The Material class represents a unique material and its relevant nuclear data (i.e., multigroup cross-sections) for neutron transport. More...

#include "src/Material.h"

Public Member Functions
	Material (int id=0, const char *name="")
	Constructor sets the ID and unique ID for the Material. More...

virtual	~Material ()
	Destructor deletes all cross-section data structures from memory.

int	getId () const
	Return the Material's user-defined ID. More...

char *	getName () const
	Return the user-defined name of the Material. More...

double	getVolume ()
	Return the aggregate volume/area of all instances of this Material. More...

int	getNumInstances ()
	Return the number of instances of this Material in the Geometry. More...

int	getNumEnergyGroups () const
	Returns the number of energy groups for this Material's nuclear data. More...

FP_PRECISION *	getSigmaT ()
	Return the array of the Material's total cross-sections. More...

FP_PRECISION	getMaxSigmaT ()
	Return the maximum of the Material's total cross-sections. More...

FP_PRECISION *	getSigmaS ()
	Return the array of the Material's scattering cross-section matrix. More...

FP_PRECISION *	getSigmaA ()
	Return the array of the Material's absorption group cross-sections. More...

FP_PRECISION *	getSigmaF ()
	Return the array of the Material's fission cross-sections. More...

FP_PRECISION *	getNuSigmaF ()
	Return the array of the Material's fission cross-sections multiplied by nu $\nu$ . More...

FP_PRECISION *	getChi ()
	Return the array of the Material's chi $\chi$ . More...

FP_PRECISION *	getFissionMatrix ()
	Return the array of the Material's fission matrix. More...

FP_PRECISION	getSigmaTByGroup (int group)
	Get the Material's total cross section for some energy group. More...

FP_PRECISION	getSigmaSByGroup (int origin, int destination)
	Get the Material's scattering cross section for some energy group. More...

FP_PRECISION	getSigmaAByGroup (int group)
	Get the Material's absorption cross section for some energy group. More...

FP_PRECISION	getSigmaFByGroup (int group)
	Get the Material's fission cross section for some energy group. More...

FP_PRECISION	getNuSigmaFByGroup (int group)
	Get the Material's nu-fission cross section for some energy group. More...

FP_PRECISION	getChiByGroup (int group)
	Get the Material's fission spectrum for some energy group. More...

FP_PRECISION	getFissionMatrixByGroup (int origin, int destination)
	Get the Material's fission matrix for some energy group. More...

bool	isFissionable ()
	Returns whether or not the Material contains a fissionable (non-zero) fission cross-section. More...

bool	isDataAligned ()
	Returns true if the data is vector aligned, false otherwise (default). More...

int	getNumVectorGroups ()
	Returns the rounded up number of energy groups to fill an integral number of vector lengths. More...

void	setName (const char *name)
	Sets the name of the Material. More...

void	setVolume (double volume)
	Set the volume/area of the Material. More...

void	incrementVolume (double volume)
	Increment the volume/area of the Material by some amount. More...

void	setNumInstances (int num_instances)
	Set the number of instances of this Material. More...

void	incrementNumInstances ()
	Increment the number of instances of this Material. More...

void	setNumEnergyGroups (const int num_groups)
	Set the number of energy groups for this Material. More...

void	setSigmaT (double *xs, int num_groups)
	Set the Material's array of total cross-sections. More...

void	setSigmaS (double *xs, int num_groups)
	Set the Material's 2D array of scattering cross-sections. More...

void	setSigmaF (double *xs, int num_groups)
	Set the Material's array of fission cross-sections. More...

void	setNuSigmaF (double *xs, int num_groups)
	Set the Material's array of fission cross-sections multiplied by $\nu$ . More...

void	setChi (double *xs, int num_groups)
	Set the Material's array of chi $\chi$ values. More...

void	setSigmaTByGroup (double xs, int group)
	Set the Material's total cross-section for some energy group. More...

void	setSigmaFByGroup (double xs, int group)
	Set the Material's fission cross-section for some energy group. More...

