INL is a U.S. Department of Energy National Laboratory
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INL/RPT-23-73834
Revision 0
Overview of Virtual Test Bed
FY23 Activities
July 2023
National Reactor Innovation Center
Abdalla Abou-Jaoude, Guillaume Giudicelli, Samuel Walker, Mauricio
Tano, Lise Charlot, Sebastian Schunert, Paolo Balestra, Dempsey Rogers,
Logan Harbour, and Derek Gaston
Idaho National Laboratory
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REVISION LOG
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No.
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SUMMARY
This milestone report highlights progress achieved within the Virtual Test
Bed (VTB) project during Fiscal Year (FY) 2023. The VTB consists of an
open-source repository of state-of-the-art simulation examples for
advanced reactor types. Its objective is to help accelerate reactor
maturation and deployment by providing ‘reference’ models that can be
openly accessed, improved upon, and re-purposed for proprietary
applications. The main accomplishments are summarized below:
1. Models on the repository were consistently maintained and new
improvements were developed to improve the useability of the
website. These new improvements include a tagging system and
testing of computationally expensive models.
2. A total of 19 new models were uploaded to the repository this FY.
They showcase a wide range of new modeling and simulation
capabilities for advanced reactors. The VTB website also now
contains two tutorials/trainings with two additional ones coming
soon.
3. Model development activities were focused this FY on potential use
cases for demonstration within the NRIC test beds. The DOME use
case was a gas-cooled microreactor, and the LOTUS use case was
a molten salt reactor. The models developed as part of this scope
are expected to help provide the foundation for potential
confirmatory analyses in the future, thus accelerating potential
timelines for reactor deployments.
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ACKNOWLEDGEMENTS
This document was sponsored by the National Reactor Innovation
Center (NRIC). NRIC is a national program funded by U. S. Department of
Energy’s Office of Nuclear Energy and is dedicated to the demonstration
and deployment of advanced nuclear energy. Neither the U.S. Government
nor any agency thereof makes any warranty, express or implied, or
assumes any legal liability or responsibility for the accuracy, completeness,
or usefulness of any information, apparatus, product, or process disclosed,
or represents that its use would not infringe on privately owned rights.
References herein to any specific commercial product, process, or service
by trade name, trademark, manufacturer, or otherwise do not necessarily
constitute or imply its endorsement, recommendation, or favoring by the
U.S. Government or any agency thereof. The views and opinions of authors
expressed herein do not necessarily state or reflect those of the U.S.
Government or any agency thereof.
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CONTENTS
SUMMARY ............................................................................................................................................................................III
ACKNOWLEDGEMENTS ................................................................................................................................................... V
ACRONYMS ........................................................................................................................................................................ XI
1. BACKGROUND AND OVERVIEW ........................................................................................................................ 1
2. REPOSITORY MAINTENANCE AND IMPROVEMENT ........................................................................................ 2
2.1 Repository Status And Maintenance ....................................................................................................... 2
2.2 Repository Improvements .......................................................................................................................... 3
3. HOSTING EXTERNAL REACTOR MODELS .......................................................................................................... 6
3.1 Molten Salt Reactors (MSR) ...................................................................................................................... 7
3.1.1 Molten Salt Fast Reactor (MSFR).................................................................................................. 7
3.2 High-Temperature Gas-cooled Reactors (HTGR) ................................................................................ 9
3.2.1 Pebble Bed Modular Reactor (PBMR) ........................................................................................ 9
3.3 Fluoride High-Temperature Reactors (FHR) ....................................................................................... 12
3.4 Liquid Metal Fast Reactors (LMFR) ....................................................................................................... 13
3.5 Microreactors (MR) .................................................................................................................................. 15
3.5.1 Heat Pipe ....................................................................................................................................... 15
3.6 Upcoming Models Slated for Inclusion on the VTB ........................................................................... 18
4. TUTORIAL DEVELOPMENT AND TRAINING ACTIVITIES ............................................................................... 19
4.1 Existing Tutorials: MultiApp and Workbench ..................................................................................... 19
4.1.1 MultiApp Tutorial ......................................................................................................................... 19
4.2 Pebble Bed Reactor Tutorial ................................................................................................................. 21
4.3 Bison Training Highlighting VTB use and Development ................................................................... 23
5. MODEL DEVELOPMENT ACTIVITIES .................................................................................................................. 24
5.1 DOME Use Case: Gas Microreactor .................................................................................................... 24
5.1.1 Balance of Plant Model .............................................................................................................. 24
5.1.2 Full Core with Bypass Flow......................................................................................................... 27
5.1.3 Remaining Work .......................................................................................................................... 27
5.2 LOTUS Use Case: Low-Powered MSR ................................................................................................. 28
5.2.1 Reactor Specifications ................................................................................................................. 28
5.2.2 Computational Codes .................................................................................................................. 29
5.2.3 Steady-State Reactor Operation ............................................................................................. 30
5.2.4 Loss of Forced-Flow Accident (LOFA) ....................................................................................... 34
5.2.5 Remaining work ............................................................................................................................ 35
6. PUBLICATIONS AND OUTREACH ...................................................................................................................... 35
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7. REFERENCES........................................................................................................................................................... 36
FIGURES
Figure 1. Current status of the VTB testing on the continuous integration platform. ............................ 2
Figure 2. Repository visits tracking over June 24 – July 7 using the GitHub activity page. .............. 3
Figure 3. Usage tracking using Google Analytics over May 1 - July 6. ................................................ 3
Figure 4. Current prototype status of the search feature. Arbitrary key and values may be
specified. ........................................................................................................................................... 4
Figure 5. The steady-state temperature distribution in the 1D MSFR primary loop model
using SAM. ........................................................................................................................................ 8
Figure 6. Steady state fluorine potential (J/mol) in the MSFR. ............................................................... 8
Figure 7. Coupled Pronghorn-SAM thermal-hydraulics result showcases the temperature
distribution (K) for steady-state primary circuit and secondary cooling of MSRE.
