GB2067343A - Generation of radio-isotopes - Google Patents

Abstract

A method of preparing a liquid containing radioisotope for radiopharmaceutical application comprises eluting a radioactive daughter isotope from a parent isotope which is adsorbed on an adsorbent by means of a physiological solution, wherein a fission- produced parent isotope is used and the eluate comprising a daughter isotope is purified with a cation-exchange material. The method produces a liquid containing a radioisotope having an exceptionally high chemical, radiochemical and radionuclidic purity in a high yield. An isotope generator system suitable for use of the said method and a reservoir for said generator system is also described and claimed.

Description

SPECIFICATION Generation of radioisotopes This invention relates to a method of preparing a liquid for radiopharmaceutical application comprising a radioisotope, and to an isotope generator suitable for preparing said liquid. More particularly the invention relates to a method of preparing a liquid for radiopharmaceutical application comprising a radioisotope by eluting from a parent isotope, which is adsorbed on an adsorbent, a radioactive daughter isotope by means of a physiological solution. The invention 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 labelling other substances such as proteins. The 99mTc isotope is generated by radioactive decay of the parent isotope 99Mo. It is known, for example from Netherlands Patent Application 7302304 (corresponding with U. S. patent 3,970,583) to adsorb the parent isotope in the form of a molybdate on a suitable adsorbent and then to elute 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 Application No.7302304.

As a result of the fast development of radiodiagnostics in the past ten years, there has come a need for a liquid for radiopharmaceutical application comprising a radioisotope, which liquid has a higher concentration of radioactive material and a greater chemical purity than the radio-diagnostics heretofore used. The present 99mtechnetium eluate is produced in an isotope generator from natural or enriched molybdenum irradiated in a nuclear reactor. The radioactive isotope 99Mo is present in this product in a very small concentration; the bulk consists of non-radioactive molybdenum and serves as a carrier for 99Mo. The dimensions of the column containing the adsorbent for the parent isotope are restricted because a column that is too large can not be eluted efficiently.This applies in particular to the withdrawal of small elution volumes from the column, which are necessary for certain purposes in which a higher isotope concentration is required. Since restrictions are imposed jupon 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 eluate cannot be obtained with prior isotope generators.

Meanwhile, radioactive isotopes, for example, radioactive molybdenum and cerium, have been produced in a different manner, namely by a fission reaction. For example, 99molybdenum 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 99Mo by a chemical separation process. Afission-produced radioisotope is purified to an acceptable degree of radionuclidic purity, but still contains traces of contaminations such as 115Cd,136Cs, 140La, 156Eu, 89Sr, 90Sr, sZr, 140Ba and actinides. 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 pharmaceutical compositions because they can seriously attack the tissues; the strontium isotopes and actinides are considered to be most toxic.

It has now been found that a liquid comprising a radioisotope suitable for radiopharmaceutical application can be produced in a high yield by elution of a fission-produced parent isotope adsorbed on a suitable adsorbent, when the eluate containing a daughter isotope is purified by means 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 Dowex or Bio-Rad 50W-X8. These strongly-acidic resins are preferably neutralized by treating them with an alkali metal base, e.g., NaOH, KOH, or with NH40H, and then washing with water. In this manner the resins are converted into the Na+, K+ or NH4+ form.

It is known from Int J. Appl. Rad. Isotopes 1978, Viol.29, pp. 91-96, that the resin Dowex 50W-X8 in the Na+ form may be used for the separation of 90Y from 90Sr. In the reaction circumstances described in this article, namely in the presence of a small quantity of EDTA, the influence of the pH on the adsorption of 90Sr was determined. From the results it appears that 90Sr is adsorbed on Dowex 50 resin at a pH of 1.5-5.5, but not at a pH of 7.0. The concentration of EDTA had no influence on the adsorption of 90Sr. These results give rise to the supposition that Dowex 50 resin is not suitable to adsorb 90Sr from a solution suitable for pharmaceutical application, namely an approximately neutral physiological saline solution.However, quite contrary to expectations it has been found that a cation-exchange resin, in particular a strongly-acidic cation-exchange resin such as Dowex or Bio-Rad 50W-X8 converted into the Na+, K+ or NH4+ form is particularly suitable to purify 99mTc produced from fission-produced 99Mo, so that a solution containing 99mTc and suitable for radiopharmaceutical application is obtained with an exceptionaliy high chemical, radiochemical and radionuclidic purity.

