Nuclear data for light charged particle induced production of emerging medical radionuclides

1.1 Radionuclides in nuclear medicine

Nuclear Medicine is a medical field that uses radionuclides both for diagnosis and therapy. The physical characteristics of radioactive decay of radionuclides determine their use in nuclear medicine [1], [2], [3]. In most cases, radionuclides must be labelled to a vector molecule, forming what is commonly called a radiopharmaceutical. The vector molecule allows to efficiently target a dedicated cell type, due to its high affinity for it, or for a specific function of the human organ. In general, radionuclides employed in nuclear medicine both for diagnostic and therapeutic purposes, should meet the following specific requirements:

1. appropriate physical properties for the selected application, in terms of half-life (T_1/2), decay mode and radiation emission energy;

2. suitable chemical properties to allow their labelling to a vector molecule with high radiochemical yields through the labelling process;

3. ability to deliver an appropriate radiation dose to the target tissue, sufficient to accomplish the intended clinical duty (diagnosis or therapy), but without causing damages to the patient;

4. an appropriate purity, as the vector molecule can be expensive or the number of targeted sites limited;

5. a reasonable price.

A new frontier of nuclear medicine is the so-called “theranostic approach” that combines therapy and diagnosis stages (preferentially PET imaging that can give a quantitative dosimetry) [4], and which includes a single radiopharmaceutical that:

1. has both diagnostic and therapeutic properties, i.e. a bioactive molecule labelled with an isotope whose radiation decay pattern is suitable both for diagnosis and therapy, such as ⁴⁷Sc, ⁶⁷Cu, ⁶⁷Ga, ¹¹¹In, ^117mSn ¹²³I, ¹³¹I, ¹⁵³Sm, ¹⁷⁷Lu, ¹⁸⁶Re, ²¹³Bi, etc.; as previously noted, the use of gamma-emitters instead of positron-emitters is recently attracting considerable interest [4];

2. can be labelled with different radioisotopes of the same element, one being utilized for diagnosis while the other one for therapy, such as ^43/44Sc/⁴⁷Sc, ^61/64Cu/⁶⁷Cu, ⁶⁸Ga/⁶⁷Ga, ⁸⁶Y/⁹⁰Y, ¹²⁴I/¹³¹I, the terbium family ^149/152Tb/^155/161Tb, etc. [5];

3. can be labelled with different isotopes each belonging to a different element but having similar chemical properties (surrogate pair), where one radionuclide is utilized for diagnosis and the other for therapy, such as ⁶⁸Ga/¹⁷⁷Lu, ⁶⁸Ga/therapeutic lanthanide, etc.

The theranostic approach allows for collecting pre-therapy information with low-dose diagnostic imaging, aimed at optimizing the following high-dose therapeutic administration to the patient [6]. This strategy is getting closer to the realization of personalized medicine, in which low-dose molecular imaging performed with SPECT/CT or PET/CT is a necessary step to provide preliminary information about biodistribution, dosimetry on the tumor and limiting or critical organs or tissues, and the maximum tolerable dose [7].

Whatever the radionuclide to be used in nuclear medicine, it is thus essential to get knowledge of the possible presence of coproduced radionuclidic impurities that can have an impact on the final product quality and in the waste management during the production process or in the use at the hospital. Moreover, the availability of the majority of theranostic radionuclides is mainly limited by the fact that most of them are currently emerging and the optimal production route still needs to be improved or defined [8, 9].

1.2 The use of accelerators for emerging medical radionuclides production

Most of the radionuclides used in nuclear medicine are artificially produced by using either nuclear reactors or accelerators delivering charged particle beams or inducing a γ-irradiation. This work is focused on light-charged particles induced reactions, even if there is an increasing interest on photo-production as for example for ⁴⁷Sc, ⁶⁷Cu, etc. [10], [11], [12].

In order to select the optimal way to produce a given radionuclide, it is mandatory to explore all possible nuclear reaction routes and thus to measure the nuclear cross sections for the concerned radioisotope, as well as of related main contaminants. Among the possible production routes, proton or light-charged particles (e.g. deuteron, ³He and ⁴He) are of great potential, mainly for positron emitters and some therapeutic radionuclides [13, 14]. In recent years, high energy (70 MeV) and multi particle accelerators have been made available from accelerators’ manufacturers. This allows to widen the range of possible nuclear reaction routes which may be explored and radionuclides that may be thus produced (e.g. ²¹¹At or ⁸²Sr, ⁶⁸Ge, etc.). Examples in Europe are the ARRONAX facility [15] and the SPES cyclotron used in the LARAMED project [16].

