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CCC-0187 SAM-CE.

SAM-CE, Time-Dependent 3-D Neutron Transport, Gamma Transport in Complex Geometry by Monte-Carlo

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1. NAME OR DESIGNATION OF PROGRAM:  SAM-CE.
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2. COMPUTERS
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Program name Package id Status Status date
SAM-CE CCC-0187/01 Tested 24-MAR-1981
SAM-CE CCC-0187/02 Tested 10-OCT-1979

Machines used:

Package ID Orig. computer Test computer
CCC-0187/01 CDC 6600 CDC 6600
CCC-0187/02 IBM 370 series IBM 370 series
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3. NATURE OF PHYSICAL PROBLEM SOLVED

The SAM-CE system comprises two Monte Carlo codes, SAM-F and SAM-A.  SAM-F supercedes the forward Monte Carlo code, SAM-C.
SAM-A is an adjoint Monte Carlo code designed to calculate the response due to fields of primary and secondary gamma radiation.
The SAM-CE system is a FORTRAN Monte Carlo computer code designed to solve the time-dependent neutron and gamma-ray transport equations in complex three-dimensional geometries.
SAM-CE is applicable for forward neutron calculations and for forward as well as adjoint primary gamma-ray calculations. In addition, SAM-CE is applicable for the gamma-ray stage of the coupled neutron-secondary gamma ray problem, which may be solved in either the forward or the adjoint mode.
Time-dependent fluxes, and flux functionals such as dose, heating, count rates, etc., are calculated as functions of energy, time and position. Multiple scoring regions are permitted and these  may be either finite volume regions or point detectors or both. Other scores of interest, e.g., collision and absorption densities,  etc., are also made.
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4. METHOD OF SOLUTION

A special feature of SAM-CE is its use of the 'combinatorial geometry' technique which affords the user geometric capabilities exceeding those available with other commonly used geometric packages.
All nuclear interaction cross section data (derived from the ENDF for neutrons and from the UNC-format library for gamma-rays) are tabulated in point energy meshes. The energy meshes for neutrons are internally derived, based on built-in convergence criteria and user- supplied tolerances. Tabulated neutron data for each distinct nuclide are in unique and appropriate energy meshes. Both resolved and unresolved resonance parameters from ENDF data files are treated automatically, and extremely precise and detailed descriptions of cross section behaviour is permitted. Such treatment avoids the ambiguities usually associated with multi-group codes, which use flux-averaged cross sections based on assumed flux distributions which may or may not be appropriate.
By use of the 'band' feature of the code, which splits cross section data into two or more energy ranges to be treated one at a time, SAM-CE affords one the ability to consider many nuclides, in a given configuration, each being described in much detail.
SAM-CE also provides the user with the opportunity to employ energy, region and angular importance sampling.
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5. RESTRICTIONS ON THE COMPLEXITY OF THE PROBLEM

Essentially no restrictions for neutron problems. For gamma-ray problems, only Compton scattering and absorption are treated.
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6. TYPICAL RUNNING TIME:  Highly problem dependent.
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7. UNUSUAL FEATURES: UNUSUAL FEATURES OF THE PROGRAM
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8. RELATED AND AUXILIARY PROGRAMS

BCDEAN - translates element data tapes produced by SAM-X from binary into BCD mode and vice versa.
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9. STATUS
Package ID Status date Status
CCC-0187/01 24-MAR-1981 Tested at NEADB
CCC-0187/02 10-OCT-1979 Tested at NEADB
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10. REFERENCES

