Use references consistently in physics modules

include/openmc/physics.h

@@ -16,55 +16,55 @@ namespace openmc {
 //==============================================================================
 
 //! Sample a nuclide and reaction and then calls the appropriate routine
-void collision(Particle* p);
+void collision(Particle& p);
 
 //! Samples an incident neutron reaction
-void sample_neutron_reaction(Particle* p);
+void sample_neutron_reaction(Particle& p);
 
 //! Samples an element based on the macroscopic cross sections for each nuclide
 //! within a material and then samples a reaction for that element and calls the
 //! appropriate routine to process the physics.
-void sample_photon_reaction(Particle* p);
+void sample_photon_reaction(Particle& p);
 
 //! Terminates the particle and either deposits all energy locally
 //! (electron_treatment = ElectronTreatment::LED) or creates secondary bremsstrahlung
 //! photons from electron deflections with charged particles (electron_treatment
 //! = ElectronTreatment::TTB).
-void sample_electron_reaction(Particle* p);
+void sample_electron_reaction(Particle& p);
 
 //! Terminates the particle and either deposits all energy locally
 //! (electron_treatment = ElectronTreatment::LED) or creates secondary bremsstrahlung
 //! photons from electron deflections with charged particles (electron_treatment
 //! = ElectronTreatment::TTB). Two annihilation photons of energy MASS_ELECTRON_EV (0.511
 //! MeV) are created and travel in opposite directions.
-void sample_positron_reaction(Particle* p);
+void sample_positron_reaction(Particle& p);
 
 //! Sample a nuclide based on their total cross sections and densities within
 //! the current material
 //!
 //! \param[in] p Particle
 //! \return Index in the data::nuclides vector
-int sample_nuclide(Particle* p);
+int sample_nuclide(Particle& p);
 
 //! Determine the average total, prompt, and delayed neutrons produced from
 //! fission and creates appropriate bank sites.
-void create_fission_sites(Particle* p, int i_nuclide, const Reaction* rx);
+void create_fission_sites(Particle& p, int i_nuclide, const Reaction* rx);
 
-int sample_element(Particle* p);
+int sample_element(Particle& p);
 
-Reaction* sample_fission(int i_nuclide, Particle* p);
+Reaction* sample_fission(int i_nuclide, Particle& p);
 
-void sample_photon_product(int i_nuclide, Particle* p, int* i_rx, int* i_product);
+void sample_photon_product(int i_nuclide, Particle& p, int* i_rx, int* i_product);
 
-void absorption(Particle* p, int i_nuclide);
+void absorption(Particle& p, int i_nuclide);
 
-void scatter(Particle* p, int i_nuclide);
+void scatter(Particle& p, int i_nuclide);
 
 //! Treats the elastic scattering of a neutron with a target.
 void elastic_scatter(int i_nuclide, const Reaction& rx, double kT,
-  Particle* p);
+  Particle& p);
 
-void sab_scatter(int i_nuclide, int i_sab, Particle* p);
+void sab_scatter(int i_nuclide, int i_sab, Particle& p);
 
 //! samples the target velocity. The constant cross section free gas model is
 //! the default method. Methods for correctly accounting for the energy
@@ -85,9 +85,9 @@ void sample_fission_neutron(int i_nuclide, const Reaction* rx, double E_in,
 
 //! handles all reactions with a single secondary neutron (other than fission),
 //! i.e. level scattering, (n,np), (n,na), etc.
-void inelastic_scatter(const Nuclide* nuc, const Reaction* rx, Particle* p);
+void inelastic_scatter(const Nuclide* nuc, const Reaction* rx, Particle& p);
 
-void sample_secondary_photons(Particle* p, int i_nuclide);
+void sample_secondary_photons(Particle& p, int i_nuclide);
 
 } // namespace openmc
 

include/openmc/physics_common.h

@@ -9,8 +9,7 @@
 namespace openmc {
 
 //! \brief Performs the russian roulette operation for a particle
-extern "C" void
-russian_roulette(Particle* p);
+void russian_roulette(Particle& p);
 
 } // namespace openmc
 #endif // OPENMC_PHYSICS_COMMON_H

include/openmc/physics_mg.h

@@ -15,7 +15,7 @@ namespace openmc {
 //! \brief samples particle behavior after a collision event.
 //! \param p Particle to operate on
 void
-collision_mg(Particle* p);
+collision_mg(Particle& p);
 
 //! \brief samples a reaction type.
 //!
@@ -23,23 +23,23 @@ collision_mg(Particle* p);
 //! fission and disappearance are treated implicitly.
 //! \param p Particle to operate on
 void
-sample_reaction(Particle* p);
+sample_reaction(Particle& p);
 
 //! \brief Samples the scattering event
 //! \param p Particle to operate on
 void
-scatter(Particle* p);
+scatter(Particle& p);
 
 //! \brief Determines the average total, prompt and delayed neutrons produced
 //! from fission and creates the appropriate bank sites.
 //! \param p Particle to operate on
 void
-create_fission_sites(Particle* p);
+create_fission_sites(Particle& p);
 
 //! \brief Handles an absorption event
 //! \param p Particle to operate on
 void
-absorption(Particle* p);
+absorption(Particle& p);
 
 } // namespace openmc
 #endif // OPENMC_PHYSICS_MG_H