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Paul Romano 2017-02-27 07:26:25 -06:00
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227
docs/source/pythonapi/examples/mdgxs-part-i.ipynb
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56
docs/source/pythonapi/examples/mgxs-part-i.ipynb
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@ -369,21 +369,21 @@
"\n",
"* `TotalXS`\n",
"* `TransportXS`\n",
"* `NuTransportXS`\n",
"* `AbsorptionXS`\n",
"* `CaptureXS`\n",
"* `FissionXS`\n",
"* `NuFissionXS`\n",
"* `KappaFissionXS`\n",
"* `ScatterXS`\n",
"* `NuScatterXS`\n",
"* `ScatterMatrixXS`\n",
"* `NuScatterMatrixXS`\n",
"* `Chi`\n",
"* `ChiPrompt`\n",
"* `InverseVelocity`\n",
"* `PromptNuFissionXS`\n",
"\n",
"Of course, we are aware that the fission cross section (`FissionXS`) can sometimes be paired with the fission neutron multiplication to become $\\nu\\sigma_f$. This can be accomodated in to the `FissionXS` class by setting the `nu` parameter to `True` as shown below.\n",
"\n",
"Additionally, scattering reactions (like (n,2n)) can also be defined to take in to account the neutron multiplication to become $\\nu\\sigma_s$. This can be accomodated in the the transport (`TransportXS`), scattering (`ScatterXS`), and scattering-matrix (`ScatterMatrixXS`) cross sections types by setting the `nu` parameter to `True` as shown below.\n",
"\n",
"These classes provide us with an interface to generate the tally inputs as well as perform post-processing of OpenMC's tally data to compute the respective multi-group cross sections. In this case, let's create the multi-group total, absorption and scattering cross sections with our 2-group structure."
]
},
@ -398,7 +398,11 @@
"# Instantiate a few different sections\n",
"total = mgxs.TotalXS(domain=cell, groups=groups)\n",
"absorption = mgxs.AbsorptionXS(domain=cell, groups=groups)\n",
"scattering = mgxs.ScatterXS(domain=cell, groups=groups)"
"scattering = mgxs.ScatterXS(domain=cell, groups=groups)\n",
"\n",
"# Note that if we wanted to incorporate neutron multiplication in the\n",
"# scattering cross section we would write the previous line as:\n",
"# scattering = mgxs.ScatterXS(domain=cell, groups=groups, nu=True)"
]
},
{
@ -520,8 +524,8 @@
" Copyright | 2011-2017 Massachusetts Institute of Technology\n",
" License | http://openmc.readthedocs.io/en/latest/license.html\n",
" Version | 0.8.0\n",
" Git SHA1 | 54b65c8bda6af5788bd762b8cf9855d1a8008238\n",
" Date/Time | 2017-02-12 13:36:24\n",
" Git SHA1 | 60a1f157dae88b62e1865a5fe3efd7ef0773a068\n",
" Date/Time | 2017-02-25 14:26:54\n",
" OpenMP Threads | 8\n",
"\n",
" ===========================================================================\n",
@ -592,7 +596,7 @@
" 42/1 1.13779 1.16177 +/- 0.00531\n",
" 43/1 1.15066 1.16143 +/- 0.00516\n",
" 44/1 1.12174 1.16026 +/- 0.00514\n",
" 45/1 1.17479 1.16068 +/- 0.00501\n",
" 45/1 1.17478 1.16068 +/- 0.00501\n",
" 46/1 1.14146 1.16014 +/- 0.00489\n",
" 47/1 1.20464 1.16135 +/- 0.00491\n",
" 48/1 1.15119 1.16108 +/- 0.00479\n",
@ -607,20 +611,20 @@
"\n",
" =======================> TIMING STATISTICS <=======================\n",
"\n",
" Total time for initialization = 4.1327E-01 seconds\n",
" Reading cross sections = 3.2638E-01 seconds\n",
" Total time in simulation = 2.2324E+00 seconds\n",
" Time in transport only = 2.1226E+00 seconds\n",
" Time in inactive batches = 3.0650E-01 seconds\n",
" Time in active batches = 1.9259E+00 seconds\n",
" Time synchronizing fission bank = 2.7640E-03 seconds\n",
" Sampling source sites = 2.0198E-03 seconds\n",
" SEND/RECV source sites = 7.0929E-04 seconds\n",
" Time accumulating tallies = 4.5355E-05 seconds\n",
" Total time for finalization = 4.1885E-04 seconds\n",
" Total time elapsed = 2.6534E+00 seconds\n",
" Calculation Rate (inactive) = 81567.1 neutrons/second\n",
" Calculation Rate (active) = 51923.2 neutrons/second\n",
" Total time for initialization = 3.5070E-01 seconds\n",
" Reading cross sections = 2.4151E-01 seconds\n",
" Total time in simulation = 2.3276E+00 seconds\n",
" Time in transport only = 2.2350E+00 seconds\n",
" Time in inactive batches = 2.5677E-01 seconds\n",
" Time in active batches = 2.0708E+00 seconds\n",
" Time synchronizing fission bank = 2.7683E-03 seconds\n",
" Sampling source sites = 2.0233E-03 seconds\n",
" SEND/RECV source sites = 7.1007E-04 seconds\n",
" Time accumulating tallies = 5.0753E-05 seconds\n",
" Total time for finalization = 3.8695E-04 seconds\n",
" Total time elapsed = 2.6857E+00 seconds\n",
" Calculation Rate (inactive) = 97364.6 neutrons/second\n",
" Calculation Rate (active) = 48290.8 neutrons/second\n",
"\n",
" ============================> RESULTS <============================\n",
"\n",
@ -901,7 +905,7 @@
" <td>6.250000e-01</td>\n",
" <td>total</td>\n",
" <td>(((total / flux) - (absorption / flux)) - (sca...</td>\n",
" <td>-2.664535e-15</td>\n",
" <td>-5.551115e-15</td>\n",
" <td>0.011292</td>\n",
" </tr>\n",
" <tr>\n",
@ -911,7 +915,7 @@
" <td>2.000000e+07</td>\n",
" <td>total</td>\n",
" <td>(((total / flux) - (absorption / flux)) - (sca...</td>\n",
" <td>-3.330669e-16</td>\n",
" <td>-1.110223e-16</td>\n",
" <td>0.002570</td>\n",
" </tr>\n",
" </tbody>\n",
@ -924,8 +928,8 @@
"1 1 6.25e-01 2.00e+07 total \n",
"\n",
" score mean std. dev. \n",
"0 (((total / flux) - (absorption / flux)) - (sca... -2.66e-15 1.13e-02 \n",
"1 (((total / flux) - (absorption / flux)) - (sca... -3.33e-16 2.57e-03 "
"0 (((total / flux) - (absorption / flux)) - (sca... -5.55e-15 1.13e-02 \n",
"1 (((total / flux) - (absorption / flux)) - (sca... -1.11e-16 2.57e-03 "
]
},
"execution_count": 22,

84
docs/source/pythonapi/examples/mgxs-part-ii.ipynb
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90
docs/source/pythonapi/examples/mgxs-part-iii.ipynb
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351
docs/source/pythonapi/examples/mgxs-part-iv.ipynb
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26
openmc/executor.py
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@ -5,21 +5,33 @@ import sys
from six import string_types
_summary_indicator = "TIMING STATISTICS"
def _run(command, output, cwd):
# Launch a subprocess
p = subprocess.Popen(command, shell=True, cwd=cwd, stdout=subprocess.PIPE,
stderr=subprocess.STDOUT, universal_newlines=True)
storage_flag = False
# Capture and re-print OpenMC output in real-time
while True:
# If OpenMC is finished, break loop
line = p.stdout.readline()
if not line and p.poll() != None:
if not line and p.poll() is not None:
break
# If user requested output, print to screen
if output:
if output == 'full':
# If user requested output, print to screen
print(line, end='')
elif output == 'summary' and _summary_indicator in line:
# If they requested a summary, look for the start of the summary
storage_flag = True
if storage_flag:
# If a summary is requested, and we have reached the summary,
# then print it
print(line, end='')
# Return the returncode (integer, zero if no problems encountered)
@ -60,7 +72,7 @@ def run_volume_calculation(output=True, openmc_exec='openmc', cwd='.'):
def run(particles=None, threads=None, geometry_debug=False,
restart_file=None, tracks=False, output=True, cwd='.',
restart_file=None, tracks=False, output='full', cwd='.',
openmc_exec='openmc', mpi_args=None):
"""Run an OpenMC simulation.
@ -79,8 +91,10 @@ def run(particles=None, threads=None, geometry_debug=False,
Path to restart file to use
tracks : bool, optional
Write tracks for all particles. Defaults to False.
output : bool, optional
Capture OpenMC output from standard out. Defaults to True.
output : {"full", "summary", "none", False}, optional
Degree of OpenMC output captured from standard out. "full" prints all
output; "summary" prints only the results summary, and "none" or False
does not show the output. Defaults to "full".
cwd : str, optional
Path to working directory to run in. Defaults to the current working
directory.

39
openmc/mesh.py
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@ -188,22 +188,22 @@ class Mesh(EqualityMixin):
return mesh
def cell_generator(self):
"""Generator function to traverse through every [i,j,k] index
of the mesh.
"""Generator function to traverse through every [i,j,k] index of the
mesh
For example the following code:
.. code-block:: python
for mesh_index in mymesh.cell_generator():
print mesh_index
print(mesh_index)
will produce the following output for a 3-D 2x2x2 mesh in mymesh::
[1, 1, 1]
[1, 1, 2]
[2, 1, 1]
[1, 2, 1]
[1, 2, 2]
[2, 2, 1]
...
@ -213,13 +213,13 @@ class Mesh(EqualityMixin):
for x in range(self.dimension[0]):
yield [x + 1, 1, 1]
elif len(self.dimension) == 2:
for x in range(self.dimension[0]):
for y in range(self.dimension[1]):
for y in range(self.dimension[1]):
for x in range(self.dimension[0]):
yield [x + 1, y + 1, 1]
else:
for x in range(self.dimension[0]):
for z in range(self.dimension[2]):
for y in range(self.dimension[1]):
for z in range(self.dimension[2]):
for x in range(self.dimension[0]):
yield [x + 1, y + 1, z + 1]
def to_xml_element(self):
@ -321,25 +321,17 @@ class Mesh(EqualityMixin):
# Build the universes which will be used for each of the [i,j,k]
# locations within the mesh.
# We will also have to build cells to assign to these universes
universes = np.ndarray(self.dimension[::-1], dtype=np.object)
# We will concurrently build cells to assign to these universes
cells = []
universes = []
for [i, j, k] in self.cell_generator():
if len(self.dimension) == 1:
universes[i - 1] = openmc.Universe()
cells.append(openmc.Cell())
universes[i - 1].add_cells([cells[-1]])
elif len(self.dimension) == 2:
universes[j - 1, i - 1] = openmc.Universe()
cells.append(openmc.Cell())
universes[j - 1, i - 1].add_cells([cells[-1]])
else:
universes[k - 1, j - 1, i - 1] = openmc.Universe()
cells.append(openmc.Cell())
universes[k - 1, j - 1, i - 1].add_cells([cells[-1]])
cells.append(openmc.Cell())
universes.append(openmc.Universe())
universes[-1].add_cell(cells[-1])
lattice = openmc.RectLattice()
lattice.lower_left = self.lower_left
lattice.universes = np.reshape(universes, self.dimension)
if self.width is not None:
lattice.pitch = self.width
@ -359,7 +351,6 @@ class Mesh(EqualityMixin):
dz = ((self.upper_right[2] - self.lower_left[2]) /
self.dimension[2])
lattice.pitch = [dx, dy, dz]
lattice.universes = universes
# Fill Cell with the Lattice
root_cell.fill = lattice

1
openmc/mgxs/library.py
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@ -1317,6 +1317,7 @@ class Library(object):
root_cell, cells = \
self.domains[0].build_cells(bc)
root.add_cell(root_cell)
geometry = openmc.Geometry()
geometry.root_universe = root
materials = openmc.Materials()

1154
openmc/mgxs/mgxs.py
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35
openmc/mgxs_library.py
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@ -351,8 +351,8 @@ class XSdata(object):
self._xs_shapes["[DG]"] = (self.num_delayed_groups,)
self._xs_shapes["[DG][G]"] = (self.num_delayed_groups,
self.energy_groups.num_groups)
self._xs_shapes["[DG'][G']"] = (self.num_delayed_groups,
self.energy_groups.num_groups)
self._xs_shapes["[DG][G']"] = (self.num_delayed_groups,
self.energy_groups.num_groups)
self._xs_shapes["[DG][G][G']"] = (self.num_delayed_groups,
self.energy_groups.num_groups,
self.energy_groups.num_groups)
@ -1074,12 +1074,12 @@ class XSdata(object):
def set_nu_fission_mgxs(self, nu_fission, temperature=294.,
nuclide='total', xs_type='macro', subdomain=None):
"""This method allows for an openmc.mgxs.NuFissionXS
"""This method allows for an openmc.mgxs.FissionXS
to be used to set the nu-fission cross section for this XSdata object.
Parameters
----------
nu_fission: openmc.mgxs.NuFissionXS
nu_fission: openmc.mgxs.FissionXS
MGXS Object containing the nu-fission cross section
for the domain of interest.
temperature : float
@ -1102,8 +1102,10 @@ class XSdata(object):
"""
check_type('nu_fission', nu_fission, (openmc.mgxs.NuFissionXS,
check_type('nu_fission', nu_fission, (openmc.mgxs.FissionXS,
openmc.mgxs.NuFissionMatrixXS))
if isinstance(nu_fission, openmc.mgxs.FissionXS):
check_value('nu', nu_fission.nu, [True])
check_value('energy_groups', nu_fission.energy_groups,
[self.energy_groups])
check_value('domain_type', nu_fission.domain_type,
@ -1124,13 +1126,13 @@ class XSdata(object):
subdomain=None):
"""Sets the prompt-nu-fission cross section.
This method allows for an openmc.mgxs.PromptNuFissionXS or
openmc.mgxs.PromptNuFissionMatrixXS to be used to set the
prompt-nu-fission cross section for this XSdata object.
This method allows for an openmc.mgxs.FissionXS or
openmc.mgxs.NuFissionMatrixXS to be used to set the prompt-nu-fission
cross section for this XSdata object.
Parameters
----------
prompt_nu_fission: openmc.mgxs.PromptNuFissionXS or openmc.mgxs.PromptNuFissionMatrixXS
prompt_nu_fission: openmc.mgxs.FissionXS or openmc.mgxs.NuFissionMatrixXS
MGXS Object containing the prompt-nu-fission cross section
for the domain of interest.
temperature : float
@ -1154,8 +1156,8 @@ class XSdata(object):
"""
check_type('prompt_nu_fission', prompt_nu_fission,
(openmc.mgxs.PromptNuFissionXS,
openmc.mgxs.PromptNuFissionMatrixXS))
(openmc.mgxs.FissionXS, openmc.mgxs.NuFissionMatrixXS))
check_value('prompt', prompt_nu_fission.prompt, [True])
check_value('energy_groups', prompt_nu_fission.energy_groups,
[self.energy_groups])
check_value('domain_type', prompt_nu_fission.domain_type,
@ -1307,12 +1309,12 @@ class XSdata(object):
def set_chi_prompt_mgxs(self, chi_prompt, temperature=294.,
nuclide='total', xs_type='macro', subdomain=None):
"""This method allows for an openmc.mgxs.ChiPrompt
to be used to set chi-prompt for this XSdata object.
"""This method allows for an openmc.mgxs.Chi to be used to set
chi-prompt for this XSdata object.
Parameters
----------
chi_prompt: openmc.mgxs.ChiPrompt
chi_prompt: openmc.mgxs.Chi
MGXS Object containing chi-prompt for the domain of interest.
temperature : float
Temperature (in units of Kelvin) of the provided dataset. Defaults
@ -1334,7 +1336,8 @@ class XSdata(object):
"""
check_type('chi_prompt', chi_prompt, openmc.mgxs.ChiPrompt)
check_type('chi_prompt', chi_prompt, openmc.mgxs.Chi)
check_value('prompt', chi_prompt.prompt, [True])
check_value('energy_groups', chi_prompt.energy_groups,
[self.energy_groups])
check_value('domain_type', chi_prompt.domain_type,
@ -1598,7 +1601,7 @@ class XSdata(object):
"""
check_type('nuscatter', nuscatter, (openmc.mgxs.NuScatterMatrixXS,
check_type('nuscatter', nuscatter, (openmc.mgxs.ScatterMatrixXS,
openmc.mgxs.MultiplicityMatrixXS))
check_value('energy_groups', nuscatter.energy_groups,
[self.energy_groups])

24
openmc/plotter.py
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@ -676,7 +676,7 @@ def calculate_mgxs(this, types, orders=None, temperature=294.,
for line in range(len(types)):
for g in range(library.energy_groups.num_groups):
data[g * 2: g * 2 + 2] = mgxs[line, g]
data[line, g * 2: g * 2 + 2] = mgxs[line, g]
return energy_grid[::-1], data
@ -775,28 +775,28 @@ def _calculate_mgxs_nuc_macro(this, types, library, orders=None,
data[i, :] = temp_data[orders[i]]
else:
data[i, :] = np.sum(temp_data[:])
elif shape in (xsdata.xs_shapes["[G'][DG]"],
xsdata.xs_shapes["[G][DG]"]):
elif shape in (xsdata.xs_shapes["[DG][G']"],
xsdata.xs_shapes["[DG][G]"]):
# Then we have an array vs groups with values for each
# delayed group. The user-provided value of orders tells us
# which delayed group we want. If none are provided, then
# we sum all the delayed groups together.
if orders[i]:
if orders[i] < len(shape[1]):
data[i, :] = temp_data[:, orders[i]]
if orders[i] < len(shape[0]):
data[i, :] = temp_data[orders[i], :]
else:
data[i, :] = np.sum(temp_data[:, :], axis=1)
elif shape == xsdata.xs_shapes["[G][G'][DG]"]:
data[i, :] = np.sum(temp_data[:, :], axis=0)
elif shape == xsdata.xs_shapes["[DG][G][G']"]:
# Then we have a delayed group matrix. We will first
# remove the outgoing group dependency
temp_data = np.sum(temp_data, axis=1)
temp_data = np.sum(temp_data, axis=-1)
# And then proceed in exactly the same manner as the
# "[G'][DG]" of "[G][DG]" shapes in the previous block.
# "[DG][G']" or "[DG][G]" shapes in the previous block.
if orders[i]:
if orders[i] < len(shape[1]):
data[i, :] = temp_data[:, orders[i]]
if orders[i] < len(shape[0]):
data[i, :] = temp_data[orders[i], :]
else:
data[i, :] = np.sum(temp_data[:, :], axis=1)
data[i, :] = np.sum(temp_data[:, :], axis=0)
elif shape == xsdata.xs_shapes["[G][G'][Order]"]:
# This is a scattering matrix with angular data
# First remove the outgoing group dependence

2
openmc/tallies.py
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@ -195,7 +195,7 @@ class Tally(object):
if isinstance(nuclide, openmc.Nuclide):
string += nuclide.name + ' '
else:
string += nuclide + ' '
string += str(nuclide) + ' '
string += '\n'

8
src/mgxs_header.F90
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@ -2340,7 +2340,7 @@ module mgxs_header
! Now need to compare this material maximum scattering order with
! the problem wide max scatt order and use whichever is lower
order = min(mat_max_order, max_order)
order = min(mat_max_order, max_order + 1)
! Ok, got our order, store the dimensionality
order_dim = order
@ -3467,9 +3467,7 @@ module mgxs_header
type(MaterialMacroXS), intent(inout) :: xs ! Resultant Mgxs Data
xs % total = this % xs(this % index_temp) % total(gin)
xs % elastic = this % xs(this % index_temp) % scatter % scattxs(gin)
xs % absorption = this % xs(this % index_temp) % absorption(gin)
xs % fission = this % xs(this % index_temp) % fission(gin)
xs % nu_fission = &
this % xs(this % index_temp) % prompt_nu_fission(gin) + &
sum(this % xs(this % index_temp) % delayed_nu_fission(:, gin))
@ -3487,12 +3485,8 @@ module mgxs_header
call find_angle(this % polar, this % azimuthal, uvw, iazi, ipol)
xs % total = this % xs(this % index_temp) % &
total(gin, iazi, ipol)
xs % elastic = this % xs(this % index_temp) % &
scatter(iazi, ipol) % obj % scattxs(gin)
xs % absorption = this % xs(this % index_temp) % &
absorption(gin, iazi, ipol)
xs % fission = this % xs(this % index_temp) % &
fission(gin, iazi, ipol)
xs % nu_fission = this % xs(this % index_temp) % &
prompt_nu_fission(gin, iazi, ipol) + &
sum(this % xs(this % index_temp) % &

2
src/tally.F90
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@ -2027,7 +2027,7 @@ contains
end do SCORE_LOOP
nullify(matxs,nucxs)
nullify(matxs, nucxs)
end subroutine score_general_mg
!===============================================================================

2
src/tracking.F90
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@ -105,9 +105,7 @@ contains
p % coord(p % n_coord) % uvw, material_xs)
else
material_xs % total = ZERO
material_xs % elastic = ZERO
material_xs % absorption = ZERO
material_xs % fission = ZERO
material_xs % nu_fission = ZERO
end if
end if