Production of high specific activity Pt isotopes from Ir alloys

Abstract

A method for producing platinum isotopes from iridium is disclosed, the method comprising the steps of providing a mixture, preferably an atomic dispersion, of a carrier material such as Al and iridium (Ir), irradiating the carrier material-Ir mixture with a neutron flux to produce an irradiated carrier material-Ir mixture containing platinum (Pt) isotopes, and isolating the Pt isotopes from the irradiated carrier material-Ir mixture.

Description

[Technical Field]

[0001] The present invention relates to a method for producing platinum isotopes from iridium-containing starting materials. The present invention also relates to platinum isotopes having high specific activity. The present invention also relates to the use of platinum isotopes in radiotherapy and / or diagnostic applications. [Background technology]

[0002] Certain Auger electron-emitting isotopes have been used in radiation therapy. Because the damage caused by Auger electron emitters is very small, effective treatment requires that the Auger electron emitter be located at or near the critical site of cancer cells, such as the cell membrane or cell nucleus (DNA) or mitochondrial DNA-RNA. In this case, the Auger electron emitter can be very effective in killing the cells.

[0003] If not properly located, Auger electron emission is unlikely to cause any severe irreparable damage to cells, even if located inside the cell. Auger electron emission within a cell does not cause damage to surrounding cells or tissues. Therefore, appropriate targeted compounds that specifically deliver Auger electron emitters to the proper location within cancer cells have the potential for effective treatment with few or no side effects.

[0004] platinum isotope 195m Pt appears to offer the desired functionality: it has a chemical affinity for binding to DNA, which has already been utilized for decades in platinum-based cancer therapies such as chemotherapy. 195m Pt emits gamma rays that may be useful for SPECT imaging. 195mCombining Pt with the chemical behavior of platinum makes it an ideal isotope for eliminating cancer cells with the potential for minimal damage to healthy cells (i.e., no or limited side effects). Pt compounds that combine cancer cell targeting with chemical binding of Pt to cellular DNA within the cell may be particularly successful in cancer therapy.

[0005] 195m Although routes to produce Pt are known, there are few methods to produce it in clinically relevant quantities with sufficiently high specific activity for therapeutic efficacy and with sufficient efficiency to accommodate the half-life of the isotope (approximately 4 days) for successful radiotherapy applications. 195m Pt is difficult to manufacture.

[0006] 194 Pt((n,γ)-reaction)) or 195 Pt((n,2n)- or (γ,n)-reaction, 195m The standard carrier-added route through the neutron source (including thermal neutron shielding to maximize yield by avoiding high cross-section burn-up of Pt) provides experimentally determined high specific activities by irradiation at HFR Petten, which may be sufficient for diagnostic applications and also provides relatively high radionuclide purity. However, for therapeutic applications, much higher, preferably maximized, specific activities are required. For this, e.g. 193 A "non-carrier addition" pathway is required via neutron irradiation and (double neutron) activation of Ir.

[0007] 193 For the non-carrier pathway via neutron irradiation of Ir, 195m A process for producing high specific activity Pt isotopes, such as Pt, is described in WO 2004015718(A1). In WO 2004015718(A1), 195m Pt is 193 Ir is neutron-activated, and as a result 194 Ir is generated, followed by194 Ir is neutron activated, 195m Generate Ir, which is 195m It is produced by decaying to Pt. 193 After irradiating an Ir-enriched iridium target in a reactor, the platinum isotopes are chemically separated from the iridium isotopes, 195m Pt]-Pt.

[0008] 193 The use of Ir-enriched iridium has several advantages: per gram of iridium 195m Higher yield of Pt. · 195m Higher specific activity of Pt. · 192 Ir, etc. 193 Other stable compounds that are more abundant at lower Ir enrichment levels 191 Unwanted by-products from the activation of Ir isotopes include long-lived isotopes that complicate the handling of irradiated targets and the recycling of spent target material, and the production of other platinum isotopes that are stable (reducing the specific activity) or unwanted (reducing the radionuclide purity of the final product).

[0009] However, this method entails several problems.

[0010] Iridium isotopes have a very large neutron absorption cross section, which leads to self-shielding of neutrons; in a solid iridium mass under neutron bombardment, the iridium on the outside of the target absorbs most of the neutrons, leaving the inside exposed to fewer neutrons, thereby reducing the amount of Pt that can be formed by neutron activation. Therefore, to maximize Pt production per unit mass of iridium, an irradiated target with a lower iridium density is desirable, so that all of the iridium is exposed to a similar high neutron flux. This reduces the need for expensive enrichment. 193 The use of Ir can be reduced and the concentration of Pt isotopes increased, which can also increase extraction efficiency.

[0011] Radioactive platinum isotopes generally emit low-energy radiation, but during irradiation, highly radioactive, high-gamma-energy emitting iridium isotopes are formed, which must be processed as part of the post-irradiation Pt extraction process. The radiation intensity, particularly from the iridium isotopes, makes them difficult to handle and requires rigorous shielding infrastructure (transport containers, hot cells, local shielding) to avoid unacceptable worker exposure, risk of radioactive release, and handling hazards.

[0012] One process for producing Pt isotopes from iridium is the melting of irradiated targets. Metallic iridium is notoriously difficult to melt. After irradiation, the melting process is made difficult by the intense radiation emitted by the iridium isotopes formed. 195m Loss of Pt radioactivity, and 195m Processing and separation is preferably carried out as soon as possible after irradiation to avoid the ingrowth of other (unwanted) Pt isotopes from the iridium, which would reduce the Pt specific activity. Therefore, a difficult dissolution process must be avoided. 195m The same is true for the follow-up steps of separating and purifying [Pt]-Pt. WO 2004015718(A1) proposes prolonged exposure to high temperature and pressure to dissolve iridium, but this is difficult to achieve practically and safely in a nuclear (radiation-shielded) environment with large amounts of high-energy radioactivity. A simple and practical method for producing Pt isotopes from iridium is needed. Summary of the Invention [Problem to be solved by the invention]

[0013] The present inventors have investigated the use of platinum isotopes, particularly 195mIt has been found that Pt can be produced at high specific activity and can be conveniently and efficiently produced by providing a target of a carrier material / iridium mixture, irradiating the mixture with neutrons to induce the formation of platinum isotopes, promoting dissolution of the target after irradiation, and subsequently isolating the platinum isotopes by an extraction, separation, and / or purification process. [Means for solving the problem]

[0014] Thus, in a first aspect, there is provided a method for producing platinum isotopes from iridium, said method comprising: - providing a mixture of a carrier material and iridium (Ir); - irradiating the carrier material-Ir mixture with a neutron flux to produce an irradiated carrier material-Ir mixture comprising a platinum (Pt) isotope; - isolating Pt isotopes from the irradiated carrier material-Ir mixture; Includes. [Brief explanation of the drawings]

[0015] [Figure 1] Figure 1 shows the cumulative recovery of Al, Ir, and Pt at different time intervals after treating beads (2.5% Ir) with 37% HCl and aqua regia (lowest recovery). [Figure 2] Figure 2 shows the recovery of Al, Ir, and Pt at different time intervals after treating beads (2.5% Ir) with 37% HCl and aqua regia (highest recovery). DETAILED DESCRIPTION OF THE INVENTION

[0016] In a first aspect, the present invention relates to a method for producing platinum isotopes from iridium, said method comprising: - providing a mixture of a carrier material and (enriched) iridium (Ir), - irradiating the carrier material-Ir mixture with a neutron flux to produce an irradiated carrier material-Ir mixture comprising a platinum (Pt) isotope; - melting the target after irradiation and isolating the Pt isotopes from the irradiated carrier material-Ir mixture; Includes.

[0017] The present invention involves the use of a carrier material-iridium mixture as the irradiation target.

[0018] The carrier material can be selected from a large group of materials and mixtures thereof, provided that they allow and advantageously enable the distribution of iridium atoms, atomic clusters or small particles in the carrier material, e.g., by heating and / or melting, have a reasonable thermal conductivity (to avoid too high temperatures and temperature gradients in the target during irradiation), have a low neutron absorption cross section to avoid self-shielding and reduce self-shielding of the iridium target as a whole, avoid the formation of problematic long-lived and / or high gamma energy neutron activation products, are advantageously meltable, and have a high vaporization temperature (to avoid possible overpressure problems during irradiation).

[0019] The carrier material may preferably be selected from the group consisting of Al, Zr, Ti, In, Si, V, Sc, Mg, Ca, Pb, and mixtures thereof, and may preferably be selected from the group consisting of Al, Ti, or Zr, and mixtures thereof.

[0020] The carrier material may be isotopically enriched to avoid, limit, or reduce certain neutron activation reactions.

[0021] The carrier material-Ir mixture is a dispersion, preferably an atomic dispersion, of Ir in the carrier material. Metallic Ir can be present as particles in the carrier material-iridium mixture. The average particle size or inclusion size of the Ir particles in the carrier material-iridium mixture is preferably less than 500 microns, preferably less than 100 microns, more preferably less than 10 microns, and even more preferably less than 1 micron. By having the iridium in the mixture exist as small particles in the bulk of a carrier material with a low neutron absorption cross section, self-shielding of neutron radiation by the target mass can be reduced and / or avoided. Shielding occurs when iridium atoms are shielded from neutrons by surrounding (iridium) atoms.

[0022] This avoidance of shadowing can be further avoided in embodiments of the present invention in which the carrier material iridium compound is provided in a form that prevents or reduces shadowing due to its high surface area to volume ratio, such as a (thin) wire, rod, plate, foil, etc.

[0023] In a particular embodiment, the carrier material-Ir mixture may be provided as an alloy or compound of Ir and the carrier material, preferably a metal compound such as an iridide. Iridide refers to an intermetallic alloy, and in the case of Al, aluminum iridide (Al n+ +Ir n- ), which, due to its unique ionogenic properties, is more soluble than Ir(0) and therefore more likely to release Pt(0) into solution. Therefore, maximizing iridide formation in the target is preferred. Similar intermetallic alloys, which may or may not be ionogenic, exist and can be formed with other carrier materials and are similarly advantageously useful.

[0024] In an alternative embodiment, the carrier material-Ir mixture may be prepared by co-precipitating the carrier material and Ir together, preferably by co-precipitation from solution.

[0025] The irradiation target can be produced by heat treating or melting iridium and a carrier material together, for example, by heating in an oven, welding, induction or electromagnetic radiation, laser, TIG welding, rewelding according to the same procedure, repeated heat treatment, possibly combined with a raw material produced by co-precipitation of pre-dissolved iridium and carrier material, or by other methods that maximize the dispersion of iridium in the carrier material. This procedure is preferred because, in the case of aluminum, aluminum-iridites can be formed, which allows iridium to be dispersed in aluminum at the atomic level. Similar compounds can be foreseen between iridium and the carrier material, which promotes the dispersion of iridium at the atomic or small particle level in the bulk carrier material. Another alternative production method can be by electrical induction.

[0026] Providing the Ir in a dispersion in a carrier material offers a number of advantages.

[0027] By combining iridium with a large amount of carrier material with a relatively small neutron absorption cross section, the effective iridium density in the target can be significantly reduced, thus reducing neutron self-shielding and activating iridium. 195m This maximizes the production of Pt. 195m Pt content can be maximized, specific activity can be increased, and / or the required concentration can be increased. 193 The amount of Ir can be minimized.

[0028] Carrier material-iridium metal alloys, such as Al-Ir, are unlikely to dissociate, form gaseous products during neutron irradiation, or undergo significant changes under neutron irradiation, allowing for direct irradiation of the target without safety concerns and without difficult target conditioning or containment procedures. Carrier material-iridium alloys or soluble carrier material-Ir compounds therefore form target materials well suited for neutron irradiation (e.g., in a reactor). Also, workup for isolating the desired Pt isotopes is greatly improved. Iridium itself is difficult to dissolve, typically requiring harsh environments such as very high temperatures, strong solvents, and / or high pressures. Dissolving iridium is greatly facilitated when iridium is dispersed in a carrier material, such as an Al or Zr matrix. Dissolution of Ir dispersed in a carrier material matrix is ​​typically rapid (a few hours) and can be carried out using standard, compact, and simple equipment. Thus, difficult, dangerous, and time-consuming procedures using highly radioactive materials, and the attendant safety hazards and worker radiation dose issues, can be addressed, mitigated, and / or avoided when using carrier material-iridium mixtures, and relatively small and simple non-pressurized dissolution (and subsequent separation / purification) equipment can be used.

[0029] The present invention is a method for detecting high specific activity [ 195m [Pt]--facilitates and improves the production of Pt, a radioisotope with great promise that remains largely unexplored due to limited availability resulting from production difficulties. The present invention maximizes Pt production by reducing self-shielding, providing a convenient and safe target for neutron irradiation that makes the melting and post-irradiation Pt extraction / purification process much easier and simpler.

[0030] The present invention allows for a simple production apparatus for dissolving the formed Pt, which, after conditioning, can be directly introduced into a chemical separation column. The (un)irradiated carrier material residues and iridium isotopes can be separated and removed to recover the desired Pt. The apparatus can be compact and implemented with simple automation (a pump, a few valves, tubing, and a small column are sufficient), allowing it to be safely and conveniently installed in a radiation-shielded environment. The simplicity and small volume of the apparatus also facilitates the use of (additional) local shielding. The Pt-containing column can be removed and used for further purification and radiochemical / radiopharmaceutical processing without the safety concerns of handling high-energy gamma-ray-emitting radioisotopes and iridium isotopes.

[0031] The amount of Ir in the mixture may be 0.5-50 wt. % calculated relative to the total weight of the mixture. Relevant factors in this regard are the level of self-shielding reduced by the large amount of carrier material, the ease of target fabrication, which requires iridium to be dispersed as small particles (preferably atomic) in the carrier material, and the ease of dissolution and subsequent separation / purification of iridium, produced platinum, and carrier material. In embodiments, the amount of Ir in the mixture is less than 50, preferably less than 10, and more preferably less than 3. In preferred embodiments, the amount may be greater than 1 wt. %, preferably greater than 2 wt. %.

[0032] The amount of carrier material in the carrier material-iridium mixture can be adjusted similarly. Thus, the amount of carrier material in the mixture, calculated based on the total weight of the mixture, can be 50 to 99.5% by weight. In an embodiment, the amount of carrier material in the mixture is greater than 50%, preferably greater than 60%, and more preferably greater than 70%. In a preferred embodiment, the amount can be greater than 80% by weight, preferably greater than 90% by weight, and even more preferably greater than 97% by weight.

[0033] In a preferred embodiment, the iridium is193 The enrichment of Ir can be achieved by methods known in the art. Preferably, 193 The Ir is highly enriched (greater than 95%, or preferably greater than 99%) to avoid the production of highly radioactive Ir isotopes and unwanted Pt isotopes.

[0034] The irradiation product obtained from the irradiation of the carrier material iridium mixture is then 195m The irradiation products may be processed to separate products containing isotopes such as Pt and / or other Pt isotopes. The carrier material may be removed from the irradiation products, for example, by melting and / or dissolution followed by separation. Alternatively, the irradiation products may be extracted with an extraction solvent to separate the irradiation products into various components, such as carrier material, unconverted iridium, and products from the irradiation process, such as platinum isotopes.

[0035] The extraction solvent is preferably capable of selectively dissolving at least one or more of the components. This method allows, for example, the carrier material to be leached from the irradiation products, leaving unconverted iridium and any products behind for further processing. If the iridium dispersion is sufficiently low and the carrier material content is high, the Pt isotopes produced due to recoil after decay when the iridium isotope is activated may be located in the carrier material mass, thereby allowing more Pt to be extracted from the carrier material than from the iridium, and facilitating the extraction process since there is no need to dissolve the iridium. The extraction solvent may also be capable of dissolving all components, which may be followed by a separation step, which may include chromatography, ion exchange columns, or electrochemical separation.

[0036] The extraction solvent can be selected from the group consisting of strong (mineral) acids and strong caustic alkalis (alkaline solutions). Aluminum dissolves in strong caustic alkalis. This can be the first step, followed by separation of products such as dissolved aluminum and unconverted iridium and Pt isotopes. Strong mineral acids are more preferred because they dissolve most metals. Particularly preferred are acids and acid mixtures such as HCl, NaBrO3 / HCl, H2O2 / HCl, and HCl / HNO3 (aqua regia), as well as mixtures thereof, preferably aqua regia. Typically, the metal is in a (partially) ionized form, which facilitates subsequent separation, for example, by selective precipitation, ion exchange, chromatography, and / or electrochemical methods.

[0037] To further aid the metal dissolution process, the (aqueous) dissolution of the irradiation product may be carried out at a temperature between 20°C and 120°C, preferably between 30°C and 90°C.

[0038] 195m The irradiation target material in solution, which may comprise one or more of Pt, other Pt isotopes, a carrier material (such as aluminum or zirconium or silicon, and as mentioned elsewhere herein), and iridium, is preferably a Pt isotope, preferably a Pt ion. 195m To isolate and / or purify the desired product, such as Pt, it can be separated into its components by (column) chromatography, ion exchange, and / or electrochemical methods. (Column) chromatography can be used to separate the components of the irradiation product. In embodiments, the components can be selectively eluted, for example, using (gradient) eluents or successively using eluents with different pHs to elute the desired (or undesired) components while retaining the desired isotope on the column, or vice versa. 195m ) selectively eluting Pt isotopes. Examples of such processes are described, for example, in WO 2004015718 A1 and can be determined by one skilled in the art.

[0039] The results of the method of the present invention are 195mPt]-Pt. 195m Pt]-Pt is 13-17 GBq per mg of Pt (all Pt isotopes produced) at the end of irradiation. 195m It has a specific activity that can be on the order of Pt. [Example]

[0040] All percentages are by weight and calculated based on the total composition unless otherwise specified.

[0041] All chemicals were available from commercial suppliers and were used as received unless otherwise noted.

[0042] Example 1: Preparation of Al / Ir and Al / Ir / Pt beads Aluminum, iridium, and platinum were weighed using a conventional balance and a microbalance, respectively. The following ratios were selected as representative: 97.5% Al, 2.50% Ir; Ir 97.5% Al, 2.25% Ir, 0.25% Pt; and 95.0% Al, 4.50% Ir, 0.50% Pt, with a total weight of 100 mg each. Generally, aluminum powder was weighed using a conventional balance, followed by the addition of iridium. Pt mesh wire was cut, weighed using a microbalance, and then added. These components were then transferred to a mold, and tablets were pressed under 500 kg pressure. Tablets prepared using a hydraulic press were welded using a tungsten inert gas welder at 47 A or 53 A at 11.7 V DC with a flow rate of 2 L argon / min. The tablets were placed into custom molds (copper or tungsten) and welded for 15 seconds on each side, with approximately 15 seconds of cooling between welds. Rewelding was performed to increase the amount of aluminum-iridium alloy to aid in the solubility of the iridium. Both Al / Ir / Pt and Al / Ir beads were fabricated and evaluated before and after irradiation, and before hot experiments (i.e., experiments with irradiated material) were performed.

[0043] Beads were also prepared using zirconium, substituting aluminum for zirconium, with a similar weight distribution, and showed the same performance and uniform target formation. The welding time and number of welding steps were increased to ensure complete melting.

[0044] Experiment 2: Extraction of Ir and / or Pt The following extraction experiments were carried out.

[0045] 2.1 Extraction of iridium from IrO2(s) powder in aqua regia at high temperatures IrO2 powder (99+%) was suspended in aqua regia (65% HNO3 and 37% HCl in a 3:1 molar ratio) and refluxed at elevated temperature in a reflux apparatus for 24 hours. Small aliquots were taken after 4 and 24 hours. ICP-OES experiments indicate that less than 1% of the iridium was recovered after this time.

[0046] 2.2. Extraction of iridium from Ir(s) powder at high temperatures using aqua regia Ir powder (99+% purity) was suspended in aqua regia (32 mL, 65% HNO3 and 37% HCl in a 3:1 molar ratio) and refluxed at elevated temperature in a reflux apparatus for 48 hours. ICP-OES analysis indicates that 0.24% of the iridium was recovered after this time.

[0047] 2.3 Extraction of Iridium from Ir(s) Powder at High Temperature and Pressure Using Aqua Regia in a Parr® Acid Digestion Vessel Several experiments were performed using metallic Ir(s) powder (99+%) and aqua regia (10:1 molar ratio; 37% HCl and 65% HNO3) in a 5 mL cup acid digestion vessel (ADV) at different reaction times (0.25 h to 5 h) and temperatures (160 °C and 217 °C) in a standard oven. None of these conditions resulted in significant recovery of iridium; less than 1% was recovered after treatment with aqua regia.

[0048] 2.4. Extraction of Iridium and Platinum from Al / Ir / Pt Beads at Room Temperature Using NaBrO3 / HCl, H2O2 / HCl, and Aqua Regia Al / Ir and Al / Ir / Pt beads (prepared by TIG welding) with diameters of 3.77 mm or 3.94 mm, with compositions of 5% or 2.5% iridium in aluminum and 0.25% Pt, were treated with HCl, followed by aqua regia, HO / HCl, or HCl / NaBrO(aq), without stirring, at room temperature (rt) in 15 mL centrifuge tubes, as shown in Table 1. Typically, the beads have dimensions in the mm range (1–10 mm).

[0049] [Table 1]

[0050] The highest recoveries of iridium and platinum were obtained using a procedure consisting of six HCl leaches followed by six aqua regia leaches: 15-25% of iridium and 22-37% of platinum were recovered using this method.

[0051] 2.5. Extraction of Iridium and Platinum from Al / Ir / Pt Beads at High Temperatures Using Aqua Regia In a one-pot procedure, Al / Ir / Pt beads were treated with 37% HCl and aqua regia at room temperature and reflux, respectively. Using standard reflux equipment, 96% Al beads were able to recover 51% of the starting amount of Ir and 71% of the starting amount of Pt after 24 hours of reflux in aqua regia. From the same batch, beads were rewelded at different currents (A) and welding times to determine whether increased intermetallic alloying could lead to increased solubility. Excellent aluminum recovery was observed for each rewelded bead, while minimal recoveries of 37% iridium and 58.1% platinum were achieved for one bead (Figure 1) and 52.2% iridium and 78% platinum for another bead (Figure 2) after 3 hours of heating under reflux (approximately 115°C).

[0052] Example 3: Column purification of lysed beads To separate the individual metals, column purification using AG50W-X4 resin was performed with a solution containing 10.7 mg of Al, 1.12 mg of Ir, and 0.1 mg of Pt. After elution with 1 M HCl and 0.2 M thiourea, 83.1–84.2% of Al, 40.4–41.8% of Ir, and 16.4–22.2% of Pt were recovered. Most of the Pt remained on the column.

