diff --git a/smr/build-assembly-long.py b/smr/build-assembly-long.py
index 623907c..79734c3 100644
--- a/smr/build-assembly-long.py
+++ b/smr/build-assembly-long.py
@@ -1,6 +1,7 @@
 #!/usr/bin/env python3
 
 import argparse
+import copy
 from math import pi, isclose
 from pathlib import Path
 
@@ -8,9 +9,10 @@ import numpy as np
 from tqdm import tqdm
 import openmc
 
-from smr.materials import materials, clone
+from smr.materials import materials, mats
 from smr.surfaces import surfs, lattice_pitch, pin_pitch, bottom_fuel_stack, \
-    top_active_core, pellet_OR, active_fuel_length
+    top_active_core, pellet_OR, clad_OR, clad_IR, guide_tube_IR, guide_tube_OR, \
+    active_fuel_length
 from smr.pins import pin_universes, make_stack
 
 
@@ -18,18 +20,14 @@ 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', dest='multipole', action='store_false',
+parser.add_argument('--no-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(clone=False, multipole=True)
+parser.set_defaults(multipole=True)
 args = parser.parse_args()
 
 # Make directory for inputs
@@ -103,22 +101,11 @@ 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='Differentiating materials / assigning volume'):
+                 desc='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 is_fuel:
+        if 'UO2 Fuel' in cell.fill.name:
             upper_right = cell.region.bounding_box[1]
             if isclose(upper_right[0], rings[0]):
                 ri, ro = 0.0, rings[0]
@@ -126,25 +113,14 @@ for cell in tqdm(geometry.get_all_material_cells().values(),
                 ri, ro = rings[0], rings[1]
             else:
                 ri, ro = rings[1], pellet_OR
-
-            if args.clone:
-                for mat in cell.fill:
-                    mat.volume = pi * (ro*ro - ri*ri) * h
+            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:
-                # 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]
+                cell.fill = fuel_mats[ri]
         else:
-            if args.clone:
-                for mat in cell.fill:
-                    mat.volume = 1.0
-            else:
-                cell.fill.volume = 1.0
+            cell.fill.volume = 1.0
 
 #### Create OpenMC "materials.xml" file
 print('Getting materials...')
@@ -174,13 +150,14 @@ settings.particles = 10000
 settings.output = {'tallies': False, 'summary': False}
 settings.source = source
 settings.sourcepoint = {'write': False}
-settings.temperature = {
-    'default': 531.5,
-    'method': 'interpolation',
-        'range': (500.0, 1300.0)
-}
+
 if args.multipole:
-    settings.temperature['multipole'] = True
-    settings.temperature['tolerance'] = 1000
+    settings.temperature = {
+        'multipole': True,
+        'tolerance': 1000,
+        'default': 531.5,
+        'method': 'interpolation',
+        'range': (500.0, 1300.0)
+    }
 
 settings.export_to_xml(str(directory / 'settings.xml'))
diff --git a/smr/build-assembly-short.py b/smr/build-assembly-short.py
index dc13456..c06c67b 100644
--- a/smr/build-assembly-short.py
+++ b/smr/build-assembly-short.py
@@ -1,6 +1,7 @@
 #!/usr/bin/env python3
 
 import argparse
+import copy
 from math import pi, isclose
 from pathlib import Path
 
@@ -8,7 +9,7 @@ import numpy as np
 from tqdm import tqdm
 import openmc
 
-from smr.materials import materials, clone
+from smr.materials import materials, mats
 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
@@ -84,6 +85,14 @@ 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':
diff --git a/smr/build-assembly.py b/smr/build-assembly.py
index d4bc166..392b114 100644
--- a/smr/build-assembly.py
+++ b/smr/build-assembly.py
@@ -1,15 +1,17 @@
 #!/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, clone
+from smr.materials import materials
 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
 
 
@@ -17,12 +19,8 @@ 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', dest='multipole', action='store_false',
+parser.add_argument('--no-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,
@@ -32,7 +30,7 @@ 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(clone=False, multipole=True)
+parser.set_defaults(multipole=True)
 args = parser.parse_args()
 
