CNL-24-011, Application for Subsequent Renewed Operating Licenses, Supplemental Information - Neutron Fluence Analyses Methodology

Application for Subsequent Renewed Operating Licenses, Supplemental Information - Neutron Fluence Analyses Methodology

ML24022A292
Person / Time
Site:	Browns Ferry
Issue date:	01/22/2024
From:	Hulvey K Tennessee Valley Authority
To:	Office of Nuclear Reactor Regulation, Document Control Desk
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ML24022A291	List: CNL-24-011, Application for Subsequent Renewed Operating Licenses, Supplemental Information - Neutron Fluence Analyses Methodology
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CNL-24-011
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Proprietary Information Withhold Under 10 CFR § 2.390 This letter is decontrolled when separated from Enclosure 2 Proprietary Information Withhold Under 10 CFR § 2.390 This letter is decontrolled when separated from Enclosure 2 1101 Market Street, Chattanooga, Tennessee 37402 CNL-24-011 January 22, 2024 10 CFR 54 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001 Browns Ferry Nuclear Plant, Units 1, 2, and 3 Renewed Facility Operating License Nos. DPR-33, DPR-52, and DPR-68 NRC Docket Nos. 50-259, 50-260, and 50-296

Subject:

Browns Ferry Nuclear Plant, Units 1, 2, and 3 - Application for Subsequent Renewed Operating Licenses, Supplemental Information - Neutron Fluence Analyses Methodology

Reference:

Letter from TVA to NRC, CNL-24-001, "Browns Ferry Nuclear Plant Units 1, 2, and 3 - Application for Subsequent Renewed Operating Licenses,"

dated January 19, 2024 (ML24019A010, ML24019A011)

By the reference Subsequent License Renewal Application Appendix E, Applicant’S Environmental Report-Operating License Renewal Stage|letter dated January 19, 2024]], the Tennessee Valley Authority (TVA) submitted a Subsequent License Renewal Application (SLRA) for the Browns Ferry Nuclear Plant (BFN),

Units 1, 2, and 3, Renewed Facility Operating Licenses in accordance with Title 10 of the Code of Federal Regulations (10 CFR), Part 54, Requirements for Renewal of Operating Licenses for Nuclear Power Plants.

Based on review of other SLRAs, it has been determined that other applicants have submitted supplemental information to support the NRC review of their neutron fluence analyses. As a result, this submittal provides supplemental information that includes the methodology utilized for the BFN SLRA, Section 4.2.1, Reactor Vessel and Internals Neutron Fluence Analyses. This supplement does not change any information in the BFN SLRA. of this letter provides the non-proprietary version of the BFN Fluence Methodology Report and Attachments 1, 2, and 3. Enclosure 2 of this letter provides the proprietary version of BFN Fluence Methodology Report and Attachments 1, 2, and 3. TransWare Enterprises Inc.

(TWE), consider portions of the information provided in Enclosure 2 of this letter to be proprietary and, therefore, exempt from public disclosure pursuant to 10 CFR 2.390, Public Inspections, Exemptions, Requests for Withholding. Affidavits for withholding information, executed by TWE, are provided in Enclosure 3. Therefore, on behalf of TWE, TVA requests that Enclosure 2 of this letter be withheld from public disclosure in accordance with the TWE affidavits and the provisions of 10 CFR 2.390.

Proprietary Information Withhold Under 10 CFR § 2.390 This letter is decontrolled when separated from Enclosure 2 U.S. Nuclear Regulatory Commission CNL-24-011 Page 2 January 22, 2024 Proprietary Information Withhold Under 10 CFR § 2.390 This letter is decontrolled when separated from Enclosure 2 Should you have any questions regarding this submittal, please contact Peter J. Donahue, Director, Subsequent License Renewal, at pjdonahue@tva.gov.

I declare under penalty of perjury that the foregoing is true and correct. Executed on this 22nd day of January 2024.

Respectfully, Kimberly D. Hulvey Director, Nuclear Regulatory Affairs

Enclosures:

1. BFN Fluence Methodology Report and Attachments 1, 2, and 3 (non-proprietary version)

2. BFN Fluence Methodology Report and Attachments 1, 2, and 3 (proprietary version)

3. Affidavits cc:

NRC Regional Administrator - Region II NRC Branch Chief - Region II NRC Senior Resident Inspector - Browns Ferry Nuclear Plant NRC Project Manager, License Renewal Projects Branch (Safety)

State Health Officer, Alabama Department of Public Health (w/o Enclosure 2)

Digitally signed by Edmondson, Carla Date: 2024.01.22 11:24:19 -05'00'

ENCLOSURE 1 BFN Fluence Methodology Report and Attachments 1, 2, and 3 (BFN-FLU-001-R-001-LNP, Rev. 1; BFN-FLU-001-R-001-LNP, Attachment 1, Revision 1; BFN-FLU-001-R-001-LNP, Attachment 2, Revision 1; and BFN-FLU-001-R-001-LNP, Attachment 3, Revision 1)

(non-proprietary version)

Non-Proprietary BFN-FLU-001-R-001-LNP Revision 1 Page i of x Topical Report BROWNS FERRY NUCLEAR PLANT FLUENCE METHODOLOGY REPORT Document Number:

BFN-FLU-001-R-001-LNP Revision 1 October 2023 Prepared by:

TransWare Enterprises Inc.

Prepared for:

Tennessee Valley Authority 1101 Market Street Chattanooga, TN 37402 Contract Number:

15904 Project Manager:

John Paul Anglin

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Controlled Copy Number: ____2____

Non-Proprietary BFN-FLU-001-R-001-LNP Revision 1 Page ii of x

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Non-Proprietary BFN-FLU-001-R-001-LNP Revision 1 Page iii of x Topical Report BROWNS FERRY NUCLEAR PLANT FLUENCE METHODOLOGY REPORT Document Number: BFN-FLU-001-R-001-LNP Revision 1 October 2023 Prepared By:

TransWare Enterprises Inc.

Project Team:

Project Manager:

E. A. Evans, Project Engineer H. J. Heppermann, Project Engineer M. E. Jewell, Project Engineer J. S. Styczynski, Project Engineer S. M. Wagstaff, Project Engineer K. E. Watkins, Project Engineer D.B. Jones 10/25/2023 D. B. Jones, Project Manager Date Reviewed By:

K.E. Watkins 10/25/2023 K. E. Watkins, Project Engineer Date K.A. Jones 10/25/2023 K. A. Jones, QA Specialist Date Approved By:

D.B. Jones 10/25/2023 D. B. Jones, Project Manager Date Prepared For:

Tennessee Valley Authority 1101 Market Street Chattanooga, TN 37402 Contract Number: 15904 Project Manager: John Paul Anglin Official signatures are on file. Contact TransWare Enterprises to request a copy.

TransWare Enterprises Inc.

• 520 Maryville Centre Dr., Suite 125

• St. Louis, Missouri 63141

• USA

Non-Proprietary BFN-FLU-001-R-001-LNP Revision 1 Page iv of x DISCLAIMER OF WARRANTIES AND LIMITATION OF LIABILITIES THE INFORMATION CONTAINED IN THIS REPORT IS BELIEVED BY TRANSWARE ENTERPRISES INC. TO BE AN ACCURATE AND TRUE REPRESENTATION 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.

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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.

Non-Proprietary BFN-FLU-001-R-001-LNP Revision 1 Page v of x CONTENTS Title Page 1

Introduction................................................................................................................... 1-1 1.1 Regulatory Requirements for Determining Fluence in Light Water Reactors...........1-1 1.2 Quality Assurance...................................................................................................1-3 2

Description of the Reactor System.............................................................................. 2-1 2.1 Overview of the Reactor System Design.................................................................2-1 2.2 Reactor System Mechanical Design Inputs.............................................................2-2 2.3 Reactor System Material Compositions..................................................................2-3 2.4 Reactor Operating Data Inputs...............................................................................2-5 2.4.1 Core Configuration and Fuel Design..........................................................2-5 2.4.2 Reactor Power History...............................................................................2-5 2.4.3 Reactor Statepoint Data.............................................................................2-6 2.4.4 Reactor Coolant Properties........................................................................2-6 3

Methodology................................................................................................................. 3-1 3.1 Computational Fluence Method..............................................................................3-1 3.2 Fluence Model........................................................................................................3-2 3.2.1 Geometry Models......................................................................................3-5 3.2.2 Reactor Core and Core Reflector...............................................................3-6 3.2.3 Reactor Core Shroud.................................................................................3-6 3.2.4 Downcomer Region...................................................................................3-7 3.2.4.1 Jet Pump Assemblies................................................................3-7 3.2.4.2 Surveillance Capsules...............................................................3-7 3.2.5 Reactor Pressure Vessel...........................................................................3-8

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3.2.6 Thermal Insulation.....................................................................................3-8 3.2.7 Inner and Outer Cavity Regions.................................................................3-8 3.2.8 Biological Shield Model..............................................................................3-8 3.2.9 Above-Core Components..........................................................................3-9 3.2.9.1 Top Guide..................................................................................3-9 3.2.9.2 Core Spray Spargers and Piping................................................3-9 3.2.10 Below-Core Components...........................................................................3-9 3.2.10.1 Core Support Plate and Rim Bolts.............................................3-9 3.2.10.2 Fuel Support Pieces................................................................. 3-10 3.2.10.3 Control Blades and Guide Tubes............................................. 3-10 3.2.11 Summary of the Geometry Modeling Approach....................................... 3-10

Non-Proprietary BFN-FLU-001-R-001-LNP Revision 1 Page vi of x 3.3 Particle Transport Calculation Parameters............................................................ 3-11 3.4 Fission Spectrum and Neutron Source................................................................. 3-11 3.5 Parametric Sensitivity Analyses............................................................................ 3-12 4

Surveillance Capsule Evaluations and Combined Uncertainty Analysis.................. 4-1 5

References.................................................................................................................... 5-1 5.1 References.............................................................................................................5-1 5.2 Glossary.................................................................................................................5-3

Non-Proprietary BFN-FLU-001-R-001-LNP Revision 1 Page vii of x LIST OF FIGURES Title Page Figure 2-1 Planar View of the Browns Ferry Reactors at the Core Mid-Plane Elevation........ 2-2 Figure 3-1 Planar View of the Browns Ferry Fluence Models at the Core Mid-Plane Elevation in Quadrant Symmetry......................................................................... 3-3 Figure 3-2 Axial View of the Browns Ferry Fluence Models.................................................. 3-4

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Non-Proprietary BFN-FLU-001-R-001-LNP Revision 1 Page ix of x LIST OF TABLES Title Page Table 2-1 Summary of Material Compositions by Region for Browns Ferry......................... 2-4

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INTRODUCTION This report provides an overview of the Browns Ferry Nuclear Plant (Browns Ferry) fast neutron fluence model and fluence methodology. Browns Ferry is three reactor units of the General Electric BWR/4 class design with core loadings of 764 fuel assemblies in each unit. The Browns Ferry reactors are operated by Tennessee Valley Authority (TVA).

Fluence evaluations performed for the Browns Ferry reactors are based upon the RAMA Fluence Methodology software [1], the RAMA Fluence Methodology Procedures Manual [2], and the RAMA Fluence Methodology Theory Manual [3]. The RAMA Fluence Methodology was developed by TransWare Enterprises under sponsorship of the Electric Power Research Institute, Inc. (EPRI) and the Boiling Water Reactor Vessel and Internals Project (BWRVIP). The RAMA Fluence Methodology (hereinafter referred to as RAMA) has received generic approval by the U.S. Nuclear Regulatory Commission for determining neutron fluence in light water reactor pressure vessels [4]. More information on the qualifications of RAMA for determining fluence in reactor pressure vessels and reactor vessel internals is presented in Section 1.1, below.

All data used to construct the Browns Ferry fluence models, define the structural and fuel materials, and develop the lifetime operating history of the reactors was provided by TVA.

1.1 Regulatory Requirements for Determining Fluence in Light Water Reactors Part 50 of Title 10 of Code of Federal Regulations (10CFR50), which is issued by the federal agencies of the United States of America, provides requirements for establishing irradiated material monitoring programs that serve to ensure the integrity of the reactor coolant pressure boundary of light water nuclear power reactors. Two appendices to Part 50 present requirements that guide fluence determinations: Appendix G, Fracture Toughness Requirements [5], and Appendix H, Reactor Vessel Material Surveillance Program Requirements [6].

Appendix G specifies fracture toughness requirements for the carbon and low-alloy ferritic materials of the pressure-retaining components of the reactor coolant pressure boundary. These requirements are to ensure adequate margins of safety during any condition of normal operation including anticipated conditions for system hydrostatic testing, to which the pressure boundary may be subjected over its service lifetime. These requirements apply to base metal, welds, and weld heat-affected zones in the materials within the reactor pressure vessel (RPV) beltline region.

Appendix H specifies the requirements for a material surveillance program that serves to monitor changes in the fracture toughness properties of the ferritic materials in the reactor beltline region.

The changes in fracture toughness properties of ferritic materials are attributed to the exposure of the material to neutron irradiation and the thermal environment.Section III of Appendix H specifies that a material surveillance program is required for light water nuclear power reactors if

Non-Proprietary BFN-FLU-001-R-001-LNP Revision 1 Page 1-2 of 1-4 the peak fast neutron fluence with energy greater than 1 MeV (E > 1 MeV) at the end of the design life of the vessel is expected to exceed 1017 n/cm2.

In compliance with Appendix H requirements, fracture toughness test data are obtained from material specimens that are exposed to neutron irradiation in surveillance capsules installed at or near the inner surface of the reactor pressure vessel. These capsules are withdrawn periodically from the reactor for measurement and analysis. Fast neutron fluence is not a measurable quantity and must be determined using analytical methods. It must be demonstrated that the analytical method used to determine the fast neutron fluence provides a conservative prediction over the beltline region of the pressure boundary when compared to the measurement data with allowances for all uncertainties in the measurement.Section III of Appendix H also allows for an Integrated Surveillance Program (ISP) in which representative materials for the reactor are irradiated in one or more other reactors of sufficiently similar design and operating features to permit accurate comparisons of the predicted amount of radiation damage.

Implementing guidelines addressing the requirements of Appendices G and H are provided in U.S. Nuclear Regulatory Commission (NRC) Regulatory Guide 1.99, Revision 2, Radiation Embrittlement of Reactor Vessel Materials [7], and Regulatory Guide 1.190, Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence [8]. Regulatory Guide 1.99 addresses the requirements of Appendix G for determining the damage fluence that is used in the evaluation of fracture toughness in light water nuclear reactor pressure vessel ferritic materials. Regulatory Guide 1.190 addresses the requirements for determining the fast neutron fluence and uncertainty in the fluence predictions that are used in fracture toughness evaluations.

RAMA is qualified against industry standard benchmarks for both boiling water reactor (BWR) and pressurized water reactor (PWR) designs. The RAMA methodology, as well as TransWares application of the methodology, have been reviewed by the NRC and given generic approval for determining fast neutron fluence in both BWR and PWR pressure vessels [4] with no discernable bias in the computed results.

The RAMA methodology has also received conditional approval for determining fast neutron fluence in light water reactor vessel internals (RVI). The Safety Evaluation (SE) issued by the U.S. Nuclear Regulatory Commission for EPRI report BWRVIP-145 [9] concludes that for applications such as IASCC, crack propagation rates and weldability determinations, the RAMA methodology can be used in determining fast neutron fluence values in the core shroud and top guidefor licensing actions provided that the calculational results are supported by sufficient justification that the proposed values are conservative for the intended application.

It is specifically noted that the limitation cited in the SE requires that sufficient justification be provided that the computed fluence for the core shroud and top guide internal components is conservative. The optimum justification would be comparisons to dosimetry measurements for the components. As noted in the SE, dosimetry measurements of internal components are not common in the industry for benchmarking purposes. Therefore, in the absence of specific dosimetry measurements for a plant, the following justifications are made:

• The comparisons to measurements determined for the Susquehanna Unit 2 core shroud and top guide components show that the RAMA Fluence Methodology over-predicted the activity measurements, therefore, resulting in the determination of conservative fluence [10].

Non-Proprietary BFN-FLU-001-R-001-LNP Revision 1 Page 1-3 of 1-4

• The computational fluence models constructed for the Browns Ferry reactors follow similar modeling techniques as those performed for the Susquehanna Unit 2 core shroud and top guide benchmark.

• The computational fluence models for the Browns Ferry reactors include accurate geometry and material representations for the central and upper core shroud shells, top guide plates, fuel structures, upper shroud plenums, and coolant water densities (saturated water and exit steam); therefore, providing best-estimate models of the reactors.

• In addition, the following conservatisms were incorporated in the Browns Ferry fluence models:

o The neutron source most important to the irradiation of the ((

))

o The total of the shroud upper plenum volume ((

))

The fast neutron fluence methods discussed in this report meet the requirements of 10CFR50 Appendices G and H and Regulatory Guides 1.190 and 1.99 Revision 2. ((

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1.2 Quality Assurance The implementation and validation of the fluence methodology presented in this report complies with the quality assurance requirements of 10CFR50 Appendix B, Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants [11], and to 10CFR21, Reporting of Defects and Noncompliance [12].

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DESCRIPTION OF THE REACTOR SYSTEM This section provides an overview of the reactor design and operating data inputs that were used to develop the computational fluence models for the Browns Ferry reactors. All reactor design and operating data inputs used to develop the models are reactor-specific and were provided by TVA. The inputs for the fluence geometry models were developed from nominal and as-built drawings for the reactor pressure vessel, vessel internals, fuel assemblies, and containment regions for each reactor.

Browns Ferry Unit 2 was previously evaluated by TransWare circa 2012 [13]. Several modifications were made to the Browns Ferry Unit 2 RAMA geometry model since the previous fluence evaluation. ((

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2.1 Overview of the Reactor System Design Browns Ferry is three General Electric BWR/4 class reactors with core loadings of 764 fuel assemblies in each unit. Figure 2-1 illustrates the basic planar configuration of the Browns Ferry reactor models at an axial elevation near the reactor core mid-plane. All of the radial regions of the reactors that are required for fluence evaluations are shown. Beginning at the center of the reactor and projecting outward, the regions include: the core region; core reflector region (bypass water); shroud wall; downcomer water region including the jet pump assemblies; RPV wall; cavity region between the RPV wall and insulation; insulation; cavity region between the insulation and biological shield; and the biological shield wall. Cladding is included on the inner RPV surface as well as the inner and outer surfaces of the biological shield wall. Also represented in Figure 2-1 are notations indicating the control rod and fuel assembly locations within the core. Note that the fuel locations are shown only for the northeast quadrant of the core region.

Non-Proprietary BFN-FLU-001-R-001-LNP Revision 1 Page 2-2 of 2-8 Figure 2-1 Planar View of the Browns Ferry Reactors at the Core Mid-Plane Elevation 2.2 Reactor System Mechanical Design Inputs The mechanical design inputs used to construct the Browns Ferry fluence geometry models are based upon ((

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((

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Non-Proprietary BFN-FLU-001-R-001-LNP Revision 1 Page 2-3 of 2-8 An important component of a computational reactor pressure vessel fluence model is the accurate description of the surveillance capsules installed in the pressure vessels. Figure 2-1 shows that the Browns Ferry reactors were initially equipped with three surveillance capsules each. The capsules were installed at an elevation around the reactor core mid-plane. Each capsule was mounted radially near the inside surface (0T) of the RPV wall. The surveillance capsules were distributed around the inner surface of the pressure vessel at the 30°, 120°, and 300° azimuths relative to reactor north, which is designated as the 0° cardinal direction. The importance of surveillance capsules in fluence analyses is that they contain flux wires that are irradiated during reactor operation. When a capsule is removed from the reactor, the irradiated flux wires are evaluated to obtain activity measurements.

For the Browns Ferry reactors, activation comparisons will be listed for the capsule containers and/or the 30° capsule flux wire holders that have been extracted for each individual reactor.

These measurements are used to validate the fluence model. The results of the measurement comparisons are provided in reactor-specific attachments to this report.

2.3 Reactor System Material Compositions Each region of the reactors is comprised of materials that include reactor fuel, metal, water, insulation, concrete, and air. Accurate material information is essential for the fluence evaluation as the material compositions determine the scattering and absorption of neutrons throughout the reactor system and, thus, affect the determination of neutron fluence in the RPV, surveillance capsules, vessel internal components, and ex-vessel structures.

Table 2-1 provides a summary of the materials for the principal components and regions of the Browns Ferry reactors. The material attributes for the metal, 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 bulk water coolant properties throughout the reactor system ((

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Non-Proprietary BFN-FLU-001-R-001-LNP Revision 1 Page 2-4 of 2-8 Table 2-1 Summary of Material Compositions by Region for Browns Ferry Region Material Composition Biological Shield Clad

((

Biological Shield Wall Cavity Regions Control Rod Guide Tubes Control Rods Core Exit Core Reflector Core Spray Sparger Nozzles Core Spray Sparger Piping Core Spray Sparger Flow Areas Core Support Plate Core Support Plate Rim Core Support Plate Rim Bolts Downcomer Region Fuel Support Pieces Fuel Hardware Regions In-Core Instrument Tube Jet Pump Assembly Hold Down Beams Jet Pump Assembly Hold Down Brackets Jet Pump Riser Pipe and Mixer Pipe Flow Areas Jet Pump Riser Pipe and Mixer Pipe Metal eel Jet Pump Riser Brace Jet Pump Riser Brace Pad Thermal Insulation Reactor Coolant / Moderator Reactor Core Reactor Pressure Vessel Clad Reactor Pressure Vessel Nozzle Forgings Reactor Pressure Vessel Wall Shroud Sparger Inlet Piping Steam Separator Standpipes Surveillance Capsule Flux Wire Holder Surveillance Capsule Specimen Top Guide

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Non-Proprietary BFN-FLU-001-R-001-LNP Revision 1 Page 2-5 of 2-8 2.4 Reactor Operating Data Inputs An accurate evaluation of reactor vessel and component fluence requires an accurate accounting of the reactor's operating history. The principal operating parameters that affect the determination of neutron fluence in light water reactors include ((

)) The following subsections provide additional information on the characterization of reactor operating data for fluence evaluations.

2.4.1 Core Configuration and Fuel Design The reactor core configuration and the fuel assembly designs loaded in the reactor determine the neutron source and spatial source distribution contributing to the irradiation of the pressure vessel, vessel internals and ex-vessel supporting structures. The Browns Ferry cores are comprised of 764 fuel assemblies. Several designs of fuel assemblies may be loaded in the reactor core in any given operating cycle. In order to determine accurate spatial fluence profiles throughout the reactor system, it is important ((

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Attachments to this report, herein made a part of this report, provide summaries of the different fuel assembly designs that have been loaded in each of the Browns Ferry reactor cores for each operating cycle.

2.4.2 Reactor Power History Reactor power history is the measure of reactor power levels ((

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Each attachment to this report provides reactor-specific summaries of the operating history of the Browns Ferry reactors for each operating cycle. Each attachment also shows the reactor-specific EFPY accumulated at the end of each cycle. The accumulated EFPY is computed from the operating data provided by TVA and is verified against power production and exposure data obtained separately for the reactors.

Non-Proprietary BFN-FLU-001-R-001-LNP Revision 1 Page 2-6 of 2-8 2.4.3 Reactor Statepoint Data Statepoints are snapshots in time that characterize the power-flow conditions of a reactor at a moment in time. Typically, several statepoints are used to represent the different operating conditions experienced by the reactor over the course of an operating cycle. The number of statepoints used to characterize a cycle of operation ((

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Core simulator data was provided by TVA to characterize the historical operating conditions of the Browns Ferry reactors. ((

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Because core simulator codes are used for a variety of core analysis functions, 10s to 100s of core calculations may be performed to track and monitor the operation of a reactor over the course of an operating cycle. ((

)) When all reactor conditions are considered, the number of core simulator statepoints selected for a fluence evaluation can ((

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A separate neutronics transport calculation is performed for each selected statepoint. ((

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2.4.4 Reactor Coolant Properties The reactor coolant water densities used in the fluence models are determined using combinations of core simulator codes and reactor heat balance data.

The water densities used for the core inlet and the reactor core region are ((

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The bulk water densities in the other regions of the reactor vessel are determined from ((

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METHODOLOGY This section provides an overview of the methodology and modeling approach used to determine fast neutron fluence for the Browns Ferry reactor pressure vessels (RPV), reactor vessel internals (RVI), and the fast neutron fluence and activations for the reactors surveillance capsules.

The fluence models for Browns Ferry are reactor-specific models that are constructed from the design inputs described in Section 2, Description of the Reactor System. The computational tools used in the fluence and activation analyses are based on the RAMA Fluence Methodology (RAMA) [3]. A general approach for using the toolset is presented in the RAMA Procedures Manual [2].

3.1 Computational Fluence Method The RAMA Fluence Methodology is a system of computer codes, a data library, and an uncertainty methodology that determines best-estimate fluence and activations in light water reactor pressure vessels and vessel internal components. The primary software that comprises the methodology includes model builder codes, a particle transport code, and a fluence calculator code.

The primary inputs for the fluence methodology are mechanical design parameters and reactor operating history data. The mechanical design inputs are obtained from ((

)). The reactor operating history data is obtained from multiple sources, such as ((

)). A variety of outputs are available from the fluence methodology that include neutron flux, fast neutron fluence, dosimetry activation, and an uncertainty analysis.

The model builder codes consist of geometry and material processor codes that generate input for the RAMA transport code. The geometry model builder code uses mechanical design inputs and meshing specifications to generate three-dimensional geometry models of the reactor. The material processor code uses reactor operating data and material property inputs to process fuel materials, structural materials, and water densities that are consistent with the geometry meshing generated by the geometry model builder code.

The RAMA transport code performs three-dimensional neutron flux calculations using a deterministic, multigroup, particle transport theory method with anisotropic scattering ((

)). The transport solver is coupled with a general geometry modeling capability ((

)). The coupling of general (arbitrary) geometry with a deterministic transport solver provides a flexible, efficient, and stable method for calculating neutron flux in light water reactor pressure vessels, vessel components, and structures. The primary inputs for the transport code include the geometry and material data generated by the model builder codes and numerical integration and convergence

Non-Proprietary BFN-FLU-001-R-001-LNP Revision 1 Page 3-2 of 3-12 parameters for the iterative transport calculation. The primary output from the transport code is the neutron flux in multigroup form for every material region mesh in the fluence model.

The fluence calculator code determines fluence and activation in the reactor pressure vessel, surveillance specimens, and vessel components over specified periods of reactor operation. The fluence calculator also includes treatments for isotopic production and decay that are required to calculate specific activities for irradiated materials, such as the dosimetry specimens in the surveillance capsules. The primary inputs to the fluence calculator include the multigroup neutron flux from the transport code, response functions for the various materials in the reactor, reactor power levels for the operating periods of interest, specification of which components to evaluate, and the energy ranges of interest for evaluating neutron fluence. ((

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The RAMA nuclear data library contains atomic mass data, nuclear cross-section data, response functions, and other nuclear constants that are needed for each of the code tools. The structure and contents of the data contained within the nuclear data file are based on the BUGLE-96 nuclear data library [15], with extended data representations derived from the VITAMIN-B6 data library [16].

The uncertainty methodology provides an assessment of the overall accuracy of the fluence and activation calculations. Variations in the ((

)) are evaluated to determine if there is a statistically significant bias in the calculated results that might affect the determination of the best-estimate fluence for the reactor. The plant-specific results are also weighted ((

)) to determine if the plant-specific model under evaluation is statistically acceptable as defined in Regulatory Guide 1.190 [8].

3.2 Fluence Model Section 2.2, Reactor System Mechanical Design Inputs, describes the design inputs that were provided by TVA for constructing the Browns Ferry reactor fluence models. These design inputs are used to develop reactor-specific, three-dimensional computational models of the Browns Ferry reactors for determining fast neutron fluence in the RPV and RVI components and for determining activation and fluence in reactor dosimetry for validating the RPV fluence predictions.

Figure 3-1 and Figure 3-2 provide general illustrations of the primary components, structures, and regions developed for the Browns Ferry fluence models. Figure 3-1 shows the planar configuration of the reactor models at an elevation corresponding to the reactor core mid-plane.

Figure 3-2 shows an axial configuration of the reactor models. ((

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((

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Figure 3-1 Planar View of the Browns Ferry Fluence Models at the Core Mid-Plane Elevation in Quadrant Symmetry

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((

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Figure 3-2 Axial View of the Browns Ferry Fluence Models

Non-Proprietary BFN-FLU-001-R-001-LNP Revision 1 Page 3-5 of 3-12 3.2.1 Geometry Models The Browns Ferry fluence models are constructed ((

)). The axial plane of the reactor models is defined by the (x,y) coordinates of the modeling system and the axial elevation at which a plane exists is defined along a perpendicular z-axis of the modeling system. This allows any point in the reactor models to be referenced by specifying the (x,y,z) coordinates for that point.

((

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This modeling approach permits a model to be developed in any level of high-definition detail, such as is necessary for fluence and activation evaluations.

Figure 2-1 illustrates a planar cross-section view of the Browns Ferry reactor design at an axial elevation corresponding to the reactor core mid-plane. It is shown for this one elevation that the reactor design is a complex geometry ((

)) When the reactor is viewed in three dimensions, the varying heights of the different components, structures, and regions create additional geometry modeling complexities. An accurate representation of these geometrical complexities in a predictive computer model is essential for calculating accurate, best-estimate fluence in the reactor pressure vessel, surveillance capsules, vessel internals, and the supporting structures inside and outside of the reactor vessel.

Figure 3-1 and Figure 3-2 provide general illustrations of the planar and axial geometry complexities that are represented in the fluence models. For comparison purposes, the planar view illustrated in Figure 3-1 corresponds to the core elevation illustrated in Figure 2-1. ((

))

Non-Proprietary BFN-FLU-001-R-001-LNP Revision 1 Page 3-6 of 3-12 3.2.2 Reactor Core and Core Reflector The reactor core contains the nuclear fuel that is the source of the neutrons that irradiate all components and structures of the reactor. The core is surrounded by a shroud structure that serves to channel the reactor coolant through the core region during reactor operation. The coolant-containing region between the core and the core shroud is the core reflector. The reactor core geometry is rectangular in design ((

))

3.2.3 Reactor Core Shroud The core shroud is a canister-like structure that surrounds the reactor core. It channels the reactor coolant and steam produced by the core into the steam separators. Axially the shroud extends almost the entire height of the model and is divided into three sections: lower, central, and upper.

The lower shroud extends from the bottom of the model to the core support plate flange, the central shroud extends from the core support plate flange to the top guide flange, and the upper shroud extends from top guide flange to the top of the shroud head rim.

((

))

Non-Proprietary BFN-FLU-001-R-001-LNP Revision 1 Page 3-7 of 3-12 Above the shroud wall is the shroud head which is penetrated by numerous steam separator standpipes. ((

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3.2.4 Downcomer Region The downcomer region lies between the core shroud and the reactor pressure vessel. The downcomer is effectively cylindrical in design, but with geometrical complexities created by the presence of the jet pump assemblies and surveillance capsules in the region. ((

))

3.2.4.1 Jet Pump Assemblies Each Browns Ferry unit has ten jet pump assemblies in the downcomer region, which provide the main recirculation flow for the core. ((

))

3.2.4.2 Surveillance Capsules The three (3) OEM surveillance capsules installed in the Browns Ferry reactors are positioned in close proximity to the RPV inner wall surface. The capsules are positioned at 30°, 120°, and 300°. ((

)) For more information concerning the specific capsules that are evaluated please see Section 2.2.

((

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3.2.5 Reactor Pressure Vessel The reactor pressure vessel and vessel cladding lie outside the downcomer region, ((

))

3.2.6 Thermal Insulation The reactor vessel thermal insulation lies in the cavity region outside the pressure vessel wall. ((

))

3.2.7 Inner and Outer Cavity Regions There are effectively two cavity regions represented in the model. The inner cavity region lies between the outer surface of the pressure vessel wall and the inner surface of the vessel insulation. The outer cavity region lies between the outer surface of the vessel insulation and inner surface of the biological shield wall cladding. ((

))

3.2.8 Biological Shield Model The biological shield (concrete) defines the outermost region of the fluence model. ((

))

Non-Proprietary BFN-FLU-001-R-001-LNP Revision 1 Page 3-9 of 3-12 3.2.9 Above-Core Components Figure 3-2 includes illustrations of other components and regions that lie above the reactor core region. ((

))

3.2.9.1 Top Guide The top guide component lies above the core region and is appropriately modeled to include discrete representations of the top guide plates. The top guide model also accounts for the fuel assembly parts and coolant flow between the plates. ((

))

3.2.9.2 Core Spray Spargers and Piping The core spray spargers include upper and lower sparger annulus pipes and a vertical inlet pipe.

((

))

3.2.10 Below-Core Components Figure 3-2 includes illustrations of other components and regions that lie below the reactor core region. ((

)) The lower shroud wall and fuel assembly components are described in previous sections, with the remaining components described in the following subsections.

3.2.10.1 Core Support Plate and Rim Bolts The core support plate includes appropriate penetrations for the fuel support pieces, control rod guide tubes, cruciform control rods, and the core support plate rim bolts. Core support plate rim bolts protrude from the top of the core support plate and traverse through the plate, rim, and core shroud lower flange. ((

))

Non-Proprietary BFN-FLU-001-R-001-LNP Revision 1 Page 3-10 of 3-12 3.2.10.2 Fuel Support Pieces The nuclear fuel assemblies loaded in the reactor are seated on fuel support pieces, which then rest in the core support plate and control blade guide tubes. ((

))

3.2.10.3 Control Blades and Guide Tubes The fluence models allow for the representation of cruciform-shaped control blades and tubular control blade guide tubes in the below-core regions of the reactors. Coolant flow paths are included in the models ((

))

3.2.11 Summary of the Geometry Modeling Approach To summarize the reactor modeling process, there are several key features that allow the reactor design to be accurately represented for RPV and RVI fluence evaluations. ((

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3.3 Particle Transport Calculation Parameters The accuracy of the transport method is based on a numerical integration technique ((

))

3.4 Fission Spectrum and Neutron Source Modern core simulator software is capable of providing three-dimensional core power distributions and fuel isotopics in high-definition detail, viz., on a pin-by-pin basis. This allows fluence models to be constructed with a high-level of modeling detail for representing unique

Non-Proprietary BFN-FLU-001-R-001-LNP Revision 1 Page 3-12 of 3-12 fission spectrum and neutron source terms for the transport calculation. ((

))

3.5 Parametric Sensitivity Analyses Several reactor-specific sensitivity analyses are performed to evaluate the accuracy and predictability of the neutral particle transport methodology for determining RPV and RVI component fluence. ((

))

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SURVEILLANCE CAPSULE EVALUATIONS AND COMBINED UNCERTAINTY ANALYSIS U.S. NRC Regulatory Guide 1.190 [8] requires that fluence calculational methods be validated by comparison to operating reactor dosimetry measurements. ((

)) The acceptance criteria provided in Regulatory Guide 1.190 is that standard deviations determined from the calculated-to-measurement comparison ratios (C/M) fall within a computed standard deviation of +/- 20%.

Attachments to this report, herein made a part of this document, present the computed activation and fluence results determined for the Browns Ferry reactor surveillance capsules and flux wires that were removed from the reactors at different exposures in the reactors operating life. The activation results form the basis for the validation and qualification of the fluence methodology for the Browns Ferry reactors in accordance with requirements of Regulatory Guide 1.190. The attachments also present the results of the combined uncertainty analysis that was performed for each of the Browns Ferry reactors.

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REFERENCES 5.1 References

1. BWRVIP-126, Revision 2: BWR Vessel and Internals Project, RAMA Fluence Methodology Software, Version 1.20. EPRI, Palo Alto, CA: 2010. 1020240.

2. BWRVIP-121-A: BWR Vessel and Internals Project, RAMA Fluence Methodology Procedures Manual. EPRI, Palo Alto, CA: 2009. 1019052.

3. BWRVIP-114-A: BWR Vessel and Internals Project, RAMA Fluence Methodology Theory Manual. EPRI, Palo Alto, CA: 2009. 1019049.

4. U.S. Nuclear Regulatory Commission. Office of Nuclear Reactor Regulation. Safety Evaluation Report with Open Items Related to the License Renewal of Seabrook Station.

Docket Number 50-443. Washington, D.C.: Office of Nuclear Reactor Regulation: 2012

5. U.S. National Archives and Records Administration. Code of Federal Regulations.

Title 10, Appendix G to Part 50 - Fracture Toughness Requirements: 2013

6. U.S. National Archives and Records Administration. Code of Federal Regulations.

Title 10, Appendix H to Part 50 - Reactor Vessel Material Surveillance Program Requirements: 2008

7. U.S. Nuclear Regulatory Commission. Office of Nuclear Regulatory Research. Regulatory Guide 1.99: Radiation Embrittlement of Reactor Vessel Materials. Revision 2. Washington, D.C.: Office of Nuclear Regulatory Research: 1988

8. U.S. Nuclear Regulatory Commission. Office of Nuclear Regulatory Research. Regulatory Guide 1.190: Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence. Washington, D.C.: Office of Nuclear Regulatory Research: 2001

9. Safety Evaluation of Proprietary EPRI Report BWR Vessel and Internals Project, Evaluation of Susquehanna Unit 2 Top Guide and Core Shroud Material Samples Using RAMA Fluence Methodology (BWRVIP-145). Matthew A. Mitchell (U.S. NRC) to Rick Libra (BWRVIP). February 7, 2008.

10. BWRVIP-145-A: BWR Vessel and Internals Project, Evaluation of Susquehanna Unit 2 Top Guide and Core Shroud Material Samples Using RAMA Fluence Methodology. EPRI, Palo Alto, CA: 2009. 1019053.

11. U.S. National Archives and Records Administration. Code of Federal Regulations.

Title 10, Appendix B to Part 50 - Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants: 2007

12. U.S. National Archives and Records Administration. Code of Federal Regulations.

Title 10, Part 21 - Reporting of Defects and Noncompliance: 2015

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13. Fluence Evaluation for Browns Ferry Unit 2 Reactor Pressure Vessel Using RAMA Fluence Methodology, TransWare Enterprises Inc., EPR-BF2-001-R-003, Revision 0, December 2012.

14. ((

))

15. Oak Ridge National Laboratory. Radiation Safety Information Computational Center.

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. Oak Ridge, TN: 1996

16. Oak Ridge National Laboratory. Radiation Safety Information Computational Center.

VITAMIN-B6: A Fine-Group Cross Section Library Based on ENDF/B-VI Release 3 for Radiation Transport Applications. RSICC Data Library Collection, DLC-184, Oak Ridge, TN: 1996

Non-Proprietary BFN-FLU-001-R-001-LNP Revision 1 Page 5-3 of 5-4 5.2 Glossary AZIMUTHAL QUADRANT SYMMETRY - A type of core and pressure vessel azimuthal representation that represents a single quadrant of the reactor that can be rotated and mirrored to represent the entire 360-degree geometry. For example, the northeast quadrant can be mirrored to represent the northwest and southeast quadrants and can be rotated to represent the southwest quadrant.

BEST-ESTIMATE NEUTRON FLUENCE - See Neutron Fluence.

BOC - An acronym for beginning-of-cycle.

CALCULATED NEUTRON FLUENCE - See Neutron Fluence.

CALCULATIONAL BIAS - A calculational adjustment based on comparisons of calculations to measurements. If a bias is determined to exist, it may be applied as a multiplicative correction to the calculated fluence to produce the best-estimate neutron fluence.

CORE BELTLINE - The axial elevations corresponding to the active fuel height of the reactor core.

DAMAGE FLUENCE - See Neutron Fluence.

DPA - An acronym for displacements per atom which is typically used to characterize material damage in ferritic steels due to neutron exposure.

EFFECTIVE FULL POWER YEARS (EFPY) - A unit of measurement representing one full year of operation at the reactors rated power level. For example, if a reactor operates for 12 months at full rated power, this represents 1.0 EFPY. If the reactor operates for 10 months at full rated power, then goes into a power uprate and continues operating for another 2 months at the new full rated power, this also represents 1.0 EFPY.

EOC - An acronym for end-of-cycle.

EXTENDED BELTLINE REGION - See RPV beltline.

FAST NEUTRON FLUENCE - Fluence accumulated by neutrons with energy greater than 1.0 MeV (E > 1.0 MeV).

NEUTRON FLUENCE - Time-integrated neutron flux reported in units of n/cm2. The term best-estimate fluence refers to the fast neutron fluence that is computed in accordance with the requirements of U.S. Nuclear Regulatory Commission Regulatory Guide 1.190. The term damage fluence, which is required for material embrittlement evaluations, refers to an adjusted fast neutron fluence that is determined using damage functions specified in U.S. Nuclear Regulatory Commission Regulatory Guide 1.99.

OEM - An acronym for Original Equipment Manufacturer.

RPV - An acronym for reactor pressure vessel. Unless otherwise noted, the reactor pressure vessel refers to the base metal material of the RPV wall (i.e., excluding clad/liner).

RPV BELTLINE - The RPV beltline is defined as that portion of the RPV adjacent to the reactor core that attains sufficient neutron radiation damage that the integrity of the pressure vessel could be compromised. For purposes of this evaluation, the fast neutron fluence threshold used to

Non-Proprietary BFN-FLU-001-R-001-LNP Revision 1 Page 5-4 of 5-4 define the traditional RPV beltline is 1.0E+17 n/cm2. The axial span of the RPV that can exceed this threshold includes the RPV shells, welds, and heat-affected zones. An extended beltline is also defined to include lower fluence regions of the pressure vessel but with higher stresses than the traditional beltline region, such as RPV nozzles. The combination of fluence and stress may result in a limiting location in the pressure vessel for determining pressure-temperature limits.

RPV ZERO ELEVATION - The RPV zero elevation is defined at the inside surface of the lowest point in the vessel bottom head, which is typically the bottom drain plug location. Axial elevations presented in this report are relative to RPV zero.

RVI - An acronym for reactor vessel internals.

Non-Proprietary BFN-FLU-001-R-001-LNP, Revision 1 Page i of x Topical Report BROWNS FERRY NUCLEAR PLANT FLUENCE METHODOLOGY REPORT Qualification of the Browns Ferry Unit 2 Reactor Fluence Model - Cycles 1 to 21 Document Number:

BFN-FLU-001-R-001-LNP, Revision 1 October 2023 Prepared by:

TransWare Enterprises Inc.

Prepared for:

Tennessee Valley Authority 1101 Market Street Chattanooga, TN 37402 Contract Number:

15904 Project Manager:

John Paul Anglin

((

))

Controlled Copy Number: ____2____

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Non-Proprietary BFN-FLU-001-R-001-LNP, Revision 1 Page iii of x Topical Report BROWNS FERRY NUCLEAR PLANT FLUENCE METHODOLOGY REPORT :

Qualification of the Browns Ferry Unit 2 Reactor Fluence Model - Cycles 1 to 21 Document Number: BFN-FLU-001-R-001-LNP, Revision 1 October 2023 Prepared By:

TransWare Enterprises Inc.

Project Team:

Project Manager:

E. A. Evans, Project Engineer H. J. Heppermann, Project Engineer M. E. Jewell, Project Engineer J. S. Styczynski, Project Engineer S. M. Wagstaff, Project Engineer K. E. Watkins, Project Engineer D.B. Jones 10/25/2023 D. B. Jones, Project Manager Date Reviewed By:

K.E. Watkins 10/25/2023 K. E. Watkins, Project Engineer Date K.A. Jones 10/25/2023 K. A. Jones, QA Specialist Date Approved By:

D.B. Jones 10/25/2023 D. B. Jones, Project Manager Date Prepared For:

Tennessee Valley Authority 1101 Market Street Chattanooga TN, 37402 Contract Number: 15904 Project Manager: John Paul Anglin Official signatures are on file. Contact TransWare Enterprises to request a copy.

TransWare Enterprises Inc.

• 520 Maryville Centre Dr., Suite 125

• St. Louis, Missouri 63141

• USA

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, Revision 1 Page iv of x DISCLAIMER OF WARRANTIES AND LIMITATION OF LIABILITIES THE INFORMATION CONTAINED IN THIS REPORT IS BELIEVED BY TRANSWARE ENTERPRISES INC. TO BE AN ACCURATE AND TRUE REPRESENTATION 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.

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, Revision 1 Page v of x CONTENTS Title Page 1

Introduction................................................................................................................... 1-1 2

Reactor Operating History........................................................................................... 2-1 3

Reactor Statepoint Data............................................................................................... 3-1 4

Surveillance Capsule Dosimetry Evaluation............................................................... 4-1 4.1 Summary of the Flux Wire Activation Analysis........................................................4-1 4.2 Comparison of Calculated Activation to Reactor-specific Measurements................4-4 4.2.1 Flux Wire Activation Analysis for the Browns Ferry Unit 2 30° Capsule......4-4 4.2.2 Cycle 7 Surveillance Capsule Activation Analysis......................................4-5 4.2.3 Cycle 16 Surveillance Capsule Activation Analysis....................................4-7 4.3 Reactor Pressure Vessel Lead Factors...................................................................4-8 5

Reactor Pressure Vessel Fluence Uncertainty Analysis............................................ 5-1 5.1 Comparison Uncertainty.........................................................................................5-1 5.1.1 Operating Reactor Comparison Uncertainty...............................................5-1 5.1.2 Benchmark Comparison Uncertainty.........................................................5-2 5.2 Analytic Uncertainty................................................................................................5-2 5.3 Combined Uncertainty............................................................................................5-3 6

References.................................................................................................................... 6-1 6.1 References.............................................................................................................6-1 7

Figures Including Dimensional Information................................................................ 7-1 7.1 Figures...................................................................................................................7-1

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, Revision 1 Page vii of x LIST OF TABLES Title Page Table 2-1 Summary of Browns Ferry Unit 2 Core Loading Inventory for Cycles 1 to 12C..... 2-1 Table 2-2 Summary of Browns Ferry Unit 2 Core Loading Inventory for Cycle 13 to 21, 22prj, and 99prj.................................................................................................... 2-2 Table 3-1 Statepoint Data for Browns Ferry Unit 2 per Cycle Basis...................................... 3-2 Table 4-1 Summary of the Fluence and Activity Comparisons for the Browns Ferry Unit 2 Dosimetry............................................................................................................. 4-3 Table 4-2 Summary of the Activity Comparisons for the Copper, Iron, and Nickel Flux Wires Removed From the Browns Ferry Unit 2 Reactor....................................... 4-3 Table 4-3 Summary of the Activity Comparisons for the Copper, Iron, and Nickel Flux Wires Removed From the Browns Ferry Unit 2 Reactor Without Cycle 7............. 4-4 Table 4-4 Comparison of the Calculated-to-Measured Activities for the Iron and Copper Flux Wires Removed From the Browns Ferry Unit 2 30° Flux Wire Holder........... 4-5 Table 4-5 Comparison of the Calculated-to-Measured Activities for the Copper, Iron, and Nickel Flux Wires Removed From the Browns Ferry Unit 2 30° Surveillance Capsule at EOC 7................................................................................................ 4-5 Table 4-6 Comparison of the Calculated-to-Measured Activities for the Copper, Iron, and Nickel Flux Wires Removed From the Browns Ferry Unit 2 120° Surveillance Capsule at EOC 16.............................................................................................. 4-7 Table 4-7 Best-Estimate Fluence Determined for the Browns Ferry Unit 2 Surveillance Capsules.............................................................................................................. 4-8 Table 4-8 Lead Factors Determined for the Surveillance Capsules Removed From Browns Ferry Unit 2.............................................................................................. 4-8 Table 5-1 Summary of Comparisons to Vessel Simulation Benchmark Measurements........ 5-2 Table 5-2 Browns Ferry Unit 2 RPV Combined Uncertainty for Energy > 1.0 MeV................ 5-3

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, Revision 1 Page ix of x LIST OF FIGURES Title Page Figure 4-1 Positioning of the Surveillance Capsules Installed in the Browns Ferry Unit 2 Reactor................................................................................................................ 4-2 Figure 7-1 Browns Ferry Unit 2 RPV Beltline Region at 64 EFPY.......................................... 7-1 Figure 7-2 Identification of the Browns Ferry Unit 2 Core Shroud Welds................................ 7-2 Figure 7-3 Axial View of the Browns Ferry Unit 2 Jet Pump Model with Weld Identifications.. 7-3

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INTRODUCTION This attachment provides the reactor operating history, comparisons to activation measurements, and uncertainty analysis that are essential for validating the Browns Ferry Nuclear Plant Unit 2 (Browns Ferry Unit 2) fluence methodology. The methodology that is used for determining the neutron fluence in the Browns Ferry Unit 2 reactor is detailed in the Browns Ferry Nuclear Plant Fluence Methodology Report [1].

The power history data presented in this report covers the time period from start of commercial operation to the end of operating Cycle 21. The reactor began commercial operation in 1974 with a rated thermal power of 3293 MWth. A power uprate occurred at the beginning of Cycle 11 to increase the rated thermal power to 3458 MWth. The rated thermal power was increased a second time to 3952 MWth at the beginning of Cycle 21. All surveillance dosimetry removed from the reactor over that time period, and which is available in the form of activation measurements, is evaluated. A combined uncertainty factor for the fluence model based on the modeling approach and measurement comparisons is determined which demonstrates that the computational fluence method used by TransWare Enterprises Inc. is qualified for use in determining neutron fluence for the Browns Ferry Unit 2 reactor pressure vessel in accordance with U.S. Nuclear Regulatory Commission (U.S. NRC) Regulatory Guide 1.190 [2].

In compliance with Regulatory Guide 1.190, it is shown in this report that the calculated-to-measured (C/M) ratio and standard deviation is 1.12 +/- 0.18 for all reactor surveillance capsule dosimetry evaluated for the Browns Ferry Unit 2 reactor. The combined uncertainty for the Browns Ferry Unit 2 reactor is determined to be 9.50%. Based upon these results, there is no discernable bias in the computed reactor pressure vessel fluence for the period Cycle 1 through the end of Cycle 21 for the Browns Ferry Unit 2 reactor.

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REACTOR OPERATING HISTORY Reactor operating history is the measure of daily reactor power levels that characterize the radiation exposure history of a reactor over its operating life. The daily power history data for the Browns Ferry Unit 2 reactor was provided by TVA in discrete form for Cycles 1 through 21.

Another important element of the reactor operating history is the fuel designs that were loaded in the reactor for each operating cycle. ((

))

Table 2-1 and Table 2-2 provide a summary of the fuel designs that were loaded in the Browns Ferry Unit 2 reactor for each operating cycle. Table 2-1 lists the fuel designs that were loaded in the reactor for Cycles 1 through 12. Table 2-2 lists the fuel designs that were loaded for Cycles 13 through 21, the fuel designs that are loaded in partial-projection Cycle 22 (22prj), and the fuel design for the equilibrium projection cycle identified as Cycle 99 (99prj). The dominant (most numerous) fuel design that was loaded in the core for each cycle is shown in bold font.

The dominant fuel design that was loaded on the core periphery is identified in blue font.

Table 2-1 Summary of Browns Ferry Unit 2 Core Loading Inventory for Cycles 1 to 12C Cycle Fuel Designs 7x7 8x8 9x9

((

))

1 168 596 2

38 594 132 3

364 168 232 4

124 168 232 240 5

123 153 488 6

70 354 336 4

7 212 388 160 4

8 4

580 176 4

9 348 216 200 10 148 200 200 216 11 48 200 516 12A 764 12B 764 12C 49 715

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((

))

13 111 281 372 14 149 335 280 15 111 653 16 764 17 764 18 764 19 504 252 8

20 219 539 6

21 758 6

22prj(1) 760 4

99prj(2) 764 (1) For purposes of fluence projections, operation of projection Cycle 22 is partially historical operation data and partially projected operating data. This cycle was provided by TVA to predict fluence for the duration of one cycle.

(2) For purposes of fluence projections, operation beyond projection Cycle 99 uses an equilibrium projection cycle featuring a full core loading of ((

)) fuel. This cycle was provided by TVA to predict fluence at the end of the extended plant license period.

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REACTOR STATEPOINT DATA The reactor operating history is defined in discrete exposure steps referred to as statepoints. A statepoint is defined as a snapshot ((

)) of a reactor core at a moment in time. ((

))

Several statepoints are generally used to represent the different operating states of a reactor core over the course of an operating cycle. The core power distribution for a statepoint is generally determined using core simulator software. Because core simulator codes are used for a variety of core analysis functions, 10s to 100s of core calculations may be performed to track and monitor the operation of a reactor. ((

))

Table 3-1 shows that a total of ((

)) statepoints were used to represent the operating states of the Browns Ferry Unit 2 reactor for the first 21 cycles of operating history. The core simulator data was provided by TVA to characterize the historical operating conditions for the Browns Ferry Unit 2 reactor. Partial projection data was also provided for Cycle 22 (22prj) and for an equilibrium cycle (99prj) for projecting fluence to the reactors end of license period.

Table 3-1 also shows the rated thermal power of the reactor for a cycle and the accumulated effective full power years (EFPY) of exposure accumulated for the cycle.

((

))

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Accumulated EFPY 1

((

3293 1.3 2

3293 2.0 3

3293 3.0 4

3293 4.4 5

3293 5.5 6

3293 6.9 7

3293 8.1 8

3293 9.4 9

3293 10.7 10 3293 12.2 11(1) 3458 14.0 12A 3458 15.0 12B 3458 15.4 12C 3458 15.7 13 3458 17.6 14 3458 19.4 15 3458 21.3 16 3458 22.9 17 3458 24.7 18 3458 26.5 19 3458 28.3 20 3458 30.2 21(2) 3952 31.9 22prj(3) 3952 33.8 99prj(4)

))

3952 64.0 (1) A power uprate was implemented at the beginning of Cycle 11 from 3293 MWth to 3458 MWth.

(2) A power uprate was implemented at the beginning of Cycle 21 from 3458 MWth to 3952 MWth.

(3) Cycle 22prj is a partial projection cycle.

(4) Cycle 99prj is a projection cycle assuming a full core loading of ((

)) fuel. This cycle will be used to project reactor fluence to the end of the reactors licensed period of operation.

EFPY is an effective measurement of the number of years that a reactor has operated at full power conditions but does not account for the changes in rated power. Therefore, one EFPY is the same measured value irrespective of the rated power of the reactor. Consequently, EFPY should not be used to interpolate between periods of different rated powers.

A separate neutronics transport calculation is performed for each statepoint listed in Table 3-1.

The neutron fluxes calculated for each statepoint are then combined with daily thermal power information to provide an integral accounting of the neutron fluence for the reactor pressure vessel, reactor vessel internals, and surveillance capsules. The periods of reactor shutdown are

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SURVEILLANCE CAPSULE DOSIMETRY EVALUATION This section presents the results of the activation analysis and the determination of fast neutron fluence for the Browns Ferry Unit 2 surveillance capsule dosimetry. Lead factors associating the peak RPV fluence with the capsule fluence are reported in Section 4.3. The results presented in this section form the basis for the validation and qualification of the fluence methodology as applied to the Browns Ferry Unit 2 reactor in accordance with the U.S Nuclear Regulatory Commission Regulatory Guide 1.190 (RG 1.190) [2].

Regulatory Guide 1.190 requires fluence calculational methods to be validated by comparisons with activation measurements from operating reactor dosimetry. It is preferred that the activation data be taken from the reactor being evaluated. ((

)) In the case for the Browns Ferry Unit 2 reactor, there is sufficient reactor-specific measurements available to qualify the calculational method ((

))

In order to report computed fluence as the best-estimate fluence, RG 1.190 requires that the standard deviation resulting from the comparison of calculated to measurement data should be 20%. It is determined that the overall calculated-to-measured (C/M) comparison ratio and standard deviation for all Browns Ferry Unit 2 reactor surveillance dosimetry is 1.12 +/- 0.18. It will be presented in the following subsections that the dosimetry measurements for the flux wires removed at the end of Cycle 7 (EOC 7) appear to be unreliable. If the EOC 7 dosimetry measurements are omitted, the calculated to measured (C/M) comparison ratio and standard deviation for the reliable dosimetry measurements is 1.00 +/- 0.12. While it is determined that the Cycle 7 dosimetry should be omitted, it is noted that both analyses, with and without the Cycle 7 results, meet the acceptance criteria per RG 1.190. Therefore, the computational fluence model for the Browns Ferry Unit 2 reactor meets the RG 1.190 criteria and, as such, no bias adjustment is required to be applied to the computed RPV fluence.

4.1 Summary of the Flux Wire Activation Analysis Three (3) surveillance capsules were initially installed in the Browns Ferry Unit 2 reactor. The capsules are mounted axially near the reactor core mid-plane elevation. The capsules are positioned near the inner surface of the reactor pressure vessel wall at the 30°, 120°, and 300° azimuths around the circumference of the reactor pressure vessel. Figure 4-1 illustrates the positioning of the surveillance capsules in the Browns Ferry Unit 2 reactor.

The 30° and 120° surveillance capsules were removed from the Browns Ferry Unit 2 reactor for testing of radiation effects on the Charpy specimens and activation analysis of the flux wires that were contained in the capsules. In addition, one (1) set of flux wires were extracted from the flux wire holder attached to the 30° surveillance capsule for activation analysis.

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, Revision 1 Page 4-2 of 4-8 Figure 4-1 Positioning of the Surveillance Capsules Installed in the Browns Ferry Unit 2 Reactor

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, Revision 1 Page 4-3 of 4-8 Table 4-1 provides a summary of the activation comparisons for each set of flux wire that were irradiated in the Browns Ferry Unit 2 reactor. The table also provides the irradiation period in terms of cycle exposure, the accumulated exposure in terms of EFPY of reactor operation, the average fast neutron fluence (E > 1.0 MeV) for each set of flux wires, and the number of specimens (viz., flux wires) evaluated for each set of flux wires removed from the reactor.

In summary, Table 4-1 shows that the overall C/M ratio and associated standard deviation for the 21 irradiated specimens was determined to be 1.12 +/- 0.18. It is presented in Section 4.2.2 that the data for the flux wires removed at EOC 7 appear to be anomalous and, therefore, should be omitted from the final analysis. When the EOC 7 results are omitted, Table 4-1 shows that the average C/M ratio and associated standard deviation is 1.00 +/- 0.12. It is noted that both results meet the acceptance criteria of RG 1.190, although the results without the EOC 7 results are considered most reliable.

Table 4-1 Summary of the Fluence and Activity Comparisons for the Browns Ferry Unit 2 Dosimetry Dosimeter Cycles of Exposure Accumulated Exposure (EFPY)

Fast Neutron Fluence

(>1 MeV, n/cm2)

Number of Specimens Calculated vs.

Measured (C/M)

Standard Deviation

()

30° Flux Wire 1 - 1 1.3 6

((

((

30° Capsule 1-7 8.1

((

9 120° Capsule 1-16 22.9

))

6

))

))

Overall Average C/M and Standard Deviation 21 1.12 0.18 Average C/M and Standard Deviation Without Cycle 7 12 1.00 0.12 Table 4-2 provides the overall C/M ratios and standard deviations considering only the copper, iron, and nickel flux wires that were irradiated in the Browns Ferry Unit 2 reactor. This table includes the anomalous EOC 7 flux wire measurement. Table 4-3 shows the same information without the EOC 7 flux wire measurements.

Table 4-2 Summary of the Activity Comparisons for the Copper, Iron, and Nickel Flux Wires Removed From the Browns Ferry Unit 2 Reactor Flux Wire Number of Specimens C/M Copper 8

((

((

Iron 8

Nickel 5

))

))

Overall Average C/M and Standard Deviation 21 1.12 0.18

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, Revision 1 Page 4-4 of 4-8 Table 4-3 Summary of the Activity Comparisons for the Copper, Iron, and Nickel Flux Wires Removed From the Browns Ferry Unit 2 Reactor Without Cycle 7 Flux Wire Number of Specimens C/M Copper 5

((

((

Iron 5

Nickel 2

))

))

Overall Average C/M and Standard Deviation Without Cycle 7 12 1.00 0.12 4.2 Comparison of Calculated Activation to Reactor-specific Measurements The comparison of calculated activations to measurements for each capsule are presented in this subsection. The fast neutron fluence and lead factors for each capsule are reported in Subsection 4.3, Reactor Pressure Vessel Lead Factors.

4.2.1 Flux Wire Activation Analysis for the Browns Ferry Unit 2 30° Capsule Copper and iron flux wires were irradiated in the Browns Ferry Unit 2 30° surveillance capsule flux wire holder during the first cycle of reactor operation. At the time of their removal, the flux wires had been irradiated for a total of 1.3 EFPY.

Table 4-4 shows the activation measurements, periods of irradiation, computed activations, and the computed-to-measured (C/M) ratios for the flux wires irradiated in the 30° flux wire holder.

Activation calculations were performed for the following reactions: 63Cu (n,) 60Co and 54Fe (n,p) 54Mn.

The average C/M ratio and associated standard deviation determined for all of the flux wires irradiated in the Browns Ferry Unit 2 30° flux wire holder is 0.89 +/- 0.05.

It is noted that the C/M results presented in Table 4-4 ((

)) are acceptable per RG 1.190.

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, Revision 1 Page 4-5 of 4-8 Table 4-4 Comparison of the Calculated-to-Measured Activities for the Iron and Copper Flux Wires Removed From the Browns Ferry Unit 2 30° Flux Wire Holder Flux Wire Dosimeter Measured (dps/mg) [3]

Calculated (dps/mg)

C/M Copper Cu-1 3.20E+00

((

((

Cu-2 3.13E+00 Cu-3 3.05E+00

))

))

Copper Average

((

))

Iron Fe-1 5.81E+01

((

((

Fe-2 5.88E+01 Fe-3 5.74E+01

))

))

Iron Average

((

))

Average C/M and Overall Standard Deviation 0.89 0.05 4.2.2 Cycle 7 Surveillance Capsule Activation Analysis Copper, iron, and nickel flux wires were irradiated in the Browns Ferry Unit 2 30° surveillance capsule during the first seven (7) cycles of reactor operation. At the time of their removal, the flux wires had been irradiated for a total of 8.1 EFPY.

Activation calculations were performed for the following reactions: 63Cu (n,) 60Co, 54Fe (n,p) 54Mn, and 58Ni (n,p) 58Co. The precise location of the individual wires within the capsule is not known; therefore, the activation calculations were performed at the center of the capsule container.

Table 4-5 compares the C/M specific activities for each iron, nickel, and copper flux wire that was irradiated in the Browns Ferry Unit 2 reactor through the end of Cycle 7. It is noted that only one measurement is provided for each flux wire. Therefore, a computed standard deviation for each flux wire is not determined.

The average C/M ratio and associated standard deviation determined for the flux wires that were irradiated in the Browns Ferry Unit 2 30° surveillance capsule is 1.29 +/- 0.07.

Table 4-5 Comparison of the Calculated-to-Measured Activities for the Copper, Iron, and Nickel Flux Wires Removed From the Browns Ferry Unit 2 30° Surveillance Capsule at EOC 7 Flux Wire Measured (dps/mg) [4]

Calculated (dps/mg)

C/M Copper 5.62E+03

((

((

Iron 6.05E+04 Nickel 1.07E+06

))

))

Average C/M and Overall Standard Deviation 1.29 0.07

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, Revision 1 Page 4-6 of 4-8 It is shown in Table 4-5 that the RAMA Fluence Methodology overpredicted the EOC 7 flux wire activation measurements per RG 1.190. This was investigated. It was found by reviewing the original capsule evaluation report [4] that the prior evaluators also observed similar overestimates for the Cycle 7 dosimetry. The evaluators reported that they used a DORT 2D transport method [no reference given] and offered the following conclusion:

The transport calculation of the surveillance capsule flux [] is about 49% higher than the dosimetry result of []. This is attributed to conservatism incorporated in the transport calculation model and may, in part, result from the use of nominal rather than as-built radius.

The RAMA Fluence Methodology, which is a best-estimate methodology, showed similar overpredictions, but not to the same degree as the original evaluators. Contrary to the prior evaluators conclusions, the overprediction does not appear to be due to the use of conservative modeling approaches. As two independent evaluators using different computational fluence methods and modeling philosophies each observed similar trends, more reasonable explanations could be:

1) The surveillance capsule or the flux wires within the capsule were not precisely documented and, therefore, biased the transport calculations. ((

))

2) The activation measurements could be anomalous for unknown reasons, including activation decay corrections or mis-calibrated measurement devices.

In either scenario, the EOC 7 measurement data is not considered reliable for validating a computational fluence method. The measurement data should, therefore, be omitted from the final analysis as unreliable. Even though the EOC 7 data is considered unreliable, it is shown in Table 4-1, Table 4-2, and Table 4-3, above, that when the EOC 7 results are averaged with the EOC 1 and EOC 16 results, the overall C/M average meets the acceptance criteria of RG 1.190.

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, Revision 1 Page 4-7 of 4-8 4.2.3 Cycle 16 Surveillance Capsule Activation Analysis Copper, iron, and nickel flux wires were irradiated in the Browns Ferry Unit 2 120° surveillance capsule during the first sixteen (16) cycles of reactor operation. At the time of their removal, the flux wires had been irradiated for a total of 22.9 EFPY.

Activation calculations were performed for the following reactions: 63Cu (n,) 60Co, 54Fe (n,p) 54Mn, and 58Ni (n,p) 58Co. The precise location of the individual wires within the capsule is not known; therefore, the activation calculations were performed at the center of the capsule container.

Table 4-6 provide comparisons of the calculated-to-measured specific activities for each iron, nickel, and copper flux wire removed from the Browns Ferry Unit 2 reactor at end of Cycle 16.

The average of the calculated-to-measured ratio for each flux wire is reported in the table below.

Table 4-6 Comparison of the Calculated-to-Measured Activities for the Copper, Iron, and Nickel Flux Wires Removed From the Browns Ferry Unit 2 120° Surveillance Capsule at EOC 16 Flux Wire Dosimeter Measured (dps/mg) [5]

Calculated (dps/mg)

C/M Copper Cu-G9 1.37E+01

((

((

Cu-G10 1.37E+01

))

))

Copper Average

((

))

Iron Fe-G9 8.00E+01

((

((

Fe-G10 8.52E+01

))

))

Iron Average

((

))

Nickel Ni-G9 1.08E+03

((

((

Ni-G10 1.12E+03

))

))

Nickel Average

((

))

Average C/M and Overall Standard Deviation 1.11 0.06

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, Revision 1 Page 4-8 of 4-8 4.3 Reactor Pressure Vessel Lead Factors Table 4-7 and Table 4-8 provide the best-estimate fast neutron fluence determined for the reactor pressure vessel and dosimeters and the associated lead factors for each evaluated time period for the Browns Ferry Unit 2 reactor.

Table 4-7 Best-Estimate Fluence Determined for the Browns Ferry Unit 2 Surveillance Capsules Evaluated Time Period Dosimeter Fluence (n/cm2)

Peak RPV Fluence (n/cm2 30° 120° 300° 0T 1/4T EOC 7(1) 2.51E+17 2.45E+17 1.72E+17 EOC 16 6.83E+17 6.72E+17 4.71E+17

1) The fluence for EOC 7 is provided for information only.

Table 4-8 Lead Factors Determined for the Surveillance Capsules Removed From Browns Ferry Unit 2 Evaluated Time Period Lead Factor (1) 30° 120° 300° 0T 1/4T 0T 1/4T 0T 1/4T EOC 7(2) 1.02 1.45 EOC 16 1.02 1.45

1) The lead factor is defined as the ratio of the fast neutron fluence at the center of the surveillance capsule to the peak fast neutron fluence at the base metal inner surface (0T) of the RPV. A second lead factor is also provided assuming the peak damage fluence at the 1/4T depth of the RPV wall.

2) The lead factors for EOC 7 are provided for information only.

The calculated-to-measured activation comparisons for the surveillance capsules presented in the previous sections show no discernable bias in the computational fluence method. Therefore, the best-estimate fluence reported for each capsule in Table 4-7 is the fast neutron fluence computed by the fluence methodology.

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REACTOR PRESSURE VESSEL FLUENCE UNCERTAINTY ANALYSIS This section presents the combined uncertainty analysis and the determination of bias for the Browns Ferry Unit 2 reactor pressure vessel (RPV) fluence evaluation. The combined uncertainty is comprised of two components. ((

))

When combined, these components provide a basis for determining the combined uncertainty (1) and bias in the computed RPV fluence.

The requirements for determining the combined uncertainty and bias for light water reactor pressure vessel fluence evaluations are provided in Regulatory Guide 1.190 [2]. The approach for determining combined uncertainty and bias for reactor pressure vessel fluence is demonstrated in Reference 6.

For pressure vessel fluence evaluations, two uncertainty factors are considered: comparison factors and uncertainty introduced by the measurement process. After analysis of these factors, it is determined that the combined uncertainty for Browns Ferry Unit 2 RPV fluence is 9.50%, and that no adjustment for bias is required for the RPV fast neutron fluence determined for the period Cycle 1 through the end of Cycle 21.

5.1 Comparison Uncertainty Comparison uncertainty factors are determined by comparing calculated activities with activity measurements. For pressure vessel fluence evaluations, two comparison uncertainty factors are considered: operating reactor comparison factors and benchmark comparison factors.

5.1.1 Operating Reactor Comparison Uncertainty TransWare has evaluated activation measurements for several BWR plants ranging from BWR/2-class plants to BWR/6-class plants. ((

))

The Browns Ferry Unit 2 reactor is a BWR/4 class design. A total of ((

)) plant-specific dosimetry measurements have been evaluated for BWR/4-764 class plants using the RAMA Fluence Methodology. The overall C/M and standard deviation for the BWR/4-764 class plant measurements is determined to be an unbiased 1.04 +/- 0.13. ((

))

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, Revision 1 Page 5-2 of 5-4 A total of over ((

)) credible activation measurement comparisons have been previously performed for a broad spectrum of BWR configurations and designs using the RAMA Fluence Methodology. The overall comparison ratio for all BWR class plants evaluated as of the date of this report is 1.01 +/- 0.10. [

))

5.1.2 Benchmark Comparison Uncertainty The benchmark comparison uncertainty is based on a set of industry standard simulation benchmark comparisons. [

)) In compliance with RG 1.190, two vessel simulation benchmarks are evaluated: the Pool Critical Assembly (PCA) and VENUS-3 experimental benchmarks.

The PCA experimental benchmark includes ((

)) activation measurements at the mid-plane elevation in various simulated reactor components. The VENUS-3 experimental benchmark includes ((

)) activation measurements at a range of elevations in various simulated reactor components. Table 5-1 summarizes the calculated-to-measurement (C/M) results determined for these vessel simulation benchmarks.

Table 5-1 Summary of Comparisons to Vessel Simulation Benchmark Measurements Benchmark Number of Measurements Average Calculated-to-Measured (C/M)

St. Dev. (1)

Pool Critical Assembly

((

((

((

VENUS-3

))

))

))

Total Simulated Vessel Comparisons 413 1.03

+/-0.05 It is noted that, in particular, the VENUS-3 experimental benchmark provides credible evidence that the RAMA Fluence Methodology accurately calculates neutron flux in three-dimensional space.

5.2 Analytic Uncertainty The calculational models used for fluence analyses are comprised of numerous analytical parameters that have associated uncertainties in their values. The uncertainty in these parameters needs to be tested for its contribution to the overall fluence uncertainty.

((

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5.3 Combined Uncertainty The combined uncertainty for the Browns Ferry Unit 2 reactor pressure vessel fluence evaluation is determined using a weighting function ((

)) Table 5-2 shows that the combined uncertainty (1) determined for the Browns Ferry Unit 2 reactor pressure vessel fluence is 9.50% for neutron energy exceeding 1.0 MeV.

Table 5-2 Browns Ferry Unit 2 RPV Combined Uncertainty for Energy > 1.0 MeV Uncertainty Term Value Combined Uncertainty (1) 9.50%

Bias None(1)

1) The bias term is less than its constituent uncertainty values, concluding that no statistically significant bias exists.

It is further shown in Table 5-2 that the combined uncertainty is well below the 20% uncertainty limit specified in RG 1.190. In accordance with RG 1.190, there is no discernable bias in the computed RPV fluence. Therefore, no adjustment to the RPV fast neutron fluence for the period corresponding to Cycle 1 through the end of Cycle 21 is required.

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REFERENCES 6.1 References

1. Browns Ferry Nuclear Plant Fluence Methodology Report, TransWare Enterprises Inc.

Document Number BFN-FLU-001-R-001, Revision 0: 2022

2. U.S. Nuclear Regulatory Commission. Office of Nuclear Regulatory Research. Regulatory Guide 1.190: Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence. Washington, D.C.: Office of Nuclear Regulatory Research: 2001

3. Analysis of the Vessel Wall Neutron Dosimeter from Browns Ferry Unit 2 Pressure Vessel, SwRI, Project 02-4884-002, September 1979.

4. Browns Ferry Steam Electric Station Unit 2 Vessel Surveillance Materials Testing and Fracture Toughness Analysis, General Electric Company, GE-NE-B1100639-01, Revision 1, August 1995.

5. Testing and Evaluation of the Browns Ferry 120 Degree Surveillance Capsule, MP Machinery and Testing LLC Report, MPM-912998, September 2012.

6. BWRVIP-189: BWR Vessel and Internals Project, Evaluation of RAMA Fluence Methodology Calculation Uncertainty, EPRI, Palo Alto, CA: 2008. 1016938.

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FIGURES INCLUDING DIMENSIONAL INFORMATION 7.1 Figures This section contains figures with proprietary dimensional information.

((

))

Figure 7-1 Browns Ferry Unit 2 RPV Beltline Region at 64 EFPY

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, Revision 1 Page 7-2 of 7-3

((

))

Figure 7-2 Identification of the Browns Ferry Unit 2 Core Shroud Welds

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((

))

Figure 7-3 Axial View of the Browns Ferry Unit 2 Jet Pump Model with Weld Identifications

Non-Proprietary BFN-FLU-001-R-001-LNP, Revision 1 Page i of x Topical Report BROWNS FERRY NUCLEAR PLANT FLUENCE METHODOLOGY REPORT Qualification of the Browns Ferry Unit 1 Reactor Fluence Model - Cycles 1 to 13 Document Number:

BFN-FLU-001-R-001-LNP, Revision 1 October 2023 Prepared by:

TransWare Enterprises Inc.

Prepared for:

Tennessee Valley Authority 1101 Market Street Chattanooga, TN 37402 Contract Number:

15904 Project Manager:

John Paul Anglin

((

))

Controlled Copy Number: ____2____

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Non-Proprietary BFN-FLU-001-R-001-LNP, Revision 1 Page iii of x Topical Report BROWNS FERRY NUCLEAR PLANT FLUENCE METHODOLOGY REPORT :

Qualification of the Browns Ferry Unit 1 Reactor Fluence Model - Cycles 1 to 13 Document Number: BFN-FLU-001-R-001-LNP, Revision 1 October 2023 Prepared By:

TransWare Enterprises Inc.

Project Team:

Project Manager:

E. A. Evans, Project Engineer H. J. Heppermann, Project Engineer M. E. Jewell, Project Engineer J. S. Styczynski, Project Engineer S. M. Wagstaff, Project Engineer K. E. Watkins, Project Engineer D.B. Jones 10/25/2023 D. B. Jones, Project Manager Date Reviewed By:

K.E. Watkins 10/25/2023 K. E. Watkins, Project Engineer Date K.A. Jones 10/25/2023 K. A. Jones, QA Specialist Date Approved By:

D.B. Jones 10/25/2023 D. B. Jones, Project Manager Date Prepared For:

Tennessee Valley Authority 1101 Market Street Chattanooga TN, 37402 Contract Number: 15904 Project Manager: John Paul Anglin Official signatures are on file. Contact TransWare Enterprises to request a copy.

TransWare Enterprises Inc.

• 520 Maryville Centre Dr., Suite 125

• St. Louis, Missouri 63141

• USA

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, Revision 1 Page iv of x DISCLAIMER OF WARRANTIES AND LIMITATION OF LIABILITIES THE INFORMATION CONTAINED IN THIS REPORT IS BELIEVED BY TRANSWARE ENTERPRISES INC. TO BE AN ACCURATE AND TRUE REPRESENTATION 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.

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, Revision 1 Page v of x CONTENTS Title Page 1

Introduction................................................................................................................... 1-1 2

Reactor Operating History........................................................................................... 2-1 3

Reactor Statepoint Data............................................................................................... 3-1 4

Surveillance Capsule Dosimetry Evaluation............................................................... 4-1 4.1 Summary of the Flux Wire Activation Analysis........................................................4-1 4.2 Comparison of Predicted Activation to Reactor-specific Measurements..................4-3 4.2.1 Flux Wire Activation Analysis for the Browns Ferry Unit 1 30° Capsule......4-3 5

Reactor Pressure Vessel Fluence Uncertainty Analysis............................................ 5-1 5.1 Comparison Uncertainty.........................................................................................5-1 5.1.1 Operating Reactor Comparison Uncertainty...............................................5-1 5.1.2 Benchmark Comparison Uncertainty.........................................................5-2 5.2 Analytic Uncertainty................................................................................................5-2 5.3 Combined Uncertainty............................................................................................5-3 6

References.................................................................................................................... 6-1 6.1 References.............................................................................................................6-1 7

Figures Including Dimensional Information................................................................ 7-1 7.1 Figures...................................................................................................................7-1

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, Revision 1 Page vii of x LIST OF TABLES Title Page Table 2-1 Summary of Browns Ferry Unit 1 Core Loading Inventory for Cycles 1 to 13........ 2-1 Table 2-2 Summary of Browns Ferry Unit 1 Core Loading Inventory for Cycle 14 and 99prj. 2-2 Table 3-1 Statepoint Data for Browns Ferry Unit 1 per Cycle Basis...................................... 3-2 Table 4-1 Summary of the Fluence and Activity Comparisons for the Browns Ferry Unit 1 Dosimetry............................................................................................................. 4-3 Table 4-2 Summary of the Activity Comparisons for the Copper and Iron Flux Wires Removed From the Browns Ferry Unit 1 Reactor................................................. 4-3 Table 4-3 Comparison of the Calculated-to-Measured Activities for the Iron and Copper Flux Wires Removed From the Browns Ferry Unit 1 30° Flux Wire Holder........... 4-4 Table 5-1 Summary of Comparisons to Vessel Simulation Benchmark Measurements........ 5-2 Table 5-2 Browns Ferry Unit 1 RPV Combined Uncertainty for Energy > 1.0 MeV................ 5-3

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, Revision 1 Page ix of x LIST OF FIGURES Title Page Figure 4-1 Positioning of the Surveillance Capsules Installed in the Browns Ferry Unit 1 Reactor................................................................................................................ 4-2 Figure 7-1 Browns Ferry Unit 1 RPV Beltline Region at 50 EFPY.......................................... 7-1 Figure 7-2 Identification of the Browns Ferry Unit 1 Core Shroud Welds................................ 7-2 Figure 7-3 Axial View of the Browns Ferry Unit 1 Jet Pump Model with Weld Identifications.. 7-3

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INTRODUCTION This attachment provides the reactor operating history, comparisons to activation measurements, and uncertainty analysis that are essential for validating the Browns Ferry Nuclear Plant Unit 1 (Browns Ferry Unit 1) fluence methodology. The methodology that is used for determining the neutron fluence in the Browns Ferry Unit 1 reactor is detailed in the Browns Ferry Nuclear Plant Fluence Methodology Report [1].

The power history data presented in this report covers the time period from start of commercial operation to the end of operating Cycle 13. The reactor began commercial operation in 1973 with a rated thermal power of 3293 MWth. A power uprate occurred at the beginning of Cycle 7 after an extended outage period to increase the rated thermal power to 3458 MWth. The rated thermal power was increased a second time to 3952 MWth at the beginning of Cycle 13. All surveillance dosimetry removed from the reactor over that time period, and which is available in the form of activation measurements, is evaluated. A combined uncertainty factor for the fluence model based on the modeling approach and measurement comparisons is determined which demonstrates that the computational fluence method used by TransWare Enterprises Inc. is qualified for use in determining neutron fluence for the Browns Ferry Unit 1 reactor pressure vessel in accordance with U.S. Nuclear Regulatory Commission (U.S. NRC) Regulatory Guide 1.190 [2].

In compliance with Regulatory Guide 1.190, it is shown in this report that the calculated-to-measured (C/M) ratio and standard deviation is 0.82 +/- 0.06 for all reactor dosimetry evaluated for the Browns Ferry Unit 1 reactor. The combined uncertainty for the Browns Ferry Unit 1 reactor is determined to be 9.79%. Based upon these results, there is no discernable bias in the computed reactor pressure vessel fluence for the period Cycle 1 through the end of Cycle 13 for the Browns Ferry Unit 1 reactor.

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REACTOR OPERATING HISTORY Reactor operating history is the measure of daily reactor power levels that characterize the radiation exposure history of a reactor over its operating life. The daily power history data for the Browns Ferry Unit 1 reactor was provided by TVA in discrete form for Cycles 1 through 13. It is noted that Browns Ferry Unit 1 did not operate for a 22-year period between Cycle 6 and Cycle 7 which has been ((

)) accounted for in the power history.

Another important element of the reactor operating history is the fuel designs that were loaded in the reactor for each operating cycle. ((

))

Table 2-1 and Table 2-2 provide a summary of the fuel designs that were loaded in the Browns Ferry Unit 1 reactor for each operating cycle. Table 2-1 lists the fuel designs that were loaded in the reactor for Cycles 1 through 13. Table 2-2 lists the fuel designs that were loaded for Cycle 14 and the fuel design for the equilibrium projection cycle identified as Cycle 99 (99prj). The dominant (most numerous) fuel design that was loaded in the core for each cycle is shown in bold font. The dominant fuel design that was loaded on the core periphery is identified in blue font.

Table 2-1 Summary of Browns Ferry Unit 1 Core Loading Inventory for Cycles 1 to 13 Cycle Fuel Designs 7x7 8x8 9x9 10x10

((

))

1 168 596 2

7 589 168 3

442 166 156 4

214 163 155 232 5

119 154 491 6

19 4

705 36 7

164 600 8

108 656 9

764 10 484 280 11 177 587 12 484 280 13 152 612

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, Revision 1 Page 2-2 of 2-2 Table 2-2 Summary of Browns Ferry Unit 1 Core Loading Inventory for Cycle 14 and 99prj Cycle Fuel Designs 10x10 11x11

((

))

14(1) 764 99prj(2) 764 (1) For purposes of fluence projections, operation of Cycle 14 is partially historical operation data and partially projected operating data. This cycle was provided by TVA to predict fluence for the duration of one cycle.

(2) For purposes of fluence projections, operation beyond Cycle 14 uses an equilibrium projection cycle featuring a full core loading of ((

)) fuel. This cycle was provided by TVA to predict fluence at the end of the extended plant license period.

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REACTOR STATEPOINT DATA The reactor operating history is defined in discrete exposure steps referred to as statepoints. A statepoint is defined as a snapshot ((

)) of a reactor core at a moment in time. ((

))

Several statepoints are generally used to represent the different operating states of a reactor core over the course of an operating cycle. The core power distribution for a statepoint is generally determined using core simulator software. Because core simulator codes are used for a variety of core analysis functions, 10s to 100s of core calculations may be performed to track and monitor the operation of a reactor. ((

))

Table 3-1 shows that a total of ((

)) statepoints were used to represent the operating states of the Browns Ferry Unit 1 reactor for the first 13 cycles of operating history. The core simulator data was provided by TVA to characterize the historical operating conditions for the Browns Ferry Unit 1 reactor. Partial projection data was also provided for Cycle 14 and for an equilibrium cycle (99prj) for projecting fluence to the reactors end of license period.

Table 3-1 also shows the rated thermal power of the reactor for a cycle and the accumulated effective full power years (EFPY) of exposure accumulated for the cycle.

((

))

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Accumulated EFPY 1

((

3293 1.4 2

3293 2.1 3

3293 2.9 4

3293 3.8 5

3293 5.1 6

3293 6.1 7(1) 3458 7.4 8

3458 9.2 9

3458 11.0 10 3458 12.7 11 3458 14.6 12 3458 16.4 13(2) 3952 18.2 14(3) 3952 20.0 99prj(4)

))

3952 50.0 (1) A power uprate was implemented at the beginning of Cycle 7 from 3293 MWth to 3458 MWth.

(2) A power uprate was implemented at the beginning of Cycle 13 from 3458 MWth to 3952 MWth.

(3) Cycle 14 is a partial projection cycle.

(4) Cycle 99prj is a projection cycle assuming a full core loading of ((

)) fuel. This cycle will be used to project reactor fluence to the end of the reactors licensed period of operation.

EFPY is an effective measure of the number of years that a reactor has operated at full power conditions. EFPY does not account for changes in rated power. Therefore, one EFPY is the same measure irrespective of the rated power of the reactor. Consequently, EFPY should not be used to interpolate between periods of different rated powers.

A separate neutronics transport calculation is performed for each statepoint listed in Table 3-1.

The neutron fluxes calculated for each statepoint are then combined with daily thermal power information to provide an integral accounting of the neutron fluence for the reactor pressure vessel, reactor vessel internals, and surveillance capsules. The periods of reactor shutdown are also accounted for in this process, particularly to allow for an accurate calculation of irradiated surveillance capsule activities.

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SURVEILLANCE CAPSULE DOSIMETRY EVALUATION This section presents the results of the activation analysis and the determination of fast neutron fluence for the Browns Ferry Unit 1 surveillance capsule dosimetry. The results presented in this section form the basis for the validation and qualification of the fluence methodology as applied to the Browns Ferry Unit 1 reactor in accordance with the U.S Nuclear Regulatory Commission Regulatory Guide 1.190 (RG 1.190)[2].

Regulatory Guide 1.190 requires fluence calculational methods to be validated by comparisons with activation measurements from operating reactor dosimetry. It is preferred that the activation data be taken from the reactor being evaluated. ((

)) In the case for the Browns Ferry Unit 1 reactor, there is sufficient reactor-specific measurements available to qualify the calculational method ((

))

In order to report computed fluence as the best-estimate fluence, RG 1.190 requires that the standard deviation resulting from the comparison of calculated to measurement data should be 20%. It is determined that overall calculated-to-measured (C/M) comparison ratio and standard deviation for all Browns Ferry Unit 1 reactor surveillance dosimetry is 0.82 +/- 0.06. Therefore, the computational fluence model for the Browns Ferry Unit 1 reactor meets the RG 1.190 criteria and, as such, no bias adjustment is required to be applied to the computed RPV fluence.

4.1 Summary of the Flux Wire Activation Analysis Three (3) surveillance capsules were initially installed in the Browns Ferry Unit 1 reactor. The capsules are mounted axially near the reactor core mid-plane elevation. The capsules are positioned near the inner surface of the reactor pressure vessel wall at the 30°, 120°, and 300° azimuths around the circumference of the reactor pressure vessel. Figure 4-1 illustrates the positioning of the surveillance capsules in the Browns Ferry Unit 1 reactor.

One (1) set of flux wires were extracted from the flux wire holder attached to the 30° surveillance capsule in the Browns Ferry Unit 1 reactor for activation analysis.

Non-Proprietary BFN-FLU-001-R-001-LNP, Revision 1 Page 4-2 of 4-4 Figure 4-1 Positioning of the Surveillance Capsules Installed in the Browns Ferry Unit 1 Reactor

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, Revision 1 Page 4-3 of 4-4 Table 4-1 provides a summary of the activation comparisons for the flux wires that were irradiated in Browns Ferry Unit 1 reactor. The table also provides the irradiation period in terms of cycle exposure, the accumulated exposure in terms of EFPY of reactor operation, the average fast neutron fluence (E > 1.0. MeV) for each set of flux wires, and the number of specimens (viz., flux wires) evaluated for each set of flux wires removed from the reactor.

In summary, Table 4-1 shows that the overall calculated-to-measured (C/M) ratio and associated standard deviation for the 6 irradiated specimens was determined to be 0.82 +/- 0.06.

Table 4-1 Summary of the Fluence and Activity Comparisons for the Browns Ferry Unit 1 Dosimetry Dosimeter Cycles of Exposure Accumulated Exposure (EFPY)

Fast Neutron Fluence

(>1 MeV, n/cm2)

Number of Specimens Calculated vs.

Measured (C/M)

Standard Deviation

()

30° Flux Wire 1 - 1 1.4 6

((

))

Overall Average C/M and Standard Deviation 6

0.82 0.06 Table 4-2 provides the overall C/M ratios and standard deviations considering only the copper and iron flux wires that were irradiated in the Browns Ferry Unit 1 reactor.

Table 4-2 Summary of the Activity Comparisons for the Copper and Iron Flux Wires Removed From the Browns Ferry Unit 1 Reactor Flux Wire Number of Specimens C/M Copper 3

((

((

Iron 3

))

))

Overall Average C/M and Standard Deviation 6

0.82 0.06 4.2 Comparison of Predicted Activation to Reactor-specific Measurements The comparison of calculated activations to measurements for each capsule are presented in this subsection.

4.2.1 Flux Wire Activation Analysis for the Browns Ferry Unit 1 30° Capsule Copper and iron flux wires were irradiated in the Browns Ferry Unit 1 30° surveillance capsule flux wire holder during the first (1) cycle of reactor operation. At the time of their removal, the flux wires had been irradiated for a total of 1.4 EFPY.

Table 4-3 shows the activation measurements, periods of irradiation, computed activations, and the C/M ratios for the flux wires irradiated in the 30° flux wire holder. Activation calculations were performed for the following reactions: 63Cu (n,) 60Co and 54Fe (n,p) 54Mn.

The average C/M ratio and associated standard deviation determined for all of the flux wires irradiated in the Browns Ferry Unit 1 30° flux wire holder is 0.82 +/- 0.06.

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, Revision 1 Page 4-4 of 4-4 It is noted that the C/M results presented in Table 4-3 ((

)) are acceptable per RG 1.190.

Table 4-3 Comparison of the Calculated-to-Measured Activities for the Iron and Copper Flux Wires Removed From the Browns Ferry Unit 1 30° Flux Wire Holder Flux Wire Dosimeter Measured (dps/mg) [3]

Calculated (dps/mg)

C/M Copper Cu-1 3.70E+00

((

((

Cu-2 3.75E+00 Cu-3 3.61E+00

))

))

Copper Average

((

))

Iron Fe-1 6.10E+01

((

((

Fe-2 5.95E+01 Fe-3 5.50E+01

))

))

Iron Average

((

))

Average C/M and Overall Standard Deviation 0.82 0.06

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REACTOR PRESSURE VESSEL FLUENCE UNCERTAINTY ANALYSIS This section presents the combined uncertainty analysis and the determination of bias for the Browns Ferry Unit 1 reactor pressure vessel (RPV) fluence evaluation. The combined uncertainty is comprised of two components. ((

))

When combined, these components provide a basis for determining the combined uncertainty (1) and bias in the computed RPV fluence.

The requirements for determining the combined uncertainty and bias for light water reactor pressure vessel fluence evaluations are provided in Regulatory Guide 1.190 [2]. The approach for determining combined uncertainty and bias for reactor pressure vessel fluence is demonstrated in Reference 4.

For pressure vessel fluence evaluations, two uncertainty factors are considered: comparison factors and uncertainty introduced by the measurement process. After analysis of these factors, it is determined that the combined uncertainty for Browns Ferry Unit 1 RPV fluence is 9.79%, and that no adjustment for bias is required for the RPV fast neutron fluence determined for the period Cycle 1 through the end of Cycle 13.

5.1 Comparison Uncertainty Comparison uncertainty factors are determined by comparing calculated activities with activity measurements. For pressure vessel fluence evaluations, two comparison uncertainty factors are considered: operating reactor comparison factors and benchmark comparison factors.

5.1.1 Operating Reactor Comparison Uncertainty TransWare has evaluated activation measurements for several BWR plants ranging from BWR/2-class plants to BWR/6-class plants. ((

))

The Browns Ferry Unit 1 reactor is a BWR/4 class design. A total of ((

)) plant-specific dosimetry measurements, which includes Browns Ferry Units 1 and 2, have been evaluated for BWR/4-764 class plants using the RAMA Fluence Methodology. The overall C/M and standard deviation for the BWR/4-764 class plant measurements is determined to be an unbiased 1.03

+/- 0.14. ((

))

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, Revision 1 Page 5-2 of 5-4 A total of over ((

)) credible activation measurement comparisons have been previously performed for a broad spectrum of BWR configurations and designs using the RAMA Fluence Methodology. The overall comparison ratio for all BWR class plants evaluated as of the date of this report is 1.01 +/- 0.10. ((

))

5.1.2 Benchmark Comparison Uncertainty The benchmark comparison uncertainty is based on a set of industry standard simulation benchmark comparisons. ((

)) In compliance with RG 1.190, two vessel simulation benchmarks are evaluated: the Pool Critical Assembly (PCA) and VENUS-3 experimental benchmarks.

The PCA experimental benchmark includes ((

)) activation measurements at the mid-plane elevation in various simulated reactor components. The VENUS-3 experimental benchmark includes ((

)) activation measurements at a range of elevations in various simulated reactor components. Table 5-1 summarizes the calculated-to-measurement (C/M) results determined for these vessel simulation benchmarks.

Table 5-1 Summary of Comparisons to Vessel Simulation Benchmark Measurements Benchmark Number of Measurements Average Calculated-to-Measured (C/M)

St. Dev. (1)

Pool Critical Assembly

((

((

((

VENUS-3

))

))

))

Total Simulated Vessel Comparisons 413 1.03

+/-0.05 It is noted that, in particular, the VENUS-3 experimental benchmark provides credible evidence that the RAMA Fluence Methodology accurately calculates neutron flux in three-dimensional space.

5.2 Analytic Uncertainty The calculational models used for fluence analyses are comprised of numerous analytical parameters that have associated uncertainties in their values. The uncertainty in these parameters needs to be tested for its contribution to the overall fluence uncertainty.

((

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5.3 Combined Uncertainty The combined uncertainty for the Browns Ferry Unit 1 reactor pressure vessel fluence evaluation is determined using a weighting function ((

)) Table 5-2 shows that the combined uncertainty (1) determined for the Browns Ferry Unit 1 reactor pressure vessel fluence is 9.79% for neutron energy exceeding 1.0 MeV.

Table 5-2 Browns Ferry Unit 1 RPV Combined Uncertainty for Energy > 1.0 MeV Uncertainty Term Value Combined Uncertainty (1) 9.79%

Bias None(1)

1) The bias term is less than its constituent uncertainty values, concluding that no statistically significant bias exists.

It is further shown in Table 5-2 that the combined uncertainty is well below the 20% uncertainty limit specified in Regulatory Guide 1.190. In accordance with Regulatory Guide 1.190, there is no discernable bias in the computed RPV fluence. Therefore, no adjustment to the RPV fast neutron fluence for the period corresponding to Cycle 1 through the end of Cycle 13 is required.

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REFERENCES 6.1 References

1. Browns Ferry Nuclear Plant Fluence Methodology Report, TransWare Enterprises Inc.

Document Number BFN-FLU-001-R-001, Revision 0: 2022

2. U.S. Nuclear Regulatory Commission. Office of Nuclear Regulatory Research. Regulatory Guide 1.190: Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence. Washington, D.C.: Office of Nuclear Regulatory Research: 2001

3. Analysis of the Vessel Wall Neutron Dosimeter from Browns Ferry Unit 1 Pressure Vessel, SwRI, Project 02-4884-001, August 1978.

4. BWRVIP-189: BWR Vessel and Internals Project, Evaluation of RAMA Fluence Methodology Calculation Uncertainty, EPRI, Palo Alto, CA: 2008. 1016938.

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FIGURES INCLUDING DIMENSIONAL INFORMATION 7.1 Figures This section contains figures with proprietary dimensional information.

((

))

Figure 7-1 Browns Ferry Unit 1 RPV Beltline Region at 50 EFPY

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((

))

Figure 7-2 Identification of the Browns Ferry Unit 1 Core Shroud Welds

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((

))

Figure 7-3 Axial View of the Browns Ferry Unit 1 Jet Pump Model with Weld Identifications

Non-Proprietary BFN-FLU-001-R-001-LNP, Revision 1 Page i of x Topical Report BROWNS FERRY NUCLEAR PLANT FLUENCE METHODOLOGY REPORT Qualification of the Browns Ferry Unit 3 Reactor Fluence Model - Cycles 1 to 20 Document Number:

BFN-FLU-001-R-001-LNP, Revision 1 October 2023 Prepared by:

TransWare Enterprises Inc.

Prepared for:

Tennessee Valley Authority 1101 Market Street Chattanooga, TN 37402 Contract Number:

15904 Project Manager:

John Paul Anglin

((

))

Controlled Copy Number: ___2_____

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Non-Proprietary BFN-FLU-001-R-001-LNP, Revision 1 Page iii of x Topical Report BROWNS FERRY NUCLEAR PLANT FLUENCE METHODOLOGY REPORT :

Qualification of the Browns Ferry Unit 3 Reactor Fluence Model - Cycles 1 to 20 Document Number: BFN-FLU-001-R-001-LNP, Revision 1 October 2023 Prepared By:

TransWare Enterprises Inc.

Project Team:

Project Manager:

E. A. Evans, Project Engineer H. J. Heppermann, Project Engineer M. E. Jewell, Project Engineer J. S. Styczynski, Project Engineer S. M. Wagstaff, Project Engineer K. E. Watkins, Project Engineer D.B. Jones 10/25/2023 D. B. Jones, Project Manager Date Reviewed By:

K.E. Watkins 10/25/2023 K. E. Watkins, Project Engineer Date K.A. Jones 10/25/2023 K. A. Jones, QA Specialist Date Approved By:

D.B. Jones 10/25/2023 D. B. Jones, Project Manager Date Prepared For:

Tennessee Valley Authority 1101 Market Street Chattanooga TN, 37402 Contract Number: 15904 Project Manager: John Paul Anglin Official signatures are on file. Contact TransWare Enterprises to request a copy.

TransWare Enterprises Inc.

• 520 Maryville Centre Dr., Suite 125

• St. Louis, Missouri 63141

• USA

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, Revision 1 Page iv of x DISCLAIMER OF WARRANTIES AND LIMITATION OF LIABILITIES THE INFORMATION CONTAINED IN THIS REPORT IS BELIEVED BY TRANSWARE ENTERPRISES INC. TO BE AN ACCURATE AND TRUE REPRESENTATION 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.

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, Revision 1 Page v of x CONTENTS Title Page 1

Introduction................................................................................................................... 1-1 2

Reactor Operating History........................................................................................... 2-1 3

Reactor Statepoint Data............................................................................................... 3-1 4

Surveillance Capsule Dosimetry Evaluation............................................................... 4-1 4.1 Summary of the Flux Wire Activation Analysis........................................................4-1 4.2 Comparison of Calculated Activation to Reactor-specific Measurements................4-3 4.2.1 Flux Wire Activation Analysis for the Browns Ferry Unit 3 30° Capsule......4-3 5

Reactor Pressure Vessel Fluence Uncertainty Analysis............................................ 5-1 5.1 Comparison Uncertainty.........................................................................................5-1 5.1.1 Operating Reactor Comparison Uncertainty...............................................5-1 5.1.2 Benchmark Comparison Uncertainty.........................................................5-2 5.2 Analytic Uncertainty................................................................................................5-2 5.3 Combined Uncertainty............................................................................................5-3 6

References.................................................................................................................... 6-1 6.1 References.............................................................................................................6-1 7

Figures Including Dimensional Information................................................................ 7-1 7.1 Figures...................................................................................................................7-1

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, Revision 1 Page vii of x LIST OF TABLES Title Page Table 2-1 Summary of Browns Ferry Unit 3 Core Loading Inventory for Cycles 1 to 11........ 2-1 Table 2-2 Summary of Browns Ferry Unit 3 Core Loading Inventory for Cycles 12 to 20 and 99prj.............................................................................................................. 2-2 Table 3-1 Statepoint Data for Browns Ferry Unit 3 per Cycle Basis...................................... 3-2 Table 4-1 Summary of the Fluence and Activity Comparisons for the Browns Ferry Unit 3 Dosimetry............................................................................................................. 4-3 Table 4-2 Summary of the Activity Comparisons for the Copper and Iron Flux Wires Removed From the Browns Ferry Unit 3 Reactor................................................. 4-3 Table 4-3 Comparison of the Calculated-to-Measured Activities for the Iron and Copper Flux Wires Removed From the Browns Ferry Unit 3 30° Flux Wire Holder........... 4-4 Table 5-1 Summary of Comparisons to Vessel Simulation Benchmark Measurements........ 5-2 Table 5-2 Browns Ferry Unit 3 RPV Combined Uncertainty for Energy > 1.0 MeV................ 5-3

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, Revision 1 Page ix of x LIST OF FIGURES Title Page Figure 4-1 Positioning of the Surveillance Capsules Installed in the Browns Ferry Unit 3 Reactor................................................................................................................ 4-2 Figure 7-1 Browns Ferry Unit 3 RPV Beltline Region at 62 EFPY.......................................... 7-1 Figure 7-2 Identification of the Browns Ferry Unit 3 Core Shroud Welds................................ 7-2 Figure 7-3 Axial View of the Browns Ferry Unit 3 Jet Pump Model with Weld Identifications.. 7-3

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INTRODUCTION This attachment provides the reactor operating history, comparisons to activation measurements, and uncertainty analysis that are essential for validating the Browns Ferry Nuclear Plant Unit 3 (Browns Ferry Unit 3) fluence methodology. The methodology that is used for determining the neutron fluence in the Browns Ferry Unit 3 reactor is detailed in the Browns Ferry Nuclear Plant Fluence Methodology Report [1].

The power history data presented in this report covers the time period from start of commercial operation to the end of operating Cycle 20. The reactor began commercial operation in 1976 with a rated thermal power of 3293 MWth. A power uprate occurred at the beginning of Cycle 9 to increase the rated thermal power to 3458 MWth. The rated thermal power was increased a second time to 3952 MWth at the beginning of Cycle 19. All surveillance dosimetry removed from the reactor over that time period, and which is available in the form of activation measurements, is evaluated. A combined uncertainty factor for the fluence model based on the modeling approach and measurement comparisons is determined which demonstrates that the computational fluence method used by TransWare Enterprises Inc. is qualified for use in determining neutron fluence for the Browns Ferry Unit 3 reactor pressure vessel in accordance with U.S. Nuclear Regulatory Commission (U.S. NRC) Regulatory Guide 1.190 [2].

In compliance with Regulatory Guide 1.190, it is shown in this report that the calculated-to-measured (C/M) ratio and standard deviation is 0.90 +/- 0.09 for all reactor surveillance capsule dosimetry evaluated for the Browns Ferry Unit 3 reactor. The combined uncertainty for the Browns Ferry Unit 3 reactor is determined to be 9.20%. Based upon these results, there is no discernable bias in the computed reactor pressure vessel fluence for the period Cycle 1 through the end of Cycle 20 for the Browns Ferry Unit 3 reactor.

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REACTOR OPERATING HISTORY Reactor operating history is the measure of daily reactor power levels that characterize the radiation exposure history of a reactor over its operating life. The daily power history data for the Browns Ferry Unit 3 reactor was provided by TVA in discrete form for Cycles 1 through 20.

Another important element of the reactor operating history is the fuel designs that were loaded in the reactor for each operating cycle. ((

))

Table 2-1 and Table 2-2 provide a summary of the fuel designs that were loaded in the Browns Ferry Unit 3 reactor for each operating cycle. Table 2-1 lists the fuel designs that were loaded in the reactor for Cycles 1 through 11. Table 2-2 lists the fuel designs that were loaded for Cycles 12 through 20 and the fuel design for the equilibrium projection cycle identified as Cycle 99 (99prj). The dominant (most numerous) fuel design that was loaded in the core for each cycle is shown in bold font. The dominant fuel design that was loaded on the core periphery is identified in blue font.

Table 2-1 Summary of Browns Ferry Unit 3 Core Loading Inventory for Cycles 1 to 11 Cycle Fuel Designs 8x8 9x9 10x10

((

))

1 764 2

556 208 3

412 208 144 4

288 208 268 5

8 208 548 6

24 492 248 7

484 248 32 8A(1) 75 397 292 8B(1) 47 425 264 9

180 292 292 10 185 579 11A 480 284 11B 481 283 (1) Due to mislabeled data transmitted by TVA, the amounts of ((

)) for Cycles 8A and 8B are incorrect. The amounts shown in the table reflect how the fuel was modeled for the transport calculations.

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, Revision 1 Page 2-2 of 2-2 Table 2-2 Summary of Browns Ferry Unit 3 Core Loading Inventory for Cycles 12 to 20 and 99prj Cycle Fuel Designs 10x10 11x11

((

))

12 181 283 300 13 169 595 14 764 15A 764 15B 764 16 764 17 764 18 460 304 19 116 648 20 764 99prj(1) 764 (1) For purposes of fluence projections, operation beyond projection Cycle 20 uses an equilibrium projection cycle featuring a full core loading of ((

)) fuel. This cycle was provided by TVA to predict fluence at the end of the extended plant license period.

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REACTOR STATEPOINT DATA The reactor operating history is defined in discrete exposure steps referred to as statepoints. A statepoint is defined as a snapshot ((

)) of a reactor core at a moment in time. ((

))

Several statepoints are generally used to represent the different operating states of a reactor core over the course of an operating cycle. The core power distribution for a statepoint is generally determined using core simulator software. Because core simulator codes are used for a variety of core analysis functions, 10s to 100s of core calculations may be performed to track and monitor the operation of a reactor. ((

))

Table 3-1 shows that a total of ((

)) statepoints were used to represent the operating states of the Browns Ferry Unit 3 reactor for the first 20 cycles of operating history. The core simulator data was provided by TVA to characterize the historical operating conditions for the Browns Ferry Unit 3 reactor. Projection data was also provided for an equilibrium cycle (99prj) for projecting fluence to the reactors end of license period. Table 3-1 also shows the rated thermal power of the reactor for a cycle and the accumulated effective full power years (EFPY) of exposure accumulated for the cycle.

((

))

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, Revision 1 Page 3-2 of 3-2 Table 3-1 Statepoint Data for Browns Ferry Unit 3 per Cycle Basis Cycle Number Number of Reactor Statepoints Rated Thermal Power (MWt)

Accumulated EFPY 1

((

3293 1.5 2

3293 2.1 3

3293 2.9 4

3293 3.6 5

3293 4.7 6

3293 4.9 7

3293 6.0 8A 3293 7.1 8B 3293 7.4 9(1) 3458 8.9 10 3458 10.8 11 3458 12.6 12 3458 14.4 13 3458 16.3 14 3458 18.0 15 3458 19.8 16 3458 21.4 17 3458 23.2 18 3458 25.1 19(2) 3952 26.8 20 3952 28.7 99prj(3)

))

3952 62.0 (1) A power uprate was implemented at the beginning of Cycle 9 from 3293 MWth to 3458 MWth.

(2) A power uprate was implemented at the beginning of Cycle 19 from 3458 MWth to 3952 MWth.

(3) Cycle 99prj is a projection cycle assuming a full core loading of ((

)) fuel. This cycle will be used to project reactor fluence to the end of the reactors licensed period of operation.

EFPY is an effective measurement of the number of years that a reactor has operated at full power conditions but does not account for the changes in rated power. Therefore, one EFPY is the same measured value irrespective of the rated power of the reactor. Consequently, EFPY should not be used to interpolate between periods of different rated powers.

A separate neutronics transport calculation is performed for each statepoint listed in Table 3-1.

The neutron fluxes calculated for each statepoint are then combined with daily thermal power information to provide an integral accounting of the neutron fluence for the reactor pressure vessel, reactor vessel internals, and surveillance capsules. The periods of reactor shutdown are also accounted for in this process, particularly to allow for an accurate calculation of irradiated surveillance capsules activities.

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SURVEILLANCE CAPSULE DOSIMETRY EVALUATION This section presents the results of the activation analysis and the determination of fast neutron fluence for the Browns Ferry Unit 3 surveillance capsule dosimetry. The results presented in this section form the basis for the validation and qualification of the fluence methodology as applied to the Browns Ferry Unit 3 reactor in accordance with the U.S Nuclear Regulatory Commission Regulatory Guide 1.190 (RG 1.190) [2].

Regulatory Guide 1.190 requires fluence calculational methods to be validated by comparisons with activation measurements from operating reactor dosimetry. It is preferred that the activation data be taken from the reactor being evaluated. ((

)) In the case for the Browns Ferry Unit 3 reactor, there is sufficient reactor-specific measurements available to qualify the calculational method ((

))

In order to report computed fluence as the best-estimate fluence, RG 1.190 requires that the standard deviation resulting from the comparison of calculated to measurement data should be 20%. It is determined that the overall calculated-to-measured (C/M) comparison ratio and standard deviation for all Browns Ferry Unit 3 reactor surveillance dosimetry is 0.90 +/- 0.09.

Therefore, the computational fluence model for the Browns Ferry Unit 3 reactor meets the RG 1.190 criteria and, as such, no bias adjustment is required to be applied to the computed RPV fluence.

4.1 Summary of the Flux Wire Activation Analysis Three (3) surveillance capsules were initially installed in the Browns Ferry Unit 3 reactor. The capsules are mounted axially near the reactor core mid-plane elevation. The capsules are positioned near the inner surface of the reactor pressure vessel wall at the 30°, 120°, and 300° azimuths around the circumference of the reactor pressure vessel. Figure 4-1 illustrates the positioning of the surveillance capsules in the Browns Ferry Unit 3 reactor.

One (1) set of flux wires were extracted from the flux wire holder attached to the 30° surveillance capsule in the Browns Ferry Unit 3 reactor for activation analysis.

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, Revision 1 Page 4-3 of 4-4 Table 4-1 provides a summary of the activation comparisons for each set of flux wire that were irradiated in the Browns Ferry Unit 3 reactor. The table also provides the irradiation period in terms of cycle exposure, the accumulated exposure in terms of EFPY of reactor operation, the average fast neutron fluence (E > 1.0 MeV) for each set of flux wires, and the number of specimens (viz., flux wires) evaluated for each set of flux wires removed from the reactor.

In summary, Table 4-1 shows that the overall C/M ratio and associated standard deviation for the 6 irradiated specimens was determined to be 0.90 +/- 0.09.

Table 4-1 Summary of the Fluence and Activity Comparisons for the Browns Ferry Unit 3 Dosimetry Dosimeter Cycles of Exposure Accumulated Exposure (EFPY)

Fast Neutron Fluence

(>1 MeV, n/cm2)

Number of Specimens Calculated vs.

Measured (C/M)

Standard Deviation

()

30° Flux Wire 1 - 1 1.5 6

((

))

Overall Average C/M and Standard Deviation 6

0.90 0.09 Table 4-2 provides the overall C/M ratios and standard deviations considering only the copper and iron flux wires that were irradiated in the Browns Ferry Unit 3 reactor.

Table 4-2 Summary of the Activity Comparisons for the Copper and Iron Flux Wires Removed From the Browns Ferry Unit 3 Reactor Flux Wire Number of Specimens C/M Copper 3

((

((

Iron 3

))

))

Overall Average C/M and Standard Deviation 6

0.90 0.09 4.2 Comparison of Calculated Activation to Reactor-specific Measurements The comparison of calculated activations to measurements for each capsule are presented in this subsection.

4.2.1 Flux Wire Activation Analysis for the Browns Ferry Unit 3 30° Capsule Copper and iron flux wires were irradiated in the Browns Ferry Unit 3 30° surveillance capsule flux wire holder during the first cycle of reactor operation. At the time of their removal, the flux wires had been irradiated for a total of 1.5 EFPY.

Table 4-3 shows the activation measurements, periods of irradiation, computed activations, and the computed-to-measured (C/M) ratios for the flux wires irradiated in the 30° flux wire holder.

Activation calculations were performed for the following reactions: 63Cu (n,) 60Co and 54Fe (n,p) 54Mn.

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, Revision 1 Page 4-4 of 4-4 The average C/M ratio and associated standard deviation determined for all of the flux wires irradiated in the Browns Ferry Unit 3 30° flux wire holder is 0.90 +/- 0.09.

It is noted that the C/M results presented in Table 4-3 ((

)) are acceptable per RG 1.190.

Table 4-3 Comparison of the Calculated-to-Measured Activities for the Iron and Copper Flux Wires Removed From the Browns Ferry Unit 3 30° Flux Wire Holder Flux Wire Dosimeter Measured (dps/mg) [3]

Calculated (dps/mg)

C/M Copper Cu-1 3.79E+00

((

((

Cu-2 3.73E+00 Cu-3 3.78E+00

))

))

Copper Average

((

))

Iron Fe-1 6.61E+01

((

((

Fe-2 6.62E+01 Fe-3 6.74E+01

))

))

Iron Average

((

))

Average C/M and Overall Standard Deviation 0.90 0.09

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REACTOR PRESSURE VESSEL FLUENCE UNCERTAINTY ANALYSIS This section presents the combined uncertainty analysis and the determination of bias for the Browns Ferry Unit 3 reactor pressure vessel (RPV) fluence evaluation. The combined uncertainty is comprised of two components. ((

))

When combined, these components provide a basis for determining the combined uncertainty (1) and bias in the computed RPV fluence.

The requirements for determining the combined uncertainty and bias for light water reactor pressure vessel fluence evaluations are provided in Regulatory Guide 1.190 [2]. The approach for determining combined uncertainty and bias for reactor pressure vessel fluence is demonstrated in Reference 4.

For pressure vessel fluence evaluations, two uncertainty factors are considered: comparison factors and uncertainty introduced by the measurement process. After analysis of these factors, it is determined that the combined uncertainty for Browns Ferry Unit 3 RPV fluence is 9.20%, and that no adjustment for bias is required for the RPV fast neutron fluence determined for the period Cycle 1 through the end of Cycle 20.

5.1 Comparison Uncertainty Comparison uncertainty factors are determined by comparing calculated activities with activity measurements. For pressure vessel fluence evaluations, two comparison uncertainty factors are considered: operating reactor comparison factors and benchmark comparison factors.

5.1.1 Operating Reactor Comparison Uncertainty TransWare has evaluated activation measurements for several BWR plants ranging from BWR/2-class plants to BWR/6-class plants. ((

))

The Browns Ferry Unit 3 reactor is a BWR/4 class design. A total of ((

)) plant-specific dosimetry measurements, including Browns Ferry Units 1, 2, and 3, have been evaluated for BWR/4-764 class plants using the RAMA Fluence Methodology. The overall C/M and standard deviation for the BWR/4-764 class plant measurements is determined to be an unbiased 1.02 +/- 0.14. ((

))

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, Revision 1 Page 5-2 of 5-4 A total of over ((

)) credible activation measurement comparisons have been previously performed for a broad spectrum of BWR configurations and designs using the RAMA Fluence Methodology. The overall comparison ratio for all BWR class plants evaluated as of the date of this report is 1.00 +/- 0.10. ((

))

5.1.2 Benchmark Comparison Uncertainty The benchmark comparison uncertainty is based on a set of industry standard simulation benchmark comparisons. ((

)) In compliance with RG 1.190, two vessel simulation benchmarks are evaluated: the Pool Critical Assembly (PCA) and VENUS-3 experimental benchmarks.

The PCA experimental benchmark includes ((

)) activation measurements at the mid-plane elevation in various simulated reactor components. The VENUS-3 experimental benchmark includes ((

)) activation measurements at a range of elevations in various simulated reactor components. Table 5-1 summarizes the calculated-to-measurement (C/M) results determined for these vessel simulation benchmarks.

Table 5-1 Summary of Comparisons to Vessel Simulation Benchmark Measurements Benchmark Number of Measurements Average Calculated-to-Measured (C/M)

St. Dev. (1)

Pool Critical Assembly

((

((

((

VENUS-3

))

))

))

Total Simulated Vessel Comparisons 413 1.03

+/-0.05 It is noted that, in particular, the VENUS-3 experimental benchmark provides credible evidence that the RAMA Fluence Methodology accurately calculates neutron flux in three-dimensional space.

5.2 Analytic Uncertainty The calculational models used for fluence analyses are comprised of numerous analytical parameters that have associated uncertainties in their values. The uncertainty in these parameters needs to be tested for its contribution to the overall fluence uncertainty.

((

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))

5.3 Combined Uncertainty The combined uncertainty for the Browns Ferry Unit 3 reactor pressure vessel fluence evaluation is determined using a weighting function ((

)) Table 5-2 shows that the combined uncertainty (1) determined for the Browns Ferry Unit 3 reactor pressure vessel fluence is 9.20% for neutron energy exceeding 1.0 MeV.

Table 5-2 Browns Ferry Unit 3 RPV Combined Uncertainty for Energy > 1.0 MeV Uncertainty Term Value Combined Uncertainty (1) 9.20%

Bias None(1)

1) The bias term is less than its constituent uncertainty values, concluding that no statistically significant bias exists.

It is further shown in Table 5-2 that the combined uncertainty is well below the 20% uncertainty limit specified in RG 1.190. In accordance with RG 1.190, there is no discernable bias in the computed RPV fluence. Therefore, no adjustment to the RPV fast neutron fluence for the period corresponding to Cycle 1 through the end of Cycle 20 is required.

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REFERENCES 6.1 References

1. Browns Ferry Nuclear Plant Fluence Methodology Report, TransWare Enterprises Inc.

Document Number BFN-FLU-001-R-001, Revision 0: 2022

2. U.S. Nuclear Regulatory Commission. Office of Nuclear Regulatory Research. Regulatory Guide 1.190: Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence. Washington, D.C.: Office of Nuclear Regulatory Research: 2001

3. Analysis of the Vessel Wall Neutron Dosimeter from Browns Ferry Unit 3 Pressure Vessel, SwRI, Project 02-4884-003, September 1979.

4. BWRVIP-189: BWR Vessel and Internals Project, Evaluation of RAMA Fluence Methodology Calculation Uncertainty, EPRI, Palo Alto, CA: 2008. 1016938.

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FIGURES INCLUDING DIMENSIONAL INFORMATION 7.1 Figures This section contains figures with proprietary dimensional information.

((

))

Figure 7-1 Browns Ferry Unit 3 RPV Beltline Region at 62 EFPY

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, Revision 1 Page 7-2 of 7-3

((

))

Figure 7-2 Identification of the Browns Ferry Unit 3 Core Shroud Welds

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, Revision 1 Page 7-3 of 7-3

((

))

Figure 7-3 Axial View of the Browns Ferry Unit 3 Jet Pump Model with Weld Identifications

Proprietary Information - Withhold Under 10 CFR § 2.390 Proprietary Information - Withhold Under 10 CFR § 2.390 ENCLOSURE 2 BFN Fluence Methodology Report and Attachments 1, 2, and 3 (BFN-FLU-001-R-001-LP, Revision 1; BFN-FLU-001-R-001-LP, Attachment 1, Revision 1; BFN-FLU-001-R-001-LP, Attachment 2, Revision 1; and BFN-FLU-001-R-001-LP, Attachment 3, Revision 1)

(proprietary version)

ENCLOSURE 3 Affidavits

Affidavit I, Kathleen A. Jones, state as follows:

1. I am the Chief Operating Officer of TransWare Enterprises Inc. (TWE) 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 the attachment TransWare Enterprises Inc. Document No. BFN-FLU-001-R-001-LP, Revision 1, Browns Ferry Nuclear Plant Fluence Methodology Report, October 2023. TWE proprietary information is indicated by enclosing it in double brackets and highlighting the proprietary text in blue. Paragraph 3 of this affidavit 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, TWE relies upon the exemption of disclosure set forth in the Freedom of Information Act (FOIA), 5 USC Sec. 552(b)(4), and the Trade Secrets Act, 18 USC Sec. 1905, and the NRC regulations 10CFR9.17(a)(4) and 2.390(a)(4) for "trade secrets and commercial or financial information obtained from a person and privileged or confidential (Exemption 4). The material for which exemption from disclosure is here sought is all confidential and commercial information, and some portions also qualify 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 Project v. Nuclear Regulatory Commission, 975F2d871 (DC Cir. 1992), and Public Citizen Health Research Group v. FDA, 704F2d1280 (DC Cir. 1983).

4. 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 TWE's competitors without license from TWE constitutes a competitive economic advantage over other companies.

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

c. Information that reveals cost or price information, production capacities, budget levels, or commercial strategies of TWE, its customers, or its suppliers.

d. Information which reveals aspects of past, present, or future TWE customer-funded development plans and programs of potential commercial value to TWE.

e. Information which discloses patentable subject matter for which it may be desirable to obtain patent protection.

The information sought to be withheld is considered to be proprietary for the reasons set forth in paragraphs 4a. and 4b., above.

5. To address 10CFR2.390 (b)(4), the information sought to be withheld is being submitted to the NRC in confidence. The information is of a sort customarily held in confidence by TWE and is in fact so held. Its initial designation as proprietary information, and the subsequent steps taken to prevent its unauthorized disclosure, are as set forth in paragraphs 6 and 7 following. The information sought to be withheld has, to the best of my knowledge and belief, consistently been held in confidence by TWE, no public disclosure has been made, and it is not available to 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 agreements which provide for maintenance of the information in confidence.

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

7. The procedure for approval of external release of such a document typically requires review by the project manager, principal engineer, and by the Quality Assurance department for technical content, competitive effect, and the determination of the accuracy of the proprietary designation. Disclosures outside TWE 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 agreements.

8. The information identified in paragraph 2 is classified as proprietary because it contains details of TWEs methodologies for fluence and uncertainty analyses.

The development of the methods used in these analyses, along with the testing, development, and approval of the supporting methodology was achieved at a significant cost, on the order of several million dollars, to TWE or its licensor.

9. Public disclosure of the information sought to be withheld is likely to cause substantial harm to TWEs competitive position and foreclose or reduce the availability of profit-making opportunities. The methodologies for fluence and uncertainty analyses are part of TWEs nuclear engineering consulting base expertise 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 TWE or its licensor.

Affidavit I, Kathleen A. Jones, state as follows:

1. I am the Chief Operating Officer of TransWare Enterprises Inc. (TWE) 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 the attachment TransWare Enterprises Inc. Document No. BFN-FLU-001-R-001-LP, Attachment 1, Revision 1, Qualification of the Browns Ferry Unit 2 Reactor Fluence Model - Cycles 1 to 21, October 2023. TWE proprietary information is indicated by enclosing it in double brackets and highlighting the proprietary text in blue. Paragraph 3 of this affidavit 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, TWE relies upon the exemption of disclosure set forth in the Freedom of Information Act (FOIA), 5 USC Sec. 552(b)(4), and the Trade Secrets Act, 18 USC Sec. 1905, and the NRC regulations 10CFR9.17(a)(4) and 2.390(a)(4) for "trade secrets and commercial or financial information obtained from a person and privileged or confidential (Exemption 4). The material for which exemption from disclosure is here sought is all confidential and commercial information, and some portions also qualify 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 Project v. Nuclear Regulatory Commission, 975F2d871 (DC Cir. 1992), and Public Citizen Health Research Group v. FDA, 704F2d1280 (DC Cir. 1983).

4. 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 TWE's competitors without license from TWE constitutes a competitive economic advantage over other companies.

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

c. Information that reveals cost or price information, production capacities, budget levels, or commercial strategies of TWE, its customers, or its suppliers.

d. Information which reveals aspects of past, present, or future TWE customer-funded development plans and programs of potential commercial value to TWE.

e. Information which discloses patentable subject matter for which it may be desirable to obtain patent protection.

The information sought to be withheld is considered to be proprietary for the reasons set forth in paragraphs 4a. and 4b., above.

5. To address 10CFR2.390 (b)(4), the information sought to be withheld is being submitted to the NRC in confidence. The information is of a sort customarily held in confidence by TWE and is in fact so held. Its initial designation as proprietary information, and the subsequent steps taken to prevent its unauthorized disclosure, are as set forth in paragraphs 6 and 7 following. The information sought to be withheld has, to the best of my knowledge and belief, consistently been held in confidence by TWE, no public disclosure has been made, and it is not available to 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 agreements which provide for maintenance of the information in confidence.

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

7. The procedure for approval of external release of such a document typically requires review by the project manager, principal engineer, and by the Quality Assurance department for technical content, competitive effect, and the determination of the accuracy of the proprietary designation. Disclosures outside TWE 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 agreements.

8. The information identified in paragraph 2 is classified as proprietary because it contains details of TWEs methodologies for fluence and uncertainty analyses.

The development of the methods used in these analyses, along with the testing, development, and approval of the supporting methodology was achieved at a significant cost, on the order of several million dollars, to TWE or its licensor.

9. Public disclosure of the information sought to be withheld is likely to cause substantial harm to TWEs competitive position and foreclose or reduce the availability of profit-making opportunities. The methodologies for fluence and uncertainty analyses are part of TWEs nuclear engineering consulting base expertise 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 TWE or its licensor.

Affidavit I, Kathleen A. Jones, state as follows:

1. I am the Chief Operating Officer of TransWare Enterprises Inc. (TWE) 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 the attachment TransWare Enterprises Inc. Document No. BFN-FLU-001-R-001-LP, Attachment 2, Revision 1, Qualification of the Browns Ferry Unit 1 Reactor Fluence Model - Cycles 1 to 13, October 2023. TWE proprietary information is indicated by enclosing it in double brackets and highlighting the proprietary text in blue. Paragraph 3 of this affidavit 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, TWE relies upon the exemption of disclosure set forth in the Freedom of Information Act (FOIA), 5 USC Sec. 552(b)(4), and the Trade Secrets Act, 18 USC Sec. 1905, and the NRC regulations 10CFR9.17(a)(4) and 2.390(a)(4) for "trade secrets and commercial or financial information obtained from a person and privileged or confidential (Exemption 4). The material for which exemption from disclosure is here sought is all confidential and commercial information, and some portions also qualify 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 Project v. Nuclear Regulatory Commission, 975F2d871 (DC Cir. 1992), and Public Citizen Health Research Group v. FDA, 704F2d1280 (DC Cir. 1983).

4. 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 TWE's competitors without license from TWE constitutes a competitive economic advantage over other companies.

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

c. Information that reveals cost or price information, production capacities, budget levels, or commercial strategies of TWE, its customers, or its suppliers.

d. Information which reveals aspects of past, present, or future TWE customer-funded development plans and programs of potential commercial value to TWE.

e. Information which discloses patentable subject matter for which it may be desirable to obtain patent protection.

The information sought to be withheld is considered to be proprietary for the reasons set forth in paragraphs 4a. and 4b., above.

5. To address 10CFR2.390 (b)(4), the information sought to be withheld is being submitted to the NRC in confidence. The information is of a sort customarily held in confidence by TWE and is in fact so held. Its initial designation as proprietary information, and the subsequent steps taken to prevent its unauthorized disclosure, are as set forth in paragraphs 6 and 7 following. The information sought to be withheld has, to the best of my knowledge and belief, consistently been held in confidence by TWE, no public disclosure has been made, and it is not available to 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 agreements which provide for maintenance of the information in confidence.

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

7. The procedure for approval of external release of such a document typically requires review by the project manager, principal engineer, and by the Quality Assurance department for technical content, competitive effect, and the determination of the accuracy of the proprietary designation. Disclosures outside TWE 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 agreements.

8. The information identified in paragraph 2 is classified as proprietary because it contains details of TWEs methodologies for fluence and uncertainty analyses.

The development of the methods used in these analyses, along with the testing, development, and approval of the supporting methodology was achieved at a significant cost, on the order of several million dollars, to TWE or its licensor.

9. Public disclosure of the information sought to be withheld is likely to cause substantial harm to TWEs competitive position and foreclose or reduce the availability of profit-making opportunities. The methodologies for fluence and uncertainty analyses are part of TWEs nuclear engineering consulting base expertise 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 TWE or its licensor.

Affidavit I, Kathleen A. Jones, state as follows:

1. I am the Chief Operating Officer of TransWare Enterprises Inc. (TWE) 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 the attachment TransWare Enterprises Inc. Document No. BFN-FLU-001-R-001-LP, Attachment 3, Revision 1, Qualification of the Browns Ferry Unit 3 Reactor Fluence Model - Cycles 1 to 20, October 2023. TWE proprietary information is indicated by enclosing it in double brackets and highlighting the proprietary text in blue. Paragraph 3 of this affidavit 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, TWE relies upon the exemption of disclosure set forth in the Freedom of Information Act (FOIA), 5 USC Sec. 552(b)(4), and the Trade Secrets Act, 18 USC Sec. 1905, and the NRC regulations 10CFR9.17(a)(4) and 2.390(a)(4) for "trade secrets and commercial or financial information obtained from a person and privileged or confidential (Exemption 4). The material for which exemption from disclosure is here sought is all confidential and commercial information, and some portions also qualify 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 Project v. Nuclear Regulatory Commission, 975F2d871 (DC Cir. 1992), and Public Citizen Health Research Group v. FDA, 704F2d1280 (DC Cir. 1983).

4. 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 TWE's competitors without license from TWE constitutes a competitive economic advantage over other companies.

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

c. Information that reveals cost or price information, production capacities, budget levels, or commercial strategies of TWE, its customers, or its suppliers.

d. Information which reveals aspects of past, present, or future TWE customer-funded development plans and programs of potential commercial value to TWE.

e. Information which discloses patentable subject matter for which it may be desirable to obtain patent protection.

The information sought to be withheld is considered to be proprietary for the reasons set forth in paragraphs 4a. and 4b., above.

5. To address 10CFR2.390 (b)(4), the information sought to be withheld is being submitted to the NRC in confidence. The information is of a sort customarily held in confidence by TWE and is in fact so held. Its initial designation as proprietary information, and the subsequent steps taken to prevent its unauthorized disclosure, are as set forth in paragraphs 6 and 7 following. The information sought to be withheld has, to the best of my knowledge and belief, consistently been held in confidence by TWE, no public disclosure has been made, and it is not available to 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 agreements which provide for maintenance of the information in confidence.

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

7. The procedure for approval of external release of such a document typically requires review by the project manager, principal engineer, and by the Quality Assurance department for technical content, competitive effect, and the determination of the accuracy of the proprietary designation. Disclosures outside TWE 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 agreements.

8. The information identified in paragraph 2 is classified as proprietary because it contains details of TWEs methodologies for fluence and uncertainty analyses.

The development of the methods used in these analyses, along with the testing, development, and approval of the supporting methodology was achieved at a significant cost, on the order of several million dollars, to TWE or its licensor.

9. Public disclosure of the information sought to be withheld is likely to cause substantial harm to TWEs competitive position and foreclose or reduce the availability of profit-making opportunities. The methodologies for fluence and uncertainty analyses are part of TWEs nuclear engineering consulting base expertise 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 TWE or its licensor.