Difference between revisions of "Output parameters"

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−
This page lists the output parameters in the main [input]_res.m output file.
+
This page lists the output parameters in the main <tt>[input]_res.m</tt> output file.
  
 
== General output parameters ==
 
== General output parameters ==
Line 70: Line 70:
 
| POP
 
| POP
 
| 1
 
| 1
−
| Population size defined using the [[Input syntax manual#set pop|set pop]] input option
+
| Population size defined using the [[Input syntax manual#set pop|set pop]] input option (criticality) or the [[Input syntax manual#set nps|set nps]] input option (external source)
 
|-
 
|-
 
| CYCLES
 
| CYCLES
Line 83: Line 83:
 
| 1
 
| 1
 
| Batching interval defined using the [[Input syntax manual#set pop|set pop]] input option
 
| Batching interval defined using the [[Input syntax manual#set pop|set pop]] input option
−
|-
 
−
| POP
 
−
|
 
−
|
 
 
|-
 
|-
 
| BATCHES
 
| BATCHES
−
|
+
| 1
−
|
+
| Number of batches defined using the [[Input syntax manual#set nps|set nps]] input option
 
|-
 
|-
 
| SRC_NORM_MODE
 
| SRC_NORM_MODE
Line 134: Line 130:
 
| IMPLICIT_REACTION_RATES
 
| IMPLICIT_REACTION_RATES
 
| 1
 
| 1
−
| Flag indicating whether or implicit reaction rates are used for group constant generation
+
| Flag indicating whether or not implicit reaction rates are used for group constant generation
 
|-
 
|-
 
| VR_ITER_IDX
 
| VR_ITER_IDX
−
|
+
| 1
−
|
+
| Variance reduction iteration index when output was printed (see [[Input syntax manual#wwin (weight window mesh definition)|wwin]] card)
 +
|-
 +
|}
 +
 
 +
=== Domain decomposition ===
 +
 
 +
{|class="wikitable" style="text-align: left;"
 +
! Parameter
 +
! Size
 +
! Description
 +
|-
 +
| DD_MODE
 +
| 1
 +
| Domain decomposition mode defined using the [[Input syntax manual#set dd|set dd]] input option
 +
|-
 +
| DD_NEUTRONS_TO_LIMBO
 +
| ''M''
 +
| Neutrons sent to limbo (transferral buffer) at each domain
 +
|-
 +
| DD_NEUTRONS_FROM_LIMBO
 +
| ''M''
 +
| Neutrons received from limbo (transferral buffer) at each domain
 
|-
 
|-
 
|}
 
|}
Line 215: Line 232:
  
 
=== File paths ===
 
=== File paths ===
 +
<u>Notes:</u>
 +
 +
* Only the first file path listed is displayed
  
 
{|class="wikitable" style="text-align: left;"
 
{|class="wikitable" style="text-align: left;"
Line 242: Line 262:
 
|-
 
|-
 
| PHOTON_PHYS_DIRECTORY
 
| PHOTON_PHYS_DIRECTORY
−
|
+
| (string)
−
|
+
| Photon physics directory path defined using the [[Input syntax manual#set pdatadir|set pdatadir]] input option
 
|-
 
|-
 
|}
 
|}
Line 258: Line 278:
 
! Size
 
! Size
 
! Description
 
! Description
 +
|-
 +
| MEAN_SRC_WGT
 +
| 2/2
 +
| Mean source weight for non-criticality calculations (neutrons/photons)
 +
|-
 +
| SOURCE_SAMPLING_EFF
 +
| 2/2
 +
| Source sampling efficiency for non-criticality calculations (neutrons/photons)
 +
|-
 +
| MEAN_SRC_WW_SPLIT
 +
| 2/2
 +
| Mean source weight-window splitting in variance reduction (neutrons/photons)
 +
|-
 +
| MEAN_SRC_WW_EFF
 +
| 2/2
 +
| Mean source weight-window sampling efficiency in variance reduction (neutrons/photons)
 +
|-
 +
| WW_BALA_ROULETE
 +
| 2/2
 +
| Mean weight-window balance due to Russian roulette in variance reduction (neutrons/photons)
 +
|-
 +
| WW_BALA_SPLIT
 +
| 2/2
 +
| Mean weight-window balance due to splitting in variance reduction (neutrons/photons)
 
|-
 
|-
 
| MIN_MACROXS
 
| MIN_MACROXS
Line 278: Line 322:
 
| 2/2
 
| 2/2
 
| Delta-tracking efficiency
 
| Delta-tracking efficiency
 +
|-
 +
| IFC_COL_EFF
 +
| 2/2
 +
| Efficiency of interface collision rejection (see [[Input syntax manual#ifc (interface file)|ifc]] card)
 
|-
 
|-
 
| REA_SAMPLING_EFF
 
| REA_SAMPLING_EFF
Line 286: Line 334:
 
| 2/2
 
| 2/2
 
| Fraction of failed reaction samples
 
| Fraction of failed reaction samples
 +
|-
 +
| TMS_SAMPLING_EFF
 +
| 2
 +
| Target motion sampling method, TMS, sampling efficiency (see [[Input syntax manual#mat (material definition)|tms]] option, in  [[Input syntax manual#mat (material definition)|mat]] card)
 
|-
 
|-
 
| TOT_COL_EFF
 
| TOT_COL_EFF
Line 294: Line 346:
 
| 2/2, 2/2
 
| 2/2, 2/2
 
| Average number of tracking loops per history and, fraction of failed tracking loops
 
| Average number of tracking loops per history and, fraction of failed tracking loops
 +
|-
 +
| TMS_FAIL_STAT
 +
| 8
 +
| TMS fail statistics: total samples, majorant fail, lower limit fail, upper limit fail (see [[Input syntax manual#mat (material definition)|tms]] option, in  [[Input syntax manual#mat (material definition)|mat]] card)
 +
|-
 +
| DBRC_EXCEED_FRAC
 +
| 1
 +
| Doppler-broadening rejection correction, DBRC, majorant exceed fraction (see [[Input syntax manual#set dbrc|set dbrc]] input option)
 
|-
 
|-
 
| AVG_TRACKS
 
| AVG_TRACKS
Line 314: Line 374:
 
| 1
 
| 1
 
| Number of lost particles
 
| Number of lost particles
 +
|-
 +
|}
 +
 +
=== STL geometries ===
 +
 +
{|class="wikitable" style="text-align: left;"
 +
! Parameter
 +
! Size
 +
! Description
 +
|-
 +
| STL_RAY_TEST
 +
| 5
 +
| STL-ray-tracing test: total, ray is too parallel to facet, intersection point is too close to edge, facet is too close to search mesh cell boundary, two facets overlap
 +
|-
 +
| STL_ENFORCE_ST
 +
| 1
 +
| Flag indicating whether or not delta tracking is enforced in STL geometries
 +
|-
 +
|}
 +
 +
=== Importance solver ===
 +
 +
{|class="wikitable" style="text-align: left;"
 +
! Parameter
 +
! Size
 +
! Description
 +
|-
 +
| NEIGHBOUR_SEARCH_FAIL
 +
| 2
 +
| Response matrix calculation fail rate
 
|-
 
|-
 
|}
 
|}
Line 374: Line 464:
 
|-
 
|-
 
| TRANSPORT_CYCLE_TIME
 
| TRANSPORT_CYCLE_TIME
−
| 1(2)
+
| 1(3)
 
| Wall-clock time spent for transport simulation
 
| Wall-clock time spent for transport simulation
 +
|-
 +
| FINIX_SOLUTION_TIME
 +
| 1
 +
| Wall-clock time spent for FINIX solution
 
|-
 
|-
 
| BURNUP_CYCLE_TIME
 
| BURNUP_CYCLE_TIME
Line 390: Line 484:
 
|-
 
|-
 
| DD_OVERHEAD_TIME
 
| DD_OVERHEAD_TIME
−
|
+
| 1
−
|
+
| Wall-clock time spent for DD algorithm (see [[Input syntax manual#set dd|set dd]] input option)
 
|-
 
|-
 
| RMX_SOLUTION_TIME
 
| RMX_SOLUTION_TIME
−
|
+
| 1
−
|
+
| Wall-clock time spent for response matrix solution (see [[Input syntax manual#wwgen (response matrix based importance map solver)|wwgen]] / [[Input syntax manual#wwin (weight window mesh definition)|wwin]] cards or [[Input syntax manual#set sca|set sca]] input option)
 
|-
 
|-
 
| LEAKAGE_CORR_SOL_TIME
 
| LEAKAGE_CORR_SOL_TIME
−
|
+
| 1
−
|
+
| Wall-clock time spent for leakage correction solution (see [[Input syntax manual#set fum|set fum]] input option)
 
|-
 
|-
 
| ESTIMATED_RUNNING_TIME
 
| ESTIMATED_RUNNING_TIME
Line 456: Line 550:
 
| IFC_MEMSIZE
 
| IFC_MEMSIZE
 
| 1
 
| 1
−
| Memory size used for data for response-matrix solver
+
| Memory size used for data for multi-physics interface data
 
|-
 
|-
 
| RMX_MEMSIZE
 
| RMX_MEMSIZE
−
|
+
| 1
−
|
+
| Memory size used for storing response matrix-wise data
 
|-
 
|-
 
| MISC_MEMSIZE
 
| MISC_MEMSIZE
Line 502: Line 596:
 
| NEUTRON_ERG_TOL
 
| NEUTRON_ERG_TOL
 
| 1
 
| 1
−
| Reconstruction tolerace for unionized energy grid
+
| Reconstruction tolerance for unionized energy grid (see [[Input syntax manual#set egrid|set egrid]] input option)
 
|-
 
|-
 
| NEUTRON_ERG_NE
 
| NEUTRON_ERG_NE
 
| 1
 
| 1
−
| Number of points in unionized energy grid
+
| Number of points in neutron unionized energy grid
 
|-
 
|-
 
| NEUTRON_EMIN
 
| NEUTRON_EMIN
 
| 1
 
| 1
−
| Minimum energy for neutron cross section data
+
| Minimum energy for neutron cross section data (see [[Input syntax manual#set egrid|set egrid]] input option)
 
|-
 
|-
 
| NEUTRON_EMAX
 
| NEUTRON_EMAX
 
| 1
 
| 1
−
| Maximum energy for neutron cross section data
+
| Maximum energy for neutron cross section data (see [[Input syntax manual#set egrid|set egrid]] input option)
 +
|-
 +
|}
 +
 
 +
=== Photon energy grid ===
 +
 
 +
{|class="wikitable" style="text-align: left;"
 +
! Parameter
 +
! Size
 +
! Description
 +
|-
 +
| PHOTON_ERG_NE
 +
| 1
 +
| Number of points in photon unionized energy grid
 +
|-
 +
| PHOTON_EMIN
 +
| 1
 +
| Minimum energy for photon cross section data (see [[Input syntax manual#set egrid|set egrid]] input option)
 +
|-
 +
| PHOTON_EMAX
 +
| 1
 +
| Maximum energy for photon cross section data (see [[Input syntax manual#set egrid|set egrid]] input option)
 
|-
 
|-
 
|}
 
|}
Line 527: Line 642:
 
| URES_DILU_CUT
 
| URES_DILU_CUT
 
| 1
 
| 1
−
| Density cut-off used for unresolved resonance probability table sampling
+
| Density cut-off used for unresolved resonance probability table sampling (see [[Input syntax manual#set ures|set ures]] input option)
 
|-
 
|-
 
| URES_EMIN
 
| URES_EMIN
Line 543: Line 658:
 
| URES_USED
 
| URES_USED
 
| 1
 
| 1
−
| Number of nuclides for which probability table sampling was used
+
| Number of nuclides for which probability table sampling was used (see [[Input syntax manual#set ures|set ures]] input option)
 
|-
 
|-
 
|}
 
|}
Line 616: Line 731:
 
| 1
 
| 1
 
| Flag indicating whether or not  implicit fission reaction mode is on (see [[Input syntax manual#set impl|set impl]] input option)
 
| Flag indicating whether or not  implicit fission reaction mode is on (see [[Input syntax manual#set impl|set impl]] input option)
 +
|-
 +
| IMPL_FISS_NUBAR
 +
| 1
 +
| Flag indicating whether or not implicit fission nubar reaction mode is on (see [[Input syntax manual#set impl|set impl]] input option)
 
|-
 
|-
 
| DOPPLER_PREPROCESSOR       
 
| DOPPLER_PREPROCESSOR       
Line 639: Line 758:
 
|}
 
|}
  
−
=== Energy deposition ===
+
=== Photon physics options ===
−
<u>Notes:</u>
 
−
 
 
−
* 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.
 
  
 
{|class="wikitable" style="text-align: left;"
 
{|class="wikitable" style="text-align: left;"
Line 649: Line 765:
 
! Description
 
! Description
 
|-
 
|-
−
| EDEP_MODE               
+
| COMPTON_EKN
−
| 1  
+
| 1
−
| Energy deposition mode (see [[Input syntax manual#set edepmode|set edepmode]] input option)
+
| Photon energy above which Klein-Nishina is used for calculating energy and direction of the scattered photons (see [[Input syntax manual#set ekn|set ekn]] input option)
 
|-
 
|-
−
| EDEP_DELAYED             
+
| COMPTON_DOPPLER
 
| 1
 
| 1
−
| Energy of delayed components in energy deposition calculations (see [[Input syntax manual#set edepdel|set edepdel]] input option)
+
| Flag indicating whether or not Doppler broadening method for the energy spectrum of the scattered photons is on (see [[Input syntax manual#set cdop|set cdop]] input option)
 
|-
 
|-
−
| EDEP_KEFF_CORR           
+
| COMPTON_EANG
−
| 1  
+
| 1
−
| Flag indicating whether or not correction for energy deposition estimates in non-critical systems (see [[Input syntax manual#set edepkcorr |set edepkcorr]] input option)
+
| Flag indicating whether or not Compton electron angular distribution model is on (see [[Input syntax manual#set cea|set cea]] input option)
 
|-
 
|-
−
| EDEP_LOCAL_EGD           
+
| PHOTON_TTB
 
| 1
 
| 1
−
| Energy distribution of delayed components in energy deposition calculations, mode 3 (see [[Input syntax manual#set edepdel|set edepdel]] input option)
+
| Flag indicating whether or not thick-target bremsstrahlung approximation for modelling electrons and positrons is on (see [[Input syntax manual#set ttb|set ttb]] input option)
 
|-
 
|-
−
| EDEP_COMP               
+
|}
−
| 9
+
 
−
| Fission energy release components: EFR, ENP, END, EGP, EGD, EB, ENU, ER, ET.
+
=== Photon production ===
 +
 
 +
{|class="wikitable" style="text-align: left;"
 +
! Parameter
 +
! Size
 +
! Description
 
|-
 
|-
−
| EDEP_CAPT_E             
+
| PHOTON_SAMPLING_MODE
 
| 1
 
| 1
−
| Additional energy release in capture reactions, mode 1 (see [[Input syntax manual#set edepmode|set edepmode]] input option)
+
| Flag indicating whether or not photon production from neutron reactions mode is on (see [[Input syntax manual#set ngamma|set ngamma]] input option)
 +
|-
 +
| PHOTON_SAMPLING_FAIL
 +
| 2
 +
| Fraction of failed photon samples
 
|-
 
|-
 
|}
 
|}
  
−
=== Radioactivity data ===
+
=== Energy deposition ===
 +
<u>Notes:</u>
 +
 
 +
* 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.
  
 
{|class="wikitable" style="text-align: left;"
 
{|class="wikitable" style="text-align: left;"
Line 682: Line 810:
 
! Description
 
! Description
 
|-
 
|-
−
| TOT_ACTIVITY             
+
| EDEP_MODE               
−
|
+
| 1
−
|
+
| Energy deposition mode (see [[Input syntax manual#set edepmode|set edepmode]] input option)
 
|-
 
|-
−
| TOT_DECAY_HEAT           
+
| EDEP_DELAYED             
−
|
+
| 1
−
|
+
| Energy of delayed components in energy deposition calculations (see [[Input syntax manual#set edepdel|set edepdel]] input option)
 
|-
 
|-
−
| TOT_SF_RATE             
+
| EDEP_KEFF_CORR           
−
|
+
| 1
−
|
+
| Flag indicating whether or not correction for energy deposition estimates in non-critical systems (see [[Input syntax manual#set edepkcorr |set edepkcorr]] input option)
 
|-
 
|-
−
| ACTINIDE_ACTIVITY       
+
| EDEP_LOCAL_EGD           
−
|
+
| 1
−
|
+
| Energy distribution of delayed components in energy deposition calculations, mode 3 (see [[Input syntax manual#set edepdel|set edepdel]] input option)
 
|-
 
|-
−
| ACTINIDE_DECAY_HEAT     
+
| EDEP_COMP               
−
|
+
| 9
−
|
+
| Fission energy release components: EFR, ENP, END, EGP, EGD, EB, ENU, ER, ET.
 
|-
 
|-
−
| FISSION_PRODUCT_ACTIVITY 
+
| EDEP_CAPT_E             
−
|
+
| 1
−
|
+
| Additional energy release in capture reactions, mode 1 (see [[Input syntax manual#set edepmode|set edepmode]] input option)
 
|-
 
|-
−
| FISSION_PRODUCT_DECAY_HEAT
+
|}
−
|
+
 
−
|
+
=== Radioactivity data ===
 +
<u>Notes:</u>
 +
 
 +
*The values are given at the current burnup point (depletion step).
 +
 
 +
{|class="wikitable" style="text-align: left;"
 +
! Parameter
 +
! Size
 +
! Description
 
|-
 
|-
−
| INHALATION_TOXICITY     
+
| TOT_ACTIVITY             
−
|
+
| 1
−
|
+
| Total activity
 
|-
 
|-
−
| INGESTION_TOXICITY       
+
| TOT_DECAY_HEAT           
−
|
+
| 1
−
|
+
| Total decay heat
 
|-
 
|-
−
| ACTINIDE_INH_TOX         
+
| TOT_SF_RATE             
−
|
+
| 1
−
|
+
| Total spontaneous fission rate
 
|-
 
|-
−
| ACTINIDE_ING_TOX           
+
| ACTINIDE_ACTIVITY       
−
|
+
| 1
−
|
+
| Actinide activity
 +
|-
 +
| ACTINIDE_DECAY_HEAT     
 +
| 1
 +
| Actinide decay heat
 +
|-
 +
| FISSION_PRODUCT_ACTIVITY 
 +
| 1
 +
| Fission product activity
 +
|-
 +
| FISSION_PRODUCT_DECAY_HEAT
 +
| 1
 +
| Fission product decay heat
 +
|-
 +
| INHALATION_TOXICITY     
 +
| 1
 +
| Total inhalation toxicity
 +
|-
 +
| INGESTION_TOXICITY       
 +
| 1
 +
| Total ingestion toxicity
 +
|-
 +
| ACTINIDE_INH_TOX         
 +
| 1
 +
| Actinide inhalation toxicity
 +
|-
 +
| ACTINIDE_ING_TOX           
 +
| 1
 +
| Actinide ingestion toxicity
 
|-
 
|-
 
| FISSION_PRODUCT_INH_TOX   
 
| FISSION_PRODUCT_INH_TOX   
−
|
+
| 1
−
|
+
| Fission product inhalation toxicity
 
|-
 
|-
 
| FISSION_PRODUCT_ING_TOX   
 
| FISSION_PRODUCT_ING_TOX   
−
|
+
| 1
−
|
+
| Fission product ingestion toxicity
 
|-
 
|-
 
| SR90_ACTIVITY             
 
| SR90_ACTIVITY             
−
|
+
| 1
−
|
+
| Sr-90 activity
 
|-
 
|-
 
| TE132_ACTIVITY             
 
| TE132_ACTIVITY             
−
|
+
| 1
−
|
+
| Te-132 activity
 
|-
 
|-
 
| I131_ACTIVITY             
 
| I131_ACTIVITY             
−
|
+
| 1
−
|
+
| I-131 activity
 
|-
 
|-
 
| I132_ACTIVITY             
 
| I132_ACTIVITY             
−
|
+
| 1
−
|
+
| I-132 activity
 
|-
 
|-
 
| CS134_ACTIVITY             
 
| CS134_ACTIVITY             
−
|
+
| 1
−
|
+
| Cs-134 activity
 
|-
 
|-
 
| CS137_ACTIVITY             
 
| CS137_ACTIVITY             
−
|
+
| 1
−
|
+
| Cs-137 activity
 
|-
 
|-
 
| PHOTON_DECAY_SOURCE       
 
| PHOTON_DECAY_SOURCE       
−
|
+
| 2
−
|
+
| Total photon decay source rate and total released gamma decay
 
|-
 
|-
 
| NEUTRON_DECAY_SOURCE       
 
| NEUTRON_DECAY_SOURCE       
−
|
+
| 1
−
|
+
| Total neutron decay source rate
 
|-
 
|-
 
| ALPHA_DECAY_SOURCE         
 
| ALPHA_DECAY_SOURCE         
−
|
+
| 1
−
|
+
| Total alpha decay source rate
 
|-
 
|-
 
| ELECTRON_DECAY_SOURCE     
 
| ELECTRON_DECAY_SOURCE     
−
|
+
| 1
−
|
+
| Total beta decay source rate
 
|-
 
|-
 
|}
 
|}
Line 814: Line 978:
 
| BURNUP                     
 
| BURNUP                     
 
| 2
 
| 2
−
| Burnup at the current step (in MWd/kgU): cumulative and real-cumulative.
+
| Burnup at the current step (in MWd/kgU): cumulative calculated from the [[Input_syntax_manual#dep_.28depletion_history.29|depletion history input]] and cumulative realized from the actual calculation.
 
|-
 
|-
 
| BURN_DAYS                 
 
| BURN_DAYS                 
Line 826: Line 990:
 
|}
 
|}
  
−
=== Analog reaction rate estimators ===
+
=== Coefficient calculation ===
  
 
{|class="wikitable" style="text-align: left;"
 
{|class="wikitable" style="text-align: left;"
Line 833: Line 997:
 
! Description
 
! Description
 
|-
 
|-
−
| CONVERSION_RATIO         
+
| COEF_IDX
−
|
+
| 2
−
|
+
| Coefficient index when output is printed and total number of coefficient calculations
 
|-
 
|-
−
| U235_FISS               
+
| COEF_BRANCH
−
|
+
| 1
−
|
+
| Branch index within coefficient calculation when output is printed
 
|-
 
|-
−
| U238_FISS               
+
| COEF_BU_STEP
−
|
+
| 1
−
|
+
| Burnup step at the given coefficient calculation when output is printed
−
|-
 
−
| U235_CAPT               
 
−
|
 
−
|
 
−
|-
 
−
| U238_CAPT               
 
−
|
 
−
|
 
−
|-
 
−
| XE135_CAPT               
 
−
|
 
−
|
 
 
|-
 
|-
 
|}
 
|}
  
−
== Particle balance ==
+
=== Analog reaction rate estimators ===
−
 
 
−
=== Neutron balance (particles/weight) ===
 
  
 
{|class="wikitable" style="text-align: left;"
 
{|class="wikitable" style="text-align: left;"
Line 868: Line 1,018:
 
! Description
 
! Description
 
|-
 
|-
−
| BALA_SRC_NEUTRON_SRC   
+
| CONVERSION_RATIO         
−
|
+
| 2
−
|
+
| Analog estimate of conversion rate, ratio between fissile production and loss rate
 
|-
 
|-
−
| BALA_SRC_NEUTRON_FISS   
+
| TH232_FISS
−
|
+
| 4
−
|
+
| Analog estimate of Th-232 fission rate (total/fraction)
 
|-
 
|-
−
| BALA_SRC_NEUTRON_NXN   
+
| U233_FISS
−
|
+
| 4
−
|
+
| Analog estimate of U-233 fission rate (total/fraction)
 
|-
 
|-
−
| BALA_SRC_NEUTRON_VR     
+
| U235_FISS               
−
|
+
| 4
−
|
+
| Analog estimate of U-235 fission rate (total/fraction)
 
|-
 
|-
−
| BALA_SRC_NEUTRON_TOT   
+
| U238_FISS               
−
|
+
| 4
−
|
+
| Analog estimate of U-238 fission rate (total/fraction)
 
|-
 
|-
−
| BALA_LOSS_NEUTRON_CAPT   
+
| PU239_FISS
−
|
+
| 4
−
|
+
| Analog estimate of Pu-239 fission rate (total/fraction)
 
|-
 
|-
−
| BALA_LOSS_NEUTRON_FISS   
+
| PU240_FISS
−
|
+
| 4
−
|
+
| Analog estimate of Pu-240 fission rate (total/fraction)
 
|-
 
|-
−
| BALA_LOSS_NEUTRON_LEAK   
+
| PU241_FISS
−
|
+
| 4
−
|
+
| Analog estimate of Pu-241 fission rate (total/fraction)
 +
|-
 +
| TH232_CAPT
 +
| 4
 +
| Analog estimate of Th-232 capture rate (total/fraction)
 +
|-
 +
| U233_CAPT
 +
| 4
 +
| Analog estimate of U-233 capture rate (total/fraction)
 +
|-
 +
| U235_CAPT               
 +
| 4
 +
| Analog estimate of U-235 capture rate (total/fraction)
 +
|-
 +
| U238_CAPT               
 +
| 4
 +
| Analog estimate of U-238 capture rate (total/fraction)
 +
|-
 +
| PU239_CAPT
 +
| 4
 +
| Analog estimate of Pu-239 capture rate (total/fraction)
 
|-
 
|-
−
| BALA_LOSS_NEUTRON_CUT   
+
| PU240_CAPT
−
|
+
| 4
−
|
+
| Analog estimate of Pu-240 capture rate (total/fraction)
 
|-
 
|-
−
| BALA_LOSS_NEUTRON_ERR   
+
| PU241_CAPT
−
|
+
| 4
−
|
+
| Analog estimate of Pu-241 capture rate (total/fraction)
 
|-
 
|-
−
| BALA_LOSS_NEUTRON_TOT   
+
| XE135_CAPT               
−
|
+
| 4
−
|
+
| Analog estimate of Xe-135 capture rate (total/fraction)
 
|-
 
|-
−
| BALA_NEUTRON_DIFF       
+
| XE135M_CAPT
−
|
+
| 4
−
|
+
| Analog estimate of Xe-135m capture rate (total/fraction)
 
|-
 
|-
 +
| SM149_CAPT
 +
| 4
 +
| Analog estimate of Sm-149 capture rate (total/fraction)
 
|}
 
|}
  
−
== Integral results ==
+
== Particle balance ==
  
−
=== Normalized total reaction rates (neutrons) ===
+
=== Neutron balance (particles/weight) ===
  
 
{|class="wikitable" style="text-align: left;"
 
{|class="wikitable" style="text-align: left;"
Line 927: Line 1,100:
 
! Description
 
! Description
 
|-
 
|-
−
| TOT_POWER               
+
| BALA_SRC_NEUTRON_SRC   
−
|
+
| 1/1
−
|
+
| Neutron produced by external source
 
|-
 
|-
−
| TOT_POWDENS             
+
| BALA_SRC_NEUTRON_FISS   
−
|
+
| 1/1
−
|
+
| Neutron produced by fission
 
|-
 
|-
−
| TOT_GENRATE             
+
| BALA_SRC_NEUTRON_NXN   
−
|
+
| 1/1
−
|
+
| Neutron produced by scattering
 
|-
 
|-
−
| TOT_FISSRATE             
+
| BALA_SRC_NEUTRON_VR     
−
|
+
| 1/1
−
|
+
| Neutron produced by variance reduction (Russian roulette, splitting)
 
|-
 
|-
−
| TOT_CAPTRATE             
+
| BALA_SRC_NEUTRON_TOT   
−
|
+
| 1/1
−
|
+
| Total neutron produced
 
|-
 
|-
−
| TOT_ABSRATE             
+
| BALA_LOSS_NEUTRON_CAPT   
−
|
+
| 1/1
−
|
+
| Neutron lost by capture
 
|-
 
|-
−
| TOT_SRCRATE             
+
| BALA_LOSS_NEUTRON_FISS   
−
|
+
| 1/1
−
|
+
| Neutron lost by fission
 
|-
 
|-
−
| TOT_FLUX                 
+
| BALA_LOSS_NEUTRON_LEAK   
−
|
+
| 1/1
−
|
+
| Neutron lost by leakage
 
|-
 
|-
−
| TOT_PHOTON_PRODRATE     
+
| BALA_LOSS_NEUTRON_CUT   
−
|
+
| 1/1
−
|
+
| Neutron lost by cut-off
 
|-
 
|-
−
| TOT_LEAKRATE             
+
| BALA_LOSS_NEUTRON_ERR   
−
|
+
| 1/1
−
|
+
| Neutron lost by failed tracking
 
|-
 
|-
−
| ALBEDO_LEAKRATE         
+
| BALA_LOSS_NEUTRON_TOT   
−
|
+
| 1/1
−
|
+
| Total neutron lost
 
|-
 
|-
−
| TOT_LOSSRATE             
+
| BALA_NEUTRON_DIFF       
−
|
+
| 1/1
−
|
+
| Difference between total neutron produced and lost
 
|-
 
|-
−
| TOT_CUTRATE             
+
|}
−
|
+
 
−
|
+
=== Photon balance (particles/weight/energy-weighted) ===
 +
 
 +
{|class="wikitable" style="text-align: left;"
 +
! Parameter
 +
! Size
 +
! Description
 +
|-
 +
| BALA_SRC_PHOTON_SRC
 +
| 1/1/1
 +
| Photon produced by external source
 +
|-
 +
| BALA_SRC_PHOTON_TTB
 +
| 1/1/1
 +
| Photon produced by bremsstrahlung
 
|-
 
|-
−
| TOT_RR                   
+
| BALA_SRC_PHOTON_ANNIH
−
|
+
| 1/1/1
−
|
+
| Photon produced by annihilation
 +
|-
 +
| BALA_SRC_PHOTON_FLUOR
 +
| 1/1/1
 +
| Photon produced by fluorescence
 +
|-
 +
| BALA_SRC_PHOTON_NREA
 +
| 1/1/1
 +
| Photon produced by neutron reaction
 +
|-
 +
| BALA_SRC_PHOTON_VR
 +
| 1/1/1
 +
| Photon produced by variance reduction (Russian roulette, splitting)
 +
|-
 +
| BALA_SRC_PHOTON_TOT
 +
| 1/1/1
 +
| Total photon produced
 +
|-
 +
| BALA_LOSS_PHOTON_CAPT
 +
| 1/1
 +
| Photon lost by capture
 
|-
 
|-
−
| TOT_XE135_ABSRATE       
+
| BALA_LOSS_PHOTON_LEAK
−
|
+
| 1/1
−
|
+
| Photon lost by leakage
 
|-
 
|-
−
| INI_FMASS               
+
| BALA_LOSS_PHOTON_CUT
−
|
+
| 1/1
−
|
+
| Photon lost by cut-off
 
|-
 
|-
−
| TOT_FMASS               
+
| BALA_LOSS_PHOTON_ERR
−
|
+
| 1/1
−
|
+
| Photon lost by failed tracking
 
|-
 
|-
−
| INI_BURN_FMASS           
+
| BALA_LOSS_PHOTON_TOT
−
|
+
| 1/1
−
|
+
| Total photon lost
 
|-
 
|-
−
| TOT_BURN_FMASS           
+
| BALA_PHOTON_DIFF
−
|
+
| 1/1
−
|
+
| Difference between total photon produced and lost
 
|-
 
|-
 
|}
 
|}
  
−
=== Equilibrium Xe-135 iteration ===
+
== Integral results ==
 +
 
 +
=== Normalized total reaction rates (neutrons) ===
 +
<u>Notes:</u>
 +
 
 +
*In burnup calculations the values correspond to total, burnable and non-burnable rates
  
 
{|class="wikitable" style="text-align: left;"
 
{|class="wikitable" style="text-align: left;"
Line 1,012: Line 1,223:
 
! Description
 
! Description
 
|-
 
|-
−
| XE135_EQUIL_CONC         
+
| TOT_POWER               
−
| 2
+
| 2(6)
−
| Averaged equilibrium Xe-135 concentration (see [[Input syntax manual#set xenon|set xenon]] input option)
+
| Total neutron fission power
 +
|-
 +
| TOT_POWDENS             
 +
| 2(6)
 +
| Total neutron fission power density
 +
|-
 +
| TOT_GENRATE             
 +
| 2(6)
 +
| Total neutron generation rate
 +
|-
 +
| TOT_FISSRATE             
 +
| 2(6)
 +
| Total neutron fission rate
 +
|-
 +
| TOT_CAPTRATE             
 +
| 2(6)
 +
| Total neutron capture rate
 +
|-
 +
| TOT_ABSRATE             
 +
| 2(6)
 +
| Total neutron absorption rate
 +
|-
 +
| TOT_SRCRATE             
 +
| 2(6)
 +
| Total neutron source rate
 
|-
 
|-
−
| I135_EQUIL_CONC         
+
| TOT_FLUX                 
−
| 2
+
| 2(6)
−
| Averaged equilibrium I-135 concentration (see [[Input syntax manual#set xenon|set xenon]] input option)
+
| Total neutron flux
 
|-
 
|-
−
|}
+
| TOT_PHOTON_PRODRATE     
−
 
+
| 4
−
=== Equilibrium Sm-149 iteration ===
+
| Total neutron-photon production rate (implicit/analog)
−
 
 
−
{|class="wikitable" style="text-align: left;"
 
−
! Parameter
 
−
! Size
 
−
! Description
 
 
|-
 
|-
−
| SM149_EQUIL_CONC         
+
| TOT_LEAKRATE             
 
| 2
 
| 2
−
| Averaged equilibrium Sm-149 concentration (see [[Input syntax manual#set samarium|set samarium]] input option)
+
| Total neutron leakage rate
 
|-
 
|-
−
| PM149_EQUIL_CONC            
+
| ALBEDO_LEAKRATE            
 
| 2
 
| 2
−
| Averaged equilibrium Pm-149 concentration (see [[Input syntax manual#set samarium|set samarium]] input option)
+
| Albedo neutron leakage rate
 
|-
 
|-
−
|}
+
| TOT_LOSSRATE             
−
 
 
−
=== Six-factor formula ===
 
−
 
 
−
{|class="wikitable" style="text-align: left;"
 
−
! Parameter
 
−
! Size
 
−
! Description
 
−
|-
 
−
| SIX_FF_ETA               
 
 
| 2
 
| 2
−
| Analog estimate of average number of neutrons emitted per thermal neutron absorbed in fuel
+
| Total neutron loss rate
 
|-
 
|-
−
| SIX_FF_F                 
+
| TOT_CUTRATE             
 
| 2
 
| 2
−
| Analog estimate of thermal utilization factor
+
| Total neutron energy cut-off rate
 
|-
 
|-
−
| SIX_FF_P                 
+
| TOT_RR                   
 
| 2
 
| 2
−
| Analog estimate of resonance escape probability
+
| Total neutron reaction rate
 
|-
 
|-
−
| SIX_FF_EPSILON           
+
| TOT_XE135_ABSRATE       
 
| 2
 
| 2
−
| Analog estimate of fast fission factor
+
| Total neutron absorption rate in Xe-135
 
|-
 
|-
−
| SIX_FF_LF               
+
| TOT_XE135M_ABSRATE
 
| 2
 
| 2
−
| Analog estimate of fast non-leakage probability
+
| Total neutron absorption rate in Xe-135m
 
|-
 
|-
−
| SIX_FF_LT               
+
| TOT_SM149_ABSRATE
 
| 2
 
| 2
−
| Analog estimate of thermal non-leakage probability
+
| Total neutron absorption rate in Sm-149
 
|-
 
|-
−
| SIX_FF_KINF             
+
| INI_FMASS               
−
| 2
+
| 1
−
| Analog estimate of six-factor ''k''<sub>inf</sub> (four-factor ''k''<sub>eff</sub>)
+
| Initial fissile mass
 
|-
 
|-
−
| SIX_FF_KEFF             
+
| TOT_FMASS               
−
| 2
+
| 1
−
| Analog estimate of six-factor ''k''<sub>eff</sub>
+
| Total fissile mass
 +
|-
 +
| INI_BURN_FMASS           
 +
| 1
 +
| Initial fissile mass in burnable materials
 +
|-
 +
| TOT_BURN_FMASS           
 +
| 1
 +
| Total fissile mass in burnable materials
 
|-
 
|-
 
|}
 
|}
  
−
=== Fission neutron and energy production ===
+
=== Normalized total reaction rates (photons) ===
  
 
{|class="wikitable" style="text-align: left;"
 
{|class="wikitable" style="text-align: left;"
Line 1,087: Line 1,316:
 
! Description
 
! Description
 
|-
 
|-
−
| NUBAR                   
+
| TOT_PHOTON_LEAKRATE
−
|
+
| 2
−
|
+
| Total photon leakage rate
 
|-
 
|-
−
| FISSE                   
+
| TOT_PHOTON_CUTRATE
−
|
+
| 2
−
|
+
| Total photon energy cut-off rate
 
|-
 
|-
−
|}
+
| PHOTOELE_CAPT_RATE
−
 
+
| 2
−
=== Criticality eigenvalues ===
+
| Photo-electric photon capture rate
−
 
 
−
{|class="wikitable" style="text-align: left;"
 
−
! Parameter
 
−
! Size
 
−
! Description
 
 
|-
 
|-
−
| ANA_KEFF                 
+
| PAIRPROD_CAPT_RATE
−
| 6
+
| 2
−
| Analog estimate of ''k''<sub>eff</sub>: total, prompt and delayed neutron contribution.
+
|Pair production photon capture rate
 
|-
 
|-
−
| IMP_KEFF                 
+
| TOT_PHOTON_LOSSRATE
 
| 2
 
| 2
−
| Implicit estimate of ''k''<sub>eff</sub>.
+
| Total photon loss rate
 
|-
 
|-
−
| COL_KEFF                 
+
| TOT_PHOTON_SRCRATE
 
| 2
 
| 2
−
| Collision estimate of ''k''<sub>eff</sub>.
+
| Total photon source rate
 
|-
 
|-
−
| ABS_KEFF                 
+
| TOT_PHOTON_RR
 
| 2
 
| 2
−
| Absorption estimate of ''k''<sub>eff</sub>.
+
| Total photon reaction rate
 
|-
 
|-
−
| ABS_KINF                 
+
| TOT_PHOTON_FLUX
 
| 2
 
| 2
−
| Absorption estimate of ''k''<sub>inf</sub>.
+
| Total photon flux
 
|-
 
|-
−
| GEOM_ALBEDO             
+
| TOT_PHOTON_HEATRATE
−
| 6
+
| 2
−
| Fixed or iterated value for albedo boundary condition for x-,y- and z-directions (see [[Input syntax manual#set bc|set bc]] or [[Input syntax manual#set iter alb|set iter alb]] input options).
+
| Total photon heating rate
 
|-
 
|-
 
|}
 
|}
  
−
=== ALF (Average lethargy of neutrons causing fission) ===
+
=== Equilibrium Xe-135 iteration ===
 +
<u>Notes:</u>
  
 +
*The averages are calculated as volume averages.
 +
*Materials with the equilibrium calculation turned off are not taken into account in the average concentration but are included in the volume if they are fissile.
 +
 
{|class="wikitable" style="text-align: left;"
 
{|class="wikitable" style="text-align: left;"
 
! Parameter
 
! Parameter
Line 1,137: Line 1,365:
 
! Description
 
! Description
 
|-
 
|-
−
| ANA_ALF                 
+
| I135_EQUIL_CONC         
 +
| 2
 +
| Average equilibrium I-135 concentration in materials with Xe-135 production rate (see [[Input syntax manual#set xenon|set xenon]] input option)
 +
|-
 +
| XE135_EQUIL_CONC         
 
| 2
 
| 2
−
| Analog estimate of average lethargy of neutrons causing fission
+
| Average equilibrium Xe-135 concentration in materials with Xe-135 production rate (see [[Input syntax manual#set xenon|set xenon]] input option)
 
|-
 
|-
−
| IMP_ALF                 
+
| XE135M_EQUIL_CONC         
 
| 2
 
| 2
−
| Implicit estimate of average lethargy of neutrons causing fission
+
| Average equilibrium Xe-135m concentration in materials with Xe-135 production rate (see [[Input syntax manual#set xenon|set xenon]] input option)
 
|-
 
|-
 
|}
 
|}
  
−
=== EALF (Energy corresponding to average lethargy of neutrons causing fission) ===
+
=== Equilibrium Sm-149 iteration ===
 +
<u>Notes:</u>
 +
 
 +
*The averages are calculated as volume averages.
 +
*Materials with the equilibrium calculation turned off are not taken into account in the average concentration but are included in the volume if they are fissile.
  
