Дисертації з теми "Numerical Methods for Neutron Transport"
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ALCARO, FABIO. "Quasi-static Methods in Neutron Transport." Doctoral thesis, Politecnico di Torino, 2012. http://hdl.handle.net/11583/2501653.
Повний текст джерелаBARBARINO, ANDREA. "Numerical Methods for Neutron Transport Calculations of Nuclear Reactors." Doctoral thesis, Politecnico di Torino, 2014. http://hdl.handle.net/11583/2561774.
Повний текст джерелаMarquez, Damian Jose Ignacio. "Multilevel acceleration of neutron transport calculations." Thesis, Atlanta, Ga. : Georgia Institute of Technology, 2007. http://hdl.handle.net/1853/19731.
Повний текст джерелаCommittee Chair: Stacey, Weston M.; Committee Co-Chair: de Oliveira, Cassiano R.E.; Committee Member: Hertel, Nolan; Committee Member: van Rooijen, Wilfred F.G.
Blackburn, Megan Satterfield. "Numerical benchmarking of a coarse-mesh transport (COMET) method for medical physics applications." Diss., Atlanta, Ga. : Georgia Institute of Technology, 2009. http://hdl.handle.net/1853/29763.
Повний текст джерелаCommittee Chair: Farzad Rahnema; Committee Co-Chair: Eric Elder; Committee Member: C.-K. Chris Wang; Committee Member: Rebecca Howell; Committee Member: Sang Cho. Part of the SMARTech Electronic Thesis and Dissertation Collection.
Byambaakhuu, Tseelmaa. "Development of Advanced Numerical Methods for Solving Neutron Transport Problems: DG-DSA and the Shishkin Mesh for Problems with Sharp Layers." The Ohio State University, 2021. http://rave.ohiolink.edu/etdc/view?acc_num=osu1618855174338701.
Повний текст джерелаBlake, Jack. "Domain decomposition methods for nuclear reactor modelling with diffusion acceleration." Thesis, University of Bath, 2016. https://ethos.bl.uk/OrderDetails.do?uin=uk.bl.ethos.698988.
Повний текст джерелаABRATE, NICOLO'. "Methods for safety and stability analysis of nuclear systems." Doctoral thesis, Politecnico di Torino, 2022. http://hdl.handle.net/11583/2971611.
Повний текст джерелаDi, Chicco Augusto. "Optimization of a calculation scheme through the parametric study of effective nuclear cross sections and application to the estimate of neutronic parameters of the ASTRID fast nuclear reactor." Master's thesis, Alma Mater Studiorum - Università di Bologna, 2018.
Знайти повний текст джерелаSheehan, B. P. "Multigrid methods for isotropic neutron transport." Connect to online resource, 2007. http://gateway.proquest.com/openurl?url_ver=Z39.88-2004&rft_val_fmt=info:ofi/fmt:kev:mtx:dissertation&res_dat=xri:pqdiss&rft_dat=xri:pqdiss:3256437.
Повний текст джерелаMuddle, John Christopher. "Advanced numerical methods for neutron star interfaces." Thesis, University of Southampton, 2015. https://eprints.soton.ac.uk/375551/.
Повний текст джерелаBaker, David James. "Characteristic-based methods for modelling neutron transport." Thesis, University of Nottingham, 2012. http://ethos.bl.uk/OrderDetails.do?uin=uk.bl.ethos.580409.
Повний текст джерелаScheben, Fynn. "Iterative methods for criticality computations in neutron transport theory." Thesis, University of Bath, 2011. https://ethos.bl.uk/OrderDetails.do?uin=uk.bl.ethos.545319.
Повний текст джерелаBennison, Tom. "Adaptive discontinuous Galerkin methods for the neutron transport equation." Thesis, University of Nottingham, 2014. http://eprints.nottingham.ac.uk/28944/.
Повний текст джерелаLanser, Debby. "Numerical methods for atmospheric flow and transport problems." [S.l. : Amsterdam : s.n.] ; Universiteit van Amsterdam [Host], 2002. http://dare.uva.nl/document/64490.
Повний текст джерелаMauger, R. L. "Nodal methods for solving the neutron transport equation for reactor analysis." Thesis, Imperial College London, 1988. http://hdl.handle.net/10044/1/47178.
Повний текст джерелаWeston, Joseph. "Numerical methods for time-resolved quantum nanoelectronics." Thesis, Université Grenoble Alpes (ComUE), 2016. http://www.theses.fr/2016GREAY040/document.
