EA042622B1 - METHOD FOR PRODUCING ISOTOPE - Google Patents

Description

Technical field

The invention relates to the production of radium-223 ( ²²³ Ra) for use in pharmaceutical production. In particular, the present invention relates to methods for the production on an industrial scale of radium-223 with a purity suitable for pharmaceutical use in human subjects.

State of the art

The lysis of certain cells can be essential in the successful treatment of a number of mammalian diseases. Typical examples are the therapy of malignancies such as sarcoma and carcinoma. However, the selective killing of certain cell types may also play a key role in the treatment of a variety of other diseases, mainly immune system diseases, hyperplastic and/or other neoplastic diseases.

The most common elective therapies currently available are surgery, chemotherapy, and external radiation. However, targeted endo-radionuclide therapy is of interest and is an emerging area with the potential to transfer highly cytotoxic effects to unwanted cell types. Currently, the most common forms of radiopharmaceuticals that are approved for use in humans are radionuclides emitting beta and/or gamma particles. However, the use of alpha-emitting radionuclides in therapy is most interesting, due to their potential for greater lysis of certain cells. In particular, one of the alpha-emitting nuclides, radium-223 ( ²²³ Ra), has a distinctly noticeable efficacy, in particular for the treatment of diseases associated with bones and the surface of bones.

The radiation range of conventional alpha particle radiation sources in the physiological environment is generally less than 100 µm, which is equivalent to only a few cell diameters. This makes this nucleus well suited for the treatment of tumors, including micrometastases, since little of the radiated energy will pass outside the target cell and thus damage to surrounding healthy tissue can be minimized (see Feinendegen et al., Radiat Res 148: 195201 (1997)). Conversely, beta particles have a range of 1 mm or more in water (see Wilbur, Antibody Immunocon Radiopharm 4: 85-96 (1991)).

The radiation energy of alpha particles is quite high compared to beta particles, gamma rays and X-rays and is usually 5-8 MeV, which is 5 to 10 times greater than that of beta particles and 20 or more times greater than the energy of gamma -rays. Thus, the release of a large amount of energy over a very short distance produces α radiation with exceptionally high linear energy transfer (LET), high relative biological efficiency (RBE), and low oxygen amplification factor (OFC) compared to gamma and beta radiation ( see Hall, Radiobiology for the radiologist, Fifth edition, Lippincott Williams & Wilkins, Philadelphia PA, USA, 2000). All this explains the exceptional cytotoxicity of alpha-emitting radionuclides and also imposes stringent requirements on the level of purity required when the isotope is needed for internal use. This is especially true when some impurities may also be emitters of alpha particles, and more specifically when sources of alpha particle radiation with a long half-life may be present, as they can potentially cause significant damage over an extended period of time.

Below is one of the radioactive decay chains leading to the formation of ²²³ Ra, which is used as a source of this isotope in small quantities. The diagram shows the element, molecular weight (Mw), decay form (form), and half-life (years (L) or days (days)) for ²²³ Ra and its two isotopic precursors. Production begins with ²²⁷ Ac, which itself is found only in trace amounts in uranium ores, being part of the natural decay chain that occurs with ²³⁵ U. One ton of uranium ore contains approximately tens of grams of actinium and thus, although ²²⁷ Ac is found in natural conditions, in most cases it is obtained by neutron irradiation with ²²⁶ Ra in a nuclear reactor.

It can be seen from the diagram that ²²⁷ Ac, with a half-life of over 20 years, is potentially hazardous

- 1 042622 impurity, taking into account the receipt of ²²³ Ra from the above chain of radioactive decay for pharmaceutical purposes. In particular, although ²²⁷ Ac itself is a source of beta radiation, its long half-life means that even very low activity is a radiative forcing with a significant lifetime, and moreover, after its decay, the resulting daughter nucleus (i.e. ²²⁷ Th) also produces 5 alpha decays and 2 beta decays before reaching the stable ²⁰⁷ Pb. See the table below:

Nuclide ^22/ Th ²²³ Ra ^{2yRn 2nd} Po ² ” R ² Bi 2U/ _T | ^2U / _pb Half-life 18.7 days 11.4 days 4.0 s 1.8ms 36.1 m 2.2 m 4.8 m stable a-energy/ MeV 6.15 5.64 6.75 7.39 6.55 β-energy (max)/MeV 1.37 1.42 Energy % 17.5 16.0 19.1 21.0 3.9 18.6 4.0

It is clear from the above decay diagram and table that more than 35 MeV of energy is released by a single decay chain of ²²⁷ Ac, presenting a significant risk of toxicity over essentially the entire half-life for any person injected with ²²⁷ Ac. As a result, the content of ²²⁷ Ac impurities in pharmaceutical ²²³ Ra is strictly limited to 45 Bq ²²⁷ Ac in 1 MBq ²²³ Ra. Thus, for practical purposes, a method that provides ²²³ Ra for pharmaceutical use should preferably provide a purity of 10 Bq ²²⁷ Ac in 1 MBq ²²³ Ra or higher to consistently meet this safety limit.

A number of studies have been published on ²²³ Ra purification, primarily in the context of the environment, where the authors attempt to accumulate ²²³ Ra from large volume samples in order to provide an analysis of the degree of environmental pollution (for example, Howitz et al., Reactive and Functional Polymers, 33 , 2536 (1997)).

There is only one previously published method that directly addresses the issue of obtaining biomedical grade ²²³ Ra, and that method is that of Larsen et al., published in WO 2000/040275. This method involves permanent absorption ^{of 227} Ac and ²²⁷ Th on the f-block of a special actinide silica gel resin (Silica Actinide Resin) containing linking groups based on P,P' dioctylmethanebisphosphonic acid on a silica support. This makes it possible to provide a relatively high frequency, less than 4x10 ^-3 % ^{of 227} As compared to ²²³ Ra, but requires a large number of manual work steps and is poorly suited for large-scale production or automation. Moreover, since the resin irreversibly sorbs the parent nucleus and the precursor nucleus, the resulting radioactive damage to the resin becomes significant if the resin is used for an industrial source (tens of years) with a lifetime of ²²⁷ As. This is especially true on an industrial scale, where isotope concentrations need to be kept as high as possible to maximize batch size and minimize manual handling.

So far, no method has been known to solve the problems of producing ²²³ Ra, such as the yield of ²²³ Ra, the speed of the purification process, automation, minimization of spent isotopes and the corresponding production or radioactive waste, or any similar problems associated with production on an industrial scale. Moreover, all known methods for producing acceptable pharmaceutical grade ²²³ Ra use specialty resins that are not guaranteed to be available and potentially difficult to claim to be reliable. Guseva et al. (Radiochemistry 46, 58-62 (2004)) proposes a basic system for producing ²²³ Ra using an ion exchange technique developed for the extraction of radium from environmental samples. However, this system has been used for very small scale production and has never been suggested or shown to provide a pharmaceutical grade material.

In light of the foregoing, there is a significant need for an improved method by which to obtain and purify ²²³ Ra for pharmaceutical use in a purity suitable for direct administration to the human subject. It would be considered an advantage if the process provided a high yield of ²²³ Ra, low losses of ²²⁷ Ac and/or ²²⁷ Th parent isotopes, and/or used a widely available separation medium. It would also be advantageous if the method were fast, suitable for relatively large (industrial scale) radioactive samples, involving only a minimum number of manual steps, and/or suitable for automation.

- 2 042622

Brief description of the invention

The applicant in this work found that by separating the initial mixture of ²²⁷ Ac/ ²²⁷ Th/ ²²³ Ra using a strong base anion exchanger followed by separation using a strong acid cation exchanger, a solution ^{of 223} Ra of very high radiochemical purity can be obtained and a number of required advantages of the method are provided.

Thus, in a first aspect of the present invention, there is provided a process for producing pharmaceutically acceptable purity ²²³ Ra, comprising:

i) obtaining the original mixture, including ²²⁷ Ac, ²²⁷ Th and ²²³ Ra;

ii) loading said feed mixture into a strong base anion resin;

iii) eluting said ²²³ Ra from said strongly basic anion resin using a first mineral acid in an alcoholic solution to obtain a first ²²³ Ra eluting solution;

iv) loading the first ²²³ Ra eluted solution into the strong acid cation exchanger and

v) elution ^{of 223} Ra from said strongly acidic cation exchanger using a second mineral acid in aqueous solution to provide a second eluted solution.

The method optionally and preferably also includes one or both of the following steps:

x) eluting said ²²⁷ Ac and ²²⁷ Th from said strongly basic anion resin using a third mineral acid in aqueous solution to recover a mixture of ²²⁷ Ac and ²²⁷ Th, and

y) storing said mixture of ²²⁷ Ac and ²²⁷ Th for a period of time sufficient to allow for an increase in ²²³ Ra through radioactive decay to re-form the original mixture of ²²⁷ Ac, ²²⁷ Th and ²²³ Ra.

Step x) is carried out at any time after step iii) of the process described above. Step y) starts immediately after the elution step iii) and can be carried out mainly on the anion exchange resin (i.e. before or without step x)) and/or after removing the ²²⁷ Ac and ²²⁷ Th mixture from the resin (i.e. after stage x)).

After growth step y) the starting mixture can be reused to produce an additional batch of ²²³ Ra and it is preferred to use one ²²⁷ Ac sample multiple times (eg more than 10 times, ie 50 to 500 times). If the mixture of ²²⁷ Ac and ²²⁷ Th does not elute from the strong base anion resin, then the process can be repeated from step iii). Preferably, however, step x) is carried out and the mixture of ²²⁷ Ac and ²²⁷ Th is eluted from the strong base anion resin. In this case, the process is repeated from step i) or step ii).

In yet another aspect of the present invention, a ²²³ Ra solution is provided comprising less than 45 Bq ²²⁷ Ac per 1 MBq ²²³ Ra, preferably a ²²³ Ra solution comprising less than 10 Bq ²²⁷ Ac per 1 MBq ²²³ Ra. Such a solution may or may not be prepared by the preferred methods described herein.

Detailed description of the invention

The most significant aspect of the present invention is the ability to desorb the initial mixture from the separating resin and re-form it with high efficiency. In particular, the present method relates to a process of long-term industrial application, while the method itself allows for multiple use of the original mixture for many years. The period of use of the initial mixture can certainly be of the order of the half-life of the resulting ²²⁷ Ac isotope, and thus possibly several tens of years (eg 10 to 50 years). There are several problems that arise as a result of this, which cannot be directed towards either the production of ²²³ Ra or the purification of the systems previously described.

The first problem arising from the potentially long shelf life of the original mixture is the stability of the ambient storage conditions. In particular, any material exposed to the original mixture probably receives more than a million beta decays per second from ²²⁷ Ac, plus about the same number of alpha decays per second from the ²²⁷ Th included, and the same number of alpha decays again from due to the increase in ²²³ Ra and due to the daughter nuclei emitting alpha particles. Which is much more concentrated than any initial system for producing ²²³ Ra proposed previously.

In particular, alpha radiation is highly ionizing and thus, over a number of years, the 10 ¹³ or more alpha decays per year that the space surrounding the original mixture will be exposed to will most likely cause significant damage to any organic components in the long term. As a result, systems such as those described in WO 2000/040275, where the feed mixture is irreversibly bonded to a separating resin, cannot be expected to be stable even when inorganic resins are used, since binders located near radionuclides are organic and susceptible to damage. This can lead to a gradual weakening of the binding ability and a possible loss of the material of the initial mixture and a decrease in the radiochemical purity ^{of 223} Ra.

In terms of potential damage due to prolonged exposure, it would be a significant advantage to be able to extract the original mixture from the separation system so that new separated material could be used intermittently. This would not only avoid wastage of the original mixture, but also ensure that the purity of the product would be as high after several disintegrations as when the system was first used. Thus, it is preferable to start the separating material generator system periodically, most preferably after each use. In the present method, this regeneration is carried out in the optional but preferred step x), which is carried out after the elution step iii) or in parallel with or after the remaining steps.

When the initial mixture is removed from the separating medium, it is important to do so with the utmost care. As noted above, a special actinide resin, the use of which is described in WO 2000/040275, does not allow you to extract the original mixture due to its irreversible sorption. This may be acceptable for laboratory or short term testing, but is a potential problem for long term industrial scale applications as described above. However, other specific materials have been proposed to separate the f-block elements from the main group elements, and they have the potential to recover the original mixture after use.

US 7,553,461 discloses a diglycomide (DGA) extractant that can be attached to a resin and used to separate f-block elements from main group elements. Unlike the previously described actinide resin, this extractant allows the initial mixture to be recovered from the f-block after separation and thus does not require constant resin stability. However, the present inventors have tested the regenerative capacity of the described DHA system and found that under conditions optimized for efficient operation, there is a loss of approximately 0.1% of the original ²²⁷ Ac and approximately 1-5% of the intermediate ²²⁷ Th isotope.

