Computer Programs

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

[ top ]

[ top ]

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-RITTS | DLC-0011/01 | Tested | 01-OCT-1970 |

Machines used:

Package ID | Orig. computer | Test computer |
---|---|---|

DLC-0011/01 | Many Computers | Many Computers |

[ top ]

3. NATURE OF PHYSICAL PROBLEM SOLVED

FORMAT: ANISN, DTF-4, DOT and MORSE.

NUMBER OF GROUPS: 100 neutron energy groups (14.92 MeV to thermal) 21 gamma-ray energy groups (14.0 to 0.01 MeV)

NUCLIDES: H, C, O, N, Na, Mg, P, S, Cl, K, and Ca, (microscopic cross sections) and 9 organic materials including 11-element standard man, 4-element standard man, skin, bone, tissue, brain, lung, red marrow, and muscle (macroscopic cross sections).

ORIGIN: ENDF/B for H, C, N, O, Na, and Mg

O5R library for Ca, S, and K

GAM-2 library for Cl

Evaluation by J. J. Ritts for P.

WEIGHTING SPECTRUM: 1/E for the top 99 groups and Maxwellian for the thermal group values.

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

DLC-11 data is suitable for neutron, gamma-ray, or coupled neutron and gamma-ray transport calculations. It is intended for use in multigroup discrete ordinates or Monte Carlo transport codes which treat anisotropic scattering by Legendre expansion up to order P3.

DLC-11 is a collection of multigroup cross section data which were compiled by J. J. Ritts for use in depth-dose calculations in anthropomorphic phantoms. For convenience the data are grouped as follows

1. A coupled 121-group (100 neutron, 21 gamma-ray) set of data for the 11 elements H, C, O, N. Na, Mg, P, S, Cl, K, and Ca. This set includes P3 coupled 121-group microscopic cross sections plus 121-group kerma factors for the 11 elements.

2. A 100-group set of neutron cross sections for the 11 elements.

3. A coupled 121-group set of macroscopic cross sections for 9 organic materials including 11-element standard man, 4-element standard man, skin, bone, tissue, brain, lung, red marrow, and muscle.

FORMAT: ANISN, DTF-4, DOT and MORSE.

NUMBER OF GROUPS: 100 neutron energy groups (14.92 MeV to thermal) 21 gamma-ray energy groups (14.0 to 0.01 MeV)

NUCLIDES: H, C, O, N, Na, Mg, P, S, Cl, K, and Ca, (microscopic cross sections) and 9 organic materials including 11-element standard man, 4-element standard man, skin, bone, tissue, brain, lung, red marrow, and muscle (macroscopic cross sections).

ORIGIN: ENDF/B for H, C, N, O, Na, and Mg

O5R library for Ca, S, and K

GAM-2 library for Cl

Evaluation by J. J. Ritts for P.

WEIGHTING SPECTRUM: 1/E for the top 99 groups and Maxwellian for the thermal group values.

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

DLC-11 data is suitable for neutron, gamma-ray, or coupled neutron and gamma-ray transport calculations. It is intended for use in multigroup discrete ordinates or Monte Carlo transport codes which treat anisotropic scattering by Legendre expansion up to order P3.

DLC-11 is a collection of multigroup cross section data which were compiled by J. J. Ritts for use in depth-dose calculations in anthropomorphic phantoms. For convenience the data are grouped as follows

1. A coupled 121-group (100 neutron, 21 gamma-ray) set of data for the 11 elements H, C, O, N. Na, Mg, P, S, Cl, K, and Ca. This set includes P3 coupled 121-group microscopic cross sections plus 121-group kerma factors for the 11 elements.

2. A 100-group set of neutron cross sections for the 11 elements.

3. A coupled 121-group set of macroscopic cross sections for 9 organic materials including 11-element standard man, 4-element standard man, skin, bone, tissue, brain, lung, red marrow, and muscle.

[ top ]

4. METHOD OF SOLUTION

The basic data sources were ENDF/B for H, C, N, O, Na, and Mg, the O5R library for Ca, S, and K, the GAM-2 library for Cl and an evaluation by Ritts for P. A 1/E spectrum was assumed for averaging the top 99 groups and a Maxwellian for averaging the thermal group values.

The gamma-ray cross sections were computed from DLC-3/HPIC using MUG.

The neutron-to-gamma-ray group transfer cross sections were generated, using POPOP-4, with account being taken for neutron capture, inelastic scattering, and other neutron reactions.

The 100-group neutron kerma factors were generated by DLC-10/ AVKER and the 21-group gamma-ray kerma factors by MUG.

