A modified power method for the multilayer multigroup two-dimensional neutron diffusion equation

Detalhes bibliográficos
Autor(a) principal: Zanette, Rodrigo
Data de Publicação: 2018
Outros Autores: Petersen, Claudio Zen, Schramm, Marcelo, Zabadal, Jorge Rodolfo Silva
Tipo de documento: Artigo
Idioma: eng
Título da fonte: Repositório Institucional da UFRGS
Texto Completo: http://hdl.handle.net/10183/222117
Resumo: In this work we present a methodology of solution of the multigroup multi-layer stationary neutron diffusion equation in two-dimensional cartesian geometry. This eigenvalue problem describes the criticality of nuclear reactor, that is, it establishes the ratio between the numbers of neutrons generated in successive fission reactions. In order to solve this problem, we use the power method to obtain the dominant eigenvalue (Keff ) and its corresponding eigenfunction. Each iteration of the power method requires the solution of a non–homogeneous diffusion problem, that usually is solved numerically, however in this work the neutron diffusion equation is solved in analytical form in each iteration. To solve this system of second order partial differential equations, we propose to use the Finite Fourier Transform in one of the spatial variables obtaining a transformed problem which is resolved by well-established methods for ordinary differential equations. After it is solved, we use the inverse Fourier Transform to reconstruct the expression of the neutron flux in the original variables. However, at each iteration of the power method it is necessary to update the source term with the neutron flux and the Keff of the previous iteration. Thus in all iterations new terms are added which becomes the process very laborious. To overcome this problem, the authors propose a methodology that approximates the neutron flux in standard form by polynomial interpolation. In order to reduce computational time we propose to subdivide the real regions of the problem into small fictitious regions. In this way, the interpolating polynomials of each region can be of low order, reducing the dimensions of the matrices involved and, consequently, computational time. The methodology is implemented to solve a heterogeneous problem and the numerical results are compared with the finite volumes method.
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spelling Zanette, RodrigoPetersen, Claudio ZenSchramm, MarceloZabadal, Jorge Rodolfo Silva2021-06-15T04:27:32Z20180306-4549http://hdl.handle.net/10183/222117001055674In this work we present a methodology of solution of the multigroup multi-layer stationary neutron diffusion equation in two-dimensional cartesian geometry. This eigenvalue problem describes the criticality of nuclear reactor, that is, it establishes the ratio between the numbers of neutrons generated in successive fission reactions. In order to solve this problem, we use the power method to obtain the dominant eigenvalue (Keff ) and its corresponding eigenfunction. Each iteration of the power method requires the solution of a non–homogeneous diffusion problem, that usually is solved numerically, however in this work the neutron diffusion equation is solved in analytical form in each iteration. To solve this system of second order partial differential equations, we propose to use the Finite Fourier Transform in one of the spatial variables obtaining a transformed problem which is resolved by well-established methods for ordinary differential equations. After it is solved, we use the inverse Fourier Transform to reconstruct the expression of the neutron flux in the original variables. However, at each iteration of the power method it is necessary to update the source term with the neutron flux and the Keff of the previous iteration. Thus in all iterations new terms are added which becomes the process very laborious. To overcome this problem, the authors propose a methodology that approximates the neutron flux in standard form by polynomial interpolation. In order to reduce computational time we propose to subdivide the real regions of the problem into small fictitious regions. In this way, the interpolating polynomials of each region can be of low order, reducing the dimensions of the matrices involved and, consequently, computational time. The methodology is implemented to solve a heterogeneous problem and the numerical results are compared with the finite volumes method.application/pdfengAnnals of nuclear energy. Oxford. Vol. 111 (Jan. 2018), p. 136-140Difusão de nêutronsTransformada de FourierNeutron diffusionFourier transformA modified power method for the multilayer multigroup two-dimensional neutron diffusion