Difference between revisions of "Input syntax manual"

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=== branch (branch definition) ===

−

'''branch''' ''NAME'' [ '''repm''' ''MAT1'' ''MAT2'' ]

+

'''branch''' ''NAME'' [ [[#branch_repm|'''repm''']] ''MAT1'' ''MAT2'' ]

−

[ '''repu''' ''UNI1'' ''UNI2'' ]

+

[ [[#branch_repu|'''repu''']] ''UNI1'' ''UNI2'' ]

−

[ '''stp''' ''MAT DENS TEMP THERM1 SABL1 SABH1 THERM2 SABL2 SABH2 ...'' ]

+

[ [[#branch_stp|'''stp''']] ''MAT DENS TEMP THERM1 SABL1 SABH1 THERM2 SABL2 SABH2 ...'' ]

−

[ '''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''' ''UNI2'' ]

+

[ [[#branch_gcu|'''gcu''']] ''UNI2'' ]

−

[ '''reptrc''' ''FILE1'' ''FILE2'' ]

+

[ [[#branch_reptrc|'''reptrc''']] ''FILE1'' ''FILE2'' ]

−

[ '''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>

| : branch name

−

|-

+

|}

+

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

+

Notes:

+

*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>''BRm,i''</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]].

+

Variation types:

+

Branch material variation (<tt>'''repm'''</tt>):

+

{|

| <tt>''MAT1''</tt>

| : name of the replaced material

Line 39:

Line 57:

| <tt>''MAT2''</tt>

| : name of the replacing material

−

|-

+

|}

+

Notes:

+

*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>''MAT1''</tt> were created using the [[#mat_.28material_definition.29|mat]] or [[#mix_.28mixture_definition.29|mix]] card of <tt>''MAT2''</tt>.

+

*The name of the material present in the geometry will still be <tt>''MAT1''</tt> after the replacement, but the material specification (composition, density, tmp, moder, rgb, etc.) will correspond to <tt>''MAT2''</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>''MAT1''</tt>) and they will automatically apply to whatever material <tt>''MAT2''</tt> replaces <tt>''MAT1''</tt> for the branch calculation.

+

*The replaced material ''MAT1'' 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 ''MAT2'' 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>):

+

{|

| <tt>''UNI1''</tt>

| : name of the replaced universe

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| <tt>''UNI2''</tt>

| : name of the replacing universe

−

|-

+

|}

+

Notes:

+

*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>''UNI1''</tt> after the replacement, but the universe contents will correspond to <tt>''UNI2''</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>''UNI1''</tt>) and they will automatically apply to whatever universe <tt>''UNI2''</tt> replaces <tt>''UNI1''</tt> for the branch calculation.

+

Branch state variation, density/temperature (<tt>'''stp'''</tt>):

+

{|

| <tt>''MAT''</tt>

| : name of the material for which density and temperature are adjusted

|-

| <tt>''DENS''</tt>

−

| : material density after adjustment (positive ~~entries for~~ atomic, negative ~~entries for~~ mass ~~densities, or "sum" to use the sum of the constituent nuclide densities~~)

+

| : material density after adjustment (positive value = atomic density [in b-1cm-1], negative value = mass density [in g/cm3])

|-

| <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>''THERMn''</tt>

−

| : ''n''th thermal scattering data associated with the material

+

| : ''n''-th thermal scattering data associated with the material

|-

| <tt>''SABLn''</tt>

−

| : name of the ''n''th S(α, β) library for temperature below the given value

+

| : name of the ''n''-th S(α, β) library for temperature below the given value

|-

| <tt>''SABHn''</tt>

−

| : name of the ''n''th S(α, β) library for temperature above the given value

+

| : name of the ''n''-th S(α, β) library for temperature above the given value

−

|-

+

|}

+

Notes:

+

*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(α, β) 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>):

+

{|

| <tt>''TGT''</tt>

| : target universe, surface or cell

Line 69:

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| <tt>''TRANS''</tt>

| : name of the applied transformation

−

|-

+

|}

+

Notes:

+

*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>):

+

{|

| <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)

−

|-

+

|}

+

Notes:

+

*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>):

+

{|

+

| <tt>''OPT''</tt>

+

| : option for setting Sm-poison concentrations (0 = set to zero, 1 = use values from restart file)

+

|}

+

Notes:

+

*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>):

+

{|

| <tt>''NSF''</tt>

| : normalization scaling factor

−

|-

+

|}

+

Notes:

+

*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>):

+

{|

+

| <tt>''UNI2''</tt>

+

|: name of the replacing universe

+

|}

+

Notes:

+

*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>):

+

{|

| <tt>''FILE1''</tt>

| : file path of the replaced transport correction curve data

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| <tt>''FILE2''</tt>

| : file path of the replacing transport correction curve data

−

|-

+

|}

+

Notes:

+

*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>):

+

{|

| <tt>''VNAME''</tt>

| : variable name

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Notes:

+

*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>):

−

*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>''~~MAT1~~''~~</tt> were created using the [[#mat_.28material_definition.29|mat]] or [[#mix_.28mixture_definition.29|mix]] card of <tt>''MAT2~~''</tt>.

+

{|

−

**The name of the material present in the geometry will still be <tt>''MAT1''</tt> after the replacement, but the material specification (composition, density, tmp, moder, rgb, etc.) ~~will correspond to~~ <tt>~~''MAT2~~</~~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>''~~MAT1~~''</tt>~~) and they will automatically apply~~ to ~~whatever material <tt>''MAT2<~~/~~sub>''</tt> replaces <tt>''MAT1''</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>''UNI1''</tt> after the replacement, but the universe contents will correspond to <tt>''UNI2''</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>~~''UNI1~~</~~sub~~>~~''</tt>) and they will automatically apply to whatever universe <tt>''UNI2''</tt> replaces <tt>''UNI1''</tt> for the branch calculation.~~

+

Notes:

−

*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(α, β) 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 card by a mixture of boron and water defined with a mix card containing the same pure water material).

+

−

*Currently (version 2.1.30) the replacing material ''MAT2</~~sub~~>~~'' of '''repm''' variation can not be included in geometry using other cards than the branch card with the '''repm''' variation~~.

+

=== casematrix (casematrix definition) ===

Line 121:

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''NBR2'' [ ''BR2,1'' ''BR2,2'' ... ''BR1,NBR2'' ]

...

−

+

Defines the casematrix for the automated burnup sequence. Input values:

Line 132:

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|-

| <tt>''HIS_BRk''</tt>

−

| : name of the ''k''th history variation branch

+

| : name of the ''k''-th history variation branch

|-

| <tt>''NBU''</tt>

Line 138:

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|-

| <tt>''BUn''</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>''NBRm''</tt>

−

| : number branches in the ''m''th dimension of the burnup matrix

+

| : number branches in the ''m''-th dimension of the burnup matrix

|-

| <tt>''BRm,i''</tt>

−

| : name of the ''i''th branch in the ''m''th dimension

+

| : name of the ''i''-th branch in the ''m''-th dimension

|}

Line 150:

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*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>''NBR1''</tt> × <tt>''NBR2''</tt> × <tt>''NBR3''</tt> × ... ). 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.

*Multiple casematrix cards can be given in a single input file.

+

*For more information, see detailed description on [[automated burnup sequence]].

=== cell (cell definition) ===

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*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.

−

*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.

*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 233:

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|-

| <tt>''BUn''</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>''NBRm''</tt>

−

| : number branches in the ''m''th dimension of the burnup matrix

+

| : number branches in the ''m''-th dimension of the burnup matrix

|-

| <tt>''BRm,i''</tt>

−

| : name of the ''i''th branch in the ''m''th dimension

+

| : name of the ''i''-th branch in the ''m''-th dimension

|}

Notes:

−

*The coef card creates a multi-dimensional coefficient matrix (of size <tt>''NBR1''</tt> × <tt>''NBR2''</tt> × <tt>''NBR3''</tt> × ... ). 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>''NBR1''</tt> × <tt>''NBR2''</tt> × <tt>''NBR3''</tt> × ... ). 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 more information, see detailed description on [[automated burnup sequence]].

Line 265:

Line 366:

|-

| <tt>''NX''</tt>

−

| : number of cells in the x direction

+

| : number of cells in the x-direction

|-

| <tt>''XMIN''</tt>

−

| : mesh lower x boundary

+

| : mesh lower x-boundary [in cm]

|-

| <tt>''XMAX''</tt>

−

| : mesh higher x boundary

+

| : mesh higher x-boundary [in cm]

|-

| <tt>''NY''</tt>

−

| : number of cells in the y direction

+

| : number of cells in the y-direction

|-

| <tt>''YMIN''</tt>

−

| : mesh lower y boundary

+

| : mesh lower y-boundary [in cm]

|-

| <tt>''YMAX''</tt>

−

| : mesh higher y boundary

+

| : mesh higher y-boundary [in cm]

|-

| <tt>''NZ''</tt>

−

| : number of cells in the z direction

+

| : number of cells in the z-direction

|-

| <tt>''ZMIN''</tt>

−

| : mesh lower z boundary

+

| : mesh lower z-boundary [in cm]

|-

| <tt>''ZMAX''</tt>

−

| : mesh higher z boundary

+

| : mesh higher z-boundary [in cm]

|}

Line 307:

Line 408:

|-

| <tt>''RMIN''</tt>

−

| : mesh inner radial boundary

+

| : mesh inner radial boundary [in cm]

|-

| <tt>''RMAX''</tt>

−

| : mesh outer radial boundary

+

| : mesh outer radial boundary [in cm]

|-

| <tt>''NPHI''</tt>

Line 329:

Line 430:

|-

| <tt>''X0''</tt>

−

| : mesh horizontal origin x-coordinate

+

| : mesh horizontal origin x-coordinate [in cm]

|-

| <tt>''Y0''</tt>

−

| : mesh horizontal origin y-coordinate

+

| : mesh horizontal origin y-coordinate [in cm]

|-

| <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>''ZMIN''</tt>

−

| : mesh lower z boundary

+

| : mesh lower z-boundary [in cm]

|-

| <tt>''ZMAX''</tt>

−

| : mesh higher z boundary

+

| : mesh higher z-boundary [in cm]

|-

| <tt>''NX''</tt>

−

| : number of cells in the x direction

+

| : number of cells in the x-direction

|-

| <tt>''NY''</tt>

−

| : number of cells in the y direction

+

| : number of cells in the y-direction

|-

| <tt>''NZ''</tt>

−

| : number of cells in the z direction

+

| : number of cells in the z-direction

|}

Line 366:

Line 467:

|-

| <tt>''X0''</tt>

−

| : mesh horizontal origin x-coordinate

+

| : mesh horizontal origin x-coordinate [in cm]

|-

| <tt>''Y0''</tt>

−

| : mesh horizontal origin y-coordinate

+

| : mesh horizontal origin y-coordinate [in cm]

|-

| <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>''ZMIN''</tt>

−

| : mesh lower z boundary

+

| : mesh lower z-boundary [in cm]

|-

| <tt>''ZMAX''</tt>

−

| : mesh higher z boundary

+

| : mesh higher z-boundary [in cm]

|-

| <tt>''NX''</tt>

−

| : number of cells in the x direction

+

| : number of cells in the x-direction

|-

| <tt>''NY''</tt>

−

| : number of cells in the y direction

+

| : number of cells in the y-direction

|-

| <tt>''NZ''</tt>

−

| : number of cells in the z direction

+

| : number of cells in the z-direction

|}

Line 402:

Line 503:

|-

| <tt>''NX''</tt>

−

| : number of cells in the x direction

+

| : number of cells in the x-direction

|-

| <tt>''NY''</tt>

−

| : number of cells in the y direction

+

| : number of cells in the y-direction

|-

| <tt>''NZ''</tt>

−

| : number of cells in the z direction

+

| : number of cells in the z-direction

|-

| <tt>''Xi''</tt>

−

| : <tt>''NX'' + 1</tt> mesh boundaries in the x direction

+

| : <tt>''NX'' + 1</tt> mesh boundaries in the x-direction [in cm]

|-

| <tt>''Yi''</tt>

−

| : <tt>''NY'' + 1</tt> mesh boundaries in the y direction

+

| : <tt>''NY'' + 1</tt> mesh boundaries in the y-direction [in cm]

|-

| <tt>''Zi''</tt>

−

| : <tt>''NZ'' + 1</tt> mesh boundaries in the z direction

+

| : <tt>''NZ'' + 1</tt> mesh boundaries in the z-direction [in cm]

|}

Line 438:

Line 539:

|-

| <tt>''Ri''</tt>

−

| : <tt>''NR'' + 1</tt> mesh boundaries in the r direction

+

| : <tt>''NR'' + 1</tt> mesh boundaries in the radial direction [in cm]

|}

−

'''datamesh''' ''NAME'' '''9''' ''NLEVEL'' ''MESH1'' ... ''MESHNLEVEL''

+

'''datamesh''' ''NAME'' '''9''' ''USELC'' ''NLEVEL'' ''MESH1'' ... ''MESHNLEVEL''

Defines a regular nested mesh that can be linked to detectors, interfaces etc.

Line 461:

Line 562:

Notes:

*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>''USELC''</tt> parameter defined in the nested mesh itself and use it in the subsequent sub-meshes, overriding the <tt>''USELC''</tt> parameter defined on those.

=== dep (depletion history) ===

Line 477:

Line 579:

The possible step types are:

−

{| class="wikitable" style="text-align: left;"

+

::{| class="wikitable" style="text-align: left;"

! Type

! Description

Line 560:

Line 662:

[ [[#det_dy|'''dy''']] ''YMIN'' ''YMAX'' ''NY'' ]

[ [[#det_dz|'''dz''']] ''ZMIN'' ''ZMAX'' ''NZ'' ]

−

[ [[#~~det_dn~~|'''dn''']] ''TYPE'' ''MIN1'' ''MAX1'' ''N1'' ''MIN2'' ''MAX2'' ''N2'' ''MIN3'' ''MAX3'' ''N3'' ]

+

[ [[#det_dn1|'''dn''']] ''TYPE'' ''MIN1'' ''MAX1'' ''N1'' ''MIN2'' ''MAX2'' ''N2'' ''MIN3'' ''MAX3'' ''N3'' ]

+

[ [[#det_dn2|'''dn''']] ''TYPE'' ''N1'' ''N2'' ''N3'' ''LIM11''...''LIM1N+1'' ''LIM21''...''LIM2N+1'' ''LIM31''...''LIM3N+1'' ]

[ [[#det_dh|'''dh''']] ''TYPE'' ''X0'' ''Y0'' ''PITCH'' ''N1'' ''N2'' ''ZMIN'' ''ZMAX'' ''NZ'' ]

[ [[#det_dumsh|'''dumsh''']] ''UNI'' ''NC'' ''CELL0'' ''BIN0'' ''CELL1'' ''BIN1'' ... ]

Line 576:

Line 679:

[ [[#det_dphb|'''dphb''']] ''PHB'' ]

[ [[#det_dmesh|'''dmesh''']] ''MESH'' ]

−

Detector definition. The first ~~parameter~~:

+

Detector definition. The two first parameters:

{|

+

| <tt>''NAME''</tt>

+

| : detector name

+

|-

| <tt>''PART''</tt>

| : 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.

+

Notes:

+

*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.

+

Detector types:

Detector response (<tt>'''dr'''</tt>):

Line 599:

Line 713:

*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]].

−

*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.

Line 610:

Line 732:

{|

| <tt>''VOL''</tt>

−

| : volume in cm3 (3D geometry) or cross-sectional area in cm2 (2D geometry)

+

| : volume [in cm3] (3D geometry) or cross-sectional area [in cm2] (2D geometry)

|}

Notes:

−

*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 cm2] (3D geometry) or the surface length [in cm] (2D geometry).

Line 662:

Line 785:

−

Cartesian mesh ~~detector~~ (<tt>'''dx'''</tt>, <tt>'''dy'''</tt> and <tt>'''dz'''</tt>):

+

Detector evenly-spaced Cartesian mesh (<tt>'''dx'''</tt>, <tt>'''dy'''</tt> and <tt>'''dz'''</tt>):

{|

| <tt>''XMIN''</tt>

−

| : minimum x-coordinate of the detector mesh

+

| : minimum x-coordinate of the detector mesh [in cm]

|-

| <tt>''XMAX''</tt>

−

| : maximum x-coordinate of the detector mesh

+

| : maximum x-coordinate of the detector mesh [in cm]

|-

| <tt>''NX''</tt>

Line 675:

Line 798:

|-

| <tt>''YMIN''</tt>

−

| : minimum y-coordinate of the detector mesh

+

| : minimum y-coordinate of the detector mesh [in cm]

|-

| <tt>''YMAX''</tt>

−

| : maximum y-coordinate of the detector mesh

+

| : maximum y-coordinate of the detector mesh [in cm]

|-

| <tt>''NY''</tt>

Line 684:

Line 807:

|-

| <tt>''ZMIN''</tt>

−

| : minimum z-coordinate of the detector mesh

+

| : minimum z-coordinate of the detector mesh [in cm]

|-

| <tt>''ZMAX''</tt>

−

| : maximum z-coordinate of the detector mesh

+

| : maximum z-coordinate of the detector mesh [in cm]

|-

| <tt>''NZ''</tt>

Line 694:

Line 817:

Notes:

−

*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~~ mesh ~~detector~~ (<tt>'''dn'''</tt>):

+

Detector evenly-spaced curvilinear mesh (<tt>'''dn'''</tt>):

{|

| <tt>''TYPE''</tt>

−

| : ~~Type~~ of curvilinear mesh - 1 = cylindrical (dimensions ''r'', ''θ'', ''z''), 2 = spherical (dimensions ''r'', ''θ'', ''φ'')

+

| : type of curvilinear mesh - 1 = cylindrical (dimensions ''r'', ''θ'', ''z''), 2 = spherical (dimensions ''r'', ''θ'', ''φ'')

|-

| <tt>''MINn''</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 (''θ'', ''φ'')].

|-

| <tt>''MAXn''</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 (''θ'', ''φ'')].

|-

| <tt>''Nn''</tt>

−

| : ~~Number~~ of bins in the ''n'' coordinate direction (the radial division will be equal ''r'', not equal volume).

+

| : number of bins in the ''n''-coordinate direction (the radial division will be equal ''r'', not equal volume).

|}

Notes:

*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.

−

~~Hexagonal~~ mesh ~~detector~~ (<tt>'''dh'''</tt>):

+

Detector unevenly-spaced mesh (<tt>'''dn'''</tt>):

+

{|

+

| <tt>''TYPE''</tt>

+

| : type of curvilinear mesh - 3 = unevenly-spaced orthogonal (dimensions ''x'', ''y'', ''z''), 4 = unevenly-spaced cylindrical (dimensions ''r'', ''θ'', ''z'')

+

|-

+

| <tt>''Nn''</tt>

+

| : number of bins in the ''n''-coordinate direction

+

|-

+

| <tt>''LIMnm''</tt>

+

| : mesh ''m''-boundary in the ''n''-coordinate direction [in cm (''r'', ''z''), in degrees (''θ'', ''φ'')].

+

|}

+

Notes:

+

*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>):

{|

| <tt>''TYPE''</tt>

−

| : ~~Type~~ of hexagonal mesh (2 = flat face perpendicular to x-axis, 3 = flat face perpendicular to y-axis)

+

| : type of hexagonal mesh (2 = flat face perpendicular to x-axis, 3 = flat face perpendicular to y-axis)

|-

| <tt>''X0''</tt>, <tt>''Y0''</tt>

−

| : coordinates of mesh center

+

| : coordinates of mesh center [in cm]

|-

| <tt>''PITCH''</tt>

−

| : mesh pitch

+

| : mesh pitch [in cm]

|-

| <tt>''N1''</tt>, <tt>''N1''</tt>

Line 735:

Line 876:

|-

| <tt>''ZMIN''</tt>

−

| : minimum z-coordinate of the detector mesh

+

| : minimum z-coordinate of the detector mesh [in cm]

|-

| <tt>''ZMAX''</tt>

−

| : maximum z-coordinate of the detector mesh

+

| : maximum z-coordinate of the detector mesh [in cm]

|-

| <tt>''NZ''</tt>

Line 748:

Line 889:

−

~~Unstructured~~ mesh ~~detector~~ (<tt>'''dumsh'''</tt>):

+

Detector unstructured mesh (<tt>'''dumsh'''</tt>):

{|

Line 763:

Line 904:

Notes:

*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 775:

Line 917:

Notes:

*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]].

+

**[[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 791:

Line 936:

−

~~Surface~~ current / flux ~~detector~~ (<tt>'''ds'''</tt>):

+

Detector current / flux surface (<tt>'''ds'''</tt>):

{|

Line 803:

Line 948:

Notes:

*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.

+

**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).

Line 813:

Line 961:

{|

−

| <tt>''COSY''</tt>

+

| <tt>''COSX''</tt>

| : component of the direction vector parallel to x-axis

|-

Line 827:

Line 975:

−

~~Super~~-imposed track-length ~~detector~~ (<tt>'''dtl'''</tt>):

+

Detector super-imposed track-length (<tt>'''dtl'''</tt>):

{|

Line 836:

Line 984:

Notes:

*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 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 847:

Line 997:

|-

| <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)

|}

Notes:

*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.

−

*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>):

+

Detector special types (<tt>'''dt'''</tt>):

{|

Line 867:

Line 1,017:

|}

−

The types are:

+

The possible special types are:

−

{|

+

::{| class="wikitable" style="text-align: left;"

+

! Type

+

! Description

+

! Notes

+

|-

| -1

−

| = cumulative spectrum

+

| cumulative spectrum

+

| use with energy binning ('''de''')

|-

| -2

−

| = division by energy width

+

| division by energy width

+

| use with energy binning ('''de''')

|-

| -3

−

| = division by lethargy width

+

| division by lethargy width

+

| use with energy binning ('''de''')

|-

| -4

−

| = sum over cell or material bins

+

| sum over cell or material bins

+

| use with cell and/or material binning ('''dc''', '''dm''')

|-

| -5

−

| = importance weighting

+

| importance weighting

+

| -

|-

| -6

−

| = sum over number of scores

+

| sum over number of scores

+

| -

|-

| 2

−

| = multiply result with another detector defined by <tt>''PARAM''</tt>

+

| multiply result with another detector defined by <tt>''PARAM''</tt>

+

| bin-compatibility

|-

| 3

−

| = divide result with another detector defined by <tt>''PARAM''</tt>

+

| divide result with another detector defined by <tt>''PARAM''</tt>

+

| bin-compatibility

+

|-

+

| 4

+

| multiply response function by (local) temperature

+

| -

+

|-

|}

Notes:

−

*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>):

+

Detector history collection flag (<tt>'''dhis'''</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

|}

Notes:

−

*When this option is set, the batch-wise results are ~~kept and~~ printed in the [[history output file]].

+

*When this option is set, the batch-wise results are printed in the history output file, <tt>[input]_stats.m</tt>.

*''Note to developers: statistical tests should be documented''

−

Detector flagging (<tt>'''dfl'''</tt>):

+

Detector score flagging (<tt>'''dfl'''</tt>):

{|

Line 921:

Line 1,086:

| <tt>''OPT''</tt>

| : 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)

+

|-

|}

Notes:

*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 logic: detector is scored if all the associated flags are set/unset

−

~~Activation detector~~ (<tt>'''da'''</tt>):

+

Detector activation (<tt>'''da'''</tt>):

{|

Line 935:

Line 1,127:

|-

| <tt>''FLX''</tt>

−

| : flux applied to activation

+

| : flux applied to activation [in 1/cm2s]

|}

Notes:

−

*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.

+

*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.

−

Functional Expansion Tally ~~detector~~ (<tt>'''dfet'''</tt>):

+

Detector Functional Expansion Tally, FET (<tt>'''dfet'''</tt>):

{|

Line 954:

Line 1,151:

|}

−

{| class="wikitable" style="text-align: left;"

+

::{| class="wikitable" style="text-align: left;"

! Geometry

! <tt>PARAMS</tt>

Line 978:

Line 1,175:

Notes:

−

*<tt>-1</tt> 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

*Specifics of this implementation:

Line 987:

Line 1,184:

***The generated FET coefficients <math>a_n</math> already have all contributions from the orthonormalization constant pre-included, i.e. <math>c_n</math> from <math>\frac{1}{c_n} = \lVert \psi_n \rVert^2 = \int_\Gamma \psi_n^2 \omega_n </math>

***Thus, an FET can be simply reconstructed/sampled from the standard functional series as: <math>F(\xi) = \sum a_n \psi_n(\xi) \omega_n(\xi)</math>

+

*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.

Line 1,012:

Line 1,211:

=== div (divisor definition) ===

−

'''div''' ''MAT'' [ '''sep''' ''LVL'' ]

+

'''div''' ''MAT'' [ [[#div_sep|'''sep''']] ''LVL'' ]

−

[ '''subx''' ''NX'' ''XMIN'' ''XMAX'' ]

+

[ [[#div_subx1|'''subx''']] ''NX'' ''XMIN'' ''XMAX'' ]

−

[ '''suby''' ''NY'' ''YMIN'' ''YMAX'' ]

+

[ [[#div_subx2|'''subx''']] ''NX'' ''X1'' ''X2'' ... ''XN+1'' ]

−

[ '''subz''' ''NZ'' ''ZMIN'' ''ZMAX'' ]

+

[ [[#div_suby1|'''suby''']] ''NY'' ''YMIN'' ''YMAX'' ]

−

[ '''subr''' ''NR'' ''RMIN'' ''RMAX'' ] ~~equal volume~~ [ '''subr''' ''-NR'' ''R0'' ''R1'' ... ''RN'' ] <~~small~~>~~manually spaced limits~~</~~small~~>

+

[ [[#div_suby2|'''suby''']] ''NY'' ''Y1'' ''Y2'' ... ''YN+1'' ]

−

[ '''subs''' ''NS'' ''S0'' ]

+

[ [[#div_subz1|'''subz''']] ''NZ'' ''ZMIN'' ''ZMAX'' ]

−

Divides a material into a number of sub-zones. ~~Input values~~:

+

[ [[#div_subz2|'''subz''']] ''NZ'' ''Z1'' ''Z2'' ... ''ZN+1'' ]

+

[ [[#div_subr1|'''subr''']] ''NR'' ''RMIN'' ''RMAX'' ]

+

[ [[#div_subr2|'''subr''']] ''NR'' ''R1'' ''R2'' ... ''RN+1'' ]

+

[ [[#div_subs1|'''subs''']] ''NS'' ''S0'' ]

+

[ [[#div_subs2|'''subs''']] ''NS'' ''S1'' ''S2'' ... ''SN+1'' ]

+

[ [[#div_peb|'''peb''']] ''PBED'' ''NUNI'' [ ''UNI1'' ... ''UNIN'' ] ]

+

[ [[#div_lims|'''lims''']] ''FLAG'' ]

+

Divides a material into a number of sub-zones. The first parameter:

{|

| <tt>''MAT''</tt>

| : name of the divided material

−

|-

+

|}

+

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

+

Notes:

+

*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 positive, the sub-division is equal volume (see below)

+

** if the number of zones associated with a sub-division is negative, the subdivision is user-defined volume (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]].

+

Sub-division types:

+

Sub-division geometry level (<tt>'''sep'''</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.)

|-

+

|}

+

Notes:

+

*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, equal volume (<tt>'''subx'''</tt>, <tt>'''suby'''</tt> and <tt>'''subz'''</tt>):

+

{|

| <tt>''NX''</tt>

−

| : number of x-zones

+

| : number of x-zones (positive value)

|-

| <tt>''XMIN''</tt>

−

| : minimum x-coordinate (cm)

+

| : minimum x-coordinate [in cm]

|-

| <tt>''XMAX''</tt>

−

| : maximum x-coordinate (cm)

+

| : maximum x-coordinate [in cm]

|-

| <tt>''NY''</tt>

−

| : number of y-zones

+

| : number of y-zones (positive value)

|-

| <tt>''YMIN''</tt>

−

| : minimum y-coordinate (cm)

+

| : minimum y-coordinate [in cm]

|-

| <tt>''YMAX''</tt>

−

| : maximum y-coordinate (cm)

+

| : maximum y-coordinate [in cm]

|-

| <tt>''NZ''</tt>

−

| : number of z-zones

+

| : number of z-zones (positive value)

|-

| <tt>''ZMIN''</tt>

−

| : minimum z-coordinate (cm)

+

| : minimum z-coordinate [in cm]

|-

| <tt>''ZMAX''</tt>

−

| : maximum z-coordinate (cm)

+

| : maximum z-coordinate [in cm]

|-

+

|}

+

Notes:

+

* An equal volume sub-division is performed in the given dimension.

+

* The value of the parameter <tt>''Nn''</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, user-defined volume (<tt>'''subx'''</tt>, <tt>'''suby'''</tt> and <tt>'''subz'''</tt>):

+

{|

+

| <tt>''NX''</tt>

+

| : number of x-zones (negative value)

+

|-

+

| <tt>''Xn''</tt>

+

| : x-coordinate boundaries [in cm]

+

|-

+

| <tt>''NY''</tt>

+

| : number of y-zones (negative value)

+

|-

+

| <tt>''Yn''</tt>

+

| : y-coordinate boundaries [in cm]

+

|-

+

| <tt>''NZ''</tt>

+

| : number of z-zones (negative value)

+

|-

+

| <tt>''Zn''</tt>

+

| : z-coordinate boundaries [in cm]

+

|-

+

|}

+

Notes:

+

* An user-defined volume sub-division is performed in the given dimension.

+

* The value of the parameter <tt>''Nn''</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, equal volume (<tt>'''subr'''</tt>):

+

{|

| <tt>''NR''</tt>

−

| : number of radial zones

+

| : number of radial-zones (positive value)

|-

| <tt>''RMIN''</tt>

−

| : minimum radial coordinate (cm)

+

| : minimum radial-coordinate [in cm]

|-

| <tt>''RMAX''</tt>

−

| : maximum radial coordinate (cm)

+

| : maximum radial-coordinate [in cm]

+

|-

+

|}

+

Notes:

+

* An equal volume radial sub-division is performed (annular-type sub-division)

+

* The value of the parameter <tt>''NR''</tt> which defines the number of zones in the given dimension must be positive.

+

Sub-division cylindrical annular mesh, user-defined volume (<tt>'''subr'''</tt>):

+

{|

+

| <tt>''NR''</tt>

+

| : number of radial-zones (negative value)

|-

| <tt>''Rn''</tt>

−

| : radial coordinate boundaries (cm)

+

| : radial-coordinate boundaries [in cm]

|-

+

|}

+

Notes:

+

* An user-defined volume radial sub-division is performed (annular-type sub-division)

+

* The value of the parameter <tt>''NR''</tt> which defines the number of zones in the given dimension must be negative.

+

Sub-division cylindrical sector mesh, equal volume (<tt>'''subs'''</tt>):

+

{|

| <tt>''NS''</tt>

−

| : number of angular ~~sectors~~

+

| : number of angular-zones (negative value)

|-

| <tt>''S0''</tt>

−

| : zero position of angular division (degrees)

+

| : zero position of angular division [in degrees]

+

|-

|}

Notes:

+

* An equal volume angular sub-division is performed (sector-type sub-division)

+

* The value of the parameter <tt>''NS''</tt> which defines the number of zones in the angular dimension must be positive.

−

*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 cylindrical sector mesh, user-defined volume (<tt>'''subs'''</tt>):

−

*~~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>''NS''</tt>

−

*The usage of <tt>''~~LVL~~''</tt> ~~is explained on page [[automated depletion zone division]].~~

+

| : number of angular-zones

−

*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~~.

+

|-

+

| <tt>''Sn''</tt>

+

| : angular-sector boundaries [in degrees]

+

|-

+

|}

+

Notes:

+

* An user-defined volume angular sub-division is performed (sector-type sub-division)

+

* The value of the parameter <tt>''NS''</tt> which defines the number of zones in the angular dimension must be negative.

+

* The manually-spaced angular-sector boundaries <tt>''Sn''</tt> must cover the full/360 degrees angular space.

+

Sub-division pebble-bed structure (<tt>'''peb'''</tt>):

+

{|

+

| <tt>''PBED''</tt>

+

| : stochastic particle / pebble-bed structure

+

|-

+

| <tt>''NUNI''</tt>

+

| : number of universes to link related to the <tt>''PBED''</tt> structure (special case: 0 = link to all)

+

|-

+

| <tt>''UNIN''</tt>

+

|: list of universes to link (non-zero number of universes)

+

|-

+

|}

+

Notes:

+

*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>):

+

{|

+

| <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) ===

−

See [[#trans (transformations)|transformations]].

+

Defines detector mesh transformations. Shortcut for "trans d".

+

Notes:

+

*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) ===

Line 1,104:

Line 1,449:

|-

|<tt>''Ei''</tt>

−

|: bin boundaries ~~(type 1)~~

+

|: bin boundaries [in MeV]

|-

|<tt>''N''</tt>

−

|: number of equi-width bins ~~(types 2 and 3)~~

+

|: number of equi-width bins

|-

|<tt>''Emin''</tt>

−

|: minimum energy ~~(types 2 and 3)~~

+

|: minimum energy [in MeV]

|-

|<tt>''Emax''</tt>

−

|: maximum energy ~~(types 2 and 3)~~

+

|: maximum energy [in MeV]

|-

|<tt>''GRID''</tt>

−

|: name of the pre-defined grid ~~(type 4)~~

+

|: name of the pre-defined grid

|}

Notes:

−

*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 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]].

−

*See separate description of [[pre-defined energy group structures]].

=== ftrans (fill transformation) ===

−

See [[#trans (transformations)|transformations]].

+

Defines fill transformations. Shortcut for "<tt>trans f</tt>".

+

Notes:

+

*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) ===

Line 1,152:

Line 1,500:

|: is the interpolation type (1 = histogram, 2 = lin-lin, 3 = lin-log, 4 = log-lin, 5 = log-log)

|-

−

|<tt>''X1'' ''F1'' ~~...~~</tt>

+

|<tt>''Xi'', ''Fi''</tt>

|: are the tabulated variable-value pairs

|}

Notes:

−

* 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.

=== hisv (history variation matrix definition) ===

Line 1,168:

Line 1,517:

{|

| <tt>''BUn''</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>''NBRn''</tt>

Line 1,180:

Line 1,529:

*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.

−

*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.

=== ifc (interface file) ===

−

'''ifc''' ''FILE'' ['''setinmat''' ''NMAT'' ''MAT1'' ''MAT2'' ... ''MATNMAT'' ]

+

'''ifc''' ''FILE'' [ [[#ifc_setinmat|'''setinmat''']] ''NMAT'' ''MAT1'' ''MAT2'' ... ''MATNMAT'' ]

−

['''setoutmat''' ''NMAT'' ''MAT1'' ''MAT2'' ... ''MATNMAT'' ]

+

[ [[#ifc_setoutmat|'''setoutmat''']] ''NMAT'' ''MAT1'' ''MAT2'' ... ''MATNMAT'' ]

−

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 1,195:

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~~.

+

Notes:

+

*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.~~

+

Optional entries:

+

Interface input materials (<tt>'''setinmat'''</tt>):

{|

−

|<tt>''NMAT''</tt>

+

| <tt>''NMAT''</tt>

−

|: number of materials to link to the interface

+

| : number of input materials to link to the interface

|-

−

|<tt>''MATi''</tt>

+

| <tt>''MATn''</tt>

−

|: name of the ''i''th material linked to the interface

+

| : name of the ''n''-th input material linked to the interface

+

|}

+

Notes:

+

*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>):

+

{|

+

| <tt>''NMAT''</tt>

+

| : number of output materials to link to the interface

|-

+

| <tt>''MATn''</tt>

+

| : name of the ''n''-th output material linked to the interface

|}

Notes:

−

* 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) ===

Line 1,231:

Line 1,600:

*The input parser starts reading and processing the new file from the point where the input card is placed. Processing of the original file continues after the new file is completed.

*The included file must contain complete input cards and options, it cannot be used to read the values of another card.

+

*Nested included file paths must refer to the original base input file or current working directory.

=== lat (regular lattice definition) ===

Line 1,248:

Line 1,618:

|-

| <tt>''X0''</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>''Y0''</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>''NX''</tt>

Line 1,260:

Line 1,630:

|-

| <tt>''PITCH''</tt>

−

| : lattice pitch

+

| : lattice pitch [in cm]

|-

| <tt>''UNIn''</tt>

Line 1,267:

Line 1,637:

Possible lattice types are:

−

{| class="wikitable" style="text-align: left;"

+

::{| class="wikitable" style="text-align: left;"

! Type

! Description

Line 1,306:

Line 1,676:

|-

| <tt>''X0''</tt>

−

| : x-coordinate of the lattice origin

+

| : x-coordinate of the lattice origin [in cm]

|-

| <tt>''Y0''</tt>

−

| : y-coordinate of the lattice origin

+

| : y-coordinate of the lattice origin [in cm]

|-

| <tt>''PITCH''</tt>

−

| : lattice pitch

+

| : lattice pitch [in cm]

|-

| <tt>''UNI1''</tt>

Line 1,319:

Line 1,689:

Possible lattice types are:

−

{| class="wikitable" style="text-align: left;"

+

::{| class="wikitable" style="text-align: left;"

! Type

! Description

Line 1,348:

Line 1,718:

|-

| <tt>''X0''</tt>

−

| : x-coordinate of the lattice origin

+

| : x-coordinate of the lattice origin [in cm]

|-

| <tt>''Y0''</tt>

−

| : y-coordinate of the lattice origin

+

| : y-coordinate of the lattice origin [in cm]

|-

| <tt>''NR''</tt>

Line 1,357:

Line 1,727:

|-

| <tt>''NS,R''</tt>

−

| : number of sectors in ~~Rth~~ ring

+

| : number of sectors in ''R''-th ring

|-

| <tt>''RADIUSR''</tt>

−

| : central radius of ~~Rth~~ ring

+

| : central radius of ''R''-th ring [in cm]

|-

| <tt>''THETAR''</tt>

−

| : angle of rotation of ~~Rth~~ ring in degrees

+

| : angle of rotation of ''R''-th ring [in degrees]

|-

| <tt>''UNIN,R''</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:

−

{| class="wikitable" style="text-align: left;"

+

::{| class="wikitable" style="text-align: left;"

! Type

! Description

Line 1,393:

Line 1,763:

|-

| <tt>''X0''</tt>

−

| : x-coordinate of the lattice origin

+

| : x-coordinate of the lattice origin [in cm]

|-

| <tt>''Y0''</tt>

−

| : y-coordinate of the lattice origin

+

| : y-coordinate of the lattice origin [in cm]

|-

| <tt>''NL''</tt>

Line 1,402:

Line 1,772:

|-

| <tt>''Zn''</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>''UNIn''</tt>

−

| : universe name filling the ~~nth~~ lattice position

+

| : universe name filling the ''n''-th lattice position

|}

Possible lattice type is:

−

{| class="wikitable" style="text-align: left;"

+

::{| class="wikitable" style="text-align: left;"

! Type

! Description

Line 1,435:

Line 1,805:

|-

| <tt>''X0''</tt>

−

| : x-coordinate of the lattice origin

+

| : x-coordinate of the lattice origin [in cm]

|-

| <tt>''Y0''</tt>

−

| : y-coordinate of the lattice origin

+

| : y-coordinate of the lattice origin [in cm]

|-

| <tt>''Z0''</tt>

−

| : z-coordinate of the lattice origin

+

| : z-coordinate of the lattice origin [in cm]

|-

| <tt>''NX''</tt>

Line 1,453:

Line 1,823:

|-

| <tt>''PITCHX''</tt>

−

| : lattice pitch in x-direction

+

| : lattice pitch in x-direction [in cm]

|-

| <tt>''PITCHY''</tt>

−

| : lattice pitch in y-direction

+

| : lattice pitch in y-direction [in cm]

|-

| <tt>''PITCHZ''</tt>

−

| : lattice pitch in z-direction

+

| : lattice pitch in z-direction [in cm]

|-

| <tt>''UNIn''</tt>

Line 1,466:

Line 1,836:

Possible lattice types are:

−

{| class="wikitable" style="text-align: left;"

+

::{| class="wikitable" style="text-align: left;"

! Type

! Description

Line 1,487:

Line 1,857:

=== ltrans (lattice transformation) ===

−

See [[#trans (transformations)|transformations]].

+

Defines lattice transformations. Shortcut for "<tt>trans l</tt>".

+

Notes:

+

*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) ===

Line 1,506:

Line 1,880:

[ ''...'' ]

−

~~Mandatory information~~:~~~~

+

Material definition. The mandatory parameters are:

{|

| <tt>''NAME''</tt>

−

| : ~~Name~~ of the material

+

| : name of the material

|-

| <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-1cm-1], negative value = mass density [in g/cm3]), or "<tt>sum</tt>" to calculate the density from given nuclide fractions

|-

| <tt>''NUCn''</tt>

−

| : Identifier of ''n''th nuclide in composition~~, e.g. "92235.03c" or "U-235.03c".~~

+

| : Identifier of ''n''-th nuclide in composition

|-

| <tt>''FRACn''</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)

|-

|}

−

~~Optional cards:~~

+

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

−

'''tmp''': ~~Material temperature for [[Doppler-broadening preprocessor routine|Doppler-preprocessor]]~~

+

Notes:

+

*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]].

+

Optional entries:

+

Material temperature for Doppler-broadening pre-processor (<tt>'''tmp'''</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''': ~~Material temperature for on-the-fly [[TMS on-the-fly temperature treatment routine‏‎|TMS temperature treatment]]~~

+

Notes:

+

*It defines the material temperature for [[Doppler-broadening preprocessor routine|Doppler-preprocessor]].

+

Material temperature for on-the-fly temperature treatment (<tt>'''tms'''</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''': ~~Temperature limits for material for [[Coupled multi-physics calculations|coupled multi-physics calculations]]~~

+

Notes:

+

*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>):

{|

| <tt>''TMIN''</tt>

−

| : ~~Lower~~ limit for material temperature

+

| : lower limit for material temperature [in K]

|-

| <tt>''TMAX''</tt>

−

| : ~~Upper~~ limit for material temperature

+

| : upper limit for material temperature [in K]

|}

−

'''rgb''': ~~Material color for [[#plot|geometry plots]]~~

+

Notes:

+

*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>):

{|

| <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>

−

| : ~~Value~~ for the green channel ~~of geometry plots~~ (between 0 and 255)

+

| : value for the green channel (between 0 and 255)

|-

| <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''': ~~Material volume~~

+

Notes:

+

*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>):

{|

| <tt>''VOL''</tt>

−

| : ~~Volume~~ of the material in cm3 (3D geometry) or cross-sectional area in cm2 (2D geometry)

+

| : volume of the material [in cm3] (3D geometry) or cross-sectional area [in cm2] (2D geometry)

|}

−

'''mass''': ~~Material mass~~

+

Notes:

+

*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>):

{|

| <tt>''MASS''</tt>

−

| : ~~Mass~~ of the material in ~~grams~~

+

| : mass of the material [in g]

|}

−

'''burn''': ~~Flag material for depletion~~

+

Notes:

+

*The material mass can be provided as an alternative to the material volume.

+

Material depletion flag (<tt>'''burn'''</tt>):

{|

| <tt>''NR''</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''': ~~Library information for decay nuclides~~

+

Notes:

+

*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>):

{|

| <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>

−

| : ~~Temperature (in Kelvin)~~ for nuclides without cross section data.

+

| : temperature for nuclides without cross section data [in K]

|}

−

'''moder''': 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.

+

Notes:

+

*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>):

{|

| <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>

−

| : ~~Nuclide~~ ZA of the thermal scatterer (e.g. 1001 for H-1).

+

| : nuclide ZA of the thermal scatterer (e.g. 1001 for H-1).

|}

Notes:

−

*~~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) ===

Line 1,623:

Line 2,048:

|-

| <tt>''XPIX''</tt>

−

| : horizontal image size in pixels

+

| : horizontal image size [in pixels]

|-

| <tt>''YPIX''</tt>

−

| : vertical image size in pixels

+

| : vertical image size [in pixels]

|-

| <tt>''SYM''</tt>

Line 1,632:

Line 2,057:

|-

| <tt>''MINn'' ''MAXn''</tt>

−

| : boundaries of the plotted region

+

| : boundaries of the plotted region [in cm]

|-

| <tt>''CMAP''</tt>

Line 1,650:

Line 2,075:

*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.

−

*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.

*Mesh plot produced by the nth mesh-card is written in file <tt>[input]_mesh[n].png</tt>.

Line 1,668:

Line 2,093:

|-

| <tt>''NUCn''</tt>

−

| : identifier of ''n''th nuclide in composition

+

| : identifier of <tt>''n''</tt>-th nuclide in composition

|-

| <tt>''λ''n</tt>

−

| : reprocessing constant of ''n''th nuclide in composition (in s-1)

+

| : reprocessing constant of <tt>''n''</tt>-th nuclide in composition [in s-1]

|}

Notes:

−

* The nuclide ID can be replaced with "all", in which case all nuclides are included with the same reprocessing fraction ''λ''.

+

* The nuclide ID can be replaced with "<tt>all</tt>", in which case all nuclides are included with the same reprocessing fraction ''λ''.

+

* The nuclide ID should follow the ZAI or ISO format (e.g., 922350 or U-235).

=== mix (mixture definition) ===

−

'''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'' ]

''MAT1'' ''F1''

''MAT2'' ''F2''

...

−

Defines a mixture of two or several materials. ~~Input~~ values:

+

Defines a mixture of two or several materials. Mandatory input values:

−

+

−

~~Mandatory information:~~

+

{|

Line 1,695:

Line 2,119:

|-

| <tt>''Fn''</tt>

−

| : material fraction (positive ~~values for~~ volume, negative ~~values for~~ mass ~~fractions~~)

+

| : material fraction (positive value = volume fraction, negative value = mass fraction)

|-

|}

−

~~Optional cards:~~

+

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

−

'''rgb''': ~~Material color for [[#plot|geometry plots]]~~

+

Notes:

+

*Mixtures can be used to define complicated material definitions consisting of two or more physical materials mixed homogeneously.

+

*The mixtures are automatically decomposed into standard materials before running the transport simulation.

+

**Alternatively, the decomposed material compositions can be written into file using the [[Installing and running Serpent#Running Serpent|<tt>''-mix''</tt>]] command line option.

+

*Inherited properties/cards:

+

**Nuclide specific thermal scattering data (see '''moder''' entry in the [[#mat_moder|mat]] card) is automatically brought from component materials to the mixture.

+

**Other input option 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.

+

**If they are to be applied to the mixture, they should be directly defined using the mixture material name (opposed to component material names) .

+

*Burnable mixtures are not supported.

+

Optional entries:

+

Mixture RGB-color (<tt>'''rgb'''</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>

−

| : ~~Value~~ for the green channel ~~of geometry plots~~ (between 0 and 255)

+

| : value for the green channel (between 0 and 255)

|-

| <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''': ~~Material volume~~

+

Notes:

+

*RGB color coding for material representation in [[#plot|geometry plots]].

+

Mixture volume (<tt>'''vol'''</tt>):

{|

| <tt>''VOL''</tt>

−

| : ~~Volume~~ of the material in cm3 (3D geometry) or cross-sectional area in cm2 (2D geometry)

+

| : volume of the material [in cm3] (3D geometry) or cross-sectional area [in cm2] (2D geometry)

|}

−

'''mass''': ~~Material mass~~

+

Mixture mass (<tt>'''mass'''</tt>):

{|

| <tt>''MASS''</tt>

−

| : ~~Mass~~ of the material in g

+

| : mass of the material [in g]

|}

−

~~Notes:~~

−

*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) ===

Line 1,763:

Line 2,198:

|-

| <tt>''R1 ... RN-1''</tt>

−

| : outer radii

+

| : outer radii [in cm]

|-

| <tt>''TYPE1 ... TYPEN-1''</tt>

| : nested surface type (different surfaces for each region)

|-

−

| <tt>''PARAMnm ... </tt>

+

| <tt>''PARAMnm'' ... </tt>

| : surface parameters

|}

Line 1,774:

Line 2,209:

Notes:

−

*The nest card defines an entire universe consisting of nested material regions. The boundaries are defined by surfaces nested inside each other. The outermost region is infinite.

+

*The nest card defines an entire universe consisting of nested material regions.

+

**The boundaries are defined by surfaces nested inside each other.

+

**The outermost region is infinite.

*The material entries can be replaced by <tt>'''fill''' ''U0''</tt>, in which case the region is filled by another universe.

−

*The first format allows defining nests in which all surfaces are of same type and centred at the origin. Only surfaces that are characterized by a single outer radius are accepted ([[Surface_types#Second-order_quadratic_surfaces|cylinders, spheres]] and some [[Surface_types#Regular_prisms|regular prisms]]). The [[#pin (pin geometry definition)|pin]] and [[#particle (particle geometry definition)|particle]] definitions are short-hand notations of the nest card.

+

*The first format allows defining nests in which all surfaces are of same type and centred at the origin.

+

**Only surfaces that are characterized by a single outer radius are accepted ([[Surface_types#Second-order_quadratic_surfaces|cylinders, spheres]] and some [[Surface_types#Regular_prisms|regular prisms]]).

+

**The [[#pin (pin geometry definition)|pin]] and [[#particle (particle geometry definition)|particle]] definitions are short-hand notations of the nest card.

*The second format allows mixing different surface types. In this case all surface parameters need to be provided after the surface type.

Line 1,796:

Line 2,235:

|-

| <tt>''R1 ... RN-1''</tt>

−

| : outer radii

+

| : outer radii [in cm]

|}

Notes:

−

*The particle card defines an entire universe consisting of nested spherical shells. The boundaries are defined by sphere surfaces. The outermost region is radially infinite.

+

*The particle card defines an entire universe consisting of nested spherical shells.

+

**The boundaries are defined by sphere surfaces.

+

**The outermost region is radially infinite.

*The material entries can be replaced by <tt>'''fill''' ''U0''</tt>, in which case the region is filled by another universe.

*Most typically used for defining TRISO fuel particles.

Line 1,807:

Line 2,248:

*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''' ''U0'' ''Ubg'' ''FILE'' [ ''OPT'' ]

+

'''pbed''' ''UNI0'' ''UNIbg'' ''FILE'' [ ''OPT'' ]

Defines a stochastic particle / pebble-bed geometry. Input values:

{|

−

| <tt>''U0''</tt>

+

| <tt>''UNI0''</tt>

| : universe name for the dispersed medium

|-

−

| <tt>''Ubg''</tt>

+

| <tt>''UNIbg''</tt>

| : background universe, i.e. universe filling the space between particles / pebbles

|-

Line 1,823:

Line 2,264:

|-

| <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~~>~~'' ''Y1<~~/~~sub>'' ''Z1'' ''R1'' ''U1''~~

+

The syntax of the file containing the particle/pebble data is:

−

''X2'' ''Y2'' ''Z2'' ''R2'' ''U2''

+

::{| class="toccolours" style="text-align: left;"

+

| ''X1'' ''Y1'' ''Z1'' ''R1'' ''UNI1''

+

|-

+

| ''X2'' ''Y2'' ''Z2'' ''R2'' ''UNI2''

+

|-

+

| ...

+

|-

+

|}

−

~~...~~

+

where:

−

+

−

~~Where~~:

+

{|

|<tt>''Xn'', ''Yn'', ''Zn''</tt>

−

|: are the coordinates

+

|: are the coordinates [in cm]

|-

|<tt>''Rn''</tt>

−

|: is the radius

+

|: is the radius [in cm]

|-

−

|<tt>''Un''</tt>

+

|<tt>''UNIn''</tt>

|: is the universe

|}

Line 1,847:

Line 2,292:

Notes:

−

*Creates a universe (<tt>''U0''</tt>), which is filled with spherical sub-universes for which the coordinates are read from a separate file.

+

*Creates a universe (<tt>''UNI0''</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.

−

*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]].

Line 1,867:

Line 2,312:

|}

−

The ~~syntax for the different~~ types ~~is as follows~~:

+

The enery-resolution types are:

'''phb''' ''NAME'' '''1''' ''INTT'' ''Emax,1'' ''R1'' ''Emax,2'' ''R2'' ...

−

~~where:~~

{|

|<tt>''INTT''</tt>

Line 1,877:

Line 2,322:

|-

|<tt>''Emax,i, Ri''</tt>

−

|: are the maximum energy-resolution tabulated pairs

+

|: are the maximum energy-resolution tabulated pairs [in MeV (energy)]

|}

Line 1,887:

Line 2,332:

'''phb''' ''NAME'' '''2''' ''INTT'' ''Emax,1'' ''FWHM1'' ''Emax,2'' ''FWHM2'' ...

−

~~where:~~

{|

|<tt>''INTT''</tt>

Line 1,893:

Line 2,337:

|-

|<tt>''Emax,i, FWHMi''</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 1,902:

Line 2,346:

'''phb''' ''NAME'' '''3''' ''a'' ''b''

−

~~where:~~

{|

|<tt>''a, b''</tt>

Line 1,910:

Line 2,353:

'''phb''' ''NAME'' '''4''' ''a'' ''b'' ''c''

−

~~where:~~

{|

|<tt>''a, b, c''</tt>

Line 1,933:

Line 2,375:

|-

| <tt>''R1 ... RN-1''</tt>

−

| : outer radii

+

| : outer radii [in cm]

|}

Notes:

−

*The pin card defines an entire universe consisting of nested annular material regions. The boundaries are defined by axially infinite cylindrical surfaces. The outermost region is radially infinite.

+

*The pin card defines an entire universe consisting of nested annular material regions.

+

**The boundaries are defined by axially infinite cylindrical surfaces.

+

**The outermost region is radially infinite.

*The material entries can be replaced by <tt>'''fill''' ''U0''</tt>, in which case the region is filled by another universe.

*Most typically used for defining fuel pins, but can also be applied to guide tubes, control rods, etc.

Line 1,956:

Line 2,400:

|-

| <tt>''XPIX''</tt>

−

| : horizontal image size in pixels

+

| : horizontal image size [in pixels]

|-

| <tt>''YPIX''</tt>

−

| : vertical image size in pixels

+

| : vertical image size [in pixels]

|-

| <tt>''POS''</tt>

−

| : position of plot plane

+

| : position of plot plane [in cm]

|-

| <tt>''MIN1''</tt>

−

| : minimum horizontal coordinate of plotted region

+

| : minimum horizontal coordinate of plotted region [in cm]

|-

| <tt>''MAX1''</tt>

−

| : maximum horizontal coordinate of plotted region

+

| : maximum horizontal coordinate of plotted region [in cm]

|-

| <tt>''MIN2''</tt>

−

| : minimum vertical coordinate of plotted region

+

| : minimum vertical coordinate of plotted region [in cm]

|-

| <tt>''MAX2''</tt>

−

| : maximum vertical coordinate of plotted region

+

| : maximum vertical coordinate of plotted region [in cm]

|-

| <tt>''Fmin''</tt>

Line 1,983:

Line 2,427:

|-

| <tt>''E''</tt>

−

| : particle energy for importance map plots

+

| : particle energy for importance map plots [in MeV]

|}

Notes:

−

*The ~~type~~ 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.

−

*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 relative dimensions of image size (<tt>''XPIX''</tt>, <tt>''YPIX''</tt>) should match that of the plotted region. Otherwise the image gets distorted.

−

*~~Void is plotted in black and special~~ colors ~~are used to plot geometry errors (red~~ = ~~overlap, green~~ = ~~undefined region).~~

+

*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).

−

*The position parameter defines the location of the plot plane on the axis perpendicular to it (~~e.g. z~~-~~coordinate for xy~~-~~type plot~~).

+

*The minimum and maximum coordinates: <tt>''MINn''</tt>, <tt>''MAXn''</tt>, define the boundaries of the plotted region (e.g. minimum and maximum x- and y-coordinates for xy-type plot).

−

*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.~~

+

** If the 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.

+

−

*Geometry plotter requires compiling the source code with [[~~Compiling~~ Serpent|GD Graphics libraries]].

+

*The second format allows to plot he importance maps read using the [[#wwin (weight window mesh definition)|wwin card]]:

−

*[[Installing and running Serpent#Running Serpent|~~Command line parameter]]~~ '-plot' stops the execution after the geometry plots are produced~~. Option~~ '-qp' invokes a quick plot mode, which does not check for overlaps~~. Option~~ '-noplot' skips the geometry plots altogether.

+

**They can be plotted on top of the geometry by setting the second value ('B') of the type parameter for:

+

*** 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>''Fmin''</tt>, <tt>''Fmax''</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)

+

| COLOR

+

| Outside cell or void-material

+

|-

+

| (0, 255, 0)

+

| COLOR

+

| No cell found at coordinates

+

|-

+

| (255, 0, 0)

+

| COLOR

+

| Overlap of multiple cells found at coordinates

+

|-

+

| (255, 0, 255)

+

| COLOR

+

| 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.

−

*~~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) ===

'''rep''' ''NAME''

−

[ '''rc''' ''SRC'' ''TGT'' ''MFLOW'' ''MODE'' ]

+

[ [[#rep_rc|'''rc''']] ''SRC'' ''TGT'' ''MFLOW'' ''MODE'' ]

−

[ '''rm''' ''MAT1'' ''MAT2'' ]

+

[ [[#rep_rm|'''rm''']] ''MAT1'' ''MAT2'' ]

−

[ '''ru''' ''UNI1'' ''UNI2'' ]

+

[ [[#rep_ru|'''ru''']] ''UNI1'' ''UNI2'' ]

−

Defines the reprocessing controllers. ~~Input values~~:

+

Defines the reprocessing controllers. The first parameter:

{|

| <tt>''NAME''</tt>

| : name of the reprocessor.

−

|-

+

|}

+

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

+

Notes:

+

*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.

+

Reprocessing regime types:

+

Reprocessing continuos regime (<tt>'''rc'''</tt>):

+

{|

| <tt>''SRC''</tt>

| : name of the source material, from which the flow is moved

Line 2,028:

Line 2,523:

| : continuous reprocessing mode

|-

+

|}

+

Notes:

+

*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

+

**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 regime works with materials, not universes. Therefore, define the universes associated with those burnable materials as surface-cell type universes.

+

*The continuous reprocessing regime 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 continuous reprocessing <tt>''MODE''</tt> defines how to incorporate the material flow into the Bateman equations:

+

::{|class="wikitable" style="text-align: center;"

+

! <tt>''MODE''</tt>

+

! Material source

+

! Material target

+

! Material flow

+

|-

+

| 0

+

| <math>N_{src}(t)=N_{src}(0)</math>

+

| <math>N_{tgt}(t)=N_{tgt}(0)+t\lambda N_{src}(0)</math>

+

| <math>\dot{N}(t)=\lambda N_{src}(0)</math>

+

|-

+

| 1

+

| <math>N_{src}(t)=N_{src}(0)e^{-\lambda t}</math>

+

| <math>N_{tgt}(t) = N_{tgt}(0) + (1-e^{-\lambda t})N_{src}(0)</math>

+

| <math>\dot{N}(t)=\lambda N_{src}(t)</math>

+

|-

+

| 2

+

| <math>N_{src}(n+1)=(1-\Delta t_{n}\lambda)N_{src}(n)</math>

+

| <math>N_{tgt}(n+1)=N_{tgt}(n)+\Delta t_{n}\lambda N_{src}(n)</math>

+

| <math>\dot{N}(n)=\lambda N_{src}(n)</math>

+

|}

+

:*<tt>''MODE''</tt> 0 : no changes at the source material and adds ''λN0'' from the source material to the target material when solving the Bateman equations (''N0'' are initial compositions).

+

:*<tt>''MODE''</tt> 1 : subtracts ''λN'' from the source material and adds it to the target material when solving the Bateman equations.

+

:*<tt>''MODE''</tt> 2 : subtracts ''λNn'' from the source material and adds it to the target material when solving the Bateman equations (compositions updated with each burnup step, ''n'').

+

Reprocessing material regime (<tt>'''rm'''</tt>):

+

{|

| <tt>''MAT1''</tt>

| : name of the replaced material

Line 2,034:

Line 2,569:

| : name of the replacing material

|-

+

|}

+

Notes:

+

*The material reprocessing regime replaces one material with another, ''MAT1'' by ''MAT2''.

+

Reprocessing universe regime (<tt>'''ru'''</tt>):

+

{|

| <tt>''UNI1''</tt>

| : name of the replaced universe

Line 2,043:

Line 2,587:

Notes:

+

*The universe reprocessing regime replaces one universe with another, ''UNI1'' by ''UNI2''.

−

*The reprocessor name identifies the reprocessing regime in the depletion calculation [[#dep|dep card]]. The syntax corresponds to '''dep pro''' ''NAME''.

+

=== sample (temperature / density data sample definition) ===

−

*Multiple reprocessing controllers/regimes can be defined within the same reprocessor definition.

+

−

*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 continuous reprocessing option defines how to incorporate the material flow into the Bateman equations:

+

−

**Reprocessing modes: <math>\begin{matrix} & \text{material source} & \text{material target} & \text{material flow} \\

+

−

~~\text{0: } & N_{src}(t)=N_{src}(0) & N_{tgt}(t)=N_{tgt}(0)+t\lambda N_{src}(0) & \dot{N}(t)=\lambda N_{src}(0) \\~~

+

−

~~\text{1: } & N_{src}(t)=N_{src}(0)e^{-\lambda t} & N_{tgt}(t) = N_{tgt}(0) + (1-e^{-\lambda t})N_{src}(0) & \dot{N}(t)=\lambda N_{src}(t) \\~~

+

−

~~\text{2: } & N_{src}(n+1)=(1-\Delta t_{n}\lambda)N_{src}(n) & N_{tgt}(n+1)=N_{tgt}(n)+\Delta t_{n}\lambda N_{src}(n) & \dot{N}(n)=\lambda N_{src}(n) \\ \end{matrix}</math>~~

+

−

**MODE 0 : no changes at the source material and adds ''λN0'' from the source material to the target material when solving the Bateman equations (''N0'' are initial compositions).

+

−

**MODE 1 : subtracts ''λN'' from the source material and adds it to the target material when solving the Bateman equations.

+

−

**MODE 2 : subtracts ''λNn'' 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, ''MAT1'' by ''MAT2''.

+

−

*The '''ru''' universe reprocessing option replaces one universe with another, ''UNI1'' by ''UNI2''.

+

−

+

−

=== sample (~~Temperature~~ / density data sample definition) ===

+

'''sample''' ''NX'' ''XMIN'' ''XMAX'' ''NY'' ''YMIN'' ''YMAX'' ''NZ'' ''ZMIN'' ''ZMAX''

Line 2,069:

Line 2,599:

{|

| <tt>''NX''</tt>

−

| : ~~Number~~ of values to sample in the x-direction.

+

| : number of values to sample in the x-direction.

|-

| <tt>''XMIN''</tt>

−

| : ~~Minimum~~ coordinate to sample from in the x-direction.

+

| : minimum coordinate to sample from in the x-direction [in cm]

|-

| <tt>''XMAX''</tt>

−

| : ~~Maximum~~ coordinate to sample from in the x-direction.

+

| : maximum coordinate to sample from in the x-direction [in cm]

|-

| <tt>''NY''</tt>

−

| : ~~Number~~ of values to sample in the y-direction.

+

| : number of values to sample in the y-direction.

|-

| <tt>''YMIN''</tt>

−

| : ~~Minimum~~ coordinate to sample from in the y-direction.

+

| : minimum coordinate to sample from in the y-direction [in cm]

|-

| <tt>''YMAX''</tt>

−

| : ~~Maximum~~ coordinate to sample from in the y-direction.

+

| : maximum coordinate to sample from in the y-direction [in cm]

|-

| <tt>''NZ''</tt>

−

| : ~~Number~~ of values to sample in the z-direction.

+

| : number of values to sample in the z-direction.

|-

| <tt>''ZMIN''</tt>

−

| : ~~Minimum~~ coordinate to sample from in the z-direction.

+

| : minimum coordinate to sample from in the z-direction [in cm]

|-

| <tt>''ZMAX''</tt>

−

| : ~~Maximum~~ coordinate to sample from in the z-direction.

+

| : maximum coordinate to sample from in the z-direction [in cm]

|-

|}

Line 2,099:

Line 2,629:

Notes:

−

*The data from each sample is written in a separate <tt>[input]_sample~~''N''~~.m</tt> ~~file~~.

+

*The data from each sample is written in a separate output file:

−

*~~Positive values for~~ the ~~density data correspond to~~ atomic densities, ~~while~~ negative ~~values correspond to~~ mass ~~densities~~.

+

**Default: <tt>[input]_sample[n].m</tt>, where "<tt>n</tt>" is the sample number.

−

*Materials with no temperature specified either in their [[~~Input syntax manual~~#mat|mat]]-card or through an interface definition will show a temperature of 0.

+

**Coupled multi-physics calculation: <tt>[input]_sample[n]_iter[i].m</tt>, where "<tt>i</tt>" is the iteration number.

+

**Burnup calculations: <tt>[input]_sample[n]_bstep[bu].m</tt>, where "<tt>bu</tt>" is the burnup step index.

+

*** Coupled multi-physics calculation: <tt>[input]_sample[n]_bstep[bu]_iter[i].m</tt>

+

**Time-dependent calculations: <tt>[input]_sample[n]_tstep[t].m</tt>, where "<tt>t</tt>" is the time step index

+

*** Coupled multi-physics calculation: <tt>[input]_sample[n]_tstep[t]_iter[i].m</tt>

+

*Sampling units:

+

** Density: positive values = atomic densities [in b-1 cm-1], negative value = mass density [in g/cm3].

+

** Temperature: [in K]

+

*Materials with no temperature specified either in their [[#mat|mat]] card or through an interface [[#ifc|ifc]] card definition will show a temperature of 0 K.

=== sens (sensitivity calculation definition) ===

Line 2,131:

Line 2,669:

|-

| <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>

Line 2,137:

Line 2,675:

|-

| <tt>''SZi''</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>

−

| : ~~Path~~ to the unstructured mesh points file

+

| : path to the unstructured mesh points file

|-

| <tt>''FACES_FILE''</tt>

−

| : ~~Path~~ to the unstructured mesh faces file

+

| : path to the unstructured mesh faces file

|-

| <tt>''OWNER_FILE''</tt>

−

| : ~~Path~~ to the unstructured mesh owner file

+

| : path to the unstructured mesh owner file

|-

| <tt>''NEIGHBOUR_FILE''</tt>

−

| : ~~Path~~ to the unstructured mesh neighbour file

+

| : path to the unstructured mesh neighbour file

|-

| <tt>''MATERIALS_FILE''</tt>

−

| : ~~Path~~ to the unstructured mesh materials file

+

| : path to the unstructured mesh materials file

|}

Line 2,177:

Line 2,715:

|-

| <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>

Line 2,186:

Line 2,724:

|-

| <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>

−

| : ~~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>''BODYi''</tt>

−

| : ~~Name~~ of solid body <tt>''i''</tt>

+

| : name of solid body <tt>''i''</tt>

|-

| <tt>''CELLi''</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>''MATi''</tt>

−

| : ~~Material~~ filling cell <tt>''i''</tt>

+

| : material filling cell <tt>''i''</tt>

|-

| <tt>''FILEi''</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>''SCALEi''</tt>

−

| : ~~Scaling~~ factor for the STL model in <tt>''FILEi''</tt>

+

| : scaling factor for the STL model in <tt>''FILEi''</tt>

|-

| <tt>''Xi''</tt>

−

| : ~~Shift~~ in x-direction to the STL model in <tt>''FILEi''</tt>

+

| : shift in x-direction to the STL model in <tt>''FILEi''</tt>

|-

| <tt>''Yi''</tt>

−

| : ~~Shift~~ in y-direction to the STL model in <tt>''FILEi''</tt>

+

| : shift in y-direction to the STL model in <tt>''FILEi''</tt>

|-

| <tt>''Zi''</tt>

−

| : ~~Shift~~ in z-direction to the STL model in <tt>''FILEi''</tt>

+

| : shift in z-direction to the STL model in <tt>''FILEi''</tt>

|}

Line 2,226:

Line 2,764:

{|

| <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 2,248:

Line 2,786:

[ [[#src_ss|'''ss''']] ''SURF'' ]

[ [[#src_sd|'''sd''']] ''U'' ''V'' ''W'' ]

+

[ [[#src_sa|'''sa''']] ''PHI'' ]

[ [[#src_se|'''se''']] ''E'' ]

[ [[#src_sb|'''sb''']] ''N'' ''INTT'' ''E1'' ''WGT1'' ''E2'' ''WGT2'' ... ]

Line 2,255:

Line 2,794:

[ [[#src_si|'''si''']] ''N'' ''P1'' ''P2'' ... ]

[ [[#src_sg|'''sg''']] ''MAT'' ''MODE'' ]

−

Source definition. The first ~~parameter~~:

+

Source definition. The two first parameters:

{|

+

| <tt>''NAME''</tt>

+

|: source name

+

|-

| <tt>''PART''</tt>

| : 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.

+

Notes:

+

*The particle type <tt>''PART''</tt> is optional in single particle simulations.

+

*A single source card may include one or several source types.

+

Source types:

Source weight (<tt>'''sw'''</tt>):

Line 2,286:

Line 2,834:

Notes:

*Setting a source cell is one of the options that can be applied to define the spatial distribution of source particles.

−

*The selection is based on rejection sampling, and if the source cell occupies a small volume of the geometry, the sampling efficiency can be increased by defining a bounding box/cylinder around the cell (using the <tt>''sx''</tt>, <tt>''sy''</tt> and <tt>''sz''</tt> or <tt>''sp''</tt>, <tt>''srad''</tt> and <tt>''sz''</tt> options, respectively).

+

*The selection is based on rejection sampling, and if the source cell occupies a small volume of the geometry, the sampling efficiency can be increased by defining a bounding box/(vertical) cylinder around the cell (using the <tt>'''sx'''</tt>, <tt>'''sy'''</tt> and <tt>'''sz'''</tt> or <tt>'''sp'''</tt>, <tt>'''srad'''</tt> and <tt>'''sz'''</tt> options, respectively).

*If no spatial distribution is defined, particles are sampled uniformly over the geometry.

Line 2,307:

Line 2,855:

Notes:

*Setting a source material is one of the options that can be applied to define the spatial distribution of source particles.

−

*The selection is based on rejection sampling, and if the source material occupies a small volume of the geometry, the sampling efficiency can be increased by defining a bounding box/cylinder around the cell (using the <tt>''sx''</tt>, <tt>''sy''</tt> and <tt>''sz''</tt> or <tt>''sp''</tt>, <tt>''srad''</tt> and <tt>''sz''</tt> options, respectively).

+

*The selection is based on rejection sampling, and if the source material occupies a small volume of the geometry, the sampling efficiency can be increased by defining a bounding box/(vertical) cylinder around the cell (using the <tt>'''sx'''</tt>, <tt>'''sy'''</tt> and <tt>'''sz'''</tt> or <tt>'''sp'''</tt>, <tt>'''srad'''</tt> and <tt>'''sz'''</tt> options, respectively).

*If no spatial distribution is defined, particles are sampled uniformly over the geometry.

Line 2,315:

Line 2,863:

{|

| <tt>''X''</tt>, <tt>''Y''</tt>, <tt>''Z''</tt>,

−

| : coordinates of the source point

+

| : coordinates of the source point [in cm]

|}

Line 2,327:

Line 2,875:

{|

| <tt>''XMIN''</tt>, <tt>''XMAX''</tt>

−

| : ~~Boundaries~~ on X-axis

+

| : boundaries on X-axis [in cm]

|-

| <tt>''YMIN''</tt>, <tt>''YMAX''</tt>

−

| : ~~Boundaries~~ on Y-axis

+

| : boundaries on Y-axis [in cm]

|-

| <tt>''ZMIN''</tt>, <tt>''ZMAX''</tt>

−

| : ~~Boundaries~~ on Z-axis

+

| : boundaries on Z-axis [in cm]

|-

| <tt>''RMIN''</tt>, <tt>''RMAX''</tt>

−

| : ~~Radial~~ boundaries

+

| : radial boundaries [in cm]

|-

|}

Notes:

−

*Source boundaries are used to define a bounding box/cylinder inside which the source particles are sampled.

+

*Source boundaries are used to define a bounding box/(vertical) cylinder inside which the source particles are sampled.

*The radial boundaries are centered around the point defined by <tt>'''sp'''</tt> and can be used in combination with <tt>'''sz'''</tt>.

*Can be used in combination with cell and material sources to increase the sampling efficiency.

Line 2,347:

Line 2,895:

−

~~Surface source~~ (<tt>'''ss'''</tt>):

+

Source surface (<tt>'''ss'''</tt>):

{|

Line 2,356:

Line 2,904:

Notes:

*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 2,370:

Line 2,919:

*If no directional dependence is defined, the direction of source particles is sampled isotropically.

+

Source angular-aperture (<tt>'''sa'''</tt>):

+

{|

+

| <tt>''PHI''</tt>

+

| : polar angle [in degrees]

+

|}

+

Notes

+

*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>):

Line 2,375:

Line 2,935:

{|

| <tt>''E''</tt>

−

| : energy of source particles

+

| : energy of source particles [in MeV]

|}

Line 2,394:

Line 2,954:

|-

| <tt>''En''</tt>

−

| : upper boundary of the energy bin

+

| : upper boundary of the energy bin [in MeV]

|-

| <tt>''WGTn''</tt>

Line 2,404:

Line 2,964:

*The bins are entered in the order of ascending energy, and weight of the first bin must be set to zero.

*Interpolation is given in a separate parameter from version 2.1.31 on.

+

*Here, a simple test input that demonstrates the [[Source energy spectrum definition test case|source spectrum definition]].

Source reaction (<tt>'''sr'''</tt>):

Line 2,412:

Line 2,973:

|-

| <tt>''MT''</tt>

−

| : reaction number

+

| : reaction number identifier

|}

Line 2,426:

Line 2,987:

{|

| <tt>''TMIN''</tt>, <tt>''TMAX''</tt>

−

| : time boundaries

+

| : time boundaries [in s]

|}

Line 2,445:

Line 3,006:

Notes:

−

*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.

Line 2,470:

Line 3,031:

{|

| <tt>''MAT''</tt>

−

| : material name or -1

+

| : material name or "-1" to refer to all radioactive materials

|-

| <tt>''MODE''</tt>

Line 2,477:

Line 3,038:

Notes:

−

*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).

−

*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 implicit sampling mode preserves the total statistical weight of emitted particles and produces a uniform source distribution over activated materials.

Line 2,488:

Line 3,048:

=== srtrans (source transformation) ===

−

See [[#trans (transformations)|transformations]].

+

Defines source transformations. Shortcut for "<tt>trans sr</tt>".

+

Notes:

+

*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) ===

−

See [[#trans (transformations)|transformations]].

+

Defines surface transformations. Shortcut for "<tt>trans s</tt>".

+

Notes:

+

*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) ===

Line 2,527:

Line 3,095:

'''thermstoch''' ''NAME'' ''TEMP'' ''LIB1'' ''LIB2''

−

Defines thermal scattering data that can be linked to nuclides using input entry '''moder''' in the [[#mat (material definition)|material cards]]. When using thermal scattering data together with TMS on-the-fly temperature treatment, the third input value of the therm card is 0. In this case, Serpent interpolates the thermal scattering data automatically to the local temperature, as defined either using the '''tms''' input entry in the material definition or via the multi-physics interface.

+

Defines thermal scattering data that can be linked to nuclides using input entry '''moder''' in the [[#mat (material definition)|material cards]]. Input values:

−

~~Input values:~~

{|

| <tt>''NAME''</tt>

Line 2,538:

Line 3,105:

|-

| <tt>''TEMP''</tt>

−

| : temperature to which the thermal scattering data is interpolated

+

| : temperature to which the thermal scattering data is interpolated [in K]

|}

Notes:

−

+

* On-the-fly thermal motion sampling (TMS) temperature treatment:

−

*~~When using on~~-the-fly ~~interpolation~~ of thermal scattering data, <tt>''LIBi''</tt> must cover the whole ~~temperature~~ range in which the materials appear in the geometry~~. I~~.e. ~~extrapolation of the~~ data is not supported.

+

**It requires the third value of the therm card to be set to "<tt>0</tt>"

−

*Thermal scattering data is interpolated using the methodology of makxsf code

+

**The thermal scattering data is automatically interpolated to the local temperature.

−

*~~The~~ interpolation can be performed using the stochastic mixing approach with '''thermstoch'''. This interpolation mode ~~is not available for~~ on-the-fly interpolation.

+

**The local temperature is either defined using:

−

*The continuous S(α β) formalism is available from version 2.1.32 on.

+

*** the '''tms''' entry in the material card (see [[#mat|mat card]])

+

*** the multi-physics interface (see [[#ifc|ifc card]]), where the temperature limits are defined using the '''tft''' entry in the material card (see [[#mat|mat card]])

+

**The thermal scattering libraries <tt>''LIBi''</tt> must cover the whole range in which the materials appear in the geometry, i.e. data extrapolation is not supported.

+

*Interpolation:

+

**Thermal scattering data is interpolated using the methodology of makxsf code<ref name="makxsf">Brown, F. B. ''"The makxsf Code with Doppler Broadening"'', Los Alamos National Laboratory Tech. Rep., LA-UR-06-7002, Los Alamos, NM, [https://mcnp.lanl.gov/pdf_files/TechReport_2006_LANL_LA-UR-06-7002_Brown.pdf 2006]</ref>.

+

**Alternatively, the interpolation can be performed using the stochastic mixing approach with the '''thermstoch''' entry.

+

***This interpolation mode doesn't support on-the-fly interpolation.

+

*The continuous S(α, β) formalism:

+

**It is available from version 2.1.32 on.

+

**The on-the-fly temperature treatment is available from version 2.2.0 and on.

=== tme (time binning definition) ===

Line 2,566:

Line 3,142:

|-

| <tt>''LIMn''</tt>

−

| : time bin boundaries in arbitrary binning

+

| : time bin boundaries in arbitrary binning [in s]

|-

| <tt>''Tmin''</tt>

−

| : minimum time boundary in uniform or log-uniform binning

+

| : minimum time boundary in uniform or log-uniform binning [in s]

|-

| <tt>''Tmax''</tt>

−

| : maximum time boundary in uniform or log-uniform binning

+

| : maximum time boundary in uniform or log-uniform binning [in s]

|}

Notes:

−

*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.

*Time binning is used with [[#det (detector definition)|detectors]] and [[Dynamic external source simulation mode|dynamic simulation mode]].

Line 2,603:

Line 3,178:

|-

| <tt>''IDX''</tt>

−

| : index number in lattice transformation

+

| : index number in lattice transformation (type L)

|-

| <tt>''LVL''</tt>

Line 2,609:

Line 3,184:

|-

| <tt>''X'',''Y'',''Z''</tt>

−

| : translation vector

+

| : translation vector [in cm]

|-

| <tt>''θx'' ''θy'' ''θz''</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>''α1'' ... ''α9''</tt>

Line 2,621:

Line 3,196:

|-

| <tt>''X0'',''Y0'',''Z0''</tt>

−

| : ~~Origin~~ of vector defining rotation axis.

+

| : origin of vector defining rotation axis [in cm]

|-

| <tt>''I'',''J'',''K''</tt>

−

| : ~~Components~~ of vector defining rotation axis.

+

| : components of vector defining rotation axis.

|-

| <tt>''β''</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 2,636:

Line 3,211:

*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>.

−

*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.

*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>''α1'' ... ''α9''</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 ''β'' 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) ===

Line 2,649:

Line 3,226:

{|

| <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>

Line 2,656:

Line 3,233:

Notes:

−

*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) ===

Line 2,675:

Line 3,254:

|-

| <tt>''IDX''</tt>

−

| : index number in lattice transformation

+

| : index number in lattice transformation (type L)

|-

| <tt>''T0''</tt>

−

| : beginning time of the transformation

+

| : beginning time of the transformation [in s]

|-

| <tt>''T1''</tt>

−

| : end time of the transformation

+

| : end time of the transformation [in s]

|-

| <tt>''TTYPE''</tt>

Line 2,687:

Line 3,266:

|-

| <tt>''VX'',''VY'',''VZ''</tt>

−

| : ~~Initial~~ velocity vector

+

| : initial velocity vector [in cm/s]

|-

| <tt>''AX'',''AY'',''AZ''</tt>

−

| : ~~Initial~~ acceleration vector

+

| : initial acceleration vector [in cm/s2]

|}

Line 2,699:

Line 3,278:

*See [[UGM 2016 Moving geometry]].

*See [[Rotating Translating STL Bunny]].

