OpenMC Monte Carlo Code
This release of OpenMC includes many bug fixes, performance improvements, and several notable new features. Some of the highlights include projection plots, pulse height tallies for photons, weight window generation, and an ability to specify continuous removal or feed of nuclides/elements during depletion. Additionally, one of the longstanding annoyances of depletion calculations, namely the need to include initial "dilute" nuclides, has been eliminated. There are also a wide array of general improvements in the Python API.
openmc.deplete.MicroXS class has been completely redesigned and improved. See further comments below under "New Features". (#2572, #2579, #2595, #2700)rectangular_prism function has been replaced by the openmc.model.RectangularPrism class and the hexagonal_prism function has been replaced by the openmc.model.HexagonalPrism class. Note that whereas the rectangular_prism and hexagonal_prism functions returned a region representing the interior of the prism, the new openmc.model.RectangularPrism and
openmc.model.HexagonalPrism classes return composite surfaces, so you need to use the unary - or + operators to obtain a region that can be assigned to a cell. (#2739)Source class has been refactored and split up into three separate classes: openmc.IndependentSource, openmc.FileSource, and openmc.CompiledSource. (#2524)vertices and centroids attributes on mesh classes now always return Cartesian coordinates and the shape of the returned arrays has changed to allow ijk indexing using a tuple (i.e., xyz = vertices[i, j, k]). (#2711)openmc.Material.decay_photon_energy attribute has been replaced by the openmc.Material.get_decay_photon_energy method. (#2715)openmc.ProjectionPlot class enables the generation of orthographic or perspective projection plots. (#1926)openmc.model.RightCircularCylinder class now supports optional filleted edges. (#2309)openmc.deplete.abc.Integrator.add_transfer_rate method. (#2358, #2564, #2626)openmc.model.CruciformPrism class has been added that provides a generalized cruciform prism composite surface. (#2457)openmc.Plot.to_vtk method. (#2464)openmc.mgxs.EnergyGroups class now allows you to alternatively pass a string of the group structure name (e.g., "CCFE-709") instead of the energy group boundaries. (#2466)openmc.Universe.plot method (addition of axis labels with units, ability to show legend and/or outlines, automatic determination of origin/width, ability to pass total number of pixels). (#2472 , #2482, #2483, #2492, #2513, #2575)openmc.BoundingBox class. (#2475)openmc.plot_xs function. (#2478)openmc Python module has been improved by deferring the import of matplotlib. (#2488)bounding_box property. (#2507, #2620, #2621)Source class has been refactored and split up into three separate classes: openmc.IndependentSource, openmc.FileSource, and openmc.CompiledSource. (#2524)openmc.Tally class now has a openmc.Tally.multiply_density attribute that indicates whether reaction rate tallies should include the number density of the nuclide of interest. (#2539)openmc.wwinp_to_wws function now supports wwinp files with cylindrical or spherical meshes. (#2556)openmc.deplete.Results class now has openmc.deplete.Results.get_mass (#2565), openmc.deplete.Results.get_activity (#2617), and openmc.deplete.Results.get_decay_heat (#2625) methods.openmc.deplete.StepResult.save method now supports a path argument. (#2567)openmc.deplete.MicroXS has been completely redesigned and improved. First, it no longer relies on the openmc.mgxs module, no longer subclasses pandas.DataFrame, and doesn't require adding initial "dilute" nuclides into material compositions. It now enables users to specify an energy group structure to collect multigroup cross sections, specify nuclides/reactions, and works with mesh domains in addition to the existing domains. A new openmc.deplete.get_microxs_and_flux function was added that improves the workflow for calculating microscopic cross sections along with fluxes. Altogether, these changes make it straightforward to switch between coupled and independent operators for depletion/activation calculations. (#2572, #2579 , #2595, #2700)openmc.Geometry class now has merge_surfaces and surface_precision arguments. (#2602)openmc.CylindricalMesh and openmc.SphericalMesh classes can now be fully formed using the constructor. (#2619)openmc.Settings.cutoff attribute. (#2631)openmc.Material.add_element method now supports a cross_sections argument that allows a cross section data source to be specified. (#2633 )openmc.Cell class now has a plot method. (#2648)openmc.Geometry class now has a plot method. (#2661)openmc.Settings.weight_window_checkpoints attribute. (#2670)openmc.Settings class now has a max_write_lost_particles attribute that can limit the number of lost particle files written. (#2688)openmc.deplete.CoupledOperator class now has a diff_volume_method argument that specifies how the volume of new materials should be determined. (#2691)openmc.DAGMCUniverse.bounding_region method now has a padding_distance argument. (#2701)openmc.Material.get_decay_photon_energy method replaces the decay_photon_energy attribute and includes an ability to eliminate low-importance points. This is facilitated by a new openmc.stats.Discrete.clip method. (#2715)openmc.model.Model.differentiate_depletable_mats method allows depletable materials to be differentiated independent of the depletion calculation itself. (#2718)This release of OpenMC includes many bug fixes, performance improvements, and several notable new features. Some of the highlights include support for MCPL source files, NCrystal thermal scattering materials, and a new openmc.stats.MeshSpatial class that allows a source distribution to be specified over a mesh. Additionally, OpenMC now allows you to export your model as a single XML file rather than separate XML files for geometry, materials, settings, and tallies.
