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Use references consistently in physics modules

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This commit is contained in:
Paul Romano 2020-04-23 10:11:43 -05:00
parent 65c0dbfe46
commit 43eae72aeb
7 changed files with 276 additions and 277 deletions
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32
include/openmc/physics.h
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@ -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