2018-msbr-reproc.bib

@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}},
	url = {http://www.sciencedirect.com/science/article/pii/S0149197014000456},
	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},
	urldate = {2017-10-22},
	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 = {Fuel cycle, Gen IV, Molten salt reactor, Neutronic performance, Nuclear systems},
	pages = {308--319},
	file = {ScienceDirect Full Text PDF:/home/andrei2/Zotero/storage/Q5WDR3ER/Serp et al. - 2014 - The molten salt reactor (MSR) in generation IV Ov.pdf:application/pdf;ScienceDirect Full Text PDF:/home/andrei2/Zotero/storage/WWISZQDL/Serp et al. - 2014 - The molten salt reactor (MSR) in generation IV Ov.pdf:application/pdf;ScienceDirect Full Text PDF:/home/andrei2/Zotero/storage/Q5UID2I9/Serp et al. - 2014 - The molten salt reactor (MSR) in generation IV Ov.pdf:application/pdf;ScienceDirect Snapshot:/home/andrei2/Zotero/storage/ZRQWCMGZ/S0149197014000456.html:text/html;ScienceDirect Snapshot:/home/andrei2/Zotero/storage/QGLMLNIW/S0149197014000456.html:text/html;ScienceDirect Snapshot:/home/andrei2/Zotero/storage/FABGBG8S/S0149197014000456.html:text/html}
}

@misc{terrapower_llc_mcfr_2018,
	title = {{MCFR} {TerraPower}},
	url = {http://terrapower.com/technologies/mcfr},
	urldate = {2018-04-04},
	author = {{TerraPower LLC}},
	month = dec,
	year = {2018},
	file = {MCFR TerraPower:/home/andrei2/Zotero/storage/VX26VEMH/mcfr.html:text/html;MCFR TerraPower:/home/andrei2/Zotero/storage/EFN8LB8Y/mcfr.html:text/html}
}

@article{leppanen_numerical_2015,
	series = {Multi-{Physics} {Modelling} of {LWR} {Static} and {Transient} {Behaviour}},
	title = {The {Numerical} {Multi}-{Physics} project ({NUMPS}) at {VTT} {Technical} {Research} {Centre} of {Finland}},
	volume = {84},
	issn = {0306-4549},
	url = {http://www.sciencedirect.com/science/article/pii/S0306454914005532},
	doi = {10.1016/j.anucene.2014.10.014},
	abstract = {The four-year Numerical Multi-Physics (NUMPS) project funded by the Academy of Finland was initiated at VTT Technical Research Centre of Finland in September 2012, for the purpose of studying and developing high-fidelity computational methods for nuclear reactor analysis. The project is built around calculation codes developed at VTT, and it aims at the coupled three-dimensional modeling of neutronics, thermal hydraulics and fuel behavior of nuclear reactors. The work involves the continuous-energy Monte Carlo code Serpent and CFD code PORFLO, together with two light-weight solvers, COSY and FINIX, coupled to Serpent at source code level. This paper is a review on the current status and development activities, reflecting the status of the NUMPS project at the beginning of its second complete year.},
	urldate = {2018-02-06},
	journal = {Annals of Nuclear Energy},
	author = {Leppanen, Jaakko and Hovi, Ville and Ikonen, Timo and Kurki, Joona and Pusa, Maria and Valtavirta, Ville and Viitanen, Tuomas},
	month = oct,
	year = {2015},
	keywords = {CFD, FINIX, Monte Carlo, Multi-physics, PORFLO},
	pages = {55--62},
	file = {ScienceDirect Full Text PDF:/home/andrei2/Zotero/storage/C3FKW8SU/Leppänen et al. - 2015 - The Numerical Multi-Physics project (NUMPS) at VTT.pdf:application/pdf;ScienceDirect Snapshot:/home/andrei2/Zotero/storage/VWAKC3VW/S0306454914005532.html:text/html}
}

@article{leppanen_serpent_2013,
	title = {Serpent -- a {Continuous}-energy {Monte} {Carlo} {Reactor} {Physics} {Burnup} {Calculation} {Code}},
	urldate = {2013-05-30},
	journal = {VTT Technical Research Centre of Finland, Espoo, Finland},
	author = {Leppanen, Jaakko},
	year = {2013},
	file = {[PDF] from vtt.fi:/home/andrei2/Zotero/storage/4XX8BJFI/Leppänen - 2012 - Serpent–a Continuous-energy Monte Carlo Reactor Ph.pdf:application/pdf}
}

@misc{rykhlevskii_arfc/saltproc:_2018,
	title = {arfc/saltproc: {Code} for online reprocessing simulation of {Molten} {Salt} {Reactor} with external depletion solver {SERPENT}},
	shorttitle = {arfc/saltproc},
	url = {https://zenodo.org/record/1196455#.WrkK_HXwZ9P},
	abstract = {The script is re-run SERPENT 2 burnup simulation, and after each step removes target isotopes with specific efficiency, keep number density of target isotopes constant and keep track of Pa-233 --{\textgreater} U-233 decaying in the separate tank.},
	urldate = {2018-03-26},
	publisher = {Zenodo},
	author = {Rykhlevskii, Andrei and Bae, Jin Whan and Huff, Kathryn},
	month = mar,
	year = {2018},
	doi = {10.5281/zenodo.1196455},
	keywords = {Fuel cycle, MSBR, MSR, online refueling, online reprocessing, salt separations, salt treatment, thorium cycle},
	file = {Zenodo Snapshot:/home/andrei2/Zotero/storage/7YGYH6A6/1196455.html:text/html}
}

@phdthesis{fiorina_molten_2013,
	type = {{PhD}},
	title = {The molten salt fast reactor as a fast spectrum candidate for thorium implementation},
	url = {https://www.politesi.polimi.it/handle/10589/74324},
	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.},
	urldate = {2013-05-28},
	school = {Politecnico Di Milano},
	author = {Fiorina, Carlo},
	month = mar,
	year = {2013},
	keywords = {unread},
	file = {[PDF] from polimi.it:/home/andrei2/Zotero/storage/NXBD45NV/FIORINA - 2013 - The molten salt fast reactor as a fast spectrum ca.pdf:application/pdf;Snapshot:/home/andrei2/Zotero/storage/GGQM6IZK/74324.html:text/html}
}

@inproceedings{alfonsi_raven_2013,
	title = {Raven as a tool for dynamic probabilistic risk assessment: {Software} overview},
	shorttitle = {Raven as a tool for dynamic probabilistic risk assessment},
	booktitle = {Proceeding of {M}\&{C}2013 {International} {Topical} {Meeting} on {Mathematics} and {Computation}},
	author = {Alfonsi, A. and Rabiti, C. and Mandelli, D. and Cogliati, J. and Kinoshita, R.},
	year = {2013},
	file = {Fulltext:/home/andrei2/Zotero/storage/XC77EAGF/Alfonsi et al. - 2013 - Raven as a tool for dynamic probabilistic risk ass.pdf:application/pdf;Snapshot:/home/andrei2/Zotero/storage/MNBLGC2D/Alfonsi et al. - 2013 - Raven as a tool for dynamic probabilistic risk ass.pdf:application/pdf}
}

@inproceedings{powers_new_2013,
	address = {Sun Valley, ID, USA},
	title = {A new approach for modeling and analysis of molten salt reactors using {SCALE}},
	url = {https://www.osti.gov/scitech/biblio/22212758},
	language = {English},
	urldate = {2017-06-26},
	publisher = {American Nuclear Society, 555 North Kensington Avenue, La Grange Park, IL 60526 (United States)},
	author = {Powers, J. J. and Harrison, T. J. and Gehin, J. C.},
	month = jul,
	year = {2013},
	file = {Powers et al. - 2013 - A new approach for modeling and analysis of molten.pdf:/home/andrei2/Zotero/storage/UBGE2IQG/Powers et al. - 2013 - A new approach for modeling and analysis of molten.pdf:application/pdf;Snapshot:/home/andrei2/Zotero/storage/RGBV8HHV/22212758.html:text/html}
}

@techreport{carter_design_1972,
	title = {{DESIGN} {AND} {COST} {STUDY} {OF} {A} {FLUORINATION}--{REDUCTIVE} {EXTRACTION}--{METAL} {TRANSFER} {PROCESSING} {PLANT} {FOR} {THE} {MSBR}.},
	url = {https://www.osti.gov/biblio/4667633},
	abstract = {The U.S. Department of Energy's Office of Scientific and Technical Information},
	language = {English},
	number = {ORNL-TM-3579},
	urldate = {2018-02-23},
	institution = {Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)},
	author = {Carter, W. L. and Nicholson, E. L.},
	month = jan,
	year = {1972},
	doi = {10.2172/4667633},
	file = {Full Text PDF:/home/andrei2/Zotero/storage/7WZLYWWQ/Carter and Nicholson - 1972 - DESIGN AND COST STUDY OF A FLUORINATION--REDUCTIVE.pdf:application/pdf;Snapshot:/home/andrei2/Zotero/storage/JRN95C2I/4667633.html:text/html}
}

@techreport{wigeland_nuclear_2014,
	title = {Nuclear {Fuel} {Cycle} {Evaluation} and {Screening} – {Final} {Report}},
	language = {en},
	number = {INL/EXT-14-31465},
	urldate = {2018-02-02},
	institution = {U.S. Department of Energy},
	author = {Wigeland, R and Taiwo, T and Ludewig, H and Todosow, M and Halsey, W},
	year = {2014},
	file = {ES Main Report.pdf:/home/andrei2/Zotero/storage/2JUQW4RD/ES Main Report.pdf:application/pdf}
}

@inproceedings{shen_progress_2017,
	address = {Washington, DC, United States},
	title = {Progress {Towards} a {Molten} {Salt} {Reactor} {Experiment} {Benchmark} {Evaluation}},
	booktitle = {Transactions of the {American} {Nuclear} {Society}},
	publisher = {American Nuclear Society},
	author = {Shen, Dan and Fratoni, Massimiliano and Aufiero, Manuele and Bidaud, Adrien and Powers, Jeffery and Ilas, Germina},
	month = nov,
	year = {2017},
	file = {trans_v117_n1_pp1359-1361.pdf:/home/andrei2/Zotero/storage/XTTI2FYQ/trans_v117_n1_pp1359-1361.pdf:application/pdf}
}

@article{pandya_implementation_2016,
	title = {Implementation, capabilities, and benchmarking of {Shift}, a massively parallel {Monte} {Carlo} radiation transport code},
	volume = {308},
	issn = {0021-9991},
	url = {http://www.sciencedirect.com/science/article/pii/S0021999115008566},
	doi = {10.1016/j.jcp.2015.12.037},
	abstract = {This work discusses the implementation, capabilities, and validation of Shift, a massively parallel Monte Carlo radiation transport package authored at Oak Ridge National Laboratory. Shift has been developed to scale well from laptops to small computing clusters to advanced supercomputers and includes features such as support for multiple geometry and physics engines, hybrid capabilities for variance reduction methods such as the Consistent Adjoint-Driven Importance Sampling methodology, advanced parallel decompositions, and tally methods optimized for scalability on supercomputing architectures. The scaling studies presented in this paper demonstrate good weak and strong scaling behavior for the implemented algorithms. Shift has also been validated and verified against various reactor physics benchmarks, including the Consortium for Advanced Simulation of Light Water Reactors' Virtual Environment for Reactor Analysis criticality test suite and several Westinghouse AP1000® problems presented in this paper. These benchmark results compare well to those from other contemporary Monte Carlo codes such as MCNP5 and KENO.},
	urldate = {2018-02-16},
	journal = {Journal of Computational Physics},
	author = {Pandya, Tara M. and Johnson, Seth R. and Evans, Thomas M. and Davidson, Gregory G. and Hamilton, Steven P. and Godfrey, Andrew T.},
	month = mar,
	year = {2016},
	keywords = {Monte Carlo methods, Neutron transport, Parallel computation},
	pages = {239--272},
	file = {ScienceDirect Full Text PDF:/home/andrei2/Zotero/storage/QLUATVRC/Pandya et al. - 2016 - Implementation, capabilities, and benchmarking of .pdf:application/pdf;ScienceDirect Snapshot:/home/andrei2/Zotero/storage/NUX5VGAS/S0021999115008566.html:text/html}
}

@inproceedings{skirpan_fuel_2017,
	address = {Washington, DC, United States},
	title = {Fuel {Cycle} {Modeling} and {Simulation} of the {Molten} {Salt} {Breeder} {Reactor}},
	booktitle = {Transactions of the {American} {Nuclear} {Society}},
	publisher = {American Nuclear Society},
	author = {Skirpan, Zachary and Betzler, Benjamin and Powers, Jeffery and Blair, Stuart},
	month = nov,
	year = {2017},
	file = {trans_v117_n1_pp1366-1368.pdf:/home/andrei2/Zotero/storage/U9KIFUF2/trans_v117_n1_pp1366-1368.pdf:application/pdf}
}

