Difference between revisions of "Output parameters"

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=== Equilibrium poison iteration ===
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=== Equilibrium Xe-135 iteration ===
  
 
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| Averaged equilibrium I-135 concentration (see [[Input syntax manual#set xenon|set xenon]] input option)
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=== Equilibrium Sm-149 iteration ===
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Revision as of 16:07, 22 June 2021

This page lists the output parameters in the main [input]_res.m output file.

Contents

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 Multi-group 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) Neutron-induced 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 delta-tracking (see the set dt input option)
ST_FRAC 2/2 Fraction of paths sampled using surface-tracking
DT_FRAC 2/2 Fraction of paths sampled using delta-tracking
DT_EFF 2/2 Delta-tracking 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 cycle-wise value
Parameter Size Description
TOT_CPU_TIME 1 Total CPU time
RUNNING_TIME 1 Total wall-clock running time
INIT_TIME 1(2) Wall-clock time spent for initialization
PROCESS_TIME 1(2) Wall-clock time spent for processing
TRANSPORT_CYCLE_TIME 1(2) Wall-clock time spent for transport simulation
BURNUP_CYCLE_TIME 1(2) Wall-clock time spent for burnup solution
BATEMAN_SOLUTION_TIME 1(2) Wall-clock time spent for solving the Bateman equations
MPI_OVERHEAD_TIME 1(2) Wall-clock time spent MPI communication
ESTIMATED_RUNNING_TIME 1(2) Estimated total wall-clock 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 material-wise data
RES_MEMSIZE 1 Memory size used for storing results
IFC_MEMSIZE 1 Memory size used for data for response-matrix 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 cut-off 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 Doppler-broadening 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 Doppler-broadening 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 pseudo-Q-value 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 Q-value.
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 non-critical 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

Parameters for burnup calculation

Parameter Size Description
BURN_MATERIALS
BURN_MODE
BURN_STEP
BURN_RANDOMIZE_DATA
BURNUP
BURN_DAYS
FIMA

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 Xe-135 iteration

Parameter Size Description
XE135_EQUIL_CONC 2 Averaged equilibrium Xe-135 concentration (see set xenon input option)
I135_EQUIL_CONC 2 Averaged equilibrium I-135 concentration (see set xenon input option)

Equilibrium Sm-149 iteration

Parameter Size Description
SM149_EQUIL_CONC 2 Averaged equilibrium Sm-149 concentration (see set samarium input option)
PM149_EQUIL_CONC 2 Averaged equilibrium Pm-149 concentration (see set samarium input option)

Six-factor 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 non-leakage probability
SIX_FF_LT 2 Analog estimate of thermal non-leakage probability
SIX_FF_KINF 2 Analog estimate of six-factor kinf (four-factor keff)
SIX_FF_KEFF 2 Analog estimate of six-factor keff

Fission neutron and energy production

Parameter Size Description
NUBAR
FISSE

Criticality eigenvalues

Parameter Size Description
ANA_KEFF
IMP_KEFF
COL_KEFF
ABS_KEFF
ABS_KINF
GEOM_ALBEDO

ALF (Average lethargy of neutrons causing fission)

Parameter Size Description
ANA_ALF
IMP_ALF

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

Parameter Size Description
ANA_EALF
IMP_EALF

Time constants

Forward-weighted delayed neutron parameters

Parameter Size Description
PRECURSOR_GROUPS
FWD_ANA_BETA_ZERO
FWD_ANA_LAMBDA

Beta-eff using Meulekamp's method

Parameter Size Description
ADJ_MEULEKAMP_BETA_EFF
ADJ_MEULEKAMP_LAMBDA

Adjoint weighted time constants using Nauchi's method

Parameter Size Description
IFP_CHAIN_LENGTH
ADJ_NAUCHI_GEN_TIME
ADJ_NAUCHI_LIFETIME
ADJ_NAUCHI_BETA_EFF
ADJ_NAUCHI_LAMBDA

Adjoint weighted time constants using IFP

Parameter Size Description
ADJ_IFP_GEN_TIME
ADJ_IFP_LIFETIME
ADJ_IFP_IMP_BETA_EFF
ADJ_IFP_IMP_LAMBDA
ADJ_IFP_ANA_BETA_EFF
ADJ_IFP_ANA_LAMBDA
ADJ_IFP_ROSSI_ALPHA

