diff --git a/docs/source/pythonapi/examples/multi-group-cross-sections.ipynb b/docs/source/pythonapi/examples/multi-group-cross-sections.ipynb index 0c0250437..cbd81f154 100644 --- a/docs/source/pythonapi/examples/multi-group-cross-sections.ipynb +++ b/docs/source/pythonapi/examples/multi-group-cross-sections.ipynb @@ -478,7 +478,7 @@ " License: http://mit-crpg.github.io/openmc/license.html\n", " Version: 0.7.0\n", " Git SHA1: 23535afa1c69644bb299bde18a094c3b99d53ae0\n", - " Date/Time: 2015-10-09 00:44:07\n", + " Date/Time: 2015-10-09 16:04:33\n", " MPI Processes: 1\n", "\n", " ===========================================================================\n", @@ -563,20 +563,20 @@ "\n", " =======================> TIMING STATISTICS <=======================\n", "\n", - " Total time for initialization = 4.1300E-01 seconds\n", - " Reading cross sections = 1.0500E-01 seconds\n", - " Total time in simulation = 1.3459E+01 seconds\n", - " Time in transport only = 1.3445E+01 seconds\n", - " Time in inactive batches = 2.1740E+00 seconds\n", - " Time in active batches = 1.1285E+01 seconds\n", - " Time synchronizing fission bank = 4.0000E-03 seconds\n", - " Sampling source sites = 3.0000E-03 seconds\n", + " Total time for initialization = 4.3000E-01 seconds\n", + " Reading cross sections = 9.0000E-02 seconds\n", + " Total time in simulation = 1.5218E+01 seconds\n", + " Time in transport only = 1.5180E+01 seconds\n", + " Time in inactive batches = 1.8800E+00 seconds\n", + " Time in active batches = 1.3338E+01 seconds\n", + " Time synchronizing fission bank = 2.0000E-03 seconds\n", + " Sampling source sites = 1.0000E-03 seconds\n", " SEND/RECV source sites = 1.0000E-03 seconds\n", " Time accumulating tallies = 1.0000E-03 seconds\n", - " Total time for finalization = 2.0000E-03 seconds\n", - " Total time elapsed = 1.3882E+01 seconds\n", - " Calculation Rate (inactive) = 11499.5 neutrons/second\n", - " Calculation Rate (active) = 8861.32 neutrons/second\n", + " Total time for finalization = 8.0000E-03 seconds\n", + " Total time elapsed = 1.5664E+01 seconds\n", + " Calculation Rate (inactive) = 13297.9 neutrons/second\n", + " Calculation Rate (active) = 7497.38 neutrons/second\n", "\n", " ============================> RESULTS <============================\n", "\n", @@ -1026,7 +1026,8 @@ "name": "stdout", "output_type": "stream", "text": [ - "[ NORMAL ] Importing ray tracing data from file...\n", + "[ NORMAL ] Ray tracing for track segmentation...\n", + "[ NORMAL ] Dumping tracks to file...\n", "[ NORMAL ] Computing the eigenvalue...\n", "[ NORMAL ] Iteration 0:\tk_eff = 0.685185\tres = 0.000E+00\n", "[ NORMAL ] Iteration 1:\tk_eff = 0.785642\tres = 3.148E-01\n", @@ -1461,7 +1462,7 @@ }, { "cell_type": "code", - "execution_count": null, + "execution_count": 35, "metadata": { "collapsed": false }, @@ -1501,11 +1502,22 @@ }, { "cell_type": "code", - "execution_count": null, + "execution_count": 36, "metadata": { "collapsed": false }, - "outputs": [], + "outputs": [ + { + "data": { + "text/plain": [ + "0" + ] + }, + "execution_count": 36, + "metadata": {}, + "output_type": "execute_result" + } + ], "source": [ "# Delete old HDF5 files\n", "!rm *.h5\n", @@ -1531,7 +1543,7 @@ }, { "cell_type": "code", - "execution_count": null, + "execution_count": 37, "metadata": { "collapsed": false }, @@ -1552,7 +1564,7 @@ }, { "cell_type": "code", - "execution_count": null, + "execution_count": 38, "metadata": { "collapsed": false }, @@ -1588,11 +1600,46 @@ }, { "cell_type": "code", - "execution_count": null, + "execution_count": 39, "metadata": { "collapsed": false }, - "outputs": [], + "outputs": [ + { + "name": "stdout", + "output_type": "stream", + "text": [ + "Multi-Group XS\n", + "\tReaction Type =\tnu-fission\n", + "\tDomain Type =\tcell\n", + "\tDomain ID =\t10000\n", + "\tNuclide =\tU-235\n", + "\tCross Sections [barns]:\n", + " Group 1 [0.821 - 20.0 MeV]:\t3.30e+00 +/- 5.91e-01%\n", + " Group 2 [0.00553 - 0.821 MeV]:\t3.97e+00 +/- 4.03e-01%\n", + " Group 3 [4e-06 - 0.00553 MeV]:\t5.48e+01 +/- 5.56e-01%\n", + " Group 4 [6.25e-07 - 4e-06 MeV]:\t8.84e+01 +/- 8.48e-01%\n", + " Group 5 [2.8e-07 - 6.25e-07 MeV]:\t2.89e+02 +/- 1.25e+00%\n", + " Group 6 [1.4e-07 - 2.8e-07 MeV]:\t4.49e+02 +/- 1.09e+00%\n", + " Group 7 [5.8e-08 - 1.4e-07 MeV]:\t6.87e+02 +/- 7.98e-01%\n", + " Group 8 [0.0 - 5.8e-08 MeV]:\t1.44e+03 +/- 5.73e-01%\n", + "\n", + "\tNuclide =\tU-238\n", + "\tCross Sections [barns]:\n", + " Group 1 [0.821 - 20.0 MeV]:\t1.06e+00 +/- 6.74e-01%\n", + " Group 2 [0.00553 - 0.821 MeV]:\t1.22e-03 +/- 8.28e-01%\n", + " Group 3 [4e-06 - 0.00553 MeV]:\t4.75e-04 +/- 7.97e+00%\n", + " Group 4 [6.25e-07 - 4e-06 MeV]:\t6.53e-06 +/- 7.56e-01%\n", + " Group 5 [2.8e-07 - 6.25e-07 MeV]:\t1.07e-05 +/- 1.22e+00%\n", + " Group 6 [1.4e-07 - 2.8e-07 MeV]:\t1.55e-05 +/- 1.09e+00%\n", + " Group 7 [5.8e-08 - 1.4e-07 MeV]:\t2.30e-05 +/- 7.97e-01%\n", + " Group 8 [0.0 - 5.8e-08 MeV]:\t4.25e-05 +/- 5.72e-01%\n", + "\n", + "\n", + "\n" + ] + } + ], "source": [ "nufission = xs_library[fuel_cell.id]['nu-fission']\n", "nufission.print_xs(xs_type='micro', nuclides=['U-235', 'U-238'])" @@ -1607,11 +1654,34 @@ }, { "cell_type": "code", - "execution_count": null, + "execution_count": 40, "metadata": { "collapsed": false }, - "outputs": [], + "outputs": [ + { + "name": "stdout", + "output_type": "stream", + "text": [ + "Multi-Group XS\n", + "\tReaction Type =\tnu-fission\n", + "\tDomain Type =\tcell\n", + "\tDomain ID =\t10000\n", + "\tCross Sections [cm^-1]:\n", + " Group 1 [0.821 - 20.0 MeV]:\t2.52e-02 +/- 6.42e-01%\n", + " Group 2 [0.00553 - 0.821 MeV]:\t1.52e-03 +/- 3.96e-01%\n", + " Group 3 [4e-06 - 0.00553 MeV]:\t2.06e-02 +/- 5.56e-01%\n", + " Group 4 [6.25e-07 - 4e-06 MeV]:\t3.32e-02 +/- 8.48e-01%\n", + " Group 5 [2.8e-07 - 6.25e-07 MeV]:\t1.09e-01 +/- 1.25e+00%\n", + " Group 6 [1.4e-07 - 2.8e-07 MeV]:\t1.69e-01 +/- 1.09e+00%\n", + " Group 7 [5.8e-08 - 1.4e-07 MeV]:\t2.58e-01 +/- 7.98e-01%\n", + " Group 8 [0.0 - 5.8e-08 MeV]:\t5.41e-01 +/- 5.73e-01%\n", + "\n", + "\n", + "\n" + ] + } + ], "source": [ "nufission = xs_library[fuel_cell.id]['nu-fission']\n", "nufission.print_xs(xs_type='macro', nuclides='sum')" @@ -1626,11 +1696,141 @@ }, { "cell_type": "code", - "execution_count": null, + "execution_count": 41, "metadata": { "collapsed": false }, - "outputs": [], + "outputs": [ + { + "data": { + "text/html": [ + "
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cellgroup ingroup outnuclidemeanstd. dev.
