Program - Brookhaven National Laboratory

transport code DRAGON. This code package includes the necessary tools for the generation of the fewgroups
nuclear data libraries by applying the data processing code NJOY and using directly data contained
in the nuclear data evaluation files. The nuclear data used in this study is from the JEFF3.1 evaluation. The
ENDF nuclear data files for the actinides 235,238 U and 239,240,241 Pu, and for the fission products (randomized
for the generation of the various DRAGON libraries) are taken from the nuclear data library TENDL-2012.
The total uncertainty (obtained separately for each of the isotopes considered by randomizing all nuclear
data in the ENDF files) on the reactor parameters has been split into different components (different nuclear
reaction channels). Both the effects of transport data and fission yields have been studied. Results show
that the combined uncertainty on reactivity as result of nuclear data uncertainties in all considered isotopes
is reasonably constant during fuel burnup and amounts to about 0.7%. At zero burnup only transport
data from 235 U and 238 U deliver an important contribution, whereas at discharge burnup 235 U and 239 Pu
represent the largest effects, followed by 238 U. Uncertainties in transport data from all 119 fission products
have only a secondary effect on the final uncertainty, although not negligible (they account for 5% relative
of the total). The uncertainties on discharged fuel inventory varies typically in the range 1-15%, depending
on the particular element, except for some fission products produced only in small concentrations for which
the uncertainties can be as high as 35%. For the important actinide isotopes the uncertainties stay below
5%. For most fission products the largest source of uncertainty is attributed to uncertainties in fission
yields of the most important fissile isotopes 235 U and 239 Pu. For the inventory of the most important
actinides the uncertainties on the transport data of 235,238 U and 239 Pu are the major contributors.
Session ND Beta delayed neutrons
Thursday March 7, 2013
Room: Empire West at 10:30 AM
ND 1 10:30 AM
Sensitivity Analysis for Reactor Stable Period Induced by Positive Reactivity Using
One-Point Adjoint Kinetic Equation
Go Chiba
Hokkaido University
In order to better predict a dynamic behavior of a nuclear fission reactor, an improvement of the delayed
neutron parameters is essential. The present paper specifies important nuclear data for a reactor kinetics:
Fission yield and decay constant data of germanium-86, some bromine isotopes, rubidium-94, yttrium-
98m and some iodine isotopes. Their importances are also quantified as sensitivities with a help of the
adjoint kinetic equation, and it is found that they are significantly dependent on an inserted reactivity
(or a reactor stable period). Moreover, dependence of sensitivities on nuclear data files are also quantified
using the latest files. Even among the currently evaluatad data files, there are large differences from a view
point of the delayed neutrons.
ND 2 11:00 AM
First Compilation and Evaluation of Beta-Delayed Neutron Emission Probabilities and
Associated Half-lives in the Non-Fission Region (A ≤ 72)
B. Singh, M. Birch
McMaster University, Hamilton, Ontario, Canada
D. Abriola
Nuclear Data Section, IAEA, Vienna, Austria
195