Production logistics of Lu-177 for radionuclide therapy

Abstract

Owing to its favourable decay characteristics 177Lu [T(1/2)=6.71 d, Ebeta(max)=497 keV] is an attractive radionuclide for various therapeutic applications. Production of 177Lu using [176Lu (n,gamma)177Lu] reaction by thermal neutron bombardment on natural as well as enriched lutetium oxide target is described. In all, approximately 4 TBq/g (108 Ci/g) of 177Lu was obtained using natural Lu target after 7 d irradiation at 3 x 10(13) n/cm2/s thermal neutron flux while it was approximately 110 TBq/g (3000 Ci/g) of 177Lu when 60.6% enriched 176Lu target was used. In both the cases, radionuclidic purity was approximately 100%, only insignificant quantity of 177mLu [T(1/2)=160.5 d, Ebeta(max)=200 keV] could be detected as the radionuclidic impurity. Production logistics using different routes of production is compared. Possible therapeutic applications of 177Lu are discussed and its merits highlighted by comparison with other therapeutic radionuclides.

Full text

Applied Radiation and Isotopes 59 (2003) 109–118
Production logistics of
177
Lu for radionuclide therapy
M.R.A. Pillai*, Sudipta Chakraborty, Tapas Das,
Meera Venkatesh, N. Ramamoorthy
Radiopharmaceuticals Division, Bhabha Atomic Research Centre, Mumbai 400 085, India
Received 9 January 2003; received in revised form 3 May 2003; accepted 14 May 2003
Abstract
Owing to its favourable decay characteristics
177
Lu [T1=2¼6:71 d, EbðmaxÞ¼497 keV] is an attractive radionuclide
for various therapeutic applications. Production of
177
Lu using [
176
Lu ðn;gÞ
177
Lu] reaction by thermal neutron
bombardment on natural as well as enriched lutetium oxide target is described. In all, B4 TBq/g (108 Ci/g) of
177
Lu was
obtained using natural Lu target after 7 d irradiation at 3 10
13
n/cm
2
/s thermal neutron flux while it was B110 TBq/g
(3000 Ci/g) of
177
Lu when 60.6% enriched
176
Lu target was used. In both the cases, radionuclidic purity was B100%,
only insignificant quantity of
177m
Lu [T1=2¼160:5d, Ebðmax)=200 keV] could be detected as the radionuclidic
impurity. Production logistics using different routes of production is compared. Possible therapeutic applications of
177
Lu are discussed and its merits highlighted by comparison with other therapeutic radionuclides.
r2003 Elsevier Ltd. All rights reserved.
Keywords: Radionuclide therapy;
177
Lu;
177m
Lu; Radionuclidic purity; Specific activity; Targeted therapy
1. Introduction
Radionuclide therapy (RNT) employing open sources
of radiotherapeutic agents is fast emerging as an
important part of nuclear medicine, primarily due to
the development of sophisticated molecular carriers
(Volkert et al., 1991;Volkert and Hoffman, 1999;
Srivastava and Dadachova, 2001;Ercan and Caglar,
2000;Jhu et al., 1998;Meredith et al., 1996;Delaloye
and Delaloye, 1995). In order to develop effective
radiopharmaceuticals for therapy, it is essential to
carefully consider the choice of appropriate radio-
nuclides as well as the carrier moiety with suitable
pharmacokinetic properties that could result in good
in vivo localization and desired excretion (Volkert et al.,
1991;Wessels and Rogus, 1984;Fritzberg et al., 1995).
The major criteria for the choice of a radionuclide for
radiotherapy are suitable decay characteristics, ease of
production and amenable chemistry. As regards the
decay characteristics, physical half-life of the radio-
nuclide should match with the biological half-life of the
radiopharmaceutical. The energy of the particulate
emission should be compatible to the volume of lesion
to be irradiated and at the same time should result in
minimal dose delivery to the tissues surrounding the site
of localization. Also, the ratio of non-penetrating to
penetrating radiation should be high (Volkert and
Hoffman, 1999;Srivastava and Dadachova, 2001;Qaim,
2001;Ehrhardt et al., 1998;Mausner et al., 1998). Other
practical considerations in selecting a radionuclide for
targeted therapy are availability in high radionuclidic
purity as well as high specific activity and production
logistics.
In recent years there is considerable interest in the
standardization of easy and economically viable produc-
tion methods for promising therapeutic radioisotopes
such as
188
Re,
186
Re,
90
Y,
153
Sm and
166
Ho. The
radionuclidic characteristics and methods of production
of these radionuclides are shown in Table 1. These
radionuclides in the form of labelled compounds or
conjugates of suitable biomolecules have already been
ARTICLE IN PRESS
*Corresponding author. Tel.: +91-22-2559-3676; fax: +91-
22-2550-5345.
E-mail address: ambi@magnum.barc.ernet.in
(M.R.A. Pillai).
0969-8043/03/$ - see front matter r2003 Elsevier Ltd. All rights reserved.
doi:10.1016/S0969-8043(03)00158-1

widely investigated (Das et al., 2000;De Jong et al.,
1998;Goeckeler et al., 1987;Ma et al., 1996;Mausner
and Srivastava, 1993;Mumper et al., 1992).
177
Lu is a
radioisotope having very good potential for use in in vivo
therapy, because of its favourable decay characteristics.
177
Lu decays with a half-life of 6.71 d by emission of b

particles with Emax of 497 keV (78.6%), 384 keV (9.1%)
and 176 keV (12.2%) to stable
177
Hf. It also emits g
photons of 113 keV (6.4%) and 208 keV (11%) (Fire-
stone, 1996), which are ideally suited for imaging the
in vivo localization with a gamma camera. The physical
half-life of
177
Lu is comparable to that of
131
I, one of the
most commonly used radioisotopes for radionuclide
therapy. The long half-life of
177
Lu provides logistic
advantage for facilitating supply to places far away from
the reactors.
177
Lu can be produced by two different routes,
namely, by irradiation of natural Lu
2
O
3
target (
176
Lu,
2.6%) or enriched (in
176
Lu) Lu
2
O
3
target, as also by
irradiation of Yb target (Yb
2
O
3
) followed by radio-
chemical separation of
177
Lu from Yb isotopes. The
above two production routes lead to the product having
different specific activities. Although the specific activity
obtained in (n;g) activation is usually low, owing to the
high thermal neutron capture cross-section of
176
Lu
(s¼2100 b) direct (n;g) activation of even natural Lu
2
O
3
powder results in reasonably high specific activity of
177
Lu. Table 2 lists the isotopic abundance of natural Lu
and all possible ðn;gÞactivation products. The specific
activity could be further enhanced considerably by using
Lu target enriched in
176
Lu, by carrying out irradiation
in a high flux reactor, as well as optimizing the duration
of irradiation. On the other hand, activation of
176
Yb
and subsequent bdecay gives no carrier added (NCA)
177
Lu. However, in this technique, radiochemical se-
paration of
177
Lu activity from irradiated Yb
2
O
3
target
is very crucial because of the radionuclidic purity
requirement.
In the present paper, we describe the production of
177
Lu by ðn;gÞactivation using natural as well as
enriched Lu
2
O
3
target and its possible uses in various
radiotherapeutic applications. Also, a comparison be-
tween the two above-mentioned routes for production of
177
Lu has been drawn and the merits of ðn;gÞactivation
as a simple and viable production route of
177
Lu are
highlighted.
2. Materials and methods
Natural Lu
2
O
3
powder (spectroscopic grade,
>99.99% pure) was obtained from Johnson Matthey
& Co. Ltd., UK. A weighed amount (typically 6 mg) of
natural Lu
2
O
3
powder was irradiated in Dhruva reactor
for 3–7 days at a thermal neutron flux of B310
13
n/
cm
2
/s. The irradiated target was dissolved in 1 M HCl by
gentle warming inside a lead-shielded plant. The
resultant solution was evaporated to near dryness and
reconstituted in double distilled water. A known aliquot
ARTICLE IN PRESS
Table 1
Decay characteristics and production routes of some important radionuclides for therapy
Radionuclide T1=2Eb;maxðMeVÞEg(keV) (%) Method of production
Nuclear reaction % Natural
abundance
s
th
(barns)
188
Re 16.9 h 2.12 155 (15)
188
W(69.4 d)–
188
Re generator — —
187
Re (n;g) 62.6 73
186
Re 90.6 h 1.07 137 (9)
185
Re (n;g) 37.4 106
90
Y 64.1 h 2.27 —
90
Sr(28.3 d)–
90
Y generator — —
89
Y(n;g) 100 1.3
153
Sm 46.3 h 0.81 103 (28)
152
Sm (n,g) 26.7 206
166
Ho 26.9 h 1.85 81 (6.4)
165
Ho (n;g) 100 66
89
Sr 50.5 d 1.49 —
88
Sr(n;g) 82.6 0.0058
117m
Sn 13.6 d 0.13, 0.15 159 (86)
116
Sn(n;g) 14.4 0.006
117
Sn(n;n=;g) 7.7 —
Table 2
Isotopic abundance of natural Lu and activation products
Isotope % Natural
abundance
s(barns) for
(n;g) reaction
Product RN and
its characteristics
175
Lu 97.4 7
176
Lu (stable)
16.4
176m
Lu (b;3.7 h)
a
176
Hf (stable)
176
Lu 2.6 2100
177
Lu (b;6.71 d)
177
Hf (stable)
7
177m
Lu (b

,
160.5 d)
b 177
Hf
(stable)
a
Eb=1.2 MeV, Eg¼88 keV.
b
Eb¼0:2MeV;Eg¼128;153;228;378;414;418 keV:
M.R.A. Pillai et al. / Applied Radiation and Isotopes 59 (2003) 109–118110