Adjoint weighted time constants using perturbation technique

Parameter Size Description
ADJ_PERT_GEN_TIME
ADJ_PERT_LIFETIME
ADJ_PERT_BETA_EFF
ADJ_PERT_ROSSI_ALPHA

Inverse neutron speed

Parameter Size Description
ANA_INV_SPD

Analog slowing-down and thermal neutron lifetime (total/prompt/delayed)

Parameter Size Description
ANA_SLOW_TIME
ANA_THERM_TIME
ANA_THERM_FRAC
ANA_DELAYED_EMTIME
ANA_MEAN_NCOL

Homogenized group constants

Notes:

  • Group constants are calculated by first homogenizing the geometry using a multi-group structure with H energy groups. The data is then collapsed into the final few-group structure with G groups using the infinite and leakage-corrected 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 multi-group structure is defined using option set micro or set fum.
  • The few-group 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 two-dimensional data (scattering matrices, ADF and pin-power 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 multi-group structure (referred to as H below)
MICRO_E H + 1 Group boundaries in the intermediate multi-group structure (in ascending order)
MACRO_NG 1 Number of energy groups in the final few-group structure (referred to as G below)
MACRO_E G + 1 Group boundaries in the final few-group structure (in descending order)

Group constants homogenized in infinite spectrum

Parameter Size Description
INF_MICRO_FLX 2H Multi-group flux spectrum (integral, un-normalized)
INF_FLX 2G Few-group 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 (group-removal + 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 P0 scattering cross section
INF_SCATT1 2G Total P1 scattering cross section
INF_SCATT2 2G Total P2 scattering cross section
INF_SCATT3 2G Total P3 scattering cross section
INF_SCATT4 2G Total P4 scattering cross section
INF_SCATT5 2G Total P5 scattering cross section
INF_SCATT6 2G Total P6 scattering cross section
INF_SCATT7 2G Total P7 scattering cross section
INF_SCATTP0 2G Total P0 scattering production cross section
INF_SCATTP1 2G Total P1 scattering production cross section
INF_SCATTP2 2G Total P2 scattering production cross section
INF_SCATTP3 2G Total P3 scattering production cross section
INF_SCATTP4 2G Total P4 scattering production cross section
INF_SCATTP5 2G Total P5 scattering production cross section
INF_SCATTP6 2G Total P6 scattering production cross section
INF_SCATTP7 2G Total P7 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: \Sigma_{1,1} \, \Sigma_{1,2} \, ... \, \Sigma_{2,1} \, \Sigma_{2,2}  \, ... where \Sigma_{g,g'} 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 reshape-command, for example:
reshape(INF_S0(idx,1:2:end), G, G);
Parameter Size Description
INF_S0 2G2 P0 scattering matrix
INF_S1 2G2 P1 scattering matrix
INF_S2 2G2 P2 scattering matrix
INF_S3 2G2 P3 scattering matrix
INF_S4 2G2 P4 scattering matrix
INF_S5 2G2 P5 scattering matrix
INF_S6 2G2 P6 scattering matrix
INF_S7 2G2 P7 scattering matrix
INF_SP0 2G2 P0 scattering production matrix
INF_SP1 2G2 P1 scattering production matrix
INF_SP2 2G2 P2 scattering production matrix
INF_SP3 2G2 P3 scattering production matrix
INF_SP4 2G2 P4 scattering production matrix
INF_SP5 2G2 P5 scattering production matrix
INF_SP6 2G2 P6 scattering production matrix
INF_SP7 2G2 P7 scattering production matrix

Diffusion parameters

Notes:

  • Calculation of sensible values for INF_TRANSPXS and INF_DIFFCOEF requires fine enough intermediate multi-group structure.
  • The cumulative migration method [2] was first developed for the OpenMC code. Currently the method works only when the homogenized region covers the entire geometry, and is surrounded by periodic or reflective boundary conditions.
  • Calculation of TRC_TRANSPXS and TRC_DIFFCOEF requires defining energy-dependent 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 out-scattering approximation)
INF_DIFFCOEF 2G Diffusion coefficient (calculated using the out-scattering approximation)
CMM_TRANSPXS 2G Transport cross section calculated using the cumulative migration method (equivalent with the in-scattering approximation)
CMM_TRANSPXS_X 2G X-component of the directional transport cross section calculated using the cumulative migration method (equivalent with the in-scattering approximation)
CMM_TRANSPXS_Y 2G Y-component of the directional transport cross section calculated using the cumulative migration method (equivalent with the in-scattering approximation)
CMM_TRANSPXS_Z 2G Z-component of the directional transport cross section calculated using the cumulative migration method (equivalent with the in-scattering approximation)
CMM_DIFFCOEF 2G Diffusion coefficient calculated using the cumulative migration method (equivalent with the in-scattering approximation)
CMM_DIFFCOEF_X 2G X-component of the directional diffusion coefficient calculated using the cumulative migration method (equivalent with the in-scattering approximation)
CMM_DIFFCOEF_Y 2G Y-component of the directional diffusion coefficient calculated using the cumulative migration method (equivalent with the in-scattering approximation)
CMM_DIFFCOEF_Z 2G Z-component of the directional diffusion coefficient calculated using the cumulative migration method (equivalent with the in-scattering approximation)
TRC_TRANSPXS 2G Transport cross section calculated by applying user-defined transport correction factors to total cross section
TRC_DIFFCOEF 2G Diffusion coefficient calculated by applying user-defined transport correction factors to total cross section

Poison cross sections

Notes:

  • Printed only if poison cross section option is on (see set poi).
Parameter Size Description
INF_I135_YIELD 2G Fission yield of I-135 (cumulative, includes all precursors)
INF_XE135_YIELD 2G Fission yield of Xe-135
INF_PM149_YIELD 2G Fission yield of Pm-149 (cumulative, includes all precursors)
INF_SM149_YIELD 2G Fission yield of Sm-149
INF_I135_MICRO_ABS 2G Microscopic absorption cross section of I-135
INF_XE135_MICRO_ABS 2G Microscopic absorption cross section of Xe-135
INF_PM149_MICRO_ABS 2G Microscopic absorption cross section of Pm-149
INF_SM149_MICRO_ABS 2G Microscopic absorption cross section of Sm-149
INF_XE135_MACRO_ABS 2G Macroscopic absorption cross section of Xe-135
INF_SM149_MACRO_ABS 2G Macroscopic absorption cross section of Sm-149

Group constants homogenized in leakage-corrected spectrum

Parameter Size Description
B1_MICRO_FLX 2H Multi-group flux spectrum (integral, un-normalized)
B1_FLX 2G Few-group 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 keff 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 (group-removal + 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 P0 scattering cross section
B1_SCATT1 2G Total P1 scattering cross section
B1_SCATT2 2G Total P2 scattering cross section
B1_SCATT3 2G Total P3 scattering cross section
B1_SCATT4 2G Total P4 scattering cross section
B1_SCATT5 2G Total P5 scattering cross section
B1_SCATT6 2G Total P6 scattering cross section
B1_SCATT7 2G Total P7 scattering cross section
B1_SCATTP0 2G Total P0 scattering production cross section
B1_SCATTP1 2G Total P1 scattering production cross section
B1_SCATTP2 2G Total P2 scattering production cross section
B1_SCATTP3 2G Total P3 scattering production cross section
B1_SCATTP4 2G Total P4 scattering production cross section
B1_SCATTP5 2G Total P5 scattering production cross section
B1_SCATTP6 2G Total P6 scattering production cross section
B1_SCATTP7 2G Total P7 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: \Sigma_{1,1} \, \Sigma_{1,2} \, ... \, \Sigma_{2,1} \, \Sigma_{2,2}  \, ... where \Sigma_{g,g'} 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 reshape-command, for example:
reshape(B1_S0(idx,1:2:end), G, G).
Parameter Size Description
B1_S0 2G2 P0 scattering matrix
B1_S1 2G2 P1 scattering matrix
B1_S2 2G2 P2 scattering matrix
B1_S3 2G2 P3 scattering matrix
B1_S4 2G2 P4 scattering matrix
B1_S5 2G2 P5 scattering matrix
B1_S6 2G2 P6 scattering matrix
B1_S7 2G2 P7 scattering matrix
B1_SP0 2G2 P0 scattering production matrix
B1_SP1 2G2 P1 scattering production matrix
B1_SP2 2G2 P2 scattering production matrix
B1_SP3 2G2 P3 scattering production matrix
B1_SP4 2G2 P4 scattering production matrix
B1_SP5 2G2 P5 scattering production matrix
B1_SP6 2G2 P6 scattering production matrix
B1_SP7 2G2 P7 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 B1 calculation mode is used, otherwise calculated from B1_DIFFCOEF)
B1_DIFFCOEF 2G Diffusion coefficient calculated from during the fundamental mode calculation (old and new B1 and P1 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).
Parameter Size Description
B1_I135_YIELD 2G Fission yield of I-135 (cumulative, includes all precursors)
B1_XE135_YIELD 2G Fission yield of Xe-135
B1_PM149_YIELD 2G Fission yield of Pm-149 (cumulative, includes all precursors)
B1_SM149_YIELD 2G Fission yield of Sm-149
B1_I135_MICRO_ABS 2G Microscopic absorption cross section of I-135
B1_XE135_MICRO_ABS 2G Microscopic absorption cross section of Xe-135
B1_PM149_MICRO_ABS 2G Microscopic absorption cross section of Pm-149
B1_SM149_MICRO_ABS 2G Microscopic absorption cross section of Sm-149
B1_XE135_MACRO_ABS 2G Macroscopic absorption cross section of Xe-135
B1_SM149_MACRO_ABS 2G Macroscopic absorption cross section of Sm-149

