Merge branch 'develop' into distribution-alias-sampling

include/openmc/cross_sections.h

@@ -62,6 +62,11 @@ extern vector<Library> libraries;
 //! libraries
 void read_cross_sections_xml();
 
+//! Read cross sections file (either XML or multigroup H5) and populate data
+//! libraries
+//! \param[in] root node of the cross_sections.xml
+void read_cross_sections_xml(pugi::xml_node root);
+
 //! Load nuclide and thermal scattering data from HDF5 files
 //
 //! \param[in] nuc_temps Temperatures for each nuclide in [K]

include/openmc/file_utils.h

@@ -3,14 +3,32 @@
 
 #include <fstream> // for ifstream
 #include <string>
+#include <sys/stat.h>
 
 namespace openmc {
 
+// TODO: replace with std::filesystem when switch to C++17 is made
+//! Determine if a path is a directory
+//! \param[in] path Path to check
+//! \return Whether the path is a directory
+inline bool dir_exists(const std::string& path)
+{
+  struct stat s;
+  if (stat(path.c_str(), &s) != 0)
+    return false;
+
+  return s.st_mode & S_IFDIR;
+}
+
 //! Determine if a file exists
 //! \param[in] filename Path to file
 //! \return Whether file exists
 inline bool file_exists(const std::string& filename)
 {
+  // rule out file being a path to a directory
+  if (dir_exists(filename))
+    return false;
+
   std::ifstream s {filename};
   return s.good();
 }

include/openmc/geometry_aux.h

@@ -10,6 +10,7 @@
 #include <vector>
 
 #include "openmc/vector.h"
+#include "openmc/xml_interface.h"
 
 namespace openmc {
 
@@ -19,8 +20,13 @@ extern std::unordered_map<int32_t, std::unordered_map<int32_t, int32_t>>
 extern std::unordered_map<int32_t, int32_t> universe_level_counts;
 } // namespace model
 
+//! Read geometry from XML file
 void read_geometry_xml();
 
+//! Read geometry from XML node
+//! \param[in] root node of geometry XML element
+void read_geometry_xml(pugi::xml_node root);
+
 //==============================================================================
 //! Replace Universe, Lattice, and Material IDs with indices.
 //==============================================================================

include/openmc/initialize.h

@@ -1,6 +1,8 @@
 #ifndef OPENMC_INITIALIZE_H
 #define OPENMC_INITIALIZE_H
 
+#include <string>
+
 #ifdef OPENMC_MPI
 #include "mpi.h"
 #endif
@@ -11,7 +13,13 @@ int parse_command_line(int argc, char* argv[]);
 #ifdef OPENMC_MPI
 void initialize_mpi(MPI_Comm intracomm);
 #endif
-void read_input_xml();
+
+//! Read material, geometry, settings, and tallies from a single XML file
+bool read_model_xml();
+//! Read inputs from separate XML files
+void read_separate_xml_files();
+//! Write some output that occurs right after initialization
+void initial_output();
 
 } // namespace openmc
 

include/openmc/material.h

@@ -12,6 +12,7 @@
 #include "openmc/bremsstrahlung.h"
 #include "openmc/constants.h"
 #include "openmc/memory.h" // for unique_ptr
+#include "openmc/ncrystal_interface.h"
 #include "openmc/particle.h"
 #include "openmc/vector.h"
 
@@ -151,12 +152,17 @@ public:
   //! \return Temperature in [K]
   double temperature() const;
 
+  //! Get pointer to NCrystal material object
+  //! \return Pointer to NCrystal material object
+  const NCrystalMat& ncrystal_mat() const { return ncrystal_mat_; };
+
   //----------------------------------------------------------------------------
   // Data
   int32_t id_ {C_NONE};                 //!< Unique ID
   std::string name_;                    //!< Name of material
   vector<int> nuclide_;                 //!< Indices in nuclides vector
   vector<int> element_;                 //!< Indices in elements vector
+  NCrystalMat ncrystal_mat_;            //!< NCrystal material object
   xt::xtensor<double, 1> atom_density_; //!< Nuclide atom density in [atom/b-cm]
   double density_;                      //!< Total atom density in [atom/b-cm]
   double density_gpcc_;                 //!< Total atom density in [g/cm^3]
@@ -221,6 +227,10 @@ double density_effect(const vector<double>& f, const vector<double>& e_b_sq,
 //! Read material data from materials.xml
 void read_materials_xml();
 
