List of Publications

@article{lindsay_moltres_2018,
    author = "Lindsay, Alexander and Huff, Kathryn",
    title = "Moltres: finite element based simulation of molten salt reactors",
    volume = "3",
    shorttitle = "Moltres",
    doi = "10.21105/joss.00298",
    abstract = "Moltres is a physics application for multiphysics modeling of fluid-fueled molten salt reactors (MSRs) (Lindsay et al. 2018). It couples equations for neutron diffusion, thermal hydraulics, and delayed neutron precursor transport. Neutron diffusion and precursor transport equations are set-up using an action system that allows the user to use an arbitrary number of neutron energy and precursor groups respectively with minimal input changes. Moltres sits on top of the Multi-physics Object-Oriented Simulation Environment (Gaston et al. 2015) and hence uses the finite element method to discretize the governing partial differential equations. In general the resulting system of non-linear algebraic equations is linearized using the Newton-Raphson method and then solved using the Portable, Extensible Toolkit for Scientific Computation (Balay et al. 2017). Assembly of the Jacobian and residual, and the linear solve are parallelized using MPI which allows Moltres to be run in massively parallel environments. Runs on the Blue Waters supercomputer at Illinois have utilized up to 608 cores. Moltres and MOOSE allow use of different basis functions for different system variables. Because of the purely diffusive nature of the neutron diffusion equations, neutron fluxes are typically discretized using continuous first-degree Lagrange polynomials and the degrees of freedom are associated with mesh nodes. The temperature variable may also be discretized with a continuous Lagrange basis, or a discontinuous basis of arbitrary degree monomials may be employed depending on the relative balance of heat convection to conduction. The purely hyperbolic precursor transport is currently discretized using constant monomials, which is equivalent to a first-order finite volume discretization. Moltres supports both segregated (through Picard iteration) and monolithic solutions of the equation system. However, due to the feedback between the power spectrum and temperature dependence of macroscopic cross-sections, monolithic solves have demonstrated superior robustness with segregated techniques often unable to converge to a solution. This result emphasizes the importance of a fully coupled multi-physics framework like the one that Moltres and MOOSE provide and suggests that iteratively coupling codes devoted to single physics (Kópházi, Lathouwers, and Kloosterman 2009) may result in limited flexibility.",
    number = "21",
    urldate = "2018-01-08",
    journal = "The Journal of Open Source Software",
    month = "January",
    year = "2018",
    pages = "1--2",
    file = "Full Text PDF:C\:\\Users\\Sun Myung\\Zotero\\storage\\MJIZZW4P\\Lindsay and Huff - 2018 - Moltres finite element based simulation of molten.pdf:application/pdf;Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\E3ARQ46H\\joss.html:text/html"
}

@article{lindsay_introduction_2018,
    author = "Lindsay, Alexander and Ridley, Gavin and Rykhlevskii, Andrei and Huff, Kathryn",
    title = "Introduction to {Moltres}: {An} application for simulation of {Molten} {Salt} {Reactors}",
    volume = "114",
    issn = "0306-4549",
    shorttitle = "Introduction to {Moltres}",
    url = "https://linkinghub.elsevier.com/retrieve/pii/S0306454917304760",
    doi = "10.1016/j.anucene.2017.12.025",
    abstract = "Moltres is a new physics application for modeling coupled physics in fluid-fuelled, molten salt reactors. This paper describes its neutronics model, thermal hydraulics model, and their coupling in the MOOSE framework. Neutron and precursor equations are implemented using an action system that allows use of an arbitrary number of groups with no change in the input card. Results for many-channel configurations in 2D-axisymmetric and 3D coordinates are presented and compared against other coupled models as well as the Molten Salt Reactor Experiment.",
    language = "en",
    urldate = "2018-01-08",
    journal = "Annals of Nuclear Energy",
    month = "April",
    year = "2018",
    keywords = "Reactor physics, Parallel computing, agent based modeling, Finite elements, Hydrologic contaminant transport, MOOSE, Multiphysics, nuclear engineering, Nuclear fuel cycle, Object orientation, repository, Simulation, Systems analysis",
    pages = "530--540",
    file = "Lindsay et al. - 2018 - Introduction to Moltres An application for simula.pdf:C\:\\Users\\Sun Myung\\Zotero\\storage\\RCWUNGTP\\Lindsay et al. - 2018 - Introduction to Moltres An application for simula.pdf:application/pdf;Lindsay et al. - 2018 - Introduction to Moltres An application for simula.pdf:C\:\\Users\\Sun Myung\\Zotero\\storage\\3GEC6NQ9\\Lindsay et al. - 2018 - Introduction to Moltres An application for simula.pdf:application/pdf;Moltres.pdf:C\:\\Users\\Sun Myung\\Zotero\\storage\\4XDXRICB\\Moltres.pdf:application/pdf;ScienceDirect Full Text PDF:C\:\\Users\\Sun Myung\\Zotero\\storage\\E2T9U5IX\\Lindsay et al. - 2018 - Introduction to Moltres An application for simula.pdf:application/pdf;ScienceDirect Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\3DT9TEY3\\S0306454917304760.html:text/html"
}

