1. NAME AND TITLE
SAM-CE: Monte Carlo Time-Dependent Complex Geometry (Combinatorial) Code System for the Solution of the Forward Neutron and Forward and Adjoint Gamma-Ray Transport Equations.
BCDEAN: Data Generator.
SAMX: Data Generator.
Libraries of Data for use in SAM-CE must be requested separately. They are available from RSIC as DLC-31 (DPL-400/GETD1, DPL-401/NEDT, and DPL-402/GPDT1).
SAM-CE follows a series of Monte Carlo codes, beginning with ADONIS and evolving through an UNC-SAM series at United Nuclear Corporation and continuing in the SAM series at MAGI. The Combinatorial Geometry technique, currently in wide usage, originated at MAGI and was incorporated in the SAM series. This package represents the last frozen version placed in RSIC.
Mathematical Applications Group, Inc., Elmsford, New York.
3. CODING LANGUAGE AND COMPUTER
Revision 5: FORTRAN IV and Assembler language; IBM 360/370 (A).
Revision 7: FORTRAN IV; CDC 6600 (B).
4. NATURE OF PROBLEM SOLVED
SAM-CE is designed to solve time-dependent neutron and gamma-ray transport equations in complex three-dimensional geometries and 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 also 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 are also made.
5. METHOD OF SOLUTION
A special feature of SAM-CE is its use of the Combinatorial Geometry technique which affords the user a variety of geometric capabilities.
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 behavior 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 the capability of considering many nuclides in a given configuration, each being described in much detail.
SAM-CE also provides the user with opportunity to employ energy, region and angular importance sampling.
6. RESTRICTIONS OR LIMITATIONS
There are essentially no restrictions for neutron problems. For gamma-ray problems, only Compton scattering and absorption are treated.
7. TYPICAL RUNNING TIME
Running time is highly problem dependent. The sample problem ran on the IBM 360/91 computer in 1 to 2 minutes.
8. COMPUTER HARDWARE REQUIREMENTS
Version A is operable on the IBM 360/370 computers with standard I/O plus 10 units of storage. The GO step uses 246K storage in overlay.
Version B is operable on the CDC 6600 computer.
9. COMPUTER SOFTWARE REQUIREMENTS
A FORTRAN IV compiler and an assembler are required. A library of ENDF/B cross sections is required (Fe cross sections are included in the code package). A non-standard ICLOCK routine is included in the package.
M. O. Cohen, W. Guber, E. Troubetzkoy, H. Lichtenstein, H. Steinberg and M. Beer, "SAM-CE: A Three-Dimensional Monte Carlo Code for the Solution of the Forward Neutron and Forward and Adjoint Gamma-Ray Transport EquationsRevision B," DNA-2830F-B (August 1973).
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-2930F (November 1971).
11. CONTENTS OF CODE PACKAGE
Included are the referenced documents and one (1.2MB) DOS diskette which contains the source code and sample problem input and output.
12. DATE OF ABSTRACT
August 1975; reviewed June 1984.
KEYWORDS: MONTE CARLO; NEUTRON; GAMMA-RAY; COMPLEX GEOMETRY; COMBINATORIAL GEOMETRY; TIME-DEPENDENT; ENDF/B FORMAT; ADJOINT