Difference between revisions of "Output parameters"
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Line 70: | Line 70: | ||
| POP | | POP | ||
| 1 | | 1 | ||
− | | Population size defined using the [[Input syntax manual#set pop|set pop]] input option | + | | Population size defined using the [[Input syntax manual#set pop|set pop]] input option (criticality) or the [[Input syntax manual#set nps|set nps]] input option (external source) |
|- | |- | ||
| CYCLES | | CYCLES | ||
Line 83: | Line 83: | ||
| 1 | | 1 | ||
| Batching interval defined using the [[Input syntax manual#set pop|set pop]] input option | | Batching interval defined using the [[Input syntax manual#set pop|set pop]] input option | ||
− | |||
− | |||
− | |||
− | |||
|- | |- | ||
| BATCHES | | BATCHES | ||
− | | | + | | 1 |
− | | | + | | Number of batches defined using the [[Input syntax manual#set nps|set nps]] input option |
|- | |- | ||
| SRC_NORM_MODE | | SRC_NORM_MODE | ||
Line 134: | Line 130: | ||
| IMPLICIT_REACTION_RATES | | IMPLICIT_REACTION_RATES | ||
| 1 | | 1 | ||
− | | Flag indicating whether or implicit reaction rates are used for group constant generation | + | | Flag indicating whether or not implicit reaction rates are used for group constant generation |
|- | |- | ||
| VR_ITER_IDX | | VR_ITER_IDX | ||
− | | | + | | 1 |
− | | | + | | Variance reduction iteration index when output was printed (see [[Input syntax manual#wwin (weight window mesh definition)|wwin]] card) |
+ | |- | ||
+ | |} | ||
+ | |||
+ | === Domain decomposition === | ||
+ | |||
+ | {|class="wikitable" style="text-align: left;" | ||
+ | ! Parameter | ||
+ | ! Size | ||
+ | ! Description | ||
+ | |- | ||
+ | | DD_MODE | ||
+ | | 1 | ||
+ | | Domain decomposition mode defined using the [[Input syntax manual#set dd|set dd]] input option | ||
+ | |- | ||
+ | | DD_NEUTRONS_TO_LIMBO | ||
+ | | ''M'' | ||
+ | | Neutrons sent to limbo (transferral buffer) at each domain | ||
+ | |- | ||
+ | | DD_NEUTRONS_FROM_LIMBO | ||
+ | | ''M'' | ||
+ | | Neutrons received from limbo (transferral buffer) at each domain | ||
|- | |- | ||
|} | |} | ||
Line 215: | Line 232: | ||
=== File paths === | === File paths === | ||
+ | <u>Notes:</u> | ||
+ | |||
+ | * Only the first file path listed is displayed | ||
{|class="wikitable" style="text-align: left;" | {|class="wikitable" style="text-align: left;" | ||
Line 242: | Line 262: | ||
|- | |- | ||
| PHOTON_PHYS_DIRECTORY | | PHOTON_PHYS_DIRECTORY | ||
− | | | + | | (string) |
− | | | + | | Photon physics directory path defined using the [[Input syntax manual#set pdatadir|set pdatadir]] input option |
|- | |- | ||
|} | |} | ||
Line 258: | Line 278: | ||
! Size | ! Size | ||
! Description | ! Description | ||
+ | |- | ||
+ | | MEAN_SRC_WGT | ||
+ | | 2/2 | ||
+ | | Mean source weight for non-criticality calculations (neutrons/photons) | ||
+ | |- | ||
+ | | SOURCE_SAMPLING_EFF | ||
+ | | 2/2 | ||
+ | | Source sampling efficiency for non-criticality calculations (neutrons/photons) | ||
+ | |- | ||
+ | | MEAN_SRC_WW_SPLIT | ||
+ | | 2/2 | ||
+ | | Mean source weight-window splitting in variance reduction (neutrons/photons) | ||
+ | |- | ||
+ | | MEAN_SRC_WW_EFF | ||
+ | | 2/2 | ||
+ | | Mean source weight-window sampling efficiency in variance reduction (neutrons/photons) | ||
+ | |- | ||
+ | | WW_BALA_ROULETE | ||
+ | | 2/2 | ||
+ | | Mean weight-window balance due to Russian roulette in variance reduction (neutrons/photons) | ||
+ | |- | ||
+ | | WW_BALA_SPLIT | ||
+ | | 2/2 | ||
+ | | Mean weight-window balance due to splitting in variance reduction (neutrons/photons) | ||
|- | |- | ||
| MIN_MACROXS | | MIN_MACROXS | ||
Line 278: | Line 322: | ||
| 2/2 | | 2/2 | ||
| Delta-tracking efficiency | | Delta-tracking efficiency | ||
+ | |- | ||
+ | | IFC_COL_EFF | ||
+ | | 2/2 | ||
+ | | Efficiency of interface collision rejection (see [[Input syntax manual#ifc (interface file)|ifc]] card) | ||
|- | |- | ||
| REA_SAMPLING_EFF | | REA_SAMPLING_EFF | ||
Line 286: | Line 334: | ||
| 2/2 | | 2/2 | ||
| Fraction of failed reaction samples | | Fraction of failed reaction samples | ||
+ | |- | ||
+ | | TMS_SAMPLING_EFF | ||
+ | | 2 | ||
+ | | Target motion sampling method, TMS, sampling efficiency (see [[Input syntax manual#mat (material definition)|tms]] option, in [[Input syntax manual#mat (material definition)|mat]] card) | ||
|- | |- | ||
| TOT_COL_EFF | | TOT_COL_EFF | ||
Line 294: | Line 346: | ||
| 2/2, 2/2 | | 2/2, 2/2 | ||
| Average number of tracking loops per history and, fraction of failed tracking loops | | Average number of tracking loops per history and, fraction of failed tracking loops | ||
+ | |- | ||
+ | | TMS_FAIL_STAT | ||
+ | | 8 | ||
+ | | TMS fail statistics: total samples, majorant fail, lower limit fail, upper limit fail (see [[Input syntax manual#mat (material definition)|tms]] option, in [[Input syntax manual#mat (material definition)|mat]] card) | ||
+ | |- | ||
+ | | DBRC_EXCEED_FRAC | ||
+ | | 1 | ||
+ | | Doppler-broadening rejection correction, DBRC, majorant exceed fraction (see [[Input syntax manual#set dbrc|set dbrc]] input option) | ||
|- | |- | ||
| AVG_TRACKS | | AVG_TRACKS | ||
Line 314: | Line 374: | ||
| 1 | | 1 | ||
| Number of lost particles | | Number of lost particles | ||
+ | |- | ||
+ | |} | ||
+ | |||
+ | === STL geometries === | ||
+ | |||
+ | {|class="wikitable" style="text-align: left;" | ||
+ | ! Parameter | ||
+ | ! Size | ||
+ | ! Description | ||
+ | |- | ||
+ | | STL_RAY_TEST | ||
+ | | 5 | ||
+ | | STL-ray-tracing test: total, ray is too parallel to facet, intersection point is too close to edge, facet is too close to search mesh cell boundary, two facets overlap | ||
+ | |- | ||
+ | | STL_ENFORCE_ST | ||
+ | | 1 | ||
+ | | Flag indicating whether or not delta tracking is enforced in STL geometries | ||
+ | |- | ||
+ | |} | ||
+ | |||
+ | === Importance solver === | ||
+ | |||
+ | {|class="wikitable" style="text-align: left;" | ||
+ | ! Parameter | ||
+ | ! Size | ||
+ | ! Description | ||
+ | |- | ||
+ | | NEIGHBOUR_SEARCH_FAIL | ||
+ | | 2 | ||
+ | | Response matrix calculation fail rate | ||
|- | |- | ||
|} | |} | ||
Line 374: | Line 464: | ||
|- | |- | ||
| TRANSPORT_CYCLE_TIME | | TRANSPORT_CYCLE_TIME | ||
− | | 1( | + | | 1(3) |
| Wall-clock time spent for transport simulation | | Wall-clock time spent for transport simulation | ||
+ | |- | ||
+ | | FINIX_SOLUTION_TIME | ||
+ | | 1 | ||
+ | | Wall-clock time spent for FINIX solution | ||
|- | |- | ||
| BURNUP_CYCLE_TIME | | BURNUP_CYCLE_TIME | ||
Line 390: | Line 484: | ||
|- | |- | ||
| DD_OVERHEAD_TIME | | DD_OVERHEAD_TIME | ||
− | | | + | | 1 |
− | | | + | | Wall-clock time spent for DD algorithm (see [[Input syntax manual#set dd|set dd]] input option) |
|- | |- | ||
| RMX_SOLUTION_TIME | | RMX_SOLUTION_TIME | ||
− | | | + | | 1 |
− | | | + | | Wall-clock time spent for response matrix solution (see [[Input syntax manual#wwgen (response matrix based importance map solver)|wwgen]] / [[Input syntax manual#wwin (weight window mesh definition)|wwin]] cards or [[Input syntax manual#set sca|set sca]] input option) |
|- | |- | ||
| LEAKAGE_CORR_SOL_TIME | | LEAKAGE_CORR_SOL_TIME | ||
− | | | + | | 1 |
− | | | + | | Wall-clock time spent for leakage correction solution (see [[Input syntax manual#set fum|set fum]] input option) |
|- | |- | ||
| ESTIMATED_RUNNING_TIME | | ESTIMATED_RUNNING_TIME | ||
Line 456: | Line 550: | ||
| IFC_MEMSIZE | | IFC_MEMSIZE | ||
| 1 | | 1 | ||
− | | Memory size used for data for | + | | Memory size used for data for multi-physics interface data |
|- | |- | ||
| RMX_MEMSIZE | | RMX_MEMSIZE | ||
− | | | + | | 1 |
− | | | + | | Memory size used for storing response matrix-wise data |
|- | |- | ||
| MISC_MEMSIZE | | MISC_MEMSIZE | ||
Line 502: | Line 596: | ||
| NEUTRON_ERG_TOL | | NEUTRON_ERG_TOL | ||
| 1 | | 1 | ||
− | | Reconstruction | + | | Reconstruction tolerance for unionized energy grid (see [[Input syntax manual#set egrid|set egrid]] input option) |
|- | |- | ||
| NEUTRON_ERG_NE | | NEUTRON_ERG_NE | ||
| 1 | | 1 | ||
− | | Number of points in unionized energy grid | + | | Number of points in neutron unionized energy grid |
|- | |- | ||
| NEUTRON_EMIN | | NEUTRON_EMIN | ||
| 1 | | 1 | ||
− | | Minimum energy for neutron cross section data | + | | Minimum energy for neutron cross section data (see [[Input syntax manual#set egrid|set egrid]] input option) |
|- | |- | ||
| NEUTRON_EMAX | | NEUTRON_EMAX | ||
| 1 | | 1 | ||
− | | Maximum energy for neutron cross section data | + | | Maximum energy for neutron cross section data (see [[Input syntax manual#set egrid|set egrid]] input option) |
|- | |- | ||
|} | |} | ||
Line 526: | Line 620: | ||
|- | |- | ||
| PHOTON_ERG_NE | | PHOTON_ERG_NE | ||
− | | | + | | 1 |
− | | | + | | Number of points in photon unionized energy grid |
|- | |- | ||
| PHOTON_EMIN | | PHOTON_EMIN | ||
− | | | + | | 1 |
− | | | + | | Minimum energy for photon cross section data (see [[Input syntax manual#set egrid|set egrid]] input option) |
|- | |- | ||
| PHOTON_EMAX | | PHOTON_EMAX | ||
− | | | + | | 1 |
− | | | + | | Maximum energy for photon cross section data (see [[Input syntax manual#set egrid|set egrid]] input option) |
|- | |- | ||
|} | |} | ||
Line 548: | Line 642: | ||
| URES_DILU_CUT | | URES_DILU_CUT | ||
| 1 | | 1 | ||
− | | Density cut-off used for unresolved resonance probability table sampling | + | | Density cut-off used for unresolved resonance probability table sampling (see [[Input syntax manual#set ures|set ures]] input option) |
|- | |- | ||
| URES_EMIN | | URES_EMIN | ||
Line 564: | Line 658: | ||
| URES_USED | | URES_USED | ||
| 1 | | 1 | ||
− | | Number of nuclides for which probability table sampling was used | + | | Number of nuclides for which probability table sampling was used (see [[Input syntax manual#set ures|set ures]] input option) |
|- | |- | ||
|} | |} | ||
Line 639: | Line 733: | ||
|- | |- | ||
| IMPL_FISS_NUBAR | | IMPL_FISS_NUBAR | ||
− | | | + | | 1 |
− | | | + | | Flag indicating whether or not implicit fission nubar reaction mode is on (see [[Input syntax manual#set impl|set impl]] input option) |
|- | |- | ||
| DOPPLER_PREPROCESSOR | | DOPPLER_PREPROCESSOR | ||
Line 672: | Line 766: | ||
|- | |- | ||
| COMPTON_EKN | | COMPTON_EKN | ||
− | | | + | | 1 |
− | | | + | | Photon energy above which Klein-Nishina is used for calculating energy and direction of the scattered photons (see [[Input syntax manual#set ekn|set ekn]] input option) |
|- | |- | ||
| COMPTON_DOPPLER | | COMPTON_DOPPLER | ||
− | | | + | | 1 |
− | | | + | | Flag indicating whether or not Doppler broadening method for the energy spectrum of the scattered photons is on (see [[Input syntax manual#set cdop|set cdop]] input option) |
|- | |- | ||
| COMPTON_EANG | | COMPTON_EANG | ||
− | | | + | | 1 |
− | | | + | | Flag indicating whether or not Compton electron angular distribution model is on (see [[Input syntax manual#set cea|set cea]] input option) |
|- | |- | ||
| PHOTON_TTB | | PHOTON_TTB | ||
− | | | + | | 1 |
− | | | + | | Flag indicating whether or not thick-target bremsstrahlung approximation for modelling electrons and positrons is on (see [[Input syntax manual#set ttb|set ttb]] input option) |
+ | |- | ||
+ | |} | ||
+ | |||
+ | === Photon production === | ||
+ | |||
+ | {|class="wikitable" style="text-align: left;" | ||
+ | ! Parameter | ||
+ | ! Size | ||
+ | ! Description | ||
+ | |- | ||
+ | | PHOTON_SAMPLING_MODE | ||
+ | | 1 | ||
+ | | Flag indicating whether or not photon production from neutron reactions mode is on (see [[Input syntax manual#set ngamma|set ngamma]] input option) | ||
+ | |- | ||
+ | | PHOTON_SAMPLING_FAIL | ||
+ | | 2 | ||
+ | | Fraction of failed photon samples | ||
|- | |- | ||
|} | |} | ||
Line 726: | Line 837: | ||
=== Radioactivity data === | === Radioactivity data === | ||
+ | <u>Notes:</u> | ||
+ | |||
+ | *The values are given at the current burnup point (depletion step). | ||
{|class="wikitable" style="text-align: left;" | {|class="wikitable" style="text-align: left;" | ||
Line 733: | Line 847: | ||
|- | |- | ||
| TOT_ACTIVITY | | TOT_ACTIVITY | ||
− | | | + | | 1 |
− | | | + | | Total activity |
|- | |- | ||
| TOT_DECAY_HEAT | | TOT_DECAY_HEAT | ||
− | | | + | | 1 |
− | | | + | | Total decay heat |
|- | |- | ||
| TOT_SF_RATE | | TOT_SF_RATE | ||
− | | | + | | 1 |
− | | | + | | Total spontaneous fission rate |
|- | |- | ||
| ACTINIDE_ACTIVITY | | ACTINIDE_ACTIVITY | ||
− | | | + | | 1 |
− | | | + | | Actinide activity |
|- | |- | ||
| ACTINIDE_DECAY_HEAT | | ACTINIDE_DECAY_HEAT | ||
− | | | + | | 1 |
− | | | + | | Actinide decay heat |
|- | |- | ||
| FISSION_PRODUCT_ACTIVITY | | FISSION_PRODUCT_ACTIVITY | ||
− | | | + | | 1 |
− | | | + | | Fission product activity |
|- | |- | ||
| FISSION_PRODUCT_DECAY_HEAT | | FISSION_PRODUCT_DECAY_HEAT | ||
− | | | + | | 1 |
− | | | + | | Fission product decay heat |
|- | |- | ||
| INHALATION_TOXICITY | | INHALATION_TOXICITY | ||
− | | | + | | 1 |
− | | | + | | Total inhalation toxicity |
|- | |- | ||
| INGESTION_TOXICITY | | INGESTION_TOXICITY | ||
− | | | + | | 1 |
− | | | + | | Total ingestion toxicity |
|- | |- | ||
| ACTINIDE_INH_TOX | | ACTINIDE_INH_TOX | ||
− | | | + | | 1 |
− | | | + | | Actinide inhalation toxicity |
|- | |- | ||
| ACTINIDE_ING_TOX | | ACTINIDE_ING_TOX | ||
− | | | + | | 1 |
− | | | + | | Actinide ingestion toxicity |
|- | |- | ||
| FISSION_PRODUCT_INH_TOX | | FISSION_PRODUCT_INH_TOX | ||
− | | | + | | 1 |
− | | | + | | Fission product inhalation toxicity |
|- | |- | ||
| FISSION_PRODUCT_ING_TOX | | FISSION_PRODUCT_ING_TOX | ||
− | | | + | | 1 |
− | | | + | | Fission product ingestion toxicity |
|- | |- | ||
| SR90_ACTIVITY | | SR90_ACTIVITY | ||
− | | | + | | 1 |
− | | | + | | Sr-90 activity |
|- | |- | ||
| TE132_ACTIVITY | | TE132_ACTIVITY | ||
− | | | + | | 1 |
− | | | + | | Te-132 activity |
|- | |- | ||
| I131_ACTIVITY | | I131_ACTIVITY | ||
− | | | + | | 1 |
− | | | + | | I-131 activity |
|- | |- | ||
| I132_ACTIVITY | | I132_ACTIVITY | ||
− | | | + | | 1 |
− | | | + | | I-132 activity |
|- | |- | ||
| CS134_ACTIVITY | | CS134_ACTIVITY | ||
− | | | + | | 1 |
− | | | + | | Cs-134 activity |
|- | |- | ||
| CS137_ACTIVITY | | CS137_ACTIVITY | ||
− | | | + | | 1 |
− | | | + | | Cs-137 activity |
|- | |- | ||
| PHOTON_DECAY_SOURCE | | PHOTON_DECAY_SOURCE | ||
− | | | + | | 2 |
− | | | + | | Total photon decay source rate and total released gamma decay |
|- | |- | ||
| NEUTRON_DECAY_SOURCE | | NEUTRON_DECAY_SOURCE | ||
− | | | + | | 1 |
− | | | + | | Total neutron decay source rate |
|- | |- | ||
| ALPHA_DECAY_SOURCE | | ALPHA_DECAY_SOURCE | ||
− | | | + | | 1 |
− | | | + | | Total alpha decay source rate |
|- | |- | ||
| ELECTRON_DECAY_SOURCE | | ELECTRON_DECAY_SOURCE | ||
− | | | + | | 1 |
− | | | + | | Total beta decay source rate |
|- | |- | ||
|} | |} | ||
Line 864: | Line 978: | ||
| BURNUP | | BURNUP | ||
| 2 | | 2 | ||
− | | Burnup at the current step (in MWd/kgU): cumulative and | + | | Burnup at the current step (in MWd/kgU): cumulative calculated from the [[Input_syntax_manual#dep_.28depletion_history.29|depletion history input]] and cumulative realized from the actual calculation. |
|- | |- | ||
| BURN_DAYS | | BURN_DAYS | ||
Line 873: | Line 987: | ||
| 3 | | 3 | ||
| Number of fissions per initial fissile atom at the current step: relative step-wise, increment step-wise, final step-wise. | | Number of fissions per initial fissile atom at the current step: relative step-wise, increment step-wise, final step-wise. | ||
+ | |- | ||
+ | |} | ||
+ | |||
+ | === Coefficient calculation === | ||
+ | |||
+ | {|class="wikitable" style="text-align: left;" | ||
+ | ! Parameter | ||
+ | ! Size | ||
+ | ! Description | ||
+ | |- | ||
+ | | COEF_IDX | ||
+ | | 2 | ||
+ | | Coefficient index when output is printed and total number of coefficient calculations | ||
+ | |- | ||
+ | | COEF_BRANCH | ||
+ | | 1 | ||
+ | | Branch index within coefficient calculation when output is printed | ||
+ | |- | ||
+ | | COEF_BU_STEP | ||
+ | | 1 | ||
+ | | Burnup step at the given coefficient calculation when output is printed | ||
|- | |- | ||
|} | |} | ||
Line 884: | Line 1,019: | ||
|- | |- | ||
| CONVERSION_RATIO | | CONVERSION_RATIO | ||
− | | | + | | 2 |
− | | | + | | Analog estimate of conversion rate, ratio between fissile production and loss rate |
+ | |- | ||
+ | | TH232_FISS | ||
+ | | 4 | ||
+ | | Analog estimate of Th-232 fission rate (total/fraction) | ||
+ | |- | ||
+ | | U233_FISS | ||
+ | | 4 | ||
+ | | Analog estimate of U-233 fission rate (total/fraction) | ||
|- | |- | ||
| U235_FISS | | U235_FISS | ||
− | | | + | | 4 |
− | | | + | | Analog estimate of U-235 fission rate (total/fraction) |
|- | |- | ||
| U238_FISS | | U238_FISS | ||
− | | | + | | 4 |
− | | | + | | Analog estimate of U-238 fission rate (total/fraction) |
+ | |- | ||
+ | | PU239_FISS | ||
+ | | 4 | ||
+ | | Analog estimate of Pu-239 fission rate (total/fraction) | ||
+ | |- | ||
+ | | PU240_FISS | ||
+ | | 4 | ||
+ | | Analog estimate of Pu-240 fission rate (total/fraction) | ||
+ | |- | ||
+ | | PU241_FISS | ||
+ | | 4 | ||
+ | | Analog estimate of Pu-241 fission rate (total/fraction) | ||
+ | |- | ||
+ | | TH232_CAPT | ||
+ | | 4 | ||
+ | | Analog estimate of Th-232 capture rate (total/fraction) | ||
+ | |- | ||
+ | | U233_CAPT | ||
+ | | 4 | ||
+ | | Analog estimate of U-233 capture rate (total/fraction) | ||
|- | |- | ||
| U235_CAPT | | U235_CAPT | ||
− | | | + | | 4 |
− | | | + | | Analog estimate of U-235 capture rate (total/fraction) |
|- | |- | ||
| U238_CAPT | | U238_CAPT | ||
− | | | + | | 4 |
− | | | + | | Analog estimate of U-238 capture rate (total/fraction) |
+ | |- | ||
+ | | PU239_CAPT | ||
+ | | 4 | ||
+ | | Analog estimate of Pu-239 capture rate (total/fraction) | ||
+ | |- | ||
+ | | PU240_CAPT | ||
+ | | 4 | ||
+ | | Analog estimate of Pu-240 capture rate (total/fraction) | ||
+ | |- | ||
+ | | PU241_CAPT | ||
+ | | 4 | ||
+ | | Analog estimate of Pu-241 capture rate (total/fraction) | ||
|- | |- | ||
| XE135_CAPT | | XE135_CAPT | ||
− | | | + | | 4 |
− | | | + | | Analog estimate of Xe-135 capture rate (total/fraction) |
|- | |- | ||
+ | | XE135M_CAPT | ||
+ | | 4 | ||
+ | | Analog estimate of Xe-135m capture rate (total/fraction) | ||
+ | |- | ||
+ | | SM149_CAPT | ||
+ | | 4 | ||
+ | | Analog estimate of Sm-149 capture rate (total/fraction) | ||
|} | |} | ||
Line 919: | Line 1,101: | ||
|- | |- | ||
| BALA_SRC_NEUTRON_SRC | | BALA_SRC_NEUTRON_SRC | ||
− | | | + | | 1/1 |
− | | | + | | Neutron produced by external source |
|- | |- | ||
| BALA_SRC_NEUTRON_FISS | | BALA_SRC_NEUTRON_FISS | ||
− | | | + | | 1/1 |
− | | | + | | Neutron produced by fission |
|- | |- | ||
| BALA_SRC_NEUTRON_NXN | | BALA_SRC_NEUTRON_NXN | ||
− | | | + | | 1/1 |
− | | | + | | Neutron produced by scattering |
|- | |- | ||
| BALA_SRC_NEUTRON_VR | | BALA_SRC_NEUTRON_VR | ||
− | | | + | | 1/1 |
− | | | + | | Neutron produced by variance reduction (Russian roulette, splitting) |
|- | |- | ||
| BALA_SRC_NEUTRON_TOT | | BALA_SRC_NEUTRON_TOT | ||
− | | | + | | 1/1 |
− | | | + | | Total neutron produced |
|- | |- | ||
| BALA_LOSS_NEUTRON_CAPT | | BALA_LOSS_NEUTRON_CAPT | ||
− | | | + | | 1/1 |
− | | | + | | Neutron lost by capture |
|- | |- | ||
| BALA_LOSS_NEUTRON_FISS | | BALA_LOSS_NEUTRON_FISS | ||
− | | | + | | 1/1 |
− | | | + | | Neutron lost by fission |
|- | |- | ||
| BALA_LOSS_NEUTRON_LEAK | | BALA_LOSS_NEUTRON_LEAK | ||
− | | | + | | 1/1 |
− | | | + | | Neutron lost by leakage |
|- | |- | ||
| BALA_LOSS_NEUTRON_CUT | | BALA_LOSS_NEUTRON_CUT | ||
− | | | + | | 1/1 |
− | | | + | | Neutron lost by cut-off |
|- | |- | ||
| BALA_LOSS_NEUTRON_ERR | | BALA_LOSS_NEUTRON_ERR | ||
− | | | + | | 1/1 |
− | | | + | | Neutron lost by failed tracking |
|- | |- | ||
| BALA_LOSS_NEUTRON_TOT | | BALA_LOSS_NEUTRON_TOT | ||
− | | | + | | 1/1 |
− | | | + | | Total neutron lost |
|- | |- | ||
| BALA_NEUTRON_DIFF | | BALA_NEUTRON_DIFF | ||
− | | | + | | 1/1 |
− | | | + | | Difference between total neutron produced and lost |
+ | |- | ||
+ | |} | ||
+ | |||
+ | === Photon balance (particles/weight/energy-weighted) === | ||
+ | |||
+ | {|class="wikitable" style="text-align: left;" | ||
+ | ! Parameter | ||
+ | ! Size | ||
+ | ! Description | ||
+ | |- | ||
+ | | BALA_SRC_PHOTON_SRC | ||
+ | | 1/1/1 | ||
+ | | Photon produced by external source | ||
+ | |- | ||
+ | | BALA_SRC_PHOTON_TTB | ||
+ | | 1/1/1 | ||
+ | | Photon produced by bremsstrahlung | ||
+ | |- | ||
+ | | BALA_SRC_PHOTON_ANNIH | ||
+ | | 1/1/1 | ||
+ | | Photon produced by annihilation | ||
+ | |- | ||
+ | | BALA_SRC_PHOTON_FLUOR | ||
+ | | 1/1/1 | ||
+ | | Photon produced by fluorescence | ||
+ | |- | ||
+ | | BALA_SRC_PHOTON_NREA | ||
+ | | 1/1/1 | ||
+ | | Photon produced by neutron reaction | ||
+ | |- | ||
+ | | BALA_SRC_PHOTON_VR | ||
+ | | 1/1/1 | ||
+ | | Photon produced by variance reduction (Russian roulette, splitting) | ||
+ | |- | ||
+ | | BALA_SRC_PHOTON_TOT | ||
+ | | 1/1/1 | ||
+ | | Total photon produced | ||
+ | |- | ||
+ | | BALA_LOSS_PHOTON_CAPT | ||
+ | | 1/1 | ||
+ | | Photon lost by capture | ||
+ | |- | ||
+ | | BALA_LOSS_PHOTON_LEAK | ||
+ | | 1/1 | ||
+ | | Photon lost by leakage | ||
+ | |- | ||
+ | | BALA_LOSS_PHOTON_CUT | ||
+ | | 1/1 | ||
+ | | Photon lost by cut-off | ||
+ | |- | ||
+ | | BALA_LOSS_PHOTON_ERR | ||
+ | | 1/1 | ||
+ | | Photon lost by failed tracking | ||
+ | |- | ||
+ | | BALA_LOSS_PHOTON_TOT | ||
+ | | 1/1 | ||
+ | | Total photon lost | ||
+ | |- | ||
+ | | BALA_PHOTON_DIFF | ||
+ | | 1/1 | ||
+ | | Difference between total photon produced and lost | ||
|- | |- | ||
|} | |} | ||
Line 971: | Line 1,214: | ||
=== Normalized total reaction rates (neutrons) === | === Normalized total reaction rates (neutrons) === | ||
+ | <u>Notes:</u> | ||
+ | |||
+ | *In burnup calculations the values correspond to total, burnable and non-burnable rates | ||
{|class="wikitable" style="text-align: left;" | {|class="wikitable" style="text-align: left;" | ||
Line 978: | Line 1,224: | ||
|- | |- | ||
| TOT_POWER | | TOT_POWER | ||
− | | | + | | 2(6) |
− | | | + | | Total neutron fission power |
|- | |- | ||
| TOT_POWDENS | | TOT_POWDENS | ||
− | | | + | | 2(6) |
− | | | + | | Total neutron fission power density |
|- | |- | ||
| TOT_GENRATE | | TOT_GENRATE | ||
− | | | + | | 2(6) |
− | | | + | | Total neutron generation rate |
|- | |- | ||
| TOT_FISSRATE | | TOT_FISSRATE | ||
− | | | + | | 2(6) |
− | | | + | | Total neutron fission rate |
|- | |- | ||
| TOT_CAPTRATE | | TOT_CAPTRATE | ||
− | | | + | | 2(6) |
− | | | + | | Total neutron capture rate |
|- | |- | ||
| TOT_ABSRATE | | TOT_ABSRATE | ||
− | | | + | | 2(6) |
− | | | + | | Total neutron absorption rate |
|- | |- | ||
| TOT_SRCRATE | | TOT_SRCRATE | ||
− | | | + | | 2(6) |
− | | | + | | Total neutron source rate |
|- | |- | ||
| TOT_FLUX | | TOT_FLUX | ||
− | | | + | | 2(6) |
− | | | + | | Total neutron flux |
|- | |- | ||
| TOT_PHOTON_PRODRATE | | TOT_PHOTON_PRODRATE | ||
− | | | + | | 4 |
− | | | + | | Total neutron-photon production rate (implicit/analog) |
|- | |- | ||
| TOT_LEAKRATE | | TOT_LEAKRATE | ||
− | | | + | | 2 |
− | | | + | | Total neutron leakage rate |
|- | |- | ||
| ALBEDO_LEAKRATE | | ALBEDO_LEAKRATE | ||
− | | | + | | 2 |
− | | | + | | Albedo neutron leakage rate |
|- | |- | ||
| TOT_LOSSRATE | | TOT_LOSSRATE | ||
− | | | + | | 2 |
