Computer Programs

NAME OR DESIGNATION OF PROGRAM, COMPUTER, NATURE OF PHYSICAL PROBLEM SOLVED, METHOD OF SOLUTION, RESTRICTIONS, TYPICAL RUNNING TIME, FEATURES, AUXILIARIES, STATUS, REFERENCES, REQUIREMENTS, LANGUAGE, OPERATING SYSTEM, OTHER RESTRICTIONS, NAME AND ESTABLISHMENT OF AUTHOR, MATERIAL, CATEGORIES

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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 |
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ZZ-DLC-14 | 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. NATURE OF PHYSICAL PROBLEM SOLVED

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 basic idea behind the distribution of this ANISN input data is to allow potential users to repeat the ANISN calculations reported in ref. (1). It is felt that it will be more economical to repeat the calculations rather than to distribute the results of the Straker-Gritzner (1) calculations.

However, the cross section part of the data can actually be used in DOT or MORSE or any transport code which will accept input cross section in the FIDO format.

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.

%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%%

The basic idea behind the distribution of this ANISN input data is to allow potential users to repeat the ANISN calculations reported in ref. (1). It is felt that it will be more economical to repeat the calculations rather than to distribute the results of the Straker-Gritzner (1) calculations.

However, the cross section part of the data can actually be used in DOT or MORSE or any transport code which will accept input cross section in the FIDO format.

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4. METHOD OF SOLUTION

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 15 MeV 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 radius 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/HPIC to an 18 group structure using MUG. The POPOP-4 code 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.

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 15 MeV 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 radius 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/HPIC to an 18 group structure using MUG. The POPOP-4 code 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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DLC-0014/01

File name | File description | Records |
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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 |

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