Note: Descriptions are shown in the official language in which they were submitted.
--1--
This invention relates to a method of preparing a
liquid for radio pharmaceutical application comprising a
radioisotope, and to an isotope generator suitable for
preparing said liquid. More particularly the invention
5 relates to a method of preparing a liquid for
radio pharmaceutical application comprising a
radioisotope by eluding from a parent isotope, which is
adsorbed on an adsorbent, a radioactive daughter isotope
by means of a physiological solution. The invention
lo also relates to an isotope generator system suitable for
conducting the above-described method as well as to a
reservoir for said generator system.
Radioisotopes having a half-life up to a few days
are used for diagnostic purposes in medicine. In order
to minimize damage to the tissues by radiation, it is
recommendable to use radioisotopes which emit only gamma
radiation. The radioisotope 99mTc is a pure gamma
radiator and has a comparatively short half-life.
Therefore, this isotope is excellently suitable for use
as a diagnostic, but also it may be used for
radioactively labeling other substances such as
proteins. The 99mTC isotope is generated by radioactive
decay of the parent isotope Moe. It is known, for
example from Netherlands Patent Application 7302304
(corresponding with l1.S. patent 3,970,583) to adsorb the
parent Isotope in the form of a molybdate on a suitable
adsorbent. and then to elude the daughter isotope 99mTc
by means of a physiological saline solution. An
apparatus suitable for producing a 99mTc-containing
liquid in this manner is an isotope generator as is also
described in the above-mentioned Netherlands Patent
Apply cation No. 7302304.
As a result of the fast development of radio-
diagnostics in the past ten years, there has come a need
US or a liquid for radio pharmaceutical application
comprising a radioisotope, which liquid has a sigher
--2--
concentration of radioactive material and a treater
chemical purity than the radio-diagnostics heretofore
used. The present 99mtechnetium equate is produced in
an isotope generator from natural or enriched molybdenum
irradiated in a nuclear reactor. The radioactive
isotope Moe is present in this product in a very small
concentration; the bulk consists of non-radioactive
molybdenum and serves as a carrier for Moe. The
dimensions of the column containing the adsorbent for
the parent isotope are restricted because a column that
is too large can not be eluded efficiently. This
applies in particular to the withdrawal of small elusion
volumes from the column, which are necessary for certain
purposes in which a higher isotope concentration is
required. Since restrictions are imposed upon the
dimensions of the column and upon the adsorption
capacity of the adsorbent, only comparatively little
parent isotope can be present in the generator, as a
result of which the required high concentration of
radioactivity in the equate cannot be obtained with
prior isotope generators.
Meanwhile, radioactive isotopes, for example,
radioactive molybdenum and curium, have been produced in
a different manner, namely by a fission reaction. For
I exclmple, molybdenum is produced by fission of 235U;
235u is irradiated in a nuclear reactor with neutrons,
after which the other fission products can be removed
from Moe by a chemical separation process. fission-
produced radioisotope is purified to an acceptable
degree of radionuclidic purity, but still contains
traces of contaminations such as 15Cd, 3 US, La,
156 89S 90Sr 95zr byway and astounds. In
addition to gamma radiation emitted by most of these
radioisotopes, these contaminations also emit
corpuscular radiation, namely alpha or beta radiation
These alpha or beta radiators are very undesirable in
so
--3
-
pharmaceutical compositions because thy can serious
attack the tissues; the strontium isotopes and
astounds are considered to be most toxic.
It has now been found that a liquid comprising a
radioisotope suitable for radio pharmaceutical
application can be produced in a high yield by elusion
of a fission-produced parent isotope adsorbed on a
suitable adsorbent, when the equate containing a
daughter isotope is purified by leans of a cation-
exchange material, preferably a cation-exchange resin.
Particularly suitable for this purpose are strongly-
acidic cation-exchange resins which are neutralized, and
have a particle size of, for example, 50-400 mesh,
preferably 100-200 mesh. As an example of a resin
suitable for this purpose may be mentioned Dower or Boo-
Red WOKS. These strongly-acidic resins are preferably
neutralized by treating them with an alkali metal base,
e.g., Noah, KOCH, or with NH40H, and then washing with
water. In this manner the resins are converted into the
No , K or NH4 form.
It is known from Into J. Apply Tad. Isotopes 1978,
*
Vol. 29, pp. 91-96, that the resin Dower WOKS in the
No form may be used for the separation of YO-YO from
Sr. In the reaction circumstances described in this
article, namely in the presence of a small quantity of
Era, the influence of the pi on the adsorption of 90Sr
was determined. From the results it appears that 90Sr
is adsorbed on Dower 50 resin at a pi of l.S-5.5, but
not at a pi of 7Ø The concentration of ETA had no
influence on the adsorption of 90Sr. These results give
rise to the supposition that Dower 50 resin is not
suitable to adsorb Sr from a solution suitable for
pharmaceutical application, namely an approx~nately
neutral physiological saline solution. However, quite
contrary to expectations it has been found that a cation-
exchange resin, in particular a strongly-acidic cation-
* Trade Mark
.
