METHOD FOR PRODUCING ACTININIUM-225 FROM RADIUM-226

Abstract

Method for the manufacture of Radium-225-containing material from Radium-226-containing materials by subjecting a starting material containing Radium-226 to neutron irradiation from a nuclear reactor to convert .sup.226Ra into Radium-225 to provide a Radium-225-containing material, characterised in that the neutron irradiation of Radium-226-containing starting material is performed in a moderated nuclear reactor; and the Radium-226-containing starting material is shielded with a thermal neutron absorption shield.

Claims

1. A method for the manufacture of Radium-225 containing material from Radium-226 containing materials, comprising: (a) subjecting a starting material containing Radium-226 to neutron irradiation from a nuclear reactor to convert Radium-226 into Radium-225 to provide a Radium-225 containing material, wherein: (i) the neutron irradiation of Radium-226 containing starting material is performed in a moderated nuclear reactor; (ii) the Radium-226 containing starting material is shielded with a thermal neutron absorption shield; (b) allowing the Radium-225 to decay into Actininium-225; and (c) isolating the Actininium-225 from the Radium-225 containing material.

2. The method according to claim 1, further comprising isolating radium isotopes from the Radium-225 containing material.

3. The method according to claim 1, wherein the moderated nuclear reactor is a moderated material test reactor.

4. The method according to claim 1, wherein the moderated nuclear reactor is a moderated high flux reactor.

5. The method according to claim 1, wherein the thermal neutron absorption shield comprises a material selected from the group consisting of boron, cadmium, gadolinium, hafnium and materials containing these elements and mixtures thereof.

6. The method according to claim 1, wherein the thermal neutron absorption shield comprises gadolinium.

7. The method according to claim 1, wherein the fast neutrons from the moderated nuclear reactor have an energy from about 0.1 MeV to about 20 MeV.

8. The method according to claim 7, wherein the fast neutrons from the moderated nuclear reactor have an energy from about 5 MeV to about 20 MeV.

9. The method according to claim 1, wherein the thermal neutrons have an energy from about 20 eV to about 1000 eV.

10. The method according to claim 1, wherein the Radium-226 containing material is provided as a metal, as oxide, as salt, halide, nitrate or carbonate, or mixtures thereof.

11. A method for the manufacture of an Actininium-225 containing material, comprising subjecting a Radium-226 containing starting material to neutron irradiation to convert Radium-226 into a Radium-225- and Actininium-225 containing material, wherein (i) the neutron irradiation of Radium-226 containing starting material is performed in a moderated nuclear reactor; and (ii) the Radium-226 containing starting material is shielded from thermal neutrons with a thermal neutron absorption shield; and (iii) the Radium-225 containing material is allowed to decay into Actininium-225 containing material.

12. The method according to claim 11, wherein the radium isotopes are isolated from the Radium-225 containing material prior to allowing the decay into the Actininium-225 containing material.

Description

DESCRIPTION OF THE DRAWINGS

[0014] FIG. 1 discloses a schematic representation of an example of a containment cylinder that holds the thermal neutron shielded radium target material.

[0015] FIG. 2 shows the quantities of .sup.225AC and .sup.225Ra over time, during and after irradiation in shielded and unshielded form

DETAILED DESCRIPTION OF THE INVENTION

[0016] The inventors have found that instead of inside a FNR, Radium-225 and Actininium-225 can be produced in high yield and selectivity when a ‘normal’ moderated material test reactor is used.

[0017] A moderated material test reactor is generating a neutron flux that contains fast neutrons (having an energy in the MeV spectrum) and (epi-) thermal neutrons (having an energy in the <KeV spectrum). The presence of thermal neutrons in the neutron spectrum of a moderated material test reactor present problems in the efficient generation of .sup.225AC since the thermal neutrons are capable of generating a variety of other isotopes in varying amounts (i.e. generates low selectivity towards the desired .sup.225Ra, thus complicating workup of the desired isotope products and burn-up of the .sup.225Ra starting material i.e. lowering yield of the desired .sup.225Ra).

[0018] From initial experimentation with the High Flux Reactor Petten (HFR), the .sup.225Ra generation rate by .sup.226Ra (n,2n) is found to be higher than hitherto known. This offers the opportunity to produce .sup.225Ra, as a source of .sup.225AC, using the fast neutron part of the neutron flux in a ‘normal’ material test reactor such as the HFR. Common moderated material test reactors such as the HFR, generally have a lower fast neutron flux than most FNR's, but the post-irradiation experiments of .sup.226Ra irradiated in the HFR show that the fast flux in a material test reactor is sufficient to produce significant amounts of .sup.225Ac.

[0019] The present inventors realised that an efficient way of making .sup.225Ac from .sup.226Ra was when only the fast neutron part of the neutron flux was used and the thermal neutron activation was (substantially) eliminated. The inventors realised that this could be achieved by using a thermal neutron absorption shield around and preferably the Radium-226 target.

[0020] The shield absorbs (a large portion of) the thermal neutrons and allows the larger part of the fast neutrons to pass and interact with the target material (.sup.226Ra).

[0021] The thermal shield is preferably formed such that the spectrum is shaped such that only or substantially neutrons that have the desired energy to convert .sup.226Ra to .sup.225Ra are not absorbed and interact with the target material. The preferred neutron energy window for the conversion of .sup.226Ra to .sup.225Ra lies in the range of from 0.1 MeV, preferably 5 MeV to 20 MeV.

[0022] The invention thus relates to the manufacture of Radium-225 containing material from Radium-226 containing materials by subjecting a starting material containing Radium-226 to neutron irradiation from (inside or near) a moderated material test nuclear reactor to convert Radium-226 into Radium-225 to provide a Radium-225 containing material, characterised in that [0023] the neutron irradiation of Radium-226-containing starting material is performed in a moderated nuclear reactor; and [0024] the Radium-226 containing starting material is shielded with a thermal neutron absorption shield.

[0025] It was found that the use of thermal neutron absorption shield during the irradiation of Radium-226 offers the following benefits: [0026] reduction of the amount and variety of highly radioactive isotopes generated by thermal neutron activation of Radium-226. Some of the highly radioactive isotopes produce high energy gamma radiation which is difficult to shield and complicate handling and processing after irradiation. The thermal neutron shield effectively reduces the formation of these particular isotopes and thereby facilitates greatly the feasibility of handling, processing and purification of the target material after irradiation. [0027] reduction of the amount and variety of isotopes generated by thermal neutron activation of Radium-226. This reduces the generation of unwanted isotopes by thermal neutron activation that may end up in the final product (improved product quality) and reduces the waste stream and complexity of the separation process (improved process quality. This specifically applies for example to .sup.227AC, which is a very long lived isotope, that should be eliminated from the final product as much as possible, but cannot be chemically separated from the desired isotope .sup.225AC. [0028] The burn-up of .sup.226Ra by thermal neutron activation is reduced, hence reducing the amounts of .sup.226Ra required for production. .sup.226Ra is scarce. [0029] The burn-up of .sup.225Ra by thermal neutron absorption (‘back-activation’ of .sup.225Ra to .sup.226Ra) is reduced, hence optimizing the .sup.225Ra production. [0030] Increasing the production of .sup.225Ra per weight unit (gr) of .sup.226Ra. [0031] Increasing the yield of .sup.225Ac per weight unit (gr) of .sup.226Ra. [0032] reducing the amount of waste produced (such as undesired isotopes)

[0033] The method of the present invention allows for existing and readily available irradiation infrastructure to be adopted, thereby allowing relative quick and cost effective implementation of this method for the generation of actinium isotopes that find widespread applicability the medical field. Because irradiation infrastructure such as a moderated material test reactor can irradiate large volumes, large quantities can be produced. The relatively long half-life of .sup.225AC of 10 days allows for distribution worldwide without losing much effectiveness, which enables centralized production. This leads to significant economic benefit compared to de-centralized production in multiple machines, using alternative .sup.225AC generation technologies.

