are

detailed cad model and simulations of the are (aircraft reactor experiment), operated by oak ridge national laboratory 1954

model

are core

docs/are.pdf lists reference of the are core design, documented in the ornl reports and located in the repository github.com/openmsr/msr-archive

work-in-progress cad model of the are can be found here on onshape.

note that this work and the cad model is under the GNU General Public License v3.0

prerequisites

CAD_to_openMC

CAD_to_openMC is an open-source package to convert CAD geometry (in the form of '.step' files) into an openmc-readable h5m file

openmc

These simulations use openmc. automated source installation scripts for linux can be found here

simulation guide

We have set up a set of helper scripts to run a set of standard simulations with the ARE-geometry: k_eff, neutron_flux, ...

All these runs first check if the necessary set of three .h5m-files exist. If they do, they ae used, of not a set will be created on the fly using the helper script are_step_to_h5m.py. Thus, should you wish to regenerate the set (or part of it) you can simply move the files out of the way or run are_step_to_h5m.py manually.

k_eff This is often the first simulation run done on a geometry which

estimates the reactivity or criticality of the reactor geometry.

energy spectrum

This run defines an overall tally in the reactor geometry that measures the energy of neutron within it, as well as a set of localized energy-scoring tallies only scoring particles within:

1. fuel-salt within the reactor core
2. the reactor vessel wall.
3. the coolant salt.

Please note that even though the ARE reactor geometry describes a fluid fuel-reactor, this simulation is a static steady-state run, meaning we presently neglect depletion/burn-up and things like delayed neutrons. It is possible to account for such effects, but is considere out-of-scope for this set of sims.

neutron_flux

This simulation run is very similar to k_eff, ecept that it adds a set of mesh-tallies on slices through the reactor core. Each plane is set to score the neutron-flux in the regions as a function of spatial coordinate, or more precisely flux binnned in cuboid volume elements in the planes. The tallies predefined in the computations are a set of XY|z\in{-50,-25,0,25,50} slices as well as slices at XZ|y=0, and YZ|z=0. Tallies, and mesh-tallies in particular, create some extra overhead slowing the simulations down somewhat - this is normal.

geometry plot

This type of run is not a simulation run as such, instead it creates a set of geometry slice plots aælong the principal axis. This is useful to make sure that you are in fact simulating the geometry you think you are. Lately, a new tool has been included the openmc-distribution: openmc-plotter, which largely renders doing this type of run manually moot. The plotter does perform a geometry plot-run in the background.

Jupyter notebooks

All this is also put together in a jupyter notebook that may be found in the subdirectory jupyter.