MURE class allow to make connections between classes. More...

#include <MURE.hxx>

Public Member Functions
Constructor &destructor
	MURE (string name="xsdir")
	Default constructor.

	MURE (const MURE &m)
	Copy constructor (call Copy method)

	~MURE ()
	Destructor (call Delete method)

void	Delete (bool DeleteConnector=true)
	Delete MURE either soft or hard and preserve ConnectorPlugin if !DeleteConnector.

void	Copy (const MURE &m)
	Copy MURE object.

Output control
void	SetConnectorPlugin (ConnectorPlugin *plugin)
	Set the output plugin's connector.

ConnectorPlugin *	GetConnectorPlugin ()
	returns the Connector's Plugin

void	SetMessageLevel (int level)
	Set the minimal level of printed messages.

int	GetMessageLevel ()

void	SetSilentDebug (bool flag=true)
	Set the Silent Debug mode on.

bool	GetSilentDebug ()

Evolution methods
void	Evolution (vector< double > T, int Start=0, bool BrutalStop=false)
	Perform the evolution.

bool	IsEvolutionRequired ()

void	SetEvolutionRequired (bool flag=true)

void	SetPower (double Power)

EvolutionSolver *	GetEvolutionSolver ()

void	AddList2DepletionTallyList (vector< ZAI * > NucleiList, int matnum)

void	AddZAI2GlobalDepletionTallyList (ZAI Nuclei)

bool	IsInDepletionList (ZAI *deplzai, int matnum)

void	PrintDepletionTallyList ()

int	GetSizeOfDepletionTallyList ()

vector< vector< ZAI * > >	GetDepletionTallyListZAI ()

vector< int >	GetDepletionTallyListINT ()

Flux normalization
void	SetTallyNormalizationFactor (double val)
	set Flux normalization value.

void	SetAllFissionEnergies (double Energy=2.0e08)
	Set the fission energy released for all nuclei.

double	GetFissionEnergies ()

void	SetFissionReleasedFile (string name)
	Set the fission energy released from a file.

string	GetFissionReleasedFile ()

MC neutron source
void	SetSource (MCSource *source)
	Set the MC particle source.

MCSource *	GetSource ()
	Get the MC particle source.

Particle modes (i.e. which particles are transported, (MCNP only))
void	SetModeN ()

void	SetModeNP ()

void	SetModeNE ()

void	SetModeNPE ()

void	SetModePE ()

void	SetModeP ()

void	SetMode (string MyMode="N")

string	GetMode ()

Reactions, spectrum, temperatures, data base
void	SetSpectrumType (string type="thermal")
	Set the spectrum type.

double	GetSpectrumType ()
	returns the average energy for a spectrum type.

void	SetTemperatureMap (double *T=nullptr, int NT=0)
	Set The TemperatureMap.

void	SetTemperatureMap (TemperatureMap *TM)

TemperatureMap *	GetTemperatureMap ()

BasePriority *	GetBasePriority ()

void	SetBasePriority (BasePriority *UserBasePriority)
	Set the BasePriority.

void	SetReactionListInitMethod (TSpecificFunctor< ReactionList > *method)
	Set the ReactionList Init Method.

TFunctor *	GetReactionListInitMethod ()

Nuclei Tree Seclection
void	KeepOnlyFissionProductSelection (string name="")
	Use only some fission products for evolution.

string	GetKeepOnlyFissionProductSelectionFN ()

bool	IsOnlyFissionProductSelection ()

void	SetReactionThreshold (double sigma=1e-2)

double	GetReactionThreshold ()

void	SetMCNucleusThreshold (double epsilon=1e-6)
	Set threshold to decide to write a nucleus in the MC input.

double	GetMCNucleusThreshold ()

void	SetExtendedReactionList ()

bool	GetExtendedReactionList ()

void	SetListOfWantedNucleiFN (string name)
	Set the List of Wanted Nuclei file name.

string	GetListOfWantedNucleiFN ()

void	SetShortestHalfLife (double time=3600)
	Set the shortest half time to consider for evolution.

double	GetShortestHalfLife ()

void	AddReactionNameForcedCut (string NewReactionNameForcedCut)
	Set Reactions that have to be cut regardless of the cross sections and for all nuclei.

vector< string >	GetReactionNamesForcedCutVector ()
	Get list of Reactions that have been set to be cut regardless of the cross sections and for all nuclei.

void	SetReactionNumbersForcedCutVector (vector< int > ReactionNumbersForcedCutUser)
	Set list of Reactions that have to be cut regardless of the cross sections and for all nuclei.

void	AddReactionNumberForcedCut (int NewReactionNumberForcedCut)
	Add a new Reaction that have to be cut regardless of the cross sections and for all nuclei.

vector< int >	GetReactionNumbersForcedCutVector ()
	Get list of Reactions that have been set to be cut regardless of the cross sections and for all nuclei.

vector< string >	GetAllPossibleReactionNames ()
	Get list of All Possible Reactions that can be taken into account by SMURE.

void	SetZAIsForcedCutVector (vector< ZAI * > ZAIsForceCutUser)

void	AddZAIForcedCut (int Z, int A, int I=0)

vector< ZAI * >	GetZAIsForcedCutVector ()
	Add a nuclei to the list of nuclei that have to be cut regardless of their half life.

void	PutElasticXSinDATA (bool RegisterElasticXSinDATA=true)

bool	IsElasticXSinDATA ()

void	SetReactionDepth (int depth=10000)
	Set the maximum number of consecutive reactions to consider for evolution.

int	GetReactionDepth ()

Data files &directories
void	SetDATADIR (string name="")
	Set the data directory path to name.

string	GetDATADIR ()

string	GetEnvVar (const std::string envName)
	Get environment variable such as DATAPATH.

string	GetDATAPATH ()
	Get environment DATAPATH variable.

void	SetEnsdfDATADIR (string name="")
	Set the ENSDF data directory path to name.

string	GetEnsdfDATADIR ()

void	SetNucleiChartFileName (string name="chart.jeff3.1.1")
	Set Name of the decay mode file.

string	GetNucleiChartFileName ()

void	SetReactionDataFileName (string name="AvailableReactionChart.dat")
	Set Name of the Available reaction file for the chart.

string	GetReactionDataFileName ()

void	SetMassDataFileName (string name="Mass.dat")
	Set name of the isotopes mass file.

string	GetMassDataFileName ()

void	SetNaturalIsotopeMassFileName (string name="NaturalIsotopeMass.dat")
	Set name of the NATURAL isotopes mass file.

string	GetNaturalIsotopeMassFileName ()

void	SetFPASCIIFileName (string name="FPavailable.dat")
	Set Name ASCII file for Fission Products.

char *	GetFPASCIIFileName ()

void	SetFPBinaryFileName (string name="FPyield.bin")
	Set Name Binary file for Fission Products.

char *	GetFPBinaryFileName ()

void	SetBaseSummaryFileName (string name="BaseSummary.dat")
	set the BaseSummary.dat file name

string	GetBaseSummaryFileName ()

void	ReadSpecialIsomerFile (map< int, map< int, map< int, ZAI * > > > &TheChart)
	Read Special Isomer File (see SpecialIsomer class)

void	SetSpecialIsomerFileName (string name="")
	Set Special Isomer File name (see SpecialIsomer class)

vector< SpecialIsomer > &	GetSpecialIsomerVector ()

MC code related methods (exec, output dir, ...)
void	SetMCExec (string Exec)
	Set Name of MC exec.

string	GetMCExec ()

void	SetUseNewRandomSeed (long int dbcn=- 1)
	Use a new Random Seed number in MC run.

long int	GetRandomSeed ()

void	SetAutoXSDIR (bool cond=true)
	Tells MURE to build automatically the xsdir/xsdata file.

bool	IsAutoXSDIR ()

void	RemoveXSDIR ()
	remove xsdir/xsdata file.

void	SetXSDIRName (string acefile="xsdir")

string	GetXSDIRName ()

void	SetMCRunDirectory (string name="MUREOutput", string option="")
	Set the MC run directory name.

string	GetMCRunDirectory ()

void	FindMCRunDirectory (string OriginalMCRunDir)
	Find the MC Run Directory according to evoluion step (PC, Partial run, ...)

void	BuildMCFile (string FileString="")
	Generate and Write input file on disk for the MC code (find missing volumes before)

string	GetMCInputBaseFileName ()

string	GetRealMCInputFileName ()

void	SetMCInputFileName (string Name)

void	SetRealMCInputFileName (string Name)

void	SetMaximumNumberOfDigitInEvolution (int maxi)
	set the maximum Number of digits to number evolution (default = 3)

int	GetMaximumNumberOfDigitInEvolution ()
	get the maximum Number of digits to number evolution (default = 3)

void	ConstructFileName (string FileName)
	Construct MC input filename.

string	ConstructFileName (string FileName, int EvoNum, int ThermoNum=- 1)
	Construct a temporary MC input filename.

bool	IsAutoNumberFileName ()

void	SetAutoNumberFileName (bool flag)

void	SetRunMultiMC (int n)

int	GetRunMultiMC ()

void	RunMultiMC (int n)
	Run n MC codes and adds MC detector output file to make one (mean and error of the n runs for all tallys &keff).

void	RunMC ()
	Run MC code!!!!

MC Parallel Run Mode
void	SetPVM (int n_processor)
	Set the processor number in PVM mode.

bool	IsPVM ()

void	SetMPI (int n_processor, string MPIExec="mpiexec")
	Set the processor number in MPICH mode.

bool	IsMPI ()

string	GetMPIExec ()

void	SetOMP (int n_thread)
	Set the processor number in OpenMP mode.

int	GetOMP ()

int	GetNProcessors ()

MC special methods
void	AddSpecialCard (string)
	Add General card to MC.

vector< string > &	GetSpecialCard ()

void	SetSpecialCard (int i, string str)

void	SetComment (string Comment="This comment is never read...")

string	GetComment ()

void	AddSpecialSurface (string)
	Add string card associated to the MC's Surface definitions.

vector< string >	GetSpecialSurfaceCard ()

Thermal-hydraulics methods
int	GetMCNPThermoRunNumber ()

void	SetMCNPThermoRunNumber (int value=0)

void	IncreaseMCNPThermoRunNumber ()

void	BuildCOBRAFiles ()
	Build COBRA files.

void	SetCOBRACalculationRequired (int NumberOfIteration)

void	SetThermoHydraulicsRequired (bool flag=true)

bool	IsThermoHydraulicsRequired ()

void	SetUserDefaultHydraulicModel ()
	Build automatic tallies in multigroup run without using the perturbative materials.

Keff methods
double	GetKeff ()

double	GetKeff_Err ()

void	SetKeff (double k)

void	SetKeff_Err (double dk)

void	UpdateKeff ()
	Read MC's Output file to set Keff value (can be used without evolution)

Volume calculation
bool	IsVolumeCalculationNeeded ()
	return true if a volume is missing for a cell in which a tally is asked

void	FindMissingVolume ()
	Find Missing volume (call ConnectorPlugin method)

void	SetVolumeNPS (int nps=400000)

int	GetVolumeNPS ()

void	SetVarianceVolumeWarning (double variance=0.005)

double	GetVarianceVolumeWarning ()

void	SetOutermostShape (Shape_ptr s)
	Set the most outer shape.