[2] ....................................................................................................................................................... 9
Figure 8. Comparison of the new SAM-based model of a pebble bed reactor versus the
Griffin Pronghorn higher fidelity model. ................................................................................... 10
Figure 9. Axisymmetric model of an aspherical particle, and time-dependent results for
that model. ...................................................................................................................................... 10
Figure 10. Cutaway of the HTTF model during the hottest point of the PG-26 transient. ............... 11
Figure 11. Time-averaged velocity field (non-dimensional) in HTTF lower plenum using
Nek. .................................................................................................................................................. 11
Figure 12. Pulse of matching initial periods for the 10-μm HEU model and the 5-μm LEU
model with (a) power density and deposited energy and (b) average feedback
temperature results. ....................................................................................................................... 12
Figure 13. Steady-state core thermal solution. ......................................................................................... 13
Figure 14. Radial power density profile at the highest axial mesh of the fuel region. .................... 14
Figure 15. Temperature gradient mapping of the duct (left) and duct bowing (right,
magnified) using the MOOSE tensor mechanics module. ....................................................... 14
Figure 16. Example of simulation results for the high-flow test case in the Toshiba 37-pin
benchmark. (a) Distribution of axial mass flow. (b) Distribution of lateral mass
flow. (c) Distribution of temperature. (d) Distribution of dynamic viscosity due to
heating. ............................................................................................................................................ 15
Figure 17. Overview of power and temperature performance of the heat pipe
microreactor model on the VTB during a transient. ................................................................. 16
Figure 18. Asymptotic temperature spatial distribution for the heat pipe failure scenario
(left) and for normal operations (right). .................................................................................... 16
Figure 19. Power density and temperature profiles in the coupled multiphysics gas
microreactor model 500 seconds after the transient. ............................................................. 17
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Figure 20. Temperature distribution in the SNAP8 model. ..................................................................... 18
Figure 21. The MOOSE MultiApp hierarchy leveraged in several VTB models. ................................ 20
Figure 22. Screenshot of a Workbench interface showcasing a VTB example simulation. .............. 21
Figure 23. Layout of the GPBR200 model. ............................................................................................... 22
Figure 24. Balance of Plant diagram. ........................................................................................................ 25
Figure 25. Power and temperature evolution during a startup transient. ........................................... 26
Figure 26. Power and temperature evolution during a load-follow transient. ................................... 26
Figure 27. Bypass cross flow and temperature in the core and bypass. ............................................. 27
Figure 28. Longitudinal cross section of primary loop modeled for an MCRE-like reactor. ............ 28
Figure 29. Continuous energy spectrum for MCRE-like reactor core with six-group energy
discretization represented in colored bins. ............................................................................... 31
Figure 30. Neutron flux distribution computed by Griffin for the fast-most spectrum (left)
and second-most-thermal spectrum (center), and continuous energy collision track
plot (right). ...................................................................................................................................... 32
Figure 31. Thermal hydraulics fields steady-state operation of MCRE including fuel salt
velocity in the vertical direction (left) and temperature field (right). Note that
while the temperature variations may appear to be substantial, they only
represent a variation within 10 K. .............................................................................................. 32
Figure 32. Thermomechanics fields in the nuclear reflector. ................................................................... 34
Figure 33. Velocity and temperature fields after 120 seconds of the beginning of the
LOFA. ............................................................................................................................................... 35
TABLES
Table 1. Overview of reactor models hosted on the VTB. New models uploaded in FY 23
are highlighted in blue. .................................................................................................................. 6
Table 2. Overview of upcoming reactor models slated to be hosted on the VTB. ............................ 18
Table 3. Overview of the Pebble Bed Reactor (PBR) modeling steps in the VTB tutorial. ............... 22
Table 4. Operating conditions for the DOME microreactor use case. .................................................. 25
Table 5. Parameters of modeled LOTUS MSR use case. ........................................................................ 29
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ACRONYMS
ANL Argonne National Laboratory
ANS American Nuclear Society
ART Advanced Reactor Technology
CFD Computational Fluid Dynamics
DOE U. S. Department of Energy
DOME Demonstration of Microreactor Experiments
EBR-II Experimental Breeder Reactor-II
FHR Fluoride High-Temperature Reactor
FY Fiscal Year
GPBR Generic Pebble Bed Reactor
HPC High-Performance Computing
HTGR High Temperature Gas Cooled Reactor
HTTF High Temperature Test Facility
INL Idaho National Laboratory
LEU low enriched uranium
LMFR Liquid Metal Fast Reactors
LOFA Loss of Flow Accident
LOTUS Laboratory for Operation and Testing in the United States
LWR Light Water Reactor
M&S Modeling and Simulation
MCRE Molten Chloride Reactor Experiment
MHTGR Modular High-Temperature Gas Reactor
MOOSE Multiphysics Object-Oriented Simulation Environment
MSFR Molten Salt Fast Reactor
MSR Molten Salt Reactor
MSRE Molten Salt Reactor Experiment
NEAMS Nuclear Energy Advanced Modeling and Simulation
NRIC National Reactor Innovation Center
NTP Nuclear Thermal Propulsion
PBMR Pebble Bed Modular Reactor
SAM System Analysis Module
SNAP8 Systems for Nuclear Auxiliary Power 8
THM Thermal Hydraulics Model
TREAT Transient Reactor Test Facility
VTB Virtual Test Bed
VTR Versatile Test Reactor
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1. BACKGROUND AND OVERVIEW
The National Reactor Innovation Center (NRIC) was created for accelerating the deployment
of novel reactor concepts. This will be achieved by providing both physical and virtual spaces for
building and testing various components, systems, and complete pilot plants. The Virtual Test Bed
(VTB) represents the virtual arm. It is being developed in collaboration with U.S. Department of
Energy’s (DOE) Nuclear Energy Advanced Modeling and Simulation (NEAMS) program.
The mission of the VTB is to accelerate the deployment of advanced reactors by facilitating
the leveraging of cutting-edge DOE advanced modeling and simulation (M&S) tools to design,
evaluate, and license reactors. This is primarily achieved by storing example challenge problems
in an externally available repository and by developing models to fill the potential
demonstrator’s M&S gaps.
Activities conducted this fiscal year (FY) within the Idaho National Laboratory (INL) workscope
centered around three key areas: (1) repository maintenance and improvement, (2) hosting
external models to the repository, and (3) developing additional modeling examples relevant to
potential demonstrators. The main accomplishments in FY23 included:
• Scope #1, Repository:
• Continued identifying and resolving deprecations in the models and codes as new updates
were pushed.
• Set up template and format for new model submission.
• Development of “model filter” capability to more easily search the ever-growing list of
models on the repository.
• Set-up HPC-link to enable testing of more computationally demanding models
• Scope #2, External Models:
• Addition of 19 new models developed by other programs (DOE NEAMS, Advanced
Reactor Technology [ART], U.S. Nuclear Regulatory Commission, etc.) to the VTB repository.
• Notable novel capabilities include: multiphysics models for gas and heat pipe-cooled
microreactors, a fuel performance model, and structural deformation models.
• The VTB hosts an increasing number of ‘tutorials’ to help initiate users on the capabilities
hosted on the VTB. This includes two existing tutorials on the MultiApp system and the
Workbench interfaces. Two additional tutorials are currently under preparation: a thermal
hydraulics tutorial for pebble bed reactors, and fuel performance modeling for a wide
range of reactor types.
• Scope #3, New Models:
• Development of an example use case for the NRIC ‘DOME’ testbed: gas-cooled
microreactor.
• Development of an example use case for the ‘LOTUS’ testbed: low-power molten salt
reactor.
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2. REPOSITORY MAINTENANCE AND IMPROVEMENT
2.1 Repository Status And Maintenance
The first major change in FY23 concerns testing practices. Regression testing was extended
across all the previously added VTB models, aside from the most computationally expensive
models which would require large “gold” files that act as reference solutions to be stored for
comparison. New models were all accepted into the repository with both syntax checks and
regression testing. The extension of regression testing to past models found issues with a restart
configuration for a PBMR400 neutronics input. This issue is in the process of being resolved with
work ongoing in the MOOSE software to transform the restart and recover system.
Over 40 pull requests were made to the repository to update the models, remove
deprecation and adapt to new syntaxes. Most of these updates were performed by the
developers of the NEAMS codes, and a little under half by the VTB technical staff. Issues fixed by
VTB personnel were usually discovered or fixed during scheduled codes updates, while issues
fixed by code developers were discovered through the automatic testing of the VTB models
performed when modifying the codes.
The current testing status of the VTB is shown in Figure 1. Most codes fully past the test suite
for the models they are supposed to run. The combined applications, BlueCrab and Direwolf, fail
a few tests each. The Direwolf failures are currently due to unused parameters in the inputs and
the executable inexplicably failing to allow it. One BlueCrab failure is tied to the restart/recover
issue previously mentioned. Three pull requests are open in the repository. One attempts to
improve the performance of an microreactor model steady-state calculation; it currently requires
further attention. Another uses the new general field transfer technology, developed by NEAMS in
FY23, in all the inputs across the repository.
Figure 1. Current status of the VTB testing on the continuous integration platform.
To better assess the impact of the VTB, utilization by internal and external collaborators
should be tracked. Github offers an interface that tracks both website traffic and “forking”, the
act of creating an independent copy of the VTB. Github reports 32 active forks of the repository
over the last year. Figure 2 indicates that between the last week of June and first week of July
2023, 50 unique visitors have browsed through the repository. To confirm these numbers, an
additional usage tracking mechanism using Google Analytics was implemented. This was set up at
the beginning of April. This data is presented in Figure 3. Over this period, nearly 600 unique
visitors have browsed the repository, with 465 located in the United States.
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Figure 2. Repository visits tracking over June 24 – July 7 using the GitHub activity page.
Figure 3. Usage tracking using Google Analytics over May 1 - July 6.