 # Make directory for inputs
@@ -51,17 +49,25 @@ if args.rings > 1:
 else:
     ring_radii = None
 assembly = assembly_universes(ring_radii, args.axial, args.depleted)
-lattice_sides = openmc.model.rectangular_prism(lattice_pitch, lattice_pitch,
-                                               boundary_type='reflective')
+lattice_sides = openmc.model.get_rectangular_prism(lattice_pitch, lattice_pitch,
+                                                   boundary_type='reflective')
 main_cell = openmc.Cell(
-    fill=assembly['Assembly (3.1%)'],
+    fill=assembly['Assembly (3.1%) 16BA'],
     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.clone:
+if args.tallies == 'mat':
     # Count the number of instances for each cell and material
     geometry.determine_paths(instances_only=True)
 
@@ -101,7 +107,7 @@ 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',
@@ -143,3 +149,7 @@ 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'))
diff --git a/smr/build-core-fresh.py b/smr/build-core-fresh.py
index 9747ae8..4265056 100644
--- a/smr/build-core-fresh.py
+++ b/smr/build-core-fresh.py
@@ -1,14 +1,17 @@
 #!/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 tqdm import tqdm
 
-from smr.materials import materials, clone
+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, \
     pellet_OR, active_fuel_length
 from smr.core import core_geometry
@@ -19,12 +22,8 @@ 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', dest='multipole', action='store_false',
+parser.add_argument('--no-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,
@@ -32,7 +31,7 @@ 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(clone=False, multipole=True)
+parser.set_defaults(multipole=True)
 args = parser.parse_args()
 
 # Make directory for inputs
@@ -54,36 +53,22 @@ 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 tqdm(geometry.get_all_cells().values(),
-                 desc='Differentiating materials / assigning volume'):
+for cell in geometry.get_all_cells().values():
     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 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
+        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
             else:
-                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]
+                cell.fill = fuel_mats[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())
@@ -108,7 +93,8 @@ 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',
@@ -119,3 +105,4 @@ if args.multipole:
     settings.temperature['tolerance'] = 1000
 
 settings.export_to_xml(str(directory / 'settings.xml'))
+
diff --git a/smr/build-core-long.py b/smr/build-core-long.py
deleted file mode 100644
index 4033d6e..0000000
--- a/smr/build-core-long.py
+++ /dev/null
@@ -1,118 +0,0 @@
-#!/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')
diff --git a/smr/build-core-short.py b/smr/build-core-short.py
index 2f5339d..317a8a0 100644
--- a/smr/build-core-short.py
+++ b/smr/build-core-short.py
@@ -1,18 +1,25 @@
 #!/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.surfaces import bottom_fuel_stack, top_active_core, \
+from smr.plots import core_plots
+from smr.surfaces import lattice_pitch, 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',
@@ -37,9 +44,6 @@ 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
@@ -109,7 +113,8 @@ 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',
diff --git a/smr/milestones.md b/smr/milestones.md
deleted file mode 100644
index bb89864..0000000
--- a/smr/milestones.md
+++ /dev/null
@@ -1,20 +0,0 @@
-# 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).
diff --git a/smr/smr/core.py b/smr/smr/core.py
index 81763fe..68f3275 100644
--- a/smr/smr/core.py
+++ b/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):
@@ -34,7 +34,6 @@ 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))
@@ -120,7 +119,6 @@ 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')
diff --git a/smr/smr/materials.py b/smr/smr/materials.py
index ee6c754..72a39d6 100644
--- a/smr/smr/materials.py
+++ b/smr/smr/materials.py
@@ -1,7 +1,5 @@
 """Instantiate the OpenMC Materials needed by the core model."""
 
-import copy
-
 import openmc
 from openmc.data import atomic_weight, atomic_mass, water_density
 
@@ -254,10 +252,3 @@ 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
diff --git a/smr/smr/reflector.py b/smr/smr/reflector.py
index 74aa4bd..ed143cc 100644
--- a/smr/smr/reflector.py
+++ b/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 smr import surfaces
+from .surfaces import lattice_pitch
 
 
 def make_reflector(name, parameters):
@@ -91,7 +91,6 @@ 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
diff --git a/smr/smr/surfaces.py b/smr/smr/surfaces.py
index b454c61..d91ec92 100644
--- a/smr/smr/surfaces.py
+++ b/smr/smr/surfaces.py
@@ -69,7 +69,8 @@ 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     = 8.466*INCHES  # ML17013A274, Table 4.1-2
+lattice_pitch     = 17*pin_pitch
 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