+

===umsh (unstructured mesh-based geometry definition) ===

+

''UNI'' ''BGUNI''

+

''MESH_SPLIT'' ''MESH_DIM'' ''SZ1'' ''SZ2'' ... ''SZMESH_DIM''

+

''POINTS_FILE''

+

''FACES_FILE''

+

''OWNER_FILE''

+

''NEIGHBOUR_FILE''

+

''MATERIALS_FILE''

+

Defines an unstructured mesh-based geometry. Input values:

+

{|

+

| <tt>''UNI''</tt>

+

| : universe name for the unstructured mesh-based geometry

+

|-

+

| <tt>''BGUNI''</tt>

+

| : name of the background universe filling all undefined space

+

|-

+

| <tt>''MESH_SPLIT''</tt>

+

| : splitting criterion for the adaptive search mesh (maximum number of geometry cells in search mesh cell)

+

|-

+

| <tt>''MESH_DIM''</tt>

+

| : number of levels in the adaptive search mesh

+

|-

+

| <tt>''SZi''</tt>

+

| : size of the search mesh at level <tt>''i''</tt>

+

|-

+

| <tt>''POINTS_FILE''</tt>

+

| : path to the unstructured mesh points file

+

|-

+

| <tt>''FACES_FILE''</tt>

+

| : path to the unstructured mesh faces file

+

|-

+

| <tt>''OWNER_FILE''</tt>

+

| : path to the unstructured mesh owner file

+

|-

+

| <tt>''NEIGHBOUR_FILE''</tt>

+

| : path to the unstructured mesh neighbour file

+

|-

+

| <tt>''MATERIALS_FILE''</tt>

+

| : path to the unstructured mesh materials file

+

|}

+

Notes:

+

*See also description of [[#solid (irregular 3D geometry definition)|solid (irregular 3D geometry definition), type 1]].

=== utrans (universe transformation) ===

−

See [[#trans (transformations)|transformations]].

+

Defines universe transformations. Shortcut for "<tt>trans u</tt>".

+

Notes:

+

*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) ===

+

'''voro''' ''UNI0'' ''UNIbg'' ''R0'' '''-1''' ''NP'' ''UNI1'' ''VF1'' [ ''UNI2'' ''VF2'' ... ]

+

'''voro''' ''UNI0'' ''UNIbg'' ''R0'' ''FILE''

+

Defines a stochastic Voronoi tessellation geometry. Input values:

+

{|

+

| <tt>''UNI0''</tt>

+

|: universe name for the Voronoi medium

+

|-

+

| <tt>''UNIbg''</tt>

+

|: background universe name filling all undefined space

+

|-

+

| <tt>''R0''</tt>

+

|: test radius [in cm]

+

|-

+

| <tt>''NP''</tt>

+

|: number of seed points

+

|-

+

| <tt>''UNIm''</tt>

+

|: sub-universe name for the ''m''-th random fragmented polyhedral zone

+

|-

+

| <tt>''VFm''</tt>

+

|: volume fraction associated to ''m''-th random fragmented polyhedral zone

+

|-

+

| <tt>''FILE''</tt>

+

|: input file containing the Voronoi data

+

|}

+

The syntax of the file containing the Voronoi seed points data is:

+

::{| class="toccolours" style="text-align: left;"

+

|-

+

| ''X1'' ''Y1'' ''Z1'' ''UNI1''

+

|-

+

| ''X2'' ''Y2'' ''Z2'' ''UNI1''

+

|-

+

| ...

+

|-

+

| ''XN'' ''YN'' ''ZN'' ''UNI1''

+

|-

+

| ''XN+1'' ''YN+1'' ''ZN+1'' ''UNI2''

+

|-

+

| ...

+

|}

+

where:

+

{|

+

| <tt>''Xn'', ''Yn'', ''Zn''</tt>

+

|: seed points coordinates [in cm]

+

|-

+

| <tt>''UNIm''</tt>

+

|: sub-universe name for the ''m''-th random zone associated to the given seed point

+

|}

+

Notes:

+

*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) ===

Line 2,747:

Line 3,443:

|-

| <tt>''MINn''</tt>

−

| : minimum mesh boundary (''n''th coordinate)

+

| : minimum mesh boundary (''n''-th coordinate)

|-

| <tt>''MAXn''</tt>

−

| : maximum mesh boundary (''n''th coordinate)

+

| : maximum mesh boundary (''n''-th coordinate)

|-

| <tt>''SZn''</tt>

−

| : number of mesh cells (''n''th coordinate)

+

| : number of mesh cells (''n''-th coordinate)

|-

| <tt>''LIMnm''</tt>

−

| : ~~mth~~ boundary of ''n''th coordinate)

+

| : mesh boundary ''m''-th (''n''-th coordinate)

|-

| <tt>''X0'', ''Y0''</tt>

Line 2,807:

Line 3,503:

|}

−

~~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.

Notes:

*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]].

*See also [[#set wwb|set wwb]] and [[#set maxsplit|set maxsplit]] for setting options for weight windows, splitting and Russian roulette.

Line 2,818:

Line 3,514:

*This capability is still very much under development. The input syntax may be revised at some point.

+

Weight-window mesh paramters:

Mesh file (<tt>'''wf'''</tt>):

Line 2,845:

Line 3,543:

|-

| <tt>''E''</tt>

−

| : energy used for renormalization

+

| : energy used for renormalization [in MeV]

|}

Line 2,945:

Line 3,643:

|-

| <tt>''DSPLi''</tt>

−

| : density split criterion (negative values ~~for~~ mass~~, positive values for atomic~~ density)

+

| : density split criterion (positive value = atomic density [in b-1cm-1], negative values = mass density [in g/cm3])

|-

| <tt>''SZi''</tt>

−

| : minimum cell dimension

+

| : minimum cell dimension [in cm]

|}

Line 2,955:

Line 3,653:

*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.

−

*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.

*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 2,968:

Line 3,666:

'''set absrate''' ''A'' [ ''MAT'' ]

−

Sets normalization to total absorption rate.

+

Sets normalization to total absorption rate. Input values:

{|

| <tt>''F''</tt>

−

| : number of neutrons absorbed per second (neutrons/s)

+

| : number of neutrons absorbed per second [in neutrons/s]

|-

| <tt>''MAT''</tt>

Line 2,981:

Line 3,679:

*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.

−

*~~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.

*See also Section 5.8 of [http://montecarlo.vtt.fi/download/Serpent_manual.pdf Serpent 1 User Manual].

Line 2,999:

Line 3,698:

*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 cross section directory 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 cross section directory files in this path if not found at the absolute.

−

*A default cross section directory file can be set by defining environment variable SERPENT_ACELIB. This file will be used if no other path is given with '''set acelib'''.

+

*A default cross section directory file can be set by defining environment variable <tt>SERPENT_ACELIB</tt>. This file will be used if no other path is given with '''set acelib'''.

=== set adf ===

'''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 3,018:

Line 3,717:

|-

| <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 3,024:

Line 3,723:

*The surface enclosing the universe can be super-imposed (i.e. not part of the geometry definition), but it must enclose the entire 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 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 3,034:

Line 3,733:

*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).

+

*Sign moments of net and partial currents are not scored for [[Surface_types#Regular_prisms|Y-type infinite/truncated hexagonal prisms]].

=== set alb ===

Line 3,055:

Line 3,755:

*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 entire 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 universe is needed only for labelling the results in the output files.

Line 3,068:

Line 3,768:

{|

| <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>

−

| : 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)

|}

Notes:

−

*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]].

Line 3,092:

Line 3,792:

*Some burnup applications require separate treatment for isotopes that are used as burnable absorbers but also produced in fission. This input parameter can be used to separate the transmutation chains.

*Isotope handled as the burnable absorber is created by duplicating the original and renaming it as ''ZAIn + 1000''.

−

*For Gd-155, for example, the fission product isotope would be assigned ZAI 641550 and the burnable absorber ZAI 642550.

+

**For Gd-155, for example, the fission product isotope would be assigned ZAI 641550 and the burnable absorber ZAI 642550.

−

*The input parameter defines the entire transmutation chain. Listing Gd-isotopes 641540 641550 641560 641570 641580 creates a transmutation path from Gd-154 to Gd-158. Listing only the main absorbers (641550 641570) produces a different result, since the capture products of Gd-155 and Gd-157 are lost.

+

*The input parameter defines the entire transmutation chain.

+

**Listing Gd-isotopes 641540 641550 641560 641570 641580 creates a transmutation path from Gd-154 to Gd-158.

+

**Listing only the main absorbers (641550 641570) produces a different result, since the capture products of Gd-155 and Gd-157 are lost.

=== set bala ===

Line 3,102:

Line 3,804:

{|

| <tt>''OPT''</tt>

−

| : ~~option~~ (0 = off, 1 = on)

+

| : probability to store particles in common queue (0 = off, >0 = on)

|}

Notes:

*Load balancing may improve OpenMP parallel scalability in calculations with significant branching (most typically related to coupled neutron/photon calculations or variance reduction).

−

*The option is ~~off~~ by default.

+

*The option is ON with <tt>''OPT''</tt>= 1 with weight-window/variance reduction calculations and dynamic/time-dependent calculation modes. Otherwise, it is set OFF by default. Before version 2.2.0, the default behavior was always OFF.

*When this option is set, the random number sequence is no longer preserved.

Line 3,165:

Line 3,867:

Notes:

*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.

*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 3,171:

Line 3,874:

*For symmetry purposes Serpent provides the [[#set usym|universe symmetry option]].

*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 ===

Line 3,201:

Line 3,903:

Notes:

−

*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.

*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 ===

Line 3,216:

Line 3,916:

{|

| <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>

−

| : weight lower-boundary

+

| : weight lower-boundary (default value: 0.2)

|-

| <tt>''WGT_HIGH''</tt>

−

| : weight upper-boundary

+

| : weight upper-boundary (default value: 10.0)

|}

Notes:

−

*The default weight limits for the branchless methods are 0.2 and 10

*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 ===

Line 3,238:

Line 3,940:

{|

| <tt>''MODE''</tt>

−

| : burnup calculation mode

+

| : burnup calculation mode (default value: 2 = CRAM)

|-

| <tt>''ORDER''</tt>

−

| : CRAM order

+

| : CRAM order (default value: 14 - 14 PFD CRAM)

|-

| <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:

−

{| class="wikitable" style="text-align: left;"

+

::{| class="wikitable" style="text-align: left;"

! Mode

! Description

Line 3,262:

Line 3,964:

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

−

|-

| <tt>2</tt>

−

|-

| <tt>4</tt>

−

|-

| <tt>6</tt>

−

|-

| <tt>8</tt>

−

|-

| <tt>10</tt>

−

|-

| <tt>12</tt>

−

|-

| <tt>14</tt>

−

|-

| <tt>16</tt>

−

|-

| <tt>-16</tt>

−

|-

| <tt>-48</tt>

|}

Notes:

−

*~~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.

+

*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.

−

*Negative values of CRAM order mean using IPF form of CRAM with order of the absolute value of the parameter.

+

*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.

−

*Positive values refer to PFD form of CRAM.

+

*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.

−

*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.

+

*Version 2.2.0 includes the sub-step method for depletion calculations involving continuous reprocessing.

−

*Positive values of <tt>''SSD''</tt> enforce usage of CRAM with given number of substeps.

+

=== set bunorm ===

+

'''set bunorm''' ''NORM''

+

Sets the burnup calculation normalization mode if it is not bound to a single material. Input values:

+

{|

+

| <tt>''NORM''</tt>

+

| : burnup calculation normalization mode (1 = all materials, 2 = burnable materials, 3 = non-burnable materials). (default value: 1)

+

|}

=== set ccmaxiter ===

Line 3,301:

Line 4,003:

{|

| <tt>''NITER''</tt>

−

| : number of iterations.

+

| : number of iterations (default value: 1 = no iteration)

|}

Notes:

−

*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.

*See [[Coupled multi-physics calculations]] for further information.

Line 3,317:

Line 4,018:

{|

| <tt>''CPOP''</tt>

−

| : total active population to simulate.

+

| : total active population to simulate (default value: [[Definitions, units and constants#Constants|INFTY]]/1E6)

|}

Notes:

−

*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.

*Only the population simulated during active cycles is included in this amount.

Line 3,335:

Line 4,035:

{|

| <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 3,349:

Line 4,049:

{|

| <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 3,362:

Line 4,062:

{|

| <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>

−

| : 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>

−

| : 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>

−

| : minimum mean time interval for scoring the CFE for photons

+

| : minimum mean time interval for scoring the CFE for photons [in s]

|}

Line 3,378:

Line 4,078:

*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.

−

*~~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 no studies have been performed on what the optimal value should be.

+

*Adjusting the distance affects both statistics and running time, but it should be noted that no studies have been performed on what the optimal value should be.

−

*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]].

*The collision flux estimator in Serpent is described in an article in Annals of Nuclear energy from 2017.<ref name="cfe">Leppänen, J.

Line 3,411:

Line 4,111:

{|

| <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>''PARAMn''</tt>

Line 3,421:

Line 4,121:

*The group constant output file <tt>[input].coe</tt> is produced when the [[automated burnup sequence]] is invoked.

*The available parameters are listed under [[Output parameters#Homogenized group constants|homogenized group constants]] in the description of the <tt>[input]_res.m</tt> output file.

−

*Detectors are identified by the name assigned to them in the [[#det (detector definition)|~~detector~~ card]].

+

*Detectors are identified by the name assigned to them in the [[#det (detector definition)|det card]].

=== set combing ===

Line 3,430:

Line 4,130:

{|

| <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)

|}

Notes:

−

*The combing method can achieve variance reduction and save computer time by keeping the population size approximately constant over time steps. In super-critical systems, it prevents the population from growing without bound ~~while, in~~ sub-critical systems, it ~~does it~~ from dying. In critical systems, it avoids the divergence of the variance of the population due to fluctuations of fission chains.

+

*The combing method can achieve variance reduction and save computer time by keeping the population size approximately constant over time steps.

+

**In super-critical systems, it prevents the population from growing without bound.

+

**In sub-critical systems, it prevents the population from dying.

+

**In critical systems, it avoids the divergence of the variance of the population due to fluctuations of fission chains.

=== set comfile ===

Line 3,452:

Line 4,155:

*Setting up a communication mode will enable the coupled calculation mode.

+

*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]]

Line 3,461:

Line 4,165:

{|

| <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>"

|}

Notes:

−

*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 ===

Line 3,475:

Line 4,179:

{|

| <tt>''LIBn''</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)

|}

Notes:

+

*It enables first-order uncertainty propagation by collapsing the covariance data with the evaluated sensitivities.

+

**It applies the Sandwich rule: <math>\mathbf{Cov}^R_X \approx ({\overline{\mathbf{S}}}^R_X)^T {\overline{\overline{\mathbf{Cov}}}}^R_{X,X} \overline{\mathbf{S}}^R_X</math>

−

*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~~.

+

:: where: <math>{\overline{\mathbf{S}}}^R_X \in \R^{N_g \times 1} </math> is the sensitivity vector containing the group sensitivities

+

::::<math>{\overline{\overline{\mathbf{Cov}}}}^R_{X,X} \in \R^{N_g \times N_g}</math> is the covariance matrix containing the group covariances

+

**The methodology is described in a related report<ref name="UncReport18">Valtavirta, V. ''"Nuclear data uncertainty propagation to Serpent generated group and time constants"'', Research report VTT-R-04681-18 [http://montecarlo.vtt.fi/download/VTT-R-04681-18_web.pdf (2018)].</ref>.

+

*It requires setting the sensitivity calculation parameters. For more information, see the detailed description on [[Sensitivity calculations|sensitivity calculations]].

=== set covlib ===

Line 3,492:

Line 4,201:

|}

+

The syntax of the file containing the covariance data is:

+

::{| class="toccolours" style="text-align: left;"

+

|-

+

| ''NG'' ''E1'' ... ''ENG+1'' ''NM''

+

|-

+

| ''ZAI1,1'' ''MT1,1'' ''ZAI1,2'' ''MT1,2''

+

|-

+

| ''COV1,1,1'' ... ''COV1,NG,NG''

+

|-

+

|...

+

|-

+

| ''ZAINM,1'' ''MTNM,1'' ''ZAINM,2'' ''MTNM,2''

+

|-

+

| ''COVNM,1,1'' ... ''COVNM,NG,NG''

+

|-

+

|}

+

where:

+

{|

+

| <tt>''NG''</tt>

+

|: number of neutron energy groups

+

|-

+

| <tt>''Eg''</tt>

+

|: energy grid boundaries [in MeV]

+

|-

+

| <tt>''NM''</tt>

+

|: number of covariance matrixes

+

|-

+

| <tt>''ZAIm,n'', ''MTm,n''</tt>

+

|: 2 × nuclide-reaction pairs defining the ''m''-th covariance matrix

+

|-

+

| <tt>''COV''m,g,g''</tt>

+

|: ''NG'' × ''NG'' covariance data corresponding to the ''m''-th matrix

+

|}

+

Notes:

+

*It enables first-order uncertainty propagation by collapsing the covariance data with the evaluated sensitivities.

+

**It applies the Sandwich rule: <math>\mathbf{Cov}^R_X \approx ({\overline{\mathbf{S}}}^R_X)^T {\overline{\overline{\mathbf{Cov}}}}^R_{X,X} \overline{\mathbf{S}}^R_X</math>

−

*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~~.

+

:: where: <math>{\overline{\mathbf{S}}}^R_X \in \R^{N_g \times 1} </math> is the sensitivity vector containing the group sensitivities

+

::::<math>{\overline{\overline{\mathbf{Cov}}}}^R_{X,X} \in \R^{N_g \times N_g}</math> is the covariance matrix containing the group covariances

+

**The methodology is described in a related report<ref name="UncReport18">Valtavirta, V. ''"Nuclear data uncertainty propagation to Serpent generated group and time constants"'', Research report VTT-R-04681-18 [http://montecarlo.vtt.fi/download/VTT-R-04681-18_web.pdf (2018)].</ref>.

+

*It requires setting the sensitivity calculation parameters. For more information, see the detailed description on [[Sensitivity calculations|sensitivity calculations]].

=== set cpd ===

−

'''set cpd''' ''DEPTH'' [ ''NZ ZMIN ZMAX'' ]

+

'''set cpd''' ''DEPTH'' [ ''NZ ZMIN ZMAX'' ] [ ''LVL1'' ''LVL2'' ]

Sets on the calculation of lattice-wise power distributions to output file <tt>[input]_core0.m</tt> on. Input values:

{|

| <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>''NZ''</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>''ZMIN''</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>''ZMAX''</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>

+

|: User-defined first level where to define the lattice-wise power distribution

|-

+

| <tt>''LVL2''</tt>

+

|: User-defined second level where to define the lattice-wise power distribution

|}

Notes:

−

* The ~~default values for~~ <tt>''~~NZ~~''</tt>, <tt>''~~ZMIN~~''</tt> ~~and <tt>''ZMAX''</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''' ''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 3,555:

Line 4,314:

*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''' ''TABLE_LIST''

+

Defines the tables included in the nuclear and material data file <tt>[input].out</tt>. Input values:

+

{|

+

| <tt>''TABLE_LIST''</tt>

+

| : 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

+

|}

+

Notes:

+

*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.

+

*For more information, see detailed description of the [[Description of output files#Nuclide and material data output|nuclear and material data output]].

=== set dbrc ===

Line 3,563:

Line 4,381:

{|

| <tt>''Emin''</tt>

−

| : Minimum energy for DBRC

+

| : Minimum energy for DBRC [in MeV]

|-

| <tt>''Emax''</tt>

−

| : Maximum energy for DBRC

+

| : Maximum energy for DBRC [in MeV]

|-

| <tt>''NUCn''</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")

|}

Notes:

−

*This description is not complete.

*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].

Line 3,580:

Line 4,397:

=== set dd ===

−

'''set dd''' ''MODE''

+

'''set dd''' ''MODE'' [ ''X0'' ''Y0'' ''α0'' ]

−

Invokes domain decomposition. Input values

+

Invokes domain decomposition. Input values:

{|

| <tt>''MODE''</tt>

−

| : decomposition mode (0 = none, 1 ~~= simple,~~ 2 = sector, 3 = sector + ~~center)~~

+

| : decomposition mode (default value: 0)

+

|-

+

| <tt>''X0''</tt>

+

| 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>''Y0''</tt>

+

| 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>''α0''</tt>

+

| 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>''X0''</tt>, <tt>''Y0''</tt> and <tt>''α0''</tt> are available

+

|-

+

| <tt>3</tt>

+

| sector-based + central division decomposition

+

| <tt>''X0''</tt>, <tt>''Y0''</tt> and <tt>''α0''</tt> are available, only applicable if MPI-tasks > 4

|}

Notes:

*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]].

−

*Decomposed materials are plotted in domain-specific colors (unless the rgb ~~option~~ in the material card is used)

+

−

*~~See~~ [[Domain decomposition|~~practical example~~]] ~~for more information~~.

+

=== set declib ===

Line 3,609:

Line 4,457:

Notes:

−

*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.

−

*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 <tt>SERPENT_DATA</tt>. The code looks for decay data files in this path if not found at the absolute.

+

*From version 2.2.0 and on, a default decay data library directory file can be set by defining environment variable <tt>SERPENT_DECLIB</tt>. This file will be used if no other path is given with '''set declib'''.

=== set decomp ===

Line 3,643:

Line 4,492:

Notes:

−

+

* 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.

Line 3,651:

Line 4,501:

'''set depmtx''' ''MODE''

−

Print burnup matrixes to <tt>~~depmtx_~~[mat][~~step~~].m</tt> file during burnup calculation.

+

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>

−

| : ~~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 3,661:

Line 4,511:

*With non-constant predictor, this option will stop the simulation up to version 2.1.31.

*With multiple substeps, only the last one is kept after the simulation up to version 2.1.31.

+

*The burnup matrix output is named <tt>depmtx_[mat][bu].m</tt> up to version 2.1.31.

=== set depout ===

Line 3,668:

Line 4,519:

{|

| <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>

−

| : 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)

|}

Notes:

*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.

+

=== set deppara ===

+

'''set deppara''' ''PARAM_LIST''

+

Defines the material- and isotopic-wise variables included in the depletion output file <tt>[input]_dep.m</tt>. Input values:

+

{|

+

| <tt>''PARAM_LIST''</tt>

+

| : 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 b-1cm-1]

+

|-

+

| <tt>mass</tt>

+

| mass density

+

| <tt>MDENS</tt>

+

| [in g/cm3]

+

|-

+

| <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

+

|}

+

Notes:

+

*For more information, see detailed description of the [[Description of output files#Burnup calculation output|burnup calculation output]].

+

=== set depstepbunorm ===

+

'''set depstepbunorm''' ''NORM''

+

Sets the depletion step normalization in burnup calculations based on energy deposition. Input values:

+

{|

+

| <tt>''NORM''</tt>

+

| depletion step normalization mode based on energy deposition (1 = all materials, 2 = burnable materials)

+

|}

+

Notes

+

* Default values (see [[#set edepmode|set edepmode]]):

+

** For energy deposition modes 0/1: the 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 ===

Line 3,686:

Line 4,623:

{|

| <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>

−

| : 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>

−

| : number of points for trapezoidal integration for homogeneous flux

+

| : number of points for trapezoidal integration for homogeneous flux (default value: 100)

|}

Notes:

−

*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]].

*The format was revised in update 2.1.27 (<tt>''DC''</tt> option was added between <tt>''MODE''</tt> and <tt>''NP''</tt>).

Line 3,705:

Line 4,640:

=== set dix ===

'''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 3,716:

Line 4,651:

''"Two practical methods for unionized energy grid construction in continuous-energy Monte Carlo neutron transport calculation."'' Ann. Nucl. Energy [https://www.sciencedirect.com/science/article/pii/S0306454909001108 '''36''' (2009) 878-885].</ref> is a method to speed-up the cross section look-up when energy grid unionization is not used for microscopic data.

*The method can be used only in [[#set_opti|optimization modes]] 1 and 3 (modes 2 and 4 are based on energy grid unionization).

+

=== set dspec ===

+

'''set dspec''' ''EGRIDp'' ''EGRIDn''

+

Sets the energy grid structure for decay spectra. Input values:

+

{|

+

| <tt>''EGRIDp''</tt>

+

| : energy grid structure for photons

+

|-

+

| <tt>''EGRIDn''</tt>

+

| : energy grid structure for neutrons

+

|}

+

Notes:

+

*The photon/neutron decay spectra is printed in the <tt>[input]_gsrc.m</tt> or <tt>[input]_nsrc.m</tt> output file, respectively.

+

*The energy group spectra only include the contribution from the discrete/line spectra.

+

*Setting "-1" instead of providing the energy grid structure disables the option for the given particle type.

=== set dt ===

Line 3,724:

Line 4,677:

{|

| <tt>''NTRSH''</tt>

−

| : probability threshold for neutrons

+

| : probability threshold for neutrons (default value: 0.9)

|-

| <tt>''GTRSH''</tt>

−

| : probability threshold for photons

+

| : probability threshold for photons (default value: 0.9)

|}

Notes:

−

+

*Tracking algorithm mode:

−

*~~Serpent uses delta~~-tracking by default for both neutrons and photons~~, but switches~~ to surface-tracking if the probability of sampling virtual collisions (ratio between material total cross section and the majorant) exceeds the given threshold.

+

**Delta-tracking is used by default for both neutrons and photons

−

*~~Default~~ probability threshold for both particle types is 0.9, i.e. delta-tracking is used if the ratio between total cross section and majorant is above 0.1.

+

**Switch to surface-tracking happens if the probability of sampling virtual collisions (ratio between material total cross section and the majorant) exceeds the given threshold.

−

*~~set dt 1 means that~~ delta-tracking is always ~~used and set dt 0 that it is never used~~.

+

*Probability threshold:

+

**The default probability threshold for both particle types is 0.9: i.e. delta-tracking is used if the ratio between total cross section and majorant is above 0.1.

+

**To enforce delta-tracking mode always: use "<tt>1</tt>".

+

**To enforce surface-tracking mode always: use "<tt>0</tt>"

*Use of delta-tracking can be enforced or blocked in individual materials using the [[#set forcedt|set forcedt]] and [[#set blockdt|set blockdt]] options

*Integral reaction rates are scored using the collision estimator of neutron flux, which has a few adjustable parameters (see [[#set cfe|set cfe]]).

Line 3,746:

Line 4,702:

{|

| <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 3,753:

Line 4,709:

'''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 3,768:

Line 4,724:

=== set ecut ===

−

'''set ecut''' ''EMINn'' ''EMINp''

+

'''set ecut''' ''EMINn'' [ ''EMINp'' ]

Sets minimum energy cut-off for neutrons and photons. Input values:

{|

−

| <tt>''EMINn''</tt>

+

| <tt>''EMINn''</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>''EMINp''</tt>

−

| : cut-off energy for photons (MeV)

+

| : cut-off energy for photons [in MeV] (default value: 1.0E-3)

|}

Notes:

−

*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.

*Versions 2.1.27 and earlier include only photon energy cut-off, which is now the second input parameter.

Line 3,792:

Line 4,747:

{|

| <tt>''DENSi''</tt>

−

| : mass density (g/cm3)

+

| : mass density [in g/cm3]

|-

| <tt>''EMINp,i''</tt>

−

| : cut-off energy for photons (MeV)

+

| : cut-off energy for photons [in MeV]

|}

Line 3,811:

Line 4,766:

|-

| <tt>''EMINp,i''</tt>

−

| : cut-off energy for photons (MeV)

+

| : cut-off energy for photons [in MeV]

|}

Line 3,817:

Line 4,772:

'''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>

−

| : 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>".

|}

Notes:

−

*Requires group constant generation to be set on.

+

*Requires group constant generation to be set on (see [[#set gcu|set gcu]]).

=== set edepdel ===

Line 3,835:

Line 4,790:

{|

| <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>

−

| : 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 3,844:

Line 4,799:

* 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.

−

* ~~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 ===

Line 3,855:

Line 4,810:

{|

| <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 3,870:

Line 4,825:

{|

| <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>

−

| : 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)

|}

−

~~Notes~~:~~~~

+

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>Efiss,i = (Qi/Q235) × H235</tt>''

+

|-

+

| <tt>1</tt>

+

| Local energy deposition based on ENDF MT 458 data

+

| ''<tt>Efiss,i = EFRi + ENPi + ENDi + EGPi + EGDi + EBi + Ecapt</tt>''

+

|-

+

| <tt>2</tt>

+

| Local photon energy deposition

+

| ''<tt>Efiss,i = EFRi + EGPi + EGDi + EBi</tt>''

+

|-

+

| <tt>3</tt>

+

| Coupled neutron-photon transport

+

| ''<tt>Efiss,i = EFRi + EBi</tt>''

+

|-

+

|}

+

Notes:

+

* 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.

−

* 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]].

=== set egrid ===

Line 3,894:

Line 4,874:

|-

| <tt>''EMIN''</tt>

−

| : minimum energy in the grid (MeV)

+

| : minimum energy in the grid [in MeV]

|-

| <tt>''EMAX''</tt>

−

| : maximum energy in the grid (MeV)

+

| : maximum energy in the grid [in MeV]

|}

Notes:

−

*~~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.

−

*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].

Line 3,914:

Line 4,894:

{|

| <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")

|}

Notes:

−

*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.

Line 3,931:

Line 4,910:

|-

| <tt>''CONDi''</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)

|}

Notes:

−

*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.

*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 3,949:

Line 4,927:

|-

| <tt>''GASi''</tt>

−

| : phase state (0 = condensed or non-gas, 1 = gas).

+

| : phase state (0 = condensed or non-gas, 1 = gas). (default value: 0)

|}

Notes:

−

*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.

Line 3,966:

Line 4,944:

|-

| <tt>''MEEi''</tt>

−

| : electrons mean excitation energy (MeV)

+

| : electrons mean excitation energy [in MeV] or "-1" (default value: -1)

|}

Notes:

−

*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.

Line 3,980:

Line 4,958:

{|

| <tt>''EGRID_E''</tt>

−

| : energy grid size (default value is 200)

+

| : energy grid size (default value: 200)

|}

Line 3,993:

Line 4,971:

{|

| <tt>''NX''</tt>

−

| : number of mesh cells in x-direction

+

| : number of mesh cells in x-direction (default value: 5)

|-

| <tt>''NY''</tt>

−

| : number of mesh cells in y-direction

+

| : number of mesh cells in y-direction (default value: 5)

|-

| <tt>''NZ''</tt>

−

| : number of mesh cells in z-direction

+

| : number of mesh cells in z-direction (default value: 5)

|-

| <tt>''XMIN''</tt>

−

| : minimum mesh boundary in x-direction

+

| : minimum mesh boundary in x-direction [in cm]

|-

| <tt>''XMAX''</tt>

−

| : maximum mesh boundary in x-direction

+

| : maximum mesh boundary in x-direction [in cm]

|-

| <tt>''YMIN''</tt>

−

| : minimum mesh boundary in y-direction

+

| : minimum mesh boundary in y-direction [in cm]

|-

| <tt>''YMAX''</tt>

−

| : maximum mesh boundary in y-direction

+

| : maximum mesh boundary in y-direction [in cm]

|-

| <tt>''ZMIN''</tt>

−

| : minimum mesh boundary in z-direction

+

| : minimum mesh boundary in z-direction [in cm]

|-

| <tt>''ZMAX''</tt>

−

| : maximum mesh boundary in z-direction

+

| : maximum mesh boundary in z-direction [in cm]

|}

Notes:

−

*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 default~~ mesh ~~size is 5 × 5 × 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]].

Line 4,052:

Line 5,029:

|-

| <tt>''En''</tt>

−

| : energy deposited per fission in MeV

+

| : energy deposited per fission [in MeV] (default value: 202.27)

|}

Line 4,058:

Line 5,035:

*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 Section 5.8 of [http://montecarlo.vtt.fi/download/Serpent_manual.pdf Serpent 1 User Manual].

Line 4,065:

Line 5,042:

'''set fissrate''' ''F'' [ ''MAT'' ]

−

+

Sets normalization to fission rate. Input values:

−

Sets normalization to fission rate.

+

{|

| <tt>''F''</tt>

−

| : number of fission reactions per second (1/s)

+

| : number of fission reactions per second [in 1/s]

|-

| <tt>''MAT''</tt>

Line 4,078:

Line 5,054:

Notes:

*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

*See also Section 5.8 of [http://montecarlo.vtt.fi/download/Serpent_manual.pdf Serpent 1 User Manual].

Line 4,085:

Line 5,062:

=== 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>".

|}

+

Notes:

+

*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''' ''F'' [ ''MAT'' ]

−

Sets normalization to total flux.

+

Sets normalization to total flux. Input values:

{|

Line 4,108:

Line 5,090:

Notes:

*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.

*See also Section 5.8 of [http://montecarlo.vtt.fi/download/Serpent_manual.pdf Serpent 1 User Manual].

Line 4,120:

Line 5,103:

{|

| <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 4,141:

Line 5,124:

{|

| <tt>''MATn''</tt>

−

| : material list ~~(type 1)~~

+

| : material list

|-

| <tt>''UNIn''</tt>

−

| : universe list ~~(type 2)~~

+

| : universe list

|-

| <tt>''LVL''</tt>

−

| : level number ~~(type 3)~~

+

| : level number

|-

| <tt>''XMIN''</tt>

−

| : minimum x-coordinate mesh boundary ~~(type 4)~~

+

| : minimum x-coordinate mesh boundary [in cm]

|-

| <tt>''XMAX''</tt>

−

| : maximum x-coordinate mesh boundary ~~(type 4)~~

+

| : maximum x-coordinate mesh boundary [in cm]

|-

| <tt>''NX''</tt>

−

| : number of x-mesh cells ~~(type 4)~~

+

| : number of x-mesh cells

|-

| <tt>''YMIN''</tt>

−

| : minimum y-coordinate mesh boundary ~~(type 4)~~

+

| : minimum y-coordinate mesh boundary [in cm]

|-

| <tt>''YMAX''</tt>

−

| : maximum y-coordinate mesh boundary ~~(type 4)~~

+

| : maximum y-coordinate mesh boundary [in cm]

|-

| <tt>''NY''</tt>

−

| : number of y-mesh cells ~~(type 4)~~

+

| : number of y-mesh cells

|-

| <tt>''ZMIN''</tt>

−

| : minimum z-coordinate mesh boundary ~~(type 4)~~

+

| : minimum z-coordinate mesh boundary [in cm]

|-

| <tt>''ZMAX''</tt>

−

| : maximum z-coordinate mesh boundary ~~(type 4)~~

+

| : maximum z-coordinate mesh boundary [in cm]

|-

| <tt>''NZ''</tt>

−

| : number of z-mesh cells ~~(type 4)~~

+

| : number of z-mesh cells

|}

Line 4,201:

Line 5,184:

'''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)

+

|}

+

Notes:

−

~~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 ===

Line 4,210:

Line 5,200:

'''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>

−

| : 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>

Line 4,222:

Line 5,212:

Notes:

−

*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.

Line 4,233:

Line 5,222:

'''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 4,243:

Line 5,232:

|-

| <tt>''MODE''</tt>

−

| : Critical spectrum calculation type (default 0, i.e. old B1 calculation)

+

| : Critical spectrum calculation type (default value: 0, i.e. old B1 calculation)

|-

| <tt>''DC''</tt>

Line 4,249:

Line 5,238:

|-

| <tt>''LIM''</tt>

−

| : Convergence criterion of keff in fundamental mode calculation calculated as the absolute value difference of keff between successive iterations (default value 1E-7)

+

| : Convergence criterion of keff in fundamental mode calculation calculated as the absolute value difference of keff between successive iterations (default value: 1E-7)

|-

| <tt>''TGT''</tt>

−

| : Target value for fundamental mode keff (default value 1.0, not used with the old B1 calculation mode)

+

| : Target value for fundamental mode keff (default value: 1.0, not used with the old B1 calculation mode)

|-

| <tt>''ITER''</tt>

−

| : Maximum number of fundamental mode calculation iterations (default value 25, not used with the old B1 calculation mode)

+

| : Maximum number of fundamental mode calculation iterations (default value: 25, not used with the old B1 calculation mode)

|-

| <tt>''INIT''</tt>

−

| : First guess for absolute value of critical B2 (default value 1E-6, not used with the old B1 calculation mode)

+

| : First guess for absolute value of critical B2 (default value: 1E-6, not used with the old B1 calculation mode)

|}

The possible values for mode are:

−

{| class="wikitable" style="text-align: left;"

+

::{| class="wikitable" style="text-align: left;"

! Mode

! Description

Line 4,282:

Line 5,271:

The possible values for FM mode multi-group diffusion coefficients are:

−

{| class="wikitable" style="text-align: left;"

+

::{| class="wikitable" style="text-align: left;"

! Mode

! Description

Line 4,325:

Line 5,314:

Notes:

−

*Photon buffer refers to pre-allocated memory block used to store photon particle data. This memory is needed for putting secondary photons in que, etc..

+

*Photon buffer refers to pre-allocated memory block used to store photon particle data.

−

*The buffer factor defines the buffer size relative to simulated batch size.

+

**This memory is needed for putting secondary photons in que, etc..

−

*The event bank refers to pre-allocated memory block used to store history data on particle events. This bank is used only with certain special options, such as [[importance detectors]] and [[track plotter]].

+

*The buffer factor <tt>''FAC''</tt> defines the buffer size relative to simulated batch size.

+

*The event bank <tt>''BNK''</tt> refers to pre-allocated memory block used to store history data on particle events.

+

**This bank is used only with certain special options, such as [[importance detectors]] and [[track plotter]].

*The default values depend on simulation mode, and there is no need to adjust the values unless the calculation terminates with an error.

*''Note to developers: event bank is now the same for both neutrons and photons.''

Line 4,334:

Line 5,325:

'''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>

−

| : 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>".

|}

Notes:

*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''

Line 4,353:

Line 5,344:

{|

| <tt>''UNIn''</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)

|}

Notes:

−

*~~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 "<tt>off</tt>" when the results are not needed (this may speed up the calculation).

*See separate description of [[Output parameters#Homogenized group constants|output parameters]].

*Super-imposed group constant generation universes (i.e. not part of the geometry definition) should not be used so that they cover (partly or totally) the same geometry region as some other group constant generation universe (super-imposed or part of the geometry definition), if group constants are generated only on a single geometry level.

+

*The spatial homogenization methodology is described in a related paper<ref name="homogenization">Leppänen, J., Pusa, M. and Fridman, E., ''"Overview of methodology for spatial homogenization in the Serpent 2 Monte Carlo code."'', Ann. Nucl. Energy, [https://doi.org/10.1016/j.anucene.2016.06.007 '''96''' (2016) 126-136]</ref>.

=== set gcut ===

Line 4,381:

Line 5,373:

'''set genrate''' ''G'' [ ''MAT'' ]

−

Sets normalization to fission neutron generation rate.

+

Sets normalization to fission neutron generation rate. Input values:

{|

| <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>

−

| : material in which the fission neutrons are generated

+

| : material in which the fission neutrons are generated. The default is all materials

|}

Line 4,395:

Line 5,387:

*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.

−

*~~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.

*See also Section 5.8 of [http://montecarlo.vtt.fi/download/Serpent_manual.pdf Serpent 1 User Manual].

+

=== set gpop ===

+

'''set gpop''' ''MAX_HIS'' ''MAX_POP'' ''C''

+

Sets the on-the-fly neutron growing population size algorithm. Input values:

+

{|

+

| <tt>''MAX_HIS''</tt>

+

| : maximum number of histories

+

|-

+

| <tt>''MAX_POP''</tt>

+

| : maximum population size

+

|-

+

| <tt>''C''</tt>

+

| : parameter to control the population growth

+

|}

+

Notes:

+

*The on-the-fly neutron growing population size algorithm is used in criticality calculations to accelerate the fission source convergence.

+

*The growing algorithm methodology is described in related paper. <ref name="gpop">Dufek, J. and Tuttelberg, K.

+

''"Monte Carlo criticality calculations accelerated by a growing neutron population."'' Ann. Nucl. Energy [https://doi.org/10.1016/j.anucene.2016.02.015 '''94''' (2016) 16-21].</ref>

=== set gsw ===

Line 4,410:

Line 5,424:

|-

| <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 4,416:

Line 5,430:

*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.

+

*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 ===

Line 4,426:

Line 5,440:

{|

| <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>".

|}

Notes:

*When invoked, Serpent collects batch-wise data on keff, fission source entropy etc. and produces a [[Description of output files#History output|separate output file]].

−

*Setting the history option also invokes the calculation of fission source (Shannon) entropy. The entropy mesh parameters can be adjusted using [[#set entr|set entr]].

+

*Setting the history option also invokes the calculation of fission source (Shannon) entropy.

+

**The entropy mesh parameters can be adjusted using [[#set entr|set entr]].

=== set ifp ===

Line 4,440:

Line 5,455:

{|

| <tt>''GEN''</tt>

−

| : Number of generations~~. Default~~ value 15.

+

| : Number of generations (default value: 15).

|}

Line 4,453:

Line 5,468:

|-

| <tt>''G''</tt>

−

| : exponential for adjusting importances

+

| : exponential for adjusting importances (default value: -0.5)

|}

Line 4,477:

Line 5,492:

=== 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 4,497:

Line 5,512:

Notes:

−

*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.

−

*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.

=== set inftrk ===

−

~~'''set''' '''inftrk''' ''LOOPn'' [ ''ERRn'' ''LOOPp'' ''ERRp'' ]~~

−

Sets parameters for terminating infinite tracking loops.

+

'''set''' '''inftrk''' ''LOOPn'' [ ''ERRn'' ''LOOPp'' ''ERRp'' ]

+

Sets parameters for terminating infinite tracking loops. Input values:

{|

| <tt>''LOOPn''</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>''ERRn''</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>''LOOPp''</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>''ERRp''</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 4,524:

Line 5,539:

Notes:

−

*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.

*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 4,530:

Line 5,545:

=== set inventory ===

−

'''set inventory''' ''~~MAT~~1'' ''~~MAT~~2'' ...

+

'''set inventory''' ''ID1'' ''ID2'' ...

+

Defines the nuclides or elements to include in the depletion output file <tt>[input]_dep.m</tt>. Input values:

+

{|

+

| <tt>''IDn''</tt>

+

| : Identifier for nuclide, or element or special entry.

+

|}

+

Notes:

+

*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 (from Serpent 1)

+

|-

+

| <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''

−

~~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>''MATn''</tt>~~

−

~~| : Nuclide or element to include~~

−

|-

| <tt>''N''</tt>

| : Number of nuclides

Line 4,548:

Line 5,614:

Notes:

−

*~~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 ===

Line 4,596:

Line 5,664:

Notes:

*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 ===

Line 4,609:

Line 5,677:

|-

| <tt>''KEFF''</tt>

−

| : target k-eff for the iteration (default value: 1.0)

+

| : target keff for the iteration (default value: 1.0)

|-

| <tt>''FX''</tt>

Line 4,627:

Line 5,695:

−

*~~See~~ [[Critical density iteration]].

+

*For more information, see the detailed description on the [[Critical density iteration]].

=== set keff ===

'''set keff''' ''KEFF''

−

Option to scale fission neutron production in external source simulations.

+

Option to scale fission neutron production in external source simulations. Input values:

{|

| <tt>''KEFF''</tt>

−

| : ~~The k-effective to use for scaling the fission~~ neutron production~~. Inverse of <tt>''KEFF''</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. Default value is 1.~~

+

| : Fission neutron production scaling factor (default value: 1.0)

|}

Notes:

−

*~~Affects~~ both prompt and delayed neutron production from fissions.

+

*The k-effective to use for scaling the fission neutron production. The inverse of <tt>''KEFF''</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.

−

*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.

+

*The option affects both prompt and delayed neutron production from fissions.

+

*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 ===

Line 4,647:

Line 5,716:

'''set lossrate''' ''L'' [ ''MAT'' ]

−

Sets normalization to total loss rate.

+

Sets normalization to total loss rate. Input values:

{|

| <tt>''L''</tt>

−

| : number of lost neutrons per second (neutrons/s)

+

| : number of lost neutrons per second [in neutrons/s]

|-

| <tt>''MAT''</tt>

Line 4,660:

Line 5,729:

*Normalization is needed to relate the Monte Carlo reaction rate estimates to a user-given parameter.

*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.

*See also Section 5.8 of [http://montecarlo.vtt.fi/download/Serpent_manual.pdf Serpent 1 User Manual].

Line 4,672:

Line 5,742:

{|

| <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 4,688:

Line 5,758:

{|

| <tt>''MAX''</tt>

−

| : maximum number of splits

+

| : maximum number of splits (default value: 1.0E4)

|-

| <tt>''MIN''</tt>

−

| : minimum survival probability in Russian roulette

+

| : minimum survival probability in Russian roulette (default value: 1.0E-18)

|}

Line 4,705:

Line 5,775:

{|

| <tt>''N''</tt>

−

| : batch size in double floats (8 bytes)

+

| : batch size in double floats, i.e. 8 bytes (default value: 1E4)

|}

Notes:

+

*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)

+

|}

+

Notes:

+

*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 proc. ANS M&C 2021. Raleigh, NC, USA. October 3-7, 2021</ref>.

=== 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:

Line 4,721:

Line 5,809:

| <tt>''NP''</tt>

| : 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 4,726:

Line 5,817:

*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 volumes.

+

**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 ===

Line 4,739:

Line 5,833:

{|

| <tt>''UNI''</tt>

−

| : universe where ~~spatial homogenization is performed~~

+

| : universe where universe where homogenized microscopic cross sections are generated

|-

| <tt>''VOL''</tt>

−

| : volume of the ~~homogenized zone~~

+

| : volume of the universe [in cm3] (3D geometry) or cross-sectional area [in cm2] (2D geometry)

|-

| <tt>''N''</tt>

Line 4,759:

Line 5,853:

Notes:

−

*When this option is set, Serpent calculates few-group microscopic cross sections for the listed nuclides and reactions.

+

*When this option is set, Serpent calculates homogenized few-group microscopic cross sections for the listed nuclides and reactions.

*The cross sections are always calculated in the actual spectrum of the problem, never with the [[#set fum|critical spectrum]].

*The results are printed in a [[Description of output files#Micro depletion output|separate output file]].

*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 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 calculation requires the [[Defining material volumes|material volumes to be correctly set]].

*Transmutation reactions to ground and isomeric states can be calculated by adding 'g' and 'm' after the reaction MT (e.g. 102m is the capture cross section corresponding to daughter nuclide being in isomeric state). Reaction rates are calculated to all states by default.

*Fission reactions corresponding to a specific yield in ENDF can be calculated by adding 1, 2, 3, 4, ... after the reaction MT (e.g. 181 is total fission cross section corresponding to first fission product yield) depending on the fission yield data of the nuclide in the data library. This parameter was different before 2.1.32.

Line 4,771:

Line 5,867:

*The nuclide identifiers are entered as ZAI, not ZA. For example, the ZAI for U-235 is 922350 and the ZAI for Am-242m is 952421.

*Parameter ''VOL'' was ''VR'' in versions before 2.1.32 (volume ratio of materials included in micro depletion to total homogenized region).

−

*Multiple set mdep cards can be given for a single ~~homogenization~~ universe.

+

*Multiple set mdep cards can be given for a single homogenized universe.

+

*The microscopic cross section calculation methodology is described in a related article.<ref name="mdep">Rintala, A., Valtavirta, V. and Leppänen, J. ''"Microscopic cross section calculation methodology in the Serpent 2 Monte Carlo code."'' Ann. Nucl. Energy [https://doi.org/10.1016/j.anucene.2021.108603 '''164''' (2021) 108603]</ref>

=== set memfrac ===

Line 4,777:

Line 5,874:

'''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>

−

| : 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)

|}

Notes:

−

* ~~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 ===

Line 4,796:

Line 5,893:

{|

| <tt>''MFPMINp''</tt>

−

| : cut-off mean-free-path for photons (cm)

+

| : cut-off mean-free-path for photons [in cm]

|}

Line 4,806:

Line 5,903:

{|

| <tt>''DENSi''</tt>

−

| : mass density (g/cm3)

+

| : mass density [g/cm3]

|-

| <tt>''MFPMINp,i''</tt>

−

| : cut-off mean-free-path for photons (cm)

+

| : cut-off mean-free-path for photons [in cm]

|}

Line 4,825:

Line 5,922:

|-

| <tt>''MFPMINp,i''</tt>

−

| : cut-off mean-free-path for photons (cm)

+

| : cut-off mean-free-path for photons [in cm]

|}

Line 4,831:

Line 5,928:

'''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>

−

| : 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>

−

| : When set to 2, results are averaged over all criticality cycles

+

| : When set to "2", results are averaged over all criticality cy

@@ Line 18: / Line 18: @@
 === 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>
 | : 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>
 | : name of the replaced material
@@ Line 39: / Line 57: @@
 | <tt>''MAT<sub>2</sub>''</tt>
 | : 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>
 | : name of the replaced universe
@@ Line 45: / Line 77: @@
 | <tt>''UNI<sub>2</sub>''</tt>
 | : 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>
 | : name of the material for which density and temperature are adjusted
 |-
 | <tt>''DENS''</tt>
-| : material density after adjustment (positive entries for atomic, negative entries for mass densities, or "sum" to use the sum of the constituent nuclide 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>
-| : 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>
-| : ''n''th thermal scattering data associated with the material
+| : ''n''-th thermal scattering data associated with the material
 |-
 | <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>
-| : 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>
 | : target universe, surface or cell
@@ Line 69: / Line 123: @@
 | <tt>''TRANS''</tt>
 | : 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>
-| : 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>
 | : 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>
 | : file path of the replaced transport correction curve data
@@ Line 81: / Line 184: @@
 | <tt>''FILE<sub>2</sub>''</tt>
 | : 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>
 | : variable name
@@ Line 90: / Line 201: @@
 <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 card by a mixture of boron and water defined with a mix card containing the same pure water material).
-*Currently (version 2.1.30) 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.
 === casematrix (casematrix definition)<span id="casematrix"></span> ===
@@ Line 121: / Line 222: @@
              ''NBR<sub>2</sub>'' [ ''BR<sub>2,1</sub>'' ''BR<sub>2,2</sub>'' ... ''BR<sub>1,NBR<sub>2</sub></sub>'' ]
              ...
 Defines the casematrix for the automated burnup sequence. Input values:
@@ Line 132: / Line 233: @@
 |-
 | <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>
@@ Line 138: / Line 239: @@
 |-
 | <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>
-| : 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>
-| : name of the ''i''th branch in the ''m''th dimension
+| : name of the ''i''-th branch in the ''m''-th dimension
 |}
@@ Line 150: / 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 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.
 *Multiple casematrix cards can be given in a single input file.
+*For more information, see detailed description on [[automated burnup sequence]].
 === cell (cell definition)<span id="cell"></span> ===
@@ Line 212: / 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>.
 *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.
 *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 233: / Line 334: @@
 |-
 | <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>
-| : 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>
-| : 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>
 *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 more information, see detailed description on [[automated burnup sequence]].
@@ Line 265: / Line 366: @@
 |-
 | <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>
-| : mesh lower x boundary
+| : mesh lower x-boundary [in cm]
 |-
 | <tt>''X<sub>MAX</sub>''</tt>
-| : mesh higher x boundary
+| : mesh higher x-boundary [in cm]
 |-
 | <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>
-| : mesh lower y boundary
+| : mesh lower y-boundary [in cm]
 |-
 | <tt>''Y<sub>MAX</sub>''</tt>
-| : mesh higher y boundary
+| : mesh higher y-boundary [in cm]
 |-
 | <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>
-| : mesh lower z boundary
+| : mesh lower z-boundary [in cm]
 |-
 | <tt>''Z<sub>MAX</sub>''</tt>
-| : mesh higher z boundary
+| : mesh higher z-boundary [in cm]
 |}
@@ Line 307: / Line 408: @@
 |-
 | <tt>''R<sub>MIN</sub>''</tt>
-| : mesh inner radial boundary
+| : mesh inner radial boundary [in cm]
 |-
 | <tt>''R<sub>MAX</sub>''</tt>
-| : mesh outer radial boundary
+| : mesh outer radial boundary [in cm]
 |-
 | <tt>''N<sub>PHI</sub>''</tt>
@@ Line 329: / Line 430: @@
 |-
 | <tt>''X<sub>0</sub>''</tt>
-| : mesh horizontal origin x-coordinate
+| : mesh horizontal origin x-coordinate [in cm]
 |-
 | <tt>''Y<sub>0</sub>''</tt>
-| : mesh horizontal origin y-coordinate
+| : mesh horizontal origin y-coordinate [in cm]
 |-
 | <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>
-| : mesh lower z boundary
+| : mesh lower z-boundary [in cm]
 |-
 | <tt>''Z<sub>MAX</sub>''</tt>
-| : mesh higher z boundary
+| : mesh higher z-boundary [in cm]
 |-
 | <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>
-| : number of cells in the y direction
+| : number of cells in the y-direction
 |-
 | <tt>''N<sub>Z</sub>''</tt>
-| : number of cells in the z direction
+| : number of cells in the z-direction
 |}
@@ Line 366: / Line 467: @@
 |-
 | <tt>''X<sub>0</sub>''</tt>
-| : mesh horizontal origin x-coordinate
+| : mesh horizontal origin x-coordinate [in cm]
 |-
 | <tt>''Y<sub>0</sub>''</tt>
-| : mesh horizontal origin y-coordinate
+| : mesh horizontal origin y-coordinate [in cm]
 |-
 | <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>
-| : mesh lower z boundary
+| : mesh lower z-boundary [in cm]
 |-
 | <tt>''Z<sub>MAX</sub>''</tt>
-| : mesh higher z boundary
+| : mesh higher z-boundary [in cm]
 |-
 | <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>
-| : number of cells in the y direction
+| : number of cells in the y-direction
 |-
 | <tt>''N<sub>Z</sub>''</tt>
-| : number of cells in the z direction
+| : number of cells in the z-direction
 |}
@@ Line 402: / Line 503: @@
 |-
 | <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>
-| : number of cells in the y direction
+| : number of cells in the y-direction
 |-
 | <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>''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>''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>''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 438: / Line 539: @@
 |-
 | <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.
@@ Line 461: / Line 562: @@
 <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.
+*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> ===
@@ Line 477: / Line 579: @@
 The possible step types are:
-{| class="wikitable" style="text-align: left;"
+::{| class="wikitable" style="text-align: left;"
 ! Type
 ! Description
@@ Line 560: / Line 662: @@
            [ [[#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_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>'' ]
+           [ [[#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_dumsh|'''dumsh''']] ''UNI'' ''N<sub>C</sub>'' ''CELL<sub>0</sub>'' ''BIN<sub>0</sub>'' ''CELL<sub>1</sub>'' ''BIN<sub>1</sub>'' ... ]
@@ Line 576: / Line 679: @@
            [ [[#det_dphb|'''dphb''']] ''PHB'' ]
            [ [[#det_dmesh|'''dmesh''']] ''MESH'' ]
-Detector definition. The first parameter:
+Detector definition. The two first parameters:
 {|
+| <tt>''NAME''</tt>
+| : detector name
+|-
 | <tt>''PART''</tt>
 | : 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>
@@ Line 599: / Line 713: @@
 *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]].
-*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.
@@ Line 610: / Line 732: @@
 {|
 | <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>
-*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 662: / 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>
-| : minimum x-coordinate of the detector mesh
+| : minimum x-coordinate of the detector mesh [in cm]
 |-
 | <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>
@@ Line 675: / Line 798: @@
 |-
 | <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>
-| : maximum y-coordinate of the detector mesh
+| : maximum y-coordinate of the detector mesh [in cm]
 |-
 | <tt>''N<sub>Y</sub>''</tt>
@@ Line 684: / Line 807: @@
 |-
 | <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>
-| : maximum z-coordinate of the detector mesh
+| : maximum z-coordinate of the detector mesh [in cm]
 |-
 | <tt>''N<sub>Z</sub>''</tt>
@@ Line 694: / Line 817: @@
 <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 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>
-| : Type of curvilinear mesh - 1 = cylindrical (dimensions ''r'', ''&theta;'', ''z''), 2 = spherical (dimensions ''r'', ''&theta;'', ''&phi;'')
+| : type of curvilinear mesh - 1 = cylindrical (dimensions ''r'', ''&theta;'', ''z''), 2 = spherical (dimensions ''r'', ''&theta;'', ''&phi;'')
 |-
 | <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>
-| : 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>
-| : Number of bins in the ''n'' coordinate direction (the radial division will be equal ''r'', not equal volume).
+| : number of bins in the ''n''-coordinate direction (the radial division will be equal ''r'', not equal volume).
 |}
 <u>Notes:</u>
 *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.
-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>
+| : 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)
+| : 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>
-| : coordinates of mesh center
+| : coordinates of mesh center [in cm]
 |-
 | <tt>''PITCH''</tt>
-| : mesh pitch
+| : mesh pitch [in cm]
 |-
 | <tt>''N<sub>1</sub>''</tt>, <tt>''N<sub>1</sub>''</tt>
@@ Line 735: / Line 876: @@
 |-
 | <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>
-| : maximum z-coordinate of the detector mesh
+| : maximum z-coordinate of the detector mesh [in cm]
 |-
 | <tt>''N<sub>Z</sub>''</tt>
@@ Line 748: / 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 763: / Line 904: @@
 <u>Notes:</u>
 *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 775: / Line 917: @@
 <u>Notes:</u>
 *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]].
+**[[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 791: / 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 803: / Line 948: @@
 <u>Notes:</u>
 *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.
+**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).
@@ Line 813: / Line 961: @@
 {|
-| <tt>''COS<sub>Y</sub>''</tt>
+| <tt>''COS<sub>X</sub>''</tt>
 | : component of the direction vector parallel to x-axis
 |-
@@ Line 827: / 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 836: / Line 984: @@
 <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.)
+*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 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 847: / Line 997: @@
 |-
 | <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>
 *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.
-*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 867: / Line 1,017: @@
 |}
-The types are:
+The possible special types are:
-{|
+::{| class="wikitable" style="text-align: left;"
+! Type
+! Description
+! Notes
+|-
 | -1
-| = cumulative spectrum
+| cumulative spectrum
+| use with energy binning ('''de''')
 |-
 | -2
-| = division by energy width
+| division by energy width
+| use with energy binning ('''de''')
 |-
 | -3
-| = division by lethargy width
+| division by lethargy width
+| use with energy binning ('''de''')
 |-
 | -4
-| = sum over cell or material bins
+| sum over cell or material bins
+| use with cell and/or material binning ('''dc''', '''dm''')
 |-
 | -5
-| = importance weighting
+| importance weighting
+| -
 |-
 | -6
-| = sum over number of scores
+| sum over number of scores
+| -
 |-
 | 2
-| = multiply result with another detector defined by <tt>''PARAM''</tt>
+| multiply result with another detector defined by <tt>''PARAM''</tt>
+| bin-compatibility
 |-
 | 3
-| = divide result with another detector defined by <tt>''PARAM''</tt>
+| divide result with another detector defined by <tt>''PARAM''</tt>
+| bin-compatibility
+|-
+| 4
+| multiply response function by (local) temperature
+| -
+|-
 |}
 <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>
-| : 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
 |}
 <u>Notes:</u>
-*When this option is set, the batch-wise results are kept and printed in the [[history output file]].
+*When this option is set, the batch-wise results are printed in the history output file, <tt>[input]_stats.m</tt>.
 *''Note to developers: statistical tests should be documented''
-Detector flagging (<tt>'''dfl'''</tt>):<span id="det_dfl"></span>
+Detector score flagging (<tt>'''dfl'''</tt>):<span id="det_dfl"></span>
 {|
@@ Line 921: / Line 1,086: @@
 | <tt>''OPT''</tt>
 | : 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>
 *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 logic:  detector 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 935: / Line 1,127: @@
 |-
 | <tt>''FLX''</tt>
-| : flux applied to activation
+| : flux applied to activation [in 1/cm<sup>2</sup>s]
 |}
 <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.
+*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.
-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 954: / Line 1,151: @@
   |}
-{| class="wikitable" style="text-align: left;"
+::{| class="wikitable" style="text-align: left;"
   ! Geometry
   ! <tt>PARAMS</tt>
@@ Line 978: / Line 1,175: @@
 <u>Notes:</u>
-*<tt>-1</tt> 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
 *Specifics of this implementation:
@@ Line 987: / Line 1,184: @@
 ***The generated FET coefficients <math>a_n</math> already have all contributions from the orthonormalization constant pre-included, i.e. <math>c_n</math> from <math>\frac{1}{c_n} = \lVert \psi_n \rVert^2 = \int_\Gamma \psi_n^2 \omega_n </math>
 ***Thus, an FET can be simply reconstructed/sampled from the standard functional series as: <math>F(\xi) = \sum a_n \psi_n(\xi) \omega_n(\xi)</math>
+*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.
@@ Line 1,012: / Line 1,211: @@
 === 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>'' ]
+          [ [[#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>'' ]
+          [ [[#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>'' ]
+         [ [[#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>0</sub>'' ''R<sub>1</sub>'' ... ''R<sub>N</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>'' ]
+         [ [[#div_subz1|'''subz''']] ''N<sub>Z</sub>'' ''Z<sub>MIN</sub>'' ''Z<sub>MAX</sub>'' ]
-Divides a material into a number of sub-zones. Input values:
+          [ [[#div_subz2|'''subz''']] ''N<sub>Z</sub>'' ''Z<sub>1</sub>'' ''Z<sub>2</sub>'' ... ''Z<sub>N+1</sub>'' ]
+         [ [[#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>
 | : 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>
-| : 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>
-| : number of x-zones
+| : number of x-zones (positive value)
 |-
 | <tt>''X<sub>MIN</sub>''</tt>
-| : minimum x-coordinate (cm)
+| : minimum x-coordinate [in cm]
 |-
 | <tt>''X<sub>MAX</sub>''</tt>
-| : maximum x-coordinate (cm)
+| : maximum x-coordinate [in cm]
 |-
 | <tt>''N<sub>Y</sub>''</tt>
-| : number of y-zones
+| : number of y-zones (positive value)
 |-
 | <tt>''Y<sub>MIN</sub>''</tt>
-| : minimum y-coordinate (cm)
+| : minimum y-coordinate [in cm]
 |-
 | <tt>''Y<sub>MAX</sub>''</tt>
-| : maximum y-coordinate (cm)
+| : maximum y-coordinate [in cm]
 |-
 | <tt>''N<sub>Z</sub>''</tt>
-| : number of z-zones
+| : number of z-zones (positive value)
 |-
 | <tt>''Z<sub>MIN</sub>''</tt>
-| : minimum z-coordinate (cm)
+| : minimum z-coordinate [in cm]
 |-
 | <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>
+| : 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>
-| : number of radial zones
+| : number of radial-zones (positive value)
 |-
 | <tt>''R<sub>MIN</sub>''</tt>
-| : minimum radial coordinate (cm)
+| : minimum radial-coordinate [in cm]
 |-
 | <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>'''subr'''</tt>):<span id="div_subr2"></span>
+{|
+| <tt>''N<sub>R</sub>''</tt>
+| : number of radial-zones (negative value)
 |-
 | <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>'''subs'''</tt>):<span id="div_subs1"></span>
+{|
 | <tt>''N<sub>S</sub>''</tt>
-| : number of angular sectors
+| : number of angular-zones (negative value)
 |-
 | <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.
-*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 cylindrical sector mesh, <u>user-defined volume</u> (<tt>'''subs'''</tt>):<span id="div_subs2"></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>''N<sub>S</sub>''</tt>
-*The usage of <tt>''LVL''</tt> is explained on page [[automated depletion zone division]].
+| : number of angular-zones
-*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.
+|-
+| <tt>''S<sub>n</sub>''</tt>
+| : angular-sector boundaries [in degrees]
+|-
+|}
+<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.
+Sub-division pebble-bed structure (<tt>'''peb'''</tt>): <span id="div_peb"></span>
+{|
+| <tt>''PBED''</tt>
+| : stochastic particle / pebble-bed structure
+|-
+| <tt>''N<sub>UNI</sub>''</tt>
+| : 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> ===
-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> ===
@@ Line 1,104: / Line 1,449: @@
 |-
 |<tt>''E<sub>i</sub>''</tt>
-|: bin boundaries (type 1)
+|: bin boundaries [in MeV]
 |-
 |<tt>''N''</tt>
-|: number of equi-width bins (types 2 and 3)
+|: number of equi-width bins
 |-
 |<tt>''E<sub>min</sub>''</tt>
-|: minimum energy (types 2 and 3)
+|: minimum energy [in MeV]
 |-
 |<tt>''E<sub>max</sub>''</tt>
-|: maximum energy (types 2 and 3)
+|: maximum energy [in MeV]
 |-
 |<tt>''GRID''</tt>
-|: name of the pre-defined grid (type 4)
+|: name of the pre-defined grid
 |}
 <u>Notes:</u>
-*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 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]].
-*See separate description of [[pre-defined energy group structures]].
 === 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> ===
@@ Line 1,152: / Line 1,500: @@
 |: is the interpolation type (1 = histogram, 2 = lin-lin, 3 = lin-log, 4 = log-lin, 5 = log-log)
 |-
-|<tt>''X<sub>1</sub>'' ''F<sub>1</sub>'' ...</tt>
+|<tt>''X<sub>i</sub>'', ''F<sub>i</sub>''</tt>
 |: are the tabulated variable-value pairs
 |}
 <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.
 === hisv (history variation matrix definition)<span id="hisv"></span> ===
@@ Line 1,168: / Line 1,517: @@
 {|
 | <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>
@@ Line 1,180: / Line 1,529: @@
 *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.
-*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.
 === 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 1,195: / 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>
-* 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> ===
@@ Line 1,231: / Line 1,600: @@
 *The input parser starts reading and processing the new file from the point where the input card is placed. Processing of the original file continues after the new file is completed.
 *The included file must contain complete input cards and options, it cannot be used to read the values of another card.
+*Nested included file paths must refer to the original base input file or current working directory.
 === lat (regular lattice definition)<span id="lat"></span> ===
@@ Line 1,248: / Line 1,618: @@
 |-
 | <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>
-| : 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>
@@ Line 1,260: / Line 1,630: @@
 |-
 | <tt>''PITCH''</tt>
-| : lattice pitch
+| : lattice pitch [in cm]
 |-
 | <tt>''UNI<sub>n</sub>''</tt>
@@ Line 1,267: / Line 1,637: @@
 Possible lattice types are:
-{| class="wikitable" style="text-align: left;"
+::{| class="wikitable" style="text-align: left;"
 ! Type
 ! Description
@@ Line 1,306: / Line 1,676: @@
 |-
 | <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>
-| : y-coordinate of the lattice origin
+| : y-coordinate of the lattice origin [in cm]
 |-
 | <tt>''PITCH''</tt>
-| : lattice pitch
+| : lattice pitch [in cm]
 |-
 | <tt>''UNI<sub>1</sub>''</tt>
@@ Line 1,319: / Line 1,689: @@
 Possible lattice types are:
-{| class="wikitable" style="text-align: left;"
+::{| class="wikitable" style="text-align: left;"
 ! Type
 ! Description
@@ Line 1,348: / Line 1,718: @@
 |-
 | <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>
-| : y-coordinate of the lattice origin
+| : y-coordinate of the lattice origin [in cm]
 |-
 | <tt>''N<sub>R</sub>''</tt>
@@ Line 1,357: / Line 1,727: @@
 |-
 | <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>
-| : central radius of Rth ring
+| : central radius of ''R''-th ring [in cm]
 |-
 | <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>
-| : 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:
-{| class="wikitable" style="text-align: left;"
+::{| class="wikitable" style="text-align: left;"
 ! Type
 ! Description
@@ Line 1,393: / Line 1,763: @@
 |-
 | <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>
-| : y-coordinate of the lattice origin
+| : y-coordinate of the lattice origin [in cm]
 |-
 | <tt>''N<sub>L</sub>''</tt>
@@ Line 1,402: / Line 1,772: @@
 |-
 | <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>
-| : universe name filling the nth lattice position
+| : universe name filling the ''n''-th lattice position
 |}
 Possible lattice type is:
-{| class="wikitable" style="text-align: left;"
+::{| class="wikitable" style="text-align: left;"
 ! Type
 ! Description
@@ Line 1,435: / Line 1,805: @@
 |-
 | <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>
-| : y-coordinate of the lattice origin
+| : y-coordinate of the lattice origin [in cm]
 |-
 | <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>
@@ Line 1,453: / Line 1,823: @@
 |-
 | <tt>''PITCH<sub>X</sub>''</tt>
-| : lattice pitch in x-direction
+| : lattice pitch in x-direction [in cm]
 |-
 | <tt>''PITCH<sub>Y</sub>''</tt>
-| : lattice pitch in y-direction
+| : lattice pitch in y-direction [in cm]
 |-
 | <tt>''PITCH<sub>Z</sub>''</tt>
-| : lattice pitch in z-direction
+| : lattice pitch in z-direction [in cm]
 |-
 | <tt>''UNI<sub>n</sub>''</tt>
@@ Line 1,466: / Line 1,836: @@
 Possible lattice types are:
-{| class="wikitable" style="text-align: left;"
+::{| class="wikitable" style="text-align: left;"
 ! Type
 ! Description
@@ Line 1,487: / Line 1,857: @@
 === 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> ===
@@ Line 1,506: / Line 1,880: @@
   [    ''...''     ]
-<u>Mandatory information:</u>
+Material definition. The mandatory parameters are:
 {|
 | <tt>''NAME''</tt>
-| : Name of the material
+| : name of the material
 |-
 | <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>
-| : 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>
-| : 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>
-| : 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_tms"></span>
 {|
 | <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>
-| : Lower limit for material temperature
+| : lower limit for material temperature [in K]
 |-
 | <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>
-| : 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>
-| : Value for the green channel of geometry plots (between 0 and 255)
+| : value for the green channel (between 0 and 255)
 |-
 | <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>
-| : 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>
-| : 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>
-| : 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>
-| : 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>
-| : 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>
-| : 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>
-| : 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>
-*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> ===
@@ Line 1,623: / Line 2,048: @@
 |-
 | <tt>''XPIX''</tt>
-| : horizontal image size in pixels
+| : horizontal image size [in pixels]
 |-
 | <tt>''YPIX''</tt>
-| : vertical image size in pixels
+| : vertical image size [in pixels]
 |-
 | <tt>''SYM''</tt>
@@ Line 1,632: / Line 2,057: @@
 |-
 | <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>
@@ Line 1,650: / Line 2,075: @@
 *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.
-*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.
 *Mesh plot produced by the nth mesh-card is written in file <tt>[input]_mesh[n].png</tt>.
@@ Line 1,668: / Line 2,093: @@
 |-
 | <tt>''NUC<sub>n</sub>''</tt>
-| :  identifier of ''n''th nuclide in composition
+| :  identifier of <tt>''n''</tt>-th nuclide in composition
 |-
 | <tt>''&lambda;''<sub>n</sub></tt>
-| : reprocessing constant of ''n''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>
-* 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).
 === 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>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 1,695: / Line 2,119: @@
 |-
 | <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.
+*The mixtures are automatically decomposed into standard materials before running the transport simulation.
+**Alternatively, the decomposed material compositions can be written into file using the [[Installing and running Serpent#Running Serpent|<tt>''-mix''</tt>]] command line option.
+*Inherited properties/cards:
+**Nuclide specific thermal scattering data (see '''moder''' entry in the [[#mat_moder|mat]] card) is automatically brought from component materials to the mixture.
+**Other input option 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.
+**If they are to be applied to the mixture, they should be directly defined using the mixture material name (opposed to component material names) .
+*Burnable mixtures are not supported.
+<u>Optional entries:</u>
+Mixture RGB-color  (<tt>'''rgb'''</tt>): <span id="mix_rgb"></span>
 {|
 | <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>
-| : Value for the green channel of geometry plots (between 0 and 255)
+| : value for the green channel (between 0 and 255)
 |-
 | <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>
-| : 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>
-| : 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> ===
@@ Line 1,763: / Line 2,198: @@
 |-
 | <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>
 | : nested surface type (different surfaces for each region)
 |-
-| <tt>''PARAM<sub>nm</sub> ... </tt>
+| <tt>''PARAM<sub>nm</sub>'' ... </tt>
 | : surface parameters
 |}
@@ Line 1,774: / Line 2,209: @@
 <u>Notes:</u>
-*The nest card defines an entire universe consisting of nested material regions. The boundaries are defined by surfaces nested inside each other. The outermost region is infinite.
+*The nest card defines an entire universe consisting of nested material regions.
+**The boundaries are defined by surfaces nested inside each other.
+**The outermost region is infinite.
 *The material entries can be replaced by <tt>'''fill''' ''U<sub>0</sub>''</tt>, in which case the region is filled by another universe.
-*The first format allows defining nests in which all surfaces are of same type and centred at the origin. Only surfaces that are characterized by a single outer radius are accepted ([[Surface_types#Second-order_quadratic_surfaces|cylinders, spheres]] and some [[Surface_types#Regular_prisms|regular prisms]]). The [[#pin (pin geometry definition)|pin]] and [[#particle (particle geometry definition)|particle]] definitions are short-hand notations of the nest card.
+*The first format allows defining nests in which all surfaces are of same type and centred at the origin.
+**Only surfaces that are characterized by a single outer radius are accepted ([[Surface_types#Second-order_quadratic_surfaces|cylinders, spheres]] and some [[Surface_types#Regular_prisms|regular prisms]]).
+**The [[#pin (pin geometry definition)|pin]] and [[#particle (particle geometry definition)|particle]] definitions are short-hand notations of the nest card.
 *The second format allows mixing different surface types. In this case all surface parameters need to be provided after the surface type.
@@ Line 1,796: / Line 2,235: @@
 |-
 | <tt>''R<sub>1</sub> ... R<sub>N-1</sub>''</tt>
-| : outer radii
+| : outer radii [in cm]
 |}
 <u>Notes:</u>
-*The particle card defines an entire universe consisting of nested spherical shells. The boundaries are defined by sphere surfaces. The outermost region is radially infinite.
+*The particle card defines an entire universe consisting of nested spherical shells.
+**The boundaries are defined by sphere surfaces.
+**The outermost region is radially infinite.
 *The material entries can be replaced by <tt>'''fill''' ''U<sub>0</sub>''</tt>, in which case the region is filled by another universe.
 *Most typically used for defining TRISO fuel particles.
@@ Line 1,807: / Line 2,248: @@
 *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:
 {|
-| <tt>''U<sub>0</sub>''</tt>
+| <tt>''UNI<sub>0</sub>''</tt>
 | : 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
 |-
@@ Line 1,823: / Line 2,264: @@
 |-
 | <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>
-|: are the coordinates
+|: are the coordinates [in cm]
 |-
 |<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
 |}
@@ Line 1,847: / Line 2,292: @@
 <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.
-*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]].
@@ Line 1,867: / Line 2,312: @@
 |}
-The syntax for the different types is as follows:
+The enery-resolution types are:
   '''phb''' ''NAME'' '''1''' ''INTT'' ''E<sub>max,1</sub>'' ''R<sub>1</sub>'' ''E<sub>max,2</sub>'' ''R<sub>2</sub>'' ...
-where:
 {|
 |<tt>''INTT''</tt>
@@ Line 1,877: / Line 2,322: @@
 |-
 |<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 1,887: / Line 2,332: @@
   '''phb''' ''NAME'' '''2''' ''INTT'' ''E<sub>max,1</sub>'' ''FWHM<sub>1</sub>'' ''E<sub>max,2</sub>'' ''FWHM<sub>2</sub>'' ...
-where:
 {|
 |<tt>''INTT''</tt>
@@ Line 1,893: / Line 2,337: @@
 |-
 |<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 1,902: / Line 2,346: @@
   '''phb''' ''NAME'' '''3''' ''a'' ''b''
-where:
 {|
 |<tt>''a, b''</tt>
@@ Line 1,910: / Line 2,353: @@
   '''phb''' ''NAME'' '''4''' ''a'' ''b'' ''c''
-where:
 {|
 |<tt>''a, b, c''</tt>
@@ Line 1,933: / Line 2,375: @@
 |-
 | <tt>''R<sub>1</sub> ... R<sub>N-1</sub>''</tt>
-| : outer radii
+| : outer radii [in cm]
 |}
 <u>Notes:</u>
-*The pin card defines an entire universe consisting of nested annular material regions. The boundaries are defined by axially infinite cylindrical surfaces. The outermost region is radially infinite.
+*The pin card defines an entire universe consisting of nested annular material regions.
+**The boundaries are defined by axially infinite cylindrical surfaces.
+**The outermost region is radially infinite.
 *The material entries can be replaced by <tt>'''fill''' ''U<sub>0</sub>''</tt>, in which case the region is filled by another universe.
 *Most typically used for defining fuel pins, but can also be applied to guide tubes, control rods, etc.
@@ Line 1,956: / Line 2,400: @@
 |-
 | <tt>''XPIX''</tt>
-| : horizontal image size in pixels
+| : horizontal image size [in pixels]
 |-
 | <tt>''YPIX''</tt>
-| : vertical image size in pixels
+| : vertical image size [in pixels]
 |-
 | <tt>''POS''</tt>
-| : position of plot plane
+| : position of plot plane [in cm]
 |-
 | <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>
-| : maximum horizontal coordinate of plotted region
+| : maximum horizontal coordinate of plotted region [in cm]
 |-
 | <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>
-| : maximum vertical coordinate of plotted region
+| : maximum vertical coordinate of plotted region [in cm]
 |-
 | <tt>''F<sub>min</sub>''</tt>
@@ Line 1,983: / Line 2,427: @@
 |-
 | <tt>''E''</tt>
-| : particle energy for importance map plots
+| : particle energy for importance map plots [in MeV]
 |}
 <u>Notes:</u>
-*The type 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.
-*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 relative dimensions of image size (<tt>''XPIX''</tt>, <tt>''YPIX''</tt>) should match that of the plotted region. Otherwise the image gets distorted.
-*Void is plotted in black and special colors are used to plot geometry errors (red = overlap, green = undefined region).
+*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).
-*The position parameter defines the location of the plot plane on the axis perpendicular to it (e.g. z-coordinate for xy-type plot).
+*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 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.
+** If the 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.
-*Geometry plotter requires compiling the source code with [[Compiling Serpent|GD Graphics libraries]].
+*The second format allows to plot he importance maps read using the [[#wwin (weight window mesh definition)|wwin card]]:
-*[[Installing and running Serpent#Running Serpent|Command line parameter]] '-plot' stops the execution after the geometry plots are produced. Option '-qp' invokes a quick plot mode, which does not check for overlaps. Option '-noplot' skips the geometry plots altogether.
+**They can be plotted on top of the geometry by setting the second value ('B') of the type parameter for:
+*** 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.
-*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''' ''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>
 | : 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>
 | : name of the source material, from which the flow is moved
@@ Line 2,028: / Line 2,523: @@
 | : continuous reprocessing mode
 |-
+|}
+<u>Notes:</u>
+*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
+**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 regime works with materials, not universes. Therefore, define the universes associated with those burnable materials as surface-cell type universes.
+*The continuous reprocessing regime 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 continuous reprocessing <tt>''MODE''</tt> defines how to incorporate the material flow into the Bateman equations:
+::{|class="wikitable" style="text-align: center;"
+! <tt>''MODE''</tt>
+! Material source
+! Material target
+! Material flow
+|-
+| 0
+| <math>N_{src}(t)=N_{src}(0)</math>
+| <math>N_{tgt}(t)=N_{tgt}(0)+t\lambda N_{src}(0)</math>
+| <math>\dot{N}(t)=\lambda N_{src}(0)</math>
+|-
+| 1
+| <math>N_{src}(t)=N_{src}(0)e^{-\lambda t}</math>
+| <math>N_{tgt}(t) = N_{tgt}(0) + (1-e^{-\lambda t})N_{src}(0)</math>
+| <math>\dot{N}(t)=\lambda N_{src}(t)</math>
+|-
+| 2
+| <math>N_{src}(n+1)=(1-\Delta t_{n}\lambda)N_{src}(n)</math>
+| <math>N_{tgt}(n+1)=N_{tgt}(n)+\Delta t_{n}\lambda N_{src}(n)</math>
+| <math>\dot{N}(n)=\lambda N_{src}(n)</math>
+|}
+:*<tt>''MODE''</tt> 0 : no changes at the source material and adds ''&lambda;N<sub>0</sub>'' from the source material to the target material when solving the Bateman equations (''N<sub>0</sub>'' are initial compositions).
+:*<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'').
+Reprocessing material regime (<tt>'''rm'''</tt>):<span id="rep_rm"></span>
+{|
 | <tt>''MAT<sub>1</sub>''</tt>
 | : name of the replaced material
@@ Line 2,034: / Line 2,569: @@
 | : 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
@@ Line 2,043: / Line 2,587: @@
 <u>Notes:</u>
+*The universe reprocessing regime replaces one universe with another, ''UNI<sub>1</sub>'' by ''UNI<sub>2</sub>''.
-*The reprocessor name identifies the reprocessing regime in the depletion calculation [[#dep|dep card]]. The syntax corresponds to '''dep pro''' ''NAME''.
+=== sample (temperature / density data sample definition)<span id="sample"></span> ===
-*Multiple reprocessing controllers/regimes can be defined within the same reprocessor definition.
-*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 continuous reprocessing option defines how to incorporate the material flow into the Bateman equations:
-**Reprocessing modes: <math>\begin{matrix} & \text{material source} & \text{material target} & \text{material flow} \\
-\text{0: } & N_{src}(t)=N_{src}(0) & N_{tgt}(t)=N_{tgt}(0)+t\lambda N_{src}(0) & \dot{N}(t)=\lambda N_{src}(0) \\
-\text{1: } & N_{src}(t)=N_{src}(0)e^{-\lambda t} & N_{tgt}(t) = N_{tgt}(0) + (1-e^{-\lambda t})N_{src}(0) & \dot{N}(t)=\lambda N_{src}(t) \\
-\text{2: } & N_{src}(n+1)=(1-\Delta t_{n}\lambda)N_{src}(n) & N_{tgt}(n+1)=N_{tgt}(n)+\Delta t_{n}\lambda N_{src}(n) & \dot{N}(n)=\lambda N_{src}(n) \\ \end{matrix}</math>
-**MODE 0 : no changes at the source material and adds ''λN<sub>0</sub>'' from the source material to the target material when solving the Bateman equations (''N<sub>0</sub>'' are initial compositions).
-**MODE 1 : subtracts ''λN'' from the source material and adds it to the target material when solving the Bateman equations.
-**MODE 2 : subtracts ''λN<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> ===
   '''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 2,069: / Line 2,599: @@
 {|
 | <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>
-| : 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>
-| : 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>
-| : Number of values to sample in the y-direction.
+| : number of values to sample in the y-direction.
 |-
 | <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>
-| : 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>
-| : Number of values to sample in the z-direction.
+| : number of values to sample in the z-direction.
 |-
 | <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>
-| : Maximum coordinate to sample from in the z-direction.
+| : maximum coordinate to sample from in the z-direction [in cm]
 |-
 |}
@@ Line 2,099: / Line 2,629: @@
 <u>Notes:</u>
-*The data from each sample is written in a separate <tt>[input]_sample''N''.m</tt> file.
+*The data from each sample is written in a separate output file:
-*Positive values for the density data correspond to atomic densities, while negative values correspond to mass densities.
+**Default: <tt>[input]_sample[n].m</tt>, where "<tt>n</tt>" is the sample number.
-*Materials with no temperature specified either in their [[Input syntax manual#mat|mat]]-card or through an interface definition will show a temperature of 0.
+**Coupled multi-physics calculation: <tt>[input]_sample[n]_iter[i].m</tt>, where "<tt>i</tt>" is the iteration number.
+**Burnup calculations: <tt>[input]_sample[n]_bstep[bu].m</tt>, where "<tt>bu</tt>" is the burnup step index.
+*** Coupled multi-physics calculation: <tt>[input]_sample[n]_bstep[bu]_iter[i].m</tt>
+**Time-dependent calculations: <tt>[input]_sample[n]_tstep[t].m</tt>, where "<tt>t</tt>" is the time step index
+*** Coupled multi-physics calculation: <tt>[input]_sample[n]_tstep[t]_iter[i].m</tt>
+*Sampling units:
+** Density: positive values = atomic densities [in b<sup>-1</sup> cm<sup>-1</sup>], negative value = mass density [in g/cm<sup>3</sup>].
+** Temperature: [in K]
+*Materials with no temperature specified either in their [[#mat|mat]] card or through an interface [[#ifc|ifc]] card definition will show a temperature of 0 K.
 === sens (sensitivity calculation definition)<span id="sens"></span> ===
@@ Line 2,131: / Line 2,669: @@
 |-
 | <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>
@@ Line 2,137: / Line 2,675: @@
 |-
 | <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>
-| : Path to the unstructured mesh points file
+| : path to the unstructured mesh points file
 |-
 | <tt>''FACES_FILE''</tt>
-| : Path to the unstructured mesh faces file
+| : path to the unstructured mesh faces file
 |-
 | <tt>''OWNER_FILE''</tt>
-| : Path to the unstructured mesh owner file
+| : path to the unstructured mesh owner file
 |-
 | <tt>''NEIGHBOUR_FILE''</tt>
-| : Path to the unstructured mesh neighbour file
+| : path to the unstructured mesh neighbour file
 |-
 | <tt>''MATERIALS_FILE''</tt>
-| : Path to the unstructured mesh materials file
+| : path to the unstructured mesh materials file
 |}
@@ Line 2,177: / Line 2,715: @@
 |-
 | <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>
@@ Line 2,186: / Line 2,724: @@
 |-
 | <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>
-| : 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>
-| : Name of solid body <tt>''i''</tt>
+| : name of solid body <tt>''i''</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>
-| : Material filling cell <tt>''i''</tt>
+| : material filling cell <tt>''i''</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>
-| : 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>
-| : 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>
-| : 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>
-| : 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 2,226: / Line 2,764: @@
 {|
 | <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 2,248: / Line 2,786: @@
            [ [[#src_ss|'''ss''']] ''SURF'' ]
            [ [[#src_sd|'''sd''']] ''U'' ''V'' ''W'' ]
+          [ [[#src_sa|'''sa''']] ''PHI'' ]
            [ [[#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>'' ... ]
@@ Line 2,255: / Line 2,794: @@
            [ [[#src_si|'''si''']] ''N'' ''P<sub>1</sub>'' ''P<sub>2</sub>'' ... ]
            [ [[#src_sg|'''sg''']] ''MAT'' ''MODE'' ]
-Source definition. The first parameter:
+Source definition. The two first parameters:
 {|
+| <tt>''NAME''</tt>
+|: source name
+|-
 | <tt>''PART''</tt>
 | : 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>
@@ Line 2,286: / Line 2,834: @@
 <u>Notes:</u>
 *Setting a source cell is one of the options that can be applied to define the spatial distribution of source particles.
-*The selection is based on rejection sampling, and if the source cell occupies a small volume of the geometry, the sampling efficiency can be increased by defining a bounding box/cylinder around the cell (using the <tt>''sx''</tt>, <tt>''sy''</tt> and <tt>''sz''</tt> or <tt>''sp''</tt>, <tt>''srad''</tt> and <tt>''sz''</tt> options, respectively).
+*The selection is based on rejection sampling, and if the source cell occupies a small volume of the geometry, the sampling efficiency can be increased by defining a bounding box/(vertical) cylinder around the cell (using the <tt>'''sx'''</tt>, <tt>'''sy'''</tt> and <tt>'''sz'''</tt> or <tt>'''sp'''</tt>, <tt>'''srad'''</tt> and <tt>'''sz'''</tt> options, respectively).
 *If no spatial distribution is defined, particles are sampled uniformly over the geometry.
@@ Line 2,307: / Line 2,855: @@
 <u>Notes:</u>
 *Setting a source material is one of the options that can be applied to define the spatial distribution of source particles.
-*The selection is based on rejection sampling, and if the source material occupies a small volume of the geometry, the sampling efficiency can be increased by defining a bounding box/cylinder around the cell (using the <tt>''sx''</tt>, <tt>''sy''</tt> and <tt>''sz''</tt> or <tt>''sp''</tt>, <tt>''srad''</tt> and <tt>''sz''</tt> options, respectively).
+*The selection is based on rejection sampling, and if the source material occupies a small volume of the geometry, the sampling efficiency can be increased by defining a bounding box/(vertical) cylinder around the cell (using the <tt>'''sx'''</tt>, <tt>'''sy'''</tt> and <tt>'''sz'''</tt> or <tt>'''sp'''</tt>, <tt>'''srad'''</tt> and <tt>'''sz'''</tt> options, respectively).
 *If no spatial distribution is defined, particles are sampled uniformly over the geometry.
@@ Line 2,315: / Line 2,863: @@
 {|
 | <tt>''X''</tt>, <tt>''Y''</tt>, <tt>''Z''</tt>,
-| : coordinates of the source point
+| : coordinates of the source point [in cm]
 |}
@@ Line 2,327: / Line 2,875: @@
 {|
 | <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>
-| : Boundaries on Y-axis
+| : boundaries on Y-axis [in cm]
 |-
 | <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>
-| : Radial boundaries
+| : radial boundaries [in cm]
 |-
 |}
 <u>Notes:</u>
-*Source boundaries are used to define a bounding box/cylinder inside which the source particles are sampled.
+*Source boundaries are used to define a bounding box/(vertical) cylinder inside which the source particles are sampled.
 *The radial boundaries are centered around the point defined by <tt>'''sp'''</tt> and can be used in combination with <tt>'''sz'''</tt>.
 *Can be used in combination with cell and material sources to increase the sampling efficiency.
@@ Line 2,347: / Line 2,895: @@
-Surface source (<tt>'''ss'''</tt>):<span id="src_ss"></span>
+Source surface (<tt>'''ss'''</tt>):<span id="src_ss"></span>
 {|
@@ Line 2,356: / Line 2,904: @@
 <u>Notes:</u>
 *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 2,370: / Line 2,919: @@
 *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>
@@ Line 2,375: / Line 2,935: @@
 {|
 | <tt>''E''</tt>
-| : energy of source particles
+| : energy of source particles [in MeV]
 |}
@@ Line 2,394: / Line 2,954: @@
 |-
 | <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>
@@ Line 2,404: / Line 2,964: @@
 *The bins are entered in the order of ascending energy, and weight of the first bin must be set to zero.
 *Interpolation is given in a separate parameter from version 2.1.31 on.
+*Here, a simple test input that demonstrates the [[Source energy spectrum definition test case|source spectrum definition]].
 Source reaction (<tt>'''sr'''</tt>):<span id="src_sr"></span>
@@ Line 2,412: / Line 2,973: @@
 |-
 | <tt>''MT''</tt>
-| : reaction number
+| : reaction number identifier
 |}
@@ Line 2,426: / Line 2,987: @@
 {|
 | <tt>''T<sub>MIN</sub>''</tt>, <tt>''T<sub>MAX</sub>''</tt>
-| : time boundaries
+| : time boundaries [in s]
 |}
@@ Line 2,445: / Line 3,006: @@
 <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.
@@ Line 2,470: / Line 3,031: @@
 {|
 | <tt>''MAT''</tt>
-| : material name or -1
+| : material name or "-1" to refer to all radioactive materials
 |-
 | <tt>''MODE''</tt>
@@ Line 2,477: / Line 3,038: @@
 <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).
-*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 implicit sampling mode preserves the total statistical weight of emitted particles and produces a uniform source distribution over activated materials.
@@ Line 2,488: / Line 3,048: @@
 === 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> ===
-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> ===
@@ Line 2,527: / Line 3,095: @@
   '''thermstoch''' ''NAME'' ''TEMP'' ''LIB<sub>1</sub>'' ''LIB<sub>2</sub>''
-Defines thermal scattering data that can be linked to nuclides using input entry '''moder''' in the [[#mat (material definition)|material cards]]. When using thermal scattering data together with TMS on-the-fly temperature treatment, the third input value of the therm card is 0. In this case, Serpent interpolates the thermal scattering data automatically to the local temperature, as defined either using the '''tms''' input entry in the material definition or via the multi-physics interface.
+Defines thermal scattering data that can be linked to nuclides using input entry '''moder''' in the [[#mat (material definition)|material cards]]. Input values:
-Input values:
 {|
 | <tt>''NAME''</tt>
@@ Line 2,538: / Line 3,105: @@
 |-
 | <tt>''TEMP''</tt>
-| : temperature to which the thermal scattering data is interpolated
+| : temperature to which the thermal scattering data is interpolated [in K]
 |}
 <u>Notes:</u>
+* On-the-fly thermal motion sampling (TMS) temperature treatment:
-*When using on-the-fly interpolation of thermal scattering data, <tt>''LIB<sub>i</sub>''</tt> must cover the whole temperature range in which the materials appear in the geometry. I.e. extrapolation of the data is not supported.
+**It requires the third value of the therm card to be set to "<tt>0</tt>"
-*Thermal scattering data is interpolated using the methodology of makxsf code
+**The thermal scattering data is automatically interpolated to the local temperature.
-*The interpolation can be performed using the stochastic mixing approach with '''thermstoch'''. This interpolation mode is not available for on-the-fly interpolation.
+**The local temperature is either defined using:
-*The continuous S(&alpha; &beta;) formalism is available from version 2.1.32 on.
+*** the '''tms''' entry in the material card (see [[#mat|mat card]])
+*** the multi-physics interface (see [[#ifc|ifc card]]), where the temperature limits are defined using the '''tft''' entry in the material card (see [[#mat|mat card]])
+**The thermal scattering libraries <tt>''LIB<sub>i</sub>''</tt> must cover the whole range in which the materials appear in the geometry, i.e. data extrapolation is not supported.
+*Interpolation:
+**Thermal scattering data is interpolated using the methodology of makxsf code<ref name="makxsf">Brown, F. B. ''"The makxsf Code with Doppler Broadening"'', Los Alamos National Laboratory Tech. Rep., LA-UR-06-7002, Los Alamos, NM, [https://mcnp.lanl.gov/pdf_files/TechReport_2006_LANL_LA-UR-06-7002_Brown.pdf 2006]</ref>.
+**Alternatively, the interpolation can be performed using the stochastic mixing approach with the '''thermstoch''' entry.
+***This interpolation mode doesn't support on-the-fly interpolation.
+*The continuous S(&alpha;, &beta;) formalism:
+**It is available from version 2.1.32 on.
+**The on-the-fly temperature treatment is available from version 2.2.0 and on.
 === tme (time binning definition)<span id="tme"></span> ===
@@ Line 2,566: / Line 3,142: @@
 |-
 | <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>
-| : 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>
-| : maximum time boundary in uniform or log-uniform binning
+| : maximum time boundary in uniform or log-uniform binning [in s]
 |}
 <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.
 *Time binning is used with [[#det (detector definition)|detectors]] and [[Dynamic external source simulation mode|dynamic simulation mode]].
@@ Line 2,603: / Line 3,178: @@
 |-
 | <tt>''IDX''</tt>
-| : index number in lattice transformation
+| : index number in lattice transformation (type L)
 |-
 | <tt>''LVL''</tt>
@@ Line 2,609: / Line 3,184: @@
 |-
 | <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>
-| : 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>
@@ Line 2,621: / Line 3,196: @@
 |-
 | <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>
-| : Components of vector defining rotation axis.
+| : components of vector defining rotation axis.
 |-
 | <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 2,636: / Line 3,211: @@
 *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>.
-*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.
 *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> ===
@@ Line 2,649: / Line 3,226: @@
 {|
 | <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>
@@ Line 2,656: / Line 3,233: @@
 <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> ===
@@ Line 2,675: / Line 3,254: @@
 |-
 | <tt>''IDX''</tt>
-| : index number in lattice transformation
+| : index number in lattice transformation (type L)
 |-
 | <tt>''T<sub>0</sub>''</tt>
-| : beginning time of the transformation
+| : beginning time of the transformation [in s]
 |-
 | <tt>''T<sub>1</sub>''</tt>
-| : end time of the transformation
+| : end time of the transformation [in s]
 |-
 | <tt>''T<sub>TYPE</sub>''</tt>
@@ Line 2,687: / Line 3,266: @@
 |-
 | <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>
-| : Initial acceleration vector
+| : initial acceleration vector [in cm/s<sup>2</sup>]
 |}
@@ Line 2,699: / Line 3,278: @@
 *See [[UGM 2016 Moving geometry]].
 *See [[Rotating Translating STL Bunny]].
+===umsh (unstructured mesh-based geometry definition)<span id="umsh"></span> ===
+ ''UNI'' ''BGUNI''
+ ''MESH_SPLIT'' ''MESH_DIM'' ''SZ<sub>1</sub>'' ''SZ<sub>2</sub>'' ... ''SZ<sub>MESH_DIM</sub>''
+ ''POINTS_FILE''
+ ''FACES_FILE''
+ ''OWNER_FILE''
+ ''NEIGHBOUR_FILE''
+ ''MATERIALS_FILE''
+Defines an unstructured mesh-based geometry. Input values:
+{|
+| <tt>''UNI''</tt>
+| : universe name for the unstructured mesh-based geometry
+|-
+| <tt>''BGUNI''</tt>
+| : name of the background universe filling all undefined space
+|-
+| <tt>''MESH_SPLIT''</tt>
+| : splitting criterion for the adaptive search mesh (maximum number of geometry cells in search mesh cell)
+|-
+| <tt>''MESH_DIM''</tt>
+| : number of levels in the adaptive search mesh
+|-
+| <tt>''SZ<sub>i</sub>''</tt>
+| : size of the search mesh at level <tt>''i''</tt>
+|-
+| <tt>''POINTS_FILE''</tt>
+| : path to the unstructured mesh points file
+|-
+| <tt>''FACES_FILE''</tt>
+| : path to the unstructured mesh faces file
+|-
+| <tt>''OWNER_FILE''</tt>
+| : path to the unstructured mesh owner file
+|-
+| <tt>''NEIGHBOUR_FILE''</tt>
+| : path to the unstructured mesh neighbour file
+|-
+| <tt>''MATERIALS_FILE''</tt>
+| : path to the unstructured mesh materials file
+|}
+<u>Notes:</u>
+*See also description of [[#solid (irregular 3D geometry definition)|solid (irregular 3D geometry definition), type 1]].
 === 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> ===
@@ Line 2,747: / Line 3,443: @@
 |-
 | <tt>''MIN<sub>n</sub>''</tt>
-| : minimum mesh boundary (''n''th coordinate)
+| : minimum mesh boundary (''n''-th coordinate)
 |-
 | <tt>''MAX<sub>n</sub>''</tt>
-| : maximum mesh boundary (''n''th coordinate)
+| : maximum mesh boundary (''n''-th coordinate)
 |-
 | <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>
-| : mth boundary of ''n''th coordinate)
+| : mesh boundary ''m''-th (''n''-th coordinate)
 |-
 | <tt>''X<sub>0</sub>'', ''Y<sub>0</sub>''</tt>
@@ Line 2,807: / Line 3,503: @@
 |}
-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>
 *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]].
 *See also [[#set wwb|set wwb]] and [[#set maxsplit|set maxsplit]] for setting options for weight windows, splitting and Russian roulette.
@@ Line 2,818: / Line 3,514: @@
 *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>
@@ Line 2,845: / Line 3,543: @@
 |-
 | <tt>''E''</tt>
-| : energy used for renormalization
+| : energy used for renormalization [in MeV]
 |}
@@ Line 2,945: / Line 3,643: @@
 |-
 | <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>
-| : minimum cell dimension
+| : minimum cell dimension [in cm]
 |}
@@ Line 2,955: / Line 3,653: @@
 *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.
-*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.
 *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 2,968: / Line 3,666: @@
   '''set absrate''' ''A'' [ ''MAT'' ]
-Sets normalization to total absorption rate.
+Sets normalization to total absorption rate. Input values:
 {|
 | <tt>''F''</tt>
-| : number of  neutrons absorbed per second (neutrons/s)
+| : number of  neutrons absorbed per second [in neutrons/s]
 |-
 | <tt>''MAT''</tt>
@@ Line 2,981: / Line 3,679: @@
 *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.
-*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]].
 *See also Section 5.8 of [http://montecarlo.vtt.fi/download/Serpent_manual.pdf Serpent 1 User Manual].
@@ Line 2,999: / Line 3,698: @@
 *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 cross section directory 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 cross section directory files in this path if not found at the absolute.
-*A default cross section directory file can be set by defining environment variable SERPENT_ACELIB. This file will be used if no other path is given with '''set acelib'''.
+*A default cross section directory file can be set by defining environment variable <tt>SERPENT_ACELIB</tt>. This file will be used if no other path is given with '''set acelib'''.
 === set adf ===
   '''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 3,018: / Line 3,717: @@
 |-
 | <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 3,024: / Line 3,723: @@
 *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 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 3,034: / Line 3,733: @@
 *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).
+*Sign moments of net and partial currents are not scored for [[Surface_types#Regular_prisms|Y-type infinite/truncated hexagonal prisms]].
 === set alb ===
@@ Line 3,055: / Line 3,755: @@
 *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 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 universe is needed only for labelling the results in the output files.
@@ Line 3,068: / Line 3,768: @@
 {|
 | <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>
-| : 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>
-*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]].
@@ Line 3,092: / Line 3,792: @@
 *Some burnup applications require separate treatment for isotopes that are used as burnable absorbers but also produced in fission. This input parameter can be used to separate the transmutation chains.
 *Isotope handled as the burnable absorber is created by duplicating the original and renaming it as ''ZAI<sub>n</sub> + 1000''.
-*For Gd-155, for example, the fission product isotope would be assigned ZAI 641550 and the burnable absorber ZAI 642550.
+**For Gd-155, for example, the fission product isotope would be assigned ZAI 641550 and the burnable absorber ZAI 642550.
-*The input parameter defines the entire transmutation chain. Listing Gd-isotopes 641540 641550 641560 641570 641580 creates a transmutation path from Gd-154 to Gd-158. Listing only the main absorbers (641550 641570) produces a different result, since the capture products of Gd-155 and Gd-157 are lost.
+*The input parameter defines the entire transmutation chain.
+**Listing Gd-isotopes 641540 641550 641560 641570 641580 creates a transmutation path from Gd-154 to Gd-158.
+**Listing only the main absorbers (641550 641570) produces a different result, since the capture products of Gd-155 and Gd-157 are lost.
 === set bala ===
@@ Line 3,102: / Line 3,804: @@
 {|
 | <tt>''OPT''</tt>
-| : option (0 = off, 1 = on)
+| : probability to store particles in common queue (0 = off, >0 = on)
 |}
 <u>Notes:</u>
 *Load balancing may improve OpenMP parallel scalability in calculations with significant branching (most typically related to coupled neutron/photon calculations or variance reduction).
-*The option is off by default.
+*The option is ON with <tt>''OPT''</tt>= 1 with weight-window/variance reduction calculations and dynamic/time-dependent calculation modes. Otherwise, it is set OFF by default. Before version 2.2.0, the default behavior was always OFF.
 *When this option is set, the random number sequence is no longer preserved.
@@ Line 3,165: / Line 3,867: @@
 <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 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.
 *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 3,171: / Line 3,874: @@
 *For symmetry purposes Serpent provides the [[#set usym|universe symmetry option]].
 *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 ===
@@ Line 3,201: / Line 3,903: @@
 <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.
 *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 ===
@@ Line 3,216: / Line 3,916: @@
 {|
 | <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>
-| : weight lower-boundary
+| : weight lower-boundary (default value: 0.2)
 |-
 | <tt>''WGT_HIGH''</tt>
-| : weight upper-boundary
+| : weight upper-boundary (default value: 10.0)
 |}
 <u>Notes:</u>
-*The default weight limits for the branchless methods are 0.2 and 10
 *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 ===
@@ Line 3,238: / Line 3,940: @@
 {|
 | <tt>''MODE''</tt>
-| : burnup calculation mode
+| : burnup calculation mode (default value: 2 = CRAM)
 |-
 | <tt>''ORDER''</tt>
-| : CRAM order
+| : CRAM order (default value: 14 - 14 PFD CRAM)
 |-
 | <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:
-{| class="wikitable" style="text-align: left;"
+::{| class="wikitable" style="text-align: left;"
 ! Mode
 ! Description
@@ Line 3,262: / Line 3,964: @@
 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
-|-
 | <tt>2</tt>
-|-
 | <tt>4</tt>
-|-
 | <tt>6</tt>
-|-
 | <tt>8</tt>
-|-
 | <tt>10</tt>
-|-
 | <tt>12</tt>
-|-
 | <tt>14</tt>
-|-
 | <tt>16</tt>
-|-
 | <tt>-16</tt>
-|-
 | <tt>-48</tt>
 |}
 <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.
+*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.
-*Negative values of CRAM order mean using IPF form of CRAM with order of the absolute value of the parameter.
+*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.
-*Positive values refer to PFD form of CRAM.
+*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.
-*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.
+*Version 2.2.0 includes the sub-step method for depletion calculations involving continuous reprocessing.
-*Positive values of <tt>''SSD''</tt> enforce usage of CRAM with given number of substeps.
+=== set bunorm ===
+ '''set bunorm''' ''NORM''
+Sets the burnup calculation normalization mode if it is not bound to a single material. Input values:
+{|
+| <tt>''NORM''</tt>
+| : burnup calculation normalization mode (1 = all materials, 2 = burnable materials, 3 = non-burnable materials). (default value: 1)
+|}
 === set ccmaxiter ===
@@ Line 3,301: / Line 4,003: @@
 {|
 | <tt>''NITER''</tt>
-| : number of iterations.
+| : number of iterations (default value: 1 = no iteration)
 |}
 <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.
 *See [[Coupled multi-physics calculations]] for further information.
@@ Line 3,317: / Line 4,018: @@
 {|
 | <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>
-*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.
 *Only the population simulated during active cycles is included in this amount.
@@ Line 3,335: / Line 4,035: @@
 {|
 | <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 3,349: / Line 4,049: @@
 {|
 | <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 3,362: / Line 4,062: @@
 {|
 | <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>
-| : 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>
-| : 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>
-| : minimum mean time interval for scoring the CFE for photons
+| : minimum mean time interval for scoring the CFE for photons [in s]
 |}
@@ Line 3,378: / Line 4,078: @@
 *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.
-*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]].
 *The collision flux estimator in Serpent is described in an article in Annals of Nuclear energy from 2017.<ref name="cfe">Leppänen, J.
@@ Line 3,411: / Line 4,111: @@
 {|
 | <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>
@@ Line 3,421: / Line 4,121: @@
 *The group constant output file <tt>[input].coe</tt> is produced when the [[automated burnup sequence]] is invoked.
 *The available parameters are listed under [[Output parameters#Homogenized group constants|homogenized group constants]] in the description of the <tt>[input]_res.m</tt> output file.
-*Detectors are identified by the name assigned to them in the [[#det (detector definition)|detector card]].
+*Detectors are identified by the name assigned to them in the [[#det (detector definition)|det card]].
