Difference between revisions of "Input syntax manual"

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=== branch (branch definition)<span id="branch"></span> ===
 
=== branch (branch definition)<span id="branch"></span> ===
  
  '''branch''' ''NAME'' [ '''repm''' ''MAT<sub>1</sub>'' ''MAT<sub>2</sub>'' ]
+
  '''branch''' ''NAME'' [ [[#branch_repm|'''repm''']] ''MAT<sub>1</sub>'' ''MAT<sub>2</sub>'' ]
             [ '''repu''' ''UNI<sub>1</sub>'' ''UNI<sub>2</sub>'' ]
+
             [ [[#branch_repu|'''repu''']] ''UNI<sub>1</sub>'' ''UNI<sub>2</sub>'' ]
             [ '''stp''' ''MAT DENS TEMP THERM<sub>1</sub> SABL<sub>1</sub> SABH<sub>1</sub> THERM<sub>2</sub> SABL<sub>2</sub> SABH<sub>2</sub> ...'' ]
+
             [ [[#branch_stp|'''stp''']] ''MAT DENS TEMP THERM<sub>1</sub> SABL<sub>1</sub> SABH<sub>1</sub> THERM<sub>2</sub> SABL<sub>2</sub> SABH<sub>2</sub> ...'' ]
             [ '''tra''' ''TGT TRANS'' ]  
+
             [ [[#branch_tra|'''tra''']] ''TGT TRANS'' ]  
             [ '''xenon''' ''OPT'' ]  
+
             [ [[#bracnh_xenon|'''xenon''']] ''OPT'' ]  
             [ '''samarium''' ''OPT'' ]  
+
             [ [[#branch_samarium|'''samarium''']] ''OPT'' ]  
             [ '''norm''' ''NSF'' ]  
+
             [ [[#branch_norm|'''norm''']] ''NSF'' ]  
             [ '''gcu''' ''UNI<sub>2</sub>'' ]  
+
             [ [[#branch_gcu|'''gcu''']] ''UNI<sub>2</sub>'' ]  
             [ '''reptrc''' ''FILE<sub>1</sub>'' ''FILE<sub>2</sub>'' ]
+
             [ [[#branch_reptrc|'''reptrc''']] ''FILE<sub>1</sub>'' ''FILE<sub>2</sub>'' ]
             [ '''var''' ''VNAME VAL'' ]
+
             [ [[#branch_var|'''var''']] ''VNAME VAL'' ]
Defines the variations invoked for a branch in the automated burnup sequence. Input values:
+
            [ [[#branch_incl|'''incl''']] ''MODFILE'' ]
 +
Defines the variations invoked for a branch in the automated burnup sequence. The first parameter:
  
 
{|
 
{|
 
| <tt>''NAME''</tt>
 
| <tt>''NAME''</tt>
 
| : branch name
 
| : branch name
|-
+
|}
 +
 
 +
The remaining parameters are defined by separate key words followed by the input values.
 +
 
 +
<u>Notes:</u>
 +
 
 +
*The branch card can be combined with the [[#coef (coefficient matrix definition)|coef card]], [[#hisv (history variation matrix definition)|hisv card]], and [[#casematrix (casematrix definition)| casematrix card]].
 +
*The branch name identifies the branch <tt>''BR<sub>m,i</sub>''</tt> in the variation matrix defined by the [[#coef (coefficient matrix definition)|coef card]], [[#hisv (history variation matrix definition)|hisv card]], and [[#casematrix (casematrix definition)| casematrix card]].
 +
*The input parameters consist of a number variations, which are invoked when the branch is applied.
 +
*A single branch card may include one or several variations.
 +
*For more information, see detailed description on the [[automated burnup sequence]].
 +
 
 +
 
 +
<u>Variation types:</u>
 +
 
 +
Branch material variation (<tt>'''repm'''</tt>):<span id="branch_repm"></span>
 +
 
 +
{|
 
| <tt>''MAT<sub>1</sub>''</tt>  
 
| <tt>''MAT<sub>1</sub>''</tt>  
 
| : name of the replaced material
 
| : name of the replaced material
Line 39: Line 57:
 
| <tt>''MAT<sub>2</sub>''</tt>  
 
| <tt>''MAT<sub>2</sub>''</tt>  
 
| : name of the replacing material
 
| : name of the replacing material
|-
+
|}
 +
 
 +
<u>Notes:</u>
 +
*The material variation can be used to replace one material with another, for example, to change coolant boron concentration.
 +
*The material replacement works as if <tt>''MAT<sub>1</sub>''</tt> were created using the [[#mat_.28material_definition.29|mat]] or [[#mix_.28mixture_definition.29|mix]] card of <tt>''MAT<sub>2</sub>''</tt>.
 +
*The name of the material present in the geometry will still be <tt>''MAT<sub>1</sub>''</tt> after the replacement, but the material specification (composition, density, tmp, moder, rgb, etc.) will correspond to <tt>''MAT<sub>2</sub>''</tt>.
 +
**This means that all other input-cards that are linked to a specific material name such as [[#det_dm|det dm]], [[#src_sm|src sm]], [[#set_trc|set trc]] and [[#set_iter_nuc|set iter nuc]] can be linked to the original material (<tt>''MAT<sub>1</sub>''</tt>) and they will automatically apply to whatever material <tt>''MAT<sub>2</sub>''</tt> replaces <tt>''MAT<sub>1</sub>''</tt> for the branch calculation.
 +
*The replaced material ''MAT<sub>1</sub>'' is also replaced inside mixtures.
 +
**This means one can not replace a material with a mixture defined with [[#mix (mixture_definition)|mix card]] containing the replaced material (for example replacing pure water defined with [[#mat (material definition)|mat card]] by a mixture of boron and water defined with a [[#mix (mixture definition)|mix card]] containing the same pure water material).
 +
*The replacing material ''MAT<sub>2</sub>'' can not be included in the geometry using other cards than the branch card, from version 2.1.30 and on.
 +
 
 +
 
 +
Branch universe variation (<tt>'''repu'''</tt>):<span id="branch_repu"></span>
 +
 
 +
{|
 
| <tt>''UNI<sub>1</sub>''</tt>  
 
| <tt>''UNI<sub>1</sub>''</tt>  
 
| : name of the replaced universe
 
| : name of the replaced universe
Line 45: Line 77:
 
| <tt>''UNI<sub>2</sub>''</tt>  
 
| <tt>''UNI<sub>2</sub>''</tt>  
 
| : name of the replacing universe
 
| : name of the replacing universe
|-
+
|}
 +
 
 +
<u>Notes:</u>
 +
*The universe variation can be used to replace one universe with another, for example, to replace empty control rod guide tubes with rodded tubes for control rod insertion in 2D geometries.
 +
*The name of the universe present in the geometry will still be <tt>''UNI<sub>1</sub>''</tt> after the replacement, but the universe contents will correspond to <tt>''UNI<sub>2</sub>''</tt>.
 +
*This means that all other input-cards that are linked to a specific universe name such as [[#det_du|det du]] and [[#src_su|src su]] can be linked to the original universe (<tt>''UNI<sub>1</sub>''</tt>) and they will automatically apply to whatever universe <tt>''UNI<sub>2</sub>''</tt> replaces <tt>''UNI<sub>1</sub>''</tt> for the branch calculation.
 +
 
 +
 
 +
Branch state variation, density/temperature (<tt>'''stp'''</tt>):<span id="branch_stp"></span>
 +
{|
 
| <tt>''MAT''</tt>
 
| <tt>''MAT''</tt>
 
| : name of the material for which density and temperature are adjusted  
 
| : name of the material for which density and temperature are adjusted  
 
|-
 
|-
 
| <tt>''DENS''</tt>  
 
| <tt>''DENS''</tt>  
| : material density after adjustment (positive entries for atomic, negative entries for mass densities)
+
| : material density after adjustment (positive value = atomic density [in b<sup>-1</sup>cm<sup>-1</sup>], negative value = mass density [in g/cm<sup>3</sup>])
 
|-
 
|-
 
| <tt>''TEMP''</tt>  
 
| <tt>''TEMP''</tt>  
| : material temperature after adjustment, or -1 if no adjustment in temperature
+
| : material temperature after adjustment [in K], or "<tt>-1</tt>" if no adjustment in temperature
 
|-
 
|-
 
| <tt>''THERM<sub>n</sub>''</tt>
 
| <tt>''THERM<sub>n</sub>''</tt>
| : ''n''th thermal scattering data associated with the material
+
| : ''n''-th thermal scattering data associated with the material
 
|-
 
|-
 
| <tt>''SABL<sub>n</sub>''</tt>
 
| <tt>''SABL<sub>n</sub>''</tt>
| : name of the ''n''th S(&alpha;, &beta;) library for temperature below the given value
+
| : name of the ''n''-th S(&alpha;, &beta;) library for temperature below the given value
 
|-
 
|-
 
| <tt>''SABH<sub>n</sub>''</tt>
 
| <tt>''SABH<sub>n</sub>''</tt>
| : name of the ''n''th S(&alpha;, &beta;) library for temperature above the given value
+
| : name of the ''n''-th S(&alpha;, &beta;) library for temperature above the given value
|-
+
|}
 +
 
 +
<u>Notes:</u>
 +
*The state variation can be used to change material density and temperature.
 +
*There are two special entries for the <tt>''DENS''</tt> entry:
 +
** "<tt>sum</tt>": to define the material density as the sum of the constituent nuclides densities (not supported from version 2.2.0 and on)
 +
** "<tt>original</tt>": to keep unmodified the material density (introduced in version 2.2.1).
 +
*The adjustment is made using the built-in [[Doppler-broadening preprocessor routine]] and tabular interpolation for S(&alpha;, &beta;) thermal scattering data.
 +
*The last three parameters of the card are provided only if the material has thermal scattering libraries attached to it (see the [[#therm (thermal scattering library definition)|therm card]]).
 +
 
 +
 
 +
Branch transformation variation (<tt>'''tra'''</tt>):<span id="branch_tra"></span>
 +
 
 +
{|
 
| <tt>''TGT''</tt>  
 
| <tt>''TGT''</tt>  
 
| : target universe, surface or cell
 
| : target universe, surface or cell
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| <tt>''TRANS''</tt>  
 
| <tt>''TRANS''</tt>  
 
| : name of the applied transformation
 
| : name of the applied transformation
|-
+
|}
 +
 
 +
<u>Notes:</u>
 +
*The transformation variation can be used to move or rotate different parts of the geometry, for example, to adjust the position of control rods in 3D geometries.
 +
*The name of the transformation <tt>''TRANS''</tt> refers to the unit (universe, cell or surface) entry in the [[#trans (transformations)|trans card]].
 +
 
 +
 
 +
Branch xenon variation (<tt>'''xenon'''</tt>):<span id="branch_xenon"></span>
 +
 
 +
{|
 
| <tt>''OPT''</tt>  
 
| <tt>''OPT''</tt>  
| : option for setting poison concentrations (0 = set to zero, 1 = use values from restart file)
+
| : option for setting Xe-poison concentrations (0 = set to zero, 1 = use values from restart file)
|-
+
|}
 +
 
 +
<u>Notes:</u>
 +
*The xenon variation can be set to enforced the Xe-135 concentration to zero. By default the concentration is read from the restart file.
 +
 
 +
 
 +
Branch samarium variation (<tt>'''samarium'''</tt>):<span id="branch_samarium"></span>
 +
 
 +
{|
 +
| <tt>''OPT''</tt>
 +
| : option for setting Sm-poison concentrations (0 = set to zero, 1 = use values from restart file)
 +
|}
 +
 
 +
<u>Notes:</u>
 +
*The samarium variation can be set to enforced the Sm-149 concentration to zero. By default the concentration is read from the restart file.
 +
 
 +
 
 +
Branch normalization variation (<tt>'''nsf'''</tt>):<span id="branch_nsf"></span>
 +
 
 +
{|
 
| <tt>''NSF''</tt>
 
| <tt>''NSF''</tt>
 
| : normalization scaling factor
 
| : normalization scaling factor
|-
+
|}
 +
 
 +
<u>Notes:</u>
 +
*The normalization variation can be used to change the normalization.
 +
*The adjustment is made applying the parameter ''NSF'' as a multiplicative scaling factor to the given normalization.
 +
 
 +
 
 +
Branch group constant variation (<tt>'''gcu'''</tt>):<span id="branch_gcu"></span>
 +
 
 +
{|
 +
| <tt>''UNI<sub>2</sub>''</tt>
 +
|: name of the replacing universe
 +
|}
 +
 
 +
<u>Notes:</u>
 +
*The group constant variation can be used to replace the universe for group constant generation.
 +
*The variation is limited to a single-valued GCU-list (see [[#set gcu|set gcu]] option).
 +
 
 +
 
 +
Branch transport-correction variation (<tt>'''reptrc'''</tt>):<span id="branch_reptrc"></span>
 +
 
 +
{|
 
| <tt>''FILE<sub>1</sub>''</tt>  
 
| <tt>''FILE<sub>1</sub>''</tt>  
 
| : file path of the replaced transport correction curve data
 
| : file path of the replaced transport correction curve data
Line 81: Line 184:
 
| <tt>''FILE<sub>2</sub>''</tt>  
 
| <tt>''FILE<sub>2</sub>''</tt>  
 
| : file path of the replacing transport correction curve data
 
| : file path of the replacing transport correction curve data
|-
+
|}
 +
 
 +
<u>Notes:</u>
 +
*The transport-correction variation can be used to replace a transport correction file with another (see [[#set trc|set trc]] option).
 +
 
 +
 
 +
Branch variable variation (<tt>'''var'''</tt>):<span id="branch_var"></span>
 +
 
 +
{|
 
| <tt>''VNAME''</tt>  
 
| <tt>''VNAME''</tt>  
 
| : variable name
 
| : variable name
Line 90: Line 201:
  
 
<u>Notes:</u>
 
<u>Notes:</u>
 +
*The variable variation can be used to pass information into the output file, which may be convenient for the post-processing of the data.
  
*The branch name identifies the branch in the coefficient matrix of the [[#coef (coefficient matrix definition)|coef card]]
+
 
*The input parameters consist of a number variations, which are invoked when the branch is applied. A single branch card may inclued one or several variations.
+
Branch user-defined variation (<tt>'''incl'''</tt>):<span id="branch_incl"></span>
*The '''repm''' variation can be used to replace one material with another, for example, to change coolant boron concentration.
+
 
**The material replacement works as if <tt>''MAT<sub>1</sub>''</tt> were created using the [[#mat_.28material_definition.29|mat]] or [[#mix_.28mixture_definition.29|mix]] card of <tt>''MAT<sub>2</sub>''</tt>.
+
{|
**The name of the material present in the geometry will still be <tt>''MAT<sub>1</sub>''</tt> after the replacement, but the material specification (composition, density, tmp, moder, rgb, etc.) will correspond to <tt>''MAT<sub>2</sub>''</tt>.
+
| <tt>''MODFILE''</tt>
**This means that all other input-cards that are linked to a specific material name such as [[#det_dm|det dm]], [[#src_sm|src sm]], [[#set_trc|set trc]] and [[#set_iter_nuc|set iter nuc]] can be linked to the original material (<tt>''MAT<sub>1</sub>''</tt>) and they will automatically apply to whatever material <tt>''MAT<sub>2</sub>''</tt> replaces <tt>''MAT<sub>1</sub>''</tt> for the branch calculation.
+
|: file path to an additional/modified input file
*The '''repu''' variation can be used to replace one universe with another, for example, to replace empty control rod guide tubes with rodded tubes for control rod insertion in 2D geometries.
+
|}
**The name of the universe present in the geometry will still be <tt>''UNI<sub>1</sub>''</tt> after the replacement, but the universe contents will correspond to <tt>''UNI<sub>2</sub>''</tt>.
+
 
**This means that all other input-cards that are linked to a specific universe name such as [[#det_du|det du]] and [[#src_su|src su]] can be linked to the original universe (<tt>''UNI<sub>1</sub>''</tt>) and they will automatically apply to whatever universe <tt>''UNI<sub>2</sub>''</tt> replaces <tt>''UNI<sub>1</sub>''</tt> for the branch calculation.
+
<u>Notes:</u>
*The '''stp''' variation can be used to change material density and temperature. The adjustment is made using the built-in [[Doppler-broadening preprocessor routine]] and tabular interpolation for S(&alpha;, &beta;) thermal scattering data.
+
*The user-defined variation can be used as a multi-purpose option to modify the base-input via the additional input file <tt>''MODFILE''</tt>.
*The last three parameters of the '''stp''' entry are provided only if the material has thermal scattering libraries attached to it (see the [[#therm (thermal scattering library definition)|therm card]]).
+
*The '''tra''' variation can be used to move or rotate different parts of the geometry, for example, to adjust the position of control rods in 3D geometries. The name of the transformation refers to the unit (universe, cell or surface) entry in the [[#trans (transformations)|trans card]].
+
*The '''xenon''' and '''samarium''' options can be set to enforce the concentrations of fission product poisons Xe-135 and Sm-149 to zero. By default the concentrations are read from the restart file.
+
*The '''norm''' variation can be used to change the normalization. The adjustment is made applying the parameter ''NSF'' as a multiplicative scaling factor to the given normalization.
+
*The '''gcu''' variation can be used to replace the universe for group constant generation. This variation is limited to a single-valued GCU-list.
+
*The '''reptrc''' variation can be used to replace a transport correction file with another.
+
*Variables can be used to pass information into output file, which may be convenient for the post-processing of the data.
+
*The branch card is used together with the [[#coef (coefficient matrix definition)|coef card]].
+
*For more information, see detailed description on the [[automated burnup sequence]].
+
*The replaced material ''MAT<sub>1</sub>'' of '''repm''' variation is also replaced inside mixtures. This means one can not replace a material with a mixture defined with [[#mix (mixture_definition)|mix card]] containing the replaced material (for example replacing pure water defined with [[#mat (material definition)|mat card]] by a mixture of boron and water defined with a [[#mix (mixture definition)|mix card]] containing the same pure water material).
+
*The replacing material ''MAT<sub>2</sub>'' of '''repm''' variation can not be included in geometry using other cards than the branch card with the '''repm''' variation, version 2.1.30.
+
*The "sum" option to define the material density as the sum of the constituent nuclide densities is not supported from version 2.2.0 and on.
+
  
 
=== casematrix (casematrix definition)<span id="casematrix"></span> ===
 
=== casematrix (casematrix definition)<span id="casematrix"></span> ===
Line 133: Line 233:
 
|-
 
|-
 
| <tt>''HIS_BR<sub>k</sub>''</tt>  
 
| <tt>''HIS_BR<sub>k</sub>''</tt>  
| : name of the ''k''th history variation branch
+
| : name of the ''k''-th history variation branch
 
|-
 
|-
 
| <tt>''NBU''</tt>
 
| <tt>''NBU''</tt>
Line 139: Line 239:
 
|-
 
|-
 
| <tt>''BU<sub>n</sub>''</tt>  
 
| <tt>''BU<sub>n</sub>''</tt>  
| : burnup steps at which the momentary variation branches are invoked
+
| : burnup steps at which the momentary variation branches are invoked (positive value = burnup [in MWd/kg], negative value = time [in d])
 
|-
 
|-
 
| <tt>''NBR<sub>m</sub>''</tt>
 
| <tt>''NBR<sub>m</sub>''</tt>
| : number branches in the ''m''th dimension of the burnup matrix
+
| : number branches in the ''m''-th dimension of the burnup matrix
 
|-
 
|-
 
| <tt>''BR<sub>m,i</sub>''</tt>  
 
| <tt>''BR<sub>m,i</sub>''</tt>  
| : name of the ''i''th branch in the ''m''th dimension
+
| : name of the ''i''-th branch in the ''m''-th dimension
 
|}
 
|}
  
Line 151: Line 251:
 
*The casematrix card performs multiple depletions with <tt>''NHIS''</tt> (different) historical variations and performs restarts similar as the [[#coef|coef]] input card.
 
*The casematrix card performs multiple depletions with <tt>''NHIS''</tt> (different) historical variations and performs restarts similar as the [[#coef|coef]] input card.
 
*The casematrix card creates a multi-dimensional coefficient matrix (of size <tt>''NBR<sub>1</sub>''</tt> &times; <tt>''NBR<sub>2</sub>''</tt> &times; <tt>''NBR<sub>3</sub>''</tt> &times; ... ). The automated burnup sequence performs a restart for each of the listed burnup points, and loops over the branch combinations defined by the coefficient matrix. This is repeated for each different depletion history.
 
*The casematrix card creates a multi-dimensional coefficient matrix (of size <tt>''NBR<sub>1</sub>''</tt> &times; <tt>''NBR<sub>2</sub>''</tt> &times; <tt>''NBR<sub>3</sub>''</tt> &times; ... ). The automated burnup sequence performs a restart for each of the listed burnup points, and loops over the branch combinations defined by the coefficient matrix. This is repeated for each different depletion history.
*Positive values in the burnup vector are interpreted as (MWd/kgU), negative values are interpreted as time steps in days.
 
 
*The casematrix card is used together with the [[#branch (branch definition)|branch card]] and [[Installing_and_running_Serpent#Running_casematrix_calculations|-casematrix]] running option.
 
*The casematrix card is used together with the [[#branch (branch definition)|branch card]] and [[Installing_and_running_Serpent#Running_casematrix_calculations|-casematrix]] running option.
 
*Multiple casematrix cards can be given in a single input file.
 
*Multiple casematrix cards can be given in a single input file.
Line 214: Line 313:
 
*There are three types of cells: material cells, filled cells and outside cells. Filled cells are identified by providing the key word <tt>'''fill'''</tt>, followed by the universe filling the cell. If the key word is missing, the third entry is interpreted as the material filling the cell. Outside cells are identified by replacing the material name with key word <tt>'''outside'''</tt>.
 
*There are three types of cells: material cells, filled cells and outside cells. Filled cells are identified by providing the key word <tt>'''fill'''</tt>, followed by the universe filling the cell. If the key word is missing, the third entry is interpreted as the material filling the cell. Outside cells are identified by replacing the material name with key word <tt>'''outside'''</tt>.
 
*Cells defined without surfaces are treated as infinite, from version 2.1.32 on.
 
*Cells defined without surfaces are treated as infinite, from version 2.1.32 on.
*Void cells can be defined by setting the material name to "void"
+
*Void cells can be defined by setting the material name to "<tt>void</tt>"
 
*When the geometry is set up, the root universe must always be defined. By default the root universe is named "0", and it can be changed with the [[#set root|set root]] option.
 
*When the geometry is set up, the root universe must always be defined. By default the root universe is named "0", and it can be changed with the [[#set root|set root]] option.
 
*Outside cells are used to define the part of the geometry that does not belong to the model. When the particle enters an outside cell, [[#set bc|boundary conditions]] are applied. It is important that the geometry model is non-re-entrant (convex) when vacuum boundary conditions are used. Delta-tracking might miss the boundary conditions in a re-entrant (concave) outer surface.
 
*Outside cells are used to define the part of the geometry that does not belong to the model. When the particle enters an outside cell, [[#set bc|boundary conditions]] are applied. It is important that the geometry model is non-re-entrant (convex) when vacuum boundary conditions are used. Delta-tracking might miss the boundary conditions in a re-entrant (concave) outer surface.
Line 235: Line 334:
 
|-
 
|-
 
| <tt>''BU<sub>n</sub>''</tt>  
 
| <tt>''BU<sub>n</sub>''</tt>  
| : burnup steps at which the branches are invoked
+
| : burnup steps at which the branches are invoked (positive value = burnup [in MWd/kg], negative value = time [in d])
 
|-
 
|-
 
| <tt>''NBR<sub>m</sub>''</tt>
 
| <tt>''NBR<sub>m</sub>''</tt>
| : number branches in the ''m''th dimension of the burnup matrix
+
| : number branches in the ''m''-th dimension of the burnup matrix
 
|-
 
|-
 
| <tt>''BR<sub>m,i</sub>''</tt>  
 
| <tt>''BR<sub>m,i</sub>''</tt>  
| : name of the ''i''th branch in the ''m''th dimension
+
| : name of the ''i''-th branch in the ''m''-th dimension
 
|}
 
|}
  
 
<u>Notes:</u>
 
<u>Notes:</u>
  
*The coef card creates a multi-dimensional coefficient matrix (of size <tt>''NBR<sub>1</sub>''</tt> &times; <tt>''NBR<sub>2</sub>''</tt> &times; <tt>''NBR<sub>3</sub>''</tt> &times; ... ). The automated burnup sequence performs a restart for each of the listed burnup points, and loops over the branch combinations defined by the coefficient matrix.
+
*The coef card creates a multi-dimensional coefficient matrix (of size <tt>''NBR<sub>1</sub>''</tt> &times; <tt>''NBR<sub>2</sub>''</tt> &times; <tt>''NBR<sub>3</sub>''</tt> &times; ... ). The automated burnup sequence performs a restart for each of the listed burnup points, and loops over the branch combinations defined by the coefficient matrix.  
*Positive values in the burnup vector are interpreted as (MWd/kgU), negative values are interpreted as time steps in days.  
+
 
*The coef card is used together with the [[#branch (branch definition)|branch]] card.
 
*The coef card is used together with the [[#branch (branch definition)|branch]] card.
 
*For multiple historical variations or historical conditions defined using a [[#branch (branch definition)|branch]] card, see the [[#casematrix (casematrix definition)|casematrix]] card.
 
*For multiple historical variations or historical conditions defined using a [[#branch (branch definition)|branch]] card, see the [[#casematrix (casematrix definition)|casematrix]] card.
Line 268: Line 366:
 
|-
 
|-
 
| <tt>''N<sub>X</sub>''</tt>
 
| <tt>''N<sub>X</sub>''</tt>
| : number of cells in the x direction
+
| : number of cells in the x-direction
 
|-
 
|-
 
| <tt>''X<sub>MIN</sub>''</tt>
 
| <tt>''X<sub>MIN</sub>''</tt>
| : mesh lower x boundary
+
| : mesh lower x-boundary [in cm]
 
|-
 
|-
 
| <tt>''X<sub>MAX</sub>''</tt>
 
| <tt>''X<sub>MAX</sub>''</tt>
| : mesh higher x boundary
+
| : mesh higher x-boundary [in cm]
 
|-
 
|-
 
| <tt>''N<sub>Y</sub>''</tt>
 
| <tt>''N<sub>Y</sub>''</tt>
| : number of cells in the y direction
+
| : number of cells in the y-direction
 
|-
 
|-
 
| <tt>''Y<sub>MIN</sub>''</tt>
 
| <tt>''Y<sub>MIN</sub>''</tt>
| : mesh lower y boundary
+
| : mesh lower y-boundary [in cm]
 
|-
 
|-
 
| <tt>''Y<sub>MAX</sub>''</tt>
 
| <tt>''Y<sub>MAX</sub>''</tt>
| : mesh higher y boundary
+
| : mesh higher y-boundary [in cm]
 
|-
 
|-
 
| <tt>''N<sub>Z</sub>''</tt>
 
| <tt>''N<sub>Z</sub>''</tt>
| : number of cells in the z direction
+
| : number of cells in the z-direction
 
|-
 
|-
 
| <tt>''Z<sub>MIN</sub>''</tt>
 
| <tt>''Z<sub>MIN</sub>''</tt>
| : mesh lower z boundary
+
| : mesh lower z-boundary [in cm]
 
|-
 
|-
 
| <tt>''Z<sub>MAX</sub>''</tt>
 
| <tt>''Z<sub>MAX</sub>''</tt>
| : mesh higher z boundary
+
| : mesh higher z-boundary [in cm]
 
|}
 
|}
  
Line 310: Line 408:
 
|-
 
|-
 
| <tt>''R<sub>MIN</sub>''</tt>
 
| <tt>''R<sub>MIN</sub>''</tt>
| : mesh inner radial boundary
+
| : mesh inner radial boundary [in cm]
 
|-
 
|-
 
| <tt>''R<sub>MAX</sub>''</tt>
 
| <tt>''R<sub>MAX</sub>''</tt>
| : mesh outer radial boundary
+
| : mesh outer radial boundary [in cm]
 
|-
 
|-
 
| <tt>''N<sub>PHI</sub>''</tt>
 
| <tt>''N<sub>PHI</sub>''</tt>
Line 332: Line 430:
 
|-
 
|-
 
| <tt>''X<sub>0</sub>''</tt>
 
| <tt>''X<sub>0</sub>''</tt>
| : mesh horizontal origin x-coordinate
+
| : mesh horizontal origin x-coordinate [in cm]
 
|-
 
|-
 
| <tt>''Y<sub>0</sub>''</tt>
 
| <tt>''Y<sub>0</sub>''</tt>
| : mesh horizontal origin y-coordinate
+
| : mesh horizontal origin y-coordinate [in cm]
 
|-
 
|-
 
| <tt>''PITCH''</tt>
 
| <tt>''PITCH''</tt>
| : mesh horizontal pitch (equal to cell flat-to-flat width)
+
| : mesh horizontal pitch (equal to cell flat-to-flat width) [in cm]
 
|-
 
|-
 
| <tt>''Z<sub>MIN</sub>''</tt>
 
| <tt>''Z<sub>MIN</sub>''</tt>
| : mesh lower z boundary
+
| : mesh lower z-boundary [in cm]
 
|-
 
|-
 
| <tt>''Z<sub>MAX</sub>''</tt>
 
| <tt>''Z<sub>MAX</sub>''</tt>
| : mesh higher z boundary
+
| : mesh higher z-boundary [in cm]
 
|-
 
|-
 
| <tt>''N<sub>X</sub>''</tt>
 
| <tt>''N<sub>X</sub>''</tt>
| : number of cells in the x direction
+
| : number of cells in the x-direction
 
|-
 
|-
 
| <tt>''N<sub>Y</sub>''</tt>
 
| <tt>''N<sub>Y</sub>''</tt>
| : number of cells in the y direction
+
| : number of cells in the y-direction
 
|-
 
|-
 
| <tt>''N<sub>Z</sub>''</tt>
 
| <tt>''N<sub>Z</sub>''</tt>
| : number of cells in the z direction
+
| : number of cells in the z-direction
 
|}
 
|}
  
Line 369: Line 467:
 
|-
 
|-
 
| <tt>''X<sub>0</sub>''</tt>
 
| <tt>''X<sub>0</sub>''</tt>
| : mesh horizontal origin x-coordinate
+
| : mesh horizontal origin x-coordinate [in cm]
 
|-
 
|-
 
| <tt>''Y<sub>0</sub>''</tt>
 
| <tt>''Y<sub>0</sub>''</tt>
| : mesh horizontal origin y-coordinate
+
| : mesh horizontal origin y-coordinate [in cm]
 
|-
 
|-
 
| <tt>''PITCH''</tt>
 
| <tt>''PITCH''</tt>
| : mesh horizontal pitch (equal to cell flat-to-flat width)
+
| : mesh horizontal pitch (equal to cell flat-to-flat width) [in cm]
 
|-
 
|-
 
| <tt>''Z<sub>MIN</sub>''</tt>
 
| <tt>''Z<sub>MIN</sub>''</tt>
| : mesh lower z boundary
+
| : mesh lower z-boundary [in cm]
 
|-
 
|-
 
| <tt>''Z<sub>MAX</sub>''</tt>
 
| <tt>''Z<sub>MAX</sub>''</tt>
| : mesh higher z boundary
+
| : mesh higher z-boundary [in cm]
 
|-
 
|-
 
| <tt>''N<sub>X</sub>''</tt>
 
| <tt>''N<sub>X</sub>''</tt>
| : number of cells in the x direction
+
| : number of cells in the x-direction
 
|-
 
|-
 
| <tt>''N<sub>Y</sub>''</tt>
 
| <tt>''N<sub>Y</sub>''</tt>
| : number of cells in the y direction
+
| : number of cells in the y-direction
 
|-
 
|-
 
| <tt>''N<sub>Z</sub>''</tt>
 
| <tt>''N<sub>Z</sub>''</tt>
| : number of cells in the z direction
+
| : number of cells in the z-direction
 
|}
 
|}
  
Line 405: Line 503:
 
|-
 
|-
 
| <tt>''N<sub>X</sub>''</tt>
 
| <tt>''N<sub>X</sub>''</tt>
| : number of cells in the x direction
+
| : number of cells in the x-direction
 
|-
 
|-
 
| <tt>''N<sub>Y</sub>''</tt>
 
| <tt>''N<sub>Y</sub>''</tt>
| : number of cells in the y direction
+
| : number of cells in the y-direction
 
|-
 
|-
 
| <tt>''N<sub>Z</sub>''</tt>
 
| <tt>''N<sub>Z</sub>''</tt>
| : number of cells in the z direction
+
| : number of cells in the z-direction
 
|-
 
|-
 
| <tt>''X<sub>i</sub>''</tt>
 
| <tt>''X<sub>i</sub>''</tt>
| : <tt>''N<sub>X</sub>'' + 1</tt> mesh boundaries in the x direction
+
| : <tt>''N<sub>X</sub>'' + 1</tt> mesh boundaries in the x-direction [in cm]
 
|-
 
|-
 
| <tt>''Y<sub>i</sub>''</tt>
 
| <tt>''Y<sub>i</sub>''</tt>
| : <tt>''N<sub>Y</sub>'' + 1</tt> mesh boundaries in the y direction
+
| : <tt>''N<sub>Y</sub>'' + 1</tt> mesh boundaries in the y-direction [in cm]
 
|-
 
|-
 
| <tt>''Z<sub>i</sub>''</tt>
 
| <tt>''Z<sub>i</sub>''</tt>
| : <tt>''N<sub>Z</sub>'' + 1</tt> mesh boundaries in the z direction
+
| : <tt>''N<sub>Z</sub>'' + 1</tt> mesh boundaries in the z-direction [in cm]
 
|}
 
|}
  
Line 441: Line 539:
 
|-
 
|-
 
| <tt>''R<sub>i</sub>''</tt>
 
| <tt>''R<sub>i</sub>''</tt>
| : <tt>''N<sub>R</sub>'' + 1</tt> mesh boundaries in the r direction
+
| : <tt>''N<sub>R</sub>'' + 1</tt> mesh boundaries in the radial direction [in cm]
 
|}
 
|}
  
  '''datamesh''' ''NAME'' '''9''' ''N<sub>LEVEL</sub>''  ''MESH<sub>1</sub>'' ... ''MESH<sub>N<sub>LEVEL</sub></sub>''  
+
  '''datamesh''' ''NAME'' '''9''' ''USE<sub>LC</sub>'' ''N<sub>LEVEL</sub>''  ''MESH<sub>1</sub>'' ... ''MESH<sub>N<sub>LEVEL</sub></sub>''  
  
 
Defines a regular nested mesh that can be linked to detectors, interfaces etc.
 
Defines a regular nested mesh that can be linked to detectors, interfaces etc.
Line 464: Line 562:
 
<u>Notes:</u>
 
<u>Notes:</u>
 
*When Serpent makes the mesh search for a specific collision point it will save the collision mesh cell temporarily so that the cell search is conducted at most once even when scoring multiple estimators using the same mesh.
 
*When Serpent makes the mesh search for a specific collision point it will save the collision mesh cell temporarily so that the cell search is conducted at most once even when scoring multiple estimators using the same mesh.
 +
*The nested data meshes (type 9) take the coordinates' level from the <tt>''USE<sub>LC</sub>''</tt> parameter defined in the nested mesh itself and use it in the subsequent sub-meshes, overriding the <tt>''USE<sub>LC</sub>''</tt> parameter defined on those.
  
 
=== dep (depletion history)<span id="dep"></span> ===
 
=== dep (depletion history)<span id="dep"></span> ===
Line 480: Line 579:
  
 
The possible step types are:
 
The possible step types are:
{| class="wikitable" style="text-align: left;"
+
::{| class="wikitable" style="text-align: left;"
 
! Type
 
! Type
 
! Description
 
! Description
Line 563: Line 662:
 
           [ [[#det_dy|'''dy''']] ''Y<sub>MIN</sub>'' ''Y<sub>MAX</sub>'' ''N<sub>Y</sub>'' ]
 
           [ [[#det_dy|'''dy''']] ''Y<sub>MIN</sub>'' ''Y<sub>MAX</sub>'' ''N<sub>Y</sub>'' ]
 
           [ [[#det_dz|'''dz''']] ''Z<sub>MIN</sub>'' ''Z<sub>MAX</sub>'' ''N<sub>Z</sub>'' ]
 
           [ [[#det_dz|'''dz''']] ''Z<sub>MIN</sub>'' ''Z<sub>MAX</sub>'' ''N<sub>Z</sub>'' ]
           [ [[#det_dn|'''dn''']] ''TYPE'' ''MIN<sub>1</sub>'' ''MAX<sub>1</sub>'' ''N<sub>1</sub>'' ''MIN<sub>2</sub>'' ''MAX<sub>2</sub>'' ''N<sub>2</sub>'' ''MIN<sub>3</sub>'' ''MAX<sub>3</sub>'' ''N<sub>3</sub>'' ] <small>1/2</small> [ [[#det_dn|'''dn''']] ''TYPE'' ''N<sub>1</sub>'' ''N<sub>2</sub>'' ''N<sub>3</sub>'' ''LIM<sub>11</sub>''...''LIM<sub>1N+1</sub>'' ''LIM<sub>21</sub>''...''LIM<sub>2N+1</sub>'' ''LIM<sub>31</sub>''...''LIM<sub>3N+1</sub>'' ] <small>3/4</small>
+
           [ [[#det_dn1|'''dn''']] ''TYPE'' ''MIN<sub>1</sub>'' ''MAX<sub>1</sub>'' ''N<sub>1</sub>'' ''MIN<sub>2</sub>'' ''MAX<sub>2</sub>'' ''N<sub>2</sub>'' ''MIN<sub>3</sub>'' ''MAX<sub>3</sub>'' ''N<sub>3</sub>'' ]
 +
          [ [[#det_dn2|'''dn''']] ''TYPE'' ''N<sub>1</sub>'' ''N<sub>2</sub>'' ''N<sub>3</sub>'' ''LIM<sub>11</sub>''...''LIM<sub>1N+1</sub>'' ''LIM<sub>21</sub>''...''LIM<sub>2N+1</sub>'' ''LIM<sub>31</sub>''...''LIM<sub>3N+1</sub>'' ]
 
           [ [[#det_dh|'''dh''']] ''TYPE'' ''X<sub>0</sub>'' ''Y<sub>0</sub>'' ''PITCH'' ''N<sub>1</sub>'' ''N<sub>2</sub>'' ''Z<sub>MIN</sub>'' ''Z<sub>MAX</sub>'' ''N<sub>Z</sub>'' ]
 
           [ [[#det_dh|'''dh''']] ''TYPE'' ''X<sub>0</sub>'' ''Y<sub>0</sub>'' ''PITCH'' ''N<sub>1</sub>'' ''N<sub>2</sub>'' ''Z<sub>MIN</sub>'' ''Z<sub>MAX</sub>'' ''N<sub>Z</sub>'' ]
 
           [ [[#det_dumsh|'''dumsh''']] ''UNI'' ''N<sub>C</sub>'' ''CELL<sub>0</sub>'' ''BIN<sub>0</sub>'' ''CELL<sub>1</sub>'' ''BIN<sub>1</sub>'' ... ]
 
           [ [[#det_dumsh|'''dumsh''']] ''UNI'' ''N<sub>C</sub>'' ''CELL<sub>0</sub>'' ''BIN<sub>0</sub>'' ''CELL<sub>1</sub>'' ''BIN<sub>1</sub>'' ... ]
Line 579: Line 679:
 
           [ [[#det_dphb|'''dphb''']] ''PHB'' ]
 
           [ [[#det_dphb|'''dphb''']] ''PHB'' ]
 
           [ [[#det_dmesh|'''dmesh''']] ''MESH'' ]
 
           [ [[#det_dmesh|'''dmesh''']] ''MESH'' ]
Detector definition. The first parameter:
+
Detector definition. The two first parameters:
 
   
 
   
 
{|
 
{|
 +
| <tt>''NAME''</tt>
 +
| : detector name
 +
|-
 
| <tt>''PART''</tt>
 
| <tt>''PART''</tt>
 
| : particle type (n = neutron, p = photon)
 
| : particle type (n = neutron, p = photon)
 
|}
 
|}
  
is optional in single particle simulations. The remaining parameters are defined by separate key words followed by the input values.
+
The remaining parameters are defined by separate key words followed by the input values.  
 +
 
 +
<u>Notes:</u>
 +
*The particle type <tt>''PART''</tt> is optional in single particle simulations.
 +
*The detectors estimates are integrated values over the space, angle, energy and time domains.
 +
*A detector with an associated discretization in space, angle, energy and/or time turns into multiple bins. Each bin results are correspondingly integrated over the discretization domain.
 +
*A single detector card may include one or several detector types. If multiple detectors are defined, the results are correspondingly divided into multiple bins.
 +
 
  
 +
<u>Detector types:</u>
  
 
Detector response (<tt>'''dr'''</tt>):<span id="det_dr"></span>
 
Detector response (<tt>'''dr'''</tt>):<span id="det_dr"></span>
Line 602: Line 713:
 
*If the detector is assigned with multiple responses, the results are divided correspondingly into separate bins.
 
*If the detector is assigned with multiple responses, the results are divided correspondingly into separate bins.
 
*The response numbers are [[ENDF reaction MT's and macroscopic reaction numbers|ENDF reaction MT's and special reaction types]].
 