Shape_ptr	GetOutermostShape ()

void	SetOutermostCell (Cell *c)

Cell *	GetOutermostCell ()

void	SetSourceSphereRadius (double R)

void	SetSourceSphereCenter (double *O)

double	GetSourceSphereRadius ()

double *	GetSourceSphereCenter ()

void	AddCellOfUnknownVolume (Cell *theCell)
	add thecell to the UnkownVolume cell vector

void	AddShapeOfUnknownArea (Shape *theShape)
	add theShep to the UnkownSurface Shape vector

vector< Cell * > &	GetCellOfUnknownVolume ()

vector< Shape * > &	GetShapeOfUnknownArea ()

Cell *	FindCellOfUnknownVolume (int CellNo)

void	SetFromVolumeCalculation (bool flag)

bool	GetFromVolumeCalculation ()

Miscellaneous methods
void	StartGUI ()
	Start the TCL GUI (Examtree.tcl) to see the Nueclei Tree.

void	IntegrateMassInEvolvingCells ()
	Sum all the evolving masses in the problem.

double	GetTotalAtoms ()
	Sum all the evolving atoms in the problem.

void	WriteGlobalNucleiTree (string fname="UnionOfAllTrees.dat")
	write the union of all nuclear trees (evolving nuclei in globalnucleivector)

void	AddToGlobalNucleiVector (Nucleus_ptr &nuc)
	add a nucleus to GlobalNucleiVector

void	AddToNucleiErrorMap (Nucleus *nuc, string ErrType)

void	SumUpNucleiError (ostream &Out=cerr)
	Write in the chosen output the list of the nuclei with problems, and the type of these problems.

map< Nucleus *, string > &	GetNucleiErrorMap ()

vector< Nucleus_ptr > &	GetGlobalNucleiVector ()

bool	IsBaseSummaryRead ()

void	ReadBaseSummary ()
	Read the whole base summary file to increase speed.

string	GetBaseSummaryLine (int i)
	Get the ith base summary line.

void	AddLineToBaseSummary (string line)
	Add a line to the read base summary.

int	GetFirstContinuousZIndexInBaseSummary (int z)

int	GetFirstThermalZIndexInBaseSummary (int z)

void	CheckExtensions ()
	Check if two different nuclei have the same extension and correct if necessary.

void	CheckNucleusDataBaseError ()
	Check for nuclei error in data base (T, base name, ...)

void	SetMCNPUserDefinedGeometry ()

bool	IsMCNPUserDefinedGeometry ()

void	SetUserGeometryInputFile (string name)
	set the file name of the user defined MCNP geometry

string	GetUserGeometryInputFile ()

void	SaveBackupMureFile (string TheFileName)
	Backup copy of the MURE source file (.cxx) in the MC run directory.

string	GetTotalRunTime ()
	Returns a string that tells how long it is since MURE's code is running.

double	GetRunTime ()

void	Dump ()
	Dump a MURE object.

void	CheckMandatoryPointeur ()
	Verify the mandatory pointeur exists (such as MCSource, EvolutionSolver and ConnectrPlugin)

double	FindFissileEnrichment (double DesiredKeff, double EnrichmentFirstGuess, Material Fertile, Material Fissile, Material *&Target)

void	SetSerpentDopplerBroadening ()

bool	IsSerpentDopplerBroadening ()

void	SetDisableViolentSimplifyNode (bool flag)

bool	GetDisableViolentSimplifyNode ()

Tally methods
void	AddTally (MureTally *T)
	Add a Tally to Mure.

void	AddMeshTally (MeshTally *MT)
	Add a MeshTally to Mure (for MCNP)

vector< MeshTally * > &	GetMeshTally ()

vector< MureTally * > &	GetTallyVector ()

void	ResetDetectorVector ()
	Remove all tallies.

MureTally *	FindTally (int TallyNo)
	return the tally that has the number TallyNo

int	NextTallyNum (int i=0)
	returns the next Tally number of type i available

void	AddTimeEnergyGrid (MureTimeEnergyGrid *TheEnergyGrid)
	add a Time and/or Energy grid for tallies

vector< MureTimeEnergyGrid * >	GetTimeEnergyGrids ()
	return the Time and/or Energy grid for tallies

MureTimeEnergyGrid *	FindTimeEnergyGrid (int num)
	return the TimeEnergyGrid of number num

bool	IsSameTimeEnergyGridForAllTallies ()

void	SetSameTimeEnergyGridForAllTallies ()
	set that all tallies must use the same TimeEnergyGrid

void	SetESpectrum ()

bool	IsESpectrum ()

void	UseEnergyBinsFile (string Filename)
	uses the filename provided for building the energy binning. (single column file with energies in eV)!

vector< double > &	GetEnergyBins ()

bool	IsUseEnergyBinsFile ()

void	SetESpectrumTallyNumber (int num)

int	GetESpectrumTallyNumber ()

Vector management
void	SetMaterialNum (int num)

int	GetMaterialNum ()

int	NextCellNum ()
	returns the next Cell number available

int	NextSurfaceNum ()
	returns the next Surface number available

int	NextMaterialNum ()
	returns the next Material number available

int	NextUniverseNum ()
	returns the next Universe number available

int	NextTransformationNum ()
	returns the next Transformation number available

void	SetLastTallyNum (int Num)
	Set the last tally number to Num (for MCNP user input geometry)

vector< Cell * > &	GetCellVector ()

void	AddCell (Cell *theCell)

Cell *	FindCell (Cell *TheCell)
	return the Cell of same number than TheCell

Cell *	FindCell (int CellNo)
	return the Cell of number CellNo

vector< Cell * > &	GetTrueCellVector ()

vector< PinCell * > &	GetPinCellVector ()

void	AddPinCell (PinCell *thePinCell)

PinCell *	FindPinCell (PinCell *pin)

PinCell *	FindPinCell (int Universe)

vector< Material * > &	GetMaterialVector ()

void	AddMaterial (Material *theMaterial)
	Add a Material to the Material Vector.

void	RemoveMaterial (Material *TheMaterial)
	Remove a Material from the Material vector.

Material *	FindMaterial (Material *TheMaterial)
	return the material of same number than TheMaterial

Material *	FindMaterial (int MaterialNumber)
	return the material of number MaterialNumber

void	DontAddMaterial ()

bool	IsAllowedToAddMaterial ()

void	AddSurfaceCard (SurfaceCard *card)
	Add a SurfaceCard to the SurfaceCard Vector.

void	RemoveSurfaceCard (vector< SurfaceCard * > ::iterator it)
	remove a Surface to the Surface Vector.

vector< SurfaceCard * > &	GetMureSurfaceCardVector ()

void	ClearSurfaceCardVector ()

void	AddTransformation (Transformation *tr)

vector< Transformation * > &	GetTransformationVector ()

Transformation *	FindTransformation (int transformation_number)
	return the transformation with this number

Friends
class	MureTrojan

fraction of delayed neutron calculation
static bool	fIsBannerPrinted = false
	Print the banner only once.

int	fStartTime
	The computer clock time at the start of the MURE run.

int	fMessageLevel
	message level treshold

bool	fSilentDebug
	generally suppress the debug messages

bool	fMaterialUpdated
	flag indicating that Materials have been updated

int	fLastCellNum
	last Cell number used

int	fLastSurfaceNum
	last Surface number used

int	fLastMaterialNum
	last Material number used

int	fLastUniverseNum
	last Universe number used

int	fLastTallyNum [8]
	last Tally of type i number used

string	fComment
	first line comment in MC file

vector< Cell * >	fCellVector
	the vector of all Cell to write

vector< PinCell * >	fPinCellVector
	the vector of all PinCell to write

vector< Material * >	fMaterialVector
	the vector of all Material to write

bool	fAddMaterial
	false if you don't want to add material in fMaterialVector

vector< Nucleus_ptr >	fGlobalNucleiVector
	Union of all the nuclei used in all the evolving materials.

vector< Cell * >	fTrueCellVector
	Vector of all true Cells (which are non void) which shall be written to MC (which fNumber is not negative)

vector< Cell * >	fCellOfUnknownVolume
	Vector of Cell of Unknown volume used in tallies.

vector< Shape * >	fShapeOfUnknownArea
	Vector of Shape_ptr of Unknown area used in tallies.

vector< SurfaceCard * >	fSurfaceCardVector
	the vector of all Surface to write (in the MC output format)

vector< MureTally * >	fTallyVector
	the vector of all Tally to write

vector< MeshTally * >	fMeshTallyVector
	the vector of all Mesh Tally to write

vector< Transformation * >	fTransformationVector
	the vector geometrical transformation

string	fDataDirectoryName
	data directory name for all data files

string	fEnsdfDataDirectoryName
	data directory name for ENSDF data (added B.LENIAU)

string	fMCRunDirectory
	name of the directory in which the MC output files are written

string	fMCInputBaseFileName
	The Base Name of the input file name generated for MC (such as "inp")

string	fMCInputFileName
	The final file name for MC (such as inpXXX for step XXX of an Evolution)

bool	fAutoNumberFileName
	whether or not use automatic numbering of evolving MC input files (e.g. XXX for inp base name)

int	fNbOfDigitInEvo
	Number of digits to number evolution (default = 3)

TemperatureMap *	fTemperatureMap
	The temperature Map.

map< Nucleus *, string >	fNucleiErrorMap
	Map to store problems about nuclei (Errors in Nucleus::FindCode, more precisely)

BasePriority *	fBasePriority
	The Base Priority wish.

double	fReactionThreshold
	the reaction ratio threshold for ReactionList

TFunctor *	fReactionListInitMethod
	ReactionList Init tMethod (e.g. InitPTO())

double	fShortestHalfTimeThres
	The shortest half life considered for a nucleus.

int	fReactionDepth
	The maximum number of consecutive reactions.

double	fMeanEnergySpectrum
	The mean energy spectrum to consider (for FP yield)

double	fMCNucleusThreshold
	the threshold to decide to write a nucleus in MC's input file

double	fSetAllFissionEnergies
	If > zero, then we will override correct fission energies with this value.

string	fFissionReleasedFN
	The name of fission energy released file.

bool	fKeepOnlyFissionProductSelection
	true if all fission products are not used for evolution

string	fKeepOnlyFissionProductSelectionFN
	The file name that contains the FP to keep.

string	fSpecialIsomerFileName
	file that explain the particular treatment of special isomers

vector< SpecialIsomer >	fSpecialIsomerVector
	vector of special Isomer treatment (like capture on Am241)

string	fListOfWantedNucleiFN
	File name of the list of wanted nuclei (and ONLY these)

vector< int >	fReactionNumbersForcedCutVector
	List of reactions type that should not be considered even if the cross section are bigger than fReactionThreshold. Given by SMURE reaction number and not MT number.

vector< ZAI * >	fZAIsForcedCutVector
	List of ZAI that should not be considered even if their half life are longer than fShortestHalfTimeThres.

bool	fRegisterElasticXSinDATA
	true if elastic scattering cross section registered in the DATA file next to other XSs

string	fXSDIRName
	the xsdir/ACElib name

bool	fIsAutoXSDIR
	Whether the XSDIr must be automtically built or not.

string	fMCExec
	name of MC exec(mcnp5, sss2, ...)

int	fVolumeNPS
	source neutron number for stochastic volume calculation

double	fVarianceVolumeWarning
	warning message threshold in stochastic volume precision

double	fSourceSphereR
	Sphere source radius used in stochastic volume calculation.

double *	fSourceSphereO
	Sphere source center used in stochastic volume calculation.

bool	fIsFromVolumeCalculation
	Flag that control copy/delete of MURE depending on the call time.

string	fNucleiChartFileName
	File where to find radioactive decays modes &branching ratios.

string	fReactionDataFile
	File where to find if reaction are available or not.

string	fMassDataFileName
	File where to find the nuclear mass data.

string	fNaturalIsotopeMassFileName
	File where to find the nuclear mass data for natural isotopes.

string	fFPASCIIFileName
	Name ASCII file for Fission Products.

string	fFPBinaryFileName
	Name Binary file for Fission Products.

string	fBaseSummaryFileName
	Name of the Base Summary file to find info on availlable data bases.

vector< string >	fBaseSummary
	Table to put the Base summary file in memory for quick access.

bool	fIsBaseSummaryRead
	Flag to say if Base summary has been read in or not.

vector< int >	fFirstContinuousZIndexInBaseSummary
	Store line number where to go to find Z entries for continuous energy in Base Summary.

vector< int >	fFirstThermalZIndexInBaseSummary
	Store line number where to go to find Z entries for thermal data in Base Summary.

string	fMode
	the mode MCNP card (N, N P, N E, P E, P, or N P E). For Serpent, only mode N is taken.

vector< string >	fSpecialSurfaceCard
	the vector of all Special Surface card

vector< string >	fSpecialCard
	the vector of all Special card to write

int	fLastTransformNum
	last Transformation used

bool	fThermoHydraulicsRequired
	If we are running MCNP many times with ThermoHydraulics.

bool	fCOBRACalculationRequired
	If we are running iteration between MCNP and COBRA.

int	fNumberOfIterationNTH
	Number of Neutronics/thermalhydraulics iterations required at each fuel depletion step.

bool	fUserDefaultHydraulicModel
	Use specific hydraulics model for CHF.

int	fMCNPThermoRunNumber
	Current ThermoHydraulics MCNP Run Number.

double	fKeff
	Current Keff of the problem.

double	fKeff_Err
	Current Error on current Keff of the problem.

bool	fSetESpectrum
	flag for tallying Energy Bins in all "True" Cells (Cells which do contain a non void material)

bool	fUseEnergyBinsFile
	flag for using a user provided data file with desired energy binning.

string	fEnergyBinsFileName
	name of the user provided data file with desired energy binning.

int	fESpectrumTallyNumber
	Number of the Tally for Energy flux spectrum.

vector< MureTimeEnergyGrid * >	fTimeEnergyGrids

vector< double >	fEnergyBins
	vector of Energies in eV read from the fEnergyBinsFileName file that will be put in a MureTimeEnergyGrid by a tally

bool	fIsSameTimeEnergyGridForAllTallies

Shape_ptr	fOutermostShape
	The outer Shape to evaluate neutron leakage or volume.

Cell *	fOutermostCell

int	fNprocessors
	Number of processor in PVM/MPICH/OpenMP MC run.

string	fMPIExec
	MPI Exec name to launch MC (mpiexec, mpirun, ...)

bool	fPVM
	True for a MC PVM parallel run.

bool	fMPI
	True for a MC MPI parallel run.

int	fOMP
	number of Threads in OpenMP parallel compilation of MC.

int	fRunMultiMC
	Number of run in a Multi MC mode.

long int	fRandomSeedGenerator
	the random seed value

bool	fForceConstRandomSeedForMultiRun

MCSource *	fReactivityCoeffsSource
	Source Used for reactivity coeffs calculations.

bool	fMUREInputGeometry
	True if the geometry is a MURE geometry and false for a user defined MCNP geometry.

string	fUserGeometryInputFile
	The input file for the user given MCNP geometry.

bool	fBetaCalculationRequired
	True if delayed neutron fraction calculation wanted.

vector< double >	fNSigmaFisPhiTotalOfNucleus
	For a nucleus : sum of reaction rates.

ConnectorPlugin *	fConnectorPlugin
	Pointer to couple MURE with MCNP, Serpent...

bool	fIsConnectorPluginCalled

bool	fEvolutionRequired
	If an Evolution is required (i.e. many MC's run steps)

EvolutionSolver *	fEvolutionSolver
	The evolution pointer connector.

bool	fSerpentDopplerBroadening

bool	fExtendedReactionList

vector< int >	fMaterialForTallyNormalization
	Material numbers to be used for partial normalization to power.

vector< ZAI * >	fGlobalZAIDepletionTallyList

vector< vector< ZAI * > >	fZAIDepletionTallyList
	List of ZAIs that should be tallied in multigroup mode for each material.

vector< int >	fMatnumDepletionTallyList
	List of ZAIs that should be tallied in multigroup mode.

bool	fDisableViolentSimplifyNode
	Material number for which tallying of isotopes should happen.

void	SetBetaCalculation (bool flag=true)

bool	IsBetaCalculation ()

bool	IsNucleusUsedForBeta (int Z, int A)
	Check on used or unused nucleus in delayed neutron fraction calculation.

void	CalculateBetaI ()
	Method called in MURE input file to calculate NOW the delayed neutron fractions.

void	SumAllCellToBetaCalculation ()
	Sums all beta by cell to obtain the total one.

void	ResetAllBetaValues ()
	put to zero all values (used at each calculation done : needed if severals calculations of beta are required in a same MURE calculation

void	SetMaterialForPartialTallyNormalization (int matnum)

void	SetMaterialForPartialTallyNormalization (vector< int > matnum)

vector< int >	GetMaterialForPartialTallyNormalization ()

void	PrintBanner ()
	Print a banner at start of MURE.

Detailed Description

MURE class allow to make connections between classes.

The aim of this class is to keep track of object/pointer used in order to take all the input information into account. One has to used the global pointer (very ugly) gMURE to interact with all MURE classes. It is a kind of Super Class knowing every body. There are 2 very important pointeurs fEvolutionSolver and fConnectorPlugin ; the first one controls all evolution parameters. One access to it via gMURE->GetEvolutionSolver(). see EvolutionSolver class for more details. The second one MUST be used in conjunction to a namespace and allows the coupling of MURE object to other codes (for example MCNP or Serpent) ; Any MURE input file MUST DEFINE a Connector and set it to MURE.

See also: User Guide

Author: PTO; Kernie; FMS; J.Hajnrych; F.Havluj; B. Leniau; Nico*; BaM; J.Wilson; WEC

Version: 2.0

Constructor & Destructor Documentation

◆ MURE() [1/2]

MURE::MURE ( string name = "xsdir" )

Default constructor.

◆ MURE() [2/2]

MURE::MURE ( const MURE & m )

Copy constructor (call Copy method)

◆ ~MURE()

MURE::~MURE ( )

Destructor (call Delete method)

Member Function Documentation

◆ AddCell()

void MURE::AddCell ( Cell * theCell )

inline

< Add a Cell to the Cell Vector. Only those cells shall be written if their fNumber attribute is positive

◆ AddCellOfUnknownVolume()

void MURE::AddCellOfUnknownVolume ( Cell * theCell )

add thecell to the UnkownVolume cell vector

◆ AddLineToBaseSummary()

void MURE::AddLineToBaseSummary ( string line )

Add a line to the read base summary.

◆ AddList2DepletionTallyList()

void MURE::AddList2DepletionTallyList	(	vector< ZAI * >	NucleiList,
		int	matnum
	)

◆ AddMaterial()

void MURE::AddMaterial ( Material * theMaterial )

Add a Material to the Material Vector.

◆ AddMeshTally()

void MURE::AddMeshTally ( MeshTally * MT )

Add a MeshTally to Mure (for MCNP)

◆ AddPinCell()

void MURE::AddPinCell ( PinCell * thePinCell )

inline

< Add a PinCell to the PinCell Vector.

◆ AddReactionNameForcedCut()

void MURE::AddReactionNameForcedCut ( string NewReactionNameForcedCut )

Set Reactions that have to be cut regardless of the cross sections and for all nuclei.

Sigma for these reactions are considered to be zero for all the rest of the simulation. Reaction are named by their SMURE name in the form similar to (N_2N) or (N_alpha) (n, gamma) is (N_gamma)

◆ AddReactionNumberForcedCut()

void MURE::AddReactionNumberForcedCut ( int NewReactionNumberForcedCut )

inline

Add a new Reaction that have to be cut regardless of the cross sections and for all nuclei.

Sigma for these reactions are considered zero. Reaction are named by their SMURE number and not their MT number. (n, gamma) is 1 and not 102

◆ AddShapeOfUnknownArea()

void MURE::AddShapeOfUnknownArea ( Shape * theShape )

add theShep to the UnkownSurface Shape vector

◆ AddSpecialCard()

void MURE::AddSpecialCard ( string card )

Add General card to MC.

Add any string card in the MC input file. For MCNP these cards are written in the General Block. This allow to define any MC card not implemented in MURE.

◆ AddSpecialSurface()

void MURE::AddSpecialSurface ( string card )

Add string card associated to the MC's Surface definitions.

These cards are printed in the Surface Block of MCNP. Allows user to define its own entries in the Surface Block of MC (after the MCNP's cell block)

◆ AddSurfaceCard()

void MURE::AddSurfaceCard ( SurfaceCard * card )

Add a SurfaceCard to the SurfaceCard Vector.

◆ AddTally()

void MURE::AddTally ( MureTally * T )

Add a Tally to Mure.

◆ AddTimeEnergyGrid()

void MURE::AddTimeEnergyGrid ( MureTimeEnergyGrid * TheEnergyGrid )

add a Time and/or Energy grid for tallies

◆ AddToGlobalNucleiVector()

void MURE::AddToGlobalNucleiVector ( Nucleus_ptr & nuc )

add a nucleus to GlobalNucleiVector

◆ AddToNucleiErrorMap()

void MURE::AddToNucleiErrorMap	(	Nucleus *	nuc,
		string	ErrType
	)

inline

< add a ZAI to NucleiErrorMap with its problem type

◆ AddTransformation()

void MURE::AddTransformation ( Transformation * tr )

inline

< Add a (geometrical) transformation to the vector

◆ AddZAI2GlobalDepletionTallyList()

void MURE::AddZAI2GlobalDepletionTallyList ( ZAI Nuclei )

◆ AddZAIForcedCut()

void MURE::AddZAIForcedCut	(	int	Z,
		int	A,
		int	I = `0`
	)

◆ BuildCOBRAFiles()

void MURE::BuildCOBRAFiles ( )

Build COBRA files.

◆ BuildMCFile()

void MURE::BuildMCFile ( string FileString = "" )

Generate and Write input file on disk for the MC code (find missing volumes before)

◆ CalculateBetaI()

void MURE::CalculateBetaI ( )

Method called in MURE input file to calculate NOW the delayed neutron fractions.

◆ CheckExtensions()

void MURE::CheckExtensions ( )

Check if two different nuclei have the same extension and correct if necessary.

◆ CheckMandatoryPointeur()

void MURE::CheckMandatoryPointeur ( )

Verify the mandatory pointeur exists (such as MCSource, EvolutionSolver and ConnectrPlugin)

◆ CheckNucleusDataBaseError()

void MURE::CheckNucleusDataBaseError ( )

Check for nuclei error in data base (T, base name, ...)

◆ ClearSurfaceCardVector()

void MURE::ClearSurfaceCardVector ( )

◆ ConstructFileName() [1/2]

void MURE::ConstructFileName ( string FileName )

Construct MC input filename.

◆ ConstructFileName() [2/2]

string MURE::ConstructFileName	(	string	FileName,
		int	EvoNum,
		int	ThermoNum = `- 1`
	)

Construct a temporary MC input filename.

◆ Copy()

void MURE::Copy ( const MURE & m )

Copy MURE object.

◆ Delete()

void MURE::Delete ( bool DeleteConnector = true )

Delete MURE either soft or hard and preserve ConnectorPlugin if !DeleteConnector.

◆ DontAddMaterial()

void MURE::DontAddMaterial ( )

inline

< The Material vector is not fill: the puspose is to avoid an error when using OpenMP for GammaSpectrum calculation in MureGui

◆ Dump()

void MURE::Dump ( )

Dump a MURE object.

◆ Evolution()

void MURE::Evolution	(	vector< double >	T,
		int	Start = `0`,
		bool	BrutalStop = `false`
	)

Perform the evolution.

The vector of times correponds to the absolute times where a Monte-Carlo (MC) transport run is performed or a cooling period begins; if the time vector does not begin by 0, this value is added. Between these times, fuel evolution is performed. This method performs evolution for each EvolutiveSystem (i.e. each evolving Material). See EvolutionSolver class for more details.

The Start argument is the step number at which step evolution should restart (default is 0). Useful only when one wishes to start off the evolution again from a given step. Of course, previous steps (before Start) must exist in the MC run directory because material compositions is read from the MURE (B)DATA_XXX output files!

