Interface-flux nodal transport method

Conference · Mon Jan 01 00:00:00 EST 1990

OSTI ID:7247722

Kohut, P

The development of the interface-flux nodal (IFN) method is presented to determine the flux distribution in reactor cells, cores and shielding. The method offers geometric flexibility, high order of spatial expansions of the node-interior sources and the node surface quantities. The surface-integral formulation is reduced to response-matrix-like global equations through coupling coefficients which are generalized expressions for escape and transmission probabilities. The spatial distribution of the neutron flux may be represented by high order polynomials using geometric basis functions determined by a least-square minimalization technique. The angular dependency of the outgoing/incoming surface flux components is treated using a general DP{sub n} expansion and the spatial variation is handled with a boundary element technique. The scattering iterations are eliminated by using an explicit expansion of the scalar fluxes. Based on the IFN method a computer code has been developed capable of calculating fixed-source and eigenvalue problems. Test problems for 1-D and 2-d X-Y and hexagonal geometries are presented including comparison with other techniques to demonstrate the validity and accuracy of the IFN method. 15 refs., 3 figs., 5 tabs.

• Research Organization: Brookhaven National Lab., Upton, NY (USA)
• Sponsoring Organization: DOE/ER
• DOE Contract Number: AC02-76CH00016
• OSTI ID: 7247722
• Report Number(s): BNL-43929; CONF-9004138-3; ON: DE90008515; TRN: 90-008857
• Resource Relation: Conference: 3. Canadian Nuclear Society international conference on simulation methods in nuclear engineering, Montreal (Canada), 18-20 Apr 1990
• Country of Publication: United States
• Language: English

Related Subjects

22 GENERAL STUDIES OF NUCLEAR REACTORS
75 CONDENSED MATTER PHYSICS
SUPERCONDUCTIVITY AND SUPERFLUIDITY
NEUTRON FLUX
NEUTRON TRANSPORT THEORY
REACTOR CORES
ACCURACY
BOUNDARY CONDITIONS
EIGENFUNCTIONS
FLEXIBILITY
FLUX DENSITY
GAUSS FUNCTION
HOMOGENEOUS PLASMA
MATHEMATICAL MODELS
MATRICES
MEASURING METHODS
ONE-DIMENSIONAL CALCULATIONS
PROBABILITY
REACTOR SAFETY
STABILITY
SURFACE PROPERTIES
TWO-DIMENSIONAL CALCULATIONS
VALIDATION
VARIATIONS
FUNCTIONS
MECHANICAL PROPERTIES
PLASMA
RADIATION FLUX
REACTOR COMPONENTS
SAFETY
TENSILE PROPERTIES
TESTING
TRANSPORT THEORY
220900* - Nuclear Reactor Technology- Reactor Safety
220200 - Nuclear Reactor Technology- Components & Accessories
656003 - Condensed Matter Physics- Interactions between Beams & Condensed Matter- (1987-)