The loss of only 0.1% of the original isotope is quite negligible in any laboratory or test equipment, but in industrial systems it is an important factor. Assuming that the stock mixture is used every 3rd week (after about 72% of the possible maximum increase in ²²³ Ra) and regeneration is required 17 times a year, this results in a total loss of 12% of the stock ²²⁷ Ac over a period of 10 years. Thus, combined with natural decay due to the half-life of the isotope of 21 years, the total decrease in activity ranges from 73% due to natural decay to 61%, including regeneration losses. At 21.8 years, this effect will be even more dramatic, reducing the expected 50% potency after half-life to about 35% and clearly reducing the useful shelf life of the system at this stage.

In the present method, the regeneration of the original mixture showed that the loss is approximately 0.01% of the original ²²⁷ Ac in each regeneration cycle. These losses are ten times less than applicants could achieve with an optimized system using a diglycomide (DGA) extractant. This applies even on an industrial scale such as a 2.5 GBq batch. Considering a 3 week cycle again, this results in only approximately 1.2% loss of ²²⁷ Ac over 10 years with the present process compared to a loss of 12% in a diglycomide (DHA) extractant column even under optimized conditions. Obviously, this allows for many additional years of useful shelf life in comparison with the resin-DHA system.

Moreover, the applicants of the present invention found that approximately 99.8% of ²²⁷ Th in the original mixture can be recovered by elution from the main anion exchange resin, as described in this work. This is also significantly better than could be achieved under optimized conditions using a diglycomide (DHA) extractant, which gives a maximum 95-99% recovery ^{of 227} Th. This not only has implications for the rate of radium growth, but also increases the life of the column, and the resulting waste will be less radioactive and thus require fewer hazardous disposal sites.

In the present invention, a possible but highly preferred step x) comprises eluting said ²²⁷ Ac and ²²⁷ Th from said strongly basic anion resin using a third mineral acid in aqueous solution, thereby providing a mixture of ²²⁷ Ac and ²²⁷ Th. This step and its related components may have the following preferred features, alone or in any possible combination, and possibly in any possible combination with any features of the other steps, as described below:

a) an optional step x) can be carried out after each instance where ²²⁷ Ac is eluted from the strong base anion resin (i.e. after each stage iii)), after each such two instances, or after each such three, four or five instances: preferably stage x ) carried out between each stage iii) elution and the subsequent case of stage iii) elution;

b) the third mineral acid may be selected from H ₂ SO ₄ , HClO ₄ and HCl, preferably HCl;

c) the third mineral acid is preferably used in such an amount that all the nitro groups attached to the anion exchange resin are washed out of the resin and the ionic groups on the resin are replaced by other ions. Preferably, this mineral acid is used in excess of the amount of ionic groups on the anion exchange resin. The concentration and volume of the third mine-

- 4 042622 ralic acids will be approximately 3 M and 10 ml using a 2 ml column and desorption of 1.2 mmol/ionic groups per ml for ²²⁷ Ac and ²²⁷ Th. If HNO3 is used for regeneration, ²²⁷ Th regeneration will not be effective, but ²²⁷ Ac will be recovered;

d) in one possible embodiment, the third mineral acid in aqueous solution does not include any significant amounts (eg less than 0.1 v/v%) of any alcohol selected from methanol, ethanol and isopropanol. The aqueous solution may be substantially free of methanol;

e) eluting said ²²⁷ Ac and ²²⁷ Th from said strong base anion resin allows more than 99.9% (eg 99.9 to 100%) ^{of the 227} Ac loaded onto the resin in step ii to be regenerated; preferably, this value is greater than 99.95, about 99.99% being the most preferred;

e) eluting said ²²⁷ Ac and ²²⁷ Th from said strong base anion resin allows more than 85% (eg 85 to 99.95%) ^{of the 227} Th loaded onto the resin in step ii to be regenerated; preferably this value is more than 95% and more preferably at least 98%; about 99.8% is most preferred.

It is also important that the DHA resin described above is only an example for providing a separation efficiency of ²²³ Ra equal to 10 ² compared to ²²⁷ Ac (US 7553461, column 19 line 9). Not only was it unexpected that a simple basic anion exchange resin, when used under conditions identical to this work, allows, for comparison, to provide a separation efficiency of at least 10 ⁴ , and moreover, this separation efficiency is essential to provide an isotope with a radiochemical purity sufficient for compliance with pharmaceutical standards.

With respect to the possible but highly preferred step y), the regeneration ^{of 223} Ra will start due to natural radioactive decay as soon as the existing ²²³ Ra is eluted in step iii). It is preferable to allow sufficient time for a significant increase ^{in 223} Ra before again separating the original mixture as described above. Preferably, the regeneration of the mixture will be sufficiently efficient (as described herein) such that the levels of ²²⁷ Ac activity and ²²⁷ Th activity will be close to equality, (eg, within 5%), and in such circumstances a time period of approximately 14 to 50 days is suitable for providing an increase in ²²³ Ra. This makes it possible to provide approximately 460 to 950 MBq from a theoretical mixture of 1 GBq ^{of 227} Ac and ²²⁷ Th. If the level of ²²⁷ Th is significantly depleted due to a decrease in regeneration, this period may be longer, in particular for values of a shorter period of time. It will not be difficult for one skilled in the art to select an appropriate growth period based on the characteristics of each particular system.

The present invention provides a method for obtaining ²²³ Ra with a purity suitable for use in endo-radionuclide therapy. Below are a number of preferred features of the system, each of which may be used in combination with any other feature where technically possible, unless otherwise noted.

The methods and all corresponding embodiments of the invention are preferably carried out on an industrial scale and thus are feasible and suitable for use at this scale while maintaining all other characteristics described herein as appropriate (such as radioisotope purity, possible methanol content, etc.). ). The industrial scale is usually a larger scale than is required for the treatment of one object and, for example, may be the purification of more than 10, preferably more than 25 and most preferably more than 45 conventional doses of ²²³ Ra. Obviously, the usual dose will depend on the application, but the intended dose may be from 0.5 to 100 MBq, preferably from 1 to 50 MBq, most preferably from about 2 to 25 MBq.

