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Merge pull request #2037 from paulromano/depletion-docstring-fixes
Collection of depletion-related fixes and enhancements
This commit is contained in:
Paul Romano
GitHub
commit
b5b928e2e0
6 changed files with 88 additions and 39 deletions
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@ -851,7 +851,7 @@ class Integrator(ABC):
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return (self.operator.prev_res[-1].time[-1],
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len(self.operator.prev_res) - 1)
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def integrate(self, final_step=True):
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def integrate(self, final_step=True, output=True):
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"""Perform the entire depletion process across all steps
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Parameters
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@ -861,12 +861,18 @@ class Integrator(ABC):
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of the last timestep.
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.. versionadded:: 0.12.1
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output : bool, optional
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Indicate whether to display information about progress
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.. versionadded:: 0.13.1
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"""
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with self.operator as conc:
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t, self._i_res = self._get_start_data()
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for i, (dt, source_rate) in enumerate(self):
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if output:
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print(f"[openmc.deplete] t={t} s, dt={dt} s, source={source_rate}")
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# Solve transport equation (or obtain result from restart)
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if i > 0 or self.operator.prev_res is None:
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conc, res = self._get_bos_data_from_operator(i, source_rate, conc)
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@ -892,6 +898,8 @@ class Integrator(ABC):
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# source rate is passed to the transport operator (which knows to
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# just return zero reaction rates without actually doing a transport
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# solve)
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if output and final_step:
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print(f"[openmc.deplete] t={t} (final operator evaluation)")
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res_list = [self.operator(conc, source_rate if final_step else 0.0)]
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Results.save(self.operator, [conc], res_list, [t, t],
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source_rate, self._i_res + len(self), proc_time)
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@ -1004,12 +1012,23 @@ class SIIntegrator(Integrator):
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self.operator.settings.particles //= self.n_steps
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return inherited
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def integrate(self):
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"""Perform the entire depletion process across all steps"""
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def integrate(self, output=True):
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"""Perform the entire depletion process across all steps
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Parameters
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----------
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output : bool, optional
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Indicate whether to display information about progress
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.. versionadded:: 0.13.1
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"""
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with self.operator as conc:
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t, self._i_res = self._get_start_data()
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for i, (dt, p) in enumerate(self):
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if output:
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print(f"[openmc.deplete] t={t} s, dt={dt} s, source={p}")
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if i == 0:
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if self.operator.prev_res is None:
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conc, res = self._get_bos_data_from_operator(i, p, conc)
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@ -6,6 +6,8 @@ from collections import OrderedDict
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import numpy as np
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from openmc import Material
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class AtomNumber:
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"""Stores local material compositions (atoms of each nuclide).
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@ -58,6 +60,12 @@ class AtomNumber:
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self.number = np.zeros((len(local_mats), len(nuclides)))
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def _get_mat_index(self, mat):
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"""Helper method for getting material index"""
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if isinstance(mat, Material):
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mat = str(mat.id)
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return self.index_mat[mat] if isinstance(mat, str) else mat
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def __getitem__(self, pos):
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"""Retrieves total atom number from AtomNumber.
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@ -75,8 +83,7 @@ class AtomNumber:
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"""
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mat, nuc = pos
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if isinstance(mat, str):
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mat = self.index_mat[mat]
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mat = self._get_mat_index(mat)
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if isinstance(nuc, str):
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nuc = self.index_nuc[nuc]
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@ -96,8 +103,7 @@ class AtomNumber:
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"""
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mat, nuc = pos
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if isinstance(mat, str):
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mat = self.index_mat[mat]
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mat = self._get_mat_index(mat)
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if isinstance(nuc, str):
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nuc = self.index_nuc[nuc]
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@ -121,11 +127,11 @@ class AtomNumber:
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if ind < self.n_nuc_burn]
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def get_atom_density(self, mat, nuc):
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"""Accesses atom density instead of total number.
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"""Return atom density of given material and nuclide
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Parameters
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----------
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mat : str, int or slice
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mat : str, int, openmc.Material or slice
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Material index.
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nuc : str, int or slice
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Nuclide index.
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@ -136,19 +142,43 @@ class AtomNumber:
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Density in [atom/cm^3]
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"""
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if isinstance(mat, str):
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mat = self.index_mat[mat]
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mat = self._get_mat_index(mat)
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if isinstance(nuc, str):
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nuc = self.index_nuc[nuc]
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return self[mat, nuc] / self.volume[mat]
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def get_atom_densities(self, mat, units='atom/b-cm'):
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"""Return atom densities for a given material
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Parameters
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----------
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mat : str, int, openmc.Material or slice
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Material index.
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units : {"atom/b-cm", "atom/cm3"}, optional
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Units for the returned concentration. Default is ``"atom/b-cm"``
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.. versionadded:: 0.13.1
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Returns
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-------
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dict
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Dictionary mapping nuclides to atom densities
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"""
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mat = self._get_mat_index(mat)
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normalization = (1.0e-24 if units == 'atom/b-cm' else 1.0) / self.volume[mat]
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return {
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name: normalization * self[mat, nuc]
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for name, nuc in self.index_nuc.items()
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}
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def set_atom_density(self, mat, nuc, val):
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"""Sets atom density instead of total number.
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Parameters
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----------
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mat : str, int or slice
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mat : str, int, openmc.Material or slice
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Material index.
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nuc : str, int or slice
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Nuclide index.
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@ -156,8 +186,7 @@ class AtomNumber:
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Array of densities to set in [atom/cm^3]
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"""
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if isinstance(mat, str):
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mat = self.index_mat[mat]
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mat = self._get_mat_index(mat)
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if isinstance(nuc, str):
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nuc = self.index_nuc[nuc]
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@ -168,7 +197,7 @@ class AtomNumber:
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Parameters
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----------
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mat : str, int or slice
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mat : str, int, openmc.Material or slice
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Material index.
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Returns
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@ -177,9 +206,7 @@ class AtomNumber:
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The slice requested in [atom].
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"""
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if isinstance(mat, str):
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mat = self.index_mat[mat]
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mat = self._get_mat_index(mat)
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return self[mat, :self.n_nuc_burn]
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def set_mat_slice(self, mat, val):
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@ -187,15 +214,13 @@ class AtomNumber:
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Parameters
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----------
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mat : str, int or slice
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mat : str, int, openmc.Material, or slice
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Material index.
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val : numpy.ndarray
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The slice to set in [atom]
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"""
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if isinstance(mat, str):
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mat = self.index_mat[mat]
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mat = self._get_mat_index(mat)
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self[mat, :self.n_nuc_burn] = val
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def set_density(self, total_density):
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@ -124,13 +124,13 @@ class CF4Integrator(Integrator):
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.. math::
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\begin{aligned}
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\mathbf{A}_1 &= h\mathbf{A}(\mathbf{n}_0) \\
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\hat{\mathbf{n}}_1 &= \exp \left ( \frac{\mathbf{A}_1}{2} \right ) \\
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\mathbf{A}_1 &= h\mathbf{A}(\mathbf{n}_i) \\
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\hat{\mathbf{n}}_1 &= \exp \left ( \frac{\mathbf{A}_1}{2} \right ) \mathbf{n}_i \\
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\mathbf{A}_2 &= h\mathbf{A}(\hat{\mathbf{n}}_1) \\
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\hat{\mathbf{n}}_2 &= \exp \left ( \frac{\mathbf{A}_2}{2} \right ) \\
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\hat{\mathbf{n}}_2 &= \exp \left ( \frac{\mathbf{A}_2}{2} \right ) \mathbf{n}_i \\
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\mathbf{A}_3 &= h \mathbf{A}(\hat{\mathbf{n}}_2) \\
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\hat{\mathbf{n}}_3 &= \exp \left ( -\frac{\mathbf{A}_1}{2} + \mathbf{A}_3
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\right ) \\
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\right ) \hat{\mathbf{n}}_1 \\
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\mathbf{A}_4 &= h\mathbf{A}(\hat{\mathbf{n}}_3) \\
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\mathbf{n}_{i+1} &= \exp \left ( \frac{\mathbf{A}_1}{4} + \frac{\mathbf{A}_2}{6}
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+ \frac{\mathbf{A}_3}{6} - \frac{\mathbf{A}_4}{12} \right )
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@ -207,7 +207,8 @@ class CELIIntegrator(Integrator):
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.. math::
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\begin{aligned}
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\mathbf{n}_{i+1}^p &= \exp \left ( h \mathbf{A}(\mathbf{n}_i ) \right ) \\
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\mathbf{n}_{i+1}^p &= \exp \left ( h \mathbf{A}(\mathbf{n}_i ) \right )
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\mathbf{n}_i \\
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\mathbf{n}_{i+1} &= \exp \left( \frac{h}{12} \mathbf{A}(\mathbf{n}_i) +
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\frac{5h}{12} \mathbf{A}(\mathbf{n}_{i+1}^p) \right)
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\exp \left( \frac{5h}{12} \mathbf{A}(\mathbf{n}_i) +
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@ -268,12 +269,12 @@ class EPCRK4Integrator(Integrator):
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.. math::
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\begin{aligned}
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\mathbf{A}_1 &= h\mathbf{A}(\mathbf{n}_0) \\
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\hat{\mathbf{n}}_1 &= \exp \left ( \frac{\mathbf{A}_1}{2} \right ) \\
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\mathbf{A}_1 &= h\mathbf{A}(\mathbf{n}_i) \\
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\hat{\mathbf{n}}_1 &= \exp \left ( \frac{\mathbf{A}_1}{2} \right ) \mathbf{n}_i \\
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\mathbf{A}_2 &= h\mathbf{A}(\hat{\mathbf{n}}_1) \\
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\hat{\mathbf{n}}_2 &= \exp \left ( \frac{\mathbf{A}_2}{2} \right ) \\
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\hat{\mathbf{n}}_2 &= \exp \left ( \frac{\mathbf{A}_2}{2} \right ) \mathbf{n}_i \\
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\mathbf{A}_3 &= h \mathbf{A}(\hat{\mathbf{n}}_2) \\
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\hat{\mathbf{n}}_3 &= \exp \left ( \mathbf{A}_3 \right ) \\
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\hat{\mathbf{n}}_3 &= \exp \left ( \mathbf{A}_3 \right ) \mathbf{n}_i \\
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\mathbf{A}_4 &= h\mathbf{A}(\hat{\mathbf{n}}_3) \\
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\mathbf{n}_{i+1} &= \exp \left ( \frac{\mathbf{A}_1}{6} + \frac{\mathbf{A}_2}{3}
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+ \frac{\mathbf{A}_3}{3} + \frac{\mathbf{A}_4}{6} \right ) \mathbf{n}_i.
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@ -123,6 +123,10 @@ class Nuclide:
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# Neutron fission yields, if present
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self._yield_data = None
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def __repr__(self):
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n_modes, n_rx = self.n_decay_modes, self.n_reaction_paths
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return f"<Nuclide: {self.name} ({n_modes} modes, {n_rx} reactions)>"
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@property
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def n_decay_modes(self):
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return len(self.decay_modes)
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@ -84,11 +84,11 @@ class Operator(TransportOperator):
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.. versionchanged:: 0.13.0
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The geometry and settings parameters have been replaced with a
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model parameter that takes an openmc.Model object
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model parameter that takes a :class:`~openmc.model.Model` object
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Parameters
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----------
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model : openmc.Model
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model : openmc.model.Model
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OpenMC model object
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chain_file : str, optional
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Path to the depletion chain XML file. Defaults to the file
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@ -163,7 +163,7 @@ class Operator(TransportOperator):
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Attributes
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----------
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model : openmc.Model
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model : openmc.model.Model
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OpenMC model object
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geometry : openmc.Geometry
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OpenMC geometry object
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@ -195,8 +195,6 @@ class Operator(TransportOperator):
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prev_res : ResultsList or None
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Results from a previous depletion calculation. ``None`` if no
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results are to be used.
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diff_burnable_mats : bool
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Whether to differentiate burnable materials with multiple instances
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cleanup_when_done : bool
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Whether to finalize and clear the shared library memory when the
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depletion operation is complete. Defaults to clearing the library.
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@ -248,7 +246,6 @@ class Operator(TransportOperator):
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)
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self.materials = model.materials
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self.diff_burnable_mats = diff_burnable_mats
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self.cleanup_when_done = True
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# Reduce the chain before we create more materials
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@ -262,8 +259,11 @@ class Operator(TransportOperator):
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self.chain = self.chain.reduce(all_isotopes, reduce_chain_level)
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# Differentiate burnable materials with multiple instances
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if self.diff_burnable_mats:
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if diff_burnable_mats:
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self._differentiate_burnable_mats()
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self.materials = openmc.Materials(
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model.geometry.get_all_materials().values()
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)
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# Clear out OpenMC, create task lists, distribute
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openmc.reset_auto_ids()
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@ -545,7 +545,7 @@ void print_results()
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fmt::print(" Leakage Fraction = {:.5f}\n",
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gt(GlobalTally::LEAKAGE, TallyResult::SUM) / n);
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}
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fmt::print("\n");
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std::cout << std::endl;
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}
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//==============================================================================
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