Difference between revisions of "Output parameters"
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Ana Jambrina (talk | contribs) m (Reverted edits by Ana Jambrina (talk) to last revision by Antti Rintala) |
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| 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 |
|- | |- | ||
| CYCLES | | CYCLES | ||
| 1 | | 1 | ||
| Number of active cycles defined using the [[Input syntax manual#set pop|set pop]] input option | | Number of active cycles defined using the [[Input syntax manual#set pop|set pop]] input option | ||
| + | | | ||
|- | |- | ||
| SKIP | | SKIP | ||
| 1 | | 1 | ||
| Number of inactive cycles defined using the [[Input syntax manual#set pop|set pop]] input option | | Number of inactive cycles defined using the [[Input syntax manual#set pop|set pop]] input option | ||
| + | | | ||
|- | |- | ||
| BATCH_INTERVAL | | BATCH_INTERVAL | ||
| 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 | ||
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| SRC_NORM_MODE | | SRC_NORM_MODE | ||
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| 1 | | 1 | ||
| Flag indicating whether or implicit reaction rates are used for group constant generation | | Flag indicating whether or implicit reaction rates are used for group constant generation | ||
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| OMP_HISTORY_PROFILE | | OMP_HISTORY_PROFILE | ||
| − | | | + | | N |
| Fraction of particle histories run for each parallel OpenMP thread | | Fraction of particle histories run for each parallel OpenMP thread | ||
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| Line 232: | Line 205: | ||
=== File paths === | === File paths === | ||
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{|class="wikitable" style="text-align: left;" | {|class="wikitable" style="text-align: left;" | ||
| Line 260: | Line 230: | ||
| (string) | | (string) | ||
| Isomeric branching ratio data file path defined using the [[Input syntax manual#set bralib|set bralib]] input option | | Isomeric branching ratio data file path defined using the [[Input syntax manual#set bralib|set bralib]] input option | ||
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|} | |} | ||
| Line 278: | Line 244: | ||
! Size | ! Size | ||
! Description | ! Description | ||
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| MIN_MACROXS | | MIN_MACROXS | ||
| Line 322: | Line 264: | ||
| 2/2 | | 2/2 | ||
| Delta-tracking efficiency | | Delta-tracking efficiency | ||
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|- | |- | ||
| REA_SAMPLING_EFF | | REA_SAMPLING_EFF | ||
| Line 334: | Line 272: | ||
| 2/2 | | 2/2 | ||
| Fraction of failed reaction samples | | Fraction of failed reaction samples | ||
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|- | |- | ||
| TOT_COL_EFF | | TOT_COL_EFF | ||
| Line 346: | Line 280: | ||
| 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 | ||
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|- | |- | ||
| AVG_TRACKS | | AVG_TRACKS | ||
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| 1 | | 1 | ||
| Number of lost particles | | Number of lost particles | ||
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|} | |} | ||
| Line 464: | Line 360: | ||
|- | |- | ||
| TRANSPORT_CYCLE_TIME | | TRANSPORT_CYCLE_TIME | ||
| − | | 1( | + | | 1(2) |
| Wall-clock time spent for transport simulation | | Wall-clock time spent for transport simulation | ||
| − | |||
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|- | |- | ||
| BURNUP_CYCLE_TIME | | BURNUP_CYCLE_TIME | ||
| Line 483: | Line 375: | ||
| Wall-clock time spent MPI communication | | Wall-clock time spent MPI communication | ||
|- | |- | ||
| − | + | | ESTIMATED_RUNNING_TIME | |
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| − | | ESTIMATED_RUNNING_TIME | ||
| 1(2) | | 1(2) | ||
| Estimated total wall-clock running time | | Estimated total wall-clock running time | ||
| Line 550: | Line 430: | ||
| IFC_MEMSIZE | | IFC_MEMSIZE | ||
| 1 | | 1 | ||
| − | | Memory size used for data for | + | | Memory size used for data for response-matrix solver |
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|- | |- | ||
| MISC_MEMSIZE | | MISC_MEMSIZE | ||
| Line 600: | Line 476: | ||
| NEUTRON_ERG_NE | | NEUTRON_ERG_NE | ||
| 1 | | 1 | ||
| − | | Number of points in | + | | Number of points in unionized energy grid |
|- | |- | ||
| NEUTRON_EMIN | | NEUTRON_EMIN | ||
| Line 609: | Line 485: | ||
| 1 | | 1 | ||
| Maximum energy for neutron cross section data | | Maximum energy for neutron cross section data | ||
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|} | |} | ||
| Line 679: | Line 534: | ||
| TOT_DOSIMETRY_NUCLIDES | | TOT_DOSIMETRY_NUCLIDES | ||
| 1 | | 1 | ||
| − | | Total number of | + | | Total number of nclides with dosimetry cross sections |
|- | |- | ||
| TOT_DECAY_NUCLIDES | | TOT_DECAY_NUCLIDES | ||
| Line 698: | Line 553: | ||
|- | |- | ||
|} | |} | ||
| − | |||
== Physics == | == Physics == | ||
| Line 731: | Line 585: | ||
| 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) | ||
| − | |||
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|- | |- | ||
| DOPPLER_PREPROCESSOR | | DOPPLER_PREPROCESSOR | ||
| Line 758: | Line 608: | ||
|} | |} | ||
| − | === | + | === 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;" | ||
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! Description | ! Description | ||
|- | |- | ||
| − | + | | EDEP_MODE | |
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| − | | EDEP_MODE | ||
| 1 | | 1 | ||
| Energy deposition mode (see [[Input syntax manual#set edepmode|set edepmode]] input option) | | Energy deposition mode (see [[Input syntax manual#set edepmode|set edepmode]] input option) | ||
| Line 837: | Line 645: | ||
=== Radioactivity data === | === Radioactivity data === | ||
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{|class="wikitable" style="text-align: left;" | {|class="wikitable" style="text-align: left;" | ||
| Line 847: | Line 652: | ||
|- | |- | ||
| TOT_ACTIVITY | | TOT_ACTIVITY | ||
| − | | | + | | |
| − | | | + | | |
|- | |- | ||
| TOT_DECAY_HEAT | | TOT_DECAY_HEAT | ||
| − | | | + | | |
| − | | | + | | |
|- | |- | ||
| TOT_SF_RATE | | TOT_SF_RATE | ||
| − | | | + | | |
| − | | | + | | |
|- | |- | ||
| ACTINIDE_ACTIVITY | | ACTINIDE_ACTIVITY | ||
| − | | | + | | |
| − | | | + | | |
|- | |- | ||
| ACTINIDE_DECAY_HEAT | | ACTINIDE_DECAY_HEAT | ||
| − | | | + | | |
| − | | | + | | |
|- | |- | ||
| FISSION_PRODUCT_ACTIVITY | | FISSION_PRODUCT_ACTIVITY | ||
| − | | | + | | |
| − | | | + | | |
|- | |- | ||
| FISSION_PRODUCT_DECAY_HEAT | | FISSION_PRODUCT_DECAY_HEAT | ||
| − | | | + | | |
| − | | | + | | |
|- | |- | ||
| INHALATION_TOXICITY | | INHALATION_TOXICITY | ||
| − | | | + | | |
| − | | | + | | |
|- | |- | ||
| INGESTION_TOXICITY | | INGESTION_TOXICITY | ||
| − | | | + | | |
| − | | | + | | |
|- | |- | ||
| ACTINIDE_INH_TOX | | ACTINIDE_INH_TOX | ||
| − | | | + | | |
| − | | | + | | |
|- | |- | ||
| ACTINIDE_ING_TOX | | ACTINIDE_ING_TOX | ||
| − | | | + | | |
| − | | | + | | |
|- | |- | ||
| FISSION_PRODUCT_INH_TOX | | FISSION_PRODUCT_INH_TOX | ||
| − | | | + | | |
| − | | | + | | |
|- | |- | ||
| FISSION_PRODUCT_ING_TOX | | FISSION_PRODUCT_ING_TOX | ||
| − | | | + | | |
| − | | | + | | |
|- | |- | ||
| SR90_ACTIVITY | | SR90_ACTIVITY | ||
| − | | | + | | |
| − | | | + | | |
|- | |- | ||
| TE132_ACTIVITY | | TE132_ACTIVITY | ||
| − | | | + | | |
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|- | |- | ||
| I131_ACTIVITY | | I131_ACTIVITY | ||
| − | | | + | | |
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|- | |- | ||
| I132_ACTIVITY | | I132_ACTIVITY | ||
| − | | | + | | |
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|- | |- | ||
| CS134_ACTIVITY | | CS134_ACTIVITY | ||
| − | | | + | | |
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|- | |- | ||
| CS137_ACTIVITY | | CS137_ACTIVITY | ||
| − | | | + | | |
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|- | |- | ||
| PHOTON_DECAY_SOURCE | | PHOTON_DECAY_SOURCE | ||
| − | | | + | | |
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|- | |- | ||
| NEUTRON_DECAY_SOURCE | | NEUTRON_DECAY_SOURCE | ||
| − | | | + | | |
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|- | |- | ||
| ALPHA_DECAY_SOURCE | | ALPHA_DECAY_SOURCE | ||
| − | | | + | | |
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|- | |- | ||
| ELECTRON_DECAY_SOURCE | | ELECTRON_DECAY_SOURCE | ||
| − | | | + | | |
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|- | |- | ||
|} | |} | ||
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|} | |} | ||
| − | === | + | === Analog reaction rate estimators === |
{|class="wikitable" style="text-align: left;" | {|class="wikitable" style="text-align: left;" | ||
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! Description | ! Description | ||
|- | |- | ||
| − | | | + | | CONVERSION_RATIO |
| − | | | + | | |
| − | | | + | | |
|- | |- | ||
| − | | | + | | U235_FISS |
| − | | | + | | |
| − | | | + | | |
|- | |- | ||
| − | + | | U238_FISS | |
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| − | | U238_FISS | ||
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| U235_CAPT | | U235_CAPT | ||
| − | | | + | | |
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|- | |- | ||
| U238_CAPT | | U238_CAPT | ||
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| XE135_CAPT | | XE135_CAPT | ||
| − | | | + | | |
| − | | | + | | |
|- | |- | ||
|} | |} | ||
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|- | |- | ||
| BALA_SRC_NEUTRON_SRC | | BALA_SRC_NEUTRON_SRC | ||
| − | | | + | | |
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|- | |- | ||
| BALA_SRC_NEUTRON_FISS | | BALA_SRC_NEUTRON_FISS | ||
| − | | | + | | |
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|- | |- | ||
| BALA_SRC_NEUTRON_NXN | | BALA_SRC_NEUTRON_NXN | ||
| − | | | + | | |
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|- | |- | ||
| BALA_SRC_NEUTRON_VR | | BALA_SRC_NEUTRON_VR | ||
| − | | | + | | |
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|- | |- | ||
| BALA_SRC_NEUTRON_TOT | | BALA_SRC_NEUTRON_TOT | ||
| − | | | + | | |
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|- | |- | ||
| BALA_LOSS_NEUTRON_CAPT | | BALA_LOSS_NEUTRON_CAPT | ||
| − | | | + | | |
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|- | |- | ||
| BALA_LOSS_NEUTRON_FISS | | BALA_LOSS_NEUTRON_FISS | ||
| − | | | + | | |
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|- | |- | ||
| BALA_LOSS_NEUTRON_LEAK | | BALA_LOSS_NEUTRON_LEAK | ||
| − | | | + | | |
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|- | |- | ||
| BALA_LOSS_NEUTRON_CUT | | BALA_LOSS_NEUTRON_CUT | ||
| − | | | + | | |
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|- | |- | ||
| BALA_LOSS_NEUTRON_ERR | | BALA_LOSS_NEUTRON_ERR | ||
| − | | | + | | |
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|- | |- | ||
| BALA_LOSS_NEUTRON_TOT | | BALA_LOSS_NEUTRON_TOT | ||
| − | | | + | | |
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|- | |- | ||
| BALA_NEUTRON_DIFF | | BALA_NEUTRON_DIFF | ||
| − | | | + | | |
| − | | | + | | |
|- | |- | ||
|} | |} | ||
| − | === | + | == Integral results == |
| + | |||
| + | === Normalized total reaction rates (neutrons) === | ||
{|class="wikitable" style="text-align: left;" | {|class="wikitable" style="text-align: left;" | ||
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! Description | ! Description | ||
|- | |- | ||
| − | | | + | | TOT_POWER |
| − | | | + | | |
| − | | | + | | |
| + | |- | ||
| + | | TOT_POWDENS | ||
| + | | | ||
| + | | | ||
|- | |- | ||
| − | | | + | | TOT_GENRATE |
| − | | | + | | |
| − | | | + | | |
|- | |- | ||
| − | | | + | | TOT_FISSRATE |
| − | | | + | | |
| − | | | + | | |
|- | |- | ||
| − | | | + | | TOT_CAPTRATE |
| − | | | + | | |
| − | | | + | | |
|- | |- | ||
| − | | | + | | TOT_ABSRATE |
| − | | | + | | |
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|- | |- | ||
| − | | | + | | TOT_SRCRATE |
| − | | | + | | |
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|- | |- | ||
| − | | | + | | TOT_FLUX |
| − | | | + | | |
| − | | | + | | |
|- | |- | ||
| − | | | + | | TOT_PHOTON_PRODRATE |
| − | | | + | | |
| − | | | + | | |
|- | |- | ||
| − | | | + | | TOT_LEAKRATE |
| − | | | + | | |
| − | | | + | | |
|- | |- | ||
| − | | | + | | ALBEDO_LEAKRATE |
| − | | | + | | |
| − | | | + | | |
|- | |- | ||
| − | | | + | | TOT_LOSSRATE |
| − | | | + | | |
| − | | | + | | |
|- | |- | ||
| − | | | + | | TOT_CUTRATE |
| − | | | + | | |
| − | | | + | | |
|- | |- | ||
| − | | | + | | TOT_RR |
| − | | | + | | |
| − | | | + | | |
|- | |- | ||
| − | | | + | | TOT_XE135_ABSRATE |
| − | + | | | |
| − | + | | | |
| + | |- | ||
| + | | INI_FMASS | ||
| + | | | ||
| + | | | ||
| + | |- | ||
| + | | TOT_FMASS | ||
| + | | | ||
| + | | | ||
| + | |- | ||
| + | | INI_BURN_FMASS | ||
| + | | | ||
| + | | | ||
| + | |- | ||
| + | | TOT_BURN_FMASS | ||
| + | | | ||
| + | | | ||
| + | |- | ||
| + | |} | ||
| − | === | + | === Equilibrium Xe-135 iteration === |
| − | |||
| − | |||
| − | |||
{|class="wikitable" style="text-align: left;" | {|class="wikitable" style="text-align: left;" | ||
| Line 1,216: | Line 981: | ||
! Description | ! Description | ||
|- | |- | ||
| − | | | + | | XE135_EQUIL_CONC |
| − | | 2( | + | | 2 |
| − | + | | Averaged equilibrium Xe-135 concentration (see [[Input syntax manual#set xenon|set xenon]] input option) | |
|- | |- | ||
| − | | | + | | I135_EQUIL_CONC |
| − | | 2( | + | | 2 |
| − | + | | Averaged equilibrium I-135 concentration (see [[Input syntax manual#set xenon|set xenon]] input option) | |
|- | |- | ||
| − | | | + | |} |
| − | + | ||
| − | + | === Equilibrium Sm-149 iteration === | |
| − | + | ||
| − | + | {|class="wikitable" style="text-align: left;" | |
| − | + | ! Parameter | |
| − | + | ! Size | |
| − | |- | + | ! Description |
| − | |||
| − | |||
| − | |||
|- | |- | ||
| − | | | + | | SM149_EQUIL_CONC |
| − | | 2( | + | | 2 |
| − | + | | Averaged equilibrium Sm-149 concentration (see [[Input syntax manual#set samarium|set samarium]] input option) | |
|- | |- | ||
| − | | | + | | PM149_EQUIL_CONC |
| − | | 2( | + | | 2 |
| − | + | | Averaged equilibrium Pm-149 concentration (see [[Input syntax manual#set samarium|set samarium]] input option) | |
|- | |- | ||
| − | | | + | |} |
| − | | | + | |
| − | + | === Six-factor formula === | |
| + | |||
| + | {|class="wikitable" style="text-align: left;" | ||
| + | ! 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 | | 2 | ||
| − | | | + | | Analog estimate of thermal utilization factor |
|- | |- | ||
| − | | | + | | SIX_FF_P |
| 2 | | 2 | ||
| − | | | + | | Analog estimate of resonance escape probability |
|- | |- | ||
| − | | | + | | SIX_FF_EPSILON |
| 2 | | 2 | ||
| − | | | + | | Analog estimate of fast fission factor |
|- | |- | ||
| − | | | + | | SIX_FF_LF |
| 2 | | 2 | ||
| − | | | + | | Analog estimate of fast non-leakage probability |
|- | |- | ||
| − | | | + | | SIX_FF_LT |
| 2 | | 2 | ||
| − | | | + | | Analog estimate of thermal non-leakage probability |
|- | |- | ||
| − | | | + | | SIX_FF_KINF |
| 2 | | 2 | ||
| − | | | + | | Analog estimate of six-factor ''k''<sub>inf</sub> (four-factor ''k''<sub>eff</sub>) |
|- | |- | ||
| − | | | + | | SIX_FF_KEFF |
| 2 | | 2 | ||
| − | | | + | | Analog estimate of six-factor ''k''<sub>eff</sub> |
| − | |||
| − | |||
| − | |||
| − | |||
| − | |||
| − | |||
| − | |||
| − | |||
| − | |||
| − | |||
| − | |||
| − | |||
| − | |||
| − | |||
| − | |||
| − | |||
|- | |- | ||
|} | |} | ||
| − | === | + | === Fission neutron and energy production === |
{|class="wikitable" style="text-align: left;" | {|class="wikitable" style="text-align: left;" | ||
| Line 1,305: | Line 1,056: | ||
! Description | ! Description | ||
|- | |- | ||
| − | | | + | | NUBAR |
| − | | | + | | |
| − | | | + | | |
|- | |- | ||
| − | | | + | | FISSE |
| − | | | + | | |
| − | | | + | | |
|- | |- | ||
| − | | | + | |} |
| − | + | ||
| − | | | + | === Criticality eigenvalues === |
| + | |||
| + | {|class="wikitable" style="text-align: left;" | ||
| + | ! Parameter | ||
| + | ! Size | ||
| + | ! Description | ||
|- | |- | ||
| − | | | + | | ANA_KEFF |
| − | | | + | | |
| − | | | + | | |
|- | |- | ||
| − | | | + | | IMP_KEFF |
| − | | | + | | |
| − | | | + | | |
|- | |- | ||
| − | | | + | | COL_KEFF |
| − | | | + | | |
| − | | | + | | |
|- | |- | ||
| − | | | + | | ABS_KEFF |
| − | | | + | | |
| − | | | + | | |
|- | |- | ||
| − | | | + | | ABS_KINF |
| − | | | + | | |
| − | | | + | | |
|- | |- | ||
| − | | | + | | GEOM_ALBEDO |
| − | | | + | | |
| − | | | + | | |
|- | |- | ||
|} | |} | ||
| − | === | + | === ALF (Average lethargy of neutrons causing fission) === |
{|class="wikitable" style="text-align: left;" | {|class="wikitable" style="text-align: left;" | ||
| Line 1,350: | Line 1,106: | ||
! Description | ! Description | ||
|- | |- | ||
| − | | | + | | ANA_ALF |
| − | | | + | | |
| − | | | + | | |
|- | |- | ||
| − | | | + | | IMP_ALF |
| − | | | + | | |
| − | | | + | | |
|- | |- | ||
|} | |} | ||
| − | === | + | === EALF (Energy corresponding to average lethargy of neutrons causing fission) === |
{|class="wikitable" style="text-align: left;" | {|class="wikitable" style="text-align: left;" | ||
| Line 1,367: | Line 1,123: | ||
! Description | ! Description | ||
|- | |- | ||
| − | | | + | | ANA_EALF |
| − | | | + | | |
| − | | | + | | |
|- | |- | ||
| − | | | + | | IMP_EALF |
| − | | | + | | |
| − | | | + | | |
|- | |- | ||
|} | |} | ||
| − | === | + | == Time constants == |
| + | |||
| + | === Forward-weighted delayed neutron parameters === | ||
{|class="wikitable" style="text-align: left;" | {|class="wikitable" style="text-align: left;" | ||
| Line 1,384: | Line 1,142: | ||
! Description | ! Description | ||
|- | |- | ||
| − | | | + | | PRECURSOR_GROUPS |
| − | | | + | | |
| − | | | + | | |
| + | |- | ||
| + | | FWD_ANA_BETA_ZERO | ||
| + | | | ||
| + | | | ||
| + | |- | ||
| + | | FWD_ANA_LAMBDA | ||
| + | | | ||
| + | | | ||
|- | |- | ||
|} | |} | ||
| − | === | + | === Beta-eff using Meulekamp's method === |
{|class="wikitable" style="text-align: left;" | {|class="wikitable" style="text-align: left;" | ||
| Line 1,397: | Line 1,163: | ||
! Description | ! Description | ||
|- | |- | ||
| − | | | + | | ADJ_MEULEKAMP_BETA_EFF |
| − | | | + | | |
| − | | | + | | |
|- | |- | ||
| − | | | + | | ADJ_MEULEKAMP_LAMBDA |
| − | | | + | | |
| − | | | + | | |
|- | |- | ||
| − | | | + | |} |
| − | | | + | |
| − | + | === Adjoint weighted time constants using Nauchi's method === | |
| + | |||
| + | {|class="wikitable" style="text-align: left;" | ||
| + | ! Parameter | ||
| + | ! Size | ||
| + | ! Description | ||
|- | |- | ||
| − | | | + | | IFP_CHAIN_LENGTH |
| − | | | + | | |
| − | | | + | | |
|- | |- | ||
| − | | | + | | ADJ_NAUCHI_GEN_TIME |
| − | | | + | | |
| − | | | + | | |
|- | |- | ||
| − | | | + | | ADJ_NAUCHI_LIFETIME |
| − | | | + | | |
| − | | | + | | |
|- | |- | ||
| − | | | + | | ADJ_NAUCHI_BETA_EFF |
| − | | | + | | |
| − | | | + | | |
|- | |- | ||
| − | | | + | | ADJ_NAUCHI_LAMBDA |
| − | | | + | | |
| − | | | + | | |
|- | |- | ||
|} | |} | ||
| − | === | + | === Adjoint weighted time constants using IFP === |
{|class="wikitable" style="text-align: left;" | {|class="wikitable" style="text-align: left;" | ||
| Line 1,438: | Line 1,209: | ||
! Description | ! Description | ||
|- | |- | ||
| − | | | + | | ADJ_IFP_GEN_TIME |
| − | | | + | | |
| − | | | + | | |
|- | |- | ||
| − | | | + | | ADJ_IFP_LIFETIME |
| − | | | + | | |
| − | | | + | | |
| + | |- | ||
| + | | ADJ_IFP_IMP_BETA_EFF | ||
| + | | | ||
| + | | | ||
| + | |- | ||
| + | | ADJ_IFP_IMP_LAMBDA | ||
| + | | | ||
| + | | | ||
| + | |- | ||
| + | | ADJ_IFP_ANA_BETA_EFF | ||
| + | | | ||
| + | | | ||
| + | |- | ||
| + | | ADJ_IFP_ANA_LAMBDA | ||
| + | | | ||
| + | | | ||
| + | |- | ||
| + | | ADJ_IFP_ROSSI_ALPHA | ||
| + | | | ||
| + | | | ||
|- | |- | ||
|} | |} | ||
| − | === | + | === Adjoint weighted time constants using perturbation technique === |
{|class="wikitable" style="text-align: left;" | {|class="wikitable" style="text-align: left;" | ||
| Line 1,455: | Line 1,246: | ||
! Description | ! Description | ||
|- | |- | ||
| − | | | + | | ADJ_PERT_GEN_TIME |
| − | | | + | | |
| − | | | + | | |
|- | |- | ||
| − | | | + | | ADJ_PERT_LIFETIME |
| − | | | + | | |
| − | | | + | | |
|- | |- | ||
| − | | | + | | ADJ_PERT_BETA_EFF |
| − | | | + | | |
| − | | | + | | |
|- | |- | ||
| − | | | + | | ADJ_PERT_ROSSI_ALPHA |
| − | | | + | | |
| − | | | + | | |
|- | |- | ||
| − | | | + | |} |
| − | + | ||
| − | + | === Inverse neutron speed === | |
| − | + | ||
| − | + | {|class="wikitable" style="text-align: left;" | |
| − | + | ! Parameter | |
| − | + | ! Size | |
| − | |- | + | ! Description |
| − | |||
| − | |||
| − | |||
|- | |- | ||
| − | | | + | | ANA_INV_SPD |
| − | + | | | |
| − | + | | | |
| − | |||
| − | |||
| − | |||
| − | |||
| − | |||
| − | |||
