EP2398023A1

EP2398023A1 - Production of molybdenum-99

Info

Publication number: EP2398023A1
Application number: EP10166604A
Authority: EP
Inventors: Kamel Abbas; Jan Kozempel; Uwe Holzwarth; Federica Simonelli; Neil Gibson
Original assignee: European Union represented by European Commission
Current assignee: European Union represented by European Commission
Priority date: 2010-06-21
Filing date: 2010-06-21
Publication date: 2011-12-21

Abstract

A method is described for the production of molybdenum-99 comprising the steps of providing a target comprising elements fissile under charged particle bombardment; and irradiating at least a portion of the target with accelerated charged particles to induce fission of said elements, thereby producing an irradiated target portion comprising molybdenum-99.

Description

Technical field

The present invention generally relates to a method for the production and the separation of molybdenum-99 (Mo-99).

Background Art

Technetium-99m (symbolized Tc-99m), the daughter product of molybdenum-99, is the most commonly utilized medical radioisotope in the world, used for approximately 20-25 million medical diagnostic procedures annually, comprising some 80% of all diagnostic nuclear medicine procedures.
Tc-99m is a metastable nuclear isomer of technetium-99 (Tc-99), which "decays" to Tc-99 by rearrangement of nucleons in its nucleus. It is a short-live gamma ray emitting isotope largely used in radioactive isotope medical tests, for example as a radioactive tracer that medical equipment can detect in the body.
The decay process starting from Mo-99 that produces Tc-99m is as follows: $Mo - 99 \to Tc - 99 m + β^{-} + {\tilde{v}}_{e}$

where β^- denotes a beta particle (electron) emitted from the nucleus, and ṽ_e denotes the emitted electron antineutrino. The half-live of this decay is about 66 h.
Tc-99m will then undergo an isomeric transition to yield Tc-99 and a monoenergetic gamma emission: $Tc - 99 m \to Tc - 99 + γ$

with a half-live of about 6 h.
Commercial quantities of molybdenum-99 have been produced so far in nuclear reactors through the uranium fission process (for example, see US 3799883 ) utilizing highly enriched uranium-235 (HEU) (>90%) and through neutron capture of Mo-98. The fission process yields Mo-99 with a large array of undesirable fission products that present significant infrastructure, health and security, liability, handling, storage, and waste issues and associated costs, while the neutron capture method produces no nuclear waste, but presents a lower yield and a much smaller specific activity of Mo-99.
A global shortage of Tc-99m is emerging because two aging nuclear reactors that provide about two-thirds of the world's supply of Mo-99, are currently shut down for repairs. In fact, nowadays only seven nuclear reactors (Table 1) are producing Mo-99 for a worldwide market and the majority of them are over 40 years old which is well above their designed operation life (35 years).

Table 1:

Country	City	Facility	Age of the reactor	World supply (%)	Power (MW)
Canada	Rolphton, Ontario	NRU Chalk River	52 years	31%	135
The Netherlands	Zijpe	HFR-Petten	47 years	33%	45
Belgium	Mol	BR2	47 years	10%	100
France	Saclay	OSIRIS	42 years	8%	70
South Africa	Pelindaba	SAFARI	43 years	3%	20
Australia	Sydney	Opal	2 years		20
Czech republic	Rez	LWR-15	55 years		10

Beside the issue related to the availability of adequate nuclear reactors, there is also an issue related to the availability and shipping of highly enriched U-235 (HEU). The use and shipping of highly enriched U-235 raises non-proliferation concerns since this material might be intercepted and abused for criminal purposes.
The use of Low Enriched Uranium (LEU) for Mo-99 production would appreciably reduce this concern. However, one disadvantage of this method is that the production of Pu-239 is increased more then 25 times (see "MAKING OF FISSION Mo-99 FROM LEU SILICIDE(S): A RADIOCHEMISTS' VIEW", Z.I. Kolar and H.Th. Wolterbeek, 2004 International Meeting on Reduced Enrichment for Research and Test Reactor, 2004 RERTR). Additionally, this method needs 5 times more uranium in the target and exhibits a lower production yield with respect to a target of highly enriched U-235 (HEU).
Further attempts have been presented for the production of Mo-99 through other channels. WO2006/028620 A2 relates to the production of Mo-99 from Zr-96 through irradiation with (charged) alpha particles, e.g. in a cyclotron or a linear accelerator. This document teaches the irradiation of the zirconium-96 target with alpha particles to yield, via an alpha particle capture/neutron emission process, an irradiated target containing Mo-99.
Finally, in view of the current global shortage in supply of Mo-99 radioisotopes, it has to be noted that even the switch from Tc-99m to other radioisotopes such as F-18 [FDG] have been considered. However, the major drawback of such a switch is the fact that all the current and well established imaging protocols with Tc-99m need to be reworked and readjusted for FDG, which will be time consuming or even not feasible at all for some imaging procedures. Furthermore, the costs of FDG radiopharmaceutical itself need to be lowered to compete with current Mo-99 based radiopharmaceuticals.

Technical problem

In view of the above, it is an object of the present invention to provide an alternative method for the production of Mo-99 which preferably does not present the above disadvantages or drawbacks, or at least to a lesser extent. The method should preferably yield the Mo-99 in commercially significant amounts.
This object is achieved by a method as claimed in claim 1.

General Description of the Invention

To achieve this object, the present invention proposes a method for the production of Mo-99 comprising the steps of

(a) providing a target fissile under charged particle bombardment
(b) irradiating at least a portion of the target with charged particles to induce fission of said elements, thereby producing an irradiated target portion comprising molybdenum-99.

Contrary to the method presented in WO2006/028620 A2, the present invention, does not start from a target element which is lighter than ⁹⁹Mo and which is irradiated with the "missing" or an excess of nucleons to yield the desired element. Hence, the present invention provides for an alternative production method of Mo-99.
In the present context, the charged particles are preferably selected from protons, deuterons, alpha or ³He²⁺ particles, but they could also be chosen among heavier charged particles that can be accelerated at the appropriate energy and intensity.
The starting materials for the target which are elements usable in the present method are selected from elements which are fissile under charged particle bombardment or combinations of two or more thereof. In the present context, the expression "elements" also comprises compounds of said elements, such as oxides, nitrides, etc. Furthermore, the elements can be in their natural abundance or enriched. Particularly preferred elements are selected from thorium-232 (Th-232), thorium-229 (Th-229), thorium-230 (Th-230), uranium-238 (U-238), uranium-235 (U-235), uranium-233 (U-233), radium-226 (Ra-226), bismuth (Bi-209), lead (Pb) or any combination thereof, even more preferably from thorium-232 (Th-232), thorium-230 (Th-230), uranium-238 (U-238), radium-226 (Ra-226) or any combination thereof. Still more preferably, the target essentially or totally consists of these preferred or even more preferred elements, either in metal, oxide, nitride form, or any combination thereof.
The starting material can be fabricated into various target configurations to enhance the production and recovery of the desired species, such as Mo-99 (or even technetium-99m). The solid or liquid target comprising the target nuclide(s) is generally loaded in a target holder adapted for use with the accelerator. The target nuclide(s) is/are in any appropriate form, preferably in metal or oxide form. The charged particle beam is preferably generated using an accelerator, such as in a cyclic and/or linear accelerator, e.g. a biomedical cyclotron.
In practice, the irradiation energy and target configuration may be selected such that the efficiency or economics of the overall process is optimal. The charged particles in the present method generally need incident energy above 15 MeV all charged particles included. The beam energy of the charged particle beam must be set for a maximum Mo-99 yield production. At the moment of the invention, there are very few experimental cross sections for charged particles induced fission, especially for Mo-99 fission product. The Mo-99 production yield is directly proportional to the charged particle beam intensity, so the higher is the intensity of the beam the higher is the production yield of Mo-99. This is to say that accelerators (cyclotrons or linear accelerators) of high beam intensities are necessary for large production of Mo-99.
The irradiation can thus be accomplished by inserting the target into any accelerator capable of producing charged particle beams of the desired energy and beam intensity. For a large commercialisation of Mo-99, accelerators of beam intensities in the range 100 µA-100 mA equipped with suitable target systems able to stand under high beam intensities would be optimum.
There are multiple advantages with respect to the method presented above. First of all, many accelerators are already available around the world, allowing for a regional production (larger production site for regional distribution) or even for a local production (smaller cyclotrons such as those operated in hospitals). Second, a significant advantage thereof relates to the decreased safety issues and constraints in relation with transport because the quality and quantity of radioactive material can be improved and reduced, respectively. Furthermore, a fast development of medical cyclotrons, including innovative isotope production methods with ion accelerators, has been observed in the last two decades, which in turn will warrant a greater security of supply of the Mo-99, because the supply is based on many rather then few production sites.
In a still further embodiment, the method further comprises one or more physical and/or chemical separation step(s) for separating molybdenum-99 from the irradiated target portion to yield Mo-99 and a purified target portion. A "purified target portion" in the present context means a target from which at least a substantial amount of the desired Mo-99 has been removed. Further separation steps may be included if necessary to eliminate possible further impurities or unwanted compounds and elements.
The physical and/or chemical separation step(s) are preferably chosen among ion exchange, electrolysis, extraction and sublimation or any combination thereof. Such step may include the oxidation or reduction of Mo-99 or the conversion to a salt form thereof.
In a further embodiment of the present method, the purified target portion is recycled in step (a) to produce further Mo-99.
Mo-99 compositions produced as described above are preferably used to "generate" Tc-99m. In fact, Tc-99m is the result of the decay of Mo-99 as a parent nuclide, which has a half-life of 66h. Hence, one embodiment of a method for producing Mo-99 includes the so-called "generation" of Tc-99m by any appropriate method.
In one example of such a method, a chromatographic generator column is charged with an alumina adsorbent. The adsorbent is then equilibrated using a salt solution, such as an ammonium nitrate or saline solution. Mo-99 is loaded on the column, typically at a pH of from about 3 to about 4. Tc-99m is then eluted from the loaded column using a salt solution. The eluted Tc-99m solution may be used without further purification. However, in certain embodiments, the Tc-99m solution can be further purified if desired, for example, by loading onto a Tc-99m concentrator column containing an anion-exchange resin known to those of ordinary skill in the art. The concentrator column typically is washed with a small amount of salt solution, followed by a small amount of deionized water. Tc-99m can also be eluted from the column using a reductive solution, such as a solution containing a complexing agent.
A further aspect of the invention is the use of the Mo-99 and/or Tc-99m obtained by a method as described above in the preparation of radiopharmaceuticals and/or radiation sources, which contain molybdenum-99 and/or technetium-99m, for use in nuclear medicine or Single photon emission computed tomography (SPECT) applications.
Medical uses for Tc-99m are, for example, bone scan, myocardial perfusion imaging (MPI), functional brain imaging, immunoscintigraphy, blood pool labeling, pyrophosphate for heart damage, sulfur colloid for spleen scan, etc.
As a conclusion, the different aspects of the present invention describe the irradiation with charged particles of a target material comprising elements which are heavier than Mo-99, the selective recovery of Mo-99 from the irradiated target material, as well as the fabrication of new or recycled targets from such recovered target material. Such a method can produce sufficient quantities of Mo-99 and thus of Tc-99m for use in the fields on nuclear medicine and/or nuclear chemistry. Finally, the above presented method allows to minimize the expenses in relation with preparing molybdenum-99 or technetium-99m.

Description of the Figure

Figure 1: A gamma-ray spectrum of a test irradiation (proton beam on Th-232) performed at the JRC-Cyclotron showing the gamma-ray peaks of the produced Mo-99.

