mit-crpg/openmc: OpenMC 0.9.0

This release of OpenMC is the first release to use a new native HDF5 cross section format rather than ACE format cross sections. Other significant new features include a nuclear data interface in the Python API (openmc.data) a stochastic volume calculation capability, a random sphere packing algorithm that can handle packing fractions up to 60%, and a new XML parser with significantly better performance than the parser used previously. CAUTION: With the new cross section format, the default energy units are now electronvolts (eV) rather than megaelectronvolts (MeV)! If you are specifying an energy filter for a tally, make sure you use units of eV now. 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: Each type of tally filter is now specified with a separate class. For example: energy_filter = openmc.EnergyFilter([0.0, 0.625, 4.0, 1.0e6, 20.0e6]) Several attributes of the Plot class have changed (color -> color_by and col_spec > colors). Plot.colors now accepts a dictionary mapping Cell or Material instances to RGB 3-tuples or string colors names, e.g.: plot.colors = { fuel: 'yellow', water: 'blue' } make_hexagon_region is now get_hexagonal_prism Several changes in Settings attributes: weight is now set as Settings.cutoff['weight'] Shannon entropy is now specified by passing a Mesh to Settings.entropy_mesh Uniform fission site method is now specified by passing a Mesh to Settings.ufs_mesh All sourcepoint_* options are now specified in a Settings.sourcepoint dictionary Resonance scattering method is now specified as a dictionary in Settings.resonance_scattering Multipole is now turned on by setting Settings.temperature['multipole'] = True The output_path attribute is now Settings.output['path'] All the openmc.mgxs.Nu* classes are gone. Instead, a nu argument was added to the constructor of the corresponding classes. New Features Stochastic volume calculations Multi-delayed group cross section generation Ability to calculate multi-group cross sections over meshes Temperature interpolation on cross section data Nuclear data interface in Python API, openmc.data Allow cutoff energy via Settings.cutoff Ability to define fuel by enrichment (see Material.add_element) Random sphere packing for TRISO particle generation, openmc.model.pack_trisos Critical eigenvalue search, openmc.search_for_keff Model container, openmc.model.Model In-line plotting in Jupyter, openmc.plot_inline Energy function tally filters, openmc.EnergyFunctionFilter Replaced FoX XML parser with pugixml Cell/material instance counting, Geometry.determine_paths Differential tallies (see openmc.TallyDerivative) Consistent multi-group scattering matrices Improved documentation and new Jupyter notebooks OpenMOC compatibility module, openmc.openmoc_compatible Bug Fixes c5df6ce146abeee0d83447aa7a1deccf354b9ade Fix mesh filter max iterator check 1cfa392bb0580b584643927fb3e57f43f28b9f12 Reject external source only if 95% of sites are rejected 3353592cb6092f9d0341e1a0393cedca778aefaf Fix bug in plotting meshlines 17c678d9f8f797a2d558162107c8eb49dded714f Make sure system_clock uses high-resolution timer 23ec0b89bb58a10017d56e651415886ac35a1c6b Fix use of S(a,b) with multipole data 7eefb7306ff0f8745af1447789d8cddc9778617f Fix several bugs in tally module 7880d4f2461a945a99ffd4fb5a88a92191a06f5c Allow plotting calculation with no boundary conditions ad2d9fff55860625f576b9589bd334b80acee337 Fix filter weight missing when scoring all nuclides 59fdcac2a91887cf5ce3960980d79f086ab51ae0 Fix use of source files for fixed source calculations 9eff5b8a2ead933a6d9839b8c473fd5c82922f59 Fix thermal scattering bugs 7848a97edcdb05bee7b8424568c75a6900b068ab Fix combined k-eff estimator producing NaN f139ce8dc12ae036e73ddf46c9ed5ba1a563be1c Fix printing bug for tallies with AggregateNuclide b8ddfacaf34aedf689cfef26ae954cf6025e1dda Bugfix for short tracks near tally mesh edges ec3cfb5bab75b67f0a86ab63cf08073305cd9663 Fix inconsistency in filter weights 5e9b06a861d4f596314eff490ad63c051f833f3a Fix XML representation for verbosity c39990accb6d0377fc05f004b0809d08e7a7f384 Fix bug tallying reaction rates with multipole on c6b67e64434c15483a26733eadbb7335b10be7ea Fix fissionable source sampling bug 48954027704d1413f62addf11bfdd072b33713fc Check for void materials in tracklength tallies f0214f4c12450c82788ed0546379d9bcae3174f2 Fixes/improvements to the ARES algorithm Contributors This release contains new contributions from the following people: @wbinventor @smharper @QingmingHe @cjosey @tjlaboss @liangjg @amandalund @nelsonag @paulromano @samuelshaner @walshjo