 
{|class="wikitable" style="text-align: left;"
 
{|class="wikitable" style="text-align: left;"
Line 1,154: Line 1,390:
 
! Description
 
! Description
 
|-
 
|-
−
| ANA_EALF                 
+
| SM149_EQUIL_CONC         
 
| 2
 
| 2
−
| Analog estimate of energy corresponding to the average lethargy of neutrons causing fission
+
| Average equilibrium Sm-149 concentration in materials with Sm-149 production rate (see [[Input syntax manual#set samarium|set samarium]] input option)
 
|-
 
|-
−
| IMP_EALF
+
| PM149_EQUIL_CONC         
 
| 2
 
| 2
−
| Implicit estimate of energy corresponding to the average lethargy of neutrons causing fission
+
| Average equilibrium Pm-149 concentration in materials with Sm-149 production rate (see [[Input syntax manual#set samarium|set samarium]] input option)
 
|-
 
|-
 
|}
 
|}
  
−
=== AFGE (Average energy of neutrons causing fission) ===
+
=== Iteration factor ===
  
 
{|class="wikitable" style="text-align: left;"
 
{|class="wikitable" style="text-align: left;"
Line 1,171: Line 1,407:
 
! Description
 
! Description
 
|-
 
|-
−
| ANA_AFGE                 
+
| ITER_FACTOR
−
| 2
 
−
| Analog estimate of average energy of neutrons causing fission
 
−
|-
 
−
| IMP_AFGE
 
 
| 2
 
| 2
−
| Implicit estimate of average energy of neutrons causing fission
+
| Iteration factor of critical density iteration (see [[Input syntax manual#set iter nuc|set iter nuc]] input option) or albedo iteration (see [[Input syntax manual#set iter alb|set iter alb]] input option)
 
|-
 
|-
 
|}
 
|}
  
−
== Time constants ==
+
=== Six-factor formula ===
 +
<u>Notes:</u>
  
−
=== Forward-weighted delayed neutron parameters ===
+
* The six-factor formula estimates are printed by default. Check the suitability of the approximation for the given spectrum (the formulation is based on a thermal spectrum).
  
 
{|class="wikitable" style="text-align: left;"
 
{|class="wikitable" style="text-align: left;"
Line 1,190: Line 1,423:
 
! Description
 
! Description
 
|-
 
|-
−
| PRECURSOR_GROUPS         
+
| SIX_FF_ETA               
−
| 1
+
| 2
−
| Number of delayed neutron precursor groups (referred to as ''D'' below)
+
| Analog estimate of average number of neutrons emitted per thermal neutron absorbed in fuel
 
|-
 
|-
−
| FWD_ANA_BETA_ZERO       
+
| SIX_FF_F                 
−
| 2''D'' + 2
+
| 2
−
| Analog estimator of physical delayed neutron fractions (number of delayed neutrons emitted in fission): total and group-wise
+
| Analog estimate of thermal utilization factor
 
|-
 
|-
−
| FWD_ANA_LAMBDA           
+
| SIX_FF_P                 
−
| 2''D'' + 2
+
| 2
−
| Analog estimator of delayed neutron precursor decay constants: total and group-wise
+
| Analog estimate of resonance escape probability
 
|-
 
|-
−
|}
+
| SIX_FF_EPSILON           
−
 
+
| 2
−
=== Beta-eff using Meulekamp's method ===
+
| Analog estimate of fast fission factor
−
 
 
−
{|class="wikitable" style="text-align: left;"
 
−
! Parameter
 
−
! Size
 
−
! Description
 
 
|-
 
|-
−
| ADJ_MEULEKAMP_BETA_EFF   
+
| SIX_FF_LF               
−
| 2''D'' + 2
+
| 2
−
| Adjoint-weighted effective delayed neutron fractions using Meulekamp's method: total and group-wise
+
| Analog estimate of fast non-leakage probability
 
|-
 
|-
−
| ADJ_MEULEKAMP_LAMBDA     
+
| SIX_FF_LT               
−
| 2''D'' + 2
+
| 2
−
| Adjoint-weighted of delayed neutron precursor decay constants using Meulekamp's method: total and group-wise
+
| Analog estimate of thermal non-leakage probability
 +
|-
 +
| SIX_FF_KINF             
 +
| 2
 +
| Analog estimate of six-factor ''k''<sub>inf</sub> (four-factor ''k''<sub>eff</sub>)
 +
|-
 +
| SIX_FF_KEFF             
 +
| 2
 +
| Analog estimate of six-factor ''k''<sub>eff</sub>
 
|-
 
|-
 
|}
 
|}
  
−
=== Adjoint weighted time constants using Nauchi's method ===
+
=== Fission neutron and energy production ===
  
 
{|class="wikitable" style="text-align: left;"
 
{|class="wikitable" style="text-align: left;"
Line 1,228: Line 1,464:
 
! Description
 
! Description
 
|-
 
|-
−
| IFP_CHAIN_LENGTH         
+
| NUBAR                   
−
| 1
+
| 2
−
| Number of generations within the iterated fission probability method
+
| Average fission neutron yield
 
|-
 
|-
−
| ADJ_NAUCHI_GEN_TIME     
+
| FISSE                   
−
| 6
+
| 2
−
| Adjoint-weighted neutron generation times using Nauchi's method: total, prompt and, delayed
+
| Average fission energy production
−
|-
 
−
| ADJ_NAUCHI_LIFETIME     
 
−
| 6
 
−
| Adjoint-weighted neutron lifetimes using Nauchi's method: total, prompt and, delayed.
 
−
|-
 
−
| ADJ_NAUCHI_BETA_EFF     
 
−
| 2''D'' + 2
 
−
| Adjoint-weighted effective delayed neutron fractions using Nauchi's method: total and group-wise
 
−
|-
 
−
| ADJ_NAUCHI_LAMBDA       
 
−
| 2''D'' + 2
 
−
| Adjoint-weighed of delayed neutron precursor decay constants using Nauchi's method: total and group-wise
 
 
|-
 
|-
 
|}
 
|}
  
−
=== Adjoint weighted time constants using IFP ===
+
=== Criticality eigenvalues / Multiplication factor external source ===
  
 
{|class="wikitable" style="text-align: left;"
 
{|class="wikitable" style="text-align: left;"
Line 1,257: Line 1,481:
 
! Description
 
! Description
 
|-
 
|-
−
| ADJ_IFP_GEN_TIME         
+
| ANA_KEFF                 
−
| 6
+
| 6(2)
−
| Adjoint-weighted neutron generation times using the iterated fission probability method: total, prompt and, delayed
+
| Analog estimate of ''k''<sub>eff</sub>: total, prompt and delayed neutron contribution.
 +
|-
 +
| IMP_KEFF                 
 +
| 2
 +
| Implicit estimate of ''k''<sub>eff</sub>.
 
|-
 
|-
−
| ADJ_IFP_LIFETIME         
+
| COL_KEFF                 
−
| 6
+
| 2
−
| Adjoint-weighted neutron lifetimes using the iterated fission probability method: total, prompt and, delayed
+
| Collision estimate of ''k''<sub>eff</sub>.
 
|-
 
|-
−
| ADJ_IFP_IMP_BETA_EFF     
+
| ABS_KEFF                 
−
| 2''D'' + 2
+
| 2
−
| Implicit estimator of adjoint-weighted effective delayed neutron fractions using the iterated fission probability method: total and group-wise
+
| Absorption estimate of ''k''<sub>eff</sub>.
 
|-
 
|-
−
| ADJ_IFP_IMP_LAMBDA       
+
| ABS_KINF                 
−
| 2''D'' + 2
+
| 2
−
| Implicit estimator of adjoint-weighted of delayed neutron precursor decay constants using the iterated fission probability method: total and group-wise
+
| Absorption estimate of ''k''<sub>inf</sub>.
 
|-
 
|-
−
| ADJ_IFP_ANA_BETA_EFF     
+
| ANA_EXT_K
−
| 2''D'' + 2
+
| 20
−
| Analog estimator of adjoint-weighted effective delayed neutron fractions using the iterated fission probability method: total and group-wise
+
| Generation-wise source multiplication factors in external source mode
 
|-
 
|-
−
| ADJ_IFP_ANA_LAMBDA       
+
| SRC_MULT
−
| 2''D'' + 2
+
| 2
−
| Analog estimator of adjoint-weighted of delayed neutron precursor decay constants using the iterated fission probability method: total and group-wise
+
| Source multiplication factor in external source mode
 
|-
 
|-
−
| ADJ_IFP_ROSSI_ALPHA     
+
| MEAN_NGEN
 
| 2
 
| 2
−
| Adjoint-weighted Rossi alpha using the iterated fission probability method
+
| Mean number of generations in external source mode
 +
|-
 +
| PROMPT_GEN_POP
 +
| ''N<sub>G</sub>''
 +
| Prompt fission population generation fraction
 +
|-
 +
| PROMPT_GEN_CUMU
 +
| ''N<sub>G</sub>''
 +
| Prompt fission cumulative generation fraction
 
|-
 
|-
−
|}
+
| PROMPT_GEN_TIMES
−
 
+
| ''N<sub>G</sub>''
−
=== Adjoint weighted time constants using perturbation technique ===
+
| Prompt fission time generation fraction
 +
|-
 +
| PROMPT_CHAIN_LENGTH
 +
| 2
 +
| Mean prompt chain length in external source mode
 +
|-
 +
| GEOM_ALBEDO             
 +
| 6
 +
| Fixed or iterated value for albedo boundary condition for x-,y- and z-directions (see [[Input syntax manual#set bc|set bc]] or [[Input syntax manual#set iter alb|set iter alb]] input options).
 +
|-
 +
|}
 +
 
 +
=== Wielandt method ===
  
 
{|class="wikitable" style="text-align: left;"
 
{|class="wikitable" style="text-align: left;"
Line 1,294: Line 1,542:
 
! Description
 
! Description
 
|-
 
|-
−
| ADJ_PERT_GEN_TIME       
+
| WIELANDT_K
−
| 2
 
−
| Adjoint-weighted neutron generation time using the perturbation technique
 
−
|-
 
−
| ADJ_PERT_LIFETIME       
 
 
| 2
 
| 2
−
| Adjoint-weighted neutron lifetime using the perturbation technique
+
| Wielandt’s method shifted eigenvalue (see [[Input syntax manual#set wie|set wie]] input option)
 
|-
 
|-
−
| ADJ_PERT_BETA_EFF       
+
| WIELANDT_P
 
| 2
 
| 2
−
| Adjoint-weighted effective delayed neutron fraction using the perturbation technique
+
| Wielandt’s method neutron banking probability (see [[Input syntax manual#set wie|set wie]] input option)
−
|-
 
−
| ADJ_PERT_ROSSI_ALPHA     
 
−
| 2
 
−
| Adjoint-weighted Rossi alpha using the perturbation technique
 
 
|-
 
|-
 
|}
 
|}
  
−
=== Inverse neutron speed  ===
+
=== ALF (Average lethargy of neutrons causing fission) ===
  
 
{|class="wikitable" style="text-align: left;"
 
{|class="wikitable" style="text-align: left;"
Line 1,319: Line 1,559:
 
! Description
 
! Description
 
|-
 
|-
−
| ANA_INV_SPD             
+
| ANA_ALF                 
 +
| 2
 +
| Analog estimate of average lethargy of neutrons causing fission
 +
|-
 +
| IMP_ALF                 
 
| 2
 
| 2
−
| Analog estimate of inverse neutron speed
+
| Implicit estimate of average lethargy of neutrons causing fission
 
|-
 
|-
 
|}
 
|}
  
−
=== Analog slowing-down and thermal neutron lifetime (total/prompt/delayed) ===
+
=== EALF (Energy corresponding to average lethargy of neutrons causing fission) ===
  
 
{|class="wikitable" style="text-align: left;"
 
{|class="wikitable" style="text-align: left;"
Line 1,332: Line 1,576:
 
! Description
 
! Description
 
|-
 
|-
−
| ANA_SLOW_TIME           
+
| ANA_EALF                 
−
| 6
+
| 2
−
| Analog estimate of slowing-down time: total, prompt and, delayed
+
| Analog estimate of energy corresponding to the average lethargy of neutrons causing fission
 
|-
 
|-
−
| ANA_THERM_TIME           
+
| IMP_EALF
−
| 6
+
| 2
−
| Analog estimate of thermal neutron lifetime: total, prompt and, delayed
+
| Implicit estimate of energy corresponding to the average lethargy of neutrons causing fission
 
|-
 
|-
−
| ANA_THERM_FRAC           
+
|}
−
| 6
+
 
−
| Analog estimate of neutron thermalisation fraction: total, prompt and, delayed
+
=== AFGE (Average energy of neutrons causing fission) ===
 +
 
 +
{|class="wikitable" style="text-align: left;"
 +
! Parameter
 +
! Size
 +
! Description
 
|-
 
|-
−
| ANA_DELAYED_EMTIME       
+
| ANA_AFGE                 
 
| 2
 
| 2
−
| Analog estimate of delayed neutron emission time
+
| Analog estimate of average energy of neutrons causing fission
 
|-
 
|-
−
| ANA_MEAN_NCOL           
+
| IMP_AFGE
−
| 4
+
| 2
−
| Analog estimate of average number of collisions per history: total and to fission
+
| Implicit estimate of average energy of neutrons causing fission
 
|-
 
|-
 
|}
 
|}
  
−
== Homogenized group constants ==
+
== Time constants ==
  
−
<u>Notes:</u>
+
=== Forward delayed neutron parameters ===
−
 
 
−
*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.<ref>Leppänen, J., Pusa, M. and Fridman, E. ''"Overview of methodology for spatial homogenization in the Serpent 2 Monte Carlo code."'' Ann. Nucl. Energy, [http://www.sciencedirect.com/science/article/pii/S0306454916303899 96 (2016) 126-136].</ref>
 
−
*The fundamental mode calculation is off by default, and invoked by the [[Input syntax manual#set fum|set fum]] option. Otherwise all values with B1 prefix are printed as zeros.
 
−
*The intermediate multi-group structure is defined using option [[Input syntax manual#set micro|set micro]] or [[Input syntax manual#set micro|set fum]].
 
−
*The few-group structure is defined using option [[Input syntax manual#set nfg|set nfg]].
 
−
*The universes in which the group constants are calculated are listed in option [[Input syntax manual#set gcu|set gcu]]. The calculation is performed for [[Input syntax manual#set root|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 (<tt>idx</tt>)
 
−
*The parameter names can be listed in the [[Input syntax manual#set coefpara|set coefpara]] option, and they will be included in the [[Automated burnup sequence#Output|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 ===
 
  
 
{|class="wikitable" style="text-align: left;"
 
{|class="wikitable" style="text-align: left;"
Line 1,374: Line 1,611:
 
! Size
 
! Size
 
! Description
 
! Description
−
|- 
 
−
| GC_UNIVERSE_NAME
 
−
| (string)
 
−
| Name of the universe where spatial homogenization was performed
 
 
|-
 
|-
−
| MICRO_NG
+
| PRECURSOR_GROUPS         
 
| 1
 
| 1
−
| Number of energy groups in the intermediate multi-group structure (referred to as ''H'' below)
+
| Number of delayed neutron precursor groups (referred to as ''D'' below)
 
|-
 
|-
−
| MICRO_E
+
| FWD_ANA_BETA_ZERO       
−
| ''H'' + 1
+
| 2''D'' + 2
−
| Group boundaries in the intermediate multi-group structure (in ascending order)
+
| Analog estimator of physical delayed neutron fractions (number of delayed neutrons emitted in fission): total, group-wise
 
|-
 
|-
−
| MACRO_NG
+
| FWD_ANA_LAMBDA           
−
| 1
+
| 2''D'' + 2
−
| Number of energy groups in the final few-group structure (referred to as ''G'' below)
+
| Analog estimator of delayed neutron precursor decay constants: total, group-wise
−
|-
 
−
| MACRO_E
 
−
| ''G'' + 1
 
−
| Group boundaries in the final few-group structure (in descending order)
 
 
|-
 
|-
 
|}
 
|}
  
−
=== Group constants homogenized in infinite spectrum ===
+
=== Adjoint-weighted time constants using Meulekamp's method ===
  
 
{|class="wikitable" style="text-align: left;"
 
{|class="wikitable" style="text-align: left;"
Line 1,403: Line 1,632:
 
! Size
 
! Size
 
! Description
 
! Description
−
|- 
 
−
| INF_MICRO_FLX
 
−
| 2''H''
 
−
| Multi-group flux spectrum (integral, un-normalized)
 
 
|-
 
|-
−
| INF_FLX
+
| ADJ_MEULEKAMP_BETA_EFF   
−
| 2''G''
+
| 2''D'' + 2
−
| Few-group flux (integral, normalized)
+
| Adjoint-weighted effective delayed neutron fractions using Meulekamp's method: total, group-wise
 
|-
 
|-
−
| INF_KINF
+
| ADJ_MEULEKAMP_LAMBDA     
−
| 2
+
| 2''D'' + 2
−
| Infinite multiplication factor
+
| Adjoint-weighted of delayed neutron precursor decay constants using Meulekamp's method: total, group-wise
 
|-
 
|-
 
|}
 
|}
  
−
==== Reaction cross sections ====
+
=== Adjoint weighted time constants using Nauchi's method ===
  
 
{|class="wikitable" style="text-align: left;"
 
{|class="wikitable" style="text-align: left;"
Line 1,425: Line 1,650:
 
! Description
 
! Description
 
|-
 
|-
−
| INF_TOT
+
| IFP_CHAIN_LENGTH         
−
| 2''G''
+
| 1
−
| Total cross section
+
| Number of generations within the iterated fission probability method
 
|-
 
|-
−
| INF_CAPT
+
| ADJ_NAUCHI_GEN_TIME     
−
| 2''G''
+
| 6
−
| Capture cross section
+
| Adjoint-weighted neutron generation times using Nauchi's method: total, prompt, delayed
 
|-
 
|-
−
| INF_FISS
+
| ADJ_NAUCHI_LIFETIME     
−
| 2''G''
+
| 6
−
| Fission cross section
+
| Adjoint-weighted neutron lifetimes using Nauchi's method: total, prompt, delayed.
 
|-
 
|-
−
| INF_NSF
+
| ADJ_NAUCHI_BETA_EFF     
−
| 2''G''
+
| 2''D'' + 2
−
| Fission neutron production cross section
+
| Adjoint-weighted effective delayed neutron fractions using Nauchi's method: total, group-wise
 
|-
 
|-
−
| INF_KAPPA
+
| ADJ_NAUCHI_LAMBDA       
−
| 2''G''
+
| 2''D'' + 2
−
| Average deposited fission energy (MeV)
+
| Adjoint-weighed of delayed neutron precursor decay constants using Nauchi's method: total, group-wise
−
|-
 
−
| INF_INVV
 
−
| 2''G''
 
−
| Inverse neutron speed (s/cm)
 
−
|-
 
−
| INF_NUBAR
 
−
| 2''G''
 
−
| Average neutron yield
 
−
|-
 
−
| INF_ABS
 
−
| 2''G''
 
−
| Absorption cross section (capture + fission)
 
−
|-
 
−
| INF_REMXS
 
−
| 2''G''
 
−
| Removal cross section (group-removal + absorption)
 
−
|-
 
−
| INF_RABSXS
 
−
| 2''G''
 
−
| Reduced absorption cross section (total - scattering production)
 
 
|-
 
|-
 
|}
 
|}
  
−
==== Fission spectra ====
+
=== Adjoint weighted time constants using IFP ===
  
 
{|class="wikitable" style="text-align: left;"
 
{|class="wikitable" style="text-align: left;"
Line 1,474: Line 1,679:
 
! Description
 
! Description
 
|-
 
|-
−
| INF_CHIT
+
| ADJ_IFP_GEN_TIME         
−
| 2''G''
+
| 6
−
| Fission spectrum (total)
+
| Adjoint-weighted neutron generation times using the iterated fission probability method: total, prompt, delayed
 +
|-
 +
| ADJ_IFP_LIFETIME         
 +
| 6
 +
| Adjoint-weighted neutron lifetimes using the iterated fission probability method: total, prompt, delayed
 +
|-
 +
| ADJ_IFP_IMP_BETA_EFF     
 +
| 2''D'' + 2
 +
| Implicit estimator of adjoint-weighted effective delayed neutron fractions using the iterated fission probability method: total, group-wise
 +
|-
 +
| ADJ_IFP_IMP_LAMBDA       
 +
| 2''D'' + 2
 +
| Implicit estimator of adjoint-weighted of delayed neutron precursor decay constants using the iterated fission probability method: total, group-wise
 +
|-
 +
| ADJ_IFP_ANA_BETA_EFF     
 +
| 2''D'' + 2
 +
| Analog estimator of adjoint-weighted effective delayed neutron fractions using the iterated fission probability method: total, group-wise
 
|-
 
|-
−
| INF_CHIP
+
| ADJ_IFP_ANA_LAMBDA       
−
| 2''G''
+
| 2''D'' + 2
−
| Fission spectrum (prompt neutrons)
+
| Analog estimator of adjoint-weighted of delayed neutron precursor decay constants using the iterated fission probability method: total, group-wise
 
|-
 
|-
−
| INF_CHID
+
| ADJ_IFP_ROSSI_ALPHA     
−
| 2''G''
+
| 2
−
| Fission spectrum (delayed neutrons)
+
| Adjoint-weighted Rossi alpha using the iterated fission probability method
 
|-
 
|-
 
|}
 
|}
  
−
==== Scattering cross sections ====
+
=== Adjoint weighted time constants using perturbation technique ===
−
 
 
−
<u>Notes:</u>
 
−
 
 
−
*Scattering production includes multiplying (n,2n), (n,3n), etc. reactions.
 
