OpenMC/include/openmc/mgxs_interface.h

//! \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();

  // Get the group index corresponding to a continuous energy
  int get_group_index(double E);

  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
  vector<double>
    default_inverse_velocity_; // approximate default inverse-velocity data
};

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