The documentation of the VTB models was modified through 14 pull requests, which notably
added direct links from documentation to models, highlighted the open-source models in the
model indexing, added instructions to download mesh files for some large models, added some
missing citations for models that were unpublished at the time of release, and kept the model
indexing up to date. Four pull requests were made by model points of contact to improve the
documentation of the model.
2.2 Repository Improvements
Now at more than 30 models, as the repository grows it becomes more and more challenging
to search for an input containing a feature of interest, a specified level of fidelity, a specific
physics solve, etc. To enable these searches, which are primordial to using the VTB as an example
before building a new model, several indexes of the models were created manually, sorted and
linked from the front page:
• By reactor type
• By code used, with a special category for fully & partially open-source models
• By features present
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• By computing needs.
This manual sorting is labor-intensive and prone to becoming outdated. Two new capabilities
were developed for MooseDocs (the markdown language used to write documentation for
MOOSE) to handle this automatically. A tagging command was written to aid in the filtering of
models. This command allows the author of a model to supply a set of key value pairs to a
dictionary which can then be used to filter down to a specific simulation. The command also
supplies a path to the directory of the tagged markdown in the dictionary. This path can then be
used to link to the corresponding simulation. For example, the High Temperature Gas Reactor
(HTGR)assembly Multiphysics simulation could be tagged within its markdown file as:
!tagger HTGR_Assembly_mps simulation_type:steady_state reactor:generic_HTGR
This would allow users to filter models by simulation type and multiphysics calculation and
provide the user with a link to the documentation to the resulting HTGR model.
The tagging command line below would allow users to filter to the Molten Salt Fast Reactor
core transient model documentation.
!tagger MSFR simulation_type:transient reactor:MSFR
The second facet of this is a simple web API (application programming interface) that parses
the dictionary and provides a search feature. Because all MOOSE websites are static, out of
cyber security concerns, the search code is executed on the visitor’s browser. The search page
currently has a limited interface, shown in Figure 4.
Figure 4. Current prototype status of the search feature. This is being expanded to capture all
models on the VTB currently.
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A template format was also added for the submission of new models to the VTB by Jun. Fang
at Argonne National Laboratory (ANL). This streamlines the process and avoids major misses in
new submissions. It also facilitates the review of new models. The minimal submission is guided to
include:
• A description of the reactor with some graphical description and historical context
• A description of the inputs starting from the general design concepts, then diving into the roles
of the individual objects in the simulation
• A description of the results and their significance
• How to run the model, notably on INL High-Performance Computing (HPC) infrastructure.
It also acts as a targeted tutorial to MooseDocs, the markdown language used to write
documentation for MOOSE, presenting examples on how to create tables, figures, paragraphs
and so on. In that regard, it is inspired by the American Nuclear Society’s (ANS) templates for
conference presentation summaries.
The VTB also began deployment of HPC enabled testing in FY23. The list of models that can
only be run using an HPC cluster is identified already in a specific indexing of the VTB. Most
models in this category are using Nek, the ANL-developed Computational Fluid Dynamics (CFD)
solver. High fidelity CFD simulations require very fine mesh and are often run on Graphics
Processing Units (GPUs) or large groups of Central Processing Units (CPU). Some other
computationally intensive models are full core heterogeneous multiphysics models, such as the heat
pipe microreactor core transient models. A script was designed to run these cases manually in an
interactive session with four cluster nodes mobilized. The codes are built manually from source
beforehand. The results are then checked by the script.
A more automated continuous integration solution to HPC model testing is currently facing the
four following challenges:
1. Automatic testing on the HPC requires building the code on one of the “head” (main) nodes, as
the codes cannot currently be built without internet access. This is a resource-intensive activity
and the head nodes are already significantly overused. The file system slowness is a regular
complaint from HPC users and building the entire suite of NEAMS tools regularly would only
aggravate the problem.
2. It also requires administrative privileges for the CIVET client hosted on HPC. This level of
permission requires great care with little tolerance for mistakes in the setup; the login node
could be brought offline (for all HPC users) in the case of any incorrect manipulations. It is
typically only granted to HPC staff.
3. The VTB is hosted on a public github repository. The combination of this fact and the
administrative rights could enable, in theory, a pathway for cyber-attackers to gain control of
an INL HPC cluster. In practice, external contributors have to be added manually to an
“allow”-list before any traffic from their code to CIVET is even allowed.
4. Current deployment work in the MOOSE team will remove these three limitations in the near
future with the deployment of pre-built containerized software solutions. It is considered more
strategic to leverage these ongoing activities than work on solutions to each limitation. They
are expected to be completed prior to the end of this FY.
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3. HOSTING EXTERNAL REACTOR MODELS
An overview of the models currently hosted is provided in Table 1. The Table showcases the
breadth of reactor types and analysis tools showcased within the repository. The intent is for users
to “mix-and-match” different combinations of use cases to suit their particular needs. New models
for FY23 are highlighted in blue.
Table 1. Overview of reactor models hosted on the VTB. New models uploaded in FY 23 are
highlighted in blue.
Type
Reactor
Simulation
Codes Used
MSR
MSFR
Core neutronics and hydraulics, steady state +
transient
Griffin; Pronghorn
Core + system neutronics and hydraulics, steady state
+ transient
Griffin; Pronghorn; SAM
Core depletion with species removal
Griffin
Core high-fidelity CFD (LES and RANS)
Nek5000
Primary loop Transient
SAM
Core steady state chemical species tracking +
thermochemistry
Griffin; Pronghorn;
Thermochimica
MSRE
Primary loop steady-state and transient
SAM
Core coarse-mesh steady-state and transient
Pronghorn; SAM
LOTUS
Molten salt reactor core hydraulics and structure;
steady-state and transient
Griffin; Pronghorn; Bison
HTGR
PBMR
Core neutronics and hydraulics, steady state +
transient
Griffin; Pronghorn
67 pebble high-fidelity conjugate heat transfer
Cardinal
Primary loop steady-state and transient
SAM
MHTGR
Primary loop steady-state
SAM
Assembly high-fidelity neutronics and thermal
hydraulics
Cardinal
TRISO fuel performance
Bison
Core neutronics steady-state
Griffin
HTTF
Hydraulics and structure transient
RELAP7/THM
Hydraulics system steady-state and transient
benchmark
SAM
High-fidelity CFD of lower plenum
Nek5000
HTR10
Core neutronics steady-state
Griffin
TREAT
Core pulsed transient
Griffin; Bison
FHR
Mk1
Core neutronics and hydraulics, steady state
Griffin; Pronghorn
Reactor bypass high-fidelity conjugate heat transfer
Cardinal
Primary loop steady-state and transient
SAM
Core + system neutronics and hydraulics, steady state
Griffin; Pronghorn; SAM
gFHR
Equilibrium pebble core with neutronics and hydraulics
Griffin; Pronghorn
LMFR Lattice
Assembly neutronics, hydraulics, and fuel
Griffin; SAM; Bison
Sub-channel hydraulics benchmark
Pronghorn
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Table 1. (continued).
INL/RPT-23-73834 7
Type
Reactor
Simulation
Codes Used
Lead-cooled reactor assembly heterogeneous
neutronics
Griffin
Duct bowing
MOOSE/Thermo-
mechanics
VTR
3D Core neutronics, hydraulics, and fuel; steady-state
Griffin; SAM; Bison
ABTR
Primary loop Transient
SAM
Micro
Heat
pipe
Core hydraulics and structure, steady-state
Sockeye; Bison
3D Core neutronics, hydraulics, and fuel; steady-state
and transient
Griffin; Sockeye; Bison
3D Core neutronics, hydraulics, and fuel with hydride
migration; steady-state and transient
Griffin; Sockeye; Bison
Gas
cooled
Assembly neutronics, hydraulics, and structure; steady-
state and transient
Griffin; SAM; Bison
System steady-state and transient
MOOSE/Thermal
hydraulics
SNAP8
2D Core neutronics and structure; steady-state
Griffin; Bison
As can be seen in Table 1, there were several new additions to the VTB repository this FY. All
reactor types have nearly double the number of models from the previous year update, with new
single and multiphysics analyses of steady state and transient scenarios. A summary of recently
added models will be discussed in more detail in the following subsections. An overview of the
models will also be presented at an ANS Winter Conference special session on the VTB.
