1. NAME AND TITLE
REX2-87: A Code For Calculating Self-Shielded Multigroup Neutron Cross Sections and Self-Shielding Factors From Preprocessed ENDF/B Basic Data Files
Indira Gandhi Center for Atomic Research, India, through NEA Data Bank, France.
3. CODING LANGUAGE AND COMPUTER
Fortran IV; VAX 8810.
4. NATURE OF PROBLEM SOLVED
REX2-87 is a computer code developed for the calculation of self-shielded multigroup average
cross sections, and self-shielding factors for total, elastic, fission and capture processes from an
ENDF/B formatted nuclear data file in which the tabulated cross sections follow linear interpolation
5. METHOD OF SOLUTION
The linearly interpolated tabulated cross sections are normally obtained with the preprocessing
codes RECENT, LINEAR, SIGMA1 (see PSR-159). REX2-87 is applied on the output of the
SIGMA1 code which gives the Doppler broadened point cross sections so that self-shielded average
cross sections and the associated self-shielding factors can be obtained corresponding to the
temperature to which the point cross sections refer. The code calculates by default the averages in the
resolved resonance region.
6. RESTRICTIONS OR LIMITATIONS
The maximum number of energy groups is 620.
7. TYPICAL RUNNING TIME
The sample input problems were tested at NEA Data Bank on a Vax 8810 computer. The total
time was 85 CPU seconds.
8. COMPUTER HARDWARE REQUIREMENTS
REX2-87 runs on the VAX 8810.
9. COMPUTER SOFTWARE REQUIREMENTS
The code was written in Fortran IV. The NEA Data Bank used the VAX Fortran (version 5.0)
under the VAX/VMS operating system.
V. Gopalakrishnan, and S. Ganesan, "REX2-87, A Code for Calculating Self-Shielded Multigroup
Neutron Cross Sections and Self-Shielding Factors From Preprocessed ENDF/B Basic Data File,"
IGCAR/NDS-18 (IAEA 0935) (November 1988).
11. CONTENTS OF CODE PACKAGE
Included are the referenced document and one DS/HD (1.2 MB) 5.25-inch diskette.
12. DATE OF ABSTRACT
KEYWORDS: CROSS SECTION PROCESSING; ENDF FORMAT; NEUTRON; NEUTRON CROSS SECTION PROCESSING; SELF-SHIELDING