Output parameters

From Serpent Wiki
Jump to: navigation, search

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: \Sigma_{1,1} \, \Sigma_{1,2} \, ... \, \Sigma_{2,1} \, \Sigma_{2,2}  \, ... where \Sigma_{g,g'} refers to scattering from group g to g'.
  • The data in the [input]_res.m file can be read into a G 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: \Sigma_{1,1} \, \Sigma_{1,2} \, ... \, \Sigma_{2,1} \, \Sigma_{2,2}  \, ... where \Sigma_{g,g'} refers to scattering from group g to g'.
  • The data in the _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: X_{\mathrm{W},1} \, X_{\mathrm{W},2} \, ... \, X_{\mathrm{S},1} \, X_{\mathrm{S},2} \, ... \, X_{\mathrm{E},1} \, X_{\mathrm{E},2} \, ... \, X_{\mathrm{N},1} \, X_{\mathrm{N},2} \, ... and the order of corner values: X_{\mathrm{NW},1} \, X_{\mathrm{NW},2} \, ... \, X_{\mathrm{NE},1} \, X_{\mathrm{NE},2} \, ... \, X_{\mathrm{SE},1} \, X_{\mathrm{SE},2} \, ... \, X_{\mathrm{SW},1} \, X_{\mathrm{SW},2} \, ... where X_{k,g} refers to parameter X on surface/corner k and energy group g.
  • The order of surface and mid-point values for Y-type hexagonal prims runs clockwise starting from the north. The corner values start from the next corner in clockwise direction.
  • The order of surface and mid-point values for X-type hexagonal prims runs counterclockwise starting from the east. Confirm the direction of corner values.
  • Note that one should confirm the order of values in the hexagonal prisms.
  • 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.
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 \times NS Inward surface currents
DF_SURF_OUT_CURR 2G \times NS Outward surface currents
DF_SURF_NET_CURR 2G \times NS Net surface currents
DF_MID_IN_CURR 2G \times NS Inward mid-point currents
DF_MID_OUT_CURR 2G \times NS Outward mid-point currents
DF_MID_NET_CURR 2G \times NS Net mid-point currents
DF_CORN_IN_CURR 2G \times NC Inward corner currents
DF_CORN_OUT_CURR 2G \times NC Outward corner currents
DF_CORN_NET_CURR 2G \times NC Net corner currents
DF_HET_VOL_FLUX 2G Heterogeneous flux over homogenized cell
DF_HET_SURF_FLUX 2G \times NS Heterogeneous surface fluxes
DF_HET_CORN_FLUX 2G \times NC Heterogeneous corner fluxes
DF_HOM_VOL_FLUX 2G Homogeneous flux over homogenized cell
DF_HOM_SURF_FLUX 2G \times NC Homogeneous surface fluxes
DF_HOM_CORN_FLUX 2G \times NC Homogeneous corner fluxes
DF_SURF_DF 2G \times NC Surface discontinuity factors
DF_CORN_DF 2G \times NC Corner discontinuity factors
DF_SGN_SURF_IN_CURR 2G \times NS Inward sign moment weighted currents
DF_SGN_SURF_OUT_CURR 2G \times NS Outward sign moment weighted currents
DF_SGN_SURF_NET_CURR 2G \times NS Net sign moment weighted currents
DF_SGN_HET_SURF_FLUX 2G \times NS Heterogeneous sign moment weighted surface fluxes
DF_SGN_HOM_SURF_FLUX 2G \times NS Homogeneous sign moment weighted surface fluxes
DF_SGN_SURF_DF 2G \times 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:
    1. If the homogenized region is surrounded by reflective boundary conditions (zero net-current), the homogeneous flux becomes flat and equal to the volume-averaged heterogeneous flux.
    2. When the net currents are non-zero, the homogeneous flux is obtained using the built-in diffusion flux solver. The form-factors (PPW_FF) are obtained by dividing the pin- and group-wise powers with the corresponding homogeneous diffusion flux (PPW_HOM_FLUX).
  • Running the diffusion flux solver currently requires ADF calculation.
  • The order of values is: X_{1,1} \, X_{1,2} \, ... \, X_{2,1} \, X_{2,2} \, ... where X_{n,g} refers to parameter X of pin n and energy group g. For example, two-group power distributions in a 17 x 17 lattice can be converted into matrix form using the reshape-command in Matlab:
P1 = reshape(PPW_POW(1, 1:4:end), 17, 17);
P2 = reshape(PPW_POW(1, 3:4:end), 17, 17);
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 \times NP Pin- and group-wise power distribution normalized to unity sum
PPW_HOM_FLUX 2G \times NP Pin- and group-wise homogeneous flux distribution
PPW_FF 2G \times NP Pin- and group-wise form factors

Albedos

Notes:

  • Calculation of albedos requires the set alb option.
  • The order of the surfaces should be the same as for the ADFs.
  • The order of ALB_IN_CURR is J_{1,1} , J_{1,2} , \ldots , J_{2,1} , J_{2,2} , \ldots where J_{k,g} refers to incoming partial current of surface k of group g.
  • The order of ALB_OUT_CURR is J_{1,1,1,1} , J_{1,1,1,2} , \ldots , J_{1,1,2,1} , J_{1,1,2,2} , \ldots , J_{1,2,1,1} , J_{1,2,1,2} , \ldots , J_{2,1,1,1} , J_{2,1,1,2} , \ldots where J_{k,g,k',g'} refers to outgoing partial current of surface k' of group g' which has entered the albedo surface through surface k and group g.
  • The order of ALB_TOT_ALB is \alpha_{1,1} , \alpha_{1,2} , \ldots ,  \alpha_{2,1} ,  \alpha_{2,2} , \ldots where \alpha_{g,g'} refers to albedo from group g to g'.
  • The order of ALB_PART_ALB is \alpha_{1,1,1,1} , \alpha_{1,1,1,2} , \ldots , \alpha_{1,1,2,1} , \alpha_{1,1,2,2} , \ldots , \alpha_{1,2,1,1} , \alpha_{1,2,1,2} , \ldots , \alpha_{2,1,1,1} , \alpha_{2,1,1,2} , \ldots where \alpha_{k,g,k',g'} refers to albedo of surface k' of group g' which has entered the albedo surface through surface k and group g.
  • For example, two-group hexagonal partial albedos can be converted into matrix form using the reshape-command in Matlab with the notation part_alb(g', k', g, k) = \alpha_{k,g,k',g'} as
part_alb = reshape(ALB_PART_ALB(1, 1:2:end), 2, 6, 2, 6)
Parameter Size Description
ALB_SURFACE (string) Name of the surface used for the calculation
ALB_FLIP_DIR 1
ALB_N_SURF 1 Number of albedo surface faces (denoted as NS below)
ALB_IN_CURR 2G \times NS Groupwise incoming partial currents of albedo surface faces
ALB_OUT_CURR 2G2 \times 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 \times 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

  1. ^ Leppänen, J., Pusa, M. and Fridman, E. "Overview of methodology for spatial homogenization in the Serpent 2 Monte Carlo code." Ann. Nucl. Energy, 96 (2016) 126-136.
  2. ^ Liu, Z., Smith, K., Forget, B. and Ortensi, J."Cumulative migration method for computing rigorous diffusion coefficients and transport cross sections from Monte Carlo." Ann. Nucl. Energy, 118 (2018) 507-516.
  3. ^ http://ttuki.vtt.fi/serpent/viewtopic.php?f=25&t=1885&p=4469