Updated depletion scripts to reflect latest OpenMC PyAPI and OpenDeplete
This commit is contained in:
Will Boyd
parent
3a30feea76
commit
377ceb928c
14 changed files with 131 additions and 1431 deletions
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@ -1,4 +1,4 @@
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from collections import OrderedDict, defaultdict
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from collections import OrderedDict
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import copy
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import os
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@ -9,56 +9,7 @@ import opendeplete
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from geometry import beavrs, openmc_geometry
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#### Create "dummy" inputs to export distribcell paths for burnable cells
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# Create OpenMC "materials.xml" file
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beavrs.write_openmc_materials()
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# Create OpenMC "geometry.xml" file
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openmc_geometry.export_to_xml()
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# Construct uniform initial source distribution over fissionable zones
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lower_left = [-21.41728, -21.41728, +192.5]
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upper_right = [+21.41728, +21.41728, +197.5]
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source = openmc.source.Source(space=openmc.stats.Box(lower_left, upper_right))
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source.space.only_fissionable = True
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# Create OpenMC "settings.xml" file
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settings_file = openmc.Settings()
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settings_file.batches = 2
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settings_file.inactive = 1
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settings_file.particles = 10
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settings_file.output = {'tallies': False}
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settings_file.source = source
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settings_file.sourcepoint_write = False
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settings_file.export_to_xml()
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# Create OpenMC "tallies.xml" file
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tallies = openmc.Tallies()
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fuel_cells = openmc_geometry.get_cells_by_name(
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name='enr radial 0: Fuel', case_sensitive=True)
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# Instantiate a "dummy" distribcell tally for each cell we wish to deplete
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for cell in fuel_cells:
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tally = openmc.Tally(name='dummy distribcell tally')
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distribcell_filter = openmc.DistribcellFilter([cell.id])
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tally.filters = [distribcell_filter]
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tally.scores = ['fission']
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tallies.append(tally)
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tallies.export_to_xml()
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# Run OpenMC to generate summary.h5 file
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openmc.run()
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# Open "summary.h5" file
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su = openmc.Summary('summary.h5')
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fuel_cells = su.openmc_geometry.get_cells_by_name(
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name='enr radial 0: Fuel', case_sensitive=True)
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#### Setup OpenDeplete Materials wrapper
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# Setup OpenDeplete Materials wrapper
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materials = opendeplete.Materials()
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materials.temperature = OrderedDict()
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materials.sab = OrderedDict()
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@ -66,10 +17,17 @@ materials.initial_density = OrderedDict()
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materials.burn = OrderedDict()
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materials.cross_sections = os.environ["OPENMC_CROSS_SECTIONS"]
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# Count the number of instances for each cell and material
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openmc_geometry.determine_paths()
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# Extract all cells filled by a fuel material
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fuel_cells = openmc_geometry.get_cells_by_name(
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name='enr radial 0: Fuel', case_sensitive=True)
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# Extract cell materials, temperatures and sab
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for cell in su.openmc_geometry.get_all_material_cells():
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for cell in openmc_geometry.get_all_material_cells().values():
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materials.burn[cell.name] = 'fuel' in cell.fill.name.lower()
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materials.temperature[cell.name] = cell.temperature[0]
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materials.temperature[cell.name] = 300
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if len(cell.fill._sab) > 0:
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materials.sab[cell.name] = cell.fill._sab[0]
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@ -84,9 +42,8 @@ for cell in fuel_cells:
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densities[nuclide][1] * 1e24
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# Determine the maximum material ID
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all_mats = su.openmc_geometry.get_all_materials()
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max_material_id = 0
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for material in all_mats:
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for material in openmc_geometry.get_all_materials().values():
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max_material_id = max(max_material_id, material.id)
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# FIXME: Automatically extract info needed to calculate burnable cell volumes
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@ -94,48 +51,45 @@ for material in all_mats:
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radius = 0.39218
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height = 5.
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# Use defaultdict since OpenDeplete assumes volumes specified for all cells
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volumes = defaultdict(lambda: 1)
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# Assign distribmats for each material
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for cell in fuel_cells:
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new_materials = []
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num_instances = len(cell.distribcell_paths)
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for i in range(num_instances):
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for i in range(cell.num_instances):
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new_material = copy.deepcopy(cell.fill)
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new_material.id = max_material_id + 1
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max_material_id += 1
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new_materials.append(new_material)
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# Store volume of burnable fuel rods cells
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volumes[new_material.id] = np.pi * radius**2 * height
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new_material.volume = np.pi * radius**2 * height
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new_material.depletable = True
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new_material.temperature = 300
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cell.fill = new_materials
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# Create dt vector for 1 month with 15 day timesteps
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dt1 = 15*24*60*60 # 15 days
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# Create dt vector for 1 month with 5 day timesteps
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dt1 = 5*24*60*60 # 5 days
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dt2 = 1.*30*24*60*60 # 1 months
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N = np.floor(dt2/dt1)
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dt = np.repeat([dt1], N)
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# Create settings variable
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settings = opendeplete.Settings()
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settings = opendeplete.OpenMCSettings()
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settings.openmc_call = ["mpirun", "openmc"]
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settings.particles = 120000
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settings.batches = 20
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settings.inactive = 10
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settings.lower_left = lower_left
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settings.upper_right = upper_right
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settings.lower_left = [-21.41728, -21.41728, +192.5]
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settings.upper_right = [+21.41728, +21.41728, +197.5]
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settings.entropy_dimension = [17*2, 17*2, 1]
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settings.power = 2.337e15 * ((17.*17.*2.) / 1.5**2) # MeV/second cm from CASMO
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# MeV/second cm from CASMO
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settings.power = 2.337e15 * ((17.*17.*2.) / 1.5**2) * height
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settings.dt_vec = dt
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settings.output_dir = 'depleted'
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op = opendeplete.Operator()
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op.geometry_fill(su.openmc_geometry, volumes, materials, settings)
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op = opendeplete.OpenMCOperator(openmc_geometry, settings)
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# Perform simulation using the MCNPX/MCNP6 algorithm
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opendeplete.integrate(op, opendeplete.ce_cm_c1)
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@ -64,9 +64,8 @@ plot_file.export_to_xml()
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#### Create OpenMC MGXS libraries
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# Get all cells filled with a "fuel" material
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mat_cells = openmc_geometry.get_all_material_cells()
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fuel_cells = []
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for cell in mat_cells:
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for cell in openmc_geometry.get_all_material_cells().values():
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if 'fuel' in cell.fill.name.lower():
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fuel_cells.append(cell)
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@ -1,287 +0,0 @@
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"""Creates a 2D 2x2 assembly colorset with periodic BCs."""
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import numpy as np
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import opencg
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import openmc
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import openmc.opencg_compatible as opencg_compatible
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from beavrs.builder import BEAVRS
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def find_assembly(assembly_name, wrap_geometry=True):
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"""Find a fuel assembly with some string name in the BEAVRS OpenCG model.
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This method extracts the fuel assembly and wraps it in an OpenCG Geometry.
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The returned geometry has reflective boundary conditions along all
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boundaries. The z-axis is bounded between z=200 and z=210 cm.
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Parameters
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----------
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assembly_name : str
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The name of the fuel assembly lattice
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wrap_geometry : bool
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If false, the fuel assembly Lattice is returned. If true, the fuel
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assembly Lattice is wrapped in an OpenCG Geometry and returned (default).
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Returns
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-------
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fuel_assembly
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The OpenCG Lattice or Geometry for the assembly or None if not found
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"""
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# Get all OpenCG Universes
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all_univ = beavrs.main_universe.get_all_universes()
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# Iterate over all Universes
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fuel_assembly = None
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for univ_id, univ in all_univ.items():
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if univ.name == assembly_name:
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fuel_assembly = univ
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# Wrap lattice in a Geometry if requested by the user
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if wrap_geometry:
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# Create a root Cell
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root_cell = opencg.Cell(name='root cell')
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root_cell.fill = fuel_assembly
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# Make mixed reflective / vacuum boundaries
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min_x = opencg.XPlane(x0=root_cell.fill.min_x, boundary='reflective')
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max_x = opencg.XPlane(x0=root_cell.fill.max_x, boundary='reflective')
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min_y = opencg.YPlane(y0=root_cell.fill.min_y, boundary='reflective')
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max_y = opencg.YPlane(y0=root_cell.fill.max_y, boundary='reflective')
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max_z = opencg.ZPlane(z0=197.5, boundary='reflective')
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min_z = opencg.ZPlane(z0=192.5, boundary='reflective')
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# Add boundaries to the root Cell
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root_cell.add_surface(surface=min_x, halfspace=+1)
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root_cell.add_surface(surface=max_x, halfspace=-1)
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root_cell.add_surface(surface=min_y, halfspace=+1)
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root_cell.add_surface(surface=max_y, halfspace=-1)
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root_cell.add_surface(surface=min_z, halfspace=+1)
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root_cell.add_surface(surface=max_z, halfspace=-1)
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# Create a root Universe
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root_univ = opencg.Universe(universe_id=0, name='root universe')
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root_univ.add_cell(root_cell)
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# Create a Geometry
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fuel_assembly = opencg.Geometry()
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fuel_assembly.root_universe = root_univ
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return fuel_assembly
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def build_two_by_two(assembly1_name, assembly2_name):
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"""Build a 2x2 fuel assembly geometry.
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This routine puts reflective boundary conditions along all boundaries.
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Parameters
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----------
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assembly1_name : str
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The BEAVRS fuel assembly to place in the bottom right and top left
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assembly2_name : str
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The BEAVRS fuel assembly to place in the bottom left and top right
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Returns
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-------
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opencg.Geometry
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A 2x2 fuel assembly OpenCG Geometry
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"""
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fuel_assembly1 = find_assembly(assembly1_name, wrap_geometry=False)
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fuel_assembly2 = find_assembly(assembly2_name, wrap_geometry=False)
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# Find the water material
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all_cells = beavrs.main_universe.get_all_cells()
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for cell_uuid, cell in all_cells.items():
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if cell.type == 'material' and cell.fill.name == 'water':
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water = cell.fill
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# Create a Cell/Universe around the first fuel assembly
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fuel_cell1 = opencg.Cell(name='assm1 cell')
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fuel_cell1.fill = fuel_assembly1
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fuel_univ1 = opencg.Universe(name='assm1 universe')
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fuel_univ1.add_cell(fuel_cell1)
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# Create a Cell/Universe around the second fuel assembly
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fuel_cell2 = opencg.Cell(name='assm2 cell')
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fuel_cell2.fill = fuel_assembly2
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fuel_univ2 = opencg.Universe(name='assm2 universe')
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fuel_univ2.add_cell(fuel_cell2)
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# Create a 3x3 lattice two fuel assemblies surrounded by a water reflector
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two_by_two_lattice = opencg.Lattice(name='reflector')
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lat_width = fuel_assembly1.max_x - fuel_assembly1.min_x
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two_by_two_lattice.width = [lat_width, lat_width, 1000.]
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two_by_two_lattice.offset = [0., 0., 0.]
