Monte Carlo N-Particle Transport (MCNP)[2] is a general-purpose, continuous-energy, generalized-geometry, time-dependent, Monte Carlo radiation transport code designed to track many particle types over broad ranges of energies and is developed by Los Alamos National Laboratory. 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. The code treats an arbitrary three-dimensional configuration of materials in geometric cells bounded by first- and second-degree surfaces and fourth-degree elliptical tori.

MCNP6 is available remotely through the CAE linux lab machines. 

If the software is not graphical, use SSH https://kb.wisc.edu/cae/80385

If the software is graphical, use ssvnc… There are some limitations we are unable to solve at this time https://kb.wisc.edu/cae/20309