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Merge pull request #1521 from paulromano/update-examples
Freshen up examples/ directory
This commit is contained in:
Patrick Shriwise
GitHub
commit
9150d30d5f
56 changed files with 430 additions and 1250 deletions
140
examples/assembly/assembly.py
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140
examples/assembly/assembly.py
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"""
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This script builds a single PWR assembly and is a slightly more advanced
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demonstration of model building using Python. The creation of two universes for
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fuel pins and guide tube pins has been separated into functions, and then the
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overall model is built by an `assembly` function. This script also demonstrates
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the use of the `Model` class, which provides some extra convenience over using
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`Geometry`, `Materials`, and `Settings` classes directly. Finally, the script
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takes two command-line flags that indicate whether to build and/or run the
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model.
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"""
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import argparse
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from math import log10
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import numpy as np
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import openmc
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# Define surfaces
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fuel_or = openmc.ZCylinder(r=0.39218, name='Fuel OR')
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clad_or = openmc.ZCylinder(r=0.45720, name='Clad OR')
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# Define materials
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fuel = openmc.Material(name='Fuel')
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fuel.set_density('g/cm3', 10.29769)
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fuel.add_nuclide('U234', 4.4843e-6)
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fuel.add_nuclide('U235', 5.5815e-4)
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fuel.add_nuclide('U238', 2.2408e-2)
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fuel.add_nuclide('O16', 4.5829e-2)
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clad = openmc.Material(name='Cladding')
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clad.set_density('g/cm3', 6.55)
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clad.add_nuclide('Zr90', 2.1827e-2)
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clad.add_nuclide('Zr91', 4.7600e-3)
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clad.add_nuclide('Zr92', 7.2758e-3)
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clad.add_nuclide('Zr94', 7.3734e-3)
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clad.add_nuclide('Zr96', 1.1879e-3)
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hot_water = openmc.Material(name='Hot borated water')
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hot_water.set_density('g/cm3', 0.740582)
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hot_water.add_nuclide('H1', 4.9457e-2)
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hot_water.add_nuclide('O16', 2.4672e-2)
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hot_water.add_nuclide('B10', 8.0042e-6)
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hot_water.add_nuclide('B11', 3.2218e-5)
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hot_water.add_s_alpha_beta('c_H_in_H2O')
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def fuel_pin():
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"""Returns a fuel pin universe."""
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fuel_cell = openmc.Cell(fill=fuel, region=-fuel_or)
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clad_cell = openmc.Cell(fill=clad, region=+fuel_or & -clad_or)
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hot_water_cell = openmc.Cell(fill=hot_water, region=+clad_or)
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univ = openmc.Universe(name='Fuel Pin')
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univ.add_cells([fuel_cell, clad_cell, hot_water_cell])
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return univ
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def guide_tube_pin():
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"""Returns a control rod guide tube universe."""
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gt_inner_cell = openmc.Cell(fill=hot_water, region=-fuel_or)
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gt_clad_cell = openmc.Cell(fill=clad, region=+fuel_or & -clad_or)
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gt_outer_cell = openmc.Cell(fill=hot_water, region=+clad_or)
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univ = openmc.Universe(name='Guide Tube')
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univ.add_cells([gt_inner_cell, gt_clad_cell, gt_outer_cell])
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return univ
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def assembly_model():
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"""Returns a single PWR fuel assembly."""
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model = openmc.model.Model()
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# Create fuel assembly Lattice
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pitch = 21.42
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assembly = openmc.RectLattice(name='Fuel Assembly')
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assembly.pitch = (pitch/17, pitch/17)
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assembly.lower_left = (-pitch/2, -pitch/2)
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# Create array indices for guide tube locations in lattice
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gt_pos = np.array([
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[2, 5], [2, 8], [2, 11],
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[3, 3], [3, 13],
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[5, 2], [5, 5], [5, 8], [5, 11], [5, 14],
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[8, 2], [8, 5], [8, 8], [8, 11], [8, 14],
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[11, 2], [11, 5], [11, 8], [11, 11], [11, 14],
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[13, 3], [13, 13],
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[14, 5], [14, 8], [14, 11]
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])
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# Create 17x17 array of universes. First we create a 17x17 array all filled
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# with the fuel pin universe. Then, we replace the guide tube positions with
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# the guide tube pin universe (note the use of numpy fancy indexing to
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# achieve this).
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assembly.universes = np.full((17, 17), fuel_pin())
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assembly.universes[gt_pos[:, 0], gt_pos[:, 1]] = guide_tube_pin()
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# Create outer boundary of the geometry to surround the lattice
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outer_boundary = openmc.model.rectangular_prism(
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pitch, pitch, boundary_type='reflective')
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# Create a cell filled with the lattice
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main_cell = openmc.Cell(fill=assembly, region=outer_boundary)
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# Finally, create geometry by providing a list of cells that fill the root
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# universe
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model.geometry = openmc.Geometry([main_cell])
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model.settings.batches = 150
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model.settings.inactive = 50
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model.settings.particles = 1000
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model.settings.source = openmc.Source(space=openmc.stats.Box(
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(-pitch/2, -pitch/2, -1),
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(pitch/2, pitch/2, 1),
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only_fissionable=True
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))
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# NOTE: We never actually created a Materials object. When you export/run
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# using the Model object, if no materials were assigned it will look through
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# the Geometry object and automatically export any materials that are
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# necessary to build the model.
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return model
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if __name__ == '__main__':
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# Set up command-line arguments for generating/running the model
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parser = argparse.ArgumentParser()
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parser.add_argument('--generate', action='store_true')
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parser.add_argument('--run', action='store_true')
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args = parser.parse_args()
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if args.generate or args.run:
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model = assembly_model()
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if args.generate:
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model.export_to_xml()
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if args.run:
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model.run()
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19
examples/custom_source/README.md
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19
examples/custom_source/README.md
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# Building a Custom Source
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To run this example, you first need to compile the custom source library, which
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requires headers from OpenMC. A CMakeLists.txt file has been set up for you that
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will search for OpenMC and build the custom library. To build the source
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library, you can run:
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mkdir build && cd build
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OPENMC_ROOT=<path_to_openmc_install> cmake ..
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make
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After this, you can build the model by running `python build_xml.py`. In the XML
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files that are created, you should see a reference to build/libsource.so, the
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custom source library that was built by CMake. The model is also set up with a
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mesh tally of the flux, so once you run `openmc`, you will get a statepoint file
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with the tally results in it. Running `python show_flux.py` will pull in the
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results from the statepoint file and display them. If all worked well, you
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should see a ring "imprint" as well as a higher flux to the right side (since
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the custom source has all particles moving in the positive x direction).
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36
examples/custom_source/build_xml.py
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36
examples/custom_source/build_xml.py
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import openmc
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# Create a single material
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iron = openmc.Material()
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iron.set_density('g/cm3', 5.0)
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iron.add_element('Fe', 1.0)
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mats = openmc.Materials([iron])
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mats.export_to_xml()
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# Create a 5 cm x 5 cm box filled with iron
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box = openmc.model.rectangular_prism(10.0, 10.0, boundary_type='vacuum')
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cell = openmc.Cell(fill=iron, region=box)
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geometry = openmc.Geometry([cell])
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geometry.export_to_xml()
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# Tell OpenMC we're going to use our custom source
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settings = openmc.Settings()
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settings.run_mode = 'fixed source'
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settings.batches = 10
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settings.particles = 1000
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source = openmc.Source()
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source.library = 'build/libsource.so'
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settings.source = source
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settings.export_to_xml()
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# Finally, define a mesh tally so that we can see the resulting flux
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mesh = openmc.RegularMesh()
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mesh.lower_left = (-5.0, -5.0)
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mesh.upper_right = (5.0, 5.0)
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mesh.dimension = (50, 50)
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tally = openmc.Tally()
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tally.filters = [openmc.MeshFilter(mesh)]
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tally.scores = ['flux']
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tallies = openmc.Tallies([tally])
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tallies.export_to_xml()
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14
examples/custom_source/show_flux.py
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14
examples/custom_source/show_flux.py
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import matplotlib.pyplot as plt
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import openmc
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# Get the flux from the statepoint
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with openmc.StatePoint('statepoint.10.h5') as sp:
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flux = sp.tallies[1].mean
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flux.shape = (50, 50)
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# Plot the flux
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fig, ax = plt.subplots()
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ax.imshow(flux, origin='lower', extent=(-5.0, 5.0, -5.0, 5.0))
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ax.set_xlabel('x [cm]')
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ax.set_ylabel('y [cm]')
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plt.show()
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33
examples/jezebel/jezebel.py
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33
examples/jezebel/jezebel.py
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import openmc
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# Create plutonium metal material
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pu = openmc.Material()
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pu.set_density('sum')
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pu.add_nuclide('Pu239', 3.7047e-02)
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pu.add_nuclide('Pu240', 1.7512e-03)
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pu.add_nuclide('Pu241', 1.1674e-04)
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pu.add_element('Ga', 1.3752e-03)
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mats = openmc.Materials([pu])
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mats.export_to_xml()
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# Create a single cell filled with the Pu metal
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sphere = openmc.Sphere(r=6.3849, boundary_type='vacuum')
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cell = openmc.Cell(fill=pu, region=-sphere)
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geom = openmc.Geometry([cell])
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geom.export_to_xml()
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# Finally, define some run settings
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settings = openmc.Settings()
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settings.batches = 200
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settings.inactive = 10
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settings.particles = 10000
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settings.export_to_xml()
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# Run the simulation
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openmc.run()
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# Get the resulting k-effective value
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n = settings.batches
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with openmc.StatePoint(f'statepoint.{n}.h5') as sp:
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keff = sp.k_combined
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print(f'Final k-effective = {keff}')
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112
examples/pincell/build_xml.py
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112
examples/pincell/build_xml.py
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from math import log10
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import numpy as np
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import openmc
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###############################################################################
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# Create materials for the problem
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uo2 = openmc.Material(name='UO2 fuel at 2.4% wt enrichment')
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uo2.set_density('g/cm3', 10.29769)
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uo2.add_element('U', 1., enrichment=2.4)
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uo2.add_element('O', 2.)
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helium = openmc.Material(name='Helium for gap')
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helium.set_density('g/cm3', 0.001598)
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helium.add_element('He', 2.4044e-4)
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zircaloy = openmc.Material(name='Zircaloy 4')
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zircaloy.set_density('g/cm3', 6.55)
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zircaloy.add_element('Sn', 0.014 , 'wo')
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zircaloy.add_element('Fe', 0.00165, 'wo')
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zircaloy.add_element('Cr', 0.001 , 'wo')
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zircaloy.add_element('Zr', 0.98335, 'wo')
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borated_water = openmc.Material(name='Borated water')
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borated_water.set_density('g/cm3', 0.740582)
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borated_water.add_element('B', 4.0e-5)
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borated_water.add_element('H', 5.0e-2)
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borated_water.add_element('O', 2.4e-2)
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borated_water.add_s_alpha_beta('c_H_in_H2O')
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# Collect the materials together and export to XML
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materials = openmc.Materials([uo2, helium, zircaloy, borated_water])
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materials.export_to_xml()
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###############################################################################
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# Define problem geometry
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# Create cylindrical surfaces
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fuel_or = openmc.ZCylinder(r=0.39218, name='Fuel OR')
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clad_ir = openmc.ZCylinder(r=0.40005, name='Clad IR')
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clad_or = openmc.ZCylinder(r=0.45720, name='Clad OR')
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# Create a region represented as the inside of a rectangular prism
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pitch = 1.25984
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box = openmc.rectangular_prism(pitch, pitch, boundary_type='reflective')
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# Create cells, mapping materials to regions
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fuel = openmc.Cell(fill=uo2, region=-fuel_or)
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gap = openmc.Cell(fill=helium, region=+fuel_or & -clad_ir)
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clad = openmc.Cell(fill=zircaloy, region=+clad_ir & -clad_or)
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water = openmc.Cell(fill=borated_water, region=+clad_or & box)
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# Create a geometry and export to XML
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geometry = openmc.Geometry([fuel, gap, clad, water])
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geometry.export_to_xml()
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###############################################################################
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# Define problem settings
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# Indicate how many particles to run
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settings = openmc.Settings()
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settings.batches = 100
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settings.inactive = 10
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settings.particles = 1000
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# Create an initial uniform spatial source distribution over fissionable zones
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lower_left = (-pitch/2, -pitch/2, -1)
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upper_right = (pitch/2, pitch/2, 1)
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uniform_dist = openmc.stats.Box(lower_left, upper_right, only_fissionable=True)
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settings.source = openmc.source.Source(space=uniform_dist)
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# For source convergence checks, add a mesh that can be used to calculate the
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# Shannon entropy
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entropy_mesh = openmc.RegularMesh()
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entropy_mesh.lower_left = (-fuel_or.r, -fuel_or.r)
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entropy_mesh.upper_right = (fuel_or.r, fuel_or.r)
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entropy_mesh.dimension = (10, 10)
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settings.entropy_mesh = entropy_mesh
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settings.export_to_xml()
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###############################################################################
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# Define tallies
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# Create a mesh that will be used for tallying
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mesh = openmc.RegularMesh()
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mesh.dimension = (100, 100)
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mesh.lower_left = (-pitch/2, -pitch/2)
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mesh.upper_right = (pitch/2, pitch/2)
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# Create a mesh filter that can be used in a tally
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mesh_filter = openmc.MeshFilter(mesh)
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# Now use the mesh filter in a tally and indicate what scores are desired
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mesh_tally = openmc.Tally(name="Mesh tally")
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mesh_tally.filters = [mesh_filter]
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mesh_tally.scores = ['flux', 'fission', 'nu-fission']
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# Let's also create a tally to get the flux energy spectrum. We start by
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# creating an energy filter
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e_min, e_max = 1e-5, 20.0e6
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groups = 500
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energies = np.logspace(log10(e_min), log10(e_max), groups + 1)
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energy_filter = openmc.EnergyFilter(energies)
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spectrum_tally = openmc.Tally(name="Flux spectrum")
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spectrum_tally.filters = [energy_filter]
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spectrum_tally.scores = ['flux']
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# Instantiate a Tallies collection and export to XML
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tallies = openmc.Tallies([mesh_tally, spectrum_tally])
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tallies.export_to_xml()
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23
examples/pincell/plot_spectrum.py
Normal file
23
examples/pincell/plot_spectrum.py
Normal file
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@ -0,0 +1,23 @@
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import matplotlib.pyplot as plt
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import openmc
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# Get results from statepoint
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with openmc.StatePoint('statepoint.100.h5') as sp:
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t = sp.get_tally(name="Flux spectrum")
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# Get the energies from the energy filter
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energy_filter = t.filters[0]
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energies = energy_filter.bins[:, 0]
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# Get the flux values
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mean = t.get_values(value='mean').ravel()
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uncertainty = t.get_values(value='std_dev').ravel()
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# Plot flux spectrum
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fix, ax = plt.subplots()
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ax.loglog(energies, mean, drawstyle='steps-post')
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ax.set_xlabel('Energy [eV]')
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ax.set_ylabel('Flux')
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ax.grid(True, which='both')
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plt.show()
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@ -1,20 +1,12 @@
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from math import log10
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import numpy as np
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import openmc
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import openmc.mgxs
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###############################################################################
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# Simulation Input File Parameters
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###############################################################################
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||||
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# OpenMC simulation parameters
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batches = 100
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inactive = 10
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particles = 1000
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||||
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||||
###############################################################################
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# Exporting to OpenMC mgxs.h5 file
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###############################################################################
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# Create multigroup data
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# Instantiate the energy group data
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groups = openmc.mgxs.EnergyGroups(group_edges=[
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@ -69,107 +61,90 @@ mg_cross_sections_file = openmc.MGXSLibrary(groups)
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mg_cross_sections_file.add_xsdatas([uo2_xsdata, h2o_xsdata])
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mg_cross_sections_file.export_to_hdf5()
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||||
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||||
###############################################################################
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||||
# Exporting to OpenMC materials.xml file
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||||
###############################################################################
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||||
# Create materials for the problem
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||||
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||||
# Instantiate some Macroscopic Data
|
||||