| − | |||
| − | |||
| − | |||
| − | |||
| − | |||
| − | |||
| − | |||
| − | |||
| − | |||
| − | |||
| − | |||
| − | |||
| − | | | ||
| − | | | ||
|- | |- | ||
|} | |} | ||
| − | === | + | === Analog slowing-down and thermal neutron lifetime (total/prompt/delayed) === |
{|class="wikitable" style="text-align: left;" | {|class="wikitable" style="text-align: left;" | ||
| Line 1,516: | Line 1,284: | ||
! Description | ! Description | ||
|- | |- | ||
| − | | | + | | ANA_SLOW_TIME |
| − | | | + | | |
| − | | | + | | |
| + | |- | ||
| + | | ANA_THERM_TIME | ||
| + | | | ||
| + | | | ||
| + | |- | ||
| + | | ANA_THERM_FRAC | ||
| + | | | ||
| + | | | ||
| + | |- | ||
| + | | ANA_DELAYED_EMTIME | ||
| + | | | ||
| + | | | ||
|- | |- | ||
| − | | | + | | ANA_MEAN_NCOL |
| − | | | + | | |
| − | | | + | | |
|- | |- | ||
|} | |} | ||
| − | == | + | == Homogenized group constants == |
| − | + | <u>Notes:</u> | |
| − | + | ||
| − | + | *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 1,549: | Line 1,326: | ||
! Size | ! Size | ||
! Description | ! 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 === |
{|class="wikitable" style="text-align: left;" | {|class="wikitable" style="text-align: left;" | ||
| Line 1,566: | Line 1,355: | ||
! Size | ! Size | ||
! Description | ! Description | ||
| + | |- | ||
| + | | INF_MICRO_FLX | ||
| + | | 2''H'' | ||
| + | | Multi-group flux spectrum (integral, un-normalized) | ||
|- | |- | ||
| − | | | + | | INF_FLX |
| − | | 2 | + | | 2''G'' |
| − | | | + | | Few-group flux (integral, normalized) |
|- | |- | ||
| − | | | + | | INF_KINF |
| 2 | | 2 | ||
| − | | | + | | Infinite multiplication factor |
|- | |- | ||
|} | |} | ||
| − | == | + | ==== Reaction cross sections ==== |
| − | |||
| − | = | ||
{|class="wikitable" style="text-align: left;" | {|class="wikitable" style="text-align: left;" | ||
| Line 1,586: | Line 1,377: | ||
! Description | ! Description | ||
|- | |- | ||
| − | | | + | | INF_TOT |
| − | | | + | | 2''G'' |
| − | + | | Total cross section | |
|- | |- | ||
| − | | | + | | INF_CAPT |
| − | | 2'' | + | | 2''G'' |
| − | | | + | | Capture cross section |
| + | |- | ||
| + | | INF_FISS | ||
| + | | 2''G'' | ||
| + | | Fission cross section | ||
|- | |- | ||
| − | | | + | | INF_NSF |
| − | | 2'' | + | | 2''G'' |
| − | | | + | | Fission neutron production cross section |
|- | |- | ||
| − | | | + | | INF_KAPPA |
| − | + | | 2''G'' | |
| − | + | | Average deposited fission energy (MeV) | |
| − | |||
| − | |||
| − | |||
| − | |||
| − | |||
|- | |- | ||
| − | | | + | | INF_INVV |
| − | | 2'' | + | | 2''G'' |
| − | | | + | | Inverse neutron speed (s/cm) |
|- | |- | ||
| − | | | + | | INF_NUBAR |
| − | | 2'' | + | | 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) | |
| − | |||
| − | |||
| − | |||
| − | |||
| − | |||
| − | |||
| − | | 2'' | ||
| − | |||
| − | |||
| − | | | ||
| − | |||
| − | |||
|- | |- | ||
|} | |} | ||
| − | === | + | ==== Fission spectra ==== |
{|class="wikitable" style="text-align: left;" | {|class="wikitable" style="text-align: left;" | ||
| Line 1,653: | Line 1,426: | ||
! Description | ! Description | ||
|- | |- | ||
| − | | | + | | INF_CHIT |
| − | | | + | | 2''G'' |
| − | | | + | | Fission spectrum (total) |
|- | |- | ||
| − | | | + | | INF_CHIP |
| − | | | + | | 2''G'' |
| − | | | + | | Fission spectrum (prompt neutrons) |
|- | |- | ||
| − | | | + | | INF_CHID |
| − | | 2'' | + | | 2''G'' |
| − | + | | Fission spectrum (delayed neutrons) | |
| − | |||
| − | |||
| − | |||
| − | | | ||
| − | |||
| − | |||
| − | |||
| − | |||
| − | |||
| − | |||
| − | |||
| − | |||
| − | |||
| − | |||
| − | |||
| − | |||
|- | |- | ||
|} | |} | ||
| − | === | + | ==== Scattering cross sections ==== |
| + | |||
| + | <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 1,690: | Line 1,451: | ||
! Description | ! Description | ||
|- | |- | ||
| − | | | + | | INF_SCATT0 |
| − | | 2 | + | | 2''G'' |
| − | | | + | | Total ''P''<sub>0</sub> scattering cross section |
|- | |- | ||
| − | | | + | | INF_SCATT1 |
| − | | 2 | + | | 2''G'' |
| − | | | + | | Total ''P''<sub>1</sub> scattering cross section |
|- | |- | ||
| − | | | + | | INF_SCATT2 |
| − | | 2 | + | | 2''G'' |
| − | | | + | | Total ''P''<sub>2</sub> scattering cross section |
| + | |- | ||
| + | | INF_SCATT3 | ||
| + | | 2''G'' | ||
| + | | Total ''P''<sub>3</sub> scattering cross section | ||
|- | |- | ||
| − | | | + | | INF_SCATT4 |
| − | | 2 | + | | 2''G'' |
| − | | | + | | Total ''P''<sub>4</sub> scattering cross section |
|- | |- | ||
| − | | | + | | INF_SCATT5 |
| − | + | | 2''G'' | |
| − | + | | Total ''P''<sub>5</sub> scattering cross section | |
| − | |||
| − | |||
| − | |||
| − | |||
| − | |||
|- | |- | ||
| − | | | + | | INF_SCATT6 |
| − | | 2 | + | | 2''G'' |
| − | | | + | | Total ''P''<sub>6</sub> scattering cross section |
|- | |- | ||
| − | | | + | | INF_SCATT7 |
| − | + | | 2''G'' | |
| − | + | | Total ''P''<sub>7</sub> scattering cross section | |
| − | |||
| − | |||
| − | |||
| − | |||
| − | |||
|- | |- | ||
| − | | | + | | INF_SCATTP0 |
| − | | | + | | 2''G'' |
| − | | | + | | Total ''P''<sub>0</sub> scattering production cross section |
|- | |- | ||
| − | | | + | | INF_SCATTP1 |
| − | | | + | | 2''G'' |
| − | | | + | | Total ''P''<sub>1</sub> scattering production cross section |
|- | |- | ||
| − | | | + | | INF_SCATTP2 |
| − | | | + | | 2''G'' |
| − | | | + | | Total ''P''<sub>2</sub> scattering production cross section |
|- | |- | ||
| − | | | + | | INF_SCATTP3 |
| − | | 2 | + | | 2''G'' |
| − | | | + | | Total ''P''<sub>3</sub> scattering production cross section |
|- | |- | ||
| − | | | + | | INF_SCATTP4 |
| − | | | + | | 2''G'' |
| − | | | + | | Total ''P''<sub>4</sub> scattering production cross section |
|- | |- | ||
| − | | | + | | 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 | |
| − | |||
| − | |||
| − | |||
| − | |||
| − | |||
| − | |||
| − | |||
| − | |||
| − | |||
| − | |||
| − | | 2'' | ||
| − | | | ||
| − | |||
| − | |||
| − | |||
| − | |||
|- | |- | ||
|} | |} | ||
| − | === | + | ==== Scattering matrices ==== |
| + | <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 [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;" | {|class="wikitable" style="text-align: left;" | ||
| Line 1,790: | Line 1,531: | ||
! Description | ! Description | ||
|- | |- | ||
| − | | | + | | INF_S0 |
| − | | 2 | + | | 2''G''<sup>2</sup> |
| − | | | + | | ''P''<sub>0</sub> scattering matrix |
|- | |- | ||
| − | | | + | | INF_S1 |
| − | + | | 2''G''<sup>2</sup> | |
| − | + | | ''P''<sub>1</sub> scattering matrix | |
| − | + | |- | |
| − | < | + | | INF_S2 |
| − | + | | 2''G''<sup>2</sup> | |
| − | + | | ''P''<sub>2</sub> scattering matrix | |
| − | |||
| − | |||
| − | |||
| − | |||
| − | |||
| − | |||
| − | |||
| − | |||
| − | |||
| − | |||
| − | |||
| − | |||
| − | |||
| − | |||
| − | |||
| − | |||
| − | |||
| − | |||
| − | |||
|- | |- | ||
| − | | | + | | INF_S3 |
| − | | | + | | 2''G''<sup>2</sup> |
| − | | | + | | ''P''<sub>3</sub> scattering matrix |
| + | |- | ||
| + | | INF_S4 | ||
| + | | 2''G''<sup>2</sup> | ||
| + | | ''P''<sub>4</sub> scattering matrix | ||
|- | |- | ||
| − | | | + | | INF_S5 |
| − | | '' | + | | 2''G''<sup>2</sup> |
| − | | | + | | ''P''<sub>5</sub> scattering matrix |
|- | |- | ||
| − | | | + | | INF_S6 |
| − | | | + | | 2''G''<sup>2</sup> |
| − | | | + | | ''P''<sub>6</sub> scattering matrix |
|- | |- | ||
| − | | | + | | INF_S7 |
| − | | ''G'' | + | | 2''G''<sup>2</sup> |
| − | | | + | | ''P''<sub>7</sub> scattering matrix |
|- | |- | ||
| − | | | + | | INF_SP0 |
| − | + | | 2''G''<sup>2</sup> | |
| − | + | | ''P''<sub>0</sub> scattering production matrix | |
| − | |||
| − | |||
| − | |||
| − | |||
| − | |||
| − | |||
| − | |||
| − | | 2'' | ||
| − | | | ||
|- | |- | ||
| − | | | + | | INF_SP1 |
| − | | 2''G'' | + | | 2''G''<sup>2</sup> |
| − | | | + | | ''P''<sub>1</sub> scattering production matrix |
|- | |- | ||
| − | | | + | | INF_SP2 |
| − | | 2 | + | | 2''G''<sup>2</sup> |
| − | | | + | | ''P''<sub>2</sub> scattering production matrix |
|- | |- | ||
| − | | | + | | INF_SP3 |
| − | + | | 2''G''<sup>2</sup> | |
| − | + | | ''P''<sub>3</sub> scattering production matrix | |
| − | |||
| − | |||
| − | |||
| − | |||
| − | |||
|- | |- | ||
| − | | | + | | INF_SP4 |
| − | | 2''G'' | + | | 2''G''<sup>2</sup> |
| − | | | + | | ''P''<sub>4</sub> scattering production matrix |
|- | |- | ||
| − | | | + | | INF_SP5 |
| − | | 2''G'' | + | | 2''G''<sup>2</sup> |
| − | | | + | | ''P''<sub>5</sub> scattering production matrix |
|- | |- | ||
| − | | | + | | INF_SP6 |
| − | | 2''G'' | + | | 2''G''<sup>2</sup> |
| − | | | + | | ''P''<sub>6</sub> scattering production matrix |
|- | |- | ||
| − | | | + | | INF_SP7 |
| − | | 2''G'' | + | | 2''G''<sup>2</sup> |
| − | | | + | | ''P''<sub>7</sub> scattering production matrix |
|- | |- | ||
| − | | | + | |} |
| − | | | + | |
| − | + | ==== 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 <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;" | ||
| + | ! Parameter | ||
| + | ! Size | ||
| + | ! Description | ||
| + | |- | ||
| + | | INF_TRANSPXS | ||
| 2''G'' | | 2''G'' | ||
| − | | | + | | Transport cross section (calculated using the out-scattering approximation) |
|- | |- | ||
| − | | | + | | INF_DIFFCOEF |
| 2''G'' | | 2''G'' | ||
| − | | | + | | Diffusion coefficient (calculated using the out-scattering approximation) |
|- | |- | ||
| − | | | + | | CMM_TRANSPXS |
| 2''G'' | | 2''G'' | ||
| − | | | + | | Transport cross section calculated using the cumulative migration method (equivalent with the in-scattering approximation) |
|- | |- | ||
| − | | | + | | CMM_TRANSPXS_X |
| 2''G'' | | 2''G'' | ||
| − | | | + | | X-component of the directional transport cross section calculated using the cumulative migration method (equivalent with the in-scattering approximation) |
|- | |- | ||
| − | | | + | | CMM_TRANSPXS_Y |
| 2''G'' | | 2''G'' | ||
| − | | | + | | Y-component of the directional transport cross section calculated using the cumulative migration method (equivalent with the in-scattering approximation) |
|- | |- | ||
| − | | | + | | CMM_TRANSPXS_Z |
| − | + | | 2''G'' | |
| − | + | | Z-component of the directional transport cross section calculated using the cumulative migration method (equivalent with the in-scattering approximation) | |
| − | + | |- | |
| − | + | | CMM_DIFFCOEF | |
| − | + | | 2''G'' | |
| − | + | | Diffusion coefficient calculated using the cumulative migration method (equivalent with the in-scattering approximation) | |
