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124
Te(p , 2n)
K. Kondo, R.M. Lambrecht, and A.P. Wolf.
Excitation functions of the
123
I
123
I production for radiopharmaceuticals-XX.
124
Te( p, 2n)
123
I and
124
Te( p,n)
124
I reactions and eect of target
enrichment on radionuclidic purity. Int. J. Appl. Radiat. Isot. 28 (1977) 395.
K. Kondo, R.M. Lambrecht, E.F. Norton, and A.P. Wolf. Cyclotron isotopes and radio-
pharmaceuticals-XXII. Improved targetry and radiochemistry for production of
124
I. Int. J. Appl. Radiat. Isot. 28 (1977) 765.
B. Scholten, Z. Kovács, F. Tárkányi, and S.M. Qaim. Excitation functions of
reactions from 6 to 31 MeV with special reference to the production of
124
124
I at a small
123
Te( p, x n)
I and
124,123
cyclotron. Appl. Radiat. Isot. 46 (1995) 255.
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24. Chattopadhyay, S., Das, M.K., Sarkar, B.R., and Ramamoorthy, N. (1997). Radiochemical
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111
In from cadmium, copper, aluminium and traces of iron: use
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25. Macdonald, N.S., Neely, H.H., Wood, R.A. etal. (1975). Methods for compact cyclotron
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27. Firouzbakht, M.L., Teng, R.R., Schlyer, D.J., and Wolf, A.P. (1987). Production of high
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123
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I production. Int. J. Appl. Radiat. Isot. 28: 967–969.
123
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124
Xe. Appl. Radiat. Isot. 42: 221–228.
123
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xenon-124: cross-sections and yields. Radiochim. Acta 56: 167–171.
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production of I-123 from Xe-124. In: Proceedings of the Sixth Workshop on Targetry and
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33. Acerbi, E., Birattari, C., Castiglioni, M., and Resmini, F. (1975). Production of
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medical purposes at the Milan AVF cyclotron. Int. J. Appl. Radiat. Isot. 26: 741–747.
34. Barrall, R.C., Beaver, J.E., Hupf, H.B., and Rubio, F.F. (1981). Production of Curie quantities of high purity I-123 with 15 MeV protons. Eur. J. Nucl. Med. 6: 411–415.
35. Clem, R.G. and Lambrecht, R.M. (1991). Enriched
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I. Nucl. Instrum. Methods A 303: 115–118.
124
Te targets for production of
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I and
36. Guillaume, M., Lambrecht, R.M., and Wolf, A.P. (1975). Cyclotron production of
123
Xe and high purity
123
I: a comparison of tellurium targets. Int. J. Appl. Radiat. Isot.
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37. Hupf, H.B., Eldridge, J.S., and Beaver, J.E. (1968). Production of iodine-123 for medical
applications. Int. J. Appl. Radiat. Isot. 19: 345–351.
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chemistry for the production of
123
I and
124
I. Int. J. Appl. Radiat. Isot. 28: 765–771.
39. Mahunka, I., Ando, L., Mikecz, P. etal. (1996). Iodine-123 production at a small cyclotron for medical use. J. Radioanal. Nucl. Chem. Lett. 213: 135–142.
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123
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124
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123
I reaction at a compact cyclotron. Int. J. Appl.
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41. Lambrecht, R.M., Mantescu, C., Redvanly, C., and Wolf, A.P. (1972). Preparation of
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42. Sajjad, M., Lambrecht, R.M., and Bakr, S. (1990). Autoradiolytic decomposition of
reductant-free Sodium
124
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Iodide. Radiochim. Acta 50: 123.
Chapter 4: Production of Radionuclides Used in SPECT 87
Chapter 5
Production
ofTherapeutic
Radionuclides
C. Shaun Loveless
1
Department of Radiology, University of Alabama at Birmingham,
1,2
and Suzanne E. Lapi
1,3
Birmingham, AL, 35294, USA
2
Department of Chemistry, Washington University in St. Louis, St.