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two_by_two_lattice.dimension = [2, 2, 1]
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two_by_two_lattice.universes = [[fuel_univ1, fuel_univ2],
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[fuel_univ2, fuel_univ1]]
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# Create a Geometry around the reflected lattice
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root_cell = opencg.Cell(name='root cell')
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root_cell.fill = two_by_two_lattice
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# Make mixed reflective / vacuum boundaries
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min_x = opencg.XPlane(x0=root_cell.fill.min_x, boundary='periodic')
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max_x = opencg.XPlane(x0=root_cell.fill.max_x, boundary='periodic')
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min_y = opencg.YPlane(y0=root_cell.fill.min_y, boundary='periodic')
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max_y = opencg.YPlane(y0=root_cell.fill.max_y, boundary='periodic')
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min_z = opencg.ZPlane(z0=192.5, boundary='reflective')
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max_z = opencg.ZPlane(z0=197.5, boundary='reflective')
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# Add boundaries to the root Cell
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root_cell.add_surface(surface=min_x, halfspace=+1)
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root_cell.add_surface(surface=max_x, halfspace=-1)
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root_cell.add_surface(surface=min_y, halfspace=+1)
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root_cell.add_surface(surface=max_y, halfspace=-1)
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root_cell.add_surface(surface=min_z, halfspace=+1)
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root_cell.add_surface(surface=max_z, halfspace=-1)
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# Create a root Universe for this fuel assembly
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root_univ = opencg.Universe(universe_id=0, name='root universe')
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root_univ.add_cell(root_cell)
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# Create an OpenCG Geometry for this fuel assembly
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two_by_two = opencg.Geometry()
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two_by_two.root_universe = root_univ
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return two_by_two
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#### Create OpenMC "materials.xml" and "geometry.xml" files
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# Instantiate a BEAVRS object
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beavrs = BEAVRS(nndc_xs=True)
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# Write all BEAVRS materials to materials.xml file
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beavrs.write_openmc_materials()
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# Extract fuel assemblies of interest from BEAVRS model
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two_by_two = build_two_by_two('Fuel 1.6% enr instr no BAs',
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'Fuel 3.1% enr instr 20')
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openmc_geometry = opencg_compatible.get_openmc_geometry(two_by_two)
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openmc_geometry.export_to_xml()
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#### Create OpenMC "settings.xml" file
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# Query the user on whether to use multipole cross sections
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multipole = input('Use multipole cross sections? (y/n): ').lower()
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multipole = True if multipole == 'y' else False
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# Construct uniform initial source distribution over fissionable zones
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lower_left = two_by_two.bounds[:3]
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upper_right = two_by_two.bounds[3:]
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source = openmc.source.Source(space=openmc.stats.Box(lower_left, upper_right))
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source.space.only_fissionable = True
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settings_file = openmc.Settings()
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settings_file.batches = 10
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settings_file.inactive = 5
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settings_file.particles = 10000
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settings_file.ptables = True
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settings_file.output = {'tallies': False}
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settings_file.source = source
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settings_file.sourcepoint_write = False
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if multipole:
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settings_file.temperature = {'multipole': True, 'tolerance': 1000}
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settings_file.export_to_xml()
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#### Create OpenMC "plots.xml" file
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# Initialize the BEAVRS color mapping scheme
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beavrs.write_openmc_plots()
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# Create a plot colored by materials
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plot = openmc.Plot()
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bounds = two_by_two.bounds
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plot.width = [two_by_two.max_x - two_by_two.min_x,
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two_by_two.max_y - two_by_two.min_y]
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plot.origin = [bounds[0] + (bounds[3] - bounds[0]) / 2.,
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bounds[1] + (bounds[4] - bounds[1]) / 2.,
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bounds[2] + (bounds[5] - bounds[2]) / 2.]
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plot.color = 'mat'
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plot.filename = '2x2-periodic'
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plot.col_spec = beavrs.plots.colspec_mat
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plot.pixels = [1000, 1000]
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plot_file = openmc.Plots([plot])
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plot_file.export_to_xml()
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#### Create OpenMC MGXS libraries
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# Get all cells filled with a "fuel" material
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mat_cells = openmc_geometry.get_all_material_cells()
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fuel_cells = []
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for cell in mat_cells:
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if 'fuel' in cell.fill.name.lower():
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fuel_cells.append(cell)
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# CASMO 70-group structure
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energy_groups = openmc.mgxs.EnergyGroups()
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energy_groups.group_edges = np.array([
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0, 0.005, 0.01, 0.015, 0.02, 0.025, 0.03, 0.035, 0.042, 0.05, 0.058, 0.067,
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0.08, 0.1, 0.14, 0.18, 0.22, 0.25, 0.28, 0.3, 0.32, 0.35, 0.4, 0.5, 0.625,
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0.78, 0.85, 0.91, 0.95, 0.972, 0.996, 1.02, 1.045, 1.071, 1.097, 1.123,
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1.15, 1.3, 1.5, 1.855, 2.1, 2.6, 3.3, 4., 9.877, 15.968, 27.7, 48.052,
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75.501, 148.73, 367.26001, 906.90002, 1.4251e3, 2.2395e3, 3.5191e3, 5.53e3,
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9.118e3, 15.03e3, 24.78e3, 40.85e3, 67.34e3, 111.e3, 183e3, 302.5e3, 500e3,
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821e3, 1.353e6, 2.231e6, 3.679e6, 6.0655e6, 2e7])
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# Initialize a 70-group "distribcell" MGXS library
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cell_mgxs_lib = openmc.mgxs.Library(openmc_geometry, by_nuclide=True)
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cell_mgxs_lib.energy_groups = energy_groups
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cell_mgxs_lib.mgxs_types = ['total', 'nu-fission', 'nu-scatter matrix', 'chi']
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cell_mgxs_lib.domain_type = 'distribcell'
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cell_mgxs_lib.domains = fuel_cells
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cell_mgxs_lib.correction = None
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cell_mgxs_lib.build_library()
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# Initialize a 70-group "material" MGXS library
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mat_mgxs_lib = openmc.mgxs.Library(openmc_geometry, by_nuclide=True)
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mat_mgxs_lib.energy_groups = energy_groups
|
||||
mat_mgxs_lib.mgxs_types = ['total', 'nu-fission', 'nu-scatter matrix', 'chi']
|
||||
mat_mgxs_lib.domain_type = 'material'
|
||||
mat_mgxs_lib.correction = None
|
||||
mat_mgxs_lib.build_library()
|
||||
|
||||
|
||||
#### Create mesh tallies for verification of pin-wise reaction rates
|
||||
|
||||
# Instantiate a tally Mesh
|
||||
mesh = openmc.Mesh(name='assembly mesh')
|
||||
mesh.type = 'regular'
|
||||
mesh.dimension = [34, 34, 1]
|
||||
mesh.lower_left = lower_left
|
||||
mesh.width = (np.array(upper_right) - np.array(lower_left))
|
||||
mesh.width[:2] /= 34
|
||||
mesh_filter = openmc.MeshFilter(mesh)
|
||||
|
||||
# Instantiate energy-integrated fission rate mesh Tally
|
||||
fission_rates = openmc.Tally(name='fission rates')
|
||||
fission_rates.filters = [mesh_filter]
|
||||
fission_rates.scores = ['fission']
|
||||
|
||||
# Instantiate energy-wise U-238 capture rate mesh Tally
|
||||
capture_rates = openmc.Tally(name='u-238 capture')
|
||||
capture_rates.filters = [mesh_filter]
|
||||
capture_rates.nuclides = ['U238']
|
||||
capture_rates.scores = ['absorption', 'fission']
|
||||
|
||||
|
||||
#### Create OpenMC "tallies.xml" file
|
||||
|
||||
# Create a "tallies.xml" file for the mesh tallies
|
||||
tallies_file = openmc.Tallies([fission_rates, capture_rates])
|
||||
cell_mgxs_lib.add_to_tallies_file(tallies_file, merge=True)
|
||||
mat_mgxs_lib.add_to_tallies_file(tallies_file, merge=True)
|
||||
tallies_file.export_to_xml()
|
||||
|
|
@ -1,4 +1,4 @@
|
|||
from collections import OrderedDict, defaultdict
|
||||
from collections import OrderedDict
|
||||
import copy
|
||||
import os
|
||||
|
||||
|
|
@ -9,56 +9,7 @@ import opendeplete
|
|||
from geometry import beavrs, openmc_geometry
|
||||
|
||||
|
||||
#### Create "dummy" inputs to export distribcell paths for burnable cells
|
||||
|
||||
# Create OpenMC "materials.xml" file
|
||||
beavrs.write_openmc_materials()
|
||||
|
||||
# Create OpenMC "geometry.xml" file
|
||||
openmc_geometry.export_to_xml()
|
||||
|
||||
# Construct uniform initial source distribution over fissionable zones
|
||||
lower_left = [-32.12592, -32.12592, 192.5]
|
||||
upper_right = [32.12592, 32.12592, 197.5]
|
||||
source = openmc.source.Source(space=openmc.stats.Box(lower_left, upper_right))
|
||||
source.space.only_fissionable = True
|
||||
|
||||
# Create OpenMC "settings.xml" file
|
||||
settings_file = openmc.Settings()
|
||||
settings_file.batches = 2
|
||||
settings_file.inactive = 1
|
||||
settings_file.particles = 10
|
||||
settings_file.output = {'tallies': False}
|
||||
settings_file.source = source
|
||||
settings_file.sourcepoint_write = False
|
||||
settings_file.export_to_xml()
|
||||
|
||||
# Create OpenMC "tallies.xml" file
|
||||
tallies = openmc.Tallies()
|
||||
fuel_cells = openmc_geometry.get_cells_by_name(
|
||||
name='enr radial 0: Fuel', case_sensitive=True)
|
||||
|
||||
# Instantiate a "dummy" distribcell tally for each cell we wish to deplete
|
||||
for cell in fuel_cells:
|
||||
tally = openmc.Tally(name='dummy distribcell tally')
|
||||
distribcell_filter = openmc.DistribcellFilter([cell.id])
|
||||
tally.filters = [distribcell_filter]
|
||||
tally.scores = ['fission']
|
||||
tallies.append(tally)
|
||||
|
||||
tallies.export_to_xml()
|
||||
|
||||
# Run OpenMC to generate summary.h5 file
|
||||
openmc.run()
|
||||
|
||||
# Open "summary.h5" file
|
||||
su = openmc.Summary('summary.h5')
|
||||
fuel_cells = su.openmc_geometry.get_cells_by_name(
|
||||
name='enr radial 0: Fuel', case_sensitive=True)
|
||||
|
||||
|
||||
#### Setup OpenDeplete Materials wrapper
|
||||
|
||||
# Setup OpenDeplete Materials wrapper
|
||||
materials = opendeplete.Materials()
|
||||
materials.temperature = OrderedDict()
|
||||
materials.sab = OrderedDict()
|
||||
|
|
@ -66,10 +17,17 @@ materials.initial_density = OrderedDict()
|
|||
materials.burn = OrderedDict()
|
||||
materials.cross_sections = os.environ["OPENMC_CROSS_SECTIONS"]
|
||||
|
||||
# Count the number of instances for each cell and material
|
||||
openmc_geometry.determine_paths()
|
||||
|
||||
# Extract all cells filled by a fuel material
|
||||
fuel_cells = openmc_geometry.get_cells_by_name(
|
||||
name='enr radial 0: Fuel', case_sensitive=True)
|
||||
|
||||
# Extract cell materials, temperatures and sab
|
||||
for cell in su.openmc_geometry.get_all_material_cells():
|
||||
for cell in openmc_geometry.get_all_material_cells().values():
|
||||
materials.burn[cell.name] = 'fuel' in cell.fill.name.lower()
|
||||
materials.temperature[cell.name] = cell.temperature[0]
|
||||
materials.temperature[cell.name] = 300
|
||||
if len(cell.fill._sab) > 0:
|
||||
materials.sab[cell.name] = cell.fill._sab[0]
|
||||
|
||||
|
|
@ -84,9 +42,8 @@ for cell in fuel_cells:
|
|||
densities[nuclide][1] * 1e24
|
||||
|
||||
# Determine the maximum material ID
|
||||
all_mats = su.openmc_geometry.get_all_materials()
|
||||
max_material_id = 0
|
||||
for material in all_mats:
|
||||
for material in openmc_geometry.get_all_materials().values():
|
||||
max_material_id = max(max_material_id, material.id)
|
||||
|
||||
# FIXME: Automatically extract info needed to calculate burnable cell volumes
|
||||
|
|
@ -94,48 +51,45 @@ for material in all_mats:
|
|||
radius = 0.39218
|
||||
height = 5.
