L-MT-23-030, Subsequent License Renewal Application Supplement 3

Subsequent License Renewal Application Supplement 3

ML23193B026
Person / Time
Site:	Monticello
Issue date:	07/04/2023
From:	Hafen S Northern States Power Company, Minnesota, Xcel Energy
To:	Office of Nuclear Reactor Regulation, Document Control Desk
Shared Package
ML23193B025	List: L-MT-23-030, Subsequent License Renewal Application Supplement 3
References
L-MT-23-030 MNT-FLU-001-R-001-LNP
Download: ML23193B026 (1)

Text

Proprietary Information Withhold Under 10 CFR 2.390 fl Xcel Energy 2807 West County Road 75 Monticello, MN 55362 July 11, 2023 L-MT-23-030 10 CFR 54.17 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 Monticello Nuclear Generating Plant Docket No. 50-263 Renewed Facility Operating License No. DPR-22 Subsequent License Renewal Application Supplement 3

References:

1) Letter from Northern States Power Company, a Minnesota corporation (NSPM), d/b/a Xcel Energy to Document Control Desk, Monticello Nuclear Generating Plant Docket No. 50-263, Renewal License Number DPR-22 Application for Subsequent Renewal Operating License dated January 9, 2023, ML23009A353.

2) Letter from Northern States Power Company, a Minnesota corporation (NSPM), d/b/a Xcel Energy to Document Control Desk, Monticello Nuclear Generating Plant Subsequent License Renewal Application Supplement 1 dated April 3, 2023, ML23094A136.

3) Letter from Northern States Power Company, a Minnesota corporation (NSPM), d/b/a Xcel Energy to Document Control Desk, Monticello Nuclear Generating Plant Subsequent License Renewal Application Supplement 2 dated June 26, 2023, ML23177A218.

Northern States Power Company, a Minnesota corporation, doing business as Xcel Energy hereafter "NSPM", is submitting a supplement to the Subsequent License Renewal Application (SLRA), listed in Reference 1.

Clarifying information regarding Tables 4.2.3-1 and 4.2.3-2 and an updated reference was provided in Supplement 1, listed in Reference 2. Clarifications to sections of the SLRA discussed in the breakout audits occurring April through June of 2023 were provided in Reference 3. This supplement includes the methodology utilized for the SLRA Section 4.2.1, Neutron Fluence Projections. This supplement does not change any information in the SLRA.

This letter is decontrolled when separated from Enclosure 3

Proprietary Information Withhold Under 10 CFR 2.390 Document Control Desk L-MT-23-030 Page 2 NSPM requests that Enclosure 3 be withheld from public disclosure pursuant to 10 CFR 2.390(a)(4) . The analysis and methodologies used in developing the reports are of commercial value and the competitive position of the owner, TransWare Enterprises Inc. (TWE) , would be harmed if disclosed . The affidavits for the TWE reports are in Enclosure 2 and the non-proprietary version of the TWE report are in Enclosure 1.

Summary of Commitments This letter makes no new commitments and no revisions to existing commitments .

I declare under penalty of perjury that the foregoing is true and correct.

Executed on July lL 2023 Sh1 Site Vice Pres* ent, Monticello Nuclear Generating Plant Northern St es Power Company - Minnesota Enclosures (3) cc: Administrator, Region 111 , USNRC Project Manager, Monticello, USNRC without Enclosure 3 Resident Inspector, Monticello, USNRC without Enclosure 3 This letter is decontrolled when separated from Enclosure 3

Enclosure 1 Non-Proprietary Monticello Nuclear Generating Plant Fluence Methodology Report (23-008 / MNT-FLU-001-R-001-LNP)

Monticello Nuclear Generating Plant Fluence Methodology Report Attachment 1 (23-008 / MNT-FLU-001-R-001-LNP) 76 pages follow

QF0549, Rev. 15 (FP-E-CAL-01) Page 1 of 2 I fj;, Xcel Energy* I Calculation Signature Sheet Approval: 602000018581 Document Information NSPM Calculation (Doc) No: 23-008 I Revision: 1

Title:

MONTICELLO NUCLEAR GENERATING PLANT FLUENCE METHODOLOGY REPORT LICENSING NON-PROPRIETARY Facility: ~ MT PI I Unit: ~ 1 2 Safety Class: ~ SR Aug Q Non SR Type: Calc Sub-Type: OTH NOTE: Print and sign name in signature blocks, as required.

Major Revisions N/A I I EC Number: 601000003566 ~ Vendor Calc:

Vendor Name or Code: Transware Vendor Doc No: MNT-FLU-001-R-001-LNP Enterprises Description of Revision: Title updated in attachement section to reflect 30 cycles (incorrectly stated 21 cylces before on vendor page iii of x)

The following calculation and attachments have been reviewed and deemed acceptable as a legible QA record ~

Prepared by: (sign) / (print) by vendor Date:

Reviewed by: (sign) / (print) Matthew Date:

Sears via SAP MOC item 600001121805 Type of Review:

Design Verification Engr Review ~ OAR EOC Method Used (For DV Only):

Review Alternate Calc Test Approved by: (sign) / (print) Paul Date:

Young via SAP MOC item 600001121804 Form retained in accordance with requirements identified in FP-G-RM-01, Quality Assurance Records Control

QF0549, Rev. 15 (FP-E-CAL-01) Page 2 of 2 I fj;, Xcel Energy* I Calculation Signature Sheet Minor Revisions N/A I I EC No: ID Vendor Calc:

Minor Rev. No:

Description of Change:

Pages Affected:

The following calculation and attachments have been reviewed and deemed acceptable as a legible QA record Prepared by: (sign) / (print) Date:

Reviewed by: (sign) / (print) Date:

Type of Review:

Design Verification Engr Review OAR EOC Method Used (For DV Only):

Review Alternate Calc Test Approved by: (sign) / (print) Date:

Summary of Verification (summary is required for Design Verification):

[gJ No Comments See attached QF0528

Superseded Calculations:

Facility Calc Document Number Title Does the Calculation:

YES [gJ No Affect piping or supports? (If YES, Attach MT Form 3544.) MONTI ONLY YES [gJ No Require Fire Protection Review? (Using QF2900, Fire Protection Program Impact Screen, determine if a Fire Protection Review is required.) If YES, document the engineering review in the EC. If NO, then attach completed QF2900 to the associated EC.

