Results 111 to 120 of about 12,472 (163)
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Nuclear Science and Engineering, 1975
(1975). Neutron Transport Theory. Nuclear Science and Engineering: Vol. 58, No. 3, pp. 340-340.
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(1975). Neutron Transport Theory. Nuclear Science and Engineering: Vol. 58, No. 3, pp. 340-340.
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Neutron Transport in Moving Media
SIAM Journal on Applied Mathematics, 1979An analytical and numerical investigation of the transport of thermal neutrons in a hot moving medium is presented. The analytical portion of the paper develops solutions to several idealized problems which are then used in the numerical section to validate the computer program and technique used for making calculations of realistic problems.
Wilson, H. L. +2 more
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Neutron stochastic transport theory with delayed neutrons
Annals of Nuclear Energy, 1987Abstract In this paper we have developed the stochastic transport theory with delayed neutrons. From this theory, the Boltzmann transport equation with delayed neutrons for the average flux emerges in a natural way without recourse to any approximation.
J.L. Muñoz-Cobo, G. Verdú
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Progress in Nuclear Energy, 1996
In connection with a nuclear reactor, we often need only simple functionals of the solution to the neutron transport equation. Such a functional is the power density distribution or the critical boron concentration. The question of how simple a model can be used to find simple functionals is investigated so that the model is defined as an input- output
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In connection with a nuclear reactor, we often need only simple functionals of the solution to the neutron transport equation. Such a functional is the power density distribution or the critical boron concentration. The question of how simple a model can be used to find simple functionals is investigated so that the model is defined as an input- output
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1971
The general concept of “transport theory” refers to a physical process whereby a particle migrates through a lattice composed of scattering centers. The first important part of this concept is that the migrating material can be considered to be a particle, in the case of interest a neutron which changes its speed of travel as its total energy changes ...
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The general concept of “transport theory” refers to a physical process whereby a particle migrates through a lattice composed of scattering centers. The first important part of this concept is that the migrating material can be considered to be a particle, in the case of interest a neutron which changes its speed of travel as its total energy changes ...
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Parallelization of neutron transport solvers
1997In this paper, we present the parallel techniques pertinent to the iterative process for solving multigroup transport problems using the collision probability technique. The flexible PVM message-passing environment was used to extend the capabilities of the lattice cell codes APOLLO-II and DRAGON. Speedups are reported for various configurations useful
Robert Roy, Žarko Stankovski
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Critical benchmarks in multigroup neutron transport
Annals of Nuclear Energy, 1989Abstract Space eigenfunction expansions are used within the multigroup neutron transport model in order to solve the criticality problem for homogeneous multiplying structures, without introducing any numerical space discretization. The solutions that can be produced by truncating the series are highly reliable, and can be considered, when calculated
COPPA, Gianni, LAPENTA G, RAVETTO, PIERO
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Spatial Multigrid for Isotropic Neutron Transport
SIAM Journal on Scientific Computing, 2007Summary: A spatial multigrid algorithm for isotropic neutron transport is presented in x-y geometry. The linear system is obtained using discrete ordinates in angle and corner balance finite differencing in space. Spatial smoothing is accomplished by a four-color block-Jacobi relaxation, where the diagonal blocks correspond to 4-cell blocks on the ...
Chang, B. +4 more
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Neutron transport studies for the Ignitor neutron diagnostics
Review of Scientific Instruments, 1992As a preliminary step for the project of the neutron diagnostics in the Ignitor tokamak, the overall neutron field in the device has been calculated with the help of the MCNP code, which solves, with the Monte Carlo method, neutron and photon transport problems in arbitrary three-dimensional geometries.
S. Rollet, P. Batistoni
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Neutron transport problems in anisotropic media
Annals of Nuclear Energy, 2001Abstract In nuclear reactor physics, neutron transport in media having anisotropic properties is of interest in modelling largely voided configurations, such as boiling-water and gas-cooled reactors. In order to treat such problems, neutronic methods need to be extended and adapted.
Lapenta, G. +5 more
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