*The response numbers are [[ENDF reaction MT's and macroscopic reaction numbers|ENDF reaction MT's and special reaction types]].
*Positive response numbers are associated with microscopic cross sections and the result is independent of material density. Materials for microscopic cross sections must consist of a single nuclide.
+
**Positive response numbers:
*Microscopic reactions to ground and isomeric states can be calculated by adding "g" or "m" at the end of the reaction number (e.g. 102g and 102m refer to radiative capture to ground and isomeric states, respectively). This option is available only for nuclides with [[#set_bralib|branching ratios]].
+
*** They are associated with microscopic cross sections
*Negative response numbers are associated with macroscopic cross sections and special types, and the result is multiplied by material density.
+
*** The detector result is independent of the material density.  
*The response material in the <tt>'''dr'''</tt> entry must not be confused with the material in the <tt>'''dm'''</tt> entry. The former defines the material for the response function, while the latter determines the volume of integration.
+
*** Materials associated to microscopic cross sections must consist of a single nuclide.
 +
***Microscopic reactions to ground and isomeric states can be calculated by adding "<tt>g</tt>" or "<tt>m</tt>" at the end of the reaction number (e.g. 102g and 102m refer to radiative capture to ground and isomeric states, respectively). This option is available only for nuclides with [[#set_bralib|branching ratios]].
 +
**Negative response numbers:
 +
*** They are associated with macroscopic cross sections and special types
 +
*** The detector result is multiplied by material density
 +
*The response material in the <tt>'''dr'''</tt> entry must not be confused with the material in the <tt>'''dm'''</tt> entry.
 +
** The former defines the material for the response function, while the latter determines the volume of integration.
 +
*The "<tt>void</tt>" entry allows the response not to be pre-assigned with a specific material (when the detector scores in a collision, the cross-section is taken from the material at the collision point - e.g., to calculate  integral reaction rates over regions composed of multiple materials)
 +
** It only can be used with negative response numbers.
 
*By default, Serpent allows a detector to have at most 10,000,000 bins.
 
*By default, Serpent allows a detector to have at most 10,000,000 bins.
  
Line 613: Line 732:
 
{|
 
{|
 
| <tt>''VOL''</tt>
 
| <tt>''VOL''</tt>
| : volume in cm<sup>3</sup> (3D geometry) or cross-sectional area in cm<sup>2</sup> (2D geometry)
+
| : volume [in cm<sup>3</sup>] (3D geometry) or cross-sectional area [in cm<sup>2</sup>] (2D geometry)
 
|}
 
|}
  
 
<u>Notes:</u>
 
<u>Notes:</u>
*The results are divided by detector volume, which is by default set to 1.
+
*The results are divided by detector bin-volume (default value: 1.0)
 +
*In the case of surface detectors, ''VOL'' represents the surface area [in cm<sup>2</sup>] (3D geometry) or the surface length [in cm] (2D geometry).
  
  
Line 665: Line 785:
  
  
Cartesian mesh detector (<tt>'''dx'''</tt>, <tt>'''dy'''</tt> and <tt>'''dz'''</tt>):<span id="det_dx"></span><span id="det_dy"></span><span id="det_dz"></span>
+
Detector evenly-spaced Cartesian mesh (<tt>'''dx'''</tt>, <tt>'''dy'''</tt> and <tt>'''dz'''</tt>):<span id="det_dx"></span><span id="det_dy"></span><span id="det_dz"></span>
  
 
{|
 
{|
 
| <tt>''X<sub>MIN</sub>''</tt>
 
| <tt>''X<sub>MIN</sub>''</tt>
| : minimum x-coordinate of the detector mesh
+
| : minimum x-coordinate of the detector mesh [in cm]
 
|-
 
|-
 
| <tt>''X<sub>MAX</sub>''</tt>
 
| <tt>''X<sub>MAX</sub>''</tt>
| : maximum x-coordinate of the detector mesh
+
| : maximum x-coordinate of the detector mesh [in cm]
 
|-
 
|-
 
| <tt>''N<sub>X</sub>''</tt>
 
| <tt>''N<sub>X</sub>''</tt>
Line 678: Line 798:
 
|-
 
|-
 
| <tt>''Y<sub>MIN</sub>''</tt>
 
| <tt>''Y<sub>MIN</sub>''</tt>
| : minimum y-coordinate of the detector mesh
+
| : minimum y-coordinate of the detector mesh [in cm]
 
|-
 
|-
 
| <tt>''Y<sub>MAX</sub>''</tt>
 
| <tt>''Y<sub>MAX</sub>''</tt>
| : maximum y-coordinate of the detector mesh
+
| : maximum y-coordinate of the detector mesh [in cm]
 
|-
 
|-
 
| <tt>''N<sub>Y</sub>''</tt>
 
| <tt>''N<sub>Y</sub>''</tt>
Line 687: Line 807:
 
|-
 
|-
 
| <tt>''Z<sub>MIN</sub>''</tt>
 
| <tt>''Z<sub>MIN</sub>''</tt>
| : minimum z-coordinate of the detector mesh
+
| : minimum z-coordinate of the detector mesh [in cm]
 
|-
 
|-
 
| <tt>''Z<sub>MAX</sub>''</tt>
 
| <tt>''Z<sub>MAX</sub>''</tt>
| : maximum z-coordinate of the detector mesh
+
| : maximum z-coordinate of the detector mesh [in cm]
 
|-
 
|-
 
| <tt>''N<sub>Z</sub>''</tt>
 
| <tt>''N<sub>Z</sub>''</tt>
Line 697: Line 817:
  
 
<u>Notes:</u>
 
<u>Notes:</u>
*The mesh detectors can be used to sub-divide the results into multiple spatial bins. For a Cartesian mesh the division is provided with separate entries in x-, y- and z- locations.
+
*The mesh detectors can be used to sub-divide the results into multiple evenly-spaced bins.  
 +
*For a Cartesian mesh the division is provided with separate entries in x-, y- and z- locations (<tt>'''dx'''</tt>, <tt>'''dy'''</tt> and <tt>'''dz'''</tt>, respectively).
  
  
Curvilinear and unevenly-spaced mesh detector (<tt>'''dn'''</tt>):<span id="det_dn"></span>
+
Detector evenly-spaced curvilinear mesh (<tt>'''dn'''</tt>):<span id="det_dn1"></span>
  
 
{|
 
{|
 
| <tt>''TYPE''</tt>  
 
| <tt>''TYPE''</tt>  
| : Type of curvilinear mesh - 1 = cylindrical (dimensions ''r'', ''&theta;'', ''z''), 2 = spherical (dimensions ''r'', ''&theta;'', ''&phi;''), 3 = unevenly-spaced orthogonal (dimensions ''x'', ''y'', ''z''), 4 = unevenly-spaced cylindrical (dimensions ''r'', ''&theta;'', ''z'')
+
| : type of curvilinear mesh - 1 = cylindrical (dimensions ''r'', ''&theta;'', ''z''), 2 = spherical (dimensions ''r'', ''&theta;'', ''&phi;'')
 
|-
 
|-
 
| <tt>''MIN<sub>n</sub>''</tt>
 
| <tt>''MIN<sub>n</sub>''</tt>
| : Minimum value of coordinate ''n'' for the mesh division (lengths in cm, angles in degrees).
+
| : minimum value of ''n''-coordinate for the mesh division [in cm (''r'', ''z''), in degrees (''&theta;'', ''&phi;'')].
 
|-
 
|-
 
| <tt>''MAX<sub>n</sub>''</tt>
 
| <tt>''MAX<sub>n</sub>''</tt>
| : Maximum value of coordinate ''n'' for the mesh division (lengths in cm, angles in degrees).
+
| : maximum value of ''n''-coordinate for the mesh division [in cm (''r'', ''z''), in degrees (''&theta;'', ''&phi;'')].
 
|-
 
|-
 
| <tt>''N<sub>n</sub>''</tt>
 
| <tt>''N<sub>n</sub>''</tt>
| : Number of bins in the ''n'' coordinate direction (the radial division will be equal ''r'', not equal volume, in evenly-spaced types 1/2).
+
| : number of bins in the ''n''-coordinate direction (the radial division will be equal ''r'', not equal volume).
|-
+
| <tt>''LIM<sub>nm</sub>''</tt>
+
| : Mesh boundary ''m'' in the ''n'' coordinate direction (lengths in cm, angles in degrees).
+
 
|}
 
|}
  
 
<u>Notes:</u>
 
<u>Notes:</u>
 
*All parameters must be provided, even for one- or two-dimensional curvilinear meshes.
 
*All parameters must be provided, even for one- or two-dimensional curvilinear meshes.
*The results are not divided by cell volume (difference to MCNP mesh tally).
 
 
*By default, the curvilinear mesh detectors use the global (universe 0) coordinate system for scoring. If the <tt>''TYPE''</tt> parameter is given as a negative value (e.g. -1) the lowest level coordinates are used instead.
 
*By default, the curvilinear mesh detectors use the global (universe 0) coordinate system for scoring. If the <tt>''TYPE''</tt> parameter is given as a negative value (e.g. -1) the lowest level coordinates are used instead.
*The syntax for curvilinear (evenly-spaced) mesh detectors (1/-1, 2/-2) differs from the unevenly-spaced mesh ones (3/-3, 4/-4).
 
  
  
Hexagonal mesh detector (<tt>'''dh'''</tt>):<span id="det_dh"></span>
+
Detector unevenly-spaced mesh (<tt>'''dn'''</tt>):<span id="det_dn2"></span>
  
 
{|
 
{|
 
| <tt>''TYPE''</tt>  
 
| <tt>''TYPE''</tt>  
| : Type of hexagonal mesh (2 = flat face perpendicular to x-axis, 3 = flat face perpendicular to y-axis)
+
| : type of curvilinear mesh - 3 = unevenly-spaced orthogonal (dimensions ''x'', ''y'', ''z''), 4 = unevenly-spaced cylindrical (dimensions ''r'', ''&theta;'', ''z'')
 +
|-
 +
| <tt>''N<sub>n</sub>''</tt>
 +
| : number of bins in the ''n''-coordinate direction
 +
|-
 +
| <tt>''LIM<sub>nm</sub>''</tt>
 +
| : mesh ''m''-boundary in the ''n''-coordinate direction [in cm (''r'', ''z''), in degrees (''&theta;'', ''&phi;'')].
 +
|}
 +
 
 +
<u>Notes:</u>
 +
*All parameters must be provided, even for one- or two-dimensional meshes.
 +
*By default, the unevenly-spaced mesh detectors use the global (universe 0) coordinate system for scoring. If the <tt>''TYPE''</tt> parameter is given as a negative value (e.g. -1) the lowest level coordinates are used instead.
 +
 
 +
 
 +
Detector hexagonal mesh (<tt>'''dh'''</tt>):<span id="det_dh"></span>
 +
 
 +
{|
 +
| <tt>''TYPE''</tt>
 +
| : type of hexagonal mesh (2 = flat face perpendicular to x-axis, 3 = flat face perpendicular to y-axis)
 
|-
 
|-
 
| <tt>''X<sub>0</sub>''</tt>, <tt>''Y<sub>0</sub>''</tt>
 
| <tt>''X<sub>0</sub>''</tt>, <tt>''Y<sub>0</sub>''</tt>
| : coordinates of mesh center
+
| : coordinates of mesh center [in cm]
 
|-
 
|-
 
| <tt>''PITCH''</tt>
 
| <tt>''PITCH''</tt>
| : mesh pitch
+
| : mesh pitch [in cm]
 
|-
 
|-
 
| <tt>''N<sub>1</sub>''</tt>, <tt>''N<sub>1</sub>''</tt>
 
| <tt>''N<sub>1</sub>''</tt>, <tt>''N<sub>1</sub>''</tt>
Line 742: Line 876:
 
|-
 
|-
 
| <tt>''Z<sub>MIN</sub>''</tt>
 
| <tt>''Z<sub>MIN</sub>''</tt>
| : minimum z-coordinate of the detector mesh
+
| : minimum z-coordinate of the detector mesh [in cm]
 
|-
 
|-
 
| <tt>''Z<sub>MAX</sub>''</tt>
 
| <tt>''Z<sub>MAX</sub>''</tt>
| : maximum z-coordinate of the detector mesh
+
| : maximum z-coordinate of the detector mesh [in cm]
 
|-
 
|-
 
| <tt>''N<sub>Z</sub>''</tt>
 
| <tt>''N<sub>Z</sub>''</tt>
Line 755: Line 889:
  
  
Unstructured mesh detector (<tt>'''dumsh'''</tt>):<span id="det_dumsh"></span>
+
Detector unstructured mesh (<tt>'''dumsh'''</tt>):<span id="det_dumsh"></span>
  
 
{|
 
{|
Line 770: Line 904:
 
<u>Notes:</u>
 
<u>Notes:</u>
 
*The polyhedral cells in [[Unstructured mesh-based geometry type|unstructured mesh based geometries]] are indexed.
 
*The polyhedral cells in [[Unstructured mesh-based geometry type|unstructured mesh based geometries]] are indexed.
*This detector option allows collecting results from the cells into an arbitrary number of bins. One or multiple cells can be mapped into a single bin.
+
*This detector option allows collecting results from the cells into an arbitrary number of bins.  
 +
*One or multiple cells can be mapped into a single bin.
  
  
Line 782: Line 917:
 
<u>Notes:</u>
 
<u>Notes:</u>
 
*The results are divided into multiple energy bins based on the grid structure.
 
*The results are divided into multiple energy bins based on the grid structure.
*Energy grid structures are defined using the [[#ene (energy grid definition)|ene card]]. [[Pre-defined energy group structures]] can not be directly used in detectors, they have to be redefined using for example the fourth type of ene card.
+
*Energy grid structures are defined using the [[#ene (energy grid definition)|ene card]].  
*The energy boundaries of photon photon pulse-height and photon heat analog detectors are solely defined by the associated energy grid and not limited by the unionized energy grid defining the model. That means that analog detectors might collect scores below the physics model minimum energy bound, without a cut-off, if the energy grid sets it.
+
**[[Pre-defined energy group structures]] can not be directly used in detectors, they have to be redefined using for example the type "<tt>4</tt>" of [[#ene|ene card]].
 +
*The energy boundaries of photon photon pulse-height and photon heat analog detectors are solely defined by the associated energy grid and not limited by the unionized energy grid defining the model.  
 +
**That means that analog detectors might collect scores below the physics model minimum energy bound, without a cut-off, if the energy grid sets it.
  
  
Line 799: Line 936:
  
  
Surface current / flux detector (<tt>'''ds'''</tt>):<span id="det_ds"></span>
+
Detector current / flux surface (<tt>'''ds'''</tt>):<span id="det_ds"></span>
  
 
{|
 
{|
Line 811: Line 948:
 
<u>Notes:</u>
 
<u>Notes:</u>
 
*With this option the detector calculates the particle flux over or current through a given surface.  
 
*With this option the detector calculates the particle flux over or current through a given surface.  
*The surface flux mode is invoked by setting the direction parameter to -2, otherwise this parameter defines the current direction with respect to surface normal.
+
*Flux mode:
*Responses are not allowed with current detectors, and with flux detectors, the material name at the collision point has to be specified (<tt>"void"</tt> is not allowed).
+
**The surface flux mode is invoked by setting the direction parameter to "<tt>-2</tt>", otherwise this parameter defines the current direction with respect to surface normal.
*The use of single-bin mesh and cell detectors is allowed to define the integration surface of the detector, from version 2.1.32 on.
+
*Current mode:
*The surface is treated separate from the geometry, and its position is always relative to the origin of the root universe. This is the case even if the surface is part of the geometry in another universe.
+
**Responses are not allowed with current detectors, and with flux detectors, the material name at the collision point has to be specified (<tt>"void"</tt> is not allowed).
 +
*The use of single-bin mesh and cell detectors is allowed to further define the surface and integration domain of the detector, from version 2.1.32 on.
 +
*The surface is treated separate from the geometry, and its position is always relative to the origin of the root universe.  
 +
**This is the case even if the surface is part of the geometry in another universe.
 
*The results are integrated over the surface area (other detectors integrate over volume).
 
*The results are integrated over the surface area (other detectors integrate over volume).
  
Line 821: Line 961:
  
 
{|
 
{|
| <tt>''COS<sub>Y</sub>''</tt>
+
| <tt>''COS<sub>X</sub>''</tt>
 
| : component of the direction vector parallel to x-axis
 
| : component of the direction vector parallel to x-axis
 
|-
 
|-
Line 835: Line 975:
  
  
Super-imposed track-length detector (<tt>'''dtl'''</tt>):<span id="det_dtl"></span>
+
Detector super-imposed track-length (<tt>'''dtl'''</tt>):<span id="det_dtl"></span>
  
 
{|
 
{|
Line 844: Line 984:
 
<u>Notes:</u>
 
<u>Notes:</u>
 
*This option can be used to apply the track-length estimator for calculating reaction rates inside regions defined by a single surface (sphere, cylinder, cuboid, etc.)
 
*This option can be used to apply the track-length estimator for calculating reaction rates inside regions defined by a single surface (sphere, cylinder, cuboid, etc.)
 +
*The surface is treated separate from the geometry, and its position is always relative to the origin of the root universe.
 +
**This is the case even if the surface is part of the geometry in another universe.
 
*The purpose of the track-length detector is to provide better statistics for special applications (activation wire measurements, etc.).
 
*The purpose of the track-length detector is to provide better statistics for special applications (activation wire measurements, etc.).
*The surface is treated separate from the geometry, and its position is always relative to the origin of the root universe. This is the case even if the surface is part of the geometry in another universe.
+
*For more information see the detailed description on [[delta- and surface-tracking]] and [[Result estimators#Implicit estimators|result estimators]].
  
  
Line 855: Line 997:
 
|-
 
|-
 
| <tt>''FRAC''</tt>
 
| <tt>''FRAC''</tt>
| : fraction of recorded scores and ascii/binary option (positive value = ascii, negative value = binary)
+
| : fraction of recorded scores and ASCII/binary option (positive value = ASCII, negative value = binary)
 
|}
 
|}
  
 
<u>Notes:</u>
 
<u>Notes:</u>
 
*This option can be used to write the scored points in a file.
 
*This option can be used to write the scored points in a file.
*When used with the surface current detector this option can provide surface source distributions for other calculations.
 
 
*The fraction parameters gives the probability that the score is written in the file and it can be used to reduce the file size in long simulations.
 
*The fraction parameters gives the probability that the score is written in the file and it can be used to reduce the file size in long simulations.
*Source files can be read using the <tt>'''sf'''</tt> entry of [[#src_sf|source cards]].
+
*When used with the surface current detector this option can provide surface source distributions for other calculations.
 +
*Source files can be read using the <tt>'''sf'''</tt> entry of [[#src_sf|source card]].
  
  
Special types (<tt>'''dt'''</tt>):<span id="det_dt"></span>
+
Detector special types (<tt>'''dt'''</tt>):<span id="det_dt"></span>
  
 
{|
 
{|
Line 875: Line 1,017:
 
|}
 
|}
  
The types are:
+
The possible special types are:
{|
+
 
 +
::{| class="wikitable" style="text-align: left;"
 +
! Type
 +
! Description
 +
! Notes
 +
|-
 
| -1  
 
| -1  
| = cumulative spectrum
+
| cumulative spectrum
 +
| use with energy binning ('''de''')
 
|-
 
|-
 
| -2
 
| -2
| = division by energy width
+
| division by energy width
 +
| use with energy binning ('''de''')
 
|-
 
|-
 
| -3
 
| -3
| = division by lethargy width
+
| division by lethargy width
 +
| use with energy binning ('''de''')
 
|-
 
|-
 
| -4
 
| -4
| = sum over cell or material bins
+
| sum over cell or material bins
 +
| use with cell and/or material binning ('''dc''', '''dm''')
 
|-
 
|-
 
| -5
 
| -5
| = importance weighting
+
| importance weighting
 +
| -
 
|-
 
|-
 
| -6
 
| -6
| = sum over number of scores
+
| sum over number of scores
 +
| -
 
|-
 
|-
 
| 2  
 
| 2  
| = multiply result with another detector defined by <tt>''PARAM''</tt>
+
| multiply result with another detector defined by <tt>''PARAM''</tt>
 +
| bin-compatibility
 
|-
 
|-
 
| 3
 
| 3
| = divide result with another detector defined by <tt>''PARAM''</tt>
+
| divide result with another detector defined by <tt>''PARAM''</tt>
 +
| bin-compatibility
 
|-
 
|-
 
| 4
 
| 4
| = multiply response function by (local) temperature
+
| multiply response function by (local) temperature
 +
| -
 +
|-
 
|}
 
|}
  
 
<u>Notes:</u>
 
<u>Notes:</u>
*Types -1, -2 and -3 are used with energy binning.
+
*Type "<tt>3</tt> can be used to calculate flux-weighted averages (microscopic and macroscopic cross sections, etc.).
*Type -4 can be used to calculate sum over multiple cell or material bins defined using the <tt>'''dc'''</tt> and <tt>'''dm'''</tt> options. By default separate bins are used for each entry.
+
*Type 3 can be used to calculate flux-weighted averages (microscopic and macroscopic cross sections, etc.).
+
*When the results are multiplied or divided by another detector, the number of bins must be compatible (single value or matching number of bins).
+
  
  
History collection option (<tt>'''dhis'''</tt>):<span id="det_dhis"></span>
+
Detector history collection flag (<tt>'''dhis'''</tt>):<span id="det_dhis"></span>
  
 
{|
 
{|
 
| <tt>''OPT''</tt>
 
| <tt>''OPT''</tt>
| : option to collect histories (0 = no, 1 = yes)
+
| : option to switch on (1/yes) or off (0/no) the collection of histories, batch-wise results
 
|}
 
|}
  
Line 924: Line 1,078:
  
  
Detector flagging (<tt>'''dfl'''</tt>):<span id="det_dfl"></span>
+
Detector score flagging (<tt>'''dfl'''</tt>):<span id="det_dfl"></span>
  
 
{|
 
{|
Line 932: Line 1,086:
 
| <tt>''OPT''</tt>
 
| <tt>''OPT''</tt>
 
| : flagging option (0 = reset if scored, 1 = set if scored, -2/2 score if set -3/3 score if not set)
 
| : flagging option (0 = reset if scored, 1 = set if scored, -2/2 score if set -3/3 score if not set)
 +
|}
 +
 +
The possible flagging options are:
 +
 +
::{| class="wikitable" style="text-align: left;"
 +
! Flag
 +
! Description
 +
! Notes
 +
|-
 +
| <tt>0</tt>
 +
| reset if scored
 +
| -
 +
|-
 +
| <tt>1</tt>
 +
| set if scored
 +
| -
 +
|-
 +
| <tt>-2/2</tt>
 +
| score if set
 +
| 2 (apply OR-type logic), -2 (apply AND-type logic)
 +
|-
 +
| <tt>-3/3</tt>
 +
| score if not set
 +
| 3 (apply OR-type logic), -3 (apply AND-type logic)
 +
|-
 
|}
 
|}
  
 
<u>Notes:</u>
 
<u>Notes:</u>
 
*Detector flagging allows limiting detector scores to histories which have already contributed to another score.
 
*Detector flagging allows limiting detector scores to histories which have already contributed to another score.
*The first two options reset or set the flag if the detector is scored, respectively. The remaining options test if the flag is set and score the detector accordingly. Positive values apply OR-type logic (detector is scored if any of the associated flags is set/unset) and negative values AND-type logic (detector is scored if all the associated flags are set/unset).
+
*Scoring logic:
 +
** OR-type logic: detector is scored if any of the associated flags is set/unset
 +
** AND-type logicdetector is scored if all the associated flags are set/unset
  
  
Activation detector (<tt>'''da'''</tt>):<span id="det_da"></span>
+
Detector activation (<tt>'''da'''</tt>):<span id="det_da"></span>
  
 
{|
 
{|
Line 946: Line 1,127:
 
|-
 
|-
 
| <tt>''FLX''</tt>
 
| <tt>''FLX''</tt>
| : flux applied to activation
+
| : flux applied to activation [in 1/cm<sup>2</sup>s]
 
|}
 
|}
  
 
<u>Notes:</u>
 
<u>Notes:</u>
*Activation detector allows performing activation calculation for materials that are not part of the geometry. The flux spectrum applied to neutron irradiation is taken from the detector scores. The absolute flux level can be set using the <tt>''FLX''</tt> parameter. If this parameter is set to -1, also the flux magnitude is taken from the detector scores.
+
*Activation detector allows performing activation calculation for materials that are not part of the geometry.  
*Requires neutron transport simulation and burnup mode. The material provided with the entry must be burnable, and cannot part of the actual geometry. Volume of the material must be defined using the <tt>'''dv'''</tt> parameter.
+
*Flux applied to activation:
 +
**The flux spectrum applied to neutron irradiation is taken from the detector scores.  
 +
**The absolute flux level can be set using the <tt>''FLX''</tt> parameter. If this parameter is set to "<tt>-1</tt>", also the flux magnitude is taken from the detector scores.
 +
*Requires neutron transport simulation and burnup mode. T
 +
*The detector associated material must be burnable, and cannot part of the actual geometry.  
 +
*The volume of the material, aka detector, must be defined using the <tt>'''dv'''</tt> parameter.
 
*Since the activated material is not part of the physical geometry, this option should be applied only to small samples and other activation calculations in which the isotopic changes do not significantly affect the neutronics.
 
*Since the activated material is not part of the physical geometry, this option should be applied only to small samples and other activation calculations in which the isotopic changes do not significantly affect the neutronics.
  
  
Functional Expansion Tally detector (<tt>'''dfet'''</tt>):<span id="det_dfet"></span>
+
Detector Functional Expansion Tally, FET (<tt>'''dfet'''</tt>):<span id="det_dfet"></span>
  
 
{|
 
{|
Line 965: Line 1,151:
 
  |}
 
  |}
  
{| class="wikitable" style="text-align: left;"
+
::{| class="wikitable" style="text-align: left;"
 
  ! Geometry
 
  ! Geometry
 
  ! <tt>PARAMS</tt>
 
  ! <tt>PARAMS</tt>
Line 989: Line 1,175:
  
 
<u>Notes:</u>
 
<u>Notes:</u>
*"-1" can be supplied as an <tt>''ORDER''</tt> <tt>PARAM</tt> to use the built-in default values
+
*"<tt>-1</tt>" can be supplied as an <tt>''ORDER''</tt> <tt>PARAM</tt> to use the built-in default values
 
*It is not recommended to configure a single FET detector to span multiple different material regions—use individual detectors for each region instead
 
*It is not recommended to configure a single FET detector to span multiple different material regions—use individual detectors for each region instead
 
*Specifics of this implementation:
 
*Specifics of this implementation:
Line 1,000: Line 1,186:
 
*From version 2.2.0  and on, FET-based detectors follow the standard normalization set in the calculation. The volume standards for detectors are set as default value for FET-based detectors, meaning detectors are not divided by the physical volume (allowing the use of volume detector '''dv''').
 
*From version 2.2.0  and on, FET-based detectors follow the standard normalization set in the calculation. The volume standards for detectors are set as default value for FET-based detectors, meaning detectors are not divided by the physical volume (allowing the use of volume detector '''dv''').
 
*In version 2.2.0, the relative error evaluation associated with FET-based detectors has been revisited.
 
*In version 2.2.0, the relative error evaluation associated with FET-based detectors has been revisited.
 +
  
 
Detector pulse-height energy broadening (<tt>'''dphb'''</tt>):<span id="det_dphb"></span>
 
Detector pulse-height energy broadening (<tt>'''dphb'''</tt>):<span id="det_dphb"></span>
Line 1,024: Line 1,211:
 
=== div (divisor definition)<span id="div"></span> ===
 
=== div (divisor definition)<span id="div"></span> ===
  
  '''div''' ''MAT'' [ '''sep''' ''LVL'' ]  
+
  '''div''' ''MAT'' [ [[#div_sep|'''sep''']] ''LVL'' ]  
         [ '''subx''' ''N<sub>X</sub>'' ''X<sub>MIN</sub>'' ''X<sub>MAX</sub>'' ] <small>equal volume</small> [ '''subr''' ''-N<sub>X</sub>'' ''X<sub>1</sub>'' ''R<sub>2</sub>'' ... ''X<sub>N+1</sub>'' ] <small>manually spaced limits</small>
+
         [ [[#div_subx1|'''subx''']] ''N<sub>X</sub>'' ''X<sub>MIN</sub>'' ''X<sub>MAX</sub>'' ]
         [ '''suby''' ''N<sub>Y</sub>'' ''Y<sub>MIN</sub>'' ''Y<sub>MAX</sub>'' ] <small>equal volume</small> [ '''suby''' ''-N<sub>Y</sub>'' ''Y<sub>1</sub>'' ''Y<sub>2</sub>'' ... ''Y<sub>N+1</sub>'' ] <small>manually spaced limits</small>
+
        [ [[#div_subx2|'''subx''']] ''N<sub>X</sub>'' ''X<sub>1</sub>'' ''X<sub>2</sub>'' ... ''X<sub>N+1</sub>'' ]  
         [ '''subz''' ''N<sub>Z</sub>'' ''Z<sub>MIN</sub>'' ''Z<sub>MAX</sub>'' ] <small>equal volume</small> [ '''subz''' ''-N<sub>Z</sub>'' ''Z<sub>1</sub>'' ''Z<sub>2</sub>'' ... ''Z<sub>N+1</sub>'' ] <small>manually spaced limits</small>
+
         [ [[#div_suby1|'''suby''']] ''N<sub>Y</sub>'' ''Y<sub>MIN</sub>'' ''Y<sub>MAX</sub>'' ]  
         [ '''subr''' ''N<sub>R</sub>'' ''R<sub>MIN</sub>'' ''R<sub>MAX</sub>'' ] <small>equal volume</small> [ '''subr''' ''-N<sub>R</sub>'' ''R<sub>1</sub>'' ''R<sub>2</sub>'' ... ''R<sub>N+1</sub>'' ] <small>manually spaced limits</small>
+
        [ [[#div_suby2|'''suby''']] ''N<sub>Y</sub>'' ''Y<sub>1</sub>'' ''Y<sub>2</sub>'' ... ''Y<sub>N+1</sub>'' ]  
         [ '''subs''' ''N<sub>S</sub>'' ''S<sub>0</sub>'' ]      <small>equal volume</small> [ '''subs''' ''-N<sub>S</sub>'' ''S<sub>1</sub>'' ''S<sub>2</sub>'' ... ''S<sub>N+1</sub>'' ] <small>manually spaced limits</small>
+
         [ [[#div_subz1|'''subz''']] ''N<sub>Z</sub>'' ''Z<sub>MIN</sub>'' ''Z<sub>MAX</sub>'' ]  
         [ '''lims''' ''FLAG'' ]
+
        [ [[#div_subz2|'''subz''']] ''N<sub>Z</sub>'' ''Z<sub>1</sub>'' ''Z<sub>2</sub>'' ... ''Z<sub>N+1</sub>'' ]  
Divides a material into a number of sub-zones. Input values:
+
         [ [[#div_subr1|'''subr''']] ''N<sub>R</sub>'' ''R<sub>MIN</sub>'' ''R<sub>MAX</sub>'' ]  
 +
        [ [[#div_subr2|'''subr''']] ''N<sub>R</sub>'' ''R<sub>1</sub>'' ''R<sub>2</sub>'' ... ''R<sub>N+1</sub>'' ]  
 +
         [ [[#div_subs1|'''subs''']] ''N<sub>S</sub>'' ''S<sub>0</sub>'' ]       
 +
        [ [[#div_subs2|'''subs''']] ''N<sub>S</sub>'' ''S<sub>1</sub>'' ''S<sub>2</sub>'' ... ''S<sub>N+1</sub>'' ]
 +
        [ [[#div_peb|'''peb''']] ''PBED'' ''N<sub>UNI</sub>'' [ ''UNI<sub>1</sub>'' ... ''UNI<sub>N</sub>'' ]  ]
 +
         [ [[#div_lims|'''lims''']] ''FLAG'' ]
 +
Divides a material into a number of sub-zones. The first parameter:
  
 
{|
 
{|
 
| <tt>''MAT''</tt>
 
| <tt>''MAT''</tt>
 
| : name of the divided material
 
| : name of the divided material
|-
+
|}
 +
 
 +
The remaining parameters are defined by separate key words followed by the input values.
 +
 
 +
<u>Notes:</u>
 +
*A single div card may include one or several sub-divisions.
 +
*As general rule:
 +
** if the number of zones associated with a sub-division is <u>positive</u>, the sub-division is <u>equal volume</u> (see below)
 +
** if the number of zones associated with a sub-division is <u>negative</u>, the subdivision is <u>user-defined volume</u> (see below)
 +
*If a material is not divided, all occurrences of it are treated as a single depletion zone (except for depleted materials defined in pin structures: pin-type division).
 +
*The use of automated instead of manual depletion zone division saves memory, which may become significant in very large burnup calculation problems (see detailed description on [[memory usage#automated depletion zone division|memory usage]]).
 +
*The volumes of the divided materials must be set manually (see [[#set mvol|set mvol]] option) or automatically, via the Monte Carlo checker-routine (see [[#set mcvol|set mcvol]] option or [[Installing and running Serpent#Running Serpent|''-checkvolumes'']] command line option). For a more detailed description, check [[Defining material volumes|Defining material volumes]]).
 +
*For more information, see detailed description on [[automated depletion zone division]].
 +
 
 +
 
 +
<u>Sub-division types:</u>
 +
 
 +
Sub-division geometry level (<tt>'''sep'''</tt>): <span id="div_sep"></span>
 +
 
 +
{|
 
| <tt>''LVL''</tt>  
 
| <tt>''LVL''</tt>  
| : geometry level at which the cell-wise division takes place (0 = no division, 1 = last level, 2 = 2nd last level, etc.)  
+
| : geometry level at which the material-wise division takes place (0 = no division, 1 = last level, 2 = 2nd last level, etc.)  
 
|-
 
|-
 +
|}
 +
 +
<u>Notes:</u>
 +
*The sub-division criterion is the geometry level.
 +
*The level number is counted backwards from the last one, i.e. level "<tt>1</tt>" is the last level.
 +
*Use examples:
 +
**to sub-divide the fuel in large LWR core into separate depletion zones on assembly-, instead of pin-basis (default division if a lattice structure is defined with no div card associated to the material)
 +
**to sub-divide HTGR fuel kernels into depletion zones on compact- or pebble-basis.
 +
 +
 +
Sub-division Cartesian mesh, <u>equal volume</u> (<tt>'''subx'''</tt>, <tt>'''suby'''</tt> and <tt>'''subz'''</tt>):<span id="div_subx1"></span><span id="div_suby1"></span><span id="div_subz1"></span>
 +
 +
{|
 
| <tt>''N<sub>X</sub>''</tt>
 
| <tt>''N<sub>X</sub>''</tt>
| : number of x-zones
+
| : number of x-zones (positive value)
 
|-
 
|-
 
| <tt>''X<sub>MIN</sub>''</tt>  
 
| <tt>''X<sub>MIN</sub>''</tt>  
| : minimum x-coordinate (cm)
+
| : minimum x-coordinate [in cm]
 
|-
 
|-
 
| <tt>''X<sub>MAX</sub>''</tt>  
 
| <tt>''X<sub>MAX</sub>''</tt>  
| : maximum x-coordinate (cm)
+
| : maximum x-coordinate [in cm]
|-
+
| <tt>''X<sub>n</sub>''</tt>
+
| : x-coordinate boundaries (cm)
+
 
|-
 
|-
 
| <tt>''N<sub>Y</sub>''</tt>
 
| <tt>''N<sub>Y</sub>''</tt>
| : number of y-zones
+
| : number of y-zones (positive value)
 
|-
 
|-
 
| <tt>''Y<sub>MIN</sub>''</tt>  
 
| <tt>''Y<sub>MIN</sub>''</tt>  
| : minimum y-coordinate (cm)
+
| : minimum y-coordinate [in cm]
 
|-
 
|-
 
| <tt>''Y<sub>MAX</sub>''</tt>  
 
| <tt>''Y<sub>MAX</sub>''</tt>  
| : maximum y-coordinate (cm)
+
| : maximum y-coordinate [in cm]
|-
+
| <tt>''Y<sub>n</sub>''</tt>
+
| : y-coordinate boundaries (cm)
+
 
|-
 
|-
 
| <tt>''N<sub>Z</sub>''</tt>
 
| <tt>''N<sub>Z</sub>''</tt>
| : number of z-zones
+
| : number of z-zones (positive value)
 
|-
 
|-
 
| <tt>''Z<sub>MIN</sub>''</tt>  
 
| <tt>''Z<sub>MIN</sub>''</tt>  
| : minimum z-coordinate (cm)
+
| : minimum z-coordinate [in cm]
 
|-
 
|-
 
| <tt>''Z<sub>MAX</sub>''</tt>  
 
| <tt>''Z<sub>MAX</sub>''</tt>  
| : maximum z-coordinate (cm)
+
| : maximum z-coordinate [in cm]
 +
|-
 +
|}
 +
 
 +
<u>Notes:</u>
 +
* An equal volume sub-division is performed in the given dimension.
 +
* The value of the parameter <tt>''N<sub>n</sub>''</tt> which defines the number of zones in the given dimension must be positive.
 +
* For a Cartesian mesh sub-division, a separate entry in x-, y-, z- directions is provided (<tt>'''subx'''</tt>, <tt>'''suby'''</tt> and <tt>'''subz'''</tt>, respectively).
 +
 
 +
 
 +
Sub-division Cartesian mesh, <u>user-defined volume</u> (<tt>'''subx'''</tt>, <tt>'''suby'''</tt> and <tt>'''subz'''</tt>):<span id="div_subx2"></span><span id="div_suby2"></span><span id="div_subz2"></span>
 +
 
 +
{|
 +
| <tt>''N<sub>X</sub>''</tt>
 +
| : number of x-zones (negative value)
 +
|-
 +
| <tt>''X<sub>n</sub>''</tt>
 +
| : x-coordinate boundaries [in cm]
 +
|-
 +
| <tt>''N<sub>Y</sub>''</tt>
 +
| : number of y-zones (negative value)
 +
|-
 +
| <tt>''Y<sub>n</sub>''</tt>
 +
| : y-coordinate boundaries [in cm]
 +
|-
 +
| <tt>''N<sub>Z</sub>''</tt>
 +
| : number of z-zones (negative value)
 
|-
 
|-
 
| <tt>''Z<sub>n</sub>''</tt>  
 
| <tt>''Z<sub>n</sub>''</tt>  
| : z-coordinate boundaries (cm)
+
| : z-coordinate boundaries [in cm]
 
|-
 
|-
 +
|}
 +
 +
<u>Notes:</u>
 +
* An user-defined volume sub-division is performed in the given dimension.
 +
* The value of the parameter <tt>''N<sub>n</sub>''</tt> which defines the number of zones in the given dimension must be negative.
 +
* For a Cartesian mesh sub-division, a separate entry in x-, y-, z- directions is provided (<tt>'''subx'''</tt>, <tt>'''suby'''</tt> and <tt>'''subz'''</tt>, respectively).
 +
 +
 +
Sub-division cylindrical annular mesh,  <u>equal volume</u> (<tt>'''subr'''</tt>):<span id="div_subr1"></span>
 +
 +
{|
 
| <tt>''N<sub>R</sub>''</tt>
 
| <tt>''N<sub>R</sub>''</tt>
| : number of radial zones
+
| : number of radial-zones (positive value)
 
|-
 
|-
 
| <tt>''R<sub>MIN</sub>''</tt>  
 
| <tt>''R<sub>MIN</sub>''</tt>  
| : minimum radial coordinate (cm)
+
| : minimum radial-coordinate [in cm]
 
|-
 
|-
 
| <tt>''R<sub>MAX</sub>''</tt>  
 
| <tt>''R<sub>MAX</sub>''</tt>  
| : maximum radial coordinate (cm)
+
| : maximum radial-coordinate [in cm]
 +
|-
 +
|}
 +
 
 +
<u>Notes:</u>
 +
* An equal volume radial sub-division is performed (annular-type sub-division)
 +
* The value of the parameter <tt>''N<sub>R</sub>''</tt> which defines the number of zones in the given dimension must be positive.
 +
 
 +
 
 +
Sub-division cylindrical annular mesh, <u>user-defined volume</u> (<tt>'''subs'''</tt>):<span id="div_subr2"></span><span id="div_suby2"></span><span id="div_subz2"></span>
 +
 
 +
{|
 +
| <tt>''N<sub>R</sub>''</tt>
 +
| : number of radial-zones (negative value)
 
|-
 
|-
 
| <tt>''R<sub>n</sub>''</tt>  
 
| <tt>''R<sub>n</sub>''</tt>  
| : radial coordinate boundaries (cm)
+
| : radial-coordinate boundaries [in cm]
 
|-
 
|-
 +
|}
 +
 +
<u>Notes:</u>
 +
* An user-defined volume radial sub-division is performed (annular-type sub-division)
 +
* The value of the parameter <tt>''N<sub>R</sub>''</tt> which defines the number of zones in the given dimension must be negative.
 +
 +
 +
Sub-division cylindrical sector mesh,  <u>equal volume</u> (<tt>'''subr'''</tt>):<span id="div_subs1"></span>
 +
 +
{|
 
| <tt>''N<sub>S</sub>''</tt>
 
| <tt>''N<sub>S</sub>''</tt>
| : number of angular sectors
+
| : number of angular-zones (negative value)
 
|-
 
|-
 
| <tt>''S<sub>0</sub>''</tt>  
 
| <tt>''S<sub>0</sub>''</tt>  
| : zero position of angular division (degrees)
+
| : zero position of angular division [in degrees]
 +
|-
 +
|}
 +
 
 +
<u>Notes:</u>
 +
* An equal volume angular sub-division is performed (sector-type sub-division)
 +
* The value of the parameter <tt>''N<sub>S</sub>''</tt> which defines the number of zones in the angular dimension must be positive.
 +
 
 +
 
 +
Sub-division cylindrical sector mesh, <u>user-defined volume</u> (<tt>'''subs'''</tt>):<span id="div_subs2"></span><span id="div_suby2"></span><span id="div_subz2"></span>
 +
 
 +
{|
 +
| <tt>''N<sub>S</sub>''</tt>
 +
| : number of angular-zones
 
|-
 
|-
 
| <tt>''S<sub>n</sub>''</tt>  
 
| <tt>''S<sub>n</sub>''</tt>  
| : angular-sector boundaries (degrees)
+
| : angular-sector boundaries [in degrees]
 
|-
 
|-
| <tt>''FLAG''</tt>
 
| : flag for mapping regions outside (material) limits to divide material - on (1/yes) or off (0/no)
 
 
|}
 
|}
  
 
<u>Notes:</u>
 
<u>Notes:</u>
 +
* An user-defined volume angular sub-division is performed (sector-type sub-division)
 +
* The value of the parameter <tt>''N<sub>S</sub>''</tt> which defines the number of zones in the angular dimension must be negative.
 +
* The manually-spaced angular-sector boundaries <tt>''S<sub>n</sub>''</tt> must cover the full/360 degrees angular space.
  
