OpenMC is a community developed open source software for simulating neutron transport, and includes a depletion module for calculating fuel burnup in nuclear reactors. Depletion calculations can be expensive as they require an iterative solution to the neutron transport equation and material composition updates. In a scenario where the material composition does not appreciably change, or we need reasonably accurate and low cost depletion, the transport solution may only need to be run once; the same cross sections used for the entire depletion calculation. We recently extended the depletion module in OpenMC to enable transport-independent depletion using multigroup cross sections and fluxes. This talk will focus on the technical details of this feature, its validation, and areas where the feature has been used, in particular in calculating shutdown dose rates for fusion power applications as well as in performing depletion for fuel cycle modeling.
Neutron radiation of the fuel inside of a nuclear reactor induces nuclear fission in certain nuclides like U235. The fission reaction causes the nucleus to break apart, releasing both energy and new nuclides, many of which are radioactive isotopes of smaller elements. This process is referred to as depletion or burnup. We model this process to design and license new reactors as depletion can effect performance and determines when the fuel must be shuffled or replaced. The typical approach to modeling depletion requires solving the neutron transport equation to obtain reaction rates which govern the material composition at the next time step. This process is repeated iteratively (transport-coupled depletion).
OpenMC is an open-source neutron transport code with a built-in depletion module. OpenMC solves the transport equation via Monte Carlo particle transport, which is accurate but expensive. OpenMC's depletion module was recently extended to enable depletion modeling without iteratively solving the transport equation (transport-independent depletion). Instead, the transport equation is solved once, and multigroup cross sections and fluxes are obtained from this solution. The multigroup cross sections and fluxes are used for every timestep of the depletion calculation. This method is accurate for the first timestep, but degrades at further timesteps. Testing with a simple model indicated errors depend on the nuclide of interest.
In this talk, I will cover:
- The physics and mathematics of depletion
- Accuracy of transport-independent depletion compared to transport-coupled depletion.
- Two applications where transport-independent depletion has been used to great effect: shutdown dose-rate calculations for fusion energy applications, and fast depletion for fuel cycle analysis.
The intended audience of this talk are folks who are interested in nuclear engineering and open-source software.
This could also be in the "Celebrating the 'Sci' in SciPy" track.