Parameters

T	Time vector at which MC runs are done or Cooling periods are started (in seconds)
Start	Number of the first MC step to restart evolution;
BrutalStop	Stop evolution when keff falls below BrutalStopValue (if 0, do not care about keff)

◆ FindCell() [1/2]

Cell * MURE::FindCell ( Cell * TheCell )

return the Cell of same number than TheCell

◆ FindCell() [2/2]

Cell * MURE::FindCell ( int CellNo )

return the Cell of number CellNo

◆ FindCellOfUnknownVolume()

Cell * MURE::FindCellOfUnknownVolume ( int CellNo )

◆ FindFissileEnrichment()

double MURE::FindFissileEnrichment	(	double	DesiredKeff,
		double	EnrichmentFirstGuess,
		Material *	Fertile,
		Material *	Fissile,
		Material *&	Target
	)

◆ FindMaterial() [1/2]

Material * MURE::FindMaterial ( int MaterialNumber )

return the material of number MaterialNumber

◆ FindMaterial() [2/2]

Material * MURE::FindMaterial ( Material * TheMaterial )

return the material of same number than TheMaterial

◆ FindMCRunDirectory()

void MURE::FindMCRunDirectory ( string OriginalMCRunDir )

Find the MC Run Directory according to evoluion step (PC, Partial run, ...)

◆ FindMissingVolume()

void MURE::FindMissingVolume ( )

Find Missing volume (call ConnectorPlugin method)

◆ FindPinCell() [1/2]

PinCell * MURE::FindPinCell ( int Universe )

◆ FindPinCell() [2/2]

PinCell * MURE::FindPinCell ( PinCell * pin )

◆ FindTally()

MureTally * MURE::FindTally ( int TallyNo )

return the tally that has the number TallyNo

◆ FindTimeEnergyGrid()

MureTimeEnergyGrid * MURE::FindTimeEnergyGrid ( int num )

return the TimeEnergyGrid of number num

◆ FindTransformation()

Transformation * MURE::FindTransformation ( int transformation_number )

return the transformation with this number

◆ GetAllPossibleReactionNames()

vector< string > MURE::GetAllPossibleReactionNames ( )

Get list of All Possible Reactions that can be taken into account by SMURE.

Reaction are named by their SMURE name in the form similar to (N_2N) or (N_alpha) (n, gamma) is (N_gamma)

◆ GetBasePriority()

BasePriority * MURE::GetBasePriority ( )

inline

< returns the BasePriority

◆ GetBaseSummaryFileName()

string MURE::GetBaseSummaryFileName ( )

inline

< return the BaseSummary.dat file name

◆ GetBaseSummaryLine()

string MURE::GetBaseSummaryLine ( int i )

Get the ith base summary line.

◆ GetCellOfUnknownVolume()

vector< Cell * > & MURE::GetCellOfUnknownVolume ( )

inline

◆ GetCellVector()

vector< Cell * > & MURE::GetCellVector ( )

inline

< Returns the Cell Vector

◆ GetComment()

string MURE::GetComment ( )

inline

< First line of MC file

◆ GetConnectorPlugin()

ConnectorPlugin * MURE::GetConnectorPlugin ( )

returns the Connector's Plugin

◆ GetDATADIR()

string MURE::GetDATADIR ( )

inline

< returns the Data directory where data files are stored

◆ GetDATAPATH()

string MURE::GetDATAPATH ( )

Get environment DATAPATH variable.

◆ GetDepletionTallyListINT()

vector< int > MURE::GetDepletionTallyListINT ( )

inline

◆ GetDepletionTallyListZAI()

vector< vector< ZAI * > > MURE::GetDepletionTallyListZAI ( )

inline

◆ GetDisableViolentSimplifyNode()

bool MURE::GetDisableViolentSimplifyNode ( )

inline

◆ GetEnergyBins()

vector< double > & MURE::GetEnergyBins ( )

inline

< Energy read from file define in UseEnergyBinsFile (Energy are in ev !!!)

◆ GetEnsdfDATADIR()

string MURE::GetEnsdfDATADIR ( )

inline

◆ GetEnvVar()

string MURE::GetEnvVar ( const std::string envName )

Get environment variable such as DATAPATH.

if env variable not set, the string is "VARIABLE.Not_Set.Error"

◆ GetESpectrumTallyNumber()

int MURE::GetESpectrumTallyNumber ( )

inline

< set the number of theEnergy spectrum global tally

◆ GetEvolutionSolver()

EvolutionSolver * MURE::GetEvolutionSolver ( )

inline

< get the EvolutionSolver, the class managing the evolution

◆ GetExtendedReactionList()

bool MURE::GetExtendedReactionList ( )

inline

◆ GetFirstContinuousZIndexInBaseSummary()

int MURE::GetFirstContinuousZIndexInBaseSummary ( int z )

inline

< Use to find faster z entries for continuous energy in Base Summary

◆ GetFirstThermalZIndexInBaseSummary()

int MURE::GetFirstThermalZIndexInBaseSummary ( int z )

inline

< Use to find faster z entries for thermal data in Base Summary

◆ GetFissionEnergies()

double MURE::GetFissionEnergies ( )

inline

< Get fission energy released used for all nuclei

◆ GetFissionReleasedFile()

string MURE::GetFissionReleasedFile ( )

inline

< Get fission energy released file used

◆ GetFPASCIIFileName()

char * MURE::GetFPASCIIFileName ( )

inline

< returns ASCII file name for FP

◆ GetFPBinaryFileName()

char * MURE::GetFPBinaryFileName ( )

inline

< returns Binary file name for FP yield

◆ GetFromVolumeCalculation()

bool MURE::GetFromVolumeCalculation ( )

inline

◆ GetGlobalNucleiVector()

vector< Nucleus_ptr > & MURE::GetGlobalNucleiVector ( )

inline

< Returns the global nuclei vector

◆ GetKeepOnlyFissionProductSelectionFN()

string MURE::GetKeepOnlyFissionProductSelectionFN ( )

inline

< return the file name that contains the FP to keep

◆ GetKeff()

double MURE::GetKeff ( )

inline

< returns the value of the total combinated estimation of the Keff of the current run

◆ GetKeff_Err()

double MURE::GetKeff_Err ( )

inline

< returns the error associated to the value of the total combinated estimation of the Keff of the current run

◆ GetListOfWantedNucleiFN()

string MURE::GetListOfWantedNucleiFN ( )

inline

< returns the List of Wanted Nuclei file name

◆ GetMassDataFileName()

string MURE::GetMassDataFileName ( )

inline

< returns the name of the isotopes masses file

◆ GetMaterialForPartialTallyNormalization()

vector< int > MURE::GetMaterialForPartialTallyNormalization ( )

inline

◆ GetMaterialNum()

int MURE::GetMaterialNum ( )

inline

< set last Material number

◆ GetMaterialVector()

vector< Material * > & MURE::GetMaterialVector ( )

inline

< returns the Material Vector.

◆ GetMaximumNumberOfDigitInEvolution()

int MURE::GetMaximumNumberOfDigitInEvolution ( )

inline

get the maximum Number of digits to number evolution (default = 3)

◆ GetMCExec()

string MURE::GetMCExec ( )

inline

<Get the name of the MC exec.

◆ GetMCInputBaseFileName()

string MURE::GetMCInputBaseFileName ( )

inline

< returns MC input Base file name (e.g. "inp" for an evo step inpXXX).

◆ GetMCNPThermoRunNumber()

int MURE::GetMCNPThermoRunNumber ( )

inline

< returns current MCNP run number for thermo

◆ GetMCNucleusThreshold()

double MURE::GetMCNucleusThreshold ( )

inline

< returns the threshold to decide to write a nucleus

◆ GetMCRunDirectory()

string MURE::GetMCRunDirectory ( )

inline

< Returns MC run directory name.

◆ GetMeshTally()

vector< MeshTally * > & MURE::GetMeshTally ( )

inline

< get the MCNP MeshTally

◆ GetMessageLevel()

int MURE::GetMessageLevel ( )

inline

< returns the minimal level of printed messages

◆ GetMode()

string MURE::GetMode ( )

inline

< returns particle transported mode

◆ GetMPIExec()

string MURE::GetMPIExec ( )

inline

< return mpiexec name

◆ GetMureSurfaceCardVector()

vector< SurfaceCard * > & MURE::GetMureSurfaceCardVector ( )

inline

< return the Surface Vector.

◆ GetNaturalIsotopeMassFileName()

string MURE::GetNaturalIsotopeMassFileName ( )

inline

< returns the name of the natural isotopes masses file

◆ GetNProcessors()

int MURE::GetNProcessors ( )

inline

< return processors number for pvm/mpich run

◆ GetNucleiChartFileName()

string MURE::GetNucleiChartFileName ( )

inline

< Get Name of the decay mode file.

◆ GetNucleiErrorMap()

map< Nucleus *, string > & MURE::GetNucleiErrorMap ( )

inline

< Returns the Nucleus Error storing map

◆ GetOMP()

int MURE::GetOMP ( )

inline

< return thread number for parallel MC run

◆ GetOutermostCell()

Cell * MURE::GetOutermostCell ( )

inline

< returns the most outer cell

◆ GetOutermostShape()

Shape_ptr MURE::GetOutermostShape ( )

inline

< returns the most outer shape_ptr

◆ GetPinCellVector()

vector< PinCell * > & MURE::GetPinCellVector ( )

inline

< Returns the PinCell Vector

◆ GetRandomSeed()

long int MURE::GetRandomSeed ( )

inline

◆ GetReactionDataFileName()

string MURE::GetReactionDataFileName ( )

inline

< Name of the Available reaction file for the chart.

◆ GetReactionDepth()

int MURE::GetReactionDepth ( )

inline

< return the maximum number of consecutive reactions

◆ GetReactionListInitMethod()

TFunctor * MURE::GetReactionListInitMethod ( )

inline

< return the fReactionListInitMethod

◆ GetReactionNamesForcedCutVector()

vector< string > MURE::GetReactionNamesForcedCutVector ( )

Get list of Reactions that have been set to be cut regardless of the cross sections and for all nuclei.

Sigma for these reactions are considered to be zero for all the rest of the simulation. Reaction are named by their SMURE name in the form similar to (N_2N) or (N_alpha) (n, gamma) is (N_gamma)

◆ GetReactionNumbersForcedCutVector()

vector< int > MURE::GetReactionNumbersForcedCutVector ( )

inline

Get list of Reactions that have been set to be cut regardless of the cross sections and for all nuclei.

Reaction are the MT number. (n, gamma) is 102, ...

◆ GetReactionThreshold()

double MURE::GetReactionThreshold ( )

inline

< Get the reaction ratio threshold for ReactionList

◆ GetRealMCInputFileName()

string MURE::GetRealMCInputFileName ( )

inline

< returns real MC input file name(eg. inpXXX).

◆ GetRunMultiMC()

int MURE::GetRunMultiMC ( )

inline

< get the number of run in multi MC

◆ GetRunTime()

double MURE::GetRunTime ( )

inline

< returns the difference between the start system clock time and now

◆ GetShapeOfUnknownArea()

vector< Shape * > & MURE::GetShapeOfUnknownArea ( )

inline

◆ GetShortestHalfLife()

double MURE::GetShortestHalfLife ( )

inline

< return the shortest half time to consider for evolution.

◆ GetSilentDebug()

bool MURE::GetSilentDebug ( )

inline

< return true for if SilentDebug is on

◆ GetSizeOfDepletionTallyList()

int MURE::GetSizeOfDepletionTallyList ( )

inline

◆ GetSource()

MCSource * MURE::GetSource ( )

Get the MC particle source.

◆ GetSourceSphereCenter()

double * MURE::GetSourceSphereCenter ( )

inline

< set the center of the source Sphere (stochastic volume calculation)

◆ GetSourceSphereRadius()

double MURE::GetSourceSphereRadius ( )

inline

< set the radius of the source Sphere (stochastic volume calculation)

◆ GetSpecialCard()

vector< string > & MURE::GetSpecialCard ( )

inline

< the vector of all Special card to write

◆ GetSpecialIsomerVector()

vector< SpecialIsomer > & MURE::GetSpecialIsomerVector ( )

inline

< return the SpecialIsomer vector

◆ GetSpecialSurfaceCard()

vector< string > MURE::GetSpecialSurfaceCard ( )

inline

< the vector of all Special Surface card

◆ GetSpectrumType()

double MURE::GetSpectrumType ( )

inline

returns the average energy for a spectrum type.

