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Статті в журналах з теми "Neutron transport problem"
Jarmouni-Idrissi, K., and L. Thevenot. "HOMOGENIZATION OF A NONLINEAR NEUTRON TRANSPORT PROBLEM." Transport Theory and Statistical Physics 31, no. 2 (May 21, 2002): 93–123. http://dx.doi.org/10.1081/tt-120003969.
Повний текст джерелаVosoughi, Naser, Akbar Salehi, Majid Shahriari, and Enzo Tonti. "Direct discrete method and its application to neutron transport problems." Nuclear Technology and Radiation Protection 18, no. 2 (2003): 12–23. http://dx.doi.org/10.2298/ntrp0302012v.
Повний текст джерелаTÜRECİ, R. Gökhan. "Machine Learning Applications to the One-speed Neutron Transport Problems." Cumhuriyet Science Journal 43, no. 4 (December 27, 2022): 726–38. http://dx.doi.org/10.17776/csj.1163514.
Повний текст джерелаTsyfra, Ivan, and Tomasz Czyżycki. "Symmetry and Solution of Neutron Transport Equations in Nonhomogeneous Media." Abstract and Applied Analysis 2014 (2014): 1–9. http://dx.doi.org/10.1155/2014/724238.
Повний текст джерелаSengupta, A. "Full range solution of half space neutron transport problem." ZAMP Zeitschrift f�r angewandte Mathematik und Physik 46, no. 1 (January 1995): 40–60. http://dx.doi.org/10.1007/bf00952255.
Повний текст джерелаKadem, Abdelouahab. "Analytical solutions for the neutron transport using the spectral methods." International Journal of Mathematics and Mathematical Sciences 2006 (2006): 1–11. http://dx.doi.org/10.1155/ijmms/2006/16214.
Повний текст джерелаBourhrara, Lahbib, and Richard Sanchez. "Existence Result for the Kinetic Neutron Transport Problem in the Presence of Delayed Neutrons." Transport Theory and Statistical Physics 35, no. 3-4 (August 2006): 137–56. http://dx.doi.org/10.1080/00411450600901748.
Повний текст джерелаOzturk, Hakan. "The influence of linear anisotropic scattering of one-speed neutrons on the critical size of a slab with reflective boundary conditions." Nuclear Technology and Radiation Protection 32, no. 3 (2017): 236–41. http://dx.doi.org/10.2298/ntrp1703236o.
Повний текст джерелаMancusi, Davide, and Andrea Zoia. "TOWARDS ZERO-VARIANCE SCHEMES FOR KINETIC MONTE-CARLO SIMULATIONS." EPJ Web of Conferences 247 (2021): 04010. http://dx.doi.org/10.1051/epjconf/202124704010.
Повний текст джерелаChen, Gen-Shun, and Anthony W. Leung. "Positive Solutions for Reactor Multigroup Neutron Transport Systems: Criticality Problem." SIAM Journal on Applied Mathematics 49, no. 3 (June 1989): 871–87. http://dx.doi.org/10.1137/0149051.
Повний текст джерелаДисертації з теми "Neutron transport problem"
Scipolo, Vittorio. "Scattered neutron tomography based on a neutron transport problem." Texas A&M University, 2004. http://hdl.handle.net/1969.1/2791.
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 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
Willert, 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.
Picoloto, Camila Becker. "Formulações espectronodais em cálculos neutrônicos multidimensionais." reponame:Biblioteca Digital de Teses e Dissertações da UFRGS, 2015. http://hdl.handle.net/10183/118888.
Повний текст джерелаNeste trabalho, uma abordagem analítica é utilizada juntamente com esquemas nodais na resolução de problemas bidimensionais de transporte de nêutrons de fonte fixa, em geometria cartesiana, definidos em meio heterogêneo, com espalhamento anisotrópico. A metodologia proposta é desenvolvida a partir da versão em ordenadas discretas da equação de transporte bidimensional, juntamente com o esquema de quadratura simétrica de nível. As equações em ordenadas discretas são integradas transversalmente, originando equações unidimensionais para os fluxos angulares médios. Tais equações unidimensionais são resolvidas pelo método ADO (Analytical Discrete Ordinates). Expressões explícitas nas variáveis espaciais são derivadas para os fluxos angulares médios em cada região em que o domínio foi subdividido. A solução em cada região é acoplada às regiões vizinhas, para fornecer a solução no domínio todo, sem a utilização de métodos iterativos. Como usual em esquemas nodais, equações auxiliares são necessárias, recebendo neste estudo dois tratamentos distintos: um em que os fluxos desconhecidos nos contornos das regiões assumem relações de proporcionalidade, com os fluxos angulares médios; e, outro, em que esses fluxos são aproximados por polinômios de ordem zero sendo, nesse caso, incorporados ao termo fonte. Resultados numéricos obtidos e comparados com disponíveis na literatura mostram a viabilidade da formulação, mantendo a eficiência computacional já verificada no tratamento de outros problemas, com o uso do método ADO.
