Computer Programs
DLC-0014 ZZ-AIR.
last modified: 01-MAR-1972 | catalog | categories | new | search |

DLC-0014 ZZ-AIR.

ZZ AIR, Group Constant Library of Secondary Gamma Transport in Air for ANISN Calculation

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1. NAME

ZZ-AIR.

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2. COMPUTERS
To submit a request, click below on the link of the version you wish to order. Rules for end-users are available here.
Program name Package id Status Status date
ZZ-AIR DLC-0014/01 Tested 01-MAR-1972

Machines used:

Package ID Orig. computer Test computer
DLC-0014/01 Many Computers Many Computers
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3. DESCRIPTION

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FORMAT: ANISN, DOT, MORSE (FIDO format)

 

NUMBER OF GROUPS: 22 neutron / 18 gamma-ray

 

NUCLIDES: air

 

ORIGIN: ENDF/B for neutron cross sections, DLC-4/HPIC for gamma-ray and DLC-12/POPLIB for secondary gamma-ray production.

 

WEIGHTING SPECTRUM: 1/E for neutron cross sections.

 

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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.

 

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 Straker-Gritzner’s calculations.

 

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4. METHODS

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.

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6. TYPICAL RUNNING TIME

Using ANISN, the problem ran in approximately 2 minutes on the IBM 360/91 computer

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9. STATUS
Package ID Status date Status
DLC-0014/01 01-MAR-1972 Tested at NEADB
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10. REFERENCES
DLC-0014/01, included references:
- 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).
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12. PROGRAMMING LANGUAGE(S) USED
No specified programming language
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15. NAME AND ESTABLISHMENT OF AUTHORS

Oak Ridge National Laboratory
Oak Ridge, Tennessee, U.S.A

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16. MATERIAL AVAILABLE
DLC-0014/01
File name File description Records
DLC0014_01.001 INFORMATION 1
DLC0014_01.002 ANISN MAIN ROUTINE (SOURCE) 305
DLC0014_01.003 SAMPLE PROBLEM DD-CARDS AND DATA 755
DLC0014_01.004 SAMPLE PROBLEM OUTPUT 2039
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17. CATEGORIES
  • Z. Data

Keywords: ENDF/B, cross sections, data, doses, gamma radiation, neutron transport theory, transport theory.