ACE Cross Sections – pyne.ace
¶
This module is for reading ACEformat cross sections. ACE stands for “A Compact ENDF” format and originated from work on MCNP. It is used in a number of other Monte Carlo particle transport codes.
ACEformat cross sections are typically generated from ENDF files through a cross section processing program like NJOY. The ENDF data consists of tabulated thermal data, ENDF/B resonance parameters, distribution parameters in the unresolved resonance region, and tabulated data in the fast region. After the ENDF data has been reconstructed and Dopplerbroadened, the ACER module generates ACEformat cross sections.
All functionality may be found in the ace
package:
from pyne import ace
The information below represents the complete specification of the classes in the ace module. For examples of usage, please refer to the User’s Guide entry for ACE Cross Sections.
Library Class¶

class
pyne.ace.
Library
¶ A Library objects represents an ACEformatted file which may contain multiple tables with data.
Parameters: filename : str
Path of the ACE library file to load.
Attributes
binary (bool) Identifies Whether the library is in binary format or not tables (dict) Dictionary whose keys are the names of the ACE tables and whose values are the instances of subclasses of AceTable (e.g. NeutronTable) verbose (bool) Determines whether output is printed to the stdout when reading a Library 
find_table
(name)¶ Returns a crosssection table with a given name.
Parameters: name : str
Name of the crosssection table, e.g. 92235.70c

read
(table_names=None)¶ Read through and parse the ACEformat library.
Parameters: table_names : None, str, or iterable, optional
Tables from the file to read in. If None, reads in all of the tables. If str, reads in only the single table of a matching name.


class
pyne.ace.
AceTable
¶ Abstract superclass of all other classes for cross section tables.

class
pyne.ace.
NeutronTable
¶ A NeutronTable object contains continuousenergy neutron interaction data read from an ACEformatted Type I table. These objects are not normally instantiated by the user but rather created when reading data using a Library object and stored within the
tables
attribute of a Library object.Parameters: name : str
ZAID identifier of the table, e.g. ‘92235.70c’.
awr : float
Atomic mass ratio of the target nuclide.
temp : float
Temperature of the target nuclide in eV.
Attributes
awr (float) Atomic mass ratio of the target nuclide. energy (list of floats) The energy values (MeV) at which reaction crosssections are tabulated. name (str) ZAID identifier of the table, e.g. 92235.70c. nu_p_energy (list of floats) Energies in MeV at which the number of prompt neutrons emitted per fission is tabulated. nu_p_type (str) Indicates how number of prompt neutrons emitted per fission is stored. Can be either “polynomial” or “tabular”. nu_p_value (list of floats) The number of prompt neutrons emitted per fission, if data is stored in “tabular” form, or the polynomial coefficients for the “polynomial” form. nu_t_energy (list of floats) Energies in MeV at which the total number of neutrons emitted per fission is tabulated. nu_t_type (str) Indicates how total number of neutrons emitted per fission is stored. Can be either “polynomial” or “tabular”. nu_t_value (list of floats) The total number of neutrons emitted per fission, if data is stored in “tabular” form, or the polynomial coefficients for the “polynomial” form. reactions (list of Reactions) A list of Reaction instances containing the cross sections, secondary angle and energy distributions, and other associated data for each reaction for this nuclide. sigma_a (list of floats) The microscopic absorption cross section for each value on the energy grid. sigma_t (list of floats) The microscopic total cross section for each value on the energy grid. temp (float) Temperature of the target nuclide in eV.

class
pyne.ace.
SabTable
¶ A SabTable object contains thermal scattering data as represented by an S(alpha, beta) table.
Parameters: name : str
ZAID identifier of the table, e.g. lwtr.10t.
awr : float
Atomic mass ratio of the target nuclide.
temp : float
Temperature of the target nuclide in eV.
Attributes
awr (float) Atomic mass ratio of the target nuclide. elastic_e_in (list of floats) Incoming energies in MeV for which the elastic cross section is tabulated. elastic_P (list of floats) Elastic scattering cross section for data derived in the incoherent approximation, or Bragg edge parameters for data derived in the coherent approximation. elastic_type (str) Describes the behavior of the elastic cross section, i.e. whether it was derived in the incoherent or coherent approximation. inelastic_e_in (list of floats) Incoming energies in MeV for which the inelastic cross section is tabulated. inelastic_sigma (list of floats) Inelastic scattering cross section in barns at each energy. name (str) ZAID identifier of the table, e.g. 92235.70c. temp (float) Temperature of the target nuclide in eV.

class
pyne.ace.
Reaction
(MT, table=None)¶ A Reaction object represents a single reaction channel for a nuclide with an associated cross section and, if present, a secondary angle and energy distribution. These objects are stored within the
reactions
attribute on subclasses of AceTable, e.g. NeutronTable.Parameters: MT : int
The ENDF MT number for this reaction. On occasion, MCNP uses MT numbers that don’t correspond exactly to the ENDF specification.
table : AceTable
The ACE table which contains this reaction. This is useful if data on the parent nuclide is needed (for instance, the energy grid at which cross sections are tabulated)
Attributes
ang_energy_in (list of floats) Incoming energies in MeV at which angular distributions are tabulated. ang_energy_cos (list of floats) Scattering cosines corresponding to each point of the angular distribution functions. ang_energy_pdf (list of floats) Probability distribution function for angular distribution. ang_energy_cdf (list of floats) Cumulative distribution function for angular distribution. e_dist_energy (list of floats) Incoming energies in MeV at which energy distributions are tabulated. e_dist_law (int) ACE law used for secondary energy distribution. IE (int) The index on the energy grid corresponding to the threshold of this reaction. MT (int) The ENDF MT number for this reaction. On occasion, MCNP uses MT numbers that don’t correspond exactly to the ENDF specification. Q (float) The Qvalue of this reaction in MeV. sigma (list of floats) Microscopic cross section for this reaction at each point on the energy grid above the threshold value. TY (int) An integer whose absolute value is the number of neutrons emitted in this reaction. If negative, it indicates that scattering should be performed in the centerofmass system. If positive, scattering should be preformed in the laboratory system. 
threshold
()¶ Return energy threshold for this reaction.