- M. O. Cohen, et al.:
  'SAM-CE - A Three-Dimensional Monte Carlo Code for the Solution of the Forward Neutron and Forward and Adjoint Gamma Ray Transport Equations'
  MR-7021, DNA 2830F (November 1971).
CCC-0187/01, included references:
- H. Lichtenstein et al.:
  The SAM-CE Monte Carlo System for Radiation Transport and
  Criticality Calculations in Complex Configurations (Revision 7.0)
  EPRI CCM-8 (July 1979).
CCC-0187/02, included references:
- Herbert A. Steinberg et al.:
  SAM-CE - A Monte Carlo Code for Three Dimensional Neutron, Gamma
  Ray and Electron Transport (Revision 5)
  MR-7052-5 (May 1977).
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11. MACHINE REQUIREMENTS

On CDC 6600 - 210000 octal locations are needed. SAM-F and SAM-A are also operable on IBM/360. Several disc files and/or tapes are also needed.
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12. PROGRAMMING LANGUAGE(S) USED
Package ID Computer language
CCC-0187/01 FORTRAN-IV
CCC-0187/02 FORTRAN-IV
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13. SOFTWARE REQUIREMENTS: OPERATING SYSTEM OR MONITOR UNDER WHICH PROGRAM IS EXECUTED
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14. OTHER PROGRAMMING OR OPERATING INFORMATION OR RESTRICTIONS

ANY OTHER PROGRAMMING OR OPERATING INFORMATION OR RESTRICTIONS
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15. NAME AND ESTABLISHMENT OF AUTHOR:      Magi, 3 Westchester Plaza, Elmsford, New York.
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16. MATERIAL AVAILABLE
CCC-0187/01
File name File description Records
CCC0187_01.001 INFORMATION 2
CCC0187_01.002 SAMF SOURCE 15090
CCC0187_01.003 SAMF COMPLEX COMBINATORIAL GEOM 1232
CCC0187_01.004 SAMF JCL 20
CCC0187_01.005 SUBROUTINE ARG AND ARPREP 223
CCC0187_01.006 GAMMA RAY DATA PROCESSOR SOURCE (11/74) 297
CCC0187_01.007 SAMF INPUT DATA 89
CCC0187_01.008 SAMF OUTPUT 454
CCC0187_01.009 BCDEAN SOURCE 314
CCC0187_01.010 BCDEAN INPUT DATA 3
CCC0187_01.011 PHOTON CROSS SECTION 17548
CCC0187_01.012 BCDEAN JCL 19
CCC0187_01.013 BCDEAN OUTPUT 13
CCC0187_01.014 SAM-X SOURCE 10616
CCC0187_01.015 SAM-X INPUT DATA 9
CCC0187_01.016 FE CROSS SECTION ENDF/B-2 (11/74) 5293
CCC0187_01.017 SAM-X JCL 47
CCC0187_01.018 SAM-X OUTPUT 4042
CCC-0187/02
File name File description Records
CCC0187_02.001 BCDEAN SOURCE (F4,BCD) 314
CCC0187_02.002 INPUT EDT 17548
CCC0187_02.003 BCDEAN SAMPLE INPUT 3
CCC0187_02.004 BCDEAN SAMPLE OUTPUT 13
CCC0187_02.005 SAM-F SOURCE (F4,BCD) 9608
CCC0187_02.006 SAM-F OVERLAY CARDS 30
CCC0187_02.007 SAM-F SAMPLE INPUT 90
CCC0187_02.008 SAM-F SAMPLE OUTPUT 558
CCC0187_02.009 SAM-X SOURCE (F4,BCD) 10619
CCC0187_02.010 ICLOCK FUNCTION (DUMMY) 4
CCC0187_02.011 SAM-X OVERLAY CARDS 31
CCC0187_02.012 ENDF 5293
CCC0187_02.013 SAM-X SAMPLE INPUT 9
CCC0187_02.014 SAM-X SAMPLE OUTPUT 4042
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17. CATEGORIES
  • F. Space - Time Kinetics, Coupled Neutronics - Hydrodynamics - Thermodynamics
  • J. Gamma Heating and Shield Design

Keywords: Monte Carlo method, absorption, collisions, cross sections, doses, gamma radiation, neutron transport theory, shielding, three-dimensional, time dependence.