| − | |||
|- | |- | ||
| − | | | + | | CMM_DIFFCOEF_X |
| 2''G'' | | 2''G'' | ||
| − | | | + | | X-component of the directional diffusion coefficient calculated using the cumulative migration method (equivalent with the in-scattering approximation) |
|- | |- | ||
| − | | | + | | CMM_DIFFCOEF_Y |
| 2''G'' | | 2''G'' | ||
| − | | | + | | Y-component of the directional diffusion coefficient calculated using the cumulative migration method (equivalent with the in-scattering approximation) |
|- | |- | ||
| − | | | + | | CMM_DIFFCOEF_Z |
| 2''G'' | | 2''G'' | ||
| − | | | + | | Z-component of the directional diffusion coefficient calculated using the cumulative migration method (equivalent with the in-scattering approximation) |
|- | |- | ||
| − | |} | + | | TRC_TRANSPXS |
| − | + | | 2''G'' | |
| − | ==== | + | | Transport cross section calculated by applying user-defined transport correction factors to total cross section |
| − | + | |- | |
| − | <u>Notes:</u> | + | | TRC_DIFFCOEF |
| + | | 2''G'' | ||
| + | | Diffusion coefficient calculated by applying user-defined transport correction factors to total cross section | ||
| + | |} | ||
| + | |||
| + | ==== 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;" | {|class="wikitable" style="text-align: left;" | ||
| Line 1,941: | Line 1,672: | ||
! Description | ! Description | ||
|- | |- | ||
| − | | | + | | INF_I135_YIELD |
| 2''G'' | | 2''G'' | ||
| − | | | + | | Fission yield of I-135 (cumulative, includes all precursors) |
|- | |- | ||
| − | | | + | | INF_XE135_YIELD |
| 2''G'' | | 2''G'' | ||
| − | | | + | | Fission yield of Xe-135 |
|- | |- | ||
| − | | | + | | INF_XE135M_YIELD |
| 2''G'' | | 2''G'' | ||
| − | | | + | | Fission yield of Xe-135m |
|- | |- | ||
| − | | | + | | INF_PM149_YIELD |
| 2''G'' | | 2''G'' | ||
| − | | | + | | Fission yield of Pm-149 (cumulative, includes all precursors) |
|- | |- | ||
| − | | | + | | INF_SM149_YIELD |
| 2''G'' | | 2''G'' | ||
| − | | | + | | Fission yield of Sm-149 |
|- | |- | ||
| − | | | + | | INF_I135_MICRO_ABS |
| 2''G'' | | 2''G'' | ||
| − | | | + | | Microscopic absorption cross section of I-135 |
|- | |- | ||
| − | | | + | | INF_XE135_MICRO_ABS |
| 2''G'' | | 2''G'' | ||
| − | | | + | | Microscopic absorption cross section of Xe-135 |
|- | |- | ||
| − | | | + | | INF_XE135M_MICRO_ABS |
| 2''G'' | | 2''G'' | ||
| − | | | + | | Microscopic absorption cross section of Xe-135m |
|- | |- | ||
| − | | | + | | INF_PM149_MICRO_ABS |
| 2''G'' | | 2''G'' | ||
| − | | | + | | Microscopic absorption cross section of Pm-149 |
|- | |- | ||
| − | | | + | | INF_SM149_MICRO_ABS |
| 2''G'' | | 2''G'' | ||
| − | | | + | | Microscopic absorption cross section of Sm-149 |
|- | |- | ||
| − | | | + | | INF_XE135_MACRO_ABS |
| 2''G'' | | 2''G'' | ||
| − | | | + | | Macroscopic absorption cross section of Xe-135 |
|- | |- | ||
| − | | | + | | INF_XE135M_MACRO_ABS |
| 2''G'' | | 2''G'' | ||
| − | | | + | | Macroscopic absorption cross section of Xe-135m |
|- | |- | ||
| − | | | + | | INF_SM149_MACRO_ABS |
| 2''G'' | | 2''G'' | ||
| − | | | + | | Macroscopic absorption cross section of Sm-149 |
| − | |||
| − | |||
| − | |||
| − | |||
| − | |||
| − | |||
| − | |||
| − | |||
| − | |||
| − | |||
| − | |||
| − | |||
|- | |- | ||
|} | |} | ||
| − | + | ==== Poison decay constants ==== | |
| − | ==== | ||
| − | |||
| − | |||
| − | |||
| − | |||
| − | |||
| − | |||
| − | |||
{|class="wikitable" style="text-align: left;" | {|class="wikitable" style="text-align: left;" | ||
| Line 2,021: | Line 1,732: | ||
! Description | ! 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 === |
| + | <span id="B1"></span> | ||
| + | |||
| + | {|class="wikitable" style="text-align: left;" | ||
| + | ! Parameter | ||
| + | ! Size | ||
| + | ! Description | ||
| + | |- | ||
| + | | B1_MICRO_FLX | ||
| + | | 2''H'' | ||
| + | | Multi-group flux spectrum (integral, un-normalized) | ||
|- | |- | ||
| − | | | + | | B1_FLX |
| − | | 2''G'' | + | | 2''G'' |
| − | | | + | | Few-group flux (integral, normalized) |
|- | |- | ||
| − | | | + | | B1_KINF |
| − | | 2 | + | | 2 |
| − | | | + | | Infinite multiplication factor |
|- | |- | ||
| − | | | + | | B1_KEFF |
| − | | 2 | + | | 2 |
| − | | | + | | Effective multiplication factor |
|- | |- | ||
| − | | | + | | B1_B2 |
| − | | 2 | + | | 2 |
| − | | | + | | Critical buckling |
|- | |- | ||
| − | | | + | | B1_ERR |
| − | | 2 | + | | 2 |
| − | + | | Absolute deviation of ''k''<sub>eff</sub> from unity | |
| − | |||
| − | |||
| − | |||
| − | | '' | ||
|- | |- | ||
|} | |} | ||
| − | ==== | + | ==== Reaction cross sections ==== |
| − | + | {|class="wikitable" style="text-align: left;" | |
| − | |||
| − | |||
| − | |||
| − | |||
| − | |||
| − | |||
| − | {|class="wikitable" style="text-align: left;" | ||
! Parameter | ! Parameter | ||
! Size | ! Size | ||
! Description | ! Description | ||
| + | |- | ||
| + | | B1_TOT | ||
| + | | 2''G'' | ||
| + | | Total cross section | ||
|- | |- | ||
| − | | | + | | B1_CAPT |
| 2''G'' | | 2''G'' | ||
| − | | | + | | Capture cross section |
|- | |- | ||
| − | | | + | | B1_FISS |
| 2''G'' | | 2''G'' | ||
| − | | | + | | Fission cross section |
|- | |- | ||
| − | | | + | | B1_NSF |
| 2''G'' | | 2''G'' | ||
| − | | | + | | Fission neutron production cross section |
|- | |- | ||
| − | | | + | | B1_KAPPA |
| 2''G'' | | 2''G'' | ||
| − | | | + | | Average deposited fission energy (MeV) |
|- | |- | ||
| − | | | + | | B1_INVV |
| 2''G'' | | 2''G'' | ||
| − | | | + | | Inverse neutron speed (s/cm) |
|- | |- | ||
| − | | | + | | B1_NUBAR |
| 2''G'' | | 2''G'' | ||
| − | | | + | | Average neutron yield |
|- | |- | ||
| − | | | + | | B1_ABS |
| 2''G'' | | 2''G'' | ||
| − | | | + | | Absorption cross section (capture + fission) |
|- | |- | ||
| − | | | + | | B1_REMXS |
| 2''G'' | | 2''G'' | ||
| − | | | + | | Removal cross section (group-removal + absorption) |
|- | |- | ||
| − | | | + | | B1_RABSXS |
| 2''G'' | | 2''G'' | ||
| − | | | + | | Reduced absorption cross section (total - scattering production) |
| + | |- | ||
| + | |} | ||
| + | |||
| + | ==== Fission spectra ==== | ||
| + | |||
| + | {|class="wikitable" style="text-align: left;" | ||
| + | ! Parameter | ||
| + | ! Size | ||
| + | ! Description | ||
|- | |- | ||
| − | | | + | | B1_CHIT |
| 2''G'' | | 2''G'' | ||
| − | | | + | | Fission spectrum (total) |
|- | |- | ||
| − | | | + | | B1_CHIP |
| 2''G'' | | 2''G'' | ||
| − | | | + | | Fission spectrum (prompt neutrons) |
|- | |- | ||
| − | | | + | | B1_CHID |
| 2''G'' | | 2''G'' | ||
| − | | | + | | Fission spectrum (delayed neutrons) |
| + | |- | ||
|} | |} | ||
| − | ==== | + | ==== Scattering cross sections ==== |
<u>Notes:</u> | <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 2,162: | Line 1,881: | ||
! Description | ! Description | ||
|- | |- | ||
| − | | | + | | B1_SCATT0 |
| 2''G'' | | 2''G'' | ||
| − | | | + | | Total ''P''<sub>0</sub> scattering cross section |
|- | |- | ||
| − | | | + | | B1_SCATT1 |
| 2''G'' | | 2''G'' | ||
| − | | | + | | Total ''P''<sub>1</sub> scattering cross section |
|- | |- | ||
| − | | | + | | B1_SCATT2 |
| 2''G'' | | 2''G'' | ||
| − | | | + | | Total ''P''<sub>2</sub> scattering cross section |
|- | |- | ||
| − | | | + | | B1_SCATT3 |
| 2''G'' | | 2''G'' | ||
| − | | | + | | Total ''P''<sub>3</sub> scattering cross section |
|- | |- | ||
| − | | | + | | B1_SCATT4 |
| 2''G'' | | 2''G'' | ||
| − | | | + | | Total ''P''<sub>4</sub> scattering cross section |
|- | |- | ||
| − | | | + | | B1_SCATT5 |
| 2''G'' | | 2''G'' | ||
| − | | | + | | Total ''P''<sub>5</sub> scattering cross section |
|- | |- | ||
| − | | | + | | B1_SCATT6 |
| 2''G'' | | 2''G'' | ||
| − | | | + | | Total ''P''<sub>6</sub> scattering cross section |
|- | |- | ||
| − | | | + | | B1_SCATT7 |
| 2''G'' | | 2''G'' | ||
| − | | | + | | Total ''P''<sub>7</sub> scattering cross section |
|- | |- | ||
| − | | | + | | B1_SCATTP0 |
| 2''G'' | | 2''G'' | ||
| − | | | + | | Total ''P''<sub>0</sub> scattering production cross section |
|- | |- | ||
| − | | | + | | B1_SCATTP1 |
| 2''G'' | | 2''G'' | ||
| − | | | + | | Total ''P''<sub>1</sub> scattering production cross section |
|- | |- | ||
| − | | | + | | B1_SCATTP2 |
| 2''G'' | | 2''G'' | ||
| − | | | + | | Total ''P''<sub>2</sub> scattering production cross section |
|- | |- | ||
| − | | | + | | B1_SCATTP3 |
| 2''G'' | | 2''G'' | ||
| − | | | + | | Total ''P''<sub>3</sub> scattering production cross section |
|- | |- | ||
| − | | | + | | B1_SCATTP4 |
| 2''G'' | | 2''G'' | ||
| − | | | + | | Total ''P''<sub>4</sub> scattering production cross section |
|- | |- | ||
| − | | | + | | B1_SCATTP5 |
| 2''G'' | | 2''G'' | ||
| − | | | + | | Total ''P''<sub>5</sub> scattering production cross section |
|- | |- | ||
| − | | | + | | B1_SCATTP6 |
| 2''G'' | | 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 ([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;" | {|class="wikitable" style="text-align: left;" | ||
| Line 2,231: | Line 1,962: | ||
! Description | ! Description | ||
|- | |- | ||
| − | | | + | | B1_S0 |
| − | | | + | | 2''G''<sup>2</sup> |
| − | | | + | | ''P''<sub>0</sub> scattering matrix |
|- | |- | ||
| − | | | + | | B1_S1 |
| − | | 1 | + | | 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'' | + | | 2''G''<sup>2</sup> |
| − | | | + | | ''P''<sub>2</sub> scattering production matrix |
|- | |- | ||
| − | | | + | | B1_SP3 |
| − | | 2 | + | | 2''G''<sup>2</sup> |
| − | | | + | | ''P''<sub>3</sub> scattering production matrix |
|- | |- | ||
| − | | | + | | B1_SP4 |
| − | | 2 | + | | 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 | + | | 2''G''<sup>2</sup> |
| − | | | + | | ''P''<sub>6</sub> scattering production matrix |
|- | |- | ||
| − | | | + | | B1_SP7 |
| − | | 2 | + | | 2''G''<sup>2</sup> |
| − | | | + | | ''P''<sub>7</sub> scattering production matrix |
|- | |- | ||
|} | |} | ||
| − | ==== | + | ==== Diffusion parameters ==== |
{|class="wikitable" style="text-align: left;" | {|class="wikitable" style="text-align: left;" | ||
| Line 2,305: | Line 2,034: | ||
! Size | ! Size | ||
! Description | ! Description | ||
| − | |- | + | |- |
| − | | | + | | B1_TRANSPXS |
| 2''G'' | | 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'' | | 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'' | | 2''G'' | ||
| − | | Fission | + | | Fission yield of I-135 (cumulative, includes all precursors) |
|- | |- | ||
| − | | | + | | B1_XE135_YIELD |
| 2''G'' | | 2''G'' | ||
| − | | Fission | + | | Fission yield of Xe-135 |
|- | |- | ||