[0034] In another aspect of the invention, it was found that when irradiating .sup.226Ra in a thermal neutron absorption shield, it was possible to separate radium (i.e. all radium isotopes in the target) from other elements in the target within days after irradiation, to eliminate unwanted impurities present in the target or unwanted impurities generated by decay. The period between end of irradiation and first purification should be at least sufficient for the .sup.227Ra (half-life 42.2 minutes) to decay. Especially .sup.227AC is an unwanted isotope with a 21.8 year half-life, which should be avoided to be introduced in patients and the environment, and could still be present in the irradiated material in unacceptable quantities, even though its generation was largely avoided by adopting a thermal neutron absorption shield in the method of the invention.

[0035] The workup (i.e. the isolation of the desired isotope) can be performed in various ways. In one embodiment by the chemical extraction (or elution) of actinium isotopes from the irradiated .sup.226Ra to remove actinium (both .sup.227AC and .sup.225Ac). With the separation being done several hours-several days after irradiation, adequate quantities of .sup.225Ra remain to generate new .sup.225AC in sufficient quantities, and in pure form, as all other actinium isotopes that could be detrimental for product quality have been removed prior to production.

[0036] In another embodiment, the workup may be performed by the extraction (or elution) of radium isotopes from the irradiated .sup.226Ra and subsequently allow the .sup.225Ra in the isolate to decay into .sup.225AC.

[0037] The thermal neutron shielding in the method of the invention can be established by encompassing the Radium-226 starting material in a (closed) thermal neutron shield. The thermal neutron shield is made of materials having a high thermal neutron cross section. The thermal neutron shield materials are preferably selected from the group of elements with high thermal neutron absorption cross-sections consisting of boron, cadmium, gadolinium, hafnium and mixtures thereof. There is a preference for a gadolinium shield.

[0038] The starting .sup.226Ra material can be provided in the desired chemical form (as a metal, oxide, salt or mixture thereof). The starting material may also be provided as a powder, and/or in combination with other elements such as Al. The starting materials may be sintered and/or pelletized. The starting material is placed in a container, preferably forming a closed containment, for example a closed ampoule, which can be made of metallic, quartz or ceramic material, and can be closed to form a containment.

[0039] The ampoule is placed in a preferably cylindrical body consisting of a material with a high thermal neutron absorption cross-section (such as the preferred gadolinium) and with a low absorption cross-section for high energy neutrons. The thermal neutron shield can be considered to form a closed containment as well. Additionally or alternatively, the target material in the ampoule may be mixed with thermal neutron absorber materials. This configuration is placed in a second containment, generally a metallic cylindrical body closed with a (welded) end-cap. The containment is cooled at the outside by the reactor coolant. In between the shield and the outer containment, a low density high thermal conductivity filler can be added in case needed, to transport heat generated in all materials and components to the coolant without high thermal gradients, to avoid overheating and reduce thermal gradients and thereby thermal stresses, melting, decomposition in the various materials and components.

[0040] The low energy (thermal) neutron flux from the material test reactor is absorbed by the thermal shield material, hence low energy (thermal) neutrons are almost absent within the shield cavity. High energy (fast) neutron (typically about 0.1 MeV, preferably 5 MeV to about 20 MeV) pass through relatively undisturbed. Therefore in this configuration a specific fast neutron spectrum is created within the neutron shield. For .sup.226Ra containing material in the shield, the .sup.226Ra(n,2n) reaction takes place in the material test reactor fast flux (which is largely undisturbed by the shield), but thermal neutron activation reactions are avoided, as thermal neutrons are effectively absorbed by the shield.

[0041] The invention can be illustrated in more detail as follows (FIG. 1):

[0042] The radium starting material (5), which can be in a variety of chemical forms such as metal, oxide, carbonate, nitride etc. is provided in a radium starting material containment unit (4). The unit (4) may be an ampoule of radiation resistant light material such as metals, quarts, ceramics. The radium-containing starting materials can be placed in the thermal neutron shield holder (3) which may be closed by an end-cap (2). The thermal neutron shield (4) holder and end-cap (3) can be made from a material with a high thermal cross-section, such as boron, cadmium, gadolinium. A preferred material for the holder and end-cap is gadolinium. Preferably the holder and the end-cap are from the same or substantially the same thermal neutron absorbing material. The thermal neutron shield preferably encloses the radium target material completely, i.e. shields it from thermal neutrons of the reactor. The thermally shielded radium target material can be placed in a containment (7), generally from a metallic material that may have a containment cap (1). The containment and containment cap can be sealingly closed, for instance by welding. The containment and containment cap are preferably from the same or substantially the same material. Between the containment and the thermal shield, a filler material (6) may be provided, typically a light weight material with a good thermal conductivity, for example graphite or aluminium.

[0043] FIG. 1 discloses a schematic representation of an example of a containment cylinder that holds the thermal neutron shielded radium target material.

[0044] FIG. 2 shows the quantities of .sup.226AC and .sup.225Ra over time, during and after irradiation in shielded and unshielded form. The unshielded case is based on an actual irradiation of 0.12 mg .sup.226Ra in a high flux position in the HFR Petten, the Netherlands. .sup.226AC content was measured at three different times and are indicated in the graph. From the measurements and supported by calculations the quantity of .sup.226AC provides the quantity of .sup.225Ra that is produced at the end of irradiation without a thermal neutron absorbing shield. The .sup.225Ra production rate will be similar or higher in case a thermal neutron absorbing shield is adopted, as the fast neutrons are not absorbed nor affected by the thermal neutron absorption shield, and the burn-up by thermal neutrons of the .sup.225Ra generated is eliminated by the thermal neutron absorption shield.

[0045] With the .sup.225Ra quantity known over time based on quantity after irradiation and decay, the process of extraction of .sup.225AC can be simulated: [0046] After 3 days all actinium is chemically removed, and only Radium isotopes .sup.225Ra and .sup.226Ra remain. This actinium can contain unacceptable amounts of the unwanted .sup.227AC isotope and may be discarded or used for other purposes [0047] From that point onward the .sup.225Ra constantly generates new .sup.225AC, and no other actinium isotope is formed or present. [0048] The .sup.225AC generated can be repetitively removed until most of the .sup.225Ra has decayed.

[0049] In the figure a tentative extraction scheme is shown, in which after periods ranging from 5 to 8 days the .sup.225AC is extracted from the Radium.

[0050] It is calculated that all Radium is extracted and purified three days after end of irradiation (i.e. at 31+3 days after start of irradiation). The extracted and purified Radium therefore contains no Actinium anymore at that point in time. The .sup.225AC subsequently generated is generated by decay from .sup.225Ra that is present in the purified Radium.

[0051] FIG. 2 shows a graph of the achieved activity of the irradiation of .sup.226Ra both in a thermal neutron shielded and a thermal neutron unshielded form (for comparative purposes). The figure shows that from the 0.12 mg .sup.226Ra irradiated in this way in a regular thermal neutron material test reactor like the HFR, it can be extrapolated that approximately a total of 6.9 GBq .sup.225Ac/g .sup.226Ra can be generated, which corresponds to 860 patient doses in case 8MBq per dose is assumed (based on normal PSMA application of .sup.225Ac). Multiple extractions of 100s of patient doses can therefore be performed based on 1 gram .sup.226Ra quantities. Adopting a thermal neutron absorption shield, provides similar or higher quantities, but the activity of unwanted isotopes is significantly reduced, facilitating handling and processing to great extent, and reducing the risk of impurities in the end product by reducing their formation during irradiation. In the method described in US20070092051, and based on theoretical calculations, the potential yield was estimated as 5 mCi (0.185 GBq) of Radium-225 per gram of Radium-226 in a fast neutron reactor. The present invention provides a significantly higher yield (about 40-fold) in a regular material test reactor such as the HFR.

[0052] There are other, known ways of producing .sup.225AC but the key benefit of the claimed method is that existing and readily available irradiation infrastructure can be used, providing significant amounts of .sup.225AC, while eliminating the formation of unwanted isotopes that complicate post-irradiation handling, processing and purification. The use of the readily available irradiation infrastructure allows relative quick and cost effective implementation. This is important for a medical isotope that is already in high demand and expected to increase in demand significantly in coming years, especially regarding the spectacular results achieved with .sup.225AC for example adopted in conjunction with PSMA-compounds for the treatment of castrate-resistant prostate cancer, and its expected efficacy for new medical applications.