This page lists the output parameters in the main [input]_res.m output file.
General output parameters
Version, title and date
Parameter

Size

Description

VERSION

(string)

Code version

COMPILE_DATE

(string)

Date when the source code was compiled

DEBUG

1

Debug flag indicating if the DEBUG option was set when the source code was compiled

TITLE

(string)

Title defined using the set title input option

CONFIDENTIAL_DATA

1

Confidentiality flag set using the set confi input option

INPUT_FILE_NAME

(string)

File name of the main input file

WORKING_DIRECTORY

(string)

Directory path where the simulation was run

HOSTNAME

(string)

Host name where the simulation was run

CPU_TYPE

(string)

CPU type of the machine where the simulation was run (parsed from /proc/cpuinfo)

CPU_MHZ

(string)

CPU clock frequency of the machine where the simulation was run (parsed from /proc/cpuinfo)

START_DATE

(string)

Date and time when the simulation was started

COMPLETE_DATE

(string)

Date and time when this output was printed

Run parameters
Parameter

Size

Description

POP

1

Population size defined using the set pop input option

CYCLES

1

Number of active cycles defined using the set pop input option


SKIP

1

Number of inactive cycles defined using the set pop input option


BATCH_INTERVAL

1

Batching interval defined using the set pop input option

SRC_NORM_MODE

1

Source normalization mode

SEED

1

Random number seed taken from system time or defined using the set seed input option

UFS_MODE

1

Uniform fission source mode defined using the set ufs input option

UFS_ORDER

1

Uniform fission exponential factor using the set ufs input option

NEUTRON_TRANSPORT_MODE

1

Flag indicating whether or not neutron transport simulation is on

PHOTON_TRANSPORT_MODE

1

Flag indicating whether or not neutron transport simulation is on

GROUP_CONSTANT_GENERATION

1

Flag indicating whether or not group constant generation is on

B1_CALCULATION

3

Flag indicating whether or not B1 calculation is on

B1_BURNUP_CORRECTION

1

Flag indicating whether or not B1 burnup correction is on

CRIT_SPEC_MODE

2

Critical spectrum modes

IMPLICIT_REACTION_RATES

1

Flag indicating whether or implicit reaction rates are used for group constant generation

Optimization
Parameter

Size

Description

OPTIMIZATION_MODE

1

Optimization mode defined using the set opti input option

RECONSTRUCT_MICROXS

1

Flag indicating whether or not microscopic cross sections are reconstructed on the unionized energy grid

RECONSTRUCT_MACROXS

1

Flag indicating whether or not macroscopic cross sections are reconstructed on the unionized energy grid

DOUBLE_INDEXING

1

Double indexing option defined using the set dix input option

MG_MAJORANT_MODE

1

Multigroup majorant mode

SPECTRUM_COLLAPSE

1

Spectrum collapse method flag (set xscalc input option)

Parallelization
Parameter

Size

Description

MPI_TASKS

1

Number of parallel MPI tasks

OMP_THREADS

1

Number of parallel OpenMP threads

MPI_REPRODUCIBILITY

1

MPI reproducibility option defined by the set repro input option

OMP_REPRODUCIBILITY

1

OpenMP reproducibility option defined by the set repro input option

OMP_HISTORY_PROFILE

N

Fraction of particle histories run for each parallel OpenMP thread

SHARE_BUF_ARRAY

1

Shared buffer flag

SHARE_RES2_ARRAY

1

Shared RES2 array flag

OMP_SHARED_QUEUE_LIM

1

Limiting value for using shared particle queue

File paths
Parameter

Size

Description

XS_DATA_FILE_PATH

(string)

Cross section directory file path defined using the set acelib input option

DECAY_DATA_FILE_PATH

(string)

Radioactive decay data file path defined using the set declib input option

SFY_DATA_FILE_PATH

(string)

Spontaneous fission yield data file path defined using the set sfylib input option

NFY_DATA_FILE_PATH

(string)

Neutroninduced fission yield data file path defined using the set nfylib input option

BRA_DATA_FILE_PATH

(string)

Isomeric branching ratio data file path defined using the set bralib input option

Misc. statistics
Collision and reaction sampling (neutrons/photons)
Notes:
 The first single/pair value corresponds to neutrons and, the second single/pair value corresponds to photons.
Parameter

Size

Description

MIN_MACROXS

2/2

Macroscopic cross section corresponding to the minimum mfp used for scoring the collision flux estimator (see the set cfe input option)

DT_THRESH

1/1

Probability threshold used for switching to deltatracking (see the set dt input option)

ST_FRAC

2/2

Fraction of paths sampled using surfacetracking

DT_FRAC

2/2

Fraction of paths sampled using deltatracking

DT_EFF

2/2

Deltatracking efficiency

REA_SAMPLING_EFF

2/2

Reaction sampling efficiency

REA_SAMPLING_FAIL

2/2

Fraction of failed reaction samples

TOT_COL_EFF

2/2

Total collision efficiency

AVG_TRACKING_LOOPS

2/2, 2/2

Average number of tracking loops per history and, fraction of failed tracking loops

AVG_TRACKS

2/2

Average number of tracks per history

AVG_REAL_COL

2/2

Average number of real collisions per history

AVG_VIRT_COL

2/2

Average number of virtual collisions per history

AVG_SURF_CROSS

2/2

Average number of surface crossings per history (NOTE: accurate only in ST mode)

LOST_PARTICLES

1

Number of lost particles

Run statistics
Parameter

Size

Description

CYCLE_IDX

1

Cycle index when output was printed

SIMULATED_HISTORIES

1

Number of simulated histories when output was printed

MEAN_POP_SIZE

1

Mean population size

MEAN_POP_WGT

1

Mean population weight

SIMULATION_COMPLETED

1

Flag indicating whether or not the simulation was completed

Running times
Notes:
 All times in minutes
 In burnup calculations the first value provides the cumulative and the second value the cyclewise value
Parameter