Louis, MI, 63130, USA
3
Department of Chemistry, University of Alabama at Birmingham,
Birmingham, AL, 35294, USA
5.1 INTRODUCTION
Therapeutic radionuclides decay by the emission of alpha particles, emission of beta
particles, or electron capture, which leads to the emission of Auger electrons. These
decay particles have high linear energy transfer (LET) and thus deposit the majority of
their energy over very short distances (μm-mm). The majority of radionuclides used for
clinical radiotherapy are neutron-rich beta emitters and are produced in nuclear reactors
by neutron capture reactions on stable targets or through ssion. Other radionuclides
that decay by alpha emission or electron capture routes may be produced via chargedparticle reactions using cyclotrons or linear accelerators. While neutron capture and
charged-particle reactions are used to produce the majority of therapeutic radionuclides,
photonuclear reactions using electron linear accelerators (eLINACs) are actively being
investigated as a potential new route to produce and supply previously inaccessible or
hard-to-obtain radionuclides. This chapter covers these dierent production pathways
and highlights commonly used isotopes produced through these routes.
Handbook of Radiopharmaceuticals: Methodology and Applications, Second Edition.
Edited by Michael R. Kilbourn and Peter J.H. Scott.
© 2021 John Wiley & Sons Ltd. Published 2021 by John Wiley & Sons Ltd.
tgt
RN/
For nuclear reactions important to the production of radionuclides, we can consider a
particle incident on a stationary target giving rise to a radioactive product and a light particle or gamma ray, as shown here
where X is the stationary target, a is the incident particle, Y is the radioactive product,
and b is the light particle or gamma ray. This can be written in short form as X(a,b)Y. For
instance, if one produces
written as
152
Sm(n,γ)
153
Sm. The reaction rate depends on the ux of particles hitting the
153
Sm via neutron bombardment of
152
Sm, the reaction can be
target, target thickness, target density, and cross-section, which is proportional to the
probability of the nuclear reaction taking place. The cross-section (σ ) is given in units of
area (m
2
); however, the smaller units of barn (1 b=10
−24
cm2) or millibarn are typically used
in practice. The reaction rate (R) in atoms/second of a neutron-induced reaction is
where φ is the neutron ux per area per second and N
is the total number of target
tgt
atoms. The corresponding reaction rate for a charged-particle reaction is
where I is the particle current incident on the target (particles/s), n is the number of
nuclei/cm
3
of target, and x is the target thickness (cm).
As these reactions often produce radioactive nuclei that are decaying, the overall
change in the number of atoms present in the target is
where λ (which can be calculated by ln(2)/t
nuclide of interest, and N is the number of atoms of the product. Eventually, at long
irradiation times, the production rate will equal the rate of decay, and the number of
radioactive nuclei in the target will be at a steady state. This is known as the saturation
point. Typically, irradiations for radionuclide production are conducted for a maximum of
two half-lives.
90 Handbook of Radiopharmaceuticals
Nt
) is the decay constant (in s−1) of the radio-
1/2
RADIONUCLIDES
Nuclear reactors use ssile material (generally
event typically results in two to three additional neutrons that sustain a chain reaction
in the reactor fuel by inducing additional ssion events. These neutrons can also be used
to produce radionuclides in targets via two general mechanisms: (i) by direct neutron
reactions such as neutron capture, or (n,p) reactions; or (ii) by inducing additional ssion
events in targets composed of ssile material.
The cross-section (probability of reaction) of neutron capture, (n,γ), reactions generally
increases as the incident neutron energy decreases. Thus, these reactions are carried out
with low-energy or slow neutrons. Generally, target materials are solid metal or oxide
powders, but they can also be liquid or gas. Neutron capture reactions typically result in a
radioactive product of the same element as the initial target material. Hence, there is typically a large amount of non-radioactive or “cold” material that cannot be chemically separated from the product of interest. The ratio of radioactive atoms to the mass of material
is relatively low, and the material is said to have low molar activity (Bq mol
the beta emitters
be produced by neutron capture on the stable isotopes
Typically, isotopically enriched targets are used to produce radionuclidically pure products. As the radioactive product is the same element as the target material, no separation
chemistry is required after irradiation, and the target is simply dissolved for downstream
chemistry[2]. The cross-sections for neutron capture reactions are typically much higher
than for other types of reactions, and thus very large quantities of radioisotopes can be
produced via this route. For example, the cross-section at low neutron energies for the
177
Lu(n,γ) reaction is ~2100 b. Thus, a 10 mg sample of
typical neutron ux of 3 × 10
sents a conversion rate of
In some cases, particles containing the target isotope can be irradiated, giving rise to
intrinsically radiolabeled particles. Radioactive gold and samarium nanoparticles (NPs)
have been produced by exposing premade stable gold or samarium particles to a neutron
ux[3, 4]. In the case of gold, solid gold particles were prepared for irradiation; however,
Hashikin etal. reported on the preparation of stable
composed of a commercially available resin, which were irradiated after preparation.
