Moltres.bib

@article{turner_virtual_2016,
	title = {The Virtual Environment for Reactor Applications ({VERA}):
                  Design and architecture},
	volume = {326},
	issn = {0021-9991},
	url = {http://www.sciencedirect.com/science/article/pii/S0021999116304156},
	doi = {10.1016/j.jcp.2016.09.003},
	shorttitle = {The Virtual Environment for Reactor Applications ({VERA})},
	abstract = {{VERA}, the Virtual Environment for Reactor Applications, is the system of physics capabilities being developed and deployed by the Consortium for Advanced Simulation of Light Water Reactors ({CASL}). {CASL} was established for the modeling and simulation of commercial nuclear reactors. {VERA} consists of integrating and interfacing software together with a suite of physics components adapted and/or refactored to simulate relevant physical phenomena in a coupled manner. {VERA} also includes the software development environment and computational infrastructure needed for these components to be effectively used. We describe the architecture of {VERA} from both software and numerical perspectives, along with the goals and constraints that drove major design decisions, and their implications. We explain why {VERA} is an environment rather than a framework or toolkit, why these distinctions are relevant (particularly for coupled physics applications), and provide an overview of results that demonstrate the use of {VERA} tools for a variety of challenging applications within the nuclear industry.},
	pages = {544--568},
	issue = {Supplement C},
	journaltitle = {Journal of Computational Physics},
	shortjournal = {Journal of Computational Physics},
	author = {Turner, John A. and Clarno, Kevin and Sieger, Matt and Bartlett, Roscoe and Collins, Benjamin and Pawlowski, Roger and Schmidt, Rodney and Summers, Randall},
	urldate = {2017-11-27},
	date = {2016-12-01},
	keywords = {Multiphysics, Simulation, Neutronics, Thermal-hydraulics, Core simulator, Coupled physics, Fluid flow, Modeling, Nuclear reactor},
	file = {ScienceDirect Full Text PDF:/home/gavin/Zotero/storage/MNPKBXBT/Turner et al. - 2016 - The Virtual Environment for Reactor Applications (.pdf:application/pdf;ScienceDirect Snapshot:/home/gavin/Zotero/storage/4BY2ABIR/S0021999116304156.html:text/html}
}

@report{engel_conceptual_1980,
	location = {Oak Ridge, {TN}, United States},
	title = {Conceptual design characteristics of a denatured molten-salt reactor with once-through fueling},
	url = {http://www.osti.gov/scitech/biblio/5352526},
	number = {{ORNL}/{TM}-7207, 5352526},
	institution = {Oak Ridge National Laboratory},
	type = {Department of Energy},
	author = {Engel, J.R. and Bauman, H.F. and Dearing, J.F. and Grimes, W.R. and {McCoy}, H.E. and Rhoades, W.A.},
	urldate = {2013-09-06},
	date = {1980-07-01},
	keywords = {read},
	file = {engel_conceptual_1980.pdf:/home/gavin/Zotero/storage/XQYT9UA4/engel_conceptual_1980.pdf:application/pdf}
}

@article{zhang_development_2009,
	title = {Development of a steady state analysis code for a molten salt reactor},
	volume = {36},
	issn = {0306-4549},
	url = {http://www.sciencedirect.com/science/article/pii/S030645490900019X},
	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.},
	pages = {590--603},
	number = {5},
	journaltitle = {Annals of Nuclear Energy},
	shortjournal = {Annals of Nuclear Energy},
	author = {Zhang, D. L. and Qiu, S. Z. and Su, G. H. and Liu, C. L.},
	urldate = {2017-11-27},
	date = {2009-05-15},
	file = {ScienceDirect Snapshot:/home/gavin/Zotero/storage/76AZTE7I/S030645490900019X.html:text/html}
}

@article{serrano-lopez_molten_2013,
	title = {Molten salts database for energy applications},
	volume = {73},
	issn = {0255-2701},
	url = {http://www.sciencedirect.com/science/article/pii/S0255270113001827},
	doi = {10.1016/j.cep.2013.07.008},
	abstract = {The growing interest in energy applications of molten salts is justified by several of their properties. Their possibilities of usage as a coolant, heat transfer fluid or heat storage substrate, require thermo-hydrodynamic refined calculations. Many researchers are using simulation techniques, such as Computational Fluid Dynamics (CFD) for their projects or conceptual designs. The aim of this work is providing a review of basic properties (density, viscosity, thermal conductivity and heat capacity) of the most common and referred salt mixtures. After checking data, tabulated and graphical outputs are given in order to offer the most suitable available values to be used as input parameters for other calculations or simulations. The reviewed values show a general scattering in characterization, mainly in thermal properties. This disagreement suggests that, in several cases, new studies must be started (and even new measurement techniques should be developed) to obtain accurate values.},
	urldate = {2014-09-23},
	journal = {Chemical Engineering and Processing: Process Intensification},
	author = {Serrano-L{\'o}pez, R. and Fradera, J. and Cuesta-L{\'o}pez, S.},
	month = nov,
	year = {2013},
	keywords = {Energy, CFD, Coolants, CSP, Properties, molten salt},
	pages = {87--102},
	file = {ScienceDirect Snapshot:/home/huff/Zotero/storage/8QI8WWKV/login.html:text/html;Untitled Attachment:/home/huff/Zotero/storage/699Z9JF2/Serrano-L{\'o}pez et al. - 2013 - Molten salts database for energy applications.html:text/html}
}

@techreport{robertson_conceptual_1971,
	title = {Conceptual {Design} {Study} of a {Single}-{Fluid} {Molten}-{Salt} {Breeder} {Reactor}.},
	url = {http://www.osti.gov/scitech/biblio/4030941},
	language = {English},
	number = {ORNL--4541},
	urldate = {2016-09-06},
	institution = {comp.; Oak Ridge National Lab., Tenn.},
	author = {Robertson, R. C.},
	month = jan,
	year = {1971},
	keywords = {Economics, design, boilers, buildings, configuration, containment, control systems, coolant loops, fused salt fuel, maintenance, n38100*  --power reactor development, operation, reactor sites, waste disposal molten salt breeder reactor/design parameters for conceptual 1000 mw(e) single fluid, MSBR},
	file = {3445600591394.pdf:/home/huff/Zotero/storage/RC7URKHP/3445600591394.pdf:application/pdf;Full Text PDF:/home/huff/Zotero/storage/DFSGI5SW/Robertson - 1971 - Conceptual Design Study of a Single-Fluid Molten-S.pdf:application/pdf;Snapshot:/home/huff/Zotero/storage/G8IEVM75/4030941.html:text/html}
}

