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Removed Nu* MGXS classes.
This commit is contained in:
Adam Nelson
parent
60a1f157da
commit
bd9b32b804
4 changed files with 224 additions and 670 deletions
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@ -369,21 +369,19 @@
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"\n",
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"* `TotalXS`\n",
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"* `TransportXS`\n",
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"* `NuTransportXS`\n",
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"* `AbsorptionXS`\n",
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"* `CaptureXS`\n",
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"* `FissionXS`\n",
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"* `NuFissionXS`\n",
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"* `KappaFissionXS`\n",
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"* `ScatterXS`\n",
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"* `NuScatterXS`\n",
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"* `ScatterMatrixXS`\n",
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"* `NuScatterMatrixXS`\n",
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"* `Chi`\n",
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"* `ChiPrompt`\n",
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"* `InverseVelocity`\n",
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"* `PromptNuFissionXS`\n",
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"\n",
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"Of course, we know that the transport (`TransportXS`), fission (`FissionXS`), scattering (`ScatterXS`), and scattering-matrix (`ScatterMatrixXS`) cross sections can potentially incorporate neutron multiplication ($\\nu$). For these types, the multpiplication can be accomodated by setting the `nu` parameter to `True` as shown below.\n",
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"\n",
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"These classes provide us with an interface to generate the tally inputs as well as perform post-processing of OpenMC's tally data to compute the respective multi-group cross sections. In this case, let's create the multi-group total, absorption and scattering cross sections with our 2-group structure."
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]
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},
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@ -398,7 +396,11 @@
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"# Instantiate a few different sections\n",
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"total = mgxs.TotalXS(domain=cell, groups=groups)\n",
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"absorption = mgxs.AbsorptionXS(domain=cell, groups=groups)\n",
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"scattering = mgxs.ScatterXS(domain=cell, groups=groups)"
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"scattering = mgxs.ScatterXS(domain=cell, groups=groups)\n",
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"\n",
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"# Note that if we wanted to incorporate neutron multiplication in the\n",
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"# scattering cross section we would write the previous line as:\n",
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"# scattering = mgxs.ScatterXS(domain=cell, groups=groups, nu=True)"
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]
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},
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{
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@ -520,8 +522,8 @@
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" Copyright | 2011-2017 Massachusetts Institute of Technology\n",
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" License | http://openmc.readthedocs.io/en/latest/license.html\n",
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" Version | 0.8.0\n",
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" Git SHA1 | 54b65c8bda6af5788bd762b8cf9855d1a8008238\n",
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" Date/Time | 2017-02-12 13:36:24\n",
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" Git SHA1 | 60a1f157dae88b62e1865a5fe3efd7ef0773a068\n",
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" Date/Time | 2017-02-25 14:26:54\n",
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" OpenMP Threads | 8\n",
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"\n",
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" ===========================================================================\n",
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@ -592,7 +594,7 @@
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" 42/1 1.13779 1.16177 +/- 0.00531\n",
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" 43/1 1.15066 1.16143 +/- 0.00516\n",
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" 44/1 1.12174 1.16026 +/- 0.00514\n",
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" 45/1 1.17479 1.16068 +/- 0.00501\n",
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" 45/1 1.17478 1.16068 +/- 0.00501\n",
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" 46/1 1.14146 1.16014 +/- 0.00489\n",
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" 47/1 1.20464 1.16135 +/- 0.00491\n",
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" 48/1 1.15119 1.16108 +/- 0.00479\n",
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@ -607,20 +609,20 @@
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"\n",
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" =======================> TIMING STATISTICS <=======================\n",
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"\n",
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" Total time for initialization = 4.1327E-01 seconds\n",
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" Reading cross sections = 3.2638E-01 seconds\n",
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" Total time in simulation = 2.2324E+00 seconds\n",
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" Time in transport only = 2.1226E+00 seconds\n",
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" Time in inactive batches = 3.0650E-01 seconds\n",
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" Time in active batches = 1.9259E+00 seconds\n",
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" Time synchronizing fission bank = 2.7640E-03 seconds\n",
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" Sampling source sites = 2.0198E-03 seconds\n",
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" SEND/RECV source sites = 7.0929E-04 seconds\n",
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" Time accumulating tallies = 4.5355E-05 seconds\n",
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" Total time for finalization = 4.1885E-04 seconds\n",
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" Total time elapsed = 2.6534E+00 seconds\n",
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" Calculation Rate (inactive) = 81567.1 neutrons/second\n",
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" Calculation Rate (active) = 51923.2 neutrons/second\n",
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" Total time for initialization = 3.5070E-01 seconds\n",
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" Reading cross sections = 2.4151E-01 seconds\n",
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" Total time in simulation = 2.3276E+00 seconds\n",
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" Time in transport only = 2.2350E+00 seconds\n",
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" Time in inactive batches = 2.5677E-01 seconds\n",
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" Time in active batches = 2.0708E+00 seconds\n",
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" Time synchronizing fission bank = 2.7683E-03 seconds\n",
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" Sampling source sites = 2.0233E-03 seconds\n",
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" SEND/RECV source sites = 7.1007E-04 seconds\n",
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" Time accumulating tallies = 5.0753E-05 seconds\n",
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" Total time for finalization = 3.8695E-04 seconds\n",
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" Total time elapsed = 2.6857E+00 seconds\n",
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" Calculation Rate (inactive) = 97364.6 neutrons/second\n",
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" Calculation Rate (active) = 48290.8 neutrons/second\n",
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"\n",
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" ============================> RESULTS <============================\n",
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"\n",
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@ -901,7 +903,7 @@
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" <td>6.250000e-01</td>\n",
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" <td>total</td>\n",
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" <td>(((total / flux) - (absorption / flux)) - (sca...</td>\n",
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" <td>-2.664535e-15</td>\n",
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" <td>-5.551115e-15</td>\n",
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" <td>0.011292</td>\n",
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" </tr>\n",
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" <tr>\n",
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@ -911,7 +913,7 @@
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" <td>2.000000e+07</td>\n",
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" <td>total</td>\n",
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" <td>(((total / flux) - (absorption / flux)) - (sca...</td>\n",
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" <td>-3.330669e-16</td>\n",
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" <td>-1.110223e-16</td>\n",
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" <td>0.002570</td>\n",
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" </tr>\n",
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" </tbody>\n",
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@ -924,8 +926,8 @@
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"1 1 6.25e-01 2.00e+07 total \n",
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"\n",
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" score mean std. dev. \n",
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"0 (((total / flux) - (absorption / flux)) - (sca... -2.66e-15 1.13e-02 \n",
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"1 (((total / flux) - (absorption / flux)) - (sca... -3.33e-16 2.57e-03 "
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"0 (((total / flux) - (absorption / flux)) - (sca... -5.55e-15 1.13e-02 \n",
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"1 (((total / flux) - (absorption / flux)) - (sca... -1.11e-16 2.57e-03 "
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]
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},
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"execution_count": 22,
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File diff suppressed because one or more lines are too long
File diff suppressed because one or more lines are too long
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@ -149,6 +149,8 @@ class MGXS(object):
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Name of the multi-group cross section
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rxn_type : str
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Reaction type (e.g., 'total', 'nu-fission', etc.)