From the above-mentioned Netherlands patent application 7302304 it is known that aluminium oxide which contains fully or partly hydrated manganese dioxide is an adsorption agent for the parent isotope 99Mo. It has been found that this material also is excellently suitable as an adsorbentforthe entirely, or substantially, molybdenum carrier-free, fission-produced 99Mo. This is not obvious as such, because the latter concerns extremely small quantities of adsorbed molybdenum which moreover contains undesired contaminants.The desired optimum elution yield strongly depends on the nature and quantity of the material to be eluted 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 elution pattern, could be obtained.

From the above it will be clearthatthe 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 eluent and with an eluate conduit, and enclosed by a generator housing. Such a system is sometimes termed "cow". The invention therefore also relates to a generator system the isotope generator of which comprises a reservoir having a supply facility for the eluent and an outlet facility for the eluate, and in which the adsorption agent for the parent isotope is present. Such a generator is known, for example, from the above-mentioned Netherlands Patent Application 7302304.

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-free, 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 eluate 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 radiopharmaceutical application. Purification afterwards of the eluate, that is to say after it has left the generator is superfluous. Such a purification afterwards generally is even impossible or at least undesired, because the daughter isotope obtained usually has too short a half-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 is admitted to one side of the reservoir. Glasswool or glass beads are frequently used on this side as a 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. Glasswool 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 elution of the generator, the quantity of parent isotope not adsorbed on the adsorption agent will contaminate the eluate.

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 sintered 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 homogenous 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 cationexchange 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 radiopharmaceutical 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.

In 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 eluent and the adsorbent for the parent isotope is present between supply facility and separation filter, the adsorbent being enclosed between the above-mentioned sintered glass filter and the separation filter. The other compartment of the reservoir comprises an outlet facility for the eluate. The cation-exchange material is present between separation filter and outlet facility, 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 fibre paper, for example, a millipore pre-filterAP 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 the parent isotope. It has been found that such a reservoir that is loaded and sterilized can be stored in an uncooled condition for more than 3 months and can be incorporated in a generator system at any desired moment during this period without any pre-treatment.

The reservoir can then be used to provide an eluate 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 system who can at any desired moment use a reservoir for his generator system without any pretreatment; 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 cross-sectional view of a favorable embodiment of the reservoir of the isotope generator according to the invention. A substantially cylindrical reservoir (4) of a suitable inert material, 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 two 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 aluminum folded cap (1, 17). The reservoir contains a slurry of the adsorbent (5) in a solution of 0.9% NaCI in water. This adsorbent consists of Al203 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 sintered glass (3) of an average porosity and a filter disk of glass fibre paper (6), namely a millipore prefilter AP 200.

The reservoir furthermore contains a slurry of the resin Bio-Rad 50W-X8 in the Na+ form (8) in a solution of 0.9% NaCI in water. The resin has been converted into the Na+ form by a treatment with NaOH succeeded by a washing with water. This resin is enclosed between a filter disk (7) of porous polyethylene engaging filter disk (6) and a filter disk (18), likewise of porous polyethylene, supported by a polycarbonate spacer ring (9).

Example 1 Ten of the above-described reservoirs were stored for 3 months and then used for the following experiment.

Each reservoir was loaded with fission-produced molybdenum-carrier4ree 95Mo in the form of sodium molybdate (pH 1-5-10) by perforating the stoppers at the ends of the reservoir, so that an inlet and an 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 autociave at 121"C 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 eluent was supplied through the inlet aperture at one end of the reservoir (at A), while the eulate was drained through the outlet aperture at the opposite end of the reservoir. The generators were eluted with sterile isotonic saline solutions (0.9% wt/vol% NaCI in water) in quantities of 4.6 or 15 ml, the average elution yields recorded in the table below being obtained.