1.3 The importance of accurate nuclear cross section determination and thick target yields (TTY) measurements in medical radionuclides production

By definition, the cross section σ(E) is the probability of a specific nuclear reaction to occur at the given energy E, i.e. the interaction probability of one bombarding particle and one target atom. This quantity is required to calculate the production yield of the radionuclide of interest. Indeed, the resulting activity at the End Of Bombardment (TTY_EOB) of a given radionuclide, produced by a particle beam hitting a thick target, which also depends on the irradiation time t_IRR and the decay constant λ [s⁻¹], is calculated considering the following formula [3, 17]:

where N _A is the Avogadro constant [mol⁻¹], N _b is the number of bombarding particles [s⁻¹], A _T is the atomic weight of the target material [g⋅mol⁻¹] and S _T(E) is the mass stopping power of the particle beam through the target material at the energy E, expressed as dE/d(ρx). The term ( 1 − e − λ t i r r ) is the so-called saturation factor (SF) that takes into account both the production of nuclei due to the considered nuclear reaction and the decay of nuclei that have already been produced. In a thick target, the impinging projectile energy decreases as it penetrates into the target thickness, thus the energy variation against target thickness, and the related variation of the cross section versus energy, must be considered.

The parameters that have an impact on the production yield (TTY) are the beam intensity, the irradiation time and the beam energy range: the beam intensity is limited by the maximum beam current available by the accelerator and also by the thermal-mechanical main features associated to the targetry; the irradiation time can be increased, keeping in mind that the SF trend keeps linear until the irradiation time remains below one T_1/2; the projectile energy range can be adapted to exploit the maximum cross section value, considering the limitation due to the coproduction of contaminants, by adjusting the incident energy and the target thickness. The presence of contaminants, especially radionuclides with close to or longer half-life than the desired radionuclide, affects the purity of the final product and may cause a dose increase to the patient.

To limit as much as possible the reaction routes which may be opened, and thus the coproduction of undesired radionuclides, it is often necessary to use expensive isotopically enriched materials. Even considering the ideal case of a fully monoisotopic target, it is foreseen the coproduction of isotopic contaminants, depending upon the beam energy range chosen. A good knowledge of production cross sections for all the occurring nuclear reaction routes is therefore a key point that has to be carefully considered. The production cross section of the concerned radionuclide, as well as those of main contaminants, the optimal irradiation parameters may be at last selected, i.e. the irradiation time and beam energy range, in order to ensure the achievement of the best final product purity given by the Specific Activity A_S [Bq⋅g⁻¹], the Radio Nuclidic Purity RNP [%] and the Isotopic Purity IP [%] parameters [2, 18].

2.1 ⁶⁷Cu

⁶⁷Cu has a half-life of 61.83 h and it is emitting β⁻ and γ radiation [36]. The energy of β⁻ particles is appropriate for therapy (141 keV as mean energy, corresponding to about 200–300 μm range in soft tissue), while the 185 keV photons are suitable for pre-therapy low dose imaging (SPECT or SPECT/CT). These physical characteristics allow the use of ⁶⁷Cu for both therapy and associated diagnostic applications, thus making ⁶⁷Cu a promising theranostic radionuclide. It can be also used as a therapeutic counterpart of the ⁶⁴Cu (12.7 h half-life and β⁻/β⁺ emitter), a PET radionuclide [5, 41, 42]. Production techniques for ⁶⁴Cu are well known and are usually based on the ⁶⁴Ni(p,n)⁶⁴Cu and ⁶⁴Ni(d,2n)⁶⁴Cu reactions routes [43], even if there are other possibilities [43, 44].

The use of ⁶⁷Cu has been prevented by the lack of regular availability in enough amounts for preclinical and clinical studies, although it is possible to produce it through many production routes [9, 45, 46]. There is currently not a clear consensus about the best nuclear reaction route to be used for a massive production, due to the impact of the coproduced isotopic ⁶⁴Cu that occurs with some specific routes [47]. Indeed, the presence of the ⁶⁴Cu annihilation photons affects the total dose received by the patient and the staff, while the emitted Auger electrons, the positron and the electrons release their energy close to the decay point, adding their contribution to the dose in tissues.

Nuclear data are therefore important to help in defining benefits and drawbacks associated with each of these production routes. Among the feasible ones are the following:

1. ⁶⁸Zn(p,2p): this production method has quite a low cross section (maximum value of around 10 mb) [40] and requires intermediate energy proton-beams (30 < E < 100 MeV), although it can be exploited also at higher energies [48], [49], [50], [51], [52], [53]. In order to limit the coproduction of other Cu-radionuclides, which directly affects the radionuclidic purity of the final product, the use of enriched ⁶⁸Zn targets is mandatory, as well as the recycling of the target material to make this production route affordable. Among the variety of radionuclides produced during the bombardment of a zinc target, ⁶⁷Ga (half-life 3.2617 d) is particularly relevant: it presents the same γ-lines of ⁶⁷Cu since they both decay into ⁶⁷Zn [36]. Moreover, they have a similar half-life, and therefore it is not possible to infer the precise activity determination of one radionuclide waiting for the decay of the other one. For these reasons a radiochemical procedure, aimed at the separation of copper from gallium elements, is thus mandatory to obtain a precise cross section for ⁶⁷Cu production [49, 54]. The separation procedure, as well as the use of different target materials (natural or enriched), selected monitor reactions or not up-to-date decay data, could explain the large discrepancies of the published data. The measurement carried out at the ARRONAX facility led to an estimated difference on the ⁶⁷Cu production yield of up to 15% in the 70–35 MeV energy range [54]. This difference has a direct impact on the planning of sustainable production of ⁶⁷Cu for medical purposes. It should also be outlined that with this production route a large amount of ⁶⁴Cu is coproduced, as indicated by the ⁶⁸Zn(p,x)⁶⁴Cu cross section recommended by the IAEA up to 100 MeV [37]. ⁶⁴Cu can be limited in the final product by waiting for its decay up to the desired value of the ⁶⁷Cu radionuclidic purity, however loosing part of the ⁶⁷Cu activity.

2. ⁷⁰Zn(p,α): this nuclear reaction, already measured up to 35 MeV by Levkovskij (1991) [50] and Kastleiner et al. (1999) [55], has a maximum cross section value of 15 mb at about 15 MeV. This production route can allow the supply of pure ⁶⁷Cu as there is no coproduction of ⁶⁴Cu below E_thr = 23.8 MeV (threshold energy associated with ⁶⁴Cu production). Considering the low cross section value this production route provides quite a low yield, i.e. 5.75 MBq/µAh for the 30–10 MeV energy range (corresponding to a 1.79 mm thick target of 100% enriched ⁷⁰Zn). Also in this case, enriched target material and recycling are mandatory.

3. ⁷⁰Zn(p,x): The use of ⁷⁰Zn targets at proton energies higher than 35 MeV allows to increase the ⁶⁷Cu yield (Figure 1 left), limiting the ⁶⁴Cu coproduction in comparison to the use of ⁶⁸Zn targets [56]. The ⁶⁷Cu production cross section reaches a value around 22 mb at 70 MeV, thus it is almost a double value than the recommended cross section by the IAEA, based on ⁶⁸Zn target. In addition to that, the ⁶⁴Cu coproduction occurring with ⁷⁰Zn targets shows a decreasing trend versus the increasing energy, while the opposite trend can be noted when using ⁶⁸Zn targets (Figure 1 right). TENDL library values [27] seem to overestimate the ⁷⁰Zn(p,x)⁶⁷Cu reaction, as well as the ⁶⁸Zn(p,2p)⁶⁷Cu cross section (Figure 1 left). On the other hand, TENDL results of the ⁶⁸Zn(p,x)⁶⁴Cu cross section slightly underestimate IAEA recommended values and show an energy shift of about 5 MeV for the peak below 30 MeV (Figure 1 right); the trend of the ⁷⁰Zn(p,x)⁶⁷Cu cross section is well described by the TENDL-data, but there is an underestimation of the experimental values by about a factor of 2.

4. ⁷⁰Zn(d,x): this nuclear reaction has already been measured in 2012 [57] and in 2021 [58] up to 29 MeV. From the most recent data it was possible to identify the cross section peak value associated with this nuclear reaction (about 30 mb at 23 MeV). The comparison between the cross sections data obtained with ⁷⁰Zn as target material for proton or deuteron beams (Figure 2) shows that the deuteron production route is more efficient than the one using protons. Furthermore, since the threshold associated with ⁶⁴Cu production is 26.5 MeV, this isotope is not produced in significant amount up to 30 MeV, allowing for a “pure” ⁶⁷Cu production. The preferred energy range through this route is thus 16–26 MeV, that corresponds to a ⁷⁰Zn thickness of 576 µm. Considering this energy range, 1 µA beam intensity, 1 h of irradiation time, and a target purity of 97.5%, the estimated production yield is 6.2 MBq [58].