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nu : bool
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If True, the cross section data will include neutron multiplication
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by_nuclide : bool
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If true, computes cross sections for each nuclide in domain
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domain : Material or Cell or Universe or Mesh
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@ -230,6 +232,7 @@ class MGXS(object):
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self._derived = False
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self._hdf5_key = None
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self._valid_estimators = ESTIMATOR_TYPES
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self._nu = False
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self.name = name
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self.by_nuclide = by_nuclide
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@ -408,6 +411,10 @@ class MGXS(object):
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def rxn_type(self):
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return self._rxn_type
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@property
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def nu(self):
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return self._nu
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@property
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def by_nuclide(self):
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return self._by_nuclide
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@ -573,6 +580,11 @@ class MGXS(object):
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cv.check_type('name', name, string_types)
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self._name = name
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@nu.setter
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def nu(self, nu):
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cv.check_type('nu', nu, bool)
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self._nu = nu
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@by_nuclide.setter
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def by_nuclide(self, by_nuclide):
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cv.check_type('by_nuclide', by_nuclide, bool)
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@ -704,7 +716,7 @@ class MGXS(object):
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elif mgxs_type == 'transport':
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mgxs = TransportXS(domain, domain_type, energy_groups)
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elif mgxs_type == 'nu-transport':
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mgxs = NuTransportXS(domain, domain_type, energy_groups)
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mgxs = TransportXS(domain, domain_type, energy_groups, nu=True)
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elif mgxs_type == 'absorption':
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mgxs = AbsorptionXS(domain, domain_type, energy_groups)
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elif mgxs_type == 'capture':
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@ -712,17 +724,17 @@ class MGXS(object):
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elif mgxs_type == 'fission':
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mgxs = FissionXS(domain, domain_type, energy_groups)
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elif mgxs_type == 'nu-fission':
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mgxs = NuFissionXS(domain, domain_type, energy_groups)
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mgxs = FissionXS(domain, domain_type, energy_groups, nu=True)
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elif mgxs_type == 'kappa-fission':
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mgxs = KappaFissionXS(domain, domain_type, energy_groups)
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elif mgxs_type == 'scatter':
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mgxs = ScatterXS(domain, domain_type, energy_groups)
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elif mgxs_type == 'nu-scatter':
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mgxs = NuScatterXS(domain, domain_type, energy_groups)
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mgxs = ScatterXS(domain, domain_type, energy_groups, nu=True)
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elif mgxs_type == 'scatter matrix':
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mgxs = ScatterMatrixXS(domain, domain_type, energy_groups)
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elif mgxs_type == 'nu-scatter matrix':
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mgxs = NuScatterMatrixXS(domain, domain_type, energy_groups)
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mgxs = ScatterMatrixXS(domain, domain_type, energy_groups, nu=True)
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elif mgxs_type == 'multiplicity matrix':
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mgxs = MultiplicityMatrixXS(domain, domain_type, energy_groups)
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elif mgxs_type == 'nu-fission matrix':
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@ -2001,6 +2013,8 @@ class MatrixMGXS(MGXS):
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Name of the multi-group cross section
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rxn_type : str
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Reaction type (e.g., 'total', 'nu-fission', etc.)
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nu : bool
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If True, the cross section data will include neutron multiplication
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by_nuclide : bool
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If true, computes cross sections for each nuclide in domain
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domain : Material or Cell or Universe or Mesh
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@ -2555,9 +2569,8 @@ class TotalXS(MGXS):
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"""
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def __init__(self, domain=None, domain_type=None,
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groups=None, by_nuclide=False, name='', num_polar=1,
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num_azimuthal=1):
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def __init__(self, domain=None, domain_type=None, groups=None,
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by_nuclide=False, name='', num_polar=1, num_azimuthal=1):
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super(TotalXS, self).__init__(domain, domain_type,
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groups, by_nuclide, name, num_polar,
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num_azimuthal)
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@ -2607,6 +2620,9 @@ class TransportXS(MGXS):
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The domain type for spatial homogenization
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groups : openmc.mgxs.EnergyGroups
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The energy group structure for energy condensation
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nu : bool
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If True, the cross section data will include neutron multiplication;
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defaults to True.
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by_nuclide : bool
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If true, computes cross sections for each nuclide in domain
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name : str, optional
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@ -2625,6 +2641,8 @@ class TransportXS(MGXS):
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Name of the multi-group cross section
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rxn_type : str
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Reaction type (e.g., 'total', 'nu-fission', etc.)
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nu : bool
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If True, the cross section data will include neutron multiplication
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by_nuclide : bool
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If true, computes cross sections for each nuclide in domain
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domain : Material or Cell or Universe or Mesh
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@ -2685,19 +2703,32 @@ class TransportXS(MGXS):
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"""
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def __init__(self, domain=None, domain_type=None,
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groups=None, by_nuclide=False, name='', num_polar=1,
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num_azimuthal=1):
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def __init__(self, domain=None, domain_type=None, groups=None, nu=False,
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by_nuclide=False, name='', num_polar=1, num_azimuthal=1):
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super(TransportXS, self).__init__(domain, domain_type,
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groups, by_nuclide, name, num_polar,
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num_azimuthal)
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self._rxn_type = 'transport'
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if not nu:
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self._rxn_type = 'transport'
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else:
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self._rxn_type = 'nu-transport'
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self._estimator = 'analog'
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self._valid_estimators = ['analog']
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self.nu = nu
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@property
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def scores(self):
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return ['flux', 'total', 'scatter-1']
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if not self.nu:
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return ['flux', 'total', 'scatter-1']
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else:
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return ['flux', 'total', 'nu-scatter-1']
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@property
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def tally_keys(self):
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if not self.nu:
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return super(TransportXS, self).tally_keys
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else:
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return ['flux', 'total', 'scatter-1']
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@property
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def filters(self):
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@ -2724,131 +2755,6 @@ class TransportXS(MGXS):
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return self._rxn_rate_tally
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class NuTransportXS(TransportXS):
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r"""A transport-corrected total multi-group cross section which
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accounts for neutron multiplicity in scattering reactions.