equationEstrangeiroinfo:eu-repo/semantics/articleinfo:eu-repo/semantics/publishedVersioninfo:eu-repo/semantics/openAccessreponame:Repositório Institucional da UFRGSinstname:Universidade Federal do Rio Grande do Sul (UFRGS)instacron:UFRGSTEXT001055674.pdf.txt001055674.pdf.txtExtracted Texttext/plain23457http://www.lume.ufrgs.br/bitstream/10183/222117/2/001055674.pdf.txt4a98e43b2fc1a4918898470934de8dffMD52ORIGINAL001055674.pdfTexto completo (inglês)application/pdf926648http://www.lume.ufrgs.br/bitstream/10183/222117/1/001055674.pdf43e84d057a4bab34a671af8c908a0898MD5110183/2221172021-06-29 04:19:50.372841oai:www.lume.ufrgs.br:10183/222117Repositório de PublicaçõesPUBhttps://lume.ufrgs.br/oai/requestopendoar:2021-06-29T07:19:50Repositório Institucional da UFRGS - Universidade Federal do Rio Grande do Sul (UFRGS)false
dc.title.pt_BR.fl_str_mv A modified power method for the multilayer multigroup two-dimensional neutron diffusion equation
title A modified power method for the multilayer multigroup two-dimensional neutron diffusion equation
spellingShingle A modified power method for the multilayer multigroup two-dimensional neutron diffusion equation
Zanette, Rodrigo
Difusão de nêutrons
Transformada de Fourier
Neutron diffusion
Fourier transform
title_short A modified power method for the multilayer multigroup two-dimensional neutron diffusion equation
title_full A modified power method for the multilayer multigroup two-dimensional neutron diffusion equation
title_fullStr A modified power method for the multilayer multigroup two-dimensional neutron diffusion equation
title_full_unstemmed A modified power method for the multilayer multigroup two-dimensional neutron diffusion equation
title_sort A modified power method for the multilayer multigroup two-dimensional neutron diffusion equation
author Zanette, Rodrigo
author_facet Zanette, Rodrigo
Petersen, Claudio Zen
Schramm, Marcelo
Zabadal, Jorge Rodolfo Silva
author_role author
author2 Petersen, Claudio Zen
Schramm, Marcelo
Zabadal, Jorge Rodolfo Silva
author2_role author
author
author
dc.contributor.author.fl_str_mv Zanette, Rodrigo
Petersen, Claudio Zen
Schramm, Marcelo
Zabadal, Jorge Rodolfo Silva
dc.subject.por.fl_str_mv Difusão de nêutrons
Transformada de Fourier
topic Difusão de nêutrons
Transformada de Fourier
Neutron diffusion
Fourier transform
dc.subject.eng.fl_str_mv Neutron diffusion
Fourier transform
description In this work we present a methodology of solution of the multigroup multi-layer stationary neutron diffusion equation in two-dimensional cartesian geometry. This eigenvalue problem describes the criticality of nuclear reactor, that is, it establishes the ratio between the numbers of neutrons generated in successive fission reactions. In order to solve this problem, we use the power method to obtain the dominant eigenvalue (Keff ) and its corresponding eigenfunction. Each iteration of the power method requires the solution of a non–homogeneous diffusion problem, that usually is solved numerically, however in this work the neutron diffusion equation is solved in analytical form in each iteration. To solve this system of second order partial differential equations, we propose to use the Finite Fourier Transform in one of the spatial variables obtaining a transformed problem which is resolved by well-established methods for ordinary differential equations. After it is solved, we use the inverse Fourier Transform to reconstruct the expression of the neutron flux in the original variables. However, at each iteration of the power method it is necessary to update the source term with the neutron flux and the Keff of the previous iteration. Thus in all iterations new terms are added which becomes the process very laborious. To overcome this problem, the authors propose a methodology that approximates the neutron flux in standard form by polynomial interpolation. In order to reduce computational time we propose to subdivide the real regions of the problem into small fictitious regions. In this way, the interpolating polynomials of each region can be of low order, reducing the dimensions of the matrices involved and, consequently, computational time. The methodology is implemented to solve a heterogeneous problem and the numerical results are compared with the finite volumes method.
publishDate 2018
dc.date.issued.fl_str_mv 2018
dc.date.accessioned.fl_str_mv 2021-06-15T04:27:32Z
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dc.identifier.issn.pt_BR.fl_str_mv 0306-4549
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dc.relation.ispartof.pt_BR.fl_str_mv Annals of nuclear energy. Oxford. Vol. 111 (Jan. 2018), p. 136-140
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