was drawn for assessment of radioactivity content and
radionuclidic purity evaluation.
For production of high specific activity
177
Lu,
isotopically enriched Lu
2
O
3
(60.6%
176
Lu) (Isoflex,
USA) was irradiated. A stock solution of enriched
target was prepared by dissolving enriched Lu
2
O
3
powder in 0.1 M HCl (1 mg ml
1
concentration). A
known aliquot of this solution was taken in a quartz
ampoule and carefully evaporated to dryness. The
ampoule was subsequently flame sealed and irradiated
after placing inside an aluminium can. The can was
irradiated at a thermal neutron flux of 3 10
13
n/cm
2
/s
for 3–7 days. The chemical processing of the irradiated
target was carried out as described above for the natural
Lu
2
O
3
target.
Radioactivity assay was carried out by measuring the
ionization current obtained when an aliquot of the batch
was placed inside a pre-calibrated well-type ion-chamber
(the calibration factor for the ion chamber for
177
Lu was
arrived at 1.143 10
14
A MBq
1
). Radionuclidic purity
was determined by recording g-ray spectrum of the
appropriately diluted solution of the irradiated target
using an HPGe detector (EGG Ortec/Canberra detec-
tor) connected to a 4 K multichannel analyser (MCA)
system. A
152
Eu reference source (Amersham Inc.) was
used for both energy and efficiency calibration. All the
nuclear data used were taken from the Table of isotopes
(Firestone, 1996). Several spectra were recorded for each
batch at regular time intervals. Samples measured
initially for the assay of
177
Lu were preserved for
complete decay of
177
Lu (over 10–15 T
1/2
of
177
Lu, i.e.
for a period of 2–3 months) and re-assayed to determine
the activity of long-lived
177m
Lu (T1=2¼160:5 d). Appro-
priately diluted sample solutions were counted for 1 h.
3. Results and discussion
The typical yields of
177
Lu from natural as well as
enriched targets for different durations of irradiation in
Dhruva reactor (3 10
13
n/cm
2
/s) are shown in Table 3.
These values are in excess of theoretically calculated
values, which are also given alongside in Table 3. This
could perhaps be attributed to the contribution from
epithermal neutrons (resonance integral=1087 b), which
is not accounted in theoretical calculations (Knapp et al.,
1995;Ramamoorthy et al., 2002). The variations in the
yield of
177
Lu between different batches are mostly due
to fluctuations in the irradiation conditions such as the
exact duration, intervening shutdown and variation of
neutron flux due to the power level of reactor operation.
The radionuclidic purity of
177
Lu produced from
either natural or enriched target was B100% as
estimated by analysing the g-ray spectrum. In a typical
g-ray spectrum of the irradiated target after chemical
processing (Fig. 1), the major gpeaks observed were 72,
113, 208, 250 and 321 keV, all of which correspond to
the photopeaks of
177
Lu (Firestone, 1996). This was
further confirmed from the decay as followed by
monitoring peak area cps values at those peaks
according to the half-life of
177
Lu. It is worthwhile to
mention that there is a possibility of formation of
177m
Lu (T1=2¼160:5 d) on thermal neutron bombard-
ment of Lu
2
O
3
target (Knapp et al., 1995;Neves et al.,
2002). However, g-ray spectrum of the irradiated Lu
target after chemical processing did not show any
significant peak corresponding to the photopeaks of
177m
Lu (128, 153, 228, 378, 414, 418 keV) (Firestone,
1996). This is expected as the radioactivity due to
177m
Lu
produced will be insignificant and below the detectable
limit on a 7 d irradiation owing to its long half-life and
comparatively low cross-section (s=7 barns) for its
formation. Attempt to assay any trace level of
177m
Lu
activity by recording g-ray spectrum of a sample aliquot,
initially having high radioactive concentration, after
complete decay of
177
Lu activity showed the presence of
trace level of
177m
Lu. The average level of radionuclidic
impurity burden in
177
Lu due to
177m
Lu was found to
be 5.5 kBq of
177m
Lu/37 MBq of
177
Lu (150 nCi/1 mCi)
at EOB.
3.1. Production logistics—yield and specific activity
Owing to its suitable decay characteristics, the
usefulness of
177
Lu in radionuclide therapy (RNT)
has already been pointed out (Volkert et al., 1991;
Srivastava and Dadachova, 2001;Liu et al., 2001;Ma
et al., 1996;Stein et al., 2001,Sola et al., 2001).
Although, natural Lu target contains only 2.6% of
176
Lu, the specific activity of
177
Lu obtained by ðn;gÞ
activation of natural Lu
2
O
3
target is reasonably high
because of the very high thermal neutron capture cross-
section. In fact the cross-section (2100 barns) is the
highest encountered among all (n;g) produced
radionuclides presently used for therapy. The high
ARTICLE IN PRESS
Table 3
Specific activities of
177
Lu produced from natural and enriched
Lu
2
O
3
target due to thermal neutron bombardment at a flux of
B310
13
n/cm
2
/s
Target Duration of
irradiation (d)
Specific activity
(at EOB) (TBq/g)
Theoretical Experimental
Natural Lu 3 1.50 2.570.3
5 2.27 3.370.2
7 2.90 4.070.3
Enriched Lu
(60.6%
176
Lu)
3 34.96 7275
5 52.91 9273
7 67.59 11075
M.R.A. Pillai et al. / Applied Radiation and Isotopes 59 (2003) 109–118 111

cross-section also ensures that there will be no con-
straints with respect to large-scale production of the
isotope. It is feasible to produce
177
Lu with high specific
activity suitable for developing agents for targeted
radiotherapy, taking the advantage of high flux reactor
and target enriched in
176
Lu.
However, a careful optimization of the time of
irradiation will have to be carried out in order to obtain
the highest specific activity. In high flux reactors the
target burn up will be considerably high due to the high
thermal neutron capture cross-section of
176
Lu and
hence, the usual assumption that the number of target
atoms remain constant during the period of irradiation
will not be valid in this case. Considering the number of
target atoms is a function of irradiation time, the
commonly used differential equation,
dN2=dt¼N1sf N2l
can be modified as,
dN2=dt¼N0esftsf N2l;
where, N
0
is the number of
176
Lu atoms used as target
(at t¼0), N
1
the number of
176
Lu atoms at any time t;
N
2
the number of
177
Lu atoms at any time t;lthe decay
constant of
177
Lu, sthe thermal neutron capture cross-
section of
176
Lu, fthe thermal neutron flux of the
reactor and tthe time of irradiation.
177
Lu activity
produced at the end of bombardment can be calculated
by the following equation, which is obtainable by solving
the modified differential equation mentioned above
A¼N0lsf
lsf esftelt

:
The
177
Lu activity produced at the end of bombardment
as a function of irradiation time at three different
thermal neutron fluxes has been calculated on the basis
of above equation and the results are shown in Fig. 2.It
is evident from the figure, that depending on neutron
flux the activity of
177
Lu produced will be maximum
after a certain duration of irradiation, beyond which the
activity will decrease owing to the high target burn up.
Higher the thermal neutron flux of the reactor, shorter
will be the time of irradiation for attaining maximum
activity. Therefore, in order to obtain maximum specific
activity using enriched
176
Lu target, the time of
irradiation must be judiciously decided as per the
neutron flux available.
ARTICLE IN PRESS
0 1000 2000
1
10
100
1000
10000
Net Counts (log scale)
Channel Number
321keV(177Lu)
250keV(177Lu) 208keV(177Lu)
113 keV(177Lu)
72 keV (177Lu)
Fig. 1. g-ray spectrum of
177
Lu.
403020100
50
40
30
20
10
0
Specific acticity (Atom %)
Irradiation time (d)
5x10
13
n/cm
2
/s
5x10
14
n/cm
2
/s
1x10
15
n/cm
2
/s
Fig. 2. Variation of
177
Lu activity with respect to duration of
irradiation at different thermal neutron fluxes.
M.R.A. Pillai et al. / Applied Radiation and Isotopes 59 (2003) 109–118112

It is also pertinent to point out that
175
Lu present in
the natural Lu target (97.4%) will also contribute to
177
Lu activity produced by undergoing successive
neutron capture. Fig. 3 shows the
177
Lu activity
produced from
175
Lu and
176
Lu in natural Lu
target along with
177
Lu activity obtainable from 100%
enriched
176
Lu target for 7 d irradiation at various
thermal neutron fluxes. Although, the
177
Lu activity
produced by double neutron capture of
175
Lu is
insignificant at relatively low neutron flux, the contribu-
tion from this route becomes quite significant with the
increase of neutron flux. This is because of the activity of
a radionuclide produced by a successive neutron capture
process being proportional to the square of the neutron
flux. It is evident from Fig. 3 that at a flux of 1 10
15
n/
cm
2
/s,
175
Luðn;gÞ
176
Luðn;gÞ
177
Lu is a significant con-
tributor to
177
Lu activity produced. Therefore, the
specific activity of
177
Lu obtainable from high flux
reactor using natural Lu target will be higher than that
expected from
176
Luðn;gÞ
177
Lu only.
3.2. Potential for use in metastatic bone pain palliation
Some promising forms of RNT do not require high
specific activity radionuclides for the formulation of the
radiotherapy agents. An important class of such radio-
therapy agents has been widely used in palliative
treatment of skeletal metastases.
153
Sm-EDTMP and
186
Re-HEDP, the two agents used most extensively in
this purpose employ ðn;gÞproduced carrier added
radioisotopes (Mausner et al., 1998;Ketring, 1987).
Although, enriched
152
Sm is normally used as the target
for the preparation of
153
Sm-EDTMP, it was demon-
strated that
153
Sm produced from even natural Sm
target can be effectively utilized in making patient dose
of
153
Sm-EDTMP (Ramamoorthy et al., 2002). How-
ever, the radionuclide will have to be used within 3–4 d,
in view of the presence of radionuclidic impurities of
154
Eu (T1=2¼8:8 y) and
155
Eu (T1=2¼4:96 y). The utility
of
153
Sm (T1=2¼46:3hÞis also limited in places not well
connected with reactor site and in countries having poor
transport logistics. Due to the 46.3 h half-life, substan-
tial quantities of the isotope produced at EOB is lost by
decay during chemical processing, preparation and
quality control of radiopharmaceuticals and subsequent
transportation. Hence, the activity to be produced in
case of
153
Sm at EOB will have to be several times higher
than that used for actual administration. This necessi-
tates the handling of large quantum of activity while
making
153
Sm products. Similarly logistics problems
affect the merit of
186
Re (T1=2¼90:6 h) also. One has to
either use highly enriched
185
Re target for irradiation or
avail of prolonged cooling periods of B120 h (B5T
1/2
of
188
Re) to let
188
Re decay to acceptable level while using
natural Re targets. Long-lived isotopes such as
89
Sr is
also very effectively used for bone pain palliation. The
major advantage with
89
Sr is that due to the long half-
life (50.5 d) the activity administrable can be signifi-
cantly lower, say 4–5 mCi, to give the required
cumulative dose. The long half-life of
89
Sr also helps
in transportation of the radiopharmaceutical across the
world. However, the production of
89
Sr in adequate
quantities is expensive due to the low cross-section of
88
Sr (5.8 mb) for thermal neutron capture and hence the
need to have very high neutron flux for irradiation and
preferably also enriched target. Again,
89
Sr is also
expected to give high bone marrow dose due to the
emission of high energy bparticles (EbðmaxÞ=
1.49 MeV). Due to the above reasons, despite being an
efficacious radiopharmaceutical, utility of
89
SrCl
2
has
remained limited. Considering that 50–70 mCi is the
patient dose for
153
Sm, 35–40 mCi for
186
Re and about
4–5 mCi for
89
Sr for the same application, the dose
requirement for
177
Lu is expected to be considerably
lower, say B15–20 mCi, than that of
153
Sm and
186
Re to
give the same cumulative dose. Taking into considera-
tion the comparatively lesser decay loss as well the lower
per patient dose, it will be possible to treat at least 10
times more patients with the same amount of activity at
EOB with
177
Lu as compared to
153
Sm. In the present
study, we have found that neutron activation of natural
Lu target at a moderate thermal neutron flux of
310
13
n/cm
2
/s for 7 d (Bhalf saturation value) pro-
duces 4 TBq/g (108 Ci/g) of
177
Lu. Hence, irradiation of
100 mg of natural Lu target at 3 10
13
n/cm
2
/s can yield
B10 Ci of
177
Lu sufficient for treating 300–500 patients.
The most significant advantage of
177
Lu will hence be
ARTICLE IN PRESS
Fig. 3.
177
Lu activity produced from
175
Lu and
176
Lu in natural
Lu target along with
177
Lu activity obtainable from 100%
enriched
176
Lu target for 7 d irradiation at various thermal
neutron fluxes.
M.R.A. Pillai et al. / Applied Radiation and Isotopes 59 (2003) 109–118 113

