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LLC - Submittal of Fluence Calculation Methodology and Results, TR-0116-20781, Revision 1
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LO-0619-66076 July 2, 2019 Docket No.52-048 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk One White Flint North 11555 Rockville Pike Rockville, MD 20852-2738

SUBJECT:

NuScale Power, LLC Submittal of Fluence Calculation Methodology and Results, TR-0116-20781, Revision 1

REFERENCES:

1. Letter from NuScale Power, LLC to Nuclear Regulatory Commission, NuScale Power, LLC Submittal of Technical Reports Supporting the NuScale Design Certification Application (NRC Project No. 0769), dated December 30, 2016 (ML17005A112)

2. NuScale Technical Report, Fluence Calculation Methodology and Results, TR-0116-20781, Revision 0, dated December 2016 (ML17005A116)

NuScale Power, LLC (NuScale) hereby submits Revision 1 of the Fluence Calculation Methodology and Results (TR-0116-20781) technical report. contains the proprietary version of the report entitled Fluence Calculation Methodology and Results. NuScale requests that the proprietary version be withheld from public disclosure in accordance with the requirements of 10 CFR § 2.390. The enclosed affidavit (Enclosure 3) supports this request. Enclosure 1 has also been determined to contain Export Controlled Information. This information must be protected from disclosure per the requirements of 10 CFR § 810. Enclosure 2 contains the nonproprietary version of the report.

This letter makes no regulatory commitments and no revisions to any existing regulatory commitments.

If you have any questions, please contact Matthew Presson at 541-452-7531 or at mpresson@nuscalepower.com.

Sincerely, Zackary W. Rad Director, Regulatory Affairs NuScale Power, LLC Distribution: Samuel Lee, NRC, OWFN-8H12 Gregory Cranston, NRC, OWFN-8H12 Bruce Bavol, NRC, OWFN-8H12 : Fluence Calculation Methodology and Results, TR-0116-20781-P, Revision 1, proprietary version : Fluence Calculation Methodology and Results, TR-0116-20781-NP, Revision 1 nonproprietary version : Affidavit of Zackary W. Rad, AF-0619-66077 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 Office 541.360-0500 Fax 541.207.3928 www.nuscalepower.com

LO-0619-66076 :

Fluence Calculation Methodology and Results, TR-0116-20781-P, Revision 1, proprietary version NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 Office 541.360-0500 Fax 541.207.3928 www.nuscalepower.com

LO-0619-66076 :

Fluence Calculation Methodology and Results, TR-0116-20781-NP, Revision 1, nonproprietary version NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 Office 541.360-0500 Fax 541.207.3928 www.nuscalepower.com

Fluence Calculation Methodology and Results TR-0116-20781-NP Rev. 1 Licensing Technical Report Fluence Calculation Methodology and Results July 2019 Revision 1 Docket: 52-048 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 www.nuscalepower.com

Fluence Calculation Methodology and Results TR-0116-20781-NP Rev. 1 Licensing Technical Report COPYRIGHT NOTICE This report has been prepared by NuScale Power, LLC and bears a NuScale Power, LLC, copyright notice. No right to disclose, use, or copy any of the information in this report, other than by the U.S.

Nuclear Regulatory Commission (NRC), is authorized without the express, written permission of NuScale Power, LLC.

The NRC is permitted to make the number of copies of the information contained in this report that is necessary for its internal use in connection with generic and plant-specific reviews and approvals, as well as the issuance, denial, amendment, transfer, renewal, modification, suspension, revocation, or violation of a license, permit, order, or regulation subject to the requirements of 10 CFR 2.390 regarding restrictions on public disclosure to the extent such information has been identified as proprietary by NuScale Power, LLC, copyright protection notwithstanding. Regarding nonproprietary versions of these reports, the NRC is permitted to make the number of copies necessary for public viewing in appropriate docket files in public document rooms in Washington, DC, and elsewhere as may be required by NRC regulations. Copies made by the NRC must include this copyright notice and contain the proprietary marking if the original was identified as proprietary.

Fluence Calculation Methodology and Results TR-0116-20781-NP Rev. 1 Licensing Technical Report Department of Energy Acknowledgement and Disclaimer This material is based upon work supported by the Department of Energy under Award Numbers DE-NE0000633, DE-NE0008742, and DE-NE0008820.

This report was prepared as an account of work sponsored by an agency of the United States Government. Neither the United States Government nor any agency thereof, nor any of their employees, makes any warranty, express or implied, or assumes any legal liability or responsibility for the accuracy, completeness, or usefulness of any information, apparatus, product, or process disclosed, or represents that its use would not infringe privately owned rights. Reference herein to any specific commercial product, process, or service by trade name, trademark, manufacturer, or otherwise does not necessarily constitute or imply its endorsement, recommendation, or favoring by the United States Government or any agency thereof. The views and opinions of authors expressed herein do not necessarily state or reflect those of the United States Government or any agency thereof.

Fluence Calculation Methodology and Results TR-0116-20781-NP Rev. 1 Licensing Technical Report CONTENTS Abstract ....................................................................................................................................... 1 1.0 Introduction ..................................................................................................................... 2 1.1 Purpose ................................................................................................................. 2 1.2 Scope .................................................................................................................... 2 1.3 Abbreviations, Acronyms, and Definitions ............................................................. 2 2.0 Background ..................................................................................................................... 3 2.1 Regulatory Requirements ...................................................................................... 3 3.0 NuScale Power Module Fluence Prediction Methodology .......................................... 5 3.1 Overview ............................................................................................................... 5 3.2 Geometry ............................................................................................................... 5 3.3 Material Compositions ........................................................................................... 9 3.4 Cross Sections .................................................................................................... 10 3.5 Neutron Source ................................................................................................... 10 3.6 Other Modeling Considerations ........................................................................... 14 4.0 Bias and Uncertainty .................................................................................................... 18 4.1 Quantified Biases and Uncertainties ................................................................... 18 4.2 Combination of Biases ........................................................................................ 19 4.3 Combination of Uncertainties .............................................................................. 20 5.0 NuScale Power Module Fluence Prediction Results.................................................. 21 6.0 Summary and Conclusions .......................................................................................... 25 7.0 References ..................................................................................................................... 26 Appendix A. Benchmarking Monte Carlo N-Particle Transport Code 6 for Fluence Applications ....................................................................................................... 27 A.1 Vulcain Experimental Nuclear Study 3 Benchmark .......................................... 27 A.1.1 Modeling ......................................................................................................... 27 Appendix B. NuScale Power Module Fluence Prediction Sensitivity Studies and Uncertainty Analysis ......................................................................................... 33 B.1 Sensitivity Studies ............................................................................................. 33 B.1.1 Homogenized Fuel Model vs Explicit Fuel Model ........................................... 33 B.1.2 Contribution of Pu239 to Neutron Source ......................................................... 33 B.1.3 Material Composition ...................................................................................... 37 B.1.4 Geometrical Tolerances .................................................................................. 37 B.1.5 Assembly Averaged Neutron Source Bias and Uncertainty ............................ 37

Fluence Calculation Methodology and Results TR-0116-20781-NP Rev. 1 Licensing Technical Report

