Description of output files

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Revision as of 21:28, 22 February 2016 by Jaakko Leppänen (Talk | contribs) (Main output file)

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Default output files

The following output files are always produced.

Main output file

The main Serpent output file is printed 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. The list of parameters is provided separately here.

Nuclide and material data

Optional 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:

  1. Nuclide index corresponding to the entries in the nuc array
  2. Reaction mt
  3. Nuclide ZAI
  4. Minimum energy of the reaction mode
  5. Maximum energy of the reaction mode
  6. Reaction rate
  7. Relative statistical error of reaction rate


  • 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.