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2.3.7} Materials

 Each geometry region is supposed to be filled with a homogeneous material, or
 with vacuum, or with "blackhole". The latter is a fictitious material used to
 terminate particle trajectories: any particle is discarded when reaching a
 blackhole boundary. Materials can be simple elements or compounds, where an
 element can have either natural composition or consist of a single nuclide,
 and compound indicates a chemical compound or a mixture or an alloy (or an
 isotopic mixture) of known composition.

 An element can be either predefined (see list in 5}) or defined by a MATERIAL
 card giving its atomic number, atomic weight, density, name and a material
 identification number > 2. The number of a non-pre-defined material can be
 chosen by the user, with the restriction that all lower numbers must also be
 defined (but not necessarily used). However, in a name-based input, it is
 convenient to leave the number blank: in this case the user does not need to
 know the number, which is assigned automatically by the code and will be used
 only internally.

 Number 1 is reserved for blackhole and 2 for ideal vacuum. There are 25
 pre-defined materials; but each of the numbers from 3 to 25 can be redefined
 freely, overriding the default definition. However, if the input is
 explicitly number-based only (via command GLOBAL), a pre-defined material can
 only be redefined using the same name by assigning to it a number equal to
 the original one. If the input is name-based, it is better to leave the
 number blank.

 Materials can also be defined with higher numbers, provided no gaps are left
 in the numbering sequence. For instance a material cannot be defined to have
 number 28 unless also 26 and 27 have been defined. Again, assigning
 explicitly a number is not necessary if the input is fully name-based.

 A compound is defined by a MATERIAL card plus as many COMPOUND cards as
 needed to describe its composition. The MATERIAL card used to define a
 compound carries only the compound name, density and, if input is explicitly
 number-based only, material number (atomic number and atomic weight having no
 meaning in this case). Some special pre-defined compounds are available: for
 them the stopping power is not calculated by a formula but is determined
 using the parameters recommended by ICRU [ICRU84]. Reference to these
 pre-defined compounds is normally by name and no MATERIAL and COMPOUND cards
 are needed: if the input is explicitly number-based only (via command
 GLOBAL), a number using a MATERIAL card needs to be assigned to them, of
 course leaving no gaps in the numbering sequence.

 Materials predefined or defined in the standard input are referred to as
 "FLUKA materials", to distinguish them from materials available in the
 low-energy neutron cross section library (called "low-energy cross section

 When transport of low-energy neutrons (E < 20 MeV) is requested (explicitly
 or by the chosen defaults), a correspondence is needed between each elemental
 (i.e. not compound) "FLUKA material" and one of the "low-energy cross section
 materials" available in the FLUKA low-energy neutron library. The default
 correspondence is set by the name and if more than one material with that
 name exist in the neutron library, the first in the list with that name (see
 Chap. 10}) is assumed by default. The default can be changed using option

 In the case of our example, only Beryllium is necessary, apart from blackhole
 and vacuum. In principle, since Beryllium is one of the pre-defined FLUKA
 materials, this part could even be omitted. However, for pedagogical reasons
 the following card is proposed, where index 5 is assigned to the target
MATERIAL 4.0 0.0 1.848 5.0 BERYLLIU Notice that the chosen name is BERYLLIU and not BERYLLIUM, in order to match the name in the list of "low-energy cross section materials" for low energy neutrons (see Chap. 10}), where material names have a maximum length of 8 characters. The standard output concerning materials is very extended. First a list of multiple scattering parameters (printed by subroutine MULMIX) is reported for each material requested. This is mostly of scarce interest for the normal user, except for a table giving for each requested material the proportion of components, both by number of atoms (normalised to 1) and by weight (normalised to material density). The same information is repeated later on in another table entitled "Material compositions". If low-energy neutron transport has been requested (explicitly or by a chosen default), the following section reports the relevant material information: in particular, a table entitled "Fluka to low en. xsec material correspondence" specifies which material in the neutron cross section library has been mapped to each input material. Note that a much more detailed cross section information can be obtained by setting a printing flag (
) in the LOW-NEUT command. The Table "Material compositions" contains information about the requested materials, those pre-defined by default and the elements used to define compounds. In addition to effective atomic number and weight, density and composition, the table shows the value of some typical quantities: inelastic and elastic scattering length for the beam particles (not valid for electron and photon beams), radiation length and inelastic scattering length for 20 MeV neutrons. The next table contains material data related to stopping power (average excitation energy, density effect parameters, and pressure in the case of gases) plus information about the implementation of various physical effects and the corresponding thresholds and tabulations. The last material table is printed just before the beginning of the history calculations, and concerns the "Correspondence of regions and EMF-FLUKA material numbers and names".

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