Task Introduction

Task Introduction#

Please allow 20 minutes for this task.

Expected outputs from this task are also in the presentation.

In this task you will use OpenMC to simulate transmutation, activation and depletion of materials under neutron irradiation. The examples progress from a simple flux-spectrum-based depletion (no transport required) through to coupled transport-depletion simulations with pulsed irradiation schedules.

Learning Outcomes

  • OpenMC can perform depletion using a precomputed multigroup flux spectrum, similar to inventory codes like FISPACT, ORIGEN and ALARA. This is fast and requires no geometry.

  • OpenMC can also perform coupled transport-depletion where the neutron flux is recalculated as the material composition evolves.

  • Activation products build up during irradiation and start to saturate at around five half-lives, when the rate of creation approaches the rate of decay.

  • Activity, decay heat and contact dose rate can be extracted from depleted materials.

  • Material evolution during irradiation affects tally results. For example, Tritium Breeding Ratio decreases as lithium-6 is burnt up.

  • Pulsed irradiation schedules can be modelled to simulate realistic reactor operating conditions.