support total nuclide

Author: GuySten

openmc/model/model.py

@@ -235,7 +235,7 @@ def _get_all_materials(self) -> dict[int, openmc.Material]:
 
         return materials
 
-    def add_kinetics_parameters_tallies(self, num_groups: int | None = None, nuclides: Sequence[str] | None = None):
+    def add_kinetics_parameters_tallies(self, num_groups: int | None = None, nuclides: Sequence[str] = ["total"]):
         """Add tallies for calculating kinetics parameters using the IFP method.
 
         This method adds tallies to the model for calculating two kinetics
@@ -249,9 +249,9 @@ def add_kinetics_parameters_tallies(self, num_groups: int | None = None, nuclide
             Number of precursor groups to filter the delayed neutron fraction.
             If None, only the total effective delayed neutron fraction is
             tallied.
-        nuclides : int, optional 
+        nuclides : Sequence[str]
             Nuclides to calculate separate kinetic parameters for.
-            If None, do not separate kinetic parameters per nuclide.
+            Defaults to ["total"].
 
         """
         if not any('ifp-time-numerator' in t.scores for t in self.tallies):
@@ -261,10 +261,9 @@ def add_kinetics_parameters_tallies(self, num_groups: int | None = None, nuclide
         if not any('ifp-beta-numerator' in t.scores for t in self.tallies):
             beta_tally = openmc.Tally(name='IFP beta numerator')
             beta_tally.scores = ['ifp-beta-numerator']
+            beta_tally.nuclides = list(nuclides)
             if num_groups is not None:
                 beta_tally.filters = [openmc.DelayedGroupFilter(list(range(1, num_groups + 1)))]
-            if nuclides:
-                beta_tally.nuclides = list(nuclides)
             self.tallies.append(beta_tally)
         if not any('ifp-denominator' in t.scores for t in self.tallies):
             denom_tally = openmc.Tally(name='IFP denominator')

src/tallies/tally.cpp

@@ -235,9 +235,12 @@ Tally::Tally(pugi::xml_node node)
         }
         break;
       case SCORE_IFP_BETA_NUM:
-        // This bool indicates if the ancestor_nuclide vector has to be filled
-        if (nuclides_[0] > -1) {
-          settings::ifp_beta_nuclide = true;
+        for (int i_nuc : nuclides_) {
+          if (i_nuc > -1) {
+            // This bool indicates if the ancestor_nuclide vector has to be
+            // filled
+            settings::ifp_beta_nuclide = true;
+          }
         }
       case SCORE_IFP_DENOM:
         if (settings::ifp_parameter == IFPParameter::None) {

src/tallies/tally_scoring.cpp

@@ -963,9 +963,9 @@ void score_general_ce_nonanalog(Particle& p, int i_tally, int start_index,
               simulation::ifp_source_delayed_group_bank[p.current_work() - 1];
             if (delayed_groups.size() == settings::ifp_n_generation) {
               int j = -1;
-              if (delayed_groups[0] > 0 && tally.nuclides_[0] == -1)
+              if (delayed_groups[0] > 0 && i_nuclide == -1)
                 j = 0;
-              if (delayed_groups[1] > 0 && tally.nuclides_[0] > -1) {
+              if (delayed_groups[1] > 0 && i_nuclide > -1) {
                 const auto& ancestor_event_nuclide = simulation::
                   ifp_source_ancestor_nuclide_bank[p.current_work() - 1];
                 if (ancestor_event_nuclide[0] == i_nuclide)

tests/regression_tests/ifp/nuclidewise/inputs_true.dat

@@ -35,7 +35,7 @@
     </tally>
     <tally id="2" name="IFP beta numerator">
       <filters>1</filters>
-      <nuclides>U235 Pu239</nuclides>
+      <nuclides>U235 Pu239 total</nuclides>
       <scores>ifp-beta-numerator</scores>
     </tally>
     <tally id="3" name="IFP denominator">

tests/regression_tests/ifp/nuclidewise/results_true.dat

@@ -8,26 +8,38 @@ tally 2:
 4.295000E-07
 9.400000E-04
 1.794000E-07
+2.780000E-03
+8.896000E-07
 5.970000E-03
 2.607900E-06
 4.550000E-03
 1.625100E-06
+9.460000E-03
+7.239800E-06
 5.070000E-03
 2.047900E-06
 2.600000E-03
 5.692000E-07
+7.420000E-03
+4.387800E-06
 1.236000E-02
 1.062980E-05
 4.020000E-03
 1.379200E-06
+1.506000E-02
+1.709240E-05
 5.690000E-03
 3.302300E-06
 3.180000E-03
 8.316000E-07
+7.230000E-03
+4.155100E-06
 2.100000E-03
 4.486000E-07
 1.300000E-03
 2.008000E-07
+3.040000E-03
+9.300000E-07
 tally 3:
 1.495260E+01
 1.490559E+01

tests/regression_tests/ifp/nuclidewise/test.py

@@ -32,7 +32,7 @@ def ifp_model():
     space = openmc.stats.Box(*cell.bounding_box)
     model.settings.source = openmc.IndependentSource(
         space=space, constraints={'fissionable': True})
-    model.add_kinetics_parameters_tallies(num_groups=6, nuclides = ["U235", "Pu239"])
+    model.add_kinetics_parameters_tallies(num_groups=6, nuclides = ["U235", "Pu239", "total"])
     return model