bibliography.bib

@article{podila_cfd_2019,
	title = {{CFD} {Simulations} of {Molten} {Salt} {Reactor} {Experiment} {Core}},
	volume = {193},
	issn = {0029-5639},
	doi = {10.1080/00295639.2019.1627177},
	abstract = {At present, no clear guidelines exist for modeling non-water-cooled small modular reactors (SMRs) despite the rising need for high-fidelity simulation tools to support regulators and the industry. Most SMR concepts currently under the Canadian prelicensing review adopted non-water-cooled–reactor technologies [molten salt reactor (MSR), gas-cooled reactor, and liquid metal–cooled reactor] that are new for Canada. There is a need for a modeling tool set that is broadly applicable for the assessment of advanced technologies used in SMRs. Computational fluid dynamics (CFD) can be used in performance evaluation and safety analysis of non-water-cooled SMRs for modeling three-dimensional (3-D) fluid flow and heat transfer in geometries of arbitrary complexity without resorting to geometry-specific empirical correlations. This study investigates the capabilities of existing models within a commercial CFD code to simulate the flow and heat transfer characteristics in a MSR configuration. The Oak Ridge National Laboratory (ORNL) Molten Salt Reactor Experiment (MSRE) configuration was simulated in this study using a stand-alone CFD approach, and CFD predictions were assessed with ORNL data. Intricate geometry details within the MSRE core were included in the computational model to study the associated geometric effects. The results obtained in this study showcased the ability of CFD to predict 3-D effects within the computational domain especially at the lower plenums. The predicted trends for the temperature rise in the fuel and moderator within the core were in good agreement with the ORNL data. The results presented in this paper constitute the first step in developing Canadian Nuclear Laboratories’ capability for CFD modeling of non-water SMRs.},
	number = {12},
	journal = {Nuclear Science and Engineering},
	author = {Podila, Krishna and Chen, Qi and Rao, Yanfei},
	month = dec,
	year = {2019},
	keywords = {Computational fluid dynamics, molten salt, Molten Salt Reactor Experiment, simulation, small modular reactor},
	pages = {1379--1393},
	file = {Full Text PDF:C\:\\Users\\Sun Myung\\Zotero\\storage\\VT8J9AME\\Podila et al. - 2019 - CFD Simulations of Molten Salt Reactor Experiment .pdf:application/pdf;Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\U57KPW8L\\00295639.2019.html:text/html},
}

@article{zhiyin_large-eddy_2015,
	title = {Large-eddy simulation: {Past}, present and the future},
	volume = {28},
	issn = {1000-9361},
	shorttitle = {Large-eddy simulation},
	doi = {10.1016/j.cja.2014.12.007},
	abstract = {Large-eddy simulation (LES) was originally proposed for simulating atmospheric flows in the 1960s and has become one of the most promising and successful methodology for simulating turbulent flows with the improvement of computing power. It is now feasible to simulate complex engineering flows using LES. However, apart from the computing power, significant challenges still remain for LES to reach a level of maturity that brings this approach to the mainstream of engineering and industrial computations. This paper will describe briefly LES formalism first, present a quick glance at its history, review its current state focusing mainly on its applications in transitional flows and gas turbine combustor flows, discuss some major modelling and numerical challenges/issues that we are facing now and in the near future, and finish with the concluding remarks.},
	language = {en},
	number = {1},
	journal = {Chinese Journal of Aeronautics},
	author = {Zhiyin, Yang},
	month = feb,
	year = {2015},
	keywords = {Gas turbine combustor, Inflow boundary condition generation methods, Large-eddy simulation (LES), Sub-grid scale (SGS) model, Turbulent flows},
	pages = {11--24},
	file = {Accepted Version:C\:\\Users\\Sun Myung\\Zotero\\storage\\TPZKAAN5\\Zhiyin - 2015 - Large-eddy simulation Past, present and the futur.pdf:application/pdf;ScienceDirect Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\KHB626M9\\S1000936114002064.html:text/html},
}

@incollection{baldwin_thin-layer_1978,
	series = {Aerospace {Sciences} {Meetings}},
	title = {Thin-layer approximation and algebraic model for separated turbulentflows},
	booktitle = {16th {Aerospace} {Sciences} {Meeting}},
	publisher = {American Institute of Aeronautics and Astronautics},
	author = {Baldwin, B. and Lomax, H.},
	month = jan,
	year = {1978},
	keywords = {Astronautics, Boundary Layer Thickness, Coefficient of Viscosity, Drag Coefficient, Eddy Viscosity Turbulence Models, Law of the Wall, Lift Coefficient, Navier Stokes Equations, Separated Flows, Skin Friction},
}

@book{smith_numerical_1967,
	address = {Long Beach, Calif.},
	title = {Numerical solution of the turbulent-boundary-layer equations},
	language = {English},
	publisher = {Douglas Aircraft Co., Douglas Aircraft Division},
	author = {Smith, A. M. O and Cebeci, Tuncer},
	year = {1967},
}

@article{prandtl_7_1925,
	title = {7. {Bericht} über {Untersuchungen} zur ausgebildeten {Turbulenz}},
	volume = {5},
	issn = {1521-4001},
	doi = {10.1002/zamm.19250050212},
	language = {en},
	number = {2},
	journal = {ZAMM - Journal of Applied Mathematics and Mechanics / Zeitschrift für Angewandte Mathematik und Mechanik},
	author = {Prandtl, L.},
	year = {1925},
	pages = {136--139},
	file = {Full Text PDF:C\:\\Users\\Sun Myung\\Zotero\\storage\\FL28EJUE\\Prandtl - 1925 - 7. Bericht über Untersuchungen zur ausgebildeten T.pdf:application/pdf;Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\XKS5CTNK\\zamm.html:text/html},
}

@book{rodi_turbulence_2017,
	address = {London},
	edition = {3},
	title = {Turbulence {Models} and {Their} {Application} in {Hydraulics}: {A} state-of-the-art review},
	isbn = {978-0-203-73489-6},
	shorttitle = {Turbulence {Models} and {Their} {Application} in {Hydraulics}},
	abstract = {This book provides an introduction to the subject of turbulence modelling in a form easy to understand for anybody with a basic background in fluid mechanics, and it summarizes the present state of the art. Individual models are described and examined for the merits and demerits which range from the simple Prandtl mixing length theory to complex second order closure schemes.},
	publisher = {Routledge},
	editor = {Rodi, Wolfgang},
	month = oct,
	year = {2017},
	file = {Rodi - 2017 - Turbulence Models and Their Application in Hydraul.pdf:C\:\\Users\\Sun Myung\\Zotero\\storage\\DM6N4PAH\\Rodi - 2017 - Turbulence Models and Their Application in Hydraul.pdf:application/pdf},
}

@article{lasher_computation_1992,
	title = {On the computation of turbulent backstep flow},
	volume = {13},
	issn = {0142727X},
	doi = {10.1016/0142-727X(92)90057-G},
	language = {en},
	number = {1},
	journal = {International Journal of Heat and Fluid Flow},
	author = {Lasher, William C. and Taulbee, Dale B.},
	month = mar,
	year = {1992},
	pages = {30--40},
	file = {Lasher and Taulbee - 1992 - On the computation of turbulent backstep flow.pdf:C\:\\Users\\Sun Myung\\Zotero\\storage\\GB78FQQG\\Lasher and Taulbee - 1992 - On the computation of turbulent backstep flow.pdf:application/pdf},
}

@article{driver_features_1985,
	title = {Features of a reattaching turbulent shear layer in divergent channelflow},
	volume = {23},
	issn = {0001-1452, 1533-385X},
	doi = {10.2514/3.8890},
	language = {en},
	number = {2},
	journal = {AIAA Journal},
	author = {Driver, David M. and Seegmiller, H. Lee},
	month = feb,
	year = {1985},
	pages = {163--171},
	file = {Driver and Seegmiller - 1985 - Features of a reattaching turbulent shear layer in.pdf:C\:\\Users\\Sun Myung\\Zotero\\storage\\Z9UNHGAV\\Driver and Seegmiller - 1985 - Features of a reattaching turbulent shear layer in.pdf:application/pdf},
}

@patent{leblanc_integral_2015,
	title = {Integral molten salt reactor},
	abstract = {Abstract: The present relates to the integration of the primary functional elements of graphite moderator and reactor vessel and/or primary heat exchangers and/or control rods into an integral molten salt nuclear reactor (IMSR). Once the design life of the IMSR is reached, for example, in the range of 3 to 10 years, it is disconnected, removed and replaced as a unit. The spent IMSR functions as the medium or long term storage of the radioactive graphite and/or heat exchangers and/or control rods and/or fuel salt contained in the vessel of the IMSR. The present also relates to a nuclear reactor that has a buffer salt surrounding the nuclear vessel. During normal operation of the nuclear reactor, the nuclear reactor operates at a temperature that is lower than the melting point of the buffer salt and the buffer salt acts as a thermal insulator. Upon loss of external cooling, the temperature of the nuclear reactor increases and melts the buffer salt, which can then transfer heat from the nuclear core to a cooled containment vessel.},
	assignee = {Terrestrial Energy Inc.},
	number = {WO2015017928 A1},
	author = {LeBlanc, David},
	month = feb,
	year = {2015},
}

@article{brovchenko_neutronic_2019,
	title = {Neutronic benchmark of the molten salt fast reactor in the frame of the {EVOL} and {MARS} collaborative projects},
	volume = {5},
	copyright = {© M. Brovchenko et al., published by EDP Sciences, 2019},
	issn = {2491-9292},
	doi = {10.1051/epjn/2018052},
	abstract = {This paper describes the neutronic benchmarks and the results obtained by the various participants of the FP7 project EVOL and the ROSATOM project MARS. The aim of the benchmarks was two-fold: first to verify and validate each of the code packages of the project partners, adapted for liquid-fueled reactors, and second to check the dependence of the core characteristics to nuclear data set for application on a molten salt fast reactor (MSFR). The MSFR operates with the thorium fuel cycle and can be started with {\textless}sup{\textgreater}233{\textless}sup/{\textgreater}U-enriched U and/or TRU elements as initial fissile load. All three compositions were covered by the present benchmark. The calculations have confirmed that the MSFR has very favorable characteristics not present in other Gen4 fast reactors, like strong negative temperature and void reactivity coefficients, a low-fissile inventory, a reduced long-lived waste production and its burning capacities of nuclear waste produced in currently operational reactors.},
	language = {en},
	journal = {EPJ Nuclear Sciences \& Technologies},
	author = {Brovchenko, Mariya and Kloosterman, Jan-Leen and Luzzi, Lelio and Merle, Elsa and Heuer, Daniel and Laureau, Axel and Feynberg, Olga and Ignatiev, Victor and Aufiero, Manuele and Cammi, Antonio and Fiorina, Carlo and Alcaro, Fabio and Dulla, Sandra and Ravetto, Piero and Frima, Lodewijk and Lathouwers, Danny and Merk, Bruno},
	month = jan,
	year = {2019},
	pages = {2},
	file = {Full Text PDF:C\:\\Users\\Sun Myung\\Zotero\\storage\\ET8QPY6L\\Brovchenko et al. - 2019 - Neutronic benchmark of the molten salt fast reacto.pdf:application/pdf;Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\XMFKGGX5\\epjn180012.html:text/html},
}

@article{chadwick_endf/b-vii.1_2011,
	series = {Special {Issue} on {ENDF}/{B}-{VII}.1 {Library}},
	title = {{ENDF}/{B}-{VII}.1 {Nuclear} {Data} for {Science} and {Technology}: {Cross} {Sections}, {Covariances}, {Fission} {Product} {Yields} and {Decay} {Data}},
	volume = {112},
	issn = {0090-3752},
	shorttitle = {{ENDF}/{B}-{VII}.1 {Nuclear} {Data} for {Science} and {Technology}},
	doi = {10.1016/j.nds.2011.11.002},
	abstract = {The ENDF/B-VII.1 library is our latest recommended evaluated nuclear data file for use in nuclear science and technology applications, and incorporates advances made in the five years since the release of ENDF/B-VII.0. These advances focus on neutron cross sections, covariances, fission product yields and decay data, and represent work by the US Cross Section Evaluation Working Group (CSEWG) in nuclear data evaluation that utilizes developments in nuclear theory, modeling, simulation, and experiment. The principal advances in the new library are: (1) An increase in the breadth of neutron reaction cross section coverage, extending from 393 nuclides to 423 nuclides; (2) Covariance uncertainty data for 190 of the most important nuclides, as documented in companion papers in this edition; (3) R-matrix analyses of neutron reactions on light nuclei, including isotopes of He, Li, and Be; (4) Resonance parameter analyses at lower energies and statistical high energy reactions for isotopes of Cl, K, Ti, V, Mn, Cr, Ni, Zr and W; (5) Modifications to thermal neutron reactions on fission products (isotopes of Mo, Tc, Rh, Ag, Cs, Nd, Sm, Eu) and neutron absorber materials (Cd, Gd); (6) Improved minor actinide evaluations for isotopes of U, Np, Pu, and Am (we are not making changes to the major actinides 235,238U and 239Pu at this point, except for delayed neutron data and covariances, and instead we intend to update them after a further period of research in experiment and theory), and our adoption of JENDL-4.0 evaluations for isotopes of Cm, Bk, Cf, Es, Fm, and some other minor actinides; (7) Fission energy release evaluations; (8) Fission product yield advances for fission-spectrum neutrons and 14 MeV neutrons incident on 239Pu; and (9) A new decay data sublibrary. Integral validation testing of the ENDF/B-VII.1 library is provided for a variety of quantities: For nuclear criticality, the VII.1 library maintains the generally-good performance seen for VII.0 for a wide range of MCNP simulations of criticality benchmarks, with improved performance coming from new structural material evaluations, especially for Ti, Mn, Cr, Zr and W. For Be we see some improvements although the fast assembly data appear to be mutually inconsistent. Actinide cross section updates are also assessed through comparisons of fission and capture reaction rate measurements in critical assemblies and fast reactors, and improvements are evident. Maxwellian-averaged capture cross sections at 30 keV are also provided for astrophysics applications. We describe the cross section evaluations that have been updated for ENDF/B-VII.1 and the measured data and calculations that motivated the changes, and therefore this paper augments the ENDF/B-VII.0 publication [M. B. Chadwick, P. Obložinský, M. Herman, N. M. Greene, R. D. McKnight, D. L. Smith, P. G. Young, R. E. MacFarlane, G. M. Hale, S. C. Frankle, A. C. Kahler, T. Kawano, R. C. Little, D. G. Madland, P. Moller, R. D. Mosteller, P. R. Page, P. Talou, H. Trellue, M. C. White, W. B. Wilson, R. Arcilla, C. L. Dunford, S. F. Mughabghab, B. Pritychenko, D. Rochman, A. A. Sonzogni, C. R. Lubitz, T. H. Trumbull, J. P. Weinman, D. A. Br, D. E. Cullen, D. P. Heinrichs, D. P. McNabb, H. Derrien, M. E. Dunn, N. M. Larson, L. C. Leal, A. D. Carlson, R. C. Block, J. B. Briggs, E. T. Cheng, H. C. Huria, M. L. Zerkle, K. S. Kozier, A. Courcelle, V. Pronyaev, and S. C. van der Marck, “ENDF/B-VII.0: Next Generation Evaluated Nuclear Data Library for Nuclear Science and Technology,” Nuclear Data Sheets 107, 2931 (2006)].},
	number = {12},
	journal = {Nuclear Data Sheets},
	author = {Chadwick, M. B. and Herman, M. and Obložinský, P. and Dunn, M. E. and Danon, Y. and Kahler, A. C. and Smith, D. L. and Pritychenko, B. and Arbanas, G. and Arcilla, R. and Brewer, R. and Brown, D. A. and Capote, R. and Carlson, A. D. and Cho, Y. S. and Derrien, H. and Guber, K. and Hale, G. M. and Hoblit, S. and Holloway, S. and Johnson, T. D. and Kawano, T. and Kiedrowski, B. C. and Kim, H. and Kunieda, S. and Larson, N. M. and Leal, L. and Lestone, J. P. and Little, R. C. and McCutchan, E. A. and MacFarlane, R. E. and MacInnes, M. and Mattoon, C. M. and McKnight, R. D. and Mughabghab, S. F. and Nobre, G. P. A. and Palmiotti, G. and Palumbo, A. and Pigni, M. T. and Pronyaev, V. G. and Sayer, R. O. and Sonzogni, A. A. and Summers, N. C. and Talou, P. and Thompson, I. J. and Trkov, A. and Vogt, R. L. and van der Marck, S. C. and Wallner, A. and White, M. C. and Wiarda, D. and Young, P. G.},
	month = dec,
	year = {2011},
	pages = {2887--2996},
	file = {ScienceDirect Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\6KGMGM5A\\S009037521100113X.html:text/html},
}

@article{ignatiev_molten_2014,
	title = {Molten salt actinide recycler and transforming system without and with {Th}-{U} support: {Fuel} cycle flexibility and key material properties},
	volume = {64},
	issn = {0306-4549},
	shorttitle = {Molten salt actinide recycler and transforming system without and with {Th}–{U} support},
	doi = {10.1016/j.anucene.2013.09.004},
	abstract = {A study is under progress to examine the feasibility of MOlten Salt Actinide Recycler and Transforming (MOSART) system without and with U–Th support fuelled with different compositions of transuranic elements (TRU) trifluorides from spent LWR fuel. New design options with homogeneous core and fuel salt with high enough solubility for transuranic elements trifluorides are being examined because of new goals. The paper has the main objective of presenting the fuel cycle flexibility of the MOSART system while accounting technical constrains and experimental data received in this study. A brief description is given of the experimental results on key physical and chemical properties of fuel salt and combined materials compatibility to satisfy MOSART system requirements.},
	number = {Supplement C},
	journal = {Annals of Nuclear Energy},
	author = {Ignatiev, V. and Feynberg, O. and Gnidoi, I. and Merzlyakov, A. and Surenkov, A. and Uglov, V. and Zagnitko, A. and Subbotin, V. and Sannikov, I. and Toropov, A. and Afonichkin, V. and Bovet, A. and Khokhlov, V. and Shishkin, V. and Kormilitsyn, M. and Lizin, A. and Osipenko, A.},
	month = feb,
	year = {2014},
	keywords = {Combined materials compatibility, Core neutronic performance, Fuel cycle flexibility, Molten salt actinide recycler and transforming system, Physical and chemical properties, Salt chemistry control},
	pages = {408--420},
	file = {Ignatiev et al. - 2014 - Molten salt actinide recycler and transforming sys.pdf:C\:\\Users\\Sun Myung\\Zotero\\storage\\7CEI4FC7\\Ignatiev et al. - 2014 - Molten salt actinide recycler and transforming sys.pdf:application/pdf},
}

@inproceedings{ignatiev_progress_2007,
	title = {Progress in development of {Li},{Be},{Na}/{F} molten salt actinide recycler and transmuter concept},
	language = {English},
	booktitle = {Proceedings of {ICAPP} 2007},
	author = {Ignatiev, V. and Feynberg, O. and Gnidoi, I. and Merzlyakov, A. and Smirnov, V. and Surenkov, A. and Tretiakov, I. and Zakirov, R. and Afonichkin, V. and Bovet, A. and Subbotin, V. and Panov, A. and Toropov, A. and Zherebtsov, A.},
	month = may,
	year = {2007},
	file = {Ignatiev et al. - 2007 - Progress in development of Li,Be,NaF molten salt .pdf:C\:\\Users\\Sun Myung\\Zotero\\storage\\4DC6FFQC\\Ignatiev et al. - 2007 - Progress in development of Li,Be,NaF molten salt .pdf:application/pdf;Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\KLWZNA6E\\search.html:text/html},
}

@article{macpherson_molten_1985,
	title = {The {Molten} {Salt} {Reactor} {Adventure}},
	volume = {90},
	issn = {ISSN 0029-5639},
	doi = {10.13182/NSE90-374},
	abstract = {A personal history of the development of molten salt reactors in the United States is presented. The initial goal was an aircraft propulsion reactor, and a molten fluoride-fueled Aircraft Reactor Experiment was operated at Oak Ridge National Laboratory in 1954. In 1956, the objective shifted to civilian nuclear power, and reactor concepts were developed using a circulating UF4-ThF4 fuel, graphite moderator, and Hastelloy N pressure boundary. The program culminated in the successful operation of the Molten Salt Reactor Experiment in 1965 to 1969. By then the Atomic Energy Commission’s goals had shifted to breeder development; the molten salt program supported on-site reprocessing development and study of various reactor arrangements that had potential to breed. Some commercial and foreign interest contributed to the program which, however, was terminated by the government in 1976. The current status of the technology and prospects for revived interest are summarized.},
	language = {en},
	number = {4},
	journal = {Nuclear Science and Engineering},
	author = {MacPherson, H. G.},
	month = aug,
	year = {1985},
	keywords = {unread},
	pages = {374--380},
	file = {[PDF] from moltensalt.org:C\:\\Users\\Sun Myung\\Zotero\\storage\\PSZWM3J8\\MacPherson - 1985 - The Molten Salt Reactor Adventure.pdf:application/pdf;nse_v90_n4_pp374-380.pdf:C\:\\Users\\Sun Myung\\Zotero\\storage\\RGPUF2H9\\nse_v90_n4_pp374-380.pdf:application/pdf;Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\7EEL6Q93\\search.html:text/html;Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\NAUGQK6J\\NSE90-374.html:text/html},
}

@techreport{fanning_sas4a/sassys-1_2017,
	title = {The {SAS4A}/{SASSYS}-1 {Safety} {Analysis} {Code} {System}, {Version} 5},
	url = {http://www.osti.gov/servlets/purl/1352187/},
	abstract = {The SAS4A/SASSYS-1 computer code is developed by Argonne National Laboratory for thermal, hydraulic, and neutronic analysis of power and flow transients in liquidmetal- cooled nuclear reactors (LMRs). SAS4A was developed to analyze severe core disruption accidents with coolant boiling and fuel melting and relocation, initiated by a very low probability coincidence of an accident precursor and failure of one or more safety systems. SASSYS-1, originally developed to address loss-of-decay-heat-removal accidents, has evolved into a tool for margin assessment in design basis accident (DBA) analysis and for consequence assessment in beyond-design-basis accident (BDBA) analysis. SAS4A contains detailed, mechanistic models of transient thermal, hydraulic, neutronic, and mechanical phenomena to describe the response of the reactor core, its coolant, fuel elements, and structural members to accident conditions. The core channel models in SAS4A provide the capability to analyze the initial phase of core disruptive accidents, through coolant heat-up and boiling, fuel element failure, and fuel melting and relocation. Originally developed to analyze oxide fuel clad with stainless steel, the models in SAS4A have been extended and specialized to metallic fuel with advanced alloy cladding. SASSYS-1 provides the capability to perform a detailed thermal/hydraulic simulation of the primary and secondary sodium coolant circuits and the balance-ofplant steam/water circuit. These sodium and steam circuit models include component models for heat exchangers, pumps, valves, turbines, and condensers, and thermal/hydraulic models of pipes and plena. SASSYS-1 also contains a plant protection and control system modeling capability, which provides digital representations of reactor, pump, and valve controllers and their response to input signal changes.},
	language = {en},
	number = {ANL/NE-16/19, 1352187},
	author = {Fanning, T. H. and Brunett, A. J. and Sumner, T.},
	month = jan,
	year = {2017},
}

@article{haubenreich_experience_1970,
	title = {Experience with the {Molten}-{Salt} {Reactor} {Experiment}},
	volume = {8},
	issn = {00295450},
	doi = {10.13182/NT8-2-118},
	number = {2},
	journal = {Nuclear Technology},
	author = {Haubenreich, Paul N. and Engel, J. R.},
	month = feb,
	year = {1970},
	pages = {118--136},
	file = {Haubenreich_Engel_MSREexperience.pdf:C\:\\Users\\Sun Myung\\Zotero\\storage\\6EIEWPDA\\Haubenreich_Engel_MSREexperience.pdf:application/pdf},
}

@article{bettis_design_1970,
	title = {The {Design} and {Performance} {Features} of a {Single}-{Fluid} {Molten}-{Salt} {Breeder} {Reactor}},
	volume = {8},
	issn = {0550-3043},
	doi = {10.13182/NT70-A28625},
	abstract = {A conceptual design has been made of a single-fluid 1000 MW(e) Molten-Salt Breeder Reactor (MSBR) power station based on the capabilities of present technology. The reactor vessel is 22ft in diameter × 20 ft high and is fabricated of Hastelloy-N with graphite as the moderator and reflector. The fuel is 233U carried in a LiF-BeF2-ThF4 mixture which is molten above 930°F. Thorium is converted to 233U in excess of fissile burnup so that bred material is a plant product. The estimated fuel yield is 3.3\% per year.The estimated construction cost of the station is comparable to PWR total construction costs. The power production cost, including fuel-cycle and graphite replacement costs, with private utility financing, is estimated to be 0.5 to 1 mill/kWh less than that for present-day light-water reactors, largely due to the low fuel-cycle cost and high plant thermal efficiency.After engineering development of the fuel purification processes and large-scale components, a practical plant similar to the one described here appears to be feasible.},
	number = {2},
	journal = {Nuclear Applications and Technology},
	author = {Bettis, E. S. and Robertson, Roy C.},
	month = feb,
	year = {1970},
	pages = {190--207},
	file = {Full Text PDF:C\:\\Users\\Sun Myung\\Zotero\\storage\\WG79G7B7\\Bettis and Robertson - 1970 - The Design and Performance Features of a Single-Fl.pdf:application/pdf},
}

@article{leppanen_calculation_2014,
	title = {Calculation of effective point kinetics parameters in the {Serpent} 2 {Monte} {Carlo} code},
	volume = {65},
	issn = {0306-4549},
	doi = {10.1016/j.anucene.2013.10.032},
	abstract = {This paper presents the methodology developed for the Serpent 2 Monte Carlo code for the calculation of adjoint-weighted reactor point kinetics parameters: effective generation time and delayed neutron fractions. The calculation routines were implemented at the Politecnico di Milano, and they are based on the iterated fission probability (IFP) method. The developed methodology is mainly intended for the modeling of small research reactor cores, and the results are validated by comparison to experimental data and MCNP5 calculations in 31 critical configurations.},
	journal = {Annals of Nuclear Energy},
	author = {Leppänen, Jaakko and Aufiero, Manuele and Fridman, Emil and Rachamin, Reuven and van der Marck, Steven},
	month = mar,
	year = {2014},
	keywords = {Monte Carlo, Adjoint-weighted time constants, Effective delayed neutron fraction, Effective generation time},
	pages = {272--279},
	file = {ScienceDirect Full Text PDF:C\:\\Users\\Sun Myung\\Zotero\\storage\\9I9C9CI7\\Leppänen et al. - 2014 - Calculation of effective point kinetics parameters.pdf:application/pdf;ScienceDirect Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\BM8IV98C\\S0306454913005628.html:text/html},
}

@article{rouch_preliminary_2014,
	title = {Preliminary thermal–hydraulic core design of the {Molten} {Salt} {Fast} {Reactor} ({MSFR})},
	volume = {64},
	issn = {0306-4549},
	doi = {10.1016/j.anucene.2013.09.012},
	abstract = {A thermal–hydraulics study of the core of the Molten Salt Fast Reactor (MSFR) is presented. The numerical simulations were carried-out using a Computation Fluid Dynamic code. The main objectives of the thermal–hydraulics studies are to design the core cavity walls in order to increase the overall flow mixing and to reduce the temperature peaking factors in the salt and on the core walls. The results of the CFD simulations show that for the chosen core design acceptable temperature distributions can be obtained by using a curved core cavity shape, inlets and outlets. The hot spot temperature is less than 10 °C above the average core outlet temperature and is located in the centre of the top wall of the core. The results show also a moderate level of sensitivity to the working point.},
	journal = {Annals of Nuclear Energy},
	author = {Rouch, H. and Geoffroy, O. and Rubiolo, P. and Laureau, A. and Brovchenko, M. and Heuer, D. and Merle-Lucotte, E.},
	month = feb,
	year = {2014},
	keywords = {CFD, Core cavity, Fuel salt temperature, MSFR, Thermal–hydraulics design},
	pages = {449--456},
	file = {ScienceDirect Full Text PDF:C\:\\Users\\Sun Myung\\Zotero\\storage\\SX2W3QA5\\Rouch et al. - 2014 - Preliminary thermal–hydraulic core design of the M.pdf:application/pdf;ScienceDirect Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\8GBUHZAG\\S0306454913004829.html:text/html},
}

@article{jones_prediction_1972,
	title = {The prediction of laminarization with a two-equation model of turbulence},
	volume = {15},
	issn = {0017-9310},
	doi = {10.1016/0017-9310(72)90076-2},
	abstract = {The paper presents a new model of turbulence in which the local turbulent viscosity is determined from the solution of transport equations for the turbulence kinetic energy and the energy dissipation rate. The major component of this work has been the provision of a suitable form of the model for regions where the turbulence Reynolds number is low. The model has been applied to the prediction of wall boundary-layer flows in which streamwise accelerations are so severe that the boundary layer reverts partially towards laminar. In all cases, the predicted hydrodynamic and heat-transfer development of the boundary layers is in close agreement with the measured behaviour. L'article présente un nouveau modèle de turbulence dans lequel la viscosité turbulente locale est déterminée à partir de la solution des équations de transport pour l'énergie cinétique de turbulence et la vitesse de dissipation d'énergie. La majeure partie de ce travail a été l'élaboration d'une forme adéquate du modèle pour des régions où le nombre de Reynolds de turbulence est faible. Le modèle a été appliqué à la prédiction des écoulements à couche limite à la paroi dans lesquels des accélérations longitudinales sont si fortes que la couche limite redevient partiellement laminaire. Dans tous les cas, le développement hydrodynamique et thermique prévu des couches limites est en parfait accord avec le comportement observé. Die Arbeit behandelt ein neuse Turbulenzmodell, bei dem die örtliche turbulente Zähigkeit aus der Lösung der Transportgleichungen für die kinetische Energie der Turbulenz und der Energiedissipation bestimmt wird. Der Hauptteil dieser Arbeit bestand darin, eine passende Form des Modells für Bereiche zu schaffen, in denen die Reynoldszahl der Turbulenz niedrig ist. Das Modell ist auf die Bestimmung von Wandgrenzschichtströmungen angewandt worden, bein denen so starke Beschleunigungen in Strömungsrichtung auftreten, dass die Grenzschicht teilweise in den laminaren Bereicht umschlägt. In allen Fällen ist die berechnete Entwicklung der hydrodynamischen und thermischen Grenzschicht in guter Übereinstimmung mit dem gemessenen Verhalten.},
	number = {2},
	journal = {International Journal of Heat and Mass Transfer},
	author = {Jones, W. P and Launder, B. E},
	month = feb,
	year = {1972},
	pages = {301--314},
	file = {Jones_Launder_Prediction_of_laminarization_with_two_equation_model_of_turbulence.pdf:C\:\\Users\\Sun Myung\\Zotero\\storage\\94WZ934Z\\Jones_Launder_Prediction_of_laminarization_with_two_equation_model_of_turbulence.pdf:application/pdf;ScienceDirect Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\B8E6WRBI\\0017931072900762.html:text/html},
}

@book{bell_nuclear_1970,
	address = {New York},
	title = {Nuclear {Reactor} {Theory}},
	language = {English},
	publisher = {Van Nostrand Reinhold Company},
	author = {Bell, George I. and Glasstone, Samuel},
	year = {1970},
	file = {nuclear-reactor-theory.pdf:C\:\\Users\\Sun Myung\\Zotero\\storage\\JP52JB8B\\nuclear-reactor-theory.pdf:application/pdf},
}

@book{lamarsh_introduction_1975,
	title = {Introduction to nuclear engineering},
	volume = {3},
	publisher = {Addison-Wesley Massachusetts},
	author = {Lamarsh, John R. and Baratta, Anthony John},
	year = {1975},
	file = {Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\D8TEQXIE\\introduction-to-nuclear-engineering-by-john-r-lamarsh.pdf:application/pdf},
}

@book{stacey_nuclear_2007,
	title = {Nuclear reactor physics},
	publisher = {Wiley. com},
	author = {Stacey, Weston M.},
	year = {2007},
	file = {Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\BQKTUZJ8\\books.html:text/html;Stacey_Nuclear_2001.pdf:C\:\\Users\\Sun Myung\\Zotero\\storage\\E3ADTKDP\\Stacey_Nuclear_2001.pdf:application/pdf},
}

@article{cammi_multi-physics_2011,
	title = {A multi-physics modelling approach to the dynamics of {Molten} {Salt} {Reactors}},
	volume = {38},
	issn = {0306-4549},
	doi = {10.1016/j.anucene.2011.01.037},
	abstract = {This paper presents a multi-physics modelling (MPM) approach developed for the study of the dynamics of the Molten Salt Reactor (MSR), which has been reconsidered as one of the future nuclear power plants in the framework of the Generation IV International Forum for its several potentialities. The proposed multi-physics modelling is aimed at the description of the coupling between heat transfer, fluid dynamics and neutronics characteristics in a typical MSR core channel, taking into account the spatial effects of the most relevant physical quantities. In particular, as far as molten salt thermo-hydrodynamics is concerned, Navier–Stokes equations are used with the turbulence treatment according to the RANS (Reynolds Averaged Navier–Stokes) scheme, while the heat transfer is taken into account through the energy balance equations for the fuel salt and the graphite. As far as neutronics is concerned, the two-group diffusion theory is adopted, where the group constants (computed by means of the neutron transport code NEWT of SCALE 5.1) are included into the model in order to describe the neutron flux and the delayed neutron precursor distributions, the system time constants, and the temperature feedback effects of both graphite and fuel salt. The developed MPM approach is implemented in the unified simulation environment offered by COMSOL Multiphysics®, and is applied to study the behaviour of the system in steady-state conditions and under several transients (i.e., reactivity insertion due to control rod movements, fuel mass flow rate variations due to the change of the pump working conditions, presence of periodic perturbations), pointing out some advantages offered with respect to the conventional approaches employed in literature for the MSRs.},
	number = {6},
	journal = {Annals of Nuclear Energy},
	author = {Cammi, Antonio and Di Marcello, Valentino and Luzzi, Lelio and Memoli, Vito and Ricotti, Marco Enrico},
	month = jun,
	year = {2011},
	keywords = {MSR, unread, read, atws, molten salt reactor, Multi-physics modelling, Reactor dynamics, Thermo-hydrodynamics},
	pages = {1356--1372},
	file = {A_multi-physics_modelling_approach_to_the_dynamics_of_Molten_Sal.mobi:C\:\\Users\\Sun Myung\\Zotero\\storage\\JWIMI3QI\\A_multi-physics_modelling_approach_to_the_dynamics_of_Molten_Sal.mobi:application/octet-stream;cammi_multi-physics_2011.pdf:C\:\\Users\\Sun Myung\\Zotero\\storage\\AHXUPQ4A\\cammi_multi-physics_2011.pdf:application/pdf;ScienceDirect Full Text PDF:C\:\\Users\\Sun Myung\\Zotero\\storage\\6FQKN2CJ\\Cammi et al. - 2011 - A multi-physics modelling approach to the dynamics.pdf:application/pdf;ScienceDirect Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\63AWQD57\\S0306454911000582.html:text/html},
}

@article{moir_recommendations_2008,
	title = {Recommendations for a restart of molten salt reactor development},
	volume = {49},
	issn = {0196-8904},
	shorttitle = {{ICENES}’2007, 13th {International} {Conference} on {Emerging} {Nuclear} {Energy} {Systems}, {June} 3–8, 2007, İstanbul, {Turkiye}},
	doi = {10.1016/j.enconman.2007.07.047},
	abstract = {The concept of the molten salt reactor (MSR) refuses to go away. The Generation-IV process lists the MSR as one of the six concepts to be considered for extending fuel resources. Good fuel utilization and good economics are required to meet the often-cited goal of 10 TWe globally and 1 TWe for the US by non-carbon energy sources in this century by nuclear fission. Strong incentives for the molten salt reactor design are its good fuel utilization, good economics, amazing fuel flexibility and promised large benefits. It can:• use thorium or uranium; • be designed with lots of graphite to have a fairly thermal neutron spectrum or without graphite moderator to have an epithermal neutron spectrum; • fission uranium isotopes and plutonium isotopes; • produces less long-lived wastes than today’s reactors by a factor of 10–100; • operate with non-weapon grade fissile fuel, or in suitable sites it can operate with enrichment between reactor-grade and weapon grade fissile fuel; • be a breeder or near breeder; • operate at temperature \>1100 °C if carbon composites are successfully developed. Enhancing 232U content in the uranium to over 500 ppm makes the fuel undesirable for weapons, but it should not detract from its economic use in liquid fuel reactors: a big advantage in nonproliferation. Economics of the MSR are enhanced by operating at low pressure and high temperature and may even lead to the preferred route to hydrogen production. The cost of the electricity produced from low enriched fuel averaged over the life of the entire process, has been predicted to be about 10\% lower than that from LWRs, and 20\% lower for high-enriched fuel, with uncertainties of about 10\%. The development cost has been estimated at about 1 B\$ (e.g., a 100 M\$/year base program for 10 years) not including construction of a series of reactors leading up to the deployment of multiple commercial units at an assumed cost of 9 B\$ (450 M\$/year over 20 years). A benefit of liquid fuel is that smaller power reactors can faithfully test features of larger reactors, thereby reducing the number of steps to commercial deployment. Assuming electricity is worth \$ 50 per MWe h, then 50 years of 10 TWe power level would be worth 200 trillion dollars. If the MSR could be developed and proven for 10 B\$ and would save 10\% over its alternative, the total savings over 50 years would be 20 trillion dollars: a good return on investment even considering discounted future savings. The incentives for the molten salt reactor are so strong and its relevance to our energy policy and national security are so compelling that one asks, “Why has the reactor not already been developed?”},
	number = {7},
	journal = {Energy Conversion and Management},
	author = {Moir, R.W.},
	month = jul,
	year = {2008},
	keywords = {Economics, read, Deployment scenario, nonproliferation, Start-up fuel},
	pages = {1849--1858},
	file = {Recommendations_for_a_restart_of_molten_salt_reactor_development.mobi:C\:\\Users\\Sun Myung\\Zotero\\storage\\JEIKHSR9\\Recommendations_for_a_restart_of_molten_salt_reactor_development.mobi:application/octet-stream;recommendations.pdf:C\:\\Users\\Sun Myung\\Zotero\\storage\\NTPNR55D\\recommendations.pdf:application/pdf;ScienceDirect Full Text PDF:C\:\\Users\\Sun Myung\\Zotero\\storage\\F9DQRBDV\\Moir - 2008 - Recommendations for a restart of molten salt react.pdf:application/pdf;ScienceDirect Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\3BWGHW7P\\S0196890407004268.html:text/html},
}

@techreport{haubenreich_msre_1964,
	title = {Msre {Design} and {Operations} {Report}. {Part} {Iii}. {Nuclear} {Analysis}},
	url = {http://www.osti.gov/scitech/biblio/4114686},
	language = {English},
	number = {ORNL-TM-730},
	institution = {Oak Ridge National Lab., Tenn.},
	author = {Haubenreich, P. N. and Engel, J. R. and Prince, B. E. and Claiborne, H. C.},
	month = feb,
	year = {1964},
	keywords = {neutrons, enrichment, graphite, Criticality, absorption, accidents, adsorption, alpha particles, beryllium, buildings, concretes, configuration, control elements, control systems, coolant loops, degassing, delayed neutrons, density, distribution, equations, excursions, failures, fertile materials, Fission products, fissionable materials, Fuels, fused salts, gases, graphite moderator, heat transfer, high temperature, liquid flow, lithium, losses, low temperature, mass, materials testing, moderators, monitoring, msre, multiplication factors, neutron flux, neutron sources, nuclear reactions, operation, personnel, planning, poisoning, power plant, reactor technology},
	file = {Full Text PDF:C\:\\Users\\Sun Myung\\Zotero\\storage\\DMIDAWSC\\Haubenreich et al. - 1964 - Msre Design and Operations Report. Part Iii. Nucle.pdf:application/pdf;Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\64Z7NT6C\\4114686.html:text/html},
}

