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You searched for subject:(Neutron transport equation). Showing records 1 – 20 of 20 total matches.

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Universitat Politècnica de València

1. Bernal García, Álvaro. Development of a 3D Modal Neutron Code with the Finite Volume Method for the Diffusion and Discrete Ordinates Transport Equations. Application to Nuclear Safety Analyses .

Degree: 2018, Universitat Politècnica de València

 El principal objetivo de esta tesis es el desarrollo de un Método Modal para resolver dos ecuaciones: la Ecuación de la Difusión de Neutrones y… (more)

Subjects/Keywords: neutron diffusion equation; neutron transport equation; finite volume method; modal method; discrete ordinates

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APA (6th Edition):

Bernal García, . (2018). Development of a 3D Modal Neutron Code with the Finite Volume Method for the Diffusion and Discrete Ordinates Transport Equations. Application to Nuclear Safety Analyses . (Doctoral Dissertation). Universitat Politècnica de València. Retrieved from http://hdl.handle.net/10251/112422

Chicago Manual of Style (16th Edition):

Bernal García, Álvaro. “Development of a 3D Modal Neutron Code with the Finite Volume Method for the Diffusion and Discrete Ordinates Transport Equations. Application to Nuclear Safety Analyses .” 2018. Doctoral Dissertation, Universitat Politècnica de València. Accessed September 24, 2020. http://hdl.handle.net/10251/112422.

MLA Handbook (7th Edition):

Bernal García, Álvaro. “Development of a 3D Modal Neutron Code with the Finite Volume Method for the Diffusion and Discrete Ordinates Transport Equations. Application to Nuclear Safety Analyses .” 2018. Web. 24 Sep 2020.

Vancouver:

Bernal García . Development of a 3D Modal Neutron Code with the Finite Volume Method for the Diffusion and Discrete Ordinates Transport Equations. Application to Nuclear Safety Analyses . [Internet] [Doctoral dissertation]. Universitat Politècnica de València; 2018. [cited 2020 Sep 24]. Available from: http://hdl.handle.net/10251/112422.

Council of Science Editors:

Bernal García . Development of a 3D Modal Neutron Code with the Finite Volume Method for the Diffusion and Discrete Ordinates Transport Equations. Application to Nuclear Safety Analyses . [Doctoral Dissertation]. Universitat Politècnica de València; 2018. Available from: http://hdl.handle.net/10251/112422


Universitat Politècnica de València

2. Vidal Ferràndiz, Antoni. Development of a finite element method for neutron transport equation approximations .

Degree: 2018, Universitat Politècnica de València

 La ecuación del transporte neutrónico describe la población de neutrones y las reacciones nucleares dentro de un reactor nuclear. Primero, introducimos esta ecuación y las… (more)

Subjects/Keywords: Finite element method; neutron transport equation; nuclear engineering

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APA (6th Edition):

Vidal Ferràndiz, A. (2018). Development of a finite element method for neutron transport equation approximations . (Doctoral Dissertation). Universitat Politècnica de València. Retrieved from http://hdl.handle.net/10251/98522

Chicago Manual of Style (16th Edition):

Vidal Ferràndiz, Antoni. “Development of a finite element method for neutron transport equation approximations .” 2018. Doctoral Dissertation, Universitat Politècnica de València. Accessed September 24, 2020. http://hdl.handle.net/10251/98522.

MLA Handbook (7th Edition):

Vidal Ferràndiz, Antoni. “Development of a finite element method for neutron transport equation approximations .” 2018. Web. 24 Sep 2020.

Vancouver:

Vidal Ferràndiz A. Development of a finite element method for neutron transport equation approximations . [Internet] [Doctoral dissertation]. Universitat Politècnica de València; 2018. [cited 2020 Sep 24]. Available from: http://hdl.handle.net/10251/98522.

Council of Science Editors:

Vidal Ferràndiz A. Development of a finite element method for neutron transport equation approximations . [Doctoral Dissertation]. Universitat Politècnica de València; 2018. Available from: http://hdl.handle.net/10251/98522


Universitat Politècnica de València

3. Carreño Sánchez, Amanda María. Integration methods for the time dependent neutron diffusion equation and other approximations of the neutron transport equation .

Degree: 2020, Universitat Politècnica de València

 [ES] Uno de los objetivos más importantes en el análisis de la seguridad en el campo de la ingeniería nuclear es el cálculo, rápido y… (more)

Subjects/Keywords: Eigenvalue problem solvers; Neutron transport equation; Finite element method; Nuclear engineering

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APA (6th Edition):

Carreño Sánchez, A. M. (2020). Integration methods for the time dependent neutron diffusion equation and other approximations of the neutron transport equation . (Doctoral Dissertation). Universitat Politècnica de València. Retrieved from http://hdl.handle.net/10251/144771

Chicago Manual of Style (16th Edition):

Carreño Sánchez, Amanda María. “Integration methods for the time dependent neutron diffusion equation and other approximations of the neutron transport equation .” 2020. Doctoral Dissertation, Universitat Politècnica de València. Accessed September 24, 2020. http://hdl.handle.net/10251/144771.

MLA Handbook (7th Edition):

Carreño Sánchez, Amanda María. “Integration methods for the time dependent neutron diffusion equation and other approximations of the neutron transport equation .” 2020. Web. 24 Sep 2020.

Vancouver:

Carreño Sánchez AM. Integration methods for the time dependent neutron diffusion equation and other approximations of the neutron transport equation . [Internet] [Doctoral dissertation]. Universitat Politècnica de València; 2020. [cited 2020 Sep 24]. Available from: http://hdl.handle.net/10251/144771.

Council of Science Editors:

Carreño Sánchez AM. Integration methods for the time dependent neutron diffusion equation and other approximations of the neutron transport equation . [Doctoral Dissertation]. Universitat Politècnica de València; 2020. Available from: http://hdl.handle.net/10251/144771


University of New Mexico

4. Weaver, Colin A. A Forward Analytic Model of Neutron Time-of-Flight Signals with Single Elastic Scattering and Beamline Attenuation for Inferring Ion Temperatures from MagLIF Experiments.

Degree: Nuclear Engineering, 2020, University of New Mexico

  A forward analytic model is required to rapidly simulate the neutron time-of-flight (nToF) signals that result from magnetized liner inertial fusion (MagLIF) experiments at… (more)

Subjects/Keywords: neutron time-of-flight; magnetized liner inertial fusion; ion temperature; neutron transport equation; analytical model; Z Pulsed Power Facility; Nuclear Engineering

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APA (6th Edition):

Weaver, C. A. (2020). A Forward Analytic Model of Neutron Time-of-Flight Signals with Single Elastic Scattering and Beamline Attenuation for Inferring Ion Temperatures from MagLIF Experiments. (Masters Thesis). University of New Mexico. Retrieved from https://digitalrepository.unm.edu/ne_etds/95

Chicago Manual of Style (16th Edition):

Weaver, Colin A. “A Forward Analytic Model of Neutron Time-of-Flight Signals with Single Elastic Scattering and Beamline Attenuation for Inferring Ion Temperatures from MagLIF Experiments.” 2020. Masters Thesis, University of New Mexico. Accessed September 24, 2020. https://digitalrepository.unm.edu/ne_etds/95.

MLA Handbook (7th Edition):

Weaver, Colin A. “A Forward Analytic Model of Neutron Time-of-Flight Signals with Single Elastic Scattering and Beamline Attenuation for Inferring Ion Temperatures from MagLIF Experiments.” 2020. Web. 24 Sep 2020.

Vancouver:

Weaver CA. A Forward Analytic Model of Neutron Time-of-Flight Signals with Single Elastic Scattering and Beamline Attenuation for Inferring Ion Temperatures from MagLIF Experiments. [Internet] [Masters thesis]. University of New Mexico; 2020. [cited 2020 Sep 24]. Available from: https://digitalrepository.unm.edu/ne_etds/95.

Council of Science Editors:

Weaver CA. A Forward Analytic Model of Neutron Time-of-Flight Signals with Single Elastic Scattering and Beamline Attenuation for Inferring Ion Temperatures from MagLIF Experiments. [Masters Thesis]. University of New Mexico; 2020. Available from: https://digitalrepository.unm.edu/ne_etds/95


University of New Mexico

5. O'Rourke, Patrick. Modeling and Simulation of Stochastic Neutron and Cumulative Deposited Fission Energy Distributions.

Degree: Nuclear Engineering, 2020, University of New Mexico

  Methods of stochastic neutron transport are investigated and applied to novel formulations for the neutron number distribution and the cumulative fission energy deposition distribution.… (more)

Subjects/Keywords: stochastic simulation algorithm; Master equations; Monte Carlo methods; Boltzmann Master equation; neutron transport; Nuclear Engineering

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APA (6th Edition):

O'Rourke, P. (2020). Modeling and Simulation of Stochastic Neutron and Cumulative Deposited Fission Energy Distributions. (Doctoral Dissertation). University of New Mexico. Retrieved from https://digitalrepository.unm.edu/ne_etds/94

Chicago Manual of Style (16th Edition):

O'Rourke, Patrick. “Modeling and Simulation of Stochastic Neutron and Cumulative Deposited Fission Energy Distributions.” 2020. Doctoral Dissertation, University of New Mexico. Accessed September 24, 2020. https://digitalrepository.unm.edu/ne_etds/94.

MLA Handbook (7th Edition):

O'Rourke, Patrick. “Modeling and Simulation of Stochastic Neutron and Cumulative Deposited Fission Energy Distributions.” 2020. Web. 24 Sep 2020.

Vancouver:

O'Rourke P. Modeling and Simulation of Stochastic Neutron and Cumulative Deposited Fission Energy Distributions. [Internet] [Doctoral dissertation]. University of New Mexico; 2020. [cited 2020 Sep 24]. Available from: https://digitalrepository.unm.edu/ne_etds/94.

Council of Science Editors:

O'Rourke P. Modeling and Simulation of Stochastic Neutron and Cumulative Deposited Fission Energy Distributions. [Doctoral Dissertation]. University of New Mexico; 2020. Available from: https://digitalrepository.unm.edu/ne_etds/94


McMaster University

6. Stone, Terry Wayne. A Modified Spherical Harmonics Approach to Solving the Neutron Transport Equation.

Degree: MEngr, 1977, McMaster University

This is Part B.

Another approach is adopted for deriving the moments equations in spherical geometry using a spherical harmonics expansion of the neutron(more)

Subjects/Keywords: spherical; harmonics; neutron; transport; equation; geometry

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APA (6th Edition):

Stone, T. W. (1977). A Modified Spherical Harmonics Approach to Solving the Neutron Transport Equation. (Masters Thesis). McMaster University. Retrieved from http://hdl.handle.net/11375/20236

Chicago Manual of Style (16th Edition):

Stone, Terry Wayne. “A Modified Spherical Harmonics Approach to Solving the Neutron Transport Equation.” 1977. Masters Thesis, McMaster University. Accessed September 24, 2020. http://hdl.handle.net/11375/20236.

MLA Handbook (7th Edition):

Stone, Terry Wayne. “A Modified Spherical Harmonics Approach to Solving the Neutron Transport Equation.” 1977. Web. 24 Sep 2020.

Vancouver:

Stone TW. A Modified Spherical Harmonics Approach to Solving the Neutron Transport Equation. [Internet] [Masters thesis]. McMaster University; 1977. [cited 2020 Sep 24]. Available from: http://hdl.handle.net/11375/20236.

