This page lists the output parameters in the main [input]_res.m output file.
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.
- The intermediate multi-group structure is defined using option set micro.
- The few-group structure is defined using option set nfg.
- The universes in which the group constants are calculated are listed in option set gcu. The calculation is performed for root universe 0 by default, and can be switched off with "set gcu -1".
- If data is produced in multiple universes within a single run, the data is assigned with different run indexes (idx)
- The parameter names can be listed in the set coefpara option, and they will be included in the group constant output file when the automated burnup sequence is invoked.
- The order in which two-dimensional data (scattering matrices, ADF and pin-power parameters) is printed in the [input].coe output file is different from what is listed below in update 2.1.24 and earlier versions.
Common parameters
Parameter
|
Size
|
Description
|
GC_UNIVERSE_NAME
|
(string)
|
Name of the universe where spatial homogenization was performed
|
MICRO_NG
|
1
|
Number of energy groups in the intermediate multi-group structure (referred to as H below)
|
MICRO_E
|
H + 1
|
Group boundaries in the intermediate multi-group structure (in ascending order)
|
MACRO_NG
|
1
|
Number of energy groups in the final few-group structure (referred to as G below)
|
MACRO_E
|
G + 1
|
Group boundaries in the final few-group structure (in descending order)
|
Group constants homogenized in infinite spectrum
Parameter
|
Size
|
Description
|
INF_MICRO_FLX
|
2H
|
Multi-group flux spectrum (integral, un-normalized)
|
INF_FLX
|
2G
|
Few-group flux (integral, normalized)
|
INF_KINF
|
2
|
Infinite multiplication factor
|
Reaction cross sections
Parameter
|
Size
|
Description
|
INF_TOT
|
2G
|
Total cross section
|
INF_CAPT
|
2G
|
Capture cross section
|
INF_FISS
|
2G
|
Fission cross section
|
INF_NSF
|
2G
|
Fission neutron production cross section
|
INF_KAPPA
|
2G
|
Average deposited fission energy (MeV)
|
INF_INVV
|
2G
|
Inverse neutron speed (s/cm)
|
INF_NUBAR
|
2G
|
Average neutron yield
|
INF_ABS
|
2G
|
Absorption cross section (capture + fission)
|
INF_REMXS
|
2G
|
Removal cross section (group-removal + absorption)
|
INF_RABSXS
|
2G
|
Reduced absorption cross section (total - scattering production)
|
Fission spectra
Parameter
|
Size
|
Description
|
INF_CHIT
|
2G
|
Fission spectrum (total)
|
INF_CHIP
|
2G
|
Fission spectrum (prompt neutrons)
|
INF_CHID
|
2G
|
Fission spectrum (delayed neutrons)
|
Scattering cross sections
Notes:
- Scattering production includes multiplying (n,2n), (n,3n), etc. reactions.
Parameter
|
Size
|
Description
|
INF_SCATT0
|
2G
|
Total P0 scattering cross section
|
INF_SCATT1
|
2G
|
Total P1 scattering cross section
|
INF_SCATT2
|
2G
|
Total P2 scattering cross section
|
INF_SCATT3
|
2G
|
Total P3 scattering cross section
|
INF_SCATT4
|
2G
|
Total P4 scattering cross section
|
INF_SCATT5
|
2G
|
Total P5 scattering cross section
|
INF_SCATT6
|
2G
|
Total P6 scattering cross section
|
INF_SCATT7
|
2G
|
Total P7 scattering cross section
|
INF_SCATTP0
|
2G
|
Total P0 scattering production cross section
|
INF_SCATTP1
|
2G
|
Total P1 scattering production cross section
|
INF_SCATTP2
|
2G
|
Total P2 scattering production cross section
|
INF_SCATTP3
|
2G
|
Total P3 scattering production cross section
|
INF_SCATTP4
|
2G
|
Total P4 scattering production cross section
|
INF_SCATTP5
|
2G
|
Total P5 scattering production cross section
|
INF_SCATTP6
|
2G
|
Total P6 scattering production cross section
|
INF_SCATTP7
|
2G
|
Total P7 scattering production cross section
|
Scattering matrices
Notes:
- Scattering production includes multiplying (n,2n), (n,3n), etc. reactions.
