# What’s New in 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 openmc.Mesh to Settings.entropy_mesh
• Uniform fission site method is now specified by passing a openmc.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.

## System Requirements¶

There are no special requirements for running the OpenMC code. As of this release, OpenMC has been tested on a variety of Linux distributions and Mac OS X. Numerous users have reported working builds on Microsoft Windows, but your mileage may vary. Memory requirements will vary depending on the size of the problem at hand (mostly on the number of nuclides and tallies in the problem).

## Bug Fixes¶

• c5df6c: Fix mesh filter max iterator check
• 1cfa39: Reject external source only if 95% of sites are rejected
• 335359: Fix bug in plotting meshlines
• 17c678: Make sure system_clock uses high-resolution timer
• 23ec0b: Fix use of S(a,b) with multipole data
• 7eefb7: Fix several bugs in tally module
• 7880d4: Allow plotting calculation with no boundary conditions
• ad2d9f: Fix filter weight missing when scoring all nuclides
• 59fdca: Fix use of source files for fixed source calculations
• 9eff5b: Fix thermal scattering bugs
• 7848a9: Fix combined k-eff estimator producing NaN
• f139ce: Fix printing bug for tallies with AggregateNuclide
• b8ddfa: Bugfix for short tracks near tally mesh edges
• ec3cfb: Fix inconsistency in filter weights
• 5e9b06: Fix XML representation for verbosity
• c39990: Fix bug tallying reaction rates with multipole on
• c6b67e: Fix fissionable source sampling bug
• 489540: Check for void materials in tracklength tallies
• f0214f: Fixes/improvements to the ARES algorithm

## Contributors¶

This release contains new contributions from the following people: