mirror of
https://github.com/openmc-dev/openmc.git
synced 2026-07-21 14:35:27 -04:00
80 lines
2.6 KiB
C++
80 lines
2.6 KiB
C++
//! \file mgxs_interface.h
|
|
//! A collection of C interfaces to the C++ Mgxs class
|
|
|
|
#ifndef OPENMC_MGXS_INTERFACE_H
|
|
#define OPENMC_MGXS_INTERFACE_H
|
|
|
|
#include "openmc/hdf5_interface.h"
|
|
#include "openmc/mgxs.h"
|
|
#include "openmc/vector.h"
|
|
|
|
namespace openmc {
|
|
|
|
//==============================================================================
|
|
// Global MGXS data container structure
|
|
//==============================================================================
|
|
|
|
class MgxsInterface {
|
|
public:
|
|
MgxsInterface() = default;
|
|
|
|
// Construct from path to cross sections file, as well as a list
|
|
// of XS to read and the corresponding temperatures for each XS
|
|
MgxsInterface(const std::string& path_cross_sections,
|
|
const vector<std::string> xs_to_read,
|
|
const vector<vector<double>> xs_temps);
|
|
|
|
// Does things to construct after the nuclides and temperatures to
|
|
// read have been specified.
|
|
void init();
|
|
|
|
// Set which nuclides and temperatures are to be read
|
|
void set_nuclides_and_temperatures(
|
|
vector<std::string> xs_to_read, vector<vector<double>> xs_temps);
|
|
|
|
// Add an Mgxs object to be managed
|
|
void add_mgxs(
|
|
hid_t file_id, const std::string& name, const vector<double>& temperature);
|
|
|
|
// Reads just the header of the cross sections file, to find
|
|
// min & max energies as well as the available XS
|
|
void read_header(const std::string& path_cross_sections);
|
|
|
|
// Calculate microscopic cross sections from nuclide macro XS
|
|
void create_macro_xs();
|
|
|
|
// Get the kT values which are used in the OpenMC model
|
|
vector<vector<double>> get_mat_kTs();
|
|
|
|
int num_energy_groups_;
|
|
int num_delayed_groups_;
|
|
vector<std::string> xs_names_; // available names in HDF5 file
|
|
vector<std::string> xs_to_read_; // XS which appear in materials
|
|
vector<vector<double>> xs_temps_to_read_; // temperatures used
|
|
std::string cross_sections_path_; // path to MGXS h5 file
|
|
vector<Mgxs> nuclides_;
|
|
vector<Mgxs> macro_xs_;
|
|
vector<double> energy_bins_;
|
|
vector<double> energy_bin_avg_;
|
|
vector<double> rev_energy_bins_;
|
|
vector<vector<double>> nuc_temps_; // all available temperatures
|
|
};
|
|
|
|
namespace data {
|
|
extern MgxsInterface mg;
|
|
}
|
|
|
|
// Puts available XS in MGXS file to globals so that when
|
|
// materials are read, the MGXS specified in a material can
|
|
// be ensured to be present in the available data.
|
|
void put_mgxs_header_data_to_globals();
|
|
|
|
// Set which nuclides and temperatures are to be read on
|
|
// mg through global data
|
|
void set_mg_interface_nuclides_and_temps();
|
|
|
|
// After macro XS have been read, materials can be marked as fissionable
|
|
void mark_fissionable_mgxs_materials();
|
|
|
|
} // namespace openmc
|
|
#endif // OPENMC_MGXS_INTERFACE_H
|