Integral formulation and numerical simulations for the neutron transport equation in X-Y geometry
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DOI: 10.1016/j.anucene.2017.10.017
Abstract: Abstract In this work the Nystrom method was used to solve the integral formulation of the neutron transport equation in X - Y geometry. Four quadrature schemes were tested, namely, Gauss–Legendre, Gauss–Lobatto, Simpson, and Boole… read more here.
Keywords: equation geometry; transport equation; neutron transport; geometry ... See more keywords
Neutron transport in anisotropic random media
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DOI: 10.1016/j.anucene.2018.04.031
Abstract: Abstract Assessing the impact of random media for eigenvalue problems plays a central role in nuclear reactor physics and criticality safety. In a recent work (Larmier et al., 2018a), we have applied a probabilistic model… read more here.
Keywords: anisotropic random; fuel; transport anisotropic; random media ... See more keywords
DIAMOND – A Method of Characteristics neutron transport code using unstructured meshing
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DOI: 10.1016/j.anucene.2019.107086
Abstract: Abstract Advanced nuclear reactor designs using more complicated and heterogeneous geometries warrant precise modeling of the neutron transport phenomenon. The use of Method of Characteristics (MOC) with unstructured meshing renders an accurate representation for tracking… read more here.
Keywords: transport; diamond; unstructured meshing; method characteristics ... See more keywords
A Rayleigh quotient method for criticality eigenvalue problems in neutron transport
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DOI: 10.1016/j.anucene.2019.107120
Abstract: Abstract The alpha- and k-effective eigenproblems describe the criticality and fundamental neutron flux mode of a nuclear system. Traditionally, the alpha-eigenvalue problem has been solved using methods that focus on supercritical systems with large, positive… read more here.
Keywords: quotient method; eigenvalue problems; criticality; neutron transport ... See more keywords
Convergence analysis of the non-linear CMFD schemes for acceleration of multi-group fixed source neutron transport calculations
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DOI: 10.1016/j.anucene.2020.108041
Abstract: Abstract The coarse mesh finite difference (CMFD) and related methods have been some of the most extensively used schemes to accelerate the convergence of neutron transport iterations. Despite its predominant application on multi-group problems, the… read more here.
Keywords: multi group; group; cmfd; neutron transport ... See more keywords
An extended half-range spherical harmonics method for first-order neutron transport equation based on variational treatment
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DOI: 10.1016/j.pnucene.2017.07.010
Abstract: Abstract A new variational approach with anisotropic scattering kernel for first order neutron transport equation based on Finite Element Method (FEM) and Double- P N ( DP N ) approximation has been introduced. In presented… read more here.
Keywords: neutron; extended half; transport equation; geometry ... See more keywords
The spectral nodal method applied to multigroup SN neutron transport problems in One-Dimensional geometry with Fixed–Source
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DOI: 10.1016/j.pnucene.2017.12.017
Abstract: Abstract The nodal methods, as deterministic models, form a class of numerical methods developed to generate accurate numerical solutions of the Boltzmann equation for neutron transport. These methods are algebraically and computationally more laborious than… read more here.
Keywords: one dimensional; method; geometry; spectral nodal ... See more keywords
A Coarse-Mesh Diffusion Synthetic Acceleration Method with Local hp Adaptation for Neutron Transport Calculations
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DOI: 10.1080/00295639.2018.1499338
Abstract: Abstract We present a local adaptive diffusion synthetic acceleration (DSA) method for neutron transport calculations. This new DSA method, called DG-DSA, solves the diffusion equation on a coarse mesh using the interior penalty discontinuous Galerkin… read more here.
Keywords: diffusion synthetic; dsa method; transport; method ... See more keywords
Application of Krylov Acceleration Technique in Method of Characteristics–Based Neutron Transport Code
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DOI: 10.1080/00295639.2018.1499340
Abstract: Abstract In our earlier work, a computer code based on Method of Characteristics (MOC) was developed to solve the neutron transport equation for mainly assembly-level lattice calculations with reflective and periodic boundary conditions and to… read more here.
Keywords: acceleration technique; method characteristics; neutron transport; code ... See more keywords
Multigroup Constant Calculation with Static α-Eigenvalue Monte Carlo for Time-Dependent Neutron Transport Simulations
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DOI: 10.1080/00295639.2020.1743578
Abstract: Multigroup constants for deterministic methods that preserve the time-dependent physics of the neutron transport equations are derived. Alternative multigroup constant weighting spectra are discuss... read more here.
Keywords: constant calculation; multigroup constant; time dependent; neutron transport ... See more keywords
A Linear Prolongating Coarse Mesh Finite Difference Acceleration of Discrete Ordinate Neutron Transport Calculation Based on Discontinuous Galerkin Finite Element Method
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DOI: 10.1080/00295639.2020.1752045
Abstract: Abstract The recently developed linear prolongation Coarse Mesh Finite Difference (lpCMFD) acceleration scheme, which employs a linear additive approach to update the scalar flux, has been shown to be more stable and effective than the… read more here.
Keywords: transport calculation; calculation; finite difference; coarse mesh ... See more keywords