- The order of values ([input].coe) or value pairs ([input]_res.m) is:
where
refers to scattering from group g to g'.
- The data in the [input]_res.m file can be read into a G by G matrix with Matlab reshape-command, for example:
reshape(INF_S0(idx,1:2:end), G, G);
Parameter
|
Size
|
Description
|
INF_S0
|
4G2
|
P0 scattering matrix
|
INF_S1
|
4G2
|
P1 scattering matrix
|
INF_S2
|
4G2
|
P2 scattering matrix
|
INF_S3
|
4G2
|
P3 scattering matrix
|
INF_S4
|
4G2
|
P4 scattering matrix
|
INF_S5
|
4G2
|
P5 scattering matrix
|
INF_S6
|
4G2
|
P6 scattering matrix
|
INF_S7
|
4G2
|
P7 scattering matrix
|
INF_SP0
|
4G2
|
P0 scattering production matrix
|
INF_SP1
|
4G2
|
P1 scattering production matrix
|
INF_SP2
|
4G2
|
P2 scattering production matrix
|
INF_SP3
|
4G2
|
P3 scattering production matrix
|
INF_SP4
|
4G2
|
P4 scattering production matrix
|
INF_SP5
|
4G2
|
P5 scattering production matrix
|
INF_SP6
|
4G2
|
P6 scattering production matrix
|
INF_SP7
|
4G2
|
P7 scattering production matrix
|
Diffusion parameters
Notes:
- Calculation of sensible values for INF_TRANSPXS and INF_DIFFCOEF requires fine enough intermediate multi-group structure.
- The cumulative migration method [2] was first developed for the OpenMC code. Currently the method works only when the homogenized region covers the entire geometry, and is surrounded by periodic or reflective boundary conditions.
- Calculation of TRC_TRANSPXS and TRC_DIFFCOEF requires defining energy-dependent correction factors using the set trc option.
- Calculation of CMM_TRANSPXS and CMM_DIFFCOEF requires that their calculation is not switched off using the set cmm option.
Parameter
|
Size
|
Description
|
INF_TRANSPXS
|
2G
|
Transport cross section (calculated using the out-scattering approximation)
|
INF_DIFFCOEF
|
2G
|
Diffusion coefficient (calculated using the out-scattering approximation)
|
CMM_TRANSPXS
|
2G
|
Transport cross section calculated using the cumulative migration method (equivalent with the in-scattering approximation)
|
CMM_TRANSPXS_X
|
2G
|
X-component of the directional transport cross section calculated using the cumulative migration method (equivalent with the in-scattering approximation)
|
CMM_TRANSPXS_Y
|
2G
|
Y-component of the directional transport cross section calculated using the cumulative migration method (equivalent with the in-scattering approximation)
|
CMM_TRANSPXS_Z
|
2G
|
Z-component of the directional transport cross section calculated using the cumulative migration method (equivalent with the in-scattering approximation)
|
CMM_DIFFCOEF
|
2G
|
Diffusion coefficient calculated using the cumulative migration method (equivalent with the in-scattering approximation)
|
CMM_DIFFCOEF_X
|
2G
|
X-component of the directional diffusion coefficient calculated using the cumulative migration method (equivalent with the in-scattering approximation)
|
CMM_DIFFCOEF_Y
|
2G
|
Y-component of the directional diffusion coefficient calculated using the cumulative migration method (equivalent with the in-scattering approximation)
|
CMM_DIFFCOEF_Z
|
2G
|
Z-component of the directional diffusion coefficient calculated using the cumulative migration method (equivalent with the in-scattering approximation)
|
TRC_TRANSPXS
|
2G
|
Transport cross section calculated by applying user-defined transport correction factors to total cross section
|
TRC_DIFFCOEF
|
2G
|
Diffusion coefficient calculated by applying user-defined transport correction factors to total cross section
|
Poison cross sections
Notes:
- Printed only if poison cross section option is on (see set poi).
