1. NAME AND TITLE OF DATA LIBRARY
MACKLIB: 100 Group Neutron Kerma Factors and Reaction Cross Sections Generated by
MACK from Data in ENDF Format.
2. NAME AND TITLE OF DATA RETRIEVAL PROGRAM
SUPREM: A Program to Edit or Convert MACKLIB Data to Forms Suitable for Input to
ANISN or DTF-IV.
Nuclear Engineering Department, University of Wisconsin, Madison, Wisconsin.
Oak Ridge National Laboratory, Oak Ridge, Tennessee.
4. HISTORICAL BACKGROUND AND INFORMATION
Studies at the University of Wisconsin Nuclear Engineering Department of calculational methods
for nuclear heating and neutronics and photonics design for Controlled Thermonuclear Research (CTR)
blankets and shields revealed the need for, and led to the development of, a computer code to calculate
neutron Kerma factors from evaluated data in ENDF format. The code, called MACK, was written
by Mohamed Abdou, then a University of Wisconsin graduate student, with assistance from
R. Q. Wright of the Computer Sciences Division of the Oak Ridge National Laboratory and members
of the RSIC staff. MACK was then used at the University of Wisconsin to generate a library of 100
group neutron Kerma factors and reaction cross sections for many materials of interest in CTR
neutronics studies. This data set, designated MACKLIB, was then made available to RSIC for general
5. APPLICATION OF THE DATA
The data can be used in conjunction with 100 group neutron transport calculations (using, for
example, the DLC-2/100G neutron cross section library) to determine the spatial distribution of
neutron heating and reaction rates. The data are in the format of dummy cross section materials which
can be used in the activity calculation option in ANISN to calculate the desired heating and reaction
rates. The data were developed for and should be particularly useful in CTR design neutronics studies.
6. SOURCE AND SCOPE OF DATA
DLC-29/MACKLIB was generated using MACK from evaluated data in the ENDF/B Version III library. The data were produced as averaged values in the 100-group GAM structure. The weighting function used for the group averaging accounts for an energy distribution of D-T neutrons around 14 MeV and is 1/E at lower energies. This function is denoted in the packaged documentation as the Los Alamos CTR spectrum. When resonance parameter data were given, resolved and unresolved resonance region contributions were calculated using the infinite dilution approximation. Doppler broadening was performed for a temperature of 300 K. The contribution to Kerma from charged particle radioactive decay was included for reaction product nuclei with half-lives less than 50 days.
The materials included in the library are:
H, D, He, 6Li, 7Li, 9Be, 10B, 11B, 12C, 16O, 23Na, 27Al, V, Cr, Fe, Ni, Cu, 63Cu, 65Cu, Nb, 181Ta, 182W,
183W, 184W, 186W, and Pb.
A more detailed description of the contents is found in the packaged documentation.
The library consists of 100 group reaction cross sections and Kerma factors. The multigroup
reaction cross sections included are those which were present on the original ENDF/B tape for a given
material. The multigroup Kerma factors are given for elastic scattering, total discrete level inelastic
scattering, total continuum inelastic scattering, (n,2n), (n,gamma), (n,charged particle), (n;n',charged
particles), and Kerma factors for the sum of all reactions. The cross section units are barns/atom and
the Kerma factor units are (eV,barns)/atom. Tables listing the reaction cross sections and Kerma
factors for each material included in DLC-29/MACKLIB are given in the packaged documentation.
7. DISCUSSION OF THE DATA RETRIEVAL PROGRAM
The retrieval program is a variation on the one provided for DLC-2/100G. It has been modified
to accept the DLC-29/MACKLIB card image format which is similar to that for DLC-2/100G. For
DLC-29/MACKLIB, however, each material is preceded by a title card, then a card with 14* in the
first 3 columns, and the data are followed by a card with a "T" in column 3. The retrieval program
can be used to edit the library. It can also be used to create an unformatted library tape or punched
cards which can be read by ANISN as cross sections for dummy materials. This facilitates the use of
the library for calculating detailed heating and reaction rates for a CTR system being analyzed with
one-dimensional discrete ordinates methods. The dummy materials containing Kerma factors and
reaction cross sections for the materials of interest can be used in "activity" calculations available with
codes such as ANISN. The input requirements for the retrieval program are described in the packaged
8. DATA FORMAT AND COMPUTER
Mixed/EBCDIC card images; IBM 360/370.
9. TYPICAL RUNNING TIME
On an IBM 360/91 computer, it requires less than 15 seconds to compile and edit data for 3
M. A. Abdou and R. W. Roussin, "MACKLIB, 100 Group Neutron Fluence-to-Kerma Factors and
Reaction Cross Sections Generated by the MACK Computer Program from Data in ENDF Format,"
ORNL-TM-3995 (August 1974).
11. CONTENTS OF LIBRARY
Included are the referenced document and one (1.2MB) DOS diskette which contains the library,
the retrieval program, the sample problem input and output.
12. DATE OF ABSTRACT
January 1974; reviewed May 1984.
KEYWORDS: KERMA FACTORS; MULTIGROUP CROSS SECTIONS; NEUTRON CROSS SECTIONS; REACTION CROSS SECTIONS