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This class can be used for both OpenMC input generation and tally data
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post-processing to compute spatially-homogenized and energy-integrated
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multi-group cross sections for multi-group neutronics calculations. At a
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minimum, one needs to set the :attr:`NuTransportXS.energy_groups` and
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:attr:`NuTransportXS.domain` properties. Tallies for the flux and
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appropriate reaction rates over the specified domain are generated
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automatically via the :attr:`NuTransportXS.tallies` property, which can then
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be appended to a :class:`openmc.Tallies` instance.
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For post-processing, the :meth:`MGXS.load_from_statepoint` will pull in the
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necessary data to compute multi-group cross sections from a
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:class:`openmc.StatePoint` instance. The derived multi-group cross section
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can then be obtained from the :attr:`NuTransportXS.xs_tally` property.
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The calculation of the transport-corrected cross section is the same as that
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for :class:`TransportXS` except that the scattering multiplicity is
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accounted for.
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Parameters
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----------
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domain : openmc.Material or openmc.Cell or openmc.Universe or openmc.Mesh
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The domain for spatial homogenization
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domain_type : {'material', 'cell', 'distribcell', 'universe', 'mesh'}
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The domain type for spatial homogenization
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groups : openmc.mgxs.EnergyGroups
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The energy group structure for energy condensation
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by_nuclide : bool
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If true, computes cross sections for each nuclide in domain
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name : str, optional
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Name of the multi-group cross section. Used as a label to identify
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tallies in OpenMC 'tallies.xml' file.
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num_polar : Integral, optional
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Number of equi-width polar angle bins for angle discretization;
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defaults to one bin
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num_azimuthal : Integral, optional
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Number of equi-width azimuthal angle bins for angle discretization;
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defaults to one bin
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Attributes
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----------
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name : str, optional
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Name of the multi-group cross section
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rxn_type : str
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Reaction type (e.g., 'total', 'nu-fission', etc.)
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by_nuclide : bool
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If true, computes cross sections for each nuclide in domain
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domain : Material or Cell or Universe or Mesh
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Domain for spatial homogenization
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domain_type : {'material', 'cell', 'distribcell', 'universe', 'mesh'}
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Domain type for spatial homogenization
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energy_groups : openmc.mgxs.EnergyGroups
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Energy group structure for energy condensation
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num_polar : Integral
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Number of equi-width polar angle bins for angle discretization
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num_azimuthal : Integral
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Number of equi-width azimuthal angle bins for angle discretization
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tally_trigger : openmc.Trigger
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An (optional) tally precision trigger given to each tally used to
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compute the cross section
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scores : list of str
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The scores in each tally used to compute the multi-group cross section
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filters : list of openmc.Filter
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The filters in each tally used to compute the multi-group cross section
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tally_keys : list of str
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The keys into the tallies dictionary for each tally used to compute
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the multi-group cross section
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estimator : 'analog'
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The tally estimator used to compute the multi-group cross section
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tallies : collections.OrderedDict
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OpenMC tallies needed to compute the multi-group cross section. The keys
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are strings listed in the :attr:`NuTransportXS.tally_keys` property and
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values are instances of :class:`openmc.Tally`.
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rxn_rate_tally : openmc.Tally
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Derived tally for the reaction rate tally used in the numerator to
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compute the multi-group cross section. This attribute is None
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unless the multi-group cross section has been computed.
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xs_tally : openmc.Tally
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Derived tally for the multi-group cross section. This attribute
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is None unless the multi-group cross section has been computed.
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num_subdomains : int
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The number of subdomains is unity for 'material', 'cell' and 'universe'
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domain types. This is equal to the number of cell instances
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for 'distribcell' domain types (it is equal to unity prior to loading
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tally data from a statepoint file).
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num_nuclides : int
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The number of nuclides for which the multi-group cross section is
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being tracked. This is unity if the by_nuclide attribute is False.
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nuclides : Iterable of str or 'sum'
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The optional user-specified nuclides for which to compute cross
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sections (e.g., 'U238', 'O16'). If by_nuclide is True but nuclides
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are not specified by the user, all nuclides in the spatial domain
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are included. This attribute is 'sum' if by_nuclide is false.
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sparse : bool
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Whether or not the MGXS' tallies use SciPy's LIL sparse matrix format
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for compressed data storage
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loaded_sp : bool
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Whether or not a statepoint file has been loaded with tally data
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derived : bool
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Whether or not the MGXS is merged from one or more other MGXS
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hdf5_key : str
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The key used to index multi-group cross sections in an HDF5 data store
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"""
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def __init__(self, domain=None, domain_type=None,
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groups=None, by_nuclide=False, name='', num_polar=1,
|
||||
num_azimuthal=1):
|
||||
super(NuTransportXS, self).__init__(domain, domain_type,
|
||||
groups, by_nuclide, name,
|
||||
num_polar, num_azimuthal)
|
||||
self._rxn_type = 'nu-transport'
|
||||
|
||||
@property
|
||||
def scores(self):
|
||||
return ['flux', 'total', 'nu-scatter-1']
|
||||
|
||||
@property
|
||||
def tally_keys(self):
|
||||
return ['flux', 'total', 'scatter-1']
|
||||
|
||||
|
||||
class AbsorptionXS(MGXS):
|
||||
r"""An absorption multi-group cross section.
|
||||
|
||||
|
|
@ -2966,9 +2872,8 @@ class AbsorptionXS(MGXS):
|
|||
|
||||
"""
|
||||
|
||||
def __init__(self, domain=None, domain_type=None,
|
||||
groups=None, by_nuclide=False, name='', num_polar=1,
|
||||
num_azimuthal=1):
|
||||
def __init__(self, domain=None, domain_type=None, groups=None,
|
||||
by_nuclide=False, name='', num_polar=1, num_azimuthal=1):
|
||||
super(AbsorptionXS, self).__init__(domain, domain_type,
|
||||
groups, by_nuclide, name, num_polar,
|
||||
num_azimuthal)
|
||||
|
|
@ -3094,9 +2999,8 @@ class CaptureXS(MGXS):
|
|||
|
||||
"""
|
||||
|
||||
def __init__(self, domain=None, domain_type=None,
|
||||
groups=None, by_nuclide=False, name='', num_polar=1,
|
||||
num_azimuthal=1):
|
||||
def __init__(self, domain=None, domain_type=None, groups=None,
|
||||
by_nuclide=False, name='', num_polar=1, num_azimuthal=1):
|
||||
super(CaptureXS, self).__init__(domain, domain_type,
|
||||
groups, by_nuclide, name, num_polar,
|
||||
num_azimuthal)
|
||||
|
|
@ -3150,6 +3054,9 @@ class FissionXS(MGXS):
|
|||
The domain type for spatial homogenization
|
||||
groups : openmc.mgxs.EnergyGroups
|
||||
The energy group structure for energy condensation
|
||||
nu : bool
|
||||
If True, the cross section data will include neutron multiplication;
|
||||
defaults to False
|
||||
by_nuclide : bool
|
||||
If true, computes cross sections for each nuclide in domain
|
||||
name : str, optional
|
||||
|
|
@ -3168,6 +3075,8 @@ class FissionXS(MGXS):
|
|||
Name of the multi-group cross section
|
||||
rxn_type : str
|
||||
Reaction type (e.g., 'total', 'nu-fission', etc.)