Council of Science Editors:

Stone TW. A Modified Spherical Harmonics Approach to Solving the Neutron Transport Equation. [Masters Thesis]. McMaster University; 1977. Available from: http://hdl.handle.net/11375/20236


University of Bath

7. Scheben, Fynn. Iterative methods for criticality computations in neutron transport theory.

Degree: PhD, 2011, University of Bath

 This thesis studies the so-called “criticality problem”, an important generalised eigenvalue problem arising in neutron transport theory. The smallest positive real eigenvalue of the problem… (more)

Subjects/Keywords: 518; linear Boltzmann equation; criticality; neutron transport; inverse iteration; inexact solves; iterative methods; eigenvalue problem

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APA (6th Edition):

Scheben, F. (2011). Iterative methods for criticality computations in neutron transport theory. (Doctoral Dissertation). University of Bath. Retrieved from https://researchportal.bath.ac.uk/en/studentthesis/iterative-methods-for-criticality-computations-in-neutron-transport-theory(286d7964-202f-43d2-b41c-4a3fb3732b52).html ; https://ethos.bl.uk/OrderDetails.do?uin=uk.bl.ethos.545319

Chicago Manual of Style (16th Edition):

Scheben, Fynn. “Iterative methods for criticality computations in neutron transport theory.” 2011. Doctoral Dissertation, University of Bath. Accessed September 24, 2020. https://researchportal.bath.ac.uk/en/studentthesis/iterative-methods-for-criticality-computations-in-neutron-transport-theory(286d7964-202f-43d2-b41c-4a3fb3732b52).html ; https://ethos.bl.uk/OrderDetails.do?uin=uk.bl.ethos.545319.

MLA Handbook (7th Edition):

Scheben, Fynn. “Iterative methods for criticality computations in neutron transport theory.” 2011. Web. 24 Sep 2020.

Vancouver:

Scheben F. Iterative methods for criticality computations in neutron transport theory. [Internet] [Doctoral dissertation]. University of Bath; 2011. [cited 2020 Sep 24]. Available from: https://researchportal.bath.ac.uk/en/studentthesis/iterative-methods-for-criticality-computations-in-neutron-transport-theory(286d7964-202f-43d2-b41c-4a3fb3732b52).html ; https://ethos.bl.uk/OrderDetails.do?uin=uk.bl.ethos.545319.

Council of Science Editors:

Scheben F. Iterative methods for criticality computations in neutron transport theory. [Doctoral Dissertation]. University of Bath; 2011. Available from: https://researchportal.bath.ac.uk/en/studentthesis/iterative-methods-for-criticality-computations-in-neutron-transport-theory(286d7964-202f-43d2-b41c-4a3fb3732b52).html ; https://ethos.bl.uk/OrderDetails.do?uin=uk.bl.ethos.545319

8. Ersoy, Ahmet. Nötron transport denkleminin çözümünde PL, TN, HN yöntemleri ve uygulamalar: The solution of the neutron transport equation with PL, TN, HN methods and applications.

Degree: Fen Fakültesi, 2019, University of Ankara

 Genel olarak yüksüz parçacık transport teorisi; plazma fiziğinde, atmosfer için yapılan hesaplarda, okyanus fiziğinde, ısı transferinde çok genel olarak kullanılan kapsamlı bir teoridir. Özel olarak,… (more)

Subjects/Keywords: Nötron transport denklemi; Saçılma fonksiyonu; Milne problemi; Neutron transport equation; The scattering function; The Milne problem

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APA (6th Edition):

Ersoy, A. (2019). Nötron transport denkleminin çözümünde PL, TN, HN yöntemleri ve uygulamalar: The solution of the neutron transport equation with PL, TN, HN methods and applications. (Doctoral Dissertation). University of Ankara. Retrieved from http://hdl.handle.net/20.500.12575/68601

Chicago Manual of Style (16th Edition):

Ersoy, Ahmet. “Nötron transport denkleminin çözümünde PL, TN, HN yöntemleri ve uygulamalar: The solution of the neutron transport equation with PL, TN, HN methods and applications.” 2019. Doctoral Dissertation, University of Ankara. Accessed September 24, 2020. http://hdl.handle.net/20.500.12575/68601.

MLA Handbook (7th Edition):

Ersoy, Ahmet. “Nötron transport denkleminin çözümünde PL, TN, HN yöntemleri ve uygulamalar: The solution of the neutron transport equation with PL, TN, HN methods and applications.” 2019. Web. 24 Sep 2020.

Vancouver:

Ersoy A. Nötron transport denkleminin çözümünde PL, TN, HN yöntemleri ve uygulamalar: The solution of the neutron transport equation with PL, TN, HN methods and applications. [Internet] [Doctoral dissertation]. University of Ankara; 2019. [cited 2020 Sep 24]. Available from: http://hdl.handle.net/20.500.12575/68601.

Council of Science Editors:

Ersoy A. Nötron transport denkleminin çözümünde PL, TN, HN yöntemleri ve uygulamalar: The solution of the neutron transport equation with PL, TN, HN methods and applications. [Doctoral Dissertation]. University of Ankara; 2019. Available from: http://hdl.handle.net/20.500.12575/68601


University of Michigan

9. Martin, William Russell. The Application Of The Finite Element Method To The Neutron Transport Equation.

Degree: PhD, Nuclear engineering, 1976, University of Michigan

Subjects/Keywords: Analysis; Application; Element; Equation; Finite; Method; Neutron; Reactor; Transport

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APA (6th Edition):

Martin, W. R. (1976). The Application Of The Finite Element Method To The Neutron Transport Equation. (Doctoral Dissertation). University of Michigan. Retrieved from http://hdl.handle.net/2027.42/127459

Chicago Manual of Style (16th Edition):

Martin, William Russell. “The Application Of The Finite Element Method To The Neutron Transport Equation.” 1976. Doctoral Dissertation, University of Michigan. Accessed September 24, 2020. http://hdl.handle.net/2027.42/127459.

MLA Handbook (7th Edition):

Martin, William Russell. “The Application Of The Finite Element Method To The Neutron Transport Equation.” 1976. Web. 24 Sep 2020.