@article{cammi_dimensional_2012,
	series = {Selected and expanded papers from {International} {Conference} {Nuclear} {Energy} for {New} {Europe} 2010, {Portoro}?, {Slovenia}, {September} 6-9, 2010},
	title = {Dimensional effects in the modelling of {MSR} dynamics: {Moving} on from simplified schemes of analysis to a multi-physics modelling approach},
	volume = {246},
	issn = {0029-5493},
	shorttitle = {Dimensional effects in the modelling of {MSR} dynamics},
	doi = {10.1016/j.nucengdes.2011.08.002},
	abstract = {The MSR (Molten Salt Reactor) is one of the six innovative concepts of nuclear reactors envisaged by the GIF-IV (Generation IV International Forum) initiative for the long term evolution of the nuclear technology, in the direction of a more sustainable, safe, proliferation resistant, and economic power generation. The MSR is characterised by a complex and highly non-linear behaviour, which requires a careful investigation, as a consequence of some unusual features like the presence of a fluid fuel and the drift of delayed neutron precursors (DNP) along the primary circuit. In this paper, the MSR primary circuit dynamics is analysed with reference to the MSRE (Molten Salt Reactor Experiment), due to the availability of both a detailed design and experimental data. Numerical models featured by increasing complexity are presented. In particular, a zero-dimensional model is developed, and two other models introducing a one-dimensional discretization for the DNP drift and/or the heat convection are also elaborated. These simplified models are then compared with a more complex model, obtained according to an innovative Multi-Physics Modelling (MPM) approach to the dynamic analysis, where the partial differential equations governing the different phenomena in a core channel are solved in a two-dimensional domain, within the same computational environment. A one-dimensional closure of the primary circuit is also provided. The MPM approach gives a unique insight into the influence of local effects on the overall dynamic behaviour of the reactor, while the variety of developed models allows a systematic investigation about the dimensional effects in the modelling of MSRs. This work represents a starting point in the set-up of a Multi-Physics (MP) simulation tool, suitable for calculations with different degrees of accuracy and physical complexity, and paves the way towards the development of MP models capable of a point-by-point coupling of all the phenomena characterising the MSRs, and the nuclear reactors in general.},
	journal = {Nuclear Engineering and Design},
	author = {Cammi, Antonio and Fiorina, Carlo and Guerrieri, Claudia and Luzzi, Lelio},
	month = may,
	year = {2012},
	pages = {12--26},
	file = {ScienceDirect Full Text PDF:C\:\\Users\\Sun Myung\\Zotero\\storage\\923SXTPA\\Cammi et al. - 2012 - Dimensional effects in the modelling of MSR dynami.pdf:application/pdf;ScienceDirect Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\9M9IPG5I\\S0029549311006273.html:text/html},
}

@article{leppanen_use_2017,
	title = {On the use of delta-tracking and the collision flux estimator in the {Serpent} 2 {Monte} {Carlo} particle transport code},
	volume = {105},
	issn = {0306-4549},
	doi = {10.1016/j.anucene.2017.03.006},
	abstract = {The Serpent Monte Carlo code was originally developed for the purpose of spatial homogenization and other computational problems encountered in the field of reactor physics. However, during the past few years the implementation of new methodologies has allowed expanding the scope of applications to new fields, including radiation transport and fusion neutronics. These applications pose new challenges for the tracking routines and result estimators, originally developed for a very specific task. The purpose of this paper is to explain how the basic collision estimator based cell flux tally in Serpent 2 is implemented, and how it is applied for calculating integral reaction rates. The methodology and its limitations are demonstrated by an example, in which the tally is applied for calculating collision rates in a problem with very low physical collision density. It is concluded that Serpent has a lot of potential to expand its scope of applications beyond reactor physics, but in order to be applied for such problems it is important that the code users understand the underlying methods and their limitations.},
	journal = {Annals of Nuclear Energy},
	author = {Leppänen, Jaakko},
	month = jul,
	year = {2017},
	keywords = {Monte Carlo, Collision flux estimator, Delta-tracking, Transport simulation},
	pages = {161--167},
	file = {ScienceDirect Full Text PDF:C\:\\Users\\Sun Myung\\Zotero\\storage\\TJF3RXPK\\Leppänen - 2017 - On the use of delta-tracking and the collision flu.pdf:application/pdf;ScienceDirect Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\GTMPG3QE\\S0306454916311367.html:text/html},
}

@misc{lindsay_moltres_2017,
	address = {University of Illinois at Urbana-Champaign},
	title = {Moltres, software for simulating {Molten} {Salt} {Reactors}},
	shorttitle = {Moltres},
	url = {https://github.com/arfc/moltres},
	abstract = {arfc/moltres: Repository for Moltres, a code for simulating Molten Salt Reactors},
	urldate = {2017-02-24},
	author = {Lindsay, Alexander},
	year = {2017},
	file = {arfc/moltres\: Repository for Moltres, a code for simulating Molten Salt Reactors:C\:\\Users\\Sun Myung\\Zotero\\storage\\WGEZI8WJ\\moltres.html:text/html},
}

@book{prince_zero-power_1968,
	title = {{ZERO}-{POWER} {PHYSICS} {EXPERIMENTS} {ON} {THE} {MOLTEN}-{SALT} {REACTOR} {EXPERIMENT}.},
	author = {Prince, B.E. and Ball, S.J. and Engel, J.R. and Haubenreich, P.N. and Kerlin, T.W.},
	month = jan,
	year = {1968},
	file = {ORNL-4233.pdf:C\:\\Users\\Sun Myung\\Zotero\\storage\\7UTAGB2T\\ORNL-4233.pdf:application/pdf},
}

@misc{github_build_2017,
	title = {Build software better, together},
	url = {https://github.com},
	abstract = {GitHub is where people build software. More than 21 million people use GitHub to discover, fork, and contribute to over 58 million projects.},
	urldate = {2017-05-11},
	journal = {GitHub},
	author = {{GitHub}},
	year = {2017},
	file = {Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\9UB445Q8\\github.com.html:text/html},
}

@techreport{gif_generation_2008,
	title = {Generation {IV} {International} {Forum} 2008 {Annual} {Report}},
	institution = {Generation IV International Forum},
	author = {{GIF}},
	year = {2008},
}

@book{duderstadt_nuclear_1976,
	address = {New York},
	edition = {1 edition},
	title = {Nuclear {Reactor} {Analysis}},
	isbn = {978-0-471-22363-4},
	abstract = {Classic textbook for an introductory course in nuclear reactor analysis that introduces the nuclear engineering student to the basic scientific principles of nuclear fission chain reactions and lays a foundation for the subsequent application of these principles to the nuclear design and analysis of reactor cores. This text introduces the student to the fundamental principles governing nuclear fission chain reactions in a manner that renders the transition to practical nuclear reactor design methods most natural. The authors stress throughout the very close interplay between the nuclear analysis of a reactor core and those nonnuclear aspects of core analysis, such as thermal-hydraulics or materials studies, which play a major role in determining a reactor design.},
	language = {English},
	publisher = {Wiley},
	author = {Duderstadt, James J. and Hamilton, Louis J.},
	month = jan,
	year = {1976},
}

@article{li_transient_2015,
	title = {Transient analyses for a molten salt fast reactor with optimized core geometry},
	volume = {292},
	issn = {0029-5493},
	doi = {10.1016/j.nucengdes.2015.06.011},
	abstract = {Molten salt reactors (MSRs) have encountered a marked resurgence of interest over the past decades, highlighted by their inclusion as one of the six candidate reactors of the Generation IV advanced nuclear power systems. The present work is carried out in the framework of the European FP-7 project EVOL (Evaluation and Viability Of Liquid fuel fast reactor system). One of the project tasks is to report on safety analyses: calculations of reactor transients using various numerical codes for the molten salt fast reactor (MSFR) under different boundary conditions, assumptions, and for different selected scenarios. Based on the original reference core geometry, an optimized geometry was proposed by Rouch et al. (2014. Ann. Nucl. Energy 64, 449) on thermal-hydraulic design aspects to avoid a recirculation zone near the blanket which accumulates heat and very high temperature exceeding the salt boiling point. Using both fully neutronics thermal-hydraulic coupled codes (SIMMER and COUPLE), we also re-confirm the efforts step by step toward a core geometry without the recirculation zone in particular as concerns the modifications of the core geometrical shape. Different transients namely Unprotected Loss of Heat Sink (ULOHS), Unprotected Loss of Flow (ULOF), Unprotected Transient Over Power (UTOP), Fuel Salt Over Cooling (FSOC) are intensively investigated and discussed with the optimized core geometry. It is demonstrated that due to inherent negative feedbacks, an MSFR plant has a high safety potential.},
	journal = {Nuclear Engineering and Design},
	author = {Li, R. and Wang, S. and Rineiski, A. and Zhang, D. and Merle-Lucotte, E.},
	month = oct,
	year = {2015},
	keywords = {L. safety and risk analysis},
	pages = {164--176},
	file = {ScienceDirect Full Text PDF:C\:\\Users\\Sun Myung\\Zotero\\storage\\TFJT7C45\\Li et al. - 2015 - Transient analyses for a molten salt fast reactor .pdf:application/pdf;ScienceDirect Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\59GD97VN\\S0029549315002617.html:text/html},
}

@article{nagy_steady-state_2014,
	title = {Steady-state and dynamic behavior of a moderated molten salt reactor},
	volume = {64},
	issn = {0306-4549},
	doi = {10.1016/j.anucene.2013.08.009},
	abstract = {The moderated Molten Salt Reactor (MSR) is an attractive breeder reactor. However, the temperature feedback coefficient of such a system can be positive due to the contribution of the moderator, an effect that can only be avoided with special measures. A previous study (Nagy et al., 2010) aimed to find a core design that is a breeder and has negative overall temperature feedback coefficient. In this paper, a coupled calculation scheme, which includes the reactor physics, heat transfer and fluid dynamics calculations is introduced. It is used both for steady-state and for dynamic calculations to evaluate the safety of the core design which was selected from the results of the previous study. The calculated feedback coefficients on the salt and graphite temperatures, power and uranium concentration prove that the core design derived in the previous optimization study is safe because the temperature feedback coefficient of the core and of the power is sufficiently negative. Transient calculations are performed to show the inherent safety of the reactor in case of reactivity insertion. As it is shown, the response of the reactor to these transients is initially dominated by the strong negative feedback of the salt. In all the presented transients, the reactor power stabilizes and the temperature of the salt never approaches its boiling point.},
	journal = {Annals of Nuclear Energy},
	author = {Nagy, K. and Lathouwers, D. and T’Joen, C. G. A. and Kloosterman, J. L. and van der Hagen, T. H. J. J.},
	month = feb,
	year = {2014},
	keywords = {Coupled calculations, Transient calculations},
	pages = {365--379},
	file = {ScienceDirect Full Text PDF:C\:\\Users\\Sun Myung\\Zotero\\storage\\92VTQVHJ\\Nagy et al. - 2014 - Steady-state and dynamic behavior of a moderated m.pdf:application/pdf;ScienceDirect Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\EAGZXEWN\\S0306454913004179.html:text/html},
}

@article{dehart_reactor_2011,
	title = {Reactor {Physics} {Methods} and {Analysis} {Capabilities} in {SCALE}},
	volume = {174},
	doi = {dx.doi.org/10.13182/NT174-196},
	number = {2},
	journal = {Nuclear Technology},
	author = {DeHart, Mark D. and Bowman, Stephen M.},
	month = may,
	year = {2011},
	pages = {196--213},
	file = {Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\BHMIBIJ9\\epubs.ans.org.html:text/html},
}

@incollection{ho_molten_2013,
	address = {Badajoz, Spain},
	edition = {2013},
	series = {Energy {Book} {Series}},
	title = {Molten salt reactors},
	isbn = {978-84-939843-7-3},
	number = {1},
	booktitle = {Materials and processes for energy: communicating current research and technological developments},
	publisher = {Formatex Research Center},
	author = {Ho, M. K. M. and Yeoh, G. H. and Braoudakis, G.},
	editor = {Méndez-Vilas, A.},
	year = {2013},
	keywords = {MSR, read, FHR},
	pages = {761--768},
	file = {ho_molten_2013.pdf:C\:\\Users\\Sun Myung\\Zotero\\storage\\NXU753GF\\ho_molten_2013.pdf:application/pdf},
}

@inproceedings{merle-lucotte_launching_2011,
	title = {Launching the thorium fuel cycle with the {Molten} {Salt} {Fast} {Reactor}},
	booktitle = {Proceedings of {ICAPP}},
	author = {Merle-Lucotte, E. and Heuer, D. and Allibert, M. and Brovchenko, M. and Capellan, N. and Ghetta, V.},
	year = {2011},
	pages = {2--5},
	file = {[PDF] researchgate.net:C\:\\Users\\Sun Myung\\Zotero\\storage\\EN5T5AGG\\Merle-Lucotte et al. - 2011 - Launching the thorium fuel cycle with the Molten S.pdf:application/pdf},
}

@article{kamei_recent_2012,
	title = {Recent {Research} of {Thorium} {Molten}-{Salt} {Reactor} from a {Sustainability} {Viewpoint}},
	volume = {4},
	copyright = {http://creativecommons.org/licenses/by/3.0/},
	doi = {10.3390/su4102399},
	abstract = {The most important target of the concept “sustainability” is to achieve fairness between generations. Its expanding interpolation leads to achieve fairness within a generation. Thus, it is necessary to discuss the role of nuclear power from the viewpoint of this definition. The history of nuclear power has been the control of the nuclear fission reaction. Once this is obtained, then the economy of the system is required. On the other hand, it is also necessary to consider the internalization of the external diseconomy to avoid damage to human society caused by the economic activity itself, due to its limited capacity. An extreme example is waste. Thus, reducing radioactive waste resulting from nuclear power is essential. Nuclear non-proliferation must be guaranteed. Moreover, the FUKUSHIMA accident revealed that it is still not enough that human beings control nuclear reaction. Further, the most essential issue for sustaining use of one technology is human resources in manufacturing, operation, policy-making and education. Nuclear power will be able to satisfy the requirements of sustainability only when these subjects are addressed. The author will review recent activities of a thorium molten-salt reactor (MSR) as a cornerstone for a sustainable society and describe its objectives and forecasts.},
	language = {en},
	number = {10},
	journal = {Sustainability},
	author = {Kamei, Takashi},
	month = sep,
	year = {2012},
	keywords = {small modular reactor, externality, molten-salt reactor, rare earth},
	pages = {2399--2418},
	file = {Full Text PDF:C\:\\Users\\Sun Myung\\Zotero\\storage\\Q54RA3YC\\Kamei - 2012 - Recent Research of Thorium Molten-Salt Reactor fro.pdf:application/pdf;Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\AMBQBKSD\\htm.html:text/html},
}

@techreport{transatomic_power_corporation_technical_2016,
	address = {Cambridge, MA, United States},
	type = {White {Paper}},
	title = {Technical {White} {Paper}},
	url = {http://www.transatomicpower.com/wp-content/uploads/2015/04/TAP-White-Paper-v2.1.pdf},
	abstract = {Transatomic Power’s advanced molten salt reactor unlocks clean,
safe, and low-cost nuclear energy. Our revolutionary design
allows us to achieve a high fuel burnup in a compact system,
solve the nuclear industry’s most pressing problems, and clear the
way for advanced nuclear power’s global deployment.},
	language = {English},
	number = {2.1},
	institution = {Transatomic Power Corporation},
	author = {{Transatomic Power Corporation}},
	month = nov,
	year = {2016},
	file = {Transatomic Power Corporation - 2016 - Technical White Paper.pdf:C\:\\Users\\Sun Myung\\Zotero\\storage\\556MUQ2G\\Transatomic Power Corporation - 2016 - Technical White Paper.pdf:application/pdf},
}

@techreport{briggs_molten-salt_1964,
	address = {Oak Ridge, TN, United States},
	type = {Technical {Report} {Archive} and {Image} {Library}},
	title = {Molten-{Salt} {Reactor} {Program} semiannual progress report for period ending {July} 31, 1964},
	url = {https://digital.library.unt.edu/ark:/67531/metadc100304/},
	abstract = {Report issued by the Oak Ridge National Laboratory discussing semiannual progress made by the Molten-Salt Reactor Program. Descriptions of design, construction, and experimental progress is presented. This report includes tables, illustrations, and photographs.},
	number = {ORNL-3708},
	institution = {Oak Ridge National Laboratory},
	author = {Briggs, R. B.},
	year = {1964},
	pages = {397},
}

@incollection{kloosterman_20_2017,
	title = {20 - {Safety} assessment of the molten salt fast reactor ({SAMOFAR})},
	isbn = {978-0-08-101126-3},
	abstract = {This chapter describes the goal, contents, and the consortium of the SAMOFAR project, which is currently the leading project in the European Union in the field of MSR research. It focuses on the safety assessment of the Molten Salt Fast Reactor and will eventually lead to an updated reactor design evaluated with a new integral safety method.},
	booktitle = {Molten {Salt} {Reactors} and {Thorium} {Energy}},
	publisher = {Woodhead Publishing},
	author = {Kloosterman, Jan L.},
	editor = {Dolan, Thomas J.},
	year = {2017},
	keywords = {Molten Salt Fast Reactor, Safety assessment},
	pages = {565--570},
	file = {ScienceDirect Full Text PDF:C\:\\Users\\Sun Myung\\Zotero\\storage\\7LLJFPR3\\Kloosterman - 2017 - 20 - Safety assessment of the molten salt fast rea.pdf:application/pdf;ScienceDirect Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\6DWXAWQD\\B9780081011263000208.html:text/html},
}

@incollection{grape_10_2017,
	title = {10 - {Nonproliferation} and safeguards aspects of the {MSR} fuel cycle},
	isbn = {978-0-08-101126-3},
	abstract = {In this chapter the reader is introduced to the concepts of nonproliferation: safeguards and security. The identified threats to the peaceful nuclear fuel cycle as well as possible targets, and the existing nuclear safeguards system are discussed. The advantages and disadvantages of the molten salt reactor (MSR) fuel cycle in terms of nonproliferation are discussed and it is compared to that of the light-water reactor fuel cycle, which is widely implemented today.},
	booktitle = {Molten {Salt} {Reactors} and {Thorium} {Energy}},
	publisher = {Woodhead Publishing},
	author = {Grape, Sophie and Hellesen, Carl},
	editor = {Dolan, Thomas J.},
	year = {2017},
	keywords = {Nuclear, nonproliferation, proliferation threat, Safeguards},
	pages = {261--279},
	file = {ScienceDirect Full Text PDF:C\:\\Users\\Sun Myung\\Zotero\\storage\\DEVXYEHF\\Grape and Hellesen - 2017 - 10 - Nonproliferation and safeguards aspects of th.pdf:application/pdf;ScienceDirect Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\7KZI9RWF\\B9780081011263000105.html:text/html},
}

@incollection{yoshioka_7_2017,
	title = {7 - {Materials}},
	isbn = {978-0-08-101126-3},
	abstract = {This chapter discusses the following materials, which are specific to molten salt reactors (MSRs). The areas covered include: molten salt, solid fuels with molten salt coolants, thorium fuel cycle, moderators, and structural materials.
Thorium fuel is also discussed in Chapters 1, Introduction and 9 Environment, waste, and resources, and reprocessing of liquid molten salt fuels is discussed in Chapter 8, Chemical processing of liquid fuel.},
	booktitle = {Molten {Salt} {Reactors} and {Thorium} {Energy}},
	publisher = {Woodhead Publishing},
	author = {Yoshioka, Ritsuo and Kinoshita, Motoyasu and Scott, Ian},
	editor = {Dolan, Thomas J.},
	year = {2017},
	keywords = {Fuel cycle, graphite, moderator, Hastelloy, structural material},
	pages = {189--207},
	file = {ScienceDirect Full Text PDF:C\:\\Users\\Sun Myung\\Zotero\\storage\\3U6LMML7\\Yoshioka et al. - 2017 - 7 - Materials.pdf:application/pdf;ScienceDirect Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\HPYVQI5L\\B9780081011263000075.html:text/html},
}

@incollection{dolan_1_2017,
	title = {1 - {Introduction}},
	isbn = {978-0-08-101126-3},
	abstract = {The United States demonstrated the feasibility of Molten Salt Reactors (MSRs) with the Aircraft Reactor Experiment (1954) and the Molten Salt Reactor Experiment (1965–69). Liquid fuel MSRs can avoid many of the problems of light water reactors (LWRs): fuel manufacture, fuel lifetime, refueling shutdowns, core melt (TMI accident), steam explosion (Chernobyl accident), hydrogen explosion (Fukushima Daiichi accident), and long-lived radioactive (actinide) waste disposal. Thorium fuel can be converted into U-233 fuel, instead of using U-235 from enrichment of natural uranium. One ton of ThO2 can generate as much energy as 293 tons of U3O8 and thorium is four times as abundant in the earth’s crust as uranium. Solid fuel MSRs could be similar to LWRs with molten salt coolant instead of water, so they could be developed quickly, but would lack the advantages of liquid fuel, which include no manufacture of fuel pellets, no fuel melt hazard, fuel burnup not limited by radiation damage, continuous refueling, actinide recycling, and fission product removal. The fuel processing plant must be developed to separate uranium, thorium, actinides, and fission products in a highly radioactive environment. Actinides generated by LWRs could be burned in MSRs, instead of being treated as radioactive wastes requiring geological disposal. Research on MSRs and thorium energy is underway in 23 countries, and reactor designs from several companies are described in this book.},
	booktitle = {Molten {Salt} {Reactors} and {Thorium} {Energy}},
	publisher = {Woodhead Publishing},
	author = {Dolan, Thomas J.},
	year = {2017},
	keywords = {MSR, refueling, uranium, core melt, hydrogen, LWR, steam, wastes},
	pages = {1--12},
	file = {ScienceDirect Full Text PDF:C\:\\Users\\Sun Myung\\Zotero\\storage\\8KGGGEMH\\Dolan - 2017 - 1 - Introduction.pdf:application/pdf;ScienceDirect Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\LKZSPXJI\\B9780081011263000014.html:text/html},
}

@incollection{dolan_27_2017,
	title = {27 - {Issues} and conclusions},
	isbn = {978-0-08-101126-3},
	abstract = {The USA demonstrated the feasibility of molten salt reactors (MSRs) with the Aircraft Reactor Experiment (1954) and the Molten Salt Reactor Experiment (1965–69). Liquid fuel MSRs can avoid many of the problems of light-water reactors (LWRs): fuel manufacture, fuel lifetime, refueling shutdowns, core melt (TMI accident), steam explosion (Chernobyl accident), hydrogen explosion (Fukushima Daichi accident), and long-lived radioactive (actinide) waste disposal. Thorium fuel can be converted into U-233 fuel, instead of using U-235 from enrichment of natural uranium. One ton of ThO2 can generate as much energy as 293 tons of U3O8, and thorium is four times as abundant in the Earth’s crust as uranium. Solid fuel MSRs could be similar to LWRs with molten salt coolant instead of water, so they could be developed quickly, but would lack the advantages of liquid fuel, which include no manufacture of fuel pellets, no fuel melt hazard, fuel burnup not limited by radiation damage, continuous refueling, actinide recycling, and fission product removal. The fuel processing plant must be developed to separate uranium, thorium, actinides, and fission products in a highly radioactive environment. Actinides generated by LWRs could be burned in MSRs, instead of being treated as radioactive waste requiring geological disposal. Research on MSRs and thorium energy is underway in 23 countries, and reactor designs from several companies are described in this book.},
	booktitle = {Molten {Salt} {Reactors} and {Thorium} {Energy}},
	publisher = {Woodhead Publishing},
	author = {Dolan, Thomas J.},
	year = {2017},
	keywords = {MSR, refueling, uranium, core melt, hydrogen, LWR, steam, wastes},
	pages = {775--777},
	file = {ScienceDirect Full Text PDF:C\:\\Users\\Sun Myung\\Zotero\\storage\\8TTBVSYL\\Dolan - 2017 - 27 - Issues and conclusions.pdf:application/pdf;ScienceDirect Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\7JB5QZW7\\B9780081011263000270.html:text/html},
}

@phdthesis{fiorina_molten_2013,
	type = {{PhD}},
	title = {The molten salt fast reactor as a fast spectrum candidate for thorium implementation},
	abstract = {The thesis work investigates the Molten Salt Fast Reactor (MSFR) technology as a possible route to combine the potential advantages of thorium use in Fast Reactors (FR) with the fuel cycle advantages fostered by a liquid fuel. The MSFR emerges as a promising reactor for its capability to operate as a flexible conversion ratio reactor. It shows good performances as U-233 breeder, though uncertainties exist on the U-233 capture cross-section in the neutron energy range of interest for the MSFR. Operation as a self-sustaining reactor fosters low consumption of natural resources, very limited waste generation, and simplified fuel management thanks to the liquid fuel. The MSFR also shows promising features in terms of radioactive waste transmutation thanks to the liquid fuel, the high specific power and the relatively hard spectrum. Safety aspects are investigated through analysis of the reactor safety parameters, and via prediction of the new reactor steady-state after accidental transient initiators. The MSFR inherent safety appears comparable to that of traditional FRs, especially considering its capability to withstand all major double-fault accidents. In addition, the MSFR presents only negative reactivity feedback coefficients, which is a unique feature among fast-spectrum reactors. The system thermal-hydraulics is also investigated in view of the internal heat generation in the working fluid. A correlation is proposed and application to the MSFR allows to exclude major impacts of decay heat on the MSFR out-of-core components, with a note of caution on the design of channels with low velocities and/or large diameters. In addition, a multi-physics model is developed to investigate the thermal-hydraulic behavior of the core, showing some points of enhancement needed in the current MSFR conceptual design. The same model is employed for investigating the reactor transient response to major accidental events, confirming the MSFR promising safety features pointed out with simpler approaches, but suggesting also possible problems related to the quick fuel temperature rise in case of a loss of heat sink.},
	school = {Politecnico Di Milano},
	author = {Fiorina, Carlo},
	month = mar,
	year = {2013},
	keywords = {unread},
	file = {[PDF] from polimi.it:C\:\\Users\\Sun Myung\\Zotero\\storage\\NXBD45NV\\FIORINA - 2013 - The molten salt fast reactor as a fast spectrum ca.pdf:application/pdf;Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\GGQM6IZK\\74324.html:text/html},
}

@article{bettis_aircraft_1957,
	title = {The {Aircraft} {Reactor} {Experiment}},
	volume = {2},
	abstract = {The ARE was operated successfully in November, 1954, at various power levels up to 2.5
MWt. The maximum steady-state fuel temperature was 1580ºF (1130 K), and there was a
differential temperature between the inlet and outlet in the NaF-ZrF4-UF4 fuel of 355ºF (200 K).
The fuel system was in operation for 241 hr before the reactor first became critical and the nuclear
operation extended over a period of 221 hr. The final 74 hr of operation were in the megawatt
range and resulted in the production of 96 MW-hr of nuclear energy. Effects of various transient
conditions on reactor operation were determined.},
	number = {6},
	journal = {Nuclear Science and Engineering},
	author = {Bettis, E. S. and Cottrell, W.B. and Mann, E.R. and Meem, J.L. and Whitman, G.D.},
	year = {1957},
	pages = {841--853},
	file = {NSE_ARE_Operation.pdf:C\:\\Users\\Sun Myung\\Zotero\\storage\\7UGTNXI5\\NSE_ARE_Operation.pdf:application/pdf},
}

@article{heuer_simulation_2010,
	title = {Simulation {Tools} and {New} {Developments} of the {Molten} {Salt} {Fast} {Reactor}},
	copyright = {© SFEN 2010},
	issn = {0335-5004},
	doi = {10.1051/rgn/20106095},
	abstract = {The CNRS has been involved in molten salt reactors since 1997. Starting from the Molten Salt Breeder Reactor project of Oak-Ridge, an innovative concept called Molten Salt Fast Reactor or MSFR has been proposed, resulting from extensive parametric studies in which various core arrangements, reprocessing performances and salt compositions were investigated to adapt the reactor in the framework of the deployment of a thorium based reactor fleet on a worldwide scale. The primary feature of the MSFR concept is the removal of the graphite moderator from the core (graphite-free core), resulting in a breeder reactor with a fast neutron spectrum and operated in the Thorium fuel cycle. MSFR has been recognized as a long term alternative to solid fuelled fast neutron systems with unique potential (negative safety coefficients, smaller fissile inventory, easy in-service inspection, simplified fuel cycle…) and has thus been selected for further studies by the Generation IV International Forum in 2008.In the MSFR, the liquid fuel processing is part of the reactor where a small side stream of the molten salt is processed for fission product removal and then returned to the reactor. This is fundamentally different from a solid fuel reactor where separate facilities produce the solid fuel and process the Spent Nuclear Fuel. Because of this design characteristic, the MSFR can thus operate with widely varying fuel composition. Thanks to this fuel composition flexibility, the MSFR concept may use as initial fissile load, {\textless}sup{\textgreater}233{\textless}sup/{\textgreater}U or enriched (between 5\% and 30\%) uranium or also the transuranic elements currently produced by PWRs in the world.Our reactor’s studies of the MSFR concept rely on numerical simulations making use of the MCNP neutron transport code coupled with a code for materials evolution which resolves the Bateman’s equations giving the population of each nucleus inside each part of the reactor at each moment. Because of MSR’s fundamental characteristics compared to classical solid-fuelled reactors, the classical Bateman equations have to be modified by adding two terms representing the reprocessing efficiencies and the fertile or fissile alimentation. We have finally coupled neutronic and reprocessing simulation codes in a numerical tool develop to calculate the evolution of the whole MSFR system. This tool is used to evaluate the extraction capacities of fission products and their location in the whole system (reactor and reprocessing unit), basis of any safety and radioprotection assessment of the reactor., Le CNRS s’intéresse aux réacteurs à sels fondus (RSF) depuis 1997. Dans le but de proposer un réacteur critique basé sur le cycle thorium pour la production d’énergie, des études plus complètes sont menées à partir de 1999. Une réévaluation complète du MSBR qui constituait alors la configuration de référence des RSF est tout d’abord effectuée, suivie d’une étude systématique d’optimisation du comportement de ce type de réacteurs en s’éloignant du design initial. Ces travaux ont permis de faire émerger un concept de réacteur innovant, qui a été sélectionné fin 2008 par le Forum International Generation IV comme représentant type des réacteurs à sels fondus et qui a désormais comme dénomination officielle : MSFR (Molten Salt Fast Reactor). Il s’agit d’un concept de réacteur surrégénérateur à spectre neutronique rapide et en cycle Thorium, qui présente une très bonne stabilité intrinsèque de fonctionnement grâce à des coefficients de sûreté tous négatifs, et un processus de retraitement du combustible in-situ simplifié acceptable d’un point de vue économique. Ce réacteur se place dans le contexte d’une filière d’utilisation du Thorium qui est un élément fertile abondant dans la nature, en association avec des éléments fissiles naturels ({\textless}sup{\textgreater}235{\textless}sup/{\textgreater}U) ou non ({\textless}sup{\textgreater}233{\textless}sup/{\textgreater}U) ou encore qui proviennent de la chaine de gestion des éléments radioactifs à vie longue des réacteurs actuels (Np, Pu, Am, …).Les études menées sont basées sur le couplage du code de transport de neutrons MCNP avec le code d’évolution des matériaux REM développé au CNRS. La résolution des équations de Bateman qui est effectuée permet de connaître la population de chaque noyau dans chaque partie du réacteur à chaque instant. Du fait des caractéristiques fondamentales des RSF qui sont très différentes de celles des réacteurs à combustible solide, ces équations doivent être modifiées pour prendre en compte la capacité d’alimentation en matière fissile et fertile, ainsi que le retraitement du sel combustible, de manière continue sans arrêt du réacteur. Nous avons finalement associé le code décrivant l’évolution du coeur et une représentation simulée du retraitement pour construire un outil numérique servant au calcul de l’évolution du système complet. Cet outil permet d’évaluer l’efficacité du retraitement vis-à-vis de l’ensemble des produits de fission (en supposant une efficacité donnée pour le procédé physicochimique utilisé), ainsi que de déterminer leur localisation dans le système. Ces informations sont essentielles pour pouvoir aborder une étude de la sûreté ou de la radioprotection du système.},
	language = {en},
	number = {6},
	journal = {Revue Générale Nucléaire},
	author = {Heuer, D. and Merle-Lucotte, E. and Allibert, M. and Doligez, X. and Ghetta, V.},
	year = {2010},
	pages = {95--100},
	file = {Heuer et al. - 2010 - Simulation Tools and New Developments of the Molte.pdf:C\:\\Users\\Sun Myung\\Zotero\\storage\\N9LPQ8AB\\Heuer et al. - 2010 - Simulation Tools and New Developments of the Molte.pdf:application/pdf;Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\ECV96N73\\rgn20106p95.html:text/html},
}

@article{kophazi_development_2009,
	title = {Development of a {Three}-{Dimensional} {Time}-{Dependent} {Calculation} {Scheme} for {Molten} {Salt} {Reactors} and {Validation} of the {Measurement} {Data} of the {Molten} {Salt} {Reactor} {Experiment}},
	volume = {163},
	abstract = {This paper presents the development, validation, and results of a three-dimensional, time- dependent, coupled-neutronics–thermal-hydraulic calculational scheme for channel-type molten salt re- actors (MSRs). The reactor physics part is based on diffusion theory, extended by a term representing the flow of the fuel through the core. The calculation of the temperature field is done by modeling all fuel channels, which are coupled to each other by a three-dimensional heat conduction equation. For the purpose of validation, the results of the MSR Experiment (MSRE) natural-circulation experiment and the thermal feedback coefficients of the reactor have been calculated and compared.
With the aid of a code system developed to implement this scheme, calculations were carried out for the normal operating state of the MSRE and some debris-induced channel-blocking-incident transients. In the case of the MSRE, it is shown that the severity of such an incident strongly depends on the degree of channel blocking and that high-temperature gradients in the moderator can connect thermally the adjacent fuel channels. Results are included for an unblocking transient (i.e., the debris suddenly exits the core, and the fuel flow reverts to the normal operating pattern), and it was demonstrated that during the unblocking large power peaks can be induced.},
	number = {2},
	journal = {Nuclear Science and Engineering},
	author = {Kophazi, J. and Lathouwers, D. and Kloosterman, J.L.},
	year = {2009},
	keywords = {unread, Molten Salt Reactor (MSR), 3D, reactor physics, Core, MSR Experiment (MSRE)},
	pages = {118--131},
	file = {Kópházi et al. - Development of a Three-Dimensional Time-Dependent .pdf:C\:\\Users\\Sun Myung\\Zotero\\storage\\ZIJ5Q643\\Kópházi et al. - Development of a Three-Dimensional Time-Dependent .pdf:application/pdf},
}

@inproceedings{aufiero_testing_2018,
	address = {Philadelphia, Pennsylvania},
	series = {Reactor {Physics}: {General}—{I}},
	title = {Testing and {Verification} of {Multiphysics} {Tools} for {Fast}-{Spectrum} {MSRs}: {The} {CNRS} {Benchmark}},
	volume = {118},
	shorttitle = {Testing and {Verification} of {Multiphysics} {Tools} for {Fast}-{Spectrum} {MSRs}},
	abstract = {Liquid fuel Molten Salt Reactors (MSRs) feature peculiar physical
phenomena that are not present in solid fuel reactors. Some of
these phenomena are not correctly resolved by legacy reactor physics
tools, and require specific treatments. Recently, research activities
within several collaborative projects related to MSRs design (e.g.,
see [1], [2]) lead to development of a number of different numerical
tools for the coupled neutronics and thermal/hydraulics analysis of
such systems. Unfortunately, especially in case of non-moderated,
fast-spectrum MSRs, very few experimental data are available for
an accurate process of verification and validation of the developed
codes.},
	language = {English},
	booktitle = {Transactions of the {American} {Nuclear} {Society}},
	publisher = {American Nuclear Society},
	author = {Aufiero, Manuele and Rubiolo, Pablo},
	month = jun,
	year = {2018},
	pages = {837--840},
	file = {Fulltext:C\:\\Users\\Sun Myung\\Zotero\\storage\\HUGQPFM3\\Aufiero and Rubiolo - Testing and Verification of Multiphysics Tools for.pdf:application/pdf;Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\SXY5L3Q3\\Aufiero and Rubiolo - Testing and Verification of Multiphysics Tools for.pdf:application/pdf},
}

@article{zanetti_geometric_2015,
	title = {A {Geometric} {Multiscale} modelling approach to the analysis of {MSR} plant dynamics},
	volume = {83},
	issn = {0149-1970},
	doi = {10.1016/j.pnucene.2015.02.014},
	abstract = {In the framework of the Generation IV International Forum (GIF-IV), six innovative concepts of nuclear reactors have been proposed as suitable to guarantee a safe, sustainable and proliferation resistant source of nuclear energy. Among these reactors, a peculiar role is played by the Molten Salt Reactor (MSR), which is the only one with a liquid and circulating fuel. This feature leads to a complex and highly coupled behaviour, which requires careful investigations, as a consequence of some unusual features like the drift of Delayed Neutron Precursors (DNP) along the primary circuit and heat transfer with a heat-generating fluid. The inherently coupled dynamics of the MSRs asks for innovative approaches to perform reliable transient analyses. The node-wise implicitly-coupled solution of the Partial Differential Equations (PDE) that govern the different phenomena in a reactor would offer in this sense an ideal solution. However, such an approach (hereinafter referred to as Multi-Physics – MP) requires a huge amount of computational power. In this work, we propose and assess a Geometric Multiscale approach on MSR, addressing the core modelling with a 3-D MP approach and the remaining part of the system – e.g., the cooling loop – with simplified 0-D models based on Ordinary Differential Equations (ODE). The aim is to conjugate the accuracy of the MP modelling approach with acceptable computation loads. Reference is made to the Molten Salt Reactor Experiment (MSRE), due to the availability of a detailed design and experimental data that are used for assessment and preliminary validation of the developed simulation tool.},
	number = {Supplement C},
	journal = {Progress in Nuclear Energy},
	author = {Zanetti, Matteo and Cammi, Antonio and Fiorina, Carlo and Luzzi, Lelio},
	month = aug,
	year = {2015},
	keywords = {Molten Salt Reactor Experiment (MSRE), Molten Salt Reactor (MSR), Geometric Multiscale approach, Multi-Physics Modelling, System dynamic behaviour},
	pages = {82--98},
	file = {ScienceDirect Full Text PDF:C\:\\Users\\Sun Myung\\Zotero\\storage\\JIPBQTRD\\Zanetti et al. - 2015 - A Geometric Multiscale modelling approach to the a.pdf:application/pdf;ScienceDirect Full Text PDF:C\:\\Users\\Sun Myung\\Zotero\\storage\\ZDHUIY95\\Zanetti et al. - 2015 - A Geometric Multiscale modelling approach to the a.pdf:application/pdf;ScienceDirect Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\NTTJSHEN\\S0149197015000487.html:text/html;ScienceDirect Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\ZZT9I5TK\\S0149197015000487.html:text/html},
}

@mastersthesis{pettersen_coupled_2016,
	title = {Coupled multi-physics simulations of the {Molten} {Salt} {Fast} {Reactor} using coarse-mesh thermal-hydraulics and spatial neutronics},
	school = {MSc thesis, September 2016 (PDF)},
	author = {Pettersen, Eirik Eide and Mikityuk, Konstantin},
	year = {2016},
	file = {[PDF] samofar.eu:C\:\\Users\\Sun Myung\\Zotero\\storage\\XVIH74PK\\Pettersen and Mikityuk - 2016 - Coupled multi-physics simulations of the Molten Sa.pdf:application/pdf;Fulltext:C\:\\Users\\Sun Myung\\Zotero\\storage\\PA885TB5\\Pettersen and Mikityuk - 2016 - Coupled multi-physics simulations of the Molten Sa.pdf:application/pdf;Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\S7SERYFF\\Pettersen and Mikityuk - 2016 - Coupled multi-physics simulations of the Molten Sa.pdf:application/pdf},
}

@article{mathieu_thorium_2006,
	title = {The thorium molten salt reactor: {Moving} on from the {MSBR}},
	volume = {48},
	issn = {0149-1970},
	shorttitle = {The thorium molten salt reactor},
	doi = {10.1016/j.pnucene.2006.07.005},
	abstract = {A re-evaluation of the molten salt breeder reactor concept has revealed problems related to its safety and to the complexity of the reprocessing considered. A reflection is carried out anew in view of finding innovative solutions leading to the thorium molten salt reactor concept. Several main constraints are established and serve as guides to parametric evaluations. These then give an understanding of the influence of important core parameters on the reactor's operation. The aim of this paper is to discuss this vast research domain and to single out the molten salt reactor configurations that deserve further evaluation.},
	number = {7},
	journal = {Progress in Nuclear Energy},
	author = {Mathieu, L. and Heuer, D. and Brissot, R. and Garzenne, C. and Le Brun, C. and Lecarpentier, D. and Liatard, E. and Loiseaux, J.-M. and Méplan, O. and Merle-Lucotte, E. and Nuttin, A. and Walle, E. and Wilson, J.},
	month = sep,
	year = {2006},
	keywords = {read, breeding, Feedback coefficient, generation IV, Neutronic, Nuclear Experiment, thorium cycle},
	pages = {664--679},
	file = {arXiv.org Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\BWFNNTBB\\0506004.html:text/html;nucl-ex/0506004 PDF:C\:\\Users\\Sun Myung\\Zotero\\storage\\P88D4PAF\\Mathieu et al. - 2005 - The Thorium Molten Salt Reactor  Moving on from t.pdf:application/pdf;ScienceDirect Full Text PDF:C\:\\Users\\Sun Myung\\Zotero\\storage\\4ENSAB4V\\Mathieu et al. - 2006 - The thorium molten salt reactor Moving on from th.pdf:application/pdf;ScienceDirect Full Text PDF:C\:\\Users\\Sun Myung\\Zotero\\storage\\WFQ6DBDV\\Mathieu et al. - 2006 - The thorium molten salt reactor Moving on from th.pdf:application/pdf;ScienceDirect Full Text PDF:C\:\\Users\\Sun Myung\\Zotero\\storage\\CSJWUU8C\\Mathieu et al. - 2006 - The thorium molten salt reactor Moving on from th.pdf:application/pdf;ScienceDirect Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\EXG3AUMK\\S0149197006000746.html:text/html;ScienceDirect Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\RGUXZFUN\\S0149197006000746.html:text/html;ScienceDirect Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\X5WR4R55\\S0149197006000746.html:text/html;thoriummsr.pdf:C\:\\Users\\Sun Myung\\Zotero\\storage\\6EDWBWGB\\thoriummsr.pdf:application/pdf;thoriummsr.pdf:C\:\\Users\\Sun Myung\\Zotero\\storage\\R9UGK2VX\\thoriummsr.pdf:application/pdf},
}