void	setNuSigmaFByGroup (double xs, int group)
	Set the Material's fission cross-section multiplied by $\nu$ for some energy group. More...

void	setSigmaSByGroup (double xs, int origin, int destination)
	Set the Material's scattering cross-section for some energy group. More...

void	setChiByGroup (double xs, int group)
	Set the Material's chi value for some energy group. More...

void	setSigmaAByGroup (double xs, int group)
	Set the Material's absorption cross-section for some energy group. More...

void	buildFissionMatrix ()
	Builds the fission matrix from chi and the fission cross-section. More...

void	transposeProductionMatrices ()
	Transposes the scattering and fission matrices. More...

void	alignData ()
	Reallocates the Material's cross-section data structures along word-aligned boundaries. More...

Material *	clone ()
	Create a duplicate of the Material. More...

std::string	toString ()
	Converts this Material's attributes to a character array representation. More...

void	printString ()
	Prints a string representation of all of the Material's attributes to the console.

Detailed Description

The Material class represents a unique material and its relevant nuclear data (i.e., multigroup cross-sections) for neutron transport.

Constructor & Destructor Documentation

◆ Material()

Material::Material	(	int	id = `0`,
		const char *	name = `""`
	)

Constructor sets the ID and unique ID for the Material.

Parameters

id	the user-specified optional Material ID
name	the user-specified optional Material name

Member Function Documentation

◆ alignData()

void Material::alignData ( )

Reallocates the Material's cross-section data structures along word-aligned boundaries.

This method is used to assist with SIMD auto-vectorization of the MOC routines in the Solver classes. Rather than using the assigned number of energy groups, this method adds "dummy" energy groups such that the total number of groups is some multiple of VEC_LENGTH (typically 4, 8, or 16). As a result, the SIMD-vectorized Solver subclasses can break up loops over energy groups in such a way to "expose" the SIMD nature of the algorithm.

◆ buildFissionMatrix()

void Material::buildFissionMatrix ( )

Builds the fission matrix from chi and the fission cross-section.

The fission matrix is constructed as the outer product of the chi and fission cross-section vectors. This routine is intended for internal use and is called by the Solver at runtime.

◆ clone()

Material * Material::clone ( )

Create a duplicate of the Material.

Returns: a pointer to the clone

◆ getChi()

FP_PRECISION * Material::getChi ( )

Return the array of the Material's chi $\chi$ .

Returns: the pointer to the Material's array of chi $\chi$ values

◆ getChiByGroup()

FP_PRECISION Material::getChiByGroup ( int group )

Get the Material's fission spectrum for some energy group.

Parameters

group the energy group

Returns: the fission spectrum

◆ getFissionMatrix()

FP_PRECISION * Material::getFissionMatrix ( )

Return the array of the Material's fission matrix.

Returns: the pointer to the Material's fission matrix array

◆ getFissionMatrixByGroup()

FP_PRECISION Material::getFissionMatrixByGroup	(	int	origin,
		int	destination
	)

Get the Material's fission matrix for some energy group.

Parameters

origin	the incoming energy group $E_{0}$
destination	the outgoing energy group $E_{1}$

Returns: the fission matrix entry $\nu\Sigma_{f}(E_{0}) * \chi(E_{1})$

◆ getId()

int Material::getId ( ) const

Return the Material's user-defined ID.

Returns: the Material's user-defined ID

◆ getMaxSigmaT()

FP_PRECISION Material::getMaxSigmaT ( )

Return the maximum of the Material's total cross-sections.

Returns: the maximum of the Material's total cross-sections.

◆ getName()

char * Material::getName ( ) const

Return the user-defined name of the Material.

Returns: the Material name

◆ getNumEnergyGroups()

int Material::getNumEnergyGroups ( ) const

Returns the number of energy groups for this Material's nuclear data.

Returns: the number of energy groups

◆ getNumInstances()

int Material::getNumInstances ( )

Return the number of instances of this Material in the Geometry.

The number of instances of this Material in the Geometry is determined during track generation.

