bibliography.bib

% CHAPTER 1: INTRODUCTION
@techreport{brey_development_2001,
	address = {Vienna, Austria},
	title = {Development {History} of the {Gas} {Turbine} {Modular} {High} {Temperature} {Reactor}},
	abstract = {The development of the high temperature gas cooled reactor (HTGR) as an environmentally agreeable
and efficient power source to support the generation of electricity and achieve a broad range of high
temperature industrial applications has been an evolutionary process spanning over four decades. This process has included ongoing major development in both the HTGR as a nuclear energy source and
associated power conversion systems from the steam cycle to the gas turbine. This paper follows the development process progressively through individual plant designs from early research of the 1950s to the present focus on the gas turbine modular HTGR.},
	number = {IAEA-TECDOC--1238},
	author = {Brey, H.L.},
	month = aug,
	year = {2001},
	file = {Brey - 2001 - Development History of the Gas Turbine Modular Hig.pdf:/home/roberto/Zotero/storage/2II77DPZ/Brey - 2001 - Development History of the Gas Turbine Modular Hig.pdf:application/pdf}
}

@techreport{iaea_current_2001,
	address = {Vienna, Austria},
	title = {Current status and future development of modular high temperature gas cooled reactor technology},
	url = {https://inis.iaea.org/collection/NCLCollectionStore/_Public/32/018/32018305.pdf?r=1},
	number = {IAEA-TECDOC--1198},
	institution = {IAEA},
	author = {IAEA},
	month = feb,
	year = {2001},
	pages = {267},
	file = {IAEA - 2001 - Current status and future development of modular h.pdf:/home/roberto/Zotero/storage/7S5NPH39/IAEA - 2001 - Current status and future development of modular h.pdf:application/pdf}
}

@misc{demkowickz_triso_2019,
	address = {Idaho Falls, ID},
	type = {{NRC} {HTGR} {Training}},
	title = {{TRISO} {Fuel}: {Design}, {Manufacturing}, and {Performance}},
	url = {https://inldigitallibrary.inl.gov/sites/sti/sti/Sort_24838.pdf},
	author = {Demkowickz, Paul},
	month = jul,
	year = {2019},
	file = {Demkowickz, Paul - 2019 - TRISO Fuel Design, Manufacturing, and Performance.pdf:/home/roberto/Zotero/storage/SW8FW2JS/Demkowickz, Paul - 2019 - TRISO Fuel Design, Manufacturing, and Performance.pdf:application/pdf}
}

@phdthesis{huning_steady_2014,
	address = {Atlanta, GA},
	type = {Masters of {Science} in {Nuclear} {Engineering}},
	title = {A {STEADY} {STATE} {THERMAL} {HYDRAULIC} {ANALYSIS} {METHOD} {FOR} {PRISMATIC} {GAS} {REACTORS}},
	school = {Georgia Institute of Technology},
	author = {Huning, Alexander J},
	month = may,
	year = {2014},
	file = {Huning - 2014 - A STEADY STATE THERMAL HYDRAULIC ANALYSIS METHOD F.pdf:/home/roberto/Zotero/storage/BZM94PJ8/Huning - 2014 - A STEADY STATE THERMAL HYDRAULIC ANALYSIS METHOD F.pdf:application/pdf}
}

@article{hales_multidimensional_2013,
	title = {Multidimensional multiphysics simulation of {TRISO} particle fuel},
	volume = {443},
	issn = {00223115},
	url = {https://linkinghub.elsevier.com/retrieve/pii/S0022311513009586},
	doi = {10.1016/j.jnucmat.2013.07.070},
	abstract = {Multidimensional multiphysics analysis of TRISO-coated particle fuel using the BISON finite element nuclear fuels code is described. The governing equations and material models applicable to particle fuel and implemented in BISON are outlined. Code verification based on a recent IAEA benchmarking exercise is described, and excellent comparisons are reported. Multiple TRISO-coated particles of increasing geometric complexity are considered. The code’s ability to use the same algorithms and models to solve problems of varying dimensionality from 1D through 3D is demonstrated. The code provides rapid solutions of 1D spherically symmetric and 2D axially symmetric models, and its scalable parallel processing capability allows for solutions of large, complex 3D models. Additionally, the flexibility to easily include new physical and material models and straightforward ability to couple to lower length scale simulations makes BISON a powerful tool for simulation of coated-particle fuel. Future code development activities and potential applications are identified.},
	language = {en},
	number = {1-3},
	urldate = {2020-06-24},
	journal = {Journal of Nuclear Materials},
	author = {Hales, J.D. and Williamson, R.L. and Novascone, S.R. and Perez, D.M. and Spencer, B.W. and Pastore, G.},
	month = nov,
	year = {2013},
	pages = {531--543},
	file = {Hales et al. - 2013 - Multidimensional multiphysics simulation of TRISO .pdf:/home/roberto/Zotero/storage/VFPEIGKR/Hales et al. - 2013 - Multidimensional multiphysics simulation of TRISO .pdf:application/pdf}
}

@techreport{ballinger_balance_2004,
	title = {Balance of {Plant} {System} {Analysis} and {Component} {Design} of {Turbo}-{Machinery} for {High} {Temperature} {Gas} {Reactor} {Systems}},
	url = {http://www.osti.gov/servlets/purl/828709/},
	language = {en},
	number = {828709},
	urldate = {2020-07-17},
	institution = {Massachusetts Institute of Technology, Northern Engineering and Research},
	author = {Ballinger, Ronald G. and Wang, Chun Yun and Kadak, Andrew and Todreas, Neil and Mirick, Bradley and Demetri, Eli and Koronowski, Martin},
	month = aug,
	year = {2004},
	doi = {10.2172/828709},
	pages = {828709},
	file = {Ballinger et al. - 2004 - Balance of Plant System Analysis and Component Des.pdf:/home/roberto/Zotero/storage/SGU4ZH2F/Ballinger et al. - 2004 - Balance of Plant System Analysis and Component Des.pdf:application/pdf}
}

@article{neylan_modular_1988,
	title = {The modular high temperature gas-cooled reactor ({MHTGR}) in the {U}.{S}.},
	volume = {109},
	issn = {00295493},
	url = {https://linkinghub.elsevier.com/retrieve/pii/002954938890146X},
	doi = {10.1016/0029-5493(88)90146-X},
	language = {en},
	number = {1-2},
	urldate = {2020-07-17},
	journal = {Nuclear Engineering and Design},
	author = {Neylan, A.J. and Graf, D.V. and Millunzi, A.C.},
	month = sep,
	year = {1988},
	pages = {99--105},
	file = {Neylan et al. - 1988 - The modular high temperature gas-cooled reactor (M.pdf:/home/roberto/Zotero/storage/I7J4LXZZ/Neylan et al. - 1988 - The modular high temperature gas-cooled reactor (M.pdf:application/pdf}
}

@article{herranz_power_2009,
	title = {Power cycle assessment of nuclear high temperature gas-cooled reactors},
	volume = {29},
	issn = {13594311},
	url = {https://linkinghub.elsevier.com/retrieve/pii/S1359431108003463},
	doi = {10.1016/j.applthermaleng.2008.08.006},
	abstract = {This century power engineering is facing up to one of the greatest challenges ever posed to humankind: the achievement of a sustainable energy system. In order to respond to this challenge, Nuclear Technology is designing a new generation of power plants termed Generation IV, among them High Temperature Gas-cooled Reactors stand out for their potential capability to achieve an excellent thermal performance. This paper investigates the thermal and economic performance of several direct Brayton cycle configurations that could be used in future HTGRs, with special attention to the effects of intercooling and reheating. Among the hypotheses and assumptions taken, the adoption of the PBMR reactor parameters and settings as a reference is particularly important. All inter-cooled layouts have shown thermal efficiencies near or even higher than 50\%, which means a substantial improvement with respect to non-intercooled baselines with no economic penalties. Reheating has been shown not to affect remarkably the thermal or economic plant performance under base-load operation, but it provides the plant with such a flexibility that allows its operation under the “load-follow” regime without heavily taxing the thermal or economic performance. Anyway, use of a multiple axes configuration instead of a single one seems to worsen plant economics and not to entail any thermal benefit.},
	language = {en},
	number = {8-9},
	urldate = {2020-06-24},
	journal = {Applied Thermal Engineering},
	author = {Herranz, L.E. and Linares, J.I. and Moratilla, B.Y.},
	month = jun,
	year = {2009},
	pages = {1759--1765},
	file = {Herranz et al. - 2009 - Power cycle assessment of nuclear high temperature.pdf:/home/roberto/Zotero/storage/RBKXUSL5/Herranz et al. - 2009 - Power cycle assessment of nuclear high temperature.pdf:application/pdf}
}

@incollection{breeze_nuclear_2014,
	title = {Nuclear {Power}},
	isbn = {978-0-08-098330-1},
	url = {https://linkinghub.elsevier.com/retrieve/pii/B978008098330100017X},
	language = {en},
	urldate = {2020-07-17},
	booktitle = {Power {Generation} {Technologies}},
	publisher = {Elsevier},
	author = {Breeze, Paul},
	year = {2014},
	doi = {10.1016/B978-0-08-098330-1.00017-X},
	pages = {353--378},
	file = {Breeze - 2014 - Nuclear Power.pdf:/home/roberto/Zotero/storage/VP3IH6MR/Breeze - 2014 - Nuclear Power.pdf:application/pdf}
}

@article{silady_licensing_1988,
	title = {The licensing experience of the {Modular} {High}-{Temperature} {Gas}-{Cooled} {Reactor} ({MHTGR})},
	volume = {16},
	issn = {03605442},
	url = {https://linkinghub.elsevier.com/retrieve/pii/036054429190120B},
	doi = {10.1016/0360-5442(91)90120-B},
	language = {en},
	number = {1-2},
	urldate = {2020-07-17},
	journal = {Energy},
	author = {Silady, F.A. and Cunliffe, J.C. and Walker, L.P.},
	month = sep,
	year = {1988},
	pages = {417--424},
	file = {Silady et al. - 1991 - The licensing experience of the Modular High-Tempe.pdf:/home/roberto/Zotero/storage/BEKGWIY8/Silady et al. - 1991 - The licensing experience of the Modular High-Tempe.pdf:application/pdf}
}

@article{rohde_development_2012,
	title = {Development and verification of the coupled {3D} neutron kinetics/thermal-hydraulics code {DYN3D}-{HTR} for the simulation of transients in block-type {HTGR}},
	volume = {251},
	issn = {00295493},
	url = {https://linkinghub.elsevier.com/retrieve/pii/S0029549311008454},
	doi = {10.1016/j.nucengdes.2011.09.051},
	abstract = {DYN3D is a nodal diffusion code for 3D steady-state and transient analysis of Light Water Reactor (LWR) cores with hexagonal or square fuel element geometry. In addition to the neutron kinetics, it comprises of a thermal-hydraulics model for flow in parallel coolant channels. Macroscopic cross section data libraries generated with variation of burn-up, reactor poisons concentrations and thermal-hydraulic feedback parameters are linked to the code. Two-group and multi-groups versions of the code are available.},
	language = {en},
	urldate = {2020-08-18},
	journal = {Nuclear Engineering and Design},
	author = {Rohde, U. and Baier, S. and Duerigen, S. and Fridman, E. and Kliem, S. and Merk, B.},
	month = oct,
	year = {2012},
	pages = {412--422},
	file = {Rohde et al. - 2012 - Development and verification of the coupled 3D neu.pdf:/home/roberto/Zotero/storage/3MM8E6T2/Rohde et al. - 2012 - Development and verification of the coupled 3D neu.pdf:application/pdf}
}

@article{ragusa_consistent_2009,
	title = {Consistent and accurate schemes for coupled neutronics thermal-hydraulics reactor analysis},
	volume = {239},
	issn = {00295493},
	url = {https://linkinghub.elsevier.com/retrieve/pii/S0029549308005645},
	doi = {10.1016/j.nucengdes.2008.11.006},
	abstract = {Conventional coupling paradigms currently used to couple different physics components in reactor analysis problems can be inconsistent in their treatment of the nonlinear terms due to the operator-split (OS) strategies employed. This leads to the usage of small time steps to maintain accuracy requirements, thereby increasing the overall computational time. This paper proposes some remedies to OS techniques that can restore consistency in the coupling of the nonlinear terms and explores high-order mono-block nonlinearly consistent techniques with time step control. The performance of the methods was studied for several transient scenarios using a 0D point-kinetics/thermal-hydraulics lumped model and a 1D neutronics/heat conduction/enthalpy balance model. The results prove that consistent approximations can be made to enhance the overall accuracy in conventional codes with simple nonintrusive techniques. Additionally, an analysis of a mono-block coupling strategy (without having recourse to an OS strategy) is carried out to assess automated time stepping control using higher order Implicit Runge–Kutta (IRK) schemes. The conclusions from these results indicate that nonlinearly consistent adaptive time stepping methods can provide better accuracy and reliability in the solution fields than constant time stepping methods, even for transients with rapid and discontinuous variations.},
	language = {en},
	number = {3},
	urldate = {2020-07-07},
	journal = {Nuclear Engineering and Design},
	author = {Ragusa, Jean C. and Mahadevan, Vijay S.},
	month = mar,
	year = {2009},
	pages = {566--579},
	file = {Ragusa and Mahadevan - 2009 - Consistent and accurate schemes for coupled neutro.pdf:/home/roberto/Zotero/storage/V9JX8K8S/Ragusa and Mahadevan - 2009 - Consistent and accurate schemes for coupled neutro.pdf:application/pdf}
}

@article{park_tightly_2010,
	title = {Tightly {Coupled} {Multiphysics} {Algorithms} for {Pebble} {Bed} {Reactors}},
	volume = {166},
	issn = {0029-5639, 1943-748X},
	url = {https://www.tandfonline.com/doi/full/10.13182/NSE09-104},
	doi = {10.13182/NSE09-104},
	abstract = {We have developed a tightly coupled multiphysics simulation tool for the pebble bed reactor (PBR) concept, a specific type of very high temperature gas-cooled reactor. The simulation tool PRONGHORN takes advantage of the Multiphysics Object-Oriented Simulation Environment library and is capable of solving multidimensional thermal-fluid and neutronics problems implicitly with a Newton-based approach. Expensive Jacobian matrix formation is alleviated via the Jacobian-free Newton-Krylov method, and physics-based preconditioning is applied to minimize Krylov iterations. Motivation for the work is provided via analysis and numerical experiments on simpler multiphysics reactor models. We then provide detail of the physical models and numerical methods in PRONGHORN. Finally, PRONGHORN’s algorithmic capability is demonstrated on a number of PBR test cases.},
	language = {en},
	number = {2},
	urldate = {2020-07-07},
	journal = {Nuclear Science and Engineering},
	author = {Park, H. and Knoll, D. A. and Gaston, D. R. and Martineau, R. C.},
	month = oct,
	year = {2010},
	pages = {118--133},
	file = {Park et al. - 2010 - Tightly Coupled Multiphysics Algorithms for Pebble.pdf:/home/roberto/Zotero/storage/CGFNBVDA/Park et al. - 2010 - Tightly Coupled Multiphysics Algorithms for Pebble.pdf:application/pdf}
}

@article{bostelmann_criticality_2016,
	title = {Criticality calculations of the {Very} {High} {Temperature} {Reactor} {Critical} {Assembly} benchmark with {Serpent} and {SCALE}/{KENO}-{VI}},
	volume = {90},
	issn = {03064549},
	url = {https://linkinghub.elsevier.com/retrieve/pii/S0306454915300177},
	doi = {10.1016/j.anucene.2015.12.008},
	abstract = {Within the framework of the IAEA Coordinated Research Project on HTGR Uncertainty Analysis in Modeling, criticality calculations of the Very High Temperature Critical Assembly experiment were performed as the validation reference to the prismatic MHTGR-350 lattice calculations. Criticality measurements performed at several temperature points at this Japanese graphitemoderated facility were recently included in the International Handbook of Evaluated Reactor Physics Benchmark Experiments, and represent one of the few data sets available for the validation of HTGR lattice physics. This work compares VHTRC criticality simulations utilizing the Monte Carlo codes Serpent and SCALE/KENO-VI. Reasonable agreement was found between Serpent and KENO-VI, but only the use of the latest ENDF cross section library release, namely the ENDF/B-VII.1 library, led to an improved match with the measured data. Furthermore, the fourth beta release of SCALE 6.2/KENO-VI showed significant improvements from the current SCALE 6.1.2 version, compared to the experimental values and Serpent.},
	language = {en},
	urldate = {2020-07-16},
	journal = {Annals of Nuclear Energy},
	author = {Bostelmann, Friederike and Hammer, Hans R. and Ortensi, Javier and Strydom, Gerhard and Velkov, Kiril and Zwermann, Winfried},
	month = apr,
	year = {2016},
	pages = {343--352},
	file = {Bostelmann et al. - 2016 - Criticality calculations of the Very High Temperat.pdf:/home/roberto/Zotero/storage/AVYQ7EU2/Bostelmann et al. - 2016 - Criticality calculations of the Very High Temperat.pdf:application/pdf}
}

@article{tak_numerical_2008,
	title = {Numerical investigation of a heat transfer within the prismatic fuel assembly of a very high temperature reactor},
	volume = {35},
	issn = {03064549},
	url = {https://linkinghub.elsevier.com/retrieve/pii/S0306454908001114},
	doi = {10.1016/j.anucene.2008.04.005},
	abstract = {The complex geometry of the hexagonal fuel blocks of the prismatic fuel assembly in a very high temperature reactor (VHTR) hinders accurate evaluations of the temperature profile within the fuel assembly without elaborate numerical calculations. Therefore, simplified models such as a unit cell model have been widely applied for the analyses and designs of prismatic VHTRs since they have been considered as effective approaches reducing the computational efforts. In a prismatic VHTR, however, the simplified models cannot consider a heat transfer within a fuel assembly as well as a coolant flow through a bypass gap between the fuel assemblies, which may significantly affect the maximum fuel temperature. In this paper, a three-dimensional computational fluid dynamics (CFD) analysis has been carried out on a typical fuel assembly of a prismatic VHTR. Thermal behaviours and heat transfer within the fuel assembly are intensively investigated using the CFD solutions. In addition, the accuracy of the unit cell approach is assessed against the CFD solutions. Two example situations are illustrated to demonstrate the deficiency of the unit cell model caused by neglecting the effects of the bypass gap flow and the radial power distribution within the fuel assembly.},
	language = {en},
	number = {10},
	urldate = {2020-07-28},
	journal = {Annals of Nuclear Energy},
	author = {Tak, Nam-il and Kim, Min-Hwan and Lee, Won Jae},
	month = oct,
	year = {2008},
	pages = {1892--1899},
	file = {Tak et al. - 2008 - Numerical investigation of a heat transfer within .pdf:/home/roberto/Zotero/storage/9XJJPRBG/Tak et al. - 2008 - Numerical investigation of a heat transfer within .pdf:application/pdf}
}

@article{gaston_moose_2009,
	title = {{MOOSE}: {A} parallel computational framework for coupled systems of nonlinear equations},
	volume = {239},
	issn = {00295493},
	shorttitle = {{MOOSE}},
	url = {https://linkinghub.elsevier.com/retrieve/pii/S0029549309002635},
	doi = {10.1016/j.nucengdes.2009.05.021},
	abstract = {Systems of coupled, nonlinear partial differential equations often arise in simulation of nuclear processes. MOOSE: Multiphysics Object Oriented Simulation Environment, a parallel computational framework targeted at solving these systems, is presented. As opposed to traditional data-flow oriented computational frameworks, MOOSE is founded on the mathematical principle of Jacobian-free Newton-Krylov (JFNK) solution methods. Utilizing the mathematical structure present in JFNK, physics are modularized into “Kernels” allowing for rapid production of new simulation tools. In addition, systems are solved fully coupled and fully implicit employing physics based preconditioning which allows for great flexibility even with large variance in time scales. A summary of the mathematics, an inspection of the structure of MOOSE, and several representative solutions from applications built on the framework are presented.},
	language = {en},
	number = {10},
	urldate = {2020-07-08},
	journal = {Nuclear Engineering and Design},
	author = {Gaston, Derek and Newman, Chris and Hansen, Glen and Lebrun-Grandié, Damien},
	month = oct,
	year = {2009},
	pages = {1768--1778},
	file = {Gaston et al. - 2009 - MOOSE A parallel computational framework for coup.pdf:/home/roberto/Zotero/storage/34CJNJ9N/Gaston et al. - 2009 - MOOSE A parallel computational framework for coup.pdf:application/pdf}
}

