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89 lines
3.7 KiB
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===================
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What's New in 0.7.1
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===================
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This release of OpenMC provides some substantial improvements over version
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0.7.0. Non-simple cell regions can now be defined through the ``|`` (union) and
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``~`` (complement) operators. Similar changes in the Python API also allow
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complex cell regions to be defined. A true secondary particle bank now exists;
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this is crucial for photon transport (to be added in the next minor release). A
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rich API for multi-group cross section generation has been added via the
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``openmc.mgxs`` Python module.
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Various improvements to tallies have also been made. It is now possible to
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explicitly specify that a collision estimator be used in a tally. A new
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``delayedgroup`` filter and ``delayed-nu-fission`` score allow a user to obtain
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delayed fission neutron production rates filtered by delayed group. Finally, the
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new ``inverse-velocity`` score may be useful for calculating kinetics
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parameters.
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.. caution:: In previous versions, depending on how OpenMC was compiled binary
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output was either given in HDF5 or a flat binary format. With this
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version, all binary output is now HDF5 which means you **must**
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have HDF5 in order to install OpenMC. Please consult the user's
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guide for instructions on how to compile with HDF5.
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-------------------
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System Requirements
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-------------------
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There are no special requirements for running the OpenMC code. As of this
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release, OpenMC has been tested on a variety of Linux distributions, Mac OS X,
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and Microsoft Windows 7. Memory requirements will vary depending on the size of
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the problem at hand (mostly on the number of nuclides in the problem).
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------------
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New Features
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------------
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- Support for complex cell regions (union and complement operators)
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- Generic quadric surface type
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- Improved handling of secondary particles
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- Binary output is now solely HDF5
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- ``openmc.mgxs`` Python module enabling multi-group cross section generation
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- Collision estimator for tallies
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- Delayed fission neutron production tallies with ability to filter by delayed
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group
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- Inverse velocity tally score
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- Performance improvements for binary search
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- Performance improvements for reaction rate tallies
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---------
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Bug Fixes
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---------
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- 299322_: Bug with material filter when void material present
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- d74840_: Fix triggers on tallies with multiple filters
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- c29a81_: Correctly handle maximum transport energy
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- 3edc23_: Fixes in the nu-scatter score
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- 629e3b_: Assume unspecified surface coefficients are zero in Python API
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- 5dbe8b_: Fix energy filters for openmc-plot-mesh-tally
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- ff66f4_: Fixes in the openmc-plot-mesh-tally script
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- 441fd4_: Fix bug in kappa-fission score
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- 7e5974_: Allow fixed source simulations from Python API
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.. _299322: https://github.com/openmc-dev/openmc/commit/299322
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.. _d74840: https://github.com/openmc-dev/openmc/commit/d74840
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.. _c29a81: https://github.com/openmc-dev/openmc/commit/c29a81
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.. _3edc23: https://github.com/openmc-dev/openmc/commit/3edc23
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.. _629e3b: https://github.com/openmc-dev/openmc/commit/629e3b
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.. _5dbe8b: https://github.com/openmc-dev/openmc/commit/5dbe8b
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.. _ff66f4: https://github.com/openmc-dev/openmc/commit/ff66f4
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.. _441fd4: https://github.com/openmc-dev/openmc/commit/441fd4
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.. _7e5974: https://github.com/openmc-dev/openmc/commit/7e5974
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------------
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Contributors
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------------
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This release contains new contributions from the following people:
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- `Will Boyd <wbinventor@gmail.com>`_
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- `Sterling Harper <sterlingmharper@mit.edu>`_
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- `Bryan Herman <hermab53@gmail.com>`_
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- `Colin Josey <cjosey@mit.edu>`_
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- `Adam Nelson <nelsonag@umich.edu>`_
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- `Paul Romano <paul.k.romano@gmail.com>`_
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- `Kelly Rowland <kellylynnerowland@gmail.com>`_
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- `Sam Shaner <samuelshaner@gmail.com>`_
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- `Jon Walsh <walshjon@mit.edu>`_
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