@incollection{leblanc_18_2017,
	title = {18 - {Integral} molten salt reactor},
	isbn = {978-0-08-101126-3},
	abstract = {The IMSR uses molten fluoride salt, a highly stable, inert liquid with robust coolant properties and high intrinsic radionuclide retention properties, for its primary fuel salt. A secondary, coolant salt loop, also using a fluoride salt (but without fuel), transfers heat away from the primary heat exchangers integrated inside the core-unit. The coolant salt loop, in turn, transfers its heat load to a solar salt loop, which is pumped out of the nuclear island to a separate building where it either heats steam generators that generate superheated steam for power generation or is used for process heat applications. The safety philosophy behind the IMSR is to produce a nuclear power plant with generation IV reactor levels of safety. For ultimate safety, there is no dependence on operator intervention, powered mechanical components, coolant injection or their support systems, such as electricity supply or instrument air in dealing with upset conditions. This is achieved through a combination of design features: the inert, stable properties of the salt; an inherently stable nuclear core; fully passive backup core and containment cooling systems; and an integral reactor architecture.},
	booktitle = {Molten {Salt} {Reactors} and {Thorium} {Energy}},
	publisher = {Woodhead Publishing},
	author = {LeBlanc, David and Rodenburg, Cyril},
	editor = {Dolan, Thomas J.},
	year = {2017},
	keywords = {IMSR, Safety, generation IV, fluoride, integral molten salt reactor, nuclear system},
	pages = {541--556},
	file = {ScienceDirect Full Text PDF:C\:\\Users\\Sun Myung\\Zotero\\storage\\TJGWFYKX\\LeBlanc and Rodenburg - 2017 - 18 - Integral molten salt reactor.pdf:application/pdf;ScienceDirect Full Text PDF:C\:\\Users\\Sun Myung\\Zotero\\storage\\BWQIA75R\\LeBlanc and Rodenburg - 2017 - 18 - Integral molten salt reactor.pdf:application/pdf;ScienceDirect Full Text PDF:C\:\\Users\\Sun Myung\\Zotero\\storage\\XW9JXU3P\\LeBlanc and Rodenburg - 2017 - 18 - Integral molten salt reactor.pdf:application/pdf;ScienceDirect Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\FIC5CLH8\\B978008101126300018X.html:text/html;ScienceDirect Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\4D5HBG5Q\\B978008101126300018X.html:text/html;ScienceDirect Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\SBH7K26M\\B978008101126300018X.html:text/html},
}

@techreport{iaea_thorium_2005,
	address = {Vienna, Austria},
	title = {Thorium {Fuel} {Cycle} - {Potential} {Benefits} and {Challenges}},
	url = {http://www-pub.iaea.org/books/IAEABooks/7192/Thorium-Fuel-Cycle-Potential-Benefits-and-Challenges},
	abstract = {Thorium Fuel Cycle Potential Benefits and Challenges},
	language = {English},
	number = {IAEA-TECDOC-1450},
	institution = {Nuclear Fuel Cycle and Materials Section International Atomic Energy Agency},
	author = {IAEA},
	month = may,
	year = {2005},
	pages = {113},
	file = {Full Text PDF:C\:\\Users\\Sun Myung\\Zotero\\storage\\4SR2PB6K\\IAEA - 2005 - Thorium Fuel Cycle - Potential Benefits and Challe.pdf:application/pdf;IAEA Report on Thorium Fuel Cycle--May 2005.pdf:C\:\\Users\\Sun Myung\\Zotero\\storage\\EI8HB6GT\\IAEA Report on Thorium Fuel Cycle--May 2005.pdf:application/pdf;Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\38XVA5P2\\Thorium-Fuel-Cycle-Potential-Benefits-and-Challenges.html:text/html},
}

@article{heuer_towards_2014,
	title = {Towards the thorium fuel cycle with molten salt fast reactors},
	volume = {64},
	issn = {0306-4549},
	doi = {10.1016/j.anucene.2013.08.002},
	abstract = {There is currently a renewed interest in molten salt reactors, due to recent conceptual developments on fast neutron spectrum molten salt reactors (MSFRs) using fluoride salts. It has been recognized as a long term alternative to solid-fueled fast neutron systems with a unique potential (large negative temperature and void coefficients, lower fissile inventory, no initial criticality reserve, simplified fuel cycle, wastes reduction etc.) and is thus one of the reference reactors of the Generation IV International Forum. In the MSFR, the liquid fuel processing is part of the reactor where a small side stream of the molten salt is processed for fission product removal and then returned to the reactor. Because of this characteristic, the MSFR can operate with widely varying fuel compositions, so that the MSFR concept may use as initial fissile load, 233U or enriched uranium or also the transuranic elements currently produced by light water reactors. This paper addresses the characteristics of these different launching modes of the MSFR and the Thorium fuel cycle, in terms of safety, proliferation, breeding, and deployment capacities of these reactor configurations. To illustrate the deployment capacities of the MSFR concept, a French nuclear deployment scenario is finally presented, demonstrating that launching the Thorium fuel cycle is easily feasible while closing the current fuel cycle and optimizing the long-term waste management via stockpile incineration in MSRs.},
	journal = {Annals of Nuclear Energy},
	author = {Heuer, D. and Merle-Lucotte, E. and Allibert, M. and Brovchenko, M. and Ghetta, V. and Rubiolo, P.},
	month = feb,
	year = {2014},
	keywords = {Thorium fuel cycle, MSFR, Deployment scenario, Generation 4 reactors, Incinerator},
	pages = {421--429},
	file = {ScienceDirect Full Text PDF:C\:\\Users\\Sun Myung\\Zotero\\storage\\RNF6R5F9\\Heuer et al. - 2014 - Towards the thorium fuel cycle with molten salt fa.pdf:application/pdf;ScienceDirect Full Text PDF:C\:\\Users\\Sun Myung\\Zotero\\storage\\8ETDPPSM\\Heuer et al. - 2014 - Towards the thorium fuel cycle with molten salt fa.pdf:application/pdf;ScienceDirect Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\TFXWLMD3\\S0306454913004106.html:text/html;ScienceDirect Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\2ILQRZ53\\S0306454913004106.html:text/html},
}

@incollection{dai_17_2017,
	title = {17 - {Thorium} molten salt reactor nuclear energy system ({TMSR})},
	isbn = {978-0-08-101126-3},
	abstract = {The thorium molten salt reactor nuclear energy system (TMSR) is designed for thorium-based nuclear energy utilization and hybrid nuclear energy application, based on a liquid-fueled thorium molten salt reactor (TMSR-LF) and a solid-fueled thorium molten salt reactor (TMSR-SF).},
	booktitle = {Molten {Salt} {Reactors} and {Thorium} {Energy}},
	publisher = {Woodhead Publishing},
	author = {Dai, Zhimin},
	editor = {Dolan, Thomas J.},
	year = {2017},
	keywords = {Molten salt reactor, thorium, hybrid nuclear energy application},
	pages = {531--540},
	file = {ScienceDirect Full Text PDF:C\:\\Users\\Sun Myung\\Zotero\\storage\\IUVWTXVJ\\Dai - 2017 - 17 - Thorium molten salt reactor nuclear energy sy.pdf:application/pdf;ScienceDirect Full Text PDF:C\:\\Users\\Sun Myung\\Zotero\\storage\\QGM7UCAA\\Dai - 2017 - 17 - Thorium molten salt reactor nuclear energy sy.pdf:application/pdf;ScienceDirect Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\SUYR248M\\B9780081011263000178.html:text/html;ScienceDirect Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\4JZSFD3K\\B9780081011263000178.html:text/html},
}

@book{massachusetts_institute_of_technology_future_2003,
	address = {Boston MA},
	title = {The {Future} of nuclear power: an interdisciplinary {MIT} {Study}.},
	isbn = {978-0-615-12420-9},
	shorttitle = {The {Future} of nuclear power},
	language = {English},
	publisher = {MIT},
	author = {{Massachusetts Institute of Technology}},
	year = {2003},
	file = {nuclearpower-full.pdf:C\:\\Users\\Sun Myung\\Zotero\\storage\\QRUQM5MF\\nuclearpower-full.pdf:application/pdf},
}

@techreport{briggs_molten-salt_1969,
	title = {Molten-salt reactor program. {Semiannual} progress report},
	language = {English},
	number = {ORNL-4396},
	institution = {Oak Ridge National Lab., Tenn.},
	author = {Briggs, R. B.},
	month = feb,
	year = {1969},
	keywords = {coolant loops, gases, msre, planning, reactor technology, reactors, air, decontamination, detection, fused salt fuel, leaks, liquids, pressure vessels, research and test reactors, research reactors, sampling, shielding, waste disposal, water coolant},
	file = {Briggs - 1969 - Molten-salt reactor program. Semiannual progress r.pdf:C\:\\Users\\Sun Myung\\Zotero\\storage\\X8DJKHWI\\Briggs - 1969 - Molten-salt reactor program. Semiannual progress r.pdf:application/pdf},
}

@techreport{euratom_final_2015,
	address = {France},
	type = {Final report},
	title = {Final {Report} {Summary} - {EVOL} ({Evaluation} and {Viability} of {Liquid} {Fuel} {Fast} {Reactor} {System}) {\textbar} {Report} {Summary} {\textbar} {EVOL} {\textbar} {FP7}{\textbar} {European} {Commission}},
	url = {https://cordis.europa.eu/result/rcn/159411_en.html},
	abstract = {Executive Summary:An innovative molten salt reactor concept, the MSFR (Molten Salt Fast Reactor) is developed by CNRS (France) since 2004. Based on the particularity of using a liquid fuel, this...},
	language = {en},
	number = {249696},
	institution = {EURATOM},
	author = {EURATOM},
	year = {2015},
	file = {Final Report Summary - EVOL (Evaluation and Viabil.pdf:C\:\\Users\\Sun Myung\\Zotero\\storage\\HB992DYS\\Final Report Summary - EVOL (Evaluation and Viabil.pdf:application/pdf;Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\67KEHCS2\\159411_en.html:text/html},
}

@article{cervi_development_2019,
	title = {Development of an {SP3} neutron transport solver for the analysis of the {Molten} {Salt} {Fast} {Reactor}},
	volume = {346},
	issn = {0029-5493},
	doi = {10.1016/j.nucengdes.2019.03.001},
	abstract = {The aim of this paper is the extension of a multiphysics OpenFOAM solver for the analysis of the Molten Salt Fast Reactor (MSFR), developed in previous works (Cervi et al., 2017, 2018). In particular, the neutronics sub-solver is improved by implementing a new module based on the SP3 approximation of the neutron transport equation. The new module is successfully tested against a Monte Carlo model of the MSFR, in order to assess its correct implementation. Then, a neutronics analysis of the MSFR is carried out on a simplified axial-symmetric model of the reactor. Particular focus is devoted to the analysis of the MSFR helium bubbling system and its effect on reactivity. The presence of bubbles inside the reactor is handled with a two-fluid thermal-hydraulics module, previously implemented into the solver. The void reactivity coefficient is evaluated on the basis of the bubble spatial distribution calculated by the multiphysics solver. Then, the results are compared to simulations carried out with uniform bubble distributions, highlighting significant differences between the two approaches. The outcomes of this work constitute a step forward in the multiphysics analysis of the Molten Salt Fast Reactor and represent a useful starting point for the optimization of the MSFR helium bubbling system, as well as for the development of appropriate control strategies.},
	journal = {Nuclear Engineering and Design},
	author = {Cervi, E. and Lorenzi, S. and Cammi, A. and Luzzi, L.},
	month = may,
	year = {2019},
	keywords = {Multiphysics, Neutron transport, Molten Salt Fast Reactor (MSFR), OpenFOAM},
	pages = {209--219},
	file = {ScienceDirect Full Text PDF:C\:\\Users\\Sun Myung\\Zotero\\storage\\2U45NTZE\\Cervi et al. - 2019 - Development of an SP3 neutron transport solver for.pdf:application/pdf;ScienceDirect Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\JDS5R3Y4\\S0029549319300354.html:text/html},
}

@article{rykhlevskii_modeling_2019,
	title = {Modeling and simulation of online reprocessing in the thorium-fueled molten salt breeder reactor},
	volume = {128},
	issn = {0306-4549},
	doi = {10.1016/j.anucene.2019.01.030},
	abstract = {In the search for new ways to generate carbon-free, reliable base-load power, interest in advanced nuclear energy technologies, particularly Molten Salt Reactors (MSRs), has resurged with multiple new companies pursuing MSR commercialization. To further develop these MSR concepts, researchers need simulation tools for analyzing liquid-fueled MSR depletion and fuel processing. However, most contemporary nuclear reactor physics software is unable to perform high-fidelity full-core depletion calculations for a reactor design with online reprocessing. This paper introduces a Python package, SaltProc, which couples with the Monte Carlo code, SERPENT2 to simulate MSR online reprocessing by modeling the changing isotopic composition of MSR fuel salt. This work demonstrates SaltProc capabilities for a full-core, high-fidelity model of the commercial Molten Salt Breeder Reactor (MSBR) concept and verifies these results to results in the literature from independent, lower-fidelity analyses.},
	journal = {Annals of Nuclear Energy},
	author = {Rykhlevskii, Andrei and Bae, Jin Whan and Huff, Kathryn D.},
	month = jun,
	year = {2019},
	keywords = {Reactor physics, Parallel computing, Molten salt reactor, Online reprocessing, agent based modeling, Finite elements, Hydrologic contaminant transport, MOOSE, Multiphysics, nuclear engineering, Nuclear fuel cycle, Object orientation, repository, Simulation, Systems analysis, Depletion, Molten salt breeder reactor, Python, Salt treatment},
	pages = {366--379},
	file = {Rykhlevskii et al. - 2019 - Modeling and simulation of online reprocessing in .pdf:C\:\\Users\\Sun Myung\\Zotero\\storage\\IQPHC25M\\Rykhlevskii et al. - 2019 - Modeling and simulation of online reprocessing in .pdf:application/pdf;ScienceDirect Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\W5RDKFMS\\S0306454919300350.html:text/html;ScienceDirect Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\P6PJ667C\\S0306454919300350.html:text/html},
}

@article{cervi_development_2019-1,
	title = {Development of a multiphysics model for the study of fuel compressibility effects in the {Molten} {Salt} {Fast} {Reactor}},
	volume = {193},
	issn = {0009-2509},
	doi = {10.1016/j.ces.2018.09.025},
	abstract = {Compressible fluid dynamics is of great practical interest in many industrial applications, ranging from chemistry to aeronautical industry, and to nuclear field as well. At the same time, modelling and simulation of compressible flows is a very complex task, requiring the development of specific approaches, in order to describe the effect of pressure on the fluid velocity field. Compressibility effects become even more important in the study of two-phase flows, due to the presence of a gaseous phase. In addition, compressibility is also expected to have a significant impact on other physics, such as chemical or nuclear reactions occurring in the mixture. In this perspective, multiphysics represents a useful approach to address this complex problem, providing a way to catch all the different physics that come into play as well as the coupling between them. In this work, a multiphysics model is developed for the analysis of the generation IV Molten Salt Fast Reactor (MSFR), with a specific focus on the compressibility effects of the fluid that acts as fuel in the reactor. The fuel mixture compressibility is expected to have an important effect on the system dynamics, especially in very rapid super-prompt-critical transients. In addition, the presence of a helium bubbling system used for online fission product removal could modify the fuel mixture compressibility, further affecting the system transient behaviour. Therefore, the MSFR represents an application of concrete interest, inherent to the analysis of compressibility effects and to the development of suitable modelling approaches. An OpenFOAM solver is developed to handle the fuel compressibility, the presence of gas bubbles in the reactor as well as the coupling between the system neutronics and fluid dynamics. The outcomes of this analysis point out that the fuel compressibility plays a crucial role in the evolution of fast transients, introducing delays in the expansion feedbacks that strongly affect the system dynamics. Moreover, it is found that the gas bubbles significantly alter the fuel compressibility, yielding even larger differences compared to the incompressible approximation usually adopted in the current MSFR solvers.},
	journal = {Chemical Engineering Science},
	author = {Cervi, E. and Lorenzi, S. and Cammi, A. and Luzzi, L.},
	month = jan,
	year = {2019},
	keywords = {Multiphysics, Reactor dynamics, Molten Salt Fast Reactor (MSFR), OpenFOAM, Compressible fluid dynamics},
	pages = {379--393},
	file = {Fulltext:C\:\\Users\\Sun Myung\\Zotero\\storage\\WNYVUELP\\S0009250918306742.html:text/html;ScienceDirect Full Text PDF:C\:\\Users\\Sun Myung\\Zotero\\storage\\WMQ8ZVWE\\Cervi et al. - 2019 - Development of a multiphysics model for the study .pdf:application/pdf;ScienceDirect Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\48G8AZZI\\S0009250918306742.html:text/html;Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\P39Y9VFL\\S0009250918306742.html:text/html},
}

@article{aufiero_extended_2013,
	title = {An extended version of the {SERPENT}-2 code to investigate fuel burn-up and core material evolution of the {Molten} {Salt} {Fast} {Reactor}},
	volume = {441},
	issn = {0022-3115},
	doi = {10.1016/j.jnucmat.2013.06.026},
	abstract = {In this work, the Monte Carlo burn-up code SERPENT-2 has been extended and employed to study the material isotopic evolution of the Molten Salt Fast Reactor (MSFR).

This promising GEN-IV nuclear reactor concept features peculiar characteristics such as the on-line fuel reprocessing, which prevents the use of commonly available burn-up codes. Besides, the presence of circulating nuclear fuel and radioactive streams from the core to the reprocessing plant requires a precise knowledge of the fuel isotopic composition during the plant operation.

The developed extension of SERPENT-2 directly takes into account the effects of on-line fuel reprocessing on burn-up calculations and features a reactivity control algorithm. It is here assessed against a dedicated version of the deterministic ERANOS-based EQL3D procedure (PSI-Switzerland) and adopted to analyze the MSFR fuel salt isotopic evolution.

Particular attention is devoted to study the effects of reprocessing time constants and efficiencies on the conversion ratio and the molar concentration of elements relevant for solubility issues (e.g., trivalent actinides and lanthanides). Quantities of interest for fuel handling and safety issues are investigated, including decay heat and activities of hazardous isotopes (neutron and high energy gamma emitters) in the core and in the reprocessing stream. The radiotoxicity generation is also analyzed for the MSFR nominal conditions.

The production of helium and the depletion in tungsten content due to nuclear reactions are calculated for the nickel-based alloy selected as reactor structural material of the MSFR. These preliminary evaluations can be helpful in studying the radiation damage of both the primary salt container and the axial reflectors.},
	number = {1–3},
	journal = {Journal of Nuclear Materials},
	author = {Aufiero, M. and Cammi, A. and Fiorina, C. and Leppänen, J. and Luzzi, L. and Ricotti, M. E.},
	month = oct,
	year = {2013},
	pages = {473--486},
	file = {ScienceDirect Full Text PDF:C\:\\Users\\Sun Myung\\Zotero\\storage\\GACISQTN\\Aufiero et al. - 2013 - An extended version of the SERPENT-2 code to inves.pdf:application/pdf;ScienceDirect Full Text PDF:C\:\\Users\\Sun Myung\\Zotero\\storage\\FK5H57EA\\Aufiero et al. - 2013 - An extended version of the SERPENT-2 code to inves.pdf:application/pdf;ScienceDirect Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\GV4RYF2E\\S0022311513008507.html:text/html;ScienceDirect Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\VTCQESRU\\S0022311513008507.html:text/html},
}

@techreport{robertson_conceptual_1971,
	title = {Conceptual {Design} {Study} of a {Single}-{Fluid} {Molten}-{Salt} {Breeder} {Reactor}.},
	language = {English},
	number = {ORNL--4541},
	institution = {ORNL},
	author = {Robertson, R. C.},
	month = jan,
	year = {1971},
	keywords = {Economics, buildings, configuration, control systems, coolant loops, operation, design, fused salt fuel, MSBR, boilers, containment, maintenance, n38100*  --power reactor development, reactor sites, waste disposal molten salt breeder reactor/design parameters for conceptual 1000 mw(e) single fluid},
	file = {Robertson - 1971 - Conceptual Design Study of a Single-Fluid Molten-S.pdf:C\:\\Users\\Sun Myung\\Zotero\\storage\\5X3PJ4F7\\Robertson - 1971 - Conceptual Design Study of a Single-Fluid Molten-S.pdf:application/pdf;Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\G8IEVM75\\4030941.html:text/html},
}

@techreport{oecd/nea_jeff-3.1.2_2014,
	title = {The {JEFF}-3.1.2 {Nuclear} {Data} {Library}},
	number = {JEFF Report 24, OECD/NEA Data Bank},
	institution = {OECD/NEA},
	author = {OECD/NEA},
	year = {2014},
}

@techreport{iea_global_2019,
	address = {Paris, France},
	title = {Global {Energy} and {CO2} {Status} {Report} 2018},
	shorttitle = {The latest trends in {Energy} and {Emissions} in 2018},
	url = {https://webstore.iea.org/global-energy-co2-status-report-2018},
	language = {en},
	institution = {International Energy Agency},
	author = {{IEA}},
	month = mar,
	year = {2019},
	pages = {29},
	file = {2018 - Global Energy and CO2 Status Report 2018.pdf:C\:\\Users\\Sun Myung\\Zotero\\storage\\Y7A8E5GK\\2018 - Global Energy and CO2 Status Report 2018.pdf:application/pdf},
}

@techreport{smith_assessment_1974,
	title = {An assessment of a 2500 {MWe} molten chloride salt fast reactor},
	language = {en},
	number = {AEEW-R956},
	institution = {United Kindom Atomic Energy Authority},
	author = {Smith, J. and Simmons, W. E.},
	month = aug,
	year = {1974},
	pages = {79},
	file = {Fulltext:C\:\\Users\\Sun Myung\\Zotero\\storage\\CS6DLW4V\\Smith and Simmons - 1974 - An assessment of a 2500 MWe molten chloride salt f.pdf:application/pdf;SIMMONS - AN ASSESSMENT OF A 2500 MWe MOLTEN CHLORIDE SALT F.pdf:C\:\\Users\\Sun Myung\\Zotero\\storage\\A4H5NGGH\\SIMMONS - AN ASSESSMENT OF A 2500 MWe MOLTEN CHLORIDE SALT F.pdf:application/pdf;Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\BTJQKB53\\Smith and Simmons - 1974 - An assessment of a 2500 MWe molten chloride salt f.pdf:application/pdf},
}

@article{zhang_review_2018,
	title = {Review of conceptual design and fundamental research of molten salt reactors in {China}},
	volume = {42},
	issn = {1099-114X},
	doi = {10.1002/er.3979},
	abstract = {Molten salt reactor (MSR) as 1 candidate of the generation IV advanced nuclear power systems attracted more attention in China due to its top ranked in fuel cycle and thorium utilization. Two types of MSR concepts were studied and developed in parallel, namely the MSR with liquid fuel and that with solid fuel. Abundant fundamental research including the neutronics modeling, thermal-hydraulics modeling, safety analysis, material investigation, molten salts technologies etc. were carried out. Some analysis software such as COUPLE and FANCY were developed. Several experimental facilities like high-temperature fluoride salt experiment loop have been constructed. Some passive residual heat removal systems were designed, and 1 test facility is under construction. The key MSR techniques including the extraction and separation of molten salt and construction of N-base alloy have been mastered. Based on these fundamental research, Chinese Academy of Sciences has completed the design of thorium-based MSRs with solid fuel and liquid fuel and is promoting their construction in the near future. In China, future efforts should be paid to the material, online fuel processing, Th-U fuel cycle, component design, and construction and thermal-hydraulic experiments for MSR, which are rather challenging nowadays.},
	language = {en},
	number = {5},
	journal = {International Journal of Energy Research},
	author = {Zhang, Dalin and Liu, Limin and Liu, Minghao and Xu, Rongshuan and Gong, Cheng and Zhang, Jun and Wang, Chenglong and Qiu, Suizheng and Su, Guanghui},
	year = {2018},
	keywords = {liquid fuel, fundamental research, solid fuel, thorium utilization},
	pages = {1834--1848},
	file = {Full Text:C\:\\Users\\Sun Myung\\Zotero\\storage\\HJ3PGQLF\\Zhang et al. - 2018 - Review of conceptual design and fundamental resear.pdf:application/pdf;Full Text PDF:C\:\\Users\\Sun Myung\\Zotero\\storage\\NS7Z2PPK\\Zhang et al. - Review of conceptual design and fundamental resear.pdf:application/pdf;Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\F6ZGQ2BL\\abstract.html:text/html;Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\MA5PMVXA\\er.html:text/html},
}

@article{romano_openmc:_2015,
	series = {Joint {International} {Conference} on {Supercomputing} in {Nuclear} {Applications} and {Monte} {Carlo} 2013, {SNA} + {MC} 2013. {Pluri}- and {Trans}-disciplinarity, {Towards} {New} {Modeling} and {Numerical} {Simulation} {Paradigms}},
	title = {{OpenMC}: {A} state-of-the-art {Monte} {Carlo} code for research and development},
	volume = {82},
	issn = {0306-4549},
	shorttitle = {{OpenMC}},
	doi = {10.1016/j.anucene.2014.07.048},
	abstract = {This paper gives an overview of OpenMC, an open source Monte Carlo particle transport code recently developed at the Massachusetts Institute of Technology. OpenMC uses continuous-energy cross sections and a constructive solid geometry representation, enabling high-fidelity modeling of nuclear reactors and other systems. Modern, portable input/output file formats are used in OpenMC: XML for input, and HDF5 for output. High performance parallel algorithms in OpenMC have demonstrated near-linear scaling to over 100,000 processors on modern supercomputers. Other topics discussed in this paper include plotting, CMFD acceleration, variance reduction, eigenvalue calculations, and software development processes.},
	number = {Supplement C},
	journal = {Annals of Nuclear Energy},
	author = {Romano, Paul K. and Horelik, Nicholas E. and Herman, Bryan R. and Nelson, Adam G. and Forget, Benoit and Smith, Kord},
	month = aug,
	year = {2015},
	keywords = {Monte Carlo, Neutron transport, OpenMC, HDF5, Parallel, XML},
	pages = {90--97},
	file = {ScienceDirect Full Text PDF:C\:\\Users\\Sun Myung\\Zotero\\storage\\9S36DPTN\\Romano et al. - 2015 - OpenMC A state-of-the-art Monte Carlo code for re.pdf:application/pdf;ScienceDirect Full Text PDF:C\:\\Users\\Sun Myung\\Zotero\\storage\\IW54R5QX\\Romano et al. - 2015 - OpenMC A state-of-the-art Monte Carlo code for re.pdf:application/pdf;ScienceDirect Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\NZ4279HS\\S030645491400379X.html:text/html;ScienceDirect Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\ALT6V9G9\\S030645491400379X.html:text/html},
}

@article{serp_molten_2014,
	title = {The molten salt reactor ({MSR}) in generation {IV}: {Overview} and perspectives},
	volume = {77},
	issn = {0149-1970},
	shorttitle = {The molten salt reactor ({MSR}) in generation {IV}},
	doi = {10.1016/j.pnucene.2014.02.014},
	abstract = {Molten Salt Reactors (MSR) with the fuel dissolved in the liquid salt and fluoride-salt-cooled High-temperature Reactors (FHR) have many research themes in common. This paper gives an overview of the international R\&D efforts on these reactor types carried out in the framework of Generation-IV. Many countries worldwide contribute to this reactor technology, among which the European Union, France, Japan, Russia and the USA, and for the past few years China and India have also contributed. In general, the international R\&D focuses on three main lines of research. The USA focuses on the FHR, which will be a nearer-term application of liquid salt as a reactor coolant, while China also focuses on solid fuel reactors as a precursor to molten salt reactors with liquid fuel and a thermal neutron spectrum. The EU, France and Russia are focusing on the development of a fast spectrum molten salt reactor capable of either breeding or transmutation of actinides from spent nuclear fuel. Future research topics focus on liquid salt technology and materials behavior, the fuel and fuel cycle chemistry and modeling, and the numerical simulation and safety design aspects of the reactor. MSR development attracts more and more attention every year, because it is generally considered as most sustainable of the six Generation-IV designs with intrinsic safety features. Continuing joint efforts are needed to advance common molten salt reactor technologies.},
	number = {Supplement C},
	journal = {Progress in Nuclear Energy},
	author = {Serp, Jerome and Allibert, Michel and Benes, Ondrej and Delpech, Sylvie and Feynberg, Olga and Ghetta, Véronique and Heuer, Daniel and Holcomb, David and Ignatiev, Victor and Kloosterman, Jan Leen and Luzzi, Lelio and Merle-Lucotte, Elsa and Uhlíř, Jan and Yoshioka, Ritsuo and Zhimin, Dai},
	month = nov,
	year = {2014},
	keywords = {Molten salt reactor, Fuel cycle, Gen IV, Neutronic performance, Nuclear systems},
	pages = {308--319},
	file = {ScienceDirect Full Text PDF:C\:\\Users\\Sun Myung\\Zotero\\storage\\WWISZQDL\\Serp et al. - 2014 - The molten salt reactor (MSR) in generation IV Ov.pdf:application/pdf;ScienceDirect Full Text PDF:C\:\\Users\\Sun Myung\\Zotero\\storage\\Q5UID2I9\\Serp et al. - 2014 - The molten salt reactor (MSR) in generation IV Ov.pdf:application/pdf;ScienceDirect Full Text PDF:C\:\\Users\\Sun Myung\\Zotero\\storage\\Q5WDR3ER\\Serp et al. - 2014 - The molten salt reactor (MSR) in generation IV Ov.pdf:application/pdf;ScienceDirect Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\QGLMLNIW\\S0149197014000456.html:text/html;ScienceDirect Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\FABGBG8S\\S0149197014000456.html:text/html;ScienceDirect Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\ZRQWCMGZ\\S0149197014000456.html:text/html;ScienceDirect Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\DLCDIEEP\\S0149197014000456.html:text/html},
}

@article{rosenthal_molten-salt_1970,
	title = {Molten-{Salt} {Reactors} - {History}, {Status}, and {Potential}},
	volume = {8},
	issn = {0550-3043},
	doi = {10.13182/NT70-A28619},
	abstract = {Molten-salt breeder reactors (MSBR’s) are being developed by the Oak Ridge National Laboratory for generating low-cost power while extending the nation’s resources of fissionable fuel. The fluid fuel in these reactors, consisting of UF4 and ThF4 dissolved in fluorides of beryllium and lithium, is circulated through a reactor core moderated by graphite. Technology developments over the past 20 years have culminated in the successful operation of the 8-MW(th) MoltenSalt Reactor Experiment (MSRE), and have indicated that operation with a molten fuel is practical, that the salt is stable under reactor conditions, and that corrosion is very low. Processing of the MSRE fuel has demonstrated the MSR processing associated with high-performance converters. New fuel processing methods under development should permit MSR’s to operate as economical breeders. These features, combined with high thermal efficiency (44\%) and low primary system pressure, give MSR converters and breeders potentially favorable economic, fuel utilization, and safety characteristics. Further, these reactors can be initially fueled with 233U, 235U, or plutonium. The construction cost of an MSBR power plant is estimated to be about the same as that of light-water reactors. This could lend to power costs 0.5 to 1.0 mill/kWh less than those for light-water reactors. Achievement of economic molten-salt breeder reactors requires the construction and operation of several reactors of increasing size and their associated processing plants.},
	number = {2},
	journal = {Nuclear Applications and Technology},
	author = {Rosenthal, M. W. and Kasten, P. R. and Briggs, R. B.},
	month = feb,
	year = {1970},
	pages = {107--117},
	file = {Full Text PDF:C\:\\Users\\Sun Myung\\Zotero\\storage\\SJ8BSHG9\\Rosenthal et al. - 1970 - Molten-Salt Reactors—History, Status, and Potentia.pdf:application/pdf;Full Text PDF:C\:\\Users\\Sun Myung\\Zotero\\storage\\SSYKDSVM\\Rosenthal et al. - 1970 - Molten-Salt Reactors—History, Status, and Potentia.pdf:application/pdf;Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\7NZ5QDPF\\NT70-A28619.html:text/html;Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\WBG97VAY\\NT70-A28619.html:text/html},
}

@article{elsheikh_safety_2013,
	title = {Safety assessment of molten salt reactors in comparison with light water reactors},
	volume = {6},
	issn = {1687-8507},
	doi = {10.1016/j.jrras.2013.10.008},
	abstract = {Molten salt reactors (MSRs) have a long history with the first design studies beginning in the 1950s at the Oak Ridge National Laboratory (ORNL). Traditionally these reactors are thought of as thermal breeder reactors running on the thorium to 233U cycle and the historical competitor to fast breeder reactors. In the recent years, there has been a growing interest in molten salt reactors, which have been considered in the framework of the Generation IV International Forum, because of their several potentialities and favorable features when compared with conventional solid-fueled reactors. MSRs meet many of the future goals of nuclear energy, in particular for what concerns an improved sustainability, an inherent safety with strong negative temperature coefficient of reactivity, stable coolant, low pressure operation that don not require expensive containment, easy to control, passive decay heat cooling and unique characteristics in terms of actinide burning and waste reduction, while benefiting from the past experience acquired with the molten salt technology. As the only liquid-fueled reactor concept, the safety basis, characteristics and licensing of an MSR are different from solid-uranium fueled light water reactors. In this paper, a historical review of the major plant systems in MSR is presented. The features of different safety characteristics of MSR power plant are reviewed and assessment in comparison to other solid fueled light water reactors LWRs.},
	number = {2},
	journal = {Journal of Radiation Research and Applied Sciences},
	author = {Elsheikh, Badawy M.},
	month = oct,
	year = {2013},
	keywords = {LWR safety, Molten salt reactor safety, Nuclear reactor accident, Nuclear safety},
	pages = {63--70},
	file = {ScienceDirect Full Text PDF:C\:\\Users\\Sun Myung\\Zotero\\storage\\R4PU37EJ\\Elsheikh - 2013 - Safety assessment of molten salt reactors in compa.pdf:application/pdf;ScienceDirect Full Text PDF:C\:\\Users\\Sun Myung\\Zotero\\storage\\FCI6DLR8\\Elsheikh - 2013 - Safety assessment of molten salt reactors in compa.pdf:application/pdf;ScienceDirect Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\98N9YQRW\\S1687850713000101.html:text/html;ScienceDirect Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\7SMEXTYL\\S1687850713000101.html:text/html},
}

@article{gehin_liquid_2016,
	title = {Liquid {Fuel} {Molten} {Salt} {Reactors} for {Thorium} {Utilization}},
	volume = {194},
	issn = {00295450},
	doi = {10.13182/NT15-124},
	abstract = {Molten salt reactors (MSRs) represent a class of reactors that use liquid salt, usually fluoride based or chloride based, as either a coolant with a solid fuel (such as fluoride salt–cooled high-temperature reactors) or as a combined coolant and fuel with the fuel dissolved in a carrier salt. For liquid-fueled MSRs, the salt can be processed online or in a batch mode to allow for removal of fission products as well as for introduction of fissile fuel and fertile materials during reactor operation. The MSR is most commonly associated with the 233U/thorium fuel cycle, as the nuclear properties of 233U combined with the online removal of parasitic absorbers enable the design of a thermal-spectrum breeder reactor. However, MSR concepts have been developed using all neutron energy spectra (thermal, intermediate, fast, and mixed-spectrum zoned concepts) and with a variety of fuels including uranium, thorium, plutonium, and minor actinides. Early MSR work was supported by a significant research and development (R\&D) program that resulted in two experimental systems operating at Oak Ridge National Laboratory in the 1950s and 1960s: the Aircraft Reactor Experiment and the Molten Salt Reactor Experiment. Subsequent design studies in the 1970s focusing on thermal-spectrum thorium-fueled systems established reference concepts for two major design variants: (1) a molten salt breeder reactor (MSBR) with multiple configurations that could breed additional fissile material or maintain self-sustaining operation and (2) a denatured molten salt reactor (DMSR) with enhanced proliferation resistance. MSRs have been selected as one of the Generation IV systems, and development activity has been seen in fast-spectrum MSRs, waste-burning MSRs, and MSRs fueled with low-enriched uranium as well as in more traditional thorium fuel cycle–based MSRs. This paper provides a historical background of MSR R\&D efforts, surveys and summarizes many of the recent developments, and provides analysis comparing thorium-based MSRs by way of example.},
	number = {2},
	journal = {Nuclear Technology},
	author = {Gehin, Jess C. and Powers, Jeffery J.},
	month = may,
	year = {2016},
	pages = {152--161},
	file = {Full Text PDF:C\:\\Users\\Sun Myung\\Zotero\\storage\\U5KHRRFR\\Gehin and Powers - 2016 - Liquid Fuel Molten Salt Reactors for Thorium Utili.pdf:application/pdf;Full Text PDF:C\:\\Users\\Sun Myung\\Zotero\\storage\\V3GV9DHE\\Gehin and Powers - 2016 - Liquid Fuel Molten Salt Reactors for Thorium Utili.pdf:application/pdf;Fulltext:C\:\\Users\\Sun Myung\\Zotero\\storage\\LT8KVIIL\\scholar.html:text/html;Liquid Fuel Molten Salt Reactors for Thorium Utilization.pdf:C\:\\Users\\Sun Myung\\Zotero\\storage\\NG377NQN\\Liquid Fuel Molten Salt Reactors for Thorium Utilization.pdf:application/pdf;Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\SKGAV2GK\\NT15-124.html:text/html;Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\N6N3FSCJ\\NT15-124.html:text/html;Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\986LM99R\\NT15-124.html:text/html},
}

@techreport{satish_petsc_2019,
	title = {{PETSc} {Users} {Manual}},
	url = {https://www.mcs.anl.gov/petsc},
	number = {ANL-95/11 - Revision 3.12},
	institution = {Argonne National Laboratory},
	author = {Satish, Balay and Shrirang, Abhyankar and Adams, Mark F. and Brown, Jed and Brune, Peter and Buschelman, Kris and Dalcin, Lisandro and Dener, Alp and Eijkhout, Victor and Gropp, William D. and Karpeyev, Dmitry and Kaushik, Dinesh and Knepley, Matthew G. and May, Dave A. and McInnes, Lois Curfman and Mills, Richard Tran and Munson, Todd and Rupp, Karl and Sanan, Patrick and Smith, Barry F. and Zampini, Stefano and Zhang, Hong},
	year = {2019},
	file = {manual.pdf:C\:\\Users\\Sun Myung\\Zotero\\storage\\4I5DK7WE\\manual.pdf:application/pdf},
}

@techreport{gif_technology_2014,
	type = {Generation {IV} {International} {Forum}},
	title = {Technology {Roadmap} {Update} for {Generation} {IV} {Nuclear} {Energy} {Systems}},
	shorttitle = {{GIF} {Technology} {Roadmap}},
	url = {https://www.gen-4.org/gif/upload/docs/application/pdf/2014-03/gif-tru2014.pdf},
	language = {en},
	number = {January 2014 Update},
	institution = {OECD Nuclear Energy Agency},
	author = {GIF},
	month = jan,
	year = {2014},
	pages = {66},
	file = {Technology Roadmap Update for Generation IV Nuclea.pdf:C\:\\Users\\Sun Myung\\Zotero\\storage\\9LRH9PYP\\Technology Roadmap Update for Generation IV Nuclea.pdf:application/pdf},
}

