Results 181 to 190 of about 11,657 (211)
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Annals of Nuclear Energy, 2020
Abstract Proper generalized decomposition (PGD) can be described as a numerical extension of separation of variables and thus be employed as a solution technique for multi-dimensional problems, where dimensions should be understood in the phase-space context of the governing law at hand.
Zachary M. Prince, Jean C. Ragusa
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Abstract Proper generalized decomposition (PGD) can be described as a numerical extension of separation of variables and thus be employed as a solution technique for multi-dimensional problems, where dimensions should be understood in the phase-space context of the governing law at hand.
Zachary M. Prince, Jean C. Ragusa
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Neutron-photon multigroup cross sections for neutron energies ⩽ 400 MeV (revision 1)
Nuclear Instruments and Methods in Physics Research Section A: Accelerators, Spectrometers, Detectors and Associated Equipment, 1986Abstract Multigroup cross sections (66 neutron groups and 22 photon groups) are described for neutron energies from thermal to 400 MeV. The elements considered are hydrogen, 10B, 11B, carbon, nitrogen, oxygen, sodium, magnesium, aluminum, silicon, sulfur, potassium, calcium, chromium, iron, nickel, tungsten, and lead. The cross section data presented
R.G. Alsmiller +2 more
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Discrete energy or multigroup finite element transport calculations in lattice physics
Progress in Nuclear Energy, 1991Abstract Finite element transport calculations are well suited for reactor lattice analysis, and in particular for benchmark problems involving tight unit cells. They model the cell accurately in two dimensions. Exact symmetry boundary conditions for the flux moments of any truncated spherical harmonics expansion of the angular flux can be imposed at
W. Rothenstein, C.R.E. de Oliveira
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An Adaptive Energy Mesh Constructor for Multigroup Library Generation for Transport Codes
Nuclear Science and Engineering, 2011Users’ demands for multigroup transport calculations are wide and diverse, encompassing routine, rough, and fast calculations as well as very precise simulations.
Pietro Mosca +3 more
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Volume 3: Nuclear Fuel and Material, Reactor Physics, and Transport Theory, 2018
The adjoint neutron flux is vital in the analysis of reactor kinetics parameters and reactor transient events. Both deterministic and Monte Carlo methods have been developed for the adjoint neutron flux calculation on the basis of multi-group cross sections which may vary significantly among different types of reactors. The iterated fission probability
Xiaotong Shang, Guanlin Shi, Kan Wang
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The adjoint neutron flux is vital in the analysis of reactor kinetics parameters and reactor transient events. Both deterministic and Monte Carlo methods have been developed for the adjoint neutron flux calculation on the basis of multi-group cross sections which may vary significantly among different types of reactors. The iterated fission probability
Xiaotong Shang, Guanlin Shi, Kan Wang
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physica status solidi (b), 1993
AbstractEnergy deposition calculations are performed due to an incident source of electrons in rectangular slab geometry using the method of discrete ordinates for directions and the multigroup treatment in energy. Using the restricted stopping power formalism to distinguish between large energy loss collisions (catastrophic) and small energy loss ...
R. P. Datta, A. K. Ray
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AbstractEnergy deposition calculations are performed due to an incident source of electrons in rectangular slab geometry using the method of discrete ordinates for directions and the multigroup treatment in energy. Using the restricted stopping power formalism to distinguish between large energy loss collisions (catastrophic) and small energy loss ...
R. P. Datta, A. K. Ray
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Canadian Journal of Physics, 2000
A low-energy neutron transport algorithm for use in space-radiation protection is developed. The algorithm is based upon a multiple energy group analysis of the straight ahead Boltzmann equation utilizing a mean value theorem for integrals. The algorithm developed is then verified by using a collocation method solution on the same straight ahead ...
M S, Clowdsley +5 more
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A low-energy neutron transport algorithm for use in space-radiation protection is developed. The algorithm is based upon a multiple energy group analysis of the straight ahead Boltzmann equation utilizing a mean value theorem for integrals. The algorithm developed is then verified by using a collocation method solution on the same straight ahead ...
M S, Clowdsley +5 more
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Nuclear Science and Engineering, 1996
A hybrid multigroup/continuous-energy Monte Carlo algorithm is developed for solving the Boltzmann-Fokker-Planck equation. This algorithm differs significantly from previous charged-particle Monte Carlo algorithms. Most importantly, it can be used to perform both forward and adjoint transport calculations, using the same basic multigroup cross-section ...
J. E. Morel +4 more
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A hybrid multigroup/continuous-energy Monte Carlo algorithm is developed for solving the Boltzmann-Fokker-Planck equation. This algorithm differs significantly from previous charged-particle Monte Carlo algorithms. Most importantly, it can be used to perform both forward and adjoint transport calculations, using the same basic multigroup cross-section ...
J. E. Morel +4 more
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Nuclear Engineering and Design, 2014
Abstract The High Temperature Engineering Test Reactor (HTTR) in Japan is a helium-cooled graphite-moderated reactor designed and operated for the future development of high-temperature gas-cooled reactors. Two detailed full-core models of HTTR have been established by using SCALE6 and MCNP5/X, respectively, to study its neutronic properties. Several
Min-Han Chiang +3 more
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Abstract The High Temperature Engineering Test Reactor (HTTR) in Japan is a helium-cooled graphite-moderated reactor designed and operated for the future development of high-temperature gas-cooled reactors. Two detailed full-core models of HTTR have been established by using SCALE6 and MCNP5/X, respectively, to study its neutronic properties. Several
Min-Han Chiang +3 more
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Annals of Nuclear Energy, 1980
Abstract The time-dependent probability distribution of neutrons in a space-independent, low-power, multiplying assembly with a source is developed in the multigroup energy approximation as forward and backward Kolmogorov equations. The relationship between these as adjoint equations is made explicit in a tensor notation and the equations developed ...
U. Salmi, J. Lewins
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Abstract The time-dependent probability distribution of neutrons in a space-independent, low-power, multiplying assembly with a source is developed in the multigroup energy approximation as forward and backward Kolmogorov equations. The relationship between these as adjoint equations is made explicit in a tensor notation and the equations developed ...
U. Salmi, J. Lewins
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