IL34751A - Production of fission product technetium 99-m generator - Google Patents

Description

ms»s TS D Ι3*π¾ 95-9 lai«S330 Φ ino, nis* PRODUCTION OF FISSION PRODUCT TECHNETIUM 99~» GENERATOR This invention relates to the production of a fission product technetium-99m generator. In one aspect, this invention relates to the production of radioactive technetium-99m in concentrations greater than heretofore known. A further aspect of this invention is directed to the production of radioactive technetium-99m which can be obtained in a high degree of purity.

Recent medical investigation has shown that technetium- 99m is an extremely useful tool for diagnosis. High purity technetium-99m is used primarily as a radioisotope in a variety of medical research and diagnosis. It is well suited for liver, lung, blood pool and tumor scanning, and is preferred over other radioactive isotopes because of its short half-life which re- suits in reduced exposure of the organs to radiation. In addition to medical uses, technetium-99m can also be employed in industrial applications, such as in the measurement of flow rates, process control, radiometric chemistry, and the like. Since the radioisotope sought to be used has such a short half-life, it is common practice to ship the users of the isotope the parent element; in this case radioactive molybdenura-99. The user then extracts the technetium from the molybdenum-99 as his needs require.

For medical diagnosis, radioactive technetium-99m is on an alumina column. The column is then sent to a hospital where the physician washes the column and obtains the 99χ0πι solution The 99Tcm solution is administered to a patient orally or by intravenou injection. The radioactivity of 9 χ0ιη localizes in brain, lung, liver, spleen, bone, and the like depending on the preparation of 99Tcm solution. The localization of the activity can be detected by typical scanning. This technique which does not the last four years and hence, the need for the 99TC™ generator is rapidly expanding.

The first ^^Tcm generator was developed at Brookhaven National Laboratory. Uranium was irradiated in a reactor and the ^Mo that was produced by nuclear fission process was separated by alumina chromatography. The purified 99jfo was again adsorbed on an alumina, column from an acidic medium and the 99-jcm in the column was recovered by dilute hydrochloric acid. This method, however, has never been used for medicine because radionuclide purity of 99Tcm solution from the Brookhaven column was not considered high enough for medical use. It contained significant amounts of radioisotope impurities such as lO^Ru, iodine isotopes, and the like.

More recently, as disclosed in U.S. Patent 3,382,152, a medical 9Tcm generator was developed by using reactor irradiated molybdenum in place of uranium target. When molybdenum is irradiated in a reactor, with a high degree of radionuclide purity is obtained by the (η, ) reaction. Furthermore, the chemical processing of the irradiated target is simple.

This method is currently widely used by radiopharmaceutical manufacturers.

However, when the Mo target is irradiated in the reactor, only an extremely small portion of the Mo is converted to radioactive 99MO by the (η,*γ) reaction. Therefore, the specific activity of activity to the total weight of elemental Mo, is small. Therefore, the active adsorption sites on alumina are virtually consumed by inactive Mo, thus requiring a larger adsorption capacity to load a high ^¾lo activity. Although several investigations have also been reported on adsorbents with a high adsorption capacity for Mo, the specific activity of (η,γ) ^Ho is proportional to the neutron flux molybdenAum in this case adversely limits the activity of a 99Tcm generator.

It is therefore an object of this invention to provide a high activity fission product technetium-99m generator. Another object of this invention is to provide a process for the production of a technetium-99m generator which is rapid and requires little or no pretreatment or post-treatment of the column substrate. A further object of this invention is to provide a generator from which technetium-99m having a concentration greater than heretofore known can be eluted. Another object of this invention is to provide a process for preparing radioactive technetium-99m in a high degree of purity and by an extremely reproducible process. A further object of this invention is to provide a process which avoids the need for separating radioactive products and other impurities. These and other objects will readily become apparent to those skilled in the art in the light of teachings herein set forth.

In its broad aspect, the present invention is directed to the production of a fission product technetium-99m generator. The process comprises the steps of (a) dissolving in an aqueous solution at pH of from about 4 to 9 an inorganic salt of fission product molybdenum-99 having a radionuclidic purity of at least 99*99%, (b) contacting a column containing an inorganic substrate which selectively retains molybdate ions with said solution to load said column, and (c) selectively eluting said column with a solvent to separate technetiui--99m from its radioactive parent molybdenum-99 that is deposited on the substrate.

Operating in the aforesaid manner provides a selective separation of technetium-99m from the fission product radioactive molybdenum-99 compound with very high efficiency, i.e., over 80 per cent. In addition the process of this invention the necessity for pretreatment and post-treatment of the substrate. In contrast to known generators which usually take at least 2 hours to prepare, the generators of this invention can be conveniently prepared in less than 5 minutes. Moreover, since fission product molybdenum-99 is employed, the resulting technetium-99m solution is of a greater concentration than heretofore possible. Prior to the instant invention the highest concentration obtained was less than 10 millicuries per milliliter. In contrast, technetium-99 can be obtained from the generators of this invention in concentrations of 1000 or a higher, millicuries per milliliter.