+//! Read material data XML node
+//! \param[in] root node of materials XML element
+void read_materials_xml(pugi::xml_node root);
+
 void free_memory_material();
 
 } // namespace openmc

include/openmc/mcpl_interface.h

@@ -0,0 +1,36 @@
+#ifndef OPENMC_MCPL_INTERFACE_H
+#define OPENMC_MCPL_INTERFACE_H
+
+#include "openmc/particle_data.h"
+#include "openmc/vector.h"
+
+#include <string>
+
+namespace openmc {
+
+//==============================================================================
+// Constants
+//==============================================================================
+
+extern "C" const bool MCPL_ENABLED;
+
+//==============================================================================
+// Functions
+//==============================================================================
+
+//! Get a vector of source sites from an MCPL file
+//
+//! \param[in] path  Path to MCPL file
+//! \return  Vector of source sites
+vector<SourceSite> mcpl_source_sites(std::string path);
+
+//! Write an MCPL source file
+//
+//! \param[in] filename  Path to MCPL file
+//! \param[in] surf_source_bank  Whether to use the surface source bank
+void write_mcpl_source_point(
+  const char* filename, bool surf_source_bank = false);
+
+} // namespace openmc
+
+#endif // OPENMC_MCPL_INTERFACE_H

include/openmc/ncrystal_interface.h

@@ -0,0 +1,94 @@
+#ifndef OPENMC_NCRYSTAL_INTERFACE_H
+#define OPENMC_NCRYSTAL_INTERFACE_H
+
+#ifdef NCRYSTAL
+#include "NCrystal/NCRNG.hh"
+#include "NCrystal/NCrystal.hh"
+#endif
+
+#include "openmc/particle.h"
+
+#include <cstdint> // for uint64_t
+#include <limits>  // for numeric_limits
+#include <string>
+
+namespace openmc {
+
+//==============================================================================
+// Constants
+//==============================================================================
+
+extern "C" const bool NCRYSTAL_ENABLED;
+
+//! Energy in [eV] to switch between NCrystal and ENDF
+constexpr double NCRYSTAL_MAX_ENERGY {5.0};
+
+//==============================================================================
+// Wrapper class an NCrystal material
+//==============================================================================
+
+class NCrystalMat {
+public:
+  //----------------------------------------------------------------------------
+  // Constructors
+  NCrystalMat() = default;
+  explicit NCrystalMat(const std::string& cfg);
+
+  //----------------------------------------------------------------------------
+  // Methods
+
+#ifdef NCRYSTAL
+  //! Return configuration string
+  std::string cfg() const;
+
+  //! Get cross section from NCrystal material
+  //
+  //! \param[in] p  Particle object
+  //! \return  Cross section in [b]
+  double xs(const Particle& p) const;
+
+  // Process scattering event
+  //
+  //! \param[in] p  Particle object
+  void scatter(Particle& p) const;
+
+  //! Whether the object holds a valid NCrystal material
+  operator bool() const;
+#else
+
+  //----------------------------------------------------------------------------
+  // Trivial methods when compiling without NCRYSTAL
+  std::string cfg() const
+  {
+    return "";
+  }
+  double xs(const Particle& p) const
+  {
+    return -1.0;
+  }
+  void scatter(Particle& p) const {}
+  operator bool() const
+  {
+    return false;
+  }
+#endif
+
+private:
+  //----------------------------------------------------------------------------
+  // Data members (only present when compiling with NCrystal support)
+#ifdef NCRYSTAL
+  std::string cfg_; //!< NCrystal configuration string
+  std::shared_ptr<const NCrystal::ProcImpl::Process>
+    ptr_; //!< Pointer to NCrystal material object
+#endif
+};
+
+//==============================================================================
+// Functions
+//==============================================================================
+
+void ncrystal_update_micro(double xs, NuclideMicroXS& micro);
+
+} // namespace openmc
+
+#endif // OPENMC_NCRYSTAL_INTERFACE_H

include/openmc/physics.h

@@ -96,8 +96,9 @@ void inelastic_scatter(const Nuclide& nuc, const Reaction& rx, Particle& p);
 
 void sample_secondary_photons(Particle& p, int i_nuclide);
 
