MCNP Guide
MCNP Terminology
Essential terms for understanding MCNP
Geometry
MCNP uses Constructive Solid Geometry (CSG): surfaces divide space into regions, and Boolean combinations of those regions define cells.
Cell
A volume of space filled with material. Cells are defined by surfaces and contain all the information MCNP needs for transport.
Hover over each highlighted field to see what it means.
1 1 -2.7 -10 11 -12 imp:n=1 $ Aluminum cylinder
MCNP Cell Card
Hover over each highlighted field to see what it means.
Surface
Mathematical boundary that divides space. Surfaces have positive and negative sides that define cell boundaries.
Universe
Reusable geometry template. Define once, use many times. Essential for modeling repeated structures like fuel assemblies.
Materials and Nuclear Data
ZAID Numbers
MCNP's system for identifying isotopes and cross-section libraries. Format: ZZAAA.nnX
AAA: Mass number (235 = U-235)
X: Data type (c = continuous energy)
c ZAID.XXc fraction
m1 92235.70c 0.05 $ U-235, 5% enrichment (atom fraction)
92238.70c 0.95 $ U-238 (remainder)Cross Section
An effective target area for nuclear interactions, measured in barns (10⁻²⁴ cm²). The macroscopic cross section (Σ = Nσ) gives probability of interaction per unit path length.
Thermal Treatment
Special handling for low-energy neutrons in materials like water. Uses mt card (e.g., mt1 lwtr.10t).
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 considered reliable, and 10 built-in statistical tests verify convergence. Cell importance (imp:n) and weight windows are the primary variance reduction tools: they focus particle population in regions that matter most to the tally result.
Common tally types
- F1 — surface current
- F2 — surface flux
- F4 — cell flux (volume-averaged)
- F5 — point detector flux
- F6 — energy deposition (heating)