uo2_data = openmc.Macroscopic('UO2')
|
||||
h2o_data = openmc.Macroscopic('LWTR')
|
||||
|
||||
# Instantiate some Materials and register the appropriate Macroscopic objects
|
||||
uo2 = openmc.Material(material_id=1, name='UO2 fuel')
|
||||
uo2 = openmc.Material(name='UO2 fuel')
|
||||
uo2.set_density('macro', 1.0)
|
||||
uo2.add_macroscopic(uo2_data)
|
||||
|
||||
water = openmc.Material(material_id=2, name='Water')
|
||||
water = openmc.Material(name='Water')
|
||||
water.set_density('macro', 1.0)
|
||||
water.add_macroscopic(h2o_data)
|
||||
|
||||
# Instantiate a Materials collection and export to XML
|
||||
materials_file = openmc.Materials([uo2, water])
|
||||
materials_file.cross_sections = "./mgxs.h5"
|
||||
materials_file.cross_sections = "mgxs.h5"
|
||||
materials_file.export_to_xml()
|
||||
|
||||
|
||||
###############################################################################
|
||||
# Exporting to OpenMC geometry.xml file
|
||||
###############################################################################
|
||||
# Define problem geometry
|
||||
|
||||
# Instantiate ZCylinder surfaces
|
||||
fuel_or = openmc.ZCylinder(surface_id=1, x0=0, y0=0, r=0.54, name='Fuel OR')
|
||||
left = openmc.XPlane(surface_id=4, x0=-0.63, name='left')
|
||||
right = openmc.XPlane(surface_id=5, x0=0.63, name='right')
|
||||
bottom = openmc.YPlane(surface_id=6, y0=-0.63, name='bottom')
|
||||
top = openmc.YPlane(surface_id=7, y0=0.63, name='top')
|
||||
# Create a surface for the fuel outer radius
|
||||
fuel_or = openmc.ZCylinder(r=0.54, name='Fuel OR')
|
||||
|
||||
left.boundary_type = 'reflective'
|
||||
right.boundary_type = 'reflective'
|
||||
top.boundary_type = 'reflective'
|
||||
bottom.boundary_type = 'reflective'
|
||||
# Create a region represented as the inside of a rectangular prism
|
||||
pitch = 1.26
|
||||
box = openmc.rectangular_prism(pitch, pitch, boundary_type='reflective')
|
||||
|
||||
# Instantiate Cells
|
||||
fuel = openmc.Cell(cell_id=1, name='cell 1')
|
||||
moderator = openmc.Cell(cell_id=2, name='cell 2')
|
||||
fuel = openmc.Cell(fill=uo2, region=-fuel_or, name='fuel')
|
||||
moderator = openmc.Cell(fill=water, region=+fuel_or & box, name='moderator')
|
||||
|
||||
# Use surface half-spaces to define regions
|
||||
fuel.region = -fuel_or
|
||||
moderator.region = +fuel_or & +left & -right & +bottom & -top
|
||||
|
||||
# Register Materials with Cells
|
||||
fuel.fill = uo2
|
||||
moderator.fill = water
|
||||
|
||||
# Instantiate Universe
|
||||
root = openmc.Universe(universe_id=0, name='root universe')
|
||||
|
||||
# Register Cells with Universe
|
||||
root.add_cells([fuel, moderator])
|
||||
|
||||
# Instantiate a Geometry, register the root Universe, and export to XML
|
||||
geometry = openmc.Geometry(root)
|
||||
# Create a geometry with the two cells and export to XML
|
||||
geometry = openmc.Geometry([fuel, moderator])
|
||||
geometry.export_to_xml()
|
||||
|
||||
|
||||
###############################################################################
|
||||
# Exporting to OpenMC settings.xml file
|
||||
###############################################################################
|
||||
# Define problem settings
|
||||
|
||||
# Instantiate a Settings object, set all runtime parameters, and export to XML
|
||||
settings_file = openmc.Settings()
|
||||
settings_file.energy_mode = "multi-group"
|
||||
settings_file.batches = batches
|
||||
settings_file.inactive = inactive
|
||||
settings_file.particles = particles
|
||||
settings = openmc.Settings()
|
||||
settings.energy_mode = "multi-group"
|
||||
settings.batches = 100
|
||||
settings.inactive = 10
|
||||
settings.particles = 1000
|
||||
|
||||
# Create an initial uniform spatial source distribution over fissionable zones
|
||||
bounds = [-0.63, -0.63, -1, 0.63, 0.63, 1]
|
||||
uniform_dist = openmc.stats.Box(bounds[:3], bounds[3:])
|
||||
settings_file.source = openmc.source.Source(space=uniform_dist)
|
||||
|
||||
settings_file.export_to_xml()
|
||||
lower_left = (-pitch/2, -pitch/2, -1)
|
||||
upper_right = (pitch/2, pitch/2, 1)
|
||||
uniform_dist = openmc.stats.Box(lower_left, upper_right, only_fissionable=True)
|
||||
settings.source = openmc.source.Source(space=uniform_dist)
|
||||
settings.export_to_xml()
|
||||
|
||||
###############################################################################
|
||||
# Exporting to OpenMC tallies.xml file
|
||||
###############################################################################
|
||||
# Define tallies
|
||||
|
||||
# Instantiate a tally mesh
|
||||
mesh = openmc.RegularMesh(mesh_id=1)
|
||||
mesh.dimension = [100, 100, 1]
|
||||
mesh.lower_left = [-0.63, -0.63, -1.e50]
|
||||
mesh.upper_right = [0.63, 0.63, 1.e50]
|
||||
# Create a mesh that will be used for tallying
|
||||
mesh = openmc.RegularMesh()
|
||||
mesh.dimension = (100, 100)
|
||||
mesh.lower_left = (-pitch/2, -pitch/2)
|
||||
mesh.upper_right = (pitch/2, pitch/2)
|
||||
|
||||
# Instantiate some tally Filters
|
||||
energy_filter = openmc.EnergyFilter([1e-5, 0.0635, 10.0, 1.0e2, 1.0e3, 0.5e6,
|
||||
1.0e6, 20.0e6])
|
||||
# Create a mesh filter that can be used in a tally
|
||||
mesh_filter = openmc.MeshFilter(mesh)
|
||||
|
||||
# Instantiate the Tally
|
||||
tally = openmc.Tally(tally_id=1, name='tally 1')
|
||||
tally.filters = [energy_filter, mesh_filter]
|
||||
tally.scores = ['flux', 'fission', 'nu-fission']
|
||||
# Now use the mesh filter in a tally and indicate what scores are desired
|
||||
mesh_tally = openmc.Tally(name="Mesh tally")
|
||||
mesh_tally.filters = [mesh_filter]
|
||||
mesh_tally.scores = ['flux', 'fission', 'nu-fission']
|
||||
|
||||
# Instantiate a Tallies collection, register all Tallies, and export to XML
|
||||
tallies_file = openmc.Tallies([tally])
|
||||
tallies_file.export_to_xml()
|
||||
# Let's also create a tally to get the flux energy spectrum. We start by
|
||||
# creating an energy filter
|
||||
e_min, e_max = 1e-5, 20.0e6
|
||||
groups = 500
|
||||
energies = np.logspace(log10(e_min), log10(e_max), groups + 1)
|
||||
energy_filter = openmc.EnergyFilter(energies)
|
||||
|
||||
spectrum_tally = openmc.Tally(name="Flux spectrum")
|
||||
spectrum_tally.filters = [energy_filter]
|
||||
spectrum_tally.scores = ['flux']
|
||||
|
||||
# Instantiate a Tallies collection and export to XML
|
||||
tallies = openmc.Tallies([mesh_tally, spectrum_tally])
|
||||
tallies.export_to_xml()
|
||||
|
|
@ -1,121 +0,0 @@
|
|||
import openmc
|
||||
|
||||
|
||||
###############################################################################
|
||||
# Simulation Input File Parameters
|
||||
###############################################################################
|
||||
|
||||
# OpenMC simulation parameters
|
||||
batches = 15
|
||||
inactive = 5
|
||||
particles = 10000
|
||||
|
||||
|
||||
###############################################################################
|
||||
# Exporting to OpenMC materials.xml file
|
||||
###############################################################################
|
||||
|
||||
# Instantiate some Materials and register the appropriate Nuclides
|
||||
moderator = openmc.Material(material_id=41, name='moderator')
|
||||
moderator.set_density('g/cc', 1.0)
|
||||
moderator.add_element('H', 2.)
|
||||
moderator.add_element('O', 1.)
|
||||
moderator.add_s_alpha_beta('c_H_in_H2O')
|
||||
|
||||
fuel = openmc.Material(material_id=40, name='fuel')
|
||||
fuel.set_density('g/cc', 4.5)
|
||||
fuel.add_nuclide('U235', 1.)
|
||||
|
||||
# Instantiate a Materials collection and export to XML
|
||||
materials_file = openmc.Materials([moderator, fuel])
|
||||
materials_file.export_to_xml()
|
||||
|
||||
|
||||
###############################################################################
|
||||
# Exporting to OpenMC geometry.xml file
|
||||
###############################################################################
|
||||
|
||||
# Instantiate ZCylinder surfaces
|
||||
surf1 = openmc.ZCylinder(surface_id=1, x0=0, y0=0, r=7, name='surf 1')
|
||||
surf2 = openmc.ZCylinder(surface_id=2, x0=0, y0=0, r=9, name='surf 2')
|
||||
surf3 = openmc.ZCylinder(surface_id=3, x0=0, y0=0, r=11, name='surf 3')
|
||||
surf3.boundary_type = 'vacuum'
|
||||
|
||||
# Instantiate Cells
|
||||
cell1 = openmc.Cell(cell_id=1, name='cell 1')
|
||||
cell2 = openmc.Cell(cell_id=100, name='cell 2')
|
||||
cell3 = openmc.Cell(cell_id=101, name='cell 3')
|
||||
cell4 = openmc.Cell(cell_id=2, name='cell 4')
|
||||
|
||||
# Use surface half-spaces to define regions
|
||||
cell1.region = -surf2
|
||||
cell2.region = -surf1
|
||||
cell3.region = +surf1
|
||||
cell4.region = +surf2 & -surf3
|
||||
|
||||
# Register Materials with Cells
|
||||
cell2.fill = fuel
|
||||
cell3.fill = moderator
|
||||
cell4.fill = moderator
|
||||
|
||||
# Instantiate Universes
|
||||
universe1 = openmc.Universe(universe_id=37)
|
||||
root = openmc.Universe(universe_id=0, name='root universe')
|
||||
cell1.fill = universe1
|
||||
|
||||
# Register Cells with Universes
|
||||
universe1.add_cells([cell2, cell3])
|
||||
root.add_cells([cell1, cell4])
|
||||
|
||||
# Instantiate a Geometry, register the root Universe, and export to XML
|
||||
geometry = openmc.Geometry(root)
|
||||
geometry.export_to_xml()
|
||||
|
||||
|
||||
###############################################################################
|
||||
# Exporting to OpenMC settings.xml file
|
||||
###############################################################################
|
||||
|
||||
# Instantiate a Settings object, set all runtime parameters, and export to XML
|
||||
settings_file = openmc.Settings()
|
||||
settings_file.batches = batches
|
||||
settings_file.inactive = inactive
|
||||
settings_file.particles = particles
|
||||
|
||||
# Create an initial uniform spatial source distribution over fissionable zones
|
||||
bounds = [-4., -4., -4., 4., 4., 4.]
|
||||
uniform_dist = openmc.stats.Box(bounds[:3], bounds[3:], only_fissionable=True)
|
||||
settings_file.source = openmc.source.Source(space=uniform_dist)
|
||||
|
||||
settings_file.export_to_xml()
|
||||
|
||||
|
||||
###############################################################################
|
||||
# Exporting to OpenMC tallies.xml file
|
||||
###############################################################################
|
||||
|
||||
# Instantiate some tally Filters
|
||||
cell_filter = openmc.CellFilter(cell2)
|
||||
energy_filter = openmc.EnergyFilter([0., 20.e6])
|
||||
energyout_filter = openmc.EnergyoutFilter([0., 20.e6])
|
||||
|
||||
# Instantiate the first Tally
|
||||
first_tally = openmc.Tally(tally_id=1, name='first tally')
|
||||
first_tally.filters = [cell_filter]
|
||||
scores = ['total', 'scatter', 'nu-scatter',
|
||||
'absorption', 'fission', 'nu-fission']
|
||||
first_tally.scores = scores
|
||||
|
||||
# Instantiate the second Tally
|
||||
second_tally = openmc.Tally(tally_id=2, name='second tally')
|
||||
second_tally.filters = [cell_filter, energy_filter]
|
||||
second_tally.scores = scores
|
||||
|
||||
# Instantiate the third Tally
|
||||
third_tally = openmc.Tally(tally_id=3, name='third tally')
|
||||
third_tally.filters = [cell_filter, energy_filter, energyout_filter]
|
||||
third_tally.scores = ['scatter', 'nu-scatter', 'nu-fission']
|
||||
|
||||
# Instantiate a Tallies collection and export to XML
|
||||
tallies_file = openmc.Tallies((first_tally, second_tally, third_tally))
|
||||
tallies_file.export_to_xml()
|
||||
|
|
@ -1,125 +0,0 @@
|
|||
import numpy as np
|
||||
import openmc
|
||||
|
||||
###############################################################################
|
||||
# Simulation Input File Parameters
|
||||
###############################################################################
|
||||
|
||||
# OpenMC simulation parameters
|
||||
batches = 15
|
||||
inactive = 5
|
||||
particles = 10000
|
||||
|
||||
|
||||
###############################################################################
|
||||
# Exporting to OpenMC materials.xml File
|
||||
###############################################################################
|
||||
|
||||
# Instantiate some Materials and register the appropriate Nuclides
|
||||
fuel1 = openmc.Material(material_id=1, name='fuel')
|
||||
fuel1.set_density('g/cc', 4.5)
|
||||
fuel1.add_nuclide('U235', 1.)
