ML24079A245

NRC Non-Light Water Reactor (Non-LWR) Vision and Strategy Verification and Validation (V&V) of the Comprehensive Reactor Analysis Bundle (Bluecrab)

ML24079A245
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Issue date:	03/20/2024
From:	Stephen Bajorek NRC/RES/DSA/CRABII
To:	Kimberly Webber NRC/RES/DSA
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0 NRC Non-Light Water Reactor (Non-LWR)

Vision and Strategy Verification and Validation (V&V) of the Comprehensive Reactor Analysis Bundle (BlueCRAB)

Rev. 23, March 19, 2024

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2 Table of Contents LIST OF FIGURES.................................................................................................................................... 3 LIST OF TABLES...................................................................................................................................... 4 NOMENCLATURE......................................................................................................................... 5 EXECUTIVE

SUMMARY

................................................................................................................ 9 ACKNOWLEDGEMENT............................................................................................................... 11

1.

Introduction................................................................................................................................... 13 1.1 References................................................................................................................................... 14

2.

The Comprehensive Reactor Analysis Bundle.............................................................................. 14 2.1 References................................................................................................................................... 16

3.

Verification.................................................................................................................................... 18 3.1 Verification for Coupled Applications............................................................................................ 19 3.2 References................................................................................................................................... 20

4.

Evaluation Model Development.................................................................................................... 23 4.1 Advanced Reactor Design Types................................................................................................. 25 4.2 Figures-of-Merit............................................................................................................................. 26 4.3 Event Scenarios........................................................................................................................... 26 4.4 Phenomena and PIRTs................................................................................................................. 27 4.6 References................................................................................................................................... 32

5.

Gas-Cooled Reactor Validation.................................................................................................... 35 5.1 References................................................................................................................................... 37

6.

Liquid Metal Cooled Reactor Validation........................................................................................ 43 6.1 References................................................................................................................................... 45

7.

Molten Salt Reactor Validation...................................................................................................... 49 7.1 References................................................................................................................................... 51

8.

Microreactor Validation................................................................................................................. 53 8.1 References................................................................................................................................... 55

9.

General Neutronics....................................................................................................................... 56 9.1 References................................................................................................................................... 58

10.

Component Validation.................................................................................................................. 62 10.1 References................................................................................................................................... 63

11.

Summary and Conclusions........................................................................................................... 66

12.

Appendix...................................................................................................................................... 71

3 LIST OF FIGURES Figure 1 The Comprehensive Reactor Analysis Bundle (BlueCRAB) for Analysis of Design Basis Events in non-LWRs......................................................................................... 16 Figure 2 Elements of the Evaluation Model Development and Assessment Process............................ 24

4 LIST OF TABLES Table 1 Verification of the BlueCRAB Codes............................................................................................ 19 Table 2 Verification Coupling Tests............................................................................................................ 20 Table 4 Characterization of non-LWR Designs.......................................................................................... 25 Table 5 Validation for Gas-Cooled Reactors............................................................................................... 35 Table 6 Gas-Cooled Reactor Benchmark Studies....................................................................................... 37 Table 7 Validation for Liquid Metal Reactors............................................................................................. 43 Table 8 Liquid Metal Reactor Benchmarks................................................................................................. 44 Table 9 Validation for Molten Salt Reactors............................................................................................... 49 Table 10 Benchmark Studies for Molten Salt Designs............................................................................... 50 Table 11 Validation for Microreactors........................................................................................................ 54 Table 12 Microreactor Benchmarks............................................................................................................ 54 Table 13 Validation of Neutronics for Advanced Non-LWRs.................................................................... 56 Table 14 Code to Code Verification of Neutronics for Advanced Non-LWRs........................................... 57 Table 15 Component Validation Tests........................................................................................................ 62

5 NOMENCLATURE Abbreviation Definition ABR Advanced Burner Reactor ANL Argonne National Laboratories ATR Advanced Test Reactor AVR Arbeitsgemeinschaft Versuchsreaktor BEPU Best Estimate Plus Uncertainty CEFR Chinese Experimental Fast Reactor CFR Code of Federal Regulations CIET Compact Integral Effects Test CRAB Comprehensive Reactor Analysis Bundle D-LOFC Depressurized Loss of Forced Cooling DBE Design Basis Event DOE U.S. Department of Energy DUFF Demonstration Using Flattop Fissions ECCS Emergency Core Cooling System EBR Experimental Breeder Reactor EM Evaluation Model EMDAP Evaluation Model Development and Assessment Process FAST Fuel Analysis under Steady-State & Transient FFTF Fast Flux Test Facility FIPD Fuels Irradiation & Physics Database GCR Gas-Cooled Reactor gPBR general Pebble Bed Reactor HELIOS Heavy Eutectic liquid metal Loop for Integral test of Operability and Safety HPR Heat Pipe Reactor HTR High Temperature Gas-cooled Reactor Test Module HTTR High Temperature Test Reactor HTTU High Temperature Test Unit IAEA International Atomic Energy Agency INL Idaho National Laboratory

6 Abbreviation Definition JAEA Japanese Atomic Energy Agency KRUSTY Kilopower Reactor Using Stirling Technology LACANES Lead Alloy-Cooled Advanced Nuclear Energy Systems LBE Lead-Bismuth Eutectic LMR Lead Cooled Modular Reactor LOFC Loss of Forced Cooling LTDF Low Temperature DRACS Facility LWR Light Water Reactor MHTGR Modular High Temperature Gas-Cooled Reactor MiGaDome Michigan Multi-Jet Gas-Mixture Dome MOOSE Multiphysics Object Oriented Simulation Environment MOX Mixed Oxide Fuel MCNP Monte Carlo N-Particle Transport Code MCSR Molten Chloride Salt Reactor MFSR Molten Fluoride Slat Reactor MSPR Molten Salt Pebble Bed Reactor MSR Molten Salt Reactor MSRE Molten Salt Reactor Experiment MSTDB Molten Salt Thermophysical Properties Database NEAMS Nuclear Energy Advanced Modeling and Simulation NGNP Next Generation Nuclear Plant NRC U.S. Nuclear Regulatory Commission NSTF Natural convection Shutdown heat removal Test Facility ORNL Oak Ridge National Laboratories PBHTX Pebble-Bed Heat Transfer Experiment P-LOFC Pressurized Loss of Forced Cooling PARCS Purdue Advanced Reactor Core Solver RCCS Reactor Cavity Cooling System RPI Rensselaer Polytechnic Institute SAM System Analysis Module SANA Selbsttatige Abfuhr der Nachwarme SCALE Standardized Computer Analyses for Licensing Evaluation

7 Abbreviation Definition SFR Sodium Fast Reactor SiMBA Simplified Microreactor Benchmark Assessment SQA Software Quality Assurance TAMU Texas A&M University THETA Thermal Hydraulic Experimental Test Article THTR Thorium High Temperature Reactor TRACE TRAC/RELAP Advanced Computational Engine TREAT Transient Test Reactor TRISO Tri-isotropic U/PLOF Unprotected/Protected Loss of Flow U/PLOHS Unprotected/Protected Loss of Heat Sink UM University of Michigan UP Upper Plenum UTK University of Tennessee - Knoxville UW University of Wisconsin VCS Vessel Cooling System VHTRC Very High-Temperature Reactor Critical VTB Virtual Test Bed V&V Verification and Validation ZPPR Zero Power Physics Reactor

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9 EXECUTIVE

SUMMARY

Verification and validation represent important components of Evaluation Model development.

An Evaluation Model is defined as the calculational framework for simulating the behavior of a reactor system during a postulated transient or design basis accident. As such, an Evaluation model can be composed of one or more computer codes and all other information needed to apply the calculational framework. Two important steps in development of an Evaluation Model are an assessment of the accuracy of calculational framework and following an appropriate quality assurance protocol during the development process. This is more generally referred to as verification and validation (V&V). These two code development functions qualify the code(s) involved for their intended applications and help quantify the accuracy of the Evaluation Model.

Regulators world-wide place significant emphasis on V&V during the licensing process. It is of particular importance for new and unique features affecting the behavior of a reactor system during hypothetical accidents and in evaluation of the performance of safety systems.

Evaluation Models used for new reactor designs and relatively untested safety features can be expected to be carefully examined and questioned during a review. Providing a clear and comprehensive V&V for an Evaluation Model is essential for an efficient regulatory review and approval.

Advanced non-light water reactors represent a new challenge to regulators and the nuclear industry. Compared to light-water reactor (LWR) analysis, there have been fewer analytical studies of these new designs and also fewer large scale integral tests to demonstrate their performance during hypothetical accidents. Evaluation Models for advanced non-LWRs must be assessed in order to develop confidence in the simulations and the large safety margins that are expected for these designs.

This report discusses the code suite proposed for non-light water reactor (non-LWR) design basis confirmatory analysis and the recommended verification and validation requirements. The (Federal) Comprehensive Reactor Analysis Bundle (BlueCRAB) is intended to be capable of performing analysis of all non-LWR reactor designs that NRC anticipates for review and Design Certification. BlueCRAB is intended for steady-state and transient system analysis. It does not simulate core disruptive events, or accident-related geometric distortions. Severe accident and source term analysis will be performed with MELCOR. Because many advanced reactor designs are expected to be robust and have large safety margins, scenarios analyzed by BlueCRAB may involve multiple failures to better understand their operation and safety features.

10 This report focusses on the codes in BlueCRAB and the validation for several general design types.

The report identifies the V&V applicable to BlueCRAB and identifies tasks yet to be completed that could further expand upon the validation basis. Central to validation are the phenomena that must be simulated. Phenomena Identification and Ranking Table (PIRT) findings are discussed in order to identify validation needs.

This report is labelled as DRAFT because it represents a snapshot in time. It is meant to document what verification and validation is complete, and to highlight work that is planned.

Future updates of this report will document new validation cases. To make it easier to quickly observe completed work from planned work, references to documents pertaining to planned work are included but highlighted in yellow.

This report is also intended to capture the large number of test facilities and databases that contribute to the validation of non-LWR analysis codes. These are included in the report sections that address non-LWRs by general category. As new test facilities are developed and new tests are made available, they will be added in the relevant section.

While code assessment is continuing, a sufficient body of work was been completed to demonstrate that BlueCRAB is capable of simulating a broad range of non-LWRs. This is an important step that provides the NRC with an independent analysis capability to examine non-LWRs for accident scenarios including unprotected loss-of-flow, unprotected loss-of-heat sink in several designs, heat pipe failure in microreactors, and rapid reactivity insertions.

BlueCRAB has been demonstrated as capable of simulating depletion and pebble tracking which is needed to determine power distributions and steady-state core conditions in pebble bed reactors. In addition, the validation includes demonstration of the self-limiting features of some reactors as core power is reduced with increases in fuel temperature and thermomechanical expansion.

The extensive verification and validation of BlueCRAB provides the NRC with a set of analysis tools that can predict with reasonable accuracy peak fuel temperatures, vessel and component structure temperatures, reactivity induced power excursions, and system heat removal in a broad range of designs.

11 ACKNOWLEDGEMENT This report documents work performed by a large number of contributors from the NRC, Universities, and from the National Laboratories. The verification and validation as well as development of the analysis codes discussed in this report represent the cooperative efforts by numerous individuals. Assistance from Idaho National Laboratory, Argonne National Laboratory, and Oak Ridge National Laboratory are acknowledged and appreciated. The efforts by Emily Shemon (ANL) and constructive comments by Paolo Balestra (INL) were especially helpful in preparation of this report.

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13

1. Introduction Verification and Validation (V&V) is a vital component of the code development process. All codes contain models and correlations that are semi-empirical. That is, they contain correlations that are based on experimental data that is applicable to a limited range of conditions. As a result the correlations are themselves valid only over that specific range.

These semi-empirical correlations often do an adequate job of predicting phenomena over the range of conditions for which there were developed but may perform poorly if extrapolated to a new range of conditions or geometry. Therefore, the NRC has typically placed a high standard on V&V to ensure the accuracy of an analysis code. Regulatory Guide 1.203 [1-1] is an example of the rigor generally expected in an Evaluation Model.

The definitions of code assessment, verification, and validation are sometimes unclear and vague. In this report code assessment is used to represent the combined activities of verification and validation. The term verification is associated with software quality assurance requirements and ensures that the models and correlations are programmed in a code as intended. The term validation is associated with simulation of the physics and demonstrates that the code accurately represents the phenomena it is intended for. As discussed by Oberkampf and Roy [1-2], verification is solving the equations right, while validation is solving the right equations. One is mathematical in purpose, while the other is associated with the physical phenomena. Both are essential to development of a code and quantifying its accuracy.

Non-light water reactors (non-LWRs) represent a new challenge to regulators, applicants and the licensing process. Compared to light-water reactors, the experimental database for validation of non-LWRs is relatively small. Some technologies such as for sodium fast reactors, have more extensive databases than other non-LWR designs such as molten salt reactors.

Codes developed elsewhere for non-LWRs are less familiar to the NRC and other organizations, thus creating some skepticism in code capability and accuracy. Verification and validation by the NRC provide a means for staff to gain familiarity with the non-LWR designs and their operations.

This report is intended to define verification and validation needs for the BlueCRAB analysis codes to be used for non-LWR design basis event (DBE) safety analysis. The focus is on thermal-fluids and reactor neutronics, and as applicable, thermo-mechanical expansion.

Validation of fuel performance is not addressed directly, although BISON is used in some of the assessment cases to take advantage of its capability to simulate thermomechanical expansion.

The BISON and FAST code validation is not covered in this report, however both codes are expected to be used in BlueCRAB applications. BISON will be used for thermal expansion and thermal conduction in structures, rather than for fuel performance. Validation of fuel performance, severe accident analysis and consequence analysis is not discussed in this report.

14 1.1 References

[1-1] U.S. NUCLEAR REGULATORY COMMISSION, TRANSIENT AND ACCIDENT ANALYSIS METHODS, REGULATORY GUIDE 1.203, December 2005.

[1-2] Oberkampf W. and Roy, C., Verification and Validation in Scientific Computing, Cambridge Press, 2010.

2. The Comprehensive Reactor Analysis Bundle To reduce costs associated with the development of analytical capability for non-LWR safety analysis, the code suite for non-LWR confirmatory analysis makes use of existing NRC codes and integrates them with several codes developed through the DOE Nuclear Energy Advanced Modelling and Simulation (NEAMS) program. Reference [2-1] describes the approach, code development needs, and requirements for verification and validation. Figure 2-1 presents a schematic showing the full suite of computer codes, known as the (Federal) Comprehensive Reactor Analysis Bundle (BlueCRAB). In Figure 2-1 the NRC developed codes are shown in gold, while those produced by the DOE are shown in light blue. For each reactor design type, only a subset of the codes would be active as part of a given Evaluation Model.

Codes that are expected to play a role include the NRC developed or sponsored codes TRACE

[2-2], FAST [2-3], PARCS [2-4] and its associated SCALE code for cross sections [2-5]. Codes developed by DOE such as MOOSE [2-6], BISON [2-7], Pronghorn [2-8], SAM [2-9], and Griffin

[2-10] are expected to be utilized extensively. The expectation is that CFD analysis will be a necessary component of analysis and would be done using a commercially available code such as FLUENT [2-11] or possibly the DOE code Nek5000 0F1 [2-12]. (In this report Nek5000 and its next generation code called NekRS are used interchangeably. NekRS is designed to take advantage of GPU based high performance computing platforms.) SERPENT [2-13] is a reactor physics code developed at the VTT Technical Research Centre of Finland, capable of calculating cross-sections and performing detailed Monte Carlo simulations. While SERPENT is currently used for cross-sections and reference calculations, in the future it is anticipated that Shift [2-14] will eventually replace SERPENT. Shift is a DOE and NRC developed Monte Carlo code that can also be used for cross-section calculations and reference calculations.

Three systems level thermal-fluids codes are used: TRACE, Pronghorn, and SAM. The TRAC and RELAP Advanced Computational Engine (TRACE) is a systems thermal-hydraulic code developed by the NRC and is well-validated for transient two-phase flow. TRACE has the 1 In this report Nek5000 and its next generation code called NekRS are used interchangeably. NekRS is designed to take advantage of GPU based high performance computing platforms.

15 capability to simulate fluid flow and heat exchange in a thermal system and includes thermophysical properties for a variety of working fluids including helium, sodium and some molten salts. For non-LWR applications TRACE will most likely be used for simulation of secondary and safety systems where water is the working fluid. Pronghorn is a MOOSE-based multi-physics reactor analysis application developed at Idaho National Laboratory. Pronghorn is a coarse-mesh multiscale T/H solver that approximates the thermal and flow physics in a complex fluid-solid structure in terms of three length scales and can be used for gas-cooled, liquid-metal cooled, and molten salt reactors. The System Analysis Module (SAM) is a MOOSE-based system analysis tool being developed at Argonne National Laboratory for advanced non-LWR safety analysis. It can provide fast-running, whole-plant transient analyses capability for sodium fast reactors and other liquid metal designs. Both Pronghorn and SAM can also simulate some molten salt coolants.

Selection of the codes in BlueCRAB was based on flexibility and the ability to use one system of codes for most if not all non-LWR analysis [2-2, 2-15]. While BlueCRAB discussed in this report for non-LWR applications it is also a tool that can be used for LWR analysis. On the left-hand side of Figure 2-1 the TRACE/PARCS/SCALE codes are fully capable of LWR analysis.

Coupling with MOOSE also enables interaction with FAST and/or BISON to investigate scenarios when either of those codes are necessary for advanced fuel analysis.

Two fuel performance codes are integrated into BlueCRAB: BISON and FAST. Both codes offer considerable capability and at this time it is not possible to identify which may be best for a particular application. For many applications, BISON and/or FAST will be used to benchmark initial conditions in the fuel. During a transient fuel performance often involving fission gas release and fuel thermal conductivity is not likely to change considerably, and the fuel can be treated as a source of heat and stored energy. Coupling the other BlueCRAB codes with BISON and FAST may be useful, providing the thermal-fluid environment in fuel performance calculations.

Reactor kinetics will be accomplished using Griffin (previously called MAMMOTH) and SERPENT (or Shift), or by PARCS and SCALE in the case of LWRs. The NRCs traditional approach using PARCS and SCALE will be used for LWR designs where diffusion theory is sufficiently accurate; however for most non-LWR designs Griffin and SERPENT (or Shift) will be applied when transport theory is needed to obtain accurate power distributions.

Sockeye [2-16] is used for heat pipe performance. Its primary purpose is to provide a transient heat pipe simulation tool for microreactor applications. Sockeye provides the capability to perform one-dimensional, two-phase simulations of a heat pipe fluid along with a two-dimensional axisymmetric model for heat conduction in the heat pipe cladding.

16 Figure 1 The Comprehensive Reactor Analysis Bundle (BlueCRAB) for Analysis of Design Basis Events in non-LWRs.

2.1 References

[2-1] USNRC, NRC Vision and Strategy: Safely Achieving Effective and Efficient Non-Light Water Reactor Mission Readiness, ADAMS ML16356A670, December 2016.

[2-2] USNRC, TRACE V5.0 Developmental Assessment Manual, ADAMS ML120060208, ML120060187, ML120060191, ML120060172, 2008.

[2-3] K. Geelhood, D. Colameco, W. Luscher, L. Kyriazidis, C. Goodson, J. Corson and J.

Whitman, "FAST-1.1: A Computer Code for Thermal-Mechanical Nuclear Fuel Analysis under Steady-state and Transients," Pacific Northwest National Laboratory, Richland, WA (2022)

17

[2-4] T. Downar, V. Seker, and A. Ward, "PARCS: Purdue Advanced Reactor Core Simulator,"

Proc. Topl. Mtg. Advances in Nuclear Analysis and Simulation (PHYSOR-2006), Vancouver, Canada, September 10-14, 2006, American Nuclear Society (2006).

[2-5] B. Rearden and M. Jessee, "SCALE Code System, ORNL/TM-2005/39, Version 6.2.3,"

UT-Battelle, LLC, Oak Ridge National Laboratory, 2018.

[2-6] D. Gaston, C. Newman, G. Hansen, and D. Lebrun-Grandie. MOOSE: A parallel computational framework for coupled systems of nonlinear equations. Nuclear Engineering and Design, 239:1768-1778, 2009.

[2-7] R. L. Williamson, J. D. Hales, S. R. Novascone, G. Pastore, D. M. Perez, B. W. Spencer, and R. C. Martineau, Overview of the BISON multidimensional fuel performance code, IAEA Technical Meeting: Modeling of Water-Cooled Fuel Including Design-Basis and Severe Accidents, Chengdu, China, October 28-November 1, 2013.

[2-8] Idaho National Laboratory, Pronghorn Manual, July 2017.

[2-9] Rui Hu, SAM Theory Manual, Argonne National Laboratory, ANL/NE-17/4, March 2017.

[2-10] Derek R. Gaston, Cody J. Permann, John W. Peterson, Andrew E. Slaughter, David Andrs, Yaqi Wang, Michael P. Short, Danielle M. Perez, Michael R. Tonks, Javier Ortensi, Ling Zou, Richard C. Martineau, Physics-based multiscale coupling for full core nuclear reactor simulation, Annals of Nuclear Energy 84 (Oct. 2015), pp. 45-54.

[2-11] ANSYS, Inc., ANSYS FLUENT User's Guide, November 2011.

[2-12] Paul Fischer, et al., Nek5000 User Documentation, ANL/MCS-TM-351, 2015.

[2-13] Leppanen, J., "SERPENT - A Continuous Energy Monte Carlo Reactor Physics Burnup Calculation Code," VTT Technical Research Centre of Finland, 2015.

[2-14] T. M. Pandya, S. R. Johnson, T. M. Evans, G. G. Davidson, S. P. Hamilton, and A. T.

Godfrey, Implementation, Capabilities, and Benchmarking of Shift, a Massively Parallel Monte Carlo Radiation Transport Code. J Comp Phys, volume 308, pp. 239-272 (2016).

[2-15] Bajorek, S. M., The U.S. Nuclear Regulatory Commission Approach to Modeling and Simulation of Advanced Non-LWRs; Preparing for the Next Nuclear Renaissance, NURETH-18, Portland, OR, 2019.

[2-16] Joshua E. Hansel, Ray A. Berry, David Andrs, Matthias S. Kunick, and Richard C.

Martineau. Sockeye: a one-dimensional, two-phase, compressible flow heat pipe application. Nuclear Technology, 207(7):1096-1117, 2021.

18

3. Verification Code verification is the process that ensures software quality. For nuclear licensing applications, applicants are required to satisfy Title 10 of the Code of Federal Regulations (CFR), Part 50, Appendix B (10 CFR 50) [3-1], and are responsible for establishing and executing the quality assurance program. Appendix B specifies that design control measures shall provide for verifying or checking the adequacy of design, including items such as the following: reactor physics, stress, thermal, hydraulic, and accident analyses; compatibility of materials; accessibility for in-service inspection, maintenance, and repair; and delineation of acceptance criteria for inspections and tests.

Developers of codes for nuclear analysis thus produce and follow Software Quality Assurance (SQA) guidelines for their organization. These SQA guidelines provide the requirements for modification and revision of codes and outline the steps necessary to document that any change to the code is implemented as intended. At the Nuclear Regulatory Commission (NRC),

the procedures for SQA are documented in NUREG-1737 [3-2]. Central to satisfying the SQA guidelines are unit testing and integration testing, which involve specific tests to examine parts of the software. These tests demonstrate that small portions of the software perform their intended functions. To accomplish this in a complex code such as TRACE, regression testing is applied. In regression testing, tests from new and existing verification are collected and re-run to ensure that recent revisions have not had unintended consequences. Regression testing compares output of a code to output from prior versions and helps to minimize the introduction of errors due to new coding.

The NEAMS program defines V&V for modeling and simulation closely following the definitions used by the Department of Defense (DoD) [3-3], the American Institute of Aeronautics and Astronautics (AIAA) [3-4] and the American Society of Mechanical Engineers (ASME) [3-5].

For NEAMS codes, the Department of Energy (DOE) requires Quality Rigor Level 1 for computational models that support nuclear system design. The objective is to ensure processes are in place so that they are NQA-1 compliant and thus meet or exceed the requirements of 10CFR50, Appendix B. Under Quality Rigor Level 1, codes are required to have daily regression testing, automated error tracking, configuration management and associated documentation.

Verification following either the NRC SQA or the NEAMS NQA-1 standards can be compared by the extent of the regression testing, and the coverage of the regression cases. Coverage is defined as the percentage of code components that the regression tests address.

For the codes that are part of the BlueCRAB, the following table characterizes the regression testing and coverage for each code.

19 Table 1 Verification of the BlueCRAB Codes Code Number of Regression Tests

Coverage, Frequency of Regression Testing NQA-1 Conformance

?

Reference(s)

MOOSE 3600+

>80%

Each update +

each integration Yes

[3-6]

TRACE 3144 Unknown Each update No

[3-7], [3-8]

SAM 900+

>80%

Each update +

each integration Yes

[3-9], [3-15]

Pronghorn 400+

>80%

Each update +

each integration Yes

[3-6]

PARCS 298

>75%

Nightly No

[3-2]

SCALE/Shift 3500+

TBD Each update +

each integration Yes

[3-10], [3-11]

Griffin 1100+

>80%

Each update +

each integration Yes

[3-6]

FAST 700+

Unknown

[3-12]

BISON 2500+

>80%

Each update +

each integration Yes

[3-6]

Nek5000 287

>80%

Each update +

each integration Yes

[3-13]

SERPENT Not available Not available Unknown Unknown

[3-14]

3.1 Verification for Coupled Applications One aspect of BlueCRAB that may be unique is the multi-physics coupling that can be enabled between two or more codes. To ensure that mass, momentum, and energy are conserved when various codes are used in a calculation in which they are coupled, it is prudent to verify the coupling. In the verification tests in Table 2 below, information passed from Code A to Code B demonstrates that the conserved quantities have been preserved.

Perhaps the more difficult of coupling tasks is when two thermal-fluid codes are joined and must pass information multidimensional regions to one-dimensional regions. This often occurs when a reactor vessel is connected to primary loop piping. While the vessel may require a complex nodalization, piping systems can use relatively simple one-dimensional representations. SAM and Pronghorn use what is termed an overlapping domain technique to ensure conservation.

20 The following table lists cases that should be established to verify conservation of mass, momentum, and energy. In the References that follow, many are listed as TBD because the report is currently incomplete or planned. These References that are not complete are highlighted in yellow to quickly identify them. In some cases the work is complete, but the documentation of was not yet identified. These are also highlighted.

Table 2 Verification Coupling Tests Code A Code B Status Ref.

BISON TRACE Complete

[3-16]

[3-17]

FAST TRACE In progress

[3-27]

SAM TRACE In progress

[3-28]

SAM Pronghorn Complete

[3-18]

[3-19]

PARCS TRACE Complete

[3-22]

Griffin Pronghorn Complete

[3-23]

Griffin SAM Complete

[3-24]

Griffin BISON Complete

[3-25]

Griffin FAST Planned

[3-26]

SAM Nek5000, Nek Complete

[3-20]

[3-21]

3.2 References

[3-1] 10CFR50, Appendix B, Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants.

[3-2] USNRC, Software Quality Assurance Procedures for NRC Thermal-Hydraulic Codes, NUREG-1737, ADAMS ML10170081, 2000.

[3-3] DoD Instruction 5000.61: Modeling and Simulation (M&S) Verification, Validation, and Accreditation (VV&A). 1996, Defense Modeling and Simulation Office, Office of the Director of Defense Research and Engineering.

[3-4] AIAA: Guide for the Verification and Validation of Computational Fluid Dynamics Simulations. 1998, American Institute of Aeronautics and Astronautics, AIAA-G-077, 1998.

[3-5] ASME: Guide for Verification and Validation of Computational Solid Mechanics. 2006, American Society of Mechanical Engineers, ASME V&V 10-2006.

21

[3-6] MOOSE Software Quality Assurance Plan (PLN-4005).

[3-7] TRACE Software Quality Assurance Plan, ML061940423

[3-8] LA-UR-96-1475, A Description of the Test Problems in the TRAC-P Standard Test Matrix.

ML072000463.

[3-9] SAM Software Quality Assurance Plan Rev. 1, SAM-SQAP-r1, Argonne National Laboratory, Dec. 2022.

[3-10] B. T. Rearden, M. T. Sieger, S. M. Bowman, and J. P. Lefebvre, Quality Assurance Plan for the SCALE Code System, SCALE-QAP-005, Rev 4, Oak Ridge National Laboratory (2013).

[3-11] B. T. Rearden, S. M. Bowman, and J. P. Lefebvre, Configuration Management Plan for the SCALE Code System, SCALE-CMP-001, Rev. 8, Oak Ridge National Laboratory (2013).

[3-12] David V. Colameco, Kenneth J. Geelhood, Ian Porter, and Lucas Kyriazidis, PNNL-28767, Software Quality Assurance Plan for the FAST Code System, Pacific Northwest National Laboratory, June 2019.

[3-13] A. Novak, L. Ibarra, and D. Shaver, Software Quality Assurance, Software Requirements and Gap Analysis for the MOOSE-Based Open-Source Multiphysics Code Cardinal, Tech. Rep.

ANL/NSE-23/83, Argonne National Laboratory, September 2023.

[3-14] TBD.

[3-15] R. Hu et al., SAM Assessment Report, Argonne National Laboratory, ANL/NSE-23/93, Dec. 2023.

[3-16] R. L. Williamson, G. Pastore, R. J. Gardner, K. A. Gamble, S. Novascone, J. Tompkins, W. Liu, LOCA Challenge Problem Final Report, CASL-U-2019-1856-000, 2019.

[3-17] Gardner, Russell J, Permann, Cody J, Bernard, Matthew, and Williamson, Richard L.

Demonstration of Bison/TRACE Coupling (CRAB) Through Validation Case LOFT L2-5. United States, INL/CON-19-52847, 2019.

22

[3-18] R. Hu, D. Nunez, G. Hu, L. Zou, G. Giudicelli, D. Andrs, S. Schunert, Development of an integrated system-and engineering-scale thermal fluids analysis capability based on SAM and Pronghorn, ANL/NSE-21/36, July 2021.

[3-19] Schunert, S., Tano Retamales, M. E., & Mohammad Jaradat, M. K. (2023). Overlapping Domain Coupling of Multidimensional and System Codes in NEAMS-Pronghorn and SAM (No.

INL/RPT-23-72874-Rev000). Idaho National Laboratory (INL), Idaho Falls, ID (United States).

[3-20] Huxford, A., Coppo Leite, V., Merzari, E., Zou, L., Petrov, V., Manera, A., 2022.

Development of Innovative Overlapping-domain Coupling between SAM and NekRS. The 19th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-19),

Virtual/Online.

[3-21] Huxford, Aaron, Victor Coppo Leite, Elia Merzari, Ling Zou, Victor Petrov, and Annalisa Manera. A hybrid domain overlapping method for coupling System Thermal Hydraulics and CFD codes. Annals of Nuclear Energy 189 (2023): 109842.

Work is complete, but appropriate reference report not identified.

Work is complete, but appropriate reference report not identified.

Work is complete, but appropriate reference report not identified.

Work is complete, but appropriate reference report not identified.

TBD.

TBD.

TBD.

[3-28]

[3-27]

[3-26]

[3-25]

[3-24]

[3-23]

[3-22]

23

4. Evaluation Model Development Code validation is the process by which a code, or codes, are demonstrated to have sufficient accuracy for simulation of a given scenario. To establish such code adequacy, the process described in Regulatory Guide 1.203 [4-1] is often applied. The Regulatory Guide 1.203 approach is general and can be applied to a very broad range of scenarios, both steady-state and transient. Central to the process is the concept of an Evaluation Model. An Evaluation Model is defined as the calculational framework for simulating the behavior of a reactor system during a postulated transient or design basis accident. As such, an Evaluation model can be composed of one or more computer codes and all other information needed to apply the calculational framework. The Evaluation Model includes all computer codes and assumptions that form the basis for modeling and simulating a transient including peripheral codes that may be used to establish initial conditions or process output.

Regulatory Guide 1.203 outlines the Evaluation Model Development and Assessment Process (EMDAP) and provides a step-by-step guide to developing and assessing code(s) for a complex analysis. The elements of the EMDAP are shown in Figure 4-1.

24 Figure 2 Elements of the Evaluation Model Development and Assessment Process.

The first element of EMDAP involves establishing requirements for the Evaluation Model. In particular, the developer should:

1. Specify the purpose of the analysis, and the type of power plant.

2. Specify the figures of merit.

3. Identify systems, components, and geometries to be modeled.

4. Identify and rank phenomena and processes.

The next few subsections discuss these requirements and their relationship to validation needs.

It is important to understand that one validates an Evaluation Model for a specific type of analysis, and not validate a code for generic applications.

25 4.1 Advanced Reactor Design Types There is a large variety of advanced non-LWR designs under consideration, and specifics of those designs are currently proprietary or simply unavailable because the designs continue to evolve. It is possible however to characterize several generic design types that share common features. Table 4 lists these generic design types that are considered for the purposes of this report. The designs include gas-cooled, liquid metal cooled and molten salt reactor concepts. Molten salt reactors can be either fuel salt or salt-cooled, and both are included in Table 4. Both thermal spectrum and fast spectrum core designs must be considered, and fuel can be TRISO, metallic, or dissolved in a molten salt.

Table 3 Characterization of non-LWR Designs Plant Type Description Expected Fuel Type HTGR High Temperature Gas Cooled (HTGR);

prismatic core, thermal spectrum TRISO (rods or plates)

PBMR Pebble Bed Modular Reactor (PBMR);

pebble bed core, thermal spectrum TRISO (pebbles)

GCFR Gas Cooled Fast Reactor (GCFR);

prismatic core, fast spectrum SIC clad UC (plates)

SFR Sodium Cooled Fast Reactor (SFR);

sodium cooled, fast spectrum Metallic (U-10Zr)

LMR Lead Cooled Modular Reactor (LMR); lead (or lead-bismuth) cooled, fast spectrum Nitride or MOX HPR Heat Pipe Cooled Reactor (HPR); heat pipe cooled micro reactor, fast spectrum Metallic (U-10Zr)

MSR Molten Salt Cooled Reactor (MSR);

prismatic core, thermal spectrum TRISO (plates)

MSPR Molten Salt Cooled Pebble Bed Reactor (MSPR); pebble bed core, thermal spectrum TRISO (pebbles)

MFSR Molten Fluoride Salt Reactor (MFSR);

fluoride fuel salt, thermal spectrum Fuel salt (liquid)

MCSR Molten Chloride Salt Reactor (MCSR);

chloride fuel salt, fast spectrum Fuel salt (liquid)

26 4.2 Figures-of-Merit The figures of merit (FOM) for advanced reactor designs are yet to be defined. The NRC has published the Advanced Reactor Design Criteria [4-2], which provide the general requirements for safety criteria. Specific FOMS, such as maximum fuel temperature, reactor power following a reactivity insertion, cooldown rate or reactor vessel temperature have not been determined to date. For the purposes of this report, it is assumed that it will be necessary to model steady-state and transient behavior of a reactor system so that the power and temperature distributions in the core are accurately predicted. In addition, it is assumed that coolant flows and temperatures in the reactor vessel and loop(s) are accurately predicted especially in the upper and lower plena where control rod drives and vessel integrity may be a concern. In the case of molten fuel salt designs, it is assumed that the distribution flow of neutron precursors must be simulated. Outside of the reactor vessel performance of reactor cavity cooling systems and decay removal systems must be modeled as these represent important heat rejection methods.

4.3 Event Scenarios Event scenarios for non-LWRs will differ from traditional NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition, Chapter 15 and Chapter 19 events due to their unique characteristics and the expected large safety margins. The new designs are likely to be able to withstand multiple failures and/or require fewer safety significant components (SSCs) in order to mitigate an accident. The analysis capability being defined in this paper is intended to simulate events in non-LWRs up to those conditions that lead to core disruption and release of fission products. While a scenario involving multiple failures is traditionally considered a beyond design basis event, for non-LWRs they may be acceptable as part of the design basis and will be analyzed by codes proposed in this report.

The event scenarios that the codes for BlueCRAB analysis will be used for include but are not limited to the following hypothetical events in gas-cooled, liquid metal cooled, molten salt, and micro reactors:

Gas-Cooled Reactors pressurized loss of force cooling (P-LOFC) accident de-pressurized loss of force cooling (D-LOFC) accident reactivity-induced transients, including ATWS events (Events that involve air-ingress and significant oxidation of the graphite, water-ingress, transport and release of graphite dust will be simulated with a traditional severe accident code.)

Liquid Metal Reactors loss of coolant with and without scram

27 loss of forced flow unprotected loss of flow unprotected loss of heat sink reactivity-induced transients, including ATWS events Molten Salt Reactors loss of forced flow unprotected loss of flow inadvertent reactivity insertion transients, including ATWS events loss of coolant over-cooling events (leading to partial solidification) station blackout loss of heat sink Heat Pipe Cooled Micro Reactors loss of heat sink inadvertent reactivity insertion transients, including ATWS events localized heat pipe failure cascading loss of heat pipes seismic event (causing reactivity increase) events related to coupling the reactor to the power conversion unit monolith temperature and stress under normal operating conditions monolith temperature and stress under postulated accident conditions.

(The monolith is the structural component that supports the fuel and control rods.)

4.4 Phenomena and PIRTs Table 4 provides a categorization of non-LWR designs as a means of identifying the Evaluation Models that will need to be developed for confirmatory analysis. While ten design types are listed, it is convenient for the purposes of discussing important phenomena to consider four overall design types: gas-cooled, liquid metal cooled, molten salt, and micro reactors cooled by heat pipes. Phenomena important to each design have been identified for each of these design types, and in some cases applicable PIRTs are available. This section briefly discusses the PIRTs and provides references used by the NRC to help establish code requirements for non-LWRs. A comprehensive review and discussion of the set of PIRTs mentioned below is beyond the scope of this report although a summary of key findings is provided at the end of the section.

Phenomena for gas-cooled reactors, both prismatic and pebble bed, were the subject of a comprehensive effort related to the Next Generation Nuclear Plant (NGNP). Expert panels considered five interrelated subject areas including thermal-fluids, neutronics and accident analysis, high temperature materials, nuclear grade graphite, process heat and cogeneration,

28 and fission product transport [4-3]. With regards to design basis event analysis, phenomena and processes of major significance due to their importance in analysis and relatively low knowledge level included:

core coolant bypass flows, power/flux profiles, outlet plenum flows, reactivity-temperature feedback coefficients, emissivity for the vessel and reactor cavity cooling system, reactor vessel cavity air circulation and heat transfer, and convection/radiation heating of upper vessel.

A separate PIRT was also developed for TRISO fuel [4-4], which discussed mechanisms and phenomena for fission product release and the evaluation of fuel performance. This report is relevant to design basis event analysis due to its identification of fission product release mechanisms and thus helps establish figures-of-merit that segregate design basis and beyond design basis analysis.

For sodium fast reactors, there has been significant efforts made to understand the phenomena and processes of importance to modeling and simulation of various accident scenarios. While not specifically a PIRT for design basis events, a recent study with many elements of a PIRT including a review of important phenomena was completed by a panel of experts [4-5]. The objective of the study was to evaluate the status of knowledge for accident analysis of sodium fast reactors (SFRs), and the panel considered a broad range of phenomena expected to be important in licensing SFR designs. Because of a fairly large quantity of experimental information, the panel was able to conclude that the knowledge level for SFRs was good and that there are no major technology gaps in preparing a safety case for an advanced SFR, so long as one stays with known technology. Potential knowledge gaps were acknowledged for advanced simulation of coupled neutronic/fluid flow dynamics, supercritical CO2 power conversion, and high minor-actinide content fuel. Reference [4-6] reports another PIRT in support of the Argonne National Laboratorys SAS4A thermal hydraulic code development.

The study identified several phenomena important to DBE analysis of SFRs, including:

single phase transient sodium flow thermal inertia pump-coast down sodium stratification transition to natural convection core cooling core flow decay heat generation reactivity due to mechanical changes in core structure reactivity feedback at high power

29 The NRC sponsored an additional review of phenomena that considered sodium fast reactors, in addition to lead-and lead-bismuth cooled systems [4-7]. Reference [4-8] provides a PIRT for a modern lead-cooled fast reactor design. The findings in [4-5] were confirmed, and largely extended to other liquid metal designs, resulting in a comprehensive list of phenomena and processes that must be modeled in analysis of a liquid metal reactor.

While important phenomena are well established for gas-cooled and liquid metal reactors, molten salt reactors represent a new and unique challenge. PIRTs applicable to these designs have however recently been developed. For a molten coolant salt design, a PIRT was developed considering two events: station blackout and a simultaneous withdrawal of control rods [4-9]. An accompanying study [4-10] considered phenomena involved in the molten salt reactor design and the need for a tight coupling between neutronics and fluid flow. Tracking of the neutron precursors is necessary. Two studies [4-9, 4-10] discussed physical phenomena and lack-of-knowledge for the particular design which utilized a fluoride salt and TRISO fuel in the form of plates. Phenomena of importance which were recognized as having low levels in knowledge base include:

thermophysical properties of coolant salt (conductivity and viscosity) wall friction in the core core flow asymmetry upper and lower plenum mixing safety system component performance chimney natural circulation and performance These studies were considered by the NRC, which then sponsored an expert panel to consider molten fuel salt reactors. This panel investigated and developed PIRTs for both fluoride (thermal spectrum) and chloride (fast spectrum) fuel salt reactors [4-11, 4-12, 4-13]. These studies determined that in addition to the phenomena in a coolant salt, there are two main challenges introduced with liquid fuel: delayed neutron precursor motion and strong coupling to salt composition. Hence, it is necessary to account for the movement of delayed neutron precursors into and out of the core, and the transit times of the fuel, fission products, and transmutation and chemistry products through the primary system. For a fuel salt reactor design, the phenomena of importance include:

delayed neutron precursor motion salt chemical composition neutron absorption in fuel salt physical properties convective heat transfer primary system flow resistances structural material performance (swelling and expansion) tritium production and transport (fluoride salts)

30 A major recommendation of one of the studies [4-11] is that modeling and simulation of MSRs will require development of computational tool(s) capable of tracking chemical inventories of constituents throughout the primary loop of the reactor facility. A comprehensive evaluation of fuel salt composition will likely require the modeling of salt chemistry, thermodynamics, mass transport, and addition and removal of chemical species.

Finally, for heat pipe cooled micro reactors, PIRTs have been developed [4-14, 4-15]. The PIRTs address reactor accident and normal operations, heat pipe design, materials, and power conversion. The PIRTs highlighted the importance of understanding:

monolith thermal stress single heat pipe failure machining and inspection of the monolith heat pipe performance reactivity and core criticality A PIRT has recently been developed for the Westinghouse eVinci microreactor [4-15].

Phenomena of importance included:

Radial thermal conductivity in graphite material in the core block Thermal conduction across the annular gap between the core block and the fuel.

Thermal conduction through the fuel compact Several phenomena affecting heat pipe performance Heat transfer and fluid flow in the primary heat exchanger References [4-3] to [4-15] provide a preliminary basis for modeling and simulation requirements for a non-LWR code suite. These PIRTs may need to be reconsidered as each of these reactor designs mature and new experimental information is made available.

The phenomena discussed in these PIRTs help to identify phenomena that are likely to require efforts to resolve in the development of a BlueCRAB analysis capability. As noted above, a comprehensive review of these entire PIRTs is beyond the scope of this report, yet there are some findings that are of particular importance with respect to code requirements and future code develop for non-LWRs. Phenomena that are significant and new, and with increased importance for non-LWRs relative to conventional LWRs, include, but are not limited to:

Thermal stratification and thermal striping Thermo-mechanical expansion and effect on reactivity Large neutron mean-free path length in fast reactors Transport of neutron pre-cursors (in fuel salt MSRs)

Solidification and plate-out (MSRs)

31 The BlueCRAB suite of codes that simulate non-LWRs need to account for these phenomena as well as many other phenomena that are important to design basis events. Development and assessment of these tools will certainly be necessary, but the basic code requirements defined in these PIRTs should be satisfied with BlueCRAB.

4.5 Scaling and Experimental Database Code validation for nuclear plant analysis often involves the question concerning scaling.

Because nuclear plants involve very high power and large dimensions, it is difficult or impossible to perform experiments at full-scale. Typically, sub-scale experiments are performed to investigate specific processes of importance so that models and correlations can be developed.

In some cases, larger scale integral tests are conducted to investigate the events that are expected in a full-scale prototype. Thus, the effects of using a scaled down experiment relative to the size of nuclear reactor or plan must be evaluated as recommended as part of EMDAP.

Challenges with code validation for non-LWRs are two-fold. First, there is a relative lack of experimental data (compared to light water reactors). Second, there is a very limited number of integral tests that simulate all of the components and processes that may occur in the proposed non-LWR designs. That is not to say that data for non-LWRs is non-existent. There is a large body of information, and code developers need to exploit this database.

While the database for non-LWRs is less than that for conventional LWRs, the EMDAP provides recommendations on how to deal with model biases and code accuracy. One of the steps of EMDAP involves the determination of model bias and uncertainties. An uncertainty analysis has the ultimate objective of providing a singular statement of uncertainty. That is, the uncertainty analysis has the objective of determining a figure-of-merit such as peak fuel temperature within a specific percentile and confidence level. However, this is accomplished when the individual uncertainty contributions of specific phenomena are determined. For phenomena with high uncertainty or if simulation models have questionable accuracy, NUREG-0800 includes guidance to use suitably conservative input parameters. Thus, weaknesses in the validation database can be addressed through uncertainty analysis and bounding assumptions.

An important step that remains is showing that the validation test data scales to the full-scale prototypes. Until full scale design details are made available, it is difficult to evaluate the completeness of the validation data currently available.

The following sections summarize the code validation important to each type of non-LWR design including gas-cooled reactors, liquid metal cooled reactors, molten salt reactors, and micro reactors. Section 5 discusses gas-cooled reactor validation, both prismatic and pebble bed. Section 6 covers liquid metal cooled reactors, including both sodium fast reactors and lead

32 cooled reactors. In Section 7 molten salt reactors are covered. Section 8 is dedicated to microreactors, primarily those cooled by heat pipes.

In Sections 5 through 8, the tables contain the notations that summarize and characterize information in the validation database. From left to right, the meaning of the column headers is as follows:

Information in the Test column identifies the facility involved.

The type of validation applicable to a given technical area are included in columns 2 through 5 where T is for thermal-hydraulics, F is for fuel performance, K is for reactor kinetics, and M is for mechanical expansion. Some tests provide validation for more than one technical area, although many focus on just a single physical process.

The Codes Involved column identifies which code or codes were applied.

The Type column refers to the experimental facility characterization as either separate effect test (SET) or integral effect test (IET). The category of integral effects tests includes operating reactors.

The Design Type column contains information about which of the major reactor designs the validation is applicable to.

The "Status column indicates which organization(s) are working on the validation including the Department of Energy and the NRC, and progress towards completion is designated as P for planned, C for complete and O for ongoing (i.e. in progress). The characterization as Complete in this table means a BlueCRAB input deck and initial calculations have been completed, although there is the possibility that the cases will eventually need to be re-run or improved.

The final column Validation Reference is meant to list the report(s) that document the validation results. The report(s) and the input decks should be publicly available, and preferably in the DOE Virtual Test Bed (VTB) [4-16]. Multiple references are acceptable (and encouraged) as improvements in the simulations can be made, and there can be several ways to model a particular test facility.

Because of the importance of neutronics to all designs, Section 9 is dedicated to validation of the Griffin code. In Section 10 validation for important system components is covered. Some of these components may be applicable to several design as well. Finally, summary and conclusions are contained in Section 11. The Appendix provides a short summary of the validation tests with references for more information.

4.6 References

[4-1] USNRC, Regulatory Guide 1.203, Transient and Accident Analysis Methods, ADAMS ML053500170, 2002.

[4-2] USNRC, GUIDANCE FOR DEVELOPING PRINCIPAL DESIGN CRITERIA FOR NON-LIGHT-WATER REACTORS, Regulatory Guide 1.232, ADAMS ML17375A611, April 2018.

33

[4-3] S.J. Ball and S.E. Fisher, Next Generation Nuclear Plant Phenomena Identification and Ranking Tables (PIRTs), ORNL/TM-2007/147, NUREG/CR-6944, Vol. 1, March 2008.

[4-4] USNRC, TRISO-Coated Particle Fuel Phenomenon Identification and Ranking Tables (PIRTs) for Fission Product Transport Due to Manufacturing, Operations, and Accidents, NUREG/CR-6844, 2004.

[4-5] R. Schmidt, et al., Sodium Fast Reactor Gap Analysis of Computer Codes and Models for Accident Analysis and Reactor Safety, SAND2011-4145, 2011.

[4-6] Ken-ichi Kawada, Ikken Sato, Yoshiharu Tobita, Werner Pfrang, Laurence Buffe, Emmanuelle Dufour, DEVELOPMENT OF PIRT (PHENOMENA IDENTIFICATION AND RANKING TABLE) FOR SAS-SFR (SAS4A) VALIDATION, ICONE22-30679, 2014.

[4-7] Lap-Yan Cheng, Michael Todosow, and David Diamond, Phenomena Important in Liquid Metal Reactor Simulations, BNL-207816-2018-INRE, ADAMS Accession No. ML18291B305 (2018).

[4-8] J. Liao, et al., "Development of Phenomena Identification and Ranking Table for Westinghouse Lead Fast Reactor's Safety," Progress in Nuclear Energy, 131, 2021.

[4-9] Hsun-Chia Lin, Sheng Zhang, David Diamond, Stephen Bajorek, Richard Christensen, Yujun Guo, David Holcomb, Graydon Yoder, Shanbin Shi, Qiuping L., Xiaodong Sun, Phenomena Identification and Ranking Table Study for Thermal Hydraulics for Advanced High Temperature Reactor, Annals of Nuclear Energy, 124 (2019) 257-269.

[4-10] Farzad Rahnema, Xiaodong Sun, Bojan Petrovic, David Diamond, Stephen Bajorek, Yujun Guo, Graydon Yoder, Dingkang Zhang, Paul Burke, Phenomena identification and categorization by the required level of multiphysics coupling in FHR modeling and simulation, Annals of Nuclear Energy, 121 (2018) 540-551.

[4-11] D. J. Diamond, N. R. Brown, R. Denning and S. Bajorek, Phenomena Important in Modeling and Simulation of Molten Salt Reactors, BNL-114869-2018-IR, Brookhaven National Laboratory, ADAMS ML18124A330, April 23, 2018.

[4-12] Diamond, D. J., Brown, N. R., Denning, R., Bajorek, S., 2018. Neutronics Phenomena Important in Modeling and Simulation of Liquid-Fuel Molten Salt Reactors, Advances in Thermal Hydraulics (ATH 2018); Nov. 11-15; Orlando, Florida, USA: American Nuclear Society.

[4-13] Bajorek, S., Diamond, D. J., Brown, N. R., Denning, R., 2018. Thermal-Hydraulics Phenomena Important in Modeling and Simulation of Liquid-Fuel Molten Salt Reactors,

34 Advances in Thermal Hydraulics (ATH 2018); Nov. 11-15; Orlando, Florida, USA: American Nuclear Society.

[4-14] J. W. Sterbentz, J. E. Werner, M. G. McKellar, A. J. Hummel, J. C. Kennedy, R. N.

Wright, J. M. Biersdorf, "Special Purpose Nuclear Reactor (5 MW) for Reliable Power at Remote Sites Assessment Report; Using Phenomena Identification and Ranking Tables (PIRTs),"

INL/EXT-16-40741, Revision 1, April 2017.

[4-15] Michael J. Patterson, Jun Liao, Megan E. Durse, William Brown, Richard F. Wright, "Development of Design Basis Safety Analysis Phenomena Identification and Ranking Table for the eVinciTM* Microreactor," Proc. of NURETH-20, Paper 40571, 2023.

[4-16] https://mooseframework.inl.gov/virtual_test_bed/

35

5. Gas-Cooled Reactor Validation This section discusses validation relevant to gas-cooled reactors with a focus on the primary system. The following table lists tests important to gas cooled reactors, including prismatic, pebble bed (thermal spectrum), and gas-cooled fast reactors. Codes that were part of the validation are listed in the table along with the design type the validation is most closely related to.

The validation in this table pertains to modeling and simulation of in-vessel phenomena and includes large scale integral test facilities and separate effects tests that are important to specific regions within the vessel. Some test facilities are listed more than once as there are multiple approaches to modeling and simulation, or the focus was on specific phenomena (neutronics as opposed to thermal-fluids for example). Validation for ex-vessel and specialized components are discussed in Section 10.

Validation exercises yet to be completed are highlighted in yellow. Those highlighted in light blue are references to models in the Virtual Test Bed (VTB) repository. These VTB models by themselves do not provide a comparison to data but are available to support other validation efforts.

Table 4 Validation for Gas-Cooled Reactors Test T

F K

M Code(s) Involved Type Design Type Status Validation Reference HTTR X

X Griffin, BISON, Pronghorn IET HTGR DOE-O

[5-1], [5-2]

[5-37], [5-38],

[5-39], [5-40]

HTTF X

SAM IET HTGR DOE-C

[5-3], [5-4],

[5-25],[5-26]

[5-28]

HTTF X

Pronghorn, Nek IET HTGR DOE-O

[5-34], [5-55]

HTR-10 X

Shift, Griffin IET PBMR DOE-C

[5-5], [5-32]

HTR-10 X

SAM IET PBMR DOE-P

[5-63]

HTR-PM X

X SAM, Pronghorn, BISON, Griffin IET PBMR DOE-O

[5-6], [5-41]

[5-42]

THTR-300 X

X SAM, Pronghorn, BISON, Griffin IET PBMR DOE-P

[5-7]

TAMU P X

SAM, Pronghorn SET PBMR MSPB DOE-C

[5-8], [5-33]

36 Missouri S&T Air Experiments X

SAM SET HTGR PBMR DOE-C

[5-8]

Purdue H-shaped helium loop X

SAM SET HTGR PBMR DOE-P

[5-21]

SANA X

Pronghorn SET PBMR MSPR DOE-C

[5-9], [5-16]

[5-17], [5-19]

[5-35], [5-36]

SANA X

SAM SET PBMR MSPB DOE-P

[5-53]

HTTU X

SAM SET PBMR DOE-C

[5-10]

[5-24]

HTTU X

Pronghorn SET PBMR DOE-C

[5-43]

HTR-PROTEUS X

Griffin SET PBMR DOE-P

[5-11]

ASTRA X

Griffin SET PBMR DOE-P

[5-12]

VHTRC X

Griffin SET HTGR DOE-P

[5-13]

TREAT X

X Griffin IET HTGR DOE-C

[5-14] [5-44]

AVR X

X SAM, Pronghorn, BISON, Griffin IET PBMR DOE-P

[5-15]

Fort Saint Vrain X

X SAM, Pronghorn, BISON, Griffin IET HTGR DOE-P

[5-59]

Peach Bottom #1 X

X SAM, Pronghorn, BISON, Griffin IET HTGR DOE-P

[5-59]

AGR Fuel X

X X

BISON, Griffin IET HTGR PBMR DOE-O

[5-18] [5-62]

[5-51] [5-52]

HTR-PM Hot Gas Mixing Experiment X

Pronghorn, Nek IET HTGR PBMR DOE-P

[5-61]

TAMU UP X

Nek SET PBMR DOE-C

[5-56]

MiGaDome X

Nek SET PBMR DOE-C

[5-56]

There are several code-to-code benchmark studies that have been conducted, often as part of international cooperative efforts. While these benchmark studies do not involve a comparison to experimental data, they exercise the analytical codes and provide some insight on performance of a reactor system. Of particular note are the MHTGR-350, a 350 MWt prismatic gas-cooled reactor design, and the PBMR-400 which is a 400 MWt pebble bed design. The gPBR-200 is another gas-cooled pebble bed representation that has been used to develop and test analytical tools. The following table lists benchmark studies relevant to gas-cooled reactors.

37 Table 5 Gas-Cooled Reactor Benchmark Studies Test T

F K

M Code(s) Involved Type Design Type Status Code-Code Verification Reference MHTGR-350 X

X SAM, Griffin IET HTGR DOE-C

[5-20], [5-27]

[5-45], [5-46]

[5-58]

gPBR-200 X

X SAM, Pronghorn, BISON Griffin, Shift IET PBMR DOE-O

[5-24] [5-29]

[5-30] [5-31]

[5-47] [5-48]

[5-49]

PBMR-400 X

X SAM, Pronghorn, BISON, Griffin, Shift IET PBMR DOE-C

[5-22] [5-23]

[5-32] [5-50]

5.1 References

[5-1] Laboure, Vincent M., et al. Multiphysics Steady-state simulation of the High Temperature Test Reactor with MAMMOTH, BISON and RELAP-7. No. INL/CON-18-52202-Rev002. Idaho National Laboratory (INL), Idaho Falls, ID (United States), 2019.

[5-2] Labouré, Vincent, et al. "Improved multiphysics model of the High Temperature Engineering Test Reactor for the simulation of loss-of-forced-cooling experiments." Annals of Nuclear Energy 189 (2023): 109838.

[5-3] T. Hua, L. Zou, and R. Hu, Code Benchmark of the HTTF Pressurized Conduction Cooldown Test Using SAM, Nuclear Science and Engineering, Volume 197, 2023 - Issue 10.

[5-4] J. Fang, T. Hua, H. Yuan, Z. J. Ooi, and L. Zou, CFD Simulations of HTTF Lower Plenum Flow Mixing Using Nek5000, ANS Winter Meeting 2022, November 13-17, 2022, Phoenix, AZ.

doi.org/10.13182/T127-39690

[5-5] Javier Ortensi, Sebastian Schunert, Yaqi Wang, Vincent Labour'e, Frederick Gleicher, Richard C. Martineau, Benchmark Analysis of the HTR-10 with the Griffin Reactor Physics Application, INL/EXT-18-45453, June 2018.

[5-6] Zhiee Jhia Ooi, Gang Yang, Travis Mui, Ling Zou, Rui Hu, Coupled SAM/Griffin Model of a Reference Pebble Bed High-Temperature Gas Cooled Reactor for Multi-Physics Simulations, ANL/NSE-23/43 Rev. 1, Argonne National Laboratory, 2023.

38

[5-7] TBD, Planned. (Simulation depends on recovering the data and consideration of a thorium fuel cycle.)

[5-8] L. Zou and R. Hu, SAM Code Validation on Frictional Pressure Drop through Pebble Beds, Argonne National Laboratory, ANL/NSE-20/5, March 2020.

[5-9] Ling Zou, April J. Novak, Richard C. Martineau, Hans D. Gougar, Validation of Pronghorn with the SANA Experiments, INL/EXT-17-44085, December 2017.

[5-10] L. Zou and R. Hu, Recent SAM Code Improvement to Heat Transfer Modeling Capabilities, Argonne National Laboratory, ANL/NSE-19/46, Dec. 2019.

[5-11] TBD, Planned. (As part of the Generation IV International Forum (GIF) - Project Arrangement on Computational Methods Validation and Benchmarks (CMVB) for The International Research and Development of the Very-High-Temperature Reactor Nuclear Energy System.)

[5-12] TBD, Planned. (Pending review of the currently available data required, and applicability for validation exercise.)

[5-13] TBD, Planned.

[5-14] Jing, Tian, et al. "Multiphysics Simulation of the NASA SIRIUS-CAL Fuel Experiment in the Transient Test Reactor Using Griffin." Energies 15.17 (2022): 6181.

[5-15] TBD, Planned. (Pending availability of data.)

[5-16] Zou, Ling, et al. Validation of Pronghorn with the SANA Experiments. No. INL/EXT 44085. Idaho National Laboratory (INL), Idaho Falls, ID (United States), 2017.

[5-17] Novak, A. J., et al. "Validation of Pronghorn friction-dominated porous media thermal-hydraulics model with the SANA experiments." Nuclear Engineering and Design 350 (2019):

182-194.

[5-18] Skerjanc, W. F., & Jiang, W. (2021). Bison As-run AGR-3/4 Irradiation Test Predictions.

Tech. Rep. INL/EXT-21-65160, INL, Idaho Falls, ID (United States), 11 2021.

[5-19] Novak, A. J., et al. Pronghorn: a porous media thermal-hydraulics core simulator and its validation with the SANA experiments. 2018.

[5-20] Prasad Vegendla, Rui Hu, Ling Zou, Multi-Scale Modeling of Thermal-Fluid Phenomena Related to Loss of Forced Circulation Transient in HTGRs, ANL-19/35, September 2019

39

[5-21] TBD, Planned.

[5-22] Balestra, Paolo, et al. Griffin/Pronghorn PBMR-400 Benchmark Results. No. INL/MIS 57522-Rev000. Idaho National Laboratory (INL), Idaho Falls, ID (United States), 2020.

[5-23] Z. Ooi, L. Zou, T. Hua, and R. Hu, "System Level Simulation of the PBMR-400 Core with SAM, ANS Winter Meeting, 2001.

[5-24] Z. Ooi, L. Zou, T. Hua, J. Fang, and R. Hu, System Level Modeling of a Generic Pebble Bed High Temperature Gas-Cooled Reactor (PB-HTGR) with SAM and Griffin, Proceedings of NURETH-20, (Summary #5217), 2023.

[5-25] Z. Ooi, T. Hua, L. Zou, and R. Hu, Simulation of the Hight Temperature Test Facility (HTTF) Core Using the 2D Ring Model with SAM, Nuclear Science and Engineering, 197:5, pp.

840-867, 2022.

[5-26] Thanh Hua, Robert Kile, Sung Nam Lee, Ling Zou, Aaron Epiney, Code Benchmark of Pressurized Conduction Cooldown Transient in the High Temperature Test Facility, 2024 International Congress on Advances in Nuclear Power Plants (ICAPP), submitted for review.

[5-27] VTB, SAM MHTGR Model and Results, https://mooseframework.inl.gov/virtual_test_bed/htgr/mhtgr_sam/

[5-28] VTB, 2D Ring Model for the High Temperature Test Facility (HTTF),

https://mooseframework.inl.gov/virtual_test_bed/htgr/httf/httf_sam_model.html

[5-29] VTB, SAM Generic PBR Model and Results, https://mooseframework.inl.gov/virtual_test_bed/htgr/generic-pbr/index.html

[5-30] Zhiee Jhia Ooi, Ling Zou, Thanh Hua, Jun Fang, Rui Hu, Modeling of a Generic Pebble Bed High-temperature Gas-cooled Reactor (PB-HTGR) with SAM, ANL/NSE-22/59, Argonne National Laboratory, September 2022.

[5-31] Zhiee Jhia Ooi, Ling Zou, Thanh Hua, Jun Fang, Rui Hu, Multi-Physics System-level Simulations of a Generic Pebble Bed High-Temperature Gas Cooled Reactor with Coupled SAM/Griffin Model, ANL/NSE-23/45, Argonne National Laboratory, July 2023.

[5-32] Pandya, Tara, et al. Modeling, Performance Assessment, and Nodal Data Analysis of TRISO-Fueled Systems with Shift. No. ORNL/TM-2022/2601. Oak Ridge National Laboratory (ORNL), Oak Ridge, TN (United States), 2022.

40

[5-33] Lee, Jieun, et al. "Validation of Pronghorn Pressure Drop Correlations Against Pebble Bed Experiments." Nuclear Technology 208.12 (2022): 1769-1805.

[5-34] Tano Retamales, Mauricio Eduardo, Vasileios Kyriakopoulos, and Sebastian Schunert. Modeling of Prismatic High Temperature Reactors in Pronghorn. No. INL/RPT 72860-Rev000. Idaho National

[5-35] Novak, A. J., et al. Pronghorn: a porous media thermal-hydraulics core simulator and its validation with the SANA experiments. 2018.

[5-36] Lee, Jieun, et al. "Pronghorn Fully Compressible Equation Set Validation Against SANA Open Plenum Experiments." Proc. HTR (2020).

[5-37] Strydom, Gerhard, Vincent M. Laboure, and Javier Ortensi. HTTR 3 D Cross Section Generation with Serpent and MAMMOTH. No. INL-EXT--18-51317-REV000. Idaho National Laboratory (INL), 2018.

[5-38] Labouré, Vincent M., et al. FY22 Status Report on the ART-GCR CMVB and CNWG International Collaborations. No. INL/RPT-22-68891-Rev000. Idaho National Laboratory (INL),

Idaho Falls, ID (United States), 2022.

[5-39] Laboure, Vincent M., et al. FY21 Status report on the CMVB and CNWG International Collaborations. No. INL/EXT-21-64241-Rev000. Idaho National Laboratory (INL), Idaho Falls, ID (United States), 2021.

[5-40] VTB, High Temperature Engineering Test Reactor (HTTR) Multiphysics Model, https://mooseframework.inl.gov/virtual_test_bed/htgr/httr/index.html

[5-41] Schunert, Sebastian, et al. Improvements in High Temperature Gas Cooled Reactor Modeling Capabilities in the Pronghorn Code. No. INL/RPT-22-69263-Rev000. Idaho National Laboratory (INL), Idaho Falls, ID (United States), 2022.

[5-42] Mohammad Jaradat, Mustafa Kamel, Sebastian Schunert, and Javier Ortensi. Modeling The DLOFC Accident Scenario of HTR-PM Equilibrium Core Using NEAMS Tools. No.

INL/CON-23-73279-Rev000. Idaho National Laboratory (INL), Idaho Falls, ID (United States),

2023.

[5-43] TBD, Planned. (Additional data are available from, Potgieter, M. C., and C. G. du Toit.

"Analysis of forced convection in the HTTU experiment using numerical codes." Nuclear Engineering and Technology (2023).)

41

[5-44] Ortensi, Javier, et al. "Validation of the Griffin application for TREAT transient modeling and simulation." Nuclear Engineering and Design 385 (2021): 111478.

[5-45] VTB, Griffin MHTGR Benchmark Model, Model Inputs, and Results https://mooseframework.inl.gov/virtual_test_bed/htgr/mhtgr_griffin/index.html

[5-46] Ortensi, Javier, and Gerhard Strydom. OECD/NEA Coupled Neutronic/Thermal-Fluids Benchmark of the MHTGR-350 MW Core Design: Results for Phase-I Exercise 1. No. INL/LTD-17-43099-Rev000. Idaho National Lab.(INL), Idaho Falls, ID (United States), 2020.

[5-47] Laboure, Vincent M., et al. Multiphysics Pebble-Bed Reactor Control Rod Withdrawal Study. No. INL/RPT-23-74341-Rev000. Idaho National Laboratory (INL), Idaho Falls, ID (United States), 2023.

[5-48] Prince, Zachary M. et al. "Sensitivity Analysis, Surrogate Modeling, and Optimization of Pebble-Bed Reactors Considering Normal and Accident Conditions", INL/RPT-23-74336, Idaho National Laboratory (INL), Idaho Falls, ID (United States), 2023.

[5-49] De Oliveira, Rodrigo et al. Verification of equilibrium core solution and decay heat modeling for the GPBR200, INL/RPT-23-03822 Idaho National Laboratory (INL), Idaho Falls, ID (United States), 2023.

[5-50] VTB, PBMR400 multiphysics numerical benchmark, https://mooseframework.inl.gov/virtual_test_bed/htgr/pbmr/index.html

[5-51] Jiang, Wen, et al. "TRISO particle fuel performance and failure analysis with BISON." Journal of Nuclear Materials 548 (2021): 152795.

[5-52] VTB, Bison TRISO model, https://mooseframework.inl.gov/virtual_test_bed/htgr/triso/index.html

[5-53] TBD.

[5-54] Yildiz, Mustafa, Botha, Gerrit, Yuan, Haomin, Merzari, Elia, Kurwitz, Richard, and Hassan, Yassin. Direct Numerical Simulation of the Flow through a Randomly Packed Pebble Bed. J. Fluids Eng. 142(4): 041405 (2020), https://doi.org/10.1115/1.4045439.

[5-55] T. Hua, J. Fang, L. Zou, "HTTF Benchmarking Activities in FY 2023", ANL/NSE-23/55, 2023, https://doi.org/10.2172/2004968

42

[5-56] V. Leite, E. Merzari, J. Mao, V. Petrov, A. Manera, "Computational Fluid Dynamic Analysis with NekRS of Mixing Phenomena in Large Enclosure", ATH22, Anaheim, CA, June 12-16, 2022

[5-57] J. Acierno, E. Merzari, "Large Eddy Simulation of Jet Interaction", ATH 22, Anaheim, CA, June 12-16, 2022

[5-58] Taiyang Zhang, Thanh Hua, Zhiee Jhia Ooi, Ling Zou and Caleb S. Brooks, "Modeling the Prismatic HTGR Core in SAM by Representative Channels," NURETH-20, Paper 5452, 2023.

[5-59] TBD, Planned. (Pending availability of data.)

[5-60] TBD, Planned. (Pending availability of data.)

[5-61] TBD, Planned. (As part of the Generation IV International Forum (GIF) - Project Arrangement on Computational Methods Validation and Benchmarks (CMVB) for The International Research and Development of the Very-High-Temperature Reactor Nuclear Energy System.)

[5-62] TBD, Planned. (As part of the Generation IV International Forum (GIF) - Project Arrangement on Computational Methods Validation and Benchmarks (CMVB) for The International Research and Development of the Very-High-Temperature Reactor Nuclear Energy System.)

43

6. Liquid Metal Cooled Reactor Validation This section discusses validation relevant to reactors cooled by liquid metals, with a focus on the primary system. Liquid metal cooled reactors include sodium-cooled, lead-cooled, and lead-bismuth cooled (fast) reactor designs. Validation for large scale integral test facilities and operating sodium fast reactors include tests in EBR-II and FFTF. Several tests investigate and provide validation data for fuel assembly heat transfer such as the Toshiba 37-pin and ORNL-19 pin test facilities.

Validation exercises yet to be completed are highlighted in yellow. Those highlighted in light blue are references to model in the Virtual Test Bed (VTB) repository. These VTB models by themselves do not provide a comparison to data but support other validation efforts.

The following table lists the important validation tests for liquid metal cooled reactors.

Table 6 Validation for Liquid Metal Reactors Test T

F K

M Code(s)

Involved Type Design Type(s)

Status Validation Reference EBR-II (SHRT)

X X

X SAM, Griffin, Pronghorn IET SFR DOE-C

[6-1],[6-15]

FFTF X

X SAM IET SFR DOE-C

[6-2], [6-33]

Phenix X

X X

SAM, Griffin IET SFR DOE-P

[6-3]

Monju X

X X

SAM, Griffin IET SFR DOE-P

[6-4]

HELIOS X

TRACE IET LMR NRC-C

[6-5]

TAMU-FA X

SAM SET SFR DOE-P

[6-6]

TAMU-FA X

Nek SET SFR DOE-C

[6-35], [6-36]

KIT-KALLA X

SAM SET LMR DOE-P

[6-7]

UW-Sodium X

SAM SET SFR DOE-O

[6-8]

UTK-Square Cavity X

SAM SET SFR DOE-C

[6-9]

CEA-Supercavna X

SAM SET SFR DOE-C

[6-10]

CEA-Supercavna X

Nek SET SFR DOE-O

[6-37]

ENEA-NACIE X

SAM SET SFR DOE-C

[6-11]

ENEA-NACIE X

Nek SET SFR DOE-O

[6-27]

ENEA-CIRCE X

SAM IET LMR DOE-P

[6-19]

KTH-TALL X

SAM SET LMR DOE-C

[6-12] [6-34]

KTH-TALL X

Nek SET LMR DOE-O

[6-40]

JAEA-PLANDT X

SAM, Nek SET SFR DOE-P

[6-13] [6-41]

44 Toshiba 37 pin X

Pronghorn (subchannel)

SET SFR DOE-C

[6-14] [6-16]

ORNL THORS 19 pin X

Pronghorn (subchannel)

SET SFR DOE-C

[6-14] [6-18]

ORNL THORS 61 pin X

Pronghorn (subchannel)

SET SFR DOE-O

[6-43] [6-42]

ORNL THORS X

Nek SET SFR DOE-O

[6-38]

PNL 7 Sleeve Blockage Benchmark X

Pronghorn (subchannel)

SET SFR DOE-C

[6-14]

GaTE X

SAM SET SFR,LMR DOE-C

[6-17]

ANL THETA X

SAM IET SFR DOE-P

[6-44]

EBR-II (X447)

X X

BISON IET SFR DOE-O

[6-28], [6-29]

EBR-II (X423)

X X

BISON IET SFR DOE-O

[6-29], [6-30]

IFR-FBTA X

X BISON SET SFR DOE-O

[6-31], [6-32]

IFR-WPF X

X BISON SET SFR DOE-O

[6-31]

ANL-MAX Nek SET LMR DOE-C

[6-39]

There are several code-to-code benchmark studies that have been conducted, often as part of international cooperative efforts. While these benchmark studies do not involve a comparison to experimental data, they exercise the analytical codes and provide some insight on performance of a reactor system. The following table lists benchmark studies relevant to liquid metal cooled reactors.

Table 7 Liquid Metal Reactor Benchmarks Test T

F K

M Code(s) Involved Design Type Status Code-Code Verification Reference IAEA Fast Reactor Working Group Verification Problems

[6-20]

SFR, LMR VP1 X

MOOSE / BISON SFR, LMR DOE-O

[6-21] [6-22]

[6-23]

VP3A X

MOOSE / BISON SFR, LMR DOE-O

[6-22], [6-24]

45

[6-25]

VP4 X

MOOSE / BISON SFR, LMR DOE-O

[6-26]

VP5 X

MOOSE / BISON SFR, LMR DOE-O

[6-26]

6.1 References

[6-1] Rui Hu and Tyler Sumner, Benchmark Simulations of the Thermal-Hydraulic Response During EBR-II Inherent Safety Tests Using SAM, Proceedings of the 2016 International Congress on Advances in Nuclear Power Plants, April 2016.

[6-2] Liu, Yang, Guojun Hu and Rui Hu. "BENCHMARK SIMULATION OF THE FFTF LOFWOS TEST #13 USING SAM." 19th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, March 6, 2022 - March 11, 2022.

[6-3] TBD, Planned.

[6-4] TBD, Planned.

[6-5] Wadim Jaeger, Victor Hugo Sanchez Espinoza, Improvements and Validation of the System Code TRACE for Lead and Lead-Alloy Cooled Fast Reactors Safety-Related Investigations, NUREG/IA-0421, 2013.

[6-6] TBD, Planned.

[6-7] TBD, Planned.

[6-8] Jeong, Yeongshin, Matthew Bucknor, Tyler Sumner, Daniel O'Grady, and Acacia Brunett, Assessment of Sodium Thermal Stratification Models Utilizing the TSTF Benchmark, Argonne National Laboratory, ANL/NSE-23/24, 2023.

[6-9] R. Hu, Three-Dimensional Flow Model Development for Thermal Mixing and Stratification Modeling in Reactor System Transient Analyses, Nuclear Engineering and Design, 345, 209-215 (2019).

[6-10] Ling Zou, Daniel Nunez and Rui Hu. "Development and Validation of SAM Multidimensional Flow Model for Thermal Mixing and Stratification Modeling." Argonne National Laboratory, ANL-NSE-20/19, June 2020.

46

[6-11] Victor Coppo Leite, Elia Merzari, Adam Dix and Ling Zou, Code Validation of SAM Using Forced and Natural-Circulation Data from NACIE-UP Benchmark, Proc. of NURETH-20, Paper 40261, 2023.

[6-12] J.V. De Kock, M.M. Stempniewicz, F. Roelofs, G. Hu, R. Hu, COMPARISON OF THE APPROACHES FOLLOWED TOWARDS MODELING OF THE TALL-3D FACILITY USING THE SYSTEMS THERMAL HYDRAULICS CODES SAM AND SPECTRA, Proceedings of NURETH-19, 2022.

[6-13] TBD, Planned.

[6-14] Kyriakopoulos, V., Tano, M. E., & Karahan, A. (2023). Demonstration of Pronghorns Subchannel Code Modeling of Liquid-Metal Reactors and Validation in Normal Operation Conditions and Blockage Scenarios. Energies, 16(6), 2592.

[6-15] Tano, M., Kyriakopoulos, V., McCay, J., & Arment, T. (2024). Validation of Pronghorns subchannel code using EBR-II shutdown heat removal tests: SHRT-17 and SHRT-45R. Nuclear Engineering and Design, 416, 112783.

[6-16] VTB, https://mooseframework.inl.gov/virtual_test_bed/sfr/subchannel/toshiba_37_pin/toshiba_37_pin.

html

[6-17] Molly Ross and Hitesh Bindra, VALIDATION OF SYSTEM ANALYSIS MODULE AXIAL MIXING MODEL FOR THERMAL STRATIFICATION AND MIXING APPLICATION IN THE UPPER PLENUM OF A LIQUID METAL REACTOR, Proceedings of NURETH-20, (Summary

1. 5218), 2023.

[6-18] VTB, https://mooseframework.inl.gov/virtual_test_bed/sfr/subchannel/ornl_19_pin/ornl_19_pin.html

[6-19] TBD, Planned.

[6-20] IAEA. Verification and validation of LMFBR static core mechanics codes part i. Technical Report IWGFR/75, International Atomic Energy Agency, 1990.

[6-21] Wozniak, Nicholas, Shemon, Emily, Grudzinski, James, and Spencer, Benjamin.

Assessment of MOOSE-Based Tools for Calculating Radial Core Expansion. United States:

ANL/NSE-21/30, 2021. Web. doi:10.2172/1808314.

47

[6-22] Nicholas Wozniak, Emily R. Shemon, MOOSE-Based Thermo-Mechanical Bowing Models with Application to Liquid-Metal Cooled Fast Reactors, ANS Winter Meeting, Phoenix, AZ, 2023.

[6-23] (IAEA VP1 VTB Model)

[6-24] Wozniak, Nicholas, and Shemon, Emily. Continued Verification of MOOSE Structural Mechanics Tools for Modeling Core Bowing Phenomena in Fast Reactors. United States: N. p.,

2022. Web. doi:10.2172/1879091.

[6-25] (IAEA VP3A VTB Model https://www.osti.gov/biblio/1879091/)

[6-26] Wozniak, Nicholas, Yang, Gang, Shemon, Emily, and Abdelhameed, Ahmed. FY23 Status Report on MOOSE-Based Approaches to Modeling Core Bowing in Fast Reactors.

United States: N. p., 2023. Web. doi:10.2172/2001073.

[6-27] TBD, In progress.

[6-28] Miao, Y., Oaks, A., Mo, K., Billone, M., Matthews, C., Zabriskie, A.X., Novascone, S. and Yacout, A.M., 2021. Metallic fuel cladding degradation model development and evaluation for BISON. Nuclear Engineering and Design, 385, p.111531.

[6-29] Miao, Y., Fassino, N., Oaks, A., Shu, S., Yacout, A.M., Matthews, C., Novascone, S.,

Implementation of low-burnup swelling focused assessment case and enhancement of FIPD-BISON integration framework, ANL/NEAMS-23/7, Argonne National Laboratory, 2023.

[6-30] Miao, Y., Oaks, A., Mo, K., Shu, S., Fassino, N., Matthews, C., Novascone, S. and Yacout, A.M., 2023. BISON-FIPD integration enhanced low-burnup SFR metallic fuel swelling model evaluation framework. Nuclear Engineering and Design, 414, p.112611.

[6-31] Miao, Y., Oaks, A., Shu, S., Fassino, N., Yacout, A.M., Matthews, C., Novascone, S.,

Enhance BISON Metal Fuel Transient Models and Enable Assessment Based on Out-of-Pile Transient Experiments, ANL/NEAMS-23/6, Argonne National Laboratory, 2023.

[6-32] Shu, S., Miao, Y., Oaks, A., Mo, K., Tomchik, C. and Yacout, A.M., 2024. Improved correlations of the fuel/cladding liquid penetration rate with the out-of-pile transient database.

Nuclear Engineering and Design, 417, p.112819.

[6-33] Liu, Yang, Travis Mui, Ziyu Xie, Rui Hu; Benchmarking FFTF LOFWOS Test# 13 using SAM code: Baseline model development and uncertainty quantification, Annals of Nuclear Energy, 192, 110010, 2023.

48

[6-34] A. Huxford, V. Coppo Leite, E. Merzari, L. Zou, V. Petrov, A. Manera, Validation of SAM-NekRS Coupling Against the TALL-3D STH/CFD Coupling Benchmark Transient, NURETH-20, Washington, DC, 2023.

[6-35] Goth, N., Jones, P., Nguyen, D. T., Vaghetto, R., Hassan, Y. A., Obabko, A., Merzari, E.,

and Fischer, P. F. Comparison of experimental and simulation results on interior subchannels of a 61-pin wire-wrapped hexagonal fuel bundle. Nucl. Eng. Design, 338, pp. 130-136, 2018.

https://doi.org/10.1016/j.nucengdes.2018.08.002.

[6-36] Dai, Dezhi, Merzari, Elia, Brockmeyer, Landon, and Shaver, Dillon. "CFD Benchmark of Pressure Drop in a 61-Pin Wire-Wrapped Assembly with Blocked Channels Using NekRS, Tech. Rep. ANL/NSE-23/6, Argonne National Laboratory, 2023, https://doi.org/10.2172/1970160

[6-37] TBD, in progress.

[6-38] TBD, in progress.

[6-39] Lomperski, S., Obabko, A., Merzari, E., Fischer, P., and Pointer, W. D. Jet stability and wall impingement flow field in a thermal striping experiment. Intl. J. Heat and Mass Transfer, 115a, pp. 1125-1136, (2017), https://doi.org/10.1016/j.ijheatmasstransfer.2017.07.076

[6-40] TBD, in progress.

[6-41] TBD, Planned.

[6-42] VTB, https://mooseframework.inl.gov/virtual_test_bed/sfr/subchannel/thors/thors.html

[6-43] TBD, Planned.

[6-44] TBD, Planned.

49

7. Molten Salt Reactor Validation There are numerous molten salt cooled reactor design concepts, including designs with fuel salts and those with solid fuel and a coolant salt. Several types of salt coolants have been proposed, with FLiBe and FLiNaK being the most common.

The Molten Salt Reactor Experiment (MSRE) is the only large-scale integral test facility for fuel salt reactors. Several prototypes are in the planning stage but have not yet produced validation information. Several small flow loops have been or are being constructed. These flow loops important in assessing codes for their ability to simulate natural circulation in molten salts, which are high Prandtl number fluids.

Unique to molten fuel salts is mobility of the fuel, neutron pre-cursors and fission products.

Assessing codes for simulation of this chemical environment is a challenge and several studies have been conducted or are in the planning stage. The molten salt thermochemical properties database (MSTDB-TC) and MSRE are being used for this purpose.

Validation exercises yet to be completed are highlighted in yellow. Those highlighted in light blue are references to models in the Virtual Test Bed (VTB) repository. These VTB models by themselves do not provide a comparison to data but are available to support other validation efforts.

Table 8 Validation for Molten Salt Reactors Test T

F K

M Code(s) Involved Type Design Type(s)

Status Validation Ref.

MSRE X

X X

SAM, Griffin IET MFSR DOE-O

[7-1] [7-2]

[7-17] [7-22]

[7-26] [7-27]

[7-28]

MSRE X

X X

Pronghorn, Griffin, SAM IET MFSR DOE-O

[7-9] [7-18]

[7-20]

CIET X

SAM IET MSPR DOE-C

[7-3] [7-23]

[7-29]

UCB-PBHTX X

SAM SET MSPR DOE-P

[7-4]

UW Flow Loop X

SAM SET

MFSR, MSPR, MSR DOE-C

[7-5]

UW NC Loop X

Pronghorn, Nek SET

MFSR, MSPR, MSR DOE-P

[7-11] [7-25]

50 OSU-LTDF X

SAM SET

MFSR, MSPR, MSR DOE-C

[7-5]

UM -

FLUSTFA X

SAM SET

MFSR, MSPR, MSR DOE-P

[7-6]

ORNL - LSTL X

SAM SET

MFSR, MSPR, MSR DOE-P

[7-7]

ORNL - LSTL X

TRACE SET

MFSR, MSPR, MSR DOE-C

[7-12]

Gallium Melting Benchmark X

Pronghorn SET

MFSR, MSPR, MSR SFR DOE-C

[7-8]

TAMU-NC X

SAM SET MSR

[7-16]

MSTDB-TC and MSRE X

X Griffin, MOOSE IET MFSR DOE-O

[7-19]

MSTDB-TC and MSRE X

MOOSE IET MFSR DOE-O

[7-21]

There are also several code-to-code benchmark and code development studies that have been conducted. While these benchmark studies do not involve a comparison to experimental data, they exercise the analytical codes and provide some insight on performance of a reactor system. The code development activities further help to enhance simulation capabilities in the absence of experimental data, such as in the tracking of chemical species in a reactor system.

The following table lists benchmark studies relevant to molten salt designs.

Table 9 Benchmark Studies for Molten Salt Designs Test T

F K M Code(s) Involved Type Design Type Status Code-Code Verification Reference Tibergas MSR Benchmark X

X Pronghorn, Griffin SET MFSR MSR DOE-C

[7-13]

Generic MSR X

Pronghorn, Nek5000 IET MFSR MSR DOE-O

[7-14]

Generic MSR Depletion Benchmark X

Griffin, Serpent SET MFSR MSR (pool)

DOE-C

[7-15]

51 7.1 References

[7-1] Fei, T., Hua, T., Feng, B., Heidet, F., and Hu, R., MSRE TRANSIENT BENCHMARKS USING SAM, EPJ Web Conf., vol. 247, p. 07008, 2021, doi: 10.1051/epjconf/202124707008.

[7-2] T. Fei, S. Shahbazi, J. Fang, and D. Shaver, Validation of NEAMS Tools Using MSRE Data, Argonne National Laboratory, ANL/NSE-22/48, Aug. 2022. doi: 10.2172/1880993.

[7-3] Ling Zou, Rui Hu, and Anne Charpentier, SAM Code Validation using the Compact Integral Effects Test (CIET) Experimental Data, ANL/NSE-19/11, June 2019.

[7-4] TBD, Planned.

[7-5] Hsun-Chia Lin, Thermal Hydraulics System-Level Code Validation and Transient Analyses for Fluoride Salt-Cooled High-Temperature Reactors, PhD thesis, University of Michigan, 2020.

[7-6] TBD, Planned.

[7-7] TBD, Planned.

[7-8] Freile, R., Tano, M. E., & Ragusa, J. C. (2024). CFD assessment of RANS turbulence modeling for solidification in internal flows against experiments and higher fidelity LBM-LES phase change model. Annals of Nuclear Energy, 197, 110275.

[7-9] TBD, Planned.

[7-10] Not used.

[7-11] TBD, Planned.

[7-12] P. Avigni, A.J. Wysocki, G.L. Yoder, "Liquid Salt Test Loop modeling using TRACE,"

Annals of Nuclear Energy, Volume 106, August 2017, Pages 170-184.

52

[7-13] Jaradat M., Choi N., Tano M., Schunert S., Abou-Jaoude A., Verification and Validation Activities of Molten Salt Reactors Multiphysics Coupling Schemes at Idaho National Laboratory, (Paper Accepted to PHYSOR 2024.)

[7-14] TBD.

[7-16] Reis, J., Seo, J. and Hassan, Y., Experimental Investigation and Numerical Validation of Natural Circulation in Molten Salt, Proceedings of NURETH-20, (Summary #5219), 2023.

[7-17] A. Leandro, F. Heidet, R. Hu, and N. Brown, Thermal hydraulic model of the molten salt experiment with the NEAMS system analysis module, Annals of Nuclear Energy, 126, pp.59-67, 2019.

[7-18] Y. Cao, J. Fang, Y. Jeong, S. Shahbazi, and T. Fei, "Multiphysics Simulation and Benchmark of MSRE," Argonne National Laboratory, ANL/NSE-23/58, September 2023.

[7-19] TBD, in progress.

[7-20] TBD, in progress.

[7-21] R. Gong and S. Shahbazi, "Demonstrating PyCalphad, ESPEI, and MSTDB-TC for MSR Applications," Argonne National Laboratory (in review), 2024.

[7-22] Hua, Thanh, Ting Fei, Bo Feng, Rui Hu, M Stempniewicz and F. Roelofs. "Benchmarking MSRE Transient Tests using SPECTRA and SAM." Proc. of NURETH-19, March 6, 2022 -

March 11, 2022.

[7-23] L. Zou, G. Hu, D. O'Grady, R. Hu, Code validation of SAM using natural-circulation experimental data from the compact integral effects test (CIET) facility, Nuclear Engineering and Design 377 (2021), 111144.

[7-24] Not used.

[7-25] TBD, planned.

[7-26] Gang Yang, Ling Zou, Rui Hu, Updated SAM Model for the Molten Salt Reactor Experiment (MSRE), ANL/NSE-23/8, Argonne National Laboratory, February 2023

[7-15] Walker, Samuel Austin, Abou Jaoude, Abdalla, Calvin, Olin W, and Tano Retamales, Mauricio Eduardo. Implementation of Isotopic Removal Capability in Griffin for Multi-Region MSR Depletion Analysis. United States: N. p., 2022. Web. https://www.osti.gov/biblio/1987468

53

[7-27] Mustafa K. Jaradat and Javier Ortensi, Thermal Spectrum Molten Salt-Fueled Reactor Reference Plant Model, INL/RPT-23-72875, Idaho National Laboratory, July 2023

[7-28] VTB, Molten Salt Fast Reactor (MSFR),

https://mooseframework.inl.gov/virtual_test_bed/msr/msfr/

[7-29] D. OGrady, T. Mui, A. Lee, L. Zou, G. Hu, R. Hu, SAM Code Enhancement, Validation, and Reference Model Development for Fluoride-salt-cooled High-temperature Reactors, ANL/NSE-21/15, Argonne National Laboratory, April 2021.

8. Microreactor Validation Microreactors are those designs with very low power, generally less than 10 MWt. Some designs are intended to be transportable and thus are small to many other advanced systems.

Designs currently proposed to the NRC for possible licensing use heat pipes to remove heat from the core. However, other microreactors may utilize other means of heat removal.

Because the microreactor designs are conceptual, very few applicable tests exist. The only test information currently available is that from the KRUSTY (Kilopower Reactor Using Stirling TechnologY) experiments, and the earlier proof of concept tests using DUFF (Demonstration Using Flattop Fissions). The MARVEL reactor is expected to be available and provide data in several years.

The following table lists the validation tests applicable to microreactors. Of note for these tests is the need to simulate thermo-mechanical expansion of the core and its effect on neutronics for fast spectrums.

As with molten salt reactor designs, this list will likely need to be increased with tests to provide data to validate heat transfer from the monolith or other supporting structures for the fuel and core. Specific validation will be needed for heat pipes to characterize steady-state and transient behavior at several levels of power.

Table 11 lists the validation cases for microreactors. Validation exercises yet to be completed are highlighted in yellow.

54 Table 10 Validation for Microreactors Test T

F K

M Code(s)

Involved Type Design Type(s)

Status Validation Ref.

KRUSTY x

x x

x

MOOSE, BISON, SAM (and/or SOCKEYE),

Griffin IET HPR DOE-C

[8-1], [8-2]

DUFF X

X X

X

MOOSE, BISON, SAM

SOCKEYE, Griffin IET HPR DOE-C

[8-3]

MARVEL X

X X

X

MOOSE, BISON, SAM

SOCKEYE, Griffin IET LMR DOE-O

[8-4]

GODIVA X

X X

MOOSE, BISON, Griffin IET HPR DOE-C

[8-6]

There are two benchmark studies applicable to microreactors. Both involve paper reactor designs which share characteristics of proposed heat pipe cooled designs being proposed but avoid use of proprietary information. While these benchmark studies do not involve a comparison to experimental data, they exercise the analytical codes and provide some insight on performance of a reactor system.

Table 11 Microreactor Benchmarks Test T

F K

M Code(s)

Involved Type Design Type Status Code-Code Verification Reference SiMBA X

X X

MOOSE, BISON,

SOCKEYE, Griffin HPR DOE-C

[8-7]

Empire X

Shift, Griffin HPR DOE-C

[8-8]

55 8.1 References

[8-1] N. Stauff, A. Abdelhameed, Y. Cao, N. Fassino, L. Ibarra, Y. Miao, K. Mo, D. Nunez, High-fidelity multiphysics load following and accidental transient modeling of microreactors using NEAMS tools, ANL/NEAMS-23/4, Sept 30, 2023.

[8-2] Yan Cao, Yinbin Miao, Kun Mo, Nicolas Stauff and Changho Lee Multiphysics Simulations of the KRUSTY Criticality Experiment Using BlueCrab. PHYSOR 2024 (accepted)

[8-3] McClure, Patrick R., David I. Poston, and David D. Dixon. Final results of Demonstration Using Flattop Fissions (DUFF) experiment. No. LA-UR-12-25165. Los Alamos National Lab.(LANL), Los Alamos, NM (United States), 2012.