Size

Description

TOT_CPU_TIME

1

Total CPU time

RUNNING_TIME

1

Total wallclock running time

INIT_TIME

1(2)

Wallclock time spent for initialization

PROCESS_TIME

1(2)

Wallclock time spent for processing

TRANSPORT_CYCLE_TIME

1(2)

Wallclock time spent for transport simulation

BURNUP_CYCLE_TIME

1(2)

Wallclock time spent for burnup solution

BATEMAN_SOLUTION_TIME

1(2)

Wallclock time spent for solving the Bateman equations

MPI_OVERHEAD_TIME

1(2)

Wallclock time spent MPI communication

ESTIMATED_RUNNING_TIME

1(2)

Estimated total wallclock running time

CPU_USAGE

1

Total CPU usage fraction

TRANSPORT_CPU_USAGE

1(2)

CPU usage fraction in transport simulation

OMP_PARALLEL_FRAC

1

Fraction of time spent in OpenMP parallel loops

Memory usage
Notes:
 All values are in megabytes
 Serpent allocates memory in fixed segments, so the allocated memory size may be larger than what is needed for the simulation
Parameter

Size

Description

AVAIL_MEM

1

Available memory size

ALLOC_MEMSIZE

1

Allocated memory size

MEMSIZE

1

Used memory size

XS_MEMSIZE

1

Memory size used for storing cross sections

MAT_MEMSIZE

1

Memory size used for storing materialwise data

RES_MEMSIZE

1

Memory size used for storing results

IFC_MEMSIZE

1

Memory size used for data for responsematrix solver

MISC_MEMSIZE

1

Memory size used for data for miscellaneous data

UNKNOWN_MEMSIZE

1

Memory size used for data for uncategorized data

UNUSED_MEMSIZE

1

Allocated memory not used for anything

Geometry parameters
Parameter

Size

Description

TOT_CELLS

1

Total number of cells

UNION_CELLS

1

Total number of cells defined using unions

Neutron energy grid
Parameter

Size

Description

NEUTRON_ERG_TOL

1

Reconstruction tolerace for unionized energy grid

NEUTRON_ERG_NE

1

Number of points in unionized energy grid

NEUTRON_EMIN

1

Minimum energy for neutron cross section data

NEUTRON_EMAX

1

Maximum energy for neutron cross section data

Unresolved resonance probability table sampling
Parameter

Size

Description

URES_DILU_CUT

1

Density cutoff used for unresolved resonance probability table sampling

URES_EMIN

1

Minimum energy for unresolved resonance range

URES_EMAX

1

Maximum energy for unresolved resonance range

URES_AVAIL

1

Number of nuclides with probability table data

URES_USED

1

Number of nuclides for which probability table sampling was used

Nuclides and reaction channels
Parameter

Size

Description

TOT_NUCLIDES

1

Total number of nuclides

TOT_TRANSPORT_NUCLIDES

1

Total number of nuclides with transport cross sections

TOT_DOSIMETRY_NUCLIDES

1

Total number of nclides with dosimetry cross sections

TOT_DECAY_NUCLIDES

1

Total number of decay nuclides (without transport cross sections)

TOT_PHOTON_NUCLIDES

1

Total number of nuclides with photon cross section data

TOT_REA_CHANNELS

1

Total number of reaction channels

TOT_TRANSMU_REA

1

Total number of transmutation reactions

Physics
Neutron physics options
Parameter

Size

Description

USE_DELNU

1

Flag indicating whether or not delayed neutron emission is on (see set delnu input option)

USE_URES

1

Flag indicating whether or not unresolved resonance probability table sampling is on (see set ures input option)

USE_DBRC

1

Flag indicating whether or not Dopplerbroadening rejection correction is on (see set dbrc input option)

IMPL_CAPT

1

Flag indicating whether or not implicit capture reaction mode is on (see set impl input option)

IMPL_NXN

1

Flag indicating whether or not implicit nxn reaction mode is on (see set impl input option)

IMPL_FISS

1

Flag indicating whether or not implicit fission reaction mode is on (see set impl input option)

DOPPLER_PREPROCESSOR

1

Flag indicating whether or not Dopplerbroadening preprocessor is on (see tmp option, in mat card)

TMS_MODE

1

Flag indicating whether or not target motion sampling is on (see tms option, in mat card)

SAMPLE_FISS

1

Flag indicating whether or not fission reactions are handled (see set nphys input option)

SAMPLE_CAPT

1

Flag indicating whether or not capture reactions are handled (see set nphys input option)

SAMPLE_SCATT

1

Flag indicating whether or not scattering reactions are handled (see set nphys input option)

Energy deposition
Notes:
 The list of fission energy release components includes: (1) EFR, kinetic energy of the fission products (following prompt neutron emission from the fission fragments); (2) ENP, kinetic energy of the prompt fission neutrons; (3) END, kinetic energy of the delayed fission neutrons; (4) EGP, total energy release by the emission of prompt gamma rays; (5) EGD, total energy release by the emission of delayed gamma rays; (6) EB, total energy release by delayed beta’s; (7) ENU, energy carried away by neutrinos; (8) ER, total energy less the energy of the neutrinos (ET  ENU), equal to the pseudoQvalue in File 3 for MT=18; (9) ET, sum of all the partial energies previously listed, corresponding to the total energy release per fission and equal the Qvalue.
Parameter

Size

Description

EDEP_MODE

1

Energy deposition mode (see set edepmode input option)

EDEP_DELAYED

1

Energy of delayed components in energy deposition calculations (see set edepdel input option)

EDEP_KEFF_CORR

1

Flag indicating whether or not correction for energy deposition estimates in noncritical systems (see set edepkcorr input option)

EDEP_LOCAL_EGD

1

Energy distribution of delayed components in energy deposition calculations, mode 3 (see set edepdel input option)

EDEP_COMP

9

Fission energy release components: EFR, ENP, END, EGP, EGD, EB, ENU, ER, ET.