|
||||
|
||||
fuel2 = openmc.Material(material_id=2, name='depleted fuel')
|
||||
fuel2.set_density('g/cc', 4.5)
|
||||
fuel2.add_nuclide('U238', 1.)
|
||||
|
||||
moderator = openmc.Material(material_id=3, name='moderator')
|
||||
moderator.set_density('g/cc', 1.0)
|
||||
moderator.add_element('H', 2.)
|
||||
moderator.add_element('O', 1.)
|
||||
moderator.add_s_alpha_beta('c_H_in_H2O')
|
||||
|
||||
# Instantiate a Materials collection and export to XML
|
||||
materials_file = openmc.Materials([fuel1, fuel2, moderator])
|
||||
materials_file.export_to_xml()
|
||||
|
||||
|
||||
###############################################################################
|
||||
# Exporting to OpenMC geometry.xml file
|
||||
###############################################################################
|
||||
|
||||
# Instantiate planar surfaces
|
||||
x1 = openmc.XPlane(surface_id=1, x0=-10)
|
||||
x2 = openmc.XPlane(surface_id=2, x0=-7)
|
||||
x3 = openmc.XPlane(surface_id=3, x0=-4)
|
||||
x4 = openmc.XPlane(surface_id=4, x0=4)
|
||||
x5 = openmc.XPlane(surface_id=5, x0=7)
|
||||
x6 = openmc.XPlane(surface_id=6, x0=10)
|
||||
y1 = openmc.YPlane(surface_id=11, y0=-10)
|
||||
y2 = openmc.YPlane(surface_id=12, y0=-7)
|
||||
y3 = openmc.YPlane(surface_id=13, y0=-4)
|
||||
y4 = openmc.YPlane(surface_id=14, y0=4)
|
||||
y5 = openmc.YPlane(surface_id=15, y0=7)
|
||||
y6 = openmc.YPlane(surface_id=16, y0=10)
|
||||
z1 = openmc.ZPlane(surface_id=21, z0=-10)
|
||||
z2 = openmc.ZPlane(surface_id=22, z0=-7)
|
||||
z3 = openmc.ZPlane(surface_id=23, z0=-4)
|
||||
z4 = openmc.ZPlane(surface_id=24, z0=4)
|
||||
z5 = openmc.ZPlane(surface_id=25, z0=7)
|
||||
z6 = openmc.ZPlane(surface_id=26, z0=10)
|
||||
|
||||
# Set vacuum boundary conditions on outside
|
||||
for surface in [x1, x6, y1, y6, z1, z6]:
|
||||
surface.boundary_type = 'vacuum'
|
||||
|
||||
# Instantiate Cells
|
||||
inner_box = openmc.Cell(cell_id=1, name='inner box')
|
||||
middle_box = openmc.Cell(cell_id=2, name='middle box')
|
||||
outer_box = openmc.Cell(cell_id=3, name='outer box')
|
||||
|
||||
# Use each set of six planes to create solid cube regions. We can then use these
|
||||
# to create cubic shells.
|
||||
inner_cube = +x3 & -x4 & +y3 & -y4 & +z3 & -z4
|
||||
middle_cube = +x2 & -x5 & +y2 & -y5 & +z2 & -z5
|
||||
outer_cube = +x1 & -x6 & +y1 & -y6 & +z1 & -z6
|
||||
outside_inner_cube = -x3 | +x4 | -y3 | +y4 | -z3 | +z4
|
||||
|
||||
# Use surface half-spaces to define regions
|
||||
inner_box.region = inner_cube
|
||||
middle_box.region = middle_cube & outside_inner_cube
|
||||
outer_box.region = outer_cube & ~middle_cube
|
||||
|
||||
# Register Materials with Cells
|
||||
inner_box.fill = fuel1
|
||||
middle_box.fill = fuel2
|
||||
outer_box.fill = moderator
|
||||
|
||||
# Instantiate root universe
|
||||
root = openmc.Universe(universe_id=0, name='root universe')
|
||||
root.add_cells([inner_box, middle_box, outer_box])
|
||||
|
||||
# Instantiate a Geometry, register the root Universe, and export to XML
|
||||
geometry = openmc.Geometry(root)
|
||||
geometry.export_to_xml()
|
||||
|
||||
|
||||
###############################################################################
|
||||
# Exporting to OpenMC settings.xml File
|
||||
###############################################################################
|
||||
|
||||
# Instantiate a Settings object, set all runtime parameters, and export to XML
|
||||
settings_file = openmc.Settings()
|
||||
settings_file.batches = batches
|
||||
settings_file.inactive = inactive
|
||||
settings_file.particles = particles
|
||||
|
||||
# Create an initial uniform spatial source distribution over fissionable zones
|
||||
uniform_dist = openmc.stats.Box(*outer_cube.bounding_box, only_fissionable=True)
|
||||
settings_file.source = openmc.source.Source(space=uniform_dist)
|
||||
|
||||
settings_file.export_to_xml()
|
||||
|
||||
###############################################################################
|
||||
# Exporting to OpenMC plots.xml File
|
||||
###############################################################################
|
||||
|
||||
plot = openmc.Plot(plot_id=1)
|
||||
plot.origin = [0, 0, 0]
|
||||
plot.width = [20, 20]
|
||||
plot.pixels = [200, 200]
|
||||
plot.color_by = 'cell'
|
||||
|
||||
# Instantiate a Plots collection and export to XML
|
||||
plot_file = openmc.Plots([plot])
|
||||
plot_file.export_to_xml()
|
||||
|
|
@ -1,138 +0,0 @@
|
|||
import openmc
|
||||
|
||||
###############################################################################
|
||||
# Simulation Input File Parameters
|
||||
###############################################################################
|
||||
|
||||
# OpenMC simulation parameters
|
||||
batches = 100
|
||||
inactive = 10
|
||||
particles = 1000
|
||||
|
||||
|
||||
###############################################################################
|
||||
# Exporting to OpenMC materials.xml file
|
||||
###############################################################################
|
||||
|
||||
|
||||
# Instantiate some Materials and register the appropriate Nuclides
|
||||
uo2 = openmc.Material(material_id=1, name='UO2 fuel at 2.4% wt enrichment')
|
||||
uo2.set_density('g/cm3', 10.29769)
|
||||
uo2.add_element('U', 1., enrichment=2.4)
|
||||
uo2.add_element('O', 2.)
|
||||
|
||||
helium = openmc.Material(material_id=2, name='Helium for gap')
|
||||
helium.set_density('g/cm3', 0.001598)
|
||||
helium.add_element('He', 2.4044e-4)
|
||||
|
||||
zircaloy = openmc.Material(material_id=3, name='Zircaloy 4')
|
||||
zircaloy.set_density('g/cm3', 6.55)
|
||||
zircaloy.add_element('Sn', 0.014 , 'wo')
|
||||
zircaloy.add_element('Fe', 0.00165, 'wo')
|
||||
zircaloy.add_element('Cr', 0.001 , 'wo')
|
||||
zircaloy.add_element('Zr', 0.98335, 'wo')
|
||||
|
||||
borated_water = openmc.Material(material_id=4, name='Borated water')
|
||||
borated_water.set_density('g/cm3', 0.740582)
|
||||
borated_water.add_element('B', 4.0e-5)
|
||||
borated_water.add_element('H', 5.0e-2)
|
||||
borated_water.add_element('O', 2.4e-2)
|
||||
borated_water.add_s_alpha_beta('c_H_in_H2O')
|
||||
|
||||
# Instantiate a Materials collection and export to XML
|
||||
materials_file = openmc.Materials([uo2, helium, zircaloy, borated_water])
|
||||
materials_file.export_to_xml()
|
||||
|
||||
|
||||
###############################################################################
|
||||
# Exporting to OpenMC geometry.xml file
|
||||
###############################################################################
|
||||
|
||||
# Instantiate ZCylinder surfaces
|
||||
fuel_or = openmc.ZCylinder(surface_id=1, x0=0, y0=0, r=0.39218, name='Fuel OR')
|
||||
clad_ir = openmc.ZCylinder(surface_id=2, x0=0, y0=0, r=0.40005, name='Clad IR')
|
||||
clad_or = openmc.ZCylinder(surface_id=3, x0=0, y0=0, r=0.45720, name='Clad OR')
|
||||
left = openmc.XPlane(surface_id=4, x0=-0.62992, name='left')
|
||||
right = openmc.XPlane(surface_id=5, x0=0.62992, name='right')
|
||||
bottom = openmc.YPlane(surface_id=6, y0=-0.62992, name='bottom')
|
||||
top = openmc.YPlane(surface_id=7, y0=0.62992, name='top')
|
||||
|
||||
left.boundary_type = 'reflective'
|
||||
right.boundary_type = 'reflective'
|
||||
top.boundary_type = 'reflective'
|
||||
bottom.boundary_type = 'reflective'
|
||||
|
||||
# Instantiate Cells
|
||||
fuel = openmc.Cell(cell_id=1, name='cell 1')
|
||||
gap = openmc.Cell(cell_id=2, name='cell 2')
|
||||
clad = openmc.Cell(cell_id=3, name='cell 3')
|
||||
water = openmc.Cell(cell_id=4, name='cell 4')
|
||||
|
||||
# Use surface half-spaces to define regions
|
||||
fuel.region = -fuel_or
|
||||
gap.region = +fuel_or & -clad_ir
|
||||
clad.region = +clad_ir & -clad_or
|
||||
water.region = +clad_or & +left & -right & +bottom & -top
|
||||
|
||||
# Register Materials with Cells
|
||||
fuel.fill = uo2
|
||||
gap.fill = helium
|
||||
clad.fill = zircaloy
|
||||
water.fill = borated_water
|
||||
|
||||
# Instantiate Universe
|
||||
root = openmc.Universe(universe_id=0, name='root universe')
|
||||
|
||||
# Register Cells with Universe
|
||||