− | | | + | | Total neutron loss rate |
|- | |- | ||
| TOT_CUTRATE | | TOT_CUTRATE | ||
− | | | + | | 2 |
− | | | + | | Total neutron energy cut-off rate |
|- | |- | ||
| TOT_RR | | TOT_RR | ||
− | | | + | | 2 |
− | | | + | | Total neutron reaction rate |
|- | |- | ||
| TOT_XE135_ABSRATE | | TOT_XE135_ABSRATE | ||
− | | | + | | 2 |
− | | | + | | Total neutron absorption rate in Xe-135 |
+ | |- | ||
+ | | TOT_XE135M_ABSRATE | ||
+ | | 2 | ||
+ | | Total neutron absorption rate in Xe-135m | ||
|- | |- | ||
| TOT_SM149_ABSRATE | | TOT_SM149_ABSRATE | ||
− | | | + | | 2 |
− | | | + | | Total neutron absorption rate in Sm-149 |
|- | |- | ||
| INI_FMASS | | INI_FMASS | ||
− | | | + | | 1 |
− | | | + | | Initial fissile mass |
|- | |- | ||
| TOT_FMASS | | TOT_FMASS | ||
− | | | + | | 1 |
− | | | + | | Total fissile mass |
|- | |- | ||
| INI_BURN_FMASS | | INI_BURN_FMASS | ||
− | | | + | | 1 |
− | | | + | | Initial fissile mass in burnable materials |
|- | |- | ||
| TOT_BURN_FMASS | | TOT_BURN_FMASS | ||
− | | | + | | 1 |
− | | | + | | Total fissile mass in burnable materials |
+ | |- | ||
+ | |} | ||
+ | |||
+ | === Normalized total reaction rates (photons) === | ||
+ | |||
+ | {|class="wikitable" style="text-align: left;" | ||
+ | ! Parameter | ||
+ | ! Size | ||
+ | ! Description | ||
+ | |- | ||
+ | | TOT_PHOTON_LEAKRATE | ||
+ | | 2 | ||
+ | | Total photon leakage rate | ||
+ | |- | ||
+ | | TOT_PHOTON_CUTRATE | ||
+ | | 2 | ||
+ | | Total photon energy cut-off rate | ||
+ | |- | ||
+ | | PHOTOELE_CAPT_RATE | ||
+ | | 2 | ||
+ | | Photo-electric photon capture rate | ||
+ | |- | ||
+ | | PAIRPROD_CAPT_RATE | ||
+ | | 2 | ||
+ | |Pair production photon capture rate | ||
+ | |- | ||
+ | | TOT_PHOTON_LOSSRATE | ||
+ | | 2 | ||
+ | | Total photon loss rate | ||
+ | |- | ||
+ | | TOT_PHOTON_SRCRATE | ||
+ | | 2 | ||
+ | | Total photon source rate | ||
+ | |- | ||
+ | | TOT_PHOTON_RR | ||
+ | | 2 | ||
+ | | Total photon reaction rate | ||
+ | |- | ||
+ | | TOT_PHOTON_FLUX | ||
+ | | 2 | ||
+ | | Total photon flux | ||
+ | |- | ||
+ | | TOT_PHOTON_HEATRATE | ||
+ | | 2 | ||
+ | | Total photon heating rate | ||
|- | |- | ||
|} | |} | ||
Line 1,065: | Line 1,360: | ||
! Size | ! Size | ||
! Description | ! Description | ||
+ | |- | ||
+ | | I135_EQUIL_CONC | ||
+ | | 2 | ||
+ | | Averaged equilibrium I-135 concentration (see [[Input syntax manual#set xenon|set xenon]] input option) | ||
|- | |- | ||
| XE135_EQUIL_CONC | | XE135_EQUIL_CONC | ||
Line 1,070: | Line 1,369: | ||
| Averaged equilibrium Xe-135 concentration (see [[Input syntax manual#set xenon|set xenon]] input option) | | Averaged equilibrium Xe-135 concentration (see [[Input syntax manual#set xenon|set xenon]] input option) | ||
|- | |- | ||
− | | | + | | XE135M_EQUIL_CONC |
| 2 | | 2 | ||
− | | Averaged equilibrium | + | | Averaged equilibrium Xe-135m concentration (see [[Input syntax manual#set xenon|set xenon]] input option) |
|- | |- | ||
|} | |} | ||
Line 1,090: | Line 1,389: | ||
| 2 | | 2 | ||
| Averaged equilibrium Pm-149 concentration (see [[Input syntax manual#set samarium|set samarium]] input option) | | Averaged equilibrium Pm-149 concentration (see [[Input syntax manual#set samarium|set samarium]] input option) | ||
+ | |- | ||
+ | |} | ||
+ | |||
+ | === Iteration factor === | ||
+ | |||
+ | {|class="wikitable" style="text-align: left;" | ||
+ | ! Parameter | ||
+ | ! Size | ||
+ | ! Description | ||
+ | |- | ||
+ | | ITER_FACTOR | ||
+ | | 2 | ||
+ | | Iteration factor of critical density iteration (see [[Input syntax manual#set iter nuc|set iter nuc]] input option) or albedo iteration (see [[Input syntax manual#set iter alb|set iter alb]] input option) | ||
|- | |- | ||
|} | |} | ||
=== Six-factor formula === | === Six-factor formula === | ||
+ | <u>Notes:</u> | ||
+ | |||
+ | * The six-factor formula estimates are printed by default. Check the suitability of the approximation for the given spectrum (the formulation is based on a thermal spectrum). | ||
{|class="wikitable" style="text-align: left;" | {|class="wikitable" style="text-align: left;" | ||
Line 1,142: | Line 1,457: | ||
|- | |- | ||
| NUBAR | | NUBAR | ||
− | | | + | | 2 |
− | | | + | | Average fission neutron yield |
|- | |- | ||
| FISSE | | FISSE | ||
− | | | + | | 2 |
− | | | + | | Average fission energy production |
|- | |- | ||
|} | |} | ||
− | === Criticality eigenvalues === | + | === Criticality eigenvalues / Multiplication factor external source === |
{|class="wikitable" style="text-align: left;" | {|class="wikitable" style="text-align: left;" | ||
Line 1,159: | Line 1,474: | ||
|- | |- | ||
| ANA_KEFF | | ANA_KEFF | ||
− | | 6 | + | | 6(2) |
| Analog estimate of ''k''<sub>eff</sub>: total, prompt and delayed neutron contribution. | | Analog estimate of ''k''<sub>eff</sub>: total, prompt and delayed neutron contribution. | ||
|- | |- | ||
Line 1,177: | Line 1,492: | ||
| 2 | | 2 | ||
| Absorption estimate of ''k''<sub>inf</sub>. | | Absorption estimate of ''k''<sub>inf</sub>. | ||
+ | |- | ||
+ | | ANA_EXT_K | ||
+ | | 20 | ||
+ | | Generation-wise source multiplication factors in external source mode | ||
+ | |- | ||
+ | | SRC_MULT | ||
+ | | 2 | ||
+ | | Source multiplication factor in external source mode | ||
+ | |- | ||
+ | | MEAN_NGEN | ||
+ | | 2 | ||
+ | | Mean number of generations in external source mode | ||
+ | |- | ||
+ | | PROMPT_GEN_POP | ||
+ | | ''N<sub>G</sub>'' | ||
+ | | Prompt fission population generation fraction | ||
+ | |- | ||
+ | | PROMPT_GEN_CUMU | ||
+ | | ''N<sub>G</sub>'' | ||
+ | | Prompt fission cumulative generation fraction | ||
+ | |- | ||
+ | | PROMPT_GEN_TIMES | ||
+ | | ''N<sub>G</sub>'' | ||
+ | | Prompt fission time generation fraction | ||
+ | |- | ||
+ | | PROMPT_CHAIN_LENGTH | ||
+ | | 2 | ||
+ | | Mean prompt chain length in external source mode | ||
|- | |- | ||
| GEOM_ALBEDO | | GEOM_ALBEDO | ||
| 6 | | 6 | ||
| Fixed or iterated value for albedo boundary condition for x-,y- and z-directions (see [[Input syntax manual#set bc|set bc]] or [[Input syntax manual#set iter alb|set iter alb]] input options). | | Fixed or iterated value for albedo boundary condition for x-,y- and z-directions (see [[Input syntax manual#set bc|set bc]] or [[Input syntax manual#set iter alb|set iter alb]] input options). | ||
+ | |- | ||
+ | |} | ||
+ | |||
+ | === Wielandt method === | ||
+ | |||
+ | {|class="wikitable" style="text-align: left;" | ||
+ | ! Parameter | ||
+ | ! Size | ||
+ | ! Description | ||
+ | |- | ||
+ | | WIELANDT_K | ||
+ | | 2 | ||
+ | | Wielandt’s method shifted eigenvalue (see [[Input syntax manual#set wie|set wie]] input option) | ||
+ | |- | ||
+ | | WIELANDT_P | ||
+ | | 2 | ||
+ | | Wielandt’s method neutron banking probability (see [[Input syntax manual#set wie|set wie]] input option) | ||
|- | |- | ||
|} | |} | ||
Line 1,237: | Line 1,597: | ||
== Time constants == | == Time constants == | ||
− | === Forward | + | === Forward delayed neutron parameters === |
{|class="wikitable" style="text-align: left;" | {|class="wikitable" style="text-align: left;" | ||
Line 1,250: | Line 1,610: | ||
| FWD_ANA_BETA_ZERO | | FWD_ANA_BETA_ZERO | ||
| 2''D'' + 2 | | 2''D'' + 2 | ||
− | | Analog estimator of physical delayed neutron fractions (number of delayed neutrons emitted in fission): total | + | | Analog estimator of physical delayed neutron fractions (number of delayed neutrons emitted in fission): total, group-wise |
|- | |- | ||
| FWD_ANA_LAMBDA | | FWD_ANA_LAMBDA | ||
| 2''D'' + 2 | | 2''D'' + 2 | ||
− | | Analog estimator of delayed neutron precursor decay constants: total | + | | Analog estimator of delayed neutron precursor decay constants: total, group-wise |
|- | |- | ||
|} | |} | ||
− | === | + | === Adjoint-weighted time constants using Meulekamp's method === |
{|class="wikitable" style="text-align: left;" | {|class="wikitable" style="text-align: left;" | ||
Line 1,267: | Line 1,627: | ||
| ADJ_MEULEKAMP_BETA_EFF | | ADJ_MEULEKAMP_BETA_EFF | ||
| 2''D'' + 2 | | 2''D'' + 2 | ||
− | | Adjoint-weighted effective delayed neutron fractions using Meulekamp's method: total | + | | Adjoint-weighted effective delayed neutron fractions using Meulekamp's method: total, group-wise |
|- | |- | ||
| ADJ_MEULEKAMP_LAMBDA | | ADJ_MEULEKAMP_LAMBDA | ||
| 2''D'' + 2 | | 2''D'' + 2 | ||
− | | Adjoint-weighted of delayed neutron precursor decay constants using Meulekamp's method: total | + | | Adjoint-weighted of delayed neutron precursor decay constants using Meulekamp's method: total, group-wise |
|- | |- | ||
|} | |} | ||
Line 1,288: | Line 1,648: | ||
| ADJ_NAUCHI_GEN_TIME | | ADJ_NAUCHI_GEN_TIME | ||
| 6 | | 6 | ||
− | | Adjoint-weighted neutron generation times using Nauchi's method: total, prompt | + | | Adjoint-weighted neutron generation times using Nauchi's method: total, prompt, delayed |
|- | |- | ||
| ADJ_NAUCHI_LIFETIME | | ADJ_NAUCHI_LIFETIME | ||
| 6 | | 6 | ||
− | | Adjoint-weighted neutron lifetimes using Nauchi's method: total, prompt | + | | Adjoint-weighted neutron lifetimes using Nauchi's method: total, prompt, delayed. |
|- | |- | ||
| ADJ_NAUCHI_BETA_EFF | | ADJ_NAUCHI_BETA_EFF | ||
| 2''D'' + 2 | | 2''D'' + 2 | ||
− | | Adjoint-weighted effective delayed neutron fractions using Nauchi's method: total | + | | Adjoint-weighted effective delayed neutron fractions using Nauchi's method: total, group-wise |
|- | |- | ||
| ADJ_NAUCHI_LAMBDA | | ADJ_NAUCHI_LAMBDA | ||
| 2''D'' + 2 | | 2''D'' + 2 | ||
− | | Adjoint-weighed of delayed neutron precursor decay constants using Nauchi's method: total | + | | Adjoint-weighed of delayed neutron precursor decay constants using Nauchi's method: total, group-wise |
|- | |- | ||
|} | |} | ||
Line 1,313: | Line 1,673: | ||
| ADJ_IFP_GEN_TIME | | ADJ_IFP_GEN_TIME | ||
| 6 | | 6 | ||
− | | Adjoint-weighted neutron generation times using the iterated fission probability method: total, prompt | + | | Adjoint-weighted neutron generation times using the iterated fission probability method: total, prompt, delayed |
|- | |- | ||
| ADJ_IFP_LIFETIME | | ADJ_IFP_LIFETIME | ||
| 6 | | 6 | ||
− | | Adjoint-weighted neutron lifetimes using the iterated fission probability method: total, prompt | + | | Adjoint-weighted neutron lifetimes using the iterated fission probability method: total, prompt, delayed |
|- | |- | ||
| ADJ_IFP_IMP_BETA_EFF | | ADJ_IFP_IMP_BETA_EFF | ||
| 2''D'' + 2 | | 2''D'' + 2 | ||
− | | Implicit estimator of adjoint-weighted effective delayed neutron fractions using the iterated fission probability method: total | + | | Implicit estimator of adjoint-weighted effective delayed neutron fractions using the iterated fission probability method: total, group-wise |
|- | |- | ||
| ADJ_IFP_IMP_LAMBDA | | ADJ_IFP_IMP_LAMBDA | ||
| 2''D'' + 2 | | 2''D'' + 2 | ||
− | | Implicit estimator of adjoint-weighted of delayed neutron precursor decay constants using the iterated fission probability method: total | + | | Implicit estimator of adjoint-weighted of delayed neutron precursor decay constants using the iterated fission probability method: total, group-wise |
|- | |- | ||
| ADJ_IFP_ANA_BETA_EFF | | ADJ_IFP_ANA_BETA_EFF | ||
| 2''D'' + 2 | | 2''D'' + 2 | ||
− | | Analog estimator of adjoint-weighted effective delayed neutron fractions using the iterated fission probability method: total | + | | Analog estimator of adjoint-weighted effective delayed neutron fractions using the iterated fission probability method: total, group-wise |
|- | |- | ||
| ADJ_IFP_ANA_LAMBDA | | ADJ_IFP_ANA_LAMBDA | ||
| 2''D'' + 2 | | 2''D'' + 2 | ||
− | | Analog estimator of adjoint-weighted of delayed neutron precursor decay constants using the iterated fission probability method: total | + | | Analog estimator of adjoint-weighted of delayed neutron precursor decay constants using the iterated fission probability method: total, group-wise |