EDEP_CAPT_E

1

Additional energy release in capture reactions, mode 1 (see set edepmode input option)

Radioactivity data
Parameter

Size

Description

TOT_ACTIVITY



TOT_DECAY_HEAT



TOT_SF_RATE



ACTINIDE_ACTIVITY



ACTINIDE_DECAY_HEAT



FISSION_PRODUCT_ACTIVITY



FISSION_PRODUCT_DECAY_HEAT



INHALATION_TOXICITY



INGESTION_TOXICITY



ACTINIDE_INH_TOX



ACTINIDE_ING_TOX



FISSION_PRODUCT_INH_TOX



FISSION_PRODUCT_ING_TOX



SR90_ACTIVITY



TE132_ACTIVITY



I131_ACTIVITY



I132_ACTIVITY



CS134_ACTIVITY



CS137_ACTIVITY



PHOTON_DECAY_SOURCE



NEUTRON_DECAY_SOURCE



ALPHA_DECAY_SOURCE



ELECTRON_DECAY_SOURCE



Normalization coefficient
Parameter

Size

Description

NORM_COEF

2/2

Proportionality constant between the simulated events and the "physical" events that the simulated events represent, for neutrons and photons.

Parameters for burnup calculation
Parameter

Size

Description

BURN_MATERIALS

1

Number of depleted materials.

BURN_MODE

1

Burnup mode: 1 = TTA, 2 = CRAM (see set bumode input option).

BURN_STEP

1

Burnup step index.

BURN_RANDOMIZE_DATA

3

Flag indicating whether or not randomize data is set on: decay constants, fission yields and decay heat (see set rnddec input option).

BURNUP

2

Burnup at the current step (in MWd/kgU): cumulative and realcumulative.

BURN_DAYS

2

Number of burn days at the current step: cumulative and stepwise.

FIMA

3

Number of fissions per initial fissile atom at the current step: relative stepwise, increment stepwise, final stepwise.

Analog reaction rate estimators
Parameter

Size

Description

CONVERSION_RATIO



U235_FISS



U238_FISS



U235_CAPT



U238_CAPT



XE135_CAPT



Particle balance
Neutron balance (particles/weight)
Parameter

Size

Description

BALA_SRC_NEUTRON_SRC



BALA_SRC_NEUTRON_FISS



BALA_SRC_NEUTRON_NXN



BALA_SRC_NEUTRON_VR



BALA_SRC_NEUTRON_TOT



BALA_LOSS_NEUTRON_CAPT



BALA_LOSS_NEUTRON_FISS



BALA_LOSS_NEUTRON_LEAK



BALA_LOSS_NEUTRON_CUT



BALA_LOSS_NEUTRON_ERR



BALA_LOSS_NEUTRON_TOT



BALA_NEUTRON_DIFF



Integral results
Normalized total reaction rates (neutrons)
Parameter

Size

Description

TOT_POWER



TOT_POWDENS



TOT_GENRATE



TOT_FISSRATE



TOT_CAPTRATE



TOT_ABSRATE



TOT_SRCRATE



TOT_FLUX



TOT_PHOTON_PRODRATE



TOT_LEAKRATE



ALBEDO_LEAKRATE



TOT_LOSSRATE



TOT_CUTRATE



TOT_RR



TOT_XE135_ABSRATE



INI_FMASS



TOT_FMASS



INI_BURN_FMASS



TOT_BURN_FMASS



Equilibrium Xe135 iteration
Parameter

Size

Description

XE135_EQUIL_CONC

2

Averaged equilibrium Xe135 concentration (see set xenon input option)

I135_EQUIL_CONC

2

Averaged equilibrium I135 concentration (see set xenon input option)

Equilibrium Sm149 iteration
Parameter

Size

Description

SM149_EQUIL_CONC

2

Averaged equilibrium Sm149 concentration (see set samarium input option)

PM149_EQUIL_CONC

2

Averaged equilibrium Pm149 concentration (see set samarium input option)

Sixfactor formula
Parameter

Size

Description

SIX_FF_ETA

2

Analog estimate of average number of neutrons emitted per thermal neutron absorbed in fuel

SIX_FF_F

2

Analog estimate of thermal utilization factor

SIX_FF_P

2

Analog estimate of resonance escape probability

SIX_FF_EPSILON

2

Analog estimate of fast fission factor

SIX_FF_LF

2

Analog estimate of fast nonleakage probability

SIX_FF_LT

2

Analog estimate of thermal nonleakage probability

SIX_FF_KINF

2

Analog estimate of sixfactor k_{inf} (fourfactor k_{eff})

SIX_FF_KEFF

2

Analog estimate of sixfactor k_{eff}

Fission neutron and energy production
Parameter

Size

Description

NUBAR



FISSE



Criticality eigenvalues
Parameter

Size

Description

ANA_KEFF

6

Analog estimate of k_{eff}: total, prompt and delayed neutron contribution.

IMP_KEFF

2

Implicit estimate of k_{eff}.

COL_KEFF

2

Collision estimate of k_{eff}.

ABS_KEFF

2

Absorption estimate of k_{eff}.

ABS_KINF

2

Absorption estimate of k_{inf}.

GEOM_ALBEDO

6

Fixed or iterated value for albedo boundary condition for x,y and zdirections (see set bc or set iter alb input options).