While not wishing to be bound by any theory regarding the mechanism of this process, a brief discussion of what is believed to take place will aid in understanding the invention.

When alumina particles are immersed In water, the surface electrical charges differ due to the pH of water.

Alumina in acid has positive electrical charges and can attract negatively charged particles by coulombic Interaction. Alumina in basic media acts as a cation exchanger. At neutral pH, the surface electrical charge Is neutral and consequently an ion exchange reaction does not take place. However, at the adsorption of molybdate on alumina in an acidic medium, a very complex adsorption equilibria takes place due to the polymerization of molybdate ion. Such a polymerisation process appears very favorable for the adsorption of Mo on a positively charged alumina surface. For example, one active point on alumina surface can adsorb 0.5 atoms of Mo at pH 6 and 2 atoms at pH 4.5. However, alumina is amphoteric, different from an ordinary ion exchange resin. Therefore, alumina loses the adsorbability for Mo when the pH of its surface reaches its isoelectrical point. Therefore, the amount of molybdenum adsorbed on the substrate capacity of molybdenum on alumina at various pH was determined.

The experimental results indicate: (1) the -adsorption capacity of alumina is 20 mg Mo/g alumina at pH lower than 4.5: (2) the adsorption capacity decreases to 2 mg Mo/g at pH 4.5-5: (3) the adsorption capacity of 2 mg Mo/g continues up to pH 6.5: (4) exceeding pH 6.5^ the system does not provide a high separation factor such as 10^" because the isoelectric point is reached.

However, the pH of the loading solution can be as high as pH 9.

The alumina column adjusts pH automatically to neutral. The above statement (3) indicate an extremely important fact that, when fission product %o ±s loaded on alumina at pH 4 to 9 where no treatment of the substrate is required, a high activity possesses^ xcm generator tha rjggses a high separation factor can be ob-tained. At pH below 4, pretreatrient and post-treatment of the substrate is necessary. By eliminating substrate treatment, the 99"rcm generators can be produced in much shorter times than presently experienced.

As previously indicated, the process of this invention utilizes fission product molybdenum-99 which has a radionuclidic purity of at least 99.99 per cent. Irradiation of compounds to produce fission product molybdenum-99 is a well known technique and can be effected by placing the proper compound in the irradiation zone of a nuclear reactor, particle generator, or neutron isotopic source. Although a variety of compounds are suitable for use in the process of this invention the preferred target is uranium-235. In the event that other compounds are employed, it is often necessary to isolate the molybdenum component after irradiation. Illustrative compounds which can be employed as the source of fission product molybdenum-99 include, among others, fissionable materials such as uranium-238, pluton-ium-239, and the like.

Thereafter, the irradiated compound is dissolved in a suitable solvent and the ^^Mo is selectively removed. The techniques to dissolve and isolate a pure molybdenum-99 as its inorganic salt are well known in the art.

The fission product molybdenum-99 in the form of an inorganic salt, such as sodium molybdate, potassium molybdate, ammonium molybdate and the like, is then dissolved in an aqueous solution at a pH of from about 4 to about 9. If necessary the pH can be adjusted to this range by the addition of acid or base. When the molybdenum salt is dissolved, the solution is then poured onto a column containing a substrate, such as alumina, zirconium oxide and the like. No washing or treatment of the column is necessary before or after loading.

The column or generator can be fabricated in accordance with the teachings of the prior art. For example, if the radioisotope is to be used for diagnostic studies, the substrate can be contained in a sterile generator such as that described in U.S. Patent No. 3,369,121.

In contrast to the work described in the literature, wherein hydrous zirconium oxide and alumina was employed, it has now been unexpectedly and surprisingly found that when an inorganic substrate such as alumina is contacted with an aqueous solution, at pH 4 to 9, of fission product inorganic molybdenum salts, the substrate selectively adsorbs molybdenum but does not appear to adsorb technetium. It is also surprising that (a) the loading capacity of the system is such that equivalent amounts of technetium are yielded when using physiological saline, (b) that the saline containing the technetium product has unexpectedly lower elemental impurities due to the molybdenum absorbing substrate, (c) that the saline contains appreciably more technetium and less molybdenum than comparable systems heretofore known and (d) the substrate needs no pretreatment or post- treatment .

The technetium-99m in the column or vessel which contains jlo- 99mTc activity can subsequently be isolated, e.g., milked, filtered, centrifuged or the like for technetium-99m as it is formed with an acidic, neutral or basic solution. Preferably, it has been observed that best results are obtained when the system is eluted with 4 milliliter portions of isotonic saline solutions. ^ subs rate/ This is done by contacting the Asmbo rae^ with the desired volume of saline and collecting the liquid portion.

Numerous variations of the preferred embodiment described above may be practiced, as will be apparent to those skilled in the art, without departing from the basic concepts of the present invention.