the easy and economical production of large quantities
of the radioisotope in relatively low flux reactor
available in several places around the world. Also,
enriched target will not be essential for making
therapeutic radiopharmaceuticals for bone pain pallia-
tion. In relatively high flux reactors (say, 510
14
n/cm
2
/s)
the specific activity could be increased to B35 TBq/g
(950 Ci/g) as obtained from theoretical calculations.
177
Lu labelled linear as well as cyclic polyaminophos-
phonic acid ligands have been prepared with high yield
at low [ligand]:[metal] ratio and evaluated in animal
models as potential agents for palliation of pain due to
bone metastasis (Chakraborty et al., 2002;Das et al.,
2002). Some of these agents have shown excellent
properties as bone seeking radiophamaceuticals.
117m
Sn(IV)-DTPA was proposed as an efficacious
bone pain palliation agent by Atkins et al. (1995);
Srivastava et al. (1998).
117m
Sn derives its therapeutic
strength from the copious Auger electron emission
following electron capture (EC) decay (T1=2¼13:6 d).
The accompanying 156 keV (86%) gamma photon
emission is not exactly preferred for RNT applications.
Nonetheless, thanks to 0.2–0.3 mm tissue range of the
Auger electron of
117m
Sn, very effective dose deposition
on bone surface as well as very high bone-to-bone
marrow ratio has been demonstrated while using
117m
Sn(IV)-DTPA (Bishayee et al., 2000).
117m
Sn is
however, difficult to produce economically in large
quantities and hence the utility has remained limited.
The excellent matching of tissue range of
177
Lu
(0.5–0.6 mm) for similar dose deposition can be inferred
from the above results reported with the use of
117m
Sn-
DTPA. The adequacy of EbB0:5 MeV of
177
Lu for
effective bone pain palliation can hence be advocated
as major attractive feature in addition to the convenient
half-life and production logistics enumerated earlier.
The mylotoxicity of
177
Lu complex will be slightly
higher than that of the Auger electron of
117m
Sn, but the
86% abundant 156 keV gamma emission of
117m
Sn
places additional burden of absorbed dose to far off
tissues/organs.
3.3. Potential for other therapeutic applications
Besides, radiolabelled particulates or microspheres for
therapy of hepatic tumour also require only low specific
activity radionuclides and
177
Lu could be used for
development of these agents.
177
Lu could be very
effective in radiation synovectomy of medium size joints
and could be a good replacement for the difficult to
produce
169
Er which is used in small joint radiation
synovectomy (Deutsch et al., 1993). It has already been
reported that, for medium size joints dose requirement
for
153
Sm labelled hydroxy apatite (HA) is B74 MBq
(B2 mCi). For
177
Lu-HA, the activity requirement is
expected to be even lower, due to the long half-life and
higher cumulative dose per mCi activity. Hence,
177
Lu
could be an attractive candidate for the preparation of
radiopharmaceuticals for radiation synovectomy.
In the case of applicability of larger lesions,
177
Lu
would suffer due to lower Eb:Thus for the current
approaches, such as, hepatocellular carcinoma,
177
Lu
would be unattractive now. Advances in diagnostic
techniques continuously helping early pick up of smaller
lesions could bring further a demand for medium b

energy products too in future. In fact, ‘‘patient tailored
RNT’’ has been reported as a distinct possibility in
future. Potential of
177
Lu applicability is thus far
reaching.
In targeted radiotherapy using peptides and other
receptor based carrier molecules, the use of high specific
activity (preferably no carrier added) radionuclide in
formulating the radiopharmaceutical is essential in order
to deliver sufficient number of radionuclides to the
target site without saturating the target (Volkert et al.,
1991;Mausner and Srivastava, 1993). The use of high
specific activity
177
Lu in targeted therapy (conjugated to
receptor specific peptides or monoclonal antibodies) has
already been proposed (Ehrhardt et al., 1998;Srivastava
and Dadachova, 2001;Meredith et al., 1996;Mausner
and Srivastava, 1993;Liu et al., 2001;Stein et al., 2001).
3.4. Specific activity aspects
No carrier added
177
Lu can be produced by irradia-
tion of enriched
176
Yb target. The nuclear reaction
leading to the formation of no carrier added
177
Lu is
176Ybðn;gÞ
s¼2:4b
177Yb T1=2¼1:9h
!
b
177Lu T1=2¼6:71 d
!
b
177Hf ðstableÞ
From theoretical calculations employing the appro-
priately modified form of the Bateman equation, it
can be shown that irradiation of 1 mg of 99% enriched
176
Yb
2
O
3
target at a reasonably high thermal neutron
flux of 5 10
14
n/cm
2
/s will produce B5.55 GBq
(B150 mCi) of
177
Lu which is carrier free having a
theoretical specific activity of 1.09 10
5
Ci/g. However,
radiochemical separation of
177
Lu activity from irra-
diated Yb
2
O
3
target is a difficult task owing to the
similarity in the chemistry of the two adjacent members
of the lanthanide series. Presence of Yb will reduce the
effective specific activity of the product. Since Yb
+3
is
an equally good complexing ion, it would interfere
during the preparation of radiopharmaceuticals. This is
one of the major impediment in the ready production of
no carrier added
177
Lu via
176
Yb (n;g;b)
177
Lu route.
Moreover, use of enriched targets with low activation
cross-section is not economical for isotope production,
as a significant part of the target will be wasted. Though,
theoretically in the present case recovery of the target is
feasible, there will be practical problems.
ARTICLE IN PRESS
M.R.A. Pillai et al. / Applied Radiation and Isotopes 59 (2003) 109–118114

