Nuclear Fuel Cross Sections and Monte Carlo Simulation - Vocabulary

Vocabulary flashcards covering key terms related to nuclear fuel cross sections, enrichment, and Monte Carlo simulations.

Cross section

A measure of the probability that a neutron will interact with a nucleus per unit path length, typically expressed in barns.

Microscopic cross section

The cross section for interactions with a single nucleus, not including number density; units are barns.

formula for macroscopic cross section (∑)

∑ = $N\sigma$

Sigma ($\sigma$ )

The symbol used to denote a cross section; commonly refers to the microscopic cross section for nuclear reactions.

Barn

A unit of cross section equal to 10^-28 square meters.

Natural uranium

Uranium as found in nature, consisting mainly of U-238 (~99.3%) and a small fraction of U-235 (~0.7%).

Enrichment

The fraction of a specified fissile isotope (e.g., U-235) in uranium; expressed as a decimal or percent.

Enrichment fraction (e)

The fraction of U-235 in the fuel; for natural uranium typically e ≈ 0.007 (0.7%).

U-235

Uranium-235, a fissile isotope capable of sustaining a nuclear chain reaction with thermal neutrons.

U-238

Uranium-238, a fertile isotope that can be converted to fissile material (e.g., Pu-239) after neutron capture.

Weighted cross section for natural uranium

The effective microscopic cross section for natural uranium: σnatural = e σ235 + (1 − e) σ_238.

Fissile

An isotope that can sustain a chain reaction with thermal neutrons (e.g., U-235, Pu-239).

Fertile

An isotope that is not fissile itself but can be converted into a fissile material via neutron capture (e.g., U-238).

Monte Carlo simulation

A stochastic computational method that simulates neutron transport by random sampling of interactions within a defined geometry.

Geometry (in Monte Carlo)

The spatial arrangement of materials used in a simulation (e.g., spherical geometry).

Leakage

Neutron loss from the system; a key consideration in reactor design and simulations.

Neutron source

A source that provides neutrons to initiate a simulation or experiment; often placed at the geometry center.

Cross-section data library

A repository of measured or calculated cross-section data (e.g., from Los Alamos National Laboratory) used in simulations.

Natural uranium cross section

The effective cross section for natural uranium, calculated as a weighted sum of the isotopes’ cross sections by their abundances.