[8-4] S. Terlizzi and I. Trivedi (2024). Progress toward the development of a high-fidelity neutronic model of MARVEL using Griffin, PHYSOR 2024.

[8-5] Not currently used.

[8-6] Yaqi Wang, Sebastian Schunert, Javier Ortensi, Frederick N. Gleicher, Benjamin A.

Baker, Mark DeHart, Richard C Martineau, "Demonstration of Griffin Fully-Coupled Multiphysics Simulation with the Godiva Benchmark Problem, " Submitted to M&C 2017 International Conference on Mathematics and Computational Methods Applied to Nuclear Science and Technology, Jeju, Korea, April 16-20, 2017.

[8-7] Stefano Terlizzi, Vincent Labouré, "Asymptotic Hydrogen Redistribution Analysis in Yttrium-Hydride-moderated Heat-Pipe-cooled Microreactors Using DireWolf", Annals of Nuclear Energy, Volume 186, 2023, 109735, ISSN 0306-4549, https://doi.org/10.1016/j.anucene.2023.109735.

56

9. General Neutronics Advanced non-LWRs include designs with thermal and fast spectrums, and thus validation of neutronics tools must cover both. Many of the tests are generic in that that may satisfy validation for several design types. The following table lists tests that have been proposed for validation of neutronics codes.

Validation exercises yet to be completed are highlighted in yellow. Those highlighted in light blue are references to models in the Virtual Test Bed (VTB) repository. These VTB models by themselves do not provide a comparison to data but are available to support other validation efforts.

Table 12 Validation of Neutronics for Advanced Non-LWRs Test Code(s)

Involved Type Design Type(s)

Status Validation Reference CEFR Griffin SET

SFR, LMR NRC-O

[9-1]

EBR-II (Runs 130B -170A)

Griffin SET

SFR, LMR DOE-O

[9-2]

Monju Startup Griffin SET SFR DOE-O

[9-3]

BFS (109-2A, 76-1A, 73-1)

Griffin SET SFR DOE-O

[9-4]

VHTRC Griffin SET HTGR DOE-O

[9-5]

HTTR Griffin IET HTGR DOE-O

[9-16]

ZPPR-6 (6,7A)

Griffin SET SFR DOE-O

[9-6]

ZPPR-21 (A-F)

Griffin SET SFR DOE-O

[9-7]

ZPPR-15 (A-D)

Griffin SET SFR DOE-O

[9-8]

TREAT Minimum Critical, M8, M2/M3 Griffin SET HTGR DOE-O

[9-9], [9-10]

[9-11] [9-12]

ASTRA Griffin SET PBMR DOE-O

[9-14]

SEFOR Griffin SET SFR DOE-P

[9-15]

HTR-PROTEUS Griffin SET PBMR DOE-P

[9-17]

GODIVA

Shift, Griffin SET HPR DOE-C

[9-13]

57 Verification of neutronics calculations can also be accomplished through a comparison of results to a Monte Carlo simulation. Monte Carlo simulation of a reactor generally serve as a reference calculation by which other results are evaluated. In addition to validation against experimental measurements, Griffin has been compared to Monte Carlo simulations. The following table lists these code-to-code comparisons and provides comments on the specific figures-of-merit.

Table 13 Code to Code Verification of Neutronics for Advanced Non-LWRs Test XS/Transport Code Design Type(s)

Status Verification Reference Comments ABR-1000 MCC3/Griffin SFR DOE-C

[9-38]

k-eff, CR worth, FA &

pin power ABTR MCC3/Griffin SFR DOE-C

[9-38] [9-39]

[9-31]

k-eff, CR worth, FA &

pin power, thermal expansion coefficients NEAMS LFR assembly MCC3/Griffin LMR DOE-C

[9-18] [9-44]

k-eff, pin power MSRE Shift/Griffin Serpent/Griffin MSRE DOE-C

[9-19] [9-32]

k-eff, XS, fission products EMPIRE Griffin SSAPI

/Griffin Serpent/Griffin Shift/Griffin HPR DOE-C

[9-20]

[9-36]

[9-37]

KRUSTY MCC3/Griffin Serpent/Griffin HPR DOE-O DOE-C

[9-21]

k-eff k-eff, power distributions NEAMS Heat Pipe Microreactor Serpent/Griffin HPR DOE-C

[9-22] [9-40]

SiMBA Serpent/Griffin HPR DOE-C

[9-23]

NEAMS Gas-cooled Microreactor Serpent/Griffin Gas-Cooled Microreactor DOE-C

[9-22] [9-24]

[9-41]

58 NEAMS Gas-cooled Microreactor Serpent/Griffin Gas-Cooled Microreactor DOE-O

[9-25]

Benchmark to be updated in FY24.

MARVEL Serpent/Griffin NaK cooled Microreator DOE-O

[9-26]

SNAP8-ER Serpent/Griffin NaK cooled Microreactor DOE-C

[9-27] [9-42]

SPR A Microreactor Serpent/Griffin HPR DOE-C

[9-28]

NRC model MHTGR-350 DRAGON/Griffin HTGR DOE-C

[9-29] [9-43]

OECD/NEA benchmark PBMR-400 VSOP/Griffin PBMR DOE-C

[9-30]

OECD/NEA benchmark gFHR DRAGON/Griffin MSPR DOE-C

[9-31]

NRC model MSRE OpenMC/Griffin MSRE DOE-C

[9-32]

NRC model ABTR MCC3/Griffin SFR DOE-C

[9-33]

NRC model HTR-PM DRAGON/Griffin PBMR DOE-C

[9-34]

NRC model eVinci like Serpent/Griffin HPR DOE-C

[9-35]

NRC model 9.1 References

[9-1] TBD.

59

[9-9] TBD.

[9-10] B. A. Baker, J. Ortensi, M. D. DeHart, Y. Wang, S. Schunert and F. N. Gleicher, "Analysis Methods and Validation Activities for Griffin Using M8 Calibration Series Data,"

INL/EXT-16-40023, Idaho National Laboratory, Idaho Falls, Idaho, September 2016.

[9-11] TBD.

[9-12] TBD.

[9-13] Pandya, Tara M., et al. "Two-step neutronics calculations with Shift and Griffin for advanced reactor systems." Annals of Nuclear Energy 173 (2022): 109131.

[9-14] TBD.

[9-15] TBD.

[9-16] TBD.

[9-17] TBD.

[9-18] C. Brennan, H. Park, E. Shemon, Demonstration of MOOSE-Based Griffin Reactor Physics Code for Heterogeneous Lead-Cooled Fast Reactor Analysis, Proceedings of PHYSOR, Pittsburgh, PA, May 15-20 2022.

[9-19] D. Hartanto, E. Davidson, A. Godfrey, Y. Cao, T. Fei, and M. E. Tano-Retamales, Molten Salt Reactor Experiment Simulation using Shift/Griffin, Technical Report, ORNL/TM-2023/3005, Oak Ridge National Laboratory, TN, United States. https://doi.org/10.2172/1997706

[9-2] TBD.

[9-3] TBD.

[9-4] TBD.

[9-5] TBD.

[9-6] TBD.

[9-7] TBD.

[9-8] TBD.

60

[9-20] C. H. Lee, H. Park, Y. Jung, and Y. Wang, High-fidelity Simulation of the Empire Micro Reactor Benchmark Problem Using Griffin with Online Cross Section Generation, M&C 2023, Niagara Falls, Ontario, Canada, August 13-17, 2023.

[9-21] Y. Cao, Y. Miao, K. Mo, N. Stauff, and C. H. Lee, Multiphysics Simulations of the KRUSTY Criticality Experiment Using BlueCrab, PHYSOR 2024, San Francisco, CA, April 21-24, 2024. (accepted)

[9-22] N. Stauff, A. Abdelhameed, Y. Cao, Kristina, Y. Miao, K. Mo, D. Nunez, Multiphysics Analysis of Load Following and Safety Transients for MicroReactors, ANL/NEAMS-22/1, September 30, 2022

[9-23] SiMBA VTB Model:

https://mooseframework.inl.gov/virtual_test_bed/microreactors/hpmr_h2/hpmr_h2_results.html

[9-24] A. Abdelhameed, Y. Cao, D. Nunez, Y. Miao, K. Mo, C.H. Lee, E. Shemon, N. Stauff High-Fidelity Multiphysics Modeling of Load Following for 3-D Gas-Cooled Microreactor Assembly using NEAMS Codes, ANS Winter Meeting, Phoenix, AZ, November 13-17 (2022)

[9-25] N. Stauff, A. Abdelhameed, Y. Cao, N. Fassino, L. Ibarra, Y. Miao, K. Mo, D. Nunez, High-fidelity multiphysics load following and accidental transient modeling of microreactors using NEAMS tools, ANL/NEAMS-23/4, Sept 30, 2023.

[9-26] S. Terlizzi and I. Trivedi. (2024). Progress toward the development of a high-fidelity neutronic model of MARVEL using Griffin, PHYSOR 2024, San Francisco, CA, April 21-24, 2024. (accepted)

[9-27] I. Naupa Aguirre, S. Garcia, B. Lindley, S. Terlizzi, A. Abou Jaoude, D. Kotlyar, Verification of the Serpent-Griffin Workflow using the SNAP 8 Experimental Reactor, International Conference on Mathematics and Computational Methods Applied to Nuclear Science and Engineering (M&C2023), Niagara Falls, CA, August 13-17, 2023.

[9-28] Hu, G., Hu, R., Kelly, J. M., and Ortensi, J. Multi-Physics Simulations of Heat Pipe Micro Reactor. ANL-NSE-19/25, 155817, 2019.

[9-29] OECD/NEA. Benchmark of the modular high-temperature gas-cooled reactor 350 MW core design: volumes I and III. Technical Report NEA/NSC/R20174, OECD/NEA, 2018.

[9-30] Paolo Balestra, Sebastian Schunert, Javier Ortensi, Mark D DeHart, Griffin/Pronghorn PBMR-400 Benchmark Results, INL/MIS-20-57522, February 2020.

[9-31] Javier Ortensi, Cole M. Mueller, Stefano Terlizzi, Guillaume Giudicelli, and Sebastian Schunert, Fluoride-Cooled High-Temperature Pebble-Bed Reactor Reference Plant Model, INL/RPT-23-72727, May 2023.

[9-32] Mustafa Jaradat and Javier Ortensi Thermal Spectrum Molten Salt-Fueled Reactor Reference Plant Model, INL/RPT-23-72875, July 2023.

61

[9-33] Stefano Terlizzi, Namjae Choi, Jackson Harter, Ishita Trivedi, and Javier Ortensi, Sodium-cooled Fast Reactor Reference Plant Model, INL/RPT-24-TDB February 2024.

[9-34] Mustafa Jaradat, Sebastian Schunert, and Javier Ortensi, Gas-Cooled High-Temperature Pebble-Bed Reactor Reference Plant Model, INL/RPT-23-72192, April 2023.

[9-35] Javier Ortensi, Joshua Hansel, Mustafa Jaradat, Stefano Terlizzi, The Monolith Heat Pipe Micro Reactor Reference Plant Mode, INL/RPT-24-TBD, April 2024.

[9-36] Lee, Changho, Jung, Yeon Sang, Zhong, Zhaopeng, Ortensi, Javier, Laboure, Vincent, Wang, Yaqi, and DeHart, Mark. Assessment of the Griffin Reactor Multiphysics Application Using the Empire Micro Reactor Design Concept. United States: N. p., 2020. Web.

doi:10.2172/1833008.

[9-37] Tara Pandya, Friederike Bostelmann, Matthew Jessee, Javier Ortensi, Two-step neutronics calculations with Shift and Griffin for advanced reactor systems, Annals of Nuclear Energy, Volume 173, 2022, https://doi.org/10.1016/j.anucene.2022.109131

[9-38] C. H. Lee, S. M. Park, S. Kumar, N. Stauff, Z. Xu, and J. Harder, Simulation of the Sodium-cooled Fast Reactor Benchmark Cores Using MC2-3/Griffin, ANS Annual Meeting, Indianapolis, IN, June 11-14, 2023.

[9-39] N. Choi, S. Terlizzi, Y. Wang, H. Park, C.H. Lee, J. Ortensi, Investigation of Ring-heterogeneous Geometry Approximation for Efficient and Practical SFR Analysis, PHYSOR 2024, San Francisco, CA, April 21-24, 2024. (accepted) https://mooseframework.inl.gov/virtual_test_bed/microreactors/mrad/index.html https://mooseframework.inl.gov/virtual_test_bed/microreactors/gcmr/index.html https://mooseframework.inl.gov/virtual_test_bed/microreactors/s8er/index.html https://mooseframework.inl.gov/virtual_test_bed/htgr/mhtgr_griffin/index.html https://mooseframework.inl.gov/virtual_test_bed/lfr/heterogeneous_single_assembly_3D/Griffin_

standalone_LFR.html

[9-44]

[9-43]

[9-42]

[9-41]

[9-40]

62

10.

Component Validation A separate, but important category of validation tests is that for special components and safety related systems. Several advanced reactor designs utilize a reactor cavity cooling system (RCCS), and some may use a decay heat removal system (DHRS). Microreactors can use heat pipes to remove primary heat. For the purposes of this report, these are considered safety related components and require special validation. The following table lists tests that are important to validation of BlueCRAB codes involved. This list is expected to increase as additional DHRS tests, pump performance tests, and other tests of safety important features are conducted. Table 15 lists the validation cases for microreactors. Validation exercises yet to be completed are highlighted in yellow.

Table 14 Component Validation Tests Test T

F K

Code(s)

Involved Type Design Type(s)

Status Validation Ref.

NSTF (air)

X SAM SET

HTGR, PBMR DOE-C

[10-1]

NSTF (air)

X TRACE SET

HTGR, PBMR NRC-C

[10-13]

NSTF (water)

X SAM SET

HTGR, PBMR DOE-P

[10-25]

NSTF (water)

X TRACE SET MSPR NRC-O

[10-2]

UW RCCS X

SAM, TRACE SET

HTGR, PBMR DOE-P

[10-3]

UW RCCS X

Nek IET HTGR MSPR DOE-O

[10-24]

RCCS (UM)

X Nek SET HTGR MSPR DOE-C

[10-23]

HTTR-VCS X

SAM, TRACE SET HTGR MSPR DOE-P

[10-4]

HTR-PM Hot Gas Mixing Experiment X

Pronghorn, Nek IET HTGR PBMR DOE-P

[10-22]

Purdue H-shaped helium loop X

SAM SET HTGR PBMR DOE-P

[10-10]

RPI HP X

Sockeye SET HPR DOE-O

[10-6]

SAFE-30 X

Sockeye SET HPR DOE-C

[10-7] [10-14]

SPHERE X

Sockeye SET HPR DOE-O

[10-8]

63 SPHERE -GC X

Sockeye SET HPR DOE-P

[10-9]

U. Michigan HP X

Sockeye SET HPR DOE-O

[10-11] [10-15]

TAMU HP X

Sockeye SET HPR DOE-P

[10-12]

Ivanoskii HP X

Sockeye SET HPR DOE-P,

[10-16]

LANL HP X

Sockeye SET HPR DOE-O

[10-17]

OECD/NEA T-junction X

Nek5000 SET n/a DOE-C

[10-18]

B&W - steam generator tube bundle X

Nek5000 SET n/a DOE-C

[10-19] [10-20]

[10-21]

Tan et al.

Twisted tube bundle X

Nek5000 SET n/a DOE-C

[10-22]

10.1 References

[10-1] B. Hollrah, M. Bucknor, D. Lisowski, Y. Hassan, R. Vaghetto, R. Hu, Benchmark Simulation of the Natural Convection Shutdown Heat Removal Test Facility Using SAM, Nuclear Technology, Vol. 206, 1337-1350, 2020.

TBD. TBD, In progress.

TBD, TBD, Planned.

TBD. TBD, Planned.

TBD. TBD, Planned.

TBD. TBD, Planned.

[10-7] Joshua E. Hansel, Ray A. Berry, David Andrs, Matthias S. Kunick, and Richard C.

Martineau. Sockeye: a one-dimensional, two-phase, compressible flow heat pipe application.

Nuclear Technology, 207(7):1096-1117, 2021.

[10-8] Joshua Hansel, Jeremy Hartvigsen, Lander Ibarra, Piyush Sabharwall, and Bo Feng.

Sockeye validation support using the sphere facility, International Conference on Physics of Reactors 2022 (PHYSOR 2022). May 2022.

TBD. TBD, Planned.

[10-9]

[10-6]

[10-5]

[10-4]

[10-3]

[10-2]

64

[10-10] TBD, Planned.

[10-11] Joshua E. Hansel, Carolina da Silva Bourdot Dutra, Pei-Hsun Huang, and Taehwan Ahn. A Sockeye Model of a Test at the Michigan Single Sodium Heat Pipe Facility. Proceedings of the ANS student conference. 2024. (Submitted)

[10-12] TBD. (No validation ref., only experiment ref.)

[10-13] Jason Thompson, TRACE Calculation Notebook for the Water-Based Natural Convection Shutdown Heat Removal Test Facility (NSTF); Single-Phase Test Series, ADAMS ML23335A021, October 2023.

[10-14] J. Hansel and L. Charlot. Sockeye heat pipe analysis code verification and validation.

In The 19th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-19).

Brussels, Belgium, March 2022.

[10-15] Carolina da Silva Bourdot Dutra, Elia Merzari, John Acierno, Adam Kraus, Annalisa Manera, Victor Petrov, Taehwan Ahn, Pei-Hsun Huang & Dillon Shaver. High-Fidelity Modeling and Experiments to Inform Safety Analysis Codes for Heat Pipe Microreactors. Nuclear Technology. 2023.

[10-16] Michael A. Shockling, Megan E. Durse, Liping Cao, John Lojek III, Rory A. F. Blunt.

Benchmark Studies of the MOOSE-based Sockeye and BISON codes for the eVinci Heat Pipe Microreactor. 20th International Topical Meeting on Nuclear Reactor Thermal Hydraulics.

Washington D.C. 2023.

[10-17] Tano, M., Sabharwall, P., and Sweetland, K. (2023). Modeling Heat Pipes with Non-Condensable Gases (No. INL/RPT-23-74866-Rev000). Idaho National Laboratory (INL), Idaho Falls, ID (United States).

[10-18] Obabko, A V, Fischer, P F, Tautges, T J, Karabasov, S, Goloviznin, V M, Zaytsev, M A, Chudanov, V V, Pervichko, V A, Aksenova, A E, CFD validation in OECD/NEA t-junction benchmark. Tech. Rep. ANL/NE-11/25, Argonne National Laboratory, 2011.

[10-19] L. Brockmeyer, J Solberg, One-Way Coupled Flow-Induced Vibration Analysis for a Twisted-Tube Heat Exchanger, LLNL-TR-795710, 2019.

[10-20] Yuan, Haomin, Solberg, Jerome, Merzari, Elia, Kraus, Adam, and Grindeanu, Iulian.

Flow-induced vibration analysis of a helical coil steam generator experiment using large eddy simulation Nucl. Eng. Design, 322, pp 547-562, (2017).

https://doi.org/10.1016/j.nucengdes.2017.07.029

65

[10-21] M. A. Yildiz, Y. Hassan, H. Yuan, E. Merzari, NUMERICAL SIMULATION OF ISOTHERMAL FLOW ACROSS SLANT FIVE-TUBE BUNDLE WITH SPECTRAL ELEMENT METHOD CODE NEK5000, Nucl. Tech., Volume 206, Issue 2, 2019, https://www.osti.gov/biblio/1607399

[10-22] D. R. Shaver, L. B. Carasik, E. Merzari, N. Salpeter, E. Blandford, Calculation of Friction Factors and Nusselt Numbers for Twisted Elliptical Tube Heat Exchangers Using Nek5000, J. Fluids Eng., Volume 141, Issue. 7, 2019, https://www.osti.gov/biblio/1542194

[10-23] J. Acierno, E. Merzari, Large Eddy Simulation of Jet Interaction, ATH 22, Anaheim, CA, June 12-16, 2022.

[10-24] TBD, Planned. (As part of the Generation IV International Forum (GIF) - Project Arrangement on Computational Methods Validation and Benchmarks (CMVB) for The International Research and Development of the Very-High-Temperature Reactor Nuclear Energy System.)

[10-25] TBD, Planned.

66

11.

Summary and Conclusions Summary Previous sections of this report have documented the verification and validation applicable and related to the BlueCRAB codes. Considerable efforts have gone into assessment of the various codes, but as can be seen in the tables summarizing the efforts to date there are many useful and important cases that need to be done. Validation for heat pipe performance and neutronic would benefit by additional work. In this summary, a characterization of the status of V&V is discussed. Because of the diversity in non-LWR designs and supporting database, this status is provided by design type.

Gas-Cooled Reactors Gas cooled reactors, both prismatic and pebble bed designs, benefit from several large-scale test facilities in addition to operating reactors to provide data for validation. Data from operating and test reactors such as the AVR, HTR-10 and HTR-PM have been used to validate the BlueCRAB codes for pebble bed reactors for both steady-state and transient conditions.

Because these reactors have nuclear cores, they provide information to examine the coupled simulation of neutronics and thermal-fluids. With operational experience and data from these facilities, BlueCRAB has been assessed for pebble bed reactors.

Pebble bed reactors present a unique set of challenges to neutronic analysis. Moving fuel and the core residence time are important factors that affect depletion and determine the power distribution. Validation of the Griffin code using the HTR-PM reactor has been done to demonstrate the capability to simulate core power in a pebble bed reactor.

The HTTR nuclear reactor and the non-nuclear integral test facility HTTF provides validation data for coupled neutronics/thermal-fluid (HTTR) and thermal-fluid only (HTTF) code assessment. SAM, Griffin, and to a lesser extent Pronghorn have been validated based on data from these facilities.

One phenomena that may not be well characterized in the pebble bed designs is bypass on the periphery of a pebble bed core. The packing fraction of pebbles near the side walls will not be as large as in the central core resulting in higher flows there. These flows are difficult to accurately measure in integral tests, and uncertainty in the bypass flow should be considered as part of an assessment. Uncertainty in the bypass flow may have a major effect on temperatures in the interior of the core.

67 Data from other operating reactors including AVR, Fort St. Vrain, and Peach Bottom could be beneficial to code validation efforts. However, availability of data and/or the lack of challenging transients in those operating reactors limit the usefulness of those validation efforts should they be made. Currently, validation of the BlueCRAB codes using these formerly operating reactors and the THTR-300 would be performed subject to data availability.

Progress has been made using separate effects tests to examine simulation capabilities for specific geometries and phenomena. Data from the SANA and TAMU P tests have enabled validation for flow in a pebble bed. Tests such as the Missouri S&T Air Experiments, TAMU UP tests, and the MigaDome tests provide data for validation of flow and heat transfer in regions of gas-cooled reactors where thermal mixing and stratification may occur.

Heat removal by reactor cavity cooling systems have been the subject of several experimental studies and those results provide important assessment data. The particular designs of the experimental facilities involved however may not be fully representative of the designs by applicants. Scaling of the physical processes in the experiments to the full-scale prototype designs will be an important consideration.

