MCNP5 is a general-purpose Monte Carlo N-Particle code that can be used for neutron, photon, electron, or coupled neutron/photon/electron transport, including the capability to calculate eigenvalues for critical systems. It is a widely used tool for nuclear engineering and radiation physics applications.
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If you want to download MCNP5 for free, you have a few options. One option is to request the code package from the Radiation Safety Information Computational Center (RSICC) at Oak Ridge National Laboratory. RSICC is the official distributor of MCNP5 and other related codes and data libraries. You can visit their website[^1^] and fill out an online registration form to request the code package. However, you may need to meet certain eligibility criteria and export control regulations to obtain the code. RSICC offers two versions of the code package: one with Fortran 90 and C source code and executables for Windows PCs, Linux PCs, and MacOSX (Package ID: C00740MNYCP08), and one with executables only (no source code) for Windows PCs, Linux PCs, Macintosh with MacOSX (Package ID: C00740MNYCP09).
Another option is to download MCNP6.1/MCNP5/MCNPX from the Research Institute of Science and Technology (RIST) in Japan. RIST is an authorized distributor of MCNP codes and data libraries for Japanese users. You can visit their website[^2^] and fill out an online registration form to request the code package. However, you may need to meet certain eligibility criteria and export control regulations to obtain the code. RIST offers two versions of the code package: one with Fortran 90 and C source code and executables for Windows PCs, Linux PCs, and MacOSX (Package ID: C00810MNYCP00), and one with executables only (no source code) for Windows PCs, Linux PCs, Macintosh with MacOSX (Package ID: C00810MNYCP01). Note that MCNP6.1 is the latest version of MCNP that combines the features of MCNP5 and MCNPX.
A third option is to use the Visual Editor for MCNPX, which is a graphical user interface that allows you to create and edit MCNP input files, run MCNP simulations, and visualize the results. The Visual Editor is developed by Randy Schwarz at Schwarz Nuclear Consulting LLC. You can visit his website[^3^] and download the latest version of the Visual Editor for MCNPX (visedX_25.exe) or MCNP6.1 (visplot61_25.exe). However, you will need to have a valid license of MCNPX or MCNP6.1 installed on your computer to use the Visual Editor.
These are some of the ways you can download MCNP5 for free. However, you should always check the license terms and conditions before using any software. MCNP5 is a proprietary code that belongs to Los Alamos National Laboratory and is subject to U.S. government regulations. You should use it only for authorized purposes and respect the intellectual property rights of the developers.
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If you are interested in the latest version of MCNP, you may want to check out MCNP6.1, which is the result of merging MCNP5 and MCNPX capabilities. MCNP6.1 can transport 37 different particle types over broad ranges of energies and includes model physics options for energies above the cross-section table range, a material burnup feature, and delayed particle production. MCNP6.1 also has expanded and/or new tally, source, and variance-reduction options as well as an improved plotting capability[^4^].
Some of the new features of MCNP6.1 include:
- Adjoint-weighted tallies for point kinetics parameters
- Mesh tallies for isotopic reaction rates
- Greatly increased limits for geometry, tally, and source specifications
- Web-based documentation
- Additional test suites for MCNP6.1
- Modifications to the regression test suite
- Modifications to criticality test suites
- Modifications to shielding validation suite
- Enhancements to the merge_mctal and merge_meshtal utilities
- MCNP6.1 build system and directory structure
- Continue runs may use different number of threads
- RAND card allowed in continue run
- The Cascade-Exciton Model (CEM) of nuclear reactions as implemented in the event-generator CEM03.03, which is used as a default choice to calculate photonuclear reactions at energies up to 1.2 GeV, reactions induced by nucleons and pions at energies up to 5 GeV, and spallation reactions at energies up to 10 GeV
- The ability to embed unstructured meshes in a constructive solid geometry (CSG) universe to form a hybrid geometry, which allows for highly complex models to be created more easily with state-of-the-art computer-aided engineering (CAE) tools and imported into MCNP once an unstructured mesh representation has been created
You can download MCNP6.1 from RSICC or RIST as mentioned above, or you can visit the official MCNP website for more information and updates. 8cf37b1e13