JPH0119102B2 - - Google Patents

Description

[Detailed description of the invention]

The present invention relates to a method for preparing a liquid for radiopharmaceutical applications comprising a radioactive isotope, and to an isotope generator suitable for preparing the liquid. More particularly, the present invention is for radiopharmaceutical applications comprising a radioactive isotope obtained by elution using a physiological solution from a parent isotope adsorbed on an adsorbent and a radioactive daughter isotope. The present invention relates to a method for preparing a liquid. Furthermore, the invention relates to an isotope generator suitable for carrying out the method, as well as a storage tank for the generator. Radioactive isotopes with a half-life of up to 2-3 days are used for diagnostic purposes in clinical practice. The use of radioisotopes that emit only gamma rays is recommended to minimize tissue damage from irradiation. The radioactive isotope ^99m Tc is a pure gamma emitter and has a relatively short half-life.
This isotope is therefore not only very suitable for diagnostic purposes, but can also be used to radioactively label other substances, such as proteins. ^{The 99m} Tc isotope is generated by the radioactive decay of the parent isotope ⁹⁹ Mo. For example, according to Dutch Patent Application No. 7302304 (corresponding to U.S. Pat. No. 3970583), the parent isotope is adsorbed in the form of a molybdate on a suitable adsorbent, and then the daughter isotope is removed by physiological saline. It is known to elute ^99m Tc. A suitable device for producing ^99m Tc-containing liquids in this manner is an isotope generator, which is also described in the aforementioned Dutch Patent Application No. 7302304. Rapid developments in diagnostic radiology over the past decade have resulted in the need for fluids for radiopharmaceutical applications comprising radioactive isotopes. This liquid has a higher concentration and chemical purity than those used in conventional diagnostic radiology. ^{The 99m} technetium eluate of the present invention is produced in an isotope generator from molybdenum enriched by natural or irradiation within a nuclear reactor. The radioactive isotope ⁹⁹ Mo is present in this product in very low concentrations, consisting mostly of non-radioactive molybdenum and serving as a carrier for ^{the 99} Mo. The size of the column containing the adsorbent for the parent isotope is limited because a column that is too large cannot elute efficiently. This is particularly true for certain purposes, namely the withdrawal of water elution volume from the column where higher isotope concentrations are required. These limitations are imposed on the dimensions of the column and the adsorption capacity of the adsorbent, so that only a relatively small amount of the parent isotope can be present in the generator, resulting in less of the required eluate. High concentrations of radioactivity cannot be obtained with prior art isotope generators. On the other hand, radioactive isotopes, such as radioactive molybdenum and cerium, have been produced in a different way, namely by nuclear fission reactions. For example, ⁹⁹ molybdenum is produced by fission of ²³⁵ U, ²³⁵
The U is irradiated with neutrons in the reactor, and other fission products can then be removed from ^{the 99} Mo by chemical separation methods. Although the fission-produced radioisotopes are purified to an acceptable degree of radionuclide purity, they still contain ¹¹⁵ Cd, ¹³⁶ Cs, ¹⁴
It has traces of contaminants such as ⁰ La, ¹⁵⁶ Eu, ⁸⁹ Sr, ⁹⁰ Sr, ⁹⁵ Zr, ¹⁴⁰ Ba and actinides. In addition to the gamma radiation emitted by most of these radioactive isotopes, these contaminants also carry out particle radiation, ie alpha or beta radiation. These alpha or beta emitting substances are highly undesirable in pharmaceutical compositions. Strontium isotopes and actinides are thought to be the most toxic because they can violently attack tissues. Now, if the eluate containing the daughter isotope is purified by a cation exchange material, preferably a cation exchange resin, a liquid comprising a radioactive isotope suitable for radiopharmaceutical applications can be obtained on a suitable adsorbent. It has been found that it can be produced in high yields by elution of the adsorbed fissile parent isotope. Particularly suitable for this purpose are strongly acidic cation exchange resins,
This is neutralized and has a particle size of e.g.
It has 400 meshes, preferably 100-200 meshes. Examples of resins suitable for this purpose include "Dowex" (trademark) and "Bio-Rad 50W-X8" (trademark). These strongly acidic resins are preferably treated with alkali metal bases such as NaOH, KOH or
Neutralized by treatment with NH ₄ OH and then washed with water. This treatment converts the resins into Na ⁺ , K ⁺ or NH ₄ ⁺ forms. Na ⁺ shaped resin “DOWEX 50W-X8”
What can be used to separate ^90Y from ^90Sr is “Int.J.Appl.Rad.Isotopes” 1978, Vol.29, 91.
Known by ~96 pages. The influence of PH on the adsorption of ⁹⁰ Sr was determined under the reaction conditions described in this paper, i.e. in the presence of small amounts of EDTA. From these results, ⁹⁰ Sr is not PH7.0, but ``Dowex 50'' at PH1.5 to 5.5.
It appears to be adsorbed onto the resin. The concentration of EDTA had no effect on the adsorption of ⁹⁰ Sr. These results show that "Dowex 50" resin
This gives rise to the presumption that adsorption of ⁹⁰ Sr from solutions suitable for pharmaceutical applications, ie approximately neutral physiological salt solutions, is not suitable. However, quite contrary to this expectation, cation exchange resins, especially strongly acidic cation exchange resins, such as Na ⁺ , K ⁺
or “Dowex” converted to _NH4 ⁺ form
or "Bio-Rad 50W-X8" is particularly suitable for purifying ^99m Tc produced from fission product ⁹⁰ Mo, resulting in a solution containing ^99m Tc and particularly suitable for radiopharmaceutical applications. It has been found that it can be obtained with a high degree of chemical, radiochemical and radionuclide purity. According to the above-mentioned Dutch patent application ^no.
It is known to be an adsorbent for Mo. Furthermore, it has been found that this material is highly suitable as an adsorbent for the fission product ⁹⁹ Mo, which is completely and substantially free of molybdenum supports. This is not obvious on its own. This is because the latter involves very small amounts of adsorbed molybdenum, which also contains undesirable contaminants. The desired optimal elution yield is highly dependent on the nature and quality of the material to be eluted, as well as on the adsorbent material present, and small differences in these respects can easily disturb this delicate equilibrium;
It is generally known that this has resulted in lower optimal yields or undesirable elution patterns. From the above it will be clear that the method according to the invention may be successfully used in an isotope generator. It is to be understood that an isotope generator means a virtual isotope generator, provided with means of connection to a reservoir with an eluent and an eluate conduit, and enclosed by a housing of the generator. Such devices are sometimes referred to as "cows." The invention therefore also relates to a generator device, which device is a reservoir with an eluent supply facility and an outlet facility for the eluate, in which an adsorbent for the parent isotope is present. It is made up of things. Such a generator is known, for example, from the aforementioned Dutch patent application No. 7302304.