|- | |- | ||
| ADJ_IFP_ROSSI_ALPHA | | ADJ_IFP_ROSSI_ALPHA | ||
Line 1,388: | Line 1,748: | ||
| ANA_SLOW_TIME | | ANA_SLOW_TIME | ||
| 6 | | 6 | ||
− | | Analog estimate of slowing-down time: total, prompt | + | | Analog estimate of slowing-down time: total, prompt, delayed |
|- | |- | ||
| ANA_THERM_TIME | | ANA_THERM_TIME | ||
| 6 | | 6 | ||
− | | Analog estimate of thermal neutron lifetime: total, prompt | + | | Analog estimate of thermal neutron lifetime: total, prompt, delayed |
|- | |- | ||
| ANA_THERM_FRAC | | ANA_THERM_FRAC | ||
| 6 | | 6 | ||
− | | Analog estimate of neutron thermalisation fraction: total, prompt | + | | Analog estimate of neutron thermalisation fraction: total, prompt, delayed |
|- | |- | ||
| ANA_DELAYED_EMTIME | | ANA_DELAYED_EMTIME | ||
Line 1,405: | Line 1,765: | ||
| 4 | | 4 | ||
| Analog estimate of average number of collisions per history: total and to fission | | Analog estimate of average number of collisions per history: total and to fission | ||
+ | |- | ||
+ | |} | ||
+ | |||
+ | === Dynamic simulation === | ||
+ | |||
+ | {|class="wikitable" style="text-align: left;" | ||
+ | ! Parameter | ||
+ | ! Size | ||
+ | ! Description | ||
+ | |- | ||
+ | | DYN_NB | ||
+ | | 1 | ||
+ | | Number of time intervals defined in the time-bin structure (see [[Input syntax manual#tme (time binning definition)|tme]] card) | ||
+ | |- | ||
+ | | DYN_TMIN | ||
+ | | 1 | ||
+ | | Minimum time boundary defined in the time-bin structure (see [[Input syntax manual#tme (time binning definition)|tme]] card) | ||
+ | |- | ||
+ | | DYN_TMAX | ||
+ | | 1 | ||
+ | | Maximum time boundary defined in the time-bin structure (see [[Input syntax manual#tme (time binning definition)|tme]] card) | ||
+ | |- | ||
+ | | DYN_TIMES | ||
+ | | 2''T'' + 2 | ||
+ | | Time-bin boundaries defined in the time-bin structure (see [[Input syntax manual#tme (time binning definition)|tme]] card) | ||
+ | |- | ||
+ | | DYN_POP | ||
+ | | 2''T'' | ||
+ | | Neutron population at the end of the time interval | ||
+ | |- | ||
+ | | DYN_PERIOD | ||
+ | | 2''T'' | ||
+ | | Reactor period based on increase/decrease of neutron population during time interval | ||
+ | |- | ||
+ | |} | ||
+ | |||
+ | === Analog mean photon lifetime === | ||
+ | |||
+ | {|class="wikitable" style="text-align: left;" | ||
+ | ! Parameter | ||
+ | ! Size | ||
+ | ! Description | ||
+ | |- | ||
+ | | ANA_LIFETIME | ||
+ | | 2 | ||
+ | | Analog estimator of photon lifetime | ||
|- | |- | ||
|} | |} | ||
Line 1,413: | Line 1,819: | ||
*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. | *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, [ | + | *The methodology used in Serpent for spatial homogenization is described in a paper published in Annals of Nuclear Energy in 2016.<ref>Leppänen, J., Pusa, M. and Fridman, E. ''"Overview of methodology for spatial homogenization in the Serpent 2 Monte Carlo code."'' Ann. Nucl. Energy, [https://doi.org/10.1016/j.anucene.2016.06.007 96 (2016) 126-136].</ref> |
*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 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 intermediate multi-group structure is defined using option [[Input syntax manual#set micro|set micro]] or [[Input syntax manual#set micro|set fum]]. | ||
Line 1,623: | Line 2,029: | ||
*Scattering production includes multiplying (n,2n), (n,3n), etc. reactions. | *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 order of values (<tt>[input].coe</tt>) or value pairs (<tt>[input]_res.m</tt>) is: <math>\Sigma_{1,1} \, \Sigma_{1,2} \, ... \, \Sigma_{2,1} \, \Sigma_{2,2} \, ...</math> where <math>\Sigma_{g,g'}</math> refers to scattering from group ''g'' to ''g'''. |
− | *The data in the [input]_res.m file can be read into a G | + | *The data in the <tt>[input]_res.m</tt> file can be read into a ''G'' × ''G'' matrix with Matlab reshape-command, for example: |
<nowiki> | <nowiki> | ||
reshape(INF_S0(idx,1:2:end), G, G);</nowiki> | reshape(INF_S0(idx,1:2:end), G, G);</nowiki> | ||
Line 1,703: | Line 2,109: | ||
<u>Notes:</u> | <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]]. | *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 | + | *The cumulative migration method <ref>Liu, Z., Smith, K., Forget, B. and Ortensi, J.''"Cumulative migration method for computing rigorous diffusion coefficients and transport cross sections from Monte Carlo."'' Ann. 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. | *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 TRC_TRANSPXS and TRC_DIFFCOEF requires defining energy-dependent correction factors using the [[Input syntax manual#set trc|set trc]] option. | ||
Line 1,768: | Line 2,174: | ||
*Printed only if poison cross section option is on (see [[Input syntax manual#set poi|set poi]]). | *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]]). | *Xe-135m values printed only if separate treatment of Xe-135m is on (see [[Input syntax manual#set poi|set poi]]). | ||
+ | *Approximate effective treatment of Pm-149 production from Pm-148 and Pm-148m is not included in Pm-149 fission yield. | ||
+ | *All values for I-135, Pm-147, Pm-148 and Pm-148m not present in all versions. | ||
{|class="wikitable" style="text-align: left;" | {|class="wikitable" style="text-align: left;" | ||
Line 1,805: | Line 2,213: | ||
| 2''G'' | | 2''G'' | ||
| Microscopic absorption cross section of Xe-135m | | Microscopic absorption cross section of Xe-135m | ||
+ | |- | ||
+ | | INF_PM147_MICRO_ABS | ||
+ | | 2''G'' | ||
+ | | Microscopic absorption cross section of Pm-147 | ||
+ | |- | ||
+ | | INF_PM148_MICRO_ABS | ||
+ | | 2''G'' | ||
+ | | Microscopic absorption cross section of Pm-148 | ||
+ | |- | ||
+ | | INF_PM148M_MICRO_ABS | ||
+ | | 2''G'' | ||
+ | | Microscopic absorption cross section of Pm-148m | ||
|- | |- | ||
| INF_PM149_MICRO_ABS | | INF_PM149_MICRO_ABS | ||
Line 1,813: | Line 2,233: | ||
| 2''G'' | | 2''G'' | ||
| Microscopic absorption cross section of Sm-149 | | Microscopic absorption cross section of Sm-149 | ||
+ | |- | ||
+ | | INF_I135_MACRO_ABS | ||
+ | | 2''G'' | ||
+ | | Macroscopic absorption cross section of I-135 | ||
|- | |- | ||
| INF_XE135_MACRO_ABS | | INF_XE135_MACRO_ABS | ||
Line 1,821: | Line 2,245: | ||
| 2''G'' | | 2''G'' | ||
| Macroscopic absorption cross section of Xe-135m | | Macroscopic absorption cross section of Xe-135m | ||
+ | |- | ||
+ | | INF_PM147_MACRO_ABS | ||
+ | | 2''G'' | ||
+ | | Macroscopic absorption cross section of Pm-147 | ||
+ | |- | ||
+ | | INF_PM148_MACRO_ABS | ||
+ | | 2''G'' | ||
+ | | Macroscopic absorption cross section of Pm-148 | ||
+ | |- | ||
+ | | INF_PM148M_MACRO_ABS | ||
+ | | 2''G'' | ||
+ | | Macroscopic absorption cross section of Pm-148M | ||
+ | |- | ||
+ | | INF_PM149_MACRO_ABS | ||
+ | | 2''G'' | ||
+ | | Macroscopic absorption cross section of Pm-149 | ||
|- | |- | ||
| INF_SM149_MACRO_ABS | | INF_SM149_MACRO_ABS | ||
Line 1,827: | Line 2,267: | ||
|- | |- | ||
|} | |} | ||
+ | |||
+ | ==== Poison universe-averaged densities ==== | ||
+ | |||
+ | <u>Notes:</u> | ||
+ | |||
+ | *The universe-averaged atomic density is defined to be such that ADENS*MICRO_ABS is equal to MACRO_ABS.<ref>Rintala, A., Valtavirta, V. and Leppänen, J.. ''Microscopic cross section calculation methodology in the Serpent 2 Monte Carlo code.'' Annals of Nuclear Energy, [https://doi.org/10.1016/j.anucene.2021.108603 164 (2021): 108603].</ref> | ||
+ | |||
+ | {|class="wikitable" style="text-align: left;" | ||
+ | ! Parameter | ||
+ | ! Size | ||
+ | ! Description | ||
+ | |- | ||
+ | | I135_ADENS | ||
+ | | 2 | ||
+ | | Universe-averaged atomic density of I-135 | ||
+ | |- | ||
+ | | XE135_ADENS | ||
+ | | 2 | ||
+ | | Universe-averaged atomic density of Xe-135 | ||
+ | |- | ||
+ | | XE135M_ADENS | ||
+ | | 2 | ||
+ | | Universe-averaged atomic density of Xe-135m | ||
+ | |- | ||
+ | | PM147_ADENS | ||
+ | | 2 | ||
+ | | Universe-averaged atomic density of Pm-147 | ||
+ | |- | ||
+ | | PM148_ADENS | ||
+ | | 2 | ||
+ | | Universe-averaged atomic density of Pm-147 | ||
+ | |- | ||
+ | | PM148M_ADENS | ||
+ | | 2 | ||
+ | | Universe-averaged atomic density of Pm-148m | ||
+ | |- | ||
+ | | PM149_ADENS | ||
+ | | 2 | ||
+ | | Universe-averaged atomic density of Pm-149 | ||
+ | |- | ||
+ | |} | ||
+ | |||
==== Poison decay constants ==== | ==== Poison decay constants ==== | ||
Line 2,054: | Line 2,536: | ||
*Scattering production includes multiplying (n,2n), (n,3n), etc. reactions. | *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 order of values (<tt>[input].coe</tt>) or value pairs (<tt>[input]_res.m</tt>) is: <math>\Sigma_{1,1} \, \Sigma_{1,2} \, ... \, \Sigma_{2,1} \, \Sigma_{2,2} \, ...</math> where <math>\Sigma_{g,g'}</math> refers to scattering from group ''g'' to ''g'''. |
− | *The data in the _res.m file can be read into a G | + | *The data in the <tt>[input]_res.m</tt> file can be read into a ''G'' × ''G'' matrix with Matlab reshape-command, for example: |
<nowiki> | <nowiki> | ||
reshape(B1_S0(idx,1:2:end), G, G)</nowiki>. | reshape(B1_S0(idx,1:2:end), G, G)</nowiki>. | ||
Line 2,218: | Line 2,700: | ||
*The output consists of total, followed by ''D'' precursor group-wise values. In earlier versions, the output was fixed to 9 values independently of the library in use, with zero values corresponding to the empty precursor groups in the library. | *The output consists of total, followed by ''D'' precursor group-wise values. In earlier versions, the output was fixed to 9 values independently of the library in use, with zero values corresponding to the empty precursor groups in the library. | ||
− | *The actual number of groups depends on the cross section library used in the calculations. JEFF-3.1, JEFF | + | *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 2,261: | Line 2,743: | ||
| (string) | | (string) | ||
| Name of the surface used for the calculation | | Name of the surface used for the calculation | ||
+ | |- | ||
+ | | DF_SURFACE_TYPE | ||
+ | | 1 | ||
+ | | Surface type | ||
+ | |- | ||
+ | | DF_SURFACE_N_PARAMS | ||
+ | | 1 | ||
+ | | Number of parameters defining the surface | ||
+ | |- | ||
+ | | DF_SURFACE_PARAMS | ||
+ | | ''N''<sub>P</sub> | ||
+ | | List of parameters defining the surface | ||
|- | |- | ||
| DF_SYM | | DF_SYM | ||
Line 2,273: | Line 2,767: | ||
| 1 | | 1 | ||
| Number of corner values (denoted as ''N''<sub>C</sub> below) | | Number of corner values (denoted as ''N''<sub>C</sub> below) | ||
+ | |- | ||
+ | | DF_N_SGN | ||
+ | | 1 | ||
+ | | Number of sign moment values (denoted as ''N''<sub>M</sub> below) | ||
|- | |- | ||
| DF_VOLUME | | DF_VOLUME | ||
Line 2,291: | Line 2,789: | ||
|- | |- | ||
| DF_SURF_IN_CURR | | DF_SURF_IN_CURR | ||
− | | ''2G'' | + | | ''2G'' × ''N<sub>S</sub>'' |
| Inward surface currents | | Inward surface currents | ||
|- | |- | ||
| DF_SURF_OUT_CURR | | DF_SURF_OUT_CURR | ||
− | | ''2G'' | + | | ''2G'' × ''N<sub>S</sub>'' |
| Outward surface currents | | Outward surface currents | ||
|- | |- | ||
| DF_SURF_NET_CURR | | DF_SURF_NET_CURR | ||
− | | ''2G'' | + | | ''2G'' × ''N<sub>S</sub>'' |
| Net surface currents | | Net surface currents | ||
|- | |- | ||
| DF_MID_IN_CURR | | DF_MID_IN_CURR | ||
− | | ''2G'' | + | | ''2G'' × ''N<sub>S</sub>'' |
| Inward mid-point currents | | Inward mid-point currents | ||
|- | |- | ||
| DF_MID_OUT_CURR | | DF_MID_OUT_CURR | ||
− | | ''2G'' | + | | ''2G'' × ''N<sub>S</sub>'' |
| Outward mid-point currents | | Outward mid-point currents | ||
|- | |- | ||
| DF_MID_NET_CURR | | DF_MID_NET_CURR | ||