ALF (Average lethargy of neutrons causing fission)
Parameter

Size

Description

ANA_ALF

2

Analog estimate of average lethargy of neutrons causing fission

IMP_ALF

2

Implicit estimate of average lethargy of neutrons causing fission

EALF (Energy corresponding to average lethargy of neutrons causing fission)
Parameter

Size

Description

ANA_EALF

2

Analog estimate of energy corresponding to the average lethargy of neutrons causing fission

IMP_EALF

2

Implicit estimate of energy corresponding to the average lethargy of neutrons causing fission

AFGE (Average energy of neutrons causing fission)
Parameter

Size

Description

ANA_AFGE

2

Analog estimate of average energy of neutrons causing fission

IMP_AFGE

2

Implicit estimate of average energy of neutrons causing fission

Time constants
Forwardweighted delayed neutron parameters
Parameter

Size

Description

PRECURSOR_GROUPS

1

Number of delayed neutron precursor groups (referred to as D below)

FWD_ANA_BETA_ZERO

2D + 2

Analog estimator of physical delayed neutron fractions (number of delayed neutrons emitted in fission): total and groupwise

FWD_ANA_LAMBDA

2D + 2

Analog estimator of delayed neutron precursor decay constants: total and groupwise

Betaeff using Meulekamp's method
Parameter

Size

Description

ADJ_MEULEKAMP_BETA_EFF

2D + 2

Adjointweighted effective delayed neutron fractions using Meulekamp's method: total and groupwise

ADJ_MEULEKAMP_LAMBDA

2D + 2

Adjointweighted of delayed neutron precursor decay constants using Meulekamp's method: total and groupwise

Adjoint weighted time constants using Nauchi's method
Parameter

Size

Description

IFP_CHAIN_LENGTH

1

Number of generations within the iterated fission probability method

ADJ_NAUCHI_GEN_TIME

6

Adjointweighted neutron generation times using Nauchi's method: total, prompt and, delayed

ADJ_NAUCHI_LIFETIME

6

Adjointweighted neutron lifetimes using Nauchi's method: total, prompt and, delayed.

ADJ_NAUCHI_BETA_EFF

2D + 2

Adjointweighted effective delayed neutron fractions using Nauchi's method: total and groupwise

ADJ_NAUCHI_LAMBDA

2D + 2

Adjointweighed of delayed neutron precursor decay constants using Nauchi's method: total and groupwise

Adjoint weighted time constants using IFP
Parameter

Size

Description

ADJ_IFP_GEN_TIME

6

Adjointweighted neutron generation times using the iterated fission probability method: total, prompt and, delayed

ADJ_IFP_LIFETIME

6

Adjointweighted neutron lifetimes using the iterated fission probability method: total, prompt and, delayed

ADJ_IFP_IMP_BETA_EFF

2D + 2

Implicit estimator of adjointweighted effective delayed neutron fractions using the iterated fission probability method: total and groupwise

ADJ_IFP_IMP_LAMBDA

2D + 2

Implicit estimator of adjointweighted of delayed neutron precursor decay constants using the iterated fission probability method: total and groupwise

ADJ_IFP_ANA_BETA_EFF

2D + 2

Analog estimator of adjointweighted effective delayed neutron fractions using the iterated fission probability method: total and groupwise

ADJ_IFP_ANA_LAMBDA

2D + 2

Analog estimator of adjointweighted of delayed neutron precursor decay constants using the iterated fission probability method: total and groupwise

ADJ_IFP_ROSSI_ALPHA

2

Adjointweighted Rossi alpha using the iterated fission probability method

Adjoint weighted time constants using perturbation technique
Parameter

Size

Description

ADJ_PERT_GEN_TIME

2

Adjointweighted neutron generation time using the perturbation technique

ADJ_PERT_LIFETIME

2

Adjointweighted neutron lifetime using the perturbation technique

ADJ_PERT_BETA_EFF

2

Adjointweighted effective delayed neutron fraction using the perturbation technique

ADJ_PERT_ROSSI_ALPHA

2

Adjointweighted Rossi alpha using the perturbation technique

Inverse neutron speed
Parameter

Size

Description

ANA_INV_SPD

2

Analog estimate of inverse neutron speed

Analog slowingdown and thermal neutron lifetime (total/prompt/delayed)
Parameter

Size

Description

ANA_SLOW_TIME

6

Analog estimate of slowingdown time: total, prompt and, delayed

ANA_THERM_TIME

6

Analog estimate of thermal neutron lifetime: total, prompt and, delayed

ANA_THERM_FRAC

6

Analog estimate of neutron thermalisation fraction: total, prompt and, delayed

ANA_DELAYED_EMTIME

2

Analog estimate of delayed neutron emission time

ANA_MEAN_NCOL

4

Analog estimate of average number of collisions per history: total and to fission

Homogenized group constants
Notes:
 Group constants are calculated by first homogenizing the geometry using a multigroup structure with H energy groups. The data is then collapsed into the final fewgroup structure with G groups using the infinite and leakagecorrected flux spectra.
 The methodology used in Serpent for spatial homogenization is described in a paper published in Annals of Nuclear Energy in 2016.^{[1]}
 The fundamental mode calculation is off by default, and invoked by the set fum option. Otherwise all values with B1 prefix are printed as zeros.
 The intermediate multigroup structure is defined using option set micro or set fum.
 The fewgroup structure is defined using option set nfg.
 The universes in which the group constants are calculated are listed in option set gcu. The calculation is performed for root universe 0 by default, and can be switched off with "set gcu 1".
 If data is produced in multiple universes within a single run, the data is assigned with different run indexes (idx)
 The parameter names can be listed in the set coefpara option, and they will be included in the group constant output file when the automated burnup sequence is invoked.
 The order in which twodimensional data (scattering matrices, ADF and pinpower parameters) is printed in the [input].coe output file is different from what is listed below in update 2.1.24 and earlier versions.
Common parameters
Parameter

Size

Description

GC_UNIVERSE_NAME

(string)

Name of the universe where spatial homogenization was performed

MICRO_NG

1

Number of energy groups in the intermediate multigroup structure (referred to as H below)

MICRO_E

H + 1

Group boundaries in the intermediate multigroup structure (in ascending order)