1261000211O-161.5600980.017801
1271000211H-10.2348770.010096
1241000212O-160.2882360.004397
1251000212H-11.5878150.007847
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1231000213H-10.0101220.000513
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\n", + "
" + ], + "text/plain": [ + " cell group in group out nuclide mean std. dev.\n", + "126 10002 1 1 O-16 1.560098 0.017801\n", + "127 10002 1 1 H-1 0.234877 0.010096\n", + "124 10002 1 2 O-16 0.288236 0.004397\n", + "125 10002 1 2 H-1 1.587815 0.007847\n", + "122 10002 1 3 O-16 0.000000 0.000000\n", + "123 10002 1 3 H-1 0.010122 0.000513\n", + "120 10002 1 4 O-16 0.000000 0.000000\n", + "121 10002 1 4 H-1 0.000000 0.000000\n", + "118 10002 1 5 O-16 0.000000 0.000000\n", + "119 10002 1 5 H-1 0.000000 0.000000" + ] + }, + "execution_count": 41, + "metadata": {}, + "output_type": "execute_result" + } + ], "source": [ "nuscatter = xs_library[moderator_cell.id]['nu-scatter']\n", "df = nuscatter.get_pandas_dataframe(xs_type='micro')\n", @@ -1646,7 +1846,7 @@ }, { "cell_type": "code", - "execution_count": null, + "execution_count": 42, "metadata": { "collapsed": false }, @@ -1674,11 +1874,22 @@ }, { "cell_type": "code", - "execution_count": null, + "execution_count": 43, "metadata": { "collapsed": false }, - "outputs": [], + "outputs": [ + { + "data": { + "image/png": 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+ "text/plain": [ + "" + ] + }, + "metadata": {}, + "output_type": "display_data" + } + ], "source": [ "# Create plot of the H-1 scattering matrix\n", "fig = plt.subplot(121)\n", @@ -1703,7 +1914,7 @@ }, { "cell_type": "code", - "execution_count": null, + "execution_count": 44, "metadata": { "collapsed": true }, @@ -1725,22 +1936,133 @@ }, { "cell_type": "code", - "execution_count": null, + "execution_count": 45, "metadata": { "collapsed": false }, - "outputs": [], + "outputs": [ + { + "name": "stdout", + "output_type": "stream", + "text": [ + "Multi-Group XS\n", + "\tReaction Type =\ttransport\n", + "\tDomain Type =\tcell\n", + "\tDomain ID =\t10000\n", + "\tNuclide =\tU-238\n", + "\tCross Sections [cm^-1]:\n", + " Group 1 [6.25e-07 - 20.0 MeV]:\t2.17e-01 +/- 4.04e-01%\n", + " Group 2 [0.0 - 6.25e-07 MeV]:\t2.53e-01 +/- 5.85e-01%\n", + "\n", + "\tNuclide =\tO-16\n", + "\tCross Sections [cm^-1]:\n", + " Group 1 [6.25e-07 - 20.0 MeV]:\t1.45e-01 +/- 4.10e-01%\n", + " Group 2 [0.0 - 6.25e-07 MeV]:\t1.75e-01 +/- 6.46e-01%\n", + "\n", + "\tNuclide =\tU-235\n", + "\tCross Sections [cm^-1]:\n", + " Group 1 [6.25e-07 - 20.0 MeV]:\t7.91e-03 +/- 1.22e+00%\n", + " Group 2 [0.0 - 6.25e-07 MeV]:\t1.82e-01 +/- 4.98e-01%\n", + "\n", + "\n", + "\n" + ] + } + ], "source": [ "condense_xs.print_xs()" ] }, { "cell_type": "code", - "execution_count": null, + "execution_count": 46, "metadata": { "collapsed": false }, - "outputs": [], + "outputs": [ + { + "data": { + "text/html": [ + "
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2100002U-235485.5135302.418761
\n", + "
" + ], + "text/plain": [ + " cell group in nuclide mean std. dev.\n", + "3 10000 1 U-238 9.589323 0.038756\n", + "4 10000 1 O-16 3.159101 0.012939\n", + "5 10000 1 U-235 21.095256 0.257787\n", + "0 10000 2 U-238 11.178844 0.065428\n", + "1 10000 2 O-16 3.800027 0.024538\n", + "2 10000 2 U-235 485.513530 2.418761" + ] + }, + "execution_count": 46, + "metadata": {}, + "output_type": "execute_result" + } + ], "source": [ "df = condense_xs.get_pandas_dataframe(xs_type='micro')\n", "df" @@ -1762,7 +2084,7 @@ }, { "cell_type": "code", - "execution_count": null, + "execution_count": 47, "metadata": { "collapsed": false }, @@ -1784,7 +2106,7 @@ }, { "cell_type": "code", - "execution_count": null, + "execution_count": 48, "metadata": { "collapsed": false }, @@ -1832,7 +2154,7 @@ }, { "cell_type": "code", - "execution_count": null, + "execution_count": 49, "metadata": { "collapsed": false }, @@ -1860,11 +2182,21 @@ }, { "cell_type": "code", - "execution_count": null, + "execution_count": 50, "metadata": { "collapsed": false }, - "outputs": [], + "outputs": [ + { + "name": "stdout", + "output_type": "stream", + "text": [ + "openmc keff = 1.227616\n", + "openmoc keff = 1.225325\n", + "bias [pcm]: -229.1\n" + ] + } + ], "source": [ "# Print report of keff and bias with OpenMC\n", "openmoc_keff = solver.getKeff()\n", @@ -1885,7 +2217,7 @@ }, { "cell_type": "code", - "execution_count": null, + "execution_count": 51, "metadata": { "collapsed": false }, @@ -1926,7 +2258,7 @@ }, { "cell_type": "code", - "execution_count": null, + "execution_count": 52, "metadata": { "collapsed": false }, @@ -1944,11 +2276,21 @@ }, { "cell_type": "code", - "execution_count": null, + "execution_count": 53, "metadata": { "collapsed": false }, - "outputs": [], + "outputs": [ + { + "name": "stdout", + "output_type": "stream", + "text": [ + "openmc keff = 1.227616\n", + "openmoc keff = 1.227096\n", + "bias [pcm]: -52.0\n" + ] + } + ], "source": [ "# Print report of keff and bias with OpenMC\n", "openmoc_keff = solver.getKeff()\n", diff --git a/openmc/mgxs/library.py b/openmc/mgxs/library.py index 6f9d40a55..a7276bc9e 100644 --- a/openmc/mgxs/library.py +++ b/openmc/mgxs/library.py @@ -12,6 +12,51 @@ if sys.version_info[0] >= 3: class Library(object): + """A multi-group cross section library for some energy group structure. + + This class can be used for both OpenMC input generation and tally data + post-processing to compute spatially-homogenized and energy-integrated + multi-group cross sections for deterministic neutronics calculations. + + This class helps automate the generation of MGXS objects for some energy + group structure and domain type. The Library serves as a collection for + MGXS objects with routines to automate the initialization of tallies for + input files, the loading of tally data from statepoint files, data storage, + energy group condensation and more. + + Parameters + ---------- + openmc_geometry : openmc.Geometry + An geometry which has been initialized with a root universe + by_nuclide : bool + If true, computes cross sections for each nuclide in each domain + mgxs_types : Iterable of str + The types of cross sections in the library (e.g., ['total', 'scatter']) + name : str, optional + Name of the multi-group cross section. library Used as a label to + identify tallies in OpenMC 'tallies.xml' file. + + Attributes + ---------- + openmc_geometry : openmc.Geometry + An geometry which has been initialized with a root universe + by_nuclide : bool + If true, computes cross sections for each nuclide in each domain + mgxs_types : Iterable of str + The types of cross sections in the library (e.g., ['total', 'scatter']) + domain_type : {'material', 'cell', 'distribcell', 'universe'} + Domain type for spatial homogenization + energy_groups : EnergyGroups + Energy group structure for energy condensation + all_mgxs : dict + MGXS objects keyed by domain ID and cross section type + statepoint : openmc.StatePoint + The statepoint with tally data used to the compute cross sections + name : str, optional + Name of the multi-group cross section library. Used as a label to + identify tallies in OpenMC 'tallies.xml' file. + + """ def __init__(self, openmc_geometry, by_nuclide=False, mgxs_types=None, name=''): @@ -23,6 +68,7 @@ class Library(object): self._domain_type = None self._energy_groups = None self._all_mgxs = {} + self._statepoint = None self.name = name self.openmc_geometry = openmc_geometry @@ -104,6 +150,10 @@ class Library(object): def all_mgxs(self): return self._all_mgxs + @property + def statepoint(self): + return self._statepoint + @openmc_geometry.setter def openmc_geometry(self, openmc_geometry): cv.check_type('openmc_geometry', openmc_geometry, openmc.Geometry) @@ -140,7 +190,13 @@ class Library(object): self._energy_groups = energy_groups def build_library(self): - """ + """Initialize MGXS objects in each domain and for each reaction type + in the library. + + This routine will populate the all_mgxs instance attribute dictionary + with MGXS subclass objects keyed by each domain ID (e.g., Material IDs) + and cross section type (e.g., 'nu-fission', 'total', etc.). + """ # Initialize MGXS for each domain and mgxs type and store in dictionary @@ -155,13 +211,20 @@ class Library(object): mgxs.create_tallies() self.all_mgxs[domain.id][mgxs_type] = mgxs - def add_to_tallies_file(self, tallies_file): - """ + def add_to_tallies_file(self, tallies_file, merge=True): + """Add all tallies from all MGXS objects to a tallies file. NOTE: This assumes that build_library() has been called - :param tallies_file: - :return: + Parameters + ---------- + tallies_file : openmc.TalliesFile + A TalliesFile object to add each MGXS' tallies to generate a + "tallies.xml" input