− | | ''2G'' | + | | ''2G'' × ''N<sub>S</sub>'' |
| Net mid-point currents | | Net mid-point currents | ||
|- | |- | ||
| DF_CORN_IN_CURR | | DF_CORN_IN_CURR | ||
− | | ''2G'' | + | | ''2G'' × ''N<sub>C</sub>'' |
| Inward corner currents | | Inward corner currents | ||
|- | |- | ||
| DF_CORN_OUT_CURR | | DF_CORN_OUT_CURR | ||
− | | ''2G'' | + | | ''2G'' × ''N<sub>C</sub>'' |
| Outward corner currents | | Outward corner currents | ||
|- | |- | ||
| DF_CORN_NET_CURR | | DF_CORN_NET_CURR | ||
− | | ''2G'' | + | | ''2G'' × ''N<sub>C</sub>'' |
| Net corner currents | | Net corner currents | ||
|- | |- | ||
Line 2,331: | Line 2,829: | ||
|- | |- | ||
| DF_HET_SURF_FLUX | | DF_HET_SURF_FLUX | ||
− | | ''2G'' | + | | ''2G'' × ''N<sub>S</sub>'' |
| Heterogeneous surface fluxes | | Heterogeneous surface fluxes | ||
|- | |- | ||
| DF_HET_CORN_FLUX | | DF_HET_CORN_FLUX | ||
− | | ''2G'' | + | | ''2G'' × ''N<sub>C</sub>'' |
| Heterogeneous corner fluxes | | Heterogeneous corner fluxes | ||
|- | |- | ||
Line 2,343: | Line 2,841: | ||
|- | |- | ||
| DF_HOM_SURF_FLUX | | DF_HOM_SURF_FLUX | ||
− | | ''2G'' | + | | ''2G'' × ''N<sub>S</sub>'' |
| Homogeneous surface fluxes | | Homogeneous surface fluxes | ||
|- | |- | ||
| DF_HOM_CORN_FLUX | | DF_HOM_CORN_FLUX | ||
− | | ''2G'' | + | | ''2G'' × ''N<sub>C</sub>'' |
| Homogeneous corner fluxes | | Homogeneous corner fluxes | ||
|- | |- | ||
| DF_SURF_DF | | DF_SURF_DF | ||
− | | ''2G'' | + | | ''2G'' × ''N<sub>S</sub>'' |
| Surface discontinuity factors | | Surface discontinuity factors | ||
|- | |- | ||
| DF_CORN_DF | | DF_CORN_DF | ||
− | | ''2G'' | + | | ''2G'' × ''N<sub>C</sub>'' |
| Corner discontinuity factors | | Corner discontinuity factors | ||
|- | |- | ||
| DF_SGN_SURF_IN_CURR | | DF_SGN_SURF_IN_CURR | ||
− | | ''2G'' | + | | ''2G'' × ''N<sub>M</sub>'' |
| Inward sign moment weighted currents | | Inward sign moment weighted currents | ||
|- | |- | ||
| DF_SGN_SURF_OUT_CURR | | DF_SGN_SURF_OUT_CURR | ||
− | | ''2G'' | + | | ''2G'' × ''N<sub>M</sub>'' |
| Outward sign moment weighted currents | | Outward sign moment weighted currents | ||
|- | |- | ||
| DF_SGN_SURF_NET_CURR | | DF_SGN_SURF_NET_CURR | ||
− | | ''2G'' | + | | ''2G'' × ''N<sub>M</sub>'' |
| Net sign moment weighted currents | | Net sign moment weighted currents | ||
|- | |- | ||
| DF_SGN_HET_SURF_FLUX | | DF_SGN_HET_SURF_FLUX | ||
− | | ''2G'' | + | | ''2G'' × ''N<sub>M</sub>'' |
| Heterogeneous sign moment weighted surface fluxes | | Heterogeneous sign moment weighted surface fluxes | ||
|- | |- | ||
| DF_SGN_HOM_SURF_FLUX | | DF_SGN_HOM_SURF_FLUX | ||
− | | ''2G'' | + | | ''2G'' × ''N<sub>M</sub>'' |
| Homogeneous sign moment weighted surface fluxes | | Homogeneous sign moment weighted surface fluxes | ||
|- | |- | ||
| DF_SGN_SURF_DF | | DF_SGN_SURF_DF | ||
− | | ''2G'' | + | | ''2G'' × ''N<sub>M</sub>'' |
| Sign moment weighted surface discontinuity factors | | Sign moment weighted surface discontinuity factors | ||
|} | |} | ||
Line 2,414: | Line 2,912: | ||
| PPW_PINS | | PPW_PINS | ||
| 1 | | 1 | ||
− | | Number of pin positions in the lattice (denoted as ''N | + | | Number of pin positions in the lattice (denoted as ''N<sub>P</sub>'' below) |
|- | |- | ||
| PPW_POW | | PPW_POW | ||
− | | ''2G'' | + | | ''2G'' × ''N<sub>P</sub>'' |
| Pin- and group-wise power distribution normalized to unity sum | | Pin- and group-wise power distribution normalized to unity sum | ||
|- | |- | ||
| PPW_HOM_FLUX | | PPW_HOM_FLUX | ||
− | | ''2G'' | + | | ''2G'' × ''N<sub>P</sub>'' |
| Pin- and group-wise homogeneous flux distribution | | Pin- and group-wise homogeneous flux distribution | ||
|- | |- | ||
| PPW_FF | | PPW_FF | ||
− | | ''2G'' | + | | ''2G'' × ''N<sub>P</sub>'' |
| Pin- and group-wise form factors | | Pin- and group-wise form factors | ||
|- | |- | ||
Line 2,439: | Line 2,937: | ||
*The order of ALB_TOT_ALB is <math>\alpha_{1,1} , \alpha_{1,2} , \ldots , \alpha_{2,1} , \alpha_{2,2} , \ldots</math> where <math>\alpha_{g,g'}</math> refers to albedo from group ''g'' to ''g'''. | *The order of ALB_TOT_ALB is <math>\alpha_{1,1} , \alpha_{1,2} , \ldots , \alpha_{2,1} , \alpha_{2,2} , \ldots</math> where <math>\alpha_{g,g'}</math> refers to albedo from group ''g'' to ''g'''. | ||
*The order of ALB_PART_ALB is <math>\alpha_{1,1,1,1} , \alpha_{1,1,1,2} , \ldots , \alpha_{1,1,2,1} , \alpha_{1,1,2,2} , \ldots , \alpha_{1,2,1,1} , \alpha_{1,2,1,2} , \ldots , \alpha_{2,1,1,1} , \alpha_{2,1,1,2} , \ldots </math> where <math>\alpha_{k,g,k',g'}</math> refers to albedo of surface ''k<nowiki>'</nowiki>'' of group ''g<nowiki>'</nowiki>'' which has entered the albedo surface through surface ''k'' and group ''g''. | *The order of ALB_PART_ALB is <math>\alpha_{1,1,1,1} , \alpha_{1,1,1,2} , \ldots , \alpha_{1,1,2,1} , \alpha_{1,1,2,2} , \ldots , \alpha_{1,2,1,1} , \alpha_{1,2,1,2} , \ldots , \alpha_{2,1,1,1} , \alpha_{2,1,1,2} , \ldots </math> where <math>\alpha_{k,g,k',g'}</math> refers to albedo of surface ''k<nowiki>'</nowiki>'' of group ''g<nowiki>'</nowiki>'' which has entered the albedo surface through surface ''k'' and group ''g''. | ||
− | * For example, two-group hexagonal partial albedos can be converted into matrix form using the reshape-command in Matlab with the notation part_alb(g', k', g, k) <math>= \alpha_{k,g,k',g'}</math> as | + | * For example, two-group hexagonal partial albedos can be converted into matrix form using the reshape-command in Matlab with the notation part_alb(''g', k', g, k'') <math>= \alpha_{k,g,k',g'}</math> as |
<nowiki> | <nowiki> | ||
part_alb = reshape(ALB_PART_ALB(1, 1:2:end), 2, 6, 2, 6)</nowiki> | part_alb = reshape(ALB_PART_ALB(1, 1:2:end), 2, 6, 2, 6)</nowiki> | ||
Line 2,458: | Line 2,956: | ||
| ALB_N_SURF | | ALB_N_SURF | ||
| 1 | | 1 | ||
− | | Number of albedo surface faces (denoted as ''N | + | | Number of albedo surface faces (denoted as ''N<sub>S</sub>'' below) |
|- | |- | ||
| ALB_IN_CURR | | ALB_IN_CURR | ||
− | | ''2G'' | + | | ''2G'' × ''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> × ''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 2,473: | Line 2,971: | ||
|- | |- | ||
| ALB_PART_ALB | | ALB_PART_ALB | ||
− | | ''2G''<sup>2</sup> | + | | ''2G''<sup>2</sup> × ''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 == | ||
− | |||
− | |||
− | |||
− | |||
== References == | == References == |
Latest revision as of 15:11, 12 June 2023
This page lists the output parameters in the main [input]_res.m output file.
Contents
- 1 General output parameters
- 2 Misc. statistics
- 2.1 Collision and reaction sampling (neutrons/photons)
- 2.2 STL geometries
- 2.3 Importance solver
- 2.4 Run statistics
- 2.5 Running times
- 2.6 Memory usage
- 2.7 Geometry parameters
- 2.8 Neutron energy grid
- 2.9 Photon energy grid
- 2.10 Unresolved resonance probability table sampling
- 2.11 Nuclides and reaction channels
- 3 Physics
- 4 Particle balance
- 5 Integral results
- 5.1 Normalized total reaction rates (neutrons)
- 5.2 Normalized total reaction rates (photons)
- 5.3 Equilibrium Xe-135 iteration
- 5.4 Equilibrium Sm-149 iteration
- 5.5 Iteration factor
- 5.6 Six-factor formula
- 5.7 Fission neutron and energy production
- 5.8 Criticality eigenvalues / Multiplication factor external source
- 5.9 Wielandt method
- 5.10 ALF (Average lethargy of neutrons causing fission)
- 5.11 EALF (Energy corresponding to average lethargy of neutrons causing fission)
- 5.12 AFGE (Average energy of neutrons causing fission)
- 6 Time constants
- 6.1 Forward delayed neutron parameters
- 6.2 Adjoint-weighted time constants 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)
- 6.8 Dynamic simulation
- 6.9 Analog mean photon lifetime
- 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 (criticality) or the set nps input option (external source) |
CYCLES | 1 | Number of active cycles defined using the set pop input option |
SKIP | 1 | Number of inactive cycles defined using the set pop input option |
BATCH_INTERVAL | 1 | Batching interval defined using the set pop input option |
BATCHES | 1 | Number of batches defined using the set nps input option |
SRC_NORM_MODE | 1 | Source normalization mode |
SEED | 1 | Random number seed taken from system time or defined using the set seed input option |
UFS_MODE | 1 | Uniform fission source mode defined using the set ufs input option |
UFS_ORDER | 1 | Uniform fission exponential factor using the set ufs input option |
NEUTRON_TRANSPORT_MODE | 1 | Flag indicating whether or not neutron transport simulation is on |
PHOTON_TRANSPORT_MODE | 1 | Flag indicating whether or not neutron transport simulation is on |
GROUP_CONSTANT_GENERATION | 1 | Flag indicating whether or not group constant generation is on |
B1_CALCULATION | 3 | Flag indicating whether or not B1 calculation is on |
B1_BURNUP_CORRECTION | 1 | Flag indicating whether or not B1 burnup correction is on |
CRIT_SPEC_MODE | 2 | Critical spectrum modes |
IMPLICIT_REACTION_RATES | 1 | Flag indicating whether or not implicit reaction rates are used for group constant generation |
VR_ITER_IDX | 1 | Variance reduction iteration index when output was printed (see wwin card) |
Domain decomposition
Parameter | Size | Description |
---|---|---|
DD_MODE | 1 | Domain decomposition mode defined using the set dd input option |
DD_NEUTRONS_TO_LIMBO | M | Neutrons sent to limbo (transferral buffer) at each domain |
DD_NEUTRONS_FROM_LIMBO | M | Neutrons received from limbo (transferral buffer) at each domain |
Optimization
Parameter | Size | Description |
---|---|---|
OPTIMIZATION_MODE | 1 | Optimization mode defined using the set opti input option |
RECONSTRUCT_MICROXS | 1 | Flag indicating whether or not microscopic cross sections are reconstructed on the unionized energy grid |
RECONSTRUCT_MACROXS | 1 | Flag indicating whether or not macroscopic cross sections are reconstructed on the unionized energy grid |
DOUBLE_INDEXING | 1 | Double indexing option defined using the set dix input option |
MG_MAJORANT_MODE | 1 | Multi-group majorant mode |
SPECTRUM_COLLAPSE | 1 | Spectrum collapse method flag (set xscalc input option) |
Parallelization
Parameter | Size | Description |
---|---|---|
MPI_TASKS | 1 | Number of parallel MPI tasks |
OMP_THREADS | 1 | Number of parallel OpenMP threads |
MPI_REPRODUCIBILITY | 1 | MPI reproducibility option defined by the set repro input option |
OMP_REPRODUCIBILITY | 1 | OpenMP reproducibility option defined by the set repro input option |
OMP_HISTORY_PROFILE | N | Fraction of particle histories run for each parallel OpenMP thread |
SHARE_BUF_ARRAY | 1 | Shared buffer flag |
SHARE_RES2_ARRAY | 1 | Shared RES2 array flag |
OMP_SHARED_QUEUE_LIM | 1 | Limiting value for using shared particle queue |
File paths
Notes:
- Only the first file path listed is displayed
Parameter | Size | Description |
---|---|---|
XS_DATA_FILE_PATH | (string) | Cross section directory file path defined using the set acelib input option |
DECAY_DATA_FILE_PATH | (string) | Radioactive decay data file path defined using the set declib input option |
SFY_DATA_FILE_PATH | (string) | Spontaneous fission yield data file path defined using the set sfylib input option |
NFY_DATA_FILE_PATH | (string) | Neutron-induced fission yield data file path defined using the set nfylib input option |
BRA_DATA_FILE_PATH | (string) | Isomeric branching ratio data file path defined using the set bralib input option |
PHOTON_PHYS_DIRECTORY | (string) | Photon physics directory path defined using the set pdatadir input option |
Misc. statistics
Collision and reaction sampling (neutrons/photons)
Notes:
- The first single/pair value corresponds to neutrons and, the second single/pair value corresponds to photons.