-4-
exchange resin such as Dixie or Byrd SUE converted
into the No , K or I~H4 form is particularly suitable
to purify 9mTc produced from fission-produced Moe, so
that a solution containing 99mTc and suitable for
radio pharmaceutical application is obtained with an
exceptionally high chemical, radio chemical and
radionuclidic purity.
From the above-mentioned Netherlands patent
application 7302304 it is known that aluminum oxide
which contains fully or partly hydrated manganese
dioxide is an adsorption agent for the parent isotope
Moe. It has been found that this material also is
excellently suitable as an adsorbent for the entirely,
or substantially, molybdenum carrier-free, fission-
produced Moe. This is not obvious as such, because the
latter concerns extremely small quantities of adsorbed
molybdenum which moreover contains undesired
contaminants. The desired optimum elusion yield
strongly depends on the nature and quantity of the
material to be eluded and the adsorbed material present,
and it is generally known that small differences in
these respects can easily disturb this subtle
equilibrium, as a result of which either a less optimum
yield, or an undesired elusion pattern could be
I obtained.
From the above it will be clear that the method
according to the invention will preferably be used in an
isotope generator system. An isotope generator system
is to be understood to mean the actual isotope generator
provided with a connection to a reservoir with fluent
and with an equate conduit, and enclosed by a generator
housing. Such a system is sometimes termed "cow". Roy
invention therefore also relates to a generator system
the isotope generator of which comprises a reservoir
hiving a supply facility for the fluent and an outlet
facility for the equate, and in which the adsorption
I . I
* trade Marc
I
--5--
agent for the parent isotope is present. Such a
generator is known, for example, from the above-
mentioned Netherlands Patent Application 730230~.
However, the generator according to the invention
comprises a fission-produced radioisotope and a cation-
exchange material. Because the fission-produced
radioisotope is fully or substantially carrier-freel a
small quantity of adsorbent for the parent isotope is
amply sufficient. As a result of this the dimensions of
the generator system can be greatly reduced, so that the
apparatus is easier to handle, both in use (in hospital
or clinical laboratory the generator system must
regularly be changed), and upon assembly by the
manufacturer. It is of great advantage that the cation-
exchange material is also present in the generator system according to the invention. As a result of this,
the equate can be purified in the generator itself so
that the liquid withdrawn from the generator and
comprising radioactive daughter isotope has a high
chemical and radionuclidic purity, hence is suitable for
radio pharmaceutical application. Purification
aEterwar~s of the equate, that is to say after it has
left eye generator, is superfluous Such a purification
afterwards generally is even impossible or at least
I undesired, because the daughter isotope obtained usually
ha owe short a halE-life to be able to stand such an
after-treatment, and also because an after-treatment in
a hospital or clinical laboratory where auxiliary means
suitable for the purpose are lacking, is out of the
question for reasons of safety.
It is usual to enclose the adsorbent for the parent
Isotope in the reservoir of the generator system between
two filters. In order to load the adsorbent with the
radioactive parent isotope, a solution of this isotope
US is admitted to one side of the reservoir. Glossily or
glass beads are frequently used on this side as a
I
.
--6--
filtering material. However, glass beads cause
channeling in the adsorbent and hence inefficient
loading and a non-uniform distribution of the parent
isotope over the adsorbent. Glossily often impedes the
loading due to too large a resistance and in addition it
tends, as also synthetic resins, for example,
polyethylene, to adsorb a little parent isotope. This
latter is very objectionable because upon elusion of the
generator, the quantity of parent isotope not adsorbed
on the adsorption agent will contaminate the equate.
As a particular aspect of the invention it has now
been found that the above-mentioned disadvantages can be
removed by the filter on that side of the generator
reservoir where the solution of the parent isotope is
admitted being made of sistered glass. It has been
found that when such a filter is used which, of course,
can also be used in the prior art isotope generators, an
efficient and homogeneous loading of the adsorbent can
very easily be achieved, while no parent isotope is
adsorbed by the filter.
The generator system according to the invention is
preferably constructed so that both the cation exchange
resin and the adsorbent for the parent isotope are
present in the same reservoir. In this embodiment in
which the dimensions of the generator can be minimized
and an optimum purity of the radio pharmaceutical
composition can be reached, the above-mentioned
advantages stand out even better, while the cost of
production can also be kept as low as possible.