  
 
{|class="wikitable" style="text-align: left;"
 
{|class="wikitable" style="text-align: left;"
Line 1,499: Line 1,716:
 
! Description
 
! Description
 
|-
 
|-
−
| INF_SCATT0
+
| ADJ_PERT_GEN_TIME       
−
| 2''G''
+
| 2
−
| Total ''P''<sub>0</sub> scattering cross section
+
| Adjoint-weighted neutron generation time using the perturbation technique
 
|-
 
|-
−
| INF_SCATT1
+
| ADJ_PERT_LIFETIME       
−
| 2''G''
+
| 2
−
| Total ''P''<sub>1</sub> scattering cross section
+
| Adjoint-weighted neutron lifetime using the perturbation technique
 
|-
 
|-
−
| INF_SCATT2
+
| ADJ_PERT_BETA_EFF       
−
| 2''G''
+
| 2
−
| Total ''P''<sub>2</sub> scattering cross section
+
| Adjoint-weighted effective delayed neutron fraction using the perturbation technique
 
|-
 
|-
−
| INF_SCATT3
+
| ADJ_PERT_ROSSI_ALPHA     
−
| 2''G''
+
| 2
−
| Total ''P''<sub>3</sub> scattering cross section
+
| Adjoint-weighted Rossi alpha using the perturbation technique
 
|-
 
|-
−
| INF_SCATT4
+
|}
−
| 2''G''
+
 
−
| Total ''P''<sub>4</sub> scattering cross section
+
=== Inverse neutron speed  ===
−
|-
+
 
−
| INF_SCATT5
+
{|class="wikitable" style="text-align: left;"
−
| 2''G''
+
! Parameter
−
| Total ''P''<sub>5</sub> scattering cross section
+
! Size
 +
! Description
 
|-
 
|-
−
| INF_SCATT6
+
| ANA_INV_SPD             
−
| 2''G''
+
| 2
−
| Total ''P''<sub>6</sub> scattering cross section
+
| Analog estimate of inverse neutron speed
 
|-
 
|-
−
| INF_SCATT7
+
|}
−
| 2''G''
+
 
−
| Total ''P''<sub>7</sub> scattering cross section
+
=== Analog slowing-down and thermal neutron lifetime (total/prompt/delayed) ===
 +
 
 +
{|class="wikitable" style="text-align: left;"
 +
! Parameter
 +
! Size
 +
! Description
 
|-
 
|-
−
| INF_SCATTP0
+
| ANA_SLOW_TIME           
−
| 2''G''
+
| 6
−
| Total ''P''<sub>0</sub> scattering production cross section
+
| Analog estimate of slowing-down time: total, prompt, delayed
 
|-
 
|-
−
| INF_SCATTP1
+
| ANA_THERM_TIME           
−
| 2''G''
+
| 6
−
| Total ''P''<sub>1</sub> scattering production cross section
+
| Analog estimate of thermal neutron lifetime: total, prompt, delayed
 
|-
 
|-
−
| INF_SCATTP2
+
| ANA_THERM_FRAC           
−
| 2''G''
+
| 6
−
| Total ''P''<sub>2</sub> scattering production cross section
+
| Analog estimate of neutron thermalisation fraction: total, prompt, delayed
 
|-
 
|-
−
| INF_SCATTP3
+
| ANA_DELAYED_EMTIME       
−
| 2''G''
+
| 2
−
| Total ''P''<sub>3</sub> scattering production cross section
+
| Analog estimate of delayed neutron emission time
 
|-
 
|-
−
| INF_SCATTP4
+
| ANA_MEAN_NCOL           
−
| 2''G''
+
| 4
−
| Total ''P''<sub>4</sub> scattering production cross section
+
| Analog estimate of average number of collisions per history: total and to fission
−
|-
 
−
| INF_SCATTP5
 
−
| 2''G''
 
−
| Total ''P''<sub>5</sub> scattering production cross section
 
−
|-
 
−
| INF_SCATTP6
 
−
| 2''G''
 
−
| Total ''P''<sub>6</sub> scattering production cross section
 
−
|-
 
−
| INF_SCATTP7
 
−
| 2''G''
 
−
| Total ''P''<sub>7</sub> scattering production cross section
 
 
|-
 
|-
 
|}
 
|}
  
−
==== Scattering matrices ====
+
=== Dynamic simulation ===
−
<u>Notes:</u>
 
  
−
*Scattering production includes multiplying (n,2n), (n,3n), etc. reactions.
+
{|class="wikitable" style="text-align: left;"
−
*The order of values ([input].coe) or value pairs ([input]_res.m) is: <math>\Sigma_{1,1} \, \Sigma_{1,2} \, ... \, \Sigma_{2,1} \, \Sigma_{2,2}  \, ...</math> where <math>\Sigma_{g,g'}</math> 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:
 
−
<nowiki>
 
−
reshape(INF_S0(idx,1:2:end), G, G);</nowiki>
 
−
 
 
−
{|class="wikitable" style="text-align: left;"
 
 
! Parameter
 
! Parameter
 
! Size
 
! Size
 
! Description
 
! Description
 
|-
 
|-
−
| INF_S0
+
| DYN_NB
−
| 2''G''<sup>2</sup>
+
| 1
−
| ''P''<sub>0</sub> scattering matrix
+
| Number of time intervals defined in the time-bin structure (see [[Input syntax manual#tme (time binning definition)|tme]] card)
 
|-
 
|-
−
| INF_S1
+
| DYN_TMIN
−
| 2''G''<sup>2</sup>
+
| 1
−
| ''P''<sub>1</sub> scattering matrix
+
| Minimum time boundary defined in the time-bin structure (see [[Input syntax manual#tme (time binning definition)|tme]] card)
−
|-  
 
−
| INF_S2
 
−
| 2''G''<sup>2</sup>
 
−
| ''P''<sub>2</sub> scattering matrix
 
 
|-
 
|-
−
| INF_S3
+
| DYN_TMAX
−
| 2''G''<sup>2</sup>
+
| 1
−
| ''P''<sub>3</sub> scattering matrix
+
| Maximum time boundary defined in the time-bin structure (see [[Input syntax manual#tme (time binning definition)|tme]] card)
 
|-
 
|-
−
| INF_S4
+
| DYN_TIMES
−
| 2''G''<sup>2</sup>
+
| 2''T'' + 2
−
| ''P''<sub>4</sub> scattering matrix
+
| Time-bin boundaries defined in the time-bin structure (see [[Input syntax manual#tme (time binning definition)|tme]] card)
 
|-
 
|-
−
| INF_S5
+
| DYN_POP
−
| 2''G''<sup>2</sup>
+
| 2''T''
−
| ''P''<sub>5</sub> scattering matrix
+
| Neutron population at the end of the time interval
 
|-
 
|-
−
| INF_S6
+
| DYN_PERIOD
−
| 2''G''<sup>2</sup>
+
| 2''T''
−
| ''P''<sub>6</sub> scattering matrix
+
| Reactor period based on increase/decrease of neutron population during time interval
 
|-
 
|-
−
| INF_S7
+
|}
−
| 2''G''<sup>2</sup>
+
 
−
| ''P''<sub>7</sub> scattering matrix
+
=== Analog mean photon lifetime ===
 +
 
 +
{|class="wikitable" style="text-align: left;"
 +
! Parameter
 +
! Size
 +
! Description
 
|-
 
|-
−
| INF_SP0
+
| ANA_LIFETIME
−
| 2''G''<sup>2</sup>
+
| 2
−
| ''P''<sub>0</sub> scattering production matrix
+
| Analog estimator of photon lifetime
 
|-
 
|-
−
| INF_SP1
+
|}
−
| 2''G''<sup>2</sup>
+
 
−
| ''P''<sub>1</sub> scattering production matrix
+
== Homogenized group constants ==
−
|-
+
 
−
| INF_SP2
+
<u>Notes:</u>
−
| 2''G''<sup>2</sup>
+
 
−
| ''P''<sub>2</sub> scattering production matrix
+
*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.<ref>Leppänen, J., Pusa, M. and Fridman, E. ''"Overview of methodology for spatial homogenization in the Serpent 2 Monte Carlo code."'' Ann. Nucl. Energy, [https://doi.org/10.1016/j.anucene.2016.06.007 96 (2016) 126-136].</ref>
−
| INF_SP3
+
*The fundamental mode calculation is off by default, and invoked by the [[Input syntax manual#set fum|set fum]] option. Otherwise all values with B1 prefix are printed as zeros.
−
| 2''G''<sup>2</sup>
+
*The intermediate multi-group structure is defined using option [[Input syntax manual#set micro|set micro]] or [[Input syntax manual#set micro|set fum]].
−
| ''P''<sub>3</sub> scattering production matrix
+
*The few-group structure is defined using option [[Input syntax manual#set nfg|set nfg]].
−
|-
+
*The universes in which the group constants are calculated are listed in option [[Input syntax manual#set gcu|set gcu]]. The calculation is performed for [[Input syntax manual#set root|root universe]] 0 by default, and can be switched off with "set gcu -1".
−
| INF_SP4
+
*If data is produced in multiple universes within a single run, the data is assigned with different run indexes (<tt>idx</tt>)
−
| 2''G''<sup>2</sup>
+
*The parameter names can be listed in the [[Input syntax manual#set coefpara|set coefpara]] option, and they will be included in the [[Automated burnup sequence#Output|group constant output file]] when the automated burnup sequence is invoked.
−
| ''P''<sub>4</sub> scattering production matrix
+
*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.
−
|-
 
−
| INF_SP5
 
−
| 2''G''<sup>2</sup>
 
−
| ''P''<sub>5</sub> scattering production matrix
 
−
|-
 
−
| INF_SP6
 
−
| 2''G''<sup>2</sup>
 
−
| ''P''<sub>6</sub> scattering production matrix
 
−
|-
 
−
| INF_SP7
 
−
| 2''G''<sup>2</sup>
 
−
| ''P''<sub>7</sub> scattering production matrix
 
−
|-
 
−
|}
 
  
−
==== Diffusion parameters ====
+
=== Common parameters ===
−
 
 
−
<u>Notes:</u>
 
−
*Calculation of sensible values for INF_TRANSPXS and INF_DIFFCOEF requires fine enough [[Input syntax manual#set micro|intermediate multi-group structure]].
 
−
*The cumulative migration method <ref name="manual">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, [https://www.sciencedirect.com/science/article/pii/S0306454917303778 118 (2018) 507-516].</ref> (CMM) was first developed for the [https://openmc.readthedocs.io 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 energy-dependent correction factors using the [[Input syntax manual#set trc|set trc]] option.
 
−
*Calculation of CMM_TRANSPXS and CMM_DIFFCOEF requires that their calculation is not switched off using the [[Input syntax manual#set cmm|set cmm]] option.
 
  
 
{|class="wikitable" style="text-align: left;"
 
{|class="wikitable" style="text-align: left;"
Line 1,658: Line 1,842:
 
! Size
 
! Size
 
! Description
 
! Description
 +
|- 
 +
| GC_UNIVERSE_NAME
 +
| (string)
 +
| Name of the universe where spatial homogenization was performed
 
|-
 
|-
−
| INF_TRANSPXS
+
| MICRO_NG
−
| 2''G''
+
| 1
−
| Transport cross section (calculated using the out-scattering approximation)
+
| Number of energy groups in the intermediate multi-group structure (referred to as ''H'' below)
 
|-
 
|-
−
| INF_DIFFCOEF
+
| MICRO_E
−
| 2''G''
+
| ''H'' + 1
−
| Diffusion coefficient (calculated using the out-scattering approximation)
+
| Group boundaries in the intermediate multi-group structure (in ascending order)
 
|-
 
|-
−
| CMM_TRANSPXS
+
| MACRO_NG
−
| 2''G''
+
| 1
−
| Transport cross section calculated using the cumulative migration method (equivalent with the in-scattering approximation)
+
| Number of energy groups in the final few-group structure (referred to as ''G'' below)
 
|-
 
|-
−
| CMM_TRANSPXS_X
+
| MACRO_E
−
| 2''G''
+
| ''G'' + 1
−
| X-component of the directional transport cross section calculated using the cumulative migration method (equivalent with the in-scattering approximation)
+
| Group boundaries in the final few-group structure (in descending order)
 
|-
 
|-
−
| CMM_TRANSPXS_Y
+
|}
−
| 2''G''
+
 
−
| Y-component of the directional transport cross section calculated using the cumulative migration method (equivalent with the in-scattering approximation)
+
=== Group constants homogenized in infinite spectrum ===
−
|-
+
 
−
| CMM_TRANSPXS_Z
+
{|class="wikitable" style="text-align: left;"
 +
! Parameter
 +
! Size
 +
! Description
 +
|- 
 +
| INF_MICRO_FLX
 +
| 2''H''
 +
| Multi-group flux spectrum (integral, un-normalized)
 +
|-
 +
| INF_FLX
 
| 2''G''
 
| 2''G''
−
| Z-component of the directional transport cross section calculated using the cumulative migration method (equivalent with the in-scattering approximation)
+
| Few-group flux (integral, normalized)
 +
|-
 +
| INF_KINF
 +
| 2
 +
| Infinite multiplication factor
 
|-
 
|-
−
| CMM_DIFFCOEF
+
|}
−
| 2''G''
+
 
−
| Diffusion coefficient calculated using the cumulative migration method (equivalent with the in-scattering approximation)
+
==== Reaction cross sections ====
 +
 
 +
{|class="wikitable" style="text-align: left;"
 +
! Parameter
 +
! Size
 +
! Description
 
|-
 
|-
−
| CMM_DIFFCOEF_X
+
| INF_TOT
 
| 2''G''
 
| 2''G''
−
| X-component of the directional diffusion coefficient calculated using the cumulative migration method (equivalent with the in-scattering approximation)
+
| Total cross section
 
|-
 
|-
−
| CMM_DIFFCOEF_Y
+
| INF_CAPT
 
| 2''G''
 
| 2''G''
−
| Y-component of the directional diffusion coefficient calculated using the cumulative migration method (equivalent with the in-scattering approximation)
+
| Capture cross section
 
|-
 
|-
−
| CMM_DIFFCOEF_Z
+
| INF_FISS
 
| 2''G''
 
| 2''G''
−
| Z-component of the directional diffusion coefficient calculated using the cumulative migration method (equivalent with the in-scattering approximation)
+
| Fission cross section
 
|-
 
|-
−
| TRC_TRANSPXS
+
| INF_NSF
 
| 2''G''
 
| 2''G''
−
| Transport cross section calculated by applying user-defined transport correction factors to total cross section
+
| Fission neutron production cross section
 
|-
 
|-
−
| TRC_DIFFCOEF
+
| INF_KAPPA
 
| 2''G''
 
| 2''G''
−
| Diffusion coefficient calculated by applying user-defined transport correction factors to total cross section
+
| Average deposited fission energy (MeV)
−
|}
 
−
 
 
−
==== Poison cross sections ====
 
−
 
 
−
<u>Notes:</u>
 
−
 
 
−
*Printed only if poison cross section option is on (see [[Input syntax manual#set poi|set poi]]).
 
−
*Xe-135m values printed only if separate treatment of Xe-135m is on (see [[Input syntax manual#set poi|set poi]]).
 