Returns: the number of material instances

◆ getNumVectorGroups()

int Material::getNumVectorGroups ( )

Returns the rounded up number of energy groups to fill an integral number of vector lengths.

Returns: The number of vector-aligned energy groups

◆ getNuSigmaF()

FP_PRECISION * Material::getNuSigmaF ( )

Return the array of the Material's fission cross-sections multiplied by nu $\nu$ .

Returns: the pointer to the Material's array of fission cross-sections multiplied by nu $\nu$

◆ getNuSigmaFByGroup()

FP_PRECISION Material::getNuSigmaFByGroup ( int group )

Get the Material's nu-fission cross section for some energy group.

Parameters

group the energy group

Returns: the nu-fission cross section

◆ getSigmaA()

FP_PRECISION * Material::getSigmaA ( )

Return the array of the Material's absorption group cross-sections.

Returns: the pointer to the Material's array of absorption cross-sections

◆ getSigmaAByGroup()

FP_PRECISION Material::getSigmaAByGroup ( int group )

Get the Material's absorption cross section for some energy group.

Parameters

group the energy group

Returns: the absorption cross section

◆ getSigmaF()

FP_PRECISION * Material::getSigmaF ( )

Return the array of the Material's fission cross-sections.

Returns: the pointer to the Material's array of fission cross-sections

◆ getSigmaFByGroup()

FP_PRECISION Material::getSigmaFByGroup ( int group )

Get the Material's fission cross section for some energy group.

Parameters

group the energy group

Returns: the fission cross section

◆ getSigmaS()

FP_PRECISION * Material::getSigmaS ( )

Return the array of the Material's scattering cross-section matrix.

Returns: the pointer to the Material's array of scattering cross-sections

◆ getSigmaSByGroup()

FP_PRECISION Material::getSigmaSByGroup	(	int	origin,
		int	destination
	)

Get the Material's scattering cross section for some energy group.

Parameters

origin	the incoming energy group
destination	the outgoing energy group

Returns: the scattering cross section

◆ getSigmaT()

FP_PRECISION * Material::getSigmaT ( )

Return the array of the Material's total cross-sections.

Returns: the pointer to the Material's array of total cross-sections

◆ getSigmaTByGroup()

FP_PRECISION Material::getSigmaTByGroup ( int group )

Get the Material's total cross section for some energy group.

Parameters

group the energy group

Returns: the total cross section

◆ getVolume()

double Material::getVolume ( )

Return the aggregate volume/area of all instances of this Material.

The volume/area of the Material is computed from track segments which overlap this Material during track generation.

Returns: the volume/area of the Material

◆ incrementNumInstances()

void Material::incrementNumInstances ( )

Increment the number of instances of this Material.

This routine is called by the TrackGenerator during track generation and segmentation.

◆ incrementVolume()

void Material::incrementVolume ( double volume )

Increment the volume/area of the Material by some amount.

This routine is called by the TrackGenerator during track generation and segmentation.

Parameters

volume the amount to increment the current volume by

◆ isDataAligned()

bool Material::isDataAligned ( )

Returns true if the data is vector aligned, false otherwise (default).

Returns: Whether or not the Material's data is vector aligned

◆ isFissionable()

bool Material::isFissionable ( )

Returns whether or not the Material contains a fissionable (non-zero) fission cross-section.

Returns: true if fissionable, false otherwise

◆ setChi()

void Material::setChi	(	double *	xs,
		int	num_groups
	)

Set the Material's array of chi $\chi$ values.

This method is a helper function to allow OpenMOC users to assign the Material's nuclear data in Python. A user must initialize a NumPy array of the correct size (e.g., a float64 array the length of the number of energy groups) as input to this function. This function then fills the NumPy array with the data values for the Material's chi distribution. An example of how this function might be called in Python is as follows:

chi = numpy.array([0.05, 0.1, 0.15, ... ])
material = openmoc.Material(openmoc.material_id())
material.setChi(chi)

Parameters

xs	the array of chi $\chi$ values
num_groups	the number of energy groups

◆ setChiByGroup()

void Material::setChiByGroup	(	double	xs,
		int	group
	)

Set the Material's chi value for some energy group.