Vancouver:

Martin WR. The Application Of The Finite Element Method To The Neutron Transport Equation. [Internet] [Doctoral dissertation]. University of Michigan; 1976. [cited 2020 Sep 24]. Available from: http://hdl.handle.net/2027.42/127459.

Council of Science Editors:

Martin WR. The Application Of The Finite Element Method To The Neutron Transport Equation. [Doctoral Dissertation]. University of Michigan; 1976. Available from: http://hdl.handle.net/2027.42/127459


University of Bath

10. Blake, Jack. Domain decomposition methods for nuclear reactor modelling with diffusion acceleration.

Degree: PhD, 2016, University of Bath

 In this thesis we study methods for solving the neutron transport equation (or linear Boltzmann equation). This is an integro-differential equation that describes the behaviour… (more)

Subjects/Keywords: 518; domain decomposition; neutron transport; Boltzmann equation; diffusion acceleration; iterative methods; nuclear; diffusion synthetic acceleration; numerical analysis; Convergence of numerical methods

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APA (6th Edition):

Blake, J. (2016). Domain decomposition methods for nuclear reactor modelling with diffusion acceleration. (Doctoral Dissertation). University of Bath. Retrieved from https://researchportal.bath.ac.uk/en/studentthesis/domain-decomposition-methods-for-nuclear-reactor-modelling-with-diffusion-acceleration(9ab4a25f-c342-4d18-b39a-40a25e8cf020).html ; https://ethos.bl.uk/OrderDetails.do?uin=uk.bl.ethos.698988

Chicago Manual of Style (16th Edition):

Blake, Jack. “Domain decomposition methods for nuclear reactor modelling with diffusion acceleration.” 2016. Doctoral Dissertation, University of Bath. Accessed September 24, 2020. https://researchportal.bath.ac.uk/en/studentthesis/domain-decomposition-methods-for-nuclear-reactor-modelling-with-diffusion-acceleration(9ab4a25f-c342-4d18-b39a-40a25e8cf020).html ; https://ethos.bl.uk/OrderDetails.do?uin=uk.bl.ethos.698988.

MLA Handbook (7th Edition):

Blake, Jack. “Domain decomposition methods for nuclear reactor modelling with diffusion acceleration.” 2016. Web. 24 Sep 2020.

Vancouver:

Blake J. Domain decomposition methods for nuclear reactor modelling with diffusion acceleration. [Internet] [Doctoral dissertation]. University of Bath; 2016. [cited 2020 Sep 24]. Available from: https://researchportal.bath.ac.uk/en/studentthesis/domain-decomposition-methods-for-nuclear-reactor-modelling-with-diffusion-acceleration(9ab4a25f-c342-4d18-b39a-40a25e8cf020).html ; https://ethos.bl.uk/OrderDetails.do?uin=uk.bl.ethos.698988.

Council of Science Editors:

Blake J. Domain decomposition methods for nuclear reactor modelling with diffusion acceleration. [Doctoral Dissertation]. University of Bath; 2016. Available from: https://researchportal.bath.ac.uk/en/studentthesis/domain-decomposition-methods-for-nuclear-reactor-modelling-with-diffusion-acceleration(9ab4a25f-c342-4d18-b39a-40a25e8cf020).html ; https://ethos.bl.uk/OrderDetails.do?uin=uk.bl.ethos.698988


University of New Mexico

11. Patel, Japan. Efficient Multiphysics Coupling for Fast Burst Reactors in Slab Geometry.

Degree: Nuclear Engineering, 2014, University of New Mexico

 In this thesis, we discuss a coupling algorithm to model simplified fast burst reactor dynamics. Kadioglu presented a tightly coupled multiphysics algorithm of diffusion neutronics… (more)

Subjects/Keywords: neutron transport; nonlinear diffusion acceleration; IMplicit EXplicit; Multiphysics; Newton method; HOLO; linear mechanics; temperature equation; fast burst reactor dynamics

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APA (6th Edition):

Patel, J. (2014). Efficient Multiphysics Coupling for Fast Burst Reactors in Slab Geometry. (Masters Thesis). University of New Mexico. Retrieved from http://hdl.handle.net/1928/24503

Chicago Manual of Style (16th Edition):

Patel, Japan. “Efficient Multiphysics Coupling for Fast Burst Reactors in Slab Geometry.” 2014. Masters Thesis, University of New Mexico. Accessed September 24, 2020. http://hdl.handle.net/1928/24503.

MLA Handbook (7th Edition):

Patel, Japan. “Efficient Multiphysics Coupling for Fast Burst Reactors in Slab Geometry.” 2014. Web. 24 Sep 2020.

Vancouver:

Patel J. Efficient Multiphysics Coupling for Fast Burst Reactors in Slab Geometry. [Internet] [Masters thesis]. University of New Mexico; 2014. [cited 2020 Sep 24]. Available from: http://hdl.handle.net/1928/24503.

Council of Science Editors:

Patel J. Efficient Multiphysics Coupling for Fast Burst Reactors in Slab Geometry. [Masters Thesis]. University of New Mexico; 2014. Available from: http://hdl.handle.net/1928/24503

12. Faure, Bastien. Development of neutronic calculation schemes for heterogeneous sodium-cooled nuclear cores in the Apollo3 code : application to the ASTRID prototype : Développement de schémas de calcul neutronique pour des coeurs nucléaires hétérogènes refroidis au sodium dans le code Apollo3.

