Program - Brookhaven National Laboratory

Concerns about the limits of worldwide uranium resources motivated initial interest in the thorium fuel
cycle. It was envisioned that as uranium reserves were depleted, thorium would supplement uranium as a
fuel in nuclear reactors. In thorium fuel cycle, Th-232 is the well known fertile nuclide. It is seen from the
detailed energy dependent neutron-induced cross-section behaviour of this isotope that it supports fission
only with very fast neutrons, i.e. beyond about 1 MeV. At lower energy it has high potential of absorption
of neutrons, and consequently converting to fissile nuclides. In that way, even though Th-232 itself is
not fissile, it transforms into fissile isotope U-233 upon absorption of a neutron. Conversion of fertile to
fissile is possible over a wide energy range, but significant conversion is possible only in an appropriate
range. In a nuclear reactor, the conversion depends on available neutrons over and above what is required
to sustain the fission chain and the inevitable losses due to leakage and parasitic absorption. It must be
noted that the newly generated nuclei undergo all probable interactions and decays, so that they have
both production and destruction routes. It is a challenge to obtain an adequate conversion through a
proper combination of fissile, fertile and other materials arranged in a carefully worked out geometry in
the design. Even before trying with different combination of these factors for achieving better conversion,
one should have a proper and quantified idea about the energy range in which a pure Th-232 isotope
have better conversion in U-233. Since all the parameters such as production and depletion cross sections
involved in the conversion vary with neutron energy due to resonance characteristics which show marked
variation with energy, it becomes very difficult to quantify and select the specific energy range/s in which
the conversion from pure Th-232 sample to U-233 is more favorable. The burnup and depletion codes
such as ORIGEN 2.1 [1] which solve the transmutation and decay equation uses a one group effective
cross section are incapable of generating the behavior of conversion from Th-232 to U-233 with energy. In
the world, none of depletion and radioactive decay computer codes use the point cross section of nuclide
directly. We have developed a computer code which solves the nuclide chain originating with Th-232 with
point cross section. This code is used to estimate the variation of conversion of pure Th-232 sample into
fissile U-233 as a function of neutron energy. In this code, Bateman equations which represent the time
rate of change in the concentration of a specific isotope are solved at every energy point of neutron energy
spectrum. The energy interval is obtained after the proper energy grid unionization process of energy
interval used in each isotope. We have taken point-wise cross section data from the basic evaluated nuclear
data file (ENDF/B-VII.0) for each isotope presented in nuclide chain originating with Th-232. Behaviour
of conversion ratio with energy and with exposure time of pure Th-232 sample is studied in our study with
the help of developed code. The results are also compared using different nuclear data libraries such as
JENDL-3.3.
[1] Allen G Croff, “ORIGEN2: A versatile Computer Code for Calculating the Nuclide Composition and
Characteristic of Nuclear material,” Nuclear technology, 62, 335-352, September 1983.
HE 5 5:00 PM
Remarks on KERMA Factors in ACE files
Chikara Konno, Kentaro Ochiai, Kosuke Takakura, Masayuki Ohta, Satoshi Sato
Japan Atomic Energy Research Institute, Tokai-mura, Naka-gun, Ibaraki-ken 319-1195, Japan
KERMA (Kinematic Energy Release in Material) factors are used as response data in order to obtain
nuclear heating in nuclear analyses. These data are not directly included in nuclear data libraries and
they in ACE (A Compact ENDF) files for the Monte Carlo radiation transport code MCNP are deduced
from cross section data for all the reactions in nuclear data libraries with the NJOY code. Many peoples
often use these data, but little is known concerning fact that most of these data are not always correct.
We will present this issue here. As well known, there are two methods to compute KERMA factors; one
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