file for OpenMC + merge : bool + Indicate whether tallies should be merged when possible. Defaults + to True. + """ cv.check_type('tallies_file', tallies_file, openmc.TalliesFile) @@ -171,17 +234,39 @@ class Library(object): for mgxs_type in self.mgxs_types: mgxs = self.get_mgxs(domain, mgxs_type) for tally_id, tally in mgxs.tallies.items(): - tallies_file.add_tally(tally, merge=True) + tallies_file.add_tally(tally, merge=merge) def load_from_statepoint(self, statepoint): - """ + """Extracts tallies in an OpenMC StatePoint with the data needed to + compute multi-group cross sections. + + This method is needed to compute cross section data from tallies + in an OpenMC StatePoint object. + + NOTE: The statepoint must first be linked with an OpenMC Summary object. + + Parameters + ---------- + statepoint : openmc.StatePoint + An OpenMC StatePoint object with tally data + + Raises + ------ + ValueError + When this method is called with a statepoint that has not been + linked with a summary object. - :param statepoint: - :return: """ cv.check_type('statepoint', statepoint, openmc.StatePoint) + if not statepoint.with_summary: + msg = 'Unable to load data from a statepoint which has not been ' \ + 'linked with a summary file' + raise ValueError(msg) + + self._statepoint = statepoint + # Load tallies for each MGXS for each domain and mgxs type for domain in self.domains: for mgxs_type in self.mgxs_types: @@ -190,11 +275,33 @@ class Library(object): mgxs.compute_xs() def get_mgxs(self, domain, mgxs_type): - """ + """Return the MGXS object for some domain and reaction rate type. + + This routine searches the library for an MGXS object for the spatial + domain and reaction rate type requ + + NOTE: This routine must be called after the build_library() routine. + + Parameters + ---------- + domain : Material or Cell or Universe or Integral + The material, cell, or universe object of interest (or its ID) + mgxs_type : {'total', 'transport', 'absorption', 'capture', 'fission', + 'nu-fission', 'scatter', 'nu-scatter', 'scatter matrix', + 'nu-scatter matrix', 'chi'} + The type of multi-group cross section object to return + + Returns + ------- + openmc.mgxs.MGXS + The MGXS object for the requested domain and reaction rate type + + Raises + ------ + ValueError + If no MGXS object can be found for the requested domain or + multi-group cross section type - :param domain: - :param mgxs_type: - :return: """ if self.domain_type == 'material': @@ -225,15 +332,38 @@ class Library(object): return self.all_mgxs[domain_id][mgxs_type] def get_condensed_library(self, coarse_groups): + """Construct an energy-condensed version of this library. + + This routine condense each of the multi-group cross sections in the + library to a coarse energy group structure. NOTE: This routine must + be called after the load_from_statepoint(...) routine loads the tallies + from the statepoint into each of the cross sections. + + Parameters + ---------- + coarse_groups : openmc.mgxs.EnergyGroups + The coarse energy group structure of interest + + Returns + ------- + Library + A new multi-group cross section library condensed to the group + structure of interest + + Raises + ------ + ValueError + When this method is called before a statepoint has been loaded + + See also + -------- + MGXS.get_condensed_xs(coarse_groups) + """ - :param coarse_groups: - :return: - """ - - if self.energy_groups is None: + if self.statepoint is None: msg = 'Unable to get a condensed coarse group cross section ' \ - 'library since the fine energy groups have not yet been set' + 'library since the statepoint has not yet been loaded' raise ValueError(msg) cv.check_type('coarse_groups', coarse_groups, openmc.mgxs.EnergyGroups) @@ -260,26 +390,42 @@ class Library(object): def build_hdf5_store(self, filename='mgxs', directory='mgxs', xs_type='macro'): """Export the multi-group cross section library to an HDF5 binary file. - This method constructs an HDF5 file which stores the multi-group - cross section data. The data is stored in a hierarchy of HDF5 groups - from the domain type, domain id, subdomain id (for distribcell domains), - nuclides and cross section types. Two datasets for the mean and standard - deviation are stored for each subdomain entry in the HDF5 file. + This method constructs an HDF5 file which stores the library's + multi-group cross