@article{lindsay_introduction_2018,
	title = {Introduction to {Moltres}: {An} application for simulation of {Molten} {Salt} {Reactors}},
	volume = {114},
	issn = {03064549},
	shorttitle = {Introduction to {Moltres}},
	url = {https://linkinghub.elsevier.com/retrieve/pii/S0306454917304760},
	doi = {10.1016/j.anucene.2017.12.025},
	abstract = {Moltres is a new physics application for modeling coupled physics in fluid-fuelled, molten salt reactors. This paper describes its neutronics model, thermal hydraulics model, and their coupling in the MOOSE framework. Neutron and precursor equations are implemented using an action system that allows use of an arbitrary number of groups with no change in the input card. Results for many-channel configurations in 2D-axisymmetric and 3D coordinates are presented and compared against other coupled models as well as the Molten Salt Reactor Experiment.},
	language = {en},
	urldate = {2020-06-26},
	journal = {Annals of Nuclear Energy},
	author = {Lindsay, Alexander and Ridley, Gavin and Rykhlevskii, Andrei and Huff, Kathryn},
	month = apr,
	year = {2018},
	pages = {530--540},
	file = {Lindsay et al. - 2018 - Introduction to Moltres An application for simula.pdf:/home/roberto/Zotero/storage/2CFNUVFZ/Lindsay et al. - 2018 - Introduction to Moltres An application for simula.pdf:application/pdf}
}

@inproceedings{paviet-hartmann_analysis_2011,
	address = {Chiba, Japan},
	title = {Analysis of {Nuclear} {Proliferation} {Resistance} {Reprocessing} and {Recycling} {Technologies}},
	booktitle = {19th {International} {Conference} on {Nuclear} {Engineering}},
	publisher = {Japan Society of Mechanical Engineers, Tokyo (Japan)},
	author = {Paviet-Hartmann, Patricia and Cerefice, Gary and Riveros Stacey, Marcela and Bakhtiar, Steven},
	month = jul,
	year = {2011},
	file = {Paviet-Hartmann et al. - 2011 - Analysis of Nuclear Proliferation Resistance Repro.pdf:/home/roberto/Zotero/storage/ULU4TUU4/Paviet-Hartmann et al. - 2011 - Analysis of Nuclear Proliferation Resistance Repro.pdf:application/pdf}
}

@article{no_review_2007,
	title = {A {Review} of {Helium} {Gas} {Turbine} {Technology} for {High}-{Temperature} {Gas}-{Cooled} {Rreactors}},
	volume = {39},
	issn = {1738-5733},
	url = {https://inis.iaea.org/search/search.aspx?orig_q=RN:39068303},
	abstract = {Current High-Temperature Gas-cooled Reactors (HTGRs) are based on a closed brayton cycle with helium gas as the working fluid. Thermodynamic performance of the axial-flow helium gas turbines is of critical concern as it considerably affects the overall cycle efficiency. Helium gas turbines pose some design challenges compared to steam or air turbomachinery because of the physical properties of helium and the uniqueness of the operating conditions at high pressure with low pressure ratio. This report present a review of the helium Brayton cycle experiences in Germany and in Japan. The design and availability of helium gas turbines for HTGR are also presented in this study. We have developed a new throughflow calculation code to calculate the design-point performance of helium gas turbines. Use of the method has been illustrated by applying it to the GTHTR300 reference.},
	language = {English},
	journal = {Nuclear Engineering and Technology},
	author = {No, Hee Cheon and Kim, Ji Hwan and Kim, Hyeun Min},
	month = feb,
	year = {2007},
	file = {No et al. - 2007 - A REVIEW OF HELIUM GAS TURBINE TECHNOLOGY FOR HIGH.pdf:/home/roberto/Zotero/storage/5QDQZ4YY/No et al. - 2007 - A REVIEW OF HELIUM GAS TURBINE TECHNOLOGY FOR HIGH.pdf:application/pdf}
}

@article{doenitz_hydrogen_1980,
	title = {{Hydrogen} {Production} {by} {High} {Temperature} {Electrolysis} {of} {Water} {Vapour}},
	volume = {5},
	url = {https://doi.org/10.1016/0360-3199(80)90114-7},
	language = {en},
	journal = {Internation Journal of Hydrogen Energy},
	author = {Doenitz, W and Schmidberger, R and Steinheil, E},
	year = {1980},
	pages = {55--63},
	file = {Doenitz et al. - HYDROGEN PRODUCTION BY HIGH TEMPERATURE ELECTROLYS.pdf:/home/roberto/Zotero/storage/GGLWUY7Z/Doenitz et al. - HYDROGEN PRODUCTION BY HIGH TEMPERATURE ELECTROLYS.pdf:application/pdf}
}

@misc{gif_gif_2019,
	title = {{GIF} {Portal} - {Home} - {VHTR}},
	copyright = {2019 Generation IV International Forum},
	note = {https://www.gen-4.org/gif/},
	language = {en},
	urldate = {2020-10-29},
	journal = {Very-High-Temperature Reactor (VHTR)},
	author = {GIF},
	year = {2019},
	file = {GIF Portal - Home - VHTR:/home/roberto/Zotero/storage/3C2H9RZE/vhtr.html:text/html}
}

@article{kadak_status_2016,
	title = {The {Status} of the {US} {High}-{Temperature} {Gas} {Reactors}},
	volume = {2},
	issn = {20958099},
	url = {https://linkinghub.elsevier.com/retrieve/pii/S2095809916301564},
	doi = {10.1016/J.ENG.2016.01.026},
	language = {en},
	number = {1},
	urldate = {2020-11-14},
	journal = {Engineering},
	author = {Kadak, Andrew C.},
	month = mar,
	year = {2016},
	pages = {119--123},
	file = {Kadak - 2016 - The Status of the US High-Temperature Gas Reactors.pdf:/home/roberto/Zotero/storage/WUGL8NPH/Kadak - 2016 - The Status of the US High-Temperature Gas Reactors.pdf:application/pdf}
}

@article{shimizu_operation_2014,
	title = {Operation and maintenance experience from the {HTTR} database},
	volume = {51},
	issn = {0022-3131, 1881-1248},
	url = {http://www.tandfonline.com/doi/abs/10.1080/00223131.2014.946568},
	doi = {10.1080/00223131.2014.946568},
	language = {en},
	number = {11-12},
	urldate = {2020-11-14},
	journal = {Journal of Nuclear Science and Technology},
	author = {Shimizu, Atsushi and Furusawa, Takayuki and Homma, Fumitaka and Inoi, Hiroyuki and Umeda, Masayuki and Kondo, Masaaki and Isozaki, Minoru and Fujimoto, Nozomu and Iyoku, Tatsuo},
	month = dec,
	year = {2014},
	pages = {1444--1451},
	file = {Shimizu et al. - 2014 - Operation and maintenance experience from the HTTR.pdf:/home/roberto/Zotero/storage/CWTBEPTU/Shimizu et al. - 2014 - Operation and maintenance experience from the HTTR.pdf:application/pdf}
}

@misc{world_nuclear_news_micro_2020,
	title = {Micro modular reactors proposed for {Idaho} and {Illinois} : {New} {Nuclear} - {World} {Nuclear} {News}},
	note = {https://world-nuclear-news.org/Articles/Micro-modular-reactors-proposed-for-Idaho-and-Illi},
	urldate = {2020-11-14},
	author = {World Nuclear News},
	month = oct,
	year = {2020},
	file = {Micro modular reactors proposed for Idaho and Illinois \: New Nuclear - World Nuclear News:/home/roberto/Zotero/storage/RW5EDQBK/Micro-modular-reactors-proposed-for-Idaho-and-Illi.html:text/html}
}

@techreport{global_first_power_project_2019,
	address = {Deep River, ON, Canada},
	type = {Project {Description}},
	title = {Project {Description} for the {Micro} {Modular} {Reactor} {Project} at {Chalk} {River}},
	copyright = {Unrestricted},
	url = {https://ceaa-acee.gc.ca/050/evaluations/proj/80182},
	abstract = {Home page for the environmental assessment of the project - Micro Modular Reactor Project at Chalk River},
	language = {eng},
	number = {CRP-LIC-01-001r2},
	urldate = {2019-07-29},
	institution = {Global First Power},
	author = {{Global First Power}},
	month = jul,
	year = {2019},
	note = {https://ceaa-acee.gc.ca/050/documents/p80182/130911E.pdf},
	pages = {1--52}
}

@incollection{barre_gas-cooled_2010,
	address = {Boston, MA},
	title = {Gas-{Cooled} {Reactors}},
	isbn = {978-0-387-98149-9},
	url = {https://doi.org/10.1007/978-0-387-98149-9_22},
	abstract = {This chapter describes the various families of reactors in which the primary fluid cooling the core is a gas, usually carbon dioxide or helium.},
	booktitle = {Handbook of {Nuclear} {Engineering}},
	publisher = {Springer US},
	author = {Barré, Bertrand},
	editor = {Cacuci, Dan Gabriel},
	year = {2010},
	doi = {10.1007/978-0-387-98149-9_22},
	pages = {2711--2748}
}

@techreport{williams_draft_1989,
	address = {Washington, DC},
	title = {Draft {Preapplication} {Safety} {Evaluation} {Report} for the {Modular} {High}-{Temperature} {Gas}-{Cooled} {Reactor}},
	url = {https://www.nrc.gov/docs/ML0527/ML052780497.pdf},
	number = {NUREG-1338},
	institution = {US NRC Office of Nuclear Regulatory Research},
	author = {Williams, P.M. and King, T.L. and Wilson, J.N.},
	year = {1989}
}

% CHAPTER 2: LITERATURE REVIEW Neutronics
@article{lewis_finite_1986,
	title = {Finite element, nodal and response matrix methods: {A} variational synthesis for neutron transport},
	volume = {18},
	issn = {01491970},
	shorttitle = {Finite element, nodal and response matrix methods},
	url = {https://linkinghub.elsevier.com/retrieve/pii/0149197086900132},
	doi = {10.1016/0149-1970(86)90013-2},
	abstract = {The variational principle used to derive finite element approximations to the even-parity neutron transport equation is modified to include the odd parity flux as a Lagrange multiplier along the interelement boundaries. The result is a functional that guarantees nodal balance over each element or node, regardless of the form of the space-angle trial functionsthat are used. Ritzprocedures are employedto yieldnodal methods that are analogous to those usedextensivelyin diffusiontheory, but without the need for ad-hoc node-to-node interpolation of transverse leakage terms. With spherical harmonics, nodal transport methods are obtained with consistent angular approximations within the nodes and across the node interfaces.The resulting equations are in response matrix form, and from their solution the detailed flux distribution as well as the node-averaged values can be obtained.},
	language = {en},
	number = {1-2},
	urldate = {2020-07-11},
	journal = {Progress in Nuclear Energy},
	author = {Lewis, E.E. and Dilber, I.},
	month = jan,
	year = {1986},
	pages = {63--74},
	file = {Lewis and Dilber - 1986 - Finite element, nodal and response matrix methods.pdf:/home/roberto/Zotero/storage/ZMLKPAWD/Lewis and Dilber - 1986 - Finite element, nodal and response matrix methods.pdf:application/pdf}
}

@article{lawrence_progress_1986,
	title = {Progress in nodal methods for the solution of the neutron diffusion and transport equations},
	volume = {17},
	issn = {01491970},
	url = {https://linkinghub.elsevier.com/retrieve/pii/014919708690034X},
	doi = {10.1016/0149-1970(86)90034-X},
	abstract = {Recent progress in the development of coarse-mesh nodal methods for the numerical solution of the neutron diffusionand transport equations is reviewed.In contrast with earlier nodal simulators, more recent nodal diffusionmethods are characterized by the systematicderivation of spatial coupling relationships that are entirely consistent with the multigroup diffusion equation. These relationships most often are derived by developing approximations to the one-dimensionalequations obtained by integrating the multidimensionaldiffusionequation over directions transverse to each coordinate axis. Both polynomial and analytic approaches to the solution of the transverse-integrated equations are discussed, and the Cartesian-geometry polynomial approach is derived in a manner which motivates the extension of this formulation to the solution of the diffusionequation in hexagonal geometry. Iterative procedures developed for the solution of the nodal equations are discussed briefly, and numerical comparisons for representative three-dimensional benchmark problems are given. The application of similar ideas to the neutron transport equation has led to the development of coarse-mesh transport schemes that combine nodal spatial approximations with angular representations based on either the standard discrete-ordinate approximation or double P, expansions of the angular dependenceof the fluxeson the surfaces of the nodes. The former methods yield improved difference approximations to the multidimensional discrete-ordinates equations, while the latter approach leads to equations similar to those obtained in interfacecurrent nodal-diffusionformulations. The relativeeflicienciesof these two approaches are discussed,and directions for future work are indicated.},
	language = {en},
	number = {3},
	urldate = {2020-07-07},
	journal = {Progress in Nuclear Energy},
	author = {Lawrence, R.D.},
	month = jan,
	year = {1986},
	pages = {271--301},
	file = {Lawrence - 1986 - Progress in nodal methods for the solution of the .pdf:/home/roberto/Zotero/storage/Y6WL9MPL/Lawrence - 1986 - Progress in nodal methods for the solution of the .pdf:application/pdf}
}

@inproceedings{duracz_nodal_1981,
	address = {Munich, Germany},
	title = {A {Nodal} {Method} in {Hexagonal} {Geometry}},
	booktitle = {International {Meeting} on {Advances} in {Mathematical} {Methods} for {Solution} of {Nuclear} {Engineering} {Problems}},
	author = {Duracz, T},
	month = apr,
	year = {1981}
}

@article{wagner_three-dimensional_1989,
	title = {Three-{Dimensional} {Nodal} {Diffusion} and {Transport} {Theory} {Methods} for {Hexagonal}- \textit{z} {Geometry}},
	volume = {103},
	issn = {0029-5639, 1943-748X},
	url = {https://www.tandfonline.com/doi/full/10.13182/NSE89-A23690},
	doi = {10.13182/NSE89-A23690},
	abstract = {Advanced nodal methods for the solution of the multigroup neutron diffusion and transport theory equations in three-dimensional hexagonal-z geometry are described. The code HEXNOD allows an accurate and efficient calculation of three-dimensional problems for fast reactors and high converter light water reactors. A unique capability of HEXNOD is the accurate solution of global three-dimensional neutron transport problems for fast reactors with very small computing times. The accuracy of the nodal diffusion and transport approximations is demonstrated by comparison with conventional finite difference methods and Monte Carlo calculations for a number of mathematical benchmark problems. Based on numerical results, it is concluded that the code HEXNOD is well suited for three-dimensional routine analysis of fast reactors and, in particular, as the neutronics module of the generalized quasi-static kinetics program HEXNODYN, which is currently being developed as part of the European accident code EAC-2.},
	language = {en},
	number = {4},
	urldate = {2020-07-08},
	journal = {Nuclear Science and Engineering},
	author = {Wagner, M. R.},
	month = dec,
	year = {1989},
	pages = {377--391},
	file = {Wagner - 1989 - Three-Dimensional Nodal Diffusion and Transport Th.pdf:/home/roberto/Zotero/storage/V62SR2L2/Wagner - 1989 - Three-Dimensional Nodal Diffusion and Transport Th.pdf:application/pdf}
}

@techreport{delp_flare_1964,
	address = {San Jose, CA},
	type = {Technical {Report}},
	title = {{FLARE}, {A} three-dimensional boiling water reactor simulatior},
	url = {https://www.osti.gov/biblio/4677068-flare-three-dimensional-boiling-water-reactor-simulator},
	number = {GEAP-4598},
	institution = {General Electric Co. Atomic Power Equipment Dept.},
	author = {Delp, D.L. and Fischer, D.L. and Harriman, J.M. and Stedwell, M.J.},
	month = jul,
	year = {1964}
}

@article{gupta_nodal_1981,
	title = {{NODAL} {METHODS} {FOR} {THREE}-{DIMENSIONAL} {SIMULATORS}},
	volume = {7},
	language = {en},
	journal = {Progress in Nuclear Energy},
	author = {Gupta, N K},
	year = {1981},
	pages = {127--149},
	file = {Gupta - NODAL METHODS FOR THREE-DIMENSIONAL SIMULATORS.pdf:/home/roberto/Zotero/storage/VUMFY6GI/Gupta - NODAL METHODS FOR THREE-DIMENSIONAL SIMULATORS.pdf:application/pdf}
}

@inproceedings{finnermann_new_1975,
	address = {Bologna, Italy},
	title = {A {New} {Computational} {Technique} for the {Solution} of {Multidimensional} {Neutron} {Diffusion} {Problems}},
	author = {Finnermann, H and Wagner, M. R.},
	month = nov,
	year = {1975}
}

@misc{finnemann_atomkernenergie_1977,
	title = {Atomkernenergie},
	author = {Finnemann, H},
	year = {1977}
}

@inproceedings{fitzpatrick_hexpedite_1992,
	address = {Chicago, IL},
	title = {{HEXPEDITE}: {A} {Net} {Current} {Multigroup} {Nodal} {Diffusion} {Method} for {Hexagonal}-z {Geometry}},
	isbn = {0003-018X},
	url = {https://www.osti.gov/biblio/6653947-hexpedite-net-current-multigroup-nodal-diffusion-method-hexagonal-geometry},
	language = {en},
	booktitle = {Joint {American} {Nuclear} {Society} ({ANS})/{European} {Nuclear} {Society} ({ENS}) international meeting on fifty years of controlled nuclear chain reaction: past, present, and future},
	publisher = {Transactions of the American Nuclear Society},
	author = {Fitzpatrick, W. E. and Ougouag, A. M.},
	month = nov,
	year = {1992}
}

@phdthesis{fitzpatrick_developments_1995,
	address = {Urbana, IL},
	title = {Developments in {Nodal} {Reactor} {Analysis} {Tools} for {Hexagonal} {Geometry}},
	school = {University of Illinois at Urbana-Champaign},
	author = {FItzpatrick, William},
	year = {1995}
}

@techreport{ortensi_deterministic_2010,
	title = {Deterministic {Modeling} of the {High} {Temperature} {Test} {Reactor}},
	url = {http://www.osti.gov/servlets/purl/989875-x5nHFd/},
	language = {en},
	number = {INL/EXT-10-18969, 989875},
	urldate = {2020-07-07},
	author = {Ortensi, J. and Cogliati, J. J. and Pope, M. A. and Ferrer, R. M. and Ougouag, A. M.},
	month = jun,
	year = {2010},
	doi = {10.2172/989875},
	pages = {INL/EXT--10--18969, 989875},
	file = {Ortensi et al. - 2010 - Deterministic Modeling of the High Temperature Tes.pdf:/home/roberto/Zotero/storage/387YBI78/Ortensi et al. - 2010 - Deterministic Modeling of the High Temperature Tes.pdf:application/pdf}
}

@inproceedings{ortensi_deterministic_2010-1,
	address = {Prague, Czech Republic},
	title = {Deterministic {Modeling} of the {High} {Temperature} {Test} {Reactor} with {DRAGON}-{HEXPEDITE}},
	booktitle = {Proceedings of {HTR} 2010},
	author = {Ortensi, J and Cogliati, J.J. and Pope, M.A. and Bess, J.D. and Ferrer, R.M. and Binghman, A. and Ougouag, A.M.},
	month = oct,
	year = {2010}
}