|
||||
nu : bool
|
||||
If True, the cross section data will include neutron multiplication
|
||||
by_nuclide : bool
|
||||
If true, computes cross sections for each nuclide in domain
|
||||
domain : Material or Cell or Universe or Mesh
|
||||
|
|
@ -3228,136 +3137,15 @@ class FissionXS(MGXS):
|
|||
|
||||
"""
|
||||
|
||||
def __init__(self, domain=None, domain_type=None,
|
||||
groups=None, by_nuclide=False, name='', num_polar=1,
|
||||
num_azimuthal=1):
|
||||
def __init__(self, domain=None, domain_type=None, groups=None, nu=False,
|
||||
by_nuclide=False, name='', num_polar=1, num_azimuthal=1):
|
||||
super(FissionXS, self).__init__(domain, domain_type,
|
||||
groups, by_nuclide, name, num_polar,
|
||||
num_azimuthal)
|
||||
self._rxn_type = 'fission'
|
||||
|
||||
|
||||
class NuFissionXS(MGXS):
|
||||
r"""A fission neutron production multi-group cross section.
|
||||
|
||||
This class can be used for both OpenMC input generation and tally data
|
||||
post-processing to compute spatially-homogenized and energy-integrated
|
||||
multi-group fission neutron production cross sections for multi-group
|
||||
neutronics calculations. At a minimum, one needs to set the
|
||||
:attr:`NuFissionXS.energy_groups` and :attr:`NuFissionXS.domain`
|
||||
properties. Tallies for the flux and appropriate reaction rates over the
|
||||
specified domain are generated automatically via the
|
||||
:attr:`NuFissionXS.tallies` property, which can then be appended to a
|
||||
:class:`openmc.Tallies` instance.
|
||||
|
||||
For post-processing, the :meth:`MGXS.load_from_statepoint` will pull in the
|
||||
necessary data to compute multi-group cross sections from a
|
||||
:class:`openmc.StatePoint` instance. The derived multi-group cross section
|
||||
can then be obtained from the :attr:`NuFissionXS.xs_tally` property.
|
||||
|
||||
For a spatial domain :math:`V` and energy group :math:`[E_g,E_{g-1}]`, the
|
||||
fission neutron production cross section is calculated as:
|
||||
|
||||
.. math::
|
||||
|
||||
\frac{\int_{r \in V} dr \int_{4\pi} d\Omega \int_{E_g}^{E_{g-1}} dE \;
|
||||
\nu\sigma_f (r, E) \psi (r, E, \Omega)}{\int_{r \in V} dr \int_{4\pi}
|
||||
d\Omega \int_{E_g}^{E_{g-1}} dE \; \psi (r, E, \Omega)}.
|
||||
|
||||
|
||||
Parameters
|
||||
----------
|
||||
domain : openmc.Material or openmc.Cell or openmc.Universe or openmc.Mesh
|
||||
The domain for spatial homogenization
|
||||
domain_type : {'material', 'cell', 'distribcell', 'universe', 'mesh'}
|
||||
The domain type for spatial homogenization
|
||||
groups : openmc.mgxs.EnergyGroups
|
||||
The energy group structure for energy condensation
|
||||
by_nuclide : bool
|
||||
If true, computes cross sections for each nuclide in domain
|
||||
name : str, optional
|
||||
Name of the multi-group cross section. Used as a label to identify
|
||||
tallies in OpenMC 'tallies.xml' file.
|
||||
num_polar : Integral, optional
|
||||
Number of equi-width polar angle bins for angle discretization;
|
||||
defaults to one bin
|
||||
num_azimuthal : Integral, optional
|
||||
Number of equi-width azimuthal angle bins for angle discretization;
|
||||
defaults to one bin
|
||||
|
||||
Attributes
|
||||
----------
|
||||
name : str, optional
|
||||
Name of the multi-group cross section
|
||||
rxn_type : str
|
||||
Reaction type (e.g., 'total', 'nu-fission', etc.)
|
||||
by_nuclide : bool
|
||||
If true, computes cross sections for each nuclide in domain
|
||||
domain : Material or Cell or Universe or Mesh
|
||||
Domain for spatial homogenization
|
||||
domain_type : {'material', 'cell', 'distribcell', 'universe', 'mesh'}
|
||||
Domain type for spatial homogenization
|
||||
energy_groups : openmc.mgxs.EnergyGroups
|
||||
Energy group structure for energy condensation
|
||||
num_polar : Integral
|
||||
Number of equi-width polar angle bins for angle discretization
|
||||
num_azimuthal : Integral
|
||||
Number of equi-width azimuthal angle bins for angle discretization
|
||||
tally_trigger : openmc.Trigger
|
||||
An (optional) tally precision trigger given to each tally used to
|
||||
compute the cross section
|
||||
scores : list of str
|
||||
The scores in each tally used to compute the multi-group cross section
|
||||
filters : list of openmc.Filter
|
||||
The filters in each tally used to compute the multi-group cross section
|
||||
tally_keys : list of str
|
||||
The keys into the tallies dictionary for each tally used to compute
|
||||
the multi-group cross section
|
||||
estimator : {'tracklength', 'collision', 'analog'}
|
||||
The tally estimator used to compute the multi-group cross section
|
||||
tallies : collections.OrderedDict
|
||||
OpenMC tallies needed to compute the multi-group cross section. The keys
|
||||
are strings listed in the :attr:`NuFissionXS.tally_keys` property and
|
||||
values are instances of :class:`openmc.Tally`.
|
||||
rxn_rate_tally : openmc.Tally
|
||||
Derived tally for the reaction rate tally used in the numerator to
|
||||
compute the multi-group cross section. This attribute is None
|
||||
unless the multi-group cross section has been computed.
|
||||
xs_tally : openmc.Tally
|
||||
Derived tally for the multi-group cross section. This attribute
|
||||
is None unless the multi-group cross section has been computed.
|
||||
num_subdomains : int
|
||||
The number of subdomains is unity for 'material', 'cell' and 'universe'
|
||||
domain types. This is equal to the number of cell instances
|
||||
for 'distribcell' domain types (it is equal to unity prior to loading
|
||||
tally data from a statepoint file).
|
||||
num_nuclides : int
|
||||
The number of nuclides for which the multi-group cross section is
|
||||
being tracked. This is unity if the by_nuclide attribute is False.
