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Monte Carlo Lectures. LA-UR-05-4983 by Forrest Brown.

Solving particle transport problems with the Monte Carlo method is simple - just simulate the particle behavior. The devil is in the details, however. This course provides a balanced approach to the theory and practice of Monte Carlo simulation codes, with lectures on transport, random number generation, random sampling, computational geometry, collision physics, tallies, statistics, eigenvalue calculations, variance reduction, and parallel algorithms. This is not a course in how to use MCNP or any other code, but rather provides in-depth coverage of the fundamental methods used in all modern Monte Carlo particle transport codes. The course content is suitable for beginners and code users, and includes much advanced material of interest to code developers. (10 lectures, 2 hrs each)

MCNP Medical Physics Input Deck DataBase (tar format)
(LA-UR-04-8518 & LA-UR-05-6921)

MCNP Medical Physics DataBase InputDecks (pdf format)

With the growing interest in using MCNP for medical physics calculations, demand has been increasing for geometric models which represent various portions of the human body. This database of analytical and voxelized (possibly based on CT data) geometries, in mcnp input deck form, helps to meet that need. They could be used for organ-specific dose calculations, code comparisons, or geometric representation studies. Neither the U.S. Government, nor Los Alamos National Security, LLC, nor Los Alamos National Laboratory, nor their respective authors makes any warranty, express or implied, or assumes any liability or responsibility for the use of these input decks.

Contributions are welcome. For more information, contact jgoorley@lanl.gov.

MCNP5 - 09/13/04 - This patch (patch-MCNP5_RSICC_1.20_to_1.30) updates the Radiation Safety Information Computational Center (RSICC) release of MCNP5 from version MCNP5_RSICC_1.20 to version MCNP5_RSICC_1.30. The issues addressed by this patch file as well as instructions on applying this patch are given in the pdf file.

MCNP5(pdf) MCNP5 (patch)

MCNP5 is available from RSICC. This is now the Standard MCNP Release.

MCNP5 - 10/23/03 - This patch (patch-MCNP5_RSICC_1.14_to_1.20) updates the Radiation Safety Information Computational Center (RSICC) release of MCNP5 from version MCNP5_RSICC_1.14 to version MCNP5_RSICC_1.20. The issues addressed by this patch file as well as instructions on applying this patch are given in the pdf file.

MCNP5(pdf) MCNP5(patch)

MCNP5 is available from RSICC. This is now the Standard MCNP Release.

Criticality Calculations with MCNP5TM: A Primer (LA-UR-04-0294)

ABSTRACT
The purpose of this primer is to assist the nuclear criticality safety analyst to perform computer calculations using the Monte Carlo code MCNP. Because of the closure of many experimental facilities, reliance on computer simulation is increasing. Often the analyst has little experience with specific codes available at his/her facility. This Primer helps the analyst understand and use the MCNP Monte Carlo code for nuclear criticality analyses. It assumes no knowledge of or particular experience with Monte Carlo codes in general or with MCNP in particular. The document begins with a QuickStart chapter that introduces the basic concepts of using MCNP. The following chapters expand on those ideas, presenting a range of problems from simple cylinders to 3-dimensional lattices for calculating keff confidence intervals. Input files and results for all problems are included. The primer can be used alone, but its best use is in conjunction with the MCNP5 manual. After completing the primer, a criticality analyst should be capable of performing and understanding a majority of the calculations that will arise in the field of nuclear criticality safety.

Monte Carlo Parameter Studies and Uncertainty Analyses with MCNP5 (LA-UR-04-04991)

A software tool called mcnp_pstudy has been developed to automate the setup, execution, and collection of results from a series of MCNP5 Monte Carlo calculations. This tool provides a convenient means of performing parameter studies, total uncertainty analyses, parallel job execution on clusters, stochastic geometry modeling, and other types of calculations where a series of MCNP5 jobs must be performed with varying problem input specifications.

Continuously Varying Material Properties and Tallies for Monte Carlo Calculations
(LA-UR-04-07321)

Using a high-order Legendre polynomial representation for material density and tallies within each cell, Monte Carlo codes can model continuous variations in material properties and results. We have demonstrated the Monte Carlo techniques for sampling the free-flight distances and performing pathlength flux tallies for this continuous representation. Application to both fixed-source and eigenvalue problems illustrates the benefits of the continuous representation as compared to conventional stepwise approximations. With these new methods, Monte Carlo codes can now be developed which are continuous in energy, angle, space, material properties, and tallied results.

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