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https://github.com/openmc-dev/openmc.git
synced 2026-07-21 14:35:27 -04:00
Added an openmc.run mode (summary, which just shows the timing and summary results; fixed the generation of cells from a mesh (incorrect ordering of universes in the lattice), xs_shapes for delayed groups had wrong ordering (fixed) in openmc.MGXSLibrary and openmc.Plotter, fixed error in multi-group plots which only plotted one group not all, fixed issue with setting max order to 0 for Legendre scattering (not legendre converted to tabular w/in OpenMC, the default), removed superfluous xs assignments in MG mode which only slowed things down
This commit is contained in:
Adam Nelson
parent
60a1f157da
commit
fc3df18eab
8 changed files with 50 additions and 52 deletions
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@ -5,21 +5,33 @@ import sys
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from six import string_types
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summary_indicator = "TIMING STATISTICS"
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def _run(command, output, cwd):
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# Launch a subprocess
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p = subprocess.Popen(command, shell=True, cwd=cwd, stdout=subprocess.PIPE,
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stderr=subprocess.STDOUT, universal_newlines=True)
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storage_flag = False
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# Capture and re-print OpenMC output in real-time
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while True:
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# If OpenMC is finished, break loop
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line = p.stdout.readline()
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if not line and p.poll() != None:
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if not line and p.poll() is not None:
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break
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# If user requested output, print to screen
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if output:
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if output == 'full':
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# If user requested output, print to screen
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print(line, end='')
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elif output == 'summary' and summary_indicator in line:
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# If they requested a summary, look for the start of the summary
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storage_flag = True
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if storage_flag:
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# If a summary is requested, and we have reached the summary,
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# then print it
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print(line, end='')
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# Return the returncode (integer, zero if no problems encountered)
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@ -44,7 +56,7 @@ def plot_geometry(output=True, openmc_exec='openmc', cwd='.'):
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def run(particles=None, threads=None, geometry_debug=False,
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restart_file=None, tracks=False, output=True, cwd='.',
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restart_file=None, tracks=False, output="full", cwd='.',
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openmc_exec='openmc', mpi_args=None):
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"""Run an OpenMC simulation.
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@ -63,8 +75,10 @@ def run(particles=None, threads=None, geometry_debug=False,
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Path to restart file to use
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tracks : bool, optional
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Write tracks for all particles. Defaults to False.
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output : bool, optional
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Capture OpenMC output from standard out. Defaults to True.
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output : {"full", "summary", "none", False}, optional
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Degree of OpenMC output captured from standard out. "full" prints all
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output; "summary" prints only the results summary, and "none" or False
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does not show the output. Defaults to "full".
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cwd : str, optional
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Path to working directory to run in. Defaults to the current working
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directory.
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@ -189,21 +189,21 @@ class Mesh(EqualityMixin):
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def cell_generator(self):
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"""Generator function to traverse through every [i,j,k] index
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of the mesh.
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of the mesh in the same order that would be done by the Fortran
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For example the following code:
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.. code-block:: python
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for mesh_index in mymesh.cell_generator():
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print mesh_index
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print(mesh_index)
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will produce the following output for a 3-D 2x2x2 mesh in mymesh::
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[1, 1, 1]
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[2, 1, 1]
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[1, 1, 2]
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[1, 2, 1]
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[1, 2, 2]
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...
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@ -213,13 +213,13 @@ class Mesh(EqualityMixin):
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for x in range(self.dimension[0]):
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yield [x + 1, 1, 1]
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elif len(self.dimension) == 2:
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for x in range(self.dimension[0]):
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for y in range(self.dimension[1]):
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for y in range(self.dimension[1]):
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for x in range(self.dimension[0]):
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yield [x + 1, y + 1, 1]
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else:
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for x in range(self.dimension[0]):
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for z in range(self.dimension[2]):
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for y in range(self.dimension[1]):
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for z in range(self.dimension[2]):
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for x in range(self.dimension[0]):
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yield [x + 1, y + 1, z + 1]
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def to_xml_element(self):
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@ -321,25 +321,17 @@ class Mesh(EqualityMixin):
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# Build the universes which will be used for each of the [i,j,k]
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# locations within the mesh.
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# We will also have to build cells to assign to these universes
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universes = np.ndarray(self.dimension[::-1], dtype=np.object)
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# We will concurrently build cells to assign to these universes
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cells = []
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universes = []
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for [i, j, k] in self.cell_generator():
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if len(self.dimension) == 1:
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universes[i - 1] = openmc.Universe()
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cells.append(openmc.Cell())
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universes[i - 1].add_cells([cells[-1]])
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elif len(self.dimension) == 2:
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universes[j - 1, i - 1] = openmc.Universe()
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cells.append(openmc.Cell())
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universes[j - 1, i - 1].add_cells([cells[-1]])
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else:
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universes[k - 1, j - 1, i - 1] = openmc.Universe()
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cells.append(openmc.Cell())
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universes[k - 1, j - 1, i - 1].add_cells([cells[-1]])
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cells.append(openmc.Cell())
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universes.append(openmc.Universe())
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universes[-1].add_cell(cells[-1])
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lattice = openmc.RectLattice()
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lattice.lower_left = self.lower_left
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lattice.universes = np.reshape(universes, self.dimension)
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if self.width is not None:
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lattice.pitch = self.width
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@ -359,7 +351,6 @@ class Mesh(EqualityMixin):
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dz = ((self.upper_right[2] - self.lower_left[2]) /
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self.dimension[2])
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lattice.pitch = [dx, dy, dz]
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lattice.universes = universes
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# Fill Cell with the Lattice
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root_cell.fill = lattice
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@ -1317,6 +1317,7 @@ class Library(object):
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root_cell, cells = \
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self.domains[0].build_cells(bc)
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root.add_cell(root_cell)
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geometry = openmc.Geometry()
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geometry.root_universe = root
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materials = openmc.Materials()
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@ -351,8 +351,8 @@ class XSdata(object):
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self._xs_shapes["[DG]"] = (self.num_delayed_groups,)
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self._xs_shapes["[DG][G]"] = (self.num_delayed_groups,
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self.energy_groups.num_groups)
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self._xs_shapes["[DG'][G']"] = (self.num_delayed_groups,
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self.energy_groups.num_groups)
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self._xs_shapes["[DG][G']"] = (self.num_delayed_groups,
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self.energy_groups.num_groups)
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self._xs_shapes["[DG][G][G']"] = (self.num_delayed_groups,
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self.energy_groups.num_groups,
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self.energy_groups.num_groups)
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@ -676,7 +676,7 @@ def calculate_mgxs(this, types, orders=None, temperature=294.,
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for line in range(len(types)):
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for g in range(library.energy_groups.num_groups):
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data[g * 2: g * 2 + 2] = mgxs[line, g]
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data[line, g * 2: g * 2 + 2] = mgxs[line, g]
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return energy_grid[::-1], data
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@ -775,28 +775,28 @@ def _calculate_mgxs_nuc_macro(this, types, library, orders=None,
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data[i, :] = temp_data[orders[i]]
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else:
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data[i, :] = np.sum(temp_data[:])
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elif shape in (xsdata.xs_shapes["[G'][DG]"],
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xsdata.xs_shapes["[G][DG]"]):
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elif shape in (xsdata.xs_shapes["[DG][G']"],
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xsdata.xs_shapes["[DG][G]"]):
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# Then we have an array vs groups with values for each
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# delayed group. The user-provided value of orders tells us
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# which delayed group we want. If none are provided, then
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# we sum all the delayed groups together.
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if orders[i]:
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if orders[i] < len(shape[1]):
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data[i, :] = temp_data[:, orders[i]]
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if orders[i] < len(shape[0]):
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data[i, :] = temp_data[orders[i], :]
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else:
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data[i, :] = np.sum(temp_data[:, :], axis=1)
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elif shape == xsdata.xs_shapes["[G][G'][DG]"]:
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data[i, :] = np.sum(temp_data[:, :], axis=0)
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elif shape == xsdata.xs_shapes["[DG][G][G']"]:
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# Then we have a delayed group matrix. We will first
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# remove the outgoing group dependency
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temp_data = np.sum(temp_data, axis=1)
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temp_data = np.sum(temp_data, axis=-1)
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# And then proceed in exactly the same manner as the
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# "[G'][DG]" of "[G][DG]" shapes in the previous block.
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# "[DG][G']" or "[DG][G]" shapes in the previous block.
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if orders[i]:
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if orders[i] < len(shape[1]):
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data[i, :] = temp_data[:, orders[i]]
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if orders[i] < len(shape[0]):
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data[i, :] = temp_data[orders[i], :]
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else:
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data[i, :] = np.sum(temp_data[:, :], axis=1)
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data[i, :] = np.sum(temp_data[:, :], axis=0)
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elif shape == xsdata.xs_shapes["[G][G'][Order]"]:
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# This is a scattering matrix with angular data
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# First remove the outgoing group dependence
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@ -2340,7 +2340,7 @@ module mgxs_header
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! Now need to compare this material maximum scattering order with
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! the problem wide max scatt order and use whichever is lower
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order = min(mat_max_order, max_order)
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order = min(mat_max_order, max_order + 1)
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! Ok, got our order, store the dimensionality
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order_dim = order
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@ -3467,9 +3467,7 @@ module mgxs_header
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type(MaterialMacroXS), intent(inout) :: xs ! Resultant Mgxs Data
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xs % total = this % xs(this % index_temp) % total(gin)
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xs % elastic = this % xs(this % index_temp) % scatter % scattxs(gin)
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xs % absorption = this % xs(this % index_temp) % absorption(gin)
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xs % fission = this % xs(this % index_temp) % fission(gin)
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xs % nu_fission = &
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this % xs(this % index_temp) % prompt_nu_fission(gin) + &
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sum(this % xs(this % index_temp) % delayed_nu_fission(:, gin))
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@ -3487,12 +3485,8 @@ module mgxs_header
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call find_angle(this % polar, this % azimuthal, uvw, iazi, ipol)
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xs % total = this % xs(this % index_temp) % &
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total(gin, iazi, ipol)
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xs % elastic = this % xs(this % index_temp) % &
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scatter(iazi, ipol) % obj % scattxs(gin)
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xs % absorption = this % xs(this % index_temp) % &
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absorption(gin, iazi, ipol)
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xs % fission = this % xs(this % index_temp) % &
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fission(gin, iazi, ipol)
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xs % nu_fission = this % xs(this % index_temp) % &
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prompt_nu_fission(gin, iazi, ipol) + &
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sum(this % xs(this % index_temp) % &
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@ -2027,7 +2027,7 @@ contains
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end do SCORE_LOOP
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nullify(matxs,nucxs)
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nullify(matxs, nucxs)
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end subroutine score_general_mg
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!===============================================================================
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@ -105,9 +105,7 @@ contains
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p % coord(p % n_coord) % uvw, material_xs)
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else
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material_xs % total = ZERO
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material_xs % elastic = ZERO
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material_xs % absorption = ZERO
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material_xs % fission = ZERO
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material_xs % nu_fission = ZERO
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end if
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end if
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