What’s New in 0.4.2¶
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).
- Ability to specify void materials.
- Option to not reduce tallies across processors at end of each batch.
- Uniform fission site method for reducing variance on local tallies.
- Reading/writing binary source files.
- Added more messages for <trace> or high verbosity.
- Estimator for diffusion coefficient.
- Ability to specify ‘point’ source type.
- Ability to change random number seed.
- Users can now specify units=’sum’ on a <density> tag. This tells the code that the total material density is the sum of the atom fractions listed for each nuclide on the material.
- a27f8f: Fixed runtime error bug when using Intel compiler with DEBUG on.
- afe121: Fixed minor bug in fission bank algorithms.
- e0968e: Force re-evaluation of cross-sections when each particle is born.
- 298db8: Fixed bug in surface currents when using energy-in filter.
- 2f3bbe: Fixed subtle bug in S(a,b) cross section calculation.
- 671f30: Fixed surface currents on mesh not encompassing geometry.
- b2c40e: Fixed bug in incoming energy filter for track-length tallies.
- 5524fd: Mesh filter now works with track-length tallies.
- d050c7: Added Bessel’s correction to make estimate of variance unbiased.
- 2a5b9c: Fixed regression in plotting.