Degree: Docteur es, Energie, Rayonnement, Plasma, 2019, Aix Marseille Université

Les réacteurs nucléaires refroidis au sodium offrent des perspectives intéressantes pour la filière nucléaire (utilisation optimale de l'uranium naturel, réduction de la radiotoxicité des déchets… (more)

Subjects/Keywords: Neutronique; Réacteur à neutrons rapides; Équation de transport; Schéma de calcul; Homogénéisation; Coeurs hétérogènes; Neutronics; Sodium-Cooled fast reactor; Neutron transport equation; Calculation scheme; Homogenization; Heterogeneous cores; 530

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APA (6th Edition):

Faure, B. (2019). Development of neutronic calculation schemes for heterogeneous sodium-cooled nuclear cores in the Apollo3 code : application to the ASTRID prototype : Développement de schémas de calcul neutronique pour des coeurs nucléaires hétérogènes refroidis au sodium dans le code Apollo3. (Doctoral Dissertation). Aix Marseille Université. Retrieved from http://www.theses.fr/2019AIXM0289

Chicago Manual of Style (16th Edition):

Faure, Bastien. “Development of neutronic calculation schemes for heterogeneous sodium-cooled nuclear cores in the Apollo3 code : application to the ASTRID prototype : Développement de schémas de calcul neutronique pour des coeurs nucléaires hétérogènes refroidis au sodium dans le code Apollo3.” 2019. Doctoral Dissertation, Aix Marseille Université. Accessed September 24, 2020. http://www.theses.fr/2019AIXM0289.

MLA Handbook (7th Edition):

Faure, Bastien. “Development of neutronic calculation schemes for heterogeneous sodium-cooled nuclear cores in the Apollo3 code : application to the ASTRID prototype : Développement de schémas de calcul neutronique pour des coeurs nucléaires hétérogènes refroidis au sodium dans le code Apollo3.” 2019. Web. 24 Sep 2020.

Vancouver:

Faure B. Development of neutronic calculation schemes for heterogeneous sodium-cooled nuclear cores in the Apollo3 code : application to the ASTRID prototype : Développement de schémas de calcul neutronique pour des coeurs nucléaires hétérogènes refroidis au sodium dans le code Apollo3. [Internet] [Doctoral dissertation]. Aix Marseille Université 2019. [cited 2020 Sep 24]. Available from: http://www.theses.fr/2019AIXM0289.

Council of Science Editors:

Faure B. Development of neutronic calculation schemes for heterogeneous sodium-cooled nuclear cores in the Apollo3 code : application to the ASTRID prototype : Développement de schémas de calcul neutronique pour des coeurs nucléaires hétérogènes refroidis au sodium dans le code Apollo3. [Doctoral Dissertation]. Aix Marseille Université 2019. Available from: http://www.theses.fr/2019AIXM0289

13. Graziano, Laurent. An axial polynomial expansion and acceleration of the characteristics method for the solution of the Neutron Transport Equation : Méthode accélérée aux caractéristiques pour la solution de l'équation du transport des neutrons, avec une approximation polynomiale axiale.

Degree: Docteur es, Énergie nucléaire, 2018, Université Paris-Saclay (ComUE)

 L'objectif de ce travail de thèse est le développement d'une approximation polynomiale axiale dans un solveur basé sur la Méthode des Caractéristiques. Le contexte, est… (more)

Subjects/Keywords: Méthode des caractéristiques; Équation du transport des neutrons; Approximation polynomiale; TDT; 3D MOC; Method of characteristics; Neutron transport equation; Polynomial approximation; TDT; 3D MOC

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APA (6th Edition):

Graziano, L. (2018). An axial polynomial expansion and acceleration of the characteristics method for the solution of the Neutron Transport Equation : Méthode accélérée aux caractéristiques pour la solution de l'équation du transport des neutrons, avec une approximation polynomiale axiale. (Doctoral Dissertation). Université Paris-Saclay (ComUE). Retrieved from http://www.theses.fr/2018SACLS389

Chicago Manual of Style (16th Edition):

Graziano, Laurent. “An axial polynomial expansion and acceleration of the characteristics method for the solution of the Neutron Transport Equation : Méthode accélérée aux caractéristiques pour la solution de l'équation du transport des neutrons, avec une approximation polynomiale axiale.” 2018. Doctoral Dissertation, Université Paris-Saclay (ComUE). Accessed September 24, 2020. http://www.theses.fr/2018SACLS389.

MLA Handbook (7th Edition):

Graziano, Laurent. “An axial polynomial expansion and acceleration of the characteristics method for the solution of the Neutron Transport Equation : Méthode accélérée aux caractéristiques pour la solution de l'équation du transport des neutrons, avec une approximation polynomiale axiale.” 2018. Web. 24 Sep 2020.

Vancouver:

Graziano L. An axial polynomial expansion and acceleration of the characteristics method for the solution of the Neutron Transport Equation : Méthode accélérée aux caractéristiques pour la solution de l'équation du transport des neutrons, avec une approximation polynomiale axiale. [Internet] [Doctoral dissertation]. Université Paris-Saclay (ComUE); 2018. [cited 2020 Sep 24]. Available from: http://www.theses.fr/2018SACLS389.

Council of Science Editors:

Graziano L. An axial polynomial expansion and acceleration of the characteristics method for the solution of the Neutron Transport Equation : Méthode accélérée aux caractéristiques pour la solution de l'équation du transport des neutrons, avec une approximation polynomiale axiale. [Doctoral Dissertation]. Université Paris-Saclay (ComUE); 2018. Available from: http://www.theses.fr/2018SACLS389

14. Hammer, Hans Rüdiger. Nonlinear Diffusion Acceleration in Voids for the Least-Squares Transport Equation.

Degree: PhD, Nuclear Engineering, 2017, Texas A&M University

 In this dissertation we present advances to the nonlinear diffusion acceleration for void regions using second order forms of the transport equation. We consider the… (more)

Subjects/Keywords: Neutron Transport; Nonlinear Diffusion Acceleration; Least-Squares Transport Equation

…x5D; Neutron speed w [ cm ] Weight for WLS equation WD Trial function space… …3 The transport equation… …14 2. SECOND ORDER FORMS OF THE TRANSPORT EQUATION . . . . . . . . . . . 16 2.1 2.2 2.3… …with LS and WLS high order transport equation using the local diffusion coefficient… …with LS and WLS high order transport equation using the local diffusion coefficient… 

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APA (6th Edition):

Hammer, H. R. (2017). Nonlinear Diffusion Acceleration in Voids for the Least-Squares Transport Equation. (Doctoral Dissertation). Texas A&M University. Retrieved from http://hdl.handle.net/1969.1/165996

Chicago Manual of Style (16th Edition):

Hammer, Hans Rüdiger. “Nonlinear Diffusion Acceleration in Voids for the Least-Squares Transport Equation.” 2017. Doctoral Dissertation, Texas A&M University. Accessed September 24, 2020. http://hdl.handle.net/1969.1/165996.