It returns 0.025eV, 500keV or 14MeV depending on the value set by MURE::SetSpectrumType

◆ GetTallyVector()

vector< MureTally * > & MURE::GetTallyVector ( )

inline

< Get the MureTally vector (abstract and must be overriden by MCNP/Serpent tallies)

◆ GetTemperatureMap()

TemperatureMap * MURE::GetTemperatureMap ( )

inline

< returns the Temperature Map

◆ GetTimeEnergyGrids()

vector< MureTimeEnergyGrid * > MURE::GetTimeEnergyGrids ( )

return the Time and/or Energy grid for tallies

◆ GetTotalAtoms()

double MURE::GetTotalAtoms ( )

Sum all the evolving atoms in the problem.

◆ GetTotalRunTime()

string MURE::GetTotalRunTime ( )

Returns a string that tells how long it is since MURE's code is running.

◆ GetTransformationVector()

vector< Transformation * > & MURE::GetTransformationVector ( )

inline

< return geometrical transformation vector

◆ GetTrueCellVector()

vector< Cell * > & MURE::GetTrueCellVector ( )

inline

< return True Cells vectors, i.e cells that are really in a geometry

◆ GetUserGeometryInputFile()

string MURE::GetUserGeometryInputFile ( )

inline

◆ GetVarianceVolumeWarning()

double MURE::GetVarianceVolumeWarning ( )

inline

< Get the warning message threshold in stochastic volume precision

◆ GetVolumeNPS()

int MURE::GetVolumeNPS ( )

inline

< Get source neutron number for stochastic volume calculation

◆ GetXSDIRName()

string MURE::GetXSDIRName ( )

inline

< return the xsdir/xsdata name

◆ GetZAIsForcedCutVector()

vector< ZAI * > MURE::GetZAIsForcedCutVector ( )

inline

Add a nuclei to the list of nuclei that have to be cut regardless of their half life.

Get list of nuclei that have to be cut regardless of their half life

◆ IncreaseMCNPThermoRunNumber()

void MURE::IncreaseMCNPThermoRunNumber ( )

inline

< This is used in ThermoCoupling.cxx

◆ IntegrateMassInEvolvingCells()

void MURE::IntegrateMassInEvolvingCells ( )

Sum all the evolving masses in the problem.

◆ IsAllowedToAddMaterial()

bool MURE::IsAllowedToAddMaterial ( )

inline

◆ IsAutoNumberFileName()

bool MURE::IsAutoNumberFileName ( )

inline

< return whether or not automatic numbering of MC/Thermo input file is performed

◆ IsAutoXSDIR()

bool MURE::IsAutoXSDIR ( )

inline

< whether or not MURE builds its own xsdir/xsdata file.

◆ IsBaseSummaryRead()

bool MURE::IsBaseSummaryRead ( )

inline

< Is the base summary has been read ?

◆ IsBetaCalculation()

bool MURE::IsBetaCalculation ( )

inline

< if delayed neutron fraction calculation wanted - Flag needed before running MC

◆ IsElasticXSinDATA()

bool MURE::IsElasticXSinDATA ( )

inline

return TRUE if the Elastic XS are tallied and register in DATA for all evolving material

◆ IsESpectrum()

bool MURE::IsESpectrum ( )

inline

< say if tally for energy flux spectrum in all cells to true.

◆ IsEvolutionRequired()

bool MURE::IsEvolutionRequired ( )

inline

< returns true if evolution is needed

◆ IsInDepletionList()

bool MURE::IsInDepletionList	(	ZAI *	deplzai,
		int	matnum
	)

◆ IsMCNPUserDefinedGeometry()

bool MURE::IsMCNPUserDefinedGeometry ( )

inline

< return true for a user defined MCNP geometry

◆ IsMPI()

bool MURE::IsMPI ( )

inline

< return whether MC's MPI is wanted

◆ IsNucleusUsedForBeta()

bool MURE::IsNucleusUsedForBeta	(	int	Z,
		int	A
	)

Check on used or unused nucleus in delayed neutron fraction calculation.

◆ IsOnlyFissionProductSelection()

bool MURE::IsOnlyFissionProductSelection ( )

inline

< return true if all fission products are not used for evolution

◆ IsPVM()

bool MURE::IsPVM ( )

inline

< return whether MC's PVM is wanted

◆ IsSameTimeEnergyGridForAllTallies()

bool MURE::IsSameTimeEnergyGridForAllTallies ( )

inline

< True if all tallies use the same TimeEnergyGrid

◆ IsSerpentDopplerBroadening()

bool MURE::IsSerpentDopplerBroadening ( )

inline

◆ IsThermoHydraulicsRequired()

bool MURE::IsThermoHydraulicsRequired ( )

inline

◆ IsUseEnergyBinsFile()

bool MURE::IsUseEnergyBinsFile ( )

inline

< say if an specific energy binning has been provided.

◆ IsVolumeCalculationNeeded()

bool MURE::IsVolumeCalculationNeeded ( )

return true if a volume is missing for a cell in which a tally is asked

◆ KeepOnlyFissionProductSelection()

void MURE::KeepOnlyFissionProductSelection ( string name = "" )

Use only some fission products for evolution.

By default, all fission products available in nuclear library are used. But because it can take a very long time to run MC (lot of tallies), a first (approximated) run can be done using a selection of FP (see UserGuide). Only the selection of FP evolved, but other FP are included for neutron flux calculation. If user want to provide an other list of evolving FP, he/she can use file with Z A of each wanted FP THIS METHOD MUST BE CALLED BEFORE ANY Material DECLARATION.

Parameters

name	: the file containing wanted fission products to erase default selection.

◆ NextCellNum()

int MURE::NextCellNum ( )

returns the next Cell number available

◆ NextMaterialNum()

int MURE::NextMaterialNum ( )

returns the next Material number available

◆ NextSurfaceNum()

int MURE::NextSurfaceNum ( )

returns the next Surface number available

◆ NextTallyNum()

int MURE::NextTallyNum ( int i = 0 )

returns the next Tally number of type i available

-For Serpent, tallies (or detectors) have no type (i.e. i=0). The numbers increase by 1. -For MCNP, tallies are defined by type (1=current, 2=flux/surf, 4=flux/cell, ...) : the number for a type increase by 10

◆ NextTransformationNum()

int MURE::NextTransformationNum ( )

returns the next Transformation number available

◆ NextUniverseNum()

int MURE::NextUniverseNum ( )

returns the next Universe number available

◆ PrintBanner()

void MURE::PrintBanner ( )

private

Print a banner at start of MURE.

◆ PrintDepletionTallyList()

void MURE::PrintDepletionTallyList ( )

◆ PutElasticXSinDATA()

void MURE::PutElasticXSinDATA ( bool RegisterElasticXSinDATA = true )

inline

Force to register the ElasticXS in DATA for all evolving materials

◆ ReadBaseSummary()

void MURE::ReadBaseSummary ( )

Read the whole base summary file to increase speed.

◆ ReadSpecialIsomerFile()

void MURE::ReadSpecialIsomerFile ( map< int, map< int, map< int, ZAI * > > > & TheChart )

Read Special Isomer File (see SpecialIsomer class)

◆ RemoveMaterial()

void MURE::RemoveMaterial ( Material * TheMaterial )

Remove a Material from the Material vector.

◆ RemoveSurfaceCard()

void MURE::RemoveSurfaceCard ( vector< SurfaceCard * > ::iterator it )

remove a Surface to the Surface Vector.

◆ RemoveXSDIR()

void MURE::RemoveXSDIR ( )

remove xsdir/xsdata file.

◆ ResetAllBetaValues()

void MURE::ResetAllBetaValues ( )

put to zero all values (used at each calculation done : needed if severals calculations of beta are required in a same MURE calculation

◆ ResetDetectorVector()

void MURE::ResetDetectorVector ( )

Remove all tallies.

◆ RunMC()

void MURE::RunMC ( )

Run MC code!!!!

◆ RunMultiMC()

void MURE::RunMultiMC ( int n )

Run n MC codes and adds MC detector output file to make one (mean and error of the n runs for all tallys &keff).

◆ SaveBackupMureFile()

void MURE::SaveBackupMureFile ( string TheFileName )

Backup copy of the MURE source file (.cxx) in the MC run directory.

◆ SetAllFissionEnergies()

void MURE::SetAllFissionEnergies ( double Energy = 2.0e08 )

Set the fission energy released for all nuclei.

By default, Fission energy released are taken from a linear interpolation of data in W.H. Walerk, Mass balance estimate of energy per fission in reactors AECL-3109, 1968. For data given in this ref., these values overide the linear fit. Using this method allow to impose a global value for fission energy released (for all fissile nuclei). THIS METHOD MUST BE CALLED BEFORE ANY Material DECLARATION.

Parameters

Energy : the fission energy released for all material in eV (Default=200MeV).

◆ SetAutoNumberFileName()

void MURE::SetAutoNumberFileName ( bool flag )

inline

< set whether or not automatic numbering of MC/Thermo input file is performed

◆ SetAutoXSDIR()

void MURE::SetAutoXSDIR ( bool cond = true )

inline

Tells MURE to build automatically the xsdir/xsdata file.

This method allow to build a specific xsdir/xsdata file (containing path to ACE lib) for the nuclei used by the MURE input file. MURE is looking for information in the BaseSummary.dat file (path to that file is set via gMURE->SetDATADIR()). If no xsdir/xsdata exists in the running directory (set via gMURE->SetMCRunDirectory()), the xsdir/xsdata is built, else the existing one is used. By default, MCNP::Connector set the name of the built file to xsdir and Serpent::Connector set it to xsdata.

◆ SetBasePriority()

void MURE::SetBasePriority ( BasePriority * UserBasePriority )

Set the BasePriority.

◆ SetBaseSummaryFileName()

void MURE::SetBaseSummaryFileName ( string name = "BaseSummary.dat" )

set the BaseSummary.dat file name

◆ SetBetaCalculation()

void MURE::SetBetaCalculation ( bool flag = true )

inline

< use if delayed neutron fraction calculation wanted - Flag needed before running MC

◆ SetCOBRACalculationRequired()

void MURE::SetCOBRACalculationRequired ( int NumberOfIteration )

inline

◆ SetComment()

void MURE::SetComment ( string Comment = "This comment is never read..." )

inline

< The First comment line of MC file

◆ SetConnectorPlugin()

void MURE::SetConnectorPlugin ( ConnectorPlugin * plugin )

Set the output plugin's connector.

The call to this method is MANDATORY ; it is use to choose which kind of output is used between MCNP or Serpent code. It MUST BE USED BEFORE ANY MCSource ; the better is to put the following lines just after the main declaration:

using namespace XXX
int main()
{
    Connector *plugin=new Connector;
    gMURE->SetConnectorPlugin(plugin);
    ...
}

The namespace XXX is either "MCNP" or "Serpent" ; this will define either a "MCNP" connector, either a "Serpent" Connector.

Parameters

plugin : a MCNP::Connector or a Serpent::Connector.

◆ SetDATADIR()

void MURE::SetDATADIR ( string name = "" )

Set the data directory path to name.

if name is not provided (default),

if the Environment variable DATADIR is set (via a setenv/export in csh/bash), the data directory path is set to this variable
else it is set to ../../data/ if name is provided, DATADIR is set to name.

◆ SetDisableViolentSimplifyNode()

void MURE::SetDisableViolentSimplifyNode ( bool flag )

inline

◆ SetEnsdfDATADIR()

void MURE::SetEnsdfDATADIR ( string name = "" )

Set the ENSDF data directory path to name.

if name is not provide, it is set to fDataDirectoryName/ENSDF/

◆ SetESpectrum()

void MURE::SetESpectrum ( )

inline

< set flag to tally for energy flux spectrum in all cells to true.

◆ SetESpectrumTallyNumber()

void MURE::SetESpectrumTallyNumber ( int num )

inline

< set the number of theEnergy spectrum global tally

◆ SetEvolutionRequired()

void MURE::SetEvolutionRequired ( bool flag = true )

inline

< say that evolution is needed or not

◆ SetExtendedReactionList()

void MURE::SetExtendedReactionList ( )

inline

◆ SetFissionReleasedFile()

void MURE::SetFissionReleasedFile ( string name )

Set the fission energy released from a file.

