Solution of the one-group time-dependent neutron transport equation in an infinite medium by polynomial reconstruction
Journal Article
·
· Nucl. Sci. Eng.; (United States)
OSTI ID:6073249
The numerical solution to the one-group time-dependent neutron transport equation in infinite plane, spherical, and cylindrical geometries is obtained via an expansion in Legendre polynomials. The computation features general anisotropic scattering, isotropic and beam sources and a power law time-dependent cross-section variation. Results for test problems are compared with previously obtained numerical solutions and with the diffusion approximation.
- Research Organization:
- University of Arizona, Department of Nuclear and Energy Engineering, Tucson, AZ 85721
- OSTI ID:
- 6073249
- Journal Information:
- Nucl. Sci. Eng.; (United States), Journal Name: Nucl. Sci. Eng.; (United States) Vol. 92:2; ISSN NSENA
- Country of Publication:
- United States
- Language:
- English
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Related Subjects
22 GENERAL STUDIES OF NUCLEAR REACTORS
220100* -- Nuclear Reactor Technology-- Theory & Calculation
ANISOTROPY
CROSS SECTIONS
DATA ANALYSIS
DIFFERENTIAL EQUATIONS
EQUATIONS
FUNCTIONS
GEOMETRY
ISOTROPY
LEGENDRE POLYNOMIALS
MATHEMATICS
NEUTRON DIFFUSION EQUATION
NEUTRON TRANSPORT THEORY
NUMERICAL SOLUTION
POLYNOMIALS
SCATTERING
TIME DEPENDENCE
TRANSPORT THEORY
220100* -- Nuclear Reactor Technology-- Theory & Calculation
ANISOTROPY
CROSS SECTIONS
DATA ANALYSIS
DIFFERENTIAL EQUATIONS
EQUATIONS
FUNCTIONS
GEOMETRY
ISOTROPY
LEGENDRE POLYNOMIALS
MATHEMATICS
NEUTRON DIFFUSION EQUATION
NEUTRON TRANSPORT THEORY
NUMERICAL SOLUTION
POLYNOMIALS
SCATTERING
TIME DEPENDENCE
TRANSPORT THEORY