Input syntax manual

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Serpent has no interactive user interface. All communication between the code and the user is handled through one or several input files and various output files.

The format of the input file is unrestricted. The file consists of white-space (space, tab or newline) separated words, containing alphanumeric characters(’a-z’, ’A-Z’, ’0-9’, ’.’, ’-’). If special characters or white spaces need to be used within the word (file names, etc.), the entire string must be enclosed within quotes.

The input file is divided into separate data blocks, denoted as cards. The file is processed one card at a time and there are no restrictions regarding the order in which the cards should be organized. The input cards are listed below. Additional options are followed by key word "set". All input cards and options are case-insensitive (note to developers: make it so). Each input card is delimited by the beginning of the next card. It is hence important that none of the parameter strings used within the card coincide with the card identifiers.

The percent-sign ('%') is used to define a comment line. Anything from this character to the end of the line is omitted when the input file is read. Unlike Serpent 1, hashtag ('#') can no longer be used to mark comment lines in Serpent 2 input. The alternative is to use C-style comment sections beginning with "/*" and ending with "*/". Everything between these delimiters is omitted, regardless of the number of newlines or special characters.

This page will contain the whole input syntax of Serpent 2, with links to more detaied descriptions where needed. For reference see also the Serpent 1 input manual.[1]


Contents

Input cards

NOTE: Serpent command words are in boldface and input parameters entered by the user in CAPITAL ITALIC. Optional input parameters are enclosed in [ square brackets ], and when the number of values is not fixed, the remaining values are marked with three dots (...).

branch (branch definition)

branch NAME [ repm MAT1 MAT2 ]
            [ repu UNI1 UNI2 ]
            [ stp MAT DENS TEMP THERM1 SABL1 SABH1 THERM2 SABL2 SABH2 ... ]
            [ tra TGT TRANS ] 
            [ var VNAME VAL ]

Defines the variations invoked for a branch in the automated burnup sequence. Input values:

NAME  : branch name
MAT1  : name of the replaced material
MAT2  : name of the replacing material
UNI1  : name of the replaced universe
UNI2  : name of the replacing universe
MAT  : name of the material for which density and temperature are adjusted
DENS  : material density after adjustment (positive entries for atomic, negative entries for mass densities, or "sum" to use the sum of the constituent nuclide densities)
TEMP  : material temperature after adjustment, or -1 if no adjustment in temperature
THERMn  : n:th thermal scattering data associated with the material
SABLn  : name of the n:th S(\alpha,\beta) library for temperature below the given value
SABHn  : name of the n:th S(\alpha,\beta) library for temperature above the given value
TGT  : target universe, surface or cell
TRANS  : name of the applied transformation
VNAME  : variable name
VAL  : variable value

Notes:

  • The branch name identifies the branch in the coefficient matrix of the 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.
  • The repm variation can be used to replace one material with another, for example, to change coolant boron concentration.
  • 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 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 last three parameters of the stp entry are provided only if the material has thermal scattering libraries attached to it (see the 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 card.
  • 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 card.
  • For more information, see detailed description on the automated burnup sequence.

cell (cell definition)

cell NAME UNI0 MAT [ SURF1 SURF2 ... ] 

Defines a material cell. Input values:

NAME  : cell name
UNI0  : universe where the cell belongs to
MAT  : material that fills the cell
SURFn  : surface list
cell NAME UNI0 fill UNI1 [ SURF1 SURF2 ... ]

Defines a filled cell. Input values:

NAME  : cell name
UNI0  : universe where the cell belongs to
UNI1  : universe that fills the cell
SURFn  : surface list
cell NAME UNI0 outside [ SURF1 SURF2 ... ]

Defines an outside cell. Input values:

NAME  : cell name
UNI0  : universe where the cell belongs to
SURFn  : surface list

Notes:

  • There are three types of cells: material cells, filled cells and outside cells. Filled cells are identified by providing the key word fill, 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 outside.
  • Void cells can be defined by setting the material name to "void"
  • 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 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, boundary conditions are applied. It is important that the geometry model is non-re-entrant when vacuum boundary conditions are used.
  • Outside cells are allowed only in the root universe. It is important that all space outside the model is defined.
  • The surface list defines the boundaries of the cell by listing the surface names (as provided in the surface card), together with the operator identifiers (nothing for intersection, ":" for union, "-" for complement and "#" for cell complement).
  • For more information, see detailed description on the universe-based geometry type in Serpent.

coef (coefficient matrix definition)

coef NBU [ BU1 BU2 ... ] 
         [ NBR1 BR1,1 BR1,2 ... ]
         [ NBR2 BR2,1 BR2,2 ... ]
         ...

Defines the coefficient matrix for the automated burnup sequence. Input values:

NBU  : number of burnup points
BUn  : burnup steps at which the branches are invoked
NBRm  : number branches in the m:th dimension of the burnup matrix
BRm,i  : name of the i:th branch in the m:th dimension

Notes:

  • The coef card creates a multi-dimensional coefficient matrix (of size NBR1 \times NBR2 \times NBR3 \times ... ). The automated burnup sequence performs a restart for each of the listed burnup points, and loops over the branch combinations defined by the coefficient matrix.
  • The coef card is used together with the branch card
  • For more information, see detailed description on automated burnup sequence.

det (detector definition)

det NAME [ PART]
         [ dr MT MAT] 
         [ dv VOL]
         [ dc CELL]
         [ du UNI]
         [ dm MAT]
         [ dl LAT] 
         [ dx XMIN XMAX NX]
         [ dy YMIN YMAX NY]
         [ dz ZMIN ZMAX NZ]
         [ dn TYPE MIN1 MAX1 N1 MIN2 MAX2 N2 MIN3 MAX3 N3]
         [ dh TYPE X0 Y0 PITCH N1 N2 ZMIN ZMAX NZ]
         [ dumsh UNI NC CELL0 BIN0 CELL1 BIN1 ...]
         [ de EGRID]
         [ di TBIN]
         [ ds SURF DIR]
         [ dir COSX COSY COSZ]
         [ dtl SURF]
         [ df FILE FRACTION]
         [ dt TYPE PARAM]
         [ dhis OPT]
         [ dfl FLAG OPT] 

Detector definition. The first parameter:

PART  : 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.


Detector response (dr):

MT  : response number
MAT  : material name or "VOID" if the material at the collision point is used

Notes:

  • 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 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.
  • Negative response numbers are associated with macroscopic cross sections and special types, and the result is multiplied by material density.
  • The response material in the dr entry must not be confused with the material in the dm entry. The former defines the material for the response function, while the latter determines the volume of integration.
  • By default, Serpent allows a detector to have at most 10,000,000 bins.


Detector volume (dv):

VOL  : 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.


Detector cell (dc):

CELL  : cell name where the detector is scored

Notes:

  • If multiple detector cells are defined, the results are correspondingly divided into multiple bins.


Detector universe (du):

UNI  : universe name where the detector is scored

Notes:

  • If multiple detector universes are defined, the results are correspondingly divided into multiple bins.


Detector material (dm):

MAT  : material name where the detector is scored

Notes:

  • If multiple detector materials are defined, the results are correspondingly divided into multiple bins.
  • The material entry defines the volume of integration, which must not be confused with the response material in the dr entry.


Detector lattice (dl):

LAT  : lattice name where the detector is scored

Notes:

  • The lattice detector automatically divides the results into multiple bins corresponding to the lattice cells.


Cartesian mesh detector (dx, dy and dz):

XMIN  : minimum x-coordinate of the detector mesh
XMAX  : maximum x-coordinate of the detector mesh
NX  : number of x-bins
YMIN  : minimum y-coordinate of the detector mesh
YMAX  : maximum y-coordinate of the detector mesh
NY  : number of y-bins
ZMIN  : minimum z-coordinate of the detector mesh
ZMAX  : maximum z-coordinate of the detector mesh
NZ  : number of z-bins

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.


Curvilinear mesh detector (dn):

TYPE  : Type of curvilinear mesh - 1 = cylindrical (dimensions r, θ, z), 2 = spherical (dimensions r, θ, φ)
MINn  : Minimum value of coordinate n for the mesh division.
MAXn  : Maximum value of coordinate n for the mesh division.
Nn  : Number of bins in the n coordinate direction (the 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).


Hexagonal mesh detector (dh):

TYPE  : Type of hexagonal mesh (2 = flat face perpendicular to x-axis, 3 = flat face perpendicular to y-axis)
X0, Y0  : coordinates of mesh center
PITCH  : mesh pitch
N1, N1  : mesh size
ZMIN  : minimum z-coordinate of the detector mesh
ZMAX  : maximum z-coordinate of the detector mesh
NZ  : number of z-bins

Notes:

  • All parameters must be provided, even for a two-dimensional hexagonal meshes.


