7 Neutronics of Lead and Bismuth
197
7.3.2 Lead Isotopes
7.3.2.1 Numerical Approach
The numerical analyses were conducted with the combined use of SRAC2006 and
MARBLE code systems: collision probability calculations (PIJ [18]), eigenvalue
calculations (CITATION [19]), sensitivity coefficient calculations (SAGEP [20]) of
SRAC2006 and uncertainty calculations (UNCERTAINTY [21]) of MARBLE shown
in Fig. 7.11, coupled with JENDL-3.3, JENDL-4.0, ENDF/B-VII.0 and JEFF-3.1
nuclear data libraries. The cross-section data set in 107-energy-group processed by
the NJOY code [22] is pre-installed with the use of each data library in the SRAC2006
and the MARBLE code systems, to conduct numeral analyses of the thermal neutron
spectrum core, such as the KUCA core. For the experimental analyses, the accuracy of deterministic (diffusion-based) calculations by CITATION was useful in the
discussion with the use of JENDL-4.0 in processing important data analyses. Also,
JENDL-4.0, as a reference library in this study, was compared with other nuclear
(1) SRAC-SAGEP/PIJ
MACRO, 107g
MICREF, 107g
(2) CITATION
Flux
Adjoint Flux
(3) SAGEP
Sensitivity coefficient
(4) UNCERTAINTY
SRAC, 107g, library
Input data for core
Covariance data, 107g
Uncertainty
Cross section adjustment
Fig. 7.11 Calculation flow of sensitivity and uncertainty analyses (Ref. [2])
197
7.3.2 Lead Isotopes
7.3.2.1 Numerical Approach
The numerical analyses were conducted with the combined use of SRAC2006 and
MARBLE code systems: collision probability calculations (PIJ [18]), eigenvalue
calculations (CITATION [19]), sensitivity coefficient calculations (SAGEP [20]) of
SRAC2006 and uncertainty calculations (UNCERTAINTY [21]) of MARBLE shown
in Fig. 7.11, coupled with JENDL-3.3, JENDL-4.0, ENDF/B-VII.0 and JEFF-3.1
nuclear data libraries. The cross-section data set in 107-energy-group processed by
the NJOY code [22] is pre-installed with the use of each data library in the SRAC2006
and the MARBLE code systems, to conduct numeral analyses of the thermal neutron
spectrum core, such as the KUCA core. For the experimental analyses, the accuracy of deterministic (diffusion-based) calculations by CITATION was useful in the
discussion with the use of JENDL-4.0 in processing important data analyses. Also,
JENDL-4.0, as a reference library in this study, was compared with other nuclear
(1) SRAC-SAGEP/PIJ
MACRO, 107g
MICREF, 107g
(2) CITATION
Flux
Adjoint Flux
(3) SAGEP
Sensitivity coefficient
(4) UNCERTAINTY
SRAC, 107g, library
Input data for core
Covariance data, 107g
Uncertainty
Cross section adjustment
Fig. 7.11 Calculation flow of sensitivity and uncertainty analyses (Ref. [2])