CCC-0718 MCNP-POLIMI V1.0.
MCNP-POLIMI v1.0, Monte Carlo N-Particle Transport Code System To Simulate Time-Analysis Quantities
NAME OR DESIGNATION OF PROGRAM, COMPUTER, DESCRIPTION OF PROGRAM OR FUNCTION, METHODS, RESTRICTIONS ON THE COMPLEXITY OF THE PROBLEM, TYPICAL RUNNING TIME, FEATURES, RELATED OR AUXILIARY PROGRAMS, STATUS, REFERENCES, HARDWARE REQUIREMENTS, LANGUAGE, SOFTWARE REQUIREMENTS, OTHER RESTRICTIONS, NAME AND ESTABLISHMENT OF AUTHORS, MATERIAL, CATEGORIES
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3. DESCRIPTION OF PROGRAM OR FUNCTION
MCNP is a general-purpose, continuous-energy, generalized geometry, time-dependent, coupled neutron-photon-electron Monte Carlo transport code system. Based on the Los Alamos National Laboratory code MCNP4C (formerly distributed as CCC-700), MCNP-PoliMi was developed to simulate time-analysis quantities. In particular, the code includes the correlation between neutron interaction and the corresponding photon production. Conversely to the technique adopted by standard MCNP, MCNP PoliMi samples secondary photons according to the neutron collision type. A post-processing code, i.e. the Matlab script 'postmain', is included and can be tailored to model specific detector characteristics. These features make MCNP-PoliMi a versatile tool to simulate particle interactions and detection processes.
MCNP is a general-purpose, continuous-energy, generalized geometry, time-dependent, coupled neutron-photon-electron Monte Carlo transport code system. Based on the Los Alamos National Laboratory code MCNP4C (formerly distributed as CCC-700), MCNP-PoliMi was developed to simulate time-analysis quantities. In particular, the code includes the correlation between neutron interaction and the corresponding photon production. Conversely to the technique adopted by standard MCNP, MCNP PoliMi samples secondary photons according to the neutron collision type. A post-processing code, i.e. the Matlab script 'postmain', is included and can be tailored to model specific detector characteristics. These features make MCNP-PoliMi a versatile tool to simulate particle interactions and detection processes.
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4. METHODS
MCNP treats an arbitrary three-dimensional configuration of materials in geometric cells bounded by first- and second-degree surfaces and some special fourth-degree surfaces. For neutrons, all reactions in a particular cross-section evaluation are accounted for. Both free gas and S(alpha, beta) thermal treatments are used. Criticality sources as well as fixed and surface sources are available. For photons, the code takes account of incoherent and coherent scattering with and without electron binding effects, the possibility of fluorescent emission following photoelectric absorption, and absorption in pair production with local emission of annihilation radiation. A very general source and tally structure is available. The tallies have extensive statistical analysis of convergence. Rapid convergence is enabled by a wide variety of variance reduction methods. Energy ranges are 0-60 MeV for neutrons (data generally only available up to 20 MeV) and 1 keV - 1 GeV for photons and electrons.
The MCNP-PoliMi code was developed to simulate each neutron-nucleus interaction as closely as possible. In particular, neutron interaction and photon production are made correlated and correct neutron and photon fission multiplicities were implemented. At each neutron and gamma collision occurring inside user-defined cells, relevant information is recorded, for example reaction type, target nucleus, energy deposited, and position.
The number of photons born in induced fission is determined by sampling from a fission photon multiplicity distribution. The number of photons that can be generated by the algorithm ranges from zero to 23. Photon energy is assigned by independently sampling from an energy distribution taken from CCC-699/MCNP-DSP. In MCNP-PoliMi, a dual-particle source is generated at each induced fission. Neutrons are emitted at the same time with multiplicity and energy distributions given by appropriate distributions from MCNP-DSP. Gamma ray multiplicities and energies have also been implemented on the basis of results found in the literature. This procedure attempts to model the induced fission chains as closely as possible.
MCNP treats an arbitrary three-dimensional configuration of materials in geometric cells bounded by first- and second-degree surfaces and some special fourth-degree surfaces. For neutrons, all reactions in a particular cross-section evaluation are accounted for. Both free gas and S(alpha, beta) thermal treatments are used. Criticality sources as well as fixed and surface sources are available. For photons, the code takes account of incoherent and coherent scattering with and without electron binding effects, the possibility of fluorescent emission following photoelectric absorption, and absorption in pair production with local emission of annihilation radiation. A very general source and tally structure is available. The tallies have extensive statistical analysis of convergence. Rapid convergence is enabled by a wide variety of variance reduction methods. Energy ranges are 0-60 MeV for neutrons (data generally only available up to 20 MeV) and 1 keV - 1 GeV for photons and electrons.
The MCNP-PoliMi code was developed to simulate each neutron-nucleus interaction as closely as possible. In particular, neutron interaction and photon production are made correlated and correct neutron and photon fission multiplicities were implemented. At each neutron and gamma collision occurring inside user-defined cells, relevant information is recorded, for example reaction type, target nucleus, energy deposited, and position.
The number of photons born in induced fission is determined by sampling from a fission photon multiplicity distribution. The number of photons that can be generated by the algorithm ranges from zero to 23. Photon energy is assigned by independently sampling from an energy distribution taken from CCC-699/MCNP-DSP. In MCNP-PoliMi, a dual-particle source is generated at each induced fission. Neutrons are emitted at the same time with multiplicity and energy distributions given by appropriate distributions from MCNP-DSP. Gamma ray multiplicities and energies have also been implemented on the basis of results found in the literature. This procedure attempts to model the induced fission chains as closely as possible.