|
||||
|
||||
# Use defaultdict since OpenDeplete assumes volumes specified for all cells
|
||||
volumes = defaultdict(lambda: 1)
|
||||
|
||||
# Assign distribmats for each material
|
||||
for cell in fuel_cells:
|
||||
new_materials = []
|
||||
num_instances = len(cell.distribcell_paths)
|
||||
|
||||
for i in range(num_instances):
|
||||
for i in range(cell.num_instances):
|
||||
new_material = copy.deepcopy(cell.fill)
|
||||
new_material.id = max_material_id + 1
|
||||
max_material_id += 1
|
||||
new_materials.append(new_material)
|
||||
|
||||
# Store volume of burnable fuel rods cells
|
||||
volumes[new_material.id] = np.pi * radius**2 * height
|
||||
new_material.volume = np.pi * radius**2 * height
|
||||
new_material.depletable = True
|
||||
new_material.temperature = 300
|
||||
|
||||
cell.fill = new_materials
|
||||
|
||||
# Create dt vector for 1 month with 15 day timesteps
|
||||
dt1 = 15*24*60*60 # 15 days
|
||||
# Create dt vector for 1 month with 5 day timesteps
|
||||
dt1 = 5*24*60*60 # 5 days
|
||||
dt2 = 1.*30*24*60*60 # 1 months
|
||||
N = np.floor(dt2/dt1)
|
||||
dt = np.repeat([dt1], N)
|
||||
|
||||
# Create settings variable
|
||||
settings = opendeplete.Settings()
|
||||
|
||||
settings = opendeplete.OpenMCSettings()
|
||||
settings.openmc_call = ["mpirun", "openmc"]
|
||||
settings.particles = 120000
|
||||
settings.batches = 30
|
||||
settings.inactive = 20
|
||||
settings.lower_left = lower_left
|
||||
settings.upper_right = upper_right
|
||||
settings.lower_left = [-32.12592, -32.12592, 192.5]
|
||||
settings.upper_right = [32.12592, 32.12592, 197.5]
|
||||
settings.entropy_dimension = [17*3, 17*3, 1]
|
||||
|
||||
settings.power = 2.337e15 * ((17.*17.*2.) / 1.5**2) # MeV/second cm from CASMO
|
||||
# MeV/second cm from CASMO
|
||||
settings.power = 2.337e15 * ((17.*17.*2.) / 1.5**2) * height
|
||||
settings.dt_vec = dt
|
||||
settings.output_dir = 'depleted'
|
||||
|
||||
op = opendeplete.Operator()
|
||||
op.geometry_fill(su.openmc_geometry, volumes, materials, settings)
|
||||
op = opendeplete.OpenMCOperator(openmc_geometry, settings)
|
||||
|
||||
# Perform simulation using the MCNPX/MCNP6 algorithm
|
||||
opendeplete.integrate(op, opendeplete.ce_cm_c1)
|
||||
|
|
|
|||
|
|
@ -67,9 +67,8 @@ plot_file.export_to_xml()
|
|||
#### Create OpenMC MGXS libraries
|
||||
|
||||
# Get all cells filled with a "fuel" material
|
||||
mat_cells = openmc_geometry.get_all_material_cells()
|
||||
fuel_cells = []
|
||||
for cell in mat_cells:
|
||||
for cell in openmc_geometry.get_all_material_cells().values():
|
||||
if 'fuel' in cell.fill.name.lower():
|
||||
fuel_cells.append(cell)
|
||||
|
||||
|
|
|
|||
|
|
@ -1,299 +0,0 @@
|
|||
"""Creates a 2D 2x2 assembly colorset with a water reflector."""
|
||||
|
||||
import numpy as np
|
||||
|
||||
import opencg
|
||||
import openmc
|
||||
import openmc.opencg_compatible as opencg_compatible
|
||||
from beavrs.builder import BEAVRS
|
||||
|
||||
|
||||
def find_assembly(assembly_name, wrap_geometry=True):
|
||||
"""Find a fuel assembly with some string name in the BEAVRS OpenCG model.
|
||||
|
||||
This method extracts the fuel assembly and wraps it in an OpenCG Geometry.
|
||||
The returned geometry has reflective boundary conditions along all
|
||||
boundaries. The z-axis is bounded between z=200 and z=210 cm.
|
||||
|
||||
Parameters
|
||||
----------
|
||||
assembly_name : str
|
||||
The name of the fuel assembly lattice
|
||||
wrap_geometry : bool
|
||||
If false, the fuel assembly Lattice is returned. If true, the fuel
|
||||
assembly Lattice is wrapped in an OpenCG Geometry and returned (default).
|
||||
|
||||
Returns
|
||||
-------
|
||||
fuel_assembly
|
||||
The OpenCG Lattice or Geometry for the assembly or None if not found
|
||||
|
||||
"""
|
||||
|
||||
# Get all OpenCG Universes
|
||||
all_univ = beavrs.main_universe.get_all_universes()
|
||||
|
||||
# Iterate over all Universes
|
||||
fuel_assembly = None
|
||||
for univ_id, univ in all_univ.items():
|
||||
if univ.name == assembly_name:
|
||||
fuel_assembly = univ
|
||||
|
||||
# Wrap lattice in a Geometry if requested by the user
|
||||
if wrap_geometry:
|
||||
|
||||
# Create a root Cell
|
||||
root_cell = opencg.Cell(name='root cell')
|
||||
root_cell.fill = fuel_assembly
|
||||
|
||||
# Make mixed reflective / vacuum boundaries
|
||||
min_x = opencg.XPlane(x0=root_cell.fill.min_x, boundary='reflective')
|
||||
max_x = opencg.XPlane(x0=root_cell.fill.max_x, boundary='reflective')
|
||||
min_y = opencg.YPlane(y0=root_cell.fill.min_y, boundary='reflective')
|
||||
max_y = opencg.YPlane(y0=root_cell.fill.max_y, boundary='reflective')
|
||||
max_z = opencg.ZPlane(z0=197.5, boundary='reflective')
|
||||
min_z = opencg.ZPlane(z0=192.5, boundary='reflective')
|
||||
|
||||
# Add boundaries to the root Cell
|
||||
root_cell.add_surface(surface=min_x, halfspace=+1)
|
||||
root_cell.add_surface(surface=max_x, halfspace=-1)
|
||||
root_cell.add_surface(surface=min_y, halfspace=+1)
|
||||
root_cell.add_surface(surface=max_y, halfspace=-1)
|
||||
root_cell.add_surface(surface=min_z, halfspace=+1)
|
||||
root_cell.add_surface(surface=max_z, halfspace=-1)
|
||||
|
||||
# Create a root Universe
|
||||
root_univ = opencg.Universe(universe_id=0, name='root universe')
|
||||
root_univ.add_cell(root_cell)
|
||||
|
||||
# Create a Geometry
|
||||
fuel_assembly = opencg.Geometry()
|
||||
fuel_assembly.root_universe = root_univ
|
||||
|
||||
return fuel_assembly
|
||||
|
||||
|
||||
def build_reflector(assembly1_name, assembly2_name):
|
||||
"""Build a 2x2 fuel assembly geometry with a water reflector on the
|
||||
bottom and right.
|
||||
|
||||
This routine puts reflective boundary conditions along min x, max y and z
|
||||
and vacuum boundary conditions along the max x and min y boundaries.
|
||||
|
||||
Parameters
|
||||
----------
|
||||
assembly1_name : str
|
||||
The BEAVRS fuel assembly to place in the bottom right and top left
|
||||
assembly2_name : str
|
||||
The BEAVRS fuel assembly to place in the bottom left and top right
|
||||
|
||||
Returns
|
||||
-------
|
||||
opencg.Geometry
|
||||
A 2x2 fuel assembly and reflector OpenCG Geometry
|
||||
|
||||
"""
|
||||
|
||||
fuel_assembly1 = find_assembly(assembly1_name, wrap_geometry=False)
|
||||
fuel_assembly2 = find_assembly(assembly2_name, wrap_geometry=False)
|
||||
|
||||
# Find the water material
|
||||
all_cells = beavrs.main_universe.get_all_cells()
|
||||
for cell_uuid, cell in all_cells.items():
|
||||
if cell.type == 'material' and cell.fill.name == 'water':
|
||||
water = cell.fill
|
||||
|
||||
# Create a Cell/Universe around the first fuel assembly
|
||||
fuel_cell1 = opencg.Cell(name='assm1 cell')
|
||||
fuel_cell1.fill = fuel_assembly1
|
||||
fuel_univ1 = opencg.Universe(name='assm1 universe')
|
||||
fuel_univ1.add_cell(fuel_cell1)
|
||||
|
||||
# Create a Cell/Universe around the second fuel assembly
|
||||
fuel_cell2 = opencg.Cell(name='assm2 cell')
|
||||
fuel_cell2.fill = fuel_assembly2
|
||||
fuel_univ2 = opencg.Universe(name='assm2 universe')
|
||||
fuel_univ2.add_cell(fuel_cell2)
|
||||
|
||||
# Create a Cell/Universe with water
|
||||
water_cell = opencg.Cell(name='water cell', fill=water)
|
||||
water_univ = opencg.Universe(name='water universe')
|
||||
water_univ.add_cell(water_cell)
|
||||
|
||||
# Create a 3x3 lattice two fuel assemblies surrounded by a water reflector
|
||||
reflector_lattice = opencg.Lattice(name='reflector')
|
||||
lat_width = fuel_assembly1.max_x - fuel_assembly1.min_x
|
||||
reflector_lattice.width = [lat_width, lat_width, 1000.]
|
||||
reflector_lattice.dimension = [3, 3, 1]
|
||||
reflector_lattice.offset = [0., 0., 0.]