| − | | | + | | B1_XE135M_YIELD |
| 2''G'' | | 2''G'' | ||
| − | | | + | | Fission yield of Xe-135m |
|- | |- | ||
| − | | | + | | B1_PM149_YIELD |
| 2''G'' | | 2''G'' | ||
| − | | | + | | Fission yield of Pm-149 (cumulative, includes all precursors) |
|- | |- | ||
| − | | | + | | B1_SM149_YIELD |
| 2''G'' | | 2''G'' | ||
| − | | | + | | Fission yield of Sm-149 |
|- | |- | ||
| − | | | + | | B1_I135_MICRO_ABS |
| 2''G'' | | 2''G'' | ||
| − | | | + | | Microscopic absorption cross section of I-135 |
|- | |- | ||
| − | | | + | | B1_XE135_MICRO_ABS |
| 2''G'' | | 2''G'' | ||
| − | | | + | | Microscopic absorption cross section of Xe-135 |
|- | |- | ||
| − | | | + | | B1_XE135M_MICRO_ABS |
| 2''G'' | | 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'' | | 2''G'' | ||
| − | | | + | | Microscopic absorption cross section of Sm-149 |
|- | |- | ||
| − | | | + | | B1_XE135_MACRO_ABS |
| 2''G'' | | 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'' | | 2''G'' | ||
| − | | | + | | Macroscopic absorption cross section of Sm-149 |
|- | |- | ||
|} | |} | ||
| − | === | + | === Delayed neutron data === |
<u>Notes:</u> | <u>Notes:</u> | ||
| − | * | + | *The output always consists of 9 values: total, followed by precursor group-wise values. If the number of groups is 6, the last two values are zero. |
| + | *The actual number of groups depends on the cross section library used in the calculations. JEFF-3.1, JEFF.3.2 and later evaluations use 8 precursor groups, while earlier evaluations, as well as all ENDF/B and JENDL data is based on 6 groups. | ||
{|class="wikitable" style="text-align: left;" | {|class="wikitable" style="text-align: left;" | ||
| Line 2,380: | Line 2,123: | ||
! Description | ! Description | ||
|- | |- | ||
| − | | | + | | BETA_EFF |
| − | | | + | | 9 |
| − | | | + | | Effective delayed neutron fraction (currently calculated using the Meulekamp method) |
|- | |- | ||
| − | | | + | | LAMBDA |
| − | | | + | | 9 |
| − | | | + | | Decay constants |
|- | |- | ||
| − | | | + | |} |
| − | | 2'' | + | |
| − | + | === 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_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_VOLUME |
| − | | | + | | 1 |
| − | | | + | | Volume (3D) or cross sectional area (2D) of the homogenized cell |
|- | |- | ||
| − | | | + | | DF_SURF_AREA |
| − | | | + | | ''N''<sub>S</sub> |
| − | + | | Area (3D) or perimeter length (2D) of the surface region | |
|- | |- | ||
| − | | | + | | DF_MID_AREA |
| − | | | + | | ''N''<sub>S</sub> |
| − | + | | Area (3D) or perimeter length (2D) of the mid-point region | |
|- | |- | ||
| − | | | + | | DF_CORN_AREA |
| − | | | + | | ''N''<sub>C</sub> |
| − | + | | Area (3D) or perimeter length (2D) of the corner region | |
|- | |- | ||
| − | | | + | | DF_SURF_IN_CURR |
| − | | | + | | ''2G'' <math>\times</math> ''N''<sub>S</sub> |
| − | + | | Inward surface currents | |
|- | |- | ||
| − | | | + | | DF_SURF_OUT_CURR |
| − | | | + | | ''2G'' <math>\times</math> ''N''<sub>S</sub> |
| − | + | | Outward surface currents | |
|- | |- | ||
| − | | | + | | DF_SURF_NET_CURR |
| − | | | + | | ''2G'' <math>\times</math> ''N''<sub>S</sub> |
| − | + | | Net surface currents | |
|- | |- | ||
| − | | | + | | DF_MID_IN_CURR |
| − | | | + | | ''2G'' <math>\times</math> ''N''<sub>S</sub> |
| − | + | | Inward mid-point currents | |
|- | |- | ||
| − | | | + | | DF_MID_OUT_CURR |
| − | | | + | | ''2G'' <math>\times</math> ''N''<sub>S</sub> |
| − | + | | Outward mid-point currents | |
|- | |- | ||
| − | | | + | | DF_MID_NET_CURR |
| − | + | | ''2G'' <math>\times</math> ''N''<sub>S</sub> | |
| − | + | | Net mid-point currents | |
| − | |||
| − | |||
| − | |||
| − | |||
| − | |||
| − | |||
| − | |||
| − | |||
| − | |||
| − | |||
| − | |||
| − | |||
| − | |||
|- | |- | ||
| − | | | + | | DF_CORN_IN_CURR |
| − | | | + | | ''2G'' <math>\times</math> ''N''<sub>C</sub> |
| − | + | | Inward corner currents | |
|- | |- | ||
| − | | | + | | DF_CORN_OUT_CURR |
| − | | | + | | ''2G'' <math>\times</math> ''N''<sub>C</sub> |
| − | + | | Outward corner currents | |
| − | |||
| − | |||
| − | |||
| − | | | ||
|- | |- | ||
| − | | | + | | DF_CORN_NET_CURR |
| − | | | + | | ''2G'' <math>\times</math> ''N''<sub>C</sub> |
| − | + | | Net corner currents | |
|- | |- | ||
| − | | | + | | DF_HET_VOL_FLUX |
| − | | | + | | ''2G'' |
| − | | | + | | Heterogeneous flux over homogenized cell |
|- | |- | ||
| − | | | + | | DF_HET_SURF_FLUX |
| − | | | + | | ''2G'' <math>\times</math> ''N''<sub>S</sub> |
| − | + | | Heterogeneous surface fluxes | |
|- | |- | ||
| − | | | + | | DF_HET_CORN_FLUX |
| − | | | + | | ''2G'' <math>\times</math> ''N''<sub>C</sub> |
| − | + | | Heterogeneous corner fluxes | |
|- | |- | ||
| − | | | + | | DF_HOM_VOL_FLUX |
| − | | | + | | ''2G'' |
| − | | | + | | Homogeneous flux over homogenized cell |
|- | |- | ||
| − | | | + | | DF_HOM_SURF_FLUX |
| − | | | + | | ''2G'' <math>\times</math> ''N''<sub>S</sub> |
| − | + | | Homogeneous surface fluxes | |
|- | |- | ||
| − | | | + | | DF_HOM_CORN_FLUX |
| − | | | + | | ''2G'' <math>\times</math> ''N''<sub>C</sub> |
| − | + | | Homogeneous corner fluxes | |
|- | |- | ||
| − | | | + | | DF_SURF_DF |
| − | | | + | | ''2G'' <math>\times</math> ''N''<sub>S</sub> |
| − | + | | Surface discontinuity factors | |
|- | |- | ||
| − | | | + | | DF_CORN_DF |
| − | | | + | | ''2G'' <math>\times</math> ''N''<sub>C</sub> |
| − | + | | Corner discontinuity factors | |
|- | |- | ||
| − | | | + | | DF_SGN_SURF_IN_CURR |
| − | | | + | | ''2G'' <math>\times</math> ''N''<sub>S</sub> |
| − | + | | Inward sign moment weighted currents | |
|- | |- | ||
| − | | | + | | DF_SGN_SURF_OUT_CURR |
| − | | | + | | ''2G'' <math>\times</math> ''N''<sub>S</sub> |
| − | + | | Outward sign moment weighted currents | |
|- | |- | ||
| − | | | + | | DF_SGN_SURF_NET_CURR |
| − | | | + | | ''2G'' <math>\times</math> ''N''<sub>S</sub> |
| − | + | | Net sign moment weighted currents | |
|- | |- | ||
| − | | | + | | DF_SGN_HET_SURF_FLUX |
| − | | | + | | ''2G'' <math>\times</math> ''N''<sub>S</sub> |
| − | + | | Heterogeneous sign moment weighted surface fluxes | |
|- | |- | ||
| − | |} | + | | DF_SGN_HOM_SURF_FLUX |
| + | | ''2G'' <math>\times</math> ''N''<sub>S</sub> | ||
| + | | Homogeneous sign moment weighted surface fluxes | ||
| + | |- | ||
| + | | DF_SGN_SURF_DF | ||
| + | | ''2G'' <math>\times</math> ''N''<sub>S</sub> | ||
| + | | Sign moment weighted surface discontinuity factors | ||
| + | |} | ||
| + | |||
| + | === Pin-power form factors === | ||
| + | |||
| + | <u>Notes:</u> | ||
| − | == | + | *Calculation of pin-power form factors requires the [[Input syntax manual#set ppw|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: | ||
| + | *#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. | ||
| + | *#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). | ||
| + | *#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 [[Input syntax manual#set adf|ADF calculation]]. | ||
| + | *The order of values is: <math>X_{1,1} \, X_{1,2} \, ... \, X_{2,1} \, X_{2,2} \, ...</math> where <math>X_{n,g}</math> refers to parameter <math>X</math> 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. | ||
{|class="wikitable" style="text-align: left;" | {|class="wikitable" style="text-align: left;" | ||
| Line 2,534: | Line 2,302: | ||
! Description | ! Description | ||
|- | |- | ||
| − | | | + | | PPW_LATTICE |
| − | | | + | | (string) |
| − | | | + | | Name of the lattice used for the calculation |
|- | |- | ||
| − | | | + | | PPW_LATTICE_TYPE |
| − | | | + | | 1 |
| − | | | + | | Lattice type (corresponds to the [[Input syntax manual#lat (regular lattice definition)|lat-card]]) |
| + | |- | ||
| + | | PPW_PINS | ||
| + | | 1 | ||
| + | | Number of pin positions in the lattice (denoted as ''N''<sub>P</sub> below) | ||
| + | |- | ||
| + | | PPW_POW | ||
| + | | ''2G'' <math>\times</math> ''N''<sub>P</sub> | ||
| + | | Pin- and group-wise power distribution normalized to unity sum | ||
| + | |- | ||
| + | | PPW_HOM_FLUX | ||
| + | | ''2G'' <math>\times</math> ''N''<sub>P</sub> | ||
| + | | Pin- and group-wise homogeneous flux distribution | ||
| + | |- | ||
| + | | PPW_FF | ||
| + | | ''2G'' <math>\times</math> ''N''<sub>P</sub> | ||
| + | | Pin- and group-wise form factors | ||
|- | |- | ||
|} | |} | ||
| − | === | + | === Albedos === |
<u>Notes:</u> | <u>Notes:</u> | ||
| + | *Calculation of albedos requires the [[Input syntax manual#set alb|set alb]] option. | ||
| + | *The order of the surfaces should be the same as for the [[#Assembly_discontinuity_factors|ADFs]]. | ||
| + | *The order of ALB_IN_CURR is <math>J_{1,1} , J_{1,2} , \ldots , J_{2,1} , J_{2,2} , \ldots</math> where <math>J_{k,g}</math> refers to incoming partial current of surface ''k'' of group ''g''. | ||
| + | *The order of ALB_OUT_CURR is <math>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 </math> where <math>J_{k,g,k',g'}</math> refers to outgoing partial current 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_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''. | ||
| + | * 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> | ||
| + | part_alb = reshape(ALB_PART_ALB(1, 1:2:end), 2, 6, 2, 6)</nowiki> | ||
| − | + | {| class="wikitable" style="text-align: left;" | |
| − | |||
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! Parameter | ! Parameter | ||
! Size | ! Size | ||
! Description | ! Description | ||
|- | |- | ||
| − | + | | ALB_SURFACE | |
| − | + | | (string) | |
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| − | | ALB_SURFACE | ||
| − | | (string) | ||
| Name of the surface used for the calculation | | Name of the surface used for the calculation | ||
|- | |- | ||
| Line 2,863: | Line 2,356: | ||
| ALB_N_SURF | | ALB_N_SURF | ||
| 1 | | 1 | ||
| − | | Number of albedo surface faces (denoted as ''N<sub>S</sub> | + | | Number of albedo surface faces (denoted as ''N''<sub>S</sub> below) |
|- | |- | ||
| ALB_IN_CURR | | ALB_IN_CURR | ||
| − | | ''2G'' | + | | ''2G'' <math>\times</math> ''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> | + | | ''2G''<sup>2</sup> <math>\times</math> ''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 2,878: | Line 2,371: | ||
|- | |- | ||
| ALB_PART_ALB | | ALB_PART_ALB | ||
| − | | ''2G''<sup>2</sup> | + | | ''2G''<sup>2</sup> <math>\times</math> ''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 == | ||
Revision as of 12:24, 25 January 2022
This page lists the output parameters in the main [input]_res.m output file.
