Merge pull request #736 from QingmingHe/no_nu_settings

No nu option like nonu in MCNP

docs/source/usersguide/input.rst

@@ -872,6 +872,19 @@ displayed. This element takes the following attributes:
 
     *Default*: 5
 
+``<create_fission_neutrons>`` Element
+-------------------------------------
+
+The ``<create_fission_neutrons>`` element indicates whether fission neutrons
+should be created or not.  If this element is set to "true", fission neutrons
+will be created; otherwise the fission is treated as capture and no fission
+neutron will be created. Note that this option is only applied to fixed source
+calculation. For eigenvalue calculation, fission will always be treated as real
+fission.
+
+  *Default*: true
+
+
 ``<volume_calc>`` Element
 -------------------------
 

openmc/settings.py

@@ -141,6 +141,8 @@ class Settings(object):
         The elastic scattering model to use for resonant isotopes
     volume_calculations : VolumeCalculation or iterable of VolumeCalculation
         Stochastic volume calculation specifications
+    create_fission_neutrons : bool
+        Indicate whether fission neutrons should be created or not.
 
     """
 
@@ -227,6 +229,8 @@ class Settings(object):
         self._volume_calculations = cv.CheckedList(
             VolumeCalculation, 'volume calculations')
 
+        self._create_fission_neutrons = None
+
     @property
     def run_mode(self):
         return self._run_mode
@@ -427,6 +431,10 @@ class Settings(object):
     def volume_calculations(self):
         return self._volume_calculations
 
+    @property
+    def create_fission_neutrons(self):
+        return self._create_fission_neutrons
+
     @run_mode.setter
     def run_mode(self, run_mode):
         if run_mode not in ['eigenvalue', 'fixed source']:
@@ -826,6 +834,12 @@ class Settings(object):
         self._volume_calculations = cv.CheckedList(
             VolumeCalculation, 'stochastic volume calculations', vol_calcs)
 
+    @create_fission_neutrons.setter
+    def create_fission_neutrons(self, create_fission_neutrons):
+        cv.check_type('Whether create fission neutrons',
+                      create_fission_neutrons, bool)
+        self._create_fission_neutrons = create_fission_neutrons
+
     def _create_run_mode_subelement(self):
 
         if self.run_mode == 'eigenvalue':
@@ -1121,6 +1135,11 @@ class Settings(object):
             for r in self.resonance_scattering:
                 elem.append(r.to_xml_element())
 
+    def _create_create_fission_neutrons_subelement(self):
+        if self._create_fission_neutrons is not None:
+            elem = ET.SubElement(self._settings_file, "create_fission_neutrons")
+            elem.text = str(self._create_fission_neutrons).lower()
+
     def export_to_xml(self, path='settings.xml'):
         """Export simulation settings to an XML file.
 
@@ -1164,6 +1183,7 @@ class Settings(object):
         self._create_dd_subelement()
         self._create_resonance_scattering_subelement()
         self._create_volume_calcs_subelement()
+        self._create_create_fission_neutrons_subelement()
 
         # Clean the indentation in the file to be user-readable
         clean_xml_indentation(self._settings_file)

src/global.F90

@@ -352,6 +352,9 @@ module global
   ! Write out initial source
   logical :: write_initial_source = .false.
 
+  ! Whether create fission neutrons or not. Only applied for MODE_FIXEDSOURCE
+  logical :: create_fission_neutrons = .true.
+
   ! ============================================================================
   ! CMFD VARIABLES
 

src/input_xml.F90

@@ -1102,6 +1102,19 @@ contains
       end select
     end if
 
+    ! Check whether create fission sites
+    if (run_mode == MODE_FIXEDSOURCE) then
+      if (check_for_node(doc, "create_fission_neutrons")) then
+        call get_node_value(doc, "create_fission_neutrons", temp_str)
+        temp_str = to_lower(temp_str)
+        if (trim(temp_str) == 'true' .or. trim(temp_str) == '1') then
+          create_fission_neutrons = .true.
+        else if (trim(temp_str) == 'false' .or. trim(temp_str) == '0') then
+          create_fission_neutrons = .false.
+        end if
+      end if
+    end if
+
     ! Close settings XML file
     call close_xmldoc(doc)
 

src/physics.F90

@@ -96,10 +96,11 @@ contains
     ! absorption (including fission)
 
     if (nuc % fissionable) then
-      call sample_fission(i_nuclide, i_reaction)
       if (run_mode == MODE_EIGENVALUE) then
+        call sample_fission(i_nuclide, i_reaction)
         call create_fission_sites(p, i_nuclide, i_reaction, fission_bank, n_bank)
-      elseif (run_mode == MODE_FIXEDSOURCE) then
+      elseif (run_mode == MODE_FIXEDSOURCE .and. create_fission_neutrons) then
+        call sample_fission(i_nuclide, i_reaction)
         call create_fission_sites(p, i_nuclide, i_reaction, &
              p % secondary_bank, p % n_secondary)
       end if

src/physics_mg.F90

@@ -73,7 +73,7 @@ contains
     if (mat % fissionable) then
       if (run_mode == MODE_EIGENVALUE) then
         call create_fission_sites(p, fission_bank, n_bank)
-      elseif (run_mode == MODE_FIXEDSOURCE) then
+      elseif (run_mode == MODE_FIXEDSOURCE .and. create_fission_neutrons) then
         call create_fission_sites(p, p % secondary_bank, p % n_secondary)
       end if
     end if

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()