|
||||
reflector_lattice.universes = [[fuel_univ1, fuel_univ2, water_univ],
|
||||
[fuel_univ2, fuel_univ1, water_univ],
|
||||
[water_univ, water_univ, water_univ]]
|
||||
|
||||
# Create a Geometry around the reflected lattice
|
||||
root_cell = opencg.Cell(name='root cell')
|
||||
root_cell.fill = reflector_lattice
|
||||
|
||||
# Make mixed reflective / vacuum boundaries
|
||||
min_x = opencg.XPlane(x0=root_cell.fill.min_x, boundary='reflective')
|
||||
max_x = opencg.XPlane(x0=root_cell.fill.max_x, boundary='vacuum')
|
||||
min_y = opencg.YPlane(y0=root_cell.fill.min_y, boundary='vacuum')
|
||||
max_y = opencg.YPlane(y0=root_cell.fill.max_y, boundary='reflective')
|
||||
min_z = opencg.ZPlane(z0=192.5, boundary='reflective')
|
||||
max_z = opencg.ZPlane(z0=197.5, boundary='reflective')
|
||||
|
||||
# Add boundaries to the root Cell
|
||||
root_cell.add_surface(surface=min_x, halfspace=+1)
|
||||
root_cell.add_surface(surface=max_x, halfspace=-1)
|
||||
root_cell.add_surface(surface=min_y, halfspace=+1)
|
||||
root_cell.add_surface(surface=max_y, halfspace=-1)
|
||||
root_cell.add_surface(surface=min_z, halfspace=+1)
|
||||
root_cell.add_surface(surface=max_z, halfspace=-1)
|
||||
|
||||
# Create a root Universe for this fuel assembly
|
||||
root_univ = opencg.Universe(universe_id=0, name='root universe')
|
||||
root_univ.add_cell(root_cell)
|
||||
|
||||
# Create an OpenCG Geometry for this fuel assembly
|
||||
reflector = opencg.Geometry()
|
||||
reflector.root_universe = root_univ
|
||||
|
||||
return reflector
|
||||
|
||||
|
||||
#### Create OpenMC "materials.xml" and "geometry.xml" files
|
||||
|
||||
# Instantiate a BEAVRS object
|
||||
beavrs = BEAVRS(nndc_xs=True)
|
||||
|
||||
# Write all BEAVRS materials to materials.xml file
|
||||
beavrs.write_openmc_materials()
|
||||
|
||||
# Extract fuel assemblies of interest from BEAVRS model
|
||||
reflector = build_reflector('Fuel 1.6% enr instr no BAs',
|
||||
'Fuel 3.1% enr instr 20')
|
||||
openmc_geometry = opencg_compatible.get_openmc_geometry(reflector)
|
||||
openmc_geometry.export_to_xml()
|
||||
|
||||
|
||||
#### Create OpenMC "settings.xml" file
|
||||
|
||||
# Query the user on whether to use multipole cross sections
|
||||
multipole = input('Use multipole cross sections? (y/n): ').lower()
|
||||
multipole = True if multipole == 'y' else False
|
||||
|
||||
# Construct uniform initial source distribution over fissionable zones
|
||||
lower_left = reflector.bounds[:3]
|
||||
upper_right = reflector.bounds[3:]
|
||||
|
||||
lat_width = (np.array(upper_right) - np.array(lower_left))
|
||||
lat_width[:2] /= 3.
|
||||
|
||||
source = openmc.source.Source(space=openmc.stats.Box(lower_left, upper_right))
|
||||
source.space.only_fissionable = True
|
||||
|
||||
settings_file = openmc.Settings()
|
||||
settings_file.batches = 10
|
||||
settings_file.inactive = 5
|
||||
settings_file.particles = 10000
|
||||
settings_file.ptables = True
|
||||
settings_file.output = {'tallies': False}
|
||||
settings_file.source = source
|
||||
settings_file.sourcepoint_write = False
|
||||
|
||||
if multipole:
|
||||
settings_file.temperature = {'multipole': True, 'tolerance': 1000}
|
||||
|
||||
settings_file.export_to_xml()
|
||||
|
||||
|
||||
#### Create OpenMC "plots.xml" file
|
||||
|
||||
# Initialize the BEAVRS color mapping scheme
|
||||
beavrs.write_openmc_plots()
|
||||
|
||||
# Create a plot colored by materials
|
||||
plot = openmc.Plot()
|
||||
bounds = reflector.bounds
|
||||
plot.width = [reflector.max_x - reflector.min_x,
|
||||
reflector.max_y - reflector.min_y]
|
||||
plot.origin = [bounds[0] + (bounds[3] - bounds[0]) / 2.,
|
||||
bounds[1] + (bounds[4] - bounds[1]) / 2.,
|
||||
bounds[2] + (bounds[5] - bounds[2]) / 2.]
|
||||
plot.color = 'mat'
|
||||
plot.filename = '2x2-reflector'
|
||||
plot.col_spec = beavrs.plots.colspec_mat
|
||||
plot.pixels = [1000, 1000]
|
||||
|
||||
plot_file = openmc.Plots([plot])
|
||||
plot_file.export_to_xml()
|
||||
|
||||
|
||||
#### Create OpenMC MGXS libraries
|
||||
|
||||
# Get all cells filled with a "fuel" material
|
||||
mat_cells = openmc_geometry.get_all_material_cells()
|
||||
fuel_cells = []
|
||||
for cell in mat_cells:
|
||||
if 'fuel' in cell.fill.name.lower():
|
||||
fuel_cells.append(cell)
|
||||
|
||||
# CASMO 70-group structure
|
||||
energy_groups = openmc.mgxs.EnergyGroups()
|
||||
energy_groups.group_edges = np.array([
|
||||
0, 0.005, 0.01, 0.015, 0.02, 0.025, 0.03, 0.035, 0.042, 0.05, 0.058, 0.067,
|
||||
0.08, 0.1, 0.14, 0.18, 0.22, 0.25, 0.28, 0.3, 0.32, 0.35, 0.4, 0.5, 0.625,
|
||||
0.78, 0.85, 0.91, 0.95, 0.972, 0.996, 1.02, 1.045, 1.071, 1.097, 1.123,
|
||||
1.15, 1.3, 1.5, 1.855, 2.1, 2.6, 3.3, 4., 9.877, 15.968, 27.7, 48.052,
|
||||
75.501, 148.73, 367.26001, 906.90002, 1.4251e3, 2.2395e3, 3.5191e3, 5.53e3,
|
||||
9.118e3, 15.03e3, 24.78e3, 40.85e3, 67.34e3, 111.e3, 183e3, 302.5e3, 500e3,
|
||||
821e3, 1.353e6, 2.231e6, 3.679e6, 6.0655e6, 2e7])
|
||||
|
||||
# Initialize a 70-group "distribcell" MGXS library
|
||||
cell_mgxs_lib = openmc.mgxs.Library(openmc_geometry, by_nuclide=True)
|
||||
cell_mgxs_lib.energy_groups = energy_groups
|
||||
cell_mgxs_lib.mgxs_types = ['total', 'nu-fission', 'nu-scatter matrix', 'chi']
|
||||
cell_mgxs_lib.domain_type = 'distribcell'
|
||||
cell_mgxs_lib.domains = fuel_cells
|
||||
cell_mgxs_lib.correction = None
|
||||
cell_mgxs_lib.build_library()
|
||||
|
||||
# Initialize a 70-group "material" MGXS library
|
||||
mat_mgxs_lib = openmc.mgxs.Library(openmc_geometry, by_nuclide=True)
|
||||
mat_mgxs_lib.energy_groups = energy_groups
|
||||
mat_mgxs_lib.mgxs_types = ['total', 'nu-fission', 'nu-scatter matrix', 'chi']
|
||||
mat_mgxs_lib.domain_type = 'material'
|
||||
mat_mgxs_lib.correction = None
|
||||
mat_mgxs_lib.build_library()
|
||||
|
||||
|
||||
#### Create mesh tallies for verification of pin-wise reaction rates
|
||||
|
||||
# Instantiate a tally Mesh
|
||||
mesh = openmc.Mesh(name='assembly mesh')
|
||||
mesh.type = 'regular'
|
||||
mesh.dimension = [34, 34, 1]
|
||||
mesh.lower_left = [lower_left[0], lower_left[1] + lat_width[1], lower_left[2]]
|
||||
mesh.width = np.array(lat_width)
|
||||
mesh.width[:2] /= 17.
|
||||
mesh_filter = openmc.MeshFilter(mesh)
|
||||
|
||||
# Instantiate energy-integrated fission rate mesh Tally
|
||||
fission_rates = openmc.Tally(name='fission rates')
|
||||
fission_rates.filters = [mesh_filter]
|
||||
fission_rates.scores = ['fission']
|
||||
|
||||
# Instantiate energy-wise U-238 capture rate mesh Tally
|
||||
capture_rates = openmc.Tally(name='u-238 capture')
|
||||
capture_rates.filters = [mesh_filter]
|
||||
capture_rates.nuclides = ['U238']
|
||||
capture_rates.scores = ['absorption', 'fission']
|
||||
|
||||
|
||||
#### Create OpenMC "tallies.xml" file
|
||||
|
||||
# Create a "tallies.xml" file for the mesh tallies
|
||||
tallies_file = openmc.Tallies([fission_rates, capture_rates])
|
||||
cell_mgxs_lib.add_to_tallies_file(tallies_file, merge=True)
|
||||
mat_mgxs_lib.add_to_tallies_file(tallies_file, merge=True)
|
||||
tallies_file.export_to_xml()
|
||||
|
|
@ -1,4 +1,4 @@
|
|||
from collections import OrderedDict, defaultdict
|
||||
from collections import OrderedDict
|
||||
import copy
|
||||
import os
|
||||
|
||||
|
|
@ -9,55 +9,7 @@ import opendeplete
|
|||
from geometry import beavrs, openmc_geometry
|
||||
|
||||
|
||||
#### Create "dummy" inputs to export distribcell paths for burnable cells
|
||||
|
||||
# Create OpenMC "materials.xml" file
|
||||
beavrs.write_openmc_materials()
|
||||
|
||||
# Create OpenMC "geometry.xml" file
|
||||
openmc_geometry.export_to_xml()
|
||||
|
||||
# Construct uniform initial source distribution over fissionable zones
|
||||
lower_left = [-10.70864, -10.70864, +192.5]
|
||||
upper_right = [+10.70864, +10.70864, +197.5]
|
||||
source = openmc.source.Source(space=openmc.stats.Box(lower_left, upper_right))
|
||||
source.space.only_fissionable = True
|
||||
|
||||
# Create OpenMC "settings.xml" file
|
||||
settings_file = openmc.Settings()
|
||||
settings_file.batches = 2
|
||||
settings_file.inactive = 1
|
||||
settings_file.particles = 10
|
||||
settings_file.output = {'tallies': False}
|
||||
settings_file.source = source
|
||||
settings_file.sourcepoint_write = False
|
||||
settings_file.export_to_xml()
|
||||
|
||||
# Create OpenMC "tallies.xml" file
|
||||
tallies = openmc.Tallies()
|
||||
fuel_cells = openmc_geometry.get_cells_by_name(
|
||||
name='enr radial 0: Fuel', case_sensitive=True)
|
||||
|
||||
# Instantiate a "dummy" distribcell tally for each cell we wish to deplete
|
||||
for cell in fuel_cells:
|
||||
tally = openmc.Tally(name='dummy distribcell tally')
|
||||
distribcell_filter = openmc.DistribcellFilter([cell.id])
|
||||
tally.filters = [distribcell_filter]
|
||||
tally.scores = ['fission']
|
||||
tallies.append(tally)
|
||||
|
||||
tallies.export_to_xml()
|
||||
|
||||
# Run OpenMC to generate summary.h5 file
|
||||
openmc.run()
|
||||
|
||||
# Open "summary.h5" file
|
||||
su = openmc.Summary('summary.h5')
|
||||
fuel_cells = su.openmc_geometry.get_cells_by_name(
|
||||
name='enr radial 0: Fuel', case_sensitive=True)
|
||||
|
||||
#### Setup OpenDeplete Materials wrapper
|
||||
|
||||
# Setup OpenDeplete Materials wrapper
|
||||
materials = opendeplete.Materials()
|
||||
materials.temperature = OrderedDict()
|
||||
materials.sab = OrderedDict()
|
||||
|
|
@ -65,10 +17,17 @@ materials.initial_density = OrderedDict()
|
|||
materials.burn = OrderedDict()
|
||||
materials.cross_sections = os.environ["OPENMC_CROSS_SECTIONS"]
|
||||
|
||||
# Count the number of instances for each cell and material
|
||||
openmc_geometry.determine_paths()
|
||||
|
||||
# Extract all cells filled by a fuel material
|
||||
fuel_cells = openmc_geometry.get_cells_by_name(
|
||||
name='enr radial 0: Fuel', case_sensitive=True)
|
||||
|
||||
# Extract cell materials, temperatures and sab
|
||||
for cell in su.openmc_geometry.get_all_material_cells():
|
||||
for cell in openmc_geometry.get_all_material_cells().values():
|
||||
materials.burn[cell.name] = 'fuel' in cell.fill.name.lower()
|
||||
materials.temperature[cell.name] = cell.temperature[0]
|
||||
materials.temperature[cell.name] = 300
|
||||
if len(cell.fill._sab) > 0:
|
||||
materials.sab[cell.name] = cell.fill._sab[0]
|
||||
|
||||
|
|
@ -83,9 +42,8 @@ for cell in fuel_cells:
|
|||
densities[nuclide][1] * 1e24
|
||||
|
||||
# Determine the maximum material ID
|
||||
all_mats = su.openmc_geometry.get_all_materials()
|
||||
max_material_id = 0
|
||||
for material in all_mats:
|
||||
for material in openmc_geometry.get_all_materials().values():
|
||||
max_material_id = max(max_material_id, material.id)
|
||||
|
||||
# FIXME: Automatically extract info needed to calculate burnable cell volumes
|
||||
|
|
@ -93,48 +51,45 @@ for material in all_mats:
|
|||
radius = 0.39218
|
||||
height = 5.