@article{park_verification_2022,
    author = "Park, Sun Myung and Munk, Madicken",
    title = "Verification of moltres for multiphysics simulations of fast-spectrum molten salt reactors",
    volume = "173",
    issn = "03064549",
    url = "https://linkinghub.elsevier.com/retrieve/pii/S0306454922001463",
    doi = "10.1016/j.anucene.2022.109111",
    abstract = "Modeling strongly coupled neutronics and thermal–hydraulics in liquid-fueled MSRs requires robust and flexible multiphysics software for accurate simulations at reasonable computational costs. In this paper, we present Moltres and its neutronics and thermal–hydraulics modeling capabilities relevant to multiphysics reactor analysis. As a MOOSE-based application, Moltres provides various multiphysics coupling schemes and time-stepping methods, including fully coupled solves with implicit time-stepping. We verified Moltres’ MSR modeling capabilities against a multiphysics numerical benchmark developed for software dedicated to modeling fast-spectrum MSRs. The results show that Moltres performed comparably to participating software packages in the benchmark; the majority of the relevant quantities fell within one standard deviation of the benchmark average. Among the participating multiphysics tools in the benchmark, Moltres agrees closest to the multiphysics tool from the Delft University of Technology due to similarities in the numerical solution techniques and meshing schemes.",
    language = "en",
    urldate = "2022-04-26",
    journal = "Annals of Nuclear Energy",
    month = "August",
    year = "2022",
    pages = "109111",
    file = "Park and Munk - 2022 - Verification of moltres for multiphysics simulatio.pdf:C\:\\Users\\Sun Myung\\Zotero\\storage\\PHNTVU4R\\Park and Munk - 2022 - Verification of moltres for multiphysics simulatio.pdf:application/pdf"
}

@inproceedings{park_multiphysics_2021,
    author = "Park, Sun Myung and Huff, Kathryn D",
    address = "Virtual Meeting",
    series = "Reactor {Analysis} {Methods} - {I}",
    title = "Multiphysics {Benchmark} {Results} from {Moltres}",
    url = "https://www.ans.org/meetings/am2021/session/view-587/",
    booktitle = "Proceedings of the 2021 {ANS} {Virtual} {Annual} {Meeting}",
    publisher = "American Nuclear Society",
    month = "June",
    year = "2021",
    note = "(Submitted before May 2021)",
    file = "Park and Huff - 2021 - Multiphysics Benchmark Results from Moltres.pdf:C\:\\Users\\Sun Myung\\Zotero\\storage\\DI37TPGK\\Park and Huff - 2021 - Multiphysics Benchmark Results from Moltres.pdf:application/pdf;Reactor Analysis Methods - I -- ANS / Meetings / 2021 ANS Virtual Annual Meeting / Technical Sessions:C\:\\Users\\Sun Myung\\Zotero\\storage\\PSC9JE66\\view-587.html:text/html"
}