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 джерелаPounders, Justin Michael. "A coarse-mesh transport method for time-dependent reactor problems." Diss., Georgia Institute of Technology, 2010. http://hdl.handle.net/1853/39586.
Повний текст джерелаByambaakhuu, 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 джерелаMosher, Scott William. "Implementation of an adaptive importance sampling technique in MCNP for monoenergetic slab problems." Thesis, Georgia Institute of Technology, 2001. http://hdl.handle.net/1853/17100.
Повний текст джерелаMilitão, Damiano da Silva. "Um modelo para a reconstrução angular e espacial analítica do problema de transporte unidimensional de partículas neutras usando um método espectro-nodal." Universidade do Estado do Rio de Janeiro, 2007. http://www.bdtd.uerj.br/tde_busca/arquivo.php?codArquivo=416.
Повний текст джерелаNesta dissertação propomos a utilização do método espectro-nodal SGF, cf. spectral Greens function, para transporte SN de partículas neutras, para determinarmos os fluxos angulares nas interfaces das regiões homogêneas do domínio espacial heterogêneo, com espalhamento linearmente anisotrópico usando preferencialmente altas ordens de quadraturas angulares. As reconstruções espaciais analíticas dos fluxos angulares são feitas no interior das regiões homogêneas, determinando as constantes arbitrárias da solução analítica local das equações SN no interior dos nodos espaciais da grade de dicretização. A seguir, utilizando essas constantes, determinamos as expressões do fluxo escalar e da corrente de nêutrons, que são substituídas na equação de transporte unidimensional em geometria retangular Cartesiana no termo de fonte por espalhamento linearmente anisotrópico. Resolvemos analiticamente a equação de transporte com os termos do fluxo escalar e corrente de nêutrons assim aproximados para estimarmos o perfil do fluxo angular de nêutrons no domínio. Esta reconstrução analítica aproximada da solução da equação de transporte de partículas neutras em geometria unidimensional Cartesiana constitui um problema inverso, na medida em que a partir da solução nodal de malha grossa fazemos primeiramente uma reconstrução analítica espacial do fluxo angular nas direções das ordenadas discretas, para em seguida procedermos à reconstrução analítica aproximada do fluxo no domínio angular.
We describe the application of the spectral Greens function SN nodal method for one-speed neutral particle transport calculations to determine the angular fluxes at the homogeneized regions within heterogeneous domains, for linearly anisotropic scattering, using preferably high-order angular quadratures. The reconstruction scheme in the space variable of the angular flux is carried out within the homogenized regions using uniform spatial grid. We determine the arbitrary constants of the analytical SN general solution inside each spatial node. Then, we determine the SN expression for the scalar flux and total current that we substitute into the analytical slab-geometry transport equation, precisely into its linearly anisotropic scattering source term. Further, we solve analytically the slab-geometry transport equation, so approximated, to obtain the flux profile within the space and angular domains. This approximate analytical reconstruction scheme of the solution of the neutral particle transport equation in slab geometry is an inverse problem, in the sense that we use accurate coarse-mesh SN numerical solution, to recover the SN analytical solution in the space variable, and then reconstruct the solution approximately in the angular domain.
Книги з теми "Neutron transport problem"
Gupta, Anurag. Krylov sub-space methods for K-eigenvalue problem in 3-D multigroup neutron transport. Mumbai: Bhabha Atomic Research Centre, 2004.
Знайти повний текст джерелаKyncl, Jan. On the problem of criticality for neutron transport equation. Řež, Czech Republic: Nuclear Research Institite Řež plc, 2003.
Знайти повний текст джерелаModak, R. S. Transport synthetic acceleration scheme for multi-dimensional neutron transport problems. Mumbai: Bhabha Atomic Research Centre, 2005.
Знайти повний текст джерелаMonte Carlo Principles and Neutron Transport Problems. Dover Publications, 2008.
Знайти повний текст джерелаKeller, Herbert. Approximate Solutions of Steady-State Neutron Transport Problems for Slabs. Creative Media Partners, LLC, 2015.
Знайти повний текст джерелаЧастини книг з теми "Neutron transport problem"
Pignedoli, Antonio. "On the Rigorous Analysis of the Problem of the Neutron Transport in a Slab Geometry And on Some Other Results." In Some Aspects of Diffusion Theory, 519–38. Berlin, Heidelberg: Springer Berlin Heidelberg, 2011. http://dx.doi.org/10.1007/978-3-642-11051-1_8.
Повний текст джерелаPrillinger, G., and M. Mattes. "The Importance of Anisotropic Scattering in High Energy Neutron Transport Problems." In Reactor Dosimetry, 287–93. Dordrecht: Springer Netherlands, 1985. http://dx.doi.org/10.1007/978-94-009-5378-9_28.
Повний текст джерелаMori, T., K. Okumura, and Y. Nagaya. "Status of JAERI’s Monte Carlo Code MVP for Neutron and Photon Transport Problems." In Advanced Monte Carlo for Radiation Physics, Particle Transport Simulation and Applications, 625–30. Berlin, Heidelberg: Springer Berlin Heidelberg, 2001. http://dx.doi.org/10.1007/978-3-642-18211-2_100.
Повний текст джерелаPignedoli, Antonio. "Transformational Methods Applied To Some One-Dimensional Problems Concerning The Equations of The Neutron Transport Theory." In Some Aspects of Diffusion Theory, 503–18. Berlin, Heidelberg: Springer Berlin Heidelberg, 2011. http://dx.doi.org/10.1007/978-3-642-11051-1_7.
Повний текст джерелаMokhtar-Kharroubi, M. "Stochastic formulations of neutron transport: Nonlinear problems." In Series on Advances in Mathematics for Applied Sciences, 215–44. WORLD SCIENTIFIC, 1997. http://dx.doi.org/10.1142/9789812819833_0010.
Повний текст джерелаJordinson, Chris. "3He transport and the solar neutrino problem." In Stellar Astrophysical Fluid Dynamics, 193–204. Cambridge University Press, 2003. http://dx.doi.org/10.1017/cbo9780511536335.014.
Повний текст джерелаТези доповідей конференцій з теми "Neutron transport problem"
Hao, Jianli, Wenzhen Chen, Shaoming Wang, and De Zhang. "Study of the Space-Time Neutron Multiplication Formula." In 18th International Conference on Nuclear Engineering. ASMEDC, 2010. http://dx.doi.org/10.1115/icone18-29279.
Повний текст джерелаWu, Zeyun, and Marvin L. Adams. "Advances in Inverse Transport Methods." In 18th International Conference on Nuclear Engineering. ASMEDC, 2010. http://dx.doi.org/10.1115/icone18-29881.
Повний текст джерелаVosoughi, Naser, Majid Shahriari, and Ali Akbar Salehi. "Direct Discrete Method for Neutronic Calculations." In 10th International Conference on Nuclear Engineering. ASMEDC, 2002. http://dx.doi.org/10.1115/icone10-22014.
Повний текст джерелаGairola, A., Hitesh Bindra, Gaurav Agarwal, and Suneet Singh. "Lattice Boltzmann Method for Solving Time-Dependent Radiation Transport and Reactor Criticality Problems." In 2016 24th International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2016. http://dx.doi.org/10.1115/icone24-60058.
Повний текст джерелаZhang, J. "A coupled thermo-mechanical and neutron diffusion numerical model for irradiated concrete." In AIMETA 2022. Materials Research Forum LLC, 2023. http://dx.doi.org/10.21741/9781644902431-4.
Повний текст джерелаKoreshi, Zafar Ullah. "Stationarity Issues in Monte Carlo Simulation for Neutron Transport." In 2013 21st International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2013. http://dx.doi.org/10.1115/icone21-15016.
Повний текст джерелаWu, Hongchun, Guoming Liu, Liangzhi Cao, and Qichang Chen. "Determinant Methods for Solving Neutron Transport Equation in Unstructured Geometry." In 18th International Conference on Nuclear Engineering. ASMEDC, 2010. http://dx.doi.org/10.1115/icone18-29442.
Повний текст джерелаYang, Wankui, Baoxin Yuan, Songbao Zhang, Haibing Guo, Yaoguang Liu, and Li Deng. "A Neutron Transport Calculation Method for Deep Penetration and its Preliminary Verification." In 2018 26th International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2018. http://dx.doi.org/10.1115/icone26-81709.
Повний текст джерелаCarreño, Amanda, Antoni Vidal Ferrándiz, Damián Ginestar Peiró, and Gumersindo Verdú. "Block strategies to compute the lambda modes associated with the neutron diffusion equation." In VI ECCOMAS Young Investigators Conference. València: Editorial Universitat Politècnica de València, 2021. http://dx.doi.org/10.4995/yic2021.2021.13470.
Повний текст джерелаXiao, Wei, Tengfei Zhang, Xiaojing Liu, Donghao He, and Qingquan Pan. "Application of Stiffness Confinement Method on Pin Resolved Variational Nodal Method and Its Implementation to C5G7-TD Benchmark Problem." In 2022 29th International Conference on Nuclear Engineering. American Society of Mechanical Engineers, 2022. http://dx.doi.org/10.1115/icone29-92399.
Повний текст джерелаЗвіти організацій з теми "Neutron transport problem"
William Charlton. Scattered Neutron Tomography Based on A Neutron Transport Inverse Problem. Office of Scientific and Technical Information (OSTI), July 2007. http://dx.doi.org/10.2172/915225.
Повний текст джерелаOndis, L. A. ,. II, L. J. Tyburski, and B. S. Moskowitz. RCPO1 - A Monte Carlo program for solving neutron and photon transport problems in three dimensional geometry with detailed energy description and depletion capability. Office of Scientific and Technical Information (OSTI), March 2000. http://dx.doi.org/10.2172/755403.
Повний текст джерелаChen, Yona, Jeffrey Buyer, and Yitzhak Hadar. Microbial Activity in the Rhizosphere in Relation to the Iron Nutrition of Plants. United States Department of Agriculture, October 1993. http://dx.doi.org/10.32747/1993.7613020.bard.
Повний текст джерела