openmc/include/openmc/simulation.h at 303b029bbdf1fb0bd1e958c29e7fd344e2828355 · openmc-dev/openmc · GitHub

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//! \file simulation.h
//! \brief Variables/functions related to a running simulation

#ifndef OPENMC_SIMULATION_H
#define OPENMC_SIMULATION_H

#include "openmc/mesh.h"
#include "openmc/particle.h"
#include "openmc/vector.h"

#include <cstdint>

namespace openmc {

constexpr int STATUS_EXIT_NORMAL {0};
constexpr int STATUS_EXIT_MAX_BATCH {1};
constexpr int STATUS_EXIT_ON_TRIGGER {2};

//==============================================================================
// Global variable declarations
//==============================================================================

namespace simulation {

extern int ct_current_file;   //!< current collision track file index
extern "C" int current_batch; //!< current batch
extern "C" int current_gen;   //!< current fission generation
extern bool ifp_delayed_on;   //!< Store delayed group IFP data?
extern bool ifp_lifetime_on;  //!< Store lifetime IFP data?
extern "C" bool initialized;  //!< has simulation been initialized?
extern "C" double keff;       //!< average k over batches
extern "C" double keff_std;   //!< standard deviation of average k
extern "C" double k_col_abs; //!< sum over batches of k_collision * k_absorption
extern "C" double
  k_col_tra; //!< sum over batches of k_collision * k_tracklength
extern "C" double
  k_abs_tra;               //!< sum over batches of k_absorption * k_tracklength
extern double log_spacing; //!< lethargy spacing for energy grid searches
extern "C" int n_lost_particles;   //!< cumulative number of lost particles
extern "C" bool need_depletion_rx; //!< need to calculate depletion rx?
extern "C" int restart_batch;      //!< batch at which a restart job resumed
extern "C" bool satisfy_triggers;  //!< have tally triggers been satisfied?
extern int ssw_current_file;       //!< current surface source file
extern "C" int total_gen;          //!< total number of generations simulated
extern double total_weight;        //!< Total source weight in a batch
extern int64_t work_per_rank;      //!< number of particles per MPI rank

extern const RegularMesh* entropy_mesh;
extern const RegularMesh* ufs_mesh;

extern vector<double> k_generation;
extern vector<int64_t> work_index;

} // namespace simulation

//==============================================================================
// Functions
//==============================================================================

//! Allocate space for source and fission banks
void allocate_banks();

//! Determine number of particles to transport per process
void calculate_work();

//! Initialize nuclear data before a simulation
void initialize_data();

//! Initialize a batch
void initialize_batch();

//! Initialize a fission generation
void initialize_generation();

//! Full initialization of a particle history
void initialize_history(Particle& p, int64_t index_source);

//! Finalize a batch
//!
//! Handles synchronization and accumulation of tallies, calculation of Shannon
//! entropy, getting single-batch estimate of keff, and turning on tallies when
//! appropriate
void finalize_batch();

//! Finalize a fission generation
void finalize_generation();

//! Determine overall generation number
extern "C" int overall_generation();

#ifdef OPENMC_MPI
void broadcast_results();
#endif

void free_memory_simulation();

//! Simulate a single particle history (and all generated secondary particles,
//!  if enabled), from birth to death
void transport_history_based_single_particle(Particle& p);

//! Simulate all particle histories using history-based parallelism
void transport_history_based();

//! Simulate all particle histories using event-based parallelism
void transport_event_based();

} // namespace openmc

#endif // OPENMC_SIMULATION_H