Unstructured mesh detector (dumsh):

UNI  : universe of the unstructured mesh based geometry
NC  : number of mesh cell bins included in the output
CELLn, BINn  : cell-bin index pairs defining the mapping

Notes:

  • The polyhedral cells in 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.


Detector energy binning (de):

EGRID  : energy grid name

Notes:

  • The results are divided into multiple energy bins based on the grid structure.
  • Energy grid structures are defined using the ene card.


Detector time binning (di):

TBIN  : time bin structure name

Notes:

  • The results are divided into multiple time bins.
  • Time bin structures are defined using the tme card.
  • Time bin division may require adjusting the average collision distance (set cfe option) to achieve sufficient statistical accuracy.


Surface current detector (ds):

SURF  : surface name
DIR  : direction with respect to surface normal (-1 = inward, 1 = outward, 0 = total)

Notes:

  • With this option the detector calculates the particle current through a give surface.
  • Responses are not allowed.
  • 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.


Detector direction vector (dir):

COSY  : component of the direction vector parallel to x-axis
COSY  : component of the direction vector parallel to y-axis
COSZ  : component of the direction vector parallel to z-axis

Notes:

  • This option multiplies the detector scores with the scalar product between the particle direction of motion and the given direction vector.


Super-imposed track-length detector (dtl):

SURF  : surface inside which the detector is scored

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 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.


Detector file (df):

FILE  : file name where the scored points are written
FRAC  : 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 sf entry of source cards.


Special types (dt):

TYPE  : special type (see below)
PARAM  : additional parameters

The types are:

-1 = cumulative spectrum
-2 = division by energy width
-3 = division by lethargy width
2 = multiply result with another detector defined by PARAM
3 = divide result with another detector defined by PARAM

Notes:

  • Types -1, -2 and -3 are used with energy binning.
  • 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 (dhis):

OPT  : option to collect histories (0 = no, 1 = yes)

Notes:

  • When this option is set, the batch-wise results are kept and printed in the history output file.
  • Note to developers: statistical tests should be documented


Detector flagging (dfl):

FLAG  : flag number (between 1 and 64)
OPT  : flagging option (0 = reset if scored, 1 = set if scored, -2/2 score if set -3/3 score if not set)

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).

div (divisor definition)

div MAT [ sep LVL ] 
        [ subx NX XMIN XMAX ]
        [ suby NY YMIN YMAX ]
        [ subz NZ ZMIN ZMAX ]
        [ subr NR RMIN RMAX ] 
        [ subs NS S0 ]

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

MAT  : name of the divided material
LVL  : geometry level at which the cell-wise division takes place (0 = no division, 1 = last level, 2 = 2nd last level, etc.)
NX  : number of x-zones
XMIN  : minimum x-coordinate (cm)
XMAX  : maximum x-coordinate (cm)
NY  : number of y-zones
YMIN  : minimum y-coordinate (cm)
YMAX  : maximum y-coordinate (cm)
NZ  : number of z-zones
ZMIN  : minimum z-coordinate (cm)
ZMAX  : maximum z-coordinate (cm)
NR  : number of radial zones
RMIN  : minimum radial coordinate (cm)
RMAX  : maximum radial coordinate (cm)
NZ  : number of angular sectors
S0  : zero position of angular division (degrees)

Notes:

  • 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.
  • Volumes of the divided materials must be set manually (see detailed description on 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).
  • For more information, see detailed description on automated depletion zone division.

ene (energy grid definition)

ene NAME 1 E0 E1 ...
ene NAME 2 N Emin Emax
ene NAME 3 N Emin Emax
ene NAME 4 GRID

Defines an energy grid structure. Input values:

NAME : energy grid name
Ei : bin boundaries (type 1)
N : number of equi-width bins (types 2 ad 3)
Emin : minimum energy (types 2 and 3)
Emax : maximum energy (types 2 and 3)
GRID : name of the pre-defined grid (type 4)

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).
  • Energy grid structures are used for several purposes, most commonly with detectors.
  • See separate description of pre-defined energy group structures.

ftrans (fill transformation)

See transformations.

fun (function definition)

fun NAME TYPE [ ... ]

Defines a function that can be used with detector responses. Input values:

NAME : function name
TYPE : function type (currently only supported type is 1 = point-wise tabular data)

The syntax for type 1 is:

fun NAME 1 INTT X1 F1 X2 F2 ...

where:

INTT : is the interpolation type (1 = histogram, 2 = lin-lin, 3 = lin-log, 4 = log-lin, 5 = log-log)
X1 F1 ... : are the tabulated variable-value pairs

Notes:

  • The defined function is linked to detector response using response number -100 (syntax: dt -100 NAME).

ifc (interface file)

ifc FILE

Links a multi-physics interface file to be used with the current input. Input values:

FILE : path to the multi-physics interface file

Notes:

include (read another input file)

include FILE

Reads another input file. Input values:

FILE  : name of the input file

Notes:

  • The include card can be used to simplify the structure of complicated inputs.
  • 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 and options, it cannot be used to read the values of another card.

iter nuc (critical density iteration)

set iter nuc CYCLES KEFF NZAI ZAI1 ZAI2 ... ZAINZAI [ NMAT MAT1 MAT2 ... MAXNMAT ]
CYCLES  : number of additional inactive cycles to run for the convergence of the iteration
KEFF  : target k-eff for the iteration
NZAI  : number of different nuclides (ZAI) included in the iteration
ZAIi  : the ZAI (e.g. 50100 for boron 10 ground state) of the nuclide to be included in the iteration
NMAT  : number of different materials included in the iteration (optional parameter)
MATi  : the name of the material to be included in the iteration

Notes:

  • If a list of materials is not given, all materials that contain the included nuclides are included in the iteration.
  • The initial density of the nuclides to be iterated should be larger than zero.
  • The critical density iteration only works for nuclides that have a reactivity effect mainly through neutron absorption. Specifically, critical densities of fissile, moderating or reflecting nuclides cannot be reliably iterated using this card.

lat (regular lattice definition)

See Section 3.6 of Serpent 1 User Manual.

mat (material definition)

See Chapter 4 of Serpent 1 User Manual.

mesh (mesh plot definition)

mesh ORI XPIX YPIX [ SYM MIN1 MAX1 MIN2 MAX2 MIN3 MAX3 ]
mesh 8 CMAP DET ORI XPIX YPIX [ SYM MIN1 MAX1 MIN2 MAX2 MIN3 MAX3 ]

Produces a png-format mesh plot of various results. Input values:

ORI  : orientation with respect to coordinate axes
XPIX  : horizontal image size in pixels
YPIX  : vertical image size in pixels
SYM  : symmetry option (not used in Serpent 2)
MINn MAXn  : boundaries of the plotted region
CMAP  : color map used for plotting detector scores (positive entry for linear, negative for log-scale)
DET  : detector name

Notes:

  • The first format produces a mesh plot where fission rate and thermal flux distribution are plotted using hot and cold color schemes, respectively. This type of mesh plot is convenient for illustrating the neutronics of thermal systems. The orientation parameter defines the coordinate axis perpendicular to the plot plane: 1 - x-axis (projection on yz-plane); 2 - y-axis (projection on xz-plane); 3 - z-axis (projection on xy-plane).
  • The second format can be used for plotting scores that contribute to a detector. The additional input parameters are the detector name and the color map used in the plot.
  • The color maps are: 1 - hot; 2 - cold; 4 - jet; 5 - black and white; 6 - hsv; 7 - spring; 8 - summer; 9 - autumn; 10 - winter; 11 - green-purple; 12 - purple-orange; 13 - blue-red. Many of these correspond to what is used in Matlab. Logarithmic scale is used if the number is given with a minus sign.
  • Detector scores are collected in the mesh cells (see the detector card and the list of ENDF reaction MT's and special reaction types for more information). The distribution is scaled according to the minimum and maximum values.
  • 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.
  • 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 n:th mesh-card is written in file [input]_mesh[n].png.

mix (mixture definition)

mix NAME
MAT1 F1
MAT2 F2
...

