Program - Brookhaven National Laboratory

is ”energy-balance method” and the other is ”kinematic method”. The energy-balance method is the best
method if the nuclear data libraries keep energy balance, but this method produces negative or extremely
large KERMA factors if the nuclear data libraries do not keep energy balance, e.g. in the case of no
gamma production data. The kinematic method produces only minimum and maximum KERMA factors,
but the calculated KERMA factors are not negative and extremely large. The maximum KERMA factors
are close to ones adequately calculated with the energy-balance method. NJOY outputs only KERMA
factors calculated with the energy-balance method to ACE files. Thus the KERMA factors in ACE files
are not correct if the nuclear data libraries do not keep energy balance. We examined KERMA factors
in the latest official ACE files; those of JENDL-4.0, ENDF/B-VII.1, JEFF-3.1.1, etc. We also checked
DPA cross section data because they are related to KERMA factors. Moreover we investigated KERMA
factors and DPA cross section data in MATXS files if they are released. We will present these results in
the conference.
HE 6 5:15 PM
Nuclear Data Uncertainty Propagation to Reactivity Coefficients of Sodium Fast Reactor
José J. Herrero, R. Ochoa, J. S. Martínez, C. J. Díez, Nuria García-Herranz, O. Cabellos
Departamento de Ingeniería Nuclear, Universidad Politécnica de Madrid, Calle José Gutiérrez Abascal 2,
28006 Madrid, Spain
Engineering of new reactor models requires computational tools capable of producing results with the
adequate level of accuracy. One source of uncertainty in the modeling arises from the employed nuclear
data. Here, sensitivity analysis of the quantities important for safety and design to the input parameters,
and posterior uncertainty propagation from the parameters to the results is a main tool to point out
which nuclear data should be improved. One such new reactor models is the European Sodium Fast
Reactor (ESFR), and a one of such design quantities is the group of reactivity coefficients due to heating
and voiding effects. Here we present uncertainty quantification from nuclear cross sections data to the
mentioned reactivity coefficients of the ESFR core model, with the objective of identifying the nuclear
reaction data where an improvement will certainly benefit the design accuracy. The ESFR full core has
been modeled for SCALE6.1, and a series of steady states have been computed with KENO-VI Monte Carlo
code using the available 238 energy groups cross sections library based on ENDF/B-VII.0 evaluation. An
adjoint calculation is also performed to apply Adjoint Sensitivity Analysis Procedure (ASAP) to obtain
sensitivities, first for each steady state k-eff, then for the reactivity coefficients between the reference and
perturbed states using SCALE6.1 tools. Propagated uncertainty data comes from the 44 energy groups
evaluation included with SCALE6.1. The research leading to these results has received funding from
the European Atomic Energy Community’s 7th Framework Programe in the ANDES project under grant
agreement No. 249671. It has also been funded by the specific collaborative agreement between CIEMAT
and UNED/UPM in the area of “High level waste transmutation.”
HE 7 5:30 PM
Assessing the Impact of Pb and Pu Cross Section Data Uncertainties on Safety Parameters
of a Low Power Lead Cooled Reactor
Erwin Alhassan, Junfeng Duan, Cecilia Gustavsson, Stephan Pomp, Henrik Sjöstrand, Michael Osterlund
Division of Applied Nuclear Physics, Department of Physics and Astronomy, Uppsala University,
Uppsala, Sweden
Dimitri Rochman, Arjan Koning
Nuclear Research and Consultancy Group (NRG), Petten, The Netherlands
129