Повний текст джерелаRecent technical progress in the field of quantum nanoelectronics have lead toexciting new experiments involving coherent single electron sources.When quantum electronic devices are manipulated on time scales shorterthan the characteristic time of flight of electrons through the device, a wholeclass of conceptually new possibilities become available. In order totreat such physical situations, corresponding advances in numerical techniquesand their software implementation are required both as a tool to aidunderstanding, and also to help when designing the next generation ofexperiments in this domain.Recent advances in numerical methods have lead to techniques for which thecomputation times scales linearly with the system volume, but as thesquare of the simulation time desired. This is particularly problematicfor cases where the characteristic dwell time of electrons in the centraldevice is much longer than the ballistic time of flight. Here, we proposean improvement to an existing wavefunction based algorithm fortreating time-resolved quantum transport which scales linearly in both thesystem volume and desired simulation time. We use this technique tostudy a number of interesting physical cases. In particular we find that theapplication of a train of voltage pulses to an electronic interferometercan be used to stabilise the dynamical modification of the interferencethat was recently proposed. We use this to perform spectroscopy on Majoranaand Andreev resonances in hybrid superconductor-nanowire structures.The numerical algorithms are implemented as an extension to the Kwantquantum transport software. This implementation is used for all the numericalresults presented here, in addition to other work, covering a wide varietyof physical applications: quantum Hall effect, Floquet topological insulators,Fabry-Perot interferometers and superconducting junction
Besselman, Michael J. "Advanced Numerical Methods in General Relativistic Magnetohydrodynamics." BYU ScholarsArchive, 2012. https://scholarsarchive.byu.edu/etd/3394.
Повний текст джерелаCarreño, Sánchez Amanda María. "Integration methods for the time dependent neutron diffusion equation and other approximations of the neutron transport equation." Doctoral thesis, Universitat Politècnica de València, 2020. http://hdl.handle.net/10251/144771.
Повний текст джерела[CAT] Un dels objectius més importants per a l'anàlisi de la seguretat en el camp de l'enginyeria nuclear és el càlcul, ràpid i precís, de l'evolució de la potència dins del nucli d'un reactor. La distribució dels neutrons pot modelar-se mitjançant l'equació del transport de Boltzmann. La solució d'aquesta equació per a un reactor realístic no pot obtenir's de manera senzilla. És per això que han de considerar-se aproximacions numèriques. En primer lloc, la tesi se centra en l'obtenció de la solució per a diversos problemes estàtics associats amb l'equació de difusió neutrònica: els modes lambda, els modes gamma i els modes alpha. Per a la discretització espacial s'ha utilitzat un mètode d'elements finits d'alt ordre. Algunes de les característiques dels problemes espectrals s'analitzaran i es compararan per a diferents reactors. Tanmateix, diversos solucionadors de problemes d'autovalors i estratègies es desenvolupen per a calcular els problemes obtinguts de la discretització espacial. La majoria dels treballs per a resoldre l'equació de difusió neutrònica estan dissenyats per a l'aproximació de dos grups d'energia i sense considerar dispersió de neutrons del grup tèrmic al grup ràpid. El principal avantatge de la metodologia exposada és que no depèn de la geometria del reactor, del tipus de problema d'autovalors ni del nombre de grups d'energia del problema. Seguidament, s'obté la solució de les equacions estacionàries d'harmònics esfèrics. La implementació d'aquestes equacions té dues principals diferències respecte a l'equació de difusió. Primer, la discretització espacial es realitza a nivell de pin a partir de l'estudi de diferents malles. Segon, el nombre de grups d'energia és, generalment, major que dos. D'aquesta forma, es desenvolupen estratègies a blocs per a optimitzar el càlcul dels problemes algebraics associats. Finalment, s'implementa un mètode modal amb actualitzacions dels modes per a integrar l'equació de difusió neutrònica dependent del temps. Es presenten i es comparen els mètodes modals basats en l'expansió dels diferents modes espacials per a diversos tipus de transitoris. A més a més, un control de pas de temps adaptatiu es desenvolupa, evitant l'actualització dels modes d'una manera fixa i adaptant el pas de temps en funció de vàries estimacions de l'error.