openmc.data.atomic_mass has been updated to AME 2020, which results in slightly different masses.openmc.CylindricalMesh and openmc.SphericalMesh classes now have an origin attribute that changes the center of the mesh. (#2256)openmc.model.Polygon class allows defining generalized 2D polygons. (#2266)openmc.data.decay_energy function and openmc.Material.get_decay_heat method enable determination of decay heat from a single nuclide or material. (#2287)openmc.stats.MeshSpatial allows a spatial source distribution to be specified with source strengths for each mesh element. (#2334)openmc.Geometry.get_surfaces_by_name method returns a list of matching surfaces in a geometry. (#2347)openmc.Settings.create_delayed_neutrons attribute controls whether delayed neutrons are created during a simulation. (#2348)openmc.deplete.Results.export_to_materials method now takes a path argument. (#2364)openmc.EnergyFilter.get_tabular method allows one to create a tabular distribution based on tally results using an energy filter. (#2371)openmc.Material class that require a volume to be set (e.g., openmc.Material.get_mass) now accept a volume argument. (#2412)This release of OpenMC includes several bug fixes, performance improvements for complex geometries and depletion simulations, and other general enhancements. Notably, a capability has been added to compute the photon spectra from decay of unstable nuclides. Alongside that, a new openmc.config configuration variable has been introduced that allows easier configuration of data sources. Additionally, users can now perform cell or material rejection when sampling external source distributions.
openmc.stats.Muir class has been replaced by a openmc.stats.muir function that returns an instance of openmc.stats.Normal.openmc.Material.get_nuclide_atom_densities method now takes an optional nuclide argument.openmc.deplete module now accept times in Julian years ('a').openmc.Universe.plot method now allows a pre-existing axes object to be passed in.openmc.RegularMesh class now has a from_domain classmethod.openmc.CylindricalMesh class now has a from_domain classmethod.openmc.mgxs module.openmc.config configuration variable has been introduced that allows data sources to be specified at runtime or via environment variables.openmc.EnergyFunctionFilter class now supports multiple interpolation schemes, not just linear-linear interpolation.openmc.DAGMCUniverse class now has material_names, n_cells, and n_surfaces attributes.openmc.data.decay_photon_energy function has been added that returns the energy spectrum of photons emitted from the decay of an unstable nuclide.openmc.Material class also has a new decay_photon_energy attribute that gives the decay photon energy spectrum from the material based on its constituent nuclides.openmc.deplete.StepResult now has a get_material method.openmc.Source class now takes a domains argument that specifies a list of cells, materials, or universes that is used to reject source sites (i.e., if the sampled sites are not within the specified domain, they are rejected).This release of OpenMC includes many bug fixes as well as improvements in geometry modeling, mesh functionality, source specification, depletion capabilities, and other general enhancements. The depletion module features a new transport operator, openmc.deplete.IndependentOperator, that allows a depletion calculation to be performed using arbitrary one-group cross sections (e.g., generated by an external solver) along with a openmc.deplete.MicroXS class for managing one-group cross sections. The track file generation capability has been significantly overhauled and a new openmc.Tracks class was introduced to allow access to information in track files from the Python API. Support has been added for new ENDF thermal scattering evaluations that use mixed coherent/incoherent elastic scattering.
The openmc.deplete.Operator class has been renamed openmc.deplete.CoupledOperator.
The openmc.deplete.ResultsList class has been renamed to openmc.deplete.Results and no longer requires you to call the from_hdf5() method in order to create it; instead, you can directly instantiate it.
A few methods that represent k-effective have been renamed for the sake of consistency:
openmc.StatePoint.k_combined is now openmc.StatePoint.keff
openmc.deplete.ResultsList.get_eigenvalue is now openmc.deplete.Results.get_keff
The openmc.stats.SphericalIndependent class, which used to accept a distribution for theta now accepts a distribution for cos_theta instead in order to more easily handle the common case of specifying a uniform spatial distribution over a sphere (also see the new openmc.stats.spherical_uniform function).
If you are building OpenMC from source, note that several of our CMake options have been changed:
| Old option | New option |
|---|---|
debug |
--- |
optimize |
--- |
profile |
OPENMC_ENABLE_PROFILE |
coverage |
OPENMC_ENABLE_COVERAGE |
openmp |
OPENMC_USE_OPENMP |
| --- | OPENMC_USE_MPI |
dagmc |
OPENMC_USE_DAGMC |
libmesh |
OPENMC_USE_LIBMESH |
The debug and optimize options have been removed; instead, use the standard CMAKE_BUILD_TYPE variable.
Two new composite surfaces: openmc.model.IsogonalOctagon and openmc.model.CylinderSector.
The DAGMCUniverse class now has a bounding_box attribute and a bounding_region method.
When translating a Region using the translate method, there is now an inplace argument.
The Material class has several new methods and attributes:
add_components methods allows you to add multiple nuclides/elements to a material with a single call by passing a dictionary.get_activity method returns the activity of a material in Bq, Bq/g, or Bq/cm³.remove_element method removes an element from a materialget_nuclide_atoms method gives the number of atoms of each nuclide in a materialAll mesh classes now have a volumes property that provides the volume of each mesh element as well as write_data_to_vtk methods.
Support for externally managed MOAB meshes or libMesh meshes.
Multiple discrete distributions can be merged with the new openmc.stats.Discrete.merge method.
The openmc.stats.spherical_uniform function creates a uniform distribution over a sphere using the SphericalIndependent class.
Univariate distributions in the openmc.stats module now have sample() methods.
An openmc_sample_external_source function has been added to the C API with a corresponding Python binding openmc.lib.sample_external_source.
The track file generation capability has been completely overhauled. Track files now include much more information, and a new openmc.Tracks class allows access to track file information from the Python API and has a write_to_vtk method for writing a VTK file. Multiple tracks are now written to a single file (one per MPI rank).
A new openmc.wwinp_to_wws function that converts weight windows from a wwinp file to a list of WeightWindows objects.
The new openmc.EnergyFilter.from_group_structure method provides a way of creating an energy filter with a group structure identified by name.
The openmc.data.Decay class now has a sources property that provides radioactive decay source distributions.
A openmc.mgxs.ReducedAbsorptionXS class produces a multigroup cross section representing "reduced" absorption (absorption less neutron production from (n,xn) reactions).
Added support in the Python API and HDF5 nuclear data format for new ENDF thermal neutron scattering evaluations with mixed coherent elastic and incoherent elastic.
CMake now relies on find_package(MPI) for a more standard means of identifying an MPI compiler configuration.
This release of OpenMC includes several noteworthy and unique features. Most importantly, mesh-based weight windows have been added and work with all supported mesh types (regular, rectilinear, cylindrical, spherical, and unstructured). Other additions include torus surfaces, an ability to place CAD-based geometries in universes, a feature to export/import physical properties, and a filter for particle time.
There is one breaking changing in the Python API. The openmc.deplete.Operator class used to accept Geometry and Settings objects as its first two arguments; users now need to pass a Model class instead.
The minimum supported Python version is now 3.6.