Step i) of the method of the invention relates to the preparation of an initial mixture comprising ²²⁷ Ac, ²²⁷ Th and ²²³ Ra. Such a mixture is essentially formed by the gradual decay of the ²²⁷ Ac sample, but one or more of the following features are also preferred when used in the invention, either alone or in any possible combination:

a) ^{a 227} Ac radioactivity of at least 500 MBq (eg 500 MBq to 50 GBq), preferably at least 1 GBq, more preferably at least 2.5 GBq;

b) ^{a 223} Ra radioactivity of at least 100 MBq (eg 100 MBq to 50 GBq), preferably at least 350 MBq, more preferably at least 1 GBq;

c) a volume of not more than 50 ml (eg 0.1 to 50 ml), preferably not more than 10 ml, more preferably not more than 5 ml.

Step ii) of the method of the invention relates to loading the feed mixture onto the strong base anion resin. This stage and the components related to it may have the following preferred features, either alone or in any possible combination, and possibly in any in any possible combination with any features of other stages as described in this work:

a) the strong base anion resin may be a polystyrene-divinylbenzene copolymer resin, preferably containing 1-95%; divinylbenzene;

b) Strong base anion resin can be RN ⁺ Me ₃ resin (type I) or R- resin

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N ⁺ Me ₂ CH ₂ CH ₂ OH (type II), preferably resin type I;

c) The strongly basic anion exchange resin may have an exchange capacity of 0.2 to 5 meq/ml, preferably 0.6 to 3 meq/ml, most preferably 1 to 1.5 meq/ml (e.g., approximately 1.2 meq/ml);

d) the strong base anion resin may have a particle size gradation of 10 to 800 mesh, preferably 50 to 600 mesh, more preferably 100 to 500 mesh (eg about 200 to 400 mesh);

e) a strong base anion resin can be used in the form of a column;

e) the volume of resin used (e.g. packed into a column) may be 10 ml or less (e.g. 0.5 to 10 ml), preferably 5 ml or less, more preferably 1 to 2.5 ml (e.g., approximately 2 ml);

g) The strong base anion resin can be DOWEX 1X8 (eg DOWEX AG 1X8) or an equivalent 200-400 mesh resin.

Step iii) of the method of the invention relates to the elution of ²²³ Ra from said strongly basic anion resin using a first mineral acid in an alcoholic solution to obtain a first ²²³ Ra eluted solution. This stage and the components related to it may have the following preferred features, either alone or in any possible combination and possibly in any possible combination with any features of other stages as described in this work:

a) the first mineral acid may be selected from H ₂ SO ₄ or HNO 3 , preferably HNO ₃ .

b) the first mineral acid can be used in a concentration of 0.1 to 5 M, such as 0.1 to 3 M or 0.1 to 1.5 M, preferably 0.3 to 0.8 M, more preferably 0.45 to 0.55 M (eg about 0.5 M), especially if the first mineral acid is HNO3; or alternatively b') the first mineral acid may be used in a concentration of 0.1 to 5 M, such as 0.1 to 3 M or 0.1 to 1.5 M, preferably 0.1 to 1.5 M , more preferably from 1.25 to 1.35 M (eg, about 1.3 M), especially if the first mineral acid is HNO ₃ ;

c) the alcoholic solution may include at least one alcohol selected from methanol, ethanol and isopropanol, preferably methanol;

d) the alcoholic solution may comprise 20 to 95% methanol, eg 75 to 90% methanol, more preferably 83 to 87% methanol (eg about 85% methanol); or alternatively d') the alcoholic aqueous solution may comprise 20 to 95% methanol, eg 75 to 90% methanol, more preferably 79 to 84% methanol (eg about 81% methanol);

e) the alcoholic solution may include about 0.5M HNO3 in about 85% aqueous methanol; or alternatively e') the alcoholic solution may include about 1.3M HNO3 in about 81% aqueous methanol;

e) ²²³ Ra can be eluted from said strong base anion resin using 10 to 100 column volumes of the first mineral acid in an aqueous alcohol solution. The preferred amount is 15 to 50 column volumes, more preferably 20 to 40 column volumes (eg, about 30 column volumes);

g) the first eluted solution preferably has an impurity level of not more than 100 (for example, from 1 to 100) Bq ²²⁷ Ac per 1 MBq ²²³ Ra, more preferably not more than 45 Bq ²²⁷ Ac per 1 MBq ²²³ Ra (for example, not more than 30) and most preferably not more than 10 Bq ²²⁷ As per 1 MBq ²²³ Ra;

h) at the stages of loading the initial mixture onto a strongly basic anion exchanger and elution to obtain the first eluted solution ^{of 223} Ra, a separation ratio determined by the ratio of ²²³ Ra to ²²⁷ As is provided at least 10,000:1 (for example, from 10,000:1 to 500,000:1), preferably at least 20,000:1, more preferably at least 30,000:1;

i) ²²³ Ra can be eluted from said strongly basic anion exchanger not in complex form, but for example in the form of a simple salt in solution (eg as a salt of a first mineral acid);

j) possibly the use of complexing agents such as DTPA (diethylenetriaminepentaacetic acid) can be avoided and in one embodiment all solutions used in step m and/or iii are substantially free of such complexing agents such as DTPA.

Step iv) of the process of the invention relates to loading ^{the 223} Ra eluted from the anion exchange resin onto the strong acid cation exchange resin. This step and the constituents related thereto may have the following preferred features, either alone or in any possible combination, and possibly in any possible combination with any of the features of the other steps as described herein:

a) the strong acid cation exchange resin may be a polystyrene-divinylbenzene copolymer resin, preferably containing 1-95% DVB;

b) the strongly acidic cation exchanger may be a SO ₃ H type resin;

c) a strongly acidic cation exchanger can have an exchange capacity from 0.2 to 5 meq / ml, preferably

- 6 042622 specifically from 0.6 to 3 meq/ml, most preferably from 1 to 2 meq/ml (for example, approximately

1.7 meq/ml);

d) The strong acid cation resin may have a particle size gradation of 10 to 800 mesh, preferably 50 to 600 mesh, more preferably 100 to 500 mesh (e.g., about 200 to

400 mesh);

e) a strong acid cation exchanger can be used in the form of a column;

e) the volume of resin used (e.g., packed into a column) may be 5 ml or less (e.g., 0.1 to 5 ml), preferably 2 ml or less, more preferably 0.2 to 1 ml (e.g., approximately 0.5 ml);

g) The strong acid cation resin can be DOWEX 50X8 or an equivalent 200-400 mesh resin.

Step v) of the method of the invention relates to the elution of ²²³ Ra from the strong base anion exchanger using a second mineral acid in aqueous solution to provide a second eluted solution. This step and the constituents related to it may have the following preferred features, either alone or in any possible combination and possibly in any possible combination with any of the features of the other steps as described herein:

a) the second mineral acid may be selected from H ₂ SO ₄ , HNO ₃ and HCl, preferably HNO 3 ;

b) the second mineral acid may be used in a concentration of 0.5 to 5 M, preferably 1 to 2 M, more preferably 1.0 to 1.6 M (eg about 1.3 M), in particular when the second the mineral acid is HNO3;

c) the aqueous solution preferably does not contain any significant amounts (eg, contains less than 0.1 vol./vol.%) of any alcohol selected from methanol, ethanol and isopropanol, preferably the aqueous solution is essentially free of methanol;

d) ²²³ Ra can be eluted from said strong acid cation exchanger using 10 to 100 column volumes of the second mineral acid in aqueous solution; preferably the amount is from 15 to 80 column volumes, more preferably from 30 to 60 column volumes;

g) the second eluted solution preferably contains an impurity level of not more than 1 (for example, from 0.0001 to 1) Bq ²²⁷ Ac per 1 MBq ²²³ Ra, more preferably not more than 0.1 Bq ²²⁷ Ac per 1 MBq ²²³ Ra and most preferably not more than 0.05 Bq ²²⁷ As per 1 MBq ²²³ Ra;

h) the second eluted solution preferably has a methanol content of at most 30 mg per ²²³ Ra dose, preferably at most 10 mg and more preferably at most 2 mg per dose;

i) in the steps of loading the first eluted solution onto the acidic cation exchanger and eluting to obtain a second ²²³ Ra eluting solution, a separation ratio defined by the ratio of ²²³ Ra to ²²⁷ Ac can be provided at least 10:1 (for example, from 10:1 to 10000:1 ), preferably at least 100:1, more preferably at least 500:1;

j) ²²³ Ra can be extracted from said strong acid cation exchanger not in a complex form, but for example in the form of a simple salt in solution (for example, in the form of a salt of the first mineral acid);

k) it is possible that the use of complexing agents such as DTPA (diethylenetriaminepentaacetic acid) can be avoided, and in one embodiment, all solutions used in step iv and/or v are substantially free of complexing agents.

In addition to these steps, the methods of the invention and all related aspects may include additional steps, such as confirming the purity of ²²³ Ra for pharmaceutical use, exchange of counterions, concentration and dilution of the solution, or control of factors such as pH and ionic strength of the solution. Thus, each of these steps forms a possible but preferred additional step in various aspects of the present invention.

In the examples below, it is determined that, with appropriate process optimization, it is possible to achieve ²²³ Ra purification to pharmaceutical grade radiochemical purity (see above, all of which are included in the disclosure where context permits) using only the strong base anion step. Thus, in an alternative aspect of the present invention, there is provided a process for producing ²²³ Ra of pharmaceutically acceptable purity, comprising:

i) obtaining the original mixture, including ²²⁷ Ac, ²²⁷ Th and ²²³ Ra;

ii) loading said feed mixture into a strong base anion resin;

iii) eluting ²²³ Ra from said strong base anion resin using a first mineral acid in an alcoholic solution to obtain a first ²²³ Ra eluting solution;

Obviously, all stages i)-iii) are preferably carried out as described in this work, and all relevant preferred aspects and embodiments described in this work are included in the invention carried out by this alternative method. The remaining steps described in this paper may possibly be used in addition to these three main steps, and alternatively or additionally, any combination of additional verification or other manipulation steps can also be done.

- 7 042622

The invention is further illustrated by reference to the following non-limiting examples and the accompanying drawings, in which:

in fig. 1 shows a continuous chromatographic process for obtaining pure radium-223 from an actinium-227 source, as described in this paper. Pharmaceutical grade ²²³ Ra is recovered from the cation exchange column;

in fig. 2 shows the device of FIG. 1 with additional capture and purification steps to confirm ²²³ Ra purity and adjust counter ions and other solution properties.

Examples

Review.

The method for separating ²²³ Ra from ²²⁷ Ac/ ²²⁷ Th of the original system was carried out as shown in FIG. 1. More specifically, the ²²⁷ Ac/ ²²⁷ Th storage vessel is designed to be loaded onto a strong base anion resin. The elution solvent is initially chosen to elute ^{the 223} Ra selectively while retaining ^{the 227} Th and ²²⁷ Ac in the anion column. These isotopes are then eluted with additional solvent and returned to the original mixture storage tank. ^{The 223} Ra eluted from the strong base anion column is then evaporated and/or resuspended as needed before being loaded into the strong acid cation column. Again, ²²³ Ra is selectively eluted using a second elution solvent to obtain ²²³ Ra of radioisotope purity suitable for pharmaceutical use. The purified ²²³ Ra is then optionally subjected to various purification and testing steps to confirm that the final solution is suitable for pharmaceutical purposes.

Each of the key steps in the above embodiment has been optimized by the applicants in order to provide a very high degree of purity of the product obtained on an industrial scale during the process with a minimum number of manual steps. It was only after optimization that it became apparent that a product of pharmaceutical grade and higher could be obtained with such a small number of simple, more automated steps.

Example 1 - anion exchange column.

1.1. The output of radium-223 from the anion column.

During the development of a procedure for the purification of radium 223, the yield of radium 223 in the process is an important issue. To be able to determine the amount of solution to be used to obtain a high yield of radium-223, pure radium-223 was loaded into a 2.0 ml column packed with 200-400 mesh Dowex 1-X8 anion resin particles.

Radium-223 was eluted from three different columns, respectively 80; 82.5 and 85% methanol. The HNO3 concentration was 0.5 M in all three experiments. The volume required to obtain approximately 95% yield of radium-223 in a given chromatographic step was determined.

In FIG. 1 shows the corresponding experimental setup for determining the yield of radium-223 from a 2 ml anion column packed with 200-400 mesh Dowex AG1-X8 anion resin particles. All radium-223 was diluted with 2 ml of elution fluid and loaded onto the column. The column was then washed with the same methanol/HNO ₃ solution.

In FIG. 2 shows the elution profile of radium-223 from a 2 ml anion column packed with 200-400 mesh AG1-X8 anion resin particles. Three different concentrations of methanol in 0.5 M HNO ₃ are shown.

It should be noted that column washes are also included in the drawing to obtain total amounts of radium-223.

Comments.

Radium-223 elutes from the anion exchange resin more slowly as the concentration of methanol increases.

1.2. Leakage of actinium-227 from the anion column.

To retain most of the actinium-227 on the anion exchange resin, initial experiments indicated that the methanol level should be 80% or higher. This is true when the concentration of nitric acid is 0.5 M. If the concentration of nitric acid is increased, the concentration of methanol can be lowered and the same result can be achieved.

Approximately 25% of atkinium-227 leaks into the radium-223 eluate from the anion column using 10 ml of 70% methanol in combination with 0.5 M HNO3. (see Table 1). If the concentration of methanol was increased to 80%, less than 0.2% actinium-227 leaked through the anion column using the same amount of eluent (10 ml).

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Table 1

Actinium-227 content in radium-223 eluate from a 2 ml anion column packed with 200-400 mesh Dowex 1-X8 particles

eluent The content of actinium-227 in the eluate, in % of the total load of actinium-227 10ml 70% methanol/0.5 Μ HNO ₃ 23.8 10ml 75% methanol/0.5 Μ HNO ₃ 1.3 10ml 80% methanol/0.5 Μ HNO ₃ < 0.2

According to these results, it is clear that the leakage of actinium-227 from the resin depended on the concentration of methanol. It is also apparent that a fairly small difference in methanol concentration has a large effect on actinium-227 leakage.

As can be seen from the table above, the concentration of methanol, when used in combination with 0.5M HNO3, must be 80% or more if all of the actinium-2277 is to be retained. Based on this information, the concentration of methanol was set to 80-85% in combination with 0.5 M HNO3 in the experiment for the study.

In FIG. Figure 3 shows a setup for assessing leakage of actinium-227 from anionic columns. All columns were packed with 200-400 mesh Dowex AG1-X8 particles. Working conditions were as follows: 0.5 M HNO3 with 80; 82.5 and 85% methanol, respectively. In all experiments, we used the initial mixture with actinium 180 MBq. All actinium-227 samples were diluted in 2 ml of eluent and loaded onto the column. The column was then washed with the same methanol/HNO ₃ solution.

table 2

The amount of actinium-227 measured in the eluate (cation column) after loading the actinium-227 source into a 2 ml column packed with _______ Dowex AG1-X8 200-400 mesh particles_______

eluent *Volume of eluent Leakage of actinium-227 eluate in Bq or % of the total load of actinium-227 **Separation ratio (actinium-227 captured in column / actinium-227 in eluate) 80% methanol 0.5 Μ HNO ₃ 35 5800 Bq/(0.0032%) 3.1Yu ⁴ 82.5% methanol 0.5 Μ HNO ₃ 50 280 Bq/(0.00016%) 6.4Yu ^b 85% methanol 0.5 Μ HNO ₃ 60 385 Bq/(0.00021%) 4.7Ί0 ^ϋ

* To achieve a 95% yield of radium-223 in the process, the volume of eluent used in this experiment is changed.

* * Measurements are carried out 40-50 days after separation.

The discussion of the results.

The results show that the combination of a larger volume and a higher concentration of methanol reduces leakage of actinium-227. The volumes used in this experiment are based on volumes that yield approximately 95% yield of radium-223.

It was observed that most of the actinium-227 was retained in the column. In order to fulfill specifications regarding the concentration of actinium-227 in the drug substance, the separation ratio, expressed as the ratio of retained actinium-227 and leaked actinium-227, should be approximately 3.1 x10 ⁴ . This criterion is fulfilled for all experiments. At 85% methanol, the criterion is met with a factor of 15.

Leakage of actinium-227 from the anion column is so low that it does not affect the recovery of actinium-227 in the process. This separation step is the first and most important step in the entire chromatographic process. This shows that it is important to ensure strict control of the methanol concentration when preparing the eluent. This is important to ensure low leakage of actinium227 from the column and thus the methanol concentration must be accurately correlated with the desired HNO3 concentration.

It is possible to use higher concentrations of methanol to obtain even better separation. However, the volumes will increase and the separation time and volume of waste will also increase.

Example 2 Cation exchange column.

The main purpose of this chromatographic stage is:

capture of radium-223 from chromatographic stage 1, removal of most of the methanol used in the first stage of separation, further purification / post-purification of radium-223 from actinium-227.

2.1. Trapping radium-223.

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In the development of the present method, it has been shown that appropriate cation exchange columns allow the binding of radium-223 when the molarity of the acid is in the range of 1 M or less. Dowex 50W-X8 cation exchange medium shows an additional increased affinity for radium-223 when methanol is present in the eluent.

During the development of this method, it was confirmed that the affinity of radium-223 for the corresponding cation exchange resin is high. When 60 ml of the radium-223 eluate is pumped through a 0.5 ml cation exchange column of 85% methanol/0.5 M HNO ₃ , all of the radium-223 is trapped.

2.2. Removal of methanol.

Since methanol is a class 2 solvent, it is preferable to keep the amount of methanol in the drug substance as low as possible. ADI (acceptable daily intake) for this solvent should not exceed 30 mg/day. The concentration of methanol in the medicinal liquid must also be less than 3000 ppm (European Agency for Evaluation of Medicines, EMA).

Approximately 99.75% of the methanol used in this process is removed simply by passing the eluent through a discard column. This represents the main methanol removal step in the process. Because the cation resin has a low volume content (eg, only 0.5 ml), the amount of methanol remaining on a given column after a given chromatographic step will be low.

It is estimated that a 0.5 ml column packed with 200-400 mesh Dowex 50W-X8 cation exchange resin will contain approximately 0.15 ml liquid or approximately 100 mg methanol when the column is dried. If all the methanol was transferred to the drug substance, more than 4 doses would have to be obtained from this eluate to meet the DDI of 30 mg/day. The eluate must also be diluted so that the concentration of methanol is less than 3000 ppm, as stated by EALS ref. 5.

2.3. Output of radium-223.

In order to provide an efficient process for purifying radium-223, it is important that the losses of radium-223 at different stages of the process are as low as possible. An experiment was performed to confirm the amounts of eluent needed to achieve 95% yield of radium-223 from a 0.5 ml cation exchange column. The cation exchange column was packed with 200-400 mesh Dowex 50W-X8 particles. The estimated HNO ₃ concentrations were 1.3; 1.45 and 1.6 M.