*The automated divisor feature can be used to sub-divide burnable materials into depletion zones, but the use is not limited to burnup mode.
+
 
*The spatial sub-division is based on either Cartesian or cylindrical mesh.
+
Sub-division pebble-bed structure (<tt>'''peb'''</tt>): <span id="div_peb"></span>
*Volumes of the divided materials must be set manually (see detailed description on [[Defining material volumes|the definition of material volumes]]).
+
 
*Using automated instead of manual depletion zone division saves memory, which may become significant in very large burnup calculation problems (see detailed description on [[memory usage#automated depletion zone division|memory usage]]).
+
{|
*For more information see detailed description on [[automated depletion zone division]].
+
| <tt>''PBED''</tt>
*The usage of <tt>''LVL''</tt> is explained on page [[automated depletion zone division]].
+
| : stochastic particle / pebble-bed structure
*The feature of mapping regions outside limits is set by default OFF.
+
|-
*The manually-spaced angular-sector boundaries <tt>''S<sub>n</sub>''</tt> should cover the full/360 degrees angular space.
+
| <tt>''N<sub>UNI</sub>''</tt>
*If a material is not divided, all occurrences of it are treated as a single depletion zone. For example, if there are multiple fuel pins with same fuel material type, and no ''div card'' is present, all pins are depleted as a single pin.
+
| : number of universes to link related to the <tt>''PBED''</tt> structure (special case: 0 = link to all)
 +
|-
 +
| <tt>''UNI<sub>N</sub>''</tt>
 +
|: list of universes to link (non-zero number of universes)
 +
|-
 +
|}
 +
 
 +
<u>Notes:</u>
 +
*The pebble bed-based sub-division divides each item in a pebble bed universe as its own item.
 +
*It features a speed-up on the depletion zone division indexing process with large number of pebble bed structures.
 +
 
 +
 
 +
Sub-division limit enforcement (<tt>'''lims'''</tt>): <span id="div_lims"></span>
 +
 
 +
{|
 +
| <tt>''FLAG''</tt>
 +
| : flag for mapping regions outside (material) limits to divide material: on (1/yes) or off (0/no). The default option is "<tt>off</tt>"
 +
|-
 +
|}
  
 
=== dtrans (detector mesh transformation)<span id="dtrans"></span> ===
 
=== dtrans (detector mesh transformation)<span id="dtrans"></span> ===
  
See [[#trans (transformations)|transformations]].
+
Defines detector mesh transformations. Shortcut for "trans d".
 +
 
 +
<u>Notes:</u>
 +
*The parameters associated with the transformation follow the standard transformation cards syntax without '''trans''' <tt>''TYPE''</tt> identifier.
 +
*See [[#trans (transformations)|transformations]].
  
 
=== ene (energy grid definition)<span id="ene"></span> ===
 
=== ene (energy grid definition)<span id="ene"></span> ===
Line 1,134: Line 1,449:
 
|-
 
|-
 
|<tt>''E<sub>i</sub>''</tt>
 
|<tt>''E<sub>i</sub>''</tt>
|: bin boundaries (type 1)
+
|: bin boundaries [in MeV]
 
|-
 
|-
 
|<tt>''N''</tt>
 
|<tt>''N''</tt>
|: number of equi-width bins (types 2 and 3)
+
|: number of equi-width bins
 
|-
 
|-
 
|<tt>''E<sub>min</sub>''</tt>
 
|<tt>''E<sub>min</sub>''</tt>
|: minimum energy (types 2 and 3)
+
|: minimum energy [in MeV]
 
|-
 
|-
 
|<tt>''E<sub>max</sub>''</tt>
 
|<tt>''E<sub>max</sub>''</tt>
|: maximum energy (types 2 and 3)
+
|: maximum energy [in MeV]
 
|-
 
|-
 
|<tt>''GRID''</tt>
 
|<tt>''GRID''</tt>
|: name of the pre-defined grid (type 4)
+
|: name of the pre-defined grid
 
|}
 
|}
  
Line 1,153: Line 1,468:
 
*The first input parameter gives the type (1 = arbitrarily defined, 2 = equal energy-width bins, 3 = equal lethargy-width bins, 4 = [[pre-defined energy group structures|pre-defined energy group structure]]).
 
*The first input parameter gives the type (1 = arbitrarily defined, 2 = equal energy-width bins, 3 = equal lethargy-width bins, 4 = [[pre-defined energy group structures|pre-defined energy group structure]]).
 
*Energy grid structures are used for several purposes, most commonly with [[#det_de|detectors]].
 
*Energy grid structures are used for several purposes, most commonly with [[#det_de|detectors]].
*See separate description of [[pre-defined energy group structures]].
 
  
 
=== ftrans (fill transformation)<span id="ftrans"></span> ===
 
=== ftrans (fill transformation)<span id="ftrans"></span> ===
  
See [[#trans (transformations)|transformations]].
+
Defines fill transformations. Shortcut for "<tt>trans f</tt>".
 +
 
 +
<u>Notes:</u>
 +
*The parameters associated with the transformation follow the standard transformation cards syntax without '''trans''' <tt>''TYPE''</tt> identifier.
 +
*See [[#trans (transformations)|transformations]].
  
 
=== fun (function definition)<span id="fun"></span> ===
 
=== fun (function definition)<span id="fun"></span> ===
Line 1,187: Line 1,505:
  
 
<u>Notes:</u>
 
<u>Notes:</u>
* The defined function is linked to detector response using [[ENDF reaction MT's and macroscopic reaction numbers|response number]] -100 (syntax: dr -100 ''NAME'').
+
* The defined function is linked to detector response using [[ENDF reaction MT's and macroscopic reaction numbers|MT -100 ]] (syntax: dr -100 ''NAME'').
 
* The defined function currently is only supported as a flux-based function, aka, flux multiplier.
 
* The defined function currently is only supported as a flux-based function, aka, flux multiplier.
  
Line 1,199: Line 1,517:
 
{|
 
{|
 
| <tt>''BU<sub>n</sub>''</tt>  
 
| <tt>''BU<sub>n</sub>''</tt>  
| : burnup steps at which the branches are invoked
+
| : burnup steps at which the branches are invoked (positive value = burnup [in MWd/kg], negative value = time [in d])
 
|-
 
|-
 
| <tt>''NBR<sub>n</sub>''</tt>
 
| <tt>''NBR<sub>n</sub>''</tt>
Line 1,211: Line 1,529:
 
*The automated burnup sequence defined by the hisv card follows the same principle as the [[#coef|coef]] input card.
 
*The automated burnup sequence defined by the hisv card follows the same principle as the [[#coef|coef]] input card.
 
*The hisv card performs multiple depletions within a single depletion calculation following the historical variation sequence, performing a restart at each of the listed burnup points, where it applies the variations defined in the listed branches for the given burnup point.
 
*The hisv card performs multiple depletions within a single depletion calculation following the historical variation sequence, performing a restart at each of the listed burnup points, where it applies the variations defined in the listed branches for the given burnup point.
*Positive values in the burnup vector are interpreted as (MWd/kgU), negative values are interpreted as time steps in days.
 
 
*The hisv card is used together with the [[#branch|branch]] card.
 
*The hisv card is used together with the [[#branch|branch]] card.
  
 
=== ifc (interface file)<span id="ifc"></span> ===
 
=== ifc (interface file)<span id="ifc"></span> ===
  
  '''ifc''' ''FILE'' ['''setinmat''' ''N<sub>MAT</sub>'' ''MAT<sub>1</sub>'' ''MAT<sub>2</sub>'' ... ''MAT<sub>N<sub>MAT</sub></sub>'' ]
+
  '''ifc''' ''FILE'' [ [[#ifc_setinmat|'''setinmat''']] ''N<sub>MAT</sub>'' ''MAT<sub>1</sub>'' ''MAT<sub>2</sub>'' ... ''MAT<sub>N<sub>MAT</sub></sub>'' ]
           ['''setoutmat''' ''N<sub>MAT</sub>'' ''MAT<sub>1</sub>'' ''MAT<sub>2</sub>'' ... ''MAT<sub>N<sub>MAT</sub></sub>'' ]
+
           [ [[#ifc_setoutmat|'''setoutmat''']] ''N<sub>MAT</sub>'' ''MAT<sub>1</sub>'' ''MAT<sub>2</sub>'' ... ''MAT<sub>N<sub>MAT</sub></sub>'' ]
  
Links a [[Multi-physics interface|multi-physics interface]] file to be used with the current input. Input values:
+
Links a [[Multi-physics interface|multi-physics interface]] file to be used with the current input. The first parameter:
  
 
{|
 
{|
Line 1,226: Line 1,543:
 
|}
 
|}
  
The optional cards are explained below.
+
The remaining parameters are defined by separate key words followed by the input values, being optional.
  
'''<tt>setinmat</tt>''' adds the possibility to link multiple input materials to the same interface, i.e. the same interface gives temperatures and densities (density factors) for multiple materials.
+
<u>Notes:</u>
 +
*See also [[Coupled multi-physics calculations]].
  
'''<tt>setoutmat</tt>''' adds the possibility to link multiple output materials to the same interface, i.e. the same interface gives temperatures and densities (density factors) for multiple materials.
+
 
 +
<u>Optional entries:</u>
 +
 
 +
Interface input materials  (<tt>'''setinmat'''</tt>): <span id="mix_setinmat"></span>
  
 
{|
 
{|
|<tt>''N<sub>MAT</sub>''</tt>
+
| <tt>''N<sub>MAT''</sub></tt>
|: number of materials to link to the interface
+
| : number of input materials to link to the interface
 
|-
 
|-
|<tt>''MAT<sub>i</sub>''</tt>
+
| <tt>''MAT<sub>n</sub>''</tt>  
|: name of the ''i''th material linked to the interface
+
| : name of the ''n''-th input material linked to the interface
 +
|}
 +
 
 +
<u>Notes:</u>
 +
*It adds the possibility to link multiple input materials to the same interface, i.e. the same interface gives temperatures and densities (density factors) for multiple materials.
 +
*If multiple input materials are linked to the interface using the option, the densities in the interface file must be given as density factors, i.e. relative to the material card density (values between 0 and 1).
 +
*If the interface is not updated, the entry is not eligible.
 +
*If the regular mesh-based interface is used and power is tallied in pin-type objects, the entry is not eligible.
 +
*The option '''<tt>setinmat</tt>''' is referred as '''<tt>setmat</tt>''' up to version 2.1.31.
 +
 
 +
 
 +
Interface output materials  (<tt>'''setoutmat'''</tt>): <span id="mix_setoutmat"></span>
 +
 
 +
{|
 +
| <tt>''N<sub>MAT''</sub></tt>
 +
| : number of output materials to link to the interface
 
|-
 
|-
 +
| <tt>''MAT<sub>n</sub>''</tt>
 +
| : name of the ''n''-th output material linked to the interface
 
|}
 
|}
  
 
<u>Notes:</u>
 
<u>Notes:</u>
* If multiple materials are linked to the interface using the '''<tt>setinmat</tt>'''/'''<tt>setoutmat</tt>''' option, the densities in the interface file must be given as density factors, i.e. relative to the material card density (values between 0 and 1).
+
*It adds the possibility to link multiple output materials to the same interface, i.e. the same interface gives temperatures and densities (density factors) for multiple materials.
* If the interface is not updated, '''<tt>setinmat</tt>'''/'''<tt>setoutmat</tt>''' options are not eligible. In the case of regular mesh-based, additionally to not updating the interface: a) the input materials cannot be specified using '''<tt>setinmat</tt>''' if power is tallied in pin-type objects; b) the output materials cannot be specified using '''<tt>setoutmat</tt>''' if power is not tallied on the same mesh.
+
*If multiple input materials are linked to the interface using the option, the densities in the interface file must be given as density factors, i.e. relative to the material card density (values between 0 and 1).
* Option '''<tt>setinmat</tt>''' was '''<tt>setmat</tt>''' in versions before 2.1.32.
+
* If the interface is not updated, the entry is not eligible.  
* See also [[Coupled multi-physics calculations]].
+
*If the regular mesh-based interface is used and if power is not tallied on the same mesh, the entry is not eligible.
  
 
=== include (read another input file)<span id="include"></span> ===
 
=== include (read another input file)<span id="include"></span> ===
Line 1,279: Line 1,617:
 
|-
 
|-
 
| <tt>''X<sub>0</sub>''</tt>
 
| <tt>''X<sub>0</sub>''</tt>
| : x-coordinate of the lattice origin (origin is in the center of the lattice).
+
| : x-coordinate of the lattice origin (origin is in the center of the lattice) [in cm].
 
|-
 
|-
 
| <tt>''Y<sub>0</sub>''</tt>
 
| <tt>''Y<sub>0</sub>''</tt>
| : y-coordinate of the lattice origin (origin is in the center of the lattice).
+
| : y-coordinate of the lattice origin (origin is in the center of the lattice) [in cm].
 
|-
 
|-
 
| <tt>''N<sub>X</sub>''</tt>
 
| <tt>''N<sub>X</sub>''</tt>
Line 1,291: Line 1,629:
 
|-
 
|-
 
| <tt>''PITCH''</tt>
 
| <tt>''PITCH''</tt>
| : lattice pitch
+
| : lattice pitch [in cm]
 
|-
 
|-
 
| <tt>''UNI<sub>n</sub>''</tt>
 
| <tt>''UNI<sub>n</sub>''</tt>
Line 1,298: Line 1,636:
  
 
Possible lattice types are:
 
Possible lattice types are:
{| class="wikitable" style="text-align: left;"
+
::{| class="wikitable" style="text-align: left;"
 
! Type
 
! Type
 
! Description
 
! Description
Line 1,337: Line 1,675:
 
|-
 
|-
 
| <tt>''X<sub>0</sub>''</tt>
 
| <tt>''X<sub>0</sub>''</tt>
| : x-coordinate of the lattice origin
+
| : x-coordinate of the lattice origin [in cm]
 
|-
 
|-
 
| <tt>''Y<sub>0</sub>''</tt>
 
| <tt>''Y<sub>0</sub>''</tt>
| : y-coordinate of the lattice origin
+
| : y-coordinate of the lattice origin [in cm]
 
|-
 
|-
 
| <tt>''PITCH''</tt>
 
| <tt>''PITCH''</tt>
| : lattice pitch
+
| : lattice pitch [in cm]
 
|-
 
|-
 
| <tt>''UNI<sub>1</sub>''</tt>
 
| <tt>''UNI<sub>1</sub>''</tt>
Line 1,350: Line 1,688:
  
 
Possible lattice types are:
 
Possible lattice types are:
{| class="wikitable" style="text-align: left;"
+
::{| class="wikitable" style="text-align: left;"
 
! Type
 
! Type
 
! Description
 
! Description
Line 1,379: Line 1,717:
 
|-
 
|-
 
| <tt>''X<sub>0</sub>''</tt>
 
| <tt>''X<sub>0</sub>''</tt>
| : x-coordinate of the lattice origin
+
| : x-coordinate of the lattice origin [in cm]
 
|-
 
|-
 
| <tt>''Y<sub>0</sub>''</tt>
 
| <tt>''Y<sub>0</sub>''</tt>
| : y-coordinate of the lattice origin
+
| : y-coordinate of the lattice origin [in cm]
 
|-
 
|-
 
| <tt>''N<sub>R</sub>''</tt>
 
| <tt>''N<sub>R</sub>''</tt>
Line 1,388: Line 1,726:
 
|-
 
|-
 
| <tt>''N<sub>S,R</sub>''</tt>
 
| <tt>''N<sub>S,R</sub>''</tt>
| : number of sectors in Rth ring
+
| : number of sectors in ''R''-th ring
 
|-
 
|-
 
| <tt>''RADIUS<sub>R</sub>''</tt>
 
| <tt>''RADIUS<sub>R</sub>''</tt>
| : central radius of Rth ring
+
| : central radius of ''R''-th ring [in cm]
 
|-
 
|-
 
| <tt>''THETA<sub>R</sub>''</tt>
 
| <tt>''THETA<sub>R</sub>''</tt>
| : angle of rotation of Rth ring in degrees
+
| : angle of rotation of ''R''-th ring [in degrees]
 
|-
 
|-
 
| <tt>''UNI<sub>N,R</sub>''</tt>
 
| <tt>''UNI<sub>N,R</sub>''</tt>
| : list of universes filling the sector positions in Rth ring
+
| : list of universes filling the sector positions in ''R''-th ring
 
|}
 
|}
  
 
Possible lattice type is:
 
Possible lattice type is:
{| class="wikitable" style="text-align: left;"
+
::{| class="wikitable" style="text-align: left;"
 
! Type
 
! Type
 
! Description
 
! Description
Line 1,424: Line 1,762:
 
|-
 
|-
 
| <tt>''X<sub>0</sub>''</tt>
 
| <tt>''X<sub>0</sub>''</tt>
| : x-coordinate of the lattice origin
+
| : x-coordinate of the lattice origin [in cm]
 
|-
 
|-
 
| <tt>''Y<sub>0</sub>''</tt>
 
| <tt>''Y<sub>0</sub>''</tt>
| : y-coordinate of the lattice origin
+
| : y-coordinate of the lattice origin [in cm]
 
|-
 
|-
 
| <tt>''N<sub>L</sub>''</tt>
 
| <tt>''N<sub>L</sub>''</tt>
Line 1,433: Line 1,771:
 
|-
 
|-
 
| <tt>''Z<sub>n</sub>''</tt>
 
| <tt>''Z<sub>n</sub>''</tt>
| : z-coordinate of the nth lattice element (lower boundary of the axial layer)
+
| : z-coordinate of the ''n''-th lattice element (lower boundary of the axial layer) [in cm]
 
|-
 
|-
 
| <tt>''UNI<sub>n</sub>''</tt>
 
| <tt>''UNI<sub>n</sub>''</tt>
| : universe name filling the nth lattice position
+
| : universe name filling the ''n''-th lattice position
 
|}
 
|}
  
 
Possible lattice type is:
 
Possible lattice type is:
{| class="wikitable" style="text-align: left;"
+
::{| class="wikitable" style="text-align: left;"
 
! Type
 
! Type
 
! Description
 
! Description
Line 1,466: Line 1,804:
 
|-
 
|-
 
| <tt>''X<sub>0</sub>''</tt>
 
| <tt>''X<sub>0</sub>''</tt>
| : x-coordinate of the lattice origin
+
| : x-coordinate of the lattice origin [in cm]
 
|-
 
|-
 
| <tt>''Y<sub>0</sub>''</tt>
 
| <tt>''Y<sub>0</sub>''</tt>
| : y-coordinate of the lattice origin
+
| : y-coordinate of the lattice origin [in cm]
 
|-
 
|-
 
| <tt>''Z<sub>0</sub>''</tt>
 
| <tt>''Z<sub>0</sub>''</tt>
| : z-coordinate of the lattice origin
+
| : z-coordinate of the lattice origin [in cm]
 
|-
 
|-
 
| <tt>''N<sub>X</sub>''</tt>
 
| <tt>''N<sub>X</sub>''</tt>
Line 1,484: Line 1,822:
 
|-
 
|-
 
| <tt>''PITCH<sub>X</sub>''</tt>
 
| <tt>''PITCH<sub>X</sub>''</tt>
| : lattice pitch in x-direction
+
| : lattice pitch in x-direction [in cm]
 
|-
 
|-
 
| <tt>''PITCH<sub>Y</sub>''</tt>
 
| <tt>''PITCH<sub>Y</sub>''</tt>
| : lattice pitch in y-direction
+
| : lattice pitch in y-direction [in cm]
 
|-
 
|-
 
| <tt>''PITCH<sub>Z</sub>''</tt>
 
| <tt>''PITCH<sub>Z</sub>''</tt>
| : lattice pitch in z-direction
+
| : lattice pitch in z-direction [in cm]
 
|-
 
|-
 
| <tt>''UNI<sub>n</sub>''</tt>
 
| <tt>''UNI<sub>n</sub>''</tt>
Line 1,497: Line 1,835:
  
 
Possible lattice types are:
 
Possible lattice types are:
{| class="wikitable" style="text-align: left;"
+
::{| class="wikitable" style="text-align: left;"
 
! Type
 
! Type
 
! Description
 
! Description
Line 1,518: Line 1,856:
 
=== ltrans (lattice transformation)<span id="ltrans"></span> ===
 
=== ltrans (lattice transformation)<span id="ltrans"></span> ===
  
See [[#trans (transformations)|transformations]].
+
Defines lattice transformations. Shortcut for "<tt>trans l</tt>".
 +
 
 +
<u>Notes:</u>
 +
*The parameters associated with the transformation follow the standard transformation cards syntax without '''trans''' <tt>''TYPE''</tt> identifier.
 +
*See [[#trans (transformations)|transformations]].
  
 
=== mat (material definition)<span id="mat"></span> ===
 
=== mat (material definition)<span id="mat"></span> ===
Line 1,537: Line 1,879:
 
  [    ''...''    ]
 
  [    ''...''    ]
  
<u>Mandatory information:</u>
+
Material definition. The mandatory parameters are:
  
 
{|
 
{|
 
| <tt>''NAME''</tt>
 
| <tt>''NAME''</tt>
| : Name of the material
+
| : name of the material
 
|-
 
|-
 
| <tt>''DENS''</tt>
 
| <tt>''DENS''</tt>
| : Density of the material (positive for atomic, negative for mass density) or '''<tt>sum</tt>''' to calculate the density from given nuclide fractions
+
| : 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>]), or "<tt>sum</tt>" to calculate the density from given nuclide fractions
 
|-
 
|-
 
| <tt>''NUC<sub>n</sub>''</tt>
 
| <tt>''NUC<sub>n</sub>''</tt>
| : Identifier of ''n''th nuclide in composition, e.g. "92235.03c" or "U-235.03c".
+
| : Identifier of ''n''-th nuclide in composition
 
|-
 
|-
 
| <tt>''FRAC<sub>n</sub>''</tt>
 
| <tt>''FRAC<sub>n</sub>''</tt>
| : Fraction of ''n''th nuclide in composition, positive values are interpreted as atomic fractions/densities, negative values as mass fractions/densities.
+
| : fraction of ''n''-th nuclide in composition (positive value = atomic fractions/density, negative values = mass fractions/density)
 
|-
 
|-
 
|}
 
|}
  
<u>Optional cards:</u>
+
The remaining parameters are defined by separate key words followed by the input values, being optional.
  
'''tmp''': <span id="mat_tmp"></span> Material temperature for [[Doppler-broadening preprocessor routine|Doppler-preprocessor]]
+
<u>Notes:</u>
 +
*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>
 
| <tt>''TEMP''</tt>
| : Temperature (in Kelvin) of the material for [[Doppler-broadening preprocessor routine|Doppler-broadening preprocessor]]
+
| : temperature of the material [in K]
 
|}
 
|}
  
'''tms''': <span id="mat_tmp"></span> Material temperature for on-the-fly [[TMS on-the-fly temperature treatment routine‏‎|TMS temperature treatment]]
+
<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_tmp"></span>
  
 
{|
 
{|
 
| <tt>''TEMP''</tt>
 
| <tt>''TEMP''</tt>
| : Temperature (in Kelvin) of the material for on-the-fly [[TMS on-the-fly temperature treatment routine‏‎|TMS temperature treatment]]
+
| : temperature of the material [in K]
 
|}
 
|}
  
'''tft''': <span id="mat_tft"></span> Temperature limits for material for [[Coupled multi-physics calculations|coupled multi-physics calculations]]
+
<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>
 
| <tt>''T<sub>MIN</sub>''</tt>
| : Lower limit for material temperature
+
| : lower limit for material temperature [in K]
 
|-
 
|-
 
| <tt>''T<sub>MAX</sub>''</tt>
 
| <tt>''T<sub>MAX</sub>''</tt>
| : Upper limit for material temperature
+
| : upper limit for material temperature [in K]
 
|}
 
|}
  
'''rgb''': <span id="mat_rgb"></span> Material color for [[#plot|geometry plots]]
+
<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]]).
 +
*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>
 
| <tt>''R''</tt>
| : Value for the red channel of [[#plot|geometry plots]] (between 0 and 255)
+
| : value for the red channel (between 0 and 255)
 
|-
 
|-
 
| <tt>''G''</tt>
 
| <tt>''G''</tt>
| : Value for the green channel of geometry plots (between 0 and 255)
+
| : value for the green channel (between 0 and 255)
 
|-
 
|-
 
| <tt>''B''</tt>
 
| <tt>''B''</tt>
| : Value for the blue channel of geometry plots (between 0 and 255)
+
| : value for the blue channel (between 0 and 255)
 
|}
 
|}
  
'''vol''': <span id="mat_vol"></span> Material volume
+
<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>
 
| <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)
+
| : volume of the material [in cm<sup>3</sup>] (3D geometry) or cross-sectional area [in cm<sup>2</sup>] (2D geometry)
 
|}
 
|}
  
'''mass''': <span id="mat_mass"></span> Material mass
+
<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>
 
| <tt>''MASS''</tt>
| : Mass of the material in grams
+
| : mass of the material [in g]
 
|}
 
|}
  
'''burn''': <span id="mat_burn"></span> Flag material for depletion
+
<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>
 
| <tt>''N<sub>R</sub>''</tt>
| : Set to 1 in order to deplete material. The depletion zone division should be done using the [[#div|div]]-card.
+
| : option to flag the material as burnable (1/yes) or non-burnable (0/no). The default option is "<tt>non-burnable</tt>"
 +
|-
 
|}
 
|}
  
'''fix''': <span id="mat_fix"></span> Library information for decay nuclides
+
<u>Notes:</u>
 +
*In order to deplete the material and include it in the burnup calculation, the flag must be set to "<tt>1</tt>
 +
*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 (not recommended)
 +
 
 +
 
 +
Material library information for nuclides without cross section data (<tt>'''fix'''</tt>): <span id="mat_fix"></span>
  
 
{|
 
{|
 
| <tt>''LIB''</tt>
 
| <tt>''LIB''</tt>
| : Library ID (e.g. "09c") for nuclides without cross section data.
+
| : library ID (e.g. "09c") for nuclides without cross section data.
 
|-
 
|-
 
| <tt>''TEMP''</tt>
 
| <tt>''TEMP''</tt>
| : Temperature (in Kelvin) for nuclides without cross section data.
+
| : temperature for nuclides without cross section data [in K]
 
|}
 
|}
  
'''moder''': <span id="mat_moder"></span> Use thermal scattering data library for a nuclide. The moder entry can be used several times to define thermal scattering libraries for multiple nuclides, such as hydrogen and deuterium in heavy water.
+
<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.
 +
 
 +
 
 +
Material associated thermal-scattering data (<tt>'''moder'''</tt>): <span id="mat_moder"></span>
  
 
{|
 
{|
 
| <tt>''THNAME''</tt>
 
| <tt>''THNAME''</tt>
| : Name of the thermal scattering data library defined using the [[#therm_and_thermstoch_.28thermal_scattering.29|therm]] card.
+
| : name of the [[#therm_and_thermstoch_.28thermal_scattering.29|thermal scattering data library]]
 
|-
 
|-
 
| <tt>''ZA''</tt>
 
| <tt>''ZA''</tt>
| : Nuclide ZA of the thermal scatterer (e.g. 1001 for H-1).
+
| : nuclide ZA of the thermal scatterer (e.g. 1001 for H-1).
 
|}
 
|}
  
 
<u>Notes:</u>
 
<u>Notes:</u>
*This description is incomplete for both the descriptions and optional settings.
+
*It links the thermal-scattering data library for a given nuclide within the material composition.
*See [[defining material volumes]] and [[#set mvol|set mvol]] regarding other ways to set the material volumes for example in burnup calculations.
+
*The thermal-scattering data library and the associated temperature treatment is defined by the [[#therm_and_thermstoch_.28thermal_scattering.29|therm]] card.
*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 nuclides without associated cross-sections, include the '''fix''' option to indicate the library and temperature of the given nuclides.
+
*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.
  
 
=== mesh (mesh plot definition)<span id="mesh"></span> ===
 
=== mesh (mesh plot definition)<span id="mesh"></span> ===
Line 1,654: Line 2,047:
 
|-
 
|-
 
| <tt>''XPIX''</tt>
 
| <tt>''XPIX''</tt>
| : horizontal image size in pixels
+
| : horizontal image size [in pixels]
 
|-
 
|-
 
| <tt>''YPIX''</tt>
 
| <tt>''YPIX''</tt>
| : vertical image size in pixels
+
| : vertical image size [in pixels]
 
|-
 
|-
 
| <tt>''SYM''</tt>
 
| <tt>''SYM''</tt>
Line 1,663: Line 2,056:
 
|-
 
|-
 
| <tt>''MIN<sub>n</sub>'' ''MAX<sub>n</sub>''</tt>
 
| <tt>''MIN<sub>n</sub>'' ''MAX<sub>n</sub>''</tt>
| : boundaries of the plotted region
+
| : boundaries of the plotted region [in cm]
 
|-
 
|-
 
| <tt>''CMAP''</tt>
 
| <tt>''CMAP''</tt>
Line 1,681: Line 2,074:
 
*Some special detector types, such as pulse-height detectors and analog photon heating detectors cannot be associated with mesh plots.
 
*Some special detector types, such as pulse-height detectors and analog photon heating detectors cannot be associated with mesh plots.
 
*The mesh plot always produces results that are integrated over space. If no boundaries are provided, the integration is carried over the entire geometry.
 
*The mesh plot always produces results that are integrated over space. If no boundaries are provided, the integration is carried over the entire geometry.
*Setting the orientation parameter of a detector mesh plot to 4 produces a plot in cylindrical coordinates. Instead of Cartesian boundaries the entered values are then the radius, angle and axial coordinate.
+
*Setting the orientation parameter of a detector mesh plot to 4 produces a plot in cylindrical coordinates. Instead of Cartesian boundaries the entered values are then the radius and axial coordinate.
 
*The symmetry option was used in Serpent 1. The parameter must be provided for Serpent 2 as well, even though it is not used. The value can be set to zero.
 
*The symmetry option was used in Serpent 1. The parameter must be provided for Serpent 2 as well, even though it is not used. The value can be set to zero.
 
*Mesh plot produced by the nth mesh-card is written in file <tt>[input]_mesh[n].png</tt>.
 
*Mesh plot produced by the nth mesh-card is written in file <tt>[input]_mesh[n].png</tt>.
Line 1,699: Line 2,092:
 
|-
 
|-
 
| <tt>''NUC<sub>n</sub>''</tt>
 
| <tt>''NUC<sub>n</sub>''</tt>
| :  identifier of <tt>''n''</tt>th nuclide in composition
+
| :  identifier of <tt>''n''</tt>-th nuclide in composition
 
|-
 
|-
 
| <tt>''&lambda;''<sub>n</sub></tt>
 
| <tt>''&lambda;''<sub>n</sub></tt>
| : reprocessing constant of <tt>''n''</tt>th nuclide in composition  (in s<sup>-1</sup>)
+
| : reprocessing constant of <tt>''n''</tt>-th nuclide in composition  [in s<sup>-1</sup>]
 
|}
 
|}
  
 
<u>Notes:</u>
 
<u>Notes:</u>
  
* The nuclide ID can be replaced with "all", in which case all nuclides are included with the same reprocessing fraction ''&lambda;''.
+
* The nuclide ID can be replaced with "<tt>all</tt>", in which case all nuclides are included with the same reprocessing fraction ''&lambda;''.
 
* The nuclide ID should follow the ZAI or ISO format (e.g., 922350 or U-235).
 
* The nuclide ID should follow the ZAI or ISO format (e.g., 922350 or U-235).
  
 
=== mix (mixture definition)<span id="mix"></span> ===
 
=== mix (mixture definition)<span id="mix"></span> ===
  
  '''mix''' ''NAME'' [ '''rgb''' ''R G B'' ]
+
  '''mix''' ''NAME'' [ [[#mix_rgb|'''rgb''']] ''R G B'' ]
           [ '''vol''' ''VOL'' ]
+
           [ [[#mix_vol|'''vol''']] ''VOL'' ]
           [ '''mass''' ''MASS'' ]
+
           [ [[#mix_mass|'''mass''']] ''MASS'' ]
 
  ''MAT<sub>1</sub>'' ''F<sub>1</sub>''
 
  ''MAT<sub>1</sub>'' ''F<sub>1</sub>''
 
  ''MAT<sub>2</sub>'' ''F<sub>2</sub>''
 
  ''MAT<sub>2</sub>'' ''F<sub>2</sub>''
 
  ...
 
  ...
Defines a mixture of two or several materials. Input values:
+
Defines a mixture of two or several materials. Mandatory input values:
 
+
<u>Mandatory information:</u>
+
  
 
{|
 
{|
Line 1,727: Line 2,118:
 
|-
 
|-
 
| <tt>''F<sub>n</sub>''</tt>
 
| <tt>''F<sub>n</sub>''</tt>
| : material fraction (positive values for volume, negative values for mass fractions)
+
| : material fraction (positive value = volume fraction, negative value = mass fraction)
 
|-
 
|-
 
|}
 
|}
  
<u>Optional cards:</u>
+
The remaining parameters are defined by separate key words followed by the input values, being optional.
  
'''rgb''': <span id="mat_rgb"></span> Material color for [[#plot|geometry plots]]
+
<u>Notes:</u>
 +
 
 +
*Mixtures can be used to define complicated material definitions consisting of two or more physical materials mixed homogeneously.
 +
*Serpent decomposes these mixtures into standard materials before running the transport simulation.
 +
*The decomposed material compositions can be written into file using the [[Installing and running Serpent#Running Serpent|<tt>''-mix''</tt>]] command line option.
 +
*Nuclide specific [[#mat_moder|thermal scattering data]] is automatically brought from component materials to the mixture.
 +
*Many other input cards such as [[#set_trc|set trc]], [[#set_iter_nuc|set iter nuc]], [[Sensitivity_calculations#Choosing_materials_to_perturb|sens pert matlist]] are not automatically inherited by the mixture from the components and should be directly defined using the mixture material name (opposed to component material names) if they are to be applied to the mixture.
 +
*Burnable mixtures are not supported.
 +
 
 +
 
 +
<u>Optional entries:</u>
 +
 
 +
Mixture RGB-color  (<tt>'''rgb'''</tt>): <span id="mix_rgb"></span>
  
 
{|
 
{|
 
| <tt>''R''</tt>
 
| <tt>''R''</tt>
| : Value for the red channel of [[#plot|geometry plots]] (between 0 and 255)
+
| : value for the red channel (between 0 and 255)
 
|-
 
|-
 
| <tt>''G''</tt>
 
| <tt>''G''</tt>
| : Value for the green channel of geometry plots (between 0 and 255)
+
| : value for the green channel (between 0 and 255)
 
|-
 
|-
 
| <tt>''B''</tt>
 
| <tt>''B''</tt>
| : Value for the blue channel of geometry plots (between 0 and 255)
+
| : value for the blue channel (between 0 and 255)
 
|}
 
|}
  
'''vol''': <span id="mat_vol"></span> Material volume
+
<u>Notes:</u>
 +
*RGB color coding for material representation in [[#plot|geometry plots]].
 +
 
 +
 
 +
Mixture volume (<tt>'''vol'''</tt>): <span id="mix_vol"></span>
  
 
{|
 
{|
 
| <tt>''VOL''</tt>
 
| <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)
+
| : volume of the material [in cm<sup>3</sup>] (3D geometry) or cross-sectional area [in cm<sup>2</sup>] (2D geometry)
 
|}
 
|}
  
'''mass''': <span id="mat_mass"></span> Material mass
+
 
 +
Mixture mass (<tt>'''mass'''</tt>): <span id="mix_mass"></span>
  
 
{|
 
{|
 
| <tt>''MASS''</tt>
 
| <tt>''MASS''</tt>
| : Mass of the material in g
+
| : mass of the material [in g]
 
|}
 
|}
 
<u>Notes:</u>
 
 
*Mixtures can be used to define complicated material definitions consisting of two or more physical materials mixed homogeneously.
 
*Serpent decomposes these mixtures into standard materials before running the transport simulation.
 
*The decomposed material compositions can be written into file using the [[Installing and running Serpent#Running Serpent|-mix]] command line option.
 
*Nuclide specific [[#mat_moder|thermal scattering data]] is automatically brought from component materials to the mixture.
 
*Many other input cards such as [[#set_trc|set trc]], [[#set_iter_nuc|set iter nuc]], [[Sensitivity_calculations#Choosing_materials_to_perturb|sens pert matlist]] are not automatically inherited by the mixture from the components and should be directly defined using the mixture material name (opposed to component material names) if they are to be applied to the mixture.
 
  
 
=== nest (nested universe definition)<span id="nest"></span> ===
 
=== nest (nested universe definition)<span id="nest"></span> ===
Line 1,795: Line 2,195:
 
|-
 
|-
 
| <tt>''R<sub>1</sub> ... R<sub>N-1</sub>''</tt>
 
| <tt>''R<sub>1</sub> ... R<sub>N-1</sub>''</tt>
| : outer radii
+
| : outer radii [in cm]
 
|-
 
|-
 
| <tt>''TYPE<sub>1</sub> ... TYPE<sub>N-1</sub>''</tt>
 
| <tt>''TYPE<sub>1</sub> ... TYPE<sub>N-1</sub>''</tt>
Line 1,828: Line 2,228:
 
|-
 
|-
 
| <tt>''R<sub>1</sub> ... R<sub>N-1</sub>''</tt>
 
| <tt>''R<sub>1</sub> ... R<sub>N-1</sub>''</tt>
| : outer radii
+
| : outer radii [in cm]
 
|}
 
|}
  
Line 1,839: Line 2,239:
 
*See also description of [[#pbed (explicit stochastic geometry)|explicit stochastic geometry type]].
 
*See also description of [[#pbed (explicit stochastic geometry)|explicit stochastic geometry type]].
  
=== pbed (explicit stochastic (pebble bed) geometry) ===
+
=== pbed (explicit stochastic (pebble bed) geometry definition) ===
  
  '''pbed''' ''U<sub>0</sub>'' ''U<sub>bg</sub>'' ''FILE'' [ ''OPT'' ]  
+
  '''pbed''' ''UNI<sub>0</sub>'' ''UNI<sub>bg</sub>'' ''FILE'' [ ''OPT'' ]  
 
Defines a stochastic particle / pebble-bed geometry. Input values:
 
Defines a stochastic particle / pebble-bed geometry. Input values:
  
 
{|
 
{|
| <tt>''U<sub>0</sub>''</tt>  
+
| <tt>''UNI<sub>0</sub>''</tt>  
 
| : universe name for the dispersed medium
 
| : universe name for the dispersed medium
 
|-
 
|-
| <tt>''U<sub>bg</sub>''</tt>
+
| <tt>''UNI<sub>bg</sub>''</tt>
 
| : background universe, i.e. universe filling the space between particles / pebbles
 
| : background universe, i.e. universe filling the space between particles / pebbles
 
|-
 
|-
Line 1,855: Line 2,255:
 
|-
 
|-
 
| <tt>''OPT''</tt>
 
| <tt>''OPT''</tt>
|: additional options (currently only '''pow''' to produce power distribution in a separate output file)
+
|: additional options (currently only supported option: "<tt>pow</tt>" = power distribution)
 
|}
 
|}
The syntax of the file containing the particle/pebble data is:
 
  
''X<sub>1</sub>'' ''Y<sub>1</sub>'' ''Z<sub>1</sub>'' ''R<sub>1</sub>'' ''U<sub>1</sub>''
+
The <u>syntax of the file</u> containing the particle/pebble data is:
  
''X<sub>2</sub>'' ''Y<sub>2</sub>'' ''Z<sub>2</sub>'' ''R<sub>2</sub>'' ''U<sub>2</sub>''
+
::{| class="toccolours" style="text-align: left;"
 +
| ''X<sub>1</sub>'' ''Y<sub>1</sub>'' ''Z<sub>1</sub>'' ''R<sub>1</sub>'' ''UNI<sub>1</sub>''
 +
|-
 +
| ''X<sub>2</sub>'' ''Y<sub>2</sub>'' ''Z<sub>2</sub>'' ''R<sub>2</sub>'' ''UNI<sub>2</sub>''
 +
|-
 +
| ...
 +
|-
 +
|}
  
...
+
where:
 
+
Where:
+
 
{|
 
{|
 
|<tt>''X<sub>n</sub>'', ''Y<sub>n</sub>'', ''Z<sub>n</sub>''</tt>
 
|<tt>''X<sub>n</sub>'', ''Y<sub>n</sub>'', ''Z<sub>n</sub>''</tt>
|: are the coordinates
+
|: are the coordinates [in cm]
 
|-
 
|-
 
|<tt>''R<sub>n</sub>''</tt>
 
|<tt>''R<sub>n</sub>''</tt>
|: is the radius
+
|: is the radius [in cm]
 
|-
 
|-
|<tt>''U<sub>n</sub>''</tt>
+
|<tt>''UNI<sub>n</sub>''</tt>
 
|: is the universe
 
|: is the universe
 
|}
 
|}
Line 1,879: Line 2,283:
 
<u>Notes:</u>
 
<u>Notes:</u>
  
*Creates a universe (<tt>''U<sub>0</sub>''</tt>), which is filled with spherical sub-universes for which the coordinates are read from a separate file.
+
*Creates a universe (<tt>''UNI<sub>0</sub>''</tt>), which is filled with spherical sub-universes for which the coordinates are read from a separate file.
*The coordinates can be defined manually, or using the [[Installing_and_running_Serpent#Particle_disperser_routine|automated disperser routine]].
+
*The coordinates can be defined manually, or using the [[Installing_and_running_Serpent#Running Serpent|<tt>''-disperse''</tt>]] command line option which launches the particle disperser routine.
 
