1. NAME AND TITLE
MARCOPOLO: Code System for Calculating the Radial and Axial Neutron Diffusion
Coefficients in One-Group and Multigroup Theory.
Beijing Institute of Nuclear Engineering, People's Republic of China, through the OECD Nuclear
Energy Agency Data Bank, Gif-sur-Yvette, France.
3. CODING LANGUAGE AND COMPUTER
Fortran IV; IBM 360/370.
4. NATURE OF PROBLEM SOLVED
MARCOPOLO calculates the radial and axial diffusion coefficients in one-group and multi-group
theory for a cylinderized cell (Wigner-Seitz theory) with several concentric zones according to the
isotropic shock or linear anisotropic shock hypotheses.
5. METHOD OF SOLUTION
MARCOPOLO allows testing of the degree of approximation of the transport correction in various
types of lattices and shows that the axial coefficient may be strongly underestimated in certain cases.
This method also allows testing of the simple formulas presented in the past for diffusion coefficients
which leads to good results. The problem of the coupling between energy groups, which appears in
the calculation of diffusion coefficients (done with zero radial and axial Laplacian), is also analyzed
by the present method; it usually appears to be weak.
6. RESTRICTIONS OR LIMITATIONS
The maximum number of concentric zones is 9.
7. TYPICAL RUNNING TIME
On an IBM 3084Q computer, the test case included in the package ran in 28 seconds of CPU time.
RSIC made no additional studies of typical running times for MARCOPOLO.
8. COMPUTER HARDWARE REQUIREMENTS
MARCOPOLO is operable on the IBM 360/370 computers. Main storage requirements were
552 K bytes.
9. COMPUTER SOFTWARE REQUIREMENTS
A Fortran IV compiler is required.
C. Yang and P. Benoist, "Scattering Anisotropy and Neutron Leakage in Reactor Lattices," Nucl. Sci. Eng. 86, 47 (1984).
C. Yang and P. Benoist, "Anisotropie du Choc and Fuites de Neutrons dans un Reseau de
Reacteur," CEA-R-5182 (September 1982).
11. CONTENTS OF CODE PACKAGE
Included are the referenced documents and one (360KB) DOS diskette which contains the source
code and sample problem input.
12. DATE OF ABSTRACT
KEYWORDS: NEUTRON CROSS SECTION PROCESSING; REACTOR PHYSICS