DOXCY - A discrete ordinates approximation of neutron transport in heterogeneous rod lattices

Journal Article · Sat Aug 01 04:00:00 EDT 1987 · Nucl. Sci. Eng.; (United States)

OSTI ID:5808101

Martens, H D; Stegemann, D

For calculating the fine flux distribution in heterogeneous fuel rod lattices, an exact treatment of the geometry and the use of a high-order approximation of the transport theory is needed. For this purpose, a discrete ordinates solution of the neutron transport equation for mixed geometry has been developed. The discretization of the space is performed in separate one-dimensional cylindrical coordinate systems, imbedded in a two-dimensional rectangular mesh grid. The geometrical link between the cylindrical and the rectangular systems is achieved by approximating the outer circle of each cylindrical system by a polygon with side numbers greater than or equal to8. Thus, each cylindrical geometry is enclosed in a two-dimensional mesh grid consisting of rectangles, trapeziums, and triangles. Because of the different orientation of the angular segmentation in XY and R coordinates, transfer coefficients are derived to calculate the directional flux distribution on the boundary between both systems. A special set of equal-weighted quadrature coefficients (EQ/sub n/) is used to get transfer coefficients, providing a fast and accurate solution. The method is realized in a program called DOXCY, which runs within the nuclear program system RSYST. The program is verified on selected benchmark problems. The numerical results are given, showing the advantage and limits of the method.

• Research Organization: Technischer Uberwachungs-Verein, Hannover e.V., Am TUV 1, D-3000 Hannover 81
• OSTI ID: 5808101
• Journal Information: Nucl. Sci. Eng.; (United States), Journal Name: Nucl. Sci. Eng.; (United States) Vol. 96:4; ISSN NSENA
• Country of Publication: United States
• Language: English

Related Subjects

22 GENERAL STUDIES OF NUCLEAR REACTORS
220100* -- Nuclear Reactor Technology-- Theory & Calculation
654003 -- Radiation & Shielding Physics-- Neutron Interactions with Matter
73 NUCLEAR PHYSICS AND RADIATION PHYSICS
BENCHMARKS
CALCULATION METHODS
COMPUTER CODES
COORDINATES
D CODES
FUEL ELEMENTS
FUEL RODS
MESH GENERATION
NEUTRON FLUX
NEUTRON TRANSPORT THEORY
NUMERICAL SOLUTION
R CODES
RADIATION FLUX
REACTOR COMPONENTS
TRANSPORT THEORY