Parameter
|
Size
|
Description
|
INF_I135_YIELD
|
2G
|
Fission yield of I-135 (cumulative, includes all precursors)
|
INF_XE135_YIELD
|
2G
|
Fission yield of Xe-135
|
INF_PM149_YIELD
|
2G
|
Fission yield of Pm-149 (cumulative, includes all precursors)
|
INF_SM149_YIELD
|
2G
|
Fission yield of Sm-149
|
INF_I135_MICRO_ABS
|
2G
|
Microscopic absorption cross section of I-135
|
INF_XE135_MICRO_ABS
|
2G
|
Microscopic absorption cross section of Xe-135
|
INF_PM149_MICRO_ABS
|
2G
|
Microscopic absorption cross section of Pm-149
|
INF_SM149_MICRO_ABS
|
2G
|
Microscopic absorption cross section of Sm-149
|
INF_XE135_MACRO_ABS
|
2G
|
Macroscopic absorption cross section of Xe-135
|
INF_SM149_MACRO_ABS
|
2G
|
Macroscopic absorption cross section of Sm-149
|
Group constants homogenized in leakage-corrected spectrum
Parameter
|
Size
|
Description
|
B1_MICRO_FLX
|
2H
|
Multi-group flux spectrum (integral, un-normalized)
|
B1_FLX
|
2G
|
Few-group flux (integral, normalized)
|
B1_KINF
|
2
|
Infinite multiplication factor
|
B1_KEFF
|
2
|
Effective multiplication factor
|
B1_B2
|
2
|
Critical buckling
|
B1_ERR
|
2
|
Absolute deviation of keff from unity
|
Reaction cross sections
Parameter
|
Size
|
Description
|
B1_TOT
|
2G
|
Total cross section
|
B1_CAPT
|
2G
|
Capture cross section
|
B1_FISS
|
2G
|
Fission cross section
|
B1_NSF
|
2G
|
Fission neutron production cross section
|
B1_KAPPA
|
2G
|
Average deposited fission energy (MeV)
|
B1_INVV
|
2G
|
Inverse neutron speed (s/cm)
|
B1_NUBAR
|
2G
|
Average neutron yield
|
B1_ABS
|
2G
|
Absorption cross section (capture + fission)
|
B1_REMXS
|
2G
|
Removal cross section (group-removal + absorption)
|
B1_RABSXS
|
2G
|
Reduced absorption cross section (total - scattering production)
|
Fission spectra
Parameter
|
Size
|
Description
|
B1_CHIT
|
2G
|
Fission spectrum (total)
|
B1_CHIP
|
2G
|
Fission spectrum (prompt neutrons)
|
B1_CHID
|
2G
|
Fission spectrum (delayed neutrons)
|
Scattering cross sections
Notes:
- Scattering production includes multiplying (n,2n), (n,3n), etc. reactions.
Parameter
|
Size
|
Description
|
B1_SCATT0
|
2G
|
Total P0 scattering cross section
|
B1_SCATT1
|
2G
|
Total P1 scattering cross section
|
B1_SCATT2
|
2G
|
Total P2 scattering cross section
|
B1_SCATT3
|
2G
|
Total P3 scattering cross section
|
B1_SCATT4
|
2G
|
Total P4 scattering cross section
|
B1_SCATT5
|
2G
|
Total P5 scattering cross section
|
B1_SCATT6
|
2G
|
Total P6 scattering cross section
|
B1_SCATT7
|
2G
|
Total P7 scattering cross section
|
B1_SCATTP0
|
2G
|
Total P0 scattering production cross section
|
B1_SCATTP1
|
2G
|
Total P1 scattering production cross section
|
B1_SCATTP2
|
2G
|
Total P2 scattering production cross section
|
B1_SCATTP3
|
2G
|
Total P3 scattering production cross section
|
B1_SCATTP4
|
2G
|
Total P4 scattering production cross section
|
B1_SCATTP5
|
2G
|
Total P5 scattering production cross section
|
B1_SCATTP6
|
2G
|
Total P6 scattering production cross section
|
B1_SCATTP7
|
2G
|
Total P7 scattering production cross section
|
Scattering matrices
Notes:
- Scattering production includes multiplying (n,2n), (n,3n), etc. reactions.
- The order of values ([input].coe) or value pairs ([input]_res.m) is:
where
refers to scattering from group g to g'.
- The data in the _res.m file can be read into a G by G matrix with Matlab reshape-command, for example:
reshape(B1_S0(idx,1:2:end), G, G).
Parameter
|
Size
|
Description
|
B1_S0
|
4G2
|
P0 scattering matrix
|
B1_S1
|
4G2
|
P1 scattering matrix
|
B1_S2
|
4G2
|
P2 scattering matrix
|
B1_S3
|
4G2
|
P3 scattering matrix
|
B1_S4
|
4G2
|
P4 scattering matrix
|
B1_S5
|
4G2
|
P5 scattering matrix
|
B1_S6
|
4G2
|
P6 scattering matrix
|
B1_S7
|
4G2
|
P7 scattering matrix
|
B1_SP0
|
4G2
|
P0 scattering production matrix
|
B1_SP1
|
4G2
|
P1 scattering production matrix
|
B1_SP2
|
4G2
|
P2 scattering production matrix
|
B1_SP3
|
4G2
|
P3 scattering production matrix
|
B1_SP4
|
4G2
|
P4 scattering production matrix
|
B1_SP5
|
4G2
|
P5 scattering production matrix
|
B1_SP6
|
4G2
|
P6 scattering production matrix
|
B1_SP7
|
4G2
|
P7 scattering production matrix
|
Diffusion parameters
Parameter
|
Size
|
Description
|
B1_TRANSPXS
|
2G
|
Transport cross section (outscattering transport cross section collapsed with the critical spectrum when old B1 calculation mode is used, otherwise calculated from B1_DIFFCOEF)
|
B1_DIFFCOEF
|
2G
|
Diffusion coefficient calculated from during the fundamental mode calculation (old and new B1 and P1 calculation modes, or flux collapsed during the FM calculation mode)
|
Poison cross sections
Notes:
- Printed only if poison cross section option is on (see set poi).
Parameter
|
Size
|
Description
|
B1_I135_YIELD
|
2G
|
Fission yield of I-135 (cumulative, includes all precursors)
|
B1_XE135_YIELD
|
2G
|
Fission yield of Xe-135
|
B1_PM149_YIELD
|
2G
|
Fission yield of Pm-149 (cumulative, includes all precursors)
|
B1_SM149_YIELD
|
2G
|
Fission yield of Sm-149
|
B1_I135_MICRO_ABS
|
2G
|
Microscopic absorption cross section of I-135
|
B1_XE135_MICRO_ABS
|
2G
|
Microscopic absorption cross section of Xe-135
|
B1_PM149_MICRO_ABS
|
2G
|
Microscopic absorption cross section of Pm-149
|
B1_SM149_MICRO_ABS
|
2G
|
Microscopic absorption cross section of Sm-149
|
B1_XE135_MACRO_ABS
|
2G
|
Macroscopic absorption cross section of Xe-135
|
B1_SM149_MACRO_ABS
|
2G
|
Macroscopic absorption cross section of Sm-149
|
Delayed neutron data
Notes:
- The output always consists of 9 values: total, followed by precursor group-wise values. If the number of groups is 6, the last two values are zero.
- The actual number of groups depends on the cross section library used in the calculations. JEFF-3.1, JEFF.3.2 and later evaluations use 8 precursor groups, while earlier evaluations, as well as all ENDF/B and JENDL data is based on 6 groups.
Parameter
|
Size
|
Description
|
BETA_EFF
|
9
|
Effective delayed neutron fraction (currently calculated using the Meulekamp method)
|
LAMBDA
|
9
|
Decay constants
|
Assembly discontinuity factors
Notes:
- Calculation of assembly discontinuity factors requires the set adf option.
- Surface flux and current tallies are used to calculate the boundary currents and fluxes. Mid-point and corner values are approximated by integrating over a small surface segment.
- The surface and volume fluxes are flux densities, i.e. they are surface or volume integrated fluxes divided by the respective surface area or volume.
- The currents are surface integrated values.
- When the homogenized region is surrounded by reflective boundary conditions (zero net-current) the homogeneous flux becomes flat and equal to the volume-averaged heterogeneous flux. When the net currents are non-zero, the homogeneous flux is obtained using the Built-in diffusion flux solver.
