Comprehensive MCNP Monte Carlo code training for particle transport simulation and dosimetry calculations in medical physics and nuclear engineering. Includes input file creation, geometry definition, sources, tallies and output results analysis for research and industrial applications
Following our exploration of various mesh tallies in MCNP, this article provides a specialized examination of the Type 3 mesh tally for energy deposition. Unlike the standard F4: tally, which measures particle flux, this tally...
VVER 1000 reactor core was simulated with MCNPX and the input file was prepared for sequence usage in the project. VVER reactor or WWER reactor (abbreviated as Water-Water Energetic Reactor) is a series of light...
The Monte Carlo method is an advanced computational technique that uses random numbers to simulate complex physical systems. Due to its ability to model the probabilistic nature of nuclear interactions, it is widely used in...
A Comprehensive Guide to the MCNP Source Card (SDEF): Definitions, Defaults, and Practical Examples Introduction In Monte Carlo N-Particle (MCNP) simulations, accurately defining the source of particles—the starting point for their random journeys—is paramount. The...
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