Fix sourcepoint write in core scripts

Fix the build-assembly.py script

smr/build-assembly-long.py

@@ -1,7 +1,6 @@
 #!/usr/bin/env python3
 
 import argparse
-import copy
 from math import pi, isclose
 from pathlib import Path
 
@@ -9,10 +8,9 @@ import numpy as np
 from tqdm import tqdm
 import openmc
 
-from smr.materials import materials, mats
+from smr.materials import materials, clone
 from smr.surfaces import surfs, lattice_pitch, pin_pitch, bottom_fuel_stack, \
-    top_active_core, pellet_OR, clad_OR, clad_IR, guide_tube_IR, guide_tube_OR, \
-    active_fuel_length
+    top_active_core, pellet_OR, active_fuel_length
 from smr.pins import pin_universes, make_stack
 
 
@@ -20,14 +18,18 @@ from smr.pins import pin_universes, make_stack
 parser = argparse.ArgumentParser()
 parser.add_argument('--multipole', action='store_true',
                     help='Use multipole cross sections')
-parser.add_argument('--no-multipole', action='store_false',
+parser.add_argument('--no-multipole', dest='multipole', action='store_false',
                     help='Do not use multipole cross sections')
+parser.add_argument('--clone', action='store_true',
+                    help='Clone materials for each cell instance')
+parser.add_argument('--no-clone', dest='clone', action='store_false',
+                    help='Do not clone materials for each cell instance')
 parser.add_argument('-a', '--axial', type=int, default=100,
                     help='Number of axial subdivisions in fuel')
 parser.add_argument('-d', '--depleted', action='store_true',
                     help='Whether UO2 compositions should represent depleted fuel')
 parser.add_argument('-o', '--output-dir', type=Path, default=None)
-parser.set_defaults(multipole=True)
+parser.set_defaults(clone=False, multipole=True)
 args = parser.parse_args()
 
 # Make directory for inputs
@@ -101,11 +103,22 @@ h = active_fuel_length / args.axial
 
 fuel_mats = {}
 
+# Count the number of instances for each cell and material
+if args.clone:
+    geometry.determine_paths(instances_only=True)
+
 for cell in tqdm(geometry.get_all_material_cells().values(),
-                 desc='Assigning volume'):
+                 desc='Differentiating materials / assigning volume'):
     if cell.fill in materials:
+        # Determine if this material is fuel
+        is_fuel = 'UO2 Fuel' in cell.fill.name
+
+        # Fill cell with list of "differentiated" materials if requested
+        if args.clone:
+            cell.fill = [clone(cell.fill) for i in range(cell.num_instances)]
+
         # Determine volume of each fuel material
-        if 'UO2 Fuel' in cell.fill.name:
+        if is_fuel:
             upper_right = cell.region.bounding_box[1]
             if isclose(upper_right[0], rings[0]):
                 ri, ro = 0.0, rings[0]
@@ -113,14 +126,25 @@ for cell in tqdm(geometry.get_all_material_cells().values(),
                 ri, ro = rings[0], rings[1]
             else:
                 ri, ro = rings[1], pellet_OR
-            if ri not in fuel_mats:
-                cell.fill = cell.fill.clone()
-                cell.fill.volume = pi * (ro*ro - ri*ri) * h
-                fuel_mats[ri] = cell.fill
+
+            if args.clone:
+                for mat in cell.fill:
+                    mat.volume = pi * (ro*ro - ri*ri) * h
             else:
-                cell.fill = fuel_mats[ri]
+                # In non-clone mode, we still need to create a copy of the
+                # material for each ring since they get different volumes
+                if ri not in fuel_mats:
+                    cell.fill = cell.fill.clone()
+                    cell.fill.volume = pi * (ro*ro - ri*ri) * h
+                    fuel_mats[ri] = cell.fill
+                else:
+                    cell.fill = fuel_mats[ri]
         else:
-            cell.fill.volume = 1.0
+            if args.clone:
+                for mat in cell.fill:
+                    mat.volume = 1.0
+            else:
+                cell.fill.volume = 1.0
 