WeightWindows class.XTorus, YTorus, and ZTorus classes.DAGMCUniverse, allowing users to combine CSG and CAD-based geometry in a single model.openmc_properties_export and openmc_properties_import with corresponding Python API bindings, openmc.lib.export_properties and openmc.lib.import_properties. These functions allow physical properties (temperatures, densities, material compositions) to be written to an HDF5 file and re-used for subsequent simulations.openmc.stats.PowerLaw univariate distributionModel class have been substantially expanded (e.g., the Model.deplete, Model.plot_geometry, and Model.rotate_cells methods).TimeFilter class that allows tallies to be filtered by the particle's time, which is now tracked.Source class now allows you to specify a time distribution.CylindricalMesh and SphericalMesh classes can be used for mesh tallies over cylidrical and spherical meshes, respectively.This release contains new contributions from the following people:
This release of OpenMC is primarily a hotfix release with numerous important bug fixes. Several tally-related enhancements have also been added.
Three tally-related enhancements were added to the code in this release:
CollisionFilter class that allows tallies to be filtered by the number of collisions a particle has undergone.translation attribute has been added to MeshFilter that allows a mesh to be translated from its original position before location checks are performed.UnstructuredMesh class now supports libMesh unstructured meshes to enable better ingration with MOOSE-based applications.This release contains new contributions from the following people:
This release of OpenMC includes an assortment of new features and many bug fixes. The openmc.deplete module incorporates a number of improvements in usability, accuracy, and performance. Other enhancements include generalized rotational periodic boundary conditions, expanded source modeling capabilities, and a capability to generate windowed multipole library files from ENDF files.
Boundary conditions have been refactored and generalized. Rotational periodic boundary conditions can now be applied to any N-fold symmetric geometry.
External source distributions have been refactored and extended. Users writing their own C++ custom sources need to write a class that derives from openmc::Source. These changes have enabled new functionality, such as:
New capability to read and write a source file based on particles that cross a surface (known as a "surface source").
Various improvements related to depletion:
reactions argument to openmc.deplete.Chain.from_endf.reaction_rate_mode and reaction_rate_opts to openmc.deplete.Operator.openmc.Materials object using the openmc.deplete.ResultsList.export_to_materials method.Multigroup current and diffusion cross sections can be generated through the openmc.mgxs.Current and openmc.mgxs.DiffusionCoefficient classes.
Added openmc.data.isotopes function that returns a list of naturally occurring isotopes for a given element.
Windowed multipole libraries can now be generated directly from the Python API using openmc.data.WindowedMultipole.from_endf.
The new openmc.write_source_file function allows source files to be generated programmatically.
RectangularParallelepiped.__pos__
This release contains new contributions from the following people:
This release of OpenMC includes an assortment of new features and many bug fixes. In particular, the openmc.deplete module has been heavily tested which has resulted in a number of usability improvements, bug fixes, and other enhancements. Energy deposition calculations, particularly for coupled neutron-photon simulations, have been improved as well.
Improvements in modeling capabilities continue to be added to the code, including the ability to rotate surfaces in the Python API, several new "composite" surfaces, a variety of new methods on openmc.Material, unstructured mesh tallies that leverage the existing DAGMC infrastructure, effective dose coefficients from ICRP-116, and a new cell instance tally filter.
All surfaces now have a rotate method that allows them to be rotated.
Several "composite" surfaces, which are actually composed of multiple surfaces but can be treated as a normal surface through the -/+ unary operators, have been added. These include:
openmc.model.RightCircularCylinder
openmc.model.RectangularParallelepiped
openmc.model.XConeOneSided (and equivalent versions for y- and z-axes)Various improvements related to depletion:
solver argument on depletion integrator classes.openmc.deplete.Chain.reduce method can automatically reduce the number of nuclides in a depletion chain.openmc.deplete.ResultsList.get_atoms now allows a user to obtain depleted material compositions in atom/b-cm.Several new methods on openmc.Material:
add_elements_from_formula method allows a user to create a material based on a chemical formula.add_element now supports the enrichment argument for non-uranium elements when only two isotopes are naturally occurring.add_element now supports adding elements by name rather than by symbol.get_elements method returns a list of elements within a material.mix_materials method allows multiple materials to be mixed together based on atom, weight, or volume fractions.The acceptable number of lost particles can now be configured through openmc.Settings.max_lost_particles and openmc.Settings.rel_max_lost_particles.
Delayed photons produced from fission are now accounted for by default by scaling the yield of prompt fission photons. This behavior can be modified through the openmc.Settings.delayed_photon_scaling attribute.