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

MATn  : material name
Fn  : material fraction (positive values for atomic, negative values for mass fractions)

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 -mix command line option.

nest (nested universe definition)

nest U TYPE
  [ MAT1 R1 ] 
  [ MAT2 R2 ] 
  ...
  [ MATN ]
nest U 
  [ MAT1 TYPE1 PARAM11 PARAM12 ... ] 
  [ MAT2 TYPE2 PARAM21 PARAM22 ... ] 
   ...
  [ MATN ] 

Defines a universe consisting of nested regions. Input values:

U  : universe name
TYPE  : nested surface type (single surface for all regions)
MAT1 ... MATN  : material regions
R1 ... RN-1  : outer radii
TYPE1 ... TYPEN-1  : nested surface type (different surfaces for each region)
PARAMnm ...  : surface parameters

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 material entries can be replaced by fill U0, 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 (cylinders, spheres and some regular prisms). The pin and 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.

particle (particle geometry definition)

particle U
       [ MAT1 R1 ] 
       [ MAT2 R2 ] 
       ...
       [ MATN ]

Defines a particle universe. Input values:

U  : universe name
MAT1 ... MATN  : material regions
R1 ... RN-1  : outer radii

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 material entries can be replaced by fill U0, in which case the region is filled by another universe.
  • Most typically used for defining TRISO fuel particles.
  • The particle card is special case of a nested universe type.
  • See also description of explicit stochastic geometry type.

pin (pin geometry definition)

pin U
  [ MAT1 R1 ] 
  [ MAT2 R2 ] 
  ...
  [ MATN ]

Defines a pin universe. Input values:

U  : universe name
MAT1 ... MATN  : material regions
R1 ... RN-1  : outer radii

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 material entries can be replaced by fill U0, 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.
  • The pin card is special case of a nested universe type.

plot (geometry plot definition)

plot TYPE XPIX YPIX [ POS MIN1 MAX1 MIN2 MAX2 ]
plot TYPE Fmin Fmax E XPIX YPIX [ POS MIN1 MAX1 MIN2 MAX2 ]

Produces a png-format geometry plot. Input values:

TYPE  : defines the plot type (orientation and plotting of boundaries)
XPIX  : horizontal image size in pixels
YPIX  : vertical image size in pixels
POS  : position of plot plane
MIN1  : minimum horizontal coordinate of plotted region
MAX1  : maximum horizontal coordinate of plotted region
MIN2  : minimum vertical coordinate of plotted region
MAX2  : maximum vertical coordinate of plotted region
Fmin  : minimum importance for importance map plots
Fmax  : maximum importance for importance map plots
E  : particle energy for importance map plots

Notes:

  • The type parameter consists of one or two concatenated values ('AB'):
    1. The first value ('A') defines the plot plane (1 = yz, 2 = xz, 3 = xy).
    2. 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.
  • Importance maps read using the 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). The input parameters then also include the minimum and maximum importance and the particle energy.
  • Each material plotted with different color. The colors are sampled randomly, unless defined using the rgb entry in the material card.
  • Void is plotted in black and special colors are used to plot geometry errors (red = overlap, green = undefined region).
  • 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 define the boundaries of the plotted region (e.g. minimum and maximum x- and y-coordinates for xy-type plot). If these coordinates are not provided, the plot is extended to the maximum dimensions of the geometry.
  • The relative dimensions of image size in pixels should match that of the plotted region. Otherwise the image gets distorted.
  • Geometry plotter requires compiling the source code with GD Graphics libraries.
  • Command line parameter '-plot' stops the execution after the geometry plots are prodused. Option '-qp' invokes a quick plot mode, which does not check for overlaps.
  • See also detailed description on geometry plotter.
  • Geometry plot produced by the n:th plot-card is written in file [input]_geom[n].png.
  • Note to developers: particle type should be included as an input parameter in importance map plots.

solid (irregular 3D geometry definition)

solid 1 UNI BGUNI
MESH_SPLIT MESH_DIM SZ1 SZ2 ... SZMESH_DIM
POINTS_FILE
FACES_FILE
OWNER_FILE
NEIGHBOUR_FILE
MATERIALS_FILE

Creates an unstructured mesh-based geometry universe. Input values are:

UNI  : universe name for the irregular geometry
BGUNI  : name of the background universe filling all undefined space
MESH_SPLIT  : Splitting criterion for the adaptive search mesh (maximum number of geometry cells in search mesh cell)
MESH_DIM  : number of levels in the adaptive search mesh
SZi  : Size of the search mesh at level i
POINTS_FILE  : Path to the unstructured mesh points file
FACES_FILE  : Path to the unstructured mesh faces file
OWNER_FILE  : Path to the unstructured mesh owner file
NEIGHBOUR_FILE  : Path to the unstructured mesh neighbour file
MATERIALS_FILE  : Path to the unstructured mesh materials file


solid 2 UNI BGUNI
MESH_SPLIT MESH_DIM SZ1 SZ2 ... SZMESH_DIM
MODE R0
body BODY1 CELL1 MAT1
file BODY1 FILE1 SCALE1 X1 Y1 Z1
file BODY1 FILE2 SCALE2 X2 Y2 Z2
...
body BODY2 CELL2 MAT2
file BODY2 FILE3 SCALE3 X3 Y3 Z3
file BODY2 FILE4 SCALE4 X4 Y4 Z4
...

Creates an STL-based geometry universe. Input values are:

UNI  : universe name for the irregular geometry
BGUNI  : name of the background universe filling all undefined space
MESH_SPLIT  : Splitting criterion for the adaptive search mesh (maximum number of geometry cells in search mesh cell)
MESH_DIM  : number of levels in the adaptive search mesh
SZi  : Size of the search mesh at level i
MODE  : Mode for handling the triangulated geometry (1 = "fast", 2 = "safe").
R0  : Radius inside which two points of the STL-geometry are joined into one.
BODYi  : Name of solid body i
CELLi  : Name of geometry cell i linked with body i
MATi  : Material filling cell i
FILEi  : Path to a file containing an STL solid model, multiple files can be linked to one body
SCALEi  : Scaling factor for the STL model in FILEi
Xi  : Shift in x-direction to the STL model in FILEi
Yi  : Shift in y-direction to the STL model in FILEi
Zi  : Shift in z-direction to the STL model in FILEi
solid 3 
INTERFACE_FILE

Creates an unstructured mesh-based geometry universe with unstructured mesh-based temperature and/or density distributions. Input values are:

INTERFACE_FILE  : Path to the interface file containing the rest of the parameters

src (source definition)

src NAME [ PART]
         [ sw WGT] 
         [ sc CELL]
         [ sm MAT]
         [ sp X Y Z] 
         [ sx XMIN XMAX]
         [ sy YMIN YMAX]
         [ sz ZMIN ZMAX]
         [ ss SURF]
         [ sd U V W]   
         [ se E]
         [ sb N E1 WGT1 E2 WGT2 ...]
         [ sr NUC MT]
         [ st TMIN TMIN]
         [ sf FILE TYPE]
         [ si N P1 P2 ...]
         [ sg MAT MODE] 

Source definition. The first parameter:

PART  : 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.


Source weight (sw):

WGT  : relative source weight

Notes:

  • When multiple sources are defined, each definition is sampled with equal probability. This probability can be changed by assigning different weights for each source.
  • The weights are automatically normalized before the calculation is started.


Source cell (sc):

CELL  : cell inside which the source points are sampled

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 around the cell using the sx, sy and sz options.
  • If no spatial distribution is defined, particles are sampled uniformly over the geometry.


Source material (sm):

MAT  : material inside which the source points are sampled

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 around the region using the sx, sy and sz options.
  • If no spatial distribution is defined, particles are sampled uniformly over the geometry.


Source point (sp):

X, Y, Z,  : coordinates of the source point

Notes:

  • Setting a point source is one of the options that can be applied to define the spatial distribution of source particles.
  • If no spatial distribution is defined, particles are sampled uniformly over the geometry.


Source boundaries (sx, sy and sz):

XMIN, XMAX  : Boundaries on X-axis
YMIN, YMAX  : Boundaries on Y-axis
ZMIN, ZMAX  : Boundaries on Z-axis

Notes:

  • Source boundaries are used to define a bounding box inside which the source particles are sampled.
  • Can be used in combination with cell and material sources to increase the sampling efficiency.
  • If no bounding box is defined, particles are sampled uniformly over the geometry.


Surface source (ss):

SURF  : surface on which the source particles are sampled

Notes:

  • The surface source is currently limited to infinite vertical cylinder (cyl) and sphere (sph) surface types.
  • Particles are started in the direction of the inward surface normal.


Source direction (sd):

U, V, W,  : direction vector of source particles

Notes:

  • The source direction option can be set to define a unidirectional source.
  • If no directional dependence is defined, the direction of source particles is sampled isotropically.


Source energy (se):

E  : energy of source particles

Notes:

  • The source energy option can be used to define a monoenergetic source.
  • The default energy of neutrons and photons is 1 MeV.
  • This option can also be used together with the source reaction option (sr).


Source energy bins (sb):

N  : number of bins
En  : upper boundary of the energy bin
WGTn  : weight of the energy bin

Notes:

  • This option allows defining an arbitrary source spectrum in the form of tabular data.
  • The bins are entered in the order of ascending energy, and weight of the first bin must be set to zero.
  • The interpolation between the bins is either linear (positive weight entries) or logarithmic (negative weight entries).


Source reaction (sr):

NUC  : nuclide name
MT  : reaction number

Notes:

  • The source reaction determines a distribution function for source energy (for example, 235U fission spectrum can be defined as: sr 92235.09c 18).
  • The reaction numbers are ENDF reaction MT's, and the data is obtained from standard cross section libraries.
  • Applies to neutrons only.
  • When the source energy parameter (se) is defined, the value is used as the energy of the incoming neutrons.


Source time (st):

TMIN, TMAX  : time boundaries

Notes:

  • This parameter defines a time interval for the sampled source particles. The starting time is sampled uniformly between the given minimum and maximum.
  • All source particles are started at time zero by default.


Source file (sf):

FILE  : file path to source file
TYPE  : file type (-1 = binary, 1 = ASCII)

Notes:

  • Source files allow defining arbitrary distributions by reading the particle coordinates, direction, energy, weight and time from a file.
  • Source files can be produced using the df entry of detector cards or the set csw option.


User-defined source routine (si):

N  : number of parameters
Pn  : parameters passed as arguments into the subroutine

Notes:

  • This option allows defining an arbitrary source distributions with a user-defined subroutine.
  • The source parameters are passed as arguments into the subroutine, together, with sampled position, direction energy, weight and time.
  • For complete description see source file "usersrc.c".
  • The subroutine may be overwritten with the blank template file when installing updates.


Radioactive decay source (sg):

MAT  : material name or -1
MODE  : sampling mode (1 = analog, 2 = implicit)

Notes:

  • Radioactive decay source combines material compositions to decay data read from ENDF format 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 and 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.
  • In version 2.1.28 the source is limited to photon line spectra.
  • The calculation produces an additional output file [input]_gsrc.m that contains the source spectra.
  • See practical example for more information.

strans (surface transformation)

See transformations.

surf (surface definition)

surf NAME TYPE [ PARAM1 PARAM2 ... ]

Defines a surface. Input values:

NAME  : is the surface name
TYPE  : is the surface type
PARAMn  : are the surface parameters

Notes:

therm and thermstoch (thermal scattering)

therm NAME LIB 
therm NAME TEMP LIB1 LIB2 
therm NAME 0 LIB1 LIB2 LIB3 ... 
thermstoch NAME TEMP LIB1 LIB2 

Defines thermal scattering data that can be linked to nuclides using input entry moder in the 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.

Input values:

NAME  : name of the thermal scattering data
LIBi  : thermal scattering data identifiers as defined in the directory file (acelib)
TEMP  : temperature to which the thermal scattering data is interpolated

Notes:

  • When using on-the-fly interpolation of thermal scattering data, LIBi must cover the whole temperature range in which the materials appear in the geometry. I.e. extrapolation of the data is not supported.
  • Thermal scattering data is interpolated using the methodology of makxsf code
  • The interpolation can be performed using the stochastic mixing approach with thermstoch. This interpolation mode is not available for on-the-fly interpolation.

tme (time binning definition)

tme NAME 1 LIM1 LIM2 ...
tme NAME 2 NB Tmin Tmax
tme NAME 3 NB Tmin Tmax

Defines a time binning structure. The second entry sets the binning type (1 = arbitrary, 2 = uniform, 3 = log-uniform). Remaining values:

NAME  : name of the time binning
NB  : number of bins
LIMn  : time bin boundaries in arbitrary binning
Tmin  : minimum time boundary in uniform or log-uniform binning
Tmax  : maximum time boundary in uniform or log-uniform binning

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 detectors and dynamic simulation mode.

trans (transformations)

trans TYPE UNIT LVL
trans TYPE UNIT X Y Z
trans TYPE UNIT X Y Z θx θy θz ORD
trans TYPE UNIT X Y Z α1 α2 α3 α4 α5 α6 α7 α8 α9 ORD
trans TYPE UNIT rot X0 Y0 Z0 I J K β

Defines surface, universe or fill transformation. Input values:

TYPE  : type of transformation (S = surface transformation, F = fill transformation, U = universe transformation)
UNIT  : surface, cell or universe name to which the transformation is applied
LVL  : level number in universe level transformation
X,Y,Z  : translation vector
θx θy θz  : rotation angles with respect to x-, y- and z-axes (in degrees)
α1 ... α9  : coefficients of the rotation matrix
ORD  : order in which translations and rotations are applied (1 = translations first, 2 = rotations first)
X0,Y0,Z0  : Origin of vector defining rotation axis.
I,J,K  : Components of vector defining rotation axis.
β  : 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.

Notes:

  • Fill transformation is applied in the universe filling the given cell.
  • Level transformation is a special type of universe transformation, in which the coordinates in the given universe are obtained relative to geometry level LVL.
  • By default rotations are applied before translations, and the order can be switched using the ORD parameter.
  • The default order of differs from MCNP.
  • 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 \vec{r'} = \bold{A} \vec{r} where \vec{r} and \vec{r'} are the position vectors before and after the operation and coefficients α1 ... α9 define the 3 by 3 matrix \bold{A}.
  • To preserve backwards compatibility, input parameters "strans", "utrans" and "ftrans" without the following type identifier are also accepted for defining surface, universe and fill transformations, respectively. To preserve compatibility with Serpent 1, parameter "trans" without type identifier defines a universe transformation.

transv and transa (velocity and acceleration transformations)

transv TYPE UNIT [tlim T0 T1 TTYPE] VX VY VZ
transa TYPE UNIT [tlim T0 T1 TTYPE] AX AY AZ


Defines surface, universe or fill transformation. Input values:

TYPE  : type of transformation (S = surface transformation, F = fill transformation, U = universe transformation)
UNIT  : surface, cell or universe name to which the transformation is applied
T0  : beginning time of the transformation
T1  : end time of the transformation
TTYPE  : transformation type after end time (1 = movement stops, 2 = transformation removed, 3 = initial acceleration and velocity removed, but velocity accumulated due to acceleration remains)
VX,VY,VZ  : Initial velocity vector
AX,AY,AZ  : Initial acceleration vector

Notes:

utrans (universe transformation)

See transformations.

wwgen (response matrix based importance map solver)

wwgen NAME LIM NI F ERG MSH 
     MIN1 MAX1 SZ1
     MIN2 MAX2 SZ2
     MIN3 MAX3 SZ3
     DET1 w1 [ DET2 w2 ... ]

Defines the parameters for importance map calculation. Input values:

NAME  : a unique name to identify the calculation
LIM  : convergence criterion (typical value 1E-12)
NI  : maximum number of iterations
F  : normalization factor / weighing flag
ERG  : energy group structure (or -1 if no energy dependence is included)
MSH  : mesh type (1 = Cartesian, 2 = Cylindrical)
MINn  : minimum mesh boundary (n:th coordinate)
MAXn  : maximum mesh boundary (n:th coordinate)
SZn  : number of mesh cells (n:th coordinate)
DETi  : detectors used as target response functions
wi  : weight factors for detector scores

Notes:

  • Multiple energy groups not yet supported (input value must be set to -1).
  • The importances are normalized in such way that the maximum importance in cells with source points is set to F.
  • If the the normalization factor is given with a negative sign, inverse detector scores are used in the importance calculation.
  • The coordinate axes 1, 2 and 3 in Cartesian mesh refer to (x,y,z) and in cylindrical mesh to (r,θ,z), with θ given in degrees.
  • The mesh must be defined slightly larger than the geometry (the mesh boundaries should not coincide with the geometry boundaries).
  • Source points located on mesh cell boundaries cause fatal errors.
  • Works only with external source simulations.
  • May not work if source distribution is biased with weight.
  • The importance mesh is printed in file [input].wwd.
  • Importance (weight window) meshes are read using the wwin card.
  • This capability is still very much under development. The input syntax may be revised at some point.

wwin (weight window mesh definition)

wwin NAME 
   [ wf FILE FMT ]
   [ wn F X Y Z E ] 
   [ wx C G ] 
        

Defines a weight window mesh for variance reduction. Input values:

NAME  : a unique name to identify the mesh
FILE  : file path and name of the importance mesh file
FMT  : file format (1 = mesh produced by Serpent importance map generator, 2 = MCNP weight window mesh file)
F  : importance for renormalization
X,Y,Z  : coordinates of point used for renormalization
E  : energy used for renormalization
C  : constant multiplier for adjusting importances
G  : exponential for adjusting importances

Notes:

  • By default the importance map is read from the mesh file and used as-is, the additional options are provided for adjustments.
  • The importances can be renormalized using the wn option by fixing the value at a given position and energy.
  • The importances can be adjusted using the wx option by constant multiplier C and exponential factor G such that F' = CF^G.
  • Currently the MCNP format only supports regular mesh type (uniform spacing).
  • Only works in external source simulation mode.
  • No source biasing is currently applied.
  • Importance (weight window) meshes can be generated by running the response matrix based solver.
  • Importance maps can be visualized using the geometry plotter.
  • This capability is still very much under development. The input syntax may be revised at some point.