 #### Create OpenMC "materials.xml" file
 print('Getting materials...')
@@ -150,14 +174,13 @@ settings.particles = 10000
 settings.output = {'tallies': False, 'summary': False}
 settings.source = source
 settings.sourcepoint = {'write': False}
-
-if args.multipole:
-    settings.temperature = {
-        'multipole': True,
-        'tolerance': 1000,
-        'default': 531.5,
-        'method': 'interpolation',
+settings.temperature = {
+    'default': 531.5,
+    'method': 'interpolation',
         'range': (500.0, 1300.0)
-    }
+}
+if args.multipole:
+    settings.temperature['multipole'] = True
+    settings.temperature['tolerance'] = 1000
 
 settings.export_to_xml(str(directory / 'settings.xml'))

smr/build-assembly-short.py

@@ -1,7 +1,6 @@
 #!/usr/bin/env python3
 
 import argparse
-import copy
 from math import pi, isclose
 from pathlib import Path
 
@@ -9,7 +8,7 @@ import numpy as np
 from tqdm import tqdm
 import openmc
 
-from smr.materials import materials, mats
+from smr.materials import materials, clone
 from smr.surfaces import surfs, lattice_pitch, pin_pitch, bottom_fuel_stack, \
     top_active_core, pellet_OR, clad_OR, clad_IR, guide_tube_IR, guide_tube_OR
 from smr.pins import pin_universes
@@ -85,14 +84,6 @@ for halfspace in surfs['lat grid box inner']:
 # Define geometry with a single assembly
 geometry = openmc.Geometry(root_universe)
 
-
-def clone(material):
-    """Perform copy of material but share nuclide densities"""
-    shared_mat = copy.copy(material)
-    shared_mat.id = None
-    return shared_mat
-
-
 #### "Differentiate" the geometry if using distribmats
 h = 10.0*pin_pitch / args.axial
 if args.tallies == 'mat':

smr/build-assembly.py

@@ -1,17 +1,15 @@
 #!/usr/bin/env python3
 
 import argparse
-import copy
 from pathlib import Path
 
 import numpy as np
 from tqdm import tqdm
 
 import openmc
-from smr.materials import materials
+from smr.materials import materials, clone
 from smr.surfaces import surfs, lattice_pitch, bottom_fuel_stack, top_active_core, pellet_OR
 from smr.assemblies import assembly_universes
-from smr.plots import assembly_plots
 from smr import inlet_temperature
 
 
@@ -19,8 +17,12 @@ from smr import inlet_temperature
 parser = argparse.ArgumentParser()
 parser.add_argument('--multipole', action='store_true',
                     help='Use multipole cross sections')
-parser.add_argument('--no-multipole', action='store_false',
+parser.add_argument('--no-multipole', dest='multipole', action='store_false',
                     help='Do not use multipole cross sections')
+parser.add_argument('--clone', action='store_true',
+                    help='Clone materials for each cell instance')
+parser.add_argument('--no-clone', dest='clone', action='store_false',
+                    help='Do not clone materials for each cell instance')
 parser.add_argument('-t', '--tallies', choices=('cell', 'mat'), default='mat',
                     help='Whether to use distribmats or distribcells for tallies')
 parser.add_argument('-r', '--rings', type=int, default=10,
@@ -30,7 +32,7 @@ parser.add_argument('-a', '--axial', type=int, default=196,
 parser.add_argument('-d', '--depleted', action='store_true',
                     help='Whether UO2 compositions should represent depleted fuel')
 parser.add_argument('-o', '--output-dir', type=Path, default=None)
-parser.set_defaults(multipole=True)
+parser.set_defaults(clone=False, multipole=True)
 args = parser.parse_args()
 