|
||||
|
||||
# Use defaultdict since OpenDeplete assumes volumes specified for all cells
|
||||
volumes = defaultdict(lambda: 1)
|
||||
|
||||
# Assign distribmats for each material
|
||||
for cell in fuel_cells:
|
||||
new_materials = []
|
||||
num_instances = len(cell.distribcell_paths)
|
||||
|
||||
for i in range(num_instances):
|
||||
for i in range(cell.num_instances):
|
||||
new_material = copy.deepcopy(cell.fill)
|
||||
new_material.id = max_material_id + 1
|
||||
max_material_id += 1
|
||||
new_materials.append(new_material)
|
||||
|
||||
# Store volume of burnable fuel rods cells
|
||||
volumes[new_material.id] = np.pi * radius**2 * height
|
||||
new_material.volume = np.pi * radius**2 * height
|
||||
new_material.depletable = True
|
||||
new_material.temperature = 300
|
||||
|
||||
cell.fill = new_materials
|
||||
|
||||
# Create dt vector for 1 month with 15 day timesteps
|
||||
dt1 = 15*24*60*60 # 15 days
|
||||
# Create dt vector for 1 month with 5 day timesteps
|
||||
dt1 = 5*24*60*60 # 5 days
|
||||
dt2 = 1.*30*24*60*60 # 1 months
|
||||
N = np.floor(dt2/dt1)
|
||||
dt = np.repeat([dt1], N)
|
||||
|
||||
# Create settings variable
|
||||
settings = opendeplete.Settings()
|
||||
|
||||
settings = opendeplete.OpenMCSettings()
|
||||
settings.openmc_call = ["mpirun", "openmc"]
|
||||
settings.particles = 30000
|
||||
settings.batches = 20
|
||||
settings.inactive = 10
|
||||
settings.lower_left = lower_left
|
||||
settings.upper_right = upper_right
|
||||
settings.lower_left = [-10.70864, -10.70864, +192.5]
|
||||
settings.upper_right = [+10.70864, +10.70864, +197.5]
|
||||
settings.entropy_dimension = [17, 17, 1]
|
||||
|
||||
settings.power = 2.337e15 * (17.**2 / 1.5**2) # MeV/second cm from CASMO
|
||||
# MeV/second cm from CASMO
|
||||
settings.power = 2.337e15 * (17.**2 / 1.5**2) * height
|
||||
settings.dt_vec = dt
|
||||
settings.output_dir = 'depleted'
|
||||
|
||||
op = opendeplete.Operator()
|
||||
op.geometry_fill(su.openmc_geometry, volumes, materials, settings)
|
||||
op = opendeplete.OpenMCOperator(openmc_geometry, settings)
|
||||
|
||||
# Perform simulation using the MCNPX/MCNP6 algorithm
|
||||
opendeplete.integrate(op, opendeplete.ce_cm_c1)
|
||||
|
|
|
|||
|
|
@ -64,9 +64,8 @@ plot_file.export_to_xml()
|
|||
#### Create OpenMC MGXS libraries
|
||||
|
||||
# Get all cells filled with a "fuel" material
|
||||
mat_cells = openmc_geometry.get_all_material_cells()
|
||||
fuel_cells = []
|
||||
for cell in mat_cells:
|
||||
for cell in openmc_geometry.get_all_material_cells().values():
|
||||
if 'fuel' in cell.fill.name.lower():
|
||||
fuel_cells.append(cell)
|
||||
|
||||
|
|
|
|||
|
|
@ -1,207 +0,0 @@
|
|||
"""Creates a 2D fuel assembly with reflective BCs."""
|
||||
|
||||
import numpy as np
|
||||
|
||||
import opencg
|
||||
import openmc
|
||||
import openmc.opencg_compatible as opencg_compatible
|
||||
from beavrs.builder import BEAVRS
|
||||
|
||||
|
||||
def find_assembly(assembly_name, wrap_geometry=True):
|
||||
"""Find a fuel assembly with some string name in the BEAVRS OpenCG model.
|
||||
|
||||
This method extracts the fuel assembly and wraps it in an OpenCG Geometry.
|
||||
The returned geometry has reflective boundary conditions along all
|
||||
boundaries. The z-axis is bounded between z=200 and z=210 cm.
|
||||
|
||||
Parameters
|
||||
----------
|
||||
assembly_name : str
|
||||
The name of the fuel assembly lattice
|
||||
wrap_geometry : bool
|
||||
If false, the fuel assembly Lattice is returned. If true, the fuel
|
||||
assembly Lattice is wrapped in an OpenCG Geometry and returned (default).
|
||||
|
||||
Returns
|
||||
-------
|
||||
fuel_assembly
|
||||
The OpenCG Lattice or Geometry for the assembly or None if not found
|
||||
|
||||
"""
|
||||
|
||||
# Get all OpenCG Universes
|
||||
all_univ = beavrs.main_universe.get_all_universes()
|
||||
|
||||
# Iterate over all Universes
|
||||
fuel_assembly = None
|
||||
for univ_id, univ in all_univ.items():
|
||||
if univ.name == assembly_name:
|
||||
fuel_assembly = univ
|
||||
|
||||
# Wrap lattice in a Geometry if requested by the user
|
||||
if wrap_geometry:
|
||||
|
||||
# Create a root Cell
|
||||
root_cell = opencg.Cell(name='root cell')
|
||||
root_cell.fill = fuel_assembly
|
||||
|
||||
# Make mixed reflective / vacuum boundaries
|
||||
min_x = opencg.XPlane(x0=root_cell.fill.min_x, boundary='reflective')
|
||||
max_x = opencg.XPlane(x0=root_cell.fill.max_x, boundary='reflective')
|
||||
min_y = opencg.YPlane(y0=root_cell.fill.min_y, boundary='reflective')
|
||||
max_y = opencg.YPlane(y0=root_cell.fill.max_y, boundary='reflective')
|
||||
max_z = opencg.ZPlane(z0=197.5, boundary='reflective')
|
||||
min_z = opencg.ZPlane(z0=192.5, boundary='reflective')
|
||||
|
||||
# Add boundaries to the root Cell
|
||||
root_cell.add_surface(surface=min_x, halfspace=+1)
|
||||
root_cell.add_surface(surface=max_x, halfspace=-1)
|
||||
root_cell.add_surface(surface=min_y, halfspace=+1)
|
||||
root_cell.add_surface(surface=max_y, halfspace=-1)
|
||||
root_cell.add_surface(surface=min_z, halfspace=+1)
|
||||
root_cell.add_surface(surface=max_z, halfspace=-1)
|
||||
|
||||
# Create a root Universe
|
||||
root_univ = opencg.Universe(universe_id=0, name='root universe')
|
||||
root_univ.add_cell(root_cell)
|
||||
|
||||
# Create a Geometry
|
||||
fuel_assembly = opencg.Geometry()
|
||||
fuel_assembly.root_universe = root_univ
|
||||
|
||||
return fuel_assembly
|
||||
|
||||
|
||||
#### Create OpenMC "materials.xml" and "geometry.xml" files
|
||||
|
||||
# Instantiate a BEAVRS object
|
||||
beavrs = BEAVRS(nndc_xs=True)
|
||||
|
||||
# Write all BEAVRS materials to materials.xml file
|
||||
beavrs.write_openmc_materials()
|
||||
|
||||
# Extract fuel assembly of interest from BEAVRS model
|
||||
assm_name = 'Fuel 1.6% enr instr no BAs'
|
||||
fuel_assembly = find_assembly(assm_name)
|
||||
openmc_geometry = opencg_compatible.get_openmc_geometry(fuel_assembly)
|
||||
openmc_geometry.export_to_xml()
|
||||
|
||||
|
||||
#### Create OpenMC "settings.xml" file
|
||||
|
||||
# Query the user on whether to use multipole cross sections
|
||||
multipole = input('Use multipole cross sections? (y/n): ').lower()
|
||||
multipole = True if multipole == 'y' else False
|
||||
|
||||
# Construct uniform initial source distribution over fissionable zones
|
||||
lower_left = fuel_assembly.bounds[:3]
|
||||
upper_right = fuel_assembly.bounds[3:]
|
||||
source = openmc.source.Source(space=openmc.stats.Box(lower_left, upper_right))
|
||||
source.space.only_fissionable = True
|
||||
|
||||
settings_file = openmc.Settings()
|
||||
settings_file.batches = 10
|
||||
settings_file.inactive = 5
|
||||
settings_file.particles = 10000
|
||||
settings_file.ptables = True
|
||||
settings_file.output = {'tallies': False}
|
||||
settings_file.source = source
|
||||
settings_file.sourcepoint_write = False
|
||||
|
||||
if multipole:
|
||||
settings_file.temperature = {'multipole': True, 'tolerance': 1000}
|
||||
|
||||
settings_file.export_to_xml()
|
||||
|
||||
|
||||
#### Create OpenMC "plots.xml" file
|
||||
|
||||
# Initialize the BEAVRS color mapping scheme
|
||||
beavrs.write_openmc_plots()
|
||||
|
||||
# Create a plot colored by materials
|
||||
plot = openmc.Plot()
|
||||
bounds = fuel_assembly.bounds
|
||||
plot.width = [fuel_assembly.max_x - fuel_assembly.min_x,
|
||||
fuel_assembly.max_y - fuel_assembly.min_y]
|
||||
plot.origin = [bounds[0] + (bounds[3] - bounds[0]) / 2.,
|
||||
bounds[1] + (bounds[4] - bounds[1]) / 2.,
|
||||
bounds[2] + (bounds[5] - bounds[2]) / 2.]