In FIG. 5 shows an experimental setup for the extraction of radium-223 in a 0.5 ml column packed with 200-400 mesh Dowex 50W-X8 particles. The eluents tested were 1.3M, 1.45M and 1.6M HNO3.

In FIG. 6 shows the elution profile of radium-223 from the experimental setup shown in FIG. 5.

Comments.

An increase in the affinity of radium-223 for resin was found at lower concentrations of HNO ₃ . The volume required to elute, for example, 95% radium-223 on the column, increases as the concentration of HNO3 decreases.

Table 3

Cumulative output of radium-223 from a small cationite column packed with Dowex 50W-X8 particles (200-400 mesh) at different concentrations of HNO ₃ . This corresponds to the data shown in Fig. 6

Radium-223 yield in % (cumulative) Radium-223 yield in % (cumulative) Radium-223 yield in % (cumulative) Volume 1.3 M HNO ₃ 1.45 _HNO3 1.6 _HNO3 0-4 ml 0.3 1.3 3.7 4-8 ml 8.3 49.9 49.2 8-12 ml 31.3 78.5 87.2 12-16 ml 52.3 91.6 98.0 16-20 ml 70.1 96.7 99.7 20-24 ml 82.8 98.7 24-28 ml 92.4 99.2 28-32 ml 96.4 Column 100 100.0 100.0

Comments.

Based on the results obtained, the elution profiles of radium-223 from a 0.5 ml column packed with Dowex 50W-X8 (200-400 mesh) particles at 1.3 M, 1.45 M and 1.6 M HNO ₃ are presented. There is a noticeable difference between the estimated lowest and highest concentrations of HNO ₃ . To achieve approximately 95% yield of radium-223, the following volumes and concentrations of HNO ₃ were used:

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1.30 M HNO ₃ : 32 ml,

1.45 M HNO ₃ : 20 ml,

1.60 M HNO3: 16 ml.

In order to be able to select the concentration of HNO ₃ in the process, it is necessary to investigate for the above investigated concentrations of HNO ₃ the separation ratio determined by the ratio of radium-223 to actinium-227. This experiment is presented below.

2.4. The separation ratio is radium-223/actinium-227.

The preceding examples regarding volume and concentration of HNO ₃ should be used to elute 95% radium-223 from a 0.5 ml column packed with Dowex 50W-X8 (200400 mesh) particles. It is also interesting to confirm the separation relation between actinium-227 and radium-223 under the same conditions.

The setup in this experiment is similar to the setup shown in Fig. 5, but with actinium-227 as a reference.

A 0.5 MBq actinium-227 sample in equilibrium with its daughter nuclei was loaded onto a 0.5 ml column.

The volumes and concentrations of HNO ₃ in this experiment are identical to those that give a 95% yield of radium-223 in the previous example.

The separation ratio obtained for radium-223 and actinium-227 is shown in FIG. 7 and in table. 4. Actinium-227 measurements are based on the growth of actinium-227's daughter nucleus, thorium-227.

In FIG. 7 shows the concentrations of HNO ₃ and the separation ratio between radium-223 and actinium-227 (Bq) in the eluate from a 0.5 ml cation exchange column packed with Dowex 50W-X8 (200400 mesh) particles.

Comments.

A linear relationship has been established between the separation ratio of radium-223 and actinium-227 and the molarity of HNO3 used.

In table. 4 shows the concentrations and volumes, as well as the separation ratio of radium-223 and actinium-227 (Bq) in the eluate from a 0.5 ml cation exchange column packed with Dowex 50W-X8 cation exchange resin particles (200-400 mesh).

Table 4

HNO ₃ concentration (M) ‘Volumes of eluent used (ml) Radium-223 Bq/actinium- 227 Bq in eluate 1.3 32 787 1.45 20 379 1.6 16 52

*Volumes used yielding 95% radium-223 from 0.5 ml cation resin. Conclusion.

The separation of radium-223 and actinium-227 on the cation resin increases with increasing molarity of HNO3. The best separation of radium-223 and actinium-227 in this case was achieved at 1.3 M HNO3. According to these results, the conditions chosen for the chromatographic step are: 1.3 M HNO ₃ and 30-35 ml eluent. Then the yield of radium-223 is approximately 95%. The separation ratio of radium-223 and actinium-227 in Bq is close to 800.

It is probably possible to use lower concentrations of acid to obtain an even better separation separation. But then the volumes will increase and the separation time and waste volume will also increase.

2.6. The purity of the eluate with radium.

If 85% methanol/0.5 M HNO ₃ is used in the anion exchange chromatography step, then the separation ratio of radium-223 and actinium-227 is approximately 4.7-105. In the cation exchange chromatography step, the separation ratio of radium-223 and actinium-227 is approximately 750 under appropriate conditions.

The total separation ratio, in Bq, of these two nuclides will be in the range: 4.7x105x750=3.5x108

3. Regeneration of actinium-227 and thorium-227.

The main purpose of this stage is the recovery of actinium-227 and thorium-227 from the separation medium for further use in a repetitive process.

Washing out the actinium-227 and thorium-227 from the anion resin back into the feed tank is an important issue. Losses of actinium-227 have a direct impact on the amount of output of radium-223 over time. Reducing the loss of thorium-227 is also important, since here the loss will give a lower gain of radium-223 from the original mixture.

In FIG. Figure 8 shows a scheme for washing actinium-227 and thorium-227 back into the container of the initial mixture.

Comments.

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Thorium-227 and actinium-227 are most likely present as a nitrate complex in the anion exchange column. Such a complex should be removed so that thorium-227 and actinium-227 can be recycled.

Using 10 ml of 3M HCl, Cl ^- is replaced by nitrate as a counterion. When all this is done, the actinium/thorium nitrate complex will no longer be present on the resin. Then actinium-227 and thorium-227 have no affinity for the resin and are eluted. There are several other factors contributing to the efficient desorption of actinium-227 and thorium-227.

1. The density of 3 M HCl is higher than the density of the methanol solution present in the particle when the desorption procedure is started. This density difference will contribute to a more efficient particle washing procedure.

2. Thorium-227 and actinium-227 also have the same charge as the ionic groups on the resin; this is also an advantage for the efficient recycling of nuclides.

3. The particle size used in this case is relatively small. A small particle size is preferred for an efficient washing procedure.

