openmc
– Basic Functionality¶
Handling nuclear data¶
openmc.XSdata |
A multi-group cross section data set providing all the multi-group data necessary for a multi-group OpenMC calculation. |
openmc.MGXSLibrary |
Multi-Group Cross Sections file used for an OpenMC simulation. |
Simulation Settings¶
openmc.Source |
Distribution of phase space coordinates for source sites. |
openmc.VolumeCalculation |
Stochastic volume calculation specifications and results. |
openmc.Settings |
Settings used for an OpenMC simulation. |
Material Specification¶
openmc.Nuclide |
A nuclide that can be used in a material. |
openmc.Element |
A natural element that auto-expands to add the isotopes of an element to a material in their natural abundance. |
openmc.Macroscopic |
A Macroscopic object that can be used in a material. |
openmc.Material |
A material composed of a collection of nuclides/elements. |
openmc.Materials |
Collection of Materials used for an OpenMC simulation. |
Cross sections for nuclides, elements, and materials can be plotted using the following function:
openmc.plot_xs |
Creates a figure of continuous-energy cross sections for this item. |
Building geometry¶
openmc.Plane |
An arbitrary plane of the form \(Ax + By + Cz = D\). |
openmc.XPlane |
A plane perpendicular to the x axis of the form \(x - x_0 = 0\) |
openmc.YPlane |
A plane perpendicular to the y axis of the form \(y - y_0 = 0\) |
openmc.ZPlane |
A plane perpendicular to the z axis of the form \(z - z_0 = 0\) |
openmc.XCylinder |
An infinite cylinder whose length is parallel to the x-axis of the form \((y - y_0)^2 + (z - z_0)^2 = R^2\). |
openmc.YCylinder |
An infinite cylinder whose length is parallel to the y-axis of the form \((x - x_0)^2 + (z - z_0)^2 = R^2\). |
openmc.ZCylinder |
An infinite cylinder whose length is parallel to the z-axis of the form \((x - x_0)^2 + (y - y_0)^2 = R^2\). |
openmc.Sphere |
A sphere of the form \((x - x_0)^2 + (y - y_0)^2 + (z - z_0)^2 = R^2\). |
openmc.Cone |
A conical surface parallel to the x-, y-, or z-axis. |
openmc.XCone |
A cone parallel to the x-axis of the form \((y - y_0)^2 + (z - z_0)^2 = R^2 (x - x_0)^2\). |
openmc.YCone |
A cone parallel to the y-axis of the form \((x - x_0)^2 + (z - z_0)^2 = R^2 (y - y_0)^2\). |
openmc.ZCone |
A cone parallel to the x-axis of the form \((x - x_0)^2 + (y - y_0)^2 = R^2 (z - z_0)^2\). |
openmc.Quadric |
A surface of the form \(Ax^2 + By^2 + Cz^2 + Dxy + Eyz + Fxz + Gx + Hy + Jz + K = 0\). |
openmc.Halfspace |
A positive or negative half-space region. |
openmc.Intersection |
Intersection of two or more regions. |
openmc.Union |
Union of two or more regions. |
openmc.Complement |
Complement of a region. |
openmc.Cell |
A region of space defined as the intersection of half-space created by quadric surfaces. |
openmc.Universe |
A collection of cells that can be repeated. |
openmc.RectLattice |
A lattice consisting of rectangular prisms. |
openmc.HexLattice |
A lattice consisting of hexagonal prisms. |
openmc.Geometry |
Geometry representing a collection of surfaces, cells, and universes. |
Many of the above classes are derived from several abstract classes:
openmc.Surface |
An implicit surface with an associated boundary condition. |
openmc.Region |
Region of space that can be assigned to a cell. |
openmc.Lattice |
A repeating structure wherein each element is a universe. |
Constructing Tallies¶
openmc.Filter |
Tally modifier that describes phase-space and other characteristics. |
openmc.UniverseFilter |
Bins tally event locations based on the Universe they occured in. |
openmc.MaterialFilter |
Bins tally event locations based on the Material they occured in. |
openmc.CellFilter |
