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Add random_ray_pincell() example model (#3735)
Co-authored-by: John Tramm <john.tramm@gmail.com>
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1 changed files with 151 additions and 34 deletions
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@ -4,7 +4,7 @@ import numpy as np
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import openmc
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PINCELL_PITCH = 1.26 # cm
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def pwr_pin_cell() -> openmc.Model:
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"""Create a PWR pin-cell model.
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@ -51,7 +51,7 @@ def pwr_pin_cell() -> openmc.Model:
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model.materials = (fuel, clad, hot_water)
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# Instantiate ZCylinder surfaces
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pitch = 1.26
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pitch = PINCELL_PITCH
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fuel_or = openmc.ZCylinder(x0=0, y0=0, r=0.39218, name='Fuel OR')
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clad_or = openmc.ZCylinder(x0=0, y0=0, r=0.45720, name='Clad OR')
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left = openmc.XPlane(x0=-pitch/2, name='left', boundary_type='reflective')
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@ -319,14 +319,14 @@ def pwr_core() -> openmc.Model:
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l100 = openmc.RectLattice(
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name='Fuel assembly (lower half)', lattice_id=100)
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l100.lower_left = (-10.71, -10.71)
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l100.pitch = (1.26, 1.26)
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l100.pitch = (PINCELL_PITCH, PINCELL_PITCH)
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l100.universes = np.tile(fuel_cold, (17, 17))
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l100.universes[tube_x, tube_y] = tube_cold
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l101 = openmc.RectLattice(
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name='Fuel assembly (upper half)', lattice_id=101)
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l101.lower_left = (-10.71, -10.71)
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l101.pitch = (1.26, 1.26)
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l101.pitch = (PINCELL_PITCH, PINCELL_PITCH)
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l101.universes = np.tile(fuel_hot, (17, 17))
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l101.universes[tube_x, tube_y] = tube_hot
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@ -350,7 +350,7 @@ def pwr_core() -> openmc.Model:
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# Define core lattices
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l200 = openmc.RectLattice(name='Core lattice (lower half)', lattice_id=200)
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l200.lower_left = (-224.91, -224.91)
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l200.pitch = (21.42, 21.42)
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l200.pitch = (17 * PINCELL_PITCH, 17 * PINCELL_PITCH)
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l200.universes = [
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[fa_cw]*21,
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[fa_cw]*21,
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@ -376,7 +376,7 @@ def pwr_core() -> openmc.Model:
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l201 = openmc.RectLattice(name='Core lattice (lower half)', lattice_id=201)
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l201.lower_left = (-224.91, -224.91)
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l201.pitch = (21.42, 21.42)
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l201.pitch = (17 * PINCELL_PITCH, 17 * PINCELL_PITCH)
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l201.universes = [
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[fa_hw]*21,
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[fa_hw]*21,
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@ -488,7 +488,7 @@ def pwr_assembly() -> openmc.Model:
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clad_or = openmc.ZCylinder(x0=0, y0=0, r=0.45720, name='Clad OR')
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# Create boundary planes to surround the geometry
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pitch = 21.42
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pitch = 17 * PINCELL_PITCH
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min_x = openmc.XPlane(x0=-pitch/2, boundary_type='reflective')
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max_x = openmc.XPlane(x0=+pitch/2, boundary_type='reflective')
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min_y = openmc.YPlane(y0=-pitch/2, boundary_type='reflective')
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@ -514,7 +514,7 @@ def pwr_assembly() -> openmc.Model:
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# Create fuel assembly Lattice
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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.pitch = (PINCELL_PITCH, PINCELL_PITCH)
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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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@ -656,24 +656,21 @@ def slab_mg(num_regions=1, mat_names=None, mgxslib_name='2g.h5') -> openmc.Model
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return model
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def random_ray_lattice() -> openmc.Model:
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"""Create a 2x2 PWR pincell asymmetrical lattic eexample.
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This model is a 2x2 reflective lattice of fuel pins with one of the lattice
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locations having just moderator instead of a fuel pin. It uses 7 group
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cross section data.
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def _generate_c5g7_materials() -> openmc.Materials:
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"""Generate materials utilizing multi-group cross sections based on the
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the C5G7 Benchmark.
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Returns
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-------
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model : openmc.Model
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A PWR 2x2 lattice model
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materials : openmc.Materials
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Materials object containing UO2 and water materials.
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Data Sources
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------------
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All cross section data are from:
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Lewis et al., "Benchmark specification for determinisitc 2D/3D MOX fuel
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assembly transport calculations without spatial homogenization"
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"""
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model = openmc.Model()
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###########################################################################
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# Create MGXS data for the problem
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# Instantiate the energy group data
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group_edges = [1e-5, 0.0635, 10.0, 1.0e2, 1.0e3, 0.5e6, 1.0e6, 20.0e6]
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groups = openmc.mgxs.EnergyGroups(group_edges)
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@ -704,9 +701,10 @@ def random_ray_lattice() -> openmc.Model:
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uo2_xsdata.set_fission([7.21206e-03, 8.19301e-04, 6.45320e-03,
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1.85648e-02, 1.78084e-02, 8.30348e-02,
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2.16004e-01])
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uo2_xsdata.set_nu_fission([2.005998e-02, 2.027303e-03, 1.570599e-02,
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4.518301e-02, 4.334208e-02, 2.020901e-01,
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5.257105e-01])
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nu_fission = np.array([2.005998e-02, 2.027303e-03, 1.570599e-02,
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4.518301e-02, 4.334208e-02, 2.020901e-01,
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5.257105e-01])
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uo2_xsdata.set_nu_fission(nu_fission)
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uo2_xsdata.set_chi([5.8791e-01, 4.1176e-01, 3.3906e-04, 1.1761e-07, 0.0000e+00,
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0.0000e+00, 0.0000e+00])
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@ -752,14 +750,28 @@ def random_ray_lattice() -> openmc.Model:
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# Instantiate a Materials collection and export to XML
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materials = openmc.Materials([uo2, water])
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materials.cross_sections = "mgxs.h5"
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return materials
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###########################################################################
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# Define problem geometry
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def _generate_subdivided_pin_cell(uo2, water) -> openmc.Universe:
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"""Create a radially and azimuthally subdivided pin cell universe. Helper
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function for random_ray_pin_cell() and random_ray_lattice()
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Parameters
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----------
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uo2 : openmc.Material
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UO2 material
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water : openmc.Material
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Water material
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Returns
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-------
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pincell : openmc.Universe
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Universe containing an unbounded pin cell
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"""
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########################################
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# Define an unbounded pincell universe
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pitch = 1.26
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# Define an unbounded pin cell universe
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# Create a surface for the fuel outer radius
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fuel_or = openmc.ZCylinder(r=0.54, name='Fuel OR')
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@ -781,7 +793,7 @@ def random_ray_lattice() -> openmc.Model:
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moderator_c = openmc.Cell(
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fill=water, region=+outer_ring_b, name='moderator outer c')
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# Create pincell universe
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# Create pin cell universe
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pincell_base = openmc.Universe()
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# Register Cells with Universe
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@ -801,19 +813,125 @@ def random_ray_lattice() -> openmc.Model:
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for i in range(8):
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azimuthal_cell = openmc.Cell(name=f'azimuthal_cell_{i}')
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azimuthal_cell.fill = pincell_base
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azimuthal_cell.region = +azimuthal_planes[i] & -azimuthal_planes[(i+1) % 8]
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azimuthal_cell.region = + \
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azimuthal_planes[i] & -azimuthal_planes[(i+1) % 8]
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azimuthal_cells.append(azimuthal_cell)
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# Create a geometry with the azimuthal universes
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pincell = openmc.Universe(cells=azimuthal_cells, name='pincell')
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return pincell
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def random_ray_pin_cell() -> openmc.Model:
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"""Create a PWR pin cell example using C5G7 cross section data.
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cross section data.
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Returns
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-------
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model : openmc.Model
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A PWR pin cell model
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"""
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model = openmc.Model()
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###########################################################################
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# Create Materials for the problem
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materials = _generate_c5g7_materials()
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uo2 = materials[0]
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water = materials[1]
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###########################################################################
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# Define problem geometry
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pincell = _generate_subdivided_pin_cell(uo2, water)
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########################################
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# Define cell containing lattice and other stuff
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pitch = PINCELL_PITCH
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box = openmc.model.RectangularPrism(pitch, pitch, boundary_type='reflective')
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pincell = openmc.Cell(fill=pincell, region=-box, name='pincell')
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# Create a geometry with the top-level cell
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geometry = openmc.Geometry([pincell])
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###########################################################################
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# Define problem settings
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# Instantiate a Settings object, set all runtime parameters, and export to XML
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settings = openmc.Settings()
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settings.energy_mode = "multi-group"
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settings.batches = 400
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settings.inactive = 200
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settings.particles = 100
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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)
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rr_source = openmc.IndependentSource(space=uniform_dist)
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settings.random_ray['distance_active'] = 100.0
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settings.random_ray['distance_inactive'] = 20.0
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settings.random_ray['ray_source'] = rr_source
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settings.random_ray['volume_normalized_flux_tallies'] = True
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###########################################################################
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# Define tallies
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# Now use the mesh filter in a tally and indicate what scores are desired
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tally = openmc.Tally(name="Pin tally")
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tally.scores = ['flux', 'fission', 'nu-fission']
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tally.estimator = 'analog'
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# Instantiate a Tallies collection and export to XML
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tallies = openmc.Tallies([tally])
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###########################################################################
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# Exporting to OpenMC model
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###########################################################################
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model.geometry = geometry
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model.materials = materials
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model.settings = settings
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model.tallies = tallies
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return model
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def random_ray_lattice() -> openmc.Model:
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"""Create a 2x2 PWR pin cell asymmetrical lattice example.
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This model is a 2x2 reflective lattice of fuel pins with one of the lattice
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locations having just moderator instead of a fuel pin. It uses C5G7
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cross section data.
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Returns
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-------
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model : openmc.Model
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A PWR 2x2 lattice model
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"""
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model = openmc.Model()
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###########################################################################
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# Create Materials for the problem
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materials = _generate_c5g7_materials()
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uo2 = materials[0]
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water = materials[1]
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###########################################################################
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# Define problem geometry
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pincell = _generate_subdivided_pin_cell(uo2, water)
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########################################
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# Define a moderator lattice universe
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moderator_infinite = openmc.Cell(fill=water, name='moderator infinite')
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moderator_infinite = openmc.Cell(name='moderator infinite')
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moderator_infinite.fill = water
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mu = openmc.Universe()
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mu.add_cells([moderator_infinite])
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pitch = PINCELL_PITCH
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lattice = openmc.RectLattice()
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lattice.lower_left = [-pitch/2.0, -pitch/2.0]
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lattice.pitch = [pitch/10.0, pitch/10.0]
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@ -837,8 +955,7 @@ def random_ray_lattice() -> openmc.Model:
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########################################
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# Define cell containing lattice and other stuff
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box = openmc.model.RectangularPrism(
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pitch*2, pitch*2, boundary_type='reflective')
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box = openmc.model.RectangularPrism(pitch*2, pitch*2, boundary_type='reflective')
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assembly = openmc.Cell(fill=lattice2x2, region=-box, name='assembly')
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