The present invention relates to a device for the production of reactor based medical radioisotopes used for example in medicine and to a related method of producing such radioisotopes. The present invention relates in particular to a device for the local production of radioisotopes. Radioisotopes are extensively used in medicine, industry, biology and agriculture, the medical uses of radioisotopes cover diagnosis and therapy as well as clinical research. With the use of radioisotopes in the diagnosis, ongoing biochemical processes can be displayed in living organs without damaging them. When injected in the body, they are called radio-tracers, because they release energy from its spontaneous radioactive decay, that can be directly detected by instruments external to the body to obtain information about the physiological, cellular and molecular process of interest.. The half-life of a radioisotope must be long enough to allow transport from production sites to end-use locations without excessive loss, and short enough to minimize the unwanted radiation dose to the patient after the procedure is complete.

Radioisotopes are produced by nuclear reactors or particle accelerators, on so-called target materials, that can be natural substances or materials enriched with stable isotopes. Nearly 80% of the radioisotopes used in nuclear medicine are currently produced by nuclear reactors. Among them, 99mTc (6.01 h half-life), a decay product of 99Mo (65.94h half-life), is the most important and widely used. This amounts to 7 million diagnoses per year in Europe and 8 million per year in the USA(about 12 kCi per week). Among ten major nuclear reactors that produce 99mTc from crude isotope 99Mo, European countries operate five - HFR (The Netherlands), OSIRIS (France), BR2 (Belgium), MARIA (Poland), and LVR-15 (Czech Republic), Canada operates - the NRU at Chalk River, South Africa - Safari-1 , at Pelindaba Site (near Pretoria), Australia - OPAL in ANSTO Research facility, Indonesia - the MPR RSG-GA in Java, and Argentina - the RA-3 in Buenos Ares. Since the early 2000s, there has been a shortage of 99mTc, widely used to diagnose cardiac and respiratory diseases, and neurological conditions. There has been frequent maintenance shutdown of the NRU and HFR, leading to shortage of 99Mo, the raw material required for production of 99mTc. In addition, a number of the producing reactors in the current ageing fleet are scheduled to be removed from the supply chain, with the earliest scheduled for 2015 (OSIRIS) and 2016 (NRU) due to technical setbacks. There is thus a need for a device and method allowing for a local production of radioisotopes, for example in hospitals, industry and/or research centers, in order to facilitate and guarantee supply of such isotopes at competitive affordable market price. Transmutation byAdiabatic Resonance Crossing (TARC) technique has been proposed by C. Rubbia (Resonance enhanced neutron captures for element activation and waste transmutation, CERN-LHC/97-0040EET, 1997; TARC collaboration, Neutron-driven nuclear transmutation by adiabatic resonance crossing, CERN-SL-99-O36EET, 1999; Abanades et al., Nuci. Instr. and Meth. A487 (2002) 577) for element activation and orfor radioactive waste transmutation.

PCT application WO 98/59347 describes in detail how the method of Adiabatic Resonance Crossing works where a material to be activated is exposed to a neutron flux by distributing it in a neutron-diffusing medium surrounding a neutron source. The method of neutron activation described in this document is intended to be a competitive alternative to reactor-driven neutron capture activation. In additbn, several isotopes which are difficult to produce by activation with the usually thermal neutrons of an ordinary reactor can be produced using the broad energy spectrum of the neutrons in the activator, extending to high energies and especially designed to make use of the large values of the cross-section in correspondence of the resonance region. A major drawback of this method and of the corresponding activator, however, is that the dimensions andor the material of the activator must be adapted to the material to be activated. In particular the distance between the samples of material to be activated and the neutron source must be dimensioned such that the energy of the neutrons reaching the material samples is attenuated to an energy adapted to the nature of the material to be activated and to the type of activation desired. For the production of different radioisotopes for medical applications, for example, there must thus be one activator per radioisotope to be produced, or the activator must be specifically modified for each radioisotope to be produced, which would be too expensive and/or too complicated for most hospitals.

The method was demonstrated with the TARC experiment at CERN in 1997 and a notable drawback is its low yield (seeS. Buono, e. a. (2007). Design and test of an accelerator driven neutron activator at the JRC). Another drawback is the enormous amount of lead and iron allocated to the device. An aim of the present invention is thus to provide a method and an activator that do not have the drawbacks of prior art activators.

An aim of the present inventbn is in particular to provide a method and an activator that enhances captures in the resonance region and allows the activation of a particular radioisotope.

Another aim of the present inventbn is to provide a method and an activator achieving high yields with a deuteron beam of low energy and low intensity.