@article{cammi_multi-physics_2011,
	title = {A multi-physics modelling approach to the dynamics of {Molten} {Salt} {Reactors}},
	volume = {38},
	issn = {0306-4549},
	url = {http://www.sciencedirect.com/science/article/pii/S0306454911000582},
	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{\textendash}Stokes equations are used with the turbulence treatment according to the RANS (Reynolds Averaged Navier{\textendash}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{\textregistered}, 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},
	urldate = {2013-05-28},
	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 = {unread, read, MSR, Molten salt reactor, atws, Multi-physics modelling, Reactor dynamics, Thermo-hydrodynamics},
	pages = {1356--1372},
	file = {A_multi-physics_modelling_approach_to_the_dynamics_of_Molten_Sal.mobi:/home/huff/Zotero/storage/JWIMI3QI/A_multi-physics_modelling_approach_to_the_dynamics_of_Molten_Sal.mobi:application/octet-stream;cammi_multi-physics_2011.pdf:/home/huff/Zotero/storage/AHXUPQ4A/cammi_multi-physics_2011.pdf:application/pdf;ScienceDirect Full Text PDF:/home/huff/Zotero/storage/6FQKN2CJ/Cammi et al. - 2011 - A multi-physics modelling approach to the dynamics.pdf:application/pdf;ScienceDirect Snapshot:/home/huff/Zotero/storage/63AWQD57/S0306454911000582.html:text/html}
}

@misc{wikipedia_molten_2016,
	title = {Molten salt reactor},
	copyright = {Creative Commons Attribution-ShareAlike License},
	url = {https://en.wikipedia.org/w/index.php?title=Molten_salt_reactor&oldid=742696102},
	abstract = {A molten salt reactor (MSR) is a class of generation IV nuclear fission reactor in which the primary nuclear reactor coolant, or even the fuel itself, is a molten salt mixture. MSRs can run at higher temperatures than water-cooled reactors for a higher thermodynamic efficiency, while staying at low vapour pressure.
The nuclear fuel may be solid or dissolved in the coolant. In many designs the nuclear fuel dissolved in the coolant is uranium tetrafluoride (UF4). The fluid becomes critical in a graphite core that serves as the moderator. Some solid-fuel designs propose ceramic fuel dispersed in a graphite matrix, with the molten salt providing low pressure, high temperature cooling. The salts are much more efficient than compressed helium (another potential coolant in Generation IV reactor designs) at removing heat from the core, reducing the need for pumping and piping and reducing the core size.
The concept was established in the 1950s. The early Aircraft Reactor Experiment (1954) was primarily motivated by the small size that the design could provide, while the Molten-Salt Reactor Experiment (1965{\textendash}1969) was a prototype for a thorium fuel cycle breeder reactor nuclear power plant. The increased research into Generation IV reactor designs included a renewed interest in the technology.},
	language = {en},
	urldate = {2016-10-06},
	journal = {Wikipedia, the free encyclopedia},
	author = {{Wikipedia}},
	month = oct,
	year = {2016},
	note = {Page Version ID: 742696102},
	file = {Snapshot:/home/huff/Zotero/storage/SBXJB67T/index.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{\textendash}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 = {K{\'o}ph{\'a}zi, J. and Lathouwers, D. and Kloosterman, J.L.},
	year = {2009},
	keywords = {3D, unread, Molten Salt Reactor (MSR), Core, MSR Experiment (MSRE), Reactor Physics},
	pages = {118--131},
	file = {K{\'o}ph{\'a}zi et al. - Development of a Three-Dimensional Time-Dependent .pdf:/home/huff/Zotero/storage/ZIJ5Q643/K{\'o}ph{\'a}zi et al. - Development of a Three-Dimensional Time-Dependent .pdf:application/pdf}
}

@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},
	urldate = {2016-09-20},
	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 = {Fuels, graphite, mass, monitoring, planning, buildings, configuration, control systems, coolant loops, operation, personnel, absorption, accidents, adsorption, alpha particles, beryllium, concretes, control elements, degassing, delayed neutrons, density, distribution, enrichment, equations, excursions, failures, fertile materials, fission products, fissionable materials, fused salts, gases, graphite moderator, heat transfer, high temperature, liquid flow, lithium, losses, low temperature, materials testing, moderators, msre, multiplication factors, neutron flux, neutron sources, neutrons, nuclear reactions, poisoning, power plant, reactor technology, criticality},
	file = {Full Text PDF:/home/huff/Zotero/storage/DMIDAWSC/Haubenreich et al. - 1964 - Msre Design and Operations Report. Part Iii. Nucle.pdf:application/pdf;Snapshot:/home/huff/Zotero/storage/64Z7NT6C/4114686.html:text/html}
}

@techreport{robertson_msre_1965,
	title = {Msre {Design} and {Operations} {Report}. {Part} {I}. {Description} of {Reactor} {Design}},
	url = {http://www.osti.gov/scitech/biblio/4654707},
	language = {English},
	number = {ORNL-TM-728},
	urldate = {2016-09-07},
	institution = {Oak Ridge National Lab., Tenn.},
	author = {Robertson, R. C.},
	month = jan,
	year = {1965},
	keywords = {reactors, waste disposal, planning, coolant loops, fused salt fuel, gases, msre, reactor technology, air, decontamination, detection, leaks, liquids, pressure vessels, research and test reactors, research reactors, sampling, shielding, water coolant},
	file = {Full Text PDF:/home/huff/Zotero/storage/VX3DIUNH/Robertson - 1965 - Msre Design and Operations Report. Part I. Descrip.pdf:application/pdf;Snapshot:/home/huff/Zotero/storage/D98K5BFA/4654707.html:text/html}
}