section data. The data is stored in a hierarchy of + HDF5 groups from the domain type, domain id, subdomain id (for + distribcell domains), nuclides and cross section types. Two datasets for + the mean and standard deviation are stored for each subdomain entry in + the HDF5 file. NOTE: This requires the h5py Python package. Parameters ---------- filename : str - Filename for the HDF5 file (default is 'mgxs') + Filename for the HDF5 file. Defaults to 'mgxs'. directory : str - Directory for the HDF5 file (default is 'mgxs') + Directory for the HDF5 file. Defaults to 'mgxs'. xs_type: {'macro', 'micro'} - Store the macro or micro cross section in units of cm^-1 or barns + Store the macro or micro cross section in units of cm^-1 or barns. + Defaults to 'macro'. + + Raises + ------ + ValueError + When this method is called before a statepoint has been loaded + + See also + -------- + MGXS.build_hdf5_store(filename, directory, xs_type) """ - # Load tallies for each MGXS for each domain and mgxs type + if self.statepoint is None: + msg = 'Unable to get a condensed coarse group cross section ' \ + 'library since a statepoint has not yet been loaded' + raise ValueError(msg) + + # Export MGXS for each domain and mgxs type to an HDF5 file for domain in self.domains: for mgxs_type in self.mgxs_types: mgxs = self.all_mgxs[domain.id][mgxs_type] diff --git a/openmc/mgxs/mgxs.py b/openmc/mgxs/mgxs.py index 940586f32..20a9f3d6b 100644 --- a/openmc/mgxs/mgxs.py +++ b/openmc/mgxs/mgxs.py @@ -242,10 +242,11 @@ class MGXS(object): energy_groups : EnergyGroups The energy group structure for energy condensation by_nuclide : bool - If true, computes cross sections for each nuclide in domain + If true, computes cross sections for each nuclide in domain. + Defaults to False name : str, optional Name of the multi-group cross section. Used as a label to identify - tallies in OpenMC 'tallies.xml' file. + tallies in OpenMC 'tallies.xml' file. Defaults to the empty string. Returns ------- @@ -351,7 +352,8 @@ class MGXS(object): A list of nuclide name strings (e.g., ['U-235', 'U-238']). The special string 'all' will return the atom densities for all nuclides in the spatial domain. The special string 'sum' will return the atom - density summed across all nuclides in the spatial domain. + density summed across all nuclides in the spatial domain. Defaults + to 'all'. Returns ------- @@ -544,22 +546,24 @@ class MGXS(object): Parameters ---------- groups : Iterable of Integral or 'all' - Energy groups of interest + Energy groups of interest. Defaults to 'all'. subdomains : Iterable of Integral or 'all' - Subdomain IDs of interest + Subdomain IDs of interest. Defaults to 'all'. nuclides : Iterable of str or 'all' or 'sum' A list of nuclide name strings (e.g., ['U-235', 'U-238']). The - special string 'all' (default) will return the cross sections for - all nuclides in the spatial domain. The special string 'sum' will - return the cross section summed over all nuclides. + special string 'all' will return the cross sections for all nuclides + in the spatial domain. The special string 'sum' will return the + cross section summed over all nuclides. Defaults to 'all'. xs_type: {'macro', 'micro'} - Return the macro or micro cross section in units of cm^-1 or barns + Return the macro or micro cross section in units of cm^-1 or barns. + Defaults to 'macro'. order_groups: {'increasing', 'decreasing'} - Return the cross section indexed according to increasing (default) - or decreasing energy groups (decreasing or increasing energies) + Return the cross section indexed according to increasing or + decreasing energy groups (decreasing or increasing energies). + Defaults to 'increasing'. value : str - A string for the type of value to return - 'mean' (default), - 'std_dev' or 'rel_err' are accepted + A string for the type of value to return - 'mean', 'std_dev' or + 'rel_err' are accepted. Defaults to 'mean'. Returns ------- @@ -739,7 +743,7 @@ class MGXS(object): Parameters ---------- subdomains : Iterable of Integral or 'all' - The subdomain IDs to average across + The subdomain IDs to average across. Defaults to 'all'. Returns ------- @@ -815,15 +819,17 @@ class MGXS(object): Parameters ---------- subdomains : Iterable of Integral or 'all' - The subdomain IDs of the cross sections to include in the report + The subdomain IDs of