Still another aim of the present inventbn is to provide an activator whose dimensbns allow its installation and use near the utilization sites of its products, for example in hospitals.

Yet another aim of the present invention is to provide an activation method and an activator allowing for the production of different isotopes in a same device, without having to adapt the dimensions and/or the material of the device. These aims and other advantages are achieved by an activator and a method according to the corresponding independent claims.

These aims and other advantages are achieved in particular with an activator for the production of radioisotopes through activation of a material to be activated by submitting said material (9) to a neutron radiation, the activator comprising a target for emitting neutrons when the target is submitted to a particle beam; a diffusor surrounding the target for allowing diffusion within the diffusor of the neutrons emitted by the target; a reflector surrounding the diffusor for avoiding neutron leakage out of the diffusor by elastic scattering; a moderator surrounding the diffusor for slowing down and reflecting neutrons that would have leaked outside the reflector; a shielding surrounding the moderator for capturing neutrons escaping through the reflector and the moderator; wherein the diffusor comprises a plurality of grooves for each holding a sample holder within the diffusor, such that a material to be activated and contained in the sample holder is submitted to the radiations of neutrons diffusing within the diffusor, wherein the plurality of grooves are positioned at distances from the target different from each other.

Preferably, the positions of the grooves within the diffusor are determined to be where the resonance neutron capture probability is maximized. The positions are for example determined with the help of simulatbn software and or through experimentation.

In embodiments, the sample holder comprises a tube and a capsule located inside the tube, the capsule being configured for containing a sample of the material to be activated.

In embodiments, the sample holder further comprises local moderators and/or reflectors located in the tube, adjacent to the capsule, for example on either side of the capsule.

The capsule is for example made of graphite. The local reflectors and/or moderators are for example made of graphite and/or of polyethylene.

The target is for example a Beryllium target. The diffusor is for example made of lead (Pb), for example of at least 98% pure lead.

The reflector is for example made of carbon.

The moderator is for example made of polyethylene.

The shielding is for example made of borated polyethylene. These aims are also achieved by a method for determining the positions of grooves for holding sample holders in such an activator, comprising the steps of: modeling the activator in a Monte Carlo simulator; simulating the emissbn of neutrons by the target within the diffusor; determining positions within the diffusor where the resonance neutron capture probability is maximized.

In embodiments, the method further comprises the steps of building the activator; positioning the grooves at the determined positions; correcting the positions based on measurement results of radioisotope yields.

In embodiments, the method further comprises the steps of positioning sample holders containing a material to be activated in grooves, wherein the grooves are positioned within the diffusor at different distances from the target; activating the target by submitting it to a particle beam. In embodiments, the method further comprises the step of locally capturing neutrons diffusing within the diffusor around the material to be activated through local reflectors and/or moderators.

The activator of the present invention and the corresponding method thus allow extending the practical utilization of activator based on the ARC technique thanks to its high neutron capture efficiency, which pemnits to produce the required amount of the radio-isotope with a relatively modest neutron generator. Furthermore, the diffusor being surrounded by the reflector, the neutrons are captured within the diffusor and form a neutron cloud therein, such that the samples may be positioned at different distances from the target, not necessarily at a distance corresponding to the optimal attenuation of their energy. Furthermore, the structural simplicity of the device proposed and its relatively modest cost and dimensions allow "local" production of short-lived radioisotopes.

The advantages of an accelerator-based approach can therefore be brought to bear thanks to the present inventbn, based on the following elements: The neutron source can produce a high flux of neutrons,

The strength of the flux can be optimised by the geometry of the diffusor and reflector surrounding the target thus boosting neutron yield,

An optimised neutron spectrum make full use of the resonance peaks for neutron capture in the epithemnal region,

The absence of water-cooling, in embodiments, makes it easier to optimise the spectrum to suit the specific isotope being produced.

The density of the neutron flux and the efficiency in its utilisatbn are essential to make an accelerator-based technology viable. The present invention makes use of a compact neutron source surrounded by a lead transparent diffusing material and adapting emerging neutrons to the neutron spectrum to enhance capture in the resonance region, as it would be necessary, for example, with 98Mo samples surrounded by the compact neutron source.