Contents
- 1 General output parameters
- 2 Misc. statistics
- 3 Physics
- 4 Particle balance
- 5 Integral results
- 5.1 Normalized total reaction rates (neutrons)
- 5.2 Equilibrium Xe-135 iteration
- 5.3 Equilibrium Sm-149 iteration
- 5.4 Six-factor formula
- 5.5 Fission neutron and energy production
- 5.6 Criticality eigenvalues
- 5.7 ALF (Average lethargy of neutrons causing fission)
- 5.8 EALF (Energy corresponding to average lethargy of neutrons causing fission)
- 6 Time constants
- 6.1 Forward-weighted delayed neutron parameters
- 6.2 Beta-eff using Meulekamp's method
- 6.3 Adjoint weighted time constants using Nauchi's method
- 6.4 Adjoint weighted time constants using IFP
- 6.5 Adjoint weighted time constants using perturbation technique
- 6.6 Inverse neutron speed
- 6.7 Analog slowing-down and thermal neutron lifetime (total/prompt/delayed)
- 7 Homogenized group constants
- 8 Miscellaneous notes for other outputs
- 9 References
General output parameters
Version, title and date
| Parameter | Size | Description |
|---|---|---|
| VERSION | (string) | Code version |
| COMPILE_DATE | (string) | Date when the source code was compiled |
| DEBUG | 1 | Debug flag indicating if the DEBUG option was set when the source code was compiled |
| TITLE | (string) | Title defined using the set title input option |
| CONFIDENTIAL_DATA | 1 | Confidentiality flag set using the set confi input option |
| INPUT_FILE_NAME | (string) | File name of the main input file |
| WORKING_DIRECTORY | (string) | Directory path where the simulation was run |
| HOSTNAME | (string) | Host name where the simulation was run |
| CPU_TYPE | (string) | CPU type of the machine where the simulation was run (parsed from /proc/cpuinfo) |
| CPU_MHZ | (string) | CPU clock frequency of the machine where the simulation was run (parsed from /proc/cpuinfo) |
| START_DATE | (string) | Date and time when the simulation was started |
| COMPLETE_DATE | (string) | Date and time when this output was printed |
Run parameters
| Parameter | Size | Description | |
|---|---|---|---|
| POP | 1 | Population size defined using the set pop input option | |
| CYCLES | 1 | Number of active cycles defined using the set pop input option | |
| SKIP | 1 | Number of inactive cycles defined using the set pop input option | |
| BATCH_INTERVAL | 1 | Batching interval defined using the set pop input option | |
| SRC_NORM_MODE | 1 | Source normalization mode | |
| SEED | 1 | Random number seed taken from system time or defined using the set seed input option | |
| UFS_MODE | 1 | Uniform fission source mode defined using the set ufs input option | |
| UFS_ORDER | 1 | Uniform fission exponential factor using the set ufs input option | |
| NEUTRON_TRANSPORT_MODE | 1 | Flag indicating whether or not neutron transport simulation is on | |
| PHOTON_TRANSPORT_MODE | 1 | Flag indicating whether or not neutron transport simulation is on | |
| GROUP_CONSTANT_GENERATION | 1 | Flag indicating whether or not group constant generation is on | |
| B1_CALCULATION | 3 | Flag indicating whether or not B1 calculation is on | |
| B1_BURNUP_CORRECTION | 1 | Flag indicating whether or not B1 burnup correction is on | |
| CRIT_SPEC_MODE | 2 | Critical spectrum modes | |
| IMPLICIT_REACTION_RATES | 1 | Flag indicating whether or implicit reaction rates are used for group constant generation |
Optimization
| Parameter | Size | Description |
|---|---|---|
| OPTIMIZATION_MODE | 1 | Optimization mode defined using the set opti input option |
| RECONSTRUCT_MICROXS | 1 | Flag indicating whether or not microscopic cross sections are reconstructed on the unionized energy grid |
| RECONSTRUCT_MACROXS | 1 | Flag indicating whether or not macroscopic cross sections are reconstructed on the unionized energy grid |
| DOUBLE_INDEXING | 1 | Double indexing option defined using the set dix input option |
| MG_MAJORANT_MODE | 1 | Multi-group majorant mode |
| SPECTRUM_COLLAPSE | 1 | Spectrum collapse method flag (set xscalc input option) |
Parallelization
| Parameter | Size | Description |
|---|---|---|
| MPI_TASKS | 1 | Number of parallel MPI tasks |
| OMP_THREADS | 1 | Number of parallel OpenMP threads |
| MPI_REPRODUCIBILITY | 1 | MPI reproducibility option defined by the set repro input option |
| OMP_REPRODUCIBILITY | 1 | OpenMP reproducibility option defined by the set repro input option |
| OMP_HISTORY_PROFILE | N | Fraction of particle histories run for each parallel OpenMP thread |
| SHARE_BUF_ARRAY | 1 | Shared buffer flag |
| SHARE_RES2_ARRAY | 1 | Shared RES2 array flag |
| OMP_SHARED_QUEUE_LIM | 1 | Limiting value for using shared particle queue |
File paths
| Parameter | Size | Description |
|---|---|---|
| XS_DATA_FILE_PATH | (string) | Cross section directory file path defined using the set acelib input option |
| DECAY_DATA_FILE_PATH | (string) | Radioactive decay data file path defined using the set declib input option |
| SFY_DATA_FILE_PATH | (string) | Spontaneous fission yield data file path defined using the set sfylib input option |
| NFY_DATA_FILE_PATH | (string) | Neutron-induced fission yield data file path defined using the set nfylib input option |
| BRA_DATA_FILE_PATH | (string) | Isomeric branching ratio data file path defined using the set bralib input option |
Misc. statistics
Collision and reaction sampling (neutrons/photons)
Notes:
- The first single/pair value corresponds to neutrons and, the second single/pair value corresponds to photons.