root.add_cells([fuel, gap, clad, water])
|
||||
|
||||
# Instantiate a Geometry, register the root Universe, and export to XML
|
||||
geometry = openmc.Geometry(root)
|
||||
geometry.export_to_xml()
|
||||
|
||||
|
||||
###############################################################################
|
||||
# Exporting to OpenMC settings.xml file
|
||||
###############################################################################
|
||||
|
||||
# Instantiate a Settings object, set all runtime parameters, and export to XML
|
||||
settings_file = openmc.Settings()
|
||||
settings_file.batches = batches
|
||||
settings_file.inactive = inactive
|
||||
settings_file.particles = particles
|
||||
|
||||
# Create an initial uniform spatial source distribution over fissionable zones
|
||||
bounds = [-0.62992, -0.62992, -1, 0.62992, 0.62992, 1]
|
||||
uniform_dist = openmc.stats.Box(bounds[:3], bounds[3:], only_fissionable=True)
|
||||
settings_file.source = openmc.source.Source(space=uniform_dist)
|
||||
|
||||
entropy_mesh = openmc.RegularMesh()
|
||||
entropy_mesh.lower_left = [-0.39218, -0.39218, -1.e50]
|
||||
entropy_mesh.upper_right = [0.39218, 0.39218, 1.e50]
|
||||
entropy_mesh.dimension = [10, 10, 1]
|
||||
settings_file.entropy_mesh = entropy_mesh
|
||||
settings_file.export_to_xml()
|
||||
|
||||
|
||||
###############################################################################
|
||||
# Exporting to OpenMC tallies.xml file
|
||||
###############################################################################
|
||||
|
||||
# Instantiate a tally mesh
|
||||
mesh = openmc.RegularMesh()
|
||||
mesh.dimension = [100, 100, 1]
|
||||
mesh.lower_left = [-0.62992, -0.62992, -1.e50]
|
||||
mesh.upper_right = [0.62992, 0.62992, 1.e50]
|
||||
|
||||
# Instantiate some tally Filters
|
||||
energy_filter = openmc.EnergyFilter([0., 4., 20.e6])
|
||||
mesh_filter = openmc.MeshFilter(mesh)
|
||||
|
||||
# Instantiate the Tally
|
||||
tally = openmc.Tally(tally_id=1, name='tally 1')
|
||||
tally.filters = [energy_filter, mesh_filter]
|
||||
tally.scores = ['flux', 'fission', 'nu-fission']
|
||||
|
||||
# Instantiate a Tallies collection and export to XML
|
||||
tallies_file = openmc.Tallies([tally])
|
||||
tallies_file.export_to_xml()
|
||||
|
|
@ -1,82 +0,0 @@
|
|||
import numpy as np
|
||||
import openmc
|
||||
|
||||
###############################################################################
|
||||
# Simulation Input File Parameters
|
||||
###############################################################################
|
||||
|
||||
# OpenMC simulation parameters
|
||||
batches = 500
|
||||
inactive = 10
|
||||
particles = 10000
|
||||
|
||||
|
||||
###############################################################################
|
||||
# Exporting to OpenMC materials.xml file
|
||||
###############################################################################
|
||||
|
||||
# Instantiate a Material and register the Nuclide
|
||||
fuel = openmc.Material(material_id=1, name='fuel')
|
||||
fuel.set_density('g/cc', 4.5)
|
||||
fuel.add_nuclide('U235', 1.)
|
||||
|
||||
# Instantiate a Materials collection and export to XML
|
||||
materials_file = openmc.Materials([fuel])
|
||||
materials_file.export_to_xml()
|
||||
|
||||
|
||||
###############################################################################
|
||||
# Exporting to OpenMC geometry.xml file
|
||||
###############################################################################
|
||||
|
||||
# Instantiate Surfaces
|
||||
surf1 = openmc.XPlane(surface_id=1, x0=-1, name='surf 1')
|
||||
surf2 = openmc.XPlane(surface_id=2, x0=+1, name='surf 2')
|
||||
surf3 = openmc.YPlane(surface_id=3, y0=-1, name='surf 3')
|
||||
surf4 = openmc.YPlane(surface_id=4, y0=+1, name='surf 4')
|
||||
surf5 = openmc.ZPlane(surface_id=5, z0=-1, name='surf 5')
|
||||
surf6 = openmc.ZPlane(surface_id=6, z0=+1, name='surf 6')
|
||||
|
||||
surf1.boundary_type = 'vacuum'
|
||||
surf2.boundary_type = 'vacuum'
|
||||
surf3.boundary_type = 'reflective'
|
||||
surf4.boundary_type = 'reflective'
|
||||
surf5.boundary_type = 'reflective'
|
||||
surf6.boundary_type = 'reflective'
|
||||
|
||||
# Instantiate Cell
|
||||
cell = openmc.Cell(cell_id=1, name='cell 1')
|
||||
|
||||
# Use surface half-spaces to define region
|
||||
cell.region = +surf1 & -surf2 & +surf3 & -surf4 & +surf5 & -surf6
|
||||
|
||||
# Register Material with Cell
|
||||
cell.fill = fuel
|
||||
|
||||
# Instantiate Universes
|
||||
root = openmc.Universe(universe_id=0, name='root universe')
|
||||
|
||||
# Register Cell with Universe
|
||||
root.add_cell(cell)
|
||||
|
||||
# Instantiate a Geometry, register the root Universe, and export to XML
|
||||
geometry = openmc.Geometry(root)
|
||||
geometry.export_to_xml()
|
||||
|
||||
|
||||
###############################################################################
|
||||
# Exporting to OpenMC settings.xml file
|
||||
###############################################################################
|
||||
|
||||
# Instantiate a Settings object, set all runtime parameters, and export to XML
|
||||
settings_file = openmc.Settings()
|
||||
settings_file.batches = batches
|
||||
settings_file.inactive = inactive
|
||||
settings_file.particles = particles
|
||||
|
||||
# Create an initial uniform spatial source distribution over fissionable zones
|
||||
uniform_dist = openmc.stats.Box(*cell.region.bounding_box,
|
||||
only_fissionable=True)
|
||||
settings_file.source = openmc.source.Source(space=uniform_dist)
|
||||
|
||||
settings_file.export_to_xml()
|
||||
|
|
@ -1,15 +0,0 @@
|
|||
<?xml version="1.0"?>
|
||||
<geometry>
|
||||
|
||||
<!-- Definition of Cells -->
|
||||
<cell id="1" universe="0" fill="37" region="-2" />
|
||||
<cell id="100" universe="37" material="40" region="-1" />
|
||||
<cell id="101" universe="37" material="41" region="1" />
|
||||
<cell id="2" universe="0" material="41" region="2 -3" />
|
||||
|
||||
<!-- Defition of Surfaces -->
|
||||
<surface id="1" type="z-cylinder" coeffs="0 0 7" />
|
||||
<surface id="2" type="z-cylinder" coeffs="0 0 9" />
|
||||
<surface id="3" type="z-cylinder" coeffs="0 0 11" boundary="vacuum" />
|
||||
|
||||
</geometry>
|
||||
|
|
@ -1,16 +0,0 @@
|
|||
<?xml version="1.0"?>
|
||||
<materials>
|
||||
|
||||
<material id="40">
|
||||
<density value="4.5" units="g/cc" />
|
||||
<nuclide name="U235" ao="1.0" />
|
||||
</material>
|
||||
|
||||
<material id="41">
|
||||
<density value="1.0" units="g/cc" />
|
||||
<nuclide name="H1" ao="2.0" />
|
||||
<nuclide name="O16" ao="1.0" />
|
||||
<sab name="c_H_in_H2O"/>
|
||||
</material>
|
||||
|
||||
</materials>
|
||||
|
|
@ -1,16 +0,0 @@
|
|||
<?xml version="1.0"?>
|
||||
<settings>
|
||||
|
||||
<run_mode>eigenvalue</run_mode>
|
||||
<batches>15</batches>
|
||||
<inactive>5</inactive>
|
||||
<particles>10000</particles>
|
||||
|
||||
<!-- Starting source -->
|
||||
<source>
|
||||
<space type="box">
|
||||
<parameters>-4 -4 -4 4 4 4</parameters>
|
||||
</space>
|
||||
</source>
|
||||
|
||||
</settings>
|
||||
|
|
@ -1,31 +0,0 @@
|
|||
<?xml version="1.0"?>
|
||||
<tallies>
|
||||
|
||||
<filter id="1" type="cell">
|
||||
<bins>100</bins>
|
||||
</filter>
|
||||
|
||||
<filter id="2" type="energy">
|
||||
<bins>0 20.0e6</bins>
|
||||
</filter>
|
||||
|
||||
<filter id="3" type="energyout">
|
||||
<bins>0 20.0e6</bins>
|
||||
</filter>
|
||||
|
||||
<tally id="1">
|
||||
<filters>1</filters>
|
||||
<scores>total scatter nu-scatter absorption fission nu-fission</scores>
|
||||
</tally>
|
||||
|
||||
<tally id="2">
|
||||
<filters>1 2</filters>
|
||||
<scores>total scatter nu-scatter absorption fission nu-fission</scores>
|
||||
</tally>
|
||||
|
||||
<tally id="3">
|
||||
<filters>1 2 3</filters>
|
||||
<scores>scatter nu-scatter nu-fission</scores>
|
||||
</tally>
|
||||
|
||||
</tallies>
|
||||
|
|
@ -1,39 +0,0 @@
|
|||
<?xml version="1.0"?>
|
||||
<geometry>
|
||||
|
||||
<!--
|
||||
This example consists of three nested boxes, and is meant to show how to
|
||||
use Boolean operators to construct complex cell regions.
