17 (322.88) 822.66 (32.88 9.83E+17 4.87E+17 9.83E+17 (323.88) 825.20 324.88 9.71E+17 4.81E+17 (324.88) 9.71E+17 827.74 325.88 9.58E+17 4.76E+17 (325.88) 9.58E+17 830.28 (328 9.46E+17 4.71 E+17 (326.88) 9.46E+17 832.82 (32.88 9.33E+17 4.66E+17 (327.88) 9.33E+17 835.36 (32.88 9.16E+17 4.58E+17 (328.88) 9.16E+17 837.90 (3298 8.96E+17 4.50E+ 17 8.96E+1 7 (329.88) 840.44 830.88 8.77E+17 4.42E+17 (330.88) 8.77E+17 842.98 (33.88 8.58E+1 7 4.34E+ 17 (331.88) 8.58E+17 845.52 8.40E+17 4.26E+17 8.40E+17 (332.88) 848.06 (33.88 8.22E+17 4.18E+17 (333.88) 8.22E+17

ENT-FLU-OO1-R-003 Revision 0 Page 7-19 of 7-21 Table 7-8 (continued)

Neutron Fluence Profile for Pilgrim Nuclear Power Station RPV Vertical Welds 1-338 A, 1-338 B, and 1-338 C at the Inner Surface for Energy >1.0 MeV at 54 EFPY Elevation cm Weld 1-338 A Fluence Weld 1-338 B Fluence Weld 1-338 C Fluence and (in) (n/cm 2) (nlcm 2) (n/cm 2) 850.60 (33.88 8.07E+17 4.04E+17 8.07E+17 (334.88) 853.14 7.92E+17 (335.88) 855.68 7.77E+17 3.78E+17 7.77E+17 (336.88) 858.22 337.88 7.63E+17 3.66E+17 7.63E+17 (337.88) 860.76 8.88 7.49E+17 3.54E+17 7.49E+17 (338.88) 863.30 (339 7.35E+17 3.41E+17 7.35E+17 (339.88)_

865.84 834.88 7.21E+17 3.27E+17 7.21E+17 (340.88) 868.38 (38 7.07E+17 3.14E+17 7.07E+17 (341.88) 870.92 6 4+1 (342 6.94E+17 3.01E+17 6.94E+17 (342.88) 873.46 6.81E+17 2.88E+17 6.81E+17 (343.88) 876.00 (34.88 6.68E+17 2.76E+17 6.68E+17 (344.88)

  • .878.54 (34.88 6.33E+17 2.60E+17 6.33E+17 (345.88) 881.08 (34.88 (346.88) 5.99E+17 2.44E+17 5.99E+17 883.6256E+7 (3.88 5.66E+17 2.29E+17 5.66E+17 (347.88) 886.16536+7 (34.88 5.36E+17 2.15E+17 5.36E+17 (348.88) 888.70 50E1 (34.88 5.07E+17 2.02E+17 5.07E+17 (349.88) 891.24 (35.88 4.73E+17 1.87E+17 4.73E+17 (350.88) 893.78 4.30E+17 1.71E+17 4.30E+17 (351.88) 896.32 3.92E+17 1.55E+17 3.92E+17 (352.88) 898.86 35E1 (35.88 3.56E+17 1.41E+17 3.56E+17 (353.88) 901.40 3.24E+17 1.29E+17 3.24E+17 (354.88) 903.94 (3.88 2.95E+17 1.17E+17 2.95E+17 (355.88) 906.48 2.65E+17 1.05E+17 2.65E+17 (356.88)

ENT-FLU-001-R-003 Revision 0 Page 7-20 of 7-21 Table 7-8 (continued)

Neutron Fluence Profile for Pilgrim Nuclear Power Station RPV Vertical Welds 1-338 A, 1-338 B, and 1-338 C at the Inner Surface for Energy >1.0 MeV at 54 EFPY Elevation cm Weld 1-338 A Fluence Weld 1-338 B Fluence Weld 1-338 C Fluence and (in) (nlcm 2) (n/cm 2) (n/cm2 )

909.02 2.37E+1 7 (357.88) 911.56 2.12E+17 8.48E+16 2.12E+17 (358.88) 914.10 1:90E+17 7.61E+16 1.90E+17 (359.88) 916.64 (36.88 17E1 1.70E+17 6.83E+16 1.70E+17 (360.88) 919.181.51E+17 6.13E+16 1.51E+17 (361.88) 921.72 1.34E+17 5.51E+16 1.34E+17 (362.88) 924.26 (36 11E1 1.18E+17 4.95E+16 1.18E+17 (363.88)_

926.801.04E+17 4.45E+16 1.04E+17 (364.88) 929.34 (36 9.16E+16 4.OOE+16 9.16E+16 (365.88) 931.88 8.08E+16 3.59E+16 8.08E+16 (366.88) 934.42 (36.88 72E1 7.25E+16 3.30E+16 7.25E+16 (367.88) 936.96 36.88 6.58E+16 3.07E+16 6.58E+16 (368.88) 939.50 (39.88 5.97E+16 2.85E+16 5.97E+16 (369.88) 942.04 (3708 5.41 E+16 2.66E+16 5.41E+16 (370.88) 944.58 371.88 4.91E+1 6 2.47E+16 4.91E+16 (371.88) 947.12 (372 4.45E+16 2.30E+16 4.45E+16 (372.88) 949.66 4.10E+16 2.18E+16 4.10E+16 (373.88) 952.20 (37.88 3.77E+16 2.07E+16 3.77E+16 (374.88) 954.74 35.88 3.47E+16 1.97E+16 3.47E+16 (375.88) 957.28 937.88 3.19E+16 1.87E+16 3.19E+16 (376.88) 959.82 937.88 2.94E+16 1.78E+16 (377.88) 2.94E+16 962.36 (37 (378.88) 2.71E+16 1.69E+16 2.71 E+16 964.90 9.88 2.51E+16 1.60E+16 (379.88) 2.51E+16

ENT-FLU-OO1-R-003 Revision 0 Page 7-21 of 7-21 Table 7-8 (continued)

Neutron Fluence Profile for Pilgrim Nuclear Power Station RPV Vertical Welds 1-338 A, 1-338 B, and 1-338 C at the Inner Surface for Energy >1.0 MeV at 54 EFPY Elevation cm Weld 1-338 A Fluence Weld 1-338 B Fluence Weld 1-338 C Fluence and (in) (n/cm2 ) 2 (n/cm ) (n/cm 2) 967.44 2.33E+16 1.53E+16 2.33E+16 (380.88) 969.98 (38.88 21E1 2.16E+16 1.45E+16 2.16E+16 (381.88) 972.52 2.OOE+16 1.38E+16 2.OOE+ 16 (382.88) 975.06 1.85E+16 1.31E+16 1.85E+16 (383.88) 977.60 (38.88 1.2+

1.72E+16 1.24E+16 1.72E+16 (384.88) 980.141.5E6 (85.88 1.59E+16 1.17E+16 1.59E+16 (385.88) 982.68 1.48E+16 1.11 E+16 1.48E+16 (386.88) 985.221 (87.88 1.37E+16 1.05E+16 1.37E+16 (387.88) 987.76 (88.88 1.27E+16 9.87E+15 1.27E+16 (388.88) 990.30 1.17E+16 9.22E+15 1.17E+16 (389.88) 992.84 90.88 1.07E+16 8.41E+15 1.07E+16 (390.88) 995.38 9.68E+15 7.67E+15 9.68E+15 (391.88) 997.92 (392 8.80E+15 7.OOE+15 8.80E+15 (392.88) 1000.46 8.OOE+15 (393.88) 1003.00 7.27E+1 5 (394.88) 1005.54 (395.88 65E1 6.50E+15 5.21E+15 6.50E+15 (395.88) 1008.08 (396.88 5.77E+15 4.62E+15 5.77E+15 (396.88) 1010.62 (397.88 5.13E+15 4.10E+15 5.13E+15 (397.88) 1010.92 (398.00 5.06E+15 4.04E+15 5.06E+15 (398.00)

ENT-FLU-OO1-R-003 Revision 0 Page 8-1 of 8-2 8

REFERENCES

1. "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence,"

Nuclear Regulatory Commission Regulatory Guide 1.190, March 2001.

2. BWRVIP-126: BWR Vessel InternalsProject,RAMA Fluence Methodology Software, Version 1.0, EPRI, Palo Alto, CA: 2003. 1007823.
5. Mechanical Input to TWFluence Calc ContractNo. 4500540948- ER02115347, Rev 0, Entergy Nuclear Northeast, September 7, 2005.
6. Documentation of OperatingInputsfor PNPS Vessel Fluence Calculations,Engineering Report No. PNPS-RPT-05-007, Rev 0, Entergy Nuclear Northeast, November 15, 2005.
7. "BUGLE-96: Coupled 47 Neutron, 20 Gamma-Ray Group Cross Section Library Derived from ENDF/B-VI for LWR Shielding and Pressure Vessel Dosimetry Applications," RSICC Data Library Collection, DLC-185, March 1996.

((I

9. BWRVIP-121: BWR Vessel InternalsProject,RAMA Fluence Methodology Procedures Manual, EPRI, Palo Alto, CA: 2003.1008062.
10. PilgrimNuclear Power Station Unit I Reactor Vessel IrradiationSurveillance Program, Southwest Research Institute, July 1981.

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ENT-FLU-OO1-R-003 Revision 0 Page A-I of A-3 A

CYCLE 4 SURVEILLANCE CAPSULE EVALUATION This appendix addresses the results of the evaluation of the Pilgrim Nuclear Power Station cycle 4 surveillance capsule results. ((

A.1 Comparison of Predicted Activation to Plant-Specific Measurements Three copper, three iron, and three nickel flux wires were irradiated in the Pilgrim Nuclear Power Station surveillance capsule during the first four cycles of operation. Activation measurements were performed following irradiation for the following reactions [10]: 63Cu(n,a)6Co,

-54Fe(n,p)aMn, and 8Ni(n,p)58 Co.((

1]

ENT-FLU-OO1-R-003 Revision 0 PageA-2 ofA-3 I((

ENT-FLU-OO1-R-003 Revision 0 Page A-3 of A-3 ft I ____________________________________ I I

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