| Parameter | Size | Description |
|---|---|---|
| MIN_MACROXS | 2/2 | Macroscopic cross section corresponding to the minimum mfp used for scoring the collision flux estimator (see the set cfe input option) |
| DT_THRESH | 1/1 | Probability threshold used for switching to delta-tracking (see the set dt input option) |
| ST_FRAC | 2/2 | Fraction of paths sampled using surface-tracking |
| DT_FRAC | 2/2 | Fraction of paths sampled using delta-tracking |
| DT_EFF | 2/2 | Delta-tracking efficiency |
| REA_SAMPLING_EFF | 2/2 | Reaction sampling efficiency |
| REA_SAMPLING_FAIL | 2/2 | Fraction of failed reaction samples |
| TOT_COL_EFF | 2/2 | Total collision efficiency |
| AVG_TRACKING_LOOPS | 2/2, 2/2 | Average number of tracking loops per history and, fraction of failed tracking loops |
| AVG_TRACKS | 2/2 | Average number of tracks per history |
| AVG_REAL_COL | 2/2 | Average number of real collisions per history |
| AVG_VIRT_COL | 2/2 | Average number of virtual collisions per history |
| AVG_SURF_CROSS | 2/2 | Average number of surface crossings per history (NOTE: accurate only in ST mode) |
| LOST_PARTICLES | 1 | Number of lost particles |
Run statistics
| Parameter | Size | Description |
|---|---|---|
| CYCLE_IDX | 1 | Cycle index when output was printed |
| SIMULATED_HISTORIES | 1 | Number of simulated histories when output was printed |
| MEAN_POP_SIZE | 1 | Mean population size |
| MEAN_POP_WGT | 1 | Mean population weight |
| SIMULATION_COMPLETED | 1 | Flag indicating whether or not the simulation was completed |
Running times
Notes:
- All times in minutes
- In burnup calculations the first value provides the cumulative and the second value the cycle-wise value
| Parameter | Size | Description |
|---|---|---|
| TOT_CPU_TIME | 1 | Total CPU time |
| RUNNING_TIME | 1 | Total wall-clock running time |
| INIT_TIME | 1(2) | Wall-clock time spent for initialization |
| PROCESS_TIME | 1(2) | Wall-clock time spent for processing |
| TRANSPORT_CYCLE_TIME | 1(2) | Wall-clock time spent for transport simulation |
| BURNUP_CYCLE_TIME | 1(2) | Wall-clock time spent for burnup solution |
| BATEMAN_SOLUTION_TIME | 1(2) | Wall-clock time spent for solving the Bateman equations |
| MPI_OVERHEAD_TIME | 1(2) | Wall-clock time spent MPI communication |
| ESTIMATED_RUNNING_TIME | 1(2) | Estimated total wall-clock running time |
| CPU_USAGE | 1 | Total CPU usage fraction |
| TRANSPORT_CPU_USAGE | 1(2) | CPU usage fraction in transport simulation |
| OMP_PARALLEL_FRAC | 1 | Fraction of time spent in OpenMP parallel loops |
Memory usage
Notes:
- All values are in megabytes
- Serpent allocates memory in fixed segments, so the allocated memory size may be larger than what is needed for the simulation
| Parameter | Size | Description |
|---|---|---|
| AVAIL_MEM | 1 | Available memory size |
| ALLOC_MEMSIZE | 1 | Allocated memory size |
| MEMSIZE | 1 | Used memory size |
| XS_MEMSIZE | 1 | Memory size used for storing cross sections |
| MAT_MEMSIZE | 1 | Memory size used for storing material-wise data |
| RES_MEMSIZE | 1 | Memory size used for storing results |
| IFC_MEMSIZE | 1 | Memory size used for data for response-matrix solver |
| MISC_MEMSIZE | 1 | Memory size used for data for miscellaneous data |
| UNKNOWN_MEMSIZE | 1 | Memory size used for data for uncategorized data |
| UNUSED_MEMSIZE | 1 | Allocated memory not used for anything |
Geometry parameters
| Parameter | Size | Description |
|---|---|---|
| TOT_CELLS | 1 | Total number of cells |
| UNION_CELLS | 1 | Total number of cells defined using unions |
Neutron energy grid
| Parameter | Size | Description |
|---|---|---|
| NEUTRON_ERG_TOL | 1 | Reconstruction tolerace for unionized energy grid |
| NEUTRON_ERG_NE | 1 | Number of points in unionized energy grid |
| NEUTRON_EMIN | 1 | Minimum energy for neutron cross section data |
| NEUTRON_EMAX | 1 | Maximum energy for neutron cross section data |
Unresolved resonance probability table sampling
| Parameter | Size | Description |
|---|---|---|
| URES_DILU_CUT | 1 | Density cut-off used for unresolved resonance probability table sampling |
| URES_EMIN | 1 | Minimum energy for unresolved resonance range |
| URES_EMAX | 1 | Maximum energy for unresolved resonance range |
| URES_AVAIL | 1 | Number of nuclides with probability table data |
| URES_USED | 1 | Number of nuclides for which probability table sampling was used |
Nuclides and reaction channels
| Parameter | Size | Description |
|---|---|---|
| TOT_NUCLIDES | 1 | Total number of nuclides |
| TOT_TRANSPORT_NUCLIDES | 1 | Total number of nuclides with transport cross sections |
| TOT_DOSIMETRY_NUCLIDES | 1 | Total number of nclides with dosimetry cross sections |
| TOT_DECAY_NUCLIDES | 1 | Total number of decay nuclides (without transport cross sections) |
| TOT_PHOTON_NUCLIDES | 1 | Total number of nuclides with photon cross section data |
| TOT_REA_CHANNELS | 1 | Total number of reaction channels |
| TOT_TRANSMU_REA | 1 | Total number of transmutation reactions |
Physics
Neutron physics options
| Parameter | Size | Description |
|---|---|---|
| USE_DELNU | 1 | Flag indicating whether or not delayed neutron emission is on (see set delnu input option) |
| USE_URES | 1 | Flag indicating whether or not unresolved resonance probability table sampling is on (see set ures input option) |
| USE_DBRC | 1 | Flag indicating whether or not Doppler-broadening rejection correction is on (see set dbrc input option) |
| IMPL_CAPT | 1 | Flag indicating whether or not implicit capture reaction mode is on (see set impl input option) |
| IMPL_NXN | 1 | Flag indicating whether or not implicit nxn reaction mode is on (see set impl input option) |
| IMPL_FISS | 1 | Flag indicating whether or not implicit fission reaction mode is on (see set impl input option) |
| DOPPLER_PREPROCESSOR | 1 | Flag indicating whether or not Doppler-broadening preprocessor is on (see tmp option, in mat card) |
| TMS_MODE | 1 | Flag indicating whether or not target motion sampling is on (see tms option, in mat card) |
| SAMPLE_FISS | 1 | Flag indicating whether or not fission reactions are handled (see set nphys input option) |
| SAMPLE_CAPT | 1 | Flag indicating whether or not capture reactions are handled (see set nphys input option) |
| SAMPLE_SCATT | 1 | Flag indicating whether or not scattering reactions are handled (see set nphys input option) |
Energy deposition
Notes:
- The list of fission energy release components includes: (1) EFR, kinetic energy of the fission products (following prompt neutron emission from the fission fragments); (2) ENP, kinetic energy of the prompt fission neutrons; (3) END, kinetic energy of the delayed fission neutrons; (4) EGP, total energy release by the emission of prompt gamma rays; (5) EGD, total energy release by the emission of delayed gamma rays; (6) EB, total energy release by delayed beta’s; (7) ENU, energy carried away by neutrinos; (8) ER, total energy less the energy of the neutrinos (ET - ENU), equal to the pseudo-Q-value in File 3 for MT=18; (9) ET, sum of all the partial energies previously listed, corresponding to the total energy release per fission and equal the Q-value.
| Parameter | Size | Description |
|---|---|---|
| EDEP_MODE | 1 | Energy deposition mode (see set edepmode input option) |
| EDEP_DELAYED | 1 | Energy of delayed components in energy deposition calculations (see set edepdel input option) |
| EDEP_KEFF_CORR | 1 | Flag indicating whether or not correction for energy deposition estimates in non-critical systems (see set edepkcorr input option) |
| EDEP_LOCAL_EGD | 1 | Energy distribution of delayed components in energy deposition calculations, mode 3 (see set edepdel input option) |
| EDEP_COMP | 9 | Fission energy release components: EFR, ENP, END, EGP, EGD, EB, ENU, ER, ET. |
| EDEP_CAPT_E | 1 | Additional energy release in capture reactions, mode 1 (see set edepmode input option) |
Radioactivity data
| Parameter | Size | Description |
|---|---|---|
| TOT_ACTIVITY | ||
| TOT_DECAY_HEAT | ||
| TOT_SF_RATE | ||
| ACTINIDE_ACTIVITY | ||
| ACTINIDE_DECAY_HEAT | ||
| FISSION_PRODUCT_ACTIVITY | ||
| FISSION_PRODUCT_DECAY_HEAT | ||
| INHALATION_TOXICITY | ||
| INGESTION_TOXICITY | ||
| ACTINIDE_INH_TOX | ||
| ACTINIDE_ING_TOX | ||
| FISSION_PRODUCT_INH_TOX | ||
| FISSION_PRODUCT_ING_TOX | ||
| SR90_ACTIVITY | ||
| TE132_ACTIVITY | ||
| I131_ACTIVITY | ||
| I132_ACTIVITY | ||
| CS134_ACTIVITY | ||
| CS137_ACTIVITY | ||
| PHOTON_DECAY_SOURCE | ||
| NEUTRON_DECAY_SOURCE | ||
| ALPHA_DECAY_SOURCE | ||
| ELECTRON_DECAY_SOURCE |
Normalization coefficient
| Parameter | Size | Description |
|---|---|---|
| NORM_COEF | 2/2 | Proportionality constant between the simulated events and the "physical" events that the simulated events represent, for neutrons and photons. |
Parameters for burnup calculation
| Parameter | Size | Description |
|---|---|---|
| BURN_MATERIALS | 1 | Number of depleted materials. |
| BURN_MODE | 1 | Burnup mode: 1 = TTA, 2 = CRAM (see set bumode input option). |
| BURN_STEP | 1 | Burnup step index. |
| BURN_RANDOMIZE_DATA | 3 | Flag indicating whether or not randomize data is set on: decay constants, fission yields and decay heat (see set rnddec input option). |
| BURNUP | 2 | Burnup at the current step (in MWd/kgU): cumulative and real-cumulative. |
| 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. |
Analog reaction rate estimators
| Parameter | Size | Description |
|---|---|---|
| CONVERSION_RATIO | ||
| U235_FISS | ||
| U238_FISS | ||
| U235_CAPT | ||
| U238_CAPT | ||
| XE135_CAPT |
Particle balance
Neutron balance (particles/weight)
| Parameter | Size | Description |
|---|---|---|
| BALA_SRC_NEUTRON_SRC | ||
| BALA_SRC_NEUTRON_FISS | ||
| BALA_SRC_NEUTRON_NXN | ||
| BALA_SRC_NEUTRON_VR | ||
| BALA_SRC_NEUTRON_TOT | ||
| BALA_LOSS_NEUTRON_CAPT | ||
| BALA_LOSS_NEUTRON_FISS | ||
| BALA_LOSS_NEUTRON_LEAK | ||
| BALA_LOSS_NEUTRON_CUT | ||
| BALA_LOSS_NEUTRON_ERR | ||
| BALA_LOSS_NEUTRON_TOT | ||
| BALA_NEUTRON_DIFF |
Integral results
Normalized total reaction rates (neutrons)
| Parameter | Size | Description |
|---|---|---|
| TOT_POWER | ||
| TOT_POWDENS | ||
| TOT_GENRATE | ||
| TOT_FISSRATE | ||
| TOT_CAPTRATE | ||
| TOT_ABSRATE | ||
| TOT_SRCRATE | ||
| TOT_FLUX | ||
| TOT_PHOTON_PRODRATE | ||
| TOT_LEAKRATE | ||
| ALBEDO_LEAKRATE | ||
| TOT_LOSSRATE | ||
| TOT_CUTRATE | ||
| TOT_RR | ||
| TOT_XE135_ABSRATE | ||
| INI_FMASS | ||
| TOT_FMASS | ||
| INI_BURN_FMASS | ||
| TOT_BURN_FMASS |
Equilibrium Xe-135 iteration
| Parameter | Size | Description |
|---|---|---|
| XE135_EQUIL_CONC | 2 | Averaged equilibrium Xe-135 concentration (see set xenon input option) |
| I135_EQUIL_CONC | 2 | Averaged equilibrium I-135 concentration (see set xenon input option) |
Equilibrium Sm-149 iteration
| Parameter | Size | Description |
|---|---|---|
| SM149_EQUIL_CONC | 2 | Averaged equilibrium Sm-149 concentration (see set samarium input option) |
| PM149_EQUIL_CONC | 2 | Averaged equilibrium Pm-149 concentration (see set samarium input option) |
Six-factor formula
| Parameter | Size | Description |
|---|---|---|
| SIX_FF_ETA | 2 | Analog estimate of average number of neutrons emitted per thermal neutron absorbed in fuel |
| SIX_FF_F | 2 | Analog estimate of thermal utilization factor |
| SIX_FF_P | 2 | Analog estimate of resonance escape probability |
| SIX_FF_EPSILON | 2 | Analog estimate of fast fission factor |
| SIX_FF_LF | 2 | Analog estimate of fast non-leakage probability |
| SIX_FF_LT | 2 | Analog estimate of thermal non-leakage probability |
| SIX_FF_KINF | 2 | Analog estimate of six-factor kinf (four-factor keff) |
| SIX_FF_KEFF | 2 | Analog estimate of six-factor keff |
Fission neutron and energy production
| Parameter | Size | Description |
|---|---|---|
| NUBAR | ||
| FISSE |
Criticality eigenvalues
| Parameter | Size | Description |
|---|---|---|
| ANA_KEFF | ||
| IMP_KEFF | ||
| COL_KEFF | ||
| ABS_KEFF | ||
| ABS_KINF | ||
| GEOM_ALBEDO |
ALF (Average lethargy of neutrons causing fission)
| Parameter | Size | Description |
|---|---|---|
| ANA_ALF | ||
| IMP_ALF |
EALF (Energy corresponding to average lethargy of neutrons causing fission)
| Parameter | Size | Description |
|---|---|---|
| ANA_EALF | ||
| IMP_EALF |
Time constants
Forward-weighted delayed neutron parameters
| Parameter | Size | Description |
|---|---|---|
| PRECURSOR_GROUPS | ||
| FWD_ANA_BETA_ZERO | ||
| FWD_ANA_LAMBDA |
Beta-eff using Meulekamp's method
| Parameter | Size | Description |
|---|---|---|
| ADJ_MEULEKAMP_BETA_EFF | ||
| ADJ_MEULEKAMP_LAMBDA |
Adjoint weighted time constants using Nauchi's method
| Parameter | Size | Description |
|---|---|---|
| IFP_CHAIN_LENGTH | ||
| ADJ_NAUCHI_GEN_TIME | ||
| ADJ_NAUCHI_LIFETIME | ||
| ADJ_NAUCHI_BETA_EFF | ||
| ADJ_NAUCHI_LAMBDA |
Adjoint weighted time constants using IFP
| Parameter | Size | Description |
|---|---|---|
| ADJ_IFP_GEN_TIME | ||
| ADJ_IFP_LIFETIME | ||
| ADJ_IFP_IMP_BETA_EFF | ||
| ADJ_IFP_IMP_LAMBDA | ||
| ADJ_IFP_ANA_BETA_EFF | ||
| ADJ_IFP_ANA_LAMBDA | ||
| ADJ_IFP_ROSSI_ALPHA |
Adjoint weighted time constants using perturbation technique
| Parameter | Size | Description |
|---|---|---|
| ADJ_PERT_GEN_TIME | ||
| ADJ_PERT_LIFETIME | ||
| ADJ_PERT_BETA_EFF | ||
| ADJ_PERT_ROSSI_ALPHA |
Inverse neutron speed
| Parameter | Size | Description |
|---|---|---|
| ANA_INV_SPD |
Analog slowing-down and thermal neutron lifetime (total/prompt/delayed)
| Parameter | Size | Description |
|---|---|---|
| ANA_SLOW_TIME | ||
| ANA_THERM_TIME | ||
| ANA_THERM_FRAC | ||
| ANA_DELAYED_EMTIME | ||
| ANA_MEAN_NCOL |
Homogenized group constants
Notes:
- Group constants are calculated by first homogenizing the geometry using a multi-group structure with H energy groups. The data is then collapsed into the final few-group structure with G groups using the infinite and leakage-corrected flux spectra.
