#include "openmc/mgxs_interface.h"

#include <string>

#include <unordered_set>

#include <fmt/format.h>

#include "openmc/cell.h"

#include "openmc/container_util.h"

#include "openmc/cross_sections.h"

#include "openmc/error.h"

#include "openmc/file_utils.h"

#include "openmc/geometry_aux.h"

#include "openmc/hdf5_interface.h"

#include "openmc/material.h"

#include "openmc/math_functions.h"

#include "openmc/nuclide.h"

#include "openmc/settings.h"

namespace openmc {

//==============================================================================

// Mgxs data loading interface methods

//==============================================================================

namespace data {

MgxsInterface mg;

}

MgxsInterface::MgxsInterface(const std::string& path_cross_sections,

const vector<std::string> xs_to_read, const vector<vector<double>> xs_temps)

{

read_header(path_cross_sections);

set_nuclides_and_temperatures(xs_to_read, xs_temps);

init();

}

void MgxsInterface::set_nuclides_and_temperatures(

vector<std::string> xs_to_read, vector<vector<double>> xs_temps)

{

// Check to remove all duplicates

xs_to_read_ = xs_to_read;

xs_temps_to_read_ = xs_temps;

if (xs_to_read_.size() != xs_temps.size())

fatal_error("The list of macro XS temperatures to read does not "

"correspond in length to the number of XS names. ");

}

void MgxsInterface::init()

{

// Check that at least some data was set to be read

if (xs_to_read_.size() == 0)

warning("No MGXS nuclides were set to be read.");

// Check if MGXS Library exists

if (!file_exists(cross_sections_path_)) {

// Could not find MGXS Library file

fatal_error(fmt::format(

"Cross sections HDF5 file '{}' does not exist!", cross_sections_path_));

}

write_message("Loading cross section data...", 5);

// Open file for reading

hid_t file_id = file_open(cross_sections_path_, 'r');

// Read filetype

std::string type;

read_attribute(file_id, "filetype", type);

if (type != "mgxs") {

fatal_error("Provided MGXS Library is not a MGXS Library file.");

}

// Read revision number for the MGXS Library file and make sure it matches

// with the current version

array<int, 2> array;

read_attribute(file_id, "version", array);

if (array != VERSION_MGXS_LIBRARY) {

fatal_error("MGXS Library file version does not match current version "

"supported by OpenMC.");

}

// ==========================================================================

// READ ALL MGXS CROSS SECTION TABLES

for (unsigned i_nuc = 0; i_nuc < xs_to_read_.size(); ++i_nuc)

add_mgxs(file_id, xs_to_read_[i_nuc], xs_temps_to_read_[i_nuc]);

file_close(file_id);

create_macro_xs();

}

//==============================================================================

void MgxsInterface::add_mgxs(

hid_t file_id, const std::string& name, const vector<double>& temperature)

{

write_message(5, "Loading {} data...", name);

// Check to make sure cross section set exists in the library

hid_t xs_grp;

if (object_exists(file_id, name.c_str())) {

xs_grp = open_group(file_id, name.c_str());

} else {

fatal_error(

fmt::format("Data for {} does not exist in provided MGXS Library", name));

}

nuclides_.emplace_back(

xs_grp, temperature, num_energy_groups_, num_delayed_groups_);

close_group(xs_grp);

}

//==============================================================================

void MgxsInterface::create_macro_xs()

{

// Get temperatures to read for each material

auto kTs = get_mat_kTs();

// Force all nuclides in a material to be the same representation.

// Therefore type(nuclides[mat->nuclide_[0]]) dictates type(macroxs).

// At the same time, we will find the scattering type, as that will dictate

// how we allocate the scatter object within macroxs.

for (int i = 0; i < model::materials.size(); ++i) {

// First we have to normalize the densities as it has not been called yet

// for MG mode

auto& mat {model::materials[i]};

mat->finalize();

if (kTs[i].size() > 0) {

// Convert atom_densities to a vector

vector<double> atom_densities(

mat->atom_density_.begin(), mat->atom_density_.end());

// Build array of pointers to nuclides's Mgxs objects needed for this

// material

vector<Mgxs*> mgxs_ptr;

for (int i_nuclide : mat->nuclide_) {

mgxs_ptr.push_back(&nuclides_[i_nuclide]);

}

macro_xs_.emplace_back(mat->name_, kTs[i], mgxs_ptr, atom_densities,

num_energy_groups_, num_delayed_groups_);

