Add SMR assembly model

.gitattributes

@@ -2,3 +2,4 @@ assembly/depleted/materials.xml filter=lfs diff=lfs merge=lfs -text
 2x2-periodic/depleted/materials.xml filter=lfs diff=lfs merge=lfs -text
 2x2-reflector/depleted/materials.xml filter=lfs diff=lfs merge=lfs -text
 smr/depleted/materials.xml filter=lfs diff=lfs merge=lfs -text
+smr/assembly/materials.xml filter=lfs diff=lfs merge=lfs -text

smr/assembly/settings.xml

@@ -0,0 +1,16 @@
+<?xml version='1.0' encoding='utf-8'?>
+<settings>
+  <run_mode>eigenvalue</run_mode>
+  <particles>10000</particles>
+  <batches>200</batches>
+  <inactive>100</inactive>
+  <source strength="1.0">
+    <space type="fission">
+      <parameters>-10.751819999999999 -10.751819999999999 36.007 10.751819999999999 10.751819999999999 236.0066</parameters>
+    </space>
+  </source>
+  <output>
+    <tallies>false</tallies>
+    <summary>false</summary>
+  </output>
+</settings>

smr/build-assembly.py

@@ -0,0 +1,123 @@
+#!/usr/bin/env python3
+
+import os
+import shutil
+import copy
+import argparse
+
+import numpy as np
+
+import openmc
+from smr.materials import materials
+from smr.surfaces import surfs, lattice_pitch, bottom_fuel_stack, top_active_core
+from smr.assemblies import assembly_universes
+
+
+# Define command-line options
+parser = argparse.ArgumentParser()
+parser.add_argument('-m', '--multipole', action='store_true',
+                    help='Whether to use multipole cross sections')
+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,
+                    help='Number of annular regions in fuel')
+parser.add_argument('-a', '--axial', type=int, default=196,
+                    help='Number of axial subdivisions in fuel')
+args = parser.parse_args()
+
+
+# Define geometry with a single assembly
+assembly = assembly_universes(args.rings, args.axial)
+lattice_sides = openmc.model.get_rectangular_prism(lattice_pitch, lattice_pitch,
+                                                   boundary_type='reflective')
+main_cell = openmc.Cell(
+    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)
+
+#### "Differentiate" the geometry if using distribmats
+if args.tallies == 'mat':
+    # Count the number of instances for each cell and material
+    geometry.determine_paths(instances_only=True)
+
+    # Extract all cells filled by a fuel material
+    fuel_mats = {m for m in materials if 'UO2 Fuel' in m.name}
+
+    for cell in geometry.get_all_material_cells().values():
+        if cell.fill in fuel_mats:
+            # Fill cell with list of "differentiated" materials
+            cell.fill = [cell.fill.clone() for i in range(cell.num_instances)]
+
+#### Create OpenMC "materials.xml" file
+all_materials = geometry.get_all_materials()
+materials = openmc.Materials(all_materials.values())
+materials.export_to_xml()
+
+
+#### Create OpenMC "geometry.xml" file
+geometry.export_to_xml()
+
+
+#### Create OpenMC "settings.xml" file
+
+# Construct uniform initial source distribution over fissionable zones
+lower_left = (-lattice_pitch/2, -lattice_pitch/2, bottom_fuel_stack)
+upper_right = (lattice_pitch/2, 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, 'summary': False}
+settings.source = source
+settings.sourcepoint_write = False
+
+if args.multipole:
+    settings.temperature = {'multipole': True, 'tolerance': 1000}
+
+settings.export_to_xml()
+
+
+####  Create OpenMC "tallies.xml" file
+tallies = openmc.Tallies()
+
+# Extract all fuel materials
+materials = geometry.get_materials_by_name(name='Fuel', matching=False)
+
+# If using distribcells, create distribcell tally needed for depletion
+if args.tallies == 'cell':
+    # Extract all cells filled by a fuel material
+    fuel_cells = []
+    for cell in geometry.get_all_cells().values():
+        if cell.fill in materials:
+            tally = openmc.Tally(name='depletion tally')
+            tally.scores = ['(n,p)', '(n,a)', '(n,gamma)',
+                            'fission', '(n,2n)', '(n,3n)', '(n,4n)']
+            tally.nuclides = cell.fill.get_nuclides()
+            tally.filters.append(openmc.DistribcellFilter([cell]))
+            tallies.append(tally)
+
+# If using distribmats, create material tally needed for depletion
+elif args.tallies == 'mat':
+    tally = openmc.Tally(name='depletion tally')
+    tally.scores = ['(n,p)', '(n,a)', '(n,gamma)',
+                    'fission', '(n,2n)', '(n,3n)', '(n,4n)']
+    tally.nuclides = materials[0].get_nuclides()
+    tally.filters = [openmc.MaterialFilter(materials)]
+    tallies.append(tally)
+
+tallies.export_to_xml()
+
+#### Move all XML files to 'fresh' directory
+
+if not os.path.exists('assembly'):
+    os.makedirs('assembly')
+
+shutil.move('materials.xml', 'assembly/materials.xml')
+shutil.move('geometry.xml', 'assembly/geometry.xml')
+shutil.move('settings.xml', 'assembly/settings.xml')
+shutil.move('tallies.xml', 'assembly/tallies.xml')

smr/build-core-depleted.py

@@ -64,7 +64,7 @@ settings.entropy_dimension = [15, 15, 1]
 # MeV/second cm from CASMO
 settings.power = 2.337e15 * ((17.*17.*37.) / 1.5**2) * height
 settings.dt_vec = dt
-settings.output_dir = 'depleted'
+settings.output_dir = 'core-depleted'
 
 op = opendeplete.OpenMCOperator(geometry, settings)
 

smr/build-core-fresh.py

@@ -111,11 +111,11 @@ tallies.export_to_xml()
 
 #### Move all XML files to 'fresh' directory
 
-if not os.path.exists('fresh'):
-    os.makedirs('fresh')
+if not os.path.exists('core-fresh'):
+    os.makedirs('core-fresh')
 
-shutil.move('materials.xml', 'fresh/materials.xml')
-shutil.move('geometry.xml', 'fresh/geometry.xml')
-shutil.move('settings.xml', 'fresh/settings.xml')
-shutil.move('tallies.xml', 'fresh/tallies.xml')
-shutil.move('plots.xml', 'fresh/plots.xml')
+shutil.move('materials.xml', 'core-fresh/materials.xml')
+shutil.move('geometry.xml', 'core-fresh/geometry.xml')
+shutil.move('settings.xml', 'core-fresh/settings.xml')
+shutil.move('tallies.xml', 'core-fresh/tallies.xml')
+shutil.move('plots.xml', 'core-fresh/plots.xml')