TWI488684B - A Method for Purification of Gallium - 69 Isotopes - Google Patents

Description

Purification and recovery method of gallium-69 isotope

A method for recovering and purifying gallium-69 from cesium-68 radioactive liquid, in particular to a panning purification and recovery of gallium-69 in the development of a strontium-68/gallium-68 PET (positive angiography) nuclear generator It can be used in the nuclear medicine industry for the development and purification process of the 68-68/gallium-68 PET (positive angiography) nuclear generator.

The elution and purification process of the known radioactive liquid agent is subjected to a pre-treatment such as dissolution and filtration, followed by a panning operation with a solution of different concentration and volume to obtain a radioactive nuclear species to be obtained, and the remaining radioactive waste liquid is subjected to radioactive waste treatment. While ignoring the value of recycling and recycling, it also increases the chance of dose-polluting the environment. If it is unable to effectively recover the eluent which still contains high content of gallium-69 and re-form it into a solid target, the cost will be difficult to reduce. And the reduction of the yield of gallium 68 nucleus and the waste of resources, etc., are necessary for improvement. A method for purifying a carrier-free gallium-67 has been disclosed, for example, in US Patent Publication No. US2011/0214995 (Method for Making Radioactive Isotopic Gallium-67); Republic of China Patent Application No. 099104331. The nuclear production conditions are 15 to 40 million electron volt (MeV) proton irradiation electroplating, and the nuclear species to be obtained are generated by a nuclear reaction. In the chemical procedure, an ion exchange resin is selected as the adsorbent, and the panning is repeated with concentrated hydrochloric acid, and finally the panning solution is subjected to evaporation treatment to obtain a target nuclear species. However, there is no recycling process for the eluent. Instead, the radioactive waste liquid treatment is directly carried out, which is a pity in terms of resource reuse.

Gallium-68 has been highly valued by the nuclear medical community because it can be used to produce related nuclear medicines by the 锗-68/gallium-68 generator system, such as US Patent Publication No. US2012/0011965 (Gallium-68). A radioisotope generator and generating method thereof); Republic of China Patent Application No. 099123148) has disclosed a gallium-68 isotope generating apparatus and method thereof that can replace a conventional time-consuming and costly particle cyclotron. In the current clinical trials of PET scintigraphy, the 锗-68/gallium-68 generator system is one of the important sources of production of related nuclear medicines, because this system can be produced from the mother nucleus (锗-68) with long half-life. A sub-nucleus with a short half-life (gallium-68). Prior to the marking, the column retention activity is eluted with hydrochloric acid of a specific volume and concentration to produce a chlorine-containing solution without acetate ions. After a long-term panning test, the results show that a very high gallium-68 nuclear species yield can be obtained with a specific volume and concentration of hydrochloric acid as a condition for each panning column. However, if the gallium-69 can be recovered and re-purified to obtain a solid target after the column is pre-purified, the efficiency of nuclear utilization can be more effectively improved, and the concept of resource reuse can be met.

In view of the above-mentioned shortcomings of the conventional washing and purifying process of the radioactive liquid, the inventors have studied the improvement of the disadvantage, so that the present invention is produced.

In order to solve the above problems, an object of the present invention is to provide a panning and purifying method for recovering gallium-69 in a 锗-68 eluting liquid (eluent), which provides a method for recovering yttrium-68 radioactive liquid contained in Gallium-69 is used for recycling.锗-68 elution solution is irradiated onto a silver substrate by protons The electroplated gallium-69/silver alloy method is obtained, and the gallium-69/silver alloy electroplating target is dissolved in an acidic solution to obtain an acidic solution containing gallium-69 and cesium-68, and then pretreated with an ion exchange resin column. Purification, the filtered gallium-69 eluent is neutralized with an alkaline solution, and then dried, sintered, etc. to form completely dried gallium oxide, and the recovered gallium oxide is dissolved and re-formed into a target. Material, you can repeat the panning action again.

Another object of the present invention is to provide a method for recycling and reusing a radioactive liquid agent, which can effectively reduce the content of gallium-69 in the waste liquid, thereby reducing the chance of the radioactive waste liquid polluting the environment and reducing the stringent requirements for environmental equipment. It can also save investment and production costs appropriately. In addition to the 锗-68/gallium-68 PET (positive angiography) nuclear generator process, the method is also indispensable for the related drug development and purification process.

To achieve a more specific understanding of the above objects, effects and features of the present invention, the following description will be made with the accompanying drawings.

S11~S17‧‧‧Steps

Fig. 1 is a flow chart of a panning purification process of a conventionally recovered radioactive liquid.

Fig. 2 is a flow chart showing the recovery and purification method of the gallium-69 isotope of the present invention.

Figure 1 is a conventional method for purifying and purifying radioactive liquid. The eluted 锗-68 scouring solution still contains high content of gallium-69. If it is not properly washed and purified, it will easily cause gallium-68. The situation in which the yield of the nuclear species is reduced and the waste of resources occurs, the content of which has been described in the prior art, and will not be described herein.

2 is a flow chart of a method for recovering and purifying a gallium-69 isotope of the present invention, After the target plated with gallium-69/silver alloy is irradiated by a cyclotron, the irradiated gallium-69/silver alloy plating target is dissolved in an acidic solution to obtain an acidic solution containing gallium-69 and strontium-68, and then The gallium-69 is pre-purified and recovered, and at least comprises the following steps:

Step S11: Conditioning is carried out by using an acidic solution through an ion exchange resin adsorption column (using Purolite S-910 resin).