−
 
 
−
{|class="wikitable" style="text-align: left;"
 
−
! Parameter
 
−
! Size
 
−
! Description
 
 
|-
 
|-
−
| INF_I135_YIELD
+
| INF_INVV
 
| 2''G''
 
| 2''G''
−
| Fission yield of I-135 (cumulative, includes all precursors)
+
| Inverse neutron speed (s/cm)
 
|-
 
|-
−
| INF_XE135_YIELD
+
| INF_NUBAR
 
| 2''G''
 
| 2''G''
−
| Fission yield of Xe-135
+
| Average neutron yield
 
|-
 
|-
−
| INF_XE135M_YIELD
+
| INF_ABS
 
| 2''G''
 
| 2''G''
−
| Fission yield of Xe-135m
+
| Absorption cross section (capture + fission)
 
|-
 
|-
−
| INF_PM149_YIELD
+
| INF_REMXS
 
| 2''G''
 
| 2''G''
−
| Fission yield of Pm-149 (cumulative, includes all precursors)
+
| Removal cross section (group-removal + absorption)
 
|-
 
|-
−
| INF_SM149_YIELD
+
| INF_RABSXS
 
| 2''G''
 
| 2''G''
−
| Fission yield of Sm-149
+
| Reduced absorption cross section (total - scattering production)
 +
|-
 +
|}
 +
 
 +
==== Fission spectra ====
 +
 
 +
{|class="wikitable" style="text-align: left;"
 +
! Parameter
 +
! Size
 +
! Description
 
|-
 
|-
−
| INF_I135_MICRO_ABS
+
| INF_CHIT
 
| 2''G''
 
| 2''G''
−
| Microscopic absorption cross section of I-135
+
| Fission spectrum (total)
 
|-
 
|-
−
| INF_XE135_MICRO_ABS
+
| INF_CHIP
 
| 2''G''
 
| 2''G''
−
| Microscopic absorption cross section of Xe-135
+
| Fission spectrum (prompt neutrons)
 
|-
 
|-
−
| INF_XE135M_MICRO_ABS
+
| INF_CHID
 
| 2''G''
 
| 2''G''
−
| Microscopic absorption cross section of Xe-135m
+
| Fission spectrum (delayed neutrons)
 
|-
 
|-
−
| INF_PM149_MICRO_ABS
+
|}
 +
 
 +
==== Scattering cross sections ====
 +
 
 +
<u>Notes:</u>
 +
 
 +
*Scattering production includes multiplying (n,2n), (n,3n), etc. reactions.
 +
 
 +
{|class="wikitable" style="text-align: left;"
 +
! Parameter
 +
! Size
 +
! Description
 +
|-
 +
| INF_SCATT0
 
| 2''G''
 
| 2''G''
−
| Microscopic absorption cross section of Pm-149
+
| Total ''P''<sub>0</sub> scattering cross section
 
|-
 
|-
−
| INF_SM149_MICRO_ABS
+
| INF_SCATT1
 
| 2''G''
 
| 2''G''
−
| Microscopic absorption cross section of Sm-149
+
| Total ''P''<sub>1</sub> scattering cross section
 
|-
 
|-
−
| INF_XE135_MACRO_ABS
+
| INF_SCATT2
 
| 2''G''
 
| 2''G''
−
| Macroscopic absorption cross section of Xe-135
+
| Total ''P''<sub>2</sub> scattering cross section
 
|-
 
|-
−
| INF_XE135M_MACRO_ABS
+
| INF_SCATT3
 
| 2''G''
 
| 2''G''
−
| Macroscopic absorption cross section of Xe-135m
+
| Total ''P''<sub>3</sub> scattering cross section
 
|-
 
|-
−
| INF_SM149_MACRO_ABS
+
| INF_SCATT4
 
| 2''G''
 
| 2''G''
−
| Macroscopic absorption cross section of Sm-149
+
| Total ''P''<sub>4</sub> scattering cross section
 
|-
 
|-
−
|}
+
| INF_SCATT5
−
==== Poison decay constants ====
+
| 2''G''
−
 
+
| Total ''P''<sub>5</sub> scattering cross section
−
{|class="wikitable" style="text-align: left;"
 
−
! Parameter
 
−
! Size
 
−
! Description
 
 
|-
 
|-
−
| PM147_LAMBDA
+
| INF_SCATT6
−
| 1
+
| 2''G''
−
| Decay constant of Pm-147
+
| Total ''P''<sub>6</sub> scattering cross section
 
|-
 
|-
−
| PM148_LAMBDA
+
| INF_SCATT7
−
| 1
+
| 2''G''
−
| Decay constant of Pm-147
+
| Total ''P''<sub>7</sub> scattering cross section
 
|-
 
|-
−
| PM148M_LAMBDA
+
| INF_SCATTP0
−
| 1
+
| 2''G''
−
| Decay constant of Pm-148m
+
| Total ''P''<sub>0</sub> scattering production cross section
 
|-
 
|-
−
| PM149_LAMBDA
+
| INF_SCATTP1
−
| 1
+
| 2''G''
−
| Decay constant of Pm-149
+
| Total ''P''<sub>1</sub> scattering production cross section
 +
|-
 +
| INF_SCATTP2
 +
| 2''G''
 +
| Total ''P''<sub>2</sub> scattering production cross section
 
|-
 
|-
−
| I135_LAMBDA
+
| INF_SCATTP3
−
| 1
+
| 2''G''
−
| Decay constant of I-135
+
| Total ''P''<sub>3</sub> scattering production cross section
 
|-
 
|-
−
| XE135_LAMBDA
+
| INF_SCATTP4
−
| 1
+
| 2''G''
−
| Decay constant of Xe-135
+
| Total ''P''<sub>4</sub> scattering production cross section
 
|-
 
|-
−
| XE135M_LAMBDA
+
| INF_SCATTP5
−
| 1
+
| 2''G''
−
| Decay constant of Xe-135m
+
| Total ''P''<sub>5</sub> scattering production cross section
 
|-
 
|-
−
| I135_BR
+
| INF_SCATTP6
−
| 1
+
| 2''G''
−
| Branching ratio of I-135 decay to Xe-135. Branching ratio of I-135 decay to Xe-135m is (1 - I135_BR).
+
| Total ''P''<sub>6</sub> scattering production cross section
 
|-
 
|-
−
|}
+
| INF_SCATTP7
−
 
+
| 2''G''
−
=== Group constants homogenized in leakage-corrected spectrum ===  
+
| Total ''P''<sub>7</sub> scattering production cross section
−
<span id="B1"></span>
+
|-
 +
|}
 +
 
 +
==== Scattering matrices ====
 +
<u>Notes:</u>
 +
 
 +
*Scattering production includes multiplying (n,2n), (n,3n), etc. reactions.
 +
*The order of values (<tt>[input].coe</tt>) or value pairs (<tt>[input]_res.m</tt>) is: <math>\Sigma_{1,1} \, \Sigma_{1,2} \, ... \, \Sigma_{2,1} \, \Sigma_{2,2}  \, ...</math> where <math>\Sigma_{g,g'}</math> refers to scattering from group ''g'' to ''g'''.
 +
*The data in the <tt>[input]_res.m</tt> file can be read into a ''G'' &times; ''G'' matrix with Matlab reshape-command, for example:
 +
<nowiki>
 +
reshape(INF_S0(idx,1:2:end), G, G);</nowiki>
  
 
{|class="wikitable" style="text-align: left;"
 
{|class="wikitable" style="text-align: left;"
Line 1,821: Line 2,046:
 
! Size
 
! Size
 
! Description
 
! Description
−
|- 
 
−
| B1_MICRO_FLX
 
−
| 2''H''
 
−
| Multi-group flux spectrum (integral, un-normalized)
 
 
|-
 
|-
−
| B1_FLX
+
| INF_S0
−
| 2''G''
+
| 2''G''<sup>2</sup>
−
| Few-group flux (integral, normalized)
+
| ''P''<sub>0</sub> scattering matrix
 
|-
 
|-
−
| B1_KINF
+
| INF_S1
−
| 2
+
| 2''G''<sup>2</sup>
−
| Infinite multiplication factor
+
| ''P''<sub>1</sub> scattering matrix
 +
|-
 +
| INF_S2
 +
| 2''G''<sup>2</sup>
 +
| ''P''<sub>2</sub> scattering matrix
 
|-
 
|-
−
| B1_KEFF
+
| INF_S3
−
| 2
+
| 2''G''<sup>2</sup>
−
| Effective multiplication factor
+
| ''P''<sub>3</sub> scattering matrix
 
|-
 
|-
−
| B1_B2
+
| INF_S4
−
| 2
+
| 2''G''<sup>2</sup>
−
| Critical buckling
+
| ''P''<sub>4</sub> scattering matrix
 
|-
 
|-
−
| B1_ERR
+
| INF_S5
−
| 2
+
| 2''G''<sup>2</sup>
−
| Absolute deviation of ''k''<sub>eff</sub> from unity
+
| ''P''<sub>5</sub> scattering matrix
 
|-
 
|-
−
|}
+
| INF_S6
−
 
+
| 2''G''<sup>2</sup>
−
==== Reaction cross sections ====
+
| ''P''<sub>6</sub> scattering matrix
−
 
 
−
{|class="wikitable" style="text-align: left;"
 
−
! Parameter
 
−
! Size
 
−
! Description
 
−
|- 
 
−
| B1_TOT
 
−
| 2''G''
 
−
| Total cross section
 
 
|-
 
|-
−
| B1_CAPT
+
| INF_S7
−
| 2''G''
+
| 2''G''<sup>2</sup>
−
| Capture cross section
+
| ''P''<sub>7</sub> scattering matrix
 
|-
 
|-
−
| B1_FISS
+
| INF_SP0
−
| 2''G''
+
| 2''G''<sup>2</sup>
−
| Fission cross section
+
| ''P''<sub>0</sub> scattering production matrix
 
|-
 
|-
−
| B1_NSF
+
| INF_SP1
−
| 2''G''
+
| 2''G''<sup>2</sup>
−
| Fission neutron production cross section
+
| ''P''<sub>1</sub> scattering production matrix
 
|-
 
|-
−
| B1_KAPPA
+
| INF_SP2
−
| 2''G''
+
| 2''G''<sup>2</sup>
−
| Average deposited fission energy (MeV)
+
| ''P''<sub>2</sub> scattering production matrix
 
|-
 
|-
−
| B1_INVV
+
| INF_SP3
−
| 2''G''
+
| 2''G''<sup>2</sup>
−
| Inverse neutron speed (s/cm)
+
| ''P''<sub>3</sub> scattering production matrix
 
|-
 
|-
−
| B1_NUBAR
+
| INF_SP4
−
| 2''G''
+
| 2''G''<sup>2</sup>
−
| Average neutron yield
+
| ''P''<sub>4</sub> scattering production matrix
 
|-
 
|-
−
| B1_ABS
+
| INF_SP5
−
| 2''G''
+
| 2''G''<sup>2</sup>
−
| Absorption cross section (capture + fission)
+
| ''P''<sub>5</sub> scattering production matrix
 
|-
 
|-
−
| B1_REMXS
+
| INF_SP6
−
| 2''G''
+
| 2''G''<sup>2</sup>
−
| Removal cross section (group-removal + absorption)
+
| ''P''<sub>6</sub> scattering production matrix
 
|-
 
|-
−
| B1_RABSXS
+
| INF_SP7
−
| 2''G''
+
| 2''G''<sup>2</sup>
−
| Reduced absorption cross section (total - scattering production)
+
| ''P''<sub>7</sub> scattering production matrix
 
|-
 
|-
 
|}
 
|}
  
−
==== Fission spectra ====
+
==== Diffusion parameters ====
 +
 
 +
<u>Notes:</u>
 +
*Calculation of sensible values for INF_TRANSPXS and INF_DIFFCOEF requires fine enough [[Input syntax manual#set micro|intermediate multi-group structure]].
 +
*The cumulative migration method (CMM) is described in related papers<ref>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. Nuc. Energy, [https://doi.org/10.1016/j.anucene.2017.10.039 '''112''' (2016) 126-136]</ref><ref>Liu, Z., Smith, K. and Forget, B. ''"Group-wise Tally Scheme of Incremental Migration Area for Cumulative Migration Method."'' In Proceedings of the PHYSOR 2018 (2018) 2512-2523</ref>.
 +
*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 energy-dependent correction factors using the [[Input syntax manual#set trc|set trc]] option.
 +
*Calculation of CMM_TRANSPXS and CMM_DIFFCOEF requires that their calculation is not switched off using the [[Input syntax manual#set cmm|set cmm]] option.
  
 
{|class="wikitable" style="text-align: left;"
 
{|class="wikitable" style="text-align: left;"
Line 1,904: Line 2,127:
 
! Description
 
! Description
 
|-
 
|-
−
| B1_CHIT
+
| INF_TRANSPXS
 
| 2''G''
 
| 2''G''
−
| Fission spectrum (total)
+
| Transport cross section (calculated using the out-scattering approximation)
 
|-
 
|-
−
| B1_CHIP
+
| INF_DIFFCOEF
 
| 2''G''
 
| 2''G''
−
| Fission spectrum (prompt neutrons)
+
| Diffusion coefficient (calculated using the out-scattering approximation)
 
|-
 
|-
−
| B1_CHID
+
| CMM_TRANSPXS
 
| 2''G''
 
| 2''G''
−
| Fission spectrum (delayed neutrons)
+
| Transport cross section calculated using the cumulative migration method (equivalent with the in-scattering approximation)
 
|-
 
|-
−
|}
+
| CMM_TRANSPXS_X
−
 
+
| 2''G''
−
==== Scattering cross sections ====
+
| X-component of the directional transport cross section calculated using the cumulative migration method (equivalent with the in-scattering approximation)
−
 
 
−
<u>Notes:</u>
 
−
 
 
−
*Scattering production includes multiplying (n,2n), (n,3n), etc. reactions.
 
−
 
 
−
{|class="wikitable" style="text-align: left;"
 
−
! Parameter
 
−
! Size
 
−
! Description
 
 
|-
 
|-
−
| B1_SCATT0
+
| CMM_TRANSPXS_Y
 
| 2''G''
 
| 2''G''
−
| Total ''P''<sub>0</sub> scattering cross section
+
| Y-component of the directional transport cross section calculated using the cumulative migration method (equivalent with the in-scattering approximation)
 
|-
 
|-
−
| B1_SCATT1
+
| CMM_TRANSPXS_Z
 
| 2''G''
 
| 2''G''
−
| Total ''P''<sub>1</sub> scattering cross section
+
| Z-component of the directional transport cross section calculated using the cumulative migration method (equivalent with the in-scattering approximation)
 
|-
 
|-
−
| B1_SCATT2
+
| CMM_DIFFCOEF
 
| 2''G''
 
| 2''G''
−
| Total ''P''<sub>2</sub> scattering cross section
+
| Diffusion coefficient calculated using the cumulative migration method (equivalent with the in-scattering approximation)
 
|-
 
|-
−
| B1_SCATT3
+
| CMM_DIFFCOEF_X
 
| 2''G''
 
| 2''G''
−
| Total ''P''<sub>3</sub> scattering cross section
+
| X-component of the directional diffusion coefficient calculated using the cumulative migration method (equivalent with the in-scattering approximation)
 
|-
 
|-
−
| B1_SCATT4
+
| CMM_DIFFCOEF_Y
 
| 2''G''
 
| 2''G''
−
| Total ''P''<sub>4</sub> scattering cross section
+
| Y-component of the directional diffusion coefficient calculated using the cumulative migration method (equivalent with the in-scattering approximation)
 
|-
 
|-
−
| B1_SCATT5
+
| CMM_DIFFCOEF_Z
 
| 2''G''
 
| 2''G''
−
| Total ''P''<sub>5</sub> scattering cross section
+
| Z-component of the directional diffusion coefficient calculated using the cumulative migration method (equivalent with the in-scattering approximation)
 
|-
 
|-
−
| B1_SCATT6
+
| TRC_TRANSPXS
 
| 2''G''
 
| 2''G''
−
| Total ''P''<sub>6</sub> scattering cross section
+
| Transport cross section calculated by applying user-defined transport correction factors to total cross section
 
|-
 
|-
−
| B1_SCATT7
+
| TRC_DIFFCOEF
 
| 2''G''
 
| 2''G''
−
| Total ''P''<sub>7</sub> scattering cross section
+
| Diffusion coefficient calculated by applying user-defined transport correction factors to total cross section
−
|-
+
|}
−
| B1_SCATTP0
+
 
−
| 2''G''
+
==== Poison cross sections ====
−
| Total ''P''<sub>0</sub> scattering production cross section
+
 
−
|-
+
<u>Notes:</u>
−
| B1_SCATTP1
+
 
 +
*Printed only if poison cross section option is on (see [[Input syntax manual#set poi|set poi]]).
 +
*Xe-135m values printed only if separate treatment of Xe-135m is on (see [[Input syntax manual#set poi|set poi]]).
 +
*Approximate effective treatment of Pm-149 production from Pm-148 and Pm-148m is not included in Pm-149 fission yield.
 +
*All values for I-135, Pm-147, Pm-148 and Pm-148m not present in all versions.
 +
 
 +
{|class="wikitable" style="text-align: left;"
 +
! Parameter
 +
! Size
 +
! Description
 +
|-
 +
| INF_I135_YIELD
 
| 2''G''
 
| 2''G''
−
| Total ''P''<sub>1</sub> scattering production cross section
+
| Fission yield of I-135 (cumulative, includes all precursors)
 
|-
 
|-
−
| B1_SCATTP2
+
| INF_XE135_YIELD
 
| 2''G''
 
| 2''G''
−
| Total ''P''<sub>2</sub> scattering production cross section
+
| Fission yield of Xe-135
 
|-
 
|-
−
| B1_SCATTP3
+
| INF_XE135M_YIELD
 
| 2''G''
 
| 2''G''
−
| Total ''P''<sub>3</sub> scattering production cross section
+
| Fission yield of Xe-135m
 
|-
 
|-
−
| B1_SCATTP4
+
| INF_PM149_YIELD
 
| 2''G''
 
| 2''G''
−
| Total ''P''<sub>4</sub> scattering production cross section
+
| Fission yield of Pm-149 (cumulative, includes all precursors)
 
|-
 
|-
−
| B1_SCATTP5
+
| INF_SM149_YIELD
 
| 2''G''
 
| 2''G''
−
| Total ''P''<sub>5</sub> scattering production cross section
+
| Fission yield of Sm-149
 
|-
 
|-
−
| B1_SCATTP6
+
| INF_I135_MICRO_ABS
 
| 2''G''
 
| 2''G''
−
| Total ''P''<sub>6</sub> scattering production cross section
+
| Microscopic absorption cross section of I-135
 
|-
 
|-
−
| B1_SCATTP7
+
| INF_XE135_MICRO_ABS
 
| 2''G''
 
| 2''G''
−
| Total ''P''<sub>7</sub> scattering production cross section
+
| Microscopic absorption cross section of Xe-135
 
|-
 
|-
−
|}
+
| INF_XE135M_MICRO_ABS
−
 
+
| 2''G''
−
==== Scattering matrices ====
+
| Microscopic absorption cross section of Xe-135m
−
 
 
−
<u>Notes:</u>
 
−
 
 
−
*Scattering production includes multiplying (n,2n), (n,3n), etc. reactions.
 
−
*The order of values ([input].coe) or value pairs ([input]_res.m) is: <math>\Sigma_{1,1} \, \Sigma_{1,2} \, ... \, \Sigma_{2,1} \, \Sigma_{2,2}  \, ...</math> where <math>\Sigma_{g,g'}</math> 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:
 
−
<nowiki>
 
−
reshape(B1_S0(idx,1:2:end), G, G)</nowiki>.
 