As previously indicated, the process of the present invention provides a simple method for the preparation of technetium-99m in a high degree of efficiency. By this process recovery of technetium-99m can be effected with isotonic saline in efficiencies as high as 95% and higher, over a pH range of about 4.0 to about 7.0 without appreciable dissolution of the substrate or removal of any molybdenum from the substrate.

A further advantage characteristic of the process of this invention, is that the substrate and/or the entire elution system can be sterilized, i.e., by autoclaving at the normally prescribed temperatures and pressures.

The radiometric analysis of the eluted technetium-99m indicates that it contains up to 957« of the available technetium-99m and £he. radionuclidic purity ¼s greater than 99.99%. The The substrate and/or the entire elution system can be sterilized by acceptable autoclave techniques with no reduction in radionuclidic impurity and no increase in the metal element impurities.

The following examples are illustrative: EXAMPLE I Five grams of uranium oxide (natural U) was irradiated in a nuclear reactor at a neutron flux of 10^ for 15 minutes. The target was dissolved in 50 cc of 37» H2O2 and 8 cc of concentrated H2SO . After dissolution, 50 cc of 6% H2SO3 was added. The sample solution was passed through an 99MO adsorption column (1 cm X 5 cm) containing 2 cc of silver-coated-charcoal (20-50 mesh) and 2 cc of charcoal. The column was washed with 60 cc of dilute sulfuric acid and 60 cc of water successively. The ^Ho retained in the column was eluted with 40 cc of hot 0.2M NaOH. eluentj The to-luant was passed through another purification column that contains 2 cc of silver-coated-charcoal at the upper part of the column and 2 cc of zirconium phosphate at the bottom. To the "MO product solution thus obtained, 5 cc of 1% NaCl - 1 MHC1 was added to make the ^Mo solution isotonic saline solution. The solution contained about 1 millicuries of ^Mo. To the isotonic saline solution obtained by the above method, 2 mg of non-radioactive Mo carrier in a form of sodium molybdate was added. The sample solution was passed through a Woelm alumina column (0.6 X 3 cm). The adsorption of 99Mo was >99.999 . The "Tcm, after is accumulated in the column, was eluted with 10 cc of isotonic saline solution and recovered :>90%. The ^Ho content in the 99xcm element was in the order of 10"^%. No other radioactive contaminants were detected either by a Ge (Li) or Nal (Tl) crystal coupled with a gamma-ray analyzer. The alumina breakthrough in the "Tcm eiugn was very small ¾1 ppm. The total EXAMPLE II T o curies of fission product 99MO was dissolved in 10 cc of isotonic saline solution. The sample solution was passed through a Woelm alumina column (1 cm diam. X 2 cm height) . The adsorption of 99Mo was > 99.995%. The recoveiy of 99Tcm was 90% in 4 cc of isotonic saline solution. The 99Mo breakthrough was in the order of 10 The concentration of the eluted technetium-99m was greater than 400 millicuries per milliliter.

Table I below shows a comparison of a generator prepared by the process of this invention with a generator prepared in accordance with the teachings of U.S. Patent 3,382,152.

Table I Comparison of Fission Product Technetium-99m Generator With Known Generator Fission Product Known Technetium-99m Description Generator Generator Time required for ■> 2 hours .5 minutes the adsorption and washing steps Amount of radio

Claims (12)

WHAT IS CLAIMED IS:

1. A process for the production of a fission product generator which comprises the steps of: a. dissolving in an aqueous solution at pH of from about 4 to about 9 an inorganic salt of fission product molybdenum-99 , said molybdenum-99 having a radionuclidic purity of at least 99.99 per cent, b. contacting a column containing an inorganic substrate which selectively retains molybdate ions with said solution to load said column, and eluttpgf c. selectively filiation said column with a solvent to separate technetium-99m from its radioactive parent molybdenum-99 that is deposited on the substrate.

2. The process of claim 1, wherein said substrate is alumina.

3. The process of claim 1 wherein said substrate is zirconium oxide.

4. The process of claim 1 wherein said eluting solvent is isotonic saline solution. ¾ n»rtwi prooooo of oloim 1 wherei -aaid oluting polyene

5. The process of claim 1 wherein said eluting solvent is water.

6. The process of claim 1 wherein said eluting solvent is a d ilute inorganic acid.

7. The process of claim 1 wherein the eluted technetium

8. . The process of claim 1 wherein eluted technetium-99m has a radioactivity concentration of greater than 25 milli-c uries per milliliter.

9. The process of claim 1 wherein the eluted tech-netium-99m has a radionuclidic purity greater than 99.99 per cent. •used for the process of claim 1

10. lA. A generator for the production of fission product technetium-99m comprised of a hollow portable body, closed at its top and bottom by pierceable autoclavable closures and contained therein a sterile, non-pyrogenic substrate having adsorbed thereon fission product molybdenum-99, said molybdenum-99 having a radionuclidic purity of at least 99.99 per cent. 10/

11. The generator of claim ί± wherein said substrate is alumina.

12. t L The generator of claim ¾ wherein said substrate is zirconium oxide. Attorney for Applies