On a further preferred embodiment the reservoir
containing both the cation-exchange resin and the
adsorbent for the parent isotope is divided into two
compartments which are separated from each other by a
filter the circumference of which adjoins the inner wall
of the reservoir. one compartment of the reservoir
comprises a supply facility for the fluent and the
so
-7- !
adsorbent for the parent isotope is present between
supply facility and separation filter, the adsorbent
being enclosed between the above-men~ioned sistered
- glass filter and the separation filter. The other
compartment of the reservoir comprises an outlet
facility for the equate. The cation-exchange material
is present between separation filter and outlet
facility, and the space between the adsorbent particles
and between the particles of the ion-exchanger being
filled with a physiological solution. A separation
filter suitable for this purpose consists of two filter
disks covering each other entirely, or substantially
entirely, the disk adjoining the adsorption agent
consisting of glass fire paper, for example, a
Millipore refilter A 200, the disk adjoining the ion
exchanger consisting of porous polyethylene.
Finally the invention relates to a reservoir for
the above-mentioned generator, which reservoir contains
both the cation-exchange material and the adsorbent for
I the parent isotope. It has been found that such a
reservoir that is loaded and sterilized can be stored in
an uncooked condition for more than 3 months and can be
incorporated in a venerator system at any desired moment
during this period without any pretreatment The
reservoir can then be used to provide an equate
containing radioactive daughter isotope in a high yield.
This is of advantage because the reservoirs can be
manufactured in stock and be shipped to the supplier of
generator systems who can at any desired moment use a
reservoir for his generator system without any pro-
treatment; this means a considerable saving of costs
The invention will be described in greater detail
with reference to the following specific example.
The drawing is a cross-sectional view of a
favorable embodiment ox the reservoir of the isotope
generator according to the invention A substantially
Jo ,,,
- * Trade Mark
58~ ,
Cal in~ric~l reservoir ( 4 ox a suitable infuriate Illat~rial,
for example glass or a polymeric material], preferably
borosilicate glass, is widened at each end and provided
with a flange portion (10,13). The openings at the t
ends of the reservoir are closed by means of rubber
stoppers (2, 14) comprising a flange portion (11, 15)
and a jacket portion (12, 16~; the flange portion of the
stopper engages the flange portion of the reservoir, the
jacket portion fitting in the opening of the reservoir.
The flange portions of stopper and reservoir are
connected together by means of a metal cap, for example
and alumina folded cap (1, 17). The reservoir contains
a slurry of the adsorbent (5) in a solution of 0.9% Nail
in water. This adsorbent consists of AYE particles
which are covered entirely or partly with a layer of
fully or partially hydrated manganese dioxide. In the
reservoir, the adsorbent is enclosed between a filter of
sistered glass (3) of an average porosity and a filter
disk of glass fire paper (6), namely a millipore pro-
filter A 200. The reservoir furthermore contains a
slurry of the resin*Bio-Rad WOKS in the No form (8)
in a solution of 0.9% luckily in water. The resin has been
converted into the Nay form by a treatment with Noah
succeeded by a washing with water. This resin is
enclosed between a filter disc (7) of porous
polyethylene engaging filter disk (6) and a filter disk
I likewise of porous polyethylene, supported by a
polycarbonate spacer ring I ¦
Example 1
Ten of the above-described reservoirs were stored
for 3 months and then wised for the following experiment.
Each reservoir was loaded with fission-produced
molybdenum-carrier-free Moe in the form of sodium
molybdate (pi 1-5-10), by perforating the stoppers at
the ends of the reservoir, so that an inlet and an
* Trade Mark
~85~
g
outlet aperture were obtained, and then causing a
solution of the radioactive sodium molybdate to flow
into the reservoir through the inlet aperture (at A).
After washing and sterilizing in an autoclave at 121C
for 30 minutes, the isotope generator thus obtained was
placed in a generator system. The radioactivity of the
isotope generator was 1,000 mCi.
When using the generator the fluent was supplied
through the inlet aperture at one end of the reservoir
(at A), while the emulate was drained through the outlet
aperture at the opposite end of the reservoir. The
generators were eluded with sterile isotonic saline
solutions (0.9% wt/vol% Nail in water) in quantities of
4.6 or 15 ml, the average elusion yields recorded in the
table below being obtained.
!
TABLE I
Properties of 4.6 ml and 15 ml 99mTc-containing equates.
Elusion volume: 4.6 ml Elusion violin: 15 ml
Louisiana Average Louisiana Elusion Average elusion
yield -(%) yield (%)
_.