The present invention will be better understood by reading the detailed des iption illustrated by the figures where:

Figure 1a is an internal schematic view of an activator according to a preferred embodiment of the invention;

Figure 1b is a general view of an activator according to a preferred embodiment of the invention; Figure 2 is a schematic cut view of the activator of figure 1 b;

Figures 3a to 3i illustrate the various elements of an activator according to a preferred embodiment of the invention and their assembly, with some partial cut views;

Figure 3j is a cut view of the assembled activator of Figs 3a-3i;

Figure 3k is a schematic cut view of a sample holder according to an embodiment of the invention;

Figure 4 shows an example of the geometric implementation of an activator according to a preferred embodiment of the invention in a simulation package; Figure 5 shows the simulated neutron flux distribution in an activator according to a preferred embodiment of the inventbn;

Figure 6 shows the simulated photon flux distribution in an activator according to a preferred embodiment of the inventbn; Figure 7 shows the neutron flux spectrum in the lead of an activator according to a preferred embodiment of the inventbn;

Figure 8 shows the neutron flux spectrum in the borated polyethylene of an activator according to a preferred embodiment of the inventbn;

Figure 9 shows the equivalent dose rates in the target and shielding of an activator according to a preferred embodiment of the inventbn for several cooling periods;

The activator of the present invention is based on the principles of theAdiabatic Resonance Crossing (ARC) method, for the industrial production for example of 99mTc and/or of other radioisotopes that are widely used for example in medical applications.

According to a preferred embodiment, and with reference to figures 1 a, 1 b, 2 and 3a-3j, the activator of the invention comprises several components having different functions:

Base Frame 1 and diffusa' support 2 (Fig. 3a)

The base frame 1 is the supporting structure of the activator. It allows installing the activator on site and preferably provides for an homogenous distribution of the weight of the activator. It preferably allows the correct positioning of the other elements of the activator relative to each other. The base frame 1 is for example made of steel, preferably with a low cobalt (Co) content, or any other suitable material. The contribution of the base frame 1 to the effect of the activator being negligible, it may be made of any material sufficiently rigid and solid for properly supporting the activator. The base frame 1 preferably comprises feet 11 that are adjustable in height, thereby allowing the correct alignment of the activator. The base frame 1 is preferably as large as the whole activator. The diffusor support 2 is placed on the base frame 1 , preferably in its center. It comprises a support platform 20 on feet 21 that keep the platform 20 at a predetemnined distance over the base frame 1. The diffusor support 2 is for example made of steel, preferably with a low cobalt (Co) content, or even preferably of a material that may not be activated by neutrons. The diffusor support 2 is configured for supporting heavy weights, in particular the diffusor support 2 is configured to support the lead forming the diffusor of the activator. In embodiments, the diffusor support 2 allows centering the lead of the diffusor relative to the rest of the activator's structure. The diffusor support 2 is preferably made of a rigid material and its thickness is preferably reduced to the minimum in order to avoid parasitic neutron captures by the support's material. The contribution of the diffusor support 2 to the effect of the activator, in particular to the production of 99Mo is negligible.

Diffusor

With reference to Fig. 1 a, the diffusor 3 is located above the base frame 1 , preferably on the support platform 20 of the diffusor support 2. The diffusor 3 is made of a neutron transparent material, for example of lead (Pb), for example of at least 99.8% pure lead, preferably of at least 99.9% pure lead, even more preferably of 99.99% pure lead. The diffusor 3 is for example made of a plurality of bricks 31 , wherein each brick 31 is for example machined to shape. In embodiments, the diffusor 3 comprises holes or grooves 30 for holding sample holders that may contain a material to be activated in order for example to produce a desired radioisotope by submitting it to neutron radiation according to the principles of the ARC method. The grooves 30 are for example machined in one or more bricks 31 of the plurality of bricks. In embodiments, each brick 31 preferably weighs at most 40kg in order to facilitate their manipulation, for example when building the activator. The one or more sample holders and the target are preferably placed in the diffusor 3, in particular when the activator is in use. The diffusor 3 is seating on the diffusor support 2. which is also installed on the base frame 1. The diffusor 3 allows neutrons produced by a neutron source located in or near the diffusor

3, for example a target of beryllium (Be), to diffuse inside of the activation volume of the activator, thereby homogenizing the corresponding neutron flux. This allows covering the entire activation volume, or at least most of it, and placing a maximum of samples to be irradiated in the activation volume, thereby maximizing the probability of capture in the irradiated samples. Diffusion of the neutron flux must be done without losing neutrons that would be captured by the diffusor material. The diffusor material must thus be selected with a very low effective neutron capture cross section. Ideally, the diffusor material should be pure lead. Lead Sb (%) Sn (%) As (%) Bi (%) Cu (%) Ag (%) Cd (%) Ni (%) Zn (%)

99.97% 0.001 0.001 0.001 0.03 0.003 0.005 0.001 0.001 ο.οω5

99.99% 0.0005 0.0005 0.0005 0.01 0.0005 0.0015 ο.οω2 0.0002 ο.οω2

Table 1 - Lead composition at 99.97% and 99.99% respectively.

Less pure lead, for example from 99.00% to 99.9%, may be selected if the proportion of impurity is negligible, i.e. the quantities of Bi, Cd, Ag, Sn, etc., are less than 10 ppm.

The minimum emerging kinetic energy T _min , or maximum energy loss, of a neutron of energy T _Q in collision with a nucleus of atomic number A is given by

Which suggests that the largest possible A minimizes the rate of energy loss. For large A, isotropic scattering is an excellent approximation. The average, logarithmic energy decrement ξ is then

< T _m ' _in > (A - ξ = -1η ≡— ₌ i - l) ² (A

in ( — + 1\ )

T _Q 2A A - 17

The logarithmic energy decrement for lead is very small: ξ = 9.54x10 ^"3 . The use of lead also allows "slowing-down" the energy neutron source in a gradual and controlled way throughout the elastic scattering of these neutrons.

Lead is a fully diffusive medium, i.e. neutrons have a relatively high average free path in lead, which implies a difficulty, in principle, to contain neutrons within a finite volume. For example in the TARC experiment, a volume of 3x3x3 m ³ was required in order to contain the produced neutrons. It implies that the neutron flux strongly reduces as one moves away from the source (Φ _η ~ 1/r ² ) The probabilities for samples to be irradiated of capturing neutrons thus also decrease in the same proportion ( _εαρ ~Φη~ 1/r ² ), which is probably why the rate of the prior art TARC yields were very low.

In order to simplify and compact the system, the size of the diffusor 3 of the invention has been significantly reduced compared to the volumes required in the prior art TARC experiments, thereby limiting the activation volume in a region near the target neutron source where the neutron flux is high.

Neutron sources, for example emitted by a Be target or resulting from the deuterons backup, have an average energy of 5.7 MeV. At this energy, the cross section (n, 2n) of lead is very high, typically of 1-2 bams, which also allows increasing the initial flux of the neutron source through the reactions (n, 2n) in the lead diffuser.

Accordingly, the diffusor 3 diffuses and homogenizes the neutron flux within the activation volume. It also allows increasing the number of neutrons through (n, Xn) and moderating these neutrons through elastic collisions.

Reflector The reflector 5 surrounds the diffusor 3, preferably on all sides. In embodiments, the reflector

5 is made of graphite, for example graphite blocks. Each block is for example made to a proper shape to surround at least part of the diffusor 3, so that the reflector 5 surrounds the diffusor 3, preferably on all sides, when the blocks are assembled around it. In embodiments, each block preferably weighs at most 40kg to allow their manipulation while the activator of the invention is constructed. The graphite blocks are for example staggered together with metallic or tissue straps.

The reflector 5 redirects neutrons that have leaked outside the activation volume of the diffusor 3 inwardly thereof. This allows containing the neutrons inside the diffusor 3, thereby increasing the neutron flux inside the activation volume and thus the probability of capture of neutrons in said activation volume. The reflector 5 is preferably made of a light nucleus material in order to increase the reflection, cf. the scattering angle equation:

Beryllium (Be) and graphite are for example particularly appropriate for forming the reflector because neutrons travelling out of the diffusor are reflected in the good energy (epi-themnal) (5 cm Be = 15 cm Graphite). Beryllium, is more effective than graphite, but it is also significantly more expensive and difficult to machine.

In embodiments, simulations have shown that a minimum thickness of 15 cm of graphite is efficient for maximizing the capture rate in 98Mo. Other dimensbns are however possible within the frame of the invention.

According to the invention, the reflector 5 allows reflecting back into the diffusor 3 neutrons that would have leaked out of it, thereby confining, or containing, the neutrons inside the diffusor 3, i.e. inside the activation volume. The reflector 5 surrounding the diffusor 3 preferably on all sides, it creates a "Ping-Pong" effect on the neutrons travelling through the relatively small diffusor 3 from one extremity to the other, thereby increasing the local neutron flux inside the activation zone.

In embodiments, the reflector 5 is made of graphite or beryllium, which furthermore allows slowing down, or moderating, the epi-themnal neutron energy, where the TARC capture resonance lies. The graphite or beryllium reflector 5 thus participates to and accelerates the TARC captures. In embodiments, for example with a reflector 5 made of graphite and having a thickness of

15 cm, the reflector 5 can not redirect all neutrons, some of the fastest neutrons having for example enough energy to pass through the reflector 5. There is thus a need for shielding around the reflector 5.

Moderator A moderator 6 made for example of polyethylene, for example of high-density polyethylene, surrounds the reflector 5. The moderator 6 allows slowing down the fast neutrons that may have passed through the reflector 5 by reflecting them back inward, toward the activation volume, through the reflector 5 again. In embodiments, the moderator 6 is made of blocks, each block being made to a proper shape to surround at least part of the reflector 5, so that the moderator 6 surrounds the reflector 5 preferably on all sides when the blocks are assembled around it. In embodiments, each block preferably weighs at most 40kg to allow their manipulation while the activator of the invention is constructed. In embodiments, the blocks, for example of polyethylene, are staggered together with metallic or tissue straps.

The moderator 6 thus slows down and reflects most neutrons that would have leaked outside the reflector 5.

Shielding

A radiological shielding 60 surrounds the moderator 6. In embodiments, the shielding 60 for example comprises a layer of high-density borated polyethylene or of any other adapted material. In embodiments, the shielding 60 is made of blocks or plates, each block or plate being made toa proper shape to surround at least part of the moderator 6, so that the shielding 60 surrounds the moderator 6 preferably on all sides when the blocks are assembled around it. In embodiments, each block or plate preferably weighs at most 40kg to allow their manipulatbn while the activator of the invention is constructed. In embodiments, the blocks or plates, for example of borated polyethylene, are staggered together with metallic or tissue straps.

Some neutrons, in particular the thermal energy neutrons, are difficult to reflect and therefore may pass through the reflector 5 and the moderator 6. They are thus rather absorbed in the shielding 60 by their capture in hydrogen or boron. The shielding 60 thus absorbs neutrons that would have leaked outside the moderator 6, thereby minimizing, preferably completely preventing, neutrons escaping from the activator of the inventbn, which allows its integration in for example a hospital environment, where radiological impact on the users should be minimized.

In embodiments, the moderator 6 and the shielding 60 are formed with the same blocks, at least part of the blocks, in particular the external blocks, being made of two materials, for example polyethylene for the moderator 6 and borated polyethylene for the shielding 60, such that when the blocks are assembled around the reflector 5, the assembled blocks form two successive layers of different material that surround each other. In still other embodiments, the blocks similarly allow fomning at once the reflector 5 and the moderator 6, or the reflector 5, the moderator 6 and the shielding 60.

Target The target assembly 7 comprises a beam tube 70 and a target 71 , for example in Beryllium.

In embodiments, a water circuit, not represented in the figures, actively cools the target 71.

The target 71 converts into neutrons charged particles that have been accelerated by cyclotron and directed to the target 71 through the beam tube 70.

In embodiments, the charged particles are deuterons with an energy of 15-25 MeV, that are broken-up by the target 71 into a proton and a neutron. The fast proton is stopped in the target 71 because the range of a 7.5 MeV proton in Beryllium (Be) is of few millimeters, and converted into heat. A small fraction of protons may be converted to neutrons through (p,n) reactions. The fast neutron escapes the beryllium target 71 and diffuse into the diffusor 3. In embodiments, a small fraction of the neutrons will be multiplied through (η,Χη) reactions inside the lead diffusor 3. The herein described choice of energy, deuteron and current, results in a cost effective way of creating a high intensity neutron source.

According to the invention, the target 71 is preferably configured to stop the protons and to remove the power deposited by the beam. Contrary to the prior artTARC method, which is based on spallation process Sample holder

According to embodiments, at least one sample holder 4 is inserted inside the diffusor 3. Each sample holder 4 comprises a tube 40 in which a sample material 9 to be irradiated is inserted. In embodiments, the tube 40 is nearly as long as the diffusor's width and or depth. The tube 40 is for example made of aluminium in order to avoid capture of neutrons by the tube itself, and closed with a SCTew-on cap of the same material, or of any other appropriate material. The sample holder 4 preferably allows placing a capsule 41 for irradiation, at different positions within the reactor, for example by inserting the sample holder 4 in different holes or grooves and by inserting the tube 40 at different depths within the diffusor 3 and/or by allowing the placement of the capsule 41 at different positions along the longitudinal axis of the tube 40. This allows maximising the capture reaction according to the local neutron flux spectrum.