TABLE 1 Properties of 4.6 ml and 15 ml 99mTc-containing eluates Elution voiume: 4.6 ml Elution Volume: 15 ml Elution Average elution Elution Average elution yield (%) yield (%) 1 89.4 1 90.8 2 89.9 2 94.1 3 89.1 3 93.9 4 88.6 4 92.4 5 89.6 5 93.3 6 89.6 6 94.2 7 87.2 7 90.9 8 89.3 8 93.1 9 89.6 9 92.7 10 88.5 10 91.9 fcontinuation Table) Analyses: Analyses: pH: 6 pH:6 radiochem. purity > 99% radiochem. purity > 99% Mn++: < 0.3 0.31lg/ml Mn++: < 0.3 Rg/ml Al+++: < 0.5 g/ml Al+++:< 0.5 g/ml labelling efficiency: labelling efficiency: -EHDP 97.8% -EHDP 98.4% -Sb2 S3 97.5% -Sb2 S3 97.0% radionuclidic purity: radionuclidic purity: -99Mo < 1 nCi/mCi 99mTc -99Mo < 2 nCi/mCi 99mTc -131| < 1 nCi/mCi 99mTc -131| < 1 nCi/mCi 99mTc -103Ru < 1 nCi/mCi 99mTc -103Ru < 1 nCi/mCi 99mTc sterility: sterile sterility: sterile apyrogenicity: pyrogen-free apyrogenicity: pyrogen-free The elution yields recorded in the above table denote the average percentage over the ten generators of the theoretically available 99mTc activity per elution, the generators being eluted ten times each on successive days.So a high elution yield was reached, namely on an average 89.06% + 2.15 and 92.17% + 1.64 of the theoretically available 99mTc activity upon elution with 4.6 and 15 ml of eluent, respectively.

Also recorded in the above table are the average results of the analysis of the eluates. The radiochemical purity was determined by means of a paper chromatographic method; the concentration of Mn++ and Al+++ was determined by means of a spectrophotometric method and a colorimetric (as quinalizarine complex) method, respectively. The labelling 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 99Mo, 1311 and 103Ru could be detected; other radionuclidic contaminations were not found in the eluate.

It will be obvious from the results shown that the long storage of the filled reservoirs had had no detrimental influence whatsoever on the elution yield and on the purity of the eluate.

In another experiment the same isotope generator could be eluted 15 times with 20 ml of a physiological saline solution without the elution yield changing or the resulting 99mtechnetium eluate being contaminated with cationic contaminants.

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: Table II Elution yield Elution yields (% of available 99mTc activity) gen. mCi 99Mo elution 1st 2nd 3rd 4th 5th 6th 7th 8th 9th 10th no. loaded volume el. el. el. el. el. el. el. el. el. el.

1 768 15 ml 92.36 92.03 92.41 94.25 93.41 89.66 87.48 92.61 90.54 89.52 3 1161 4.6ml 90.45 91.09 89.82 89.61 91.14 88.14 87.17 90.70 89.32 88.96 4 1103 4.6ml 90.25 91.15 89.65 89.09 91.30 88.45 85.69 94.15 88.62 88.59 5 1833 4.6ml 92.02 91.03 90.19 89.76 88.99 90.74 86.24 91.93 89.06 87.12 6 1089 15 ml 92.50 93.27 91.82 92.10 91.47 89.41 88.05 94.18 90.46 89.40 7 1826 4.6ml 90.44 89.91 90.21 90.15 89.91 91.69 86.36 93.22 89.21 87.49 9 1810 15 ml 92.24 93.50 91.95 92.41 91.93 92.34 88.81 93.68 91.42 88.86 10 1815 4.6ml 91.52 90.89 90.45 89.91 91.40 89.80 86.63 92.01 89.98 88.01 TABLE Ill Chemical purity of eluate 1st eluates 10th eluates gen. no.

pH yg Mn/ml yg Al/ml pH g Mn/ml Cig Al/ml 1 5.8 < 0.3 < 1 5.8 0.55 < 1 3 5.9 5.9 < 0.30 4 5.8 5.8 < 0.30 5 5.9 6.0 < 0.30 6 5.8 5.9 0.35 7 5.8 5.8 < 0.30 9 6.0 5.9 0.78 10 5.8 5.8 < 0.30 Radiochemical purity - 1st eluates > 99% Labelling properties - 1st eluates Sb2S3 colloid labelled with 2 ml from 4.6 ml eluate 96.0% 4 ml from 4.6 ml eluate 93.9% 2mlfrom15 mleluate 100.0% 4mlfrom15 ml eluate 99.8% TABLE IV 99Mo4n eluate Preliminary test detector: Nal/single channel analyzer 99Mo-breakthrough (nCi 99Mo/mCi 99mTc) gen. no. 1st 2nd 3rd 4th 5th 6th 7th 8th 9th 10th el. el. el. el. el. el. el. el. el. el.