|
||||
-->
|
||||
|
||||
<surface id="1" type="x-plane" coeffs="-10" boundary="vacuum" />
|
||||
<surface id="2" type="x-plane" coeffs="-7" />
|
||||
<surface id="3" type="x-plane" coeffs="-4" />
|
||||
<surface id="4" type="x-plane" coeffs="4" />
|
||||
<surface id="5" type="x-plane" coeffs="7" />
|
||||
<surface id="6" type="x-plane" coeffs="10" boundary="vacuum" />
|
||||
|
||||
<surface id="11" type="y-plane" coeffs="-10" boundary="vacuum" />
|
||||
<surface id="12" type="y-plane" coeffs="-7" />
|
||||
<surface id="13" type="y-plane" coeffs="-4" />
|
||||
<surface id="14" type="y-plane" coeffs="4" />
|
||||
<surface id="15" type="y-plane" coeffs="7" />
|
||||
<surface id="16" type="y-plane" coeffs="10" boundary="vacuum" />
|
||||
|
||||
<surface id="21" type="z-plane" coeffs="-10" boundary="vacuum" />
|
||||
<surface id="22" type="z-plane" coeffs="-7" />
|
||||
<surface id="23" type="z-plane" coeffs="-4" />
|
||||
<surface id="24" type="z-plane" coeffs="4" />
|
||||
<surface id="25" type="z-plane" coeffs="7" />
|
||||
<surface id="26" type="z-plane" coeffs="10" boundary="vacuum" />
|
||||
|
||||
<!-- Innermost cube -->
|
||||
<cell id="1" material="1" region="3 -4 13 -14 23 -24" />
|
||||
|
||||
<!-- Middle cubic shell -->
|
||||
<cell id="2" material="2" region="2 -5 12 -15 22 -25 (-3 | 4 | -13 | 14 | -23 | 24)" />
|
||||
|
||||
<!-- Outermost cubic shell -->
|
||||
<cell id="3" material="3" region="1 -6 11 -16 21 -26 ~(2 -5 12 -15 22 -25)" />
|
||||
|
||||
</geometry>
|
||||
|
|
@ -1,21 +0,0 @@
|
|||
<?xml version="1.0"?>
|
||||
<materials>
|
||||
|
||||
<material id="1">
|
||||
<density value="4.5" units="g/cc" />
|
||||
<nuclide name="U235" ao="1.0" />
|
||||
</material>
|
||||
|
||||
<material id="2">
|
||||
<density value="4.5" units="g/cc" />
|
||||
<nuclide name="U238" ao="1.0" />
|
||||
</material>
|
||||
|
||||
<material id="3">
|
||||
<density value="1.0" units="g/cc" />
|
||||
<nuclide name="O16" ao="1.0" />
|
||||
<nuclide name="H1" ao="2.0" />
|
||||
<sab name="c_H_in_H2O" />
|
||||
</material>
|
||||
|
||||
</materials>
|
||||
|
|
@ -1,9 +0,0 @@
|
|||
<?xml version="1.0"?>
|
||||
<plots>
|
||||
<plot id="1" type="slice">
|
||||
<color_by>cell</color_by>
|
||||
<origin>0. 0. 0.</origin>
|
||||
<width>20. 20.</width>
|
||||
<pixels>200 200</pixels>
|
||||
</plot>
|
||||
</plots>
|
||||
|
|
@ -1,15 +0,0 @@
|
|||
<?xml version="1.0"?>
|
||||
<settings>
|
||||
|
||||
<!-- Parameters for k-eigenvalue calculation -->
|
||||
<run_mode>eigenvalue</run_mode>
|
||||
<batches>15</batches>
|
||||
<inactive>5</inactive>
|
||||
<particles>10000</particles>
|
||||
|
||||
<!-- Starting source -->
|
||||
<source>
|
||||
<space type="box" parameters="-10. -10. -10. 10. 10. 10." />
|
||||
</source>
|
||||
|
||||
</settings>
|
||||
|
|
@ -1,15 +0,0 @@
|
|||
<?xml version="1.0"?>
|
||||
<geometry>
|
||||
|
||||
<!-- Definition of Cells -->
|
||||
<cell id="1" universe="0" fill="37" region="-2" />
|
||||
<cell id="100" universe="37" material="40" region="-1" />
|
||||
<cell id="101" universe="37" material="41" region="1" />
|
||||
<cell id="2" universe="0" material="41" region="2 -3" />
|
||||
|
||||
<!-- Defition of Surfaces -->
|
||||
<surface id="1" type="z-cylinder" coeffs="0 0 7" />
|
||||
<surface id="2" type="z-cylinder" coeffs="0 0 9" />
|
||||
<surface id="3" type="z-cylinder" coeffs="0 0 11" boundary="vacuum" />
|
||||
|
||||
</geometry>
|
||||
|
|
@ -1,16 +0,0 @@
|
|||
<?xml version="1.0"?>
|
||||
<materials>
|
||||
|
||||
<material id="40">
|
||||
<density value="4.5" units="g/cc" />
|
||||
<nuclide name="U235" ao="1.0" />
|
||||
</material>
|
||||
|
||||
<material id="41">
|
||||
<density value="1.0" units="g/cc" />
|
||||
<nuclide name="H1" ao="2.0" />
|
||||
<nuclide name="O16" ao="1.0" />
|
||||
<sab name="c_H_in_H2O"/>
|
||||
</material>
|
||||
|
||||
</materials>
|
||||
|
|
@ -1,14 +0,0 @@
|
|||
<?xml version="1.0"?>
|
||||
<settings>
|
||||
|
||||
<run_mode>fixed source</run_mode>
|
||||
<batches>10</batches>
|
||||
<inactive>0</inactive>
|
||||
<particles>100000</particles>
|
||||
|
||||
<!-- Starting source -->
|
||||
<source>
|
||||
<library>build/libsource.so</library>
|
||||
</source>
|
||||
|
||||
</settings>
|
||||
|
|
@ -1,17 +0,0 @@
|
|||
<?xml version="1.0"?>
|
||||
<tallies>
|
||||
|
||||
<filter id="1" type="cell">
|
||||
<bins>100</bins>
|
||||
</filter>
|
||||
|
||||
<filter id="2" type="energy">
|
||||
<bins>0 20.0e6</bins>
|
||||
</filter>
|
||||
|
||||
<tally id="3">
|
||||
<filters>1 2 </filters>
|
||||
<scores>flux</scores>
|
||||
</tally>
|
||||
|
||||
</tallies>
|
||||
|
|
@ -1,43 +0,0 @@
|
|||
<?xml version="1.0"?>
|
||||
<geometry>
|
||||
|
||||
<cell id="1" fill="6" region="1 -2 3 -4" />
|
||||
<cell id="2" universe="5" fill="4" region="1 -2 3 -4" />
|
||||
<cell id="101" universe="1" material="1" region="-5" />
|
||||
<cell id="102" universe="1" material="2" region="5" />
|
||||
<cell id="201" universe="2" material="1" region="-6" />
|
||||
<cell id="202" universe="2" material="2" region="6" />
|
||||
<cell id="301" universe="3" material="1" region="-7" />
|
||||
<cell id="302" universe="3" material="2" region="7" />
|
||||
|
||||
<!-- 4 x 4 assembly -->
|
||||
<lattice id="4">
|
||||
<dimension>2 2</dimension>
|
||||
<lower_left>-1.0 -1.0</lower_left>
|
||||
<pitch>1.0 1.0</pitch>
|
||||
<universes>
|
||||
1 2
|
||||
2 3
|
||||
</universes>
|
||||
</lattice>
|
||||
|
||||
<!-- 4 x 4 core -->
|
||||
<lattice id="6">
|
||||
<dimension>2 2</dimension>
|
||||
<lower_left>-2.0 -2.0</lower_left>
|
||||
<pitch>2.0 2.0</pitch>
|
||||
<universes>
|
||||
5 5
|
||||
5 5
|
||||
</universes>
|
||||
</lattice>
|
||||
|
||||
<surface id="1" type="x-plane" coeffs="-2.0" boundary="vacuum" />
|
||||
<surface id="2" type="x-plane" coeffs="2.0" boundary="vacuum" />
|
||||
<surface id="3" type="y-plane" coeffs="-2.0" boundary="vacuum" />
|
||||
<surface id="4" type="y-plane" coeffs="2.0" boundary="vacuum" />
|
||||
<surface id="5" type="z-cylinder" coeffs="0.0 0.0 0.4" />
|
||||
<surface id="6" type="z-cylinder" coeffs="0.0 0.0 0.3" />
|
||||
<surface id="7" type="z-cylinder" coeffs="0.0 0.0 0.2" />
|
||||
|
||||
</geometry>
|
||||
|
|
@ -1,17 +0,0 @@
|
|||
<?xml version="1.0"?>
|
||||
<materials>
|
||||
|
||||
<!-- Definition of materials -->
|
||||
<material id="1">
|
||||
<density value="4.5" units="g/cc" />
|
||||
<nuclide name="U235" ao="1.0" />
|
||||
</material>
|
||||
|
||||
<material id="2">
|
||||
<density value="1.0" units="g/cc" />
|
||||
<nuclide name="H1" ao="2.0" />
|
||||
<nuclide name="O16" ao="1.0" />
|
||||
<sab name="c_H_in_H2O" />
|
||||
</material>
|
||||
|
||||
</materials>
|
||||
|
|
@ -1,10 +0,0 @@
|
|||
<?xml version="1.0"?>
|
||||
<plots>
|
||||
|
||||
<plot id="1" color_by="material">
|
||||
<origin>0. 0. 0.</origin>
|
||||
<width>4.0 4.0</width>
|
||||
<pixels>400 400</pixels>
|
||||
</plot>
|
||||
|
||||
</plots>
|
||||
|
|
@ -1,17 +0,0 @@
|
|||
<?xml version="1.0"?>
|
||||
<settings>
|
||||
|
||||
<!-- Parameters for k-eigenvalue calculation -->
|
||||
<run_mode>eigenvalue</run_mode>
|
||||
<batches>20</batches>
|
||||
<inactive>10</inactive>
|
||||
<particles>10000</particles>
|
||||
|
||||
<!-- Starting source -->
|
||||
<source>
|
||||
<space type="box">
|
||||
<parameters>-1 -1 -1 1 1 1</parameters>
|
||||
</space>
|
||||
</source>
|
||||
|
||||
</settings>
|
||||
|
|
@ -1,20 +0,0 @@
|
|||
<?xml version="1.0"?>
|
||||
<tallies>
|
||||
|
||||
<mesh id="1">
|
||||
<type>regular</type>
|
||||
<dimension>4 4</dimension>
|
||||
<lower_left>-2.0 -2.0</lower_left>
|
||||
<width>1.0 1.0</width>
|
||||
</mesh>
|
||||
|
||||
<filter id="1" type="mesh">
|
||||
<bins>1</bins>