The sample capsule 41 comprising the material 9 to be irradiated is for example made of graphite or of any other appropriate neutron reflecting material. In embodiments, the capsule 41 is placed at a determined position within the sample holder 4, for example at a determined longitudinal position within the tube 41 , while the rest of the sample holder 4 is filled with lead 42, preferably of the same composition as the lead of the diffusor 3. The sample holder 4 is inserted within the diffusor 3 through a preferably minimal opening of the radioprotective shielding 6 in order to minimize the potential radiation dose received by the user while inserting and r extracting the sample.

In embodiments, the sample holder 4 comprises two or more capsules 41 separated from each other by lead 42. The two or more capsules are for example placed each at a different position along the longitudinal axis of the tube and separated by lead which is inserted within the tube between the capsules.

According to the inventbn, the activation volume of the reactor is spread throughout the entire diffusor 3. Samples 9 of material to be activated, i.e. irradiated, are placed in capsules 41 made for example of graphite and then placed in sample holders 4 that may be distributed inside the entire diffusor volume. This allows matching the level of neutron flux and energy to a given sample material to be activated by placing the sample material 9 in the most appropriate region of the diffusor 3.

For instance, samples 9 with small cross-sections at high energy can be located near the target 71 where the flux is high and neutrons are fast, while samples with large cross-sectbns close to thermal energies can be located near the reflector 5 that surrounds the diffusor 3.

The activator of the invention thus allows activation of different sample materials having different characteristics by placing them each in different locations inside the activation zone, i.e. inside the diffusor 3. The activator of the invention furthermore allows the simultaneous activation of different material by placing each corresponding sample holder in the position(s) most appropriate to the activation of the particular material .

In embodiments, the sample holders 4 allow inserting local reflectors and/or moderators 43, made for example of graphite, polyethylene, beryllium, etc., in order to locally tune the neutrons' spectra and/or energy to the particular resonance region of the sample 9 to be activated. This allows for example using the entire volume of the diffusor 3 for the activation of a particular material by adapting the local reflector and/or moderator 43 to the tuning required in the particular area of the activation zone. The local reflector and/or moderator 43 comprises for example the capsule 41 that contains the material 9 to be activated, the capsule 41 being for example made of a reflecting and/or moderating material.

The local moderator and/or local reflector 43, acts similarly to what was des ibed above in relation with the reflector 5 and/or the moderator 6. Accordingly, the local moderator and/or reflector 43 allows locally increasing the neutron flux inside the sample holder 4 through the multiple reflection process, i.e. the "Ping-Pong" or boosting effect described above, and/or allows slowing down the neutrons to the appropriate energy level . The dimensions and composition of the local reflector and/or local moderator 43 are preferably adapted to the sample's characteristics, in particular to its position of resonances.

In an embodiment, the assembled activator comprises a lead diffusor 3 having for example dimensions of 0.6 x 0.6 x 0.8 m ³ and weighing approximately 3.27 tons, a reflector 5, for example a graphite reflector with a thickness of 0.15 m, a moderator 6, made for example of 0.3 m of polyethylene and a shielding 60 made of 0.03 m of borated polyethylene. According to this embodiment, the assembled activator weighs approximately 8.2 tons with the base frame. This embodiment is described herein as an illustrative but in noway limiting example. Other dimensions, proportions, weights and/or materials are possible within the frame of the invention.

According to embodiments, several positions for inserting the sample holders 4 in the activator assembly have been considered. Figure 1b for example illustrates an embodiment with a total of ten different positions. The positions of the holes 30 for the sample holders 4 are preferably located at positions that have been determined, for example with the help of simulation software, to be where the resonance neutron capture probability is maximized.

In embodiments, the purity of the lead of the diffusor 3 is very high, for example approximately 99.99%, in order to avoid unexpected activation of unwanted materials within the diffusor. According to the invention, the activator comprises a reflector 5 surrounding the diffusor 3, a moderator 6 surrounding the reflector 5 and a shielding 60 surrounding the moderator 6. In embodiments, the reflector 5 comprises a graphite layer, the moderator 6 comprises a polyethylene layer and the shielding 60 comprises a layer of borated polyethylene. The reflector 5, the moderator 6 and the shielding 60 preferably completely surround the diffusor 3. The reflector 5, for example a graphite reflector, surrounding the diffusor 3, for example a lead diffusor, reduces neutron leakage out of the diffusor 3 by elastic scattering. Indeed, most neutrons travelling inside the diffusor 3 towards the reflector 5 hit the reflector 5 and elastically bounce back towards the diffusor 3. The moderator 6, for example a polyethylene moderator, reduces the energy of the neutrons that travelled through the reflector 5 and redirects them towards the reflector 5. The shielding 60 for example comprises a layer of borated polyethylene, made for example of sheets of borated polyethylene glued to the outer surface of the moderator 6 to further enhance radiation protection. The shielding 60 thus serves to efficiently capture the remainder of the neutron radiation. The target 71 , for example a beryllium target, is preferably placed at the center of the diffusor 3, and or of the activator assembly, at the end of the tube 70, for example a proton beam transport tube. The target 71 for example produce a high neutron flux when bombarded for example with a 65 MeV proton beam (40 μΑ). The neutrons produced by the target 71 diffuse in the diffusor 3. Elastic scattering on the diffusor's material, for example lead, progressively slows down the neutrons, such that their energies 'scan' the region where neutron resonances occur (in the eV range).

