Interface-flux nodal transport method

Conference · Sun Jan 01 04:00:00 EST 1989 · Transactions of the American Nuclear Society; (USA)

OSTI ID:6504886

Kohut, P ^[1]

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1. Brookhaven National Lab., Upton, NY (USA)

The need for the solution of the neutron transport equation in large geometries was the basis of the early development of nodal transport methods using primarily the integral formulation. Most of these early techniques were based on the interface-current method and were extremely successful in solving complex geometries on the fuel assembly level, generally employing flat or linear spatial expansions of the node interior fluxes and surface currents. In this paper, the development of an interface-flux nodal method is presented. The method offers geometric flexibility and as such is not confined to any particular geometry. It is capable of high-order spatial expansion of the node-interior fluxes and the node-surface quantities and, by using an explicit expansion of the scalar fluxes, the iteration on the scattering source is eliminated. This method is based on the surface-integral formulation of the transport problem and employs an approach of a point-to-point transport on each nodal surface. The resulting response-matrix-like equations express the outgoing flux components in terms of the incoming values and iteration techniques that are used to get the final solution. Numerical results of a number of benchmark problems are presented to illustrate the accuracy of the method.

• OSTI ID: 6504886
• Report Number(s): CONF-891103--
• Conference Information: Journal Name: Transactions of the American Nuclear Society; (USA) Journal Volume: 60
• Country of Publication: United States
• Language: English

Related Subjects

654003* -- Radiation & Shielding Physics-- Neutron Interactions with Matter
73 NUCLEAR PHYSICS AND RADIATION PHYSICS
ACCURACY
ALGORITHMS
BENCHMARKS
COMPUTER CODES
DIFFERENTIAL EQUATIONS
DISCRETE ORDINATE METHOD
EIGENVALUES
EQUATIONS
GEOMETRY
INTEGRAL EQUATIONS
ITERATIVE METHODS
MATHEMATICAL LOGIC
MATHEMATICS
MESH GENERATION
NEUTRON DIFFUSION EQUATION
NEUTRON FLUX
NEUTRON TRANSPORT THEORY
ONE-DIMENSIONAL CALCULATIONS
QUADRATURES
RADIATION FLUX
TRANSPORT THEORY
TWO-DIMENSIONAL CALCULATIONS