Python API¶

OpenMC includes a rich Python API that enables programmatic pre- and post-processing. The easiest way to begin using the API is to take a look at the example Jupyter notebooks provided. However, this assumes that you are already familiar with Python and common third-party packages such as NumPy. If you have never programmed in Python before, there are many good tutorials available online. We recommend going through the modules from Codecademy and/or the Scipy lectures. The full API documentation serves to provide more information on a given module or class.

openmc – Basic Functionality¶

Handling nuclear data¶

Classes¶

 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.

Functions¶

 openmc.ace.ascii_to_binary Convert an ACE file in ASCII format (type 1) to binary format (type 2).

Simulation Settings¶

 openmc.Source Distribution of phase space coordinates for source sites. openmc.ResonanceScattering Specification of the elastic scattering model for resonant isotopes openmc.Settings Settings file used for an OpenMC simulation.

Material Specification¶

 openmc.Nuclide A nuclide that can be used in a material. openmc.Element A natural element used in a material via . openmc.Macroscopic A Macroscopic object that can be used in a material. openmc.Material A material composed of a collection of nuclides/elements that can be assigned to a region of space. openmc.Materials Collection of Materials used for an OpenMC simulation.

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.

One function is also available to create a hexagonal region defined by the intersection of six surface half-spaces.

 openmc.make_hexagon_region Create a hexagon region from six surface planes.

Constructing Tallies¶

 openmc.Filter A filter used to constrain a tally to a specific criterion, e.g. openmc.Mesh A structured Cartesian mesh in two or three dimensions openmc.Trigger A criterion for when to finish a simulation based on tally uncertainties. 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.

Plotting¶

 openmc.Plot Definition of a finite region of space to be plotted, either as a slice plot in two dimensions or as a voxel plot in three dimensions. openmc.Plots Collection of Plots used for an OpenMC simulation.

Running OpenMC¶

 openmc.run Run an OpenMC simulation. openmc.plot_geometry Run OpenMC in plotting mode

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 Information summarizing the 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.

openmc.stats – Statistics¶

Univariate Probability Distributions¶

 openmc.stats.Univariate Probability distribution of a single random variable. openmc.stats.Discrete Distribution characterized by a probability mass function. openmc.stats.Uniform Distribution with constant probability over a finite interval [a,b] openmc.stats.Maxwell Maxwellian distribution in energy. openmc.stats.Watt Watt fission energy spectrum. openmc.stats.Tabular Piecewise continuous probability distribution.

Angular Distributions¶

 openmc.stats.UnitSphere Distribution of points on the unit sphere. openmc.stats.PolarAzimuthal Angular distribution represented by polar and azimuthal angles openmc.stats.Isotropic Isotropic angular distribution. openmc.stats.Monodirectional Monodirectional angular distribution.

Spatial Distributions¶

 openmc.stats.Spatial Distribution of locations in three-dimensional Euclidean space. openmc.stats.CartesianIndependent Spatial distribution with independent x, y, and z distributions. openmc.stats.Box Uniform distribution of coordinates in a rectangular cuboid. openmc.stats.Point Delta function in three dimensions.

openmc.mgxs – Multi-Group Cross Section Generation¶

Energy Groups¶

 openmc.mgxs.EnergyGroups An energy groups structure used for multi-group cross-sections.

Multi-group Cross Sections¶

 openmc.mgxs.MGXS An abstract multi-group cross section for some energy group structure within some spatial domain. openmc.mgxs.AbsorptionXS An absorption multi-group cross section. openmc.mgxs.CaptureXS A capture multi-group cross section. openmc.mgxs.Chi The fission spectrum. openmc.mgxs.FissionXS A fission multi-group cross section. openmc.mgxs.KappaFissionXS A recoverable fission energy production rate multi-group cross section. openmc.mgxs.MultiplicityMatrixXS The scattering multiplicity matrix. openmc.mgxs.NuFissionXS A fission neutron production multi-group cross section. openmc.mgxs.NuFissionMatrixXS A fission production matrix multi-group cross section. openmc.mgxs.NuScatterXS A scattering neutron production multi-group cross section. openmc.mgxs.NuScatterMatrixXS A scattering production matrix multi-group cross section for one or more Legendre moments. openmc.mgxs.ScatterXS A scattering multi-group cross section. openmc.mgxs.ScatterMatrixXS A scattering matrix multi-group cross section for one or more Legendre moments. openmc.mgxs.TotalXS A total multi-group cross section. openmc.mgxs.TransportXS A transport-corrected total multi-group cross section.

Multi-group Cross Section Libraries¶

 openmc.mgxs.Library A multi-group cross section library for some energy group structure.

openmc.model – Model Building¶

TRISO Fuel Modeling¶

Classes¶

 openmc.model.TRISO Tristructural-isotopic (TRISO) micro fuel particle

Functions¶

 openmc.model.create_triso_lattice Create a lattice containing TRISO particles for optimized tracking.