3.1 Molten Salt Reactors (MSR)
The VTB hosts several example simulations for Molten Salt Reactors (MSRs). Two main design
variants are represented, the fast spectrum MSFR, [1] and the thermal spectrum MSRE [2].
Additionally, a new model for the first MSR experiment to go into LOTUS was also developed
and is discussed in more detail in Section 5.2.
3.1.1 Molten Salt Fast Reactor (MSFR)
While several models for the MSFR were already hosted in the VTB (see Table 1), including a
Pronghorn-SAM model, a standalone 1D SAM model was recently uploaded to the VTB , and is
showcased in Figure 5. This model demonstrates the point kinetics capabilities within the code to
allow comparisons against the higher fidelity Griffin-Pronghorn-SAM model on the repository.
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Figure 5. The steady-state temperature distribution in the 1D MSFR primary loop model using
SAM.
Additionally, a new chemical species tracking and thermochemistry model of the MSFR was
recently finalized. This model uses a steady state multiphysics solutions from Griffin and
Pronghorn, and then solves the steady state thermochemistry of the MSFR system given the
temperature, pressure, and elemental spatial distributions as seen in Figure 6.
Figure 6. Steady state fluorine potential (J/mol) in the MSFR.
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3.1.1.1 Molten Salt Reactor Experiment (MSRE)
A Pronghorn-SAM coupled model of the MSRE primary loop was developed and shown in
Figure 7. Pronghorn was used for the reactor core and primary loop and was coupled to a SAM
primary loop model using a “domain overlap” approach. The SAM model captures both the
primary and secondary loop of the system. Future work intends to couple the Pronghorn model to
Griffin in a similar fashion as the Pronghorn-Griffin-SAM model of the MSFR that was previously
uploaded to the VTB. [1]
Figure 7. Coupled Pronghorn-SAM thermal-hydraulics result showcases the temperature
distribution (K) for steady-state primary circuit and secondary cooling of MSRE. [2]
3.2 High-Temperature Gas-cooled Reactors (HTGR)
Several high-temperature gas reactor (HTGR) models are already hosted in the VTB as seen
in Table 1, with several variations represented. New additions were made to the PBMR model [3]
and the High Temperature Test Facility (HTTF) benchmark [4] specifically. Additionally, new fuel
performance analyses for Tri-stuctural ISOtropic (TRISO) particle fuel using Bison were also
added, [5] as well as a pulsed low enriched uranium (LEU) fuel performance model analogous to
the Transient Reactor Test Facility (TREAT) reactor. [6]
3.2.1 Pebble Bed Modular Reactor (PBMR)
A SAM-based model of the PBMR-400 was uploaded to the VTB. As in the new MSFR use
case, it can be used to compare against existing Griffin-Pronghorn models of the reactor designs
on the VTB as shown in the transient response of Figure 8.
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Figure 8. Comparison of the new SAM-based model of a pebble bed reactor versus the
Griffin Pronghorn higher fidelity model.
3.2.1.1 TRISO Fuel Performance
New 1D and 2D aspherical TRISO models in the fuel performance code Bison were also
included recently in the VTB which demonstrate the effect of burnup and the resulting stress on
TRISO fuel as seen in Figure 9. [5] These models are helpful since they can be coupled with other
VTB reactor models that use TRISO fuel.
Figure 9. Axisymmetric model of an aspherical particle, and time-dependent results for that
model.
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3.2.1.2 High Temperature Test Facility (HTTF)
Several new models of the HTTF benchmark were uploaded to the repository recently This
included system models using SAM and the MOOSE thermal-hydraulics module (THM) (shown in
Figure 10) as well as high-fidelity CFD simulations using Nek (shown in Figure 11). These models
are especially helpful since they use benchmark data to validate these codes capabilities at
replicating real world results.
Figure 10. Cutaway of the HTTF model during the hottest point of the PG-26 transient.
Figure 11. Time-averaged velocity field (non-dimensional) in HTTF lower plenum using Nek.
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3.2.1.3 Transient Reactor Test Facility (TREAT)
Last, a multiphysics model of pulsed fuel performance analogous to being irradiated in the
TREAT reactor has also been uploaded to the VTB which uses the neutronics code Griffin and the
fuel performance code Bison. [6] The results are showcased in Figure 12.
Figure 12. Pulse of matching initial periods for the 10-μm HEU model and the 5-μm LEU model
with (a) power density and deposited energy and (b) average feedback temperature results.
3.3 Fluoride High-Temperature Reactors (FHR)
A model showcasing the new equilibrium pebble shuffling in the gFHR model was added to
the VTB. The simulation coupled Griffin-Pronghorn to assess the impact of pebble depletion on the
temperature distribution in a pebble-bed core. [7] The steady state core thermal solution is shown
in Figure 13.
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Figure 13. Steady-state core thermal solution.
3.4 Liquid Metal Fast Reactors (LMFR)
Three new capabilities for modeling LMFR reactors were recently uploaded to the VTB. This
included a detailed heterogeneous neutronics model for a lead-cooled fast reactor assembly [8]
(shown in Figure 14), an assembly duct bowing benchmark problem that uses openly available
MOOSE modules [9] (shown in Figure 15), and new subchannel benchmarks for LMFR lattices [10]
(shown in Figure 16).
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Figure 14. Radial power density profile at the highest axial mesh of the fuel region.
Figure 15. Temperature gradient mapping of the duct (left) and duct bowing (right, magnified)
using the MOOSE tensor mechanics module.
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Figure 16. Example of simulation results for the high-flow test case in the Toshiba 37-pin
benchmark. (a) Distribution of axial mass flow. (b) Distribution of lateral mass flow. (c) Distribution
of temperature. (d) Distribution of dynamic viscosity due to heating.
3.5 Microreactors (MR)
Several new additions were added to showcase microreactor simulation capabilities. These
can be broadly grouped between heat-pipe based systems and gas-cooled systems. Last, a
Systems for Nuclear Auxiliary Power 8 (SNAP8) reactor model for space use was also developed
and uploaded to the VTB.
3.5.1 Heat Pipe
The original heat pipe microreactor model on the VTB was lacking neutronics coupling. The
most recent coupling showcases a complete multiphysics simulation including neutronics (Griffin),
heat pipe hydraulics (Sockeye), and thermomechanics (Bison). [11] Both steady-state and transient
models were uploaded in the VTB. The transient scenario considers a cascading heat pipe failure
accident as seen in Figure 17.
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Figure 17. Overview of power and temperature performance of the heat pipe microreactor
model on the VTB during a transient.
Additionally, a second heat pipe microreactor model with hydride migration was also
included and selected results can be seen in Figure 18. [12] The model is similar to the one above,
but captures the feedback that resulted from hydrogen migrating due to temperature and power
effects.
Figure 18. Asymptotic temperature spatial distribution for the heat pipe failure scenario (left) and
for normal operations (right).
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3.5.1.1 Gas-Cooled
The gas-cooled reactor example model on the VTB consists of a 3D assembly model for
prismatic HTGR-type reactor but with hydride moderating material as seen in Figure 19. [13]
Griffin is leveraged for neutronics, SAM for the system hydraulics, and Bison for the
thermomechanics. Steady-state simulations are included alongside transient ones – namely flow
blockage and reactivity insertion event.
Figure 19. Power density and temperature profiles in the coupled multiphysics gas microreactor
model 500 seconds after the transient.
3.5.1.2 SNAP8 Experimental Reactor (S8ER)
Last, a multiphysics model of the NaK cooled SNAP8 experimental reactor was also uploaded
to the VTB. [14] This model uses Griffin and Bison to solve the neutronics and heat transfer in this
legacy micro reactor design. The steady state temperature solution is shown in Figure 20.