Parameter | Size | Description |
---|---|---|
MEAN_SRC_WGT | 2/2 | Mean source weight for non-criticality calculations (neutrons/photons) |
SOURCE_SAMPLING_EFF | 2/2 | Source sampling efficiency for non-criticality calculations (neutrons/photons) |
MEAN_SRC_WW_SPLIT | 2/2 | Mean source weight-window splitting in variance reduction (neutrons/photons) |
MEAN_SRC_WW_EFF | 2/2 | Mean source weight-window sampling efficiency in variance reduction (neutrons/photons) |
WW_BALA_ROULETE | 2/2 | Mean weight-window balance due to Russian roulette in variance reduction (neutrons/photons) |
WW_BALA_SPLIT | 2/2 | Mean weight-window balance due to splitting in variance reduction (neutrons/photons) |
MIN_MACROXS | 2/2 | Macroscopic cross section corresponding to the minimum mfp used for scoring the collision flux estimator (see the set cfe input option) |
DT_THRESH | 1/1 | Probability threshold used for switching to delta-tracking (see the set dt input option) |
ST_FRAC | 2/2 | Fraction of paths sampled using surface-tracking |
DT_FRAC | 2/2 | Fraction of paths sampled using delta-tracking |
DT_EFF | 2/2 | Delta-tracking efficiency |
IFC_COL_EFF | 2/2 | Efficiency of interface collision rejection (see ifc card) |
REA_SAMPLING_EFF | 2/2 | Reaction sampling efficiency |
REA_SAMPLING_FAIL | 2/2 | Fraction of failed reaction samples |
TMS_SAMPLING_EFF | 2 | Target motion sampling method, TMS, sampling efficiency (see tms option, in mat card) |
TOT_COL_EFF | 2/2 | Total collision efficiency |
AVG_TRACKING_LOOPS | 2/2, 2/2 | Average number of tracking loops per history and, fraction of failed tracking loops |
TMS_FAIL_STAT | 8 | TMS fail statistics: total samples, majorant fail, lower limit fail, upper limit fail (see tms option, in mat card) |
DBRC_EXCEED_FRAC | 1 | Doppler-broadening rejection correction, DBRC, majorant exceed fraction (see set dbrc input option) |
AVG_TRACKS | 2/2 | Average number of tracks per history |
AVG_REAL_COL | 2/2 | Average number of real collisions per history |
AVG_VIRT_COL | 2/2 | Average number of virtual collisions per history |
AVG_SURF_CROSS | 2/2 | Average number of surface crossings per history (NOTE: accurate only in ST mode) |
LOST_PARTICLES | 1 | Number of lost particles |
STL geometries
Parameter | Size | Description |
---|---|---|
STL_RAY_TEST | 5 | STL-ray-tracing test: total, ray is too parallel to facet, intersection point is too close to edge, facet is too close to search mesh cell boundary, two facets overlap |
STL_ENFORCE_ST | 1 | Flag indicating whether or not delta tracking is enforced in STL geometries |
Importance solver
Parameter | Size | Description |
---|---|---|
NEIGHBOUR_SEARCH_FAIL | 2 | Response matrix calculation fail rate |
Run statistics
Parameter | Size | Description |
---|---|---|
CYCLE_IDX | 1 | Cycle index when output was printed |
SIMULATED_HISTORIES | 1 | Number of simulated histories when output was printed |
MEAN_POP_SIZE | 1 | Mean population size |
MEAN_POP_WGT | 1 | Mean population weight |
SIMULATION_COMPLETED | 1 | Flag indicating whether or not the simulation was completed |
Running times
Notes:
- All times in minutes
- In burnup calculations the first value provides the cumulative and the second value the cycle-wise value
Parameter | Size | Description |
---|---|---|
TOT_CPU_TIME | 1 | Total CPU time |
RUNNING_TIME | 1 | Total wall-clock running time |
INIT_TIME | 1(2) | Wall-clock time spent for initialization |
PROCESS_TIME | 1(2) | Wall-clock time spent for processing |
TRANSPORT_CYCLE_TIME | 1(3) | Wall-clock time spent for transport simulation |
FINIX_SOLUTION_TIME | 1 | Wall-clock time spent for FINIX solution |
BURNUP_CYCLE_TIME | 1(2) | Wall-clock time spent for burnup solution |
BATEMAN_SOLUTION_TIME | 1(2) | Wall-clock time spent for solving the Bateman equations |
MPI_OVERHEAD_TIME | 1(2) | Wall-clock time spent MPI communication |
DD_OVERHEAD_TIME | 1 | Wall-clock time spent for DD algorithm (see set dd input option) |
RMX_SOLUTION_TIME | 1 | Wall-clock time spent for response matrix solution (see wwgen / wwin cards or set sca input option) |
LEAKAGE_CORR_SOL_TIME | 1 | Wall-clock time spent for leakage correction solution (see set fum input option) |
ESTIMATED_RUNNING_TIME | 1(2) | Estimated total wall-clock running time |
CPU_USAGE | 1 | Total CPU usage fraction |
TRANSPORT_CPU_USAGE | 1(2) | CPU usage fraction in transport simulation |
OMP_PARALLEL_FRAC | 1 | Fraction of time spent in OpenMP parallel loops |
Memory usage
Notes:
- All values are in megabytes
- Serpent allocates memory in fixed segments, so the allocated memory size may be larger than what is needed for the simulation
Parameter | Size | Description |
---|---|---|
AVAIL_MEM | 1 | Available memory size |
ALLOC_MEMSIZE | 1 | Allocated memory size |
MEMSIZE | 1 | Used memory size |
XS_MEMSIZE | 1 | Memory size used for storing cross sections |
MAT_MEMSIZE | 1 | Memory size used for storing material-wise data |
RES_MEMSIZE | 1 | Memory size used for storing results |
IFC_MEMSIZE | 1 | Memory size used for data for multi-physics interface data |
RMX_MEMSIZE | 1 | Memory size used for storing response matrix-wise data |
MISC_MEMSIZE | 1 | Memory size used for data for miscellaneous data |
UNKNOWN_MEMSIZE | 1 | Memory size used for data for uncategorized data |
UNUSED_MEMSIZE | 1 | Allocated memory not used for anything |
Geometry parameters
Parameter | Size | Description |
---|---|---|
TOT_CELLS | 1 | Total number of cells |
UNION_CELLS | 1 | Total number of cells defined using unions |
Neutron energy grid
Parameter | Size | Description |
---|---|---|
NEUTRON_ERG_TOL | 1 | Reconstruction tolerance for unionized energy grid (see set egrid input option) |
NEUTRON_ERG_NE | 1 | Number of points in neutron unionized energy grid |
NEUTRON_EMIN | 1 | Minimum energy for neutron cross section data (see set egrid input option) |
NEUTRON_EMAX | 1 | Maximum energy for neutron cross section data (see set egrid input option) |
Photon energy grid
Parameter | Size | Description |
---|---|---|
PHOTON_ERG_NE | 1 | Number of points in photon unionized energy grid |
PHOTON_EMIN | 1 | Minimum energy for photon cross section data (see set egrid input option) |
PHOTON_EMAX | 1 | Maximum energy for photon cross section data (see set egrid input option) |
Unresolved resonance probability table sampling
Parameter | Size | Description |
---|---|---|
URES_DILU_CUT | 1 | Density cut-off used for unresolved resonance probability table sampling (see set ures input option) |
URES_EMIN | 1 | Minimum energy for unresolved resonance range |
URES_EMAX | 1 | Maximum energy for unresolved resonance range |
URES_AVAIL | 1 | Number of nuclides with probability table data |
URES_USED | 1 | Number of nuclides for which probability table sampling was used (see set ures input option) |
Nuclides and reaction channels
Parameter | Size | Description |
---|---|---|
TOT_NUCLIDES | 1 | Total number of nuclides |
TOT_TRANSPORT_NUCLIDES | 1 | Total number of nuclides with transport cross sections |
TOT_DOSIMETRY_NUCLIDES | 1 | Total number of nuclides with dosimetry cross sections |
TOT_DECAY_NUCLIDES | 1 | Total number of decay nuclides (without transport cross sections) |
TOT_PHOTON_NUCLIDES | 1 | Total number of nuclides with photon cross section data |
TOT_REA_CHANNELS | 1 | Total number of reaction channels |
TOT_TRANSMU_REA | 1 | Total number of transmutation reactions |
Physics
Neutron physics options
Parameter | Size | Description |
---|---|---|
USE_DELNU | 1 | Flag indicating whether or not delayed neutron emission is on (see set delnu input option) |
USE_URES | 1 | Flag indicating whether or not unresolved resonance probability table sampling is on (see set ures input option) |
USE_DBRC | 1 | Flag indicating whether or not Doppler-broadening rejection correction is on (see set dbrc input option) |
IMPL_CAPT | 1 | Flag indicating whether or not implicit capture reaction mode is on (see set impl input option) |
IMPL_NXN | 1 | Flag indicating whether or not implicit nxn reaction mode is on (see set impl input option) |
IMPL_FISS | 1 | Flag indicating whether or not implicit fission reaction mode is on (see set impl input option) |
IMPL_FISS_NUBAR | 1 | Flag indicating whether or not implicit fission nubar reaction mode is on (see set impl input option) |
DOPPLER_PREPROCESSOR | 1 | Flag indicating whether or not Doppler-broadening preprocessor is on (see tmp option, in mat card) |
TMS_MODE | 1 | Flag indicating whether or not target motion sampling is on (see tms option, in mat card) |
SAMPLE_FISS | 1 | Flag indicating whether or not fission reactions are handled (see set nphys input option) |
SAMPLE_CAPT | 1 | Flag indicating whether or not capture reactions are handled (see set nphys input option) |
SAMPLE_SCATT | 1 | Flag indicating whether or not scattering reactions are handled (see set nphys input option) |
Photon physics options
Parameter | Size | Description |
---|---|---|
COMPTON_EKN | 1 | Photon energy above which Klein-Nishina is used for calculating energy and direction of the scattered photons (see set ekn input option) |
COMPTON_DOPPLER | 1 | Flag indicating whether or not Doppler broadening method for the energy spectrum of the scattered photons is on (see set cdop input option) |
COMPTON_EANG | 1 | Flag indicating whether or not Compton electron angular distribution model is on (see set cea input option) |
PHOTON_TTB | 1 | Flag indicating whether or not thick-target bremsstrahlung approximation for modelling electrons and positrons is on (see set ttb input option) |
Photon production
Parameter | Size | Description |
---|---|---|
PHOTON_SAMPLING_MODE | 1 | Flag indicating whether or not photon production from neutron reactions mode is on (see set ngamma input option) |
PHOTON_SAMPLING_FAIL | 2 | Fraction of failed photon samples |
Energy deposition
Notes:
- The list of fission energy release components includes: (1) EFR, kinetic energy of the fission products (following prompt neutron emission from the fission fragments); (2) ENP, kinetic energy of the prompt fission neutrons; (3) END, kinetic energy of the delayed fission neutrons; (4) EGP, total energy release by the emission of prompt gamma rays; (5) EGD, total energy release by the emission of delayed gamma rays; (6) EB, total energy release by delayed beta’s; (7) ENU, energy carried away by neutrinos; (8) ER, total energy less the energy of the neutrinos (ET - ENU), equal to the pseudo-Q-value in File 3 for MT=18; (9) ET, sum of all the partial energies previously listed, corresponding to the total energy release per fission and equal the Q-value.
Parameter | Size | Description |
---|---|---|
EDEP_MODE | 1 | Energy deposition mode (see set edepmode input option) |
EDEP_DELAYED | 1 | Energy of delayed components in energy deposition calculations (see set edepdel input option) |
EDEP_KEFF_CORR | 1 | Flag indicating whether or not correction for energy deposition estimates in non-critical systems (see set edepkcorr input option) |
EDEP_LOCAL_EGD | 1 | Energy distribution of delayed components in energy deposition calculations, mode 3 (see set edepdel input option) |
EDEP_COMP | 9 | Fission energy release components: EFR, ENP, END, EGP, EGD, EB, ENU, ER, ET. |
EDEP_CAPT_E | 1 | Additional energy release in capture reactions, mode 1 (see set edepmode input option) |
Radioactivity data
Notes:
- The values are given at the current burnup point (depletion step).
Parameter | Size | Description |
---|---|---|
TOT_ACTIVITY | 1 | Total activity |
TOT_DECAY_HEAT | 1 | Total decay heat |
TOT_SF_RATE | 1 | Total spontaneous fission rate |
ACTINIDE_ACTIVITY | 1 | Actinide activity |
ACTINIDE_DECAY_HEAT | 1 | Actinide decay heat |
FISSION_PRODUCT_ACTIVITY | 1 | Fission product activity |
FISSION_PRODUCT_DECAY_HEAT | 1 | Fission product decay heat |
INHALATION_TOXICITY | 1 | Total inhalation toxicity |
INGESTION_TOXICITY | 1 | Total ingestion toxicity |
ACTINIDE_INH_TOX | 1 | Actinide inhalation toxicity |
ACTINIDE_ING_TOX | 1 | Actinide ingestion toxicity |
FISSION_PRODUCT_INH_TOX | 1 | Fission product inhalation toxicity |
FISSION_PRODUCT_ING_TOX | 1 | Fission product ingestion toxicity |
SR90_ACTIVITY | 1 | Sr-90 activity |
TE132_ACTIVITY | 1 | Te-132 activity |
I131_ACTIVITY | 1 | I-131 activity |
I132_ACTIVITY | 1 | I-132 activity |
CS134_ACTIVITY | 1 | Cs-134 activity |
CS137_ACTIVITY | 1 | Cs-137 activity |
PHOTON_DECAY_SOURCE | 2 | Total photon decay source rate and total released gamma decay |
NEUTRON_DECAY_SOURCE | 1 | Total neutron decay source rate |
ALPHA_DECAY_SOURCE | 1 | Total alpha decay source rate |
ELECTRON_DECAY_SOURCE | 1 | Total beta decay source rate |
Normalization coefficient
Parameter | Size | Description |
---|---|---|
NORM_COEF | 2/2 | Proportionality constant between the simulated events and the "physical" events that the simulated events represent, for neutrons and photons. |
Parameters for burnup calculation
Parameter | Size | Description |
---|---|---|
BURN_MATERIALS | 1 | Number of depleted materials. |
BURN_MODE | 1 | Burnup mode: 1 = TTA, 2 = CRAM (see set bumode input option). |
BURN_STEP | 1 | Burnup step index. |
BURN_RANDOMIZE_DATA | 3 | Flag indicating whether or not randomize data is set on: decay constants, fission yields and decay heat (see set rnddec input option). |
BURNUP | 2 | Burnup at the current step (in MWd/kgU): cumulative calculated from the depletion history input and cumulative realized from the actual calculation. |
BURN_DAYS | 2 | Number of burn days at the current step: cumulative and step-wise. |
FIMA | 3 | Number of fissions per initial fissile atom at the current step: relative step-wise, increment step-wise, final step-wise. |
Coefficient calculation
Parameter | Size | Description |
---|---|---|
COEF_IDX | 2 | Coefficient index when output is printed and total number of coefficient calculations |
COEF_BRANCH | 1 | Branch index within coefficient calculation when output is printed |
COEF_BU_STEP | 1 | Burnup step at the given coefficient calculation when output is printed |
Analog reaction rate estimators
Parameter | Size | Description |
---|---|---|
CONVERSION_RATIO | 2 | Analog estimate of conversion rate, ratio between fissile production and loss rate |
TH232_FISS | 4 | Analog estimate of Th-232 fission rate (total/fraction) |
U233_FISS | 4 | Analog estimate of U-233 fission rate (total/fraction) |
U235_FISS | 4 | Analog estimate of U-235 fission rate (total/fraction) |
U238_FISS | 4 | Analog estimate of U-238 fission rate (total/fraction) |
PU239_FISS | 4 | Analog estimate of Pu-239 fission rate (total/fraction) |
PU240_FISS | 4 | Analog estimate of Pu-240 fission rate (total/fraction) |
PU241_FISS | 4 | Analog estimate of Pu-241 fission rate (total/fraction) |
TH232_CAPT | 4 | Analog estimate of Th-232 capture rate (total/fraction) |
U233_CAPT | 4 | Analog estimate of U-233 capture rate (total/fraction) |
U235_CAPT | 4 | Analog estimate of U-235 capture rate (total/fraction) |
U238_CAPT | 4 | Analog estimate of U-238 capture rate (total/fraction) |
PU239_CAPT | 4 | Analog estimate of Pu-239 capture rate (total/fraction) |
PU240_CAPT | 4 | Analog estimate of Pu-240 capture rate (total/fraction) |
PU241_CAPT | 4 | Analog estimate of Pu-241 capture rate (total/fraction) |
XE135_CAPT | 4 | Analog estimate of Xe-135 capture rate (total/fraction) |
XE135M_CAPT | 4 | Analog estimate of Xe-135m capture rate (total/fraction) |
SM149_CAPT | 4 | Analog estimate of Sm-149 capture rate (total/fraction) |
Particle balance
Neutron balance (particles/weight)
Parameter | Size | Description |
---|---|---|
BALA_SRC_NEUTRON_SRC | 1/1 | Neutron produced by external source |
BALA_SRC_NEUTRON_FISS | 1/1 | Neutron produced by fission |
BALA_SRC_NEUTRON_NXN | 1/1 | Neutron produced by scattering |
BALA_SRC_NEUTRON_VR | 1/1 | Neutron produced by variance reduction (Russian roulette, splitting) |
BALA_SRC_NEUTRON_TOT | 1/1 | Total neutron produced |
BALA_LOSS_NEUTRON_CAPT | 1/1 | Neutron lost by capture |
BALA_LOSS_NEUTRON_FISS | 1/1 | Neutron lost by fission |
BALA_LOSS_NEUTRON_LEAK | 1/1 | Neutron lost by leakage |
BALA_LOSS_NEUTRON_CUT | 1/1 | Neutron lost by cut-off |
BALA_LOSS_NEUTRON_ERR | 1/1 | Neutron lost by failed tracking |
BALA_LOSS_NEUTRON_TOT | 1/1 | Total neutron lost |
BALA_NEUTRON_DIFF | 1/1 | Difference between total neutron produced and lost |
Photon balance (particles/weight/energy-weighted)
Parameter | Size | Description |
---|---|---|
BALA_SRC_PHOTON_SRC | 1/1/1 | Photon produced by external source |
BALA_SRC_PHOTON_TTB | 1/1/1 | Photon produced by bremsstrahlung |
BALA_SRC_PHOTON_ANNIH | 1/1/1 | Photon produced by annihilation |
BALA_SRC_PHOTON_FLUOR | 1/1/1 | Photon produced by fluorescence |
BALA_SRC_PHOTON_NREA | 1/1/1 | Photon produced by neutron reaction |
BALA_SRC_PHOTON_VR | 1/1/1 | Photon produced by variance reduction (Russian roulette, splitting) |
BALA_SRC_PHOTON_TOT | 1/1/1 | Total photon produced |
BALA_LOSS_PHOTON_CAPT | 1/1 | Photon lost by capture |
BALA_LOSS_PHOTON_LEAK | 1/1 | Photon lost by leakage |
BALA_LOSS_PHOTON_CUT | 1/1 | Photon lost by cut-off |
BALA_LOSS_PHOTON_ERR | 1/1 | Photon lost by failed tracking |
BALA_LOSS_PHOTON_TOT | 1/1 | Total photon lost |
BALA_PHOTON_DIFF | 1/1 | Difference between total photon produced and lost |
Integral results
Normalized total reaction rates (neutrons)
Notes:
- In burnup calculations the values correspond to total, burnable and non-burnable rates
Parameter | Size | Description |
---|---|---|
TOT_POWER | 2(6) | Total neutron fission power |
TOT_POWDENS | 2(6) | Total neutron fission power density |
TOT_GENRATE | 2(6) | Total neutron generation rate |
TOT_FISSRATE | 2(6) | Total neutron fission rate |
TOT_CAPTRATE | 2(6) | Total neutron capture rate |
TOT_ABSRATE | 2(6) | Total neutron absorption rate |
TOT_SRCRATE | 2(6) | Total neutron source rate |
TOT_FLUX | 2(6) | Total neutron flux |
TOT_PHOTON_PRODRATE | 4 | Total neutron-photon production rate (implicit/analog) |
TOT_LEAKRATE | 2 | Total neutron leakage rate |
ALBEDO_LEAKRATE | 2 | Albedo neutron leakage rate |
TOT_LOSSRATE | 2 | Total neutron loss rate |
TOT_CUTRATE | 2 | Total neutron energy cut-off rate |
TOT_RR | 2 | Total neutron reaction rate |
TOT_XE135_ABSRATE | 2 | Total neutron absorption rate in Xe-135 |
TOT_XE135M_ABSRATE | 2 | Total neutron absorption rate in Xe-135m |
TOT_SM149_ABSRATE | 2 | Total neutron absorption rate in Sm-149 |
INI_FMASS | 1 | Initial fissile mass |
TOT_FMASS | 1 | Total fissile mass |
INI_BURN_FMASS | 1 | Initial fissile mass in burnable materials |
TOT_BURN_FMASS | 1 | Total fissile mass in burnable materials |
Normalized total reaction rates (photons)
Parameter | Size | Description |
---|---|---|
TOT_PHOTON_LEAKRATE | 2 | Total photon leakage rate |
TOT_PHOTON_CUTRATE | 2 | Total photon energy cut-off rate |
PHOTOELE_CAPT_RATE | 2 | Photo-electric photon capture rate |
PAIRPROD_CAPT_RATE | 2 | Pair production photon capture rate |
TOT_PHOTON_LOSSRATE | 2 | Total photon loss rate |
TOT_PHOTON_SRCRATE | 2 | Total photon source rate |
TOT_PHOTON_RR | 2 | Total photon reaction rate |
TOT_PHOTON_FLUX | 2 | Total photon flux |
TOT_PHOTON_HEATRATE | 2 | Total photon heating rate |
Equilibrium Xe-135 iteration
Parameter | Size | Description |
---|---|---|
I135_EQUIL_CONC | 2 | Averaged equilibrium I-135 concentration (see set xenon input option) |
XE135_EQUIL_CONC | 2 | Averaged equilibrium Xe-135 concentration (see set xenon input option) |
XE135M_EQUIL_CONC | 2 | Averaged equilibrium Xe-135m 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) |
Iteration factor
Parameter | Size | Description |
---|---|---|
ITER_FACTOR | 2 | Iteration factor of critical density iteration (see set iter nuc input option) or albedo iteration (see set iter alb input option) |
Six-factor formula
Notes:
- The six-factor formula estimates are printed by default. Check the suitability of the approximation for the given spectrum (the formulation is based on a thermal spectrum).
