refactoring

include/openmc/constants.h

@@ -108,7 +108,7 @@ constexpr std::array<const char*, 39> SUBSHELLS =  {
 // Void material and nuclide
 // TODO: refactor and remove
 constexpr int MATERIAL_VOID {-1};
-constexpr int NUCLIDE_VOID  {-1};
+constexpr int NUCLIDE_NONE  {-1};
 
 // ============================================================================
 // CROSS SECTION RELATED CONSTANTS
@@ -128,7 +128,7 @@ constexpr int TEMPERATURE_INTERPOLATION {2};
 
 // Reaction types
 // TODO: Convert to enum
-constexpr int NONE_REACTION {0};
+constexpr int REACTION_NONE {0};
 constexpr int TOTAL_XS {1};
 constexpr int ELASTIC {2};
 constexpr int N_NONELASTIC {3};

include/openmc/particle.h

@@ -261,7 +261,7 @@ public:
 
   // Post-collision physical data
   int n_bank_ {0};        //!< number of fission sites banked
-  int n_bank_second_ {0}; //!< number of secondary particles
+  int n_bank_second_ {0}; //!< number of secondary particles banked
   double wgt_bank_ {0.0}; //!< weight of fission sites banked
   int n_delayed_bank_[MAX_DELAYED_GROUPS];  //!< number of delayed fission
                                             //!< sites banked

include/openmc/physics.h

@@ -44,7 +44,7 @@ void sample_positron_reaction(Particle* p);
 //!
 //! \param[in] p Particle
 //! \return Index in the data::nuclides vector
-int sample_nuclide(Particle* p);
+int sample_nuclide(const Particle* p);
 
 //! Determine the average total, prompt, and delayed neutrons produced from
 //! fission and creates appropriate bank sites.

src/particle.cpp

@@ -161,8 +161,8 @@ Particle::transport()
 
     // Reset event variables
     event_ = EVENT_KILL;
-    event_nuclide_ = NUCLIDE_VOID;
-    event_mt_ = NONE_REACTION;
+    event_nuclide_ = NUCLIDE_NONE;
+    event_mt_ = REACTION_NONE;
 
     // If the cell hasn't been determined based on the particle's location,
     // initiate a search for the current cell. This generally happens at the

src/physics.cpp

@@ -81,6 +81,9 @@ void sample_neutron_reaction(Particle* p)
   // Sample a nuclide within the material
   int i_nuclide = sample_nuclide(p);
 
+  // Save which nuclide particle had collision with
+  p->event_nuclide_ = i_nuclide;
+
   // Create fission bank sites. Note that while a fission reaction is sampled,
   // it never actually "happens", i.e. the weight of the particle does not
   // change when sampling fission sites. The following block handles all
@@ -394,7 +397,7 @@ void sample_positron_reaction(Particle* p)
   p->event_ = EVENT_ABSORB;
 }
 
-int sample_nuclide(Particle* p)
+int sample_nuclide(const Particle* p)
 {
   // Sample cumulative distribution function
   double cutoff = prn() * p->macro_xs_.total;
@@ -411,11 +414,7 @@ int sample_nuclide(Particle* p)
 
     // Increment probability to compare to cutoff
     prob += atom_density * p->neutron_xs_[i_nuclide].total;
-    if (prob >= cutoff) {
-      // Save which nuclide particle had collision with
-      p->event_nuclide_ = i_nuclide;
-      return i_nuclide;
-    }
+    if (prob >= cutoff) return i_nuclide;
   }
 
   // If we reach here, no nuclide was sampled
@@ -443,7 +442,7 @@ int sample_element(Particle* p)
     // Increment probability to compare to cutoff
     prob += sigma;
     if (prob > cutoff) {
-      // Save which nuclide particle had collision with
+      // Save which nuclide particle had collision with for tally purpose
       p->event_nuclide_ = mat->nuclide_[i];
 
       return i_element;