−
 
 
−
{|class="wikitable" style="text-align: left;"
 
−
! Parameter
 
−
! Size
 
−
! Description
 
 
|-
 
|-
−
| B1_S0
+
| INF_PM147_MICRO_ABS
−
| 2''G''<sup>2</sup>
+
| 2''G''
−
| ''P''<sub>0</sub> scattering matrix
+
| Microscopic absorption cross section of Pm-147
 
|-
 
|-
−
| B1_S1
+
| INF_PM148_MICRO_ABS
−
| 2''G''<sup>2</sup>
+
| 2''G''
−
| ''P''<sub>1</sub> scattering matrix
+
| Microscopic absorption cross section of Pm-148
−
|-  
 
−
| B1_S2
 
−
| 2''G''<sup>2</sup>
 
−
| ''P''<sub>2</sub> scattering matrix
 
 
|-
 
|-
−
| B1_S3
+
| INF_PM148M_MICRO_ABS
−
| 2''G''<sup>2</sup>
+
| 2''G''
−
| ''P''<sub>3</sub> scattering matrix
+
| Microscopic absorption cross section of Pm-148m
 
|-
 
|-
−
| B1_S4
+
| INF_PM149_MICRO_ABS
−
| 2''G''<sup>2</sup>
+
| 2''G''
−
| ''P''<sub>4</sub> scattering matrix
+
| Microscopic absorption cross section of Pm-149
 
|-
 
|-
−
| B1_S5
+
| INF_SM149_MICRO_ABS
−
| 2''G''<sup>2</sup>
+
| 2''G''
−
| ''P''<sub>5</sub> scattering matrix
+
| Microscopic absorption cross section of Sm-149
 
|-
 
|-
−
| B1_S6
+
| INF_I135_MACRO_ABS
−
| 2''G''<sup>2</sup>
+
| 2''G''
−
| ''P''<sub>6</sub> scattering matrix
+
| Macroscopic absorption cross section of I-135
 
|-
 
|-
−
| B1_S7
+
| INF_XE135_MACRO_ABS
−
| 2''G''<sup>2</sup>
+
| 2''G''
−
| ''P''<sub>7</sub> scattering matrix
+
| Macroscopic absorption cross section of Xe-135
 
|-
 
|-
−
| B1_SP0
+
| INF_XE135M_MACRO_ABS
−
| 2''G''<sup>2</sup>
+
| 2''G''
−
| ''P''<sub>0</sub> scattering production matrix
+
| Macroscopic absorption cross section of Xe-135m
 
|-
 
|-
−
| B1_SP1
+
| INF_PM147_MACRO_ABS
−
| 2''G''<sup>2</sup>
+
| 2''G''
−
| ''P''<sub>1</sub> scattering production matrix
+
| Macroscopic absorption cross section of Pm-147
 
|-
 
|-
−
| B1_SP2
+
| INF_PM148_MACRO_ABS
−
| 2''G''<sup>2</sup>
+
| 2''G''
−
| ''P''<sub>2</sub> scattering production matrix
+
| Macroscopic absorption cross section of Pm-148
 
|-
 
|-
−
| B1_SP3
+
| INF_PM148M_MACRO_ABS
−
| 2''G''<sup>2</sup>
+
| 2''G''
−
| ''P''<sub>3</sub> scattering production matrix
+
| Macroscopic absorption cross section of Pm-148M
 
|-
 
|-
−
| B1_SP4
+
| INF_PM149_MACRO_ABS
−
| 2''G''<sup>2</sup>
+
| 2''G''
−
| ''P''<sub>4</sub> scattering production matrix
+
| Macroscopic absorption cross section of Pm-149
 
|-
 
|-
−
| B1_SP5
+
| INF_SM149_MACRO_ABS
−
| 2''G''<sup>2</sup>
+
| 2''G''
−
| ''P''<sub>5</sub> scattering production matrix
+
| Macroscopic absorption cross section of Sm-149
−
|-
 
−
| B1_SP6
 
−
| 2''G''<sup>2</sup>
 
−
| ''P''<sub>6</sub> scattering production matrix
 
−
|-
 
−
| B1_SP7
 
−
| 2''G''<sup>2</sup>
 
−
| ''P''<sub>7</sub> scattering production matrix
 
 
|-
 
|-
 
|}
 
|}
  
−
==== Diffusion parameters ====
+
==== Poison universe-averaged densities ====
 +
 
 +
<u>Notes:</u>
 +
 
 +
*The universe-averaged atomic density is defined to be such that ADENS*MICRO_ABS is equal to MACRO_ABS.<ref>Rintala, A., Valtavirta, V. and Leppänen, J.. ''Microscopic cross section calculation methodology in the Serpent 2 Monte Carlo code.'' Annals of Nuclear Energy, [https://doi.org/10.1016/j.anucene.2021.108603 164 (2021): 108603].</ref>
  
 
{|class="wikitable" style="text-align: left;"
 
{|class="wikitable" style="text-align: left;"
Line 2,083: Line 2,287:
 
! Description
 
! Description
 
|-
 
|-
−
| B1_TRANSPXS
+
| I135_ADENS
−
| 2''G''
+
| 2
−
| Transport cross section (outscattering transport cross section collapsed with the critical spectrum when old B<sub>1</sub> calculation mode is used, otherwise calculated from B1_DIFFCOEF)
+
| Universe-averaged atomic density of I-135
 +
|-
 +
| XE135_ADENS
 +
| 2
 +
| Universe-averaged atomic density of Xe-135
 +
|-
 +
| XE135M_ADENS
 +
| 2
 +
| Universe-averaged atomic density of Xe-135m
 +
|-
 +
| PM147_ADENS
 +
| 2
 +
| Universe-averaged atomic density of Pm-147
 +
|-
 +
| PM148_ADENS
 +
| 2
 +
| Universe-averaged atomic density of Pm-147
 +
|-
 +
| PM148M_ADENS
 +
| 2
 +
| Universe-averaged atomic density of Pm-148m
 
|-
 
|-
−
| B1_DIFFCOEF
+
| PM149_ADENS
−
| 2''G''
+
| 2
−
| Diffusion coefficient calculated from during the fundamental mode calculation (old and new B<sub>1</sub> and P<sub>1</sub> calculation modes, or flux collapsed during the FM calculation mode)
+
| Universe-averaged atomic density of Pm-149
 
|-
 
|-
 
|}
 
|}
  
−
==== Poison cross sections ====
+
==== Poison decay constants ====
−
 
 
−
<u>Notes:</u>
 
−
 
 
−
*Printed only if poison cross section option is on (see [[Input syntax manual#set poi|set poi]]).
 
−
*Xe-135m values printed only if separate treatment of Xe-135m is on (see [[Input syntax manual#set poi|set poi]]).
 
  
 
{|class="wikitable" style="text-align: left;"
 
{|class="wikitable" style="text-align: left;"
Line 2,105: Line 2,324:
 
! Description
 
! Description
 
|-
 
|-
−
| B1_I135_YIELD
+
| PM147_LAMBDA
−
| 2''G''
+
| 1
−
| Fission yield of I-135 (cumulative, includes all precursors)
+
| Decay constant of Pm-147
 
|-
 
|-
−
| B1_XE135_YIELD
+
| PM148_LAMBDA
−
| 2''G''
+
| 1
−
| Fission yield of Xe-135
+
| Decay constant of Pm-147
 
|-
 
|-
−
| B1_XE135M_YIELD
+
| PM148M_LAMBDA
−
| 2''G''
+
| 1
−
| Fission yield of Xe-135m
+
| Decay constant of Pm-148m
 
|-
 
|-
−
| B1_PM149_YIELD
+
| PM149_LAMBDA
−
| 2''G''
+
| 1
−
| Fission yield of Pm-149 (cumulative, includes all precursors)
+
| Decay constant of Pm-149
 
|-
 
|-
−
| B1_SM149_YIELD
+
| I135_LAMBDA
−
| 2''G''
+
| 1
−
| Fission yield of Sm-149
+
| Decay constant of I-135
 
|-
 
|-
−
| B1_I135_MICRO_ABS
+
| XE135_LAMBDA
−
| 2''G''
+
| 1
−
| Microscopic absorption cross section of I-135
+
| Decay constant of Xe-135
 
|-
 
|-
−
| B1_XE135_MICRO_ABS
+
| XE135M_LAMBDA
−
| 2''G''
+
| 1
−
| Microscopic absorption cross section of Xe-135
+
| Decay constant of Xe-135m
 
|-
 
|-
−
| B1_XE135M_MICRO_ABS
+
| I135_BR
−
| 2''G''
+
| 1
−
| Microscopic absorption cross section of Xe-135m
+
| Branching ratio of I-135 decay to Xe-135. Branching ratio of I-135 decay to Xe-135m is (1 - I135_BR).
−
|-
 
−
| B1_PM149_MICRO_ABS
 
−
| 2''G''
 
−
| Microscopic absorption cross section of Pm-149
 
−
|-
 
−
| B1_SM149_MICRO_ABS
 
−
| 2''G''
 
−
| Microscopic absorption cross section of Sm-149
 
−
|-
 
−
| B1_XE135_MACRO_ABS
 
−
| 2''G''
 
−
| Macroscopic absorption cross section of Xe-135
 
−
|-
 
−
| B1_XE135M_MACRO_ABS
 
−
| 2''G''
 
−
| Macroscopic absorption cross section of Xe-135m
 
−
|-
 
−
| B1_SM149_MACRO_ABS
 
−
| 2''G''
 
−
| Macroscopic absorption cross section of Sm-149
 
 
|-
 
|-
 
|}
 
|}
  
−
=== Delayed neutron data ===
+
=== Group constants homogenized in leakage-corrected spectrum ===  
−
 
+
<span id="B1"></span>
−
<u>Notes:</u>
 
−
 
 
−
*The output consists of total, followed by ''D'' precursor group-wise 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. 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.
 
  
 
{|class="wikitable" style="text-align: left;"
 
{|class="wikitable" style="text-align: left;"
Line 2,170: Line 2,365:
 
! Size
 
! Size
 
! Description
 
! Description
 +
|- 
 +
| B1_MICRO_FLX
 +
| 2''H''
 +
| Multi-group flux spectrum (integral, un-normalized)
 
|-
 
|-
−
| BETA_EFF
+
| B1_FLX
−
| 2''D'' + 2
+
| 2''G''
−
| Effective delayed neutron fraction (currently calculated using the Meulekamp method)
+
| Few-group flux (integral, normalized)
 +
|-
 +
| B1_KINF
 +
| 2
 +
| Infinite multiplication factor
 +
|-
 +
| B1_KEFF
 +
| 2
 +
| Effective multiplication factor
 +
|-
 +
| B1_B2
 +
| 2
 +
| Critical buckling
 
|-
 
|-
−
| LAMBDA
+
| B1_ERR
−
| 2''D'' + 2
+
| 2
−
| Decay constants
+
| Absolute deviation of ''k''<sub>eff</sub> from unity
 
|-
 
|-
 
|}
 
|}
  
−
=== Assembly discontinuity factors ===
+
==== Reaction cross sections ====
  
−
<u>Notes:</u>
+
{|class="wikitable" style="text-align: left;"
 +
! Parameter
 +
! Size
 +
! Description
 +
|- 
 +
| B1_TOT
 +
| 2''G''
 +
| Total cross section
 +
|-
 +
| B1_CAPT
 +
| 2''G''
 +
| Capture cross section
 +
|-
 +
| B1_FISS
 +
| 2''G''
 +
| Fission cross section
 +
|-
 +
| B1_NSF
 +
| 2''G''
 +
| Fission neutron production cross section
 +
|-
 +
| B1_KAPPA
 +
| 2''G''
 +
| Average deposited fission energy (MeV)
 +
|-
 +
| B1_INVV
 +
| 2''G''
 +
| Inverse neutron speed (s/cm)
 +
|-
 +
| B1_NUBAR
 +
| 2''G''
 +
| Average neutron yield
 +
|-
 +
| B1_ABS
 +
| 2''G''
 +
| Absorption cross section (capture + fission)
 +
|-
 +
| B1_REMXS
 +
| 2''G''
 +
| Removal cross section (group-removal + absorption)
 +
|-
 +
| B1_RABSXS
 +
| 2''G''
 +
| Reduced absorption cross section (total - scattering production)
 +
|-
 +
|}
 +
 
 +
==== Fission spectra ====
  
−
*Calculation of assembly discontinuity factors requires the [[Input syntax manual#set adf|set adf]] option.
+
{|class="wikitable" style="text-align: left;"
−
*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.
+
! Parameter
−
*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.
+
! Size
−
*The currents are surface integrated values.
+
! Description
−
*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]].
+
| B1_CHIT
−
*The calculation currently supports only a limited number of [[surface types]]: infinite planes and square and hexagonal prisms.
+
| 2''G''
−
*The order of surface and mid-point values for square prisms is: <math>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} \, ... </math> and the order of corner values: <math>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} \, ... </math> where <math>X_{k,g}</math> refers to parameter <math>X</math> on surface/corner ''k'' and energy group ''g''.
+
| Fission spectrum (total)
−
*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.
+
| B1_CHIP
−
*The sign moment weighted parameters are calculated only for [[surface types]] sqc, rect and hexxc.
+
| 2''G''
−
*The convention of sign moment directions follows that of the nodal neutronics program Ants.
+
| Fission spectrum (prompt neutrons)
−
*The [[ADF symmetry options]] on [[Input syntax manual#set adf|set adf]] card are currently not used for sign moment weighted parameters.
+
|-
−
 
+
| B1_CHID
−
{|class="wikitable" style="text-align: left;"
+
| 2''G''
−
! Parameter
+
| Fission spectrum (delayed neutrons)
−
! Size
+
|-
−
! Description
+
|}
−
|-
+
 
−
| DF_SURFACE
+
==== Scattering cross sections ====
−
| (string)
+
 
−
| Name of the surface used for the calculation
+
<u>Notes:</u>
−
|-
+
 
−
| DF_SYM
+
*Scattering production includes multiplying (n,2n), (n,3n), etc. reactions.
−
| 1
+
 
−
| Symmetry option defined in the input
+
{|class="wikitable" style="text-align: left;"
−
|-
+
! Parameter
−
| DF_N_SURF
+
! Size
−
| 1
+
! Description
−
| Number of surface values (denoted as ''N''<sub>S</sub> below)
+
|-
 +
| B1_SCATT0
 +
| 2''G''
 +
| Total ''P''<sub>0</sub> scattering cross section
 +
|-
 +
| B1_SCATT1
 +
| 2''G''
 +
| Total ''P''<sub>1</sub> scattering cross section
 +
|-
 +
| B1_SCATT2
 +
| 2''G''
 +
| Total ''P''<sub>2</sub> scattering cross section
 +
|-
 +
| B1_SCATT3
 +
| 2''G''
 +
| Total ''P''<sub>3</sub> scattering cross section
 +
|-
 +
| B1_SCATT4
 +
| 2''G''
 +
| Total ''P''<sub>4</sub> scattering cross section
 +
|-
 +
| B1_SCATT5
 +
| 2''G''
 +
| Total ''P''<sub>5</sub> scattering cross section
 +
|-
 +
| B1_SCATT6
 +
| 2''G''
 +
| Total ''P''<sub>6</sub> scattering cross section
 +
|-
 +
| B1_SCATT7
 +
| 2''G''
 +
| Total ''P''<sub>7</sub> scattering cross section
 +
|-
 +
| B1_SCATTP0
 +
| 2''G''
 +
| Total ''P''<sub>0</sub> scattering production cross section
 +
|-
 +
| B1_SCATTP1
 +
| 2''G''
 +
| Total ''P''<sub>1</sub> scattering production cross section
 +
|-
 +
| B1_SCATTP2
 +
| 2''G''
 +
| Total ''P''<sub>2</sub> scattering production cross section
 +
|-
 +
| B1_SCATTP3
 +
| 2''G''
 +
| Total ''P''<sub>3</sub> scattering production cross section
 +
|-
 +
| B1_SCATTP4
 +
| 2''G''
 +
| Total ''P''<sub>4</sub> scattering production cross section
 +
|-
 +
| B1_SCATTP5
 +
| 2''G''
 +
| Total ''P''<sub>5</sub> scattering production cross section
 +
|-
 +
| B1_SCATTP6
 +
| 2''G''
 +
| Total ''P''<sub>6</sub> scattering production cross section
 +
|-
 +
| B1_SCATTP7
 +
| 2''G''
 +
| Total ''P''<sub>7</sub> scattering production cross section
 +
|-
 +
|}
 +
 
 +
==== Scattering matrices ====
 +
 
 +
<u>Notes:</u>
 +
 
 +
*Scattering production includes multiplying (n,2n), (n,3n), etc. reactions.
 +
*The order of values (<tt>[input].coe</tt>) or value pairs (<tt>[input]_res.m</tt>) is: <math>\Sigma_{1,1} \, \Sigma_{1,2} \, ... \, \Sigma_{2,1} \, \Sigma_{2,2}  \, ...</math> where <math>\Sigma_{g,g'}</math> refers to scattering from group ''g'' to ''g'''.
 +
*The data in the <tt>[input]_res.m</tt> file can be read into a ''G'' &times; ''G'' matrix with Matlab reshape-command, for example:
 +
<nowiki>
 +
reshape(B1_S0(idx,1:2:end), G, G)</nowiki>.
 +
 
 +
{|class="wikitable" style="text-align: left;"
 +
! Parameter
 +
! Size
 +
! Description
 +
|-
 +
| B1_S0
 +
| 2''G''<sup>2</sup>
 +
| ''P''<sub>0</sub> scattering matrix
 +
|-
 +
| B1_S1
 +
| 2''G''<sup>2</sup>
 +
| ''P''<sub>1</sub> scattering matrix
 +
|-
 +
| B1_S2
 +
| 2''G''<sup>2</sup>
 +
| ''P''<sub>2</sub> scattering matrix
 +
|-
 +
| B1_S3
 +
| 2''G''<sup>2</sup>
 +
| ''P''<sub>3</sub> scattering matrix
 +
|-
 +
| B1_S4
 +
| 2''G''<sup>2</sup>
 +
| ''P''<sub>4</sub> scattering matrix
 +
|-
 +
| B1_S5
 +
| 2''G''<sup>2</sup>
 +
| ''P''<sub>5</sub> scattering matrix
 +
|-
 +
| B1_S6
 +
| 2''G''<sup>2</sup>
 +
| ''P''<sub>6</sub> scattering matrix
 +
|-
 +
| B1_S7
 +
| 2''G''<sup>2</sup>
 +
| ''P''<sub>7</sub> scattering matrix
 +
|-
 +
| B1_SP0
 +
| 2''G''<sup>2</sup>
 +
| ''P''<sub>0</sub> scattering production matrix
 +
|-
 +
| B1_SP1
 +
| 2''G''<sup>2</sup>
 +
| ''P''<sub>1</sub> scattering production matrix
 +
|-
 +
| B1_SP2
 +
| 2''G''<sup>2</sup>
 +
| ''P''<sub>2</sub> scattering production matrix
 +
|-
 +
| B1_SP3
 +
| 2''G''<sup>2</sup>
 +
| ''P''<sub>3</sub> scattering production matrix
 +
|-
 +
| B1_SP4
 +
| 2''G''<sup>2</sup>
 +
| ''P''<sub>4</sub> scattering production matrix
 +
|-
 +
| B1_SP5
 +
| 2''G''<sup>2</sup>
 +
| ''P''<sub>5</sub> scattering production matrix
 +
|-
 +
| B1_SP6
 +
| 2''G''<sup>2</sup>
 +
| ''P''<sub>6</sub> scattering production matrix
 +
|-
 +
| B1_SP7
 +
| 2''G''<sup>2</sup>
 +
| ''P''<sub>7</sub> scattering production matrix
 +
|-
 +
|}
 +
 
 +
==== Diffusion parameters ====
 +
 
 +
{|class="wikitable" style="text-align: left;"
 +
! Parameter
 +
! Size
 +
! Description
 +
|-
 +
| B1_TRANSPXS
 +
| 2''G''
 +
| Transport cross section (outscattering transport cross section collapsed with the critical spectrum when old B<sub>1</sub> calculation mode is used, otherwise calculated from B1_DIFFCOEF)
 +
|-
 +
| B1_DIFFCOEF
 +
| 2''G''
 +
| Diffusion coefficient calculated from during the fundamental mode calculation (old and new B<sub>1</sub> and P<sub>1</sub> calculation modes, or flux collapsed during the FM calculation mode)
 +
|-
 +
|}
 +
 
 +
==== Poison cross sections ====
 +
 
 +
<u>Notes:</u>
 +
 
 +
*Printed only if poison cross section option is on (see [[Input syntax manual#set poi|set poi]]).
 +
*Xe-135m values printed only if separate treatment of Xe-135m is on (see [[Input syntax manual#set poi|set poi]]).
 +
 
 +
{|class="wikitable" style="text-align: left;"
 +
! Parameter
 +
! Size
 +
! Description
 +
|-
 +
| B1_I135_YIELD
 +
| 2''G''
 +
| Fission yield of I-135 (cumulative, includes all precursors)
 +
|-
 +
| B1_XE135_YIELD
 +
| 2''G''
 +
| Fission yield of Xe-135
 +
|-
 +
| B1_XE135M_YIELD
 +
| 2''G''
 +
| Fission yield of Xe-135m
 +
|-
 +
| B1_PM149_YIELD
 +
| 2''G''
 +
| Fission yield of Pm-149 (cumulative, includes all precursors)
 +
|-
 +
| B1_SM149_YIELD
 +
| 2''G''
 +
| Fission yield of Sm-149
 +
|-
 +
| B1_I135_MICRO_ABS
 +
| 2''G''
 +
| Microscopic absorption cross section of I-135
 +
|-
 +
| B1_XE135_MICRO_ABS
 +
| 2''G''
 +
| Microscopic absorption cross section of Xe-135
 +
|-
 +
| B1_XE135M_MICRO_ABS
 +
| 2''G''
 +
| Microscopic absorption cross section of Xe-135m
 +
|-
 +
| B1_PM149_MICRO_ABS
 +
| 2''G''
 +
| Microscopic absorption cross section of Pm-149
 +
|-
 +
| B1_SM149_MICRO_ABS
 +
| 2''G''
 +
| Microscopic absorption cross section of Sm-149
 +
|-
 +
| B1_XE135_MACRO_ABS
 +
| 2''G''
 +
| Macroscopic absorption cross section of Xe-135
 +
|-
 +
| B1_XE135M_MACRO_ABS
 +
| 2''G''
 +
| Macroscopic absorption cross section of Xe-135m
 +
|-
 +
| B1_SM149_MACRO_ABS
 +
| 2''G''
 +
| Macroscopic absorption cross section of Sm-149
 +
|-
 +
|}
 +
 
