What’s New in 0.7.1¶
This release of OpenMC provides some substantial improvements over version
0.7.0. Non-simple cell regions can now be defined through the |
(union) and
~
(complement) operators. Similar changes in the Python API also allow
complex cell regions to be defined. A true secondary particle bank now exists;
this is crucial for photon transport (to be added in the next minor release). A
rich API for multi-group cross section generation has been added via the
openmc.mgxs
Python module.
Various improvements to tallies have also been made. It is now possible to
explicitly specify that a collision estimator be used in a tally. A new
delayedgroup
filter and delayed-nu-fission
score allow a user to obtain
delayed fission neutron production rates filtered by delayed group. Finally, the
new inverse-velocity
score may be useful for calculating kinetics
parameters.
Caution
In previous versions, depending on how OpenMC was compiled binary output was either given in HDF5 or a flat binary format. With this version, all binary output is now HDF5 which means you must have HDF5 in order to install OpenMC. Please consult the user’s guide for instructions on how to compile with HDF5.
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, Mac OS X, and Microsoft Windows 7. Memory requirements will vary depending on the size of the problem at hand (mostly on the number of nuclides in the problem).
New Features¶
Support for complex cell regions (union and complement operators)
Generic quadric surface type
Improved handling of secondary particles
Binary output is now solely HDF5
openmc.mgxs
Python module enabling multi-group cross section generationCollision estimator for tallies
Delayed fission neutron production tallies with ability to filter by delayed group
Inverse velocity tally score
Performance improvements for binary search
Performance improvements for reaction rate tallies
Bug Fixes¶
299322: Bug with material filter when void material present
d74840: Fix triggers on tallies with multiple filters
c29a81: Correctly handle maximum transport energy
3edc23: Fixes in the nu-scatter score
629e3b: Assume unspecified surface coefficients are zero in Python API
5dbe8b: Fix energy filters for openmc-plot-mesh-tally
ff66f4: Fixes in the openmc-plot-mesh-tally script
441fd4: Fix bug in kappa-fission score
7e5974: Allow fixed source simulations from Python API
Contributors¶
This release contains new contributions from the following people: