Radioanalytical prediction of radiative capture in 99Moproduction via transmutation adiabatic resonance crossingby cyclotron
Abdollah Khorshidi • Mahdi Sadeghi •
Ali Pazirandeh • Claudio Tenreiro •
Yacine Kadi
Received: 12 June 2013 / Published online: 8 October 2013
� Akademiai Kiado, Budapest, Hungary 2013
Abstract In this study, the transmutation adiabatic reso-
nance crossing (TARC) concept was estimated in 99Mo
radioisotope production via radiative capture reaction in
two designs. The TARC method was composed of mod-
erating neutrons in lead or a composition of lead and water.
Additionally, the target was surrounded by a moderator
assembly and a graphite reflector district. Produced neu-
trons were investigated by (p,xn) interactions with 30 MeV
and 300 lA proton beam on tungsten, beryllium, and tan-
talum targets. The 99Mo production yield was related to the
moderator property, cross section, and sample positioning
inside the distinct region of neutron storage as must be
proper to achieve gains. Gathered thermal flux of neutrons
can contribute to molybdenum isotope production. More-
over, the sample positioning to gain higher production
yield was dependent on a greater flux in the length of
thermal neutrons and region materials inside the moderator
or reflector. When the sample radial distance from Be was
38 cm inside the graphite region using a lead moderator
design, the production yield had the greatest value of
activity, compared with the other regions, equal to
608.72 MBq/g. Comparison of the two designs using a Be
target revealed that the maximum yield occurred inside the
graphite region for the first design at 38 cm and inside the
lead region for the second design at 10 cm. The results and
modeling of the new neutron activator were very encour-
aging and seem to confirm that the TARC concept can be
used for 99Mo production in nuclear medicine.
Keywords Low energy proton � Moderator �Transmutation � Flux � 98Mo position � MCNP5
Introduction
The Adiabatic Resonance Crossing (ARC) concept was
first proposed by Nobel Laureate Carlo Rubbia [1] for the
transmutation by neutron capture of long-lived nuclear
waste, and for radioisotope production for nuclear medi-
cine applications. The ARC effect is based on the fact that
the transport of high energy neutrons in a medium with
high atomic number, low absorption cross section and high
elastic scattering cross section, is characterized by a short
mean free path and low energy loss per collision. The
concept was experimentally tested at CERN in the TARC
experiment using a 3.5 GeV proton beam and 334 tons of
high-purity lead [2–6]. The TARC experiment is well-
known with respect to the field of minor actinides trans-
mutation, but it also opened new perspectives regarding the
use of ion-beam generated neutrons for production of
medical radioisotopes that are currently produced in
nuclear reactors.99mTc is the product of the decay of the 65.94 h 99Mo.
99mTc decays to the ground state of 99Tc mainly by the
emission of 140.5 keV gamma-rays. 99Mo is produced in
A. Khorshidi (&) � A. Pazirandeh
Department of Nuclear Engineering, Science and Research
Branch, Islamic Azad University, Tehran, Iran
e-mail: [email protected]
M. Sadeghi (&)
Radiation Application Research School, Nuclear Science and
Technology Research Institute, Tehran, Iran
e-mail: [email protected]
C. Tenreiro
Faculty of Engineering, University of Talca, 2 Norte 685, Talca,
Chile
Y. Kadi
CERN, 1211, Geneva 23, Switzerland
123
J Radioanal Nucl Chem (2014) 299:303–310
DOI 10.1007/s10967-013-2749-7
reactors by the neutron fission of enriched uranium via the235U(n,fission)99Mo reaction. Meanwhile, the TARC
method consists of moderating neutrons in lead, which has
the lowest capture cross section in the fast neutron field. In
natural lead, fast neutrons are moderated in small energy
degradation steps; therefore neutron capture in the thermal
range is enhanced. Industrial 99Mo radioisotope is pro-
duced regularly using critical reactors. An accelerator-dri-
ven neutron activator can emulate higher saturation activity
of 98Mo(n,c)99Mo reaction [6–15].
In this study, 99Mo production was investigated via
neutron activation in two differently designed neutron
activators using 9Be, 181Ta and 184W targets according to
proton reactions by MCNP5 [16] simulations. The low
energy and low current proton with 30 MeV energy and the
300 lA current interact on the target to produce neutrons to
drive the neutron activator. The investigation was made
with an aim of efficiently utilizing generated neutrons for
the production of thermal neutrons in regions of the
designed moderators.
Neutron storage regions acting as moderator buffers in
two layers were surrounded by a graphite reflector. The
flux of gathered neutrons and diverse positions of 98Mo
samples in different regions were simulated to evaluate
saturation activity.
Materials and methods
Target properties in (p,xn) reaction
Low energy protons from an accelerator with 30 MeV and
300 lA based on the Karaj cyclotron (Cyclon30, IBA,
Belgium) in Iran with a gaussian beam were simulated in
interactions on 9Be, 181Ta, and 184W targets. These target
materials were chosen because of their high melting points
(1,287, 3,017, and 3,422 �C respectively) and reasonably
low activation levels. The target was designed in a cylin-
drical shape with a radius of 0.83 cm and length of 4 cm
length. Deposited proton energy in the targets depended on
the total stopping power, current, and energy of the
protons.
MCNP5 is a general-purpose Monte Carlo code for
simulation of protons, neutrons and photons or coupled
proton/neutron/photon transport [16]. The Monte Carlo
code MCNP5 from Los Alamos National Laboratories was
used to investigate the TARC method. When the code ran
by the defined protons, the protons tracked it through the
simulated design and particle history weights were binned
into the energy spectra. In the simulation of proton energy
loss inside the target, an *F4 tally (The unit of F4 tally
output is particle/cm2/s, as well as adding an asterisk tally
of F4 changes the units into an energy tally as MeV/cm2)
was used to determine lost energy per cm3, and the beam
intensity was 3.28E-4 proton/cm2/s. The target was divi-
ded into different segments with a thickness of 0.01 cm in
39 layers.
Moderators, reflector and 98Mo position designs
The target was surrounded by a moderator assembly.
Transmutation was prepared in two moderator designs,
including natural lead and water regions. In the first design,
the whole lead moderator was considered a spherical model
with a 30 cm radius around the target. In the second design,
the lead and water moderators were arranged with a 15 cm
thickness around the target. Figure 1 shows these two
designs in addition to 98Mo locations, which were settled in
x, y, and z directions of the beam axis. The beam axis is
perpendicular to the base of the cylindrical target as well as
the target places in origin. Figure 1a shows the first design
including the natural lead moderator and 98Mo samples in x
and y directions at diverse regions. Figure 1b shows the
second design including the lead and water moderators as
well as settled samples.
Induced neutrons slowed down inside the moderator
region at first. Then the graphite reflector, which had a
35 cm thickness around the moderator assembly, saved the
derived neutrons in both designs. The graphite was pre-
ferred as a reflector because of its high elastic scattering
cross section and low absorption cross section to store the
neutrons in a superior flux. Moreover, boron with a 5 cm
thickness was used as the absorbent matter to enclose the
reflector.98Mo samples were spheres with a 1 cm radius and were
arranged at distances of 15 and 25 cm from the target
inside the lead moderator region and at 38 and 45 cm
inside the graphite reflector region for Design 1. In Design
2, samples were positioned at distances of 10 and 14 cm
from the target inside the lead region at 20 and 28 cm
inside the water region, and at 38 and 45 cm inside the
graphite reflector. The 99Mo production yield related to the
neutron flux in thermal range must be proper to achieve
gains. Therefore, using a desirable system and location of
samples inside lead or graphite regions for production yield
is a significant result.
Moderators cross sections characteristics
Radioisotope production yield relates to the cross section
and position of the sample inside the distinct region of the
neutron storage. Therefore, the investigation of the reaction
activity and flux transpires by way of determining cross
section characteristics. Absorption and elastic cross sec-
tions of the derived neutrons from the target were calcu-
lated from the ENDF/B-VII.1 libraries. The elastic cross
304 J Radioanal Nucl Chem (2014) 299:303–310
123
section in whole neutron energy was higher than the
absorption cross section. Therefore, lead was chosen as the
moderator material because of its greater elastic charac-
teristics of absorption. The elastic cross section comparison
for lead and water moderators indicated that water had the
highest cross section at the thermal range of neutron
energies without any significant resonance peaks; lead,
however, included a higher resonance peak from lower
neutron energy.
Results and discussions
Proton energy loss and neutron cross section
inside the target
The target was divided into different segments with a
0.01 cm thickness, and the proton energy loss decreased
along the target. The energy loss of a proton incident on the
target surface was related to the different segments that the
Incident protonTarget
Lead
Graphite
Boron
30 cm35 cm
5 cm
Shield
98Mo sample
Incident protonTarget
Water
Graphite
Boron
15 cm35 cm
5 cm
Shield
Lead15 cm
98Mo sample
a
b
Fig. 1 a The first design
containing lead moderator.
b The second design containing
lead and water moderators
J Radioanal Nucl Chem (2014) 299:303–310 305
123
particle felt during its trajectory. Figure 2 displays the
deposited proton energy for the three targets. Inserted
proton energy inside the Be target was begun in
2.64E?16 MeV/cm2 at the first layer, and then this quan-
tity lessened until the 13th layer in order of E?15. This
significant effect for the Be target occurred because of the
low atomic mass and lower melting point compared to that
of the Ta and W targets. The proton energy inside the Ta
and W targets decreased toward zero at the 13th and 15th
layers respectively. Inside the Be target, however, the
deposited energy increased at the 14th layer, and then
energy loss decreased along the target. As shown in Fig. 2,
the proton energy loss inside the W target was greater than
that of the Ta and Be targets because of its lesser projected
range in 30 MeV incident protons as well as its elevated
density and melting point. Absorption and elastic cross
sections of generated neutrons inside the target were con-
sidered. For the Be target the elastic cross section was
between 1 and 110 barns, but the absorption cross section
was smaller than the elastic cross section at whole neutron
energy ranges of between 10E-5 and 1 barns. These cross
sections indicated more resonance peaks of produced
neutrons inside the Ta and W targets, which measured
between 10E-5 to 0.001 MeV and 10E-4 to 0.01 MeV
neutron energy respectively. The resonance peaks of
induced neutrons inside the Ta and W targets increased the
interaction probability and produced superior heat per
centimetre within the target compared to the Be target;
therefore a cooling system of water must be designed
which transfers heat and keeps the target mechanically
stable and cool.
The interaction of the proton beam in the target resulted
in target disintegration because of the proton energy range
and target characteristics. The values of the total number of
neutrons per second and per microampere in the target
were produced by proton bombardment on specified tar-
gets, and the simulated results were 0.038, 0.036, and 0.041
from the Be, W, and Ta targets respectively.
Neutron flux evaluation inside diverse regions
Neutrons produced by interaction between the incident
protons and target nucleus were moderated through lead
and water for both designs. Volumetric flux was estimated
via F4 tally inside diverse regions. Figure 3a displays the
volumetric flux (n/cm2/s) inside the lead moderator,
graphite reflector, and boron absorber regions at the ther-
mal range of neutron energy via the Ta target for the first
design. The thermal flux of neutrons inside the lead and
boron regions were around each other, but inside the
graphite region, they were non-contiguous with a maxi-
mum thermal flux between 0.2E-7 and 0.4E-7 MeV of
Fig. 2 Deposited proton energy
inside the Be, W, and Ta targets
per cm2 for 39 layers
306 J Radioanal Nucl Chem (2014) 299:303–310
123
neutron energy. Figure 3b shows the thermal neutron flux
via Ta target for the second design. The flux in whole
thermal energy was greater inside the lead region than the
graphite region. The thermal flux inside the water moder-
ator had a maximum peak at thermal energy of derived
neutrons, and the thermal neutron flux inside the boron was
non-contiguous with a maximum thermal flux at 0.6E-
7 MeV of neutron energy.
Table 1 demonstrates the volumetric flux (n/cm2/s) for
the three targets inside diverse regions at different neutron
energies for the first design. As is evident, the thermal flux
inside the lead, graphite, and boron regions via Be was
greater than that via the Ta and W targets. Moreover,
Table 2 demonstrates the volumetric flux for the second
design. The thermal flux inside the lead and water mod-
erators were in order of E ? 10 n/cm2/s with greater values
via the Be target. Meanwhile, the thermal flux inside the
graphite and boron regions were in order of E?9 and
E?8 n/cm2/s, respectively.
Flux estimation inside diverse regions by two schemes
can contribute to 98Mo position for the attainment of higher
activity and 99Mo production yield via radiative capture.
Comparison of the two designs revealed that the second
design gathered a greater neutron flux inside the natural
lead region, but the neutron flux inside the graphite region
had lesser values than in the first design.
Radiative capture yield
The production yield from radiative capture inside 98Mo
samples is dependent on the neutron flux in thermal ranges
and samples placement. The yield must also be proper to
achieve greater activity. Therefore, using a desirable design
and location of samples inside lead or graphite regions for
production yield is a significant result. The 99Mo activity
per gram versus 98Mo placement from the target in origin
for both designs and three targets is shown in Fig. 4. When
the sample radial distance from Be was 38 cm inside the
graphite region using the first design, the production yield
had the greatest value of activity equal to 608.72 MBq/g.
Comparison of the first and second designs for the Be
target revealed that maximum yield occurred inside the
graphite region at 38 cm and inside the lead region at
10 cm. The production yield via the W target for the first
and second designs transpired at 45 cm inside the graphite
region, equal to 523.44 MBq/g, and at 14 cm inside the
lead region, equal to 558.81 MBq/g.
These simulations included absorption, scattering, cap-
ture, escape, and neutron production in a neutron activator,
Fig. 3 a Neutron flux (n/cm2/s) inside lead (solid line), graphite (non
contiguous dash line) and boron (contiguous dash line) regions at
thermal range of neutron energy via Ta target for 1th design.
b Neutron flux (n/cm2/s) inside lead (solid line), water (one histogram
dash line), graphite (contiguous dash line) and boron (non contiguous
dash line) regions at thermal range of neutron energy via Ta target for
2th design
Table 1 Volumetric flux (n/cm2/s) of induced neutrons from proton interaction with three targets in diverse ranges of neutron energies for the
first design
Target Material Thermal 0–0.1 eV Epithermal 0.1 eV–5 keV Fast 5 keV–1 MeV Total 0–30 MeV
9Be Lead 1.94E?10 4.57E?10 1.06E?11 1.20E?11
Graphite 1.75E?10 2.49E?10 2.73E?10 2.76E?10
Boron 2.88E?09 3.30E?09 3.31E?09 3.32E?09181Ta Lead 1.31E?10 3.26E?10 8.05E?10 8.68E?10
Graphite 1.27E?10 1.78E?10 1.94E?10 1.95E?10
Boron 1.96E?09 2.25E?09 2.25E?09 2.25E?09184W Lead 1.32E?10 3.00E?10 7.62E?10 8.23E?10
Graphite 1.20E?10 1.69E?10 1.85E?10 1.85E?10
Boron 1.89E?09 2.14E?09 2.14E?09 2.14E?09
J Radioanal Nucl Chem (2014) 299:303–310 307
123
and also included radiative capture in molybdenum sam-
ples which are an output of the code. The activity of 99Mo
with a c peak at 739.50 keV and 12.13 % intensity was
considered. The emitted photons inside the sample volume
per second and the activity generated in molybdenum
samples were tabulated in Table 3 for the three targets and
two designs. When the material changed from lead to
graphite in the first design, the emitted photons and activity
increased via the Be and Ta targets. This increase for the
Ta target was greater than the increase for the Be target.
Changing the moderator for the second design led to lesser
values in emitted photons and activity. Moreover, this
reduction at 28 cm inside both the water moderator and the
graphite reflector kept up continually.
99Mo is produced generally in reactors by the neutron
fission of enriched uranium via the 235U(n,fission)99Mo
reaction. Assuming that a specific in the fissium with an
atomic fraction k and that the exposure time texp equal to
one half-life of compound, the initial activity for 1 kg of
activated sample is 2.5 9 10-10 S0kgf (GBq/kg) which S0
is neutronic source (n/s) and gf is fissium efficiency (kg-1).