 +
=== Delayed neutron data ===
 +
 
 +
<u>Notes:</u>
 +
 
 +
*The output consists of total, followed by ''D'' precursor group-wise 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. 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.
 +
 
 +
{|class="wikitable" style="text-align: left;"
 +
! Parameter
 +
! Size
 +
! Description
 +
|-
 +
| BETA_EFF
 +
| 2''D'' + 2
 +
| Effective delayed neutron fraction (currently calculated using the Meulekamp method)
 +
|-
 +
| LAMBDA
 +
| 2''D'' + 2
 +
| Decay constants
 +
|-
 +
|}
 +
 
 +
=== Assembly discontinuity factors ===
 +
 
 +
<u>Notes:</u>
 +
 
 +
*Calculation of assembly discontinuity factors requires the [[Input syntax manual#set adf|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: <math>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} \, ... </math> and the order of corner values: <math>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} \, ... </math> where <math>X_{k,g}</math> refers to parameter <math>X</math> 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 [[Input syntax manual#set adf|set adf]] card are currently not used for sign moment weighted parameters.
 +
 
 +
{|class="wikitable" style="text-align: left;"
 +
! Parameter
 +
! Size
 +
! Description
 +
|-
 +
| DF_SURFACE
 +
| (string)
 +
| Name of the surface used for the calculation
 +
|-
 +
| DF_SURFACE_TYPE
 +
| 1
 +
| Surface type
 +
|-
 +
| DF_SURFACE_N_PARAMS
 +
| 1
 +
| Number of parameters defining the surface
 +
|-
 +
| DF_SURFACE_PARAMS
 +
| ''N''<sub>P</sub>
 +
| List of parameters defining the surface
 +
|-
 +
| DF_SYM
 +
| 1
 +
| Symmetry option defined in the input
 +
|-
 +
| DF_N_SURF
 +
| 1
 +
| Number of surface values (denoted as ''N''<sub>S</sub> below)
 +
|-
 +
| DF_N_CORN
 +
| 1
 +
| Number of corner values (denoted as ''N''<sub>C</sub> below)
 
|-
 
|-
−
| DF_N_CORN
+
| DF_N_SGN
 
| 1
 
| 1
−
| Number of corner values (denoted as ''N''<sub>C</sub> below)
+
| Number of sign moment values (denoted as ''N''<sub>M</sub> below)
 
|-
 
|-
 
| DF_VOLUME
 
| DF_VOLUME
Line 2,237: Line 2,797:
 
|-
 
|-
 
| DF_SURF_IN_CURR
 
| DF_SURF_IN_CURR
−
| ''2G'' <math>\times</math> ''N''<sub>S</sub>
+
| ''2G'' &times; ''N<sub>S</sub>''
 
| Inward surface currents
 
| Inward surface currents
 
|-
 
|-
 
| DF_SURF_OUT_CURR
 
| DF_SURF_OUT_CURR
−
| ''2G'' <math>\times</math> ''N''<sub>S</sub>
+
| ''2G'' &times; ''N<sub>S</sub>''
 
| Outward surface currents
 
| Outward surface currents
 
|-
 
|-
 
| DF_SURF_NET_CURR
 
| DF_SURF_NET_CURR
−
| ''2G'' <math>\times</math> ''N''<sub>S</sub>
+
| ''2G'' &times; ''N<sub>S</sub>''
 
| Net surface currents
 
| Net surface currents
 
|-
 
|-
 
| DF_MID_IN_CURR
 
| DF_MID_IN_CURR
−
| ''2G'' <math>\times</math> ''N''<sub>S</sub>
+
| ''2G'' &times; ''N<sub>S</sub>''
 
| Inward mid-point currents
 
| Inward mid-point currents
 
|-
 
|-
 
| DF_MID_OUT_CURR
 
| DF_MID_OUT_CURR
−
| ''2G'' <math>\times</math> ''N''<sub>S</sub>
+
| ''2G'' &times; ''N<sub>S</sub>''
 
| Outward mid-point currents
 
| Outward mid-point currents
 
|-
 
|-
 
| DF_MID_NET_CURR
 
| DF_MID_NET_CURR
−
| ''2G'' <math>\times</math> ''N''<sub>S</sub>
+
| ''2G'' &times; ''N<sub>S</sub>''
 
| Net mid-point currents
 
| Net mid-point currents
 
|-
 
|-
 
| DF_CORN_IN_CURR
 
| DF_CORN_IN_CURR
−
| ''2G'' <math>\times</math> ''N''<sub>C</sub>
+
| ''2G'' &times; ''N<sub>C</sub>''
 
| Inward corner currents
 
| Inward corner currents
 
|-
 
|-
 
| DF_CORN_OUT_CURR
 
| DF_CORN_OUT_CURR
−
| ''2G'' <math>\times</math> ''N''<sub>C</sub>
+
| ''2G'' &times; ''N<sub>C</sub>''
 
| Outward corner currents
 
| Outward corner currents
 
|-
 
|-
 
| DF_CORN_NET_CURR
 
| DF_CORN_NET_CURR
−
| ''2G'' <math>\times</math> ''N''<sub>C</sub>
+
| ''2G'' &times; ''N<sub>C</sub>''
 
| Net corner currents
 
| Net corner currents
 
|-
 
|-
Line 2,277: Line 2,837:
 
|-
 
|-
 
| DF_HET_SURF_FLUX
 
| DF_HET_SURF_FLUX
−
| ''2G'' <math>\times</math> ''N''<sub>S</sub>
+
| ''2G'' &times; ''N<sub>S</sub>''
 
| Heterogeneous surface fluxes
 
| Heterogeneous surface fluxes
 
|-
 
|-
 
| DF_HET_CORN_FLUX
 
| DF_HET_CORN_FLUX
−
| ''2G'' <math>\times</math> ''N''<sub>C</sub>
+
| ''2G'' &times; ''N<sub>C</sub>''
 
| Heterogeneous corner fluxes
 
| Heterogeneous corner fluxes
 
|-
 
|-
Line 2,289: Line 2,849:
 
|-
 
|-
 
| DF_HOM_SURF_FLUX
 
| DF_HOM_SURF_FLUX
−
| ''2G'' <math>\times</math> ''N''<sub>S</sub>
+
| ''2G'' &times; ''N<sub>S</sub>''
 
| Homogeneous surface fluxes
 
| Homogeneous surface fluxes
 
|-
 
|-
 
| DF_HOM_CORN_FLUX
 
| DF_HOM_CORN_FLUX
−
| ''2G'' <math>\times</math> ''N''<sub>C</sub>
+
| ''2G'' &times; ''N<sub>C</sub>''
 
| Homogeneous corner fluxes
 
| Homogeneous corner fluxes
 
|-
 
|-
 
| DF_SURF_DF
 
| DF_SURF_DF
−
| ''2G'' <math>\times</math> ''N''<sub>S</sub>
+
| ''2G'' &times; ''N<sub>S</sub>''
 
| Surface discontinuity factors
 
| Surface discontinuity factors
 
|-
 
|-
 
| DF_CORN_DF
 
| DF_CORN_DF
−
| ''2G'' <math>\times</math> ''N''<sub>C</sub>
+
| ''2G'' &times; ''N<sub>C</sub>''
 
| Corner discontinuity factors
 
| Corner discontinuity factors
 
|-
 
|-
 
| DF_SGN_SURF_IN_CURR  
 
| DF_SGN_SURF_IN_CURR  
−
| ''2G'' <math>\times</math> ''N''<sub>S</sub>
+
| ''2G'' &times; ''N<sub>M</sub>''
 
| Inward sign moment weighted currents
 
| Inward sign moment weighted currents
 
|-
 
|-
 
| DF_SGN_SURF_OUT_CURR
 
| DF_SGN_SURF_OUT_CURR
−
| ''2G'' <math>\times</math> ''N''<sub>S</sub>
+
| ''2G'' &times; ''N<sub>M</sub>''
 
| Outward sign moment weighted currents
 
| Outward sign moment weighted currents
 
|-
 
|-
 
| DF_SGN_SURF_NET_CURR
 
| DF_SGN_SURF_NET_CURR
−
| ''2G'' <math>\times</math> ''N''<sub>S</sub>
+
| ''2G'' &times; ''N<sub>M</sub>''
 
| Net sign moment weighted currents
 
| Net sign moment weighted currents
 
|-
 
|-
 
| DF_SGN_HET_SURF_FLUX
 
| DF_SGN_HET_SURF_FLUX
−
| ''2G'' <math>\times</math> ''N''<sub>S</sub>
+
| ''2G'' &times; ''N<sub>M</sub>''
 
| Heterogeneous sign moment weighted surface fluxes
 
| Heterogeneous sign moment weighted surface fluxes
 
|-
 
|-
 
| DF_SGN_HOM_SURF_FLUX
 
| DF_SGN_HOM_SURF_FLUX
−
| ''2G'' <math>\times</math> ''N''<sub>S</sub>
+
| ''2G'' &times; ''N<sub>M</sub>''
 
| Homogeneous sign moment weighted surface fluxes
 
| Homogeneous sign moment weighted surface fluxes
 
|-
 
|-
 
| DF_SGN_SURF_DF
 
| DF_SGN_SURF_DF
−
| ''2G'' <math>\times</math> ''N''<sub>S</sub>
+
| ''2G'' &times; ''N<sub>M</sub>''
 
| Sign moment weighted surface discontinuity factors
 
| Sign moment weighted surface discontinuity factors
 
|}
 
|}
Line 2,360: Line 2,920:
 
| PPW_PINS
 
| PPW_PINS
 
| 1
 
| 1
−
| Number of pin positions in the lattice (denoted as ''N''<sub>P</sub> below)
+
| Number of pin positions in the lattice (denoted as ''N<sub>P</sub>'' below)
 
|-
 
|-
 
| PPW_POW
 
| PPW_POW
−
| ''2G'' <math>\times</math> ''N''<sub>P</sub>
+
| ''2G'' &times; ''N<sub>P</sub>''
 
| Pin- and group-wise power distribution normalized to unity sum
 
| Pin- and group-wise power distribution normalized to unity sum
 
|-
 
|-
 
| PPW_HOM_FLUX
 
| PPW_HOM_FLUX
−
| ''2G'' <math>\times</math> ''N''<sub>P</sub>
+
| ''2G'' &times; ''N<sub>P</sub>''
 
| Pin- and group-wise homogeneous flux distribution
 
| Pin- and group-wise homogeneous flux distribution
 
|-
 
|-
 
| PPW_FF
 
| PPW_FF
−
| ''2G'' <math>\times</math> ''N''<sub>P</sub>
+
| ''2G'' &times; ''N<sub>P</sub>''
 
| Pin- and group-wise form factors
 
| Pin- and group-wise form factors
 
|-
 
|-
Line 2,385: Line 2,945:
 
*The order of ALB_TOT_ALB is <math>\alpha_{1,1} , \alpha_{1,2} , \ldots ,  \alpha_{2,1} ,  \alpha_{2,2} , \ldots</math> where <math>\alpha_{g,g'}</math> refers to albedo from group ''g'' to ''g'''.
 
*The order of ALB_TOT_ALB is <math>\alpha_{1,1} , \alpha_{1,2} , \ldots ,  \alpha_{2,1} ,  \alpha_{2,2} , \ldots</math> where <math>\alpha_{g,g'}</math> refers to albedo from group ''g'' to ''g'''.
 
*The order of ALB_PART_ALB is <math>\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 </math> where <math>\alpha_{k,g,k',g'}</math> refers to albedo of surface ''k<nowiki>'</nowiki>'' of group ''g<nowiki>'</nowiki>'' which has entered the albedo surface through surface ''k'' and group ''g''.
 
*The order of ALB_PART_ALB is <math>\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 </math> where <math>\alpha_{k,g,k',g'}</math> refers to albedo of surface ''k<nowiki>'</nowiki>'' of group ''g<nowiki>'</nowiki>'' 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) <math>= \alpha_{k,g,k',g'}</math> as
+
* 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'') <math>= \alpha_{k,g,k',g'}</math> as
 
  <nowiki>
 
  <nowiki>
 
part_alb = reshape(ALB_PART_ALB(1, 1:2:end), 2, 6, 2, 6)</nowiki>
 
part_alb = reshape(ALB_PART_ALB(1, 1:2:end), 2, 6, 2, 6)</nowiki>
Line 2,404: Line 2,964:
 
| ALB_N_SURF
 
| ALB_N_SURF
 
| 1
 
| 1
−
| Number of albedo surface faces (denoted as ''N''<sub>S</sub> below)
+
| Number of albedo surface faces (denoted as ''N<sub>S</sub>'' below)
 
|-
 
|-
 
| ALB_IN_CURR
 
| ALB_IN_CURR
−
| ''2G'' <math>\times</math> ''N''<sub>S</sub>
+
| ''2G'' &times; ''N<sub>S</sub>''
 
| Groupwise incoming partial currents of albedo surface faces
 
| Groupwise incoming partial currents of albedo surface faces
 
|-
 
|-
 
| ALB_OUT_CURR
 
| ALB_OUT_CURR
−
| ''2G''<sup>2</sup> <math>\times</math> ''N''<sub>S</sub><sup>2</sup>
+
| ''2G''<sup>2</sup> &times; ''N<sub>S</sub>''<sup>2</sup>
 
| Outgoing group to group and face to face outgoing partial currents  
 
| Outgoing group to group and face to face outgoing partial currents  
 
|-
 
|-
Line 2,419: Line 2,979:
 
|-
 
|-
 
| ALB_PART_ALB
 
| ALB_PART_ALB
−
| ''2G''<sup>2</sup> <math>\times</math> ''N''<sub>S</sub><sup>2</sup>
+
| ''2G''<sup>2</sup> &times; ''N<sub>S</sub>''<sup>2</sup>
 
| Partial group to group and face to face albedos  
 
| Partial group to group and face to face albedos  
 
|}
 
|}
  
 
== Miscellaneous notes for other outputs ==
 
== 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. <ref>
 
−
http://ttuki.vtt.fi/serpent/viewtopic.php?f=25&t=1885&p=4469</ref>
 
  
 
== References ==
 
== References ==

Latest revision as of 08:30, 8 October 2024

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 (criticality) or the set nps input option (external source)
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
BATCHES 1 Number of batches defined using the set nps 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 not implicit reaction rates are used for group constant generation
VR_ITER_IDX 1 Variance reduction iteration index when output was printed (see wwin card)

Domain decomposition

Parameter Size Description
DD_MODE 1 Domain decomposition mode defined using the set dd input option
DD_NEUTRONS_TO_LIMBO M Neutrons sent to limbo (transferral buffer) at each domain
DD_NEUTRONS_FROM_LIMBO M Neutrons received from limbo (transferral buffer) at each domain

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

Notes:

  • Only the first file path listed is displayed
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
PHOTON_PHYS_DIRECTORY (string) Photon physics directory path defined using the set pdatadir 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
MEAN_SRC_WGT 2/2 Mean source weight for non-criticality calculations (neutrons/photons)
SOURCE_SAMPLING_EFF 2/2 Source sampling efficiency for non-criticality calculations (neutrons/photons)
MEAN_SRC_WW_SPLIT 2/2 Mean source weight-window splitting in variance reduction (neutrons/photons)
MEAN_SRC_WW_EFF 2/2 Mean source weight-window sampling efficiency in variance reduction (neutrons/photons)
WW_BALA_ROULETE 2/2 Mean weight-window balance due to Russian roulette in variance reduction (neutrons/photons)
WW_BALA_SPLIT 2/2 Mean weight-window balance due to splitting in variance reduction (neutrons/photons)
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
IFC_COL_EFF 2/2 Efficiency of interface collision rejection (see ifc card)
REA_SAMPLING_EFF 2/2 Reaction sampling efficiency
REA_SAMPLING_FAIL 2/2 Fraction of failed reaction samples
TMS_SAMPLING_EFF 2 Target motion sampling method, TMS, sampling efficiency (see tms option, in mat card)
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
TMS_FAIL_STAT 8 TMS fail statistics: total samples, majorant fail, lower limit fail, upper limit fail (see tms option, in mat card)
DBRC_EXCEED_FRAC 1 Doppler-broadening rejection correction, DBRC, majorant exceed fraction (see set dbrc input option)
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

STL geometries

Parameter Size Description
STL_RAY_TEST 5 STL-ray-tracing test: total, ray is too parallel to facet, intersection point is too close to edge, facet is too close to search mesh cell boundary, two facets overlap
STL_ENFORCE_ST 1 Flag indicating whether or not delta tracking is enforced in STL geometries

Importance solver

Parameter Size Description
NEIGHBOUR_SEARCH_FAIL 2 Response matrix calculation fail rate

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(3) Wall-clock time spent for transport simulation
FINIX_SOLUTION_TIME 1 Wall-clock time spent for FINIX solution
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
DD_OVERHEAD_TIME 1 Wall-clock time spent for DD algorithm (see set dd input option)
RMX_SOLUTION_TIME 1 Wall-clock time spent for response matrix solution (see wwgen / wwin cards or set sca input option)
LEAKAGE_CORR_SOL_TIME 1 Wall-clock time spent for leakage correction solution (see set fum input option)
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 multi-physics interface data
RMX_MEMSIZE 1 Memory size used for storing response matrix-wise data
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 tolerance for unionized energy grid (see set egrid input option)
NEUTRON_ERG_NE 1 Number of points in neutron unionized energy grid
NEUTRON_EMIN 1 Minimum energy for neutron cross section data (see set egrid input option)
NEUTRON_EMAX 1 Maximum energy for neutron cross section data (see set egrid input option)

Photon energy grid

Parameter Size Description
PHOTON_ERG_NE 1 Number of points in photon unionized energy grid
PHOTON_EMIN 1 Minimum energy for photon cross section data (see set egrid input option)
PHOTON_EMAX 1 Maximum energy for photon cross section data (see set egrid input option)

Unresolved resonance probability table sampling

Parameter Size Description
URES_DILU_CUT 1 Density cut-off used for unresolved resonance probability table sampling (see set ures input option)
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 (see set ures input option)

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 nuclides 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)
IMPL_FISS_NUBAR 1 Flag indicating whether or not implicit fission nubar 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)

Photon physics options

Parameter Size Description
COMPTON_EKN 1 Photon energy above which Klein-Nishina is used for calculating energy and direction of the scattered photons (see set ekn input option)
COMPTON_DOPPLER 1 Flag indicating whether or not Doppler broadening method for the energy spectrum of the scattered photons is on (see set cdop input option)
COMPTON_EANG 1 Flag indicating whether or not Compton electron angular distribution model is on (see set cea input option)
PHOTON_TTB 1 Flag indicating whether or not thick-target bremsstrahlung approximation for modelling electrons and positrons is on (see set ttb input option)

Photon production

Parameter Size Description
PHOTON_SAMPLING_MODE 1 Flag indicating whether or not photon production from neutron reactions mode is on (see set ngamma input option)
PHOTON_SAMPLING_FAIL 2 Fraction of failed photon samples

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

Notes:

  • The values are given at the current burnup point (depletion step).
Parameter Size Description
TOT_ACTIVITY 1 Total activity
TOT_DECAY_HEAT 1 Total decay heat
TOT_SF_RATE 1 Total spontaneous fission rate
ACTINIDE_ACTIVITY 1 Actinide activity
ACTINIDE_DECAY_HEAT 1 Actinide decay heat
FISSION_PRODUCT_ACTIVITY 1 Fission product activity
FISSION_PRODUCT_DECAY_HEAT 1 Fission product decay heat
INHALATION_TOXICITY 1 Total inhalation toxicity
INGESTION_TOXICITY 1 Total ingestion toxicity
ACTINIDE_INH_TOX 1 Actinide inhalation toxicity
ACTINIDE_ING_TOX 1 Actinide ingestion toxicity
FISSION_PRODUCT_INH_TOX 1 Fission product inhalation toxicity
FISSION_PRODUCT_ING_TOX 1 Fission product ingestion toxicity
SR90_ACTIVITY 1 Sr-90 activity
TE132_ACTIVITY 1 Te-132 activity
I131_ACTIVITY 1 I-131 activity
I132_ACTIVITY 1 I-132 activity
CS134_ACTIVITY 1 Cs-134 activity
CS137_ACTIVITY 1 Cs-137 activity
PHOTON_DECAY_SOURCE 2 Total photon decay source rate and total released gamma decay
NEUTRON_DECAY_SOURCE 1 Total neutron decay source rate
ALPHA_DECAY_SOURCE 1 Total alpha decay source rate
ELECTRON_DECAY_SOURCE 1 Total beta decay source rate

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 calculated from the depletion history input and cumulative realized from the actual calculation.
BURN_DAYS 2 Number of burn days at the current step: cumulative and step-wise.
FIMA 3 Number of fissions per initial fissile atom at the current step: relative step-wise, increment step-wise, final step-wise.