@misc{merle_concept_2017,
	title = {Concept of {Molten} {Salt} {Fast} {Reactor}},
	language = {en},
	author = {Merle, Elsa},
	year = {2017},
	file = {Merle - 2017 - Concept of Molten Salt Fast Reactor.pdf:C\:\\Users\\Sun Myung\\Zotero\\storage\\DZP6CNWB\\Merle - 2017 - Concept of Molten Salt Fast Reactor.pdf:application/pdf},
}

@techreport{kedl_fluid_1970,
	title = {Fluid dynamic studies of the {Molten}-{Salt} {Reactor} {Experiment} ({MSRE}) core},
	url = {http://www.osti.gov/servlets/purl/4080377/},
	abstract = {In the MSRE reactor vessel, fluid fuel is circulated at 1200 gpm down through an annular region and up through 1140 passages in the graphite core. The core design was based on preliminary tests in a one-fifth scale model, followed by detailed measurements with water solutions in a full-scale mockup of the reactor vessel and internals. This report describes the models, the testing, and the data from which velocity, pressure drop and flow patterns are deduced. It also describes how the measurements were extrapolated to molten salt at 1200'F in the actual reactor. The few observations possible in the reactor were consistent with the predicted behavior.},
	language = {en},
	number = {ORNL-TM--3229, 4080377},
	institution = {Oak Ridge National Lab.},
	author = {Kedl, R. J.},
	year = {1970},
	pages = {ORNL--TM--3229, 4080377},
	file = {Fulltext:C\:\\Users\\Sun Myung\\Zotero\\storage\\RMW8RPRA\\Kedl - 1970 - Fluid dynamic studies of the Molten-Salt Reactor E.pdf:application/pdf;Kedl - 1970 - FLUID DYNAMIC STUDIES OF THE MOLTEN-SALT REACTOR E.pdf:C\:\\Users\\Sun Myung\\Zotero\\storage\\R5WZZLLV\\Kedl - 1970 - FLUID DYNAMIC STUDIES OF THE MOLTEN-SALT REACTOR E.pdf:application/pdf;Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\W946QWRZ\\Kedl - 1970 - Fluid dynamic studies of the Molten-Salt Reactor E.pdf:application/pdf},
}

@misc{noauthor_trelis_2018,
	address = {American Fork, UT},
	title = {Trelis ({Version} 16.5)},
	url = {http://csimsoft.com},
	publisher = {csimsoft},
	year = {2018},
}

@article{leppanen_serpent_2014,
	title = {The {Serpent} {Monte} {Carlo} code: {Status}, development and applications in 2013},
	volume = {82},
	issn = {0306-4549},
	shorttitle = {The {Serpent} {Monte} {Carlo} code},
	doi = {10.1016/j.anucene.2014.08.024},
	abstract = {The Serpent Monte Carlo reactor physics burnup calculation code has been developed at VTT Technical Research Centre of Finland since 2004, and is currently used in over 100 universities and research organizations around the world. This paper presents the brief history of the project, together with the currently available methods and capabilities and plans for future work. Typical user applications are introduced in the form of a summary review on Serpent-related publications over the past few years.},
	journal = {Annals of Nuclear Energy},
	author = {Leppanen, Jaakko and Pusa, Maria and Viitanen, Tuomas and Valtavirta, Ville and Kaltiaisenaho, Toni},
	month = aug,
	year = {2014},
	keywords = {Burnup calculation, Homogenization, Monte Carlo, Reactor physics},
	pages = {142--150},
	file = {ScienceDirect Full Text PDF:C\:\\Users\\Sun Myung\\Zotero\\storage\\83V2KXJ9\\Leppänen et al. - 2015 - The Serpent Monte Carlo code Status, development .pdf:application/pdf;ScienceDirect Full Text PDF:C\:\\Users\\Sun Myung\\Zotero\\storage\\VT4QRTXR\\Leppänen et al. - 2015 - The Serpent Monte Carlo code Status, development .pdf:application/pdf;ScienceDirect Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\AQAB5875\\S0306454914004095.html:text/html;ScienceDirect Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\9AVEULJX\\S0306454914004095.html:text/html},
}

@inproceedings{zhang_couple_2014,
	title = {{COUPLE}, {A} {Time}-{Dependent} {Coupled} {Neutronics} and {Thermal}-{Hydraulics} {Code}, and its {Application} to {MSFR}},
	doi = {10.1115/ICONE22-30609},
	language = {en},
	publisher = {American Society of Mechanical Engineers Digital Collection},
	author = {Zhang, Dalin and Zhai, Zhi-Gang and Rineiski, Andrei and Guo, Zhangpeng and Wang, Chenglong and Xiao, Yao and Qiu, Suizheng},
	month = nov,
	year = {2014},
	file = {Full Text PDF:C\:\\Users\\Sun Myung\\Zotero\\storage\\VUK7EPRC\\Zhang et al. - 2014 - COUPLE, A Time-Dependent Coupled Neutronics and Th.pdf:application/pdf;Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\PDPPHN33\\255491.html:text/html},
}

@techreport{robertson_msre_1965,
	title = {{MSRE} {DESIGN} {AND} {OPERATIONS} {REPORT}. {PART} {I}. {DESCRIPTION} {OF} {REACTOR} {DESIGN}},
	url = {https://www.osti.gov/biblio/4654707},
	abstract = {The U.S. Department of Energy's Office of Scientific and Technical Information},
	language = {English},
	number = {ORNL-TM-728},
	institution = {Oak Ridge National Lab., Tenn.},
	author = {Robertson, R. C.},
	month = jan,
	year = {1965},
	file = {Robertson - MSRE DESIGN AND OpERclTIONS REPORT PART I DESCRIPT.pdf:C\:\\Users\\Sun Myung\\Zotero\\storage\\CEDRFF7D\\Robertson - MSRE DESIGN AND OpERclTIONS REPORT PART I DESCRIPT.pdf:application/pdf;Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\TYHQ8QDQ\\4654707.html:text/html;Submitted Version:C\:\\Users\\Sun Myung\\Zotero\\storage\\F6XQCCD3\\Robertson - 1965 - MSRE DESIGN AND OPERATIONS REPORT. PART I. DESCRIP.pdf:application/pdf},
}

@article{tiberga_results_2020,
	title = {Results from a multi-physics numerical benchmark for codes dedicated to molten salt fast reactors},
	volume = {142},
	issn = {0306-4549},
	doi = {10.1016/j.anucene.2020.107428},
	abstract = {Verification and validation of multi-physics codes dedicated to fast-spectrum molten salt reactors (MSR) is a very challenging task. Existing benchmarks are meant for single-physics codes, while experimental data for validation are absent. This is concerning, given the importance numerical simulations have in the development of fast MSR designs. Here, we propose the use of a coupled numerical benchmark specifically designed to assess the physics-coupling capabilities of the aforementioned codes. The benchmark focuses on the specific characteristics of fast MSRs and features a step-by-step approach, where physical phenomena are gradually coupled to easily identify sources of error. We collect and compare the results obtained during the benchmarking campaign of four multi-physics tools developed within the SAMOFAR project. Results show excellent agreement for all the steps of the benchmark. The benchmark generality and the broad spectrum of results provided constitute a useful tool for the testing and development of similar multi-physics codes.},
	language = {en},
	journal = {Annals of Nuclear Energy},
	author = {Tiberga, Marco and de Oliveira, Rodrigo Gonzalez Gonzaga and Cervi, Eric and Blanco, Juan Antonio and Lorenzi, Stefano and Aufiero, Manuele and Lathouwers, Danny and Rubiolo, Pablo},
	month = jul,
	year = {2020},
	keywords = {Molten salt reactor, Benchmark, Neutronics, Thermal-hydraulics, Multi-physics, Code-to-code comparison, Fast-spectrum},
	pages = {107428},
	file = {ScienceDirect Full Text PDF:C\:\\Users\\Sun Myung\\Zotero\\storage\\7AYW94UF\\Tiberga et al. - 2020 - Results from a multi-physics numerical benchmark f.pdf:application/pdf;ScienceDirect Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\4NR9AK9R\\S0306454920301262.html:text/html},
}

@article{fiorina_modelling_2014,
	title = {Modelling and analysis of the {MSFR} transient behaviour},
	volume = {64},
	issn = {0306-4549},
	doi = {10.1016/j.anucene.2013.08.003},
	abstract = {Molten Salt Reactors (MSRs) were conceived at the early stages of nuclear energy in view of the favourable features fostered by a liquid fuel. They were developed as graphite-moderated thorium-fuelled breeder reactors till the seventies, when studies on this reactor concept were mostly abandoned in favour of the liquid–metal fast breeder concepts. A decade ago, the MSRs were included among the six GEN-IV systems and a core optimization process allowing for the GEN-IV main objectives led toward a fast-spectrum MSR concept (MSFR – Molten Salt Fast Reactor). Albeit advantageous in terms of U-233 breeding and/or radio-active waste burning, the new concept lacks the notable know-how available for the thermal-spectrum MSR technology. The present paper preliminarily investigates the MSR dynamics, based on the conceptual MSFR design currently available. A primary objective is to benchmark two different models of the MSFR primary circuit, both of them including a detailed and fully-coupled (node-wise) representation of turbulent fuel-salt flow, neutron diffusion, and delayed-neutron precursor diffusion and convection. A good agreement is generally observed between the adopted models, though some discrepancies exist in the temperature-field, with ensuing mild impacts on the reactor dynamics. The performed analyses are also used for a preliminary characterization of the MSFR steady-state and accidental transient response. Some points of enhancement needed in the MSFR conceptual design are identified, mainly related to in-core velocity and temperature fields. The reactor response following major accidental transient initiators suggests a generally benign behaviour of this reactor concept.},
	number = {Supplement C},
	journal = {Annals of Nuclear Energy},
	author = {Fiorina, Carlo and Lathouwers, Danny and Aufiero, Manuele and Cammi, Antonio and Guerrieri, Claudia and Kloosterman, Jan Leen and Luzzi, Lelio and Ricotti, Marco Enrico},
	month = feb,
	year = {2014},
	keywords = {Molten Salt Reactor (MSR), Molten Salt Fast Reactor (MSFR), Safety, Dynamics},
	pages = {485--498},
	file = {ScienceDirect Full Text PDF:C\:\\Users\\Sun Myung\\Zotero\\storage\\IXFXRASN\\Fiorina et al. - 2014 - Modelling and analysis of the MSFR transient behav.pdf:application/pdf;ScienceDirect Full Text PDF:C\:\\Users\\Sun Myung\\Zotero\\storage\\XPVTBLQW\\Fiorina et al. - 2014 - Modelling and analysis of the MSFR transient behav.pdf:application/pdf;ScienceDirect Full Text PDF:C\:\\Users\\Sun Myung\\Zotero\\storage\\8I8JU2JZ\\Fiorina et al. - 2014 - Modelling and analysis of the MSFR transient behav.pdf:application/pdf;ScienceDirect Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\9WNRAF3E\\S0306454913004118.html:text/html;ScienceDirect Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\U2ZUJ8KI\\S0306454913004118.html:text/html;ScienceDirect Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\275TZZMW\\S0306454913004118.html:text/html},
}

@article{aufiero_development_2014,
	title = {Development of an {OpenFOAM} model for the {Molten} {Salt} {Fast} {Reactor} transient analysis},
	volume = {111},
	issn = {0009-2509},
	doi = {10.1016/j.ces.2014.03.003},
	abstract = {In the paper, the development of a multiphysics model for the transient analysis of non-moderated Molten Salt Reactors is discussed. Particular attention is devoted to the description of the adopted time integration and physics coupling strategies. The proposed model features the adoption of an implicit Runge–Kutta scheme and the coupling among neutron diffusion, Reynolds-Averaged Navier–Stokes equations for mass and momentum conservation, and energy and delayed neutron precursor balance equations, in order to accurately catch thermal feedbacks on neutronics. The solver is aimed at performing fast-running simulations of the full-core three-dimensional Molten Salt Fast Reactor geometry. The neutronics modelling is assessed against Monte Carlo simulations and the results of a simplified case study are compared to those from multiphysics tools previously developed. As an example of the capability of the model, an unprotected MSFR single pump failure accidental scenario is simulated and discussed. The main purpose of the present model is to serve as fast-running computational tool in the phase of design optimization of fuel loop components. More in general, it is of valuable help in the study of reactor physics of circulating-fuel systems.},
	journal = {Chemical Engineering Science},
	author = {Aufiero, Manuele and Cammi, Antonio and Geoffroy, Olivier and Losa, Mario and Luzzi, Lelio and Ricotti, Marco E. and Rouch, Hervé},
	month = may,
	year = {2014},
	keywords = {Multiphysics, Reactor dynamics, Molten Salt Reactor (MSR), Molten Salt Fast Reactor (MSFR), OpenFOAM},
	pages = {390--401},
	file = {ScienceDirect Full Text PDF:C\:\\Users\\Sun Myung\\Zotero\\storage\\GFZ6CW4E\\Aufiero et al. - 2014 - Development of an OpenFOAM model for the Molten Sa.pdf:application/pdf;ScienceDirect Full Text PDF:C\:\\Users\\Sun Myung\\Zotero\\storage\\NI9HRTWS\\Aufiero et al. - 2014 - Development of an OpenFOAM model for the Molten Sa.pdf:application/pdf;ScienceDirect Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\TGWQHMHK\\S0009250914001146.html:text/html;ScienceDirect Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\L4BZULIM\\S0009250914001146.html:text/html;ScienceDirect Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\YJ3GUNQ6\\S0009250914001146.html:text/html},
}

@inproceedings{merle_optimized_2007,
	address = {Nice, France},
	title = {Optimized transition from the reactors of second and third generations to the {Thorium} {Molten} {Salt} {Reactor}},
	abstract = {Molten salt reactors, in the configuration presented here and called Thorium Molten Salt Reactor (TMSR), are particularly well suited to fulfil the criteria chosen by the Generation IV forum, and may be operated in simplified and safe conditions in the Th/233U fuel cycle with fluoride salts. Amongst all MSR configurations in the thorium cycle, many studies have highlighted the configurations with no moderator in the core as simple and very promising. Since 233U does not exist on earth and is not being produced today, we aim at designing a critical MSR able to burn the Plutonium and the Minor Actinides produced in the current operating reactors, and consequently to convert this Plutonium into 233U. This leads to closing the current fuel cycle thanks to TMSRs started with transuranic elements on a Thorium base, i.e. started in the Th/Pu fuel cycle, similarly to fast neutron reactors operated in the U/Pu fuel cycle. We will detail optimizations of this transition between the reactors of second and third generations to the Thorium cycle. Such a transition is based on a fleet of TMSRs with no moderator in the core, including TMSRs started with Plutonium and TMSRs directly started with 233U. We developed parametric studies to optimize these TMSRs, amongst which the study presented here, based on one of the main TMSR parameters: the percentage of heavy nuclei in the fuel salt of the TMSR configuration, which modifies the moderation ratio of the reactor and thus influences both the initial fissile inventory and the spectrum of the reactor. We analyze the characteristics of each reactorconfiguration, in terms of deterministic safety parameters, fissile matter inventory, salt reprocessing, radiotoxicity and waste production, and finally deployment capacities.},
	booktitle = {{ICAPP} 2007},
	author = {Merle, Elsa and Heuer, Daniel and Allibert, M. and Ghetta, Veronique and Brun, C. and Mathieu, Ludovic and Brissot, R. and Liatard, E.},
	month = may,
	year = {2007},
	file = {Full Text PDF:C\:\\Users\\Sun Myung\\Zotero\\storage\\NHDHPHZR\\Merle et al. - 2007 - Optimized transition from the reactors of second a.pdf:application/pdf},
}

@article{aji_experimental_2020,
	title = {An {Experimental} and {Numerical} {Study} of {Wall} {Effect} on {Freeze} {Valve} {Performance} in a {Molten} {Salt} {Reactor}},
	volume = {6},
	issn = {2332-8983},
	doi = {10.1115/1.4045180},
	language = {en},
	number = {2},
	journal = {Journal of Nuclear Engineering and Radiation Science},
	author = {Aji, Indarta Kuncoro and Tatsuya, Tokushima and Kinoshita, Motoyasu and Okawa, Tomio},
	month = apr,
	year = {2020},
	file = {Full Text PDF:C\:\\Users\\Sun Myung\\Zotero\\storage\\RE75F4ZC\\Aji et al. - 2020 - An Experimental and Numerical Study of Wall Effect.pdf:application/pdf;Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\AKVDJVFM\\An-Experimental-and-Numerical-Study-of-Wall-Effect.html:text/html},
}

@article{brovchenko_design-related_2013,
	title = {Design-{Related} {Studies} for the {Preliminary} {Safety} {Assessment} of the {Molten} {Salt} {Fast} {Reactor}},
	volume = {175},
	doi = {10.13182/NSE12-70},
	abstract = {Salt reactors are liquid fuel reactors so that they are flexible in operation, but they are very different from solid fuel reactors in the approach to safety. This study concentrates on the specific concept named Molten Salt Fast Reactor (MSFR). Since this new nuclear technology is in development, safety is an essential point to be considered all along the research and development studies. After a short description of the MSFR systems, necessary to device accidental scenarios, this paper will focus on the decay heat evaluation of such a reactor. Among different contributions, the decay heat of fission products in the MSFR is evaluated to be low (3\% of nominal power), mainly due to the reprocessing during the reactor operation. As a result, the contribution of the actinides is significant (0.5\% of nominal power). However, the decay heat of the fission products is important, and among the different uncertainty sources, the fission yield uncertainties are pointed out. The unprotected loss of heat sink transients are studied in this paper. It appears that slow transients are favorable ( {\textgreater}1 min) to minimize the temperature increase of the fuel salt. This work will be the basis of further safety studies as well as an essential parameter for the design of the draining system.},
	journal = {Nuclear Science and Engineering},
	author = {Brovchenko, Mariya and Heuer, Daniel and Merle, Elsa and Allibert, M. and Ghetta, Veronique and Laureau, Axel and Rubiolo, P.},
	month = nov,
	year = {2013},
	pages = {329--339},
	file = {Full Text PDF:C\:\\Users\\Sun Myung\\Zotero\\storage\\ZGLHTVHA\\Brovchenko et al. - 2013 - Design-Related Studies for the Preliminary Safety .pdf:application/pdf},
}

@article{delpech_reactor_2009,
	series = {Fluorine \& {Nuclear} {Energy}},
	title = {Reactor physic and reprocessing scheme for innovative molten salt reactor system},
	volume = {130},
	issn = {0022-1139},
	doi = {10.1016/j.jfluchem.2008.07.009},
	abstract = {The molten salt reactor is one of the six concepts retained by the Generation IV forum in 2001. Based on the MSRE and MSBR concepts developed by ORNL in the 60s which involve a liquid fuel constituted of fluorine molten salt at a temperature close to 600°C, new developments with innovative approach and technology have been realized which contribute to strongly improve the concept. The thorium breeder potentiality is closely related to the use of a liquid fuel which is able to be periodically treated. A reprocessing scheme has been established to treat used fuel by extraction of fission products. According to the Gen IV philosophy for closed cycle nuclear reactor, the actinides are sent back in the reactor core. In this way, the wastes radiotoxicity is strongly decreased and the use of natural resource is optimized. This paper describes an innovative reactor concept, the TMSR-NM (non-moderated thorium molten salt reactor), from the nuclear physic point of view and the different steps involving in the reprocessing scheme from the chemical point of view.},
	language = {en},
	number = {1},
	journal = {Journal of Fluorine Chemistry - J FLUORINE CHEM},
	author = {Delpech, S. and Merle-Lucotte, E. and Heuer, D. and Allibert, M. and Ghetta, V. and Le-Brun, C. and Doligez, X. and Picard, G.},
	month = jan,
	year = {2009},
	keywords = {Molten salt reactor, Thorium fuel cycle, Electrochemistry, Reactor physic, Pyrochemistry},
	pages = {11--17},
	file = {Delpech et al. - 2009 - Reactor physic and reprocessing scheme for innovat.pdf:C\:\\Users\\Sun Myung\\Zotero\\storage\\LT55IZ94\\Delpech et al. - 2009 - Reactor physic and reprocessing scheme for innovat.pdf:application/pdf;ScienceDirect Full Text PDF:C\:\\Users\\Sun Myung\\Zotero\\storage\\A7CI75JN\\Delpech et al. - 2009 - Reactor physic and reprocessing scheme for innovat.pdf:application/pdf;ScienceDirect Full Text PDF:C\:\\Users\\Sun Myung\\Zotero\\storage\\B9S54ZKY\\Delpech et al. - 2009 - Reactor physic and reprocessing scheme for innovat.pdf:application/pdf;ScienceDirect Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\Z32IFZE3\\S0022113908002054.html:text/html;ScienceDirect Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\FWJKF4PF\\S0022113908002054.html:text/html},
}

@article{krepel_dyn3d-msr_2007,
	title = {{DYN3D}-{MSR} spatial dynamics code for molten salt reactors},
	volume = {34},
	issn = {0306-4549},
	doi = {10.1016/j.anucene.2006.12.011},
	abstract = {The development of spatial dynamics code for molten salt reactors (MSRs) is reported in this paper. The graphite-moderated channel type MSR – one of the ‘Generation IV’ concepts – was selected for the numerical simulation. It has several peculiarities (e.g. the drift of delayed neutrons precursors), which disable the use of standard dynamics codes. Therefore, the own DYN3D-MSR code was developed. It is based on the light water reactor code DYN3D and it allows transients simulation by 3D neutronics and parallel channel thermal-hydraulics. The neutronics and thermal-hydraulics were modified for the MSR peculiarities, where the experience from DYN1D-MSR development was exploited. The code was validated on experimental results from the MSRE experiment done in Oak Ridge National Laboratory and by the comparison with other codes especially with the 1D version. However, by the 3D code transients can be simulated, where space-dependant efforts are relevant, like local blockage of fuel channels or local temperature perturbations.},
	language = {en},
	number = {6},
	journal = {Annals of Nuclear Energy},
	author = {Krepel, Jiří and Rohde, Ulrich and Grundmann, Ulrich and Weiss, Frank-Peter},
	month = jun,
	year = {2007},
	pages = {449--462},
	file = {Křepel et al. - 2007 - DYN3D-MSR spatial dynamics code for molten salt re.pdf:C\:\\Users\\Sun Myung\\Zotero\\storage\\AMVNY7E6\\Křepel et al. - 2007 - DYN3D-MSR spatial dynamics code for molten salt re.pdf:application/pdf;ScienceDirect Full Text PDF:C\:\\Users\\Sun Myung\\Zotero\\storage\\T5JKV5U6\\Křepel et al. - 2007 - DYN3D-MSR spatial dynamics code for molten salt re.pdf:application/pdf;ScienceDirect Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\Q24HJQHD\\S0306454907000527.html:text/html},
}

@article{moir_cost_2002,
	title = {Cost of {Electricity} from {Molten} {Salt} {Reactors}},
	volume = {138},
	issn = {0029-5450},
	doi = {10.13182/NT02-A3281},
	abstract = {The cost of electricity is estimated for a molten salt reactor based on evaluations at the Oak Ridge National Laboratory (ORNL) and compared to the ORNL pressurized water reactor and coal plant estimates of the same pre-1980 vintage plants. The results were 3.8, 4.1, and 4.2 ¢/kWh for the molten salt reactor, pressurized water reactor, and coal. Surprisingly, such cost estimates have never before been published for the molten salt reactor.},
	number = {1},
	journal = {Nuclear Technology},
	author = {Moir, R. W.},
	month = apr,
	year = {2002},
	pages = {93--95},
	file = {Full Text PDF:C\:\\Users\\Sun Myung\\Zotero\\storage\\P44YDAUT\\Moir - 2002 - Cost of Electricity from Molten Salt Reactors.pdf:application/pdf;Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\AGWM8DBI\\NT02-A3281.html:text/html},
}

@techreport{graham_development_2019,
	title = {Development of {Molten} {Salt} {Reactor} {Modeling} and {Simulation} {Capabilities} in {VERA}},
	url = {https://www.osti.gov/biblio/1559657-development-molten-salt-reactor-modeling-simulation-capabilities-vera},
	abstract = {The U.S. Department of Energy's Office of Scientific and Technical Information},
	language = {English},
	institution = {Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)},
	author = {Graham, Aaron (ORCID:0000000232458441) and Collins, Benjamin S. (ORCID:0000000241918868) and Salko Jr, Robert (ORCID:0000000253566679) and Taylor, Robert Z. and Gentry, Cole A. (ORCID:0000000237730707)},
	month = sep,
	year = {2019},
	file = {Full Text PDF:C\:\\Users\\Sun Myung\\Zotero\\storage\\MXBR8DZ9\\Graham et al. - 2019 - Development of Molten Salt Reactor Modeling and Si.pdf:application/pdf;Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\6UQXW5EZ\\1559657-development-molten-salt-reactor-modeling-simulation-capabilities-vera.html:text/html},
}

@article{dittus_heat_1930,
	title = {Heat transfer in automobile radiators of the tubular type},
	volume = {2},
	language = {en},
	number = {13},
	journal = {University of California Publications in Engineering},
	author = {Dittus, F. W. and Boelter, L. M. K.},
	month = oct,
	year = {1930},
	pages = {443--461},
	file = {ScienceDirect Full Text PDF:C\:\\Users\\Sun Myung\\Zotero\\storage\\32CHLSWJ\\Dittus and Boelter - 1985 - Heat transfer in automobile radiators of the tubul.pdf:application/pdf;ScienceDirect Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\NACS27ES\\073519338590003X.html:text/html},
}

@techreport{u.s._doe_nuclear_energy_research_advisory_committee_technology_2002,
	title = {A technology roadmap for generation {IV} nuclear energy systems},
	url = {https://www.gen-4.org/gif/jcms/c_40481/technology-roadmap},
	number = {GIF-002-00},
	institution = {U.S. DOE Nuclear Energy Research Advisory Committee and the Generation IV International Forum},
	author = {{U.S. DOE Nuclear Energy Research Advisory Committee}},
	month = dec,
	year = {2002},
	file = {genivroadmap2002.pdf:C\:\\Users\\Sun Myung\\Zotero\\storage\\75S9TRA6\\genivroadmap2002.pdf:application/pdf},
}

@misc{nrc_trace_2007,
	title = {{TRACE} {V5}.0 {User}'s {Manual}},
	url = {https://www.nrc.gov/docs/ML1200/ML120060239.pdf},
	urldate = {2020-07-22},
	author = {NRC},
	year = {2007},
	file = {ML120060239.pdf:C\:\\Users\\Sun Myung\\Zotero\\storage\\VSTVM8S7\\ML120060239.pdf:application/pdf},
}

@book{ott_introductory_1985,
	title = {Introductory nuclear reactor dynamics},
	isbn = {978-0-89448-029-4},
	abstract = {This text presents the theory and methods of prediction that are the heart of nuclear reactor safety. Time-dependent reactor behavior is explained in both mathematical and physical terms. This book also explains the logic behind the working formulas and calculational methods for reactor transients and illustrates typical dynamic responses. The classical concept of point kinetics is developed in three steps, with discussion of various solutions to kinetics problems. Each chapter includes homework problems and review questions.},
	language = {en},
	publisher = {American Nuclear Society},
	author = {Ott, Karl Otto and Neuhold, Robert J.},
	month = dec,
	year = {1985},
	keywords = {Technology \& Engineering / Power Resources / Nuclear, Nuclear reactors, Nuclear reactor kinetics, Science / Life Sciences / Cytology},
}

@book{wilcox_turbulence_2006,
	title = {Turbulence {Modeling} for {CFD}},
	isbn = {978-1-928729-08-2},
	language = {en},
	publisher = {DCW Industries},
	author = {Wilcox, David C.},
	year = {2006},
}

@article{moorthi_review_2018,
	title = {A review of sub-channel thermal hydraulic codes for nuclear reactor core and future directions},
	volume = {332},
	issn = {0029-5493},
	doi = {10.1016/j.nucengdes.2018.03.012},
	abstract = {Thermal hydraulic analysis of nuclear reactor core is mainly performed using the sub-channel analysis codes to estimate different thermal hydraulic safety margins. The safety margins and the operating power limits of nuclear reactor core under different conditions of primary system i.e. system pressure, coolant inlet temperature, coolant flow rate and thermal power and its distributions are considered as the key parameters for sub-channel analysis. Considering the complexity of rod bundle geometry, different turbulent scales and due to their limitations of computational resources, performing the full scale computational fluid dynamic (CFD) analysis of nuclear reactor core is a cumbersome and time consuming task. Hence, the thermal hydraulic safety margins of most of the reactors operating in the world are carried out using the sub-channel analysis codes. In these codes, the governing equations of mass, momentum and energy are solved in control volumes which are connected in both radial and axial directions. The flow distributions in the rod bundle geometry are estimated by considering lateral momentum balance and the inter channel mixing models to account for the cross flow between the adjacent sub-channels. The accurate estimations of the local conditions of the sub-channels are required to predict fuel temperature, critical heat flux ratio (CHFR) and critical power ratio (CPR). In this paper, various experiments conducted on the different geometries of sub-channels related inter sub-channel mixing in rod bundles are identified. A comprehensive review of sub-channel thermal hydraulic codes used for the analysis of nuclear reactor core is presented. This review covers various aspects of experimental, analytical and computational works related to rod bundles carried out in the past and brings out future directions derived from earlier research works.},
	language = {en},
	journal = {Nuclear Engineering and Design},
	author = {Moorthi, A. and Kumar Sharma, Anil and Velusamy, K.},
	month = jun,
	year = {2018},
	keywords = {Thermal hydraulics, COBRA, Rod bundle, Sub-channel},
	pages = {329--344},
	file = {ScienceDirect Full Text PDF:C\:\\Users\\Sun Myung\\Zotero\\storage\\LFCRK9G5\\Moorthi et al. - 2018 - A review of sub-channel thermal hydraulic codes fo.pdf:application/pdf;ScienceDirect Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\TLAKYQNQ\\S0029549318302681.html:text/html},
}

@incollection{bartosiewicz_612_2019,
	title = {6.1.2 - {Large}-eddy simulation: {Application} to liquid metal fluid flow and heat transfer},
	isbn = {978-0-08-101980-1},
	shorttitle = {6.1.2 - {Large}-eddy simulation},
	language = {en},
	booktitle = {Thermal {Hydraulics} {Aspects} of {Liquid} {Metal} {Cooled} {Nuclear} {Reactors}},
	publisher = {Woodhead Publishing},
	author = {Bartosiewicz, Y. and Duponcheel, M.},
	editor = {Roelofs, Ferry},
	month = jan,
	year = {2019},
	pages = {245--271},
	file = {ScienceDirect Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\7MXRDHG9\\B978008101980100017X.html:text/html},
}

@misc{us_department_of_commerce_global_2020,
	title = {Global {Monitoring} {Laboratory} - {Carbon} {Cycle} {Greenhouse} {Gases}},
	url = {https://www.esrl.noaa.gov/gmd/ccgg/trends/global.html#global},
	abstract = {The Global Monitoring Laboratory conducts research on greenhouse gas and carbon cycle feedbacks, changes in clouds, aerosols, and surface radiation, and recovery of stratospheric ozone.},
	language = {EN-US},
	urldate = {2020-06-12},
	author = {US Department of Commerce, NOAA},
	year = {2020},
	file = {Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\9BI38IX5\\global.html:text/html},
}

@incollection{mcmichael_global_2004,
	title = {Global {Climate} {Change} ({Chapter} 20)},
	language = {en},
	booktitle = {Comparative {Quantification} of {Health} {Risks}},
	publisher = {World Health Organization},
	author = {McMichael, Anthony J. and Campbell-Lendrum, Diarmid and Kovats, Sari and Edwards, Sally and Wilkinson, Paul and Wilson, Theresa and Nicholls, Robert and Hales, Simon and Tanser, Frank and Le Sueur, David and Schlesinger, Michael and Andronova, Natasha},
	year = {2004},
	pages = {1543--1649},
	file = {Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\26V8H7TG\\climate-change-and-nutrition-archive.html:text/html},
}

@article{laureau_transient_2017,
	title = {Transient coupled calculations of the {Molten} {Salt} {Fast} {Reactor} using the {Transient} {Fission} {Matrix} approach},
	volume = {316},
	issn = {0029-5493},
	doi = {10.1016/j.nucengdes.2017.02.022},
	abstract = {In this paper we present transient studies of the Molten Salt Fast Reactor (MSFR). This generation IV reactor is characterized by a liquid fuel circulating in the core cavity, requiring specific simulation tools. An innovative neutronic approach called “Transient Fission Matrix” is used to perform spatial kinetic calculations with a reduced computational cost through a pre-calculation of the Monte Carlo spatial and temporal response of the system. Coupled to this neutronic approach, the Computational Fluid Dynamics code OpenFOAM is used to model the complex flow pattern in the core. An accurate interpolation model developed to take into account the thermal hydraulics feedback on the neutronics including reactivity and neutron flux variation is presented. Finally different transient studies of the reactor in normal and accidental operating conditions are detailed such as reactivity insertion and load following capacities. The results of these studies illustrate the excellent behavior of the MSFR during such transients.},
	number = {Supplement C},
	journal = {Nuclear Engineering and Design},
	author = {Laureau, A. and Heuer, D. and Merle-Lucotte, E. and Rubiolo, P. R. and Allibert, M. and Aufiero, M.},
	month = may,
	year = {2017},
	keywords = {MSFR, Neutronics, Transient Fission Matrix, Thermal hydraulics, Transient calculation},
	pages = {112--124},
	file = {ScienceDirect Full Text PDF:C\:\\Users\\Sun Myung\\Zotero\\storage\\K5JUZTZ6\\Laureau et al. - 2017 - Transient coupled calculations of the Molten Salt .pdf:application/pdf;ScienceDirect Full Text PDF:C\:\\Users\\Sun Myung\\Zotero\\storage\\GEYY7UJ9\\Laureau et al. - 2017 - Transient coupled calculations of the Molten Salt .pdf:application/pdf;ScienceDirect Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\IKKCKIDB\\S002954931730081X.html:text/html;ScienceDirect Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\5B25JVQI\\S002954931730081X.html:text/html},
}

@mastersthesis{park_advancement_2020,
	address = {Urbana, IL},
	title = {Advancement and {Verification} of {Moltres} for {Molten} {Salt} {Reactor} {Safety} {Analysis}},
	copyright = {Copyright 2020 Sun Myung Park},
	language = {English},
	school = {University of Illinois at Urbana-Champaign},
	author = {Park, Sun Myung},
	month = aug,
	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},
}

@inproceedings{park_safety_2019,
	address = {Seattle, WA, United States},
	title = {Safety {Analysis} of the {Molten} {Salt} {Fast} {Reactor} {Fuel} {Composition} using {Moltres}},
	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.},
	booktitle = {Proceedings of {GLOBAL} {International} {Fuel} {Cycle} {Conference}},
	publisher = {American Nuclear Society},
	author = {Park, Sun Myung and Rykhlevskii, Andrei and Huff, Kathryn},
	month = sep,
	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},
}

@article{petti_future_2018,
	title = {The future of nuclear energy in a carbon-constrained world},
	language = {en},
	journal = {Massachusetts Institute of Technology Energy Initiative (MITEI)},
	author = {Petti, David},
	year = {2018},
	pages = {272},
	file = {The Future of Nuclear Energy in a Carbon-Constrain.pdf:C\:\\Users\\Sun Myung\\Zotero\\storage\\VIFFZS43\\The Future of Nuclear Energy in a Carbon-Constrain.pdf:application/pdf;The Future of Nuclear Energy in a Carbon-Constrain.pdf:C\:\\Users\\Sun Myung\\Zotero\\storage\\R2ITYFBM\\The Future of Nuclear Energy in a Carbon-Constrain.pdf:application/pdf},
}

@article{kirk_libmesh_2006,
	title = {{libMesh} : a {C}++ library for parallel adaptive mesh refinement/coarsening simulations},
	volume = {22},
	issn = {0177-0667, 1435-5663},
	shorttitle = {{libMesh}},
	doi = {10.1007/s00366-006-0049-3},
	abstract = {In this paper we describe the libMesh (http://libmesh.sourceforge.net) framework for parallel adaptive finite element applications. libMesh is an open-source software library that has been developed to facilitate serial and parallel simulation of multiscale, multiphysics applications using adaptive mesh refinement and coarsening strategies. The main software development is being carried out in the CFDLab (http://cfdlab.ae.utexas.edu) at the University of Texas, but as with other open-source software projects; contributions are being made elsewhere in the US and abroad. The main goals of this article are: (1) to provide a basic reference source that describes libMesh and the underlying philosophy and software design approach; (2) to give sufficient detail and references on the adaptive mesh refinement and coarsening (AMR/C) scheme for applications analysts and developers; and (3) to describe the parallel implementation and data structures with supporting discussion of domain decomposition, message passing, and details related to dynamic repartitioning for parallel AMR/C. Other aspects related to C++ programming paradigms, reusability for diverse applications, adaptive modeling, physics-independent error indicators, and similar concepts are briefly discussed. Finally, results from some applications using the library are presented and areas of future research are discussed.},
	language = {en},
	number = {3-4},
	journal = {Engineering with Computers},
	author = {Kirk, Benjamin S. and Peterson, John W. and Stogner, Roy H. and Carey, Graham F.},
	month = dec,
	year = {2006},
	pages = {237--254},
	file = {Full Text PDF:C\:\\Users\\Sun Myung\\Zotero\\storage\\2Z9B88E4\\Kirk et al. - 2006 - libMesh a C++ library for parallel adaptive mesh .pdf:application/pdf;Kirk et al. - 2006 - libMesh  a C++ library for parallel adaptive mesh.pdf:C\:\\Users\\Sun Myung\\Zotero\\storage\\GLYFYTV8\\Kirk et al. - 2006 - libMesh  a C++ library for parallel adaptive mesh.pdf:application/pdf;Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\CJ6ZDS26\\10.html:text/html},
}

@unpublished{diamond_phenomena_2018,
	title = {Phenomena {Important} in {Modeling} and {Simulation} of {Molten} {Salt} {Reactors}.},
	language = {en},
	author = {Diamond, David J. and Brown, Nicholas and Denning, Richard and Bajorek, Stephen},
	month = apr,
	year = {2018},
	file = {Phenomena Important in Modeling and Simulation of .pdf:C\:\\Users\\Sun Myung\\Zotero\\storage\\9QC54KIH\\Phenomena Important in Modeling and Simulation of .pdf:application/pdf},
}

@article{brown_thermal-hydraulic_2020,
	title = {Thermal-{Hydraulic} and {Neutronic} {Phenomena} {Important} in {Modeling} and {Simulation} of {Liquid}-{Fuel} {Molten} {Salt} {Reactors}},
	volume = {206},
	issn = {0029-5450},
	doi = {10.1080/00295450.2019.1590077},
	abstract = {We discuss liquid-fuel molten salt (cooled) reactors (MSRs); how they will operate under normal, transient, and accident conditions; and the results of an expert elicitation to determine the corresponding thermal-hydraulic and neutronic phenomena important to understanding their behavior. Identifying these phenomena will enable the U.S. Nuclear Regulatory Commission (NRC), U.S. Department of Energy, and industry to develop or identify modeling functionalities and tools required to carry out confirmatory and licensing analyses that examine the validity and accuracy of an applicant’s calculations and help determine the margin of safety in plant design. The NRC frequently does an expert elicitation using a Phenomena Identification and Ranking Table (PIRT) to identify and evaluate the state of knowledge of important modeling phenomena. However, few details about the design of these reactors and the sequence of events during accidents are known, so the process used was considered a preliminary PIRT. A panel comprising a group of subject matter experts met to define phenomena that would need to be modeled and considered the impact/importance of each phenomenon with respect to specific figures of merit (FoMs) (e.g., salt temperature, velocity, and composition). Each FoM reflected a potential impact on radionuclide release or loss of a barrier to release. The panel considered what the path forward might be with respect to being able to model the phenomenon in a simulation code. Results are explained for both thermal and fast spectrum designs, with an emphasis on the thermal-hydraulic takeaways.It was concluded that compared to light water reactors, the lack of high-pressure operation, energetic break flow, depressurization, and quench front tracking may simplify some aspects of an MSR analysis. However, MSRs have new phenomena both for a license applicant and NRC confirmatory analysis. There is a need for enhanced understanding of physical properties for MSRs that encompass several individual thermophysical properties, including thermal conductivity, viscosity, specific heat, density, optical properties, thermodynamic properties, volatilities, solubilities, etc. Salt composition is closely linked to both these properties and the neutronics of the system. Additionally, the large number of MSR concepts and system designs means that there is wide variation in the potential modeling needs for these systems.},
	number = {2},
	journal = {Nuclear Technology},
	author = {Brown, Nicholas R. and Diamond, David J. and Bajorek, Stephen and Denning, Richard},
	month = feb,
	year = {2020},
	keywords = {flouride-salt–cooled high-temperature reactors, advanced reactors, molten salt (cooled) reactors, Phenomena Identification and Ranking Table},
	pages = {322--338},
	file = {Full Text PDF:C\:\\Users\\Sun Myung\\Zotero\\storage\\PRGZCZHZ\\Brown et al. - 2020 - Thermal-Hydraulic and Neutronic Phenomena Importan.pdf:application/pdf;Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\6YN78NFY\\00295450.2019.html:text/html},
}

@article{gaston_physics-based_2015,
	series = {Multi-{Physics} {Modelling} of {LWR} {Static} and {Transient} {Behaviour}},
	title = {Physics-based multiscale coupling for full core nuclear reactor simulation},
	volume = {84},
	issn = {0306-4549},
	doi = {10.1016/j.anucene.2014.09.060},
	abstract = {Numerical simulation of nuclear reactors is a key technology in the quest for improvements in efficiency, safety, and reliability of both existing and future reactor designs. Historically, simulation of an entire reactor was accomplished by linking together multiple existing codes that each simulated a subset of the relevant multiphysics phenomena. Recent advances in the MOOSE (Multiphysics Object Oriented Simulation Environment) framework have enabled a new approach: multiple domain-specific applications, all built on the same software framework, are efficiently linked to create a cohesive application. This is accomplished with a flexible coupling capability that allows for a variety of different data exchanges to occur simultaneously on high performance parallel computational hardware. Examples based on the KAIST-3A benchmark core, as well as a simplified Westinghouse AP-1000 configuration, demonstrate the power of this new framework for tackling—in a coupled, multiscale manner—crucial reactor phenomena such as CRUD-induced power shift and fuel shuffle.},
	journal = {Annals of Nuclear Energy},
	author = {Gaston, Derek R. and Permann, Cody J. and Peterson, John W. and Slaughter, Andrew E. and Andrš, David and Wang, Yaqi and Short, Michael P. and Perez, Danielle M. and Tonks, Michael R. and Ortensi, Javier and Zou, Ling and Martineau, Richard C.},
	month = oct,
	year = {2015},
	keywords = {Multiphysics, Full core reactor simulation, Multiphysics coupling},
	pages = {45--54},
	file = {Gaston et al. - 2015 - Physics-based multiscale coupling for full core nu.pdf:C\:\\Users\\Sun Myung\\Zotero\\storage\\K84PKGIK\\Gaston et al. - 2015 - Physics-based multiscale coupling for full core nu.pdf:application/pdf;ScienceDirect Full Text PDF:C\:\\Users\\Sun Myung\\Zotero\\storage\\SGQXWGJV\\Gaston et al. - 2015 - Physics-based multiscale coupling for full core nu.pdf:application/pdf;ScienceDirect Full Text PDF:C\:\\Users\\Sun Myung\\Zotero\\storage\\C4TXYTMN\\Gaston et al. - 2015 - Physics-based multiscale coupling for full core nu.pdf:application/pdf;ScienceDirect Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\PI8FZ93W\\S030645491400543X.html:text/html;ScienceDirect Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\K6WTZCEL\\S030645491400543X.html:text/html},
}

@article{lindsay_introduction_2018,
	title = {Introduction to {Moltres}: {An} application for simulation of {Molten} {Salt} {Reactors}},
	volume = {114},
	issn = {0306-4549},
	shorttitle = {Introduction to {Moltres}},
	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},
	journal = {Annals of Nuclear Energy},
	author = {Lindsay, Alexander and Ridley, Gavin and Rykhlevskii, Andrei and Huff, Kathryn},
	month = apr,
	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},
}

@techreport{huff_identifying_2019,
	address = {Urbana, IL},
	type = {Technical {Report}},
	title = {Identifying {MSR} {Multiphysics} {Modeling} {Challenges}},
	shorttitle = {Contract {DE}-{AC07}-{05ID14517}},
	url = {https://zenodo.org/record/335456},
	abstract = {Solid fueled nuclear reactors differ from liquid fueled reactors with re- spect to many physical domains including neutronics, thermal hydraulics, and materials performance. Accordingly, liquid fueled reactors challenge the capabilities of conventional computational tools designed for modeling and simulating the physics of solid fueled reactors.},
	language = {english},
	number = {UIUC-ARFC-2019-01},
	institution = {University of Illinois at Urbana-Champaign},
	author = {Huff, Kathryn D.},
	collaborator = {Sabharwall, Piyush},
	month = feb,
	year = {2019},
	keywords = {nuclear, molten salt reactor, multiphysics},
	pages = {1--12},
	file = {uiuc-arfc-2019-01.pdf:C\:\\Users\\Sun Myung\\Zotero\\storage\\CSVHID3A\\uiuc-arfc-2019-01.pdf:application/pdf;Zenodo Full Text PDF:C\:\\Users\\Sun Myung\\Zotero\\storage\\RI2BAY9M\\Kathryn Huff - 2019 - Identifying MSR Multiphysics Modeling Challenges.pdf:application/pdf},
}

@article{ragusa_consistent_2009,
	title = {Consistent and accurate schemes for coupled neutronics thermal-hydraulics reactor analysis},
	volume = {239},
	issn = {0029-5493},
	doi = {10.1016/j.nucengdes.2008.11.006},
	abstract = {Conventional coupling paradigms currently used to couple different physics components in reactor analysis problems can be inconsistent in their treatment of the nonlinear terms due to the operator-split (OS) strategies employed. This leads to the usage of small time steps to maintain accuracy requirements, thereby increasing the overall computational time. This paper proposes some remedies to OS techniques that can restore consistency in the coupling of the nonlinear terms and explores high-order mono-block nonlinearly consistent techniques with time step control. The performance of the methods was studied for several transient scenarios using a 0D point-kinetics/thermal-hydraulics lumped model and a 1D neutronics/heat conduction/enthalpy balance model. The results prove that consistent approximations can be made to enhance the overall accuracy in conventional codes with simple nonintrusive techniques. Additionally, an analysis of a mono-block coupling strategy (without having recourse to an OS strategy) is carried out to assess automated time stepping control using higher order Implicit Runge–Kutta (IRK) schemes. The conclusions from these results indicate that nonlinearly consistent adaptive time stepping methods can provide better accuracy and reliability in the solution fields than constant time stepping methods, even for transients with rapid and discontinuous variations.},
	language = {en},
	number = {3},
	journal = {Nuclear Engineering and Design},
	author = {Ragusa, Jean C. and Mahadevan, Vijay S.},
	month = mar,
	year = {2009},
	pages = {566--579},
	file = {Ragusa and Mahadevan - 2009 - Consistent and accurate schemes for coupled neutro.pdf:C\:\\Users\\Sun Myung\\Zotero\\storage\\V9JX8K8S\\Ragusa and Mahadevan - 2009 - Consistent and accurate schemes for coupled neutro.pdf:application/pdf;ragusa_consistent_2009.pdf:C\:\\Users\\Sun Myung\\Zotero\\storage\\JM35MP6U\\ragusa_consistent_2009.pdf:application/pdf;ScienceDirect Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\6MEVWPM4\\login.html:text/html},
}

@mastersthesis{fairhurst-agosta_multi-physics_2020,
	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},
	author = {Fairhurst-Agosta, Roberto},
	month = dec,
	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},
}

@techreport{koning_jeff-31_2006,
	title = {The {JEFF}-3.1 nuclear data library: {JEFF} report 21},
	shorttitle = {The {JEFF}-3.1 nuclear data library},
	language = {en},
	number = {21},
	institution = {OECD Nuclear Energy Agency},
	author = {Koning, Arjan and Forrest, Robin and Kellett, Mark and Mills, Robert and Henriksson, Hans and Rugama, Yolanda},
	year = {2006},
	file = {Koning and OECD Nuclear Energy Agency - 2006 - The JEFF-3.1 nuclear data library JEFF report 21.pdf:C\:\\Users\\Sun Myung\\Zotero\\storage\\KNHM5TBB\\Koning and OECD Nuclear Energy Agency - 2006 - The JEFF-3.1 nuclear data library JEFF report 21.pdf:application/pdf},
}

@misc{office_of_nuclear_energy_advanced_nodate,
	title = {Advanced {Reactor} {Demonstration} {Program}},
	url = {https://www.energy.gov/ne/nuclear-reactor-technologies/advanced-reactor-demonstration-program},
	abstract = {Advanced Reactor Demonstration Program},
	language = {en},
	urldate = {2021-01-28},
	journal = {Energy.gov},
	author = {Office of Nuclear Energy},
	file = {Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\XQBTMHYZ\\advanced-reactor-demonstration-program.html:text/html},
}

@misc{cordis_severe_nodate,
	title = {Severe {Accident} {Modeling} and {Safety} {Assessment} for {Fluid}-fuel {Energy} {Reactors} {\textbar} {SAMOSAFER} {Project} {\textbar} {H2020} {\textbar} {CORDIS} {\textbar} {European} {Commission}},
	url = {https://cordis.europa.eu/project/id/847527},
	urldate = {2021-01-28},
	author = {CORDIS},
	file = {Severe Accident Modeling and Safety Assessment for Fluid-fuel Energy Reactors | SAMOSAFER Project | H2020 | CORDIS | European Commission:C\:\\Users\\Sun Myung\\Zotero\\storage\\5QDPPWYC\\847527.html:text/html},
}

@article{wang_review_2020,
	title = {Review on neutronic/thermal-hydraulic coupling simulation methods for nuclear reactor analysis},
	volume = {137},
	issn = {0306-4549},
	doi = {10.1016/j.anucene.2019.107165},
	abstract = {In an operating nuclear reactor system, various physical phenomena of different properties are intimately linked. These multiphysics phenomena include neutronics (N), thermal-hydraulics (TH), materials science, and other subjects. Among them, the interaction between neutronics and thermal-hydraulics is of great significance in reactor design and safety analysis. In this work, different N/TH coupling methods are reviewed, including loose and tight coupling. For the studies on loose coupling, in which two physical fields are decoupled, the current research status is summarized and classified based on the coupling methods of neutronics and thermal-hydraulics. The studies of tight coupling are introduced based on multiphysics coupling platforms. The investigation shows that the number and objectives of loose coupling studies are more abundant and extensive than those of tight coupling. This indicates that loose coupling strategies are the mainstream coupling solutions in recent research. Furthermore, the solution approaches of N/TH coupling are reviewed with respect to the aspects of performance improvement and application studies, including the operator splitting (OS), Picard iteration, and Jacobian-Free Newton–Krylov (JFNK) methods. A comprehensive study of the solution approaches shows that most of the current loose coupling numerical simulations adopt the Picard iteration method, because it has higher calculation accuracy than the OS method. In contrast to the decoupling approaches such as the OS and Picard iteration methods, the JFNK method updates all physical quantities synchronously, which makes it more accurate. Hence, there are broad application prospects for N/TH tight coupling of the JFNK method.},
	language = {en},
	journal = {Annals of Nuclear Energy},
	author = {Wang, Jincheng and Wang, Qin and Ding, Ming},
	month = mar,
	year = {2020},
	keywords = {JFNK method, Loose coupling method, N/TH coupling method, Operator splitting method, Picard iteration method, Tight coupling method},
	pages = {107165},
	file = {ScienceDirect Full Text PDF:C\:\\Users\\Sun Myung\\Zotero\\storage\\F5VHAFBU\\Wang et al. - 2020 - Review on neutronicthermal-hydraulic coupling sim.pdf:application/pdf;ScienceDirect Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\QKBQWM8B\\S0306454919306759.html:text/html},
}

@mastersthesis{lee_neutronics_2020,
	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},
	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},
	author = {Lee, Alvin J. H.},
	month = dec,
	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},
}

@techreport{terrapower_mcfr_2020,
	title = {{MCFR} {Factsheet}},
	url = {https://www.terrapower.com/wp-content/uploads/2020/08/TP_2020_MCFR_Technology_082020.pdf},
	author = {Terrapower},
	year = {2020},
	file = {TP_2020_MCFR_Technology_082020.pdf:C\:\\Users\\Sun Myung\\Zotero\\storage\\2IIEJR37\\TP_2020_MCFR_Technology_082020.pdf:application/pdf},
}

@article{vijayan_conceptual_2015,
	title = {Conceptual design of {Indian} molten salt breeder reactor},
	volume = {85},
	issn = {0973-7111},
	doi = {10.1007/s12043-015-1070-0},
	abstract = {The third stage of Indian nuclear power programme envisages the use of thorium as the fertile material with 233U, which would be obtained from the operation of Pu/Th based fast reactors in the later part of the second stage. Thorium-based reactors have been designed in many configurations, from light water-cooled designs to high-temperature liquid metal-cooled options. Another option, which holds promise, is the molten salt-fuelled reactor, which can be configured to give significant breeding ratios. A crucial part for achieving reasonable breeding in such reactors is the need to reprocess the salt continuously, either online or in batch mode. India has recently started carrying out fundamental studies so as to arrive at a conceptual design of Indian molten salt breeder reactor (IMSBR). Presently, various design options and possibilities are being studied from the point of view of reactor physics and thermal hydraulic design. In parallel, fundamental studies on natural circulation and corrosion behaviour of various molten salts have also been initiated.},
	language = {en},
	number = {3},
	journal = {Pramana},
	author = {VIJAYAN, P. K. and BASAK, A. and DULERA, I. V. and VAZE, K. K. and BASU, S. and SINHA, R. K.},
	month = sep,
	year = {2015},
	pages = {539--554},
	file = {Springer Full Text PDF:C\:\\Users\\Sun Myung\\Zotero\\storage\\FNJC92AY\\VIJAYAN et al. - 2015 - Conceptual design of Indian molten salt breeder re.pdf:application/pdf},
}

@article{peterson_overview_2018,
	title = {Overview of the {Incompressible} {Navier}-{Stokes} simulation capabilities in the {MOOSE} {Framework}},
	volume = {119},
	doi = {10.1016/j.advengsoft.2018.02.004},
	abstract = {The Multiphysics Object Oriented Simulation Environment (MOOSE) framework is a high-performance, open source, C++ finite element toolkit developed at Idaho National Laboratory. MOOSE was created with the aim of assisting domain scientists and engineers in creating customizable, high-quality tools for multiphysics simulations. While the core MOOSE framework itself does not contain code for simulating any particular physical application, it is distributed with a number of physics "modules" which are tailored to solving e.g. heat conduction, phase field, and solid/fluid mechanics problems. In this report, we describe the basic equations, finite element formulations, software implementation, and regression/verification tests currently available in MOOSE's navier\_stokes module for solving the Incompressible Navier-Stokes (INS) equations.},
	journal = {Advances in Engineering Software},
	author = {Peterson, John and Lindsay, Alexander and Kong, Fande},
	month = may,
	year = {2018},
	keywords = {65N30, Mathematics - Numerical Analysis},
	pages = {68--92},
	file = {arXiv\:1710.08898 PDF:C\:\\Users\\Sun Myung\\Zotero\\storage\\UHKRDGN5\\Peterson et al. - 2017 - Overview of the Incompressible Navier-Stokes simul.pdf:application/pdf;arXiv.org Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\DIYPZSJQ\\1710.html:text/html;Full Text PDF:C\:\\Users\\Sun Myung\\Zotero\\storage\\KXTKFUB8\\Peterson et al. - 2018 - Overview of the Incompressible Navier-Stokes simul.pdf:application/pdf},
}

@article{boyd_multigroup_2019,
	title = {Multigroup {Cross}-{Section} {Generation} with the {OpenMC} {Monte} {Carlo} {Particle} {Transport} {Code}},
	volume = {205},
	issn = {0029-5450},
	doi = {10.1080/00295450.2019.1571828},
	abstract = {High-fidelity deterministic transport codes require highly accurate multigroup cross sections (MGXS). Monte Carlo is increasingly cited as a reactor-agnostic approach to MGXS generation since it is unconstrained by the engineering-based approximations that limit the applicability of deterministic MGXS generation tools. This paper introduces a new framework that uses the OpenMC Monte Carlo code to generate MGXS for use in multigroup transport codes. The openmc.mgxs module is built atop OpenMC’s Python application programming interface to process tally data output by the OpenMC executable. This paper validates the module to generate MGXS that enable the multigroup OpenMOC transport code to compute eigenvalues to within 50 pcm and fission rates to within 1\% of reference solutions for two heterogeneous pressurized water reactor benchmarks.},
	number = {7},
	journal = {Nuclear Technology},
	author = {Boyd, William and Nelson, Adam and Romano, Paul K. and Shaner, Samuel and Forget, Benoit and Smith, Kord},
	month = jul,
	year = {2019},
	keywords = {OpenMC, Monte Carlo neutron transport, multigroup cross-section generation},
	pages = {928--944},
	file = {Boyd et al. - 2019 - Multigroup Cross-Section Generation with the OpenM.pdf:C\:\\Users\\Sun Myung\\Zotero\\storage\\5WND9NK5\\Boyd et al. - 2019 - Multigroup Cross-Section Generation with the OpenM.pdf:application/pdf;Boyd et al. - Multigroup Cross-Section Generation with the OpenM.pdf:C\:\\Users\\Sun Myung\\Zotero\\storage\\D79F5CNH\\Boyd et al. - Multigroup Cross-Section Generation with the OpenM.pdf:application/pdf;Full Text PDF:C\:\\Users\\Sun Myung\\Zotero\\storage\\2JHN2FJD\\Boyd et al. - 2019 - Multigroup Cross-Section Generation with the OpenM.pdf:application/pdf;Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\D7ASSDRX\\00295450.2019.html:text/html},
}

@incollection{kuhlmann_lid-driven_2018,
	title = {The {Lid}-{Driven} {Cavity}},
	isbn = {978-3-319-91493-0},
	abstract = {The lid-driven cavity is an important fluid mechanical system serving as a benchmark for testing numerical methods and for studying fundamental aspects of incompressible flows in confined volumes which are driven by the tangential motion of a bounding wall. A comprehensive review is provided of lid-driven cavity flows focusing on the evolution of the flow as the Reynolds number is increased. Understanding the flow physics requires to consider pure two-dimensional flows, flows which are periodic in one space direction as well as the full three-dimensional flow. The topics treated range from the characteristic singularities resulting from the discontinuous boundary conditions over flow instabilities and their numerical treatment to the transition to chaos in a fully confined cubical cavity. In addition, the streamline topology of two-dimensional time-dependent and of steady three-dimensional flows are covered, as well as turbulent flow in a fully confined lid-driven cube. Finally, an overview on various extensions of the lid-driven cavity is given.},
	booktitle = {Computational {Methods} in {Applied} {Sciences}},
	author = {Kuhlmann, Hendrik and Romanò, Francesco},
	month = apr,
	year = {2018},
	file = {Full Text PDF:C\:\\Users\\Sun Myung\\Zotero\\storage\\GYBGM3QT\\Kuhlmann and Romanò - 2018 - The Lid-Driven Cavity.pdf:application/pdf},
}

@article{shi_gen-foam_2021,
	title = {Gen-foam {Model} and {Benchmark} of {Delayed} {Neutron} {Precursor} {Drift} in the {Molten} {Salt} {Reactor} {Experiment}},
	volume = {247},
	copyright = {© The Authors, published by EDP Sciences, 2021},
	issn = {2100-014X},
	doi = {10.1051/epjconf/202124706040},
	abstract = {The effective delayed neutron fraction is an important reactor kinetics parameter. In flowing liquid-fuel reactors, this differs from the delayed neutron fraction because of the emission of delayed neutrons with a lower energy spectrum than prompt and the delayed neutron precursor (DNP) drift due to the fuel movement. In general, neglecting delayed neutron precursor drift leads to an over-estimation of the effective delayed neutron fraction. Nevertheless, the capability to simulate this peculiar phenomenon is not available in most reactor physics tools. In this project, a multi-physics approach to modeling DNP drift is developed using the GeN-Foam toolkit, and it benchmarked against available experimental data from the Molten Salt Reactor Experiment (MSRE). GeN-Foam couples a neutron diffusion solver with a thermal-hydraulics solver. Additionally, a new function was added for solving adjoint multi-group diffusion eigenvalue problems and calculating effective delayed neutron fraction. For benchmarking, an R-Z model of the MSRE was developed in GeN-Foam. The porous media model was applied, and cross sections were generated using the Monte Carlo code Serpent-2 with ENDF/B-VII.1 nuclear data library. In order to evaluate the impact of DNP drift, two steady-state conditions (stationary and flowing salt at 1200 gpm) were simulated. A reactivity change of -241 pcm was calculated using GeN-Foam for the MSRE between static and flowing fuel, which is in a good agreement with the experimental value of -212 pcm. The total effective delayed neutron fraction change was calculated to be -230 pcm vs. -304 pcm reported for the MSRE and analytical calculated during the experimental campaign. Three transient accidents were also analyzed.},
	language = {en},
	journal = {EPJ Web of Conferences},
	author = {Shi, Jun and Fratoni, Massimiliano},
	year = {2021},
	pages = {06040},
	file = {Full Text PDF:C\:\\Users\\Sun Myung\\Zotero\\storage\\EMKLDDCE\\Shi and Fratoni - 2021 - GEN-FOAM MODEL AND BENCHMARK OF DELAYED NEUTRON PR.pdf:application/pdf;Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\PSY22SMX\\epjconf_physor2020_06040.html:text/html},
}

@techreport{doe_office_2021,
	title = {Office of {Nuclear} {Energy} {Strategic} {Vision}},
	shorttitle = {{DOE}-{NE} {Strategic} {Vision}},
	url = {https://www.energy.gov/sites/prod/files/2021/01/f82/DOE-NE Strategic Vision -Web - 01.08.2021.pdf},
	language = {en},
	institution = {Office of Nuclear Energy},
	author = {DOE, Department of Energy},
	month = jan,
	year = {2021},
	file = {DOE-NE Strategic Vision -Web - 01.08.2021.pdf:C\:\\Users\\Sun Myung\\Zotero\\storage\\SFSYKIHW\\DOE-NE Strategic Vision -Web - 01.08.2021.pdf:application/pdf},
}

@article{keyes_multiphysics_2013,
	title = {Multiphysics simulations: {Challenges} and opportunities},
	volume = {27},
	issn = {1094-3420},
	shorttitle = {Multiphysics simulations},
	doi = {10.1177/1094342012468181},
	abstract = {We consider multiphysics applications from algorithmic and architectural perspectives, where “algorithmic” includes both mathematical analysis and computational complexity, and “architectural” includes both software and hardware environments. Many diverse multiphysics applications can be reduced, en route to their computational simulation, to a common algebraic coupling paradigm. Mathematical analysis of multiphysics coupling in this form is not always practical for realistic applications, but model problems representative of applications discussed herein can provide insight. A variety of software frameworks for multiphysics applications have been constructed and refined within disciplinary communities and executed on leading-edge computer systems. We examine several of these, expose some commonalities among them, and attempt to extrapolate best practices to future systems. From our study, we summarize challenges and forecast opportunities.},
	language = {en},
	number = {1},
	journal = {The International Journal of High Performance Computing Applications},
	author = {Keyes, David E and McInnes, Lois C and Woodward, Carol and Gropp, William and Myra, Eric and Pernice, Michael and Bell, John and Brown, Jed and Clo, Alain and Connors, Jeffrey and Constantinescu, Emil and Estep, Don and Evans, Kate and Farhat, Charbel and Hakim, Ammar and Hammond, Glenn and Hansen, Glen and Hill, Judith and Isaac, Tobin and Jiao, Xiangmin and Jordan, Kirk and Kaushik, Dinesh and Kaxiras, Efthimios and Koniges, Alice and Lee, Kihwan and Lott, Aaron and Lu, Qiming and Magerlein, John and Maxwell, Reed and McCourt, Michael and Mehl, Miriam and Pawlowski, Roger and Randles, Amanda P and Reynolds, Daniel and Rivière, Beatrice and Rüde, Ulrich and Scheibe, Tim and Shadid, John and Sheehan, Brendan and Shephard, Mark and Siegel, Andrew and Smith, Barry and Tang, Xianzhu and Wilson, Cian and Wohlmuth, Barbara},
	month = feb,
	year = {2013},
	keywords = {Multiphysics, implicit and explicit algorithms, loose and tight coupling., multimodel, multirate, multiscale, strong and weak coupling},
	pages = {4--83},
	file = {SAGE PDF Full Text:C\:\\Users\\Sun Myung\\Zotero\\storage\\FZHVHMA9\\Keyes et al. - 2013 - Multiphysics simulations Challenges and opportuni.pdf:application/pdf},
}

@article{permann_moose_2020,
	title = {{MOOSE}: {Enabling} massively parallel multiphysics simulation},
	volume = {11},
	issn = {2352-7110},
	shorttitle = {{MOOSE}},
	doi = {10.1016/j.softx.2020.100430},
	abstract = {Harnessing modern parallel computing resources to achieve complex multiphysics simulations is a daunting task. The Multiphysics Object Oriented Simulation Environment (MOOSE) aims to enable such development by providing simplified interfaces for specification of partial differential equations, boundary conditions, material properties, and all aspects of a simulation without the need to consider the parallel, adaptive, nonlinear, finite element solve that is handled internally. Through the use of interfaces and inheritance, each portion of a simulation becomes reusable and composable in a manner that allows disparate research groups to share code and create an ecosystem of growing capability that lowers the barrier for the creation of multiphysics simulation codes. Included within the framework is a unique capability for building multiscale, multiphysics simulations through simultaneous execution of multiple sub-applications with data transfers between the scales. Other capabilities include automatic differentiation, scaling to a large number of processors, hybrid parallelism, and mesh adaptivity. To date, MOOSE-based applications have been created in areas of science and engineering such as nuclear physics, geothermal science, magneto-hydrodynamics, seismic events, compressible and incompressible fluid flow, microstructure evolution, and advanced manufacturing processes.},
	language = {en},
	journal = {SoftwareX},
	author = {Permann, Cody J. and Gaston, Derek R. and Andrš, David and Carlsen, Robert W. and Kong, Fande and Lindsay, Alexander D. and Miller, Jason M. and Peterson, John W. and Slaughter, Andrew E. and Stogner, Roy H. and Martineau, Richard C.},
	month = jan,
	year = {2020},
	keywords = {Multiphysics, Parallel, Finite-element, Framework, Multiscale},
	pages = {100430},
	file = {ScienceDirect Full Text PDF:C\:\\Users\\Sun Myung\\Zotero\\storage\\Q27QTRPG\\Permann et al. - 2020 - MOOSE Enabling massively parallel multiphysics si.pdf:application/pdf;ScienceDirect Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\ZGBGE5S9\\S2352711019302973.html:text/html},
}

@incollection{allibert_7_2016,
	series = {Woodhead {Publishing} {Series} in {Energy}},
	title = {7 - {Molten} salt fast reactors},
	isbn = {978-0-08-100149-3},
	abstract = {Liquid-fueled reactors exhibit unusual and interesting properties compared with solid-fueled reactors, requesting a revision of some well-known conception and safety rules. Thus emphasis in this chapter is placed on such differences and the need for innovative approaches. The molten salt fast reactor, based on a fast spectrum and seen as a long-term alternative to solid-fueled fast reactors, fulfills the Generation IV criteria and has been studied for almost a decade, mainly by calculations and determination of basic physical and chemical properties in the European Union and Russian Federation. The main characteristics of this concept are presented and discussed, including transient simulation, chemistry and material issues, safety analysis, the research roadmap, and laboratory-scale experiments.},
	language = {en},
	booktitle = {Handbook of {Generation} {IV} {Nuclear} {Reactors}},
	publisher = {Woodhead Publishing},
	author = {Allibert, M. and Aufiero, M. and Brovchenko, M. and Delpech, S. and Ghetta, V. and Heuer, D. and Laureau, A. and Merle-Lucotte, E.},
	editor = {Pioro, Igor L.},
	month = jan,
	year = {2016},
	keywords = {Reactor physics, Molten salt reactor, Thorium fuel cycle, Fuel cycle, Materials, Safety, Chemistry, Generation IV, Liquid fuel},
	pages = {157--188},
	file = {ScienceDirect Full Text PDF:C\:\\Users\\Sun Myung\\Zotero\\storage\\M6NZSPWL\\Allibert et al. - 2016 - 7 - Molten salt fast reactors.pdf:application/pdf;ScienceDirect Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\8LHTXNSL\\B9780081001493000070.html:text/html},
}

@article{jayaram_overview_1987,
	title = {An overview of world thorium resources, incentives for further exploration and forecast for thorium requirements in the near future},
	language = {en},
	author = {Jayaram, K. M. V.},
	year = {1987},
	file = {Jayaram - 1987 - An overview of world thorium resources, incentives.pdf:C\:\\Users\\Sun Myung\\Zotero\\storage\\T5AZ5RE9\\Jayaram - 1987 - An overview of world thorium resources, incentives.pdf:application/pdf;Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\F9XXPP3G\\search.html:text/html},
}

@misc{zou_research_2019,
	address = {Delft, Netherlands},
	title = {Research {Progress} of {TMSR} {Design}},
	url = {http://samofar.eu/wp-content/uploads/2019/07/2019-TMSR-SAMOFAR%E2%80%94%E2%80%94Yang-ZOU-PDF-version-1.pdf},
	urldate = {2021-08-06},
	author = {Zou, Yang},
	month = jul,
	year = {2019},
	file = {2019-TMSR-SAMOFAR——Yang-ZOU-PDF-version-1.pdf:C\:\\Users\\Sun Myung\\Zotero\\storage\\GYQ4QWJA\\2019-TMSR-SAMOFAR——Yang-ZOU-PDF-version-1.pdf:application/pdf},
}

@techreport{fratoni_molten_2020,
	title = {Molten {Salt} {Reactor} {Experiment} {Benchmark} {Evaluation}},
	url = {https://www.osti.gov/biblio/1617123},
	abstract = {The U.S. Department of Energy's Office of Scientific and Technical Information},
	language = {English},
	number = {DOE-UCB-8542},
	institution = {Univ. of California, Berkeley, CA (United States)},
	author = {Fratoni, Massimiliano (ORCID:0000000304520508) and Shen, Dan and Powers, Jeff and Ilas, Germina},
	month = mar,
	year = {2020},
	file = {Full Text PDF:C\:\\Users\\Sun Myung\\Zotero\\storage\\VNC7LFTY\\Fratoni et al. - 2020 - Molten Salt Reactor Experiment Benchmark Evaluatio.pdf:application/pdf;Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\WH22LH4T\\1617123.html:text/html},
}

@book{hallegatte_shock_2016,
	title = {Shock {Waves}: {Managing} the {Impacts} of {Climate} {Change} on {Poverty}},
	isbn = {978-1-4648-0673-5},
	shorttitle = {Shock {Waves}},
	abstract = {Ending poverty and stabilizing climate change will be two unprecedented global achievements and two major steps toward sustainable development. But the two objectives cannot be considered in isolation: they need to be jointly tackled through an integrated strategy. This report brings together those two objectives and explores how they can more easily be achieved if considered together. It examines the potential impact of climate change and climate policies on poverty reduction. Shock Waves also provides guidance on how to create a "win-win" situation so that climate change policies contribute to poverty reduction and poverty-reduction policies contribute to climate change mitigation and resilience building. The key finding of the report is that climate change represents a significant obstacle to the sustained eradication of poverty, but future impacts on poverty are determined by policy choices: rapid, inclusive, and climate-informed development can prevent most short-term impacts whereas immediate pro-poor, emissions-reduction policies can drastically limit long-term ones.},
	language = {en},
	publisher = {World Bank Publications},
	author = {Hallegatte, Stéphane},
	year = {2016},
	keywords = {Business \& Economics / Development / Economic Development, Business \& Economics / Development / Sustainable Development, Social Science / Poverty \& Homelessness},
}

@article{forsberg_market_2020,
	title = {Market {Basis} for {Salt}-{Cooled} {Reactors}: {Dispatchable} {Heat}, {Hydrogen}, and {Electricity} with {Assured} {Peak} {Power} {Capacity}},
	volume = {206},
	issn = {0029-5450},
	shorttitle = {Market {Basis} for {Salt}-{Cooled} {Reactors}},
	doi = {10.1080/00295450.2020.1743628},
	abstract = {Energy markets are changing because of (1) the addition of nondispatchable wind and solar electric generating capacity and (2) the goal of a low-carbon energy system. The large-scale addition of wind and solar photovoltaics results in low wholesale electricity prices at times of high wind and solar output and high prices at times of low wind and solar input. The goal of a low-carbon energy system requires a replacement energy production system with assured peak energy production capacity.To minimize costs, capital-intensive nuclear reactors should operate at base load. To maximize revenue (minimize sales at times of low prices and maximize sales at times of high prices), the power cycle should provide variable heat and electricity. This requires the power cycle to (1) include heat storage that enables peak heat and electricity output that may be several times base-load reactor output and (2) provide assured peak power production. Assured peak power production requires the capability to efficiently burn low-carbon fuels such as hydrogen and biofuels. Alternatively, nuclear systems with base-load reactors can be built to produce peak electricity and storable hydrogen for industry, biofuels, and other markets. All power reactors with appropriate system designs can meet these requirements.The lowest-cost technologies for heat storage, assured peak power production, and hydrogen production require high-temperature heat. This economically favors salt-cooled reactors with the average temperature of delivered heat of about 650°C versus heat delivered at lower average temperatures from other reactors such as light water reactors: 280°C, sodium-cooled reactors: 500°C, and high-temperature helium-cooled reactors: 550°C. Salt-cooled reactors include (1) Fluoride-salt-cooled High-temperature Reactors (FHRs) with solid fuel and clean salt, (2) Molten Salt Reactors (MSRs) with fuel dissolved in the salt, and (3) fusion reactors with salt blankets. Future energy markets, nuclear systems (heat storage, assured peak energy production capacity, and hydrogen production) designed for such markets and the power cycle technologies that economically favor salt reactors are described.},
	number = {11},
	journal = {Nuclear Technology},
	author = {Forsberg, Charles W.},
	month = nov,
	year = {2020},
	keywords = {molten salt reactor, hydrogen, Heat storage, fluoride-salt-cooled high-temperature reactor, power cycles},
	pages = {1659--1685},
	file = {Full Text PDF:C\:\\Users\\Sun Myung\\Zotero\\storage\\P9VBPVXA\\Forsberg - 2020 - Market Basis for Salt-Cooled Reactors Dispatchabl.pdf:application/pdf;Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\NYLGYC9T\\00295450.2020.html:text/html},
}

@techreport{iea_net_2021,
	type = {Special {Report}},
	title = {Net {Zero} by 2050 - {A} {Roadmap} for the {Global} {Energy} {Sector}},
	url = {https://iea.blob.core.windows.net/assets/405543d2-054d-4cbd-9b89-d174831643a4/NetZeroby2050-ARoadmapfortheGlobalEnergySector_CORR.pdf},
	language = {en},
	institution = {International Energy Agency},
	author = {IEA},
	month = jun,
	year = {2021},
	pages = {224},
	file = {Net Zero by 2050 - A Roadmap for the Global Energy.pdf:C\:\\Users\\Sun Myung\\Zotero\\storage\\LY5ERVXQ\\Net Zero by 2050 - A Roadmap for the Global Energy.pdf:application/pdf},
}

@book{pope_turbulent_2000,
	title = {Turbulent {Flows}},
	abstract = {Cambridge Core - Nonlinear Science and Fluid Dynamics - Turbulent Flows -  by Stephen B. Pope},
	language = {en},
	author = {Pope, Stephen B.},
	month = aug,
	year = {2000},
	file = {Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\SNDTZWFG\\C58EFF59AF9B81AE6CFAC9ED16486B3A.html:text/html},
}

@inproceedings{fiorina_creation_2018,
	title = {Creation of an {OpenFOAM} {Fuel} {Performance} {Class} {Based} on {FRED} and {Integration} {Into} the {GeN}-{Foam} {Multi}-{Physics} {Code}},
	doi = {10.1115/ICONE26-81574},
	language = {en},
	publisher = {American Society of Mechanical Engineers Digital Collection},
	author = {Fiorina, Carlo and Pautz, Andreas and Mikityuk, Konstantin},
	month = oct,
	year = {2018},
	file = {Full Text PDF:C\:\\Users\\Sun Myung\\Zotero\\storage\\YJUREMLU\\Fiorina et al. - 2018 - Creation of an OpenFOAM Fuel Performance Class Bas.pdf:application/pdf;Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\FT6XYEX2\\272722.html:text/html},
}

@article{zhang_development_2009,
	title = {Development of a steady state analysis code for a molten salt reactor},
	volume = {36},
	issn = {0306-4549},
	doi = {10.1016/j.anucene.2009.01.004},
	abstract = {The molten salt reactor (MSR), which is one of the ‘Generation IV’ concepts, can be used for transmutation, and production of electricity, hydrogen and fissile fuels. In this study, a single-liquid-fueled MSR is designed for conceptual research, in which no solid material is present in the core as moderator, except for the external reflector. The fuel salt flow makes the MSR neutronics different from that of conventional reactors using solid fuels, and couples the flow and heat transfer strongly. Therefore, it is necessary to study the core characteristics with due attention to the coupling among flow, heat transfer and neutronics. The standard turbulent model is adopted to establish the flow and heat transfer model, while the diffusion theory is used for the neutronics model, which consists of two-group neutron diffusion equations for fast and thermal neutron fluxes, and balance equations for six groups of delayed neutron precursors. These two models which are coupled through the temperature and heat source are coded in a microcomputer program. The distributions of the velocity, temperature, neutron fluxes, and delayed neutron precursors under the rated condition are obtained. In addition, the effects of the inflow temperature, inflow velocity, and the fuel salt residence time out of the core are discussed in detail. The results provide some valuable information for the research and design of the new generation molten salt reactors.},
	language = {en},
	number = {5},
	journal = {Annals of Nuclear Energy},
	author = {Zhang, D. L. and Qiu, S. Z. and Su, G. H. and Liu, C. L.},
	month = may,
	year = {2009},
	keywords = {unread},
	pages = {590--603},
	file = {ScienceDirect Full Text PDF:C\:\\Users\\Sun Myung\\Zotero\\storage\\DDA9ZJDQ\\Zhang et al. - 2009 - Development of a steady state analysis code for a .pdf:application/pdf;ScienceDirect Full Text PDF:C\:\\Users\\Sun Myung\\Zotero\\storage\\KS4ZFFFG\\Zhang et al. - 2009 - Development of a steady state analysis code for a .pdf:application/pdf;ScienceDirect Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\8DB2MI7D\\S030645490900019X.html:text/html;ScienceDirect Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\W6E73ACI\\S030645490900019X.html:text/html;zhang_development_2009.pdf:C\:\\Users\\Sun Myung\\Zotero\\storage\\6NFMMZAE\\zhang_development_2009.pdf:application/pdf},
}

@article{nicolino_coupled_2008,
	title = {Coupled dynamics in the physics of molten salt reactors},
	volume = {35},
	issn = {0306-4549},
	doi = {10.1016/j.anucene.2007.06.015},
	abstract = {The present paper is devoted to the analysis of the coupled thermo-fluid and neutronic dynamics of fast fluid-fuel multiplying nuclear systems. A completely coupled model is needed since in some fast reactors designs, the velocity pattern could be very complicated and strongly affected by the neutron dynamics via the heat source from fission reactions. Furthermore, the neutron dynamics is strongly affected by the thermohydrodynamics via the motion of precursors and by feedback effects. The methods typical of solid fuel reactors of previous generations are not sufficient to handle these more highly coupled concepts. In the preset paper, we consider the coupling between neutronics and thermohydrodynamics with simple but realistic hypotheses assumed to model the evolution of all the variables involved in the calculation. The numerical scheme used represents the current state of the art in the solution of non-linear systems: the Newton–Krylov algorithm. Several calculations are presented to demonstrate the ability of the methods described here to study the behavior of molten salt reactors in both steady state and transient situations.},
	language = {en},
	number = {2},
	journal = {Annals of Nuclear Energy},
	author = {Nicolino, Claudio and Lapenta, Giovanni and Dulla, Sandra and Ravetto, Piero},
	month = feb,
	year = {2008},
	pages = {314--322},
	file = {ScienceDirect Full Text PDF:C\:\\Users\\Sun Myung\\Zotero\\storage\\78544BSJ\\Nicolino et al. - 2008 - Coupled dynamics in the physics of molten salt rea.pdf:application/pdf;ScienceDirect Full Text PDF:C\:\\Users\\Sun Myung\\Zotero\\storage\\HFETG65I\\Nicolino et al. - 2008 - Coupled dynamics in the physics of molten salt rea.pdf:application/pdf;ScienceDirect Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\GZI753P9\\S0306454907001454.html:text/html;ScienceDirect Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\FPYD47S6\\S0306454907001454.html:text/html},
}