MACRO_NG

1

Number of energy groups in the final fewgroup structure (referred to as G below)

MACRO_E

G + 1

Group boundaries in the final fewgroup structure (in descending order)

Group constants homogenized in infinite spectrum
Parameter

Size

Description

INF_MICRO_FLX

2H

Multigroup flux spectrum (integral, unnormalized)

INF_FLX

2G

Fewgroup flux (integral, normalized)

INF_KINF

2

Infinite multiplication factor

Reaction cross sections
Parameter

Size

Description

INF_TOT

2G

Total cross section

INF_CAPT

2G

Capture cross section

INF_FISS

2G

Fission cross section

INF_NSF

2G

Fission neutron production cross section

INF_KAPPA

2G

Average deposited fission energy (MeV)

INF_INVV

2G

Inverse neutron speed (s/cm)

INF_NUBAR

2G

Average neutron yield

INF_ABS

2G

Absorption cross section (capture + fission)

INF_REMXS

2G

Removal cross section (groupremoval + absorption)

INF_RABSXS

2G

Reduced absorption cross section (total  scattering production)

Fission spectra
Parameter

Size

Description

INF_CHIT

2G

Fission spectrum (total)

INF_CHIP

2G

Fission spectrum (prompt neutrons)

INF_CHID

2G

Fission spectrum (delayed neutrons)

Scattering cross sections
Notes:
 Scattering production includes multiplying (n,2n), (n,3n), etc. reactions.
Parameter

Size

Description

INF_SCATT0

2G

Total P_{0} scattering cross section

INF_SCATT1

2G

Total P_{1} scattering cross section

INF_SCATT2

2G

Total P_{2} scattering cross section

INF_SCATT3

2G

Total P_{3} scattering cross section

INF_SCATT4

2G

Total P_{4} scattering cross section

INF_SCATT5

2G

Total P_{5} scattering cross section

INF_SCATT6

2G

Total P_{6} scattering cross section

INF_SCATT7

2G

Total P_{7} scattering cross section

INF_SCATTP0

2G

Total P_{0} scattering production cross section

INF_SCATTP1

2G

Total P_{1} scattering production cross section

INF_SCATTP2

2G

Total P_{2} scattering production cross section

INF_SCATTP3

2G

Total P_{3} scattering production cross section

INF_SCATTP4

2G

Total P_{4} scattering production cross section

INF_SCATTP5

2G

Total P_{5} scattering production cross section

INF_SCATTP6

2G

Total P_{6} scattering production cross section

INF_SCATTP7

2G

Total P_{7} scattering production cross section

Scattering matrices
Notes:
 Scattering production includes multiplying (n,2n), (n,3n), etc. reactions.
 The order of values ([input].coe) or value pairs ([input]_res.m) is: where refers to scattering from group g to g'.
 The data in the [input]_res.m file can be read into a G by G matrix with Matlab reshapecommand, for example:
reshape(INF_S0(idx,1:2:end), G, G);
Parameter

Size

Description

INF_S0

2G^{2}

P_{0} scattering matrix

INF_S1

2G^{2}

P_{1} scattering matrix

INF_S2

2G^{2}

P_{2} scattering matrix

INF_S3

2G^{2}

P_{3} scattering matrix

INF_S4

2G^{2}

P_{4} scattering matrix

INF_S5

2G^{2}

P_{5} scattering matrix

INF_S6

2G^{2}

P_{6} scattering matrix

INF_S7

2G^{2}

P_{7} scattering matrix

INF_SP0

2G^{2}

P_{0} scattering production matrix

INF_SP1

2G^{2}

P_{1} scattering production matrix

INF_SP2

2G^{2}

P_{2} scattering production matrix

INF_SP3

2G^{2}

P_{3} scattering production matrix

INF_SP4

2G^{2}

P_{4} scattering production matrix

INF_SP5

2G^{2}

P_{5} scattering production matrix

INF_SP6

2G^{2}

P_{6} scattering production matrix

INF_SP7

2G^{2}

P_{7} scattering production matrix

Diffusion parameters
Notes:
 Calculation of sensible values for INF_TRANSPXS and INF_DIFFCOEF requires fine enough intermediate multigroup structure.
 The cumulative migration method ^{[2]} (CMM) was first developed for the OpenMC code.
 CMM diffusion coefficients and transport cross sections are reasonable only when they are calculated over entire geometry (homogenized region covers the entire geometry and is surrounded by periodic or reflective boundary conditions). This means that e.g. pin cell CMM diffusion coefficients can not be calculated from a 2D fuel assembly calculation.
 Calculation of TRC_TRANSPXS and TRC_DIFFCOEF requires defining energydependent correction factors using the set trc option.
 Calculation of CMM_TRANSPXS and CMM_DIFFCOEF requires that their calculation is not switched off using the set cmm option.
Parameter

Size

Description

INF_TRANSPXS

2G

Transport cross section (calculated using the outscattering approximation)

INF_DIFFCOEF

2G

Diffusion coefficient (calculated using the outscattering approximation)

CMM_TRANSPXS

2G

Transport cross section calculated using the cumulative migration method (equivalent with the inscattering approximation)

CMM_TRANSPXS_X

2G

Xcomponent of the directional transport cross section calculated using the cumulative migration method (equivalent with the inscattering approximation)

CMM_TRANSPXS_Y

2G

Ycomponent of the directional transport cross section calculated using the cumulative migration method (equivalent with the inscattering approximation)

CMM_TRANSPXS_Z

2G

Zcomponent of the directional transport cross section calculated using the cumulative migration method (equivalent with the inscattering approximation)

CMM_DIFFCOEF

2G

Diffusion coefficient calculated using the cumulative migration method (equivalent with the inscattering approximation)

CMM_DIFFCOEF_X

2G

Xcomponent of the directional diffusion coefficient calculated using the cumulative migration method (equivalent with the inscattering approximation)