Code-to-code benchmarks for gas-cooled reactors have been used to investigate code performance and identify modeling needs and capabilities. The MHTGR-350, gPRB-200, and PBMR-400 reactors have been used for that purpose. These benchmark exercises have been useful to the overall BlueCRAB code development effort although they do not have a direct comparison to using experimental data for validation.

In summary, BlueCRAB code assessment for gas-cooled reactor designs has been completed with the data that is available to-date. An experimental database exists and enables codes to be examined over a range of geometric scales. There are several integral test facilities and small reactors, complimented with small scale tests to examine flow and heat transfer in various regions of a gas-cooled system. The state of assessment for gas-cooled rectors could be improved upon with access to data in some of the international facilities the U.S. currently does not have.

Liquid Metal Reactors Like gas-cooled reactors, sodium-cooled reactor designs have been investigated with operational facilities including EBR-II and the FFTF. Both of these sodium-cooled reactors were used to conduct transient events to demonstrate adequate safety margins. Internationally, reactors including Monju and Phenix contribute operational experience in sodium fast reactor design. Additional operational information from the BN-600 (and other BN series reactors)

68 would be beneficial; however, access to much of this international database is restricted and will likely not be available for code assessment.

While EBR-II and FFTF provide useful data for code assessment, the design and scaling of these reactor systems may not be closely representative of expected applications. Fuel for the FFTF was comprised of metal oxide as opposed to metallic, and FFTF was a loop-design rather than a pool-type design in which major components reside within the main reactor vessel. EBR-II was a pool-type design, but its total power of 65 MWt is lower than the power in expected applications. Thus, scaling should be an important consideration in applying EBR-II assessment to a full scale SFR.

Liquid metal fast reactor fuel assemblies generally have wire-wrapped rods and a tight lattice with a pitch to diameter ratio that is much smaller than in LWRs. Validation for heat transfer and fluid flow is necessary and there are several separate effects tests that have provided data.

Tests performed in the ORNL 19-pin and 61-pin bundles along with the Toshiba 37-pin bundle have been used to validate subchannel analysis of SFR fuel assemblies. The TAMU-FA provides fluid flow information in a 61-pin bundle with and without blockages.

BlueCRAB validation for thermal stratification in sodium pools and the upper plenum are not specifically addressed. Some applicable validation on mixing and thermal striping has been accomplished through the ANL-MAX and UTK Square Cavity simulations but more is needed to be able to model and simulate prototypical geometries and conditions.

Liquid metals other than sodium have been part of validation though they involve fewer tests.

Tests involving lead, lead-bismuth, and gallium have been the subject of validation using HELIOS, KIT-KALLA, ENEA-CIRCE, and GaTE.

Thermomechanical expansion is a phenomenon of significant importance in fast reactors due to the negative reactivity resulting from core radial expansion. To partly address this, structural mechanics is being validated using the IAEA Fast Reactor Working Group Verification Problems. This may need to be supplemented by additional work to validate core plate expansion in a prototypical geometry.

In summary for liquid metal cooled reactors the experimental database for validation is sufficient for most transients and phenomena expected in SFR design basis type scenarios. Assembly flow blockage scenarios are partly covered by several separate effects tests with and without blockages; however, BlueCRAB is not validated for sodium boiling or fuel assembly distortion.

Molten Salt Reactors

69 Molten salt reactors face several validation challenges due to unique phenomena that occur.

Tracking precursors in a fuel salt reactor might possibly be the most unique process that involves the coupled simulation of both neutronics and thermal-fluids. For molten fuel salt reactors, the MSRE is the "go to" facility for code validation, and the only operational facility for a fuel salt design. There are several other facilities that provide coolant salt data and have been used for validation such as CIET and the UW Flow Loop. Other flow loops are planned or have not yet been utilized for code validation.

Salt solidification is a phenomenon that is "new" and more of a concern due to the high melting point in salts. (This can also be a process of interest in liquid metal reactors, although they have much lower melting points than salts.) Data for validation of salt freezing models in simulation codes is scare, thus in this report only the Gallium Melting Benchmark has so far been used for validation of BlueCRAB. Additional data on salt solidification and possibly the performance of "freeze plugs" for some designs may be needed.

Thermophysical properties is an area of high uncertainty in molten salts. In general, the properties such as viscosity, thermal conductivity, specific heat and density are not known to the accuracy as for those of the more common fluids including helium and sodium. As fission products are produced in a fuel salt, or if corrosion products accumulate in coolant salts, estimation of the thermophysical properties becomes more difficult. Validation of BlueCRAB thermophysical models is on-going making use of recent work to develop the Molten Salt Thermophyscial Properties Database (MSTDB). Continuation of this validation exercise is necessary to quantify accuracy of the modeling.

In summary, more work is needed to build the experimental database and conduct assessment of the BlueCRAB simulation tools for molten fuel salt reactor designs. Only the MSRE provides large-scale system data for a fuel salt. As new fuel salt designs are proposed and investigated it may be found that the MSRE does not necessarily scale well with differing geometry or operating conditions. Uncertainties in the thermophysical properties for salts not fully investigated and included in the MSTDB database remain a challenge for BlueCRAB validation of fuel salt reactors.

Microreactors These low power designs are expected to be completely passive and have a very large safety margin. Initial microreactor designs use heat pipes to exchange heat with a secondary system.

However, since initiation of BlueCRAB development, "microreactor" designs have diversified. In addition to designs cooled by heat pipes, gas-cooled and liquid metals are also being proposed as coolants. In this report, the emphasis has been placed on heat pipe designs although the BlueCRAB system of codes is considered capable of modeling other types of microreactors.

70 Regardless of the cooling mechanism, data for validation of microreactor analysis BlueCRAB codes is limited. KRUSTY and DUFF were both heat pipe cooled reactors which provide some useful validation data which has been used for code assessment. However, the designs of these reactors differ considerably from applicant expected designs. (Both KRUSTY and DUFF were highly enriched fast reactors.)

Heat pipe validation data is available, but assessment of Sockeye is currently incomplete. New data is being developed through the DOE NEUP program and is expected to be useful in validation of heat pipe performance. The expected database should be sufficient for validation.

Currently verification and validation for heat pipe microreactors depends on (1) use of benchmarks such as SiMBA and EMPIRE and (2) the future availability of small-scale prototypes being proposed by the industry and including DOE funded MAGNET and MARVEL test reactors. Thus, more work to validate microreactor designs is needed as data becomes available. BlueCRAB is considered to have the capability to simulate microreactor designs including those cooled by heat pipes and single phase fluids (helium, liquid sodium, liquid NaK).

Heat pipe simulations using Sockeye have reasonable agreement with existing data and simulation of single phase fluids through a monolith is not expected to present major problems.

Never-the-less, additional experimental data would be beneficial.

Conclusions Verification and validation of the BlueCRAB analytical codes has made significant progress.

Verification plans are in place for most codes, and validation has been performed for the design types most likely to become near-term licensing candidates. Additional validation is in progress or planned. Integral and separate effects tests are available to validate codes for physical processes important to non-LWRs although there are some where the database could stand improvement. Molten fuel salt and microreactors in particular have limited experimental databases.

Scaling of the test facilities for validation data to full-scale prototypes has not been addressed in this report. As advanced reactor designs are finalized, scaling is an important step to ensure the test results appropriately represent the full-scale design.

While additional validation should be performed, it should be noted that no unresolvable gaps exist in the validation database. Phenomena that are poorly understood can be bounded by conservative assumptions if necessary. Phenomena with high uncertainties exist, and these can be identified through application of Best Estimate Plus Uncertainty (BEPU) which is an element of EMDAP. Thus, the BlueCRAB system of codes is tentatively ready for supporting independent calculations to assist licensing reviews by the NRC.

71

12.

Appendix There are a number of experimental facilities and benchmark problems that contribute to non-LWR code validation. This section provides a brief summary of the facility with references on where more detailed information can be obtained.

ABR-1000: The Advanced Burner Reactor (ABR) is a benchmark to evaluate the reactivity coefficients for a safety analysis, the shutdown margin, and the reconstructed pin power distribution in a design similar to the TerraPower Natrium reactor.

Reference:

Stauff, Nicolas, Fei, Ting, and Smith, Mike. ARDP Natrium Neutronic Methodology:

Argonne Neutronic Assessment of ABR-1000. United States: N. p., 2022. Web.

doi:10.2172/1875808.

ANL-MAX: ANL-MAX is a large rectangular tank enclosure with two differentially heated hexagonal jets. It uses air as a working fluid and includes temperature measurements along the top of the tank for thermal striping analysis.

References:

W. D. Pointer, S. Lomperski, and P. Fischer, Validation of CFD methods for advanced SFR design: upper plenum thermal striping and stratification. In ICONE17 (2009).

S. Lomperski and D. Pointer, Max CFD validation data set for twin isothermal jets. ANL-NE-12/42, (2012).

ANL THETA: The Thermal Hydraulic Experimental Test Article (THETA) is a 500 L liquid sodium facility that is used to develop components and instrumentation as well as acquire experimental data for validation of reactor thermal hydraulic and safety analysis codes.

Reference:

Matthew Weathered, Christopher Grandy, Derek Kultgen, Edward Kent, Jordan Rein, Alex Grannan, Evan Ogren, Thermal Hydraulic Experimental Test Article - Fiscal Year 2023 Final Report, ANL-ART-278, Argonne National Laboratory, September 2023.

72 ASTRA: ASTRA is a zero-power critical facility located at the Kurchatov Institute in Russia. It has been configured to represent a PBR core and was intended for experimental investigation of HTGR neutronics. Several series of critical experiments were performed.

Reference:

Ponomarev-Stepnoi, N.N., Glushkov, E.S., Kompaniets, G.V., Polyakov, Graphite Annular Core Assemblies with Spherical Fuel Elements Containing Coated UO2 Fuel Particles, NEA/NCS/DOC(95)03/III. In: International Handbook of Evaluated Criticality Safety Benchmark Experiments, 2007 Edition, IEU-COMP-THERM-008, 2007.

ATR 94CIC: The Advanced Test Reactor (ATR) is a high neutron flux research facility that is operated at the Idaho National Laboratory (INL) in the United States. The critical configuration from the 1994 core-internals change-out (94-CIC) provides a well-documented benchmark.

Reference:

Mark D. DeHart and Gray S. Chang, EVALUATION OF CORE PHYSICS ANALYSIS METHODS FOR CONVERSION OF THE INL ADVANCED TEST REACTOR TO LOWENRICHMENT FUEL, PHYSOR 2012 - Advances in Reactor Physics - Linking Research, Industry, and Education Knoxville, Tennessee, USA, April 15-20, 2012 AVR: The AVR is a pebble bed demonstration HTGR plant in Julich, West Germany that began generating electricity in December 1967. Its purpose is to demonstrate the feasibility of an HTGR with pebble fuel elements and high operating temperatures. The core is fueled with about 100,000 graphite pebbles containing coated fuel particles. It used a variety of fuel pebbles, which eventually evolved into the TRISO pebbles. AVR provided a great deal of operating experience and experimental data, including some accident simulation data.

References:

Egon Ziermann and Gunther Ivens, Final Report on the Power Operation of the AVR Experimental Nuclear Power Station, Available from: Forschungszentrum (Research Center) Julich GmbH - Central Library D-52425 Jolich - Federal Republic of Germany, ADAMS ML082130449, 1997.

Klaus Kruger and John Cleveland, L0SS-0F-C00LANT ACCIDENT EXPERIMENT AT THE AVR GAS-COOLED REACTOR, CONF-8911401 Baeumer, R., et al., 1990. AVR - experimental high-temperature reactor. 21 years of successful operation for a future energy technology.

BFS:

The Institute of Physics and Power Engineering (IPPE) in Russia for conducted reactor physics experiments in four critical assemblies were constructed in the IPPE BFS-1 or BFS-2

73 facilities (called BFS-73-1, BFS-75-1, BFS-76-1A, and BFS-109-2A). These represented either the metal uranium fuel (U-10Zr) loaded SFR concept or the current PGSFR design (BFS 1A). (BFS stands for Bolshoy Fizicheskiy Stand which translates to Big Physical Facility in English.)

References:

Briggs, J Blair, Tsibulya, Anatoly, Matveenko, Igor, and Rozhikhin, Yevgeniy. BFS, a Legacy to the International Reactor Physics, Criticality Safety, and Nuclear Data Communities. United States: N. p., 2012.

BN-600: The BN-600 reactor is a sodium-cooled fast breeder reactor with a core power of 1470 MWt (600 Mwe). The plant is a pool type LMFBR, where the reactor, coolant pumps, intermediate heat exchangers and associated piping are all located in a common liquid sodium pool. The reactor system is housed in a concrete rectilinear building and provided with filtration and gas containment.

Reference:

INTERNATIONAL ATOMIC ENERGY AGENCY, BN-600 Hybrid Core Benchmark Analyses, IAEA-TECDOC-1623, IAEA, Vienna (2010)

B&W steam generator tube bundle: The steam generator tube bundle was a full-scale helical tube bundle representing a 150-degree sector of the outer layers of a steam generator. It consisted of a rectangular array of tubes 5-1/2 tubes wide by 16 tubes deep. The tubes were installed in a 4000-gpm water flow loop and the 5 outer layers were allowed to vibrate freely.

Reference:

R. P. Glasser, Experimental Evaluation of Helical Consolidated Nuclear Steam Generator (CNSG) Tubes and Supports, Tech. Rep. MA-RD-920-76019, The Babcock and Wilcox Company, November 1975.

CEA-Supercavna: The CEA-Supercavna facility was designed to investigate the interaction of sodium flow and thermal stratification. These phenomena are important in the hot sodium pool of a SFR upper plenum. The facility consists of a rectangular cavity with heated side walls, and flow is driven by a wall-bounded cold jet at the bottom of the cavity.

References:

R. Vidil, D. Grand and F. Leroux, Interaction of recirculation and stable stratification in a rectangular cavity filled with sodium, Nuclear Engineering and Design 105 321-332, 1988.

74 Vidil, R., Astegiano, J.C., Grand, Marechal, A., 1981. Experimental and numerical studies of mixed convection in a cavity, Cases of sodium and water, 7th International Heat Transfer Conference, Munich, Germany.

CEFR: The Chinese Experimental Fast Reactor (CEFR) is a 65 MWt sodium cooled fast reactor. The design has a high neutron leakage core fueled with uranium oxide and stainless-steel radial reflector. Neutronics Benchmark of CEFR Start-Up Tests will include evaluation of the criticality, control rod worth, reactivity effects, and neutron spectral characteristics. The recorded experimental data from the CEFR start-up provide an opportunity for validation of the physical models and neutronics simulation codes.

References:

Huo, X., Hu, Y., Chen, Xiaoliang, Xu, L., Zhang, J., Chen, Xiaoxian, Cao, P., 2019.

Technical Speci"cations for neutronics benchmark of CEFR start-up tests (CRP-I31032).

KY-IAEA-CEFRCRP-001, Department of Reactor Engineering, China Institute of Atomic Energy.

Huo, X., Yu, H., Hu, Y., Yang, X., Yang, Y., Chen, Y., ZHOU, K., FAN, Z., CHEN, X., XU, L., ZHANG, J., 2017. IAEA CRP Proposal for Benchmark Analysis on Physical Start-Up Experiments of China Experimental Fast Reactor, in: IAEA FR2017, IAEA-CN245-501.

Ekaterinburg, Russian Federation.

CIET: The Compact Integral Effects Test (CIET) facility investigates the thermal-hydraulic behavior of "uoride salt-cooled high-temperature reactors (FHRs) under forced-and natural-circulation operation. CIET tests are used for natural-circulation-driven decay heat removal in FHRs, under a set of reference licensing basis events. A simulant "uid, DOWTHERM A oil, is used and scaling evaluations show it is a good simulant of high temperature salts.

Reference:

N. Zweibaum, Z. Guo, J. C. Kendrick, and P. F. Peterson, Design of the Compact Integral Effects Test Facility and Validation of Best-Estimate Models for Fluoride Salt-Cooled High-Temperature Reactors, NUCLEAR TECHNOLOGY

• VOLUME 196

• 000 -000

• DECEMBER 2016.

C5G7: This is OECD/NEAs 2-D and 3-D MOX/UOX small core neutronic benchmark problem.

Reference:

E. E. Lewis, et al., Benchmark Specifications for Deterministic MOX Fuel Assembly Transport Calculation without Spatial Homogenization, NEA, 2003.

75 DUFF: The Demonstration Using Flattop Fissions (DUFF) reactor was a LANL design that was built and tested. DUFF included a heat pipe and power conversion system to couple to Flattop with the end goal of demonstrating electrical power production using nuclear technology was applicable to space application.

Reference:

Gibson, Marc A.; Briggs, Maxwell H.; Sanzi, James L.; Brace, Michael H. (December 2013). "Heat Pipe Powered Stirling Conversion for the Demonstration Using Flattop Fission (DUFF) Test" Nuclear and Emerging Technologies for Space (NETS-2013).

EBR-II (SHRT): The EBR-II plant is located in Idaho and operated between 1964 and 1994, and was used for experiments designed to demonstrate passive safety in liquid metal reactors.

It was rated for a thermal power of 62.5 MW with an electric output of approximately 20 MW.

The Shutdown Heat Removal Test (SHRT) program was carried out with the objectives to support plant design, provide test data for validation of computer codes for design, licensing and operation of LMRs, and demonstrate passive reactor shutdown and decay heat removal in response to protected and unprotected transients.

References:

T. Sumner and T. Y. C. Wei, Benchmark Specifications and Data Requirements for EBR-II Shutdown Heat Removal Tests SHRT-17 and SHRT-45R, ANL-ARC-226 (Rev 1), 2012.

EBR-II (X447):

Fuels Irradiation & Physics Database (FIPD) and the Integral Fast Reactor Materials Information System (IMIS) to supply PIE data for comparison with simulations of EBR-II experiments X447/X447A.

Reference:

A. Oaks, K. Mo, W. Mohamed, A. Yacout, "Development of FIPD: The EBR-II fuels irradiation & physics database," Top Fuel 2019: Light Water Reactor Fuel Performance Conference (2019).

EBR-II (X423): Fuels Irradiation & Physics Database (FIPD) and the Integral Fast Reactor Materials Information System (IMIS) to supply PIE data for comparison with simulations of EBR-II experiments X447/X447A.

Reference:

76 A. Oaks, K. Mo, W. Mohamed, A. Yacout, "Development of FIPD: The EBR-II fuels irradiation & physics database," Top Fuel 2019: Light Water Reactor Fuel Performance Conference (2019).

ENEA-CIRCE: CIRCE is a large-scale test facility designed for verification of key operating principles of the80 MW Experimental Accelerator-Driven System (XADS) currently being designed in Italy.

Reference:

P. Turroni, et al., The CIRCE Test Facility, ANS Winter Meeting, Reno, Nevada, USA, 2001.

ENEA-NACIE: A 19-pin wire-wrapped facility using LBE as a coolant. Rods can be heated in multiple configurations. The fuel-pin test section is instrumented with 67 thermocouples at different elevations along the heated length, both attached to the walls of selected pins as well as in the center of some subchannels.

Reference:

Di Piazza, M. Angelucci, R. Marinari, M. Tarantino, N. Forgione, Heat transfer on HLM cooled wire-spaced fuel pin bundle simulator in the NACIE-UP facility, Nucl. Eng. Des., 300 (2016), pp. 256-267.

EMPIRE: The Empire core design was developed at LANL to fill the need for a design test bed for a heat pipe cooled microreactor. The 2 MWt core is composed of 18 unit assemblies.

Control drums in the reflector provide long-term reactivity control, while a shutdown rod in the center of the core provides a means of rapid reactivity control.

Reference:

A.J. FALLGREN, D. V. RAO, and H. TRELLUE, Heatpipe Reactor Design for Special Purpose Applications, Los Alamos National Laboratory (2018).

Christopher Matthews, Vincent Laboure, Mark DeHart, Joshua Hansel, David Andrs, Yaqi Wang, Javier Ortensi & Richard C. Martineau (2021) Coupled Multiphysics Simulations of Heat Pipe Microreactors Using DireWolf, Nuclear Technology, 207:7, 1142-1162, DOI:

10.1080/00295450.2021.1906474 "eVinci - like: Westinghouse is developing the eVinci microreactor which is expected to have an electric power output up to 15 MW with more than three years without refueling. It features

77 High-Assay Low--Enriched Uranium (HALEU) TRISO (TRi-structural ISOtropic) fuel. The core design is constituted by a solid monolithic block with tree types of channels for fuel, neutron moderators and heat pipes. The model developed and evaluated is based on public information, and only approximates the Westinghouse design.

Reference:

Arafat, Y., Van Wyk, J., 2019. eVinci micro reactor-our next disruptive technology.

Nucl. Plant J. 34 - 37.

Alex Levinsky, Jurie J. van Wyk, Yasir Arafat, Matthew C. Smith, "Westinghouse eVinci Reactor for Off-Grid Markets," Transactions of the American Nuclear Society, Vol. 119, Orlando, Florida, November 11-15, 2018.

FFTF: The Fast Flux Test Facility (FFTF) is a 400 MW, oxide fueled, liquid sodium test reactor.

The reactor has three primary and secondary loops. Reactor heat is removed by 12 air-cooled dump heat exchangers (DHX). Loss of forced cooling tests were conducted in the FFTF facility and are the subject of an IAEA international benchmark exercise.

References:

Moisseytsev, A. Rivas, T. Kim, T. Sumner, and N. Stauff, FY20 FFTF Benchmark Analysis, ANL-ART-208, 2020.

T. Sumner, et. Al. Benchmark Specification for FFTF LOFWOS Test #13, ANL-ART-102 (Rev 1), Nuclear Engineering Division, Argonne National Laboratory, December 22, 2017.

Gallium Melting Benchmark:

Study that examined flowing and freezing in initially emptied pipes.

Reference:

Somers-Neal, S., Pegarkov, A., Matida, E., Tang, V., Kaya, T., 2021. Experimental and numerical investigation of solidification of gallium in an initially emptied horizontal pipe flow.

J. Nucl. Eng. Radiat. Sci. 7 (3).

Fort St Vrain:

The Fort St. Vrain Nuclear Power Plant was a commercial nuclear power station located in northern Colorado, which operated from 1979 to 1989. It was a 840 MWt (330 Mwe) high-temperature gas reactor. The reactor fuel was a combination of fissile uranium and fertile thorium microspheres dispersed within a prismatic graphite matrix.

Reference:

78 H.L. Brey, Fort St. Vrain Operation and Future Energy, 16 (1-2) (1991), pp. 49-58.

Godiva: Godiva was an assembly at LANL that provides data for criticality calculations and has been used in several neutronic code assessments. It was fueled with 93% HEU.

References:

G. E. Hansen and II. C. Paxton, "Reevaluated Critical Specifications of Some Los Alamos Fast-Neutron Systems," Los Alamos National Laboratory report LA-4208 (1969).

H. C. Paxton, "Los Alamos Critical Data," Los Alamos National Laboratory Report LA-3067-MS, 1975.

gPBR-200: The gPBR-200 is a 200 MW general pebble bed reactor that is based on current and past designs such as the Xe-100, HTR-PM, PBMR-400. It is used for development of code capabilities for a pebble bed design.

Reference:

R. Stewart, D. Reger, and P. Balestra, Demonstrate Capability of NEAMS Tools to Generate Reactor Kinetics Parameters for Pebble Bed HTGRs Transient Modeling, INL/EXT-21-64176, Idaho National Laboratory, 2021.