[0053] Example 4: Concentration 193 Ir irradiation The quartz tubes, whose lengths (inner diameter 0.7 mm, outer diameter 1.6 mm) varied from 20 to 47 mm, contained 94.34% of the sample, ranging from 78.1 to 1024.7 μg. 193 Ir was loaded using a Mettler Toledo microbalance. The tubes were cut and sealed. All tubes were placed in graphite inserts and irradiated in the high flux reactor Petten for two weeks. The activation of these samples was then measured using gamma spectroscopy. 94.34% 193 Ir irradiated at moderate neutron flux in the HFR Petten has a specific activity of 13-17 GBq / mg Pt (all isotopes produced). 195m It was found that Pt could be obtained.

[0054] In these experiments, Pt isotopes and 195m It should be noted that the amount of Pt was significantly underestimated in neutron activation calculations using some of the common activation cross section libraries, which 194 Ir and 195 These libraries show significant uncertainties and underestimations for the neutron absorption cross section of Ir, which can differ by up to a factor of 20.

[0055] Example 5 Al / 193 Repeating the above experiment with Ir beads, 195mPt is obtained in good yields, in certain cases higher than expected, which is attributed to the recoil effect, which causes the Pt isotope to leave the Ir atomic environment and become embedded in the aluminum rather than the Ir, making it easier to isolate in higher yields.

Claims

1. A method for producing platinum isotopes from iridium, comprising the steps of: - providing a mixture of a carrier material and iridium (Ir); - irradiating said carrier material-Ir mixture with a neutron flux to produce an irradiated carrier material-Ir mixture comprising platinum (Pt) isotopes; - isolating Pt isotopes from the irradiated carrier material-Ir mixture; A method comprising:

2. 2. The method of claim 1, wherein the carrier material is selected from the group consisting of Al, Zr, Ti, In, Si, V, Sc, Mg, Ca, Pb, and mixtures thereof, preferably selected from the group consisting of Al, Ti, or Zr, and mixtures thereof.

3. The method of claim 1 or claim 2, wherein the carrier material-Ir mixture is a dispersion, preferably an atomic dispersion, of Ir in the carrier material.

4. 4. The method of any one of claims 1 to 3, wherein the Ir is dispersed in the carrier material in particles having an average particle size or atomic cluster size or precipitate size of less than 500 microns, preferably less than 100 microns, more preferably less than 10 microns, and even more preferably less than 1 micron.

5. The method according to any one of claims 1 to 4, wherein the carrier material-Ir mixture is an alloy or compound of Ir and the carrier material, preferably a metal compound such as iridide.

6. The method of any one of claims 1 to 5, wherein the carrier material-Ir mixture is made by melting the carrier material and Ir together.

7. The method according to any one of claims 1 to 6, wherein the fusing is by heat treatment, preferably by welding, more preferably by TIG welding.

8. 8. The method according to any one of claims 1 to 7, wherein the amount of Ir is between 0.5 and 50% by weight, calculated relative to the total weight of the mixture.

9. The method according to any one of claims 1 to 8, wherein the amount of carrier material is from 50 to 99.5% by weight, calculated relative to the total weight of the mixture.

10. Ir is １９３ 10. The method of any one of claims 1 to 9, wherein the Ir is enriched iridium.

11. The Pt isotope is １９５ｍ The method of any one of claims 1 to 10, comprising Pt.

12. The method according to any one of claims 1 to 11, wherein the carrier material is removed from the irradiation product.

13. The method according to any one of claims 1 to 12, wherein the irradiation product is extracted with an extraction solvent.

14. The isolation step preferably comprises column chromatography, １９５ｍ The method according to any one of claims 1 to 13, further comprising a step of capturing Pt. Request 15 High specific activity with a specific activity of 13 to 17 GBq / mg Pt, obtainable by the method according to any one of claims 1 to 14. １９５ｍ Pt.

Images