@inproceedings{park_safety_2019,
    author = "Park, Sun Myung and Rykhlevskii, Andrei and Huff, Kathryn",
    address = "Seattle, WA, United States",
    title = "Safety {Analysis} of the {Molten} {Salt} {Fast} {Reactor} {Fuel} {Composition} using {Moltres}",
    url = "http://epubs.ans.org/?a=47030",
    doi = "10.31224/osf.io/7ce89",
    abstract = "The Molten Salt Fast Reactor (MSFR) has garnered much interest for its inherent safety and sustainbility features. The MSFR can adopt a closed thorium fuel cycle for sustainable operation through the breeding of 233 U from 232 Th. The fuel composition changes significantly over the course of its lifespan. In this study, we investigated the steady state and transient behavior of the MSFR using Moltres, a coupled neutronics/thermal-hydraulics code developed within the Multiphysics Object Oriented Simulation Environment (MOOSE) framework. Three different fuel compositions, start-up, early-life, and equilibrium, were examined for potentially dangerous core temperature excursions during a unprotected loss of heat sink (ULOHS) accident. The six-group and total neutron flux distributions showed good agreement with SERPENT and published MSFR results, while the temperature distribution and total power showed discrepancies which can be attributed toknown sources of error. For the transient behavior under the ULOHS scenario, while the transition time towards the new steady state core temperature is also in good agreement with existing MSFR simulations by Fiorina et al., Moltres under-estimated the temperature rise by a factor of ten, due to the same sources of error affecting the steady state results. While an MSFR loaded with start-up fuel composition operates at a higher temperature than with the other two fuel compositions, all three cases were shown to be inherently safe due to thestrong negative temperature feedback.",
    urldate = "2019-11-05",
    booktitle = "Proceedings of {GLOBAL} {International} {Fuel} {Cycle} {Conference}",
    publisher = "American Nuclear Society",
    month = "September",
    year = "2019",
    file = "Park et al. - 2019 - Safety Analysis of Molten Salt Fast Reactor Fuel C.pdf:C\:\\Users\\Sun Myung\\Zotero\\storage\\EXYUTUGW\\Park et al. - 2019 - Safety Analysis of Molten Salt Fast Reactor Fuel C.pdf:application/pdf"
}

@mastersthesis{park_advancement_2020,
    author = "Park, Sun Myung",
    address = "Urbana, IL",
    title = "Advancement and {Verification} of {Moltres} for {Molten} {Salt} {Reactor} {Safety} {Analysis}",
    copyright = "Copyright 2020 Sun Myung Park",
    url = "https://www.ideals.illinois.edu/handle/2142/108542",
    language = "English",
    school = "University of Illinois at Urbana-Champaign",
    month = "August",
    year = "2020",
    file = "Park - 2020 - Advancement and Verification of Moltres for Molten.pdf:C\:\\Users\\Sun Myung\\Zotero\\storage\\JYYYTBJ7\\Park - 2020 - Advancement and Verification of Moltres for Molten.pdf:application/pdf"
}