............................................................................................................................ 47 B.1.6 Core Power ..................................................................................................... 47 B.1.7 Radial Power Profile ....................................................................................... 47 B.1.8 Axial Power Profile.......................................................................................... 49 B.1.9 Boron Concentration ....................................................................................... 52 B.1.10 Nuclear Cross Section Data and Transport Code......................................... 52 B.1.11 Monte Carlo Method ..................................................................................... 52 B.1.12 Water Density ............................................................................................... 52 B.1.13 Axial Coolant Density Bias............................................................................ 53 B.1.14 Tally Mesh Size ............................................................................................. 56 Appendix C. Regulatory Guide 1.190 Alternatives ............................................................... 61 TABLES Table 1-1 Abbreviations and Acronyms ................................................................................. 2 Table 1-2 Definitions .............................................................................................................. 2 Table 3-1 Lifetime exposure averaged core axial power profile .......................................... 12 Table 3-2 Lifetime exposure averaged assembly averaged radial power profile................. 13 Table 4-1 List of quantified systematic biases (+ or -) and random uncertainties(+/-) ......... 18 Table 5-1 Best estimate of fluence expected to be experienced in various NuScale Power Module components and locations ...................................................................... 21 Table A-1 Vulcain Experimental Nuclear Study 3 experimental versus calculated results .................................................................................................................. 32 Table B-1 The averaged fast neutron flux in pin lattice of fuel assembly G3 from SIMULATE5, cycle 8 (unit: x1013 n/cm2 sec) ....................................................... 40 Table B-2 The averaged fast neutron flux in pin lattice of fuel assembly G4 from SIMULATE5, cycle 8 (unit: x1013 n/cm2 sec) ....................................................... 41 Table B-3 The averaged fast neutron flux in pin lattice of fuel assembly F6 from SIMULATE5, cycle 8 (unit: x1013 n/cm2 sec) ....................................................... 42 Table B-4 The averaged fast neutron flux in pin lattice of fuel assembly E7 from SIMULATE5, cycle 8 (unit: x1013 n/cm2 sec) ....................................................... 43 Table B-5 The averaged fast neutron flux in pin lattice of fuel assembly G3 from MCNP6, cycle 8 (unit: x1013 n/cm2 sec) ............................................................................. 44 Table B-6 The averaged fast neutron flux in pin lattice of fuel assembly G4 from MCNP6, cycle 8 (unit: x1013 n/cm2 sec) ............................................................................. 45 Table B-7 The averaged fast neutron flux in pin lattice of fuel assembly F6 from MCNP6, cycle 8 (unit: x1013 n/cm2 sec) ............................................................................. 46 Table B-8 The averaged fast neutron flux in pin lattice of fuel assembly E7 from MCNP6, cycle 8 (unit: x1013 n/cm2 sec) ............................................................................. 47 Table B-9 Average Axial Power Profiles .............................................................................. 50 Table B-10 Variance and Standard Deviation for the Axial Power Profiles ............................ 51 Table B-11 Coolant Water Axial Density Variations ............................................................... 55 Table B-12 Peak Fluence Results for Axially Varied Coolant Density .................................... 56 Table B-13 Peak Fluence Results for smaller tally subdivisions ............................................ 60

Fluence Calculation Methodology and Results TR-0116-20781-NP Rev. 1 Licensing Technical Report Table C-1 Alternatives to particular Regulatory Guide 1.190 regulatory positions ............... 61 FIGURES Figure 3-1 Vertical cross-sectional view of the lower section of the NuScale Power Module ................................................................................................................... 7 Figure 3-2 Vertical cross-sectional view of the Monte Carlo N-Particle Transport Code 6 fluence homogenized model ................................................................................. 8 Figure 3-3 Horizontal cross-sectional view of the Monte Carlo N-Particle Transport Code 6 fluence homogenized model ................................................................................. 9 Figure 3-4 Fuel assembly naming index ............................................................................... 11 Figure 3-5 Vertical cross-sectional view of the reactor pressure vessel ............................... 15 Figure 3-6 Horizontal cross-sectional view of the reactor pressure vessel mesh tally.......... 16 Figure 3-7 Horizontal cross-sectional view of the containment vessel mesh tally ............... 17 Figure A-1 Horizontal cross-sectional view of the Vulcain Experimental Nuclear Study 3 benchmark geometry ........................................................................................... 28 Figure A-2 Vertical cross-sectional view of the Monte Carlo N-Particle Transport Code 6 model of the Vulcain Experimental Nuclear Study 3 benchmark ......................... 29 Figure A-3 Horizontal cross-sectional view of the inner and outer baffle of the Monte Carlo N-Particle Transport Code 6 model of the Vulcain Experimental Nuclear Study 3 benchmark ........................................................................................................ 30

Fluence Calculation Methodology and Results TR-0116-20781-NP Rev. 1 Abstract The methodology developed by NuScale Power, LLC, to calculate the neutron fluence for the NuScale Power Module reactor pressure vessel and containment vessel is provided by this Technical Report. Estimations of the bias and uncertainty associated with these fluence calculations, derived from benchmarking and sensitivity studies, are presented along with associated end of life fluence predictions for the NuScale reactor pressure vessel, containment vessel, and other locations.

NuScales fluence methodology uses the Monte Carlo N-Particle Transport Code 6 and is based on the guidance found in Regulatory Guide 1.190, Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence (Reference 7.1). Alternatives to particular Regulatory Guide 1.190 regulatory positions are described and justified. Measured data from the VENUS-3 pressure vessel simulator benchmark is used to validate the NuScale methodology.

Fluence Calculation Methodology and Results TR-0116-20781-NP Rev. 1 1.0 Introduction 1.1 Purpose The purpose of this report is to provide the methodology used to calculate the neutron fluence for the NuScale Power Module (NPM) reactor pressure vessel (RPV) and containment vessel (CNV). This report also provides the estimations of the bias and uncertainty associated with these fluence calculations, derived from benchmarking and sensitivity studies, along with associated end of life fluence predictions for the NuScale RPV, CNV, and other locations.

1.2 Scope This report covers the methodology for predicting the end of life fluence for the NuScale RPV and NuScale CNV as well as the associated results of applying the methodology to support the Final Safety Analysis Report Section 4.3 of the NuScale Design Certification Application (DCA). The testing program associated with confirming these fluence predictions in the operating plant, the methodology for adjusting best estimate fluence predictions throughout an NPMs operating life, and the effects on material properties caused by the fluence are outside of the scope of this report.

1.3 Abbreviations, Acronyms, and Definitions Table 1-1 Abbreviations and Acronyms Term Definition BN bottom nozzle CMS core management software CNV containment vessel CRA control rod assembly DCA Design Certification Application MCNP Monte Carlo N-Particle Transport Code MeV megaelectronvolt NPM NuScale Power Module RG Regulatory Guide RPV reactor pressure vessel TN top nozzle VENUS-3 Vulcain Experimental Nuclear Study 3 Table 1-2 Definitions Term Definition Fluence In the context of this report, the term fluence is always taken to mean the fast neutron fluence, which is the time integrated flux of neutrons with an energy greater than one megaelectronvolt (MeV).

Fluence Calculation Methodology and Results TR-0116-20781-NP Rev. 1 2.0 Background Neutron fluence is known to affect the material properties of RPV materials. The extent of the effect is influenced by the magnitude of the fluence, among other factors.

Regulatory Guide (RG) 1.190, Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence, provides guidance for calculating pressure vessel neutron fluence. NuScales fluence calculation methodology is based on RG 1.190.

Descriptions of, and justifications for, alternatives to certain applicable portions of RG 1.190 regulatory positions are provided in Appendix C.

The NuScale CNV is in close proximity to the RPV compared to a typical large light water reactor and the same methodology used to calculate RPV fluence is taken to be directly applicable to calculating CNV fluence.

2.1 Regulatory Requirements This report in conjunction with the NuScale Pressure and Temperature Limits Methodology Technical Report (Reference 7.5) and Final Safety Analysis Report Sections 4.3 and 5.3 address the regulatory requirements pertaining to vessel fluence analysis and surveillance.

The regulatory requirements pertaining to vessel fluence analysis and surveillance are as follows:

10 CFR Part 50 Appendix A, General Design Criterion 14 as it relates to ensuring an extremely low probability of abnormal leakage, of rapidly propagating failure, and of gross rupture of the reactor coolant pressure boundary, in part, insofar as it considers calculations of neutron fluence.

General Design Criterion 31 as it relates to ensuring that the reactor coolant pressure boundary will behave in a nonbrittle manner and that the probability of rapidly propagating fracture is minimized, in part, insofar as it considers calculations of fluence.

Appendix G, to 10 CFR Part 50, as it relates to RPV material fracture toughness requirements, in part, insofar as it considers calculations of neutron fluence.

Appendix H, to 10 CFR Part 50, as it relates to RPV material surveillance program requirements, in part, insofar as it considers calculations of neutron fluence.

10 CFR 50.61 as it relates to fracture toughness criteria for pressurized water reactors relevant to pressurized thermal shock events, in part, insofar as it considers calculations of neutron fluence.

Fluence Calculation Methodology and Results TR-0116-20781-NP Rev. 1 Standard Review Plan Section 4.3 and Design Specific Review Standard Sections 5.3.2 and 5.3.3 provide the following applicable NRC acceptance criteria for the vessel fluence analysis methodology:

There is reasonable assurance that the proposed design limits can be met for the expected range of reactor operation, taking into account analysis uncertainties.

There is reasonable assurance that during normal operation the design limits will not be exceeded.

The acceptance criteria of RG 1.190, Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence.

The acceptance criteria of RG 1.99, Radiation Embrittlement of Reactor Vessel Materials.

Fluence Calculation Methodology and Results TR-0116-20781-NP Rev. 1 3.0 NuScale Power Module Fluence Prediction Methodology 3.1 Overview NuScales fluence calculation methodology uses Monte Carlo N-Particle Transport Code 6 version 1.0 (MCNP6), which was released in 2013 by Los Alamos National Laboratory and merges MCNP5 and MCNPX functions. The MCNP6 code is a general-purpose Monte Carlo method code that can be used for neutron, photon, electron, or coupled neutron/photon/electron transport. The code treats an arbitrary three-dimensional configuration of materials in geometric cells. The Monte Carlo method has the advantage of allowing an exact representation of the reactors three dimensional geometry. In addition, the Monte Carlo method allows a continuous energy description of the nuclear cross-sections and flux solution.

NuScale calculates three dimensional exposure and power distribution data for each fuel assembly using core management software (CMS) codes CASMO5 and SIMULATE5.

CASMO5 is a lattice physics code that characterizes reactor fuel assembly designs.

SIMULATE5 is a three-dimensional core simulator code for core design and core follow calculations. Information from CASMO5 and SIMULATE5 is used as inputs to the MCNP6 based fluence calculation.

3.2 Geometry Calculations are run on a three-dimensional MCNP6 model.

An illustration of the vertical cross sectional view of the lower section of the NPM is shown in Figure 3-1. The vertical cross sectional view of the MCNP6 NuScale best estimate fluence model is presented in Figure 3-2 and the horizontal cross sectional view is presented in Figure 3-3.

The NuScale best estimate fluence model is representative of the standard NPM design submitted as part of the DCA with the following general exceptions and modeling simplifications.

The geometry is specified using cold dimensions and thermal expansion is not modeled. Thermal expansion for hot full power dimensions is accounted for in NuScales Studsvik Scandpower CMS codes (SIMULATE5 and CASMO5), whose outputs are used as inputs to establish the neutron source distribution in the MCNP6 model. The effect of this modeling simplification and the effect of this difference between MCNP6 and CMS treatment of cold dimensions on the fluence estimate is discussed in Section B.1.3 and Section B.1.4.

The NuScale best estimate fluence model contains an axially homogenized representation of the active fuel region of the fuel assemblies. This modeling simplification was implemented for consistency because fuel assembly power information was taken from NuScales SIMULATE5 models output, which is a homogenized model. A sensitivity study comparing this homogenized treatment to an

Fluence Calculation Methodology and Results TR-0116-20781-NP Rev. 1 MCNP6 model that explicitly models the fuel across ((2(a),(c) nodes is presented in Section B.1.1.

Each fuel assembly consists of ((

                                                                                                     }}2(a),(c). The active fuel pin region consists of a ((
                                                                   }}2(a),(c). On the basis of engineering judgment, the impact of this modeling simplification on the fluence estimates is negligible.

The TN skirt and upper core plate are modeled explicitly as part of the fuel assembly for assemblies that do not contain control rod assemblies (CRAs). ((

                                                                      }}2(a),(c). On the basis of engineering judgment, the impact of this modeling simplification on the fluence estimates is negligible.

The calculation of the homogenized plenums composition is based on the cross-sectional area within the guide and instrument tube, rather than the actual volume of the tube. This difference will lead to a higher zircalloy fraction versus water fraction.

On the basis of engineering judgment, the impact of this modeling simplification on the fluence estimates is negligible.

The RPV bottom core support block was not explicitly modeled. The RPV beltline region is the main region of interest for the vessel fluence estimation. On the basis of engineering judgment, the impact of this modeling simplification on the RPV beltline region fluence estimates is negligible.
Only ((

                      }}2(a),(c). The effect of this minor error on the fluence estimate is discussed in Section B.1.12.

All water densities in the NuScale best estimate fluence model are ((

                                                                                           }}2(a),(c). The effect of this modeling simplification on the fluence estimate is discussed in Section B.1.12.