Input options

Input options are used to set various calculation parameters that are not included in the main input cars. Each option is identified by key word "set". Optional values are enclosed within square brackets.

set acelib

set acelib LIB1 [ LIB2 LIB3 ... ]

Sets the cross section directory file paths. Input values:

LIBn  : file paths to directory files

Notes:

  • 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.

set adf

set adf UNI SURF SYM 

Sets parameters for the calculation of assembly discontinuity factors. Input values:

UNI  : universe where spatial homogenization is performed
SURF  : surface enclosing the universe
SYM  : symmetry option (see separate list)

Notes:

  • The surface enclosing the universe can be super-imposed (i.e. not part of the geometry definition), but it must enclose the entire universe.
  • When the universe is surrounded by zero net-current (reflective) boundary conditions, the ADF's 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 built-in diffusion flux solver.
  • Calculation parameters for the diffusion flux solver can be set using the set dfsol option.
  • Calculation of ADF's is currently allowed only for planes and infinite square and hexagonal prisms.
  • Symmetry options are used to average out the statistical variation in the ADF's, which might otherwise lead to systematic errors in core calculations. It is important that the options are used only when the geometry has the corresponding symmetry.
  • See separate description of output parameters in the [input]_res.m.
  • The ADF's can be printed to the group constant output with set coefpara.

set alb

set alb UNI SURF DIR 

Sets parameters for calculating albedos. Input values:

UNI  : universe where spatial homogenization is performed
SURF  : surface for which the albedos are calculated
DIR  : current direction (-1 = inward, 1 = outward)

Notes:

  • When this option is set, Serpent calculates both total albedos (ratio of currents) and partial albedos (response matrix).
  • The current direction is given relative to the surface normal vectors
  • The universe is needed only for labelling the results in the output files.
  • See separate description of output parameters.

set arr

set arr MODEN [ MODEG ]

Sets analog reaction rate calculation on or off. Input values:

MODEN  : mode for neutrons (0 = no reactions included, 1 = include only reactions that affect neutron balance, 2 = include all reactions)
MODEG  : mode for photons (0 = no reactions included, 1 = include all reactions)

Notes:

  • Analog reaction rates are calculated by counting sampled events and printed in a separate output file.
  • See detailed description on the reaction rate output file.

set bc

set bc MODE

Sets the boundary conditions for all outer boundaries of the geometry. Input values:

MODE  : boundary type (1 = vacuum, 2 = reflective, 3 = periodic)
set bc MODE ALB

Sets the boundary conditions with albedo for all outer boundaries of the geometry. Input values:

MODE  : boundary type (1 = vacuum, 2 = reflective, 3 = periodic)
ALB  : albedo
set bc MODEX MODEY MODEZ

Sets the boundary conditions separately for x-, y- and z-directions. Input values:

MODEX  : boundary type in x-direction (1 = vacuum, 2 = reflective, 3 = periodic)
MODEY  : boundary type in y-direction (1 = vacuum, 2 = reflective, 3 = periodic)
MODEZ  : boundary type in z-direction (1 = vacuum, 2 = reflective, 3 = periodic)
set bc MODEX MODEY MODEZ ALB

Sets the boundary conditions with albedo separately for x-, y- and z-directions. Input values:

MODEX  : boundary type in x-direction (1 = vacuum, 2 = reflective, 3 = periodic)
MODEY  : boundary type in y-direction (1 = vacuum, 2 = reflective, 3 = periodic)
MODEZ  : boundary type in z-direction (1 = vacuum, 2 = reflective, 3 = periodic)
ALB  : albedo

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.
  • Albedo boundary conditions are invoked by multiplying the particle weight with factor ALB each time a reflective or periodic boundary is hit.
  • Repeated boundary conditions (reflective or periodic) are based on universe transformations, which limits outer boundary to surfaces that form regular lattices (square and hexagonal prisms, rectangles, cubes and cuboids).
  • Repeated boundary conditions are applied on the first surface of outside cells (see definition of outside cells in the cell card)
  • For symmetry purposes Serpent provides the universe symmetry option.
  • For more information, see detailed description on boundary conditions.

set blockdt

set blockdt MAT1 MAT2 ...

Defines the list of materials where delta-tracking is never used. Input values:

MATn  : material names

Notes:

  • This option is used to override selection of tracking mode based on the probability threshold (see set dt) in individual materials.
  • Use of delta-tracking can be forced in individual materials using set forcedt.
  • For more information on tracking modes, see the detailed description on delta- and surface-tracking.
  • Note to developers: should have different lists for neutrons and photons?

set bralib

set bralib LIB1 [ LIB2 LIB3 ... ]

Sets isomeric branching data library file paths. Input values:

LIBn  : library file paths

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 constant branching ratios by default. The default values can be overridden using the 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.

set ccmaxiter

set ccmaxiter NITER

Sets the maximum number of coupled calculation iterations. Input values:

NITER  : number of iterations.

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) has been simulated.
  • See Coupled multi-physics calculations for further information.

set ccmaxpop

set ccmaxpop CPOP

Sets the maximum total live population to simulate in a coupled calculation. Input values:

CPOP  : total active population to simulate.

Notes:

  • Default maximum population is INFTY/1e6.
  • The iteration is stopped when either the maximum number of iterations (set with set ccmaxiter) or the maximum active neutron population has been simulated.
  • Only the population simulated during active cycles is included in this amount.
  • This is mostly useful if the neutron population per iteration is not constant.
  • See Coupled multi-physics calculations for further information.

set cfe

set cfe LN [ TN LG TG ]

Defines the minimum mean distance for scoring the collision flux estimator (CFE) for photons and neutrons. Input values:

LN  : minimum mean distance for scoring the CFE for neutrons
TN  : minimum mean time interval for scoring the CFE for neutrons
LG  : minimum mean distance for scoring the CFE for photons
TG  : minimum mean time interval for scoring the CFE for photons

Notes:

  • The use of delta-tracking necessitates the use of CFE for scoring the integral reaction rates. The scoring is based on both real and virtual collision to improve the statistics in low density regions (and short time intervals).
  • 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.
  • 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.
  • The collision flux estimator in Serpent is described in an article in Annals of Nuclear energy from 2017.[2]
  • In version 2.1.27 and earlier the name of this input option was "set minxs".

set coefpara

set coefpara FMT [ PARAM1 PARAM2 ... ]

Defines the parameters included in the separate group constant output file. Input values:

FMT  : output format, currently used for including or excluding statistical errors (0 = not included, 1 = included)
PARAMn  : list of parameters or detectors included in the file

Notes:

set comfile

set comfile INFILE OUTFILE

Defines the communication files used in the file-based coupled calculation communications. Input values:

INFILE  : Path to inwards communication file (signals to Serpent).
OUTFILE  : Path to outwards communication file (signals from Serpent).

Notes:

  • Setting up a communication mode will enable the coupled calculation mode.
  • For more information see: External coupling

set confi

set confi OPT

Sets confidentiality flag on or off. Input values:

OPT  : option to set confidentiality flag on (1/yes) or off (0/no)

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 calculation title and the value of variable CONFIDENTIAL_DATA in the [input]_res.m output file is set to 1.

set csw

set csw FILE

Writes source points in criticality source simulation into a file. Input values:

FILE  : file name where the source points are written

Notes:

  • Only source points from active cycles are included.

set declib

set declib LIB1 [ LIB2 LIB3 ... ]

Sets the decay data library file paths. Input values:

LIBn  : library file paths

Notes:

  • Decay libraries are standard ENDF format 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.

set delnu

set delnu OPT

Sets delayed neutron emission on or off. Input values:

OPT  : option to switch delayed neutron emission on (1/yes) or off (0/no)

Notes:

  • Delayed neutron emission is on by default in neutron criticality source and off by default in external source simulation mode.
  • See separate description of physics options in Serpent for differences to other codes.

set depout

set depout MODE

Controls which burnable material compositions are printed into the _dep.m output file in case of divided materials. Input values:

MODE  : value indicating, which materials to output to the _dep.m file (1 = only partials, 2 = only parents, 3 = both)

Notes:

  • Parent materials refer to materials defined by mat cards, and partials to depletion zones created automatically using the div card.
  • Default is 2 (only parents).

set dfsol

set dfsol MODE [ DC NP ]

Options for homogeneous diffusion flux solver. Input values:

MODE  : boundary conditions for solver (1 = include net currents at boundary surfaces and corners, 2 = include only surface currents)
DC  : type of diffusion coefficient used in the calculation (1 = INF_DIFFCOEF, 2 = TRC_DIFFCOEF)
NP  : number of points for trapezoidal integration for homogeneous flux

Notes:

  • This input option is used to control how the deterministic diffusion flux solver used to obtain assembly discontinuity factors (set adf) and pin power distributions (set ppw) is run
  • Default mode is 1 (include both surfaces and corners in solution)
  • Default diffusion coefficient is INF_DIFFCOEF, option 2 requires the set trc option.
  • Default number of points for trapezoidal integration is 100
  • See also separate description of the built-in diffusion flux solver.
  • The format was revised in update 2.1.27 (DC option was added between MODE and NP).

set dynsrc

set dynsrc PATH [ MODE ]

Links previously generated steady state source distributions to be used in a transient simulation with delayed neutron emission.

PATH  : The path of the previously generated source file (without the .main suffix)
MODE  : Precursor tracking mode (0 = mesh based, 1 = point-wise)

Notes:

  • Four source files will be required PATH.main, PATH.prec, PATH.live and PATH.precpoints

set dt

set dt NTRSH [ GTRSH ]

Sets probability threshold for delta-tracking. Input values:

NTRSH  : probability threshold for neutrons
GTRSH  : probability threshold for photons

Notes:

  • 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.
  • 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.
  • set dt 1 means that delta-tracking is always used and set dt 0 that it is never used.
  • Use of delta-tracking can be enforced or blocked in individual materials using the set forcedt and set blockdt options
  • Integral reaction rates are scored using the collision estimator of neutron flux, which has a few adjustable parameters (see set cfe).
  • For more information on tracking modes, see the detailed description on delta- and surface-tracking.

set ecut

set ecut EMINn EMINp

Sets minimum energy cut-off for neutrons and photons. Input values:

EMINn  : cut-off energy for neutrons (MeV)
EMINp  : cut-off energy for photons (MeV)

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.

set entr

set entr NX NY NZ [ XMIN XMAX YMIN YMAX ZMIN ZMAX ]

Defines the mesh structure used for calculating fission source entropy. Input values:

NX  : number of mesh cells in x-direction
NY  : number of mesh cells in y-direction
NZ  : number of mesh cells in z-direction
XMIN  : minimum mesh boundary in x-direction
XMAX  : maximum mesh boundary in x-direction
YMIN  : minimum mesh boundary in y-direction
YMAX  : maximum mesh boundary in y-direction
ZMIN  : minimum mesh boundary in z-direction
ZMAX  : maximum mesh boundary in z-direction

Notes:

  • Shannon entropy is used to monitor fission source convergence by recording the distribution of source points on mesh.
  • The calculation is invoked by setting the generation history record option on (set his).
  • The default mesh size is 5x5x5, extending 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.

set fininitfile

set fininitfile FILEPATH

Links a file containing an initial fuel behavior solution used as a starting point for a coupled calculation with the FINIX fuel behavior module. Input values:

FILEPATH  : path to the file

Notes:

  • The file should be a type 6 fuel behavior interface containing a previous solution from a coupled FINIX calculation.
  • The axial and radial nodalization as well as the included fuel rods should be the same in the file as in the calculation.

set fissh

set fissh ZAI1 E1 ZAI2 E2 ...

Overrides default fission heating values. Input values:

ZAIn  : nuclide identifiers
En  : energy deposited per fission in MeV

Notes:

  • 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.

set forcedt

set forcedt MAT1 MAT2 ...

Defines the list of materials where delta-tracking is always used. Input values:

MATn  : material names

Notes:

  • This option is used to override selection of tracking mode based on the probability threshold (see set dt) in individual materials.
  • Use of delta-tracking can be blocked in individual materials using set blockdt.
  • For more information on tracking modes, see the detailed description on delta- and surface-tracking.
  • Note to developers: should have different lists for neutrons and photons?

set fsp

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.

OPT  : option to switch fission source passing on (1/yes) or off (0/no)
NSKIP  : number of inactive generations on subsequent steps

Notes:

  • Fission source passing is turned off by default.
  • Number of inactive generations is taken from set pop card on the first step and from set fsp on all later steps.

set fum

set fum ERG [ BTCH ]

Activates fundamental mode for calculation of macroscopic data in few-group structure in critical spectrum.

ERG  : Intermediate multi-group structure for calculation of the leakage-corrected critical spectrum
BTCH  : When set to 2, results are averaged over all criticality cycles

Notes:

  • When invoked, Serpent applies a B1 leakage correction on group constants and produces a second set of output parameters.
  • The multi-group structure may be an energy grid defined using the ene card or a name of a pre-defined energy group structure.
  • Setting the fundamental mode calculation overrides the default 70-group structure used for macroscopic data calculation in infinite spectrum.
  • Averaging the results over all cycles may improve convergence and speed up the calculation, but all information on statistical errors is lost.

set gbuf

set gbuf FAC [ BNK ]

Sets the size of photon buffer and event bank. Input values:

FAC  : factor (> 1) defining the buffer size
BNK  : event bank size

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..
  • The buffer factor defines the buffer size relative to simulated batch size.
  • 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 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.

set gcu

set gcu UNI1 [ UNI2 UNI3 ... ]

Sets the universes for group constant generation. Input values:

UNIn  : universe where group constants are generated or -1 to switch group constant generation off.

Notes:

  • Group constants are by default generated in the root universe (universe 0).
  • Group constant generation should be switched off when the results are not needed (this may speed up the calculation).
  • See separate description of output parameters.

set gcut

set gcut GMAX

Sets generation cut-off for neutrons. Input values:

gmax  : number of simulated generations before cut-off

Notes:

  • The generation cut-off can be used in neutron external source simulations, to limit the length of fission chains.
  • Applicable only to neutron external source simulation (invoked using set nps)
  • Generation or time cut-off (set tcut) is always needed for neutron external source simulations in super-critical systems.

set his

set his OPT

Sets batch history record on or off. Input values:

OPT  : option to switch batch history record on (1/yes) or off (0/no)

Notes:

  • When invoked, Serpent collects batch-wise data on keff, fission source entropy etc. and produces a 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 impl

set impl ICAPT [INXN INUBAR]

Sets implicit reaction modes on or off.

ICAPT  : option to switch implicit capture reactions on (1/yes) or off (0/no)
INXN  : option to switch implicit nxn reactions on (1/yes) or off (0/no)
INUBAR  : number of fission neutrons to emit in each fission (nonzero = implicit treatment, 0 = analog treatment)

Notes:

  • Group constant generation requires implicit nxn reactions to be set on.
  • If an implicit nubar is given, the weights of the fission neutrons are scaled to conserve the physical number of fission neutrons.
  • 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.

LOOPn  : number of neutron tracking loops interpreted as a geometry error
ERRn  : flag to terminate neutron tracking when an infinite loop occurs (0 = no, 1 = yes)
LOOPp  : number of photon tracking loops interpreted as a geometry error
ERRp  : flag to terminate photon tracking when an infinite loop occurs (0 = no, 1 = yes)

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.
  • 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.

set inventory

set inventory MAT1 MAT2 ...
set inventory top N PARA

Specifies which nuclides or elements to include in the [input]_dep.m output file. Input values:

MATn  : Nuclide or element to include
N  : Number of nuclides
PARAM  : Parameter name

Notes:

  • MAT can be: an isotope, e.g. "U-235"; an element, e.g. "Zr"; or a special entry. Special entries include:
    • "act" -- actinides (Z>89)
    • "fp" -- fission products
    • "dp" -- decay products below thorium in the natural actinide decay series
    • "ng" -- noble gases (in the fission product range, He and Rn excluded)
  • The second syntax allows listing the most significant contributors to a given result. The parameters are:
    • "mass" -- contribution to mass fraction
    • "activity" -- contribution to activity
    • "sf" -- contribution to spontaneous fission rate
    • "gsrc" -- contribution to gamma emission rate
    • "dh" -- contribution to decay heat
    • "ingttox" -- contribution to ingestion toxicity
    • "inhtox" -- contribution to inhallation toxicity
  • for example "top 10 dh" gives the top 10 contributors to decay heat.
  • Since most significant contributors may change over time, the output may contain more nuclides than is requested in the input card.
  • The calculation of top contributors does not work with the "-rdep" command line option.

set isobra

set isobra ZAI1 MT1 FG1
           ZAI2 MT2 FG2
           ...

Defines constant branching ratios to isomeric states. Input values:

ZAIn  : isotope ZAI (10000 x Z + 10 x A + I)
MTn  : ENDF reaction MT
FGn  : fraction of reactions leading to the ground state of the product nuclide

Notes:

  • Serpent uses 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 overrides the constant ratios.

set lost

set lost LIM

Option to treat undefined geometry regions as void. Input values:

LIM  : maximum number of collisions allowed in undefined regions or -1 if no limit is set.

Notes:

  • This option allows the calculation to proceed even if the geometry routine encounters undefined cells.
  • The option is intended, for example, for complex geometries in which the boundaries of adjacent cells may not fully coincide.
  • When the option is set, the number of lost particles is printed in variable LOST_PARTICLES in the [input]_res.m output file.
  • The option should not be used to get away with errors in incomplete or poorly defined geometries.

set mcvol

set mcvol NP

Runs the Monte Carlo volume-checker routine to set material volumes before running the transport simulation. Input values:

NP  : number of points sampled in the geometry

Notes:

  • The 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.
  • The volume checker can also be used to produce a separate input file for the volume entries (see detailed description on defining material volumes).

set mdep

set mdep UNI VR N MAT1 MAT2 ... MATN 
         ZAI1 MT1
         ZAI2 MT2
         ...

Sets parameters for calculating cross sections for micro-depletion. Input values:

UNI  : universe where spatial homogenization is performed
VR  : volme ratio of materials included in micro depletion to total homogenized region
N  : number of materials included in the calculation
MAT1 MAT2 ... MATN  : material names
ZAIi  : Nuclide identifier (ZAI)
MTi  : ENDF reaction MT

Notes:

  • When this option is set, Serpent calculates multi-group microscopic cross sections for the listed nuclides and reactions.
  • The results are printed in a separate output file.
  • If the number of materials is zero, the calculation is carried over all burnable materials.
  • The listed materials must be enclosed inside the homogenized universe.
  • Transmutation reactions to ground and isomeric states can be calculated by adding 'g' and 'm' after the reaction MT. Reaction rates are calculated to all states by default.
  • Fission reactions corresponding to a specific yield can be calculated by adding 1 (=thermal), 2 (=fast) or 3 (=high-energy) after the reaction MT (e.g 181 is total fission cross section corresponding to thermal fission product yield).
  • Some actinides are missing the total fission channel, and setting the MT to 18 produces sum over MT's 19, 20, 21 and 38 (from version 2.1.29 on).

set micro

set micro ERG [ BTCH ]

Defines the intermediate multi-group structure used for group constant generation.

ERG  : Intermediate multi-group structure used for group constant generation
BTCH  : When set to 2, results are averaged over all criticality cycles

Notes:

  • Serpent uses an intermediate multi-group structure to calculate homogenized few-group constants. This input parameter is used to override the default 70-group structure used in the calculation.
  • The multi-group structure may be an energy grid defined using the ene card or a name of a pre-defined energy group structure.
  • Averaging the results over all cycles may improve convergence and speed up the calculation, but all information on statistical errors is lost.

set memfrac

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:

FRAC  : the fraction of system memory that Serpent is allowed to use (between 0 and 1)

Notes:

  • The default fraction is 0.8.
  • The fraction can also be set via a SERPENT_MEM_FRAC environmental variable.
  • Serpent tries to read the system total memory from the first line of the file /proc/meminfo

set minxs

See set cfe.

set mvol

set mvol MAT1 ZONE1,1 VOL1,1
         MAT1 ZONE1,2 VOL1,2
         ...
         MAT2 ZONE2,1 VOL2,1
         ...

Sets the volumes of material regions. Input values:

MATm  : name of material m
ZONEm,n  : index of zone n in material m
VOLm,n  : volume of zone n in material m

Notes:

set nbuf

set nbuf FAC [ BNK ]

Sets the size of neutron buffer and event bank. Input values:

FAC  : factor (> 1) defining the buffer size
BNK  : event bank size

Notes:

  • Neutron buffer refers to pre-allocated memory block used to store neutron particle data. This memory is needed for banking fission neutrons for the next generation and putting secondary neutrons in que.
  • The buffer factor defines the buffer size relative to simulated batch size.
  • 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 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.

set nfg

set nfg ERG

Defines the few-group structure used for group constant generation.

ERG  : Few-group structure used for group constant generation

Notes:

  • The few-group structure may be an energy grid defined using the ene card or a name of a pre-defined energy group structure.
  • The default is a two-group structure with boundary between fast and thermal group set to 0.625 eV.
  • Serpent uses an intermediate multi-group structure in the calculation. The default structure consists of 70 groups, and can be changed using the set micro option.
  • The few-group structure must be a sub-set of the intermediate multi-group structure.
  • See also: group constant output.

set nfylib

set nfylib LIB1 [ LIB2 LIB3 ... ]

Sets the neutron-induced fission yield library file paths. Input values:

LIBn  : library file paths

Notes:

  • Fission yield libraries are standard ENDF format files containing neutron-induced fission yield 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 fission yield data files in this path if not found at the absolute.

set ngamma

set ngamma MODE WMIN

Sets the coupled neutron-photon transport simulation on. Input values:

MODE  : simulation mode (0 = off, 1 = implicit, 2 = analog)
WMIN  : weight limit for implicit mode

Notes:

  • This input card invokes the production of prompt gammas in neutron reactions. The coupled simulation mode requires that both neutron and photon data libraries are defined.
  • In analog mode, the average number of secondary photons produced per collision is defined by the ratio of photon production cross section to material total. Each emitted photon assumes the weight of the incident neutron.
  • The implicit mode can be used to produce more photons by allowing variation in their statistical weight. The weight limit defines the minimum allowed weight of emitted photons. This method is close to what is used in MCNP.
  • Both calculation modes produce photons in all collisions without any correlation to the sampled reaction mode.

set nphys

set nphys FISS [ CAPT SCATT ]

Option to set reaction modes for neutrons on and off. Input values:

FISS  : option to handle fission (0 = not handled, 1 = handled)
CAPT  : option to handle capture (0 = not handled, 1 = handled)
SCATT  : option to handle scattering (0 = not handled, 1 = handled)

Notes:

  • All reaction modes are handled by default.
  • If fission is switched off, it is handled as capture.

set nps

set nps PP [ BTCH TBI ]

Sets parameters for simulated particle population in external source mode. Input values:

PP  : total number of particles
BTCH  : number of batches
TBI  : time binning for dynamic mode

Notes:

  • The total number of particles is divided by the given number of batches to give the number of particles per batch.
  • Using the nps card sets the mode to external source simulation. Criticality source simulation for neutrons is invoked using set pop.
  • If time binning is provided, the simulation is run in the dynamic mode with sequential population control. The bin structure is defined using the tme card.
  • Running an external source simulation requires a source, defined by the src card. Source definition also sets the transported particle type.
  • Neutron external source simulations are limited to sub-critical systems, unless dynamic mode, time cut-off (set tcut) or generation cut-off (set gcut) is invoked.
  • Neutron external source simulations in multiplying systems may require adjusting the neutron buffer (set nbuf).
  • Delayed neutron emission is switched off by default in neutron external source simulation (for compatibility with MCNP). Delayed neutrons can be included with set delnu.
  • In transient simulations, where an initial transient source is linked using the set dynsrc option, PP particles are sampled for each time interval.

set outp

set outp INT

Sets the interval (in cycles) for writing simulation output to files. Input values:

INT  : number of cycles after which the output-files are updated

Notes:

  • Default value is 50.
  • In coupled transient simulations the interval refers to time steps rather than batches. By default, the output is written after every 50th time step.
  • Affects files such as _res.m, _det.m as well as mesh plots.

set pdatadir

set pdatadir DIR

Sets the file path for auxiliary photon data. Input values:

DIR  : file path for directory where the data is located

Notes:

set poi

set poi OPT [ VR ]

Switches the calculation of poison cross sections on or off. Input values:

OPT  : option to switch the calculation of poison cross sections on (1/yes) or off (0/no)
VR  : ratio of fuel volume to the volume of the homogenized zone

Notes:

  • Poison cross sections include the fission yields and microscopic and macroscopic absorption cross sections of fission product poisons 135Xe and 149Sm, as well as their precursors. The data is part of the homogenized group constant output.
  • The calculation requires setting the decay and fission yield library file paths.
  • The volume ratio is required for calculating microscopic absorption cross sections.

set pop

set pop NPG NGEN NSKIP [ K0 BTCH  ]

Sets parameters for simulated neutron population in criticality source mode. Input values:

NPG  : number of neutrons per generation
NGEN  : number of active generations
NSKIP  : number of inactive generations
K0  : initial guess for keff
BTCH  : batching interval

Notes:

  • The simulation is first run for a number of inactive generations to allow the fission source to converge. This is followed by a number of active generations, during which the results are collected. The statistics are divided in batches, and by default each generation forms its own batch.
  • Using the pop card sets the mode to criticality source simulation. External source simulation is invoked using set nps.
  • Convergence of fission source can be monitored using Shannon entropy (input parameters set his and set entr).
  • Initial guess for keff is 1.0 by default. Setting the value manually may get the simulation going if it terminates on the first generation because of poor initial guess. The value does not affect fission source convergence.
  • See detailed descriptions on fission source convergence and statistical effects of batching.

set ppid

set ppid PID

Defines the external code process identifier (PID) number to be used for communication in the POSIX-based coupled calculation communications. Input values:

PID  : Process identifier (PID) of the (parent) process that Serpent should communicate with.

Notes:

  • Setting up a communication mode will enable the coupled calculation mode.
  • For more information see: External coupling

set ppw

set ppw UNI LAT

Turns on the calculation of pin-power form factors. The results are printed in variable PPW_POW_FRAC in the [input]_res.m output file. Input values:

UNI  : The universe where the power distribution is calculated.
LAT  : The lattice where the calculation is performed.

Notes:

set printm

set printm MODE [FRAC]

Print material compositions to input.bumat<step> file during burnup calculation.

