Description of output files

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Default output files

The following output files are always produced.

Main output file

The main output file contains all results calculated by default during the transport cycle. The file is written in Matlab-readable format in file:



[input]  : is the name of the input file

In calculations involving multiple transport cycles (such burnup calculation) the file is appended after each cycle. When the file is read into Matlab, each parameter is read into a variable (scalar or vector). A run index “idx” is assigned to each block of results, and the output data from different cycles are read into different rows (turning scalar variables into vectors and vector variables into matrices).

The list of parameters is provided separately here.

Nuclide and material data

Nuclear and material data is collected in in file:



[input]  : is the name of the input file

Basically the file lists all nuclides and their reactions as they are read from the nuclear data libraries. The material data includes isotopic compositions and densities, as well as volumes and masses if available. The format is self-explanatory.

Additional output files

The following output files are produced by invoking various input options.

Group constant output

Group constant data is printed separately in file:



[input]  : is the name of the input file

The file is designed to be read by post-processing scripts, and the format is described together with the automated burnup sequence.

Reaction rate output

Calculation of analog reaction rates by counting the number of sampled interactions is invoked using the set arr option. The output is printed in file:



[input]  : is the name of the input file
[n]  : is the burnup index (zero for first step or if no burnup calculation is run)

The data is printed in Matlab format in two variables: string array "nuc", which contains the nuclide identifiers (, and table "rr", consisting one row for each reaction and 7 columns:


where the values are:

IDX : Nuclide index corresponding to the entries in the nuc array
MT : Reaction mt
ZAI : Nuclide ZAI
EMIN : Minimum energy of the reaction mode
EMAX : Maximum energy of the reaction mode
RR : Reaction rate
ERR : Relative statistical error


  • The values are normalized microscopic reaction rates integrated over all materials and energies.
  • Neutron transport mode includes either reactions that affect neutron balance (absorption, fission, neutron-multiplying scattering) or all reactions, depending on the value of the input option.
  • All reaction modes are included in photon transport mode.

Burned material output

Burned materials' isotopic compositions and densities at each burnup step can be printed using the set printm option. The output will be in files of the form:



[input]  : is the name of the input file
[n]  : is the burnup index (zero for first step or if no burnup calculation is run)

The data will be printed in a serpent-compatible material definition format.

Additionally, depleted material output can be found in files of the form:


This file will contain matlab-compatible material data with variables named as:


Every variable is a matrix with columns corresponding to the materials requested in the set inventory option; rows correspond to different burnup steps.


[material]  : is the name of a material in the calculation
[aspect]  : is some aspect of the material.

The aspects of materials given in the *_dep.m file are:

ADENS : Atom density in b^-1 cm^-1
MDENS : mass density in g/cm^3
A : Activity in becquerels
H : Decay heat in watts
SF : Spontaneous fission in fissions per second
GSRC :  ??
ING_TOX : Ingestion toxicity in sieverts
INH_TOX : Inhalation toxicity in sieverts
VOLUME : Material volume in cm^3
BURNUP : Burnup in (units?)


  • For 2D geometries, values are on a per axial length basis.