Added test for distribcell tallies in an asymmetric lattice

tests/test_asymmetric_lattice/inputs_true.dat

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+dd39c0ae6327e6e74cb077d56e37c112611b95c4c10d96203e672b3e7f928211cc991ec7ebbf9eeadabd968dcdcb651b250233169b62d43ef6994ab9a46cb34a

tests/test_asymmetric_lattice/results_true.dat

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+7e75ad5b7979e65e52ce564bfbd8fac819013ef8ba35fe184569b452dd9a1ba98d267b6e33d357fdd1c943f201125ff8a4f8601147b87036525870528063dcae

tests/test_asymmetric_lattice/test_asymmetric_lattice.py

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+#!/usr/bin/env python
+
+import os
+import sys
+import glob
+import hashlib
+sys.path.insert(0, os.pardir)
+from testing_harness import PyAPITestHarness
+import openmc
+from openmc.source import Source
+from openmc.stats import Box
+
+
+class AsymmetricLatticeTestHarness(PyAPITestHarness):
+
+    def _build_inputs(self):
+        """Build an axis-asymmetric lattice of fuel assemblies"""
+
+        # Build full core geometry from underlying input set
+        self._input_set.build_default_materials_and_geometry()
+
+        # Extract all universes from the full core geometry
+        geometry = self._input_set.geometry.geometry
+        all_univs = geometry.get_all_universes()
+        print(all_univs.keys())
+
+        # Extract universes encapsulating fuel and water assemblies
+        water = all_univs[7]
+        fuel = all_univs[8]
+
+        # Construct a 3x3 lattice of fuel assemblies
+        core_lat = openmc.RectLattice(name='3x3 Core Lattice', lattice_id=202)
+        core_lat.dimension = (3, 3)
+        core_lat.lower_left = (-32.13, -32.13)
+        core_lat.pitch = (21.42, 21.42)
+        core_lat.universes = [[fuel,  water, water],
+                              [fuel,  fuel,  fuel],
+                              [water, water, water]]
+
+        # Create bounding surfaces
+        min_x = openmc.XPlane(x0=-32.13, boundary_type='reflective')
+        max_x = openmc.XPlane(x0=+32.13, boundary_type='reflective')
+        min_y = openmc.YPlane(y0=-32.13, boundary_type='reflective')
+        max_y = openmc.YPlane(y0=+32.13, boundary_type='reflective')
+        min_z = openmc.ZPlane(z0=0, boundary_type='reflective')
+        max_z = openmc.ZPlane(z0=+32.13, boundary_type='reflective')
+
+        # Define root universe
+        root_univ = openmc.Universe(universe_id=0, name='root universe')
+        root_cell = openmc.Cell(cell_id=1)
+        root_cell.region = +min_x & -max_x & +min_y & -max_y & +min_z & -max_z
+        root_cell.fill = core_lat
+        root_univ.add_cell(root_cell)
+
+        # Over-ride geometry in the input set with this 3x3 lattice
+        self._input_set.geometry.geometry.root_universe = root_univ
+
+        # Initialize a "distribcell" filter for the cold fuel pin cell
+        distrib_filter = openmc.Filter(type='distribcell', bins=[27])
+
+        # Initialize the tallies
+        tally = openmc.Tally(name='distribcell tally', tally_id=27)
+        tally.add_filter(distrib_filter)
+        tally.add_score('nu-fission')
+
+        # Initialize the tallies file
+        tallies_file = openmc.TalliesFile()
+        tallies_file.add_tally(tally)
+
+        # Assign the tallies file to the input set
+        self._input_set.tallies = tallies_file
+
+        # Specify summary output and correct source sampling box
+        self._input_set.build_default_settings()
+        self._input_set.settings.output = {'summary': True}
+        self._input_set.settings.source = Source(space=Box(
+            [0, 0, 0], [32.13, 32.13, 32.13]))
+
+        # Write input XML files
+        self._input_set.export()
+
+    def _get_results(self, hash_output=True):
+        """Digest info in statepoint and summary and return as a string."""
+
+        # Read the statepoint file
+        statepoint = glob.glob(os.path.join(os.getcwd(), self._sp_name))[0]
+        sp = openmc.StatePoint(statepoint)
+
+        # Read the summary file
+        summary = glob.glob(os.path.join(os.getcwd(), 'summary.h5'))[0]
+        su = openmc.Summary(summary)
+        sp.link_with_summary(su)
+
+        # Extract the tally of interest
+        tally = sp.get_tally(name='distribcell tally')
+
+        # Create a string of all mean, std. dev. values for both tallies
+        outstr = ''
+        outstr += ', '.join(map(str, tally.mean.flatten())) + '\n'
+        outstr += ', '.join(map(str, tally.std_dev.flatten())) + '\n'
+
+        # Extract fuel assembly lattices from the summary
+        all_cells = su.openmc_geometry.get_all_cells()
+        fuel = all_cells[80].fill
+        core = all_cells[1].fill
+
+        # Append a string of lattice distribcell offsets to the string
+        outstr += ', '.join(map(str, fuel.offsets.flatten())) + '\n'
+        outstr += ', '.join(map(str, core.offsets.flatten())) + '\n'
+
+        # Hash the results if necessary
+        if hash_output:
+            sha512 = hashlib.sha512()
+            sha512.update(outstr.encode('utf-8'))
+            outstr = sha512.hexdigest()
+
+        return outstr
+
+    def _cleanup(self):
+        super(AsymmetricLatticeTestHarness, self)._cleanup()
+        f = os.path.join(os.getcwd(), 'tallies.xml')
+        if os.path.exists(f): os.remove(f)
+
+
+if __name__ == '__main__':
+    harness = AsymmetricLatticeTestHarness('statepoint.10.h5', True)
+    harness.main()