Fission Product Yield Data for the Transmutation of Minor Actinide ...

terms. Destruction terms are only related to the
concentration of the nuclide and include both
radioactive decay and reactions that transform the
nuclide. The production terms are related to the
concentrations of the other nuclides present that by
radioactive decay processes or induced reactions
lead directly to the nuclide of interest.
In traditional inventory calculations the
destruction terms considered are the radioactive
decay of the nuclide and the main neutron induced
reactions of the nuclide (n,g), (n,f) and (n,2n), and
sometimes (n,p), (n,d) and (n,a). The production
terms include radioactive decay, fission products
from the fission of actinides, and the neutron
reactions of all nuclides that generate the nuclide of
interest.
There are three ways in which nuclide
inventory could be determined. The first and most
commonly used is a numerical solution to the
complete set of differential equations (e.g. FISPIN
[3.4.7], ORIGEN [3.4.8]). An alternative is an
analytical solution of simplified nuclide chains (e.g.
FISP [3.4.9]). The final alternative used in some
special cases, typically where nuclide cross-sections
are small and precursor half-lives are short, is direct
calculation from the fission rate and the cumulative
fission product yield.
In all of these calculations, the neutron
induced fission product yields are from a limited set
of actinides that significantly contribute to the
fission rate during the irradiation. The fission rates
are calculated from the neutron flux, spectra and
fission cross-sections. The production rate for each
fission product is calculated by summing over the
product of the independent yield and the fission
rate for each significant fissioning nuclide.
3.4.2.3. Requirements for fission product yield data
for minor actinide transmutation
There are many scenarios proposed for
transmuting minor actinides. Some of these involve
accelerated particle beams bombarding a target
either inside or on the outer boundary of a reactor
core to produce neutrons. Under these circumstances,
the traditional yield sets can no longer be
assumed to be adequate because the neutron
spectra in or near the target may differ significantly
from that in traditional fast or thermal reactors. In
some other scenarios, the minor actinides are
separated and placed in fuel or irradiation
assemblies within conventional thermal or fast
reactors. The current yield sets for such cases may
be adequate, although the higher fission rates in
minor actinides may require improved accuracy to
predict safety related parameters adequately. Thus
the following only considers accelerator driven
systems, which require a new evaluation approach.
Fission in the particle beam target can be
induced by either the incident particle or a wide
range of secondary energetic particles, including
neutrons, protons, electrons, photons and heavy
ions. The energy range of these particles can extend
to several hundred MeV, and even potentially into
the GeV region. The range of fission reactions at
these energies is much larger than in conventional
reactors. For example, natural lead and tungsten
isotopes can undergo fission, although in practice
their cross-sections will be much lower than typical
‘thermal’ or ‘fast’ neutron cross-sections for
actinides. A code, such as MCNPX [3.4.10], will be
required in this region to calculate the particle
fluxes and the resultant fission rates. It will then be
necessary to estimate the production rate of fission
products to allow the solution of the nuclide
concentration equations.
In the fuel or irradiation targets, the fission
rate is expected to be dominated by neutron
induced reactions as charged particles will not travel
far in matter. It will be necessary to estimate the
neutron flux in each region and use neutron crosssections
to determine the fission rates from each
fissile nuclide, again using codes like MCNPX. As in
the target region, it will be necessary to estimate the
production rate of fission products to allow the
solution of the nuclide concentration equations.
Both cases require tabulations of the yields for all
significant fissile species for a wide range of neutron
or charged particle energies. In practice, the calculations
will require three stages.
Firstly, the particle fluxes and the resultant
fission rates for all fissionable nuclides need to be
determined as a function of energy or within a set of
energy groups. These calculations would be carried
out by a multi-particle transport code such as
MCNPX.
Secondly, the yield tabulations need to be
condensed into average product yields per induced
fission by weighted averaging of the energy
dependent fission product yields with the fission
rate for each energy, fissionable nuclide and particle
inducing fission. This will require the particle fluxes
and fission rates from the transport code. As these
parameters may be time dependent, it may be
necessary to produce time dependent average
fission product yields sets. This is reminiscent of the
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