 # Make directory for inputs
@@ -49,25 +51,17 @@ if args.rings > 1:
 else:
     ring_radii = None
 assembly = assembly_universes(ring_radii, args.axial, args.depleted)
-lattice_sides = openmc.model.get_rectangular_prism(lattice_pitch, lattice_pitch,
-                                                   boundary_type='reflective')
+lattice_sides = openmc.model.rectangular_prism(lattice_pitch, lattice_pitch,
+                                               boundary_type='reflective')
 main_cell = openmc.Cell(
-    fill=assembly['Assembly (3.1%) 16BA'],
+    fill=assembly['Assembly (3.1%)'],
     region=lattice_sides & +surfs['lower bound'] & -surfs['upper bound']
 )
 root_univ = openmc.Universe(cells=[main_cell])
 geometry = openmc.Geometry(root_univ)
 
-
-def clone(material):
-    """Perform copy of material but share nuclide densities"""
-    shared_mat = copy.copy(material)
-    shared_mat.id = None
-    return shared_mat
-
-
 #### "Differentiate" the geometry if using distribmats
-if args.tallies == 'mat':
+if args.clone:
     # Count the number of instances for each cell and material
     geometry.determine_paths(instances_only=True)
 
@@ -107,7 +101,7 @@ settings.inactive = 100
 settings.particles = 10000
 settings.output = {'tallies': False, 'summary': False}
 settings.source = source
-settings.sourcepoint_write = False
+settings.sourcepoint = {'write': False}
 settings.temperature = {
     'default': inlet_temperature,
     'method': 'interpolation',
@@ -149,7 +143,3 @@ elif args.tallies == 'mat':
     tallies.append(tally)
 
 tallies.export_to_xml(str(directory / 'tallies.xml'))
-
-# Create plots
-plots = assembly_plots(main_cell.fill)
-plots.export_to_xml(str(directory / 'plots.xml'))

smr/build-core-fresh.py

@@ -1,17 +1,14 @@
 #!/usr/bin/env python3
 
-import os
-import shutil
-import copy
 import argparse
 from math import pi
 from pathlib import Path
 
 import numpy as np
-
 import openmc
-from smr.materials import materials
-from smr.plots import core_plots
+from tqdm import tqdm
+
+from smr.materials import materials, clone
 from smr.surfaces import lattice_pitch, bottom_fuel_stack, top_active_core, \
     pellet_OR, active_fuel_length
 from smr.core import core_geometry
@@ -22,8 +19,12 @@ from smr import inlet_temperature
 parser = argparse.ArgumentParser()
 parser.add_argument('--multipole', action='store_true',
                     help='Use multipole cross sections')
-parser.add_argument('--no-multipole', action='store_false',
+parser.add_argument('--no-multipole', dest='multipole', action='store_false',
                     help='Do not use multipole cross sections')
+parser.add_argument('--clone', action='store_true',
+                    help='Clone materials for each cell instance')
+parser.add_argument('--no-clone', dest='clone', action='store_false',
+                    help='Do not clone materials for each cell instance')
 parser.add_argument('-r', '--rings', type=int, default=10,
                     help='Number of annular regions in fuel')
 parser.add_argument('-a', '--axial', type=int, default=196,
@@ -31,7 +32,7 @@ parser.add_argument('-a', '--axial', type=int, default=196,
 parser.add_argument('-d', '--depleted', action='store_true',
                     help='Whether UO2 compositions should represent depleted fuel')
 parser.add_argument('-o', '--output-dir', type=Path, default=None)
-parser.set_defaults(multipole=True)
+parser.set_defaults(clone=False, multipole=True)
 args = parser.parse_args()
 
 # Make directory for inputs
@@ -53,22 +54,36 @@ geometry = core_geometry(ring_radii, args.axial, args.depleted)
 h = active_fuel_length / args.axial
 fuel_mats = {}
 
