What is MCNP?

MCNP (Monte Carlo N-Particle) tracks neutrons, photons, electrons, and other particles through three-dimensional geometry. You describe the model with text cards, name a source and materials, and the code follows individual histories from birth to termination, sampling at every interaction. The average over many histories is the answer — flux, reaction rate, dose, or k-effective — with a quantified uncertainty.

Every input file has three sections. Cell cards define the physical regions and what fills them. Surface cards describe the geometric boundaries that shape those regions. Data cards supply materials, nuclear data, the source, and tallies. This path walks through all three in order, from simple complete decks toward assemblies and shielding.

Common uses include reactor core physics and criticality safety, shielding and dose, medical scenarios where electron and photon transport matter, and detection or space environment estimates. Licensing and design work often lean on it when the geometry or continuous-energy physics would fight a mesh-based deterministic code.

Each history starts at the source and ends when the particle is absorbed, escapes, or is killed by variance reduction. At every step MCNP samples the distance to the next collision and the outcome — scatter, absorb, fission, or escape. Millions of such histories build a statistically meaningful result without a spatial mesh or multi-group energy structure.

Two problem types cover most of the work. A fixed-source run has a known source in energy, position, and direction: the SDEF card defines it and NPS sets how many histories to run. Shielding, dose, and detector response usually look like this. A criticality (k-eigenvalue) run iterates neutron generations until the fission source converges; KCODE sets particles per cycle, inactive cycles, and total cycles, and KSRC places the initial source points.

MCNP transports neutrons, photons, and electrons — and combinations via the MODE card. Interaction probabilities come from evaluated nuclear data (ENDF) processed into continuous-energy ACE files. The ZAID extension identifies the library and temperature together: in the files LANL ships with MCNP6, .80c is ENDF/B-VII.1 at room temperature and .83c is the same evaluation at 1200 K, while ENDF/B-VIII.0 lives at .00c–.06c. The resemblance between .80c and "VIII.0" is a coincidence. Bound thermal scattering needs S(α,β) data on MT cards. For library selection in more detail, see Nuclear Data & Cross Sections.

Length is centimetres, energy is MeV, and time is shakes (10⁻⁸ s), and there is no card that changes any of them. The shake is the one that surprises people: a 1 MeV neutron travels about 14 cm in a shake, so time bins written as though they were seconds are wrong by eight orders of magnitude and still run without complaint.

Densities on cell cards carry the meaning in their sign. Negative means g/cm³, positive means atoms/barn·cm — and because both are legal, writing the wrong one produces a model that runs to completion and reports a confident wrong answer. The same logic applies to material fractions later on, where negative means weight fraction and positive means atom fraction.

Coordinates are right-handed Cartesian throughout. Cylindrical and spherical surface types (cz, so, and the rest) exist, but their parameters are still given in Cartesian terms — there is no (r, θ, z) input mode.

MCNP geometry is constructive solid geometry: surfaces divide space, and Boolean combinations of sense-signed surfaces define cells. The card below has the five fields every filled cell needs — hover each highlighted token for a field-by-field explanation.

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

A surface is a mathematical boundary with a positive and a negative side; the signed numbers on the cell card pick which side the cell occupies. Common types are planes (px, py, pz), spheres (so), and cylinders (cz). A universe is a reusable geometry template — define it once and place it with fill= — which is how repeated structures such as fuel assemblies stay manageable. Cell cards are covered in full on Cell Cards.

Every isotope on a material card is a ZAID in the form ZZAAA.nnX: ZZ is the atomic number, AAA the mass number, nn the library identifier, and X the data class (c for continuous-energy neutrons). Hover either line below for a field-by-field breakdown.

c   ZAID.XXc  fraction
m1  92235.70c  0.045   $ U-235, 4.5% enrichment (atom fraction)
    92238.70c  0.955   $ U-238 (remainder)

A cross section is the effective target area for an interaction, in barns (10⁻²⁴ cm²); the macroscopic cross section Σ = Nσ is the interaction probability per unit path length. Below roughly thermal energies in bound materials such as water or graphite, free-gas scattering is not enough — an MT card attaches an S(α,β) table (for light water at room temperature on this path, lwtr.10t). Material cards and library choice are the subject of Material Cards.

Tallies extract quantities from the run — flux, current, dose, or reaction rates. MCNP reports a relative error with each result; values below 0.05 are generally treated as reliable, and ten built-in statistical tests check convergence. Cell importance (imp:n) and weight windows steer the particle population toward regions that matter to the tally.

The tally number chooses both the quantity and where it is scored. The five you will meet first are f1 for current across a surface, f2 for flux averaged over a surface, f4 for flux averaged over a cell volume, f5 for flux at a point, and f6 for energy deposition. The distinction between current and flux is not cosmetic: current is directional and flux is not, which is why angular binning is available on f1 and on nothing else.

The running example is a PWR pin cell: UO₂ fuel at 4.5% enrichment on library .70c, a helium gap, zirconium cladding, and light water with lwtr.10t. The full deck — cells, surfaces, materials, and a KCODE criticality setup — is assembled on Example: Pin Cell. The pages between here and there teach each card type in the order you will need it.

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 documents below, and download the approved package — processing typically takes 1–2 weeks.

• Export Control Form — signed compliance statement
• Reference Letter — from advisor or supervisor
• Institutional Proof — student ID or employment letter
• Research Statement — brief description of intended use

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