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

@techreport{briggs_molten-salt_1966,
	title = {Molten-salt reactor program. {Semiannual} progress report},
	language = {English},
	number = {ORNL-3936},
	urldate = {2016-09-07},
	institution = {Oak Ridge National Lab., Tenn.},
	author = {Briggs, R. B.},
	month = jun,
	year = {1966},
	keywords = {air, coolant loops, decontamination, detection, fused salt fuel, gases, leaks, liquids, msre, planning, pressure vessels, reactor technology, reactors, research and test reactors, research reactors, sampling, shielding, waste disposal, water coolant},
	file = {Full Text PDF:/home/andrei2/Zotero/storage/VX3DIUNH/Robertson - 1965 - Msre Design and Operations Report. Part I. Descrip.pdf:application/pdf;ORNL-3936.pdf:/home/andrei2/Zotero/storage/3J2SRDCQ/ORNL-3936.pdf:application/pdf;ORNL-3936.pdf:/home/andrei2/Zotero/storage/874UTQ39/ORNL-3936.pdf:application/pdf;Snapshot:/home/andrei2/Zotero/storage/D98K5BFA/4654707.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},
	url = {http://www.sciencedirect.com/science/article/pii/S030645491300474X},
	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},
	urldate = {2017-10-04},
	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:/home/andrei2/Zotero/storage/7CEI4FC7/Ignatiev et al. - 2014 - Molten salt actinide recycler and transforming sys.pdf:application/pdf}
}

@article{kelly_removal_1965,
	title = {{REMOVAL} {OF} {RARE}-{EARTH} {FISSION} {PRODUCTS} {FROM} {MOLTEN}-{SALT} {REACTOR} {FUELS} {BY} {DISTILLATION}},
	volume = {Vol: 8},
	url = {https://www.osti.gov/biblio/4620744-removal-rare-earth-fission-products-from-molten-salt-reactor-fuels-distillation},
	abstract = {The U.S. Department of Energy's Office of Scientific and Technical Information},
	language = {English},
	urldate = {2018-02-13},
	journal = {Transactions of the American Nuclear Society (U.S.)},
	author = {Kelly, M. J.},
	month = may,
	year = {1965},
	file = {Snapshot:/home/andrei2/Zotero/storage/XF83GRR8/4620744-removal-rare-earth-fission-products-from-molten-salt-reactor-fuels-distillation.html:text/html}
}

@article{cathers_uranium_1957,
	title = {Uranium {Recovery} for {Spent} {Fuel} by {Dissolution} in {Fused} {Salt} and {Fluorination}},
	volume = {2},
	issn = {0029-5639},
	url = {https://doi.org/10.13182/NSE57-A35491},
	doi = {10.13182/NSE57-A35491},
	abstract = {A promising nonaqueous process for the recovery of uranium from spent fuel elements is under development at Oak Ridge National Laboratory. This process consists of dissolution of the fuel element in a fluoride melt by hydrofluorination at 600 to 700°C, direct fluorination with fluorine for the production and volatilization of UF6, with further decontamination of the product UF6 from fission product activity being secured in a NaF absorption-desorption step. Good decontamination is obtained in the fluorination step due to the low volatility of most of the fission product fluorides. An over-all decontamination factor greater than 106 with adequate uranium recovery has been demonstrated in laboratory scale tests using a double bed procedure for the NaF step. A pilot plant has been constructed for testing the process with various heterogeneous fuel elements. The engineering and operational features of the pilot plant are described.},
	number = {6},
	urldate = {2018-02-13},
	journal = {Nuclear Science and Engineering},
	author = {Cathers, G. I.},
	month = nov,
	year = {1957},
	pages = {768--777},
	file = {Full Text PDF:/home/andrei2/Zotero/storage/EW8RTSJL/Cathers - 1957 - Uranium Recovery for Spent Fuel by Dissolution in .pdf:application/pdf;Snapshot:/home/andrei2/Zotero/storage/N9DFLIWS/NSE57-A35491.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:/home/andrei2/Zotero/storage/7UGTNXI5/NSE_ARE_Operation.pdf:application/pdf}
}

@article{dagum_openmp:_1998,
	title = {{OpenMP}: an industry standard {API} for shared-memory programming},
	volume = {5},
	issn = {1070-9924},
	shorttitle = {{OpenMP}},
	doi = {10.1109/99.660313},
	abstract = {At its most elemental level, OpenMP is a set of compiler directives and callable runtime library routines that extend Fortran (and separately, C and C++ to express shared memory parallelism. It leaves the base language unspecified, and vendors can implement OpenMP in any Fortran compiler. Naturally, to support pointers and allocatables, Fortran 90 and Fortran 95 require the OpenMP implementation to include additional semantics over Fortran 77. OpenMP leverages many of the X3H5 concepts while extending them to support coarse grain parallelism. The standard also includes a callable runtime library with accompanying environment variables},
	number = {1},
	journal = {IEEE Computational Science and Engineering},
	author = {Dagum, L. and Menon, R.},
	month = jan,
	year = {1998},
	keywords = {allocatables, ANSI standards, application program interfaces, callable runtime library, callable runtime library routines, coarse grain parallelism, Coherence, compiler directives, Computer architecture, environment variables, Fortran 90, Fortran 95, Fortran compiler, Hardware, industry standard API, Message passing, OpenMP, Parallel processing, parallel programming, pointers, Power system modeling, Scalability, shared memory parallelism, shared memory programming, shared memory systems, software portability, software reviews, software standards, Software systems, X3H5 concepts},
	pages = {46--55},
	file = {IEEE Xplore Abstract Record:/home/andrei2/Zotero/storage/XBRYMUEH/660313.html:text/html;IEEE Xplore Full Text PDF:/home/andrei2/Zotero/storage/HJPAIV6K/Dagum and Menon - 1998 - OpenMP an industry standard API for shared-memory.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},
	url = {http://www.sciencedirect.com/science/article/pii/S0022113908002054},
	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.},
	number = {1},
	urldate = {2018-02-09},
	journal = {Journal of Fluorine Chemistry},
	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 = {Electrochemistry, Reactor physic, Thorium fuel cycle},
	pages = {11--17},
	file = {ScienceDirect Full Text PDF:/home/andrei2/Zotero/storage/A7CI75JN/Delpech et al. - 2009 - Reactor physic and reprocessing scheme for innovat.pdf:application/pdf;ScienceDirect Snapshot:/home/andrei2/Zotero/storage/Z32IFZE3/S0022113908002054.html:text/html}
}

@inproceedings{powers_reactor_2013,
	title = {Reactor physics analysis of thorium fuel cycles using molten salt reactors},
	volume = {109},
	language = {en},
	urldate = {2018-02-05},
	booktitle = {Transactions of the {American} {Nuclear} {Society}},
	publisher = {ANS},
	author = {Powers, J. J. and Worrall, A. and Gehin, J. C. and Harrison, T. J. and Sunny, E.E.},
	year = {2013},
	pages = {1457},
	file = {Snapshot:/home/andrei2/Zotero/storage/CPB98KUV/287163391_Reactor_physics_analysis_of_thorium_fuel_cycles_using_molten_salt_reactors.html:text/html;trans_v109_n1_pp1457-1460.pdf:/home/andrei2/Zotero/storage/JZ6JHT46/trans_v109_n1_pp1457-1460.pdf:application/pdf}
}

@techreport{anon_plutonium_1989,
	title = {Plutonium fuel an assessment. {Report} by an expert group},
	url = {http://inis.iaea.org/Search/search.aspx?orig_q=RN:21076679},
	language = {en},
	number = {INIS-XN--254},
	urldate = {2018-02-02},
	institution = {Nuclear Energy Agency},
	author = {Anon},
	year = {1989},
	file = {Full Text PDF:/home/andrei2/Zotero/storage/BBIEDPVK/Anon - 1989 - Plutonium fuel an assessment. Report by an expert .pdf:application/pdf;Full Text PDF:/home/andrei2/Zotero/storage/FVSMCJI9/Anon - 1989 - Plutonium fuel an assessment. Report by an expert .pdf:application/pdf;Snapshot:/home/andrei2/Zotero/storage/NB2V9N57/search.html:text/html;Snapshot:/home/andrei2/Zotero/storage/DVWNMNWK/search.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–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},
	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 = {atws, molten salt reactor, MSR, Multi-physics modelling, Reactor dynamics, read, Thermo-hydrodynamics, unread},
	pages = {1356--1372},
	file = {A_multi-physics_modelling_approach_to_the_dynamics_of_Molten_Sal.mobi:/home/andrei2/Zotero/storage/JWIMI3QI/A_multi-physics_modelling_approach_to_the_dynamics_of_Molten_Sal.mobi:application/octet-stream;cammi_multi-physics_2011.pdf:/home/andrei2/Zotero/storage/AHXUPQ4A/cammi_multi-physics_2011.pdf:application/pdf;ScienceDirect Full Text PDF:/home/andrei2/Zotero/storage/6FQKN2CJ/Cammi et al. - 2011 - A multi-physics modelling approach to the dynamics.pdf:application/pdf;ScienceDirect Snapshot:/home/andrei2/Zotero/storage/63AWQD57/S0306454911000582.html:text/html}
}

@article{leblanc_molten_2010,
	title = {Molten salt reactors: {A} new beginning for an old idea},
	volume = {240},
	issn = {0029-5493},
	shorttitle = {Molten salt reactors},
	url = {http://www.sciencedirect.com/science/article/pii/S0029549310000191},
	doi = {10.1016/j.nucengdes.2009.12.033},
	abstract = {Molten salt reactors have seen a marked resurgence of interest over the past decade, highlighted by their inclusion as one of six Generation IV reactor types. The most active development period however was between the mid 1950s and early 1970s at Oak Ridge National Laboratories (ORNL) and any new re-examination of this concept must bear in mind the far different priorities then in place. High breeding ratios and short doubling times were paramount and this guided the evolution of the Molten Salt Breeder Reactor (MSBR) program. As the inherent advantages of the molten salt concept have become apparent to an increasing number of researchers worldwide it is important to not simply look to continue where ORNL left off but to return to basics in order to offer the best design using updated goals and abilities. A major potential change to the traditional Single Fluid, MSBR design and a subject of this presentation is a return to the mode of operation that ORNL proposed for the majority of its MSR program. That being the Two Fluid design in which separate salts are used for fissile 233UF4 and fertile ThF4. Oak Ridge abandoned this promising route due to what was known as the “plumbing problem”. It will be shown that a simple yet crucial modification to core geometry can solve this problem and enable the many advantages of the Two Fluid design. In addition, another very promising route laid out by ORNL was simplified Single Fluid converter reactors that could obtain far superior lifetime uranium utilization than LWR or CANDU without the need for any fuel processing beyond simple chemistry control. Updates and potential improvements to this very attractive concept will also be explored.},
	number = {6},
	urldate = {2013-05-28},
	journal = {Nuclear Engineering and Design},
	author = {LeBlanc, David},
	month = jun,
	year = {2010},
	keywords = {MSR, read},
	pages = {1644--1656},
	file = {newbeginning.pdf:/home/andrei2/Zotero/storage/6NNAEPAG/newbeginning.pdf:application/pdf;newbeginning.pdf:/home/andrei2/Zotero/storage/CF65ST62/newbeginning.pdf:application/pdf;ScienceDirect Full Text PDF:/home/andrei2/Zotero/storage/Q94W579M/LeBlanc - 2010 - Molten salt reactors A new beginning for an old i.pdf:application/pdf;ScienceDirect Full Text PDF:/home/andrei2/Zotero/storage/44BSHSZU/LeBlanc - 2010 - Molten salt reactors A new beginning for an old i.pdf:application/pdf;ScienceDirect Full Text PDF:/home/andrei2/Zotero/storage/48BFI2SE/LeBlanc - 2010 - Molten salt reactors A new beginning for an old i.pdf:application/pdf;ScienceDirect Snapshot:/home/andrei2/Zotero/storage/U572AVDM/S0029549310000191.html:text/html;ScienceDirect Snapshot:/home/andrei2/Zotero/storage/TAMCTNDG/S0029549310000191.html:text/html;ScienceDirect Snapshot:/home/andrei2/Zotero/storage/GJQN7FJR/S0029549310000191.html:text/html}
}