By default, Fission energy released are taken from a linear interpolation of data in W.H. Walerk, Mass balance estimate of energy per fission in reactors AECL-3109, 1968.

Use this method to give explicit values for nuclei described in the provided file. The file must contain Z A fission_energy_released_in_eV for each wanted isotopes. THIS METHOD MUST BE CALLED BEFORE ANY Material DECLARATION.

Parameters

name	: the file containing fission energy released.

◆ SetFPASCIIFileName()

void MURE::SetFPASCIIFileName ( string name = "FPavailable.dat" )

Set Name ASCII file for Fission Products.

This file has contains for some fissiles, the FP yield record position and availlable energies in the binary file. It has been made by the MURE/Evo/FP/GenerateFPYield ; data are extracted from endfb6-y.asc. WARNING: The data directory path has to be set by SetDATADIR().

◆ SetFPBinaryFileName()

void MURE::SetFPBinaryFileName ( string name = "FPyield.bin" )

Set Name Binary file for Fission Products.

This file contains for some fissiles, the FP yield at availlable energies. It has been made by the MURE/Evo/FP/GenerateFPYield ; data are extracted from endfb6-y.asc. WARNING: The data directory path has to be set by SetDATADIR().

◆ SetFromVolumeCalculation()

void MURE::SetFromVolumeCalculation ( bool flag )

inline

◆ SetKeff()

void MURE::SetKeff ( double k )

inline

< Set the Keff to k

◆ SetKeff_Err()

void MURE::SetKeff_Err ( double dk )

inline

< Set the Keff error to dk

◆ SetLastTallyNum()

void MURE::SetLastTallyNum ( int Num )

Set the last tally number to Num (for MCNP user input geometry)

◆ SetListOfWantedNucleiFN()

void MURE::SetListOfWantedNucleiFN ( string name )

inline

Set the List of Wanted Nuclei file name.

WARNING: USE WITH CARE!!!

This method allows user to specify the wanted nuclei for a evolution: ONLY nuclei defined as "Z A" in this file are used in the evolution. If nuclei X and Y are in the file and if the initial composition contains only X, then Y will be present in the evolution tree if and only if Y is a direct daugther (reaction or decay) of X. But if X->T->Y (decay or reaction) then Y will not be produced. The given list of nuclei will also imposed the allowed reactions: for examples, capture on X is allowed only if X(Z, A) and Y(Z, A+1) is in the file ; THUS TO ALLOW FISSION YOU MUST PROVIDE AT LEAST A SMALL SET OF FP in in the file.

YOU ALSO MUST DEFINE IN THIS FILE THE NUCLEI OF YOUR INITIAL COMPOSITIONS.

◆ SetMassDataFileName()

void MURE::SetMassDataFileName ( string name = "Mass.dat" )

Set name of the isotopes mass file.

This has been made from "The 2003 Atomic Mass Evaluation" of the National Nuclear Data Center of BNL. WARNING: The data directory path has to be set by SetDATADIR().

◆ SetMaterialForPartialTallyNormalization() [1/2]

void MURE::SetMaterialForPartialTallyNormalization ( int matnum )

◆ SetMaterialForPartialTallyNormalization() [2/2]

void MURE::SetMaterialForPartialTallyNormalization ( vector< int > matnum )

inline

◆ SetMaterialNum()

void MURE::SetMaterialNum ( int num )

inline

< set last Material number

◆ SetMaximumNumberOfDigitInEvolution()

void MURE::SetMaximumNumberOfDigitInEvolution ( int maxi )

inline

set the maximum Number of digits to number evolution (default = 3)

◆ SetMCExec()

void MURE::SetMCExec ( string Exec )

inline

Set Name of MC exec.

By default, MCNP::Connector set it to "mcnp5" and Serpent::Connector set it to "sss2"

◆ SetMCInputFileName()

void MURE::SetMCInputFileName ( string Name )

inline

< Set MC Input Base File Name (e.g. "inp" for an evo step inpXXX).

◆ SetMCNPThermoRunNumber()

void MURE::SetMCNPThermoRunNumber ( int value = 0 )

inline

< returns current MCNP run number for thermo

◆ SetMCNPUserDefinedGeometry()

void MURE::SetMCNPUserDefinedGeometry ( )

inline

< Tell that it is a user defined MCNP geometry

◆ SetMCNucleusThreshold()

void MURE::SetMCNucleusThreshold ( double epsilon = 1e-6 )

inline

Set threshold to decide to write a nucleus in the MC input.

When writing a evolving material, a nucleus is written only if its normalized proportion p *Sigma_tot is greater than the threshold epsilon

◆ SetMCRunDirectory()

void MURE::SetMCRunDirectory	(	string	name = `"MUREOutput"`,
		string	option = `""`
	)

Set the MC run directory name.

This is used for evolution.

Parameters

name	: Name of the directory where all input (for MC) and output files are written.
option	: what to do with an eventually already existing directory: if option="create", the existing directory is removed (and its content is lost) ; if option="keep", the existing directory (and its content) is kept else [default] a question is asked.

◆ SetMessageLevel()

void MURE::SetMessageLevel ( int level )

inline

Set the minimal level of printed messages.

This method control the MURE output message verbosity. Low level correspond to a complete verbose output. Possible value are

LOG_LEVEL_DEBUG or 0,
LOG_LEVEL_WARNING or 2,
LOG_LEVEL_IMPORTANT_WARNING or 3,
LOG_LEVEL_ALERT or 4,
and  LOG_LEVEL_ERROR or 5.

In general the level 3(=LOG_LEVEL_IMPORTANT_WARNING) is enough to have all important informations. The default value is 2(=LOG_LEVEL_WARNING). If you call this method in the middle of your code, message level will be default up to the line where it is called: put it at the begining of your main to be effective from the beginning.

Parameters

level : the level of message

◆ SetMode()

void MURE::SetMode ( string MyMode = "N" )

inline

< Set a general run mode (for MCNP only)

◆ SetModeN()

void MURE::SetModeN ( )

inline

< Set the Neutron mode (default and only possibility for Serpent)

◆ SetModeNE()

void MURE::SetModeNE ( )

inline

< Set the Neutron Electron mode (ignored for Serpent output)

◆ SetModeNP()

void MURE::SetModeNP ( )

inline

< Set the Neutron Photon mode (ignored for Serpent output)

◆ SetModeNPE()

void MURE::SetModeNPE ( )

inline

< Set the Neutron Photon Electron mode (ignored for Serpent output)

◆ SetModeP()

void MURE::SetModeP ( )

inline

< Set the Photon mode (ignored for Serpent output)

◆ SetModePE()

void MURE::SetModePE ( )

inline

< Set the Photon Electron mode (ignored for Serpent output)

◆ SetMPI()

void MURE::SetMPI	(	int	n_processor,
		string	MPIExec = `"mpiexec"`
	)

Set the processor number in MPICH mode.

Set the processor number in MPICH mode. In order to run MC code in parallel (MPICH), you have to specify the wanted processor number to MC's exec.

TO BE NOTICED : this is effective ONLY if the MC exec is compiled with MPI support.

Parameters

n_processor	: number of processor to use
MPIExec	: MPI exec name to launch mcnp/serpent (mpirun, mpiexec, ...).

◆ SetNaturalIsotopeMassFileName()

void MURE::SetNaturalIsotopeMassFileName ( string name = "NaturalIsotopeMass.dat" )

Set name of the NATURAL isotopes mass file.

This file has been made, by hand, from NNCD nuclear wallet card, to give mass to natural isotopes (i.e. with A=0). To be noticed that these isotopes cannot evolve WARNING: The data directory path has to be set by SetDATADIR().

◆ SetNucleiChartFileName()

void MURE::SetNucleiChartFileName ( string name = "chart.jeff3.1.1" )

Set Name of the decay mode file.

Name of the ascii file containing all decay modes (including spontaneus fission) of all nuclei. The default value, chart.jeff3.1.1 is a compilation from Jeff 3.1/ensdf/LNHB and Nubase-2003 library ; it is extracted from M.A. Kellett, O. Bersillon, R.W. Mills, NUCLEAR ENERGY AGENCY, NEA No. 6287, OECD 2009. This file contains Z A nucleus_symbol isomeric_state period in second (or 0 for stable) decay_mode_i:branching_ration_in-%. (or s for stable).

User can provide a other file that have the same format (look to this file for example). WARNING: The data directory path has to be set by SetDATADIR().

◆ SetOMP()

void MURE::SetOMP ( int n_thread )

inline

Set the processor number in OpenMP mode.

Set the processor number in OpenMP mode. In order to run MC code in multithreading (OpenMP), you have to specify the number of thread for the MC's exec.

TO BE NOTICED : this is effective ONLY if the MC exec is compiled with multi-threading support. This can be used together with MPI.

Parameters

n_thread : number of thread to use

< set the number of threads for parallel MC run

◆ SetOutermostCell()

void MURE::SetOutermostCell ( Cell * c )

inline

◆ SetOutermostShape()

void MURE::SetOutermostShape ( Shape_ptr s )

inline

Set the most outer shape.

The most external shape has to be defined when using for example poison control ; indeed, a keff evaluation is done using formula keff=(creation)/(absorption+leakage). The leakage are evaluated using F2 tally on the most outer shape of the geometry (i.e. the whole exterior). It can be also used for volume calculation

Parameters

s	: the most outer Shape_ptr (exterior)

◆ SetPower()

void MURE::SetPower ( double Power )

inline

< set the reactor power (in watts)

◆ SetPVM()

void MURE::SetPVM ( int n_processor )

Set the processor number in PVM mode.

Set the processor number in PVM mode. In order to run MC in parallel (PVM), you have to specify the wanted processor number to MC's PVM exec. In case of MCNP4B the method for calling MCNP's PVM exec is different (don't use "x" in proc number: for MCNP4B, one use mcnp task n_processor n=... whereas for mcnp4C and above, one use mcnp task n_processorx1 n=...).

TO BE NOTICED : this is effective ONLY if the MC exec is compiled with PVM support.

Parameters

n_processor : number of processor to use

◆ SetReactionDataFileName()

void MURE::SetReactionDataFileName ( string name = "AvailableReactionChart.dat" )

Set Name of the Available reaction file for the chart.

This file is built by MURE/Evo/utils/CheckReaction and has to be put in the data directory. The aim of this file is to take all isotopes of "chart.jeff3.1.1" or file provided by user and to say whether or not cross-sections are available (cross-sections are read in BaseSummary.dat). WARNING: The data directory path has to be set by SetDATADIR().

◆ SetReactionDepth()

void MURE::SetReactionDepth ( int depth = 10000 )

inline

Set the maximum number of consecutive reactions to consider for evolution.

Set the maximum number of consecutive reactions to consider for evolution. Stop the reaction "propagation" after the ith consecutive reactions in order to cut nuclei trees. The default (10000) is so big that it allows all reactions (i.e. no cut on nuclei tree)

Parameters

depth : the maximum number of consecutive reactions(default=10000)

◆ SetReactionListInitMethod()

void MURE::SetReactionListInitMethod ( TSpecificFunctor< ReactionList > * method )

inline

Set the ReactionList Init Method.

The ReactionList init methods is used by NucleiTree to find the allow reaction for a given ZAI. Up to now, you can use among 4 methods: ReactionList::InitJon(), ReactionList::InitFrancoTest(), ReactionList::InitFissionTest() and ReactionList::InitPTO(). The default is the latter one and it is based on automatic reaction detections according to binary nuclear data libraries. The way of given an other ReactionList Init method is: gMURE->SetReactionListInitMethod(new TSpecificFunctor<ReactionList>(0, &ReactionList::InitJon)); it has to be used before any cell definition.

Parameters

method : a pointer on the ReactionList Init Method

◆ SetReactionNumbersForcedCutVector()

void MURE::SetReactionNumbersForcedCutVector ( vector< int > ReactionNumbersForcedCutUser )

inline

Set list of Reactions that have to be cut regardless of the cross sections and for all nuclei.