*Can be used for modelling stochastic particle / pebble-bed geometries in multiple levels.
 
*Can be used for modelling stochastic particle / pebble-bed geometries in multiple levels.
*If the power distribution option is set, the pebble/particle-wise distribution is written in file <tt>[U<sub>0</sub>].out</tt>.
+
*If the "<tt>pow</tt>" (power distribution) option is set, the pebble/particle-wise distribution is written in file <tt>[''FILE'']_pow[bu].m</tt>, where "<tt>bu</tt>" is the burnup step, from version 2.2.1 and on (in previous versions, <tt>[''FILE''].out</tt>).
 
*See also [[Collection_of_example_input_files#Simple_burnup_examples|HTGR geometry examples]].
 
*See also [[Collection_of_example_input_files#Simple_burnup_examples|HTGR geometry examples]].
  
Line 1,909: Line 2,313:
 
|-
 
|-
 
|<tt>''E<sub>max,i</sub>, R<sub>i</sub>''</tt>
 
|<tt>''E<sub>max,i</sub>, R<sub>i</sub>''</tt>
|: are the maximum energy-resolution tabulated pairs
+
|: are the maximum energy-resolution tabulated pairs [in MeV (energy)]
 
|}
 
|}
  
Line 1,925: Line 2,329:
 
|-
 
|-
 
|<tt>''E<sub>max,i</sub>, FWHM<sub>i</sub>''</tt>
 
|<tt>''E<sub>max,i</sub>, FWHM<sub>i</sub>''</tt>
|: are the maximum energy-full width at half maximum pairs
+
|: are the maximum energy-full width at half maximum pairs [in MeV (energy)]
 
|}
 
|}
  
Line 1,965: Line 2,369:
 
|-
 
|-
 
| <tt>''R<sub>1</sub> ... R<sub>N-1</sub>''</tt>
 
| <tt>''R<sub>1</sub> ... R<sub>N-1</sub>''</tt>
| : outer radii
+
| : outer radii [in cm]
 
|}
 
|}
  
Line 1,988: Line 2,392:
 
|-
 
|-
 
| <tt>''XPIX''</tt>
 
| <tt>''XPIX''</tt>
| : horizontal image size in pixels
+
| : horizontal image size [in pixels]
 
|-
 
|-
 
| <tt>''YPIX''</tt>
 
| <tt>''YPIX''</tt>
| : vertical image size in pixels
+
| : vertical image size [in pixels]
 
|-
 
|-
 
| <tt>''POS''</tt>
 
| <tt>''POS''</tt>
| : position of plot plane
+
| : position of plot plane [in cm]
 
|-
 
|-
 
| <tt>''MIN<sub>1</sub>''</tt>
 
| <tt>''MIN<sub>1</sub>''</tt>
| : minimum horizontal coordinate of plotted region
+
| : minimum horizontal coordinate of plotted region [in cm]
 
|-
 
|-
 
| <tt>''MAX<sub>1</sub>''</tt>
 
| <tt>''MAX<sub>1</sub>''</tt>
| : maximum horizontal coordinate of plotted region
+
| : maximum horizontal coordinate of plotted region [in cm]
 
|-
 
|-
 
| <tt>''MIN<sub>2</sub>''</tt>
 
| <tt>''MIN<sub>2</sub>''</tt>
| : minimum vertical coordinate of plotted region
+
| : minimum vertical coordinate of plotted region [in cm]
 
|-
 
|-
 
| <tt>''MAX<sub>2</sub>''</tt>
 
| <tt>''MAX<sub>2</sub>''</tt>
| : maximum vertical coordinate of plotted region
+
| : maximum vertical coordinate of plotted region [in cm]
 
|-
 
|-
 
| <tt>''F<sub>min</sub>''</tt>
 
| <tt>''F<sub>min</sub>''</tt>
Line 2,015: Line 2,419:
 
|-
 
|-
 
| <tt>''E''</tt>
 
| <tt>''E''</tt>
| : particle energy for importance map plots
+
| : particle energy for importance map plots [in MeV]
 
|}
 
|}
  
Line 2,021: Line 2,425:
  
 
*The <tt>''TYPE''</tt> parameter consists of one or two concatenated values ('AB'):  
 
*The <tt>''TYPE''</tt> parameter consists of one or two concatenated values ('AB'):  
*#The first value ('A') defines the plot plane (1 = yz, 2 = xz, 3 = xy).  
+
**The first value ('A') defines the plot plane (1 = yz, 2 = xz, 3 = xy).  
*#The second value ('B') defines which boundaries are plotted (0 = no boundaries, 1 = cell boundaries, 2 = material boundaries, 3 = both). If the second value in is not provided, material boundaries are plotted.
+
**The second value ('B') defines which boundaries are plotted (0 = no boundaries, 1 = cell boundaries, 2 = material boundaries, 3 = both).  
*Importance maps read using the [[#wwin (weight window mesh definition)|wwin card]] can be plotted on top of the geometry by setting the second value ('B') of the type parameter to 4 (linear color scheme) or 5 (logarithmic color scheme) for cell importances, and to 6 (linear color scheme) or 7 (logarithmic color scheme) for source importances. The input parameters then also include the minimum and maximum importance and the particle energy. If importance maps are provided for both neutrons and photons, they can be plotted by entering positive and negative energy values, respectively.  
+
***If the second value in is not provided, material boundaries are plotted.
*If both, minimum and maximum importance values are set to "-1", Serpent automatically adjusts them based on the weight-window mesh data, from version 2.2.0 and on.
+
*The relative dimensions of image size (<tt>''XPIX''</tt>, <tt>''YPIX''</tt>) should match that of the plotted region. Otherwise the image gets distorted.
*Each material plotted with different color. The colors are sampled randomly, unless defined using the '''rgb''' entry in the [[#mat (material definition)|material card]].
+
*The position parameter <tt>''POS''</tt> defines the location of the plot plane on the axis perpendicular to it (e.g. z-coordinate for xy-type plot).
*Void is plotted in black and special colors are used to plot geometry errors (red = overlap, green = undefined region).
+
*The minimum and maximum coordinates: <tt>''MIN<sub>n</sub>''</tt>, <tt>''MAX<sub>n</sub>''</tt>, define the boundaries of the plotted region (e.g. minimum and maximum x- and y-coordinates for xy-type plot).
*The position parameter defines the location of the plot plane on the axis perpendicular to it (e.g. z-coordinate for xy-type plot).
+
** If the coordinates are not provided, the plot is extended to the maximum dimensions of the geometry.
*The minimum and maximum coordinates define the boundaries of the plotted region (e.g. minimum and maximum x- and y-coordinates for xy-type plot). If these coordinates are not provided, the plot is extended to the maximum dimensions of the geometry.
+
 
*The relative dimensions of image size in pixels should match that of the plotted region. Otherwise the image gets distorted.
+
*The second format allows to plot he importance maps read using the [[#wwin (weight window mesh definition)|wwin card]]:
*Geometry plotter requires compiling the source code with [[Compiling Serpent|GD Graphics libraries]].
+
**They can be plotted on top of the geometry by setting the second value ('B') of the type parameter for:
*[[Installing and running Serpent#Running Serpent|Command line parameter]] <tt>''-plot''</tt> stops the execution after the geometry plots are produced. Option <tt>''-qp''</tt> invokes a quick plot mode, which does not check for overlaps. Option <tt>''-noplot''</tt> skips the geometry plots altogether.
+
*** Cell importances: 4 (linear color scheme) or 5 (logarithmic color scheme)
 +
*** Source importances: 6 (linear color scheme) or 7 (logarithmic color scheme)
 +
**The input parameters include the minimum and maximum importance (<tt>''F<sub>min</sub>''</tt>, <tt>''F<sub>max</sub>''</tt>) and the particle energy, <tt>''E''</tt>.  
 +
***If importance maps are provided for both neutrons and photons, they can be plotted by entering positive and negative energy values, respectively.  
 +
***If both, minimum and maximum importance values are set to "-1", Serpent automatically adjusts them based on the weight-window mesh data, from version 2.2.0 and on.
 +
****If the calculation fails on providing those minimum and maximum values due to the weight-window evaluation, the values are set by default to (1E-200, 1E+200).
 +
**''Note to developers: particle type should be included as an input parameter in importance map plots.''
 +
 
 +
*Material colors:
 +
**Each material plotted with different color.
 +
**The colors are sampled randomly, unless defined using the '''rgb''' entry in the material card (see [[#mat (material definition)|mat]]) or mixture card (see [[#mix (mixture definition)|mix]])
 +
**Special RGB-colors:
 +
 
 +
::{|class="wikitable" style="text-align: left;"
 +
! RGB value
 +
! Color
 +
! Description
 +
|-
 +
| (0, 0, 0)
 +
| <span style="color:#000; background:#000">COLOR</span>
 +
| Outside cell or void-material
 +
|-
 +
| (0, 255, 0)
 +
| <span style="color:#00FF00; background:#00FF00">COLOR</span>
 +
| No cell found at coordinates
 +
|-
 +
| (255, 0, 0)
 +
| <span style="color:#FF0000; background:#FF0000">COLOR</span>
 +
| Overlap of multiple cells found at coordinates
 +
|-
 +
| (255, 0, 255)
 +
| <span style="color:#FF00FF; background:#FF00FF">COLOR</span>
 +
| Undefined material density factor at coordinates
 +
|}
 +
 
 +
*Geometry plotter requires compiling the source code with [[Installing and running Serpent#GD Graphics library|GD Graphics libraries]].
 +
*Command line options:
 +
**[[Installing and running Serpent#Running Serpent|<tt>''-plot''</tt>]] stops the execution after the geometry plots are produced
 +
**[[Installing and running Serpent#Running Serpent|<tt>''-qp''</tt>]] invokes a quick plot mode, which does not check for overlaps
 +
**[[Installing and running Serpent#Running Serpent|<tt>''-noplot''</tt>]] skips the geometry plots altogether
 
*See also [[Visualizing the results#Geometry plotter|detailed description]] on geometry plotter.
 
*See also [[Visualizing the results#Geometry plotter|detailed description]] on geometry plotter.
*Geometry plot produced by the nth plot-card is written in file <tt>[input]_geom[n].png</tt>.  
+
*The geometry plot produced by the ''n''-th plot-card is written in file <tt>[input]_geom[n].png</tt>.
*''Note to developers: particle type should be included as an input parameter in importance map plots.''
+
  
 
=== rep (reprocessor definition)<span id="rep"></span> ===
 
=== rep (reprocessor definition)<span id="rep"></span> ===
  
 
  '''rep''' ''NAME''
 
  '''rep''' ''NAME''
     [ '''rc''' ''SRC'' ''TGT'' ''MFLOW'' ''MODE'' ]
+
     [ [[#rep_rc|'''rc''']] ''SRC'' ''TGT'' ''MFLOW'' ''MODE'' ]
     [ '''rm''' ''MAT<sub>1</sub>'' ''MAT<sub>2</sub>'' ]
+
     [ [[#rep_rm|'''rm''']] ''MAT<sub>1</sub>'' ''MAT<sub>2</sub>'' ]
     [ '''ru''' ''UNI<sub>1</sub>'' ''UNI<sub>2</sub>'' ]
+
     [ [[#rep_ru|'''ru''']] ''UNI<sub>1</sub>'' ''UNI<sub>2</sub>'' ]
  
Defines the reprocessing controllers. Input values:
+
Defines the reprocessing controllers. The first parameter:
  
 
{|
 
{|
 
| <tt>''NAME''</tt>  
 
| <tt>''NAME''</tt>  
 
| : name of the reprocessor.
 
| : name of the reprocessor.
|-
+
|}
 +
 
 +
The remaining parameters are defined by separate key words followed by the input values.
 +
 
 +
<u>Notes:</u>
 +
*The reprocessor name identifies the reprocessing regime in the depletion calculation [[#dep|dep card]]. The syntax corresponds to '''dep pro''' ''NAME''.
 +
*Multiple reprocessing controllers/regimes can be defined within the same reprocessor definition.
 +
 
 +
 
 +
<u>Reprocessing regime types:</u>
 +
 
 +
Reprocessing continuos regime (<tt>'''rc'''</tt>):<span id="rep_rc"></span>
 +
{|
 
| <tt>''SRC''</tt>
 
| <tt>''SRC''</tt>
 
| : name of the source material, from which the flow is moved
 
| : name of the source material, from which the flow is moved
Line 2,060: Line 2,514:
 
| <tt>''MODE''</tt>
 
| <tt>''MODE''</tt>
 
| : continuous reprocessing mode
 
| : continuous reprocessing mode
|-
 
| <tt>''MAT<sub>1</sub>''</tt>
 
| : name of the replaced material
 
|-
 
| <tt>''MAT<sub>2</sub>''</tt>
 
| : name of the replacing material
 
|-
 
| <tt>''UNI<sub>1</sub>''</tt>
 
| : name of the replaced universe
 
|-
 
| <tt>''UNI<sub>2</sub>''</tt>
 
| : name of the replacing universe
 
 
|-
 
|-
 
|}
 
|}
  
 
<u>Notes:</u>
 
<u>Notes:</u>
 
*The reprocessor name identifies the reprocessing regime in the depletion calculation [[#dep|dep card]]. The syntax corresponds to '''dep pro''' ''NAME''.
 
 
*The nuclides identifier of those included in both source <tt>''SRC''</tt> and target <tt>''TGT''</tt> materials in reprocessors should follow the same format, either ZA.ID or ISO.ID (for nuclides with cross sections), or ZAI (for nuclides without associated cross sections, and adding the '''fix''' entry to the [[#mat (material definition)|mat card]]).
 
*The nuclides identifier of those included in both source <tt>''SRC''</tt> and target <tt>''TGT''</tt> materials in reprocessors should follow the same format, either ZA.ID or ISO.ID (for nuclides with cross sections), or ZAI (for nuclides without associated cross sections, and adding the '''fix''' entry to the [[#mat (material definition)|mat card]]).
*The (continuous) reprocessing implementation works with materials, not universes. Therefore, define the universes associated with those burnable materials as surface-cell type universes.
+
*The continuous reprocessing regime works with materials, not universes. Therefore, define the universes associated with those burnable materials as surface-cell type universes.
*Multiple reprocessing controllers/regimes can be defined within the same reprocessor definition.
+
*The continuous reprocessing regime can be used to define the material flow between the source and the target materials.  
*The '''rc''' continuous reprocessing option can be used to define the material flow between the source and the target materials.  
+
 
**The material flow is defined using the [[#mflow|mflow card]].
 
**The material flow is defined using the [[#mflow|mflow card]].
 
**The continuous reprocessing <tt>''MODE''</tt> defines how to incorporate the material flow into the Bateman equations:
 
**The continuous reprocessing <tt>''MODE''</tt> defines how to incorporate the material flow into the Bateman equations:
Line 2,110: Line 2,549:
 
:*<tt>''MODE''</tt> 1 : subtracts ''&lambda;N'' from the source material and adds it to the target material when solving the Bateman equations.
 
:*<tt>''MODE''</tt> 1 : subtracts ''&lambda;N'' from the source material and adds it to the target material when solving the Bateman equations.
 
:*<tt>''MODE''</tt> 2 : subtracts ''&lambda;N<sub>n</sub>'' from the source material and adds it to the target material when solving the Bateman equations (compositions updated with each burnup step, ''n'').
 
:*<tt>''MODE''</tt> 2 : subtracts ''&lambda;N<sub>n</sub>'' from the source material and adds it to the target material when solving the Bateman equations (compositions updated with each burnup step, ''n'').
*The '''rm''' material reprocessing option replaces one material with another, ''MAT<sub>1</sub>'' by ''MAT<sub>2</sub>''.
 
*The '''ru''' universe reprocessing option replaces one universe with another, ''UNI<sub>1</sub>'' by ''UNI<sub>2</sub>''.
 
  
=== sample (Temperature / density data sample definition)<span id="sample"></span> ===
+
 
 +
Reprocessing material regime (<tt>'''rm'''</tt>):<span id="rep_rm"></span>
 +
 
 +
{|
 +
| <tt>''MAT<sub>1</sub>''</tt>
 +
| : name of the replaced material
 +
|-
 +
| <tt>''MAT<sub>2</sub>''</tt>
 +
| : name of the replacing material
 +
|-
 +
|}
 +
 
 +
<u>Notes:</u>
 +
*The material reprocessing regime replaces one material with another, ''MAT<sub>1</sub>'' by ''MAT<sub>2</sub>''.
 +
 
 +
 
 +
Reprocessing universe regime (<tt>'''ru'''</tt>):<span id="rep_ru"></span>
 +
 
 +
{|
 +
| <tt>''UNI<sub>1</sub>''</tt>
 +
| : name of the replaced universe
 +
|-
 +
| <tt>''UNI<sub>2</sub>''</tt>
 +
| : name of the replacing universe
 +
|-
 +
|}
 +
 
 +
<u>Notes:</u>
 +
*The universe reprocessing regime replaces one universe with another, ''UNI<sub>1</sub>'' by ''UNI<sub>2</sub>''.
 +
 
 +
=== sample (temperature / density data sample definition)<span id="sample"></span> ===
  
 
  '''sample'''  ''N<sub>X</sub>'' ''X<sub>MIN</sub>'' ''X<sub>MAX</sub>''  ''N<sub>Y</sub>'' ''Y<sub>MIN</sub>'' ''Y<sub>MAX</sub>''  ''N<sub>Z</sub>'' ''Z<sub>MIN</sub>'' ''Z<sub>MAX</sub>''
 
  '''sample'''  ''N<sub>X</sub>'' ''X<sub>MIN</sub>'' ''X<sub>MAX</sub>''  ''N<sub>Y</sub>'' ''Y<sub>MIN</sub>'' ''Y<sub>MAX</sub>''  ''N<sub>Z</sub>'' ''Z<sub>MIN</sub>'' ''Z<sub>MAX</sub>''
Line 2,123: Line 2,590:
 
{|
 
{|
 
| <tt>''N<sub>X</sub>''</tt>  
 
| <tt>''N<sub>X</sub>''</tt>  
| : Number of values to sample in the x-direction.
+
| : number of values to sample in the x-direction.
 
|-
 
|-
 
| <tt>''X<sub>MIN</sub>''</tt>  
 
| <tt>''X<sub>MIN</sub>''</tt>  
| : Minimum coordinate to sample from in the x-direction.
+
| : minimum coordinate to sample from in the x-direction [in cm]
 
|-
 
|-
 
| <tt>''X<sub>MAX</sub>''</tt>  
 
| <tt>''X<sub>MAX</sub>''</tt>  
| : Maximum coordinate to sample from in the x-direction.
+
| : maximum coordinate to sample from in the x-direction [in cm]
 
|-
 
|-
 
| <tt>''N<sub>Y</sub>''</tt>  
 
| <tt>''N<sub>Y</sub>''</tt>  
| : Number of values to sample in the y-direction.
+
| : number of values to sample in the y-direction.
 
|-
 
|-
 
| <tt>''Y<sub>MIN</sub>''</tt>  
 
| <tt>''Y<sub>MIN</sub>''</tt>  
| : Minimum coordinate to sample from in the y-direction.
+
| : minimum coordinate to sample from in the y-direction [in cm]
 
|-
 
|-
 
| <tt>''Y<sub>MAX</sub>''</tt>  
 
| <tt>''Y<sub>MAX</sub>''</tt>  
| : Maximum coordinate to sample from in the y-direction.
+
| : maximum coordinate to sample from in the y-direction [in cm]
 
|-
 
|-
 
| <tt>''N<sub>Z</sub>''</tt>  
 
| <tt>''N<sub>Z</sub>''</tt>  
| : Number of values to sample in the z-direction.
+
| : number of values to sample in the z-direction.
 
|-
 
|-
 
| <tt>''Z<sub>MIN</sub>''</tt>  
 
| <tt>''Z<sub>MIN</sub>''</tt>  
| : Minimum coordinate to sample from in the z-direction.
+
| : minimum coordinate to sample from in the z-direction [in cm]
 
|-
 
|-
 
| <tt>''Z<sub>MAX</sub>''</tt>  
 
| <tt>''Z<sub>MAX</sub>''</tt>  
| : Maximum coordinate to sample from in the z-direction.
+
| : maximum coordinate to sample from in the z-direction [in cm]
 
|-
 
|-
 
|}
 
|}
Line 2,185: Line 2,652:
 
|-
 
|-
 
| <tt>''MESH_SPLIT''</tt>
 
| <tt>''MESH_SPLIT''</tt>
| : Splitting criterion for the adaptive search mesh (maximum number of geometry cells in search mesh cell)
+
| : splitting criterion for the adaptive search mesh (maximum number of geometry cells in search mesh cell)
 
|-
 
|-
 
| <tt>''MESH_DIM''</tt>
 
| <tt>''MESH_DIM''</tt>
Line 2,191: Line 2,658:
 
|-
 
|-
 
| <tt>''SZ<sub>i</sub>''</tt>
 
| <tt>''SZ<sub>i</sub>''</tt>
| : Size of the search mesh at level <tt>''i''</tt>
+
| : size of the search mesh at level <tt>''i''</tt>
 
|-
 
|-
 
| <tt>''POINTS_FILE''</tt>
 
| <tt>''POINTS_FILE''</tt>
| : Path to the unstructured mesh points file
+
| : path to the unstructured mesh points file
 
|-
 
|-
 
| <tt>''FACES_FILE''</tt>
 
| <tt>''FACES_FILE''</tt>
| : Path to the unstructured mesh faces file
+
| : path to the unstructured mesh faces file
 
|-
 
|-
 
| <tt>''OWNER_FILE''</tt>
 
| <tt>''OWNER_FILE''</tt>
| : Path to the unstructured mesh owner file
+
| : path to the unstructured mesh owner file
 
|-
 
|-
 
| <tt>''NEIGHBOUR_FILE''</tt>
 
| <tt>''NEIGHBOUR_FILE''</tt>
| : Path to the unstructured mesh neighbour file
+
| : path to the unstructured mesh neighbour file
 
|-
 
|-
 
| <tt>''MATERIALS_FILE''</tt>
 
| <tt>''MATERIALS_FILE''</tt>
| : Path to the unstructured mesh materials file
+
| : path to the unstructured mesh materials file
 
|}
 
|}
  
Line 2,231: Line 2,698:
 
|-
 
|-
 
| <tt>''MESH_SPLIT''</tt>
 
| <tt>''MESH_SPLIT''</tt>
| : Splitting criterion for the adaptive search mesh (maximum number of geometry cells in search mesh cell)
+
| : splitting criterion for the adaptive search mesh (maximum number of geometry cells in search mesh cell)
 
|-
 
|-
 
| <tt>''MESH_DIM''</tt>
 
| <tt>''MESH_DIM''</tt>
Line 2,240: Line 2,707:
 
|-
 
|-
 
| <tt>''MODE''</tt>
 
| <tt>''MODE''</tt>
| : Mode for handling the triangulated geometry (1 = "fast", 2 = "safe").
+
| : mode for handling the triangulated geometry (1 = "fast", 2 = "safe").
 
|-
 
|-
 
| <tt>''R0''</tt>
 
| <tt>''R0''</tt>
| : Radius inside which two points of the STL-geometry are joined into one.
+
| : radius inside which two points of the STL-geometry are joined into one.
 
|-
 
|-
 
| <tt>''BODY<sub>i</sub>''</tt>
 
| <tt>''BODY<sub>i</sub>''</tt>
| : Name of solid body <tt>''i''</tt>
+
| : name of solid body <tt>''i''</tt>
 
|-
 
|-
 
| <tt>''CELL<sub>i</sub>''</tt>
 
| <tt>''CELL<sub>i</sub>''</tt>
| : Name of geometry cell <tt>''i''</tt> linked with body <tt>''i''</tt>
+
| : name of geometry cell <tt>''i''</tt> linked with body <tt>''i''</tt>
 
|-
 
|-
 
| <tt>''MAT<sub>i</sub>''</tt>
 
| <tt>''MAT<sub>i</sub>''</tt>
| : Material filling cell <tt>''i''</tt>
+
| : material filling cell <tt>''i''</tt>
 
|-
 
|-
 
| <tt>''FILE<sub>i</sub>''</tt>
 
| <tt>''FILE<sub>i</sub>''</tt>
| : Path to a file containing an STL solid model, multiple files can be linked to one body
+
| : path to a file containing an STL solid model, multiple files can be linked to one body
 
|-
 
|-
 
| <tt>''SCALE<sub>i</sub>''</tt>
 
| <tt>''SCALE<sub>i</sub>''</tt>
| : Scaling factor for the STL model in <tt>''FILE<sub>i</sub>''</tt>
+
| : scaling factor for the STL model in <tt>''FILE<sub>i</sub>''</tt>
 
|-
 
|-
 
| <tt>''X<sub>i</sub>''</tt>
 
| <tt>''X<sub>i</sub>''</tt>
| : Shift in x-direction to the STL model in <tt>''FILE<sub>i</sub>''</tt>
+
| : shift in x-direction to the STL model in <tt>''FILE<sub>i</sub>''</tt>
 
|-
 
|-
 
| <tt>''Y<sub>i</sub>''</tt>
 
| <tt>''Y<sub>i</sub>''</tt>
| : Shift in y-direction to the STL model in <tt>''FILE<sub>i</sub>''</tt>
+
| : shift in y-direction to the STL model in <tt>''FILE<sub>i</sub>''</tt>
 
|-
 
|-
 
| <tt>''Z<sub>i</sub>''</tt>
 
| <tt>''Z<sub>i</sub>''</tt>
| : Shift in z-direction to the STL model in <tt>''FILE<sub>i</sub>''</tt>
+
| : shift in z-direction to the STL model in <tt>''FILE<sub>i</sub>''</tt>
 
|}
 
|}
  
Line 2,280: Line 2,747:
 
{|
 
{|
 
| <tt>''INTERFACE_FILE''</tt>
 
| <tt>''INTERFACE_FILE''</tt>
| : Path to the [[Multi-physics_interface#Unstructured_mesh_based_interface_and_geometry_definition_.28type_9.29|interface file]] containing the rest of the parameters
+
| : path to the [[Multi-physics_interface#Unstructured_mesh_based_interface_and_geometry_definition_.28type_9.29|interface file]] containing the rest of the parameters
 
|}
 
|}
  
Line 2,302: Line 2,769:
 
           [ [[#src_ss|'''ss''']] ''SURF'' ]
 
           [ [[#src_ss|'''ss''']] ''SURF'' ]
 
           [ [[#src_sd|'''sd''']] ''U'' ''V'' ''W'' ]   
 
           [ [[#src_sd|'''sd''']] ''U'' ''V'' ''W'' ]   
 +
          [ [[#src_sa|'''sa''']] ''PHI'' ]
 
           [ [[#src_se|'''se''']] ''E'' ]
 
           [ [[#src_se|'''se''']] ''E'' ]
 
           [ [[#src_sb|'''sb''']] ''N'' ''INTT'' ''E<sub>1</sub>'' ''WGT<sub>1</sub>'' ''E<sub>2</sub>'' ''WGT<sub>2</sub>'' ... ]
 
           [ [[#src_sb|'''sb''']] ''N'' ''INTT'' ''E<sub>1</sub>'' ''WGT<sub>1</sub>'' ''E<sub>2</sub>'' ''WGT<sub>2</sub>'' ... ]
Line 2,309: Line 2,777:
 
           [ [[#src_si|'''si''']] ''N'' ''P<sub>1</sub>'' ''P<sub>2</sub>'' ... ]
 
           [ [[#src_si|'''si''']] ''N'' ''P<sub>1</sub>'' ''P<sub>2</sub>'' ... ]
 
           [ [[#src_sg|'''sg''']] ''MAT'' ''MODE'' ]
 
           [ [[#src_sg|'''sg''']] ''MAT'' ''MODE'' ]
Source definition. The first parameter:
+
Source definition. The two first parameters:
 
   
 
   
 
{|
 
{|
 +
| <tt>''NAME''</tt>
 +
|: source name
 +
|-
 
| <tt>''PART''</tt>
 
| <tt>''PART''</tt>
 
| : particle type (n = neutron, p = photon)
 
| : particle type (n = neutron, p = photon)
 
|}
 
|}
  
is optional in single particle simulations. The remaining parameters are defined by separate key words followed by the input values.
+
The remaining parameters are defined by separate key words followed by the input values.
 +
 
 +
<u>Notes:</u>
 +
*The particle type <tt>''PART''</tt> is optional in single particle simulations.
 +
*A single source card may include one or several source types.
 +
 
  
 +
<u>Source types:</u>
  
 
Source weight (<tt>'''sw'''</tt>):<span id="src_sw"></span>
 
Source weight (<tt>'''sw'''</tt>):<span id="src_sw"></span>
Line 2,369: Line 2,846:
 
{|
 
{|
 
| <tt>''X''</tt>, <tt>''Y''</tt>, <tt>''Z''</tt>,
 
| <tt>''X''</tt>, <tt>''Y''</tt>, <tt>''Z''</tt>,
| : coordinates of the source point
+
| : coordinates of the source point [in cm]
 
|}
 
|}
  
Line 2,381: Line 2,858:
 
{|
 
{|
 
| <tt>''X<sub>MIN</sub>''</tt>, <tt>''X<sub>MAX</sub>''</tt>
 
| <tt>''X<sub>MIN</sub>''</tt>, <tt>''X<sub>MAX</sub>''</tt>
| : Boundaries on X-axis
+
| : boundaries on X-axis [in cm]
 
|-
 
|-
 
| <tt>''Y<sub>MIN</sub>''</tt>, <tt>''Y<sub>MAX</sub>''</tt>
 
| <tt>''Y<sub>MIN</sub>''</tt>, <tt>''Y<sub>MAX</sub>''</tt>
| : Boundaries on Y-axis
+
| : boundaries on Y-axis [in cm]
 
|-
 
|-
 
| <tt>''Z<sub>MIN</sub>''</tt>, <tt>''Z<sub>MAX</sub>''</tt>
 
| <tt>''Z<sub>MIN</sub>''</tt>, <tt>''Z<sub>MAX</sub>''</tt>
| : Boundaries on Z-axis
+
| : boundaries on Z-axis [in cm]
 
|-
 
|-
 
| <tt>''R<sub>MIN</sub>''</tt>, <tt>''R<sub>MAX</sub>''</tt>
 
| <tt>''R<sub>MIN</sub>''</tt>, <tt>''R<sub>MAX</sub>''</tt>
| : Radial boundaries
+
| : radial boundaries [in cm]
 
|-
 
|-
 
|}
 
|}
Line 2,401: Line 2,878:
  
  
Surface source (<tt>'''ss'''</tt>):<span id="src_ss"></span>
+
Source surface (<tt>'''ss'''</tt>):<span id="src_ss"></span>
  
 
{|
 
{|
Line 2,410: Line 2,887:
 
<u>Notes:</u>
 
<u>Notes:</u>
 
*The surface source is currently limited to infinite vertical cylinder (<tt>'''cyl'''</tt>) and sphere (<tt>'''sph'''</tt>) surface types.
 
*The surface source is currently limited to infinite vertical cylinder (<tt>'''cyl'''</tt>) and sphere (<tt>'''sph'''</tt>) surface types.
*Particles are started in the direction of the inward surface normal.
+
*The default behavior is that particles are started in the direction of the outward surface normal.
 +
*Positive and negative surface entries refer to neutrons being emitted in the direction of the positive and negative surface normal, respectively (meaning: positive = outward, negative = inward - same convention as for the surface detectors).
  
  
Line 2,424: Line 2,902:
 
*If no directional dependence is defined, the direction of source particles is sampled isotropically.
 
*If no directional dependence is defined, the direction of source particles is sampled isotropically.
  
 +
 +
Source angular-aperture (<tt>'''sa'''</tt>):<span id="src_sa"></span>
 +
 +
{|
 +
| <tt>''PHI''</tt>
 +
| : polar angle [in degrees]
 +
|}
 +
 +
<u>Notes</u>
 +
*The source angular-aperture option can be set to define the semi-aperture with respect a direction.
 +
*The option requires the definition of a unidirectional source (<tt>'''sd'''</tt>).
  
 
Source energy (<tt>'''se'''</tt>):<span id="src_se"></span>
 
Source energy (<tt>'''se'''</tt>):<span id="src_se"></span>
Line 2,429: Line 2,918:
 
{|
 
{|
 
| <tt>''E''</tt>
 
| <tt>''E''</tt>
| : energy of source particles
+
| : energy of source particles [in MeV]
 
|}
 
|}
  
Line 2,448: Line 2,937:
 
|-
 
|-
 
| <tt>''E<sub>n</sub>''</tt>
 
| <tt>''E<sub>n</sub>''</tt>
| : upper boundary of the energy bin
+
| : upper boundary of the energy bin [in MeV]
 
|-
 
|-
 
| <tt>''WGT<sub>n</sub>''</tt>
 
| <tt>''WGT<sub>n</sub>''</tt>
Line 2,467: Line 2,956:
 
|-
 
|-
 
| <tt>''MT''</tt>
 
| <tt>''MT''</tt>
| : reaction number
+
| : reaction number identifier
 
|}
 
|}
  
Line 2,481: Line 2,970:
 
{|
 
{|
 
| <tt>''T<sub>MIN</sub>''</tt>, <tt>''T<sub>MAX</sub>''</tt>
 
| <tt>''T<sub>MIN</sub>''</tt>, <tt>''T<sub>MAX</sub>''</tt>
| : time boundaries
+
| : time boundaries [in s]
 
|}
 
|}
  
Line 2,500: Line 2,989:
  
 
<u>Notes:</u>
 
<u>Notes:</u>
*Source files allow defining arbitrary distributions by reading the particle coordinates, direction, energy, weight and time from a file.
+
*Source files allow defining arbitrary distributions by reading the particle coordinates, direction, energy, weight and time from a file: [<tt> x y z u v w E wgt t </tt>] .
 
*Source files can be produced using the <tt>'''df'''</tt> entry of [[#det_df|detector cards]], or the [[#set csw|set csw]] or [[#set gsw|set gsw]] options.
 
*Source files can be produced using the <tt>'''df'''</tt> entry of [[#det_df|detector cards]], or the [[#set csw|set csw]] or [[#set gsw|set gsw]] options.
  
Line 2,525: Line 3,014:
 
{|
 
{|
 
| <tt>''MAT''</tt>
 
| <tt>''MAT''</tt>
| : material name or -1
+
| : material name or "-1" to refer to all radioactive materials
 
|-
 
|-
 
| <tt>''MODE''</tt>
 
| <tt>''MODE''</tt>
Line 2,532: Line 3,021:
  
 
<u>Notes:</u>
 
<u>Notes:</u>
*Radioactive decay source combines material compositions to decay data read from ENDF format libraries and forms the normalized source distribution automatically.
+
*Radioactive decay source combines material compositions to decay data read from ENDF format<ref name="endf" /> libraries and forms the normalized source distribution automatically.
 
*Material compositions can be defined manually, or read from binary restart files produced by a burnup or activation calculation (see the [[#set rfw|set rfw]] and [[#set rfr|set rfr]] options).
 
*Material compositions can be defined manually, or read from binary restart files produced by a burnup or activation calculation (see the [[#set rfw|set rfw]] and [[#set rfr|set rfr]] options).
*If the material name is replaced by -1, source points are started from all radioactive materials.
 
 
*The analog sampling mode preserves the average number of particles produced in radioactive decay, but may lead to poor sampling efficiency in geometries with both low and high-active materials.
 
*The analog sampling mode preserves the average number of particles produced in radioactive decay, but may lead to poor sampling efficiency in geometries with both low and high-active materials.
 
*The implicit sampling mode preserves the total statistical weight of emitted particles and produces a uniform source distribution over activated materials.
 
*The implicit sampling mode preserves the total statistical weight of emitted particles and produces a uniform source distribution over activated materials.
Line 2,543: Line 3,031:
 
=== srtrans (source transformation)<span id="srtrans"></span> ===
 
=== srtrans (source transformation)<span id="srtrans"></span> ===
  
See [[#trans (transformations)|transformations]].
+
Defines source transformations. Shortcut for "<tt>trans sr</tt>".
 +
 
 +
<u>Notes:</u>
 +
*The parameters associated with the transformation follow the standard transformation cards syntax without '''trans''' <tt>''TYPE''</tt> identifier.
 +
*See [[#trans (transformations)|transformations]].
  
 
=== strans (surface transformation)<span id="strans"></span> ===
 
=== strans (surface transformation)<span id="strans"></span> ===
  
See [[#trans (transformations)|transformations]].
+
Defines surface transformations. Shortcut for "<tt>trans s</tt>".
 +
 
 +
<u>Notes:</u>
 +
*The parameters associated with the transformation follow the standard transformation cards syntax without '''trans''' <tt>''TYPE''</tt> identifier.
 +
*See [[#trans (transformations)|transformations]].
  
 
=== surf (surface definition)<span id="surf"></span> ===
 
=== surf (surface definition)<span id="surf"></span> ===
Line 2,593: Line 3,089:
 
|-
 
|-
 
| <tt>''TEMP''</tt>
 
| <tt>''TEMP''</tt>
| : temperature to which the thermal scattering data is interpolated
+
| : temperature to which the thermal scattering data is interpolated [in K]
 
|}
 
|}
  
Line 2,622: Line 3,118:
 
|-
 
|-
 
| <tt>''LIM<sub>n</sub>''</tt>
 
| <tt>''LIM<sub>n</sub>''</tt>
| : time bin boundaries in arbitrary binning
+
| : time bin boundaries in arbitrary binning [in s]
 
|-
 
|-
 
| <tt>''T<sub>min</sub>''</tt>
 
| <tt>''T<sub>min</sub>''</tt>
| : minimum time boundary in uniform or log-uniform binning
+
| : minimum time boundary in uniform or log-uniform binning [in s]
 
|-
 
|-
 
| <tt>''T<sub>max</sub>''</tt>
 
| <tt>''T<sub>max</sub>''</tt>
| : maximum time boundary in uniform or log-uniform binning
+
| : maximum time boundary in uniform or log-uniform binning [in s]
 
|}
 
|}
  
 
<u>Notes:</u>
 
<u>Notes:</u>
  
*The entered values are in seconds
 
 
*The first limit in the arbitrary type (type = 1), is the lower bound of the first bin. The second limit is the upper bound of the first bin and so on.
 
*The first limit in the arbitrary type (type = 1), is the lower bound of the first bin. The second limit is the upper bound of the first bin and so on.
 
*Time binning is used with [[#det (detector definition)|detectors]] and [[Dynamic external source simulation mode|dynamic simulation mode]].
 
*Time binning is used with [[#det (detector definition)|detectors]] and [[Dynamic external source simulation mode|dynamic simulation mode]].
Line 2,665: Line 3,160:
 
|-
 
|-
 
| <tt>''X'',''Y'',''Z''</tt>
 
| <tt>''X'',''Y'',''Z''</tt>
| : translation vector
+
| : translation vector [in cm]
 
|-
 
|-
 
| <tt>''&theta;<sub>x</sub>'' ''&theta;<sub>y</sub>'' ''&theta;<sub>z</sub>''</tt>
 
| <tt>''&theta;<sub>x</sub>'' ''&theta;<sub>y</sub>'' ''&theta;<sub>z</sub>''</tt>
| : rotation angles with respect to x-, y- and z-axes (in degrees)
+
| : rotation angles with respect to x-, y- and z-axes [in degrees]
 
|-
 
|-
 
| <tt>''&alpha;<sub>1</sub>'' ... ''&alpha;<sub>9</sub>''</tt>
 
| <tt>''&alpha;<sub>1</sub>'' ... ''&alpha;<sub>9</sub>''</tt>
Line 2,677: Line 3,172:
 
|-
 
|-
 
| <tt>''X<sub>0</sub>'',''Y<sub>0</sub>'',''Z<sub>0</sub>''</tt>
 
| <tt>''X<sub>0</sub>'',''Y<sub>0</sub>'',''Z<sub>0</sub>''</tt>
| : Origin of vector defining rotation axis.
+
| : origin of vector defining rotation axis [in cm]
 
|-
 
|-
 
| <tt>''I'',''J'',''K''</tt>
 
| <tt>''I'',''J'',''K''</tt>
| : Components of vector defining rotation axis.
+
| : components of vector defining rotation axis.
 
|-
 
|-
 
| <tt>''&beta;''</tt>
 
| <tt>''&beta;''</tt>
| : Angle around rotation axis defined by a vector (in degrees). '''NB: In Serpent 2.1.29 positive values correspond to rotation to the negative mathematical direction and vice versa.'''
+
| : angle around rotation axis defined by a vector [in degrees].
 
|-
 
|-
 
|}
 
|}
Line 2,692: Line 3,187:
 
*Level transformation is a special type of universe transformation, in which the coordinates in the given universe are obtained relative to geometry level <tt>''LVL''</tt>.
 
*Level transformation is a special type of universe transformation, in which the coordinates in the given universe are obtained relative to geometry level <tt>''LVL''</tt>.
 
*Lattice transformation requires to provide the index for the transformation  <tt>''IDX''</tt>.
 