Form retained in accordance with requirements identified in FP-G-RM-01, Quality Assurance Records Control

TABLE OF CONTENTS CALCULATION 23008 REV 1 Item No. Number of Pages QF0549Calculation Cover Sheet 2 Table of Contents 1 Calculation 71 Certificate of ConformanceCalculation 1 Certificate of ConformanceAttachment 1 Total = 76 pages

transware Non-Proprietary MNT-FLU-001-R-001-LNP EPl T ERPR I S ES Revision 0 Page i ofx Topical Report MONTICELLO NUCLEAR GENERATING PLANT FLUENCE METHODOLOGY REPORT C I Document Number: MNT-FLU-001-R-001-LNP Revision 0 I

April 2023 Prepared by: TransWare Enterprises Inc_

Prepared for: Xcel Energy 2807 W County Road 75 Monticello, MN 55362-9601 I I Contract Number: CW45422 Supervisor: Paul Young Project Manager: Max Smith trans tu a.re E NT E RPRIS ES Controlled Copy Number: 2 1565 Mediterranean Dr Sycamore, Illinois 60178-3141 815-895-4700

• www.transware.net

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transware Non-Proprietary MNT-FLU-001-R-001-LNP ENTERPRISES Revision 0 Page iii ofx Topical Report MONTICELLO NUCLEAR GENERATING PLANT FLUENCE METHODOLOGY REPORT Document Number: MNT-FLU-001-R-001-LNP Revision 0 April 2023 Prepared By: TransWare Enterprises Inc.

Project Team: E. A. Evans, Project Engineer H. J. Heppennann, Project Engineer M. E. Jewell, Project Engineer S. M. Wagstaff, Project Engineer K. E. Watkins, Project Engineer Project Manager:

D. ~&-eel Manager t//<i:l/B Date Reviewed By: I./L:io~3 K. E. Watkins, Project Engineer Date

~ecialist Approved By:

~Ei!';:jectMmlager Prepared For: Xcel Energy 2807 W County Road 75 Monticello, MN 55362-9601 Contract Number: CW45422 Supervisor: Paul Young Project Manager: Max Smith TransWare Enterprises Inc.* 1565 Mediterranean Dr.* Sycamore, Illinois 60178-3141

• USA

+1-815-895-4700

• www.transware.net

transwa.re Non-Proprietary MNT-FLU-001-R-001-LNP Revision 0 ENTERPRISES 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.

transwa.re Non-Proprietary MNT-FLU-001-R-001-LNP Revision 0 ENTERPRISES 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 Pumps ................................................................................... 3-7 3.2.4.2 Surveillance Capsules ................................................................ 3-7 3.2.5 Reactor Pressure Vessel ............................................................................ 3-8

(( ))

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 ........................................................................... 3-8 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 ................................................................... 3-9 3.2.10.3 Control Blades and Guide Tubes.............................................. 3-10

transwa.re Non-Proprietary MNT-FLU-001-R-001-LNP Revision 0 ENTERPRISES Page vi of x 3.2.11 Summary of the Geometry Modeling Approach ........................................ 3-10 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

transwa.re Non-Proprietary MNT-FLU-001-R-001-LNP Revision 0 ENTERPRISES Page vii of x LIST OF FIGURES Title Page Figure 2-1 Planar View of the Monticello Reactor at the Core Mid-Plane Elevation ............... 2-2 Figure 3-1 Planar View of the Monticello Fluence Model at the Core Mid-Plane Elevation in Quadrant Symmetry ........................................................................... 3-3 Figure 3-2 Axial View of the Monticello Fluence Model........................................................... 3-4

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transwa.re Non-Proprietary MNT-FLU-001-R-001-LNP Revision 0 ENTERPRISES Page ix of x LIST OF TABLES Title Page Table 2-1 Summary of Material Compositions by Region for Monticello ............................... 2-4

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INTRODUCTION This report provides an overview of the Monticello Nuclear Generating Plant (Monticello) fast neutron fluence model and fluence methodology. Monticello is a reactor unit of the General Electric BWR/3 class design with a core loading of 484 fuel assemblies in the unit. The Monticello reactor is owned and operated by Xcel Energy (Xcel).

Fluence evaluations performed for the Monticello reactor 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 (hereinafter referred to as RAMA) was developed by TransWare Enterprises Inc. under sponsorship of the Electric Power Research Institute, Inc. (EPRI) and the Boiling Water Reactor Vessel and Internals Project (BWRVIP).

All data used to construct the Monticello fluence model, define the structural and fuel materials, and develop the lifetime operating history of the reactor was provided by Xcel [4].

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 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.

transwa.re Non-Proprietary MNT-FLU-001-R-001-LNP Revision 0 ENTERPRISES Page 1-2 of 1-4 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 [9] 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 [10] 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 [11].

• The computational fluence models constructed for the Monticello reactor follows similar modeling techniques as that performed for the Susquehanna Unit 2 core shroud and top guide benchmark.

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• The computational fluence model for the Monticello reactor includes 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 a best-estimate model of the reactor.

• In addition, the following conservatisms were incorporated in the Monticello fluence model:

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

))

o The total of the shroud upper plenum volume ((

))

It is further noted that the vessel components below the core region are treated in a manner similar to the above-core components. That is, the neutron source important to the irradiation of the ((

))

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. ((

)).

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 [12], and to 10CFR21, Reporting of Defects and Noncompliance [13].

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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 model for the Monticello reactor. All reactor design and operating data inputs used to develop the model are reactor-specific and were provided by Xcel [4]. The inputs for the fluence geometry model were developed from nominal and as-built drawings for the reactor pressure vessel, vessel internals, fuel assemblies, and containment regions for the reactor.

Monticello was previously evaluated by TransWare in 2007 [14]. Several modifications were made to the Monticello RAMA geometry model since the previous fluence evaluation. ((

))

2.1 Overview of the Reactor System Design Monticello is a General Electric BWR/3 class reactor with a core loading of 484 fuel assemblies.

Figure 2-1 illustrates the basic planar configuration of the Monticello reactor model at an axial elevation near the reactor core mid-plane. All of the radial regions of the reactor 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 pumps; 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.

trans-ware Non -Proprietary MNT-FLU-001-R-001 -LNP Revision 0 C U T E R PR I SE S Pag e 2-2 of 2- 8 Surveillance Capsule Jet Pump Assembly (Mixer-Riser-Mixer)

Downcomer 1,-n--1--1-_ Reactor Pressure Vessel & Clad Thermal Insulation Drawing not lo scale F = Fuel bundle locatio ns.

(Locations shown only for the 0-90 degree quadra nt )

+ = Control rod locations 180° Biological Shield & Clad Figure 2-1 Planar View of the Monticello Reactor at the Core Mid-Plane Elevation 2.2 Reactor System Mechanical Design Inputs The mechanical design inputs used to construct the Monticello fluence geometry model is based upon ((

))

((

))

transware

[ ~I Non-Proprietary MNT-FLU-001-R-001-LNP Revision 0 T C R P R I S E S Page 2-3 of2-8 An important component of computational reactor pressure vessel fluence models is the accurate description of the surveillance capsules installed in the pressure vessels. Figure 2-1 shows that the Monticello reactor was initially equipped with three surveillance capsules. The capsules were installed at an elevation around the reactor core mid-plane. Each capsule was mounted radially near the inside surface (OT) 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 Monticello reactor, activation comparisons will be listed for the 120° and 300° capsules and the 30° capsule flux wire holder. These measurements are used to validate the fluence model. (( ))

2.3 Reactor System Material Compositions Each region of the reactor 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 Monticello reactor. 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, except for the core region, are determined assuming rated power and flow conditions.

The coolant properties remain constant unless there is a reported change in system heat balance conditions that affect the water properties in the reactor. The nuclear fuel compositions and coolant properties in the reactor core region change continuously during reactor operation. The fuel and coolant properties in the core region are updated for each reactor statepoint condition based on the actual or predicted operating states of the reactor. Water properties immediately above and below the core region ((

)) .

transware E ,i T E n P R I S E S Non-Proprietary NINT-FLU-00l-R-001-LNP Revision 0 Page 2-4 of2-8 Table 2-1 Summary of Material Compositions by Region for Monticello 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 Suppo11 Plate Rim Bolts Downcomer Region Fuel Support Pieces Fuel Hardware Regions In-Core Instrument Tube Jet Pump Hold Down Beams Jet Pump Hold Down Brackets Jet Pump Riser and Mixer Flow Areas Jet Pump Riser and Mixer Metal Jet Pump Riser Braces Jet Pump Riser Brace Pads Thennal Insulation Reactor Coolant I Moderator Reactor Core Reactor Pressure Vessel Clad Reactor Pressure Vessel Nozzle Forgings Reactor Pressure Vessel Wall Shroud Sparger Inlet Piping Steam Separator Standpipes Smveillance Capsule Flux Wire Holder Smveillance Capsule Specimen Top Guide ])

transwa.re Non-Proprietary MNT-FLU-001-R-001-LNP Revision 0 ENTERPRISES 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 Monticello core is comprised of 484 fuel assemblies in a fixed configuration. 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 ((

)).

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

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

)) that a reactor experiences over its operating life. The power history data used in the Monticello fluence evaluation includes daily thermal power levels for each cycle. The power history for Monticello also accounts for periods of reactor shutdown for refueling outages and other events that affect the activation and decay of dosimetry data. Power projection data is also needed to predict the reactor fluence at the end of (( )) licensed operation.

Each attachment to this report provides reactor-specific summaries of the operating history of the Monticello reactor for each operating cycle. The 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 Xcel and is verified against power production and exposure data obtained separately for the reactor [4].

transwa.re Non-Proprietary MNT-FLU-001-R-001-LNP Revision 0 ENTERPRISES 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 ((

))

Power history data was provided by Xcel to characterize the historical operating conditions for the Monticello reactor. ((

))

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 ((

)).

A separate neutronics transport calculation is performed for each reactor statepoint. The neutron fluxes calculated for each statepoint are then combined with the appropriate daily power history data described in Section 2.4.2 ((

)). The periods of reactor shutdown are also accounted for in this process ((

)).

2.4.4 Reactor Coolant Properties The reactor coolant water densities used in the fluence model is 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 derived ((

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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 Monticello reactor pressure vessel (RPV), reactor vessel internals (RVI), and the fast neutron fluence and activations for the reactors surveillance capsules.

The fluence model for Monticello is a reactor-specific model that is 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 ((

)) [15]. The transport solver is coupled with a general geometry modeling capability based on combinatorial geometry techniques. 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

transwa.re Non-Proprietary MNT-FLU-001-R-001-LNP Revision 0 ENTERPRISES 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. ((

))

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 [16], with extended data representations derived from the VITAMIN-B6 data library [17].

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

)) 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 describes the design inputs that were provided by Xcel for constructing the Monticello reactor fluence model. These design inputs are used to develop the reactor-specific, three-dimensional computational model of the Monticello reactor 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 Monticello fluence model. Figure 3-1 shows the planar configuration of the reactor model at an elevation corresponding to the reactor core mid-plane.

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

))

The figures are intended only to provide a perspective for the layout of the model, and specifically how the various components, structures, and regions lie relative to the reactor core region (i.e., the neutron source). (( ))

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((

))

Figure 3-1 Planar View of the Monticello Fluence Model at the Core Mid-Plane Elevation in Quadrant Symmetry

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((

))

Figure 3-2 Axial View of the Monticello Fluence Model

transwa.re Non-Proprietary MNT-FLU-001-R-001-LNP Revision 0 ENTERPRISES Page 3-5 of 3-12 3.2.1 Geometry Models The Monticello fluence model is constructed on a Cartesian coordinate system using a generalized three-dimensional geometry modeling technique based on combinatorial geometry.

The axial plane of the reactor model 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 model to be referenced by specifying the (x,y,z) coordinates for that point.

((

))

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

Figure 2-1 illustrates a planar cross-section view of the Monticello 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 model. For comparison purposes, the planar view illustrated in Figure 3-1 corresponds to the core elevation illustrated in Figure 2-1. ((

))

As previously noted, Figure 3-1 and Figure 3-2 are not drawn precisely to scale and are intended only to provide a perspective of how the various components, structures, and regions of the reactor are positioned relative to the reactor core region. The following subsections provide additional information on the constituent models developed for the individual components, structures, and regions of the fluence model.

transwa.re Non-Proprietary MNT-FLU-001-R-001-LNP Revision 0 ENTERPRISES 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.

((

))

transwa.re Non-Proprietary MNT-FLU-001-R-001-LNP Revision 0 ENTERPRISES Page 3-7 of 3-12 Above the shroud wall is the shroud head which is penetrated by numerous steam separator standpipes. ((

))

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 jet pumps and surveillance capsules in the region. ((

))

3.2.4.1 Jet Pumps The Monticello reactor has ten jet pump assemblies in the downcomer region, which provide the main recirculation flow for the core. ((

))

3.2.4.2 Surveillance Capsules Three (3) OEM surveillance capsules were initially installed in the Monticello reactor. The capsules 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. ((

))

3.2.9 Above-Core Components Figure 3-2 includes illustrations of other components and regions that lie above the reactor core region. ((

)).

transwa.re Non-Proprietary MNT-FLU-001-R-001-LNP Revision 0 ENTERPRISES Page 3-9 of 3-12 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, in-core instrumentation tubes, cruciform control rods, and the core support plate rim bolts. ((

))

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. ((

))

transwa.re Non-Proprietary MNT-FLU-001-R-001-LNP Revision 0 ENTERPRISES Page 3-10 of 3-12 3.2.10.3 Control Blades and Guide Tubes The fluence model allows for the representation of cruciform-shaped control blades and tubular control blade guide tubes in the below-core regions of a reactor. Coolant flow paths are included in the model ((

))

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 fission spectrum and neutron source terms for the transport calculation. ((

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))

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 Monticello 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 Monticello reactor in accordance with requirements of Regulatory Guide 1.190. The attachments also present the results of the combined uncertainty analysis that was performed for the Monticello reactor.

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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. Design Information Transmittal, Documents for Transware Fluence Calculation, Approval Number 602000005221: June 2022

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. 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

10. 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.

11. 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.

12. 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

transwa.re Non-Proprietary MNT-FLU-001-R-001-LNP Revision 0 ENTERPRISES Page 5-2 of 5-4

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

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

14. Monticello Reactor Pressure Vessel Fluence Evaluation at 32 EFPY and 54 EFPY, TransWare Enterprises Inc., EPR-VIP-005-R-002, Revision 0, March 2009.

15. ((

))

16. 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

17. 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

transwa.re Non-Proprietary MNT-FLU-001-R-001-LNP Revision 0 ENTERPRISES 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

transwa.re Non-Proprietary MNT-FLU-001-R-001-LNP Revision 0 ENTERPRISES 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.

transware rurrRPR 1 srs Non-Proprietary MNT-FLU-001-R-001-LNP Attachment 1, Revision 1 Page i ofx Topical Report MONTICELLO NUCLEAR GENERATING I PLANT FLUENCE METHODOLOGY REPORT i .ll

' i I 1*

l Attachment 1 Qualification of the Monticello Reactor Fluence Model - Cycles 1 to 30 I I I,

Document Number: MNT-FLU-001-R-001-LN P Attachment 1, Revision 1 I

June 2023 Prepared by: TransWare Enterprises Inc.

Prepared for: Xcel Energy 2807 W County Road 75 Monticello, MN 55362-9601 Contract Number: CW45422 Supervisor: Paul Young Project Manager: Max Smith transwa.re E NT ERP RIS ES Controlled Copy Number: _2_ _

1565 Mediterranean Dr Sycamore, Illinois 60178-3141 815*895-4700

• www.transware.net

transwa.re Non-Proprietary MNT-FLU-00I-R-001-LNP Attachment I, Revision I E NTE R P RIS E S Page ii ofx

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trans 1 . , I? Non-Proprietary MNT-FLU-001-R-001-LNP CNTCn~n,~c~

Attachment 1, Revision I Page iii ofx Topical Report MONTICELLO NUCLEAR GENERATING PLANT FLUENCE METHODOLOGY REPORT Attachment 1:

Qualification of the Monticello Reactor Fluence Model - Cycles 1 to 30 Document Number: MNT-FLU-001-R-001-LNP Attachment 1, Revision 1 June 2023 Prepared By: TransWare Enterprises Inc.

Project Team: E. A. Evans, Project Engineer H. J. Heppermann, Project Engineer M. E. Jewell, Project Engineer S. M. Wagstaff, Project Engineer K. E. Watkins, Project Engineer Project Manager:

. ~7efroject Manager Reviewed By: lC~ - '13 K . E. Watkins, Project Engineer Date K. A. J eeialist Approved By: ,,-,..=~~-.t... ~}Z9/-z.::,

D.B. Jo Date' '

Prepared For: Xcel Energy 2807 W County Road 75 Monticello, MN 55362-9601 Contract Number: CW45422 Supervisor: Paul Young Project Manager: Max Smith Trans Ware Enterprises Inc.

• 1565 Mediterranean Dr.

• Sycamore, Illinois 60178-3141

• USA

+1-815-895-4700

• www .transware.net

transwa.re Non-Proprietary MNT-FLU-001-R-001-LNP Attachment 1, Revision 1 ENT E R P R I S E S 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.

transwa.re Non-Proprietary MNT-FLU-001-R-001-LNP Attachment 1, Revision 1 ENT E R P R I S E S 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 Unit-specific Measurements........................4-3 4.2.1 Flux Wire Activation Analysis for the Monticello 30° Capsule .....................4-3 4.2.2 Cycle 23 Surveillance Capsule Activation Analysis ....................................4-4 4.2.3 Cycle 30 Surveillance Capsule Activation Analysis ....................................4-5 4.3 Reactor Pressure Vessel Lead Factors...................................................................4-6 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

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transwa.re Non-Proprietary MNT-FLU-001-R-001-LNP Attachment 1, Revision 1 ENT E R P R I S E S Page vii of x LIST OF TABLES Title Page Table 2-1 Summary of Monticello Core Loading Inventory for Cycles 1 to 23 ....................... 2-2 Table 2-2 Summary of Monticello Core Loading Inventory for Cycles 24 to 30, 31prj, and 90prj ..................................................................................................................... 2-3 Table 3-1 Statepoint Data for Monticello per Cycle Basis ..................................................... 3-2 Table 4-1 Summary of the Fluence and Activity Comparisons for the Monticello Dosimetry . 4-3 Table 4-2 Summary of the Activity Comparisons for the Copper, Iron, and Nickel Flux Wires Removed From the Monticello Reactor ...................................................... 4-3 Table 4-3 Comparison of the Calculated-to-Measured Activities for the Copper and Iron Flux Wires Removed From the Monticello 30° Flux Wire Holder .......................... 4-4 Table 4-4 Comparison of the Calculated-to-Measured Activities for the Copper, Iron, and Nickel Flux Wires Removed From the Monticello 300° Surveillance Capsule at EOC 23 ................................................................................................................ 4-5 Table 4-5 Comparison of the Calculated-to-Measured Activities for the Copper, Iron, and Nickel Flux Wires Removed From the Monticello 120° Surveillance Capsule at EOC 30 ................................................................................................................ 4-6 Table 4-6 Best-Estimate Fluence Determined for the Monticello Surveillance Capsules ....... 4-6 Table 4-7 Lead Factors Determined for the Monticello Surveillance Capsules ...................... 4-7 Table 5-1 Summary of Comparisons to Vessel Simulation Benchmark Measurements ........ 5-2 Table 5-2 Monticello RPV Combined Uncertainty for Energy > 1.0 MeV ............................... 5-3

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transwa.re Non-Proprietary MNT-FLU-001-R-001-LNP Attachment 1, Revision 1 ENT E R P R I S E S Page ix of x LIST OF FIGURES Title Page Figure 4-1 Positioning of the Surveillance Capsules Installed in the Monticello Reactor ........ 4-2

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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 Monticello Nuclear Generating Plant (Monticello) fluence methodology. The methodology that is used for determining the neutron fluence in the Monticello reactor is detailed in the Monticello Nuclear Generating 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 30. The reactor began commercial operation in 1971 with a rated thermal power of 1670 MWth. A power uprate occurred during Cycle 19 to increase the rated thermal power to 1775 MWth. The rated thermal power was increased a second time to 2004 MWth during Cycle 27. 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 Monticello 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 1.04 +/- 0.11 for all reactor dosimetry evaluated for the Monticello reactor. The combined uncertainty for the Monticello reactor is determined to be 11.60%. 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 30 for the Monticello 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 Monticello reactor was provided by Xcel in discrete form for Cycles 1 through 30.

Another important element of the reactor operating history is the fuel designs that were loaded in the reactor for each operating cycle. Each fuel design has a different power signature in the core and, therefore, results in different spatial power, exposure, and fuel isotopic distributions throughout the reactor core region.

Table 2-1 and Table 2-2 provide a summary of the fuel designs that were loaded in the Monticello reactor for each operating cycle. Table 2-1 lists the fuel designs that were loaded in the reactor for Cycles 1 through 23. Table 2-2 lists the fuel designs that were loaded for Cycles 24 through 30, the fuel designs that are loaded in projection Cycle 31 (31prj), and the fuel design for the equilibrium projection cycle identified as Cycle 90 (90prj). The dominant 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.

transware fNTrnPR I SfS Non-Proprietary MNT-FLU-00l-R-001-LNP Attachment 1, Revision 1 Page 2-2 of2-4 Table 2-1 Summary of Monticello Core Loading Inventory for Cycles 1 to 23 Fuel Designs Cycle 7x7 8x8 9x9 10x10

(( ))

1 484 2 484 3 368 116 4 288 196 5 20 464 6 484 7 372 112 8 272 212 9 184 300 10 84 400 11 92 276 116 12 44 200 240 13 124 240 120 14 236 120 128 15 100 120 128 136 16 . 92 128 264 17 108 368 8 18 332 8 140 4 19A 204 8 268 4 19B 204 8 268 4 20 68 412 4 21 376 4 104 22 244 240 23 92 392

transware f~ l Tff?PR I S f S Non-Proprietary MNT-FLU-00l-R-001-LNP Attachment 1, Revision l Page 2-3 of2-4 Table 2-2 Summary of Monticello Core Loading Inventory for Cycles 24 to 30, 31prj, and 90prj Fuel Designs Cycle 10x10 11x11

(( ))

24 484 25 484 26 484 27 484 28 484 29 336 148 30 168 316 31prj<1> 8 4761 21 90prj1 3) 484

1) ((

))

2) ((

))

3) For pmposes of flueuce projections to end of license, Cycle 90 is an equilibrium projection cycle featuring a core loading of (( )). This cycle was provided by Xcel to predict fluence at the end of the extended plant license period.

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transwa.re ENTER P RI S ES Non-Proprietary MNT-FLU-001-R-001 -LNP Attachment I, Revision 1 Page 3-1 of 3-2 3

REACTOR STATEPOINT DATA The reactor operating history is defined in discrete exposure steps referred to as statepoints. A statepoint is defined as a snapshot in time that characterizes a steady-state power-flow condition of a reactor core at a moment in time. The importance of statepoints ((

))

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, 10' s to 100 's of core calculations may be performed to track and monitor the operation of a reactor. ((

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

Power history data for Cycles 1 through 30 was provided by Xcel to characterize the historical operating conditions for the Monticello reactor. Table 3-1 shows that a total of (( ))

statepoints were used to represent the operating states of the Monticello reactor for the first 30 .

cycles of operating history. Projection data was also provided for Cycle 31 ("31 p1j") and for an equilibrium cycle ("90p1j") for projecting fluence to the reactor's end of license period.

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

((

))

A separate neutronics transpo11 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 [f

)).

transware rnrrRPR 1 srs Non-Proprietary MNT-FLU-00l-R-001-LNP Attaclunent l , Revision l Page 3-2 of 3-2 Table 3-1 Statepoint Data for Monticello per Cycle Basis Number of Reactor Rated Thermal Power Cycle Number Accumulated EFPY Statepoints (MWt) 1 (( 1670 1.2 2 1670 1.9 3 1670 2.4 4 1670 2.8 5 1670 4.4 6 1670 5.2 7 1670 6.4 8 1670 7.2 9 1670 8.2 10 1670 9.2 11 1670 10.3 12 1670 11.5 13 1670 12.9 14 1670 14.2 15 1670 15.8 16 1670 17.2 17 1670 18.6 18 1670 20.1 19A<1> 1670 21.7 19B 1775 20 1775 23.2 21 1775 24.5 22 1775 26.3 23 1775 28.2 24 1775 30.0 25 1775 31.7 26 1775 33.3 27A< 2 l 1775 34.7 27B 2004 28 2004 36.6 29 2004 38.4 30 2004 40.3 31prj<3> 2004 42.2 90prj<4> 11 2004 72.0 (1) A power uprate was implemented m.idcycle of Cycle 19 from 1670 MWth to 1775 MWth.

(2) A power uprate was implemented m.idcycle of Cycle 27 from 1775 MWth to 2004 MWth.

(3) Cycle 3lp1j is a projection cycle (( )). This cycle will be used to project reactor fluence for the duration of one cycle.

(4) Cycle 90p1j 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 reactor' s licensed period or operation.

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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 Monticello surveillance capsule dosimetry. Lead factors associating the peak RPV fluence with the capsule fluence are also reported. The results presented in this section form the basis for the validation and qualification of the fluence methodology as applied to the Monticello reactor in accordance with the Regulatory Guide 1.190 [1].

Regulatory Guide 1.190 requires that 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 Monticello reactor, there is sufficient plant-specific measurements available to qualify the calculational method (( )).

In order to report computed fluence as the best-estimate fluence, Regulatory Guide 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 the Monticello reactor is 1.04 +/- 0.11. Therefore, the computational fluence model for the Monticello reactor meets the Regulatory Guide 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 Monticello 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 Monticello reactor.

The 120° and 300° surveillance capsules were removed from the Monticello 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 set of flux wires were extracted from the flux wire holder attached to the 30° surveillance capsule for activation analysis.

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I Surveillance Capsule Drawing not to scale F = Fuel bundle locations.

(Locations shown only for the 0-90 degree quadrant)

+ = Control rod locations 180° Biological Shield & Clad Figure 4-1 Positioning of the Surveillance Capsules Installed in the Monticello Reactor Table 4-1 provides a summary of the activation comparisons for each set of flux wires that were irradiated in the Monticello reactor. The table shows that the overall calculated-to-measured (C/M) ratio and associated standard deviation for 30 irradiated specimens was determined to be 1.04 +/- 0.11.

Table 4-1 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 Me V) for each set of flux wires, and the number of specimens (viz., flux wires) evaluated for each surveillance capsule removed from the Monticello reactor.

transware r ~J T r R p R I s r s Non-Proprietary MNT-FLU-00l-R-001-LNP Attachment 1, Revision 1 Page 4-3 of 4-8 Table 4-1 Summary of the Fluence and Activity Comparisons for the Monticello Dosimetry Accumulated Fast Neutron Calculated vs. Standard Cycles of Number of Dosimeter Exposure Fluence Measured Deviation Exposure Specimens (EFPY) (>1 MeV, n/cm2) (C/M) (a) 30° Flux Wire 1- 1 1.2 -- 6 (( u 300° Capsule 1-23 28.2 (( 12 120° Capsule 1-30 40.3 11 12 n 11 Overall Average C/M and Standard Deviation 30 1.04 0.11 Table 4-2 provides the overall calculated-to-measured ratios and standard deviations for the copper, iron, and nickel flux wires that were irradiated in the Monticello reactor.

Table 4-2 Summary of the Activity Comparisons for the Copper, Iron, and Nickel Flux Wires Removed From the Monticello Reactor Number of Flux Wire C/M a Specimens Copper 11 (( ((

Iron 11 Nickel 8 l] 11 Overall Average C/M 30 1.04 0.11 and Standard Deviation 4.2 Comparison of Predicted Activation to Unit-specific Measurements The comparison of calculated activations to measurements for the dosimehy are presented in this subsection. Fluence and lead factors for each capsule are repoited in Subsection 4.3 , Reactor Pressure Vessel Lead Factors.

4.2.1 Flux Wire Activation Analysis for the Monticello 30° Capsule Copper and iron flux wires were inadiated in the Monticello 30° smveillance capsule flux wire holder during the first cycle of reactor operation. At the time of their removal, the flux wires had been inadiated for a total of 1.2 EFPY.

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

Activation calculations were performed for the following reactions: 54Fe (n,p) 54Mn and 63 Cu (n,a) 6°Co.

The average C/M ratio and associated standard deviation cletemiined for all of the flux wires inadiated in the Monticello 30° flux wire holder is 0.97 +/- 0.06.

transware r T r R p R I s r s Non-Proprietary MNT-FLU-00l-R-001-LNP Attachment 1, Revision 1 fl Page 4-4 of 4-8 Table 4-3 Comparison of the Calculated-to-Measured Activities for the Copper and Iron Flux Wires Removed From the Monticello 30° Flux Wire Holder Measured Calculated Flux Wire Dosimeter C/M a (dps/mg) (3) (dps/mg)

Copper Cu-1 3.49E+03 (( (( -

Cu-2 3.55E+03 -

Cu-3 3.51E+03 )) )) -

Copper Average (( ))

Iron Fe-1 7.12E+04 (( (( -

Fe-2 7.08E+04 -

Fe-3 7.33E+04 )) )) -

Iron Average (( ))

Average C/M and Overall Standard Deviation 0.97 0.06 4.2.2 Cycle 23 Surveillance Capsule Activation Analysis Iron, nickel, and copper flux wires were inadiated in the Monticello 300° surveillance capsules during the first twenty-three (23) cycles of reactor operation. At the time of their removal, the flux wires had been inadiated for a total of 28.2 EFPY.

Activation calculations were perfom1ed for the following reactions: 54Fe (n,p) 54Mn, 58 Ni (n,p) 58 Co, and 63Cu (n,a) 60co. The precise location of the individual wires within the capsule is not known; therefore, the activation calculations were perfo1med at the center of the capsule container. Table 4-4 provide comparisons of the calculated-to-measmed specific activities for each iron, nickel, and copper flux wire removed from the Monticello reactor at the end of Cycle 23.

The average C/M ratio and associated standard deviation dete1mined for all of the flux wires i1rndiated in the Monticello 300° flux wire holder is 1.09 +/- 0.12.

transware r~I TfJ?PR I S r ~

Non-Proprietary MNT-FLU-001-R-001-LNP Attachment 1, Revision 1 Page 4-5 of 4-8 Table 4-4 Comparison of the Calculated-to-Measured Activities for the Copper, Iron, and Nickel Flux Wires Removed From the Monticello 300° Surveillance Capsule at EOC 23 Measured Calculated Flux Wire Dosimeter C/M a (dps/mg) [4] (dps/mg)

Copper P4-Cu

• 9.81E+02 (( (( -

PS-Cu 9.69E+02 -

P9-Cu 1.14E+03 -

P10-Cu 1.07E+03 )) )) -

Copper Average (( ))

Iron P4-Fe 8.1SE+01 (( (( -

PS-Fe 8.08E+01 -

P9-Fe 9.00E+01 -

P10-Fe 8.S6E+01 )) )) -

Iron Average (( ))

Nickel P4-Ni 9.81E+02 (( (( -

PS-Ni 9.69E+02 -

P9-Ni 1.14E+03 -

P10-Ni 1.07E+03 )) )) -

Nickel Average (( ))

Average C/M and Overall Standard Deviation 1.09 0.12 4.2.3 Cycle 30 Surveillance Capsule Activation Analysis Copper, iron, and nickel flux wires were inadiated in the Monticello 120° smveillance capsules during the first thirty (30) cycles of reactor operation. At the time of their removal, the flux wires had been irradiated for a total of 40.3 EFPY.

Activation calculations were performed for the following reactions: 63 Cu (n,a) 6°Co, 54 Fe (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 perfonned at the center of the capsule container.

Table 4-5 provides comparisons of the calculated-to-measured specific activities for each iron, nickel, and copper flux wire removed from the Monticello reactor at end of Cycle 30.

The average C/M ratio and associated standard deviation detennined for all of the flux wires i1rndiated in the Monticello 120° flux wire holder is 1.02 +/- 0.10.

transware Non-Proprietary MNT-FLU-001-R-001-LNP Attachment 1, Revision 1 f~ I T f J t P R I S f S Page 4-6 of 4-8 Table 4-5 Comparison of the Calculated-to-Measured Activities for the Copper, Iron, and Nickel Flux Wires Removed From the Monticello 120° Surveillance Capsule at EOC 30 Measured Calculated Flux Wire Dosimeter C/M CJ (dps/mg) (5) (dps/mg)

Copper G4-Cu 1.98E+01 (( (( -

GS-Cu 1.76E+01 -

G9-Cu 1.97E+01 -

G10-Cu 1.84E+01 11 11 -

Copper Average (( ))

Iron G4-Fe 1.19E+02 (( (( -

GS-Fe 1.0SE+02 -

G9-Fe 1.18E+02 -

G10-Fe 1.17E+02 11 11 -

Iron Average (( ))

Nickel G4-Ni 1.46E+03 (( (( -

GS-Ni 1.29E+03 -

G9-Ni 1.S6E+03 -

G10-Ni 1.S7E+03 11 11 -

Nickel Average (( ))

Average C/M and Overall Standard Deviation 1.02 0.10 4.3 Reactor Pressure Vessel Lead Factors Table 4-6 and Table 4-7 provide the best-estimate fast neuh'on fluence detennined for the reactor pressure vessel and dosimeters and the associated lead factors for each evaluated time period for the Monticello reactor.

Table 4-6 Best-Estimate Fluence Determined for the Monticello Surveillance Capsules Evaluated Dosimeter Fluence (n/cm2) Peak RPV Fluence (n/cm 2)

Time Period 120° 300° OT 1/4T EOC23 -- (( )) (( 11 (( ))

EOC 30 (( l] -- (( 11 (( 11

transware r ~1 rrRPR 1 s r s Non-Proprietary MNT-FLU-001-R-001-LNP Attachment l , Revision 1 Page 4-7 of 4-8 Table 4-7 Lead Factors Determined for the Monticello Surveillance Capsules Lead Factor Evaluated Time 120° 300° Period OT 1/4T OT 1/4T EOC23 - -- 0.40 0.54 EOC 30 0.40 0.55 -- --

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 (OT) of the RPV. A second lead factor is also provided assuming the peak damage fluence at the l/4T depth of the RPV wall.

The calculated-to-measmed activation comparisons for the smveillance capsules presented in the previous sections show no disceruable bias in the computational fluence method. Therefore, the best-estimate fluence repo11ed for each capsule in Table 4-6 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 Monticello 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 (la) and bias in the computed RPV fluence.

The requirements for determining the combined unce1tainty 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 unce1tainty introduced by the measurement process. After analysis of these factors, it is determined that the combined unce1tainty for Monticello RPV fluence is 11.60%, and that no adjustment for bias is required for the RPV fast neutron fluence determined for the period Cycle I through the end of Cycle 30.

5.1 Comparison Uncertainty Comparison uncertainty factors are determined by comparing calculated activities with activity measurements. For pressure vessel fluence evaluations, two comparison unce1iainty factors are considered: operating reactor comparison factors and benchmark comparison factors.

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

))

The Monticello reactor is a BWR/3 class design. ((

)) The overall C/M and standard deviation for the BWR/3 class plant measurements is determined to be an unbiased 1.05 +/- 0.14. ((

)) The overall comparison ratio for all BWR class plants evaluated as of the date of this report is 1.01 +/- 0.10. ((

transware r~1 r r n PR 1 srs Non-Proprietary ?vINT-FLU-001-R-001-LNP Attachment 1 Revision 1 Page 5-2 of 5-4

))

5. 1.2 Benchmark Comparison Uncertainty The benchmark comparison 1mcertainty is based on a set of industry standard simulation benchmark comparisons. ((

)) 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 sullllllarizes the calculated-to-measurement (C/M) results dete1mined for these vessel simulation benchmarks.

Table 5-1 Summary of Comparisons to Vessel Simulation Benchmark Measurements Average Number of Benchmark Calculated-to- St. Dev. (1 a)

Measurements Measured (C/M)

Pool Critical Assembly ((

VENUS-3 ))

Total Simulated Vessel 413 1.03 +/-0.05 Comparisons 5.2 Analytic Uncertainty The calculational models used for fluence analyses are comprised of numerous analytical parameters that have associated 1mcertainties in their values. The 1mcertainty in these parameters needs to be tested for its contribution to the overall fluence unce11ainty.

((

transware r ~, r T r. p R I s F s Non-Proprietary MNT-FLU-001-R-001-LNP Attachment 1, Revision 1 Page 5-3 of 5-4

))

5.3 Combined Uncertainty The combined uncertainty for the reactor pressure vessel fluence evaluation is detenuined with a weighting fimction ((

)). Table 5-2 shows that the combined uncertainty (lcr) detennined for the Monticello reactor pressure vessel fluence is 11.60% for neutron energy exceeding 1.0 MeV.

It is shown in Table 5-2 that the combined lmce11ainty is well below the 20% lmce1iainty limit specified in Regulat01y Guide 1.190. In accordance with Regulato1y Guide 1.190, there is no discemable bias in the computed RPV fluence. Therefore, no adjustment to the RPV fast neutron fluence for the period coITesponding to Cycle 1 thrnugh the end of Cycle 30 is required.

Table 5-2 Monticello RPV Combined Uncertainty for Energy > 1.0 MeV Uncertainty Term Value Combined Uncertainty (1cr) 11.60%

Bias None< 1l l) The bias tennis less than its constituent u11ce11ainty values, concluding that no statistically significant bias exists.

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REFERENCES 6.1 References

1. "Monticello Nuclear Generating Plant Fluence Methodology Report," Trans Ware Enterprises Inc. Document Number MNT-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. "Neutron Dosimeter Fast Fluence Measurement at Monticello Nuclear Power Station,"

Letter from Roy E. Willis, GE to C. E. Larson, Monticello: 1973

4. "Testing and Evaluation of the Monticello 300 Degree Surveillance Capsule," MP Machinery and Testing LLC Repo11, MPM-908991: 2008

5. "Testing and Evaluation of the Monticello Nuclear Generating Plant 120 Degree Surveillance Capsule," MP Machinery and Testing LLC Rep011, MPM-1221220: 2021

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

transwa.re E t\J T E R P RI S E S 1565 Mediterranean Drive

• Sycamore, Illinois 60178 phone 815.895.4700 fax 815.895.4704 www.transware.net CERTIFICATE OF CONFORMANCE

REFERENCE:

Xcel Energy Agreement No. CW45422 TransWare Enterprises Inc. hereby certifies that the product identified below is being furnished in accordance with the TransWare Quality Assurance Program Manual, Revision 3, dated October 2, 2022. TransWare also certifies that the product is in conformance with all requirements prescribed in the referenced purchase order, 10CFR50 Appendix B, 10CFR21, and applicable codes and standards as specified in TransWare Enterprises Inc.

Project Plan No. MNT-FLU-001-Q-001, Revision 0, dated June 24, 2022.

Steve Wagstaff,J\Man;ge Date TransWare Enterprises Inc.

Product:

1. TransWare Report Number MNT-FLU-001-R-001, Revision 1. "Monticello Nuclear Generating Plant Fluence Methodology Report", April 2023.

2. TransWare Report Number MNT-FLU-001-R-001, Attachment 1, Revision 1.

"Qualification of the Monticello Reactor Fluence Model - Cycles 1 to 30," April 2023

3. TransWare Report Number MNT-FLU-001-R-001-LP, Revision 0. "Monticello Nuclear Generating Plant Fluence Methodology Report," April 2023.

4. TransWare Report Number MNT-FLU-001-R-001-LP, Revision 0. "Qualification of the Monticello Reactor Fluence Model - Cycles 1 to 30," April 2023.

5. TransWare Report Number MNT-FLU-001-R-001-LNP, Revision 0. "Monticello Nuclear Generating Plant Fluence Methodology Report," April 2023.

6. TransWare Report Number MNT-FLU-001-R-001-LNP, Revision 0. "Qualification of the Monticello Reactor Fluence Model - Cycles 1 to 30," April 2023.

transw ENTERPR I SES CERTIFICATE OF CONFORMANCE

REFERENCE:

Xcel Energy Agreement No. CW45422 TransWare Enterprises Inc. hereby certifies that the product identified below is being furnished in accordance with the TransWare Quality Assurance Program Manual, Revision 3, dated October 2, 2022. Trans Ware also certifies that the product is in conformance with all requirements prescribed in the referenced purchase order, 10CFRS0 Appendix B, 10CFR21, and applicable codes and standards as specified in TransWare Enterprises Inc.

Project Plan No. MNT-FLU-001 -Q-001, Revision 0, dated June 24, 2022.

Steve Wagstaff,QAa;ager Date TransWare Enterprises Inc.

Product:

1. TransWare Report Number MNT-FLU-001-R-001, Attachment 1, Revision 2.

"Qualification of the Monticello Reactor Fluence Model - Cycles 1 to 30," June 2023.

2. TransWare Report Number MNT-FLU-001-R-001-LP, Attachment 1, Revision 1.

"Qualification of the Monticello Reactor Fluence Model - Cycles 1 to 30," June 2023.

3. TransWare Report Number MNT-FLU-001-R-001-LNP, Attachment 1, Revision 1.

"Qualification of the Monticello Reactor Fluence Model - Cycles 1 to 30," June 2023.

Enclosure 2 Affidavits 6 pages follow

Affidavit I, Kenneth E. Watkins, state as follows:

1. I am the Vice President 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. MNT-FLU-001-R-001-LP, Revision 0, Monticello Nuclear Generating Plant Fluence Methodology Report, April 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.

The precise value of the expertise to devise an evaluation process and apply the correct analytical methodology is difficult to quantify, but it is clearly substantial.

TWE' s competitive advantage will be lost if its competitors are able to use the results of the TWE experience to normalize or verify their own process or if they are able to claim an equivalent understanding by demonstrating that they can arrive at the same or similar conclusions.

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

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

Executed at Sycamore, Illinois, this 21 st day of April 2023.

Q I ,U- ~ d_~ci_~

_g;: _...,.f. ~---- .::P

~~ Kenneth E. Watkins TransWare Enterprises Inc.

IAL SEAL" IS KEEF I STATE OF IWNOIS IRES 8/3/2025

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. MNT-FLU-001-R-001-LP, Attachment 1, Revision 1, Qualification of the Monticello Reactor Fluence Model - Cycles 1 to 30, June 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.

The precise value of the expertise to devise an evaluation process and apply the correct analytical methodology is difficult to quantify, but it is clearly substantial.

TWE's competitive advantage will be lost if its competitors are able to use the results of the TWE experience to normalize or verify their own process or if they are able to claim an equivalent understanding by demonstrating that they can arrive at the same or similar conclusions.

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

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

Executed at St. Louis, Missouri, this 29th day of June 2023.

STATE(),- Ml~-:.vuhl COUNTY OF Sf- UM) (o..,.,,y KathleenAJones TransWare Enterprises Inc.

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