NAME OR DESIGNATION OF PROGRAM, COMPUTER, NATURE OF PHYSICAL PROBLEM SOLVED, METHOD OF SOLUTION, RESTRICTIONS ON THE COMPLEXITY OF THE PROBLEM, TYPICAL RUNNING TIME, UNUSUAL FEATURES OF THE PROGRAM, RELATED AND AUXILIARY PROGRAMS, STATUS, REFERENCES, MACHINE REQUIREMENTS, LANGUAGE, OPERATING SYSTEM OR MONITOR UNDER WHICH PROGRAM IS EXECUTED, ANY OTHER PROGRAMMING OR OPERATING INFORMATION OR RESTRICTIONS, NAME AND ESTABLISHMENT OF AUTHOR, MATERIAL, CATEGORIES

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Program name | Package id | Status | Status date |
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MACK-IV | PSR-0132/01 | Tested | 01-MAY-1979 |

Machines used:

Package ID | Orig. computer | Test computer |
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PSR-0132/01 | IBM 360 series | IBM 360 series |

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3. NATURE OF PHYSICAL PROBLEM SOLVED

The principal purpose of the program is in calculating pointwise neutron energy release parameters (fluence-to-kerma factors) at an arbitrary energy mesh from nuclear data in ENDF/B format (2). The kerma factors are of prime importance for calculating heating and dose rates in any nuclear system. The program processes all reactions significant to energy deposition. In addition, the program calculates energy group kerma factors and group cross sections by reactions (group constants not transfer matrices) averaged over an arbirary input weighting function or any of the 'built-in' functions. When resonance data is available, the code calculates the contribution from the resolved and unresolved resonance parameters. The pointwise cross sections, pointwise kerma factors, energy group cross sections and energy group kerma factors can be printed, punched, and/or saved on tape for all reactions and the sum as selected by input. The pointwise kerma factors can be saved for later use (3) to generate group kerma factors for a different energy group structure or possibly for inclusion in the ENDF/B evaluation for the nuclide with the appropriate MT numbers in the 300's series (2).

The principal purpose of the program is in calculating pointwise neutron energy release parameters (fluence-to-kerma factors) at an arbitrary energy mesh from nuclear data in ENDF/B format (2). The kerma factors are of prime importance for calculating heating and dose rates in any nuclear system. The program processes all reactions significant to energy deposition. In addition, the program calculates energy group kerma factors and group cross sections by reactions (group constants not transfer matrices) averaged over an arbirary input weighting function or any of the 'built-in' functions. When resonance data is available, the code calculates the contribution from the resolved and unresolved resonance parameters. The pointwise cross sections, pointwise kerma factors, energy group cross sections and energy group kerma factors can be printed, punched, and/or saved on tape for all reactions and the sum as selected by input. The pointwise kerma factors can be saved for later use (3) to generate group kerma factors for a different energy group structure or possibly for inclusion in the ENDF/B evaluation for the nuclide with the appropriate MT numbers in the 300's series (2).

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4. METHOD OF SOLUTION

The expressions for the energy release per reaction are obtained from a solution of the kinematics of nuclear reactions. The anisotropy of elastic and inelastic scattering is considered. The contribution to energy deposition from radioactive decay of the residual nucleus can be added by reaction and is calculated using Fermi theory in the case of beta decay. In the resolved resonance region, MACK accepts either single or multi level Breit-Wigner parameters. Doppler broadening is performed at an arbirary input temperature. The unresolved resonance treatment includes some shielding effects through a 1/sigmat weighting. The energy group kerma factors and cross sections are calculated by averaging the pointwise data over either a user supplied input spectrum or 'built-in' weighting functions. The program calculates the contribution to the energy release from all reactions and the accuracy of the kerma factor calculation is set only by the availability of the required nuclear data.

The expressions for the energy release per reaction are obtained from a solution of the kinematics of nuclear reactions. The anisotropy of elastic and inelastic scattering is considered. The contribution to energy deposition from radioactive decay of the residual nucleus can be added by reaction and is calculated using Fermi theory in the case of beta decay. In the resolved resonance region, MACK accepts either single or multi level Breit-Wigner parameters. Doppler broadening is performed at an arbirary input temperature. The unresolved resonance treatment includes some shielding effects through a 1/sigmat weighting. The energy group kerma factors and cross sections are calculated by averaging the pointwise data over either a user supplied input spectrum or 'built-in' weighting functions. The program calculates the contribution to the energy release from all reactions and the accuracy of the kerma factor calculation is set only by the availability of the required nuclear data.

