adding back files to be reviewed

include/openmc/physics.h

@@ -0,0 +1,93 @@
+#ifndef OPENMC_PHYSICS_H
+#define OPENMC_PHYSICS_H
+
+#include "openmc/bank.h"
+#include "openmc/nuclide.h"
+#include "openmc/particle.h"
+#include "openmc/position.h"
+#include "openmc/reaction.h"
+
+#include <vector>
+
+namespace openmc {
+
+//==============================================================================
+// Non-member functions
+//==============================================================================
+
+//! Sample a nuclide and reaction and then calls the appropriate routine
+void collision(Particle* p);
+
+//! Samples an incident neutron reaction
+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);
+
+//! Terminates the particle and either deposits all energy locally
+//! (electron_treatment = ELECTRON_LED) or creates secondary bremsstrahlung
+//! photons from electron deflections with charged particles (electron_treatment
+//! = ELECTRON_TTB).
+void sample_electron_reaction(Particle* p);
+
+//! Terminates the particle and either deposits all energy locally
+//! (electron_treatment = ELECTRON_LED) or creates secondary bremsstrahlung
+//! photons from electron deflections with charged particles (electron_treatment
+//! = ELECTRON_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);
+
+//! 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(const 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,
+  std::vector<Particle::Bank>& bank);
+
+int sample_element(Particle* p);
+
+Reaction* sample_fission(int i_nuclide, const Particle* p);
+
+void sample_photon_product(int i_nuclide, const Particle* p, int* i_rx, int* i_product);
+
+void absorption(Particle* p, int i_nuclide);
+
+void scatter(Particle*, 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);
+
+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
+//! dependence of cross sections in treating resonance elastic scattering such
+//! as the DBRC and a new, accelerated scheme are also implemented here.
+Direction sample_target_velocity(const Nuclide* nuc, double E, Direction u,
+   Direction v_neut, double xs_eff, double kT);
+
+//! samples a target velocity based on the free gas scattering formulation, used
+//! by most Monte Carlo codes, in which cross section is assumed to be constant
+//! in energy. Excellent documentation for this method can be found in
+//! FRA-TM-123.
+Direction sample_cxs_target_velocity(double awr, double E, Direction u, double kT);
+
+void sample_fission_neutron(int i_nuclide, const Reaction* rx, double E_in, Particle::Bank* site);
+
+//! 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 sample_secondary_photons(Particle* p, int i_nuclide);
+
+} // namespace openmc
+
+#endif // OPENMC_PHYSICS_H