CMM_DIFFCOEF_Y

2G

Ycomponent of the directional diffusion coefficient calculated using the cumulative migration method (equivalent with the inscattering approximation)

CMM_DIFFCOEF_Z

2G

Zcomponent of the directional diffusion coefficient calculated using the cumulative migration method (equivalent with the inscattering approximation)

TRC_TRANSPXS

2G

Transport cross section calculated by applying userdefined transport correction factors to total cross section

TRC_DIFFCOEF

2G

Diffusion coefficient calculated by applying userdefined transport correction factors to total cross section

Poison cross sections
Notes:
 Printed only if poison cross section option is on (see set poi).
 Xe135m values printed only if separate treatment of Xe135m is on (see set poi).
Parameter

Size

Description

INF_I135_YIELD

2G

Fission yield of I135 (cumulative, includes all precursors)

INF_XE135_YIELD

2G

Fission yield of Xe135

INF_XE135M_YIELD

2G

Fission yield of Xe135m

INF_PM149_YIELD

2G

Fission yield of Pm149 (cumulative, includes all precursors)

INF_SM149_YIELD

2G

Fission yield of Sm149

INF_I135_MICRO_ABS

2G

Microscopic absorption cross section of I135

INF_XE135_MICRO_ABS

2G

Microscopic absorption cross section of Xe135

INF_XE135M_MICRO_ABS

2G

Microscopic absorption cross section of Xe135m

INF_PM149_MICRO_ABS

2G

Microscopic absorption cross section of Pm149

INF_SM149_MICRO_ABS

2G

Microscopic absorption cross section of Sm149

INF_XE135_MACRO_ABS

2G

Macroscopic absorption cross section of Xe135

INF_XE135M_MACRO_ABS

2G

Macroscopic absorption cross section of Xe135m

INF_SM149_MACRO_ABS

2G

Macroscopic absorption cross section of Sm149

Poison decay constants
Parameter

Size

Description

PM147_LAMBDA

1

Decay constant of Pm147

PM148_LAMBDA

1

Decay constant of Pm147

PM148M_LAMBDA

1

Decay constant of Pm148m

PM149_LAMBDA

1

Decay constant of Pm149

I135_LAMBDA

1

Decay constant of I135

XE135_LAMBDA

1

Decay constant of Xe135

XE135M_LAMBDA

1

Decay constant of Xe135m

I135_BR

1

Branching ratio of I135 decay to Xe135. Branching ratio of I135 decay to Xe135m is (1  I135_BR).

Group constants homogenized in leakagecorrected spectrum
Parameter

Size

Description

B1_MICRO_FLX

2H

Multigroup flux spectrum (integral, unnormalized)

B1_FLX

2G

Fewgroup flux (integral, normalized)

B1_KINF

2

Infinite multiplication factor

B1_KEFF

2

Effective multiplication factor

B1_B2

2

Critical buckling

B1_ERR

2

Absolute deviation of k_{eff} from unity

Reaction cross sections
Parameter

Size

Description

B1_TOT

2G

Total cross section

B1_CAPT

2G

Capture cross section

B1_FISS

2G

Fission cross section

B1_NSF

2G

Fission neutron production cross section

B1_KAPPA

2G

Average deposited fission energy (MeV)

B1_INVV

2G

Inverse neutron speed (s/cm)

B1_NUBAR

2G

Average neutron yield

B1_ABS

2G

Absorption cross section (capture + fission)

B1_REMXS

2G

Removal cross section (groupremoval + absorption)

B1_RABSXS

2G

Reduced absorption cross section (total  scattering production)

Fission spectra
Parameter

Size

Description

B1_CHIT

2G

Fission spectrum (total)

B1_CHIP

2G

Fission spectrum (prompt neutrons)

B1_CHID

2G

Fission spectrum (delayed neutrons)

Scattering cross sections
Notes:
 Scattering production includes multiplying (n,2n), (n,3n), etc. reactions.
Parameter

Size

Description

B1_SCATT0

2G

Total P_{0} scattering cross section

B1_SCATT1

2G

Total P_{1} scattering cross section

B1_SCATT2

2G

Total P_{2} scattering cross section

B1_SCATT3

2G

Total P_{3} scattering cross section

B1_SCATT4

2G

Total P_{4} scattering cross section

B1_SCATT5

2G

Total P_{5} scattering cross section

B1_SCATT6

2G

Total P_{6} scattering cross section

B1_SCATT7

2G

Total P_{7} scattering cross section

B1_SCATTP0

2G

Total P_{0} scattering production cross section

B1_SCATTP1

2G

Total P_{1} scattering production cross section

B1_SCATTP2

2G

Total P_{2} scattering production cross section

B1_SCATTP3

2G

Total P_{3} scattering production cross section

B1_SCATTP4

2G

Total P_{4} scattering production cross section

B1_SCATTP5

2G

Total P_{5} scattering production cross section

B1_SCATTP6

2G

Total P_{6} scattering production cross section

B1_SCATTP7

2G

Total P_{7} scattering production cross section

Scattering matrices
Notes:
 Scattering production includes multiplying (n,2n), (n,3n), etc. reactions.
 The order of values ([input].coe) or value pairs ([input]_res.m) is: where refers to scattering from group g to g'.
 The data in the _res.m file can be read into a G by G matrix with Matlab reshapecommand, for example:
reshape(B1_S0(idx,1:2:end), G, G).
Parameter

Size

Description

B1_S0

2G^{2}

P_{0} scattering matrix

B1_S1

2G^{2}

P_{1} scattering matrix

B1_S2

2G^{2}

P_{2} scattering matrix

B1_S3

2G^{2}

P_{3} scattering matrix

B1_S4

2G^{2}

P_{4} scattering matrix

B1_S5

2G^{2}

P_{5} scattering matrix

B1_S6

2G^{2}

P_{6} scattering matrix

B1_S7

2G^{2}

P_{7} scattering matrix

B1_SP0

2G^{2}

P_{0} scattering production matrix

B1_SP1

2G^{2}

P_{1} scattering production matrix

B1_SP2

2G^{2}

P_{2} scattering production matrix

B1_SP3

2G^{2}

P_{3} scattering production matrix

B1_SP4

2G^{2}

P_{4} scattering production matrix

B1_SP5

2G^{2}

P_{5} scattering production matrix

B1_SP6

2G^{2}

P_{6} scattering production matrix

B1_SP7

2G^{2}

P_{7} scattering production matrix

Diffusion parameters
Parameter

Size

Description

B1_TRANSPXS

2G

Transport cross section (outscattering transport cross section collapsed with the critical spectrum when old B_{1} calculation mode is used, otherwise calculated from B1_DIFFCOEF)

B1_DIFFCOEF

2G

Diffusion coefficient calculated from during the fundamental mode calculation (old and new B_{1} and P_{1} calculation modes, or flux collapsed during the FM calculation mode)

Poison cross sections
Notes:
 Printed only if poison cross section option is on (see set poi).
 Xe135m values printed only if separate treatment of Xe135m is on (see set poi).
Parameter

Size

Description

B1_I135_YIELD

2G

Fission yield of I135 (cumulative, includes all precursors)

B1_XE135_YIELD

2G

Fission yield of Xe135

B1_XE135M_YIELD

2G

Fission yield of Xe135m

B1_PM149_YIELD

2G

Fission yield of Pm149 (cumulative, includes all precursors)

B1_SM149_YIELD

2G

Fission yield of Sm149

B1_I135_MICRO_ABS

2G

Microscopic absorption cross section of I135

B1_XE135_MICRO_ABS

2G

Microscopic absorption cross section of Xe135

B1_XE135M_MICRO_ABS

2G

Microscopic absorption cross section of Xe135m

B1_PM149_MICRO_ABS

2G

Microscopic absorption cross section of Pm149

B1_SM149_MICRO_ABS

2G

Microscopic absorption cross section of Sm149

B1_XE135_MACRO_ABS

2G

Macroscopic absorption cross section of Xe135

B1_XE135M_MACRO_ABS

2G

Macroscopic absorption cross section of Xe135m

B1_SM149_MACRO_ABS

2G

Macroscopic absorption cross section of Sm149

Delayed neutron data
Notes:
 The output consists of total, followed by D precursor groupwise values. In earlier versions, the output was fixed to 9 values independently of the library in use, with zero values corresponding to the empty precursor groups in the library.
 The actual number of groups depends on the cross section library used in the calculations. JEFF3.1, JEFF.3.2 and later evaluations use 8 precursor groups, while earlier evaluations, as well as all ENDF/B and JENDL data is based on 6 groups.
Parameter

Size

Description

BETA_EFF

2D + 2

Effective delayed neutron fraction (currently calculated using the Meulekamp method)

LAMBDA

2D + 2

Decay constants

Assembly discontinuity factors
Notes:
 Calculation of assembly discontinuity factors requires the set adf option.
 Surface flux and current tallies are used to calculate the boundary currents and fluxes. Midpoint and corner values are approximated by integrating over a small surface segment.
 The surface and volume fluxes are flux densities, i.e. they are surface or volume integrated fluxes divided by the respective surface area or volume.
 The currents are surface integrated values.
 The net current is defined as current in subtracted with current out.
 When the homogenized region is surrounded by reflective boundary conditions (zero netcurrent) the homogeneous flux becomes flat and equal to the volumeaveraged heterogeneous flux. When the net currents are nonzero, the homogeneous flux is obtained using the Builtin diffusion flux solver.
 The calculation currently supports only a limited number of surface types: infinite planes and square and hexagonal prisms.
 The order of surface and midpoint values for square prisms is: and the order of corner values: where refers to parameter on surface/corner k and energy group g.
 The order of surface values for Ytype hexagonal prims runs clockwise starting from the north, i.e. N, NE, SE, S, SW, NW. The corner values run counterclockwise starting from east, i.e. E, NE, NW, W, SW, SE.
 The order of surface values for Xtype hexagonal prims runs counterclockwise starting from the east, i.e. E, NE, NW, W, SW, SE. The corner values run clockwise starting from north, i.e. N, NE, SE, S, SW, NW.
 The sign moment weighted parameters are calculated only for surface types sqc, rect and hexxc.
 The convention of sign moment directions follows that of the nodal neutronics program Ants.
 The ADF symmetry options on set adf card are currently not used for sign moment weighted parameters.
Parameter

Size

Description

DF_SURFACE

(string)

Name of the surface used for the calculation

DF_SYM

1

Symmetry option defined in the input

DF_N_SURF

1

Number of surface values (denoted as N_{S} below)

DF_N_CORN

1

Number of corner values (denoted as N_{C} below)