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6. TYPICAL RUNNING TIME
Running time varies greatly depending on problem parameters. The cf.inp sample problem describes a Cf-252 point source placed at one meter from a plastic scintillator detector. The detector is cell no.1, and the collision data is recorded for this detector. The program will execute this problem in 8.48 minutes on a 2.2 GHz Pentium 4.
Running time varies greatly depending on problem parameters. The cf.inp sample problem describes a Cf-252 point source placed at one meter from a plastic scintillator detector. The detector is cell no.1, and the collision data is recorded for this detector. The program will execute this problem in 8.48 minutes on a 2.2 GHz Pentium 4.
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8. RELATED OR AUXILIARY PROGRAMS
This package includes a test library of cross sections for running the sample problems, but the test library is not suitable for real problems. The user is advised to select available ENDF-based MCNP libraries (60C), or those with the most detailed photon production description for the particular problem. DLC-200 / MCNPDATA (or equivalent) is recommended for use with MCNP-PoliMi.
This package includes a test library of cross sections for running the sample problems, but the test library is not suitable for real problems. The user is advised to select available ENDF-based MCNP libraries (60C), or those with the most detailed photon production description for the particular problem. DLC-200 / MCNPDATA (or equivalent) is recommended for use with MCNP-PoliMi.
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10. REFERENCES
- S. A. Pozzi, E. Padovani, and M. Marseguerra:
MCNP-PoliMi: a Monte Carlo code for correlation measurements, Nuclear Instruments and Methods in Physics Research (Section A), Volume 513, Issue 3, pages 550-558 (November 2003)
- M. Marseguerra, E. Padovani, S. A.Pozzi, M. Da Ros:
Phenomenological simulation of detector response for safeguards experiments, Nuclear Instruments and Methods in Physics Research, (Section B), Volume 213 pages 289-293 (January 2004)
- S. A. Pozzi, E. Padovani, J.K. Mattingly, and J.T. Mihalczo:
MCNP-PoliMi Evaluation of Time Dependent Coincidence between Detectors for Fissile Metal Vs. Oxide Determination, Institute of Nuclear Materials Management 43rd annual meeting, Orlando, Florida (June 23-27, 2002)
- S. A. Pozzi, J.K. Mattingly, J.T. Mihalczo and E. Padovani:
Validation of the MCNP-PoliMi code for the simulation of nuclear safeguards experiments on uranium and plutonium metal, Nuclear Mathematical and Computational Sciences M&C2003, April 6-11, 2003, Gatlinburg, Tennessee, USA
- S. A. Pozzi and J.T. Mihalczo:
Monte Carlo Evaluation of the Improvements in Nuclear Materials Identification System (NMIS) resulting from a DT Neutron Generator, Institute of Nuclear Materials Management 43rd annual meeting, June 23-27, 2002, Orlando, Florida, USA
- S. A. Pozzi, E. Padovani, and M. Marseguerra:
MCNP-PoliMi: a Monte Carlo code for correlation measurements, Nuclear Instruments and Methods in Physics Research (Section A), Volume 513, Issue 3, pages 550-558 (November 2003)
- M. Marseguerra, E. Padovani, S. A.Pozzi, M. Da Ros:
Phenomenological simulation of detector response for safeguards experiments, Nuclear Instruments and Methods in Physics Research, (Section B), Volume 213 pages 289-293 (January 2004)
- S. A. Pozzi, E. Padovani, J.K. Mattingly, and J.T. Mihalczo:
MCNP-PoliMi Evaluation of Time Dependent Coincidence between Detectors for Fissile Metal Vs. Oxide Determination, Institute of Nuclear Materials Management 43rd annual meeting, Orlando, Florida (June 23-27, 2002)
- S. A. Pozzi, J.K. Mattingly, J.T. Mihalczo and E. Padovani:
Validation of the MCNP-PoliMi code for the simulation of nuclear safeguards experiments on uranium and plutonium metal, Nuclear Mathematical and Computational Sciences M&C2003, April 6-11, 2003, Gatlinburg, Tennessee, USA
- S. A. Pozzi and J.T. Mihalczo:
Monte Carlo Evaluation of the Improvements in Nuclear Materials Identification System (NMIS) resulting from a DT Neutron Generator, Institute of Nuclear Materials Management 43rd annual meeting, June 23-27, 2002, Orlando, Florida, USA
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13. SOFTWARE REQUIREMENTS
The included executable (file Pcexe\Main.exe) was created with Compaq Visual Fortran Professional Edition 6.6 on a Pentium IV under Windows XP Professional. The developers also tested the code under WINDOWS 98 SE on INTEL Pentium III, under WINDOWS XP Home on AMD Athlon XP, and under Windows 2000 Pro on Pentium 4. The MDAS parameter inside the code was set to 16,000,000 (16 Mwords). The plot feature is not included.
The included executable (file Pcexe\Main.exe) was created with Compaq Visual Fortran Professional Edition 6.6 on a Pentium IV under Windows XP Professional. The developers also tested the code under WINDOWS 98 SE on INTEL Pentium III, under WINDOWS XP Home on AMD Athlon XP, and under Windows 2000 Pro on Pentium 4. The MDAS parameter inside the code was set to 16,000,000 (16 Mwords). The plot feature is not included.
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Keywords: Monte Carlo method, coupled neutron gamma cross sections, neutron, photons, scintillation detectors, shielding, time analysis.