Parameters

xs	the chi value ( $\Chi$ )
group	the energy group

◆ setName()

void Material::setName ( const char * name )

Sets the name of the Material.

Parameters

name	the Material name string

◆ setNumEnergyGroups()

void Material::setNumEnergyGroups ( const int num_groups )

Set the number of energy groups for this Material.

Parameters

num_groups the number of energy groups.

◆ setNumInstances()

void Material::setNumInstances ( int num_instances )

Set the number of instances of this Material.

Parameters

num_instances the number of instances of this Material in the Geometry

◆ setNuSigmaF()

void Material::setNuSigmaF	(	double *	xs,
		int	num_groups
	)

Set the Material's array of fission cross-sections multiplied by $\nu$ .

Parameters

xs	the array of fission cross-sections multiplied by nu $\nu$
num_groups	the number of energy groups

◆ setNuSigmaFByGroup()

void Material::setNuSigmaFByGroup	(	double	xs,
		int	group
	)

Set the Material's fission cross-section multiplied by $\nu$ for some energy group.

This method is a helper function to allow OpenMOC users to assign the Material's nuclear data in Python. A user must initialize a NumPy array of the correct size (e.g., a float64 array the length of the number of energy groups) as input to this function. This function then fills the NumPy array with the data values for the Material's nu*fission cross-sections. An example of how this function might be called in Python is as follows:

nu_sigma_f = numpy.array([0.05, 0.1, 0.15, ... ])
material = openmoc.Material(openmoc.material_id())
material.setNuSigmaF(nu_sigma_f)

Parameters

xs	the fission cross-section multiplied by nu $\nu$
group	the energy group

◆ setSigmaAByGroup()

void Material::setSigmaAByGroup	(	double	xs,
		int	group
	)

Set the Material's absorption cross-section for some energy group.

The absorption cross section is computed by OpenMOC when needed, but the user can set it manually, to replace absorption by another cross section as absorption is not needed when determining Keff from fission rates.

Parameters

xs	the absorption cross-section
group	the energy group

◆ setSigmaF()

void Material::setSigmaF	(	double *	xs,
		int	num_groups
	)

Set the Material's array of fission cross-sections.

This method is a helper function to allow OpenMOC users to assign the Material's nuclear data in Python. A user must initialize a NumPy array of the correct size (e.g., a float64 array the length of the number of energy groups) as input to this function. This function then fills the NumPy array with the data values for the Material's fission cross-sections. An example of how this function might be called in Python is as follows:

sigma_f = numpy.array([0.05, 0.1, 0.15, ... ])
material = openmoc.Material(openmoc.material_id())
material.setSigmaF(sigma_f)

Parameters

xs	the array of fission cross-sections
num_groups	the number of energy groups

◆ setSigmaFByGroup()

void Material::setSigmaFByGroup	(	double	xs,
		int	group
	)

Set the Material's fission cross-section for some energy group.

Parameters

xs	the fission cross-section
group	the energy group

◆ setSigmaS()

void Material::setSigmaS	(	double *	xs,
		int	num_groups_squared
	)

Set the Material's 2D array of scattering cross-sections.

The array should be passed to OpenMOC as a 1D array in column-major order. This assumes the standard convention, where column index is the origin group and the row index is the destination group. That is, the array should be ordered as follows: 1 -> 1 1 -> 2 1 -> 3 ... 2 -> 1 2 -> 2 ...

Note that if the scattering matrix is defined in NumPy by the standard convention, "flat" will put the matrix into row major order. Thus, one should transpose the matrix before flattening.

For cache efficiency, the transpose of the input is actually stored in OpenMOC.