A trigger can now be specified for a volume calculation via the openmc.VolumeCalculation.set_trigger method.
The openmc.stats.SphericalIndependent and openmc.stats.CylindricalIndependent classes allow a user to specify source distributions based on spherical or cylindrical coordinates.
Custom external source distributions can be used via the openmc.Source.library attribute.
Unstructured mesh class, openmc.UnstructuredMesh, that can be used in tallies.
The openmc.CellInstanceFilter class allows one or more instances of a repeated cell to be tallied. This is effectively a more flexible version of the existing openmc.DistribcellFilter class.
The openmc.data.dose_coefficients function provides effective dose coefficients from ICRP-116 and can be used in conjunction with openmc.EnergyFunctionFilter in a tally.
__repr__ methods
This release contains new contributions from the following people:
This release of OpenMC adds several major new features: depletion, photon transport, and support for CAD geometries through DAGMC. In addition, the core codebase has been rewritten in C++14 (it was previously written in Fortran 2008). This makes compiling the code considerably simpler as no Fortran compiler is needed.
Functional expansion tallies are now supported through several new tally filters that can be arbitrarily combined:
openmc.LegendreFilter
openmc.SpatialLegendreFilter
openmc.SphericalHarmonicsFilter
openmc.ZernikeFilter
openmc.ZernikeRadialFilter
Note that these filters replace the use expansion scores like scatter-P1. Instead, a normal scatter score should be used along with a openmc.LegendreFilter.
The interface for random sphere packing has been significantly improved. A new openmc.model.pack_spheres function takes a region and generates a random, non-overlapping configuration of spheres within the region.
White boundary conditions can be applied to surfaces
Support for rectilinear meshes through openmc.RectilinearMesh.
The Geometry, Materials, and Settings classes now have a from_xml method that will build an instance from an existing XML file.
Predefined energy group structures can be found in openmc.mgxs.GROUP_STRUCTURES.
New tally scores: H1-production, H2-production, H3-production, He3-production, He4-production, heating, heating-local, and damage-energy.
Switched to cell-based neighor lists (PR 1140)
Two new probability distributions that can be used for source distributions: openmc.stats.Normal and openmc.stats.Muir
The openmc.data module now supports reading and sampling from ENDF File 32 resonance covariance data (PR 1024).
Several new convenience functions/methods have been added:
openmc.model.cylinder_from_points function creates a cylinder given two points passing through its center and a radius.openmc.Plane.from_points function creates a plane given three points that pass through it.openmc.model.pin function creates a pin cell universe given a sequence of concentric cylinders and materials.All surface classes now have coefficient arguments given as lowercase names.
The order of arguments in surface classes has been changed so that coefficients are the first arguments (rather than the optional surface ID). This means you can now write::
x = openmc.XPlane(5.0, 'reflective')
zc = openmc.ZCylinder(0., 0., 10.)
The Mesh class has been renamed openmc.RegularMesh.
The get_rectangular_prism function has been renamed openmc.model.rectangular_prism.
The get_hexagonal_prism function has been renamed openmc.model.hexagonal_prism.
Python bindings to the C/C++ API have been move from openmc.capi to openmc.lib.
This release contains new contributions from the following people:
This release of OpenMC includes several new features, performance improvements, and bug fixes compared to version 0.9.0. Notably, a C API has been added that enables in-memory coupling of neutronics to other physics fields, e.g., burnup calculations and thermal-hydraulics. The C API is also backed by Python bindings in a new openmc.capi package. Users should be forewarned that the C API is still in an experimental state and the interface is likely to undergo changes in future versions.
The Python API continues to improve over time; several backwards incompatible changes were made in the API which users of previous versions should take note of:
To indicate that nuclides in a material should be treated such that elastic scattering is isotropic in the laboratory system, there is a new Material.isotropic property:
mat = openmc.Material()
mat.add_nuclide('H1', 1.0)
mat.isotropic = ['H1']
To treat all nuclides in a material this way, the Material.make_isotropic_in_lab method can still be used.
The initializers for openmc.Intersection and openmc.Union now expect an iterable.
Auto-generated unique IDs for classes now start from 1 rather than 10000.
NOTE: This is the last release of OpenMC that will support Python 2.7. Future releases of OpenMC will require Python 3.4 or later.
This release contains new contributions from the following people: