1. NAME AND TITLE
NANICK: Infinitely-Diluted Multigroup Cross-Section Generator -- from ENDF/B.
Soreq Nuclear Research Center, Yavne, Israel.
3. CODING LANGUAGE AND COMPUTER
Fortran IV; IBM 360/370.
4. NATURE OF PROBLEM SOLVED
NANICK generates flux-averaged, infinite-dilution, group constants. These are sigmat, sigmac,
sigmain, sigmael, sigmaf, nu x sigmaf, sigman,2n, mu, xi, with usual meaning of the symbols, and the
scattering matrix. It is supplemented by PSR-121/NASIF-NARES for temperature dependent shielding
5. METHOD OF SOLUTION
NANICK reads, from the ENDF/B II, III, IV, or V tape, the data for the processed material in any form or option in which these data come. It can treat up to an entire ENDF/B tape in one run. It searches for the first requested material, processes it, searches for the second, and so on, until the list of materials is exhausted.
In the resolved resonance region, the exact contribution of each cold Breit-Wigner resonance to
each energy group is taken into account. In the unresolved region, the cross sections are treated in
accordance with Appendix D of ENDF-102. The smooth-file is treated by the Centipede method. The
elastic scattering matrix is treated by the ABN method using xi in file 3 of ENDF/B (excluding
hydrogen, which is computed exactly). The inelastic scattering matrix is computed from the data in
files 3 and 5.
6. RESTRICTIONS OR LIMITATIONS
Up to 30 energy groups can be handled. Only Breit-Wigner resonances are treated. The lethargy
width of a group should be larger than the xi of the isotope for the ABN method of computing the
elastic scattering matrix.
7. TYPICAL RUNNING TIME
On an IBM 370/165 computer, about one minute is required for 235U and 14 seconds for 238U.
8. COMPUTER HARDWARE REQUIREMENTS
NANICK is operable on the IBM 360/370 computers.
9. COMPUTER SOFTWARE REQUIREMENTS
A Fortran IV compiler is required.
a. Included in the documentation:
Y. Gur, "NANICK - A Program for the Computation of Infinitely Diluted Multigroup Cross
Sections from ENDF/B-V Nuclear Data Files," Informal paper (April 1978).
b. Background information:
D. Garber, C. Dunford, and S. Pearlstein, "Data Formats and Procedures for the ENDF Neutron Cross Section Library, ENDF-102" (1975).
I. I. Bondarenko, ed., "Group Constants for Nuclear Reactor Calculations," New York:
Consultants Bureau (1964).
11. CONTENTS OF CODE PACKAGE
Included are the referenced document (10.a) and one (1.2MB) DOS diskette which contains the
source code and sample problem input and output.
12. DATE OF ABSTRACT
December 1978; revised March 1984.
KEYWORDS: ENDF FORMAT; MULTIGROUP CROSS SECTION PROCESSING