Terminology - MCNP

Core Transport Concepts

MCNP simulates particle transport through matter. Understanding these fundamental concepts is essential for creating effective models.

History

One complete particle simulation from birth to death. Includes the original particle and all secondary particles it creates.

Example: Neutron → scatters → creates gamma → both absorbed

Track

Path segment between collision points. Each history contains many tracks as the particle moves through the geometry.

Properties: position, direction, energy, time

Geometry Building Blocks

MCNP uses Constructive Solid Geometry (CSG) to build complex models from simple mathematical surfaces.

Cell

A volume of space filled with material. Cells are defined by surfaces and contain all the information MCNP needs for transport.

mcnp

1  1  -2.7  -10 11 -12  imp:n=1  $ Aluminum cylinder
c  |  |   |    |  |  |    |
c  |  |   |    |  |  |    +-- Importance
c  |  |   |    |  |  +-- Surface list
c  |  |   |    +-- Defines geometry
c  |  |   +-- Density (g/cm³)
c  |  +-- Material number
c  +-- Cell number

Surface

Mathematical boundary that divides space. Surfaces have positive and negative sides that define cell boundaries.

Common types: planes (px, py, pz), spheres (so), cylinders (cz)

Universe

Reusable geometry template. Define once, use many times. Essential for modeling repeated structures like fuel assemblies.

Usage: fill=1 (places universe 1 in cell)

Materials and Nuclear Data

ZAID Numbers

MCNP's system for identifying isotopes and cross-section libraries. Format: ZZAAA.nnX

ZZ: Atomic number (92 = Uranium)
AAA: Mass number (235 = U-235)

nn: Library ID (70 = ENDF/B-VII.0)
X: Data type (c = continuous energy)

mcnp

m1  92235.70c  0.05    $ U-235, 5% abundance
    92238.70c  0.95    $ U-238, 95% abundance

Cross Section

Probability of interaction per unit path length. Higher cross sections mean more likely interactions.

Thermal Treatment

Special handling for low-energy neutrons in materials like water. Uses mt card (e.g., mt1 lwtr.10t).

Tallies and Results

Tallies extract useful information from the simulation. They calculate quantities like flux, dose, or reaction rates.

Tally Types

F1: Surface current
F2: Surface flux
F4: Cell flux
F5: Point detector
F6: Energy deposition

Statistical Quality

MCNP provides relative error estimates. Good results typically have relative errors < 0.05.

10 statistical tests verify result quality

Variance Reduction

Techniques to improve simulation efficiency by focusing computational effort where it's most needed.

Importance

Cell parameter that controls particle population. Higher importance means more particles in that region.

Weight Windows

Advanced population control that maintains particle weights within specified ranges throughout the geometry.

Common Input Cards

MCNP input consists of cards that specify geometry, materials, sources, and tallies. Here are the most important ones.

Geometry Cards

Cell cards: Define volumes
Surface cards: Define boundaries
Transform cards: Rotate/translate

Physics Cards

M cards: Material definitions
SDEF: Source definition
MODE: Particle types

Quick Reference

Don't memorize everything at once. Focus on cells, surfaces, and materials first. The other concepts will make more sense as you build your first models.

MCNP Guide

MCNP Terminology