All temperatures of components in the NuScale best estimate fluence model are

(( }}2(a),(c). On the basis of engineering judgment, the impact of this modeling simplification on the fluence estimates is small relative to the effect of using a single water coolant density for the primary coolant. © Copyright 2019 by NuScale Power, LLC 6

Fluence Calculation Methodology and Results TR-0116-20781-NP Rev. 1 ((

                                                                                                              }}2(a),(c)

Figure 3-2 Vertical cross-sectional view of the Monte Carlo N-Particle Transport Code 6 fluence homogenized model

  © Copyright 2019 by NuScale Power, LLC 8

Fluence Calculation Methodology and Results TR-0116-20781-NP Rev. 1 Figure 3-3 Horizontal cross-sectional view of the Monte Carlo N-Particle Transport Code 6 fluence homogenized model 3.3 Material Compositions The material composition information used in the MCNP6 NuScale best estimate fluence model is based on the typical isotopic contents associated with the materials associated with the NPM design. Cold dimensions are used and thermal expansion is not taken into account in the determination of material densities. The effect of this modeling simplification on the fluence estimate is discussed in Section B.1.3 and Section B.1.4. © Copyright 2019 by NuScale Power, LLC 9

Fluence Calculation Methodology and Results TR-0116-20781-NP Rev. 1 The core composition of the MCNP6 base model is based on the core composition of the SIMULATE5 base model core design. The NuScale best estimate fluence model does not contain any Pu239 since it is based on a fresh core (beginning of cycle of Cycle 1). A bias and uncertainty to account for the contribution of Pu239 buildup to fluence is derived in Section B.1.2. The material composition of the homogenized active fuel is composed of fuel at an averaged 3.5 percent enrichment, fuel cladding, borated water, and guide tubes. 3.4 Cross Sections NuScales MCNP6 based fluence calculation methodology uses the ENDF/B-VII.1 nuclear data for continuous energy cross section libraries. A .92c file extension was used to represent isotopic cross-section data with a temperature at (( }}2(a),(c). The ENDF/B-VII.1 data libraries have cross-sections processed at selected temperatures ((

                                        }}2(a),(c). The MAKXSF code was used to derive the ((
                         }} 2(a),(c) library from ((                                        }}2(a),(c) and ((
                       }} 2(a),(c) libraries.

The temperature card TMP is used in MCNP6 to provide the time-dependent cell thermal temperatures that are necessary for the free-gas thermal treatment of low-energy neutron transport at the correct material temperatures. The temperature card TMP requires inputs to be in units of megaelectronvolts so a conversion is performed. For example, NuScale uses (( }}2(a),(c) as the averaged temperature of moderator and this temperature in K is converted to megaelectronvolts as shown in Equation 3-1. (( 8.617385 10 557.2 ° = 4.801866 10 () }}2(a),(c) Eq. 3-1 3.5 Neutron Source For the NuScale best estimate fluence model, the energy spectrum of the fission neutrons emitted from the fuel assemblies is taken as the Watt fission spectrum for U235 . Sensitivity studies on the effect of Pu239 buildup are presented in Section B.1.2. There are no delayed neutrons modeled since the fission modeling is turned off by using the NONU card in MCNP6 input decks for neutron transport. For the purpose of the NuScale best estimate of fast neutron fluence, the delayed neutron contribution to fast neutron fluence is negligible. For the purposes of this report, the fuel assemblies are refered to according to the naming index shown in Figure 3-4. © Copyright 2019 by NuScale Power, LLC 10

Fluence Calculation Methodology and Results TR-0116-20781-NP Rev. 1 Figure 3-4 Fuel assembly naming index SIMULATE5 was used to calculate the core average axial power profile associated with each cycle in a lifetime refueling scheme for (( }}2(a),(c). The axial power profiles associated with each cycle were averaged to produce the 8-cycle exposure averaged axial power profile shown in Table 3-1. Table 3-1 was used to establish the vertical sampling of the neutron source used in the MCNP6 NuScale best estimate fluence model. © Copyright 2019 by NuScale Power, LLC 11

Fluence Calculation Methodology and Results TR-0116-20781-NP Rev. 1 Table 3-1 Lifetime exposure averaged core axial power profile ((

                                                                                                  }}2(a),(c),ECI SIMULATE5 was used to calculate the assembly averaged radial power profile associated with each cycle in an 8-cycle refueling scheme. The assembly averaged radial power profile associated with each cycle were averaged to produce the 8-cycle exposure averaged radial power profile shown in Table 3-2. The radial sampling of the neutron source used in the MCNP6 NuScale best estimate fluence model is based on Table 3-2.
  © Copyright 2019 by NuScale Power, LLC 12

Fluence Calculation Methodology and Results TR-0116-20781-NP Rev. 1 Table 3-2 Lifetime exposure averaged assembly averaged radial power profile ((

                                                                                                    }}2(a),(c),ECI MCNP6 produces flux results that are on a per source particle basis and part of converting to final reported results involves establishing the source intensity. The total fission neutron source intensity S (neutrons/second) in the NPM at a given power is determined by Equation 3-2:

10 ( )

Eq. 3-2 1.602 x 10 ( ) where,

          = Average number of neutrons produced per fission in NuScale module (neutrons/fission); calculated from results in the MCNP6 output file to be =2.46 at initial cycle for a fresh core with 3.5 percent U235 enrichment at hot zero power, P = Fission power (MW); taken to be 160 MW based on NPMs thermal power rating, Keff = Effective multiplication factor; taken to be 1.000 for critical light water reactor, and

Fluence Calculation Methodology and Results TR-0116-20781-NP Rev. 1 Qave = The average recoverable energy per fission for all fissionable materials (MeV/fission); taken to be 198 MeV/fission as a best estimate based on other low enriched uranium systems. The calculated fission neutron intensity for the NPM is estimated as: 2.46 160 10 19 neutrons

                   =                                             = 1.24 x 10 second 1.602 x 10            1.000  198 A factor of 1.8 x 109 seconds (57 effective full power years) is then used to convert from flux to fluence based on a 60-year operating life with a 95 percent power capacity factor.

3.6 Other Modeling Considerations There is no upper limit placed on the neutron source energy and neutrons are treated with implicit capture in the NuScale best estimate fluence model. A lower cut off energy of 0.9 MeV is utilized. Since there are no processes modeled that would result in a higher energy neutron, the implementation of the 0.9 MeV lower cut off energy makes no difference to the >1 MeV neutron fluence results. A series of cylindrical mesh tallies and surface tallies are used to specify the locations of interest where fluence is calculated throughout the MCNP6 model. Example illustrations of mesh tallies used in the calculation of RPV and CNV fluence are shown in Figures 3-5 through Figure 3-7, including naming and numbering conventions for the axial and azimuthal segments. The effect of the tally region volumes impact on final fluence results is discussed in Section B.1.14. ((

                                                                                        }}2(a),(c)

Fluence Calculation Methodology and Results TR-0116-20781-NP Rev. 1 ((

                                                                                                          }}2(a),(c)

Figure 3-5 Vertical cross-sectional view of the reactor pressure vessel

  © Copyright 2019 by NuScale Power, LLC 15

Fluence Calculation Methodology and Results TR-0116-20781-NP Rev. 1 ((

                                                                                                               }}2(a),(c)

Figure 3-6 Horizontal cross-sectional view of the reactor pressure vessel mesh tally

  © Copyright 2019 by NuScale Power, LLC 16

Fluence Calculation Methodology and Results TR-0116-20781-NP Rev. 1 ((

                                                                                                             }}2(a),(c)

Figure 3-7 Horizontal cross-sectional view of the containment vessel mesh tally

  © Copyright 2019 by NuScale Power, LLC 17

Fluence Calculation Methodology and Results TR-0116-20781-NP Rev. 1 4.0 Bias and Uncertainty 4.1 Quantified Biases and Uncertainties Appendix A describes the NuScale best estimate fluence prediction benchmarking work. Appendix B describes sensitivity analysis associated with the best estimate fluence calculation. A summary of the relevant results associated with the NuScale best estimate fluence bias and uncertainty, and a reference to the applicable report section, is provided in Table 4-1. Table 4-1 List of quantified systematic biases (+ or -) and random uncertainties(+/-) ((

                                                                                                     }}2(a),(c)

Fluence Calculation Methodology and Results TR-0116-20781-NP Rev. 1 ((

                                                                                                   }}2(a),(c) 4.2      Combination of Biases The analytical bias (also known as  per RG 1.190 terminology) is comprised of all known uncertainties that are biased in a certain direction compared to the best estimate fluence calculation. For the NuScale best estimate fluence calculation,  is calculated as the algebraic summation of all systematic biases presented in Table 4-1, excluding
         , as shown in Equation 4-1.
                                    =  +    +     +                                         Eq. 4-1 A tendency for NuScales MCNP6 based fluence calculation methodology to ((
                                                            }}2(a),(c).

Fluence Calculation Methodology and Results TR-0116-20781-NP Rev. 1 (( = + = +2.82% + 1.84% Eq. 4-2 11.51% + 1.72% + 3.03% = }}2(a),(c) 4.3 Combination of Uncertainties Independent random uncertainties are all uncertainties that have no specific direction associated with them with respect to their effect on the final fluence estimate. The overall uncertainty ( ) is established per Equation 4-3 for the NuScale best estimate fluence MCNP6 model.

                                              =    ( )                                                 Eq. 4-3 Where  is the square root of the sum of the squares of all random uncertainties in Table 4-1, as shown in Equation 4-4.
  =          + + + +                +    +    +       +      +       +      +                Eq. 4-4

(( =

                                                                          =           }}2(a),(c)

Substituting the value established for back into Equation 4-3 gives Equation 4-5. Equation 4-5 is used to establish overall uncertainties given in Table 5-1. (( = (17.63%) }}2(a),(c) Eq. 4-5 Note that a single ((

                        }}2(a),(c). See Section B.1.11 for more details.

Fluence Calculation Methodology and Results TR-0116-20781-NP Rev. 1 5.0 NuScale Power Module Fluence Prediction Results Table 5-1 presents the results of the MCNP6 NuScale best estimate fluence model. These results are of the highest fluence found on the particular component after comparing results across various axial and radial locations. To obtain the Best Estimate Neutron Fluence the total bias ( ) of (( }}2(a),(c), established in Section 4.2, must be applied to the MCNP6 Calculated Neutron Fluence results shown in Table 5-1. Having thus derived the "Best Estimate Neutron Fluence", the overall uncertainty (( }}2(a),(c) established in Section 4.3, should also be accounted for. The five representative statistical tests used to ensure tally convergence discussed in RG 1.190 were satisfied for the results in Table 5-1. Table 5-1 Best estimate of fluence expected to be experienced in various NuScale Power Module components and locations ((

                                                                                                       }}2(a),(c),ECI
  © Copyright 2019 by NuScale Power, LLC 21

Fluence Calculation Methodology and Results TR-0116-20781-NP Rev. 1 ((

                                                                       }}2(a),(c),ECI
  © Copyright 2019 by NuScale Power, LLC 22

Fluence Calculation Methodology and Results TR-0116-20781-NP Rev. 1 ((

                                                                         }}2(a),(c),ECI
  © Copyright 2019 by NuScale Power, LLC 23

Fluence Calculation Methodology and Results TR-0116-20781-NP Rev. 1 ((

                                                                                                                           }}2(a),(c),ECI (1) Results from Table 5-1 have not yet applied the total bias and overall uncertainty established in Sections 4.2 and 4.3. To translate the MCNP6 calculated neutron fluence estimates to the best estimate neutron fluence expected in the NuScale Power Module, BT and  should be accounted for.
  © Copyright 2019 by NuScale Power, LLC 24

Fluence Calculation Methodology and Results TR-0116-20781-NP Rev. 1 6.0 Summary and Conclusions A best estimate neutron fluence calculation for the NPM was performed through the use of the MCNP6 code based on RG 1.190. Alternatives to particular RG 1.190 Regulatory Positions are provided in Appendix C. Biases and uncertainties associated with the MCNP6 best estimate neutron fluence model are reported in Table 4-1, which were established through benchmarking against the VENUS-3 experiment and NPM-specific sensitivity studies associated with key MCNP6 modeling simplifications and inputs. The peak RPV beltline cladding surface and CNV beltline at 1/4-T fluence over a 60-year NPM operating life (assumed 95 percent capacity factor) was calculated to be ((

                                                    }}2(a),(c),ECI as reported in Table 5-1. Neutron fluence estimates provided in this report are acceptable for supporting Final Safety Analysis Report Section 4.3 of the NuScale DCA and meet the regulatory guidance and requirements discussed in Section 2.1 of this report.

Fluence Calculation Methodology and Results TR-0116-20781-NP Rev. 1 7.0 References 7.1 U.S. Nuclear Regulatory Commission, Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence, Regulatory Guide 1.190, March 2001. 7.2 D.I. Poston and H.R. Trellue, "User's Manual, Version 2.0 for Monteburns, Version 5B," Los Alamos National Laboratory, 1999. 7.3 Radiation Safety Information Computational Center Oak Ridge National Laboratory, "Shielding Integral Benchmark and Database," DCL-237, SINBAD-2013.12, Version December 2013. 7.4 NEA Nuclear Science Committee, Prediction of Neutron Embrittlement in the Reactor Pressure Vessel: VENUS-1 and VENUS-3 Benchmarks, OECD, 2000. 7.5 NuScale Power, LLC, "Pressure and Temperature Limits Methodology Topical Report," TR-1015-18177, Revision 2. © Copyright 2019 by NuScale Power, LLC 26

Fluence Calculation Methodology and Results TR-0116-20781-NP Rev. 1 Appendix A. Benchmarking Monte Carlo N-Particle Transport Code 6 for Fluence Applications A.1 Vulcain Experimental Nuclear Study 3 Benchmark This appendix presents a description of benchmarking work performed to demonstrate that MCNP6 can perform neutron flux determinations that compare favorably with expected or experimental results. The benchmarking work shown in this appendix is also used to establish the bias and uncertainty stemming from use of the MCNP6 transport code and associated cross section data. A.1.1 Modeling MCNP6 code version 1.0 was used to create a model of the third configuration in the Vulcain Experimental Nuclear Study, commonly known as VENUS-3. The VENUS-3 pressure vessel fluence benchmark is based on documentation from the Shielding Integral Benchmark Archive and Database from the Radiation Safety Information Computational Center (Reference 7.3). The VENUS-3 benchmark provides reaction rates associated with various detector types for the core barrel of an experimental reactor setup. The VENUS-3 benchmark is considered to be generally applicable to the NPM. The basic configuration of the VENUS-3 benchmark is shown in Figure A-1. © Copyright 2019 by NuScale Power, LLC 27

Fluence Calculation Methodology and Results TR-0116-20781-NP Rev. 1 Figure A-1 Horizontal cross-sectional view of the Vulcain Experimental Nuclear Study 3 benchmark geometry The MCNP6 model is based on the MCNP model supplied as part of the VENUS-3 benchmark collection in Reference 7.3, which used an earlier version of MCNP. This model was reviewed for correctness and updated as needed for use with the current MCNP version MCNP6. The ENDF/B-VII.1 libraries associated with 293.6 degrees K (.80c extension) were used for all materials. In addition, a light water S(,) library based on the © Copyright 2019 by NuScale Power, LLC 28

Fluence Calculation Methodology and Results TR-0116-20781-NP Rev. 1 ENDF/B-VII.1, lwtr.20t, is used for those materials containing water. The benchmark used a U235 Watt fission spectrum. Portions of the NuScale MCNP6 model of the VENUS-3 benchmark are shown in Figure A-2 and Figure A-3. Figure A-2 Vertical cross-sectional view of the Monte Carlo N-Particle Transport Code 6 model of the Vulcain Experimental Nuclear Study 3 benchmark © Copyright 2019 by NuScale Power, LLC 29

Fluence Calculation Methodology and Results TR-0116-20781-NP Rev. 1 Figure A-3 Horizontal cross-sectional view of the inner and outer baffle of the Monte Carlo N-Particle Transport Code 6 model of the Vulcain Experimental Nuclear Study 3 benchmark A variety of experimental results were provided as part of the VENUS-3 collection of data, but the results of specific interest to this benchmark are the results associated with the core barrel only. These results are based on nickel, indium, and aluminum reaction rates Ni58(n,p), In115(n,n), and Al27(n,), respectively. © Copyright 2019 by NuScale Power, LLC 30

Fluence Calculation Methodology and Results TR-0116-20781-NP Rev. 1 Based on the energy thresholds associated with the reaction rates, the In115(n,n) reaction rates are associated with the neutron flux greater than 1 MeV, the Ni58(n,p) reaction rates are associated with neutron fluxes greater than 3 MeV, and the Al27(n,) reaction rates are associated with neutron fluxes greater than 8 MeV. The relative experimental uncertainties for the reaction rates in the core barrel for the VENUS-3 data were reported to be 9 percent for Ni58(n,p), 7 percent for In115(n,n), and 14 percent for Al27(n,) in Section 6.1 of Reference 7.4. The relative difference between the reported experimental (Exp) values for these reaction rates and the MCNP6 calculated values (Calc) was established for each data point provided in the VENUS-3 benchmark, relative to the experimental value, using Equation A-1.The average relative difference of all experimental versus calculated values and standard deviations are reported in Table A-1. (%) = Eq. A-1 The In115(n,n) reaction rate comparisons were judged to provide the best comparison to the overall neutron flux since it has the lowest threshold energy of ~1 MeV. The Ni58(n,p) and 27Al(n,) reaction rates have higher thresholds, 3 MeV and 8 MeV, respectively. The In115(n,n) results also have the lowest experimental uncertainty associated with them. Further, the In115(n,n) results are the only results from the NuScale VENUS-3 benchmark that indicate MCNP6 has a tendency to ((

                                                               }}2(a),(c) compared to incorporating the Ni58(n,p)  or Al27(n,)   based benchmark results.

((

                                                                                                     }}2(a),(c)

Fluence Calculation Methodology and Results TR-0116-20781-NP Rev. 1 Table A-1 Vulcain Experimental Nuclear Study 3 experimental versus calculated results ((

                                                                                              }}2(a),(c)

((

                                                                                    }}2(a),(c)

The results of this benchmark demonstrate that MCNP6 can perform neutron flux determinations that compare favorably with expected or experimental results. The results show good agreement between MCNP6 and the benchmark results. © Copyright 2019 by NuScale Power, LLC 32

Fluence Calculation Methodology and Results TR-0116-20781-NP Rev. 1 Appendix B. NuScale Power Module Fluence Prediction Sensitivity Studies and Uncertainty Analysis This appendix presents sensitivity studies and an uncertainty analysis associated with the NPM fluence prediction calculations. This will be combined with Appendix A findings in Section 4.0 of this report in order to properly present results with total uncertainty in Section 5.0 of this report. B.1 Sensitivity Studies B.1.1 Homogenized Fuel Model vs Explicit Fuel Model The best estimate fluence predictions presented in Table 5-1 were based on a homogenized fuel model. ((

                                                        }}2(a),(c).

B.1.2 Contribution of Pu239 to Neutron Source As discussed in Section 3.3, the MCNP6 NuScale best estimate fluence model does not contain any plutonium because it is based on a fresh core. ((

                                                                                                       }}2(a),(c)

Fluence Calculation Methodology and Results TR-0116-20781-NP Rev. 1 ((

                                                                          }}2(a),(c)

Fluence Calculation Methodology and Results TR-0116-20781-NP Rev. 1 ((

                     }}2(a),(c)

Fluence Calculation Methodology and Results TR-0116-20781-NP Rev. 1 ((

                                                            }}2(a),(c)

Fluence Calculation Methodology and Results TR-0116-20781-NP Rev. 1 B.1.3 Material Composition The uncertainty in fluence estimates associated with differences between the as built and operating NPM material chemical compositions and densities compared to how these characteristics were modeled in the NuScale best estimate fluence model is assumed to be ((

                                                          }}2(a),(c)

B.1.4 Geometrical Tolerances The uncertainty in fluence estimates associated with differences between as built and operating NPM dimensions and dimensions modeled in the NuScale best estimate fluence model is assumed to be ((

                                                          }}2(a),(c)

B.1.5 Assembly Averaged Neutron Source Bias and Uncertainty The MCNP6 NuScale best estimate fluence model uses an assembly averaged pin power profile instead of an explicit pin-wise power profile. ((

                                                                                            }}2(a),(c)

Fluence Calculation Methodology and Results TR-0116-20781-NP Rev. 1 ((

                                      }}2(a),(c)

Fluence Calculation Methodology and Results TR-0116-20781-NP Rev. 1 ((

                               }}2(a),(c)

Fluence Calculation Methodology and Results TR-0116-20781-NP Rev. 1 Table B-1 The averaged fast neutron flux in pin lattice of fuel assembly G3 from SIMULATE5, cycle 8 (unit: x1013 n/cm2 sec) ((

                                                                                                      }}2(a),(c),ECI
  © Copyright 2019 by NuScale Power, LLC 40

Fluence Calculation Methodology and Results TR-0116-20781-NP Rev. 1 Table B-2 The averaged fast neutron flux in pin lattice of fuel assembly G4 from SIMULATE5, cycle 8 (unit: x1013 n/cm2 sec) ((

                                                                                                     }}2(a),(c),ECI
  © Copyright 2019 by NuScale Power, LLC 41

Fluence Calculation Methodology and Results TR-0116-20781-NP Rev. 1 Table B-3 The averaged fast neutron flux in pin lattice of fuel assembly F6 from SIMULATE5, cycle 8 (unit: x1013 n/cm2 sec) ((

                                                                                                    }}2(a),(c),ECI
  © Copyright 2019 by NuScale Power, LLC 42

Fluence Calculation Methodology and Results TR-0116-20781-NP Rev. 1 Table B-4 The averaged fast neutron flux in pin lattice of fuel assembly E7 from SIMULATE5, cycle 8 (unit: x1013 n/cm2 sec) ((