5. ⁶⁴Ni(α,p): this nuclear reaction has been measured since the 1960s up to 50 MeV [50, 59–62] and it presents the maximum value of about 38 mb at 22 MeV. Since the threshold for ⁶⁴Cu production is 23.7 MeV, this radioisotope is not produced in large amount up to 30 MeV allowing, also in this case, a “pure” production of ⁶⁷Cu [63]. The main advantage of this production route is the fact that enriched ⁶⁴Ni target (natural abundance equal to 0.9255%) is well mastered as it is the target used for ⁶⁴Cu production. A pure ⁶⁷Cu production requires the use of enriched ⁶⁴Ni and recycling, making the enrichment level of the target material a crucial aspect (currently it is commercially available with an enrichment level >99%).

Figure 1:

⁶⁷Cu (left) and ⁶⁴Cu (right) production cross sections induced by proton-beams on ⁶⁸Zn and ⁷⁰Zn targets.

Figure 2:

⁶⁷Cu production cross sections induced by deuteron and proton beams on ⁷⁰Zn targets.

To compare the charged-particles induced nuclear reactions, Table 2 presents the ⁶⁷Cu (and ⁶⁴Cu) production yields calculated using the IAEA tool ISOTOPIA [26], assuming 100% enriched materials and the same irradiation conditions (⁶⁷Cu SF = 24%, ⁶⁴Cu SF = 73%) considering proton, deuteron and α-beams on ⁷⁰Zn, ⁶⁸Zn, and ⁶⁴Ni targets [61].

The values reported in Table 2 show that with a hypothetical fully enriched target material (^68/70Zn or ⁶⁴Ni) the most convenient route to obtain ⁶⁷Cu (without ⁶⁴Cu coproduction) is by using deuteron beams and ⁷⁰Zn targets. The final ⁶⁷Cu quality using this route will be close to that obtained by photo-production [64, 65]. If some ⁶⁴Cu coproduction is acceptable for clinical application, the ⁶⁸Zn(p,2p)⁶⁷Cu reaction seems to be a better option since it provides a larger ⁶⁷Cu yield in comparison with the ⁷⁰Zn(d,x)⁶⁷Cu route mostly because more intense proton beams can be available (hundreds of µA as compared to tens of µA for deuterons). With the same motivation, also the high-energy ⁷⁰Zn(p,x)⁶⁷Cu route seems to be interesting. The α-beam production route may benefit soon from the newly available multiparticle 30 MeV accelerators providing tens of µA. The first two machines have been installed in Europe (Jülich and Warsaw) and some others are available in Japan. In the future, linear accelerators that can provide very intense deuteron and α-beam (mA range) are soon foreseen, however in this case the major challenge will be the design of specific targets able to withstand such high currents.

2.2 ⁴⁷Sc

⁴⁷Sc (half-life 3.3492 d) is one of the radionuclides of interest in theranostic applications thanks to its β⁻ and γ-radiation (β⁻ mean energy 162 keV, 100% intensity; E_γ = 159.381 keV, 68.3% intensity) which makes it suitable for both therapeutic and diagnostic purposes [41, 42]. The interest in ⁴⁷Sc is also due to the possibility to pair it with a β⁺ emitter isotope, such as ^44gSc and ⁴³Sc, allowing to carry out also PET imaging [66]. ⁴⁷Sc can be produced via different nuclear reaction routes by using cyclotrons, nuclear reactors, and electron linear accelerators [5]; moreover, ⁴⁷Sc may be produced either directly or by the decay of its parent radionuclide, ⁴⁷Ca, as shown in Table 3. Particular care must be paid to the other Sc-isotopes produced during irradiation, since they cannot be chemically separated from ⁴⁷Sc. Among the Sc-isotopic contaminants, the most critical one is ⁴⁶Sc, due to its long half-life (half-life 83.79 d) and γ emissions (889.277 keV with 99.9840% intensity and 1120.545 keV with 99.9870% intensity).

Considering the proton induced-reactions, the:

1. ^natTi(p,x)⁴⁷Sc cross section has been widely studied by many research groups, but no energy range has been found in which the ⁴⁷Sc is produced without ⁴⁶Sc. Considering the composition of ^natTi (natural abundances: ⁵⁰Ti 5.18%, ⁴⁹Ti 5.41%, ⁴⁸Ti 73.72%, ⁴⁷Ti 7.44%, and ⁴⁶Ti 8.25%) it is not possible to extract from the whole cross sections the contribution of each titanium isotope to the production of the specific radionuclides of interest. Since no data are available for the ⁴⁹Ti(p,x)⁴⁷Sc cross section and only a few data sets are published for the ⁵⁰Ti(p,x)⁴⁷Sc [45] and the ⁴⁸Ti(p,2p)⁴⁷Sc [50, 67, 68] reactions, both studied up to 85 MeV, it will be desirable to add new experimental data in order to evaluate if one of these production routes can be competitive.