HELIOS: The LACANES benchmark deals with the characterization of the local pressure losses, due to form and frictional losses, at the HELIOS facility for lead alloy coolants. Several benchmark participants calculated the losses measured during an isothermal forced convection case.

Reference:

Jaeger, W., Sanchez-Espinoza, V. H., "Uncertainty and sensitivity analysis for the HELIOS loop within the LACANES benchmark," Proceedings of the 2010 International Congress on Advances in Nuclear Power Plants - ICAPP '10, 2010.

HTTF: The High Temperature Test Facility (HTTF) is a one-quarter scale integral test facility representing the Modular High Temperature Gas Reactor (MHTGR). The HTTF core consists of 10 hexagonal core blocks made from a cast ceramic. The facility is electrically heated by a network of 210 graphite rods with a total power of approximately 2.2 MWt.

Reference:

79 Brian Woods, OSU High Temperature Test Facility Design Technical Report, Revision 2, DOE-OSU-14517-116-10244, 2019.

HTR-10: The HTR-10 is a 10 MWt pebble bed reactor in China that provides startup test data for neutronics code assessment. The core contains approximately 27000 spherical pebbles with a 60 mm diameter. Active core cooling is not necessary for decay heat removal during accidents. It is quite suf"cient to discharge the decay heat by means of passive heat transport mechanisms (such as heat conduction, radiation, nature convention) to a simple cavity cooler outside the reactor pressure vessel.

References:

Z. Wu, D. Lin, and D. Zhong, The design features of the HTR-10, Nuclear Engineering and Design, Vol. 218, No. 1-3, pp. 25-32, 2002.

Gao Zuying, and Shi Lei, Thermal hydraulic transient analysis of the HTR-10, Nuclear Engineering and Design 218 (2002) 65-80.

HTR-PM: There are two HTR-PM reactors at the Shidao Bay power plant in China. These are gas-cooled pebble bed reactors with a power of 250 MWt. The core consists of a cylindrical pebble bed region that is surrounded by upper, lower, and radial reflectors. The core is randomly packed with approximately 420,000 fuel pebbles containing TRISO particles.

References:

Z. Wu, D. Lin, and D. Zhong, The design features of the HTR-10, Nuclear Engineering and Design, Vol. 218, No. 1-3, pp. 25-32, 2002.

Y. Zheng, M. M. Stempniewicz, Z. Chen, and L. Shi, Study on the DLOFC and PLOFC accidents of the 200 Mwe pebble-bed modular high temperature gas-cooled reactor with TINTE and spectra codes, Annals of Nuclear Energy, vol. 120, pp. 763-777, 2018.

D. Tian, L. Shi, L. Sun, Z. Zhang, Z. Zhang, and Z. Zhang, Installation of the Graphite Internals in HTR-PM, Nuclear Engineering and Design, vol. 363, p. 110585, 2020.

Zhang, J., Li, F., Sun, Y., 2017. Physical analysis of the initial core and running-in phase for pebble-bed reactor HTR-PM. Sci. Technol. Nucl. Inst. 8918424.

HTR-PM Hot Gas Mixing Experiment: A hot gas mixing experiment of HTR-PM reactor core outlet is proposed to measure and analyze the actual heat mixing performance and flow resistance property of this mixing structure. The design criteria and parameters of the hot gas mixing experiment are determined according to similarity criterion.

Reference:

80 Zhou Yangping, Li Fu, Hao Pengfei, He Feng, Shi Lei, Thermal hydraulic analysis for hot gas mixing structure of HTR-PM, Nuclear Engineering and Design, Volume 271, May 2014, Pages 510-514.

HTR-PROTEUS: PROTEUS was a zero-power research reactor located at the Paul Scherrer Institute (PSI) configured as a pebble bed reactor (PBR) critical facility and given the designation HTR-PROTEUS. It provided experimental benchmark data to support computational assessment of high-temperature gas-cooled reactors (HTGRs). The series consisted of 17 critical configurations and various reactor physics measurements.

References:

T. WILLIAMS, LEU-HTR PROTEUS: Configuration and Critical Balances for the Cores of the HTR-PROTEUS Experimental Programme, TM-41-95-18, Paul Scherrer Institute (1996).

D. MATHEWS and T. WILLIAMS, LEU-HTR PROTEUS System Component Description, TM-41-93-43, Paul Scherrer Institute (1996).

J. D. Bess, HTR-PROTEUS Pebble Bed Experimental Program Cores 9 and 10: Columnar Hexagonal Point-On-Point Packing with a 1:1 Moderator-to-Fuel Pebble Ratio, PROTEUS-GCR-EXP-004, in International Handbook of Evaluated Reactor Physics Benchmark Experiments, NEA/ NSC/DOC(2006)1, OECD.

HTTR: The HTTR is a helium gas cooled and graphite-moderated HTGR with the rated thermal power of 30 MW and the maximum reactor outlet coolant temperature of 950 °C. Tests, including loss of forced cooling tests, were conducted to demonstrate shutdown and inherent safety. The HTTR system consisted of the reactor pressure vessel, main cooling system and helium-helium heat exchanger. The reactor core was graphite moderated with prismatic fuel elements.

References:

S. SAITO, et al., Design of High Temperature Engineer-ing Test Reactor (HTTR), JAERI 1332, Japan Atomic Energy Research Institute (1994).

M. SHINDO, F. OKAMOTO, K. KUNITOMI, S. FU-JITA, and K. SAWA, Safety Characteristics of the High Temperature Engineering Test Reactor, Nuclear Engineering and Design, 132, 39 (1991).

81 HTTR-VCS: The vessel cooling system (VCS) in the HTTR corresponds to the reactor cavity cooling system (RCCS), and it mainly cools the reactor pressure vessel (RPV) by thermal radiation. Tests with the VCS compliment LOFC test without VCS.

Reference:

K. Kunitomi, S. Nakagawa, M. Shinozaki, Passive heat removal by vessel cooling system of HTTR during no forced cooling accidents, Nuclear Engineering and Design, Volume 166, Issue 2, 2 October 1996, Pages 179-190.

HTTU: The non-nuclear High Temperature Test Unit (HTTU) facility was designed in investigate thermal-fluid phenomena in a pebble bed. The test section consisted of approximately 25000 graphite spheres randomly packed in an annular configuration. The spheres had an outer diameter of 60 mm. Heat was supplied in the inner reflector and cooled by a water jacket in the outer reflector.

References:

Rousseau, P.G., van Staden, M., 2008. Introduction to the PBMR heat transfer test facility. Nuclear Engineering and Design 238, 3060-3072.

P.G. Rousseau, C.G. du Toit, W. van Antwerpen, H.J. van Antwerpen, Separate effects tests to determine the effective thermal conductivity in the PBMR HTTU test facility, Nuclear Engineering and Design 271 (2014) 444-458.

IAEA Fast Reactor Working Group Verification Problems:

Reference:

IAEA. Verification and validation of LMFBR static core mechanics codes part i. Technical Report IWGFR/75, International Atomic Energy Agency, 1990.

IAEA 3D PWR: Neutronic benchmark for 3D core with 177 fuel assemblies and 64 reflector assemblies. The active height of a fuel assembly is 340 cm with 20 cm think axial reflectors each.

82

Reference:

Benchmark Problem Book, ANL-7416, Suppl. 2, Argonne National Laboratory, 1977.

IFR-FBTA:

The Integral Fast Reactor (IFR) program, active from 1984 to 1994, produced a large collection of data relating to the performance of metallic fuels in sodium-cooled fast reactors (SFRs). To evaluate fuel/cladding compatibility, tests were conducted under both steady-state and transient conditions.

Reference:

Hanchung Tsai, "A VERSATILE APPARATUS FOR STUDYING IRRADIATED FUEL BEHAVIOR," CONF-891103-54, (1989).

IFR-WPF: A series of ex-reactor heating tests on low burnup U-26wt. %Pu-10wt. %Zr metallic fuel for the PRISM reactor was conducted to evaluate fuel/cladding metallurgical interaction and its effect en cladding integrity at elevated temperatures.

Reference:

Hanchung Tsai, Yung Y. Liu, Da-Yung Wang, and John. M. Kramer, "BEHAVIOR OF LOW-BURNUP ME7ALLIC FUELS FOR THE INTEGRAL FAST REACTOR AT ELEVATED TEMPERATURES IN EX-REACTOR TESTS," ANL/CP-11935, July 1991.

Ivanoskii HP: The Ivanovskii experiments directly measured the vapor temperature in a 70 cm long sodium heat pipe operating at 315 W and 845 W at approximately 820 K vapor temperature.

Reference:

M.N. Ivanovskii, V.P. Sorokin, I.V. Yagodkin, The Physical Principles of Heat Pipes, translation, Oxford University Press, New York, NY, USA (1982).

JAEA-PLANDT: A sodium test loop to support tests for wire-wrapped fuel assembly mixing and heat transfer, inter-wrapper heat removal, and thermal striping.

References:

83 A. Ono, et al., An experimental study on natural circulation decay heat removal system for a loop type fast reactor, J. Nucl. Sci. Tech., Vol.53, No.9, pp.1385-1396, 2016.

H. Kamide, et al., Investigation of core thermohydraulics in fast reactors

- interwrapper flow during natural circulation, Nucl. Tech., Vol.133, pp.77-91, 2001.

KAIST-3A: 3-D MOX neutronic benchmark problem.

Reference:

N. Z. Cho, Benchmark Problem 3A: MOX Fuel-Loaded Small PWR Core, MOX Fuel with Zoning, 7 Group Homogenized Cells, KAIST, 2000.

KIT-KALLA: The Karlsruhe Liquid metal Laboratory KALLA located at the Institute for Nuclear and Energy Technologies (IKET) of the Forschungszentrum Karlsruhe (FZK) summarizes experimental activities using heavy liquid metals (HLM) as operation fluid. The experimental field covered by KALLA necessitates a modular structure consisting of several individual facilities, in which single HLM specific issues are investigated experimentally. Loop experiments include THESYS (Technologies for HEavy metal SYStems) and THEADES (THErmalhydraulics and Ads DESign).

Reference:

J. Pacio, M. Daubner, F. Fellmoser, T. Wetzel, Experimental study of the influence of inter-wrapper flow on liquid-metal cooled fuel assemblies," Nuclear Engineering and Design, Volume 352, October 2019.

KRUSTY: The Kilopower Reactor Using Stirling Technology (KRUSTY) experiment provides data for a heat pipe cooled reactor. KRUSTY was operated at powers up to 5.5 kWt and includes data for several transients.

References:

Poston, David I., et al, KRUSTY reactor design, Nuclear Technology 206.sup1 (2020):

S13-S30 Poston, David I., et al. "Results of the KRUSTY nuclear system test." Nuclear Technology 206.sup1 (2020): S89-S117

84 KTH-TALL: The Karlsruhe Liquid metal Laboratory KALLA located at the Institute for Nuclear and Energy Technologies (IKET) of the Forschungszentrum Karlsruhe (FZK) summarizes experimental activities using heavy liquid metals (HLM) as operation fluid. The experimental field covered by KALLA necessitates a modular structure consisting of several individual facilities, in which single HLM specific issues are investigated experimentally. Loop experiments include THESYS (Technologies for HEavy metal SYStems) and THEADES (THErmalhydraulics and Ads DESign).

Reference:

D. Grishchenko, A. Papukchiev, C. Liu, C. Geffray, M. Polidori, K. Kp, M. Jeltsov, P.

Kudinov, "TALL-3D open and blind benchmark on natural circulation instability," Nuclear Engineering and Design, Volume 358, March 2020.

LANL HP: Numerous tests on heat pipe performance have been conducted at Los Alamos National Laboratory. The reference below cites many of those tests.

Reference:

Sweetland, Katrina, "Heat Pipes with Arbitrary Boundary Conditions. " PhD diss., University of Tennessee, 2022. https://trace.tennessee.edu/utk_graddiss/7419 MARVEL: The sodium-potassium-cooled microreactor is designed to generate 85 kilowatts of thermal energy. The fuel for MARVEL is similar to the TRIGA fuel used in university reactors for research and hands-on training.

Reference:

U.S. Department of Energy, Draft Environmental Assessment for the Microreactor Applications Research, Validation and Evaluation (MARVEL) Project at Idaho National Laboratory, DOE/EA-2146, 2021.

MHTGR-350: The MHTGR-350 is a prismatic HTGR design developed by General Atomics.

The reactor vessel contains the reactor core, reflectors and associated neutron control systems, core support structures, and shutdown cooling heat exchanger and motor-driven circulator. The steam generator vessel houses a helically coiled steam generator bundle as well as the motor-driven main circulator. The reactor vessel is uninsulated to provide for decay heat removal under loss-of-forced-circulation conditions. In such events, heat is transported to the passive

85 Reactor Cavity Cooling System (RCCS), which circulates outside air by natural circulation within enclosed panels surrounding the reactor vessel.

References:

A.J. Neylan, D.V. Graf, A.C. Millunzi, The Modular High Temperature Gas-Cooled Reactor (MHTGR) in the U.S., Nuc. Eng. & Des. Vol. 109, pp.99-105, 1988.

Sung Nam Lee, Nam-il Tak, Hong-Sik Lim, Chang Keun Jo, Volkan Seker, Thomas J.

Downar, "Benchmark calculations for verification of thermo-fluid analysis codes using simplified model of the OECD/NEA MHTGR-350 core design," Annals of Nuclear Energy, Volume 139, May 2020.

MiGaDome: The Michigan Multi-Jet Gas-Mixture Dome (MiGaDome) is a model of an HTGR upper plenum. The purpose of the facility is to study mixing in large enclosures widely seen in nuclear reactor designs and to provide high-resolution experimental data for code validation and development.

References:

S. Che, J. Mao, X. Sun, A. Manera, and V. Petrov, Design of high-resolution experiments for extended LOFC accidents in HTGRs, Proceedings of the 19th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-19) (2022)

A. Manera et al., Challenge problem 3: benchmark specifications for mixing in large enclosure and thermal stratification, Tech. Rep. ANL/NSE-22/8 (2022)

A. Manera et al., NEAMS IRP Challenge Problem 3: Mixing in Large Enclosures and Thermal Strati-fication, Proc. 20th Int. Topl. Mtg. Nuclear Reactor Thermal Hydraulics (NURETH-20) (2023)

Missouri S&T Air Experiments: A test facility was constructed at Missouri S&T to study the gas dynamics and heat transfer characteristics in a packed pebble bed for Gen-IV HTGRs.

Among investigated topics, the pressure drop through pebble beds is the main research objective of one of the separate-effect tests.

References:

86 R. Abdullmohsin, Gas Dynamics and Heat Transfer in a Packed Pebble-Bed Reactor for the 4th Generation Nuclear Eenergy, Doctoral Dissertation, Missouri University of Science and Technology, Fall 2013.

R. Abdullmohsin and M. H. Al-Dahhan, Pressure Drop and Fluid Flow Characteristics in a Packed Pebble Bed Reactor, Nuclear Technology, 198:17-25, 2017.

Monju: Monju was a sodium cooled, MOX-fueled, loop-type reactor with three primary coolant loops, designed to produce 280 MWe from 714 MWt. It had a breeding ratio of approximately 1.2.

Natural circulation tests were conducted in Monju.

Analysis of the occurrence of thermal strati"cation in the upper plenum of the prototype fast breeder reactor (FBR) Monju is the subject of a Coordinated Research Programs (CRP) organized by IAEA.

Reference:

Yoshikawa, S., Minami, M., 2008. Data Description for Coordinated Research Project on Benchmark Analyses of Sodium Natural Convection in the Upper Plenum of the MONJU Reactor Vessel under Supervisory of Technical Working Group on Fast Reactors.

International Atomic Energy Agency, JAEA-Data/Code 2008-024.

Sofu, T., Thomas, J., 2011. Analysis of thermal strati"cation in the upper plenum of the Monju reactor vessel. In: Proc. of the 14th International Meeting on Nuclear Thermalhydraulics, Toronto, Canada, NURETH-14-510.

Hiroyasu Mochizuki, Hiroki Yao, "Analysis of thermal strati"cation in the upper plenum of the Monju reactor," Nuclear Engineering and Design 270 (2014) 48-59.

MSRE: The Molten-Salt Reactor Experiment (MSRE) was a 7.4 MWt experimental molten fuel salt reactor at the Oak Ridge National Laboratory (ORNL) that operated between 1965 and 1969. The experimental program included a variety of criticality and reactivity measurements at zero-and full-power operation for molten-salt fuels using 235U or 233U as fissile material.

References:

P. Haubenreich, J. Engel, Experience with the molten-salt reactor experiment, Nucl. Appl.

Technol., 8 (2), 1970.

Robertson, R.C., "MSRE Design and Operations Report Part I, Description of the Reactor Design," ORNL-TM-728, 1965.

87 MSTDB-TC:

The Molten Salt Thermal Properties Database Thermochemical (MSTDBTC) has been developed to support thermodynamic modeling of fluoride-and chloride-based systems for molten salt reactors. MSTDBTC Ver. 1.2 contains models for 96 pseudo-binary systems, 37 pseudo-ternary systems, 42 solid solutions, 229 stoichiometric compounds, and 130 gaseous species.

References:

Besmann, T.M. Schorne-Pinto, J. Developing Practical Models of Complex Salts for Molten Salt Reactors. Thermo 2021, 1, 168-178. https://doi.org/10.3390/thermo1020012 Johnathon C. Ard, Jacob A. Yingling, Kaitlin E. Johnson, Juliano Schorne-Pinto, Mina Aziziha, Clara M. Dixon, Matthew S. Christian, Jacob W. McMurray, Theodore M.

Besmann, Development of the Molten Salt Thermal Properties Database Thermochemical (MSTDBTC), example applications, and LiClRbCl and UF3UF4 system assessments, Journal of Nuclear Materials, Volume 563, May 2022.

Ard, J.C.; Yingling, J.A.; Johnson, K.E.; Schorne-Pinto, J.; Christian, M.S.; McMurray, J.W.;

Besmann, T.M. Development and Applications of the Molten Salt Thermal Properties Database-Thermochemical (MSTDB-TC). (Currently in preparation.)

NEAMS Gas-Cooled Microreactor: This is a NEAMS-designed gas cooled microreactor assembly designed by Nicolas Stauff & Ahmed Abdelhameed at ANL. Specifications are documented in the VTB model as well as in the references below.

References:

N. Stauff, A. Abdelhameed, Y. Cao, Kristina, Y. Miao, K. Mo and D. Nunez.

"Multiphysics Analysis of Load Following and Safety Transients for MicroReactors."

Argonne National Laboratory, ANL/NEAMS-22/1, September 30, 2022.

Ahmed Abdelhameed, Yan Cao, Daniel Nunez, Yinbin Miao, Kun Mo, Changho Lee, Emily Shemon, Nicolas E. Stauff High-Fidelity Multiphysics Modeling of Load Following for 3-D Gas-Cooled Microreactor Assembly using NEAMS Codes ANS Winter 2022 meeting, November (2022)

VTB (assembly model):

https://mooseframework.inl.gov/virtual_test_bed/microreactors/gcmr/index.html

88 NEAMS Heat Pipe Cooled Microreactor: This is a NEAMS-designed heat pipe cooled microreactor core, designed by ANL as a modelling exercise. The model of a 1/6 full-core heat pipe cooled microreactor concept is intended to demonstrate multiphysics transient simulations including a load following transient and a scenario initiated by the forced failure of a single heat-pipe.

References:

N. Stauff, A. Abdelhameed, Y. Cao, Kristina, Y. Miao, K. Mo and D. Nunez.

"Multiphysics Analysis of Load Following and Safety Transients for MicroReactors."

Argonne National Laboratory, ANL/NEAMS-22/1, September 30, 2022.

VTB (full core model):

https://mooseframework.inl.gov/virtual_test_bed/microreactors/mrad/index.html NEAMS LFR Assembly: This is a lead-cooled fast reactor fuel assembly based on Westinghouses LFR inner fuel assembly prototype core from a few years ago. Details are in the following reference.

References:

G. Grasso, A. Levinsky, F. Franceschini, and P. Ferroni. A mox-fuel core configuration for the Westinghouse lead fast reactor. Proceedings of International Congress on Advances in Nuclear Power Plants, France, May, 2019.

VTB (assembly model):

https://mooseframework.inl.gov/virtual_test_bed/lfr/heterogeneous_single_assembly_3D/Gr iffin_standalone_LFR.html NSTF: The Natural convection Shutdown heat removal Test Facility (NSTF) is a large scale thermal hydraulics test facility that has been built at Argonne National Laboratory (ANL). The facility has principally been designed for testing of Reactor Cavity Cooling System (RCCS) concepts that rely on natural convection cooling for either air or water-based systems.

Reference:

89 D. Lisowski, M. Farmer, et al., Design Report for the 1/2 Scale Air-Cooled RCCS Tests in the Natural convection Shutdown Heat Removal Test Facility (NSTF), ANL-SMR-8, Argonne National Laboratory, June 2014.

D. Lisowski, C. Gerardi, D. Kilsdonk, et al., Final Project Report on RCCS Testing with the Air-based NSTF, ANL-ART-47, Argonne National Laboratory, August 2016.

Q. Lv et al., Testing and Modeling of a Large Scale RCCS Operating at Single-Phase Normal Conditions, 19th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-19), Brussels, Belgium, 2022.

Q. Lv et al., Testing of A Large Scale RCCS Operating at Two Phase Accident Conditions, in Proceedings of Advances in Thermal Hydraulics (ATH2022), pp. 796-810, 2022.

OECD/NEA T-Junction: A thermal mixing experiment featuring a pipe t-junction where hot and cold legs meet. Velocity and temperature measurements are taken downstream of the junction to provide data to evaluate predictions of turbulent mixing.

Reference:

B.L. Smith, et al., "Report of the OECD/NEA-Vattenfall T-Junction Benchmark exercise,"

NEA/CSNI/R(2011)5, 2011.

ORNL (THORS) 19-pin:

The ORNL Thermal-Hydraulic Out-of-Reactor Safety facility (THORS) 19-pin tests were used for studying the thermal-hydraulic flow characteristics in SFR assemblies. The measurements included distribution of temperatures at the exit of the fuel assembly, duct walls and rod bundle. Bundles both with and without blockages were investigated.

References:

Fontana, M.; MacPherson, R.; Gnadt, P.; Parsly, L.; Wantland, J. Temperature distribution in the duct wall and at the exit of a 19-rod simulated LMFBR fuel assembly (FFM Bundle 2A). Nucl. Technol. 1974, 24, 176-200.

J.T.Han, Blockages in LMFBR Fuel Assemblies - A Review of Experimental and Theoretical Studies, ORNL/TM-5839, 1977.

90 M. H. Fontana et al., Effect of Partial Blockages in Simulated LMFBR Fuel Assemblies, ORNL/TM-4324 (December 1973).

M. H. Fontana et al., Thermal-Hydraulic Effects of Partial Blockages in Simulated LMFBR Fuel Assemblies with Applications to the CRBR, ORNL/TM-4779 (July 1975).

ORNL (THORS) 61-pin: Boiling tests were conducted in a 61-pin bundle (Bundle 9) in the (ORNL)Thermal-Hydraulic Out-of-Reactor Safety facility (THORS)

References:

Gnadt, P A, Anderson, A H, Clapp, N E, Montgomery, B H, Collins, C W, and Stulting, R D. THORS: a high-temperature sodium test facility rated at 2. 0 MW. [LMFBR]. CONF-790816-24, 1979.