 === set combing ===
@@ Line 3,430: / Line 4,130: @@
 {|
 | <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)
 |}
 <u>Notes:</u>
-*The combing method can achieve variance reduction and save computer time by keeping the population size approximately constant over time steps. In super-critical systems, it prevents the population from growing without bound while, in sub-critical systems, it does it from dying. In critical systems, it avoids the divergence of the variance of the population due to fluctuations of fission chains.
+*The combing method can achieve variance reduction and save computer time by keeping the population size approximately constant over time steps.
+**In super-critical systems, it prevents the population from growing without bound.
+**In sub-critical systems, it prevents the population from dying.
+**In critical systems, it avoids the divergence of the variance of the population due to fluctuations of fission chains.
 === set comfile ===
@@ Line 3,452: / Line 4,155: @@
 *Setting up a communication mode will enable the coupled calculation mode.
+*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]]
@@ Line 3,461: / Line 4,165: @@
 {|
 | <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>
-*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 ===
@@ Line 3,475: / Line 4,179: @@
 {|
 | <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)
 |}
 <u>Notes:</u>
+*It enables first-order uncertainty propagation by collapsing the covariance data with the evaluated sensitivities.
+**It applies the <u>Sandwich rule</u>: <math>\mathbf{Cov}^R_X \approx ({\overline{\mathbf{S}}}^R_X)^T {\overline{\overline{\mathbf{Cov}}}}^R_{X,X} \overline{\mathbf{S}}^R_X</math>
-*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.
+:: where: <math>{\overline{\mathbf{S}}}^R_X \in \R^{N_g \times 1} </math> is the sensitivity vector containing the group sensitivities
+::::<math>{\overline{\overline{\mathbf{Cov}}}}^R_{X,X} \in \R^{N_g \times N_g}</math> is the covariance matrix containing the group covariances
+**The methodology is described in a related report<ref name="UncReport18">Valtavirta, V. ''"Nuclear data uncertainty propagation to Serpent generated group and time constants"'', Research report VTT-R-04681-18 [http://montecarlo.vtt.fi/download/VTT-R-04681-18_web.pdf (2018)].</ref>.
+*It requires setting the sensitivity calculation parameters. For more information, see the detailed description on [[Sensitivity calculations|sensitivity calculations]].
 === set covlib ===
@@ Line 3,492: / Line 4,201: @@
 |}
+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>
 <u>Notes:</u>
+*It enables first-order uncertainty propagation by collapsing the covariance data with the evaluated sensitivities.
+**It applies the <u>Sandwich rule</u>: <math>\mathbf{Cov}^R_X \approx ({\overline{\mathbf{S}}}^R_X)^T {\overline{\overline{\mathbf{Cov}}}}^R_{X,X} \overline{\mathbf{S}}^R_X</math>
-*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.
+:: where: <math>{\overline{\mathbf{S}}}^R_X \in \R^{N_g \times 1} </math> is the sensitivity vector containing the group sensitivities
+::::<math>{\overline{\overline{\mathbf{Cov}}}}^R_{X,X} \in \R^{N_g \times N_g}</math> is the covariance matrix containing the group covariances
+**The methodology is described in a related report<ref name="UncReport18">Valtavirta, V. ''"Nuclear data uncertainty propagation to Serpent generated group and time constants"'', Research report VTT-R-04681-18 [http://montecarlo.vtt.fi/download/VTT-R-04681-18_web.pdf (2018)].</ref>.
+*It requires setting the sensitivity calculation parameters. For more information, see the detailed description on [[Sensitivity calculations|sensitivity calculations]].
 === set cpd ===
-  '''set cpd''' ''DEPTH'' [ ''N<sub>Z</sub> Z<sub>MIN</sub> Z<sub>MAX</sub>'' ]
+  '''set cpd''' ''DEPTH'' [ ''N<sub>Z</sub> Z<sub>MIN</sub> Z<sub>MAX</sub>'' ] [ ''LVL1'' ''LVL2'' ]
 Sets on the calculation of lattice-wise power distributions to output file <tt>[input]_core0.m</tt> on. Input values:
 {|
 | <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>
-| : 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>
-| : 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>
-| : 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>
+|: User-defined first level where to define the lattice-wise power distribution
 |-
+| <tt>''LVL2''</tt>
+|: User-defined second level where to define the lattice-wise power distribution
 |}
 <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''' ''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 3,555: / Line 4,314: @@
 *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''' ''TABLE_LIST''
+Defines the tables included in the nuclear and material data file <tt>[input].out</tt>. Input values:
+{|
+| <tt>''TABLE_LIST''</tt>
+| : 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>
+*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.
+*For more information, see detailed description of the [[Description of output files#Nuclide and material data output|nuclear and material data output]].
 === set dbrc ===
@@ Line 3,563: / Line 4,381: @@
 {|
 | <tt>''E<sub>min</sub>''</tt>
-| : Minimum energy for DBRC
+| : Minimum energy for DBRC [in MeV]
 |-
 | <tt>''E<sub>max</sub>''</tt>
-| : Maximum energy for DBRC
+| : Maximum energy for DBRC [in MeV]
 |-
 | <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>
-*This description is not complete.
 *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].
@@ Line 3,580: / Line 4,397: @@
 === set dd ===
-  '''set dd''' ''MODE''
+  '''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>
-| : decomposition mode (0 = none, 1 = simple, 2 = sector, 3 = sector + center)
+| : decomposition mode (default value: 0)
+|-
+| <tt>''X<sub>0</sub>''</tt>
+| 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>
+| 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>
+| 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>
 *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]].
-*Decomposed materials are plotted in domain-specific colors (unless the rgb option in the material card is used)
-*See [[Domain decomposition|practical example]] for more information.
 === set declib ===
@@ Line 3,609: / Line 4,457: @@
 <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.
-*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 <tt>SERPENT_DATA</tt>. The code looks for decay data files in this path if not found at the absolute.
+*From version 2.2.0 and on, a default decay data library directory file can be set by defining environment variable <tt>SERPENT_DECLIB</tt>. This file will be used if no other path is given with '''set declib'''.
 === set decomp ===
@@ Line 3,643: / Line 4,492: @@
 <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.
@@ Line 3,651: / Line 4,501: @@
   '''set depmtx''' ''MODE''
-Print burnup matrixes to <tt>depmtx_[mat][step].m</tt> file during burnup calculation.
+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>
-| : 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 3,661: / Line 4,511: @@
 *With non-constant predictor, this option will stop the simulation up to version 2.1.31.
 *With multiple substeps, only the last one is kept after the simulation up to version 2.1.31.
+*The burnup matrix output is named <tt>depmtx_[mat][bu].m</tt> up to version 2.1.31.
 === set depout ===
@@ Line 3,668: / Line 4,519: @@
 {|
 | <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>
-| : 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>
 *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.
+=== set deppara ===
+  '''set deppara''' ''PARAM_LIST''
+Defines the material- and isotopic-wise variables included in the depletion output file <tt>[input]_dep.m</tt>. Input values:
+{|
+| <tt>''PARAM_LIST''</tt>
+| : 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 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>
+*For more information, see detailed description of the [[Description of output files#Burnup calculation output|burnup calculation output]].
+=== set depstepbunorm ===
+ '''set depstepbunorm''' ''NORM''
+Sets the depletion step normalization in burnup calculations based on energy deposition. Input values:
+{|
+| <tt>''NORM''</tt>
+| depletion step normalization mode based on energy deposition (1 = all materials, 2 = burnable materials)
+|}
+<u>Notes</u>
+* Default values (see [[#set edepmode|set edepmode]]):
+** For energy deposition modes 0/1:  the 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 ===
@@ Line 3,686: / Line 4,623: @@
 {|
 | <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>
-| : 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>
-| : number of points for trapezoidal integration for homogeneous flux
+| : number of points for trapezoidal integration for homogeneous flux (default value: 100)
 |}
 <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]].
 *The format was revised in update 2.1.27 (<tt>''DC''</tt> option was added between <tt>''MODE''</tt> and <tt>''NP''</tt>).
@@ Line 3,705: / Line 4,640: @@
 === set dix ===
   '''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 3,716: / Line 4,651: @@
 ''"Two practical methods for unionized energy grid construction in continuous-energy Monte Carlo neutron transport calculation."'' Ann. Nucl. Energy [https://www.sciencedirect.com/science/article/pii/S0306454909001108 '''36''' (2009) 878-885].</ref> is a method to speed-up the cross section look-up when energy grid unionization is not used for microscopic data.
 *The method can be used only in [[#set_opti|optimization modes]] 1 and 3 (modes 2 and 4 are based on energy grid unionization).
+=== set dspec ===
+ '''set dspec''' ''EGRID<sub>p</sub>'' ''EGRID<sub>n</sub>''
+Sets the energy grid structure for decay spectra. Input values:
+{|
+| <tt>''EGRID<sub>p</sub>''</tt>
+| : energy grid structure for photons
+|-
+| <tt>''EGRID<sub>n</sub>''</tt>
+| : energy grid structure for neutrons
+|}
+<u>Notes:</u>
+*The photon/neutron decay spectra is printed in the <tt>[input]_gsrc.m</tt> or <tt>[input]_nsrc.m</tt> output file, respectively.
+*The energy group spectra only include the contribution from the discrete/line spectra.
+*Setting "-1" instead of providing the energy grid structure disables the option for the given particle type.
 === set dt ===
@@ Line 3,724: / Line 4,677: @@
 {|
 | <tt>''NTRSH''</tt>
-| : probability threshold for neutrons
+| : probability threshold for neutrons (default value: 0.9)
 |-
 | <tt>''GTRSH''</tt>
-| : probability threshold for photons
+| : probability threshold for photons (default value: 0.9)
 |}
 <u>Notes:</u>
+*Tracking algorithm mode:
-*Serpent uses delta-tracking by default for both neutrons and photons, but switches to surface-tracking if the probability of sampling virtual collisions (ratio between material total cross section and the majorant) exceeds the given threshold.
+**Delta-tracking is used by default for both neutrons and photons
-*Default probability threshold for both particle types is 0.9, i.e. delta-tracking is used if the ratio between total cross section and majorant is above 0.1.
+**Switch to surface-tracking happens if the probability of sampling virtual collisions (ratio between material total cross section and the majorant) exceeds the given threshold.
-*set dt 1 means that delta-tracking is always used and set dt 0 that it is never used.
+*Probability threshold:
+**The default probability threshold for both particle types is 0.9: i.e. delta-tracking is used if the ratio between total cross section and majorant is above 0.1.
+**To enforce delta-tracking mode always: use "<tt>1</tt>".
+**To enforce surface-tracking mode always: use "<tt>0</tt>"
 *Use of delta-tracking can be enforced or blocked in individual materials using the [[#set forcedt|set forcedt]] and [[#set blockdt|set blockdt]] options
 *Integral reaction rates are scored using the collision estimator of neutron flux, which has a few adjustable parameters (see [[#set cfe|set cfe]]).
@@ Line 3,746: / Line 4,702: @@
 {|
 | <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 3,753: / Line 4,709: @@
   '''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 3,768: / Line 4,724: @@
 === 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:
 {|
 | <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>
-| : cut-off energy for photons (MeV)
+| : cut-off energy for photons [in MeV] (default value: 1.0E-3)
 |}
 <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.
 *Versions 2.1.27 and earlier include only photon energy cut-off, which is now the second input parameter.
@@ Line 3,792: / Line 4,747: @@
 {|
 | <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>
-| : cut-off energy for photons (MeV)
+| : cut-off energy for photons [in MeV]
 |}
@@ Line 3,811: / Line 4,766: @@
 |-
 | <tt>''EMIN<sub>p,i</sub>''</tt>
-| : cut-off energy for photons (MeV)
+| : cut-off energy for photons [in MeV]
 |}
@@ Line 3,817: / Line 4,772: @@
   '''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>
-| : 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>
-*Requires group constant generation to be set on.
+*Requires group constant generation to be set on (see [[#set gcu|set gcu]]).
 === set edepdel ===
@@ Line 3,835: / Line 4,790: @@
 {|
 | <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>
-| : 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 3,844: / Line 4,799: @@
 * 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.
-* 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 ===
@@ Line 3,855: / Line 4,810: @@
 {|
 | <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 3,870: / Line 4,825: @@
 {|
 | <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>
-| : 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.
-* 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]].
 === set egrid ===
@@ Line 3,894: / Line 4,874: @@
 |-
 | <tt>''EMIN''</tt>
-| : minimum energy in the grid (MeV)
+| : minimum energy in the grid [in MeV]
 |-
 | <tt>''EMAX''</tt>
-| : maximum energy in the grid (MeV)
+| : maximum energy in the grid [in MeV]
 |}
 <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.
-*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].
@@ Line 3,914: / Line 4,894: @@
 {|
 | <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>
-*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.
@@ Line 3,931: / Line 4,910: @@
 |-
 | <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>
-*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.
 *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 3,949: / Line 4,927: @@
 |-
 | <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>
-*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.
@@ Line 3,966: / Line 4,944: @@
 |-
 | <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>
-*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.
@@ Line 3,980: / Line 4,958: @@
 {|
 | <tt>''EGRID_E''</tt>
-| : energy grid size (default value is 200)
+| : energy grid size (default value: 200)
 |}
@@ Line 3,993: / Line 4,971: @@
 {|
 | <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>
-| : number of mesh cells in y-direction
+| : number of mesh cells in y-direction (default value: 5)
 |-
 | <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>
-| : minimum mesh boundary in x-direction
+| : minimum mesh boundary in x-direction [in cm]
 |-
 | <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>
-| : minimum mesh boundary in y-direction
+| : minimum mesh boundary in y-direction [in cm]
 |-
 | <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>
-| : minimum mesh boundary in z-direction
+| : minimum mesh boundary in z-direction [in cm]
 |-
 | <tt>''Z<sub>MAX</sub>''</tt>
-| : maximum mesh boundary in z-direction
+| : maximum mesh boundary in z-direction [in cm]
 |}
 <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 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]].
@@ Line 4,052: / Line 5,029: @@
 |-
 | <tt>''E<sub>n</sub>''</tt>
-| : energy deposited per fission in MeV
+| : energy deposited per fission [in MeV] (default value: 202.27)
 |}
@@ Line 4,058: / Line 5,035: @@
 *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 Section 5.8 of [http://montecarlo.vtt.fi/download/Serpent_manual.pdf Serpent 1 User Manual].
@@ Line 4,065: / Line 5,042: @@
   '''set fissrate''' ''F'' [ ''MAT'' ]
+Sets normalization to fission rate. Input values:
-Sets normalization to fission rate.
 {|
 | <tt>''F''</tt>
-| : number of fission reactions per second (1/s)
+| : number of fission reactions per second [in 1/s]
 |-
 | <tt>''MAT''</tt>
@@ Line 4,078: / Line 5,054: @@
 <u>Notes:</u>
 *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]].
 *See also Section 5.8 of [http://montecarlo.vtt.fi/download/Serpent_manual.pdf Serpent 1 User Manual].
@@ Line 4,085: / Line 5,062: @@
 === 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''' ''F'' [ ''MAT'' ]
-Sets normalization to total flux.
+Sets normalization to total flux. Input values:
 {|
@@ Line 4,108: / Line 5,090: @@
 <u>Notes:</u>
 *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]].
 *See also Section 5.8 of [http://montecarlo.vtt.fi/download/Serpent_manual.pdf Serpent 1 User Manual].
@@ Line 4,120: / Line 5,103: @@
 {|
 | <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 4,141: / Line 5,124: @@
 {|
 | <tt>''MAT<sub>n</sub>''</tt>
-| : material list (type 1)
+| : material list
 |-
 | <tt>''UNI<sub>n</sub>''</tt>
-| : universe list (type 2)
+| : universe list
 |-
 | <tt>''LVL''</tt>
-| : level number (type 3)
+| : level number
 |-
 | <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>
-| : maximum x-coordinate mesh boundary (type 4)
+| : maximum x-coordinate mesh boundary [in cm]
 |-
 | <tt>''N<sub>X</sub>''</tt>
-| : number of x-mesh cells (type 4)
+| : number of x-mesh cells
 |-
 | <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>
-| : maximum y-coordinate mesh boundary (type 4)
+| : maximum y-coordinate mesh boundary [in cm]
 |-
 | <tt>''N<sub>Y</sub>''</tt>
-| : number of y-mesh cells (type 4)
+| : number of y-mesh cells
 |-
 | <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>
-| : maximum z-coordinate mesh boundary (type 4)
+| : maximum z-coordinate mesh boundary [in cm]
 |-
 | <tt>''N<sub>Z</sub>''</tt>
-| : number of z-mesh cells (type 4)
+| : number of z-mesh cells
 |}
@@ Line 4,201: / Line 5,184: @@
   '''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 ===
@@ Line 4,210: / Line 5,200: @@
   '''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>
-| : 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>
@@ Line 4,222: / Line 5,212: @@
 <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.
@@ Line 4,233: / Line 5,222: @@
   '''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 4,243: / Line 5,232: @@
 |-
 | <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>
@@ Line 4,249: / Line 5,238: @@
 |-
 | <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>
-| : 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>
-| : 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>
-| : 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:
-{| class="wikitable" style="text-align: left;"
+::{| class="wikitable" style="text-align: left;"
 ! Mode
 ! Description
@@ Line 4,282: / Line 5,271: @@
 The possible values for FM mode multi-group diffusion coefficients are:
-{| class="wikitable" style="text-align: left;"
+::{| class="wikitable" style="text-align: left;"
 ! Mode
 ! Description
@@ Line 4,325: / Line 5,314: @@
 <u>Notes:</u>
-*Photon buffer refers to pre-allocated memory block used to store photon particle data. This memory is needed for putting secondary photons in que, etc..
+*Photon buffer refers to pre-allocated memory block used to store photon particle data.
-*The buffer factor defines the buffer size relative to simulated batch size.
+**This memory is needed for putting secondary photons in que, etc..
-*The event bank refers to pre-allocated memory block used to store history data on particle events. This bank is used only with certain special options, such as [[importance detectors]] and [[track plotter]].
+*The buffer factor <tt>''FAC''</tt> defines the buffer size relative to simulated batch size.
+*The event bank <tt>''BNK''</tt> refers to pre-allocated memory block used to store history data on particle events.
+**This bank is used only with certain special options, such as [[importance detectors]] and [[track plotter]].
 *The default values depend on simulation mode, and there is no need to adjust the values unless the calculation terminates with an error.
 *''Note to developers: event bank is now the same for both neutrons and photons.''
@@ Line 4,334: / Line 5,325: @@
   '''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>
-| : 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>
 *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''
@@ Line 4,353: / Line 5,344: @@
 {|
 | <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>
-*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 "<tt>off</tt>" when the results are not needed (this may speed up the calculation).
 *See separate description of [[Output parameters#Homogenized group constants|output parameters]].
 *Super-imposed group constant generation universes (i.e. not part of the geometry definition) should not be used so that they cover (partly or totally) the same geometry region as some other group constant generation universe (super-imposed or part of the geometry definition), if group constants are generated only on a single geometry level.
+*The spatial homogenization methodology is described in a related paper<ref name="homogenization">Leppänen, J., Pusa, M. and Fridman, E., ''"Overview of methodology for spatial homogenization in the Serpent 2 Monte Carlo code."'', Ann. Nucl. Energy, [https://doi.org/10.1016/j.anucene.2016.06.007 '''96''' (2016) 126-136]</ref>.
 === set gcut ===
@@ Line 4,381: / Line 5,373: @@
   '''set genrate''' ''G'' [ ''MAT'' ]
-Sets normalization to fission neutron generation rate.
+Sets normalization to fission neutron generation rate. Input values:
 {|
 | <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>
-| : material in which the fission neutrons are generated
+| : material in which the fission neutrons are generated. The default is all materials
 |}
@@ Line 4,395: / Line 5,387: @@
 *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.
-*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]].
 *See also Section 5.8 of [http://montecarlo.vtt.fi/download/Serpent_manual.pdf Serpent 1 User Manual].
+=== set gpop ===
+ '''set gpop''' ''MAX_HIS'' ''MAX_POP'' ''C''
+Sets the on-the-fly neutron growing population size algorithm. Input values:
+{|
+| <tt>''MAX_HIS''</tt>
+| : maximum number of histories
+|-
+| <tt>''MAX_POP''</tt>
+| : maximum population size
+|-
+| <tt>''C''</tt>
+| : parameter to control the population growth
+|}
+<u>Notes:</u>
+*The on-the-fly neutron growing population size algorithm is used in criticality calculations to accelerate the fission source convergence.
+*The growing algorithm methodology is described in related paper.  <ref name="gpop">Dufek, J. and Tuttelberg, K.
+''"Monte Carlo criticality calculations accelerated by a growing neutron population."'' Ann. Nucl. Energy [https://doi.org/10.1016/j.anucene.2016.02.015 '''94''' (2016) 16-21].</ref>
 === set gsw ===
@@ Line 4,410: / Line 5,424: @@
 |-
 | <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 4,416: / Line 5,430: @@
 *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.
+*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 ===
@@ Line 4,426: / Line 5,440: @@
 {|
 | <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>".
 |}
 <u>Notes:</u>
 *When invoked, Serpent collects batch-wise data on k<sub>eff</sub>, fission source entropy etc. and produces a [[Description of output files#History output|separate output file]].
-*Setting the history option also invokes the calculation of fission source (Shannon) entropy. The entropy mesh parameters can be adjusted using [[#set entr|set entr]].
+*Setting the history option also invokes the calculation of fission source (Shannon) entropy.
+**The entropy mesh parameters can be adjusted using [[#set entr|set entr]].
 === set ifp ===
@@ Line 4,440: / Line 5,455: @@
 {|
 | <tt>''GEN''</tt>
-| : Number of generations. Default value 15.
+| : Number of generations (default value: 15).
 |}
@@ Line 4,453: / Line 5,468: @@
 |-
 | <tt>''G''</tt>
-| : exponential for adjusting importances
+| : exponential for adjusting importances (default value: -0.5)
 |}
@@ Line 4,477: / Line 5,492: @@
 === 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 4,497: / Line 5,512: @@
 <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.
-*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.
 === 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>
-| : 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>
-| : 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>
-| : 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>
-| : 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 4,524: / Line 5,539: @@
 <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.
 *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 4,530: / Line 5,545: @@
 === 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 (from Serpent 1)
+|-
+| <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''
-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>
 | : Number of nuclides
@@ Line 4,548: / Line 5,614: @@
 <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 ===
@@ Line 4,596: / Line 5,664: @@
 <u>Notes:</u>
 *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 ===
@@ Line 4,609: / Line 5,677: @@
 |-
 | <tt>''KEFF''</tt>
-| : target k-eff for the iteration (default value: 1.0)
+| : target k<sub>eff</sub> for the iteration (default value: 1.0)
 |-
 | <tt>''F<sub>X</sub>''</tt>
@@ Line 4,627: / Line 5,695: @@
 {{:Critical density iteration}}
-*See [[Critical density iteration]].
+*For more information, see the detailed description on the [[Critical density iteration]].
 === set keff ===
   '''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>
-| : 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. Default value is 1.
+| : Fission neutron production scaling factor (default value: 1.0)
 |}
 <u>Notes:</u>
-*Affects both prompt and delayed neutron production from fissions.
+*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.
-*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.
+*The option affects both prompt and delayed neutron production from fissions.
+*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 ===
@@ Line 4,647: / Line 5,716: @@
   '''set lossrate''' ''L'' [ ''MAT'' ]
-Sets normalization to total loss rate.
+Sets normalization to total loss rate. Input values:
 {|
 | <tt>''L''</tt>
-| : number of lost neutrons per second (neutrons/s)
+| : number of lost neutrons per second [in neutrons/s]
 |-
 | <tt>''MAT''</tt>
@@ Line 4,660: / Line 5,729: @@
 *Normalization is needed to relate the Monte Carlo reaction rate estimates to a user-given parameter.
 *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]].
 *See also Section 5.8 of [http://montecarlo.vtt.fi/download/Serpent_manual.pdf Serpent 1 User Manual].
@@ Line 4,672: / Line 5,742: @@
 {|
 | <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 4,688: / Line 5,758: @@
 {|
 | <tt>''MAX''</tt>
-| : maximum number of splits
+| : maximum number of splits (default value: 1.0E4)
 |-
 | <tt>''MIN''</tt>
-| : minimum survival probability in Russian roulette
+| : minimum survival probability in Russian roulette (default value: 1.0E-18)
 |}
@@ Line 4,705: / Line 5,775: @@
 {|
 | <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>
+*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 proc. ANS M&C 2021. Raleigh, NC, USA. October 3-7, 2021</ref>.
 === 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:
@@ Line 4,721: / Line 5,809: @@
 | <tt>''NP''</tt>
 | : 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 4,726: / Line 5,817: @@
 *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 volumes.
+**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 ===
@@ Line 4,739: / Line 5,833: @@
 {|
 | <tt>''UNI''</tt>
-| : universe where spatial homogenization is performed
+| : universe where universe where homogenized microscopic cross sections are generated
 |-
 | <tt>''VOL''</tt>
-| : volume of the homogenized zone
+| : 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>
@@ Line 4,759: / Line 5,853: @@
 <u>Notes:</u>
-*When this option is set, Serpent calculates few-group microscopic cross sections for the listed nuclides and reactions.
+*When this option is set, Serpent calculates homogenized few-group microscopic cross sections for the listed nuclides and reactions.
 *The cross sections are always calculated in the actual spectrum of the problem, never with the [[#set fum|critical spectrum]].
 *The results are printed in a [[Description of output files#Micro depletion output|separate output file]].
 *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 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 calculation requires the [[Defining material volumes|material volumes to be correctly set]].
 *Transmutation reactions to ground and isomeric states can be calculated by adding 'g' and 'm' after the reaction MT (e.g. 102m is the capture cross section corresponding to daughter nuclide being in isomeric state). Reaction rates are calculated to all states by default.
 *Fission reactions corresponding to a specific yield in ENDF can be calculated by adding 1, 2, 3, 4, ... after the reaction MT (e.g. 181 is total fission cross section corresponding to first fission product yield) depending on the fission yield data of the nuclide in the data library. This parameter was different before 2.1.32.
@@ Line 4,771: / Line 5,867: @@
 *The nuclide identifiers are entered as ZAI, not ZA. For example, the ZAI for U-235 is 922350 and the ZAI for Am-242m is 952421.
 *Parameter ''VOL'' was ''VR'' in versions before 2.1.32 (volume ratio of materials included in micro depletion to total homogenized region).
-*Multiple set mdep cards can be given for a single homogenization universe.
+*Multiple set mdep cards can be given for a single homogenized universe.
+*The microscopic cross section calculation methodology is described in a related article.<ref name="mdep">Rintala, A., Valtavirta, V. and Leppänen, J. ''"Microscopic cross section calculation methodology in the Serpent 2 Monte Carlo code."'' Ann. Nucl. Energy [https://doi.org/10.1016/j.anucene.2021.108603 '''164''' (2021) 108603]</ref>
 === set memfrac ===
@@ Line 4,777: / Line 5,874: @@
   '''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>
-| : 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>
-* 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 ===
@@ Line 4,796: / Line 5,893: @@
 {|
 | <tt>''MFPMIN<sub>p</sub>''</tt>
-| : cut-off mean-free-path for photons (cm)
+| : cut-off mean-free-path for photons [in cm]
 |}
@@ Line 4,806: / Line 5,903: @@
 {|
 | <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>
-| : cut-off mean-free-path for photons (cm)
+| : cut-off mean-free-path for photons [in cm]
 |}
@@ Line 4,825: / Line 5,922: @@
 |-
 | <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 4,831: / Line 5,928: @@
   '''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>
-| : 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>
-| : When set to 2, results are averaged over all criticality cycles
+| : When set to "2", results are averaged over all criticality cy