2. ⁴⁸Ca(p,2n)⁴⁷Sc excitation function was measured in 2019 up to 30 MeV [69] reaching a value of 1100 mb at about 18 MeV. However, it must be taken into account that, despite the large cross section value, the natural abundance of ⁴⁸Ca is only 0.187%, thus making this production route feasible with highly-enriched targets only (enrichment levels available up to 97%), that are anyway very expensive (recycling is mandatory).

3. ^natV(p,x)⁴⁷Sc cross section has been measured by many authors [28, 50, 70–74] up to 200 MeV, reaching a maximum value of about 12 mb at 34 MeV and for energies higher than 80 MeV. A recent work reported new experimental data for ⁴⁷Sc and contaminants coproduction using ^natV targets, compared with theoretical estimations obtained by varying the level densities using the TALYS code [28]. Considering ^natV targets it is possible to limit the coproduction of the ⁴⁶Sc by using the energy range 35–19 MeV. The theoretical ^47/46Sc production estimates based on the experimental results and the well-known DOTA-folate conjugate cm10 ([⁴⁷Sc]-cm10) biodistribution, allowed to evaluate the dose increase due to the presence of the ⁴⁶Sc contaminant [75]. The calculations showed that for E_p < 35 MeV and specific irradiation conditions considered, the effective Dose Increase (DI) due to the presence of contaminants was maintained within the 10% limit required for clinical use [75, 76]. However, to limit the presence of the ⁴⁶Sc contaminant, the calculated ⁴⁷Sc yield in the 35–19 MeV energy range is rather low, i.e. 0.1 GBq (irradiation conditions: 1 µA proton beams and 24 h of irradiation time), still suitable for preclinical studies [75].

Regarding the indirect reactions induced by protons to obtain the ⁴⁷Ca → ⁴⁷Sc precursor system, only few data are available in the international database [40]. The ⁵¹V(p,x)⁴⁷Ca cross section was measured up to 250 MeV with a maximum value of about 0.2 mb making this route unfeasible [69]. No experimental data are present for the ⁵⁰Ti(p,x)⁴⁷Ca and the ⁴⁹Ti(p,x)⁴⁷Ca nuclear reactions, whose threshold energies are 24.5 and 21.2 MeV, respectively. On the other hand, the trend of the ⁴⁸Ca(p,x)⁴⁷Ca cross section was measured up to 19 MeV in 2019 [69], showing a maximum value of about 240 mb at 17 MeV. The low natural abundance of the target material (0.187%) has to be reminded that would make this route very expensive.

Considering the deuteron-induced reactions to produce ⁴⁷Sc, the:

1. ^natTi(d,x)⁴⁷Sc excitation function has been measured by several authors [77], [78], [79], [80], [81], [82], [83], [84], showing a good agreement in the increasing trend of the reaction cross section from low energy up to 50 MeV. Also the co-production of ⁴⁶Sc has been measured by different research teams and it is not possible to indicate a deuteron energy range where there is a pure ⁴⁷Sc production [77], [78], [79], [80], [81], [82], [83], [84], [85], [86].

2. ⁵⁰Ti(d,αn)⁴⁷Sc cross section was measured in 1964 [87], showing an increasing trend up to about 65 mb at 17.5 MeV; unfortunately, there are no data on the coproduction of ⁴⁶Sc, whose production threshold energy is 4.6 MeV.

3. ⁴⁹Ti(d,α)⁴⁷Sc excitation function was measured in the same experimental campaign in 1964 [87], depicting a peak of about 45 mb at around 10 MeV. In this case, the ⁴⁶Sc coproduction was measured in the same energy range (7–20 MeV), obtaining an increasing trend with a plateau value of about 25 mb from 17.5 MeV up to 20 MeV; the lowest ⁴⁹Ti(d,αn)⁴⁶Sc cross section value is at 12 MeV, amounting to ca. 7 mb.

4. ⁴⁸Ti(d,x)⁴⁷Sc cross section has never been measured (E_thr = 6.2 MeV). In the 1960s the ⁴⁸Ti(d,α)⁴⁶Sc excitation function was measured by different authors [87], [88], [89], giving a peak of 40 mb at around 12 MeV. It can be noted that there is a large discrepancy in the low energy region, around 7 MeV, with the value reported by Hall et al. [89] very close to 0 mb and the value measured by Chen et al. [87] around 24 mb; considering the agreement in the values reported by Chen et al. and Anders et al. [88], one may conclude that the 24 mb value is more reliable.

5. ⁴⁷Ti(d,2p)⁴⁷Sc excitation function was also measured in 1964 by Chen et al. [87], showing an increasing trend from 10 MeV up to 20 MeV, with a maximum value of ca. 45 mb. There are no data regarding the ⁴⁶Sc coproduction, whose threshold energy is 5.2 MeV.