@article{gehin_liquid_2016,
	title = {Liquid {Fuel} {Molten} {Salt} {Reactors} for {Thorium} {Utilization}},
	volume = {194},
	issn = {00295450},
	url = {https://doi.org/10.13182/NT15-124},
	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},
	urldate = {2016-05-06},
	journal = {Nuclear Technology},
	author = {Gehin, Jess C. and Powers, Jeffery J.},
	month = may,
	year = {2016},
	pages = {152--161},
	file = {Full Text PDF:/home/andrei2/Zotero/storage/U5KHRRFR/Gehin and Powers - 2016 - Liquid Fuel Molten Salt Reactors for Thorium Utili.pdf:application/pdf;Fulltext:/home/andrei2/Zotero/storage/LT8KVIIL/scholar.html:text/html;Liquid Fuel Molten Salt Reactors for Thorium Utilization.pdf:/home/andrei2/Zotero/storage/NG377NQN/Liquid Fuel Molten Salt Reactors for Thorium Utilization.pdf:application/pdf;Snapshot:/home/andrei2/Zotero/storage/N6N3FSCJ/NT15-124.html:text/html;Snapshot:/home/andrei2/Zotero/storage/SKGAV2GK/NT15-124.html:text/html}
}

@article{macpherson_molten_1985,
	title = {The {Molten} {Salt} {Reactor} {Adventure}},
	volume = {90},
	issn = {ISSN 0029-5639},
	url = {http://moltensalt.org/references/static/downloads/pdf/MSadventure.pdf},
	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},
	urldate = {2013-09-06},
	journal = {Nuclear Science and Engineering},
	author = {MacPherson, H. G.},
	month = aug,
	year = {1985},
	keywords = {unread},
	pages = {374--380},
	file = {[PDF] from moltensalt.org:/home/andrei2/Zotero/storage/PSZWM3J8/MacPherson - 1985 - The Molten Salt Reactor Adventure.pdf:application/pdf;nse_v90_n4_pp374-380.pdf:/home/andrei2/Zotero/storage/RGPUF2H9/nse_v90_n4_pp374-380.pdf:application/pdf;Snapshot:/home/andrei2/Zotero/storage/7EEL6Q93/search.html:text/html;Snapshot:/home/andrei2/Zotero/storage/NAUGQK6J/NSE90-374.html:text/html}
}

@article{forsberg_molten-salt-cooled_2003,
	title = {Molten-{Salt}-{Cooled} {Advanced} {High}-{Temperature} {Reactor} for {Production} of {Hydrogen} and {Electricity}},
	volume = {144},
	issn = {0029-5450},
	url = {https://doi.org/10.13182/NT03-1},
	doi = {10.13182/NT03-1},
	abstract = {The molten-salt-cooled Advanced High-Temperature Reactor (AHTR) is a new reactor concept designed to provide very high-temperature (750 to 1000°C) heat to enable efficient low-cost thermochemical production of hydrogen (H2) or production of electricity. This paper provides an initial description and technical analysis of its key features. The proposed AHTR uses coated-particle graphite-matrix fuel similar to that used in high-temperature gas-cooled reactors (HTGRs), such as the General Atomics gas turbine-modular helium reactor. However, unlike the HTGRs, the AHTR uses a molten-salt coolant and a pool configuration, similar to that of the General Electric Super Power Reactor Inherently Safe Module (S-PRISM) liquid-metal reactor. Because the boiling points for molten fluoride salts are near 1400°C, the reactor can operate at very high temperatures and atmospheric pressure. For thermochemical H2 production, the heat is delivered at the required near-constant high temperature and low pressure. For electricity production, a multireheat helium Brayton (gas-turbine) cycle, with efficiencies {\textgreater}50\%, is used. The low-pressure molten-salt coolant, with its high heat capacity and natural circulation heat transfer capability, creates the potential for robust safety (including fully passive decay-heat removal) and improved economics with passive safety systems that allow higher power densities and scaling to large reactor sizes [{\textgreater}1000 MW(electric)].},
	number = {3},
	journal = {Nuclear Technology},
	author = {Forsberg, Charles W. and Peterson, Per F. and Pickard, Paul S.},
	month = dec,
	year = {2003},
	keywords = {High-Temperature Reactors (HTR), Hydrogen Production, Molten-Salt, read},
	pages = {289--302},
	file = {Full Text PDF:/home/andrei2/Zotero/storage/IT6HVA59/Forsberg et al. - 2003 - Molten-Salt-Cooled Advanced High-Temperature React.pdf:application/pdf;Full Text PDF:/home/andrei2/Zotero/storage/BMPBFFW3/Forsberg et al. - 2003 - Molten-Salt-Cooled Advanced High-Temperature React.pdf:application/pdf;NT-144-3-289-Forsberg.pdf:/home/andrei2/Zotero/storage/EU4E9BA2/NT-144-3-289-Forsberg.pdf:application/pdf;Snapshot:/home/andrei2/Zotero/storage/7Y7JLWDI/NT03-1.html:text/html;Snapshot:/home/andrei2/Zotero/storage/MZN3C8AE/NT03-1.html:text/html}
}

@techreport{ruggieri_eranos_2006,
	title = {{ERANOS} 2.1: international code system for {GEN} {IV} fast reactor analysis},
	institution = {American Nuclear Society, 555 North Kensington Avenue, La Grange Park, IL 60526 (United States)},
	author = {Ruggieri, JM and Tommasi, J and Lebrat, JF and Suteau, C and Plisson-Rieunier, D and De Saint Jean, C and Rimpault, G and Sublet, JC},
	year = {2006}
}

@mastersthesis{pettersen_coupled_2016,
	title = {Coupled multi-physics simulations of the {Molten} {Salt} {Fast} {Reactor} using coarse-mesh thermal-hydraulics and spatial neutronics},
	url = {http://samofar.eu/wp-content/uploads/2016/11/MScThesis-eirikEidePettersen.pdf},
	urldate = {2016-11-29},
	school = {MSc thesis, September 2016 (PDF)},
	author = {Pettersen, Eirik Eide and Mikityuk, Konstantin},
	year = {2016},
	file = {[PDF] samofar.eu:/home/andrei2/Zotero/storage/XVIH74PK/Pettersen and Mikityuk - 2016 - Coupled multi-physics simulations of the Molten Sa.pdf:application/pdf;Fulltext:/home/andrei2/Zotero/storage/PA885TB5/Pettersen and Mikityuk - 2016 - Coupled multi-physics simulations of the Molten Sa.pdf:application/pdf;Snapshot:/home/andrei2/Zotero/storage/S7SERYFF/Pettersen and Mikityuk - 2016 - Coupled multi-physics simulations of the Molten Sa.pdf:application/pdf}
}

@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},
	url = {http://www.sciencedirect.com/science/article/pii/S0022311513008507},
	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},
	urldate = {2017-01-04},
	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:/home/andrei2/Zotero/storage/GACISQTN/Aufiero et al. - 2013 - An extended version of the SERPENT-2 code to inves.pdf:application/pdf;ScienceDirect Full Text PDF:/home/andrei2/Zotero/storage/FK5H57EA/Aufiero et al. - 2013 - An extended version of the SERPENT-2 code to inves.pdf:application/pdf;ScienceDirect Snapshot:/home/andrei2/Zotero/storage/GV4RYF2E/S0022311513008507.html:text/html;ScienceDirect Snapshot:/home/andrei2/Zotero/storage/VTCQESRU/S0022311513008507.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},
	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 “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},
	urldate = {2017-03-18},
	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, Thermal hydraulics, Transient calculation, Transient Fission Matrix},
	pages = {112--124},
	file = {ScienceDirect Full Text PDF:/home/andrei2/Zotero/storage/K5JUZTZ6/Laureau et al. - 2017 - Transient coupled calculations of the Molten Salt .pdf:application/pdf;ScienceDirect Full Text PDF:/home/andrei2/Zotero/storage/GEYY7UJ9/Laureau et al. - 2017 - Transient coupled calculations of the Molten Salt .pdf:application/pdf;ScienceDirect Snapshot:/home/andrei2/Zotero/storage/IKKCKIDB/S002954931730081X.html:text/html;ScienceDirect Snapshot:/home/andrei2/Zotero/storage/5B25JVQI/S002954931730081X.html:text/html}
}

@inproceedings{scopatz_pyne:_2012,
	address = {San Diego, CA, USA},
	title = {{PyNE}: {Python} for {Nuclear} {Engineering}},
	volume = {107},
	abstract = {PyNE, or 'Python for Nuclear Engineering' 1 , is a nascent
free and open source C++/Cython/Python package for perform-
ing common nuclear engineering tasks. This is intended as a
base level tool kit - akin to SciPy or Biopython - for common
algorithms in the nuclear science and engineering domain.
The remainer of this paper is composed of a discussion of the
difficulties which prevented PyNE from being written earlier,
a listing of the first cut capabilities, and a description of why
PyNE has thus far been successful and what future features are
currently planned.},
	booktitle = {Transactions of the {American} {Nuclear} {Society}},
	publisher = {American Nuclear Society},
	author = {Scopatz, Anthony and Romano, Paul K. and Wilson, Paul P. H. and Huff, Kathryn D.},
	month = nov,
	year = {2012},
	file = {trans_v107_n1_pp985-987.pdf:/home/andrei2/Zotero/storage/TQ9E3XGC/trans_v107_n1_pp985-987.pdf:application/pdf}
}

@article{doligez_coupled_2014,
	title = {Coupled study of the {Molten} {Salt} {Fast} {Reactor} core physics and its associated reprocessing unit},
	volume = {64},
	issn = {0306-4549},
	url = {http://www.sciencedirect.com/science/article/pii/S0306454913004799},
	doi = {10.1016/j.anucene.2013.09.009},
	abstract = {Molten Salt Reactors (MSRs) are liquid-fuel reactors, in which the fuel is also the coolant and flows through the core. A particular configuration presented in this paper called the Molten Salt Fast Reactor consists in a Molten Salt Reactor with no moderator inside the core and a salt composition that leads to a fast neutron spectrum. Previous studies showed that this concept (previously called Thorium Molten Salt Reactor – Nonmoderated) has very promising characteristics. The liquid fuel implies a special reprocessing. Each day a small amount of the fuel salt is extracted from the core for on-site reprocessing. To study such a reactor, the materials evolution within the core has to be coupled to the reprocessing unit, since the latter cleans the salt quasi continuously and feeds the reactor. This paper details the issues associated to the numerical coupling of the core and the reprocessing. It presents how the chemistry is introduced inside the classical Bateman equation (evolution of nuclei within a neutron flux) in order to carry a numerical coupled study. To achieve this goal, the chemistry has to be modeled numerically and integrated to the equations of evolution. This paper presents how is it possible to describe the whole concept (reactor+reprocessing unit) by a system of equations that can be numerically solved. Our program is a connection between MCNP and a homemade evolution code called REM. Thanks to this tool; constraints on the fuel reprocessing were identified. Limits are specified to preserve the good neutronics properties of the MSFR. In this paper, we show that the limit rate for the reprocessing is 2.5l of fuel salt a day, which means that the fuel should be reprocessed within 7000days approximately if there is a specific way to control the redox potential of the salt. Finally, a last part of this paper analyzes the impact of chemical parameter uncertainties on the reprocessing performance.},
	number = {Supplement C},
	urldate = {2017-10-22},
	journal = {Annals of Nuclear Energy},
	author = {Doligez, X. and Heuer, D. and Merle-Lucotte, E. and Allibert, M. and Ghetta, V.},
	month = feb,
	year = {2014},
	keywords = {Molten salt, Neutronics, Reprocessing influence, thorium cycle},
	pages = {430--440},
	file = {ScienceDirect Full Text PDF:/home/andrei2/Zotero/storage/9JYPVX26/Doligez et al. - 2014 - Coupled study of the Molten Salt Fast Reactor core.pdf:application/pdf;ScienceDirect Full Text PDF:/home/andrei2/Zotero/storage/XLLWXDGR/Doligez et al. - 2014 - Coupled study of the Molten Salt Fast Reactor core.pdf:application/pdf;ScienceDirect Snapshot:/home/andrei2/Zotero/storage/FUF9ASZ2/S0306454913004799.html:text/html;ScienceDirect Snapshot:/home/andrei2/Zotero/storage/Y6RKHT6D/S0306454913004799.html:text/html}
}

@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/andrei2/Zotero/storage/5PX5BA6R/Weller et al. - 1998 - A tensorial approach to computational continuum me.pdf:application/pdf;Snapshot:/home/andrei2/Zotero/storage/JIM2283D/1.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.},
	journal = {Annals of Nuclear Energy},
	author = {Ridley, Gavin and Chvala, Ondrej},
	month = dec,
	year = {2017},
	keywords = {Chemistry control, Depletion, DMSR, On-line refueling},
	pages = {265--281},
	file = {1-s2.0-S030645491730021X-main.pdf:/home/andrei2/Zotero/storage/X6755AXC/1-s2.0-S030645491730021X-main.pdf:application/pdf;ScienceDirect Full Text PDF:/home/andrei2/Zotero/storage/XTWJM8BV/Ridley and Chvala - 2017 - A method for predicting fuel maintenance in once-t.pdf:application/pdf;ScienceDirect Snapshot:/home/andrei2/Zotero/storage/SISVF65Z/S030645491730021X.html:text/html;ScienceDirect Snapshot:/home/andrei2/Zotero/storage/UJCBPM9Z/S030645491730021X.html:text/html}
}

