Introduction to MCNP

What is MCNP?

MCNP (Monte Carlo N-Particle) is a widely used radiation transport code from Los Alamos National Laboratory. It follows neutrons, photons, electrons, and other particles through three-dimensional models—valuable for reactor work, shielding, criticality, medical physics, and detection systems.

It uses statistical particle tracking (Monte Carlo) rather than solving deterministic transport equations on a mesh, which suits complex CSG geometry and continuous-energy nuclear data.

Development history

Monte Carlo transport at Los Alamos dates to early weapons-era work; the lineage became the MCNP family. Major releases added electron transport (MCNP4-era), parallel options (MCNP5), and the merged multi-particle feature set in MCNP6.

Key capabilities

Continuous-energy data, thermal scattering, tallies for flux and reaction rates, repeated structures via universes and lattices, and variance reduction for deep penetration or rare events. Geometry is built from surfaces and Boolean cell definitions—the tutorials here walk through cards and patterns step by step.

Applications

• Nuclear power: core physics, shielding, criticality, fuel and assembly analysis; widely used in licensing support.
• Medical physics: therapy and device scenarios where electron and photon transport matter.
• Security & space: detection concepts, shielding, and environment dose estimates.

Getting started

You will need access to MCNP and evaluated nuclear data through appropriate channels (export control applies). This guide follows the sidebar order: installation and terminology, then input structure, examples, and advanced topics.

Related

• Installation — get MCNP running
• First input file — cell, surface, data cards
• Pin cell example — kcode, materials, geometry
• Code comparison — MCNP vs OpenMC vs SERPENT