                                                                                                      }}2(a),(c),ECI
  © Copyright 2019 by NuScale Power, LLC 43

Fluence Calculation Methodology and Results TR-0116-20781-NP Rev. 1 Table B-5 The averaged fast neutron flux in pin lattice of fuel assembly G3 from MCNP6, cycle 8 (unit: x1013 n/cm2 sec) ((

                                                                                                      }}2(a),(c),ECI
  © Copyright 2019 by NuScale Power, LLC 44

Fluence Calculation Methodology and Results TR-0116-20781-NP Rev. 1 Table B-6 The averaged fast neutron flux in pin lattice of fuel assembly G4 from MCNP6, cycle 8 (unit: x1013 n/cm2 sec) ((

                                                                                                     }}2(a),(c),ECI
  © Copyright 2019 by NuScale Power, LLC 45

Fluence Calculation Methodology and Results TR-0116-20781-NP Rev. 1 Table B-7 The averaged fast neutron flux in pin lattice of fuel assembly F6 from MCNP6, cycle 8 (unit: x1013 n/cm2 sec) ((

                                                                                                      }}2(a),(c),ECI
  © Copyright 2019 by NuScale Power, LLC 46

Fluence Calculation Methodology and Results TR-0116-20781-NP Rev. 1 Table B-8 The averaged fast neutron flux in pin lattice of fuel assembly E7 from MCNP6, cycle 8 (unit: x1013 n/cm2 sec) ((

                                                                                                                   }}2(a),(c),ECI B.1.6      Core Power The uncertainty of the core power level is directly proportional to the uncertainty of the fluence estimates. ((
                                                                                                   }}2(a),(c)

B.1.7 Radial Power Profile Uncertainty in the radial power profile is directly proportional to the uncertainty of the fluence estimates. The radial power profile uncertainty ( is estimated by ((

                                                                                                        }}2(a),(c)
  © Copyright 2019 by NuScale Power, LLC 47

Fluence Calculation Methodology and Results TR-0116-20781-NP Rev. 1 ((

                     }}2(a),(c)

((

                                                                                             }}2(a),(c),ECI Figure B-1 Time Weighted Averages and Standard Deviations for Radial Power Profile
© Copyright 2019 by NuScale Power, LLC 48

Fluence Calculation Methodology and Results TR-0116-20781-NP Rev. 1 B.1.8 Axial Power Profile A single, time averaged axial profile was utilized in the MCNP6 NuScale best estimate fluence model. Variations in the axial power profile could impact fluence estimates. ((

                                                                                      }}2(a),(c)

Fluence Calculation Methodology and Results TR-0116-20781-NP Rev. 1 ((

                          }}2(a),(c)

Table B-9 Average Axial Power Profiles ((

                                                                             }}2(a),(c),ECI
  © Copyright 2019 by NuScale Power, LLC 50

Fluence Calculation Methodology and Results TR-0116-20781-NP Rev. 1 Table B-10 Variance and Standard Deviation for the Axial Power Profiles ((

                                                                                                     }}2(a),(c)
  © Copyright 2019 by NuScale Power, LLC 51

Fluence Calculation Methodology and Results TR-0116-20781-NP Rev. 1 B.1.9 Boron Concentration The best estimate fluence prediction MCNP6 model assumed a boron concentration of ((

                            }}2(a),(c)

The concentration of soluble boron in the primary coolant will vary over the course of the fuel cycle, ((

                                                              }}2(a),(c)

B.1.10 Nuclear Cross Section Data and Transport Code There is uncertainty associated with the various cross sections taken from the ENDF/B-VII.1 nuclear data library and there is uncertainty associated with the use of the transport code MCNP6. ((

                                                                                     }}2(a),(c).

B.1.11 Monte Carlo Method In Monte Carlo analysis, a calculational uncertainty ( ) is introduced as a result of the finite number of particle histories sampled. ((

                                       }}2(a),(c)

B.1.12 Water Density ((

                                                                                                 }}2(a),(c)

Fluence Calculation Methodology and Results TR-0116-20781-NP Rev. 1 ((

                                                                   }}2(a),(c).

B.1.13 Axial Coolant Density Bias The coolant in the MCNP6 NuScale best estimate fluence model was modeled as ((

                                                                                              }}2(a),(c)

Fluence Calculation Methodology and Results TR-0116-20781-NP Rev. 1 ((

                                      }}2(a),(c)

Fluence Calculation Methodology and Results TR-0116-20781-NP Rev. 1 Table B-11 Coolant Water Axial Density Variations ((

                                                                                                }}2(a),(c)

((

                                 }}2(a),(c)
  © Copyright 2019 by NuScale Power, LLC 55

Fluence Calculation Methodology and Results TR-0116-20781-NP Rev. 1 ((

                                                                                          }}2(a),(c)

Table B-12 Peak Fluence Results for Axially Varied Coolant Density ((

                                                                                                 }}2(a),(c)

B.1.14 Tally Mesh Size An MCNP6 model was created to evaluate the uncertainty in the peak neutron fluence values due to decreasing the size of the tally subdivision used for determining the neutron fluence values on the belt line surfaces of the RPV and CNV. This section presents the results from this model and the determination of the uncertainty, . ((

                               }}2(a),(c)

Fluence Calculation Methodology and Results TR-0116-20781-NP Rev. 1 ((

                                                                                                       }}2(a),(c)

Figure B-2 Max neutron fluence versus tally degree midpoint for RPV inner clad surface

  © Copyright 2019 by NuScale Power, LLC 57

Fluence Calculation Methodology and Results TR-0116-20781-NP Rev. 1 ((

                                                                                                       }}2(a),(c)

Figure B-3 Max neutron fluence versus tally degree midpoint for RPV 0-T surface

  © Copyright 2019 by NuScale Power, LLC 58

Fluence Calculation Methodology and Results TR-0116-20781-NP Rev. 1 ((

                                                                                                       }}2(a),(c)

Figure B-4 Max neutron fluence versus tally degree midpoint for CNV 0-T surface

  © Copyright 2019 by NuScale Power, LLC 59

Fluence Calculation Methodology and Results TR-0116-20781-NP Rev. 1 Table B-13 Peak Fluence Results for smaller tally subdivisions ((

                                                                                                  }}2(a),(c)

Fluence Calculation Methodology and Results TR-0116-20781-NP Rev. 1 Appendix C. Regulatory Guide 1.190 Alternatives Regulatory Guide (RG) 1.190, Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence provides guidance for calculating pressure vessel neutron fluence. The NuScale fluence calculation methodology described in this report utilized some alternative approaches to those recommended in RG 1.190. This appendix describes and justifies these alternatives in Table C-1. The descriptions in Table C-1 are summaries or excerpts of specific portions of particular regulatory positions of RG 1.190. Table C-1 Alternatives to particular Regulatory Guide 1.190 regulatory positions RG 1.190 Description of Regulatory Position Description of Alternative and Justification Regulatory Position 1.1.1 Regional temperatures should be All materials in the NuScale best estimate included in the input data. fluence model are taken to be at ((

                                                                                                }}2(a),(c) The effect of the latter was accounted for in Section B.1.13.

