openmc: OpenMC 0.7.1
Romano, Paul, Will Boyd, Sterling M. Harper, Nelson, Adam, Jonathan A. Walsh, Sam, Nicholas E. Horelik, Colin J. Josey, John Lyu, tpviitan, Ellis, Matt, Locke, Kelly Rowland, John Xia, Bryan R. Herman, Anton Leontiev, Anthony Scopatz · Zenodo (CERN European Organization for Nuclear Research) · 2015
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. 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 generation Collision 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 2993228a12db933da376c0d8b9c63b9f99d60359 Bug with material filter when void material present d748406f20e52d74df5ae5c567d3cf7ce19bfeb6 Fix triggers on tallies with multiple filters c29a811f3e7dc8fe62e0341ab3df4d86696649a0 Correctly handle maximum transport energy 3edc2389e588f4f2348c00105a4fb067dbb54364 Fixes in the nu-scatter score 629e3b21569d990e00f76ca93a79366ab511816a Assume unspecified surface coefficients are zero in Python API 5dbe8b75db638b28b1d1714a47cc16b879061b36 Fix energy filters for openmc-plot-mesh-tally ff66f41d89fcb5dcd4b9c523bdfcc741d6bd4025 Fixes in the openmc-plot-mesh-tally script 441fd4f00dfb3cd6f79abc0ad2887b04dd5dbfd8 Fix bug in kappa-fission score 7e5974a23356c4c9fc547a64ee483acda1fc34ea Allow fixed source simulations from Python API Contributors This release contains new contributions from the following people: @bhermanmit @cjosey @kellyrowland @nelsonag @paulromano @samuelshaner @smharper @walshjon @wbinventor