Detailed Description of the Invention

The invention proposes the production of Mo-99 by bombarding a heavy material target with accelerated charged particles. Under the bombardment the fission products are produced and Mo-99 is among these products. Hence, in a preferred embodiment of the invention, the steps for the production of Mo-99 may be summarized as follows:

1 - Preparation of the target material made of elements or combinations of the heavy elements mentioned above. The target can be solid as a thick disk or liquid. The target should be encapsulated appropriately to avoid any dispersion of fission products.
2 - Irradiation of this target with accelerated charged particles in a cyclotron or linear accelerator. This generally includes the installation of a suitable target system equipped with a cooling system and a remote uploading and down loading of the target from the beam line to the "hot" cell where the radiochemical process for the extraction of Mo-99 takes place.
3 - The radiochemical process that can preferably be performed by programmable robot. This process preferably includes the reprocessing of the irradiation target and extraction of the fissile material for further irradiation. The radiochemical process is well established as it is quiet similar to that the extraction of Mo-99 from a nuclear reactor irradiated U-235.
4 - Preparation of high purity Mo-99 as a radiochemical which can include its loading in a chromatographic column.

Taking in consideration some available experimental fission cross sections, a calculation estimates that with an irradiation of a Th-232 disk of 800 µm thickness with a proton beam of 22 MeV would yield 3,7 MBq/µA.h of Mo-99 which makes for an irradiation of the similar Th-232 disk with 1 mA and 10 h a yield of 37 GBq (1 Ci) which could satisfy the needs of several hospitals.
A test irradiation of 1 h performed at the JRC-cyclotron (Ispra, Italy) of a disk of Th-232 of 1 mm thickness (about 0.5 g of Th-232), carried out with a proton beam of 30 MeV energy and about 1 µA intensity yielded about 4 MBq at the end of bombardment which confirms the order of magnitude of the production level of Mo-99 from Th-232 fission with proton beam (energy below 30 MeV).
A gamma-ray spectrum of the activated Th-232 showing the gamma-ray peaks of Mo-99 is presented in Figure 1. In addition, the Mo-99 was also identified with its half-life deduced from the measurement of Mo-99 activity at different times during several days after the irradiation test. As the fission cross section is increasing versus the proton beam energy, a significantly higher yield of Mo-99 is expected with charged particle beam of higher energies (>30 MeV). With a proton irradiation of Th-232 with higher beam intensities (>100 µA) with suitable beam energies, Mo-99 yields of tens of Ci (tens of 37 GBq) can be reached.

Claims

A method for the production of molybdenum-99 comprising the steps of
(a) providing a target comprising elements or combination thereof which are fissile under charged particle bombardment,

(b) irradiating at least a portion of the target with charged particles to induce fission of said elements, thereby producing an irradiated target portion comprising molybdenum-99.
The method as claimed in claim 1, wherein the charged particles are selected from protons, deuterons, alpha/³He²⁺ particles.
The method as claimed in any one of claims 1 or 2, wherein said elements that are fissile under charged particle bombardment selected from thorium-232 (Th-232), thorium-229 (Th-229), thorium-230 (Th-230), uranium-238 (U-238), uranium-235 (U-235), uranium-233 (U-233), radium-226 (Ra-226), bismuth (Bi-209), lead (Pb) or any combination thereof, even more preferably from thorium-232 (Th-232), thorium-230 (Th-230), uranium-238 (U-238), radium-226 (Ra-226) or any combination thereof.
The method as claimed in any one of claims 1 to 3, wherein the charged particles have an incident energy of a minimum of 15MeV for any of the used charged particle beam.
The method as claimed in any one of claims 1 to 4, wherein the irradiation is effected in a cyclic and/or linear accelerator.
The method as claimed in any one of claims 1 to 5, further comprising one or more physical and/or chemical separation step(s) for separating molybdenum-99 from the irradiated target portion to yield molybdenum-99 and a purified target portion.
The method as claimed in claim 6, further comprising one or more physical and/or chemical separation step(s) for separating technetium-99m produced by decay from molybdenum-99.
The method as claimed in claim 6 or 7, wherein the physical and/or chemical separation step(s) are chosen among ion exchange, electrolysis, extraction and sublimation or combinations thereof.
The method as claimed in any one of claims 6 to 8, further comprising a step of recycling the purified target portion in step (a).
Use of the molybdenum-99 obtained as claimed in any of claims 1 to 9 and/or the technetium-99m obtained as claimed in any of claims 7 to 9 in the preparation of radiopharmaceuticals and/or radiation sources, which contain molybdenum-99 and/or technetium-99m.