@techreport{downar_parcs_2010,
	address = {Ann Arbor, MI},
	type = {Theory {Manual}},
	title = {{PARCS} v3.0 {U}.{S}. {NRC} {Core} {Neutronics} {Simulator}.},
	language = {en},
	institution = {University of Michigan},
	author = {Downar, T and Xu, Y and Seker, V},
	year = {2010},
	pages = {129},
	file = {Downar - PARCS v3.0 U.S. NRC Core Neutronics Simulator.pdf:C\:\\Users\\Sun Myung\\Zotero\\storage\\DI2FT6AP\\Downar - PARCS v3.0 U.S. NRC Core Neutronics Simulator.pdf:application/pdf},
}

@article{de_zwaan_static_2007,
	title = {Static design of a liquid-salt-cooled pebble bed reactor ({LSPBR})},
	volume = {34},
	issn = {0306-4549},
	doi = {10.1016/j.anucene.2006.11.008},
	abstract = {A renewed interest has been raised for liquid-salt-cooled nuclear reactors. The excellent heat transfer properties of liquid-salt coolants provide several benefits, like lower fuel temperatures, higher average coolant temperature, increased core power density and better decay heat removal, and thus higher achievable core power. In order to benefit from the on-line refueling capability of a pebble bed reactor, the liquid salt pebble bed reactor (LSPBR) is proposed. This is a high temperature pebble bed reactor with a fuel design similar to existing HTRs, but using a liquid-salt as coolant. In this paper, the selection criteria for the liquid-salt coolant are described. Based on its neutronic properties, LiF–BeF2 (flibe) was selected for the LSPBR. Two designs of the LSPBR were considered: a cylindrical core and an annular core with a graphite inner reflector. Coupled neutronic thermal-hydraulic calculations were performed to obtain the steady state power distribution and the corresponding fuel temperature distribution. Calculations were performed to investigate the decay heat removal capability in a protected loss-of-forced cooling accident. The maximum allowable power that can be produced with the LSPBR is hereby determined.},
	language = {en},
	number = {1},
	journal = {Annals of Nuclear Energy},
	author = {de Zwaan, S. J. and Boer, B. and Lathouwers, D. and Kloosterman, J. L.},
	month = jan,
	year = {2007},
	pages = {83--92},
	file = {ScienceDirect Full Text PDF:C\:\\Users\\Sun Myung\\Zotero\\storage\\62KZ3WDB\\de Zwaan et al. - 2007 - Static design of a liquid-salt-cooled pebble bed r.pdf:application/pdf;ScienceDirect Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\D37AJ697\\S0306454906002295.html:text/html},
}

@article{boer_validation_2010,
	title = {Validation of the {DALTON}-{THERMIX} {Code} {System} with {Transient} {Analyses} of the {HTR}-10 and {Application} to the {PBMR}},
	volume = {170},
	issn = {0029-5450},
	doi = {10.13182/NT10-A9485},
	abstract = {The DALTON-THERMIX code system has been developed for safety analysis and core optimization of pebble-bed reactors. The code system consists of the new three-dimensional diffusion code DALTON, which is coupled to the existing thermal-hydraulic code THERMIX. These codes are linked to a database procedure for the generation of neutron cross sections using SCALE-5.The behavior of pebble-bed reactors during a loss of forced cooling (LOFC) transient is of particular interest since the absence of forced cooling could lead to a significant increase of the temperature of the coated particle fuel. Therefore, the reactor power may be constrained during normal operation to limit the temperature.For validation purposes, calculation results of normal operation, an LOFC transient, and a control rod withdrawal transient without SCRAM have been compared with experimental data obtained in the High Temperature Reactor–10 (HTR-10). The code system has been applied to the 400-MW(thermal) pebble bed modular reactor (PBMR) design, including the analysis of three different LOFC transients. Theses results are verified by a comparison with the results of the existing TINTE code system.It was found that the code system is capable of modeling both small (HTR-10) and large (PBMR) pebble-bed reactors and therefore provides a flexible tool for safety analysis and core optimization of future reactor designs. The analyses of the LOFC transients show that the peak fuel temperature is only slightly elevated (less than +100°C) as compared to its nominal value in the HTR-10 but reaches a maximum value of 1648°C during the depressurized LOFC case of the PBMR benchmark, which is significantly higher than the peak fuel temperature (976°C) during normal operation.},
	number = {2},
	journal = {Nuclear Technology},
	author = {Boer, B. and Lathouwers, D. and Kloosterman, J. L. and Hagen, T. H. J. J. Van Der and Strydom, G.},
	month = may,
	year = {2010},
	pages = {306--321},
	file = {Full Text PDF:C\:\\Users\\Sun Myung\\Zotero\\storage\\BVWYSKV5\\Boer et al. - 2010 - Validation of the DALTON-THERMIX Code System with .pdf:application/pdf;Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\XGK4NEIN\\NT10-A9485.html:text/html},
}

@phdthesis{blanco_neutronic_2020,
	type = {{PhD} {Thesis}},
	title = {Neutronic, thermohydraulic and thermomechanical coupling for the modeling of criticality accidents in nuclear systems},
	abstract = {This thesis was developed within in the framework of the multi-scale and multi-physics models for the simulation of criticality accidents, carried jointly between the CNRS and the IRSN. A multi-physics and multi-scale approach aims to produce a numerical model taking into account all the relevant physical phenomena existing in nuclear systems as well as their coupling. This approach makes possible to improve the predictive capacities of the single physics models and to numerically study the behavior of a nuclear system under conditions that would be difficult to achieve or reproduce by experiments. The multi-scale / multi-physics approach is, therefore, particularly useful for the study of nuclear reactor criticality accidents, or more generally, for all nuclear systems where a tight coupling exists between neutronics, mechanics (of solids and fluids) and heat transfers.The objectives of the thesis were, firstly, to develop a new numerical scheme for the coupling between the neutronic code Serpent 2 (Monte Carlo code) and the Computational Fluid Dynamics (CFD) code OpenFOAM. Secondly, to develop the physical models that allow greater flexibility for criticality accidents studies in terms of type of transients, systems and phenomena considered. Among the various physical models developed during the work, it can be mentioned the transient neutronic models based on a quasi-static Monte Carlo approach and on the deterministic SP1 and SP3 methods. A porous medium model was also developed during the work to allow performing studies on nuclear systems containing a solid nuclear fuel cooled by a fluid. The numerical implementation of the multi-physics coupling was performed in C/C++ in the OpenFOAM code. This code is very well suited to numerically solve continuous mechanics problems using a finite volume method. It also provides very large library of CFD algorithms (RANS, LES et DNS). The thesis work specially focused on the study of the strategy to be followed to implement the quasi-static method numerically with a Monte Carlo type code in the same platform through internal coupling.The performances of the coupling and the developed models were studied for different scenarios and nuclear systems: the transient Godiva experiments, an international benchmark for multi-physics codes for Molten Salts Reactors and the case of a hypothetical criticality accident in a Boiling Water Reactor (BWR) spent fuel pool. These diverse scenarios and systems were selected because they are characterized by presenting a multitude of highly coupled physical phenomena which required a very careful modeling. One can mention: the Doppler and fuel density effects, the thermal expansion and thermomechanical stresses, the presence of laminar or turbulent flows in the coolant or liquid fuel, the delayed neutrons precursors convection, and the energy and mass transfers and the phase change in porous media. The different comparisons between the multi-physics tool and the available data show a very good agreement and confirm that the selected approach is pertinent for the study of criticality accidents and allows obtaining very good precision and flexibility while maintaining satisfactory computational costs.},
	language = {en},
	school = {Université Grenoble Alpes},
	author = {Blanco, Juan Antonio},
	month = dec,
	year = {2020},
	file = {Full Text PDF:C\:\\Users\\Sun Myung\\Zotero\\storage\\PWC7UK9X\\Blanco - 2020 - Neutronic, thermohydraulic and thermomechanical co.pdf:application/pdf;Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\MCS7NXPH\\tel-03172978.html:text/html},
}

@article{chapelle_inf-sup_1993,
	title = {The inf-sup test},
	volume = {47},
	issn = {0045-7949},
	doi = {10.1016/0045-7949(93)90340-J},
	abstract = {We briefly review the inf-sup condition for the finite element solution of problems in incompressible elasticity, and then propose a numerical test on whether the inf-sup condition is passed. The evaluation of elements with this test is simple, and various results are presented. This inf-sup test will prove useful for many discretizations of constrained variational problems.},
	language = {en},
	number = {4},
	journal = {Computers \& Structures},
	author = {Chapelle, D. and Bathe, K. J.},
	month = jun,
	year = {1993},
	pages = {537--545},
	file = {ScienceDirect Full Text PDF:C\:\\Users\\Sun Myung\\Zotero\\storage\\LPZPUCQC\\Chapelle and Bathe - 1993 - The inf-sup test.pdf:application/pdf;ScienceDirect Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\WTBAFE57\\004579499390340J.html:text/html},
}

@article{hughes_new_1986,
	title = {A new finite element formulation for computational fluid dynamics: {V}. {Circumventing} the babuška-brezzi condition: a stable {Petrov}-{Galerkin} formulation of the stokes problem accommodating equal-order interpolations},
	volume = {59},
	issn = {0045-7825},
	shorttitle = {A new finite element formulation for computational fluid dynamics},
	doi = {10.1016/0045-7825(86)90025-3},
	abstract = {A new Petrov-Galerkin formulation of the Stokes problem is proposed. The new formulation possesses better stability properties than the classical Galerkin/variational method. An error analysis is performed for the case in which both the velocity and pressure are approximated by C0 interpolations. Combinations of C0 interpolations which are unstable according to the Babuška-Brezzi condition (e.g., equal-order interpolations) are shown to be stable and convergent within the present framework. Calculations exhibiting the good behavior of the methodology are presented.},
	language = {en},
	number = {1},
	journal = {Computer Methods in Applied Mechanics and Engineering},
	author = {Hughes, Thomas J. R. and Franca, Leopoldo P. and Balestra, Marc},
	month = nov,
	year = {1986},
	pages = {85--99},
	file = {ScienceDirect Full Text PDF:C\:\\Users\\Sun Myung\\Zotero\\storage\\2GD7L83S\\Hughes et al. - 1986 - A new finite element formulation for computational.pdf:application/pdf;ScienceDirect Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\XVXDPTRA\\0045782586900253.html:text/html},
}

@article{brooks_streamline_1982,
	title = {Streamline upwind/{Petrov}-{Galerkin} formulations for convection dominated flows with particular emphasis on the incompressible {Navier}-{Stokes} equations},
	volume = {32},
	issn = {0045-7825},
	doi = {10.1016/0045-7825(82)90071-8},
	abstract = {A new finite element formulation for convection dominated flows is developed. The basis of the formulation is the streamline upwind concept, which provides an accurate multidimensional generalization of optimal one-dimensional upwind schemes. When implemented as a consistent Petrov-Galerkin weighted residual method, it is shown that the new formulation is not subject to the artificial diffusion criticisms associated with many classical upwind methods. The accuracy of the streamline upwind/Petrov-Galerkin formulation for the linear advection diffusion equation is demonstrated on several numerical examples. The formulation is extended to the incompressible Navier-Stokes equations. An efficient implicit pressure/explicit velocity transient algorithm is developed which accomodates several treatments of the incompressibility constraint and allows for multiple iterations within a time step. The effectiveness of the algorithm is demonstrated on the problem of vortex shedding from a circular cylinder at a Reynolds number of 100.},
	language = {en},
	number = {1},
	journal = {Computer Methods in Applied Mechanics and Engineering},
	author = {Brooks, Alexander N. and Hughes, Thomas J. R.},
	month = sep,
	year = {1982},
	pages = {199--259},
	file = {ScienceDirect Full Text PDF:C\:\\Users\\Sun Myung\\Zotero\\storage\\FQ6XS4RK\\Brooks and Hughes - 1982 - Streamline upwindPetrov-Galerkin formulations for.pdf:application/pdf;ScienceDirect Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\ILNUHQDT\\0045782582900718.html:text/html},
}

@article{wang_rattlesnake_2021,
	title = {Rattlesnake: {A} {MOOSE}-{Based} {Multiphysics} {Multischeme} {Radiation} {Transport} {Application}},
	volume = {0},
	issn = {0029-5450},
	shorttitle = {Rattlesnake},
	doi = {10.1080/00295450.2020.1843348},
	abstract = {Advanced reactor concepts span the spectrum from heat pipe–cooled microreactors, through thermal and fast molten-salt reactors, to gas- and salt-cooled pebble bed reactors. The modeling and simulation of each of these reactor types comes with their own geometrical complexities and multiphysics challenges. However, the common theme for all nuclear reactors is the necessity to be able to accurately predict neutron distribution in the presence of multiphysics feedback. We argue that the current standards of modeling and simulation, which couple single-physics, single-reactor-focused codes via ad hoc methods, are not sufficiently flexible to address the challenges of modeling and simulation for advanced reactors. In this work, we present the Multiphysics Object Oriented Simulation Environment (MOOSE)–based radiation transport application Rattlesnake. The use of Rattlesnake for the modeling and simulation of nuclear reactors represents a paradigm shift away from makeshift data exchange methods, as it is developed based on the MOOSE platform with its very natural form of shared data distribution. Rattlesnake is well equipped for addressing the geometric and multiphysics challenges of advanced reactor concepts because it is a flexible finite element tool that leverages the multiphysics capabilities inherent in MOOSE. This paper focuses on the concept and design of Rattlesnake. We also demonstrate the capabilities and performance of Rattlesnake with a set of problems including a microreactor, a molten-salt reactor, a pebble bed reactor, the Advanced Test Reactor at the Idaho National Laboratory, and two benchmarks: a multiphysics version of the C5G7 benchmark and the LRA benchmark.},
	number = {0},
	journal = {Nuclear Technology},
	author = {Wang, Yaqi and Schunert, Sebastian and Ortensi, Javier and Laboure, Vincent and DeHart, Mark and Prince, Zachary and Kong, Fande and Harter, Jackson and Balestra, Paolo and Gleicher, Frederick},
	month = apr,
	year = {2021},
	keywords = {MOOSE, multiphysics, reactor physics, radiation transport, Rattlesnake},
	pages = {1--26},
	file = {Full Text PDF:C\:\\Users\\Sun Myung\\Zotero\\storage\\P959F3R3\\Wang et al. - 2021 - Rattlesnake A MOOSE-Based Multiphysics Multischem.pdf:application/pdf;Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\CDETVBF8\\00295450.2020.html:text/html},
}

@inproceedings{abou-jaoude_coupled_2020,
	title = {Coupled {Multiphysics} {Simulation} of {Pool}-{Type} {Molten} {Salt} {Reactors} {Using} {Griffin}/{Pronghorn}},
	doi = {10.13182/T123-33466},
	author = {Abou-Jaoude, A. and Hermosillo, A. and Martin, Nicolas and Schunert, Sebastian and Wang, Y. and Balestra, Paolo},
	month = jan,
	year = {2020},
	pages = {1587--1590},
	file = {Abou-Jaoude et al. - 2020 - Coupled Multiphysics Simulation of Pool-Type Molte.pdf:C\:\\Users\\Sun Myung\\Zotero\\storage\\TJHCY6NV\\Abou-Jaoude et al. - 2020 - Coupled Multiphysics Simulation of Pool-Type Molte.pdf:application/pdf},
}

@misc{tiberga_results_2019,
	title = {Results of the {CNRS} benchmark},
	url = {https://zenodo.org/record/3648639},
	abstract = {Raw results obtained by each partner performing the CNRS benchmark.},
	urldate = {2021-10-11},
	publisher = {Zenodo},
	author = {Tiberga, M. and De Oliveira, R. G. G. and Cervi, E. and Blanco, J. A. and Lorenzi, S. and Aufiero, M. and Lathouwers, D. and Rubiolo, P.},
	month = dec,
	year = {2019},
	file = {Zenodo Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\PKN7A9TW\\3648639.html:text/html},
}

@misc{terrapower_terrapower_2021,
	title = {{TerraPower} {Molten} {Chloride} {Fast} {Reactor} {Technology} {Fact} {Sheet}},
	url = {https://www.terrapower.com/our-work/molten-chloride-fast-reactor-technology/},
	urldate = {2021-10-08},
	publisher = {https://www.terrapower.com/our-work/molten-chloride-fast-reactor-technology/},
	author = {TerraPower},
	month = apr,
	year = {2021},
	file = {TP_2021_MCFR_Technology.pdf:C\:\\Users\\Sun Myung\\Zotero\\storage\\P8R7G8EI\\TP_2021_MCFR_Technology.pdf:application/pdf},
}

@misc{park_results_2021,
	title = {Results from {Moltres} for the {CNRS} {Benchmark}},
	url = {https://zenodo.org/record/5534964},
	abstract = {Raw data from running the CNRS benchmark with Moltres.},
	language = {eng},
	urldate = {2021-09-28},
	publisher = {Zenodo},
	author = {Park, Sun Myung and Munk, Madicken and Huff, Kathryn D.},
	month = sep,
	year = {2021},
	file = {Zenodo Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\ME29K27S\\5534964.html:text/html},
}

@inproceedings{tiberga_discontinuous_2019,
	address = {Portland, OR},
	title = {A {DISCONTINUOUS} {GALERKIN} {FEM} {MULTI}-{PHYSICS} {SOLVER} {FOR} {THE} {MOLTEN} {SALT} {FAST} {REACTOR}},
	abstract = {Numerical simulations of fast MSRs constitute a challenging task. In fact, classical codes employed in reactor physics cannot be used, and new dedicated multi-physics tools must be developed, to capture the unique features of these systems: the strong coupling between neutronics and thermal-hydraulics due to the use of a liquid fuel, the effects on reactor kinetics induced by the precursors drift, the internal heat generation, and the shape of the core having no fuel pins as a repeated structure. In this work, we present a novel multi-physics tool being developed at TU Delft. The coupling is realized between an SN radiation transport code (PHANTOM-SN) and a RANS solver (DGFlows). Both in-house tools are based on a Discontinuous Galerkin Finite Element space discretization, characterized by local conservation, high-order accuracy, and allowing for high geometric flexibility. Implicit discretization in time is performed adopting Backward Differentiation Formulae. Cross sections are computed on an element base, starting from the local average temperature and a set of libraries generated at reference temperatures with Monte Carlo or deterministic codes. Comparison of the results obtained performing a suitable numerical benchmark created at LPSC/CNRS/Grenoble with those available in literature shows that the multi-physics tool is able to capture the unique phenomena characterizing fast liquid-fueled systems.},
	booktitle = {Proceedings of the {International} {Conference} on {Mathematics} and {Computational} {Methods} applied to {Nuclear} {Science} and {Engineering} ({M}\&{C} 2019)},
	publisher = {American Nuclear Society},
	author = {Tiberga, Marco and Lathouwers, D. and Kloosterman, Jan},
	month = aug,
	year = {2019},
	file = {Full Text PDF:C\:\\Users\\Sun Myung\\Zotero\\storage\\RA4MFL2T\\Tiberga et al. - 2019 - A DISCONTINUOUS GALERKIN FEM MULTI-PHYSICS SOLVER .pdf:application/pdf},
}

@article{altahhan_preliminary_2020,
	title = {Preliminary design and analysis of {Liquid} {Fuel} {Molten} {Salt} {Reactor} using multi-physics code {GeN}-{Foam}},
	volume = {369},
	issn = {0029-5493},
	doi = {10.1016/j.nucengdes.2020.110826},
	abstract = {In this study, a hypothetical fast spectrum Liquid Fuel Molten Salt Reactor (LFMSR) core is modeled using the multiphysics C++ code GeN-Foam (General Nuclear Foam). GeN-Foam is based on OpenFOAM, a C++ open-source library for solution of continuum mechanics problems. The code utilizes a unified fine/coarse mesh approach, modeling different physics such as neutron kinetics, thermal-hydraulics based on porous fluid equations, and structural thermal-mechanics. A steady state analysis of a simplified two-dimensional (2D) LFMSR model has been performed assuming rotational symmetry to cross verify the code with the commercial ANSYS Computational Fluid Dynamics (CFD) code Fluent. The calculations showed very good agreement between the two codes allowing progression to a three-dimensional (3D) model simulation. A coupled neutron kinetics and CFD steady state analysis of a right-cylindrical 3D LFMSR core has been performed modeling one quarter of the core while using symmetry boundaries to reduce the computational time. Mixed uranium and plutonium chloride fuel has been selected in this preliminary study. Both 2D and 3D simulations showed appearance of recirculation zones within the right-cylinder core. These zones can be a challenge for LFMSR control and materials. A new hyperboloid design is proposed to remove recirculation zones, which is based on eight symmetrical loops. An Unprotected Loss of Flow accident (ULOF), in which the pump head is instantaneously reduced to zero, has been selected to demonstrate the safety characteristics of the reactor in one of the most challenging possible situations for LFMSR. The obtained results (e.g., reduced total precursors concentration at the core inlet and reduction of the core nominal power following the transient) confirm that GeN-Foam is capable of performing coupled LFMSR transient analysis and can be used for design analysis and optimization. Although the current design needs further assessment and development, it shows encouraging performance under ULOF conditions paving the way to the next step in the optimization process.},
	language = {en},
	journal = {Nuclear Engineering and Design},
	author = {Altahhan, Muhammad Ramzy and Bhaskar, Sandesh and Ziyad, Devshibhai and Balestra, Paolo and Fiorina, Carlo and Hou, Jason and Smith, Nicholas and Avramova, Maria},
	month = dec,
	year = {2020},
	keywords = {Nuclear energy, Multiphysics, CFD, OpenFOAM, Molten Salt Reactor, Liquid fuel, GeN-Foam},
	pages = {110826},
	file = {ScienceDirect Full Text PDF:C\:\\Users\\Sun Myung\\Zotero\\storage\\5WFVL2AZ\\Ramzy Altahhan et al. - 2020 - Preliminary design and analysis of Liquid Fuel Mol.pdf:application/pdf;ScienceDirect Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\XW4XVUG8\\S0029549320303204.html:text/html},
}

@inproceedings{fiorina_detailed_2019,
	title = {{DETAILED} {MODELLING} {OF} {THE} {EXPANSION} {REACTIVITY} {FEEDBACK} {IN} {FAST} {REACTORS} {USING} {OpenFOAM}},
	abstract = {The correct prediction of expansion-related reactivity feedbacks is paramount to the safety analysis of fast reactors. Traditionally, these feedbacks were evaluated via uniform radial and axial expansions based on average diagrid and fuel temperatures, respectively. The resulting smearing of the neutronic effects entails some limitations in terms of representation of local effects and overall accuracy. In addition, the thermo-mechanical behavior of a core is a complex phenomenon that cannot be reduced to simple uniform axial and radial expansions. Common examples of complex deformation patterns are the core compaction and the core flowering, which are known to have potentially dramatic impacts on core safety. In this paper, a model is developed that allows for a detailed thermo-mechanical analysis of the core, and that is capable of using this information for a locally detailed deformation of the geometry employed in the neutronic model. The capabilities of the model are discussed, together with its present limitations and some needs for improvements.},
	author = {Fiorina, Carlo and Radman, Stefan and Koc, Muhammed-Zahid and Pautz, Andreas},
	month = dec,
	year = {2019},
	file = {Full Text PDF:C\:\\Users\\Sun Myung\\Zotero\\storage\\3YKPXURV\\Fiorina et al. - 2019 - DETAILED MODELLING OF THE EXPANSION REACTIVITY FEE.pdf:application/pdf},
}

@article{fiorina_gen-foam_2015,
	title = {{GeN}-{Foam}: a novel {OpenFOAM}® based multi-physics solver for {2D}/{3D} transient analysis of nuclear reactors},
	volume = {294},
	issn = {00295493},
	shorttitle = {{GeN}-{Foam}},
	doi = {10.1016/j.nucengdes.2015.05.035},
	language = {en},
	journal = {Nuclear Engineering and Design},
	author = {Fiorina, Carlo and Clifford, Ivor and Aufiero, Manuele and Mikityuk, Konstantin},
	month = dec,
	year = {2015},
	pages = {24--37},
	file = {Fiorina et al. - 2015 - GeN-Foam a novel OpenFOAM® based multi-physics so.pdf:C\:\\Users\\Sun Myung\\Zotero\\storage\\LRSSJTUE\\Fiorina et al. - 2015 - GeN-Foam a novel OpenFOAM® based multi-physics so.pdf:application/pdf},
}

@article{leppanen_development_2013,
	title = {{DEVELOPMENT} {OF} {A} {DYNAMIC} {SIMULATION} {MODE} {IN} {SERPENT} 2 {MONTE} {CARLO} {CODE}},
	abstract = {This paper presents a dynamic neutron transport mode, currently being implemented in the Serpent 2 Monte Carlo code for the purpose of simulating short reactivity transients with temperature feedback. The transport routine is introduced and validated by comparison to MCNP5 calculations. The method is also tested in combination with an internal temperature feedback module, which forms the inner part of a multi-physics coupling scheme in Serpent 2. The demo case for the coupled calculation is a reactivity-initiated accident (RIA) in PWR fuel.},
	language = {en},
	author = {Leppanen, Jaakko},
	year = {2013},
	pages = {11},
	file = {Leppanen - 2013 - DEVELOPMENT OF A DYNAMIC SIMULATION MODE IN SERPEN.pdf:C\:\\Users\\Sun Myung\\Zotero\\storage\\C9JY6R2J\\Leppanen - 2013 - DEVELOPMENT OF A DYNAMIC SIMULATION MODE IN SERPEN.pdf:application/pdf},
}

@article{fiorina_development_2016,
	title = {Development and verification of the neutron diffusion solver for the {GeN}-{Foam} multi-physics platform},
	volume = {96},
	issn = {03064549},
	doi = {10.1016/j.anucene.2016.05.023},
	abstract = {The Laboratory for Reactor Physics and Systems Behaviour at the PSI and the EPFL has been developing in recent years a new code system for reactor analysis based on OpenFOAMÒ. The objective is to supplement available legacy codes with a modern tool featuring state-of-the-art characteristics in terms of scalability, programming approach and flexibility. As part of this project, a new solver has been developed for the eigenvalue and transient solution of multi-group diffusion equations. Several features distinguish the developed solver from other available codes, in particular: object oriented programming to ease code modification and maintenance; modern parallel computing capabilities; use of general unstructured meshes; possibility of mesh deformation; cell-wise parametrization of cross-sections; and arbitrary energy group structure. In addition, the solver is integrated into the GeN-Foam multi-physics solver. The general features of the solver and its integration with GeN-Foam have already been presented in previous publications. The present paper describes the diffusion solver in more details and provides an overview of new features recently implemented, including the use of acceleration techniques and discontinuity factors. In addition, a code verification is performed through a comparison with Monte Carlo results for both a thermal and a fast reactor system.},
	language = {en},
	journal = {Annals of Nuclear Energy},
	author = {Fiorina, Carlo and Kerkar, Nordine and Mikityuk, Konstantin and Rubiolo, Pablo and Pautz, Andreas},
	month = oct,
	year = {2016},
	pages = {212--222},
	file = {Fiorina et al. - 2016 - Development and verification of the neutron diffus.pdf:C\:\\Users\\Sun Myung\\Zotero\\storage\\96Y33ZJI\\Fiorina et al. - 2016 - Development and verification of the neutron diffus.pdf:application/pdf},
}

@article{griffiths_no_1997,
	title = {The ‘{No} {Boundary} {Condition}’ {Outflow} {Boundary} {Condition}},
	volume = {24},
	issn = {1097-0363},
	doi = {10.1002/(SICI)1097-0363(19970228)24:4<393::AID-FLD505>3.0.CO;2-O},
	abstract = {We use a one-dimensional model problem of advection– diffusion to investigate the treatment recently advocated by Papanastasiou and colleagues to deal with boundary conditions at artificial outflow boundaries. Using finite elements of degree p, we show that their treatment is equivalent to imposing the condition that the (p+1 )st derivative of the dependent variable should vanish at a point close to the outflow. This is then shown to lead to errors of order 𝒪((h+1/Pe)1.6p+1) in the numerical solutions (where h is the maximum element size and Pe is the global Peclet number), which is superior to the errors of order 𝒪(hp+1+1/Pe) obtained using a standard no-flux outflow condition. These findings are verified by numerical experiments. © 1997 by John Wiley and Sons, Ltd.},
	language = {en},
	number = {4},
	journal = {International Journal for Numerical Methods in Fluids},
	author = {Griffiths, David F.},
	year = {1997},
	keywords = {advection–diffusion, finite elements, outflow boundary conditions},
	pages = {393--411},
	file = {Full Text PDF:C\:\\Users\\Sun Myung\\Zotero\\storage\\UUQ2GFCE\\Griffiths - 1997 - The ‘No Boundary Condition’ Outflow Boundary Condi.pdf:application/pdf;Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\TQNRDBYQ\\(SICI)1097-0363(19970228)244393AID-FLD5053.0.html:text/html},
}

@article{brown_endfb-viii0_2018,
	series = {Special {Issue} on {Nuclear} {Reaction} {Data}},
	title = {{ENDF}/{B}-{VIII}.0: {The} 8th {Major} {Release} of the {Nuclear} {Reaction} {Data} {Library} with {CIELO}-project {Cross} {Sections}, {New} {Standards} and {Thermal} {Scattering} {Data}},
	volume = {148},
	issn = {0090-3752},
	shorttitle = {{ENDF}/{B}-{VIII}.0},
	doi = {10.1016/j.nds.2018.02.001},
	abstract = {We describe the new ENDF/B-VIII.0 evaluated nuclear reaction data library. ENDF/B-VIII.0 fully incorporates the new IAEA standards, includes improved thermal neutron scattering data and uses new evaluated data from the CIELO project for neutron reactions on 1H, 16O, 56Fe, 235U, 238U and 239Pu described in companion papers in the present issue of Nuclear Data Sheets. The evaluations benefit from recent experimental data obtained in the U.S. and Europe, and improvements in theory and simulation. Notable advances include updated evaluated data for light nuclei, structural materials, actinides, fission energy release, prompt fission neutron and γ-ray spectra, thermal neutron scattering data, and charged-particle reactions. Integral validation testing is shown for a wide range of criticality, reaction rate, and neutron transmission benchmarks. In general, integral validation performance of the library is improved relative to the previous ENDF/B-VII.1 library.},
	language = {en},
	journal = {Nuclear Data Sheets},
	author = {Brown, D. A. and Chadwick, M. B. and Capote, R. and Kahler, A. C. and Trkov, A. and Herman, M. W. and Sonzogni, A. A. and Danon, Y. and Carlson, A. D. and Dunn, M. and Smith, D. L. and Hale, G. M. and Arbanas, G. and Arcilla, R. and Bates, C. R. and Beck, B. and Becker, B. and Brown, F. and Casperson, R. J. and Conlin, J. and Cullen, D. E. and Descalle, M. -A. and Firestone, R. and Gaines, T. and Guber, K. H. and Hawari, A. I. and Holmes, J. and Johnson, T. D. and Kawano, T. and Kiedrowski, B. C. and Koning, A. J. and Kopecky, S. and Leal, L. and Lestone, J. P. and Lubitz, C. and Márquez Damián, J. I. and Mattoon, C. M. and McCutchan, E. A. and Mughabghab, S. and Navratil, P. and Neudecker, D. and Nobre, G. P. A. and Noguere, G. and Paris, M. and Pigni, M. T. and Plompen, A. J. and Pritychenko, B. and Pronyaev, V. G. and Roubtsov, D. and Rochman, D. and Romano, P. and Schillebeeckx, P. and Simakov, S. and Sin, M. and Sirakov, I. and Sleaford, B. and Sobes, V. and Soukhovitskii, E. S. and Stetcu, I. and Talou, P. and Thompson, I. and van der Marck, S. and Welser-Sherrill, L. and Wiarda, D. and White, M. and Wormald, J. L. and Wright, R. Q. and Zerkle, M. and Žerovnik, G. and Zhu, Y.},
	month = feb,
	year = {2018},
	pages = {1--142},
	file = {ScienceDirect Full Text PDF:C\:\\Users\\Sun Myung\\Zotero\\storage\\F3YIPC7R\\Brown et al. - 2018 - ENDFB-VIII.0 The 8th Major Release of the Nuclea.pdf:application/pdf},
}

@article{plompen_joint_2020,
	title = {The joint evaluated fission and fusion nuclear data library, {JEFF}-3.3},
	volume = {56},
	issn = {1434-601X},
	doi = {10.1140/epja/s10050-020-00141-9},
	abstract = {The joint evaluated fission and fusion nuclear data library 3.3 is described. New evaluations for neutron-induced interactions with the major actinides \$\${\textasciicircum}\{235\}{\textbackslash}hbox \{U\}\$\$, \$\${\textasciicircum}\{238\}{\textbackslash}hbox \{U\}\$\$and \$\${\textasciicircum}\{239\}{\textbackslash}hbox \{Pu\}\$\$, on \$\${\textasciicircum}\{241\}{\textbackslash}hbox \{Am\}\$\$and \$\${\textasciicircum}\{23\}{\textbackslash}hbox \{Na\}\$\$, \$\${\textasciicircum}\{59\}{\textbackslash}hbox \{Ni\}\$\$, Cr, Cu, Zr, Cd, Hf, W, Au, Pb and Bi are presented. It includes new fission yields, prompt fission neutron spectra and average number of neutrons per fission. In addition, new data for radioactive decay, thermal neutron scattering, gamma-ray emission, neutron activation, delayed neutrons and displacement damage are presented. JEFF-3.3 was complemented by files from the TENDL project. The libraries for photon, proton, deuteron, triton, helion and alpha-particle induced reactions are from TENDL-2017. The demands for uncertainty quantification in modeling led to many new covariance data for the evaluations. A comparison between results from model calculations using the JEFF-3.3 library and those from benchmark experiments for criticality, delayed neutron yields, shielding and decay heat, reveals that JEFF-3.3 performes very well for a wide range of nuclear technology applications, in particular nuclear energy.},
	language = {en},
	number = {7},
	journal = {The European Physical Journal A},
	author = {Plompen, A. J. M. and Cabellos, O. and De Saint Jean, C. and Fleming, M. and Algora, A. and Angelone, M. and Archier, P. and Bauge, E. and Bersillon, O. and Blokhin, A. and Cantargi, F. and Chebboubi, A. and Diez, C. and Duarte, H. and Dupont, E. and Dyrda, J. and Erasmus, B. and Fiorito, L. and Fischer, U. and Flammini, D. and Foligno, D. and Gilbert, M. R. and Granada, J. R. and Haeck, W. and Hambsch, F.-J. and Helgesson, P. and Hilaire, S. and Hill, I. and Hursin, M. and Ichou, R. and Jacqmin, R. and Jansky, B. and Jouanne, C. and Kellett, M. A. and Kim, D. H. and Kim, H. I. and Kodeli, I. and Koning, A. J. and Konobeyev, A. Yu. and Kopecky, S. and Kos, B. and Krása, A. and Leal, L. C. and Leclaire, N. and Leconte, P. and Lee, Y. O. and Leeb, H. and Litaize, O. and Majerle, M. and Márquez Damián, J. I. and Michel-Sendis, F. and Mills, R. W. and Morillon, B. and Noguère, G. and Pecchia, M. and Pelloni, S. and Pereslavtsev, P. and Perry, R. J. and Rochman, D. and Röhrmoser, A. and Romain, P. and Romojaro, P. and Roubtsov, D. and Sauvan, P. and Schillebeeckx, P. and Schmidt, K. H. and Serot, O. and Simakov, S. and Sirakov, I. and Sjöstrand, H. and Stankovskiy, A. and Sublet, J. C. and Tamagno, P. and Trkov, A. and van der Marck, S. and Álvarez-Velarde, F. and Villari, R. and Ware, T. C. and Yokoyama, K. and Žerovnik, G.},
	month = jul,
	year = {2020},
	pages = {181},
	file = {Springer Full Text PDF:C\:\\Users\\Sun Myung\\Zotero\\storage\\WXUQHJPX\\Plompen et al. - 2020 - The joint evaluated fission and fusion nuclear dat.pdf:application/pdf},
}

@misc{the_openfoam_foundation_ltd_openfoam_2021,
	address = {England},
	title = {{OpenFOAM}},
	url = {https://openfoam.org/},
	urldate = {2021-06-24},
	publisher = {The OpenFOAM Foundation Ltd},
	author = {{The OpenFOAM Foundation Ltd}},
	year = {2021},
	file = {Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\UPAPF2JA\\openfoam.org.html:text/html},
}

@misc{comsol_ab_comsol_nodate,
	address = {Stockholm, Sweden},
	title = {{COMSOL} {Multiphysics}®},
	url = {www.comsol.com},
	publisher = {COMSOL AB},
	author = {{COMSOL AB}},
}

@article{schunert_control_2019,
	title = {Control rod treatment for {FEM} based radiation transport methods},
	volume = {127},
	issn = {0306-4549},
	doi = {10.1016/j.anucene.2018.11.054},
	abstract = {This paper presents a novel control rod treatment for finite element method (FEM) based reactor physics simulations. The new method falls into the class of homogenization methods which treat material interfaces in mesh elements by defining a representative, homogenized nuclear cross section set. In contrast to existing homogenization based cusping treatments that attempt to approximate flux-volume weighting of the cross sections, we propose to exactly integrate terms appearing in the variational form of the transport or diffusion equation over mixed material elements. This goal is achieved by defining smoothly varying cross sections in elements that straddle material boundaries. It is shown that the smoothly varying cross sections can be obtained by projecting the piecewise set of cross sections onto the set of Legendre polynomials up to a finite order. Implementation of smoothly varying cross sections is supported within the Rattlesnake radiation multiphysics code at the Idaho National Laboratory. Homogenization methods traditionally suffer from cusping effects that originate from overestimation of absorption cross sections in strongly neutron-absorbing materials as found in control rods. We demonstrate that the presented method is (1) highly effective in removing cusping effects and (2) provides accurate results based on the C5G7-TD benchmark, a time-dependent version of the C5G7 benchmark.},
	language = {en},
	journal = {Annals of Nuclear Energy},
	author = {Schunert, Sebastian and Wang, Yaqi and Ortensi, Javier and Laboure, Vincent and Gleicher, Frederick and DeHart, Mark and Martineau, Richard},
	month = may,
	year = {2019},
	keywords = {Reactor physics, Control rod cusping, Finite element method, Multiphysics object oriented simulation environment (MOOSE), Radiation transport, Rattlesnake transport code},
	pages = {293--302},
	file = {ScienceDirect Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\6HX8JTII\\S0306454918306534.html:text/html},
}

@article{park_verification_2022,
	title = {Verification of moltres for multiphysics simulations of fast-spectrum molten salt reactors},
	volume = {173},
	issn = {03064549},
	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},
	journal = {Annals of Nuclear Energy},
	author = {Park, Sun Myung and Munk, Madicken},
	month = aug,
	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},
}

@incollection{shultis_chapter_2016,
	title = {Chapter 10: {Principles} of {Nuclear} {Reactors}},
	isbn = {978-1-315-18318-3},
	abstract = {Since December 2, 1942, when the first man-made nuclear reactor produced a selfsustaining chain reaction, several hundred different types of reactor systems have been constructed. Despite the many possible differences in design, there are a number of general features which all reactors have in common. The heart of every reactor is an active core in which the fission chain reaction is sustained. The active core contains (1) fissile fuel which through its fissioning is the main source of neutrons, (2) moderator material if the fission neutrons are to slow down, (3) coolant if the heat generated by the fissions is to be removed from the core, and (4) structural material which maintains the physical integrity of the core. Surrounding the active core is usually either a reflector whose purpose is to scatter neutrons back towards the core or a blanket region which captures neutrons leaking from the core to produce useful isotopes such as 60Co or 239Pu. The reactor core and reflector/blanket are, in turn, surrounded by a shield to minimize radiation reaching personnel and equipment near the reactor. Finally, all reactors must have a means of control to allow the chain reaction to be started up, maintained at some desired level, and safely shut down.},
	language = {en},
	booktitle = {Fundamentals of {Nuclear} {Science} and {Engineering}},
	publisher = {CRC Press},
	author = {Shultis, J. Kenneth and Faw, Richard E.},
	month = nov,
	year = {2016},
	pages = {361},
	file = {Full Text PDF:C\:\\Users\\Sun Myung\\Zotero\\storage\\8QFRFHV2\\2016 - Fundamentals of Nuclear Science and Engineering.pdf:application/pdf;Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\TBS9X8CC\\fundamentals-nuclear-science-engineering-kenneth-shultis-richard-faw.html:text/html},
}

@techreport{scherer_determination_1976,
	address = {Germany},
	title = {Determination of equivalent cross sections for representation of control rod regions in diffusion calculations},
	abstract = {The representation of control rod regions in reactor calculations requires a combination
of transport and diffusion theory calculations A method is described which produces
equivalent cross sections for a rodded region These cross sections used in a diffusion
theory calcualtion yield the same rod efficiency and reaction rate distribution as
the transport theory calculation for the explicit heterogeneous control rod The description
of the method is complemented by sample problems (orig)},
	author = {Scherer, W. and Neef, H.J.},
	year = {1976},
	pages = {24},
	file = {Scherer and Neef - 1976 - Determination of equivalent cross sections for rep.pdf:C\:\\Users\\Sun Myung\\Zotero\\storage\\65T4PK2F\\Scherer and Neef - 1976 - Determination of equivalent cross sections for rep.pdf:application/pdf},
}