@article{park_whole_2015,
	title = {Whole core analysis of molten salt breeder reactor with online fuel reprocessing: {Whole} core analysis of {MSBR} with online fuel reprocessing},
	issn = {0363907X},
	shorttitle = {Whole core analysis of molten salt breeder reactor with online fuel reprocessing},
	url = {http://doi.wiley.com/10.1002/er.3371},
	doi = {10.1002/er.3371},
	language = {en},
	urldate = {2016-11-10},
	journal = {International Journal of Energy Research},
	author = {Park, Jinsu and Jeong, Yongjin and Lee, Hyun Chul and Lee, Deokjung},
	month = jul,
	year = {2015},
	pages = {n/a--n/a},
	file = {Whole core analysis of molten salt breeder reactor with online fuel reprocessing.pdf:/home/huff/Zotero/storage/TPRZSD4Q/Whole core analysis of molten salt breeder reactor with online fuel reprocessing.pdf:application/pdf}
}

@misc{nestor_reactor_1960,
	type = {Report},
	title = {{REACTOR} {PHYSICS} {CALCULATIONS} {FOR} {THE} {MSRE}},
	url = {https://digital.library.unt.edu/ark:/67531/metadc868473/},
	abstract = {A compilation is presented of results obtained to date from a number of reactor physics calculations for the molten salt reactor experiment (MSRE). Included are one-dimensional multigroup and two-dimensional twogroup calculations of critical mass, flux, and power density distributions; gamma heating in the core can, reactor vessel, and core support grid; drain tank criticality; and an estimate of the beta , gamma , and delayed neutron dose rates due to fission products in the fuel contained in the pump bowl. For a cylindrical core 54 in. in diameter and 66 in. high, graphite-mcderated with 8 vol\% fuel salt, the calculated critical loading is 0.76 mole\% uranium (93.3\% U/sup 235/), which is equivalent to a critical mass of 16 kg. At a reactor power of 10 mw, the peak power density in the core assuming a homogeneous mixure of fuel salt and graphite is 10 watts/cm/sup 3/, the average power density is 4 watts/cm/sup 3/. The computed peak thermal flux is 7.3 x 10/sup 13/ neutrons/cm/sup 2/ sec and the average is 2.5 x l0/sup 13/ neutrons/cm/sup 2/ sec. Gamma heating prcduces a power density of 0.2 watts/cm/sup 3/ in the core wall at the midplane and 0.4 watts/cm/sup 3/ in the support grid at the bottom of the core at the reactor center line. (auth)},
	language = {English},
	urldate = {2017-05-18},
	journal = {Other Information: Orig. Receipt Date: 31-DEC-60},
	author = {Nestor, Jr},
	month = jul,
	year = {1960},
	file = {Full Text PDF:/home/huff/Zotero/storage/F57AWU5X/Nestor - 1960 - REACTOR PHYSICS CALCULATIONS FOR THE MSRE.pdf:application/pdf;Snapshot:/home/huff/Zotero/storage/T28V6VJF/metadc868473.html:text/html}
}

@techreport{bucholz_scale:_1982,
	type = {{NUREG}},
	title = {{SCALE}: {A} modular code system for performing standardized computer analyses for licensing evaluation},
	shorttitle = {{SCALE}},
	url = {http://www.osti.gov/scitech/biblio/5360496},
	number = {NUREG/CR-0200-Vol.2-Bk.2; ORNL/NUREG/CSD-2-Vol.2-Bk.2 ON: DE82013370; TRN: 82-013156},
	urldate = {2016-10-06},
	institution = {Oak Ridge National Lab., TN (USA)},
	author = {Bucholz, J. A.},
	year = {1982},
	file = {Snapshot:/home/huff/Zotero/storage/ERDE7T4B/5360496.html:text/html}
}

@misc{wikipedia_flibe_2016,
	title = {{FLiBe}},
	copyright = {Creative Commons Attribution-ShareAlike License},
	url = {https://en.wikipedia.org/w/index.php?title=FLiBe&oldid=742978740},
	abstract = {FLiBe is a molten salt made from a mixture of lithium fluoride (LiF) and beryllium fluoride (BeF2). It is both a nuclear reactor coolant and solvent for fertile or fissile material. It served both purposes in the Molten-Salt Reactor Experiment (MSRE).
The 2:1 mixture forms a stoichiometric compound, Li2BeF4, which has a melting point of 459 {\textdegree}C, a boiling point of 1430 {\textdegree}C, and a density of 1.94 g/cm3. Its volumetric heat capacity is 4540 kJ/m3K, which is similar to that of water, more than four times that of sodium, and more than 200 times that of helium at typical reactor conditions. Its appearance is white to transparent, with crystalline grains in a solid state, morphing into a completely clear liquid upon melting. However, soluble fluorides such as UF4 and NiF2, can dramatically change the color salt in both solid and liquid state. This made spectrophotometry a viable analysis tool, and it was employed extensively during the MSRE Operations
The eutectic mixture is slightly greater than 50\% BeF2 and has a melting point of 360 {\textdegree}C. This mixture was never used in practice due to the overwhelming increase in viscosity caused by the BeF2 addition in the eutectic mixture. BeF2, which behaves as a glass, is only fluid in salt mixtures containing enough molar percent of Lewis base. Lewis bases, such as the alkali fluorides, will donate fluoride ions to the beryllium, breaking the glassy bonds which increase viscosity. In FLiBe, beryllium fluoride is able to sequester two fluoride ions from two lithium fluorides in a liquid state, converting it into the tetrafluorberyllate ion BeF4-2.},
	language = {en},
	urldate = {2016-10-19},
	journal = {Wikipedia},
	author = {{Wikipedia}},
	month = oct,
	year = {2016},
	note = {Page Version ID: 742978740},
	file = {Snapshot:/home/huff/Zotero/storage/9UZE8MM3/index.html:text/html}
}

@techreport{satish_balay_petsc_2015,
	title = {{PETSc} {Users} {Manual}},
	url = {http://www.mcs.anl.gov/petsc},
	number = {ANL-95/11 - Revision 3.6},
	institution = {Argonne National Laboratory},
	author = {{Satish Balay} and {Shrirang Abhyankar} and {Mark Adams} and {Jed Brown} and {Peter Brune} and {Kris Buschelman} and {Lisandro Dalcin} and {Victor Eijkhout} and {William Grop} and {Dinesh Kaushik} and {Matthew Knepley} and {Lois Curfman McInnes} and {Karl Rupp} and {Barry Smith} and {Stefano Zampini} and {Hong Zhang}},
	year = {2015},
	file = {PetScDocv3pt6.pdf:/home/huff/Zotero/storage/BKHJAPZA/PetScDocv3pt6.pdf:application/pdf}
}