*Lattice transformation requires to provide the index for the transformation  <tt>''IDX''</tt>.
*Source transformation is inverted. transformation is inverted compared to how surface, universe, etc. are handled.
+
*Source transformation is inverted compared to how surface, universe, etc. are handled.
 
*By default translations are applied before rotations, and the order can be switched using the <tt>''ORD''</tt> parameter.
 
*By default translations are applied before rotations, and the order can be switched using the <tt>''ORD''</tt> parameter.
 
*Rotations can be defined either by providing the three angles with respect to the three coordinate axes, or by defining the rotation matrix. In the second case Serpent applies vector multiplication <math>\vec{r'} = \bold{A} \vec{r}</math> where <math>\vec{r}</math> and <math>\vec{r'}</math> are the position vectors before and after the operation and coefficients <tt>''&alpha;<sub>1</sub>'' ... ''&alpha;<sub>9</sub>''</tt> define the 3 by 3 matrix <math>\bold{A}</math>.
 
*Rotations can be defined either by providing the three angles with respect to the three coordinate axes, or by defining the rotation matrix. In the second case Serpent applies vector multiplication <math>\vec{r'} = \bold{A} \vec{r}</math> where <math>\vec{r}</math> and <math>\vec{r'}</math> are the position vectors before and after the operation and coefficients <tt>''&alpha;<sub>1</sub>'' ... ''&alpha;<sub>9</sub>''</tt> define the 3 by 3 matrix <math>\bold{A}</math>.
*To preserve backwards compatibility, input parameters "strans", "utrans", "ftrans", "ltrans", "dtrans" and "srtrans" without the following type identifier are also accepted for defining surface, universe, fill, lattice, detector mesh and source transformations, respectively. To preserve compatibility with Serpent 1, parameter "trans" without type identifier defines a universe transformation.
+
*In Serpent 2.1.29, a positive value of ''&beta;'' corresponds to rotation to the negative mathematical direction and vice versa.
 +
*To preserve backwards compatibility, input parameters "<tt>strans</tt>", "<tt>utrans</tt>", "<tt>ftrans</tt>", "<tt>ltrans</tt>", "<tt>dtrans</tt>" and "<tt>srtrans</tt>" without the following type identifier are also accepted for defining surface, universe, fill, lattice, detector mesh and source transformations, respectively.  
 +
*To preserve compatibility with Serpent 1, parameter "<tt>trans</tt>" without type identifier defines a universe transformation.
  
 
=== transb (burnup transformation)<span id="transb"></span> ===
 
=== transb (burnup transformation)<span id="transb"></span> ===
Line 2,705: Line 3,202:
 
{|
 
{|
 
| <tt>''STEP''</tt>
 
| <tt>''STEP''</tt>
| : step in units of burnup (positive values) or days (negative values)
+
| : depletion step (positive value = burnup [in MWd/kg], negative value = time [in d])
 
|-
 
|-
 
| <tt><''trans''></tt>
 
| <tt><''trans''></tt>
Line 2,712: Line 3,209:
  
 
<u>Notes:</u>
 
<u>Notes:</u>
*The parameters associated with the transformation follow the standard transformation cards syntax without "trans" identifier.
+
*The parameters associated with the transformation follow the standard transformation cards syntax without '''trans''' identifier.
 +
*See [[#trans (transformations)|transformations]].
 +
*Geometry plots associated with burnup transformations are featured from version 2.2.1 and on.
  
 
=== transv and transa (velocity and acceleration transformations)<span id="transa"></span><span id="transv"></span> ===
 
=== transv and transa (velocity and acceleration transformations)<span id="transa"></span><span id="transv"></span> ===
Line 2,734: Line 3,233:
 
|-
 
|-
 
| <tt>''T<sub>0</sub>''</tt>
 
| <tt>''T<sub>0</sub>''</tt>
| : beginning time of the transformation
+
| : beginning time of the transformation [in s]
 
|-
 
|-
 
| <tt>''T<sub>1</sub>''</tt>
 
| <tt>''T<sub>1</sub>''</tt>
| : end time of the transformation
+
| : end time of the transformation [in s]
 
|-
 
|-
 
| <tt>''T<sub>TYPE</sub>''</tt>
 
| <tt>''T<sub>TYPE</sub>''</tt>
Line 2,743: Line 3,242:
 
|-
 
|-
 
| <tt>''V<sub>X</sub>'',''V<sub>Y</sub>'',''V<sub>Z</sub>''</tt>
 
| <tt>''V<sub>X</sub>'',''V<sub>Y</sub>'',''V<sub>Z</sub>''</tt>
| : Initial velocity vector
+
| : initial velocity vector [in cm/s]
 
|-
 
|-
 
| <tt>''A<sub>X</sub>'',''A<sub>Y</sub>'',''A<sub>Z</sub>''</tt>
 
| <tt>''A<sub>X</sub>'',''A<sub>Y</sub>'',''A<sub>Z</sub>''</tt>
| : Initial acceleration vector
+
| : initial acceleration vector [in cm/s<sup>2</sup>]
 
|}
 
|}
  
Line 2,805: Line 3,304:
 
=== utrans (universe transformation)<span id="utrans"></span> ===
 
=== utrans (universe transformation)<span id="utrans"></span> ===
  
See [[#trans (transformations)|transformations]].
+
Defines universe transformations. Shortcut for "<tt>trans u</tt>".
 +
 
 +
<u>Notes:</u>
 +
*The parameters associated with the transformation follow the standard transformation cards syntax without '''trans''' <tt>''TYPE''</tt> identifier.
 +
*See [[#trans (transformations)|transformations]].
 +
 
 +
=== voro (stochastic Voronoi tessellation geometry definition)<span id="voro"></span> ===
 +
 
 +
'''voro''' ''UNI<sub>0</sub>'' ''UNI<sub>bg</sub>'' ''R<sub>0</sub>'' '''-1''' ''NP'' ''UNI<sub>1</sub>'' ''VF<sub>1</sub>'' [ ''UNI<sub>2</sub>'' ''VF<sub>2</sub>'' ... ]
 +
 
 +
'''voro'''  ''UNI<sub>0</sub>'' ''UNI<sub>bg</sub>'' ''R<sub>0</sub>'' ''FILE''
 +
 
 +
Defines a stochastic Voronoi tessellation geometry. Input values:
 +
 
 +
{|
 +
| <tt>''UNI<sub>0</sub>''</tt>
 +
|: universe name for the Voronoi medium
 +
|-
 +
| <tt>''UNI<sub>bg</sub>''</tt>
 +
|: background universe name filling all undefined space
 +
|-
 +
| <tt>''R<sub>0</sub>''</tt>
 +
|: test radius [in cm]
 +
|-
 +
| <tt>''NP''</tt>
 +
|: number of seed points
 +
|-
 +
| <tt>''UNI<sub>m</sub>''</tt>
 +
|: sub-universe name for the ''m''-th random fragmented polyhedral zone
 +
|-
 +
| <tt>''VF<sub>m</sub>''</tt>
 +
|: volume fraction associated to ''m''-th random fragmented polyhedral zone
 +
|-
 +
| <tt>''FILE''</tt>
 +
|: input file containing the Voronoi data
 +
|}
 +
 
 +
The <u>syntax of the file</u> containing the Voronoi seed points data is:
 +
 
 +
::{| class="toccolours" style="text-align: left;"
 +
|-
 +
| ''X<sub>1</sub>'' ''Y<sub>1</sub>'' ''Z<sub>1</sub>'' ''UNI<sub>1</sub>''
 +
|-
 +
| ''X<sub>2</sub>'' ''Y<sub>2</sub>'' ''Z<sub>2</sub>'' ''UNI<sub>1</sub>''
 +
|-
 +
| ...
 +
|-
 +
| ''X<sub>N</sub>'' ''Y<sub>N</sub>'' ''Z<sub>N</sub>'' ''UNI<sub>1</sub>''
 +
|-
 +
| ''X<sub>N+1</sub>'' ''Y<sub>N+1</sub>'' ''Z<sub>N+1</sub>'' ''UNI<sub>2</sub>''
 +
|-
 +
| ...
 +
|
 +
|}
 +
 
 +
where:
 +
{|
 +
| <tt>''X<sub>n</sub>'', ''Y<sub>n</sub>'', ''Z<sub>n</sub>''</tt>
 +
|: seed points coordinates [in cm]
 +
|-
 +
| <tt>''UNI<sub>m</sub>''</tt>
 +
|: sub-universe name for the ''m''-th random zone associated to the given seed point
 +
|}
 +
 
 +
<u>Notes:</u>
 +
*The input consists of a list of seed points and associated sub-universes filling the Voronoi cells. Alternatively, the number of seeds points and volume fractions of each zone can be provided, letting Serpent sample the positions randomly.
 +
**The advantage of the first option is that the distribution can be defined explicitly, taking into account, for example, the varying level of fragmentation closer to the boundaries.
 +
*The cell search and surfaces distances are based on search mesh and local short-list of points to reduce the computational effort. The search mesh is conditioned by the test radius, which should enclose the Voronoi polyhedral cells.
 +
**Too small radius may result in geometry errors as some points are excluded from all the search mesh cells in which they should be.
 +
**Too large radius may results in including points in cells that do not actually intersect with the polyhedral boundary.
 +
*The <tt>''DENS''</tt> parameter in the [[#set mcvol|mcvol]] input option can be switched "<tt>on</tt>" to compensate the non-preservation of the volume fractions provided as input due to the randomness of the seed points.
 +
**It applies calculated scaling factors to material densities preserving the original masses (scaling factor = volume MC routine / volume given)
  
 
=== wwgen (response matrix based importance map solver)<span id="wwgen"></span> ===
 
=== wwgen (response matrix based importance map solver)<span id="wwgen"></span> ===
Line 2,850: Line 3,420:
 
|-
 
|-
 
| <tt>''MIN<sub>n</sub>''</tt>
 
| <tt>''MIN<sub>n</sub>''</tt>
| : minimum mesh boundary (''n''th coordinate)
+
| : minimum mesh boundary (''n''-th coordinate)
 
|-
 
|-
 
| <tt>''MAX<sub>n</sub>''</tt>
 
| <tt>''MAX<sub>n</sub>''</tt>
| : maximum mesh boundary (''n''th coordinate)
+
| : maximum mesh boundary (''n''-th coordinate)
 
|-
 
|-
 
| <tt>''SZ<sub>n</sub>''</tt>
 
| <tt>''SZ<sub>n</sub>''</tt>
| : number of mesh cells (''n''th coordinate)
+
| : number of mesh cells (''n''-th coordinate)
 
|-
 
|-
 
| <tt>''LIM<sub>nm</sub>''</tt>
 
| <tt>''LIM<sub>nm</sub>''</tt>
| : mesh boundary ''m''th (''n''th coordinate)
+
| : mesh boundary ''m''-th (''n''-th coordinate)
 
|-
 
|-
 
| <tt>''X<sub>0</sub>'', ''Y<sub>0</sub>''</tt>
 
| <tt>''X<sub>0</sub>'', ''Y<sub>0</sub>''</tt>
Line 2,910: Line 3,480:
 
|}
 
|}
  
identifies the mesh. The remaining parameters are defined by separate key words followed by the input values.
+
The remaining parameters are defined by separate key words followed by the input values.
  
 
<u>Notes:</u>
 
<u>Notes:</u>
  
 
*Only works in external source simulation mode.
 
*Only works in external source simulation mode.
*Importance (weight window) meshes can be generated by running the [[#wwgen (response matrix based importance map solver)|response matrix based solver]], or read in MCNP WWINP format.
+
*Importance (weight window) meshes can be generated by running the [[#wwgen (response matrix based importance map solver)|response matrix based solver]], or read in MCNP WWINP format<ref>Kulesza, J. A. (ed.), ''“MCNP code version 6.3.0 Theory & User Manual: Appendix A Mesh-Based WWINP, WWOUT, and WWONE File Format,”'' LA-UR-22-30006, Rev. 1, Los Alamos National Laboratory [https://mcnp.lanl.gov/pdf_files/TechReport_2022_LANL_LA-UR-22-30006Rev.1_KuleszaAdamsEtAl.pdf (2022)].</ref>.
 
*Importance maps can be visualized using the [[#plot (geometry plot definition)|geometry plotter]].
 
*Importance maps can be visualized using the [[#plot (geometry plot definition)|geometry plotter]].
 
*See also [[#set wwb|set wwb]] and [[#set maxsplit|set maxsplit]] for setting options for weight windows, splitting and Russian roulette.
 
*See also [[#set wwb|set wwb]] and [[#set maxsplit|set maxsplit]] for setting options for weight windows, splitting and Russian roulette.
Line 2,921: Line 3,491:
 
*This capability is still <u>very much under development</u>. The input syntax may be revised at some point.
 
*This capability is still <u>very much under development</u>. The input syntax may be revised at some point.
  
 +
 +
<u>Weight-window mesh paramters:</u>
  
 
Mesh file (<tt>'''wf'''</tt>):<span id="wwin_wf"></span>
 
Mesh file (<tt>'''wf'''</tt>):<span id="wwin_wf"></span>
Line 2,948: Line 3,520:
 
|-
 
|-
 
| <tt>''E''</tt>  
 
| <tt>''E''</tt>  
| : energy used for renormalization
+
| : energy used for renormalization [in MeV]
 
|}
 
|}
  
Line 3,048: Line 3,620:
 
|-
 
|-
 
| <tt>''DSPL<sub>i</sub>''</tt>
 
| <tt>''DSPL<sub>i</sub>''</tt>
| : density split criterion (negative values for mass, positive values for atomic density)
+
| : density split criterion (positive value = atomic density [in b<sup>-1</sup>cm<sup>-1</sup>], negative values = mass density [in g/cm<sup>3</sup>])
 
|-
 
|-
 
| <tt>''SZ<sub>i</sub>''</tt>
 
| <tt>''SZ<sub>i</sub>''</tt>
| : minimum cell dimension
+
| : minimum cell dimension [in cm]
 
|}
 
|}
  
Line 3,058: Line 3,630:
 
*The adaptive mesh option (<tt>''ITP''</tt> = 2 or 3) starts with a coarse base mesh, and refines the resolution iteratively.
 
*The adaptive mesh option (<tt>''ITP''</tt> = 2 or 3) starts with a coarse base mesh, and refines the resolution iteratively.
 
*There are two adaptive mesh options. In the geometry-based option (<tt>''ITP''</tt> = 2) Serpent covers the geometry with <tt>''NTRK''</tt> random tracks and splits cells according to density criteria. In the tracking-based option (<tt>''ITP''</tt> = 3) the tracks are started from the source instead. The procedure is repeated <tt>''NLOOP''</tt> times.
 
*There are two adaptive mesh options. In the geometry-based option (<tt>''ITP''</tt> = 2) Serpent covers the geometry with <tt>''NTRK''</tt> random tracks and splits cells according to density criteria. In the tracking-based option (<tt>''ITP''</tt> = 3) the tracks are started from the source instead. The procedure is repeated <tt>''NLOOP''</tt> times.
*Cell splitting is defined using the <tt>''NX'', ''NY''</tt> and <tt>''NZ''</tt> options. For example <tt>''NX''</tt> = 2, <tt>''NY''</tt> = 2, <tt>''NZ''</tt> = 2 results in each cell being split to 8 sub-cells (octree mesh). For 2D meshes the <tt>''NZ''</tt> parameter must be set to 1.
+
*Cell splitting is defined using the <tt>''NX'', ''NY''</tt> and <tt>''NZ''</tt> options. For example <tt>''NX''</tt> = 2, <tt>''NY''</tt> = 2, <tt>''NZ''</tt> = 2 results in each cell being split to 8 sub-cells (octree mesh). For 2D meshes the <tt>''NZ''</tt> parameter must be set to "1".
 
*Splitting is carried out recursively, until limiting criteria are met. The <tt>''DSPL''</tt> parameters define upper density boundaries and minimum cell sizes for stopping the splits.
 
*Splitting is carried out recursively, until limiting criteria are met. The <tt>''DSPL''</tt> parameters define upper density boundaries and minimum cell sizes for stopping the splits.
 
*The importance split criterion defines the maximum relative difference between the importances of two adjacent cells. If the criterion is not met, both cells are split.
 
*The importance split criterion defines the maximum relative difference between the importances of two adjacent cells. If the criterion is not met, both cells are split.
Line 3,071: Line 3,643:
 
  '''set absrate''' ''A'' [ ''MAT'' ]
 
  '''set absrate''' ''A'' [ ''MAT'' ]
  
Sets normalization to total absorption rate.
+
Sets normalization to total absorption rate. Input values:
  
 
{|
 
{|
 
| <tt>''F''</tt>
 
| <tt>''F''</tt>
| : number of  neutrons absorbed per second (neutrons/s)
+
| : number of  neutrons absorbed per second [in neutrons/s]
 
|-
 
|-
 
| <tt>''MAT''</tt>
 
| <tt>''MAT''</tt>
Line 3,084: Line 3,656:
 
*Normalization is needed to relate the Monte Carlo reaction rate estimates to a user-given parameter.
 
*Normalization is needed to relate the Monte Carlo reaction rate estimates to a user-given parameter.
 
*Absorption includes all reactions in which the incident neutron is lost, i.e. all capture reactions and fission.
 
*Absorption includes all reactions in which the incident neutron is lost, i.e. all capture reactions and fission.
*Neutron transport simulations are by default normalized to unit total loss rate.
+
*The default normalization:
*Photon transport simulations are by default normalized to unit total source rate.
+
**It is set to unit total loss rate (neutron transport) and to unit total source rate (photon transport).
 +
**In coupled neutron-photon transport simulations the normalization is driven by the neutron normalization.
 
*For other normalization options, see: [[#set power|set power]], [[#set powdens|set powdens]], [[#set flux|set flux]], [[#set genrate|set genrate]], [[#set fissrate|set fissrate]], [[#set lossrate|set lossrate]], [[#set srcrate|set srcrate]], [[#set sfrate|set sfrate]].
 
*For other normalization options, see: [[#set power|set power]], [[#set powdens|set powdens]], [[#set flux|set flux]], [[#set genrate|set genrate]], [[#set fissrate|set fissrate]], [[#set lossrate|set lossrate]], [[#set srcrate|set srcrate]], [[#set sfrate|set sfrate]].
 
*See also Section 5.8 of [http://montecarlo.vtt.fi/download/Serpent_manual.pdf Serpent 1 User Manual].
 
*See also Section 5.8 of [http://montecarlo.vtt.fi/download/Serpent_manual.pdf Serpent 1 User Manual].
Line 3,108: Line 3,681:
  
 
  '''set adf''' ''UNI SURF SYM [ENF]''  
 
  '''set adf''' ''UNI SURF SYM [ENF]''  
Sets parameters for the calculation of assembly discontinuity factors (ADFs). Input values:
+
Sets parameters for the calculation of assembly discontinuity factors (ADFs) and related net and partial currents. Input values:
  
 
{|
 
{|
Line 3,121: Line 3,694:
 
|-
 
|-
 
| <tt>''ENF''</tt>
 
| <tt>''ENF''</tt>
| : option to skip diffusion flux solver and enforce flat homogeneous flux distribution based on mean heterogeneous flux (1/yes). Default is (0/no).
+
| : option to switch on (1/yes) or off (0/no) skipping the diffusion flux solver and enforce flat homogeneous flux distribution based on mean heterogeneous flux. The default option is "<tt>off</tt>"
 
|}
 
|}
  
Line 3,127: Line 3,700:
  
 
*The surface enclosing the universe can be super-imposed (i.e. not part of the geometry definition), but it must enclose the <u>entire</u> universe.
 
*The surface enclosing the universe can be super-imposed (i.e. not part of the geometry definition), but it must enclose the <u>entire</u> universe.
*The surface is super-imposed on the geometry, i.e. its parameters (coordinates) are relative to the [[Input_syntax_manual#set_root|root universe]] (default 0)
+
*The surface is super-imposed on the geometry, i.e. its parameters (coordinates) are relative to the [[Input_syntax_manual#set_root|root universe]].
 
*When the universe is surrounded by zero net-current (reflective) boundary conditions, the ADFs are calculated as the ratios of surface- and volume-averaged heterogeneous flux.  
 
*When the universe is surrounded by zero net-current (reflective) boundary conditions, the ADFs are calculated as the ratios of surface- and volume-averaged heterogeneous flux.  
 
*When the net current is non-zero, the calculation is based on the ratio of surface-averaged homogeneous and heterogeneous flux. The homogeneous flux is obtained from a [[Diffusion flux solver|built-in diffusion flux solver]].
 
*When the net current is non-zero, the calculation is based on the ratio of surface-averaged homogeneous and heterogeneous flux. The homogeneous flux is obtained from a [[Diffusion flux solver|built-in diffusion flux solver]].
Line 3,137: Line 3,710:
 
*ADFs are calculated in the [[#set nfg|few-group structure used for group constant generation]].
 
*ADFs are calculated in the [[#set nfg|few-group structure used for group constant generation]].
 
*The <tt>''ENF''</tt> parameter should be switched on only in rare cases (and you should know what you are doing).
 
*The <tt>''ENF''</tt> parameter should be switched on only in rare cases (and you should know what you are doing).
 +
*Sign moments of net and partial currents are not scored for [[Surface_types#Regular_prisms|Y-type infinite/truncated hexagonal prisms]].
  
 
=== set alb ===
 
=== set alb ===
Line 3,158: Line 3,732:
 
*When this option is set, Serpent calculates both total albedos (ratio of currents) and partial albedos (response matrix).
 
*When this option is set, Serpent calculates both total albedos (ratio of currents) and partial albedos (response matrix).
 
*The surface enclosing the universe can be super-imposed (i.e. not part of the geometry definition), but it must enclose the <u>entire</u> universe.
 
*The surface enclosing the universe can be super-imposed (i.e. not part of the geometry definition), but it must enclose the <u>entire</u> universe.
*The surface is super-imposed on the geometry, i.e. its parameters (coordinates) are relative to the [[Input_syntax_manual#set_root|root universe]] (default 0)
+
*The surface is super-imposed on the geometry, i.e. its parameters (coordinates) are relative to the [[Input_syntax_manual#set_root|root universe]]
 
*The current direction is given relative to the surface normal vectors
 
*The current direction is given relative to the surface normal vectors
 
*The universe is needed only for labelling the results in the output files.
 
*The universe is needed only for labelling the results in the output files.
Line 3,171: Line 3,745:
 
{|
 
{|
 
| <tt>''MODEN''</tt>  
 
| <tt>''MODEN''</tt>  
| : mode for neutrons (0 = no reactions included, 1 = include only reactions that affect neutron balance, 2 = include all reactions)
+
| : mode for neutrons (0 = no reactions included, 1 = include only reactions that affect neutron balance, 2 = include all reactions). (default value: 0)
 
|-
 
|-
 
| <tt>''MODEG''</tt>
 
| <tt>''MODEG''</tt>
| : mode for photons (0 = no reactions included, 1 = include all reactions)
+
| : mode for photons (0 = no reactions included, 1 = include all reactions). (default value: 0)
 
|}
 
|}
  
 
<u>Notes:</u>
 
<u>Notes:</u>
  
*Analog reaction rates are calculated by counting sampled events and printed in a separate output file.
+
*Analog reaction rates are calculated by counting sampled events and printed in a separate output file <tt>[input]_arr[bu].m</tt>, where "<tt>bu</tt>" is the burnup step.
 
*See detailed description on the [[Description of output files#Reaction rate output|reaction rate output file]].
 
*See detailed description on the [[Description of output files#Reaction rate output|reaction rate output file]].
  
Line 3,268: Line 3,842:
 
<u>Notes:</u>
 
<u>Notes:</u>
 
*The boundary conditions can be set either for all directions at once (single parameter) or x-, y- and z-directions separately (three parameters). Albedos are provided by adding one more parameter in the list.
 
*The boundary conditions can be set either for all directions at once (single parameter) or x-, y- and z-directions separately (three parameters). Albedos are provided by adding one more parameter in the list.
*The default boundary condition is vacuum (= 1) in all directions.
+
*The boundary condition type numbers can also be given as strings, with "black" = 1, "reflective" = 2 and "periodic" = 3.
 +
*The default boundary condition is "vacuum" (= 1) in all directions.
 
*Albedo boundary conditions are invoked by multiplying the particle weight with factor <tt>''ALB''</tt> each time a reflective or periodic boundary is hit.
 
*Albedo boundary conditions are invoked by multiplying the particle weight with factor <tt>''ALB''</tt> each time a reflective or periodic boundary is hit.
 
*Repeated boundary conditions (reflective or periodic) are based on universe transformations, which limits outer boundary to surfaces that form regular lattices (square and hexagonal prisms, rectangles, cubes and cuboids).
 
*Repeated boundary conditions (reflective or periodic) are based on universe transformations, which limits outer boundary to surfaces that form regular lattices (square and hexagonal prisms, rectangles, cubes and cuboids).
Line 3,274: Line 3,849:
 
*For symmetry purposes Serpent provides the [[#set usym|universe symmetry option]].
 
*For symmetry purposes Serpent provides the [[#set usym|universe symmetry option]].
 
*For more information, see [[Boundary conditions|detailed description on boundary conditions]].
 
*For more information, see [[Boundary conditions|detailed description on boundary conditions]].
*The boundary condition type numbers can also be given as strings, with black = 1, reflective = 2 and periodic = 3.
 
  
 
=== set blockdt ===
 
=== set blockdt ===
Line 3,304: Line 3,878:
  
 
<u>Notes:</u>
 
<u>Notes:</u>
 
*Isomeric branching data libraries are standard ENDF format files containing energy-dependent branching ratios. The data is read from ENDF files 9 and 10.
 
*Serpent uses [[Default isomeric branching ratios|constant branching ratios]] by default. The default values can be overridden using the [[#set isobra|set isobra]] option. Energy-dependent data read read from ENDF format files overrides the constant ratios.
 
 
*If the file path contains special characters it is advised to enclose it within quotes.
 
*If the file path contains special characters it is advised to enclose it within quotes.
 
*A default directory path can be set by defining environment variable SERPENT_DATA. The code looks for decay data files in this path if not found at the absolute.
 
*A default directory path can be set by defining environment variable SERPENT_DATA. The code looks for decay data files in this path if not found at the absolute.
*See also: [[Branching ratio example|example input]]
+
*Isomeric branching data libraries are standard ENDF format<ref name="endf">Trkov, A., Herman, M. and Brown, D. A. ''"ENDF-6 Formats Manual."'' CSEWG Document ENDF-102 / BNL-90365-2009 Rev. 2 [https://www.nndc.bnl.gov/endf-b8.0/endf-manual-viii.0.pdf (2018)]</ref> files containing energy-dependent branching ratios. The data is read from ENDF files 9 and 10.
*Example data from the [http://virtual.vtt.fi/virtual/montecarlo/misc/sss_jeff31a.bra JEFF-3.1 activation file]
+
*Serpent uses [[Default isomeric branching ratios|constant branching ratios]] by default. The default values can be overridden using the [[#set isobra|set isobra]] option. Energy-dependent data read from ENDF format files overrides the constant ratios.
 +
*See a practical example on how to evaluate branching ratios: [[Branching ratio example|example input]], where the isomeric branching  data library corresponds to the JEFF-3.1 activation file [[File:JEFF-3.1_activation_file.tgz]].
  
 
=== set branchless ===
 
=== set branchless ===
Line 3,319: Line 3,891:
 
{|
 
{|
 
| <tt>''OPT''</tt>
 
| <tt>''OPT''</tt>
| :  option to switch calculation on (1/yes) or off (0/no). Default is off.
+
| :  option to switch calculation on (1/yes) or off (0/no). The default option is "<tt>off</tt>".
 
|-
 
|-
 
| <tt>''WGT_LOW''</tt>
 
| <tt>''WGT_LOW''</tt>
Line 3,331: Line 3,903:
 
*The branchless algorithm suppresses the variability due to the simultaneous propagation of the several branches associated to a fission event
 
*The branchless algorithm suppresses the variability due to the simultaneous propagation of the several branches associated to a fission event
 
*The branchless method uses analog scattering combined with forced fission so that after each collision, the neutron is either a scattering neutron or a fission neutron. In a non-multiplying method, the branchless method behaves as implicit capture.
 
*The branchless method uses analog scattering combined with forced fission so that after each collision, the neutron is either a scattering neutron or a fission neutron. In a non-multiplying method, the branchless method behaves as implicit capture.
 +
*The branchless method sets the following simulation configuration: reaction sampling ([[#set nphys|set nphys 1 1 1]]), reaction modes ([[#set impl|set impl 0 1 1]]), and population control ([[#set combing|set combing 1]]), overriding any user-defined option.
 +
*The current implementation does not support the use of the branchless collision method combined with the unresolved resonance probability table sampling (see [[#set ures|set ures]]).
  
 
=== set bumode ===
 
=== set bumode ===
Line 3,340: Line 3,914:
 
{|
 
{|
 
| <tt>''MODE''</tt>
 
| <tt>''MODE''</tt>
| : burnup calculation mode
+
| : burnup calculation mode (default value: 2 = CRAM)
 
|-
 
|-
 
| <tt>''ORDER''</tt>
 
| <tt>''ORDER''</tt>
| : CRAM order
+
| : CRAM order (default value: 14 - 14 PFD CRAM)
 
|-
 
|-
 
| <tt>''SSD''</tt>
 
| <tt>''SSD''</tt>
| : number of substeps for CRAM decay steps (default or 0: use TTA)
+
| : number of substeps for CRAM decay steps (default value: 0 = use TTA)
 
|}
 
|}
  
 
The possible settings for mode are:
 
The possible settings for mode are:
  
{| class="wikitable" style="text-align: left;"
+
::{| class="wikitable" style="text-align: left;"
 
! Mode
 
! Mode
 
! Description
 
! Description
Line 3,364: Line 3,938:
 
The CRAM order parameter can only be given when choosing the CRAM mode. The possible settings for CRAM order are:
 
The CRAM order parameter can only be given when choosing the CRAM mode. The possible settings for CRAM order are:
  
{| class="wikitable" style="text-align: left;"
+
::{| class="wikitable" style="text-align: left;"
 
! CRAM order
 
! CRAM order
|-
 
 
| <tt>2</tt>
 
| <tt>2</tt>
|-
 
 
| <tt>4</tt>
 
| <tt>4</tt>
|-
 
 
| <tt>6</tt>
 
| <tt>6</tt>
|-
 
 
| <tt>8</tt>
 
| <tt>8</tt>
|-
 
 
| <tt>10</tt>
 
| <tt>10</tt>
|-
 
 
| <tt>12</tt>
 
| <tt>12</tt>
|-
 
 
| <tt>14</tt>
 
| <tt>14</tt>
|-
 
 
| <tt>16</tt>
 
| <tt>16</tt>
|-
 
 
| <tt>-16</tt>
 
| <tt>-16</tt>
|-
 
 
| <tt>-48</tt>
 
| <tt>-48</tt>
 
|}
 
|}
  
 
<u>Notes:</u>
 
<u>Notes:</u>
*The default setting for the burnup calculation mode is CRAM.
+
*Positive values refer to PFD form of CRAM. Negative values of CRAM order mean using IPF form of CRAM with order of the absolute value of the parameter.
*Default value for the CRAM order is 14 resulting in order 14 PFD CRAM.
+
*Negative values of CRAM order mean using IPF form of CRAM with order of the absolute value of the parameter.
+
*Positive values refer to PFD form of CRAM.
+
 
*Decay calculations (see [[#dep (depletion history)|dep (depletion history)]]) and burnup calculations with very low flux are always calculated with TTA disregarding this input before version 2.1.32. The latter, very low flux condition, only applies to calculations not involving continuous reprocessing.
 
*Decay calculations (see [[#dep (depletion history)|dep (depletion history)]]) and burnup calculations with very low flux are always calculated with TTA disregarding this input before version 2.1.32. The latter, very low flux condition, only applies to calculations not involving continuous reprocessing.
*Positive values of <tt>''SSD''</tt> enforce usage of CRAM with given number of substeps.
+
*Positive values of <tt>''SSD''</tt> enforce usage of CRAM with given number of substeps. A zero value of <tt>''SSD''</tt> enforces usage of TTA.
 
*The Serpent 1 <tt>''MODE''</tt> 3, a variation TTA method, in which cyclic transmutation chains are handled by inducing small variations in the coefficients instead of solving the extended TTA equations, is overwritten by the standard TTA method <tt>''MODE''</tt> 1.
 
*The Serpent 1 <tt>''MODE''</tt> 3, a variation TTA method, in which cyclic transmutation chains are handled by inducing small variations in the coefficients instead of solving the extended TTA equations, is overwritten by the standard TTA method <tt>''MODE''</tt> 1.
 
*Version 2.2.0 includes the sub-step method for depletion calculations involving continuous reprocessing.
 
*Version 2.2.0 includes the sub-step method for depletion calculations involving continuous reprocessing.
Line 3,406: Line 3,967:
 
{|
 
{|
 
| <tt>''NORM''</tt>
 
| <tt>''NORM''</tt>
| : burnup calculation normalization mode (1 = all materials, 2 = burnable materials, 3 = non-burnable materials)
+
| : burnup calculation normalization mode (1 = all materials, 2 = burnable materials, 3 = non-burnable materials). (default value: 1)
 
|}
 
|}
 
<u>Notes:</u>
 
* The default normalization for burnup calculations includes all materials.
 
  
 
=== set ccmaxiter ===
 
=== set ccmaxiter ===
Line 3,419: Line 3,977:
 
{|
 
{|
 
| <tt>''NITER''</tt>
 
| <tt>''NITER''</tt>
| : number of iterations.
+
| : number of iterations (default value: 1 = no iteration)
 
|}
 
|}
  
 
<u>Notes:</u>
 
<u>Notes:</u>
  
*Default maximum number of iterations is 1 (no iteration).
 
 
*The iteration is stopped when either the maximum number of iterations or the maximum active neutron population (set with [[#set ccmaxpop|set ccmaxpop]]) has been simulated.
 
*The iteration is stopped when either the maximum number of iterations or the maximum active neutron population (set with [[#set ccmaxpop|set ccmaxpop]]) has been simulated.
 
*See [[Coupled multi-physics calculations]] for further information.
 
*See [[Coupled multi-physics calculations]] for further information.
Line 3,435: Line 3,992:
 
{|
 
{|
 
| <tt>''CPOP''</tt>
 
| <tt>''CPOP''</tt>
| : total active population to simulate.
+
| : total active population to simulate (default value: [[Definitions, units and constants#Constants|INFTY]]/1E6)
 
|}
 
|}
  
 
<u>Notes:</u>
 
<u>Notes:</u>
  
*Default maximum population is [[Definitions, units and constants#Constants|INFTY]]/1e6.
 
 
*The iteration is stopped when either the maximum number of iterations (set with [[#set ccmaxiter|set ccmaxiter]]) or the maximum active neutron population has been simulated.
 
*The iteration is stopped when either the maximum number of iterations (set with [[#set ccmaxiter|set ccmaxiter]]) or the maximum active neutron population has been simulated.
 
*Only the population simulated during active cycles is included in this amount.
 
*Only the population simulated during active cycles is included in this amount.
Line 3,453: Line 4,009:
 
{|
 
{|
 
| <tt>''OPT''</tt>
 
| <tt>''OPT''</tt>
| : option to set Doppler broadening method off (0/no) or on (1/yes). The default option is on.
+
| : option to set Doppler broadening method off (0/no) or on (1/yes). The default option is "<tt>on</tt>".
 
|}
 
|}
  
Line 3,467: Line 4,023:
 
{|
 
{|
 
| <tt>''OPT''</tt>
 
| <tt>''OPT''</tt>
| : option to set the Compton electron angular distribution model off (0/no) or on (1/yes). The default option is on.
+
| : option to set the Compton electron angular distribution model off (0/no) or on (1/yes). The default option is "<tt>on</tt>".
 
|}
 
|}
  
Line 3,480: Line 4,036:
 
{|
 
{|
 
| <tt>''LN''</tt>
 
| <tt>''LN''</tt>
| : minimum mean distance for scoring the CFE for neutrons
+
| : minimum mean distance for scoring the CFE for neutrons [in cm] (default value: 20.0)
 
|-
 
|-
 
| <tt>''TN''</tt>
 
| <tt>''TN''</tt>
| : minimum mean time interval for scoring the CFE for neutrons
+
| : minimum mean time interval for scoring the CFE for neutrons [in s]
 
|-
 
|-
 
| <tt>''LG''</tt>
 
| <tt>''LG''</tt>
| : minimum mean distance for scoring the CFE for photons
+
| : minimum mean distance for scoring the CFE for photons [in cm] (default value: 20.0)
 
|-
 
|-
 
| <tt>''TG''</tt>
 
| <tt>''TG''</tt>
| : minimum mean time interval for scoring the CFE for photons
+
| : minimum mean time interval for scoring the CFE for photons [in s]
 
|}
 
|}
  
Line 3,496: Line 4,052:
 
*The minimum mean distance is the statistical mean-free-path (mfp) of collisions that contribute to the CFE. Collisions are more frequent if the physical mfp is shorter.
 
*The minimum mean distance is the statistical mean-free-path (mfp) of collisions that contribute to the CFE. Collisions are more frequent if the physical mfp is shorter.
 
*In time-dependent simulations it may be more convenient to define the minimum mean time between two collisions, to get sufficient statistics for short time bins.
 
*In time-dependent simulations it may be more convenient to define the minimum mean time between two collisions, to get sufficient statistics for short time bins.
*The default minimum mean scoring distance is 20 cm for both neutrons and photons. Adjusting the distance affects both statistics and running time, but it should be noted that <u>no studies have been performed on what the optimal value should be</u>.
+
*Adjusting the distance affects both statistics and running time, but it should be noted that <u>no studies have been performed on what the optimal value should be</u>.
*Only one criterion can be provided for each particle type. If distance is given, time must be set to -1 and vice versa.
+
*Only one criterion can be provided for each particle type. If distance is given, time must be set to "-1" and vice versa.
 
*For more information on tracking modes and CFE, see the detailed descriptions on [[delta- and surface-tracking]] and [[Result estimators#Implicit estimators|result estimators]].
 
*For more information on tracking modes and CFE, see the detailed descriptions on [[delta- and surface-tracking]] and [[Result estimators#Implicit estimators|result estimators]].
 
*The collision flux estimator in Serpent is described in an article in Annals of Nuclear energy from 2017.<ref name="cfe">Leppänen, J.
 
*The collision flux estimator in Serpent is described in an article in Annals of Nuclear energy from 2017.<ref name="cfe">Leppänen, J.
Line 3,529: Line 4,085:
 
{|
 
{|
 
| <tt>''FMT''</tt>
 
| <tt>''FMT''</tt>
| : output format, currently used for including or excluding statistical errors (0 = not included, 1 = included)
+
| : output format, currently used for including or excluding statistical errors (0 = not included, 1 = included). (default value: 0)
 
|-
 
|-
 
| <tt>''PARAM<sub>n</sub>''</tt>
 
| <tt>''PARAM<sub>n</sub>''</tt>
Line 3,548: Line 4,104:
 
{|
 
{|
 
| <tt>''MODE''</tt>
 
| <tt>''MODE''</tt>
| : combing population-control mode (0 = none, 1 = weight-based, 2 = emission-based)
+
| : combing population-control mode (0 = none, 1 = weight-based, 2 = emission-based). (default value: 0)
 
|}
 
|}
  
Line 3,580: Line 4,136:
 
{|
 
{|
 
| <tt>''OPT''</tt>
 
| <tt>''OPT''</tt>
| : option to set confidentiality flag on (1/yes) or off (0/no)
+
| : option to set confidentiality flag on (1/yes) or off (0/no). The default option is "<tt>off</tt>"
 
|}
 
|}
  
 
<u>Notes:</u>
 
<u>Notes:</u>
  
*This option can be used to label calculations as confidential. If the option is set, text "(CONFIDENTIAL)" is printed in the run-time output next to the [[#set title|calculation title]] and the value of variable CONFIDENTIAL_DATA in the <tt>[input]_res.m</tt> output file is set to 1.
+
*This option can be used to label calculations as confidential. If the option is set, text "(CONFIDENTIAL)" is printed in the run-time output next to the [[#set title|calculation title]] and the value of variable CONFIDENTIAL_DATA in the <tt>[input]_res.m</tt> output file is set to "1".
  
 
=== set coverxlib ===
 
=== set coverxlib ===
Line 3,594: Line 4,150:
 
{|
 
{|
 
| <tt>''LIB<sub>n</sub>''</tt>
 
| <tt>''LIB<sub>n</sub>''</tt>
| : file paths to multi-group covariance data files in the COVERX format (ASCII or binary)
+
| : file paths to multi-group covariance data files in the COVERX format<ref name="coverx">Wieselquist, W. A. and Lefebvre, R. A (ed.), ''"SCALE 6.3.1 User Manual: Sensitivity and Uncertainty Analysis - Appendix 6.3.4.1.6. COVERX format"'', ORNL/TM-SCALE-6.3.1, UT-Battelle, LLC, Oak Ridge National Laboratory, Oak Ridge, TN [https://scale-manual.ornl.gov/tsunami-ip-appAB.html#coverx-format (2023)]</ref>  (ASCII or binary)
 
|}
 
|}
  
Line 3,611: Line 4,167:
 
|}
 
|}
  
<u>Notes:</u>
+
The <u>syntax of the file</u> containing the covariance data is:
  
 +
::{| class="toccolours" style="text-align: left;"
 +
|-
 +
| ''NG'' ''E<sub>1</sub>'' ... ''E<sub>NG+1</sub>'' ''NM''
 +
|-
 +
| ''ZAI<sub>1,1</sub>'' ''MT<sub>1,1</sub>'' ''ZAI<sub>1,2</sub>'' ''MT<sub>1,2</sub>''
 +
|-
 +
| ''COV<sub>1,1,1</sub>'' ... ''COV<sub>1,NG,NG</sub>''
 +
|-
 +
|...
 +
|-
 +
| ''ZAI<sub>NM,1</sub>'' ''MT<sub>NM,1</sub>'' ''ZAI<sub>NM,2</sub>'' ''MT<sub>NM,2</sub>''
 +
|-
 +
| ''COV<sub>NM,1,1</sub>'' ... ''COV<sub>NM,NG,NG</sub>''
 +
|-
 +
|}
 +
 +
where:
 +
 +
{|
 +
| <tt>''NG''</tt>
 +
|: number of neutron energy groups
 +
|-
 +
| <tt>''E<sub>g</sub>''</tt>
 +
|: energy grid boundaries [in MeV]
 +
|-
 +
| <tt>''NM''</tt>
 +
|: number of covariance matrixes
 +
|-
 +
| <tt>''ZAI<sub>m,n</sub>'', ''MT<sub>m,n</sub>''</tt>
 +
|: 2 &times; nuclide-reaction pairs defining the ''m''-th covariance matrix
 +
|-
 +
| <tt>''COV''<sub>m,g,g</sub>''</tt>
 +
|: ''NG'' &times; ''NG'' covariance data corresponding to the ''m''-th matrix
 +
|}
 +
 +
<u>Notes:</u>
 
*If covariance data is linked when running [[Sensitivity calculations]], Serpent will automatically apply the sandwich rule using the calculated sensitivity vectors and propagate the covariance data to uncertainties of the sensitivity responses.
 