P.A. Gnadt, J.J. Carbajo, J.F.

Dearing,

A.E. Levin, R.E. MacPherson, B.H. Montgomery, S.

D. Rose, R.H. Thornton, J.L. Wantland, Sodium boiling experiments in the THORS facility, Nuclear Engineering and Design, Volume 82, Issues 2-3, 2 October 1984, Pages 241-280.

ORNL - LSTL: The ORNL liquid salt test loop (LSTL) is a test facility for the development and demonstration of high-temperature fluoride-salt technology. The major components include a centrifugal pump to circulate the salt, a salt-to-air heat exchanger, three tanks, pressure control and trace heating systems, and associated instrumentation.

Reference:

G.L. Yoder, et al., High-Temperature Fluoride Salt Test Loop, Oak Ridge National Laboratory (2015), ORNL/TM-2012/430.

OSU-LTDF: The Low Temperature DRACS Facility (LTDF) was therefore designed and built at the Ohio State University. Distilled water was used as a surrogate for the FLiBe in the primary loop and KF-ZrF4 in the DRACS loop (i.e., the secondary loop in the LTDF). Data from the LTDF was later used for assessment of SAM and RELAP.

91

Reference:

Lv, Q., H.C. Lin, I.H. Kim, X. Sun, R.N. Christensen, T.E. Blue, G.L. Yoder, D.F. Wilson, P.

Sabharwall, DRACS thermal performance evaluation for FHR, Annu. Nucl. Energy 77, 115-128, (2015).

Lv, Q., H.C. Lin, S. Shi, X. Sun, R.N. Christensen, T.E. Blue, G.L. Yoder, D.F. Wilson, P.

Sabharwall, Experimental Study of DRACS Steady-State and Transient Performance, In:

Proc. of ICAPP16, San Francisco, CA, April 17-20, (2016).

Lv, Q., H.C. Lin., X. Sun, R.N. Christensen, T.E. Blue, G.L. Yoder, D.F. Wilson, P.

Sabharwall, Experimental Study of DRACS Thermal Performance in a Low-Temperature Test Facility. Nucl. Tech. 196 (2), 319-337, (2016).

PBMR-400: The PBMR-400 benchmarks were an international exercise sponsored by the Nuclear Energy Agency (NEA) of the Organization for Economic Cooperation and Development (OECD) that investigated Pebble Bed Modular Reactor (PBMR) coupled neutronics/thermal-hydraulics transients.

Reference:

OECD/NEA/NSC PBMR Coupled Neutronic/Thermal Hydraulics Transient Benchmark The PBMR-400 Core Design, Draft V03, September 2005.

Phenix:

Phenix is a prototype pool type 250 MW SFR built by Commissariat l'Energie Atomique et aux Energies Alternatives (CEA) of France, and operated from 1973 to 2009. The end-of-life benchmark provides data on control rod efficiency and core power shape due to insertion and withdrawal of control rods.

Reference:

International Atomic Energy Agency, 2013. Benchmark Analyses on the Natural Circulation Test Performed during the PHENIX End-of-life Experiments, IAEA-TECDOC-1703, IAEA, Vienna.

D. Tenchine, D. Pialla, P. Gauthé, et al. Natural convection test in Phenix reactor and associated CATHARE calculation, Nucl. Eng. Des., 253 (none) (2012), pp. 23-31

92 PNL 7 Sleeve Blockage Benchmark: PNLs 7 x 7 sleeve blockage facility was designed to investigate the turbulent flow phenomena near postulated sleeve blockages in a model nuclear fuel rod bundle. The sleeve blockages were characteristic of fuel-clad swelling or ballooning, which could occur during loss-of-coolant accidents in pressurized-water reactors.

Reference:

Creer, J.; Rowe, D.; Bates, J.; Sutey, A. Effects of Sleeve Blockages on Axial Velocity and Intensity of Turbulence in an Unheated 7 x 7 Rod Bundle. [PWR]; Technical Report; Battelle Pacific Northwest Labs.: Richland, WA, USA, 1976.

Purdue H-Shaped Helium Loop: Experimental studies were conducted with an H-shaped test facility to investigate diffusion and air ingress in a natural circulation loop. Results suggested that diffusion may not be the dominant mass transfer process during the air ingress.

References:

Gould, Daniel, Franken, Daniel, Bindra, Hitesh, Kawaji, Masahiro, 2017, Transition from molecular diffusion to natural circulation mode air-ingress in high temperature helium loop, Ann. Nucl. Energy 107, 103-109.

Daniel Franken, Daniel Gould, Prashant K. Jain, Hitesh Bindra, Numerical study of air ingress transition to natural circulation in a high temperature helium loop, Annals of Nuclear Energy 111 (2018) 371-378.

RPI HP: A Low-Temperature Heat Pipe Test Facility (LTHPF) was designed and constructed at RPI to study heat pipe performance. The facility uses low-temperature working fluids, such as water, ethanol, ammonia, and FC-72, to simulate the phenomena of interest in high-temperature heat pipes. The LTHPF consists of a single heat pipe oriented in the vertical direction.

Reference:

Experimental Investigation of Heat Pipe Flow Dynamics and Performance. Ilyas Yilgor, Shanbin Shi. 20th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-20), 2023.

93 SAFE-30: The Safe Affordable Fission Engine (SAFE) was a small experimental nuclear fission reactors for electricity production in space. It was heat pipe cooled.

References:

Hrbud, Ivana; Van Dyke, Melissa; Houts, Mike; Goodfellow, Keith, End-to-End demonstrator of the Safe Affordable Fission Engine (SAFE) 30: Power conversion and ion engine operation, Space Technology and Applications International Forum (STAIF 2002).

Vol. 608. Albuquerque, New Mexico: American Institute of Physics. pp. 906-911.

R.S. Reid, J.T. Sena, and A.L. Martinez, Sodium Heat Pipe Module Test for the SAFE-30 Reactor Prototype Los Alamos National Laboratory report LA-UR-00-4728 (2000).

SANA: The SANA tests simulated heat transfer and coolant flow in a pebble bed geometry and represents an important assessment of the simulation of convection/conduction/radiation in a packed bed. The SANA facility consisted of a cylindrical steel vessel containing about 9500 non-fueled spheres, with either helium or nitrogen coolant and either 6 cm electric graphite pebbles, 3 cm matrix graphite pebbles, or 6.5 cm aluminum oxide pebbles.

Reference.:

Stocker, B., Niessen, H., 1996. Data Sets of the SANA Experiment 1994-1996. Technical Report, Forschungszentrum Julich.

SEFOR:

The Southwest Experimental Fast Oxide Reactor (SEFOR) reactor was designed to provide a Doppler measurement in an environment representative of an operating LMFBR, with respect to the neutron spectrum, the fuel temperature range, the reactor composition and the fuel microstructure. Standard fuel for SEFOR was mixed oxide.

Reference:

BROOKHAVEN NATIONAL LABORATORY, CROSS SECTION EVALUATION WORKING GROUP BENCHMARK SPECIFICATIONS, BNL 19302, VOL. II, ENDF-202, 1974.

https://www.nndc.bnl.gov/endfdocs/ENDF-202-Vol2.pdf SiMBA:

The Simplified Microreactor Benchmark Assessment (SiMBA) is for a conceptual design that Idaho National Laboratory developed based on the Empire specification and used

94 as the assessment problem. The SiMBA problem is a 2-MW microreactor composed of 18 hexagonal assemblies arranged into two rings. The tops and bottoms of these 160-cm-high assemblies are surrounded by 20-cm-high axial beryllium reflectors.

Reference:

Matthews, C., et al., 2021b. Coupled multiphysics simulations of heat pipe microreactors using DireWolf. Nucl. Technol. 207 (7), 1142-1162.

SNAP8-ER: This is a Systems for Nuclear Auxiliary Power (SNAP) space microreactor similar to the MARVEL design (NaK cooled, UZrH fuel).

References:

AID. Snap 8 summary report. 9 1973.

URL: https://www.osti.gov/biblio/4393793, doi:10.2172/4393793.

Samuel Garcia, Isaac Naupa, Dan Kotlyar, and Ben Lindley. Validation of snap8 criticality configuration experiments using serpent. In Proceedings of ANS Winter Conference. 1 2022.

Isaac Naupa, Samuel Garcia, Stefano Terlizzi, Dan Kotlyar, and Ben Lindley. Validation of snap8 criticality configuration experiments using neams tools. In Submitted to Proceedings of M&C. 1 2022 VTB: https://mooseframework.inl.gov/virtual_test_bed/microreactors/s8er/index.html SPR A Microreactor: This reference plant model was developed from the INL proposed Design A that was detailed in [Sterbentz et al, 2018].

Reference:

Sterbentz, J.W., et al, Preliminary Assessment of Two Alternative Core Design Concepts for the Special Purpose Reactor, Nuclear Science and Technology Div., Idaho National Laboratory, INL/EXT-17-43212, Rev. 1, Idaho Falls, ID, May 2018.

SPHERE: The Single Primary Heat Extraction and Removal Emulator (SPHERE) is a separate effects facility at Idaho National Laboratory (INL) designed to study the thermal performance of heat pipes under a wide range of heating values and operating temperatures. Gap conductance

95 tests obtained data on the heat losses through the annular gap between the outer wall of the heat pipe and the inner diameter of a stainless steel core block through radiative and conductive heat transfer with varying gas compositions

Reference:

Zachary Sellers, Minseop Song, Jeremy Hartvigsen, and Piyush Sabharwall, SPHERE Gap Conductance Test, INL/RPT-22-66992, September 2022.

SPHERE-GC: A second series of tests were conducted in the SPHERE facility with additional data on gap conductance between the heat pipe and surrounding core block.

Reference:

Currently in preparation.

TAMU P: Pebble bed pressure drop tests were conducted at Texas A&M University that provide data on hydraulic losses for flow through a porous matrix. The experimental setup was used to measure single-phase flow conditions in a packed bed. A cylindrical column was constructed out of PMMA and randomly packed with spherical PMMA beads. The column dimensions were 12.065-cm diameter and 152.4-cm height. Diameters of the spherical beads used as packing material were 0.635, 1.27, 1.905, and 3.302 cm. Air and water were used as working fluids.

References:

Yassin A. Hassan and Changwoo Kang (2012) Pressure Drop in a Pebble Bed Reactor Under High Reynolds Number, Nuclear Technology, 180:2, 159-173, DOI: 10.13182/NT12-A1463 C.W. Kang, Pressure Drop in a Pebble Bed Reactor, Masters Thesis, Texas A&M University, August 2010.

TAMU-FA: The Texas A&M University (TAMU) fuel assembly (FA) tests included adiabatic assemblies with unblocked channels, blockages on two interior subchannels, and blockages both on interior and exterior subchannels. Tests were performed on a 61-pin facility using P-

96 cymene as the surrogate fluid in a refractive index matched system, which enables time-resolved particle image velocimetry (TR-PIV) measurements of velocity.

References:

N. Goth et al., Comparison of experimental and simulation results on interior subchannels of a 61-pin wire-wrapped hexagonal fuel bundle, Nuclear Engineering and Design, 338, pp.

130-136 (2018); doi.org/10.1016/j.nucengdes.2018.08.002

[T. Nguyen, L. White, R. Vaghetto, and Y. Hassan, High-fidelity velocity measurements in a totally blocked interior subchannel of a wire-wrapped 61-pin hexagonal fuel bundle, Nuclear Engineering and Design, 353, pp. 110234 (2019);

doi.org/10.1016/j.nucengdes.2019.110234 M. Childs, R. Muyshondt, R. Vaghetto, D. T. Nguyen, and Y. Hassan, Experimental study on the effect of localized blockages on the friction factor of a 61-pin wire-wrapped bundle, Journal of Fluids Engineering, 142 (11) (2020); doi.org/10.1115/1.4048140 TAMU HP:

Heat pipe tests at Texas A&M University used optical fiber-distributed temperature sensors to measure the internal and external temperature distributions of a water-cooled heat pipe. See: (NEUP Project 20-19735) Experiments for Modeling and Validation of Liquid-Metal Heat Pipe Simulation Tools for Micro-Reactor.

Reference:

Seo Joseph, Kim Hansol, Yassin A. Hassan, "Experimental study on the startup of the annular wick type heat pipe using fiber optical temperature measurement technique featured," Physics of Fluids, Volume 35, Issue 5, May 2023.

TAMU-NC:

A natural convection loop was designed based on literature review and previous results of molten salt loop facilities. Hitec salt (NaNO3-NaNO2-KNO3, 7-49-44 mol%) was used as the fluid. Instantaneous velocity measurements were obtained through the vertical test section using PIV.

Reference:

Reis, J., Seo, J. and Hassan, Y., Experimental Investigation and Numerical Validation of Natural Circulation in Molten Salt, Proceedings of NURETH-20, (Summary #5219), 2023.

97 TAMU UP: Experimental measurements were performed in the upper plenum of a high-temperature gas-cooled reactor (HTGR) that was scaled, designed, and assembled at Texas A&M University. The goal was to investigate flow mixing, thermal stratification, and plume impingement in the upper plenum of the HTGR under a loss-of-forced-coolant accident and provide a high-fidelity experimental database for validating computational fluid dynamics (CFD) and system codes.

Reference:

Anas Alwafi, Thien Nguyen, Yassin Hassan, N.K. Anand, Experimental analysis of a non-isothermal confined impinging single plume using time-resolved particle image velocimetry and planar laser induced fluorescence measurements, International Journal of Heat and Mass Transfer, Volume, 193, September 2022.

Tan et al. Twisted Tube Bundle: The twisted tube bundle is an experimental facility consisting of 37 twisted elliptical tubes with water as the working fluid on both the tube-side and shell-side.

The tubes were arranged in a triangular lattice and instrumented to measure heat transfer from the twisted tubes on the shell-side.

Reference:

Tan, X.-h., Zhu, D.-s., Zhou, G.-y., and Zeng, L.-d., 2013. Heat transfer and pressure drop performance of twisted oval heat exchanger. Applied Thermal Engineering, 50, pp. 374-383.

THTR-300: The THTR-300 was a thorium high-temperature nuclear reactor rated at 300 MW electric (THTR-300) in Hamm-Uentrop, Germany. It started operating in 1983, synchronized with the grid in 1985, operated at full power in February 1987 and was shut down September 1, 1989. The THTR-300 served as a prototype high-temperature reactor (HTR) to use the TRISO pebble fuel produced by the AVR, an experimental pebble bed operated by VEW.

Reference:

R. Bumer, I. Kalinowski, E. Rhler, J. Schning, W. Wachholz, "Construction and operating experience with the 300-MW THTR nuclear power plant," Nuclear Engineering and Design, Volume 121, Issue 2, 2 July 1990, Pages 155-166.

Oehme, H., Schoening, J., 1970. DESIGN, FEATURES, AND engineering STATUS OF THE THTR 300 MWe prototype POWER STATION. URL https://www.osti.gov/biblio/4121847.

98 Tibergas MSR benchmark: This is a multi-physics benchmark featuring a molten salt system whose characteristics (neutron spectrum, strong temperature feedback, salt composition, precursors movement) that make it a simple representation of the MSFR. The geometry is a 2 m by 2 m cavity filled with molten salt at an initial temperature of 900 K. The fuel salt is a LiF-BeF2-UF4, whose fluid properties are considered constant with temperature and uniform in space.

References:

Marco Tiberga, Rodrigo Gonzalez Gonzaga de Oliveira, Eric Cervi, Juan Antonio Blanco, Stefano Lorenzi, Manuele Aufiero, Danny Lathouwers, Pablo Rubiolo, "Results from a multi-physics numerical benchmark for codes dedicated to molten salt fast reactors,"

Annals of Nuclear Energy, Volume 142, July 2020.

Aufiero, M., 2015. Serpent-OpenFOAM coupling for criticality accidents modelling -

Definition of a benchmark for MSRs multiphysics modelling. In: Serpent and Multiphysics meeting. Grenoble, France.

Toshiba 37-pin: The Toshiba 37-pin benchmark is based on liquid-sodium experiments conducted by the Toshiba Corporation Nuclear Engineering Laboratory in Japan. The Toshiba bundle is larger than the ORNL 19-pin benchmark, with one more outer ring of heated rods. The specific power per rod is less than the ORNL 19-pin benchmark, but the rods have a slightly larger diameter and a larger axial heated length.

Reference:

Namekawa, F.; Ito, A.; Mawatari, K. Buoyancy effects on wire-wrapped rod bundle heat transfer in an LMFBR fuel assembly. AIChE Symp. Ser. 1984, 80, 128-133.

TREAT M8CAL: TREAT is a homogeneous, air-cooled, graphite-moderated and graphite-reflected reactor. The TREAT calibration test provides neutronic validation relevant to prismatic gas-cooled cores.

Reference:

T. H. Bauer W. R. Robinson. The M8 Power Calibration Experiment (M8CAL). Technical Report ANL-IFR-232, Argonne National Laboratory, May 1994.

99 UCB-PBHTX: The Pebble-Bed Heat Transfer Experiment (PBHTX) is a scaled facility designed to measure heat transfer coefficients within a pebble-bed test section for the conditions applicable to the PB-FHR. A simulant oil, Dowtherm A, is used as the heat transfer fluid, which matches the Prandtl number of FLiBe at temperatures lower than the PB-FHR conditions.

Reference:

Huddar, Lakshana Ravindranath, Heat Transfer in Pebble-Bed Nuclear Reactor Cores Cooled by Fluoride Salts, Ph.D. Thesis, University of California, Berkeley, 2016.

UM-FLUSTFA: The UM-FLUSTFA is a small integral test facility to investigate heat transfer and fluid flow in a molten salt. FLUSTFA is comprised of a reservoir tank for salt storage and melting, a primary molten salt loop, a secondary molten salt loop, a closed air loop, and a chilled water loop. This salt facility uses FLiNaK (LiF-NaF-KF: 46.5-11.5-42 mol%) as the working fluids for both the primary and secondary salt loops, and operates up to 700 °C and near the atmospheric pressure.

References:

Sheng Zhang, Study of a Passive Decay Heat Removal System and Tritium Mitigation for Fluoride-salt-cooled High-temperature Reactors, PhD. Thesis, University of Michigan 2020.

Zhang, S., Lin, H.-C., Chen, M., Shi, S., Che, S., Burak, A., Sun, X., and Lv, Q., Design and Construction of an Integral-Effect Test Facility FLUSTFA for Molten Salt Reactor Applications, Proceedings of the 20th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-20), August 20-25, 2023, Washington, DC, USA.

UM RCCS: An experimental facility built at the University of Michigan was aimed at the investigation of mixing in the upper plenum of the air-cooled RCCS design. The facility was equipped with state-of-the-art high-resolution instrumentation to achieve CFD grade experiments.

Reference:

Manera, Annalisa, Corradini, Michael, Petrov, Victor, Anderson, Mark, Tompkins, Casey, and Nunez, Daniel. Model validation using CFD-grade experimental database for NGNP

100 U. Michigan HP: The Experimental and Computational Multi-phase Flow (ECMF) laboratory at the University of Michigan has a single sodium heat pipe separate-effect facility (MISOH1). A comprehensive set of experiments to further investigate the operation of sodium heat pipes was conducted in the facility. The facility allows for the control of input power at the evaporator region, cooling intensity at the condenser region and heat pipe inclination angle.

References:

T. AHN, P.-H. HUANG, J. DIAZ, A. MANERA, & V. PETROV, Experimental Study on Start-up Characteristics of a Sodium-filled Heat Pipe, using in-house High-Resolution and High-Speed Radiation-Based Imaging System, 19th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-19) (2022).

Pei-Hsun Huang, Taehwan Ahn, Annalisa Manera, Victor Petrov, "Investigation of key parameters for the operation of a sodium heat pipe with visualization using X-ray radiography," Applied Thermal Engineering, Volume 236, Part E, 15 January 2024.

UTK-Square Cavity:

The University of Tennessee (UTK) Square Cavity study provides a series of benchmarks for CFD evaluation for flow in a cavity.

Reference:

Grubert, Marcel Andre, "Development of a Potentially Accurate and Efficient LES CFD Algorithm to Predict Heat and Mass Transport in Inhabited Spaces. " PhD diss., University of Tennessee, 2006. https://trace.tennessee.edu/utk_graddiss/1672 UW Flow Loop: Small flow loop(s) in operation at Univ. of Wisconsin may provide data on heat transfer and flow in a molten salt loop.

Reference:

Ambrosek, J., M. Anderson, et al. (2009). "Current status of knowledge of the fluoride salt (FLiNaK) heat transfer." Nuclear Technology 165(Copyright 2009, The Institution of Engineering and Technology): 166-173.

"Materials corrosion in molten LiF-NaF-KF salt", Luke Olson, James Ambrosek, Kumar Sridharan, Mark Anderson, Todd Allen, Journal of Fluorine Chemistry, 130 (2009)67-73 Reactor Cavity Cooling Systems with water and air. United States: N. p., 2018. Web.

doi:10.2172/1420273.

101 UW RCCS Tests: An RCCS 1/4 scale air-cooled facility constructed at the University of Wisconsin - Madison has been used to compliment the tests from the Argonne NSTF. The experiment consisted of a test section with three riser ducts heated via radiation.

Reference:

M. Corradini, "Experimental studies of NGNP reactor cavity cooling system with water,"

Nuclear Energy University Programs - US Department of Energy (2012)

UW-Sodium: There are two sodium loops located at the University of Wisconsin-Madison (UW). The loops were built to study chemistry, materials science, and instrumentation development under sodium fast reactor operating conditions.

Reference:

M. HVASTA, Designing and Optimizing a Moving Magnet Pump for Liquid Sodium Systems, Ph.D. Thesis, UNIVERSITY OF WISCONSIN - MADISON 2013.

VHTRC: The Very High Temperature Critical Assembly (VHTRC) is a graphite-moderated thermal critical assembly that was utilized to study temperature effects on the assembly characteristics up to a temperature of 200°C. The VHTRC was the subject of an international benchmark study. The VHTRC assembly consists of two hexagonal-prism half assemblies that are covered with a thermal neutron absorber, steel frames, and heat insulation. The half assemblies consist of hexagonal graphite reflector blocks (except for the outermost trapezoidal blocks) which contain holes for fuel rods, control and safety rods, electric heaters and other uses.

References:

F. Reitsma, G. Strydom, F. Bostelmann, K. Ivanov, The IAEA Coordinated Research Program on HTGR Uncertainty Analysis: Phase I Status and Initial Results, Proceedings of HTR 2014, Weihai, China, 2014.

NEA, International Handbook of Evaluated Reactor Physics Benchmark Experiments, NEA/NSC/DOC, NEA No. 7140, 2013.

102 ZPPR: A series of criticality tests were performed by Argonne National Lab in the 1960s. Tests ZPPR-6, -15, and -21 are recommended. Test ZPPR-15 appears to be suitable for fuel with enrichments expected for modern fast reactors.

Reference:

McFarlane, H. F.; Collins, P. J.; Carpenter, S. G.; Olsen, D. N.; Smith, D. M.; Schaefer, R.

W. (Argonne National Lab., Idaho Falls, ID (USA)) et al. Analysis and evaluation of ZPPR (Zero Power Physics Reactor) critical experiments for a 100 kilowatt-electric space reactor, article, January 1, 1990; Idaho Falls, Idaho.