@article{jeong_equilibrium_2016,
	title = {Equilibrium core design methods for molten salt breeder reactor based on two-cell model},
	volume = {53},
	issn = {0022-3131, 1881-1248},
	url = {http://www.tandfonline.com/doi/full/10.1080/00223131.2015.1062812},
	doi = {10.1080/00223131.2015.1062812},
	language = {en},
	number = {4},
	urldate = {2017-04-06},
	journal = {Journal of Nuclear Science and Technology},
	author = {Jeong, Yongjin and Park, Jinsu and Lee, Hyun Chul and Lee, Deokjung},
	month = apr,
	year = {2016},
	pages = {529--536},
	file = {Equilibrium core design methods for molten salt breeder reactor based on two cell model.pdf:/home/andrei2/Zotero/storage/4IRQCB82/Equilibrium core design methods for molten salt breeder reactor based on two cell model.pdf:application/pdf}
}

@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 = {Leppanen, Jaakko and Pusa, Maria and Viitanen, Tuomas and Valtavirta, Ville and Kaltiaisenaho, Toni},
	month = aug,
	year = {2015},
	keywords = {Burnup calculation, Homogenization, Monte Carlo, Reactor physics},
	pages = {142--150},
	file = {ScienceDirect Full Text PDF:/home/andrei2/Zotero/storage/VT4QRTXR/Leppänen et al. - 2015 - The Serpent Monte Carlo code Status, development .pdf:application/pdf;ScienceDirect Full Text PDF:/home/andrei2/Zotero/storage/83V2KXJ9/Leppänen et al. - 2015 - The Serpent Monte Carlo code Status, development .pdf:application/pdf;ScienceDirect Snapshot:/home/andrei2/Zotero/storage/9AVEULJX/S0306454914004095.html:text/html;ScienceDirect Snapshot:/home/andrei2/Zotero/storage/AQAB5875/S0306454914004095.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},
	abstract = {Current interest in advanced nuclear energy and molten salt reactor (MSR) concepts has enhanced interest in building the tools necessary to analyze these systems. A Python script known as ChemTriton has been developed to simulate equilibrium MSR fuel cycle performance by modeling the changing isotopic composition of an irradiated fuel salt using SCALE for neutron transport and depletion calculations. Improved capabilities in ChemTriton include a generic geometry capable of modeling multi-zone and multi-fluid systems, enhanced time-dependent feed and separations, and a critical concentration search. Although more generally applicable, the capabilities developed to date are illustrated in this paper in three applied problems: (1) simulating the startup of a thorium-based MSR fuel cycle (a likely scenario requires the first of these MSRs to be started without available 233U); (2) determining the effect of the removal of different fission products on MSR operations; and (3) obtaining the equilibrium concentration of a mixed-oxide light-water reactor fuel in a two-stage fuel cycle with a sodium fast reactor. The third problem is chosen to demonstrate versatility in an application to analyze the fuel cycle of a non-MSR system. In the first application, the initial fuel salt compositions fueled with different sources of fissile material are made feasible after (1) removing the associated nonfissile actinides after much of the initial fissile isotopes have burned and (2) optimizing the thorium concentration to maintain a critical configuration without significantly reducing breeding capability. In the second application, noble metal, volatile gas, and rare earth element fission products are shown to have a strong negative effect on criticality in a uranium-fueled thermal-spectrum MSR; their removal significantly increases core lifetime (by 30\%) and fuel utilization. In the third application, the fuel of a mixed-oxide light-water reactor approaches an equilibrium composition after 20 depletion steps, demonstrating the potential for the longer time scales required to achieve equilibrium for solid-fueled systems over liquid fuel systems. This time to equilibrium can be reduced by starting with an initial fuel composition closer to that of the equilibrium fuel, reducing the need to handle time-dependent fuel compositions.},
	language = {en},
	number = {Supplement C},
	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, Fuel cycle, salt separations, salt treatment},
	pages = {489--503},
	file = {ScienceDirect Full Text PDF:/home/andrei2/Zotero/storage/DICQR39F/Betzler et al. - 2017 - Molten salt reactor neutronics and fuel cycle mode.pdf:application/pdf;ScienceDirect Full Text PDF:/home/andrei2/Zotero/storage/57FKUK6R/Betzler et al. - 2017 - Molten salt reactor neutronics and fuel cycle mode.pdf:application/pdf;ScienceDirect Snapshot:/home/andrei2/Zotero/storage/HWDLR9ZN/S0306454916309185.html:text/html;ScienceDirect Snapshot:/home/andrei2/Zotero/storage/6ZXURJ4E/S0306454916309185.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–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 = {Molten Salt Fast Reactor (MSFR), Molten Salt Reactor (MSR), Multiphysics, OpenFOAM, Reactor dynamics},
	pages = {390--401},
	file = {ScienceDirect Full Text PDF:/home/andrei2/Zotero/storage/GFZ6CW4E/Aufiero et al. - 2014 - Development of an OpenFOAM model for the Molten Sa.pdf:application/pdf;ScienceDirect Snapshot:/home/andrei2/Zotero/storage/TGWQHMHK/S0009250914001146.html:text/html;ScienceDirect Snapshot:/home/andrei2/Zotero/storage/L4BZULIM/S0009250914001146.html:text/html}
}

@article{cui_transition_2017,
	title = {Transition toward thorium fuel cycle in a molten salt reactor by using plutonium},
	volume = {28},
	issn = {1001-8042, 2210-3147},
	url = {https://link.springer.com/article/10.1007/s41365-017-0303-y},
	doi = {10.1007/s41365-017-0303-y},
	abstract = {The molten salt reactor (MSR), as one of the Generation IV advanced nuclear systems, has attracted a worldwide interest due to its excellent performances in safety, economics, sustainability, and proliferation resistance. The aim of this work is to provide and evaluate possible solutions to fissile 233U233U{\textasciicircum}\{233\} {\textbackslash}hbox \{U\} production and further the fuel transition to thorium fuel cycle in a thermal MSR by using plutonium partitioned from light water reactors spent fuel. By using an in-house developed tool, a breeding and burning (B\&B) scenario is first introduced and analyzed from the aspects of the evolution of main nuclides, net 233U233U{\textasciicircum}\{233\} {\textbackslash}hbox \{U\} production, spectrum shift, and temperature feedback coefficient. It can be concluded that such a Th/Pu to Th/233UTh/233U{\textbackslash}hbox \{Th\}/{\textasciicircum}\{233\} {\textbackslash}hbox \{U\} transition can be accomplished by employing a relatively fast fuel reprocessing with a cycle time less than 60 days. At the equilibrium state, the reactor can achieve a conversion ratio of about 0.996 for the 60-day reprocessing period (RP) case and about 1.047 for the 10-day RP case. The results also show that it is difficult to accomplish such a fuel transition with limited reprocessing (RP is 180 days), and the reactor operates as a converter and burns the plutonium with the help of thorium. Meanwhile, a pre-breeding and burning (PB\&B) scenario is also analyzed briefly with respect to the net 233U233U{\textasciicircum}\{233\} {\textbackslash}hbox \{U\} production and evolution of main nuclides. One can find that it is more efficient to produce 233U233U{\textasciicircum}\{233\} {\textbackslash}hbox \{U\} under this scenario, resulting in a double time varying from about 1.96 years for the 10-day RP case to about 6.15 years for the 180-day RP case.},
	language = {en},
	number = {10},
	urldate = {2017-09-18},
	journal = {Nuclear Science and Techniques},
	author = {Cui, De-Yang and Xia, Shao-Peng and Li, Xiao-Xiao and Cai, Xiang-Zhou and Chen, Jin-Gen},
	month = oct,
	year = {2017},
	pages = {152},
	file = {Full Text PDF:/home/andrei2/Zotero/storage/CSJN358R/Cui et al. - 2017 - Transition toward thorium fuel cycle in a molten s.pdf:application/pdf;Snapshot:/home/andrei2/Zotero/storage/HJUGJ2AV/s41365-017-0303-y.html:text/html;Snapshot:/home/andrei2/Zotero/storage/DJZ6B5G8/s41365-017-0303-y.html:text/html}
}

@article{heuer_towards_2014,
	title = {Towards the thorium fuel cycle with molten salt fast reactors},
	volume = {64},
	issn = {0306-4549},
	url = {http://www.sciencedirect.com/science/article/pii/S0306454913004106},
	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.},
	urldate = {2017-09-13},
	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 = {Deployment scenario, Generation 4 reactors, Incinerator, MSFR, Thorium fuel cycle},
	pages = {421--429},
	file = {ScienceDirect Full Text PDF:/home/andrei2/Zotero/storage/RNF6R5F9/Heuer et al. - 2014 - Towards the thorium fuel cycle with molten salt fa.pdf:application/pdf;ScienceDirect Full Text PDF:/home/andrei2/Zotero/storage/8ETDPPSM/Heuer et al. - 2014 - Towards the thorium fuel cycle with molten salt fa.pdf:application/pdf;ScienceDirect Snapshot:/home/andrei2/Zotero/storage/TFXWLMD3/S0306454913004106.html:text/html;ScienceDirect Snapshot:/home/andrei2/Zotero/storage/2ILQRZ53/S0306454913004106.html:text/html}
}

@article{lung_perspectives_1998,
	title = {Perspectives of the thorium fuel cycle},
	volume = {180},
	issn = {0029-5493},
	url = {http://www.sciencedirect.com/science/article/pii/S0029549397002963},
	doi = {10.1016/S0029-5493(97)00296-3},
	abstract = {In the 1960s and 70s, great interest developed in the thorium fuel cycle as a supplement to limited uranium reserves. A great amount of work was carried out and many interesting developments resulted, among which were prototype high temperature reactors. It was shown that thorium could be used practically in any type of existing reactor. These initiatives have been virtually brought to a halt for various reasons, except in India, which has continued with its thorium activities. Recently however, new considerations have revived interest in thorium. An overview of the specificities of the thorium fuel cycle was carried out and the most interesting projects launched in the 1960s are recalled. The new concepts are analyzed briefly and some lines of thought for the future are proposed.},
	number = {2},
	urldate = {2017-08-29},
	journal = {Nuclear Engineering and Design},
	author = {Lung, Michel and Gremm, Otto},
	month = mar,
	year = {1998},
	pages = {133--146},
	file = {ScienceDirect Full Text PDF:/home/andrei2/Zotero/storage/Y9HK4BPH/Lung and Gremm - 1998 - Perspectives of the thorium fuel cycle.pdf:application/pdf;ScienceDirect Full Text PDF:/home/andrei2/Zotero/storage/EACJHK7A/Lung and Gremm - 1998 - Perspectives of the thorium fuel cycle.pdf:application/pdf;ScienceDirect Snapshot:/home/andrei2/Zotero/storage/5QSP4FFV/S0029549397002963.html:text/html;ScienceDirect Snapshot:/home/andrei2/Zotero/storage/XXZUSNYH/S0029549397002963.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}
}

@incollection{jorgensen_19_2017,
	title = {19 - {ThorCon} reactor},
	isbn = {978-0-08-101126-3},
	url = {https://www.sciencedirect.com/science/article/pii/B9780081011263000191},
	abstract = {ThorCon is a molten salt reactor design targeting the developing world. It uses existing technology from MSRE to allow rapid development. It uses shipbuilding techniques to allow rapid deployment. The result is a capital cost of \$1.2B/1GWe, operating costs.},
	urldate = {2017-11-21},
	booktitle = {Molten {Salt} {Reactors} and {Thorium} {Energy}},
	publisher = {Woodhead Publishing},
	author = {Jorgensen, Lars},
	editor = {Dolan, Thomas J.},
	year = {2017},
	doi = {10.1016/B978-0-08-101126-3.00019-1},
	keywords = {Molten salt reactor, shipbuilding, thorium},
	pages = {557--564},
	file = {ScienceDirect Full Text PDF:/home/andrei2/Zotero/storage/8PPAYZRV/Jorgensen - 2017 - 19 - ThorCon reactor.pdf:application/pdf;ScienceDirect Full Text PDF:/home/andrei2/Zotero/storage/ES4PSRJZ/Jorgensen - 2017 - 19 - ThorCon reactor.pdf:application/pdf;ScienceDirect Snapshot:/home/andrei2/Zotero/storage/XAW388RZ/B9780081011263000191.html:text/html;ScienceDirect Snapshot:/home/andrei2/Zotero/storage/SQ9MYSGC/B9780081011263000191.html:text/html}
}

@incollection{dai_17_2017,
	title = {17 - {Thorium} molten salt reactor nuclear energy system ({TMSR})},
	isbn = {978-0-08-101126-3},
	url = {https://www.sciencedirect.com/science/article/pii/B9780081011263000178},
	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).},
	urldate = {2017-11-21},
	booktitle = {Molten {Salt} {Reactors} and {Thorium} {Energy}},
	publisher = {Woodhead Publishing},
	author = {Dai, Zhimin},
	editor = {Dolan, Thomas J.},
	year = {2017},
	doi = {10.1016/B978-0-08-101126-3.00017-8},
	keywords = {hybrid nuclear energy application, Molten salt reactor, thorium},
	pages = {531--540},
	file = {ScienceDirect Full Text PDF:/home/andrei2/Zotero/storage/QGM7UCAA/Dai - 2017 - 17 - Thorium molten salt reactor nuclear energy sy.pdf:application/pdf;ScienceDirect Full Text PDF:/home/andrei2/Zotero/storage/IUVWTXVJ/Dai - 2017 - 17 - Thorium molten salt reactor nuclear energy sy.pdf:application/pdf;ScienceDirect Snapshot:/home/andrei2/Zotero/storage/4JZSFD3K/B9780081011263000178.html:text/html;ScienceDirect Snapshot:/home/andrei2/Zotero/storage/SUYR248M/B9780081011263000178.html:text/html}
}

@incollection{leblanc_18_2017,
	title = {18 - {Integral} molten salt reactor},
	isbn = {978-0-08-101126-3},
	url = {https://www.sciencedirect.com/science/article/pii/B978008101126300018X},
	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.},
	urldate = {2017-11-21},
	booktitle = {Molten {Salt} {Reactors} and {Thorium} {Energy}},
	publisher = {Woodhead Publishing},
	author = {LeBlanc, David and Rodenburg, Cyril},
	editor = {Dolan, Thomas J.},
	year = {2017},
	doi = {10.1016/B978-0-08-101126-3.00018-X},
	keywords = {fluoride, generation IV, IMSR, integral molten salt reactor, nuclear system, Safety},
	pages = {541--556},
	file = {ScienceDirect Full Text PDF:/home/andrei2/Zotero/storage/TJGWFYKX/LeBlanc and Rodenburg - 2017 - 18 - Integral molten salt reactor.pdf:application/pdf;ScienceDirect Full Text PDF:/home/andrei2/Zotero/storage/BWQIA75R/LeBlanc and Rodenburg - 2017 - 18 - Integral molten salt reactor.pdf:application/pdf;ScienceDirect Full Text PDF:/home/andrei2/Zotero/storage/XW9JXU3P/LeBlanc and Rodenburg - 2017 - 18 - Integral molten salt reactor.pdf:application/pdf;ScienceDirect Snapshot:/home/andrei2/Zotero/storage/FIC5CLH8/B978008101126300018X.html:text/html;ScienceDirect Snapshot:/home/andrei2/Zotero/storage/4D5HBG5Q/B978008101126300018X.html:text/html;ScienceDirect Snapshot:/home/andrei2/Zotero/storage/SBH7K26M/B978008101126300018X.html:text/html}
}

@inproceedings{doe_technology_2002,
	title = {A technology roadmap for generation {IV} nuclear energy systems},
	booktitle = {Nuclear {Energy} {Research} {Advisory} {Committee} and the {Generation} {IV} {International} {Forum}},
	author = {DoE, U. S.},
	year = {2002},
	pages = {48--52},
	file = {genivroadmap2002.pdf:/home/andrei2/Zotero/storage/75S9TRA6/genivroadmap2002.pdf:application/pdf}
}

@inproceedings{betzler_modeling_2016,
	title = {Modeling and simulation of the start-up of a thorium-based molten salt reactor},
	booktitle = {Proc. {Int}. {Conf}. {PHYSOR}},
	author = {Betzler, Benjamin R. and Powers, J. J. and Worrall, A.},
	year = {2016},
	file = {Fulltext:/home/andrei2/Zotero/storage/H8Q3FER4/Betzler et al. - 2016 - Modeling and simulation of the start-up of a thori.pdf:application/pdf;Snapshot:/home/andrei2/Zotero/storage/MUZMZU8A/Betzler et al. - 2016 - Modeling and simulation of the start-up of a thori.pdf:application/pdf}
}

@article{haubenreich_experience_1970,
	title = {Experience with the {Molten}-{Salt} {Reactor} {Experiment}},
	volume = {8},
	issn = {00295450},
	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/andrei2/Zotero/storage/6EIEWPDA/Haubenreich_Engel_MSREexperience.pdf:application/pdf}
}

@inproceedings{gentry_initial_2017,
	address = {Washington D.C., United States},
	title = {Initial {Benchmarking} of {ChemTriton} and {MPACT} {MSR} {Modeling} {Capabilities}},
	booktitle = {Transactions of the {American} {Nuclear} {Society}},
	publisher = {American Nuclear Society, 555 North Kensington Avenue, La Grange Park, IL 60526 (United States)},
	author = {Gentry, Cole and Betzler, Benjamin and Collins, Benjamin},
	month = nov,
	year = {2017},
	file = {414.pdf:/home/andrei2/Zotero/storage/2VF8QHCA/414.pdf:application/pdf}
}

@techreport{noauthor_world_2017,
	title = {World {Energy} {Outlook} 2017},
	url = {https://www.iea.org/weo2017/},
	urldate = {2018-01-08},
	institution = {IEA},
	month = nov,
	year = {2017},
	file = {WEO 2017:/home/andrei2/Zotero/storage/ILRJTFMD/weo2017.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},
	url = {http://www.sciencedirect.com/science/article/pii/S1687850713000101},
	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},
	urldate = {2018-01-08},
	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:/home/andrei2/Zotero/storage/R4PU37EJ/Elsheikh - 2013 - Safety assessment of molten salt reactors in compa.pdf:application/pdf;ScienceDirect Snapshot:/home/andrei2/Zotero/storage/98N9YQRW/S1687850713000101.html:text/html}
}

@article{cherubini_co2_2011,
	title = {{CO}2 emissions from biomass combustion for bioenergy: atmospheric decay and contribution to global warming},
	volume = {3},
	issn = {1757-1707},
	shorttitle = {{CO}2 emissions from biomass combustion for bioenergy},
	url = {http://onlinelibrary.wiley.com/doi/10.1111/j.1757-1707.2011.01102.x/abstract},
	doi = {10.1111/j.1757-1707.2011.01102.x},
	abstract = {Carbon dioxide (CO2) emissions from biomass combustion are traditionally assumed climate neutral if the bioenergy system is carbon (C) flux neutral, i.e. the CO2 released from biofuel combustion approximately equals the amount of CO2 sequestered in biomass. This convention, widely adopted in life cycle assessment (LCA) studies of bioenergy systems, underestimates the climate impact of bioenergy. Besides CO2 emissions from permanent C losses, CO2 emissions from C flux neutral systems (that is from temporary C losses) also contribute to climate change: before being captured by biomass regrowth, CO2 molecules spend time in the atmosphere and contribute to global warming. In this paper, a method to estimate the climate impact of CO2 emissions from biomass combustion is proposed. Our method uses CO2 impulse response functions (IRF) from C cycle models in the elaboration of atmospheric decay functions for biomass-derived CO2 emissions. Their contributions to global warming are then quantified with a unit-based index, the GWPbio. Since this index is expressed as a function of the rotation period of the biomass, our results can be applied to CO2 emissions from combustion of all the different biomass species, from annual row crops to slower growing boreal forest.},
	language = {en},
	number = {5},
	urldate = {2018-01-08},
	journal = {GCB Bioenergy},
	author = {Cherubini, Francesco and Peters, Glen P. and Berntsen, Terje and Strømman, Anders H. and Hertwich, Edgar},
	month = oct,
	year = {2011},
	keywords = {bioenergy, carbon neutral, CO2 accounting, global warming potential, LCA},
	pages = {413--426},
	file = {Full Text PDF:/home/andrei2/Zotero/storage/YWBXS3T8/Cherubini et al. - 2011 - CO2 emissions from biomass combustion for bioenerg.pdf:application/pdf;Snapshot:/home/andrei2/Zotero/storage/A8SB9ZVB/abstract.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}},
	url = {https://www.sciencedirect.com/science/article/pii/S0306454917304760},
	doi = {10.1016/j.anucene.2017.12.025},
	abstract = {Moltres is a new physics application for modeling coupled physics in fluid-fuelled, molten salt reactors. This paper describes its neutronics model, thermal hydraulics model, and their coupling in the MOOSE framework. Neutron and precursor equations are implemented using an action system that allows use of an arbitrary number of groups with no change in the input card. Results for many-channel configurations in 2D-axisymmetric and 3D coordinates are presented and compared against other coupled models as well as the Molten Salt Reactor Experiment.},
	urldate = {2018-01-08},
	journal = {Annals of Nuclear Energy},
	author = {Lindsay, Alexander and Ridley, Gavin and Rykhlevskii, Andrei and Huff, Kathryn},
	month = apr,
	year = {2018},
	keywords = {agent based modeling, Finite elements, Hydrologic contaminant transport, MOOSE, Multiphysics, nuclear engineering, Nuclear fuel cycle, Object orientation, Parallel computing, Reactor physics, repository, Simulation, Systems analysis},
	pages = {530--540},
	file = {Moltres.pdf:/home/andrei2/Zotero/storage/4XDXRICB/Moltres.pdf:application/pdf;ScienceDirect Full Text PDF:/home/andrei2/Zotero/storage/E2T9U5IX/Lindsay et al. - 2018 - Introduction to Moltres An application for simula.pdf:application/pdf;ScienceDirect Snapshot:/home/andrei2/Zotero/storage/3DT9TEY3/S0306454917304760.html:text/html}
}

@techreport{eschbach_possible_1966,
	type = {Report},
	title = {{POSSIBLE} {OPTIMUM} {USE} {OF} {THORIUM} {AND} {URANIUM} {EMPLOYING} {CROSSED}-{PROGENY} {FUEL} {CYCLES}},
	language = {English},
	number = {BNWL-289},
	urldate = {2018-02-02},
	institution = {Pacific Northwest Laboratory},
	author = {Eschbach, E. A. and Deonigi, D. E.},
	month = may,
	year = {1966},
	doi = {10.2172/4505731},
	file = {Full Text PDF:/home/andrei2/Zotero/storage/246C4U3A/Eschbach and Deonigi - 1966 - POSSIBLE OPTIMUM USE OF THORIUM AND URANIUM EMPLOY.pdf:application/pdf;Snapshot:/home/andrei2/Zotero/storage/CSY76CR2/metadc1029446.html:text/html}
}

@article{natarajan_fast_2007,
	title = {Fast {Reactor} {Fuel} {Reprocessing} {Technology} in {India}},
	volume = {44},
	issn = {0022-3131},
	url = {http://www.tandfonline.com/doi/abs/10.1080/18811248.2007.9711299},
	doi = {10.1080/18811248.2007.9711299},
	abstract = {The limited indigenous uranium resource coupled with the need of energy independence necessitated the Department of Atomic Energy (DAE) in India to opt for a three-stage nuclear power program comprising of PHWRs in Stage 1, FBRs in Stage 2 and Th-U233 based reactors in Stage 3, respectively to meet the energy needs of the growing Indian economy. Presently the country has stepped into the 2nd Stage. The transition from the Stage 1 to 2 was fruitful thanks to the 2-decade long R\&D experience gained in Fast Reactor Fuel Reprocessing (FRFR). Closing the fast reactor fuel cycle through FRFR was inevitable for the success of the Indian Nuclear Power Program (INPP). The latest achievement by the center was the successful reprocessing of 100 GWd/t burnup mixed carbide fuel with 70\% Pu, discharged from the Fast Breeder Test Reactor (FBTR) which is also located in the same center. The designs of the various equipments and process flow sheet had stemmed from the above experiences thereby increasing the confidence level in the future plans of the Department. In this paper, an overview of the experiences in FRFR with glimpses of the various developmental activities towards the future plants is spelt out.},
	number = {3},
	urldate = {2018-02-02},
	journal = {Journal of Nuclear Science and Technology},
	author = {NATARAJAN, Rajamani and RAJ, Baldev},
	month = mar,
	year = {2007},
	keywords = {centrifugal extractor, fast reactor, fast reactor fuel reprocessing, SIMPSEX, single pin chopper},
	pages = {393--397},
	file = {Full Text PDF:/home/andrei2/Zotero/storage/WR2MX8SC/NATARAJAN and RAJ - 2007 - Fast Reactor Fuel Reprocessing Technology in India.pdf:application/pdf;Snapshot:/home/andrei2/Zotero/storage/6BKRZKXN/18811248.2007.html:text/html}
}

@article{bagla_thorium_2015,
	title = {Thorium seen as nuclear's new frontier},
	volume = {350},
	copyright = {Copyright © 2015, American Association for the Advancement of Science},
	issn = {0036-8075, 1095-9203},
	url = {http://science.sciencemag.org/content/350/6262/726},
	doi = {10.1126/science.350.6262.726},
	abstract = {In the wake of the meltdowns in Fukushima, Japan, in March 2011, several nations are taking a close look at the unheralded radioactive element, thorium, as a nuclear fuel—the theme of a conference in Mumbai, India, last month that drew participants from 30 countries. Compared with uranium, the standard reactor fuel, thorium is more abundant and harder to divert to weapons production, and it yields less radioactive waste. But thorium can't simply be swapped in for uranium in standard reactors. Taking up the engineering gauntlet, several nations are pursuing thorium reactors in which the fuel is dissolved in a bath of molten salt. India, meanwhile, plans to have a power reactor using solid thorium fuel running within 10 years.
Unsung reactor fuel is more abundant than uranium and, proponents say, safer.
Unsung reactor fuel is more abundant than uranium and, proponents say, safer.},
	language = {en},
	number = {6262},
	urldate = {2018-02-02},
	journal = {Science},
	author = {Bagla, Pallava},
	month = nov,
	year = {2015},
	pmid = {26564825},
	pages = {726--727},
	file = {726.full.pdf:/home/andrei2/Zotero/storage/7MG4WPB8/726.full.pdf:application/pdf;Snapshot:/home/andrei2/Zotero/storage/3IFNNRTU/726.html:text/html;Snapshot:/home/andrei2/Zotero/storage/FX5XH7Q4/726.html:text/html}
}

@article{ahmad_neutronics_2015,
	title = {Neutronics calculations for denatured molten salt reactors: {Assessing} resource requirements and proliferation-risk attributes},
	volume = {75},
	issn = {0306-4549},
	shorttitle = {Neutronics calculations for denatured molten salt reactors},
	url = {http://www.sciencedirect.com/science/article/pii/S0306454914003995},
	doi = {10.1016/j.anucene.2014.08.014},
	abstract = {Molten salt reactors (MSRs) are often advocated as a radical but worthwhile alternative to traditional reactor concepts based on solid fuels. This article builds upon the existing research into MSRs to model and simulate the operation of thorium-fueled single-fluid and two-fluid reactors. The analysis is based on neutronics calculations and focuses on denatured MSR systems. Resource utilization and basic proliferation-risk attributes are compared to those of standard light-water reactors. Depending on specific design choices, even fully denatured reactors could reduce uranium and enrichment requirements by a factor of 3–4. Overall, denatured single-fluid designs appear as the most promising candidate technology minimizing both design complexity and overall proliferation risks despite being somewhat less attractive from the perspective of resource utilization.},
	urldate = {2018-02-01},
	journal = {Annals of Nuclear Energy},
	author = {Ahmad, Ali and McClamrock, Edward B. and Glaser, Alexander},
	month = jan,
	year = {2015},
	keywords = {Denatured fuel, Proliferation resistance},
	pages = {261--267},
	file = {ScienceDirect Full Text PDF:/home/andrei2/Zotero/storage/6K4GFVDU/Ahmad et al. - 2015 - Neutronics calculations for denatured molten salt .pdf:application/pdf;ScienceDirect Snapshot:/home/andrei2/Zotero/storage/D7KZYFPM/S0306454914003995.html:text/html}
}

@article{sheu_depletion_2013,
	title = {Depletion analysis on long-term operation of the conceptual {Molten} {Salt} {Actinide} {Recycler} \& {Transmuter} ({MOSART}) by using a special sequence based on {SCALE}6/{TRITON}},
	volume = {53},
	issn = {0306-4549},
	url = {http://www.sciencedirect.com/science/article/pii/S0306454912004173},
	doi = {10.1016/j.anucene.2012.10.017},
	abstract = {A special sequence based on SCALE6/TRITON was developed to perform fuel cycle analysis of the Molten Salt Actinide Recycler \& Transmuter (MOSART), with emphasis on the simulation of its dynamic refueling and salt reprocessing scheme during long-term operation. MOSART is one of conceptual designs in the molten salt reactor (MSR) category of the Generation-IV systems. This type of reactors is distinguished by the use of liquid fuel circulating in and out of the core, which offers many unique advantages but complicates the modeling and simulation of core behavior using conventional reactor physics codes. The TRITON control module in SCALE6 can perform reliable depletion and decay analysis for many reactor physics applications due to its problem-dependent cross-section processing and rigorous treatment of neutron transport. In order to accommodate a simulation of on-line refueling and reprocessing scenarios, several in-house programs together with a run script were developed to integrate a series of stepwise TRITON calculations; the result greatly facilitates the neutronics analyses of long-term MSR operation. Using this method, a detailed reexamination of the MOSART operation in 30years was performed to investigate the neutronic characteristics of the core design, the change of fuel salt composition from start-up to equilibrium, the effects of various salt reprocessing scenarios, the performance of actinide transmutation, and the radiotoxicity reduction.},
	urldate = {2018-02-01},
	journal = {Annals of Nuclear Energy},
	author = {Sheu, R. J. and Chang, C. H. and Chao, C. C. and Liu, Y. -W. H.},
	month = mar,
	year = {2013},
	keywords = {Depletion analysis, Molten Salt Actinide Recycler \& Transmuter (MOSART), Molten salt reactors (MSRs), SCALE6/TRITON},
	pages = {1--8},
	file = {ScienceDirect Full Text PDF:/home/andrei2/Zotero/storage/Y43277ZR/Sheu et al. - 2013 - Depletion analysis on long-term operation of the c.pdf:application/pdf;ScienceDirect Snapshot:/home/andrei2/Zotero/storage/8X9GMCZ7/S0306454912004173.html:text/html}
}

@article{fiorina_investigation_2013,
	title = {Investigation of the {MSFR} core physics and fuel cycle characteristics},
	volume = {68},
	issn = {0149-1970},
	url = {http://www.sciencedirect.com/science/article/pii/S0149197013001236},
	doi = {10.1016/j.pnucene.2013.06.006},
	abstract = {The adoption of Th fuel in fast reactors is being reconsidered due to the potential favorable impact on actinide waste management and resource availability. A closed Th cycle leads to an actinide inventory with lower radiotoxicity and heat load for the first several thousands of years. Due to the typically low TRansUranic (TRU) Conversion Ratio (CR), Th can also be advantageous to expedite the consumption of legacy TRU. One of the main obstacles to the implementation of Th is the highly radioactive recycled fuel which requires remote handling under heavy shielding, inevitably penalizing economics and challenging conventional pin-based fuel manufacturing. From this perspective, the development of liquid-fuelled reactors, with Molten Salt Reactors regarded as the most promising, appears particularly attractive as fuel handling would be greatly simplified. The present paper investigates the fuel cycle performances of the reference GEN-IV Molten Salt Fast Reactor (MSFR) in terms of isotope evolution, radiotoxicity generation and safety-related parameters. Similarly to most MSR concepts proposed in the past, the MSFR is based on the fluoride molten salt technology, but it features the novelty of a fast neutron spectrum. Calculations are performed using state-of-the-art equilibrium-cycle methodologies, i.e., the ERANOS-based EQL3D procedure developed at the Paul Scherrer Institut and extended to the simulation of the MSFR. Selected results have been benchmarked with the Monte Carlo code PSG2/SERPENT. These results have also been used for the assessment of a diffusion module based on the COMSOL multi-physics toolkit, which is the subject of current studies aimed at efficiently simulating the peculiar MSFR transient behavior.},
	urldate = {2018-02-01},
	journal = {Progress in Nuclear Energy},
	author = {Fiorina, Carlo and Aufiero, Manuele and Cammi, Antonio and Franceschini, Fausto and Krepel, Jiri and Luzzi, Lelio and Mikityuk, Konstantin and Ricotti, Marco Enrico},
	month = sep,
	year = {2013},
	keywords = {Fast-spectrum molten salt reactor, Safety-related parameters, Thorium fuel cycle, Waste management},
	pages = {153--168},
	file = {ScienceDirect Full Text PDF:/home/andrei2/Zotero/storage/K7YYXUQF/Fiorina et al. - 2013 - Investigation of the MSFR core physics and fuel cy.pdf:application/pdf;ScienceDirect Snapshot:/home/andrei2/Zotero/storage/CD82YK4U/S0149197013001236.html:text/html}
}

@article{heuer_simulation_2010,
	title = {Simulation {Tools} and {New} {Developments} of the {Molten} {Salt} {Fast} {Reactor}},
	copyright = {© SFEN 2010},
	issn = {0335-5004},
	url = {https://rgn.publications.sfen.org/articles/rgn/abs/2010/06/rgn20106p95/rgn20106p95.html},
	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},
	urldate = {2018-02-01},
	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 = {Snapshot:/home/andrei2/Zotero/storage/ECV96N73/rgn20106p95.html:text/html}
}

@article{nuttin_potential_2005,
	title = {Potential of thorium molten salt reactorsdetailed calculations and concept evolution with a view to large scale energy production},
	volume = {46},
	issn = {0149-1970},
	url = {http://www.sciencedirect.com/science/article/pii/S0149197004000794},
	doi = {10.1016/j.pnucene.2004.11.001},
	abstract = {We discuss here the concept of Thorium Molten Salt Reactor dedicated to future nuclear energy production. The fuel of such reactors being liquid, it can be easily reprocessed to overcome neutronic limits. In the late sixties, the MSBR project showed that breeding is possible with thorium in a thermal spectrum, provided that an efficient pyrochemical reprocessing is added. With tools developed around the Monte Carlo MCNP code, we first re-evaluate the performance of a MSBR-like reference system with 232Th/233U fuel. We find an important reduction of inventories and induced radiotoxicities at equilibrium compared to other fuel cycles, with a doubling time of about thirty years. We then study how to start this interesting reference system with theplutonium from PWR spent fuel. Such a transition appears slow and difficult, since it is very sensitive to the fissile quality of the plutonium used. Deployment scenarios of 232Th/233U MSBR-like systems from the existing French PWRs demonstrate the advantage of an upstream 233U production in other reactors, allowing a direct start of the MSBR-like systems with 233U. This finally leads us to explore alternatives to some MSBR features, for energy production with 232Th/233U fuel from the start. We thus test different options, especially in terms of core neutronics optimization and reprocessing unit adaptation.},
	number = {1},
	urldate = {2018-02-01},
	journal = {Progress in Nuclear Energy},
	author = {Nuttin, A. and Heuer, D. and Billebaud, A. and Brissot, R. and Le Brun, C. and Liatard, E. and Loiseaux, J. -M. and Mathieu, L. and Meplan, O. and Merle-Lucotte, E. and Nifenecker, H. and Perdu, F. and David, S.},
	month = jan,
	year = {2005},
	keywords = {MCNP, pyrochemistry, thorium fuel},
	pages = {77--99},
	file = {ScienceDirect Full Text PDF:/home/andrei2/Zotero/storage/G5G2ZZJT/Nuttin et al. - 2005 - Potential of thorium molten salt reactorsdetailed .pdf:application/pdf;ScienceDirect Snapshot:/home/andrei2/Zotero/storage/V3M2G23K/S0149197004000794.html:text/html}
}

@techreport{croff_users_1980,
	title = {User's manual for the {ORIGEN}2 computer code},
	language = {en},
	number = {ORNL/TM--7175},
	urldate = {2018-02-01},
	institution = {Oak Ridge National Lab.},
	author = {Croff, A. G.},
	year = {1980},
	file = {Full Text PDF:/home/andrei2/Zotero/storage/H399VWKG/Croff - 1980 - User's manual for the ORIGEN2 computer code.pdf:application/pdf;Snapshot:/home/andrei2/Zotero/storage/VE3SQH6Q/search.html:text/html}
}

@article{bowman_scale_2011,
	title = {{SCALE} 6: {Comprehensive} {Nuclear} {Safety} {Analysis} {Code} {System}},
	volume = {174},
	shorttitle = {{SCALE} 6},
	url = {http://epubs.ans.org/?a=11717},
	doi = {dx.doi.org/10.13182/NT10-163},
	number = {2},
	urldate = {2018-02-01},
	journal = {Nuclear Technology},
	author = {Bowman, S. M.},
	month = may,
	year = {2011},
	pages = {126--148},
	file = {Snapshot:/home/andrei2/Zotero/storage/QK6CS7SS/a_11717.html:text/html}
}

@article{gauld_isotopic_2011,
	title = {Isotopic {Depletion} and {Decay} {Methods} and {Analysis} {Capabilities} in {SCALE}},
	volume = {174},
	url = {http://epubs.ans.org/?a=11719},
	doi = {dx.doi.org/10.13182/NT11-3},
	number = {2},
	urldate = {2018-02-01},
	journal = {Nuclear Technology},
	author = {Gauld, Ian C. and Radulescu, Georgeta and Ilas, Germina and Murphy, Brian D. and Williams, Mark L. and Wiarda, Dorothea},
	month = may,
	year = {2011},
	pages = {169--195},
	file = {Snapshot:/home/andrei2/Zotero/storage/VEJPQE5K/a_11719.html:text/html}
}