6. ⁴⁸Ca(d,3n)⁴⁷Sc and ⁴⁶Ca(d,n)⁴⁷Sc cross sections were not measured, and also not for ⁴⁶Sc coproduction; the threshold energies are respectively 11.4 and 22.5 MeV for ⁴⁷Sc and ⁴⁶Sc production with ⁴⁸Ca targets, 0 and 4.8 MeV in case of ⁴⁶Ca targets.

7. ^natV(d,x)⁴⁷Sc cross section has been measured up to 90 MeV by three research teams, with agreeing results [90], [91], [92]. The cross section starts at 25 MeV and it reaches the maximum value of about 30 mb at ca. 40 MeV, then it decreases to 20 mb at 60 MeV with almost a flat trend up to the maximum energy. Unfortunately, the co-production of ⁴⁶Sc can be avoided only for deuteron energies <30 MeV. Below 40 MeV there is a low co-production of ⁴⁶Sc, since its production cross section is still at 5 mb, but it rapidly increases up to about 50 mb at ca. 55 MeV, keeping a quite flat trend up to 90 MeV.

Regarding the indirect reactions induced by deuterons to obtain the ⁴⁷Ca, it is possible to irradiate ^natV targets: however, the production cross section is quite low (<0.4 mb) in the entire 30–90 MeV energy range investigated [91, 92].

Considering the α-induced reactions, the:

1. ⁴⁴Ca(α,p)⁴⁷Sc cross section has been measured up to 45 MeV reaching a maximum value for ⁴⁷Sc production of about 120 mb at 14 MeV [50]; the same author also measured the coproduction of ⁴⁶Sc, finding a peak of about 400 mb at around 25–30 MeV. Although the natural abundance of ⁴⁴Ca is only 2.086%, the advantage in the use of α-beams lies in the short range of α particles in Ca targets, allowing the use of a limited amount of enriched ⁴⁴Ca target material, compared to the use of proton and deuteron beams. The work published by Minegishi et al. [93] also reports the ⁴⁷Sc yield of about 11 MBq at the end of preparation (approximately 1.5 h from the EOB), obtained by irradiating a [⁴⁴Ca]CaO (⁴⁴Ca = 97.0 atom%, ⁴⁰Ca = 2.89 atom%) target of 200 mg in the energy range 28→0 MeV at 10 eμA, for 2 h irradiation. To obtain pure ⁴⁷Sc it is necessary to take into account the expected large coproduction of ⁴⁶Sc, not reported in the work by Minegishi et al. [93].

2. ⁴⁶Ca(α,t)⁴⁷Sc cross sections are not known [40]; the threshold energies are respectively 12.3 and 23.9 MeV for ⁴⁷Sc and ⁴⁶Sc production.

In conclusion, some charged-particle induced reactions need to be measured to verify if they are feasible as ⁴⁷Sc production routes, e.g. ^48/49/50Ti(p,x)⁴⁷Sc, ⁴⁷Ca → ⁴⁷Sc, ^47/48/49/50Ti(d,x)⁴⁷Sc, ⁴⁶Ca/⁴⁸Ca(d,x)⁴⁷Sc and ⁴⁶Ca(α,x). From all the data available it turns out that the only ways to get ⁴⁷Sc without ⁴⁶Sc using light charged particle induced reactions lead to low production yield, not suitable for clinical use. It is thus necessary to study how to add a mass separation step to remove ⁴⁶Sc. Also in this case further cross section measurements will be needed to identify a precise impurity profile associated with each possible production route and it will be necessary to carefully look at mass separation technique, presumably coupled to laser ionization to be able to extract pure ⁴⁷Sc with high efficiency [94, 95].

2.3 ⁸⁹Zr

Thanks to its nuclear characteristics (T_1/2 = 78.41 h, 22.3% positron emission with a maximum energy of 900 keV), zirconium-89 (⁸⁹Zr) is a very promising radionuclide for immuno-PET [96]. Different experimental results obtained to produce zirconium-89 from a ⁸⁹Y target [97] may be compared by using both proton and deuteron beams [50]. The proton thick target yield is higher up to 26.5 MeV leading to higher production yield, whereas above 26.5 MeV the achievable yield with protons is lower than that obtained with deuterons [96, 98]. At the same time, a lower amount of radionuclidic impurities is produced by deuteron irradiation. For example, to keep a radionuclidic purity of 99.9%, it is possible to use a higher beam energy for deuterons (20.5 MeV) than for protons (16.2 MeV) which turns out in about 10% higher production yield for deuteron.