*If covariance data is linked when running [[Sensitivity calculations]], Serpent will automatically apply the sandwich rule using the calculated sensitivity vectors and propagate the covariance data to uncertainties of the sensitivity responses.
  
Line 3,622: Line 4,214:
 
{|
 
{|
 
| <tt>''DEPTH''</tt>
 
| <tt>''DEPTH''</tt>
| : The number of levels included. 1 is the first lattice calculated from universe 0 usually corresponding to assembly-wise distribution. 2 includes the first two levels usually corresponding to the assembly- and pin-wise distributions.
+
| : The number of lattice-levels included.  
 
|-
 
|-
 
| <tt>''N<sub>Z</sub>''</tt>
 
| <tt>''N<sub>Z</sub>''</tt>
| : Number of equal sized axial bins into which the lattices are divided.
+
| : Number of equal sized axial bins into which the lattices are divided (default value: 1)
 
|-
 
|-
 
| <tt>''Z<sub>MIN</sub>''</tt>
 
| <tt>''Z<sub>MIN</sub>''</tt>
| : Minimum z-coordinate for the axial division.
+
| : Minimum z-coordinate for the axial division [in cm] (default value: [[Definitions, units and constants#Constants|-INFTY]])
 
|-
 
|-
 
| <tt>''Z<sub>MAX</sub>''</tt>
 
| <tt>''Z<sub>MAX</sub>''</tt>
| : Maximum z-coordinate for the axial division.
+
| : Maximum z-coordinate for the axial division [in cm] (default value: [[Definitions, units and constants#Constants|INFTY]])
 
|-
 
|-
 
| <tt>''LVL1''</tt>
 
| <tt>''LVL1''</tt>
Line 3,641: Line 4,233:
  
 
<u>Notes:</u>
 
<u>Notes:</u>
* The default values for <tt>''N<sub>Z</sub>''</tt>, <tt>''Z<sub>MIN</sub>''</tt> and <tt>''Z<sub>MAX</sub>''</tt> are 1, -INFTY and INFTY, respectively.
+
*The interpretation of the number of levels included is as follows:
 +
** <tt>''DEPTH''</tt> 1: includes the first level from the root universe, which "usually" corresponds to the assembly-wise distribution.
 +
** <tt>''DEPTH''</tt> 2:  includes the first two levels from the root universe, which "usually" corresponds to the assembly- and pin-wise distributions.
  
 
=== set cpop ===
 
=== set cpop ===
  
 
  '''set cpop''' ''NPG'' ''NGEN'' ''NSKIP'' [ ''NSKIP2'' ]
 
  '''set cpop''' ''NPG'' ''NGEN'' ''NSKIP'' [ ''NSKIP2'' ]
Sets parameters for simulated neutron population for corrector neutron transport solutions in burnup calculation. Typically used with the [[Stochastic Implicit Euler burnup scheme|SIE burnup scheme]]. Input values
+
Sets parameters for simulated neutron population for corrector neutron transport solutions in burnup calculation. Typically used with the [[Stochastic Implicit Euler burnup scheme|SIE burnup scheme]]. Input values:
  
 
{|
 
{|
Line 3,679: Line 4,273:
  
 
*Only source points from active cycles are included.
 
*Only source points from active cycles are included.
 +
*From version 2.2.1 and on, multi-step depletion source files can be generated <tt>[''FILE'']_[bu]</tt>, where "<tt>bu</tt>" is the burnup step. Otherwise, simply, <tt>[''FILE'']</tt>.
  
 
=== set dataout ===
 
=== set dataout ===
  
 
  '''set dataout''' ''TABLE_LIST''
 
  '''set dataout''' ''TABLE_LIST''
Defines the tables included in the nuclear and material data file <tt>[input].out.m</tt>. Input values:
+
Defines the tables included in the nuclear and material data file <tt>[input].out</tt>. Input values:
  
 
{|
 
{|
 
| <tt>''TABLE_LIST''</tt>
 
| <tt>''TABLE_LIST''</tt>
| : list of tables
+
| : list of tables (default value: <tt>all/0</tt>)
 +
|}
 +
 
 +
Possible list of tables:
 +
Possible key-words/variables are:
 +
::{| class="wikitable" style="text-align: left;"
 +
! Key-word
 +
! Table ID
 +
! Description
 +
|-
 +
| <tt>0, all</tt>
 +
|
 +
| include all available tables
 +
|-
 +
| <tt>1, nuc_summary</tt>
 +
| Table 1: Summary of nuclide data
 +
|
 +
|-
 +
| <tt>2, nuc_readec</tt>
 +
| Table 2: Reaction and decay data
 +
|
 +
|-
 +
| <tt>3, nuc_nfy</tt>
 +
| Table 3: Fission yield data
 +
| only in burnp mode
 +
|-
 +
| <tt>4, nuc_lostpath</tt>
 +
| Table 4: Lost transmutation paths
 +
| only in burnup mode
 +
|-
 +
| <tt>5, mat_summary</tt>
 +
| Table 1: Summary of material compositions
 +
|
 +
|-
 +
| <tt>8, allnuc</tt>
 +
|
 +
| (nuclide) Tables 1-4
 +
|-
 +
| <tt>9, allmat</tt>
 +
|
 +
| (material) Tables 1
 +
|-
 +
| <tt>-1</tt>
 +
|
 +
| omit the <tt>[input].out</tt> file
 
|}
 
|}
  
 
<u>Notes:</u>
 
<u>Notes:</u>
*The file lists all nuclides and their reactions as they are read from the nuclear data libraries. The material data includes isotopic compositions and densities, as well as volumes and masses if available. (See the [[Description of output files#Nuclide and material data output|description of the nuclear and material data output]]).
+
*The output file data is divided into two sections: nuclear data (Tables 1-4) and material data (Table 1). Respectively, they include all the nuclides and their reactions as they are read from the nuclear data libraries, and the material data includes isotopic compositions and densities, as well as volumes and masses if available.  
*The data is divided into two sections: nuclear data (Tables 1-4) and material data (Table 5), and the tables are the following:
+
*For more information, see detailed description of the [[Description of output files#Nuclide and material data output|nuclear and material data output]].
**Table 1: Summary of nuclide data
+
**Table 2: Reaction and decay data
+
**Table 3: Fission yield data (only in burnup mode)
+
**Table 4: Lost transmutation paths (only in burnup mode)
+
**Table 5: Summary of material compositions
+
*The entry-names to input the tables are: "all"/0 = Tables 1-5, "nuc_summary"/1 = Tables 1, "nuc_readec"/2 = Table 2, "nuc_nfy"/3 = Table 3, "nuc_lostpath"/4 = Table 4, "mat_summary"/5 = Table 5, "allnuc"/8 = Tables 1-4, "allmat"/9 = Table 5.
+
*The default behavior is to include all the tables/data available. To omit the <tt>[input].out</tt> file, select "-1".
+
  
 
=== set dbrc ===
 
=== set dbrc ===
Line 3,708: Line 4,340:
 
{|
 
{|
 
| <tt>''E<sub>min</sub>''</tt>
 
| <tt>''E<sub>min</sub>''</tt>
| : Minimum energy for DBRC
+
| : Minimum energy for DBRC [in MeV]
 
|-
 
|-
 
| <tt>''E<sub>max</sub>''</tt>
 
| <tt>''E<sub>max</sub>''</tt>
| : Maximum energy for DBRC
+
| : Maximum energy for DBRC [in MeV]
 
|-
 
|-
 
| <tt>''NUC<sub>n</sub>''</tt>
 
| <tt>''NUC<sub>n</sub>''</tt>
| : Nuclide identifiers for which to apply DBRC to, with 0 K cross section data, e.g. "92238.00c"
+
| : zero-kelvin nuclide identifiers for which to apply DBRC (e.g. "92238.00c")
 
|}
 
|}
  
 
<u>Notes:</u>
 
<u>Notes:</u>
*This description is not complete.
 
 
*Use of DBRC requires 0 K cross section data.
 
*Use of DBRC requires 0 K cross section data.
 
*See also Section 5.6 of [http://montecarlo.vtt.fi/download/Serpent_manual.pdf Serpent 1 User Manual].
 
*See also Section 5.6 of [http://montecarlo.vtt.fi/download/Serpent_manual.pdf Serpent 1 User Manual].
Line 3,726: Line 4,357:
  
 
  '''set dd''' ''MODE'' [ ''X<sub>0</sub>'' ''Y<sub>0</sub>'' ''&alpha;<sub>0</sub>'' ]
 
  '''set dd''' ''MODE'' [ ''X<sub>0</sub>'' ''Y<sub>0</sub>'' ''&alpha;<sub>0</sub>'' ]
Invokes domain decomposition. Input values
+
Invokes domain decomposition. Input values:
  
 
{|
 
{|
 
| <tt>''MODE''</tt>
 
| <tt>''MODE''</tt>
| : decomposition mode (0 = none, 1 = simple, 2 = sector, 3 = sector + center)
+
| : decomposition mode (default value: 0)
 
|-  
 
|-  
 
| <tt>''X<sub>0</sub>''</tt>
 
| <tt>''X<sub>0</sub>''</tt>
| x-coordinate of the domain decomposition origin (origin is the centre of the angular division)
+
| x-coordinate of the domain decomposition origin (centre of the radial division, initial position of the angular division) [in cm] (default value: 0.0)
 
|-
 
|-
 
| <tt>''Y<sub>0</sub>''</tt>
 
| <tt>''Y<sub>0</sub>''</tt>
| y-coordinate of the domain decomposition origin (origin is the centre of the angular division)
+
| y-coordinate of the domain decomposition origin (centre of the radial division, initial position of the angular division) [in cm] (default value: 0.0)
 
|-
 
|-
 
| <tt>''&alpha;<sub>0</sub>''</tt>
 
| <tt>''&alpha;<sub>0</sub>''</tt>
| angular position of the domain decomposition origin (origin is the initial position of the angular division)  
+
| angular position of the domain decomposition origin [in degrees] (default value: 0.0)
 +
|}
 +
 
 +
The possible modes are:
 +
 
 +
::{| class="wikitable" style="text-align: left;"
 +
! Mode
 +
! Description
 +
! Notes
 +
|-
 +
| <tt>0</tt>
 +
| none
 +
|
 +
|-
 +
| <tt>1</tt>
 +
| depletion zone indexing-based decomposition
 +
| not recommended
 +
|-
 +
| <tt>2</tt>
 +
| sector-based decomposition
 +
| <tt>''X<sub>0</sub>''</tt>, <tt>''Y<sub>0</sub>''</tt> and <tt>''&alpha;<sub>0</sub>''</tt> are available
 +
|-
 +
| <tt>3</tt>
 +
| sector-based + central division decomposition
 +
| <tt>''X<sub>0</sub>''</tt>, <tt>''Y<sub>0</sub>''</tt> and <tt>''&alpha;<sub>0</sub>''</tt> are available, only applicable if MPI-tasks > 4
 
|}
 
|}
  
 
<u>Notes:</u>
 
<u>Notes:</u>
 
*Domain decomposition works in MPI mode by separating burnable materials into different parallel tasks.
 
*Domain decomposition works in MPI mode by separating burnable materials into different parallel tasks.
*Number of domains is given by the number of MPI tasks
+
*The number of domains is given by the number of MPI tasks.
*Only burnable materials separated into depletion zones using the "div sep" option are decomposed
+
*Only burnable materials separated into depletion zones using the '''sep''' entry in the [[#div|div card]] are decomposed
*<tt>''MODE''</tt> 1 decomposes the geometry based on the automatically assigned depletion zone indexes (not recommended).
+
*Decomposed materials are plotted in domain-specific colors (unless the '''rgb''' entry in the [[#mat (material definition)|mat card]] is used)
*<tt>''MODE''</tt> 2 decomposes the zones into sectors and <tt>''MODE''</tt> 3 adds a central zone if the number of domains is greater than 4.
+
*For more information, see the detail description and practical example on the [[Domain decomposition|Domain decomposition]].
*The additional input options, <tt>''X<sub>0</sub>''</tt>, <tt>''Y<sub>0</sub>''</tt> and <tt>''&alpha;<sub>0</sub>''</tt>, are available for <tt>''MODE''</tt> 2 and 3 (default values are zero).
+
*Decomposed materials are plotted in domain-specific colors (unless the '''rgb''' entry in the [[#mat (material definition)|material card]] is used)
+
*See [[Domain decomposition|practical example]] for more information.
+
  
 
=== set declib ===
 
=== set declib ===
Line 3,764: Line 4,416:
 
<u>Notes:</u>
 
<u>Notes:</u>
  
*Decay libraries are standard ENDF format files containing decay data.
+
*Decay libraries are standard ENDF format<ref name="endf" /> files containing decay data.
 
*If the file path contains special characters it is advised to enclose it within quotes.
 
*If the file path contains special characters it is advised to enclose it within quotes.
 
*A default directory path can be set by defining environment variable <tt>SERPENT_DATA</tt>. The code looks for decay data files in this path if not found at the absolute.
 
*A default directory path can be set by defining environment variable <tt>SERPENT_DATA</tt>. The code looks for decay data files in this path if not found at the absolute.
Line 3,799: Line 4,451:
  
 
<u>Notes:</u>
 
<u>Notes:</u>
 
+
* Default values:
*Delayed neutron emission is on by default in neutron criticality source and off by default in (static/dynamic) external source simulation mode.
+
** Criticality source mode: the delayed neutron emission is "<tt>on</tt>"
*In time-dependent calculations, driven by the [[#set dynsrc|set dynsrc]] option, precursor based delayed neutron emission is included in the calculation: off at fission, but on at delayed nubar in total nubar.
+
** External (static/dynamic) source mode: the delayed neutron emission is "<tt>off</tt>"
 +
*In time-dependent calculations, driven by the [[#set dynsrc|set dynsrc]] option, precursor based delayed neutron emission is included in the calculation: "<tt>off</tt>" at fission, but "<tt>on</tt>" at delayed nubar in total nubar.
 
*See separate description of [[physics options in Serpent]] for differences to other codes.
 
*See separate description of [[physics options in Serpent]] for differences to other codes.
  
Line 3,807: Line 4,460:
 
  '''set depmtx''' ''MODE''
 
  '''set depmtx''' ''MODE''
  
Print burnup matrixes to <tt>[input]_depmtx_[mat]_[bu]_[ss].m</tt> file during burnup calculation, where "<tt>bu</tt>" is the burnup step and "<tt>ss</tt>" is the substep.
+
Print burnup matrixes to <tt>[input]_depmtx_[mat]_[bu]_[ss].m</tt> file during burnup calculation, where "<tt>bu</tt>" is the burnup step and "<tt>ss</tt>" is the substep. Input values:
  
 
{|
 
{|
 
| <tt>''MODE''</tt>
 
| <tt>''MODE''</tt>
| : Set printing on (1/yes) or off (0/no).
+
| : option to switch on (1/yes) or off (0/no) the printing of burnup matrixes. The default value is "<tt>off</tt>"
 
|}
 
|}
  
Line 3,825: Line 4,478:
 
{|
 
{|
 
| <tt>''MODE''</tt>
 
| <tt>''MODE''</tt>
| : value indicating, which materials to output to the <tt>[input]_dep.m</tt> file (1 = only partials, 2 = only parents, 3 = both)
+
| : value indicating, which materials to output to the <tt>[input]_dep.m</tt> file (1 = only partials, 2 = only parents, 3 = both). (default value: 2)
 
|-
 
|-
 
|<tt>''STEP''</tt>
 
|<tt>''STEP''</tt>
| : value indicating the print-out interval of the <tt>[input]_dep.m</tt> file (0 = final step, 1 = all steps, 2 =none)
+
| : value indicating the print-out interval of the <tt>[input]_dep.m</tt> file (0 = final step, 1 = all steps, 2 =none). (default value: 1)
 
|}
 
|}
  
 
<u>Notes:</u>
 
<u>Notes:</u>
 
*Parent materials refer to materials defined by [[#mat (material definition)|mat cards]], and partials to depletion zones created automatically using the [[#div (divisor definition)|div card]].
 
*Parent materials refer to materials defined by [[#mat (material definition)|mat cards]], and partials to depletion zones created automatically using the [[#div (divisor definition)|div card]].
*Default mode is 2 (only parents) and default print-out interval step is 1 (all steps).
+
*Print-out interval step option 2, no <tt>[input]_dep.m</tt> generation, can be combined with post-processing re-depletion: [[Installing and running Serpent#Running Serpent|<tt>''-rdep''</tt>]] command line option.
*Print-out interval step option 2, no <tt>[input]_dep.m</tt> generation, can be combined with post-processing re-depletion: "-rdep" command line option.
+
 
*This option when is used with the domain decomposition feature, [[#set dd|set dd]], in a mode different from 2, generates multiple depletion files which are named adding <tt>_dd[mpiid]</tt> (domain decomposition identifier) to the standard file name. Each of them contains the partial materials information of the given domain/MPI task.
 
*This option when is used with the domain decomposition feature, [[#set dd|set dd]], in a mode different from 2, generates multiple depletion files which are named adding <tt>_dd[mpiid]</tt> (domain decomposition identifier) to the standard file name. Each of them contains the partial materials information of the given domain/MPI task.
  
Line 3,840: Line 4,492:
  
 
   '''set deppara''' ''PARAM_LIST''
 
   '''set deppara''' ''PARAM_LIST''
Defines the variables included in the depletion output file <tt>[input]_dep.m</tt>. Input values:
+
Defines the material- and isotopic-wise variables included in the depletion output file <tt>[input]_dep.m</tt>. Input values:
  
 
{|
 
{|
 
| <tt>''PARAM_LIST''</tt>
 
| <tt>''PARAM_LIST''</tt>
| : list of variables
+
| : list of variables (default value: "<tt>all</tt>")
 +
|}
 +
 
 +
Possible key-words/variables are:
 +
::{| class="wikitable" style="text-align: left;"
 +
! Key-word
 +
! Quantity
 +
! Output ID
 +
! Description
 +
|-
 +
| <tt>atom</tt>
 +
| atom density
 +
| <tt>ADENS</tt>
 +
| [in in b<sup>-1</sup> cm<sup>-1</sup>]
 +
|-
 +
| <tt>mass</tt>
 +
| mass density
 +
| <tt>MDENS</tt>
 +
| [in g/cm<sup>3</sup>]
 +
|-
 +
| <tt>activity</tt>
 +
| activity
 +
| <tt>A</tt>
 +
| [in Bq]
 +
|-
 +
| <tt>dh</tt>
 +
| decay heat
 +
| <tt>H</tt>
 +
| [in W]
 +
|-
 +
| <tt>sf</tt>
 +
| spontaneous fission rate
 +
| <tt>SF</tt>
 +
| [in fissions/s]
 +
|-
 +
| <tt>gsrc</tt>
 +
| photon emission rate
 +
| <tt>GSRC</tt>
 +
| [in photons/s]
 +
|-
 +
| <tt>ingtox</tt>
 +
| ingestion toxicity
 +
| <tt>ING_TOX</tt>
 +
| [in Sv]
 +
|-
 +
| <tt>inhtox</tt>
 +
| inhalation toxicity
 +
| <tt>INH_TOX</tt>
 +
| [in Sv]
 +
|-
 +
| <tt>all</tt>
 +
|
 +
|
 +
| include full-set of variables
 +
|-
 +
| <tt>none</tt>
 +
|
 +
|
 +
| exclude full-set of variables
 
|}
 
|}
  
 
<u>Notes:</u>
 
<u>Notes:</u>
*The default behavior is to include the full-set of variables: atomic density, mass density, activity, decay heat, spontaneous fission rate, photon emission rate, ingestion toxicity, and inhalation toxicity (see the [[Description of output files#Burnup calculation output|description of the depletion output]]).
+
*For more information, see detailed description of the [[Description of output files#Burnup calculation output|burnup calculation output]].
*The entry-names to input the variables are: "atom" = atomic density, "mass" = mass density, "activity" = activity, "dh" = decay heat, "sf" = spontaneous fission rate, "gsrc" = photon emission rate, "ingtox" = ingestion toxicity, "inhtox" = inhalation toxicity.
+
*Alternatively, the "all" entry would include the full-set of variables, and the "none" entry would exclude the full-set of variables.
+
  
 
=== set depstepbunorm ===
 
=== set depstepbunorm ===
Line 3,864: Line 4,572:
  
 
<u>Notes</u>
 
<u>Notes</u>
* By default, for energy deposition modes 0/1, the normalization includes only "burnable" materials while for energy deposition modes 2/3, the normalization includes "all" materials (see [[#set edepmode|set edepmode]]).
+
* Default values (see [[#set edepmode|set edepmode]]):
 +
** For energy deposition modes 0/1the normalization includes only "<tt>burnable</tt>" materials - mode 2.
 +
** For energy deposition modes 2/3: the normalization includes "<tt>all materials</tt>" - mode 1.
  
 
=== set dfsol ===
 
=== set dfsol ===
Line 3,872: Line 4,582:
 
{|
 
{|
 
| <tt>''MODE''</tt>
 
| <tt>''MODE''</tt>
| : boundary conditions for solver (1 = include net currents at boundary surfaces and corners, 2 = include only surface currents)
+
| : boundary conditions for solver (1 = include net currents at boundary surfaces and corners, 2 = include only surface currents). (default value: 1)
 
|-
 
|-
 
| <tt>''DC''</tt>
 
| <tt>''DC''</tt>
| : type of diffusion coefficient used in the calculation (1 = INF_DIFFCOEF, 2 = TRC_DIFFCOEF)
+
| : type of diffusion coefficient used in the calculation (1 = INF_DIFFCOEF, 2 = TRC_DIFFCOEF). (default value: 1)
 
|-
 
|-
 
| <tt>''NP''</tt>
 
| <tt>''NP''</tt>
| : number of points for trapezoidal integration for homogeneous flux
+
| : number of points for trapezoidal integration for homogeneous flux (default value: 100)
 
|}
 
|}
  
 
<u>Notes:</u>
 
<u>Notes:</u>
*This input option is used to control how the deterministic diffusion flux solver used to obtain assembly discontinuity factors [[#set adf|(set adf)]] and pin power distributions [[#set ppw|(set ppw)]] is run.
+
*This input option is used to control how the deterministic diffusion flux solver used to obtain assembly discontinuity factors ([[#set adf|set adf]]) and pin power distributions ([[#set ppw|set ppw]]) is run.
*Default mode is 1 (include both surfaces and corners in solution).
+
*The "<tt>TRC_DIFFCOEF</tt>" diffusion coefficient requires the [[#set trc|set trc]] option.
*Default diffusion coefficient is INF_DIFFCOEF, option 2 requires the [[#set trc|set trc]] option.
+
*Default number of points for trapezoidal integration is 100.
+
 
*See also separate description of the [[Diffusion flux solver|built-in diffusion flux solver]].
 
*See also separate description of the [[Diffusion flux solver|built-in diffusion flux solver]].
 
*The format was revised in update 2.1.27 (<tt>''DC''</tt> option was added between <tt>''MODE''</tt> and <tt>''NP''</tt>).
 
*The format was revised in update 2.1.27 (<tt>''DC''</tt> option was added between <tt>''MODE''</tt> and <tt>''NP''</tt>).
Line 3,891: Line 4,599:
 
=== set dix ===
 
=== set dix ===
 
  '''set dix''' ''OPT''
 
  '''set dix''' ''OPT''
Sets double indexing for cross section energy grid look-up on or off:
+
Sets double indexing for cross section energy grid look-up on or off. Input values:
  
 
{|
 
{|
Line 3,906: Line 4,614:
  
 
  '''set dspec''' ''EGRID<sub>p</sub>'' ''EGRID<sub>n</sub>''
 
  '''set dspec''' ''EGRID<sub>p</sub>'' ''EGRID<sub>n</sub>''
Sets the energy grid structure for decay spectra.
+
Sets the energy grid structure for decay spectra. Input values:
  
 
{|
 
{|
Line 3,950: Line 4,658:
 
{|
 
{|
 
| <tt>''OPT</tt>
 
| <tt>''OPT</tt>
| : option to switch on (1/yes) or off (0/no) the store/write dynamic data into a file. Default is on.
+
| : option to switch on (1/yes) or off (0/no) the store/write dynamic data into a file. The default option is "<tt>on</tt>".
 
|}
 
|}
  
Line 3,957: Line 4,665:
 
  '''set dynsrc''' ''PATH'' [ ''MODE'' ]
 
  '''set dynsrc''' ''PATH'' [ ''MODE'' ]
  
Links previously generated steady state source distributions to be used in a transient simulation with delayed neutron emission.
+
Links previously generated steady state source distributions to be used in a transient simulation with delayed neutron emission. Input values:
  
 
{|
 
{|
Line 3,972: Line 4,680:
 
=== set ecut ===
 
=== set ecut ===
  
  '''set ecut''' ''EMIN<sub>n</sub>'' ''EMIN<sub>p<sub>''
+
  '''set ecut''' ''EMIN<sub>n</sub>'' [ ''EMIN<sub>p<sub>'' ]
 
Sets minimum energy cut-off for neutrons and photons. Input values:
 
Sets minimum energy cut-off for neutrons and photons. Input values:
  
 
{|
 
{|
| <tt>''EMIN<sub>n</sub>''</tt>
+
| <tt>''EMIN<sub>n</sub>''</tt>  
| : cut-off energy for neutrons (MeV)
+
| : cut-off energy for neutrons [in MeV] (default value: [[Definitions, units and constants#Constants|-INFTY]]/"no cut-off")
 
|-
 
|-
 
| <tt>''EMIN<sub>p</sub>''</tt>
 
| <tt>''EMIN<sub>p</sub>''</tt>
| : cut-off energy for photons (MeV)
+
| : cut-off energy for photons [in MeV] (default value: 1.0E-3)
 
|}
 
|}
  
 
<u>Notes:</u>
 
<u>Notes:</u>
  
*The default cut-of energy for photons is 1 keV. Neutron energy cut-off is switched off by default.
 
 
*Using energy cut-off for neutrons may lead to non-physical results, since fission and up-scattering may not be accurately modeled.
 
*Using energy cut-off for neutrons may lead to non-physical results, since fission and up-scattering may not be accurately modeled.
 
*Versions 2.1.27 and earlier include only photon energy cut-off, which is now the second input parameter.
 
*Versions 2.1.27 and earlier include only photon energy cut-off, which is now the second input parameter.
Line 3,996: Line 4,703:
 
{|
 
{|
 
| <tt>''DENS<sub>i</sub>''</tt>
 
| <tt>''DENS<sub>i</sub>''</tt>
| : mass density (g/cm<sup>3</sup>)
+
| : mass density [in g/cm<sup>3</sup>]
 
|-
 
|-
 
| <tt>''EMIN<sub>p,i</sub>''</tt>
 
| <tt>''EMIN<sub>p,i</sub>''</tt>
| : cut-off energy for photons (MeV)
+
| : cut-off energy for photons [in MeV]
 
|}
 
|}
  
Line 4,015: Line 4,722:
 
|-
 
|-
 
| <tt>''EMIN<sub>p,i</sub>''</tt>
 
| <tt>''EMIN<sub>p,i</sub>''</tt>
| : cut-off energy for photons (MeV)
+
| : cut-off energy for photons [in MeV]
 
|}
 
|}
  
Line 4,021: Line 4,728:
  
 
  '''set eddi''' ''OPT''
 
  '''set eddi''' ''OPT''
Switches on the calculation of Eddington factors. Input values:
+
Option that enables the calculation of Eddington factors. Input values:
  
 
{|
 
{|
 
| <tt>''OPT''</tt>
 
| <tt>''OPT''</tt>
| : option to switch calculation on (1) or off (0)
+
| : option to switch calculation on (1/yes) or off (0/no). The default option is "<tt>off</tt>".
 
|}
 
|}
  
 
<u>Notes:</u>
 
<u>Notes:</u>
  
*Requires group constant generation to be set on.
+
*Requires group constant generation to be set on (see [[#set gcu|set gcu]]).
  
 
=== set edepdel ===
 
=== set edepdel ===
Line 4,039: Line 4,746:
 
{|
 
{|
 
| <tt>''OPT''</tt>
 
| <tt>''OPT''</tt>
| : include (1) or exclude (0) the energy of delayed components in energy deposition estimates
+
| : include (1/yes) or exclude (0/no) the energy of delayed components in energy deposition estimates (default value: 1)
 
|-
 
|-
 
| <tt>''LOCAL_EGD''</tt>
 
| <tt>''LOCAL_EGD''</tt>
| : deposit the energy of the delayed fission gammas to fission sites (1) or with the same distribution as the prompt fission gammas (0) (this option is used only in energy deposition mode 3)
+
| : deposit the energy of the delayed fission gammas to fission sites ("<tt>1</tt>") or with the same distribution as the prompt fission gammas ("<tt>0</tt>"). (default value: 0)
 
|}
 
|}
  
Line 4,048: Line 4,755:
  
 
* Delayed components include delayed neutrons, delayed fission gammas and delayed betas.
 
* Delayed components include delayed neutrons, delayed fission gammas and delayed betas.
* The energy of delayed neutrons can be excluded using this option only in energy deposition mode 1.
 
 
* The energy of the delayed components is deposited at the time of fission so the time dependence of the energy deposition is not accounted for properly in transient simulations.
 
* The energy of the delayed components is deposited at the time of fission so the time dependence of the energy deposition is not accounted for properly in transient simulations.
* Default options are to include delayed components (1) and to deposit the energy of the delayed fission gammas with the same distribution as the prompt fission gammas (0).
+
* The energy of delayed neutrons can be excluded using this option only in energy deposition mode "<tt>3</tt>" (see [[#set edepmode| edepmode]] option).
 +
* Option to deposit the energy of the delayed fission gammas with the same distribution as the prompt fission gammas works only in criticality source simulations.
  
 
=== set edepkcorr ===
 
=== set edepkcorr ===
Line 4,059: Line 4,766:
 
{|
 
{|
 
| <tt>''OPT''</tt>
 
| <tt>''OPT''</tt>
| : apply (1) or do not apply (0) correction
+
| : option to switch the correction on (1/yes) or off (0/no). The default option is "<tt>on</tt>".
 
|}
 
|}
  
Line 4,074: Line 4,781:
 
{|
 
{|
 
| <tt>''MODE''</tt>
 
| <tt>''MODE''</tt>
| : energy deposition mode (0, 1, 2 or 3)
+
| : energy deposition mode: 0, 1, 2 or 3 (default value: 0)
 
|-
 
|-
 
| <tt>''E_CAPT''</tt>
 
| <tt>''E_CAPT''</tt>
| : additional energy release in capture reactions given in MeVs per fission (used only in energy deposition mode 1)
+
| : additional energy release in capture reactions given [in MeVs per fission] (default value: 0.0)
 
|}
 
|}
  
<u>Notes:</u>
+
The possible setting for mode are:
  
* The energy deposition modes are described in related paper. <ref name="edep" />.  
+
::{| class="wikitable" style="text-align: left;"
 +
! Mode
 +
! Description
 +
! Evaluation
 +
|-
 +
| <tt>0</tt>
 +
| Constant energy deposition per fission
 +
| ''<tt>E<sub>fiss,i</sub> = (Q<sub>i</sub>/Q<sub>235</sub>) &times; H<sub>235</sub></tt>''
 +
|-
 +
| <tt>1</tt>
 +
| Local energy deposition based on ENDF MT 458 data
 +
| ''<tt>E<sub>fiss,i</sub> = EFR<sub>i</sub> + ENP<sub>i</sub> + END<sub>i</sub> + EGP<sub>i</sub> + EGD<sub>i</sub> + EB<sub>i</sub> + E<sub>capt</sub></tt>''
 +
|-
 +
| <tt>2</tt>
 +
| Local photon energy deposition
 +
| ''<tt>E<sub>fiss,i</sub> = EFR<sub>i</sub> + EGP<sub>i</sub> + EGD<sub>i</sub> + EB<sub>i</sub></tt>''
 +
|-
 +
| <tt>3</tt>
 +
| Coupled neutron-photon transport
 +
| ''<tt>E<sub>fiss,i</sub> = EFR<sub>i</sub> + EB<sub>i</sub></tt>''
 +
|-
 +
|}
 +
 
 +
<u>Notes:</u>
 +
* The energy deposition modes are described in related paper  <ref name="edep" />.  
 +
* The additional energy release in capture reactions is only used in energy deposition mode "<tt>1</tt>"
 
* The choice of energy deposition mode affects also the normalization of the results, if normalization to total power or power density is used.
 
* The choice of energy deposition mode affects also the normalization of the results, if normalization to total power or power density is used.
* Energy deposition modes 1, 2 and 3 require data which is not available in the standard ACE-format cross section files used by Serpent. Separately distributed ACE-files containing additional data are required to use these modes.
+
* Energy deposition modes "<tt>1</tt>", "<tt>2</tt>" and "<tt>3</tt>" require data which is not available in the standard ACE-format cross section files used by Serpent. [https://vtt.sharefile.eu/d-s3cad105a3d874f988bfb3f1d905fca13 Separately distributed ACE-files] (file endfb71_edep.tar.gz) containing additional data are required to use these modes.
* Default option is 0 which corresponds to the methodology used before version 2.1.31.
+
* KERMA coefficients used in energy deposition modes "<tt>2</tt>" and "<tt>3</tt>" are not Doppler-broadened correctly by the built-in preprocessor. See [[Doppler-broadening preprocessor routine|Doppler-broadening preprocessor]].
* KERMA coefficients used in energy deposition modes 2 and 3 are not Doppler-broadened correctly by the built-in preprocessor. See [[Doppler-broadening preprocessor routine|Doppler-broadening preprocessor]].
+
  
 
=== set egrid ===
 
=== set egrid ===
Line 4,099: Line 4,830:
 
|-
 
|-
 
| <tt>''EMIN''</tt>
 
| <tt>''EMIN''</tt>
| : minimum energy in the grid (MeV)
+
| : minimum energy in the grid [in MeV]
 
|-
 
|-
 
| <tt>''EMAX''</tt>
 
| <tt>''EMAX''</tt>
| : maximum energy in the grid (MeV)
+
| : maximum energy in the grid [in MeV]
 
|}
 
|}
  
 
<u>Notes:</u>
 
<u>Notes:</u>
*The default fractional reconstruction tolerance is 0.0 in transport calculation mode and 5E-5 in burnup calculation mode.
+
* Default values:
 +
** Fraction reconstruction tolerance: 0.0 (transport calculation mode), 5.0E-05 (burnup calculation mode)
 +
** Energy grid boundaries: [1.0E-11, 20.0] (neutrons), [1.0E-03, 100.0] (photons)
 
*A higher energy grid reconstruction tolerance means lower memory consumption and possibly higher computation speed but also reduced accuracy of the calculation.
 
*A higher energy grid reconstruction tolerance means lower memory consumption and possibly higher computation speed but also reduced accuracy of the calculation.
*The default minimum energy is 1E-11 MeV (neutrons) and 1E-03 MeV (photons).
 
*The default maximum energy is 20.0 MeV (neutrons) and 100.0 MeV (photons).
 
 
*See also Section 5.3 of [http://montecarlo.vtt.fi/download/Serpent_manual.pdf Serpent 1 User Manual].
 
*See also Section 5.3 of [http://montecarlo.vtt.fi/download/Serpent_manual.pdf Serpent 1 User Manual].
  
Line 4,119: Line 4,850:
 
{|
 
{|
 
| <tt>''E''</tt>
 
| <tt>''E''</tt>
| : energy cut-off for modelling energy and direction of the scattered photon (MeV)
+
| : energy cut-off for modelling energy and direction of the scattered photon [in MeV] (default value: [[Definitions, units and constants#Constants|INFTY]]/"no cut-off")
 
|}
 
|}
  
 
<u>Notes:</u>
 
<u>Notes:</u>
*The default value of <tt>''E''</tt> is set to INFTY (1E+37).
 
 
*The Klein-Nishina equation is used above <tt>''E''</tt> for calculating both the energy and direction of the scattered photon. Below <tt>''E''</tt>, the Doppler brodening method is used if switched on. Otherwise, the incoherent scattering function approximation is in use.
 
*The Klein-Nishina equation is used above <tt>''E''</tt> for calculating both the energy and direction of the scattered photon. Below <tt>''E''</tt>, the Doppler brodening method is used if switched on. Otherwise, the incoherent scattering function approximation is in use.
  
Line 4,136: Line 4,866:
 
|-
 
|-
 
| <tt>''COND<sub>i</sub>''</tt>
 
| <tt>''COND<sub>i</sub>''</tt>
| : conductivity state (0 = non-conductor, 1 = conductor, 2 = conduction electron dependent).
+
| : conductivity state (0 = non-conductor, 1 = conductor, 2 = conduction electron dependent). (default value: 2)
 
|}
 
|}
  
 
<u>Notes:</u>
 
<u>Notes:</u>
*The default option is 2 or conduction electron dependent.
 
 
*If the material is set as a conductor and no-conduction electrons are found, the material conductivity state is overwritten to non-conductor.
 
*If the material is set as a conductor and no-conduction electrons are found, the material conductivity state is overwritten to non-conductor.
 
*Option 2, conduction electron dependent, establishes that a single element material is a conductor if conduction electrons are found, otherwise is a non-conductor. A compound is always a non-conductor.
 
*Option 2, conduction electron dependent, establishes that a single element material is a conductor if conduction electrons are found, otherwise is a non-conductor. A compound is always a non-conductor.
Line 4,154: Line 4,883:
 
|-
 
|-
 
| <tt>''GAS<sub>i</sub>''</tt>
 
| <tt>''GAS<sub>i</sub>''</tt>
| : phase state (0 = condensed or non-gas, 1 = gas).
+
| : phase state (0 = condensed or non-gas, 1 = gas). (default value: 0)
 
|}
 
|}
  
 
<u>Notes:</u>
 
<u>Notes:</u>
*The default option is 0/gas and only affects mixtures.  
+
*The default option "0/condensed" only affects mixtures.  
 
*The gas phase does not affect the mean excitation energy of a single material: if [[#set elmee|set elmee]] option is set for a material, the material is considered as non-gas.
 
*The gas phase does not affect the mean excitation energy of a single material: if [[#set elmee|set elmee]] option is set for a material, the material is considered as non-gas.
  
Line 4,171: Line 4,900:
 
|-
 
|-
 
| <tt>''MEE<sub>i</sub>''</tt>
 
| <tt>''MEE<sub>i</sub>''</tt>
| : electrons mean excitation energy (MeV)
+
| : electrons mean excitation energy [in MeV] or "-1" (default value: -1)
 
|}
 
|}
  
 
<u>Notes:</u>
 
<u>Notes:</u>
*The default value is -1.0 interpreted as calculated during runtime for compounds and extracted from data for single element materials.  
+
*The default value "-1" is interpreted as calculated during runtime for compounds and extracted from data for single element materials.  
 
*The maximum mean excitation energy for electrons is 1 MeV.
 
*The maximum mean excitation energy for electrons is 1 MeV.
  
Line 4,185: Line 4,914:
 
{|
 
{|
 
| <tt>''EGRID_E''</tt>
 
| <tt>''EGRID_E''</tt>
| : energy grid size (default value is 200)
+
| : energy grid size (default value: 200)
 
|}
 
|}
  
Line 4,198: Line 4,927:
 
{|
 
{|
 
| <tt>''N<sub>X</sub>''</tt>
 
| <tt>''N<sub>X</sub>''</tt>
| : number of mesh cells in x-direction
+
| : number of mesh cells in x-direction (default value: 5)
 
|-
 
|-
 
| <tt>''N<sub>Y</sub>''</tt>
 
| <tt>''N<sub>Y</sub>''</tt>
| : number of mesh cells in y-direction
+
| : number of mesh cells in y-direction (default value: 5)
 
|-
 
|-
 
| <tt>''N<sub>Z</sub>''</tt>
 
| <tt>''N<sub>Z</sub>''</tt>
| : number of mesh cells in z-direction
+
| : number of mesh cells in z-direction (default value: 5)
 
|-
 
|-
 
| <tt>''X<sub>MIN</sub>''</tt>
 
| <tt>''X<sub>MIN</sub>''</tt>
| : minimum mesh boundary in x-direction
+
| : minimum mesh boundary in x-direction [in cm]
 
|-
 
|-
 
| <tt>''X<sub>MAX</sub>''</tt>
 
| <tt>''X<sub>MAX</sub>''</tt>
| : maximum mesh boundary in x-direction
+
| : maximum mesh boundary in x-direction [in cm]
 
|-
 
|-
 
| <tt>''Y<sub>MIN</sub>''</tt>
 
| <tt>''Y<sub>MIN</sub>''</tt>
| : minimum mesh boundary in y-direction
+
| : minimum mesh boundary in y-direction [in cm]
 
|-
 
|-
 
| <tt>''Y<sub>MAX</sub>''</tt>
 
| <tt>''Y<sub>MAX</sub>''</tt>
| : maximum mesh boundary in y-direction
+
| : maximum mesh boundary in y-direction [in cm]
 
|-
 
|-
 
| <tt>''Z<sub>MIN</sub>''</tt>
 
| <tt>''Z<sub>MIN</sub>''</tt>
| : minimum mesh boundary in z-direction
+
| : minimum mesh boundary in z-direction [in cm]
 
|-
 
|-
 
| <tt>''Z<sub>MAX</sub>''</tt>
 
| <tt>''Z<sub>MAX</sub>''</tt>
| : maximum mesh boundary in z-direction
+
| : maximum mesh boundary in z-direction [in cm]
 
|}
 
|}
  
 
<u>Notes:</u>
 
<u>Notes:</u>
*Shannon entropy is used to monitor fission source convergence by recording the distribution of source points on mesh.
+
*Shannon entropy is used to monitor fission source convergence, in criticality source simulations, by recording the distribution of source points on mesh.
 