@mastersthesis{fairhurst-agosta_multi-physics_2020,
    author = "Fairhurst-Agosta, Roberto",
    address = "Urbana, IL",
    title = "Multi-{Physics} and {Technical} {Analysis} of {High}-{Temperature} {Gas}-{Cooled} {Reactors} for {Hydrogen} {Production}",
    copyright = "Copyright 2020 Roberto Fairhurst Agosta",
    abstract = "The future energy needs require the development of clean energy sources to ease the increasing environmental concerns. High-Temperature Gas-cooled Reactors have several desirable features that make them ideal candidates for the near-future large-scale deployment. Some of these features are a high temperature and high thermal cycle efficiency, which enable a wide range of process heat applications, such as hydrogen production. Implementing hydrogen economies can decarbonize the transport and power sectors, offering an alternative to ease climate change. This work uses Moltres as the primary simulation tool. Although Moltres original development targeted Molten Salt Reactors, this work studies Moltres applicability to multi-physics simulations of prismatic High-Temperature Gas-cooled Reactors. Multi-physics simulations are necessary for assessing reactor safety characteristics. Ensuring Moltres’ multi-physics modeling capabilities requires assessing the independent modeling capabilities of the different physical phenomena. Therefore, this thesis breaks down the analysis into three parts: stand-alone neutronics, stand-alone thermal-fluids, and coupled neutronics/thermal-fluids. Regarding stand-alone neutronics, several analyses compare the results calculated by Moltres and Serpent on an MHTGR-350 model. The first analysis studies the energy group structure effects on the simulation of a fuel column. The results of the study suggest using a 15-energy group structure for attaining a desirable accuracy. The following analysis focuses on the full-core problem and compares different aspects of the simulations, concluding that Moltres obtains reasonably accurate results. The final study on stand-alone neutronics describes Moltres results of Phase I Exercise 1 of the OECD/NEA MHTGR-350 Benchmark. The benchmark exercise proved to be a modeling challenge, requiring the implementation of several approximations. For the most part, this thesis demonstrates Moltres’ capability to simulate stand-alone neutronics of prismatic High-Temperature Gas-cooled Reactors. Regarding stand-alone thermal-fluids, several studies compare Moltres results to previously published results. These studies focus on local models such as the unit cell and the fuel column problems, for which Moltres temperature results differ by less than 2\\% from the published results. Further studies analyze the possibility of extending the thermal-fluids model implemented in the previous problems to a full-core simulation, finding a high memory requirement imposed by the simulations. The full-core simulations focus on Phase I Exercise 2 of the benchmark, for which the implementation of a two-level approach in Moltres was necessary. The study’s temperatures were within an 11.3\\% difference to the published results, concluding that further analysis is required. Regarding coupled neutronics/thermal-fluids, the analysis describes Phase I Exercise 3 of the benchmark. The exercise uses a simplified model that helps visualize some of the essential aspects of multi-physics simulations in Moltres. This exercise finds some areas of improvement in Moltres’ model and sets a basis for future work. This thesis aligns with the University of Illinois’ goals to reduce carbon emissions from its campus’s electricity generation and transportation sectors. This work focuses on two main analysis by introducing a nuclear reactor coupled to a hydrogen plant as a solution. The first analysis evaluates the conversion of the university fleet and the mass transit transport system in Urbana-Champaign to Fuel Cell Electric Vehicles. The second analysis investigates the duck curve phenomenon in the university’s grid and introduces a mitigation strategy that may reduce the reliance on dispatchable sources. These studies emphasize how nuclear energy and hydrogen production can potentially mitigate climate change.",
    language = "English (US)",
    school = "University of Illinois at Urbana-Champaign",
    month = "December",
    year = "2020",
    file = "Fairhurst-Agosta - Multi-Physics and Technical Analysis of High-Tempe.pdf:C\:\\Users\\Sun Myung\\Zotero\\storage\\PTK9576U\\Fairhurst-Agosta - Multi-Physics and Technical Analysis of High-Tempe.pdf:application/pdf"
}

