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Added a bit more documentation for nu, merged the Prompt* MGXS types with their total counterparts using the same method used for nu, and udpated mdgxs-part-i.
This commit is contained in:
Adam Nelson
parent
5e313cf5f1
commit
413e04c5cf
2 changed files with 201 additions and 507 deletions
File diff suppressed because one or more lines are too long
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@ -151,6 +151,8 @@ class MGXS(object):
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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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prompt : bool
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If true, computes cross sections which only includes prompt neutrons
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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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@ -233,6 +235,7 @@ class MGXS(object):
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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._prompt = False
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self.name = name
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self.by_nuclide = by_nuclide
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@ -415,6 +418,10 @@ class MGXS(object):
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def nu(self):
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return self._nu
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@property
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def prompt(self):
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return self._prompt
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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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@ -585,6 +592,11 @@ class MGXS(object):
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cv.check_type('nu', nu, bool)
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self._nu = nu
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@prompt.setter
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def prompt(self, prompt):
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cv.check_type('prompt', prompt, bool)
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self._prompt = prompt
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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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@ -742,13 +754,14 @@ class MGXS(object):
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elif mgxs_type == 'chi':
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mgxs = Chi(domain, domain_type, energy_groups)
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elif mgxs_type == 'chi-prompt':
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mgxs = ChiPrompt(domain, domain_type, energy_groups)
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mgxs = Chi(domain, domain_type, energy_groups, prompt=True)
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elif mgxs_type == 'inverse-velocity':
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mgxs = InverseVelocity(domain, domain_type, energy_groups)
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elif mgxs_type == 'prompt-nu-fission':
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mgxs = PromptNuFissionXS(domain, domain_type, energy_groups)
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mgxs = FissionXS(domain, domain_type, energy_groups, prompt=True)
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elif mgxs_type == 'prompt-nu-fission matrix':
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mgxs = PromptNuFissionMatrixXS(domain, domain_type, energy_groups)
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mgxs = NuFissionMatrixXS(domain, domain_type, energy_groups,
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prompt=True)
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mgxs.by_nuclide = by_nuclide
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mgxs.name = name
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@ -2612,6 +2625,9 @@ class TransportXS(MGXS):
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\sigma_{tr} &= \frac{\langle \sigma_t \phi \rangle - \langle \sigma_{s1}
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\phi \rangle}{\langle \phi \rangle}
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To incorporate the effect of scattering multiplication in the above
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relation, the `nu` parameter can be set to `True`.
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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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@ -3046,6 +3062,16 @@ class FissionXS(MGXS):
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\sigma_f (r, E) \psi (r, E, \Omega)}{\int_{r \in V} dr \int_{4\pi}
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d\Omega \int_{E_g}^{E_{g-1}} dE \; \psi (r, E, \Omega)}.
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To incorporate the effect of neutron multiplication in the above
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relation, the `nu` parameter can be set to `True`.
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This class can also be used to gather a prompt-nu-fission cross section
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(which only includes the contributions from prompt neutrons). This is
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accomplished by setting the :attr:`FissionXS.prompt` attribute to `True`.
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Since the prompt-nu-fission cross section requires neutron multiplication,
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the `nu` parameter will automatically be set to `True` if `prompt` is also
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`True`.
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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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@ -3057,6 +3083,10 @@ class FissionXS(MGXS):
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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 False
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prompt : bool
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If true, computes cross sections which only includes prompt neutrons;
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defaults to False which includes prompt and delayed in total. Setting
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this to True will also set nu 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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@ -3077,6 +3107,8 @@ class FissionXS(MGXS):
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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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prompt : bool
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If true, computes cross sections which only includes prompt neutrons
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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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@ -3138,15 +3170,20 @@ class FissionXS(MGXS):
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"""
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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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prompt=False, by_nuclide=False, name='', num_polar=1,
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num_azimuthal=1):
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super(FissionXS, 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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if not nu:
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self._rxn_type = 'fission'
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if not prompt:
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if not nu:
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self._rxn_type = 'fission'
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else:
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self._rxn_type = 'nu-fission'
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self.nu = nu
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else:
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self._rxn_type = 'nu-fission'
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self._rxn_type = 'prompt-nu-fission'
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self.nu = True
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class KappaFissionXS(MGXS):
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r"""A recoverable fission energy production rate multi-group cross section.
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@ -3305,6 +3342,9 @@ class ScatterXS(MGXS):
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\Omega) \right ]}{\int_{r \in V} dr \int_{4\pi} d\Omega
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\int_{E_g}^{E_{g-1}} dE \; \psi (r, E, \Omega)}.
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To incorporate the effect of scattering multiplication from (n,xn)
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reactions in the above relation, the `nu` parameter can be set to `True`.
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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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@ -3455,6 +3495,8 @@ class ScatterMatrixXS(MatrixMGXS):
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\phi \rangle - \delta_{gg'} \sum_{g''} \langle \sigma_{s,1,g''\rightarrow
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g} \phi \rangle}{\langle \phi \rangle}
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To incorporate the effect of neutron multiplication from (n,xn) reactions
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in the above relation, the `nu` parameter can be set to `True`.
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Parameters
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----------
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@ -4481,6 +4523,11 @@ class NuFissionMatrixXS(MatrixMGXS):
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\nu\sigma_{f,g'\rightarrow g} &= \frac{\langle \nu\sigma_{f,g'\rightarrow
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g} \phi \rangle}{\langle \phi \rangle}
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This class can also be used to gather a prompt-nu-fission cross section
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(which only includes the contributions from prompt neutrons). This is
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accomplished by setting the :attr:`NuFissionMatrixXS.prompt` attribute to
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`True`.
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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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@ -4489,6 +4536,9 @@ class NuFissionMatrixXS(MatrixMGXS):
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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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prompt : bool
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If true, computes cross sections which only includes prompt neutrons;
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defaults to False which includes prompt and delayed in total
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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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@ -4507,6 +4557,8 @@ class NuFissionMatrixXS(MatrixMGXS):
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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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prompt : bool
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If true, computes cross sections which only includes prompt neutrons
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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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@ -4568,12 +4620,17 @@ class NuFissionMatrixXS(MatrixMGXS):
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"""
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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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prompt=False, by_nuclide=False, name='', num_polar=1,
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num_azimuthal=1):
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super(NuFissionMatrixXS, self).__init__(domain, domain_type,
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groups, by_nuclide, name,
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num_polar, num_azimuthal)
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self._rxn_type = 'nu-fission'
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self._hdf5_key = 'nu-fission matrix'
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if not prompt:
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self._rxn_type = 'nu-fission'
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self._hdf5_key = 'nu-fission matrix'
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else:
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self._rxn_type = 'prompt-nu-fission'
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self._hdf5_key = 'prompt-nu-fission matrix'
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self._estimator = 'analog'
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self._valid_estimators = ['analog']
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self.nu = True
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@ -4610,6 +4667,10 @@ class Chi(MGXS):
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\chi_g &= \frac{\langle \nu\sigma_{f,g' \rightarrow g} \phi \rangle}
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{\langle \nu\sigma_f \phi \rangle}
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This class can also be used to gather a prompt-chi (which only includes the
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outgoing energy spectrum of prompt neutrons). This is accomplished by
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setting the :attr:`Chi.prompt` attribute to `True`.
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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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@ -4618,6 +4679,9 @@ class Chi(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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prompt : bool
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If true, computes cross sections which only includes prompt neutrons;
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defaults to False which includes prompt and delayed in total
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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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@ -4636,6 +4700,8 @@ class Chi(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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prompt : bool
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If true, computes cross sections which only includes prompt neutrons
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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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@ -4697,13 +4763,18 @@ class Chi(MGXS):
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"""
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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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prompt=False, by_nuclide=False, name='', num_polar=1,
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num_azimuthal=1):
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super(Chi, self).__init__(domain, domain_type, groups, by_nuclide,
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name, num_polar, num_azimuthal)
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self._rxn_type = 'chi'
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if not prompt:
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self._rxn_type = 'chi'
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else:
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self._rxn_type = 'chi-prompt'
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self._estimator = 'analog'
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self._valid_estimators = ['analog']
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self.nu = True
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self.prompt = prompt
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@property
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def _dont_squeeze(self):
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@ -4717,7 +4788,10 @@ class Chi(MGXS):
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@property
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def scores(self):
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return ['nu-fission', 'nu-fission']
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if not self.prompt:
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return ['nu-fission', 'nu-fission']
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else:
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return ['prompt-nu-fission', 'prompt-nu-fission']
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@property
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def filters(self):
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@ -5132,135 +5206,6 @@ class Chi(MGXS):
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return '%'
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class ChiPrompt(Chi):
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r"""The prompt fission spectrum.
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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:`ChiPrompt.energy_groups` and
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:attr:`ChiPrompt.domain` properties. Tallies for the flux and appropriate
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reaction rates over the specified domain are generated automatically via the
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:attr:`ChiPrompt.tallies` property, which can then be appended to a
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: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:`ChiPrompt.xs_tally` property.
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For a spatial domain :math:`V` and energy group :math:`[E_g,E_{g-1}]`, the
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fission spectrum is calculated as:
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.. math::
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\langle \nu^p \sigma_{f,g' \rightarrow g} \phi \rangle &= \int_{r \in V}
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dr \int_{4\pi} d\Omega' \int_0^\infty dE' \int_{E_g}^{E_{g-1}} dE \;
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\chi(E)^p \nu^p \sigma_f (r, E') \psi(r, E', \Omega')\\
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\langle \nu^p \sigma_f \phi \rangle &= \int_{r \in V} dr \int_{4\pi}
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d\Omega' \int_0^\infty dE' \int_0^\infty dE \; \chi(E) \nu^p \sigma_f (r,
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E') \psi(r, E', \Omega') \\
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\chi_g^p &= \frac{\langle \nu^p \sigma_{f,g' \rightarrow g} \phi \rangle}
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{\langle \nu^p \sigma_f \phi \rangle}
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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:`ChiPrompt.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., 'U-238', 'O-16'). 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, groups=None,
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by_nuclide=False, name='', num_polar=1, num_azimuthal=1):
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super(ChiPrompt, self).__init__(domain, domain_type, groups,
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by_nuclide, name, num_polar,
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num_azimuthal)
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self._rxn_type = 'chi-prompt'
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@property
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def scores(self):
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return ['prompt-nu-fission', 'prompt-nu-fission']
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class InverseVelocity(MGXS):
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r"""An inverse velocity multi-group cross section.
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@ -5408,253 +5353,3 @@ class InverseVelocity(MGXS):
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else:
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raise ValueError('Unable to return the units of InverseVelocity'
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' for xs_type other than "macro"')
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|
||||
class PromptNuFissionXS(MGXS):
|
||||
r"""A prompt 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 cross sections for multi-group neutronics calculations. At a
|
||||
minimum, one needs to set the :attr:`PromptNuFissionXS.energy_groups` and
|
||||
:attr:`PromptNuFissionXS.domain` properties. Tallies for the flux and
|
||||
appropriate reaction rates over the specified domain are generated
|
||||
automatically via the :attr:`PromptNuFissionXS.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:`PromptNuFissionXS.xs_tally` property.
|
||||
|
||||
For a spatial domain :math:`V` and energy group :math:`[E_g,E_{g-1}]`, the
|
||||
fission spectrum 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^p (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:`PromptNuFissionXS.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., 'U-238', 'O-16'). 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(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):
|
||||
r"""A prompt fission neutron production matrix 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 cross sections for multi-group neutronics calculations. At a
|
||||
minimum, one needs to set the :attr:`PromptNuFissionMatrixXS.energy_groups`
|
||||
and :attr:`PromptNuFissionMatrixXS.domain` properties. Tallies for the flux
|
||||
and appropriate reaction rates over the specified domain are generated
|
||||
automatically via the :attr:`PromptNuFissionMatrixXS.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:`PromptNuFissionMatrixXS.xs_tally`
|
||||
property.
|
||||
|
||||
For a spatial domain :math:`V` and energy group :math:`[E_g,E_{g-1}]`, the
|
||||
fission spectrum is calculated as:
|
||||
|
||||
.. math::
|
||||
|
||||
\langle \nu\sigma_{f,g'\rightarrow g} \phi \rangle &= \int_{r \in V} dr
|
||||
\int_{4\pi} d\Omega' \int_{E_{g'}}^{E_{g'-1}} dE' \int_{E_g}^{E_{g-1}} dE
|
||||
\; \chi(E) \nu\sigma_f^p (r, E') \psi(r, E', \Omega')\\
|
||||
\langle \phi \rangle &= \int_{r \in V} dr \int_{4\pi} d\Omega
|
||||
\int_{E_g}^{E_{g-1}} dE \; \psi (r, E, \Omega) \\
|
||||
\nu\sigma_{f,g'\rightarrow g} &= \frac{\langle \nu\sigma_{f,g'\rightarrow
|
||||
g}^p \phi \rangle}{\langle \phi \rangle}
|
||||
|
||||
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:`PromptNuFissionXS.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., 'U-238', 'O-16'). 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(PromptNuFissionMatrixXS, self).__init__(domain, domain_type,
|
||||
groups, by_nuclide, name,
|
||||
num_polar, num_azimuthal)
|
||||
self._rxn_type = 'prompt-nu-fission'
|
||||
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