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10.4} The FLUKA neutron cross section library


 As explained in 3}, an unformatted cross section data set must be available
 for low-energy neutron transport. For a description of the algorithms used
 for tracking low-energy neutrons, see the beginning of this Chapter. Other
 useful information can be found in the Notes to options LOW--NEUT, LOW--MAT
 and LOW--BIAS}.
 The Legendre expansion used in Fluka is P5, i.e. at each collision
 the polar scattering angle is sampled from three discrete values, such that
 the first 6 moments of the angular distribution are preserved (the
 azimuthal angle is sampled instead from a uniform distribution between
 0 and 2*pi). The energy group structure depends on the cross section
 set used. Here below the group structure of the currently available sets
 is reported, and a list of the materials they contain.

 The default FLUKA neutron cross section library (originally prepared by
 G. Panini of ENEA [Cuc91]) contains more than 250 different materials (natural
 elements or single nuclides), selected for their interest in physics, dosimetry
 and accelerator engineering. This library has a larger number of groups and a
 better resolution in the thermal energy range in respect to the original one.
* Start_Devel_seq
 Both sets are being continuously enriched and updated on the basis of the most
 recent evaluations (ENDF/B, JEF, JENDL, etc.).
* End_Devel_seq
 The preparation of the library involves the use of a specialised code [NJOY] and
 several ad-hoc programs written to adjust the output to the particular structure
 of these libraries. The library is continuously enriched and updated on the
 basis of the most recent evaluations (ENDF/B, JEF, JENDL etc.).
 The library format is similar to that known as ANISN (or FIDO) format, but it
 has been modified to include kerma factor data, residual nuclei and partial
 exclusive cross sections when available. The latter are not used directly by
 FLUKA, but can be folded over calculated spectra to get reaction rates and
 induced activities.
 More materials can be made available on request, if good evaluations are
 available. Some cross sections are available in the library at two or three
 different temperatures, mainly in view of simulations of calorimeters containing
 cryogenic scintillators. Doppler broadening is taken into account.

* Start_Devel_seq
 In some cases, data from different evaluated files are available for the
 same material. For hydrogen, cross sections are provided for both gas
 hydrogen and hydrogen with different molecular bonds. The maximum energy is
 0.0196 GeV for the ENEA set, and 0.020 GeV for the 260 group set.
* End_Devel_seq

 Note that the energy groups are numbered in order of DECREASING energy
 (group 1 corresponds to the highest energy).

* Start_Devel_seq
 The ENEA cross section set has 72 neutron energy groups and 22 gamma groups.
* End_Devel_seq
 The default FLUKA neutron cross section library has 260 neutron groups and 42
 gamma groups. Gamma energy groups are used only for (n,gamma) production, since
 transport of photons in FLUKA is continuous in energy and angle and is performed
 through the EMF module).

* Start_Devel_seq
 For temperatures different from those for which a material has been
 prepared, an approximate approach is possible, provided the cross sections
 of that material follow closely a 1/v dependence (see option MAT-PROP). In this
 approximation, the thermal group velocities, absorption probabilities,
 gamma generation probabilities, fission probabilities and kermas are
 rescaled according to 1/v. No modification is made to the elastic
 cross section and hence to the downscattering matrix: this can be a
 very bad approximation, mostly for light materials. No modification is
 applied for possible Doppler broadening effects on resonances for
 thermal and epithermal neutrons: again this can be a bad approximation.
 The total cross section is rescaled according to the modified absorption and
 fission ones.

 The averaging inside each energy group of the ENEA library has been done
 according to the weighting function used by the VITAMIN-J cross section
 library [Sar90], using - from low to high energies - a Maxwellian at the
 relevant temperature, a 1/E spectrum in the intermediate energy range, a
 fission spectrum and again a 1/E spectrum.
 No special averaging has been done for the 260-group library, since the
 group width is much narrower.
* End_Devel_seq

 Hydrogen cross sections, which have a particular importance in neutron
 slowing-down, are available also for different types of molecular binding
 (free, H2O, CH2).

 At present, the FLUKA libraries contain only single isotopes or elements
 of natural isotopic composition, although the possibility exists to
 include in future also pre-mixed materials.

 Neutron energy deposition in most materials is calculated by means of
 kerma factors (including contributions from low-energy fission). However,
 recoil protons and protons from N(n,p) reaction are produced and transported

 Each material is identified by an alphanumeric name (a string
 not longer than 8 characters, all in upper case), and by three integer
 identifiers. Correspondence with FLUKA materials (pre-defined or
 user-defined) is based on any combination of name and zero or more
 identifiers. In case of ambiguity, the first material in the list
 fulfilling the combination is selected. (See command LOW-MAT for more details).

 The convention generally used (but there may be exceptions) for the 3
 identifiers is:
* Atomic number
* Mass number, or natural isotopic composition if negative
(exceptions are possible in order to distinguish between data from different sources referring to the same nuclide)
* Neutron temperature in degrees Kelvin
10.4.1} 260 neutron, 42 gamma group library -------------------------------------------

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