Parameter | Size | Description |
---|---|---|
SIX_FF_ETA | 2 | Analog estimate of average number of neutrons emitted per thermal neutron absorbed in fuel |
SIX_FF_F | 2 | Analog estimate of thermal utilization factor |
SIX_FF_P | 2 | Analog estimate of resonance escape probability |
SIX_FF_EPSILON | 2 | Analog estimate of fast fission factor |
SIX_FF_LF | 2 | Analog estimate of fast non-leakage probability |
SIX_FF_LT | 2 | Analog estimate of thermal non-leakage probability |
SIX_FF_KINF | 2 | Analog estimate of six-factor k_{inf} (four-factor k_{eff}) |
SIX_FF_KEFF | 2 | Analog estimate of six-factor k_{eff} |
Fission neutron and energy production
Parameter | Size | Description |
---|---|---|
NUBAR | 2 | Average fission neutron yield |
FISSE | 2 | Average fission energy production |
Criticality eigenvalues / Multiplication factor external source
Parameter | Size | Description |
---|---|---|
ANA_KEFF | 6(2) | Analog estimate of k_{eff}: total, prompt and delayed neutron contribution. |
IMP_KEFF | 2 | Implicit estimate of k_{eff}. |
COL_KEFF | 2 | Collision estimate of k_{eff}. |
ABS_KEFF | 2 | Absorption estimate of k_{eff}. |
ABS_KINF | 2 | Absorption estimate of k_{inf}. |
ANA_EXT_K | 20 | Generation-wise source multiplication factors in external source mode |
SRC_MULT | 2 | Source multiplication factor in external source mode |
MEAN_NGEN | 2 | Mean number of generations in external source mode |
PROMPT_GEN_POP | N_{G} | Prompt fission population generation fraction |
PROMPT_GEN_CUMU | N_{G} | Prompt fission cumulative generation fraction |
PROMPT_GEN_TIMES | N_{G} | Prompt fission time generation fraction |
PROMPT_CHAIN_LENGTH | 2 | Mean prompt chain length in external source mode |
GEOM_ALBEDO | 6 | Fixed or iterated value for albedo boundary condition for x-,y- and z-directions (see set bc or set iter alb input options). |
Wielandt method
Parameter | Size | Description |
---|---|---|
WIELANDT_K | 2 | Wielandt’s method shifted eigenvalue (see set wie input option) |
WIELANDT_P | 2 | Wielandt’s method neutron banking probability (see set wie input option) |
ALF (Average lethargy of neutrons causing fission)
Parameter | Size | Description |
---|---|---|
ANA_ALF | 2 | Analog estimate of average lethargy of neutrons causing fission |
IMP_ALF | 2 | Implicit estimate of average lethargy of neutrons causing fission |
EALF (Energy corresponding to average lethargy of neutrons causing fission)
Parameter | Size | Description |
---|---|---|
ANA_EALF | 2 | Analog estimate of energy corresponding to the average lethargy of neutrons causing fission |
IMP_EALF | 2 | Implicit estimate of energy corresponding to the average lethargy of neutrons causing fission |
AFGE (Average energy of neutrons causing fission)
Parameter | Size | Description |
---|---|---|
ANA_AFGE | 2 | Analog estimate of average energy of neutrons causing fission |
IMP_AFGE | 2 | Implicit estimate of average energy of neutrons causing fission |
Time constants
Forward delayed neutron parameters
Parameter | Size | Description |
---|---|---|
PRECURSOR_GROUPS | 1 | Number of delayed neutron precursor groups (referred to as D below) |
FWD_ANA_BETA_ZERO | 2D + 2 | Analog estimator of physical delayed neutron fractions (number of delayed neutrons emitted in fission): total, group-wise |
FWD_ANA_LAMBDA | 2D + 2 | Analog estimator of delayed neutron precursor decay constants: total, group-wise |
Adjoint-weighted time constants using Meulekamp's method
Parameter | Size | Description |
---|---|---|
ADJ_MEULEKAMP_BETA_EFF | 2D + 2 | Adjoint-weighted effective delayed neutron fractions using Meulekamp's method: total, group-wise |
ADJ_MEULEKAMP_LAMBDA | 2D + 2 | Adjoint-weighted of delayed neutron precursor decay constants using Meulekamp's method: total, group-wise |
Adjoint weighted time constants using Nauchi's method
Parameter | Size | Description |
---|---|---|
IFP_CHAIN_LENGTH | 1 | Number of generations within the iterated fission probability method |
ADJ_NAUCHI_GEN_TIME | 6 | Adjoint-weighted neutron generation times using Nauchi's method: total, prompt, delayed |
ADJ_NAUCHI_LIFETIME | 6 | Adjoint-weighted neutron lifetimes using Nauchi's method: total, prompt, delayed. |
ADJ_NAUCHI_BETA_EFF | 2D + 2 | Adjoint-weighted effective delayed neutron fractions using Nauchi's method: total, group-wise |
ADJ_NAUCHI_LAMBDA | 2D + 2 | Adjoint-weighed of delayed neutron precursor decay constants using Nauchi's method: total, group-wise |
Adjoint weighted time constants using IFP
Parameter | Size | Description |
---|---|---|
ADJ_IFP_GEN_TIME | 6 | Adjoint-weighted neutron generation times using the iterated fission probability method: total, prompt, delayed |
ADJ_IFP_LIFETIME | 6 | Adjoint-weighted neutron lifetimes using the iterated fission probability method: total, prompt, delayed |
ADJ_IFP_IMP_BETA_EFF | 2D + 2 | Implicit estimator of adjoint-weighted effective delayed neutron fractions using the iterated fission probability method: total, group-wise |
ADJ_IFP_IMP_LAMBDA | 2D + 2 | Implicit estimator of adjoint-weighted of delayed neutron precursor decay constants using the iterated fission probability method: total, group-wise |
ADJ_IFP_ANA_BETA_EFF | 2D + 2 | Analog estimator of adjoint-weighted effective delayed neutron fractions using the iterated fission probability method: total, group-wise |
ADJ_IFP_ANA_LAMBDA | 2D + 2 | Analog estimator of adjoint-weighted of delayed neutron precursor decay constants using the iterated fission probability method: total, group-wise |
ADJ_IFP_ROSSI_ALPHA | 2 | Adjoint-weighted Rossi alpha using the iterated fission probability method |
Adjoint weighted time constants using perturbation technique
Parameter | Size | Description |
---|---|---|
ADJ_PERT_GEN_TIME | 2 | Adjoint-weighted neutron generation time using the perturbation technique |
ADJ_PERT_LIFETIME | 2 | Adjoint-weighted neutron lifetime using the perturbation technique |
ADJ_PERT_BETA_EFF | 2 | Adjoint-weighted effective delayed neutron fraction using the perturbation technique |
ADJ_PERT_ROSSI_ALPHA | 2 | Adjoint-weighted Rossi alpha using the perturbation technique |
Inverse neutron speed
Parameter | Size | Description |
---|---|---|
ANA_INV_SPD | 2 | Analog estimate of inverse neutron speed |
Analog slowing-down and thermal neutron lifetime (total/prompt/delayed)
Parameter | Size | Description |
---|---|---|
ANA_SLOW_TIME | 6 | Analog estimate of slowing-down time: total, prompt, delayed |
ANA_THERM_TIME | 6 | Analog estimate of thermal neutron lifetime: total, prompt, delayed |
ANA_THERM_FRAC | 6 | Analog estimate of neutron thermalisation fraction: total, prompt, delayed |
ANA_DELAYED_EMTIME | 2 | Analog estimate of delayed neutron emission time |
ANA_MEAN_NCOL | 4 | Analog estimate of average number of collisions per history: total and to fission |
Dynamic simulation
Parameter | Size | Description |
---|---|---|
DYN_NB | 1 | Number of time intervals defined in the time-bin structure (see tme card) |
DYN_TMIN | 1 | Minimum time boundary defined in the time-bin structure (see tme card) |
DYN_TMAX | 1 | Maximum time boundary defined in the time-bin structure (see tme card) |
DYN_TIMES | 2T + 2 | Time-bin boundaries defined in the time-bin structure (see tme card) |
DYN_POP | 2T | Neutron population at the end of the time interval |
DYN_PERIOD | 2T | Reactor period based on increase/decrease of neutron population during time interval |
Analog mean photon lifetime
Parameter | Size | Description |
---|---|---|
ANA_LIFETIME | 2 | Analog estimator of photon lifetime |
Homogenized group constants
Notes:
- Group constants are calculated by first homogenizing the geometry using a multi-group structure with H energy groups. The data is then collapsed into the final few-group structure with G groups using the infinite and leakage-corrected flux spectra.
- The methodology used in Serpent for spatial homogenization is described in a paper published in Annals of Nuclear Energy in 2016.^{[1]}
- The fundamental mode calculation is off by default, and invoked by the set fum option. Otherwise all values with B1 prefix are printed as zeros.
- The intermediate multi-group structure is defined using option set micro or set fum.
- The few-group structure is defined using option set nfg.
- The universes in which the group constants are calculated are listed in option set gcu. The calculation is performed for root universe 0 by default, and can be switched off with "set gcu -1".
- If data is produced in multiple universes within a single run, the data is assigned with different run indexes (idx)
- The parameter names can be listed in the set coefpara option, and they will be included in the group constant output file when the automated burnup sequence is invoked.
- The order in which two-dimensional data (scattering matrices, ADF and pin-power parameters) is printed in the [input].coe output file is different from what is listed below in update 2.1.24 and earlier versions.
Common parameters
Parameter | Size | Description |
---|---|---|
GC_UNIVERSE_NAME | (string) | Name of the universe where spatial homogenization was performed |
MICRO_NG | 1 | Number of energy groups in the intermediate multi-group structure (referred to as H below) |
MICRO_E | H + 1 | Group boundaries in the intermediate multi-group structure (in ascending order) |
MACRO_NG | 1 | Number of energy groups in the final few-group structure (referred to as G below) |
MACRO_E | G + 1 | Group boundaries in the final few-group structure (in descending order) |
Group constants homogenized in infinite spectrum
Parameter | Size | Description |
---|---|---|
INF_MICRO_FLX | 2H | Multi-group flux spectrum (integral, un-normalized) |
INF_FLX | 2G | Few-group flux (integral, normalized) |
INF_KINF | 2 | Infinite multiplication factor |
Reaction cross sections
Parameter | Size | Description |
---|---|---|
INF_TOT | 2G | Total cross section |
INF_CAPT | 2G | Capture cross section |
INF_FISS | 2G | Fission cross section |
INF_NSF | 2G | Fission neutron production cross section |
INF_KAPPA | 2G | Average deposited fission energy (MeV) |
INF_INVV | 2G | Inverse neutron speed (s/cm) |
INF_NUBAR | 2G | Average neutron yield |
INF_ABS | 2G | Absorption cross section (capture + fission) |
INF_REMXS | 2G | Removal cross section (group-removal + absorption) |
INF_RABSXS | 2G | Reduced absorption cross section (total - scattering production) |
Fission spectra
Parameter | Size | Description |
---|---|---|
INF_CHIT | 2G | Fission spectrum (total) |
INF_CHIP | 2G | Fission spectrum (prompt neutrons) |
INF_CHID | 2G | Fission spectrum (delayed neutrons) |
Scattering cross sections
Notes:
- Scattering production includes multiplying (n,2n), (n,3n), etc. reactions.
Parameter | Size | Description |
---|---|---|
INF_SCATT0 | 2G | Total P_{0} scattering cross section |
INF_SCATT1 | 2G | Total P_{1} scattering cross section |
INF_SCATT2 | 2G | Total P_{2} scattering cross section |
INF_SCATT3 | 2G | Total P_{3} scattering cross section |
INF_SCATT4 | 2G | Total P_{4} scattering cross section |
INF_SCATT5 | 2G | Total P_{5} scattering cross section |
INF_SCATT6 | 2G | Total P_{6} scattering cross section |
INF_SCATT7 | 2G | Total P_{7} scattering cross section |
INF_SCATTP0 | 2G | Total P_{0} scattering production cross section |
INF_SCATTP1 | 2G | Total P_{1} scattering production cross section |
INF_SCATTP2 | 2G | Total P_{2} scattering production cross section |
INF_SCATTP3 | 2G | Total P_{3} scattering production cross section |
INF_SCATTP4 | 2G | Total P_{4} scattering production cross section |
INF_SCATTP5 | 2G | Total P_{5} scattering production cross section |
INF_SCATTP6 | 2G | Total P_{6} scattering production cross section |
INF_SCATTP7 | 2G | Total P_{7} scattering production cross section |
Scattering matrices
Notes:
- Scattering production includes multiplying (n,2n), (n,3n), etc. reactions.
- The order of values ([input].coe) or value pairs ([input]_res.m) is: where refers to scattering from group g to g'.
