Difference between revisions of "Description of output files"

From Serpent Wiki
Jump to: navigation, search
(Nuclide and material data)
(Nuclide and material data)
Line 29: Line 29:
 
| : is the name of the input file
 
| : 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.
  
 
== Optional output files ==
 
== Optional output files ==

Revision as of 23:45, 22 February 2016

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]_res.m

Where:

[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

Nuclear and material data is collected in in file:

[input].out

Where:

[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.

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].coe

Where:

[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]_arr[n].m

Where:

[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 (ZA.id), and table "rr", consisting one row for each reaction and 7 columns:

IDX MT ZAI EMIN EMAX RR ERR

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

Notes:

  • 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.