4. It is also important that the flow rate is lower (1-2 ml/min in this case) than the rate used in the separation process, so that the contact time during the desorption procedure increases. The recommended flow rate is less than half the separation rate (eg, 0.5 ml/min or less).

Evaporation of 10 ml of HCl must be carried out before the subsequent separation/elution of radium-223 from the original mixture.

Example 3 - industrial production of a drug substance from a source of actinium-227 2.5 GBq.

After the initial experiments, an experiment was carried out on an industrial scale. The initial mixture contained approximately 2.5 GBq actinium-227. The amount of radium-223 in the batch during separation was estimated to be 1.2 GBq.

In FIG. 9 shows a setup for an industrial scale experiment to obtain a drug substance.

Comments.

The separation was carried out and the results confirmed the assumptions. The method made it possible to obtain a medicinal substance in the amount of approximately 1100 MBq of radium-223. This corresponds to an overall process yield of 92% since the total amount of radium-223 in the feed mixture at separation time is estimated to be 1.2 GBq. The purity was determined to meet the limits of pharmaceutical control requirements, and recovery of the parent isotope was high, as detailed below.

Measurements with HDVCH (High Purity Germanium Detector).

In table. 5 shows the levels of thorium-227, radium-223 and actinium-227 in various fractions/columns from the lot. Measurements were taken at different times after separation. It is likely that the level given for actinium-227 is overestimated, since complete decay of thorium-227 has not been obtained.

In table. 5 shows the concentrations of actinium-227, thorium-227 and radium-223 in various fractions from columns in a commercial batch.

Table 5

Radium-223 concentration on the day of separation, calculated in Bq* Thorium-227 concentration on the day of separation, calculation in Bq** Actinium-227 concentration on the day of separation, calculation in Bq*** Anion column <2360 1.86-10 ⁶ 2.96Ί O ⁵ **** Small cation column 0.5 ml 1.9Yu ⁷ 9928 619 Radium eluate 1.1-10" No measurements No sizes

* The radium concentration was calculated from measurements on the 13th day after separation.

**Thorium-227 is calculated based on the fact that all measurements of thorium-227 arise from the decay of thorium-227. Calculations are based on measurements 77-80 days after separation.

*** Actinium-227 concentrations are based on all thorium-227 originating from actinium-227. Calculations are based on measurements 77 days after separation. During this time, 92.5% of the increase in thorium227 from actinium-227 occurs.

**** The concentration of actinium-227 on the anion exchanger is calculated from the level of thorium-227 measured 126 knocks after separation.

Actinium-227 losses.

The most important thing in the method is to reduce the loss of actinium-227 to a minimum. The amount of actinium-227 remaining in the anion exchange column was calculated from measurements of the level of thorium-227 on

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Claims (2)

126 days after separation. According to these measurements, the amount of actinium-227 remaining in the column is approximately 3x10 ⁵ Bq, or 0.3 MBq. The column load was 2500 MBq. Based on these figures, the loss is approximately 0.012%. No significant actinium-227 was observed in the other columns/solutions in the process. Regeneration of thorium-227. Approximately 1.8 x 10 ⁶ Bq of thorium-227 was measured on the anion exchange column after desorption. No significant level of thorium-227 was measured in any other column or solutions. Based on these data, the recovery of thorium-227 in the process is over 99.5%. Loss of radium-223 in the process. In accordance with the development of the method, approximately 95% of the radium-223 should be eluted in the first chromatographic step from the anion exchange column. Such a yield cannot be measured directly, since the remaining radium-223 in the anion-exchange column is washed back into the container of the initial mixture along with actinium-227 and thorium-227. It is possible to measure the content of radium-223 in all columns, as well as liquid fractions used in the method. In table. 6 shows measurements on different liquid fractions and columns. The loss of radium-223 at various stages is also calculated as a percentage of the total amount of radium-223 produced. In table. 6 shows the % radium-223 in different fractions/columns on the day of separation. The results are calculated based on germanium measurements on the 13th day after separation. Table 6 Columns/Fractions Radium-223 relative to the total amount of radium-223 in the drug substance (%) Anion column before HCI wash No measurements Anion column after HCI wash 0.000 Small cation column 0.5 ml 1.685 Waste 1 0.000 Conclusion. The loss of radium-223 in the form of waste and in the column is low. As expected, large losses in a small cation column. The method is effective in obtaining radium-223 with a high yield. Measurement of the second (cation exchange) resin. The content of actinium-227 in the eluate from a 0.5 ml cation exchange column can be calculated. This calculation is based on the knowledge that 750 Bq of actinium-227 remains in a small cation exchange column for every 1 Bq of actinium-227 eluted. This ratio is approximately 750 for 30 ml of 1.3 M HNO3. After 77 days, the percentage increase in thorium-227 from actinium-227 is 92%. The amount of thorium-227 measured on the cation exchange resin was less than 572 Bq. If all thorium-227 originates from actinium227, which is the worst-case scenario, the maximum amount of actinium-227 on the column is: 572 Bq/0.92 = 619 Bq of actinium-227. The total amount of actinium-227 in a batch of 1100 MBq of radium-223 eluate from the cation exchange column will be: Level of actinium-227 measured on the column / Separation ratio of retained actinium-227 / eluted actinium-227 in the column: 619 Bq / 750 = 0.82 Bq. The final purity of the drug substance. The amount of radium-223 in the drug substance: 1100 MBq. The amount of actinium-227 in the eluate: 0.82 Bq. Bq of actinium-227/MBq of radium-223 = 0.82 Bq/1100 MBq = 0.00075. Characteristics: 45 Bq of actinium-227 per MBq of radium-223: 45 Bq/MBq. Characteristics satisfy the coefficient 45 / 0.00075 = 60000 CLAIM 1 . ²²³ Ra, including less than 0.05 Bq ²²⁷ Ac per 1 MBq ²²³ Ra. 2. ²²³ Ra according to claim 1, obtained or obtained by a method including: i) obtaining the original mixture, including ²²⁷ Ac, ²²⁷ Th and ²²³ Ra; ii) loading said feed mixture into a strong base anion resin; iii) eluting said ²²³ Ra from said strongly basic anion resin using a first mineral acid in an alcoholic solution to obtain a first ²²³ Ra eluting solution; iv) loading ²²³ Ra from the first eluted ²²³ Ra solution into the strong acid cation exchanger; And v) elution ^{of 223} Ra from said strongly acidic cation exchange resin using a second mini- -