What is MCNP?

Introduction to MCNP

Key capabilities

MCNP is the industry-standard Monte Carlo radiation transport code used worldwide in reactor design, shielding analysis, criticality safety, and medical physics. If you are new to MCNP, here is the core idea: you describe a 3D model using text-based input cards, specify a particle source and materials, and MCNP simulates millions of individual particle histories — each neutron or photon is tracked from birth to termination, with random sampling at every interaction. The statistical average across all histories gives you fluxes, reaction rates, dose, or k-effective, each with a quantified uncertainty.

Every MCNP input file has three sections. Cell cards define the physical regions in your model and what fills them. Surface cards describe the geometric boundaries (planes, cylinders, spheres) that shape those regions. Data cards provide materials, nuclear data, source definitions, and instructions on what to measure (tallies). This guide walks through all three sections in order, starting with the simplest complete examples and building toward reactor assemblies and shielding problems.

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.

Basic Concepts

How MCNP simulates particles

MCNP tracks individual particle histories from source to termination using random sampling at each interaction — scatter, absorb, fission, or escape. Running millions of histories builds a statistically meaningful result with quantified uncertainty. This approach handles complex geometry and continuous-energy physics without the mesh or group-structure approximations that deterministic codes require.

Units and coordinate system

Problem types

Particles and nuclear data

MCNP transports neutrons, photons, and electrons — and combinations via the MODE card. Interaction probabilities come from evaluated nuclear data libraries (ENDF) processed into the continuous-energy ACE format. The ZAID extension identifies the library and temperature: .80c is ENDF/B-VIII.0 at room temperature, .71c is ENDF/B-VII.1. Thermal neutron scattering in bound materials requires S(α,β) data via MT cards.

For details on library extensions and cross-section selection, see Nuclear Data & Cross Sections.

Key Terminology

Geometry

MCNP uses Constructive Solid Geometry (CSG): surfaces divide space into regions, and Boolean combinations of those regions define cells. The cell card below shows the five fields every cell must have — hover each highlighted token for a field-by-field explanation.

1  1  -2.7  -10 11 -12  imp:n=1  $ Aluminum cylinder

Materials and nuclear data

c   ZAID.XXc  fraction
m1  92235.70c  0.05    $ U-235, 5% enrichment (atom fraction)
    92238.70c  0.95    $ U-238 (remainder)

Tallies and variance reduction

Tallies extract quantities from the simulation — flux, current, dose, or reaction rates. MCNP reports a relative error for each tally; values below 0.05 are generally reliable, and 10 built-in statistical tests verify convergence. Cell importance (imp:n) and weight windows focus particle population in regions that matter most to the tally result.

Installing MCNP

Obtaining MCNP

MCNP is export-controlled and distributed by the Radiation Safety Information Computational Center (RSICC) at Oak Ridge National Laboratory. Register at rsicc.ornl.gov with an institutional email, submit the required documents below, and download the approved package — processing typically takes 1–2 weeks.

Install guides

Select your operating system for the full step-by-step guide, including system requirements, build instructions, cross-section data setup, and a verification test.