*The calculation is invoked by setting the generation history record option on ([[#set his|set his]]).
 
*The calculation is invoked by setting the generation history record option on ([[#set his|set his]]).
*The default mesh size is 5 &times; 5 &times; 5, extending over the entire geometry.
+
*If no mesh boundaries are specified, the mesh extends over the entire geometry.
*Monitoring fission source convergence make sense only in criticality source mode.
+
 
*For more information, see detailed description on [[fission source convergence]].
 
*For more information, see detailed description on [[fission source convergence]].
  
Line 4,257: Line 4,985:
 
|-
 
|-
 
| <tt>''E<sub>n</sub>''</tt>
 
| <tt>''E<sub>n</sub>''</tt>
| : energy deposited per fission in MeV
+
| : energy deposited per fission [in MeV] (default value: 202.27)
 
|}
 
|}
  
Line 4,263: Line 4,991:
  
 
*The energy deposited per fission includes additional energy released in capture reactions when fission neutrons are absorbed.
 
*The energy deposited per fission includes additional energy released in capture reactions when fission neutrons are absorbed.
*By default the energy release per U-235 fission is set to 202.27 MeV, and the values for other actinides scaled based the Q-values found in the cross section libraries.
+
*The energy release per fission value for other actinides is scaled based the Q-values found in the cross section libraries in reference with the U-235 value set.
 
*See also [[Input syntax manual#set U235H|set U235H]].
 
*See also [[Input syntax manual#set U235H|set U235H]].
 
*See also Section 5.8 of [http://montecarlo.vtt.fi/download/Serpent_manual.pdf Serpent 1 User Manual].
 
*See also Section 5.8 of [http://montecarlo.vtt.fi/download/Serpent_manual.pdf Serpent 1 User Manual].
Line 4,270: Line 4,998:
  
 
  '''set fissrate''' ''F'' [ ''MAT'' ]
 
  '''set fissrate''' ''F'' [ ''MAT'' ]
 
+
Sets normalization to fission rate. Input values:
Sets normalization to fission rate.
+
  
 
{|
 
{|
 
| <tt>''F''</tt>
 
| <tt>''F''</tt>
| : number of fission reactions per second (1/s)
+
| : number of fission reactions per second [in 1/s]
 
|-
 
|-
 
| <tt>''MAT''</tt>
 
| <tt>''MAT''</tt>
Line 4,283: Line 5,010:
 
<u>Notes:</u>
 
<u>Notes:</u>
 
*Normalization is needed to relate the Monte Carlo reaction rate estimates to a user-given parameter.
 
*Normalization is needed to relate the Monte Carlo reaction rate estimates to a user-given parameter.
*Neutron transport simulations are by default normalized to unit total loss rate.
+
*The default normalization:
*Photon transport simulations are by default normalized to unit total source rate.
+
**It is set to unit total loss rate (neutron transport) and to unit total source rate (photon transport).
 +
**In coupled neutron-photon transport simulations the normalization is driven by the neutron normalization
 
*For other normalization options, see: [[#set power|set power]], [[#set powdens|set powdens]], [[#set flux|set flux]], [[#set genrate|set genrate]], [[#set absrate|set absrate]], [[#set lossrate|set lossrate]], [[#set srcrate|set srcrate]], [[#set sfrate|set sfrate]].
 
*For other normalization options, see: [[#set power|set power]], [[#set powdens|set powdens]], [[#set flux|set flux]], [[#set genrate|set genrate]], [[#set absrate|set absrate]], [[#set lossrate|set lossrate]], [[#set srcrate|set srcrate]], [[#set sfrate|set sfrate]].
 
*See also Section 5.8 of [http://montecarlo.vtt.fi/download/Serpent_manual.pdf Serpent 1 User Manual].
 
*See also Section 5.8 of [http://montecarlo.vtt.fi/download/Serpent_manual.pdf Serpent 1 User Manual].
Line 4,290: Line 5,018:
 
=== set fissye ===
 
=== set fissye ===
  
  '''set fissye''' ''OPT''
+
  '''set fissye''' ''INTT''
Option to include energy-dependent fission yields. Input values:
+
Sets the energy-dependent interpolation scheme for the fission yields. Input values:
  
 
  {|
 
  {|
| <tt>''OPT''</tt>
+
| <tt>''INTT''</tt>
| : include (1/yes) or exclude (0/no) energy-dependent fission yields. Default is 1.
+
| : energy-dependent interpolation (0 = none, 1 = linear-linear, 2 = histogram). The default option is "<tt>1/linear-linear</tt>".
 
|}
 
|}
 +
 +
<u>Notes:</u>
 +
*The default option "1/linear-linear" is based on the two-dimensional interpolation scheme dictated by the ENDF data (File 8: Decay and Fission Product Yields - sec. 0.5.2.2)<ref name="endf">Trkov, A., Herman, M. and Brown, D. A. ''"ENDF-6 Formats Manual."'' CSEWG Document ENDF-102 / BNL-90365-2009 Rev. 2 [https://www.nndc.bnl.gov/endf-b8.0/endf-manual-viii.0.pdf (2018)]</ref>. The interpolation is defined by the neutron energy spectrum.
 +
*The option "0/none" excludes the energy-dependency from the fission yields, i.e. single-value defined at the lower limit.
 +
*The  option "2/histogram" implies that the function is constant and equal to the value given at the lower limit of the interval (e.g., thermal, epithermal, fast values) in connection with how the data were measured in thermal and fast systems. (Note that the option is under evaluation).
  
 
=== set flux===
 
=== set flux===
  
 
  '''set flux''' ''F'' [ ''MAT'' ]
 
  '''set flux''' ''F'' [ ''MAT'' ]
Sets normalization to total flux.
+
Sets normalization to total flux. Input values:
  
 
{|
 
{|
Line 4,313: Line 5,046:
 
<u>Notes:</u>
 
<u>Notes:</u>
 
*Normalization is needed to relate the Monte Carlo reaction rate estimates to a user-given parameter.
 
*Normalization is needed to relate the Monte Carlo reaction rate estimates to a user-given parameter.
*Neutron transport simulations are by default normalized to unit total loss rate.
+
*The default normalization:
*Photon transport simulations are by default normalized to unit total source rate.
+
**It is set to unit total loss rate (neutron transport) and to unit total source rate (photon transport).
 +
**In coupled neutron-photon transport simulations the normalization is driven by the neutron normalization.
 
*For other normalization options, see: [[#set power|set power]], [[#set powdens|set powdens]], [[#set genrate|set genrate]], [[#set fissrate|set fissrate]], [[#set absrate|set absrate]], [[#set lossrate|set lossrate]], [[#set srcrate|set srcrate]], [[#set sfrate|set sfrate]].
 
*For other normalization options, see: [[#set power|set power]], [[#set powdens|set powdens]], [[#set genrate|set genrate]], [[#set fissrate|set fissrate]], [[#set absrate|set absrate]], [[#set lossrate|set lossrate]], [[#set srcrate|set srcrate]], [[#set sfrate|set sfrate]].
 
*See also Section 5.8 of [http://montecarlo.vtt.fi/download/Serpent_manual.pdf Serpent 1 User Manual].
 
*See also Section 5.8 of [http://montecarlo.vtt.fi/download/Serpent_manual.pdf Serpent 1 User Manual].
Line 4,325: Line 5,059:
 
{|
 
{|
 
| <tt>''OPT''</tt>
 
| <tt>''OPT''</tt>
| :  option to switch calculation on (1/yes) or off (0/no). Default is off.
+
| :  option to switch calculation on (1/yes) or off (0/no). The default option is "<tt>off</tt>".
 
|}
 
|}
  
Line 4,346: Line 5,080:
 
{|
 
{|
 
| <tt>''MAT<sub>n</sub>''</tt>
 
| <tt>''MAT<sub>n</sub>''</tt>
| : material list (type 1)
+
| : material list
 
|-
 
|-
 
| <tt>''UNI<sub>n</sub>''</tt>
 
| <tt>''UNI<sub>n</sub>''</tt>
| : universe list (type 2)
+
| : universe list
 
|-
 
|-
 
| <tt>''LVL''</tt>
 
| <tt>''LVL''</tt>
| : level number (type 3)
+
| : level number
 
|-
 
|-
 
| <tt>''X<sub>MIN</sub>''</tt>
 
| <tt>''X<sub>MIN</sub>''</tt>
| : minimum x-coordinate mesh boundary (type 4)
+
| : minimum x-coordinate mesh boundary [in cm]
 
|-
 
|-
 
| <tt>''X<sub>MAX</sub>''</tt>
 
| <tt>''X<sub>MAX</sub>''</tt>
| : maximum x-coordinate mesh boundary (type 4)
+
| : maximum x-coordinate mesh boundary [in cm]
 
|-
 
|-
 
| <tt>''N<sub>X</sub>''</tt>
 
| <tt>''N<sub>X</sub>''</tt>
| : number of x-mesh cells (type 4)
+
| : number of x-mesh cells  
 
|-
 
|-
 
| <tt>''Y<sub>MIN</sub>''</tt>
 
| <tt>''Y<sub>MIN</sub>''</tt>
| : minimum y-coordinate mesh boundary (type 4)
+
| : minimum y-coordinate mesh boundary [in cm]
 
|-
 
|-
 
| <tt>''Y<sub>MAX</sub>''</tt>
 
| <tt>''Y<sub>MAX</sub>''</tt>
| : maximum y-coordinate mesh boundary (type 4)
+
| : maximum y-coordinate mesh boundary [in cm]
 
|-
 
|-
 
| <tt>''N<sub>Y</sub>''</tt>
 
| <tt>''N<sub>Y</sub>''</tt>
| : number of y-mesh cells (type 4)
+
| : number of y-mesh cells
 
|-
 
|-
 
| <tt>''Z<sub>MIN</sub>''</tt>
 
| <tt>''Z<sub>MIN</sub>''</tt>
| : minimum z-coordinate mesh boundary (type 4)
+
| : minimum z-coordinate mesh boundary [in cm]
 
|-
 
|-
 
| <tt>''Z<sub>MAX</sub>''</tt>
 
| <tt>''Z<sub>MAX</sub>''</tt>
| : maximum z-coordinate mesh boundary (type 4)
+
| : maximum z-coordinate mesh boundary [in cm]
 
|-
 
|-
 
| <tt>''N<sub>Z</sub>''</tt>
 
| <tt>''N<sub>Z</sub>''</tt>
| : number of z-mesh cells (type 4)
+
| : number of z-mesh cells
 
|}
 
|}
  
Line 4,406: Line 5,140:
  
 
  '''set fpcut''' ''FPCUT''
 
  '''set fpcut''' ''FPCUT''
 +
Sets the fission product yield cut-off. Input values:
  
Fission product yield cut-off <tt>''FPCUT''</tt> acts on cumulative fission yields, and the given number is the lower limit for the maximum cumulative yield in each mass chain. So "set fpcut 1E-2" means that every mass chain (nuclides with same mass number) with all cumulative yields below 1% are discarded from the calculation.
+
{|
 +
| <tt>''FPCUT''</tt>
 +
|: fission product yield cut-off (default value: 0.0)
 +
|}
 +
 
 +
<u>Notes:</u>
  
The cut-off is set to zero by default. Setting the <tt>''FPCUT''</tt> cut-off to a higher value (~1E-4 or so) is an effective way to reduce the number of nuclides in the calculation, but at some point it will start affecting the results.
+
*Fission product yield cut-off <tt>''FPCUT''</tt> acts on cumulative fission yields, and the given number is the lower limit for the maximum cumulative yield in each mass chain. So "set fpcut 1E-2" means that every mass chain (nuclides with same mass number) with all cumulative yields below 1% are discarded from the calculation.
 +
*Setting the <tt>''FPCUT''</tt> cut-off to a higher value (~1E-4 or so) is an effective way to reduce the number of nuclides in the calculation, but at some point it will start affecting the results.
  
 
=== set fsp ===
 
=== set fsp ===
Line 4,415: Line 5,156:
 
  '''set fsp''' ''OPT'' ''NSKIP''
 
  '''set fsp''' ''OPT'' ''NSKIP''
  
Sets fission source passing between two transport simulations in burnup or coupled calculation. The fission source at the end of one transport calculation is used as the initial source for the next transport calculation.
+
Sets fission source passing between two transport simulations in burnup or coupled calculation. The fission source at the end of one transport calculation is used as the initial source for the next transport calculation. Input values:
  
 
{|
 
{|
 
| <tt>''OPT''</tt>
 
| <tt>''OPT''</tt>
| : option to switch fission source passing on (1/yes) or off (0/no)  
+
| : option to switch fission source passing on (1/yes) or off (0/no). The default option is "<tt>off</tt>".
 
|-
 
|-
 
| <tt>''NSKIP''</tt>
 
| <tt>''NSKIP''</tt>
Line 4,427: Line 5,168:
 
<u>Notes:</u>
 
<u>Notes:</u>
  
*Fission source passing is turned off by default.
 
 
*Number of inactive generations is taken from [[#set pop|set pop]] card on the first step and from [[#set fsp|set fsp]] on all later steps.
 
*Number of inactive generations is taken from [[#set pop|set pop]] card on the first step and from [[#set fsp|set fsp]] on all later steps.
  
Line 4,438: Line 5,178:
 
  '''set fum''' ''ERG'' [ ''BTCH MODE DC LIM TGT ITER INIT'' ]
 
  '''set fum''' ''ERG'' [ ''BTCH MODE DC LIM TGT ITER INIT'' ]
  
Activates fundamental mode calculation for collapsing intermediate multi-group constant data into few-group constants with a critical spectrum.
+
Activates fundamental mode calculation for collapsing intermediate multi-group constant data into few-group constants with a critical spectrum. Input values:
  
 
{|
 
{|
Line 4,448: Line 5,188:
 
|-
 
|-
 
| <tt>''MODE''</tt>
 
| <tt>''MODE''</tt>
| : Critical spectrum calculation type (default 0, i.e. old B<sub>1</sub> calculation)
+
| : Critical spectrum calculation type (default value: 0, i.e. old B<sub>1</sub> calculation)
 
|-
 
|-
 
| <tt>''DC''</tt>
 
| <tt>''DC''</tt>
Line 4,454: Line 5,194:
 
|-
 
|-
 
| <tt>''LIM''</tt>
 
| <tt>''LIM''</tt>
| : Convergence criterion of k<sub>eff</sub> in fundamental mode calculation calculated as the absolute value difference of k<sub>eff</sub> between successive iterations (default value 1E-7)
+
| : Convergence criterion of k<sub>eff</sub> in fundamental mode calculation calculated as the absolute value difference of k<sub>eff</sub> between successive iterations (default value: 1E-7)
 
|-
 
|-
 
| <tt>''TGT''</tt>
 
| <tt>''TGT''</tt>
| : Target value for fundamental mode k<sub>eff</sub> (default value 1.0, not used with the old B<sub>1</sub> calculation mode)
+
| : Target value for fundamental mode k<sub>eff</sub> (default value: 1.0, not used with the old B<sub>1</sub> calculation mode)
 
|-
 
|-
 
| <tt>''ITER''</tt>
 
| <tt>''ITER''</tt>
| : Maximum number of fundamental mode calculation iterations (default value 25, not used with the old B<sub>1</sub> calculation mode)
+
| : Maximum number of fundamental mode calculation iterations (default value: 25, not used with the old B<sub>1</sub> calculation mode)
 
|-
 
|-
 
| <tt>''INIT''</tt>
 
| <tt>''INIT''</tt>
| : First guess for absolute value of critical B<sup>2</sup> (default value 1E-6, not used with the old B<sub>1</sub> calculation mode)
+
| : First guess for absolute value of critical B<sup>2</sup> (default value: 1E-6, not used with the old B<sub>1</sub> calculation mode)
 
|}
 
|}
  
 
The possible values for mode are:
 
The possible values for mode are:
  
{| class="wikitable" style="text-align: left;"
+
::{| class="wikitable" style="text-align: left;"
 
! Mode
 
! Mode
 
! Description
 
! Description
Line 4,487: Line 5,227:
 
The possible values for FM mode multi-group diffusion coefficients are:
 
The possible values for FM mode multi-group diffusion coefficients are:
  
{| class="wikitable" style="text-align: left;"
+
::{| class="wikitable" style="text-align: left;"
 
! Mode
 
! Mode
 
! Description
 
! Description
Line 4,539: Line 5,279:
  
 
  '''set gct''' ''OPT''
 
  '''set gct''' ''OPT''
Switches on the calculation of group constant statistics test. Input values:
+
Option that enables the calculation of group constant statistics tests. Input values:
  
 
{|
 
{|
 
| <tt>''OPT''</tt>
 
| <tt>''OPT''</tt>
| : option to switch calculation on (1/yes) or off (0/no). Default is off.
+
| : option to switch calculation on (1/yes) or off (0/no). The default option is "<tt>off</tt>".
 
|}
 
|}
  
 
<u>Notes:</u>
 
<u>Notes:</u>
 
*When this option is set, the batch-wise statistical tests are printed in the file <tt>[input]_stat.m</tt>.
 
*When this option is set, the batch-wise statistical tests are printed in the file <tt>[input]_stat.m</tt>.
*Requires group constant generation to be set on.
+
*Requires group constant generation to be set on (see [[#set gcu|set gcu]]).
 
*''Note to developers: statistical tests should be documented''
 
*''Note to developers: statistical tests should be documented''
  
Line 4,558: Line 5,298:
 
{|
 
{|
 
| <tt>''UNI<sub>n</sub>''</tt>
 
| <tt>''UNI<sub>n</sub>''</tt>
| : universe where group constants are generated or -1 to switch group constant generation off.
+
| : universe where group constants are generated or "-1" to switch group constant generation off (default value: 0 = root universe)
 
|}
 
|}
  
 
<u>Notes:</u>
 
<u>Notes:</u>
*Group constants are by default generated in the root universe (universe 0).
+
*By default, group constants are generated in the root universe.
 
*Group constant generation should be switched off when the results are not needed (this may speed up the calculation).
 
*Group constant generation should be switched off when the results are not needed (this may speed up the calculation).
 
*See separate description of [[Output parameters#Homogenized group constants|output parameters]].
 
*See separate description of [[Output parameters#Homogenized group constants|output parameters]].
Line 4,586: Line 5,326:
 
  '''set genrate''' ''G'' [ ''MAT'' ]
 
  '''set genrate''' ''G'' [ ''MAT'' ]
  
Sets normalization to fission neutron generation rate.
+
Sets normalization to fission neutron generation rate. Input values:
  
 
{|
 
{|
 
| <tt>''G''</tt>
 
| <tt>''G''</tt>
| : number of fission neutrons emitted per second (in neutrons/s)
+
| : number of fission neutrons emitted per second [in neutrons/s]
 
|-
 
|-
 
| <tt>''MAT''</tt>
 
| <tt>''MAT''</tt>
| : material in which the fission neutrons are generated
+
| : material in which the fission neutrons are generated. The default is all materials
 
|}
 
|}
  
Line 4,600: Line 5,340:
 
*If the material name is omitted, the value corresponds to total fission neutron generation rate in the system.
 
*If the material name is omitted, the value corresponds to total fission neutron generation rate in the system.
 
*The neutron generation rate includes only prompt and delayed neutrons emitted in fission.
 
*The neutron generation rate includes only prompt and delayed neutrons emitted in fission.
*Neutron transport simulations are by default normalized to unit total loss rate.
+
*The default normalization:
*Photon transport simulations are by default normalized to unit total source rate.
+
**It is set to unit total loss rate (neutron transport) and to unit total source rate (photon transport).
 +
**In coupled neutron-photon transport simulations the normalization is driven by the neutron normalization.
 
*For other normalization options, see: [[#set power|set power]], [[#set powdens|set powdens]], [[#set flux|set flux]], [[#set fissrate|set fissrate]], [[#set absrate|set absrate]], [[#set lossrate|set lossrate]], [[#set srcrate|set srcrate]], [[#set sfrate|set sfrate]].
 
*For other normalization options, see: [[#set power|set power]], [[#set powdens|set powdens]], [[#set flux|set flux]], [[#set fissrate|set fissrate]], [[#set absrate|set absrate]], [[#set lossrate|set lossrate]], [[#set srcrate|set srcrate]], [[#set sfrate|set sfrate]].
 
*See also Section 5.8 of [http://montecarlo.vtt.fi/download/Serpent_manual.pdf Serpent 1 User Manual].
 
*See also Section 5.8 of [http://montecarlo.vtt.fi/download/Serpent_manual.pdf Serpent 1 User Manual].
Line 4,608: Line 5,349:
 
   
 
   
 
  '''set gpop''' ''MAX_HIS'' ''MAX_POP'' ''C''
 
  '''set gpop''' ''MAX_HIS'' ''MAX_POP'' ''C''
Sets the on-the-fly neutron growing population size algorithm.
+
Sets the on-the-fly neutron growing population size algorithm. Input values:
  
 
{|
 
{|
Line 4,636: Line 5,377:
 
|-
 
|-
 
| <tt>''OPT''</tt>
 
| <tt>''OPT''</tt>
| : option to include secondary photons in transport calculation (1/yes), (0/no)
+
| : option to include (1/yes) or exclude (0/no) secondary photons in transport calculation. The default option is "<tt>off</tt>".
 
|}
 
|}
  
Line 4,642: Line 5,383:
  
 
*Applicable only to coupled neutron-photon transport simulation (invoked using [[#set ngamma|set ngamma]]).
 
*Applicable only to coupled neutron-photon transport simulation (invoked using [[#set ngamma|set ngamma]]).
*Default photon-production (from neutron reactions) transported simulation option is switched off (0/no).
 
 
*Only source points from active cycles are included in criticality source simulations.
 
*Only source points from active cycles are included in criticality source simulations.
 +
*From version 2.2.1 and on, multi-step depletion source files can be generated <tt>[''FILE'']_[bu]</tt>, where "<tt>bu</tt>" is the burnup step. Otherwise, simply, <tt>[''FILE'']</tt>.
  
 
=== set his ===
 
=== set his ===
Line 4,652: Line 5,393:
 
{|
 
{|
 
| <tt>''OPT''</tt>
 
| <tt>''OPT''</tt>
| : option to switch batch history record on (1/yes) or off (0/no)
+
| : option to switch batch history record on (1/yes) or off (0/no). The default option is "<tt>off</tt>".
 
|}
 
|}
  
Line 4,666: Line 5,407:
 
{|
 
{|
 
| <tt>''GEN''</tt>
 
| <tt>''GEN''</tt>
| : Number of generations. Default value 15.
+
| : Number of generations (default value: 15).
 
|}
 
|}
  
Line 4,703: Line 5,444:
  
 
=== set impl ===
 
=== set impl ===
'''set''' '''impl''' ''ICAPT'' [ ''INXN'' ''INUBAR'' ''ILEAK'' ]
 
  
Sets implicit reaction modes on or off.
+
'''set''' '''impl''' ''ICAPT'' [ ''INXN'' ''INUBAR'' ''ILEAK'' ]
 +
Sets implicit reaction modes on or off. Input values:
  
 
{|
 
{|
Line 4,723: Line 5,464:
 
<u>Notes:</u>
 
<u>Notes:</u>
  
*Group constant generation requires implicit nxn reactions to be set on.
+
*Group constant generation (see [[#set gcu|set gcu]]) requires implicit nxn reactions to be set "<tt>on</tt>".
 
*If an implicit nubar is given, the weights of the fission neutrons are scaled to conserve the physical number of fission neutrons.
 
*If an implicit nubar is given, the weights of the fission neutrons are scaled to conserve the physical number of fission neutrons.
*Implicit leakage requires group constant generation to be set on.
+
*Implicit leakage requires group constant generation (see [[#set gcu|set gcu]]) to be set "<tt>on</tt>".
 
*See separate description of [[physics options in Serpent]] for differences to other codes.
 
*See separate description of [[physics options in Serpent]] for differences to other codes.
  
 
=== set inftrk ===
 
=== set inftrk ===
'''set''' '''inftrk''' ''LOOP<sub>n</sub>'' [ ''ERR<sub>n</sub>'' ''LOOP<sub>p</sub>'' ''ERR<sub>p</sub>'' ]
 
  
Sets parameters for terminating infinite tracking loops.
+
'''set''' '''inftrk''' ''LOOP<sub>n</sub>'' [ ''ERR<sub>n</sub>'' ''LOOP<sub>p</sub>'' ''ERR<sub>p</sub>'' ]
 +
Sets parameters for terminating infinite tracking loops. Input values:
  
 
{|
 
{|
 
| <tt>''LOOP<sub>n</sub>''</tt>
 
| <tt>''LOOP<sub>n</sub>''</tt>
| : number of neutron tracking loops interpreted as a geometry error
+
| : number of neutron tracking loops interpreted as a geometry error (default value 1E6)
 
|-
 
|-
 
| <tt>''ERR<sub>n</sub>''</tt>
 
| <tt>''ERR<sub>n</sub>''</tt>
| : flag to terminate neutron tracking when an infinite loop occurs (0 = no, 1 = yes)
+
| : flag to terminate neutron tracking when an infinite loop occurs: on (0/no) or off (1/ yes). The default option is "<tt>on</tt>"
 
|-
 
|-
 
| <tt>''LOOP<sub>p</sub>''</tt>
 
| <tt>''LOOP<sub>p</sub>''</tt>
| : number of photon tracking loops interpreted as a geometry error
+
| : number of photon tracking loops interpreted as a geometry error (default value 1E6)
 
|-
 
|-
 
| <tt>''ERR<sub>p</sub>''</tt>
 
| <tt>''ERR<sub>p</sub>''</tt>
| : flag to terminate photon tracking when an infinite loop occurs (0 = no, 1 = yes)
+
| : flag to terminate photon tracking when an infinite loop occurs: on (0/no) or off (1/ yes). The default option is "<tt>on</tt>"
 
|-
 
|-
 
|}
 
|}
Line 4,750: Line 5,491:
 
<u>Notes:</u>
 
<u>Notes:</u>
  
*Serpent checks for tracking loop length to avoid simulation being stuck in an infinite loop. The simulation is terminated by default if no material is found after the particle has been moved forward 1000000 times.
+
*Serpent checks for tracking loop length to avoid simulation being stuck in an infinite loop.
 
*Long loops can occur by chance in complicated geometries, and this parameter allows continuing the simulation without terminating with error message.
 
*Long loops can occur by chance in complicated geometries, and this parameter allows continuing the simulation without terminating with error message.
 
*Even if the problem can be solved by switching the infinite loop error off, it is advised to check the geometry for possible errors.
 
*Even if the problem can be solved by switching the infinite loop error off, it is advised to check the geometry for possible errors.
Line 4,756: Line 5,497:
 
=== set inventory ===
 
=== set inventory ===
  
  '''set inventory''' ''MAT<sub>1</sub>'' ''MAT<sub>2</sub>'' ...
+
  '''set inventory''' ''ID<sub>1</sub>'' ''ID<sub>2</sub>'' ...
 +
 
 +
Defines the nuclides or elements to include in the depletion output file <tt>[input]_dep.m</tt>. Input values:
 +
 
 +
{|
 +
| <tt>''ID<sub>n</sub>''</tt>
 +
| : Identifier for nuclide, or element or special entry.
 +
|}
 +
 
 +
<u>Notes:</u>
 +
*Nuclides are entered using element symbol and mass number (e.g U-235, Am-242m, etc.) or ZAI (922350, 952421, etc.). In the ZAI format the last digit refers to the isomeric state (0 = ground state, 1 = isomeric state).
 +
*Elements are entered using symbol or numerical (U, 92, etc.). The output includes the sum over all isotopes.
 +
*Special entries include:
 +
 
 +
::{| class="wikitable" style="text-align: left;"
 +
! Key-word
 +
! Description
 +
|-
 +
| <tt>all</tt>
 +
| all nuclides
 +
|-
 +
| <tt>accident</tt>
 +
| nuclides with significant health impact & high migration probability in accident conditions<ref>''"Chernobyl: Assessment of Radiological and Health Impacts"'',  OECD/NEA2002 [https://www.oecd-nea.org/jcms/pl_13598/chernobyl-assessment-of-radiological-and-health-impacts-2002 (2002)]</ref>
 +
|-
 +
| <tt>actinides</tt>
 +
| actinides (Z>88) for which cross section data are found in JEFF-3.1.1
 +
|-
 +
| <tt>burnupcredit</tt>
 +
| nuclides commonly considered in burnup credit criticality analyses for PWR fuels <ref name="soar">''"Spent Nuclear Fuel Assay Data for Isotopic Validation"'', NEA/NSC/WPNCS/DOC(2011)5 [http://www.oecd-nea.org/science/wpncs/ADSNF/SOAR_final.pdf (2011)]</ref>
 +
|-
 +
| <tt>burnupindicators</tt>
 +
| burnup indicators (commonly measured from spent fuel) <ref name="soar"/>
 +
|-
 +
| <tt>cosi6</tt>
 +
| inventory list used by the COSI6 code (excluding lumped fission products)
 +
|-
 +
| <tt>lanthanides</tt>
 +
| lanthanides (56<Z<72) for which cross section data are found in JEFF-3.1.1
 +
|-
 +
| <tt>longterm</tt>
 +
| relevant radionuclides in long-term waste analyses <ref>"''The identification of radionuclides relevant to long-term waste management in th United Kingdom''", Nirex Report no. N/105 [https://webarchive.nationalarchives.gov.uk/ukgwa/20211004151355/https://rwm.nda.gov.uk/publication/the-identification-of-radionuclides-relevant-to-long-term-waste-management-in-the-united-kingdom-nirex-report-n105-november-2004/ (2004)]</ref>
 +
|-
 +
| <tt>minoractinides</tt>
 +
| minor actinides (actinides - thorium - uranium - plutonium) for which cross section data are found in JEFF-3.1.1
 +
|-
 +
| <tt>fp</tt>
 +
| fission products
 +
|-
 +
| <tt>dp</tt>
 +
| actinide decay products
 +
|-
 +
| <tt>ng</tt>
 +
| noble gases
 +
|}
 +
*The detailed list of nuclides associated with each special entry: [[Pre-defined inventory lists|Pre-defined inventory lists]]
  
 
  '''set inventory''' '''top''' ''N'' ''PARA''
 
  '''set inventory''' '''top''' ''N'' ''PARA''
  
Specifies which nuclides or elements to include in the <tt>[input]_dep.m</tt> output file. Input values:
+
Defines the criterion or variable based on which to include the most significant contributors under that category in the depletion output file <tt>[input]_dep.m</tt>. Input values:
  
 
{|
 
{|
| <tt>''MAT<sub>n</sub>''</tt>
 
| : Nuclide or element to include
 
|-
 
 
| <tt>''N''</tt>
 
| <tt>''N''</tt>
 
| : Number of nuclides
 
| : Number of nuclides
Line 4,774: Line 5,566:
  
 
<u>Notes:</u>
 
<u>Notes:</u>
*MAT can be: an isotope, element, or a special entry.
+
*The contribution criterion is based on the variables evaluated in the depletion calculation and outputted in the bunurp output.
*Isotopes are entered using element symbol and mass number (e.g U-235, Am-242m, etc.) or ZAI (922350, 952421, etc.). In the ZAI format the last digit refers to the isomeric state (0 = ground state, 1 = isomeric state).
+
*The possible key-words/variables are:
*Elements are entered using symbol or numerical (U, 92, etc.). The output includes the sum over all isotopes.
+
::{| class="wikitable" style="text-align: left;"
*Special entries include:
+
! Key-word
**"all" -- All nuclides.
+
! Description
**"accident" -- Nuclides with significant health impact & high migration probability in accident conditions. Source [http://www.oecd-nea.org/rp/chernobyl/c02.html Chernobyl: Assessment of Radiological and Health Impact]
+
|-
**"actinides" -- Actinides (Z>88) for which cross section data are found in JEFF-3.1.1
+
| <tt>mass</tt>
**"burnupcredit" -- Nuclides commonly considered in burnup credit criticality analyses for PWR fuels [http://www.oecd-nea.org/science/wpncs/ADSNF/SOAR_final.pdf]
+
| contribution to mass fraction
**"burnupindicators" --  Burnup indicators (commonly measured from spent fuel) [http://www.oecd-nea.org/science/wpncs/ADSNF/SOAR_final.pdf]
+
|-
** "cosi6" -- Inventory list used by the COSI6 code (excluding lumped fission products)
+
| <tt>activity</tt>
** "lanthanides" -- Lanthanides (56<Z<72) for which cross section data are found in JEFF-3.1.1
+
| contribution to activity
** "longterm" -- Relevant radionuclides in long-term waste analyses <ref>"''The identification of radionuclides relevant to long-term waste management in th United Kingdom''", Nirex Report no. N/105 (2004)</ref>
+
|-
** "minoractinides" -- Minor actinides (actinides - thorium - uranium - plutonium) for which cross section data are found in JEFF-3.1.1
+
| <tt>dh</tt>
** "fp" -- Fission products
+
| contribution to decay heat
** "dp" -- Actinide decay products
+
|-
** "ng" -- Noble gases
+
| <tt>sf</tt>
*The second syntax allows listing the most significant contributors to a given result. The parameters are:
+
| contribution to spontaneous fission rate
**"mass" -- contribution to mass fraction
+
|-
**"activity" -- contribution to activity
+
| <tt>gsrc</tt>
**"sf" -- contribution to spontaneous fission rate
+
| contribution to gamma emission rate
**"gsrc" -- contribution to gamma emission rate
+
|-
**"dh" -- contribution to decay heat
+
| <tt>ingtox</tt>
**"ingtox" -- contribution to ingestion toxicity
+
| contribution to ingestion toxicity
**"inhtox" -- contribution to inhallation toxicity
+
|-
*for example "top 10 dh" gives the top 10 contributors to decay heat.
+
| <tt>inhtox</tt>
*Since most significant contributors may change over time, the output may contain more nuclides than is requested in the input card.
+
| contribution to inhalation toxicity
*The calculation of top contributors does not work with the "-rdep" command line option.
+
|}
 +
*For example: "top 10 dh" gives the top 10 contributors to decay heat.
 +
*The special entries and the calculation of top contributors do not work with the re-depletion [[Installing and running Serpent#Running Serpent|<tt>''-rdep''</tt>]] command line option.
  
 
=== set isobra ===
 
=== set isobra ===
Line 4,822: Line 5,616:
 
<u>Notes:</u>
 
<u>Notes:</u>
 
*Serpent uses [[Default isomeric branching ratios|constant branching ratios]] by default. This option overrides the default values.
 
*Serpent uses [[Default isomeric branching ratios|constant branching ratios]] by default. This option overrides the default values.
*Energy-dependent data read read from ENDF format files defined by the [[#set bralib|set bralib]] overrides the constant ratios.
+
*Energy-dependent data read read from ENDF format<ref name="endf" /> files defined by the [[#set bralib|set bralib]] overrides the constant ratios.
  
 
=== set iter alb ===
 
=== set iter alb ===
Line 4,858: Line 5,652:
  
 
  '''set keff''' ''K<sub>EFF</sub>''
 
  '''set keff''' ''K<sub>EFF</sub>''
Option to scale fission neutron production in external source simulations.
+
Option to scale fission neutron production in external source simulations. Input values:
  
 
{|
 
{|
 
| <tt>''K<sub>EFF</sub>''</tt>
 
| <tt>''K<sub>EFF</sub>''</tt>
| : Fission neutron production scaling factor. Default value is 1.
+
| : Fission neutron production scaling factor (default value: 1.0)
 
|}
 
|}
  
 
<u>Notes:</u>
 
<u>Notes:</u>
*The k-effective to use for scaling the fission neutron production. Inverse of <tt>''K<sub>EFF</sub>''</tt> will be used as a multiplicative constant for the nubar, i.e. a value of 2.0 will cut the fission neutron production in half.  
+
*The k-effective to use for scaling the fission neutron production. The inverse of <tt>''K<sub>EFF</sub>''</tt> is used as a multiplicative constant for the nubar, i.e. a value of 2.0 will cut the fission neutron production in half.  
*Affects both prompt and delayed neutron production from fissions.
+
*The option affects both prompt and delayed neutron production from fissions.
*In versions '''prior to 2.1.31''': does not affect delayed neutron precursor production, which will cause unexpected behaviour in [[Transient simulations]] that track delayed neutron precursor concentrations.
+
*In versions prior to 2.1.31, it does not affect delayed neutron precursor production, which will cause unexpected behaviour in [[Transient simulations]] that track delayed neutron precursor concentrations.
  
 
=== set lossrate ===
 
=== set lossrate ===
Line 4,874: Line 5,668:
 
  '''set lossrate''' ''L'' [ ''MAT'' ]
 
  '''set lossrate''' ''L'' [ ''MAT'' ]
  
Sets normalization to total loss rate.
+
Sets normalization to total loss rate. Input values:
  
 
{|
 
{|
 
| <tt>''L''</tt>
 
| <tt>''L''</tt>
| : number of lost neutrons per second (neutrons/s)
+
| : number of lost neutrons per second [in neutrons/s]
 
|-
 
|-
 
| <tt>''MAT''</tt>
 
| <tt>''MAT''</tt>
Line 4,887: Line 5,681:
 
*Normalization is needed to relate the Monte Carlo reaction rate estimates to a user-given parameter.
 
*Normalization is needed to relate the Monte Carlo reaction rate estimates to a user-given parameter.
 
*Loss rate includes absorption rate and leakage.
 
*Loss rate includes absorption rate and leakage.
*Neutron transport simulations are by default normalized to unit total loss rate.
+
*The default normalization:
*Photon transport simulations are by default normalized to unit total source rate.
+
**It is set to unit total loss rate (neutron transport) and to unit total source rate (photon transport).
 +
**In coupled neutron-photon transport simulations the normalization is driven by the neutron normalization.
 
*For other normalization options, see: [[#set power|set power]], [[#set powdens|set powdens]], [[#set powdens|set flux]], [[#set genrate|set genrate]], [[#set fissrate|set fissrate]], [[#set absrate|set absrate]], [[#set srcrate|set srcrate]], [[#set sfrate|set sfrate]].
 
*For other normalization options, see: [[#set power|set power]], [[#set powdens|set powdens]], [[#set powdens|set flux]], [[#set genrate|set genrate]], [[#set fissrate|set fissrate]], [[#set absrate|set absrate]], [[#set srcrate|set srcrate]], [[#set sfrate|set sfrate]].
 
*See also Section 5.8 of [http://montecarlo.vtt.fi/download/Serpent_manual.pdf Serpent 1 User Manual].
 
*See also Section 5.8 of [http://montecarlo.vtt.fi/download/Serpent_manual.pdf Serpent 1 User Manual].
Line 4,899: Line 5,694:
 
{|
 
{|
 
| <tt>''LIM''</tt>
 
| <tt>''LIM''</tt>
| : maximum number of collisions allowed in undefined regions or -1 if no limit is set.
+
| : maximum number of collisions allowed in undefined regions or "<tt>-1</tt>" if no limit is set (default value: 0)
 
|}
 
|}
  
Line 4,915: Line 5,710:
 
{|
 
{|
 
| <tt>''MAX''</tt>
 
| <tt>''MAX''</tt>
| : maximum number of splits
+
| : maximum number of splits (default value: 1.0E4)
 
|-
 
|-
 
| <tt>''MIN''</tt>
 
| <tt>''MIN''</tt>
| : minimum survival probability in Russian roulette
+
| : minimum survival probability in Russian roulette (default value: 1.0E-18)
 
|}
 
|}
  
Line 4,932: Line 5,727:
 
{|
 
{|
 
| <tt>''N''</tt>
 
| <tt>''N''</tt>
| : batch size in double floats (8 bytes)
+
| : batch size in double floats, i.e. 8 bytes (default value: 1E4)
 
|}
 
|}
  
 
<u>Notes:</u>
 
<u>Notes:</u>
 +
*The batch size determines the division of data blocks in MPI data transfer. The value may have some effect on parallel performance.
  
