4 Running the Program The first thing that I will mention is that the User’s Manuals are extremely helpful. These manuals have loads of information in them and can help you through the entire process that this tutorial takes you through.
We have added a source definition component to the Moritz geometry editing and display program. Gamma line emission can be imported from a library based on the Brookhaven National Laboratory Nuclear Data Center. The user can select one or more isotopes from a tree view list. The abundances can be corrected for radioactive decay between two times. The line emission is then converted to MCNP SDEF format. Tabulated data, such as an X-ray tube spectrum, can be read and converted to MCNP format. Interactive tools are available for defining source volumes, direction, cone source opening angle, and bias direction; these items can be shown together with the geometry. All source definition items can be entered exactly in dialog fields. Source definitions can be read from existing input files. The tools will expedite and verify source definition and ensure accuracy and do not require knowledge of the MCNP SDEF syntax.
- View MCNP5manualVOLI.pdf from NUCLEAR 22.615 at Massachusetts Institute of Technology. LA-UR-03-1987 Approved for public release; distribution is unlimited Title: MCNP A.
- Volume II: MCNP User's Guide Chapters 1, 3, 4, 5 and Appendices A, B, I, J, K Volume II(LA-CP-03-0245) provides detailed specifications for MCNP5 input and options, numerous example problems, and a discussion of the output generated by MCNP5. The first chapter is a primer on basic MCNP5 use.
| Developer(s) | LANL |
|---|---|
| Stable release | |
| Written in | Fortran 90 |
| Operating system | Cross-platform |
| Type | Computational physics |
| License | MCNPX Single-User Software License (proprietary) |
| Website | mcnp.lanl.gov |
Monte Carlo N-Particle Transport Code (MCNP) is a software package for simulating nuclear processes. It is developed by Los Alamos National Laboratory since at least 1957[2] with several further major improvements. It is distributed within the United States by the Radiation Safety Information Computational Center in Oak Ridge, TN and internationally by the Nuclear Energy Agency in Paris, France. It is used primarily for the simulation of nuclear processes, such as fission, but has the capability to simulate particle interactions involving neutrons, photons, and electrons among other particles. 'Specific areas of application include, but are not limited to, radiation protection and dosimetry, radiation shielding, radiography, medical physics, nuclear criticality safety, detector design and analysis, nuclear oil well logging, accelerator target design, fission and fusion reactor design, decontamination and decommissioning.'
Monte Carlo N-Particle eXtended[edit]
Monte Carlo N-Particle eXtended (MCNPX) was also developed at Los Alamos National Laboratory, and is capable of simulating particle interactions of 34 different types of particles (nucleons and ions) and 2000+ heavy ions at nearly all energies,[3] including those simulated by MCNP.
Mcnp5 Manual
Both codes can be used to judge whether or not nuclear systems are critical and to determine doses from sources, among other things.
MCNP6 is a merger of MCNP5 and MCNPX.[3]
See also[edit]
Notes[edit]
Mcnp6 User Manual Pdf
- ^'MCNP6.2 Release notes'(PDF). LANL. 2018-02-05. Retrieved 2018-02-15.
- ^Cashwell, E.D.; Everett, C.J. (1959). A Practical Manual on the Monte Carlo Method for Random Walk Problems(PDF). London: Pergamon Press.
- ^ abJames, M.R. 'MCNPX 2.7.x - New Features Being Developed'(PDF).
External links[edit]