- The data in the [input]_res.m file can be read into a G × G matrix with Matlab reshape-command, for example:
reshape(INF_S0(idx,1:2:end), G, G);
Parameter | Size | Description |
---|---|---|
INF_S0 | 2G^{2} | P_{0} scattering matrix |
INF_S1 | 2G^{2} | P_{1} scattering matrix |
INF_S2 | 2G^{2} | P_{2} scattering matrix |
INF_S3 | 2G^{2} | P_{3} scattering matrix |
INF_S4 | 2G^{2} | P_{4} scattering matrix |
INF_S5 | 2G^{2} | P_{5} scattering matrix |
INF_S6 | 2G^{2} | P_{6} scattering matrix |
INF_S7 | 2G^{2} | P_{7} scattering matrix |
INF_SP0 | 2G^{2} | P_{0} scattering production matrix |
INF_SP1 | 2G^{2} | P_{1} scattering production matrix |
INF_SP2 | 2G^{2} | P_{2} scattering production matrix |
INF_SP3 | 2G^{2} | P_{3} scattering production matrix |
INF_SP4 | 2G^{2} | P_{4} scattering production matrix |
INF_SP5 | 2G^{2} | P_{5} scattering production matrix |
INF_SP6 | 2G^{2} | P_{6} scattering production matrix |
INF_SP7 | 2G^{2} | P_{7} scattering production matrix |
Diffusion parameters
Notes:
- Calculation of sensible values for INF_TRANSPXS and INF_DIFFCOEF requires fine enough intermediate 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).
- Approximate effective treatment of Pm-149 production from Pm-148 and Pm-148m is not included in Pm-149 fission yield.
- All values for I-135, Pm-147, Pm-148 and Pm-148m not present in all versions.
Parameter | Size | Description |
---|---|---|
INF_I135_YIELD | 2G | Fission yield of I-135 (cumulative, includes all precursors) |
INF_XE135_YIELD | 2G | Fission yield of Xe-135 |
INF_XE135M_YIELD | 2G | Fission yield of Xe-135m |
INF_PM149_YIELD | 2G | Fission yield of Pm-149 (cumulative, includes all precursors) |
INF_SM149_YIELD | 2G | Fission yield of Sm-149 |
INF_I135_MICRO_ABS | 2G | Microscopic absorption cross section of I-135 |
INF_XE135_MICRO_ABS | 2G | Microscopic absorption cross section of Xe-135 |
INF_XE135M_MICRO_ABS | 2G | Microscopic absorption cross section of Xe-135m |
INF_PM147_MICRO_ABS | 2G | Microscopic absorption cross section of Pm-147 |
INF_PM148_MICRO_ABS | 2G | Microscopic absorption cross section of Pm-148 |
INF_PM148M_MICRO_ABS | 2G | Microscopic absorption cross section of Pm-148m |
INF_PM149_MICRO_ABS | 2G | Microscopic absorption cross section of Pm-149 |
INF_SM149_MICRO_ABS | 2G | Microscopic absorption cross section of Sm-149 |
INF_I135_MACRO_ABS | 2G | Macroscopic absorption cross section of I-135 |
INF_XE135_MACRO_ABS | 2G | Macroscopic absorption cross section of Xe-135 |
INF_XE135M_MACRO_ABS | 2G | Macroscopic absorption cross section of Xe-135m |
INF_PM147_MACRO_ABS | 2G | Macroscopic absorption cross section of Pm-147 |
INF_PM148_MACRO_ABS | 2G | Macroscopic absorption cross section of Pm-148 |
INF_PM148M_MACRO_ABS | 2G | Macroscopic absorption cross section of Pm-148M |
INF_PM149_MACRO_ABS | 2G | Macroscopic absorption cross section of Pm-149 |
INF_SM149_MACRO_ABS | 2G | Macroscopic absorption cross section of Sm-149 |
Poison universe-averaged densities
Notes:
- The universe-averaged atomic density is defined to be such that ADENS*MICRO_ABS is equal to MACRO_ABS.^{[3]}
Parameter | Size | Description |
---|---|---|
I135_ADENS | 2 | Universe-averaged atomic density of I-135 |
XE135_ADENS | 2 | Universe-averaged atomic density of Xe-135 |
XE135M_ADENS | 2 | Universe-averaged atomic density of Xe-135m |
PM147_ADENS | 2 | Universe-averaged atomic density of Pm-147 |
PM148_ADENS | 2 | Universe-averaged atomic density of Pm-147 |
PM148M_ADENS | 2 | Universe-averaged atomic density of Pm-148m |
PM149_ADENS | 2 | Universe-averaged atomic density of Pm-149 |
Poison decay constants
Parameter | Size | Description |
---|---|---|
PM147_LAMBDA | 1 | Decay constant of Pm-147 |
PM148_LAMBDA | 1 | Decay constant of Pm-147 |
PM148M_LAMBDA | 1 | Decay constant of Pm-148m |
PM149_LAMBDA | 1 | Decay constant of Pm-149 |
I135_LAMBDA | 1 | Decay constant of I-135 |
XE135_LAMBDA | 1 | Decay constant of Xe-135 |
XE135M_LAMBDA | 1 | Decay constant of Xe-135m |
I135_BR | 1 | Branching ratio of I-135 decay to Xe-135. Branching ratio of I-135 decay to Xe-135m is (1 - I135_BR). |
Group constants homogenized in leakage-corrected spectrum
Parameter | Size | Description |
---|---|---|
B1_MICRO_FLX | 2H | Multi-group flux spectrum (integral, un-normalized) |
B1_FLX | 2G | Few-group flux (integral, normalized) |
B1_KINF | 2 | Infinite multiplication factor |
B1_KEFF | 2 | Effective multiplication factor |
B1_B2 | 2 | Critical buckling |
B1_ERR | 2 | Absolute deviation of k_{eff} from unity |
Reaction cross sections
Parameter | Size | Description |
---|---|---|
B1_TOT | 2G | Total cross section |
B1_CAPT | 2G | Capture cross section |
B1_FISS | 2G | Fission cross section |
B1_NSF | 2G | Fission neutron production cross section |
B1_KAPPA | 2G | Average deposited fission energy (MeV) |
B1_INVV | 2G | Inverse neutron speed (s/cm) |
B1_NUBAR | 2G | Average neutron yield |
B1_ABS | 2G | Absorption cross section (capture + fission) |
B1_REMXS | 2G | Removal cross section (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 P_{0} scattering cross section |
B1_SCATT1 | 2G | Total P_{1} scattering cross section |
B1_SCATT2 | 2G | Total P_{2} scattering cross section |
B1_SCATT3 | 2G | Total P_{3} scattering cross section |
B1_SCATT4 | 2G | Total P_{4} scattering cross section |
B1_SCATT5 | 2G | Total P_{5} scattering cross section |
B1_SCATT6 | 2G | Total P_{6} scattering cross section |
B1_SCATT7 | 2G | Total P_{7} scattering cross section |
B1_SCATTP0 | 2G | Total P_{0} scattering production cross section |
B1_SCATTP1 | 2G | Total P_{1} scattering production cross section |
B1_SCATTP2 | 2G | Total P_{2} scattering production cross section |
B1_SCATTP3 | 2G | Total P_{3} scattering production cross section |
B1_SCATTP4 | 2G | Total P_{4} scattering production cross section |
B1_SCATTP5 | 2G | Total P_{5} scattering production cross section |
B1_SCATTP6 | 2G | Total P_{6} scattering production cross section |
B1_SCATTP7 | 2G | Total P_{7} scattering production cross section |
Scattering matrices
Notes:
- Scattering production includes multiplying (n,2n), (n,3n), etc. reactions.
- The order of values ([input].coe) or value pairs ([input]_res.m) is: where refers to scattering from group g to g'.
- The data in the [input]_res.m file can be read into a G × G matrix with Matlab reshape-command, for example:
reshape(B1_S0(idx,1:2:end), G, G).
Parameter | Size | Description |
---|---|---|
B1_S0 | 2G^{2} | P_{0} scattering matrix |
B1_S1 | 2G^{2} | P_{1} scattering matrix |
B1_S2 | 2G^{2} | P_{2} scattering matrix |
B1_S3 | 2G^{2} | P_{3} scattering matrix |
B1_S4 | 2G^{2} | P_{4} scattering matrix |
B1_S5 | 2G^{2} | P_{5} scattering matrix |
B1_S6 | 2G^{2} | P_{6} scattering matrix |
B1_S7 | 2G^{2} | P_{7} scattering matrix |
B1_SP0 | 2G^{2} | P_{0} scattering production matrix |
B1_SP1 | 2G^{2} | P_{1} scattering production matrix |
B1_SP2 | 2G^{2} | P_{2} scattering production matrix |
B1_SP3 | 2G^{2} | P_{3} scattering production matrix |
B1_SP4 | 2G^{2} | P_{4} scattering production matrix |
B1_SP5 | 2G^{2} | P_{5} scattering production matrix |
B1_SP6 | 2G^{2} | P_{6} scattering production matrix |
B1_SP7 | 2G^{2} | P_{7} scattering production matrix |
Diffusion parameters
Parameter | Size | Description |
---|---|---|
B1_TRANSPXS | 2G | Transport cross section (outscattering transport cross section collapsed with the critical spectrum when old B_{1} calculation mode is used, otherwise calculated from B1_DIFFCOEF) |
B1_DIFFCOEF | 2G | Diffusion coefficient calculated from during the fundamental mode calculation (old and new B_{1} and P_{1} calculation modes, or flux collapsed during the FM calculation mode) |
Poison cross sections
Notes:
- Printed only if poison cross section option is on (see set poi).
- Xe-135m values printed only if separate treatment of Xe-135m is on (see set poi).
Parameter | Size | Description |
---|---|---|
B1_I135_YIELD | 2G | Fission yield of I-135 (cumulative, includes all precursors) |
B1_XE135_YIELD | 2G | Fission yield of Xe-135 |
B1_XE135M_YIELD | 2G | Fission yield of Xe-135m |
B1_PM149_YIELD | 2G | Fission yield of Pm-149 (cumulative, includes all precursors) |
B1_SM149_YIELD | 2G | Fission yield of Sm-149 |
B1_I135_MICRO_ABS | 2G | Microscopic absorption cross section of I-135 |
B1_XE135_MICRO_ABS | 2G | Microscopic absorption cross section of Xe-135 |
B1_XE135M_MICRO_ABS | 2G | Microscopic absorption cross section of Xe-135m |
B1_PM149_MICRO_ABS | 2G | Microscopic absorption cross section of Pm-149 |
B1_SM149_MICRO_ABS | 2G | Microscopic absorption cross section of Sm-149 |
B1_XE135_MACRO_ABS | 2G | Macroscopic absorption cross section of Xe-135 |
B1_XE135M_MACRO_ABS | 2G | Macroscopic absorption cross section of Xe-135m |
B1_SM149_MACRO_ABS | 2G | Macroscopic absorption cross section of Sm-149 |
Delayed neutron data
Notes:
- The output consists of total, followed by D precursor group-wise values. In earlier versions, the output was fixed to 9 values independently of the library in use, with zero values corresponding to the empty precursor groups in the library.
- The actual number of groups depends on the cross section library used in the calculations. JEFF-3.1, JEFF-3.2 and later evaluations use 8 precursor groups, while earlier evaluations, as well as all ENDF/B and JENDL data is based on 6 groups.
Parameter | Size | Description |
---|---|---|
BETA_EFF | 2D + 2 | Effective delayed neutron fraction (currently calculated using the Meulekamp method) |
LAMBDA | 2D + 2 | Decay constants |
Assembly discontinuity factors
Notes:
- Calculation of assembly discontinuity factors requires the set adf option.
- Surface flux and current tallies are used to calculate the boundary currents and fluxes. Mid-point and corner values are approximated by integrating over a small surface segment.
- The surface and volume fluxes are flux densities, i.e. they are surface or volume integrated fluxes divided by the respective surface area or volume.
- The currents are surface integrated values.
- The net current is defined as current in subtracted with current out.
- When the homogenized region is surrounded by reflective boundary conditions (zero net-current) the homogeneous flux becomes flat and equal to the volume-averaged heterogeneous flux. When the net currents are non-zero, the homogeneous flux is obtained using the Built-in diffusion flux solver.
- The calculation currently supports only a limited number of surface types: infinite planes and square and hexagonal prisms.
- The order of surface and mid-point values for square prisms is: 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_SURFACE_TYPE | 1 | Surface type |
DF_SURFACE_N_PARAMS | 1 | Number of parameters defining the surface |
DF_SURFACE_PARAMS | N_{P} | List of parameters defining the surface |
DF_SYM | 1 | Symmetry option defined in the input |
DF_N_SURF | 1 | Number of surface values (denoted as N_{S} below) |
DF_N_CORN | 1 | Number of corner values (denoted as N_{C} below) |
DF_N_SGN | 1 | Number of sign moment values (denoted as N_{M} below) |
DF_VOLUME | 1 | Volume (3D) or cross sectional area (2D) of the homogenized cell |
DF_SURF_AREA | N_{S} | Area (3D) or perimeter length (2D) of the surface region |
DF_MID_AREA | N_{S} | Area (3D) or perimeter length (2D) of the mid-point region |
DF_CORN_AREA | N_{C} | Area (3D) or perimeter length (2D) of the corner region |
DF_SURF_IN_CURR | 2G × N_{S} | Inward surface currents |
DF_SURF_OUT_CURR | 2G × N_{S} | Outward surface currents |
DF_SURF_NET_CURR | 2G × N_{S} | Net surface currents |
DF_MID_IN_CURR | 2G × N_{S} | Inward mid-point currents |
DF_MID_OUT_CURR | 2G × N_{S} | Outward mid-point currents |
DF_MID_NET_CURR | 2G × N_{S} | Net mid-point currents |
DF_CORN_IN_CURR | 2G × N_{C} | Inward corner currents |
DF_CORN_OUT_CURR | 2G × N_{C} | Outward corner currents |
DF_CORN_NET_CURR | 2G × N_{C} | Net corner currents |
DF_HET_VOL_FLUX | 2G | Heterogeneous flux over homogenized cell |
DF_HET_SURF_FLUX | 2G × N_{S} | Heterogeneous surface fluxes |
DF_HET_CORN_FLUX | 2G × N_{C} | Heterogeneous corner fluxes |
DF_HOM_VOL_FLUX | 2G | Homogeneous flux over homogenized cell |
DF_HOM_SURF_FLUX | 2G × N_{S} | Homogeneous surface fluxes |
DF_HOM_CORN_FLUX | 2G × N_{C} | Homogeneous corner fluxes |
DF_SURF_DF | 2G × N_{S} | Surface discontinuity factors |
DF_CORN_DF | 2G × N_{C} | Corner discontinuity factors |
DF_SGN_SURF_IN_CURR | 2G × N_{M} | Inward sign moment weighted currents |
DF_SGN_SURF_OUT_CURR | 2G × N_{M} | Outward sign moment weighted currents |
DF_SGN_SURF_NET_CURR | 2G × N_{M} | Net sign moment weighted currents |
DF_SGN_HET_SURF_FLUX | 2G × N_{M} | Heterogeneous sign moment weighted surface fluxes |
DF_SGN_HOM_SURF_FLUX | 2G × N_{M} | Homogeneous sign moment weighted surface fluxes |
DF_SGN_SURF_DF | 2G × N_{M} | 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 N_{P} below) |
PPW_POW | 2G × N_{P} | Pin- and group-wise power distribution normalized to unity sum |
PPW_HOM_FLUX | 2G × N_{P} | Pin- and group-wise homogeneous flux distribution |
PPW_FF | 2G × N_{P} | 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 N_{S} below) |
ALB_IN_CURR | 2G × N_{S} | Groupwise incoming partial currents of albedo surface faces |
ALB_OUT_CURR | 2G^{2} × N_{S}^{2} | Outgoing group to group and face to face outgoing partial currents |
ALB_TOT_ALB | 2G^{2} | Total group to group albedos for the entire albedo surface |
ALB_PART_ALB | 2G^{2} × N_{S}^{2} | Partial group to group and face to face albedos |
Miscellaneous notes for other outputs
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.
- ^ Rintala, A., Valtavirta, V. and Leppänen, J.. Microscopic cross section calculation methodology in the Serpent 2 Monte Carlo code. Annals of Nuclear Energy, 164 (2021): 108603.