1. NAME AND TITLE
FDMXPC: Code System for Calculation of Neutron Transmission and Other Functionals from Evaluated Data in ENDF Format.
Central Research Institute of Physics, Institute of Atomic Energy Research, Budapest, Hungary.
3. CODING LANGUAGE AND COMPUTER/OPERATING SYSTEM
Fortran-77; PCAT; PC-386; PC-486.
4. NATURE OF PROBLEM SOLVED
FDMXPC processes evaluated neutron cross-section data in ENDF format to group-averaged functionals mainly for the modelling of neutron transmission experiments. It is also possible to treat mixtures of isotopes. This release of the code has been extended to use ENDF-6 format.
5. METHOD OF SOLUTION
Breit-Wigner, Reich-Moore or Adler-Adler neutron cross-section formula are integrated numerically over specified energy intervals by means of Romberg's integration method.
6. RESTRICTIONS OR LIMITATIONS
7. TYPICAL RUNNING TIME
Very different, depending upon the problem specification.
8. COMPUTER HARDWARE REQUIREMENTS
Personal computer minimum AT with 286 processor, math coprocessor is recommended.
9. COMPUTER SOFTWARE REQUIREMENTS
Minimum: DOS 4.0, Microsoft Fortran compiler 5.0, or any other equivalent system or compiler.
P. Vertes, "FDMXPC README."
P. Vertes, "Calculation of Transmission and Other Functionals from Evaluated Data in ENDF Format by Means of Personal Computers," KFKI-1991-10/G (1991).
P. Vertes, "Processing of Evaluated Neutron Data Files in ENDF Format on Personal Computers," Informal Notes (1991).
11. CONTENTS OF CODE PACKAGE
Included in the package are the referenced documents and 1 DS/HD disk which includes the source files, test case input data and output.
12. DATE OF ABSTRACT
KEYWORDS: ENDF FORMAT; MICROCOMPUTER; NEUTRON CROSS SECTION PROCESSING; MULTIGROUP CROSS-SECTION PROCESSING