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 2.01 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 5 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 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 2.18 10 1.82 1.42 1.61 1.33 1.15 1.25 1.66 1.35 1.03 1.53 TABLE V Radionuclidic purity of eluate Definitive test detector: Ge(Li)/Nuclear Data System 4410 analyzer and ND 812 minicomputer 1st eluates 10th eluates gen. no 131 InCi/mCi Tc 103Ru nCi/mCi Tc 99Mo nCi/mCi Tc 131InCi/mCi Tc 103Ru nCi/mCi Tc 99Mo nCi/mCi Tc 1 not detectable not detectable not detectable 3 " " " 4 " " " 5 not detectable not detectable 0.29 " " " 6 " " 3.70 " " " 7 " " 1.71 " " " 9 0.005 " 1.27 " " " 10 0.004 " not detectable " " " Sterility - sterile Apyrogenicity - pyrogen-free

Claims (20)

1. A method of preparing a liquid for a radiopharmaceutical application comprising a radioisotope, by eluting from a parent isotope, which is adsorbed on an adsorbent, a radioactive daughter isotope by means of a physiological solution, wherein a fission-produced parent isotope is used and the eluate containing a daughter isotope is purified with a cation-exchange material.

2. A method as claimed in claim 1, wherein molybdenum-99 is used as a parent isotope.

3. A method as claimed in claim 1 or 2, wherein aluminium oxide which contains fully or partly hydrated manganese dioxide is used as an adsorbent for the parent isotope.

4. A method as claimed in any one of the preceding claims, wherein a cation-exchange resin is used as a cation-exchange material.

5. A method as claimed in claim 4, wherein a strongly acid cation-exchange resin which has been neutralised is used as a resin.

6. A method as claimed in claim 5, wherein a strongly acid cation-exchange resin which has been converted into a Na+, K+ or NH4+ form is used as a resin.

7. A method as claimed in any one of the claims 4 to 6, wherein a cation-exchange resin is used having a particle size of 50-400 mesh.

8. A method as claimed in claim 7, wherein the resin particle size is 100 -200 mesh.

9. An isotope generator system suitable for use of the method as claimed in any one of the preceding claims, which generator system comprises a reservoir which has a supply facility for the eluent and an outlet facility for the eluate, and in which the adsorption agent for the parent isotope is present, and wherein the parent isotope is a fission-produced isotope which is fully or substantially carrier-free and that a cation-exchange material is present in the generator system.

10. A generator system as claimed in claim 9, in which the adsorbent for the parent isotope in the reservoir is enclosed between two filters and on that side of the reservoir where during the loading of the adsorbent the solution of the parent isotope is admitted, the filter consists of sintered glass.

11. A generator system as claimed in claim 9 or 10, wherein the cation-exchange material is present in the same reservoir as the eluent for the parent isotope.

12. A generator system as claimed in claim 11, wherein the reservoir 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, in which in one compartment which comprises a supply facility for the eluentthe adsorbent for the parent isotope is present between supply facility and separation filter, the adsorbent being enclosed between a sintered glass filter and the separation filter, and in which in the other compartment, which comprises an outlet facility for the eluate, the cation-exchange material is present between separation filter and outlet facility, the space between the adsorbent particles and between the particles of the ion exchanger being filled with a physiological solution.

13. A generator system as claimed in claim 12, wherein the separation filter consists of two disks covering each other entirely or substantially entirely, the disk adjoining the adsorbent consisting of glass fibre paper and the disk adjoining the ion exchanger consisting of porous polyethylene.

14. A reservoir for a generator system as claimed in claim 10, which reservoir comprises two filters between which the adsorbent for the parent isotope is enclosed, and wherein, on that side of the reservoir where during the loading of the adsorbent the solution of the parent isotope is admitted, the filter consists of sintered glass.

15. A reservoir for a generator system as claimed in claim 11, wherein the reservoir comprises both an adsorbent for the parent isotope and a cation-exchange material.

16. A reservoir for a generator system as claimed in claim 12 or 13, wherein the reservoir 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, in which in one compartment which comprises a supply facility for the eluent, the adsorbent for the parent isotope is present between supply facility and separation filter, the adsorbent being enclosed between a sintered glass filter and the separation filter, and in which in the other compartment, which comprises an outlet facility for the eluate, the cation-exchange material is present between separation filter and outlet facility, the space between the adsorbent particles and between the particles of the ion exchanger being filled with a physiological solution.

17. A reservoir for a radio isotope generator, which reservoir comprises two filters between which the adsorbent for the parent isotope is enclosed, and wherein on that side of the reservoir where during the loading of the adsorbent the solution of the parent isotope is admitted, the filter consists of sintered glass.

18. A method of preparing a liquid for a radiopharmaceutical application according to claim 1 substantially as described herein with reference to the examples.

19. An isotope generator system according to claim 9 substantially as described herein.

20. A reservoir for a radio isotope generator substantially as described herein with reference to the accompanying drawing.