+# Count the number of instances for each cell and material
+if args.clone:
+    geometry.determine_paths(instances_only=True)
+
 fuel_volume = pi * pellet_OR**2 * h / args.rings
-for cell in geometry.get_all_cells().values():
+for cell in tqdm(geometry.get_all_cells().values(),
+                 desc='Differentiating materials / assigning volume'):
     if cell.fill in materials:
+        # Determine if this material is fuel
+        name = cell.fill.name
+        is_fuel = 'UO2 Fuel' in name
+
         # Determine volume of each fuel material
-        if 'UO2 Fuel' in cell.fill.name:
-            r_o = cell.region.bounding_box[1][0]
-            if r_o not in fuel_mats:
-                cell.fill = cell.fill.clone()
-                cell.fill.volume = fuel_volume
-                fuel_mats[r_o] = cell.fill
+        if is_fuel:
+            if args.clone:
+                # Fill cell with list of "differentiated" materials if requested
+                cell.fill = [clone(cell.fill) for i in range(cell.num_instances)]
+                for mat in cell.fill:
+                    mat.volume = fuel_volume
             else:
-                cell.fill = fuel_mats[r_o]
+                r_o = cell.region.bounding_box[1][0]
+                if (name, r_o) not in fuel_mats:
+                    cell.fill = cell.fill.clone()
+                    cell.fill.volume = fuel_volume
+                    fuel_mats[name, r_o] = cell.fill
+                else:
+                    cell.fill = fuel_mats[name, r_o]
         else:
             cell.fill.volume = 1.0
 
-
 #### Create OpenMC "materials.xml" file
 all_materials = geometry.get_all_materials()
 materials = openmc.Materials(all_materials.values())
@@ -93,8 +108,7 @@ settings.inactive = 100
 settings.particles = 10000
 settings.output = {'tallies': False, 'summary': False}
 settings.source = source
-settings.sourcepoint_write = False
-
+settings.sourcepoint = {'write': False}
 settings.temperature = {
     'default': inlet_temperature,
     'method': 'interpolation',
@@ -105,4 +119,3 @@ if args.multipole:
     settings.temperature['tolerance'] = 1000
 
 settings.export_to_xml(str(directory / 'settings.xml'))
-

smr/build-core-long.py

@@ -0,0 +1,118 @@
+#!/usr/bin/env python3
+
+import argparse
+from math import pi, isclose
+from pathlib import Path
+
+import openmc
+from smr.materials import materials
+from smr.surfaces import bottom_fuel_stack, top_active_core, \
+    pellet_OR, pin_pitch, clad_IR, clad_OR, active_fuel_length
+from smr.core import core_geometry
+from smr import inlet_temperature
+import smr.surfaces
+
+# Define command-line options
+parser = argparse.ArgumentParser()
+parser.add_argument('--multipole', action='store_true',
+                    help='Use multipole cross sections')
+parser.add_argument('--no-multipole', dest='multipole', action='store_false',
+                    help='Do not use multipole cross sections')
+parser.add_argument('-a', '--axial', type=int, default=100,
+                    help='Number of axial subdivisions in fuel')
+parser.add_argument('-d', '--depleted', action='store_true',
+                    help='Whether UO2 compositions should represent depleted fuel')
+parser.add_argument('-o', '--output-dir', type=Path, default=None)
+parser.set_defaults(multipole=True)
+args = parser.parse_args()
+
+# Make directory for inputs
+if args.output_dir is None:
+    if args.depleted:
+        directory = Path('core-long-depleted')
+    else:
+        directory = Path('core-long-fresh')
+else:
+    directory = args.output_dir
+directory.mkdir(exist_ok=True)
+
+# Modify lattice pitch
+smr.surfaces.lattice_pitch = lattice_pitch = 17*smr.surfaces.pin_pitch
+
+ring_radii = [0.1*pin_pitch, 0.2*pin_pitch]
+
+geometry = core_geometry(ring_radii, args.axial, args.depleted)
+
+h = active_fuel_length / args.axial
+fuel_mats = {}
+
+for cell in geometry.get_all_cells().values():
+    if cell.fill in materials:
+        # Determine volume of each fuel material
+        name = cell.fill.name
+        if 'UO2 Fuel' in name:
+            upper_right = cell.region.bounding_box[1][0]
+            if isclose(upper_right, ring_radii[0]):
+                ri, ro = 0.0, ring_radii[0]
+            elif isclose(upper_right, ring_radii[1]):
+                ri, ro = ring_radii[0], ring_radii[1]
+            else:
+                ri, ro = ring_radii[1], pellet_OR
+            if (name, ri) not in fuel_mats:
+                cell.fill = cell.fill.clone()
+                cell.fill.volume = pi * (ro*ro - ri*ri) * h
+                fuel_mats[name, ri] = cell.fill
+            else:
+                cell.fill = fuel_mats[name, ri]
+        elif name == 'Helium':
+            cell.fill.volume = pi * (clad_IR**2 - pellet_OR**2) * h
+        elif name == 'M5':
+            # Clad is not subdivided
+            cell.fill.volume = pi * (clad_OR**2 - clad_IR**2) * active_fuel_length
+        else:
+            cell.fill.volume = 1.0
+
+
+#### Create OpenMC "materials.xml" file
+all_materials = geometry.get_all_materials()
+materials = openmc.Materials(all_materials.values())
+materials.export_to_xml(str(directory / 'materials.xml'))
+
+#### Create OpenMC "geometry.xml" file
+geometry.export_to_xml(str(directory / 'geometry.xml'))
+
+
+#### Create OpenMC "settings.xml" file
+
+# 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 = 20_000_000
+settings.output = {'tallies': False, 'summary': False}
+settings.source = source
+settings.sourcepoint = {'write': False}
+settings.temperature = {
+    'default': inlet_temperature,
+    'method': 'interpolation',
+    'range': (300.0, 1500.0),
+}
+if args.multipole:
+    settings.temperature['multipole'] = True
+    settings.temperature['tolerance'] = 1000
+
+settings.export_to_xml(str(directory / 'settings.xml'))
+
+# Check assembly power distribution
+core_lattice = geometry.get_cells_by_fill_name('Main core')[0].fill
+mesh = openmc.RegularMesh.from_rect_lattice(core_lattice)
+assembly_power = openmc.Tally()
+assembly_power.filters = [openmc.MeshFilter(mesh)]
+assembly_power.scores = ['nu-fission']
+tallies = openmc.Tallies([assembly_power])
+tallies.export_to_xml(directory / 'tallies.xml')

smr/build-core-short.py

@@ -1,25 +1,18 @@
 #!/usr/bin/env python3
 
-import os
-import shutil
-import copy
 import argparse
 from math import pi, isclose
 from pathlib import Path
 
-import numpy as np
-
 import openmc
 from smr.materials import materials
-from smr.plots import core_plots
-from smr.surfaces import lattice_pitch, bottom_fuel_stack, top_active_core, \
+from smr.surfaces import bottom_fuel_stack, top_active_core, \
     pellet_OR, surfs, pin_pitch, clad_IR, clad_OR
 import smr.surfaces
 import smr.pins
 from smr.core import core_geometry
 from smr import inlet_temperature
 
-
 # Define command-line options
 parser = argparse.ArgumentParser()
 parser.add_argument('--multipole', action='store_true',
@@ -44,6 +37,9 @@ else:
     directory = args.output_dir
 directory.mkdir(exist_ok=True)
 
+# Modify lattice pitch
+smr.surfaces.lattice_pitch = lattice_pitch = 17*smr.surfaces.pin_pitch
+
 # Modify fuel length
 length = 3. * pin_pitch
 smr.surfaces.active_fuel_length = length
@@ -113,8 +109,7 @@ settings.inactive = 100
 settings.particles = 10000
 settings.output = {'tallies': False, 'summary': False}
 settings.source = source
-settings.sourcepoint_write = False
-
+settings.sourcepoint = {'write': False}
 settings.temperature = {
     'default': inlet_temperature,
     'method': 'interpolation',