The DLC-11 cross sections represent a P3 approximation to elastic (or Compton) scattering angular distributions. The 100 neutron groups cover an energy range from 14.92 MeV to thermal. For gamma-rays, 21 energy groups cover the range from 14.0 to 0.01 MeV. The group structures are given in ref. 2.

The basic data sources were ENDF/B for H, C, N, O, Na, and Mg, the O5R library for Ca, S, and K, the GAM-2 library for Cl and an evaluation by Ritts for P. A 1/E spectrum was assumed for averaging the top 99 groups and a Maxwellian for averaging the thermal group values.

The gamma-ray cross sections were computed from DLC-3/HPIC using MUG.

The neutron-to-gamma-ray group transfer cross sections were generated, using POPOP-4, with account being taken for neutron capture, inelastic scattering, and other neutron reactions.

The 100-group neutron kerma factors were generated by DLC-10/ AVKER and the 21-group gamma-ray kerma factors by MUG.

The DLC-11 cross sections represent a P3 approximation to elastic (or Compton) scattering angular distributions. The 100 neutron groups cover an energy range from 14.92 MeV to thermal. For gamma-rays, 21 energy groups cover the range from 14.0 to 0.01 MeV. The group structures are given in ref. 2.

[ top ]

[ top ]

[ top ]

[ top ]

8. RELATED AND AUXILIARY PROGRAMS

The retrieval program JRMACRO is used to read microscopic, multigroup, PN expansion cross section data, 'mix' this data into macroscopic cross section data as needed, and write the resulting set in a suitable output format. The cross section data considered here is of the general type used by particle transport computer codes such as DTF-4, ANISN, DOT, and MORSE.

JRMACRO accepts input cross sections by means of cards, tape written in card image format, an unformatted tape (binary), or a combination of the above. Formatted input (card or card image tape) is first read and stored on a tape or disk before any mixing of cross section data is performed.

The output cross section data may be in the form of a card image tape, an unformatted tape, or both.

The retrieval program JRMACRO is used to read microscopic, multigroup, PN expansion cross section data, 'mix' this data into macroscopic cross section data as needed, and write the resulting set in a suitable output format. The cross section data considered here is of the general type used by particle transport computer codes such as DTF-4, ANISN, DOT, and MORSE.

JRMACRO accepts input cross sections by means of cards, tape written in card image format, an unformatted tape (binary), or a combination of the above. Formatted input (card or card image tape) is first read and stored on a tape or disk before any mixing of cross section data is performed.

The output cross section data may be in the form of a card image tape, an unformatted tape, or both.

[ top ]

DLC-0011/01, included references:

-R.J. Ritts, R.W. Roussin, I.J.Brown;"JRMACRO: A Program for Converting Microscopic, Multigroup, P,

Expansion Cross Section Data into Corresponding

Macroscopic Data for Mixtures or Compounds"

ORNL-TM-3052

- J.J. Ritts, R.W. Roussin ;

"121 Group Coupled Neutron and Gamma-Ray

Cross-Section Library for Transport Codes (100 Neutron,

21 Gamma-Ray Groups)" Informal notes.

- R.W. Roussin ;

"Using ANISN to Reduce the DLC-2 100 Group

Cross-Section Data to a Smaller Number of Groups"

ORNL-TM-3049

- J.J. Ritts, M.Solomito, P.N.Stevens;

"The calculation of Neutron-Induced Physical Doses

in Human Tissues" (ORNL-TM-2991)

[ top ]

[ top ]

[ top ]

[ top ]

[ top ]

DLC-0011/01

File name | File description | Records |
---|---|---|

DLC0011_01.001 | INFORMATION | 11 |

DLC0011_01.002 | X-SEC LIBRARY FILE 1 | 36475 |

DLC0011_01.003 | X-SEC LIBRARY FILE 2 | 13522 |

DLC0011_01.004 | X-SEC LIBRARY FILE 3 | 8445 |

DLC0011_01.005 | JRMACRO ASSEMBLER ROUTINE (SOURCE) | 39 |

DLC0011_01.006 | JRMACRO SOURCE PROGRAM | 328 |

DLC0011_01.007 | JRMACRO SAMPLE PROBLEM DD-CARDS AND DATA | 21 |

DLC0011_01.008 | JRMACRO SAMPLE PROBLEM PUNCHED OUTPUT | 1469 |

DLC0011_01.009 | JRMACRO SAMPLE PROBLEM OUTPUT | 921 |

[ top ]

- C. Static Design Studies
- G. Radiological Safety, Hazard and Accident Analysis
- Z. Data.

Keywords: angular distribution, cross sections, data, elastic scattering, gamma radiation, multigroup, neutrons, transport theory.