`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.Settings` |
Settings used for an OpenMC simulation. |

The following function can be used for generating a source file:

`openmc.write_source_file` |
Write a source file using a collection of source particles |

## 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.CellInstanceFilter` |
Bins tally events based on which cell instance a particle is in. |

`openmc.SurfaceFilter` |
Filters particles by surface crossing |

`openmc.MeshFilter` |
Bins tally event locations onto a regular, rectangular mesh. |

`openmc.MeshSurfaceFilter` |
Filter events by surface crossings on 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.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.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.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.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. |

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()`

.