Delayed neutron data

Notes:

  • The output always consists of 9 values: total, followed by precursor group-wise values. If the number of groups is 6, the last two values are zero.
  • The actual number of groups depends on the cross section library used in the calculations. JEFF-3.1, 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 9 Effective delayed neutron fraction (currently calculated using the Meulekamp method)
LAMBDA 9 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. Mid-point 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 net-current) the homogeneous flux becomes flat and equal to the volume-averaged heterogeneous flux. When the net currents are non-zero, the homogeneous flux is obtained using the Built-in 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 mid-point values for square prisms is: X_{\mathrm{W},1} \, X_{\mathrm{W},2} \, ... \, X_{\mathrm{S},1} \, X_{\mathrm{S},2} \, ... \, X_{\mathrm{E},1} \, X_{\mathrm{E},2} \, ... \, X_{\mathrm{N},1} \, X_{\mathrm{N},2} \, ... and the order of corner values: X_{\mathrm{NW},1} \, X_{\mathrm{NW},2} \, ... \, X_{\mathrm{NE},1} \, X_{\mathrm{NE},2} \, ... \, X_{\mathrm{SE},1} \, X_{\mathrm{SE},2} \, ... \, X_{\mathrm{SW},1} \, X_{\mathrm{SW},2} \, ... where X_{k,g} refers to parameter X on surface/corner k and energy group g.
  • The order of surface values for Y-type 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 X-type 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 NS below)
DF_N_CORN 1 Number of corner values (denoted as NC below)
DF_VOLUME 1 Volume (3D) or cross sectional area (2D) of the homogenized cell
DF_SURF_AREA NS Area (3D) or perimeter length (2D) of the surface region
DF_MID_AREA NS Area (3D) or perimeter length (2D) of the mid-point region
DF_CORN_AREA NC Area (3D) or perimeter length (2D) of the corner region
DF_SURF_IN_CURR 2G \times NS Inward surface currents
DF_SURF_OUT_CURR 2G \times NS Outward surface currents
DF_SURF_NET_CURR 2G \times NS Net surface currents
DF_MID_IN_CURR 2G \times NS Inward mid-point currents
DF_MID_OUT_CURR 2G \times NS Outward mid-point currents
DF_MID_NET_CURR 2G \times NS Net mid-point currents
DF_CORN_IN_CURR 2G \times NC Inward corner currents
DF_CORN_OUT_CURR 2G \times NC Outward corner currents
DF_CORN_NET_CURR 2G \times NC Net corner currents
DF_HET_VOL_FLUX 2G Heterogeneous flux over homogenized cell
DF_HET_SURF_FLUX 2G \times NS Heterogeneous surface fluxes
DF_HET_CORN_FLUX 2G \times NC Heterogeneous corner fluxes
DF_HOM_VOL_FLUX 2G Homogeneous flux over homogenized cell
DF_HOM_SURF_FLUX 2G \times NS Homogeneous surface fluxes
DF_HOM_CORN_FLUX 2G \times NC Homogeneous corner fluxes
DF_SURF_DF 2G \times NS Surface discontinuity factors
DF_CORN_DF 2G \times NC Corner discontinuity factors
DF_SGN_SURF_IN_CURR 2G \times NS Inward sign moment weighted currents
DF_SGN_SURF_OUT_CURR 2G \times NS Outward sign moment weighted currents
DF_SGN_SURF_NET_CURR 2G \times NS Net sign moment weighted currents
DF_SGN_HET_SURF_FLUX 2G \times NS Heterogeneous sign moment weighted surface fluxes
DF_SGN_HOM_SURF_FLUX 2G \times NS Homogeneous sign moment weighted surface fluxes
DF_SGN_SURF_DF 2G \times NS Sign moment weighted surface discontinuity factors