- The calculation currently supports only a limited number of surface types: infinite planes and square and hexagonal prisms.
- The order of surface and mid-point values for square prisms is:
and the order of corner values:
where
refers to parameter
on surface/corner k and energy group g.
- The order of surface values for Y-type hexagonal prims runs clockwise starting from the north, i.e. N, NE, SE, S, SW, NW. The corner values run counterclockwise starting from east, i.e. E, NE, NW, W, SW, SE.
- The order of surface values for X-type hexagonal prims runs counterclockwise starting from the east, i.e. E, NE, NW, W, SW, SE. The corner values run clockwise starting from north, i.e. N, NE, SE, S, SW, NW.
- The sign moment weighted parameters are calculated only for surface types sqc, rect and hexxc.
- The convention of sign moment directions follows that of the nodal neutronics program Ants.
- The ADF symmetry options on set adf card are currently not used for sign moment weighted parameters.
Parameter
|
Size
|
Description
|
DF_SURFACE
|
(string)
|
Name of the surface used for the calculation
|
DF_SYM
|
1
|
Symmetry option defined in the input
|
DF_N_SURF
|
1
|
Number of surface values (denoted as NS below)
|
DF_N_CORN
|
1
|
Number of corner values (denoted as NC below)
|
DF_VOLUME
|
1
|
Volume (3D) or cross sectional area (2D) of the homogenized cell
|
DF_SURF_AREA
|
NS
|
Area (3D) or perimeter length (2D) of the surface region
|
DF_MID_AREA
|
NS
|
Area (3D) or perimeter length (2D) of the mid-point region
|
DF_CORN_AREA
|
NC
|
Area (3D) or perimeter length (2D) of the corner region
|
DF_SURF_IN_CURR
|
2G NS
|
Inward surface currents
|
DF_SURF_OUT_CURR
|
2G NS
|
Outward surface currents
|
DF_SURF_NET_CURR
|
2G NS
|
Net surface currents
|
DF_MID_IN_CURR
|
2G NS
|
Inward mid-point currents
|
DF_MID_OUT_CURR
|
2G NS
|
Outward mid-point currents
|
DF_MID_NET_CURR
|
2G NS
|
Net mid-point currents
|
DF_CORN_IN_CURR
|
2G NC
|
Inward corner currents
|
DF_CORN_OUT_CURR
|
2G NC
|
Outward corner currents
|
DF_CORN_NET_CURR
|
2G NC
|
Net corner currents
|
DF_HET_VOL_FLUX
|
2G
|
Heterogeneous flux over homogenized cell
|
DF_HET_SURF_FLUX
|
2G NS
|
Heterogeneous surface fluxes
|
DF_HET_CORN_FLUX
|
2G NS
|
Heterogeneous corner fluxes
|
DF_HOM_VOL_FLUX
|
2G
|
Homogeneous flux over homogenized cell
|
DF_HOM_SURF_FLUX
|
2G NC
|
Homogeneous surface fluxes
|
DF_HOM_CORN_FLUX
|
2G NC
|
Homogeneous corner fluxes
|
DF_SURF_DF
|
2G NS
|
Surface discontinuity factors
|
DF_CORN_DF
|
2G NC
|
Corner discontinuity factors
|
DF_SGN_SURF_IN_CURR
|
2G NS
|
Inward sign moment weighted currents
|
DF_SGN_SURF_OUT_CURR
|
2G NS
|
Outward sign moment weighted currents
|
DF_SGN_SURF_NET_CURR
|
2G NS
|
Net sign moment weighted currents
|
DF_SGN_HET_SURF_FLUX
|
2G NS
|
Heterogeneous sign moment weighted surface fluxes
|
DF_SGN_HOM_SURF_FLUX
|
2G NS
|
Homogeneous sign moment weighted surface fluxes
|
DF_SGN_SURF_DF
|
2G NS
|
Sign moment weighted surface discontinuity factors
|
Pin-power form factors
Notes:
- Calculation of pin-power form factors requires the set ppw option.
- The power distribution is calculated by tallying the few-group fission energy deposition in each lattice position and dividing the values with the total energy produced in the universe (sum over all values of PPW_POW equals 1).
- The calculation of form factors depends on the boundary conditions:
- If the homogenized region is surrounded by reflective boundary conditions (zero net-current), the homogeneous flux becomes flat and equal to the volume-averaged heterogeneous flux.
- When the net currents are non-zero, the homogeneous flux is obtained using the built-in diffusion flux solver. The form-factors (PPW_FF) are obtained by dividing the pin- and group-wise powers with the corresponding homogeneous diffusion flux (PPW_HOM_FLUX).
- Running the diffusion flux solver currently requires ADF calculation.
- The order of values is:
where
refers to parameter
of pin n and energy group g. For example, two-group power distributions in a 17 x 17 lattice can be converted into matrix form using the reshape-command in Matlab:
P1 = reshape(PPW_POW(1, 1:4:end), 17, 17);
P2 = reshape(PPW_POW(1, 3:4:end), 17, 17);
- Symmetry used in the lattice may result in some pin powers and form factors to be for example 1/2, 1/4 or 1/8 of their true value, which have to be corrected during post processing of the values.
Parameter
|
Size
|
Description
|
PPW_LATTICE
|
(string)
|
Name of the lattice used for the calculation
|
PPW_LATTICE_TYPE
|
1
|
Lattice type (corresponds to the lat-card)
|
PPW_PINS
|
1
|
Number of pin positions in the lattice (denoted as NP below)
|
PPW_POW
|
2G NP
|
Pin- and group-wise power distribution normalized to unity sum
|
PPW_HOM_FLUX
|
2G NP
|
Pin- and group-wise homogeneous flux distribution
|
PPW_FF
|
2G NP
|
Pin- and group-wise form factors
|
Albedos
Notes:
- Calculation of albedos requires the set alb option.
- The order of the surfaces should be the same as for the ADFs.
- The order of ALB_IN_CURR is
where
refers to incoming partial current of surface k of group g.
- The order of ALB_OUT_CURR is
where
refers to outgoing partial current of surface k' of group g' which has entered the albedo surface through surface k and group g.
- The order of ALB_TOT_ALB is
where
refers to albedo from group g to g'.
- The order of ALB_PART_ALB is
where
refers to albedo of surface k' of group g' which has entered the albedo surface through surface k and group g.
- For example, two-group hexagonal partial albedos can be converted into matrix form using the reshape-command in Matlab with the notation part_alb(g', k', g, k)
as
part_alb = reshape(ALB_PART_ALB(1, 1:2:end), 2, 6, 2, 6)
Parameter
|
Size
|
Description
|
ALB_SURFACE
|
(string)
|
Name of the surface used for the calculation
|
ALB_FLIP_DIR
|
1
|
|
ALB_N_SURF
|
1
|
Number of albedo surface faces (denoted as NS below)
|
ALB_IN_CURR
|
2G NS
|
Groupwise incoming partial currents of albedo surface faces
|
ALB_OUT_CURR
|
2G2 NS2
|
Outgoing group to group and face to face outgoing partial currents
|
ALB_TOT_ALB
|
2G2
|
Total group to group albedos for the entire albedo surface
|
ALB_PART_ALB
|
2G2 NS2
|
Partial group to group and face to face albedos
|
Miscellaneous notes for other outputs
Delayed neutrons accounted for in ANA_KEFF
Since Serpent 2.1.23, ANA_KEFF estimator is calculated separately for delayed neutrons. The first two values are total, 3-4 are prompt neutron multiplication only and 5-6 delayed neutron multiplication only. [3]
References
- ^ Leppänen, J., Pusa, M. and Fridman, E. "Overview of methodology for spatial homogenization in the Serpent 2 Monte Carlo code." Ann. Nucl. Energy, 96 (2016) 126-136.
- ^ Liu, Z., Smith, K., Forget, B. and Ortensi, J."Cumulative migration method for computing rigorous diffusion coefficients and transport cross sections from Monte Carlo." Ann. Nucl. Energy, 118 (2018) 507-516.
- ^
http://ttuki.vtt.fi/serpent/viewtopic.php?f=25&t=1885&p=4469