the cross sections to include in the report. + Defaults to 'all'. nuclides : Iterable of str or 'all' or 'sum' The nuclides of the cross-sections to include in the report. This may be a list of nuclide name strings (e.g., ['U-235', 'U-238']). - The special string 'all' (default) will report the cross sections - for all nuclides in the spatial domain. The special string 'sum' - will report the cross sections summed over all nuclides. + The special string 'all' will report the cross sections for all + nuclides in the spatial domain. The special string 'sum' will report + the cross sections summed over all nuclides. Defaults to 'all'. xs_type: {'macro', 'micro'} - Return the macro or micro cross section in units of cm^-1 or barns + Return the macro or micro cross section in units of cm^-1 or barns. + Defaults to 'macro'. """ @@ -912,14 +918,15 @@ class MGXS(object): Parameters ---------- filename : str - Filename for the HDF5 file (default is 'mgxs') + Filename for the HDF5 file. Defaults to 'mgxs'. directory : str - Directory for the HDF5 file (default is 'mgxs') + Directory for the HDF5 file. Defaults to 'mgxs'. xs_type: {'macro', 'micro'} - Store the macro or micro cross section in units of cm^-1 or barns + Store the macro or micro cross section in units of cm^-1 or barns. + Defaults to 'macro'. append : boolean If true, appends to an existing HDF5 file with the same filename - directory (if one exists) + directory (if one exists). Defaults to True. Raises ------ @@ -1030,15 +1037,16 @@ class MGXS(object): Parameters ---------- filename : str - Filename for the exported file (default is 'mgxs') + Filename for the exported file. Defaults to 'mgxs'. directory : str - Directory for the exported file (default is 'mgxs') + Directory for the exported file. Defaults to 'mgxs'. format : {'csv', 'excel', 'pickle', 'latex'} - The format for the exported data file + The format for the exported data file. Defaults to 'csv'. groups : Iterable of Integral or 'all' - Energy groups of interest + Energy groups of interest. Defaults to 'all'. xs_type: {'macro', 'micro'} - Store the macro or micro cross section in units of cm^-1 or barns + Store the macro or micro cross section in units of cm^-1 or barns. + Defaults to 'macro'. """ @@ -1100,15 +1108,17 @@ class MGXS(object): Parameters ---------- groups : Iterable of Integral or 'all' - Energy groups of interest + Energy groups of interest. Defaults to 'all'. nuclides : Iterable of str or 'all' or 'sum' The nuclides of the cross-sections to include in the dataframe. This may be a list of nuclide name strings (e.g., ['U-235', 'U-238']). - The special string 'all' (default) will include the cross sections - for all nuclides in the spatial domain. The special string 'sum' - will include the cross sections summed over all nuclides. + The special string 'all' will include the cross sections for all + nuclides in the spatial domain. The special string 'sum' will + include the cross sections summed over all nuclides. Defaults + to 'all'. xs_type: {'macro', 'micro'} - Return macro or micro cross section in units of cm^-1 or barns + Return macro or micro cross section in units of cm^-1 or barns. + Defaults to 'macro'. summary : None or Summary An optional Summary object to be used to construct columns for distribcell tally filters (default is None). The geometric @@ -1628,24 +1638,26 @@ class ScatterMatrixXS(MGXS): Parameters ---------- in_groups : Iterable of Integral or 'all' - Incoming energy groups of interest + Incoming energy groups of interest. Defaults to 'all'. out_groups : Iterable of Integral or 'all' - Outgoing energy groups of interest + Outgoing energy groups of interest. Defaults to 'all'. subdomains : Iterable of Integral or 'all' - Subdomain IDs of interest + Subdomain IDs of interest. Defaults to 'all'. nuclides : Iterable of str or 'all' or 'sum' A list of nuclide name strings (e.g., ['U-235', 'U-238']). The - special string 'all' (default) will return the cross sections for - all nuclides in the spatial domain. The special string 'sum' will - return the cross section summed over all nuclides. + special string 'all' will return the cross sections for all nuclides + in the spatial domain. The special string 'sum' will return the + cross section summed over all nuclides. Defaults to 'all'. xs_type: {'macro', 'micro'} - Return the macro or