smr/milestones.md

@@ -0,0 +1,20 @@
+# Milestone Models
+
+- **AD-SE-08-61, Coupled Multiphysics Driver Implementation** --- This milestone
+  used the singlerod short/long problems. Generating the model was done with the
+  script `tests/singlerod/make_openmc_model.py` from the ENRICO repository
+  (there is a `--short` command line option to generate the short version)
+
+- **AD-SE-08-66, Coupled Assembly Analysis** --- This milestone used the
+  assembly short and long (v2) problems. Generating the model was done with
+  `smr/build-assembly-long.py -a 100 --clone` on the ecp-benchmarks repository
+  (git commit `631fefe`, after pull request #14). In these models, materials are
+  fully differentiated across each fuel ring/axial segment.
+
+- **AD-SE-08-73, Full core coupled-physics simulation** --- This milestone used
+  the core-short and core-long (90 layer) models. Generating the models was done
+  with `smr/build-core-short.py` and `smr/build-assembly-long.py -a 90` on the
+  ecp-benchmarks repository (git commit `c7b89db`, after pull request #16). In
+  these models, materials are not differentiated and no grid spacers are
+  present. The lattice pitch is modified to be exactly 17 times the pin pitch
+  (slightly different than NuScale specification).

smr/smr/core.py

@@ -5,9 +5,9 @@ import numpy as np
 import openmc
 
 from .materials import mats
-from .surfaces import surfs, lattice_pitch
 from .reflector import reflector_universes
 from .assemblies import assembly_universes
+from smr import surfaces
 
 
 def core_geometry(ring_radii, num_axial, depleted):
@@ -34,6 +34,7 @@ def core_geometry(ring_radii, num_axial, depleted):
 
     # Construct main core lattice
     core = openmc.RectLattice(name='Main core')
+    lattice_pitch = surfaces.lattice_pitch
     core.lower_left = (-9*lattice_pitch/2, -9*lattice_pitch/2)
     core.pitch = (lattice_pitch, lattice_pitch)
     universes = np.tile(reflector['solid'], (9, 9))
@@ -119,6 +120,7 @@ def core_geometry(ring_radii, num_axial, depleted):
     core.universes = universes
 
     root_univ = openmc.Universe(universe_id=0, name='root universe')
+    surfs = surfaces.surfs
 
     # Cylinder filled with core lattice
     cell = openmc.Cell(name='Main core')

smr/smr/materials.py

@@ -1,5 +1,7 @@
 """Instantiate the OpenMC Materials needed by the core model."""
 
+import copy
+
 import openmc
 from openmc.data import atomic_weight, atomic_mass, water_density
 
@@ -252,3 +254,10 @@ mats['UO2 3.1 depleted'] = mat
 # Construct a collection of Materials to export to XML
 
 materials = openmc.Materials(mats.values())
+
+
+def clone(material):
+    """Perform copy of material but share nuclide densities"""
+    shared_mat = copy.copy(material)
+    shared_mat.id = None
+    return shared_mat

smr/smr/reflector.py

@@ -12,7 +12,7 @@ converted to actual dimensions by scaling according to the width of an assembly.
 import openmc
 
 from .materials import mats
-from .surfaces import lattice_pitch
+from smr import surfaces
 
 
 def make_reflector(name, parameters):
@@ -91,6 +91,7 @@ def reflector_universes():
 
     # All pixel widths are scaled according to the actual width of an assembly
     # divided by the width of an assembly in pixels
+    lattice_pitch = surfaces.lattice_pitch
     scale = lattice_pitch/width
 
     # Physical positions

smr/smr/surfaces.py

@@ -69,8 +69,7 @@ spacer_height = 1.750*INCHES  # ML17013A274, Figure 4.2-7
 # assembly parameters
 assembly_length   = 95.89*INCHES  # ML17013A274, Table 4.1-2
 pin_pitch         = 0.496*INCHES  # ML17013A274, Table 4.1-2
-#lattice_pitch     = 8.466*INCHES  # ML17013A274, Table 4.1-2
-lattice_pitch     = 17*pin_pitch
+lattice_pitch     = 8.466*INCHES  # ML17013A274, Table 4.1-2
 grid_strap_side   = 21.47270
 top_nozzle_height = 3.551*INCHES  # ML17013A274, Figure 4.2-2
 top_nozzle_width  = 8.406*INCHES  # ML17013A274, Figure 4.2-2