Results 161 to 170 of about 24,771 (208)
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Journal of Mathematical Physics, 1972
We investigate the nature of the approximations involved in the multigroup treatment of the time-dependent neutron transport equation by using the method of approximating sequences of Banach spaces. We prove that solutions of the multigroup system converge, in a suitable sense, to the corresponding solutions of the exact transport equation.
A. Belleni-Morante, G. Busoni
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We investigate the nature of the approximations involved in the multigroup treatment of the time-dependent neutron transport equation by using the method of approximating sequences of Banach spaces. We prove that solutions of the multigroup system converge, in a suitable sense, to the corresponding solutions of the exact transport equation.
A. Belleni-Morante, G. Busoni
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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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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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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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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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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 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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Neutron Transport in a Slab with Moving Boundaries
SIAM Journal on Applied Mathematics, 1976We study a monoenergetic neutron transport initial value problem in a nonhomogeneous slab with time-dependent half-thickness. By using the theory of semigroups, we prove the existence of a unique positive solution $u = u( t )$ belonging to the Hilbert space of square summable functions.
Belleni-Morante, A., Farano, R.
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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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