@book{methani_evaluation_2003,
	title = {Evaluation of high temperature gas cooled reactor performance: benchmark analysis related to initial testing of the {HTTR} and {HTR}-10},
	isbn = {978-92-0-116203-8},
	shorttitle = {Evaluation of high temperature gas cooled reactor performance},
	language = {en},
	author = {Methani, M and {International Atomic Energy Agency. IAEA}},
	year = {2003},
	note = {OCLC: 1039268451},
	file = {Methani and International Atomic Energy Agency. IAEA - 2003 - Evaluation of high temperature gas cooled reactor .pdf:/home/roberto/Zotero/storage/TCEWWPWD/Methani and International Atomic Energy Agency. IAEA - 2003 - Evaluation of high temperature gas cooled reactor .pdf:application/pdf}
}

@techreport{lawrence_dif3d_1983,
	address = {Lemont, IL},
	title = {The {DIF3D} {Nodal} {Neutronics} {Option} for {Two}- and {Three}-{Dimensional} {Diffusion}-{Theory} {Calculations} in {Hexagonal} {Geometry}},
	url = {https://inis.iaea.org/collection/NCLCollectionStore/_Public/17/069/17069657.pdf},
	language = {en},
	number = {ANL-83-1 ON: DE83011019},
	institution = {Argonne National Laboratory (ANL)},
	author = {Lawrence, R},
	month = mar,
	year = {1983},
	file = {Lawrence - 1983 - The DIF3D Nodal Neutronics Option for Two- and Thr.pdf:/home/roberto/Zotero/storage/6XIEGTS3/Lawrence - 1983 - The DIF3D Nodal Neutronics Option for Two- and Thr.pdf:application/pdf}
}

@misc{palmiotti_variant_1995,
	title = {{VARIANT}: {VARIational} {Anisotropic} {Nodal} {Transport} for {Multidimensional} {Cartesian} and {Hexagoanl} {Geometry} {Calculation}},
	author = {Palmiotti, G. and Lewis, E.E. and Carrico, C.B.},
	month = oct,
	year = {1995}
}

@techreport{downar_parcs_2004,
	address = {W. Lafayette, Indiana},
	title = {{PARCS} v2.6 {US} {NRC} {Core} {Neutronics} {Simulator} {THEORY} {MANUAL}},
	url = {https://engineering.purdue.edu/PARCS/Code/Manual/Theory/PDF/PARCS_TheoryManual.pdf},
	institution = {School of Nuclear Engineering Purdue University},
	author = {Downar, T and Lee, D and Xu, Y and Kozlowski, T},
	month = jun,
	year = {2004}
}

@article{wang_modified_2018,
	title = {A modified, hybrid nodal-integral/finite-element method for {3D} convection-diffusion problems in arbitrary geometries},
	volume = {122},
	issn = {00179310},
	url = {https://linkinghub.elsevier.com/retrieve/pii/S0017931017312802},
	doi = {10.1016/j.ijheatmasstransfer.2018.01.087},
	abstract = {A modified, hybrid nodal-integral/finite-element method (NI-FEM) is developed to solve the threedimensional (3D), steady-state, convection-diffusion problems in arbitrary geometries. The hybrid NI-FEM takes advantage of the high efficiency of the conventional nodal-integral method (NIM) and the flexible mesh generation of the finite element method (FEM), which is applicable to arbitrary geometries. In this method, the computational domain is discretized into, for 3D problems, four-node tetrahedral elements and eight-node cuboid elements. The cuboid elements are used to discretize the interior region and regions adjacent to boundaries that are parallel to planes formed by two of the axes, while the tetrahedral elements are used to discretize the remaining irregular regions. The conventional NIM is used to develop difference equations for the transverse-averaged variables on the interface between two adjacent cuboid elements. The FEM is applied to develop algebraic equations for the node temperatures of tetrahedral elements. On the interface between two different kinds of elements, the transverseaveraged variable for the cuboid element is obtained by averaging the node values of its adjacent tetrahedral elements, while the heat flux for the tetrahedral element is calculated using the corresponding transverse-averaged variables of its adjacent cuboid element. The hybrid NI-FEM is developed to be solved using a matrix formulation for the entire domain rather than an iterative procedure, detailed derivations of which for the 3D case are presented in this paper. Using the NI-FEM developed here, 2D and 3D convection-diffusion test problems are solved, and the numerical results are compared to the exact (manufactured) solutions to determine the order, accuracy and efficiency of the method. Numerical scheme is found to be of, as expected, second order.},
	language = {en},
	urldate = {2020-08-27},
	journal = {International Journal of Heat and Mass Transfer},
	author = {Wang, Pengfei and {Rizwan-uddin}},
	month = jul,
	year = {2018},
	pages = {99--116},
	file = {Wang and Rizwan-uddin - 2018 - A modified, hybrid nodal-integralfinite-element m.pdf:/home/roberto/Zotero/storage/66ZJ6UY4/Wang and Rizwan-uddin - 2018 - A modified, hybrid nodal-integralfinite-element m.pdf:application/pdf}
}

@techreport{lee_status_2006,
	address = {Lemont, IL},
	title = {Status of {Reactor} {Physics} {Activities} on {Cross} {Section} {Generation} and {Functionalization} for the {Prismatic} {Very} {High} {Temperature} {Reactor}, and {Development} of {Spatially}-{Heterogeneous} {Codes}},
	url = {https://publications.anl.gov/anlpubs/2006/09/57340.pdf},
	number = {ANL-GenIV-075},
	institution = {ANL},
	author = {Lee and Zhong and Taiwo and Yang and Smith and Palmiotti},
	month = aug,
	year = {2006}
}

@article{kang_finite_1973,
	title = {Finite {Element} {Methods} for {Reactor} {Analysis}},
	volume = {51},
	issn = {0029-5639, 1943-748X},
	url = {https://www.tandfonline.com/doi/full/10.13182/NSE73-A23278},
	doi = {10.13182/NSE73-A23278},
	language = {en},
	number = {4},
	urldate = {2020-08-31},
	journal = {Nuclear Science and Engineering},
	author = {Kang, C. M. and Hansen, K. F.},
	month = aug,
	year = {1973},
	pages = {456--495},
	file = {Kang and Hansen - 1973 - Finite Element Methods for Reactor Analysis.pdf:/home/roberto/Zotero/storage/VDZTILBS/Kang and Hansen - 1973 - Finite Element Methods for Reactor Analysis.pdf:application/pdf}
}

@article{cavdar_finite_2004,
	title = {A finite element/boundary element hybrid method for 2-{D} neutron diffusion calculations},
	volume = {31},
	issn = {03064549},
	url = {https://linkinghub.elsevier.com/retrieve/pii/S0306454904000799},
	doi = {10.1016/j.anucene.2004.04.006},
	abstract = {A finite element-boundary element hybrid method has been developed for one or two group neutron diffusion calculations. A linear or bilinear finite element formulation for the reactor core and a linear boundary element technique for the reflector which are combined through interface continuity conditions constitute the basis of the developed method. The present formulation is restricted to two-dimensional geometries and has been implemented in the developed computer program. Via comparisons with analytical solutions, the proposed method has been validated. Further comparisons against the pure finite and boundary element formulations show that the proposed method constitutes a viable alternative for the numerical solution of neutron diffusion problems of both the external neutron source and multiplication eigenvalue determination variety.},
	language = {en},
	number = {14},
	urldate = {2020-07-11},
	journal = {Annals of Nuclear Energy},
	author = {Cavdar, S and Ozgener, H.A},
	month = sep,
	year = {2004},
	pages = {1555--1582}
}

@incollection{lewis_finite_1981,
	address = {Boston, MA},
	title = {Finite {Element} {Approximation} to the {Even}-{Parity} {Transport} {Equation}},
	isbn = {978-1-4613-9919-3},
	url = {https://doi.org/10.1007/978-1-4613-9919-3_3},
	abstract = {The finite element method is a procedure for reducing partial differential equations to sets of simultaneous algebraic equations suitable for solution on a digital computer. The method consists of several steps. The differential equation first is cast into the form of a variational principle, and the domain of the resulting functional is partitioned into a number of subdomains or finite elements. The dependent variable then is approximated within each element by a simple polynomial, and these are linked across inter-element boundaries by appropriate continuity conditions. Such an approximation may be represented as a sum of piecewise polynomial trial functions with unknown coefficients that typically correspond to the values of the dependent variable and possibly its derivatives at mesh points defined by the finite element structure. The requirement that the functional be stationary with respect to variations in these coefficients then leads to the desired set of simultaneous equations.},
	booktitle = {Advances in {Nuclear} {Science} and {Technology}},
	publisher = {Springer US},
	author = {Lewis, E. E.},
	editor = {Lewins, Jeffery and Becker, Martin},
	year = {1981},
	doi = {10.1007/978-1-4613-9919-3_3},
	pages = {155--225}
}

@inproceedings{lee_development_2008,
	address = {Washington, DC USA},
	title = {{DEVELOPMENT} {OF} {CAPP} {CODE} {BASED} {ON} {THE} {FINITE} {ELEMENT} {METHOD} {FOR} {THE} {ANALYSIS} {OF} {VHTR} {CORES}},
	booktitle = {Proceedings of the 4th {International} {Topical} {Meeting} on {High} {Temperature} {Reactor} {Technology}},
	author = {Lee, H.C. and Jo, C.K. and Noh, J.M.},
	month = oct,
	year = {2008}
}

@inproceedings{lee_development_2011,
	address = {Taebaek, Korea},
	title = {Development of the {CAPP} code for the {Analysis} of {Block} {Type} {VHTRs}},
	booktitle = {Transactions of the {Korean} {Nuclear} {Society} {Spring} {Meeting}},
	author = {Lee, Hyun Chul and Jo, Chag Keun and Noh, Jae Man},
	month = may,
	year = {2011},
	file = {Lee et al. - 2011 - Development of the CAPP code for the Analysis of B.pdf:/home/roberto/Zotero/storage/TG97L98Q/Lee et al. - 2011 - Development of the CAPP code for the Analysis of B.pdf:application/pdf}
}

@inproceedings{casal_helios_1998,
	address = {Pittsburg, PA},
	title = {{HELIOS}: {Geometric} {Capabilities} of a {New} {Fuel}-{Assembly} {Program}},
	volume = {II, Sect. 10.2.1, 1-13},
	booktitle = {Proc. {Int}. {Top}. {Mtg}. {Adv}. {Math}. {Comp}. {Reac}. {Phys}},
	author = {Casal, J.J. and Stamml'er, R.J.J. and Villarino, E.A. and Ferri, A.A.},
	year = {1998}
}

@article{tak_cappgamma_2016,
	title = {{CAPP}/{GAMMA}+ code system for coupled neutronics and thermo-fluid simulation of a prismatic {VHTR} core},
	volume = {92},
	issn = {03064549},
	url = {https://linkinghub.elsevier.com/retrieve/pii/S0306454916300603},
	doi = {10.1016/j.anucene.2016.01.050},
	abstract = {There exists a large body of publications on neutronics or thermal-fluids analyses of a prismatic gascooled reactor core. They mainly focus on stand-alone analyses, although the neutronics and thermalfluids behavior affect each other. In order to consider the interaction between the neutronics and thermal-fluids behavior, a coupled analysis is desirable. However, only a few studies are available in the open literature on the coupled analysis of a prismatic gas-cooled reactor core. In this work, a code system for a coupled neutronics and thermo-fluids simulation was developed using CAPP (reactor physics code) and GAMMA+ (thermo-fluid system code). Using the code system developed, coupled neutronics and thermo-fluid simulations were carried out for a prismatic very high temperature reactor (VHTR) core to extend the knowledge about the coupled behavior. In particular, the effect of the bypass flow modeling on the coupled behavior was analyzed.},
	language = {en},
	urldate = {2020-09-04},
	journal = {Annals of Nuclear Energy},
	author = {Tak, Nam-il and Lee, Hyun Chul and Lim, Hong Sik and Han, Tae Young},
	month = jun,
	year = {2016},
	pages = {228--242},
	file = {Tak et al. - 2016 - CAPPGAMMA+ code system for coupled neutronics and.pdf:/home/roberto/Zotero/storage/8637GKH2/Tak et al. - 2016 - CAPPGAMMA+ code system for coupled neutronics and.pdf:application/pdf}
}

@article{lim_gamma_2006,
	title = {{GAMMA} {Multidimensional} {Multicomponent} {Mixture} {Analysis} to {Predict} {Air} {Ingress} {Phenomena} in an {HTGR}},
	volume = {152},
	issn = {0029-5639, 1943-748X},
	url = {https://www.tandfonline.com/doi/full/10.13182/NSE06-5},
	doi = {10.13182/NSE06-5},
	abstract = {We developed a multidimensional GAs Multicomponent Mixture Analysis (GAMMA) code in order to investigate chemical reaction behaviors related to an air ingress accident and the thermofluid transients in high-temperature gas-cooled reactors. The implicit continuous Eulerian technique is adopted for the reduction of a 10N × 10N matrix into an N × N pressure difference matrix and fast transient computation. In the validation with a high-temperature engineering test reactor (HTTR)-simulated air ingress experiment, the onset times of natural convection are accurately predicted within a 10\% deviation. Small internal leaks in the HTTR-simulated test facility have been found to significantly affect the consequence of air ingress. In all the simulated cases for a SANA-1 afterheat removal test, the predictions of GAMMA are in a high level of agreement with the measured temperature profiles and are comparable to the results of other codes (TINTE, THERMIX/DIREKT, and TRIO-EF).},
	language = {en},
	number = {1},
	urldate = {2020-09-07},
	journal = {Nuclear Science and Engineering},
	author = {Lim, Hong Sik and No, Hee Cheon},
	month = jan,
	year = {2006},
	pages = {87--97},
	file = {Lim and No - 2006 - GAMMA Multidimensional Multicomponent Mixture Anal.pdf:/home/roberto/Zotero/storage/MSCGPL25/Lim and No - 2006 - GAMMA Multidimensional Multicomponent Mixture Anal.pdf:application/pdf}
}

@article{yuk_time-dependent_2020,
	title = {Time-dependent neutron diffusion analysis using finite element method for a block-type {VHTR} core design},
	volume = {360},
	issn = {0029-5493},
	url = {http://www.sciencedirect.com/science/article/pii/S0029549320300078},
	doi = {https://doi.org/10.1016/j.nucengdes.2020.110512},
	abstract = {The very high-temperature gas-cooled reactor (VHTR) has attracted interest owing to its high passive safety. A reactor analysis tool is required to analyze the safety characteristics of a VHTR in detail, and such a tool should be capable of a transient calculation for a reactivity insertion accident such as control rod ejection or withdrawal. This work presents a transient analysis of a block-type VHTR core. Time-dependent neutron diffusion equation is solved by the finite element method. A simplified thermal fluid analysis tool is also implemented to consider thermal feedback. In addition, a new method is introduced to resolve the control rod cusping effect without additional computation or mesh reconstruction. The above methods are applied to a reactor physics code CAPP, developed at Korea Atomic Energy Research Institute (KAERI). The methods are tested on three-dimensional VHTR problems that include control rods and show encouraging results.},
	journal = {Nuclear Engineering and Design},
	author = {Yuk, Seungsu and Cho, Jin Young and Jo, Chang Keun and Tak, Nam-il and Lim, Hong Sik},
	year = {2020},
	keywords = {Block-type VHTR, CAPP, Control rod cusping, Multi-physics analysis, Transient analysis},
	pages = {110512}
}

@misc{joo_resolution_1984,
	title = {Resolution of the control rod cusping problem for nodal methods},
	author = {Joo, Han-Sem},
	month = feb,
	year = {1984}
}

@techreport{novak_pronghorn_2018,
	address = {Idaho Falls, Idaho},
	title = {Pronghorn {Theory} {Manual}},
	url = {https://inldigitallibrary.inl.gov/sites/sti/sti/Sort_4439.pdf},
	language = {en},
	number = {INL/EXT-18-44453},
	institution = {INL},
	author = {Novak, A and Zou, L and Peterson, J and Andrs, D and Kelly, J and Slaybaugh, R and Martineau, R and Gougar, H},
	month = feb,
	year = {2018}
}

@misc{novak_pronghorn_2018-1,
	title = {Pronghorn: {A} {Porous} {Media} {Thermal}-{Hydraulics} {Core} {Simulator} and its {Validation} with the {SANA} {Experiments}},
	url = {https://www.osti.gov/servlets/purl/1478375},
	language = {en},
	publisher = {INL},
	author = {Novak, A J and Zou, L and Peterson, J W and Martineau, R C and Slaybaugh, R N},
	month = apr,
	year = {2018},
	file = {Novak et al. - 2018 - Pronghorn A Porous Media Thermal-Hydraulics Core .pdf:/home/roberto/Zotero/storage/SPF7BG2M/Novak et al. - 2018 - Pronghorn A Porous Media Thermal-Hydraulics Core .pdf:application/pdf}
}

@techreport{wang_rattlesnake_2019,
	address = {Idaho Falls, ID},
	title = {Rattlesnake {Theory} {Manual}},
	url = {https://rattlesnake.inl.gov/SiteAssets/SitePages/manuals/theory.pdf},
	number = {INL/EXT-17-42103},
	institution = {INL},
	author = {Wang, Yaqi and Schunert, Sebastian and Laboure, Vincent},
	month = apr,
	year = {2019}
}

@techreport{j_ortensi_relap-7_2012,
	title = {{RELAP}-7 and {PRONGHORN} {Initial} {Integration} {Plan}},
	url = {http://www.osti.gov/servlets/purl/1048408/},
	language = {en},
	number = {INL/EXT-12-26016, 1048408},
	urldate = {2020-09-07},
	author = {{J. Ortensi} and {D. Andrs} and {A.A. Bingham} and {R.C. Martineau} and {J.W. Peterson}},
	month = may,
	year = {2012},
	doi = {10.2172/1048408},
	pages = {INL/EXT--12--26016, 1048408},
	file = {J. Ortensi et al. - 2012 - RELAP-7 and PRONGHORN Initial Integration Plan.pdf:/home/roberto/Zotero/storage/LIQRGF4X/J. Ortensi et al. - 2012 - RELAP-7 and PRONGHORN Initial Integration Plan.pdf:application/pdf}
}

@techreport{j_ortensi_initial_2012,
	title = {Initial {Coupling} of the {RELAP}-7 and {PRONGHORN} {Applications}},
	url = {http://www.osti.gov/servlets/purl/1060984/},
	language = {en},
	number = {INL/EXT-12-27350, 1060984},
	urldate = {2020-09-08},
	institution = {INL},
	author = {{J. Ortensi} and {D. Andrs} and {A.A. Bingham} and {R.C. Martineau} and {J.W. Peterson}},
	month = oct,
	year = {2012},
	doi = {10.2172/1060984},
	pages = {INL/EXT--12--27350, 1060984},
	file = {J. Ortensi et al. - 2012 - Initial Coupling of the RELAP-7 and PRONGHORN Appl.pdf:/home/roberto/Zotero/storage/2RKDC8FP/J. Ortensi et al. - 2012 - Initial Coupling of the RELAP-7 and PRONGHORN Appl.pdf:application/pdf}
}

@techreport{andrs_relap-7_2012,
	address = {Idaho Falls, ID},
	title = {{RELAP}-7 level 2 milestone report: {Demonstration} of a steady state single phase pwr simlulation with relap-7},
	url = {https://core.ac.uk/download/pdf/205952457.pdf},
	number = {INL/EXT-12-25924},
	institution = {INL},
	author = {Andrs, D and Berry, R and Gaston, R and Martineau, R and Peterson, J and Zhang, H and Zhao, H and Zou, L},
	year = {2012}
}