DF_VOLUME

1

Volume (3D) or cross sectional area (2D) of the homogenized cell

DF_SURF_AREA

N_{S}

Area (3D) or perimeter length (2D) of the surface region

DF_MID_AREA

N_{S}

Area (3D) or perimeter length (2D) of the midpoint region

DF_CORN_AREA

N_{C}

Area (3D) or perimeter length (2D) of the corner region

DF_SURF_IN_CURR

2G N_{S}

Inward surface currents

DF_SURF_OUT_CURR

2G N_{S}

Outward surface currents

DF_SURF_NET_CURR

2G N_{S}

Net surface currents

DF_MID_IN_CURR

2G N_{S}

Inward midpoint currents

DF_MID_OUT_CURR

2G N_{S}

Outward midpoint currents

DF_MID_NET_CURR

2G N_{S}

Net midpoint currents

DF_CORN_IN_CURR

2G N_{C}

Inward corner currents

DF_CORN_OUT_CURR

2G N_{C}

Outward corner currents

DF_CORN_NET_CURR

2G N_{C}

Net corner currents

DF_HET_VOL_FLUX

2G

Heterogeneous flux over homogenized cell

DF_HET_SURF_FLUX

2G N_{S}

Heterogeneous surface fluxes

DF_HET_CORN_FLUX

2G N_{C}

Heterogeneous corner fluxes

DF_HOM_VOL_FLUX

2G

Homogeneous flux over homogenized cell

DF_HOM_SURF_FLUX

2G N_{S}

Homogeneous surface fluxes

DF_HOM_CORN_FLUX

2G N_{C}

Homogeneous corner fluxes

DF_SURF_DF

2G N_{S}

Surface discontinuity factors

DF_CORN_DF

2G N_{C}

Corner discontinuity factors

DF_SGN_SURF_IN_CURR

2G N_{S}

Inward sign moment weighted currents

DF_SGN_SURF_OUT_CURR

2G N_{S}

Outward sign moment weighted currents

DF_SGN_SURF_NET_CURR

2G N_{S}

Net sign moment weighted currents

DF_SGN_HET_SURF_FLUX

2G N_{S}

Heterogeneous sign moment weighted surface fluxes

DF_SGN_HOM_SURF_FLUX

2G N_{S}

Homogeneous sign moment weighted surface fluxes

DF_SGN_SURF_DF

2G N_{S}

Sign moment weighted surface discontinuity factors

Pinpower form factors
Notes:
 Calculation of pinpower form factors requires the set ppw option.
 The power distribution is calculated by tallying the fewgroup fission energy deposition in each lattice position and dividing the values with the total energy produced in the universe (sum over all values of PPW_POW equals 1).
 The calculation of form factors depends on the boundary conditions:
 If the homogenized region is surrounded by reflective boundary conditions (zero netcurrent), the homogeneous flux becomes flat and equal to the volumeaveraged heterogeneous flux.
 When the net currents are nonzero, the homogeneous flux is obtained using the builtin diffusion flux solver. The formfactors (PPW_FF) are obtained by dividing the pin and groupwise powers with the corresponding homogeneous diffusion flux (PPW_HOM_FLUX).
 However, if the net currents are nonzero, but the sum of the net currents is equal to zero, the volumeaveraged heterogeneous flux is used as the homogeneous flux, which is not an accurate approximation. This case is for example when modeling hexagonal fuel assemblies with other than 30 or 60 degree symmetries with periodic boundary conditions.
 Running the diffusion flux solver currently requires ADF calculation.
 The order of values is: where refers to parameter of pin n and energy group g. For example, twogroup power distributions in a 17 x 17 lattice can be converted into matrix form using the reshapecommand in Matlab:
P1 = reshape(PPW_POW(1, 1:4:end), 17, 17);
P2 = reshape(PPW_POW(1, 3:4:end), 17, 17);
 Symmetry used in the lattice may result in some pin powers and form factors to be for example 1/2, 1/4 or 1/8 of their true value, which have to be corrected during post processing of the values.
Parameter

Size

Description

PPW_LATTICE

(string)

Name of the lattice used for the calculation

PPW_LATTICE_TYPE

1

Lattice type (corresponds to the latcard)

PPW_PINS

1

Number of pin positions in the lattice (denoted as N_{P} below)

PPW_POW

2G N_{P}

Pin and groupwise power distribution normalized to unity sum

PPW_HOM_FLUX

2G N_{P}

Pin and groupwise homogeneous flux distribution

PPW_FF

2G N_{P}

Pin and groupwise form factors

Albedos
Notes:
 Calculation of albedos requires the set alb option.
 The order of the surfaces should be the same as for the ADFs.
 The order of ALB_IN_CURR is where refers to incoming partial current of surface k of group g.
 The order of ALB_OUT_CURR is where refers to outgoing partial current of surface k' of group g' which has entered the albedo surface through surface k and group g.
 The order of ALB_TOT_ALB is where refers to albedo from group g to g'.
 The order of ALB_PART_ALB is where refers to albedo of surface k' of group g' which has entered the albedo surface through surface k and group g.
 For example, twogroup hexagonal partial albedos can be converted into matrix form using the reshapecommand in Matlab with the notation part_alb(g', k', g, k) as
part_alb = reshape(ALB_PART_ALB(1, 1:2:end), 2, 6, 2, 6)
Parameter

Size

Description

ALB_SURFACE

(string)

Name of the surface used for the calculation

ALB_FLIP_DIR

1


ALB_N_SURF

1

Number of albedo surface faces (denoted as N_{S} below)

ALB_IN_CURR

2G N_{S}

Groupwise incoming partial currents of albedo surface faces

ALB_OUT_CURR

2G^{2} N_{S}^{2}

Outgoing group to group and face to face outgoing partial currents

ALB_TOT_ALB

2G^{2}

Total group to group albedos for the entire albedo surface

ALB_PART_ALB

2G^{2} N_{S}^{2}

Partial group to group and face to face albedos

Miscellaneous notes for other outputs
Delayed neutrons accounted for in ANA_KEFF
Since Serpent 2.1.23, ANA_KEFF estimator is calculated separately for delayed neutrons. The first two values are total, 34 are prompt neutron multiplication only and 56 delayed neutron multiplication only. ^{[3]}
References
 ^ Leppänen, J., Pusa, M. and Fridman, E. "Overview of methodology for spatial homogenization in the Serpent 2 Monte Carlo code." Ann. Nucl. Energy, 96 (2016) 126136.
 ^ Liu, Z., Smith, K., Forget, B. and Ortensi, J."Cumulative migration method for computing rigorous diffusion coefficients and transport cross sections from Monte Carlo." Ann. Nucl. Energy, 118 (2018) 507516.
 ^
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