If the target be 20 %-enriched metallic Uranium of a mass
of 33 kg, and the exposure time set to 10 days the
asymptotic yield calculated to 51.3 GBq/kg [1–5]. In the
current study, 99Mo production was simulated via neutron
activation by the 98Mo(n,c)99Mo reaction in two new
designs of neutron activators using 9Be, 181Ta, and 184W
targets according to (p,xn) reactions. Neutron flux
Fig. 4 99Mo activity (MBq/g)
by radiative capture inside 98Mo
samples at varied radial
distances from the target
Table 2 Volumetric flux (n/cm2/s) of induced neutrons from proton interaction with three targets in diverse ranges of neutron energies for the
second design
Target Material Thermal 0–0.1 eV Epithermal 0.1 eV–5 keV Fast 5 keV–1 MeV Total 0–30 MeV
9Be Lead 3.68E?10 5.52E?10 1.25E?11 1.76E?11
Water 1.96E?10 2.55E?10 2.90E?10 3.12E?10
Graphite 2.67E?09 3.23E?09 3.39E?09 3.47E?09
Boron 3.40E?08 4.05E?08 4.10E?08 4.14E?08181Ta Lead 2.90E?10 4.60E?10 1.19E?11 1.61E?11
Water 1.41E?10 1.85E?10 2.09E?10 2.16E?10
Graphite 1.50E?09 1.72E?09 1.76E?09 1.78E?09
Boron 1.81E?08 2.03E?08 2.03E?08 2.03E?08184W Lead 2.74E?10 4.19E?10 1.09E?11 1.33E?11
Water 1.34E?10 1.76E?10 1.99E?10 2.07E?10
Graphite 1.44E?09 1.68E?09 1.72E?09 1.74E?09
Boron 1.77E?08 2.00E?08 2.00E?08 2.00E?08
308 J Radioanal Nucl Chem (2014) 299:303–310
123
estimation inside diverse regions of the two designs can
contribute to 98Mo placement for the attainment of higher
activity and 99Mo production yield via radiative capture.
Comparison of the two designs revealed that the second
design gathered a greater neutron flux inside the natural
lead region, but the neutron flux inside the graphite region
had lesser values than the first design.
The TARC method was investigated via Fermi’s age
theory for the slowing down of neutrons from their ori-
ginal energy to their final thermal energy through colli-
sions with the moderator’s nuclei. This collision and
energy loss is described as a continuous process, which is
valid if the energy loss is small compared with neutron
energy.
Neutronic flux around the target goes with 1/r inside the
lead moderator as opposed to 1/r2 inside air (for the bare
target), and this is the power of TARC, that is ‘‘amplifying
geometrically’’ the neutron source strength in lead as a
heavy nuclei. Transport of fast neutrons in the lead mod-
erator by a short mean free path and low energy loss per
collision decelerates the fast neutrons. Additionally, (n,2n)
and (n,3n) interactions transpire in the moderator. Fast
neutrons can have interactions with the moderator to pro-
duce induction neutrons related to their Q value. Moreover,
since the capture cross section at thermal energy and the
capture resonance integral of natural lead are much larger
than the corresponding values of most of the light nuclides,
it is safe to say that neutrons are captured during the
slowing down process in natural lead with a higher prob-
ability compared to the slowing down process in light
materials. Therefore, only a small part of neutrons will
slow down to thermal energy. This means that thermal
neutron flux in natural lead will be much lower than in light
materials.
Conclusion
In this paper, The TARC method was composed of mod-
erating neutrons in lead or a composition of lead and water.
Additionally, the target was surrounded by a moderator
assembly and a graphite reflector district. When the sample
radial distance from Be was 38 cm inside the graphite
region using the first design, the production yield had the
greatest value of activity compared to the others, equal to
608.72 MBq/g. Comparison of the first and second designs
for the Be target revealed that the maximum yield occurred
inside the graphite region at 38 cm and inside the lead
region at 10 cm. The 99Mo production yield is related to
the cross section and sample placement inside the distinct
region of neutron storage as must be proper to achieve
gains. The results and modeling of the new neutron acti-
vator were very encouraging and seem to confirm that the
TARC concept can be used for 99Mo production in nuclear
medicine.
Acknowledgments This research was supported by Direccion de
Investigacion, Universidad de Talca, project VAC 600 556.
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