MCNP6/X Tutorial & Guide

Master MCNP - The industry-standard Monte Carlo code for nuclear and radiation transport simulations. Free comprehensive tutorials for MCNP6 and MCNPX with practical examples.

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Last Updated: February 2026

Why engineers choose MCNP

MCNP (Monte Carlo N-Particle) is the reference code for criticality safety, shielding, and advanced reactor analysis. Maintained by Los Alamos National Laboratory for more than 60 years, it underpins licensing packages, research programs, and regulatory guidance across the globe.

Regulatory Credibility

Accepted worldwide for licensing, criticality safety, and shielding analyses with decades of benchmark validation.

Coupled Physics

Supports multi-particle (n, p, e, γ) transport with variance-reduction controls for challenging shielding cases.

Extensive Geometry Library

Robust constructive solid geometry (CSG) surfaces plus repeated structures for lattices, universes, and hierarchical models.

Rich Nuclear Data

Ships with multiple ENDF/B evaluations, S(α,β) treatments, photon libraries, and easy hooks for custom ACE data.

Quick start guides

Installation Guide

Set up MCNP on Windows, Linux, and macOS with RSICC packages.

MCNP Basics

Understand cards, decks, and the Monte Carlo workflow.

Geometry & Cells

Master cell and surface cards for complex models.

Material & Data Cards

Define compositions, densities, and cross-section references.

Sample MCNP Input File

Below is a complete MCNP input file for a simple fuel pin cell model. This example demonstrates the three main sections: Cell Cards, Surface Cards, and Data Cards.

mcnp
c Simple Pin Cell Model
c Cell Cards
1 1 -10.4 -1     imp:n=1  $ Fuel
2 2 -6.55 1 -2   imp:n=1  $ Cladding
3 3 -1.0  2 -3   imp:n=1  $ Moderator
4 0       3      imp:n=0  $ Outside world

c Surface Cards
1 cz 0.4096  $ Fuel radius
2 cz 0.4750  $ Cladding outer radius
3 cz 0.6617  $ Cell boundary radius

c Data Cards
m1   92235.70c 0.05   $ UO2 Fuel
     92238.70c 0.95
     8016.70c  2.0
m2   40000.70c 1.0    $ Zirconium Cladding
m3   1001.70c  2.0    $ Water Moderator
     8016.70c  1.0
kcode 1000 1.0 20 100
ksrc 0 0 0

Common Modifications

  • Adjust enrichment (U-235/U-238 ratio)
  • Modify pin dimensions
  • Add burnable poisons
  • Change boundary conditions

Key Features Shown

  • Cell definitions
  • Material specifications
  • Criticality calculation setup
  • Geometry modeling

How to Read the Cell Cards

  • Column 1: Cell ID. This is the number you reference later (e.g., universe fills).
  • Material & density: "1 -10.4" means material 1 at -10.4 g/cm³ (negative = mass density).
  • Surface logic: "-1" means the particle is inside surface 1. A positive index (e.g., "1") indicates outside.
  • Importance: imp:n=1 keeps neutrons alive in that region, while imp:n=0 kills them (used for vacuum boundaries).

Surface & Data Highlights

  • CZ surfaces: Cylinders aligned with the z-axis. Surface 1 is the fuel radius, 2 is the cladding.
  • Material cards: m1 lists nuclides with fractions. Make sure the sum of atom fractions matches your stoichiometry.
  • KCODE: kcode 1000 1.0 20 100 requests 1000 neutrons per cycle, an initial guess of 1.0, 20 inactive, 100 active cycles.
  • KSRC: Defines the initial source position. Here it's centered at (0,0,0).

Nuclear data tips

MCNP relies on evaluated nuclear data files (ENDF/B, JEFF, JENDL, ENDF/VIII) that are processed for specific temperatures and moderators. The suffix on each isotope (for example .70c) tells reviewers which evaluation and temperature you selected.

Match the library to your scenario: .72c or .74c for hot fuel, S(α,β) treatments for light-water moderators, or specialized photon libraries for detector studies. Accurate nuclear data choices often improve agreement with benchmarks more than any geometry tweak.

Document the library selections in comments so future reviewers know the temperature, evaluation, and processing basis for each material.

Frequently Asked Questions

How do I get access to MCNP?

MCNP is export-controlled software available through RSICC. Academic institutions can request access through their organization. Commercial licenses are also available.

Which MCNP version should I use?

MCNP6.3 is recommended for most users as the latest stable release. MCNPX is preferred for certain particle physics applications, while MCNP5 remains widely used in established workflows.

What are the system requirements?

MCNP runs on Windows, Linux, and Mac OS. Minimum requirements include 4GB RAM and a modern multi-core processor. GPU acceleration is available in newer versions.

Advanced capabilities

Variance Reduction

Leverage weight windows, source biasing, and DXTRAN spheres to focus computing power on tallies that matter.

Criticality & Safety

Execute k-effective studies, shutdown margin checks, and criticality alarm evaluations with high confidence.

MCNP applications

Reactor Physics

  • Fuel assembly design and lattice physics
  • Control rod worth & shutdown margin
  • Flux/power distribution benchmarking
  • Burnup swing and reactivity effects

Radiation Protection

  • Shielding optimization and dose mapping
  • Spent fuel cask analysis
  • Detector response and calibration
  • Accelerator and medical shielding studies

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