Editing Input syntax manual (section)

=== mat (material definition)<span id="mat"></span> ===

See Chapter 4 of <ref name="manual" />.

 '''mat''' ''NAME DENS'' [ [[#mat_tmp|'''tmp''']] ''TEMP'' ]
               [ [[#mat_tms|'''tms''']] ''TEMP'' ]
               [ [[#mat_tft|'''tft''']] ''T<sub>MIN</sub>'' ''T<sub>MAX</sub>'' ]
               [ [[#mat_rgb|'''rgb''']] ''R G B'' ]
               [ [[#mat_vol|'''vol''']] ''VOL'' ]
               [ [[#mat_mass|'''mass''']] ''MASS'' ]
               [ [[#mat_burn|'''burn''']] ''N<sub>R</sub>'' ]
               [ [[#mat_fix|'''fix''']] ''ID'' ''TEMP'' ] 
               [ [[#mat_moder|'''moder''']] ''THNAME'' ''ZA'' ]
   ''NUC<sub>1</sub> FRAC<sub>1</sub>'' 
 [ ''NUC<sub>2</sub> FRAC<sub>2</sub>'' ]
 [    ''...''     ]

Material definition. The mandatory parameters are:

{|
| <tt>''NAME''</tt>
| : name of the material
|-
| <tt>''DENS''</tt>
| : density of the material (positive value = atomic density [in b<sup>-1</sup>cm<sup>-1</sup>], negative value = mass density [in g/cm<sup>3</sup>])
|-
| <tt>''NUC<sub>n</sub>''</tt>
| : Identifier of ''n''-th nuclide in composition
|-
| <tt>''FRAC<sub>n</sub>''</tt>
| : fraction of ''n''-th nuclide in composition (positive value = atomic fractions/density, negative values = mass fractions/density)
|-
|}

The remaining parameters are defined by separate key words followed by the input values, being optional.

<u>Notes:</u>
*There is a special entry for the <tt>''DENS''</tt> parameter:
** "<tt>sum</tt>": to calculate the density from given nuclide fractions
*The nuclide identifier for nuclides with associated cross-sections corresponds to ZZAAA.ID and, for nuclides without associated cross-sections, e.g., decay nuclides, to ZZAAAI. 
**The identifiers include ''Z'', the atomic number; ''A'', the mass number of the nuclide; ''I'', the isomeric state (0 = ground state, 1 = metastable state); and ''ID'', the library identifier.
*For more information, see the detailed description on [[Definitions, units and constants#Definitions|Definitions]].


<u>Optional entries:</u>

Material temperature for Doppler-broadening pre-processor (<tt>'''tmp'''</tt>): <span id="mat_tmp"></span>

{|
| <tt>''TEMP''</tt>
| : temperature of the material [in K]
|}

<u>Notes:</u>
*It defines the material temperature for [[Doppler-broadening preprocessor routine|Doppler-preprocessor]]. 


Material temperature for on-the-fly temperature treatment (<tt>'''tms'''</tt>): <span id="mat_tms"></span>

{|
| <tt>''TEMP''</tt>
| : temperature of the material [in K]
|}

<u>Notes:</u>
*It defines the material temperature for on-the-fly [[TMS on-the-fly temperature treatment routine‏‎|TMS temperature treatment]].


Material temperature for coupled multi-physics calculations (<tt>'''tft'''</tt>): <span id="mat_tft"></span>

{|
| <tt>''T<sub>MIN</sub>''</tt>
| : lower limit for material temperature [in K]
|-
| <tt>''T<sub>MAX</sub>''</tt>
| : upper limit for material temperature [in K]
|}

<u>Notes:</u>
*It sets the temperature limits for material for [[Coupled multi-physics calculations|coupled multi-physics calculations]].
*It is used to define the minimum and maximum temperature for the TMS-treatment directly from the interface files (see [[#ifc|ifc]] card).
*For more information, see the detailed description on the [[multi-physics interface| Multi-physics interface]].


Material RGB-color (<tt>'''rgb'''</tt>): <span id="mat_rgb"></span>

{|
| <tt>''R''</tt>
| : value for the red channel (between 0 and 255)
|-
| <tt>''G''</tt>
| : value for the green channel (between 0 and 255)
|-
| <tt>''B''</tt>
| : value for the blue channel (between 0 and 255)
|}

<u>Notes:</u>
*It assigns a dedicated RGB-color to the material for the material representation in [[#plot|geometry plots]].
*If the entry is not provided, the material color is sampled randomly.


Material volume (<tt>'''vol'''</tt>): <span id="mat_vol"></span>

{|
| <tt>''VOL''</tt>
| : volume of the material [in cm<sup>3</sup>] (3D geometry) or cross-sectional area [in cm<sup>2</sup>] (2D geometry)
|}

<u>Notes:</u>
*It defines the material volume.
*Alternatives ways to provide the material volume includes:
** [[#set mvol|set mvol]] option, used to define the material volumes manually
** [[#set mcvol|set mcvol]] option, used to define the material volumes automatically using the Monte Carlo checker routine at runtime.
** [[Installing and running Serpent#Monte Carlo volume calculation routine|<tt>''-checkvolumes''</tt>]] command line option, used to evaluate the material volumes in an independent run.
*For more information, see the detailed description on [[defining material volumes|material volumes definition]].


Material mass (<tt>'''mass'''</tt>): <span id="mat_mass"></span>

{|
| <tt>''MASS''</tt>
| : mass of the material [in g]
|}

<u>Notes:</u>
*The material mass can be provided as an alternative to the material volume.


Material depletion flag (<tt>'''burn'''</tt>): <span id="mat_burn"></span>

{|
| <tt>''N<sub>R</sub>''</tt>
| : option to flag the material as burnable (1/yes) or non-burnable (0/no). The default option is "<tt>non-burnable</tt>"
|-
|}

<u>Notes:</u>
*In order to deplete the material and include it in the burnup calculation, <u>the flag must be set to "<tt>1</tt>"</u>
*The depletion zone division should be done using the [[#div|div]] card. However, 
** if a material is defined within a pin-structure, Serpent, by default if no [[#div|div]] card is associated to the material, sub-divides the material in a pin-type level.
** in Serpent 1, the "flag" is interpreted as the number of annular regions (<u>not recommended</u>)


Material library information for nuclides without cross section data and their decay products (<tt>'''fix'''</tt>): <span id="mat_fix"></span>

{|
| <tt>''LIB''</tt>
| : library ID (e.g. "09c") for nuclides without cross section data.
|-
| <tt>''TEMP''</tt>
| : temperature for nuclides without cross section data [in K]
|}

<u>Notes:</u>
*It defines the library properties: identifier and temperature for the nuclides without cross section data, e.g. decay nuclides, within the material composition.
*Decay products from these nuclides may have cross section data and will inherit the library ID and temperature based on this card.


Material associated thermal-scattering data (<tt>'''moder'''</tt>): <span id="mat_moder"></span>

{|
| <tt>''THNAME''</tt>
| : name of the [[#therm|thermal scattering data library]]
|-
| <tt>''ZA''</tt>
| : nuclide ZA of the thermal scatterer (e.g. 1001 for H-1).
|}

<u>Notes:</u>
*It links the thermal-scattering data library for a given nuclide within the material composition.
*The thermal-scattering data library and the associated temperature treatment is defined by the [[#therm|therm]] card.
*A single material can include multiple "<tt>moder</tt>" entries to define thermal-scattering libraries form multiple nuclides, such as H-H20 and D-D20 in semi-heavy water.