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Figure 20. Temperature distribution in the SNAP8 model.
3.6 Upcoming Models Slated for Inclusion on the VTB
Additionally, there are several new models currently under development that are slated to be
uploaded to the VTB once the appropriate permissions have been granted. These models are
listed in Table 2. They notably include new reactor types like nuclear thermal propulsion (NTP)
systems and light water reactor (LWR) models that use NEAMS codes. Additionally, the inclusion
of more detailed multiphysics models for each reactor type and new material degradation and
aging models utilizing Grizzly will be included.
Table 2. Overview of upcoming reactor models slated to be hosted on the VTB.
Type
Reactor
Simulation
Codes Used
MSR
MSFR
Core thermal hydraulics with NEK5000 informed
Turbulence models
Pronghorn; Nek5000
0D Griffin + Thermochemistry Depletion
Griffin; Thermochimica
Core steady state chemical species tracking,
thermochemistry, and 3D resolved depletion
Griffin; Pronghorn;
Thermochimica
CNRS
CNRS numerical benchmark – steady state
neutronics and thermal hydraulics
Griffin; Pronghorn
MSRE
Core neutronics and thermal hydraulics with system,
steady state + transient
Griffin;Pronghorn; SAM
SILENE
Liquid fuel SILENE multiphysics benchmark
Griffin; Pronghorn
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Table 2. (continued).
INL/RPT-23-73834 19
Type
Reactor
Simulation
Codes Used
HTGR
HTTR
Core neutronics benchmark
Griffin
PBMR
Updated steady state benchmark and new
benchmark transients
Griffin; Pronghorn
MHTGR
Vessel aging and degradation
Grizzly
HTR-PM
Steady state neutronics and thermal hydraulics with
depletion
Griffin; Pronghorn
HTR-10
Thermal hydraulics steady state and transients
Griffin; Pronghorn;
HTTF
Core thermal hydraulics steady-state and transient
Pronghorn
LMFR
Fuel
Performance
Metallic Fuel performance
Bison
UN Fuel performance
Bison
EBR-II Subchannel thermal hydraulics
MOOSE
THM/Pronghorn
ABTR
Core heterogenous neutronics
Griffin
Core neutronics and thermal hydraulics
Griffin; Pronghorn
Micro Gas Cooled
2D Core neutronics and heat conduction
Griffin; MOOSE
System level thermal hydraulics
MOOSE THM
NTP
gNTP
NTP system extender cycle model
MOOSE THM
LWR
Fuel
Performance
LWR fuel performance Bison
Vessel
Degradation
Concrete degradation and aging
Grizzly
Pressure vessel degradation and aging
Grizzly
4. TUTORIAL DEVELOPMENT AND TRAINING ACTIVITIES
Being a host of cutting-edge advanced reactor simulation models, it was quickly realized that
user adoption of the tools showcased on the VTB can be expanded with tutorials. The VTB
currently contains two tutorials: the first is on the MOOSE MultiApp system that enables
multiphysics simulations by transferring data efficiently between codes and the second is on the
Workbench interface that can facilitate the setting up and running of problem sets on the INL HPC
clusters. Work is underway to include two new sets of tutorials on the VTB: the first is in
collaboration with the DOE Advanced Reactor Technology (ART) campaign on how to setup gas-
cooled reactor models and the second is on leveraging Bison for fuel performance modeling.
4.1 Existing Tutorials: MultiApp and Workbench
4.1.1 MultiApp Tutorial
Most examples showcased on the VTB leverage several physics that are coupled in various
different fashions. As such, it was felt important to include a tutorial on the MOOSE MultiApp
system that enables these the types of data transfers. Figure 21 shows the overall hierarchy of
how the system organizes different “application,” each containing its separate set of physics.
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Figure 21. The MOOSE MultiApp hierarchy leveraged in several VTB models.
The application at the top level of the MultiApp hierarchy drives the coupled simulation. This
application is termed the parent (or main) application. The parent application can have any
number of MultiApp objects. These objects may be added to any individual application to invoke
one or more child applications. A child application may then solve completely different physics
from the application above it, which may or may not be the parent application depending on its
level in the hierarchy. The MultiApp system allows a tree of simulations to be constructed with
large numbers of different applications potentially with different time and space scales.
Additional information on this tutorial can be found here:
https://mooseframework.inl.gov/virtual_test_bed/resources/multiapps.html
4.1.1.1 Workbench Tutorial
Recently, another tutorial was added to the VTB. A tight integration between the NEAMS
workbench and the VTB was put in place to enable more seamless execution of simulations. This
integration now permits users to easily run VTB examples (or their own models) using the
BlueCRAB binary on the INL HPC platform.
The tutorial itself walks new users through cloning the VTB repository, launching a NEAMS
workbench session on the INL HPC, running multiphysics examples, and analyzing the results all
within the NEAMS workbench Graphical User Interface (GUI). A screenshot explaining the
Workbench interface is shown in Figure 22. Additional information on the tutorial can be found
here: https://mooseframework.inl.gov/virtual_test_bed/resources/neams-workbench.html
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Figure 22. Screenshot of a Workbench interface showcasing a VTB example simulation.
4.2 Pebble Bed Reactor Tutorial
An upcoming tutorial is being prepared for the VTB in collaboration with the ART program.
The goal of this pebble bed reactor tutorial is to provide new users of Pronghorn with step-by-
step instructions on how to construct gas-cooled pebble reactor thermal-hydraulics models. The
tutorial starts out with a channel flow through a porous bed and gradually develops a realistic
pebble bed model – a simplified version of the generic PBR (GPBR200) [11] - in about 12 steps.
The final exercise is envisioned to tie together the thermal-hydraulics model with the stochastic
tools module to perform a sensitivity analysis of maximum fuel temperatures to thermal properties
of graphite. The tutorial emphasizes best practices for building Pronghorn models such as checking
energy and momentum balances.
The GPBR200 is a 200MW gas-cooled pebble bed reactor based on open-source literature
of current and past designs. The geometry of the GPBR200 model in Pronghorn is depicted in
Figure 23. The final model in the tutorial will be a slightly simplified version of the GPBR200
model. The difference will be:
• The cone will be removed in favor of a homogenized bottom reflector.
• The discharge chute will not be modeled explicitly.
• The barrel will be omitted and the periphery past the graphite reflector will consist of a single
gap and the reactor pressure vessel.
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Figure 23. Layout of the GPBR200 model.
The steps to completing the GPBR200 tutorial are listed in Table 3. Most of the steps for the
tutorials have been completed and are ready for submission to the VTB. The remaining steps in
progress are expected to be completed before the end of the FY.
Table 3. Overview of the Pebble Bed Reactor (PBR) modeling steps in the VTB tutorial.
Step
Description
Status
1
Creation of an axially symmetric flow channel with a uniform porosity
of 0.39. No pressure drop or heat source is imposed. Conservation of
mass is checked. No temperature equation is solved.
Complete
2
Pressure drop is added to the model created in step 1. The ‘KTA’ drag
correlation is used for that purpose. The expected pressure drop is
computed using a hand calculation and using appropriate
postprocessors from the code. The values are compared
Complete
3
Equations for solid and fluid temperatures are added. Heat transfer
between solid and fluid is added in the bed and a heat source is
introduced in the solid in the pebble bed. Conservation of energy is
checked.
Complete
4
The geometry is extended to include the cavity above the pebble bed
that has a porosity of 1. This introduces a porosity jump in the problem
that is treated using Pronghorn’s Bernoulli treatment
In progress
5
Top, bottom, and side reflectors are added. The reflectors are the first
region where only solid heat conduction is solved.
In progress
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Step
Description
Status
6
The riser channel, hot and cold plena, and flow through the bottom
reflector are added. Flow now enters the top plenum via the riser
channel radially inward and leaves the bottom plenum radially
outward. This is the first geometry where the flow is not a channel flow.