Sigma for these reactions are considered zero. Reaction are named by their SMURE number and not their MT number. (n, gamma) is 1 and not 102

◆ SetReactionThreshold()

void MURE::SetReactionThreshold ( double sigma = 1e-2 )

inline

< Set the reaction ratio threshold for ReactionList

◆ SetRealMCInputFileName()

void MURE::SetRealMCInputFileName ( string Name )

inline

< Set Real MC Input File Name (e.g. inpXXX).

◆ SetRunMultiMC()

void MURE::SetRunMultiMC ( int n )

inline

< set the number of run in multi MC runs

◆ SetSameTimeEnergyGridForAllTallies()

void MURE::SetSameTimeEnergyGridForAllTallies ( )

set that all tallies must use the same TimeEnergyGrid

◆ SetSerpentDopplerBroadening()

void MURE::SetSerpentDopplerBroadening ( )

inline

◆ SetShortestHalfLife()

void MURE::SetShortestHalfLife ( double time = 3600 )

inline

Set the shortest half time to consider for evolution.

Replace each nucleus (ZAI) of half life time shorter than this time by its daugthers. Erase your local ReactionList directory when you use/modify the time of this method.

Parameters

time	: the shortest half time in second (default=3600s)

◆ SetSilentDebug()

void MURE::SetSilentDebug ( bool flag = true )

inline

Set the Silent Debug mode on.

If SetMessageLevel is set to LOG_LEVEL_DEBUG, all LOG_DEBUG are printed. But if SetSilentDebug is used, only LOG_DEBUG of methods where a "int DODEBUG=1" is inserted are printed. Thus, using LOG_DEBUG, avoid to comment all "cout" when no debugging is wanted.

◆ SetSource()

void MURE::SetSource ( MCSource * source )

Set the MC particle source.

The source declaration should be either a MCNPSource, either a SerpentSource ; for the latter only neutron are transported. You MUST use a SerpentSource withe the Serpent namespace and a MCNPSource with the MCNP namespace.

Parameters

source : either a MCNPSource or a SerpentSource

◆ SetSourceSphereCenter()

void MURE::SetSourceSphereCenter ( double * O )

inline

< set the center of the source Sphere (stochastic volume calculation)

◆ SetSourceSphereRadius()

void MURE::SetSourceSphereRadius ( double R )

inline

< set the radius of the source Sphere (stochastic volume calculation)

◆ SetSpecialCard()

void MURE::SetSpecialCard	(	int	i,
		string	str
	)

inline

< the vector of all Special card to write

◆ SetSpecialIsomerFileName()

void MURE::SetSpecialIsomerFileName ( string name = "" )

Set Special Isomer File name (see SpecialIsomer class)

◆ SetSpectrumType()

void MURE::SetSpectrumType ( string type = "thermal" )

Set the spectrum type.

The spectrum could be "thermal", "fast", or "spallation". This allows to take fission yield at 0.025eV, 500keV or 14MeV respectively. (To be noticed, if FP yield are not allways given for all these energies ; then the closest energy is chosen).

Parameters

type	: the spectrum type (thermal", "fast", or "spallation")

◆ SetTallyNormalizationFactor()

void MURE::SetTallyNormalizationFactor ( double val )

inline

set Flux normalization value.

The call to this method, by default, do not keep the power constant, but the flux.

Parameters

val	: the number of particule per sec

◆ SetTemperatureMap() [1/2]

void MURE::SetTemperatureMap	(	double *	T = `nullptr`,
		int	NT = `0`
	)

Set The TemperatureMap.

if NT=0, default temperature map is built (19 temperatures from 0, 300, 400...1900K) .

Parameters

T	: array of temperature (the map)
NT	: size of this array

◆ SetTemperatureMap() [2/2]

void MURE::SetTemperatureMap ( TemperatureMap * TM )

◆ SetThermoHydraulicsRequired()

void MURE::SetThermoHydraulicsRequired ( bool flag = true )

inline

◆ SetUseNewRandomSeed()

void MURE::SetUseNewRandomSeed ( long int dbcn = - 1 )

Use a new Random Seed number in MC run.

Set the random seed to dbcn if >0, else take a random value=1+2*rand().

Parameters

dbcn	: the random seed use in the MC run.

◆ SetUserDefaultHydraulicModel()

void MURE::SetUserDefaultHydraulicModel ( )

inline

Build automatic tallies in multigroup run without using the perturbative materials.

Automatic tallies used in an reaction rates calculation (phi(E)) are build It doesn't build perturbatives materials which are used in a case of a fuel depletion IT MUST BE CALLED CAREFULLY it is used in some case of thermal coupling when the number of materials are very important and can overload the maximum number of materials allowed in MCNP (100, 000) Updates phi(E) in multigroup run.

Only if BuildTHTalliesMultiGroup() is used

◆ SetUserGeometryInputFile()

void MURE::SetUserGeometryInputFile ( string name )

set the file name of the user defined MCNP geometry

◆ SetVarianceVolumeWarning()

void MURE::SetVarianceVolumeWarning ( double variance = 0.005 )

inline

< set the warning message threshold in stochastic volume precision

◆ SetVolumeNPS()

void MURE::SetVolumeNPS ( int nps = 400000 )

inline

< Set source neutron number for stochastic volume calculation

◆ SetXSDIRName()

void MURE::SetXSDIRName ( string acefile = "xsdir" )

inline

< set the xsdir/xsdata name

◆ SetZAIsForcedCutVector()

void MURE::SetZAIsForcedCutVector ( vector< ZAI * > ZAIsForceCutUser )

inline

Set list of nuclei that have to be cut regardless of their half life

◆ StartGUI()

void MURE::StartGUI ( )

Start the TCL GUI (Examtree.tcl) to see the Nueclei Tree.

◆ SumAllCellToBetaCalculation()

void MURE::SumAllCellToBetaCalculation ( )

Sums all beta by cell to obtain the total one.

◆ SumUpNucleiError()

void MURE::SumUpNucleiError ( ostream & Out = cerr )

Write in the chosen output the list of the nuclei with problems, and the type of these problems.

◆ UpdateKeff()

void MURE::UpdateKeff ( )

Read MC's Output file to set Keff value (can be used without evolution)

◆ UseEnergyBinsFile()

void MURE::UseEnergyBinsFile ( string Filename )

uses the filename provided for building the energy binning. (single column file with energies in eV)!

◆ WriteGlobalNucleiTree()

void MURE::WriteGlobalNucleiTree ( string fname = "UnionOfAllTrees.dat" )

write the union of all nuclear trees (evolving nuclei in globalnucleivector)

Friends And Related Symbol Documentation

◆ MureTrojan

friend class MureTrojan

friend

Member Data Documentation

◆ fAddMaterial

bool MURE::fAddMaterial

private

false if you don't want to add material in fMaterialVector

◆ fAutoNumberFileName

bool MURE::fAutoNumberFileName

private

whether or not use automatic numbering of evolving MC input files (e.g. XXX for inp base name)

◆ fBasePriority

BasePriority* MURE::fBasePriority

private

The Base Priority wish.

◆ fBaseSummary

vector< string > MURE::fBaseSummary

private

Table to put the Base summary file in memory for quick access.

◆ fBaseSummaryFileName

string MURE::fBaseSummaryFileName

private

Name of the Base Summary file to find info on availlable data bases.

◆ fBetaCalculationRequired

bool MURE::fBetaCalculationRequired

private

True if delayed neutron fraction calculation wanted.

◆ fCellOfUnknownVolume

vector< Cell *> MURE::fCellOfUnknownVolume

private

Vector of Cell of Unknown volume used in tallies.

◆ fCellVector

vector< Cell *> MURE::fCellVector

private

the vector of all Cell to write

◆ fCOBRACalculationRequired

bool MURE::fCOBRACalculationRequired

private

If we are running iteration between MCNP and COBRA.

◆ fComment

string MURE::fComment

private

first line comment in MC file

◆ fConnectorPlugin

ConnectorPlugin* MURE::fConnectorPlugin

private

Pointer to couple MURE with MCNP, Serpent...

◆ fDataDirectoryName

string MURE::fDataDirectoryName

private

data directory name for all data files

◆ fDisableViolentSimplifyNode

bool MURE::fDisableViolentSimplifyNode

private

Material number for which tallying of isotopes should happen.

disable the violent symplification in Node::Simplify for 1st attempt to obtain cell volumes

◆ fEnergyBins

vector< double > MURE::fEnergyBins

private

vector of Energies in eV read from the fEnergyBinsFileName file that will be put in a MureTimeEnergyGrid by a tally

◆ fEnergyBinsFileName

string MURE::fEnergyBinsFileName

private

name of the user provided data file with desired energy binning.

◆ fEnsdfDataDirectoryName

string MURE::fEnsdfDataDirectoryName

private

data directory name for ENSDF data (added B.LENIAU)

◆ fESpectrumTallyNumber

int MURE::fESpectrumTallyNumber

private

Number of the Tally for Energy flux spectrum.

◆ fEvolutionRequired

bool MURE::fEvolutionRequired

private

If an Evolution is required (i.e. many MC's run steps)

◆ fEvolutionSolver

EvolutionSolver* MURE::fEvolutionSolver

private

The evolution pointer connector.

◆ fExtendedReactionList

bool MURE::fExtendedReactionList

private

◆ fFirstContinuousZIndexInBaseSummary

vector< int > MURE::fFirstContinuousZIndexInBaseSummary

private

Store line number where to go to find Z entries for continuous energy in Base Summary.

◆ fFirstThermalZIndexInBaseSummary

vector< int > MURE::fFirstThermalZIndexInBaseSummary

private

Store line number where to go to find Z entries for thermal data in Base Summary.

◆ fFissionReleasedFN

string MURE::fFissionReleasedFN

private

The name of fission energy released file.

◆ fForceConstRandomSeedForMultiRun

bool MURE::fForceConstRandomSeedForMultiRun

private

◆ fFPASCIIFileName

string MURE::fFPASCIIFileName

private

Name ASCII file for Fission Products.

◆ fFPBinaryFileName

string MURE::fFPBinaryFileName

private

Name Binary file for Fission Products.

◆ fGlobalNucleiVector

vector< Nucleus_ptr > MURE::fGlobalNucleiVector

private

Union of all the nuclei used in all the evolving materials.

◆ fGlobalZAIDepletionTallyList

vector<ZAI *> MURE::fGlobalZAIDepletionTallyList

private

◆ fIsAutoXSDIR

bool MURE::fIsAutoXSDIR

private

Whether the XSDIr must be automtically built or not.

◆ fIsBannerPrinted

bool MURE::fIsBannerPrinted = false

staticprivate

Print the banner only once.

◆ fIsBaseSummaryRead

bool MURE::fIsBaseSummaryRead

private

Flag to say if Base summary has been read in or not.

◆ fIsConnectorPluginCalled

bool MURE::fIsConnectorPluginCalled

private

◆ fIsFromVolumeCalculation

bool MURE::fIsFromVolumeCalculation

private

Flag that control copy/delete of MURE depending on the call time.

◆ fIsSameTimeEnergyGridForAllTallies

bool MURE::fIsSameTimeEnergyGridForAllTallies

private

◆ fKeepOnlyFissionProductSelection

bool MURE::fKeepOnlyFissionProductSelection

private

true if all fission products are not used for evolution

◆ fKeepOnlyFissionProductSelectionFN

string MURE::fKeepOnlyFissionProductSelectionFN

private

The file name that contains the FP to keep.

◆ fKeff

double MURE::fKeff

private

Current Keff of the problem.

◆ fKeff_Err

double MURE::fKeff_Err

private

Current Error on current Keff of the problem.

◆ fLastCellNum

int MURE::fLastCellNum

private

last Cell number used

◆ fLastMaterialNum

int MURE::fLastMaterialNum

private

last Material number used

◆ fLastSurfaceNum

int MURE::fLastSurfaceNum

private

last Surface number used

◆ fLastTallyNum

int MURE::fLastTallyNum[8]

private

last Tally of type i number used

◆ fLastTransformNum

int MURE::fLastTransformNum

private

last Transformation used

◆ fLastUniverseNum

int MURE::fLastUniverseNum

private

last Universe number used

◆ fListOfWantedNucleiFN

string MURE::fListOfWantedNucleiFN

private

File name of the list of wanted nuclei (and ONLY these)

◆ fMassDataFileName

string MURE::fMassDataFileName

private

File where to find the nuclear mass data.

