Description of output files
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 list of tables included in the file can be handled in the set dataout option. The format is self-explanatory.
The data is divided into two sections: nuclear data (Tables 1-4) and material data (Table 5), as follows:
- Table 1: Summary of nuclide data
- Table 2: Reaction and decay data
- Table 3: Fission yield data
- Table 4: Lost transmutation paths
- Table 5: Summary of material compositions
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 columns corresponding to different burnup steps. The list of variables included in the file can be handled in the set deppara option. The data types 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 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 cm^{3} |
BURNUP | : | Burnup in MWd/kgU |
Notes:
- For 2D geometries, values are on a per axial length basis.
- Two 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 ZAI MT EMIN EMAX RR ERR
where the values are:
IDX | : | Nuclide index corresponding to the entries in the nuc array |
ZAI | : | Nuclide identifier (ZAI) |
MT | : | ENDF reaction MT |
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) |
or in branch calculation:
[input]_mdx[n]b[m].m
Where:
[input] | : is the name of the input file |
[n] | : is the burnup index |
[m] | : is the branch index |
Flux
The data includes few-group flux densities printed in table FLUX_[u], where [u] is the universe for which the calculation is carried out. The columns are:
FLUX_{1} ERR_{1} FLUX_{2} ERR_{2} ..
where the values are:
FLUX_{g} | : | Few-group flux density (Few-group flux (integral, normalized) divided by the universe volume given in the set mdep card). |
ERR_{g} | : | Associated relative statistical error |
Cross sections
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 I N ERR_{N} XS_{1} ERR_{1} XS_{2} ERR_{2} ..
where the values are:
ZAI | : | Nuclide identifier (ZAI) |
MT | : | ENDF reaction MT |
I | : | Special flag (isomeric state or fission yield distribution number) |
N | : | Nuclide density smeared to homogenized volume |
ERR_{N} | : | Associated relative statistical error |
XS_{g} | : | Microscopic cross section |
ERR_{g} | : | Associated relative statistical error |
RR | : | Reaction rate |
ERR | : | Relative statistical error |
Fission yields
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:
- Product identifier -1 denotes that the fission yield is considered as a lost nuclide in Serpent (not included in the transmutation paths).
- Since Serpent version 2.2.1, the product identifier might be different than the actual nuclide in cross section data, e.g. in case there are yields for both the ground state and the metastable state (e.g. As-74 and As-74m in some cross section libraries) but the metastable state is not included in the transmutation paths. Then both yields are set to produce the ground state nuclide (e.g. As-74).
Decay data
Decay data of decaying nuclides are additionally printed in variable dec. The columns in the decay data table are:
1. ZAI 2. decay constant (1/s) 3. specific decay energy (J) 4. reaction type 5. branch fraction 6. product ZAI
Neutron-induced reaction data
Neutron induced reaction data of nuclides involved in the calculation are additionally printed in variable area. The columns in the neutron-induced data table are:
1. ZAI 2. reaction MT 3. reaction metastable flag 4. reaction product, ZAI 5. reaction Q-value (J)
Notes:
- For fission reactions the special flag corresponds to a fission product yield distribution, which are tabulated for different energies.
- For transmutation reactions the special flag indicates the isomeric state of the product nuclide (0 = ground state, 1 = isomeric state).
- Nuclide densities were not present before 2.1.32.
- Additional product nuclides (e.g. H-1) can be determined from the reaction type. Each number in the reaction type corresponds to one simultaneous reaction.
- Neutron-induced reaction data were not included before 2.2.0.
History output
Certain cycle-wise results are stored when the the set his option is invoked. The results are printed in file:
[input]_his[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 output consists of tables corresponding to different parameters. The first column lists the cycle index, which is then followed by the results grouped in three columns that provide the cycle-wise value, the cumulative mean and the corresponding relative statistical error. If the parameter has two values, the number of columns is 7 (cycle index + two groups of three columns of results), and so on.
By default the output includes the following variables:
Parameter | Description |
---|---|
HIS_IMP_KEFF | Implicit estimator of k_{eff} |
HIS_ANA_KEFF | Analog estimator of k_{eff} (total, prompt and delayed) |
HIS_COL_KEFF | Collision estimator of k_{eff} |
HIS_MEAN_POP_SIZE | Mean simulated population size |
HIS_MEAN_POP_WGT | Mean simulated population weight |
HIS_TRANSPORT_RUNTIME | Transport cycle running time (wall-clock and CPU time) |
HIS_TRANSPORT_CPU_USAGE | Mean CPU usage (ratio of CPU and wall-clock time) |
HIS_ENTR_SPT | Shannon entropy of source point distribution (total, x, y and z) |
HIS_ENTR_SWG | Shannon entropy of source weight distribution (total, x, y and z) |
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.