07-09, 11:25–11:55 (US/Pacific), Ballroom
OpenMC is an open source, community-developed, Monte Carlo tool for neutron transport simulations, featuring a depletion module for fuel burnup calculations in nuclear reactors and a Python API. Depletion calculations can be expensive as they require solving the neutron transport and bateman equations in each timestep to update the neutron flux and material composition, respectively. Material properties such as temperature and density govern material cross sections, which in turn govern reaction rates. The reaction rates can effect the neutron population. In a scenario where there is no significant change in the material properties or composition, the transport simulation may only need to be run once; the same cross sections are 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 briefly touch on areas where the feature has been used. Two recent use cases will be highlighted. The first use case calculates shutdown dose rates for fusion power applications, and the second performs depletion for fission reactor fuel cycle modeling.
Neutrons can induce nuclear fission in fissile nuclides like 235U. The fission reaction causes the nucleus to break apart, releasing both energy and new nuclides, many of which are radioactive isotopes of smaller elements. In nuclear reactors, which are fueled by fissionable material, this process is referred to as depletion, or burnup. Nuclear engineers model this process to design and license new reactors. Depletion can affect reactor physics and 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 the neutron flux, which changes reaction rates in the fuel. These reaction rates inform the material composition at the next time step. This process is repeated each time step and is called 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 solving the transport equation at each time step. Instead, the transport equation is solved once. From this solution, the flux of neutrons and their cross sections are discretized in energy and normalized. These are called multigroup fluxes and cross sections, respectively, and are used during every time step of the depletion calculation. The process is called transport-independent depletion. This method is accurate for the first time step, but degrades at further time steps. Testing with a simple model indicates that the error with respect to transport-coupled depletion depends on the nuclide of interest.
In this talk, I will cover:
- A brief background on the physics and mathematics of depletion,
- the accuracy of transport-independent depletion compared to transport-coupled depletion,
- and 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 people interested in nuclear reactor systems and open source software.
Oleksandr is a PhD student at the University of Illinois Urbana-Champaign working on numerical methods for time-dependent neutron transport.