@misc{flynn_1000_1972,
	title = {1000 {MW}(e) {MOLTEN} {SALT} {BREEDER} {REACTOR} {CONCEPTUAL} {DESIGN} {STUDY} {FINAL} {REPORT} {TASK} {I} [{Disc} 2] - {TID}-26156.pdf},
	url = {http://moltensalt.org/references/static/downloads/pdf/TID-26156.pdf},
	urldate = {2017-04-06},
	author = {Flynn, T.A. and deBoisblanc, D.R.},
	month = jul,
	year = {1972}
}

@article{turk_yt:_2011,
	title = {yt: {A} {Multi}-code {Analysis} {Toolkit} for {Astrophysical} {Simulation} {Data}},
	volume = {192},
	issn = {0067-0049},
	shorttitle = {yt},
	url = {http://adsabs.harvard.edu/abs/2011ApJS..192....9T},
	doi = {10.1088/0067-0049/192/1/9},
	abstract = {The analysis of complex multiphysics astrophysical simulations presents a unique and rapidly growing set of challenges: reproducibility, parallelization, and vast increases in data size and complexity chief among them. In order to meet these challenges, and in order to open up new avenues for collaboration between users of multiple simulation platforms, we present yt (available at http://yt.enzotools.org/) an open source, community-developed astrophysical analysis and visualization toolkit. Analysis and visualization with yt are oriented around
physically relevant quantities rather than quantities native to
astrophysical simulation codes. While originally designed for handling Enzo's structure adaptive mesh refinement data, yt has been extended to work with several different simulation methods and simulation codes including Orion, RAMSES, and FLASH. We report on its methods for reading, handling, and visualizing data, including projections,
multivariate volume rendering, multi-dimensional histograms, halo finding, light cone generation, and topologically connected isocontour identification. Furthermore, we discuss the underlying algorithms yt uses for processing and visualizing data, and its mechanisms for parallelization of analysis tasks.},
	language = {en},
	number = {1},
	urldate = {2017-02-27},
	journal = {The Astrophysical Journal Supplement Series},
	author = {Turk, Matthew J. and Smith, Britton D. and Oishi, Jeffrey S. and Skory, Stephen and Skillman, Samuel W. and Abel, Tom and Norman, Michael L.},
	month = jan,
	year = {2011},
	keywords = {cosmology: theory, methods: data analysis, methods: numerical},
	pages = {9},
	file = {IOP Full Text PDF:/home/huff/Zotero/storage/MPG8Z49A/Turk et al. - 2011 - yt A Multi-code Analysis Toolkit for Astrophysica.pdf:application/pdf;NASA/ADS Full Text PDF:/home/huff/Zotero/storage/W2Z6F89H/Turk et al. - 2011 - yt A Multi-code Analysis Toolkit for Astrophysica.pdf:application/pdf}
}

@misc{lindsay_moltres_2017,
	title = {Moltres, a code for simulating {Molten} {Salt} {Reactors}},
	url = {https://github.com/arfc/moltres},
	abstract = {moltres - Repository for Moltres, a code for simulating Molten Salt Reactors},
	urldate = {2017-04-10},
	journal = {GitHub},
	author = {Lindsay, Alexander},
	year = {2017},
	note = {https://github.com/arfc/moltres},
	file = {Snapshot:/home/huff/Zotero/storage/FZUHJB6V/moltres.html:text/html}
}

@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:/home/huff/Zotero/storage/9UB445Q8/github.com.html:text/html}
}

@techreport{thorcon_-able_2017,
	address = {Stevenson, WA},
	title = {The {Do}-able {Molten} {Salt} {Reactor}},
	language = {English},
	institution = {ThorCon USA Inc},
	author = {{ThorCon}},
	month = feb,
	year = {2017},
	file = {thorcon_msr.pdf:/home/huff/Zotero/storage/FG3D7JRT/thorcon_msr.pdf:application/pdf}
}

@techreport{transatomic_power_corporation_neutronics_2016,
	address = {Cambridge, MA, United States},
	type = {White {Paper}},
	title = {Neutronics {Overview}},
	url = {http://www.transatomicpower.com/new-neutronics-white-paper-2/},
	abstract = {For nuclear energy to be a viable source of baseload power, new nuclear reactor designs must address existing concerns
about waste storage by increasing fuel utilization and reducing overall waste production. This paper outlines the ways
in which the Transatomic Power 1250 MWth molten salt reactor design takes advantage of its liquid fuel in order to
address these challenges. By employing continuous fission product removal and core geometry modification, the TAP
MSR achieves more than twice the fuel utilization of light water reactors (LWRs). When using 5\% enriched uranium
{\textendash} the maximum enrichment readily available in the current supply chain {\textendash} this increased efficiency leads to an
approximate 53\% waste reduction compared to LWRs. Using higher enrichments, up to the 20\% Low Enriched
Uranium (LEU) limit, this reduction reaches 83\%.},
	language = {English},
	number = {1.1},
	institution = {Transatomic Power Corporation},
	author = {{Transatomic Power Corporation}},
	month = nov,
	year = {2016},
	file = {transatomic_Neutronics-White-Paper-v1.1.pdf:/home/huff/Zotero/storage/NISQFVGT/transatomic_Neutronics-White-Paper-v1.1.pdf:application/pdf}
}

@techreport{gif_generation_2015,
	title = {Generation {IV} {International} {Forum} 2015 {Annual} {Report}},
	language = {English},
	institution = {Generation IV International Forum},
	author = {{GIF}},
	year = {2015},
	file = {gif_2015_annual_report_-_final_e-book_v2_sept2016.pdf:/home/huff/Zotero/storage/RN7M8GRE/gif_2015_annual_report_-_final_e-book_v2_sept2016.pdf:application/pdf}
}

