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93 lines
3.6 KiB
C++
93 lines
3.6 KiB
C++
#ifndef OPENMC_PHYSICS_H
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#define OPENMC_PHYSICS_H
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#include "openmc/bank.h"
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#include "openmc/nuclide.h"
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#include "openmc/particle.h"
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#include "openmc/position.h"
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#include "openmc/reaction.h"
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#include <vector>
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namespace openmc {
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//==============================================================================
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// Non-member functions
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//==============================================================================
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//! Sample a nuclide and reaction and then calls the appropriate routine
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void collision(Particle* p);
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//! Samples an incident neutron reaction
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void sample_neutron_reaction(Particle* p);
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//! Samples an element based on the macroscopic cross sections for each nuclide
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//! within a material and then samples a reaction for that element and calls the
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//! appropriate routine to process the physics.
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void sample_photon_reaction(Particle* p);
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//! Terminates the particle and either deposits all energy locally
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//! (electron_treatment = ELECTRON_LED) or creates secondary bremsstrahlung
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//! photons from electron deflections with charged particles (electron_treatment
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//! = ELECTRON_TTB).
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void sample_electron_reaction(Particle* p);
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//! Terminates the particle and either deposits all energy locally
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//! (electron_treatment = ELECTRON_LED) or creates secondary bremsstrahlung
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//! photons from electron deflections with charged particles (electron_treatment
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//! = ELECTRON_TTB). Two annihilation photons of energy MASS_ELECTRON_EV (0.511
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//! MeV) are created and travel in opposite directions.
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void sample_positron_reaction(Particle* p);
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//! Sample a nuclide based on their total cross sections and densities within
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//! the current material
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//!
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//! \param[in] p Particle
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//! \return Index in the data::nuclides vector
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int sample_nuclide(const Particle* p);
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//! Determine the average total, prompt, and delayed neutrons produced from
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//! fission and creates appropriate bank sites.
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void create_fission_sites(Particle* p, int i_nuclide, const Reaction* rx,
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std::vector<Particle::Bank>& bank);
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int sample_element(Particle* p);
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Reaction* sample_fission(int i_nuclide, const Particle* p);
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void sample_photon_product(int i_nuclide, const Particle* p, int* i_rx, int* i_product);
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void absorption(Particle* p, int i_nuclide);
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void scatter(Particle*, int i_nuclide);
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//! Treats the elastic scattering of a neutron with a target.
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void elastic_scatter(int i_nuclide, const Reaction& rx, double kT,
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Particle* p);
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void sab_scatter(int i_nuclide, int i_sab, Particle* p);
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//! samples the target velocity. The constant cross section free gas model is
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//! the default method. Methods for correctly accounting for the energy
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//! dependence of cross sections in treating resonance elastic scattering such
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//! as the DBRC and a new, accelerated scheme are also implemented here.
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Direction sample_target_velocity(const Nuclide* nuc, double E, Direction u,
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Direction v_neut, double xs_eff, double kT);
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//! samples a target velocity based on the free gas scattering formulation, used
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//! by most Monte Carlo codes, in which cross section is assumed to be constant
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//! in energy. Excellent documentation for this method can be found in
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//! FRA-TM-123.
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Direction sample_cxs_target_velocity(double awr, double E, Direction u, double kT);
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void sample_fission_neutron(int i_nuclide, const Reaction* rx, double E_in, Particle::Bank* site);
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//! handles all reactions with a single secondary neutron (other than fission),
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//! i.e. level scattering, (n,np), (n,na), etc.
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void inelastic_scatter(const Nuclide* nuc, const Reaction* rx, Particle* p);
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void sample_secondary_photons(Particle* p, int i_nuclide);
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} // namespace openmc
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#endif // OPENMC_PHYSICS_H
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