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: abkhorshidi@yahoo.com

M. Sadeghi (&)

Radiation Application Research School, Nuclear Science and

Technology Research Institute, Tehran, Iran

e-mail: msadeghi@nrcam.org

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

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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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