Coefficient calculation

Parameter Size Description
COEF_IDX 2 Coefficient index when output is printed and total number of coefficient calculations
COEF_BRANCH 1 Branch index within coefficient calculation when output is printed
COEF_BU_STEP 1 Burnup step at the given coefficient calculation when output is printed

Analog reaction rate estimators

Parameter Size Description
CONVERSION_RATIO 2 Analog estimate of conversion rate, ratio between fissile production and loss rate
TH232_FISS 4 Analog estimate of Th-232 fission rate (total/fraction)
U233_FISS 4 Analog estimate of U-233 fission rate (total/fraction)
U235_FISS 4 Analog estimate of U-235 fission rate (total/fraction)
U238_FISS 4 Analog estimate of U-238 fission rate (total/fraction)
PU239_FISS 4 Analog estimate of Pu-239 fission rate (total/fraction)
PU240_FISS 4 Analog estimate of Pu-240 fission rate (total/fraction)
PU241_FISS 4 Analog estimate of Pu-241 fission rate (total/fraction)
TH232_CAPT 4 Analog estimate of Th-232 capture rate (total/fraction)
U233_CAPT 4 Analog estimate of U-233 capture rate (total/fraction)
U235_CAPT 4 Analog estimate of U-235 capture rate (total/fraction)
U238_CAPT 4 Analog estimate of U-238 capture rate (total/fraction)
PU239_CAPT 4 Analog estimate of Pu-239 capture rate (total/fraction)
PU240_CAPT 4 Analog estimate of Pu-240 capture rate (total/fraction)
PU241_CAPT 4 Analog estimate of Pu-241 capture rate (total/fraction)
XE135_CAPT 4 Analog estimate of Xe-135 capture rate (total/fraction)
XE135M_CAPT 4 Analog estimate of Xe-135m capture rate (total/fraction)
SM149_CAPT 4 Analog estimate of Sm-149 capture rate (total/fraction)

Particle balance

Neutron balance (particles/weight)

Parameter Size Description
BALA_SRC_NEUTRON_SRC 1/1 Neutron produced by external source
BALA_SRC_NEUTRON_FISS 1/1 Neutron produced by fission
BALA_SRC_NEUTRON_NXN 1/1 Neutron produced by scattering
BALA_SRC_NEUTRON_VR 1/1 Neutron produced by variance reduction (Russian roulette, splitting)
BALA_SRC_NEUTRON_TOT 1/1 Total neutron produced
BALA_LOSS_NEUTRON_CAPT 1/1 Neutron lost by capture
BALA_LOSS_NEUTRON_FISS 1/1 Neutron lost by fission
BALA_LOSS_NEUTRON_LEAK 1/1 Neutron lost by leakage
BALA_LOSS_NEUTRON_CUT 1/1 Neutron lost by cut-off
BALA_LOSS_NEUTRON_ERR 1/1 Neutron lost by failed tracking
BALA_LOSS_NEUTRON_TOT 1/1 Total neutron lost
BALA_NEUTRON_DIFF 1/1 Difference between total neutron produced and lost

Photon balance (particles/weight/energy-weighted)

Parameter Size Description
BALA_SRC_PHOTON_SRC 1/1/1 Photon produced by external source
BALA_SRC_PHOTON_TTB 1/1/1 Photon produced by bremsstrahlung
BALA_SRC_PHOTON_ANNIH 1/1/1 Photon produced by annihilation
BALA_SRC_PHOTON_FLUOR 1/1/1 Photon produced by fluorescence
BALA_SRC_PHOTON_NREA 1/1/1 Photon produced by neutron reaction
BALA_SRC_PHOTON_VR 1/1/1 Photon produced by variance reduction (Russian roulette, splitting)
BALA_SRC_PHOTON_TOT 1/1/1 Total photon produced
BALA_LOSS_PHOTON_CAPT 1/1 Photon lost by capture
BALA_LOSS_PHOTON_LEAK 1/1 Photon lost by leakage
BALA_LOSS_PHOTON_CUT 1/1 Photon lost by cut-off
BALA_LOSS_PHOTON_ERR 1/1 Photon lost by failed tracking
BALA_LOSS_PHOTON_TOT 1/1 Total photon lost
BALA_PHOTON_DIFF 1/1 Difference between total photon produced and lost

Integral results

Normalized total reaction rates (neutrons)

Notes:

  • In burnup calculations the values correspond to total, burnable and non-burnable rates
Parameter Size Description
TOT_POWER 2(6) Total neutron fission power
TOT_POWDENS 2(6) Total neutron fission power density
TOT_GENRATE 2(6) Total neutron generation rate
TOT_FISSRATE 2(6) Total neutron fission rate
TOT_CAPTRATE 2(6) Total neutron capture rate
TOT_ABSRATE 2(6) Total neutron absorption rate
TOT_SRCRATE 2(6) Total neutron source rate
TOT_FLUX 2(6) Total neutron flux
TOT_PHOTON_PRODRATE 4 Total neutron-photon production rate (implicit/analog)
TOT_LEAKRATE 2 Total neutron leakage rate
ALBEDO_LEAKRATE 2 Albedo neutron leakage rate
TOT_LOSSRATE 2 Total neutron loss rate
TOT_CUTRATE 2 Total neutron energy cut-off rate
TOT_RR 2 Total neutron reaction rate
TOT_XE135_ABSRATE 2 Total neutron absorption rate in Xe-135
TOT_XE135M_ABSRATE 2 Total neutron absorption rate in Xe-135m
TOT_SM149_ABSRATE 2 Total neutron absorption rate in Sm-149
INI_FMASS 1 Initial fissile mass
TOT_FMASS 1 Total fissile mass
INI_BURN_FMASS 1 Initial fissile mass in burnable materials
TOT_BURN_FMASS 1 Total fissile mass in burnable materials

Normalized total reaction rates (photons)

Parameter Size Description
TOT_PHOTON_LEAKRATE 2 Total photon leakage rate
TOT_PHOTON_CUTRATE 2 Total photon energy cut-off rate
PHOTOELE_CAPT_RATE 2 Photo-electric photon capture rate
PAIRPROD_CAPT_RATE 2 Pair production photon capture rate
TOT_PHOTON_LOSSRATE 2 Total photon loss rate
TOT_PHOTON_SRCRATE 2 Total photon source rate
TOT_PHOTON_RR 2 Total photon reaction rate
TOT_PHOTON_FLUX 2 Total photon flux
TOT_PHOTON_HEATRATE 2 Total photon heating rate

Equilibrium Xe-135 iteration

Notes:

  • The averages are calculated as volume averages.
  • Materials with the equilibrium calculation turned off are not taken into account in the average concentration but are included in the volume if they are fissile.
Parameter Size Description
I135_EQUIL_CONC 2 Average equilibrium I-135 concentration in materials with Xe-135 production rate (see set xenon input option)
XE135_EQUIL_CONC 2 Average equilibrium Xe-135 concentration in materials with Xe-135 production rate (see set xenon input option)
XE135M_EQUIL_CONC 2 Average equilibrium Xe-135m concentration in materials with Xe-135 production rate (see set xenon input option)

Equilibrium Sm-149 iteration

Notes:

  • The averages are calculated as volume averages.
  • Materials with the equilibrium calculation turned off are not taken into account in the average concentration but are included in the volume if they are fissile.
Parameter Size Description
SM149_EQUIL_CONC 2 Average equilibrium Sm-149 concentration in materials with Sm-149 production rate (see set samarium input option)
PM149_EQUIL_CONC 2 Average equilibrium Pm-149 concentration in materials with Sm-149 production rate (see set samarium input option)

Iteration factor

Parameter Size Description
ITER_FACTOR 2 Iteration factor of critical density iteration (see set iter nuc input option) or albedo iteration (see set iter alb input option)

Six-factor formula

Notes:

  • The six-factor formula estimates are printed by default. Check the suitability of the approximation for the given spectrum (the formulation is based on a thermal spectrum).
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 2 Average fission neutron yield
FISSE 2 Average fission energy production

Criticality eigenvalues / Multiplication factor external source

Parameter Size Description
ANA_KEFF 6(2) Analog estimate of keff: total, prompt and delayed neutron contribution.
IMP_KEFF 2 Implicit estimate of keff.
COL_KEFF 2 Collision estimate of keff.
ABS_KEFF 2 Absorption estimate of keff.
ABS_KINF 2 Absorption estimate of kinf.
ANA_EXT_K 20 Generation-wise source multiplication factors in external source mode
SRC_MULT 2 Source multiplication factor in external source mode
MEAN_NGEN 2 Mean number of generations in external source mode
PROMPT_GEN_POP NG Prompt fission population generation fraction
PROMPT_GEN_CUMU NG Prompt fission cumulative generation fraction
PROMPT_GEN_TIMES NG Prompt fission time generation fraction
PROMPT_CHAIN_LENGTH 2 Mean prompt chain length in external source mode
GEOM_ALBEDO 6 Fixed or iterated value for albedo boundary condition for x-,y- and z-directions (see set bc or set iter alb input options).

Wielandt method

Parameter Size Description
WIELANDT_K 2 Wielandt’s method shifted eigenvalue (see set wie input option)
WIELANDT_P 2 Wielandt’s method neutron banking probability (see set wie input option)

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

Forward 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, group-wise
FWD_ANA_LAMBDA 2D + 2 Analog estimator of delayed neutron precursor decay constants: total, group-wise

Adjoint-weighted time constants using Meulekamp's method

Parameter Size Description
ADJ_MEULEKAMP_BETA_EFF 2D + 2 Adjoint-weighted effective delayed neutron fractions using Meulekamp's method: total, group-wise
ADJ_MEULEKAMP_LAMBDA 2D + 2 Adjoint-weighted of delayed neutron precursor decay constants using Meulekamp's method: total, group-wise

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 Adjoint-weighted neutron generation times using Nauchi's method: total, prompt, delayed
ADJ_NAUCHI_LIFETIME 6 Adjoint-weighted neutron lifetimes using Nauchi's method: total, prompt, delayed.
ADJ_NAUCHI_BETA_EFF 2D + 2 Adjoint-weighted effective delayed neutron fractions using Nauchi's method: total, group-wise
ADJ_NAUCHI_LAMBDA 2D + 2 Adjoint-weighed of delayed neutron precursor decay constants using Nauchi's method: total, group-wise

Adjoint weighted time constants using IFP

Parameter Size Description
ADJ_IFP_GEN_TIME 6 Adjoint-weighted neutron generation times using the iterated fission probability method: total, prompt, delayed
ADJ_IFP_LIFETIME 6 Adjoint-weighted neutron lifetimes using the iterated fission probability method: total, prompt, delayed
ADJ_IFP_IMP_BETA_EFF 2D + 2 Implicit estimator of adjoint-weighted effective delayed neutron fractions using the iterated fission probability method: total, group-wise
ADJ_IFP_IMP_LAMBDA 2D + 2 Implicit estimator of adjoint-weighted of delayed neutron precursor decay constants using the iterated fission probability method: total, group-wise
ADJ_IFP_ANA_BETA_EFF 2D + 2 Analog estimator of adjoint-weighted effective delayed neutron fractions using the iterated fission probability method: total, group-wise
ADJ_IFP_ANA_LAMBDA 2D + 2 Analog estimator of adjoint-weighted of delayed neutron precursor decay constants using the iterated fission probability method: total, group-wise
ADJ_IFP_ROSSI_ALPHA 2 Adjoint-weighted Rossi alpha using the iterated fission probability method

Adjoint weighted time constants using perturbation technique

Parameter Size Description
ADJ_PERT_GEN_TIME 2 Adjoint-weighted neutron generation time using the perturbation technique
ADJ_PERT_LIFETIME 2 Adjoint-weighted neutron lifetime using the perturbation technique
ADJ_PERT_BETA_EFF 2 Adjoint-weighted effective delayed neutron fraction using the perturbation technique
ADJ_PERT_ROSSI_ALPHA 2 Adjoint-weighted Rossi alpha using the perturbation technique

Inverse neutron speed

Parameter Size Description
ANA_INV_SPD 2 Analog estimate of inverse neutron speed

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

Parameter Size Description
ANA_SLOW_TIME 6 Analog estimate of slowing-down time: total, prompt, delayed
ANA_THERM_TIME 6 Analog estimate of thermal neutron lifetime: total, prompt, delayed
ANA_THERM_FRAC 6 Analog estimate of neutron thermalisation fraction: total, prompt, 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

Dynamic simulation

Parameter Size Description
DYN_NB 1 Number of time intervals defined in the time-bin structure (see tme card)
DYN_TMIN 1 Minimum time boundary defined in the time-bin structure (see tme card)
DYN_TMAX 1 Maximum time boundary defined in the time-bin structure (see tme card)
DYN_TIMES 2T + 2 Time-bin boundaries defined in the time-bin structure (see tme card)
DYN_POP 2T Neutron population at the end of the time interval
DYN_PERIOD 2T Reactor period based on increase/decrease of neutron population during time interval

Analog mean photon lifetime

Parameter Size Description
ANA_LIFETIME 2 Analog estimator of photon lifetime

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 × 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 (CMM) is described in related papers[2][3].
  • 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 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).
  • Xe-135m values printed only if separate treatment of Xe-135m is on (see set poi).
  • Approximate effective treatment of Pm-149 production from Pm-148 and Pm-148m is not included in Pm-149 fission yield.
  • All values for I-135, Pm-147, Pm-148 and Pm-148m not present in all versions.
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_XE135M_YIELD 2G Fission yield of Xe-135m
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_XE135M_MICRO_ABS 2G Microscopic absorption cross section of Xe-135m
INF_PM147_MICRO_ABS 2G Microscopic absorption cross section of Pm-147
INF_PM148_MICRO_ABS 2G Microscopic absorption cross section of Pm-148
INF_PM148M_MICRO_ABS 2G Microscopic absorption cross section of Pm-148m
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_I135_MACRO_ABS 2G Macroscopic absorption cross section of I-135
INF_XE135_MACRO_ABS 2G Macroscopic absorption cross section of Xe-135
INF_XE135M_MACRO_ABS 2G Macroscopic absorption cross section of Xe-135m
INF_PM147_MACRO_ABS 2G Macroscopic absorption cross section of Pm-147
INF_PM148_MACRO_ABS 2G Macroscopic absorption cross section of Pm-148
INF_PM148M_MACRO_ABS 2G Macroscopic absorption cross section of Pm-148M
INF_PM149_MACRO_ABS 2G Macroscopic absorption cross section of Pm-149
INF_SM149_MACRO_ABS 2G Macroscopic absorption cross section of Sm-149

Poison universe-averaged densities

Notes:

  • The universe-averaged atomic density is defined to be such that ADENS*MICRO_ABS is equal to MACRO_ABS.[4]
Parameter Size Description
I135_ADENS 2 Universe-averaged atomic density of I-135
XE135_ADENS 2 Universe-averaged atomic density of Xe-135
XE135M_ADENS 2 Universe-averaged atomic density of Xe-135m
PM147_ADENS 2 Universe-averaged atomic density of Pm-147
PM148_ADENS 2 Universe-averaged atomic density of Pm-147
PM148M_ADENS 2 Universe-averaged atomic density of Pm-148m
PM149_ADENS 2 Universe-averaged atomic density of Pm-149

Poison decay constants

Parameter Size Description
PM147_LAMBDA 1 Decay constant of Pm-147
PM148_LAMBDA 1 Decay constant of Pm-147
PM148M_LAMBDA 1 Decay constant of Pm-148m
PM149_LAMBDA 1 Decay constant of Pm-149
I135_LAMBDA 1 Decay constant of I-135
XE135_LAMBDA 1 Decay constant of Xe-135
XE135M_LAMBDA 1 Decay constant of Xe-135m
I135_BR 1 Branching ratio of I-135 decay to Xe-135. Branching ratio of I-135 decay to Xe-135m is (1 - I135_BR).

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 [input]_res.m file can be read into a G × 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).
  • Xe-135m values printed only if separate treatment of Xe-135m 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_XE135M_YIELD 2G Fission yield of Xe-135m
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_XE135M_MICRO_ABS 2G Microscopic absorption cross section of Xe-135m
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_XE135M_MACRO_ABS 2G Macroscopic absorption cross section of Xe-135m
B1_SM149_MACRO_ABS 2G Macroscopic absorption cross section of Sm-149

Delayed neutron data

Notes:

  • The output consists of total, followed by D precursor group-wise 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. 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 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. 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_SURFACE_TYPE 1 Surface type
DF_SURFACE_N_PARAMS 1 Number of parameters defining the surface
DF_SURFACE_PARAMS NP List of parameters defining the surface
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_N_SGN 1 Number of sign moment values (denoted as NM 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 × NS Inward surface currents
DF_SURF_OUT_CURR 2G × NS Outward surface currents
DF_SURF_NET_CURR 2G × NS Net surface currents
DF_MID_IN_CURR 2G × NS Inward mid-point currents
DF_MID_OUT_CURR 2G × NS Outward mid-point currents
DF_MID_NET_CURR 2G × NS Net mid-point currents
DF_CORN_IN_CURR 2G × NC Inward corner currents
DF_CORN_OUT_CURR 2G × NC Outward corner currents
DF_CORN_NET_CURR 2G × NC Net corner currents
DF_HET_VOL_FLUX 2G Heterogeneous flux over homogenized cell
DF_HET_SURF_FLUX 2G × NS Heterogeneous surface fluxes
DF_HET_CORN_FLUX 2G × NC Heterogeneous corner fluxes
DF_HOM_VOL_FLUX 2G Homogeneous flux over homogenized cell
DF_HOM_SURF_FLUX 2G × NS Homogeneous surface fluxes
DF_HOM_CORN_FLUX 2G × NC Homogeneous corner fluxes
DF_SURF_DF 2G × NS Surface discontinuity factors
DF_CORN_DF 2G × NC Corner discontinuity factors
DF_SGN_SURF_IN_CURR 2G × NM Inward sign moment weighted currents
DF_SGN_SURF_OUT_CURR 2G × NM Outward sign moment weighted currents
DF_SGN_SURF_NET_CURR 2G × NM Net sign moment weighted currents
DF_SGN_HET_SURF_FLUX 2G × NM Heterogeneous sign moment weighted surface fluxes
DF_SGN_HOM_SURF_FLUX 2G × NM Homogeneous sign moment weighted surface fluxes
DF_SGN_SURF_DF 2G × NM 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 × NP Pin- and group-wise power distribution normalized to unity sum
PPW_HOM_FLUX 2G × NP Pin- and group-wise homogeneous flux distribution
PPW_FF 2G × 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 × NS Groupwise incoming partial currents of albedo surface faces
ALB_OUT_CURR 2G2 × 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 × NS2 Partial group to group and face to face albedos

Miscellaneous notes for other outputs

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. Nuc. Energy, 112 (2016) 126-136
  3. ^ Liu, Z., Smith, K. and Forget, B. "Group-wise Tally Scheme of Incremental Migration Area for Cumulative Migration Method." In Proceedings of the PHYSOR 2018 (2018) 2512-2523
  4. ^ Rintala, A., Valtavirta, V. and Leppänen, J.. Microscopic cross section calculation methodology in the Serpent 2 Monte Carlo code. Annals of Nuclear Energy, 164 (2021): 108603.
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