*The batch size determines the division of data blocks in MPI data transfer. The value may have some effect on parallel performance.  
+
=== set mcleak ===
*The batch size is set to 10000 by default.
+
 
 +
'''set mcleak''' ''OPT'' [ ''ERG'' ]
 +
Option that enables the implicit leakage correction via MC-Fundamental Mode. Input values:
 +
 
 +
{|
 +
| <tt>''OPT''</tt>
 +
|: option to switch the calculation on (1/yes) or off (0/no). The default option is "<tt>off</tt>"
 +
|-
 +
| <tt>''ERG''</tt>
 +
|: intermediate multi-group structure used for group constant generation (default value: [[Pre-defined energy group structures#Default multi-group structure|Default multi-group structure]], 70 energy-groups)
 +
|}
 +
 
 +
<u>Notes:</u>
 +
*Requires group constant generation to be set on (see [[#set gcu|set gcu]]).
 +
*The calculation mode deactivates any other leakage correction option defined in [[#set fum|set fum]].
 +
*The Monte Carlo transport process is modified to produce inherently critical spectrum burnup calculations and, therefore, all results estimates are leakage corrected.
 +
*The FM-leakage-modified transport equation is solved with continuous energy Monte Carlo and the data is directly tallied to the preferred few-group structure with exception of the diffusion coefficients. The diffusion coefficients used in the evaluation are based on a multi-group CMM approach.
 +
*The implicit leakage correction via MC-Fundamental Mode methodology is described in a related paper<ref name="mckeak">Valtavirta, V. and Leppänen, J. ''"A novel Monte Carlo leakage correction for Serpent 2"'', In ''proceedings of'' ANS M&C 2021. Raleigh, NC, USA. October 3-7, 2021</ref>.
  
 
=== set mcvol ===
 
=== set mcvol ===
  
  '''set mcvol''' ''NP''
+
  '''set mcvol''' ''NP'' [ ''DENS'' ]
 
Runs the Monte Carlo volume-checker routine to set material volumes before running the transport simulation. Input values:
 
Runs the Monte Carlo volume-checker routine to set material volumes before running the transport simulation. Input values:
  
Line 4,948: Line 5,761:
 
| <tt>''NP''</tt>
 
| <tt>''NP''</tt>
 
| : number of points sampled in the geometry
 
| : number of points sampled in the geometry
 +
|-
 +
| <tt>''DENS''</tt>
 +
| : option to set material density adjustment on (1/yes) or off (0/no). The default option is "<tt>off</tt>"
 
|}
 
|}
  
Line 4,953: Line 5,769:
  
 
*The [[Installing and running Serpent#Monte Carlo volume calculation routine|Monte Carlo based volume calculation routine]] works by sampling random points in the geometry, and counting the number of hits in every material.
 
*The [[Installing and running Serpent#Monte Carlo volume calculation routine|Monte Carlo based volume calculation routine]] works by sampling random points in the geometry, and counting the number of hits in every material.
*When invoked, all material volumes are overridden by the results given by the checker routine.
+
*When invoked, all materials given volumes are overridden by the results given by the checker routine (MC-estimated volume).
*The volume checker can also be used to produce a separate input file for the volume entries (see detailed description on [[defining material volumes]]).
+
*The [[Installing and running Serpent#Running Serpent|''-checkvolumes'']] command line option can also be used to produce a separate input file for the volume entries (see detailed description on [[defining material volumes]]).
 +
*The <tt>''DENS''</tt> option enables the adjustment of material densities by the ratio of given and MC-estimated volume. The adjustment is only applied to materials with given volumes. (Implemented to preserve masses in Voronoi geometries ([[#voro (stochastic Voronoi tessellation geometry definition)|voro]] card), but could be also applied to account for thermal expansion).
  
 
=== set mdep ===
 
=== set mdep ===
Line 4,969: Line 5,786:
 
|-
 
|-
 
| <tt>''VOL''</tt>
 
| <tt>''VOL''</tt>
| : volume of the universe
+
| : volume of the universe [in cm<sup>3</sup>] (3D geometry) or cross-sectional area [in cm<sup>2</sup>] (2D geometry)
 
|-
 
|-
 
| <tt>''N''</tt>
 
| <tt>''N''</tt>
Line 4,991: Line 5,808:
 
*The results can be included in the [[Description_of_output_files#Group_constant_output|group constant output]] by adding <tt>MDEP_XS</tt> in the [[#set_coefpara|set coefpara]] list.
 
*The results can be included in the [[Description_of_output_files#Group_constant_output|group constant output]] by adding <tt>MDEP_XS</tt> in the [[#set_coefpara|set coefpara]] list.
 
*If the number of materials is zero, the calculation is carried over all burnable materials.
 
*If the number of materials is zero, the calculation is carried over all burnable materials.
*If the list of nuclides and reactions is substituted by "all", Serpent will generate the micro-depletion output including all the nuclides and all reactions involved in the calculation at beginning of the  simulation aiming to the extract the constant data (stopping the calculation right after).
+
*If the list of nuclides and reactions is substituted by "all", Serpent will generate a micro-depletion output <tt>[input]_mdep.inc</tt> including all the nuclides and all reactions involved in the calculation at beginning of the  simulation aiming to the extract the constant data (stopping the calculation right after).
 
*The listed materials must be enclosed inside the homogenized universe.
 
*The listed materials must be enclosed inside the homogenized universe.
 
*The calculation requires the [[Defining material volumes|material volumes to be correctly set]].
 
*The calculation requires the [[Defining material volumes|material volumes to be correctly set]].
Line 5,007: Line 5,824:
 
  '''set memfrac''' ''FRAC''
 
  '''set memfrac''' ''FRAC''
  
Defines the fraction of total system memory Serpent can allocate to its use. If the fraction is exceeded, the simulation will abort. This is mainly to avert the use of swap-memory, which can make the system unresponsive. Input values:
+
Defines the fraction of total system memory Serpent can allocate to its use. Input values:
  
 
{|
 
{|
 
| <tt>''FRAC''</tt>
 
| <tt>''FRAC''</tt>
| : the fraction of system memory that Serpent is allowed to use (between 0 and 1)
+
| : the fraction of system memory that Serpent is allowed to use (default value: 0.8)
 
|}
 
|}
  
 
<u>Notes:</u>
 
<u>Notes:</u>
* The default fraction is 0.8.
+
*Serpent tries to read the system total memory from the first line of the file /proc/meminfo
* The fraction can also be set via a SERPENT_MEM_FRAC environmental variable.
+
*The fraction can also be set via a <tt>SERPENT_MEM_FRAC</tt> environmental variable.
* Serpent tries to read the system total memory from the first line of the file /proc/meminfo
+
*If the fraction is exceeded, the simulation will abort. This is mainly to avert the use of swap-memory, which can make the system unresponsive.
  
 
=== set mfpcut ===
 
=== set mfpcut ===
Line 5,026: Line 5,843:
 
{|
 
{|
 
| <tt>''MFPMIN<sub>p</sub>''</tt>
 
| <tt>''MFPMIN<sub>p</sub>''</tt>
| : cut-off mean-free-path for photons (cm)
+
| : cut-off mean-free-path for photons [in cm]
 
|}
 
|}
  
Line 5,036: Line 5,853:
 
{|
 
{|
 
| <tt>''DENS<sub>i</sub>''</tt>
 
| <tt>''DENS<sub>i</sub>''</tt>
| : mass density (g/cm<sup>3</sup>)
+
| : mass density [g/cm<sup>3</sup>]
 
|-
 
|-
 
| <tt>''MFPMIN<sub>p,i</sub>''</tt>
 
| <tt>''MFPMIN<sub>p,i</sub>''</tt>
| : cut-off mean-free-path for photons (cm)
+
| : cut-off mean-free-path for photons [in cm]
 
|}
 
|}
  
Line 5,055: Line 5,872:
 
|-
 
|-
 
| <tt>''MFPMIN<sub>p,i</sub>''</tt>
 
| <tt>''MFPMIN<sub>p,i</sub>''</tt>
| : cut-off mean-free-path for photons (cm)
+
| : cut-off mean-free-path for photons [in cm]
 
|}
 
|}
  
Line 5,061: Line 5,878:
  
 
  '''set micro''' ''ERG'' [ ''BTCH'' ]
 
  '''set micro''' ''ERG'' [ ''BTCH'' ]
Defines the intermediate multi-group structure used for group constant generation.
+
Defines the intermediate multi-group structure used for group constant generation. Input values:
  
 
{|
 
{|
 
| <tt>''ERG''</tt>
 
| <tt>''ERG''</tt>
| : Intermediate multi-group structure used for group constant generation
+
| : Intermediate multi-group structure used for group constant generation (default value: [[Pre-defined energy group structures#Default multi-group structure|Default multi-group structure]], 70 energy-groups)
 
|-
 
|-
 
| <tt>''BTCH''</tt>
 
| <tt>''BTCH''</tt>
| : When set to 2, results are averaged over all criticality cycles
+
| : When set to "2", results are averaged over all criticality cycles
 
|}
 
|}
  
Line 5,089: Line 5,906:
 
{|
 
{|
 
| <tt>''OPT''</tt>
 
| <tt>''OPT''</tt>
| : option to switch calculation on (1/yes) or off (0/no). Default is off.
+
| : option to switch calculation on (1/yes) or off (0/no). The default option is "<tt>off</tt>"
 
|}
 
|}
  
Line 5,103: Line 5,920:
 
{|
 
{|
 
| <tt>''MAT<sub>m</sub>''</tt>
 
| <tt>''MAT<sub>m</sub>''</tt>
| : name of material ''m''
+
| : name of ''m''-th material
 
|-
 
|-
 
| <tt>''ZONE<sub>m,n</sub>''</tt>
 
| <tt>''ZONE<sub>m,n</sub>''</tt>
| : index of zone ''n'' in material ''m''
+
| : index of ''n''-th zone in ''m''-th material
 
|-
 
|-
 
| <tt>''VOL<sub>m,n</sub>''</tt>
 
| <tt>''VOL<sub>m,n</sub>''</tt>
| : volume of zone ''n'' in material ''m''
+
| : volume of ''n''-th zone in ''m''-th material [in cm<sup>3</sup>] (3D geometry) or cross sectional area [in cm<sup>2</sup>] (2D geometry)
 
|}
 
|}
  
 
<u>Notes:</u>
 
<u>Notes:</u>
  
*This option is used to define material volumes manually. The input card is also produced when the [[Installing and running Serpent#Monte Carlo volume calculation routine|Monte Carlo based volume checker routine]] is invoked.
+
*This option is used to define material volumes manually.  
*The zone index is related to [[automated depletion zone division]], invoked by the [[#div (divisor definition)|div]] card. If no division is used, the index must be set to zero for non-burnable materials. For burnable materials the indexing starts from 1. If no division is used, the corresponding index can be found using the [[Installing and running Serpent#Monte Carlo volume calculation routine|Monte Carlo based volume checker routine]].
+
*The input card, as a separate file, is also produced when the [[Installing and running Serpent#Monte Carlo volume calculation routine|<tt>''-checkvolumes''</tt>]] command line option is launched, invoking the Monte Carlo checker routine. The file can be added to the input using the [[#include|include card]].
*Another option to define material volumes is to use the vol entry in the [[#mat (material definition)|material card]].
+
*The zone index is related to [[automated depletion zone division]], invoked by the [[#div (divisor definition)|div]] card.  
 +
**If no division is used, the index must be set to "0" for non-burnable materials.  
 +
**For burnable materials the indexing starts from "1".
 +
**The corresponding index can be found using the [[Installing and running Serpent#Monte Carlo volume calculation routine|<tt>''-checkvolumes''</tt>]] command line option (listing all materials and their associated indexes)
 +
*Alternative options to define the material volumes are: manually enter the volume via the '''vol''' entry in the [[#mat (material definition)|mat card]] or automatically evaluation via [[#set mcvol|set mcvol]] option, invoking at runtime the Monte Carlo checker routine.
 
*For more infomation, see detailed description on [[Defining material volumes|the definition of material volumes]].
 
*For more infomation, see detailed description on [[Defining material volumes|the definition of material volumes]].
  
Line 5,145: Line 5,966:
 
  '''set nfg''' ''ERG''
 
  '''set nfg''' ''ERG''
  
Defines the few-group structure used for group constant generation.
+
Defines the few-group structure used for group constant generation. Input values
  
 
{|
 
{|
 
| <tt>''ERG''</tt>
 
| <tt>''ERG''</tt>
| : Name of the few-group structure used for group constant generation
+
| : Name of the few-group structure used for group constant generation (default value: [[Pre-defined energy group structures#Default 2-group structure|Default 2-group structure]], 2 energy-groups)
 
|}
 
|}
  
 
  '''set nfg''' ''NE''
 
  '''set nfg''' ''NE''
  
Defines the few-group structure used for group constant generation. This syntax should not be used anymore.
+
Defines the few-group structure used for group constant generation. (This syntax should not be used anymore). Input values:
  
 
{|
 
{|
Line 5,163: Line 5,984:
 
  '''set nfg''' ''NE E<sub>NE-1</sub> E<sub>NE-2</sub> ... E<sub>1</sub>''
 
  '''set nfg''' ''NE E<sub>NE-1</sub> E<sub>NE-2</sub> ... E<sub>1</sub>''
  
Defines the few-group structure used for group constant generation. This syntax should not be used anymore.
+
Defines the few-group structure used for group constant generation. (This syntax should not be used anymore). Input values:
  
 
{|
 
{|
Line 5,170: Line 5,991:
 
|-
 
|-
 
| <tt>''E<sub>N</sub>''</tt>
 
| <tt>''E<sub>N</sub>''</tt>
| : Energy group boundary value between groups ''N'' and ''N+1''. Values have to be given in ascending order.
+
| : Energy group boundary value between groups ''N'' and ''N+1'' [in MeV]. Values have to be given in ascending order.
 
|}
 
|}
  
Line 5,176: Line 5,997:
  
 
*The few-group structure may be an energy grid defined using the [[#ene (energy grid definition)|ene card]] or a name of a [[pre-defined energy group structures|pre-defined energy group structure]].
 
*The few-group structure may be an energy grid defined using the [[#ene (energy grid definition)|ene card]] or a name of a [[pre-defined energy group structures|pre-defined energy group structure]].
*The default is a two-group structure with boundary between fast and thermal group set to 0.625 eV.
 
*Serpent uses an intermediate multi-group structure in the calculation. The default structure consists of 70 groups, and can be changed using the [[#set micro|set micro]] or [[#set fum|set fum]] options.
 
*Note that in general the intermediate multi-group structure should have more groups than the few-group structure to get reasonable results for [[#set fum|leakage corrected group constants]] and out-scatter diffusion coefficients.
 
 
*The few-group structure must be a sub-set of the intermediate multi-group structure.
 
*The few-group structure must be a sub-set of the intermediate multi-group structure.
 +
*The default is a two-group structure with boundary between fast and thermal group set to 0.625 eV (see [[Pre-defined energy group structures#Default 2-group structure|Default 2-group structure]])).
 +
**Serpent uses an intermediate multi-group structure in the calculation. The default structure consists of 70 groups (see [[Pre-defined energy group structures#Default multi-group structure|Default multi-group structure]]), and can be changed using the [[#set micro|set micro]] or [[#set fum|set fum]] options.
 +
**Note that in general the intermediate multi-group structure should have more groups than the few-group structure to get reasonable results for [[#set fum|leakage corrected group constants]] and out-scatter diffusion coefficients.
 
*See also: [[Output parameters#Homogenized group constants|group constant output]].
 
*See also: [[Output parameters#Homogenized group constants|group constant output]].
 
*The few-group structure will also be used for example for [[#set adf|assembly discontinuity factors]] and [[#set alb|albedos]].
 
*The few-group structure will also be used for example for [[#set adf|assembly discontinuity factors]] and [[#set alb|albedos]].
Line 5,195: Line 6,016:
 
<u>Notes:</u>
 
<u>Notes:</u>
  
*Fission yield libraries are standard ENDF format files containing neutron-induced fission yield data.
+
*Fission yield libraries are standard ENDF format<ref name="endf" /> files containing neutron-induced fission yield data.
 
*If the file path contains special characters it is advised to enclose it within quotes.
 
*If the file path contains special characters it is advised to enclose it within quotes.
 
*A default directory path can be set by defining environment variable <tt>SERPENT_DATA</tt>. The code looks for fission yield data files in this path if not found at the absolute.
 
*A default directory path can be set by defining environment variable <tt>SERPENT_DATA</tt>. The code looks for fission yield data files in this path if not found at the absolute.
Line 5,207: Line 6,028:
 
{|
 
{|
 
| <tt>''MODE''</tt>
 
| <tt>''MODE''</tt>
| : simulation mode (0 = off, 1 = analog, 2 = implicit)
+
| : simulation mode (0 = off, 1 = analog, 2 = implicit). The default option is "<tt>off</tt>"
 
|-
 
|-
 
| <tt>''WMIN''</tt>
 
| <tt>''WMIN''</tt>
Line 5,231: Line 6,052:
 
{|
 
{|
 
| <tt>''FISS''</tt>
 
| <tt>''FISS''</tt>
| : option to handle fission (0 = not handled, 1 = handled)
+
| : option to handle fission (0 = not handled, 1 = handled). The default option is "<tt>on</tt>"
 
|-
 
|-
 
| <tt>''CAPT''</tt>
 
| <tt>''CAPT''</tt>
| : option to handle capture (0 = not handled, 1 = handled)
+
| : option to handle capture (0 = not handled, 1 = handled). The default option is "<tt>on</tt>"
 
|-
 
|-
 
| <tt>''SCATT''</tt>
 
| <tt>''SCATT''</tt>
| : option to handle scattering (0 = not handled, 1 = handled)
+
| : option to handle scattering (0 = not handled, 1 = handled). The default option is "<tt>on</tt>"
 
|}
 
|}
  
 
<u>Notes:</u>
 
<u>Notes:</u>
 
+
*If fission is switched "<tt>off</tt>", it is handled as capture.
*All reaction modes are handled by default.
+
*If fission is switched off, it is handled as capture.
+
  
 
=== set nps ===
 
=== set nps ===
Line 5,255: Line 6,074:
 
|-
 
|-
 
| <tt>''BTCH''</tt>
 
| <tt>''BTCH''</tt>
| : number of batches
+
| : number of batches (default value: 200)
 
|-
 
|-
 
| <tt>''TBI''</tt>
 
| <tt>''TBI''</tt>
Line 5,264: Line 6,083:
  
 
*The total number of particles is divided by the given number of batches to give the number of particles per batch.
 
*The total number of particles is divided by the given number of batches to give the number of particles per batch.
*Default number of batches is 200.
+
*Using the nps card sets the mode to external source simulation.  
*Using the nps card sets the mode to external source simulation. Criticality source simulation for neutrons is invoked using [[#set pop|set pop]]. (The two cards are mutually exclusive).
+
**Criticality source simulation for neutrons is invoked using [[#set pop|set pop]]. (The two cards are mutually exclusive).
*If time binning is provided, the simulation is run in the [[Dynamic external source simulation mode|dynamic mode]] with sequential population control. The bin structure is defined using the [[#tme (time binning definition)|tme card]].
+
 
*Running an external source simulation requires a source, defined by the [[#src|src card]]. Source definition also sets the transported particle type.
 
*Running an external source simulation requires a source, defined by the [[#src|src card]]. Source definition also sets the transported particle type.
*Neutron external source simulations are limited to sub-critical systems, unless dynamic mode, time cut-off ([[#set tcut|set tcut]]) or generation cut-off ([[#set gcut|set gcut]]) is invoked.
+
*Delayed neutron emission is switched "<tt>off</tt> by default in neutron external source simulation. Delayed neutrons can be included with [[#set delnu|set delnu]].
*Neutron external source simulations in multiplying systems may require adjusting the neutron buffer ([[#set nbuf|set nbuf]]).
+
*If time binning is provided, the simulation is run in the [[Dynamic external source simulation mode|dynamic mode]] with sequential population control. The bin structure is defined using the [[#tme (time binning definition)|tme card]].
*Delayed neutron emission is switched off by default in neutron external source simulation. Delayed neutrons can be included with [[#set delnu|set delnu]].
+
 
*In transient simulations, where an initial transient source is linked using the [[#set dynsrc|set dynsrc]] option, <tt>''PP''</tt> particles are sampled for each time interval.
 
*In transient simulations, where an initial transient source is linked using the [[#set dynsrc|set dynsrc]] option, <tt>''PP''</tt> particles are sampled for each time interval.
 +
*Neutron external source simulations:
 +
**are limited to sub-critical systems, unless dynamic mode, time cut-off ([[#set tcut|set tcut]]) or generation cut-off ([[#set gcut|set gcut]]) is invoked.
 +
**in multiplying systems, may require adjusting the neutron buffer ([[#set nbuf|set nbuf]]).
  
 
=== set opti ===
 
=== set opti ===
Line 5,281: Line 6,101:
 
{|
 
{|
 
| <tt>''MODE''</tt>
 
| <tt>''MODE''</tt>
| : optimization mode
+
| : optimization mode (default value: 4)
 
|}
 
|}
  
 
The possible settings for mode are:
 
The possible settings for mode are:
{| class="wikitable" style="text-align: left;"
+
::{| class="wikitable" style="text-align: left;"
 
! <tt>''MODE''</tt>
 
! <tt>''MODE''</tt>
 
! Description
 
! Description
 +
! Usage
 
|-
 
|-
 
| <tt>1</tt>
 
| <tt>1</tt>
| Minimum optimization and small memory usage. Suitable for very large burnup calculation problems involving tens or hundreds of thousands of depletion zones.
+
| Minimum optimization and small memory usage
 +
| Suitable for very large burnup calculation problems involving tens or hundreds of thousands of depletion zones
 
|-
 
|-
 
| <tt>2</tt>
 
| <tt>2</tt>
| Good performance in burnup calculations involving several thousand depletion zones. Suitable for research reactor applications, but not the best choice for group constant generation.
+
| Good performance in burnup calculations involving several thousand depletion zones  
 +
|Suitable for research reactor applications, but not the best choice for group constant generation
 
|-
 
|-
 
| <tt>3</tt>
 
| <tt>3</tt>
| Similar to mode 4, but lower memory demand. CPU time required by burnup and processing routines increases steeply along with the number of depletion zones, which makes the mode better suited for small burnup calculation problems.
+
| Similar to mode 4, but lower memory demand
 +
| Suitable for small burnup calculation problems
 
|-
 
|-
 
| <tt>4</tt>
 
| <tt>4</tt>
| Maximum performance at the cost of memory usage. Suitable for group constant generation and 2D assembly burnup calculations with a limited number of depletion zones.
+
| Maximum performance at the cost of memory usage
 +
|Suitable for group constant generation and 2D assembly burnup calculations with a limited number of depletion zones
 
|}
 
|}
  
 
<u>Notes:</u>
 
<u>Notes:</u>
*Default value for the optimization mode is 4.
 
 
*The mode 4 is essentially the same as the methodology in Serpent 1.
 
*The mode 4 is essentially the same as the methodology in Serpent 1.
 
*The methodology is described in a paper presented in PHYSOR 2012<ref name="opti">Leppänen, J. and Isotalo, A. ''"Burnup calculation methodology in the Serpent 2 Monte Carlo code."'' In proc. PHYSOR 2012, Knoxville, TN, Apr. 15-20, 2012.</ref>.
 
*The methodology is described in a paper presented in PHYSOR 2012<ref name="opti">Leppänen, J. and Isotalo, A. ''"Burnup calculation methodology in the Serpent 2 Monte Carlo code."'' In proc. PHYSOR 2012, Knoxville, TN, Apr. 15-20, 2012.</ref>.
Line 5,314: Line 6,138:
 
{|
 
{|
 
| <tt>''INT''</tt>
 
| <tt>''INT''</tt>
| : number of cycles after which the output-files are updated
+
| : number of cycles after which the output-files are updated (default value: 50)
 
|}
 
|}
  
 
<u>Notes:</u>
 
<u>Notes:</u>
*Default value is 50.
+
*In coupled transient simulations the interval refers to time steps rather than batches.
*In coupled transient simulations the interval refers to time steps rather than batches. By default, the output is written after every 50th time step.
+
 
*Affects files such as <tt>[input]_res.m</tt> and <tt>[input]_det.m</tt> as well as mesh plots.
 
*Affects files such as <tt>[input]_res.m</tt> and <tt>[input]_det.m</tt> as well as mesh plots.
  
Line 5,345: Line 6,168:
 
{|
 
{|
 
| <tt>''MODE''</tt>
 
| <tt>''MODE''</tt>
| : time integration method
+
| : time integration method (default value: 1)
 
|-
 
|-
 
| <tt>''PRED''</tt>
 
| <tt>''PRED''</tt>
Line 5,354: Line 6,177:
 
|-
 
|-
 
| <tt>''SSP''</tt>
 
| <tt>''SSP''</tt>
| : number of substeps for predictor steps
+
| : number of substeps for predictor steps (default value: 1)
 
|-
 
|-
 
| <tt>''SSC''</tt>
 
| <tt>''SSC''</tt>
| : number of substeps for corrector steps
+
| : number of substeps for corrector steps (default value: 1)
 
|}
 
|}
  
 
The possible settings for mode are:
 
The possible settings for mode are:
{| class="wikitable" style="text-align: left;"
+
::{| class="wikitable" style="text-align: left;"
 
! Mode
 
! Mode
 
! Predictor method
 
! Predictor method
 
! Corrector method
 
! Corrector method
 +
! Notes
 
|-
 
|-
 
| <tt>0, CE</tt>
 
| <tt>0, CE</tt>
 
| Constant extrapolation
 
| Constant extrapolation
 
| -
 
| -
 +
| Serpent 1 without substeps - "Euler's method"
 
|-
 
|-
 
| <tt>1, CELI</tt>
 
| <tt>1, CELI</tt>
 
| Constant extrapolation
 
| Constant extrapolation
 
| Linear interpolation
 
| Linear interpolation
 +
| Serpent 1 without substeps - "old predictor-corrector method"
 
|-
 
|-
 
| <tt>2, LE</tt>
 
| <tt>2, LE</tt>
 
| Linear extrapolation
 
| Linear extrapolation
 +
| -
 
| -
 
| -
 
|-
 
|-
Line 5,381: Line 6,208:
 
| Linear extrapolation
 
| Linear extrapolation
 
| Linear interpolation
 
| Linear interpolation
 +
| -
 
|-
 
|-
 
| <tt>4, LEQI</tt>
 
| <tt>4, LEQI</tt>
 
| Linear extrapolation
 
| Linear extrapolation
 
| Quadratic interpolation
 
| Quadratic interpolation
 +
| -
 
|-
 
|-
 
| <tt>5</tt>
 
| <tt>5</tt>
 
| Constant or linear extrapolation
 
| Constant or linear extrapolation
 
| Linear or quadratic interpolation
 
| Linear or quadratic interpolation
 +
| -
 
|-
 
|-
 
| <tt>6, CECE</tt>
 
| <tt>6, CECE</tt>
 
| Constant extrapolation
 
| Constant extrapolation
 
| Constant backwards extrapolation
 
| Constant backwards extrapolation
 +
| -
 
|}
 
|}
  
 
<u>Notes:</u>
 
<u>Notes:</u>
*<tt>''MODE''</tt> 0 is also known as "Euler's method".
 
*<tt>''MODE''</tt> 1 is also known as the "old predictor-corrector method".
 
*<tt>''MODE''</tt> 0 and 1 without substeps corresponds to the methods used in Serpent 1.
 
*The default numbers of predictor and corrector substeps (<tt>''SSP''</tt> and <tt>''SSC''</tt>, respectively) are 1 for applicable methods.
 
*The default time integration scheme corresponds to <tt>''MODE''</tt> 1 for applicable methods.
 
 
*Number of substeps could not be given for constant predictor or corrector before 2.1.32.
 
*Number of substeps could not be given for constant predictor or corrector before 2.1.32.
 
*Decay calculations were always calculated with single substep disregarding this input before 2.1.32.
 
*Decay calculations were always calculated with single substep disregarding this input before 2.1.32.
Line 5,418: Line 6,244:
 
<u>Notes:</u>
 
<u>Notes:</u>
  
*Serpent uses auxiliary data files for the modelling of photon interaction physics.
+
*Serpent uses auxiliary data files for the modelling of photon interaction physics: [http://montecarlo.vtt.fi/download/photon_data.zip (photon_data.zip)].
*For more information, see [http://serpent.vtt.fi/mediawiki/index.php/Installing_and_running_Serpent#Setting_up_the_data_libraries instructions on setting up the data libraries].
+
*For more information, see [[Installing_and_running_Serpent#Setting_up_the_data_libraries|instructions on setting up the data libraries]].
  
 
=== set poi ===
 
=== set poi ===
Line 5,428: Line 6,254:
 
{|
 
{|
 
| <tt>''OPT''</tt>
 
| <tt>''OPT''</tt>
| : option to switch the calculation of poison cross sections on (1/yes) or off (0/no)
+
| : option to switch the calculation of poison cross sections on (1/yes) or off (0/no). The default option is "<tt>off</tt>"
 
|-
 
|-
 
| <tt>''VOL''</tt>
 
| <tt>''VOL''</tt>
Line 5,434: Line 6,260:
 
|-
 
|-
 
| <tt>''XE135M''</tt>
 
| <tt>''XE135M''</tt>
| : option to treat <sup>135m</sup>Xe separate from ground state <sup>135</sup>Xe (0 = lumped with <sup>135</sup>Xe, 1 = separate treatment)
+
| : option to treat <sup>135m</sup>Xe separate from ground state <sup>135</sup>Xe (0 = lumped with <sup>135</sup>Xe, 1 = separate treatment). (default value: 0)
 
|}
 
|}
  
Line 5,440: Line 6,266:
 
*Poison cross sections include the fission yields and microscopic and macroscopic absorption cross sections of fission product poisons <sup>135</sup>Xe and <sup>149</sup>Sm, as well as the fission yields and microscopic absorption cross sections of their precursors. The data is part of the [[Output parameters#Homogenized group constants|homogenized group constant output]].
 
*Poison cross sections include the fission yields and microscopic and macroscopic absorption cross sections of fission product poisons <sup>135</sup>Xe and <sup>149</sup>Sm, as well as the fission yields and microscopic absorption cross sections of their precursors. The data is part of the [[Output parameters#Homogenized group constants|homogenized group constant output]].
 
*The calculation requires setting the [[#set declib|decay]] and [[#set nfylib|fission yield]] library file paths.
 
*The calculation requires setting the [[#set declib|decay]] and [[#set nfylib|fission yield]] library file paths.
*The calculation requires the [[Defining material volumes|material volumes to be correctly set]].
 
 
*The volume is required for calculating microscopic absorption cross sections matching macroscopic absorption cross sections with poison nuclide densities smeared to the homogenization volume.
 
*The volume is required for calculating microscopic absorption cross sections matching macroscopic absorption cross sections with poison nuclide densities smeared to the homogenization volume.
*Separate treatment for <sup>135m</sup>Xe requires cross sections for this isotope.
+
**The calculation requires the material volumes correctly set (see [[Defining material volumes|Defining material volumes]]).
*Parameter ''VOL'' was ''VR'' (optional) in versions before 2.1.32 (ratio of fuel volume to the volume of the homogenized zone).
+
**Parameter ''VOL'' was ''VR'' (optional) in versions before 2.1.32 (ratio of fuel volume to the volume of the homogenized zone).
 +
*Separate treatment for <sup>135m</sup>Xe requires cross sections for this isotope (from version 2.2.0 and on).
 +
*From version 2.1.32 and on, the use of the 'set poi' card to evaluate the fission poison cross sections should be limited to homogenized universes enclosing all fissionable materials. Otherwise, use the micro-depletion [[#set mdep|set mdep]] input option.
  
 
=== set pop ===
 
=== set pop ===
Line 5,461: Line 6,288:
 
|-
 
|-
 
| <tt>''K0''</tt>
 
| <tt>''K0''</tt>
| : initial guess for ''k''<sub>eff</sub>
+
| : initial guess for ''k''<sub>eff</sub> (default value: 1.0)
 
|-
 
|-
 
| <tt>''BTCH''</tt>
 
| <tt>''BTCH''</tt>
| : batching interval (default value is 1 generation per batch)
+
| : batching interval (default value: 1)
 
|-
 
|-
 
| <tt>''NEIG''</tt>
 
| <tt>''NEIG''</tt>
| : number of independent parallel eigenvalue calculations (default is 1)
+
| : number of independent parallel eigenvalue calculations (default value: 1)
 
|}
 
|}
  
 
<u>Notes:</u>
 
<u>Notes:</u>
 
*The simulation is first run for a number of inactive generations to allow the fission source to converge. This is followed by a number of active generations, during which the results are collected. The statistics are divided in batches, and by default each generation forms its own batch.
 
*The simulation is first run for a number of inactive generations to allow the fission source to converge. This is followed by a number of active generations, during which the results are collected. The statistics are divided in batches, and by default each generation forms its own batch.
*Using the pop card sets the mode to criticality source simulation. External source simulation is invoked using [[#set nps|set nps]]. (The two cards are mutually exclusive).
+
*Using the pop card sets the mode to criticality source simulation.  
 +
**External source simulation is invoked using [[#set nps|set nps]]. (The two cards are mutually exclusive).
 
*Convergence of fission source can be monitored using Shannon entropy (input parameters [[#set his|set his]] and [[#set entr|set entr]]).
 
*Convergence of fission source can be monitored using Shannon entropy (input parameters [[#set his|set his]] and [[#set entr|set entr]]).
*Initial guess for ''k''<sub>eff</sub> is 1.0 by default. Setting the value manually may get the simulation going if it terminates on the first generation because of poor initial guess. The value does not affect fission source convergence.
+
*Setting an initial guess value manually may get the simulation going if it terminates on the first generation because of poor initial guess. The value does not affect fission source convergence.
 
*See detailed descriptions on [[fission source convergence]] and [[statistical effects of batching]].
 
*See detailed descriptions on [[fission source convergence]] and [[statistical effects of batching]].
 
*The number of neutrons per generation also affects the memory usage together with [[#set nbuf|set nbuf]]. With a large value, lowering the buffer size might be necessary for the simulation to be runnable.
 
*The number of neutrons per generation also affects the memory usage together with [[#set nbuf|set nbuf]]. With a large value, lowering the buffer size might be necessary for the simulation to be runnable.
Line 5,481: Line 6,309:
  
 
  '''set powdens''' ''PDE'' [ ''MAT'' ]
 
  '''set powdens''' ''PDE'' [ ''MAT'' ]
Sets normalization to power density
+
Sets normalization to power density. Input values:
  
 
{|
 
{|
 
| <tt>''PDE''</tt>
 
| <tt>''PDE''</tt>
| : power density (in kW/g)
+
| : power density [in kW/g] (typical value LWR: 20E-3 ... 50E-3)
 
|-
 
|-
 
| <tt>''MAT''</tt>
 
| <tt>''MAT''</tt>
| : material in which the given power is produced
+
| : material in which the given power is produced. The default option is all materials.
 
|}
 
|}
  
Line 5,494: Line 6,322:
 
*Normalization is needed to relate the Monte Carlo reaction rate estimates to a user-given parameter.
 
*Normalization is needed to relate the Monte Carlo reaction rate estimates to a user-given parameter.
 
*If the material name is omitted, the value corresponds to average power density produced in the system.
 
*If the material name is omitted, the value corresponds to average power density produced in the system.
*Power density is given in units of kW/g, not W/g used in several other programs. The typical order of magnitude for this parameter in LWR calculations is 20E-3 ... 50E-3.
+
*The default normalization:
*Neutron transport simulations are by default normalized to unit total loss rate.
+
**It is set to unit total loss rate (neutron transport) and to unit total source rate (photon transport).  
*Photon transport simulations are by default normalized to unit total source rate.
+
**In coupled neutron-photon transport simulations the normalization is driven by the neutron normalization.
 
*For other normalization options, see: [[#set power|set power]], [[#set flux|set flux]], [[#set genrate|set genrate]], [[#set fissrate|set fissrate]], [[#set absrate|set absrate]], [[#set lossrate|set lossrate]], [[#set srcrate|set srcrate]], [[#set sfrate|set sfrate]].
 
*For other normalization options, see: [[#set power|set power]], [[#set flux|set flux]], [[#set genrate|set genrate]], [[#set fissrate|set fissrate]], [[#set absrate|set absrate]], [[#set lossrate|set lossrate]], [[#set srcrate|set srcrate]], [[#set sfrate|set sfrate]].
 
*See also Section 5.8 of [http://montecarlo.vtt.fi/download/Serpent_manual.pdf Serpent 1 User Manual].
 
*See also Section 5.8 of [http://montecarlo.vtt.fi/download/Serpent_manual.pdf Serpent 1 User Manual].
Line 5,503: Line 6,331:
  
 
  '''set power''' ''P'' [ ''MAT'' ]
 
  '''set power''' ''P'' [ ''MAT'' ]
Sets normalization to total fission power
+
Sets normalization to total fission power. Input values:
  
 
{|
 
{|
 
| <tt>''P''</tt>
 
| <tt>''P''</tt>
| : fission power (in W)
+
| : fission power [in W]
 
|-
 
|-
 
| <tt>''MAT''</tt>
 
| <tt>''MAT''</tt>
| : material in which the given power is produced
+
| : material in which the given power is produced. The default option is all materials.
 
|}
 
|}
  
Line 5,516: Line 6,344:
 
*Normalization is needed to relate the Monte Carlo reaction rate estimates to a user-given parameter.
 
*Normalization is needed to relate the Monte Carlo reaction rate estimates to a user-given parameter.
 
*If the material name is omitted, the value corresponds to total fission power produced in the system.
 
*If the material name is omitted, the value corresponds to total fission power produced in the system.
*Neutron transport simulations are by default normalized to unit total loss rate.
+
*The default normalization:
*Photon transport simulations are by default normalized to unit total source rate.
+
**It is set to unit total loss rate (neutron transport) and to unit total source rate (photon transport).
 +
**In coupled neutron-photon transport simulations the normalization is driven by the neutron normalization.
 
*For other normalization options, see: [[#set powdens|set powdens]], [[#set flux|set flux]], [[#set genrate|set genrate]], [[#set fissrate|set fissrate]], [[#set absrate|set absrate]], [[#set lossrate|set lossrate]], [[#set srcrate|set srcrate]], [[#set sfrate|set sfrate]].
 
*For other normalization options, see: [[#set powdens|set powdens]], [[#set flux|set flux]], [[#set genrate|set genrate]], [[#set fissrate|set fissrate]], [[#set absrate|set absrate]], [[#set lossrate|set lossrate]], [[#set srcrate|set srcrate]], [[#set sfrate|set sfrate]].
 
*See also Section 5.8 of [http://montecarlo.vtt.fi/download/Serpent_manual.pdf Serpent 1 User Manual].
 
*See also Section 5.8 of [http://montecarlo.vtt.fi/download/Serpent_manual.pdf Serpent 1 User Manual].
Line 5,528: Line 6,357:
 
{|
 
{|
 
| <tt>''PID''</tt>
 
| <tt>''PID''</tt>
| : Process identifier (PID) of the (parent) process that Serpent should communicate with.
+
| : Process identifier (PID) of the (parent) process that Serpent should communicate with (theoretical maximum for 64-bit system: 2<sup>22</sup>)
 
|}
 
|}
  
Line 5,534: Line 6,363:
  
 
*Setting up a communication mode will enable the coupled calculation mode.
 
*Setting up a communication mode will enable the coupled calculation mode.
*<tt>''PID''</tt> 2<sup>22</sup> theoretical maximum for 64-bit system
 
 
*The communication options [[#set comfile|set comfile]], [[#set ppid|set ppid]] and [[#set pport|set pport]] are mutually exclusive, aka, multiple signalling modes are not allowed.
 
*The communication options [[#set comfile|set comfile]], [[#set ppid|set ppid]] and [[#set pport|set pport]] are mutually exclusive, aka, multiple signalling modes are not allowed.
 
*For more information see: [[Coupled_multi-physics_calculations#External_coupling|External coupling]]
 
*For more information see: [[Coupled_multi-physics_calculations#External_coupling|External coupling]]
Line 5,545: Line 6,373:
 
{|
 
{|
 
| <tt>''PORT''</tt>
 
| <tt>''PORT''</tt>
 +
| : user given parent process port
 
|}
 
|}
  
Line 5,577: Line 6,406:
 
  '''set precsrcf''' ''FACTOR''
 
  '''set precsrcf''' ''FACTOR''
  
Set number of pointwise precursors to hold in memory in [[transient simulations]]. A multiplicative factor for the number of neutrons per batch, i.e. setting ''<tt>FACTOR</tt>'' to 10 when running 1000 neutrons per batch will keep 10000 pointwise precursors in memory.
+
Set number of pointwise precursors to hold in memory in [[transient simulations]]. Input values:
  
 
{|
 
{|
 
| <tt>''FACTOR''</tt>
 
| <tt>''FACTOR''</tt>
| : The factor to multiply the number of neutrons per batch to obtain the number of pointwise precursors to hold in memory.
+
| : The factor to multiply the number of neutrons per batch to obtain the number of pointwise precursors to hold in memory (default value: 10.0)
 
|}
 
|}
  
 
<u>Notes:</u>
 
<u>Notes:</u>
  
*Default value is 10.
+
*A multiplicative factor for the number of neutrons per batch, i.e. setting ''<tt>FACTOR</tt>'' to 10 when running 1000 neutrons per batch will keep 10000 pointwise precursors in memory.
 
*The number of pointwise precursors held in memory is controlled to the requested number at time interval boundaries. The physical number of precursors is conserved.
 
*The number of pointwise precursors held in memory is controlled to the requested number at time interval boundaries. The physical number of precursors is conserved.
 
*Storing a too low of a number of pointwise precursors can lead to undersampling of certain precursor groups or parts of geometry.
 
*Storing a too low of a number of pointwise precursors can lead to undersampling of certain precursor groups or parts of geometry.
Line 5,594: Line 6,423:
 
  '''set precthresh''' ''THRESHOLD''
 
  '''set precthresh''' ''THRESHOLD''
  
Set the weight threshold for creating and storing a new delayed neutron precursor in [[transient simulations]].
+
Set the weight threshold for creating and storing a new delayed neutron precursor in [[transient simulations]]. Input values:
  
 
{|
 
{|
 
| <tt>''THRESHOLD''</tt>
 
|