What’s New in 0.11.0¶
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:
Note that these filters replace the use expansion scores like
Instead, a normal
scatter score should be used along with a
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
Predefined energy group structures can be found in
New tally scores:
Switched to cell-based neighor lists (PR 1140)
Two new probability distributions that can be used for source distributions:
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.
Python API Changes¶
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.)
Meshclass has been renamed
get_rectangular_prismfunction has been renamed
get_hexagonal_prismfunction has been renamed
Python bindings to the C/C++ API have been move from
This release contains new contributions from the following people: