Neutron Flux Calculator

Neutron flux is the number of neutrons passing through a unit area per unit time, expressed in n/cm²·s. It is measured using activation detectors, fission chambers, or self-powered neutron detectors placed inside or near a reactor core. The flux characterizes the intensity of the neutron field at a given location.

Thermal neutrons have energies below ~0.5 eV and are in thermal equilibrium with the surrounding medium. Epithermal neutrons fall between ~0.5 eV and ~0.1 MeV. Fast neutrons have energies above ~0.1 MeV and originate directly from fission reactions. Each region interacts with materials differently and drives distinct nuclear reactions.

The Cadmium ratio is the ratio of the activation rate of a bare detector to that of the same detector wrapped in cadmium foil. Cadmium absorbs neutrons with energies below ~0.5 eV, so this ratio lets you separate thermal and epithermal flux contributions. A higher Cd ratio indicates a harder (more epithermal-dominant) neutron spectrum.

Neutron fluence is the time-integrated neutron flux, measured in n/cm². While flux describes the instantaneous intensity of the neutron field, fluence measures the cumulative neutron exposure over the irradiation period. Fluence is the key parameter for predicting material activation and radiation damage.

Self-shielding occurs when a sample absorbs a significant fraction of the incident neutrons, reducing the effective flux seen by the interior of the sample. This correction factor (Gth) depends on the sample's macroscopic thermal neutron cross-section, its thickness, and density. Thin or dilute samples typically have Gth values close to 1.

At higher elevations, the atmospheric shielding from cosmic rays is reduced, increasing the cosmic-ray neutron background. This effect is significant for sensitive low-background measurements or electronics reliability testing (per JESD89B). Sea-level measurements serve as the reference baseline for cosmic neutron flux comparisons.

Thermal reactors (e.g. light-water reactors) moderate neutrons to thermal energies, producing a large thermal flux component. Fast reactors have no moderator and maintain a predominantly high-energy fission spectrum. Mixed-spectrum reactors fall between these extremes and are common in research reactor settings near reflector regions.

Neutron fluence is used to predict radiation damage in structural materials, fuel burnup in reactor fuel elements, and the degree of radioactive activation in irradiated samples. Damage correlations such as displacement per atom (dpa) are directly proportional to the fast fluence experienced by the material.