It is also possible to investigate other routes, such as the reactions ^87,88Sr(α,xn)⁸⁹Zr [50] that can be the object of further investigation as they seem competitive in the production of ⁸⁹Zr with possibly high radionuclidic purity.

2.4 ¹⁰³Pd

¹⁰³Pd (T_1/2 = 16.991 d) decays almost exclusively (99.90%) by electron capture (EC) to ^103mRh (T_1/2 = 56.12 min) which de-excites through internal transition (IT) [36]. As a result of these processes (EC and IT) Auger-electrons and X-rays are emitted which are ideally suited for cancer therapy. Considering also the de-excitation of the “daughter” nuclide ^103mRh, every 100 decays of ¹⁰³Pd are accompanied by the emission of about 263 Auger electrons, 188 low-energy conversion electrons and 97 X-rays. A comparison between the IAEA curve for the deuteron production and the IAEA curve for the proton production of ¹⁰³Pd [99] shows that the yields are comparable up to 12 MeV. For higher particle energies the achievable thick-target yield with deuterons is higher.

There are only two works reporting the production of ¹⁰³Pd by particles other than protons and deuterons [100], [101], [102]. Some of these reactions, mainly ¹⁰¹Ru(α,2n)¹⁰³Pd and ¹⁰³Rh(³He,x)¹⁰³Pd, may be worth the effort for further investigation in order to evaluate the achievable specific activity and the radionuclidic purities.

2.5 ^186gRe

^186gRe is a β⁻/γ emitter radionuclide with a half-life of 90.64 h. The β⁻ end-point energies of 1.07 and 0.93 MeV suggest that this radionuclide is a good candidate for therapy of cancers with small tumour dimensions (from few millimetres to a few centimetres) [103]. Moreover, its γ^-emission at 137.15 keV (very close to that of ^99mTc) is in the energy range suitable for SPECT imaging (i.e., sufficiently high energy to penetrate the body and sufficiently low to be collimated by thin collimators and detected by thin detectors) [41, 42].

Very high specific activity ^186gRe can be produced by either proton [104], [105], [106], [107], [108], [109] or deuteron [77, 110–116] cyclotron irradiation [117], provided an enriched target is used. It is possible to compare the experimental results obtained to produce rhenium-186g from a ^natW target using proton and deuteron beams. The cross section values for the deuteron beams are considerably greater than for the proton beams. The maximum value that is possible to obtain with deuterons is almost 9 times greater than the maximum value for the protons. This effect is also confirmed by the study of thick target yield [118]. It is possible to compare the integrated thin target yields of ^186gRe by using proton or deuteron beam: the two curves are coincident till energy of almost 8 MeV. From this energy up to 9 MeV, the yield of the proton bombardment is slightly larger than the deuteron one but starting from 9 MeV the deuteron yield grows much faster than that of the proton irradiation.

Even if there is a very short energy range in which proton irradiation is competitive with the deuteron one, there is no reason to prefer proton beams. In fact, the advantage of using deuterons over protons for the preparation of ^186gRe is double: more ^186gRe is formed and less ¹⁸⁶W is needed for the target, due to larger -dE/dx in the case of deuterons, by a factor of 4. This lower amount of target material is advantageous both to lower production cost and to simplify the dissolution of the target and its separation from the produced ^186gRe, with advantages also for waste and radioprotection purposes. This makes the deuteron route at facilities with medium-size cyclotrons the best choice for ^186gRe production.

It is also possible to investigate other routes [119], [120], [121] among which the ¹⁹²Os(p,α3n)^186gRe [119] and ¹⁸⁶W(α,x)^186gRe [120, 121] reactions may be of some interest. However, there is a high probability that other rhenium isotopes are produced and will affect the specific activity and the radionuclidic purity of the final product.

2.6 ⁹⁷Ru

The radionuclide ⁹⁷Ru, with a half-life of T_1/2 = 2.83 d, emits low-energy high-intensity γ lines (215.7 keV with 85.8% intensity and 324.5 keV with 10.8%) which have favourable characteristics for SPECT imaging. This can be beneficial to select a patient that may respond to new chemotherapy agents developed with ruthenium [122]. It decays by electron capture (EC) leading to Auger emission that can be used for Auger therapy. It can form a theranostic matched pair with ¹⁰³Ru (T_1/2 = 39.26 d) that decays to the short-lived Auger emitter ^103mRh (T_1/2 = 56.12 min), a promising γ-free therapeutic agent.