@techreport{fen_modelling_1992,
	address = {Germany},
	title = {Modelling of neutron absorbers in high temperature reactors by combined transport-diffusion methods},
	abstract = {Today, the neutron-physical description of strong neutron absorbing materials for
control and shut-down of nuclear power plants is performed using combined transport
and diffusion methods Two of these approaches are described and compared in this paper
The method of equivalent cross-sections has been developed at the KFA-Juelich Institute
for Safety Research and Reactor Technology (ISR) and was widely used for all german
HTR reactor concepts The Obninsk Institute for Nuclear Power Engineering (OINPE) in
UdSSR has worked out the response-matrix method for rodded regions Both, equivalent
cross-sections and response-matrix are results of transport calculations They are
subsequently used in multidimensional diffusion type reactor codes for the representation
of absorber regions The analysis of typical examples related to the modular pebble-bed
HTR showed a good agreement of the results of both methods relative to each other
and relative to pure transport calculations such as S-N and Monte-Carlo (orig)},
	author = {Fen, V. and Lebedev, M. and Sarytchev, V. and Scherer, W.},
	year = {1992},
	pages = {46},
}

@article{mulder_neutronics_2020,
	title = {Neutronics characteristics of a 165 {MWth} {Xe}-100 reactor},
	volume = {357},
	issn = {00295493},
	doi = {10.1016/j.nucengdes.2019.110415},
	abstract = {The Xe-100 is a high temperature, helium-cooled, graphite moderated, pebble bed reactor (HTGR) featuring a 6times through, multi-pass (MEDUL = MEhrfachDUrchLauf) fuelling scheme. It is designed for delivering heat in the form of superheated steam at 565 °C and 16.5 MPa. This steam can be used to generate electricity, provide process heat for petrochemical application, or for cogeneration applications, such as the treatment of contaminated water resources, whilst simultaneously generating electricity.},
	language = {en},
	journal = {Nuclear Engineering and Design},
	author = {Mulder, E.J. and Boyes, W.A.},
	month = feb,
	year = {2020},
	pages = {110415},
	file = {Mulder and Boyes - 2020 - Neutronics characteristics of a 165 MWth Xe-100 re.pdf:C\:\\Users\\Sun Myung\\Zotero\\storage\\7E28SQJJ\\Mulder and Boyes - 2020 - Neutronics characteristics of a 165 MWth Xe-100 re.pdf:application/pdf},
}

@article{reitsma_evaluating_2003,
	title = {Evaluating the control rod modelling approach used in the {South} {African} {PBMR}: comparison of {VSOP} calculations with {ASTRA} experiments},
	volume = {222},
	issn = {00295493},
	shorttitle = {Evaluating the control rod modelling approach used in the {South} {African} {PBMR}},
	doi = {10.1016/S0029-5493(03)00009-8},
	abstract = {The problem of modelling a highly absorbing region in a diffusion calculation is well known and many methods have been developed to accommodate the transport effects in diffusion theory. In this work the use of the equivalent cross-sections method for pebble bed type reactors is evaluated by applying it to calculations of control rod (CR) experiments performed at the ASTRA Critical Facility at the Russian Research Centre, Kurchatov Institute in Moscow. The measured reactivity worths of the CRs situated in the side reflector are compared with the calculated values making use of equivalent diffusion parameters in VSOP. Results obtained were favourable for CRs situated within the first ring of reflector blocks, with larger errors obtained for CRs situated further from the core. An additional method that was investigated is the use of equivalent boron concentrations (EBCs) to represent the absorber region. This is shown to be useful if applied correctly and with care especially in the case of differential CR worth. Practical difficulties exist with both approaches, which makes the investigation of an alternative method, which should remove these shortcomings, attractive.},
	language = {en},
	number = {2-3},
	journal = {Nuclear Engineering and Design},
	author = {Reitsma, F. and Naidoo, D.},
	month = jun,
	year = {2003},
	pages = {147--159},
	file = {Reitsma and Naidoo - 2003 - Evaluating the control rod modelling approach used.pdf:C\:\\Users\\Sun Myung\\Zotero\\storage\\IE2VNYR4\\Reitsma and Naidoo - 2003 - Evaluating the control rod modelling approach used.pdf:application/pdf},
}

@techreport{ougouag_transport_2010,
	title = {Transport {Corrections} in {Nodal} {Diffusion} {Codes} for {HTR} {Modeling}},
	url = {http://www.osti.gov/servlets/purl/993162-BIrnVE/},
	language = {en},
	number = {INL/EXT-10-19729, 993162},
	institution = {Idaho National Laboratory},
	author = {Ougouag, Abderrafi M. and Gleicher, Frederick N. and Ferrer, Rodolfo M.},
	month = aug,
	year = {2010},
	pages = {INL/EXT--10--19729, 993162},
	file = {Abderrafi M. Ougouag and Frederick N. Gleicher - 2010 - Transport Corrections in Nodal Diffusion Codes for.pdf:C\:\\Users\\Sun Myung\\Zotero\\storage\\Q33J5SU2\\Abderrafi M. Ougouag and Frederick N. Gleicher - 2010 - Transport Corrections in Nodal Diffusion Codes for.pdf:application/pdf},
}

@article{xuhua_application_2009,
	title = {Application of the discontinuity factor theory to accelerate routine rod worth calculations for modular {HTRs}},
	volume = {239},
	issn = {00295493},
	doi = {10.1016/j.nucengdes.2008.10.018},
	abstract = {A modular high temperature gas cooled reactor (HTR) is equipped with strong absorbers and many void areas in the control rod regions which are located in the graphitic side reflector. When using diffusion calculations, a special skill is required to homogenize these ex-core absorber regions. These difficulties arise from the fact that these absorber regions do not contain fission sources while the boundaries of the absorber regions are exposed to very strong neutron currents. This paper shows that a mere conventional homogenization technique is not enough to obtain acceptable results. It is shown that a discontinuity factor-corrected diffusion-solution can be found which yields very good, satisfactory results for fast, whole core diffusion calculations. The verification is done by comparing time-consuming transport calculations [SN-codes or multi-group MCNP-codes] with a fast running diffusion code, corrected by the newly proposed homogenization technique.},
	language = {en},
	number = {2},
	journal = {Nuclear Engineering and Design},
	author = {Xuhua, Zhou and Fu, Li and Dengying, Wang},
	month = feb,
	year = {2009},
	pages = {281--288},
	file = {Xuhua et al. - 2009 - Application of the discontinuity factor theory to .pdf:C\:\\Users\\Sun Myung\\Zotero\\storage\\FIH7QJUF\\Xuhua et al. - 2009 - Application of the discontinuity factor theory to .pdf:application/pdf},
}

@article{guo_improvement_2016,
	title = {Improvement of discontinuity factor for strong absorber region},
	volume = {306},
	issn = {00295493},
	doi = {10.1016/j.nucengdes.2016.04.030},
	abstract = {At Institute of Nuclear and New Energy Technology (INET) the discontinuity factor corrected diffusion method with the homogenization technology was developed and applied in the control rod worth calculation of the pebble bed high temperature gas cooled reactor. But the result with the normal procedure is not accurate enough for a strong absorber. The numerical analysis shows that the strong absorber still has great influence on the flux distribution in the nearby graphite region, so that the flux distribution obtained by the normal diffusion method does not agree with the transport result. Thus, two improvements were proposed in this paper. First, instead of the neutron flux in the middle of the fine mesh, the surface flux of the absorber region was calculated through the net current in the boundary of the region; and then, while the discontinuity factor of the homogenized absorber region should be calculated, the discontinuity factor of the neighboring graphite region on the other side of the interface should also be calculated to eliminate the influence of the strong absorber. The numerical results demonstrate that, based on the improved method, the accuracy of heterogeneous transport calculation can be achieved by a diffusion calculation.},
	language = {en},
	journal = {Nuclear Engineering and Design},
	author = {Guo, Jiong and Li, Fu and Zhang, Han and Zhou, Xiafeng and Fan, Kai and Wang, Lidong and Lu, Jianan},
	month = sep,
	year = {2016},
	pages = {133--137},
	file = {Guo et al. - 2016 - Improvement of discontinuity factor for strong abs.pdf:C\:\\Users\\Sun Myung\\Zotero\\storage\\S335VLVF\\Guo et al. - 2016 - Improvement of discontinuity factor for strong abs.pdf:application/pdf},
}

@article{wen_improved_2022,
	title = {An improved homogenization method for the treatment of strong absorbers in pebble-bed {HTGR}},
	volume = {170},
	issn = {03064549},
	doi = {10.1016/j.anucene.2022.108983},
	abstract = {In pebble-bed high temperature gas-cooled reactors (HTGRs), strong absorbers including control rods and absorber balls are placed in the reflector. The strong absorber regions need to be homogenized for the convenience of downstream whole-core neutronics diffusion calculation. Previously, a homogenization method based on the generalized equivalence theory (GET) was studied. This paper proposes an improved homogenization method by employing a more rigorous discontinuity factor (DF) model, for more accurate treatment of the strong absorbers in the HTGR reflector. The proposed method has been implemented in the PANGU code developed by Institute of Nuclear and New Energy Technology (INET), Tsinghua University. Numerical results suggest that the newly proposed homogenization method achieves good agreement with the heterogeneous Monte Carlo transport solution and experimental result.},
	language = {en},
	journal = {Annals of Nuclear Energy},
	author = {Wen, Yutong and She, Ding and Shi, Lei},
	month = jun,
	year = {2022},
	pages = {108983},
	file = {Wen et al. - 2022 - An improved homogenization method for the treatmen.pdf:C\:\\Users\\Sun Myung\\Zotero\\storage\\RGDA8QZ9\\Wen et al. - 2022 - An improved homogenization method for the treatmen.pdf:application/pdf},
}

@techreport{teuchert_vsop-computer_1980,
	address = {Germany},
	title = {{VSOP}-computer code system for reactor physics and fuel cycle simulation},
	abstract = {VSOP (Very Superior Old Programs) is a system of codes linked together for the simulation
of reactor life histories It comprises neutron cross section libraries and processing
routines, repeated neutron spectrum evaluation, 2-D diffusion calculation based on
neutron flux synthesis with depletion and shutdown features, incore and out-of-pile
fuel management, fuel cycle cost analysis, and thermal hydraulics (at present restricted
to Pebble Bed HTRs) Various techniques have been employed to accelerate the iterative
processes and to optimize the internal data transfer A limitation of the storage requirement
to 360 K-bites is achieved by an overlay structure The code system has been used extensively
for comparison studies of reactors, their fuel cycles, and related detailed features
Beside its use in research and development work for the high temperature reactor the
system has been applied successfully to LWR and Heavy Water Reactors (orig)},
	author = {Teuchert, E. and Hansen, U. and Haas, K.A.},
	year = {1980},
	pages = {90},
}

@techreport{teuchert_vsop94_1994,
	address = {Germany},
	title = {{VSOP}('94) computer code system for reactor physics and fuel cycle simulation},
	abstract = {VSOP (Very Superior Old Programs) is a system of codes linked together for the simulation
of reactor life histories and temporary in-depth research It comprises neutron cross
section libraries and processing routines, repeated neutron spectrum evaluation, 2-D
diffusion calculation with depletion and shut-down features, in-core and out-of-pile
fuel management, fuel cycle cost analysis, and thermal hydraulics (at present restricted
to HTR's) Various techniques have been employed to accelerate the iterative processes
and to optimize the internal data transfer The storage requirement is confined to
17 M-Bytes The code system has extensively been used for comparison studies of reactors,
their fuel cycles, simulation of safety features, developmental research, and reactor
assessments Beside its use in research and development work for the gas cooled High
Temperature Reactor the code has succesfully been applied to Light Water Reactors,
Heavy Water Reactors, and hybride systems with different moderators (orig)},
	number = {0944-2952},
	author = {Teuchert, E. and Haas, K.A. and Ruetten, H.J. and Brockmann, H. and Gerwin, H. and Ohlig, U. and Scherer, W.},
	year = {1994},
	pages = {199},
}

@article{mosher_incident_2006,
	title = {The {Incident} {Flux} {Response} {Expansion} {Method} for {Heterogeneous} {Coarse} {Mesh} {Transport} {Problems}},
	volume = {35},
	issn = {0041-1450, 1532-2424},
	doi = {10.1080/00411450600878615},
	language = {en},
	number = {1-2},
	journal = {Transport Theory and Statistical Physics},
	author = {Mosher, Scott W. and Rahnema, Farzad},
	month = jul,
	year = {2006},
	pages = {55--86},
}

@techreport{rahnema_advanced_2011,
	title = {An {Advanced} {Integrated} {Diffusion}/{Transport} {Method} for the {Design}, {Analysis} and {Optimization} of the {Very}-{High}-{Temperature} {Reactors}},
	url = {http://www.osti.gov/servlets/purl/1011245/},
	language = {en},
	number = {DOE/ID/14821, 1011245},
	author = {Rahnema, Farzad and Zhang, Dingkang and Ougouag, Abderrafi and Gleicher, Frederick},
	month = apr,
	year = {2011},
	pages = {DOE/ID/14821, 1011245},
	file = {Rahnema et al. - 2011 - An Advanced Integrated DiffusionTransport Method .pdf:C\:\\Users\\Sun Myung\\Zotero\\storage\\QIPYJ2FY\\Rahnema et al. - 2011 - An Advanced Integrated DiffusionTransport Method .pdf:application/pdf},
}

@article{ronen_accurate_2004,
	title = {Accurate {Relations} between the {Neutron} {Current} {Densities} and the {Neutron} {Fluxes}},
	volume = {146},
	issn = {0029-5639, 1943-748X},
	doi = {10.13182/NSE04-A2407},
	abstract = {Accurate relations between neutron current densities and neutron flux are obtained using the integral transport equation. Using these relations and Fick’s Law, diffusion constants can be calculated. These diffusion constants are better than those usually used for the cases in which Sa /Ss is not small.},
	language = {en},
	number = {2},
	journal = {Nuclear Science and Engineering},
	author = {Ronen, Yigal},
	month = feb,
	year = {2004},
	pages = {245--247},
	file = {Ronen - 2004 - Accurate Relations between the Neutron Current Den.pdf:C\:\\Users\\Sun Myung\\Zotero\\storage\\982YMN87\\Ronen - 2004 - Accurate Relations between the Neutron Current Den.pdf:application/pdf},
}

@inproceedings{tomatis_application_2011,
	address = {Brazil},
	title = {Application of a numerical transport correction in diffusion calculations},
	abstract = {Full core calculations by ordinary transport methods can demand considerable computational
time, hardly acceptable in the industrial work frame However, the trend of next generation
nuclear cores goes toward more heterogeneous systems, where transport phenomena of
neutrons become very important On the other hand, using diffusion solvers is more
practical allowing faster calculations, but a specific formulation of the diffusion
coefficient is requested to reproduce the scalar flux with reliable physical accuracy
In this paper, the Ronen method is used to evaluate numerically the diffusion coefficient
in the slab reactor The new diffusion solution is driven toward the solution of the
integral neutron transport equation by non linear iterations Better estimates of currents
are computed and diffusion coefficients are corrected at node interfaces, still assuming
Fick's law This method enables obtaining closer results to the transport solution
by a common solver in multigroup diffusion (author)},
	author = {Tomatis, Daniele and Dall'Osso, Aldo},
	year = {2011},
}

@article{tomatis_ronen_2021,
	title = {On the {Ronen} {Method} in {Simple} 1-{D} {Geometries} for {Neutron} {Transport} {Theory} {Solutions}},
	volume = {50},
	issn = {2332-4309},
	doi = {10.1080/23324309.2020.1843496},
	abstract = {In this work, we apply the Ronen method to obtain highly-accurate approximations to the solution of the neutron transport equation in simple homogeneous problems. Slab, cylindrical, and spherical geometries are studied. This method demands successive resolutions of the diffusion equation, where the local diffusion constants are modified in order to reproduce new estimates of the currents by a transport operator. The diffusion solver employs here finite differences and the transport-corrected currents are forced in the numerical scheme by means of drift terms, like in the CMFD scheme. Boundary conditions are discussed introducing proper approximations to save the particle balance in case of reflection in the slab. The solution from the Ronen iterations is compared against reference results provided by the collision probability method. More accurate estimates of the currents are provided by integral transport equations using first flight escape probabilities. Slow convergence on the scalar flux is analyzed, although the results match the reference solutions in the limit of fine meshes and far from the bare boundary.},
	number = {2},
	journal = {Journal of Computational and Theoretical Transport},
	author = {Tomatis, Daniele and Gross, Roy and Gilad, Erez},
	month = feb,
	year = {2021},
	keywords = {neutron transport, CMFD, collision probability method, Ronen method},
	pages = {134--157},
	file = {Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\6XJJDHPA\\23324309.2020.html:text/html},
}

@article{gross_high-accuracy_2020,
	title = {High-accuracy neutron diffusion calculations based on integral transport theory},
	volume = {135},
	issn = {2190-5444},
	doi = {10.1140/epjp/s13360-020-00216-y},
	abstract = {In this paper, the Ronen method is developed, implemented, and applied to resolve the neutron flux and the criticality eigenvalue in simple one-dimensional homogeneous and heterogeneous problems. The Ronen method is based on iterative calculations of correction factors to use in a multigroup diffusion model, where the factors are actually given by the integral transport equation. In particular, spatially dependent diffusion constants are modified locally in order to reproduce new estimates of the surface currents obtained by the integral transport operator. The diffusion solver employed in this study uses finite differences, and the transport-corrected currents are introduced into the numerical scheme as drift terms. The corrected solutions are compared against reference results from a discrete ordinate code. The results match well with the reference solutions, especially in the limit of fine meshes, but slow convergence of the scalar flux is reported.},
	language = {en},
	number = {2},
	journal = {The European Physical Journal Plus},
	author = {Gross, Roy and Tomatis, Daniele and Gilad, Erez},
	month = feb,
	year = {2020},
	pages = {235},
	file = {Gross et al. - 2020 - High-accuracy neutron diffusion calculations based.pdf:C\:\\Users\\Sun Myung\\Zotero\\storage\\ITS242IC\\Gross et al. - 2020 - High-accuracy neutron diffusion calculations based.pdf:application/pdf},
}

@book{lewis_computational_1984,
	address = {United States},
	title = {Computational methods of neutron transport},
	abstract = {This books presents a balanced overview of the major methods currently available for
obtaining numerical solutions in neutron and gamma ray transport It focuses on methods
particularly suited to the complex problems encountered in the analysis of reactors,
fusion devices, radiation shielding, and other nuclear systems Derivations are given
for each of the methods showing how the transport equation is reduced to sets of algebraic
equations suitable for solution on a digital computer The limitations of the methods
and their suitability for different classes of problems are discussed in terms of
computer memory, time requirements, and accuracy},
	publisher = {John Wiley and Sons, Inc},
	author = {Lewis, E.E. and Miller, W.F.},
	year = {1984},
}

@article{pounders_diffusion_2009,
	title = {On the {Diffusion} {Coefficients} for {Reactor} {Physics} {Applications}},
	volume = {163},
	issn = {0029-5639, 1943-748X},
	doi = {10.13182/NSE163-243},
	abstract = {The definition of the multigroup diffusion coefficient for reactor physics problems is not unique; rather, it is based on limiting approximations made to the Boltzmann transport equation. In this paper, we present several new diffusion closures in an attempt to gain increased accuracy over the standard P1based diffusion theory. First, the Levermore-Pomraning flux-limited diffusion theory is applied to reactor physics problems both in its original form and in a new modified form that makes the methodology more robust with respect to the energy variable. Additionally, two novel definitions of the diffusion coefficient are introduced that permit a neutron flux that is greater than first order in angle. These various diffusion theories are completed by developing consistent boundary conditions for each case. Diffusion theory solutions are computed for each unique closure and are compared against transport theory analytically for a simple half-space problem and numerically for a suite of simplified one-dimensional reactor problems. Conclusions and observations are made for each diffusion method in terms of its underlying assumptions and accuracy of the benchmark solutions.},
	language = {en},
	number = {3},
	journal = {Nuclear Science and Engineering},
	author = {Pounders, Justin M. and Rahnema, Farzad},
	month = nov,
	year = {2009},
	pages = {243--262},
	file = {Pounders and Rahnema - 2009 - On the Diffusion Coefficients for Reactor Physics .pdf:C\:\\Users\\Sun Myung\\Zotero\\storage\\EW97CFDR\\Pounders and Rahnema - 2009 - On the Diffusion Coefficients for Reactor Physics .pdf:application/pdf},
}

@article{davison_influence_1951,
	title = {Influence of a {Black} {Sphere} and of a {Black} {Cylinder} upon the {Neutron} {Density} in an {Infinite} {Non}-{Capturing} {Medium}},
	volume = {64},
	issn = {0370-1298},
	doi = {10.1088/0370-1298/64/10/305},
	abstract = {For a black body inserted into a homogeneous scattering medium, the neutron density at large distances from the body IS determined using the perturbation method for the case when the scattering in the scattering medium is assumed elastic and isotropic capture is absent and the following four types of the black body are considered : (1) a sphere of radius small compared with the mean free path in the scattering mehum; (ii)a sphere of radius large compared with the mean free path; (in) an infinitelylong cylinder ofradius small compared with the mean free path; (IV)an infinitely long cylinder of rados large compared with the mean free path. Combining the results for the above limiting Cases with the results obtained by other methods the distribution is also estimated for a sphere and an infinitely long cylinder of intermediate radii.},
	language = {en},
	number = {10},
	journal = {Proceedings of the Physical Society. Section A},
	author = {Davison, B},
	month = oct,
	year = {1951},
	pages = {881--902},
	file = {Davison - 1951 - Influence of a Black Sphere and of a Black Cylinde.pdf:C\:\\Users\\Sun Myung\\Zotero\\storage\\B34G9D4G\\Davison - 1951 - Influence of a Black Sphere and of a Black Cylinde.pdf:application/pdf},
}

@article{maynard_blackness_1959,
	title = {Blackness {Theory} and {Coefficients} for {Slab} {Geometry}},
	volume = {6},
	issn = {0029-5639, 1943-748X},
	doi = {10.13182/NSE59-A25657},
	language = {en},
	number = {3},
	journal = {Nuclear Science and Engineering},
	author = {Maynard, C. W.},
	month = sep,
	year = {1959},
	pages = {174--186},
	file = {Maynard - 1959 - Blackness Theory and Coefficients for Slab Geometr.pdf:C\:\\Users\\Sun Myung\\Zotero\\storage\\BP66U23Q\\Maynard - 1959 - Blackness Theory and Coefficients for Slab Geometr.pdf:application/pdf},
}

@techreport{mendelson_two-dimensional_1969,
	title = {{TWO}-{DIMENSIONAL} {BLACKNESS} {THEORY} {BOUNDARY} {CONDITIONS}: {RZ} {GEOMETRY}.},
	shorttitle = {{TWO}-{DIMENSIONAL} {BLACKNESS} {THEORY} {BOUNDARY} {CONDITIONS}},
	url = {https://www.osti.gov/biblio/4757861-two-dimensional-blackness-theory-boundary-conditions-rz-geometry},
	abstract = {The U.S. Department of Energy's Office of Scientific and Technical Information},
	language = {English},
	number = {KAPL-P-3830; CONF-690609-9},
	institution = {Knolls Atomic Power Lab. (KAPL), Niskayuna, NY (United States)},
	author = {Mendelson, M. R. and Storm, M. L. and Gazda, S. A.},
	month = oct,
	year = {1969},
	file = {Full Text PDF:C\:\\Users\\Sun Myung\\Zotero\\storage\\LR6CEGIJ\\Mendelson et al. - 1969 - TWO-DIMENSIONAL BLACKNESS THEORY BOUNDARY CONDITIO.pdf:application/pdf;Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\8X6M6VXE\\4757861-two-dimensional-blackness-theory-boundary-conditions-rz-geometry.html:text/html},
}

@article{spinks_extrapolation_1965,
	title = {The {Extrapolation} {Distance} at the {Surface} of a {Grey} {Cylindrical} {Control} {Rod}},
	volume = {22},
	issn = {0029-5639, 1943-748X},
	doi = {10.13182/NSE65-A19765},
	language = {en},
	number = {1},
	journal = {Nuclear Science and Engineering},
	author = {Spinks, N.},
	month = may,
	year = {1965},
	pages = {87--93},
	file = {Spinks - 1965 - The Extrapolation Distance at the Surface of a Gre.pdf:C\:\\Users\\Sun Myung\\Zotero\\storage\\I43G6Y3W\\Spinks - 1965 - The Extrapolation Distance at the Surface of a Gre.pdf:application/pdf},
}

@techreport{bretscher_computing_1997,
	address = {Lemont, IL},
	title = {{COMPUTING} {CONTROL} {ROD} {WORTHS} {IN} {THERMAL} {RESEARCH} {REACTORS}},
	abstract = {Simple diffusion theory cannot be used to evaluate control rod worths in thermal reactors because of steep flux gradients that occur within the absorber material. However, reliable control rod worths can be calculated within the framework of diffusion theory if the control material is characterized by a set of mesh-dependent effective diffusion parameters or if group-dependent current-to-flux ratios are specified on the absorber surface.},
	language = {en},
	number = {ANL/RERTR/TM-5},
	institution = {Argonne National Laboratory (ANL)},
	author = {Bretscher, M M},
	month = feb,
	year = {1997},
	pages = {39},
	file = {Bretscher - COMPUTING CONTROL ROD WORTHS IN THERMAL RESEARCH R.pdf:C\:\\Users\\Sun Myung\\Zotero\\storage\\A2FF45BQ\\Bretscher - COMPUTING CONTROL ROD WORTHS IN THERMAL RESEARCH R.pdf:application/pdf},
}

@article{marshak_note_1947,
	title = {Note on the {Spherical} {Harmonic} {Method} {As} {Applied} to the {Milne} {Problem} for a {Sphere}},
	volume = {71},
	issn = {0031-899X},
	doi = {10.1103/PhysRev.71.443},
	language = {en},
	number = {7},
	journal = {Physical Review},
	author = {Marshak, R. E.},
	month = apr,
	year = {1947},
	pages = {443--446},
	file = {Marshak - 1947 - Note on the Spherical Harmonic Method As Applied t.pdf:C\:\\Users\\Sun Myung\\Zotero\\storage\\CG2L2WVX\\Marshak - 1947 - Note on the Spherical Harmonic Method As Applied t.pdf:application/pdf},
}

@article{pellaud_extrapolation_1968,
	title = {The {Extrapolation} {Distance} for a {Black} or {Grey} {Cylindrical} {Neutron} {Absorber} by the {Spherical} {Harmonics} {Method}},
	volume = {33},
	issn = {0029-5639, 1943-748X},
	doi = {10.13182/NSE68-A20655},
	language = {en},
	number = {2},
	journal = {Nuclear Science and Engineering},
	author = {Pellaud, Bruno},
	month = aug,
	year = {1968},
	pages = {169--186},
	file = {Pellaud - 1968 - The Extrapolation Distance for a Black or Grey Cyl.pdf:C\:\\Users\\Sun Myung\\Zotero\\storage\\ZXETFPDL\\Pellaud - 1968 - The Extrapolation Distance for a Black or Grey Cyl.pdf:application/pdf},
}

@article{smith_nodal_1983,
	title = {Nodal method storage reduction by nonlinear iteration},
	issn = {0003-018X},
	journal = {Transactions of the American Nuclear Society},
	author = {Smith, K.S.},
	year = {1983},
	pages = {265--266},
}

@article{smith_assembly_1986,
	title = {Assembly homogenization techniques for light water reactor analysis},
	volume = {17},
	issn = {01491970},
	doi = {10.1016/0149-1970(86)90035-1},
	abstract = {Recent progress in development and application of advanced assembly homogenization methods for light water reactor analysis is reviewed. Practical difficulties arising from conventional flux-weighting approximations are discussed and numerical examples given. The mathematical foundations for homogenization methods are outlined. Two methods, Equivalence Theory and Generalized Equivalence Theory which are theoretically capable of eliminating homogenization error are reviewed. Practical means of obtaining approximate homogenized parameters are presented and numerical examples are used to contrast the two methods.},
	language = {en},
	number = {3},
	journal = {Progress in Nuclear Energy},
	author = {Smith, K.S.},
	month = jan,
	year = {1986},
	pages = {303--335},
	file = {Smith - 1986 - Assembly homogenization techniques for light water.pdf:C\:\\Users\\Sun Myung\\Zotero\\storage\\U8ZNKAX6\\Smith - 1986 - Assembly homogenization techniques for light water.pdf:application/pdf},
}

@article{yang_development_2022,
	title = {Development of coupled {PROTEUS}-{NODAL} and {SAM} code system for multiphysics analysis of molten salt reactors},
	volume = {168},
	issn = {03064549},
	doi = {10.1016/j.anucene.2021.108889},
	abstract = {This paper presents the coupled code system of PROTEUS-NODAL and SAM developed for multiphysics simulation of molten salt reactors (MSRs). The coupling was made under the MOOSE framework by developing a MOOSE wrapper of PROTEUS-NODAL and a master application that controls PROTEUSNODAL neutronics and SAM thermal-fluidic calculations. The coupled code system was verified and validated using the Molten Salt Fast Reactor (MSFR) benchmark problem and the available experimental data of the Molten Salt Reactor Experiments (MSRE). Steady-state solutions for the MSFR benchmark agreed well with those of manually coupled calculations of PROTEUS-NODAL and ANSYS CFX, and transient solutions agreed well with the open literature results. The validation test results for the pump startup, pump coast-down, and natural circulation tests of MSRE agreed well with the measured values. These results indicate that the coupled code system of PROTEUS-NODAL and SAM can be used reliably for steady-state and transient analyses of MSRs.},
	language = {en},
	journal = {Annals of Nuclear Energy},
	author = {Yang, Gang and Jaradat, Mustafa K. and Sik Yang, Won and Lee, Changho},
	month = apr,
	year = {2022},
	pages = {108889},
	file = {Yang et al. - 2022 - Development of coupled PROTEUS-NODAL and SAM code .pdf:C\:\\Users\\Sun Myung\\Zotero\\storage\\YDRXBTHY\\Yang et al. - 2022 - Development of coupled PROTEUS-NODAL and SAM code .pdf:application/pdf},
}

@techreport{koebke_new_1980,
	address = {International Atomic Energy Agency (IAEA)},
	title = {A new approach to homogenization and group condensation},
	abstract = {A coarse-mesh few group neutron flux distribution in a reactor core should conserve
several integral quantities of the heterogeneous representation in every homogenized
region The homogenization theory described in this paper will be called ''equivalence
theory'' All the information needed for the equivalent coarse-mesh calculation is
taken from multigroup spectral calculations of heterogeneous assemblies in their ''typical''
environment A self-consistent system of differential equations and boundary conditions
for the determination of equivalent group parameters is derived The theory is based
on the postulate, that the integral reaction rates, average fluxes and average leakages
are conserved The progress achieved is demonstrated by applying the proposed scheme
to an idealized heterogeneous BWR two-group problem defined by Henry and Worley For
this problem equivalent group parameters in a two-group and a one-group representation
are determined The numerical solution of the equivalence equations for this reactor
problem reproduces the integral quantities within five digits},
	author = {Koebke, K.},
	year = {1980},
	pages = {303--322},
}

@inproceedings{cabellos_generalized_1996,
	address = {Japan},
	title = {Generalized effects in two-group cross sections and discontinuity factors for {PWR}'s},
	abstract = {In order to improve the internal core 3-D calculation, a extended parametrization
of 2-group cross sections and discontinuity factors have been developed and implemented
in the SEANAP system We have introduced additional 'generalized state' variables:
a new spectral history index, equivalencing the different burnup paths in parameter
space, and new boundary indexes, to account for the intragroup spectral effects due
to changes in the boundary conditions by the different kinds of neighbor cells or
nodes The new library per cell type including fuel pins at different locations or
with gadolinium, water tubes, control rods and burnable absorber tubes, is used by
COBAYA for cell-by-cell 2D diffusion calculations (author)},
	author = {Cabellos, O. and Ahnert, C. and Aragones, J.M.},
	year = {1996},
	pages = {615},
}

@inproceedings{park_pin-cell_2002,
	title = {Pin-{Cell} {Homogenization} via {Generalized} {Equivalence} {Theory} and {Embedding} {Assembly} {Calculation}},
	volume = {85},
	language = {English},
	booktitle = {Transactions of the {American} {Nuclear} {Society}},
	publisher = {American Nuclear Society},
	author = {Park, K. W. and Park, C. J. and Lee, K. T. and Cho, Nam-Zin and Kim, Y. H.},
	year = {2002},
	pages = {334},
	file = {Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\AFFRCJZU\\79713.html:text/html},
}

@article{aragones_linear_1986,
	title = {A {Linear} {Discontinuous} {Finite} {Difference} {Formulation} for {Synthetic} {Coarse}-{Mesh} {Few}-{Group} {Diffusion} {Calculations}},
	volume = {94},
	issn = {0029-5639, 1943-748X},
	doi = {10.13182/NSE86-A18343},
	abstract = {A linear discontinuous finite difference formulation to solve the diffusion equations in coarse mesh and few groups is developed. The correction factors for heterogeneities, coarse mesh, and spectral effects are general interface flux discontinuity factors that can be explicitly calculated (synthetized) from detailed diffusion or transport solutions in fine mesh (heterogeneous) and multigroups, preserving the integrated fluxes and interface net currents. The stability is explicitly established for general synthetizations and for specific fine to coarse mesh and group reductions. Computing methods have been implemented for one-group (grey) synthetic diffusion acceleration, two-dimensional nodal/local solutions, and three-dimensional nodal simulation of pressurized water reactor cores. Results demonstrate the simplicity and stability of the formulation, a regular behavior of the correction factors, an outstanding acceleration performance, and high potential for parallel and vector computing.},
	language = {en},
	number = {4},
	journal = {Nuclear Science and Engineering},
	author = {Aragonés, José M. and Ahnert, Carol},
	month = dec,
	year = {1986},
	pages = {309--322},
	file = {Aragonés and Ahnert - 1986 - A Linear Discontinuous Finite Difference Formulati.pdf:C\:\\Users\\Sun Myung\\Zotero\\storage\\RIQRDVDG\\Aragonés and Ahnert - 1986 - A Linear Discontinuous Finite Difference Formulati.pdf:application/pdf},
}

@techreport{kavenoky_sph_1978,
	address = {France},
	title = {The {SPH} homogeneization method},
	abstract = {The homogeneization of a uniform lattice is a rather well understood topic while difficult
problems arise if the lattice becomes irregular The SPH homogeneization method is
an attempt to generate homogeneized cross sections for an irregular lattice Section
1 summarizes the treatment of an isolated cylindrical cell with an entering surface
current (in one velocity theory); Section 2 is devoted to the extension of the SPH
method to assembly problems Finally Section 3 presents the generalisation to general
multigroup problems Numerical results are obtained for a PXR rod bundle assembly in
Section 4},
	author = {Kavenoky, Alain},
	year = {1978},
	pages = {8},
}

@article{hebert_consistent_1991,
	title = {A {Consistent} {Technique} for the {Global} {Homogenization} of a {Pressurized} {Water} {Reactor} {Assembly}},
	volume = {109},
	issn = {0029-5639, 1943-748X},
	doi = {10.13182/NSE109-360},
	abstract = {Proposals are made for improving the homogeneous diffusion procedure for reactor operating and design calculations. The procedure is based on the use of homogeneous properties for the assemblies and on high-order discretization for the reactor diffusion calculations. It proposes to improve the radial leakage model incorporated in the flux calculation and to introduce a superhomogeneisation equivalence technique between the flux and cross-section edit calculations to yield homogeneous diffusion properties consistent with the control rod worth calculations.},
	language = {en},
	number = {4},
	journal = {Nuclear Science and Engineering},
	author = {Hébert, Alain and Benoist, Pierre},
	month = dec,
	year = {1991},
	pages = {360--372},
	file = {Hébert and Benoist - 1991 - A Consistent Technique for the Global Homogenizati.pdf:C\:\\Users\\Sun Myung\\Zotero\\storage\\RWKEFIN7\\Hébert and Benoist - 1991 - A Consistent Technique for the Global Homogenizati.pdf:application/pdf},
}

@article{yamamoto_cell_2004,
	title = {Cell homogenization methods for pin-by-pin core calculations tested in slab geometry},
	volume = {31},
	issn = {03064549},
	doi = {10.1016/j.anucene.2003.12.001},
	abstract = {In this paper, performances of spatial homogenization methods for fuel or non-fuel cells are compared in slab geometry in order to facilitate pin-by-pin core calculations. Since the spatial homogenization methods were mainly developed for fuel assemblies, systematic study of their performance for the cell-level homogenization has not been carried out. Importance of celllevel homogenization is recently increasing since the pin-by-pin mesh core calculation in actual three-dimensional geometry, which is less approximate approach than current advanced nodal method, is getting feasible. Four homogenization methods were investigated in this paper; the flux-volume weighting, the generalized equivalence theory, the superhomogenization (SPH) method and the nonlinear iteration method. The last one, the nonlinear iteration method, was tested as the homogenization method for the first time. The calculations were carried out in simplified colorset assembly configurations of PWR, which are simulated by slab geometries, and homogenization performances were evaluated through comparison with the reference cell-heterogeneous calculations. The calculation results revealed that the generalized equivalence theory showed best performance. Though the nonlinear iteration method can significantly reduce homogenization error, its performance was not as good as that of the generalized equivalence theory. Through comparison of the results obtained by the generalized equivalence theory and the superhomogenization method, important byproduct was obtained; deficiency of the current superhomogenization method, which could be improved by incorporating the ‘‘cell-level discontinuity factor between assemblies’’, was clarified.},
	language = {en},
	number = {8},
	journal = {Annals of Nuclear Energy},
	author = {Yamamoto, Akio and Kitamura, Yasunori and Yamane, Yoshihiro},
	month = may,
	year = {2004},
	pages = {825--847},
	file = {Yamamoto et al. - 2004 - Cell homogenization methods for pin-by-pin core ca.pdf:C\:\\Users\\Sun Myung\\Zotero\\storage\\W4BT9RSZ\\Yamamoto et al. - 2004 - Cell homogenization methods for pin-by-pin core ca.pdf:application/pdf},
}

@article{ortensi_newton_2018,
	title = {A {Newton} solution for the {Superhomogenization} method: {The} {PJFNK}-{SPH}},
	volume = {111},
	issn = {0306-4549},
	shorttitle = {A {Newton} solution for the {Superhomogenization} method},
	doi = {10.1016/j.anucene.2017.09.027},
	abstract = {This work presents two novel topics regarding the Superhomogenization method: 1) the formalism for the implementation of the method with the linear Boltzmann Transport Equation, and 2) a Newton algorithm for the solution of the nonlinear problem that arises from the method. These new ideas have been implemented in a continuous finite element discretization in the MAMMOTH reactor physics application. The traditional solution strategy for this nonlinear problem uses a Picard, fixed-point iterative process whereas the new implementation relies on MOOSE’s Preconditioned Jacobian-Free Newton Krylov method to allow for a direct solution. The PJFNK-SPH can converge problems that were either intractable or very difficult to converge with the traditional iterative approach, including geometries with reflectors and vacuum boundary conditions. This is partly due to the underlying Scalable Nonlinear Equations Solvers in PETSc, which are integral to MOOSE and offer Newton damping, line search and trust region methods. The PJFNK-SPH has been implemented and tested for various discretizations of the transport equation included in the Rattlesnake transport solver. Speedups of five times for diffusion and ten to fifteen times for transport were obtained when compared to the traditional Picard approach. The three test problems cover a wide range of applications including a standard Pressurized Water Reactor lattice with control rods, a Transient Reactor Test facility control rod supercell and a prototype fast-thermal reactor. The reference solutions and initial cross sections were obtained from the Serpent 2 Monte Carlo code. The SPH-corrected cross sections yield eigenvalues that are near exact, relative to reference solutions, for reflected geometries, even with reflector regions. In geometries with vacuum boundary conditions the accuracy is problem dependent and solutions can be within a few to a few hundred pcm. The root-mean-square error in the power distribution is below 0.8\% of the reference Monte Carlo. There is little benefit from SPH-corrected transport in typical scoping calculations, but for more detailed analyses it can yield superior convergence of the solution in some of the test problems. This PJFNK-SPH approach is currently being used in the modeling of the Transient Test Reactor at Idaho National Laboratory, where full reactor core SPH-corrected cross sections are employed to reduce the homogenization errors in transient or multi-physics calculations. This base implementation of the PJFNK-SPH provides an extremely robust solver and a springboard to further improve the Superhomogenization method in order to better preserve neutron currents, one of the primary deficiencies of the method.},
	language = {en},
	journal = {Annals of Nuclear Energy},
	author = {Ortensi, Javier and Wang, Yaqi and Laurier, Alexandre and Schunert, Sebastian and Hébert, Alain and DeHart, Mark},
	month = jan,
	year = {2018},
	keywords = {Homogenization, MOOSE, Rattlesnake, Equivalence, PJFNK, Superhomogenization},
	pages = {579--594},
	file = {ScienceDirect Full Text PDF:C\:\\Users\\Sun Myung\\Zotero\\storage\\P6G2S7Z5\\Ortensi et al. - 2018 - A Newton solution for the Superhomogenization meth.pdf:application/pdf;ScienceDirect Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\GBDHKWR6\\S0306454917303079.html:text/html},
}