|
||||
nuclides : Iterable of str or 'sum'
|
||||
The optional user-specified nuclides for which to compute cross
|
||||
sections (e.g., 'U238', 'O16'). If by_nuclide is True but nuclides
|
||||
are not specified by the user, all nuclides in the spatial domain
|
||||
are included. This attribute is 'sum' if by_nuclide is false.
|
||||
sparse : bool
|
||||
Whether or not the MGXS' tallies use SciPy's LIL sparse matrix format
|
||||
for compressed data storage
|
||||
loaded_sp : bool
|
||||
Whether or not a statepoint file has been loaded with tally data
|
||||
derived : bool
|
||||
Whether or not the MGXS is merged from one or more other MGXS
|
||||
hdf5_key : str
|
||||
The key used to index multi-group cross sections in an HDF5 data store
|
||||
|
||||
"""
|
||||
|
||||
def __init__(self, domain=None, domain_type=None,
|
||||
groups=None, by_nuclide=False, name='', num_polar=1,
|
||||
num_azimuthal=1):
|
||||
super(NuFissionXS, self).__init__(domain, domain_type,
|
||||
groups, by_nuclide, name, num_polar,
|
||||
num_azimuthal)
|
||||
self._rxn_type = 'nu-fission'
|
||||
if not nu:
|
||||
self._rxn_type = 'fission'
|
||||
else:
|
||||
self._rxn_type = 'nu-fission'
|
||||
|
||||
|
||||
class KappaFissionXS(MGXS):
|
||||
|
|
@ -3479,9 +3267,8 @@ class KappaFissionXS(MGXS):
|
|||
|
||||
"""
|
||||
|
||||
def __init__(self, domain=None, domain_type=None,
|
||||
groups=None, by_nuclide=False, name='', num_polar=1,
|
||||
num_azimuthal=1):
|
||||
def __init__(self, domain=None, domain_type=None, groups=None,
|
||||
by_nuclide=False, name='', num_polar=1, num_azimuthal=1):
|
||||
super(KappaFissionXS, self).__init__(domain, domain_type,
|
||||
groups, by_nuclide, name,
|
||||
num_polar, num_azimuthal)
|
||||
|
|
@ -3526,6 +3313,9 @@ class ScatterXS(MGXS):
|
|||
The domain type for spatial homogenization
|
||||
groups : openmc.mgxs.EnergyGroups
|
||||
The energy group structure for energy condensation
|
||||
nu : bool
|
||||
If True, the cross section data will include neutron multiplication;
|
||||
defaults to False
|
||||
by_nuclide : bool
|
||||
If true, computes cross sections for each nuclide in domain
|
||||
name : str, optional
|
||||
|
|
@ -3544,6 +3334,8 @@ class ScatterXS(MGXS):
|
|||
Name of the multi-group cross section
|
||||
rxn_type : str
|
||||
Reaction type (e.g., 'total', 'nu-fission', etc.)
|
||||
nu : bool
|
||||
If True, the cross section data will include neutron multiplication
|
||||
by_nuclide : bool
|
||||
If true, computes cross sections for each nuclide in domain
|
||||
domain : Material or Cell or Universe or Mesh
|
||||
|
|
@ -3604,142 +3396,21 @@ class ScatterXS(MGXS):
|
|||
|
||||
"""
|
||||
|
||||
def __init__(self, domain=None, domain_type=None,
|
||||
groups=None, by_nuclide=False, name='', num_polar=1,
|
||||
num_azimuthal=1):
|
||||
def __init__(self, domain=None, domain_type=None, groups=None, nu=False,
|
||||
by_nuclide=False, name='', num_polar=1, num_azimuthal=1):
|
||||
super(ScatterXS, self).__init__(domain, domain_type,
|
||||
groups, by_nuclide, name, num_polar,
|
||||
num_azimuthal)
|
||||
self._rxn_type = 'scatter'
|
||||
if not nu:
|
||||
self._rxn_type = 'scatter'
|
||||
else:
|
||||
self._rxn_type = 'nu-scatter'
|
||||
# Only analog estimators are valid so change from the defaults
|
||||
# to reflect this
|
||||
self._estimator = 'analog'
|
||||
self._valid_estimators = ['analog']
|
||||
|
||||
|
||||
class NuScatterXS(MGXS):
|
||||
r"""A scattering neutron production multi-group cross section.
|
||||
|
||||
The neutron production from scattering is defined as the average number of
|
||||
neutrons produced from all neutron-producing reactions except for fission.
|
||||
|
||||
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 multi-group neutronics calculations. At a
|
||||
minimum, one needs to set the :attr:`NuScatterXS.energy_groups` and
|
||||
:attr:`NuScatterXS.domain` properties. Tallies for the flux and appropriate
|
||||
reaction rates over the specified domain are generated automatically via the
|
||||
:attr:`NuScatterXS.tallies` property, which can then be appended to a
|
||||
:class:`openmc.Tallies` instance.
|
||||
|
||||
For post-processing, the :meth:`MGXS.load_from_statepoint` will pull in the
|
||||
necessary data to compute multi-group cross sections from a
|
||||
:class:`openmc.StatePoint` instance. The derived multi-group cross section
|
||||
can then be obtained from the :attr:`NuScatterXS.xs_tally` property.
|
||||
|
||||
For a spatial domain :math:`V` and energy group :math:`[E_g,E_{g-1}]`, the
|
||||
scattering neutron production cross section is calculated as:
|
||||
|
||||
.. math::
|
||||
|
||||
\frac{\int_{r \in V} dr \int_{4\pi} d\Omega \int_{E_g}^{E_{g-1}} dE \;
|
||||
\sum_i \upsilon_i \sigma_i (r, E) \psi (r, E, \Omega)}{\int_{r \in V} dr
|
||||
\int_{4\pi} d\Omega \int_{E_g}^{E_{g-1}} dE \; \psi (r, E, \Omega)}.
|
||||
|
||||
where :math:`\upsilon_i` is the multiplicity of the :math:`i`-th scattering
|
||||
reaction.
|
||||
|
||||
Parameters
|
||||
----------
|
||||
domain : openmc.Material or openmc.Cell or openmc.Universe or openmc.Mesh
|
||||
The domain for spatial homogenization
|
||||
domain_type : {'material', 'cell', 'distribcell', 'universe', 'mesh'}
|
||||
The domain type for spatial homogenization
|
||||
groups : openmc.mgxs.EnergyGroups
|
||||
The energy group structure for energy condensation
|
||||
by_nuclide : bool
|
||||
If true, computes cross sections for each nuclide in domain
|
||||
name : str, optional
|
||||
Name of the multi-group cross section. Used as a label to identify
|
||||
tallies in OpenMC 'tallies.xml' file.