When the radioactive product is created from a nuclear reaction, the resulting nucleus
has a signicant amount of kinetic energy. By careful selection of the target material
and purication process, this recoil energy can be used to partially purify the radioactive
177
Lu (t
=6.65 days), 90Y (t
1/2
13
n cm−2 would yield 3.58 GBq of
176
Lu to
177
Lu of only 1.4% and thus results in low molar activity.
235
U) to generate neutrons. Each ssion
−1
). For example,
=2.67 days), and
1/2
176
176
Lu irradiated for three days at a
152
Sm incorporated into particles
186
Re (t
=3.72 days) can
1/2
Lu, 89Y, and
177
185
Re, respectively[1].
Lu. However, this repre-
Chapter 5: Production ofTherapeutic Radionuclides 91
nuclides from the stable target material, even if the target material is of the same
element. This process, discovered by L. Szilard and T.A. Chalmers in 1934 and published in
Nature, was used to separate
C
I. This method of capturing a recoiling radioactive atom has been used to generate
2H5
radionuclides with higher molar activity (Bq mol
128
I (t
=24.9 minutes) from neutron-irradiated ethyl iodide,
1/2
−1
) than would otherwise be possible from
direct neutron capture reactions[5].
Radionuclide products may also be generated from the decay of a radionuclide resulting from a direct neutron capture event. Typically, the “parent” radionuclide is produced
in the nuclear reaction, whereas the “daughter” radionuclide is the decay product used
for downstream chemistry and applications. In this manner, one can obtain high-purity
radionuclides with little or no non-radioactive contaminant (also referred to as no-carrier-
added). For example, as discussed previously,
177
Lu can be produced via the
176
Lu(n,γ)
177
Lu
reaction. However, this radionuclide can also be produced via neutron irradiation of Yb:
176
Yb(n,γ)
177
Yb → β− →
177
Lu[6]. In the case of receptor-targeted therapy where a high-
molar-activity product is desirable, this route may be preferable to the direct neutron
capture reaction. The challenge with this method is the lower yields: the nal product
must be separated from the target material, and the enriched target material must be
recycled[6].
Radionuclides with high molar activity can also be produced with higher energy or fast
neutrons, which induce (n,p) reactions. Typically, the production yields for these reactions
are much lower than the direct neutron capture reactions. As an example, several groups
have reported the production of
In a similar study,
47
Sc (t
=3.35 days) production rates via the 47Ti(n,p) reaction have also
1/2
64
Cu (t
=12.7 hours) via the 64Zn(n,p)64Cu reaction[7].
1/2
been reported[8]. In these proof of principle studies, Kolsky etal. reported the purica-
47
tion of
Sc from neutron-irradiated natural Ti and enriched targets by cation-exchange
chromatography using Dowex AG 50W resin. The optimized technique resulted in the
near-quantitative recovery (>95%) of
to 10 days at a neutron ux of 1.5 × 10
47
Sc with yields of 3.2 GBq for irradiation times of up
14
n cm−2[9].
In rare cases, radionuclides may be produced via double neutron capture. For example,
the radionuclide generator
188
(n,γ)
W reaction[10, 11]. This production route requires a very high neutron ux and is
188W/188
Re can be produced in this manner via the
186
W(n,γ)
only feasible at a few sites in the world.
Examples of therapeutic radionuclides and characteristics that can be produced by
neutron-induced reactions are shown in Table5.1.
Neutrons can also be used to induce ssion in target materials. Many radionuclides are
produced via ssion events where the
products: one with mass number (A)=80–110 and the other with A=130–150. One of the
most commonly used therapeutic radionuclides,
ssion of uranium.
92 Handbook of Radiopharmaceuticals
90
Sr (t
235
U nucleus splits asymmetrically into two ssion
131
I (t
=8 days), is mainly produced via
1/2
=28.8 years), the parent of the therapeutic 90Y (t
1/2
=2.67 days),
1/2
Primary
Radionuclide Half-life
47
Sc 3. 35 d β
90
Y 64 .0 5 h β
105
Rh 35.37 β
103
Pd 16 .9 9 d EC Auger
111
Ag 7.45 d β
117m
Sn 13 .8 d IT (1) Auger
131
I 8 .03 d β
149
Pm 53 .1 h β
153
Sm 4 6. 3 h β
166
Ho 2 6. 8 h β
177
Lu 6.65 d β
186
Re 3.72 d β
195m
Pt 4 .0 1 d β
decay mode
−
−
−
−
−
−
−
−
−
−
−
Mean particle
energy (keV) Production route
143
204
934
70
74
179
224
279
360
69
97
192
256
369
200
225
264
651
694
47.6
112
149
306
359
Auger
47
Ti(n,p)47Sc
89
Y(n,γ)90Y
104
Ru(n,γ)
102
Pd(n,γ)
110
Pd(n,γ)
116
Sn(n,γ)
130
Te(n,γ)
148
Nd(n,γ)
152
Sm(n,γ)
165
Ho(n,γ)
176
Lu(n,γ)
176
Yb(n,γ)
185
Re(n,γ)
194
Pt(n,γ)
195
Pt(n,n’γ)
194
Ir(n,γ)
111
117m
131
177
177
195m
195m
105
103
149
153
166
186
195m
Ru →
Pd
Pd →
Sn
Te →
Nd →
Sm
Ho
Lu
Yb →
Re
Pt
Pt
Ir →
105
111
131
177
195m
149
Table 5.1 Radionu-
clides produced by
neutron-induced
reactions.
Rh
Ag
I
Pm
Lu
Pt
can also be produced via ssion. This reaction route results in the coproduction of a
substantial number of radionuclides and requires signicant chemistry in order to pro-
vide radiochemically pure products. Further, the number of available nuclear reactors
that routinely produce radionuclides by ssion is small. In many cases, these research
or commercial reactors are nearing or have exceeded their operating lifetime and serve
as the primary supply point for critically important medical radionuclides. The future
decommissioning of these reactors and concern over the use of highly enriched uranium
targets has generated interest in developing new methods of producing therapeutic
radionuclides commonly derived via ssion.
Chapter 5: Production ofTherapeutic Radionuclides 93
Charged particles for the production of therapeutic radionuclides can be accelerated via
cyclotrons or linear accelerators. As the stopping power (MeV mm
is much higher than neutrons or photons, smaller targets can typically be used. However,
the substantial heat deposited by the beam in the target material should be considered
when designing targets for the production of radionuclides via these reactions.
Charged-particle reactions typically lead to proton-rich radionuclides, which are
mainly used for diagnostic purposes. However, high-energy proton-induced reactions on
heavy target nuclei and alpha particle-induced reactions can be used to produce some
alpha emitters. Additionally, recent studies with lower-energy medical cyclotrons have
shown the feasibility of producing radionuclides that decay by electron capture, which
results in the emission of Auger electrons. The emitted Auger electrons can be used for
therapeutic applications due to their high LET.
A handful of beta-emitting radionuclides are accessible via charged-particle reactions. Of
67
these,
duced via high-energy proton bombardment on enriched zinc metal or oxide targets via
the
relatively low (10 mb), high yields can be obtained with high-current irradiations. Separation of this isotope from the target material and coproduced cobalt radionuclides is typi-
cally accomplished via ion-exchange techniques.
capture on stable rhenium targets in a reactor, this isotope can also be produced via
proton or deuteron bombardment of enriched
irradiation of natural isotopic composition targets results in the production of a variety
of radioactive rhenium products, enriched targets are typically used for the production
of this radionuclide. For the proton or deuteron production of
gets, the cross-section reaches a maximum of 80 mb at 10 MeV and 600 mb at 15 MeV,
respectively[16]. Targets prepared of tungsten metal, oxides, carbides, and suldes have
been reported[14, 15]. Purication of the radionuclide from the target material can be
accomplished via dry distillation techniques or wet chemical separation using liquid-liquid
extraction or ion-exchange methods[15, 17–19].
teron irradiation of Ti or Ca targets via a number of dierent reactions:
46
tion route has a signicantly higher cross-section, the enriched target material is more
Cu (t
=2.58 days) is likely the most widely studied. This radionuclide can be pro-
1/2
68
Zn(p,2p)67Cu reaction[12, 13]. While the maximum cross-section for this reaction is
While low-specic-activity
The production of
47
186
Re (t
=3.72 days) can be readily produced via neutron
1/2
186
W or Osmium targets[14, 15]. As the
Sc (t
=3.35 days) has been reported via the proton or deu-
1/2
Ca(d,n)47Sc, 44Ca(α,p)47Sc, 48Ti(p,2p)47Sc, and 50Ti(p,α)47Sc[9, 20–22]. While the Ca produc-
−1
) of charged particles
186
Re from tungsten tar-
48
Ca(p,2n)47Sc,
94 Handbook of Radiopharmaceuticals