} else {

// Preserve the ordering of materials by including a blank entry

macro_xs_.emplace_back();

}

}

}

//==============================================================================

vector<vector<double>> MgxsInterface::get_mat_kTs()

{

vector<vector<double>> kTs(model::materials.size());

for (const auto& cell : model::cells) {

// Skip non-material cells

if (cell->fill_ != C_NONE)

continue;

for (int j = 0; j < cell->material_.size(); ++j) {

// Skip void materials

int i_material = cell->material_[j];

if (i_material == MATERIAL_VOID)

continue;

// Get temperature of cell (rounding to nearest integer)

double sqrtkT =

cell->sqrtkT_.size() == 1 ? cell->sqrtkT_[j] : cell->sqrtkT_[0];

double kT = sqrtkT * sqrtkT;

// Add temperature if it hasn't already been added

if (!contains(kTs[i_material], kT)) {

kTs[i_material].push_back(kT);

}

}

}

return kTs;

}

//==============================================================================

void MgxsInterface::read_header(const std::string& path_cross_sections)

{

// Save name of HDF5 file to be read to struct data

cross_sections_path_ = path_cross_sections;

// Check if MGXS Library exists

if (!file_exists(cross_sections_path_)) {

// Could not find MGXS Library file

fatal_error(fmt::format(

"Cross section HDF5 file '{}' does not exist", cross_sections_path_));

}

write_message("Reading cross sections HDF5 file...", 5);

// Open file for reading

hid_t file_id = file_open(cross_sections_path_, 'r', true);

ensure_exists(file_id, "energy_groups", true);

read_attribute(file_id, "energy_groups", num_energy_groups_);

if (attribute_exists(file_id, "delayed_groups")) {

read_attribute(file_id, "delayed_groups", num_delayed_groups_);

} else {

num_delayed_groups_ = 0;

}

ensure_exists(file_id, "group structure", true);

read_attribute(file_id, "group structure", rev_energy_bins_);

// Reverse energy bins

std::copy(rev_energy_bins_.crbegin(), rev_energy_bins_.crend(),

std::back_inserter(energy_bins_));

// Create average energies

for (int i = 0; i < energy_bins_.size() - 1; ++i) {

energy_bin_avg_.push_back(0.5 * (energy_bins_[i] + energy_bins_[i + 1]));

}

// Add entries into libraries for MG data

xs_names_ = group_names(file_id);

if (xs_names_.empty()) {

fatal_error("At least one MGXS data set must be present in mgxs "

"library file!");

}

// Close MGXS HDF5 file

file_close(file_id);

}

void put_mgxs_header_data_to_globals()

{

// Get the minimum and maximum energies

int neutron = static_cast<int>(ParticleType::neutron);

data::energy_min[neutron] = data::mg.energy_bins_.back();

data::energy_max[neutron] = data::mg.energy_bins_.front();

// Save available XS names to library list, so that when

// materials are read, the specified mgxs can be confirmed

// as present

for (auto& name : data::mg.xs_names_) {

Library lib {};

lib.type_ = Library::Type::neutron;

lib.materials_.push_back(name);

data::libraries.push_back(lib);

}

}

void set_mg_interface_nuclides_and_temps()

{

// Get temperatures from global data

vector<vector<double>> nuc_temps(data::nuclide_map.size());

vector<vector<double>> dummy;

get_temperatures(nuc_temps, dummy);

// Build vector of nuclide names which are to be read

vector<std::string> nuclide_names(data::nuclide_map.size());

for (const auto& kv : data::nuclide_map) {

nuclide_names[kv.second] = kv.first;

}

std::unordered_set<std::string> already_read;

// Loop over materials to find xs and temperature to be read

for (const auto& mat : model::materials) {

for (int i_nuc : mat->nuclide_) {

std::string& name = nuclide_names[i_nuc];

if (already_read.find(name) == already_read.end()) {

data::mg.xs_to_read_.push_back(name);

data::mg.xs_temps_to_read_.push_back(nuc_temps[i_nuc]);

already_read.insert(name);

}

}

}

}

void mark_fissionable_mgxs_materials()

{

// Loop over all files

for (const auto& mat : model::materials) {

for (int i_nuc : mat->nuclide_) {

if (data::mg.nuclides_[i_nuc].fissionable) {

mat->fissionable_ = true;

}

}

}

}

} // namespace openmc