Step S12: The acidic solution containing gallium-69 and strontium-68 is pre-purified through the ion exchange resin adsorption column to filter out the gallium-69 eluent.

Step S13: recovering the gallium-69 eluent obtained by filtration, neutralizing the gallium-69 eluent with an alkaline solution, and monitoring the gallium-69 eluent by a pH meter or an acid-base test paper. The pH value is changed, and the pH of the gallium-69 eluate is slowly adjusted to neutral.

Step S14: Controlling the addition of the alkaline solution to the pH-pH of the gallium-69 eluate is less than 7, and the precipitation phenomenon gradually occurs at the bottom, and the addition of the alkaline solution is continued until the bottom precipitation is completed.

Step S15: The upper clear liquid is separated from the bottom precipitate gallium-69, and all of the precipitate gallium-69 is collected and dried to be completely dry.

Step S16: The completely dried precipitate gallium-69 is sintered to form a gallium-69 chemical structure of the precipitate gallium-69 to complete the gallium-69 recovery procedure in the gallium-69 eluate.

Step S17: Dissolving the recovered gallium oxide and re-forming the target to be recycled again.

The following is one of the examples of the recovery and purification method of the gallium-69 isotope of the present invention.

Step S11: First, 9 to 12 N (optimum condition: 10 N) of nitric acid is passed through an ion exchange resin adsorption column (using Purolite S-910 resin) for conditioning.

Step S12: dissolving the electroplating target after the proton irradiation, and eluting the gallium-containing-69, 锗-68 The acidic solution was pre-purified by a conditional ion exchange resin adsorption column to filter out the gallium-69 eluent.

Step S13: recovering the gallium-69 eluent obtained by filtration, neutralizing the gallium-69 eluate with an equivalent concentration of 1N or 2N sodium hydroxide, and monitoring the gallium by a pH meter or an acid-base test paper. -69 The pH value of the eluent was changed, and the pH of the gallium-69 eluate was slowly adjusted to neutral.

Step S14: controlling the pH of the sodium hydroxide to the gallium-69 eluent to be less than 7, and gradually forming a precipitate at the bottom. When the pH of the liquid to be washed is close to 6.5-6.8, the bottom precipitation is completed.

Step S15: The upper clear liquid was separated from the bottom precipitate gallium-69, and all of the precipitate gallium-69 was collected and dried in an oven at about 100 ° C until completely dried.

Step S16: The completely dried precipitate gallium-69 is sintered at about 1100 ° C, and the precipitate gallium-69 forms a chemical structure of gallium oxide, and the gallium-69 recovery procedure in the gallium-69 eluate is completed.

Step S17: Dissolving the recovered gallium oxide and re-forming the target to be recycled again.

It can be seen from the above that the gallium-69 purification and recovery method of the present invention can effectively filter the radioactive liquid agent by eluting and purifying, thereby not only improving production efficiency, but also reducing environmental pollution, thereby effectively reducing cost and complying with resource recycling. The policy of re-use has indeed been industrially utilized, novel and progressive.

The above description is only a preferred embodiment of the invention and is not intended to limit the scope of the invention. That is, the equivalent changes and modifications made by the scope of the patent application of the present invention are covered by the scope of the invention.

S11~S17‧‧‧Steps

Claims (6)

A method for recovering and purifying gallium-69 isotope by irradiating a target plated with gallium-69/silver alloy with a cyclotron, and dissolving the irradiated gallium-69/silver alloy electroplating target with an acidic solution to obtain a gallium-containing target -69, an acidic solution of 锗-68, and then pre-purifying and recovering gallium-69, which comprises at least the following steps: Step S11: first conditioned by an acidic solution through an ion exchange resin adsorption column; Step S12: The acidic solution of gallium-69 and strontium-68 is pre-purified by the ion exchange resin adsorption column to filter out the gallium-69 eluent; step S13: recovering the gallium-69 eluate obtained by filtration to be alkaline The solution is neutralized with the gallium-69 eluate, and the pH value of the gallium-69 eluate is monitored by a pH meter or an acid-base test paper, and the alkaline solution is slowly added to adjust the gallium-69 eluent. The pH is neutral to the step; step S14: controlling the alkaline solution to be added to the gallium-69 eluent until the pH is less than 7, the precipitation gradually occurs at the bottom, and the addition of the alkaline solution is continued until the bottom precipitation is completed; step S15: the upper portion is The clear liquid is separated from the bottom precipitate gallium-69, collected The precipitate gallium-69 is dried to completely dry; step S16: the completely dried precipitate gallium-69 is sintered, and the precipitate gallium-69 forms a chemical structure of gallium oxide to complete the gallium-69 wash-off The gallium-69 recovery procedure in the liquid; step S17: dissolving the gallium oxide and re-forming the target to be recycled again. According to the recovery method of the gallium-69 isotope according to the first item of the patent application scope, wherein the acidic solution is nitric acid, the equivalent concentration is 9~10N, and the optimum condition is 10N. A method for recovering and purifying a gallium-69 isotope according to the first aspect of the patent application, wherein the alkaline solution is sodium hydroxide and has an equivalent concentration of 1 N or 2 N. According to the recovery and purification method of the gallium-69 isotope according to the first aspect of the patent application, wherein the pH of the step S14 is completed, the pH of the pH is about 6.8 to 6.5. The method for recovering and purifying gallium-69 isotope according to item 1 of the patent application scope, wherein the drying temperature in step S15 is about 100 °C. The method for recovering and purifying gallium-69 isotope according to item 1 of the patent application scope, wherein the sintering temperature of step S16 is about 1100 °C.