add a create_fission_neutrons unit test

tests/test_create_fission_neutrons/inputs_true.dat

@@ -0,0 +1 @@
+c3581501d1486293c255390251d20584431a198fbb7c07dc2879dc95dc96e26786d3913ce0db8007f542468fd137ff3267824844c03b4687fdad81ea568c92d7

tests/test_create_fission_neutrons/results_true.dat

@@ -0,0 +1,3 @@
+tally 1:
+sum = 2.056839E+02
+sum_sq = 4.244628E+03

tests/test_create_fission_neutrons/test_create_fission_neutrons.py

@@ -0,0 +1,80 @@
+#!/usr/bin/env python
+
+import os
+import sys
+sys.path.insert(0, os.pardir)
+from testing_harness import PyAPITestHarness
+import openmc
+
+
+class CreateFissionNeutronsTestHarness(PyAPITestHarness):
+    def _build_inputs(self):
+        # Material is composed of H-1 and U-235
+        mat = openmc.Material(material_id=1, name='mat')
+        mat.set_density('atom/b-cm', 0.069335)
+        mat.add_nuclide('H1', 40.0)
+        mat.add_nuclide('U235', 1.0)
+        materials_file = openmc.Materials([mat])
+        materials_file.export_to_xml()
+
+        # Cell is box with reflective boundary
+        x1 = openmc.XPlane(surface_id=1, x0=-1)
+        x2 = openmc.XPlane(surface_id=2, x0=1)
+        y1 = openmc.YPlane(surface_id=3, y0=-1)
+        y2 = openmc.YPlane(surface_id=4, y0=1)
+        z1 = openmc.ZPlane(surface_id=5, z0=-1)
+        z2 = openmc.ZPlane(surface_id=6, z0=1)
+        for surface in [x1, x2, y1, y2, z1, z2]:
+            surface.boundary_type = 'reflective'
+        box = openmc.Cell(cell_id=1, name='box')
+        box.region = +x1 & -x2 & +y1 & -y2 & +z1 & -z2
+        box.fill = mat
+        root = openmc.Universe(universe_id=0, name='root universe')
+        root.add_cell(box)
+        geometry = openmc.Geometry(root)
+        geometry.export_to_xml()
+
+        # Set the running parameters
+        settings_file = openmc.Settings()
+        settings_file.run_mode = 'fixed source'
+        settings_file.batches = 10
+        settings_file.particles = 100
+        settings_file.create_fission_neutrons = False
+        bounds = [-1, -1, -1, 1, 1, 1]
+        uniform_dist = openmc.stats.Box(bounds[:3], bounds[3:])
+        watt_dist = openmc.stats.Watt()
+        settings_file.source = openmc.source.Source(space=uniform_dist,
+                                                    energy=watt_dist)
+        settings_file.export_to_xml()
+
+        # Create tallies
+        tallies = openmc.Tallies()
+        tally = openmc.Tally(1)
+        tally.scores = ['flux']
+        tallies.append(tally)
+        tallies.export_to_xml()
+
+    def _get_results(self):
+        """Digest info in the statepoint and return as a string."""
+        # Read the statepoint file.
+        sp = openmc.StatePoint(self._sp_name)
+
+        # Write out tally data.
+        outstr = ''
+        t = sp.get_tally()
+        outstr += 'tally {0}:\n'.format(t.id)
+        outstr += 'sum = {0:12.6E}\n'.format(t.sum[0, 0, 0])
+        outstr += 'sum_sq = {0:12.6E}\n'.format(t.sum_sq[0, 0, 0])
+
+        return outstr
+
+    def _cleanup(self):
+        super(CreateFissionNeutronsTestHarness, self)._cleanup()
+        f = os.path.join(os.getcwd(), 'tallies.xml')
+        if os.path.exists(f):
+            os.remove(f)
+
+
+if __name__ == '__main__':
+    harness = CreateFissionNeutronsTestHarness('statepoint.10.h5', True)
+    harness.main()