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

@article{scarlat_design_2014,
	title = {Design and licensing strategies for the fluoride-salt-cooled, high-temperature reactor ({FHR}) technology},
	volume = {77},
	issn = {0149-1970},
	url = {http://www.sciencedirect.com/science/article/pii/S0149197014001838},
	doi = {10.1016/j.pnucene.2014.07.002},
	abstract = {Fluoride-salt-cooled, high-temperature reactor (FHR) technology combines the robust coated-particle fuel of high-temperature, gas-cooled reactors with the single phase, high volumetric heat capacity coolant of molten salt reactors and the low-pressure pool-type reactor configuration of sodium fast reactors. FHRs have the capacity to deliver heat at high average temperature, and thus to achieve higher thermal efficiency than light water reactors. Licensing of the passive safety systems used in FHRs can use the same framework applied successfully to passive advanced light water reactors, and earlier work by the NGNP and PBMR projects provide an appropriate framework to guide the design of safety-relevant FHR systems. This paper provides a historical review of the development of FHR technology, describes ongoing development efforts, and presents design and licensing strategies for FHRs. A companion review article describes the phenomenology, methods and experimental program in support of FHR.},
	urldate = {2017-05-11},
	journal = {Progress in Nuclear Energy},
	author = {Scarlat, Raluca O. and Laufer, Michael R. and Blandford, Edward D. and Zweibaum, Nicolas and Krumwiede, David L. and Cisneros, Anselmo T. and Andreades, Charalampos and Forsberg, Charles W. and Greenspan, Ehud and Hu, Lin-Wen and Peterson, Per F.},
	month = nov,
	year = {2014},
	keywords = {Licensing framework and reactor design, Liquid fluoride salt, Passive safety systems, reactor design, molten salt},
	pages = {406--420},
	file = {ScienceDirect Full Text PDF:/home/huff/Zotero/storage/BEIT5HHS/Scarlat et al. - 2014 - Design and licensing strategies for the fluoride-s.pdf:application/pdf;ScienceDirect Snapshot:/home/huff/Zotero/storage/429E8ZBF/S0149197014001838.html:text/html}
}

@article{betzler_molten_2017,
	title = {Molten salt reactor neutronics and fuel cycle modeling and simulation with {SCALE}},
	volume = {101},
	issn = {03064549},
	url = {http://linkinghub.elsevier.com/retrieve/pii/S0306454916309185},
	doi = {10.1016/j.anucene.2016.11.040},
	language = {en},
	urldate = {2017-04-06},
	journal = {Annals of Nuclear Energy},
	author = {Betzler, Benjamin R. and Powers, Jeffrey J. and Worrall, Andrew},
	month = mar,
	year = {2017},
	keywords = {Depletion, Molten salt reactors, Salt separations, Salt treatment, fuel cycle},
	pages = {489--503}
}

@article{fiorina_modelling_2014,
	title = {Modelling and analysis of the {MSFR} transient behaviour},
	volume = {64},
	issn = {0306-4549},
	url = {http://www.sciencedirect.com/science/article/pii/S0306454913004118},
	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{\textendash}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 {\textendash} 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},
	urldate = {2017-10-03},
	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:/home/huff/Zotero/storage/IXFXRASN/Fiorina et al. - 2014 - Modelling and analysis of the MSFR transient behav.pdf:application/pdf;ScienceDirect Snapshot:/home/huff/Zotero/storage/9WNRAF3E/S0306454913004118.html:text/html}
}

@article{ridley_method_2017,
	title = {A method for predicting fuel maintenance in once-through {MSRs}},
	volume = {110},
	issn = {0306-4549},
	url = {http://www.sciencedirect.com/science/article/pii/S030645491730021X},
	doi = {10.1016/j.anucene.2017.06.043},
	abstract = {Liquid fuel molten salt reactors allow reactivity control by material addition. This paper presents a method to adjust material flows in a molten salt reactor to keep the core critical, and to maintain desired reduction-oxidation potential in the core salt melt. The method is aimed at low-enriched uranium fueled thermal systems. It is developed as a Python library and uses Serpent2 Monte-Carlo transport and depletion code. A toy 300MW(th) reactor with a FLiBe carrier salt is employed to demonstrate the performance of the method over 10 full power years. Results of the calculation are presented, including material flows, conversion ratio, effective delayed neutron fraction, and expected limits on trifluoride concentrations and graphite lifetime are investigated. This method lays a foundation for future studies including fuel cycle performance of molten salt reactors and dynamic behavior of the core during depletion.},
	urldate = {2017-09-14},
	journal = {Annals of Nuclear Energy},
	author = {Ridley, Gavin and Chvala, Ondrej},
	month = dec,
	year = {2017},
	keywords = {DMSR, Depletion, Serpent, Chemistry control, On-line refueling, molten salt reactor},
	pages = {265--281},
	file = {ScienceDirect Full Text PDF:/home/huff/Zotero/storage/F27JL6PA/Ridley and Chvala - 2017 - A method for predicting fuel maintenance in once-t.pdf:application/pdf;ScienceDirect Snapshot:/home/huff/Zotero/storage/9BPX5LGR/S030645491730021X.html:text/html}
}

@article{moser_lattice_2017,
	title = {Lattice optimization for graphite moderated molten salt reactors using low-enriched uranium fuel},
	volume = {110},
	issn = {0306-4549},
	url = {http://www.sciencedirect.com/science/article/pii/S0306454917301627},
	doi = {10.1016/j.anucene.2017.06.015},
	abstract = {The family of Generation-IV reactor concepts comprises of multiple promising designs, one of which is the molten salt reactor. These reactors have traditionally been chosen for the possible use of the thorium-uranium fuel cycle and used 7LiF-BeF2 carrier salt. This particular salt choice however presents several challenges due to the cost of highly depleted 7Li isotope for the carrier salt, tritium production, and beryllium toxicity. Additionally, lack of developed and accepted safeguards methodology for thorium fuel cycle presents a barrier. While none of these issues are insurmountable, alternatives are worth investigating. The purpose of this paper is to analyze the more cost effective and regulatory amenable fuel salt choices by using low-enriched uranium fuel in the form of UF4. Several eutectic mixtures are examined that avoid the use of 7Li and Be while maintaining a melting point low enough to be compatible with standard structural materials. The optimal conditions for hexagonal lattice arrangements using nuclear graphite moderation are discussed for multiple fuel salt choices. The aim of this study is to present options of using simpler molten salt reactor alternatives focused on thermal single-fluid low-enriched uranium converter concepts.},
	urldate = {2017-09-13},
	journal = {Annals of Nuclear Energy},
	author = {Moser, Dallas and Wheeler, Alexander and Chv{\'a}la, Ond{\v r}ej},
	month = dec,
	year = {2017},
	keywords = {DMSR, MCNP, Serpent, molten salt reactor},
	pages = {1--10},
	file = {ScienceDirect Full Text PDF:/home/huff/Zotero/storage/289SFDB2/Moser et al. - 2017 - Lattice optimization for graphite moderated molten.pdf:application/pdf;ScienceDirect Snapshot:/home/huff/Zotero/storage/JLFQ2WUD/S0306454917301627.html:text/html}
}