@techreport{xu_mcode_2008,
	title = {{MCODE}, {Version} 2.2: an {MCNP}-{ORIGEN} depletion program},
	shorttitle = {{MCODE}, {Version} 2.2},
	institution = {Massachusetts Institute of Technology. Center for Advanced Nuclear Energy Systems. Nuclear Fuel Cycle Program},
	author = {Xu, Z. and Hejzlar, Pavel},
	year = {2008},
	file = {Snapshot:/home/andrei2/Zotero/storage/W8DJTURN/75242.html:text/html}
}

@article{goluoglu_monte_2011,
	title = {Monte {Carlo} {Criticality} {Methods} and {Analysis} {Capabilities} in {SCALE}},
	volume = {174},
	issn = {0029-5450},
	url = {http://ans.tandfonline.com/doi/10.13182/NT10-124},
	doi = {10.13182/NT10-124},
	abstract = {This paper describes the Monte Carlo codes KENO V.a and KENO-VI in SCALE that are primarily used to calculate multiplication factors and flux distributions of fissile systems. Both codes allow explicit geometric representation of the target systems and are used internationally for safety analyses involving fissile materials. KENO V.a has limiting geometric rules such as no intersections and no rotations. These limitations make KENO V.a execute very efficiently and run very fast. On the other hand, KENO-VI allows very complex geometric modeling. Both KENO codes can utilize either continuous-energy or multigroup cross-section data and have been thoroughly verified and validated with ENDF libraries through ENDF/B-VII.0, which has been first distributed with SCALE 6. Development of the Monte Carlo solution technique and solution methodology as applied in both KENO codes is explained in this paper. Available options and proper application of the options and techniques are also discussed. Finally, performance of the codes is demonstrated using published benchmark problems.},
	number = {2},
	urldate = {2018-02-01},
	journal = {Nuclear Technology},
	author = {Goluoglu, Sedat and Petrie, Lester M. and Dunn, Michael E. and Hollenbach, Daniel F. and Rearden, Bradley T.},
	month = may,
	year = {2011},
	pages = {214--235},
	file = {Monte Carlo Criticality Methods and Analysis Capabilities in SCALE.pdf:/home/andrei2/Zotero/storage/RJVRWDY6/Monte Carlo Criticality Methods and Analysis Capabilities in SCALE.pdf:application/pdf;Snapshot:/home/andrei2/Zotero/storage/JJD2GYTH/NT10-124.html:text/html}
}

@article{goorley_mcnp6_2013,
	title = {{MCNP}6 {User}’s {Manual}, {Version} 1.0},
	journal = {LA-CP-13-00634, Los Alamos National Laboratory},
	author = {Goorley, John T. and James, M. R. and Booth, T. E.},
	year = {2013}
}

@misc{noauthor_mcnp_2004,
	title = {{MCNP} - {A} {General} {Monte} {Carlo} {N}-{Particle} {Transport} {Code}},
	url = {http://mcnp.lanl.gov},
	publisher = {Los Alamos National Laboratory},
	year = {2004}
}

@techreport{bauman_rod:_1971,
	title = {Rod: {A} {Nuclear} and {Fuel}-{Cycle} {Analysis} {Code} for {Circulating}-{Fuel} {Reactors}.},
	shorttitle = {Rod},
	language = {English},
	number = {ORNL-TM-3359},
	urldate = {2018-01-31},
	institution = {Oak Ridge National Lab., Tenn.},
	author = {Bauman, H. F. and Cunningham III, G. W. and Lucius, J. L. and Kerr, H. T. and Craven, C. W. Jr},
	month = jan,
	year = {1971},
	doi = {10.2172/4741221},
	keywords = {computer program for calculation of optimum, computer program for calculation of optimum design parameters of, computer programs/rod, computers, fluid fueled reactors, fluid- fueled/fuel cycle for, fluid-fueled/core for, for core design and fuel cycle calculations for fluid fueled reactors, fortran, fortran 4 for ibm 360, Fuel cycle, ibm 360, molten salt reactors, molten-salt/core for, molten-salt/fuel cycle for, n38150* -power reactor development-fuels, programming, reactor core, reactor fuels/cycle for fluid and molten salt, reactors},
	file = {Full Text PDF:/home/andrei2/Zotero/storage/HVB93XTF/Bauman et al. - 1971 - Rod A Nuclear and Fuel-Cycle Analysis Code for Ci.pdf:application/pdf}
}

@techreport{kee_mrpp:_1976,
	title = {{MRPP}: multiregion processing plant code},
	shorttitle = {{MRPP}},
	language = {en},
	number = {ORNL/TM--4210},
	urldate = {2018-01-31},
	institution = {Oak Ridge National Lab.},
	author = {Kee, C. W. and McNeese, L. E.},
	year = {1976},
	file = {Full Text PDF:/home/andrei2/Zotero/storage/WLNIC9JF/Kee and McNeese - 1976 - MRPP multiregion processing plant code.pdf:application/pdf;Snapshot:/home/andrei2/Zotero/storage/4BNASWUD/search.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 = {boilers, buildings, configuration, containment, control systems, coolant loops, design, Economics, fused salt fuel, maintenance, MSBR, n38100*  --power reactor development, operation, reactor sites, waste disposal molten salt breeder reactor/design parameters for conceptual 1000 mw(e) single fluid},
	file = {3445600591394.pdf:/home/andrei2/Zotero/storage/RC7URKHP/3445600591394.pdf:application/pdf;Full Text PDF:/home/andrei2/Zotero/storage/DFSGI5SW/Robertson - 1971 - Conceptual Design Study of a Single-Fluid Molten-S.pdf:application/pdf;Snapshot:/home/andrei2/Zotero/storage/G8IEVM75/4030941.html:text/html}
}

@techreport{noauthor_report_1959,
	title = {Report of the {Fluid} {Fuel} {Reactors} {Task} {Force}},
	institution = {US-AEC TASK FORCE},
	month = feb,
	year = {1959}
}

@article{rosenthal_molten-salt_1970,
	title = {Molten-{Salt} {Reactors}—{History}, {Status}, and {Potential}},
	volume = {8},
	issn = {0550-3043},
	url = {https://doi.org/10.13182/NT70-A28619},
	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},
	urldate = {2018-01-23},
	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:/home/andrei2/Zotero/storage/SJ8BSHG9/Rosenthal et al. - 1970 - Molten-Salt Reactors—History, Status, and Potentia.pdf:application/pdf;Snapshot:/home/andrei2/Zotero/storage/7NZ5QDPF/NT70-A28619.html:text/html}
}

@techreport{noauthor_key_2015,
	type = {Text},
	title = {Key {World} {Energy} {Statistics}, {Statistics} / {Key} {World} {Energy} {Statistics} / 2015},
	url = {http://www.oecd-ilibrary.org/energy/key-world-energy-statistics-2015_key_energ_stat-2015-en},
	abstract = {The IEA produced its first handy, pocket-sized summary of key energy data in 1997 and every year since then it has been more and more successful. Key World En},
	language = {en},
	urldate = {2018-01-08},
	institution = {IEA},
	year = {2015},
	file = {keyworld2015.pdf:/home/andrei2/Zotero/storage/R5R2I4HY/keyworld2015.pdf:application/pdf;Snapshot:/home/andrei2/Zotero/storage/XQ5WD37Z/key-world-energy-statistics-2015_key_energ_stat-2015-en.html:text/html}
}

@techreport{noauthor_key_2017,
	title = {Key {World} {Energy} {Statistics} 2017},
	url = {https://www.iea.org/publications/freepublications/publication/key-world-energy-statistics.html},
	urldate = {2018-01-08},
	institution = {IEA},
	year = {2017},
	file = {KeyWorld2017.pdf:/home/andrei2/Zotero/storage/Y55L8WPP/KeyWorld2017.pdf:application/pdf;Publication\: Key World Energy Statistics:/home/andrei2/Zotero/storage/94PVCCG4/key-world-energy-statistics.html:text/html}
}

@inproceedings{rykhlevskii_online_2017,
	address = {Washington, DC, United States},
	title = {Online reprocessing simulation for thorium-fueled molten salt breeder reactor},
	abstract = {The current paper presents a single-cell model developed using the continuous-energy Serpent 2 Monte Carlo reactor physics software. It was employed to establish a Serpent- based method for finding the equilibrium core composition and core depletion of the Molten Salt Breeder Reactor (MSBR).},
	booktitle = {Transactions of the {American} {Nuclear} {Society}},
	publisher = {American Nuclear Society},
	author = {Rykhlevskii, Andrei and Lindsay, Alexander and Huff, Kathryn D.},
	month = nov,
	year = {2017},
	file = {msbr_reproc.pdf:/home/andrei2/Zotero/storage/PASSAITQ/msbr_reproc.pdf:application/pdf}
}

@inproceedings{rykhlevskii_full-core_2017,
	address = {Washington, DC, United States},
	title = {Full-core analysis of thorium-fueled {Molten} {Salt} {Breeder} {Reactor} using the {SERPENT} 2 {Monte} {Carlo} code},
	abstract = {We used the continuous-energy Serpent 2 Monte Carlo
particle transport code to calulate whole-core depletion in the thermal spectrum Molten Salt Breeder Reactor
(MSBR) . We then compare these results with existing MCNP6
results with a more simplified geometric model. This
neutronics model is of sufficient fidelity to inform optimization
of fuel salt composition, fuel utilization, neutron fluxes, and
spectrum evaluation. Moreover, this model will be employed
for depeletion calculations, generation of problem-oriented
homogenized nuclear data (multi-group cross sections and
diffusion constants) for deterministic reactor codes, and multiphysics
simulations.},
	booktitle = {Transactions of the {American} {Nuclear} {Society}},
	publisher = {American Nuclear Society},
	author = {Rykhlevskii, Andrei and Lindsay, Alexander and Huff, Kathryn D.},
	month = nov,
	year = {2017},
	file = {full-core_msbr_model.pdf:/home/andrei2/Zotero/storage/3F8DA752/full-core_msbr_model.pdf:application/pdf}
}

@inproceedings{powers_inventory_2014,
	address = {Kyoto, Japan},
	title = {An inventory analysis of thermal-spectrum thorium-fueled molten salt reactor concepts},
	booktitle = {{PHYSOR} 2014},
	publisher = {JAEA-CONF--2014-003},
	author = {Powers, Jeffrey J. and Gehin, Jess C. and Worrall, Andrew and Harrison, T. Jay and Sunny, Eva E.},
	month = oct,
	year = {2014},
	file = {1159468.pdf:/home/andrei2/Zotero/storage/F7V5UJWY/1159468.pdf:application/pdf;Snapshot:/home/andrei2/Zotero/storage/XJD36U2S/search.html:text/html}
}

@misc{the_hdf_group_hierarchical_1997,
	title = {Hierarchical data format, version 5},
	url = {https://www.hdfgroup.org/solutions/hdf5/},
	urldate = {2018-02-23},
	author = {The HDF Group},
	year = {1997},
	file = {Snapshot:/home/andrei2/Zotero/storage/QR7UPGNY/hdf5.html:text/html}
}

@misc{sorensen_one-fluid_2006,
	type = {Blog},
	title = {One-{Fluid} {MSBR} {Chemical} {Processing} -},
	url = {http://energyfromthorium.com/history/msbr-design-one-fluid-chem/},
	abstract = {Research into reductive extraction for bismuth and rare-earths (lanthanides) made ORNL chemical engineers particularly optimistic about the application of these processes to molten-salt breeder reactors.  At the same time data was coming back to reactor designers on the dimensional stability of graphite at various temperatures in fast neutrons.  ...},
	language = {en-US},
	urldate = {2018-02-27},
	journal = {THORIUM},
	author = {Sorensen, Kirk},
	year = {2006},
	file = {Snapshot:/home/andrei2/Zotero/storage/R8VCF84N/msbr-design-one-fluid-chem.html:text/html}
}

@inproceedings{jeong_development_2014,
	address = {Kyoto, Japan},
	title = {Development of {Computer} {Code} {Packages} for {Molten} {Salt} {Reactor} {Core} {Analysis}},
	booktitle = {{PHYSOR} 2014},
	author = {Jeong, Yongjin and Choi, Sooyoung and Lee, Deokjung},
	month = sep,
	year = {2014},
	file = {Full Text PDF:/home/andrei2/Zotero/storage/QRWWV5RQ/Jeong et al. - 2014 - Development of Computer Code Packages for Molten S.pdf:application/pdf;Snapshot:/home/andrei2/Zotero/storage/Q8DACV96/272960553_Development_of_Computer_Code_Packages_for_Molten_Salt_Reactor_Core_Analysis.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},
	urldate = {2018-02-02},
	institution = {Nuclear Fuel Cycle and Materials Section International Atomic Energy Agency},
	author = {IAEA},
	month = may,
	year = {2005},
	pages = {113},
	file = {Full Text PDF:/home/andrei2/Zotero/storage/4SR2PB6K/IAEA - 2005 - Thorium Fuel Cycle - Potential Benefits and Challe.pdf:application/pdf;IAEA Report on Thorium Fuel Cycle--May 2005.pdf:/home/andrei2/Zotero/storage/EI8HB6GT/IAEA Report on Thorium Fuel Cycle--May 2005.pdf:application/pdf;Snapshot:/home/andrei2/Zotero/storage/38XVA5P2/Thorium-Fuel-Cycle-Potential-Benefits-and-Challenges.html:text/html}
}

