1. NAME AND TITLE OF DATA LIBRARY
AIR: Sample Input to ANISN for Calculation of Neutron and Secondary Gamma-Ray
Transport in Air.
2. NAME AND TITLE OF DATA RETRIEVAL PROGRAM
No retrieval program is included in the package.
Oak Ridge National Laboratory, Oak Ridge, Tennessee.
4. HISTORICAL BACKGROUND AND INFORMATION
The ANISN input data (including the air cross section data) were developed for use by E. A.
Straker and M. L. Gritzner of Oak Ridge National Laboratory, Oak Ridge, Tennessee, for the
calculation of neutron and secondary gamma-ray transport in infinite homogeneous air. Their results
are reported in the packaged documentation.
5. APPLICATION OF THE DATA
The basic idea behind the distribution of this ANISN input data is to allow potential users to repeat the ANISN calculations reported in the packaged documentation. It is felt that it will be more economical to repeat the calculations rather than to distribute the results of the authors' calculations.
However, the cross section part of the data can actually be used in CCC-89/DOT or
CCC-127/MORSE or any transport code which will accept input cross section data in the FIDO
6. SOURCE AND SCOPE OF DATA
The sample input data for ANISN are for a P5, S16 calculation of the transport of neutrons and secondary gamma rays from a 12.2 to 15MeV point neutron source in an infinite air medium. The source is actually uniformly distributed in the first interval (500 cm radius) of a spherical medium of air with a radius of 3005 meters.
The problem is set up for calculating various "detector responses" by means of the "activity" option available with ANISN. This is accomplished by providing a cross section table for a "material" which has detector responses in certain table positions. Then the inclusion of appropriate input data for 22$ and 23$ arrays causes the group fluxes to be multiplied by the group response function values to give the desired answer. The neutron detector responses calculated by this sample problem are Henderson tissue dose, Snyder-Neufeld dose, tissue kerma, and air kerma. The gamma-ray response functions calculated are Henderson tissue dose and air kerma.
The neutron cross sections were first reduced from point data from ENDF/B to a 104 fine-group
structure with a modified version of CSP, assuming a 1/E weighting factor. The gamma-ray data were
reduced from point data from DLC-4/HPICO to an 18 group structure using MUG. PSR-11/POPOP4
was used to convert secondary gamma-ray production data from DLC-12/POPLIB to neutron-to-gamma-ray group transfer cross sections. The coupled set (104 neutron, 18 gamma-ray groups) was
then collapsed to 22 neutron and 18 gamma-ray groups with ANISN, using as the weighting function
the spectrum from a spatially uniform source of 14 MeV neutrons in an infinite air medium with a
density of 1.11 mg/cc. The resulting data are coupled macroscopic multigroup, P5 expansion cross
sections for air punched on cards and suitable for input to the ANISN code.
7. DISCUSSION OF THE DATA RETRIEVAL PROGRAM
8. DATA FORMAT COMPUTER
BCD/EBCDIC card images; IBM 360/370. The data are written in the special format used for card
input to the ANISN code.
9. TYPICAL RUNNING TIME
Using CCC-82/ANISN-CEA, the problem ran in approximately 2 minutes on the IBM 360/91
E. A. Straker and M. L. Gritzner, "Science Applications, Inc., Memorandum," Informal note (July 1973).
E. A. Straker and M. L. Gritzner, "Neutron and Secondary Gamma-Ray Transport in Infinite
Homogeneous Air" ORNL-4464 (December 1969).
11. CONTENTS OF LIBRARY
Included are the referenced documents and one tape cartridge in TAR format which contains the
source code and sample problem input and output.
12. DATE OF ABSTRACT
June 1971; reviewed May 1984.
KEYWORDS: AIR CROSS SECTIONS; ANISN FORMAT; BENCHMARK PROBLEM CROSS SECTIONS; COUPLED NEUTRON-GAMMA-RAY CROSS SECTIONS; MULTIGROUP CROSS SECTIONS