[EN] One of the most important targets in nuclear safety analyses is the fast and accurate computation of the power evolution inside of the reactor core. The distribution of neutrons can be described by the neutron transport Boltzmann equation. The solution of this equation for realistic nuclear reactors is not straightforward, and therefore, numerical approximations must be considered. First, the thesis is focused on the attainment of the solution for several steady-state problems associated with neutron diffusion problem: the $\lambda$-modes, the $\gamma$-modes and the $\alpha$-modes problems. A high order finite element method is used for the spatial discretization. Several characteristics of each type of spectral problem are compared and analyzed on different reactors. Thereafter, several eigenvalue solvers and strategies are investigated to compute efficiently the algebraic eigenvalue problems obtained from the discretization. Most works devoted to solve the neutron diffusion equation are made for the approximation of two energy groups and without considering up-scattering. The main property of the proposed methodologies is that they depend on neither the reactor geometry, the type of eigenvalue problem nor the number of energy groups. After that, the solution of the steady-state simplified spherical harmonics equations is obtained. The implementation of these equations has two main differences with respect to the neutron diffusion. First, the spatial discretization is made at level of pin. Thus, different meshes are studied. Second, the number of energy groups is commonly bigger than two. Therefore, block strategies are developed to optimize the computation of the algebraic eigenvalue problems associated. Finally, an updated modal method is implemented to integrate the time-dependent neutron diffusion equation. Modal methods based on the expansion of the different spatial modes are presented and compared in several types of transients. Moreover, an adaptive time-step control is developed that avoids setting the time-step with a fixed value and it is adapted according to several error estimations.
Carreño Sánchez, AM. (2020). Integration methods for the time dependent neutron diffusion equation and other approximations of the neutron transport equation [Tesis doctoral no publicada]. Universitat Politècnica de València. https://doi.org/10.4995/Thesis/10251/144771
TESIS
Käser, Martin Andreas. "Adaptive methods for the numerical simulation of transport processes." [S.l.] : [s.n.], 2003. http://deposit.ddb.de/cgi-bin/dokserv?idn=970272855.
Повний текст джерелаNelson, Adam Gregory Ivanov Kostadin N. "Monte Carlo methods for neutron transport on Graphics Processing Units using CUDA." [University Park, Pa.] : Pennsylvania State University, 2009. http://etda.libraries.psu.edu/theses/approved/WorldWideIndex/ETD-4605/index.html.
Повний текст джерелаNenna, Luca. "Numerical Methods for Multi-Marginal Optimal Transportation." Thesis, Paris Sciences et Lettres (ComUE), 2016. http://www.theses.fr/2016PSLED017/document.
Повний текст джерелаIn this thesis we aim at giving a general numerical framework to approximate solutions to optimal transport (OT) problems. The general idea is to introduce an entropic regularization of the initialproblems. The regularized problem corresponds to the minimization of a relative entropy with respect a given reference measure. Indeed, this is equivalent to find the projection of the joint coupling with respect the Kullback-Leibler divergence. This allows us to make use the Bregman/Dykstra’s algorithm and solve several variational problems related to OT. We are especially interested in solving multi-marginal optimal transport problems (MMOT) arising in Physics such as in Fluid Dynamics (e.g. incompressible Euler equations à la Brenier) and in Quantum Physics (e.g. Density Functional Theory). In these cases we show that the entropic regularization plays a more important role than a simple numerical stabilization. Moreover, we also give some important results concerning existence and characterization of optimal transport maps (e.g. fractal maps) for MMOT
Douglass, Steven James. "Consistent energy treatment for radiation transport methods." Diss., Georgia Institute of Technology, 2012. http://hdl.handle.net/1853/47612.
Повний текст джерелаWagner, Carsten. "Transport phenomena in complex turbulent flows : numerical and experimental methods." kostenfrei, 2007. http://e-collection.ethbib.ethz.ch/view/eth:30077.
Повний текст джерелаMercier, Olivier. "Numerical methods for set transport and related partial differential equations." Thesis, McGill University, 2013. http://digitool.Library.McGill.CA:80/R/?func=dbin-jump-full&object_id=119767.
Повний текст джерелаDans plusieurs situations, la simulation de systèmes physiques requiert de suivre l'évolution d'un ensemble. Cet ensemble peut être un bout de tissu dans le vent, la frontière entre une masse d'eau et l'air, ou même le front d'un feu brûlant à travers une forêt. D'un point de vue numérique, transporter de tels ensembles peut être difficile, et des algorithmes pour accomplir cette tâche plus efficacement et avec plus de précision sont toujours en demande. Dans ce mémoire, nous présentons plusieurs méthodes pour suivre l'évolution d'ensembles dans un champ de vecteur donné. Nous appliquons aussi ces techniques à divers systèmes physiques où le champ vectoriel est couplé de manière non linéaire aux ensembles évolués.
Murphy, Steven. "Methods for solving discontinuous-Galerkin finite element equations with application to neutron transport." Phd thesis, Toulouse, INPT, 2015. http://oatao.univ-toulouse.fr/14650/1/murphy.pdf.
Повний текст джерелаWalsh, Jonathan A. (Jonathan Alan). "Computational methods for efficient nuclear data management in Monte Carlo neutron transport simulations." Thesis, Massachusetts Institute of Technology, 2014. http://hdl.handle.net/1721.1/95570.
Повний текст джерелаCataloged from PDF version of thesis.
Includes bibliographical references (pages 127-133).
