The number density of any fission product nuclide in the reactor is governed by the following differential equation:
Let us explain each term carefully:
| Term | Symbol | Meaning (in plain English) |
|---|---|---|
| (1) Decay in | Nuclide is produced when its precursor nuclide(s) decay into it. is the decay constant for that specific decay path, and is the number of precursor atoms present. | |
| (2) Capture in | Nuclide is produced when another nuclide absorbs a neutron and is transformed into . is the capture cross-section for that reaction, and is the neutron flux. | |
| (3) Fission production | Nuclide is produced directly from fission. is the fission yield of nuclide from the fissioning of nuclide (e.g. U), is the fission cross-section of nuclide , and is the neutron flux. | |
| (4) Decay out | Nuclide is removed by its own radioactive decay. is its decay constant. | |
| (5) Burn-up | Nuclide is removed when it absorbs a neutron (burn-up). is its neutron absorption cross-section. |
In practice, these coupled equations are solved numerically using computer codes such as FISPIN (BNFL) or ORIGEN (Oak Ridge National Laboratory). These codes handle hundreds of isotopes, time-varying flux, and multi-group cross-sections.
After Shutdown
When the reactor shuts down, the neutron flux drops to zero. All terms containing vanish, and the equation simplifies to:
This is simply the decay chain equation --- nuclides are produced only from the decay of their precursors and are removed only by their own decay.