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.SourceParticle	Source particle
openmc.VolumeCalculation	Stochastic volume calculation specifications and results.
openmc.WeightWindows	Mesh-based weight windows
openmc.Settings	Settings used for an OpenMC simulation.

openmc.write_source_file	Write a source file using a collection of source particles
openmc.wwinp_to_wws	Create WeightWindows instances from a wwinp file

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 z-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.XTorus	A torus of the form \((x - x_0)^2/B^2 + (\sqrt{(y - y_0)^2 + (z - z_0)^2} - A)^2/C^2 - 1 = 0\).
openmc.YTorus	A torus of the form \((y - y_0)^2/B^2 + (\sqrt{(x - x_0)^2 + (z - z_0)^2} - A)^2/C^2 - 1 = 0\).
openmc.ZTorus	A torus of the form \((z - z_0)^2/B^2 + (\sqrt{(x - x_0)^2 + (y - y_0)^2} - A)^2/C^2 - 1 = 0\).
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.BoundingBox	Axis-aligned bounding box.
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.DAGMCUniverse	A reference to a DAGMC file to be used in the model.
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 occurred in.
openmc.MaterialFilter	Bins tally event locations based on the Material they occurred in.
openmc.CellFilter	Bins tally event locations based on the Cell they occurred in.
openmc.CellFromFilter	Bins tally on which cell the particle came from.
openmc.CellBornFilter	Bins tally events based on which cell the particle was born in.
openmc.CellInstanceFilter	Bins tally events based on which cell instance a particle is in.
openmc.CollisionFilter	Bins tally events based on the number of collisions.
openmc.SurfaceFilter	Filters particles by surface crossing
openmc.MeshFilter	Bins tally event locations by mesh elements.
openmc.MeshSurfaceFilter	Filter events by surface crossings on a 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.LegendreFilter	Score Legendre expansion moments up to specified order.
openmc.SpatialLegendreFilter	Score Legendre expansion moments in space up to specified order.
openmc.SphericalHarmonicsFilter	Score spherical harmonic expansion moments up to specified order.
openmc.TimeFilter	Bins tally events based on the particle's time.
openmc.ZernikeFilter	Score Zernike expansion moments in space up to specified order.
openmc.ZernikeRadialFilter	Score the \(m = 0\) (radial variation only) Zernike moments up to specified order.
openmc.ParticleFilter	Bins tally events based on the Particle type.
openmc.RegularMesh	A regular Cartesian mesh in one, two, or three dimensions
openmc.RectilinearMesh	A 3D rectilinear Cartesian mesh
openmc.CylindricalMesh	A 3D cylindrical mesh
openmc.SphericalMesh	A 3D spherical mesh
openmc.UnstructuredMesh	A 3D unstructured mesh
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.

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.ParticleTrack	Particle track information
openmc.StatePoint	State information on a simulation at a certain point in time (at the end of a given batch).
openmc.Summary	Summary of model used in a simulation.
openmc.Track	Tracks resulting from a single source particle
openmc.Tracks	Collection of particle tracks

openmc.voxel_to_vtk

Converts a voxel HDF5 file to a VTK file

The following classes and functions are used for functional expansion reconstruction.

openmc.ZernikeRadial

Create radial only Zernike polynomials given coefficients and domain.

openmc.legendre_from_expcoef

Return a Legendre series object based on expansion coefficients.

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.

Coarse Mesh Finite Difference Acceleration

CMFD is implemented in OpenMC and allows users to accelerate fission source convergence during inactive neutron batches. To use CMFD, the openmc.cmfd.CMFDRun class executes OpenMC through the C API, solving the CMFD system between fission generations and modifying the source weights. Note that the openmc.cmfd module is not imported by default with the openmc namespace and needs to be imported explicitly.

openmc.cmfd.CMFDMesh	A structured Cartesian mesh used for CMFD acceleration.
openmc.cmfd.CMFDRun	Class for running CMFD acceleration through the C API.

At the minimum, a CMFD mesh needs to be specified in order to run CMFD. Once the mesh and other optional properties are set, a simulation can be run with CMFD turned on using openmc.cmfd.CMFDRun.run().