@techreport{briggs_summary_1967,
	title = {{SUMMARY} {OF} {THE} {OBJECTIVES}, {THE} {DESIGN}, {AND} {A} {PROGRAM} {OF} {DEVELOPMENT} {OF} {MOLTEN}-{SALT} {BREEDER} {REACTORS}.},
	number = {ORNL-TM-1851},
	institution = {Oak Ridge National Lab., Tenn.},
	author = {Briggs, R. B.},
	year = {1967}
}

@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 – 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.},
	number = {6},
	urldate = {2016-08-17},
	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 = {ScienceDirect Full Text PDF:/home/andrei2/Zotero/storage/T5JKV5U6/Křepel et al. - 2007 - DYN3D-MSR spatial dynamics code for molten salt re.pdf:application/pdf;ScienceDirect Snapshot:/home/andrei2/Zotero/storage/Q24HJQHD/S0306454907000527.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 = {3D, Core, Molten Salt Reactor (MSR), MSR Experiment (MSRE), reactor physics, unread},
	pages = {118--131},
	file = {Kópházi et al. - Development of a Three-Dimensional Time-Dependent .pdf:/home/andrei2/Zotero/storage/ZIJ5Q643/Kópházi et al. - Development of a Three-Dimensional Time-Dependent .pdf:application/pdf}
}

@inproceedings{kasten_mosel_1964,
	address = {Geneva},
	title = {The {MOSEL} {Reactor} {Concept}},
	booktitle = {Proc. {Third} {International} {Conference} on the {Peaceful} {Uses} of {Atomic} {Energy}},
	publisher = {Third Intern. Conf. on the Peaceful Uses of Atomic Energy},
	author = {Kasten, P. R.},
	year = {1964},
	file = {Fulltext:/home/andrei2/Zotero/storage/7NL8X7XC/Kasten - The MOSEL reactor concept.pdf:application/pdf;Fulltext:/home/andrei2/Zotero/storage/HUF4E5YB/Kasten - 1964 - The MOSEL-Reactor-Concept.pdf:application/pdf;Snapshot:/home/andrei2/Zotero/storage/K7TA4CP4/Kasten - The MOSEL reactor concept.pdf:application/pdf;Snapshot:/home/andrei2/Zotero/storage/CKIFS8S6/Kasten - 1964 - The MOSEL-Reactor-Concept.pdf:application/pdf}
}

@article{park_whole_2015,
	title = {Whole core analysis of molten salt breeder reactor with online fuel reprocessing},
	volume = {39},
	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},
	abstract = {The simulation of whole core depletion and continuous reprocessing of a molten salt breeder reactor (MSBR) was performed. The MSBR model was built using MCNP6, and the depletion and reprocessing simulations were modeled using CINDER90 and a Python script. The Python script was introduced to implement online reprocessing of molten salt fuel and the feeding of new fertile material with 3-day depletion intervals during the simulations. The simulation started with the reference composition from the original Oak Ridge National Laboratory MSBR design. Equilibrium compositions were obtained from depletion and reprocessing simulations 7000 days worth of data. The MSBR whole core analysis was performed at the initial and equilibrium core conditions, for various reactor design parameters such as multiplication factors, neutron flux distributions, temperature coefficients, rod worths, and power distributions. The neutronic core characteristics were analyzed using a four-factor formula applied to the two zones of the core separately. Copyright © 2015 John Wiley \& Sons, Ltd.},
	language = {en},
	number = {12},
	urldate = {2016-11-10},
	journal = {International Journal of Energy Research},
	author = {Park, Jinsu and Jeong, Yongjin and Lee, Hyun Chul and Lee, Deokjung},
	month = oct,
	year = {2015},
	keywords = {equilibrium state, MCNP, MSBR, MSR, online reprocessing, whole core},
	pages = {1673--1680},
	file = {Full Text PDF:/home/andrei2/Zotero/storage/S3L2KPS6/Park et al. - 2015 - Whole core analysis of molten salt breeder reactor.pdf:application/pdf;Snapshot:/home/andrei2/Zotero/storage/B4H7QFMR/abstract.html:text/html;Whole core analysis of molten salt breeder reactor with online fuel reprocessing.pdf:/home/andrei2/Zotero/storage/TPRZSD4Q/Whole core analysis of molten salt breeder reactor with online fuel reprocessing.pdf:application/pdf}
}

@article{laureau_transient_2015,
	title = {Transient {Fission} {Matrix}: {Kinetic} calculation and kinetic parameters βeff and Λeff calculation},
	volume = {85},
	issn = {0306-4549},
	shorttitle = {Transient {Fission} {Matrix}},
	url = {http://www.sciencedirect.com/science/article/pii/S0306454915003825},
	doi = {10.1016/j.anucene.2015.07.023},
	abstract = {In this paper, we present a time dependent version of the fission matrix method called Transient Fission Matrix (TFM) that we developed to perform accurate spatial kinetic calculations with a reduced computational cost. The matrices are generated once using the Monte Carlo neutron transport MCNP prior to the kinetic calculations. This article is devoted to the description of this approach with two different sets of equations depending on the aimed precision. We tested the TFM approach in several instances. In the calculation of the kinetic parameters such as βeff and Λeff on the Flattop and Jezebel experiments, the results obtained are compatible with adjoint weighted calculations performed with the Monte Carlo neutron transport code SERPENT. Likewise, our calculated value of the αRossi observable is also compatible with the experimental measurement. Finally a time–space prompt neutron propagation induced by a neutron pulse on Flattop has been calculated using the TFM approach and compared to a reference time dependent SERPENT calculation. Again, the results obtained demonstrate the good characterization of the time–space system response by the TFM approach.},
	journal = {Annals of Nuclear Energy},
	author = {Laureau, A. and Aufiero, M. and Rubiolo, P. R. and Merle-Lucotte, E. and Heuer, D.},
	year = {2015},
	keywords = {Kinetics, Neutronics, Transient, Transient Fission Matrix},
	pages = {1035--1044},
	file = {Fulltext:/home/andrei2/Zotero/storage/4JC699FZ/S0306454915003825.html:text/html;ScienceDirect Snapshot:/home/andrei2/Zotero/storage/AXDDPT7V/S0306454915003825.html:text/html;Snapshot:/home/andrei2/Zotero/storage/QHMMQAUH/S0306454915003825.html:text/html}
}

@article{betzler_fuel_2018,
	title = {Fuel cycle and neutronic performance of a spectral shift molten salt reactor design},
	volume = {119},
	issn = {0306-4549},
	url = {https://www.sciencedirect.com/science/article/pii/S0306454918302287},
	doi = {10.1016/j.anucene.2018.04.043},
	abstract = {The fuel cycle performance and core design of the Transatomic Power liquid-fueled molten salt reactor concept is analyzed. This advanced reactor concept uses configurable zirconium hydride moderator rod assemblies to shift the neutron spectrum in the core from intermediate at beginning of life to thermal at end of life. With a harder spectrum during the early years of reactor operation, this spectral shift design drives captures in fertile 238U. The converted fissile plutonium makes up over 50\% of the fissile material in the fuel salt over the last half (∼15 years) of reactor operation. A softer spectrum late in reactor life helps drive the fuel to a burnup of 90 GWd/MTU. Continuously changing physics necessitates time-dependent analyses resolved over long timescales (i.e., months to years), as this concept does not meet an equilibrium condition. The spectral shift and molten salt reactor material feeds and removals enable this concept to perform better in fuel cycle metrics, increasing resource utilization by more than 50\% compared with a typical light water reactor (i.e., from ∼0.6\% to ∼1\%). These metrics are compared to similar fuel cycles using alternate technologies. Additional core design and analysis challenges associated with the spectral shift and use of molten salt reactor technology are identified and discussed.},
	urldate = {2018-06-06},
	journal = {Annals of Nuclear Energy},
	author = {Betzler, Benjamin R. and Robertson, Sean and Davidson, Eva E. and Powers, Jeffrey J. and Worrall, Andrew and Dewan, Leslie and Massie, Mark},
	month = sep,
	year = {2018},
	keywords = {Fuel cycle, Molten salt reactor, Spectral shift, Zirconium hydride},
	pages = {396--410},
	file = {ScienceDirect Full Text PDF:/home/andrei2/Zotero/storage/SZFV9Q2N/Betzler et al. - 2018 - Fuel cycle and neutronic performance of a spectral.pdf:application/pdf;ScienceDirect Full Text PDF:/home/andrei2/Zotero/storage/8MP5IAP8/Betzler et al. - 2018 - Fuel cycle and neutronic performance of a spectral.pdf:application/pdf;ScienceDirect Full Text PDF:/home/andrei2/Zotero/storage/T6DVVTNP/Betzler et al. - 2018 - Fuel cycle and neutronic performance of a spectral.pdf:application/pdf;ScienceDirect Snapshot:/home/andrei2/Zotero/storage/CVPEBMPH/S0306454918302287.html:text/html;ScienceDirect Snapshot:/home/andrei2/Zotero/storage/8H55YE7H/S0306454918302287.html:text/html;ScienceDirect Snapshot:/home/andrei2/Zotero/storage/I298VS3S/S0306454918302287.html:text/html;Snapshot:/home/andrei2/Zotero/storage/CPT9CPPE/S0306454918302287.html:text/html}
}

@techreport{derstine_dif3d:_1984,
	title = {{DIF}3D: {A} code to solve one-, two-, and three-dimensional finite-difference diffusion theory problems},
	shorttitle = {{DIF}3D},
	institution = {Argonne National Lab.},
	author = {Derstine, K. L.},
	year = {1984},
	file = {Full Text:/home/andrei2/Zotero/storage/TAYF3RHZ/Derstine - 1984 - DIF3D A code to solve one-, two-, and three-dimen.pdf:application/pdf;Snapshot:/home/andrei2/Zotero/storage/99AHLCPI/search.html:text/html}
}

@article{zhou_fuel_2018,
	title = {Fuel cycle analysis of molten salt reactors based on coupled neutronics and thermal-hydraulics calculations},
	volume = {114},
	journal = {Annals of Nuclear Energy},
	author = {Zhou, Shengcheng and Yang, Won Sik and Park, Tongkyu and Wu, Hongchun},
	year = {2018},
	pages = {369--383},
	file = {Full Text:/home/andrei2/Zotero/storage/JT7J2PUA/S0306454917303754.html:text/html}
}

@techreport{forget_integral_2018,
	title = {Integral {Full} {Core} {Multi}-{Physics} {PWR} {Benchmark} with {Measured} {Data}},
	institution = {Massachusetts Institute of Technology},
	author = {Forget, Benoit and Smith, Kord and Kumar, Shikhar and Rathbun, Miriam and Liang, Jingang},
	year = {2018},
	file = {Full Text:/home/andrei2/Zotero/storage/WU654HVA/Forget et al. - 2018 - Integral Full Core Multi-Physics PWR Benchmark wit.pdf:application/pdf;Snapshot:/home/andrei2/Zotero/storage/BANRFSJJ/1432668.html:text/html}
}

@inproceedings{betzler_implementation_2017,
	address = {Jeju, Korea},
	title = {Implementation of {Molten} {Salt} {Reactor} {Tools} in {SCALE}},
	booktitle = {Proc. {M}\&{C} 2017 - {International} {Conference} on {Mathematics} \& {Computational} {Methods} {Applied} to {Nuclear} {Science} and {Engineering}},
	author = {Betzler, Benjamin R. and Powers, J. J. and Brown, N. R. and Rearden, B. T.},
	month = apr,
	year = {2017},
	file = {Betzler et al. - 2017 - Implementation of Molten Salt Reactor Tools in SCA.pdf:/home/andrei2/Zotero/storage/ZFPCRTYR/Betzler et al. - 2017 - Implementation of Molten Salt Reactor Tools in SCA.pdf:application/pdf}
}

@inproceedings{ashraf_nuclear_2018,
	title = {Nuclear fuel optimization for molten salt fast reactor},
	volume = {1133},
	booktitle = {Journal of {Physics}: {Conference} {Series}},
	publisher = {IOP Publishing},
	author = {Ashraf, O. and Smirnov, A. D. and Tikhomirov, G. V.},
	year = {2018},
	pages = {012026},
	file = {Full Text:/home/andrei2/Zotero/storage/AUEG4MZV/Ashraf et al. - 2018 - Nuclear fuel optimization for molten salt fast rea.pdf:application/pdf;Snapshot:/home/andrei2/Zotero/storage/P53F37KF/meta.html:text/html}
}