◆ fMaterialForTallyNormalization

vector<int> MURE::fMaterialForTallyNormalization

private

Material numbers to be used for partial normalization to power.

◆ fMaterialUpdated

bool MURE::fMaterialUpdated

private

flag indicating that Materials have been updated

◆ fMaterialVector

vector< Material *> MURE::fMaterialVector

private

the vector of all Material to write

◆ fMatnumDepletionTallyList

vector<int> MURE::fMatnumDepletionTallyList

private

List of ZAIs that should be tallied in multigroup mode.

◆ fMCExec

string MURE::fMCExec

private

name of MC exec(mcnp5, sss2, ...)

◆ fMCInputBaseFileName

string MURE::fMCInputBaseFileName

private

The Base Name of the input file name generated for MC (such as "inp")

◆ fMCInputFileName

string MURE::fMCInputFileName

private

The final file name for MC (such as inpXXX for step XXX of an Evolution)

◆ fMCNPThermoRunNumber

int MURE::fMCNPThermoRunNumber

private

Current ThermoHydraulics MCNP Run Number.

◆ fMCNucleusThreshold

double MURE::fMCNucleusThreshold

private

the threshold to decide to write a nucleus in MC's input file

◆ fMCRunDirectory

string MURE::fMCRunDirectory

private

name of the directory in which the MC output files are written

◆ fMeanEnergySpectrum

double MURE::fMeanEnergySpectrum

private

The mean energy spectrum to consider (for FP yield)

◆ fMeshTallyVector

vector< MeshTally *> MURE::fMeshTallyVector

private

the vector of all Mesh Tally to write

◆ fMessageLevel

int MURE::fMessageLevel

private

message level treshold

◆ fMode

string MURE::fMode

private

the mode MCNP card (N, N P, N E, P E, P, or N P E). For Serpent, only mode N is taken.

◆ fMPI

bool MURE::fMPI

private

True for a MC MPI parallel run.

◆ fMPIExec

string MURE::fMPIExec

private

MPI Exec name to launch MC (mpiexec, mpirun, ...)

◆ fMUREInputGeometry

bool MURE::fMUREInputGeometry

private

True if the geometry is a MURE geometry and false for a user defined MCNP geometry.

◆ fNaturalIsotopeMassFileName

string MURE::fNaturalIsotopeMassFileName

private

File where to find the nuclear mass data for natural isotopes.

◆ fNbOfDigitInEvo

int MURE::fNbOfDigitInEvo

private

Number of digits to number evolution (default = 3)

◆ fNprocessors

int MURE::fNprocessors

private

Number of processor in PVM/MPICH/OpenMP MC run.

◆ fNSigmaFisPhiTotalOfNucleus

vector< double > MURE::fNSigmaFisPhiTotalOfNucleus

private

For a nucleus : sum of reaction rates.

◆ fNucleiChartFileName

string MURE::fNucleiChartFileName

private

File where to find radioactive decays modes &branching ratios.

◆ fNucleiErrorMap

map< Nucleus *, string > MURE::fNucleiErrorMap

private

Map to store problems about nuclei (Errors in Nucleus::FindCode, more precisely)

◆ fNumberOfIterationNTH

int MURE::fNumberOfIterationNTH

private

Number of Neutronics/thermalhydraulics iterations required at each fuel depletion step.

◆ fOMP

int MURE::fOMP

private

number of Threads in OpenMP parallel compilation of MC.

◆ fOutermostCell

Cell* MURE::fOutermostCell

private

◆ fOutermostShape

Shape_ptr MURE::fOutermostShape

private

The outer Shape to evaluate neutron leakage or volume.

◆ fPinCellVector

vector< PinCell *> MURE::fPinCellVector

private

the vector of all PinCell to write

◆ fPVM

bool MURE::fPVM

private

True for a MC PVM parallel run.

◆ fRandomSeedGenerator

long int MURE::fRandomSeedGenerator

private

the random seed value

◆ fReactionDataFile

string MURE::fReactionDataFile

private

File where to find if reaction are available or not.

◆ fReactionDepth

int MURE::fReactionDepth

private

The maximum number of consecutive reactions.

◆ fReactionListInitMethod

TFunctor* MURE::fReactionListInitMethod

private

ReactionList Init tMethod (e.g. InitPTO())

◆ fReactionNumbersForcedCutVector

vector< int > MURE::fReactionNumbersForcedCutVector

private

List of reactions type that should not be considered even if the cross section are bigger than fReactionThreshold. Given by SMURE reaction number and not MT number.

◆ fReactionThreshold

double MURE::fReactionThreshold

private

the reaction ratio threshold for ReactionList

◆ fReactivityCoeffsSource

MCSource* MURE::fReactivityCoeffsSource

private

Source Used for reactivity coeffs calculations.

◆ fRegisterElasticXSinDATA

bool MURE::fRegisterElasticXSinDATA

private

true if elastic scattering cross section registered in the DATA file next to other XSs

◆ fRunMultiMC

int MURE::fRunMultiMC

private

Number of run in a Multi MC mode.

◆ fSerpentDopplerBroadening

bool MURE::fSerpentDopplerBroadening

private

◆ fSetAllFissionEnergies

double MURE::fSetAllFissionEnergies

private

If > zero, then we will override correct fission energies with this value.

◆ fSetESpectrum

bool MURE::fSetESpectrum

private

flag for tallying Energy Bins in all "True" Cells (Cells which do contain a non void material)

◆ fShapeOfUnknownArea

vector< Shape *> MURE::fShapeOfUnknownArea

private

Vector of Shape_ptr of Unknown area used in tallies.

◆ fShortestHalfTimeThres

double MURE::fShortestHalfTimeThres

private

The shortest half life considered for a nucleus.

◆ fSilentDebug

bool MURE::fSilentDebug

private

generally suppress the debug messages

◆ fSourceSphereO

double* MURE::fSourceSphereO

private

Sphere source center used in stochastic volume calculation.

◆ fSourceSphereR

double MURE::fSourceSphereR

private

Sphere source radius used in stochastic volume calculation.

◆ fSpecialCard

vector< string > MURE::fSpecialCard

private

the vector of all Special card to write

◆ fSpecialIsomerFileName

string MURE::fSpecialIsomerFileName

private

file that explain the particular treatment of special isomers

◆ fSpecialIsomerVector

vector< SpecialIsomer > MURE::fSpecialIsomerVector

private

vector of special Isomer treatment (like capture on Am241)

◆ fSpecialSurfaceCard

vector< string > MURE::fSpecialSurfaceCard

private

the vector of all Special Surface card

◆ fStartTime

int MURE::fStartTime

private

The computer clock time at the start of the MURE run.

◆ fSurfaceCardVector

vector< SurfaceCard *> MURE::fSurfaceCardVector

private

the vector of all Surface to write (in the MC output format)

◆ fTallyVector

vector< MureTally *> MURE::fTallyVector

private

the vector of all Tally to write

◆ fTemperatureMap

TemperatureMap* MURE::fTemperatureMap

private

The temperature Map.

◆ fThermoHydraulicsRequired

bool MURE::fThermoHydraulicsRequired

private

If we are running MCNP many times with ThermoHydraulics.

◆ fTimeEnergyGrids

vector< MureTimeEnergyGrid *> MURE::fTimeEnergyGrids

private

◆ fTransformationVector

vector< Transformation *> MURE::fTransformationVector

private

the vector geometrical transformation

◆ fTrueCellVector

vector< Cell *> MURE::fTrueCellVector

private

Vector of all true Cells (which are non void) which shall be written to MC (which fNumber is not negative)

◆ fUseEnergyBinsFile

bool MURE::fUseEnergyBinsFile

private

flag for using a user provided data file with desired energy binning.

◆ fUserDefaultHydraulicModel

bool MURE::fUserDefaultHydraulicModel

private

Use specific hydraulics model for CHF.

◆ fUserGeometryInputFile

string MURE::fUserGeometryInputFile

private

The input file for the user given MCNP geometry.

◆ fVarianceVolumeWarning

double MURE::fVarianceVolumeWarning

private

warning message threshold in stochastic volume precision

◆ fVolumeNPS

int MURE::fVolumeNPS

private

source neutron number for stochastic volume calculation

◆ fXSDIRName

string MURE::fXSDIRName

private

the xsdir/ACElib name

◆ fZAIDepletionTallyList

vector< vector<ZAI *> > MURE::fZAIDepletionTallyList

private

List of ZAIs that should be tallied in multigroup mode for each material.

◆ fZAIsForcedCutVector

vector< ZAI *> MURE::fZAIsForcedCutVector

private

List of ZAI that should not be considered even if their half life are longer than fShortestHalfTimeThres.

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

/gpr/meplan/SMURE/source/include/MURE.hxx
/gpr/meplan/SMURE/source/src/MURE_buildfile.cxx
/gpr/meplan/SMURE/source/src/MURE_constructor.cxx
/gpr/meplan/SMURE/source/src/MURE_evol.cxx
/gpr/meplan/SMURE/source/src/MURE_misc.cxx

Public Member Functions

Friends

fraction of delayed neutron calculation

Detailed Description

Constructor & Destructor Documentation

◆ MURE() [1/2]

◆ MURE() [2/2]

◆ ~MURE()

Member Function Documentation

◆ AddCell()

◆ AddCellOfUnknownVolume()

◆ AddLineToBaseSummary()

◆ AddList2DepletionTallyList()

◆ AddMaterial()

◆ AddMeshTally()

◆ AddPinCell()

◆ AddReactionNameForcedCut()

◆ AddReactionNumberForcedCut()

◆ AddShapeOfUnknownArea()

◆ AddSpecialCard()

◆ AddSpecialSurface()

◆ AddSurfaceCard()

◆ AddTally()

◆ AddTimeEnergyGrid()

◆ AddToGlobalNucleiVector()

◆ AddToNucleiErrorMap()

◆ AddTransformation()

◆ AddZAI2GlobalDepletionTallyList()

◆ AddZAIForcedCut()

◆ BuildCOBRAFiles()

◆ BuildMCFile()

◆ CalculateBetaI()

◆ CheckExtensions()

◆ CheckMandatoryPointeur()

◆ CheckNucleusDataBaseError()

◆ ClearSurfaceCardVector()

◆ ConstructFileName() [1/2]

◆ ConstructFileName() [2/2]

◆ Copy()

◆ Delete()

◆ DontAddMaterial()

◆ Dump()

◆ Evolution()

◆ FindCell() [1/2]

◆ FindCell() [2/2]

◆ FindCellOfUnknownVolume()

◆ FindFissileEnrichment()

◆ FindMaterial() [1/2]

◆ FindMaterial() [2/2]

◆ FindMCRunDirectory()

◆ FindMissingVolume()

◆ FindPinCell() [1/2]

◆ FindPinCell() [2/2]

◆ FindTally()

◆ FindTimeEnergyGrid()

◆ FindTransformation()

◆ GetAllPossibleReactionNames()

◆ GetBasePriority()

◆ GetBaseSummaryFileName()

◆ GetBaseSummaryLine()

◆ GetCellOfUnknownVolume()

◆ GetCellVector()

◆ GetComment()

◆ GetConnectorPlugin()

◆ GetDATADIR()

◆ GetDATAPATH()

◆ GetDepletionTallyListINT()

◆ GetDepletionTallyListZAI()

◆ GetDisableViolentSimplifyNode()

◆ GetEnergyBins()

◆ GetEnsdfDATADIR()

◆ GetEnvVar()

◆ GetESpectrumTallyNumber()

◆ GetEvolutionSolver()

◆ GetExtendedReactionList()

◆ GetFirstContinuousZIndexInBaseSummary()

◆ GetFirstThermalZIndexInBaseSummary()

◆ GetFissionEnergies()

◆ GetFissionReleasedFile()

◆ GetFPASCIIFileName()