`openmc.data` – Nuclear Data Interface¶

Core Classes¶

`openmc.data.IncidentNeutron`	Continuous-energy neutron interaction data.
`openmc.data.Reaction`	A nuclear reaction
`openmc.data.Product`	Secondary particle emitted in a nuclear reaction
`openmc.data.Tabulated1D`	A one-dimensional tabulated function.
`openmc.data.FissionEnergyRelease`	Energy relased by fission reactions.
`openmc.data.ThermalScattering`	A ThermalScattering object contains thermal scattering data as represented by an S(alpha, beta) table.
`openmc.data.CoherentElastic`	Coherent elastic scattering data from a crystalline material
`openmc.data.FissionEnergyRelease`	Energy relased by fission reactions.
`openmc.data.DataLibrary`	Collection of cross section data libraries.
`openmc.data.Decay`	Radioactive decay data.
`openmc.data.FissionProductYields`	Independent and cumulative fission product yields.
`openmc.data.WindowedMultipole`	Resonant cross sections represented in the windowed multipole format.

`openmc.data.atomic_mass`	Return atomic mass of isotope in atomic mass units.
`openmc.data.linearize`	Return a tabulated representation of a function of one variable.
`openmc.data.thin`	Check for (x,y) points that can be removed.
`openmc.data.write_compact_458_library`	Read ENDF files, strip the MF=1 MT=458 data and write to small HDF5.

`openmc.data.AngleEnergy`	Distribution in angle and energy of a secondary particle.
`openmc.data.KalbachMann`	Kalbach-Mann distribution
`openmc.data.CorrelatedAngleEnergy`	Correlated angle-energy distribution
`openmc.data.UncorrelatedAngleEnergy`	Uncorrelated angle-energy distribution
`openmc.data.NBodyPhaseSpace`	N-body phase space distribution
`openmc.data.LaboratoryAngleEnergy`	Laboratory angle-energy distribution
`openmc.data.AngleDistribution`	Angle distribution as a function of incoming energy
`openmc.data.EnergyDistribution`	Abstract superclass for all energy distributions.
`openmc.data.ArbitraryTabulated`	Arbitrary tabulated function given in ENDF MF=5, LF=1 represented as
`openmc.data.GeneralEvaporation`	General evaporation spectrum given in ENDF MF=5, LF=5 represented as
`openmc.data.MaxwellEnergy`	Simple Maxwellian fission spectrum represented as
`openmc.data.Evaporation`	Evaporation spectrum represented as
`openmc.data.WattEnergy`	Energy-dependent Watt spectrum represented as
`openmc.data.MadlandNix`	Energy-dependent fission neutron spectrum (Madland and Nix) given in
`openmc.data.DiscretePhoton`	Discrete photon energy distribution
`openmc.data.LevelInelastic`	Level inelastic scattering
`openmc.data.ContinuousTabular`	Continuous tabular distribution

`openmc.data.Resonances`	Resolved and unresolved resonance data
`openmc.data.ResonanceRange`	Resolved resonance range
`openmc.data.SingleLevelBreitWigner`	Single-level Breit-Wigner resolved resonance formalism data.
`openmc.data.MultiLevelBreitWigner`	Multi-level Breit-Wigner resolved resonance formalism data.
`openmc.data.ReichMoore`	Reich-Moore resolved resonance formalism data.
`openmc.data.RMatrixLimited`	R-matrix limited resolved resonance formalism data.
`openmc.data.ParticlePair`
`openmc.data.SpinGroup`	Resonance spin group
`openmc.data.Unresolved`	Unresolved resonance parameters as identified by LRU=2 in MF=2.

`openmc.data.ace.Library`	A Library objects represents an ACE-formatted file which may contain multiple tables with data.
`openmc.data.ace.Table`	ACE cross section table

openmc.data.ace.ascii_to_binary Convert an ACE file in ASCII format (type 1) to binary format (type 2).

openmc.data.endf.Evaluation ENDF material evaluation with multiple files/sections

`openmc.data.endf.float_endf`	Convert string of floating point number in ENDF to float.
`openmc.data.endf.get_cont_record`	Return data from a CONT record in an ENDF-6 file.
`openmc.data.endf.get_evaluations`	Return a list of all evaluations within an ENDF file.
`openmc.data.endf.get_head_record`	Return data from a HEAD record in an ENDF-6 file.
`openmc.data.endf.get_tab1_record`	Return data from a TAB1 record in an ENDF-6 file.
`openmc.data.endf.get_tab2_record`
`openmc.data.endf.get_text_record`	Return data from a TEXT record in an ENDF-6 file.

`openmc.data.njoy.run`	Run NJOY with given commands
`openmc.data.njoy.make_pendf`	Generate ACE file from an ENDF file
`openmc.data.njoy.make_ace`	Generate incident neutron ACE file from an ENDF file
`openmc.data.njoy.make_ace_thermal`	Generate thermal scattering ACE file from ENDF files