This thesis presents the development and analysis of computational methods for efficiently accessing and utilizing nuclear data in Monte Carlo neutron transport code simulations. Using the OpenMC code, profiling studies are conducted in order to determine the types of nuclear data that are used in realistic reactor physics simulations, as well as the frequencies with which those data are accessed. The results of the profiling studies are then used to motivate the conceptualization of a nuclear data server algorithm aimed at reducing on-node memory requirements through the use of dedicated server nodes for the storage of infrequently accessed data. A communication model for this algorithm is derived and used to make performance predictions given data access frequencies and assumed system hardware parameters. Additionally, a new, accelerated approach for rejection sampling the free gas resonance elastic scattering kernel that reduces the frequency of zero-temperature elastic scattering cross section data accesses is derived and implemented. Using this new approach, the runtime overhead incurred by an exact treatment of the free gas resonance elastic scattering kernel is reduced by more than 30% relative to a standard sampling procedure used by Monte Carlo codes. Finally, various optimizations of the commonly-used binary energy grid search algorithm are developed and demonstrated. Investigated techniques include placing kinematic constraints on the range of the searchable energy grid, index lookups on unionized material energy grids, and employing energy grid hash tables. The accelerations presented routinely result in overall code speedup by factors of 1.2-1.3 for simulations of practical systems.
by Jonathan A. Walsh.
S.M.
J, Labossière-Hickman Travis. "Modeling and simulation of The Transient Reactor Test Facility using modern neutron transport methods." Thesis, Massachusetts Institute of Technology, 2019. https://hdl.handle.net/1721.1/123360.
Повний текст джерелаThesis: S.M., Massachusetts Institute of Technology, Department of Nuclear Science and Engineering, 2019
Cataloged from student-submitted PDF version of thesis.
Includes bibliographical references (pages 111-113).
The Transient Reactor Test Facility (TREAT) has regained the interest of the nuclear engineering community in recent years. While TREAT's design makes it uniquely suited to transient fuel testing, it also makes the reactor very challenging to model and simulate. In this thesis, we build a Monte Carlo model of TREAT's Minimum Critical Mass core to examine the effects of fuel impurities, calculate a reference solution, and analyze a number of multigroup cross section generation approaches. Several method of characteristics (MOC) simulations employing these cross sections are then converged in space and angle, corrected for homogenization, and compared to the Monte Carlo reference solution. The thesis concludes with recommendations for future analysis of TREAT using MOC.
by Travis J. Labossière-Hickman.
S.M.
S.M. Massachusetts Institute of Technology, Department of Nuclear Science and Engineering
Ford, Wesley. "The Advancement of Stable, Efficient and Parallel Acceleration Methods for the Neutron Transport Equation." Thesis, Université Paris-Saclay (ComUE), 2019. http://www.theses.fr/2019SACLX105/document.
Повний текст джерелаIn this paper we propose a new library of non-linear techniques for accelerating the discrete-ordinates transport equation. Two new types of nonlinear acceleration methods called Spatially Variant Rebalancing Method (SVRM) and Response Matrix Acceleration (RMA), respectively, are proposed and investigated. The first method, SVRM, is based on the computation of the zeroth and first order spatial variation of the neutron balance equation. RMA, is a DP0 method that uses knowledge of the transport operator to form a consistent relationship. Two distinct variants of RMA, called Explicit-RMA (E-RMA) and Balance (B-RMA), respectively, are derived. The convergence properties of both acceleration methods are investigated for two different iteration schemes of the method of characteristics (MOC) transport operator for a 1D slab, using spectral and Fourier analysis. Based off the results of the 1D comparison, only RMA and CMFD were implemented in the library. The performance of RMA is compared to CMFD using the C5G7, ZPPR, and UH12 3D benchmarks. Both parallel and sequential solving schemes are considered. Analysis of the results indicates that both variants of RMA have improved effectiveness and stability relative to CMFD, for optically diffusive materials. Moreover, RMA shows great improvement in stability and effectiveness when the geometry is spatially decomposed. To achieve optimal numerical performance, a combination of RMA and CMFD is suggested. Further investigation into the use and improvement of RMA is proposed. As well, many ideas for extending the features of the library are presented
Simon, Stefan [Verfasser]. "Numerical Methods for Optimal Transport and Elastic Shape Optimization / Stefan Simon." Bonn : Universitäts- und Landesbibliothek Bonn, 2019. http://d-nb.info/1201727898/34.
Повний текст джерелаJobic, Yann. "Numerical approach by kinetic methods of transport phenomena in heterogeneous media." Thesis, Aix-Marseille, 2016. http://www.theses.fr/2016AIXM4723/document.