@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://www.osti.gov/scitech/servlets/purl/4466863},
	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}
}

@article{weller_tensorial_1998,
	title = {A tensorial approach to computational continuum mechanics using object-oriented techniques},
	volume = {12},
	url = {http://scitation.aip.org/content/aip/journal/cip/12/6/10.1063/1.168744},
	number = {6},
	journal = {Computers in physics},
	author = {Weller, Henry G. and Tabor, G. and Jasak, Hrvoje and Fureby, C.},
	year = {1998},
	pages = {620--631},
	file = {[PDF] semanticscholar.org:/home/huff/Zotero/storage/5PX5BA6R/Weller et al. - 1998 - A tensorial approach to computational continuum me.pdf:application/pdf;Snapshot:/home/huff/Zotero/storage/JIM2283D/1.html:text/html}
}

@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}},
	url = {https://link.springer.com/article/10.1007/s00366-006-0049-3},
	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},
	urldate = {2017-04-11},
	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:/home/huff/Zotero/storage/2Z9B88E4/Kirk et al. - 2006 - libMesh a C++ library for parallel adaptive mesh .pdf:application/pdf;Snapshot:/home/huff/Zotero/storage/CJ6ZDS26/10.html:text/html}
}

@article{dehart_reactor_2011,
	title = {Reactor {Physics} {Methods} and {Analysis} {Capabilities} in {SCALE}},
	volume = {174},
	url = {http://epubs.ans.org/?a=11720},
	doi = {dx.doi.org/10.13182/NT174-196},
	number = {2},
	urldate = {2017-04-10},
	journal = {Nuclear Technology},
	author = {DeHart, Mark D. and Bowman, Stephen M.},
	month = may,
	year = {2011},
	pages = {196--213},
	file = {Snapshot:/home/huff/Zotero/storage/BHMIBIJ9/epubs.ans.org.html:text/html}
}

@article{leppanen_serpent_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 = {The {Serpent} {Monte} {Carlo} code: {Status}, development and applications in 2013},
	volume = {82},
	issn = {0306-4549},
	shorttitle = {The {Serpent} {Monte} {Carlo} code},
	url = {http://www.sciencedirect.com/science/article/pii/S0306454914004095},
	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.},
	urldate = {2017-04-10},
	journal = {Annals of Nuclear Energy},
	author = {Lepp{\"a}nen, Jaakko and Pusa, Maria and Viitanen, Tuomas and Valtavirta, Ville and Kaltiaisenaho, Toni},
	month = aug,
	year = {2015},
	keywords = {Monte Carlo, Serpent, Burnup calculation, Homogenization, reactor physics},
	pages = {142--150},
	file = {ScienceDirect Full Text PDF:/home/huff/Zotero/storage/83V2KXJ9/Lepp{\"a}nen et al. - 2015 - The Serpent Monte Carlo code Status, development .pdf:application/pdf;ScienceDirect Snapshot:/home/huff/Zotero/storage/AQAB5875/S0306454914004095.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},
	url = {http://www.sciencedirect.com/science/article/pii/S030645491400543X},
	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{\textemdash}in a coupled, multiscale manner{\textemdash}crucial reactor phenomena such as CRUD-induced power shift and fuel shuffle.},
	urldate = {2017-04-10},
	journal = {Annals of Nuclear Energy},
	author = {Gaston, Derek R. and Permann, Cody J. and Peterson, John W. and Slaughter, Andrew E. and Andr{\v s}, 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 = {ScienceDirect Full Text PDF:/home/huff/Zotero/storage/SGQXWGJV/Gaston et al. - 2015 - Physics-based multiscale coupling for full core nu.pdf:application/pdf;ScienceDirect Snapshot:/home/huff/Zotero/storage/PI8FZ93W/S030645491400543X.html:text/html}
}

@article{guo_effects_2013,
	title = {The effects of core zoning on optimization of design analysis of molten salt reactor},
	volume = {265},
	issn = {0029-5493},
	url = {http://www.sciencedirect.com/science/article/pii/S002954931300544X},
	doi = {10.1016/j.nucengdes.2013.09.036},
	abstract = {The molten salt reactor (MSR) is one of six advanced reactor types in the frame of the Generation 4 International Forum. In this study, a multiple-channel analysis code (MAC) is developed to analyze thermal-hydraulics behavior and MCNP4c is used to study the neutronics behavior of Molten Salt Reactor Experiment (MSRE). The MAC calculates thermal-hydraulic parameters, namely temperature distribution, flow distribution and pressure drop. The MCNP4c performs the analysis of effective multiplication factor, neutron flux, power distribution and conversion ratio. In this work, the modification of core configuration is achieved by different core zoning and various fuel channel diameters, contributing to flat flux distribution. Specifically, the core is divided into two regions and the effects of different core zoning on the both neutronics and thermal-hydraulic behavior of moderated molten salt reactor are investigated. We conclude that the flat flux distribution cannot always guarantee better performance in thermal-hydraulic perspective and can decreases the graphite lifetime significantly.},
	journal = {Nuclear Engineering and Design},
	author = {Guo, Zhangpeng and Wang, Chenglong and Zhang, Dalin and Chaudri, Khurrum Saleem and Tian, Wenxi and Su, Guanghui and Qiu, Suizheng},
	month = dec,
	year = {2013},
	pages = {967--977},
	file = {ScienceDirect Full Text PDF:/home/huff/Zotero/storage/NH6H24GQ/Guo et al. - 2013 - The effects of core zoning on optimization of desi.pdf:application/pdf;ScienceDirect Snapshot:/home/huff/Zotero/storage/EV6DDG4D/S002954931300544X.html:text/html}
}