Samples of material 9 to be activated are properly located at various positions in the diffusor 3, preferably where the resonance neutron capture probability is maximized. In embodiments, the samples of materials to be activated are placed in capsules 41 , made for example of graphite and/or of polyethylene, that are positioned in tubes 40, made for example of aluminum, which are in turn inserted in the diffusor 3, through the shielding 60, the moderator 6 and the reflector 5. The graphite and/or polyethylene of the capsules 41 , that surrounds the sample material to be activated, allows optimizing neutron capture in the samples by creating a local reflector and or moderator. The capsule 41 for example comprises layers of graphite and/or polyethylene of predetermined thickness in order to achieve the desired concentration of neutrons through reflection inside the capsule andor the desired energy of the neutrons by moderating the neutrons that travel inside the capsule 41.

The method of the invention using the activator of the invention will be described below in relation with the production of radioisotopes. In embodiments, thanks to its relatively modest dimensions and simple mechanical structure, the activator of the inventbn is aimed at the "local" production of short-lived radioisotopes, for example at the production in hospitals of radioisotopes for medical use. This is made possible thanks to the high efficiency neutron capture achievable by the ARC method that allows obtaining the required number of radioisotopes with a modest neutron source.

One of the medical isotopes most widely used today is 99mTc, which are widely used for bone scans, cardiac perfusion studies and other diagnostic procedures.

The use of "generators" such as 99Mo/99mTc, includes more long-lfved parents (2.75 days 99Mo), which breaks down into shorter-lived daughter (6-hour 99mTc). In the corresponding exemplary but not limiting embodiments, the activator comprises:

- a centrally located beryllium neutron convertertarget 71 that emits neutrons for example under a proton beam or a deuteron beam;

- a neutron diffusor 3 made of highly pure lead, for example 100% Pb Nat, surrounding the beryllium target 71 , for distributing the neutrons homogeneously in space and energy through multiple elastic scattering.

- an isotope sample 9 suitable for an η-γ reaction leading to the production of medically useful radioisotopes;

- a carbon reflector 5 surrounding the lead diffusor 3 serving to reduce neutron leakage out of the diffusor 3 by elastic scattering;

- a polyethylene moderator 6 and a borated polyethylene shielding 60 surrounding the reflector 5 to moderate the energy of the neutrons escaping through the reflector 5 and/or capture neutrons escaping through the moderator 6, thereby improving the reaction yield and reducing the dose rate to the environment during irradiation.

- a carbon block 43 and/or or capsule 41 , surrounding the isotope sample 9 to enhance neutron capture in the isotope sample through local moderatbn.

In variant embodiments, the material of the diffusor 3 is for example 3N (99.98%Pb) lead or 4N lead (99.99%Pb) instead of 100% pure lead. 3N lead has a small negative impact on the radioisotope yield, but is four times less expensive than 4N lead.

Pure lead (99.99%) is preferred to ensure that impurities have a negligible effect on the neutron flux. In practice, however, the experiment calls for high-purity lead devoid of such known impurities as silver, antimony, and cadmium. All the measurements were combined to obtain the best estimate of the impurity concentrations in lead.

Optionally, the production yield of radioisotopes was improved by optimizing the local moderators 43 consisting for example of a carbon block surrounding each sample 9 of material to be activated. Optimization implied for example varying the thickness and/or the position of the carbon layer relative to the sample.

.Experiments demonstrated that varying the positions inside the diffusor 3 of the samples 9 of materials had an influence on the production yield of radioisotopes. In embodiments, the initial positions determined through simulation of the neutron radiatbn inside the diffusor 3 is thus adjusted experimentally to optimize the production yield of radioisotopes.

In experiments, activation yields (Ho, Re, Lu, and 98Mo) after irradiation in different positions and different beam energies inside the activator, has been measured by γ-ray spectrometry using calibrated Ge detectors.