Повний текст джерелаA novel kinetic scheme satisfying an entropy condition is developed, tested and implemented for the simulation of practical problems. The construction of this new entropic scheme is presented. A classical hyperbolic system is approximated by a discrete velocity vector kinetic scheme (with the simplified BGK collisional operator), but applied to an inviscid compressible gas dynamics system with a small Mach number parameter, according to the approach of Carfora and Natalini (2008). The numerical viscosity is controlled, and tends to the physical viscosity of the Navier-Stokes system. The proposed numerical scheme is analyzed and formulated as an explicit finite volume flux vector splitting (FVS) scheme that is very easy to implement. It is close in spirit to Lattice Boltzmann schemes, but it has the advantage to satisfy a discrete entropy inequality under a CFL condition and a subcharacteristic stability condition involving a cell Reynolds number. The new scheme is proved to be second-order accurate in space. We show the efficiency of the method in terms of accuracy and robustness on a variety of classical benchmark tests. Some physical problems have been studied in order to show the usefulness of both schemes. The LB code was successfully used to determine the longitudinal dispersion of metallic foams, with the use of a novel indicator. The entropic code was used to determine the permeability tensor of various porous media, from the Fontainebleau sandstone (low porosity) to a redwood tree sample (high porosity). These results are pretty accurate. Finally, the entropic framework is applied to the advection-diffusion equation as a passive scalar
Mosher, Scott William. "A Variational Transport Theory Method for Two-Dimensional Reactor Core Calculations." Diss., Georgia Institute of Technology, 2004. http://hdl.handle.net/1853/5070.
Повний текст джерелаWillert, Jeffrey Alan. "Hybrid Deterministic/Monte Carlo Methods for Solving the Neutron Transport Equation and k-Eigenvalue Problem." Thesis, North Carolina State University, 2013. http://pqdtopen.proquest.com/#viewpdf?dispub=3575891.
Повний текст джерелаThe goal of this thesis is to build hybrid deterministic/Monte Carlo algorithms for solving the neutron transport equation and associated k-eigenvalue problem. We begin by introducing and deriving the transport equation before discussing a series of deterministic methods for solving the transport equation. To begin we consider moment-based acceleration techniques for both the one and two-dimensional fixed source problems. Once this machinery has been developed, we will apply similar techniques for computing the dominant eigenvalue of the neutron transport equation. We'll motivate the development of hybrid methods by describing the deficiencies of deterministic methods before describing Monte Carlo methods and their advantages. We conclude the thesis with a chapter describing the detailed implementation of hybrid methods for both the fixed-source and k-eigenvalue problem in both one and two space dimensions. We'll use a series of test problems to demonstrate the effectiveness of these algorithms before hinting at some possible areas of future work.
Bello, Kelani. "Modeling multiphase solid transport velocity in long subsea tiebacks : numerical and experimental methods." Thesis, Robert Gordon University, 2013. http://hdl.handle.net/10059/3138.
Повний текст джерелаLeroy, Thomas. "Reduced models and numerical methods for kinetic equations applied to photon transport." Thesis, Paris 6, 2016. http://www.theses.fr/2016PA066047/document.
Повний текст джерелаThe modeling of inertial confinement experiments involves kinetic equations whose discretization can become very costly. The research of reduced models allows to decrease the size and the complexity of these systems. The mathematical justification of such reduced models becomes an important issue. In this work we study several reduced models for the transfer equation in several contexts, from the theoretical and numerical point of view. In particular we study the relativistic transfer equation in the non-equilibrium diffusion regime, and we prove the convergence of the solution of this equation to the solution of a drift diffusion equation, in which the Doppler effects are modeled by a frequency transport term. This transport equation is discretized by a new class of well-balanced schemes, and we show that these schemes are consistant as the wave velocity tends to zero, by opposition to the Greenberg-Leroux type schemes. We also study several original reduced models for the Compton scattering (inelastic electron-photon collision). A hierarchy of nonlinear kinetic equations generalizing the Kompaneets equation for anisotropic distributions are derived and their properties are studied. The M_1 and P_1 angular moments models are derived from one of these equations, and we show that the anisotropic part of a radiation beam can modify the Bose condensation phenomena observed by caflisch and Levermore. This work ends with the reports of two side projects. The first one is a technical proof of the uniform convergence of the Gosse-Toscani scheme on unstructured meshes. This scheme is asymptotic preserving, since it preserves at the discrete level the diffusion limit of the hyperbolic heat equation, and this proof on unstructured meshes in 2D is original. The second one is devoted to the derivation of a kinetic model for the electron-ion Bremsstrahlung that preserves the thermal limit
Duerigen, Susan. "Neutron transport in hexagonal reactor cores modeled by trigonal-geometry diffusion and simplified P3 nodal methods." Forschungszentrum Dresden, 2013. http://nbn-resolving.de/urn:nbn:de:bsz:d120-qucosa-124665.
Повний текст джерелаTallarek, Ulrich. "Electrokinetic flow and transport in porous media: Experimental methods, numerical analysis, and applications." [S.l. : s.n.], 2004. http://deposit.ddb.de/cgi-bin/dokserv?idn=974460923.
Повний текст джерелаOlbrant, Edgar [Verfasser]. "Models and numerical methods for time- and energy-dependent particle transport / Edgar Olbrant." Aachen : Hochschulbibliothek der Rheinisch-Westfälischen Technischen Hochschule Aachen, 2012. http://d-nb.info/1023980002/34.
Повний текст джерелаWatson, Aaron Michael. "The WN adaptive method for numerical solution of particle transport problems." Texas A&M University, 2005. http://hdl.handle.net/1969.1/3133.
Повний текст джерелаDallan, Eleonora. "Numerical and experimental methods for stream and wetland modelling." Doctoral thesis, Università degli studi di Padova, 2019. http://hdl.handle.net/11577/3422714.
Повний текст джерелаHerrera, Paulo Andres Ricci. "Particle and streamline numerical methods for conservative and reactive transport simulations in porous media." Thesis, University of British Columbia, 2009. http://hdl.handle.net/2429/15967.
Повний текст джерелаBouloutas, Efthimios T. "Improved numerical methods for modeling flow and transport processes in partially saturated porous media." Thesis, Massachusetts Institute of Technology, 1989. http://hdl.handle.net/1721.1/14355.
Повний текст джерелаM.I.T. copy lacks leaf 258.
Includes bibliographical references (leaves 264-275).
by Efthimios T. Bouloutas.
Ph.D.
Pasdunkorale, Arachchige Jayantha. "Accurate finite volume methods for the numerical simulation of transport in highly anistropic media." Thesis, Queensland University of Technology, 2003.
Знайти повний текст джерелаYasseri, Saam. "Generalized spatial homogenization method in transport theory and high order diffusion theory energy recondensation methods." Diss., Georgia Institute of Technology, 2013. http://hdl.handle.net/1853/51727.
Повний текст джерелаOliveira, Anabela Pacheco de Pacheco de Oliveira Anabela De Oliveira Anabela Pacheco. "A comparison of Eulerian-Lagrangian methods for the solution of the transport equation /." Full text open access at:, 1994. http://content.ohsu.edu/u?/etd,208.
Повний текст джерелаTeaca, Bogdan. "Numerical simulations of transport processes in magnetohydrodynamic turbulence." Doctoral thesis, Universite Libre de Bruxelles, 2010. http://hdl.handle.net/2013/ULB-DIPOT:oai:dipot.ulb.ac.be:2013/210082.
Повний текст джерелаL’objectif principal de cette thèse est d’analyser le transport d’énergie inter-échelles en utilisant une simulation numérique directe d’un écoulement turbulent MHD. Les propriétés de localité du transport de l’énergie entre les échelles pour un écoulement anisotropique ou isotropique, généré par la présence d’un champ magnétique constant, sont renforcées. Un objectif secondaire est d’établir un cadre de travail pour l’étude du transport de particules test chargées dans un champ électromagnétique turbu-lent, i.e. généré par le mouvement d’un fluide conducteur, qui possède des structures à plusieurs ordres de grandeur. La structure de la thèse est présentée ci-dessous.
Dans la première partie, composée des deux premiers chapitres, l’auteur présente les notions de turbu-lences, aussi bien hydrodynamiques que MHD. Ces deux chapitres sont des synthèses.
La deuxième partie est la principale source de nouveaux résultats. Le chapitre 3 présente les méthodes numériques pour la résolution des équations, les méthodes pseudo-spectrales. Un nouveau type de force est introduit, imposant un niveau de dissipation pour tous les invariants. Dans le chapitre 4, il est effectué une analyse du transfert d'énergie entre ordres de grandeur pour les turbulences MHD. Pour explorer ces transferts d'énergie, le domaine spectral est décomposé en une série de coques de même nombre d'onde. Le transfert moyen d'énergie entre ces coques est analysé. Les transferts d'énergie s'avèrent être surtout locaux en ordre de grandeur, alors qu'une contribution non locale existe due à la force. En présence d'un champ magnétique, l'écoulement développe une direction préférentielle, une anisotropie, où une idée nouvelle de décomposition de l'espace spectral en structures annulaires est présentée. Utilisant cette décomposition annulaire on trouve que le transfert entre anneaux est local, surtout dans les anneaux de direction perpendiculaire au champ magnétique. Pour les turbulences isotropiques, dans le chapitre 5, la localité des flux d'énergie est explorée par le biais de fonctions de localité. Dans le cas de la turbulence MHD, nous avons un comportement non local plus prononcé.
La dernière partie, les chapitres 6 et 7, présente le formalisme de suivi des trajectoires de particules chargées évoluant dans un champ électromagnétique turbulent. L'influence de la méthode d'interpola-tion du solveur de particules est étudiée avant la présentation des concepts liés au transport de particu-les et aux régimes de diffusion. L'adiabatisme du mouvement des particules chargées est discuté et le transport de particules chargées dans un champ magnétique turbulent est montré en exemple.
Doctorat en sciences, Spécialisation physique
info:eu-repo/semantics/nonPublished
Morato, Rafet Sergio. "Contributions to solve the Multi-group Neutron Transport equation with different Angular Approaches." Doctoral thesis, Universitat Politècnica de València, 2021. http://hdl.handle.net/10251/159271.
Повний текст джерела[CA] La forma més exacta de conèixer el desplaçament dels neutrons a través d'un mitjà material s'aconsegueix resolent l'Equació del Transport Neutrònic. Tres diferents aproximacions d'esta equació s'han investigat en aquesta tesi: Equació del Transport Neutrònic resolta pel mètode d'Ordenades Discretes, Equació de la Difusió i Equació d'Ármonics Esfèrics Simplificats. Per a resoldre estes equacions s'estudien diferents esquemes del Mètode de Diferències Finites. La solució a estes equacions descriu la població de neutrons i les reaccions ocasionades dins d'un reactor nuclear. Al seu torn, estes variables estan relacionades amb el flux i la potència, paràmetres fonamentals per a l'Anàlisi de Seguretat Nuclear. La tesi introduïx la definició de les equacions mencionades i en particular es detallen per a l'estat estacionari. Es planteja el Mètode Modal com a solució als problemes d'autovalors definits per les dites equacions. Primer es desenvolupen diversos algoritmes per a la resolució de l'estat estacionari de l'Equació del Transport de Neutrons amb el Mètode d'Ordenades Discretes per a la discretiztació angular i el Mètode de Diferències Finites per a la discretització espacial. S'ha implementat una formulació capaç de resoldre el problema d'autovalors per a qualsevol nombre de grups energètics amb upscattering i anisotropia. Diverses quadratures utilitzades per este mètode en la seua resolució angular han sigut estudiades i implementades per a qualsevol orde d'aproximació d'Ordenades Discretes. A més, una altra formulació es desenvolupa per a la solució del problema font de l'Equació del Transport Neutrònic. A continuació, es du a terme un algoritme que permet resoldre l'Equació de la Difusió de Neutrons amb dos variants del mètode de Diferències Finites, una centrada en cel·la i una altra en vèrtex o node. S'utilitza també el Mètode Modal calculant qualsevol nombre d'autovalors per a diversos grups d'energia i amb upscattering. També s'implementen els dos esquemes del Mètode de Diferències Finites anteriorment mencionats en el desenvolupament de diferents algoritmes per a resoldre les Equacions d'Harmònics Esfèrics Simplificats. A més, s'ha realitzat una anàlisi de diferents aproximacions de les condicions de contorn. Finalment, s'han realitzat càlculs de la constant de multiplicació, els modes subcrítics, el flux neutrònic i la potència per a diferents tipus de reactors nuclears. Estes variables resulten essencials en Anàlisi de Seguretat Nuclear. A més, s'han realitzat diferents estudis de sensibilitat de paràmetres com la grandària de malla, orde utilitzat en quadratures o tipus de quadratures.
[EN] The most accurate way to know the movement of the neutrons through matter is achieved by solving the Neutron Transport Equation. Three different approaches to solve this equation have been investigated in this thesis: Discrete Ordinates Neutron Transport Equation, Neutron Diffusion Equation and Simplified Spherical Harmonics Equations. In order to solve the equations, different schemes of the Finite Differences Method were studied. The solution of these equations describes the population of neutrons and the occurred reactions inside a nuclear system. These variables are related with the flux and power, fundamental parameters for the Nuclear Safety Analysis. The thesis introduces the definition of the mentioned equations. In particular, they are detailed for the steady state case. The Modal Method is proposed as a solution to the eigenvalue problems determined by the equations. First, several algorithms for the solution of the steady state of the Neutron Transport Equation with the Discrete Ordinates Method for the angular discretization and Finite Difference Method for spatial discretization are developed. A formulation able to solve eigenvalue problems for any number of energy groups, with scattering and anisotropy has been developed. Several quadratures used by this method for the angular discretization have been studied and implemented for any order of approach of the discrete ordinates. Furthermore, an adapted formulation has been developed as a solution of the source problem for the Neutron Transport Equation. Next, an algorithm is carried out that allows to solve the Neutron Diffusion Equation with two variants of the Finite Difference Method, one with cell centered scheme and another edge entered. The Modal method is also used for calculating any number of eigenvalues for several energy groups and upscattering. Both Finite Difference schemes mentioned before are also implemented to solve the Simplified Spherical Harmonics Equations. Moreover, an analysis of different approaches of the boundary conditions is performed. Finally, calculations of the multiplication factor, subcritical modes, neutron flux and the power for different nuclear reactors were carried out. These variables result essential in Nuclear Safety Analysis. In addition, several sensitivity studies of parameters like mesh size, quadrature order or quadrature type were performed.
Me gustaría dar las gracias al Ministerio de Economía, Industria y Competitividad y a la Agencia Estatal de Investigación de España por la concesión de mi contrato predoctoral de formación de personal investigador con referencia BES-2016-076782. La ayuda económica proporcionada por este contrato fue esencial para el desarrollo de esta tesis, así como para el financiamiento de una estancia.
Morato Rafet, S. (2020). Contributions to solve the Multi-group Neutron Transport equation with different Angular Approaches [Tesis doctoral]. Universitat Politècnica de València. https://doi.org/10.4995/Thesis/10251/159271
TESIS
Owens, Alex. "Discontinuous isogeometric analysis methods for the first order form of the neutron transport equation with discrete ordinate angular discretisation." Thesis, Imperial College London, 2017. http://hdl.handle.net/10044/1/52924.
Повний текст джерелаFan, Jianhua. "Numerical study of particle transport and deposition in porous media." Thesis, Rennes, INSA, 2018. http://www.theses.fr/2018ISAR0003/document.
Повний текст джерелаThe objective of the present research was to numerically investigate the transport and deposition of particles in porous media at the pore scale. Firstly, a developed coupled lattice Boltzmann method (LBM) and discrete element method (DEM) is used to simulate the fluid-particle flow. LBM is employed to describe the fluid flow around fibers whereas DEM is used to deal with the particle dynamics. The corresponding method is two-way coupling in the sense that particle motion affects the fluid flow and reciprocally. It allowed us to predict the capture efficiency and pressure drop at the initial stage of filtration process. The quality factor is also calculated for determining the filtration performance. Secondly, we focus on the study the capture efficiency of single fiber with circular, diamond and square cross-section, respectively. The results of LBM-DEM for filtration process of single circular fiber agree well with the empirical correlation. The impaction of particles on the front side of square-shaped fiber is more favorable than those on circular and diamond cases. However, diamond fiber exhibits a good filtration performance. Then the variations of quality factor due to the different orientation angle and aspect ratio of rectangular fiber were studied using LBM-DEM. For each case, we have found the optimal value of the windward area to which corresponds a maximum value of the quality factor. The comparison of the performance of the different forms of fibers shows that the largest quality factor is obtained for square fiber oriented with angle π/4.Finally, the influence of the arrangement of fiber on filtration performance is analyzed by considering the staggered configuration. Simulations conducted for several particle size and density show that the diamond with staggered array performs better for large particles and high particle-to-fluid density ratio in terms of quality factor. The present study provide an insight to optimize the filtration process and predict filtration performance
Schramm, Marcelo. "An algorithm for multi-group two-dimensional neutron diffusion kinetics in nuclear reactor cores." reponame:Biblioteca Digital de Teses e Dissertações da UFRGS, 2016. http://hdl.handle.net/10183/142510.
Повний текст джерелаThe objective of this thesis is to introduce a new methodology for two{dimensional multi{ group neutron diffusion kinetics in a reactor core. The presented methodology uses a polyno- mial approximation in a rectangular homogeneous domain with non{homogeneous boundary conditions. As it consists on a truncated Taylor series, its error estimates varies with the size of the rectangle. The coefficients are obtained mainly by their relations with the independent term, which is determined by the differential equation. These relations are obtained by the boundary conditions only, and these relations are proven linear independent. A numerical scheme is made to assure faster convergence. The procedures done for one homogeneous rectangle are used to construct the solution of global orthogonal geometry with step{wise constant parameters steady state and time dependent problems by the iterative SOR algo- rithm. The dominant eigenvalue and its eigenfunction are obtained by the power method in the eigenvalue problem. The solution for the time dependent cases uses the modi ed Euler method in the time variable. Four classic test cases are considered for illustration.
Hou, Jingming [Verfasser]. "Robust Numerical Methods for Shallow Water Flows and Advective Transport Simulation on Unstructured Grids / Jingming Hou." Aachen : Shaker, 2013. http://d-nb.info/1051573904/34.
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