Bins tally event locations based on the Cell they occured in. |
openmc.CellFromFilter |
Bins tally on which Cell the neutron came from. |
openmc.CellbornFilter |
Bins tally events based on which Cell the neutron was born in. |
openmc.SurfaceFilter |
Bins particle currents on Mesh surfaces. |
openmc.MeshFilter |
Bins tally event locations onto a regular, rectangular mesh. |
openmc.EnergyFilter |
Bins tally events based on incident particle energy. |
openmc.EnergyoutFilter |
Bins tally events based on outgoing particle energy. |
openmc.MuFilter |
Bins tally events based on particle scattering angle. |
openmc.PolarFilter |
Bins tally events based on the incident particle’s direction. |
openmc.AzimuthalFilter |
Bins tally events based on the incident particle’s direction. |
openmc.DistribcellFilter |
Bins tally event locations on instances of repeated cells. |
openmc.DelayedGroupFilter |
Bins fission events based on the produced neutron precursor groups. |
openmc.EnergyFunctionFilter |
Multiplies tally scores by an arbitrary function of incident energy. |
openmc.Mesh |
A structured Cartesian mesh in one, two, or three dimensions |
openmc.Trigger |
A criterion for when to finish a simulation based on tally uncertainties. |
openmc.TallyDerivative |
A material perturbation derivative to apply to a tally. |
openmc.Tally |
A tally defined by a set of scores that are accumulated for a list of nuclides given a set of filters. |
openmc.Tallies |
Collection of Tallies used for an OpenMC simulation. |
Coarse Mesh Finite Difference Acceleration¶
openmc.CMFDMesh |
A structured Cartesian mesh used for Coarse Mesh Finite Difference (CMFD) acceleration. |
openmc.CMFD |
Parameters that control the use of coarse-mesh finite difference acceleration in OpenMC. |
Geometry Plotting¶
openmc.Plot |
Definition of a finite region of space to be plotted. |
openmc.Plots |
Collection of Plots used for an OpenMC simulation. |
Running OpenMC¶
openmc.run |
Run an OpenMC simulation. |
openmc.calculate_volumes |
Run stochastic volume calculations in OpenMC. |
openmc.plot_geometry |
Run OpenMC in plotting mode |
openmc.plot_inline |
Display plots inline in a Jupyter notebook. |
openmc.search_for_keff |
Function to perform a keff search by modifying a model parametrized by a single independent variable. |
Post-processing¶
openmc.Particle |
Information used to restart a specific particle that caused a simulation to fail. |
openmc.StatePoint |
State information on a simulation at a certain point in time (at the end of a given batch). |
openmc.Summary |
Summary of geometry, materials, and tallies used in a simulation. |
Various classes may be created when performing tally slicing and/or arithmetic:
openmc.arithmetic.CrossScore |
A special-purpose tally score used to encapsulate all combinations of two tally’s scores as an outer product for tally arithmetic. |
openmc.arithmetic.CrossNuclide |
A special-purpose nuclide used to encapsulate all combinations of two tally’s nuclides as an outer product for tally arithmetic. |
openmc.arithmetic.CrossFilter |
A special-purpose filter used to encapsulate all combinations of two tally’s filter bins as an outer product for tally arithmetic. |
openmc.arithmetic.AggregateScore |
A special-purpose tally score used to encapsulate an aggregate of a subset or all of tally’s scores for tally aggregation. |
openmc.arithmetic.AggregateNuclide |
A special-purpose tally nuclide used to encapsulate an aggregate of a subset or all of tally’s nuclides for tally aggregation. |
openmc.arithmetic.AggregateFilter |
A special-purpose tally filter used to encapsulate an aggregate of a subset or all of a tally filter’s bins for tally aggregation. |