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Added create_fission_sites and absorption to physics_mg.cpp

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This commit is contained in:
Adam G Nelson 2018-10-09 20:35:25 -04:00
parent 960f71e276
commit ef9908e774
5 changed files with 183 additions and 164 deletions
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2
include/openmc/capi.h
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@ -149,6 +149,8 @@ extern "C" {
extern int32_t n_tallies;
extern int32_t n_universes;
extern bool openmc_simulation_initialized;
extern double global_tally_absorption;
#pragma omp threadprivate(global_tally_absorption)
// Variables that are shared by necessity (can be removed from public header
// later)

12
include/openmc/physics_mg.h
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@ -4,7 +4,9 @@
#ifndef OPENMC_PHYSICS_MG_H
#define OPENMC_PHYSICS_MG_H
#include "openmc/capi.h"
#include "openmc/particle.h"
#include "openmc/nuclide.h"
namespace openmc {
@ -16,5 +18,15 @@ namespace openmc {
extern "C" void
scatter(Particle* p, const double* energy_bin_avg);
//! \brief Determines the average total, prompt and delayed neutrons produced
//! from fission and creates the appropriate bank sites.
extern "C" void
create_fission_sites(Particle* p, Bank* bank_array, int64_t& size_bank,
int64_t& bank_array_size, MaterialMacroXS& material_xs);
//! \brief Handles an absorption event
extern "C" void
absorption(Particle* p, MaterialMacroXS& material_xs);
} // namespace openmc
#endif // OPENMC_PHYSICS_MG_H