|
||||
</filter>
|
||||
|
||||
<tally id="1">
|
||||
<filters>1</filters>
|
||||
<scores>total</scores>
|
||||
</tally>
|
||||
|
||||
</tallies>
|
||||
|
|
@ -1,32 +0,0 @@
|
|||
<?xml version="1.0"?>
|
||||
<geometry>
|
||||
|
||||
<cell id="1" fill="5" region="1 -2 3 -4" />
|
||||
<cell id="101" universe="1" material="1" region="-5" />
|
||||
<cell id="102" universe="1" material="2" region="5" />
|
||||
<cell id="201" universe="2" material="1" region="-6" />
|
||||
<cell id="202" universe="2" material="2" region="6" />
|
||||
<cell id="301" universe="3" material="1" region="-7" />
|
||||
<cell id="302" universe="3" material="2" region="7" />
|
||||
|
||||
<lattice id="5">
|
||||
<dimension>4 4</dimension>
|
||||
<lower_left>-2.0 -2.0</lower_left>
|
||||
<pitch>1.0 1.0</pitch>
|
||||
<universes>
|
||||
1 2 1 2
|
||||
2 3 2 3
|
||||
1 2 1 2
|
||||
2 3 2 3
|
||||
</universes>
|
||||
</lattice>
|
||||
|
||||
<surface id="1" type="x-plane" coeffs="-2.0" boundary="vacuum" />
|
||||
<surface id="2" type="x-plane" coeffs="2.0" boundary="vacuum" />
|
||||
<surface id="3" type="y-plane" coeffs="-2.0" boundary="vacuum" />
|
||||
<surface id="4" type="y-plane" coeffs="2.0" boundary="vacuum" />
|
||||
<surface id="5" type="z-cylinder" coeffs="0.0 0.0 0.4" />
|
||||
<surface id="6" type="z-cylinder" coeffs="0.0 0.0 0.3" />
|
||||
<surface id="7" type="z-cylinder" coeffs="0.0 0.0 0.2" />
|
||||
|
||||
</geometry>
|
||||
|
|
@ -1,17 +0,0 @@
|
|||
<?xml version="1.0"?>
|
||||
<materials>
|
||||
|
||||
<!-- Definition of materials -->
|
||||
<material id="1">
|
||||
<density value="4.5" units="g/cc" />
|
||||
<nuclide name="U235" ao="1.0" />
|
||||
</material>
|
||||
|
||||
<material id="2">
|
||||
<density value="1.0" units="g/cc" />
|
||||
<nuclide name="H1" ao="2.0" />
|
||||
<nuclide name="O16" ao="1.0" />
|
||||
<sab name="c_H_in_H2O" />
|
||||
</material>
|
||||
|
||||
</materials>
|
||||
|
|
@ -1,10 +0,0 @@
|
|||
<?xml version="1.0"?>
|
||||
<plots>
|
||||
|
||||
<plot id="1" color_by="material">
|
||||
<origin>0. 0. 0.</origin>
|
||||
<width>4.0 4.0</width>
|
||||
<pixels>400 400</pixels>
|
||||
</plot>
|
||||
|
||||
</plots>
|
||||
|
|
@ -1,17 +0,0 @@
|
|||
<?xml version="1.0"?>
|
||||
<settings>
|
||||
|
||||
<!-- Parameters for k-eigenvalue calculation -->
|
||||
<run_mode>eigenvalue</run_mode>
|
||||
<batches>20</batches>
|
||||
<inactive>10</inactive>
|
||||
<particles>10000</particles>
|
||||
|
||||
<!-- Starting source -->
|
||||
<source>
|
||||
<space type="box">
|
||||
<parameters>-1 -1 -1 1 1 1</parameters>
|
||||
</space>
|
||||
</source>
|
||||
|
||||
</settings>
|
||||
|
|
@ -1,20 +0,0 @@
|
|||
<?xml version="1.0"?>
|
||||
<tallies>
|
||||
|
||||
<mesh id="1">
|
||||
<type>regular</type>
|
||||
<dimension>4 4</dimension>
|
||||
<lower_left>-2.0 -2.0</lower_left>
|
||||
<width>1.0 1.0</width>
|
||||
</mesh>
|
||||
|
||||
<filter id="1" type="mesh">
|
||||
<bins>1</bins>
|
||||
</filter>
|
||||
|
||||
<tally id="1">
|
||||
<filters>1</filters>
|
||||
<scores>total</scores>
|
||||
</tally>
|
||||
|
||||
</tallies>
|
||||
|
|
@ -1,27 +0,0 @@
|
|||
<?xml version="1.0"?>
|
||||
<geometry>
|
||||
|
||||
<!--
|
||||
This is a simple pin cell model based on dimensions from the MIT BEAVRS
|
||||
(Benchmarking for Evaluation and Validation of Reactor Simulations)
|
||||
benchmark.
|
||||
-->
|
||||
|
||||
<!-- Surfaces for fuel, gap, cladding. Dimensions from Figure 2 in BEAVRS -->
|
||||
<surface id="1" type="z-cylinder" coeffs="0. 0. 0.39218" /> <!-- Fuel OR -->
|
||||
<surface id="2" type="z-cylinder" coeffs="0. 0. 0.40005" /> <!-- Clad IR -->
|
||||
<surface id="3" type="z-cylinder" coeffs="0. 0. 0.45720" /> <!-- Clad OR -->
|
||||
|
||||
<!-- Reflective surfaces on outside of pin-cell. The lattice pitch is 1.25984
|
||||
cm (taken from Table 2 in BEAVRS). -->
|
||||
<surface id="4" type="x-plane" coeffs="-0.62992" boundary="reflective" />
|
||||
<surface id="5" type="x-plane" coeffs=" 0.62992" boundary="reflective" />
|
||||
<surface id="6" type="y-plane" coeffs="-0.62992" boundary="reflective" />
|
||||
<surface id="7" type="y-plane" coeffs=" 0.62992" boundary="reflective" />
|
||||
|
||||
<cell id="1" material="1" region=" -1" /> <!-- UO2 Fuel -->
|
||||
<cell id="2" material="2" region="1 -2" /> <!-- Helium gap -->
|
||||
<cell id="3" material="3" region="2 -3" /> <!-- Zircaloy cladding -->
|
||||
<cell id="4" material="4" region="3 4 -5 6 -7" /> <!-- Borated water -->
|
||||
|
||||
</geometry>
|
||||
|
|
@ -1,67 +0,0 @@
|
|||
<?xml version="1.0"?>
|
||||
<materials>
|
||||
|
||||
<!--
|
||||
Since O-18 is not present in ENDF/B-VII, it was necessary to combine the
|
||||
atom densities for O-17 and O-18 in any materials containing Oxygen.
|
||||
-->
|
||||
|
||||
<!-- UO2 fuel at 2.4 wt% enrichment -->
|
||||
<material id="1">
|
||||
<density value="10.29769" units="g/cm3" />
|
||||
<nuclide name="U234" ao="4.4843e-06" />
|
||||
<nuclide name="U235" ao="5.5815e-04" />
|
||||
<nuclide name="U238" ao="2.2408e-02" />
|
||||
<nuclide name="O16" ao="4.5829e-02" />
|
||||
<nuclide name="O17" ao="1.1164e-04" />
|
||||
</material>
|
||||
|
||||
<!-- Helium for gap -->
|
||||
<material id="2">
|
||||
<density value="0.001598" units="g/cm3" />
|
||||
<nuclide name="He4" ao="2.4044e-04" />
|
||||
</material>
|
||||
|
||||
<!-- Zircaloy 4 -->
|
||||
<material id="3">
|
||||
<density value="6.55" units="g/cm3" />
|
||||
<nuclide name="O16" ao="3.0743e-04" />
|
||||
<nuclide name="O17" ao="7.4887e-07" />
|
||||
<nuclide name="Cr50" ao="3.2962e-06" />
|
||||
<nuclide name="Cr52" ao="6.3564e-05" />
|
||||
<nuclide name="Cr53" ao="7.2076e-06" />
|
||||
<nuclide name="Cr54" ao="1.7941e-06" />
|
||||
<nuclide name="Fe54" ao="8.6699e-06" />
|
||||
<nuclide name="Fe56" ao="1.3610e-04" />
|
||||
<nuclide name="Fe57" ao="3.1431e-06" />
|
||||
<nuclide name="Fe58" ao="4.1829e-07" />
|
||||
<nuclide name="Zr90" ao="2.1827e-02" />
|
||||
<nuclide name="Zr91" ao="4.7600e-03" />
|
||||
<nuclide name="Zr92" ao="7.2758e-03" />
|
||||
<nuclide name="Zr94" ao="7.3734e-03" />
|
||||
<nuclide name="Zr96" ao="1.1879e-03" />
|
||||
<nuclide name="Sn112" ao="4.6735e-06" />
|
||||
<nuclide name="Sn114" ao="3.1799e-06" />
|
||||
<nuclide name="Sn115" ao="1.6381e-06" />
|
||||
<nuclide name="Sn116" ao="7.0055e-05" />
|
||||
<nuclide name="Sn117" ao="3.7003e-05" />
|
||||
<nuclide name="Sn118" ao="1.1669e-04" />
|
||||
<nuclide name="Sn119" ao="4.1387e-05" />
|
||||
<nuclide name="Sn120" ao="1.5697e-04" />
|
||||
<nuclide name="Sn122" ao="2.2308e-05" />
|
||||
<nuclide name="Sn124" ao="2.7897e-05" />
|
||||
</material>
|
||||
|
||||
<!-- Borated water at 975 ppm -->
|
||||
<material id="4">
|
||||
<density value="0.740582" units="g/cm3" />
|
||||
<nuclide name="B10" ao="8.0042e-06" />
|
||||
<nuclide name="B11" ao="3.2218e-05" />
|
||||
<nuclide name="H1" ao="4.9457e-02" />
|
||||
<nuclide name="H2" ao="7.4196e-06" />
|
||||
<nuclide name="O16" ao="2.4672e-02" />
|
||||
<nuclide name="O17" ao="6.0099e-05" />
|
||||
<sab name="c_H_in_H2O" />
|
||||
</material>
|
||||
|
||||
</materials>
|
||||
|
|
@ -1,32 +0,0 @@
|
|||
<?xml version="1.0"?>
|
||||
<settings>
|
||||
|
||||
<!-- Define how many particles to run and for how many batches -->
|
||||
<run_mode>eigenvalue</run_mode>
|
||||
<batches>100</batches>
|
||||
<inactive>10</inactive>
|
||||
<particles>1000</particles>
|
||||
|
||||
<!-- The starting source is a uniform distribution over the entire pin
|
||||
cell. Note that since this is effectively a 2D model, the z coordinates
|
||||
are inconsequential -->
|
||||
<source>
|
||||
<space type="box">
|
||||
<parameters>
|
||||
-0.62992 -0.62992 -1.
