What’s New in 0.4.1¶
There are no special requirements for running the OpenMC code. As of this release, OpenMC has been tested on a variety of Linux distributions as well as Mac OS X. However, it has not been tested yet on any releases of Microsoft Windows. Memory requirements will vary depending on the size of the problem at hand (mostly on the number of nuclides in the problem).
A batching method has been implemented so that statistics can be calculated based on multiple generations instead of a single generation. This can help to overcome problems with underpredicted variance in problems where there is correlation between successive fission source iterations.
Users now have the option to select a non-unionized energy grid for problems with many nuclides where the use of a unionized grid is not feasible.
Improved plotting capability (Nick Horelik). The plotting input is now in
Added multiple estimators for k-effective and added a global tally for leakage.
Moved cross section-related output into cross_sections.out.
Improved timing capabilities.
Can now use more than 2**31 - 1 particles per generation.
Improved fission bank synchronization method. This also necessitated changing the source bank to be of type Bank rather than of type Particle.
Added HDF5 output (not complete yet).
Major changes to tally implementation.
b206a8: Fixed subtle error in the sampling of energy distributions.
800742: Fixed error in sampling of angle and rotating angles.
a07c08: Fixed bug in linear-linear interpolation during sampling energy.
a75283: Fixed energy and energyout tally filters to support many bins.
95cfac: Fixed error in cell neighbor searches.
83a803: Fixed bug related to probability tables.