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@misc{github_build_2017,
title = {Build software better, together},
url = {https://github.com},
abstract = {GitHub is where people build software. More than 21 million people use GitHub to discover, fork, and contribute to over 58 million projects.},
urldate = {2017-05-11},
journal = {GitHub},
author = {{GitHub}},
year = {2017},
file = {Snapshot:/home/sunmyung/Zotero/storage/9UB445Q8/github.com.html:text/html},
}
@article{cammi_multi-physics_2011,
title = {A multi-physics modelling approach to the dynamics of {Molten} {Salt} {Reactors}},
volume = {38},
issn = {0306-4549},
url = {http://www.sciencedirect.com/science/article/pii/S0306454911000582},
doi = {10.1016/j.anucene.2011.01.037},
abstract = {This paper presents a multi-physics modelling (MPM) approach developed for the study of the dynamics of the Molten Salt Reactor (MSR), which has been reconsidered as one of the future nuclear power plants in the framework of the Generation IV International Forum for its several potentialities. The proposed multi-physics modelling is aimed at the description of the coupling between heat transfer, fluid dynamics and neutronics characteristics in a typical MSR core channel, taking into account the spatial effects of the most relevant physical quantities. In particular, as far as molten salt thermo-hydrodynamics is concerned, Navier–Stokes equations are used with the turbulence treatment according to the RANS (Reynolds Averaged Navier–Stokes) scheme, while the heat transfer is taken into account through the energy balance equations for the fuel salt and the graphite. As far as neutronics is concerned, the two-group diffusion theory is adopted, where the group constants (computed by means of the neutron transport code NEWT of SCALE 5.1) are included into the model in order to describe the neutron flux and the delayed neutron precursor distributions, the system time constants, and the temperature feedback effects of both graphite and fuel salt. The developed MPM approach is implemented in the unified simulation environment offered by COMSOL Multiphysics®, and is applied to study the behaviour of the system in steady-state conditions and under several transients (i.e., reactivity insertion due to control rod movements, fuel mass flow rate variations due to the change of the pump working conditions, presence of periodic perturbations), pointing out some advantages offered with respect to the conventional approaches employed in literature for the MSRs.},
number = {6},
urldate = {2013-05-28},
journal = {Annals of Nuclear Energy},
author = {Cammi, Antonio and Di Marcello, Valentino and Luzzi, Lelio and Memoli, Vito and Ricotti, Marco Enrico},
month = jun,
year = {2011},
keywords = {MSR, atws, read, unread, molten salt reactor, Multi-physics modelling, Reactor dynamics, Thermo-hydrodynamics},
pages = {1356--1372},
annote = {RANS ke, excellent review of previous work},
file = {A_multi-physics_modelling_approach_to_the_dynamics_of_Molten_Sal.mobi:/home/sunmyung/Zotero/storage/JWIMI3QI/A_multi-physics_modelling_approach_to_the_dynamics_of_Molten_Sal.mobi:application/octet-stream;cammi_multi-physics_2011.pdf:/home/sunmyung/Zotero/storage/AHXUPQ4A/cammi_multi-physics_2011.pdf:application/pdf;ScienceDirect Full Text PDF:/home/sunmyung/Zotero/storage/6FQKN2CJ/Cammi et al. - 2011 - A multi-physics modelling approach to the dynamics.pdf:application/pdf;ScienceDirect Snapshot:/home/sunmyung/Zotero/storage/63AWQD57/S0306454911000582.html:text/html},
}
@article{rouch_preliminary_2014,
title = {Preliminary thermal–hydraulic core design of the {Molten} {Salt} {Fast} {Reactor} ({MSFR})},
volume = {64},
issn = {0306-4549},
url = {http://www.sciencedirect.com/science/article/pii/S0306454913004829},
doi = {10.1016/j.anucene.2013.09.012},
abstract = {A thermal–hydraulics study of the core of the Molten Salt Fast Reactor (MSFR) is presented. The numerical simulations were carried-out using a Computation Fluid Dynamic code. The main objectives of the thermal–hydraulics studies are to design the core cavity walls in order to increase the overall flow mixing and to reduce the temperature peaking factors in the salt and on the core walls. The results of the CFD simulations show that for the chosen core design acceptable temperature distributions can be obtained by using a curved core cavity shape, inlets and outlets. The hot spot temperature is less than 10 °C above the average core outlet temperature and is located in the centre of the top wall of the core. The results show also a moderate level of sensitivity to the working point.},
urldate = {2016-08-22},
journal = {Annals of Nuclear Energy},
author = {Rouch, H. and Geoffroy, O. and Rubiolo, P. and Laureau, A. and Brovchenko, M. and Heuer, D. and Merle-Lucotte, E.},
month = feb,
year = {2014},
keywords = {CFD, Core cavity, Fuel salt temperature, MSFR, Thermal–hydraulics design},
pages = {449--456},
file = {ScienceDirect Full Text PDF:/home/sunmyung/Zotero/storage/SX2W3QA5/Rouch et al. - 2014 - Preliminary thermal–hydraulic core design of the M.pdf:application/pdf;ScienceDirect Snapshot:/home/sunmyung/Zotero/storage/8GBUHZAG/S0306454913004829.html:text/html},
}
@article{wang_molten_2006,
series = {13th {International} {Conference} on {Nuclear} {Energy13th} {International} {Conference} on {Nuclear} {Energy}},
title = {Molten salt related extensions of the {SIMMER}-{III} code and its application for a burner reactor},
volume = {236},
issn = {0029-5493},
url = {http://www.sciencedirect.com/science/article/pii/S002954930600330X},
doi = {10.1016/j.nucengdes.2006.04.022},
abstract = {Molten salt reactors (MSRs) can be used as effective burners of plutonium (Pu) and minor actinides (MAs) from light water reactor (LWR) spent fuel. In this paper a study was made to examine the thermal hydraulic behaviour of the conceptual design of the molten salt advanced reactor transmuter (MOSART) [Ignatiev, V., Feynberg, O., Myasnikov, A., Zakirov, R., 2003a. Neutronic properties and possible fuel cycle of a molten salt transmuter. Proceedings of the 2003 ANS/ENS International Winter Meeting (GLOBAL 2003), Hyatt Regency, New Orleans, LA, USA 16–20 November 2003]. The molten salt fuel is a ternary NaF–LiF–BeF2 system fuelled with ca. 1 mol\% typical compositions of transuranium-trifluorides (PuF3, etc.) from light water reactor spent fuel. The MOSART reactor core does not contain graphite structure elements to guide the flow, so the neutron spectrum is rather hard in order to improve the burning performance. Without those structure elements in the core, the molten salt in core flows freely and the flow pattern could be potentially complicated and may affect significantly the fuel temperature distribution in the core. Therefore, some optimizations of the salt flow pattern may be needed. Here, the main attention has been paid to the fluid dynamic simulations of the MOSART core with the code SIMMER-III [Kondo, Sa., Morita, K., Tobita, Y., Shirakawa, K., 1992. SIMMER-III: an advanced computer program for LMFBR severe accident analysis. Proceedings of the ANP’ 92, Tokyo, Japan; Kondo, Sa., Tobita, Y., Morita, K., Brear, D.J., Kamiyama, K., Yamano, H., Fujita, S., Maschek, W., Fischer, E.A., Kiefhaber, E., Buckel, G., Hesselschwerdt, E., Flad, M., Costa, P., Pigny, S., 1999. Current status and validation of the SIMMER-III LMFR safety analysis code. Proceedings of the ICONE-7, Tokyo, Japan], which was originally developed for the safety assessment of sodium-cooled fast reactors and recently extended by the authors for the thermo-hydraulic and neutronic models so as to describe the molten salt reactors. For the adaptation to molten salt reactor, a complete equation of state (EOS) for this liquid fuel had to be developed and implemented into the SIMMER-III code. Through those simulations it was concluded that the thermal hydraulic behaviour appeared to be very important in molten salt reactors concerning design, operation and safety. A flow distribution plate design was found effective to optimize the flow pattern in the core region. Further investigations are under way to obtain optimal flow fields without exceeding design limits.},
number = {14–16},
urldate = {2016-08-17},
journal = {Nuclear Engineering and Design},
author = {Wang, Shisheng and Rineiski, Andrei and Maschek, Werner},
month = aug,
year = {2006},
pages = {1580--1588},
file = {ScienceDirect Full Text PDF:/home/sunmyung/Zotero/storage/WI6S7IWU/Wang et al. - 2006 - Molten salt related extensions of the SIMMER-III c.pdf:application/pdf;ScienceDirect Full Text PDF:/home/sunmyung/Zotero/storage/PV9ZJET6/Wang et al. - 2006 - Molten salt related extensions of the SIMMER-III c.pdf:application/pdf;ScienceDirect Snapshot:/home/sunmyung/Zotero/storage/8J63MGCR/S002954930600330X.html:text/html;ScienceDirect Snapshot:/home/sunmyung/Zotero/storage/CEFLR9U7/S002954930600330X.html:text/html},
}
@mastersthesis{pettersen_coupled_2016,
title = {Coupled multi-physics simulations of the {Molten} {Salt} {Fast} {Reactor} using coarse-mesh thermal-hydraulics and spatial neutronics},
url = {http://samofar.eu/wp-content/uploads/2016/11/MScThesis-eirikEidePettersen.pdf},
urldate = {2016-11-29},
school = {MSc thesis, September 2016 (PDF)},
author = {Pettersen, Eirik Eide and Mikityuk, Konstantin},
year = {2016},
file = {[PDF] samofar.eu:/home/sunmyung/Zotero/storage/XVIH74PK/Pettersen and Mikityuk - 2016 - Coupled multi-physics simulations of the Molten Sa.pdf:application/pdf;Fulltext:/home/sunmyung/Zotero/storage/PA885TB5/Pettersen and Mikityuk - 2016 - Coupled multi-physics simulations of the Molten Sa.pdf:application/pdf;Snapshot:/home/sunmyung/Zotero/storage/S7SERYFF/Pettersen and Mikityuk - 2016 - Coupled multi-physics simulations of the Molten Sa.pdf:application/pdf},
}
@article{cammi_dimensional_2012,
series = {Selected and expanded papers from {International} {Conference} {Nuclear} {Energy} for {New} {Europe} 2010, {Portoro}?, {Slovenia}, {September} 6-9, 2010},
title = {Dimensional effects in the modelling of {MSR} dynamics: {Moving} on from simplified schemes of analysis to a multi-physics modelling approach},
volume = {246},
issn = {0029-5493},
shorttitle = {Dimensional effects in the modelling of {MSR} dynamics},
url = {http://www.sciencedirect.com/science/article/pii/S0029549311006273},
doi = {10.1016/j.nucengdes.2011.08.002},
abstract = {The MSR (Molten Salt Reactor) is one of the six innovative concepts of nuclear reactors envisaged by the GIF-IV (Generation IV International Forum) initiative for the long term evolution of the nuclear technology, in the direction of a more sustainable, safe, proliferation resistant, and economic power generation. The MSR is characterised by a complex and highly non-linear behaviour, which requires a careful investigation, as a consequence of some unusual features like the presence of a fluid fuel and the drift of delayed neutron precursors (DNP) along the primary circuit. In this paper, the MSR primary circuit dynamics is analysed with reference to the MSRE (Molten Salt Reactor Experiment), due to the availability of both a detailed design and experimental data. Numerical models featured by increasing complexity are presented. In particular, a zero-dimensional model is developed, and two other models introducing a one-dimensional discretization for the DNP drift and/or the heat convection are also elaborated. These simplified models are then compared with a more complex model, obtained according to an innovative Multi-Physics Modelling (MPM) approach to the dynamic analysis, where the partial differential equations governing the different phenomena in a core channel are solved in a two-dimensional domain, within the same computational environment. A one-dimensional closure of the primary circuit is also provided. The MPM approach gives a unique insight into the influence of local effects on the overall dynamic behaviour of the reactor, while the variety of developed models allows a systematic investigation about the dimensional effects in the modelling of MSRs. This work represents a starting point in the set-up of a Multi-Physics (MP) simulation tool, suitable for calculations with different degrees of accuracy and physical complexity, and paves the way towards the development of MP models capable of a point-by-point coupling of all the phenomena characterising the MSRs, and the nuclear reactors in general.},
urldate = {2017-01-04},
journal = {Nuclear Engineering and Design},
author = {Cammi, Antonio and Fiorina, Carlo and Guerrieri, Claudia and Luzzi, Lelio},
month = may,
year = {2012},
pages = {12--26},
file = {ScienceDirect Full Text PDF:/home/sunmyung/Zotero/storage/923SXTPA/Cammi et al. - 2012 - Dimensional effects in the modelling of MSR dynami.pdf:application/pdf;ScienceDirect Snapshot:/home/sunmyung/Zotero/storage/9M9IPG5I/S0029549311006273.html:text/html},
}
@article{krepel_dyn1d-msr_2005,
title = {{DYN1D}-{MSR} dynamics code for molten salt reactors},
volume = {32},
issn = {0306-4549},
url = {http://www.sciencedirect.com/science/article/pii/S0306454905001702},
doi = {10.1016/j.anucene.2005.07.007},
abstract = {This paper reports about the DYN1D-MSR code development and dynamics studies of the molten salt reactors (MSR) – one of the ‘Generation IV International Forum’ concepts. In this forum the graphite-moderated channel type MSR based on the previous Oak Ridge National Laboratory research is considered.
The liquid molten salt serves as a fuel and coolant, simultaneously and causes two physical peculiarities: the fission energy is released predominantly directly into the coolant and the delayed neutrons precursors are drifted by the fuel flow. The drift causes the spread of delayed neutrons distribution to the non-core parts of primary circuit and it can lead to a reactivity loss or gain in the case of fuel flow acceleration or deceleration, respectively. Therefore, specific 3D tool based on in house code DYN3D was developed in FZR. The code DYN3D-MSR is based on the solution of two-group neutron diffusion equation by the help of a nodal expansion method and it includes models of delayed neutrons drift and specific MSR heat release distribution.
In this paper the development and verification of 1D version DYN1D-MSR of the code is described. The code has been validated with the experimental data gained from the molten salt reactor experiment performed in the Oak Ridge and after the validation it was applied to several typical transients (overcooling of fuel at the core inlet, reactivity insertion, and the fuel pump trip).},
number = {17},
urldate = {2017-03-31},
journal = {Annals of Nuclear Energy},
author = {Krepel, Jiri and Grundmann, Ulrich and Rohde, Ulrich and Weiss, Frank-Peter},
month = nov,
year = {2005},
pages = {1799--1824},
file = {ScienceDirect Full Text PDF:/home/sunmyung/Zotero/storage/F6V3M36C/Krepel et al. - 2005 - DYN1D-MSR dynamics code for molten salt reactors.pdf:application/pdf;ScienceDirect Snapshot:/home/sunmyung/Zotero/storage/DXKS96PF/S0306454905001702.html:text/html},
}
@article{laureau_transient_2017,
title = {Transient coupled calculations of the {Molten} {Salt} {Fast} {Reactor} using the {Transient} {Fission} {Matrix} approach},
volume = {316},
issn = {0029-5493},
url = {http://www.sciencedirect.com/science/article/pii/S002954931730081X},
doi = {10.1016/j.nucengdes.2017.02.022},
abstract = {In this paper we present transient studies of the Molten Salt Fast Reactor (MSFR). This generation IV reactor is characterized by a liquid fuel circulating in the core cavity, requiring specific simulation tools. An innovative neutronic approach called “Transient Fission Matrix” is used to perform spatial kinetic calculations with a reduced computational cost through a pre-calculation of the Monte Carlo spatial and temporal response of the system. Coupled to this neutronic approach, the Computational Fluid Dynamics code OpenFOAM is used to model the complex flow pattern in the core. An accurate interpolation model developed to take into account the thermal hydraulics feedback on the neutronics including reactivity and neutron flux variation is presented. Finally different transient studies of the reactor in normal and accidental operating conditions are detailed such as reactivity insertion and load following capacities. The results of these studies illustrate the excellent behavior of the MSFR during such transients.},
number = {Supplement C},
urldate = {2017-03-18},
journal = {Nuclear Engineering and Design},
author = {Laureau, A. and Heuer, D. and Merle-Lucotte, E. and Rubiolo, P. R. and Allibert, M. and Aufiero, M.},
month = may,
year = {2017},
note = {JC0004},
keywords = {Neutronics, MSFR, Thermal hydraulics, Transient calculation, Transient Fission Matrix},
pages = {112--124},
file = {ScienceDirect Full Text PDF:/home/sunmyung/Zotero/storage/GEYY7UJ9/Laureau et al. - 2017 - Transient coupled calculations of the Molten Salt .pdf:application/pdf;ScienceDirect Full Text PDF:/home/sunmyung/Zotero/storage/K5JUZTZ6/Laureau et al. - 2017 - Transient coupled calculations of the Molten Salt .pdf:application/pdf;ScienceDirect Snapshot:/home/sunmyung/Zotero/storage/5B25JVQI/S002954931730081X.html:text/html;ScienceDirect Snapshot:/home/sunmyung/Zotero/storage/IKKCKIDB/S002954931730081X.html:text/html},
}
@misc{lindsay_moltres_2017,
address = {University of Illinois at Urbana-Champaign},
title = {Moltres, software for simulating {Molten} {Salt} {Reactors}},
shorttitle = {Moltres},
url = {https://github.com/arfc/moltres},
abstract = {arfc/moltres: Repository for Moltres, a code for simulating Molten Salt Reactors},
urldate = {2017-02-24},
author = {Lindsay, Alexander},
year = {2017},
note = {https://github.com/arfc/moltres},
file = {arfc/moltres\: Repository for Moltres, a code for simulating Molten Salt Reactors:/home/sunmyung/Zotero/storage/WGEZI8WJ/moltres.html:text/html},
}
@article{nagy_steady-state_2014,
title = {Steady-state and dynamic behavior of a moderated molten salt reactor},
volume = {64},
issn = {0306-4549},
url = {http://www.sciencedirect.com/science/article/pii/S0306454913004179},
doi = {10.1016/j.anucene.2013.08.009},
abstract = {The moderated Molten Salt Reactor (MSR) is an attractive breeder reactor. However, the temperature feedback coefficient of such a system can be positive due to the contribution of the moderator, an effect that can only be avoided with special measures. A previous study (Nagy et al., 2010) aimed to find a core design that is a breeder and has negative overall temperature feedback coefficient. In this paper, a coupled calculation scheme, which includes the reactor physics, heat transfer and fluid dynamics calculations is introduced. It is used both for steady-state and for dynamic calculations to evaluate the safety of the core design which was selected from the results of the previous study. The calculated feedback coefficients on the salt and graphite temperatures, power and uranium concentration prove that the core design derived in the previous optimization study is safe because the temperature feedback coefficient of the core and of the power is sufficiently negative. Transient calculations are performed to show the inherent safety of the reactor in case of reactivity insertion. As it is shown, the response of the reactor to these transients is initially dominated by the strong negative feedback of the salt. In all the presented transients, the reactor power stabilizes and the temperature of the salt never approaches its boiling point.},
journal = {Annals of Nuclear Energy},
author = {Nagy, K. and Lathouwers, D. and T’Joen, C. G. A. and Kloosterman, J. L. and van der Hagen, T. H. J. J.},
month = feb,
year = {2014},
keywords = {Coupled calculations, Transient calculations},
pages = {365--379},
file = {ScienceDirect Full Text PDF:/home/sunmyung/Zotero/storage/92VTQVHJ/Nagy et al. - 2014 - Steady-state and dynamic behavior of a moderated m.pdf:application/pdf;ScienceDirect Snapshot:/home/sunmyung/Zotero/storage/EAGZXEWN/S0306454913004179.html:text/html},
}
@article{leppanen_serpent_2014,
title = {The {Serpent} {Monte} {Carlo} code: {Status}, development and applications in 2013},
volume = {82},
issn = {0306-4549},
shorttitle = {The {Serpent} {Monte} {Carlo} code},
doi = {10.1016/j.anucene.2014.08.024},
abstract = {The Serpent Monte Carlo reactor physics burnup calculation code has been developed at VTT Technical Research Centre of Finland since 2004, and is currently used in over 100 universities and research organizations around the world. This paper presents the brief history of the project, together with the currently available methods and capabilities and plans for future work. Typical user applications are introduced in the form of a summary review on Serpent-related publications over the past few years.},
urldate = {2017-04-10},
journal = {Annals of Nuclear Energy},
author = {Leppanen, Jaakko and Pusa, Maria and Viitanen, Tuomas and Valtavirta, Ville and Kaltiaisenaho, Toni},
month = aug,
year = {2014},
keywords = {Reactor physics, Monte Carlo, Burnup calculation, Homogenization},
pages = {142--150},
file = {ScienceDirect Full Text PDF:/home/sunmyung/Zotero/storage/83V2KXJ9/Leppänen et al. - 2015 - The Serpent Monte Carlo code Status, development .pdf:application/pdf;ScienceDirect Full Text PDF:/home/sunmyung/Zotero/storage/VT4QRTXR/Leppänen et al. - 2015 - The Serpent Monte Carlo code Status, development .pdf:application/pdf;ScienceDirect Snapshot:/home/sunmyung/Zotero/storage/AQAB5875/S0306454914004095.html:text/html;ScienceDirect Snapshot:/home/sunmyung/Zotero/storage/9AVEULJX/S0306454914004095.html:text/html},
}
@article{gaston_physics-based_2015,
series = {Multi-{Physics} {Modelling} of {LWR} {Static} and {Transient} {Behaviour}},
title = {Physics-based multiscale coupling for full core nuclear reactor simulation},
volume = {84},
issn = {0306-4549},
url = {http://www.sciencedirect.com/science/article/pii/S030645491400543X},
doi = {10.1016/j.anucene.2014.09.060},
abstract = {Numerical simulation of nuclear reactors is a key technology in the quest for improvements in efficiency, safety, and reliability of both existing and future reactor designs. Historically, simulation of an entire reactor was accomplished by linking together multiple existing codes that each simulated a subset of the relevant multiphysics phenomena. Recent advances in the MOOSE (Multiphysics Object Oriented Simulation Environment) framework have enabled a new approach: multiple domain-specific applications, all built on the same software framework, are efficiently linked to create a cohesive application. This is accomplished with a flexible coupling capability that allows for a variety of different data exchanges to occur simultaneously on high performance parallel computational hardware. Examples based on the KAIST-3A benchmark core, as well as a simplified Westinghouse AP-1000 configuration, demonstrate the power of this new framework for tackling—in a coupled, multiscale manner—crucial reactor phenomena such as CRUD-induced power shift and fuel shuffle.},
urldate = {2017-04-10},
journal = {Annals of Nuclear Energy},
author = {Gaston, Derek R. and Permann, Cody J. and Peterson, John W. and Slaughter, Andrew E. and Andrš, David and Wang, Yaqi and Short, Michael P. and Perez, Danielle M. and Tonks, Michael R. and Ortensi, Javier and Zou, Ling and Martineau, Richard C.},
month = oct,
year = {2015},
keywords = {Full core reactor simulation, Multiphysics, Multiphysics coupling},
pages = {45--54},
file = {Gaston et al. - 2015 - Physics-based multiscale coupling for full core nu.pdf:/home/sunmyung/Zotero/storage/K84PKGIK/Gaston et al. - 2015 - Physics-based multiscale coupling for full core nu.pdf:application/pdf;ScienceDirect Full Text PDF:/home/sunmyung/Zotero/storage/SGQXWGJV/Gaston et al. - 2015 - Physics-based multiscale coupling for full core nu.pdf:application/pdf;ScienceDirect Full Text PDF:/home/sunmyung/Zotero/storage/C4TXYTMN/Gaston et al. - 2015 - Physics-based multiscale coupling for full core nu.pdf:application/pdf;ScienceDirect Snapshot:/home/sunmyung/Zotero/storage/PI8FZ93W/S030645491400543X.html:text/html;ScienceDirect Snapshot:/home/sunmyung/Zotero/storage/K6WTZCEL/S030645491400543X.html:text/html},
}
@article{fiorina_modelling_2014,
title = {Modelling and analysis of the {MSFR} transient behaviour},
volume = {64},
issn = {0306-4549},
url = {http://www.sciencedirect.com/science/article/pii/S0306454913004118},
doi = {10.1016/j.anucene.2013.08.003},
abstract = {Molten Salt Reactors (MSRs) were conceived at the early stages of nuclear energy in view of the favourable features fostered by a liquid fuel. They were developed as graphite-moderated thorium-fuelled breeder reactors till the seventies, when studies on this reactor concept were mostly abandoned in favour of the liquid–metal fast breeder concepts. A decade ago, the MSRs were included among the six GEN-IV systems and a core optimization process allowing for the GEN-IV main objectives led toward a fast-spectrum MSR concept (MSFR – Molten Salt Fast Reactor). Albeit advantageous in terms of U-233 breeding and/or radio-active waste burning, the new concept lacks the notable know-how available for the thermal-spectrum MSR technology. The present paper preliminarily investigates the MSR dynamics, based on the conceptual MSFR design currently available. A primary objective is to benchmark two different models of the MSFR primary circuit, both of them including a detailed and fully-coupled (node-wise) representation of turbulent fuel-salt flow, neutron diffusion, and delayed-neutron precursor diffusion and convection. A good agreement is generally observed between the adopted models, though some discrepancies exist in the temperature-field, with ensuing mild impacts on the reactor dynamics. The performed analyses are also used for a preliminary characterization of the MSFR steady-state and accidental transient response. Some points of enhancement needed in the MSFR conceptual design are identified, mainly related to in-core velocity and temperature fields. The reactor response following major accidental transient initiators suggests a generally benign behaviour of this reactor concept.},
number = {Supplement C},
urldate = {2017-10-03},
journal = {Annals of Nuclear Energy},
author = {Fiorina, Carlo and Lathouwers, Danny and Aufiero, Manuele and Cammi, Antonio and Guerrieri, Claudia and Kloosterman, Jan Leen and Luzzi, Lelio and Ricotti, Marco Enrico},
month = feb,
year = {2014},
keywords = {Safety, Molten Salt Reactor (MSR), Dynamics, Molten Salt Fast Reactor (MSFR)},
pages = {485--498},
annote = {2d rans},
file = {ScienceDirect Full Text PDF:/home/sunmyung/Zotero/storage/IXFXRASN/Fiorina et al. - 2014 - Modelling and analysis of the MSFR transient behav.pdf:application/pdf;ScienceDirect Full Text PDF:/home/sunmyung/Zotero/storage/XPVTBLQW/Fiorina et al. - 2014 - Modelling and analysis of the MSFR transient behav.pdf:application/pdf;ScienceDirect Full Text PDF:/home/sunmyung/Zotero/storage/8I8JU2JZ/Fiorina et al. - 2014 - Modelling and analysis of the MSFR transient behav.pdf:application/pdf;ScienceDirect Snapshot:/home/sunmyung/Zotero/storage/9WNRAF3E/S0306454913004118.html:text/html;ScienceDirect Snapshot:/home/sunmyung/Zotero/storage/U2ZUJ8KI/S0306454913004118.html:text/html;ScienceDirect Snapshot:/home/sunmyung/Zotero/storage/275TZZMW/S0306454913004118.html:text/html},
}
@article{li_transient_2015,
title = {Transient analyses for a molten salt fast reactor with optimized core geometry},
volume = {292},
issn = {0029-5493},
url = {http://www.sciencedirect.com/science/article/pii/S0029549315002617},
doi = {10.1016/j.nucengdes.2015.06.011},
abstract = {Molten salt reactors (MSRs) have encountered a marked resurgence of interest over the past decades, highlighted by their inclusion as one of the six candidate reactors of the Generation IV advanced nuclear power systems. The present work is carried out in the framework of the European FP-7 project EVOL (Evaluation and Viability Of Liquid fuel fast reactor system). One of the project tasks is to report on safety analyses: calculations of reactor transients using various numerical codes for the molten salt fast reactor (MSFR) under different boundary conditions, assumptions, and for different selected scenarios. Based on the original reference core geometry, an optimized geometry was proposed by Rouch et al. (2014. Ann. Nucl. Energy 64, 449) on thermal-hydraulic design aspects to avoid a recirculation zone near the blanket which accumulates heat and very high temperature exceeding the salt boiling point. Using both fully neutronics thermal-hydraulic coupled codes (SIMMER and COUPLE), we also re-confirm the efforts step by step toward a core geometry without the recirculation zone in particular as concerns the modifications of the core geometrical shape. Different transients namely Unprotected Loss of Heat Sink (ULOHS), Unprotected Loss of Flow (ULOF), Unprotected Transient Over Power (UTOP), Fuel Salt Over Cooling (FSOC) are intensively investigated and discussed with the optimized core geometry. It is demonstrated that due to inherent negative feedbacks, an MSFR plant has a high safety potential.},
journal = {Nuclear Engineering and Design},
author = {Li, R. and Wang, S. and Rineiski, A. and Zhang, D. and Merle-Lucotte, E.},
month = oct,
year = {2015},
keywords = {L. safety and risk analysis},
pages = {164--176},
file = {ScienceDirect Full Text PDF:/home/sunmyung/Zotero/storage/TFJT7C45/Li et al. - 2015 - Transient analyses for a molten salt fast reactor .pdf:application/pdf;ScienceDirect Snapshot:/home/sunmyung/Zotero/storage/59GD97VN/S0029549315002617.html:text/html},
}
@article{dehart_reactor_2011,
title = {Reactor {Physics} {Methods} and {Analysis} {Capabilities} in {SCALE}},
volume = {174},
url = {http://epubs.ans.org/?a=11720},
doi = {dx.doi.org/10.13182/NT174-196},
number = {2},
urldate = {2017-04-10},
journal = {Nuclear Technology},
author = {DeHart, Mark D. and Bowman, Stephen M.},
month = may,
year = {2011},
pages = {196--213},
file = {Snapshot:/home/sunmyung/Zotero/storage/BHMIBIJ9/epubs.ans.org.html:text/html},
}
@article{zanetti_geometric_2015,
title = {A {Geometric} {Multiscale} modelling approach to the analysis of {MSR} plant dynamics},
volume = {83},
issn = {0149-1970},
url = {https://www.sciencedirect.com/science/article/pii/S0149197015000487},
doi = {10.1016/j.pnucene.2015.02.014},
abstract = {In the framework of the Generation IV International Forum (GIF-IV), six innovative concepts of nuclear reactors have been proposed as suitable to guarantee a safe, sustainable and proliferation resistant source of nuclear energy. Among these reactors, a peculiar role is played by the Molten Salt Reactor (MSR), which is the only one with a liquid and circulating fuel. This feature leads to a complex and highly coupled behaviour, which requires careful investigations, as a consequence of some unusual features like the drift of Delayed Neutron Precursors (DNP) along the primary circuit and heat transfer with a heat-generating fluid. The inherently coupled dynamics of the MSRs asks for innovative approaches to perform reliable transient analyses. The node-wise implicitly-coupled solution of the Partial Differential Equations (PDE) that govern the different phenomena in a reactor would offer in this sense an ideal solution. However, such an approach (hereinafter referred to as Multi-Physics – MP) requires a huge amount of computational power. In this work, we propose and assess a Geometric Multiscale approach on MSR, addressing the core modelling with a 3-D MP approach and the remaining part of the system – e.g., the cooling loop – with simplified 0-D models based on Ordinary Differential Equations (ODE). The aim is to conjugate the accuracy of the MP modelling approach with acceptable computation loads. Reference is made to the Molten Salt Reactor Experiment (MSRE), due to the availability of a detailed design and experimental data that are used for assessment and preliminary validation of the developed simulation tool.},
number = {Supplement C},
urldate = {2017-02-08},
journal = {Progress in Nuclear Energy},
author = {Zanetti, Matteo and Cammi, Antonio and Fiorina, Carlo and Luzzi, Lelio},
month = aug,
year = {2015},
keywords = {Molten Salt Reactor (MSR), Geometric Multiscale approach, Molten Salt Reactor Experiment (MSRE), Multi-Physics Modelling, System dynamic behaviour},
pages = {82--98},
file = {ScienceDirect Full Text PDF:/home/sunmyung/Zotero/storage/JIPBQTRD/Zanetti et al. - 2015 - A Geometric Multiscale modelling approach to the a.pdf:application/pdf;ScienceDirect Full Text PDF:/home/sunmyung/Zotero/storage/ZDHUIY95/Zanetti et al. - 2015 - A Geometric Multiscale modelling approach to the a.pdf:application/pdf;ScienceDirect Snapshot:/home/sunmyung/Zotero/storage/NTTJSHEN/S0149197015000487.html:text/html;ScienceDirect Snapshot:/home/sunmyung/Zotero/storage/ZZT9I5TK/S0149197015000487.html:text/html},
}
@article{aufiero_development_2014,
title = {Development of an {OpenFOAM} model for the {Molten} {Salt} {Fast} {Reactor} transient analysis},
volume = {111},
issn = {0009-2509},
url = {http://www.sciencedirect.com/science/article/pii/S0009250914001146},
doi = {10.1016/j.ces.2014.03.003},
abstract = {In the paper, the development of a multiphysics model for the transient analysis of non-moderated Molten Salt Reactors is discussed. Particular attention is devoted to the description of the adopted time integration and physics coupling strategies. The proposed model features the adoption of an implicit Runge–Kutta scheme and the coupling among neutron diffusion, Reynolds-Averaged Navier–Stokes equations for mass and momentum conservation, and energy and delayed neutron precursor balance equations, in order to accurately catch thermal feedbacks on neutronics. The solver is aimed at performing fast-running simulations of the full-core three-dimensional Molten Salt Fast Reactor geometry. The neutronics modelling is assessed against Monte Carlo simulations and the results of a simplified case study are compared to those from multiphysics tools previously developed. As an example of the capability of the model, an unprotected MSFR single pump failure accidental scenario is simulated and discussed. The main purpose of the present model is to serve as fast-running computational tool in the phase of design optimization of fuel loop components. More in general, it is of valuable help in the study of reactor physics of circulating-fuel systems.},
journal = {Chemical Engineering Science},
author = {Aufiero, Manuele and Cammi, Antonio and Geoffroy, Olivier and Losa, Mario and Luzzi, Lelio and Ricotti, Marco E. and Rouch, Hervé},
month = may,
year = {2014},
keywords = {Reactor dynamics, Molten Salt Reactor (MSR), Multiphysics, Molten Salt Fast Reactor (MSFR), OpenFOAM},
pages = {390--401},
file = {ScienceDirect Full Text PDF:/home/sunmyung/Zotero/storage/GFZ6CW4E/Aufiero et al. - 2014 - Development of an OpenFOAM model for the Molten Sa.pdf:application/pdf;ScienceDirect Full Text PDF:/home/sunmyung/Zotero/storage/NI9HRTWS/Aufiero et al. - 2014 - Development of an OpenFOAM model for the Molten Sa.pdf:application/pdf;ScienceDirect Snapshot:/home/sunmyung/Zotero/storage/TGWQHMHK/S0009250914001146.html:text/html;ScienceDirect Snapshot:/home/sunmyung/Zotero/storage/L4BZULIM/S0009250914001146.html:text/html;ScienceDirect Snapshot:/home/sunmyung/Zotero/storage/YJ3GUNQ6/S0009250914001146.html:text/html},
}
@article{romano_openmc:_2015,
series = {Joint {International} {Conference} on {Supercomputing} in {Nuclear} {Applications} and {Monte} {Carlo} 2013, {SNA} + {MC} 2013. {Pluri}- and {Trans}-disciplinarity, {Towards} {New} {Modeling} and {Numerical} {Simulation} {Paradigms}},
title = {{OpenMC}: {A} state-of-the-art {Monte} {Carlo} code for research and development},
volume = {82},
issn = {0306-4549},
shorttitle = {{OpenMC}},
url = {http://www.sciencedirect.com/science/article/pii/S030645491400379X},
doi = {10.1016/j.anucene.2014.07.048},
abstract = {This paper gives an overview of OpenMC, an open source Monte Carlo particle transport code recently developed at the Massachusetts Institute of Technology. OpenMC uses continuous-energy cross sections and a constructive solid geometry representation, enabling high-fidelity modeling of nuclear reactors and other systems. Modern, portable input/output file formats are used in OpenMC: XML for input, and HDF5 for output. High performance parallel algorithms in OpenMC have demonstrated near-linear scaling to over 100,000 processors on modern supercomputers. Other topics discussed in this paper include plotting, CMFD acceleration, variance reduction, eigenvalue calculations, and software development processes.},
number = {Supplement C},
urldate = {2017-11-22},
journal = {Annals of Nuclear Energy},
author = {Romano, Paul K. and Horelik, Nicholas E. and Herman, Bryan R. and Nelson, Adam G. and Forget, Benoit and Smith, Kord},
month = aug,
year = {2015},
keywords = {HDF5, Monte Carlo, Neutron transport, OpenMC, Parallel, XML},
pages = {90--97},
file = {ScienceDirect Full Text PDF:/home/sunmyung/Zotero/storage/9S36DPTN/Romano et al. - 2015 - OpenMC A state-of-the-art Monte Carlo code for re.pdf:application/pdf;ScienceDirect Full Text PDF:/home/sunmyung/Zotero/storage/IW54R5QX/Romano et al. - 2015 - OpenMC A state-of-the-art Monte Carlo code for re.pdf:application/pdf;ScienceDirect Snapshot:/home/sunmyung/Zotero/storage/NZ4279HS/S030645491400379X.html:text/html;ScienceDirect Snapshot:/home/sunmyung/Zotero/storage/ALT6V9G9/S030645491400379X.html:text/html},
}
@incollection{dai_17_2017,
title = {17 - {Thorium} molten salt reactor nuclear energy system ({TMSR})},
isbn = {978-0-08-101126-3},
url = {https://www.sciencedirect.com/science/article/pii/B9780081011263000178},
abstract = {The thorium molten salt reactor nuclear energy system (TMSR) is designed for thorium-based nuclear energy utilization and hybrid nuclear energy application, based on a liquid-fueled thorium molten salt reactor (TMSR-LF) and a solid-fueled thorium molten salt reactor (TMSR-SF).},
urldate = {2017-11-21},
booktitle = {Molten {Salt} {Reactors} and {Thorium} {Energy}},
publisher = {Woodhead Publishing},
author = {Dai, Zhimin},
editor = {Dolan, Thomas J.},
year = {2017},
doi = {10.1016/B978-0-08-101126-3.00017-8},
keywords = {Molten salt reactor, hybrid nuclear energy application, thorium},
pages = {531--540},
file = {ScienceDirect Full Text PDF:/home/sunmyung/Zotero/storage/IUVWTXVJ/Dai - 2017 - 17 - Thorium molten salt reactor nuclear energy sy.pdf:application/pdf;ScienceDirect Full Text PDF:/home/sunmyung/Zotero/storage/QGM7UCAA/Dai - 2017 - 17 - Thorium molten salt reactor nuclear energy sy.pdf:application/pdf;ScienceDirect Snapshot:/home/sunmyung/Zotero/storage/SUYR248M/B9780081011263000178.html:text/html;ScienceDirect Snapshot:/home/sunmyung/Zotero/storage/4JZSFD3K/B9780081011263000178.html:text/html},
}
@incollection{kloosterman_20_2017,
title = {20 - {Safety} assessment of the molten salt fast reactor ({SAMOFAR})},
isbn = {978-0-08-101126-3},
url = {https://www.sciencedirect.com/science/article/pii/B9780081011263000208},
abstract = {This chapter describes the goal, contents, and the consortium of the SAMOFAR project, which is currently the leading project in the European Union in the field of MSR research. It focuses on the safety assessment of the Molten Salt Fast Reactor and will eventually lead to an updated reactor design evaluated with a new integral safety method.},
urldate = {2017-11-21},
booktitle = {Molten {Salt} {Reactors} and {Thorium} {Energy}},
publisher = {Woodhead Publishing},
author = {Kloosterman, Jan L.},
editor = {Dolan, Thomas J.},
year = {2017},
doi = {10.1016/B978-0-08-101126-3.00020-8},
keywords = {Molten Salt Fast Reactor, Safety assessment},
pages = {565--570},
file = {ScienceDirect Full Text PDF:/home/sunmyung/Zotero/storage/7LLJFPR3/Kloosterman - 2017 - 20 - Safety assessment of the molten salt fast rea.pdf:application/pdf;ScienceDirect Snapshot:/home/sunmyung/Zotero/storage/6DWXAWQD/B9780081011263000208.html:text/html},
}
@article{lindsay_introduction_2018,
title = {Introduction to {Moltres}: {An} application for simulation of {Molten} {Salt} {Reactors}},
volume = {114},
issn = {0306-4549},
shorttitle = {Introduction to {Moltres}},
url = {https://linkinghub.elsevier.com/retrieve/pii/S0306454917304760},
doi = {10.1016/j.anucene.2017.12.025},
abstract = {Moltres is a new physics application for modeling coupled physics in fluid-fuelled, molten salt reactors. This paper describes its neutronics model, thermal hydraulics model, and their coupling in the MOOSE framework. Neutron and precursor equations are implemented using an action system that allows use of an arbitrary number of groups with no change in the input card. Results for many-channel configurations in 2D-axisymmetric and 3D coordinates are presented and compared against other coupled models as well as the Molten Salt Reactor Experiment.},
language = {en},
urldate = {2018-01-08},
journal = {Annals of Nuclear Energy},
author = {Lindsay, Alexander and Ridley, Gavin and Rykhlevskii, Andrei and Huff, Kathryn},
month = apr,
year = {2018},
keywords = {agent based modeling, Finite elements, Hydrologic contaminant transport, MOOSE, Multiphysics, nuclear engineering, Nuclear fuel cycle, Object orientation, Parallel computing, Reactor physics, repository, Simulation, Systems analysis},
pages = {530--540},
annote = {2d prescribed},
file = {Lindsay et al. - 2018 - Introduction to Moltres An application for simula.pdf:/home/sunmyung/Zotero/storage/RCWUNGTP/Lindsay et al. - 2018 - Introduction to Moltres An application for simula.pdf:application/pdf;Lindsay et al. - 2018 - Introduction to Moltres An application for simula.pdf:/home/sunmyung/Zotero/storage/3GEC6NQ9/Lindsay et al. - 2018 - Introduction to Moltres An application for simula.pdf:application/pdf;Moltres.pdf:/home/sunmyung/Zotero/storage/4XDXRICB/Moltres.pdf:application/pdf;ScienceDirect Full Text PDF:/home/sunmyung/Zotero/storage/E2T9U5IX/Lindsay et al. - 2018 - Introduction to Moltres An application for simula.pdf:application/pdf;ScienceDirect Snapshot:/home/sunmyung/Zotero/storage/3DT9TEY3/S0306454917304760.html:text/html},
}
@techreport{oecd/nea_jeff-3.1.2_2014,
title = {The {JEFF}-3.1.2 {Nuclear} {Data} {Library}},
number = {JEFF Report 24, OECD/NEA Data Bank},
institution = {OECD/NEA},
author = {OECD/NEA},
year = {2014},
}
@article{krepel_dyn3d-msr_2007,
title = {{DYN3D}-{MSR} spatial dynamics code for molten salt reactors},
volume = {34},
issn = {0306-4549},
url = {http://www.sciencedirect.com/science/article/pii/S0306454907000527},
doi = {10.1016/j.anucene.2006.12.011},
abstract = {The development of spatial dynamics code for molten salt reactors (MSRs) is reported in this paper. The graphite-moderated channel type MSR – one of the ‘Generation IV’ concepts – was selected for the numerical simulation. It has several peculiarities (e.g. the drift of delayed neutrons precursors), which disable the use of standard dynamics codes. Therefore, the own DYN3D-MSR code was developed. It is based on the light water reactor code DYN3D and it allows transients simulation by 3D neutronics and parallel channel thermal-hydraulics. The neutronics and thermal-hydraulics were modified for the MSR peculiarities, where the experience from DYN1D-MSR development was exploited. The code was validated on experimental results from the MSRE experiment done in Oak Ridge National Laboratory and by the comparison with other codes especially with the 1D version. However, by the 3D code transients can be simulated, where space-dependant efforts are relevant, like local blockage of fuel channels or local temperature perturbations.},
language = {en},
number = {6},
urldate = {2016-08-17},
journal = {Annals of Nuclear Energy},
author = {Krepel, Jiří and Rohde, Ulrich and Grundmann, Ulrich and Weiss, Frank-Peter},
month = jun,
year = {2007},
pages = {449--462},
annote = {1D TH example},
file = {Křepel et al. - 2007 - DYN3D-MSR spatial dynamics code for molten salt re.pdf:/home/sunmyung/Zotero/storage/AMVNY7E6/Křepel et al. - 2007 - DYN3D-MSR spatial dynamics code for molten salt re.pdf:application/pdf;ScienceDirect Full Text PDF:/home/sunmyung/Zotero/storage/T5JKV5U6/Křepel et al. - 2007 - DYN3D-MSR spatial dynamics code for molten salt re.pdf:application/pdf;ScienceDirect Snapshot:/home/sunmyung/Zotero/storage/Q24HJQHD/S0306454907000527.html:text/html},
}
@article{kophazi_development_2009,
title = {Development of a {Three}-{Dimensional} {Time}-{Dependent} {Calculation} {Scheme} for {Molten} {Salt} {Reactors} and {Validation} of the {Measurement} {Data} of the {Molten} {Salt} {Reactor} {Experiment}},
volume = {163},
abstract = {This paper presents the development, validation, and results of a three-dimensional, time- dependent, coupled-neutronics–thermal-hydraulic calculational scheme for channel-type molten salt re- actors (MSRs). The reactor physics part is based on diffusion theory, extended by a term representing the flow of the fuel through the core. The calculation of the temperature field is done by modeling all fuel channels, which are coupled to each other by a three-dimensional heat conduction equation. For the purpose of validation, the results of the MSR Experiment (MSRE) natural-circulation experiment and the thermal feedback coefficients of the reactor have been calculated and compared.
With the aid of a code system developed to implement this scheme, calculations were carried out for the normal operating state of the MSRE and some debris-induced channel-blocking-incident transients. In the case of the MSRE, it is shown that the severity of such an incident strongly depends on the degree of channel blocking and that high-temperature gradients in the moderator can connect thermally the adjacent fuel channels. Results are included for an unblocking transient (i.e., the debris suddenly exits the core, and the fuel flow reverts to the normal operating pattern), and it was demonstrated that during the unblocking large power peaks can be induced.},
number = {2},
journal = {Nuclear Science and Engineering},
author = {Kophazi, J. and Lathouwers, D. and Kloosterman, J.L.},
year = {2009},
keywords = {unread, Molten Salt Reactor (MSR), 3D, reactor physics, Core, MSR Experiment (MSRE)},
pages = {118--131},
file = {Kópházi et al. - Development of a Three-Dimensional Time-Dependent .pdf:/home/sunmyung/Zotero/storage/ZIJ5Q643/Kópházi et al. - Development of a Three-Dimensional Time-Dependent .pdf:application/pdf},
}
@inproceedings{aufiero_testing_2018,
address = {Philadelphia, Pennsylvania},
series = {Reactor {Physics}: {General}—{I}},
title = {Testing and {Verification} of {Multiphysics} {Tools} for {Fast}-{Spectrum} {MSRs}: {The} {CNRS} {Benchmark}},
volume = {118},
shorttitle = {Testing and {Verification} of {Multiphysics} {Tools} for {Fast}-{Spectrum} {MSRs}},
url = {http://ansannual.org/wp-content/2018/Data/pdfs/382-25331.pdf},
abstract = {Liquid fuel Molten Salt Reactors (MSRs) feature peculiar physical
phenomena that are not present in solid fuel reactors. Some of
these phenomena are not correctly resolved by legacy reactor physics
tools, and require specific treatments. Recently, research activities
within several collaborative projects related to MSRs design (e.g.,
see [1], [2]) lead to development of a number of different numerical
tools for the coupled neutronics and thermal/hydraulics analysis of
such systems. Unfortunately, especially in case of non-moderated,
fast-spectrum MSRs, very few experimental data are available for
an accurate process of verification and validation of the developed
codes.},
language = {English},
booktitle = {Transactions of the {American} {Nuclear} {Society}},
publisher = {American Nuclear Society},
author = {Aufiero, Manuele and Rubiolo, Pablo},
month = jun,
year = {2018},
pages = {837--840},
file = {Fulltext:/home/sunmyung/Zotero/storage/HUGQPFM3/Aufiero and Rubiolo - Testing and Verification of Multiphysics Tools for.pdf:application/pdf;Snapshot:/home/sunmyung/Zotero/storage/SXY5L3Q3/Aufiero and Rubiolo - Testing and Verification of Multiphysics Tools for.pdf:application/pdf},
}
@techreport{euratom_final_2015,
address = {France},
type = {Final report},
title = {Final {Report} {Summary} - {EVOL} ({Evaluation} and {Viability} of {Liquid} {Fuel} {Fast} {Reactor} {System}) {\textbar} {Report} {Summary} {\textbar} {EVOL} {\textbar} {FP7}{\textbar} {European} {Commission}},
url = {https://cordis.europa.eu/result/rcn/159411_en.html},
abstract = {Executive Summary:An innovative molten salt reactor concept, the MSFR (Molten Salt Fast Reactor) is developed by CNRS (France) since 2004. Based on the particularity of using a liquid fuel, this...},
language = {en},
number = {249696},
urldate = {2018-05-22},
institution = {EURATOM},
author = {EURATOM},
year = {2015},
file = {Final Report Summary - EVOL (Evaluation and Viabil.pdf:/home/sunmyung/Zotero/storage/HB992DYS/Final Report Summary - EVOL (Evaluation and Viabil.pdf:application/pdf;Snapshot:/home/sunmyung/Zotero/storage/67KEHCS2/159411_en.html:text/html},
}
@article{cervi_development_2019,
title = {Development of a multiphysics model for the study of fuel compressibility effects in the {Molten} {Salt} {Fast} {Reactor}},
volume = {193},
issn = {0009-2509},
url = {http://www.sciencedirect.com/science/article/pii/S0009250918306742},
doi = {10.1016/j.ces.2018.09.025},
abstract = {Compressible fluid dynamics is of great practical interest in many industrial applications, ranging from chemistry to aeronautical industry, and to nuclear field as well. At the same time, modelling and simulation of compressible flows is a very complex task, requiring the development of specific approaches, in order to describe the effect of pressure on the fluid velocity field. Compressibility effects become even more important in the study of two-phase flows, due to the presence of a gaseous phase. In addition, compressibility is also expected to have a significant impact on other physics, such as chemical or nuclear reactions occurring in the mixture. In this perspective, multiphysics represents a useful approach to address this complex problem, providing a way to catch all the different physics that come into play as well as the coupling between them. In this work, a multiphysics model is developed for the analysis of the generation IV Molten Salt Fast Reactor (MSFR), with a specific focus on the compressibility effects of the fluid that acts as fuel in the reactor. The fuel mixture compressibility is expected to have an important effect on the system dynamics, especially in very rapid super-prompt-critical transients. In addition, the presence of a helium bubbling system used for online fission product removal could modify the fuel mixture compressibility, further affecting the system transient behaviour. Therefore, the MSFR represents an application of concrete interest, inherent to the analysis of compressibility effects and to the development of suitable modelling approaches. An OpenFOAM solver is developed to handle the fuel compressibility, the presence of gas bubbles in the reactor as well as the coupling between the system neutronics and fluid dynamics. The outcomes of this analysis point out that the fuel compressibility plays a crucial role in the evolution of fast transients, introducing delays in the expansion feedbacks that strongly affect the system dynamics. Moreover, it is found that the gas bubbles significantly alter the fuel compressibility, yielding even larger differences compared to the incompressible approximation usually adopted in the current MSFR solvers.},
urldate = {2018-10-11},
journal = {Chemical Engineering Science},
author = {Cervi, E. and Lorenzi, S. and Cammi, A. and Luzzi, L.},
month = jan,
year = {2019},
keywords = {Reactor dynamics, Multiphysics, Molten Salt Fast Reactor (MSFR), OpenFOAM, Compressible fluid dynamics},
pages = {379--393},
file = {Fulltext:/home/sunmyung/Zotero/storage/WNYVUELP/S0009250918306742.html:text/html;ScienceDirect Full Text PDF:/home/sunmyung/Zotero/storage/WMQ8ZVWE/Cervi et al. - 2019 - Development of a multiphysics model for the study .pdf:application/pdf;ScienceDirect Snapshot:/home/sunmyung/Zotero/storage/48G8AZZI/S0009250918306742.html:text/html;Snapshot:/home/sunmyung/Zotero/storage/P39Y9VFL/S0009250918306742.html:text/html},
}
@misc{comsol_ab_comsol_nodate,
address = {Stockholm, Sweden},
title = {{COMSOL} {Multiphysics}®},
url = {www.comsol.com},
publisher = {COMSOL AB},
author = {{COMSOL AB}},
}
@article{cervi_development_2019-1,
title = {Development of an {SP3} neutron transport solver for the analysis of the {Molten} {Salt} {Fast} {Reactor}},
volume = {346},
issn = {0029-5493},
url = {http://www.sciencedirect.com/science/article/pii/S0029549319300354},
doi = {10.1016/j.nucengdes.2019.03.001},
abstract = {The aim of this paper is the extension of a multiphysics OpenFOAM solver for the analysis of the Molten Salt Fast Reactor (MSFR), developed in previous works (Cervi et al., 2017, 2018). In particular, the neutronics sub-solver is improved by implementing a new module based on the SP3 approximation of the neutron transport equation. The new module is successfully tested against a Monte Carlo model of the MSFR, in order to assess its correct implementation. Then, a neutronics analysis of the MSFR is carried out on a simplified axial-symmetric model of the reactor. Particular focus is devoted to the analysis of the MSFR helium bubbling system and its effect on reactivity. The presence of bubbles inside the reactor is handled with a two-fluid thermal-hydraulics module, previously implemented into the solver. The void reactivity coefficient is evaluated on the basis of the bubble spatial distribution calculated by the multiphysics solver. Then, the results are compared to simulations carried out with uniform bubble distributions, highlighting significant differences between the two approaches. The outcomes of this work constitute a step forward in the multiphysics analysis of the Molten Salt Fast Reactor and represent a useful starting point for the optimization of the MSFR helium bubbling system, as well as for the development of appropriate control strategies.},
urldate = {2019-03-23},
journal = {Nuclear Engineering and Design},
author = {Cervi, E. and Lorenzi, S. and Cammi, A. and Luzzi, L.},
month = may,
year = {2019},
keywords = {Neutron transport, Multiphysics, Molten Salt Fast Reactor (MSFR), OpenFOAM},
pages = {209--219},
file = {ScienceDirect Full Text PDF:/home/sunmyung/Zotero/storage/2U45NTZE/Cervi et al. - 2019 - Development of an SP3 neutron transport solver for.pdf:application/pdf;ScienceDirect Snapshot:/home/sunmyung/Zotero/storage/JDS5R3Y4/S0029549319300354.html:text/html},
}
@techreport{gif_technology_2014,
type = {Generation {IV} {International} {Forum}},
title = {Technology {Roadmap} {Update} for {Generation} {IV} {Nuclear} {Energy} {Systems}},
shorttitle = {{GIF} {Technology} {Roadmap}},
url = {https://www.gen-4.org/gif/upload/docs/application/pdf/2014-03/gif-tru2014.pdf},
language = {en},
number = {January 2014 Update},
institution = {OECD Nuclear Energy Agency},
author = {GIF},
month = jan,
year = {2014},
note = {https://www.gen-4.org/gif/upload/docs/application/pdf/2014-03/gif-tru2014.pdf},
pages = {66},
file = {Technology Roadmap Update for Generation IV Nuclea.pdf:/home/sunmyung/Zotero/storage/9LRH9PYP/Technology Roadmap Update for Generation IV Nuclea.pdf:application/pdf},
}
@article{brovchenko_neutronic_2019,
title = {Neutronic benchmark of the molten salt fast reactor in the frame of the {EVOL} and {MARS} collaborative projects},
volume = {5},
copyright = {© M. Brovchenko et al., published by EDP Sciences, 2019},
issn = {2491-9292},
url = {https://www.epj-n.org/articles/epjn/abs/2019/01/epjn180012/epjn180012.html},
doi = {10.1051/epjn/2018052},
abstract = {This paper describes the neutronic benchmarks and the results obtained by the various participants of the FP7 project EVOL and the ROSATOM project MARS. The aim of the benchmarks was two-fold: first to verify and validate each of the code packages of the project partners, adapted for liquid-fueled reactors, and second to check the dependence of the core characteristics to nuclear data set for application on a molten salt fast reactor (MSFR). The MSFR operates with the thorium fuel cycle and can be started with {\textless}sup{\textgreater}233{\textless}sup/{\textgreater}U-enriched U and/or TRU elements as initial fissile load. All three compositions were covered by the present benchmark. The calculations have confirmed that the MSFR has very favorable characteristics not present in other Gen4 fast reactors, like strong negative temperature and void reactivity coefficients, a low-fissile inventory, a reduced long-lived waste production and its burning capacities of nuclear waste produced in currently operational reactors.},
language = {en},
urldate = {2019-09-16},
journal = {EPJ Nuclear Sciences \& Technologies},
author = {Brovchenko, Mariya and Kloosterman, Jan-Leen and Luzzi, Lelio and Merle, Elsa and Heuer, Daniel and Laureau, Axel and Feynberg, Olga and Ignatiev, Victor and Aufiero, Manuele and Cammi, Antonio and Fiorina, Carlo and Alcaro, Fabio and Dulla, Sandra and Ravetto, Piero and Frima, Lodewijk and Lathouwers, Danny and Merk, Bruno},
month = jan,
year = {2019},
pages = {2},
file = {Full Text PDF:/home/sunmyung/Zotero/storage/ET8QPY6L/Brovchenko et al. - 2019 - Neutronic benchmark of the molten salt fast reacto.pdf:application/pdf;Snapshot:/home/sunmyung/Zotero/storage/XMFKGGX5/epjn180012.html:text/html},
}
@inproceedings{park_safety_2019,
address = {Seattle, WA, United States},
title = {Safety {Analysis} of the {Molten} {Salt} {Fast} {Reactor} {Fuel} {Composition} using {Moltres}},
url = {http://epubs.ans.org/?a=47030},
doi = {10.31224/osf.io/7ce89},
abstract = {The Molten Salt Fast Reactor (MSFR) has garnered much interest for its inherent safety and sustainbility features. The MSFR can adopt a closed thorium fuel cycle for sustainable operation through the breeding of 233 U from 232 Th. The fuel composition changes significantly over the course of its lifespan. In this study, we investigated the steady state and transient behavior of the MSFR using Moltres, a coupled neutronics/thermal-hydraulics code developed within the Multiphysics Object Oriented Simulation Environment (MOOSE) framework. Three different fuel compositions, start-up, early-life, and equilibrium, were examined for potentially dangerous core temperature excursions during a unprotected loss of heat sink (ULOHS) accident. The six-group and total neutron flux distributions showed good agreement with SERPENT and published MSFR results, while the temperature distribution and total power showed discrepancies which can be attributed toknown sources of error. For the transient behavior under the ULOHS scenario, while the transition time towards the new steady state core temperature is also in good agreement with existing MSFR simulations by Fiorina et al., Moltres under-estimated the temperature rise by a factor of ten, due to the same sources of error affecting the steady state results. While an MSFR loaded with start-up fuel composition operates at a higher temperature than with the other two fuel compositions, all three cases were shown to be inherently safe due to thestrong negative temperature feedback.},
urldate = {2019-11-05},
booktitle = {Proceedings of {GLOBAL} {International} {Fuel} {Cycle} {Conference}},
publisher = {American Nuclear Society},
author = {Park, Sun Myung and Rykhlevskii, Andrei and Huff, Kathryn},
month = sep,
year = {2019},
file = {Park et al. - 2019 - Safety Analysis of Molten Salt Fast Reactor Fuel C.pdf:/home/sunmyung/Zotero/storage/EXYUTUGW/Park et al. - 2019 - Safety Analysis of Molten Salt Fast Reactor Fuel C.pdf:application/pdf},
}
@inproceedings{zhang_couple_2014,
title = {{COUPLE}, {A} {Time}-{Dependent} {Coupled} {Neutronics} and {Thermal}-{Hydraulics} {Code}, and its {Application} to {MSFR}},
url = {https://asmedigitalcollection.asme.org/ICONE/proceedings/ICONE22/45936/V003T05A019/255491},
doi = {10.1115/ICONE22-30609},
language = {en},
urldate = {2020-01-29},
publisher = {American Society of Mechanical Engineers Digital Collection},
author = {Zhang, Dalin and Zhai, Zhi-Gang and Rineiski, Andrei and Guo, Zhangpeng and Wang, Chenglong and Xiao, Yao and Qiu, Suizheng},
month = nov,
year = {2014},
file = {Full Text PDF:/home/sunmyung/Zotero/storage/VUK7EPRC/Zhang et al. - 2014 - COUPLE, A Time-Dependent Coupled Neutronics and Th.pdf:application/pdf;Snapshot:/home/sunmyung/Zotero/storage/PDPPHN33/255491.html:text/html},
}
@article{lecarpentier_neutronic_2003,
title = {A {Neutronic} {Program} for {Critical} and {Nonequilibrium} {Study} of {Mobile} {Fuel} {Reactors}: {The} {Cinsf1D} {Code}},
volume = {143},
issn = {0029-5639},
shorttitle = {A {Neutronic} {Program} for {Critical} and {Nonequilibrium} {Study} of {Mobile} {Fuel} {Reactors}},
url = {https://doi.org/10.13182/NSE03-A2316},
doi = {10.13182/NSE03-A2316},
abstract = {Molten salt reactors (MSRs) have the distinction of having a liquid fuel that is also the coolant. The transport of delayed-neutron precursors by the fuel modifies the precursors’ equation. As a consequence, it is necessary to adapt the methods currently used for solid fuel reactors to achieve critical or kinetics calculations for an MSR. A program is presented for which this adaptation has been carried out within the framework of the two-energy-group diffusion theory with one dimension of space. This program has been called Cinsf1D (Cinétique pour réacteur à sels fondus 1D).},
number = {1},
urldate = {2020-02-17},
journal = {Nuclear Science and Engineering},
author = {Lecarpentier, David and Carpentier, Vincent},
month = jan,
year = {2003},
pages = {33--46},
file = {Full Text PDF:/home/sunmyung/Zotero/storage/7UYFXLSR/Lecarpentier and Carpentier - 2003 - A Neutronic Program for Critical and Nonequilibriu.pdf:application/pdf;Snapshot:/home/sunmyung/Zotero/storage/2DQBSVAU/NSE03-A2316.html:text/html},
}
@article{nicolino_coupled_2008,
title = {Coupled dynamics in the physics of molten salt reactors},
volume = {35},
issn = {0306-4549},
url = {http://www.sciencedirect.com/science/article/pii/S0306454907001454},
doi = {10.1016/j.anucene.2007.06.015},
abstract = {The present paper is devoted to the analysis of the coupled thermo-fluid and neutronic dynamics of fast fluid-fuel multiplying nuclear systems. A completely coupled model is needed since in some fast reactors designs, the velocity pattern could be very complicated and strongly affected by the neutron dynamics via the heat source from fission reactions. Furthermore, the neutron dynamics is strongly affected by the thermohydrodynamics via the motion of precursors and by feedback effects. The methods typical of solid fuel reactors of previous generations are not sufficient to handle these more highly coupled concepts. In the preset paper, we consider the coupling between neutronics and thermohydrodynamics with simple but realistic hypotheses assumed to model the evolution of all the variables involved in the calculation. The numerical scheme used represents the current state of the art in the solution of non-linear systems: the Newton–Krylov algorithm. Several calculations are presented to demonstrate the ability of the methods described here to study the behavior of molten salt reactors in both steady state and transient situations.},
language = {en},
number = {2},
urldate = {2020-02-17},
journal = {Annals of Nuclear Energy},
author = {Nicolino, Claudio and Lapenta, Giovanni and Dulla, Sandra and Ravetto, Piero},
month = feb,
year = {2008},
pages = {314--322},
annote = {2D TH with VORTICITY},
file = {ScienceDirect Full Text PDF:/home/sunmyung/Zotero/storage/78544BSJ/Nicolino et al. - 2008 - Coupled dynamics in the physics of molten salt rea.pdf:application/pdf;ScienceDirect Full Text PDF:/home/sunmyung/Zotero/storage/HFETG65I/Nicolino et al. - 2008 - Coupled dynamics in the physics of molten salt rea.pdf:application/pdf;ScienceDirect Snapshot:/home/sunmyung/Zotero/storage/GZI753P9/S0306454907001454.html:text/html;ScienceDirect Snapshot:/home/sunmyung/Zotero/storage/FPYD47S6/S0306454907001454.html:text/html},
}
@article{zhang_development_2009,
title = {Development of a steady state analysis code for a molten salt reactor},
volume = {36},
issn = {0306-4549},
url = {http://www.sciencedirect.com/science/article/pii/S030645490900019X},
doi = {10.1016/j.anucene.2009.01.004},
abstract = {The molten salt reactor (MSR), which is one of the ‘Generation IV’ concepts, can be used for transmutation, and production of electricity, hydrogen and fissile fuels. In this study, a single-liquid-fueled MSR is designed for conceptual research, in which no solid material is present in the core as moderator, except for the external reflector. The fuel salt flow makes the MSR neutronics different from that of conventional reactors using solid fuels, and couples the flow and heat transfer strongly. Therefore, it is necessary to study the core characteristics with due attention to the coupling among flow, heat transfer and neutronics. The standard turbulent model is adopted to establish the flow and heat transfer model, while the diffusion theory is used for the neutronics model, which consists of two-group neutron diffusion equations for fast and thermal neutron fluxes, and balance equations for six groups of delayed neutron precursors. These two models which are coupled through the temperature and heat source are coded in a microcomputer program. The distributions of the velocity, temperature, neutron fluxes, and delayed neutron precursors under the rated condition are obtained. In addition, the effects of the inflow temperature, inflow velocity, and the fuel salt residence time out of the core are discussed in detail. The results provide some valuable information for the research and design of the new generation molten salt reactors.},
language = {en},
number = {5},
urldate = {2020-02-17},
journal = {Annals of Nuclear Energy},
author = {Zhang, D. L. and Qiu, S. Z. and Su, G. H. and Liu, C. L.},
month = may,
year = {2009},
keywords = {unread},
pages = {590--603},
annote = {RANS ke},
file = {ScienceDirect Full Text PDF:/home/sunmyung/Zotero/storage/KS4ZFFFG/Zhang et al. - 2009 - Development of a steady state analysis code for a .pdf:application/pdf;ScienceDirect Full Text PDF:/home/sunmyung/Zotero/storage/DDA9ZJDQ/Zhang et al. - 2009 - Development of a steady state analysis code for a .pdf:application/pdf;ScienceDirect Snapshot:/home/sunmyung/Zotero/storage/W6E73ACI/S030645490900019X.html:text/html;ScienceDirect Snapshot:/home/sunmyung/Zotero/storage/8DB2MI7D/S030645490900019X.html:text/html;zhang_development_2009.pdf:/home/sunmyung/Zotero/storage/6NFMMZAE/zhang_development_2009.pdf:application/pdf},
}
@article{tiberga_results_2020,
title = {Results from a multi-physics numerical benchmark for codes dedicated to molten salt fast reactors},
volume = {142},
issn = {0306-4549},
url = {http://www.sciencedirect.com/science/article/pii/S0306454920301262},
doi = {10.1016/j.anucene.2020.107428},
abstract = {Verification and validation of multi-physics codes dedicated to fast-spectrum molten salt reactors (MSR) is a very challenging task. Existing benchmarks are meant for single-physics codes, while experimental data for validation are absent. This is concerning, given the importance numerical simulations have in the development of fast MSR designs. Here, we propose the use of a coupled numerical benchmark specifically designed to assess the physics-coupling capabilities of the aforementioned codes. The benchmark focuses on the specific characteristics of fast MSRs and features a step-by-step approach, where physical phenomena are gradually coupled to easily identify sources of error. We collect and compare the results obtained during the benchmarking campaign of four multi-physics tools developed within the SAMOFAR project. Results show excellent agreement for all the steps of the benchmark. The benchmark generality and the broad spectrum of results provided constitute a useful tool for the testing and development of similar multi-physics codes.},
language = {en},
urldate = {2020-03-17},
journal = {Annals of Nuclear Energy},
author = {Tiberga, Marco and de Oliveira, Rodrigo Gonzalez Gonzaga and Cervi, Eric and Blanco, Juan Antonio and Lorenzi, Stefano and Aufiero, Manuele and Lathouwers, Danny and Rubiolo, Pablo},
month = jul,
year = {2020},
keywords = {Neutronics, Molten salt reactor, Benchmark, Thermal-hydraulics, Multi-physics, Code-to-code comparison, Fast-spectrum},
pages = {107428},
file = {ScienceDirect Full Text PDF:/home/sunmyung/Zotero/storage/7AYW94UF/Tiberga et al. - 2020 - Results from a multi-physics numerical benchmark f.pdf:application/pdf;ScienceDirect Snapshot:/home/sunmyung/Zotero/storage/4NR9AK9R/S0306454920301262.html:text/html},
}
@techreport{graham_development_2019,
title = {Development of {Molten} {Salt} {Reactor} {Modeling} and {Simulation} {Capabilities} in {VERA}},
url = {https://www.osti.gov/biblio/1559657-development-molten-salt-reactor-modeling-simulation-capabilities-vera},
abstract = {The U.S. Department of Energy's Office of Scientific and Technical Information},
language = {English},
urldate = {2020-06-28},
institution = {Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)},
author = {Graham, Aaron (ORCID:0000000232458441) and Collins, Benjamin S. (ORCID:0000000241918868) and Salko Jr, Robert (ORCID:0000000253566679) and Taylor, Robert Z. and Gentry, Cole A. (ORCID:0000000237730707)},
month = sep,
year = {2019},
file = {Full Text PDF:/home/sunmyung/Zotero/storage/MXBR8DZ9/Graham et al. - 2019 - Development of Molten Salt Reactor Modeling and Si.pdf:application/pdf;Snapshot:/home/sunmyung/Zotero/storage/6UQXW5EZ/1559657-development-molten-salt-reactor-modeling-simulation-capabilities-vera.html:text/html},
}
@techreport{nrc_trace_2007,
title = {{TRACE} {V5}.0 {User}'s {Manual}},
url = {https://www.nrc.gov/docs/ML1200/ML120060239.pdf},
urldate = {2020-07-22},
author = {NRC},
year = {2007},
file = {ML120060239.pdf:/home/sunmyung/Zotero/storage/VSTVM8S7/ML120060239.pdf:application/pdf},
}
@mastersthesis{park_advancement_2020,
address = {Urbana, IL},
title = {Advancement and {Verification} of {Moltres} for {Molten} {Salt} {Reactor} {Safety} {Analysis}},
copyright = {Copyright 2020 Sun Myung Park},
url = {https://www.ideals.illinois.edu/handle/2142/108542},
language = {English},
school = {University of Illinois at Urbana-Champaign},
author = {Park, Sun Myung},
month = aug,
year = {2020},
file = {Park - 2020 - Advancement and Verification of Moltres for Molten.pdf:/home/sunmyung/Zotero/storage/JYYYTBJ7/Park - 2020 - Advancement and Verification of Moltres for Molten.pdf:application/pdf},
}
@article{altahhan_preliminary_2020,
title = {Preliminary design and analysis of {Liquid} {Fuel} {Molten} {Salt} {Reactor} using multi-physics code {GeN}-{Foam}},
volume = {369},
issn = {0029-5493},
url = {http://www.sciencedirect.com/science/article/pii/S0029549320303204},
doi = {10.1016/j.nucengdes.2020.110826},
abstract = {In this study, a hypothetical fast spectrum Liquid Fuel Molten Salt Reactor (LFMSR) core is modeled using the multiphysics C++ code GeN-Foam (General Nuclear Foam). GeN-Foam is based on OpenFOAM, a C++ open-source library for solution of continuum mechanics problems. The code utilizes a unified fine/coarse mesh approach, modeling different physics such as neutron kinetics, thermal-hydraulics based on porous fluid equations, and structural thermal-mechanics. A steady state analysis of a simplified two-dimensional (2D) LFMSR model has been performed assuming rotational symmetry to cross verify the code with the commercial ANSYS Computational Fluid Dynamics (CFD) code Fluent. The calculations showed very good agreement between the two codes allowing progression to a three-dimensional (3D) model simulation. A coupled neutron kinetics and CFD steady state analysis of a right-cylindrical 3D LFMSR core has been performed modeling one quarter of the core while using symmetry boundaries to reduce the computational time. Mixed uranium and plutonium chloride fuel has been selected in this preliminary study. Both 2D and 3D simulations showed appearance of recirculation zones within the right-cylinder core. These zones can be a challenge for LFMSR control and materials. A new hyperboloid design is proposed to remove recirculation zones, which is based on eight symmetrical loops. An Unprotected Loss of Flow accident (ULOF), in which the pump head is instantaneously reduced to zero, has been selected to demonstrate the safety characteristics of the reactor in one of the most challenging possible situations for LFMSR. The obtained results (e.g., reduced total precursors concentration at the core inlet and reduction of the core nominal power following the transient) confirm that GeN-Foam is capable of performing coupled LFMSR transient analysis and can be used for design analysis and optimization. Although the current design needs further assessment and development, it shows encouraging performance under ULOF conditions paving the way to the next step in the optimization process.},
language = {en},
urldate = {2020-10-09},
journal = {Nuclear Engineering and Design},
author = {Altahhan, Muhammad Ramzy and Bhaskar, Sandesh and Ziyad, Devshibhai and Balestra, Paolo and Fiorina, Carlo and Hou, Jason and Smith, Nicholas and Avramova, Maria},
month = dec,
year = {2020},
keywords = {CFD, Multiphysics, OpenFOAM, Nuclear energy, Molten Salt Reactor, GeN-Foam, Liquid fuel},
pages = {110826},
file = {ScienceDirect Full Text PDF:/home/sunmyung/Zotero/storage/5WFVL2AZ/Ramzy Altahhan et al. - 2020 - Preliminary design and analysis of Liquid Fuel Mol.pdf:application/pdf;ScienceDirect Snapshot:/home/sunmyung/Zotero/storage/XW4XVUG8/S0029549320303204.html:text/html},
}
@article{wang_review_2020,
title = {Review on neutronic/thermal-hydraulic coupling simulation methods for nuclear reactor analysis},
volume = {137},
issn = {0306-4549},
url = {http://www.sciencedirect.com/science/article/pii/S0306454919306759},
doi = {10.1016/j.anucene.2019.107165},
abstract = {In an operating nuclear reactor system, various physical phenomena of different properties are intimately linked. These multiphysics phenomena include neutronics (N), thermal-hydraulics (TH), materials science, and other subjects. Among them, the interaction between neutronics and thermal-hydraulics is of great significance in reactor design and safety analysis. In this work, different N/TH coupling methods are reviewed, including loose and tight coupling. For the studies on loose coupling, in which two physical fields are decoupled, the current research status is summarized and classified based on the coupling methods of neutronics and thermal-hydraulics. The studies of tight coupling are introduced based on multiphysics coupling platforms. The investigation shows that the number and objectives of loose coupling studies are more abundant and extensive than those of tight coupling. This indicates that loose coupling strategies are the mainstream coupling solutions in recent research. Furthermore, the solution approaches of N/TH coupling are reviewed with respect to the aspects of performance improvement and application studies, including the operator splitting (OS), Picard iteration, and Jacobian-Free Newton–Krylov (JFNK) methods. A comprehensive study of the solution approaches shows that most of the current loose coupling numerical simulations adopt the Picard iteration method, because it has higher calculation accuracy than the OS method. In contrast to the decoupling approaches such as the OS and Picard iteration methods, the JFNK method updates all physical quantities synchronously, which makes it more accurate. Hence, there are broad application prospects for N/TH tight coupling of the JFNK method.},
language = {en},
urldate = {2021-01-12},
journal = {Annals of Nuclear Energy},
author = {Wang, Jincheng and Wang, Qin and Ding, Ming},
month = mar,
year = {2020},
keywords = {JFNK method, Loose coupling method, N/TH coupling method, Operator splitting method, Picard iteration method, Tight coupling method},
pages = {107165},
file = {ScienceDirect Full Text PDF:/home/sunmyung/Zotero/storage/F5VHAFBU/Wang et al. - 2020 - Review on neutronicthermal-hydraulic coupling sim.pdf:application/pdf;ScienceDirect Snapshot:/home/sunmyung/Zotero/storage/QKBQWM8B/S0306454919306759.html:text/html},
}
@techreport{doe_office_2021,
title = {Office of {Nuclear} {Energy} {Strategic} {Vision}},
shorttitle = {{DOE}-{NE} {Strategic} {Vision}},
url = {https://www.energy.gov/sites/prod/files/2021/01/f82/DOE-NE Strategic Vision -Web - 01.08.2021.pdf},
language = {en},
institution = {Office of Nuclear Energy},
author = {DOE, Department of Energy},
month = jan,
year = {2021},
file = {DOE-NE Strategic Vision -Web - 01.08.2021.pdf:/home/sunmyung/Zotero/storage/SFSYKIHW/DOE-NE Strategic Vision -Web - 01.08.2021.pdf:application/pdf},
}
@misc{cordis_severe_nodate,
title = {Severe {Accident} {Modeling} and {Safety} {Assessment} for {Fluid}-fuel {Energy} {Reactors} {\textbar} {SAMOSAFER} {Project} {\textbar} {H2020} {\textbar} {CORDIS} {\textbar} {European} {Commission}},
url = {https://cordis.europa.eu/project/id/847527},
urldate = {2021-01-28},
author = {CORDIS},
file = {Severe Accident Modeling and Safety Assessment for Fluid-fuel Energy Reactors | SAMOSAFER Project | H2020 | CORDIS | European Commission:/home/sunmyung/Zotero/storage/5QDPPWYC/847527.html:text/html},
}
@techreport{koning_jeff-31_2006,
title = {The {JEFF}-3.1 nuclear data library: {JEFF} report 21},
shorttitle = {The {JEFF}-3.1 nuclear data library},
language = {en},
number = {21},
institution = {OECD Nuclear Energy Agency},
author = {Koning, Arjan and Forrest, Robin and Kellett, Mark and Mills, Robert and Henriksson, Hans and Rugama, Yolanda},
year = {2006},
note = {OCLC: 1039431841},
file = {Koning and OECD Nuclear Energy Agency - 2006 - The JEFF-3.1 nuclear data library JEFF report 21.pdf:/home/sunmyung/Zotero/storage/KNHM5TBB/Koning and OECD Nuclear Energy Agency - 2006 - The JEFF-3.1 nuclear data library JEFF report 21.pdf:application/pdf},
}
@mastersthesis{fairhurst-agosta_multi-physics_2020,
address = {Urbana, IL},
title = {Multi-{Physics} and {Technical} {Analysis} of {High}-{Temperature} {Gas}-{Cooled} {Reactors} for {Hydrogen} {Production}},
copyright = {Copyright 2020 Roberto Fairhurst Agosta},
abstract = {The future energy needs require the development of clean energy sources to ease the increasing environmental concerns. High-Temperature Gas-cooled Reactors have several desirable features that make them ideal candidates for the near-future large-scale deployment. Some of these features are a high temperature and high thermal cycle efficiency, which enable a wide range of process heat applications, such as hydrogen production. Implementing hydrogen economies can decarbonize the transport and power sectors, offering an alternative to ease climate change.
This work uses Moltres as the primary simulation tool. Although Moltres original development targeted Molten Salt Reactors, this work studies Moltres applicability to multi-physics simulations of prismatic High-Temperature Gas-cooled Reactors. Multi-physics simulations are necessary for assessing reactor safety characteristics. Ensuring
Moltres’ multi-physics modeling capabilities requires assessing the independent modeling capabilities of the different physical phenomena. Therefore, this thesis breaks down the analysis into three parts: stand-alone neutronics, stand-alone thermal-fluids, and coupled neutronics/thermal-fluids.
Regarding stand-alone neutronics, several analyses compare the results calculated by Moltres and Serpent on an MHTGR-350 model. The first analysis studies the energy group structure effects on the simulation of a fuel column. The results of the study suggest using a 15-energy group structure for attaining a desirable accuracy. The following analysis focuses on the full-core problem and compares different aspects of the simulations, concluding that Moltres obtains reasonably accurate results. The final study on stand-alone neutronics describes Moltres results of Phase I Exercise 1 of the OECD/NEA MHTGR-350 Benchmark. The benchmark exercise proved to be a modeling challenge, requiring the implementation of several approximations. For the most part, this thesis demonstrates Moltres’ capability to simulate stand-alone neutronics of prismatic High-Temperature Gas-cooled Reactors.
Regarding stand-alone thermal-fluids, several studies compare Moltres results to previously published results.
These studies focus on local models such as the unit cell and the fuel column problems, for which Moltres temperature results differ by less than 2\% from the published results. Further studies analyze the possibility of extending the thermal-fluids model implemented in the previous problems to a full-core simulation, finding a high memory
requirement imposed by the simulations. The full-core simulations focus on Phase I Exercise 2 of the benchmark, for which the implementation of a two-level approach in Moltres was necessary. The study’s temperatures were within an 11.3\% difference to the published results, concluding that further analysis is required.
Regarding coupled neutronics/thermal-fluids, the analysis describes Phase I Exercise 3 of the benchmark. The exercise uses a simplified model that helps visualize some of the essential aspects of multi-physics simulations in Moltres. This exercise finds some areas of improvement in Moltres’ model and sets a basis for future work.
This thesis aligns with the University of Illinois’ goals to reduce carbon emissions from its campus’s electricity generation and transportation sectors. This work focuses on two main analysis by introducing a nuclear reactor coupled to a hydrogen plant as a solution. The first analysis evaluates the conversion of the university fleet and the mass transit transport system in Urbana-Champaign to Fuel Cell Electric Vehicles. The second analysis investigates the duck curve phenomenon in the university’s grid and introduces a mitigation strategy that may reduce the reliance on dispatchable sources. These studies emphasize how nuclear energy and hydrogen production can potentially mitigate climate change.},
language = {English (US)},
school = {University of Illinois at Urbana-Champaign},
author = {Fairhurst-Agosta, R.},
month = dec,
year = {2020},
file = {Fairhurst-Agosta - Multi-Physics and Technical Analysis of High-Tempe.pdf:/home/sunmyung/Zotero/storage/PTK9576U/Fairhurst-Agosta - Multi-Physics and Technical Analysis of High-Tempe.pdf:application/pdf},
}
@article{peterson_overview_2018,
title = {Overview of the {Incompressible} {Navier}-{Stokes} simulation capabilities in the {MOOSE} {Framework}},
volume = {119},
doi = {10.1016/j.advengsoft.2018.02.004},
abstract = {The Multiphysics Object Oriented Simulation Environment (MOOSE) framework is a high-performance, open source, C++ finite element toolkit developed at Idaho National Laboratory. MOOSE was created with the aim of assisting domain scientists and engineers in creating customizable, high-quality tools for multiphysics simulations. While the core MOOSE framework itself does not contain code for simulating any particular physical application, it is distributed with a number of physics "modules" which are tailored to solving e.g. heat conduction, phase field, and solid/fluid mechanics problems. In this report, we describe the basic equations, finite element formulations, software implementation, and regression/verification tests currently available in MOOSE's navier\_stokes module for solving the Incompressible Navier-Stokes (INS) equations.},
journal = {Advances in Engineering Software},
author = {Peterson, John and Lindsay, Alexander and Kong, Fande},
month = may,
year = {2018},
keywords = {65N30, Mathematics - Numerical Analysis},
pages = {68--92},
annote = {Comment: 54 pages, 16 figures, includes peer reviewer revisions},
file = {arXiv\:1710.08898 PDF:/home/sunmyung/Zotero/storage/UHKRDGN5/Peterson et al. - 2017 - Overview of the Incompressible Navier-Stokes simul.pdf:application/pdf;arXiv.org Snapshot:/home/sunmyung/Zotero/storage/DIYPZSJQ/1710.html:text/html;Full Text PDF:/home/sunmyung/Zotero/storage/KXTKFUB8/Peterson et al. - 2018 - Overview of the Incompressible Navier-Stokes simul.pdf:application/pdf},
}
@article{shi_gen-foam_2021,
title = {Gen-foam {Model} and {Benchmark} of {Delayed} {Neutron} {Precursor} {Drift} in the {Molten} {Salt} {Reactor} {Experiment}},
volume = {247},
copyright = {© The Authors, published by EDP Sciences, 2021},
issn = {2100-014X},
url = {https://www.epj-conferences.org/articles/epjconf/abs/2021/01/epjconf_physor2020_06040/epjconf_physor2020_06040.html},
doi = {10.1051/epjconf/202124706040},
abstract = {The effective delayed neutron fraction is an important reactor kinetics parameter. In flowing liquid-fuel reactors, this differs from the delayed neutron fraction because of the emission of delayed neutrons with a lower energy spectrum than prompt and the delayed neutron precursor (DNP) drift due to the fuel movement. In general, neglecting delayed neutron precursor drift leads to an over-estimation of the effective delayed neutron fraction. Nevertheless, the capability to simulate this peculiar phenomenon is not available in most reactor physics tools. In this project, a multi-physics approach to modeling DNP drift is developed using the GeN-Foam toolkit, and it benchmarked against available experimental data from the Molten Salt Reactor Experiment (MSRE). GeN-Foam couples a neutron diffusion solver with a thermal-hydraulics solver. Additionally, a new function was added for solving adjoint multi-group diffusion eigenvalue problems and calculating effective delayed neutron fraction. For benchmarking, an R-Z model of the MSRE was developed in GeN-Foam. The porous media model was applied, and cross sections were generated using the Monte Carlo code Serpent-2 with ENDF/B-VII.1 nuclear data library. In order to evaluate the impact of DNP drift, two steady-state conditions (stationary and flowing salt at 1200 gpm) were simulated. A reactivity change of -241 pcm was calculated using GeN-Foam for the MSRE between static and flowing fuel, which is in a good agreement with the experimental value of -212 pcm. The total effective delayed neutron fraction change was calculated to be -230 pcm vs. -304 pcm reported for the MSRE and analytical calculated during the experimental campaign. Three transient accidents were also analyzed.},
language = {en},
urldate = {2021-05-28},
journal = {EPJ Web of Conferences},
author = {Shi, Jun and Fratoni, Massimiliano},
year = {2021},
note = {JC0007},
pages = {06040},
file = {Full Text PDF:/home/sunmyung/Zotero/storage/EMKLDDCE/Shi and Fratoni - 2021 - GEN-FOAM MODEL AND BENCHMARK OF DELAYED NEUTRON PR.pdf:application/pdf;Snapshot:/home/sunmyung/Zotero/storage/PSY22SMX/epjconf_physor2020_06040.html:text/html},
}
@incollection{kuhlmann_lid-driven_2018,
title = {The {Lid}-{Driven} {Cavity}},
isbn = {978-3-319-91493-0},
abstract = {The lid-driven cavity is an important fluid mechanical system serving as a benchmark for testing numerical methods and for studying fundamental aspects of incompressible flows in confined volumes which are driven by the tangential motion of a bounding wall. A comprehensive review is provided of lid-driven cavity flows focusing on the evolution of the flow as the Reynolds number is increased. Understanding the flow physics requires to consider pure two-dimensional flows, flows which are periodic in one space direction as well as the full three-dimensional flow. The topics treated range from the characteristic singularities resulting from the discontinuous boundary conditions over flow instabilities and their numerical treatment to the transition to chaos in a fully confined cubical cavity. In addition, the streamline topology of two-dimensional time-dependent and of steady three-dimensional flows are covered, as well as turbulent flow in a fully confined lid-driven cube. Finally, an overview on various extensions of the lid-driven cavity is given.},
booktitle = {Computational {Methods} in {Applied} {Sciences}},
author = {Kuhlmann, Hendrik and Romanò, Francesco},
month = apr,
year = {2018},
doi = {10.1007/978-3-319-91494-7_8},
note = {Journal Abbreviation: Computational Methods in Applied Sciences},
file = {Full Text PDF:/home/sunmyung/Zotero/storage/GYBGM3QT/Kuhlmann and Romanò - 2018 - The Lid-Driven Cavity.pdf:application/pdf},
}
@inproceedings{abou-jaoude_coupled_2020,
title = {Coupled {Multiphysics} {Simulation} of {Pool}-{Type} {Molten} {Salt} {Reactors} {Using} {Griffin}/{Pronghorn}},
doi = {10.13182/T123-33466},
author = {Abou-Jaoude, A. and Hermosillo, A. and Martin, Nicolas and Schunert, Sebastian and Wang, Y. and Balestra, Paolo},
month = jan,
year = {2020},
pages = {1587--1590},
file = {Abou-Jaoude et al. - 2020 - Coupled Multiphysics Simulation of Pool-Type Molte.pdf:/home/sunmyung/Zotero/storage/TJHCY6NV/Abou-Jaoude et al. - 2020 - Coupled Multiphysics Simulation of Pool-Type Molte.pdf:application/pdf},
}
@article{wang_rattlesnake_2021,
title = {Rattlesnake: {A} {MOOSE}-{Based} {Multiphysics} {Multischeme} {Radiation} {Transport} {Application}},
volume = {0},
issn = {0029-5450},
shorttitle = {Rattlesnake},
url = {https://doi.org/10.1080/00295450.2020.1843348},
doi = {10.1080/00295450.2020.1843348},
abstract = {Advanced reactor concepts span the spectrum from heat pipe–cooled microreactors, through thermal and fast molten-salt reactors, to gas- and salt-cooled pebble bed reactors. The modeling and simulation of each of these reactor types comes with their own geometrical complexities and multiphysics challenges. However, the common theme for all nuclear reactors is the necessity to be able to accurately predict neutron distribution in the presence of multiphysics feedback. We argue that the current standards of modeling and simulation, which couple single-physics, single-reactor-focused codes via ad hoc methods, are not sufficiently flexible to address the challenges of modeling and simulation for advanced reactors. In this work, we present the Multiphysics Object Oriented Simulation Environment (MOOSE)–based radiation transport application Rattlesnake. The use of Rattlesnake for the modeling and simulation of nuclear reactors represents a paradigm shift away from makeshift data exchange methods, as it is developed based on the MOOSE platform with its very natural form of shared data distribution. Rattlesnake is well equipped for addressing the geometric and multiphysics challenges of advanced reactor concepts because it is a flexible finite element tool that leverages the multiphysics capabilities inherent in MOOSE. This paper focuses on the concept and design of Rattlesnake. We also demonstrate the capabilities and performance of Rattlesnake with a set of problems including a microreactor, a molten-salt reactor, a pebble bed reactor, the Advanced Test Reactor at the Idaho National Laboratory, and two benchmarks: a multiphysics version of the C5G7 benchmark and the LRA benchmark.},
number = {0},
urldate = {2021-06-21},
journal = {Nuclear Technology},
author = {Wang, Yaqi and Schunert, Sebastian and Ortensi, Javier and Laboure, Vincent and DeHart, Mark and Prince, Zachary and Kong, Fande and Harter, Jackson and Balestra, Paolo and Gleicher, Frederick},
month = apr,
year = {2021},
note = {Publisher: Taylor \& Francis
\_eprint: https://doi.org/10.1080/00295450.2020.1843348},
keywords = {MOOSE, multiphysics, reactor physics, radiation transport, Rattlesnake},
pages = {1--26},
file = {Full Text PDF:/home/sunmyung/Zotero/storage/P959F3R3/Wang et al. - 2021 - Rattlesnake A MOOSE-Based Multiphysics Multischem.pdf:application/pdf;Snapshot:/home/sunmyung/Zotero/storage/CDETVBF8/00295450.2020.html:text/html},
}
@article{brooks_streamline_1982,
title = {Streamline upwind/{Petrov}-{Galerkin} formulations for convection dominated flows with particular emphasis on the incompressible {Navier}-{Stokes} equations},
volume = {32},
issn = {0045-7825},
url = {https://www.sciencedirect.com/science/article/pii/0045782582900718},
doi = {10.1016/0045-7825(82)90071-8},
abstract = {A new finite element formulation for convection dominated flows is developed. The basis of the formulation is the streamline upwind concept, which provides an accurate multidimensional generalization of optimal one-dimensional upwind schemes. When implemented as a consistent Petrov-Galerkin weighted residual method, it is shown that the new formulation is not subject to the artificial diffusion criticisms associated with many classical upwind methods. The accuracy of the streamline upwind/Petrov-Galerkin formulation for the linear advection diffusion equation is demonstrated on several numerical examples. The formulation is extended to the incompressible Navier-Stokes equations. An efficient implicit pressure/explicit velocity transient algorithm is developed which accomodates several treatments of the incompressibility constraint and allows for multiple iterations within a time step. The effectiveness of the algorithm is demonstrated on the problem of vortex shedding from a circular cylinder at a Reynolds number of 100.},
language = {en},
number = {1},
urldate = {2021-06-21},
journal = {Computer Methods in Applied Mechanics and Engineering},
author = {Brooks, Alexander N. and Hughes, Thomas J. R.},
month = sep,
year = {1982},
pages = {199--259},
file = {ScienceDirect Full Text PDF:/home/sunmyung/Zotero/storage/FQ6XS4RK/Brooks and Hughes - 1982 - Streamline upwindPetrov-Galerkin formulations for.pdf:application/pdf;ScienceDirect Snapshot:/home/sunmyung/Zotero/storage/ILNUHQDT/0045782582900718.html:text/html},
}
@article{hughes_new_1986,
title = {A new finite element formulation for computational fluid dynamics: {V}. {Circumventing} the babuška-brezzi condition: a stable {Petrov}-{Galerkin} formulation of the stokes problem accommodating equal-order interpolations},
volume = {59},
issn = {0045-7825},
shorttitle = {A new finite element formulation for computational fluid dynamics},
url = {https://www.sciencedirect.com/science/article/pii/0045782586900253},
doi = {10.1016/0045-7825(86)90025-3},
abstract = {A new Petrov-Galerkin formulation of the Stokes problem is proposed. The new formulation possesses better stability properties than the classical Galerkin/variational method. An error analysis is performed for the case in which both the velocity and pressure are approximated by C0 interpolations. Combinations of C0 interpolations which are unstable according to the Babuška-Brezzi condition (e.g., equal-order interpolations) are shown to be stable and convergent within the present framework. Calculations exhibiting the good behavior of the methodology are presented.},
language = {en},
number = {1},
urldate = {2021-06-22},
journal = {Computer Methods in Applied Mechanics and Engineering},
author = {Hughes, Thomas J. R. and Franca, Leopoldo P. and Balestra, Marc},
month = nov,
year = {1986},
pages = {85--99},
file = {ScienceDirect Full Text PDF:/home/sunmyung/Zotero/storage/2GD7L83S/Hughes et al. - 1986 - A new finite element formulation for computational.pdf:application/pdf;ScienceDirect Snapshot:/home/sunmyung/Zotero/storage/XVXDPTRA/0045782586900253.html:text/html},
}
@article{chapelle_inf-sup_1993,
title = {The inf-sup test},
volume = {47},
issn = {0045-7949},
url = {https://www.sciencedirect.com/science/article/pii/004579499390340J},
doi = {10.1016/0045-7949(93)90340-J},
abstract = {We briefly review the inf-sup condition for the finite element solution of problems in incompressible elasticity, and then propose a numerical test on whether the inf-sup condition is passed. The evaluation of elements with this test is simple, and various results are presented. This inf-sup test will prove useful for many discretizations of constrained variational problems.},
language = {en},
number = {4},
urldate = {2021-06-22},
journal = {Computers \& Structures},
author = {Chapelle, D. and Bathe, K. J.},
month = jun,
year = {1993},
pages = {537--545},
file = {ScienceDirect Full Text PDF:/home/sunmyung/Zotero/storage/LPZPUCQC/Chapelle and Bathe - 1993 - The inf-sup test.pdf:application/pdf;ScienceDirect Snapshot:/home/sunmyung/Zotero/storage/WTBAFE57/004579499390340J.html:text/html},
}
@inproceedings{tiberga_discontinuous_2019,
address = {Portland, OR},
title = {A {DISCONTINUOUS} {GALERKIN} {FEM} {MULTI}-{PHYSICS} {SOLVER} {FOR} {THE} {MOLTEN} {SALT} {FAST} {REACTOR}},
abstract = {Numerical simulations of fast MSRs constitute a challenging task. In fact, classical codes employed in reactor physics cannot be used, and new dedicated multi-physics tools must be developed, to capture the unique features of these systems: the strong coupling between neutronics and thermal-hydraulics due to the use of a liquid fuel, the effects on reactor kinetics induced by the precursors drift, the internal heat generation, and the shape of the core having no fuel pins as a repeated structure. In this work, we present a novel multi-physics tool being developed at TU Delft. The coupling is realized between an SN radiation transport code (PHANTOM-SN) and a RANS solver (DGFlows). Both in-house tools are based on a Discontinuous Galerkin Finite Element space discretization, characterized by local conservation, high-order accuracy, and allowing for high geometric flexibility. Implicit discretization in time is performed adopting Backward Differentiation Formulae. Cross sections are computed on an element base, starting from the local average temperature and a set of libraries generated at reference temperatures with Monte Carlo or deterministic codes. Comparison of the results obtained performing a suitable numerical benchmark created at LPSC/CNRS/Grenoble with those available in literature shows that the multi-physics tool is able to capture the unique phenomena characterizing fast liquid-fueled systems.},
booktitle = {Proceedings of the {International} {Conference} on {Mathematics} and {Computational} {Methods} applied to {Nuclear} {Science} and {Engineering} ({M}\&{C} 2019)},
publisher = {American Nuclear Society},
author = {Tiberga, Marco and Lathouwers, D. and Kloosterman, Jan},
month = aug,
year = {2019},
file = {Full Text PDF:/home/sunmyung/Zotero/storage/RA4MFL2T/Tiberga et al. - 2019 - A DISCONTINUOUS GALERKIN FEM MULTI-PHYSICS SOLVER .pdf:application/pdf},
}
@phdthesis{blanco_neutronic_2020,
type = {phdthesis},
title = {Neutronic, thermohydraulic and thermomechanical coupling for the modeling of criticality accidents in nuclear systems},
url = {https://tel.archives-ouvertes.fr/tel-03172978},
abstract = {This thesis was developed within in the framework of the multi-scale and multi-physics models for the simulation of criticality accidents, carried jointly between the CNRS and the IRSN. A multi-physics and multi-scale approach aims to produce a numerical model taking into account all the relevant physical phenomena existing in nuclear systems as well as their coupling. This approach makes possible to improve the predictive capacities of the single physics models and to numerically study the behavior of a nuclear system under conditions that would be difficult to achieve or reproduce by experiments. The multi-scale / multi-physics approach is, therefore, particularly useful for the study of nuclear reactor criticality accidents, or more generally, for all nuclear systems where a tight coupling exists between neutronics, mechanics (of solids and fluids) and heat transfers.The objectives of the thesis were, firstly, to develop a new numerical scheme for the coupling between the neutronic code Serpent 2 (Monte Carlo code) and the Computational Fluid Dynamics (CFD) code OpenFOAM. Secondly, to develop the physical models that allow greater flexibility for criticality accidents studies in terms of type of transients, systems and phenomena considered. Among the various physical models developed during the work, it can be mentioned the transient neutronic models based on a quasi-static Monte Carlo approach and on the deterministic SP1 and SP3 methods. A porous medium model was also developed during the work to allow performing studies on nuclear systems containing a solid nuclear fuel cooled by a fluid. The numerical implementation of the multi-physics coupling was performed in C/C++ in the OpenFOAM code. This code is very well suited to numerically solve continuous mechanics problems using a finite volume method. It also provides very large library of CFD algorithms (RANS, LES et DNS). The thesis work specially focused on the study of the strategy to be followed to implement the quasi-static method numerically with a Monte Carlo type code in the same platform through internal coupling.The performances of the coupling and the developed models were studied for different scenarios and nuclear systems: the transient Godiva experiments, an international benchmark for multi-physics codes for Molten Salts Reactors and the case of a hypothetical criticality accident in a Boiling Water Reactor (BWR) spent fuel pool. These diverse scenarios and systems were selected because they are characterized by presenting a multitude of highly coupled physical phenomena which required a very careful modeling. One can mention: the Doppler and fuel density effects, the thermal expansion and thermomechanical stresses, the presence of laminar or turbulent flows in the coolant or liquid fuel, the delayed neutrons precursors convection, and the energy and mass transfers and the phase change in porous media. The different comparisons between the multi-physics tool and the available data show a very good agreement and confirm that the selected approach is pertinent for the study of criticality accidents and allows obtaining very good precision and flexibility while maintaining satisfactory computational costs.},
language = {en},
urldate = {2021-06-23},
school = {Université Grenoble Alpes [2020-....]},
author = {Blanco, Juan Antonio},
month = dec,
year = {2020},
file = {Full Text PDF:/home/sunmyung/Zotero/storage/PWC7UK9X/Blanco - 2020 - Neutronic, thermohydraulic and thermomechanical co.pdf:application/pdf;Snapshot:/home/sunmyung/Zotero/storage/MCS7NXPH/tel-03172978.html:text/html},
}
@article{boer_validation_2010,
title = {Validation of the {DALTON}-{THERMIX} {Code} {System} with {Transient} {Analyses} of the {HTR}-10 and {Application} to the {PBMR}},
volume = {170},
issn = {0029-5450},
url = {https://doi.org/10.13182/NT10-A9485},
doi = {10.13182/NT10-A9485},
abstract = {The DALTON-THERMIX code system has been developed for safety analysis and core optimization of pebble-bed reactors. The code system consists of the new three-dimensional diffusion code DALTON, which is coupled to the existing thermal-hydraulic code THERMIX. These codes are linked to a database procedure for the generation of neutron cross sections using SCALE-5.The behavior of pebble-bed reactors during a loss of forced cooling (LOFC) transient is of particular interest since the absence of forced cooling could lead to a significant increase of the temperature of the coated particle fuel. Therefore, the reactor power may be constrained during normal operation to limit the temperature.For validation purposes, calculation results of normal operation, an LOFC transient, and a control rod withdrawal transient without SCRAM have been compared with experimental data obtained in the High Temperature Reactor–10 (HTR-10). The code system has been applied to the 400-MW(thermal) pebble bed modular reactor (PBMR) design, including the analysis of three different LOFC transients. Theses results are verified by a comparison with the results of the existing TINTE code system.It was found that the code system is capable of modeling both small (HTR-10) and large (PBMR) pebble-bed reactors and therefore provides a flexible tool for safety analysis and core optimization of future reactor designs. The analyses of the LOFC transients show that the peak fuel temperature is only slightly elevated (less than +100°C) as compared to its nominal value in the HTR-10 but reaches a maximum value of 1648°C during the depressurized LOFC case of the PBMR benchmark, which is significantly higher than the peak fuel temperature (976°C) during normal operation.},
number = {2},
urldate = {2021-06-23},
journal = {Nuclear Technology},
author = {Boer, B. and Lathouwers, D. and Kloosterman, J. L. and Hagen, T. H. J. J. Van Der and Strydom, G.},
month = may,
year = {2010},
note = {Publisher: Taylor \& Francis
\_eprint: https://doi.org/10.13182/NT10-A9485},
pages = {306--321},
file = {Full Text PDF:/home/sunmyung/Zotero/storage/BVWYSKV5/Boer et al. - 2010 - Validation of the DALTON-THERMIX Code System with .pdf:application/pdf;Snapshot:/home/sunmyung/Zotero/storage/XGK4NEIN/NT10-A9485.html:text/html},
}
@article{de_zwaan_static_2007,
title = {Static design of a liquid-salt-cooled pebble bed reactor ({LSPBR})},
volume = {34},
issn = {0306-4549},
url = {https://www.sciencedirect.com/science/article/pii/S0306454906002295},
doi = {10.1016/j.anucene.2006.11.008},
abstract = {A renewed interest has been raised for liquid-salt-cooled nuclear reactors. The excellent heat transfer properties of liquid-salt coolants provide several benefits, like lower fuel temperatures, higher average coolant temperature, increased core power density and better decay heat removal, and thus higher achievable core power. In order to benefit from the on-line refueling capability of a pebble bed reactor, the liquid salt pebble bed reactor (LSPBR) is proposed. This is a high temperature pebble bed reactor with a fuel design similar to existing HTRs, but using a liquid-salt as coolant. In this paper, the selection criteria for the liquid-salt coolant are described. Based on its neutronic properties, LiF–BeF2 (flibe) was selected for the LSPBR. Two designs of the LSPBR were considered: a cylindrical core and an annular core with a graphite inner reflector. Coupled neutronic thermal-hydraulic calculations were performed to obtain the steady state power distribution and the corresponding fuel temperature distribution. Calculations were performed to investigate the decay heat removal capability in a protected loss-of-forced cooling accident. The maximum allowable power that can be produced with the LSPBR is hereby determined.},
language = {en},
number = {1},
urldate = {2021-06-23},
journal = {Annals of Nuclear Energy},
author = {de Zwaan, S. J. and Boer, B. and Lathouwers, D. and Kloosterman, J. L.},
month = jan,
year = {2007},
pages = {83--92},
file = {ScienceDirect Full Text PDF:/home/sunmyung/Zotero/storage/62KZ3WDB/de Zwaan et al. - 2007 - Static design of a liquid-salt-cooled pebble bed r.pdf:application/pdf;ScienceDirect Snapshot:/home/sunmyung/Zotero/storage/D37AJ697/S0306454906002295.html:text/html},
}
@techreport{downar_parcs_2010,
address = {Ann Arbor, MI},
type = {Theory {Manual}},
title = {{PARCS} v3.0 {U}.{S}. {NRC} {Core} {Neutronics} {Simulator}.},
language = {en},
institution = {University of Michigan},
author = {Downar, T and Xu, Y and Seker, V},
year = {2010},
pages = {129},
file = {Downar - PARCS v3.0 U.S. NRC Core Neutronics Simulator.pdf:/home/sunmyung/Zotero/storage/DI2FT6AP/Downar - PARCS v3.0 U.S. NRC Core Neutronics Simulator.pdf:application/pdf},
}
@misc{the_openfoam_foundation_ltd_openfoam_2021,
address = {England},
title = {{OpenFOAM}},
url = {https://openfoam.org/},
urldate = {2021-06-24},
publisher = {The OpenFOAM Foundation Ltd},
author = {{The OpenFOAM Foundation Ltd}},
year = {2021},
file = {Snapshot:/home/sunmyung/Zotero/storage/UPAPF2JA/openfoam.org.html:text/html},
}
@inproceedings{fiorina_creation_2018,
title = {Creation of an {OpenFOAM} {Fuel} {Performance} {Class} {Based} on {FRED} and {Integration} {Into} the {GeN}-{Foam} {Multi}-{Physics} {Code}},
url = {https://asmedigitalcollection.asme.org/ICONE/proceedings/ICONE26/51456/V003T02A027/272722},
doi = {10.1115/ICONE26-81574},
language = {en},
urldate = {2021-06-24},
publisher = {American Society of Mechanical Engineers Digital Collection},
author = {Fiorina, Carlo and Pautz, Andreas and Mikityuk, Konstantin},
month = oct,
year = {2018},
file = {Full Text PDF:/home/sunmyung/Zotero/storage/YJUREMLU/Fiorina et al. - 2018 - Creation of an OpenFOAM Fuel Performance Class Bas.pdf:application/pdf;Snapshot:/home/sunmyung/Zotero/storage/FT6XYEX2/272722.html:text/html},
}
@article{permann_moose_2020,
title = {{MOOSE}: {Enabling} massively parallel multiphysics simulation},
volume = {11},
issn = {2352-7110},
shorttitle = {{MOOSE}},
url = {https://www.sciencedirect.com/science/article/pii/S2352711019302973},
doi = {10.1016/j.softx.2020.100430},
abstract = {Harnessing modern parallel computing resources to achieve complex multiphysics simulations is a daunting task. The Multiphysics Object Oriented Simulation Environment (MOOSE) aims to enable such development by providing simplified interfaces for specification of partial differential equations, boundary conditions, material properties, and all aspects of a simulation without the need to consider the parallel, adaptive, nonlinear, finite element solve that is handled internally. Through the use of interfaces and inheritance, each portion of a simulation becomes reusable and composable in a manner that allows disparate research groups to share code and create an ecosystem of growing capability that lowers the barrier for the creation of multiphysics simulation codes. Included within the framework is a unique capability for building multiscale, multiphysics simulations through simultaneous execution of multiple sub-applications with data transfers between the scales. Other capabilities include automatic differentiation, scaling to a large number of processors, hybrid parallelism, and mesh adaptivity. To date, MOOSE-based applications have been created in areas of science and engineering such as nuclear physics, geothermal science, magneto-hydrodynamics, seismic events, compressible and incompressible fluid flow, microstructure evolution, and advanced manufacturing processes.},
language = {en},
urldate = {2021-08-12},
journal = {SoftwareX},
author = {Permann, Cody J. and Gaston, Derek R. and Andrš, David and Carlsen, Robert W. and Kong, Fande and Lindsay, Alexander D. and Miller, Jason M. and Peterson, John W. and Slaughter, Andrew E. and Stogner, Roy H. and Martineau, Richard C.},
month = jan,
year = {2020},
keywords = {Multiphysics, Parallel, Finite-element, Framework, Multiscale},
pages = {100430},
file = {ScienceDirect Full Text PDF:/home/sunmyung/Zotero/storage/Q27QTRPG/Permann et al. - 2020 - MOOSE Enabling massively parallel multiphysics si.pdf:application/pdf;ScienceDirect Snapshot:/home/sunmyung/Zotero/storage/ZGBGE5S9/S2352711019302973.html:text/html},
}
@misc{park_results_2021,
title = {Results from {Moltres} for the {CNRS} {Benchmark}},
url = {https://zenodo.org/record/5534964},
abstract = {Raw data from running the CNRS benchmark with Moltres.},
language = {eng},
urldate = {2021-09-28},
publisher = {Zenodo},
author = {Park, Sun Myung and Munk, Madicken and Huff, Kathryn D.},
month = sep,