|
||||
0.62992 0.62992 1.
|
||||
</parameters>
|
||||
</space>
|
||||
</source>
|
||||
|
||||
<!-- To assess convergence of the source distribution, we need to define the
|
||||
bounds for a mesh over which the Shannon entropy should be
|
||||
calculated. The extent in the z direction is made arbitrarily large. -->
|
||||
<mesh id="1">
|
||||
<lower_left>-0.39218 -0.39218 -1.e50</lower_left>
|
||||
<upper_right>0.39218 0.39218 1.e50</upper_right>
|
||||
<dimension>10 10 1</dimension>
|
||||
</mesh>
|
||||
<entropy_mesh>1</entropy_mesh>
|
||||
|
||||
</settings>
|
||||
|
|
@ -1,23 +0,0 @@
|
|||
<?xml version="1.0"?>
|
||||
<tallies>
|
||||
|
||||
<mesh id="2" type="regular">
|
||||
<dimension>100 100 1</dimension>
|
||||
<lower_left>-0.62992 -0.62992 -1.e50</lower_left>
|
||||
<upper_right>0.62992 0.62992 1.e50</upper_right>
|
||||
</mesh>
|
||||
|
||||
<filter id="1" type="mesh">
|
||||
<bins>2</bins>
|
||||
</filter>
|
||||
|
||||
<filter id="2" type="energy">
|
||||
<bins>0. 4. 20.0e6</bins>
|
||||
</filter>
|
||||
|
||||
<tally id="1">
|
||||
<filters>1 2</filters>
|
||||
<scores>flux fission nu-fission</scores>
|
||||
</tally>
|
||||
|
||||
</tallies>
|
||||
|
|
@ -1,10 +0,0 @@
|
|||
<?xml version='1.0' encoding='utf-8'?>
|
||||
<geometry>
|
||||
<cell id="1" material="1" name="cell 1" region="-1" universe="0" />
|
||||
<cell id="2" material="2" name="cell 2" region="1 4 -5 6 -7" universe="0" />
|
||||
<surface coeffs="0 0 0.54" id="1" name="Fuel OR" type="z-cylinder" />
|
||||
<surface boundary="reflective" coeffs="-0.63" id="4" name="left" type="x-plane" />
|
||||
<surface boundary="reflective" coeffs="0.63" id="5" name="right" type="x-plane" />
|
||||
<surface boundary="reflective" coeffs="-0.63" id="6" name="bottom" type="y-plane" />
|
||||
<surface boundary="reflective" coeffs="0.63" id="7" name="top" type="y-plane" />
|
||||
</geometry>
|
||||
|
|
@ -1,12 +0,0 @@
|
|||
<?xml version='1.0' encoding='utf-8'?>
|
||||
<materials>
|
||||
<cross_sections>./mgxs.h5</cross_sections>
|
||||
<material id="1" name="UO2 fuel">
|
||||
<density units="macro" value="1.0" />
|
||||
<macroscopic name="UO2" />
|
||||
</material>
|
||||
<material id="2" name="Water">
|
||||
<density units="macro" value="1.0" />
|
||||
<macroscopic name="LWTR" />
|
||||
</material>
|
||||
</materials>
|
||||
Binary file not shown.
|
|
@ -1,28 +0,0 @@
|
|||
<?xml version="1.0"?>
|
||||
<plots>
|
||||
|
||||
<plot>
|
||||
<id>1</id>
|
||||
<filename>mat</filename>
|
||||
<color_by>material</color_by>
|
||||
<origin>0 0 0</origin>
|
||||
<width>1.26 1.26</width>
|
||||
<type>slice</type>
|
||||
<pixels>1000 1000 </pixels>
|
||||
<color id="1" rgb="255 0 0" />
|
||||
<color id="2" rgb="0 0 0" />
|
||||
<color id="3" rgb="0 255 0" />
|
||||
<color id="4" rgb="0 0 255" />
|
||||
</plot>
|
||||
|
||||
<plot>
|
||||
<id>2</id>
|
||||
<filename>cell</filename>
|
||||
<color_by>cell</color_by>
|
||||
<origin>0 0 0</origin>
|
||||
<width>1.26 1.26</width>
|
||||
<type>slice</type>
|
||||
<pixels>1000 1000 </pixels>
|
||||
</plot>
|
||||
|
||||
</plots>
|
||||
|
|
@ -1,13 +0,0 @@
|
|||
<?xml version='1.0' encoding='utf-8'?>
|
||||
<settings>
|
||||
<run_mode>eigenvalue</run_mode>
|
||||
<particles>1000</particles>
|
||||
<batches>100</batches>
|
||||
<inactive>10</inactive>
|
||||
<source strength="1.0">
|
||||
<space type="box">
|
||||
<parameters>-0.63 -0.63 -1 0.63 0.63 1</parameters>
|
||||
</space>
|
||||
</source>
|
||||
<energy_mode>multi-group</energy_mode>
|
||||
</settings>
|
||||
|
|
@ -1,18 +0,0 @@
|
|||
<?xml version='1.0' encoding='utf-8'?>
|
||||
<tallies>
|
||||
<mesh id="1" type="regular">
|
||||
<dimension>100 100 1</dimension>
|
||||
<lower_left>-0.63 -0.63 -1e+50</lower_left>
|
||||
<upper_right>0.63 0.63 1e+50</upper_right>
|
||||
</mesh>
|
||||
<filter id="1" type="energy">
|
||||
<bins>1e-05 0.0635 10.0 100.0 1000.0 500000.0 1000000.0 20000000.0</bins>
|
||||
</filter>
|
||||
<filter id="2" type="mesh">
|
||||
<bins>1</bins>
|
||||
</filter>
|
||||
<tally id="1" name="tally 1">
|
||||
<filters>1 2</filters>
|
||||
<scores>flux fission nu-fission</scores>
|
||||
</tally>
|
||||
</tallies>
|
||||
|
|
@ -1,19 +0,0 @@
|
|||
<?xml version="1.0"?>
|
||||
<geometry>
|
||||
|
||||
<!-- Definition of Cells -->
|
||||
<cell id="1">
|
||||
<universe>0</universe>
|
||||
<material>1</material>
|
||||
<region>1 -2 3 -4 5 -6</region>
|
||||
</cell>
|
||||
|
||||
<!-- Defition of Surfaces -->
|
||||
<surface id="1" type="x-plane" coeffs="-1" boundary="vacuum" />
|
||||
<surface id="2" type="x-plane" coeffs="1" boundary="vacuum" />
|
||||
<surface id="3" type="y-plane" coeffs="-1" boundary="reflective" />
|
||||
<surface id="4" type="y-plane" coeffs="1" boundary="reflective" />
|
||||
<surface id="5" type="z-plane" coeffs="-1" boundary="reflective" />
|
||||
<surface id="6" type="z-plane" coeffs="1" boundary="reflective" />
|
||||
|
||||
</geometry>
|
||||
|
|
@ -1,9 +0,0 @@
|
|||
<?xml version="1.0"?>
|
||||
<materials>
|
||||
|
||||
<material id="1">
|
||||
<density value="4.5" units="g/cc" />
|
||||
<nuclide name="U235" ao="1.0" />
|
||||
</material>
|
||||
|
||||
</materials>
|
||||
|
|
@ -1,17 +0,0 @@
|
|||
<?xml version="1.0"?>
|
||||
<settings>
|
||||
|
||||
<!-- Parameters for k-eigenvalue calculation -->
|
||||
<run_mode>eigenvalue</run_mode>
|
||||
<batches>500</batches>
|
||||
<inactive>10</inactive>
|
||||
<particles>10000</particles>
|
||||
|
||||
<!-- Starting source -->
|
||||
<source>
|
||||
<space type="box">
|
||||
<parameters>-1 -1 -1 1 1 1</parameters>
|
||||
</space>
|
||||
</source>
|
||||
|
||||
</settings>
|
||||
|
|
@ -612,10 +612,6 @@ class Settings(object):
|
|||
@entropy_mesh.setter
|
||||
def entropy_mesh(self, entropy):
|
||||
cv.check_type('entropy mesh', entropy, RegularMesh)
|
||||
if entropy.dimension:
|
||||
cv.check_length('entropy mesh dimension', entropy.dimension, 3)
|
||||
cv.check_length('entropy mesh lower-left corner', entropy.lower_left, 3)
|
||||
cv.check_length('entropy mesh upper-right corner', entropy.upper_right, 3)
|
||||
self._entropy_mesh = entropy
|
||||
|
||||
@trigger_active.setter
|
||||
|
|
|
|||
Loading…
Add table
Add a link
Reference in a new issue