micro cross section in units of cm^-1 or barns + Return the macro or micro cross section in units of cm^-1 or barns. + Defaults to 'macro'. order_groups: {'increasing', 'decreasing'} - Return the cross section indexed according to increasing (default) - or decreasing energy groups (decreasing or increasing energies) + Return the cross section indexed according to increasing or + decreasing energy groups (decreasing or increasing energies). + Defaults to 'increasing'. value : str - A string for the type of value to return - 'mean' (default), - 'std_dev' or 'rel_err' are accepted + A string for the type of value to return - 'mean', 'std_dev', or + 'rel_err' are accepted. Defaults to the empty string. Returns ------- @@ -1756,15 +1768,17 @@ class ScatterMatrixXS(MGXS): Parameters ---------- subdomains : Iterable of Integral or 'all' - The subdomain IDs of the cross sections to include in the report + The subdomain IDs of the cross sections to include in the report. + Defaults to 'all'. nuclides : Iterable of str or 'all' or 'sum' The nuclides of the cross-sections to include in the report. This may be a list of nuclide name strings (e.g., ['U-235', 'U-238']). - The special string 'all' (default) will report the cross sections - for all nuclides in the spatial domain. The special string 'sum' - will report the cross sections summed over all nuclides. + The special string 'all' will report the cross sections for all + nuclides in the spatial domain. The special string 'sum' will report + the cross sections summed over all nuclides. Defaults to 'all'. xs_type: {'macro', 'micro'} - Return the macro or micro cross section in units of cm^-1 or barns + Return the macro or micro cross section in units of cm^-1 or barns. + Defaults to 'macro'. """ @@ -1952,23 +1966,24 @@ class Chi(MGXS): Parameters ---------- groups : Iterable of Integral or 'all' - Energy groups of interest + Energy groups of interest. Defaults to 'all'. subdomains : Iterable of Integral or 'all' - Subdomain IDs of interest + Subdomain IDs of interest. Defaults to 'all'. nuclides : Iterable of str or 'all' or 'sum' A list of nuclide name strings (e.g., ['U-235', 'U-238']). The - special string 'all' (default) will return the cross sections for - all nuclides in the spatial domain. The special string 'sum' will - return the cross section summed over all nuclides. + special string 'all' will return the cross sections for all nuclides + in the spatial domain. The special string 'sum' will return the + cross section summed over all nuclides. Defaults to 'all'. xs_type: {'macro', 'micro'} This parameter is not relevant for chi but is included here to mirror the parent MGXS.get_xs(...) class method order_groups: {'increasing', 'decreasing'} - Return the cross section indexed according to increasing (default) - or decreasing energy groups (decreasing or increasing energies) + Return the cross section indexed according to increasing or + decreasing energy groups (decreasing or increasing energies). + Defaults to 'increasing'. value : str - A string for the type of value to return - 'mean' (default), - 'std_dev' or 'rel_err' are accepted + A string for the type of value to return - 'mean', 'std_dev', or + 'rel_err' are accepted. Defaults to 'mean'. Returns ------- @@ -2082,15 +2097,17 @@ class Chi(MGXS): Parameters ---------- groups : Iterable of Integral or 'all' - Energy groups of interest + Energy groups of interest. Defaults to 'all'. nuclides : Iterable of str or 'all' or 'sum' The nuclides of the cross-sections to include in the dataframe. This may be a list of nuclide name strings (e.g., ['U-235', 'U-238']). - The special string 'all' (default) will include the cross sections - for all nuclides in the spatial domain. The special string 'sum' - will include the cross sections summed over all nuclides. + The special string 'all' will include the cross sections for all + nuclides in the spatial domain. The special string 'sum' will + include the cross sections summed over all nuclides. Defaults to + 'all'. xs_type: {'macro', 'micro'} - Return macro or micro cross section in units of cm^-1 or barns + Return macro or micro cross section in units of cm^-1 or barns. + Defaults to 'macro'. summary : None or Summary An optional Summary object to be used to construct columns for distribcell tally filters (default is None). The geometric