@mastersthesis{lee_neutronics_2020,
    author = "Lee, Alvin J. H.",
    address = "Urbana, IL",
    title = "Neutronics and {Thermal}-{Hydraulics} {Analysis} of {TransAtomic} {Power} {Molten} {Salt} {Reactor} ({TAP} {MSR}) {Core} {Under} {Load} {Following} {Operations}",
    shorttitle = "Neutronics and thermal-hydraulics analysis of transatomic power molten salt reactor ({TAP} {MSR}) core under load following operations",
    url = "https://www.ideals.illinois.edu/bitstream/handle/2142/109415/LEE-THESIS-2020.pdf?sequence=1\&isAllowed=y",
    abstract = "This work analyzed the neutronics and thermal-hydraulics behavior of the Transatomic Power Molten Salt Reactor (TAP MSR) core under load-following operations using a Monte Carlo code, Serpent 2, as well as a UIUC-developed MOOSE-based code, Moltres. A simulation method was developed to determine the operational bounds of the TAP MSR core and its transient behavior under rapid power ramps using both fresh fuel salt and fuel salt with equilibrium 135Xe. The thermal-hydraulics investigation studied the potential of exceeding material temperature constraints under simple advection flow versus a flow simulated with Incompressible Navier-Stokes physics. Finally, the effects of gas entrainment on the reactor core behavior were investigated. This study concludes that the TAP MSR core is able to perform rapid load following operations without exceeding its thermal safety constraints. The findings in this work were derived from research performed under the DOE ARPA-E MEITNER project, DE-AR0000983.",
    school = "University of Illinois at Urbana-Champaign",
    month = "December",
    year = "2020",
    file = "Lee - 2020 - Neutronics and Thermal-Hydraulics Analysis of Tran.pdf:C\:\\Users\\Sun Myung\\Zotero\\storage\\73JPRPQ3\\Lee - 2020 - Neutronics and Thermal-Hydraulics Analysis of Tran.pdf:application/pdf"
}

@mastersthesis{pater_multiphysics_2019,
    author = "Pater, Mateusz",
    title = "Multiphysics simulations of {Molten} {Salt} {Reactors} using the {Moltres} code",
    copyright = "http://creativecommons.org/licenses/by-nc-sa/3.0/es/",
    url = "https://upcommons.upc.edu/handle/2117/173747",
    language = "cat",
    urldate = "2022-05-04",
    school = "Universitat Politècnica de Catalunya",
    month = "November",
    year = "2019",
    note = "Accepted: 2019-12-11T11:03:14Z Publisher: Universitat Politècnica de Catalunya",
    keywords = "Àrees temàtiques de la UPC::Física, Nuclear engineering--Safety measures, Reactors nuclears -- Mesures de seguretat -- Simulació per ordinador",
    file = "Full Text PDF:C\:\\Users\\Sun Myung\\Zotero\\storage\\NWWJN9XA\\Pater - 2019 - Multiphysics simulations of Molten Salt Reactors u.pdf:application/pdf;Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\QZJV3C6Z\\173747.html:text/html"
}

@phdthesis{chee_fluoride-salt-cooled_2022,
    author = "Chee, Gwendolyn Jin Yi",
    address = "Urbana, IL",
    type = "Dissertation",
    title = "Fluoride-{Salt}-{Cooled} {High} {Temperature} {Reactor} {Design} {Optimization} with {Evolutionary} {Algorithms}",
    copyright = "Copyright 2021 Gwendolyn Jin Yi Chee",
    url = "https://github.com/arfc/2022-chee-dissertation",
    abstract = "Additive manufacturing of reactor core components removes the geometric constraints required by conventional manufacturing, such as slabs as fuel planks and cylinders as fuel rods. Due to the expansion of the potential design space facilitated through additive manufacturing, reactor designers need to find methods, such as generative design, to explore the design space efficiently. In this defense, I will show that I successfully applied evolutionary algorithms to conduct generative reactor design optimization for a fluoride-salt-cooled high-temperature reactor (FHR). I achieved this through three distinct research efforts: 1) furthering our understanding of the FHR design’s complexities through neutronics and temperature modeling, 2) creating an open-source tool that enables generative design reactor optimization with evolutionary algorithms, and 3) applying the tool to the FHR to optimize for non-conventional geometries and fuel distributions",
    school = "University of Illinois at Urbana-Champaign",
    month = "August",
    year = "2022",
    file = "2022-chee-dissertation-pres.pdf:C\:\\Users\\Sun Myung\\Zotero\\storage\\Y93FYTHR\\2022-chee-dissertation-pres.pdf:application/pdf;Chee - 2021 - Fluoride-Salt-Cooled High Temperature Reactor Desi.pdf:C\:\\Users\\Sun Myung\\Zotero\\storage\\EXF7M46A\\Chee - 2021 - Fluoride-Salt-Cooled High Temperature Reactor Desi.pdf:application/pdf"
}