Known by the number. However, the generator according to the invention comprises a fissile radioisotope and a cation exchange material. Since the fission-produced radioisotope is completely or substantially free of carriers, a small amount of adsorbent for the parent isotope is sufficient. As a result, the dimensions of the generator can be greatly reduced, so that the device can be Its handling is easy. It is also a great advantage that cation exchange substances are present in the generator according to the invention. As a result, it is possible to purify the eluate within the generator itself, so that the liquid withdrawn from the generator and comprising the radioactive daughter isotope has a high degree of chemical and radionuclide purity; Therefore, it is suitable for radiopharmaceutical applications. Further purification of the eluate, ie after leaving the generator, is unnecessary. Such further purification is generally not even possible;
Or at least not desirable. It is important to note that the resulting daughter isotopes usually have too short a half-life to be able to carry out such post-treatments, and that they cannot be used in hospitals or clinical laboratories that lack suitable auxiliary means for the purpose. This is because post-processing is out of the question for safety reasons. The adsorbent for the parent isotope in the reservoir of the generator is usually enclosed between two filters. To load the adsorbent with radioactive parent isotope, a solution of this isotope is passed through one side of the reservoir. Glass wool or glass beads are often used as overfill material on this side. However, the glass beads create channels in the adsorbent resulting in poor packing and non-uniform distribution of the parent isotope on the adsorbent. Glass wool often prevents its filling due to too high a resistance, which also causes synthetic resins, such as polyethylene, to adsorb very little of the parent isotope. This latter situation is highly undesirable. This is because, with respect to generator elution, the amount of parent isotope that is not adsorbed to the adsorbent will contaminate the eluate. It has been found that, as a particular feature of the invention, the above-mentioned disadvantages can be eliminated by means of a filter consisting of semi-molten glass, placed on the reservoir side of the generator through which the solution of the isotope according to the invention passes. There is. When using such a filter, which can of course also be used in isotope generators according to the prior art, an effective and homogeneous loading of the adsorbent is easily achieved, while the filter makes it possible to isolate the parent isotope. is not adsorbed at all. The generator 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, the dimensions of the generator can be minimized and it is possible to reach an optimal purity of the radiopharmaceutical composition, better preserving the above-mentioned advantages while reducing production costs to the possible. can 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, and the periphery of the compartment is separated from the reservoir by a filter. is connected to the inner wall of One compartment of the reservoir comprises an eluent supply facility and an adsorbent for the parent isotope located between the supply facility and the separation filter, the adsorbent being located between the aforementioned semi-molten glass filter and the separation filter. enclosed in between. Another compartment of the reservoir comprises an eluate outlet facility.
A cation exchange material is present between the separation filter and the outlet equipment, and the voids between the adsorbent particles as well as the voids between the ion exchanger particles are filled with a physiological solution. Separating filters suitable for this purpose consist of two filter disks that completely or substantially completely cover each other, the disks in contact with the adsorbent being made of glass fiber paper, e.g. ) consists of a prefilter AP200, the disk in contact with the ion exchanger consisting of porous polyethylene. Finally, the invention relates to a storage tank for the aforementioned generator,
This reservoir contains both the cation exchange material and the adsorbent for the parent isotope. This filled and sterilized reservoir can be stored unrefrigerated for more than three months and can be incorporated into the generator at any time within this period without any pretreatment. The reservoir can then be used to obtain a high yield of eluate containing the radioactive daughter isotope. This is an advantage. This is because the storage tanks can be manufactured in stock and shipped to the generator supplier, who can remove the generator when desired without any pre-treatment. This is because a storage tank can be used for this purpose, resulting in significant cost savings using this method. The present invention will be explained in more detail with reference to the following specific examples. The accompanying drawing is a cross-sectional view of a preferred embodiment of an isotope generator reservoir according to the invention, in which a suitable inert material, such as glass or a polymeric material,
A generally cylindrical reservoir 4, preferably made of borosilicate glass, is enlarged at each end and is provided with flanges 10,13. The openings at the two ends of the reservoir are closed by rubber stops 2, 14 comprising flange parts 11, 15 and jacket parts 12, 16. The flange portion of the stopper engages the flange portion of the reservoir, and the jacket portion fits into the opening of the reservoir. The flange part of the stopper and the reservoir are connected together by a metal cap, for example an aluminum bent cap 1,17. The reservoir contains a slurry of adsorbent 5 in a solution consisting of 0.9% NaCl in water.
This adsorbent consists of Al ₂ O ₃ granules which are totally or partially covered with a layer of fully or partially hydrated manganese dioxide. In the reservoir, the adsorbent is enclosed between a filter made of semi-fused glass 3 with average porosity and a filter disc made of glass fiber paper 6, ie millipore prefilter AP200. The reservoir further contains a slurry consisting of Na ⁺ form 8 resin "Bio-Rad 50W-X8" in a 0.9% NaCl solution in water. The resin has been converted to the Na ⁺ form by treatment with NaOH with subsequent washing with water. This resin is used for filter disk 6
A filter disc 7 of porous polyethylene which engages with the filter disc 18 is likewise comprised of porous polyethylene and is supported by a spacer ring 9 of polycarbonate. Example 1 Ten reservoirs described above were stored for 3 months and then used for the experiments described below. Each storage tank is made of sodium molybdate (PH1-5-5) by drilling stoppers at both ends.
10) Fission-produced molybdenum in the form - carrier-free
It was filled with 99Mo, resulting in inlet and outlet holes, and then a solution of radioactive sodium molybdate was flowed into the reservoir through the inlet hole (at A). After washing and sterilization in an autoclave at 121° C. for 30 minutes, the isotope generator thus obtained was placed in a generator. The radioactivity of the isotope generator was 1000 mCi. When using a generator, the eluent is placed at one end of the reservoir (A
The eluate was fed through the inlet hole at the other end of the reservoir, while the eluate exited through the outlet hole at the other end of the reservoir. The generator contains a sterile isotonic salt solution (0.9% wt/vol% in water).
NaCl) was eluted in volumes ranging from 4.6 to 15 ml. The average elution yields recorded in the table below were obtained.