Planned
7
The control rod bypass channel in the side reflector is added. The
control rod channel is an engineered flow region and is modeled as a
porous flow region. The flow branches in the upper cavity from the
main flow and merges back in the hot plenum.
Planned
9
Core outer structure including the gaps between reflector and baffle,
and baffle and reactor pressure vessel are added. The structures may
be simplified from Figure 23.
Planned
10
Steady-state simulation setup.
Including post-processing (get average and maximum T solid in the
core) and storing initial conditions for transient
Planned
11
Transient simulation (PLOFC) setup and analysis.
Planned
12
Use stochastic tools module to compute sensitivity of bed average and
max temperature to graphite thermal conductivity
Planned
4.3 Bison Training Highlighting VTB use and Development
The NEAMS program has committed to submitting fuel performance modeling examples to the
NRIC VTB by the end of FY23. Some cases may be added in early FY24. These examples fall
under four categories: light water reactor (LWR) fuels, TRISO fuel particles, metallic fuel (U-Zr
and U-Pu-Zr), and uranium nitride (UN) fuels.
In the LWR space, two additional examples are planned using state-of-the-art models that
have been validated against appropriate experiments in the Bison assessment suite: (1) normal
operation to high burnup followed by a loss of coolant accident (LOCA)-like transient and (2)
Cr2O3-doped UO2 during normal operation.
For TRISO fuel, a variety of additional examples are planned to demonstrate the versatility
of Bison in supporting 1-, 2-, or 3-dimensional analyses depending on the physics of interest.
Examples for TRISO will be focused in the two main areas of thermomechanical performance and
fission product diffusion.
For metallic fuel, two cases are planned based upon existing Experimental Breeder Reactor
(EBR)-II experiments. These experiments will be supported by the EBR-II Fuels Irradiation and
Physics Database (FIPD). One will be a binary system of U-Zr and the second will be for ternary
U-Pu-Zr fuel. Finally, a single example will be added for UN fuel for demonstrating how to set up
a Bison simulation using the latest models. UN development is still in its infancy in Bison and most
of the models are empirical in nature.
All examples added will be accompanied by detailed descriptive documentation, Bison
models and materials selected for use in the model, the reason for their selection, and their ranges
of applicability (i.e., temperature, flux).
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5. MODEL DEVELOPMENT ACTIVITIES
Model development activities at the VTB were previously focused on MSR and FHR concepts.
Those were previously identified as key gaps in the modeling and simulation space [8] that were
not investigated by other programs. As NRIC engages with potential demonstrators, a new
priority list was developed in coordination with the leadership team:
1. DOME use case: gas-cooled microreactors
2. LOTUS use case: low-powered MSR
The models were developed in collaboration with ANL. Their activities will be summarized in a
separate milestone report. This report will focus primarily on the INL activities.
5.1 DOME Use Case: Gas Microreactor
The model development activities focused on providing examples of high-fidelity core models
and incorporating the balance of plant to these models. The balance of plant effects are
important for both transient analysis and for normal operation. To this end, the INL activities
focused on three different models:
1. Collaborating with the ANL team on the existing gas-cooled microreactor assembly model to
include a model using THM, the MOOSE thermal-hydraulics module instead of SAM for the
thermal-hydraulics analysis. This will enable this model to be coupled with a direct Brayton
cycle using the THM components. An assembly model was provided to the ANL team and will
not be discussed in this report.
2. A full balance of plant model using THM using a simplified core.
3. A thermal model of a full core including a subchannel model to assess the bypass flow effects.
5.1.1 Balance of Plant Model
The system is a primary loop with a high-temperature gas-cooled reactor coupled to an open-
air Brayton cycle through a heat exchanger. The code used is THM, which is open-source. A
diagram of the system is given in Figure 24.
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Figure 24. Balance of Plant diagram.
The core design is taken from [15] and consists of 55 hexagonal assemblies, using a graphite
matrix, which acts as a moderator. The core is 2 m long, and the sides of the hexagonal
assemblies measure 0.11 m. Each assembly has 42 fuel channels of TRISO particles in a graphite
matrix and 18 coolant channels. This design uses helium as a coolant. The helium extracts heat
from the core and releases it in the heat exchanger. Finally, a pump compensates for the loss of
pressure.
The secondary loop is a recuperated open-air Brayton cycle. Air is pumped in the loop by a
compressor. It is then heated by the exhaust gases in the recuperator and then in the heat
exchanger. The gas goes through a turbine, transfers a part of its residual heat to the
recuperator, and is finally released outside. Energy is extracted from the hot air by the turbine to
spin the generator shaft and produce electricity. A motor starts the shaft rotation. When the
turbine torque is high enough to compensate for the compressor and generator torque, the motor
is turned off. The operating conditions are given in Table 4.
Table 4. Operating conditions for the DOME microreactor use case.
Total core power
15 MWth
Primary mass flow rate
9.4 kg/s
Core inlet temperature
890 K
Core outlet temperature
1190 K
Primary system pressure
9MPa
Secondary mass flow rate
20 kg/s
Compressor pressure ratio
8.9
Turbine pressure ratio
2.9
Generator power
2 MWe
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For simplicity, the core is modeled using a single flow channel combining all the cooling
channels coupled with a representative heat structure. The core power is prescribed as a function
of time. The circulator is modeled using the homologous performance curves. A motor is connected
to the same shaft and the motor torque is controlled using a Proportional–Integral–Derivative
(PID) controller to match the nominal mass flow rate in the loop. On the secondary side, the
compressor, turbine, and generator share the same shaft. The performance curves provide the
efficiency and pressure ratio as a function of the shaft speed and flow rate. The generator is
modeled using a negative torque, proportional to the shaft speed. The torque and inertia of each
component connected to the shaft make contributions to the shaft speed equation.
A startup and a load-follow transient were performed. Figure 25 shows the power extracted
by the coolant in the core, the power generated and temperature evolution during a startup
transient.
Figure 25. Power and temperature evolution during a startup transient.
The load-follow transient starts from the steady-state conditions, then the reactor power is
lowered to 80% of its nominal value, then set back to 100%. Figure 26 shows the resulting
powers and temperatures. The power conversion system responds well to the power modulation,
and the generated power is 1.6 MW.
Figure 26. Power and temperature evolution during a load-follow transient.
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5.1.2 Full Core with Bypass Flow
In this model, the core design is similar to the one used for the balance of plant. A detailed
mesh of the core was created using the MOOSE reactor module. For compatibility with the
Pronghorn Subchannel code, four more assemblies are added on the corners. Gaps for the
bypass channels are introduced in the mesh. Power is prescribed in the fuel; the 3D heat
conduction problem is solved in the core. The full core model is coupled to the subchannel
simulation through a convective boundary condition. It is assumed that 5% of the total coolant flow
goes through the bypass. The resulting crossflow in the bypass is shown in Figure 27 with the
bypass and core temperatures.
(a) Bypass crossflow
(b) Bypass temperature
(c) Mid-core temperature
Figure 27. Bypass cross flow and temperature in the core and bypass.
5.1.3 Remaining Work
Remaining work includes:
• Develop a model coupling a high-fidelity multiphysics core model with the balance of plant
model that was developed this FY.
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• Couple the full core model with the Pronghorn Navier-stokes capability to include thermal
mixing in the plenum.
5.2 LOTUS Use Case: Low-Powered MSR
The Molten Chloride Reactor Experiment (MCRE) is planned to be demonstrated at LOTUS.
Even though this is a low power reactor, a multiphysics model of the reactor was built to ensure
the reactor integrity under operational transients and to evaluate sequences potentially leading
to accident conditions. The VTB scope intended to demonstrate the capability to model this type of
candidate reactor for LOTUS.
The approach integrates a comprehensive 3D multiphysics model, utilizing three key simulation
tools: Griffin for neutronics simulations, Pronghorn for thermal-hydraulics simulations, and BISON
for thermomechanics simulations. This integration enables the accurate representation of
steady-state and transient reactor operations. Behavior of the MSR during a loss of flow accident
were also investigated to offer insights into the system's transient response.
