Abstract
We have obtained results for a large sodium-cooled fast breeder reactor benchmark using data from the ENDF/B-VI and from Revision 1 of the JEF-1 (JEF-1.1) evaluation. The required cross sections were processed with the NJOY code system (Version 89.62) and homogenized with the spectrum cell code MICROX-2. Multigroup transport-theory calculations in 33 neutron groups (forward and adjoint) were performed using the two-dimensional code TWODANT and kinetic parameters were determined using the first-order perturbation-theory code PERT-V. We calculated eigenvalues, neutron balance data, global and regional breeding and conversion ratios, central rate ratios and reactivity worths with and without sodium, effective delayed neutron fraction and inhour reactivity, regional sodium void reactivity, and isothermal core fuel Doppler-reactivities. In particular, it is shown that good agreement (generally within one standard deviation) is achieved between these results and the average values over sixteen benchmark solutions obtained in the past. The eigenvalues predicted with ENDF/B-VI are up to 0.7% larger than those calculated with JEF-1.1 cross sections. This discrepancy is mainly due to different inelastic scattering cross sections for {sup 23}Na and {sup 238}U, and to different fast fission and nubar data for {sup 239}Pu. (author) 5 figs., 30 tabs., 24 refs.
Pelloni, S
[1]
- Paul Scherrer Inst. (PSI), Villigen (Switzerland)
Citation Formats
Pelloni, S.
Comparison calculation of a large sodium-cooled fast breeder reactor using the cell code MICROX-2 in connection with ENDF/B-VI and JEF-1.1 neutron data.
Switzerland: N. p.,
1992.
Web.
Pelloni, S.
Comparison calculation of a large sodium-cooled fast breeder reactor using the cell code MICROX-2 in connection with ENDF/B-VI and JEF-1.1 neutron data.
Switzerland.
Pelloni, S.
1992.
"Comparison calculation of a large sodium-cooled fast breeder reactor using the cell code MICROX-2 in connection with ENDF/B-VI and JEF-1.1 neutron data."
Switzerland.
@misc{etde_10157694,
title = {Comparison calculation of a large sodium-cooled fast breeder reactor using the cell code MICROX-2 in connection with ENDF/B-VI and JEF-1.1 neutron data}
author = {Pelloni, S}
abstractNote = {We have obtained results for a large sodium-cooled fast breeder reactor benchmark using data from the ENDF/B-VI and from Revision 1 of the JEF-1 (JEF-1.1) evaluation. The required cross sections were processed with the NJOY code system (Version 89.62) and homogenized with the spectrum cell code MICROX-2. Multigroup transport-theory calculations in 33 neutron groups (forward and adjoint) were performed using the two-dimensional code TWODANT and kinetic parameters were determined using the first-order perturbation-theory code PERT-V. We calculated eigenvalues, neutron balance data, global and regional breeding and conversion ratios, central rate ratios and reactivity worths with and without sodium, effective delayed neutron fraction and inhour reactivity, regional sodium void reactivity, and isothermal core fuel Doppler-reactivities. In particular, it is shown that good agreement (generally within one standard deviation) is achieved between these results and the average values over sixteen benchmark solutions obtained in the past. The eigenvalues predicted with ENDF/B-VI are up to 0.7% larger than those calculated with JEF-1.1 cross sections. This discrepancy is mainly due to different inelastic scattering cross sections for {sup 23}Na and {sup 238}U, and to different fast fission and nubar data for {sup 239}Pu. (author) 5 figs., 30 tabs., 24 refs.}
place = {Switzerland}
year = {1992}
month = {Feb}
}
title = {Comparison calculation of a large sodium-cooled fast breeder reactor using the cell code MICROX-2 in connection with ENDF/B-VI and JEF-1.1 neutron data}
author = {Pelloni, S}
abstractNote = {We have obtained results for a large sodium-cooled fast breeder reactor benchmark using data from the ENDF/B-VI and from Revision 1 of the JEF-1 (JEF-1.1) evaluation. The required cross sections were processed with the NJOY code system (Version 89.62) and homogenized with the spectrum cell code MICROX-2. Multigroup transport-theory calculations in 33 neutron groups (forward and adjoint) were performed using the two-dimensional code TWODANT and kinetic parameters were determined using the first-order perturbation-theory code PERT-V. We calculated eigenvalues, neutron balance data, global and regional breeding and conversion ratios, central rate ratios and reactivity worths with and without sodium, effective delayed neutron fraction and inhour reactivity, regional sodium void reactivity, and isothermal core fuel Doppler-reactivities. In particular, it is shown that good agreement (generally within one standard deviation) is achieved between these results and the average values over sixteen benchmark solutions obtained in the past. The eigenvalues predicted with ENDF/B-VI are up to 0.7% larger than those calculated with JEF-1.1 cross sections. This discrepancy is mainly due to different inelastic scattering cross sections for {sup 23}Na and {sup 238}U, and to different fast fission and nubar data for {sup 239}Pu. (author) 5 figs., 30 tabs., 24 refs.}
place = {Switzerland}
year = {1992}
month = {Feb}
}