@article{tosolin_synthesis_2018,
	title = {Synthesis of plutonium trifluoride by hydro-fluorination and novel thermodynamic data for the {PuF3}-{LiF} system},
	volume = {503},
	issn = {0022-3115},
	doi = {10.1016/j.jnucmat.2018.02.037},
	abstract = {PuF3 was synthetized by hydro-fluorination of PuO2 and subsequent reduction of the product by hydrogenation. The obtained PuF3 was analysed by X-Ray Diffraction (XRD) and found phase-pure. High purity was also confirmed by the melting point analysis using Differential Scanning Calorimetry (DSC). PuF3 was then used for thermodynamic assessment of the PuF3-LiF system. Phase equilibrium points and enthalpy of fusion of the eutectic composition were measured by DSC. XRD analyses of selected samples after DSC measurement confirm that after solidification from the liquid, the system returns to a mixture of LiF and PuF3.},
	language = {en},
	journal = {Journal of Nuclear Materials},
	author = {Tosolin, A. and Souček, P. and Beneš, O. and Vigier, J. -F. and Luzzi, L. and Konings, R. J. M.},
	month = may,
	year = {2018},
	keywords = {Molten salt reactor, Thermodynamics, Actinide fluorides, Fluorination, PuF synthesis, PuF-LiF phase diagram},
	pages = {171--177},
	file = {ScienceDirect Full Text PDF:C\:\\Users\\Sun Myung\\Zotero\\storage\\E6FBPGPE\\Tosolin et al. - 2018 - Synthesis of plutonium trifluoride by hydro-fluori.pdf:application/pdf;ScienceDirect Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\AHCVN22S\\S0022311517314757.html:text/html},
}

@article{duran-klie_dynamic_2016,
	title = {Dynamic {Reference} {Electrode} development for redox potential measurements in fluoride molten salt at high temperature},
	volume = {195},
	issn = {0013-4686},
	doi = {10.1016/j.electacta.2016.02.042},
	abstract = {Measurement of redox potential in fluoride media is a major problem due to the difficulty to design a reference electrode with high stability, high mechanical resistance and high accuracy. In the frame of molten salt reactor studies, a dynamic reference electrode (DRE) is developed to measure redox potential in fluoride molten salt at high temperature. DRE is based on the in-situ generation of a transient redox system. The choice of the redox couple corresponds to the cathodic limit of the molten salt considered. As a preliminary step, the demonstration of feasibility of generating a DRE was done in LiF-NaF-KF (46.5–11.5–42mol\%) media at 500°C. In this salt, the reference redox system generated by coulometry at applied current is KF/K, metallic potassium being electrodeposited on a tungsten wire electrode. The validation of the DRE response and the experimental optimization parameters for DRE generation were realized by following the NiF2/Ni redox potential evolution as a function of NiF2 concentration in the fused salt. The current value applied for DRE generation was optimized. It depends on the amount of metallic cations contained in the fused salt and which can be electrochemically reduced simultaneously during the DRE generation. The current corresponding to the DRE generation has to be 4 times greater than the current corresponding to the reduction of the other elements.},
	language = {en},
	journal = {Electrochimica Acta},
	author = {Durán-Klie, Gabriela and Rodrigues, Davide and Delpech, Sylvie},
	month = mar,
	year = {2016},
	keywords = {molten salt reactor, corrosion},
	pages = {19--26},
	file = {ScienceDirect Full Text PDF:C\:\\Users\\Sun Myung\\Zotero\\storage\\TU7JR65E\\Durán-Klie et al. - 2016 - Dynamic Reference Electrode development for redox .pdf:application/pdf;ScienceDirect Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\ZWKTGIJG\\S0013468616303073.html:text/html},
}

@article{soucek_synthesis_2017,
	title = {Synthesis of {UF4} and {ThF4} by {HF} gas fluorination and re-determination of the {UF4} melting point},
	volume = {200},
	issn = {0022-1139},
	doi = {10.1016/j.jfluchem.2017.05.011},
	abstract = {Basic thermodynamic and electrochemical data of pure actinide fluorides and their mixtures are required for the design and safety assessment of any presently studied molten salt reactor concept based on molten fluoride salt fuel. Since the actinide fluorides are usually not produced commercially, they have to be prepared from the available input materials, typically oxides. In this work, a specially designed facility for synthesis of pure actinide fluorides using pure HF gas is described, as well as a complete procedure of synthesis and characterisation of pure UF4 and ThF4. The fluorination installation consists of a glove box kept under a purified argon atmosphere, a high temperature horizontal fluorination reactor and a HF supply gas line connected to the glove box. The fluorides were synthesised from high specific surface oxides prepared from the respective oxalates by low temperature calcination. The fluorination was partly stationary and partly in a HF gas flow, based on a heterogeneous powder-gas reaction at high temperatures. The products were characterised by X-ray diffraction and differential scanning calorimetry, which confirmed high purity products obtained by this method. Moreover, the melting point of UF4 was revised using a pure sample and a new value is suggested.},
	language = {en},
	journal = {Journal of Fluorine Chemistry},
	author = {Souček, Pavel and Beneš, Ondřej and Claux, Benoit and Capelli, Elisa and Ougier, Michel and Tyrpekl, Václav and Vigier, Jean-Francois and Konings, Rudy J. M.},
	month = aug,
	year = {2017},
	keywords = {Molten salt reactor, Actinide fluorides, Fluorination, Hydrogen fluorine, ThF synthesis, UF synthesis},
	pages = {33--40},
	file = {ScienceDirect Full Text PDF:C\:\\Users\\Sun Myung\\Zotero\\storage\\8FZCJMDP\\Souček et al. - 2017 - Synthesis of UF4 and ThF4 by HF gas fluorination a.pdf:application/pdf;ScienceDirect Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\HDBBF2HI\\S0022113917301690.html:text/html},
}

@article{rodrigues_pyrochemical_2015,
	title = {Pyrochemical reprocessing of molten salt fast reactor fuel: focus on the reductive extraction step},
	volume = {60},
	shorttitle = {Pyrochemical reprocessing of molten salt fast reactor fuel},
	doi = {10.1515/nuka-2015-0153},
	abstract = {Abstract The nuclear fuel reprocessing is a prerequisite for nuclear energy to be a clean and sustainable energy. In the case of the molten salt reactor containing a liquid fuel, pyrometallurgical way is an obvious way. The method for treatment of the liquid fuel is divided into two parts. In-situ injection of helium gas into the fuel leads to extract the gaseous fission products and a part of the noble metals. The second part of the reprocessing is performed by ‘batch’. It aims to recover the fissile material and to separate the minor actinides from fission products. The reprocessing involves several chemical steps based on redox and acido-basic properties of the various elements contained in the fuel salt. One challenge is to perform a selective extraction of actinides and lanthanides in spent liquid fuel. Extraction of actinides and lanthanides are successively performed by a reductive extraction in liquid bismuth pool containing metallic lithium as a reductive reagent. The objective of this paper is to give a description of the several steps of the reprocessing retained for the molten salt fast reactor (MSFR) concept and to present the initial results obtained for the reductive extraction experiments realized in static conditions by},
	language = {en},
	number = {4},
	journal = {Nukleonika},
	author = {Rodrigues, Davide and Durán-Klie, Gabriela and Delpech, Sylvie},
	month = nov,
	year = {2015},
	pages = {907--914},
	file = {Full Text PDF:C\:\\Users\\Sun Myung\\Zotero\\storage\\67SXNIXI\\Rodrigues et al. - 2015 - Pyrochemical reprocessing of molten salt fast reac.pdf:application/pdf;Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\4NZDNM4N\\nuka-2015-0153.html:text/html},
}

@article{tosolin_vaporization_2019,
	title = {Vaporization behaviour of a {PuF3}-containing fuel mixture for the {Molten} {Salt} {Fast} {Reactor}},
	volume = {527},
	issn = {0022-3115},
	doi = {10.1016/j.jnucmat.2019.151780},
	abstract = {The mixture LiF-ThF4-UF4-PuF3 (77.5–6.6-12.3–3.6 mol\%), a fuel option for the Molten Salt Fast Reactor (MSFR), has been measured by differential scanning calorimetry (DSC) for determination of phase transitions, and by Knudsen effusion mass spectrometry (KEMS) for measurement of partial vapour pressures. The boiling point of the mixture was determined by extrapolation of total vapour pressure to 1 bar. Thermodynamic calculations were performed and compared with the experimental results. Novel experimental data on pure plutonium trifluoride are presented: melting point, vaporization enthalpy, vapour pressure and ionization energies by electron impact.},
	language = {en},
	journal = {Journal of Nuclear Materials},
	author = {Tosolin, A. and Colle, J. -Y. and Mastromarino, S. and Souček, P. and Luzzi, L. and Konings, R. J. M. and Beneš, O.},
	month = dec,
	year = {2019},
	keywords = {Molten salt reactor, Fluoride fuel, Knudsen effusion mass spectrometry, Plutonium trifluoride, Vapour pressure},
	pages = {151780},
	file = {ScienceDirect Full Text PDF:C\:\\Users\\Sun Myung\\Zotero\\storage\\N7YJ5H87\\Tosolin et al. - 2019 - Vaporization behaviour of a PuF3-containing fuel m.pdf:application/pdf;ScienceDirect Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\MUJISLBC\\S0022311519308293.html:text/html},
}

@article{tano_progress_2017,
	title = {Progress in modeling solidification in molten salt coolants},
	volume = {25},
	issn = {0965-0393},
	doi = {10.1088/1361-651X/aa8345},
	abstract = {Molten salts have been proposed as heat carrier media in the nuclear and concentrating solar power plants. Due to their high melting temperature, solidification of the salts is expected to occur during routine and accidental scenarios. Furthermore, passive safety systems based on the solidification of these salts are being studied. The following article presents new developments in the modeling of eutectic molten salts by means of a multiphase, multicomponent, phase-field model. Besides, an application of this methodology for the eutectic solidification process of the ternary system LiF–KF–NaF is presented. The model predictions are compared with a newly developed semi-analytical solution for directional eutectic solidification at stable growth rate. A good qualitative agreement is obtained between the two approaches. The results obtained with the phase-field model are then used for calculating the homogenized properties of the solid phase distribution. These properties can then be included in a mixture macroscale model, more suitable for industrial applications.},
	language = {en},
	number = {7},
	journal = {Modelling and Simulation in Materials Science and Engineering},
	author = {Tano, Mauricio and Rubiolo, Pablo and Doche, Olivier},
	month = aug,
	year = {2017},
	pages = {074001},
	file = {Submitted Version:C\:\\Users\\Sun Myung\\Zotero\\storage\\KGAHM2JW\\Tano et al. - 2017 - Progress in modeling solidification in molten salt.pdf:application/pdf},
}

@article{tiberga_preliminary_2019,
	title = {Preliminary investigation on the melting behavior of a freeze-valve for the {Molten} {Salt} {Fast} {Reactor}},
	volume = {132},
	issn = {0306-4549},
	doi = {10.1016/j.anucene.2019.06.039},
	abstract = {This paper focuses on the freeze-plug, a key safety component of the Molten Salt Fast Reactor, one of the Gen. IV nuclear reactors that must excel in safety, reliability, and sustainability. The freeze-plug is a valve made of frozen fuel salt, designed to melt when an event requiring the core drainage occurs. Melting and draining must be passive, relying on decay heat and gravity, and must occur before the reactor incurs structural damage. In this work, we preliminarily investigate the freeze-plug melting behavior, assessing the influence of various design configurations and parameters (e.g., sub-cooling, recess depth). We used COMSOL Multiphysics® to simulate melting, adopting an apparent heat capacity method. Results show that single-plug designs generally outperform multi-plug ones, where melting is inhibited by the formation of a frozen layer on top of the metal grate hosting the plugs. The layer thickness strongly depends on sub-cooling and recess depth. For multi-plug designs, the P/D ratio has a negligible influence on melting and can therefore be chosen to optimize the draining time. The absence of significant mixing in the pipe region above the plug leads to acceptable melting times (i.e., {\textless}1000 s) only for distances from the core up to 0.1 m, considered insufficient to host all the cooling equipment on the outside of the draining pipe and to protect the plug from possible large temperature oscillations in the core. Consequently, we conclude that the current freeze-plug design based only on decay heat to melt is likely to be unfeasible. A design improvement, preserving passivity and studied within the SAMOFAR project (http://samofar.eu/), consists in accelerating melting via heat stored in steel masses adjacent to the draining pipe.},
	language = {en},
	journal = {Annals of Nuclear Energy},
	author = {Tiberga, Marco and Shafer, Devaja and Lathouwers, Danny and Rohde, Martin and Kloosterman, Jan Leen},
	month = oct,
	year = {2019},
	keywords = {Molten Salt Fast Reactor, Apparent heat capacity method, Design improvement, Freeze-plug, Melting simulation, Passive safety device},
	pages = {544--554},
	file = {ScienceDirect Full Text PDF:C\:\\Users\\Sun Myung\\Zotero\\storage\\PIA6NZ9H\\Tiberga et al. - 2019 - Preliminary investigation on the melting behavior .pdf:application/pdf;ScienceDirect Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\WCGGPP7W\\S0306454919303573.html:text/html},
}

@article{giraud_development_2019,
	title = {Development of a cold plug valve with fluoride salt},
	volume = {5},
	copyright = {© J. Giraud et al., published by EDP Sciences, 2019},
	issn = {2491-9292},
	doi = {10.1051/epjn/2019005},
	abstract = {Experimental studies have been developed on a new freeze plug concept for safety valves in facilities using molten salt. They are designed to allow the closure of an upstream circuit by solidifying the molten salt in a section of the device and to passively melt in case of a loss of electric power, thus releasing the upper fluid. The working principle of these cold plug designs relies on the control of the heat transfer balance inside the device, which determines whether the salt inside the cold plug solidifies or melts. The device is mainly composed of steel masses that are dimensioned to provide sufficient thermal heat storage to melt the salt and thus open the cold plug after the electric power is stopped. The final goal of the work is to provide useful recommendations and guidelines for the design of a cold plug for the emergency draining system of a molten salt reactor. Some numerical thermal simulations were performed with ANSYS mechanical (Finite Element Method) to be compared with results of the experiments and to make extrapolations for a new component to be used in a reactor.},
	language = {en},
	journal = {EPJ Nuclear Sciences \& Technologies},
	author = {Giraud, Julien and Ghetta, Veronique and Rubiolo, Pablo and Retamales, Mauricio Tano},
	year = {2019},
	pages = {9},
	file = {Full Text PDF:C\:\\Users\\Sun Myung\\Zotero\\storage\\5RNC2PTC\\Giraud et al. - 2019 - Development of a cold plug valve with fluoride sal.pdf:application/pdf;Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\SKDAI6ZE\\epjn190007.html:text/html},
}

@article{di_ronco_eulerian_2021,
	title = {An {Eulerian} {Single}-{Phase} {Transport} {Model} for {Solid} {Fission} {Products} in the {Molten} {Salt} {Fast} {Reactor}: {Development} of an {Analytical} {Solution} for {Verification} {Purposes}},
	volume = {9},
	issn = {2296-598X},
	shorttitle = {An {Eulerian} {Single}-{Phase} {Transport} {Model} for {Solid} {Fission} {Products} in the {Molten} {Salt} {Fast} {Reactor}},
	abstract = {Nuclear reactor modeling has been shifting, over the last decades, towards full-core multiphysics analysis due to the ever-increasing safety requirements and complexity of the designs of innovative systems. This is particularly true for liquid-fuel reactor concepts such as the Molten Salt Fast Reactor (MSFR), given their strong intrinsic coupling between thermal-hydraulics, neutronics and fuel chemistry. In the MSFR, fission products (FPs) are originated within the liquid fuel and are carried by the fuel flow all over the reactor core and through pumping and heat exchange systems. Some of FP species, in the form of solid precipitates, can represent a major design and safety challenge, e.g., due to deposition on solid boundaries, and their distribution in the core is relevant to the design and safety analysis of the reactor. In this regard it is essential, both for the design and the safety assessment of the reactor, the capability to model the transport of solid FPs and their deposition to the boundary (e.g., wall or heat exchanger structures). To this aim, in this study, models of transport of solid FPs in the MSFR are developed and verified. An Eulerian single-phase transport model is developed and integrated in a consolidated multiphysics model of the MSFR based on the open-source CFD library OpenFOAM. In particular, general mixed-type deposition boundary conditions are considered, to possibly describe different kinds of particle-wall interaction mechanisms. For verification purposes, analytical solutions for simple case studies are derived ad hoc based on the extension of the classic Graetz problem to linear decay, distributed source terms and mixed-type boundary conditions. The results show excellent agreement between the two models, and highlight the effects of decay and deposition phenomena of various intensity. The resulting approach constitutes a computationally efficient tool to extend the capabilities of CFD-based multiphysics MSFR calculations towards the simulation of solid fission products transport.},
	journal = {Frontiers in Energy Research},
	author = {Di Ronco, Andrea and Lorenzi, Stefano and Giacobbo, Francesca and Cammi, Antonio},
	year = {2021},
	file = {Full Text PDF:C\:\\Users\\Sun Myung\\Zotero\\storage\\2U4ISW9L\\Di Ronco et al. - 2021 - An Eulerian Single-Phase Transport Model for Solid.pdf:application/pdf},
}

@article{di_ronco_multiphysics_2022,
	title = {Multiphysics analysis of {RANS}-based turbulent transport of solid fission products in the {Molten} {Salt} {Fast} {Reactor}},
	volume = {391},
	issn = {0029-5493},
	doi = {10.1016/j.nucengdes.2022.111739},
	abstract = {The analysis of innovative reactor concepts such as the Molten Salt Fast Reactor (MSFR) requires the development of new modeling and simulation tools. In the case of the MSFR, the strong intrinsic coupling between thermal–hydraulics, neutronics and fuel chemistry has lead to the adoption of the multiphysics approach as a state-of-the-art paradigm. One of the peculiar aspects of liquid-fuel reactors such as the MSFR is the mobility of fission products (FPs) in the reactor circuit. Some FP species appear in form of solid precipitates carried by the fuel flow and can deposit on reactor boundaries (e.g., heat exchangers), potentially representing design issues related to the degradation of heat exchange performance or radioactive hotspots. Other precipitates might be present in the primary system as well, e.g. due to oxidation of fissile species which might lead to local criticality issues. The integration of transport models for solid particles in multiphysics codes is therefore relevant for the prediction of deposited fractions. To this aim, a previously developed Eulerian single-phase transport model is employed to analyze the distribution of solid FPs in two simplified MSFR-related geometries. The effect of physical parameters and of distributed particle sources on the numerical requirements needed to resolve particle concentration gradients at reactor boundaries in the considered geometries is investigated with the aid of analytical results. Analytical estimates of concentration gradients, even though not in full agreement with simulations, prove useful to drive the choice of adequate mesh refinements. Furthermore, the influence of different RANS turbulence modeling approaches on the prediction of particle distributions and deposition is tested. Results show a limited influence of the choice of turbulence models and parameters on the deposited fraction and on concentration gradients. As a result, it is found that deposition rates are scarcely affected by the choice of turbulent Schmidt number, with lower diffusivities being compensated by larger gradients. Large deposited fractions are indeed predicted, suggesting the efficiency of FPs transport mechanisms in the reactor and the need for the integration of adequate FPs transport models within state-of-the-art MSFR codes.},
	language = {en},
	journal = {Nuclear Engineering and Design},
	author = {Di Ronco, Andrea and Lorenzi, Stefano and Giacobbo, Francesca and Cammi, Antonio},
	month = may,
	year = {2022},
	keywords = {Multiphysics, MSFR, Molten salt, Fission products, OpenFOAM, Turbulence modeling},
	pages = {111739},
	file = {ScienceDirect Full Text PDF:C\:\\Users\\Sun Myung\\Zotero\\storage\\NDHCDHXG\\Di Ronco et al. - 2022 - Multiphysics analysis of RANS-based turbulent tran.pdf:application/pdf;ScienceDirect Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\8MDMZCRH\\S0029549322000930.html:text/html},
}

@article{kalilainen_evaporation_2020,
	title = {Evaporation of materials from the molten salt reactor fuel under elevated temperatures},
	volume = {533},
	issn = {0022-3115},
	doi = {10.1016/j.jnucmat.2020.152134},
	abstract = {The release of fission product and salt compounds from a molten salt reactor fuel under accident conditions was investigated with coupled computer simulations. The thermodynamic modeling of the salt and fission product mixture was performed in The Gibbs Energy Minimization Software GEMS and the obtained compound vapor pressures were exchanged with the severe accident code MELCOR, where the evaporation from a salt surface located at the bottom of a confinement building was simulated. The fuel salt considered in the simulations was LiF-ThF4-UF4 with fission products Cs and I. The composition of the fuel salt material was obtained from an equilibrium fuel cycle simulation of the salt using the EQL0D routine coupled to the Serpent 2 code. The results were compared to simulations using pure compound vapor pressures in the evaporation simulations. It was observed that by modeling the salt mixing the release of fission products and salt materials was reduced when compared to the pure compound simulations. The mixing effects in the salt, when compared to the pure compound simulation also affected evaporation temperatures and therefore the timing of the release of compounds. In an additional simulation in which the depressurization of the confinement was considered, the total evaporated mass of compounds increased due to increased mass transfer at the salt surface. The simulation process described in this paper can be used for a more comprehensive accident analysis of molten salt reactors once the detailed description of the reactor confinement and accident sequences are available and more fission product elements have been added to the analysis.},
	language = {en},
	journal = {Journal of Nuclear Materials},
	author = {Kalilainen, Jarmo and Nichenko, Sergii and Krepel, Jiri},
	month = may,
	year = {2020},
	keywords = {Molten salt reactor, Severe accident, Fission product release},
	pages = {152134},
	file = {ScienceDirect Full Text PDF:C\:\\Users\\Sun Myung\\Zotero\\storage\\3KE6RIAF\\Kalilainen et al. - 2020 - Evaporation of materials from the molten salt reac.pdf:application/pdf;ScienceDirect Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\ASIT9VV9\\S0022311519314631.html:text/html},
}

@article{jaradat_development_2021,
	title = {Development and validation of {PROTEUS}-{NODAL} transient analyses capabilities for molten salt reactors},
	volume = {160},
	issn = {0306-4549},
	doi = {10.1016/j.anucene.2021.108402},
	abstract = {This paper presents the PROTEUS-NODAL capabilities developed for modeling and simulation of molten salt reactors (MSRs) with flowing fuel. Steady-state and transient solvers were developed to solve the coupled neutron flux and delayed neutron precursor concentration equations for a given velocity field. A variational nodal solver for cylindrical geometries was also developed. For thermal feedback calculations, a standalone thermal-hydraulics solver for flowing fuel was implemented. Verification and validation tests were performed using the Molten Salt Fast Reactor (MSFR) benchmark and the available experimental data of the Molten Salt Reactor Experiments (MSRE). The MSFR solution of PROTEUS-NODAL agreed well with the open literature solutions for all transients. The PROTEUS-NODAL results for static and transient experiments of the MSRE were in good agreement with measured values. These results indicate that the PROTEUS-NODAL code can be used reliably for steady-state and transient analyses of fast and thermal spectrum MSRs.},
	language = {en},
	journal = {Annals of Nuclear Energy},
	author = {Jaradat, Mustafa K. and Park, Hansol and Yang, Won Sik and Lee, Changho},
	month = sep,
	year = {2021},
	keywords = {Molten salt reactor, Coupled neutronics and thermal-hydraulics, Delayed neutron precursor, Flowing fuel, Transient fixed source problem},
	pages = {108402},
	file = {ScienceDirect Full Text PDF:C\:\\Users\\Sun Myung\\Zotero\\storage\\JIZSZQXH\\Jaradat et al. - 2021 - Development and validation of PROTEUS-NODAL transi.pdf:application/pdf;ScienceDirect Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\RHIBZM8H\\S0306454921002784.html:text/html},
}

@inproceedings{sargent_verification_2010,
	title = {Verification and validation of simulation models},
	doi = {10.1109/WSC.2010.5679166},
	abstract = {In this paper we discuss verification and validation of simulation models. Four different approaches to deciding model validity are described; two different paradigms that relate verification and validation to the model development process are presented; various validation techniques are defined; conceptual model validity, model verification, operational validity, and data validity are discussed; a way to document results is given; a recommended procedure for model validation is presented; and model accreditation is briefly discussed.},
	booktitle = {Proceedings of the 2010 {Winter} {Simulation} {Conference}},
	author = {Sargent, Robert G.},
	month = dec,
	year = {2010},
	keywords = {Programming, Mathematical model, Analytical models, Accuracy, Data models, Computational modeling, Computers},
	pages = {166--183},
	file = {IEEE Xplore Abstract Record:C\:\\Users\\Sun Myung\\Zotero\\storage\\6HTQB464\\5679166.html:text/html;IEEE Xplore Full Text PDF:C\:\\Users\\Sun Myung\\Zotero\\storage\\5SF95AWR\\Sargent - 2010 - Verification and validation of simulation models.pdf:application/pdf},
}

@inproceedings{delpech_benchmark_2003,
	title = {Benchmark of dynamic simulation tools for molten salt reactors},
	language = {en},
	publisher = {American Nuclear Society},
	author = {Delpech, M. and Dulla, S. and Garzenne, C. and Kophazi, J. and Krepel, J. and Brun, C. Le and Lecarpentier, D. and Mattioda, F. and Ravetto, P. and Rineiski, A. and Schikorr, M. and Szieberth, M.},
	month = nov,
	year = {2003},
	pages = {2182},
	file = {Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\KH8AGWKB\\in2p3-00020303.html:text/html},
}

@article{laureau_unmoderated_2022,
	title = {Unmoderated molten salt reactors design optimisation for power stability},
	volume = {177},
	issn = {0306-4549},
	doi = {10.1016/j.anucene.2022.109265},
	abstract = {The critical region of unmoderated molten salt reactors consists in a cavity filled with a liquid fuel. The lack of internal structure implies a complex flow structure of the circulating fuel salt. A preliminary core shape optimization has been performed during the EVOL European project to limit recirculation and hotspots. This optimization was based on a Reynolds Averaged Navier Stokes (RANS) approach, but the latter only provides time-averaged values for velocity and temperature. However, the power stability is sensitive to thermal fluctuations induced by the flow turbulence itself, even at steady state without pump flow rate or heat extraction variation. This phenomenon is studied using a Detached Eddies Simulation approach to solve the turbulence in the reactor and get a time dependent temperature distribution and then the reactivity fluctuations. A new geometry is proposed to limit the total power fluctuations from 7.5\% for the preconceptual EVOL geometry down to 1.2\%.},
	language = {en},
	journal = {Annals of Nuclear Energy},
	author = {Laureau, A. and Bellè, A. and Allibert, M. and Heuer, D. and Merle, E. and Pautz, A.},
	month = nov,
	year = {2022},
	keywords = {Turbulence, Molten salt reactor, Core design, DES, Reactivity fluctuation},
	pages = {109265},
	file = {ScienceDirect Full Text PDF:C\:\\Users\\Sun Myung\\Zotero\\storage\\P2NYYI35\\Laureau et al. - 2022 - Unmoderated molten salt reactors design optimisati.pdf:application/pdf;ScienceDirect Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\BV9RMG4G\\S0306454922003000.html:text/html},
}

@article{cui_development_2021,
	title = {Development of a steady state analysis code for molten salt reactor based on nodal expansion method},
	volume = {151},
	issn = {0306-4549},
	doi = {10.1016/j.anucene.2020.107950},
	abstract = {The unique feature of molten salt reactor (MSR) is that heavy metal fuel and fission products can be dissolved in a molten fluoride salt to form a eutectic mixture which acts as both fuel and coolant. The fission energy from fissile fuel is released immediately into the fuel salt in the core, and the delayed neutron precursors (DNP) have a loss due to fuel circulation in the primary loop, which both make MSR be characterized by the strong interplay between the neutronics and thermal hydraulics (TH) and imply that the software developed for traditional solid-fueled reactors is not applicable to MSR. In this study, a new coupled neutronics and TH code named TMSR3D is developed for evaluating the steady-state characteristics of liquid-fueled MSR. In this code, the fourth order standard nodal expansion method (NEM) is implemented to solve the neutron diffusion equations, and the parallel multi-channel model is adopted to take into account the feedbacks of local temperature to group constants. Besides, a convection term is introduced into the traditional balance equations for DNP to consider the influence of fuel salt flow. Some experiment data of Molten Salt Reactor Experiment (MSRE) are chosen to verify the code, and the core characteristics including the distributions of neutron fluxes, DNP and temperature under the rated condition are simulated and discussed. Furthermore, the impacts of the inlet mass flow rate, inlet fuel temperature and the fuel salt residence time out of the core are analyzed in detail. The numerical results indicate that the TMSR3D code can provide a reliable description of the neutronics-TH coupling phenomena and of the steady state analysis for graphite-moderated channel-type MSRs.},
	language = {en},
	journal = {Annals of Nuclear Energy},
	author = {Cui, Y. and Chen, J. G. and Dai, M. and Cai, X. Z.},
	month = feb,
	year = {2021},
	keywords = {Molten salt reactor, Nodal expansion method, Parallel multi-channel model, TMSR3D},
	pages = {107950},
	file = {ScienceDirect Full Text PDF:C\:\\Users\\Sun Myung\\Zotero\\storage\\9NV85NCT\\Cui et al. - 2021 - Development of a steady state analysis code for mo.pdf:application/pdf;ScienceDirect Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\66AT6K6L\\S0306454920306460.html:text/html},
}

@article{pomraning_flux-limited_1984,
	title = {Flux-{Limited} {Diffusion} {Theory} with {Anisotropic} {Scattering}},
	volume = {86},
	issn = {0029-5639},
	doi = {10.13182/NSE84-A18634},
	abstract = {The Chapman-Enskog procedure of kinetic theory is used to obtain a flux-limited diffusion theory for linear transport problems. The analysis is carried out within the context of multigroup theory, with an arbitrary scattering matrix involving anisotropic effects for both within-group and group-to group scattering. Aside from the question of flux limiting, a new definition of the transport cross section arises naturally from the analysis.},
	number = {4},
	journal = {Nuclear Science and Engineering},
	author = {Pomraning, G. C},
	month = apr,
	year = {1984},
	pages = {335--343},
	file = {Full Text PDF:C\:\\Users\\Sun Myung\\Zotero\\storage\\XTHUN36R\\Pomraning - 1984 - Flux-Limited Diffusion Theory with Anisotropic Sca.pdf:application/pdf},
}

@misc{noauthor_joblib_nodate,
	title = {Joblib: running {Python} functions as pipeline jobs — joblib 1.3.0.dev0 documentation},
	url = {https://joblib.readthedocs.io/en/latest/},
	urldate = {2023-02-09},
	file = {Joblib\: running Python functions as pipeline jobs — joblib 1.3.0.dev0 documentation:C\:\\Users\\Sun Myung\\Zotero\\storage\\SLAB328E\\latest.html:text/html},
}

@article{reynolds_analysis_2023,
	title = {An {Analysis} of {Transport} {Effects} in {Steady}-{State} {Simulation} of the {Molten} {Salt} {Reactor} {Experiment}},
	volume = {197},
	issn = {0029-5639},
	doi = {10.1080/00295639.2022.2097565},
	abstract = {We use the deterministic neutron transport code QuasiMolto to simulate steady-state operation of the Molten Salt Reactor Experiment (MSRE). Comparisons are made to similar results from the MOST benchmark, the MOOSE-based code Moltres, and the design calculations for the MSRE. In the course of these comparisons, we calculate a value of 0.1799 for the graphite-to-fuel power density ratio, which differs significantly from that seen in other works. We also find uniform graphite heating inadequate to reproduce the characteristic graphite temperature distribution of the MSRE. Leveraging the multilevel projective methodology of QuasiMolto, the influence of transport effects on the modeled problem is found to produce average and maximum group flux variations of 2\% to 5\% and 30\%, respectively, with a 12\% variation in the reactivity loss due to delayed neutron precursor drift.},
	number = {1},
	journal = {Nuclear Science and Engineering},
	author = {Reynolds, Aaron J. and Palmer, Todd S.},
	month = jan,
	year = {2023},
	keywords = {Molten Salt Reactor Experiment, circulating fuel reactors, graphite-to-fuel power density ratio, multilevel method, transport effects},
	pages = {45--73},
	file = {Full Text PDF:C\:\\Users\\Sun Myung\\Zotero\\storage\\AKE4XB3E\\Reynolds and Palmer - 2023 - An Analysis of Transport Effects in Steady-State S.pdf:application/pdf},
}

@misc{ornl_first-ever_2023,
	title = {First-ever {MSR} benchmark published, based on {ORNL}’s past and present research},
	url = {https://www.ornl.gov/molten-salt-reactor/measurements},
	urldate = {2023-03-20},
	author = {ORNL},
	month = mar,
	year = {2023},
	file = {Measurements | Molten Salt Reactor | ORNL:C\:\\Users\\Sun Myung\\Zotero\\storage\\Z7ZFEUQQ\\measurements.html:text/html},
}

@article{lindsay_20_2022,
	title = {2.0 - {MOOSE}: {Enabling} massively parallel multiphysics simulation},
	volume = {20},
	issn = {2352-7110},
	shorttitle = {2.0 - {MOOSE}},
	doi = {10.1016/j.softx.2022.101202},
	abstract = {The last 2 years have been a period of unprecedented growth for the MOOSE community and the software itself. The number of monthly visitors to the website has grown from just over 3,000 to now averaging 5,000. In addition, over 1,800 pull requests have been merged since the beginning of 2020, and the new discussions forum has averaged 600 unique visitors per month. The previous publication has been cited over 200 times since it was published 2 years ago. This paper serves as an update on some of the key additions and changes to the code and ecosystem over the last 2 years, as well as recognizing contributions from the community.},
	language = {en},
	journal = {SoftwareX},
	author = {Lindsay, Alexander D. and Gaston, Derek R. and Permann, Cody J. and Miller, Jason M. and Andrš, David and Slaughter, Andrew E. and Kong, Fande and Hansel, Joshua and Carlsen, Robert W. and Icenhour, Casey and Harbour, Logan and Giudicelli, Guillaume L. and Stogner, Roy H. and German, Peter and Badger, Jacob and Biswas, Sudipta and Chapuis, Leora and Green, Christopher and Hales, Jason and Hu, Tianchen and Jiang, Wen and Jung, Yeon Sang and Matthews, Christopher and Miao, Yinbin and Novak, April and Peterson, John W. and Prince, Zachary M. and Rovinelli, Andrea and Schunert, Sebastian and Schwen, Daniel and Spencer, Benjamin W. and Veeraraghavan, Swetha and Recuero, Antonio and Yushu, Dewen and Wang, Yaqi and Wilkins, Andy and Wong, Christopher},
	month = dec,
	year = {2022},
	keywords = {Multiphysics, Finite-element, Framework, Object-oriented},
	pages = {101202},
	file = {ScienceDirect Full Text PDF:C\:\\Users\\Sun Myung\\Zotero\\storage\\A6UYVLR6\\Lindsay et al. - 2022 - 2.0 - MOOSE Enabling massively parallel multiphys.pdf:application/pdf},
}

@article{liu_sensitivityuncertainty_2020,
	title = {Sensitivity/uncertainty comparison and similarity analysis between {TMSR}-{LF1} and {MSR} models},
	volume = {122},
	issn = {0149-1970},
	doi = {10.1016/j.pnucene.2020.103289},
	abstract = {TMSR-LF1 (Thorium-based Molten Salt Reactor-Liquid Fuel) is the reactor chosen to be developed in the “Thorium molten salt reactor nuclear energy system” project of the Strategic Priority Program launched by the Chinese Academy of Sciences. The software and data library adopted in its design needs to be verified and validated by using benchmarks. In order to guarantee the validity of the verification results, the being-designed reactor should have similarity (index Ck and E ≥ 0.8) with the benchmark reactor. Using TSUNAMI module and in SCALE 6.1, the Sensitivity/Uncertainty (S/U) analyses and similarity assessments were performed in TMSR-LF1 and the candidate benchmark reactors: MSRE (Molten Salt Reactor Experiment), MSBR (Molten Salt Breeder Reactor) and the Critical Experiment Device (hereafter the device). Uncertainty was calculated using both SCALE and TMSR 44-group covariance data. It is found that the total eigenvalue uncertainties deduced from covariance data in TMSR-LF1 and the candidate benchmark reactors satisfy the ±1\% design accuracy requirement proposed by M. Ishikawa in 2008, except in MSBR. To meet the accuracy of ±0.3\%, targeted experiments for TMSR are needed. The covariance data of 7Li, C-graphite, 58Ni and 19F were the main contributors to the uncertainty of effective multiplication factor as well as the covariance data of fuel nuclides. The calculated similarity indices indicated that there was a good similarity (Ck and E {\textgreater} 0.9) between TMSR-LF1 and MSRE, and a lesser similarity between TMSR-LF1 and the device (Ck and E {\textgreater} 0.8). The uncertainties should be improved and more benchmarks for the direct verification and validation should be provided by relevant experiments.},
	language = {en},
	journal = {Progress in Nuclear Energy},
	author = {Liu, Yafen and Yan, Rui and Zou, Yang and Yu, Shihe and Zhou, Bo and Kang, Xuzhong and Hu, Jifeng and Cai, Xiangzhou},
	month = apr,
	year = {2020},
	keywords = {MSR, Covariance data, SCALE 6.1/TSUNAMI, Sensitivity/uncertainty, Similarity, TMSR-LF1},
	pages = {103289},
	file = {ScienceDirect Full Text PDF:C\:\\Users\\Sun Myung\\Zotero\\storage\\WWXG7TRB\\Liu et al. - 2020 - Sensitivityuncertainty comparison and similarity .pdf:application/pdf;ScienceDirect Snapshot:C\:\\Users\\Sun Myung\\Zotero\\storage\\6CUHAVYI\\S0149197020300482.html:text/html},
}

@article{virtanen_scipy_2020,
	title = {{SciPy} 1.0: fundamental algorithms for scientific computing in {Python}},
	volume = {17},
	copyright = {2020 The Author(s)},
	issn = {1548-7105},
	shorttitle = {{SciPy} 1.0},
	doi = {10.1038/s41592-019-0686-2},
	abstract = {SciPy is an open-source scientific computing library for the Python programming language. Since its initial release in 2001, SciPy has become a de facto standard for leveraging scientific algorithms in Python, with over 600 unique code contributors, thousands of dependent packages, over 100,000 dependent repositories and millions of downloads per year. In this work, we provide an overview of the capabilities and development practices of SciPy 1.0 and highlight some recent technical developments.},
	language = {en},
	number = {3},
	journal = {Nature Methods},
	author = {Virtanen, Pauli and Gommers, Ralf and Oliphant, Travis E. and Haberland, Matt and Reddy, Tyler and Cournapeau, David and Burovski, Evgeni and Peterson, Pearu and Weckesser, Warren and Bright, Jonathan and van der Walt, Stéfan J. and Brett, Matthew and Wilson, Joshua and Millman, K. Jarrod and Mayorov, Nikolay and Nelson, Andrew R. J. and Jones, Eric and Kern, Robert and Larson, Eric and Carey, C. J. and Polat, Ilhan and Feng, Yu and Moore, Eric W. and VanderPlas, Jake and Laxalde, Denis and Perktold, Josef and Cimrman, Robert and Henriksen, Ian and Quintero, E. A. and Harris, Charles R. and Archibald, Anne M. and Ribeiro, Antônio H. and Pedregosa, Fabian and van Mulbregt, Paul},
	month = mar,
	year = {2020},
	keywords = {Biophysical chemistry, Computational biology and bioinformatics, Technology},
	pages = {261--272},
	file = {Full Text PDF:C\:\\Users\\Sun Myung\\Zotero\\storage\\GK9II5WK\\Virtanen et al. - 2020 - SciPy 1.0 fundamental algorithms for scientific c.pdf:application/pdf},
}