This section provides a concise description of the MSR concept studied, an overview of the
computational tools employed, and presents results encompassing steady-state operation as well
as the response to a loss-of-flow accident.
5.2.1 Reactor Specifications
The proposed design of the open-pool MSR is visually illustrated in Figure 28 and
accompanied by a grid indicating the reactor's dimensions. The design specifications were
obtained from Reference [16]. The geometry of the model encompasses a primary open core
cavity, a pump, and interconnected piping facilitating the flow of the liquid nuclear fuel between
the reactor and the pump. Enveloping the main core cavity is a reflector, which plays a crucial
role in the reactor's performance.
Figure 28. Longitudinal cross section of primary loop modeled for an MCRE-like reactor.
In this MSR configuration, the circulation of the liquid nuclear fuel takes place from the bottom
piping, ascending into the reactor cavity, and then flowing into the pump. To counteract the
reactivity-temperature coefficient that arises due to thermal expansion, a carefully designed
reflector is clamped to the reactor wall. The reflector, currently represented as a solid monolithic
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block in the model, is expected to be constructed using multiple blocks in practical applications to
minimize the impact of thermal deformations and ensure optimal performance.
Notably, the model does not explicitly incorporate control rods, relying instead on normalizing
the fission productions using the effective multiplication factor to maintain the reactor at criticality.
Additionally, a mixing plate is positioned at the entrance of the reactor cavity to promote uniform
flow distribution within the reactor, optimizing the overall operation.
To provide a comprehensive understanding of the MSR design, several key design parameters
are outlined in Table 5. The composition of the fuel salt employed in this reactor design is based
on the eutectic point of the UCl3-NaCl system, offering specific advantages and characteristics.
It is important to emphasize that, due to the relatively small size of the core, a higher fuel
enrichment level fuel is necessary to ensure a sufficiently large reactivity margin above criticality
during operation. In its current configuration without control rods inserted, the reactor exhibits a
reactivity excess of approximately 12,000 pcm. The mass flow rate within the reactor is
regulated at 50 kg/s, resulting in a noticeable pressure drop of 81 kPa across the reactor loop.
The fuel salt circulates with an approximate circulation time of 3 seconds, showcasing the dynamic
nature of the system.
For a more comprehensive understanding, Table 5 provides detailed thermophysical
properties of the fuel salt, enabling a thorough analysis of the reactor's performance and
behavior under different operating conditions. Last, it should be noted that the reflector material
employed in this design consists of MgO and does not undergo any specific purification processes
for its constituent elements, striking a balance between practicality and performance in the overall
reactor design.
Table 5. Parameters of modeled LOTUS MSR use case.
Parameter [Unit]
Value
Core Power [MW]
10
Operation Temperature [K]
900
Rated Mass Flow Rate [kg/s]
50
Fuel Salt Composition [mol%]
UCl3 [33.3%] - NaCl [66.7%]
Fuel Enrichment 235U [wt%]
93.2
Salt Density [kg/m3]
4212.6 1.
Specific Heat [J/(kg.K)]
8900. .
Thermal Conductivity [W/(m.K)]
5. . × 10
Dynamic Viscosity [Pa.s]
1. × 10
.×
.
5.2.2 Computational Codes
The neutronics modeling of the MSR utilizes a diffusion model, implemented with the Griffin
computational code. To derive the cross sections and effective parameters for the delay neutron
families, data condensation is performed using the Monte Carlo Serpent 2 code, known for its
accuracy and reliability in nuclear simulations.
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For the comprehensive thermal-hydraulics modeling of the MSR, including the reactor cavity,
pump, and connecting piping, the coarse-mesh thermal-hydraulics code Pronghorn is employed.
This code enables detailed analysis of the thermal and hydraulic behavior within the reactor
system, capturing intricate phenomena and providing valuable insights into the overall
performance.
In parallel, the thermomechanics model, responsible for studying the structural behavior and
integrity of the reactor reflector, is implemented using the Bison code. By considering the
thermomechanical interactions and responses, Bison plays a crucial role in assessing the structural
integrity and overall safety of the MSR design.
To ensure seamless integration and efficient interaction among the Griffin, Pronghorn, and
Bison codes, the BlueCRAB MOOSE suite is employed. This suite acts as a powerful platform that
enables the internal Picard iterations to be performed across all codes, establishing a tight
coupling between the different modeling aspects. The integration facilitated by BlueCRAB
MOOSE enhances the accuracy and consistency of the simulations, allowing for a comprehensive
analysis of the MSR's behavior under various operating conditions.
By combining the capabilities of Griffin, Pronghorn, and Bison within the BlueCRAB MOOSE
suite, this modeling approach provides a robust framework for studying the neutronics, thermal-
hydraulics, and thermomechanics aspects of the MSR system. It enables a holistic understanding of
the MCRE’s performance, contributing to the advanced understanding of the reactor dynamics and
safety evaluations.
5.2.3 Steady-State Reactor Operation
The continuous neutronics calculations for the MSR core are performed using the Serpent 2
code. Subsequently, the obtained cross sections are condensed into six energy groups to enable
efficient diffusion neutronics modeling in Griffin. To ensure comprehensive coverage, the cross
sections are tabulated at 20K intervals within the temperature range of 800 to 1,000 K. For
temperature values outside the tabulated points, Griffin internally employs interpolation
techniques to estimate the cross sections.
In Figure 29, the computed neutronics spectrum for the reactor core, obtained through
Serpent 2, is visualized alongside the six-group discretization represented by colored windows.
The process of energy bin optimization aims to minimize errors in reactivity and ensure an
accurate representation of the neutron flux shape. As a result, a finer discretization is
implemented at higher energies, aligning with the increased neutron flux observed within that
energy range.
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Figure 29. Continuous energy spectrum for MCRE-like reactor core with six-group energy
discretization represented in colored bins.
Following the energy bin optimization, a thorough comparison between Griffin and Serpent 2
reference data is conducted. The resulting error in reactivity is determined to be ± 76 pcm
across the entire temperature range, reflecting the close agreement between Griffin and the
accurate Serpent 2 calculations. Furthermore, the average L2 difference in the neutron flux
between Griffin and Serpent 2 results amounts to 0.7% ± 0.6% . These small discrepancies
further demonstrate the reliability and accuracy of the Griffin diffusion neutronics model in
capturing the behavior of the MSR across a wide range of temperatures.
By utilizing the advanced capabilities of Serpent 2 and the tailored discretization approach in
Griffin, this comprehensive analysis ensures precise and reliable neutronics calculations for the
MSR. The close agreement between Griffin and Serpent 2 data in terms of reactivity and neutron
flux attests to the robustness of the modeling methodology, laying a solid foundation for further
investigations and optimizations in MSR design and operation.
Figure 30 showcases the computed fast-most flux (group 1) and thermal-most flux (group 6)
obtained from the Griffin neutronics model. As expected, the fast flux demonstrates a
considerable amount of leakage, while the thermal flux diffuses into the reactor's reflector region.
Furthermore, Figure 30 includes the collision track plot, a continuous energy neutron flux estimator
computed by Serpent 2. Qualitatively, the predicted flux shape closely resembles the fast flux
due to the predominantly fast neutron characteristics of the neutron spectrum in the reactor.
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Figure 30. Neutron flux distribution computed by Griffin for the fast-most spectrum (left) and
second-most-thermal spectrum (center), and continuous energy collision track plot (right).
The power distribution within the reactor core cavity exhibits a concentration toward its central
region. It is important to highlight that these neutronics results have already incorporated the
feedback effects arising from temperature, density, and advection of delayed neutron precursors.
These feedback effects are accurately computed using the Pronghorn thermal-hydraulics code,
enabling a comprehensive analysis of the interplay between neutronics and thermal-hydraulics
phenomena within the MSR system.