MODE  : Set printing on (1/yes) or off (0/no).
FRAC  : Optional atomic fraction for printing out decay nuclides (nuclides with no transport cross sections). Value 0.0 will print out all decay nuclides, while 1.0 will not print out any decay nuclides. Value FRAC will print out nuclides whose atomic fraction in the material is greater than or equal to FRAC.

Notes:

  • Nuclides without transport data are excluded from the output by default.

set relfactor

set relfactor FAC

Sets the underrelaxation factor for the power relaxation used in coupled multi-physics calculations. Input values:

FAC  : underrelaxation factor

Notes:

  • Setting the underrelaxation factor to 0, disables power relaxation altogether. The power distribution written to the output-files will be unrelaxed and only based on the most recent iteration.

set rfr

set rfr STEP FILE
set rfr idx I FILE

Reads material compositions from a binary restart file. Input values:

STEP  : burnup step from which the compositions are obtained
I  : burnup step index from which the compositions are obtained
FILE  : name of the binary restart file

Notes:

  • Positive values for step are interpreted as burnup units (MWd/kgU) and negative values as time units (days)
  • This option can be used together with the set rfw feature for applying changes in the modeled system during burnup calculation

set rfw

set rfw OPT FILE

Writes material compositions in burnup calculation into a binary restart file. Input values:

OPT  : option to switch writing on (1/yes) or off (0/no)
FILE  : name of the binary restart file

Notes:

  • Restart file writing is switched off by default.
  • This option can be used together with the set rfr feature for applying changes in the modeled system during burnup calculation

set root

set root UNI

Sets the root universe. Input values:

UNI  : universe name

Notes:

  • Root universe is the universe at the lowest level of the geometry hierarchy, and must always be defined. By default the root is set to "0".
  • For more information, see detailed description on the universe-based geometry type in Serpent.

set runtme

set runtme T

Sets the maximum running time for the transport simulation. Input values:

T  : wall-clock running time (minutes)

Notes:

  • When defined, the transport simulation is terminated after the maximum time is reached.
  • Setting the parameter does not override the set pop or set nps option.

set savesrc

set savesrc PATH [ PN PP NX NY NZ ]

Sets up the creation of an initial source to be used in a dynamic simulation with delayed neutron emission.

PATH  : The path of the source file to be created
PN  : Proportion of tentative neutrons to save (default 1.0). Only affects criticality source simulations.
PP  : Proportion of tentative precursors to save (default 1.0). Only affects criticality source simulations.
NX  : Precursor mesh size in x-direction (default 1)
NY  : Precursor mesh size in y-direction (default 1)
NZ  : Precursor mesh size in z-direction (default 1)

Notes:

  • Four source files will be generated PATH.main, PATH.prec, PATH.live and PATH.precpoints
  • If used in criticality source simulation, the system should be critical and the values will correspond to steady state values.
  • If used in a dynamic simulation, the values will correspond to end-of-simulation values
  • If you are getting a warning from function WriteSourceFile "P larger than 1", you should lower the PN value.
  • See Transient simulations.

set seed

set seed RNG

Sets the seed value for the random number sequence. Input values:

RNG  : seed value used for the random number sequence

Notes:

  • By default, Serpent gets the RNG seed from system time. This option overrides the value.

set spd

set spd Vn Vp

Overrides the speed of simulated particles. Input values:

Vn  : speed of neutrons (cm/s)
Vp  : speed of photons (cm/s)

Notes:

  • This option is intended for adjusting particle speeds for better visualization in track plot animations.
  • Adjusting the speed (obviously) results in incorrect estimates for all time constants.
  • Exceeding the speed of light causes a fatal error in debug mode.

set sfylib

set sfylib LIB1 [ LIB2 LIB3 ... ]

Sets the spontaneous fission yield library file paths. Input values:

LIBn  : library file paths

Notes:

  • Spontaneous fission yield libraries are standard ENDF format files containing spontaneous fission yield data.
  • If the file path contains special characters it is advised to enclose it within quotes

set title

set title NAME

Sets a title for the calculation. Input values:

NAME  : title used for the calculation

Notes:

  • The title is printed in the run-time output. If the title is not set, the input file name is printed instead.

set tcut

set tcut Tmax

Sets time cut-off for neutrons and photons. Input values:

Tmax  : time limit for simulated particle histories (in seconds)

Notes:

  • The time cut-off can be used in both neutron and photon external source simulations, to limit the length of particle histories.
  • Time or generation cut-off (set gcut) is always needed for neutron external source simulations in super-critical systems.
  • Time cut-off is automatically set in the dynamic external source simulation mode.
  • Note to developers: this should take independent values for photons and neutrons

set tpa

set tpa Tmin Tmax TAIL NF EVB

Sets parameters for track plot animation. Input values:

Tmin  : starting time of track plot animation (in seconds)
Tmax  : end time of track plot animation (in seconds)
TAIL  : tail length of plotted particles (in cm)
NF  : number of frames
EVB  : event bank size

Notes:

  • The track plot animation works with the geometry plotter by creating a number of frames that visualize the motion of particles through the geometry.
  • The track plotter is invoked using the "-tracks N" command line option, where N is the number of simulated particle histories (if the set tpa option is not defined, the code plots particle tracks in a geometry plot output).
  • The routine produces NF frames [input]_trck[n]_frame[m].png, where [input] is the input file name, [n] is an index corresponding to the geometry plot and [m] is the frame index. The frames can be converted into a gif animation using tools like Imagemagick.
  • The plotted particles have a tail to better visualize their movement from one collision to the next. The length of this tail is given in centimeters (similar to geometry dimensions). Neutrons are plotted in magenta, photons are plotted in green.
  • Storing the simulated collision points requires additional memory, determined by the event bank size. If the given size is insufficient the calculation is terminated with an error message ("Event bank is empty").
  • Producing track plot animations may require adjusting the particle buffers (set nbuf and set gbuf).
  • The calculation may require a lot of memory for storing the complete simulated histories in neutron-multiplying systems or when particles are split by variance reduction.
  • Motion of thermal and fast neutrons cannot be captured simultaneously because of the several orders of magnitude difference in their speed. Thermal systems are best visualized by enforcing all neutrons to travel at the same speed using the set spd option.
  • See also detailed description on track plotter and track plot animation.

set transnorm

set transnorm FACTOR

Mulitplies the normalization from a transient source linked with set dynsrc by a factor. Input values:

FACTOR  : factor to multiply the normalization with.

Notes:

  • The normalization in transient simulations is, by default, same as in the transient source generation calculation.
  • See separate description of transient simulations.

set trc

set trc MAT FILE Emin

Defines transport correction for the calculation of diffusion parameters. Input values:

MAT  : material to which the correction is applied
FILE  : file path to correction curve data
Emin  : minimum energy above which the correction is applied

Notes:

  • The method calculates transport cross sections by multiplying material total cross sections by a user-defined energy-dependent transport correction ratio before collapsing the energy variable. These transport cross sections are used for calculating diffusion coefficients: D = 1/3\Sigma_\mathrm{tr}.
  • If the correction ratios are properly defined, the transport cross section is equivalent with the in-scattering approximation.
  • The out-scattering approximation is used for materials without defined correction factors and energies below the cut-off.
  • The produced homogenized cross sections are written in variables TRC_TRANSPXS and TRC_DIFFCOEF in the [input]_res.m and [input].coe output files.
  • See separate description of correction factor file format.

set ures

set ures OPT [ NUC1 NUC2 ... ]

Sets unresolved resonance probability table sampling on or off. Input values:

OPT  : option to switch probability table sampling on (1/yes) or off (0/no)
NUCn  : list of nuclides to which the option is applied to

Notes:

  • Probability table sampling is switched off by default
  • Note to developers: add description of dilution cut-off
  • See separate description of physics options in Serpent for differences to other codes.

set usym

set usym UNI AX BC X0 Y0 θ0 θw 

Defines a universe symmetry. Input values:

UNI  : universe name
AX  : symmetry axis (1 = x, 2 = y, 3 = z)
BC  : boundary condition (2 = reflective, 3 = periodic)
X0  : x-coordinate of the origin
Y0  : y-coordinate of the origin
θ0  : azimuthal position where the symmetry segment starts (degrees)
θw  : width of the segment (degrees)

Notes:

set xsplot

set xsplot NP Emin Emax 

Prints cross section data to [input]_xs0.m file. Input values:

NP  : Number of energy points to print.
Emin  : Lower boundary for the energy points.
Emax  : Upper boundary for the energy points.

The cross sections will be printed out at NP logarithmically spaced points between the energy boundaries.

References

  1. ^ Leppänen, J. "Serpent – a Continuous-energy Monte Carlo Reactor Physics Burnup Calculation Code." User manual, June 18, 2015.
  2. ^ Leppänen, J. "On the use of delta-tracking and the collision flux estimator in the Serpent 2 Monte Carlo particle transport code." Ann. Nucl. Energy 105 (2017) 161-167.