|
||||
plot.color = 'mat'
|
||||
plot.filename = 'assembly'
|
||||
plot.col_spec = beavrs.plots.colspec_mat
|
||||
plot.pixels = [1000, 1000]
|
||||
|
||||
plot_file = openmc.Plots([plot])
|
||||
plot_file.export_to_xml()
|
||||
|
||||
|
||||
#### Create OpenMC MGXS libraries
|
||||
|
||||
# Get all cells filled with a "fuel" material
|
||||
mat_cells = openmc_geometry.get_all_material_cells()
|
||||
fuel_cells = []
|
||||
for cell in mat_cells:
|
||||
if 'fuel' in cell.fill.name.lower():
|
||||
fuel_cells.append(cell)
|
||||
|
||||
# CASMO 70-group structure
|
||||
energy_groups = openmc.mgxs.EnergyGroups()
|
||||
energy_groups.group_edges = np.array([
|
||||
0, 0.005, 0.01, 0.015, 0.02, 0.025, 0.03, 0.035, 0.042, 0.05, 0.058, 0.067,
|
||||
0.08, 0.1, 0.14, 0.18, 0.22, 0.25, 0.28, 0.3, 0.32, 0.35, 0.4, 0.5, 0.625,
|
||||
0.78, 0.85, 0.91, 0.95, 0.972, 0.996, 1.02, 1.045, 1.071, 1.097, 1.123,
|
||||
1.15, 1.3, 1.5, 1.855, 2.1, 2.6, 3.3, 4., 9.877, 15.968, 27.7, 48.052,
|
||||
75.501, 148.73, 367.26001, 906.90002, 1.4251e3, 2.2395e3, 3.5191e3, 5.53e3,
|
||||
9.118e3, 15.03e3, 24.78e3, 40.85e3, 67.34e3, 111.e3, 183e3, 302.5e3, 500e3,
|
||||
821e3, 1.353e6, 2.231e6, 3.679e6, 6.0655e6, 2e7])
|
||||
|
||||
# Initialize a 70-group "distribcell" MGXS library
|
||||
cell_mgxs_lib = openmc.mgxs.Library(openmc_geometry, by_nuclide=True)
|
||||
cell_mgxs_lib.energy_groups = energy_groups
|
||||
cell_mgxs_lib.mgxs_types = ['total', 'nu-fission', 'nu-scatter matrix', 'chi']
|
||||
cell_mgxs_lib.domain_type = 'distribcell'
|
||||
cell_mgxs_lib.domains = fuel_cells
|
||||
cell_mgxs_lib.correction = None
|
||||
cell_mgxs_lib.build_library()
|
||||
|
||||
# Initialize a 70-group "material" MGXS library
|
||||
mat_mgxs_lib = openmc.mgxs.Library(openmc_geometry, by_nuclide=True)
|
||||
mat_mgxs_lib.energy_groups = energy_groups
|
||||
mat_mgxs_lib.mgxs_types = ['total', 'nu-fission', 'nu-scatter matrix', 'chi']
|
||||
mat_mgxs_lib.domain_type = 'material'
|
||||
mat_mgxs_lib.correction = None
|
||||
mat_mgxs_lib.build_library()
|
||||
|
||||
|
||||
#### Create mesh tallies for verification of pin-wise reaction rates
|
||||
|
||||
# Instantiate a tally Mesh
|
||||
mesh = openmc.Mesh(name='assembly mesh')
|
||||
mesh.type = 'regular'
|
||||
mesh.dimension = [17, 17, 1]
|
||||
mesh.lower_left = lower_left
|
||||
mesh.width = (np.array(upper_right) - np.array(lower_left))
|
||||
mesh.width[:2] /= 17
|
||||
mesh_filter = openmc.MeshFilter(mesh)
|
||||
|
||||
# Instantiate energy-integrated fission rate mesh Tally
|
||||
fission_rates = openmc.Tally(name='fission rates')
|
||||
fission_rates.filters = [mesh_filter]
|
||||
fission_rates.scores = ['fission']
|
||||
|
||||
# Instantiate energy-wise U-238 capture rate mesh Tally
|
||||
capture_rates = openmc.Tally(name='u-238 capture')
|
||||
capture_rates.filters = [mesh_filter]
|
||||
capture_rates.nuclides = ['U238']
|
||||
capture_rates.scores = ['absorption', 'fission']
|
||||
|
||||
|
||||
#### Create OpenMC "tallies.xml" file
|
||||
|
||||
# Create a "tallies.xml" file for the mesh tallies
|
||||
tallies_file = openmc.Tallies([fission_rates, capture_rates])
|
||||
cell_mgxs_lib.add_to_tallies_file(tallies_file, merge=True)
|
||||
mat_mgxs_lib.add_to_tallies_file(tallies_file, merge=True)
|
||||
tallies_file.export_to_xml()
|
||||
|
|
@ -1,4 +1,4 @@
|
|||
from collections import OrderedDict, defaultdict
|
||||
from collections import OrderedDict
|
||||
import copy
|
||||
import os
|
||||
|
||||
|
|
@ -9,55 +9,7 @@ import opendeplete
|
|||
from geometry import beavrs, openmc_geometry
|
||||
|
||||
|
||||
#### Create "dummy" inputs to export distribcell paths for burnable cells
|
||||
|
||||
# Create OpenMC "materials.xml" file
|
||||
beavrs.write_openmc_materials()
|
||||
|
||||
# Create OpenMC "geometry.xml" file
|
||||
openmc_geometry.export_to_xml()
|
||||
|
||||
# Construct uniform initial source distribution over fissionable zones
|
||||
lower_left = [-0.62992, -0.62992, -10.0]
|
||||
upper_right = [+0.62992, +0.62992, +10.0]
|
||||
source = openmc.source.Source(space=openmc.stats.Box(lower_left, upper_right))
|
||||
source.space.only_fissionable = True
|
||||
|
||||
# Create OpenMC "settings.xml" file
|
||||
settings_file = openmc.Settings()
|
||||
settings_file.batches = 2
|
||||
settings_file.inactive = 1
|
||||
settings_file.particles = 10
|
||||
settings_file.output = {'tallies': False}
|
||||
settings_file.source = source
|
||||
settings_file.sourcepoint_write = False
|
||||
settings_file.export_to_xml()
|
||||
|
||||
# Create OpenMC "tallies.xml" file
|
||||
tallies = openmc.Tallies()
|
||||
fuel_cells = openmc_geometry.get_cells_by_name(
|
||||
name='enr radial 0: Fuel', case_sensitive=True)
|
||||
|
||||
# Instantiate a "dummy" distribcell tally for each cell we wish to deplete
|
||||
for cell in fuel_cells:
|
||||
tally = openmc.Tally(name='dummy distribcell tally')
|
||||
distribcell_filter = openmc.DistribcellFilter([cell.id])
|
||||
tally.filters = [distribcell_filter]
|
||||
tally.scores = ['fission']
|
||||
tallies.append(tally)
|
||||
|
||||
tallies.export_to_xml()
|
||||
|
||||
# Run OpenMC to generate summary.h5 file
|
||||
openmc.run()
|
||||
|
||||
# Open "summary.h5" file
|
||||
su = openmc.Summary('summary.h5')
|
||||
fuel_cells = su.openmc_geometry.get_cells_by_name(
|
||||
name='enr radial 0: Fuel', case_sensitive=True)
|
||||
|
||||
#### Setup OpenDeplete Materials wrapper
|
||||
|
||||
# Setup OpenDeplete Materials wrapper
|
||||
materials = opendeplete.Materials()
|
||||
materials.temperature = OrderedDict()
|
||||
materials.sab = OrderedDict()
|
||||
|
|
@ -65,10 +17,17 @@ materials.initial_density = OrderedDict()
|
|||
materials.burn = OrderedDict()
|
||||
materials.cross_sections = os.environ["OPENMC_CROSS_SECTIONS"]
|
||||
|
||||
# Count the number of instances for each cell and material
|
||||
openmc_geometry.determine_paths()
|
||||
|
||||
# Extract all cells filled by a fuel material
|
||||
fuel_cells = openmc_geometry.get_cells_by_name(
|
||||
name='enr radial 0: Fuel', case_sensitive=True)
|
||||
|
||||
# Extract cell materials, temperatures and sab
|
||||
for cell in su.openmc_geometry.get_all_material_cells():
|
||||
for cell in openmc_geometry.get_all_material_cells().values():
|
||||
materials.burn[cell.name] = 'fuel' in cell.fill.name.lower()
|
||||
materials.temperature[cell.name] = cell.temperature[0]
|
||||
materials.temperature[cell.name] = 300
|
||||
if len(cell.fill._sab) > 0:
|
||||
materials.sab[cell.name] = cell.fill._sab[0]
|
||||
|
||||
|
|
@ -83,9 +42,8 @@ for cell in fuel_cells:
|
|||
densities[nuclide][1] * 1e24
|
||||
|
||||
# Determine the maximum material ID
|
||||
all_mats = su.openmc_geometry.get_all_materials()
|
||||
max_material_id = 0
|
||||
for material in all_mats:
|
||||
for material in openmc_geometry.get_all_materials().values():
|
||||
max_material_id = max(max_material_id, material.id)
|
||||
|
||||
# FIXME: Automatically extract info needed to calculate burnable cell volumes
|
||||
|
|
@ -93,48 +51,45 @@ for material in all_mats:
|
|||
radius = 0.39218
|
||||
height = 5.
|
||||
|
||||
# Use defaultdict since OpenDeplete assumes volumes specified for all cells
|
||||
volumes = defaultdict(lambda: 1)
|
||||
|
||||
# Assign distribmats for each material
|
||||
for cell in fuel_cells:
|
||||
new_materials = []
|
||||
num_instances = len(cell.distribcell_paths)
|
||||
|
||||
for i in range(num_instances):
|
||||
for i in range(cell.num_instances):
|
||||
new_material = copy.deepcopy(cell.fill)
|
||||
new_material.id = max_material_id + 1
|
||||
max_material_id += 1
|
||||
new_materials.append(new_material)
|
||||
|
||||
# Store volume of burnable fuel rods cells
|
||||
volumes[new_material.id] = np.pi * radius**2 * height
|
||||
new_material.volume = np.pi * radius**2 * height
|
||||
new_material.depletable = True
|
||||
new_material.temperature = 300
|
||||
|
||||
cell.fill = new_materials
|
||||
|
||||
# Create dt vector for 1 month with 15 day timesteps
|
||||
dt1 = 15*24*60*60 # 15 days
|
||||
# Create dt vector for 1 month with 5 day timesteps
|
||||
dt1 = 5*24*60*60 # 5 days
|
||||
dt2 = 1.*30*24*60*60 # 1 months
|
||||
N = np.floor(dt2/dt1)
|
||||
dt = np.repeat([dt1], N)
|
||||
|
||||
# Create settings variable
|
||||
settings = opendeplete.Settings()
|
||||
|
||||
settings = opendeplete.OpenMCSettings()
|
||||
settings.openmc_call = ["mpirun", "openmc"]
|
||||
settings.particles = 10000
|
||||
settings.batches = 20
|
||||
settings.inactive = 10
|
||||
settings.lower_left = lower_left
|
||||
settings.upper_right = upper_right
|
||||
settings.lower_left = [-0.62992, -0.62992, -10.0]
|
||||
settings.upper_right = [+0.62992, +0.62992, +10.0]
|
||||
settings.entropy_dimension = [1, 1, 1]
|
||||
|
||||
settings.power = 2.337e15 * (1.**2 / 1.5**2) # MeV/second cm from CASMO
|
||||
# MeV/second cm from CASMO
|
||||
settings.power = 2.337e15 * (1.**2 / 1.5**2) * height
|
||||
settings.dt_vec = dt
|
||||
settings.output_dir = 'depleted'
|
||||
|
||||
op = opendeplete.Operator()
|
||||
op.geometry_fill(su.openmc_geometry, volumes, materials, settings)
|
||||
op = opendeplete.OpenMCOperator(openmc_geometry, settings)
|
||||
|
||||
# Perform simulation using the MCNPX/MCNP6 algorithm
|
||||
opendeplete.integrate(op, opendeplete.ce_cm_c1)
|
||||
|
|
|
|||
|
|
@ -1,164 +0,0 @@
|
|||
"""Creates a 2D fuel pin cell with reflective BCs."""