-//！Split or Roulette particles based their weight and the lower weight window
-// bound.
+//! Split or Roulette particles based their weight and the lower weight window
+//! bound.
+//
 //! \param[in] p, particle to be split or rouletted with the weight window.
 void split_particle(Particle& p);
 

include/openmc/plot.h

@@ -279,6 +279,10 @@ void voxel_finalize(hid_t dspace, hid_t dset, hid_t memspace);
 //! Read plot specifications from a plots.xml file
 void read_plots_xml();
 
+//! Read plot specifications from an XML Node
+//! \param[in] XML node containing plot info
+void read_plots_xml(pugi::xml_node root);
+
 //! Clear memory
 void free_memory_plot();
 

include/openmc/settings.h

@@ -28,6 +28,7 @@ extern bool check_overlaps;       //!< check overlaps in geometry?
 extern bool confidence_intervals; //!< use confidence intervals for results?
 extern bool
   create_fission_neutrons; //!< create fission neutrons (fixed source)?
+extern bool create_delayed_neutrons; //!< create delayed fission neutrons?
 extern "C" bool cmfd_run;  //!< is a CMFD run?
 extern bool
   delayed_photon_scaling;   //!< Scale fission photon yield to include delayed
@@ -47,7 +48,9 @@ extern "C" bool run_CE;            //!< run with continuous-energy data?
 extern bool source_latest;         //!< write latest source at each batch?
 extern bool source_separate;       //!< write source to separate file?
 extern bool source_write;          //!< write source in HDF5 files?
+extern bool source_mcpl_write;     //!< write source in mcpl files?
 extern bool surf_source_write;     //!< write surface source file?
+extern bool surf_mcpl_write;       //!< write surface mcpl file?
 extern bool surf_source_read;      //!< read surface source file?
 extern bool survival_biasing;      //!< use survival biasing?
 extern bool temperature_multipole; //!< use multipole data?
@@ -127,9 +130,12 @@ extern double weight_survive;      //!< Survival weight after Russian roulette
 //==============================================================================
 
 //! Read settings from XML file
-//! \param[in] root XML node for <settings>
 void read_settings_xml();
 
+//! Read settings from XML node
+//! \param[in] root XML node for <settings>
+void read_settings_xml(pugi::xml_node root);
+
 void free_memory_settings();
 
 } // namespace openmc

include/openmc/source.h

@@ -102,6 +102,7 @@ class FileSource : public Source {
 public:
   // Constructors
   explicit FileSource(std::string path);
+  explicit FileSource(const vector<SourceSite>& sites) : sites_ {sites} {}
 
   // Methods
   SourceSite sample(uint64_t* seed) const override;

include/openmc/tallies/tally.h

@@ -48,10 +48,10 @@ public:
   void set_nuclides(const vector<std::string>& nuclides);
 
   //! returns vector of indices corresponding to the tally this is called on
-  const vector<int32_t>& filters() const { return filters_; } 
+  const vector<int32_t>& filters() const { return filters_; }
 
   //! \brief Returns the tally filter at index i
-  int32_t filters(int i) const { return filters_[i]; } 
+  int32_t filters(int i) const { return filters_[i]; }
 
   void set_filters(gsl::span<Filter*> filters);
 
@@ -178,6 +178,10 @@ extern double global_tally_leakage;
 //! Read tally specification from tallies.xml
 void read_tallies_xml();
 
+//! Read tally specification from an XML node
+//! \param[in] root node of tallies XML element
+void read_tallies_xml(pugi::xml_node root);
+
 //! \brief Accumulate the sum of the contributions from each history within the
 //! batch to a new random variable
 void accumulate_tallies();