|
||||
num_polar : Integral, optional
|
||||
Number of equi-width polar angle bins for angle discretization;
|
||||
defaults to one bin
|
||||
num_azimuthal : Integral, optional
|
||||
Number of equi-width azimuthal angle bins for angle discretization;
|
||||
defaults to one bin
|
||||
|
||||
Attributes
|
||||
----------
|
||||
name : str, optional
|
||||
Name of the multi-group cross section
|
||||
rxn_type : str
|
||||
Reaction type (e.g., 'total', 'nu-fission', etc.)
|
||||
by_nuclide : bool
|
||||
If true, computes cross sections for each nuclide in domain
|
||||
domain : Material or Cell or Universe or Mesh
|
||||
Domain for spatial homogenization
|
||||
domain_type : {'material', 'cell', 'distribcell', 'universe', 'mesh'}
|
||||
Domain type for spatial homogenization
|
||||
energy_groups : openmc.mgxs.EnergyGroups
|
||||
Energy group structure for energy condensation
|
||||
num_polar : Integral
|
||||
Number of equi-width polar angle bins for angle discretization
|
||||
num_azimuthal : Integral
|
||||
Number of equi-width azimuthal angle bins for angle discretization
|
||||
tally_trigger : openmc.Trigger
|
||||
An (optional) tally precision trigger given to each tally used to
|
||||
compute the cross section
|
||||
scores : list of str
|
||||
The scores in each tally used to compute the multi-group cross section
|
||||
filters : list of openmc.Filter
|
||||
The filters in each tally used to compute the multi-group cross section
|
||||
tally_keys : list of str
|
||||
The keys into the tallies dictionary for each tally used to compute
|
||||
the multi-group cross section
|
||||
estimator : 'analog'
|
||||
The tally estimator used to compute the multi-group cross section
|
||||
tallies : collections.OrderedDict
|
||||
OpenMC tallies needed to compute the multi-group cross section. The keys
|
||||
are strings listed in the :attr:`NuScatterXS.tally_keys` property and
|
||||
values are instances of :class:`openmc.Tally`.
|
||||
rxn_rate_tally : openmc.Tally
|
||||
Derived tally for the reaction rate tally used in the numerator to
|
||||
compute the multi-group cross section. This attribute is None
|
||||
unless the multi-group cross section has been computed.
|
||||
xs_tally : openmc.Tally
|
||||
Derived tally for the multi-group cross section. This attribute
|
||||
is None unless the multi-group cross section has been computed.
|
||||
num_subdomains : int
|
||||
The number of subdomains is unity for 'material', 'cell' and 'universe'
|
||||
domain types. This is equal to the number of cell instances
|
||||
for 'distribcell' domain types (it is equal to unity prior to loading
|
||||
tally data from a statepoint file).
|
||||
num_nuclides : int
|
||||
The number of nuclides for which the multi-group cross section is
|
||||
being tracked. This is unity if the by_nuclide attribute is False.
|
||||
nuclides : Iterable of str or 'sum'
|
||||
The optional user-specified nuclides for which to compute cross
|
||||
sections (e.g., 'U238', 'O16'). If by_nuclide is True but nuclides
|
||||
are not specified by the user, all nuclides in the spatial domain
|
||||
are included. This attribute is 'sum' if by_nuclide is false.
|
||||
sparse : bool
|
||||
Whether or not the MGXS' tallies use SciPy's LIL sparse matrix format
|
||||
for compressed data storage
|
||||
loaded_sp : bool
|
||||
Whether or not a statepoint file has been loaded with tally data
|
||||
derived : bool
|
||||
Whether or not the MGXS is merged from one or more other MGXS
|
||||
hdf5_key : str
|
||||
The key used to index multi-group cross sections in an HDF5 data store
|
||||
|
||||
"""
|
||||
|
||||
def __init__(self, domain=None, domain_type=None,
|
||||
groups=None, by_nuclide=False, name='', num_polar=1,
|
||||
num_azimuthal=1):
|
||||
super(NuScatterXS, self).__init__(domain, domain_type,
|
||||
groups, by_nuclide, name, num_polar,
|
||||
num_azimuthal)
|
||||
self._rxn_type = 'nu-scatter'
|
||||
self._estimator = 'analog'
|
||||
self._valid_estimators = ['analog']
|
||||
self.nu = nu
|
||||
|
||||
|
||||
class ScatterMatrixXS(MatrixMGXS):
|
||||
|
|
@ -3793,6 +3464,9 @@ class ScatterMatrixXS(MatrixMGXS):
|
|||
The domain type for spatial homogenization
|
||||
groups : openmc.mgxs.EnergyGroups
|
||||
The energy group structure for energy condensation
|
||||
nu : bool
|
||||
If True, the cross section data will include neutron multiplication;
|
||||
defaults to False
|
||||
by_nuclide : bool
|
||||
If true, computes cross sections for each nuclide in domain
|
||||
name : str, optional
|
||||
|
|
@ -3824,6 +3498,8 @@ class ScatterMatrixXS(MatrixMGXS):
|
|||
Name of the multi-group cross section
|
||||
rxn_type : str
|
||||
Reaction type (e.g., 'total', 'nu-fission', etc.)
|
||||
nu : bool
|
||||
If True, the cross section data will include neutron multiplication
|
||||
by_nuclide : bool
|
||||
If true, computes cross sections for each nuclide in domain
|
||||
domain : Material or Cell or Universe or Mesh
|
||||
|
|
@ -3884,20 +3560,24 @@ class ScatterMatrixXS(MatrixMGXS):
|
|||
|
||||
"""
|
||||
|
||||
def __init__(self, domain=None, domain_type=None,
|
||||
groups=None, by_nuclide=False, name='', num_polar=1,
|
||||
num_azimuthal=1):
|
||||
def __init__(self, domain=None, domain_type=None, groups=None, nu=False,
|
||||
by_nuclide=False, name='', num_polar=1, num_azimuthal=1):
|
||||
super(ScatterMatrixXS, self).__init__(domain, domain_type,
|
||||
groups, by_nuclide, name,
|
||||
num_azimuthal)
|
||||
self._rxn_type = 'scatter'
|
||||
if not nu:
|
||||
self._rxn_type = 'scatter'
|
||||
self._hdf5_key = 'scatter matrix'
|
||||
else:
|
||||
self._rxn_type = 'nu-scatter'
|
||||
self._hdf5_key = 'nu-scatter matrix'
|
||||
self._correction = 'P0'
|
||||
self._scatter_format = 'legendre'
|
||||
self._legendre_order = 0
|
||||
self._histogram_bins = 16
|
||||
self._hdf5_key = 'scatter matrix'
|
||||
self._estimator = 'analog'
|
||||
self._valid_estimators = ['analog']
|
||||
self.nu = nu
|
||||
|
||||
def __deepcopy__(self, memo):
|
||||
clone = super(ScatterMatrixXS, self).__deepcopy__(memo)
|
||||
|
|
@ -4601,128 +4281,6 @@ class ScatterMatrixXS(MatrixMGXS):
|
|||
print(string)
|
||||
|
||||
|
||||
class NuScatterMatrixXS(ScatterMatrixXS):
|
||||
"""A scattering production matrix multi-group cross section for one or
|
||||
more Legendre moments.