@techreport{ellis_initial_2014,
	title = {Initial {RattleSnake} {Calculations} of the {Hot} {Zero} {Power} {BEAVRS}},
	url = {http://www.osti.gov/servlets/purl/1126731/},
	language = {en},
	number = {INL/EXT--13-30903, 1126731},
	urldate = {2020-09-08},
	author = {Ellis, Matthew and Ortensi, J. and Wang, Y. and Smith, Kord and Martineau, R. C.},
	month = jan,
	year = {2014},
	doi = {10.2172/1126731},
	pages = {INL/EXT--13--30903, 1126731},
	file = {Ellis et al. - 2014 - Initial RattleSnake Calculations of the Hot Zero P:/home/roberto/Zotero/storage/JEY6N34Q/Ellis et al. - 2014 - Initial RattleSnake Calculations of the Hot Zero P:application/pdf}
}

@techreport{strydom_inl_2013,
	address = {Idaho Falls, ID},
	title = {{INL} {Results} for {Phase} {I} and {III} of the {OECD}/{NEA} {MHTGR}-350 {Benchmark}},
	url = {https://inldigitallibrary.inl.gov/sites/sti/sti/6007188.pdf},
	language = {en},
	number = {INL/EXT-13-30176},
	institution = {INL},
	author = {Strydom, Gerhard and Ortensi, Javier and Sen, Sonat and Hammer, Hans},
	month = sep,
	year = {2013}
}

@inproceedings{wang_krylov_2011,
	address = {Rio de Janeiro, Brazil},
	title = {Krylov {Solvers} {Preconditioned} with the {Low}-{Order} {Red}-{Black} {Algorithm} for {The} {Pn} {Hybrid} {FEM} for the {INSTANT} {Code}},
	booktitle = {International {Conference} on {Mathematics} and {Computational} {Methods} {Applied} to {Nuclear} {Science} and {Engineering}},
	author = {Wang, Y and Rabiti, C and Palmiotti, G},
	month = may,
	year = {2011}
}

@inproceedings{damian_vhtr_2008,
	address = {Washington, DC, USA},
	title = {{VHTR} {Core} {Preliminary} {Analysis} {Using} {NEPHTIS3} / {CAST3M} {Coupled} {Modelling}},
	isbn = {978-0-7918-4854-8 978-0-7918-3834-1},
	url = {https://asmedigitalcollection.asme.org/HTR/proceedings/HTR2008/48548/419/334907},
	doi = {10.1115/HTR2008-58052},
	abstract = {Along with the GFR another gas-cooled reactor identified in the Gen IV technology roadmap, the VHTR is studied in France. Some models have been developed at CEA relying on existing computational tools essentially dedicated to the prismatic block type reactor. These models simulate normal operating conditions and accidental reactor transients by using neutronic [1], thermal-hydraulic, system analysis codes [2], and their coupling [3,4].},
	language = {en},
	urldate = {2020-09-09},
	booktitle = {Fourth {International} {Topical} {Meeting} on {High} {Temperature} {Reactor} {Technology}, {Volume} 1},
	publisher = {ASMEDC},
	author = {Damian, Frederic},
	month = jan,
	year = {2008},
	pages = {419--429},
	file = {Damian - 2008 - VHTR Core Preliminary Analysis Using NEPHTIS3  CA.pdf:/home/roberto/Zotero/storage/CBLY44MG/Damian - 2008 - VHTR Core Preliminary Analysis Using NEPHTIS3  CA.pdf:application/pdf}
}

@inproceedings{loubiere_apollo2_1999,
	address = {Madrid, Spain},
	title = {{APOLLO2} {Twelve} {Years} {Later}},
	booktitle = {International {Conference} on {Mathematical} and {Computation}, {Reactor} {Physics} and {Environmental} {Analysis} in {Nuclear} {Applications}},
	author = {Loubiere, S. and Sanchez, R. and Coste, A. and Hebert, A. and Stankovski, Z. and Van der Gucht, C. and Zmijarevic, I.},
	month = sep,
	year = {1999}
}

@inproceedings{cavalier_presentation_2005,
	address = {Paris, France},
	title = {Presentation of the {HTR} {Neutronic} {Code} {System} {NEPHTIS}},
	volume = {INIS-FR--4515},
	url = {https://inis.iaea.org/search/search.aspx?orig_q=RN:37057810},
	booktitle = {{ENC} 2005: {European} nuclear conference. {Nuclear} power for the 21. century: from basic research to high-tech industry},
	author = {Cavalier, C and Trakas, Christos and Stepnik, Bertrand and Damian, Frederic and Groizard, Mathilde and Raepsaet, Xavier},
	year = {2005},
	pages = {12}
}

@misc{iaea_advanced_1992,
	title = {Advanced calculational methods for power reactors and {LWR} core design parameters},
	author = {IAEA},
	month = dec,
	year = {1992}
}

@inproceedings{krebs_calculational_1990,
	address = {Cadarache, France},
	title = {Calculational {Methods} for {PWRs}},
	booktitle = {Specialists meeting on advanced calculational methods for power reactors},
	publisher = {IAEA Specialists Meeting},
	author = {Krebs, J. and Laigle, M. and Lenain, R. and Mathonniere, G. and Nicolas, A.},
	month = sep,
	year = {1990}
}

@techreport{lautard_cronos_1990,
	address = {CEA Centre d'Etudes Nucleaires de Saclay},
	title = {{CRONOS}, {A} {Modular} {Computational} {System} for {Neutronic} {Core} {Calculations}},
	url = {https://inis.iaea.org/search/search.aspx?orig_q=RN:24033460},
	number = {IAEA-TECDOC--678},
	institution = {IAEA Specialists Meeting},
	author = {Lautard, J.J. and Loubiere, S. and Magnaud, C},
	month = sep,
	year = {1990}
}

@misc{ragusa_feasibility_2001,
	title = {Feasibility of the {Integration} of {CRONOS}, a 3-{D} {Neutronics} {Code}, into {Real}-{Time} {Simulators}},
	author = {Ragusa, J.C.},
	month = apr,
	year = {2001}
}

@misc{boer_assessment_2008,
	title = {Assessment of {VHTR} core design with regard to fuel temperature {Normal} operation},
	author = {Boer, B. and Kloosterman, J.L. and Damian, F.},
	month = jul,
	year = {2008}
}

@article{studer_cast3marcturus_2007,
	title = {{CAST3M}/{ARCTURUS}: {A} coupled heat transfer {CFD} code for thermal–hydraulic analyzes of gas cooled reactors},
	volume = {237},
	issn = {00295493},
	shorttitle = {{CAST3M}/{ARCTURUS}},
	url = {https://linkinghub.elsevier.com/retrieve/pii/S0029549307002099},
	doi = {10.1016/j.nucengdes.2007.03.016},
	abstract = {The safety of gas cooled reactors (High Temperature Reactors (HTR), Very High Temperature Reactors (VHTR) or Gas Cooled Fast Reactors (GFR)) must be ensured by systems (active or passive) which maintain loads on component (fuel) and structures (vessel, containment) within acceptable limits under accidental conditions. To achieve this objective, thermal–hydraulics computer codes are necessary tools to design, enhance the performance and ensure a high safety level of the different reactors. Some key safety questions are related to the evaluation of decay heat removal and containment pressure and thermal loads. This requires accurate simulations of conduction, convection, thermal radiation transfers and energy storage. Coupling with neutronics is also an important modeling aspect for the determination of representative parameters such as neutronics coefficient (Doppler coefficient, Moderator coeffcient, . . .), critical position of control rods, reactivity insertion aspects, . . .. For GFR, the high power density of the core and its necessary reduced dimension cannot rely only on passive systems for decay heat removal. Therefore, forced convection using active safety systems (gas blowers, heat exchangers, . . .) are highly recommended. Nevertheless, in case of station black-out, the safety demonstration of the concept should be guaranteed by natural circulation heat removal. This could be performed by keeping a relatively high back-up pressure for pure helium convection and also by heavy gas injection. So, it is also necessary to model mixing of different gases, the on-set of natural convection and the pressure and thermal loads onto the proximate or guard containment. In this paper, we report on the developments of the CAST3M/ARCTURUS thermal–hydraulics (Lumped Parameter and CFD) code developed at CEA, including its coupling to the neutronics code CRONOS2 and the system code CATHARE. Elementary validation cases are detailed, as well as application of the code to benchmark problems such as the HTR-10 thermal–hydraulic exercise. Examples of containment thermal–hydraulics calculations for fast reactor design (GFR) are also detailed. © 2007 Elsevier B.V. All rights reserved.},
	language = {en},
	number = {15-17},
	urldate = {2020-09-12},
	journal = {Nuclear Engineering and Design},
	author = {Studer, E. and Beccantini, A. and Gounand, S. and Dabbene, F. and Magnaud, J.P. and Paillère, H. and Limaiem, I. and Damian, F. and Golfier, H. and Bassi, C. and Garnier, J.C.},
	month = sep,
	year = {2007},
	pages = {1814--1828}
}

@mastersthesis{han_sensitivity_2008,
	address = {University Park, Pennsylvania},
	title = {Sensitivity {Study} on the {Energy} {Group} {Structure} for {High} {Temperature} {Reactor} {Analysis}},
	url = {https://etda.libraries.psu.edu/files/final_submissions/3001},
	school = {Pennsylvania State University},
	author = {Han, James Sanggene},
	month = may,
	year = {2008}
}


@misc{grimesey_combinepc-portable_1990,
	address = {Idaho Falls, ID},
	title = {{COMBINE}/{PC}-{A} {Portable} {ENDF}/{B} {Version} 5 {Neutron} {Spectrum} and {Cross} {Section} {Generation} {Program}},
	url = {https://www.osti.gov/biblio/7161236},
	publisher = {EG and G Idaho, Inc.},
	author = {Grimesey, R.A. and Nigg, D.W. and Curtis, R.L.},
	month = apr,
	year = {1990}
}

@misc{grimesey_combinepc-portable_1994,
	title = {{COMBINE}/{PC}-{A} {Portable} {ENDF}/{B} {Version} 6 {Neutron} {Spectrum} and {Cross} {Section} {Generation} {Program}},
	author = {Grimesey, R.A.},
	year = {1994}
}

@phdthesis{bandini_three-dimensional_1990,
	type = {Ph.{D}.},
	title = {A {Three}-dimensional {Transient} {Neutronics} {Routine} for the {TRAC}-{PF1} {Reactor} {Thermal} {Hydraulic} {Computer} {Code}},
	school = {The Pennsylvania State University},
	author = {Bandini, B},
	year = {1990}
}

@techreport{x-5_monte_carlo_team_mcnp_2003,
	address = {Los Alamos, NM},
	title = {{MCNP} — {A} {General} {Monte} {Carlo} {N}-{Particle} {Transport} {Code}, {Version} 5},
	number = {LA-UR-03-1987},
	institution = {LANL},
	author = {X-5 Monte Carlo Team},
	month = apr,
	year = {2003}
}

@book{merrill_nuclear_1973,
	title = {Nuclear {Design} {Methods} and {Experimental} {Data} in {Use} at {Gulf} {General} {Atomic}},
	url = {https://books.google.com/books?id=4sc3MgAACAAJ},
	publisher = {Gulf Environmental Systems Company},
	author = {Merrill, M.H. and Company, Gulf Environmental Systems},
	year = {1973}
}

@techreport{strydom_results_2015,
	title = {Results for {Phase} {I} of the {IAEA} {Coordinated} {Research} {Program} on {HTGR} {Uncertainties}},
	url = {http://www.osti.gov/servlets/purl/1173079/},
	language = {en},
	number = {INL/EXT--14-32944, 1173079},
	urldate = {2020-09-20},
	institution = {INL},
	author = {Strydom, Gerhard and Bostelmann, Friederike and Yoon, Su Jong},
	month = jan,
	year = {2015},
	doi = {10.2172/1173079},
	pages = {INL/EXT--14--32944, 1173079},
	file = {Strydom et al. - 2015 - Results for Phase I of the IAEA Coordinated Resear:/home/roberto/Zotero/storage/2T45E4IF/Strydom et al. - 2015 - Results for Phase I of the IAEA Coordinated Resear:application/pdf}
}

% CHAPTER2: LITERATURE REVIEW Thermal-Hydraulics
@misc{ansys_cfx_2006,
	address = {USA},
	title = {{CFX} 11: {User}'s {Manual} {Version} 11},
	author = {ANSYS},
	year = {2006}
}

@misc{general_atomics_gas_1996,
	title = {Gas {Turbine}-{Modular} {Helium} {Reactor} ({GT}-{MHR}) {Conceptual} {Design} {Description} {Report}},
	author = {General Atomics},
	year = {1996}
}

@techreport{poiter_gas_1996,
	address = {San Diego, CA},
	title = {Gas {Turbine}-{Modular} {Helium} {Reactor} ({GT}-{MHR}) {Conceptual} {Design} {Description} {Report}},
	url = {https://www.nrc.gov/docs/ML0224/ML022470282.pdf},
	number = {910720},
	institution = {General Atomic},
	author = {Poiter, R.C. and Shenoy, A.},
	year = {1996}
}

@article{no_multi-component_2007,
	title = {Multi-component diffusion analysis and assessment of {GAMMA} code and improved {RELAP5} code},
	volume = {237},
	issn = {00295493},
	url = {https://linkinghub.elsevier.com/retrieve/pii/S0029549306006200},
	doi = {10.1016/j.nucengdes.2006.10.020},
	abstract = {A loss-of-coolant accident (LOCA) has been considered a critical event for very high temperature gas-cooled reactor (VHTR). Following helium depressurization, it is anticipated that unless countermeasures are taken, air will enter the core through the break by molecular diffusion and ultimately by natural convection leading to oxidation of the in-core graphite structure. Thus, without any mitigating features, a LOCA will lead to an air ingress event, which will lead to exothermic chemical reactions of graphite with oxygen, potentially resulting in significant increases of the core temperature.},
	language = {en},
	number = {10},
	urldate = {2020-07-31},
	journal = {Nuclear Engineering and Design},
	author = {No, Hee Cheon and Lim, Hong S. and Kim, Jong and Oh, Chang and Siefken, Larry and Davis, Cliff},
	month = may,
	year = {2007},
	pages = {997--1008},
	file = {No et al. - 2007 - Multi-component diffusion analysis and assessment .pdf:/home/roberto/Zotero/storage/JULB2FKA/No et al. - 2007 - Multi-component diffusion analysis and assessment .pdf:application/pdf}
}

@inproceedings{reza_design_2006,
	address = {Reno, NV},
	title = {Design of an {Alternative} {Coolant} {Inlet} {Flow} {Configuration} for the {Modular} {Helium} {Reactor}},
	booktitle = {International {Congress} on {Advances} in {Nuclear} {Power} {Plants} ({ICAPP} 06)},
	author = {Reza, SM Mohsin and Harvego, Edwin and Richards, Matt and Shenoy, Arkal and Pddicord, Kenneth},
	month = jun,
	year = {2006},
	file = {Reza et al. - 2006 - Design of an Alternative Coolant Inlet Flow Config.pdf:/home/roberto/Zotero/storage/SJI7EZK6/Reza et al. - 2006 - Design of an Alternative Coolant Inlet Flow Config.pdf:application/pdf}
}

@inproceedings{riemke_relap5-3d_2005,
	address = {Miami, FL},
	title = {{RELAP5}-{3D} {Code} {Includes} {Athena} {Features} and {Models}},
	booktitle = {International {Conference} on {Nuclear} {Engineering} – {ICONE} 14-89217},
	publisher = {ASME},
	author = {Riemke, R. and Davis, C. and Schultz, R.},
	month = jul,
	year = {2005}
}

@article{nakano_conceptual_2008,
	title = {Conceptual {Reactor} {Design} {Study} of {Very} {High} {Temperature} {Reactor} ({VHTR}) with {Prismatic}-{Type} {Core}},
	volume = {2},
	journal = {Journal of Power and Energy Systems},
	author = {Nakano and Tsuji and Tazawa},
	year = {2008},
	file = {Nakano et al. - 2008 - Conceptual Reactor Design Study of Very High Tempe.pdf:/home/roberto/Zotero/storage/YVKXMCX6/Nakano et al. - 2008 - Conceptual Reactor Design Study of Very High Tempe.pdf:application/pdf}
}

@inproceedings{maruyama_verification_1989,
	address = {Karlsruhe, Germany},
	title = {Verification of in-core thermal and hydraulic analysis code {FLOWNET}/{TRUMP} for the high temperature engineering test reactor ({HTTR}) at {JAERI}},
	volume = {22},
	isbn = {3-7650-1115-0},
	language = {Japanese},
	author = {Maruyama, S and Sudo, Y and Saito, S and Kiso, Y and Hayakawa, H},
	year = {1989}
}

@misc{peterson_tac2d_1969,
	title = {{TAC2D} {A} {General} {Purpose} {Two}-dimensional {Heat} {Transfer} {Computer} {Code} {User}'s {Manual}},
	publisher = {General Atomics},
	author = {Peterson, J},
	year = {1969}
}

@inproceedings{in_three-dimensional_2006,
	address = {Gapyoung, Korea},
	title = {Three-{Dimensional} {Analysis} of the {Hot}-{Spot} {Fuel} {Temperature} in {Pebble} {Bed} and {Prismatic} {Modular} {Reactors}},
	language = {en},
	booktitle = {Spring meeting of the {Korean} {Nuclear} {Society}},
	author = {In, W.K. and Lee, S.W. and Lim, H.S. and Lee, W.J.},
	month = may,
	year = {2006},
	file = {In et al. - 2006 - Three-Dimensional Analysis of the Hot-Spot Fuel Te.pdf:/home/roberto/Zotero/storage/H2M2Y7SQ/In et al. - 2006 - Three-Dimensional Analysis of the Hot-Spot Fuel Te.pdf:application/pdf}
}

@misc{ansys_cfx_2005,
	address = {USA},
	title = {{CFX} 10: {User}'s {Manual} {Version} 10},
	author = {ANSYS},
	year = {2005}
}

@article{richards_thermal_2007,
	title = {Thermal {Hydraulic} {Optimization} of a {VHTR} {Block}-{Type} {Core}},
	language = {en},
	author = {Richards, Matt and Lee, Won-Jae and Kim, Yong-Hee and Tak, Nam-IL and Reza, Mohsin},
	year = {2007},
	pages = {9},
	file = {Richards et al. - 2007 - THERMAL HYDRAULIC OPTIMIZATION OF A VHTR BLOCK-TYP.pdf:/home/roberto/Zotero/storage/GHENT97E/Richards et al. - 2007 - THERMAL HYDRAULIC OPTIMIZATION OF A VHTR BLOCK-TYP.pdf:application/pdf}
}

@article{cioni_3d_2005,
	title = {{3D} thermal-hydraulic calculations of a modular block-type {HTR} core},
	volume = {236},
	issn = {00295493},
	url = {https://linkinghub.elsevier.com/retrieve/pii/S0029549306000240},
	doi = {10.1016/j.nucengdes.2005.10.024},
	abstract = {This paper deals with 3D numerical simulation of assemblies of a high temperature helium cooled reactor core. The present aim is to investigate an emergency situation due to the blocking of few Helium channels in the core. In this situation, some Helium channels for coolant passage are blocked and the related issue is the temperature distribution within the assemblies, as well as the locations of the local extrema in graphite and fuel media, that must not exceed 1600 ◦C. Two configurations are investigated, depending on the situation of the blocked fuel assembly in the core. The blocked assembly is surrounded by six unblocked fuel assemblies in the first situation and by five unblocked fuel assemblies and one reflector assembly in the second situation.},
	language = {en},
	number = {5-6},
	urldate = {2020-08-04},
	journal = {Nuclear Engineering and Design},
	author = {Cioni, O. and Marchand, M. and Geffraye, G. and Ducros, F.},
	month = mar,
	year = {2005},
	pages = {565--573},
	file = {Cioni et al. - 2006 - 3D thermal-hydraulic calculations of a modular blo.pdf:/home/roberto/Zotero/storage/NK32VCF7/Cioni et al. - 2006 - 3D thermal-hydraulic calculations of a modular blo.pdf:application/pdf}
}

@inproceedings{bieder_priceles_2000,
	address = {Montreal QC, Canada},
	title = {{PRICELES}: an object oriented code for industrial large eddy simulations.},
	booktitle = {8th {Annual} {Conference} of the {CFD} {Society} of {Canada}},
	author = {Bieder, U and Calvin,, C and Emonot, P},
	month = jun,
	year = {2000}
}

@article{simoneau_three-dimensional_2007,
	title = {Three-dimensional simulation of the coupled convective, conductive, and radiative heat transfer during decay heat removal in an {HTR}},
	volume = {237},
	issn = {00295493},
	url = {https://linkinghub.elsevier.com/retrieve/pii/S0029549307002294},
	doi = {10.1016/j.nucengdes.2007.03.010},
	abstract = {A three-dimensional model has been constructed to simulate the passive heat removal in a modular prismatic-block high temperature reactor during a loss of active cooling accident. This model, developed using the STAR-CD general computational fluid dynamics code, solves the combined conductive, convective, and radiative heat transfer within a 30◦ section of the core and reactor vessel. To accommodate the different spatial scales, it uses homogeneous equivalent media to represent the coolant flow and the prismatic fuel blocks. A customized procedure that manages solving alternatively the dynamic and thermal fields permits the computation of very long transients, which typically are performed for 100 or more hours of simulated time.},
	language = {en},
	number = {15-17},
	urldate = {2020-08-02},
	journal = {Nuclear Engineering and Design},
	author = {Simoneau, Jan-Patrice and Champigny, Julien and Mays, Brian and Lommers, Lewis},
	month = sep,
	year = {2007},
	pages = {1923--1937},
	file = {Simoneau et al. - 2007 - Three-dimensional simulation of the coupled convec.pdf:/home/roberto/Zotero/storage/YRN7XQWB/Simoneau et al. - 2007 - Three-dimensional simulation of the coupled convec.pdf:application/pdf}
}

@article{sato_computational_2010,
	title = {Computational fluid dynamic analysis of core bypass flow phenomena in a prismatic {VHTR}},
	volume = {37},
	issn = {03064549},
	url = {https://linkinghub.elsevier.com/retrieve/pii/S0306454910001581},
	doi = {10.1016/j.anucene.2010.04.021},
	abstract = {The core bypass flow in a prismatic very high temperature reactor (VHTR) is an important design consideration and can have considerable impact on the condition of reactor core internals including fuels. The interstitial gaps are an inherent presence in the reactor core because of tolerances in manufacturing the blocks and the inexact nature of their installation. Furthermore, the geometry of the graphite blocks changes over the lifetime of the reactor because of thermal expansion and irradiation damage. The occurrence of hot spots in the core and lower plenum and hot streaking in the lower plenum (regions of very hot gas flow) are affected by bypass flow.},
	language = {en},
	number = {9},
	urldate = {2020-08-04},
	journal = {Annals of Nuclear Energy},
	author = {Sato, Hiroyuki and Johnson, Richard and Schultz, Richard},
	month = sep,
	year = {2010},
	pages = {1172--1185},
	file = {Sato et al. - 2010 - Computational fluid dynamic analysis of core bypas.pdf:/home/roberto/Zotero/storage/67CGBAVA/Sato et al. - 2010 - Computational fluid dynamic analysis of core bypas.pdf:application/pdf}
}

@misc{fluent_fluent_2006,
	address = {Lebanon, NH},
	title = {{FLUENT} {Version} 6.3.26},
	publisher = {FLUENT Inc.},
	author = {FLUENT},
	year = {2006}
}

@article{takada_core_2004,
	title = {Core thermal–hydraulic design},
	volume = {233},
	issn = {00295493},
	url = {https://linkinghub.elsevier.com/retrieve/pii/S0029549304002365},
	doi = {10.1016/j.nucengdes.2004.07.009},
	abstract = {The core thermal–hydraulic design for the HTTR is carried out to evaluate the maximum fuel temperature at normal operation and anticipated operation occurrences. To evaluate coolant flow distribution and maximum fuel temperature, we use the experimental results such as heat transfer coefficient, pressure loss coefficient obtained by mock-up test facilities. Furthermore, we evaluated hot spot factors of fuel temperatures conservatively.},
	language = {en},
	number = {1-3},
	urldate = {2020-09-15},
	journal = {Nuclear Engineering and Design},
	author = {Takada, Eiji and Nakagawa, Shigeaki and Fujimoto, Nozomu and Tochio, Daisuke},
	month = oct,
	year = {2004},
	pages = {37--43},
	file = {Takada et al. - 2004 - Core thermal–hydraulic design.pdf:/home/roberto/Zotero/storage/6ZCMTIV2/Takada et al. - 2004 - Core thermal–hydraulic design.pdf:application/pdf}
}

@article{wu_investigating_2010,
	title = {Investigating the advantages and disadvantages of realistic approach and porous approach for closely packed pebbles in {CFD} simulation},
	volume = {240},
	issn = {00295493},
	url = {https://linkinghub.elsevier.com/retrieve/pii/S0029549310000397},
	doi = {10.1016/j.nucengdes.2010.01.015},
	abstract = {A pebble bed geometry is usually adopted for high-temperature gas-cooled reactors (HTGRs), which exhibits inherently safe performance, high conversion efficiency, and low power density design. It is important to understand the thermal-hydraulic characteristics of HTGR core for optimum design and safe operation. Therefore, this study investigates the thermal-hydraulic behaviors in a segment of pebbles predicted by the Reynolds-averaged Navier–Stokes (RANS) computational fluid dynamics (CFD) model using porous and realistic approaches for the complicated geometry. The advantages of each approach’s methodology for the closely packed pebble geometry can be revealed by comparing the calculated results. In an engineering application, a CFD simulation with the porous approach for the pebble geometry can quickly and reasonably capture the averaged behaviors of the thermal-hydraulic parameters as the gas flows through the core, including the pressure drop and temperature increase. However, it is necessary to utilize the realistic approach for this complicated geometry to obtain the detailed and localized characteristics within the fluid and solid fuel regions. The present simulation results can provide useful information to help CFD researchers to determine an appropriate approach to be used when investigating the thermal-hydraulic characteristics within the reactor core of a closely packed pebble bed.},
	language = {en},
	number = {5},
	urldate = {2020-09-17},
	journal = {Nuclear Engineering and Design},
	author = {Wu, C.Y. and Ferng, Y.M. and Chieng, C.C. and Liu, C.C.},
	month = may,
	year = {2010},
	pages = {1151--1159},
	file = {Wu et al. - 2010 - Investigating the advantages and disadvantages of .pdf:/home/roberto/Zotero/storage/TX8ADXFD/Wu et al. - 2010 - Investigating the advantages and disadvantages of .pdf:application/pdf}
}

@techreport{shenoy_htgr_1974,
	address = {San Diego, CA},
	title = {{HTGR} {Core} {Thermal} {Design} {Methods} and {Analysis}},
	number = {GA-A12985 (GA-LTR-17)},
	institution = {General Atomic Co.},
	author = {Shenoy, A.S. and McEachern, D.W.},
	month = dec,
	year = {1974}
}

@article{tak_practical_2012,
	title = {A practical method for {Whole}-{Core} {Thermal} {Analysis} of a {Prismatic} {Gas}-{Cooled} {Reactor}},
	doi = {10.13182/NT12-A13480},
	journal = {Nuclear Technology},
	author = {Tak, Nam-Il and Kim, Min-Hwan and Lim, Hong Sik and Noh, Jae Man},
	month = mar,
	year = {2012}
}

@article{tak_development_2014,
	title = {{DEVELOPMENT} {OF} {A} {CORE} {THERMO}-{FLUID} {ANALYSIS} {CODE} {FOR} {PRISMATIC} {GAS} {COOLED} {REACTORS}},
	volume = {46},
	issn = {1738-5733},
	url = {http://www.sciencedirect.com/science/article/pii/S1738573315301042},
	doi = {https://doi.org/10.5516/NET.02.2014.020},
	abstract = {A new computer code, named CORONA (Core Reliable Optimization and thermo-fluid Network Analysis), was developed for the core thermo-fluid analysis of a prismatic gas cooled reactor. The CORONA code is targeted for whole-core thermo-fluid analysis of a prismatic gas cooled reactor, with fast computation and reasonable accuracy. In order to achieve this target, the development of CORONA focused on (1) an efficient numerical method, (2) efficient grid generation, and (3) parallel computation. The key idea for the efficient numerical method of CORONA is to solve a three-dimensional solid heat conduction equation combined with one-dimensional fluid flow network equations. The typical difficulties in generating computational grids for a whole core analysis were overcome by using a basic unit cell concept. A fast calculation was finally achieved by a block-wise parallel computation method. The objective of the present paper is to summarize the motivation and strategy, numerical approaches, verification and validation, parallel computation, and perspective of the CORONA code.},
	number = {5},
	journal = {Nuclear Engineering and Technology},
	author = {TAK, NAM-IL and LEE, SUNG NAM and KIM, MIN-HWAN and LIM, HONG SIK and NOH, JAE MAN},
	year = {2014},
	keywords = {CORONA, Fluid Flow Network, Fuel Temperature, Gas Cooled Reactor, Prismatic Core, VHTR},
	pages = {641 -- 654}
}

@inproceedings{tak_coupled_2016,
	address = {Gyeongju, Korea},
	title = {Coupled {Neutronics} and {Thermo}-{Fluid} {Simulation} of {Full}-{Core} {Prismatic} {Gas}-{Cooled} {Reactor} {Using} {DeCART}/{CORONA} {Code} {System}},
	language = {en},
	booktitle = {Transactions of the {Korean} {Nuclear} {Society} {Autumn} {Meeting}},
	author = {Tak, Nam-il and Han, Tae Young and Lee, Hyun Chul},
	month = oct,
	year = {2016},
	pages = {4},
	file = {Tak et al. - 2016 - Coupled Neutronics and Thermo-Fluid Simulation of .pdf:/home/roberto/Zotero/storage/CE3TVM43/Tak et al. - 2016 - Coupled Neutronics and Thermo-Fluid Simulation of .pdf:application/pdf}
}

@misc{kaeri_decart_2007,
	title = {{DeCART} v1.2 {User}'s {Manual}},
	author = {KAERI},
	year = {2007}
}

@article{travis_thermalhydraulics_2013,
	title = {Thermal–hydraulics analyses for 1/6 prismatic {VHTR} core and fuel element with and without bypass flow},
	volume = {67},
	issn = {01968904},
	url = {https://linkinghub.elsevier.com/retrieve/pii/S0196890412004396},
	doi = {10.1016/j.enconman.2012.11.003},
	abstract = {Presented are the results of thermal–hydraulics analyses for a Very High Temperature Reactor (VHTR) 1/6 core and hexagonal fuel element, with and without helium-coolant bypass flow in interstitial gaps and in the control rod channels. Owing to the complexity and massive size of the VHTR, a full core analysis requires extensive and parallelized computation capabilities and a long time (weeks to months) to complete. These demanding requirements are mostly due to the 3-D computational fluid dynamics (CFD) simulation of the helium flow in the 10-m long coolant channels in the reactor core. Results demonstrate the effectiveness of coupling a 1-D helium flow in the channels together with a recently developed and validated convective heat transfer correlation, to a 3-D heat conduction in the graphite and fuel compacts in the core fuel elements. This simplified thermal–hydraulics analysis methodology for VHTR markedly reduces computation time and memory requirements, while maintaining accurate results. The helium bypass flow in interstitial gaps between fuel elements in the VHTR core provides additional cooling of the edge regions, but increases the temperature near the center of the prismatic fuel elements. Results also show that helium ‘‘bleed’’ flow through the control rod channels minimally affects the temperature distribution within the fuel elements. The heat generation in the corner burnable poison rods in the VHTR core fuel elements affects the temperature only in the close vicinity of the rods.},
	language = {en},
	urldate = {2020-09-18},
	journal = {Energy Conversion and Management},
	author = {Travis, Boyce W. and El-Genk, Mohamed S.},
	month = mar,
	year = {2013},
	pages = {325--341},
	file = {Travis and El-Genk - 2013 - Thermal–hydraulics analyses for 16 prismatic VHTR.pdf:/home/roberto/Zotero/storage/G3VLRHGF/Travis and El-Genk - 2013 - Thermal–hydraulics analyses for 16 prismatic VHTR.pdf:application/pdf}
}

@misc{cd-adapco_star-ccm_2012,
	title = {{STAR}-{CCM}+ release 7.04},
	author = {CD-adapco},
	year = {2012}
}

@misc{cd-adapco_star-cd_2004,
	address = {London},
	title = {{STAR}-{CD} {Version} 3.24 {User} {Guide}},
	author = {CD-adapco},
	year = {2004}
}

@phdthesis{huning_novel_2016,
	address = {Atlanta, GA},
	title = {A {NOVEL} {CORE} {ANALYSIS} {METHOD} {FOR} {PRISMATIC} {HIGH} {TEMPERATURE} {GAS} {REACTORS}},
	url = {https://smartech.gatech.edu/handle/1853/58615},
	school = {Georgia Institute of Technology},
	author = {Huning, Alexander},
	month = aug,
	year = {2016}
}

@techreport{macdonald_ngnp_2003,
	address = {Idaho Falls},
	title = {{NGNP} {Point} {Design} - {Results} of the {Initial} {Neutronics} and {Thermal}-{Hydraulic} {Assessments} {During} {Fy}-03},
	number = {INEEL/EXT-03-00870 Rev. 1},
	institution = {INL},
	author = {MacDonald, Philip E},
	month = sep,
	year = {2003}
}

@book{melese_thermal_1984,
	address = {La Grange Park, Ill., USA},
	title = {Thermal and flow design of helium-cooled reactors},
	isbn = {978-0-89448-027-0},
	language = {en},
	publisher = {American Nuclear Society},
	author = {Melese, Gilbert and Katz, Robert},
	collaborator = {American Nuclear Society and United States},
	year = {1984},
	keywords = {Design and construction, Gas cooled reactors, Helium},
	file = {Melese and Katz - 1984 - Thermal and flow design of helium-cooled reactors.pdf:/home/roberto/Zotero/storage/BYXWYJGD/Melese and Katz - 1984 - Thermal and flow design of helium-cooled reactors.pdf:application/pdf}
}

@article{bieder_qualification_2008,
	title = {Qualification of the {CFD} code {Trio}\_U for full scale reactor applications},
	volume = {238},
	issn = {00295493},
	url = {https://linkinghub.elsevier.com/retrieve/pii/S0029549307003366},
	doi = {10.1016/j.nucengdes.2007.02.040},
	abstract = {The article presents a procedure to qualify the Trio U code for the prediction of the boron concentration at the core inlet of a French 900 MWe pressurized water reactor under accidental conditions (inherent dilution problem).1 The objective of this procedure is to ensure that the validation calculations are performed with the same modelling hypotheses as the full scale reactor analysis, for which usually no experimental data are available. A density driven ROCOM experiment as well as an UPTF Tram-C3 experiment have been used for the qualification of the Trio U code. Both experiments present similar thermal hydraulic conditions as the reactor case. The predicted boron concentration at the core inlet of the reactor shows that the potential return to criticality might not be excluded in the case of a small break LOCA. Further neutronic calculations are necessary to confirm this result.},
	language = {en},
	number = {3},
	urldate = {2020-08-04},
	journal = {Nuclear Engineering and Design},
	author = {Bieder, Ulrich and Graffard, Estelle},
	month = mar,
	year = {2008},
	pages = {671--679},
	file = {Bieder and Graffard - 2008 - Qualification of the CFD code Trio_U for full scal.pdf:/home/roberto/Zotero/storage/MIBIDLMT/Bieder and Graffard - 2008 - Qualification of the CFD code Trio_U for full scal.pdf:application/pdf}
}

@techreport{ball_next_2008,
	title = {Next {Generation} {Nuclear} {Plant} {Phenomena} {Identification} and {Ranking} {Tables} ({PIRTs}) {Volume} 1: {Main} {Report}},
	shorttitle = {Next {Generation} {Nuclear} {Plant} {Phenomena} {Identification} and {Ranking} {Tables} ({PIRTs}) {Volume} 1},
	url = {http://www.osti.gov/servlets/purl/1001275-2PPpIH/},
	language = {en},
	number = {ORNL/TM-2007/147, 1001275},
	urldate = {2020-08-04},
	author = {Ball, Sydney J},
	month = mar,
	year = {2008},
	doi = {10.2172/1001275},
	pages = {ORNL/TM--2007/147, 1001275},
	file = {Ball - 2008 - Next Generation Nuclear Plant Phenomena Identifica.pdf:/home/roberto/Zotero/storage/367N5UUH/Ball - 2008 - Next Generation Nuclear Plant Phenomena Identifica.pdf:application/pdf}
}

@inproceedings{johnson_development_2008,
	address = {Washington, DC, USA},
	title = {Development of a {CFD} {Analysis} {Plan} for the {First} {VHTR} {Standard} {Problem}},
	isbn = {978-0-7918-4854-8 978-0-7918-3834-1},
	url = {https://asmedigitalcollection.asme.org/HTR/proceedings/HTR2008/48548/567/334847},
	doi = {10.1115/HTR2008-58283},
	abstract = {Data from a scaled model of a portion of the lower plenum of the helium-cooled very high temperature reactor (VHTR) are under consideration for acceptance as a computational fluid dynamics (CFD) validation data set or standard problem. A CFD analysis will help determine if the scaled model is a suitable geometry for validation data. The present article describes the development of an analysis plan for the CFD model. The plan examines the boundary conditions that should be used, the extent of the computational domain that should be included and which turbulence models need not be examined against the data. Calculations are made for a closely related 2D geometry to address these issues. It was found that a CFD model that includes only the inside of the scaled model in its computational domain is adequate for CFD calculations. The realizable k{\textasciitilde}İ model was found not to be suitable for this problem because it did not predict vortex-shedding.},
	language = {en},
	urldate = {2020-08-05},
	booktitle = {Fourth {International} {Topical} {Meeting} on {High} {Temperature} {Reactor} {Technology}, {Volume} 1},
	publisher = {ASMEDC},
	author = {Johnson, Richard W.},
	month = sep,
	year = {2008},
	pages = {567--574},
	file = {Johnson - 2008 - Development of a CFD Analysis Plan for the First V.pdf:/home/roberto/Zotero/storage/8S8AXBPT/Johnson - 2008 - Development of a CFD Analysis Plan for the First V.pdf:application/pdf}
}

@article{anderson_analysis_2007,
	title = {Analysis of the hot gas flow in the outlet plenum of the very high temperature reactor using coupled {RELAP5}-{3D} system code and a {CFD} code},
	volume = {238},
	issn = {00295493},
	url = {https://linkinghub.elsevier.com/retrieve/pii/S0029549307003925},
	doi = {10.1016/j.nucengdes.2007.06.008},
	abstract = {The very high temperature reactor (VHTR) system behavior should be predicted during normal operating conditions and postulated accident conditions. The plant accident scenario and the passive safety behavior should be accurately predicted. Uncertainties in passive safety behavior could have large effects on the resulting system characteristics. Due to these performance issues in the VHTR, there is a need for development, testing and validation of design tools to demonstrate the feasibility of the design concepts and guide the improvement of the plant components. One of the identified design issues for the gas-cooled reactor is the thermal mixing of the coolant exiting the core into the outlet plenum. Incomplete thermal mixing may give rise to thermal stresses in the downstream components. To provide flow details, the analysis presented in this paper was performed by coupling a VHTR model generated in a thermal hydraulic systems code to a computational fluid dynamics (CFD) outlet plenum model. The outlet conditions obtained from the systems code VHTR model provide the inlet boundary conditions to the CFD outlet plenum model. By coupling the two codes in this manner, the important three-dimensional flow effects in the outlet plenum are well modeled while avoiding modeling the entire reactor with a computationally expensive CFD code. The values of pressure, mass flow rate and temperature across the coupled boundary showed differences of less than 5\% in every location except for one channel. The coupling auxiliary program used in this analysis can be applied to many different cases requiring detailed three-dimensional modeling in a small portion of the domain.},
	language = {en},
	number = {1},
	urldate = {2020-08-05},
	journal = {Nuclear Engineering and Design},
	author = {Anderson, Nolan and Hassan, Yassin and Schultz, Richard},
	month = jun,
	year = {2007},
	pages = {274--279},
	file = {Anderson et al. - 2008 - Analysis of the hot gas flow in the outlet plenum .pdf:/home/roberto/Zotero/storage/HUBTUX5T/Anderson et al. - 2008 - Analysis of the hot gas flow in the outlet plenum .pdf:application/pdf}
}