1.1.1 and In the absence of plant-specific Uncertainty between the as built and operating 1.4.1 information, conservative estimates of and as modeled design was accounted for (( the variations in the material compositions and dimensions should be made and accounted for in the }}2(a),(c) as discussed in Sections B.1.3 determination of the fluence and B.1.4. uncertainty. 1.1.1 The input data should account for (( axial and radial variations in water density. }}2(a),(c) The effect of this modeling simplification is accounted for in Section B.1.13. © Copyright 2019 by NuScale Power, LLC 61

Fluence Calculation Methodology and Results TR-0116-20781-NP Rev. 1 RG 1.190 Description of Regulatory Position Description of Alternative and Justification Regulatory Position 1.2 The peripheral assemblies, which Assembly averaged power profiles obtained from contribute the most to the vessel core depletion calculations were used in the fluence, have strong radial power MCNP6 NuScale best estimate fluence model. A gradients, and these gradients should sensitivity study to establish the effect of this not be neglected. Peripheral modeling simplification on the NuScale fluence assembly pin-wise neutron source estimates is discussed in Section B.1.5. distributions obtained from core depletion calculations should be used. 1.3.2 The bias introduced by the neutron The MCNP6 NuScale best estimate fluence energy cutoff technique should be model implements a cutoff energy threshold of estimated by comparison with an 0.9 MeV. An additional study involving an unbiased calculation. MCNP6 model without a cutoff energy threshold is unnecessary. Since there are no processes modeled that would result in a higher energy neutron, the use of a 0.9 MeV cutoff energy threshold makes no difference to the >1 MeV fluence results. 1.3.3 The capsule fluence is extremely (( sensitive to the representation of the capsule geometry and internal water region (if present), and the adequacy of the capsule representation and mesh must be demonstrated using }}2(a),(c) sensitivity calculations. 1.4.2 The fluence calculation methods must The pressure vessel simulator benchmark be validated against (1) operating VENUS-3 is used to validate the NuScale fluence reactor measurements or both, (2) a calculation methodology (see Appendix A). The pressure vessel simulator benchmark, VENUS-3 benchmark results are adequate to and (3) the fluence calculation validate the NuScale fluence calculation benchmark. methodology. © Copyright 2019 by NuScale Power, LLC 62

LO-0619-66076 : Affidavit of Zackary W. Rad, AF-0619-66077 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 Office 541.360-0500 Fax 541.207.3928 www.nuscalepower.com

NuScale Power, LLC AFFIDAVIT of Zackary W. Rad I, Zackary W. Rad, state as follows: (1) I am the Director of Regulatory Affairs of NuScale Power, LLC (NuScale), and as such, I have been specifically delegated the function of reviewing the information described in this Affidavit that NuScale seeks to have withheld from public disclosure, and am authorized to apply for its withholding on behalf of NuScale (2) I am knowledgeable of the criteria and procedures used by NuScale in designating information as a trade secret, privileged, or as confidential commercial or financial information. This request to withhold information from public disclosure is driven by one or more of the following: (a) The information requested to be withheld reveals distinguishing aspects of a process (or component, structure, tool, method, etc.) whose use by NuScale competitors, without a license from NuScale, would constitute a competitive economic disadvantage to NuScale. (b) The information requested to be withheld consists of supporting data, including test data, relative to a process (or component, structure, tool, method, etc.), and the application of the data secures a competitive economic advantage, as described more fully in paragraph 3 of this Affidavit. (c) Use by a competitor of the information requested to be withheld would reduce the competitors expenditure of resources, or improve its competitive position, in the design, manufacture, shipment, installation, assurance of quality, or licensing of a similar product. (d) The information requested to be withheld reveals cost or price information, production capabilities, budget levels, or commercial strategies of NuScale. (e) The information requested to be withheld consists of patentable ideas. (3) Public disclosure of the information sought to be withheld is likely to cause substantial harm to NuScales competitive position and foreclose or reduce the availability of profit-making opportunities. The accompanying report reveals distinguishing aspects about the process and method by which NuScale develops its fluence calculation methodology and results. NuScale has performed significant research and evaluation to develop a basis for this process and method and has invested significant resources, including the expenditure of a considerable sum of money. The precise financial value of the information is difficult to quantify, but it is a key element of the design basis for a NuScale plant and, therefore, has substantial value to NuScale. If the information were disclosed to the public, NuScale's competitors would have access to the information without purchasing the right to use it or having been required to undertake a similar expenditure of resources. Such disclosure would constitute a misappropriation of NuScale's intellectual property, and would deprive NuScale of the opportunity to exercise its competitive advantage to seek an adequate return on its investment. (4) The information sought to be withheld is in the enclosed report entitled Fluence Calculation Methodology and Results. The enclosure contains the designation Proprietary" at the top of each page containing proprietary information. The information considered by NuScale to be proprietary is identified within double braces, "(( }}" in the document. (5) The basis for proposing that the information be withheld is that NuScale treats the information as a trade secret, privileged, or as confidential commercial or financial information. NuScale relies upon AF-0619-66077 Page 1 of 2

the exemption from disclosure set forth in the Freedom of Information Act ("FOIA"), 5 USC § 552(b)(4), as well as exemptions applicable to the NRC under 10 CFR §§ 2.390(a)(4) and 9.17(a)(4). (6) Pursuant to the provisions set forth in 10 CFR § 2.390(b)(4), the following is provided for consideration by the Commission in determining whether the information sought to be withheld from public disclosure should be withheld: (a) The information sought to be withheld is owned and has been held in confidence by NuScale. (b) The information is of a sort customarily held in confidence by NuScale and, to the best of my knowledge and belief, consistently has been held in confidence by NuScale. The procedure for approval of external release of such information typically requires review by the staff manager, project manager, chief technology officer or other equivalent authority, or the manager of the cognizant marketing function (or his delegate), for technical content, competitive effect, and determination of the accuracy of the proprietary designation. Disclosures outside NuScale are limited to regulatory bodies, customers and potential customers and their agents, suppliers, licensees, and others with a legitimate need for the information, and then only in accordance with appropriate regulatory provisions or contractual agreements to maintain confidentiality. (c) The information is being transmitted to and received by the NRC in confidence. (d) No public disclosure of the information has been made, and it is not available in public sources. All disclosures to third parties, including any required transmittals to NRC, have been made, or must be made, pursuant to regulatory provisions or contractual agreements that provide for maintenance of the information in confidence. (e) Public disclosure of the information is likely to cause substantial harm to the competitive position of NuScale, taking into account the value of the information to NuScale, the amount of effort and money expended by NuScale in developing the information, and the difficulty others would have in acquiring or duplicating the information. The information sought to be withheld is part of NuScale's technology that provides NuScale with a competitive advantage over other firms in the industry. NuScale has invested significant human and financial capital in developing this technology and NuScale believes it would be difficult for others to duplicate the technology without access to the information sought to be withheld. I declare under penalty of perjury that the foregoing is true and correct. Executed on July 2, 2019. Zackary W. Rad AF-0619-66077 Page 2 of 2}}

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