Description of output files
Contents
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]_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. 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].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.
Burnup calculation output
Output from burnup calculations is printed in file:
[input]_dep.m
This file contains Matlab format data in several variables of form:
MAT_[material]_[data]
Where:
| [material] | : is the name of a material in the calculation |
| [data] | : is the data type |
Every variable is a matrix with rows corresponding to the nuclides requested in the set inventory option and colums corresponding to different burnup steps. The data types are:
| ADENS | : | Atom density in b-1 cm-1 |
| MDENS | : | mass density in g/cm3 |
| A | : | Activity in becquerels |
| H | : | Decay heat in Watts |
| SF | : | Spontaneous fission rate in fissions per second |
| GSRC | : | Photon emission rate in photons per second |
| ING_TOX | : | Ingestion toxicity in sieverts |
| INH_TOX | : | Inhalation toxicity in sieverts |
| VOLUME | : | Material volume in cm3 |
| BURNUP | : | Burnup in MWd/kgU |
Notes:
- For 2D geometries, values are on a per axial length basis.
- To additional rows are printed for each data array: data for lost nuclides (reaction products without nuclide data) and total.
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].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.
Micro depletion output
Microscopic few-group cross sections calculated for the purpose of micro-depletion (set mdep) option are printed in file:
[input]_mdx[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 includes few-group cross sections printed in table XS_[u], where [u] is the universe for which the calculation is carried out. The columns are:
ZAI MT N XS1 ERR1 XS2 ERR2 ..
where the values are:
| ZAI | : | Nuclide identifier (ZAI) |
| MT | : | ENDF reaction MT |
| N | : | Special flag (isomeric state or fission yield distribution number) |
| XSg | : | Microscopic cross section |
| ERRg | : | Associated relative statistical error |
| RR | : | Reaction rate |
| ERR | : | Relative statistical error |
Actinide fission yields are additionally printed in variables NFY_[ZAI]_[n], where [ZAI] is the nuclide identifier and [n] is the yield distribution number. Each yield corresponds to an energy, printed in variable NFY_[ZAI]_[n]E. The columns in the fission yield distribution are:
ZAI FI FC
where the values are:
| ZAI | : | Product identifier |
| FI | : | Independent yield |
| FC | : | Cumulative yield |
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.
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].bumat[n]
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 will be printed in a serpent-compatible material definition format.