89~4 1 9().8
89.9 2 94.1
3 8~.1 3 93.9
88.6 4 92.4
89.6 5 93.3
6 89.6 6 ~34.2
7 87.2 7 9~).9
8 89.3 8 93.1
9 89.6 9 92.7
I 10 88.5 10 91.9
~S89~3
,--
-pa-
(continuation table)
Analyses A aliases:
phi 6 phi
radiochem. purity 99% radiochem. purity > 99%
My ; < Owe gel My : < Owe g/ml
Al : C 0.5~kg/ml Al : < 0.5 Amelia
labeling efficiency: labeling efficiency:
-EHDP 97.8% -EHDP 98.4%
-sb2 So 97-5% sb2 3
radionuclidic purity: radionuclidic Purity:
- My C 1 nCijmCi To - Moe 2 nCi/mCi 99mTc
_131I < 1 nCi/mCi 99mTc _131I < 1 nCi/mCi 9 To
-Roy 1 nCi/mCi 99mTc -103R~ 1 nci/mci 99 To
sterility: sterile sterility: sterile
lo apyrogenicity: pyrogen-free apyrogenicity: ~yrogen-free
~589~3
, Jo
--10--
The elusion yields recorded in the above table denote
the average percentage over the ten generators of the
theoretically available 99mTc activity per elusion, the
- generators being eluded ten times each on successive
days. So a high elusion yield was reached, namely on an
average 89.06% + 2.15 and 92.17% + 1.64 of the
theoretically available 99mTC activity upon elusion with
4.6 and 15 ml of fluent, respectively.
Also recorded in the above table are the average
results of the analysis of the equates. The
radio chemical purity was determined by means of a paper
chromatographic method; the concentration of My++ and
Al ++ was determined by means of a spectrophotometric
method and a calorimetric (as quinalizarine complex)
method respectively. The labeling efficiency of
ethylene hydroxy diphosphonate (EHDP) and Sb2S3 proves
that the resulting 99mTc is excellently suitable for
preparing 99mTc-labelled compounds and may hence be used
for all desired applications. The radionuclidity purity
was determined by means of a gamma analyzer. At most,
traces of the radioisotopes Moe, 131I and ROY could
be detected; other radionuclidic contaminations were not
found in the equate.
It will be obvious from the results shown that the
I long Syria of the willed reservoirs had had no
detrimental influence whatsoever on the elusion yield
and on the purity ox the equate.
In another experiment the same isotope generator
could be eluded 15 times with 20 ml of a physiological
saline solution without the elusion yield changing or
the resulting 99mtechnetium equate being contaminated
with cat ionic contaminants.
So
- 11 -
Example 2
Ten additional columns substantially similar to
those employed in Example 1 but with less dead volume
were used to generate 99mTc. The reduced dead volume
is achieved by dispensing with the spacer ring and
utilizing a cylindrical, non-flanged reservoir in
constructing the column. The results are shown in the
following tables:
--12--
TABLE I I
Elusion Yield
. . . __ ___
N ED to N O ED I
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-13-
TABLE III
Chemical purity of equate
. _ _ _ . I
sty equates Thea equates
gun no. _ .
_ pi go Mn/ml go Al/ml pi go Mn/ml go Al/ml
1 5.8<0.3 < l 5.80.55 < l
3 5.9 5.9<0.30
4 5.8 5.8<0.30
5.9 6.0< 0.30
6 5.8 5.90.35
7 5.8 5.8<0.30
9 6.0 5.90.78
5.8 5.8< 0.30 ,
Radio chemical purity - sty equates >99%
Lulling properties - sty equates Sb2S3 killed
labeled with 2 ml from 4.6 ml equate 96.0%
4 ml from I ml equate 93.9%
2 ml from 15 ml equate 100.0%
ml from 15 ml equate 99.8%
Jo
58~8
- -
TABLE IV
Mooney equate
Preliminary test detector: NaI/single channel analyzer
99Mo-breakthrough (nCi mums 99mTc)
sty end 3rd Thea Thea Thea Thea Thea Thea Thea
gun no. of. of. of. of. of. of. of. of. of. of.
- _ _ ___
1 1.78 1.47 1.72 0.91 1.47 0.78 0.31 1.24 0.90 1.88
__ _ . ._. _ ..._ .
3 1.80 1.70 2.00 1.52 1.23 1.41 2.01 2001 0.34 1.76
. _ _.... __.
4 3.21 2.83 2.31 2.19 2.43 1.86 2.37 3.35 1.03 2.35
. _ _ _ ,__ .
1.05 1.26 1.19 0.87 1.00 0.92 1.06 0.62 0.89 0.63
_ _ . . _ _ _,
6 5.40 3.91 3.47 3.01 3.08 2.37 2.68 1.53 2.66 1.73
__ _ _ . _ _ I _____ _ . _ . . _~_
7 2.90 2.37 3.09 2.12 2.30 1.47 1.74 1.89 1.31 0.99
__ _ __ _ __ _ __, . .. _ .
9 2.50 2.32 1.84 1.59 1.51 1.60 1.81 1.42 0.94 2.1~
._ _ _ . __ . ------- --- ----- - -I
1.82 1.~2 1 61 1.33 1.15 1.25 1.66 1.3~ 1.03 1.53
- I 'I
/ ,~.,
-15~ 35
TABLE V
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