- The isotope 99mTc (T1/2 = 6h) is produced in the nuclear reaction 98Μο(η,γ)99Μο followed by disintegration to 99mTc. (Major 99Mo γ at 739.5 KeV).

- The isotope 186Re (T1/2 = 90.64h) is produced in the nuclear reaction 185Re(n,y)186Re (Major γ at 137 keV).

- The isotope 188Re (T1/2 = 18.59m) is produced in the nuclear reaction 187Re(n,Y)188Re (Major γ at 8.65 and 61.1 keV).

- The isotope 166Ho (T112 = 1.2d) is produced in the nuclear reaction 165Ηο(η,γ)166Ho (Major Y at 6.95 and 80.6 keV).

- The isotope 177Lu (T112 = 6.6d) is produced in the nuclear reaction 176Lu(n,Y)177Lu (Major Y at 208.4 keV).

Naturally occurring lutetium (Lu) is composed of one stable isotope Lu-175 (97.41 % natural abundance).

Figure 4 illustrates a modeled activator according to a preferred embodiment of the invention for use in a Monte-Carlo simulatbn, wherein same reference numbers designate same or equivalent elements as in the other figures. The precision of a Monte Carlo simulation depends strongly on the three-dimensional geometrical description of the simulated system and neutron cross-sectbn is taken from the latest compilations available.

In the illustrated simulated example, the neutrons produced in the target are for example captured in 98Mo targets to produce 99Mo.

The lead diffusor 3 is for example surrounded by a graphite reflector 5 with a density p = 1.7 g/cm ³ , to reflect the neutrons back to the Mo samples, polyethylene moderator 6, for example C2H ₄ with a density p = 0.93 g/cm ³ , and borated polyethylene shielding 60 with for example 10% of 10B with density p = 1.05 g/cm ³ .

The simulation is done using a deuteron beam of 40 μΑ. Figure 5 shows that the neutron fluxes go up to 10 ¹⁴ neutrons/cn /s in the target and decrease to approximately 10 ⁶ neutrons/cn /s after the shielding 6. The photon fluxes range from 10 ¹³ photons/cn /s in the target to approximately 10 ¹³ photons/cn /s outside the shielding, as illustrated in figure 6.

The neutron flux in the diffusor of the activator is shown in figure 7. The effects of neutron moderation are remarked in the lead, with a reduction in the flux of 7.5 MeV neutrons, and two "shoulders" centered at 0.05 eV and about 500 keV.

Outside the shielding, the benefit of using a polyethylene moderator with a shielding of borated polyethylene is illustrated in figure 8: the "shoulder" of low-energy neutrons disappeared completely, due to neutron absorption, and the fluxes are in general much lower when compared to the neutron fluxes in the lead.

The residual equivalent dose rates (Sv/h) in the assembly are presented in figure 9, for six cooling periods afterthe stoppage of the deuteron beam: one second, one hour, two hours, ten hours, 1 day and seven days. The results are presented twice, with average values for 200 cm and 8 cm thicknesses in the x direction. The dose rates outside the shielding 60 range from approximately 10 to 100 mSv/h one second after the beam stops to approximately 1 Sv/h ten days after, The highest residual dose rates are below the assembly, due to the activation of the stainless steel of the base frame 1.

The composition of the stainless steel used in the simulations (5235) is presented in the table below. c Mn P s N Cu Fe

0.17% 1.4% 0.04% 0.04% 0.012% 0.55% 97.788%

In order to confirm that these high residual dose rates are mostly due to the activation of stainless steel, simulations were performed again, replacing the stainless steel by lead in the structural components 1 , 2. The Monte Carlo simulations results confirm that the stainless steel has indeed a sizable contribution to the residual dose rates outside the shielding 60. Without this contribution, a dose rate of 1 Sv/h outside the shielding could be reached within one day after the stoppage of the beam, instead of seven days. Furthermore, the distribution of the dose rates is more "isotropic", centered on the target.

One second afterthe stoppage of the beam, the activities of 99Mo in the sample range from 1.97x10 ⁷ Bq/cm ³ to 6.73x10 ⁷ Bq/cm ³ .

The invention has been des ibed herein in relation with illustrative but in no way limiting examples of preferred embodiments. Other embodiments are however possible within the scope of the inventbn. In particular, the dimensions of the activator and its elements are given by way of example but may differ in various embodiments. The materials used for the varbus elements may also be modified within the scope of the present invention, provided that the replacement materials provide the same or similar function to said elements.