Production of ⁹⁷Ru can be done by irradiation of ⁹⁹Tc with protons [123]. Unfortunately, Tc as a radioelement (T_1/2 = 2.111⋅10⁵ y) is not easily available and it requires a dedicated infrastructure to be used as target material. Other reactions with proton-beams can be measured [124], [125], [126], but it seems more attractive to explore the use of an α particle beam coupled to a molybdenum target that is easily available as natural molybdenum or as enriched material [127], [128], [129]. Recently, new cross section values [130] were measured toward higher energy in coherence with commercially available cyclotrons [15] which are able to deliver α-beam up to 70 MeV. These data allow the optimization of ^natMo(⍺,x)⁹⁷Ru production route and constrain the nuclear codes to explore the possibility to use enriched Mo targets. Figure 3 shows recent data reported in the literature for the cross sections of ^natMo(⍺,x) reactions to produce ⁹⁷Ru [123, 130–132].

Figure 3:

Measured cross section for ^natMo(⍺,x)⁹⁷Ru reaction [130] compared with data available in the literature [40, 131, 132].

A plateau is reached above 20 MeV coming from the fact that ^natMo is a mixture of six stable isotopes with comparable abundance. If enriched ⁹⁵Mo or ⁹⁶Mo targets are used instead, one can increase the production by a factor of five–six thanks to the natural abundances of these isotopes, that are respectively 15.84 and 16.67%, and considering that the beam energy can be defined adequately to favour the (α,2n) and (α,3n) reactions. Using a ^natMo target with a thickness of 100 mg/cm² and a 30 MeV beam energy, 3.5 MBq/µAh can be obtained. This example shows the benefit of having an α beam to allow for an alternative route when the target material is not easily available (see also [14]).

2.7 ²¹¹At

²¹¹At (Z = 85) is an α emitter that is promising for targeted α therapy. It has a half-life T_1/2 = 7.214 h and it emits one α particle per decay, either directly decaying to ²⁰⁷Bi (41.80%, Eα = 5.869 MeV) or indirectly through EC (58.2%) to ²¹¹Po, that quickly (T_1/2 = 0.516 s) decays to ²⁰⁷Pb (E_α = 7.3695 MeV). In the emitted radiation by ²¹¹At there are also X-rays that can be used for imaging purpose. As the heavier stable element in the radionuclide chart is ²⁰⁹Bi (Z = 83) and there is no long lived radionuclide in the close vicinity of ²¹¹At, it is not possible to use low or medium energy proton or deuteron beams to produce it. The only possibility is to rely on an α particles beam impinging on a bismuth target (Z = 83). Bismuth is by chance monoisotopic, which makes target fabrication not expensive and easy to manufacture by evaporation under vacuum.

The energy threshold for ²¹¹At production through the (α,2n) reaction is 20.7 MeV. In Figure 4, the family curves of integrated thick target Yields Y(E,ΔE) and the loci of the maxima, corresponding to optimal irradiation conditions for production, are presented. From these curves, it is clear that increasing the incident energy allows increasing the ²¹¹At production. Unfortunately, ²¹⁰At (T_1/2 = 8.1 h) can also be produced by irradiation of a bismuth target with an alpha beam starting at 28.613 MeV (energy threshold of this reaction). This radionuclide must be avoided as much as possible, as it decays to the bone seeker ²¹⁰Po (T_1/2 = 138 d) that decays by α emission and it is highly radiotoxic. One possibility is to restrict the beam energy below the energy threshold of ²¹⁰At production but this will limit the production yield of ²¹¹At at the same time. The rise of the ²¹⁰At production cross section is slow and there is room to optimize the irradiation parameters if one can define the acceptable limit of ²¹⁰At in the final ²¹¹At product. This point is of fairly high importance as beam delivered by accelerators have some energy dispersion. In Table 4 the measured ²¹⁰At/²¹¹At activity ratio are presented [133], [134], [135], showing how the ratio evolves as a function of the beam energy close to ²¹⁰At production energy threshold. From these data, it is clear how the choice of an incident energy of 28.8 MeV, slightly higher than the threshold energy (28.613 MeV) for the ²⁰⁹Bi(α,3n)²¹⁰At reaction, allows to produce limited quantities of the ²¹⁰At radionuclide impurity, remaining below 1%, considered as a limit value.

Figure 4:

Integrated thick target yield for ²¹¹At as a function of energy and energy loss into the target, with the loci of maxima corresponding to the couples (E,ΔE) for optimal irradiation conditions to produce the radionuclide with higher as possible yield, radionuclidic purity and specific activity [134].

More precise cross section values close to the ²¹⁰At threshold may help optimize production yield as a function of ²¹¹At purity. In parallel, some toxicology studies on the impact of ²¹⁰At content in the final ²¹¹At product are needed to get a consensus on the acceptable level of ²¹⁰At for medical application.