@article{zanetti_geometric_2015,
	title = {A {Geometric} {Multiscale} modelling approach to the analysis of {MSR} plant dynamics},
	volume = {83},
	issn = {0149-1970},
	url = {https://www.sciencedirect.com/science/article/pii/S0149197015000487},
	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 {\textendash} 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 {\textendash} e.g., the cooling loop {\textendash} 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.},
	urldate = {2017-02-08},
	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 (MSR), Geometric Multiscale approach, Molten Salt Reactor Experiment (MSRE), Multi-Physics Modelling, System dynamic behaviour},
	pages = {82--98},
	file = {ScienceDirect Full Text PDF:/home/huff/Zotero/storage/JIPBQTRD/Zanetti et al. - 2015 - A Geometric Multiscale modelling approach to the a.pdf:application/pdf;ScienceDirect Snapshot:/home/huff/Zotero/storage/NTTJSHEN/S0149197015000487.html:text/html}
}

@article{dulla_neutron_2004,
	title = {Neutron kinetics of fluid{\textendash}fuel systems by the quasi-static method},
	volume = {31},
	issn = {0306-4549},
	url = {http://www.sciencedirect.com/science/article/pii/S0306454904000982},
	doi = {10.1016/j.anucene.2004.05.004},
	abstract = {The quasi-static method for the neutron kinetics of nuclear reactors is generalized for application to neutron multiplying systems fueled by a fluid multiplying material, typically a mixture of fissile molten salts. The method is derived by the application of factorization formulae for both the neutron density and the delayed precursor concentrations and the projection of the balance equations upon a weighting function. A physically meaningful weight can be assumed as the solution of the adjoint model, which is constructed for the situation considered, including delayed neutrons. The quasi-static scheme is then applied to calculations of some transients for a typical configuration of a molten-salt reactor, in a multigroup diffusion model with a one-dimensional slug-flow velocity field. The physical features associated to the motion of the fissile material are highlighted.},
	number = {15},
	urldate = {2016-09-20},
	journal = {Annals of Nuclear Energy},
	author = {Dulla, S. and Ravetto, P. and Rostagno, M. M.},
	month = oct,
	year = {2004},
	pages = {1709--1733},
	file = {ScienceDirect Full Text PDF:/home/huff/Zotero/storage/FNHFCRSH/Dulla et al. - 2004 - Neutron kinetics of fluid{\textendash}fuel systems by the quas.pdf:application/pdf;ScienceDirect Snapshot:/home/huff/Zotero/storage/QTQT4S83/S0306454904000982.html:text/html}
}

@article{haubenreich_experience_1970,
	title = {Experience with the {Molten}-{Salt} {Reactor} {Experiment}},
	volume = {8},
	issn = {00295450},
	url = {http://www.ans.org/pubs/journals/nt/a_28620},
	doi = {10.13182/NT8-2-118},
	number = {2},
	urldate = {2016-09-06},
	journal = {Nuclear Technology},
	author = {Haubenreich, Paul N. and Engel, J. R.},
	month = feb,
	year = {1970},
	pages = {118--136},
	file = {Haubenreich_Engel_MSREexperience.pdf:/home/huff/Zotero/storage/6EIEWPDA/Haubenreich_Engel_MSREexperience.pdf:application/pdf}
}

@article{krepel_dyn3d-msr_2007,
	title = {{DYN}3D-{MSR} spatial dynamics code for molten salt reactors},
	volume = {34},
	issn = {0306-4549},
	url = {http://www.sciencedirect.com/science/article/pii/S0306454907000527},
	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 {\textendash} one of the {\textquoteleft}Generation IV{\textquoteright} concepts {\textendash} 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.},
	number = {6},
	urldate = {2016-08-17},
	journal = {Annals of Nuclear Energy},
	author = {K{\v r}epel, Ji{\v r}{\'i} and Rohde, Ulrich and Grundmann, Ulrich and Weiss, Frank-Peter},
	month = jun,
	year = {2007},
	pages = {449--462},
	file = {ScienceDirect Full Text PDF:/home/huff/Zotero/storage/T5JKV5U6/K{\v r}epel et al. - 2007 - DYN3D-MSR spatial dynamics code for molten salt re.pdf:application/pdf;ScienceDirect Snapshot:/home/huff/Zotero/storage/Q24HJQHD/S0306454907000527.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},
	url = {http://www.sciencedirect.com/science/article/pii/S0009250914001146},
	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{\textendash}Kutta scheme and the coupling among neutron diffusion, Reynolds-Averaged Navier{\textendash}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, Herv{\'e}},
	month = may,
	year = {2014},
	keywords = {Reactor dynamics, Molten Salt Reactor (MSR), Molten Salt Fast Reactor (MSFR), Multiphysics, OpenFOAM},
	pages = {390--401},
	file = {ScienceDirect Full Text PDF:/home/huff/Zotero/storage/GFZ6CW4E/Aufiero et al. - 2014 - Development of an OpenFOAM model for the Molten Sa.pdf:application/pdf;ScienceDirect Snapshot:/home/huff/Zotero/storage/L4BZULIM/S0009250914001146.html:text/html;ScienceDirect Snapshot:/home/huff/Zotero/storage/TGWQHMHK/S0009250914001146.html:text/html}
}

@techreport{mark_mitchell_markummitchell/engauge-digitizer:_2017,
	title = {markummitchell/engauge-digitizer: {Version} 10.1 {Export} {Improvements}},
	shorttitle = {markummitchell/engauge-digitizer},
	url = {https://zenodo.org/record/832444},
	institution = {Zenodo},
	author = {Mark Mitchell and Baurzhan Muftakhidinov and Tobias Winchen and Zbigniew J{\k e}drzejewski-Szmek and The Gitter Badger and badshah400},
	month = jul,
	year = {2017},
	note = {DOI: 10.5281/zenodo.832444},
	file = {Zenodo Snapshot:/home/huff/Zotero/storage/234FAE57/832444.html:text/html}
}

@patent{leblanc_integral_2015,
	title = {Integral molten salt reactor},
	url = {http://www.google.com/patents/WO2015017928A1},
	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},
	urldate = {2017-05-11},
	author = {Leblanc, David},
	month = feb,
	year = {2015}
}

@patent{hyde_liquid_2015,
	title = {Liquid fuel nuclear fission reactor},
	url = {http://www.google.com/patents/US9183953},
	abstract = {Disclosed embodiments include nuclear fission reactors, nuclear fission fuel pins, methods of operating a nuclear fission reactor, methods of fueling a nuclear fission reactor, and methods of fabricating a nuclear fission fuel pin.},
	nationality = {United States},
	assignee = {Terrapower, Llc},
	number = {US9183953 B2},
	urldate = {2017-05-11},
	author = {Hyde, Roderick A. and McWhirter, Jon D.},
	month = nov,
	year = {2015}
}

@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},
	url = {http://www.sciencedirect.com/science/article/pii/S002954931730081X},
	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 {\textquotedblleft}Transient Fission Matrix{\textquotedblright} 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},
	urldate = {2017-11-21},
	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 = {Neutronics, MSFR, Thermal hydraulics, Transient calculation, Transient Fission Matrix},
	pages = {112--124},
	file = {ScienceDirect Full Text PDF:/home/huff/Zotero/storage/GEYY7UJ9/Laureau et al. - 2017 - Transient coupled calculations of the Molten Salt .pdf:application/pdf;ScienceDirect Snapshot:/home/huff/Zotero/storage/5B25JVQI/S002954931730081X.html:text/html}
}

@article{de_zwaan_static_2007,
	title = {Static design of a liquid-salt-cooled pebble bed reactor ({LSPBR})},
	volume = {34},
	number = {1},
	journal = {Annals of Nuclear Energy},
	author = {De Zwaan, S. J. and Boer, B. and Lathouwers, D. and Kloosterman, J. L.},
	year = {2007},
	pages = {83--92},
	file = {Fulltext:/home/huff/Zotero/storage/6Y9JW99W/S0306454906002295.html:text/html;Snapshot:/home/huff/Zotero/storage/SRXR2CFA/S0306454906002295.html:text/html}
}

@article{van_der_linden_coupled_2012,
	title = {Coupled neutronics and computational fluid dynamics for the molten salt fast reactor},
	journal = {Delft University of Technology},
	author = {van der Linden, Erik},
	year = {2012}
}

@article{biondo_quality_2014,
	title = {Quality assurance within the {PyNE} open source toolkit},
	volume = {111},
	journal = {Transactions of the American Nuclear Society},
	author = {Biondo, Elliott and Scopatz, Anthony and Gidden, Matthew and Slaybaugh, Rachel and Bates, Cameron and Wilson, Paul PH},
	year = {2014},
	file = {Fulltext:/home/huff/Zotero/storage/SB5TMMJ2/Biondo et al. - 2014 - Quality assurance within the PyNE open source tool.pdf:application/pdf;Snapshot:/home/huff/Zotero/storage/VVP5V367/Biondo et al. - 2014 - Quality assurance within the PyNE open source tool.pdf:application/pdf}
}

@article{bates_pyne_2014,
	title = {{PyNE} progress report},
	volume = {111},
	journal = {Transactions of the American Nuclear Society},
	author = {Bates, Cameron R. and Biondo, Elliott and Huff, Kathryn and Kiesling, Kalin and Scopatz, Anthony and Carlsen, Robert and Davis, Andrew and Gidden, Matthew and Haines, Tim and Howland, Joshua},
	year = {2014},
	file = {Fulltext:/home/huff/Zotero/storage/LDIWXJRV/Bates et al. - 2014 - PyNE progress report.pdf:application/pdf;Snapshot:/home/huff/Zotero/storage/VP45CSRG/Bates et al. - 2014 - PyNE progress report.pdf:application/pdf}
}

@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}},
	url = {http://www.sciencedirect.com/science/article/pii/S030645491400379X},
	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},
	urldate = {2017-11-22},
	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 = {HDF5, Monte Carlo, Neutron transport, OpenMC, Parallel, XML},
	pages = {90--97},
	file = {ScienceDirect Full Text PDF:/home/huff/Zotero/storage/9S36DPTN/Romano et al. - 2015 - OpenMC A state-of-the-art Monte Carlo code for re.pdf:application/pdf;ScienceDirect Snapshot:/home/huff/Zotero/storage/NZ4279HS/S030645491400379X.html:text/html}
}

@article{boyd_openmoc_2014,
	title = {The {OpenMOC} method of characteristics neutral particle transport code},
	volume = {68},
	issn = {0306-4549},
	url = {http://www.sciencedirect.com/science/article/pii/S0306454913006634},
	doi = {10.1016/j.anucene.2013.12.012},
	abstract = {The method of characteristics (MOC) is a numerical integration technique for partial differential equations, and has seen widespread use for reactor physics lattice calculations. The exponential growth in computing power has finally brought the possibility for high-fidelity full core MOC calculations within reach. The OpenMOC code is being developed at the Massachusetts Institute of Technology to investigate algorithmic acceleration techniques and parallel algorithms for MOC. OpenMOC is a free, open source code written using modern software languages such as C/C++ and CUDA with an emphasis on extensible design principles for code developers and an easy to use Python interface for code users. The present work describes the OpenMOC code and illustrates its ability to model large problems accurately and efficiently.},
	number = {Supplement C},
	urldate = {2017-11-22},
	journal = {Annals of Nuclear Energy},
	author = {Boyd, William and Shaner, Samuel and Li, Lulu and Forget, Benoit and Smith, Kord},
	month = jun,
	year = {2014},
	keywords = {Criticality, High performance computing, Method of characteristics, Neutron transport, Nonlinear diffusion acceleration, Open source},
	pages = {43--52},
	file = {ScienceDirect Full Text PDF:/home/huff/Zotero/storage/MX322CHB/Boyd et al. - 2014 - The OpenMOC method of characteristics neutral part.pdf:application/pdf;ScienceDirect Snapshot:/home/huff/Zotero/storage/MDQMMU6K/S0306454913006634.html:text/html}
}

@article{heil2008solvers,
  title={Solvers for large-displacement fluid--structure interaction problems: segregated versus monolithic approaches},
  author={Heil, Matthias and Hazel, Andrew L and Boyle, Jonathan},
  journal={Computational Mechanics},
  volume={43},
  number={1},
  pages={91--101},
  year={2008},
  publisher={Springer}
}