MLA Handbook (7th Edition):

Hammer, Hans Rüdiger. “Nonlinear Diffusion Acceleration in Voids for the Least-Squares Transport Equation.” 2017. Web. 24 Sep 2020.

Vancouver:

Hammer HR. Nonlinear Diffusion Acceleration in Voids for the Least-Squares Transport Equation. [Internet] [Doctoral dissertation]. Texas A&M University; 2017. [cited 2020 Sep 24]. Available from: http://hdl.handle.net/1969.1/165996.

Council of Science Editors:

Hammer HR. Nonlinear Diffusion Acceleration in Voids for the Least-Squares Transport Equation. [Doctoral Dissertation]. Texas A&M University; 2017. Available from: http://hdl.handle.net/1969.1/165996


Texas A&M University

15. Hrabal, Craig Anthony. 2DBTOR: a toroidal geometry neutron diffusion code.

Degree: MS, nuclear engineering, 2012, Texas A&M University

Subjects/Keywords: nuclear engineering.; Major nuclear engineering.; Nuclear fusion.; Heat equation.; Neutron transport theory.

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APA (6th Edition):

Hrabal, C. A. (2012). 2DBTOR: a toroidal geometry neutron diffusion code. (Masters Thesis). Texas A&M University. Retrieved from http://hdl.handle.net/1969.1/ETD-TAMU-1990-THESIS-H873

Chicago Manual of Style (16th Edition):

Hrabal, Craig Anthony. “2DBTOR: a toroidal geometry neutron diffusion code.” 2012. Masters Thesis, Texas A&M University. Accessed September 24, 2020. http://hdl.handle.net/1969.1/ETD-TAMU-1990-THESIS-H873.

MLA Handbook (7th Edition):

Hrabal, Craig Anthony. “2DBTOR: a toroidal geometry neutron diffusion code.” 2012. Web. 24 Sep 2020.

Vancouver:

Hrabal CA. 2DBTOR: a toroidal geometry neutron diffusion code. [Internet] [Masters thesis]. Texas A&M University; 2012. [cited 2020 Sep 24]. Available from: http://hdl.handle.net/1969.1/ETD-TAMU-1990-THESIS-H873.

Council of Science Editors:

Hrabal CA. 2DBTOR: a toroidal geometry neutron diffusion code. [Masters Thesis]. Texas A&M University; 2012. Available from: http://hdl.handle.net/1969.1/ETD-TAMU-1990-THESIS-H873


University of Michigan

16. Patton, Bruce Wayne. Application of Krylov subspace iterative techniques to the numerical solution of the neutron transport equation.

Degree: PhD, Nuclear engineering, 1996, University of Michigan

 The application of the Generalized Minimal Residual Method (GMRES) with right preconditioning is examined as an alternative to both standard and accelerated transport sweeps for… (more)

Subjects/Keywords: Application; Equati; Equation; Iterative; Krylov; Neutron; Numerical; Solution; Subspace; Techniques; Transport

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APA (6th Edition):

Patton, B. W. (1996). Application of Krylov subspace iterative techniques to the numerical solution of the neutron transport equation. (Doctoral Dissertation). University of Michigan. Retrieved from http://hdl.handle.net/2027.42/130088

Chicago Manual of Style (16th Edition):

Patton, Bruce Wayne. “Application of Krylov subspace iterative techniques to the numerical solution of the neutron transport equation.” 1996. Doctoral Dissertation, University of Michigan. Accessed September 24, 2020. http://hdl.handle.net/2027.42/130088.

MLA Handbook (7th Edition):

Patton, Bruce Wayne. “Application of Krylov subspace iterative techniques to the numerical solution of the neutron transport equation.” 1996. Web. 24 Sep 2020.

Vancouver:

Patton BW. Application of Krylov subspace iterative techniques to the numerical solution of the neutron transport equation. [Internet] [Doctoral dissertation]. University of Michigan; 1996. [cited 2020 Sep 24]. Available from: http://hdl.handle.net/2027.42/130088.

Council of Science Editors:

Patton BW. Application of Krylov subspace iterative techniques to the numerical solution of the neutron transport equation. [Doctoral Dissertation]. University of Michigan; 1996. Available from: http://hdl.handle.net/2027.42/130088


Universidade do Estado do Rio de Janeiro

17. Ralph dos Santos Mansur. Solução analítica da equação unidimensional de transporte de nêutrons monoenergéticos com espalhamento linearmente anisotrópico e aproximação sintética de difusão.

Degree: Master, 2011, Universidade do Estado do Rio de Janeiro

Nesta dissertação, são apresentados os seguintes modelos matemáticos de transporte de nêutrons: a equação linearizada de Boltzmann e a equação da difusão de nêutrons monoenergéticos… (more)

Subjects/Keywords: ENGENHARIA NUCLEAR; Diffusion; Scilab; Modelagem Computacional; Equação de Transporte de Nêutrons; Método Espectronodal; Difusão; Teoria do transporte de nêutrons; Modelos matemáticos; Spectral Nodal Method; Neutron Transport Equation; Computational Modeling; Scilab

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APA (6th Edition):

Mansur, R. d. S. (2011). Solução analítica da equação unidimensional de transporte de nêutrons monoenergéticos com espalhamento linearmente anisotrópico e aproximação sintética de difusão. (Masters Thesis). Universidade do Estado do Rio de Janeiro. Retrieved from http://www.bdtd.uerj.br/tde_busca/arquivo.php?codArquivo=5873 ;

Chicago Manual of Style (16th Edition):

Mansur, Ralph dos Santos. “Solução analítica da equação unidimensional de transporte de nêutrons monoenergéticos com espalhamento linearmente anisotrópico e aproximação sintética de difusão.” 2011. Masters Thesis, Universidade do Estado do Rio de Janeiro. Accessed September 24, 2020. http://www.bdtd.uerj.br/tde_busca/arquivo.php?codArquivo=5873 ;.