Pin-power form factors

Notes:

  • Calculation of pin-power form factors requires the set ppw option.
  • The power distribution is calculated by tallying the few-group 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:
    1. If the homogenized region is surrounded by reflective boundary conditions (zero net-current), the homogeneous flux becomes flat and equal to the volume-averaged heterogeneous flux.
    2. When the net currents are non-zero, the homogeneous flux is obtained using the built-in diffusion flux solver. The form-factors (PPW_FF) are obtained by dividing the pin- and group-wise powers with the corresponding homogeneous diffusion flux (PPW_HOM_FLUX).
    3. However, if the net currents are non-zero, but the sum of the net currents is equal to zero, the volume-averaged 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: X_{1,1} \, X_{1,2} \, ... \, X_{2,1} \, X_{2,2} \, ... where X_{n,g} refers to parameter X of pin n and energy group g. For example, two-group power distributions in a 17 x 17 lattice can be converted into matrix form using the reshape-command 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 lat-card)
PPW_PINS 1 Number of pin positions in the lattice (denoted as NP below)
PPW_POW 2G \times NP Pin- and group-wise power distribution normalized to unity sum
PPW_HOM_FLUX 2G \times NP Pin- and group-wise homogeneous flux distribution
PPW_FF 2G \times NP Pin- and group-wise 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 J_{1,1} , J_{1,2} , \ldots , J_{2,1} , J_{2,2} , \ldots where J_{k,g} refers to incoming partial current of surface k of group g.
  • The order of ALB_OUT_CURR is J_{1,1,1,1} , J_{1,1,1,2} , \ldots , J_{1,1,2,1} , J_{1,1,2,2} , \ldots , J_{1,2,1,1} , J_{1,2,1,2} , \ldots , J_{2,1,1,1} , J_{2,1,1,2} , \ldots where J_{k,g,k',g'} 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 \alpha_{1,1} , \alpha_{1,2} , \ldots ,  \alpha_{2,1} ,  \alpha_{2,2} , \ldots where \alpha_{g,g'} refers to albedo from group g to g'.
  • The order of ALB_PART_ALB is \alpha_{1,1,1,1} , \alpha_{1,1,1,2} , \ldots , \alpha_{1,1,2,1} , \alpha_{1,1,2,2} , \ldots , \alpha_{1,2,1,1} , \alpha_{1,2,1,2} , \ldots , \alpha_{2,1,1,1} , \alpha_{2,1,1,2} , \ldots where \alpha_{k,g,k',g'} refers to albedo of surface k' of group g' which has entered the albedo surface through surface k and group g.
  • For example, two-group hexagonal partial albedos can be converted into matrix form using the reshape-command in Matlab with the notation part_alb(g', k', g, k) = \alpha_{k,g,k',g'} 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 NS below)
ALB_IN_CURR 2G \times NS Groupwise incoming partial currents of albedo surface faces
ALB_OUT_CURR 2G2 \times NS2 Outgoing group to group and face to face outgoing partial currents
ALB_TOT_ALB 2G2 Total group to group albedos for the entire albedo surface
ALB_PART_ALB 2G2 \times NS2 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, 3-4 are prompt neutron multiplication only and 5-6 delayed neutron multiplication only. [3]

References

  1. ^ 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) 126-136.
  2. ^ 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) 507-516.
  3. ^ http://ttuki.vtt.fi/serpent/viewtopic.php?f=25&t=1885&p=4469