|
||||
|
||||
import numpy as np
|
||||
|
||||
import opencg
|
||||
import openmc
|
||||
import openmc.opencg_compatible as opencg_compatible
|
||||
from beavrs.builder import BEAVRS
|
||||
|
||||
|
||||
def find_pin(pin_name, wrap_geometry=True):
|
||||
"""Find a fuel pin with some string name in the BEAVRS OpenCG model.
|
||||
|
||||
This method extracts the pin cell and wraps it in an OpenCG Geometry.
|
||||
The returned geometry has reflective boundary conditions along the x and y
|
||||
boundaries. The z-axis left unbounded.
|
||||
|
||||
Parameters
|
||||
----------
|
||||
pin_name : str
|
||||
The name of the fuel pin universe
|
||||
wrap_geometry : bool
|
||||
If false, the pin cell Universe is returned. If true, the pin cell
|
||||
Universe is wrapped in an OpenCG Geometry and returned (default).
|
||||
|
||||
Returns
|
||||
-------
|
||||
opencg.Universe
|
||||
The OpenCG Universe or Geometry for this fuel pin or None if not found
|
||||
|
||||
"""
|
||||
|
||||
# Get all OpenCG Universes
|
||||
all_univ = beavrs.main_universe.get_all_universes()
|
||||
|
||||
# Iterate over all Universes
|
||||
fuel_pin = None
|
||||
for univ_id, univ in all_univ.items():
|
||||
if univ._name == pin_name:
|
||||
fuel_pin = univ
|
||||
|
||||
# Wrap pin cell Universe in a Geometry if requested by the user
|
||||
if wrap_geometry:
|
||||
|
||||
# Make reflective boundaries
|
||||
pin_pitch = 0.62992
|
||||
min_x = opencg.XPlane(x0=-pin_pitch, boundary='reflective')
|
||||
max_x = opencg.XPlane(x0=pin_pitch, boundary='reflective')
|
||||
min_y = opencg.YPlane(y0=-pin_pitch, boundary='reflective')
|
||||
max_y = opencg.YPlane(y0=pin_pitch, boundary='reflective')
|
||||
|
||||
# Create a root Cell
|
||||
root_cell = opencg.Cell(name='root cell')
|
||||
root_cell.fill = fuel_pin
|
||||
|
||||
# Add boundaries to the root Cell
|
||||
root_cell.add_surface(surface=min_x, halfspace=+1)
|
||||
root_cell.add_surface(surface=max_x, halfspace=-1)
|
||||
root_cell.add_surface(surface=min_y, halfspace=+1)
|
||||
root_cell.add_surface(surface=max_y, halfspace=-1)
|
||||
|
||||
# Create a root Universe
|
||||
root_univ = opencg.Universe(universe_id=0, name='root universe')
|
||||
root_univ.add_cell(root_cell)
|
||||
|
||||
# Create a Geometry
|
||||
fuel_pin = opencg.Geometry()
|
||||
fuel_pin.root_universe = root_univ
|
||||
|
||||
return fuel_pin
|
||||
|
||||
|
||||
#### Create OpenMC "materials.xml" and "geometry.xml" files
|
||||
|
||||
# User-specified enrichment of 1.6, 2.4 or 3.1 percent
|
||||
enrichment = 1.6
|
||||
|
||||
# Instantiate a BEAVRS object
|
||||
beavrs = BEAVRS(nndc_xs=True)
|
||||
|
||||
# Write all BEAVRS materials to materials.xml file
|
||||
beavrs.write_openmc_materials()
|
||||
|
||||
# Extract fuel pin of interest from BEAVRS model
|
||||
pin_name = 'Fuel rod active region - {}% enr'.format(enrichment)
|
||||
pin_geometry = find_pin(pin_name)
|
||||
openmc_geometry = opencg_compatible.get_openmc_geometry(pin_geometry)
|
||||
openmc_geometry.export_to_xml()
|
||||
|
||||
|
||||
#### Create OpenMC "settings.xml" file
|
||||
|
||||
# Query the user on whether to use multipole cross sections
|
||||
multipole = input('Use multipole cross sections? (y/n): ').lower()
|
||||
multipole = True if multipole == 'y' else False
|
||||
|
||||
# Construct uniform initial source distribution over fissionable zones
|
||||
lower_left = pin_geometry.bounds[:2] + [-10.]
|
||||
upper_right = pin_geometry.bounds[3:5] + [10.]
|
||||
source = openmc.source.Source(space=openmc.stats.Box(lower_left, upper_right))
|
||||
source.space.only_fissionable = True
|
||||
|
||||
settings_file = openmc.Settings()
|
||||
settings_file.batches = 10
|
||||
settings_file.inactive = 5
|
||||
settings_file.particles = 10000
|
||||
settings_file.ptables = True
|
||||
settings_file.output = {'tallies': False}
|
||||
settings_file.source = source
|
||||
settings_file.sourcepoint_write = False
|
||||
|
||||
if multipole:
|
||||
settings_file.temperature = {'multipole': True, 'tolerance': 1000}
|
||||
|
||||
settings_file.export_to_xml()
|
||||
|
||||
|
||||
#### Create OpenMC "plots.xml" file
|
||||
|
||||
# Initialize the BEAVRS color mapping scheme
|
||||
beavrs.write_openmc_plots()
|
||||
|
||||
# Create a plot colored by materials
|
||||
plot = openmc.Plot()
|
||||
bounds = pin_geometry.bounds
|
||||
plot.width = [pin_geometry.max_x - pin_geometry.min_x,
|
||||
pin_geometry.max_y - pin_geometry.min_y]
|
||||
plot.origin = [bounds[0] + (bounds[3] - bounds[0]) / 2.,
|
||||
bounds[1] + (bounds[4] - bounds[1]) / 2.,
|
||||
bounds[2] + (bounds[5] - bounds[2]) / 2.]
|
||||
plot.color = 'mat'
|
||||
plot.filename = 'fuel-pin'
|
||||
plot.col_spec = beavrs.plots.colspec_mat
|
||||
plot.pixels = [1000, 1000]
|
||||
|
||||
plot_file = openmc.Plots([plot])
|
||||
plot_file.export_to_xml()
|
||||
|
||||
|
||||
#### Create OpenMC MGXS library and "tallies.xml" file
|
||||
|
||||
# CASMO 70-group structure
|
||||
energy_groups = openmc.mgxs.EnergyGroups()
|
||||
energy_groups.group_edges = np.array([
|
||||
0, 0.005, 0.01, 0.015, 0.02, 0.025, 0.03, 0.035, 0.042, 0.05, 0.058, 0.067,
|
||||
0.08, 0.1, 0.14, 0.18, 0.22, 0.25, 0.28, 0.3, 0.32, 0.35, 0.4, 0.5, 0.625,
|
||||
0.78, 0.85, 0.91, 0.95, 0.972, 0.996, 1.02, 1.045, 1.071, 1.097, 1.123,
|
||||
1.15, 1.3, 1.5, 1.855, 2.1, 2.6, 3.3, 4., 9.877, 15.968, 27.7, 48.052,
|
||||
75.501, 148.73, 367.26001, 906.90002, 1.4251e3, 2.2395e3, 3.5191e3, 5.53e3,
|
||||
9.118e3, 15.03e3, 24.78e3, 40.85e3, 67.34e3, 111.e3, 183e3, 302.5e3, 500e3,
|
||||
821e3, 1.353e6, 2.231e6, 3.679e6, 6.0655e6, 2e7])
|
||||
|
||||
# Initialize a 70-group MGXS library
|
||||
mgxs_lib = openmc.mgxs.Library(openmc_geometry, by_nuclide=True)
|
||||
mgxs_lib.energy_groups = energy_groups
|
||||
mgxs_lib.mgxs_types = ['total', 'nu-fission', 'nu-scatter matrix', 'chi']
|
||||
mgxs_lib.domain_type = 'material'
|
||||
mgxs_lib.correction = None
|
||||
mgxs_lib.build_library()
|
||||
|
||||
# Create a "tallies.xml" file for the MGXS Library
|
||||
tallies_file = openmc.Tallies()
|
||||
mgxs_lib.add_to_tallies_file(tallies_file, merge=True)
|
||||
tallies_file.export_to_xml()
|
||||
|
|
@ -1,4 +1,4 @@
|
|||
from collections import OrderedDict, defaultdict
|
||||
from collections import OrderedDict
|
||||
import copy
|
||||
import os
|
||||
|
||||
|
|
@ -10,64 +10,7 @@ from smr.materials import materials
|
|||
from smr.surfaces import lattice_pitch, bottom_fuel_stack, top_active_core
|
||||
from smr.core import geometry
|
||||
|
||||
#### Create "dummy" inputs to export distribcell paths for burnable cells
|
||||
|
||||
# Create OpenMC "geometry.xml" file
|
||||
geometry.export_to_xml()
|
||||
|
||||
# Create OpenMC "materials.xml" file
|
||||
materials.export_to_xml()
|
||||
|
||||
# Construct uniform initial source distribution over fissionable zones
|
||||
lower_left = [-7.*lattice_pitch/2., -7.*lattice_pitch/2., bottom_fuel_stack]
|
||||
upper_right = [+7.*lattice_pitch/2., +7.*lattice_pitch/2., top_active_core]
|
||||
source = openmc.source.Source(space=openmc.stats.Box(lower_left, upper_right))
|
||||
source.space.only_fissionable = True
|
||||
|
||||
# Create OpenMC "settings.xml" file
|
||||
settings = openmc.Settings()
|
||||
settings.batches = 2
|
||||
settings.inactive = 1
|
||||
settings.particles = 10
|
||||
settings.output = {'tallies': False}
|
||||
settings.source = source
|
||||
settings.sourcepoint_write = False
|
||||
settings.export_to_xml()
|
||||
|
||||
# Create OpenMC "tallies.xml" file
|
||||
tallies = openmc.Tallies()
|
||||
fuel_cells = geometry.get_cells_by_name(
|
||||
name='(1.6%) (0)', case_sensitive=True)
|
||||
fuel_cells.extend(geometry.get_cells_by_name(
|
||||
name='(2.4%) (0)', case_sensitive=True))
|
||||
fuel_cells.extend(geometry.get_cells_by_name(
|
||||
name='(3.1%) (0)', case_sensitive=True))
|
||||
|
||||
# Instantiate a "dummy" distribcell tally for each cell we wish to deplete
|
||||
for cell in fuel_cells:
|
||||
tally = openmc.Tally(name='dummy distribcell tally')
|
||||
distribcell_filter = openmc.DistribcellFilter([cell.id])
|
||||
tally.filters = [distribcell_filter]
|
||||
tally.scores = ['fission']
|
||||
tallies.append(tally)
|
||||
|
||||
tallies.export_to_xml()
|
||||
|
||||
# Run OpenMC to generate summary.h5 file
|
||||
openmc.run()
|
||||
|
||||
# Open "summary.h5" file
|
||||
su = openmc.Summary('summary.h5')
|
||||
fuel_cells = su.openmc_geometry.get_cells_by_name(
|
||||
name='(1.6%) (0)', case_sensitive=True)
|
||||
fuel_cells.extend(su.openmc_geometry.get_cells_by_name(
|
||||
name='(2.4%) (0)', case_sensitive=True))
|
||||
fuel_cells.extend(su.openmc_geometry.get_cells_by_name(
|
||||
name='(3.1%) (0)', case_sensitive=True))
|
||||
|
||||
|
||||
#### Setup OpenDeplete Materials wrapper
|
||||
|
||||
# Setup OpenDeplete Materials wrapper
|
||||
materials = opendeplete.Materials()
|
||||
materials.temperature = OrderedDict()
|
||||
materials.sab = OrderedDict()
|
||||
|
|
@ -75,10 +18,21 @@ materials.initial_density = OrderedDict()
|
|||
materials.burn = OrderedDict()
|
||||
materials.cross_sections = os.environ["OPENMC_CROSS_SECTIONS"]
|
||||
|
||||
# Count the number of instances for each cell and material
|
||||
geometry.determine_paths()
|
||||
|
||||
# Extract all cells filled by a fuel material
|
||||
fuel_cells = geometry.get_cells_by_name(
|
||||
name='(1.6%) (0)', case_sensitive=True)
|
||||
fuel_cells.extend(geometry.get_cells_by_name(
|
||||
name='(2.4%) (0)', case_sensitive=True))
|
||||
fuel_cells.extend(geometry.get_cells_by_name(
|
||||
name='(3.1%) (0)', case_sensitive=True))
|
||||
|
||||
# Extract cell materials, temperatures and sab
|
||||
for cell in su.openmc_geometry.get_all_material_cells():
|
||||
for cell in geometry.get_all_material_cells().values():
|
||||
materials.burn[cell.name] = 'fuel' in cell.fill.name.lower()
|
||||
materials.temperature[cell.name] = cell.temperature[0]
|
||||
materials.temperature[cell.name] = 300
|
||||
if len(cell.fill._sab) > 0:
|
||||
materials.sab[cell.name] = cell.fill._sab[0]
|
||||
|
||||
|
|
@ -93,58 +47,56 @@ for cell in fuel_cells:
|
|||
densities[nuclide][1] * 1e24
|
||||
|
||||
# Determine the maximum material ID
|
||||
all_mats = su.openmc_geometry.get_all_materials()
|
||||
max_material_id = 0
|
||||
for material in all_mats:
|
||||
for material in geometry.get_all_materials().values():
|
||||
max_material_id = max(max_material_id, material.id)
|
||||
|
||||
# FIXME: Automatically extract info needed to calculate burnable cell volumes
|
||||
# Fuel rod geometric parameters
|
||||
radius = 0.39218
|
||||
height = 5.