Table: The elution yields recorded in the table above are logically obtained per elution. Showing the average percentage over 10 generators of ^99m Tc titer. Continuously each day, the generator elutes 10 times each. As a result, a high elution yield was achieved. i.e. eluent 4.6 and
^99m theoretically obtained by elution using 15ml
Mean Tc titer 89.06% ± 2.15 and 92.17% ± 1.64
reached. Also recorded in the above table are the average analysis results of the eluate. Radiochemical purity was determined by paper chromatography and Mn ⁺⁺ and
Al ⁺⁺⁺ was measured by spectrophotometric method and colorimetric (quinalizarin complex) method, respectively. The labeling efficiency of ethylene hydroxy diphosphonate (EHDP) and Sb ₂ S ₃ indicates that the obtained ^99m Tc is very suitable for preparing ^99m Tc-labeled compounds and can therefore be used for any desired application. proves that. Radionuclide purity was determined using a gamma spectrometer. At best, traces of the radioisotopes ⁹⁹ Mo, ¹³¹ I and ¹⁰³ Ru were detectable, and no other radionuclide contamination was found in the eluate. From the results shown above, long-term storage of the filled storage tank did not have any adverse effects on the elution yield and purity of the eluate. In another experiment, the same isotope generator was run 15 times using 20 ml of physiological saline solution without changing the elution yield or contaminating the resulting ^99m technetium eluate with cationic contaminants. It was possible to elute. Example 2 Ten additional columns, substantially similar to those used in Example 1, but with lower dead volume, were used to generate ^99m Tc. This reduced empty volume allows for the elimination of spacer rings in column construction and the utilization of cylindrical, non-flange storage vessels. The results are shown in the table below.