@techreport{johnson_cfd_2009,
	address = {Idaho Falls, ID},
	title = {{CFD} {Analysis} of {Core} {Bypass} {Phenomena}},
	language = {en},
	number = {INL/EXT-09-16882},
	institution = {INL},
	author = {Johnson, Richard W and Sato, Hiroyuki and Schultz, Richard R},
	month = nov,
	year = {2009},
	pages = {60},
	file = {Johnson et al. - CFD Analysis of Core Bypass Phenomena.pdf:/home/roberto/Zotero/storage/FDCHXP7V/Johnson et al. - CFD Analysis of Core Bypass Phenomena.pdf:application/pdf}
}

@article{baxi_evaluation_2008,
	title = {Evaluation of alternate power conversion unit designs for the {GT}-{MHR}},
	volume = {238},
	issn = {00295493},
	url = {https://linkinghub.elsevier.com/retrieve/pii/S0029549308001398},
	doi = {10.1016/j.nucengdes.2007.12.021},
	abstract = {General Atomics in the USA and Experimental Design Bureau of Machine Building (OKBM) in the Russian Federation have jointly developed a nuclear power plant design, the gas turbine modular helium reactor (GT-MHR). There have been considerable improvements during the last 10 years, which resulted in a more effective, efficient and safe design. The existing design is based on a 600 MW(t) reactor cooled by helium at a pressure of about 7 MPa. The power conversion unit (PCU) uses reactor outlet helium with a temperature of 850 ◦C in a direct Brayton cycle to achieve the cycle efficiency of about 48\%. The PCU consists of a gas turbine, a recuperator, a precooler, low-pressure and high-pressure compressors, an intercooler, and a generator. The turbomachine (TM) includes the turbine, compressors and generator mounted on a single vertical shaft. TM shaft rotation speed is 4400 rpm. The shaft of generator is connected to the turbine shaft by a flexible coupling. The required grid frequency of generated electricity is achieved by a converter. All PCU components are enclosed in a single vessel. TM uses radial and axial electromagnetic bearings (EMB) for support. Catcher bearings (CB) are provided as redundant support for the TM rotor in case of EMBs failure. Several alternative PCU designs were analyzed on the basis of current progress in technologies, new world experience, and experience accumulated in the process of GT-MHR design development. Results of these analyses will be taken into account when a final PCU design is selected.},
	language = {en},
	number = {11},
	urldate = {2020-11-14},
	journal = {Nuclear Engineering and Design},
	author = {Baxi, C.B. and Shenoy, A. and Kostin, V.I. and Kodochigov, N.G. and Vasyaev, A.V. and Belov, S.E. and Golovko, V.F.},
	month = nov,
	year = {2008},
	pages = {2995--3001}
}

@techreport{doe-ne_technology_2002,
	title = {A {Technology} {Roadmap} for {Generation} {IV} {Nuclear} {Energy} {Systems}},
	url = {http://www.osti.gov/servlets/purl/859029-304XRr/},
	language = {en},
	number = {GIF-002-00, 859029},
	urldate = {2020-11-15},
	author = {DOE-NE and GIF},
	month = dec,
	year = {2002},
	doi = {10.2172/859029},
	pages = {GIF--002--00, 859029},
	file = {DOENE (USDOE Office of Nuclear Energy, Science and Technology (NE)) - 2002 - A Technology Roadmap for Generation IV Nuclear Ene.pdf:/home/roberto/Zotero/storage/3P3KQCHH/DOENE (USDOE Office of Nuclear Energy, Science and Technology (NE)) - 2002 - A Technology Roadmap for Generation IV Nuclear Ene.pdf:application/pdf}
}

% CHAPTER 2: LITERATURE REVIEW Multi-physics
@techreport{oecd_nea_benchmark_2017,
	title = {Benchmark of the {Modular} {High}-{Temperature} {Gas}-{Cooled} {Reactor} ({MHTGR})-350 {MW} {Core} {Design}},
	number = {NEA/NSC/R(2017)4},
	institution = {OECD},
	author = {OECD/NEA},
	year = {2017}
}

@techreport{oecd_nea_coupled_2020,
	title = {Coupled {Neutronic}/{Thermal}- {Fluid} {Benchmark} of the {MHTGR}-350 {MW} {Core} {Design}: {Results} for {Phase} {I} {Exercise} 1},
	number = {NEA/NSC/R(2019)5-JT03458400},
	institution = {OECD/NEA},
	author = {OECD/NEA},
	month = feb,
	year = {2020}
}

@inproceedings{tyobeka_htgr_2011,
	address = {Rio de Janeiro, Brazil},
	title = {{HTGR} {REACTOR} {PHYSICS}, {THERMAL}-{HYDRAULICS} {AND} {DEPLETION} {UNCERTAINTY} {ANALYSIS}: {A} {PROPOSED} {IAEA} {COORDINATED} {RESEARCH} {PROJECT}},
	booktitle = {International {Conference} on {Mathematics} and {Computational} {Methods} {Applied} to {Nuclear} {Science} and {Engineering}},
	publisher = {American Nuclear Society},
	author = {Tyobeka, Bismark and Reitsma, Frederik and Ivanov, Kostadin},
	month = may,
	year = {2011}
}

@techreport{iaea_evaluation_2003,
	address = {Vienna, Austria},
	title = {Evaluation of high temperature gas cooled reactor performance: {Benchmark} analysis related to initial testing of the {HTTR} and {HTR}-10},
	number = {IAEA-TECDOC-1382},
	institution = {IAEA},
	author = {IAEA},
	month = nov,
	year = {2003}
}

@inproceedings{reitsma_oecd-neansc_2008,
	address = {Interlaken, Switzerland},
	title = {The {OECD}-{NEA}/{NSC} {PBMR400} {MW} coupled neutronics thermal hydraulics transient benchmark - {Steady}-state results and status},
	volume = {41},
	isbn = {978-3-9521409-5-6},
	url = {https://inis.iaea.org/search/search.aspx?orig_q=RN:41119087},
	booktitle = {Proceedings of {PHYSOR}-2008 {Conference}, {International} {Conference} on the {Physics} of {Reactors} “{Nuclear} {Power}: {A} {Sustainable} {Resource}”},
	author = {Reitsma, F and Han, J and Ivanov, K and Sartori, E},
	month = sep,
	year = {2008}
}

% Thermo-mechanics
@article{kang_thermo-mechanical_2012,
	title = {Thermo-mechanical analysis of the prismatic fuel assembly of {VHTR} in normal operational condition},
	volume = {44},
	issn = {03064549},
	url = {https://linkinghub.elsevier.com/retrieve/pii/S030645491200031X},
	doi = {10.1016/j.anucene.2012.01.015},
	abstract = {A Very High Temperature Reactor (VHTR) is designed to generate high temperature which is typically around 950 °C at the outlet. The high temperature results in thermal stress with dimensional changes in addition to the mechanical stress caused by the external loads. It is necessary to assess the structural integrity of the graphite blocks and fuel compacts for thermal stress and dimensional changes; however, precise assessment has not been conducted because of the complexity of the fuel assembly geometry and absence of accurate temperature distribution. In the present study, a finite element analysis (FEA) of each fuel element in a 10-story fuel assembly was carried out by using the temperature distribution which was calculated from a previous study of computational fluid dynamics (CFD) analysis on the fuel assembly. The results showed that the maximum stress level in the fuel elements reaches 32\% of the tensile strength. The maximum stress point was at the circumference of the bottom surface of a graphite plug on which the thermal expansion of the fuel compacts pushes up the graphite plug. Mitigation of the maximum stress by placing a small gap between the graphite plug and the fuel compact was investigated and the effect of the gap was confirmed. Stress evaluation was performed according to a draft of standard for nuclear graphite. All stress components lied within the limits and the structural integrity was regarded to be attained.},
	language = {en},
	urldate = {2020-08-06},
	journal = {Annals of Nuclear Energy},
	author = {Kang, Ji-Ho and Tak, Nam-il and Kim, Min-Hwan},
	month = jun,
	year = {2012},
	pages = {76--86},
	file = {Kang et al. - 2012 - Thermo-mechanical analysis of the prismatic fuel a.pdf:/home/roberto/Zotero/storage/4KWR2ZWX/Kang et al. - 2012 - Thermo-mechanical analysis of the prismatic fuel a.pdf:application/pdf}
}

@techreport{taiwo_summary_2005,
	title = {Summary of {Generation}-{IV} {Transmutation} {Impacts}},
	number = {ANL-AFCI-150},
	institution = {ANL},
	author = {Taiwo, T and Hill, R},
	month = jun,
	year = {2005},
	file = {Taiwo and Hill - 2005 - Summary of Generation-IV Transmutation Impacts.pdf:/home/roberto/Zotero/storage/CAW88VCG/Taiwo and Hill - 2005 - Summary of Generation-IV Transmutation Impacts.pdf:application/pdf}
}

@book{nield_convection_1999,
	address = {New York, USA},
	title = {Convection in {Porous} {Media}},
	publisher = {Springer-Verlag},
	author = {Nield, D.A. and Bejan, A.},
	year = {1999}
}

@article{boeer_fast_1992,
	title = {Fast analytical flux reconstruction method for nodal space-time nuclear reactor analysis},
	volume = {24},
	issn = {0306-4549},
	abstract = {The reconstruction of the fine-mesh neutron flux distribution using coarse-mesh computational results is an important step in a nodal reactor analysis procedure. In this paper the problem is reformulated as a one-dimensional interpolation along the node boundaries followed by an approximate analytical solution of the associated Dirichlet problem. Finite difference techniques are employed for the calculation of boundary values and their derivatives for Cartesian and hexagonal fuel assemblies. Apart from the fact that this approach can be used for an arbitrary number of energy groups in any geometry, an additional advantage is that it yields an estimate of the order of accuracy. The interior solution is based on weak-element approximations to elliptic problems and uses in its simplest variant as points of support only flux corner values in addition to the known surface fluxes. The high computational accuracy and efficiency is demonstrated by some results for Cartesian and hexagonal configurations.},
	number = {9},
	journal = {Annals of Nuclear Energy (Oxford)},
	author = {Boeer, R and Finnemann, H},
	year = {1992}
}

% CHAPTER 3: Methodology
@article{kirk_libmesh_2006,
	title = {{libMesh} : a {C}++ library for parallel adaptive mesh refinement/coarsening simulations},
	volume = {22},
	issn = {0177-0667, 1435-5663},
	shorttitle = {{libMesh}},
	url = {http://link.springer.com/10.1007/s00366-006-0049-3},
	doi = {10.1007/s00366-006-0049-3},
	abstract = {In this paper we describe the libMesh (http://libmesh.sourceforge.net)framework for parallel adaptive finite element applications. libMesh is an open-source software library that has been developed to facilitate serial and parallel simulation of multiscale, multiphysics applications using adaptive mesh refinement and coarsening strategies. The main software development is being carried out in the CFDLab (http://cfdlab.ae.utexas.edu) at the University of Texas, but as with other open-source software projects; contributions are being made elsewhere in the US and abroad. The main goals of this article are: (1) to provide a basic reference source that describes libMesh and the underlying philosophy and software design approach; (2) to give sufficient detail and references on the adaptive mesh refinement and coarsening (AMR/ C) scheme for applications analysts and developers; and (3) to describe the parallel implementation and data structures with supporting discussion of domain decomposition, message passing, and details related to dynamic repartitioning for parallel AMR/C. Other aspects related to C++ programming paradigms, reusability for diverse applications, adaptive modeling, physics-independent error indicators, and similar concepts are briefly discussed. Finally, results from some applications using the library are presented and areas of future research are discussed.},
	language = {en},
	number = {3-4},
	urldate = {2020-08-20},
	journal = {Engineering with Computers},
	author = {Kirk, Benjamin S. and Peterson, John W. and Stogner, Roy H. and Carey, Graham F.},
	month = dec,
	year = {2006},
	pages = {237--254},
	file = {Kirk et al. - 2006 - libMesh  a C++ library for parallel adaptive mesh.pdf:/home/roberto/Zotero/storage/A94AUHE5/Kirk et al. - 2006 - libMesh  a C++ library for parallel adaptive mesh.pdf:application/pdf}
}

@techreport{balay_petsc_2016,
	address = {Lemont, IL},
	title = {{PETSc} {Users} {Manual}},
	url = {https://publications.anl.gov/anlpubs/2016/05/127241.pdf},
	number = {ANL-95/11 Rev 3.14},
	institution = {Argonne National Laboratory (ANL)},
	author = {Balay and Brune and Buschelman and Gropp and Karpeyev and Knepley and Curfman McInnes and Rupp and Smith and Zhang},
	month = apr,
	year = {2016}
}

@article{knoll_jacobian-free_2004,
	title = {Jacobian-free {Newton}–{Krylov} methods: a survey of approaches and applications},
	volume = {193},
	issn = {00219991},
	shorttitle = {Jacobian-free {Newton}–{Krylov} methods},
	url = {https://linkinghub.elsevier.com/retrieve/pii/S0021999103004340},
	doi = {10.1016/j.jcp.2003.08.010},
	abstract = {Jacobian-free Newton–Krylov (JFNK) methods are synergistic combinations of Newton-type methods for superlinearly convergent solution of nonlinear equations and Krylov subspace methods for solving the Newton correction equations. The link between the two methods is the Jacobian-vector product, which may be probed approximately without forming and storing the elements of the true Jacobian, through a variety of means. Various approximations to the Jacobian matrix may still be required for preconditioning the resulting Krylov iteration. As with Krylov methods for linear problems, successful application of the JFNK method to any given problem is dependent on adequate preconditioning. JFNK has potential for application throughout problems governed by nonlinear partial differential equations and integro-differential equations. In this survey paper, we place JFNK in context with other nonlinear solution algorithms for both boundary value problems (BVPs) and initial value problems (IVPs). We provide an overview of the mechanics of JFNK and attempt to illustrate the wide variety of preconditioning options available. It is emphasized that JFNK can be wrapped (as an accelerator) around another nonlinear fixed point method (interpreted as a preconditioning process, potentially with significant code reuse). The aim of this paper is not to trace fully the evolution of JFNK, nor to provide proofs of accuracy or optimal convergence for all of the constituent methods, but rather to present the reader with a perspective on how JFNK may be applicable to applications of interest and to provide sources of further practical information.},
	language = {en},
	number = {2},
	urldate = {2020-07-04},
	journal = {Journal of Computational Physics},
	author = {Knoll, D.A. and Keyes, D.E.},
	month = jan,
	year = {2004},
	pages = {357--397},
	file = {Knoll and Keyes - 2004 - Jacobian-free Newton–Krylov methods a survey of a.pdf:/home/roberto/Zotero/storage/YXT3BD8A/Knoll and Keyes - 2004 - Jacobian-free Newton–Krylov methods a survey of a.pdf:application/pdf}
}

@article{gaston_physics-based_2015,
	title = {Physics-based multiscale coupling for full core nuclear reactor simulation},
	volume = {84},
	issn = {03064549},
	url = {https://linkinghub.elsevier.com/retrieve/pii/S030645491400543X},
	doi = {10.1016/j.anucene.2014.09.060},
	abstract = {Numerical simulation of nuclear reactors is a key technology in the quest for improvements in efficiency, safety, and reliability of both existing and future reactor designs. Historically, simulation of an entire reactor was accomplished by linking together multiple existing codes that each simulated a subset of the relevant multiphysics phenomena. Recent advances in the MOOSE (Multiphysics Object Oriented Simulation Environment) framework have enabled a new approach: multiple domain-specific applications, all built on the same software framework, are efficiently linked to create a cohesive application. This is accomplished with a flexible coupling capability that allows for a variety of different data exchanges to occur simultaneously on high performance parallel computational hardware. Examples based on the KAIST-3A benchmark core, as well as a simplified Westinghouse AP-1000 configuration, demonstrate the power of this new framework for tackling—in a coupled, multiscale manner—crucial reactor phenomena such as CRUD-induced power shift and fuel shuffle.},
	language = {en},
	urldate = {2020-07-04},
	journal = {Annals of Nuclear Energy},
	author = {Gaston, Derek R. and Permann, Cody J. and Peterson, John W. and Slaughter, Andrew E. and Andrš, David and Wang, Yaqi and Short, Michael P. and Perez, Danielle M. and Tonks, Michael R. and Ortensi, Javier and Zou, Ling and Martineau, Richard C.},
	month = oct,
	year = {2015},
	pages = {45--54},
	file = {Gaston et al. - 2015 - Physics-based multiscale coupling for full core nu.pdf:/home/roberto/Zotero/storage/YCBH9985/Gaston et al. - 2015 - Physics-based multiscale coupling for full core nu.pdf:application/pdf}
}

@phdthesis{leppanen_development_2007,
	address = {Espoo, Finland},
	title = {Development of a {New} {Monte} {Carlo} {Reactor} {Physics} {Code}},
	abstract = {Monte Carlo neutron transport codes are widely used in various reactor physics applications, traditionally related to criticality safety analyses, radiation shielding problems, detector modelling and validation of deterministic transport codes. The main advantage of the method is the capability to model geometry and interaction physics without major approximations. The disadvantage is that the modelling of complicated systems is very computing-intensive, which restricts the applications to some extent. The importance of Monte Carlo calculation is likely to increase in the future, along with the development in computer capacities and parallel calculation.},
	language = {en},
	school = {Helsinki University of Technology},
	author = {Leppänen, Jaakko},
	month = jun,
	year = {2007},
	file = {Leppänen - Development of a New Monte Carlo Reactor Physics C.pdf:/home/roberto/Zotero/storage/PFB9JB6W/Leppänen - Development of a New Monte Carlo Reactor Physics C.pdf:application/pdf}
}

@techreport{leppanen_serpent_2015,
	title = {Serpent – a {Continuous}-energy {Monte} {Carlo} {Reactor} {Physics} {Burnup} {Calculation} {Code}},
	shorttitle = {User's {Manual}},
	url = {http://montecarlo.vtt.fi/download/Serpent_manual.pdf},
	institution = {VTT Technical Research Centre of Finland},
	author = {Leppänen, Jaakko},
	month = jun,
	year = {2015},
	file = {Leppänen - 2015 - Serpent – a Continuous-energy Monte Carlo Reactor .pdf:/home/roberto/Zotero/storage/F45I2E7H/Leppänen - 2015 - Serpent – a Continuous-energy Monte Carlo Reactor .pdf:application/pdf}
}

@misc{shenoy_htgr_2010,
	title = {{HTGR} {Technology} {Coursefor} the {Nuclear} {Regulatory} {Commission}},
	url = {https://art.inl.gov/NGNP/Training%20Modules%20%20HTGR%20Fundamentals/Module%202a%20-%20History%20and%20Evolution%20of%20HTGRs.pdf},
	author = {Shenoy, Arkal},
	month = may,
	year = {2010}
}

@article{leppanen_calculation_2014,
	title = {Calculation of effective point kinetics parameters in the {Serpent} 2 {Monte} {Carlo} code},
	volume = {65},
	issn = {0306-4549},
	url = {http://www.sciencedirect.com/science/article/pii/S0306454913005628},
	doi = {10.1016/j.anucene.2013.10.032},
	abstract = {This paper presents the methodology developed for the Serpent 2 Monte Carlo code for the calculation of adjoint-weighted reactor point kinetics parameters: effective generation time and delayed neutron fractions. The calculation routines were implemented at the Politecnico di Milano, and they are based on the iterated fission probability (IFP) method. The developed methodology is mainly intended for the modeling of small research reactor cores, and the results are validated by comparison to experimental data and MCNP5 calculations in 31 critical configurations.},
	urldate = {2016-09-14},
	journal = {Annals of Nuclear Energy},
	author = {Leppänen, Jaakko and Aufiero, Manuele and Fridman, Emil and Rachamin, Reuven and van der Marck, Steven},
	month = mar,
	year = {2014},
	keywords = {Adjoint-weighted time constants, Effective delayed neutron fraction, Effective generation time, Monte Carlo},
	pages = {272--279},
	file = {ScienceDirect Full Text PDF:/home/roberto/Zotero/storage/JIC4MNVB/Leppänen et al. - 2014 - Calculation of effective point kinetics parameters.pdf:application/pdf;ScienceDirect Snapshot:/home/roberto/Zotero/storage/YF2RDBXU/S0306454913005628.html:text/html}
}

@misc{gougar_prismatic_2010,
	title = {{PRISMATIC} {COUPLED} {NEUTRONICS}/{THERMAL} {FLUIDS} {TRANSIENT} {BENCHMARK} {OF} {THE} {MHTGR}-350 {MW} {CORE} {DESIGN}},
	author = {Gougar, Hans and Pope, Michael and Strydom, Gerhard and Baxter, Alan},
	year = {2010},
	file = {Gougar et al. - 2010 - PRISMATIC COUPLED NEUTRONICSTHERMAL FLUIDS TRANSI.pdf:/home/roberto/Zotero/storage/V5J54QP3/Gougar et al. - 2010 - PRISMATIC COUPLED NEUTRONICSTHERMAL FLUIDS TRANSI.pdf:application/pdf}
}

@techreport{hawari_development_2018,
	title = {Development and {Deployment} {Assessment} of a {Melt}-{Down} {Proof} {Modular} {Micro} {Reactor} ({MDP}-{MMR})},
	url = {http://www.osti.gov/servlets/purl/1431209/},
	language = {en},
	number = {14--6782, 1431209},
	urldate = {2019-08-01},
	institution = {North Carolina State University},
	author = {Hawari, Ayman I. and Venneri, Francesco},
	month = apr,
	year = {2018},
	doi = {10.2172/1431209},
	pages = {14--6782, 1431209},
	file = {Hawari and Venneri - 2018 - Development and Deployment Assessment of a Melt-Do.pdf:/home/roberto/Zotero/storage/VP7KT3D2/Hawari and Venneri - 2018 - Development and Deployment Assessment of a Melt-Do.pdf:application/pdf}
}

@incollection{massimo_basic_1976,
	title = {{BASIC} {ASPECTS} {OF} {TRANSPORT} {AND} {DIFFUSION} {THEORY}},
	isbn = {978-0-08-019616-9},
	url = {https://linkinghub.elsevier.com/retrieve/pii/B9780080196169500111},
	language = {en},
	urldate = {2020-08-31},
	booktitle = {Physics of {High}-{Temperature} {Reactors}},
	publisher = {Elsevier},
	author = {Massimo, Luigi},
	year = {1976},
	doi = {10.1016/B978-0-08-019616-9.50011-1},
	pages = {18--42}
}

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

@book{white_viscous_2006,
	edition = {Third},
	title = {Viscous {Fluid} {Flow}},
	publisher = {McGraw Hill},
	author = {White, F},
	year = {2006}
}

@article{churchill_friction-factor_1977,
	title = {Friction-{Factor} {Equation} {Spans} {All} {Fluid}-{Flow} {Regimes}},
	journal = {Chemical Engineering Journal},
	author = {Churchill, Stuart W.},
	year = {1977},
	file = {_.pdf:/home/roberto/Zotero/storage/3Y7NP6RA/_.pdf:application/pdf}
}

@techreport{hawari_development_2018,
	title = {Development and {Deployment} {Assessment} of a {Melt}-{Down} {Proof} {Modular} {Micro} {Reactor} ({MDP}-{MMR})},
	url = {http://www.osti.gov/servlets/purl/1431209/},
	language = {en},
	number = {14--6782, 1431209},
	urldate = {2020-10-24},
	author = {Hawari, Ayman I. and Venneri, Francesco},
	month = apr,
	year = {2018},
	doi = {10.2172/1431209},
	pages = {14--6782, 1431209},
	file = {Hawari and Venneri - 2018 - Development and Deployment Assessment of a Melt-Do.pdf:/home/roberto/Zotero/storage/A5VGTWV2/Hawari and Venneri - 2018 - Development and Deployment Assessment of a Melt-Do.pdf:application/pdf}
}

@techreport{stainsby_investigation_2008,
	title = {Investigation of {Local} {Heat} {Transfer} {Phenomena} in a {Prismatic} {Modular} {Reactor} {Core}},
	institution = {AMEC NSS Limited},
	author = {Stainsby, R.},
	year = {2008},
	file = {Stainsby - 2008 - Investigation of Local Heat Transfer Phenomena in .pdf:/home/roberto/Zotero/storage/9G5H2N4M/Stainsby - 2008 - Investigation of Local Heat Transfer Phenomena in .pdf:application/pdf}
}

@book{hetrick_dynamics_1993,
	edition = {1st},
	title = {Dynamics of {Nuclear} {Reactors}},
	isbn = {978-0-89448-453-7},
	language = {en},
	publisher = {American Nuclear Society},
	author = {Hetrick, D},
	year = {1993}
}

% CHAPTER 4: NEUTRONICS
@article{geuzaine_gmsh_2009,
	title = {Gmsh: a three-dimensional finite element mesh generator with built-in pre- and post-processing facilities.},
	copyright = {John Wiley and Sons, Ltd.},
	url = {https://gmsh.info/},
	journal = {International Journal for Numerical Methods in Engineering},
	author = {Geuzaine, Christophe and Remacle, Jean-François},
	year = {2009}
}

@techreport{marleau_user_2016,
	address = {Montreal QC, Canada},
	type = {{TECHNICAL} {REPORT}},
	title = {A {USER} {GUIDE} {FOR} {DRAGON} {VERSION4}},
	url = {https://www.polymtl.ca/merlin/downloads/IGE294.pdf},
	language = {en},
	number = {IGE–294},
	institution = {Ecole Polytechnique de Montreal},
	author = {Marleau, G and Hebert, A and Roy, R},
	month = aug,
	year = {2016},
	pages = {318},
	file = {Marleau et al. - A USER GUIDE FOR DRAGON VERSION4.pdf:/home/roberto/Zotero/storage/PA9XVA2N/Marleau et al. - A USER GUIDE FOR DRAGON VERSION4.pdf:application/pdf}
}

@mastersthesis{pater_multiphysics_2019,
	address = {Copenhagen, Denmark},
	title = {Multiphysics simulations of {Molten} {Salt} {Reactors} using the {Moltres} code},
	school = {Universite Paris-Saclay},
	author = {Pater, Mateusz},
	month = sep,
	year = {2019}
}





% CHAPTER 5: THERMAL-FLUIDS

@incollection{lemmon_thermophysical_2019,
	address = {Gaithersburg, MD},
	title = {Thermophysical {Properties} of {Fluid} {Systems}},
	booktitle = {{NIST} {Chemistry} {WebBook}, {NIST} {Standard} {Reference} {Database} {Number} 69},
	publisher = {National Institute of Standards and Technology},
	author = {Lemmon, E.W. and McLinden, M.O. and Friend, D.G.},
	editor = {Linstrom, P.J. and Mallard, W.G.},
	year = {2019}
}




% CHAPTER 6: HYDROGEN
@article{burke_impact_2018,
	title = {The {Impact} of {Electricity} on {Economic} {Development}: {A} {Macroeconomic} {Perspective}},
	volume = {12},
	issn = {19321473},
	shorttitle = {The {Impact} of {Electricity} on {Economic} {Development}},
	url = {http://www.nowpublishers.com/article/Details/IRERE-0101},
	doi = {10.1561/101.00000101},
	abstract = {We find that electricity use and access are strongly correlated with economic development, as theory would suggest. Despite large empirical literatures and suggestive case evidence, there are, however, few methodologically strong studies that establish causal effects on an economy-wide basis. There is some evidence that reliability of electricity supply is important for economic growth. We propose that future research focuses on identifying the causal effects of electricity reliability, infrastructure, and access on economic growth; testing the replicability of the literature; and deepening our theoretical understanding of how lack of availability of electricity can be a constraint to growth.},
	language = {en},
	number = {1},
	urldate = {2020-05-09},
	journal = {International Review of Environmental and Resource Economics},
	author = {Burke, Paul J. and Stern, David I. and Bruns, Stephan B.},
	month = nov,
	year = {2018},
	pages = {85--127},
	file = {Burke et al. - 2018 - The Impact of Electricity on Economic Development.pdf:/home/roberto/Zotero/storage/CFML8PAU/Burke et al. - 2018 - The Impact of Electricity on Economic Development.pdf:application/pdf}
}

@misc{us_epa_sources_2020,
	title = {Sources of {Greenhouse} {Gas} {Emissions}},
	note = {https://www.epa.gov/ghgemissions/sources-greenhouse-gas-emissions},
	language = {en},
	journal = {Sources of Greenhouse Gas Emissions},
	author = {US EPA},
	month = jan,
	year = {2020}
}

@techreport{ming_resource_2019,
	address = {San Francisco, CA},
	title = {Resource {Adequacy} in the {Pacific} {Northwest}},
	url = {https://www.ethree.com/wp-content/uploads/2019/03/E3_Resource_Adequacy_in_the_Pacific-Northwest_March_2019.pdf},
	language = {en},
	urldate = {2020-05-09},
	institution = {Energy and Environmental Economics, Inc.},
	author = {Ming, Zach and Olson, Arne and Jiang, Huai and Mogadali, Manohar and Schlag, Nick},
	month = mar,
	year = {2019},
	file = {Ming et al. - 2019 - Resource Adequacy in the Pacific Northwest.pdf:/home/roberto/Zotero/storage/N4ZU4BPB/Ming et al. - 2019 - Resource Adequacy in the Pacific Northwest.pdf:application/pdf}
}

@misc{us_department_of_energy_confronting_2017,
	title = {Confronting the {Duck} {Curve}: {How} to {Address} {Over}-{Generation} of {Solar} {Energy}},
	note = {https://www.energy.gov/eere/articles/confronting-duck-curve-how-address-over-generation-solar-energy},
	urldate = {2020-05-09},
	journal = {Confronting the Duck Curve: How to Address Over-Generation of Solar Energy},
	author = {U.S. Department of Energy},
	month = oct,
	year = {2017}
}

@misc{bouillon_prepared_2014,
	title = {Prepared {Statement} of {Brad} {Bouillon} on behalf of the {California} {Independent} {System} {Operator} {Corporation}},
	urldate = {2020-05-09},
	publisher = {Federal Energy Rregulatory Commission},
	author = {Bouillon, Brad},
	month = jun,
	year = {2014},
	file = {Bouillon - 2014 - Prepared Statement of Brad Bouillon on behalf of t.pdf:/home/roberto/Zotero/storage/2XKN48ZV/Bouillon - 2014 - Prepared Statement of Brad Bouillon on behalf of t.pdf:application/pdf}
}

@techreport{nagashima_japans_2018,
	title = {{JAPAN}'s {HYDROGEN} {STRATEGY} {AND} {ITS} {ECONOMIC} {AND} {GEOPOLITICAL} {IMPLICATIONS}},
	url = {https://www.ifri.org/sites/default/files/atoms/files/nagashima_japan_hydrogen_2018_.pdf},
	institution = {IFRI},
	author = {Nagashima, Monica},
	month = oct,
	year = {2018}
}

@techreport{isee_illinois_2015,
	address = {Urbana, IL},
	type = {Full {Report}},
	title = {Illinois {Climate} {Action} {Plan} ({iCAP})},
	url = {https://sustainability.illinois.edu/campus-sustainability/icap/},
	abstract = {This Plan represents a roadmap to a new, prosperous, and sustainable future for the University of Illinois at Urbana-Champaign campus. It outlines strategies, initiatives, and targets toward meeting the stated goal of carbon neutrality by 2050.},
	language = {en-US},
	number = {2015},
	urldate = {2019-07-03},
	institution = {University of Illinois at Urbana-Champaign},
	author = {{iSEE}},
	year = {2015},
	file = {Institute for Sustainability, Energy, and Environment - 2015 - Illinois Climate Action Plan (iCAP).pdf:/home/roberto/Zotero/storage/3LANJYFK/Institute for Sustainability, Energy, and Environment - 2015 - Illinois Climate Action Plan (iCAP).pdf:application/pdf;Snapshot:/home/roberto/Zotero/storage/XRF3SWYQ/icap.html:text/html;2015 - Illlinois Climate Action Plan.pdf:/home/roberto/Zotero/storage/XUELX8V7/2015 - Illlinois Climate Action Plan.pdf:application/pdf}
}

@misc{hi2h2_highly_2007,
	title = {Highly {Efficient}, {High} {Temperature}, {Hydrogen} {Production} by {Water} {Electrolysis}},
	url = {http://www.hi2h2.com/elec-heat_curve.htm},
	urldate = {2020-05-09},
	journal = {Highly Efficient, High Temperature, Hydrogen Production by Water Electrolysis},
	author = {Hi2H2},
	month = jan,
	year = {2007}
}

@techreport{j_e_obrien_high_2010,
	title = {High {Temperature} {Electrolysis} for {Hydrogen} {Production} from {Nuclear} {Energy} {TechnologySummary}},
	shorttitle = {High {Temperature} {Electrolysis} for {Hydrogen} {Production} from {Nuclear} {Energy} },
	url = {http://www.osti.gov/servlets/purl/978368-eWxk5T/},
	abstract = {The Department of Energy, Office of Nuclear Energy, has requested that a Hydrogen Technology Down-Selection be performed to identify the hydrogen production technology that has the best potential for timely commercial demonstration and for ultimate deployment with the Next Generation Nuclear Plant (NGNP). An Independent Review Team (IRT) has been assembled to execute the down-selection. This report has been prepared to provide the members of the Independent Review Team with detailed background information on the High Temperature Electrolysis (HTE) process, hardware, and state of the art. The Idaho National Laboratory has been serving as the lead lab for HTE research and development under the Nuclear Hydrogen Initiative. The INL HTE program has included small-scale experiments, detailed computational modeling, system modeling, and technology demonstration. Aspects of all of these activities are included in this report. In terms of technology demonstration, the INL successfully completed a 1000-hour test of the HTE Integrated Laboratory Scale (ILS) technology demonstration experiment during the fall of 2008. The HTE ILS achieved a hydrogen production rate in excess of 5.7 Nm3/hr, with a power consumption of 18 kW. This hydrogen production rate is far larger than has been demonstrated by any of the thermochemical or hybrid processes to date.},
	language = {en},
	number = {INL/EXT-09-16140, 978368},
	urldate = {2020-05-11},
	institution = {INL},
	author = {{J. E. O'Brien} and {C. M. Stoots} and {J. S. Herring} and {M. G. McKellar} and {E. A. Harvego} and {M. S. Sohal} and {K. G. Condie}},
	month = feb,
	year = {2010},
	doi = {10.2172/978368},
	pages = {INL/EXT--09--16140, 978368},
	file = {J. E. O'Brien et al. - 2010 - High Temperature Electrolysis for Hydrogen Product.pdf:/home/roberto/Zotero/storage/7ISBKZ53/J. E. O'Brien et al. - 2010 - High Temperature Electrolysis for Hydrogen Product.pdf:application/pdf}
}

@misc{doe_office_of_energy_efficiency_and_renewable_energy_hydrogen_2020,
	title = {Hydrogen {Production} {Processes}},
	note = {https://www.energy.gov/eere/fuelcells/hydrogen-production-processes},
	abstract = {Hydrogen Production Processes},
	language = {en},
	urldate = {2020-06-30},
	journal = {Energy.gov},
	author = {DOE Office of Energy Efficiency {and} Renewable Energy},
	year = {2020},
	file = {Snapshot:/home/roberto/Zotero/storage/MXS5Z63R/hydrogen-production-processes.html:text/html}
}

@misc{harper_us_2012,
	title = {U.{S}. {Department} of {Energy} - {Hydrogen} from {Natural} {Gas} and {Coal}: {The} {Road} to a {Sustainable} {Energy} {Future}},
	author = {Harper, H. Wayne},
	year = {2012},
	file = {Harper - 2012 - U.S. Department of Energy - Hydrogen from Natural .pdf:/home/roberto/Zotero/storage/36I933LJ/Harper - 2012 - U.S. Department of Energy - Hydrogen from Natural .pdf:application/pdf}
}

@article{el-shafie_hydrogen_2019,
	title = {Hydrogen {Production} {Technologies} {Overview}},
	volume = {07},
	issn = {2327-588X, 2327-5901},
	url = {http://www.scirp.org/journal/doi.aspx?DOI=10.4236/jpee.2019.71007},
	doi = {10.4236/jpee.2019.71007},
	abstract = {Hydrogen energy became the most significant energy as the current demand gradually starts to increase. Hydrogen energy is an important key solution to tackle the global temperature rise. The key important factor of hydrogen production is the hydrogen economy. Hydrogen production technologies are commercially available, while some of these technologies are still under development. This paper reviews the hydrogen production technologies from both fossil and non-fossil fuels such as (steam reforming, partial oxidation, auto thermal, pyrolysis, and plasma technology). Additionally, water electrolysis technology was reviewed. Water electrolysis can be combined with the renewable energy to get eco-friendly technology. Currently, the maximum hydrogen fuel productions were registered from the steam reforming, gasification, and partial oxidation technologies using fossil fuels. These technologies have different challenges such as the total energy consumption and carbon emissions to the environment are still too high. A novel non-fossil fuel method [ammonia NH3] for hydrogen production using plasma technology was reviewed. Ammonia decomposition using plasma technology without and with a catalyst to produce pure hydrogen was considered as compared case studies. It was showed that the efficiency of ammonia decomposition using the catalyst was higher than ammonia decomposition without the catalyst. The maximum hydrogen energy efficiency obtained from the developed ammonia decomposition system was 28.3\% with a hydrogen purity of 99.99\%. The development of ammonia decomposition processes is continues for hydrogen production, and it will likely become commercial and be used as a pure hydrogen energy source.},
	language = {en},
	number = {01},
	urldate = {2020-03-16},
	journal = {Journal of Power and Energy Engineering},
	author = {El-Shafie, Mostafa and Kambara, Shinji and Hayakawa, Yukio},
	year = {2019},
	pages = {107--154},
	file = {El-Shafie et al. - 2019 - Hydrogen Production Technologies Overview.pdf:/home/roberto/Zotero/storage/UI2YHAQX/El-Shafie et al. - 2019 - Hydrogen Production Technologies Overview.pdf:application/pdf}
}

@article{kalamaras_hydrogen_2013,
	title = {Hydrogen {Production} {Technologies}: {Current} {State} and {Future} {Developments}},
	volume = {2013},
	issn = {2314-4009},
	shorttitle = {Hydrogen {Production} {Technologies}},
	url = {http://www.hindawi.com/archive/2013/690627/},
	doi = {10.1155/2013/690627},
	abstract = {Hydrogen (H
              2
              ) is currently used mainly in the chemical industry for the production of ammonia and methanol. Nevertheless, in the near future, hydrogen is expected to become a significant fuel that will largely contribute to the quality of atmospheric air. Hydrogen as a chemical element (H) is the most widespread one on the earth and as molecular dihydrogen (H
              2
              ) can be obtained from a number of sources both renewable and nonrenewable by various processes. Hydrogen global production has so far been dominated by fossil fuels, with the most significant contemporary technologies being the steam reforming of hydrocarbons (e.g., natural gas). Pure hydrogen is also produced by electrolysis of water, an energy demanding process. This work reviews the current technologies used for hydrogen (H
              2
              ) production from both fossil and renewable biomass resources, including reforming (steam, partial oxidation, autothermal, plasma, and aqueous phase) and pyrolysis. In addition, other methods for generating hydrogen (e.g., electrolysis of water) and purification methods, such as desulfurization and water-gas shift reactions are discussed.},
	language = {en},
	urldate = {2020-03-08},
	journal = {Conference Papers in Energy},
	author = {Kalamaras, Christos M. and Efstathiou, Angelos M.},
	year = {2013},
	pages = {1--9},
	file = {Kalamaras and Efstathiou - 2013 - Hydrogen Production Technologies Current State an.pdf:/home/roberto/Zotero/storage/QHZNQISH/Kalamaras and Efstathiou - 2013 - Hydrogen Production Technologies Current State an.pdf:application/pdf}
}