We have produced
177
Lu having specific activity of
B110 TBq/g (3 10
3
Ci/g) by irradiation of commer-
cially available enriched Lu
2
O
3
powder at a flux of
310
13
n/cm
2
/s for 7 d (Table 3). From theoretical
calculations, the specific activity of
177
Lu will be
800 TBq/g (2.17 10
4
Ci/g) in a relatively high flux
reactor (say, 5 10
14
n/cm
2
/s) under the same irradia-
tion conditions. In other words, it translates to a specific
activity of B0.20 atoms of
177
Lu per atom of all
lutetium nuclides, resulting in 20% of the maximum
achievable theoretical specific activity. This is expected
to be sufficient for the preparation of targeted therapy
agents involving minimum number of carrier molecules
capable of delivering required dose to the target (Volkert
et al., 1991;Mausner and Srivastava, 1993). As for
example, in the formulation of
177
Lu labelled peptide for
targeted tumour therapy, 10 mg of peptide (say,
B1000 D molecular weight) could incorporate >1 mg
of Lu. With a specific activity of 2.17 10
4
Ci/g, even
1mg corresponds to 21.7 mCi of
177
Lu activity. Similarly,
in radioimmunotherapy (RIT), assuming the conjuga-
tion of an average of one lutetium atom per monoclonal
antibody (Mab) molecule, 1 mg Mab preparation could
hold>1 mgof
177
Lu corresponding to an activity of
>20 mCi sufficient to deliver the required therapeutic
dose (Volkert et al., 1991;Wessels and Rogus, 1984;
Mausner and Srivastava, 1993). The major advantage of
the present method of production is that a simple and
quick post irradiation chemical processing gives radio-
nuclidically as well as radiochemically pure product.
3.5. Comparison with other therapeutic radionuclides
The use of no carrier added (NCA) bemitting
radionuclides in radioimmunotherapy (RIT) and other
forms of targeted therapy is a very popular practice
(Volkert et al., 1991;Srivastava and Dadachova, 2001;
Mausner and Srivastava, 1993). Most of the NCA
radionuclides are produced in the reactors either via
indirect reactions or obtained from generator systems
where the longer-lived parent radionuclide may be
obtained by direct neutron activation. Table 4 gives a
list of most widely used radionuclides in NCA levels in
targeted therapy.
188
Re and
90
Y are the two most
attractive generator-produced radionuclides for RNT
applications.
188
Re is available in the NCA form from a
188
W–
188
Re generator installed at hospital radiophar-
macy (Kamioski et al., 1994). However, the availability
of
188
W, the parent radionuclide, in adequate quantity
and specific activity is restricted, since it is produced by a
double neutron capture reaction. Only very few reactors
in the world having thermal neutron flux of the order of
>5 10
14
n/cm
2
/s (such as, HFIR of ORNL, USA,
MIR.M1 and SM reactor of Russian fedaration, BR-2
reactor of Belgium) are capable of producing reasonable
quantities of
188
W for the preparation of
188
W–
188
Re
generator.
90
Sr–
90
Y generator system is now being used
to provide NCA
90
Y for targeted therapy (Venkatesh
et al., 2001). However, the possible radionuclidic
contaminant
90
Sr (T1=2¼28:3 y, a natural bone seeker)
and other trace metals in the eluted
90
Y is a very crucial
factor.
105
Rh is another isotope having suitable decay
properties for therapeutic applications and can be made
available in NCA form through
104
Ru (n;g;b)
105
Rh
route. However, production of
105
Rh with high radio-
chemical and radionuclidic purity involves multi-step
chemical preparations and purification that prohibit its
widespread application (Volkert et al., 1991;Grazman
and Troutner, 1988;Unni and Pillai, 2002). Moreover,
due to poor activation cross-section, large quantities of
the target need to be handled during irradiation and
chemical processing.
67
Cu is also very convenient for
RIT and several bifunctional chelating agents have been
developed for its conjugation to monoclonal antibodies
and other biomolecules (Rogers et al., 1996). However,
the cross-section for (n;p) production of
67
Cu
(s¼0:0012 b) is too small to enthuse reactor production
in sufficient quantities for widespread therapeutic
applications (O’Brien, 1969).
Even though many bemitting radionuclides show
considerable promise for therapy,
131
I(T1=2¼8:03 d,
EbðmaxÞ¼0:81 MeV) continues to play a significant role
ARTICLE IN PRESS
Table 4
Radionuclides (produced in NCA form) most widely used in targeted therapy
Radionuclide T1=2ðhÞEb(max) (MeV) Eg(keV) Source
67
Cu 62 0.57 184 (48%)
67
Zn ðn;pÞ
67
Cu,
92 (23%) s=0.0012 b
105
Rh 35.5 0.57 319 (19%)
104
Ru (n;g;b)
105
Rh,
306 (5%) s=0.5 b
90
Y 64.1 2.27 —
235
U(n;f)
90
Sr-
90
Sr(28.3y)–
90
Y generator
188
Re 16.9 2.12 155 (15%)
186
W(n;g)
187
W(n;g)
188
W-
188
W(69.4 d)–
188
Re generator
47
Sc 80.2 0.60 159 (68%)
46
Ca (n;g;b)
47
Sc, s¼0:7b
47
Ti (n;p)
47
Sc (En>1 MeV)
M.R.A. Pillai et al. / Applied Radiation and Isotopes 59 (2003) 109–118 115