This method is a helper function to allow OpenMOC users to assign the Material's nuclear data in Python. A user must initialize a NumPy array of the correct size (e.g., a float64 array the length of the square of the number of energy groups) as input to this function. This function then fills the NumPy array with the data values for the Material's scattering cross-sections. An example of how this function might be called in Python is as follows:

sigma_s = numpy.array([[0.05,    0,    0,     ... ],
                       [0.10, 0.08,   ...     ... ],
                                   ...
                       [...        ...        ... ]])
sigma_s = numpy.transpose(sigma_s)
material = openmoc.Material(openmoc.material_id())
material.setSigmaS(sigma_s.flat)

Parameters

xs	the array of scattering cross-sections
num_groups_squared	the number of energy groups squared

◆ setSigmaSByGroup()

void Material::setSigmaSByGroup	(	double	xs,
		int	origin,
		int	destination
	)

Set the Material's scattering cross-section for some energy group.

Parameters

xs	the scattering cross-section
origin	the column index in the scattering matrix
destination	the row index in the scattering matrix

◆ setSigmaT()

void Material::setSigmaT	(	double *	xs,
		int	num_groups
	)

Set the Material's array of total cross-sections.

This method is a helper function to allow OpenMOC users to assign the Material's nuclear data in Python. A user must initialize a NumPy array of the correct size (e.g., a float64 array the length of the number of energy groups) as input to this function. This function then fills the NumPy array with the data values for the Material's total cross-sections. An example of how this function might be called in Python is as follows:

sigma_t = numpy.array([0.05, 0.1, 0.15, ...])
material = openmoc.Material(openmoc.material_id())
material.setSigmaT(sigma_t)

     NOTE: This routine will override any zero-valued cross-sections
     (e.g., in void or gap regions) with a minimum value of 1E-10 to
     avoid numerical issues in the MOC solver.

Parameters

xs	the array of total cross-sections
num_groups	the number of energy groups

◆ setSigmaTByGroup()

void Material::setSigmaTByGroup	(	double	xs,
		int	group
	)

Set the Material's total cross-section for some energy group.

Parameters

xs	the total cross-section
group	the energy group

◆ setVolume()

void Material::setVolume ( double volume )

Set the volume/area of the Material.

Parameters

volume the volume/area of the Material

◆ toString()

std::string Material::toString ( )

Converts this Material's attributes to a character array representation.

The character array returned includes the user-defined ID, and each of the absorption, total, fission, nu multiplied by fission and scattering cross-sections and chi for all energy groups.

Returns: character array of this Material's attributes

◆ transposeProductionMatrices()

void Material::transposeProductionMatrices ( )

Transposes the scattering and fission matrices.

This routine is used by the Solver when performing adjoint flux caclulations.

The documentation for this class was generated from the following files:

/home/guillaume/Research/OpenMOC/src/Material.h
/home/guillaume/Research/OpenMOC/src/Material.cpp

Public Member Functions

Detailed Description

Constructor & Destructor Documentation

◆ Material()

Member Function Documentation

◆ alignData()

◆ buildFissionMatrix()

◆ clone()

◆ getChi()

◆ getChiByGroup()

◆ getFissionMatrix()

◆ getFissionMatrixByGroup()

◆ getId()

◆ getMaxSigmaT()

◆ getName()

◆ getNumEnergyGroups()

◆ getNumInstances()

◆ getNumVectorGroups()

◆ getNuSigmaF()

◆ getNuSigmaFByGroup()

◆ getSigmaA()

◆ getSigmaAByGroup()

◆ getSigmaF()

◆ getSigmaFByGroup()

◆ getSigmaS()

◆ getSigmaSByGroup()

◆ getSigmaT()

◆ getSigmaTByGroup()

◆ getVolume()

◆ incrementNumInstances()

◆ incrementVolume()

◆ isDataAligned()

◆ isFissionable()

◆ setChi()

◆ setChiByGroup()

◆ setName()

◆ setNumEnergyGroups()

◆ setNumInstances()

◆ setNuSigmaF()

◆ setNuSigmaFByGroup()

◆ setSigmaAByGroup()

◆ setSigmaF()

◆ setSigmaFByGroup()

◆ setSigmaS()

◆ setSigmaSByGroup()

◆ setSigmaT()

◆ setSigmaTByGroup()

◆ setVolume()

◆ toString()

◆ transposeProductionMatrices()