@book{cea_gas-cooled_2006,
	address = {Paris},
	title = {Gas-cooled nuclear reactors},
	isbn = {978-2-281-11343-3},
	language = {en},
	publisher = {CEA},
	editor = {CEA},
	year = {2006},
	note = {OCLC: 123900736},
	keywords = {Gas cooled reactors},
	file = {France - 2006 - Gas-cooled nuclear reactors.pdf:/home/roberto/Zotero/storage/A55T7BMW/France - 2006 - Gas-cooled nuclear reactors.pdf:application/pdf}
}

@article{yildiz_efficiency_2006,
	title = {Efficiency of hydrogen production systems using alternative nuclear energy technologies},
	volume = {31},
	issn = {03603199},
	url = {https://linkinghub.elsevier.com/retrieve/pii/S0360319905000583},
	doi = {10.1016/j.ijhydene.2005.02.009},
	abstract = {Nuclear energy can be used as the primary energy source in centralized hydrogen production through high-temperature thermochemical processes, water electrolysis, or high-temperature steam electrolysis. Energy efficiency is important in providing hydrogen economically and in a climate friendly manner. High operating temperatures are needed for more efficient thermochemical and electrochemical hydrogen production using nuclear energy. Therefore, high-temperature reactors, such as the gas-cooled, molten-salt-cooled and liquid-metal-cooled reactor technologies, are the candidates for use in hydrogen production. Several candidate technologies that span the range from well developed to conceptual are compared in our analysis. Among these alternatives, high-temperature steam electrolysis (HTSE) coupled to an advanced gas reactor cooled by supercritical CO2 (S-CO2) and equipped with a supercritical CO2 power conversion cycle has the potential to provide higher energy efficiency at a lower temperature range than the other alternatives.},
	language = {en},
	number = {1},
	urldate = {2020-06-30},
	journal = {International Journal of Hydrogen Energy},
	author = {Yildiz, B and Kazimi, M},
	month = jan,
	year = {2006},
	pages = {77--92},
	file = {Yildiz and Kazimi - 2006 - Efficiency of hydrogen production systems using al.pdf:/home/roberto/Zotero/storage/SILUYPEW/Yildiz and Kazimi - 2006 - Efficiency of hydrogen production systems using al.pdf:application/pdf}
}

@techreport{benjamin_russ_sulfur_2009,
	title = {Sulfur {Iodine} {Process} {Summary} for the {Hydrogen} {Technology} {Down}-{Selection}},
	url = {http://www.osti.gov/servlets/purl/1047207/},
	language = {en},
	number = {INL/EXT-12-25773, 1047207},
	urldate = {2019-11-28},
	institution = {INL},
	author = {{Benjamin Russ}},
	month = may,
	year = {2009},
	doi = {10.2172/1047207},
	file = {Benjamin Russ - 2009 - Sulfur Iodine Process Summary for the Hydrogen Tec.pdf:/home/roberto/Zotero/storage/3PX424SU/Benjamin Russ - 2009 - Sulfur Iodine Process Summary for the Hydrogen Tec.pdf:application/pdf}
}

@techreport{brown_high_2003,
	title = {{HIGH} {EFFICIENCY} {GENERATION} {OF} {HYDROGEN} {FUELS} {USING} {NUCLEAR} {POWER}},
	url = {https://www.osti.gov/biblio/814014},
	abstract = {The U.S. Department of Energy's Office of Scientific and Technical Information},
	language = {English},
	urldate = {2018-05-17},
	institution = {GENERAL ATOMICS (US)},
	author = {Brown, Lc and Besenbruch, Ge and Lentsch, Rd and Schultz, Kr and Funk, Jf and Pickard, Ps and Marshall, Ac and Showalter, Sk},
	month = jun,
	year = {2003},
	doi = {10.2172/814014},
	file = {Full Text PDF:/home/roberto/Zotero/storage/XC4LF8Y3/Brown et al. - 2003 - HIGH EFFICIENCY GENERATION OF HYDROGEN FUELS USING.pdf:application/pdf;Snapshot:/home/roberto/Zotero/storage/IMPB86HE/814014.html:text/html}
}

@techreport{us-doe_ultimate_2019,
	address = {Washington D.C.},
	type = {Fact {Sheet}},
	title = {The {Ultimate} {Fast} {Facts} {Guide} to {Nuclear} {Energy}},
	shorttitle = {Nuclear {Fast} {Facts}},
	url = {https://www.energy.gov/ne/downloads/ultimate-fast-facts-guide-nuclear-energy},
	abstract = {The Ultimate Fast Facts Guide to Nuclear Energy},
	language = {en},
	number = {DOE/NE-0150},
	urldate = {2019-10-17},
	institution = {Department of Energy Office of Nuclear Energy},
	author = {US-DOE},
	month = jan,
	year = {2019},
	note = {https://www.energy.gov/ne/downloads/ultimate-fast-facts-guide-nuclear-energy},
	pages = {16},
	file = {Snapshot:/home/roberto/Zotero/storage/ESI5XYZY/ultimate-fast-facts-guide-nuclear-energy.html:text/html;Ultimate Fast Facts Guide-PRINT.pdf:/home/roberto/Zotero/storage/59S4TUKG/Ultimate Fast Facts Guide-PRINT.pdf:application/pdf}
}

@inproceedings{dotson_optimal_2020,
	address = {Raleigh, N.C.},
	title = {Optimal {Sizing} of a {Micro}-{Reactor} for {Embedded} {Grid} {Systems}},
	abstract = {There has been significant work recently to develop en- ergy system concepts that incorporate a mixture of nuclear energy, variable renewable energy (VRE) and energy storage techniques. These systems are often referred to as nuclear hy- brid energy systems (NHES) and are generally robust, reliable, economically appealing, and have low to zero greenhouse gas (GHG) emissions [1, 2, 3, 4]. In 2015, the University of Illi- nois at Urbana-Champaign (UIUC) made a commitment to reach carbon neutrality by 2050 in the Illinois Climate Ac- tion Plan (iCAP) [5]. This plan identified nuclear energy as a potential contributor to this goal. In this work we use the RAVEN framework to find the optimal size for a micro-reactor that would minimize the levelized cost of electricity (LCOE) for the UIUC embedded grid. The greatest source economic variability is contained in the capital costs of a micro-reactor. Thus we examine both scenarios where the reactor is provided at no cost to the university and purchased at full price.},
	booktitle = {Transactions of the {American} {Nuclear} {Society} {Student} {Conference}},
	publisher = {American Nuclear Society},
	author = {Dotson, Samuel G. and Huff, Kathryn D.},
	month = mar,
	year = {2020},
	file = {Dotson and Huff - 2020 - Optimal Sizing of a Micro-Reactor for Embedded Gri.pdf:/home/roberto/Zotero/storage/XL93Z3I5/Dotson and Huff - 2020 - Optimal Sizing of a Micro-Reactor for Embedded Gri.pdf:application/pdf}
}

@misc{world_nuclear_association_small_2020,
	title = {Small nuclear power reactors - {World} {Nuclear} {Association}},
	note = {https://www.world-nuclear.org/information-library/nuclear-fuel-cycle/nuclear-power-reactors/small-nuclear-power-reactors.aspx},
	urldate = {2020-07-01},
	author = {World Nuclear Association},
	month = jun,
	year = {2020},
	file = {Small nuclear power reactors - World Nuclear Association:/home/roberto/Zotero/storage/LPCHKT5I/small-nuclear-power-reactors.html:text/html}
}

@misc{mtd_irecords_2019,
	title = {{Champaign}-{Urbana} {Mass} {Transit} {District} {Public} {Records}},
	language = {en},
	author = {Champaign-Urbana Mass Transit District},
	note = {https://mtd.org/inside/public-information/documents/},
	month = dec,
	year = {2019}
}

@misc{uiuc_personnal_communication,
	title = {{Personal} {Communication}},
	language = {en},
	author = {Pete Varney},
	month = jan,
	year = {2020}
}

@misc{alternative_fuels_data_center_fuel_2014,
	title = {Fuel {Properties} {Comparison}},
	note = {https://afdc.energy.gov/fuels/properties},
	language = {en},
	author = {Alternative Fuels Data Center},
	month = oct,
	year = {2014},
	file = {Fuel Properties Comparison.pdf:/home/roberto/Zotero/storage/JCQKISRJ/Fuel Properties Comparison.pdf:application/pdf}
}

@online{usnc_mmr_2019,
	title = {{MMR} - {USNC}},
	copyright = {USNC},
	shorttitle = {{MMR}},
	note = {https://usnc.com/MMR.html},
	language = {en},
	journal = {MMR™ Energy System},
	author = {USNC},
	year = {2019}
}

@article{hernandez_micro_2019,
	title = {Micro heat pipe nuclear reactor concepts: {Analysis} of fuel cycle performance and environmental impacts},
	volume = {126},
	issn = {0306-4549},
	shorttitle = {Micro heat pipe nuclear reactor concepts},
	url = {http://www.sciencedirect.com/science/article/pii/S0306454918306509},
	doi = {10.1016/j.anucene.2018.11.050},
	abstract = {The demand for more practical and innovative nuclear reactor designs capable of providing clean energy to meet new global demands is revolutionizing the nuclear industry. In this paper we present the fuel cycle and neutronics analysis of a design expected to be similar to the Westinghouse eVinci™ heat pipe (HP) reactor. The concept considered here uses low enriched urania rods and potassium liquid metal-cooled HPs to remove heat from the core. The fuel and HPs are contained in steel monolith to form the core which is surrounded by an alumina reflector. There is no need for forced circulation to remove heat from the active core region, thus eliminating the use of pumps, valves and tube piping. The concept is designed to operate safely and provide a reliable autonomous power supply in support of off-grid missions. This study shows that the HP reactor design can operate for more than 10 years without refueling. Further, we show that the current design of the HP reactor concept is best suited to serve as a nuclear battery rather than a centralized power source. We also studied the nuclear fuel cycle performance of the HP reactor concept in a once-through fuel cycle. The natural resource utilization, waste output, and environmental impact, as defined in a reference study from the literature, did not perform as well on a GWe per year energy basis compared to light water reactors with less than 5\% enriched uranium in a once-through fuel cycle. Parasitic absorption in the steel monolith structure and neutron leakage lower the reactivity of the core, which decreases the discharge burnup of the fuel. Analysis of modifications of the monolith material and size of the reference core configuration showed that the neutron leakage impact on the small sized HP core is the most limiting factor on the fuel cycle performance.},
	urldate = {2019-07-25},
	journal = {Annals of Nuclear Energy},
	author = {Hernandez, Richard and Todosow, Michael and Brown, Nicholas R.},
	month = apr,
	year = {2019},
	keywords = {Fuel cycle performance, Heat pipe reactor, Modular microreactor},
	pages = {419--426},
	file = {Hernandez et al. - 2019 - Micro heat pipe nuclear reactor concepts Analysis.pdf:/home/roberto/Zotero/storage/YJCBGFEG/Hernandez et al. - 2019 - Micro heat pipe nuclear reactor concepts Analysis.pdf:application/pdf;ScienceDirect Snapshot:/home/roberto/Zotero/storage/QQ7YJY8K/S0306454918306509.html:text/html}
}

@misc{harlan_x-energy_2018,
	address = {Rockville, MD},
	title = {X-energy {Xe}-100 {Reactor} initial {NRC} meeting},
	url = {https://adamswebsearch2.nrc.gov/webSearch2/main.jsp?AccessionNumber=ML18253A109},
	language = {English},
	urldate = {2019-09-20},
	author = {Harlan, Bowers},
	month = sep,
	year = {2018},
	file = {X-energy Xe-100 Reactor initial NRC meeting:/home/roberto/Zotero/storage/HC5YWNJD/X-energy Xe-100 Reactor initial NRC meeting.pdf:application/pdf}
}

@techreport{ding_design_2011,
	title = {Design of a {U}-{Battery}},
	url = {https://www.u-battery.com/_/uploads/AExecSummary.pdf},
	abstract = {This report presents a feasibility study for the design of an intrinsically safe modular nuclear power
generation system that combines quick-developed till commercial design using proven technology with the
basic features to profit from economy of number. The investigation shows that the proposed 10MWth U-
Battery® design is very promising to fulfill all the above requirements.
The study is executed by the Delft University of Technology together with the University of Manchester and
is sponsored by URENCO and Koopman \& Witteveen.},
	number = {PNR-131-2011-014},
	institution = {Urenco, and Koopman and Witteveen},
	author = {Ding, Ming and Kloosterman, J. L. and Kooijman, Theo and Linssen, Rik},
	month = nov,
	year = {2011},
	file = {Ding et al. - 2011 - Design of a U-Battery.pdf:/home/roberto/Zotero/storage/2KGCXW7Y/Ding et al. - 2011 - Design of a U-Battery.pdf:application/pdf}
}

@misc{star_core_nuclear_star_2015,
	title = {Star {Core} {Spec} {Sheet}},
	url = {https://starcorenuclear.ca/wp-content/uploads/2015/12/StarCoreSpecSheet.pdf},
	urldate = {2020-06-01},
	publisher = {starcore-nuclear},
	author = {StarCore-Nuclear},
	month = dec,
	year = {2015},
	file = {Star Core Nuclear - 2015 - Star Core Spec Sheet.pdf:/home/roberto/Zotero/storage/EHHUK5ZX/Star Core Nuclear - 2015 - Star Core Spec Sheet.pdf:application/pdf}
}

@misc{energy_information_administration_how_2014,
	title = {How much carbon dioxide is produced by burning gasoline and diesel fuel?},
	note = {http://www.patagoniaalliance.org/wp-content/uploads/2014/08/How-much-carbon-dioxide-is-produced-by-burning-gasoline-and-diesel-fuel-FAQ-U.S.-Energy-Information-Administration-EIA.pdf},
	publisher = {US Energy Information Administration},
	author = {Energy Information Administration},
	month = apr,
	year = {2014},
	file = {Energy Information Administration - 2014 - How much carbon dioxide is produced by burning gas.pdf:/home/roberto/Zotero/storage/IWXJSWBX/Energy Information Administration - 2014 - How much carbon dioxide is produced by burning gas.pdf:application/pdf}
}

@misc{us_energy_information_administration_electric_2020,
	title = {Electric {Power} {Monthly} with data for {February} 2020},
	url = {https://www.eia.gov/electricity/monthly/epm_table_grapher.php?t=epmt_6_02_a},
	language = {en},
	urldate = {2020-04-26},
	author = {US Energy Information Administration},
	month = apr,
	year = {2020},
	file = {2020 - Electric Power Monthly with data for February 2020.pdf:/home/roberto/Zotero/storage/6X9V3QQ7/2020 - Electric Power Monthly with data for February 2020.pdf:application/pdf}
}

@techreport{usdrive_hydrogen_2017,
	title = {Hydrogen {Production} {Tech} {Team} {Roadmap}},
	url = {https://www.energy.gov/sites/prod/files/2017/11/f46/HPTT%20Roadmap%20FY17%20Final_Nov%202017.pdf},
	institution = {US-DOE},
	author = {USDRIVE},
	month = nov,
	year = {2017}
}

@misc{helmeth_high_2020,
	title = {High temperature electrolysis cell ({SOEC})},
	note = {http://www.helmeth.eu/index.php/},
	urldate = {2020-02-10},
	journal = {High temperature electrolysis cell (SOEC)},
	author = {HELMETH},
	month = feb,
	year = {2020},
	file = {High temperature electrolysis cell (SOEC):/home/roberto/Zotero/storage/VUY6L5H7/high-temperature-electrolysis-cell-soec.html:text/html}
}

@inproceedings{obrien_status_2010,
	title = {Status of the {INL} {High}- {Temperature} {Electrolysis} {Research} {Program} – {Experimental} and {Modeling}},
	url = {https://inis.iaea.org/search/search.aspx?orig_q=RN:41083941},
	abstract = {This paper provides a status update on the high-temperature electrolysis (HTE) research and development program at the Idaho National Laboratory (INL), with an overview of recent large-scale system modeling results and the status of the experimental program. System analysis results have been obtained using the commercial code UniSim, augmented with a custom high-temperature electrolyzer module. The process flow diagrams for the system simulations include an advanced nuclear reactor as a source of high-temperature process heat, a power cycle and a coupled steam electrolysis loop. Several reactor types and power cycles have been considered, over a range of reactor coolant outlet temperatures.},
	language = {en},
	booktitle = {Fourth {Information} {Exchange} {Meeting} on the {Nuclear} {Production} of {Hydrogen}},
	author = {O’Brien, J E and Herring, J S and Stoots, C M and McKellar, M G and Harvego, E A and Condie, K G and Housley, G K and Hartvigsen, J J},
	year = {2010},
	pages = {13},
	file = {O’Brien et al. - Status of the INL High- Temperature Electrolysis R.pdf:/home/roberto/Zotero/storage/3YF7CUCX/O’Brien et al. - Status of the INL High- Temperature Electrolysis R.pdf:application/pdf}
}

@article{ursua_hydrogen_2012,
	title = {Hydrogen {Production} {From} {Water} {Electrolysis}: {Current} {Status} and {Future} {Trends}},
	volume = {100},
	issn = {0018-9219},
	shorttitle = {Hydrogen {Production} {From} {Water} {Electrolysis}},
	url = {http://ieeexplore.ieee.org/document/5898382/},
	doi = {10.1109/JPROC.2011.2156750},
	abstract = {This paper reviews water electrolysis technologies for hydrogen production and also surveys the state of the art of water electrolysis integration with renewable energies. First, attention is paid to the thermodynamic and electrochemical processes to better understand how electrolysis cells work and how they can be combined to build big electrolysis modules. The electrolysis process and the characteristics, advantages, drawbacks, and challenges of the three main existing electrolysis technologies, namely alkaline, polymer electrolyte membrane, and solid oxide electrolyte, are then discussed. Current manufacturers and the main features of commercially available electrolyzers are extensively reviewed. Finally, the possible configurations allowing the integration of water electrolysis units with renewable energy sources in both autonomous and grid-connected systems are presented and some relevant demonstration projects are commented.},
	language = {en},
	number = {2},
	urldate = {2020-05-09},
	journal = {Proceedings of the IEEE},
	author = {Ursua, Alfredo and Gandia, Luis M. and Sanchis, Pablo},
	month = feb,
	year = {2012},
	pages = {410--426},
	file = {Ursua et al. - 2012 - Hydrogen Production From Water Electrolysis Curre.pdf:/home/roberto/Zotero/storage/8EEJC7SN/Ursua et al. - 2012 - Hydrogen Production From Water Electrolysis Curre.pdf:application/pdf}
}

@techreport{doe_energy_efficiency_and_renewable_energy_fuel_2015,
	title = {Fuel {Cells} [fact sheet]},
	language = {en},
	institution = {DOE},
	author = {Energy Efficiency {and} Renewable Energy},
	month = nov,
	year = {2015},
	file = {DOE Energy Efficiency and Renewable Energy - 2015 - Fuel cells fact sheet.pdf:/home/roberto/Zotero/storage/26PXGV9Z/DOE Energy Efficiency and Renewable Energy - 2015 - Fuel cells fact sheet.pdf:application/pdf}
}