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5. RESTRICTIONS ON THE COMPLEXITY OF THE PROBLEM

Because of the variable dimensioning technique used in the program, the principal restriction on the size of the problem is the availability of sufficient core storage. Problems with up to about 1500 energy points can be run in less than 65k words of core storage. Core storage requirements are not affected by the number of reactions processed for the nuclides or the number of nuclides in the run.

The code recognizes almost all of the multiplicity of data formats allowed by ENDF.

Because of the variable dimensioning technique used in the program, the principal restriction on the size of the problem is the availability of sufficient core storage. Problems with up to about 1500 energy points can be run in less than 65k words of core storage. Core storage requirements are not affected by the number of reactions processed for the nuclides or the number of nuclides in the run.

The code recognizes almost all of the multiplicity of data formats allowed by ENDF.

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6. TYPICAL RUNNING TIME

Running time depends on (a) number of resonances, (b) Doppler broadening, (c) size of the energy mesh selected for calculating the pointwise kerma factors, (d) number of groups, and (e) number of reactions processed. For nuclides with no resonance data, the typical running time is approximately 2 to 3 minutes on the UNIVAC-1108 for 1000 energy points and 100 groups. The running time for nuclides with resonance data depends strongly on the number of resonances and Doppler broadening and varies from 3 to 15 minutes on the UNIVAC-1108. Running times quoted are lower by approximately a factor of 4 for the IBM 360/91.

Running time depends on (a) number of resonances, (b) Doppler broadening, (c) size of the energy mesh selected for calculating the pointwise kerma factors, (d) number of groups, and (e) number of reactions processed. For nuclides with no resonance data, the typical running time is approximately 2 to 3 minutes on the UNIVAC-1108 for 1000 energy points and 100 groups. The running time for nuclides with resonance data depends strongly on the number of resonances and Doppler broadening and varies from 3 to 15 minutes on the UNIVAC-1108. Running times quoted are lower by approximately a factor of 4 for the IBM 360/91.

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10. REFERENCES

- M.A. Abdou, C.W. Maynard, and R.Q. Wright:

'MACK, A Computer Program to Calculate Neutron Energy Release

Parameters (Fluence-to-Kerma Factors) and Multigroup Neutron

Reaction Cross Sections from Nuclear Data in ENDF Format'

ORNL-TM-3994.

- M.K. Drake, Editor

'Data Formats and Procedures for the ENDF Neutron Cross Section

Library'

BNL-50274 (ENDF-102, Vol. 1) (October 1970).

- M.A. Abdou and R.W. Roussin:

'MACKLIB, Neutron Fluence-to-Kerma Factor Library Generated with

MACK from Nuclear Data in ENDF Format'

ORNL-TM-3995.

- M.A. Abdou, C.W. Maynard, and R.Q. Wright:

'MACK, A Computer Program to Calculate Neutron Energy Release

Parameters (Fluence-to-Kerma Factors) and Multigroup Neutron

Reaction Cross Sections from Nuclear Data in ENDF Format'

ORNL-TM-3994.

- M.K. Drake, Editor

'Data Formats and Procedures for the ENDF Neutron Cross Section

Library'

BNL-50274 (ENDF-102, Vol. 1) (October 1970).

- M.A. Abdou and R.W. Roussin:

'MACKLIB, Neutron Fluence-to-Kerma Factor Library Generated with

MACK from Nuclear Data in ENDF Format'

ORNL-TM-3995.

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PSR-0132/01

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

PSR0132_01.001 | INFORMATION | 5 |

PSR0132_01.002 | LIBRARY DATA (BCD) | 5277 |

PSR0132_01.003 | SOURCE PROGRAM (F4,EBCDIC) | 10082 |

PSR0132_01.004 | TIMMING ROUTINE (F4,EBCDIC) | 24 |

PSR0132_01.005 | DD CARDS FOR THE EXECUTION | 19 |

PSR0132_01.006 | OVERLAY CARDS | 29 |

PSR0132_01.007 | SAMPLE PROB. CASE 1 -- INPUT DATA | 71 |

PSR0132_01.008 | SAMPLE PROB. CASE 1 -- PRINTED OUTPUT | 6608 |

PSR0132_01.009 | SAMPLE PROB. CASE 2 -- INPUT DATA | 18 |

PSR0132_01.010 | SAMPLE PROB. CASE 2 -- PRINTED OUTPUT | 3902 |

PSR0132_01.011 | SAMPLE PROB. CASE 2 -- PUNCHED OUTPUT | 137 |

PSR0132_01.012 | SAMPLE PROB. CASE 3 -- INPUT DATA | 10 |

PSR0132_01.013 | SAMPLE PROB. CASE 3 -- PRINTED OUTPUT | 5966 |

Keywords: Breit-Wigner formula, Doppler broadening, ENDF/B, absorption, anisotropic scattering, cross sections, decay, group constants, neutron reactions, resonance.