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【table】

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[Table] Non-pyrogenic - Contains no pyrogens

[Brief explanation of drawings]

The figure is a cross-sectional view showing a preferred embodiment of a storage tank for an isotope generator according to the present invention. 3...Semi-molten glass, 4...Storage tank, 5...Adsorbent, 6...Glass fiber paper, 7, 18...Filter disk, 8
…Na ⁺ shape resin.

Claims (1)

[Claims] 1. A radioactive isotope obtained by elution using a physiological solution from molybdenum-99 as a parent isotope and radioactive technetium-99m as a daughter isotope adsorbed on an adsorbent. It is used to prepare a liquid for radiopharmaceutical application consisting of a storage tank having eluent supply equipment and eluent outlet equipment, and in which an adsorbent for the parent isotope is present. An isotope generator for isotope generation, characterized in that the parent isotope is a fission-produced isotope and does not contain a carrier, and a cation exchange resin is present in the generator. Device. 2. The isotope generator according to claim 1, wherein the cation exchange substance as a parent isotope eluent is present in the storage tank. 3. The isotope generator according to claim 2, wherein the storage tank contains both the parent isotope adsorbent and the cation exchange material. 4. Radiopharmaceutical applications comprising a radioactive isotope obtained by elution using a physiological solution from molybdenum-99 as the parent isotope and radioactive technetium-99m as the daughter isotope adsorbed on an adsorbent. In an isotope generator, which is used to prepare a liquid for the parent isotope, and is equipped with a storage tank having eluent supply equipment and eluent outlet equipment, and in which an adsorbent for the parent isotope is present. , the storage tank is separated into two compartments, each compartment is separated from each other by a separation filter, the periphery of the separation filter is in contact with the inner wall of the storage tank, and one compartment is equipped with the supply equipment. The adsorbent for the parent isotope is contained between the molten glass filter provided on the side of the supply equipment and the separation filter, and the other compartment is equipped with the outlet equipment, and further contains the adsorbent for the parent isotope. An isotope isotope characterized in that a cation exchange material is present between the separation filter and the outlet equipment, and the voids between the adsorbed particles as well as the voids between the ion exchanger particles are filled with a physiological solution. Generator. 5. said separation filter consists of two disks completely covering each other, the disk in contact with said adsorbent consists of glass fibers and the disk adjacent to said ion exchanger consists of porous 5. The isotope generator according to claim 4, which is made of high quality polyethylene. 6. The isotope according to claim 4 or 5, wherein aluminum oxide containing partially or completely hydrated manganese dioxide is used as the adsorbent for the parent isotope. body generator. 7 Claims 4 to 6 in which a cation exchange resin is used as the cation exchange material
The isotope generator according to any one of the above items. 8. The isotope generator according to claim 7, wherein a neutralized strongly acidic cation exchange resin is used as the resin. 9. The isotope generator according to claim 8, wherein a strongly acidic cation exchange resin converted into Na ⁺ , K ⁺ or NH4 ⁺ type is used as the resin. 10 Particle size is 50-400 mesh, preferably 100-400 mesh
Any one of claims 8 to 9 using a cation exchange resin having 200 meshes
Isotope generator described in Section 1.