The visualization of the computed fluxes and the inclusion of the collision track plot provide
valuable insights into the behavior and characteristics of neutron transport within the MSR. The
concentration of power in the central region of the core cavity highlights the significance of
optimizing the design and placement of control mechanisms and safety features. These results
combined with the integration of thermal-hydraulics feedback, contribute to a more
comprehensive understanding of the MSR's performance, facilitating the development of safer
and more efficient reactor designs.
The power distribution obtained from the neutronics model is transferred to the Pronghorn
thermal-hydraulics model, leading to results showcased in Figure 31. This figure presents the
vertical velocity distribution and temperature profile within the reactor cavity.
Figure 31. Thermal hydraulics fields steady-state operation of MCRE including fuel salt velocity in
the vertical direction (left) and temperature field (right). Note that while the temperature
variations may appear to be substantial, they only represent a variation within 10 K.
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The vertical velocity field demonstrates the downward movement of the liquid nuclear fuel as
it enters the reactor through the bottom pipe, following the flow path indicated by the outermost
pipe. Subsequently, the fuel enters the reactor cavity and undergoes partial equalization upon
reaching the lower mixing grid after the expansion of the reactor cavity. However, the current
configuration exhibits a higher velocity toward the right side of the reactor cavity due to
increased pressure resulting from injection through the bottom pipe. As a result, non-uniform
temperature fields are observed (Figure 31), with the right side exhibiting relatively cooler
temperatures due to faster fuel circulation, while the left side experiences hotter temperatures
due to slower circulation in that region.
Given the large mass flow rate within the reactor and the relatively low thermal power, the
overall temperature rise across the core is only 2K. It is worth noting that the pipes not in contact
with the neutron reflector are treated as adiabatic in this analysis. In practical implementations,
achieving this adiabatic condition is facilitated by insulation and thermal blankets applied to
these pipes, effectively minimizing heat exchange with the surroundings and maintaining stable
reactor performance.
By examining the vertical velocity and temperature distribution in the reactor cavity, this
analysis provides crucial insights into the flow behavior and thermal characteristics of the MSR
system. Understanding these factors enables the optimization of reactor design, fuel circulation,
and cooling mechanisms, contributing to enhanced safety and performance. Furthermore, the
negligible temperature rise across the core highlights the efficient heat transfer and effective
temperature control within the MSR.
Conjugated heat transfer is conducted between the reactor core and the surrounding reflector
to capture the intricate thermal interactions. The process involves employing Picard iterations to
ensure temperature consistency between the thermal-hydraulics and thermomechanics fields at the
interface where the reactor and reflector meet.
In the thermal analysis, the reflector is assumed to be cooled through natural convection to the
atmosphere via the presence of reflector insulation. The free convection coefficient is set at
3 W/(m2.K), while the external convection temperature is fixed at 300 K to approximate typical
ambient conditions.
Figure 32 displays the obtained results for the temperature and von Mises stress distributions
within the reflector. Notably, the skewed temperature profile, with higher temperatures toward
the top-left side, is consistently maintained in the reflector. Consequently, this non-uniform
temperature distribution induces corresponding non-uniform deformation of the reflector, with
preferential expansion on the top-left side. The highest stresses are observed internally within the
reflector, precisely in the region where it is clamped to the reactor wall.
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Figure 32. Thermomechanics fields in the nuclear reflector.
Note that, based on the current assumptions and analysis, the reflector is expected to
experience plastic deformation in its internal portion where contact with the reactor walls occurs.
This plastic deformation is a result of the applied thermal and mechanical loads, emphasizing the
significance of considering thermomechanical effects in reactor design and safety assessments.
The feedback of thermomechanics on neutronics must be considered as well. As the reflector
expands due to temperature variations, a larger flux of reflected neutrons enters the reactor
core, resulting in a positive reactivity-temperature coefficient for the reflector due to thermal
expansion. This effect competes with the Doppler effect, which increases neutron absorption in the
reflector and leads to a negative reactivity coefficient.
At the operating temperature, the reactivity-temperature feedback coefficient for the
reflector is found to be negative, with a value of -5.9 pcm/K. This coefficient is relatively smaller
compared to the reactivity-temperature feedback coefficient of the fuel salt, which is determined
to be -9.6 pcm/K. The disparity primarily arises from the expansion of the neutron reflector,
which is absent in the fuel salt. The additional volume of fuel salt produced by thermal expansion
is accommodated in the expansion vessel located outside of the reactor, ensuring the stability and
integrity of the overall system.
5.2.4 Loss of Forced-Flow Accident (LOFA)
After investigating the steady-state operation of the Molten Salt Reactor (MSR), an analysis of
a LOFA with heater still operational is conducted to assess the reactor's response under such
transient conditions. In this scenario, the reactor pump is assumed to initially stop, leading to a rise
in temperature within the system. Consequently, the reactor undergoes a transition to a subcritical
state, where the power generated within the reactor is primarily derived from residual nuclear
heating. Due to the gradual decrease in flow caused by inertia, a small circulation speed is
established through natural convection within the reactor. Eventually, a new equilibrium condition is
reached after approximately 120 seconds.
Figure 33 illustrates the results of the equilibrium vertical velocity and temperature fields
obtained from the LOFA analysis. The control tube at the center of the reactor has been included
in the transient model as it plays a role in limiting natural convection. As heat is predominantly lost
through thermal conduction to the reactor reflector, the temperature field becomes more uniform,
with the influence of temperature convection diminishing. However, it is worth noting that the top
part of the reactor core remains hotter than the bottom part due to the gradual heating of the
flow as it circulates from the bottom to the top of the reactor.
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Figure 33. Velocity and temperature fields after 120 seconds of the beginning of the LOFA.
Within the reactor core, the established natural convection gives rise to a few reticulation
eddies. On average, a positive vertical circulation speed across the core is observed, indicating
the presence of flow motion induced by natural convection. However, it is important to highlight
that the average circulation speed within the reactor core is approximately 100 times smaller
than the speed observed when the pump was actively promoting flow during normal operation.
This decrease in velocity is expected, considering the reactor's relatively low power output and
the limited height difference between the hottest and coldest regions within the reactor.
The LOFA analysis provides valuable insights into the behavior and dynamics of the MSR
under transient conditions. The observed temperature field and natural convection patterns
contribute to understanding the thermal response and heat transfer characteristics during such
accidents. By simulating and analyzing LOFA scenarios, engineers and researchers can further
optimize safety measures, improve reactor designs, and ensure the integrity and resilience of MSR
systems.
5.2.5 Remaining work
Remaining work includes the following:
• Develop a model for thermal-radiation cooling of the nuclear reactor core
• Improve pump modeling in the reactor simulation, which is currently treated as a volumetric
momentum source.
6. PUBLICATIONS AND OUTREACH
In an effort to increase the visibility of the VTB, the team held several outreach activities with
a wide range of stakeholders. These activities are briefly summarized below. The team had
several outreach activities with varying stakeholders listed below:
• Industry presentations: Southern Company, TerraPower-Natrium, TerraPower-MCRE,
Westinghouse, Radiant.
• Government outreach: US Nuclear Regulatory Commission, UK Fusion Programmes.
• Universities: Georgia Tech, Massachusetts Institute of Technology, University of Texas
Austin.
During FY23, the VTB project outputted several publications both in the form of conferences
and journal publications. They are summarized below:
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• Two special sessions on the VTB were held at ANS Winter conference in November 2022.
This included 10 paper submissions sponsored by the VTB.
• A journal publication on the VTB:
• G. Guidicelli et al., “The Virtual Test (VTB) Repository: A Library of Reference Reactor
Models Using NEAMS Tools”, Nuclear Science and Engineering, 1-17, DOI:
10.1080/00295639.2022.2142440, (2023)
• The team has also been working diligently on further publications: Another special session
at the ANS Winter 2023 meeting on the VTB is being organized (November 2023).
Multiple papers have been submitted for the conference and are being reviewed.
7. REFERENCES
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[12] S. Terlizzi and V. Laboure. 2023. “Asymptotic hydrogen redistribution analysis in yttrium-
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