MLA Handbook (7th Edition):

Mansur, Ralph dos Santos. “Solução analítica da equação unidimensional de transporte de nêutrons monoenergéticos com espalhamento linearmente anisotrópico e aproximação sintética de difusão.” 2011. Web. 24 Sep 2020.

Vancouver:

Mansur RdS. Solução analítica da equação unidimensional de transporte de nêutrons monoenergéticos com espalhamento linearmente anisotrópico e aproximação sintética de difusão. [Internet] [Masters thesis]. Universidade do Estado do Rio de Janeiro; 2011. [cited 2020 Sep 24]. Available from: http://www.bdtd.uerj.br/tde_busca/arquivo.php?codArquivo=5873 ;.

Council of Science Editors:

Mansur RdS. Solução analítica da equação unidimensional de transporte de nêutrons monoenergéticos com espalhamento linearmente anisotrópico e aproximação sintética de difusão. [Masters Thesis]. Universidade do Estado do Rio de Janeiro; 2011. Available from: http://www.bdtd.uerj.br/tde_busca/arquivo.php?codArquivo=5873 ;

18. Koeze, D.J. (author). Goal-Oriented Angular Adaptive Algorithm using Sensitivity Analysis for the Transport Equation and Boltzmann-Fokker-Planck Equation.

Degree: 2012, Delft University of Technology

In this work we examined the discretised form of Boltzmann-like transport, i.e. the neutron trans- port equation and the Boltzmann-Fokker-Plank (BFP) equation with the discontinuous… (more)

Subjects/Keywords: Neutron Transport Equation; sensitivity analysis; adaptive algorithm; Discontinuous Galerkin method

…and absorption, with the surrounding material. The neutron transport equation, or the… …surrounding, healthy tissue. Nuclear physics, and more specifically the neutron transport equation… …reactor designs. This section contains some basic remarks on the neutron transport equation… …not take part in any important reactions. The neutron transport equation therefore considers… …linear test space for the one dimensional transport equation. . . . . . Two cases of different… 

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APA (6th Edition):

Koeze, D. J. (. (2012). Goal-Oriented Angular Adaptive Algorithm using Sensitivity Analysis for the Transport Equation and Boltzmann-Fokker-Planck Equation. (Masters Thesis). Delft University of Technology. Retrieved from http://resolver.tudelft.nl/uuid:3e96836d-39f6-46d2-a2a0-f3962cc9d803

Chicago Manual of Style (16th Edition):

Koeze, D J (author). “Goal-Oriented Angular Adaptive Algorithm using Sensitivity Analysis for the Transport Equation and Boltzmann-Fokker-Planck Equation.” 2012. Masters Thesis, Delft University of Technology. Accessed September 24, 2020. http://resolver.tudelft.nl/uuid:3e96836d-39f6-46d2-a2a0-f3962cc9d803.

MLA Handbook (7th Edition):

Koeze, D J (author). “Goal-Oriented Angular Adaptive Algorithm using Sensitivity Analysis for the Transport Equation and Boltzmann-Fokker-Planck Equation.” 2012. Web. 24 Sep 2020.

Vancouver:

Koeze DJ(. Goal-Oriented Angular Adaptive Algorithm using Sensitivity Analysis for the Transport Equation and Boltzmann-Fokker-Planck Equation. [Internet] [Masters thesis]. Delft University of Technology; 2012. [cited 2020 Sep 24]. Available from: http://resolver.tudelft.nl/uuid:3e96836d-39f6-46d2-a2a0-f3962cc9d803.

Council of Science Editors:

Koeze DJ(. Goal-Oriented Angular Adaptive Algorithm using Sensitivity Analysis for the Transport Equation and Boltzmann-Fokker-Planck Equation. [Masters Thesis]. Delft University of Technology; 2012. Available from: http://resolver.tudelft.nl/uuid:3e96836d-39f6-46d2-a2a0-f3962cc9d803

19. Odry, Nans. Méthode de décomposition de domaine avec parallélisme hybride et accélération non linéaire pour la résolution de l'équation du transport Sn en géométrie non-structurée : Domain decomposition method using a hybrid parallelism and a low-order acceleration for solving the Sn transport equation on unstructured geometry.

Degree: Docteur es, Energie, Rayonnement, Plasma, 2016, Aix Marseille Université

Les schémas de calcul déterministes permettent une modélisation à moindre coût du comportement de la population de neutrons en réacteur, mais sont traditionnellement construits sur… (more)

Subjects/Keywords: Equation du transport des neutrons; Schémas déterministes; Apollo3; Méthode de Décomposition de Domaine; Parallélisme hybride; MPI/OpenMP; Méthode d'accélération; Coarse Mesh Rebalance; Neutron transport equation; Deterministic schemes; Apollo3; Domain Decomposition Method; Hybrid parallelism; MPI/OpenMP; Acceleration technique; Coarse Mesh Rebalance; 530

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APA (6th Edition):

Odry, N. (2016). Méthode de décomposition de domaine avec parallélisme hybride et accélération non linéaire pour la résolution de l'équation du transport Sn en géométrie non-structurée : Domain decomposition method using a hybrid parallelism and a low-order acceleration for solving the Sn transport equation on unstructured geometry. (Doctoral Dissertation). Aix Marseille Université. Retrieved from http://www.theses.fr/2016AIXM4058

Chicago Manual of Style (16th Edition):

Odry, Nans. “Méthode de décomposition de domaine avec parallélisme hybride et accélération non linéaire pour la résolution de l'équation du transport Sn en géométrie non-structurée : Domain decomposition method using a hybrid parallelism and a low-order acceleration for solving the Sn transport equation on unstructured geometry.” 2016. Doctoral Dissertation, Aix Marseille Université. Accessed September 24, 2020. http://www.theses.fr/2016AIXM4058.

MLA Handbook (7th Edition):

Odry, Nans. “Méthode de décomposition de domaine avec parallélisme hybride et accélération non linéaire pour la résolution de l'équation du transport Sn en géométrie non-structurée : Domain decomposition method using a hybrid parallelism and a low-order acceleration for solving the Sn transport equation on unstructured geometry.” 2016. Web. 24 Sep 2020.

Vancouver:

Odry N. Méthode de décomposition de domaine avec parallélisme hybride et accélération non linéaire pour la résolution de l'équation du transport Sn en géométrie non-structurée : Domain decomposition method using a hybrid parallelism and a low-order acceleration for solving the Sn transport equation on unstructured geometry. [Internet] [Doctoral dissertation]. Aix Marseille Université 2016. [cited 2020 Sep 24]. Available from: http://www.theses.fr/2016AIXM4058.

Council of Science Editors:

Odry N. Méthode de décomposition de domaine avec parallélisme hybride et accélération non linéaire pour la résolution de l'équation du transport Sn en géométrie non-structurée : Domain decomposition method using a hybrid parallelism and a low-order acceleration for solving the Sn transport equation on unstructured geometry. [Doctoral Dissertation]. Aix Marseille Université 2016. Available from: http://www.theses.fr/2016AIXM4058


Université Paris-Sud – Paris XI

20. Lenain, Roland. Amélioration des méthodes de calcul de cœurs de réacteurs nucléaires dans APOLLO3 : décomposition de domaine en théorie du transport pour des géométries 2D et 3D avec une accélération non linéaire par la diffusion : Contribution to the development of methods for nuclear reactor core calculations with APOLLO3 code : domain decomposition in transport theory for 2D and 3D geometries with nonlinear diffusion acceleration.

Degree: Docteur es, Physique, 2015, Université Paris-Sud – Paris XI

Ce travail de thèse est consacré à la mise en œuvre d’une méthode de décomposition de domaine appliquée à l’équation du transport. L’objectif de ce… (more)

Subjects/Keywords: Neutronique; Équation du transport des neutrons; Méthode des caractéristiques courtes; IDT; Décomposition de domaine; Algorithme de Jacobi parallèle par bloc multigroupe; Coarse Mesh Finite Difference; Parallélisme hybride; APOLLO3; Neutronics; Neutron transport equation; Method of short characteristics; IDT; Domain decomposition method; Parallel Multigroup-Block Jacobi algorithm; Coarse Mesh Finite Difference; Hybrid parallelism; APOLLO3

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APA · Chicago · MLA · Vancouver · CSE | Export to Zotero / EndNote / Reference Manager

APA (6th Edition):

Lenain, R. (2015). Amélioration des méthodes de calcul de cœurs de réacteurs nucléaires dans APOLLO3 : décomposition de domaine en théorie du transport pour des géométries 2D et 3D avec une accélération non linéaire par la diffusion : Contribution to the development of methods for nuclear reactor core calculations with APOLLO3 code : domain decomposition in transport theory for 2D and 3D geometries with nonlinear diffusion acceleration. (Doctoral Dissertation). Université Paris-Sud – Paris XI. Retrieved from http://www.theses.fr/2015PA112180

Chicago Manual of Style (16th Edition):

Lenain, Roland. “Amélioration des méthodes de calcul de cœurs de réacteurs nucléaires dans APOLLO3 : décomposition de domaine en théorie du transport pour des géométries 2D et 3D avec une accélération non linéaire par la diffusion : Contribution to the development of methods for nuclear reactor core calculations with APOLLO3 code : domain decomposition in transport theory for 2D and 3D geometries with nonlinear diffusion acceleration.” 2015. Doctoral Dissertation, Université Paris-Sud – Paris XI. Accessed September 24, 2020. http://www.theses.fr/2015PA112180.

MLA Handbook (7th Edition):

Lenain, Roland. “Amélioration des méthodes de calcul de cœurs de réacteurs nucléaires dans APOLLO3 : décomposition de domaine en théorie du transport pour des géométries 2D et 3D avec une accélération non linéaire par la diffusion : Contribution to the development of methods for nuclear reactor core calculations with APOLLO3 code : domain decomposition in transport theory for 2D and 3D geometries with nonlinear diffusion acceleration.” 2015. Web. 24 Sep 2020.

Vancouver:

Lenain R. Amélioration des méthodes de calcul de cœurs de réacteurs nucléaires dans APOLLO3 : décomposition de domaine en théorie du transport pour des géométries 2D et 3D avec une accélération non linéaire par la diffusion : Contribution to the development of methods for nuclear reactor core calculations with APOLLO3 code : domain decomposition in transport theory for 2D and 3D geometries with nonlinear diffusion acceleration. [Internet] [Doctoral dissertation]. Université Paris-Sud – Paris XI; 2015. [cited 2020 Sep 24]. Available from: http://www.theses.fr/2015PA112180.

Council of Science Editors:

Lenain R. Amélioration des méthodes de calcul de cœurs de réacteurs nucléaires dans APOLLO3 : décomposition de domaine en théorie du transport pour des géométries 2D et 3D avec une accélération non linéaire par la diffusion : Contribution to the development of methods for nuclear reactor core calculations with APOLLO3 code : domain decomposition in transport theory for 2D and 3D geometries with nonlinear diffusion acceleration. [Doctoral Dissertation]. Université Paris-Sud – Paris XI; 2015. Available from: http://www.theses.fr/2015PA112180

.