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 B1 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 B1 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 .coef output 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:
- 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.
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 B1 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 (.coe) or value pairs (_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 (calculated from diffusion coefficient)
|
B1_DIFFCOEF
|
2G
|
Diffusion coefficient
|
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.
- Fluxes and currents are normalized average 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 and mid-point values for hexagonal prims runs clockwise starting from the north (Y-type) or east (X-type) face. The corner values start from the next corner in clockwise direction.
- Note to developers: the description may be wrong for for X-type hexagonal prism.
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
|
NS
|
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 flux
|
DF_HET_CORN_FLUX
|
2G NC
|
Heterogeneous corner flux
|
DF_HOM_VOL_FLUX
|
2G
|
Homogeneous flux over homogenized cell
|
DF_HOM_SURF_FLUX
|
2G NC
|
Homogeneous surface flux
|
DF_HOM_CORN_FLUX
|
2G NC
|
Homogeneous corner flux
|
DF_SURF_DF
|
2G NC
|
Surface discontinuity factors
|
DF_CORN_DF
|
2G NC
|
Corner 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. Variables PPW_HOM_FLUX and PPW_FF are then omitted.
- 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
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);
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-wise power distribution
|
PPW_HOM_FLUX
|
2G NP
|
Pin-wise homogeneous flux distribution
|
PPW_FF
|
2G NP
|
Pin-wise form factors
|
Albedos
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. and Forget, B. "A Cumulative Migration Method for Computing Rigorous Transport Cross Sections and Diffusion Coefficients for LWR Lattices with Monte Carlo." In proc. PHYSOR 2016, Sun Valley, Idaho, May 1-5, 2016.