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

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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.

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.

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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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).

- 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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Package ID | Computer language |
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CCC-0187/01 | FORTRAN-IV |

CCC-0187/02 | FORTRAN-IV |

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CCC-0187/01

File name | File description | Records |
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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 |
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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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- 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.