for RNT (Volkert et al., 1991). Though
131
I has a tissue
penetration range which is well suited for the treatment
of small tumours, the accompanying 364 keV gemission
with high abundance (81%) is a major drawback for its
use in therapeutic purposes.
177
Lu can be considered as a
viable alternative of
131
I for the therapy of non-thyroid
small sized tumours. Relatively low energy major g
emission (208 keV) with low abundance (11%) is a
distinct advantage in favour of
177
Lu.
177
Lu will also
give much lower external dose emanating from the
patient thereby enabling early discharge of the patient
from the isolation ward, which could save money as well
as making such isolation wards available for other
patient use. One of the advantages cited for the routine
use of
131
I is its availability with relatively high specific
activity (B17 atom%) (Volkert et al., 1991) from
commercial sources. However, our theoretical calcula-
tions show that,
177
Lu also can be produced with
comparable specific activity (B20 atom%) from a
moderately high flux reactor (B510
14
n/cm
2
/s) by 7
days irradiation of commercially available 60.6%
enriched
176
Lu target. It may also be possible to get
176
Lu with higher enrichment. The production of
131
I
needs more elaborate chemical separation and conse-
quent waste disposal measures, as it is produced either
by irradiation of natural tellurium target or from fission
of
235
U.
177
Lu, produced via neutron activation of enriched
176
Lu target is an attractive candidate for targeted
therapy applications, since it satisfies specific activity as
well as radionuclidic purity requirements and can be
made available in large quantities very easily in the
desired chemical form. A comparison of specific
activities of a few potential therapeutic radionuclides
produced by neutron activation or from radionuclide
generators is given in Table 5. Though atom% of
generator produced isotopes is theoretically 100%, in
actual practice the decay product of the radionuclide
though inactive will interfere in complexation thereby
reducing the effective specific activity.
4. Conclusions
177
Lu has got very good potential as a therapeutic
radionuclide especially in developing countries with
limited facilities of reactor irradiation and for indigen-
ous production capability of radiopharmaceuticals. The
high thermal neutron cross-section of
176
Luðn;gÞ
177
Lu
reaction facilitates large-scale production, while rela-
tively longer half-life provides logistic advantage for
production, radiochemical processing and transporta-
tion of finished radiopharmaceuticals. The present
studies show B4 TBq/g (108 Ci/g) and B110 TBq/g
(3000 Ci/g) of
177
Lu activity could be produced by
thermal neutron bombardment at a flux of 3 10
13
n/cm
2
/s for a period of 7 days using natural and enriched
(60.6%
176
Lu) Lu
2
O
3
targets, respectively.
177
Lu pro-
duced by neutron irradiation of natural Lu
2
O
3
target in
medium to high flux reactor could be used in the
development of bone pain palliation and radiation
synovectomy agents for small and medium size joints,
wherein the specific activity requirement is low. On
the other hand,
177
Lu produced with enriched target
will have adequate specific activity for labelling
peptides, antibodies etc. for targeted radiotherapy, such
that NCA
177
Lu production through tedious radio-
chemical separation from irradiated Yb target can be
avoided.
Acknowledgements
The authors acknowledge International Atomic En-
ergy Agency (IAEA), Vienna, for providing Lu target
enriched in
176
Lu for the present studies. The valuable
technical support of our colleagues, Mr. S. V. Thakare
and Mr. K. C. Jagadeesan, Radiopharmaceuticals
Division, Bhabha Atomic Research Centre in carrying
the irradiations in the Dhruva reactor is gratefully
acknowledged.
ARTICLE IN PRESS
Table 5
Theoretical specific activities of potential therapeutic radionuclides produced by neutron bombardment at a flux of 5 10
14
n/cm
2
/s for
7 d or from radionuclide generator
Isotope Source (enrichment) Cross-section (barns) Specific activity
Ci/g Atom%
188
Re
188
W/
188
Re generator — 9.80 10
5
100
188
Re
187
Re (n;g)
188
Re (99%) 73 3.14 10
3
0.32
186
Re
185
Re (n;g)
186
Re (99%) 106 3.35 10
3
1.76
90
Y
90
Sr/
90
Y generator 5.44 10
5
100
153
Sm
152
Sm (n;g)
153
Sm (98%) 206 8.60 10
3
2.21
166
Ho
165
Ho (n;g)
166
Ho (100%) 66 3.16 10
3
0.45
177
Lu
176
Lu (n;g)
177
Lu (60.6%) 2100 2.17 10
4
19.74
M.R.A. Pillai et al. / Applied Radiation and Isotopes 59 (2003) 109–118116

References
Atkins, H.L., Mausner, L.F., Srivastava, S.C., Meinken, G.E.,
Cabahug, C.J., D’Alessandro, T., 1995. Sn-117m (4+)
DTPA for palliation of painful osseous metastases: a pilot
study. J. Nucl. Med. 36, 725–729.
Bishayee, A., Rao, D.V., Srivastava, S.C., Bouchet, L.G.,
Bolch, W.E., Howell, R.W., 2000. Marrow-sparing effects
of
117m
Sn(4+) diethylenetriaminepentaacetic acid for
radionuclide therapy of bone cancer. J. Nucl. Med. 40,
2043–2050.
Chakraborty, S., Das, T., Unni, P.R., Sarma, H.D., Samuel,
G., Banerjee, S., Venkatesh, M., Ramamoorthy, N., Pillai,
M.R.A., 2002.
177
Lu–labeled polyaminophonates as poten-
tial agents for bone pain palliation. Nucl. Med. Commun.
23, 67–74.
Das, T., Banerjee, S., Samuel, G., Sarma, H.D., Ramamoorthy,
N., Pillai, M.R.A., 2000.
188
Re-Ethylene Dicysteine: a novel
agent for possible use in endovascular radiation therapy.
Nucl. Med. Commun. 21, 939–945.
Das, T., Chakraborty, S., Unni, P.R., Banerjee, S., Samuel, G.,
Sarma, H.D., Venkatesh, M., Pillai, M.R.A., 2002.
177
Lu–labeled cyclic polyaminophonates as potential agents
for bone pain palliation. Appl. Radiat. Isot. 57, 177–184.
De Jong, M., Bernard, B.F., De Bruin, E., Van Gameren, A.,
Bakker, W.H., Visser, T.J., Macke, H.R., Krenning, E.P.,
1998. Internalization of radiolabelled [DTPA
0
] octreotide
and [DTPA
0
, Tyr
3
] octreotide: peptides for somatostatin
receptor-targeted scintigraphy and radionuclide therapy.
Nucl. Med. Commun. 19, 283–288.
Delaloye, A.B., Delaloye, B., 1995. Radiolabeled monoclonal
antibodies in tumor imaging and therapy: out of fashion?
Eur. J. Nucl. Med. 22, 571–580.
Deutsch, E., Brodack, J.W.K., Duetsch, F., 1993. Radiation
Synovectomy revisited. Eur. J. Nucl. Med. 20, 1113–1127.
Ehrhardt, G.J., Ketring, A.R., Ayers, L.M., 1998. Reactor-
produced radionuclides at the University of Missouri
research reactor. Appl. Radiat. Isot. 49, 295–297.
Ercan, M.T., Caglar, M., 2000. Therapeutic radiopharmaceu-
ticals. Curr. Pharmaceut. Des. 6, 1085–1121.
Firestone, R., 1996. In: Shirley, V.S. (Eds.), Table of isotopes,
8th Edition. Wiley, New York.
Fritzberg, A.R., Gustavson, L.M., Hylarides, M.D., Reno,
J.M., 1995. In: Weiner, D.B., Williams, M.V. (Eds.),
Chemical and Stuctural Approaches to Rational Drug
Design. CRC Press, Boca Raton, FL, p. 125.
Grazman, B., Troutner, D.E., 1988.
105
Rh as a potential
radiotherapeutic agent. Appl. Radiat. Isot. 39, 257–260.
Goeckeler, W.F., Edwards, B., Volkert, W.A., Holmes, R.A.,
Simon, J., Winston, D., 1987. Skeletal localisation of
samarium-153 chelates: potential therapeutic bone agents.
J. Nucl. Med. 28, 495–504.
Jhu, H., Jain, R.K., Baxter, L.T., 1998. Tumor pretargeting for
radioimmunodetection and radioimmunotherapy. J. Nucl.
Med. 39, 65–76.
Kamioski, H., Mirzadeh, S., Lambrecht, R.M., Knapp Jr., R.,
Dadachova, K., 1994.
188
Re generator for biomedical
applications. Radiochim. Acta. 65, 39–46.
Ketring, A.R., 1987.
153
Sm-EDTMP and
186
Re-HEDP as
bone therapeutic radiopharmaceuticals. Nucl. Med. Biol.
14, 223–232.
Knapp, F.F., Ambrose, K.R., Beets, A.L., Luo, H., McPher-
son, D.W., Mirzadeh, S., 1995. Nuclear medicine program
progress report for quarter ending September 30, 1995, Oak
Ridge National Laboratory, pp. 6–16.
Liu, S., Cheung, E., Ziegler, M.C., Rajopadhye, M., Edwards,
D.S., 2001.
90
Y and
177
Lu labeling of a DOTA-conjugated
vitronectin receptor antagonist useful for tumor therapy.
Bioconjug. Chem. 12, 559–568.
Ma, D., Ketring, A.R., Ehrhardt, G.J., Jia, W., 1996.
Production of radiolanthanides and radiotherapy research
at MURR. J. Radioanal. Nucl. Chem. Articles 206,
119–126.
Mausner, L.F., Srivastava, S.C., 1993. Selection of radio-
nuclides for radioimmunotherapy. Med. Phys. 20, 503–509.
Mausner, L.F., Kolsky, K.L., Joshi, V., Srivastava, S.C., 1998.
Radionuclide development at BNL for nuclear medicine
therapy. Appl. Radiat. Isot. 49, 285–294.
Meredith, R.F., Partridge, E.E., Alvarez, R.D., 1996. Intraper-
itonial radioimmunotherapy of ovarian cancer with lute-
tium-177-CC49. J. Nucl. Med. 37, 1491–1496.
Mumper, R.J., Ryo, U.Y., Jay, M., 1992. Neutron activated
holmium-166- Poly (l-Lactic acid) microspheres: a potential
agent for internal radiation therapy of hepatic tumours.
J. Nucl. Med. 32, 2139–2143.
Neves, M., Kling, A., Lambrecht, R.M., 2002. Radionuclide
production for therapeutic radiopharmaceuticals. Appl.
Radiat. Isot. 57, 657–664.
O’Brien Jr., H.A., 1969. The preparation of
67
Cu from
67
Zn in a
nuclear reactor. Int. J. Appl. Radiat. Isot. 20, 121–124.
Qaim, S.M., 2001. Therapeutic radionuclides and nuclear data.
Radiochim. Acta 89, 297–302.
Ramamoorthy, N., Saraswathy, P., Das, M.K., Mehra, K.S.,
Ananthakrishnan, M., 2002. Production logistics and radio-
nuclidic purity aspects of
153
Sm for radionuclide therapy.
Nucl. Med. Commun. 23, 83–89.
Rogers, B.E., Anderson, C.J., Connett, J.M., 1996. Comparison
of bifunctional chelates for radiolabeling monoclonal
antibodies with copper radioisotopes: biodistribution and
metabolism. Bioconj. Chem. 7, 511–522.
Sola, G.A.R., Arguelles, M.G., Bottazzini, D.L., Furnari, J.C.,
Parada, I.G., Rojo, A., Ruiz, H.V., 2000. Lutetium-
177-EDTMP for bone pain palliation. Preparation, biodis-
tribution, and preclinical studies. Radiochim. Acta 88,
157–161.
Srivastava, S.C., Atkins, H.L., Krishnamurthy, G.T., Zanzi, I.,
Silberstein, E., Meinkein, G., Swailem, F., Mausner, L.F.,
D’Alessandro, T., Cabahug, C.J., Lau, Y., Park, T.,
Madajewicz, S., 1998. Treatment of metastatic bone pain
with tin 117 m stannic DTPA: a phase I/II clinical study.
Clin. Cancer Res. 4, 61–68.
Srivastava, S.C., Dadachova, E., 2001. Recent advances in
radionuclide therapy. Semin. Nucl. Med. 31, 330–341.
Stein, R., Govindan, S.V., Chen, S., Reed, L., Griffiths, G.L.,
Hansen, H.J., Goldenberg, D.M., 2001. Radioimmunother-
apy of a human lung cancer xenograft with monoclonal
antibody RS7: evaluation of
177
Lu and comparison of its
efficacy with that of
90
Y residualizing
131
I. J. Nucl. Med. 42,
967–974.
Unni, P.R., Pillai, M.R.A., 2002. Preparation and radio-
chemical separation of Rh-105 for radiopharmaceutical
applications. Radiochim. Acta 90, 363–369.
ARTICLE IN PRESS
M.R.A. Pillai et al. / Applied Radiation and Isotopes 59 (2003) 109–118 117

Venkatesh, M., Pandey, U., Dhami, P.S., Kannan, R.,
Achuthan, P.V., Chitnis, R.R., Gopalkrishnan, V., Banerjee,
S., Samuel, G., Pillai, M.R.A., Ramanujam, A., 2001.
Complexation studies with
90
Y from a novel
90
Sr-
90
Y
generator. Radiochim Acta 89, 413–417.
Volkert, W.A., Hoffman, T.J., 1999. Therapeutic radiophama-
ceuticals. Chem. Rev. 99, 2269–2292.
Volkert, W.A., Goeckeler, W.F., Ehrhardt, G.J., Ketring,
A.R., 1991. Therapeutic radionuclides: production and
decay property considerations. J. Nucl. Med. 32,
174–185.
Wessels, B.W., Rogus, R.D., 1984. Radionuclide selection and
model absorbed dose calculations for radiolabelled tumor
associated antibodies. Med. Phys. 11, 638–645.
ARTICLE IN PRESS
M.R.A. Pillai et al. / Applied Radiation and Isotopes 59 (2003) 109–118118