An analytical solution of the one-dimensional neutron diffusion kinetic equation in cartesian geometry
Creators
- 1. Universidade Federal do Rio Grande do Sul (UFRGS), Porto Alegre, RS (Brazil). Programa de Pos-Graduacao em Engenharia Mecanica
- 2. Comissao Nacional de Energia Nuclear (CNEN), Rio de Janeiro, RJ (Brazil)
Description
In this work we report an analytical solution for the monoenergetic neutron diffusion kinetic equation in cartesian geometry. Bearing in mind that the equation for the delayed neutron precursor concentration is a first order linear differential equation in the time variable, to make possible the application of the GITT approach to the kinetic equation, we introduce a fictitious diffusion term multiplied by a positive small value ε. By this procedure, we are able to solve this set of equations. Indeed, applying the GITT technique to the modified diffusion kinetic equation, we come out with a matrix differential equation which has a well known analytical solution when ε goes to zero. We report numerical simulations as well study of numerical convergence of the results attained. (author)
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41115752.pdf
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Additional details
Publishing Information
- Imprint Pagination
- [7 p.]
- Report number
- INIS-BR--7445
Conference
- Title
- International nuclear atlantic conference. Innovations in nuclear technology for a sustainable future; 16. Brazilian national meeting on reactor physics and thermal hydraulics; 9. Brazilian national meeting on nuclear applications; 1. Meeting on nuclear industry
- Acronym
- INAC 2009
- Dates
- 27 Sep - 2 Oct 2009
- Place
- Rio de Janeiro, RJ (Brazil)
INIS
- Country of Publication
- Brazil
- Country of Input or Organization
- Brazil
- INIS RN
- 41115752
- Subject category
- S22: GENERAL STUDIES OF NUCLEAR REACTORS;
- Resource subtype / Literary indicator
- Conference, Numerical Data
- Descriptors DEI
- ANALYTICAL SOLUTION; DELAYED NEUTRONS; EIGENVALUES; EXPERIMENTAL DATA; LAPLACE TRANSFORMATION; NEUTRON DIFFUSION EQUATION; NEUTRON FLUX; ONE-DIMENSIONAL CALCULATIONS; RESPONSE MATRIX METHOD
- Descriptors DEC
- BARYONS; CALCULATION METHODS; DATA; DIFFERENTIAL EQUATIONS; DIFFUSION EQUATIONS; ELEMENTARY PARTICLES; EQUATIONS; FERMIONS; FISSION NEUTRONS; HADRONS; INFORMATION; INTEGRAL TRANSFORMATIONS; MATHEMATICAL SOLUTIONS; NEUTRONS; NUCLEONS; NUMERICAL DATA; PARTIAL DIFFERENTIAL EQUATIONS; RADIATION FLUX; REACTOR KINETICS EQUATIONS; TRANSFORMATIONS
Optional Information
- Notes
- Published only in CD-Rom. Code: r16_91_fullpaper.pdf