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248 lines
8.1 KiB
C++
248 lines
8.1 KiB
C++
#ifndef OPENMC_MATERIAL_H
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#define OPENMC_MATERIAL_H
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#include <string>
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#include <unordered_map>
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#include "openmc/span.h"
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#include "pugixml.hpp"
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#include "xtensor/xtensor.hpp"
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#include <hdf5.h>
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#include "openmc/bremsstrahlung.h"
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#include "openmc/constants.h"
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#include "openmc/memory.h" // for unique_ptr
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#include "openmc/ncrystal_interface.h"
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#include "openmc/particle.h"
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#include "openmc/vector.h"
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namespace openmc {
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//==============================================================================
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// Global variables
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//==============================================================================
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class Material;
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namespace model {
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extern std::unordered_map<int32_t, int32_t> material_map;
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extern vector<unique_ptr<Material>> materials;
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} // namespace model
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//==============================================================================
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//! A substance with constituent nuclides and thermal scattering data
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//==============================================================================
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class Material {
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public:
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//----------------------------------------------------------------------------
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// Types
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struct ThermalTable {
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int index_table; //!< Index of table in data::thermal_scatt
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int index_nuclide; //!< Index in nuclide_
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double fraction; //!< How often to use table
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};
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//----------------------------------------------------------------------------
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// Constructors, destructors, factory functions
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Material() {};
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explicit Material(pugi::xml_node material_node);
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~Material();
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//----------------------------------------------------------------------------
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// Methods
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void calculate_xs(Particle& p) const;
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//! Assign thermal scattering tables to specific nuclides within the material
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//! so the code knows when to apply bound thermal scattering data
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void init_thermal();
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//! Set up mapping between global nuclides vector and indices in nuclide_
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void init_nuclide_index();
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//! Finalize the material, assigning tables, normalize density, etc.
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void finalize();
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//! Write material data to HDF5
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void to_hdf5(hid_t group) const;
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//! Export physical properties to HDF5
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//! \param[in] group HDF5 group to write to
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void export_properties_hdf5(hid_t group) const;
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//! Import physical properties from HDF5
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//! \param[in] group HDF5 group to read from
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void import_properties_hdf5(hid_t group);
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//! Add nuclide to the material
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//
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//! \param[in] nuclide Name of the nuclide
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//! \param[in] density Density of the nuclide in [atom/b-cm]
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void add_nuclide(const std::string& nuclide, double density);
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//! Set atom densities for the material
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//
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//! \param[in] name Name of each nuclide
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//! \param[in] density Density of each nuclide in [atom/b-cm]
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void set_densities(
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const vector<std::string>& name, const vector<double>& density);
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//! Clone the material by deep-copying all members, except for the ID,
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// which will get auto-assigned to the next available ID. After creating
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// the new material, it is added to openmc::model::materials.
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//! \return reference to the cloned material
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Material& clone();
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//----------------------------------------------------------------------------
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// Accessors
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//! Get density in [atom/b-cm]
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//! \return Density in [atom/b-cm]
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double density() const { return density_; }
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//! Get density in [g/cm^3]
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//! \return Density in [g/cm^3]
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double density_gpcc() const { return density_gpcc_; }
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//! Get name
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//! \return Material name
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const std::string& name() const { return name_; }
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//! Set name
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void set_name(const std::string& name) { name_ = name; }
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//! Set total density of the material
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//
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//! \param[in] density Density value
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//! \param[in] units Units of density
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void set_density(double density, const std::string& units);
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//! Set temperature of the material
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void set_temperature(double temperature) { temperature_ = temperature; };
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//! Get nuclides in material
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//! \return Indices into the global nuclides vector
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span<const int> nuclides() const
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{
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return {nuclide_.data(), nuclide_.size()};
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}
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//! Get densities of each nuclide in material
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//! \return Densities in [atom/b-cm]
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span<const double> densities() const
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{
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return {atom_density_.data(), atom_density_.size()};
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}
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//! Get ID of material
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//! \return ID of material
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int32_t id() const { return id_; }
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//! Assign a unique ID to the material
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//! \param[in] Unique ID to assign. A value of -1 indicates that an ID
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//! should be automatically assigned.
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void set_id(int32_t id);
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//! Get whether material is fissionable
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//! \return Whether material is fissionable
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bool fissionable() const { return fissionable_; }
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bool& fissionable() { return fissionable_; }
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//! Get volume of material
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//! \return Volume in [cm^3]
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double volume() const;
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//! Get temperature of material
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//! \return Temperature in [K]
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double temperature() const;
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//! Whether or not the material is depletable
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bool depletable() const { return depletable_; }
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bool& depletable() { return depletable_; }
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//! Get pointer to NCrystal material object
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//! \return Pointer to NCrystal material object
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const NCrystalMat& ncrystal_mat() const { return ncrystal_mat_; };
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//----------------------------------------------------------------------------
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// Data
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int32_t id_ {C_NONE}; //!< Unique ID
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std::string name_; //!< Name of material
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vector<int> nuclide_; //!< Indices in nuclides vector
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vector<int> element_; //!< Indices in elements vector
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NCrystalMat ncrystal_mat_; //!< NCrystal material object
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xt::xtensor<double, 1> atom_density_; //!< Nuclide atom density in [atom/b-cm]
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double density_; //!< Total atom density in [atom/b-cm]
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double density_gpcc_; //!< Total atom density in [g/cm^3]
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double volume_ {-1.0}; //!< Volume in [cm^3]
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vector<bool> p0_; //!< Indicate which nuclides are to be treated with
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//!< iso-in-lab scattering
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// To improve performance of tallying, we store an array (direct address
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// table) that indicates for each nuclide in data::nuclides the index of the
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// corresponding nuclide in the nuclide_ vector. If it is not present in the
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// material, the entry is set to -1.
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vector<int> mat_nuclide_index_;
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// Thermal scattering tables
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vector<ThermalTable> thermal_tables_;
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unique_ptr<Bremsstrahlung> ttb_;
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private:
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//----------------------------------------------------------------------------
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// Private methods
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//! Calculate the collision stopping power
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void collision_stopping_power(double* s_col, bool positron);
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//! Initialize bremsstrahlung data
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void init_bremsstrahlung();
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//! Normalize density
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void normalize_density();
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void calculate_neutron_xs(Particle& p) const;
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void calculate_photon_xs(Particle& p) const;
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//----------------------------------------------------------------------------
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// Private data members
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int64_t index_;
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bool depletable_ {false}; //!< Is the material depletable?
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bool fissionable_ {
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false}; //!< Does this material contain fissionable nuclides
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//! \brief Default temperature for cells containing this material.
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//!
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//! A negative value indicates no default temperature was specified.
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double temperature_ {-1};
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};
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//==============================================================================
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// Non-member functions
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//==============================================================================
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//! Calculate Sternheimer adjustment factor
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double sternheimer_adjustment(const vector<double>& f,
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const vector<double>& e_b_sq, double e_p_sq, double n_conduction,
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double log_I, double tol, int max_iter);
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//! Calculate density effect correction
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double density_effect(const vector<double>& f, const vector<double>& e_b_sq,
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double e_p_sq, double n_conduction, double rho, double E, double tol,
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int max_iter);
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//! Read material data from materials.xml
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void read_materials_xml();
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//! Read material data XML node
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//! \param[in] root node of materials XML element
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void read_materials_xml(pugi::xml_node root);
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void free_memory_material();
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} // namespace openmc
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#endif // OPENMC_MATERIAL_H
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