- The methodology used in Serpent for spatial homogenization is described in a paper published in Annals of Nuclear Energy in 2016.[1]
- The fundamental mode calculation is off by default, and invoked by the set fum option. Otherwise all values with B1 prefix are printed as zeros.
- The intermediate multi-group structure is defined using option set micro or set fum.
- The few-group structure is defined using option set nfg.
- The universes in which the group constants are calculated are listed in option set gcu. The calculation is performed for root universe 0 by default, and can be switched off with "set gcu -1".
- If data is produced in multiple universes within a single run, the data is assigned with different run indexes (idx)
- The parameter names can be listed in the set coefpara option, and they will be included in the group constant output file when the automated burnup sequence is invoked.
- The order in which two-dimensional data (scattering matrices, ADF and pin-power parameters) is printed in the [input].coe output file is different from what is listed below in update 2.1.24 and earlier versions.
Common parameters
| Parameter | Size | Description |
|---|---|---|
| GC_UNIVERSE_NAME | (string) | Name of the universe where spatial homogenization was performed |
| MICRO_NG | 1 | Number of energy groups in the intermediate multi-group structure (referred to as H below) |
| MICRO_E | H + 1 | Group boundaries in the intermediate multi-group structure (in ascending order) |
| MACRO_NG | 1 | Number of energy groups in the final few-group structure (referred to as G below) |
| MACRO_E | G + 1 | Group boundaries in the final few-group structure (in descending order) |
Group constants homogenized in infinite spectrum
| Parameter | Size | Description |
|---|---|---|
| INF_MICRO_FLX | 2H | Multi-group flux spectrum (integral, un-normalized) |
| INF_FLX | 2G | Few-group flux (integral, normalized) |
| INF_KINF | 2 | Infinite multiplication factor |
Reaction cross sections
| Parameter | Size | Description |
|---|---|---|
| INF_TOT | 2G | Total cross section |
| INF_CAPT | 2G | Capture cross section |
| INF_FISS | 2G | Fission cross section |
| INF_NSF | 2G | Fission neutron production cross section |
| INF_KAPPA | 2G | Average deposited fission energy (MeV) |
| INF_INVV | 2G | Inverse neutron speed (s/cm) |
| INF_NUBAR | 2G | Average neutron yield |
| INF_ABS | 2G | Absorption cross section (capture + fission) |
| INF_REMXS | 2G | Removal cross section (group-removal + absorption) |
| INF_RABSXS | 2G | Reduced absorption cross section (total - scattering production) |
Fission spectra
| Parameter | Size | Description |
|---|---|---|
| INF_CHIT | 2G | Fission spectrum (total) |
| INF_CHIP | 2G | Fission spectrum (prompt neutrons) |
| INF_CHID | 2G | Fission spectrum (delayed neutrons) |
Scattering cross sections
Notes:
- Scattering production includes multiplying (n,2n), (n,3n), etc. reactions.
| Parameter | Size | Description |
|---|---|---|
| INF_SCATT0 | 2G | Total P0 scattering cross section |
| INF_SCATT1 | 2G | Total P1 scattering cross section |
| INF_SCATT2 | 2G | Total P2 scattering cross section |
| INF_SCATT3 | 2G | Total P3 scattering cross section |
| INF_SCATT4 | 2G | Total P4 scattering cross section |
| INF_SCATT5 | 2G | Total P5 scattering cross section |
| INF_SCATT6 | 2G | Total P6 scattering cross section |
| INF_SCATT7 | 2G | Total P7 scattering cross section |
| INF_SCATTP0 | 2G | Total P0 scattering production cross section |
| INF_SCATTP1 | 2G | Total P1 scattering production cross section |
| INF_SCATTP2 | 2G | Total P2 scattering production cross section |
| INF_SCATTP3 | 2G | Total P3 scattering production cross section |
| INF_SCATTP4 | 2G | Total P4 scattering production cross section |
| INF_SCATTP5 | 2G | Total P5 scattering production cross section |
| INF_SCATTP6 | 2G | Total P6 scattering production cross section |
| INF_SCATTP7 | 2G | Total P7 scattering production cross section |
Scattering matrices
Notes:
- Scattering production includes multiplying (n,2n), (n,3n), etc. reactions.
- The order of values ([input].coe) or value pairs ([input]_res.m) is:
where 
refers to scattering from group g to g'. - The data in the [input]_res.m file can be read into a G by G matrix with Matlab reshape-command, for example:
reshape(INF_S0(idx,1:2:end), G, G);
| Parameter | Size | Description |
|---|---|---|
| INF_S0 | 2G2 | P0 scattering matrix |
| INF_S1 | 2G2 | P1 scattering matrix |
| INF_S2 | 2G2 | P2 scattering matrix |
| INF_S3 | 2G2 | P3 scattering matrix |
| INF_S4 | 2G2 | P4 scattering matrix |
| INF_S5 | 2G2 | P5 scattering matrix |
| INF_S6 | 2G2 | P6 scattering matrix |
| INF_S7 | 2G2 | P7 scattering matrix |
| INF_SP0 | 2G2 | P0 scattering production matrix |
| INF_SP1 | 2G2 | P1 scattering production matrix |
| INF_SP2 | 2G2 | P2 scattering production matrix |
| INF_SP3 | 2G2 | P3 scattering production matrix |
| INF_SP4 | 2G2 | P4 scattering production matrix |
| INF_SP5 | 2G2 | P5 scattering production matrix |
| INF_SP6 | 2G2 | P6 scattering production matrix |
| INF_SP7 | 2G2 | P7 scattering production matrix |
Diffusion parameters
Notes:
- Calculation of sensible values for INF_TRANSPXS and INF_DIFFCOEF requires fine enough intermediate multi-group structure.
- The cumulative migration method [2] (CMM) was first developed for the OpenMC code.
- CMM diffusion coefficients and transport cross sections are reasonable only when they are calculated over entire geometry (homogenized region covers the entire geometry and is surrounded by periodic or reflective boundary conditions). This means that e.g. pin cell CMM diffusion coefficients can not be calculated from a 2D fuel assembly calculation.
- Calculation of TRC_TRANSPXS and TRC_DIFFCOEF requires defining 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).
| 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_PM149_MICRO_ABS | 2G | Microscopic absorption cross section of Pm-149 |
| INF_SM149_MICRO_ABS | 2G | Microscopic absorption cross section of Sm-149 |
| INF_XE135_MACRO_ABS | 2G | Macroscopic absorption cross section of Xe-135 |
| INF_XE135M_MACRO_ABS | 2G | Macroscopic absorption cross section of Xe-135m |
| INF_SM149_MACRO_ABS | 2G | Macroscopic absorption cross section of Sm-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: where refers to scattering from group g to g'.
- The data in the _res.m file can be read into a G by G matrix with Matlab 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 always consists of 9 values: total, followed by precursor group-wise values. If the number of groups is 6, the last two values are zero.
- The actual number of groups depends on the cross section library used in the calculations. JEFF-3.1, JEFF.3.2 and later evaluations use 8 precursor groups, while earlier evaluations, as well as all ENDF/B and JENDL data is based on 6 groups.
| Parameter | Size | Description |
|---|---|---|
| BETA_EFF | 9 | Effective delayed neutron fraction (currently calculated using the Meulekamp method) |
| LAMBDA | 9 | Decay constants |
Assembly discontinuity factors
Notes:
- Calculation of assembly discontinuity factors requires the set adf option.
- Surface flux and current tallies are used to calculate the boundary currents and fluxes. Mid-point and corner values are approximated by integrating over a small surface segment.
- The surface and volume fluxes are flux densities, i.e. they are surface or volume integrated fluxes divided by the respective surface area or volume.
- The currents are surface integrated values.
- The net current is defined as current in subtracted with current out.
- When the homogenized region is surrounded by reflective boundary conditions (zero net-current) the homogeneous flux becomes flat and equal to the volume-averaged heterogeneous flux. When the net currents are non-zero, the homogeneous flux is obtained using the Built-in diffusion flux solver.
- The calculation currently supports only a limited number of surface types: infinite planes and square and hexagonal prisms.
- The order of surface and mid-point values for square prisms is:
and the order of corner values: 
where 
refers to parameter 
on surface/corner k and energy group g. - The order of surface values for Y-type hexagonal prims runs clockwise starting from the north, i.e. N, NE, SE, S, SW, NW. The corner values run counterclockwise starting from east, i.e. E, NE, NW, W, SW, SE.
- The order of surface values for X-type hexagonal prims runs counterclockwise starting from the east, i.e. E, NE, NW, W, SW, SE. The corner values run clockwise starting from north, i.e. N, NE, SE, S, SW, NW.
- The sign moment weighted parameters are calculated only for surface types sqc, rect and hexxc.
- The convention of sign moment directions follows that of the nodal neutronics program Ants.
- The ADF symmetry options on set adf card are currently not used for sign moment weighted parameters.
| Parameter | Size | Description |
|---|---|---|
| DF_SURFACE | (string) | Name of the surface used for the calculation |
| DF_SYM | 1 | Symmetry option defined in the input |
| DF_N_SURF | 1 | Number of surface values (denoted as NS below) |
| DF_N_CORN | 1 | Number of corner values (denoted as NC below) |
| DF_VOLUME | 1 | Volume (3D) or cross sectional area (2D) of the homogenized cell |
| DF_SURF_AREA | NS | Area (3D) or perimeter length (2D) of the surface region |
| DF_MID_AREA | NS | Area (3D) or perimeter length (2D) of the mid-point region |
| DF_CORN_AREA | NC | Area (3D) or perimeter length (2D) of the corner region |
| DF_SURF_IN_CURR | 2G  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 NS
|
Inward sign moment weighted currents |
| DF_SGN_SURF_OUT_CURR | 2G NS
|
Outward sign moment weighted currents |
| DF_SGN_SURF_NET_CURR | 2G NS
|
Net sign moment weighted currents |
| DF_SGN_HET_SURF_FLUX | 2G NS
|
Heterogeneous sign moment weighted surface fluxes |
| DF_SGN_HOM_SURF_FLUX | 2G NS
|
Homogeneous sign moment weighted surface fluxes |
| DF_SGN_SURF_DF | 2G NS
|
Sign moment weighted surface discontinuity factors |
Pin-power form factors
Notes:
- Calculation of pin-power form factors requires the set ppw option.
- The power distribution is calculated by tallying the few-group fission energy deposition in each lattice position and dividing the values with the total energy produced in the universe (sum over all values of PPW_POW equals 1).
- The calculation of form factors depends on the boundary conditions:
- 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.
- 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).
- 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:
where 
refers to parameter 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
where 
refers to incoming partial current of surface k of group g. - The order of ALB_OUT_CURR is
where 
refers to outgoing partial current of surface k' of group g' which has entered the albedo surface through surface k and group g. - The order of ALB_TOT_ALB is
where 
refers to albedo from group g to g'. - The order of ALB_PART_ALB is
where 
refers to albedo of surface k' of group g' which has entered the albedo surface through surface k and group g. - For example, 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)
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
Delayed neutrons accounted for in ANA_KEFF
Since Serpent 2.1.23, ANA_KEFF estimator is calculated separately for delayed neutrons. The first two values are total, 3-4 are prompt neutron multiplication only and 5-6 delayed neutron multiplication only. [3]
References
- ^ 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.
- ^ Liu, Z., Smith, K., Forget, B. and Ortensi, J."Cumulative migration method for computing rigorous diffusion coefficients and transport cross sections from Monte Carlo." Ann. Nucl. Energy, 118 (2018) 507-516.
- ^ http://ttuki.vtt.fi/serpent/viewtopic.php?f=25&t=1885&p=4469