|
||||
|
||||
# Use defaultdict since OpenDeplete assumes volumes specified for all cells
|
||||
volumes = defaultdict(lambda: 1)
|
||||
height = 200.
|
||||
|
||||
# Assign distribmats for each material
|
||||
for cell in fuel_cells:
|
||||
new_materials = []
|
||||
num_instances = len(cell.distribcell_paths)
|
||||
|
||||
for i in range(num_instances):
|
||||
for i in range(cell.num_instances):
|
||||
new_material = copy.deepcopy(cell.fill)
|
||||
new_material.id = max_material_id + 1
|
||||
max_material_id += 1
|
||||
new_materials.append(new_material)
|
||||
|
||||
# Store volume of burnable fuel rods cells
|
||||
volumes[new_material.id] = np.pi * radius**2 * height
|
||||
new_material.volume = np.pi * radius**2 * height
|
||||
new_material.depletable = True
|
||||
new_material.temperature = 300
|
||||
|
||||
cell.fill = new_materials
|
||||
|
||||
# Create dt vector for 1 month with 15 day timesteps
|
||||
dt1 = 15*24*60*60 # 15 days
|
||||
# Create dt vector for 1 month with 5 day timesteps
|
||||
dt1 = 5*24*60*60 # 5 days
|
||||
dt2 = 1.*30*24*60*60 # 1 months
|
||||
N = np.floor(dt2/dt1)
|
||||
dt = np.repeat([dt1], N)
|
||||
|
||||
# Create settings variable
|
||||
settings = opendeplete.Settings()
|
||||
|
||||
settings = opendeplete.OpenMCSettings()
|
||||
settings.openmc_call = ["mpirun", "openmc"]
|
||||
settings.particles = 1000000
|
||||
settings.batches = 200
|
||||
settings.inactive = 100
|
||||
settings.lower_left = lower_left
|
||||
settings.upper_right = upper_right
|
||||
settings.lower_left = \
|
||||
[-7.*lattice_pitch/2., -7.*lattice_pitch/2., bottom_fuel_stack]
|
||||
settings.upper_right = \
|
||||
[+7.*lattice_pitch/2., +7.*lattice_pitch/2., top_active_core]
|
||||
settings.entropy_dimension = [15, 15, 1]
|
||||
|
||||
settings.power = 2.337e15 * ((17.*17.*37.) / 1.5**2) # MeV/second cm from CASMO
|
||||
# MeV/second cm from CASMO
|
||||
settings.power = 2.337e15 * ((17.*17.*37.) / 1.5**2) * height
|
||||
settings.dt_vec = dt
|
||||
settings.output_dir = 'depleted'
|
||||
|
||||
op = opendeplete.Operator()
|
||||
op.geometry_fill(su.openmc_geometry, volumes, materials, settings)
|
||||
op = opendeplete.OpenMCOperator(geometry, settings)
|
||||
|
||||
# Perform simulation using the MCNPX/MCNP6 algorithm
|
||||
opendeplete.integrate(op, opendeplete.ce_cm_c1)
|
||||
|
|
|
|||
|
|
@ -51,9 +51,8 @@ plots.export_to_xml()
|
|||
#### Create OpenMC MGXS libraries
|
||||
|
||||
# Get all cells filled with a "fuel" material
|
||||
mat_cells = geometry.get_all_material_cells()
|
||||
fuel_cells = []
|
||||
for cell in mat_cells:
|
||||
for cell in geometry.get_all_material_cells().values():
|
||||
if 'fuel' in cell.fill.name.lower():
|
||||
fuel_cells.append(cell)
|
||||
|
||||
|
|
|
|||
109
smr/build-xml.py
109
smr/build-xml.py
|
|
@ -1,109 +0,0 @@
|
|||
import numpy as np
|
||||
|
||||
import openmc
|
||||
from smr.materials import materials
|
||||
from smr.plots import plots
|
||||
from smr.surfaces import lattice_pitch, bottom_fuel_stack, top_active_core
|
||||
from smr.core import geometry
|
||||
|
||||
|
||||
#### Create OpenMC "geometry.xml" file
|
||||
geometry.export_to_xml()
|
||||
|
||||
|
||||
#### Create OpenMC "materials.xml" file
|
||||
materials.export_to_xml()
|
||||
|
||||
|
||||
#### Create OpenMC "settings.xml" file
|
||||
|
||||
# Query the user on whether to use multipole cross sections
|
||||
multipole = input('Use multipole cross sections? (y/n): ').lower()
|
||||
multipole = (multipole == 'y')
|
||||
|
||||
# Construct uniform initial source distribution over fissionable zones
|
||||
lower_left = [-7.*lattice_pitch/2., -7.*lattice_pitch/2., bottom_fuel_stack]
|
||||
upper_right = [+7.*lattice_pitch/2., +7.*lattice_pitch/2., top_active_core]
|
||||
source = openmc.source.Source(space=openmc.stats.Box(lower_left, upper_right))
|
||||
source.space.only_fissionable = True
|
||||
|
||||
settings = openmc.Settings()
|
||||
settings.batches = 200
|
||||
settings.inactive = 100
|
||||
settings.particles = 10000
|
||||
settings.output = {'tallies': False}
|
||||
settings.source = source
|
||||
settings.sourcepoint_write = False
|
||||
|
||||
if multipole:
|
||||
settings.temperature = {'multipole': True, 'tolerance': 1000}
|
||||
|
||||
settings.export_to_xml()
|
||||
|
||||
|
||||
#### Create OpenMC "plots.xml" file
|
||||
plots.export_to_xml()
|
||||
|
||||
|
||||
#### Create OpenMC MGXS libraries
|
||||
|
||||
# Get all cells filled with a "fuel" material
|
||||
mat_cells = geometry.get_all_material_cells()
|
||||
fuel_cells = []
|
||||
for cell in mat_cells:
|
||||
if 'fuel' in cell.fill.name.lower():
|
||||
fuel_cells.append(cell)
|
||||
|
||||
# CASMO 8-group structure
|
||||
energy_groups = openmc.mgxs.EnergyGroups()
|
||||
energy_groups.group_edges = np.array([0., 0.058e-6, 0.14e-6, 0.28e-6,
|
||||
0.625e-6, 4.e-6, 5.53e-3, 821.e-3, 20.])
|
||||
|
||||
# Initialize a 70-group "distribcell" MGXS library
|
||||
cell_mgxs_lib = openmc.mgxs.Library(geometry, by_nuclide=True)
|
||||
cell_mgxs_lib.energy_groups = energy_groups
|
||||
cell_mgxs_lib.mgxs_types = ['total', 'nu-fission', 'nu-scatter matrix', 'chi']
|
||||
cell_mgxs_lib.domain_type = 'distribcell'
|
||||
cell_mgxs_lib.domains = fuel_cells
|
||||
cell_mgxs_lib.correction = None
|
||||
cell_mgxs_lib.build_library()
|
||||
|
||||
# Initialize a 70-group "material" MGXS library
|
||||
mat_mgxs_lib = openmc.mgxs.Library(geometry, by_nuclide=True)
|
||||
mat_mgxs_lib.energy_groups = energy_groups
|
||||
mat_mgxs_lib.mgxs_types = ['total', 'nu-fission', 'nu-scatter matrix', 'chi']
|
||||
mat_mgxs_lib.domain_type = 'material'
|
||||
mat_mgxs_lib.correction = None
|
||||
mat_mgxs_lib.build_library()
|
||||
|
||||
|
||||
#### Create mesh tallies for verification of pin-wise reaction rates
|
||||
|
||||
# Instantiate a tally Mesh
|
||||
mesh = openmc.Mesh(name='assembly mesh')
|
||||
mesh.type = 'regular'
|
||||
mesh.dimension = [7*17, 7*17, 100]
|
||||
mesh.lower_left = lower_left
|
||||
mesh.width = (np.array(upper_right) - np.array(lower_left))
|
||||
mesh.width[:2] /= (7*17)
|
||||
mesh_filter = openmc.MeshFilter(mesh)
|
||||
|
||||
# Instantiate energy-integrated fission rate mesh Tally
|
||||
fission_rates = openmc.Tally(name='fission rates')
|
||||
fission_rates.filters = [mesh_filter]
|
||||
fission_rates.scores = ['fission']
|
||||
|
||||
# Instantiate energy-wise U-238 capture rate mesh Tally
|
||||
capture_rates = openmc.Tally(name='u-238 capture')
|
||||
capture_rates.filters = [mesh_filter]
|
||||
capture_rates.nuclides = ['U238']
|
||||
capture_rates.scores = ['absorption', 'fission']
|
||||
|
||||
|
||||
#### Create OpenMC "tallies.xml" file
|
||||
|
||||
# Create a "tallies.xml" file for the mesh tallies
|
||||
tallies_file = openmc.Tallies([fission_rates, capture_rates])
|
||||
cell_mgxs_lib.add_to_tallies_file(tallies_file, merge=True)
|
||||
mat_mgxs_lib.add_to_tallies_file(tallies_file, merge=True)
|
||||
tallies_file.export_to_xml()
|
||||
Loading…
Add table
Add a link
Reference in a new issue