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PSR-0132 MACK.

MACK, Fluence to Kerma Generator from ENDF/B

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1. NAME OR DESIGNATION OF PROGRAM:  MACK.
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2. COMPUTERS
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Program name Package id Status Status date
MACK-IV PSR-0132/01 Tested 01-MAY-1979

Machines used:

Package ID Orig. computer Test computer
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).
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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.
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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.
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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.
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7. UNUSUAL FEATURES OF THE PROGRAM

MACK is the first code to calculate neutron fluence-to-kerma factors from nuclear data in ENDF format.
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8. RELATED AND AUXILIARY PROGRAMS

Since the code has a 'built-in' resonance treatment and recognizes all of the multiplicity of ENDF data formats, it does not depend on any other code. Only a nuclear data library in ENDF format is required.
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9. STATUS
Package ID Status date Status
PSR-0132/01 01-MAY-1979 Tested at NEADB
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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.
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11. MACHINE REQUIREMENTS

Approximately 65k words of core storage. One scratch tape or disk is always needed in addition to the standard I/O devices. One or two additional tapes may be needed depending on the characteristics of the problem (1).
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12. PROGRAMMING LANGUAGE(S) USED
Package ID Computer language
PSR-0132/01 FORTRAN-IV
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13. OPERATING SYSTEM OR MONITOR UNDER WHICH PROGRAM IS EXECUTED

With
UNIVAC-1108 : EXEC-8. With IBM 360 : O/360 and FORTRAN-H compiler.
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14. ANY OTHER PROGRAMMING OR OPERATING INFORMATION OR RESTRICTIONS

The program presently consists of about 5700 FORTRAN statements in 69 subroutines. A 4-segment overlay structure is presently used.
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15. NAME AND ESTABLISHMENT OF AUTHOR: Oak Ridge National Laboratory
Oak Ridge, Tennessee, U.S.A.
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16. MATERIAL AVAILABLE
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
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17. CATEGORIES
  • B. Spectrum Calculations, Generation of Group Constants and Cell Problems

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