|
||||
|
||||
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 multi-group neutronics calculations. At a
|
||||
minimum, one needs to set the :attr:`NuScatterMatrixXS.energy_groups` and
|
||||
:attr:`NuScatterMatrixXS.domain` properties. Tallies for the flux and
|
||||
appropriate reaction rates over the specified domain are generated
|
||||
automatically via the :attr:`NuScatterMatrixXS.tallies` property, which can
|
||||
then be appended to a :class:`openmc.Tallies` instance.
|
||||
|
||||
For post-processing, the :meth:`MGXS.load_from_statepoint` will pull in the
|
||||
necessary data to compute multi-group cross sections from a
|
||||
:class:`openmc.StatePoint` instance. The derived multi-group cross section
|
||||
can then be obtained from the :attr:`NuScatterMatrixXS.xs_tally` property.
|
||||
|
||||
The calculation of the scattering-production matrix is the same as that for
|
||||
:class:`ScatterMatrixXS` except that the scattering multiplicity is
|
||||
accounted for.
|
||||
|
||||
Parameters
|
||||
----------
|
||||
domain : openmc.Material or openmc.Cell or openmc.Universe or openmc.Mesh
|
||||
The domain for spatial homogenization
|
||||
domain_type : {'material', 'cell', 'distribcell', 'universe', 'mesh'}
|
||||
The domain type for spatial homogenization
|
||||
groups : openmc.mgxs.EnergyGroups
|
||||
The energy group structure for energy condensation
|
||||
by_nuclide : bool
|
||||
If true, computes cross sections for each nuclide in domain
|
||||
name : str, optional
|
||||
Name of the multi-group cross section. Used as a label to identify
|
||||
tallies in OpenMC 'tallies.xml' file.
|
||||
num_polar : Integral, optional
|
||||
Number of equi-width polar angle bins for angle discretization;
|
||||
defaults to one bin
|
||||
num_azimuthal : Integral, optional
|
||||
Number of equi-width azimuthal angle bins for angle discretization;
|
||||
defaults to one bin
|
||||
|
||||
Attributes
|
||||
----------
|
||||
correction : 'P0' or None
|
||||
Apply the P0 correction to scattering matrices if set to 'P0'
|
||||
legendre_order : int
|
||||
The highest legendre moment in the scattering matrix (default is 0)
|
||||
name : str, optional
|
||||
Name of the multi-group cross section
|
||||
rxn_type : str
|
||||
Reaction type (e.g., 'total', 'nu-fission', etc.)
|
||||
by_nuclide : bool
|
||||
If true, computes cross sections for each nuclide in domain
|
||||
domain : Material or Cell or Universe or Mesh
|
||||
Domain for spatial homogenization
|
||||
domain_type : {'material', 'cell', 'distribcell', 'universe', 'mesh'}
|
||||
Domain type for spatial homogenization
|
||||
energy_groups : openmc.mgxs.EnergyGroups
|
||||
Energy group structure for energy condensation
|
||||
num_polar : Integral
|
||||
Number of equi-width polar angle bins for angle discretization
|
||||
num_azimuthal : Integral
|
||||
Number of equi-width azimuthal angle bins for angle discretization
|
||||
tally_trigger : openmc.Trigger
|
||||
An (optional) tally precision trigger given to each tally used to
|
||||
compute the cross section
|
||||
scores : list of str
|
||||
The scores in each tally used to compute the multi-group cross section
|
||||
filters : list of openmc.Filter
|
||||
The filters in each tally used to compute the multi-group cross section
|
||||
tally_keys : list of str
|
||||
The keys into the tallies dictionary for each tally used to compute
|
||||
the multi-group cross section
|
||||
estimator : 'analog'
|
||||
The tally estimator used to compute the multi-group cross section
|
||||
tallies : collections.OrderedDict
|
||||
OpenMC tallies needed to compute the multi-group cross section. The keys
|
||||
are strings listed in the :attr:`NuScatterMatrixXS.tally_keys` property
|
||||
and values are instances of :class:`openmc.Tally`.
|
||||
rxn_rate_tally : openmc.Tally
|
||||
Derived tally for the reaction rate tally used in the numerator to
|
||||
compute the multi-group cross section. This attribute is None
|
||||
unless the multi-group cross section has been computed.
|
||||
xs_tally : openmc.Tally
|
||||
Derived tally for the multi-group cross section. This attribute
|
||||
is None unless the multi-group cross section has been computed.
|
||||
num_subdomains : int
|
||||
The number of subdomains is unity for 'material', 'cell' and 'universe'
|
||||
domain types. This is equal to the number of cell instances
|
||||
for 'distribcell' domain types (it is equal to unity prior to loading
|
||||
tally data from a statepoint file).
|
||||
num_nuclides : int
|
||||
The number of nuclides for which the multi-group cross section is
|
||||
being tracked. This is unity if the by_nuclide attribute is False.
|
||||
nuclides : Iterable of str or 'sum'
|
||||
The optional user-specified nuclides for which to compute cross
|
||||
sections (e.g., 'U238', 'O16'). If by_nuclide is True but nuclides
|
||||
are not specified by the user, all nuclides in the spatial domain
|
||||
are included. This attribute is 'sum' if by_nuclide is false.
|
||||
sparse : bool
|
||||
Whether or not the MGXS' tallies use SciPy's LIL sparse matrix format
|
||||
for compressed data storage
|
||||
loaded_sp : bool
|
||||
Whether or not a statepoint file has been loaded with tally data
|
||||
derived : bool
|
||||
Whether or not the MGXS is merged from one or more other MGXS
|
||||
hdf5_key : str
|
||||
The key used to index multi-group cross sections in an HDF5 data store
|
||||
|
||||
"""
|
||||
|
||||
def __init__(self, domain=None, domain_type=None,
|
||||
groups=None, by_nuclide=False, name='', num_polar=1,
|
||||
num_azimuthal=1):
|
||||
super(NuScatterMatrixXS, self).__init__(domain, domain_type,
|
||||
groups, by_nuclide, name,
|
||||
num_polar, num_azimuthal)
|
||||
self._rxn_type = 'nu-scatter'
|
||||
self._hdf5_key = 'nu-scatter matrix'
|
||||
|
||||
|
||||
class MultiplicityMatrixXS(MatrixMGXS):
|
||||
r"""The scattering multiplicity matrix.
|
||||
|
||||
|
|
@ -4847,15 +4405,15 @@ class MultiplicityMatrixXS(MatrixMGXS):
|
|||
|
||||
"""
|
||||
|
||||
def __init__(self, domain=None, domain_type=None,
|
||||
groups=None, by_nuclide=False, name='', num_polar=1,
|
||||
num_azimuthal=1):
|
||||
def __init__(self, domain=None, domain_type=None, groups=None,
|
||||
by_nuclide=False, name='', num_polar=1, num_azimuthal=1):
|
||||
super(MultiplicityMatrixXS, self).__init__(domain, domain_type, groups,
|
||||
by_nuclide, name, num_polar,
|
||||
num_azimuthal)
|
||||
self._rxn_type = 'multiplicity matrix'
|
||||
self._estimator = 'analog'
|
||||
self._valid_estimators = ['analog']
|
||||
self.nu = True
|
||||
|
||||
@property
|
||||
def scores(self):
|
||||
|
|
@ -5009,9 +4567,8 @@ class NuFissionMatrixXS(MatrixMGXS):
|
|||
|
||||
"""
|
||||
|
||||
def __init__(self, domain=None, domain_type=None,
|
||||
groups=None, by_nuclide=False, name='', num_polar=1,
|
||||
num_azimuthal=1):
|
||||
def __init__(self, domain=None, domain_type=None, groups=None,
|
||||
by_nuclide=False, name='', num_polar=1, num_azimuthal=1):
|
||||
super(NuFissionMatrixXS, self).__init__(domain, domain_type,
|
||||
groups, by_nuclide, name,
|
||||
num_polar, num_azimuthal)
|
||||
|
|
@ -5019,6 +4576,7 @@ class NuFissionMatrixXS(MatrixMGXS):
|
|||
self._hdf5_key = 'nu-fission matrix'
|
||||
self._estimator = 'analog'
|
||||
self._valid_estimators = ['analog']
|
||||
self.nu = True
|
||||
|
||||
|
||||
class Chi(MGXS):
|
||||
|
|
@ -5138,14 +4696,14 @@ class Chi(MGXS):
|
|||
|
||||
"""
|
||||
|
||||
def __init__(self, domain=None, domain_type=None,
|
||||
groups=None, by_nuclide=False, name='', num_polar=1,
|
||||
num_azimuthal=1):
|
||||
def __init__(self, domain=None, domain_type=None, groups=None,
|
||||
by_nuclide=False, name='', num_polar=1, num_azimuthal=1):
|
||||
super(Chi, self).__init__(domain, domain_type, groups, by_nuclide,
|
||||
name, num_polar, num_azimuthal)
|
||||
self._rxn_type = 'chi'
|
||||
self._estimator = 'analog'
|
||||
self._valid_estimators = ['analog']
|
||||
self.nu = True
|
||||
|
||||
@property
|
||||
def _dont_squeeze(self):
|
||||
|
|
@ -5691,9 +5249,8 @@ class ChiPrompt(Chi):
|
|||
|
||||
"""
|
||||
|
||||
def __init__(self, domain=None, domain_type=None,
|
||||
groups=None, by_nuclide=False, name='', num_polar=1,
|
||||
num_azimuthal=1):
|
||||
def __init__(self, domain=None, domain_type=None, groups=None,
|
||||
by_nuclide=False, name='', num_polar=1, num_azimuthal=1):
|
||||
super(ChiPrompt, self).__init__(domain, domain_type, groups,
|
||||
by_nuclide, name, num_polar,
|
||||
num_azimuthal)
|
||||
|
|
@ -5820,9 +5377,8 @@ class InverseVelocity(MGXS):
|
|||
|
||||
"""
|
||||
|
||||
def __init__(self, domain=None, domain_type=None,
|
||||
groups=None, by_nuclide=False, name='', num_polar=1,
|
||||
num_azimuthal=1):
|
||||
def __init__(self, domain=None, domain_type=None, groups=None,
|
||||
by_nuclide=False, name='', num_polar=1, num_azimuthal=1):
|
||||
super(InverseVelocity, self).__init__(domain, domain_type,
|
||||
groups, by_nuclide, name,
|
||||
num_polar, num_azimuthal)
|
||||
|
|
@ -5850,8 +5406,8 @@ class InverseVelocity(MGXS):
|
|||
if xs_type == 'macro':
|
||||
return 'second/cm'
|
||||
else:
|
||||
raise ValueError('Unable to return the units of InverseVelocity for'
|
||||
' xs_type other than "macro"')
|
||||
raise ValueError('Unable to return the units of InverseVelocity'
|
||||
' for xs_type other than "macro"')
|
||||
|
||||
|
||||
class PromptNuFissionXS(MGXS):
|
||||
|
|
@ -5966,13 +5522,13 @@ class PromptNuFissionXS(MGXS):
|
|||
|
||||
"""
|
||||
|
||||
def __init__(self, domain=None, domain_type=None,
|
||||
groups=None, by_nuclide=False, name='', num_polar=1,
|
||||
num_azimuthal=1):
|
||||
def __init__(self, domain=None, domain_type=None, groups=None,
|
||||
by_nuclide=False, name='', num_polar=1, num_azimuthal=1):
|
||||
super(PromptNuFissionXS, self).__init__(domain, domain_type, groups,
|
||||
by_nuclide, name, num_polar,
|
||||
num_azimuthal)
|
||||
self._rxn_type = 'prompt-nu-fission'
|
||||
self.nu = True
|
||||
|
||||
|
||||
class PromptNuFissionMatrixXS(MatrixMGXS):
|
||||
|
|
@ -6092,9 +5648,8 @@ class PromptNuFissionMatrixXS(MatrixMGXS):
|
|||
|
||||
"""
|
||||
|
||||
def __init__(self, domain=None, domain_type=None,
|
||||
groups=None, by_nuclide=False, name='', num_polar=1,
|
||||
num_azimuthal=1):
|
||||
def __init__(self, domain=None, domain_type=None, groups=None,
|
||||
by_nuclide=False, name='', num_polar=1, num_azimuthal=1):
|
||||
super(PromptNuFissionMatrixXS, self).__init__(domain, domain_type,
|
||||
groups, by_nuclide, name,
|
||||
num_polar, num_azimuthal)
|
||||
|
|
@ -6102,3 +5657,4 @@ class PromptNuFissionMatrixXS(MatrixMGXS):
|
|||
self._hdf5_key = 'prompt-nu-fission matrix'
|
||||
self._estimator = 'analog'
|
||||
self._valid_estimators = ['analog']
|
||||
self.nu = True
|
||||
|
|
|
|||
Loading…
Add table
Add a link
Reference in a new issue